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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML20212H5491999-06-18018 June 1999 Non-proprietary Rev 4 to HI-981983, Licensing Rept for Storage Capacity Expansion of Oyster Creek Spent Fuel Pool ML20237D5691998-08-31031 August 1998 Rev 0 to MPR-1957, Design Submittal for Oyster Creek Core Plate Wedge Modification ML20237D5711998-08-18018 August 1998 Rev 0 to SE-000222-002, Core Plate Wedge Installation ML20247F1891998-05-0505 May 1998 Risk Evaluation of Post-LOCA Containment Overpressure Request ML20151Y4651998-03-31031 March 1998 Non-proprietary Version of Rev 1 to GENE-E21-00143, ECCS Suction Strainer Hydraulic Sizing Rept ML20217G0481997-01-0202 January 1997 Rev 0 to Risk Assessment of Deferred Oyster Creek Projects ML20133A7901996-12-20020 December 1996 Volumes 1 & 2 to Plant Referenced Simulator Certification ML20134C8851996-10-0404 October 1996 Core Spray Sys Insp Program -16R ML20117E7331996-07-31031 July 1996 Pressure-Temp Curves Per RG 1.99,Rev 2 for Oncpp ML20248D1121996-07-31031 July 1996 Development of Equipment Seismic Fragilities for OCNGS Ipeee ML20117J0491996-05-0808 May 1996 At&T Round Cell Nuclear Util User'S Council Charter ML20117F1041996-04-11011 April 1996 Response to NRC GL 95-07 Pressure Locking & Thermal Binding of Safety Related Power Operated Gate Valves ML20101L8041996-03-0101 March 1996 USI A-46 Seismic Evaluation Rept ML20140B3041995-08-16016 August 1995 Rev 2 to Removal of Ice Condenser Vent Rod & Containment Spray/E5 Water Heat Exchanger Monitoring Sys ML20248D0491994-12-31031 December 1994 Assessment of Potential for Liquefaction & Permanent Ground Displacements at Designated Facilities,Oyster Creek Nuclear Generating Station ML20078E5781994-11-0909 November 1994 Rev 0 to Safety Evaluation SE-403037-001, Reactor Vessel Shroud Repair ML20248D0901994-10-31031 October 1994 Seismic Fragilities of Civil Structures at Oyster Creek Nuclear Generating Station ML20069M8311994-06-0202 June 1994 Considerations Associated W/Changing Electromatic Relief Valve (EMRV) Setpoints ML20058E2601993-09-20020 September 1993 Addl Stress Info for Drywell Shield Wall from Structural Evaluation of Spent Fuel Pool ML20059H9581993-06-30030 June 1993 Design Criteria for Soil-Structure Interaction Analysis of Reactor/Containment Bldg at Gpun Oyster Creek Nuclear Generating Station ML20012G6051993-02-26026 February 1993 Referenced Simulator Certification Suppl Rept. ML20125A9971992-11-30030 November 1992 Rev 2 to Drf 00664 Index 9-4, ASME Section Viii Evaluation of Oyster Creek Drywell for W/O Sand Case,Part 2,Stability Analysis ML20106A7631992-08-31031 August 1992 Vessel Fracture Mechanics Analysis for Upper Shelf Energy Requirement ML20128B7591992-06-29029 June 1992 Structural Evaluation of Spent Fuel Pool ML20096F2091992-05-13013 May 1992 Investigation Rept. Rept Provides Info for Second Phase of Investigation Into Alleged Falsification of Operator Log & Tour Repts at Plant ML20096C4101992-05-0505 May 1992 Investigative Rept ML20090A9171992-01-31031 January 1992 Cycle 13R Outage IGSCC Activities ML20085H4501991-08-31031 August 1991 Rev 2 to Licensing Basis for Oyster Creek Long-Term Solution to Reactor Instability ML20090A9511991-06-30030 June 1991 Full Structural Weld Overlay Design for Oyster Creek Shutdown Cooling & Core Spray Sys ML20081L0591991-06-18018 June 1991 Rev 1 to Oyster Creek Plant-Specific Oxygen Generation Following Loca ML20079K8441991-04-30030 April 1991 Vols 1 & 2 to Leak Before Break Evaluation of Isolation Condenser Sys Piping Outside Containment ML20072U4751991-04-13013 April 1991 Core Spray Sys Insp Program ML20087E7781991-02-28028 February 1991 Index 9-4, ASME Section Viii Evaluation of Oyster Creek Drywell for W/O Sand Case,Part 2,Stability Analysis ML20029B3861991-02-28028 February 1991 ASME Section Viii Evaluation of Oyster Creek for W/O Sand Case,Part I,Stress Analysis. ML20070M5691991-02-28028 February 1991 Effects of Internal Pressure on Axial Compression Strength of Cylinders ML20066D3981990-11-30030 November 1990 Draft Pressure-Temp Curves,Per Reg Guide 1.99,Rev 2 ML20087E7671990-11-30030 November 1990 Index 9-2, ASME Section Viii Evaluation of Oyster Creek Drywell,Part 2,Stability Analysis ML20029A3181990-11-30030 November 