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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML18066A4671999-03-31031 March 1999 Rev 0 to SIR-99-032, Flaw Tolerance & Leakage Evaluation Spent Fuel Pool Heat Exchanger E-53B Nozzle Palisades Nuclear Plant. ML20249C4951998-06-17017 June 1998 Rev 1 to EA-GEJ-98-01, Palisades Cycle 14 Disposition of Events Review ML18066A3411998-04-22022 April 1998 Rev 0 to EMF-98-013, Palisades Cycle 14:Disposition & Analysis of SRP Chapter 15 Events. ML20217C2741998-03-31031 March 1998 Independent Review - Is Consumers Energy Method (W Method) of Determining Palisades Nuclear Plant Best Estimate Fluence by Combining Transport Calculation & Dosimetry Measurements Technically Sound & Does It Meet Intent of Pts ML18065B1641998-02-0505 February 1998 Rev 0 to Regression Analysis for Containment Prestressing Sys at 25th Year Surveillance. ML20197J3891997-12-18018 December 1997 25th Year Physical Surveillance of Palisades Npp ML20217C2571997-12-16016 December 1997 Review of Neutron Fluence Data for Palisades Reactor Pressure Vessel ML18067A6351997-07-0909 July 1997 Excerpt from Ampacity Evaluation for Open Air Cable Trays W/Percent Fill Greater than 30% of Useable Cross Sectional Area. ML18067A6381997-07-0909 July 1997 Excerpt from Ampacity Evaluation for Continuously Energized Power Cables Routed Through Fire Stops, Revision 1 ML18067A6371997-07-0808 July 1997 Excerpt from Ampacity Evaluation for Duct Runs Containing Continuously Energized Power Cables, Revision 1 ML18067A6361997-06-26026 June 1997 Excerpt from Ampacity Evaluation for Continuously Energized Power Cables in Open Air Conduits, Revision 1 ML18066A8581997-01-31031 January 1997 Rev 2 to C-PAL-96-1063-01, Operability Assessment for Transient Conditions at Palisades Nuclear Plant in Response to GL 96-06. ML18065B0471996-07-12012 July 1996 TR on Use of Mcbend Code for Calculation of Neutron Fluences in PVs of Lwrs. ML18065A7571996-05-22022 May 1996 Rev 1 to IPEEE Rept, Per GL 88-20 ML20108C1671996-04-0101 April 1996 Nonproprietary Version of Fluence Calculations for Palisades Plant ML18065A5971996-03-23023 March 1996 Evaluation of Effects of Fire on West Wall of Turbine Lube Oil Room Adjacent to Pipe Tunnel Between TB & FW Purity Bldg. ML18065A6011996-03-22022 March 1996 Evaluation of Effects of Fire on West Wall of CCW Pump Room (Fire Area 16). ML20100D7491996-01-18018 January 1996 Rev 0 to Evaluation of Effects of Fire on West Wall of TB Lube Oil Room Adjacent to Pipe Tunnel Between TB & FW Purity Bldg ML18065A4481995-12-14014 December 1995 Radiological Consequences for Palisades Max Hypothetical Accident & Loss of Coolant Accident. ML18064A8321995-06-30030 June 1995 IPE of External Events (Ipeee). ML20085H2801995-05-23023 May 1995 Security Investigation Rept ML18064A7801995-05-19019 May 1995 Rept of SQUG Assessment at Palisades Nuclear Plant for Resolution of USI A-46. ML20078P7021995-01-27027 January 1995 Investigative Rept ML18064A4121994-08-22022 August 1994 Pressure-Temp Curves & LTOP Setpoint Curve for Max Reactor Vessel Fluence of 2.192 X 10^19 Neutrons/cm^2. ML20070J8001994-07-15015 July 1994 Final Rept Containment Sump Check Valves Weld Overlay Repair Implementation Evaluation Palisades Nuclear Plant ML18059B0041994-04-0505 April 1994 Rev 1 to EDG Fuel Supply Sys Storage Tank Tornado Protection Overview of EDG Fuel Supply Sys, Incorporating CARB Comments of 940318 & 24 ML20064E5301994-03-0606 March 1994 Evaluation of Effectiveness of Code Case N-504-1 Repair for Proposed Root Causes for Containment Sump Suction Check Valves ML20064E4451994-03-0505 March 1994 Check Valve Leak Root Cause,Engineering Analysis & Repair/Replacement Options ML18059A5161993-10-31031 October 1993 Nonproprietary Exam...Sections of Pressurizer PORV Line Safe-End Failure from Palisades Nuclear Generating Station. ML20058P1361993-10-31031 October 1993 Crack Propagation Analysis for Circumferential Cracks in Alloy 600 Nozzle Safe-Ends ML18059A4821993-10-25025 October 1993 Evaluation of Potential Interference Between TE-0102 Nozzle & Thermowell. ML18059A4831993-10-25025 October 1993 Structural Evaluation for Machined Thermawell for TE-0101. ML20059D8811993-10-23023 October 1993 Justification of Weld Mods to Pressurizer Temperature Nozzles for TE-0101 & TE-0102 ML18059A4811993-10-22022 October 1993 Acceptability of Partial Severing of TE-0101 Nozzle. ML18059A4801993-10-19019 October 1993 Structural Analysis of Temperature Nozzle Weld Mods for Consumers Power Palisades Pressurizer. ML18059A4791993-10-15015 October 1993 Half Bead Welding for Mods