ML042080336
| ML042080336 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/02/2004 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| -RFPFR, BFN-TS-405, TAC MB5733, TAC MB5734, TAC MB5735 | |
| Download: ML042080336 (226) | |
Text
I Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 July 2, 2004 BFN-TS-405 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN, P1-35 Washington, D.C. 20555-0001 Gentlemen:
In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 -
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
RELATED TO TECHNICAL SPECIFICATIONS (TS) CHANGE NO. TS-405 -
ALTERNATIVE SOURCE TERM (AST) (TAC NOS. MB5733, MB5734, MB5735)
This letter provides additional information requested by NRC in support of TS-405. TS-405, which was submitted on July 31, 2002, requested a license amendment and TS changes for application of AST methodology for BFN Units 1, 2, and 3.
NRC provided the RAI on June 21, 2004. The questions were discussed with the staff in teleconferences on June 15, and 21, 2004. Enclosure 1 provides TVA's response to each of the staffs questions.
The AST Loss-of-Coolant Accident (LOCA) analysis assumes an alternative leakage treatment pathway to the main condenser as a dose mitigation feature.
Establishing this pathway requires a seismic ruggedness evaluation and Pr-ed o recyedc rpaw
I,1 U.S. Nuclear Regulatory Commission Page 2 July 2, 2004 possible modifications to the main steam system piping. During a June 15, 2004, teleconference NRC requested that TVA provide the Unit 1 seismic ruggedness evaluation for review as part of the approval of the Unit I AST LOCA analysis.
Enclosure 2 provides the Unit 1 MSIV Seismic Ruggedness Evaluation consisting of the "MSIV Seismic Ruggedness Verification for Browns Ferry Nuclear Plant Unit 1," and TVA's, "Seismic Evaluation Report." The evaluation is based on General Electric NEDC-31858, "BWROG Report For Increase In The MSIV Leakage Rate Limits and Elimination of Leakage Control System. Main steam line ruggedness provides an alternate leakage pathway to the main condenser as a dose mitigation feature and is required to support Unit 1 AST LOCA analysis.
NRC intends to complete the review of the Unit I MSIV Seismic Ruggedness Evaluation in time to support the approval of TS-405. However, if for some unforeseen reason, the review of the seismic ruggedness evaluation cannot be completed to support issuance of the Safety Evaluation, TVA proposes the following Unit I license condition.
TVA has performed an analysis showing the main steam line seismic ruggedness and has submitted the results of the analysis to NRC. The analysis is based on General Electric NEDC-31858, "BWROG Report For Increase InThe MSIV Leakage Rate Limits and Elimination of Leakage Control System." TVA will not implement Unit 1 Loss-of-Coolant Accident (LOCA) analysis portion of the Unit I AST analysis until NRC completes review of the seismic ruggedness evaluation.
U.S. Nuclear Regulatory Commission Page 3 July 2, 2004 There are no regulatory commitments contained in this letter. If you have any questions about this, please telephone me at (256) 729-2636.
Pursuant to 28 U. S. C. § 1746 (1994), 1 declare under penalty of perjury that the foregoing is true and correct. Executed on July 2, 2004.
Sincerely, and dustry Affairs En losures/
Response To The une 21, 2004, Request For Additional Information I) Relating Technical Specifications Changie No. TS-405 irce Term (AST)
.t U.S. Nuclear Regulatory Commission Page 4 July 2, 2004 Enclosures cc (Enclosures):
State Health Officer Alabama State Department of Public health RSA Tower - Administration Suite 1552 P.O. Box 30310 Montgomery, Alabama 36130-3017 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 Mr. Stephen J. Cahill, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, AL 35611-6970 Kahtan N. Jabbour, Senior Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Eva A. Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
ENCLOSURE I TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 RESPONSE TO THE JUNE 21, 2004, REQUEST FOR ADDITIONAL INFORMATION (RAI) RELATING TO TECHNICAL SPECIFICATIONS (TS) CHANGE No. TS-405 ALTERNATIVE SOURCE TERM (AST)
This enclosure provides additional information requested by NRC in support of TS-405.
NRC provided the RAls on June 21, 2004. These were discussed with the staff in teleconferences on June 15, and 21, 2004.
NRC Request 1 At the bottom of Page 5 of the replacement pages for Enclosure 2, there is a reference to a "new control room X/Q value associated with an instantaneous ground level puff release." A similar reference occurs on page 36, paragraph 3.1.1.2, and in Table 3-5. These references were present on the original pages as well. In a letter dated March 26, 2003, Tennessee Valley Authority (TVA) agreed to use the current licensing basis X/Q value rather than pursuing the new instantaneous puff model. Please confirm that this is still the position of TVA and that failure to remove these references in the updated pages was an oversight.
TVA Response I Failure to remove reference to the instantaneous puff release was not an oversight on TVA's part. TVA will continue to use the current X/Q licensing basis rather than pursuing the instantaneous puff model shown on page 5 of the Safety Assessment.
In the March 26, 2003 letter (Reference 1), TVA indicated the new puff release X/Q dose is conservative when compared to a calculation based on the current Updated Final Safety Analysis Report (UFSAR) and the existing UFSAR X/Q would continue to be the licensing basis for the main steam line break accident.
In Enclosure 5 of the TS-405 submittal (Reference 2), TVA proposed changes to the UFSAR associated with the implementation of AST, which included the use of the puff model X/Q. However, following approval of AST by NRC, the final changes to the UFSAR will not include discussions of the new Main Steam Line Break puff release dispersion X/Q values, but rather will retain the current UFSAR X/Q values.
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NRC Request 2 Please discuss the status of the regulatory commitment made in your letter dated December 9, 2002, regarding the seismic ruggedness analyses and modifications for Unit 1 alternative main steam isolation valve (MISV) leakage treatment path. When would this analysis be available for staff review?
TVA Response 2 The seismic ruggedness analysis for the Unit 1 MSIV leakage treatment pathway is complete. The Unit I MSIV Seismic Ruggedness Evaluation consisting of the "MSIV Seismic Ruggedness Verification At Browns Ferry Nuclear Plant Unit 1,"
and TVA's, "Seismic Evaluation Report" are provided in Enclosure 2. Required modifications to Unit I will be implemented prior to restart from its extended outage that began in March of 1985.
NRC Request 3 The original submittal requested, in part, cancellation of License Condition 2.C.(4) for Units 2 and 3. However, by submitting the updated analyses the condition was satisfied. The May 17, 2004, letter did not request that the condition also be cancelled for Unit 1, now that the Unit 1 analyses have been submitted. Thus, it is the staffs understanding that the license condition remains effective for Unit 1. Please confirm the staffs understanding.
TVA Response 3 Unit 1 License Condition 2.C.(4) continues to remain in effect. The subject matter of the Unit 1 License Condition 2.C.(4) and the Units 2 and 3 License Condition 2.C.(4) are not related. The Units 2 and 3 License Condition 2.C.(4) was issued as part of Amendment 254, Power Uprate (Reference 3). The condition required TVA to perform analysis of the design basis loss-of-coolant accident (LOCA), confirm off-site and on-site dose limits, obtain NRC approval of the results and make necessary modifications. In response to this license condition, by letter dated March 30, 1999 (Reference 4), TVA forwarded the results of the LOCA analysis. By letter dated August 3,1999 (Reference 5),
NRC stated that TVA had resolved the portion of the condition which required TVA to submit the dose consequences for a LOCA, however; the condition would remain open until the required modifications are completed. These modifications were completed as required by the license condition. Accordingly, the original TS 405 submittal requested deletion of License Condition 2.C.(4) for Units 2 and 3.
Unit 1 License Condition 2.C.(4) was issued as part of Amendment 234, "Request To Convert To Improved Standard Technical Specifications Consistent With NUREG-1433," to allow approval of Amendment 234 without all the supporting analysis complete for Unit 1. The TS 405 submittal referenced the El -2
Unit 1 License Condition 2.C.(4) that would allow approval of AST for Unit I without the supporting analysis. These supporting analyses were submitted in of TVA's May 17, 2004 letter (Reference 6). As such, Unit 1 License Condition 2.C.(4) no longer needed to allow approval of AST.
Since all of the analyses associated with Amendment 234 that were the subject of Unit 1 License Condition 2.C.(4) have not been submitted, it remains effective.
TVA proposed a plan for resolution of Unit I License Condition 2.C.(4) in a letter dated June 16, 2004 (Reference 12).
NRC Request 4 Please confirm that, for Browns Ferry Units 1,2, and 3, all equipment required to meet Title 10, Code of Federal Regulations (10 CFR), Section 50.49, are qualified to total radiation dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design-basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including the dose-rate effects pursuant to 10 CFR 50.49(e)(4). Alternatively, either, (a) confirm the radiation doses used for the environmental qualification analyses at both current licensed thermal power and extended power uprate conditions have been adjusted upward to include the increased radiation doses resulting from the proposed alternative source term (AST), or (b) describe the new method (versus type testing) for demonstrating components are qualified pursuant with the requirements of 10 CFR 50.49 and provide justification that components are qualified for the increased radiation doses resulting from the proposed AST pursuant with the requirements of General Design Criteria 4 of 10 CFR Part 50, Appendix A and 10 CFR 50.49.
TVA Response 4 As previously provided in our March 26, 2003 (Referencel), reply to the October 15, 2002, RAI, TVA has elected to retain the TID-14844 assumptions for performing the required environmental qualification (EQ) analyses. The radiation doses used for the EQ analyses at both current licensed thermal power and Extended Power Uprate (EPU) conditions are adjusted upward from the original values based on the determined source terms of the ORIGEN computer code for the respective power level. The BFN AST analysis considers the source term from the Fission Product Inventory shown on Table 2-2 of the Safety Assessment. A reactor thermal power of 4031 MWt (102 percent of 3952 MWt) is also used.
Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Section 1.3.5, states that licensees may use either the AST or the TID-14844 assumptions for performing the required EQ analyses. It further states that no plant modifications are El-3
required to address the impact of the difference in source term characteristics (i.e., AST vs TID 14844) on EQ doses pending the outcome of the generic issue.
The issue was subsequently evaluated by Generic Safety Issue 187.
Supplement 25 to NUREG-0933, "A Prioritization of Generic Safety Issues,"
dated June 30, 2001, provides the resolution of Generic Issue 187, 'The Potential Impact of Postulated Cesium Concentration on Equipment Qualification." The supporting description states that the Sandia National Laboratories' report, "Evaluation of Radiological Consequences of Design Basis Accidents at Operating Reactors Using the Revised Source Term," dated September 28, 1998, showed that 1) for equipment exposed to the containment atmosphere, the TID-14844 source term and the gap and in-vessel releases in the AST produced similar integrated doses, and 2) for equipment exposed to suppression pool water, the integrated doses calculated with the AST remain enveloped by those calculated with TID-14844 for the first 145 days post accident for a BWR, including the 30 percent vs. 1 percent release of cesium.
There are no instruments in the suppression pool that are within the EQ program.
The BFN EQ program is based on qualification for 100 days post accident.
Based on the above, the continued use of the TID-14844 source term provides integrated doses for equipment which envelope those that would be calculated using AST. Therefore, following implementation of AST, BFN will continue to meet their commitment to 10 CFR 50.49 by using a radiation environment associated with the most severe design basis accident.
With regard to the BFN EQ program qualification time of 100 days post accident.
An internal memo issued in January 1983, established a maximum post accident operating time of 100 days within TVA. This is publicized in various correspondences with NRC. Enclosure 3 of a May 20, 1983, letter (Reference 7) that provided a Justification For Continued Operation for various equipment indicated a maximum post LOCA operating time of 100 days. In a meeting with NRC on September 16, 1985, as shown in October 9, 1985, NRC meeting summary (Reference 8), TVA stated the EQ program is based on 100 days post LOCA operation. In NRC inspection reports for the Browns Ferry EQ program dated September 1, 1988 (Reference 9), and December 14, 1990 (Reference 10), the inspectors reviewed areas such as required post-accident operating time compared to the duration of time equipment was qualified and determined that the EQ program meets the requirements of 10 CFR 50.49.
TVA has subsequently submitted other TS changes, including Power Uprate, using 100 days qualification post accident as a basis for the EQ program (Reference 3).
NRC Reauest 5 In TVA's December 9, 2002 letter, TVA responded to staff questions regarding assumptions used in assessing MSIV leakage deposition (NRC request 5, page El -4
EI-17). Please confirm that the responses provided, including piping length, surface areas, etc., are valid for BFN Unit I as well.
TVA Response 5 The Unit I configuration including piping length, surface areas etc., are the same as the Units 2 and 3 configuration. Therefore, the TVA response to NRC Request 5 in TVA's December 9, 2002 letter (Reference 11), is applicable to the Unit 1 configurations.
NRC Request 6 In the response to NRC Request 3 page E1-8 of the May 17, 2004 letter, the licensee stated:
"The CS [core spray] ECCS [emergency core cooling system] injection flow delivered to the reactor vessel mixes with the sodium pentaborate (SP) and spills from the reactor vessel through the recirculation line break flushing SP to the suppression pool."
The pH control of the suppression pool is dependent on the transfer of a quantity of SP to balance the acids produced by radiolytic action and other sources.
From the information provided, it appears that the SP is mixed primarily in the core region and lower head regions of the vessel, although it is unclear what degree of mixing of water containing SP with injected water would be required to assure that sufficient SP is transported out the break and to the suppression pool. The following questions are related to this concern:
NRC Request 6.a In the event that both low pressure coolant injection (LPCI) trains are operating, does the core level remain essentially at the top of the jet pumps with the LPCI water being spilled into the annulus from the top of the jet pumps and out the break with no mixing with the SP containing water?
TVA Response 6.a For a large break LOCA of the BFN recirculation system piping, there will always be one loop (two pumps) of the CS system available for ECCS injection flow, both short term (< 10 minutes) and long term (> 10 minutes). In the BFN LOCA analysis, a large break LOCA of the recirculation system piping with the limiting single failure constitutes the maximum loss of ECCS makeup capacity. As a consequence of the single failure, a complete train each of Residual Heat Removal (RHR) Low Pressure Coolant Injection (LPCI) and CS ECCS injection capability is lost. This results in one loop (two pumps) of CS and one loop (two pumps) of RHR LPCI remaining available. However, the remaining available loop of RHR LPCI flow is assumed lost directly through the break.
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Inthe event that RHR LPCI flow exists in the short term (< 10 minutes), the RHR LPCI flow would contribute to core reflood, and then overflow the jet pumps, enter the annulus, and exit the vessel through the break. BFN Updated Final Safety Analysis Report (UFSAR) Section 14.6.3.3.2.1 states that the RHR pumps are switched from the LPCI mode to the containment cooling mode at 10 minutes. Therefore, in the long term, RHR flow has no influence on the CS ECCS injection capability to maintain water level and continue the flushing action of SP to the suppression pool. However, although LOCA containment analysis dictates redirection of available LPCI capacity to containment cooling in the long term, LPCI flow would assist in providing core reflood with the resulting contribution to the lower head region of the reactor vessel providing additional flow for the lower head mixing and vessel coolant exit through the break as discussed in the response to RAI-6.c.
NRC Request 6.b In the event that one train of LPCI is operating (through single failure or isolation) is the momentum of injection capable of forcing flow into the lower head region and out the non active jet pumps in the reverse direction? Are controls needed to isolate one LPCI after SLC injection since it would promote mixing and sweeping of SP water from the vessel?
TVA Response 6.b No controls are needed for the isolation of RHR LPCI injection in the event it should be operating since it does not substantially affect the mixing and sweeping SP from the vessel as credited in the analysis. As stated in the response to RAI-6.a, in the event that RHR LPCI flow exists in the short term
(< 10 minutes), the RHR LPCI flow would contribute to core reflood, and then overflow the jet pumps, enter the annulus, and exit the vessel through the break.
BFN UFSAR Section 14.6.3.3.2.1 states that the RHR pumps are switched from the LPCI mode to the containment cooling mode at 600 seconds, or 10 minutes.
Since LPCI is a short term operation in LOCA analyses, no controls are needed to isolate RHR LPCI flow in the event it should be operating. The CS ECCS capability to maintain water level and continue the flushing of SP to the suppression pool long term are discussed in the response to RAI-6.c.
NRC Request 6.c The addition of core spray water would force some SP mixed water out of the jet pumps, but steaming due to decay heat could reduce this. What is the amount of SP mixed water that is removed from the vessel by CS ECCS operation? If this is the only mixing process, how long would it take to transport sufficient SP to the suppression pool? Would substantial SP remain in the vessel and be unavailable for pH control in the suppression pool? Does the chemical analysis El -6
need to assume a quantity of SP remains in the vessel as part of its balance with the acid producing mechanisms?
TVA Response 6.c The attached figure is a simplified representation of the BFN reactor vessel and internal components.
For a large break LOCA of the BFN recirculation system piping, there will always be one loop (two pumps) of the CS system available for ECCS injection flow at an assumed LOCA analysis flow rate of 5,600 gpm. This post LOCA CS ECCS injection takes suction from the suppression pool and discharges directly above the active reactor core region inside the core shroud. Based upon the design basis Reactor Core Isolation Cooling (RCIC) system flow rate of 600 gpm to offset maximum decay heat (by steaming), the net CS ECCS injection liquid flow rate is effectively 5,000 gpm immediately after the event. This net injection rate will be conservatively used throughout the event.
The free fluid volume of the entire reactor vessel and the recirculation system is approximately 109,300 gallons of liquid. With a net CS ECCS injection rate of 5,000 gpm, this volume would be refilled approximately every 22 minutes.
The free fluid volumes of the reactor vessel lower plenum and lower head zones are Zone A - 5,199 gallons, Zone B - 12,440 gallons, and Zone C - 8,468 gallons, for an approximate total of 26,107 gallons. At a net CS ECCS injection rate of 5,000 gpm, the lower plenum and lower head liquid volume would be displaced approximately every 5.2 minutes.
The Standby Liquid Control (SLC) system differential pressure and liquid control line (shown in the attached figure as Detail A) serves a dual function within the reactor vessel to inject SP solution into the coolant stream and to sense the differential pressure across the core support assembly during normal operation.
The line enters the reactor vessel at a point below the core shroud as two concentric pipes. Inthe lower plenum, the two pipes separate, with the inner pipe being the SP injection line. As shown, this pipe is a vertical perforated sparger that disperses the SP in the lower shroud immediately below the reactor core plate and above the jet pump outlets.
Considering the Technical Specification minimum SLC single pump flow rate of 39 gpm, the credited AST SP volume of 4,000 gallons will be injected in less than two hours. The injection of the SP is concurrent with the CS ECCS injection occurring in reverse flow through the active core region to the lower plenum and lower head zones. The flow path of least resistance for the two flows is to travel axially along the guide tubes in Zones C and B until reaching the region below the guide tubes where the control rod drive mechanism housings are located (Zone A). There, the much smaller diameter of the drive mechanism housings El-7
provides a radial flow path with a significantly larger flow area and lower losses than the guide tube region of the lower plenum. It is not expected that substantial radial flow would occur across the guide tubes.
The colder SP water has a slightly higher density than the CS ECCS water in the lower plenum and head (specific gravity of about 1.1). As a result, the SP water would fall or be entrained with the CS ECCS flow from the injection point to the region below the jet pump outlet. Here it would either mix with the CS ECCS water that is flowing radially across this region and be drawn upward into the jet pumps or be drawn off via the reactor vessel bottom head drain line.
The BFN reactor pressure vessel (RPV) includes a two inch bottom head drain line that is designed to provide a constant flow of reactor coolant from the RPV bottom head to the Reactor Water Cleanup System during reactor operation to obtain a representative temperature measurement in the bottom region of the RPV and for removal of cold water accumulation from the control rod drive system in-leakage from the bottom region of the RPV. The active flow path of this line is from the vessel bottom head to the Reactor Water Cleanup (RWCU) system that takes suction from recirculation loop A. The BFN large recirculation break LOCA analysis includes this bottom head drain line as a contribution to break flow during the LOCA event and this flow path remains active post LOCA for continued flow to the break. This lower vessel head flow path will provide additional assurance that SP will not stagnate in the lower head, but will be flushed out of the break to the suppression pool.
The suppression pool pH analysis for BFN shows that no credit for suppression pool buffering is required until greater than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> post-LOCA. Prior to this time, the acid added to the suppression pool from radiolysis of water and radiolysis of cable is not enough to neutralize the CsOH that will exist. Based on this analysis, BFN conservatively assumes that SP injection will occur no later than two hours after accident initiation and adequate mixing occurs prior to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Inaccordance with the AST pH analysis, an amount of SP would remain in the reactor vessel (and recirculation system) in equilibrium with the suppression pool SP concentration. The BFN suppression pool pH chemical analysis volumetric determination of pH conservatively includes the initial maximum volume of the suppression pool as well as the reactor coolant inventory to be borated to arrive at the final equilibrium SP concentration of the pool to be considered for pH control.
Because of the configuration of the RPV, the path CS ECCS injection takes through the RPV, and the configuration of the SLC injection piping the post LOCA flow through the lower head zone provides adequate volume turnover along with flushing of the SP out of the lower head zone to the suppression pool through either the jet pumps into the annulus or through the bottom head drain.
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NRC Request 7 The staff considers the Administrative controls to restore isolation of the secondary containment and to terminate venting in the event of a refueling accident as an important defense in depth measure. Inwhat document will these administrative controls be located? Other licensees have indicated that (1) designated personnel will be aware of which openings would require closure, (2) specific responsibilities for closure would be assigned, and (3) obstructions that could prevent closure would be easily removable. Insome cases, licensees have specified a time to achieve closure and have added the administrative control definition to their technical specifications. Please provide the staff with additional information as to the content of the administrative controls.
TVA Response 7 The three BFN Units share a common secondary containment. BFN TS 3.6.4.1 and 3.6.4.2 require the secondary containment and the associated isolation valves to be operable in Modes 1,2, and 3. Therefore, all three BFN Units would need to be in Modes 4 or 5 for secondary containment not to be operable.
Additionally, the AST analyses do not take credit for secondary containment during the movement of irradiated fuel and during core alterations. TS-405 shows offsite doses following a refueling accident to be well below regulatory limits.
As a defense in depth measure, TVA plans to revise the following procedures:
The BFN General Operating Instruction for fuel movements during refueling will be revised to verify that prior to moving irradiated fuel, if secondary containment is not required to be operable, that it can be reestablished.
The BFN Technical Instruction for secondary containment penetration breach analysis currently contains actions to be taken in the event secondary containment cannot be maintained due to a breech exceeding the available margin. These steps include stationing an Auxiliary Unit Operator at the breach location that is responsible for closing the breach if instructed by the control room. This instruction will be revised to require calculating the size of the breaches in secondary containment even when the TSs do not require secondary containment.
The BFN Abnormal Operating Instruction for fuel damage during refueling provides the symptoms, automatic actions and operator actions for a fuel damage accident, including a dropped fuel bundle. Steps will be added to this instruction to ensure that secondary containment is intact or promptly restored following a postulated fuel handling accident.
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References:
- 1. TVA letter to NRC dated March 26, 2003. "Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 Response to Request for Additional Information Relating to Technical Specifications Change No. TS-405 - Alternative Source Term."
- Units 1, 2, and 3 - License Amendment - Alternative Source Term."
- 3. NRC letter to TVA dated September 8,1998, "Issuance of Amendments RE:
Power Uprate - Browns Ferry Plant, Units 2, and 3."
- 4. TVA letter to NRC dated March 30, 1999, "Browns Ferry Nuclear Plant (BFN) - Resolution of Control Room Emergency Ventilation (CREV) System Issues with Regard to License Condition Associated with Units 2 and 3 Power Uprate Operating License Amendments 254 and 214."
- 5. NRC letter to TVA dated August 3,1999, "Safety Evaluation Supplement, Browns Ferry Nuclear Plant Units 2 and 3 - Radiological Dose Calculations Associated with Power Uprate License Amendment Nos. 254 and 214."
- Units 1, 2, and 3 - Response to Request for Additional Information and Unit 1 Analysis Results Related to Technical Specifications Change No. TS-405 - Alternative Source Term."
- 7. TVA letter to NRC dated May 20,1983, This letter provided TVA's response to NRC regarding the Environmental qualification of equipment required by 10 CFR 50.49(g) for Browns Ferry Nuclear Plant.
- 8. NRC letter to TVA dated October 9,1985, "Summary of Meeting On Environmental Qualifications of TVA Plants."
- 9. NRC letter to TVA dated September 1,1988, "Environmental Qualification (10 CFR 50.49) Inspection (50-259, 260, 296/88-11)."
- 10. NRC letter to TVA dated December, 14,1990, "Environmental Qualification (10 CFR 50.49) Inspection (50-259, 260, 296/90-22)."
11.TVA letter to NRC dated December 9, 2002, "Browns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3 - Response to Request for Additional Information Relating to Technical Specifications Change No. TS-405 alternative Source Term."
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References:
12.TVA letter to NRC dated June 16, 2004, "Browns Ferry Nuclear Plant (BFN)
- Unit 1 - Plan for Satisfying License Condition 2.C.(4)."
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ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNIT 1 BFN UNIT 1 MSIV SEISMIC RUGGEDNESS EVALUATION
- MSIV Seismic Ruggedness Verification At Browns Ferry Nuclear Plant,
- Seismic Evaluation Report
- .1 _ _ _ _ _ _ _ _ _ _ _
t w MSIV SEISMIC RUGGEDNESS VERIFICATION AT BROWNS FERRY NUCLEAR PLANT UNIT 1 May 2004 Prepared for:
BROWNS FERRY UNIT 1 RESTART PROJECT Preparedby:
FACILITY RISK CONSULTANTS, INC.
FACIfI'Y RISK CONSULTANTS! INC.
EXECUTIVE
SUMMARY
This report summarizes the seismic ruggedness verification program for increasing the Main Steam Isolation Valve (MSIV) leakage rate limits at Browns Ferry Nuclear Plant Unit 1 (BFN-1). Key engineering attributes of the seismic verification program consisted of the following:
- Review of the MSIV seismic verification boundaries;
- In-plant screening walkdown evaluations and identification of potential outliers;
- Further evaluations and resolution of potential outliers;
- Recommendations for plant modifications to resolve outliers;
- Work order requests to address general maintenance and housekeeping items.
BFN-1 MSIV seismic ruggedness verification program was performed in accordance with the recommendations of the General Electric BWR Owners Group Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems. MSIV leakage boundaries were established to ensure that the main steam piping from the outboard MSIV's to the turbine stop valves and the main steam drain line from the outboard MSIV's to and including the condenser itself are capable of being a pressure retaining boundary following a Design Basis Earthquake (DBE).
In-plant screening walkdown evaluations of piping and associated components within the seismic verification boundaries were performed in accordance with Walkdown Instruction WI-BFN-0-CEB-07, "EngineeringWalkdown Instruction for MSIV Seismic Ruggedness Verification." A total of 15 subsystems within the seismic boundaries were included.
Screening evaluations focused on key design attributes and seismic issues such as pipe spans and support integrity; seismic interaction Issues including proximity impact and falling concerns; differential displacement of structures, equipment and piping; performance of seismic verification boundary components, among others. Screening tools such as seismic deflection estimates and charts for various plant features; pipe flexibility and seismic anchor movement evaluation charts; support and anchorage capacity screening charts; and others were used in the in-plant screening walkdown evaluations.
Certain configurations identified during the in-plant screening walkdowns as not meeting the screening criteria were documented in the Potential Outlier Sheet (POS) as potential Page i FACILITY RISK CONSULTANTS, INC.
outliers and for further evaluation and disposition. Walkdown results, including a total of 54 potential outliers identified, are documented in the Walkdown Data Packages (WDP's) for the respective subsystems.
Potential outliers identified during the in-plant screening walkdowns were further evaluated to the acceptance criteria of TVA Design Criteria BFN-50-C-7306, "Qualification Criteria for Seismic Class 11 Piping, Pipe Supports, and Components." Further evaluations and bounding analyses of these potential outliers consisted of hand calculations using basic engineering mechanics techniques for simple configurations, and rigorous piping analyses (TPIPE computer program) for more complex piping configurations. A total of 15 outliers were found to have not met the acceptance criteria. Plant modifications were designed and several Design Change Notices (DCN's) were issued to implement the changes so that all of these concerns were resolved. Furthermore, 15 maintenance and/or housekeeping items were also identified for corrective actions. Maintenance work order requests were issued to address these items.
Overall program results for the MSIV seismic ruggedness verification program for Browns Ferry Unit 1 are summarized in the following table:
Subsysem Desriptio PN,.of!:*,.2 ssem P,..ote~ntila Design.i. ,--Maint,.
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I :.
Nt. Outliiers WRds MS Drain Line - Turbine Bldg. Main Steam Tunnel 4 0 0 MS Lines - Turbine Bldg. 2 0 1 MS Drain Line - Reactor Bldg. Main Steam Vault 2 0 0 HPCI/RCIC/Aux. Boiler Drains 8 2 1 MS Pressure Transmitters PT 1-72, 76, 82, 86 & 93 6 4 2 MS Sample Lines to Sampling Station 4 2 2 MS Bypass 1 0 0 MS Stop Valve Above Seat Drains 3 0 3 MS to Steam Seal Regulator 6 1 1 Steam Supply to RFP Turbines 6 1 2 Page ii FACILITY RISK CONSULTANTS. INC.
. ... I 11 Steam Supply to Steam Jet Air Ejectors (SJAE's) 6 2 I 12 Steam Supply to Off-Gas Preheaters 1 1 0 13 SJAE's Drain to Condenser 1 1 0 14 MS Drain Line (Turbine Bldg.) to Condenser 3 1 2 15 Condenser 1 0 0 Total 54 15 15 Page iii FACILITY RISK CONSULTANTS. INC.
TABLE OF CONTENTS Page Executive Summary .................. ....................................... i
- 1. INTRODUCTION............................................................................ 1-1 1.1 Report Organization ......................................................... 1-1
- 2. PROGRAM SCOPE ......................................................... 2-1 2.1 Seismic Verification Boundary ............................ 2-1 2.2 Seismic Walkdown Scope ....................................................... 2-2
- 3. SEISMIC VERIFICATION WALKDOWN ............................................... 3-1 3.1 Seismic Verification Review Guidelines . .3-2 3.2 Piping, Pipe Support and Equipment Attributes . .3-2 3.2.1 Piping and Pipe Support Design Attributes .3-3 3.2.2 Vulnerable Pipe Joints .3-3 3.3.3 Other Potential Seismic Vulnerabilities .3-3 3.3 Equipment Design Attributes .. 3-4 3.4 Anchorage Design Attributes .. 3-5 3.4.1 Expansion Anchor Bolts Review Guidelines .3-5 3.4.2 Welded Anchorage Review Guidelines .3-5 3.5 Seismic Anchor Movement .. 3-6 3.6 Seismic Interaction Review (Il/I and Proximity) . .3-6 3.7 Seismic Verification Boundary Valves . .3-7 3.8 Walkdown .. 3-8 3.9 Documentation .. 3-8
- 4. WALKDOWN OPEN ITEMS ......................................................... 4-1 4.1 Main Steam Drain Line - Turbine Bldg. Main Steam Tunnel .4-1 4.2 Main Steam - Turbine Bldg .4-2 4.3 Main Steam Drain Line - Reactor Bldg. Main Steam Vault .4-3 Page iv FACILITY RISK CONSULTANTS, INC.