1990 Site Specific Response Spectra,Oyster Creek Nuclear Generating Station,Response to NRC Questions. ML20065M0981990-11-30030 November 1990 Ge/Teledyne Rept TR-7377-1, Justification for Use of Section III Subsection Ne Guidance in Evaluating Oyster Creek Drywell ML20065M1021990-11-30030 November 1990 Rev 0 to Index 9-1, ASME Section Viii Evaluation of Oyster Creek Drywall Part 1 Stress Analysis ML20065Q3891990-11-30030 November 1990 Rev 0 to Index 9-2, ASME Section Viii Evaluation of Oyster Creek Drywell,Part 2 Stability Analysis ML20062G8601990-11-0101 November 1990 Rev 1 to Design of Ultrasonic Test Insp Plan for Drywell Containment Using Statistical Inference Methods ML20056A0021990-07-26026 July 1990 Trust Agreement Jcp&L Qualified Trust ML19332F4831989-11-30030 November 1989 Offsite Dose Consequences Resulting from Pipe Break Between Drywell & Isolation Condensers at Oyster Creek Nuclear Generating Station. ML19324B8111989-10-30030 October 1989 Rev 0 to Oyster Creek Cycle 12R Outage IGSCC Activities. ML19325C8101989-09-30030 September 1989 Site Specific Response Spectra Oyster Creek Nuclear Generating Station. ML20058C6491989-07-31031 July 1989 Rept on Weld Overlay Procedure Development for Waterbacked Nozzle-to-Safe End Welds ML20062G8491989-02-0101 February 1989 Rev 1 to Statistical Analysis of Drywell Thickness Data ML20246N8491988-12-31031 December 1988 Rev 3 to Oyster Creek Nuclear Generating Station Mark I Containment Long Term Program,Addendum to MPR-734 Plant- Unique Analysis Rept Torus Attached Piping ML20207J1511988-08-10010 August 1988 Rev 0 to Response to Generic Ltr 88-01 & NUREG-0313,Rev 2 1999-06-18
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217K4451999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Oyster Creek Nuclear Generating Station.With ML20211P6731999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Oyster Creek Nuclear Generating Station.With ML20211A7051999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Oyster Creek Nuclear Station.With ML20209G0631999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Oyster Creek Nuclear Generating Station.With ML20212H5491999-06-18018 June 1999 Non-proprietary Rev 4 to HI-981983, Licensing Rept for Storage Capacity Expansion of Oyster Creek Spent Fuel Pool ML20195E7961999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Oyster Creek Nuclear Generating Station.With ML20206N7431999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Oyster Creek Nuclear Generating Station.With ML20205P5401999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Oyster Creek Nuclear Generating Station.With ML20204C8201999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Oyster Creek Nuclear Generating Station.With ML20199E4671998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Oyster Creek Nuclear Generating Station.With ML20195E8321998-12-31031 December 1998 10CFR50.59(b) Rept of Changes to Oyster Creek Sys & Procedures, for Period of June 1997 to Dec 1998.With ML20198D2091998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Oyster Creek Nuclear Generating Station.With ML20195J8591998-11-12012 November 1998 Rev 11 to 1000-PLN-7200.01, Gpu Nuclear Operational QA Plan ML20195C4271998-11-0606 November 1998 Safety Evaluation Supporting Proposed Ocnpp Mod to Install Core Support Plate Wedges to Structurally Replace Lateral Resistance Provided by Rim Hold Down Bolts for One Operating Cycle ML20155J3021998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Oyster Creek Nuclear Generating Station.With ML20154R4981998-10-20020 October 1998 Core Spray Sys Insp Program - 17R ML20154L3051998-10-14014 October 1998 Safety Evaluation Accepting Licensee Request to Defer Insp of 79 Welds from One Fuel Cycle at 17R Outage ML20154Q3371998-09-30030 September 1998 Rev 8 to Oyster Creek Cycle 17,COLR ML20154L5571998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Oyster Creek Nuclear Generating Station.With ML20151V6311998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Oyster Creek Nuclear Generating Station.With ML20237D5691998-08-31031 August 1998 Rev 0 to MPR-1957, Design Submittal for Oyster Creek Core Plate Wedge Modification ML20237D5711998-08-18018 August 1998 Rev 0 to SE-000222-002, Core Plate Wedge Installation