3F0599-02, Rev 0 to LAR 249 to License DPR-72,proposing Alternate Repair Criteria for Axial Tube End crack-like Indications in Upper & Lower Tubesheets of Otsgs.B&Wog Proprietary Topical Rept BAW-2346P,Rev 0 Encl.Rept Withheld,Per 10CFR2.790
| ML20206H631 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/05/1999 |
| From: | Holden J FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20137N284 | List: |
| References | |
| 3F0599-02, 3F599-2, NUDOCS 9905110265 | |
| Download: ML20206H631 (28) | |
Text
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' Finrida Power S FJaa^A." anYgeENo.OPRM May 5,1999 3F0599-02 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
License Amendment Request #249, Revision 0 Once Through Steam Generator Tube Surveillance Program, Alternate Repair Criteria for Axial Tube End Crack Indications
Dear Sir:
Pursuant to 10CFR50.90, Florida Power Corporation (FPC) hereby submits a request for an amendment to the Crystal River Unit 3 (CR-3) Operating License No. DPR-72. The attached License Amendment Request (LAR) #249, Revision 0, proposes an alternate repair criteria (ARC) for axial tube end crack-like indications in the upper and lower tubesheets of the CR-3 Once Through Steam Generators (OTSGs). The ARC will allow leaving OTSG tubes with axially c-ien:ed tube end cracks located within the clad region of the tube-to-tubesheet roll joint in-service.
The technical basis for the alternate repair criteria is contained in Babcock & Wilcox Owners Group (B&WOG) proprietary Topical Report BAW-2346P, Revision 0 (Attachment F). An Affidavit from the B&WOG is also included in Attachment F. The Affidavit sets on which the information may be withheld from public disclosure by the NRC and addresses the considerations listed in 10CFR2.790(b)(4). Accordingly, FPC requests that the report contained in Attachment F be withheld from public disclosure. A non-proprietary version of Topical Report BAW-2346P will be submitted under separate cover.
Topical Report BAW-2346P includes leak rate values based on the worst case bounding main steam break (MSLB) tube loads for Babcock & Wilcox plants. Framatome Technologies Incorporated (FTI) is re-calculating the MSLB tube loads for CR-3. The re-calculated MSLB tube loads will be used to determine the leak rates for tubes with axial tube end cracks during a MSLB accident. The specific leak rates for CR-3 based on the MSLB tube loads will be included in an addendum to BAW-2346P. FPC will submit the addendum to the NRC by June 2,1999. ,5 9905110265 990505
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U.S. Nuclear Regulatory Commission
~ 3F0599-02 Page 2 of 3 Leaving axially oriented tube end cracks in-service as proposed in this LAR has no impact on safety. Implementation of the ARC will allow sound tubes to remain in service, reduce personnel radiation exposure, and reduce the outage time due to OTSG repairs. FPC intends to implement the proposed LAR during Refueling Outage 11R planned for October 1,1999. FPC is respectfully requesting NRC approval of this LAR by September 15, 1999, if you have any questions regarding this submittal, please contact Mr. Sid Powell, Manager, Nuclear Licensing at (352) 563-4883.
Sincerely, U
Jo m J. Holden Vice President and Site Director JJH/lve xc: Regional Administrator, Region II Senior Resident Inspector NRR Project Manager Attachments:
A. License Amendment Request, No Significant Hazards Consideration Evaluation, and Environmental Impact Evaluation B. Table 1, List ofIn-Service Tubes With Axially Oriented TEC C. Proposed Technical Specification Change Pages, Strikeout / Shaded D. Proposed Technical Specification Change Pages, Revision Bars E. List of Regulatory Commitments F.
B&WOG Proprietary Version, BAW-2346P, Revision 0, " Alternate Repait Criteria for Tube End Cracking in the Tube-To-Tubesheet Roll Joint of Once Through Steam Generators" I
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U.S. Nucicar Regulatory Commission
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3F0599-02 Page 3 of 3 STATE OF FLORIDA COUNTY OF CITRUS l
John J. Holden states that he is the Vice President and Site Director for Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear l Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
I Joh J. Holden Vice President and Site Director Sworn to and subscribed before me this I day of /MA'1 ,1999,by
' V John J. Holden.
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Signature of Notary Public State of Florida _ _ - - - - - -
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Personally Produced Known -OR- Identification
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l FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 '
DOCKET NO. 50-302/ LICENSE NO. DPR-72 ATTACHMENT A LICENSE AMENDMENT REQUEST #249, REVISION 0 ONCE THROUGH STEAM GENERATOR TUBE SURVEILLANCE l PROGRAM, ALTERNATE REPAIR CRITERIA FOR AXIAL TUBE END CRACK INDICATIONS License Amendment Request, No Significant Hazards Consideration Evaluation, and Environmental Impact Evaluation
U.S. Nuclear Regulatory Commission Attachment A
. 3F0599-02 Page 1 of 17 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 LICENSE AMENDMENT REQUEST #249, REVISION 0 ONCE THROUGH STEAM GENERATOR TUBE SURVEILLANCE PROGRAM, ALTERNATE REPAIR CRITERIA FOR AXIAL q TUBE END CRACK INDICATIONS J
LICENSE DOCUMENT INVOLVED: Improved Technical Specifications (ITS) 4 PORTIONS: ITS 5.6.2.10, " Steam Generator (OTSG) Tube Surveillance Program" ITS 5.7.2, "Special Reports"
SUMMARY
OF CHANGES TO THE ITS: 1 Florida Power Corporation (FPC) is proposing to change the ITS for Crystal River Unit 3 (CR-3) by adding an alternate repair criteria (ARC) for indications identified as axially oriented tube end cracks (TEC). Application of the ARC will result in leaving tubes in-service with TEC. The proposed changes to the ITS include a definition of TEC, criteria for leaving the tubes with TEC in-service, growth monitoring, inspection, and reporting requirements. ,
i BACKGROUND- 1 In the summer of 1997, FPC performed extensive eddy current testing (ECT) of both Once Through Steam Generators (OTSG) at CR-3. The scope of examination included inspecting 100% of the upper tubesheet (hot leg) roll transitions and roll expansions using a rotating coil eddy current inspection probe. The rotating coil probe included both motorized rotating pancake (MRPC) and Plus Point coils to achieve maximum inspection sensitivity.
