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Category:Report
MONTHYEARNLS2024065, 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report2024-10-0707 October 2024 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report NLS2024055, Data from Metamic Coupon Sampling Program2024-07-29029 July 2024 Data from Metamic Coupon Sampling Program ML23129A2792023-04-20020 April 2023 1 to Updated Safety Analysis Report, Plant Unique Analysis Report Mark I Containment Program NLS2023010, Inservice Inspection OAR-1 Owners Activity Report for Cooper Nuclear Station2023-02-0909 February 2023 Inservice Inspection OAR-1 Owners Activity Report for Cooper Nuclear Station NLS2023005, Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 20222023-01-19019 January 2023 Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 2022 NLS2022050, Notification of Revision to Cooper Nuclear Station Emergency Response Data System Data Point Library2022-11-0909 November 2022 Notification of Revision to Cooper Nuclear Station Emergency Response Data System Data Point Library NLS2022042, 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report2022-10-0707 October 2022 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report NLS2022003, Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 20212022-01-20020 January 2022 Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 2021 ML19262G9042019-11-0808 November 2019 Staff Assessment of Flood Hazard Integrated Assessment (Public) NLS2019040, 3-EN-DC-147, Rev. 5C1, Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (Efpy). (Non-proprietary)2019-08-0707 August 2019 3-EN-DC-147, Rev. 5C1, Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 54 Effective Full-Power Years (Efpy). (Non-proprietary) NLS2019028, Enclosure 2 - Calculation Package, File No. 1801303.301, Cooper Nuclear Station Core Shroud H3 Weld Evaluation - 2018, Revision 02018-10-24024 October 2018 Enclosure 2 - Calculation Package, File No. 1801303.301, Cooper Nuclear Station Core Shroud H3 Weld Evaluation - 2018, Revision 0 ML18184A2732018-07-18018 July 2018 Staff Review of Mitigation Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter (CAC No. MF7819; EPID L-2016-JLD-0006) NLS2018024, Report 004N7680-R1-NP, Gnf Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR, Coo(Er Nuclear Station Cycle 31.2018-04-30030 April 2018 Report 004N7680-R1-NP, Gnf Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR, Coo(Er Nuclear Station Cycle 31. NLS2018003, Submittal of Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 20172018-01-15015 January 2018 Submittal of Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 2017 NLS2017070, Seismic Mitigating Strategies Assessment Report for the Reevaluated Seismic Hazard Information2017-08-24024 August 2017 Seismic Mitigating Strategies Assessment Report for the Reevaluated Seismic Hazard Information ML17244A2812017-05-0808 May 2017 16C4384-RPT-005, Rev 005, 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation. NLS2016070, Completion of Required Action by NRC Order EA-12-049 - Mitigation Strategies for Beyond-Design-Basis External Events2017-01-0404 January 2017 Completion of Required Action by NRC Order EA-12-049 - Mitigation Strategies for Beyond-Design-Basis External Events NLS2016066, Completion of Required Action by NRC Order EA-12-051 - Reliable Spent Fuel Pool Instrumentation2016-12-20020 December 2016 Completion of Required Action by NRC Order EA-12-051 - Reliable Spent Fuel Pool Instrumentation ML16351A2472016-12-12012 December 2016 Review of SIA Calculation 1601004.301, Cooper High Pressure RHRSW Thinning Evaluation Per Code Case N-513 ML17018A1522016-11-0101 November 2016 Pressure and Temperature Limits Report, Revision 1 NLS2016058, 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report of Evaluations for 08/01/2014 - 07/31/20162016-10-0707 October 2016 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report of Evaluations for 08/01/2014 - 07/31/2016 ML16351A2512016-09-15015 September 2016 SW-E-3-2851-3, Ultrasonic Thickness Measurement System. ML16351A2502016-09-13013 September 2016 SW-Z4-2851-7, Altrasonic Thickness Measurement System. NLS2016021, Gnf Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR - Cooper Nuclear Station Cycle 30 (Non-Proprietary)2016-04-21021 April 2016 Gnf Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR - Cooper Nuclear Station Cycle 30 (Non-Proprietary) ML16084A1832016-03-0202 March 2016 Er 15-019, Rev. 1, Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY) (Non-Proprietary Version of Er 15-015). NLS2015128, Submittal of 10 CFR 71.95 Report Involving Energysolutions Certificate of Compliance No. 91682015-10-29029 October 2015 Submittal of 10 CFR 71.95 Report Involving Energysolutions Certificate of Compliance No. 9168 ML15259A3422015-08-24024 August 2015 10 CFR 71.95 Report Evaluation Form; Submitted by Fenok, Erwin Resin Solutions, North Anna Power Stations, Et Al NLS2015073, Er 15-019, Rev. 0, Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY) (Non-Proprietary), Enclosure 22015-08-0606 August 2015 Er 15-019, Rev. 0, Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY) (Non-Proprietary), Enclosure 2 NLS2015091, 10 CFR 71.95 Report Involving Energysolutions Certificate of Compliance No. 91682015-07-28028 July 2015 10 CFR 71.95 Report Involving Energysolutions Certificate of Compliance No. 9168 NLS2015043, Expedited Seismic Evaluation Process Report2015-04-29029 April 2015 Expedited Seismic Evaluation Process Report ML15006A2342015-02-11011 February 2015 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-13-109 (Severe Accident Capable Hardened Vents) ML17018A1532014-12-31031 December 2014 BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program Data Source Book and Plant Evaluations NLS2016014, Er 15-019, Supplement 1, BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations.2014-12-31031 December 2014 Er 15-019, Supplement 1, BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. NLS2014074, Submittal of 10 CFR 71.95 Report Involving Areva Certificate of Compliance No. 92332014-07-30030 July 2014 Submittal of 10 CFR 71.95 Report Involving Areva Certificate of Compliance No. 9233 ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 NLS2014044, and ISFSI - Thirty-Day Notification Pursuant to 10 CFR 72.212, Conditions of General License Issued Under Section 72.2102014-05-13013 May 2014 and ISFSI - Thirty-Day Notification Pursuant to 10 CFR 72.212, Conditions of General License Issued Under Section 72.210 NLS2014027, Enclosure - Seismic Hazard Evaluation and Screening Report for Cooper Nuclear Station2014-03-20020 March 2014 Enclosure - Seismic Hazard Evaluation and Screening Report for Cooper Nuclear Station ML14007A6502014-02-11011 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14037A1672014-02-10010 February 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for Cooper Nuclear Station, TAC No.: MF0972 NLS2012130, Problem Identification and Resolution P. 1(c) Substantive Cross-Cutting Issue2012-12-13013 December 2012 Problem Identification and Resolution P. 1(c) Substantive Cross-Cutting Issue ML12340A2752012-11-27027 November 2012 Engineering Evaluation 12-E18, Revision 0, Attachment E Through J ML12340A2732012-11-27027 November 2012 Engineering Evaluation 12-E18, Revision 0, Attachment D - Area Walk-By Checklists ML12340A2692012-11-27027 November 2012 Engineering Evaluation 12-E18, Revision 0, Attachment C - Page 76 of 322 ML12340A2682012-11-27027 November 2012 CNS Memo DED12-0003 Response to 10 CFR 50.54(f) Section 2.3 Seismic, Enclosing Engineering Evaluation 12-E18, Revision 0 NLS2012124, Enclosure to NLS2012124, Cooper Nuclear Station Flooding Walkdown Report2012-11-26026 November 2012 Enclosure to NLS2012124, Cooper Nuclear Station Flooding Walkdown Report NLS2012085, CFR 50.59(d)(2) Summary Report Covering Time Period from August 1, 2012 to July 31, 20122012-10-10010 October 2012 CFR 50.59(d)(2) Summary Report Covering Time Period from August 1, 2012 to July 31, 2012 NLS2012089, Submittal of Nuclear Material Transaction Report2012-08-29029 August 2012 Submittal of Nuclear Material Transaction Report ML1216000292012-06-11011 June 2012 Review of 60-day Response to Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 9.3 of the Near Term Task Force Review of Fukushima Dai-ichi Accident NLS2012040, Enclosure 2, Gnf S-0000-0140-2518-R0-NP, Gnf Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR - Cooper Cycle 282012-05-30030 May 2012 Enclosure 2, Gnf S-0000-0140-2518-R0-NP, Gnf Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR - Cooper Cycle 28 NLS2012006, Nebraska Public Power District - Cooper Nuclear Station, License Amendment Request to Revise the Fire Protection Licensing Basis to NFPA 805 Per 10 CFR 50.48(c)2012-04-24024 April 2012 Nebraska Public Power District - Cooper Nuclear Station, License Amendment Request to Revise the Fire Protection Licensing Basis to NFPA 805 Per 10 CFR 50.48(c) 2024-07-29