ML20237B0131998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Oyster Creek Nuclear Generating Station ML20236R0511998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Oyster Creek Nuclear Generating Station ML20249B2981998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Oyster Creek Nuclear Station ML20248F3531998-05-21021 May 1998 Part 21 Rept Re Electronic Equipment Repaired or Reworked by Integrated Resources,Inc from Approx 930101-980501.Caused by 1 Capacitor in Each Unit Being Installed W/Reverse Polarity. Policy of Second Checking All Capacitors Is Being Adopted ML20247F1891998-05-0505 May 1998 Risk Evaluation of Post-LOCA Containment Overpressure Request ML20247G0581998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Oyster Creek Nuclear Generating Station ML20216K0341998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Oyster Creek Nuclear Generating Station ML20151Y4651998-03-31031 March 1998 Non-proprietary Version of Rev 1 to GENE-E21-00143, ECCS Suction Strainer Hydraulic Sizing Rept ML20217A4631998-03-23023 March 1998 Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively ML20212E2291998-03-0404 March 1998 Rev 11 to 1000-PLN-7200,01, Gpu Nuclear Operational QAP, Consisting of Revised Pages & Pages for Which Pagination Affected ML20216J0841998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Oyster Creek Nuclear Generating Station ML20203B2781998-02-16016 February 1998 10CFR50.59(b) Rept of Changes to Oyster Creek Systems & Procedures ML20203A3801998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Oyster Creek Nuclear Generation Station ML20198P1791997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Oyster Creek Nuclear Station ML20217C7591997-12-31031 December 1997 1997 Annual Environmental Operating Rept for Oyster Creek Nuclear Generating Station ML20197E9131997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Oyster Creek Nuclear Station ML20199E4561997-11-13013 November 1997 Safety Evaluation Accepting Ampacity Derating Analysis in Response to NRC RAI Re GL-92-08, Thermo-Lag 330-1 Fire Barriers, for Plant ML20199D4381997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for Oyster Creek Nuclear Station ML20202E8511997-10-21021 October 1997 Rev 0 to Scenario 47, Gpu Nuclear Oyster Creek Nuclear Generating Station Emergency Preparedness (Nrc/Fema Evaluated) 1997 Biennial Exercise. Pages 49 & 59 of Incoming Submittal Were Not Included ML20211M9481997-10-0303 October 1997 Supplemental Part 21 Rept Re Condition Effected Emergency Svc Water Pumps Supplied by Bw/Ip Intl Inc to Gpu Nuclear, Oyster Creek Nuclear Generation Station.No Other Nuclear Generating Stations Effected by Notification ML20198J7361997-09-30030 September 1997 Monthly Operating Rept for Sept 1997 for Oyster Creek Nuclear Generating Station ML20211B7461997-09-24024 September 1997 Part 21 Rept Re Failure of Emergency Service Water Pump Due to Threaded Flange Attaching Column to Top Series Case Failure.Caused by Dissimilar Metals.Pumps in High Ion Svc Will Be Upgraded to 316 Stainless Steel Matl ML20210V0181997-08-31031 August 1997 Monthly Operating Rept for Aug 1997 for Oyster Creek Nuclear Generating Station ML20210L2961997-07-31031 July 1997 Monthly Operating Rept for Jul 1997 for Oyster Creek Nuclear Station ML20149F9961997-07-18018 July 1997 Safety Evaluation Re Gpu Nuclear Operational Quality Assurance Plan,Rev 10 for Three Mile Island Nuclear Generating Station,Unit 1 & Oyster Creek Nuclear Generating Station ML20196H0111997-07-11011 July 1997 Special Rept 97-001:on 970620,removed High Range Radioactive Noble Gas Effluent Monitor (Stack Ragems) from Service to Allow Secondary Calibr IAW Master Surveillance Schedule. Completed Calibr on 970628 & Returned Stack Ragems to Svc ML20210L3081997-06-30030 June 1997 Corrected Page to MOR for June 1997 for Oyster Creek Nuclear Generating Station ML20141H2051997-06-30030 June 1997 Monthly Operating Rept for June 1997 for Oyster Creek Nuclear Station 1999-09-30
[Table view] |
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EVALUATION TO INCREASE NORMAL DRYWELL TEMPERATURE OPERATING LIMIT TOPICAL REPORT # 024 i
PROJECT NUMBER: 5400-40715 1 ,
P. S. SMITH July 18, 1985 APPROVALS:
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@ TR 024 Rev. O Page i TABLE OF CONTENTS PAGE 1.0 ABSTRACT 3 2.0' INTRODUCTION 4 3.0 METH000 LOGY 5 4.0 RESULTS 7