Inspections at other plants with Babcock & Wilcox (B&W) designed steam generators revealed crack-like indications near the ends of the expansion roll of the tubes' upper tube-to-tubesheet joints. These crack-like indications were originally defined as tube end anomalies (TEA) and were dispositioned to be outside of the pressure boundary portion of the tube. The tube pressure boundary was considered to be the area of roll contact with the tubesheet up to and including the carbon steel-to-Inconel clad interface for the upper and lower tube sheets. !
In June 1998, FPC requested an exigent License Amendment Request (LAR)#228 to leave the tubes with TEAS in-service until an outage of sufficient duration to disposition the indications.
Based on LAR #228, the NRC issued License Amendment No.169 in July 1998. The NRC approval allowed tubes with TEA to remain in-service until the next outage of sufficient duration.
U.S. Nuclear Regulatory Commission Attachment .A
. 3F0599-02 Page 2 of 17 On February 8,1999, FPC submitted Revision 1 to Special Report 97-05, "Once Through Steam Generator (OTSG) Notifications Required Prior to MODE 4, and Completed Results of OTSG Tube Inservice Inspection Performed During the Current Outage (90 Day Report)."
Special Report 97-05, Revision 1, included the results of the re-analysis of the OTSG tubes that were previously identified with TEAS within the pressure boundary.
A calculation performed for CR-3 in 1979 determined that the acceptable primary-to-secondary pressure boundary for the OTSGs is a 0.719 inches roll joint, measured from the end of the roll transition area and which may extend into the clad region. As indicated in LAR #228, an activity was planned to define the pressure boundary through analytical techniques. Testing and analysis for this ARC determined that the pressure boundary for CR-3 is from the primary face of the clad on the upper tubesheet to the primary face of the clad on the lower tubesheet (Figure 2).
l Tube End Anomalies (TEA) are now referred to as Tube End Cracks (TEC). Topical Report BAW-2346P, Revision 0, " Alternate Repair Criteria for Tube End Cracking in the Tube-to-Tubesheet Roll Joint of Once Through Steam Generators," (Attachment F), provides the technical justification for the ARC. The deterministic assessment contained in the Topical Report addresses operational leakage, accident leakage and structural (no burst) criteria for OTSG tube integrity for leaving tubes with TEC in-service with no impact on safety. Table 1 in Attachment B lists the tubes CR-3 currently has in-service with axially oriented TEC.
General Description of the Once Through Steam Generator CR-3 is a B&W designed pressurized water reactor with Model 177 FA OTSGs (Figure 1).
For this design, the primary coolant enters the steam generators at the top of the tubes (the hot leg) and exits the bottom of the tubes (the cold leg), where the primary coolant is directed back to the reactor coolant pumps and the reactor vessel. The functions of the OTSGs are to provide:
A pressure boundary between the reactor coolant and the secondary side fluid, confine the fission products and activation products within the reactor coolant system, provide heat transfer capability and a heat sink to remove the reactor coolant heat produced during power operations, and provide normal and auxiliary feedwater flow paths and heat transfer capability for normal and emergency cooldown.
The OTSG tubes are mill annealed Alloy 600 (Inconel) which have been sensitized as a result of the full vessel post-fabrication heat treatment. The original tube-to-tubesheet joint consists of a roll expansion of one to two inches in length with a seal weld (fillet) between the tube and the primary side tubesheet cladding. The tubesheets are 24 inche-s thick carbon steel with a minimum primary side Inconel clad of approximately 5/16 inches, based on design drawings (the clad thickness will be measured ultrasonically during Refueling Outage 11R). Each OTSG has l
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U.S. Nuclear Regulatory Commission Attachment A 3F0599-02 Page 3 of 17 15,531 tubes with a nominal outer diameter of 0.625 inches and a nominal wall thickness of 0.034 inches. I ITS and FSAR Information Improved Technical Specifications (ITS) Section 3.4.12 contains RCS operational leakage limits and surveillance requirements. Normal operational primary-to-secondary leakage is limited to 150 gallons per day, per License Amendment No.158, dated October 28,1997. This leakage limit was established such that tubes initially leaking during normal operation do not contribute excessively to total leakage during accident conditions. Operational leakage is monitored daily by plant procedures.
Final Safety Analysis Report (FSAR) Section 4.3.4 provides an overview of the design basis for the OTSGs. Related FSAR accident scenarios are described in Sections 14.2.2.1, " Steam Line Failure Accident," and 14.2.2.2, " Steam Generator Tube Rupture Accident." The Steam Line Failure Accident is more limiting for site radiation releases because of the potential for primary-to-secondary leakage release directly to the environment. The safety analysis for this ;
accident assumes one gallon per minute (gpm) leakage through a single OTSG.
Application of ARC l The number of tubes with TEC left in-service by the proposed ARC will be determined using the method described in Topical Report BAW-2346P. The operational leakage limit is 150 ,
gallons per day, as established in the ITS. Therefore, the calculation for projected leakage !
from axial TEC left in-service will be based on the accident leakage limit. The projected accident leakage due to TEC left in-service shall not exceed one gallon per minute, minus the I operational leakage limit of 150 gallons per day.
ITS 5.6.2.10, " Steam Generator (OTSG) Tube Surveillance Program" Description of Specification Change Add ITS 5.6.2.10.2.f to include the ARC for axially oriented TEC in the upper and lower tubesheets. The changes to Pages 5.0-14 and 5.014A will read as follows.
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- f. Tubes in-service with axially oriented tube end cracks (TEC) are identified in l the OTSG Inservice Inspection Surveillance Procedure. The portion of the !
tube with the axial TEC must be inspected using the motorized rotating coil eddy current technique during each subsequent inspection. Tubes identified with axial TEC during inservice inspections will be added to the existing list of tubes in the OTSG Inservice Inspection Surveillance Procedure. Tubes with axial TEC will not be included when calculating the inspection category of the OTSG. The inspection data for tubes with axially oriented TEC indications j shall be compared to the previous inspection data to monitor the indications for growth.
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U.S. Nuclear Regulatory Commission Attachment A
. 3F0599-02 Page 4 of 17 No credit is to be taken for this inspection for meeting the minimum sample size requirement for random sample inspection.
Tubes with axially oriented TEC may be left in-service using the method described in Topical Report BAW-2346P provided the combined projected leakage from all primary-to-secondary leakage, including axial TEC indications left in-service, does not exceed the Main Steam Line Break (MSLB) accident leakage limit of one gallon per minute, minus 150 gallons per day, per OTSG.