TABLE OF CONTENTS (CONT'D)
Page 4.4 HPCI/RCIC/Aux. Boiler Drains ................................................ 4-4 4.5 Main Steam Pressure Transmitters PT 1-72, 76, 82, 86 & 93 ........ 4-5 4.6 Main Steam Sample Lines to Sampling Station ............ .............. 4-6 4.7 Main Steam Bypass ....................... ......................... 4-7 4.8 Main Steam Stop Valve Above Seat Drains ............... ................ 4-7 4.9 Main Steam to Steam Seal Regulator ....................................... 4-8 4.10 Steam Supply to RFP Turbines ............................................... 4-8 4.11 Steam Supply to Steam Jet Air Ejectors (SJAE's) ....................... 4-9 4.12 Steam Supply to Off-Gas Preheaters ....................................... 4-10 4.13 SJAE's Drain to Condenser .. ................ 4-11 4.14 MS Drain Line (Turbine Bldg.) to Condenser ............................. 4-11 4.15 Condenser ......................................... 4-12 4.16 Summary of MSIV Seismic Walkdown Evaluations ...................... 4-13
- 5. POTENTIAL OUTLIER RESOLUTION . . .5-1 5.1 Outlier Evaluation Guidelines .. 5-1 5.1.1 Seismic Demand .5-1 5.1.2 Equipment Anchorage Acceptance Criteria.5-2 5.1.3 Pipe Support Acceptance Criteria .5-2 5.1.4 Pipe Stress Acceptance Criteria .5-3 5.2 Outlier Evaluation Results .5-3
- 6.
SUMMARY
AND RECOMMENDATIONS . .6-1 6.1 Plant Modifications .6-1 6.2 Maintenance and Housekeeping Items .6-1
- 7. REFERENCES. .............................................................................. 7-1 Page v FACILIW RISK CONSULTANTS, INC.
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TABLE OF CONTENTS (CONT'D)
APPENDICES Page APPENDIX A: BFN-1 MSIV Seismic Ruggedness Verification Walkdown Data Packages......................................................................... A-1 APPENDIX B: BFN-1 MSIV Seismic Ruggedness Verification Outlier Resolution Calculation Packages ........................................................ B-1 APPENDIX C: BFN-1 MSIV Seismic Ruggedness Verification Plant Modification Design Calculation Packages and Maint. Work Orders ........ C-1 FIGURES 2-1 BFN-1 MSIV Seismic Verification Boundary ...................... .................... 2-6 2-2 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 1 ......... ........ 2-7 2-3 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 2 ......... ........ 2-8 2-4 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 3 ......... ........ 2-9 2-5 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 4 .2-10 2-6 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 5 .2-11 2-7 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 6 ................. 2-12 2-8 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 7 .2-13 2-9 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 8.2-14 2-10 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 9 .2-15 2-11 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 10 .2-16 2-12 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 11 .2-17 2-13 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 12 .2-18 2-14 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 13 .2-19 2-15 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 14 .2-20 2-16 BFN-1 MSIV Seismic Walkdown by Subsystems - Package 15 .2-21 3-1 Comparison of Database Site Spectra to BFN Design Basis Ground Spectrum ..... 3-10 Page vi FACILITY RISK CONSULTANTS, INC.
TABLE OF CONTENTS (CONT'D)
TABLES Paae 2-1 BFN Unit 1 Mechanical Flow Diagrams ........................ ...................... 2-3 2-2 BFN Unit 1 MSIV Leakage Boundary Points ................. ....................... 2-4 5-1 Outlier Resolution Summary for BFN-1 MSIV Seismic Ruggedness 5-5 Verification Program .
6-1 Summary of BFN-1 MSIV Seismic Ruggedness Verification Program ........ 6-2 6-2 Summary of Plant Modifications, BFN-1 MSIV Seismic Ruggedness Verification Program ............................... ........................... 6-3 6-3 Summary of Misc. Maintenance & Housekeeping Items, BFN-1 MSIV Seismic Ruggedness Verification Program .................. ........................ 6-5 Page vii FACLITY RISK CONSULTANTS, INC.
- 1. INTRODUCTION This report summarizes the work performed for supplemental plant-specific Main Steam Piping Seismic Verification. The verification program was performed in accordance with the recommendation of the General Electric BWR Owners Group Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems (Reference 7-1).
The objective of the supplemental plant specific walkdown verification was to identify specific design conditions associated with poor piping and component seismic performance. The walkdown was directed toward identification of the following specific phenomenon and issues:
Piping Support Integrity Issues
- Seismic Interaction Issues of:
- Failure and Falling (Il/I Concerns)
- Displacement and Proximity Impact Issues
- Differential Displacement of Structures, Equipment and Piping
- Performance of Seismic Verification Boundary Components 1.1 Report Organization This report is organized as follows. Chapter 2 presents the scope of the program.
Technical bases and evaluation approach for in-plant screening evaluations, including walkdown implementation, are described In Chapter 3. Summary of in-plant screening walkdowns and evaluations, including open items identified, is presented in Chapter 4, while that of potential outlier resolution is in Chapter 5. Overall summary and recommendations, including plant modifications and maintenance work orders, are provided in Chapter 6. References are listed in Chapter 7 of the report.
There are 3 appendices to this report. Appendix A contains a listing of walkdown data packages (WDP) which document the in-plant screening evaluations performed by the Walkdown Teams. Appendix B contains a list of engineering calculations which document the resolution of potential outliers by further analyses. A list of calculations Page 1-1 FACTLIY RISK CONSULTANTS, INC.
containing the engineering design of plant modifications for the resolution of BFN-1 MSIV outliers, as well as maintenance work order requests, are provided in Appendix C.
Page 1-2 FACILITY RISK CONSULTANTS, INC.
-90
- 2. PROGRAM SCOPE The walkdown scope included the drain path that will be established to convey leakage past the outboard Main Steam Isolation Valves (MSIV's) to the isolated condenser and includes piping, instrumentation, valves and equipment that would be required to maintain the drain pathway.
2.1 Seismic Verification Boundary BFN-1 MSIV Leakage Containment Boundaries are described in Reference 7-2, and are shown in Figure 2-1. The associated flow diagrams are listed in Table 2-1, and the piping isolation boundaries defining the seismic verification boundary are shown in Table 2-2. Note that Unit 1 seismic boundaries are generally similar to those for Units 2 and 3.
The scope of the seismic verification walkdown presented in this report generally consists of the-following portions of the Main Steam (MS) system beyond the outboard MSIV's:
- 1. Main Steam drain path to the condenser for any leakage past the isolated outboard MSIV's.
- 2. Main Steam piping from the outboard MSIV's to the Main Steam Stop Valves (MSV's).
- 3. Main Steam bypass piping from the Main Steam lines to the Bypass Valve Chest.
- 4. Main Condensers.
- 5. Additional piping and instrumentation within the Seismic Verification Boundary includes:
Stop Valve Above Seat Drains to Condenser Steam Sample System HPCI/RCIC Steam Drains to Main Steam Auxiliary Boiler Drains to Main Steam Main Steam Instrumentation Steam to Steam Seal Regulator Main Steam Supply to Reactor Feed Pumps Main Steam Supply to Steam Jet Air Ejectors Main Steam Supply to Off-Gas Preheaters Page 2-1 FACILUTY RISK CONSULTANTS, INC.
w : __ _ _; . 6--Y 2.2 Seismic Walkdown Scope The scope of the seismic verification walkdown included consideration of design conditions which in past earthquake experience have been associated with piping damage and could contribute to pressure boundary failure and inventory release. These conditions include support failure, falling of non-seismically designed plant features (Il/1), proximity impact, and differential seismic anchor motion of structures, piping or equipment. The scope and extent of these conditions are consistent with the guidelines as specified in Reference 7-1 and are described in greater details in Chapters 3 and 4 of this report.
BFN-1 MSIV seismic ruggedness verification walkdowns were performed considering a total of 15 subsystems or portions within the MSIV seismic verification boundary, as follows:
- 1. Main Steam Drain Line - Turbine Bldg. Main Steam Tunnel
- 2. Main Steam - Turbine Bldg.
- 3. Main Steam Drain Line - Reactor Bldg. Main Steam Vault
- 4. HPCI/RCIC/Aux. Boiler Drains
- 5. Main Steam Pressure Transmitters PT 1-72, 76, 82, 86 & 93
- 6. Main Steam Sample Lines to Sampling Station
- 7. Main Steam Bypass
- 8. Main Steam Stop Valve Above Seat Drains
- 9. Main Steam to Steam Seal Regulator
- 10. Steam Supply to RFP Turbines
- 11. Steam Supply to Steam Jet Air Ejectors (SJAE's)
- 12. Steam Supply to Off-Gas Preheaters
- 13. SJAE's Drain to Condenser
- 14. Main Steam Drain Line (Turbine Bldg.) to Condenser
- 15. Condenser Figures 2-2 to 2-16 highlight the extent of the seismic walkdown scope for each of the above 15 subsystems or portions within the BFN-1 MSIV seismic verification boundary Page 2-2 FA CJLJY RISK CONSULTANTS, 1NC.
1-11 . ---- I- II . -
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'1-47E807-1 Turbine Drains and Miscellaneous Piping 1-47E807-2 Turbine Drains and Miscellaneous Piping 1-47E812-1 High Pressure Coolant Injection System 1-47E813-1 Reactor Core Isolation Cooling System 1-47E81 5-1 0-47E81 1Auxiliary Boiler System 1-47E610-43-1 Sampling and Water Quality System (Mechanical Control Diagram)
Page 2-3 FACILuTY RISK CONSULTANTSt INC.
Table 2-2 BFN UNIT 1 MSIV LEAKAGE BOUNDARY POINTS Leakage Flow Boundary Diagran/ Comment Point Drawing 1-FCV-1-15 1-47E801-1 Outboard MSIV for Main Steam Line A 1-FCV-1-27 1-47E801-1 Outboard MSIV for Main Steam Line B 1-FCV-1-38 1-47E801-1 Outboard MSIV for Main Steam Line C 1-FCV-1 -52 1-47E801-1 Outboard MSIV for Main Steam Line D 1-FCV-1-56 1-47E801-1 Outboard containment isolation valve for Primary Containment steam drains 1-1 -527 1-47E801-1 Normally closed Main Steam Drain manual isolation valves 1-43-631 1-47E610-43-1 Normally closed Main Steam Sample System manual isolation valve 1-43-631A 1-47E610-43-1 Normally closed Main Steam Sample System manual isolation valve 1-43-632 1-47E610-43-1 Normally closed Main Steam Sample System manual isolation valve 1-FCV-1 -74 1-47E801-2 Main Turbine Stop Valve for Steam Line A 1-FCV-1-78 1-47E801-2 Main Turbine Stop Valve for Steam Line B 1-FCV-1 -84 1-47E801 -2 Main Turbine Stop Valve for Steam Line C 1-FCV-1-88 1-47E801-2 Main Turbine Stop Valve for Steam Une D 1-FCV-1 -61 1-FCV-1-62 1-FCV-1 -63 1-FCV-1 -64 1-FCV-1-65 1-47E801-2 Main Steam Bypass Valve Chest 1-FCV-1 -66 1-FCV-1-67 1-FCV-1 -68 1-FCV-1 -69 1-FCV-1-127 1-47E801-2 RFP Turbine A Stop Valve 1-FCV-1-135 1-47E801-2 RFP Turbine B Stop Valve 1-FCV-1-143 1-47E801-2 RFP Turbine CStop Valve 1-FCV-6-153 1-47E807-2 Normally closed motor operated isolation valve - RFP Turbine A 1-FCV-6-155 1-47E8O7-2 Normally closed motor operated isolation valve - RFP Turbine B 1-FCV-6-157 1-47Es07-2 Normally closed motor operated isolation valve - RFP Turbine C 1-FCV-6-122 1-47E807-2 Normally closed motor operated isolation valve - RFP Turbine A 1-FCV-6-127 1-47Es07-2 Normally closed motor operated isolation valve - RFP Turbine B Page 2-4 FACILITY RISK CONSULTANTS, INC.
______ - - -.'- -- - .. . - -60. - . - -. .- '.. I, -
Table 2-2 (Cont.)
BFN UNIT 1 MSIV LEAKAGE BOUNDARY POINTS Leakage Flow Boundary Diagram/ Comment Point Drawing 1-FCV-6-132 1-47E807-2 Normally closed motor operated isolation valve - RFP Turbine C 1-PCV-1-151 1-47E801-2 Normally open air operated isolation valve - SJAE A 1-PCV-1-166 1-47E801-2 Normally open air operated isolation valve - SJAE A 1-6-826 1-47E805-3 Check Valve - SJAE A 1-PCV-1 -153 1-47E801 -2 Normally open air operated isolation valve - SJAE B 1-PCV-1-167 1-47E801-2 Normally open air operated isolation valve - SJAE B 1-6-822 1-47E805-3 Check Valve - SJAE B 1-SHV-1-741 ... New manual isolation valve for Off-Gas Preheater - per DCN 51112 1-CKV-1 -742 ... New check valve for Off-Gas Preheater - per DCN 51112 1-SHV-1-743 --- New manual isolation valve for Off-Gas Preheater - per DCN 51112 1-CKV-1 -744 --- New check valve for Off-Gas Preheater - per DCN 51112 1-FCV-73-6B 1-47E812-1 Normally open air operated isolation valve - HPCI 1-FCV-71-6B 1-47E813-1 Normally open air operated isolation valve - RCIC 1-12-635 1-47E815-3 Normally closed manual isolation valve - Aux. Boiler 1-12-637 1-47E815-3 Normally closed manual isolation valve - Aux. Boiler 1-12-623 1-47E815-3 Normally closed manual isolation valve - Aux. Boiler 1-12-625 1-47E815-3 Normally closed manual isolation valve - Aux. Boiler 1-12-824 0-47E815-1 Normally closed manual isolation valve - Aux. Boiler 1-FCV-1-145 1-47E807-2 Normally closed motor operated isolation valve - Steam Seal Regulator 1-FCV-1-154 1-47E807-2 Normally closed motor operated isolation valve - Steam Seal Regulator 1-PCV-1 -147 1-47E807-2 Air operated pressure control valve/relief valve - Steam Seal Regulator 1-FCV-6-1 00 1-47E807-1 Normally closed motor operated isolation valve - Stop valve above seat drains 1-FCV-6-101 1-47E807-1 Normally closed motor operated isolation valve - Stop valve above seat drains 1-FCV-6-102 1-47E807-1 Normally closed motor operated isolation valve - Stop valve above seat drains 1-FCV-6-103 1-47E807-1 Normally closed motor operated isolation valve - Stop valve above seat drains Condenser A --- The condenser is the ultimate boundary for the MSIV leakage path.
Condenser B --- The condenser is the ultimate boundary for the MSIV leakage path.
Condenser C ... The condenser Is the ultimate boundary for the MSIV leakage path.
Misc. 1-47E801-1 Miscellaneous test, vent, drain and instrument connections Components 1-47E801-2 Page 2-5 F.a rit irv Pisq rnNfqi t TA NT-q 1Nc-
Figure 2-1: BFN Unit 1 MSIV Seismic Verification Boundary Page 2-6 FACILITY RISK CONSULTANTS, INC.
BFN UNIT I MSIV SEISMIC VERIFICATION WALKDOWN BOUNDARY Figure 2-2: BFN Unit 1 MSIV Seismic Walkdown Scope by Subsystems - Package 1 Page 2-7 FACILITY RISK CONSULTANTS INC.
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BFN UNIT 1 MSIV SEISMIC VERIFICATION WALKDOWN BOUNDARY Figure 2-10: BFN Unit 1 MSIV Seismic Walkdown Scope by Subsystems - Package 9 Page 2-15 FACILITY RISK CONSULTANTS} INC.
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4 BFN UNIT 1 MSIV SEISMIC VERIFICATION WALKDOWN BOUNDARY Figure 2-13: BFN Unit 1 MSIV Seismic Walkdown Scope by Subsystems - Package 12 Page 2-1 8 FACILITY RISK CONSULTANTS, INC.
BFN UNIT 1 MSIV SEISMIC VERIFICATION WALKDOWN BOUNDARY Figure 2-14: BFN Unit 1 MSIV Seismic Walkdown Scope by Subsystems - Package 13 Page 2-19 FACILITY RISK CONSULTANTS, INC.
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I BFN UNIT 1 MSIV SEISMIC VERIFICATION WALKDOWN BOUNDARY Figure 2-16: BFN Unit 1 MSIV Seismic Walkdown Scope by Subsystems - Package 15 Page 2-21 FACItuTY RISK CONSULTANnS, INC.
- 3. SEISMIC VERIFICATION WALKDOWN Very few components of nuclear plant systems are unique to nuclear facilities. Nuclear plant systems include piping, tubing, conduit, and many other items that are common components of conventional power plants and industrial facilities. Seismic experience data based methods have been developed which address the question of adequacy of seismic performance of equipment and commodities not designed, procured and installed to current nuclear seismic criteria (Reference 7-3). By reviewing the performance of facilities that contain equipment similar to that found in nuclear plants, conclusions can be drawn about the performance of nuclear plant equipment and associated components during and after earthquake events. Extensive work has been performed documenting the performance of power plant equipment and the common sources of seismic damage to equipment and piping (References 7-1, 7-4 and 7-5). These and other similar studies form the basis for the walkdown.
Equipment, piping and tubing systems in the seismic experience data base have performed very well in earthquakes, even though they were typically designed for deadweight and operating loads only, with little or no consideration for seismic loads (Reference 7-5).
Earthquake experience data base methods provide the basis for review of the main steam piping and equipment.
The catalog of earthquake experience data on which the walkdown screening evaluation was based includes the El Centro Steam Plant, Valley Steam Plant, Glendale Power Plant, Burbank Power Plant, Humboldt Bay Plant, PALCO Co-generation Plant, Coolwater Plant, Ormond Beach Plant, Moss Landing Steam Plants and several other facilities affected by the 1987 Whittier earthquakes, and the 1992 Cape Mendocino (Humboldt Bay) and Landers-Big Bear earthquakes in Califomia.
A comparison of the selected earthquake experience database site spectra with the Browns Ferry Design Basis Earthquake (DBE) ground spectrum, amplified by 1.6 to account for soil-founded structures such as the Turbine Building, is shown in Figure 3-1. The earthquake experience database plants have experienced strong motions substantially in excess of the soil-amplified Browns Ferry DBE (1.6 x DBE).
Page 3-1 FAILITY RISK CONSULTANTS, INC.
Further discussion on the appropriate seismic demand used for Turbine Building is presented in Section 5.1.1. Also, additional details regarding the earthquake experience database plants can be found in Appendix D, Section 4.1, of the BWROG Report GE NEDC-31 858P (Reference 7-1).
3.1 Seismic Verification Review Guidelines Various design attributes of the as-installed scope of equipment, piping, and tubing were reviewed and evaluated by the Seismic Walkdown Teams to insure that the BFN installations are representative of data base design practice and that components are free of known seismic vulnerabilities. Earthquake experience has identified conditions that have resulted in failure of piping and tubing systems and components. The conditions evaluated in this walkdown review included:
- Piping, Pipe Support and Equipment Design Attributes
- Seismic Anchor Motion Issues
- Seismic Interaction Issues (Il/I & Proximity)
- Boundary Valve Design Attributes 3.2 Piping, Pipe Support and Equipment Attributes Earthquake experience data base representation of the piping and tubing systems can be verified by the following design and installation attributes:
Piping and tubing installations are in conformance with industry-standard practices (e.g., ANSI B31.1 spans for piping, standard industrial supports for piping and tubing).
Piping or tubing system does not display known seismic vulnerabilities or exhibit seismically sensitive characteristics.
Page 3-2 FACILTY RISK CONSULTANTS, !NC.
3.2.1 Piping and Pipe Support Design Attributes The Seismic Walkdown Team reviewed the piping and tubing systems for conditions associated with past poor performance. Support failure in past earthquakes has rarely resulted in piping failure unless multiple supports over long runs fail. Support types which have demonstrated poor seismic performance include:
- One-way bracket or stanchion supports that could allow pipe to slide off the support and fall.
- Supports attached by beam clamps without restraining straps.
- Short threaded rods that are fixed against rotation may be vulnerable to low cycle fatigue.
- Other support integrity issues such as damaged, deteriorated or altered parts that could result in non-ductile behavior or significant weakness in the load path.
Rod hanger attachments that act as pinned members (e.g. clevis, eyes, etc.) and welded structural steel pipe supports have not been observed in seismic events to be vulnerable unless there are major design or construction flaws in their attachment to the supporting structure.
3.2.2 Vulnerable Piping Joints Threaded piping connections, unsupported bellows or expansion joints and mechanical joints (e.g., mechanical type couplings, bell and spigot joints, etc.) have been observed in past earthquakes to be more vulnerable to seismic loads than welded piping joints. The Seismic Walkdown Team reviewed flexibly supported piping to identify segments which contain weak or brittle joints. Bolted flange connections have performed well and are not considered to be weak or vulnerable to seismic loads.
3.2.3 Other Potential Seismic Vulnerabilities Other piping and tubing design attributes which are inconsistent with good design practice and which may contribute to poor seismic performance include:
Page 3-3 FACILTY RISK CONSULTANTS, IIVC.
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Piping with dead weight support spacing greatly in excess of ANSI/
ASME B31.1 suggested spans (Reference 7-16). Tubing with excessive sagging or support spans greatly in excess of 6'-O'.
- Heavy in-line masses (e.g., valves, accumulators, filters, strainers).
- Piping inadequately restrained adjacent to expansion joints.
- Non-standard fittings, such as mitered elbows or unreinforced branch connections, or unusual attachments that could cause excessive localized stresses.
- Presence of severe corrosion.
Walkdowns reviewed piping and supports to identify designs which contained any of these attributes.
3.3 Equipment Design Attributes The equipment reviewed in the seismic verification walkdown includes the main condenser and equipment which acts as terminal anchor points for piping and tubing systems. Typical equipment components include heat exchangers, vessels and some measuring instrumentation such as transmitters, and gauges. In general the walkdown review included:
- Review the equipment for known failure modes and sources of seismic damage which may affect the seismic performance of the equipment and sub components.
Check for unusual or non typical arrangements of the devices within the equipment or of items external to the equipment.
- Assess the anchorage and presence of an adequate load path. For instrumentation, this included evaluation of the instrument rack and/or other terminal equipment.
Page 3-4 FACILITY RISK CONSULTANTS, INC.
The details of the review varied according to the type of equipment and location within the plant. The extent of review and information gathering for evaluation of Seismic Verification Boundary components, equipment required for structural integrity, etc. were determined based on the judgment and experience of the Seismic Walkdown Team.
3.4 Anchorage Design Attributes Anchorage of equipment and supports were reviewed for adequacy on a visual basis. Details of piping and equipment anchorage were documented in the walkdown notes only as required for review of outliers.
3.4.1 ExDansion Anchor Bolt Review Guidelines Visual review of expansion anchor bolts considered the following:
- The concrete appears sound with no significant cracks in the vicinity of the anchor bolt.
- Gaps between the equipment base and the concrete surface were negligible.
- The bolt spacing is greater than about 10 times the bolt diameter.
- The distance between the bolt and any free concrete surface is greater than approximately 10 times the bolt diameter.
3.4.2 Welded Anchorage Review Guidelines Welded anchorages were reviewed for adequacy, good installation practices and workmanship on a visual basis. This visual review included:
Review for weld burn-through on thin sections.
Review of weld thickness, against the thickness of thinner part being connected.
Review for plug welds subjected to tension loads.
Page 3-5 FACLITY RISK CONSULTANTS, INC.
3.5 Seismic Anchor Movement The experience data base includes instances of seismic damage to piping, tubing and supports that were attributed to seismic anchor movement. Damage was the result of excessive movement of terminal end equipment, differential movement between supports in adjacent buildings, and excessive movements imposed on branch lines by flexible headers.
As a result of these observed vulnerabilities, the following attributes were evaluated by the Seismic Walkdown Team during the piping walkdown:
- System configurations at building joints and between buildings were reviewed to insure adequate piping system flexibility to accommodate seismically-induced differential building movement.
- Fittings which can be adversely affected by seisrtically-induced differential movement (e.g., bellows, etc.) were evaluated for adequate flexibility. This included conditions with rigid connections at multiple structures.
- Piping attached to unanchored or poorly anchored equipment was considered an outlier. Stiff piping attached to flexible equipment was evaluated to verify that the piping will not act as an equipment anchorage.
- Conditions where stiffly supported branch lines were attached to flexibly-supported (e.g., rod-hung) mainlines or headers were identified as outliers. The Seismic Walkdown Team evaluated these configurations for potential damage due to seismically-induced differential movement.
3.6 Seismic Interaction Review (IVI and Proximity)
The seismic interaction review was a visual inspection of structures, piping, or equipment adjacent to the components under evaluation. The seismic interaction review identified seismically induced failures (11/1) and displacements of adjacent structures, piping, or Page 3-6 FACLTY RISK CONSULTANTS, INC.
-'- p equipment (proximity) that could adversely affect the required seismic performance of the system and components under consideration.
The Seismic Walkdown Team identified and evaluated all credible and significant interaction hazards in the immediate vicinity of the item being evaluated. Evaluation of interaction effects considered detrimental effects on the capability of equipment and systems to function, taking into account equipment attributes such as mass, size, support configuration, and material hardness in conjunction with the physical relationships of interacting equipment, systems, and structures. In the evaluation of proximity effects involving overhead or adjacent equipment failure and interactions, the effects of intervening structures and equipment which would preclude impact were also considered.
Unusual circumstances and details have led to damaging interaction of plant features during past earthquakes. In the interaction review, the Seismic Walkdown Team looked for unusual impact situations, and lack of proper anchorage or bracing of adjacent equipment. All credible interactions that could affect the required performance of the piping, tubing and equipment reviewed were identified and documented, as appropriate.
3.7 Seismic Verification Boundary Valves Screening guidelines are provided for valves which are either relied upon to establish, or are within the Seismic Verification Boundary. The guidelines are consistent with the SQUG Generic Implementation Procedure (GIP, Reference 7-3) and include provisions for air-operated diaphragm valves, spring-operated pressure relief valves and piston-operated valves of light-weight construction. Screening guidelines for motor-operated valves and substantial piston-operated valves are also provided. Evaluation of valves included review of power and control utilities to insure adequate slack is provided to accommodate anticipated seismic motions. Supports located on the valve operator were reviewed to insure that they were accompanied by supports on the valve body or piping adjacent to the valve body.
Reviews were also performed to insure that the valve body and operator were supported by a common structure to prevent differential displacement. Piping or tubing less than 1-inch in diameter with in-line eccentric masses such as motor or air operated valves were checked for support at or near the valve.
Page 3-7 FACILITY RISK CONSULTANTS, INC.
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3.8 Walkdown BFN-1 MSIV seismic ruggedness verification boundary was divided into 15 subsystems or portions for walkdown purposes (see Section 2.2 above). Seismic verification walkdown of the main steam lines, various drain paths, and associated components and appendages within the Seismic Verification Boundary were conducted as part of BFN Unit 1 restart project, and were performed by Seismic Walkdown Teams consisted of Messrs. John 0.
Dizon, Stephen J. Eder, Robert D. Hookway and Michael W. Whited of FACILITY RISK CONSULTANTS, Inc. All of the MSIV Seismic Verification Walkdown Team members are degreed engineers; each has over ten to twenty years of experience in structural or mechanical engineering and/or earthquake engineering application to nuclear power plants, and is familiar with the earthquake experience methodology.
The above engineers have performed complete MSIV Seismic Verification Walkdowns in accordance with the recommendations of the BWROG Report GE NEDC-31 858P at several other plants, including BFN Units 2 and 3.
3.9 Documentation The seismic verification walkdowns and evaluations were performed in accordance with applicable Walkdown Instruction WI-BFN-0-CEB-07 (References 7-6), and following the guidelines contained in the BWROG Report GE NEDC-31 858P (References 7-1). Screening evaluation guidelines were based on TVA Design Criteria BFN-50-C-7306 (References 7-7) for Class II piping and components. Walkdown evaluations utilized existing plant documentation, as available, including:
Systems flow diagrams or P&ID's identifying piping and equipment within the verification review boundaries; Piping isometric drawings; Piping support sketches and piping layout drawings, as needed; Page 3-8 FACILITY RISK CONSULTANTS, INC.
In-plant screening tools such as piping deflection charts, pipe flexibility charts, standard support hardware capacities, anchorage capacities, and others (Reference 7-8), as applicable; Walkdown and evaluation results from previous MSIV seismic ruggedness programs conducted for BFN Units 2 and 3 (References 7-9 and 7-10).
The walkdown review of piping and supports was primarily visual for qualitative attributes of the systems. Only physical system attributes which could be visually verified with available access, and without system disassembly were reviewed. Where indicated, additional details of the system design, installation and construction were collected and are documented on piping isometrics and in walkdown field notes, as appropriate.
Conditions which did not conform to the walkdown screening guidelines or which were judged by the Seismic Walkdown Team to require further review were identified and documented as potential outliers in the Potential Outlier Sheet (POS). Some photographs were taken for informational use in subsequent evaluation phases of the project.
Results of the walkdown for each of the 15 portions or subsystems within the BFN-1 MSIV seismic verification boundary are documented in the respective walkdown data package (WDP). A listing of the walkdown data packages with unique identifier number can be found in Appendix A of this report.
Page 3-9 FAcILuY RISK CONSULTANTS, INC.
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- 4. WALKDOWN OPEN ITEMS Screening walkdown evaluations for the 15 subsystems or portions included in the BFN-1 MSIV Seismic Verification Boundary were performed in accordance with Walkdown Instruction WI-BFN-0-CEB-07 (Reference 7-6). Screening evaluations focused on certain key attributes of the associated piping and components, as discussed earlier in Chapter 3, to ensure pressure boundary integrity. Available screening tools (see Reference 7-8) such as seismic deflection estimates and charts for various plant features, pipe flexibility and seismic anchor movement evaluation charts, support and anchorage capacity screening charts, and others, that are developed based on the TVA Design Criteria BFN-50-C-7306 (Reference 7-7) were utilized during the walkdowns.
Results of in-plant screening walkdown evaluations for the 15 subsystems included in the MSIV seismic boundary are documented in details in the respective walkdown data packages (WDP's). Conditions which did not conform to the walkdown screening guidelines or which were judged by the Seismic Walkdown Team to require further review were identified and documented as potential outliers in the Potential Outlier Sheet (POS) which are also included in the WDP's. A listing of the walkdown data packages with unique identifier number can be found in Appendix A of this report.
The following sections provide a brief discussion of the walkdown evaluations and identified open items or potential outliers associated with each of the 15 subsystems.
4.1 Main Steam Drain Line - Turbine Bldg. Main Steam Tunnel (Pkg. 01)
The Main Steam Drain Line, which is the primary MSIV leakage drain path, originates in the Reactor Building MSIV Vault (Package 03, Section 4.3) and continues through the Turbine Building Main Steam Tunnel (Package 01, this section). The drain then enters a guard pipe which is embedded below the slab and crosses below the gap between the Steam Tunnel and the Turbine Pedestal Structure, continues into the Turbine Structure and terminates at the main condenser (Package 14, Section 4.14). The Main Steam Drain Line also has a 4-inch diameter vent line that begins in the Turbine Building Main Steam Tunnel, rises up to Elevation 593' and continues through the Turbine Structure to the south end of Condenser A where it runs down to rejoin the Main Steam Drain Line Page 4-1 FACILTY RISK CQNSULTANSt INC.
before it enters the condenser (also included in Package 14). The lines are shown on Flow Diagram 1-47E801 -1.
The portion of piping in this package included the Main Steam drain lines in the Turbine Building Steam Tunnel on Elevation 565', i.e., from the blowout panels at column line N to just north of column line K. Several potential outliers were identified, as listed below:
Main Steam Drain Taps 1-1 Flexibility concerns associated with the MS lines and Drains.
MOV 's 1-FCV-1-57 & 1-2 Excessive mass and extended valve operators.
1-FCV-1-58 Main Steam and HPCI 1-3 Flexibility concerns associated with the MS and HPCI Drains.
Drains Main Steam Bypass 1-4 Proximity interaction concerns between bypass line and Drain adjacent 2-way support.
4.2 Main Steam - Turbine Building (Package 02)
This package includes the four Main Steam Lines (A, B, C & D) inside the Turbine Building that penetrate from the blowout panels at column line N and terminate at the Main Steam Stop Valves 1-FCV-1 -74, -78, -84 and -88, which are considered as seismic boundary valves. These lines are shown on Flow Diagrams 1-47E801-1 and 1-47E801-2.
The Main Steam lines come from the outboard MSIV's and are anchored downstream inside the Reactor Building MSIV vault. These lines are vertically supported by variable springs along their length until the Stop Valves, which are supported by vertical struts.