5.0 CONCLUSION
S 15
6.0 REFERENCES
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Id TR 024 Rev. O Page 2 LIST OF TABLES AND FIGURES PAGE TABLE 1 ANALYSIS ASSUMPTIONS 6 TABLE 2 PEAK VALUE
SUMMARY
14 FIGURE 1 DRYWELL TEMPERATURE PROFILE FOR DBA LOCA (0-1600 SEC) 9 FIGURE 2 DRYHELL TEMPERATURE PROFILE FOR DBA LOCA (0-160 SEC) 10 FIGURE 3 DRYWELL TEMPERATURE PROFILE FOR EEQ MSLB (0-80 SEC) 11 FIGURE 4 DRYHELL TEMPERATURE PROFILE FOR EEQ MSLB (0-1600 SEC) 12 FIGURE 5 DRYWELL TEMPERATURE PROFILE FOR REMOTE SHUTDOWN 13 PANEL ANALYSIS f
42 TR 024 Rev. O Page 3 1.0 ABSTRACT This report provides a technical basis for increasing the Emergency Operating Procedures entry condition based on bulk drywell temperature from 135'F to 150*F. The normal limit for power operation can be raised to 150*F.
A series of CONTEMPT analyses were performed to determine the sensitivity of containment response for three events starting from different initial drywell temperatures. The analyses covered the containment resprase to the design basis LOCA; MSLB for defining the EQ temperature profile inside the drywell, and the loss of drywell cooling case which defines the limiting drywell temperature condition for the Renote Shutdown Panel study. The temperature profiles for each of these cases do not change significantly when the initial drywell temperature is raised to 150*F.
In all cases, the peak temperatures and pressure are either unchanged or still within the acceptable ranges. Thus, it is concluded that it is acceptable to raise the E0P entry condition on bulk drywell temperature to 150*F and to permit normal plant operations up to this value of bulk drywell temperature.
p N TR 024 Rev. O Page 4
2.0 INTRODUCTION
This report is being prepared to provide technical justification for increasing the entry condition to the Containment Control section of the Emergency Operating Procedures (EOPs) based on drywell bulk temperature.
The intent of this report is to justify raising the entry condition from 135'F to 150*F.
Recently, the operator has begun computing drywell bulk temperature using the algorithm contained in Plant Administration Procedure 106 (Reference 1). The calculated bulk drywell temperature was usually near or a few degrees above the E0P entry condition value of 135'F. In order to prevent the operator from continuously being in the E0Ps, it was proposed to raise the E0P entry condition to 150*F.
l
Q 7 TR 024 Rev. O Page 5 3.0 METHODOLOGY The justification to raise the drywell temperature E0P entry condition to 150*F was based on the results from a series of analyses using the CONTEMPT code (Reference 2). The three cases that were analyzed are the
'following:
Design basis LOCA 0.75 ft* MSLB, containment spray initiated at 10 minutes (EEQ Inside Containment Temperature Profile Analysis)
Loss of drywell cooling, 50*F/hr cooldown begun at 10 minutes (Remote Shutdown Panel Performance Analysis) ,
The design basis LOCA was chosen since this break maximizes the energy input into the containment. The steam line break case defines the temperature profile for the environmental qualification of electrical equipment laside the drywell (Reference 3). The last case was used as the limiting case for drywell temperature as part of the remote shutdown panel study (Reference 4). The assumptions used for these cases are listed in Table 1.
47 h TR 024 Rev. O Page 6 TABLE 1 ANALYSIS ASSUMPTIONS DBA LOCA EE0 MSLB RSD PANEL CALC CASE 1A CASE IB CASE 2A CASE 2B CASE 3A CASE 3B Initial drywell pressure, psig 1.2 1.2 1.2 1.2 0.5 0.5 Initial torus pressure, psig 0.0 0.0 0.0 0.0 0.0 0.0 Initial drywell & torus humidity 0.01 0.01 0.01 0.01 0.01 0.01 Initial drywell temperature, F 135 150 135 150 135 150 Initial torus temperature, F 120 120 120 120 85 85
I
-Y TR 024 Rev. O Page 7 4.0 RESULTS The results from the six CONTEMPT analyses are summarized in Figures 1 through 5 and the peak values for drywell pressure and temperature are given in Table 2. Floure i shows the drywell atmosphere response for the DBA LOCA over 1600 second period. The long term temperature profile is identical irrespective of the initial drywell temperature. The early portion of the transient shows a slight deviation between the two cases.