If the plant is required to shut down due to primary-to-secondary leakage and the cause is determined to be degradation of the TEC portion of the tubes, 100% of the tubes with TEC in that OTSG shall be examined in the location of the TEC. If more than 1% of the examined tubes are defective tubes, 100% of the tubes with TEC in the other OTSG shall be examined in the location of the TEC.
Tubes with crack-like indications within the carbon steel portion of the tubesheet shall be repaired or removed from service using the appropriate approved method. Tubes with circumferentially oriented TEC or volumetric indications within the Inconel clad region of the tubesheet shall be repaired or removed from service using the appropriate approved method. I J
Reason for Request
FPC identified 820 tubes with TEC during the 1997 OTSG inspection. Based on Topical Report BA'W 2346P, leaving the tubes in-service with TEC will not adversely affect the safe operation of the OTSGs. Addition of the ARC to the Steam Generator Program will allow tubes with TEC indications of axial orientation, located within the clad region of the tube-to-tubesheet rolled joint, including the portion of the tube protruding from the cladding, to remain in-service without repair. Leaving axially oriented tube end cracks in-service as proposed in this LAR has no impact on safety. Implementation of the ARC will allow sound tubes to j remain in service, reduce personnel radiation exposure, and reduce the outage time due to OTSG repairs. !
I Description of Specification Change I Add ITS 5.6.2.10.4.a.11 to include a description of Tube End Cracks (TEC). The change to ,
Page 5.0-17 will read as follows:
]
- 11. Tube End Cracks (TEC) are those crack-like eddy current Indications, circumferentially and/or axially oriented, that are within the Inconel clad I region of the primary face of the upper and lower tubesheets, but do not !
extend into the carbon steel-to-Inconel clad interface. Axial TEC indications l
U.S. Nuclear Regulatory Commission Attachment A
. 3F0599-02 Page 5 of 17 are those indications whose major dimension is within 45 degrees of the vertical axis of the tube.
Reason for Request
This change is to include a description of TEC to the ITS to clearly identify TEC and the location for applying the proposed ARC.
Description of Specification Change Add ITS 5.7.2.c.5 to include Special Reporting requirements for tubes left in-service using the proposed ARC. The change to Page 5.0-29 will read as follows:
- 5. Number of tubes and axially oriented TEC indications left in-service and the projected accident leakage using the ARC for TEC.
Reason for Request
This change will report the number of tubes and axially oriented TEC indications left in-service using the proposed ARC in the OTSG 90-day inspection report.
Evaluation of Requests Topical Report BAW-2346P provides the technical justification and qualification of the proposed ARC. The report and qualification focused on the primary-to-secondary side leakage through a 100% through-wall indication within the primary side clad region. Testing and analysis was performed by Framatome Technologies, Incorporated (FTI). The testirig and ,
analysis as presented in BAW-2346P used materials and conditions similar to those found in l the plant for steady state operation as well as the most limiting transient condition for primary-to-secondary leakage.
To demonstrate that operation with axial TEC in-service is not a safety concern, it was necessary to evaluate the primary-to-secondary leakage for normal power operation and transient conditions to determine the structural significance of the axially oriented TEC. The original tube roll joint is considered a leak-limiting joint.
The main steam line break accident (MSLB) is the most limiting transient where primary-to-secondary leakage could occur due to the loss of pressure on the secondary side of the OTSG during the transient. Therefore, the MSLB is the most limiting transient for the application of this ARC.
FTl developed the leakage values using the MSLB axial tube loads to produce the bounding radial tubesheet hole dilations for B&W plants. The mock-ups were tested utilizing these axial loads to determine the radially applied leakage values. The leakage values are found in B.AW-2346P, Table 7-5 and Table 7-6.
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U.S. Nuclear Regulatory Commission Attachment A
. 3F0599-02 Page 6 of 17 Leakage rates for TEC indications will be based on CR-3 specific axial tube loads generated during the MSLB. FPC will use these leakage rates when calculating the leakage limit for TEC indications when implementing the proposed ARC. The application of the leak rates is discussed later in this submittal.
1 Limitations This ARC will allow tubes with indications of axial orientation located within the clad region of the tube-to-tubesheet rolled joint, or in the portion of the tube protruding from the cladding, to remain in-service without repair. The limitations of this ARC are:
- This ARC does not apply to tubes with circumferential, mixed mode, or volumetric j indications.
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- This ARC applies only to tubes with axial indications adjacent to the tubesheet cladding or I in the tube end protruding from the cladding. Tubes with any portion of an axial indication that extends into to the carbon steel tubesheet must be repaired. Therefore, it is required to locate, by eddy current inspection, the clad-to-carbon steel interface (CCI).
- The combined primary-to-secondary leakage from all sources, including TEC indications left in-service, shall not exceed one gallon per minute minus operational leakage of 150 gallons per day per OTSG.
- The tubesheet cladding thickness, as determined by ultrasonic measurement, must be less i than 0.625 inches in order for this ARC to be applied.
- This ARC will not be applied to the tubes identified during a 1978 visual inspection as having Class IV damage to tube seal welds due to a loose part. For details, see FPC to i NRC letter,3F0578-01, dated May 2,1978.
Analysis and Testing FTl performed a finite element analysis of the general structural behavior to determine tube axial loading and tubesheet hole dilation parameters for input to the mock-up leakage testing.
Mock-ups were representative of the tube-to-tubesheet joints in the OTSGs, including the tubesheet, tube, original roll, and seal weld. The tube had a Electric Discharge Machined (EDM) notch to represent a TEC and was then rolled into the tubesheet. The tube was pressurized to represent primary side pressure. The secondary side was at atmospheric pressure to represent a MSLB. Loading was applied to the tubesheet mock-up to create tubesheet hole dilation to simulate tubesheet bowing. Leakage was measured based on a pressure drop in the tube. Details of the testing and analysis are described in BAW-2346P.
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U.S. Nuclear Regulatory Commission Attachment A l
. 3F0599-02 Page 7 of 17 f
1 Leakage ,
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The analysis and testing, as detailed in BAW-2346P, established a leak rate for each TEC l
based on the radial position of the tube in the tubesheet, axial tube loading, and tubesheet hole l dilation. For each TEC indication, a leak rate will be assigned and combined with other primary-to-secondary leakage.