Two (2) potential outliers were identified and are listed below:
MS Balancing Header } 2-1 1Upper pipe clamp nut is missing. I MS Turbine 2-2 MS Stop/Control Valves are not covered in the valve Stop/Control Valves screening criteria, thus, seismic adequacy of these valves needs to be verified.
Page 4-2 FACLITY RtSK CONSULTANTS, INC.
4.3 Main Steam Drain Line - Reactor Building Main Steam Vault (Pkg. 03)
Included in this package are the four Main Steam Lines which originate at the containment penetrations and run to the outboard MSIV's 1-FCV-1 -15, -27, -38 and -52, which are considered as seismic boundary valves. The Main Steam Lines then run to anchors (one for each line) embedded in a reinforced concrete wall. The lines continue through the MSIV Vault and enter the Turbine Building through the blowout panel wall. The Main Steam Piping is shown on Flow Diagram 1-47E801 -1. The Main Steam Lines in the MSIV Vault are seismically analyzed from the containment penetrations to the anchors in the concrete wall, which are also the only supports for the lines. The Main Steam lines beyond the anchors were found to be seismically adequate. The only potential outlier identified during the walkdown relates to the seismic adequacy of the MSIV's (not bounded by seismic experience data).
Also included in this package is a single 3-inch diameter steam drain that runs from its containment penetration through an anchor and to the normally closed outboard containment isolation valve for Primary Containment steam drains, 1-FCV-1 -56, which is a seismic boundary valve. The piping continues through the Reactor Building MSIV Vault and exits to the Turbine Building through the blowout panel wall at the north end of the Vault (Package 04, Section 4.4). This drain line is shown on Flow Diagram 1-47E801 -1.
This drain line was found to be seismically adequate.
In the Reactor Building MSIV Vault, the 3-inch diameter drain line is primarily supported on stanchion supports that provide only vertical restraint and are detailed to accommodate thermal growth of the piping. Piping and supports in this area are considered to be acceptable. Valve 1-FCV-1 -56 was identified as a potential outlier due to its extended valve operator not meeting the screening criteria for operator height.
MSIV 's 1-FCV-1-15, 3-1 MSIV 's are not covered in the valve screening criteria, thus,
-27, -38 & -52 seismic adequacy of these valves needs to be verified.
MOV 1-FCV-1-56 3-2 Excessive mass and extended valve operators.
Page 4-3 FACILITY RISK CONSULTANTS, INC.
4.4 HPCVRCIC/Aux. Boiler Drains (Package 04)
HPCI and RCIC Steam Drains route steam leakage from the HPCI and RCIC Steam Supply Lines to the Main Steam Drain Line in the Turbine Building Steam Tunnel. These line segments are shown on Flow Diagrams 1-47E812-1, 1-47E813-1, and 1-47E801-1.
The RCIC Steam Drain Line originates in the Northwest comer room of the Reactor Building (Boundary valve 1-FCV-071 -06B) and is routed over the top of the Torus to its connection into the HPCI Steam Drain Line just below the penetration through the floor into the MSIV Vault. The HPCI Steam Drain Line starts in the HPCI room (Boundary valve 1-FCV-073-06B) and is routed above the Torus to the point where it penetrates the floor into the MSIV Vault. Both drains have branches near their points of origin that lead from Auxiliary Boiler Drain piping. The Auxiliary Boiler piping is shown on Flow Diagrams 0-47E815-1 and 1-47E815-3. The normally closed manual isolation valves (1-12-824, 1-12-635/637 and 1-12-623/625) that serve as the seismic verification boundary for the Auxiliary Boiler drains are located in the HPCI room and the Northwest corner room of the Reactor Building. These drain lines were found to be generally well supported in these areas.
From the penetration above the Torus through the floor of the Reactor Building MSIV Vault, the HPCI piping is routed through the MSIV Vault and into the Turbine Building Steam Tunnel through the blowout panels. The HPCI Steam Drain continues alongside the Main Steam Drain to the point where it connects with the Main Steam Drain between valves 1-FCV-1 -58 and -59. The HPCI piping is supported throughout the MSIV Vault and the Turbine Building Steam Tunnel on floor-mounted, custom-fabricated supports which act as horizontal and vertical rigid restraints and are detailed to accommodate thermal growth of the piping.
Several potential outliers were identified on these drain lines. Conditions include pipe overspan, support deficiencies, proximity interactions and piping flexibility issues, as listed below:
RCIC Drain 4-1 Excessive pipe overspan condition.
RCIC Drain 4-2 Pipe support with questionable capacity.
Aux. Steam Boiler Drain 4-3 Missing rod hanger hardware (eye nut).
Page 4-4 FACILITY RISK CONSULTANTS, INC.
-- z' - - .. I I . .' . - ___
HPCI Drain 4-4 Multiple pipe overspan conditions on the 1"6 drain line.
HPCI Line 4-5 Multiple pipe overspan conditions on the 2"6 line.
HPCI Line and 4-6 Proximity interaction of the 2'6 HPCI line and the valve.
Valve 1-73-222 HPCI Drain 4-7 Several floor-mounted supports were noted to be damaged (potentially due to thermal effects). Also, anchor bolt spacing violations on these typical supports.
HPCI Drain 4-8 Flexibility concern on 2"0 HPCI line due to differential displacement between RB and TB 4.5 Main Steam Pressure Transmitters PT 1-72, 76, 82, 86 & 93 (Pkg. 05)
The Main Steam Pressure Transmitters which tap off from the four Main Steam Lines upstream from the Main Steam Manifold (Balancing Header) and downstream from the Stop Valves are shown on Flow Diagram 1-47E801-2. The instrument lines to the pressure transmitters are routed on 1/2-inch diameter stainless steel piping from the pipe taps off the Main Steam Lines and the Manifold passing under the steel grating at Elevation 606'-3" in the Turbine Building to a penetration through the south wall of the area. The piping then runs to Instrument Racks 1-25-112 and 1-25-113C located on the south corridor of the Turbine Building at Elevation 586'.
Support configurations underneath the steel grating are generally rigid, consisting of strut channel welded to the underside of the grating beams. As such, several potential outliers were identified on these lines, mainly due to the limited flexibility in the piping and support configuration to accommodate the Main Steam header movements, as evidenced from the loose or missing clamps on the existing strut channel supports, broken overhead welds, and bent pipes noted during the walkdown. The identified potential outliers are tabulated below.
Page 4-5 FACILITY RISK CONSULTANTS, INC.
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SS Piping to PT 1-93 5-5 l Overspan condition due to broken support.
SS Piping to PT 1-93 5-6 l Flexibility concern on Y2"6 piping along the wall.
4.6 Main Steam Sample Lines to Sampling Station (Package 06)
The Main Steam Sample Lines tap off from each of the four Main Steam Lines upstream of the Main Steam Stop Valves. The lines are shown on Flow Diagram 1-47E801-1 and Mechanical Control Diagram 1-47E610-43-1. The sample lines begin as a 1-inch diameter piping and transition to one-half-inch diameter tubing. The tubing is routed through the Turbine Building and exits the area through the south wall to Instrument Rack 1-25-149 in the Turbine Building south corridor, Elevation 586'. On the rack, the four tubes connect into a single tube with three branches. Seismic verification boundary valves 1-43-631 and 1-43-631 A terminate two of the branches on the rack. One one-quarter-inch diameter tube runs from the rack through the constant temperature bath and on to the grab sample station and normally closed valve 1-43-632. The tubing is typically supported with tubing clamps attached to strut channels mounted to the wall.
Also included in this package are the Pressure Sensing lines for PT-1 6A and B which tap off from the 1-inch diameter piping taps for the Steam Line A sample lines. This piping transitions to tubing and runs west into the Turbine Lube Oil Pump area. The transmitters are attached to wall-mounted steel plate bracket supports and are enclosed by a mesh cage. The tubing lines are adequately supported using U-bolts and fabricated steel angle brackets mounted to the wall with expansion anchors.
Four (4)potential outliers were identified which include unanchored terminal equipment, broken support potentially due to inadequate pipe flexibility, and support hardware deficiencies. These potential outliers are tabulated below.
Sample Line A 6-1 Missing tubing clamp.
Sample Station 6-2 Unanchored constant temperature bath cabinet.
Sample Line A to 1-PT- 6-3 Flexibility concerns on sample lines to 1-PT-1-1 6A/B, involving 1-1 6A/B broken support.
Sample Line A to 1-PT- 6-4 Missing nuts on U-bolt support.
1-1 6ANB Page 4-6 FACITY RISK CONSULTANTS, INC.
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4.7 Main Steam Bypass (Package 07)
The Main Steam Bypass Valve Chest is located on a loop from the Main Steam Header upstream of the Stop Valves. The Bypass Valves and piping are shown on Flow Diagram 1-47E801-2. This package includes only the portion of Main Steam piping from the Main Steam Manifold to and including the Bypass Valve chest assembly which is a seismic boundary valve (1-FCV-1 -61 to -69). The bypass valve chest is supported with trapeze spring hangers on each end of the valve assembly. An outlier pertaining to the seismic adequacy of the Main Steam Bypass Valve was identified, as listed below.
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MS Bypass Valve 7-1 MS Bypass valve chest is not covered in the valve screening Assembly criteria, thus, its seismic adequacy needs to be verified.
4.8 Main Steam Stop Valve Above Seat Drains (Package 08)
The 1-inch diameter Main Steam Stop Valve Above Seat Drains originate at each of the four stop valves and are shown on Flow Diagram 1-47E807-1. The drains run down through the stop valve platform grating at Elevation 601'-6" and continue east to the four verification boundary valves (1-FCV-6-1 00 through -103) under the east end of the platform. The piping is typically supported with rod hangers attached from steel angle members welded to the platform steel. The verification boundary valves have operators which are independently supported by rod hangers to the platform supporting beam.
Several potential outliers were identified for this system; one has to do with proximity interaction concern and the rest are maintenance items. The potential outliers are tabulated below.
1-FCV-6-101, -102 & 8-1 Proximity interactions of valve yokes and beam.
-103 SV-D Above Seat Drain 8-2 Broken pipe strap on support.
Piping 1-FCV-6-1 00 8-3 Broken flex conduit connection to operator.
Page 4-7 FACILITY RISK CONSULTANTSt INC.
4.9 Main Steam to Steam Seal Regulator (Package 09)
Steam is taken from the west side of the Manifold between the Stop and Bypass valves and routed to the Steam Seal Feed valves. The line is shown on Flow Diagrams 1-47E801 -2 and 1-47E807-2. This package includes the 4-inch diameter steam line off the Manifold to the air operated pressure control/relief valve 1-PCV-1 -147 (Steam Seal Regulator) and the normally closed motor operated isolation valves 1-FCV-1 -145 (Steam Seal Bypass Valve) and 1-FCV-1 -154 (Aux. Steam Isolation Valve) which are all considered as seismic boundary valves. Air instrumentation line for valve 1-PCV-1 -147 routed to Panel 1-25-102 located along the west corridor wall at Elevation 617' of the Turbine Building is also included in this package.
Several potential outliers were identified during the walkdown of this system, including valve performance issues, piping overspan, pipe flexibility and falling interaction concems.
The potential outliers are listed below.
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De"s~cripId ' Penia~tirpdto Main Steam to 9-1 Overspan condition on the 4"6 MS line to the supply to steam 1-FCV-1-146 seal MOV 1-FCV-1-146.
Main Steam to 9-2 Loose rod hanger eye nuts and disengaged rod.
1-FCV-1 -146 1-PCV-1-147 Air Tubing 9-3 Inadequate flexibility for air line connected to the top of valve.
1-PCV-1 -147 9-4 Falling interaction concern from nearby block wall at El. 617'.
Instrumentation 1-PCV-1 -147 9-5 Extended valve operator.
1-FCV-1-146, -145 & 9-6 Extended valve operator, and substantial operator weight.
-154 4.10 Steam Supply to RFP Turbines (Package 10)
Steam is taken from the east side of the Manifold between the Stop and Bypass valves and routed to the turbine drives for the Reactor Feed Pumps at Elevation 617'. The lines are shown on Flow Diagrams 1-47E801-2 and 1-47E807-2. The 6-inch pipe runs below the turbine operating deck from the Main Steam line and penetrates up through the turbine deck (El. 617') where it branches to 4-inch diameter and into the RFP rooms (A, B
& C). The line reduces to 3-inch diameter and continues back down through the turbine deck floor penetration to the SJAE's (Package 11, Section 4.11) and Off-Gas Preheaters Page 4-8 FACILTY RISK CONSULTANTS, INC.
(Package 12, Section 4.12). The line is typically supported on one-way stanchion supports which are susceptible to sliding off the supports during a DBE seismic event.
The 4-inch branch lines that enter into the RFP rooms terminate at the respective RFP Turbine HP Stop Valves, 1-FCV-6-127, -135 and -143 (RFP-A, -B and -C), which are seismic boundary valves. In addition, HP Stop Valve Above Seat Drain Valves 1-FCV 122, -127 and -132, as well as HP Steam Isolation Valves 1-FCV-6-153, -155 & -157 are also considered as seismic boundary valves. Instrument tubing to Panels 1-25-1 0OA and 1-25-101 B located on the turbine deck are also included in this package.
As such, several supports on the line were identified as potential outliers for further evaluation of the as-installed configuration. Other potential outliers identified include piping overspan, large in-line masses, and seismic interaction concerns. Potential outliers for this system are tabulated below.
Subsytem PS Decripton o Steam Supply Line 10-1 Stanchion supports for the steam supply header at El. 617' (Total of 6).
1 support for the 6"6 line near H/T4; 2 supports for the 3"6 line near D/T4; and 1 support for each of the 4"6 branch line to RFP Turbine compartments.
Steam Supply Line 10-2 Turbine Bldg. overhead crane.
RFP Stop Valve Above 10-3 Valve operators with large eccentric mass on Y2"6 and 4"6 Seat Drains lines.
Tubing to Pi 1-134 10-4 Missing or broken clamps (2 places) on tubing to PI 1-134.
Also, tubing is bent and sagging.
Steam Supply Line 10-5 Missing nut on rod hanger support.
Steam Supply Line 10-6 Overspan conditions on the 6"i steam supply line.
4.11 Steam Supply to Steam Jet Air Ejectors (Package 11)
The Steam Supply line to the Steam Jet Air Ejectors (SJAE's) Is shown on Flow Diagram 1-47E801-2. The portions of piping included in this package begin with the 3-inch Steam Supply line that drops from the turbine deck floor penetration near column lines T3 and D (past RFP rooms, refer to Section 4.10 above), runs north along the Off-Gas pipe chase, Page 4-9 FACILITY RISK CONSULTANTS, INC.
branches eastward into the SJAE rooms and terminate at the seismic boundary valves 1-PCV-1-151 & -166 in SJAE Room A and 1-PCV-1-153 & -167 in SJAE Room B.
Potential outliers identified for this package, consist of mainly piping overspan on 1-inch diameter steam drains, are listed below.
SJAE'sl1A &1B 11-1 Need to verify the seismic adequacy of SJAE anchorage.
MOV 1-FCV-6-114 11-2 Broken flex conduit connection to valve motor operator.
MS Supply to SJAE's 11-3 Overspan condition on the 3"6 supply line.
MS Supply Drain at 11-4 Overspan condition on the 1"6 steam supply drain line SJAE 1B downstream of 1-FCV-1-114.
Steam Trap Drain at 11-5 Overspan condition on the 1"0 steam trap drain between 1-SJAE 1A FCV-1-172 and 1-PCV-1-166.
Steam Trap Drain at 11-6 Overspan condition on the 1"6 steam trap drain between 1-SJAE I B FCV-1-173 and 1-PCV-1-167.
4.12 Steam Supply to Off-Gas Preheaters (Package 12)
The Steam Supply line to the Off-Gas Preaheaters is also shown on Flow Diagram 1-47E801-2, and is a continuation of the same steam supply line from RFP's and SJAE's above. The portion of piping included in this package consists of the 2-inch diameter line (reduced from the 3-inch diameter past the SJAE's) along the Off-Gas pipe chase and enters eastward into the Off-Gas room through a masonry block wall. It is to be noted that new manual isolation valves 1-SHV-1 -741 & -743 and check valves 1-CKV-1 -742 & -744 will be installed in Unit 1 per DCN 51112, consistent with the modifications performed for Units 2 and 3 as part of their MSIV seismic ruggedness programs. These new sets of valves are now the boundary points for the Off-Gas Preheaters.
The only potential outlier for this package, concerning the seismic falling interaction between the piping and its components and the nearby masonry block wall, will then be resolved by this plant modification.
Descripto fPotnilOtir~rdt Steam Supply Lines to 12-1 Lines penetrate into the Off-Gas Preheater room through Off-Gas Preheaters masonry block wall.
Page 4-10 FACILITY RISK CONSULTANTS, INC.
w e~Cb;r . . t- : : - -S . - -. -
4.13 SJAE's Drain to Condenser (Package 13)
The Steam Jet Air Ejectors (SJAE's) drain to Condenser is shown on Flow Diagrams 1-47E801-2 and 1-47E805-3. The drains originate in the SJAE rooms, at boundary valves 1-6-826 & 1-6-822 downstream of the SJAE's in Rooms A and B, respectively, and tie back into a single 1-1/2 inch diameter line which is routed through the Off-Gas pipe chase and connects into an 8-inch diameter collector pipe that is directly attached to the Condenser 1C shell. The collector pipe has several other drain lines attached to it, thus it was considered to be a potential outlier.
't..X OS .D...... t.....f......
- k. ...... X:* He :O . : ..*:: *:: ::-.*. . .*.B
- . . ;.*m * . {. > ,.*:.*
<. : if:: .:$.* .*2..*.* ~*..
SJAE's Drain to 13-1 Drain to condenser ties into a multi-system collector.
Condenser 1C 4.14 Main Steam Drain Line (Turbine Building) to Condenser (Pkg. 14)
This package includes the 4-inch diameter Main Steam Drain line downstream of normally closed motor operated 1-FCV-1 -59 on Elevation 565' in the Steam Tunnel which is the primary drain path to the Condenser. Past valve 1-FCV-1-59, the MS drain line drops into a small pipe chase, turns north and enters an 8-inch diameter guard pipe that is embedded below the slab. The drain line and the guard pipe continue under the gap between the Steam Vault and the Turbine Structure and re-enter the Turbine Building in the Moisture Separator Level Control Reservoir Room. Since the drain line runs through the guard pipe at ground level, differential building displacements will be minor and, based on a review of drawings, adequate flexibility has been provided to accommodate the movements. The guard pipe ends at the room wall and the drain line runs through the room, exits through the north wall into the condenser bay and then enters Condenser 1A.
Also included in the package is a 4-inch vent line which branches off the above mentioned 4-inch diameter MS drain between valves 1-FCV-1-58 and -59 and rises up to Elevation 593' where it crosses the building separation between the Turbine Building Steam Vault and the Turbine Pedestal Structure. The vent line has adequate flexibility to accommodate anticipated differential seismic displacements between the two structures.
The vent line continues on to the south end of Condenser 1A and runs down the gap Page 4-11 FACILITY RISK CONSULTANTS. INC.
between the concrete structure and the condenser before connecting back to the 4-inch MS drain prior to its entering the Condenser 1A.
The above lines are shown on Flow Diagram 1-47E-801-1. Support systems for these lines in the Turbine Building are representative of commercial design practice and were determined to be adequate during the walkdown. Potential outliers associated with this package are tabulated below.
Susstemi POSssy esscripio dig Desriptto~ Po Pteinitijaltutier Condition Main Steam Drain Line 14-1 Spring hanger rod detached from 4"i drain line.
to Condenser 1A 1-FCV-1-59 14-2 Broken flex conduit connection at base.
1-FCV-1 -59 14-3 Valve operator (motor) in close proximity to the side rail of steel ladder.
4.15 Condenser (Package 15)
Condensers are the ultimate heat sink for the MSIV alternate leakage treatment drain path. The Main Condenser anchorage was reviewed during the walkdown. Each of the three Condensers is mounted on five concrete pedestals, four at the comers and one in the center of the condenser. The concrete pedestals were observed to be in good condition. Confirmation of the condenser seismic capacity and anchorage adequacy is required to ensure its structural integrity during a DBE seismic event.
Condensers 1A,B& 15-1 Need to verify the seismic adequacy of the condenser and its IC anchorage configurations.
Page 4-12 FACILTY RISK CONSULTANTSt INC.
4.16 Summary of MSIV Seismic Walkdown Evaluations A total of fifty-four (54) potential outliers were identified for further evaluation and resolution from the seismic walkdown of the above 15 subsystems under BFN-1 MSIV seismic ruggedness verification program.
Detailed description of the potential outlier conditions, including as-built data, sketches and/or photos are documented in the corresponding walkdown data packages (WDP's) for the respective plant areas. A listing of all walkdown data packages generated under the BFN-1 MSIV seismic ruggedness verification program is provided in Appendix A.
Page 4-13 FACILiTY RISK CONSULTANTS, INC.
- 5. POTENTIAL OUTLIER RESOLUTION Potential outliers identified during the in-plant screening walkdowns, as documented in the respective Potential Outlier Sheet (POS) and tabulated in Table 4-1, were further evaluated for appropriate resolution. Further evaluations and bounding analyses of these potential outliers consisted of hand calculations using basic engineering mechanics techniques for simple configurations, and rigorous piping analyses using TPIPE computer program (Reference 7-11) for more complex piping configurations.
Outlier evaluation guidelines and acceptance criteria are discussed in Section 5.1 below. Results of the outlier evaluation are summarized in Section 5.2.
5.1 Outlier Evaluation Guidelines Conditions which do not meet the above in-plant screening guidelines or which were judged by the Walkdown Team members to require further reviews are documented as "Potential Outliers". Technical bases and methods used for further analysis and evaluation of these potential outliers are based on industry standard engineering practices, and are consistent with those recommended in BWROG Report GE NEDC-31858P (Reference 7-1), Generic Implementation Procedure (GIP, Reference 7-3) and TVA Design Criteria BFN-50-C-7306 (Reference 7-7), as applicable. Realistic effects of non-linear behavior due to design features and phenomena such as proximity impact with other plant features, interferences and small clearances to stiff structures, geometric restoring forces, wall penetration sealants, and support ductile behavior are considered in the analysis as appropriate.
5.1.1 Seismic Demand Turbine Building at Browns Ferry Nuclear Plant is classified as a Class II structure per BFN FSAR (Reference 7-11). As such, no dynamic response analysis was performed for the building. The building below the operating floor (El. 617 ft.) is a reinforced concrete framed structure supported on steel H-piles to bedrock. The superstructure above the operating floor consists of welded steel rigid braced frames. As a Class II structure, Turbine Building was designed to seismic zone 1 and a wind speed of 100 Page 5-1 FACIUTY RISK CONSULTANTSt INC.
- ~~~~ .. - -
mph per Uniform Building Code (Reference 7-13), utilizing industry-standard design guidelines such as AISC (Reference 7-14) and ACI (Reference 7-15).
Majority of the MSIV alternate leakage treatment piping and associated components within the MSIV seismic verification boundary, including the condensers, are located in the Turbine Building. Since no in-structure response spectra were available for the Turbine Building, horizontal seismic demand for components located within about 40 feet of the Turbine Building effective grade elevation (El. 568 ft.) is taken as the BFN 5%
damped design basis DBE input spectrum (0.2g) scaled by 1.6 for soil amplification per BFN FSAR (Reference 7-11), and 1.5 for building amplification per GIP (Reference 7-3).
For components located above 40 feet of the Turbine Building effective grade elevation, an additional amplification factor of 1.5 is conservatively applied. In the vertical direction, seismic demand is taken as 2/3 that of the horizontal direction, with a soil amplification factor of 1.1 instead of 1.6 per BFN FSAR (Reference 7-11).
5.1.2 Equipment Anchorage Acceptance Criteria Unanchored, unrestrained, marginally or inadequately anchored equipment components identified as potential outliers are subject to further evaluations. Anchorage capacities are based on those provided in Appendix C of the Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment (Reference 7-3) with appropriate reduction factors as applicable. Piping and tubing attached to equipment components with flexible support systems, such as vibration isolators, are evaluated for seismic anchor movement.
5.1.3 Pipe Support Acceptance Criteria Pipe support anchorage loads are verified against capacities as provided in Appendix C of the GIP (Reference 7-3). Support components that may exhibit non-ductile behavior are accepted based the following stress allowables:
Flexural and tensile stresses: lesser of 0.7 Su and 1.2 Sy Shear stresses: lesser of 0.42 Su and 0.72 Sy Bolt stresses: greater of 0.7 Su and minimum specified Sy Page 5-2 FACIOTY RISK CONSULTANTSf INf
Where, S, is the material's ultimate strength Sy is the material's yield strength When test data are available, acceptable loads based on test data consider mean less one standard deviation capacity.
Pipe supports not meeting the above criteria may be accepted if adjacent supports and the resulting pipe span can resist dead loads with a factor of safety of 2.0. In-plant considerations regarding other consequences of support failure such as falling and excessive deflection are made when using this provision.
5.1.4 Pipe Stress Acceptance Criteria Pipe stresses induced by the combination of normal operating loads (dead load and pressure) and seismic loads (DBE inertial loads and DBE seismic anchor movements) are limited to 2.0 Sy.
Majority of the piping and components within the MSIV seismic verification boundary are associated with the Main Steam system which is high energy (i.e., high pressure and temperature), thus the effects from these loads may be significant. Although most piping within the seismic boundary are typically supported by rod hangers, thermal effects are included in the evaluations when judged to be significant, as in the case of rigid support systems, or when addressing proximity interaction and piping flexibility issues.
5.2 Outlier Evaluation Results Results of further evaluations and analyses to resolve the identified potential outliers are presented in Table 5-1 by the designated subsystems. A listing of the calculation packages associated with further evaluations and bounding analyses of potential outliers identified under the BFN-1 MSIV seismic ruggedness verification program is provided in Appendix B of this report.
For those outliers not meeting the evaluation acceptance criteria as described in the above Section 5.1, plant modifications are designed in accordance with TVA Design Criteria BFN-50-C-7306 (Reference 7-7) and implemented accordingly to resolve these outlier conditions. Other miscellaneous maintenance and/or housekeeping items are resolved through the issuance of work orders. Discussions of the plant modifications Page 5-3 FACILTY RISK CONSULTANSt INC.
,<e-41 Fs; * . - ., ,.._~,I ____
including general maintenance and/or housekeeping items are presented in Chapter 6 of this report.
Page 5-4 FACILTY RISK CONSULTANTSt INC.
TABLE 5-1 OUTLIER RESOLUTION
SUMMARY
BFN-1 MSIV SEISMIC RUGGEDNESS VERIFICATION PROGRAM 00tidra
- ._~ sismic: issu Syste;m De scp Noft lon off Wtion P l. D tlV tAF 1 1.0 Main Steam Drain Line - Turbine Bldg. Main Steam Vault Main Steam Drain Taps 1-1 Flexibility concerns associated with the MS lines - As-installed configurations were evaluated and found and Drains. to be acceptable per calculation.
MOV 's1-FCV-1 -57 &58 1-2 Excessive mass and extended valve operators. i As-installed configurations were evaluated and found to be acceptable per calculation.
Main Steam and HPCI Drains 1-3 Flexibility concerns associated with the MS and - As-installed configurations were evaluated and found I_ HPCI Drains. to be acceptable per calculation.
Main Steam Bypass Drain 1-4 Proximity interaction concerns between bypass - As-installed configurations were evaluated and found
_I_ line and adjacent 2-way support. to be acceptable per calculation.
2.0 Main Steam Lines - Turbine Bldg. = = = =
MS Balancing Header 2-1 Upper pipe damp nut ismissing. Work order 98-001447-001 has been in place. (WO)
MS Turbine StopControl Valves 2-2 MS stopcontrol valves are not covered inthe _ - T Seismic adequacy of these valves is presented ina valve screening criteria, thus, seismic adequacy calculation.
of these valves needs to be verified.
3.0 Main Steam Drain Line - Reactor Building MSIV Vault MSIV 's 1-FCV-1 -15, 27, 38 & 52 3-1 MSIV s are not covered in the valve screening _ 7 Seismic adequacy of these valves ispresented in a criteria, thus, seismic adequacy of these valves calculation.
needs to be verified.
MOV 1-FCV-1-56 3-2 Excessive mass and extended valve operators. _ - As-installed configurations were evaluated and found
_. to be acceptable per calculation.
4.0 HPCI/RCIC/Aux. Boiler Drains RCIC Drain 4-1 Excessive pipe overspan condition. - - Overspan condition found to be unacceptable by calculation. Additional support to be installed per calculation. (M)
RCIC Drain 4-2 Pipe support with questionable capacity. l - - Pipe support OK per calculation.
Aux. Steam Boler 4-3 Missing rod hanger hardware (eye nut). l _ Issue work order to re-install the missing nut. (WO)
HPCI Drain 4-4 Multiple pipe overspan conditions on the 1"w - - As-installed configurations were evaluated and found drain line. I to be acceptable per calculation.
HPCI Line 4-5 Multiple pipe overspan conditions on the 2"' - - As-installed configurations were evaluated and found
_ line. l to be acceptable per calculation.
HPCI Une and Valve 1-73-222 4-6 Proximity interaction of the 2"l HPCI line and - - As-installed configurations were evaluated and found
_ the valve. - to be acceptable per calculation.
Page 5-5 FACILiTY RISK CONSULAN, INC.
TABLE 5-1 (CONT'D)
OUTLIER RESOLUTION
SUMMARY
BFN-1 MSIV SEISMIC RUGGEDNESS VERIFICATION PROGRAM
!;OUI. mid'i, smcsutllee-er*.........Sei = ...........
=
i]" D e"c"qfe"...
m"' "" e" a .....
4.0 HPCl/RCIC/Aux. Boiler Drains (cont'd)
HPCI Drain 4-7 Several floor-mounted supports were noted to Existing support anchorage capacities are adequate per be damaged (potentially due to thermal effects). calculation. However, the damaged supports located Also, anchor bolt spacing violations on these just passed the blowout panel needs to be repaired and typical supports. longer lugs are required to avoid pipe from sliding off the existing supports due to thermal movements. (M)
Use 3"thermal movement as a guide for the design of
__ __ support mods.
HPCI Drain 4-8 Flexibility concern on 2"a HPCI line due to i As-installed configurations were evaluated and found to Idifferential displacement between RB and TB be acceptable per calculation.
5.0 MS PT 1-72, -76, -82, -86 &-93 SS Tubing to I-PT-1-72 &76 5-1 Overspan conditions for both lines. Also, - - Replace existing support at N-end with rod hanger, and flexibility issue exists on tubing to 1-PT-1 -72. _ add new rod hanger support at S-end. (M)
SS Tubing to 1-PT-1 -82 5-2 Overspan conditions due to broken weld on the 7T -T Replace both existing supports at N-and S-ends with support and missing clamps. rod hangers. (M)
SS Tubing to 1-PT-1 -86 5-3 Overspan conditions due to broken weld on the T _T Replace both existing supports at N-and S-ends with I support and loose clamp. rod hangers. 8U)
SS Tubing to 1-PT-1-86 5-4 Flexibility concern on 1 a"tubing. N PER 040016-000 was generated to identify the required work order to remove existing clamp on support. (WO)
SS Tubing to PT 1-93 5-5 Overspan condition due to broken support. _ Replace existing support with rod hanger. (M)
SS Tubing to PT 1-93 5-6 Flexibility concern on 19a tubing along the wall. 7 1. PER 04-000616-000 was generated to identify the required work order to remove existing clamp on the
_ _ spacersupport. (WO) 6.0 Main Steam Sample Lines to Sampling Station Sample Une A 6-1 Missing tubing clamp. - _ PER 04- 000616-000 was generated to identify the required work order to re-install the missing damp.