This difference is shown in more detail on Figure 2. This figure shows that the post-peak atmospheric temperature profile for Case 1A (initial temperature - 135*F) is greater than that for Case 18. This is caused by condensation in the drywell atmosphere so that energy is transferred to the drywell liquid region, resulting in a slightly higher 11guld temperature profile for Case IB for the first 10 minutes of the transient. It can be seen from Figure 2 that the peak drywell temperature for the DBA LOCA is not dependent upon the initial drywell temperature. This behavior is expected since the difference in drywell atmosphere energy between initial temperatures of 135'F and 150*F is small relative to the amount of energy deposited into the drywell by the blowdown.
The profile of drywell atmosphere temperature for the steam line break case is shown in Figures 3 and 4. Figure 4 shows that the long term temperature profile does not depend on the initial drywell temperature.
Figure 3 depicts the short term temperature increase. The peak b
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I/ TR 024 Rev. O Page 8 temperature is roughly 5*F higher for the case starting from an initial temperature of 150*F when compared to the case starting from 135'F. This difference becomes smaller so that by the end of the first minute of the event, the temperature profiles are the same.
Figure 5 illustrates th'e temperature profiles for the loss of drywell coolers cases. Once the RPV cooldown has begun, the difference in the two profiles is nearly constant and its magnitude is the difference in the initial drywell temperatures for the two cases. The peak drywell temperature of 226*F is reached at roughly 6.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. This value is acceptable since it is well below the design temperature for the drywell liner of 281*F. The increased temperature profile does not cause any detectable total drywell pressure difference because of the reduced initial mass of non-condensibles for the 150*F case. Thus, there is no increased chance of isolating the containment or initiating ADS based on a high drywell pressure signal.
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TR 024 Rev. O Page 14 TABLE 2 PEAK VALUE
SUMMARY
CASE DRYWELL PRESSURE. PSIA DRYWELL ATMOSPHERE TEMPERATURE. 'F IA 55.4 9 6.6 seconds 287.6 9 6.6 seconds 1B 55.2 9 6.7 seconds 287.3 9 6.7 seconds 2A 39.9 9 18.5 seconds 345.4 9 18.5 seconds 2B 39.1 9 21 seconds 347.9 9 18 seconds 3A 16.6 9 22370 seconds 211.1 9 22040 seconds 38 16.5 9 19370 seconds 226.2 9 22620 seconds I
h TR 024 Rev. O Page 15
5.0 CONCLUSION
The results for the LOCA and MSLB cases do not show a noticeable increase in the drywell atmospheric temperature profile when the initial drywell temperature is raised from 135*F to 150*F. In addition, the response of the other containment parameters are not dependent upon which of the two initial drywell temperatures are used.
The loss of coolers case indicates a temperature profile which is higher by roughly the difference in the initial drywell temperatures. However, this profile is acceptable with respect to remaining below the design .
temperature for the drywell line- and remaining below the containment isolation setpoint on high drywell pressure.
Also, since the temperature profiles do not change significantly, the findings of TDR 564 (Reference 5) on the YARNAY level instrument response will still be valid.
Thus, it is concluded that the upper limit for drywell temperature for normal operations may be raised to 150*F. Similarly, the entry condition into the Containment Control section of the E0Ps may be raised to the same value.
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.h TR 024 Rev. O Page 16
6.0 REFERENCES
i
- 1. " Conduct of Operations", Oyster Creek Nuclear Generating Station f Procedure 106, Rev. 32.
- 2. G. F. Niederauer et. al., " CONTEMPT-EI/28C - A Computer Program for Predicting Containment Pressure - Temperature Transients", Energy Incorporated, EI-81-03, February 1981.
- 3. N. G. Trikouros, P. S. Smith, "0yster Creek Containment Temperature e for Environmental Qualification of Equipment", GPUN TDh #180, October 1980.
- 4. A. A. Irant, P. W. Lynches, "0yster Creek Appendix R Drywell Temperature Analysts for Remote Shutdown System", GPUN TDR #660, February 1905.
- 5. P. W. Lynches, " Preliminary Transient Response Analysis for YARWAY RPV Level Measurement Instrumentation at Oyster Creek", GPUN TDR #564, July 1984.
4 - 6. P. S. Smith, " Increase of E0P Entry Condition on Bulk Drywell Temperature". GPUN TDR 670, March 1985.
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