In LAR #228, FPC stated that the combined accident leakage, including TEC, was calculated to be approximately 0.031 gallons per minute from the bounding OTSG. This calculation was based on the MSLB accident for the "B" OTSG. The MSLB accident analysis did not take into consideration tubesheet bowing or tubesheet hole dilation. FTI is currently re-calculating the axial tube loads for CR-3 for MSLB conditions. The specific leakage rates for CR-3 based on the MSLB tube loads will be included in an addendum to BAW-2346P. FPC will submit the addendum to the NRC by June 2,1999. FPC will use the CR-3 specific leakage rates when calculating the predicted leakage of TEC indications remaining in-service to ensure that the predicted accident leakage is less than one gallon per minute.
Application of Leak ~ Rates For each axially oriented TEC indication, a leak rate based on the radial location of the tube will be established and documented. A leak rate for operational and accident conditions will be established using the following typical protocol;
- Determine radial position of the tube.
- Apply the leakage value from BAW-2346P for each TEC.
- Calculate the total projected leakage from all TEC.
- Calculate the leakage based on the Probability of Detection (POD) for TEC not detected.
- Combine the total leakage from identified TEC and POD /TEC with all other known sources of primary-to-secondary leakage. j e If the sum exceeds the accident leakage limit, repair tubes with TEC as necessary until the i projected leakage is less than the leakage limits. The remaining leakage attributable to i TEC will be added to the end-of-cycle operational assessment.
Inspection Technique The Eddy Current (ECT) technique used to support this ARC is motorized rotating pancake coil (MRPC) using either a PlusPoint or pancake type coil. This technique, which is able to differentiate indications relative to the carbon steel and clad interface, will be relied upon for detection and to determine flaw orientation. Length and depth sizing of indications is not required for the purposes of this ARC. The conservative assumption for this ARC is that any detected axial TEC is 100% through-wall. Topical Report BAW-2346P provides details which show why MRPC is the best technique to be used for landmark assessment.
U.S. Nuclear Regulatory Commission Attachment A
. 3F0599-02 Page 8 of 17 MRPC was used for detection of the TEC in the upper tubesheet roll joint of the OTSG tubes during the 1997 CR-3 OTSG eddy current inspection. The tubes that were identified with TEC during the 1998 re-analysis will be examined during each subsequent inspection. Future inspections may identify additional tubes with TEC due to the probability of detection (POD) and ECT inspection uncertainties. These newly identified tubes will not be included in the population to calculate the inspection category of the OTSG (C-1, C-2, or C-3) during the inspection, but will be added to the list of tubes with TEC for subsequent inspections. The data shall be compared with the previous inspection data to monitor the indications for growth. No credit will be taken for tubes with TEC when calculating the 20% random selection for inspecting the OTSG.
Probability of Detection (POD)
The POD reflects the ability of the inspection method to detect all of the TEC that exist in the OTSG tubing. The adjusted population, minus detected indications for tubes that have been plugged or repaired, should constitute for purposes of the tube integrity analyses the total number of indications remaining in-service. Therefore, the number of indications found during the ECT inspection, for each radial zone must be increased by the inverse of the POD.
This resulting number of indications must be multiplied by the associated leak rate to obtain the total leak rate for the radial position. The combined total leakage from all primary-to-secondary sources, including TEC indications left in-service, shall not exceed the MSLB accident leakage limit.
Based on the ECT data, the TEC are believed to initiate on the inside surface of the tube.
They are typically short, axially oriented, and located in the rolled portion of the tube near the heat affected zone created by the tube-to-tubesheet weld. The rolling process and weld create residual stress that makes the material more susceptible to primary water stress corrosion cracking (PWSCC). For this reason it is believed that TEC are PWSCC initiated.
The PlusPoint technique has been qualified per Appendix H of the EPRI Steam Generator Examination Guidelines for detection and sizing of axial PWSCC. The POD for PlusPoint is 0.84 based on a 90% confidence for cracks greater than 50% through-wall. The 0.115 diameter pancake coil technique has also been qualified per Appendix H. The POD for 0.115 diameter pancake coil technique is 0.85 based on a 90% confidence for cracks greater than 50% through-wall. For this ARC, the frequency distribution of TEC indications (based on radial location) found during ECT inspection will be scaled upward by a factor of 1/ POD to account for non-detected indications that could potentially leak during accident conditions during the next cycle of operation.
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U.S. Nuclear Regulatory Commission Attachment A
. 3F0599-02 Page 9 of 17 Tube Pull and In-situ Pressure Testing Due to the location of TEC, data analysis from pulled tubes does not exist. TEC are within the seal weld and Inconel clad region of the pressure boundary. Extraction of this area would damage the TEC, rendering the sample useless for evaluation. Also in-situ pressure testing ;
would not be possible due to the positioning of the testing fixture. TEC are contained within .
the tubesheet, therefore, the tubesheet physically limits tube deformation and prevents tube burst or tube rupture.
Growth Monitoring From the perspective of this ARC, the length of the indication in the cladding is not of significance, as long as the indication does not extend past the clad-carbon steel interface (CCI) upon initial detection. Tubes that are identified with axially oriented TEC indications that extend into the carbon steel during the inspection must be repaired or removed from service.
All leak rate data for this ARC has been calculated assuming that the crack spans the full length of the cladding (0.625 inches maximum). Topical Report BAW-2346P concluded that the growth rate of the indications is insignificant and no adjustment in leak rate needs to be made to account for growth over an inspection cycle.
For this ARC, the growth of the indication is measured relative to the clad-to-carbon steel interface. Growth monitoring is based on the relative distance from the tip of the indication closest to the interface (designated as the Flaw Tip Location, FTL) to the CCI. The distance between the FTL and the CCI will be monitored for change between the subsequent inspections. The length of the indication in the cladding is not of significance, as long as the identified indication does not extend past the CCI during the current inspection. The leak rate data for this ARC has assumed that the crack may be as long as 0.625 inches. Therefore, growth of the indication is allowed as long as the propagation of the indication does not extend ,
past the maximum length of 0.625 inches from the primary face of the cladding. This would i allow the predicted growth of the TEC to extend beyond the CCI into the carbon steel. The testing and qualification of the ARC, as detailed in BAW-2346P, used a maximum indication length 0.625 inches, therefore, the testing bounds the leakage through an indication of a maximum of 0.625 inches.