(WO)
Sample Station 6-2 Unanchored constant temperature bath cabinet. T = = Provide positive anchorage to the cabinet. (M)
Sample Une Ato 1-PT- -1l6NB 6-3 Flexibility concerns on sample lines to 1-PT-i- T i Replace existing wall-mounted support with rod I16NB, involving broken support. hangers. (M)
Sample Une A to I -PT- -16A/B 6-4 Missing nuts on U-bolt support. - PER 04-000616-000 was generated to identify the I__ __I.Missingnuts required work order to re-install the missing nuts. (WO)
Page 5-6 FACILITY RISK CONSULTANTSJ, IN.
'..(i
,Kl I r., !",
I. .
I.,
I., ,
TABLE 5-1 (CONT'D) I .
OUTLIER RESOLUTION
SUMMARY
BFN-1 MSIV SEISMIC RUGGEDNESS VERIFICATION PROGRAM r
Sytm%,1t~I~~Ii Imm sciotlon6 of Cod Itio I FIP
'VI . -'
eco0 7.0 Main Steam Bypass MS Bypass Valve Assembly 7-1 MS Bypass valve chest isnot covered inthe Seismic adequacy of the bypass valve assembly is 8.0 MS Stop Valve Above Seat Drains valve screening criteria, thus, its seismic adequacy needs to be verified. L l presented in a calculation.
Seis 1-FCV-6-101, -102 & -103 8-1 Proximity interactions of valve yokes and beam. As-installed configurations were evaluated in a calculation. Valves 101 & 102 were found to be OK as-is, but not for valve 103. Initiate work request to cope the bottom flange of the WF beam supporting the steel grating at El. 601'-6". (WO)
Note that a lateral (E-1W) clearance of 3Xminimum shouldbe provided to preclude potentialseismic l_ _impact between the valve and nearby platform steel.
SV-D Above Seat Drain Piping 8-2 Broken pipe strap on support. - _ PER 04-000616-000 was generated to identify the required work order to replace broken pipe strap.
_ _-) . O_
1-FCV-6-100 8-3 Broken flex conduit connection to operator. _T PER 04-000617-000 was generated to identify the required work order to repair broken flex conduit.
(WO) 9.0 Steam to Steam Seal Regulator Main Steam to 1-FCV-1 -146 9-1 Overspan condition on the 4"6 MS line to the - - - As-installed configurations were evaluated and found supply to steam seal MOV 1-FCV-i -146. . _ to be acceptable per calculation.
Main Steam to 1-FCV-1 *146 9-2 Loose rod hanger eye nuts and disengaged rod. q T PER 04-000616-00 was generated to identify the required work orders (2)to correct rod hanger
_ ______ ____ ___ ___ ___ ___ ____ ___ __ - __ hardware deficiencies. (W O) 1-PCV-1 -147 Air Tubing 9-3 Inadequate flexibility for air line connected to As-installed configurations were evaluated and found the top of valve. _ to be acceptable per calculation.
1-PCV-1-147 Instrumentation 9-4 Falling interaction concern from nearby block T
- Re-route tubing and protect instrumentation panel, as wall at El. 617'. necessary, to preclude falling interactions. (M) 1-PCV-1 -147 9-5 Extended valve operator. _- T As-installed configurations were evaluated and found
_to__ t be acceptable per calculation.
1-FCV-1-146,145 & 154 9-6 Extended valve operator, and substantial i As-installed configurationswere evaluated and found
_ operator weight. I to be acceptable per calculation.
Page 5-7 FACILITY RISK CONSULTANTS, INC.
TABLE 5-1 (CONT'D)
OUTLIER RESOLUTION
SUMMARY
BFN-1 MSIV SEISMIC RUGGEDNESS VERIFICATION PROGRAM 1i... ............. ........
.~~tr~
sritn~Dscripio ofonndiio A Fj' p ecommnendd 10.0 Steam Supply to RFP Turbines Steam Supply Une 10-1 Stanchion supports for the steam supply header T Modify existing stanchion supports for the postulated at El. 617' (Total of 6). DBE seismic movements. Atotal of 3supports require 1support for the 6"6 line near -/T4; modifications - one inthe S-end (H/T4) and two inthe 2supports for the 3"6 line near D/T4; and N-end (D/T4) of the steam supply header at the 1support for each of the 4"6 branch line to RFP Turbine deck, El. 617. Note that similar type supports Turbine compartments. inside each of the RFP Turbine rooms are OK as-is.
(M)
Steam Supply Une 10-2 Turbine Bldg. overhead crane. Resolved per calculation.
RFP Stop Valve Above Seat 10-3 Valve operators with large eccentric mass on T As-installed configurations were evaluated and found Drains . '/2"6 and 3/4"6 lines. to be acceptable per calculation.
Tubing to Pi 1-134 10-4 Missing or broken clamps (2places) on tubing T T - PER 04-000616-000 was generated to identify the to PI 1134. Also, tubing isbent and sagging. required work order to re-install tubing clamps and
_ repair bent tubing. (WO)
Steam Supply Une 10-5 Missing nut on rod hanger support. PER 04-000616-000 was generated to identify the
_ _ __ _ __ __ _required work order to re-install the missing nut. (WO)
Steam Supply Une 10-6 Overspan conditions on the 6"d steam supply T X As-installed configurations were evaluated and found line. to be acceptable per calculation.
11.0 Steam Supply to SJAE's SJAE 's 1A &1B 11-1 Need to verify the seismic adequacy of SJAE Seismic adequacy of SJAE anchorage was verified per anchorage. calculation.
MOV 1-FCV-6-114 11-2 Broken flex conduit connection to valve motor . - T PER 04-000617-00 was generated to identify the operator, required work order to repair the broken flex conduit connection to valve motor operator. (WO)
MS Supply to SJAE's 11-3 Overspan condition on the 3"6 supply line. _T _T As-installed configurations were evaluated and found to be acceptable per calculation.
MS Supply Drain at SJAE 1B 11-4 Overspan condition on the 1"d steam supply NT T As-installed configurations were evaluated and found drain line downstream of 1 FCV-1 -114. to be acceptable per calculation.
Steam Trap Drain at SJAE 1A 11-5 Overspan condition on the 1"6 steam trap drain - T - Overspan condition was found to be unacceptable per between 1FCV-1 -172 and 1-PCV-1 -166. calculation. Add anew support near T4 wall (rod hanger). (M)
Steam Trap Drain at SJAE 1B 11-6 Overspan condition on the 1"6 steam trap drain V - Overspan condition was found to be unacceptable per between 1FCV-1 -173 and 1-PCV-1 -167. calculation. Add anew support near T4 wall (rod
___ _ hanger). (M)
Page 5-8 FACIuTY RISK CONSULTANTS, INC.
TABLE 5-1 (CONT'D)
OUTLIER RESOLUTION
SUMMARY
BFN-1 MSIV SEISMIC RUGGEDNESS VERIFICATION PROGRAM
... . . .. I I.
ptlort of..........:~ ~ ~ Fecnnene ieoito 1.LJ Otld1tl Oupply lr urr-us I-rrnd(rz Steam Supply Unes to Off-Gas 12-1 Lines penetrate into the Off-Gas Preheater - - Re-route pipes to preclude falling interactions. Note Preheaters room through masonry block wall. that pipe re-routing will be done inconjunction with the installation of new isolation valves (1-SHV-1-741
&743 and 1-CKV-1-742 & 744) under DCN 51112.
(M) 13.0 SJAE's Drain to Condenser SJAE's Drain to Condenser 1C 13-1 Drain to condenser ties into a multi-system _ Re-route drain line to achieve a direct path to et 1..DraitoIcollector. condenser. (M) 14.0 Main Steam Drain Line to Condenser Main Steam Drain Line to 14-1 Spring hanger rod detached from 4"i drain line. T _ PER 04-000616-000 was generated to identify the Condenser IA required maintenance work order to re-connect the detached rod to pipe. (WO) 1-FCV-1 -59 14-2 Broken flex conduit connection at base. T PER 04-M17600000 was generated to identify the required work order to repair broken flex conduit.
.___ _ __ (WO) 1-FCV-1-59 14-3 Valve operator (motor) in close proximity to the - As-installed configurations were evaluated and found side rail of steel ladder. to be unacceptable per calculation. Valve should be re-oriented with about 6"dearance to the nearest fixed structure to the north. (M)
Note that the 46 MS drain and vent pipes are to be replaced under DCN 51112. As such, provide 6"
_ clearance when installing the new pipes.
15.0 Condensers Condensers 1A, lB & 1C 15-1 Need to verify the seismic adequacy of the f r Condenser and its anchorage configurations were l condenser and its anchorage configurations. . l l evaluated and found to be acceptable per calculation.
- Seismic Issue Notations:
A= Anchorage/Support Capacity; F= Failure/Falling Interaction; P= Proximity & Impact; D= Differential Displacement; V= Valve Screening &Performance Note: Items that require engineering design modifications are listed as (M), and are to be resolved by various DCN's.
Items that require resolution by work order requests as listed as (WO).
Page 5-9 FACILITY RISK CONSULTANTS, INC.
- 6.
SUMMARY
AND RECOMMENDATIONS A total of fifty-four (54) potential outliers were identified in the MSIV seismic ruggedness verification walkdown for BFN-1. Majority of these potential outliers were resolved by performing further analyses and evaluations to the acceptance criteria of TVA Design Criteria BFN-50-C-7306 (Reference 7-7). For the remaining outliers that did not meet the above evaluation criteria, plant design modifications are developed for appropriate resolution. In addition, work requests are initiated for those outliers that fall into the general category of maintenance and housekeeping items, such as missing support hardware, broken parts, proximity interaction and flexibility concerns and others. Table 6-1 presents a summary of the BFN-1 MSIV seismic ruggedness verification program.
6.1 Plant Modifications Plant design modifications are developed for outliers not meeting the acceptance requirements of TVA Design Criteria BFN-50-C-7306 (Reference 7-7). Design modifications ranged from simple support hardware modifications to addition of new pipe supports, rerouting of piping and others. A total of fifteen (15) plant design modifications were implemented for the resolution of these MSIV seismic ruggedness outliers. A brief description of each of the plant modifications is provided in Table 6-2.
Engineering designs of the plant modifications are documented in various calculation packages listed in Appendix C of this report. These plant modifications will be implemented under several Design Change Notices (DCN's 51112, 51669, and 51126).
6.2 Maintenance and Housekeeping Items In addition to the plant modifications discussed in Section 6.1 above, a total of fifteen (15) miscellaneous maintenance and housekeeping items were identified for appropriate actions and resolution. These maintenance items, along with a brief description, are tabulated in Table 6-3. These items will be disposed through maintenance work requests.
Page 6-1 FACILITY RISK CONSULTANTS, INC.
"I I -
f ; e TABLE 6-1 Summary of BFN-1 MSIV Seismic Ruggedness Verification Program
...ltem.s 1 MS Drain Line -Turbine Bldg. Main Steam Tunnel 4 0 0 2 MS Lines - Turbine Bldg. 2 0 1 3 MS Drain Line - Reactor Bldg. Main Steam Vault 2 0 0 4 HPCI/RCIC/Aux. Boiler Drains 8 2 1 5 MS Pressure Transmitters PT 1-72, 76, 82, 86 & 93 6 4 2 6 MS Sample Lines to Sampling Station 4 2 2 7 MS Bypass 1 0 0 8 MS Stop Valve Above Seat Drains 3 0 3 9 MS to Steam Seal Regulator 6 1 1 10 Steam Supply to RFP Turbines 6 1 2 11 Steam Supply to Steam Jet Air Ejectors (SJAE's) 6 2 1 12 Steam Supply to Off-Gas Preheaters 1 1 0 13 SJAE's Drain to Condenser 1 1 0 14 MS Drain Line (Turbine Bldg.) to Condenser 3 1 2 15 Condenser 1 0 0 Total 54 15 15 Page 6-2 FACILITY RISK CONSULTANTS, INC.
m, : ;- . 4-- - - ; I. .
TABLE 6-2 Summary of Plant Modifications BFN-1 MSIV Seismic Ruggedness Verification Program 1 POS Add one (1) new support. 1-47B456-2100 4-1 see note (3) 2 POS Add four (4) new supports. 1-47B400-2030 4-7 to -2033 3 POS Add two (2) new supports. 1-47&400-2026 5-1 & -2027 4 POS Add two (2) new supports. 1-471400-2022 5-2 & -2023 5 P0S Add two (2) new supports. 1-47B3400-2024 5-3 & -2025 6 P05 Add one (1) new support. 1-4783400-2021 5-5 7 POS Provide anchorage to the Constant 1-488879-1 6-2 Temperature Bath Cabinet. see note (4) 8 POS Add one (1) new support. 1-473400-2020 6-3 9 POS Reroute tubing to preclude seismic falling see note (2) 9-4 interaction.
10 POS Add three (3) new supports. 1-47B400-2034-1 & -2, and 10-1 1-471400-2035 & -2036 Page 6-3 FACLTY RISK CONSULTANTS, INC.
TABLE 6-2 (cont'd)
Summary of Plant Modifications BFN-1 MSIV Seismic Ruggedness Verification Program 11 POS Add one (1) new support. 1-47B400-2028 11-5 12 POS Add one (1) new support. 1-47B400-2029 11-6 13 POS Reroute piping to preclude seismic falling interaction. see note (2) 12-1 14 POS Reroute drain line directly to the condenser. see note (2) 13-1 15 POS Relocate valve to preclude seismic proximity see note (2) 14-3 interaction.
(1) Detailed description and as-built information of these items, including photos and/or sketches, can be found in the Potential Outlier Sheet (POS) contained in the respective MSIV seismic ruggedness Walkdown Data Packages (WDP's).
(2) Refer to DCN 51112, unless noted otherwise.
(3) Refer to DCN 51669.
(4) Refer to DCN 51126.
Page 6-4 FACILITY RISK CONSULTANTS, INC.
TABLE 6-3 Summary of Misc. Maintenance & Housekeeping Items BFN-1 MSIV Seismic Ruggedness Verification Program Item... ef X.c Recommen"de"d A Ao ...
... ~g. ...... S 1 POS Replace missing upper pipe clamp Work Order WO # 98-001447-001 2-1 nut. was issued.
2 POS Replace missing eye nut. Work Order WO # 04-712923-000 4-3 was issued.
3 POS Remove existing clamp on pipe Work OrderWO # 04-712923-000 5-4 support. was issued.
(See also DCA 51112-131) 4 POS Remove existing clamp on pipe Work OrderWO # 04-712923-000 5-6 support. was issued.
(See also DCA 51112-132) 5 POS Replace missing tubing clamp. Work Order WO # 04-712923-000 6-1 was issued.
6 POS Replace missing nut on U-bolt Work Order WO # 04-712923-000 6-4 support. was issued.
7 POS Cope the bottom flange of the WF PIC 61400 was added to DCN 8-1 beam to provide 3" clearance (E-W) to 51112 to initiate the work.
preclude seismic proximity interaction.
8 POS Replace broken pipe strap on support. Work OrderWO # 04-712923-000 8-2 was issued.
9 POS Repair broken flex conduit connection. Work Order WO # 04-712926-000 8-3 was issued.
10 POS Replace loose rod hanger eye nuts Work OrderWO # 04-712923-000 9-2 and disengaged rod. was issued.
Page 6-5 FACILY RISK CONSULTANTS, INC.
.-_-C'-'.- ". ". f - , " .. . . '_ - .
TABLE 6-3 (cont'd)
Summary of Misc. Maintenance & Housekeeping Items BFN-1 MSIV Seismic Ruggedness Verification Program 11 POS Replace missing or broken tubing Work Order WO # 04-712923-000 10-4 clamps and repair bent tubing. was issued.
12 POS Replace missing nut on rod hanger Work Order WO # 04-712923-000 10-5 support. was issued.
13 POS Repair broken flex conduit connection. Work Order WO # 04-712926-000 11-2 was issued.
14 POS Reinstall detached rod to pipe Work Order WO # 04-712923-000 14-1 support. was issued.
15 POS Repair broken flex conduit connection. Work Order WO # 04-712926-000 14-2 was issued.
- Detailed description and as-built information of these items, including photos and/or sketches, can be found in the Potential Outlier Sheet (POS) contained in the respective MSIV seismic ruggedness Walkdown Data Packages (WDP's).
Page 6-6 FACIUTY RISK CONSULTANTS, INC.
- 7. REFERENCES 7-1 BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, GE NEDC-31858P, Revision 2. September 1993.
7-2 TVA calculation, "MSIV Leakage Containment System Boundaries, Physical Properties, System 001." Rev. 3.
7-3 "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment", Revision 2A, March 1993. Seismic Qualification Utility Group (SQUG).
7-4 EPRI Report NP-7149. March 1991. "Summary of the Seismic Adequacy of Twenty Classes of Equipment Required for Safe Shutdown of Nuclear Plants."
Electric Power Research Institute, Palo Alto, Califomia.
7-5 EPRI Report RP-2635-1. February 1987. OPiping Seismic Adequacy Criteria Recommendation Based on Performance During and After Earthquakes."
Volumes 1 and 2. Electric Power Research Institute, Palo Alto, Califomia.
7-6 TVA Walkdown Instruction WI-BFN-0-CEB-07, "Engineering Walkdown Instruction for MSIV Seismic Ruggedness Verification." Revision 0.
7-7 TVA Detailed Design Criteria BFN-50-C-7306, "Qualification Criteria for Seismic Class II Piping, Pipe Supports, and Components." Revision 1.
7-8 TVA calculation, "In-Plant Screening Tools for Seismic Il/I Spray Walkdown Evaluations." Revision 1.
7-9 TVA calculation, "Main Steam Seismic Ruggedness Verification." Revision 5.
7-10 TVA calculation, "Main Steam Seismic Ruggedness Evaluation." Revision 1.
7-11 "TPIPE Program User Manual." Version 16, July 1994.
7-12 TVA Browns Ferry Nuclear Plant Final Safety Analysis Report (FSAR).
Page 7-1 FACILITY RISK CONSULTANTS, INC.
g -1 ., , ., -y,. -W. -- - --,, , - - - - -- - -, -
. - -.- - 1. .
- 7. REFERENCES (CONT'D) 7-13 Uniform Building Code.
7-14 American Institute of Steel Construction (AISC). "Manual of Steel Construction." 6th Edition.
7-15 American Concrete Institute (ACI). "Building Code Requirements for Reinforced Concrete." ACI 318-1963 Edition.
7-16 USAS B31.1.0, "Power Piping." 1967 Edition.
Page 7-2 FACILI7Y RISK CONSULTANTS, INC.
APPENDIX A:
BFN-1 MSIV SEISMIC RUGGEDNESS VERIFICATION WALKDOWN DATA PACKAGES Page A-1 FACILITY RISK CONSULTANTS, INC.
, .- -------- . . ... . . .- . . .. -- - . . - . . - -1. .. - '.. . .
BFN-1 MSIV Seismic Ruggedness Verification Walkdown Data Packages
- ,W lkdw. Daa..
Pakg . .,.,{5,(WDP)Sf .ys.< i':'::s:
BFN1 -CEB-MSIV-01 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 1 BFN1-CEB-MSIV-02 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 2 BFN1-CEB-MSIV-03 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 3 BFN1-CEB-MSIV-04 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 4 BFN1-CEB-MSIV-05 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1- Package 5 BFN1-CEB-MSIV-06 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 6 BFN1-CEB-MSIV-07 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 7 BFN1-CEB-MSIV-08 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 8 BFN1-CEB-MSIV-09 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 9 BFN1-CEB-MSIV-10 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 10 BFN1-CEB-MSIV-11 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 11 BFN1-CEB-MSIV-12 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 12 BFN1-CEB-MSIV-13 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 13 BFN1-CEB-MSIV-14 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 14 BFN1 -CEB-MSIV-15 MSIV Seismic Ruggedness Walkdown Screening Evaluation 0 Documentation for BFN Unit 1 - Package 15 Page A-2 FACILitrY RISK CONSULTANTS, INC.
APPENDIX B:
BFN-1 MSIV SEISMIC RUGGEDNESS VERIFICATION OUTLIER RESOLUTION CALCULATION PACKAGES Page B-1 FACILImY RISK CONSULTANSt INC.
BFN-1 MSIV Seismic Ruggedness Verification Outlier Resolution Calculation Packages Main Steam Drain Line, MSIV Ruggedness Seismic Analysis - 0 Resolution of POS 1-1, 1-3 & 1-4 HPCI Drain Line, MSIV Ruggedness Seismic Analysis - Resolution of 0 POS 4-6,4-7 & 4-8 Misc. Main Steam Valve Performance Issues, MSIV Ruggedness 1 Seismic Analysis - Resolution of POS 2-2, 3-1 & 7-1 Misc. Extended Valve Operator Issues, MSIV Ruggedness Seismic 0 Analysis - Resolution of POS 1-2, 3-2, 9-5, 9-6 & 10-3 Misc. Piping Overspan Issues, MSIV Ruggedness Seismic Analysis - 0 Resolution of POS 4-4, 4-5, 9-1, 10-6, 11-3, 11-4,11-5 & 11-6 Misc. Seismic Interaction Issues, MSIV Ruggedness Seismic Analysis 0
- Resolution of POS 10-1, 10-2 & 11-1 Misc. Valve Proximity Interaction Issues, MSIV Ruggedness Seismic 0 Analysis - Resolution of POS 8-1, 9-3 & 14-3 Seismic Verification of Condenser and its Anchorage, MSIV 0 Ruggedness Seismic Analysis - Resolution of POS 15-1 RCIC Line, MSIV Ruggedness Seismic Analysis - Resolution of POS 1 4-1 RCIC Line, MSIV Ruggedness Seismic Analysis - Resolution of POS 0 4-2 Page B-2 FACILIy RISK CONSULTANTS, INC.
APPENDIX C:
BFN-1 MSIV SEISMIC RUGGEDNESS VERIFICATION PLANT MODIFICATION DESIGN CALCULATION PACKAGES AND MAINTENANCE WORK ORDERS Page C-1 FACILITY RISK CONSULTANTS, INC.
-~~ *--. --
_____________ - .:- .t-.- n--B--.. i . <-.-.-------- *- e . .-e BFN-1 MSIV Seismic Ruggedness Verification Plant Modification Design Calculation Packages and Work Orders
... X.. .
Desc ip io of Wrk rde 2-1 (W) WO 98-001447-001 Replace missing upper pipe clamp nut 4-1 (M) Calculation RCIC Line, MSIV Ruggedness Seismic Analysis - Resolution of POS 4-1, Rev. 1 4-3 (W) WO 04-712923-000 Replace missing eye nut 4-7 (M) Calculation Main Steam Seismic Ruggedness Verification, Unit 1, Rev. 0 5-1 (M) Calculation Main Steam Seismic Ruggedness Verification, Unit 1, Rev. 0 5-2 (M) Calculation Main Steam Seismic Ruggedness Verification, Unit 1, Rev. 0 5-3 (M) Calculation Main Steam Seismic Ruggedness Verification, Unit 1, Rev. 0 5-4 (W) WO 04-712923-000 Remove pipe clamp on existing support (see also DCA 51112-131) 5-5 (M) Calculation Main Steam Seismic Ruggedness Verification, Unit 1, Rev. C 5-6 (W) WO 04-712923-000 Remove pipe clamp on existing support (see also DCA 51112-132) 6-1 (W) WO 04-712923-000 Replace missing tubing clamp 6-2 (M) Calculation Design of Anchorage for Bath and Circulator Cabinet for Unit 1, Rev. 0 6-3 (M) Calculation Main Steam Seismic Ruggedness Verification, Unit 1, Rev. 0 6-4 (W) WO 04-712923-000 Replace missing nut on U-bolt support
- (M) - Plant Modifications (W) - Work Page C-2 FACILITY RISK CONSULTANTS, INC.
t'"- a>.-
BFN-1 MSIV Seismic Ruggedness Verification Plant Modification Design Calculation Packages and Work Orders 8-1 (W) Refer to DCN 51112 Cope bottom flange of the WF beam to provide PIC 61400 minimum of 3" clearance in the E-W direction 8-2 (W) WO 04-712923-000 Replace broken pipe strap on the support 8-3 (W) WO 04-712926-000 Replace broken flex conduit connection 9-2 (W) WO 04-712923-000 Replace loose rod hanger eye nuts and the disengaged rod 9-4 (M) Refer to DCN 51112 Reroute instrument tubing 10-1 (M) Calculation Main Steam Seismic Ruggedness Verification, Unit 1, Rev. 0 Calculation Main Steam Seismic Ruggedness Verification, Units 2 & 3, Rev. 5 10-4 (W) WO 04-712923-000 Replace missing or broken tubing clamps and repair bent tubing 10-5 (W) WO 04-712923-000 Replace missing nut on rod hanger support 11-2 (W) WO 04-712926-000 Replace broken flex conduit connection 11-5 (M) Calculation Main Steam Seismic Ruggedness Verification, Unit 1, Rev. 0 11-6 (M) Calculation Main Steam Seismic Ruggedness Verification, Unit 1, Rev. 0 12-1 (M) Refer to DCN 51112 Reroute piping 13-1 (M) Refer to DCN 51112 Reroute drain line to condenser 14-1 (W) WO 04-712923-000 Reinstall detached rod to pipe support 14-2 (W) WO 04-712926-000 Replace broken flex conduit connection 14-3 (M) Refer to DCN 51112 Relocate valve
- (M)-PlantModifications (W)-WorkOrders Page C-3 FACILITY RISK CONSULTANTS, INC.
TVAN CALCULATION COVERSHEET/CCRIS UPDATE Page I of 2 Page I REV 0 EDMS/RIMS NO. EDMS TYPE: EDMS ACCESSION NO (IWA for REV. 0)
R14 990909 102 Calculations(Nuclear) W7 8 0406 0 3 0 67 CalcTitle: SEISMIC EVALUATION REPORT CALC ID lYPE I ORG PLANT I BRANCH CUR REV lNEWE CURRENT CN NUC BFN CEB CDN0 001 99 0113 001 002 lPREVISION NEW CN NUC BFN CEB Entire calc 0 Selected pages 0 No CCRIS Changes O ACTION NEW O DELETE 0 SUPERSEDE 0 CCRIS UPDATE ONLY E (For cal revision, CCRIS REVISION 0 RENAME 0 DUPLICATE 0 (Verifier Approval Signatures Not been reviewed and no Require CCRIS changes required)
UNITS SYSTEMS UNIDS 000 001 WNA DCN.EDC.N/A APPLICABLE DESIGN DOCUMENT(S) CLASSIFICATION 51112 N/A D OUALITY SAFETY RELATED? UNVERIFIED SPECIAL REQUIREMENTS DESIGN OUTPUT SARITS AFFECTED RELATED? (II yes, OR = yes) ASSUMPTION AND/OR LIMITING CONDITIONS? ATTACHMENT?
Yes No 0 Yes O No O Yes [I No O Yes O No Yes O No Q2 Yes O No O PREPARER ID PREPARER PHONE NO PREPARING ORG tBRANCH)I VERIFICATION METHOD NEW METHOD OF ANALYSIS John 0. Dizon (256) 729.7000 x18376 CEB Design Review 0 Yes 0 No PREP E SIGNATUVB DATE CHECKER IGNATURE DATE
- 0.
- J.O. Dizon 6102104 S.J. Eder 6/02104 VER!FIER SIGNATURE DATE APPROVAL SIGNA61URE A DATE
-5ta&-r- S.J. Eder 6/02104 . 4/j(% C /- ____RC_____
STATEMENT OF PROBLEWABSTRACT This calculation documents the MSIV Seismic Ruggedness Verification for Browns Ferry Nuclear Plant, Units 1, 2 & 3 for the Increased MSIV Leakage Tech Spec Change.
MICROFICHEJEFICHE Yes 0 No 0 FICHE NUMBER(S)
O LOAD INTO EDMS AND DESTROY 09 LOAD INTO EDMS AND RETURN CALCULATION TO CALCULATION LIBRARY. ADDRESS: BFN Calculation Ubrary a LOAD INTO EDMS AND RETURN CALCULATION TO:
TVA 40532 (07-20011 Page 1 of 2 NEDP-2-1107-09-20011
T:'. -- - _.. - .
TVAN CALCULATION COVERSHEETICCRIS UPDATE Page 2 of 2 Page IA CALC ID l TYPE l ORG l PLANT l BRANCH I NUMBER I REV CN NUC BFN CEB I CDNO 001 99 0113 002 ALTERNATE CALCULATION IDENTIFICATION BLDG ROOM ELEV COORD/AZIM FIRM Print Report Yes 00 N/A N/A N/A Bechtel CATEGORIES F06 KEY NOUNS (A-add, D-delete)
ACTION KEY NOUN AID KEY NOUN fAtD)
A MSIV A RUGGEDNESS__________________
A CONDENSER _
A JANCHORAGE CROSS-REFERENCES (A-add, C-change, D-delete)
ACTION XREF XREF XREF XREF - XREF XREF (AICID CODE TYPE PLANT BRANCH NUMBER REV A P CN BFN CEB CDNO001 98 0038 A P CN BFN CEB CDN0 001 98 0039 A P CN BFN CEB CDN1 000 2004 0041 I u~I1_JF CCRIS ONLY UPDATES:
Following are required onl when makIng keyword/cross reference CCRIS updates and page 1 of form NEDP-2-1 Is not Included:
PREPARER SIGNATURE PREPARER PHONE NO.
DATE EDMS ACCESSION NO.
I CHECKER SIGNATURE 78 0 4 06 03 I DATE 06 7
-_ -_ n 6r -If2.-4 _.
T VA 4053Ze p07-Z0 Ij Fage Z 01 z 11"Mr-&- LU- ~-CW-- '
TVAN CALCULATION RECORD OF REVISION Page 1 of 1 Page II TVAN CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER CDND 001 99 0113 Title HCLPF CALCULATIONS OF SELECTED BLOCKWALLS FOR SEISMIC IPEEE PROGRAM Revision DESCRIPTION OF REVISION No. I 000 Initial Issue Total Number of Pages = 125 001 Minor revision to correct Table of Contents page, change Attachment Section 11.0 to 9.0 and delete reference to Appendices and minor changes to the EQE 'Seismic Evaluation Report", page A40 show valves 71-68 and 73-68 as open, and add Reference 10 on page A75 and add reference to Reference 10 to pages A44 and A46.
Total Number of Pages = 125 002 Add pages I, iA,ii & iii; delete page 2 (previous revision log); minor revision to pages 4, 5 & 6 to reference Attachment D; and add Attachment D to this revision.
The impact to the FSAR and Tech Specs has been addressed by the 10 CFR 50.59 review for DCN 51112.
Total Number of Pages = 131 V
T 4070 (- 00 TVA 40709 [12-20001 PAA 1 0? 1 Page I at 1 Aa NtDP-2-Z- [ I 4-V4-ezw-J
.1
- . - )o-TVAN CALCULATION VERIFICATION FORM PageI of 1 Page III TVAN CALCULATION VERIFICATION FORM Calculation Identifier CDNO 001 99 0113 Revision 002 Method of verification used:
- 1. Design Review 0
- 2. Alternate Calculation Verifier Date 6/02/04
- 3. Qualification Test oStephen J.Eder Comments:
This revision of the calculation has been reviewed by the Design Review Methodology and has been determined to be technically adequate based on the design Input information contained herein using accepted handbook andlor computer applications, and sound engineering practices and techniques, supplemented by applicable industry-standard guidelines.
TVA 40533 [07-2001) Page I of 1 NEOP-2-4 [07-09-2001)
OA Record TVAN CALCULAT Title Seismic Evaluation Report Preparing Organization Key Nouns (For EDM)
NE/CEB Seismic, Component Qualification, Piping, Pipe Support Calculation Identifier Each time these calculations are Issued, preparer must ensure that the original (R0)
RIMS/EDM accession number Is filled In.
CD-N0001-990113 Rev (for EDM use) EDM Accession Number Applicable Design Document(s) RO BFN-50-C-7100 __ _ _ 14 990909 A 102 BFN-50-C-7107, BFN-50-C-7306 RI I tL 4 0 0 .0 6.1. 3 1 0. 6 UNID System(s) 001 R2 R3 RO RI R2 R3 Quality Related? Yes No I a ni DCN, EDC, NIA 7--V8/ , . Safety related? If yes, mark Yes No NA "A Quality Related yes 0 U F Caramante _
Prepared 45;0/3 7 Checked l A.- ;. ^.These calculations contain Yes No unverified assumption(s) 0 U e£.4wr.
sign $.PL5L. that must be verified later?