[Table view] Category:Miscellaneous
MONTHYEARNLS2024065, 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report2024-10-0707 October 2024 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report NLS2024055, Data from Metamic Coupon Sampling Program2024-07-29029 July 2024 Data from Metamic Coupon Sampling Program NLS2023005, Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 20222023-01-19019 January 2023 Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 2022 NLS2022042, 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report2022-10-0707 October 2022 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report NLS2022003, Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 20212022-01-20020 January 2022 Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 2021 ML19262G9042019-11-0808 November 2019 Staff Assessment of Flood Hazard Integrated Assessment (Public) ML18184A2732018-07-18018 July 2018 Staff Review of Mitigation Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter (CAC No. MF7819; EPID L-2016-JLD-0006) NLS2018003, Submittal of Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 20172018-01-15015 January 2018 Submittal of Annual Report of Changes and Errors in Emergency Core Cooling System Evaluation Models for 2017 ML17244A2812017-05-0808 May 2017 16C4384-RPT-005, Rev 005, 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation. NLS2016070, Completion of Required Action by NRC Order EA-12-049 - Mitigation Strategies for Beyond-Design-Basis External Events2017-01-0404 January 2017 Completion of Required Action by NRC Order EA-12-049 - Mitigation Strategies for Beyond-Design-Basis External Events NLS2016066, Completion of Required Action by NRC Order EA-12-051 - Reliable Spent Fuel Pool Instrumentation2016-12-20020 December 2016 Completion of Required Action by NRC Order EA-12-051 - Reliable Spent Fuel Pool Instrumentation ML16351A2472016-12-12012 December 2016 Review of SIA Calculation 1601004.301, Cooper High Pressure RHRSW Thinning Evaluation Per Code Case N-513 NLS2016058, 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report of Evaluations for 08/01/2014 - 07/31/20162016-10-0707 October 2016 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2) Summary Report of Evaluations for 08/01/2014 - 07/31/2016 NLS2016021, Gnf Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR - Cooper Nuclear Station Cycle 30 (Non-Proprietary)2016-04-21021 April 2016 Gnf Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR - Cooper Nuclear Station Cycle 30 (Non-Proprietary) NLS2015128, Submittal of 10 CFR 71.95 Report Involving Energysolutions Certificate of Compliance No. 91682015-10-29029 October 2015 Submittal of 10 CFR 71.95 Report Involving Energysolutions Certificate of Compliance No. 9168 ML15259A3422015-08-24024 August 2015 10 CFR 71.95 Report Evaluation Form; Submitted by Fenok, Erwin Resin Solutions, North Anna Power Stations, Et Al ML15006A2342015-02-11011 February 2015 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-13-109 (Severe Accident Capable Hardened Vents) NLS2014074, Submittal of 10 CFR 71.95 Report Involving Areva Certificate of Compliance No. 92332014-07-30030 July 2014 Submittal of 10 CFR 71.95 Report Involving Areva Certificate of Compliance No. 9233 NLS2014044, and ISFSI - Thirty-Day Notification Pursuant to 10 CFR 72.212, Conditions of General License Issued Under Section 72.2102014-05-13013 May 2014 and ISFSI - Thirty-Day Notification Pursuant to 10 CFR 72.212, Conditions of General License Issued Under Section 72.210 NLS2012124, Enclosure to NLS2012124, Cooper Nuclear Station Flooding Walkdown Report2012-11-26026 November 2012 Enclosure to NLS2012124, Cooper Nuclear Station Flooding Walkdown Report NLS2012085, CFR 50.59(d)(2) Summary Report Covering Time Period from August 1, 2012 to July 31, 20122012-10-10010 October 2012 CFR 50.59(d)(2) Summary Report Covering Time Period from August 1, 2012 to July 31, 2012 NLS2012089, Submittal of Nuclear Material Transaction Report2012-08-29029 August 2012 Submittal of Nuclear Material Transaction Report