Current data indicates that growth rate of the indications toward the CCI is insignificant, and l
no adjustment in leak rate needs to be made to account for growth over an inspection cycle, j However, plant specific average growth rates of the entire population of axial TEC remaining !
in-service will be evaluated as part of condition monitoring and operational assessments, and l adjusted as necessary.
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- 3F0599-02 Page 10 of 17 Performance Monitoring All indications remaining in-service as a result of this ARC will be inspected in each planned future inspection outage in order to ensure that the requirements of the ARC are continually satisfied. The inspection will be conducted in accordance with the CR-3 Steam Generator Eddy Current Inspection Guidelines, as supplemented by the requirements in this report. In addition, primary-to-secondary leakage monitoring during normal power operation serves to ensure that the tubes are not degrading at a rate significantly higher than assumed in this application. Therefore, it is concluded that sufficient monitoring measures are in place to ensure the continued satisfactory performance of any tubes with indications left in-service as a result of this ARC.
Failure Modes With regard to tube burst or rupture considerations, the tubes with TEC are contained within the tubesheet. Therefore, the tubesheet physically limits tube deformation and prevents tube burst or tube rupture. Axial cracks may diminish the axial load carrying capability of the tube, however, the tube will not slip because the tube-to-tubesheet weld has been conservatively analyzed to be strong enough to carry the loads.
The presence of the tube-to-tubesheet weld and the support of the tubesheet provides adequate structural integrity for tubes with TEC. Therefore, the most important technical issue for the subject indications becomes the evaluation of their postulated leakage.
Dose Assessment The CR-3 Final Safety Analysis Report (FSAR) MSLB analysis assumes one gpm leakage in one steam generator as an initial condition. The dose consequences resulting from the MSLB accident meet the acceptance criteria defined in 10CFR100 and bound the potential leakage calculated from leaving the TEC in-service in accordance with the conditions specified for application of the ARC.
The OTSGs are capable of performing their intended safety function during normal operation and j postulated accident conditions with the TEC in-service. Additionally, since the potential leakage from leaving the TEC in-service is bounded by the current CR-3 safety analyses, the proposed ARC will not be of potential detriment to the public health and safety.
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J U.S. Nuclear Regulatory Commission Attachment A
. 3F0599-02 Page 11 of 17 l
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Risk Assessment of Severe Accidents i The NRC has established five key principles that should be met in order to implement risk-informed decision making relative to license basis changes. These principles are listed .in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk- l Informed Decisions on Plant-Specific Changes to the Licensing Basis" (July 1998). These principles are as follows:
- 1. The proposed change meets the current regulations unless it is explicitly related to a requested exemption or rule change, i.e., a " specific exemption" under 10 CFR 50.12 or a " petition for rulemaking" under 10 CFR 2.802.
- 2. The proposed change is consistent with the defense-in-depth philosophy. ll
- 3. The proposed change maintains sufficient safety margins.
- 4. When proposed changes result in an increase in core damage frequency or I risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement.
- 5. The impact of the proposed change should be monitored using performance measurement strategies.
The implementation of the ARC relative to each of these principles is discussed in the following paragraphs.
- Satisfaction of Current Regulation The analyses and testing performed in support of this ARC have demonstrated that, if the ARC is applied consistent with the requirements set forth in Topical Report BAW- i 2346P, the deterministic structural integrity criteria of the plant's current licensing basis is satisfied. This includes appropriate margins for failure under normal operating conditions and postulated accidents. In addition, the impact on postulated leakage during a MSLB has been assessed and it has been shown that the leakage will be less than the limit established to satisfy 10 CFR 100 limits for off-site dose. Therefore, it is concluded that the proposed change to CR-3's licensing basis meets all current regulations set forth for implementation of the proposed ARC.
- Defense-in-Depth As discussed earlier, the structural integrity of the tubes is maintained because the tubes with TEC are contained within the tubesheet and the tubesheet physically limits tube deformation and prevents tube burst or tube rupture. Leakage that could be attributed to
U.S. Nuclear Regulatory Commission Attachment A
- 3F0599-02 Page 12 of 17 these indications during a MSLB has been conservatively estimated and shown to be less than the acceptance criteria when the ARC is applied in accordance with the requirements in BAW-2346P. Furthermore, the ARC has no effect on the remaining containment structures, or on any plant process or procedure that would increase the likelihood or consequences of any accident. Therefore, it is concluded that the defense-in-depth design attributes are satisfactorily maintained under application of this ARC.
- Safety Margins It has been shown that the portion of the tubes where this ARC is applied cannot burst due to the support provided by the tubesheet. In addition, analysis has shown that the structural integrity of the tube-to-tubesheet joint is maintained with safety margins.
Therefore, it is concluded that safety margins consistent with the design basis have been maintained.
. Effect of ARC on Core Damage Frequency or Consequences of a Severe Accident The ARC for tube end cracks will be applied only to axial indications that are located in the portion of tubing that is roll expanded into the tubesheet. Burst of these indications is not possible due to the constraint provided by the tubesheet. The thermal challenge conditions associated with a severe accident cannot increase the probability of tube burst at this location. Therefore, application of this ARC will not increase the probability of tube rupture during a postulated severe accident.
- Performance Monitoring All indications remaining in-service as a result of this ARC will be inspected in each planned future inspection outage in order to ensure that the requirements of the ARC are continually satisfied. The inspection will be conducted in accordance with the CR-3 Steam Generator Inspection Guidelines. In addition, primary-to-secondary leakage monitoring during normal operation serves to ensure that the tubes are not degrading at a rate significantly higher than that assumed in this application. Therefore, it is concluded that sufficient monitoring measures are in place to ensure the continued satisfactory performance of any tubes with indications left in-service as a result of this ARC.
Conclusion FPC has evaluated Topical Report BAW-2346P and concluded that the methodology described in BAW-2346P for leaving tubes in-service with TEC meets the accident leakage and structural integrity requirements with no significant increase in risk or a reduction in safe operation of the OTSGs.
i
U.S. Nuclear Regulatory Commission Attachment A 3F0599-02 Page 13 of 17 NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION An evaluation of this proposed LAR has been performed in accordance with 10CFR50.91(a)(1) regarding significant hazard considerations, using the standards in 10CFR50.92(c). A discussion of these standards as they relate to this LAR follows-(1) Involve a sigmficant increase in the probability or consequences of an accident previously evaluated.