Design These calculations contain Yes No Verified special requirements andlor a
_______ AS conditions?
.Z Dlimiting Approved R These calculations contain Yes No
- t U 4odt a design output a
___ __ __ __L_ _.attachment?
Approval Date pproval_____ 6 00 Calculation Classification D SAR Yes 0 No w Yes 0 No ' Yes O No 0 Yes O No 0 Microfiche generated Yes No Affected? O U Revision Entire cafe U Entire calc B0 Entire calc 0 Entire calc a Number applicability . Selected pgsO Selected pgsO Selected pgs C)
Statement of Problem:
The Main Steam piping downstream of the outboard MSIV's is desired to be capable of with standing an earthquake so that any leakage through the MSIV's from the Reactor side can be contained and diverted to the main condenser. This calculation supports the MSIV leakage tech spec change at BFN.
Abstract This calculation is documents the "Browns Ferry Nuclear Plant - Increased MSIV Leakage Tech Spec Change Submittal Seismic Evaluation Report", (200918-R-002); August 1999, By EQE International, Oakland, CA. Additionally, Bounding Calculations (200918-C-002) *Seismic evaluation for the Condensers' and (200918-C-001) 'Seismic verification of the main steam drain piping and supports associated with the MSIV alternate leakage treatment pathway' are contained in this calculation.
I.l mo CA0467 t 46A f 3 Microfilm and return calculation to Calculation Library. Address: POB-1A-BFN 0 Microfilm and destroy.
o Microfilm and return calculation to:
TVA 40532 102-1999] NEDP-2-1 [02-19-1999]
This Page Added By Revision 0
Page _3-TVAN CALCULATION DESIGN VERIFICATION FORM Calculation Identifier CD-NOOOI-990113 Revision 1 Method of design verification used:
- 1. Design Review E
- 2. Alternate Calculation D
- 3. Qualification Test a_
Comments:
Minor changes that are not technical in nature therefore a detailed design review within a strict interpretation of the design review was not required.
- `AW TVA40533 [01-19991 NEDP-2-4 101-08.19991 This replaced by revision I
. .. I . .
TVAN CALCULATION TABLE OF CONTENTS Calculation Identifier: CD-NOOOI-990113 l Revision: I Page 4 TABLE OF CONTENTS SECTION TITLE PAGE Coversheet 1 Revision Log 2 Independent Review Form 3 Table of Contents 4 1.0 Purpose 5 2.0 Assumptions 5 3.0 Requirements/Limiting Conditions 5 4.0 References 5 5.0 Design Input Data 5 6.0 Supporting Graphics 5 7.0 Computations and Analysis 5 8.0 Summary of Results 5 9.0 Attachments 6
'A' - Browns Ferry Nuclear Plant - Increased MSIV Leakage Tech Spec Change Submittal - Seismic Evaluation Report 75 pages B - Seismic Evaluation for the Condensers 15 pages ACT - Seismic Verification of the Main Steam Piping and Supports Associated with the MSIV Alternate Leakage Treatment Pathway 29 pages 4p,/- Mwlfil AAbF *f*4 tW4 I Ke;,Iv 9c<5)tjw dS',Of I3 TVA 40710101.19991 -r,. eA.-r RL,5?.jCFP 9~ Izv NEDP-2-3 101-08-1999]
___ CALCULATION SHEET Document: CD-NOOOI-990113 l Rev.: 0 Plant: BFN 0 Page: -
Subject:
Seismic Evaluation Report Prepared By:. AC77 Date:. 9?/9/l((
Checked By: Date:_
SECTION 1.0: PURPOSE The purpose of this calculation is to document the "Browns Ferry Nuclear Plant -
Increased MSIV Leakage Tech Spec Change Submittal Seismic Evaluation Report",
(200918-R-002); August 1999, By EQE International, Oakland, CA. Additionally, Bounding Calculations (200918-C-002) "Seismic evaluation for the Condensers" and (200918-C-001) "Seismic verification of the main steam drain piping and supports associated with the MSIV alternate leakage treatment pathway" are contained within this calculation.
SECTION 2.0: ASSUMPTIONS:
There are no unverified assumptions in this calculation.
SECTION 3.0: REQUIREMENTS/LIMITING CONDITION:
This calculation does not generate any requirements or limiting conditions which limit system or plant operation from that currently documented in the design, place special requirements on a safety evaluation, or place special requirements on the physical configuration that are generally outside the stated purpose of the calculation.
SECTION
4.0 REFERENCES
See Attachment A, Section 5.; Attachment B, Section 3.0 And Attachment C, Section 3.0 for references used by EQE. alko 5 a V.?' of kAtic t.l SECTION 5.0 DESIGN INPUT DATA:
See Attachment A ,B, And C, design input data is identified where it has been used.
Alo oU { r4 ects Il'sh t
SECTION 6.0 SUPPORTING GRAPHICS:
See Attachment A, sheet 5 TABLES and FIGURES; Attachment B, sheet 2 TABLES and FIGURES, and Attachment C, sheet 2 TABLES and FIGURES for graphics used by EQE.
5 odVo QC4( s oil farw.-es of4 lq, tci eLs s W AtiPfac WT D.IA SECTION 7.0 COMPUTATIONS AND ANALYSIS:
See Attachment A, Section 5. for computations and analysis performed by EQE X1p 0~k7 "e^U-t-o-Lzc4* ot at reos ;D A2k--A SECTION K.0
SUMMARY
See Attachment A, Section 4.1.3 5.0 antion 5.0.
,S thio PagtAchde t B, Sey t in tchmnnt C, Section 5 This Page Added By Revision 0
CALCULATION SHEET SECTION 9.0 ATTACHMENTS:
Attachment A, "Browns Ferry Nuclear Plant - Increased MSIV Leakage Tech Spec Change Submittal Seismic Evaluation Report", (200918-R-002); August 1999, By EQE International, Oakland, CA.
Attachment B, Bounding Calculation (200918-C-002) "Seismic evaluation for the Condensers"; August 1999, By EQE International, Oakland, CA.
Attachment C, Bounding Calculation (200918-C-001) 'Seismic verification of the main steam drain piping and supports associated with the MSIV alternate leakage treatment pathway"; August 1999, By EQE International, Oakland, CA.
('4tV A g o~
farS 5,';
aC at I
/ASII 55IOW S\/.$c rogam IA This Page Added By Revision 0
CD-NOOO1-990113 Page 9I Attachment A 20091 a-R-0R Revision 0 August 31, 1999 Page 1 of 75 d
Browns Ferry Nuclear Plant Increased MSIV Leakage Tech Spec Change Submittal Seismic Evaluation Report Preparedfor.
TENNESSEE VALLEY AUTHORITY EQE Job Number: 501 47.1.6 &200918n01
CD-NO001-990113 Paige 0?
Attachment Pl 20091 8-R-002 Revision 0 August 31, 1999 Page 2 of 75 APPROVAL COVER SHEET twTrNi fONAI.
Title:
Browns Ferry Nuclear Plant Increased MSIV Leakage Tech Spec Change Submittal - Seismic Evaluation Report Report Number: 200918-R-002 Client: Tennessee Valley Authority Project Number: 200918.01 Revision Approval Number Date Prepared Reviewed Approved 0 8/31/99 ESE
CD-NOOO1-990113 Page-0 Attachment -. 20091 8-R-002 Revision 0 August 31, 1999 Page 3 of 75 TABLE OF REVISIONS Revson No. . :Description of Revision .Date 0 Initial Issue August 31, 1999
--- mixnQ1 R~mstbrat-doc
[OE
CD-Noooi-990113 Page I? I Attachment 200918-R-002 Revision 0 August 31,1999 Page 4 of 75 TABLE OF CONTENTS Paae
- 1. INTRODUCTION ........................................................ 8
- 2. SEISMIC EXPERIENCE DATABASE COMPARISONS .......................... 9 2.1 Seismic Ground Motions ........................................................ 9 2.2 Piping, Equipment, and Other Plant Features .............. ................. 11
- 3. SEISMIC VERIFICATION WALKDOWNS ................................................ 24 3.1 Seismic Verification Review Guidelines ......................................... 24 3.1.1 Piping, Pipe Support and Equipment Design Attributes ..... 25 3.1.2 Seismic Anchor Movement Issues ....................... .............. 26 3.1.3 Seismic Interaction Issues (Il/I and Proximity) ........... ......... 26 3.1.4 Valve Design Attributes .................................................. 26 3.2 Seismic Verification Boundary .................................................. 26 3.3 Walkdown Results ......................... ......................... 28 3.3.1 Unit 3 Seismic Walkdown .................................................. 28 3.3.2 Unit 2 Seismic Walkdown .................................................. 29 3.3.3 Additional Seismic Walkdown ............................................ 30
- 4. SEISMIC ASSESSMENTS ........................................................ 41 4.1 Outlier Resolution ......................................... 41 4.1.1 Seismic Demand .41 4.1.2 Seismic Capacity .42 4.1.3 Summary of Results .42 4.2 Alternate Leakage Treatment Piping and Supports ....................... 43 4.3 Turbine Building .......... 44 4.4 Condenser ................................................... 46
- 5. REFERENCES ........................................................ 75 PA2009 18R-00R1\subrpt.doc
CD-NOOO1-990113 Page A Attachment 20091 8-R-002 Revision 0 August 31, 1999 Page 5 of 75 TABLES Paie 3-1 Browns Ferry MSIV Leakage Boundary Flow Diagrams ............ ............... 32 3-2 BFN MSIV Leakage Boundary Points ...................................................... 33 3-3 Browns Ferry Unit 3 MSIV Walkdown Outliers ........................................... 36 3-4 Browns Ferry Unit 2 MSIV Walkdown Outliers ........................................... 38 4-1 Browns Ferry Unit 2 MSIV Outliers Resolution Summary ............ .............. 48 4-2 Browns Ferry Unit 3 MSIV Outliers Resolution Summary ............ .............. 50 4-3 Design Basis for Browns Ferry ALT Related Piping and Supports ............ 53 4-4 Seismic Experience Database Piping Data ............................................... 56 4-5 Comparison of Browns Ferry and Selected Database Piping Parameter .. 62 4-6 Bounding Evaluation of Typical Support Configurations ............. ............... 63 4-7 Browns Ferry Turbine Building Design Basis ............................................ 64 4-8 Comparison of Browns Ferry and Selected Database Condensers ........... 65 FIGURES 2-1 Comparison of Browns Ferry DBE Ground Spectrum and Selected Database Site Spectra .13 2-2 Comparison of Browns Ferry DBE and Valley Steam Plant Ground Spectra .14 2-3 Comparison of Browns Ferry DBE and Burbank Power Plant Ground Spectra .15 2-4 Comparison of Browns Ferry DBE and El Centro Steam Plant Ground Spectra .16 2-5 Comparison of Browns Ferry DBE and Moss Landing Power Plant Ground Spectra .17 PA\20091 8.R*01I\subrpt.doc
CD-NOOOI-990113 Page 4 Attachment P 200918-R-002 Revision 0 August 31, 1999 Page 6 of 75 FIGURES (CONT.)
Page 2-6 Comparison of Browns Ferry DBE and Humboldt Bay Nuclear Power Plant Ground Spectra .......................... ............................ 18 2-7 Comparison of Browns Ferry DBE and Coolwater Power Plant Ground Spectra ...................................................... 19 2-8 Comparison of Browns Ferry DBE and Commerce Refuge to Energy Plant Ground Spectra ............................ .......................... 20 2-9 Comparison of Browns Ferry DBE and Grayson Power Plant Ground Spectra ...................................................... 21 2-10 Comparison of Browns Ferry DBE and Ormond Beach Power Plant Ground Spectra ............... ....................................... 22 2-11 Comparison of Browns Ferry DBE and PALCO Cogeneration Plant Ground Spectra ............... ....................................... 23 3-1 Browns Ferry MSIV Seismic Verification Boundary .................................... 40 4-1 Comparison of Browns Ferry and Selected Database Piping DAt Ratios ...................................................... 67 4-2 Size Comparison of Browns Ferry Condenser with Selected Database Condensers ............... ....................................... 68 4-3 Weight Comparison of Browns Ferry Condenser with Selected Database Condensers ................ ....................................... 69 4-4 Height Comparison of Browns Ferry Condenser with Selected Database Condensers ............... ....................................... 70 4-5 Plan Dimension Comparison of Browns Ferry Condenser with Selected Database Condensers ...................................................... 71 4-6 Schematic Plan View of Browns Ferry Condenser Anchorage .......... ........ 72 PA2009 I 8-R-001\subrpt.doc
- - . - - .- -- --- - - I CD-NO001-990113 Page - v Attachment P 20091 8-R-002 Revision 0 August 31, 1999 Page 7 of 75 FIGURES (CONT.)
Pane 4-7 Comparison of Browns Ferry and Selected Database Condenser Anchorage to Seismic Demand for Direction Parallel to the Turbine Generator Axis ................. 73 4-8 Comparison of Browns Ferry and Selected Database Condenser Anchorage to Seismic Demand for Direction Transverse to the Turbine Generator Axis ............................................................................. 74 n -- ea I A--n.tk IR subot~doc ESE
CD-NOOO1-990113 Page 29 R0 Attachment fib 200918-R002 Revision 0 August 31, 1999 Page 8 of 75
- 1. INTRODUCTION This report summarizes the engineering activities performed for the supplemental plant specific Main Steam piping seismic verification to support the increased Main Steam Isolation Valve (MSIV) leakage tech spec change at Browns Ferry Nuclear Plant (BFN).
The verification program was performed in accordance with the recommendations of the General Electric Boiling Water Reactor Owners Group (BWROG) Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems (Reference 1).
The U.S. Nuclear Regulatory Commission (NRC) has reviewed the BWROG report and issued a safety evaluation report (SER) on its application for addressing the MSIV leakage issues (Reference 2), subject to certain limitations.
Engineering activities associated with the supplemental plant specific seismic verification program, as recommended in the BWROG report, consist of the following key elements:
Seismic Experience Database Comparisons Seismic Verification Walkdowns
- Seismic Assessments of Selected Components Detailed discussions of each of these activities are presented in the following sections of the report.
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CD-NOOO1-990113 Page 200918-R-002 Attincbment Revision 0 August 31, 1999 Page 9 of 75
- 2. SEISMIC EXPERIENCE DATABASE COMPARISONS The seismic experience data are derived from an extensive database on the performance of power plants and industrial facilities in past strong-motion earthquakes.
These performance data are compiled by EQE for the Seismic Qualification Utility Group, the Electric Power Research Institute and others, and included over 100 facilities in more than 60 earthquakes that have occurred around the world from 1934 to present.
Of interest to the MSIV leakage issues are the performance of the non-seismically analyzed main steam system piping, related components and supports, and condensers.
The BWROG Report (Reference 1) summarizes data on the performance of main steam piping and condensers in past strong-motion earthquakes and compares these piping and condensers with those in typical U.S. GE Mark 1,11, and IlIl nuclear plants. The earthquake experience data and similarity comparisons are then used to draw conclusions on how the GE piping and condensers would perform in a design basis earthquake (DBE).
The following sections present experience database comparisons that are plant-specific to Browns Ferry Nuclear Plant for use to support the increased MSIV leakage tech spec change submittal.
2.1 SEISMIC GROUND MOTIONS Ground motion estimates of 13 database sites were reviewed and accepted by the NRC staff for inclusion in the BWROG's earthquake experience database, and are presented in the referenced NRC Safety Evaluation Report (SER, Reference 2). To establish applicability of the BWROG's earthquake experience-based methodology for demonstrating the seismic ruggedness of non-seismically analyzed main steam piping and associated components at Browns Ferry, comparisons of the ground response spectra of selected database facilities with BFN design basis ground spectrum were made.
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CD-NOOO1-990113 Page 201R0 Attachment 200918R002 Revision 0 August 31, 1999 Page 10 of 75 The majority of the MSIV alternate leakage treatment (ALT) path and associated piping systems and the condensers at Browns Ferry are located in the lower elevations of the Turbine Building. BFN Turbine Building is classified as a Class 11structure, hence, no dynamic analysis of the building was performed. The building below the operating floor is a reinforced concrete framed structure supported on steel H-piles to bedrock. The horizontal ground spectrum is conservatively taken as the BFN 5% damped design basis DBE input spectrum (0.2g Housner spectrum defined at rock outcrop) and scaled by 1.6 to account for soil amplification.
A composite comparison of the ground response spectra of selected earthquake experience database facilities with the Browns Ferry design basis DBE ground spectrum is shown in Figure 2-1. The selected ground motions include the following 10 sites from among the 13 database facilities reviewed and accepted by the NRC:
- Valley Steam Plant - USGS estimate 1971 San Fernando Earthquake(M6.6)
- Burbank Power Plant - USGS estimate 1971 San Fernando Earthquake (M6.6)
- El Centro Steam Plant - E/W direction 1979 Imperial Valley Earthquake(M6.6)
- Moss Landing Power Plant - PG&E estimate 1989 Loma Prieta Earthquake (M7.1)
- Humboldt Bay Nuclear Power Plant - Average 1975 Ferndale Earthquake (M5.5)
- Coolwater Power Plant - Longitudinal direction 1992 Landers Earthquake (M7.3)
- Commerce Refuge to Energy Plant (LA Bulk Mail) - N/S direction 1987 Whittier Narrows Earthquake(M5.9)
- Grayson Power Plant (Glendale) - N200E direction 1971 San Fernando Earthquake (M6.6)
- Ormond Beach Power Plant - N270E direction 1973 Point Mugu Earthquake (M5.8)
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CD-N0ool-990113 Page AJll Attachment rj 200918-R-002 Revision 0 August 31, 1999 Page 11 of 75 PALCO Cogeneration Plant (Rio Dell) - Average 1992 Petrolia Earthquake (M6.9)
The individual comparison plots of the 5% damped ground spectra of the above 10 database facilities with the Browns Ferry DBE ground spectrum are shown in Figures 2-2 to 2-11. In general, the earthquake experience database sites have experienced strong ground motions that are in excess of the Browns Ferry DBE at the frequency range of interest (i.e., about 1 Hz. and above), with the exception of the Ormond Beach site. Many of the database site ground motions envelope the conservatively estimated BFN DBE ground spectrum by large factors in various frequency bands within the 1 Hz.
and above range.
Based on the above observations and comparison, it is concluded that the Browns Ferry DBE ground spectrum is generally bounded by those of the earthquake experience database sites at the frequencies of interest. Hence, the use of earthquake experience-based approach for demonstrating the seismic ruggedness of non-seismically analyzed main steam piping and associated components at Browns Ferry, consistent with the BWROG's recommendations and limitations of the SER, is appropriate.
2.2 PIPING, EQUIPMENT AND OTHER PLANT FEATURES The main steam piping and condensers in the earthquake experience database exhibited substantial seismic ruggedness, even when they are typically not designed to resist earthquakes. This is a common conclusion in studies of this type on other plant items such as welded steel piping in general, anchored equipment such as motor control centers, pumps, valves, structures, and so forth. That is, with limited exceptions, normal industrial construction and equipment typically have substantial inherent seismic ruggedness, even when they are not designed for earthquakes. No failures of the main steam piping were found. Anchored condensers have also performed well in past earthquakes with damage limited to minor internal tube leakage.
The BWROG Report (Reference 1) contains detailed discussions and comparisons of main steam piping and condenser design in several earthquake experience database P:\emp\2009 I 8\subrpt.doc S
CD-NOOO1-990113 Page A1 2 Attachment i 20091 8-R-002 Revision 0 August 31, 1999 Page 12 of 75 sites and example GE Mark 1,11, and Ill plants in the U.S. The general conclusions of these comparisons are as follows:
GE plant designs are similar to or more rugged than those in the earthquake experience database that exhibited good earthquake performance; The possibility of significant failure in GE BWR main steam piping or condensers in the event of an eastern U.S. design basis earthquake is highly unlikely; and that Any such failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.
Plant-specific comparisons of the main steam piping, related components and supports, and condensers at Browns Ferry with those in the selected earthquake experience database facilities are provided in Section 4 of this report.
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20091 8-R-002 Revision 0 August 31, 1999 Page 13 of 75 1.6 1.4 1.2 of 1 I 0
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20091 8-R-002 Revision 0 August 31, 1999 Page 14 of 75 Valley Steam Plant, CA (1971 San Fernando Earthquake)
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20091 8-R-002 Revision 0 August 31, 1999 Page 15 of 75 Burbank Power Plant, CA (1971 San Fernando Earthquake) 2.4 - -
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200918-R-002 Revision 0 August 31, 1999 Page 16 of 75 El Centro Steam Plant, CA (1979 Imperial Valley Earthquake) 2.4 2
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20091 8-R-002 Revision 0 August 31, 1999 Page 17 of 75 Moss Landing Power Plant, CA (1989 Loma Pr!eta Earthquake) 2.4 -
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20091 8-R-002 Revision 0 August31, 1999 Page 18 of 75 Humboldt Bay Nuclear Power Plant, CA (1975 Ferndale Earthquake) 2.4 2
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200918-R-002 Revision 0 August31, 1999 Page 19 of 75 Coolwater Power Plant, CA (1992 Landers Earthquake) 2.4 - -
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200918-R-002 Revision 0 August 31, 1999 Page 20 of 75 Commerce Refuge to Energy Plant, CA (1987 Whittier Narrows Earthquake) 2.4 2
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20091 8-R-002 Revision 0 August 31, 1999 Page 21 of 75 Grayson Power Plant, Glendale, CA (1971 San Fernanado Earthquake)
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20091 8-R-002 Revision 0 August 31, 1999 Page 22 of 75 Ormond Beach Power Plant, CA (1973 Point Mugu Earthquake) 2.4 2
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20091 8-R-002 Revision 0 August31, 1999 Page 23 of 75 PALCO Cogeneratlon Plant, CA (1992 Petrolla Earthquake) a 0
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CD-NOOO1-990113 Page q2q 2 8 Atachment R PRevision 0 August 31, 1999 Page 24 of 75
- 3. SEISMIC VERIFICATION WALKDOWNS Very few components of nuclear plant systems are unique to the nuclear facilities.
Nuclear plant systems include equipment, piping, tubing, conduit, and many other items that are common components of conventional power plants and industrial facilities.
Seismic experience data based methods have been developed to address seismic issues associated with the adequate performance of these equipment and commodities not designed, procured and installed to current nuclear seismic criteria. By reviewing the performance of the database facilities that contain equipment similar to that found in nuclear plants, conclusions can be drawn about the performance of nuclear plant equipment during and after earthquake events.
Extensive work has been performed documenting the performance of power plant equipment performance and the common sources of seismic damage to equipment and piping. In general, equipment, piping and tubing systems in the seismic experience database have performed very well in earthquakes, even though they were typically designed for deadweight and operating loads only, with little or no consideration for seismic loads. Performance of piping and equipment in.past earthquakes are summarized in Appendix D of the BWROG Report (Reference 1). Earthquake experience-based methods provide the basis for the seismic review of the main steam piping and equipment within the MSIV alternate leakage treatment (ALT) boundary at BFN.
3.1 SEISMIC VERIFICATION REVIEW GUIDELINES Various design attributes of the as-installed scope of equipment, piping, and tubing were reviewed and evaluated by the Seismic Walkdown Teams to ensure that the BFN installations are representative of database design practice and that components are free of known seismic vulnerabilities. Earthquake experience has identified conditions that have resulted in failure of piping and tubing systems and components. The conditions evaluated in the walkdown reviews included:
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CD-NOOO1-990113 Page_ 2 1 -
Attachment Attach j ent~.Revision 200eO91,8-R-002 0
August 31, 1999 Page 25 of 75 Piping, Pipe Support and Equipment Design Attributes Seismic Anchor Movement Issues Seismic Interaction Issues (Il/I & Proximity)
- Valve Design Attributes The above design attributes and conditions are briefly discussed below.
3.1.1 Piping, Pipe Support and Equipment Design Attributes The Seismic Walkdown reviewed the piping and tubing systems, and associated supports to ensure that the design attributes and conditions are consistent with good design and industry standard practices. The systems were also screened to ensure that they are free from known seismic vulnerabilities identified from earthquake experience data. These design attributes include:
- Piping with dead weight support spacing greatly in excess of the B31.1 suggested spans, or tubing with excessive sagging.
- Heavy, unsupported in-line components.
- Non-standard fittings or unusual attachments that could cause excessive localized stresses.
Pipe supports that exhibit non-ductile behavior.
- Presence of severe corrosion.
In addition, anchorage of terminal equipment to piping and tubing systems were reviewed for adequacy.
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CD-NOOO1-990113 Page f.A Attachment a 200918-R-002 Revision 0 August 31, 1999 Page 26 of 75 3.1.2 Seismic Anchor Movement Issues The experience database includes instances of seismic damage to piping, tubing and supports that were attributed to seismic anchor movement. Damage was the result of excessive movement of terminal end equipment, differential movement between supports in adjacent buildings, and excessive movements imposed on branch lines by flexible headers. These attributes were evaluated during the piping walkdowns.
3.1.3 Seismic Interaction Issues (Il/I and Proximity)
The seismic interaction review was a visual inspection of structures, piping, or equipment adjacent to the components under evaluation. The seismic interaction review evaluated conditions where seismically induced failures (Il/I) and displacements of adjacent structures, piping, or equipment (proximity) could adversely affect the required seismic performance of the system and components under consideration.
3.1.4 Valve Design Attributes Screening guidelines are provided for valves that are relied upon to establish the ALT pathway or are part of the Seismic Verification Boundary. The guidelines are consistent with the SQUG Generic Implementation Procedure (GIP, Reference 5) and include provisions for air-operated diaphragm valves, spring-operated pressure relief valves, piston-operated valves of light-weight construction, motor-operated valves, and substantial piston-operated valves.
3.2 SEISMIC VERIFICATION BOUNDARY The walkdown scope included the Main Steam drain path that will be established to convey leakage past the outboard Main Steam Isolation Valves (MSIV) to the isolated condenser and includes piping, instrumentation, valves and equipment that would be required to maintain the drain pathway.
The Seismic Verification Boundary for the MSIV Alternate Leakage Treatment path was developed in consultation with TVA Browns Ferry Systems Engineering, and is shown in P:\temp\2009 18\subrpt.doc
CD-NO001-990113 Page 209 -
Attachmentfl 20091 8R-002 Revision 0 August 31, 1999 Page 27 of 75 Figure 3-1. The associated flow diagrams are listed on Table 3-1, and the piping isolation boundaries defining the seismic verification boundary are shown on Table 3-2.
The Seismic Verification Boundary generally consists of the following portions of the Main Steam (MS) system beyond the outboard MSIV's:
- 1. Main Steam drain path to the condenser for any leakage past the isolated outboard MSIVs.
- 2. Main Steam piping from the outboard MSIV to the Main Steam Stop Valves (MSV).
- 3. Main Steam Bypass piping from the Main Steam lines to the Bypass Valve chest.
- 4. Main Condensers.
- 5. Additional piping and instrumentation within the Seismic Verification Boundary includes:
. Stop Valve Above Seat Drains to Condenser
- Steam Sample System
- HPCI/RCIC Steam Drains to Main Steam
- Auxiliary Boiler Drains to Main Steam
- Main Steam Instrumentation
- Main Steam Supply to the Reactor Feed Pumps
- Main Steam Supply to the Steam Jet Air Ejectors
- Main Steam Supply to the Off-Gas Preheaters The above Seismic Verification Boundary was originally developed for Unit 3 seismic walkdown. The Unit 2 Seismic Verification Boundary was less than that shown above for Unit 3. The original Unit 2 boundary assumed the addition of an isolation valve to isolate the steam path to the RFP Turbines and that the steam feed shutoff valve 8-575 would be qualified as an isolation boundary to the Steam Seal system. The Unit 2 Seismic Verification Boundary will be expanded and additional walkdown will be performed during the Unit 2 Cycle 11 outage to remove the assumptions of the isolation valves noted above, hence, eliminating the unit differences with Unit 3 Seismic Verification Boundary.
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CD-NO001-990113 Page A00 Attachment A 200918-R-002 Revision 0 August 31, 1999 Page 28 of 75 3.3 WALKDOWN RESULTS Field walkdowns of the main steam lines, ALT drain path and associated appendages within the Seismic Verification Boundary were conducted during the Unit 3 recovery outage in April 1995, and during the Unit 2 refueling outage in April 1996 by EQE engineers. Plant specific guidance, systems expertise and support were provided by BFN Site Engineering staff. All members of the MSIV Seismic Verification Walkdown Teams are degreed engineers, have ten to twenty years of experience in structural engineering and/or earthquake engineering application to nuclear power plants, and are familiar with the earthquake experience methodology. EQE engineers have performed the complete. MSIV Seismic Verification Walkdowns in accordance with the recommendations of the GE NEDC-31858P (BWROG Report, Reference 1) at several other plants.
Results of the Seismic Verification Walkdowns, including the identified walkdown open items or 'Outliers", are discussed in detail in References 3 and 4 for Browns Ferry, Unit 3 and Unit 2, respectively. A brief summary of the walkdown results is presented below, with walkdown outliers summarized in Table 3-3 and 3-4 for Browns Ferry, Unit 3 and Unit2, respectively.
3.3.1 Unit 3 Seismic Walkdown The main steam drain piping included in the Unit 3 MSIV alternate leakage treatment (ALT) path to the condenser generally conform to ANSI B31.1 design guidelines. Piping are typically insulated, and constructed from carbon steel, SA-106 Grade B, with butt-welded or socket-welded joints. In addition, pipe supports consist of a combination of rigid struts and U-bolt brackets, floor-mounted stanchions, and spring or rod hangers.
The as-installed configurations are inherently rugged and are similar to those found in the earthquake experience database facilities that have performed well during past earthquakes.
The piping systems within the Unit 3 MSIV Seismic Verification Boundary were divided into the following 13 portions for walkdown purposes:
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CD-NOOO1-990113 Page q 20091 R Attnchent ~Revision 0 August 31, 1999 Page 29 of 75
- 1. Main Steam drain line in the Turbine Building
- 2. Main Steam lines in the Turbine Building
- 3. Main Steam and Main Steam drain lines in the Reactor Building MSIV vault
- 4. HPCI/RCIC/Auxiliary Boiler drains in the Reactor Building and above the Torus
- 5. Main Steam PT instrumentation lines
- 6. Main Steam sampling lines to the Sample Station
- 7. Main Steam bypass lines
- 8. Main Steam stop valve above seat drains
- 9. Steam supply to Steam Seal Regulators
- 10. Steam supply to RFP Turbines
- 11. Steam supply to Steam Jet Air Ejectors
- 12. Steam supply to Off-Gas Preheaters
- 13. Condensers Conditions not meeting the Seismic Verification Review guidelines, as discussed in Section 3.1 of this report, were identified and documented as "Outliers" for further evaluation and resolution by the Seismic Walkdown Teams. These conditions included limited numbers of piping overspans, equipment anchorage or support integrity issues, proximity or falling interaction concerns, flexibility concerns due to seismic anchor movements or differential displacements, boundary valve integrity issues, and general maintenance or housekeeping items. Table 3-3 presents a summary of Unit 3 MSIV walkdown outliers.
3.3.2 Unit 2 Seismic Walkdown Similar to Unit 3, the main steam drain piping included in the Unit 2 MSIV alternate leakage treatment (ALT) path to the condenser generally conform to ANSI B31.1 design guidelines. Piping are typically insulated, and constructed from carbon steel, SA-106 Grade B, with butt-welded or socket-welded joints. Pipe supports consist of a combination of rigid struts and U-bolt brackets, floor-mounted stanchions, and spring or rod hangers. The as-installed configurations are inherently rugged and are similar to those found in the earthquake experience database facilities that have performed well during past earthquakes.