ML1216000292012-06-11011 June 2012 Review of 60-day Response to Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 9.3 of the Near Term Task Force Review of Fukushima Dai-ichi Accident ML1127000692011-09-26026 September 2011 Enclosure 2, Mfn 10-245 R4, Description of the Evaluation and Surveillance Recommendations for BWR/2-5 Plants NLS2009099, SAMA Meteorological Anomaly Related to the Cooper Nuclear Station License Renewal Application2009-12-0707 December 2009 SAMA Meteorological Anomaly Related to the Cooper Nuclear Station License Renewal Application NLS2008106, Submittal of Program for Maintenance of Irradiated Fuel2008-12-23023 December 2008 Submittal of Program for Maintenance of Irradiated Fuel NLS2008096, Corrected Semiannual Fitness for Duty Program Performance Report for the Period of January 1, 2008 Through June 30, 20082008-10-22022 October 2008 Corrected Semiannual Fitness for Duty Program Performance Report for the Period of January 1, 2008 Through June 30, 2008 NLS2008059, Submittal of Revised Root Cause Investigation - Diesel Generator Amphenol Connector2008-07-21021 July 2008 Submittal of Revised Root Cause Investigation - Diesel Generator Amphenol Connector NLS2008033, Response to Request for Additional Information for Question I.2 Regarding License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate2008-03-12012 March 2008 Response to Request for Additional Information for Question I.2 Regarding License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate NLS2008019, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate2008-03-0606 March 2008 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate ML0917502992008-01-0101 January 2008 Lr - NPPD 2008 Integrated Resource Plan NLS2006090, Core Operating Limits Report, Cycle 24, Revision 02006-11-0505 November 2006 Core Operating Limits Report, Cycle 24, Revision 0 NLS2006084, CFR 50.59(d)(2) Summary Report2006-10-12012 October 2006 CFR 50.59(d)(2) Summary Report NLS2005107, Reporting of Changes and Errors in ECCS Evaluation Models Cooper Nuclear Station2005-12-28028 December 2005 Reporting of Changes and Errors in ECCS Evaluation Models Cooper Nuclear Station ML0520901582005-08-0202 August 2005 2004 External Stakeholder Response; 2004 Reactor Oversight Process External Survey - Attachment ML0522700732005-04-18018 April 2005 Event Notification Report for April 18, 2005 ML0727008492005-01-31031 January 2005 Caldon Experience in Nuclear Feedwater Flow Measurement NLS2004117, 1OCFR50.59(d)(2) Summary Report2004-10-14014 October 2004 1OCFR50.59(d)(2) Summary Report NLS2004105, Initial Actions Summary Report and Response to NRC Generic Letter 2003-01, Control Room Habitability.2004-09-30030 September 2004 Initial Actions Summary Report and Response to NRC Generic Letter 2003-01, Control Room Habitability. ML0619802152004-08-0606 August 2004 CNS Assessment of Survivability of the Service Water Pumps When Gland Waste Water Was Lost ML0622205762004-07-27027 July 2004 Johnston Pump Company Letter to Mr. Kent Sutton ML0619801532004-07-16016 July 2004 Excerpts from Historic Work Orders ML0532204612004-06-14014 June 2004 Power Reactor Status Report for 6/14/04 ML0618607342004-05-11011 May 2004 Comments Concerning Loss of Gland Water to the Service Water Pumps at Cooper Nuclear Station ML0618607312004-05-0606 May 2004 Comments Concerning Loss of Gland Water to the Service Water Pumps at Cooper Nuclear Station ML0618701942004-04-30030 April 2004 Excerpt from Missouri Nuclear Accident Plan ML0619801092004-03-17017 March 2004 CNS Roots Cause Investigation - Service Water Gland Water Valve Mis-positioning Event, SCR 2004-0077 ML0619801012004-03-10010 March 2004 CNS Clearance Order, Section SWB-1-4324147 SW-STNR-B ML0626304082004-03-0101 March 2004 PSEG Nuclear, LLC Salem/Hope Creek Safety Culture Assessment ML0619800892004-02-18018 February 2004 CNS Notification 10294449 2024-07-29
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Comments Concerning Loss of Gland Water to the Service Water Pumps at Cooper Nuclear Station Randall Noon, P.E., preparer Cooper Nuclear Station May 6, 2004 1-- /7
\Comments About Gland Water Loss 2 Premise It is postulated that a valve supplying gland water to a service water pump is closed, and the service water pump is deprived of gland water. How would this affect the service water pump?