This LAR proposes to implement an alternate repair criteria (ARC) for Once Through Steam Generator (OTSG) tubes with axial tube end crack (TEC) indications. Application of the ARC will allow tubes with axially oriented TEC to remain in-service in accordance with specific conditions. Based on a combination of structural analyses, mock-up testing and inservice inspections, as detailed in Topical Report BAW-2346P, allowing tubes with TEC indications to remain in-service is safe and justified.
Potential leakage from tubes with TEC will be bounded by the main steam line break (MSLB) evaluation presented in the Final Safety Analysis Report (FSAR). The proposed change requires inspections during subsequent outages of tubes remaining in-service with the TEC indications. The addition of this inspection does not change any accident initiators. The proposed inspection of these indications during the subsequent OTSG inservice inspections assures continuous monitoring of these tubes such that degradation of tubes containing TEC indications will be detected. Therefore, this change does not involve a significant increase in the probability or consequences of any accident previously evaluated. 1 (2) Create the possibility of a new or dWerent kind of accident from any accident previously evaluated.
l The proposed alternate repair criteria for axial TEC indications introduces no new failure j modes or accident scenarios. Topical Report BAW-2346P demonstrated structural and leakage integrity for all normal operating and accident conditions for Crystal River Unit 3 (CR-3). Furthermore, leaving TEC in-service does not change the design or operating characteristics of the OTSGs. In the unlikely event that a tube with a TEC should fail and sever completely, the tube would remain engaged in the tubesheet bore, preventing interaction with other surrounding tubes. In this case, leakage is bounded by the steam generator tube rupture (SGTR) accident analysis. Therefore, this change does not create a possibility of a new or different kind of accident from any previously evaluated.
(3) Involve a sigmficant reduction in a margin ofsafety.
The mechanical joint is constrained within the tubesheet bore; thus, there is no additional risk associated with tube rupture. ITS Bases 3.4.12 contains relevant information l
l l
l
U.S. Nuclear Regulatory Commission Attachment A 3F0599-02 Page 14 of 17 pertaining to limitations on Reactor Coolant System leakage. The accident leakage is shown to be less than one gallon per minute primary-to-secondary leakage. Therefore, the FSAR analyzed accident scenarios remain bounding, and the use of the proposed alternate repair criteria does not reduce the margin of safety.
ENVIRONMENTAL IMPACT EVALUATION 10CFR51.22(9) provides criteria for identification of licensing and regulatory actions eligible for categorical exclusion from performing an enviromnental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not 1
(i) involve a significant hazards consideration, (ii) result in a significant change in the types or significant increase in the amounts of any eftiuents that may be released offsite, and (iii) result in a significant increase in individual or cumulative occupational radiation j exposure.
l FPC has reviewed this proposed LAR and concludes it meets the eligibility criteria for categorical exclusion set forth in 10CFR51.22(9). Pursuant to 10CFR51.22, no environmental impact statement or environmental assessment needs to be prepared in connection with this request.
i
U.S. Nuclear Regulatory Commission Attachment A 3F0599-02 Page 15 of 17 Figure 1:
OTSG Longitudinal Section Primary inlet Nozzle inspection opening H E "** , , Upper Tubesheet Vent & Level Sensing Connection llf ll
, 3,; , ,
& Nozzles (7) M '
l Upper Cylindrical Bame i
h f, f Tube Support Plate
-...i 1+
N. Mdifr I i Steam Outlet Nozzle (2)
Handhole (2 to 8)
Se in ne on (2) m
-2/
Main Feedwater Risers lk d -
-Il h! l No le Lower High Level Sensing Connection (2) p,. f .f
. L h Extemal Main Foedwater Header Downcomer Temperature Sensing Connection (2)
Tubes -
-}
Shell 4
Lower Cylindrical Bame .
Upper Low Level Sensing Connection (2)
[ .
~ ~ " "" NRWIN,'HlM' Handholes (7)
Dmin (2)
_p - Sens C nnection Lower Hemispherical Head \ Lower Tubosha t Primary Outlet Nozzle (2) i Manway Support Skirt // p k Inspection Opening 96C1290 Primary Drain
U.S. Nuclear Regulatory Commission Attachment A 3F0599-02 Page 16 of 17 Figure 2:
Typical OTSG Tube to Tubesheet Joint 1
1 1
Ak Tube-to-Tubesheet Weld, 1.
~0.187" 0.051" minimum leg y Primary dL JL 4----
F. ace of the Clad 5/16" * * < Cladding v Clad / Carbon 8 Interface
. a". w:-.~ o. ;..,n mo '
-sse .-
fg aze:f [ W *-
.y: ..+ nn .; ; ,
3.qQft i, 1" min 4 Tubesheet V,;hl[@v,a<,. ,
- m. :
/: ,v . .: 6 r g 5 'J 2 .f. Jl
, 3. . , ./.e;W. ,, ?n;t_ . ,
xbiqq45CS u
St.= dagt<
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- Nominal Value. Actualclad thickness varies
1 U.S. Nuclear Regulatory Commission Attachment A 3F0599-02 Page 17 of 17 Figure 3:
Typical Tube End Crack in the Tube-to-Tubesheet Joint Asial Tube End Cracking Typical Seal Weld lleat Affected 'Z.one Seal Weld
/ 4 .......> _
[
Inconel Cladding
.. , . . . g.. . .