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CD-NOOO1-990113 Page 201R0 Attachment J9 20091 8-R002 Revision 0 August 31, 1999 Page 30 of 75 The piping systems within the scope of the original Unit 2 MSIV Seismic Verification Walkdown Boundary were divided into the following 11 portions for walkdown purposes:
- 1. Main Steam drain line in the Turbine Building
- 2. Main Steam lines in the Turbine Building
- 3. Main Steam and Main Steam drain lines in the Reactor Building MSIV vault
- 4. HPCI/RCIC/Auxiliary Boiler drains in the Reactor Building and above the Torus
- 5. Main Steam PT instrumentation lines
- 6. Main Steam sampling lines to the Sample Station
- 7. Main Steam bypass lines
- 8. Main Steam stop valve above seat drains
- 9. Steam supply to Steam Feed valve 8-575 (proposed isolation boundary)
- 10. Steam supply to RFP Turbines (with proposed manual isolation valve to be located on the Turbine Building operating deck, El. 617')
- 13. Condensers Conditions not meeting the Seismic Verification Review guidelines, as discussed in Section 3.1 of this report, were identified and documented as "Outliers" for further evaluation and resolution by the Seismic Walkdown Teams. As in the Unit 3 walkdown, these conditions included limited numbers of piping overspans, equipment anchorage or support integrity issues, proximity or falling interaction concerns, flexibility concerns due to seismic anchor movements or differential displacements, boundary valve integrity issues, and general maintenance or housekeeping items. Table 3-4 presents a summary of the Unit 2 MSIV walkdown outliers.
As mentioned in Section 3.2 above, the original Unit 2 Seismic Verification Boundary will be expanded and additional walkdown will be performed during the Unit 2 Cycle 11 outage to remove the assumptions of the isolation valves, hence, eliminating the unit differences with Unit 3 Seismic Verification Boundary.
3.3.3 Additional Seismic Walkdown As mentioned in Section 3.2 above, the Unit 2 Seismic Verification Boundary will be expanded to include portions of the steam supply lines from the Main Steam Header to P:\20091 6R-001\subrpt.doc
CD-NOOO1-990113 Page 311 2 Attachment 200918-R-002 Revision 0 August 31, 1999 Page 31 of 75 the turbine drives for the Reactor Feed Pumps, the Steam Jet Air Ejectors, the Off-Gas Preheaters, and the Steam Seal Regulators, i.e., extension of piping portions 9 and 10, and portions 11 and 12, as in the Unit 3 walkdown scope. The resulting Unit 2 Seismic Verification Boundary will then be consistent with that of Unit 3, hence, eliminating any unit differences between them. Additional seismic verification walkdown for the expanded scope will be performed during the Unit 2 Cycle 11 outage to verify the seismic ruggedness of the MS piping and associated components, and all identified outliers will be resolved during the same outage. Design Change Notice (DCN) will address any physical changes to restore the drain path into compliance.
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CD-Noool-990113 Page 62 Attachment A~ 20091 8-R-002 Revision 0 August 31, 1999 Page 32 of 75 Table 3-1 BROWNS FERRY MSIV LEAKAGE BOUNDARY FLOW DIAGRAMS Drawing Number System Description Unit 2 2-47E801-1 Main Steam System 2-47E801-2 Main Steam System 2-47E805-3 Heater Drains & Vents and Miscellaneous Piping Systems 2-47.807-1 Turbine Drains and Miscellaneous Piping Systems 2-47E807-2 Turbine Drains and Miscellaneous Piping Systems 2-47E812-1 High Pressure Coolant Injection Systen 2-47E813-1 Reactor Core Isolation Cooling System 0-47E815-1 Auxiliary Boiler System 2-47E815-4 2-47E610-43-1 Sampling and Water Quality System Unit 3 3-47E801-1 Main Steam System 3-47E801 -2 Main Steam System 3-47E805-3 Heater Drains & Vents and Miscellaneous Piping Systems 3-47E807-1 Turbine Drains and Miscellaneous Piping Systems
.3-47E807-2 Turbine Drains and Miscellaneous Piping Systems 3-47E812-1 High Pressure Coolant Injection System 3-47E813-1 Reactor Core Isolation Cooling System 3-47E815-5 Auxiliary Boiler System 3-47E610-43-6 Sampling and Water Quality System P;\2O091I8-R*001\subrpt.doc ESE
. . . . . .. - .- - . . - .. -- I -I- - - . - - . ...- . . .. .. - - . .
CD-NOOO1-990113 Page 1)33 Attachment A 20091 8-R-002 Revision 0 August 31, 1999 Page 33 of 75 Table 3-2 BFN MSIV LEAKAGE BOUNDARY POINTS Leakage Flow Diagram/
Boundary Point* Drawing I Comment FCV-1-15 47E801-1 MSIV for Main Steam Line A FCV-1 -27 47E801-1 MSIV for Main Steam Line B FCV-1-38 47E801-1 MSIV for Main Steam Line C FCV-1-52 47E801.1 MSIV for Main Steam Line D FCV-1-56 47E801-1 Outboard Containment Isolation valve for Primary Containment steam drains 1-521 47E801-1 Normally closed Main Steam Drain manual isolation 1-527 valves43-631 2-47E610-43-1 Normally closed Main Steam Sample System manual 3-47E610-43-6 isolation valve 43-631A 2-47E610-43-1 Normally closed Main Steam Sample System manual 3-47E610-43-6 isolation valve 43-632 2-47E610-43-1 Normally closed Main Steam Sample System manual 3-47E610-43-6 isolation valve FCV-1-74 47E801-2 Main Turbine Stop Valve for Steam Line A FCV-1-78 47E801-2 Main Turbine Stop Valve for Steam Line B FCV-1-84 47E801-2 Main Turbine Stop Valve for Steam Line C FCV-1 -88 47E801-2 Main Turbine Stop Valve for Steam Line D FCV-1 -61 FCV-1 -62 FCV-1 -63 FCV-1-64 FCV-1-65 47E801-2 Main Steam Bypass Valve Chest FCV-1 -66 FCV-1 -67 FCV-1 -68 FCV-1-69 FCV-73-6B 47E812-1 Normally open air operated isolation valve - HPCI FCV-71-6B 47E813-1 Normally open air operated isolation valve - RCIC 12-635 2-47E815-4 Normally closed manual isolation valve - Aux. Boiler 3-478815-5 P:\20091 B-R-OOIsubrpt.doc ESE
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Attachment -a 200918-R-002 Revision 0 August31, 1999 Page 34 of 75 Table 3-2 (CONT.)
BFN MSIV LEAKAGE BOUNDARY POINTS Leakage Flow Diagram/
F Boundary Point' Drawing
- Comment 12-637 2-47E815-4 Normally closed manual isolation valve - Aux. Boiler 3-47B815-5 12-623 2-47E815-4 Normally closed manual isolation valve - Aux. Boiler 3-47B815 5 12-625 2-47E815-4 Normally closed manual isolation valve - Aux. Boiler 3-47B815-5 2-12-822 0-47E815-1 Normally closed manual isolation valve - Aux. Boiler (Unit 2 only)
FCV-6-100 47E807-1 Normally closed motor operated isolation valve - Stop valve above seat drains FCV-6-101 47E807-1 Normally closed motor operated isolation valve - Stop valve above seat drains FCV-6-102 47E807-1 Normally closed motor operated isolation valve - Stop valve above seat drains FCV-6-103 47E807-1 Normally closed motor operated isolation valve - Stop valve above seat drains FCV-1-127 47E801-2 Reactor Feed Pump Turbine A Stop Valve FCV-1-135 47E801-2 Reactor Feed Pump Turbine B Stop Valve FCV-1-143 47E801-2 Reactor Feed Pump Turbine C Stop Valve FCV-6-153 47E807-2 Normally closed motor operated isolation valve - RFP FCV-6-155 47E807-2 Normally closed motor operated isolation valve - RFP FCV-6-157 47E807-2 Normally closed motor operated isolation valve - RFP FCV-6-122 47E807-2 Normally closed motor operated isolation valve - RFP FCV-6-127 47E807-2 Normally closed motor operated isolation valve - RFP FCV-6-132 47E807-2 Normally closed motor operated isolation valve - RFP PCV-1-151 47E801-2 Normally open air operated isolation valve - SJAE PCV-1 -166 47E801-2 Normally open air operated isolation valve - SJAE PCV-1-153 47E801-2 Normally open air operated isolation valve - SJAE PCV-1-167 47E801-2 Normally open air operated isolation valve - SJAE 6-826 47E805-3 Check valve - SJAE 6-822 47E805-3 Check valve - SJAE PA2009 18-R-00 1~subrpt.doc ME
CD-NOOO1-990113 Page 3A Attachment P 20091 8-R-002 Revision 0 August 31, 1999 Page 35 of 75 Table 3-2 (CONT.)
BFN MSIV LEAKAGE BOUNDARY POINTS Leakage Boundary Point' Flow Diagram)
Drawing* J Comment FCV-1 -145 47E807-2 Normally closed motor operated isolation valve - Steam Seal Regulator FCV-1 -154 47E807-2 Normally closed motor operated isolation valve - Steam Seal Regulator FCV-1-147 47E807-2 Air operated pressure regulating valve - Steam Seal Regulator CKV-1 -742 47E801-2 Check valve (NEW) - Off-Gas Preheater A CKV-1 -744 47E801-2 Check valve (NEW) - Off-Gas Preheater B Condenser A --- The condenser is the ultimate boundary for the MSIV
__ leakage path.
Condenser B --- The condenser is the ultimate boundary for the MSIV leakage path.
Condenser C --- The condenser is the ultimate boundary for the MSIV leakage path.
Miscellaneous test, 47E801-1 vent, drain and 47E801-2 instrument connections NOTE:
- Boundary component ID's and flow diagram/drawing nos. are generally applicable to both Units 2 and 3, unless noted otherwise specifically (i.e., 2- for Unit 2; 3- for Unit 3; and 0- for common)
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CD-NOOO1-990113 Page .A3i6 200918-R-002 Attachment pA Revision 0 August 31, 1999 Page 36 of 75 Table 3-3 BROWNS FERRY UNIT 3 MSIV WALKDOWN "OUTLIERS" SYSTEM DESCRIPTION I ID1 I OUTLIER 2 A I ID[V Main Steam Drain Line-Turbine 1 Bldg.
MS Drain Taps 1-1 MS Line differential motion MS Drain Taps 1-2 Impact with conduit supports X FCV 1-58 1-3 Extended valve operators X FCV 1-58/59 Conduit 1-4 Unknown routing at TB/RB joint Main Steam Lines - Turbine 2 Bldg.
MS Stop Valves 2-1 Valve performance X X MSH-17 2-2 Missing eyebolt nut X MSH-17,18 & 19 2-3 Grating clearance Main Steam Drain Line- MSIV 3 Vault FCV 1-15,27,38 & 52 3-1 Valve performance X X FCV 1-56 3-2 Manual operator X X
HPCL/RCICDrain 4 HPCI Drain at MS drain 4-1 Inadequate bending leg X connection x
MS PT 1-72, 76, 82, 86 & 93 5 MS instrument tubing 5-1 Overspan on 1" pipe to PT 1-86 X 1/2 Line to PT 1-86 5-2 Interaction with steel & pipe Main Steam Sample to Station 6 x Sample lines B & D 6-1 Missing tubing support clamps X Sample lines A, B, C, D 6-2 Inadequate flex legs at MS line PT 16AIB 6-3 Inadequate flex legs at MS line x
Sample Station 6-4 Temperature bath anchorage X Main Steam Bypass 7 X Main Steam Bypass Valve 7-1 Valve performance X x SVAbove Seat Drains 8 FCV 6-100,101,102,103 8-1 Short rod hangers X Steam to Steam Seal Regulator 9 MS to FCV 1-146 9-1 Overspan piping PCV 1-147 9-2 Handwheel proximity to WF X
PCV 1-147 airline 9-3 Inadequate flexibility & blockwall XI PCV 1-147 9-4 Extended valve operator . _ _
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CD-NOOOI-990113 Pagefj7 Attaichment - 20091 8-R-002 Revision 0 August 31, 1999 Page 37 of 75 Table 3-3 (CONT.)
BROWNS FERRY UNIT 3 MSIV WALKDOWN "OUTLIERS" SYSTEM DESCRIPTION l ID1 OUTLIER 2 lAI FI P l lV Steam Supply to RFP Turbines 10 Steam supply line 10-1 Inadequate flex leg at MS X header Steam supply line 10-2 Stanchion supports X Steam supply line 10-3 TB crane overhead X RFP Stop Valve above seat 10-4 Large mass on the 1/2 & 314 X drains inch lines Tubing to Pi 1-126 10-5 Missing tubing clamps - X overspan Steam Supply to SJAE 11 SJAE 3A/B 11-1 Anchorage X SJAE 3B 11-2 Loose anchor bolt nut X Drain to Condenser 11-3 Drain ties to multi system X collector Steam to Off-Gas Preheaters 12 PCV 1-175A/B 12-1 Masonry wall X Steam supply line to FCV 12-2 Vert. restraint of line at X 1-178A1B FCV 1-178 PCV 1-175AJB, FCV 1-178A/B 12-3 Valve performance X Condenser 13 Condenser and anchorage 13-1 Evaluate condenser/anchorage X adequacy KEY TO ISSUES:
A Anchorage or Support Capacity F Failure and Falling (I1/I)
P Proximity and Impact D Differential Displacement V Valve Screening NOTES:
1 - ID - Refers to MSIV Walkdown package identifier.
2 - Outliers" are plant conditions which require further evaluation.
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. , , , '. - ._. . I CD-NOOO1-990113 Page -CP 36 Attachment -4 20091 8-R-002 Revision 0 August 31, 1999 Page 38 of 75 Table 3-4 BROWNS FERRY UNIT 2 MSIV WALKDOWN "OUTLIERS" SYSTEM DESCRIPTION IID1 OUTLIER2 IAIF IP D I IV Main Steam Drain Line-Turbine I Bldg.
MS Drain Taps 1-1: MS Line differential motion FCV-1 -58 1-2: Extended valve operators 'X Main Steam Lines - Turbine Bldg. 2 MS Stop Valves 2-1: Valve performance X X RB MSIV Vault - MS and MS Drain 3 FCV-1-15, -27, -38 & -52 3-1: Valve performance X HPCI/RCIC/Aux. Boller Drains 4A HPCI Drain at MS drain connection 4-1: Inadequate bending leg X HPCI Drain in RB Steam Vault 4-2: Piping overspan X HPCI Drain in RB SE Corner Rm 4-3: Piping overspan X HPCLIRCIC/Aux. Boiler Drains 4B HPCI & Aux. Boiler drain lines 4-4: Miscellaneous maintenance X supports items HPCI Drain above the Torus 4-5: Piping overspan X RCIC Drain above the Torus 4-6: Inadequate support X MS PT-1-72, -76, -82 & -86 5 1/2" PT Piping from Steam Lines 5-1: Interaction with platform steel Main Steam Sampling 6 PT-16AtB Piping 6-1: Interaction with Feedwater X piping Sample lines A, B. C, D 6-2: Inadequate flex legs at MS line PT-1 6ANB 6-3: Inadequate flex legs at MS line Sample Station 6-4: Temperature bath X anchorage PT-1 6AB 6-5: Interaction with oil drum x Main Steam Bypass 7 Main Steam Bypass Valve 7-1: Valve performance X X SVAbove Seat Drains 8 FCV-6-100, -101, -102 & -103 8-1: Short rod hangers X 1' Drain Piping from Steam Line D 8-2: Interaction with MS x pipinq/steel PAtemp\20091 8\subrpt.doc ESE
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CD-Nooo1-990113 Page- j3cj Attachment 4 20091 8-R-002 Revision 0 August 31, 1999 Page 39 of 75 Table 3-4 (CONT.)
BROWNS FERRY UNIT 2 MSIV WALKDOWN "OUTLIERS" SYSTEM DESCRIPTION ID1 OUTLIER 2 TA FJP D V Steam to Steam Feed Valve 9 Rod Hanger Downstream of Valve 9-1: Disengaged rod hanger X 8-575 Verification Boundary Valve 8-575 9-2: Normally open manual X (Proposed in the Original Scope) valve Steam Supply to RFP Turbines 10 Steam supply line 10-1: Inadequate flex leg at MS X header Steam supply line 10-2: Stanchion supports X Steam supply line 10-3: TB overhead crane X X Verification Boundary Valve 1I-RFPT 10-4: Installation of valve X (Proposed in the Original Scope)
Condensers 13 Condenser anchorage 13-1: Evaluate anchorage X _
KEY TO ISSUES:
A Anchorage or Support Capacity F Failure and Falling (Il/I)
P Proximity and Impact D Differential Displacement V Valve Screening NOTES:
1 - ID - Refers to MSIV Walkdown package identifier.
2 - Outliers' are plant conditions which require further evaluation.
P:'demp\2009 18Rsubrpt.doc ESE
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CD-N0001-990113 Page- ,,9L Attachment 200918-R-002 Revision 0 August 31, 1999 Page 41 of 75
- 4. SEISMIC ASSESSMENTS As part of the supplemental plant specific seismic verification program to support the increased MSIV leakage tech spec change at BFN, various engineering evaluations and assessments were performed to verify the seismic adequacy of the Alternate Leakage Treatment (ALT) piping, related components and supports, and condensers. The following sections discuss the technical bases and methods used in these evaluations and assessments. Results of the seismic evaluations are also presented.
4.1 OUTLIER RESOLUTION Conditions which did not meet the walkdown screening guidelines (Section 3.1) or which were judged by the Seismic Walkdown Team to require further review were documented as 'Outliers" during the Units 2 and 3 Seismic Verification Walkdowns at Browns Ferry Nuclear Plant. For BFN Unit 3, the walkdown outliers have been resolved on a deterministic basis and dispositioned as described in more detail below. The proposed resolution for Unit 2 outliers will follow similar Unit 3 approaches and/or utilize existing Unit 3 analyses, as applicable. The Unit 3 outlier resolution are documented in BFN calculations (References 6 and 7).
4.1.1 Seismic Demand The BFN Turbine Building is classified as a Class 11structure, hence, no dynamic analysis of the building was performed and no in-structure response spectra were available for the structure. For seismic evaluations and outlier resolution, the horizontal seismic demand for components located within about 40 feet of the Turbine Building effective grade elevation (EL. 568') is conservatively taken as the BFN 5%damped design basis DBE input spectrum (0.2g Housner curve) scaled by 1.6 to account for soil amplification per BFN General Design Criteria (Reference 8) for soil founded structures, and 1.5 for building amplification per GIP. For components located above 40 feet of the Turbine Building effective grade elevation, an additional amplification factor of 1.5 is conservatively applied. In the vertical direction, seismic demand is taken as 2/3 that of the horizontal direction, with a soil amplification factor of 1.1 instead of 1.6.
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Attachment 4 20091 8-R-002 Revision 0 August 31, 1999 Page 42 of 75 4.1.2 Seismic Capacity For outlier resolution and evaluation of ALT piping, and related components and supports, the following load combinations and stress allowables, as applicable, were used:
Component Load Combination Stress Allowables Piping D + P + I +A 2.0 Sy (Primary + Secondary)
Pipe Supports D + T + I+ A AISC Equipment D+I AISC, GIP Anchorage Valve 3g load check GIP where, D- Dead load P- Pressure load T- Thermal load I- Seismic (DBE) inertial load A - Load due to seismic anchor movement Sy - Material yield strength at temperature AISC - American Institute of Steel Construction GIP - Generic Implementation Procedure 4.1.3 Summary of Results Table 4-1 provides a summary of the proposed resolution methods for the outliers associated with the Unit 2 MSIV Seismic Verification Walkdown. Similarly, the results of the resolution of outliers associated with the Unit 3 MSIV Seismic Verification Walkdown are summarized in Table 4-2.
P:A2009 1 8-R-001\subrpt.doc
[ME
CD-NOOO1-990113 Page_ j913 Attachment 200918-R-002 Revision 0 August 31, 1999 Page 43 of 75 As mentioned in Section 3.3.3 above, additional Unit 2 Seismic Verification Walkdown for the expanded scope will be performed during its Cycle 11 outage to verify the seismic ruggedness of the MS piping and associated components. Any additional outliers identified during this walkdown will be addressed and resolved within the same outage period. Design Change Notice (DCN) will address any physical changes to restore the drain path into compliance.
4.2 ALTERNATE LEAKAGE TREATMENT PIPING AND SUPPORTS Majority of the MSIV alternate leakage treatment (ALT) piping systems and related components at Browns Ferry, i.e., those portions downstream of the outboard Main Steam Isolation Valves (MSIV's) and the outboard Main Steam Drain Isolation Valve (MSDIV), are located in the Turbine Building and are not designated as Seismic Class I systems. In general, these piping systems are not seismically analyzed, and are typically designed to the requirements of USAS 831.1-1967.
As part of the plant specific seismic verification of the non-seismic ALT piping, related supports and components using the earthquake experience-based approach as outlined in the BWROG Report, the following reviews were performed to demonstrate that the piping and related supports fall within the bounds of the experience database:
Review of the design codes and standards, piping design parameters, and support configurations.
- . Seismic verification walkdown to identify potential piping concerns.
The Browns Ferry ALT piping systems consist of welded steel pipe and standard support components. Support spacing generally meets the B31.1 recommended span.
The design bases for the portions of piping associated with the ALT pathway to the condensers are tabulated in Table 4-3. Table 4-4 presents a general summary of the piping data that constitute the seismic experience data. Comparison of Browns Ferry and selected database piping parameters is presented in Table 4-5, along with Figure 4-1, which presents a comparison of D/t ratios of the BFN ALT drain piping with those PA20091 8-R-001\subrpt.doc
CD-NOOO1-990113 Page Aq4 Attachment A 200918-R-002 Revision 0 August 31,,1999 Page 44 of 75 found in the database. Overall, the BFN piping design is similar to and well represented by those found in the experience database sites that have shown to perform well in past earthquakes.
Browns Ferry FSAR does not reference Appendix A to 10 CFR Part 100. The seismic adequacy of the ALT piping is addressed by performing seismic verification walkdowns to identify specific design attributes associated with poor seismic performance, following the guidelines outlined in Section 3.1 of this report. Bounding evaluations were performed for typical support configurations using evaluation criteria as discussed in Section 4.1. Table 4-6 summarizes the results of the support and anchorage evaluations for the selected bounding configurations (Reference 1 0).
+/--.e r 1 xCAL e/it The seismic evaluations, consisting of verification walkdowns, bounding support evaluations, and resolution of the identified walkdown outliers, provide reasonable assurance that the ALT drain path piping, related supports and components will remain functional in the event of a Design Basis Earthquake (DBE) at Browns Ferry.
4.3 TURBINE BUILDING Performance of the turbine building and other non-seismic structures during a seismic event is of interest to the MSIV leakage issue only to the extent that the building structure and its internal components should survive and not degrade the capabilities of the selected main steam and condenser pathways. A BWROG (Reference 1) survey of this type of industrial structures has, in general, confirmed that excellent past seismic performance exists. There are no known cases of structural collapse of either turbine buildings at power stations or structures of similar construction.
The majority of the MSIV alternate leakage treatment (ALT) piping and the condensers at Browns Ferry are located in the Turbine Building, while small portions of the ALT piping are located in the Reactor Building which is a seismically designed, Class I structure. BFN Turbine Building is classified as a Class 11structure in the BFN FSAR. The BFN Design Criteria for Class 11structures are that they shall not degrade the integrity of any Class I structure. Those portions of Class 11structures required to remain structurally competent in order to support the operation of Class I
-_ i2- -, A , -r, g;-e P'.z-l I/
CD-NOOO1-990113 Page -J5 Attachment A 200918-R-002 Revision 0 August 31, 1999 Page 45 of.75 structures or equipment shall be designed for earthquake in accordance to the Uniform Building Code. Table 4-7 provides the design basis of the BFN Turbine Building and the applicable design codes used.
BFN Turbine Building below the operating floor at El. 617 feet is a reinforced concrete framed structure supported on steel H-piles to the bedrock at El. 519 feet. Piles are spaced far enough apart within each cluster to ensure that the maximum average unit bearing stress on the rock area is limited to 500 psi. Stresses in the piles are limited to one third of the yield stress. The concrete beams and slabs are designed to ACI 318-63 code using the working stress method. Similarly, the columns are also designed to ACI 318-63 code using the working stress method and checked by the ultimate strength design method using a load factor of 1.8.
The superstructure above the operating deck consists of transverse welded steel rigid frames spanning approximately 107 feet. An expansion joint is provided between a two-bay frame for Units 1 and 2, and a single-bay frame for Unit 3. For longitudinal expansion, the superstructure is provided with joints by using double rows of frames spaced at 4 feet apart. The steel frames, which form the Turbine Building structure above the concrete structure, are braced to provide rigidity in the direction of the Reactor Building as well as to provide support for the turbine cranes. These frames are designed to resist lateral forces from the overhead cranes and wind loads, in addition to supporting the vertical dead and live loads. The design of the steel superstructure is based on 1963 AISC code. All material conforms to ASTM-36, except for anchor rods which are ASTM A-307 steel. Shop connections are ASTM A-502 Gr. 1 rivets or welded, and field connections are ASTM A-325 high-strength bolts.
Based on the above design bases for the BFN Turbine Building, and the excellent seismic performance of this similar type of industrial structure in past strong-motion earthquakes as documented in the BWROG Report, the Browns Ferry Turbine Building is expected to remain structurally intact following a DBE.
CD-NO001-990113 Page 44~
Attachment A 200918-R-002 Revision 0 August31, 1999 Page 46 of 75 4.4 CONDENSER The BFN condensers consist of three single-pass, single pressure, radial flow type surface condensers. Each condenser is located beneath each of the three low pressure turbines, and is structurally independent. Table 4-8 lists the design data for BFN condensers and for the two experience database sites listed in the BWROG Report. In addition, design characteristic comparisons of the BFN condensers with the selected database condensers are shown in Figures 4-2 to 4-5. The BFN condenser design data is comparable to the data for these two database sites. The BFN condensers were also evaluated for structural integrity subject to seismic DBE loads. Results of the evaluation indicate that the condenser shell stresses are small. Maximum stress ratios, based on AISC allowables, are 0.12 for combined axial and bending and .0.10 for shear (Reference,)3^ r,10/s ChLck otr . 1Fl The condenser support anchorage consists of a center key and six support feet that are arranged as shown in Figure 4-6. The center support is a fixed anchor, and consists of a built-up wide flange H section embedded 4 feet into the concrete pedestal which is connected to the Turbine Building base mat, and welded to the bottom plate of the condenser. The support plates consist of 2 to 3 anchors of 2- to 2-1/2- inch diameter bolts. Each anchor bolt has greater than 5 feet nominal length with approximately 48 inches of embedment into the concrete pedestal which is connected to the Turbine Building base mat. These supports are designed to resist vertical operating loads, and are slotted radially from the center key to allow for thermal growth. Shear forces are transferred to the wide flange shaped anchor in the center and to the anchor bolts and shear keys to the support feet and carried through the concrete pedestal to the Turbine Building base mat.
The BFN condenser anchorage was compared with the performance of similar condenser in the earthquake experience database. The shear areas of the condenser anchorage, in the directions parallel and transverse to the turbine generator axis, divided by the seismic demand, were used to compare with those presented in the BWROG Report (Reference 1), and are shown in Figures 4-7 and 4-8, respectively.
The BFN condenser anchorage shear area to seismic demand is substantially greater than the selected database sites. The condenser support anchorage was also
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CD-NOOOI-990113 Page ,44q Attachment 200918-R-002 Revision 0 August 31, 1999 Page 47 of 75 evaluated and the results indicate that the combined seismic DBE and operational demand is less than the anchorage capacity based on the AISC allowables. Maximum stress ratios are 0.70 for bolt tension in the perimeter support feet, and 0.86 for shear in the center support built-up section (Reference 7).
The above comparisons of the condenser seismic experience data and the anchorage capacity evaluations demonstrate that the conclusions presented in the BWROG Report (Reference 1) can be applied to the BFN condensers. That is, a significant failure of the condenser in the event of a DBE at BFN is highly unlikely and contrary to the large body of historical earthquake experience data.
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CD-NOOO1-990113 Page - jfL Attachment - 20091 8-R-002 Revision 0 August 31, 1999 Page 48 of 75 Table 4-1 BROWNS FERRY UNIT 2 MSIV "OUTLIERS" RESOLUTION
SUMMARY
SYSTEM j OUTLIER l OUTLIER RESOLUTION DESCRIPTION DESCRIPTION METHOD Main Steam Drain Line-Turbine Bldg.
MS Drain Taps 1-1 MS line differential Modify supports per motion DCN FCV 1-58 1-2 Extended valve To be resolved per BFN operators Catc. CD-N0001-980038 Main Steam Lines MS Stop Valves 2-1 Valve performance To be resolved per BFN I
l__ CaIc. CD-NO001 -980038 Main Steam Drain Line- MSIV Vault FCV 1-15, 27, 38 & 52 3-1 Valve performance To be resolved per BFN Calc. CD-NOOO1-980039 HPCL/RCIC/Aux. Boiler Drains
-MSiV Pit HPCI Drain at MS drain 4-1 Inadequate bending Modify supports per connection leg DCN HPCI Drain in RB Steam Vault 4-2 Piping overspan Install new supports per DCN HPCI Drain in RB SE Corner 4-3 Piping overspan Install new supports per Room DCN HPCI/Aux. Boiler drain line 4-4 Misc. maintenance Misc. maintenance items supports items to be addressed by WR C340989 HPCI Drain above the Torus 4-5 Piping overspan Install new supports per
._ DCN RCIC Drain above the Torus 4-6 Inadequate support Modify support per DCN (RCIC-09) -
MS PT 1-72, 76, 82, 86 & 93 .
1/2 in. PT Piping from 5-1 Interaction with Re-route Steam Lines platform steel pipingA/nstrumentation line per DCN Main Steam Sample to Station __ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
PT-16ANB Piping 6-1 Interaction with Re-route piping and Feedwater piping modify support per DCN Sample lines A, B. C, D 6-2 Inadequate flex legs Remove existing at MS line supports and install new supports per DCN PA2009 I 8-R-00R1subrpt.doc ESE
CD-NOOOI-9901 13 Page A49 Attachment t 11R 20091 8-R-002 Revision 0 August31, 1999 Page 49 of 75 Table 4-1 (CONT.)
BROWNS FERRY UNIT 2 MSIV "OUTLIERS" RESOLUTION
SUMMARY
SYSTEM OUTLIER OUTLIER RESOLUTION DESCRIPTION DESCRIPTION METHOD Main Steam Sample to Station (cont.)
PT 16AB 6-3 Inadequate flex legs Modify supports per at MS line DCN Sample Station 6-4 Temperature bath Provide equipment anchorage anchorage per DCN PT-16A1B 6-5 Interaction with oil Initiate Work Request to drum relocate the oil drum Main Steam Bypass .
Main Steam Bypass Valve 7-1 Valve performance To be resolved per BFN
.__ ICatc. CD-NOOO1-980038 SVAbove Seat Drains FCV 6-100,101, 102, 103 8-1 Short rod hangers Modify rod hangers per DCN 1" Drain piping from Steam 8-2 Interaction with MS Re-route drain piping Line D piping/steel and modify support per DCN Steam to Steam Seal Regulator Rod hanger downstream 9-1 Disengaged rod Maintenance item to be of Valve 8-575 hanger addressed by WR C341864 Verification Boundary 9-2 Valve performance Walkdown scope to be Valve 8-575 (Proposed) expanded to remove the assumption Steam Supply to RFP Turbines Steam supply line 10-1 Inadequate flex leg at Modify supports per MS header DCN Steam supply line. 10-2 Stanchion supports Modify supports per DCN Steam supply line 10-3 TB overhead crane To be resolved per BFN Catc. CD-NO001-980039 Verification Boundary 10-4 Installation of Walkdown scope to be Valve 1-RFPT (Proposed) boundary valve expanded to remove the assumption Condenser Condenser and anchorage 13-1 Evaluate To be resolved per BFN adequacy I condenser/anchorage I Calc. CD-NO001-980038 PA20091I -R-001\subrpt.doc ESE
CD-NOOOI-990113 Page Ago Attachment A 200918-R-002 Revision 0 August 31, 1999 Page 50 of 75 Table 4-2 BROWNS FERRY UNIT 3 MSIV "OUTLIERS" RESOLUTION
SUMMARY
SYSTEM DESCRIPTION OUTLIER l OUTLIER l RESOLUTION l DESCRIPTION METHOD Main Steam Drain Line-Turbine Bldg.