Background Information The service water pumps at Cooper Nuclear Station are Byron-Jackson, mixed flow, single stage, vertical, irrigation type pumps. They operate at 1180 rpm, have a rating of 8000 gpm, and are driven by a 300 hp electric motor.
Loss of gland water does not directly affect the motor-driver.
Loss of gland water does not directly affect the lower bronze bearing at the impeller. It is immersed in the river and will self lubricate sufficiently to be unaffected by the loss of gland water.
Loss of water does not directly affect the impeller and wear ring. They are immersed in the river.
Loss of gland water primarily affects the rubber, Cutlass bushings that are spaced along the length of the shaft.
There are eleven Cutless bushings, more or less evenly spaced from the bottom to the top of the pump shaft. The bushings have an inside diameter of 2.193 to 2.199 inches, and are 7.5 inches long. There are ten flutes. The shaft that fits through the bushing is nominally 2.1875 inches in diameter.
The bushing material that contacts the pump shaft is nitrile rubber, or buna-N.
The maximum service temperature for nitrile is 250 degrees F. Nitrile rubber is often used for rotary shaft seals and o-rings. The tensile strength at room temperature is about 2000 psi.
The rubber bushings in the service water pump do not carry significant bearing loads. Their purpose is to minimize lateral movement of the pump shaft, since it is relatively long. As is noted in Kent's Mechanical Engineers' Handbook, 12th Edition, "Design and Production" Volume, page 12-46:
\Comments About Gland Water Loss 3 Rubber is used to line bearings where only water is available as lubricantand where bearing pressureis light. It is useful in guide bearingbushingsfor vertical revolving shafts, and in stern tube bearingsin ships.
Because the pump shaft is relatively long, about 45 feet, any centerline or mass distribution eccentricities in the pump shaft will cause the shaft to vibrate or wobble when it rotates. If the Cutless bushings were not present, lateral motion would be greatest at the center of the shaft, and least at the two ends.
By placing a bushing at the center of the shaft, the maximum lateral motion at that point is reduced to the clearance dimension. This effectively divides the pump shaft then into two shafts with smaller wobbles at their midpoints, or at the 1/4 and 3/4 points along the shaft. With eleven Cutless bushings, the pump shaft is effectively eleven short shafts that wobble slightly at their midpoints.
Due to the stiffness of the shaft between bushings, the maximum lateral motion of the pump shaft is essentially the clearance between the shaft and the bushings.
The pumps originally were equipped with bronze bushings. However, this was changed to rubber Cutless type bushings since rubber is more resilient to impact.
Since rubber is much softer than bronze, occasional contact with the rubber surface does not scratch or gall the 410 hardened stainless steel pump shaft.
Typically, when gland water is supplied to a service water pump, water flows downward through the bushings. It flows through the fluted slots in the bushing, and it flows through the clearance between the shaft and the bushing I.D.
A typical gland water flow rate is 5 to 10 gpm. Five gpm is equivalent to 0.668 cubic feet of water per minute. Since the annular space between the enveloping tube and the pump shaft is about 7.75 square inches, the flow velocity in this space is about 12.4 ft/min or 0.207 ft/sec at 5 gpm, or 0.414 ft/sec at 10 gpm.
This is very slow, and is just slightly higher than flow velocities associated with natural convection with modest temperature differentials. Because of the constriction due to the bushings, flow in the area of the bushing is faster than that estimated for the annular space between bushings.
When there is gland water, water in the clearance between the shaft and the Cutless bushings forms a typical lubrication wedge. This wedge also helps "cushion" and dampen lateral motion of the pump shaft. The surface speed of the shaft at 1180 rpm is 11.26 ft/sec.
\Comments About Gland Water Loss 4 The 410 hardened stainless steel shaft has a hardness of about 225 BHN.
Consequently, it has a tensile strength of about 106,000 psi and a yield strength of about 92,000 psi. At 250 degrees F, the 410 stainless steel shaft has not significantly lost any material strength.
Discussion If gland water is removed while the pump is running, the following will occur.