1 Clad Region /. ', >
- 4 %
.-. Carbon Steel Tube Sheet (24" v; '
Thick) 1 l
1 Tubesheet-to-Clad interface l
l i
1 a
l l
FLORIDA POWER CORPORATION l CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 1
ATTACHMENT B LICENSE AMENDMENT REQUEST #249, REVISION 0 ONCE THROUGH STEAM GENERATOR TUBE SURVEILLANCE PROGRAM, ALTERNATE REPAIR CRITERIA FOR AXIAL TUBE END CRACK INDICATIONS Table 1 !
l List of In-Service Tubes with Axially Oriented TEC (not including tubes with circumferential (TEC) l
U.S. Nuclear Regulatory Commission Attachment B 3F0599-02 Page 1 of 6 Table 1 1
List of In-Service Tubes with Axially Oriented TEC l (not including tubes with circumferential TEC) l l Row l Tube l S/G l l Row l Tube l S/G l l Row l Tube lS/G l 1 6 A 19 72 A 31 69 A 2 21 A 19 74 A 31 96 A 2 23 A 19 84 A 32 90 A l 2 25 A 20 85 A 32 102 A 2 22 A 20 73 A 34 9G A 2 24 A 20 77 A 35 90 A 3 1 A 20 84 A 36 1 A l 3 4 A 21 61 A 36 95 A 3 34 A 21 64 A 36 97 A 3 3 A 21 74 A 36 99 A 4 38 A 21 76 A 36 110 A 5 33 A 21 83 A 36 86 A 7 43 A 21 85 A 36 103 A 8 1 A 21 90 A 36 112 A_
8 2 A 21 77 A 36 100 A 8 56 A 21 84 A 37 94 A 9 61 A 22 63 A 37 99 A l 9 60 A 22 65 A 37 109 A 11 67 A 22 82 A 37 110 A 12 70 A 22 86 A 38 95 A 13 37 A 22 92 A 38 98 A 13 68 A 22 93 A 38 99 A 13 73 A 22 91 A 38 101 A 14 55 A 23 76 A 38 115 A 15 1 A 23 77 A 39 99 A 15 68 A 24 79 A 39 111 A 15 72 A 24 53 A 39 1 A 15 77 A 24 77 A 40 101 A 15 78 A 25 67 A 40 113 A 16 1 A 25 81 A 40 77 A 16 80 A 25 97 A 41 111 A 17 70 A 25 98 A 41 112 A 17 72 A 26 92 A 41 91 A 17 81 A 27 82 A 41 98 A 17 82 A 27 100 A 42 117 A !
18 59 A 27 97 A 42 111 A 18 60 A 28 80 A 42 115 A i 18 74 A 28 89 A 42 116 A l 18 76 A 28 91 A 43 90 A !
18 85 A 28 94 A 44 107 A 18 79 A 28 101 A 45 93 A 18 84 A 28 83 A 47 107 A l 19 73 A 29 83 A 47 121 A 19 76 A 29 58 A 48 13 A ,
1 l
4 3 U.S. Nuclear Regulatory Commission Attachment B 3F0599-02 Page 2 of 6 Table 1 List of In-Service Tubes with Axially Oriented TEC (not including tubes with circumferential TEC) l Row l Tube l S/G l l Row l Tube l S/G l l Row l Tube lS/G l 48 118 A 103 1 A 142 64 A 48 123 A 103 124 A 142 65 A 49 104 A 103 113 A 143 62 A 50 123 A 105 121 A 144 57 A 51 94 A 106 53 A 144 1 A 51 111 A 107 109 A 144 56 A 51 118 A 111 116 A 145 53 A 51 120 A 112 117 A 145 54 A 51 122 A 113 116 A 147 46 A 51 124 A 115 100 A 148 40 A 52 117 A 115 109 A 149 32 A 52 120 A 115 111 A 149 31 A 53 115 A 115 114 A 150 2 A 54 101 A 116 112 A 150 25 A 55 126 A 117 78 A 150 27 A 56 127 A 117 108 A 150 24 A 57 87 A 118 107 A 150 26 A 58 1 A 119 108 A 151 1 A 59 124 A 120 107 A 151 3 A 59 109 A 121 106 A 151 6 A 60 97 A 122 99 A 151 7 A 60 120 A 122 105 A 151 8 A 61 16 A 123 102 A 151 10 A 65 101 A 123 103 A 151 14 A 65 8 A 123 104 A 151 5 A 67 97 A 126 99 A 151 9 A 67 101 A 127 98 A 151 12 A 67 102 A 132 85 A 69 1 A 132 84 A 1 7 8 70 131 A 134 84 A 1 10 B 71 88 A 134 85 A 1 2 B 72 130 A 135 82 A 1 15 B 74 125 A 136 50 A 1 9 B 80 131 A 136 55 A 1 2 B 82 130 A 136 80 A 1 1 B 90 129 A 137 47 A 2 1 B 91 78 A 137 50 A 3 4 8 92 79 A 137 53 A 3 3 8 94 57 A 137 61 A 3 2 B 95 69 A 141 40 A 3 1 8 98 115 A 141 67 A 3 34 8 100 75 A 141 1 A 4 37 B 101 120 A 141 68 A 4 4 B 102 123 A 142 1 A 4 3 B
U.S. Nuclear Regulatory Commission Attachment B 3F0599-02 Page 3 of 6 Table 1 List of In-Service Tubes with Axially Oriented TEC (not including tubes with circumferential TEC) l Row l Tube l S/G l l Row l Tube l S/G l l Row l Tube lS/G l 5 18 B 24 24 8 38 14 8 5 7 B 25 58 8 38 115 B 1 7 34 8 25 97 8 39 27 B 7 3 B 26 94 8 39 116 B 7 54 B 26 4 8 40 117 8 8 57 8 26 99 8 41 33 8 9 57 8 27 44 B 42 18 B 9 60 B 27 95 B 42 111 B 9 1 B 27 98 B 42 117 8 9 13 8 27 100 B 43 83 8 9 2 B 28 95 B 43 1 B 11 68 B 28 30 B 43 12 B 12 2 B 28 101 B 44 13 8 12 1 B 29 19 8 44 12 8 13 56 B 29 94 8 44 7 B 15 69 B 29 103 8 44 77 8 17 22 B 29 104 8 44 1 B 17 82 B 30 49 B 45 29 B 18 1 B 30 12 B 45 11 B 18 85 B 30 97 8 45 85 B ,
19 47 B 30 105 B~ 45 8 B !