MS Drain Taps 1-1 MS line differential Relocated three motion supports per DCN T40871A and BFN Calc.
No. CD-NO001-980039 MS Drain Taps 1-2 Impact with conduit Resolved per BFN Calc.
supports No. CD-NO001-980038 FCV 1-58 1-3 Extended valve Resolved per BFN Calc.
operators No. CD-NOOO1-980038 FCV 1-58/59 Conduit 1-4 Unknown routing at Resolved per BFN Calc.
TB/RB joint No. CD-NO001-980038 Main Steam Lines MS Stop Valves 2-1 Valve performance Resolved per BFN Calc.
_ No. CD-NO001 -980038 MSH-17 2-2 Missing eyebolt nut Nut replaced per WR C164362 MSH-17,18 & 19 2-3 Grating clearances Modified grating clearances per DCN T40871 A Main Steam Drain Line-MSIV Vault .
FCV 1-15, 27, 38 & 52 3-1 Valve performance Resolved per BFN Calc.
__ No. CD-NOOO1-980039 FCV 1-56 3-2 Manual operator Valve replaced by DCN
._ W17935A HPCI'RCIC/Aux. Boiler Drains - MSIV Pit HPCI Drain at MS drain 4-1 Inadequate bending Modified two supports connection leg per DCN T40871A and BFN Calc. No.
CD-NO001-980039 MS PT 1-72, 76, 82, 86 & 93 MS instrument tubing 5-1 Overspan on 1" pipe to Missing clamp replaced PT 1-86 per DCN T40871A 1/2 in. Line to PT 1-86 5-2 Interaction with steel & Re-route pipe piping/instrumentation line per DCN T40871A and BFN Calc. No. CD-NOOO -980039 P.1temp\20091ft\ubrpt.doc ESE
CD-NOOOI-9901 13 Page P i/
Attachment : 14 2009 18-R-002 Revision 0 August 31, 1999 Page 51 of 75 Table 4-2 (CONT.)
BROWNS FERRY UNIT 3 MSIV "OUTLIERS" RESOLUTION
SUMMARY
SYSTEM DESCRIPTION l OUTLIER l OUTLIER l RESOLUTION METHOD DESCRIPTION Main Steam Sample to Station Sample lines B & D 6-1 Missing tubing support Missing clamps replaced per clamps WR C193204 Sample lines A, B, C, D 6-2 Inadequate flex legs at Added four supports and MS line removed four supports per DCN T40871A and BFN Calc. No. CD-NOOO1-980039 PT 16A/B 6-3 Inadequate flex legs at Modified two supports per MS line DCN T40871A and BFN Calc. No. CD-NO001-980039 Sample Station 6-4 Temperature bath Anchorage provided per anchorage DCN T40871A and BFN
__ Calc. No. CD-NO001-980039 Main Steam Bypass Main Steam Bypass Valve 7-1 Valve performance Resolved per BFN Catc. No.
I CD-N0001-980038 SVAbove Seat Drains FCV 6-100, 101, 102, 103 8-1 Short rod hangers Modified rod hangers per DCN T40871A and BFN Calc. No. CD-NO001-980039 Steam to Steam Seal Regulator MS to FCV 1-1 46 9-1 Overspan piping Resolved per BFN Calc. No.
._ ._ . CD-NOOO1-980039 PCV 1-147 9-2 Hand wheel in proximity Resolved per BFN Catc. No.
to WF section
__ blockwall interaction CD-NOOOI-980039 PCV 1-147 9-4 Extended valve Resolved per BFN Calc. No.
operator CD-NO001-980039 PR\temp\2009 18\subrpt.doc ESE
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CD-NOOOI-990113 PoeAgz 20091 8-R-002 Attachment - Revision 0 August 31, 1999 Page 52 of 75
. ... Table 4-2 (CONT.)
BROWNS FERRY UNIT 3 MSIV "OUTLIERS" RESOLUTION
SUMMARY
SYSTEM DESCRIPTION OUTLIER l OUTLIER l RESOLUTION METHOD DESCRIPTION Steam Supply to RFP Turbines Steam supply line 10-1 Inadequate flex leg at Remove hanger per DCN MS header T40871A and BFN Calc. No.
CD-N0001 -980039 Steam supply line 10-2 Stanchion supports Replace two spring hangers per DCN T40871A and BFN Calc. No. CD-N0001-980039 Steam supply line 10-3 TB crane overhead Resolved per BFN Calc. No.
CD-NO001-980039 RFT Stop Valve above 10-4 Lass mass on 1/2 and Resolved per BFN CaIc. No.
seat drains 3/4 inch lines CD-NO001-980039 Tubing to Pi 1-126 10-5 Missing tubing clamps - Missing clamps replaced per overspan WR-C193201 Steam Supply to SJAE's SJAE 3A/B .11-1 Anchorage and cracked Anchorage resolved per BFN pedestal Calc. No. CD-N0001-980039; Cracked concrete pedestal repaired per WR-C193206 SJAE 3B 11-2 Loose anchor bolt nut Re-torqued loose nut per WR-C193205 Drain to Condenser 11-3 Drain ties to multi- Re-route piping per DCN system collector T40871A and BFN Cabc. No.
CD-NOOO1 -980039 Steam Supply to Off-Gas Preheaters PCV 1-175A/B 12-1 Masonry wall To be resolved by the proposed installation of NEW boundary valves to Preheaters A & B Steam supply line to 12-2 Vertical restraint of line Resolved per BFN Calc. No.
FCV 1-178A1B_ at FCV 1-178 CD-NO001-980039 PCV 1-175A/B, 12-3 Valve performance To be resolved by the FCV 1-1 78A/B proposed installation of NEW boundary valves to Preheaters A & B Condenser Condenser and anchorage 13-1 Evaluate Resolved per BFN Calc.
adequacy condenser/anchorage No. CD-N0001 -980038 PAtemip\20091 8\subrpt.doc ESE
200918-R-002 Revision 0 August 31, 1999 Page 53 of 75 Table 4-3 Design Basis for Browns Ferry ALT Related Piping and Supports .
Piping Design Design Pipe Pipe ..Piping Typical Piping Description Temp. Press. Size Sch. DIt Material Support Types Design CF) (psig) (NPS) . Basis MS Lines from outboard MSIV's to 562 1146 24 80 20 ASTM A-106 Spring hangers USAS MS Header and to Turbine Stop 1 160 5 Grade B Vertical struts B31.1- 1967 Valves 0
Main Steam Header 562 1146 24 80 20 ASTM A-155 Spring hangers USAS Grade KC-70 B31.1- 1967 MS Stop Valve Above Seat 562 1146 1 160 5 ASTM A-106 Rod hangers USAS El0 Leak-off Grade B B31.1- 1967 (= W Turbine Bypass Valve Header 562 1146 18 80 19 ASTM A-106 Rigid supports USAS II Grade B Rod and Spring hangers B31.1- 1967 r
MS Steam Supply to RFP Turbine 562 1146 6 80 15 ASTM A-106 Rod and Spring hangers USAS Stop Valves 4 80 13 Grade B Stanchion supports B31.1- 1967 MS Steam Supply from MS Header 562 1146 3 160 8 AStM A-106 Rod and Spring hangers USAS to SJAE's to the Condenser 2 160 7 Grade B B31.1- 1967 1-1/2 160 7 1 160 5 PA20091 8-R-00 I~subrpt.doc
20091 8-R-002 Revision 0 August 31. 1999 Page 54 of 75 Table 4-3 (CONT.)
Design Basis for Browns Ferry ALT Related Piping and Supports Piping Design Design Pipe Pipe l Piping Typical Piping Description Temp. Press. Size Sch. l D/t Materlal Support Types Design (F) (psig) (NPS) .
l Basis MS Steam Supply to Steam Seal 562 1146 4 80 13 ASTM A-106 Rod hangers USAS Regulators Grade B B31.1- 1967 MS Steam Supply from MS Header 562 1146 2 160 7 ASTM A-106 Rod hangers USAS 0t to the Off-Gas Preheaters A & B Grade B B31.1- 1967 > %Z
_ %O 2 160 7 ASTM A-335 New piping associated with the Grade P11 proposed installation of new boundary S o valves to Preheaters A &B MS Outboard Drains from MS Lines 562 1146 3 160 8 ASTM A-I 06 Stanchion supports USAS eb to the Main Drain Line 2 160 7 Grade B 831.1- 1967 1 160 5 *Y>o o'
3 160 8 ASTM A-333 -1bZ 2 160 7 Grade I Main Drain Line to the Condenser 562/ 1146/ 4 80 13 ASTM A-106 Rod and Spring USAS 450 400 3 160 8 Grade B hangers B31.1- 1967 1 160 5 Stanchion supports PA2009 18-ROO1'subrpt.doc
20091 8-R-002 Revision 0 August 31, 1999 Page 55 of 75 Table 4-3 (CONT.) II 11 Design Basis for Browns Ferry ALT Related Piping and Supports II II Piping Design Design Pipe Pipe Piping Typical Piping I Description Temp. Press. Size Sch. D/t Material Support Types Design I (OF) (psig) (NPS) Basis HPCI Drain to MS Drain; 450 400 2 160 7 ASTM A-106 Rigid supports USAS I RCIC Drain to HPCI Drain; 1 160 5 Grade B B31.1- 1967 Aux. Boiler Drains to HPCIIRCIC/ l 2
Reactor Building Drain Line 270 415 1 160 5 Misc. PT Instrument Lines 562 1146 1 160 5 ASTM A-106 Rigid supports USAS Sample Lines to Sample Station Grade B B31.1- 1967 0 4 .049" - ASTM A-213 Rigid supports -- =_
3 t tubing (wall ) SS Gr. TP-304 (tube clamps) cm CD ItD E
P:A200918-R-00O1ubpt.dc
CD-NO001-990113 Page_ 96 Attachment A*t 200918-R-002 Revision 0 August 31, 1999 Page 56 of 75 Table 4-4 Seismic Experience Database Piping Data Pipe Size Pipe Wall Facility (NPS) O.D. Schedule Thickness DAt (inch) (inch) 24 24.0 20 0.375 64 20 20.0 20 0.375 53 18 18.0 30 0.437 41 16 16.0 30 0.375 43 14 14.0 30 0.375 37 12 12.75 40 0.406 31 12 12.75 30 0.330 39 10 10.75 160 1.125 10 8 8.625 160 0.906 10 Valley Steam Plant 6 6.625 40 0.280 24 Units 1 & 2 4 4.50 160 0.531 8 4 . 4.50 40 0.237 19 3 3.50 160 0.437 8 3 3.50 80 0.300 12 3 3.50 40 0.216 16 2 2.375 160 0.343 7 2- 2.375 40 0.154 15 11/2 1.90 160 0.281 7 11/2 1.90 40 0.145 13 1 1.315 40 0.133 10
.__ 1.05 160 0.218 5 3/'4 1.05 40 0.113 9 P:\temp\20091 8\subrpt~doc ESE
CD-NOOOI-990113 PageA ;
Attachment ft 200918-R-002 Revision 0 August 31, 1999 Page 57 of 75 Table 4-4 (CONT.)
Seismic Experience Database Piping Data Pipe Size Pipe Wall Facility (NPS) O.D. Schedule Thickness DAt (inch) (inch) 20 20.0 STD 0.375 53 18 18.0 160 1.781 10 18 18.0 XS 0.500 36 18 18.0 STD 0.375 48 14 14.0 40 0.437 32 14 14.0 STD 0.375 37 12 12.75 160 1.312 10 12 12.75 STD 0.375 34 10 10.75 40 0.365 29 8 8.625 160 0.906 10 8 8.625 120 0.718 12 8 8.625 40 0.322 27 6 6.625 120 0.562 12 6 6.625 40 0.280 24 El Centro 4 4.50 80 0.337 13 Steam Plant 4 4.50 40 0.237 19 3 3.50 160 0.437 8 3 3.50 80 0.300 12 3 3.50 40 0.216 16 2 2.375 160 0.343 7 2 2.375 80 0.218 11 2 2.375 40 0.154 15 1Y2 1.90 160 0.281 7 1YM2 1.90 80 0.200 10 11/2 1.90 40 0.145 13 1 1.315 80 0.179 7 1 1.315 40 0.133 10
- h4 1.05 80 0.154 7 3/4 1.05 40 0.113 9 P;\Iemnp\209 I MsubrPt.doc ESE
I-
. - I ... . - --. . . I . . . -... ; 1. . . . . - - .. .- - . . --. . - . .
CD-NOOOl-990113 Page -Ags Attachment k 20091 8-R-002 Revision 0 August 31, 1999 Page 58 of 75 Table 4-4 (CONT.)
Seismic Experience Database Piping Data Pipe Size Pipe Wall Facility (NPS) O.D. Schedule Thickness D/t (inch) (inch) 16 16.0 . 1.394 11 12 12.75 . 1.148 11 8 8.625 160 0.906 10 8 8.625 30 0.277 31 6 6.625 160 0.562 12 6 6.625 40 0.280 24 4 4.50 160 . 0.531 8 4 4.50 80 0.337 13 Moss Landing 4 4.50 40 0.237 19 Units 1, 2 & 3 3-- 3.50 160 0.437 8 3 3.50 80 0.300 12 3 3.50 40 0.216 16 2 2.375 160 0.343 7 2 2.375 80 0.218 11 2 2.375 40 0.154 15 1% 1.90 160 0.281 7 1% 1.90 80 0.200 10 1 1.315 160 0.250 5 1 1.315 80 0.179 7 3/4 1.05 160 0.218 5 3/4 1.05 80 0.154 7 P:\temo\20091 8subrpt.doc ME
CD-NOOOl-990113 Page A o.....
Attachment 20091 8-R-002 Revision 0 August 31,1999 Page 59 of 75 Table 4-4 (CONT.)
Seismic Experience Database Piping Data Pipe Size Pipe Wall Facility (NPS) O.D. Schedule Thickness DIt (inch) (
(inch) 24 24.0 40 0.687 35 24 24.0 . 1.066 23
-- 18.8 -- 2.287 8 16 16.0 40 0.500 32 16 16.0 *- 0.902 18
- 13.2 -- 1.668 8 8 8.625 160 0.906 10 8 8.625 40 0.322 27 6 6.625 160 0.562 12 6 6.625 40 0.280 24 4 4.50 160 0.531 8 4 4.50 80 0.337 13 4 4.50 40 0.237 19 Moss Landing 3 3.50 160 0.437 8 Units 4 5 3 3.50 80 0.300 12
_3--- 3.50 40 0.216 .16 2 2.375 160 0.343 7 2 2.375 80 0.218 11 2 2.375 40 0.154 15 111/22 1.90 160 0.281 7 1%/2 1.90 . 80 0.200 10 1% 1.90 40 0.145 13 1 1.315 160 0.250 5 1 1.315 80 0.179 7 1 1.315 40 0.133 10 3/4 1.05 160 0.218 5 34 1.05 80 0.154 7 3 1.05 40 0.113 9 a .\*Ar\YnQ I R\kubrDt.doc ROM~
CD-NOOOI-9901 13 Page Attachment A 200918-R-002 Revision 0 August31,1999 Page 60 of 75 Table 4-4 (CONT.)
Seismic Experience Database Piping Data Pipe Size Pipe Wall Facility (NPS) O.D. Schedule Thickness DA (inch) (inch) 30 30.0 0.632 47 26 26.0 1.128 23 18 18.0 3.444 5 12 12.75 2.444 5 12 12.75 0.601 21 8 8.625 -- 1.650 5 8 8.625 40 0.322 27 6 6.625 -- 1.268 5 6 6.625 40 0.280 24 4 4.50 -- 0.861 5 4 4.50 80 0.337 13 4 4.50 40 0.237 19 3 3.50 80 0.300 12 3 3.50 40 0.216 16 2V2 2.875 -- 0.550 5 2'/2 2.875 80 0.276 10 Moss Landing 2122.875 40 0.178 16 Units 6 & 7 2 2.375 -- 0.519 5 2 2.375 80 0.218 11 2 2.375 40 0.154 15 11/2 1.90 0.428 4
- 1. 1.90 80 0.200 10 1Y2 1.90 40 0.145 13 1 1.315 -- 0.301 4 1 1.315 80 0.179 7 1 1.315 40 0.133 10
. ./4 1.05 160 0.218 5 3/4 1.05 80 0.154 7 3/4 1.05 40 0.113 9 1/2 1.05 -. 0.210 4
_/4 0.54 0.153 4 P:\tempWO09 18Nsubrpt.doc M02
CD-NOOO1-990113 Page 461 Attachment P 200918-R-002 Revision 0 August 31, 1999 Page 61 of 75 Table 4-4 (CONT.)
Seismic Experience Database Piping Data Pipe Size Pipe Wall Facility (NPS) O.D. Schedule Thickness DAt (inch) (inch)
Ormond Beach 30 30.0 1.298 23 Units 1 & 2 30 30.0 0.719 42 21 21.0 3.793 6 Humboldt Bay 12 12.75 80 0.687 19 Unit 3 10 10.75 80 0.593 18
_*6 6.625 80 0.432 15 P:\temD\2009 i 6\subrpt.doc EAE
....;.J CD-NOOO1-990113 Page oO6 Attachment A 20091 8-R-002 Revision 0 August 31, 1999 Page 62 of 75 Table 4-5 Comparison of Browns Ferry and Selected Database Piping Parameters Piping Parameter Browns Ferry Database Sites Pipe Diameter 1.315 - 24.0 1.05 - 30.0 (inch)
Wall Thickness 0.25 - 1.218 0.113 - 3.793 (inch)
Diameter-to-Thickness Ratio 5 - 20 4 - 64 (D/t)
P:\temp\20091 8\subrpt.doc
[mE
CD-NO001-990113 Page A 23 Attachment 200918R002 Revision 0 August 31, 1999 Page 63 of 75 Table 4-6 Bounding Evaluations of Typical Support Configurations Support Type Critical Component Stress Ratio I
Cantilever bracket Anchor bolts .73 Rod hanger Overhead weld .70 attachment PA\20091 8.R.001\subrpt.doc . ESE
CD-Noooi-99om1 Page Ajq Attachment A"' 20091 8-R-002 Revision 0 August 31, 1999 Page 64 of 75 Table 4-7 Browns Ferry Turbine Building Design Basis Design Attribute Description Lateral Force Resisting The Turbine Building above the operating deck is framed by System Above the transverse welded steel rigid frames with fixed bases and Operating Deck braced in the direction of the Reactor Building to provide the resistance to lateral forces.
Lateral Force Resisting The Turbine Building below the operating deck is a reinforced System Below the concrete structure. Concrete walls serve as shear walls for Operating Deck the lateral loads in the direction of the Reactor Building.
Design Codes General: Uniform Building Code (UBC)
Concrete: American Concrete Institute (ACI 318-1963)
Steel: American Institute of Steel Construction (AISC) -1963 Seismic Design Basis UBC zone 1 Wind Design Basis Wind speed of 100 mph PA2009 I8.R.001\subrpt.doc ESE
CD-NODOI-990113 Page~
Attnchment..J 200918-R-002 Revision 0 August31,1999 Page 65 of 75 Table 4-8 Comparison of Browns Ferry and Selected Database Condensers Design Moss Landing Ormond Beach Browns Attributes Units 6 & 7 Units 1 & 2 Ferry Condenser Ingersoll-Rand Southwestern Foster Wheeler Manufacturer Flow Type Single Pass Single Pass Single Pass Condenser Dimensions 65 ft. x 36 ft. 52 ft. x 27 ft. 58 ft. x 32 ft.
(LxWxH) x 47 ft. x 20 ft. x 47 ft.
Condenser 435,000 sq. ft. 210,000 sq. ft. 222,000 sq. ft.
Surface Area Condenser Shell Cu Bearing Cu Bearing ASTM A-285C Material ASTM A-285C ASTM A-285C Condenser Shell 3/4" 3/4" 7/8" Thickness Condenser Operating 3,115 kips 1,767 kips 2,076 kips Weight Tube Material Al-Brass 90-10 Cu-Ni Al-6XN Tube Size 1" dia. 1" dia. 7/8" dia.
Tube Length 65 ft. 53 ft. 50 ft.
Tube Wall Thickness 18 BWG 20 BWG 22 BWG P:~temp=0O91 ft'~ubrpt.doc ESE
CD-NOOOI-990113 Page AC; Attachment A 2Re 91 8-R-002 vision 0 Auclust 31, 1999 Pag le 66 of 75 Table 4-8 (CONT.)
Comparison of Browns Ferry and Selected Database Condensers Design Moss Landing Ormond Beach Browns Attributes Units 6 & 7 Units 1 & 2 Ferry Number of Tubes 25,590 15,220 19,480 Tube Sheet Muntz Muntz ASTM A-285C Material Tube Sheet 1-1/2" 1-1/4" 1-1/4" Thickness No. of Tube Support 15 14 15 Plates Tube Support Not Given Cu Bearing ASTM A-285C Plate Material ASTM A-285C Tube Support 3/4" 5/8" 7/8" Plate Thickness Tube Support 48 in. 36 in. 39 in.
Plate Spacing Water Box Material 2% Ni Cast Iron Cu Bearing ASTM A-285C ASTM A-48 ASTM A-285C Class 30 Expansion Joint Rubber Belt Stainless Steel Rubber Belt Hotwell Capacity 20,000 gal. 34,338 gal. 28,000 gal. (max.)
P*~ttmn\90091 8%SUbrvt.doc MAE
CD7-60001-990113 Page JA6t7 Attachment A 20091 8-R-002 Revision 0 August 31, 1999 Page 67 of 75 65 64 M Browns Ferry 2 25 '
El Database 20 20 - 19 15 16 15 15 - 13 Dit 13 13 10 t 8 7 7 0 8 5
5+ 5 ' 5 5 5 4 4 I I I I I 0 1 1 1/2 2 3 4 6 18 and Above Pipe Size (NPS)
Figure 4-1 Comparison of Browns Ferry and Selected Database Piping D/t Ratios PA\200918.R-OO1\subrpt.doc ESE
- . -- .-- - . --.- I -
CD-NOOOI-990113 Page 4 2009 18-R-002 Attachment ,A Revision 0 August31, 1999 Page 68 of 75 Size Comparison of Browns Ferry Condenser with Selected Database Condensers Browns Ferry 22ll000 Ormond Beach 21QOOO tAbss Landing 435.000
.IItI I I I I 0 50,000 100.000 150,000 -200.000 250.000 300,000 350.000 400,000 450.000 500,000 H-eat Transfer Area (tt)
Figure 4-2 Size Comparison of Browns Ferry Condenser with Selected Database Condensers P.Ntemp\2009 I B\subrpt.doc ESE
CD-NOOOI-990113 Page - A
- 20091 8-R-002 Attachment a Revision 0 August31, 1999 Page 69 of 75 Weight Comparison of Browns Ferry Condenser with Selected Database Condensers Browns Ferry 2.070,000 Ormond Beach 1,767,500 Moss Landing ,115.
0 500.000 1,000.000 1.500,000 2.000,000 2,500,000 3,000,000 3.500.000 Weight (Ibs)
Figure 4-3 Weight Comparison of Browns Ferry Condenser with Selected Database Condensers P:\temp\20091 8\subrpt.doc ESE
- .. d -.
CD-NOOO1-990113 Page -P99 20091 8-R-002 Attachment -A Revision 0 August31, 1999 Page 70 of 75 Height Comparison of Browns Ferry Condenser Height Comparison of Browns Ferry Condenser with Selected Database Condensers Brow ns Ferry 47 Ormond Beach __ 20 Moss Landing 47 Il MO I f-'-I 47I 0 10 20 30 40 50 Height (ft)
Figure 4:4 Height Comparison of Browns Ferry Condenser with Selected Database Condensers P.Memp\2009 I 8subrpt-doc . ESE
. . . - 4 CD-NOOO1-990113 Page A?9 Attachment A 20091 8-R-002 Revision 0 August 31, 1999 Page 71 of 75 LN1 Moss Landing 6 & 7 Il (65ft x 36 ft)
I Ormond Beach (52ft x 27ft)
Browns Ferry (50ft x 32ft)
Figure 4-5 Plan Dimension Comparison of Browns Ferry Condenser with Selected Database Condensers Egg
CD-Noool-990113 Pnge- AY Attachment 4 20091 8-R-002 Revision 0 August 31, 1999 Page 72 of 75 j
J.
I
.1 J.
I J.
Anchor bolts with slotted holes cirected from center anchor plate Anchor bolts with slotted holes perpendicular Fixed anchor plate Figure 4-6 Schematic Plan View of Browns Ferry Condenser Anchorage PANtemp\2O09 I 8\.subrpt-doc ESE
. I4 - .. I.' .. , Z%. -
I CD-NOOO-1990113 Page-A 73 20091 8-R-002 Attachment /,4 Revision 0 August 31, 1999 Page 73 of 75 Comparison of Browns Ferry Condenser Anchorage with Selected Database Condensers 0.0002 0
0 l3 Upper Bound 0.0001 +
l Lower Bound pa U, 0 Moss Landing El Centro Browns Ferry I Parallel to Turbine Generator Axis Figure 4-7 Comparison of Browns Ferry and Selected Database Condenser Anchorage to Seismic Demand for Direction Parallel to the Turbine Generator Axis-*
PAternp%2009 16\subrpt.doc . EVE
CD-Noool-990113 Page-,49q 20091 8-R-002 Attachment A Revision 0 August:31, 1999 Page 74 of 75 Com parison of Browns Ferry Condenser Anchorage with Selected Database Condensers c 0.00014 E
E 0.00012 a) o 0.0001 E 0.00008 OUpper Bound
, 0.00006 C') El Lower Bound w 0.00004 1 1 1]
a,0.00002 C))
Moss Landing El Centro Browns Ferry 1 Transverse to Turbine Generator Axis Figure 4-8 Comparison of Browns Ferry and Selected Database Condenser Anchorage to Seismic Demand for Direction Transverse to the Turbine Generator Axis P.\temp\20091I \subrpt.doc ESE
CD-NOOO1-990113 Page A15
- Attachment~ 200918-R-002 Revision 0 August 31,1999 Page 75 of 75
- 5. REFERENCES
- 1. "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems", GE NEDC-31858P, Revision 2, September 1993.
- 2. Safety Evaluation of GE Topical Report, NEDC-31858P, Revision 2, 'BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems",
U.S. Nuclear Regulatory Commission, March 3,1999.
- 3. 'Browns Ferry - Unit 3, MSIV Seismic Verification Summary Report", EQE Report No.
200621-R-001, Revision 0, September 1998.
- 4. 'Browns Ferry - Unit 2, MSIV Seismic Verification Summary Report", EQE Report No.
200918-R-001, Revision 0, August 1999.
- 5. Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Rev. 2A, March 1993, Prepared by Winston & Strawn, EQE, et al., for the Seismic Qualification Utility Group (SQUG).
- 6. BFN Calculation No. CD-NOOOI-980039, "Main Steam Seismic Ruggedness Verification".
- 7. BFN Calculation No. CD-NOOO1-980038, 'Main Steam Seismic Ruggedness Evaluation".
- 8. BFN General Design Criteria, BFN-50-C-7102, 'Seismic Design', Revision 3.
- 9. BFN Detailed Design Criteria, BFN-50-C-7306, "Qualification Criteria for Seismic Class 11Piping, Pipe Supports, and Components', Revision 1.
- 10. BFN Calculation No. CD-NO001-990113, uMSIV Seismic Evaluation Report". RIl
CD-NOOOI-990113 PagejJ.
Attachinent-k
- CALCULATION COVER SHEET INTrUlJZfONAL Calculation No. 200918-C-002 Project: TVA BFN MSIV LEAKAGE TECH SPEC CHANGE Calculation
Title:
ADDITIONAL SEISMIC EVALUATIONS FOR THE BFN CONDENSERS
References:
See Section 3.0 Attachments:
Total Number of Pages (Including Cover Sheet): l /5 l Revision Approval Number Date Description of Revision , Originator Checker Approver 0 8/30/99 ORIGINAL ISSUE F.R. BEIGI J.O. DIZON J.O. DIZON 7/ O .
J:Xbfnpmsiv\c3Ic91I802.doc
CD-NO001-990113 Page e, Attachment j SHEET NO. 2.
a JOB NO. 200918 JOB BFN MSIVTECH SPEC CHANGE BY DATE g/2 /II CALC. NO. C-002 SUBJECT ADDi~iONALSEISMIC EVALUATIONS FOR CHK 22-DATE _Oj/)
THE BFN CONDENSERS JMRtNAT7aNAL Table of Contents Page 1.0 PURPOSE ..................... 3 2.0 SCOPE & METHODOLOGY ..................... 3
3.0 REFERENCES
..................... 4 4.0 SEISMIC EVALUATIONS ..................... 5
5.0 CONCLUSION
.................... 15 Tables Table 1 - Comparison of Browns Ferry and Selected Database Condensers .10 Figures Figure 1 - Comparison of Browns Ferry DBE Ground Spectrum with Selected Database Site Ground Spectra 12 Figure 2- Comparison of Browns Ferry DBE and Moss Landing Power Plant Ground Spectra .13 Figure 3- Comparison of Browns Ferry DBE and Ormond Beach Power Plant Ground Spectra .14 J:\bfnpmsiv\calc91802.doc
CD-NOOOI-990113 Pagejj3 Attachment R SHEET NO. 3 JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE BY DATE L 2 -se CALC. NO. C-002 SUBJECT ADDITIONAL SEISMIC EVALUATIONS FOR CHK DATE e l
- m,> THE BFN CONDENSERS
_NTUNWA11NAL 1.0 PURPOSE The purpose of this calculation is to document the results of the additional seismic evaluation performed on the BFN condensers, as part of the seismic adequacy verification of the components associated with the MSIV Alternate Leakage Treatment (ALT) pathway.
2.0 SCOPE &METHODOLOGY The BFN condensers are the terminal boundary points of the MSIV alternate leakage treatment (ALT) pathway, hence, they are necessary to maintain structural integrity following a Design Basis Earthquake (DBE). The condensers are located in the Turbine Building and are not designated as Seismic Class I systems.
As part of the plant specific seismic verification of the non-seismic components using the earthquake experience-based approach as outlined in the BWROG Report (Reference 1), the following reviews are performed to demonstrate that the BFN condensers fall within the bounds of the experience database and/or exhibit adequate seismic capacity:
- Review of the condenser design codes and standards, design characteristics and parameters, and support/anchorage configurations.
- Verification walkdown to identify potential seismic interaction concerns.
- Engineering evaluations of the condenser and support configurations.
The BFN condensers are evaluated using both seismic experience data from past earthquakes and engineering analysis. Analytical evaluations of the condenser and support anchorage are performed In accordance with the guidelines in the Generic Implementation Procedure (GIP, Reference 5), and the general requirements of the American Institute of Steel Construction (AISC, Reference 6), as applicable.
J~bfnpmsiv\calc9 1802.doc
CD-NOOOI-990113 Page BY Attachment 'R SHEE NO.4 JOB NO. 200918 JOB BFN MSIVTECH SPEC CHANGE BY DATE y CALC. NO. C-002 SUBJECT ADDITIONAL SEISMIC EVALUATIONS FOR CHK DATE E THE BFN CONDENSERS
.JT'IjWA11ONAL
3.0 REFERENCES
- 1. "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems", GE NEDC-31858P, Revision 2, September 1993.
- 2. Safety Evaluation of GE Topical Report, NEDC-31858P, Revision 2, 'BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems", U.S.
Nuclear Regulatory Commission, March 3,1999.
- 3. 'Browns Ferry - Unit 2, MSIV Seismic Verification Summary Report", EQE Report No.
200918-R-001, Revision 0, August 1999.
- 4. "Browns Ferry - Unit 3, MSIV Seismic Verification Summary Report", EQE Report No.
200621-R-001, Revision 0, September 1998.