The level of water in the enveloping tube will drop to a point slightly below that of the river. The level will not exactly match that of the river because the pressure at the bottom hub of the impeller, that is, at the "eye" of the impeller, will be slightly less than the ambient pressure at the same elevation elsewhere in the "E" bay. When the pump is operating at full load, this decrease in pressure will be about 0.75 psi. This figure is based upon the loss of pressure head at that elevation that is due to flow velocity near the inlet of the pump.
Consequently, the water level in the enveloping tube will be about 1.73 feet lower than the level of the river. Assuming that the river is at a "typical" elevation of perhaps 880', then the lower half of the enveloping tube will still have water in it. Thus, about half of the bushings will still have water for lubrication.
The upper bushings, however, will likely eventually becomes dry and will loose the wedge of water between the rotating shaft and the Cutless bushings. Thus, dry contact between the shaft and the pump shaft may occur.
Frictional contact between the bushings and smooth, polished shaft will eventually cause the surface of the bushing to heat up and form a hard glaze on the surface. Heated nitrile rubber become brittle and hard. Further, as the nitrile rubber heats up, it volatilizes and outgases. The net result of out gassing is that the heated rubber will lose volume and shrink. This is, of course, why overheated rubber exhibits deep cracking.
If the temperature exceeds 400 degrees F, the nitrile rubber will have lost so much material strength that further frictional contact with the shaft will simply tear away material on its surface. This process will continue until the inside diameter of the bushings is sufficiently wallowed out to match the total lateral motion of the shaft.
\Comments About Gland Water Loss 5 It has been hypothesized that the Cutless bushings will "grab" the shaft and cause it to seize. This is improbable unless there is significant misalignment. The shaft can not contact all the rubber in the bushing, that is, the shaft can not be "grabbed" by the rubber like a prony brake clinching it. At most, with a continuously applied lateral force, the shaft can only contact half of a bushing's inside diameter, that is, 180 degrees of the inside diameter, at one instant.
Further, in order for the shaft to be "grabbed" a lateral force must press the shaft continuously into the rubber at one spot. However, the lateral motion caused by centerline eccentricity or mass distribution eccentricity is sinusoidal. It does not press the shaft continually into the bushing in one spot. Instead, it "bangs" around in the hole as the shaft rotates. Actual contact between the shaft and the bushing is intermittent.
It is true that drag on the shaft will be increased if the bushings run dry.
However, the pump motor-driver is rated for 300 hp with 15% excess. It has more than enough torque to overcome the increase in drag.
This was demonstrated in December 2001, when the impeller and bowl of the "D" service water pump where jammed together sufficiently to prevent rotation of the impeller. The torque of the motor completely twisted off the 410 hardened stainless steel coupling.
If the lateral force on each bushing, when it is dry, were about 100 pounds, which is grossly high, the frictional force between the shaft and each bushing will be 50 pounds. This will create a frictional torque drag of 55 in-lbs or 4.6 ft-lbs at each bushing. Since perhaps six of the bushings will eventually become dry, this is an increase in torque of 27.3 ft-lbs. At running load, the pump motor driver is capable of developing 1,334 ft-lbs of torque. Thus, the drag of six dry bushings will "steal" about 2% of the motor's torque.
Typical vibration measurements on the service water pumps indicate that the vibrational velocity is about 0.1 or 0.2 inches per second. The service water pumps are allowed to have a vibration level of about 0.7 in/sec before they have to be serviced.
At 0.1 in/sec, the peak-to-peak displacement is about 0.0008 inches. At 0.7 in/sec, the peak-to-peak displacement is about 0.0056 inches.
\Comments About Gland Water Loss 6 The maximum allowable diametrical clearance between the shaft and the bushings is 0.026 inches. The initial clearance can be as tight at 0.006 inches, but would normally be around 0.010 or 0.012 inches.
If the pump were allowed to operate until the 0.7 in/sec vibration level were reached due to sideways wobble, the vibrational "shake" pattern of the shaft still fits inside the diametric clearance of the bushing.
If allowed to continue further that 0.7 in/ sec, the bounding limitation appears to be the vibrational tolerance of the lower bearing in the driver motor. Increased vibrational levels in the lower bearing of the motor-driver can cause its life to shorten. If the induced vibrations were sufficiently severe to cause high radial loading, the bearing will degrade, become damage, and eventually fail.