19 18 B 31 106 B 46 10 B l 19 1 B 32 44 8 46 75 B l 19 86 B 32 42 B 46 23 B I 20 84 B 32 9 B 47 78 B I 20 85 B 32 107 8 47 25 B 21 89 B 33 25 B 47 14 8 21 1 B 34 28 8 47 3 8 21 90 B 34 27 B 47 122 8 22 16 B 34 23 B 48 91 B 22 90 8 34 56 B 48 31 B 22 1 B 34 59 B 48 17 8 22 1 B 34 95 B 48 3 B 22 93 B 35 57 B 48 50 B 23 46 8 35 67 B 49 92 B 23 44 B 35 27 8 50 123 B 23 39 B 35 83 8 51 116 B 23 37 8 35 106 8 51 122 B 23 21 B 35 108 8 51 1 B 23 16 B 36 112 B 51 52 B 23 92 B 37 23 B 52 90 B 23 59 8 37 12 B 52 116 8 23 94 B 37 103 B 52 1 B 24 33 8 37 114 8 52 90 B
I l*. .
l U.S. Nuclear Regulatory Commission Attachment B l- 3F0599-02 Page 4 of 6 j Table 1 List of In-Service Tubes with Axially Oriented TEC l (not including tubes with circumferential TEC) l Row l Tube l S/G l l Row l Tube l S/G l l Row l Tube lS/G l 53 119 B 75 57 8 99 93 8 53 5 B 77 96 B 99 6 B 54 48 8 78 57 B 99 93 B 54 46 B 78 34 B 100 70 B 54 42 B 78 33 B 100 72 B 54 127 8 78 126 B 100 33 B 55 122 8 78 50 B 100 28 B 56 1 B 78 48 8 100 53 B 56 88 8 78 41 B 100 1 B 56 5 B 79 86 B 100 31 B 57 13 8 80 40 B 101 80 8 58 84 8 80 111 B 101 85 B 58 119 B 83 34 8 101 1 B 58 120 8 83 126 B 102 72 B 58 59 114 107 8
8 84 84 59 25 8
B 102 102 76 5
B B
]
{
59 124 8 84 9 0 _ 103 74 8 )
60 26 B 85 58 8 103 76 B 60 126 8 85 4 B 103 1 B 60 28 8 86 65 B 104 76 B 61 121 8 86 8 8 104 18 B 63 37 8 86 60 B 104 1 B 66 50 B 87 10 B 107 3 8 66 120 B 87 117 8 108 39 B 66 111 8 87 125 B 108 112 8 67 40 B 89 8 B 109 72 B 67 11 B 90 16 B 109 117 B .
67 9 8 90 118 B 110 63 8 68 37 8 90 1 B 110 35 B 68 42 8 92 14 B 110 34 8 68 41 8 92 88 8 110 110 8 69 52 B 93 82 B 110 1 B 69 50 8 93 7 B 111 62 B 69 108 8 93 124 B 111 116 B 69 87 8 94 69 B 112 60 B 69 105 B 94 53 B 112 1 B 69 111 8 94 87 B 112 13 B 70 49 8 94 16 B 112 1 B 70 23 8 94 7 B 112 117 B 71 51 B 95 53 8 113 1 B l
72 130 B 96 42 B 114 1 B 74 55 B 97 99 B 115 49 8 74 72 B 98 83 B 115 27 B 75 69 8 98 1 B 115 1 B
^
U.S. Nuclear Regulatory Commission Attachment B 3F0599-02 Page 5 of 6 l Table 1 1 List of In-Service Tubes with Axially Oriented TEC (not including tubes with circumferential TEC) l Row l Tube l S/G l l Row l Tube lS/Gl l Row l Tube lS/G l 115 55 B 127 10 B 132 26 B 116 30 B 127 5 B 132 71 B 117 88 B 127 98 B 132 13 B 117 89 B 128 24 B 132 2 B 117 108 8 128 23 B 132 84 8 118 98 8 128 19 B 132 1 B 118 107 B 128 78 B 132 23 B 119 88 B 128 8 8 133 44 B 119 108 8 128 81 B 133 22 B 120 65 B 129 27 B 133 66 B 120 65 B 129 26 B 133 16 B 120 30 B 129 23 8 133 14 8 121 34 8 129 76 B 133 74 8 121 31 8 129 79 B 133 76 8 121 26 B 129 10 B 133 8 B 121 25 __ B 129 4 B 133 1 B 121 106 B 130 28 B 133 15 B
]
122 66 8 130 25 B 133 72 B j 122 29 B 130 69 8 133 12 B l 122 25 B 130 22 B 134 40 B J' 122 12 8 130 18 B 134 27 B 123 28 B 130 77 B 134 16 8 123 22 B 130 78 B 134 73 8 1 123 18 8 130 78 8 134 74 8 123 17 8 130 12 8 134 72 B i 123 10 8 130 2 B 134 79 8 !
123 103 8 130 93 0 135 23 B )
123 54 8 130 93 B 135 70 8 j 123 1 B 130 78 B 135 12 8 124 33 B 130 93 B 135 76 B !
124 89 8 131 30 B 135 14 B l 124 29 B 131 25 B 136 20 B j 125 59 B 131 24 8 136 72 B 125 29 8 131 23 8 136 74 8 ;
125 28 B 131 21 B 136 14 B ,
125 81 B 131 16 B 137 39 B j 125 11 B 131 14 B 137 29 B 125 10 B 131 78 B 137 17 B 125 31 B 131 35 B 137 8 8 126 30 B, 131 68 B 137 39 B 126 83 8 131 74 8 137 62 B 126 39 B 131 18 8 137 14 8 l 126 10 B 132 30 B 137 9 B l 127 80 B 132 29 B 137 77 B l l
\
U.S. Nuclear Regulatory Commission Attachment B l 3F0599-02 Page 6 of 6 l Table 1 List of In-Service Tubes with Axially Oriented TEC (not including tubes with circumferential TEC) l j
l l Row l Tube l S/G l l Row l Tube l S/G l l Row l Tube lS/G l j 1
138 17 8 144 22 B 8 144 18 B
]8 "A"OTSG=247 tiibes 138 35 B 144 13 B "B"OTSG=481 tubes )
139 13 B 144 2 B 139 12 B 144 1 B 139 74 B 145 26 B 140 23 8 145 21 B 140 22 B 145 2 B l 140 16 B 145 1 B 140 15 B 146 1 B !
140 12 B 146 1 B 140 9 B 147 39 8 l
140 20 B 147 23 B 140 8
]
19 148 17 8 i 140 2 B 148 4 8 140 70 B 149 31 B 140 21 B 149 32 B 141 14 8 149 33 8 l 141 63 8 149 1 B l 141 5 8 151 4 B 141 66 B 151 13 B 141 2 8 151 3 B 141 67 8 151 5 B 141 18 8 151 2 8 141 17 8 142 2 B 142 27 8 142 26 B 142 17 8 142 64 8 142 1 B 143 21 B 143 11 B 143 61 B 143 1 B 143 12 B 143 3 8 143 60 B 143 2 B 143 62 B 144 26 B 144 12 B 144 23 B