- 5. "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment", Rev. 2A, March 1993, Prepared by Winston & Strawn, EQE, et al., for the Seismic Qualification Utility Grotjp (SQUG).
- 6. AISC, "Manual of Steel Construction", Eighth Edition, 1980.
- 7. TVA Calculation No. CD-NO001 -980039, 'Main Steam Seismic Ruggedness Verification".
- 8. TVA Calculation No. CD-NO001-980038, "Main Steam Seismic Ruggedness Evaluation".
- 9. ASME, "Boiler and Pressure Vessel Code,Section III, Division I, Appendices",
1980 Edition.
J~\bfnpmsiv\calc91 802.doc
CD-NOOOI-990113 Page B*
Attachment fiP SHEET NO. .
a JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE BY DATE XC/St CALC. NO. C-002 SUBJECT ADDITIONAL SEISMIC EVALUATIONS FOR CHK DATE §/3 one THE BFN CONDENSERS
.M11XArAXJAL 4.0 SEISMIC EVALUATIONS The BFN condensers consist of three single-pass, single pressure, radial flow type surface condensers. Each condenser is located beneath each of the three low pressure turbines, and is structurally independent. Table 1 lists the design data for BFN condensers and for the two experience database sites listed in the BWROG Report (i.e., Moss Landing 6 & 7, and Ormond Beach 1 & 2). Design characteristic comparisons of the BFN condensers with the above two selected database condensers are presented in details in Reference 8. These include size (surface area), weight, height, and plan comparisons. The BFN condenser design data is comparable to the data for these two database sites.
The BFN condenser anchorage was compared with the performance of similar condenser in the earthquake experience database. The shear areas of the condenser anchorage, in the directions parallel and transverse to the turbine generator axis, divided by the seismic demand, were used to compare with those presented in the BWROG Report (Reference 1). The BFN condenser anchorage shear area to seismic demand is substantially greater than the selected database sites. The condenser support anchorage was also evaluated and the results indicate that the combined seismic DBE and operational demand is less than the anchorage capacity based on the AISC allowables. Maximum stress ratios are 0.70 for bolt tension in the perimeter support feet, and 0.86 for shear in the center support built-up section. Detailed description of the BFN condenser support anchorage and anchorage evaluations are presented in Reference 8.
A composite comparison of the ground response spectra of selected earthquake experience database sites with the conservatively estimated BFN DBE ground spectrum (i.e., 0.2g Housner input spectrum at rock outcrop scaled by 1.6 to account for soil amplification) is shown in Figure 1. In general, the earthquake experience database sites have experienced strong ground motions that are in excess of the BFN DBE at the frequency range of interest (.e., about 1 Hz.
and above), with the exception of the Ormond Beach site. Many of the database site ground motions envelope the conservatively estimated BFN DBE ground spectrum by large factors In various frequency bands within the 1 Hz. and above range. Figures 2 and 3 show the Individual comparison plots of the conservatively estimated BFN DBE ground spectrum with the Moss Landing and Ormond Beach site spectra, respectively.
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CD-Noool-990113 Pa e _ _
Attachment S SHEETr NO.6 JOB NO. 200918 JOB BFN MSIVTECH SPEC CHANGE BY DATE CALC.NO. C-002 SUBJECT ADDITIONAL SEISMIC EVALUATIONSFOR CHK DATE THE BFN CONDENSERS
,0MMAnNAfONAL The Ormond Beach Power Plant was affected by the magnitude 5.8, Point Mugu Earthquake in 1973, which was considered to be a relatively moderate earthquake, and was substantially lower than the 1989 Loma Prieta Earthquake (Magnitude 7.1) as experienced in the Moss Landing Power Plant as well as those experienced by most of the other database sites.
To ensure that adequate seismic margins exist in the BFN condensers in the event of a plant DBE, additional seismic evaluation was performed to verify the overall structural integrity of the condensers, as shown in pages 7 to 9 of this calculation. Results of the evaluation indicate that the condenser shell stresses due to the seismic DBE loads are small. Maximum stress ratios, based on AISC allowables, are 0.12 for combined axial and bending and 0.10 for shear.
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.I ' J9T S CdJOB NO. 200918 JOB 8FN MSIV TECH SPEC CHANGE BY DATE _At v /SA CALM. NO. C-002 SUBJECT ADDITIONAL SEISMIC EVALUATIONS FOR CHK DATE tL3 i tI:
THE BFN CONDENSERS jPMNAIUNAJAL Table 1 Comparison of Browns Ferry and Selected Database Condensers Design Moss Landing Ormond Beach Browns Attributes Units 6 & 7 Units 1 & 2 Ferry Condenser Ingersoll-Rand Southwestern Foster Wheeler Manufacturer Flow Type Single Pass Single Pass Single Pass Condenser Dimensions 65 ft. x 36 ft. 52 ft. x 27 ft. 58 ft. x 32 ft.
(LxWxH) x 47 ft. x 20 ft. x 47 ft.
Condenser 435,000 sq. ft. 210,000 sq. ft. 222,000 sq. ft.
Surface Area Condenser Shell Cu Bearing Cu Bearing ASTM A-285C Material ASTM A-285C ASTM A-285C Condenser Shell 3/4" 3/4" 7/8" Thickness Condenser Operating 3,115 kips 1,767 kips 2,076 kips Weight Tube Material Al-Brass 90-10 Cu-Ni Al-6XN Tube Size 1" dia. 1" dia. 7/8" dia.
Tube Length 65 ft. 53 ft. 50 ft.
Tube Wall Thickness 18 BWG 20 BWG 22 BWG J:\bfnpmsiv\calc9 1802.doc
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SHEET NO. /1 JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE By DATE r/S - 1fe9 CALC. NO. C-002 SUBJECT ADDITIONAL SEISMIC EVALUATIONS FOR CHK 5I DATE THE BFN CONDENSERS
,PhflRHA17W~AL Table 1 (cont.)
Comparison of Browns Ferry and Selected Database Condensers
- Design Moss Landing Ormond Beach Browns Attributeg Units 6 & 7. Uni 1 &
Number of Tubes 25,590 15,220 19,480 Tube Sheet Muntz Muntz ASTM A-285C Material Tube Sheet 1-1/2" 1-1/4" 1-1/4" Thickness No. of Tube Support 15 14 15 Plates Tube Support Not Given Cu Bearing ASTM A-285C Plate Material ASTM A-285C Tube Support 3/4" 5/8" 7/8" Plate Thickness Tube Support 48 in. 36 in. 39 in.
Plate Spacing Water Box Material 2% Ni Cast Iron Cu Bearing ASTM A-285C ASTM A-48 ASTM A-285C Class 30 Expansion Joint Rubber Belt Stainless Steel Rubber Belt Hotwell Capacity 20,000 gal. 34,338 gal. 28,000 gal. (max.)
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.KTEXNAflaNAL Figure 1 Comparison of Browns Ferry DBE Ground Spectrum with Selected Database Site Spectra e
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SHEET NO. /3 JOB NO. 200918 JOB BFN MSIVTECH SPEC CHANGE BY DATE j/jjf CALC. NO. C-002 SUBJECT ADDmONAL SEISMIC EVALUATIONS FOR CHK DATE _g2 i*g THE BFN CONDENSERS bNTTRNA1CHAL Figure 2 Comparison of Browns Ferry DBE and Moss Landing Power Plant Ground Spectra 2.4 2
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CALC. NO. C-002 SUBJECT ADDITIONAL SEISMIC EVALUATIONS FOR CHK DATE _________
THE BFN CONDENSERS WMTANAr"OAL
5.0 CONCLUSION
S The comparisons of the condenser seismic experience data, supplemented by the additional condenser evaluation and the anchorage capacity evaluations demonstrate that the conclusions presented in the BWROG Report (Reference 1) can be applied to the BFN condensers. That is, a significant failure of the condenser in the event of a DBE at BFN is highly unlikely and contrary to the large body of historical earthquake experience data.
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CD-NOOOI-990113 Page C,1 Attachment C
- CALCULATION COVER SHEET I Tx ONAL Calculation No. 200918-C-001 Project: TVA BFN MSIV LEAKAGE TECH SPEC CHANGE Calculation
Title:
SEISMIC VERIFICATION OF THE MS DRAIN PIPING AND SUPPORTS ASSOCIATED WITH THE MSIV ALTERNATE LEAKAGE TREATMENT PATHWAY
References:
See Section 3.0 Attachments:
Total Number of Pages (Including Cover Sheet): I :1 I Revision Approval Number Date Description of Revision Originator Checker Approver 0 8/30/99 ORIGINAL ISSUE F.R. BEIGI J.O. DIZON J.O. DIZON
_ _7
... , -- .-- 1-1n 0toJ ago
CD-NOOO1-990113 Page CZ Attachment C SHEET NO. 2 JOB NO. 200918 JOB BFN MS1V`TECH SPEC CHANGE BY DATE R S CALC. NO. C-001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK DATE /qq PIPING AND SUPPORTS 11KNfAMONAL Table of Contents Page 1.0 PURPOSE ......................................... 3 2.0 SCOPE & METHODOLOGY .......................................... ;.3
3.0 REFERENCES
.......................................... 4 4.0 SEISMIC EVALUATIONS .......................................... 5
5.0 CONCLUSION
S ......................................... 29 Tables Table 1 - Design Basis for Browns Ferry ALT Related Piping and Supports ........................ 16 Table 2 - Seismic Experience Database Piping Data .19 Table 3 - Comparison of Browns Ferry and Selected Database Piping Parameters .25 Table 4 - Bounding Evaluations of Typical Support Configurations .26 Figures Figure 1 - Comparison of Database Site Spectra to Browns Ferry DBE Ground Spectrum .... 27 Figure 2 - Comparison of Browns Ferry and Selected Database Piping D/t Ratios .28
--- o'.'. .. ' . I CD-NOOO1-990113 Page C3 Attachment C tPUL:1-Rd SHEET NO. 3
_.n JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE BY 7 DATE 4-Z3.-4 M__- CALC. NO. C-001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK vl DATE ____ ___
PIPING AND SUPPORTS M
KINAMCNAL 1.0 PURPOSE The purpose of this calculation is to document the seismic adequacy verification of the main steam drain piping and related supports that are associated with the MSIV Alternate Leakage Treatment (ALT) pathway.
2.0 SCOPE & METHODOLOGY The MSIV alternate leakage treatment (ALT) piping systems and related components at Browns Ferry, i.e., those portions downstream of the outboard Main Steam Isolation Valves (MSIV's) and the outboard Main Steam Drain Isolation Valve (MSDIV), are located in the Turbine Building and are not designated as Seismic Class I systems.
As part of the plant specific seismic verification of the non-seismic ALT piping, related supports and components using the earthquake experience-based approach as outlined in the BWROG Report (Reference 1), the following reviews will be performed to demonstrate that the piping and related supports fall within the bounds of the experience database:
- Review of the design codes and standards, piping design parameters, and support configurations.
- Seismic verification walkdown to identify potential piping concerns.
- Seismic evaluations of selected bounding support configurations.
Support evaluations will be performed in accordance to the general requirements of the American Institute of Steel Construction (AISC, Reference 6).
CD-NOOOI-990113 Page CL Attachment C SHEET NO. 4 JOB NO. 200918 JOB BFN MSIVTECH SPEC CHANGE BY ?8 DATE L_______
CALC. NO. C.001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK DATE ______
PIPING AND SUPPORTS
3.0 REFERENCES
- 1. "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems", GE NEDC-31858P, Revision 2, September 1993.
- 2. Safety Evaluation of GE Topical Report, NEDC-31858P, Revision 2, UBWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems", U.S.
Nuclear Regulatory Commission, March 3, 1999.
- 3. "Browns Ferry - Unit 2, MSIV Seismic Verification Summary Report", EQE Report No.
200918-R-001, Revision 0, August 1999.
- 4. "Browns Ferry - Unit 3, MSIV Seismic Verification Summary Report", EQE Report No.
200621-R-001, Revision 0, September 1998.
- 5. "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment", Rev. 2A, March 1993, Prepared by Winston & Strawn, EQE, et al., for the Seismic Qualification Utility Group (SQUG).
- 6. AISC, "Manual of Steel Construction", Eighth Edition, 1980.
- 7. USAS B31.1 - Power Piping, 1967. Also, ANSI/ASME B31.1 - Power Piping, 1983.
- 8. TVA Calculation No. CD-NO001-980039, "Main Steam Seismic Ruggedness Verification".
- 9. TVA Calculation No. CD-NO001 -980038, "Main Steam Seismic Ruggedness Evaluation".
CD-Noool-990113 Pa...e C i Attachment SHEET NO. 5 JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE BY DATE CALC. NO. C-001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK asFo<7 DATE'27H PIPING AND SUPPORTS U4AMICNAL 4.0 SEISMIC EVALUATIONS In general, the Browns Ferry ALT piping systems are typically designed to the requirements of USAS B31.1-1967 code (Reference 7), and consist of welded steel pipe and standard support components. Support spacing generally meets the B31.1 recommended span. The design bases for the portions of piping associated with the ALT pathway to the condensers are tabulated in Table 1. Table 2 presents a general summary of the piping data that constitute the seismic experience data. Figure 1 shows the comparison of the selected database site spectra with Browns Ferry DBE ground spectrum which indicates that the BFN DBE ground spectrum is generally bounded by those of the earthquake experience database sites at the frequencies of interest. Hence, the use of earthquake experience-based approaching for demonstrating the seismic ruggedness of non-seismically analyzed piping and related components at BFN, consistent with the BWROG's recommendations, is appropriate. Comparison of Browns Ferry and selected database piping parameters is presented in Table 3, along with Figure 2, which presents a comparison of D/t ratios of the BFN ALT drain piping with those found In the database. Overall, the BFN piping design is similar to and well represented by those found in the experience database sites that have shown to perform well in past earthquakes.
Browns Ferry FSAR does not reference Appendix A to 10 CFR Part 100. As such, bounding analysis for the selected portion of the ALT piping system is not required (Reference 2). The seismic adequacy of the ALT piping is addressed by performing seismic verification walkdowns to identify specific design attributes associated with poor seismic performance, following the guidelines as presented in the BWROG Report (Reference 1). The results of the walkdowns, including the resolution of the identified outliers, were presented in the respective MSIV Walkdown Summary Reports for Units 2 and 3 (References 3 and 4).
Furthermore, bounding evaluations are performed for typical support configurations as shown in pages 6 to 15 of this calculation. Table 4 summarizes the results of the support and anchorage evaluations.
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0 mo Design Basis for Browns Ferry ALT Related Piping and Supports >C- i 03 Piping Design Deslgn.1PS p e. ;**, Piping: : ;Typica ' .. asIg Description Temp. Prs.Prs ie~~ch~ .D -tI.T:atra.: ' SDesign' .~~zr (psig). NS .,. :,. .
-.4 MS Lines from outboard MSIVWs to MS Header and to Turbine Stop 562 1146 24 1
80 160 20 5
ASTM A-106 GradeB Spring hangers Vertical struts USAS B31.1- 1967 g
Valves -a ,o 00 oz 3> 4o 0
Main Steam Header 562 1146 24 80 20 ASTM A-155 Spring hangers USAS 0 j)
- i 6 Grade KC-70 B31.1- 1967 o C)rn 0
3D-I MS Stop Valve Above Seat 562 1146 1 ISO 5 ASTM A- 06 Rod hangers USAS C := W Leak-off Grade B B31.1- 1967 I 0) 0 -4 z
At ;I n-q-
-4 Turbine Bypass Valve Header 562 1146 18 80 19 ASTM A-106 Grade B Rigid supports Rod and Spring hangers USAS B31.1- 1967
-L co xa. co D
MS Steam Supply to RFP Turbine 562 1146 6 80 15 ASTM A- 06 Rod and Spring hangers USAS TN Stop Valves 4 80 13 Grade B Stanchion supports B31.1- 1967 MS Steam Supply from MS Header 562 1146 3 160 8 ASTM A-106 Rod and Spring hangers USAS to SJAE's to the Condenser 2 160 7 Grade B B31.1- 1967 1-112 160 7 .0 I C 1 160 5 A -<
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aa Design Basis for Browns Ferry ALT Related Piping and Supports '-0 co Piping Design Design -Pipe::.. Plpe. .. , : Piping Typical Piping
-.. cn0 Description Temp. Press.; Slze9 , Sch.D/t . .Material.. . Support!oTypes Design c- o F (psig) (NMS) ..... - Basis m w H 0
MS Steam Supply to Steam Seal 562 1146 4 80 13 ASTM A-106 Rod hangers USAS 4 I en n C,1 Regulators Grade 8 B31.1- 1967
- CD MS Steam Supply trom MS Header 562 1146 2 160 7 ASTM A-106 Rod hangers USAS to the OII-Gas Preheaters A & B Grade B B31.1- 1967 sa M,,3 3n_ -
-`I 2 160 7 ASTM A-335 Now piping associated with the w i! rn Grade P11 proposed Installation of new boundary valves to Preheaters A & B =CD
-L 0e e MS Outboard Drains from MS Unes 562 1146 3 160 8 ASTM A-106 Stanchion supports , USAS.
0 0
- S
- ts to the Main Drain Line 2 160 7 Grade B B31.1-.1967 1 160 5 I.
3 160 8 ASTM A-333 2 160 7 Grade 1 Main Drain Une to the Condenser 562/ 1146/ 4 E0 13 ASTM A-106 Rod and Spring USAS 450 400 3 160 8 Grade B hangers B31.1- 1967 0 1 160 5 Stanchion supports I _<
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C-Design Basis for Browns Ferry ALT Related Piping and Supports F03 Piping Design. Design. Pipe. -,Plpe i-. Pig : .':...,,Typical,- *..,Ppi z nz Description Temp. 'Press.. Size Sch.' D11t .- 5vMaterial.. :, .SupportTypes 4. :.:Design CF) (psig). (NPS). ;. , :S - - -i,:
Cfl8 HPCI Drain to MS Drain; 450' 400 2 160 7 ASTM A-106 Rigid supports USAS u5!
RCIC Drain to HPCI Drain; 1 160 5 Grade B B31.1- 1967 *4
'C Aux. Boiler Drains to HPCllRCIC/ 4e-. > n Reactor Building Drain Line 270 415 1 160 5 0-.C
-4 o r, D-&
Misc. PT Instrument Lines 562 1146 1 160 5 ASTM A-106 Rigid supports USAS mmC Sample Lines to Sample Station Grade B B31.1- 1967 (n CO
.A4. .049' - ASTM Ak213 Rigid supports _
0 I-D
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tubing (wail t) SS Gr. TP-304 (tube clamps) 0 I *.
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CD-NOOOI-990113 Page C I4 Attachment C SHEET NO. 11 JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE BY _ 2?2c; DATE t CALC, NO. C-001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK ri U DATE /
PIPING AND SUPPORTS Table 2 Seismic Experience Database Piping Data Pipe Size Pipe . Wail Facility (NPS) O.D. Schedule Thickness DAt (inch) (inch) 24 24.0 20 0.375 64 20 20.0 20 0.375 53 18 18.0 30 . 0.437 41 16 16.0 30 0.375 43 14 14.0 30 0.375 37 12 12.75 40 0.406 31 12 12.75 30 0.330 39
.10 10.75 160 1.125 10 8 8.625 160 0.906 10 Valley Steam Plant 6 6.625 40 0.280 24 Units 1 & 2 4 4.50 160 0.531 8 4 4.50 40 0.237 19 3 3.50 160 0.437 8 3 3.50 80 0.300 12 3 3.50 40 0.216 16 2 2.375 160 0.343 7 2 2.375 40 0.154 15 1/1 1.90 160 0.281 7 1Yh1.90 40 0.145 13 1 1.315 40 0.133 10 3/4 1.05 160 0.218 5 3/K 1.05 40 0.113 9
.. - . -- w --
CD-NOOO1-990113 Page C aL Attachment _
SHEET NO. Zo JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE BY ?3~ DATE g X CALC. NO. C-00t SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK DATE .&p Ai11I PIPING AND SUPPORTS Table 2 (cont.)
Seismic Experience Database Piping Data Pipe Size Pipe Wall Facility (NPS) O.D. Schedule Thickness D/t (inch) (inch) 20 20.0 STD 0.375 53 18 18.0 160 1.781 10 18 18.0 XS 0.500 36 18 18.0 STD 0.375 48 14 14.0 40 0.437 32 14 14.0 STD 0.375 37 12 12.75 160 1.312 10 12 12.75 STD 0.375 34 10 10.75 40 0.365 29 8 8.625 160 0.906 10 8 8.625 120 0.718 12 8 8.625 40 0.322 27 6 6.625 120 0.562 12
_____6 6.625 40 0.280 24 El Centro 4 4.50 80 0.337 13 Steam Plant 4 4.50 40 0.237 19 3 3.50 160 0.437 8 3 3.50 80 0.300 12 3 3.50 .40 0.216 16 2 2.375 160 0.343 7 2 2.375 80 0.218 11 2 2.375 40 0.154 15 11/2' 1.90 160 0.281 7 13i 1.90 80 0.200 10 1 1/22 1.90 40 0.145 13 1 1.315 80 0.179 7 1 1.315 40 0.133 10 3h 1.05 80 0.154 7 h4 1.05 40 0.113 9
. . = o A
- j _ _
CD-NOOO1-990113 Page Cal Attachment C.
SHEET NO. 21 JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE BY DATE It n
.Z CALC. NO. C-o001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK c, DATE g0o§q PIPING AND SUPPORTS A.717OM.NAt Table 2 (cont.)
Seismic Experience Database Piping Data Pipe Size Pipe Wall Facility (NPS) O.D. Schedule Thickness . D/t
- (inch) (inch) 16 16.0 1.394 11 12 12.75 - 1.148 11 8 8.625 160 . 0.906 10 8 8.625 30 0.277 31 6 6.625 160 0.562 12 6 6.625 40 0.280 24 4 4.50 160 0.531 8 4 4.50 80 0.337 13 Moss Landing 4 4.50 40 0.237 19 Units 1, 2 & 3 3 3.50 160 0.437 8 3 3.50 80 0.300 12 3 3.50 40 0.216 16 2 2.375 160 0.343 7 2 2.375 80 0.218 11 2 2.375 40 0.154 15 1% 1.90 160 0.281 7 1% 1.90 80 0.200 10 1 1.315 160 0.250 5
. 1 1.315 80 0.179 7 3/4__ 1.05 160 0.218 5
. 3/4 1.05 80 0.154 7 JAbfnpmsiv\calc91801 .doc
CD-NOOO1-990113 Page Attachment C REA
.Earn_ JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE BY air SHEET NO. Z L DATE -
CALC. NO. C-001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK AZs IDATE _Z 7jf lox ammwmmn
.rZXNAMONAL PIPING AND SUPPORTS Table 2 (cont.)
Seismic Experience Database Piping Data
. . Pipe Size Pipe Wall Facility (NPS) O.D. Schedule Thickness Dt (inch) (inch) 24 24.0 40 0.687 35 24 24.0 -- 1.066 23
_ 0 18.8 - 2.287 8 16 16.0 40 0.500 32 16 16.0 - 0.902 18
- 13.2 - 1.668 8 8 8.625 160 0.906 10 8 8.625 40 0.322 27 6 6.625 160 0.562 12 6 6.625 40 0.280 24 4 4.5Q 160 0.531 8 4 4.50 80 0.337 13 4 4.50 40 0.237 19 Moss Landing 3 3.50 160 0.437 8 Units 4 & 5 3 3.50 80 0.300 12 3 3.50 40 0.216 16 2 2.375 160 0.343 7 2 2.375 80 0.218 11 2 2.375 40 0.154 15 1 '21.90 160 0.281 7
.1% 1.90 80 0.200 10 1'1 1.90 40 0.145 13 1 1.315 160 0.250 5 1 1.315 80 0.179 7 1 1.315 40 0.133 10 3/4 1.05 160 0.218 5 4 1.05 80 0.154 7 3/4 1.05 40 0.113 9 T.%hfnrnmstv\cajc9j801 .doc
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CD-NOOO1-990113 Page C23 Attachment a SHEET NO. 23 JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE BY air DATE T E9_-S
?UP4A11aTd CALC. NO. C-001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK jTC-P DATE $ffiD1 PIPING AND SUPPORTS Table 2 (cont.)
Seismic Experience Database Piping Data Pipe Size Pipe Wall Facility (NPS) O.D. Schedule Thickness D/t (inch) (inch) 30 30.0 0.632 47 26 26.0 1.128 23 18 18.0 - 3.444 5 12 12.75 2.444 5 12 12.75 - 0.601 21 8 8.625 -- 1.650 5 8 8.625 40 0.322 27 6 6.625 -- 1.268 5 6 6.625 40 0.280 24 4 4.50 -- 0.861 5 4 4.50 80 0.337 13
... ___4 4.50 40 0.237 19 3 3.50 80 0.300 12 3 3.50 40 0.216 16 2/2 2.875 - 0.550 5 Moss Landing 2V2 2.875 80 0.276 10 UnitsL2%
6 &7 2.875 40 . 0.178 16 2 2.375 - 0.519 5 2 2.375 '80 0.218 11 2 2.375 40 0.154 15 1____ 1.90 -- 0.428 4 1Y2 1.90 80 0.200 10 112 1.90 40 0.145 13 1 1.315 -- 0.301 4 1 1.315 80 0.179 7 1 1.315 40 0.133 10 3 1.05 160 0.218 5 34 1.05 80 0.154 7 33/44 1.05 40 0.113 9 1/2 1.05 -- 0.210 4
___0.54 . 0.153 4 JI:%fnamsiv\calc91801.doc
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. ;; . z;77 . - . - - . .. . . .;.. -1. - . .. . .. . . ..
CD-NOOO1-990113 Page C 2-/
Attachment C SHEET NO. 24 JOB NO. JOB EFN MSIV TECH SPEC CHANGE BY ;?13 -' DATE V - T?.q p7=-
200918 CALC. NO. C.001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK <\D(J DATE olr.
PIPING AND SUPPORTS Table 2 (cont.)
Seismic Experience Database Piping Data Pipe Size .. Pipe. : ..Wall Facility . ;- (NPS) O.D. Schedule Thickhess DAt
_ > ;. : -(in c ) " .. - (inch)
Ormond Beach 30 30.0 . 1.298 23 Units 1 & 2 30 30.0 0.719 42 21 21.0 3.793 6 Humboldt Bay 12 12.75 80 0.687 19 Unit 3 10 10.75 80 0.593 18 6 6.625 80 0.432 15 J.\bfnpmsiv\calc9I801 .doc
1-1 "- - , - .-
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CD-NOOO1-990113 Page CZ*
Attachment C SHEET NO. __f JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE BY _ DATE g-Lz-.q CALC. NO. C-001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK LIJ DATE g/Dj PIPING AND SUPPORTS MAMCNAL Table 3 Comparison of Browns Ferry and Selected Database Piping Parameters Piping Parameter Browns Ferry Database Sites Pipe Diameter 1.315 - 24.0 1.05 - 30.0 (inch)
Wall Thickness 0.25 - 1.218 .. 0.113 - 3.793 (inch)
Diameter-to-Thickness Ratio 5 - 20 4 - 64 (D/t)
I
- .\;t__>:A141 fineA
CD-NOOO1-9901 13 Page a Z6 Attachment ' C SHEET NO. Z JOB NO. 20317 E JOB BFN MSIV TECH SPEC CHANGE BY I DATE ,X
- 2, qq CALC. NO. C-O01 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK < o DATE ________
PIPING AND SUPPORTS tWRNAMONAL Table 4 Bounding Evaluations of Typical Support Configurations N L~. .'.Sup~pqrte . - S C.rtial.Coripnnglir s Rt ::
Cantilever bracket Anchor bolts .73 Rod hanger Overhead weld .70 attachment
.. . . . _ . An. .
- -* ' . -x
- .' , v v .
CD-NO001-990113 Page -C P Attachment C SHEET 1C. 7 JO3 NO. 2DZ: JOB SFN MSIV TECH SPEC CHANGE EY=V 73 DATE Jr-CALC. NO. C-001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK 4- DATE . /2199 PIPING AND SUPPORTS l2KAZONAL Figure 1 Comparison of Database Site Spectra to Browns Ferry DBE Ground Spectra
- 1.6 1.4 -
1.2-1-
.2 c 0.8 a
CC0.6.
0.4 0.2 0
I, -1 1 Frequency, Hz. 10 100
,- . .if. oef A.
CD-NOOO1-990113 Page ..c Z Attachment C SHEET NO. 4 JOB NO. 200916 2 JOB BFN MSIV TECH SPEC CHANGE BY ;2C,I..* DATE r.2.3-qq CALC. NO. C-001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK I f DATE PIPING AND SUPPORTS UINATIONAL Figure 2 Comparison of Browns Ferry and Selected Database Piping D/t Ratios 65 - -64 BErowns Ferry4 .
25 -
Eli Database 24 a20 20 -
19 15 16 15 15 - 13 Dit 13 13 10 -
7 7 55 5
4 4 5 5 5 I I 1 0 I I Y2 2 3 4 6 18 and Above Pipe Size (NPS)
TA1,f'n-e,%A,.1&01Rni rin,
I - - - - . . ..
CD-NOOO1-990113 Page 2C Z
Attachment C SHEETNO.
JOB NO. 200918 JOB BFN MSIV TECH SPEC CHANGE , BY .2 DATE jr- zt s CALC. NO. C-001 SUBJECT SEISMIC VERIFICATION OF THE MSIV ALT CHK DATE DA E e/3o/
PIPING AND SUPPORTS
£IRAMONIAL
5.0 CONCLUSION
S Based on the results of the seismic verification walkdowns and bounding support evaluations, and upon the resolution of the identified walkdown outliers, it is reasonable to assume that the ALT piping, related supports and components have adequate seismic capacity in the event of a Design Basis Earthquake (DBE) at Browns Ferry.
J:\bfnpnsiv\calc91801 .doc
.1 . . .. .. I .
CAILCULATION SHEET Document: CDNO 001 99 0113 Rev.002 Plant BFN Unit(s): 0
Subject:
SEISMIC EVALUATION REPORT Prepared : J.O. Dizon Date: 6102104 Checked: S.J. Eder Date: 6/02/04 ATTACHMENT D MSIV SEISMIC EVALUATION REPORT APPLICATION TO BFN UNIT 1 MSIV SEISMIC RUGGEDNESS VERIFICATION PROGRAM Page D-1 of 3
CALCULATION SHEET D.1 PURPOSE The purpose of this Attachment is to map the applicable sections of Attachments A, B and C of this calculation for use in the BFN Unit 1 Increased MSIV Leakage Tech Spec Change Submittal.
D.2 APPLICATION TO BFN UNIT 1 The following table presents a map purpose 1Yes N/A 2.1 Yes N/A 2.2 Yes --- NIA 3.1 Yes N/A 3.2 No 2.1 ...
A 3.3 No 4.1 to4.16 ...
4.1 No 5.1 to 5.2 ...
4.2 Yes - -- N/A 4.3 Yes - -- N/A 4.4 Yes
- Ref. D-2 of Attachment D 5 Yes ^ . References inAttachment D 1 Yes --- N/A 2 Yes --- N/A B 3 Yes * ... References in Attachment D 4 Yes ' . Ref. D-2 of Attachment D 5 Yes ... N/A 1 Yes . N/A 2 Yes . N/A C 3 Yes
- Ref. D-1 of Attachment D 4 Yes ' .. Ref. D-1 of Attachment D 5 Yes . . N/A
- Supplemented by the additional reference(s) listed.
Page D-2 of 3
CALCULATION SHEET D.3 REFERENCES D-1 Facility Risk Consultants, Inc., Report No. TVANBFN-01-R-003, "MSIV Seismic Ruggedness Verification at Browns Ferry Nuclear Plant Unit 1", Revision 0, May 13,2004. RIMS No. W87040520002.
D-2 TVA Calculation No. CDN1-000-2004-0041, uSeismic Verification of Condenser and its Anchorage, MSIV Ruggedness Seismic Analysis -
Resolution of POS 15-1", Revision 000.
Page D-3 of 3