to TE-0101 & TE-0102. ML18059A4221993-10-0707 October 1993 Pressurizer Safe End Crack Engineering Analysis & Root Cause Evaluation. ML18059A3751993-08-31031 August 1993 Rev 1 to Palisades Cycle 11:Disposition & Analysis of SRP Chapter 15 Events. ML18059B0191993-07-31031 July 1993 Detailed Site Study,Berrien County,Mi, Final Rept ML18064A4271993-06-30030 June 1993 Wind Tunnel Predictions of Control Room Intake Concentrations from Three Sources of Radioactive Materials at Palisades Nuclear Power Plant, (CPP-Project 93-0907) ML18058B8661993-05-13013 May 1993 Resolution of Anchor Bolt Design Issues. ML18058B3911992-12-21021 December 1992 Cycle 11:Disposition & Analysis of Standard Review Plan Chapter 15 Events. ML18058B4281992-11-30030 November 1992 Vols 1,2 & 3 of Palisades Nuclear Plant Ipe. ML18058A5391992-06-16016 June 1992 Twentieth Yr Physical Surveillance of Palisades Nuclear Plant. ML20086P8551991-12-0909 December 1991 Criticality Safety Analysis for Palisades Spent Fuel Storage Pool NUS Racks ML20086P8571991-12-0909 December 1991 Criticality Safety Analysis for Palisades New Fuel Storage Array ML18057B3521991-10-31031 October 1991 Large Break Loca/Eccs Analysis W/Increased Radial Peaking & Reduced ECCS Flow. ML18057A8591991-03-31031 March 1991 Benchmarking & Validation of In-House DOT Calculation Methodology. ML20081K7741990-08-14014 August 1990 Incore Detector Algorithm (Pidal) Analysis of Quadrant Power Tilt Uncertainties ML18057A2611990-06-11011 June 1990 Simulator Certification Submittal. 1999-03-31
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML18066A6901999-11-0101 November 1999 Rev 5 to Palisades Nuclear Plant Colr. ML18066A6761999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Palisades Nuclear Plant ML18066A6271999-09-0202 September 1999 LER 98-011-01:on 981217,inadequate Lube Oil Collection Sys for Primary Coolant Pumps Was Noted.Caused by Design Change Not Containing Appropriate Level of Rigor.Exemption from 10CFR50,App R Was Requested.With 990902 Ltr ML18066A6351999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Palisades Nuclear Plant ML18066A6771999-08-31031 August 1999 Operating Data Rept Page of MOR for Aug 1999 for Palisades Nuclear Plant ML18066A6221999-08-20020 August 1999 LER 99-002-00:on 990722,TS Surveillance Was Not Completed within Specified Frequency.Caused by Failure to Incorporate Revised Frequency Into Surveillance Schedule in Timely Manner.Verified Implementation.With 990820 Ltr ML18066A6061999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Palisades Nuclear Plant.With 990803 Ltr ML18066A5201999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Palisades Nuclear Plant.With 990702 Ltr ML18066A4841999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Palisades Nuclear Plant.With 990603 Ltr ML18066A6371999-04-30030 April 1999 Revised Monthly Operating Rept for Apr 1999 for Palisades Nuclear Plant ML18068A5941999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Palisades Nuclear Plant.With 990503 Ltr ML18066A4161999-04-0101 April 1999 Rev 4 to COLR, for Palisades Nuclear Plant ML18066A4501999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Palisades Nuclear Plant.With 990402 Ltr ML18066A4671999-03-31031 March 1999 Rev 0 to SIR-99-032, Flaw Tolerance & Leakage Evaluation Spent Fuel Pool Heat Exchanger E-53B Nozzle Palisades Nuclear Plant. ML18068A5351999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Palisades Nuclear Plant.With 990302 Ltr ML18066A3931999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Palisades Nuclear Plant.With 990202 Ltr ML18066A3781999-01-20020 January 1999 LER 98-013-00:on 981222,safeguards Transfer Tap Changer Failure Caused Inadvertant DG Start.Caused by Failed Motor Contactor.Contactor Was Replaced.With 990120 Ltr ML20206F6131998-12-31031 December 1998 1998 Consumers Energy Co Annual Rept. with ML18066A3651998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Palisades Nuclear Plant.With 990105 Ltr ML18066A3421998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Palisades Nuclear Plant.With 981202 Ltr ML18066A3301998-11-11011 November 1998 Part 21 Rept Re Potential Safety Hazard Associated with Wrist Pin Assemblies for FM-Alco 251 Engines at Palisades Nuclear Power Plant.Caused by Insufficient Friction Fit Between Pin & Sleeve.Supplier of Pin Will No Longer Be Used ML18068A4921998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Palisades Nuclear Plant.With 981103 Ltr ML18068A4851998-10-29029 October 1998 LER 97-011-01:on 971012,starting of Primary Coolant Pump with SG Temps Greater than Cold Leg Temps Occurred.Caused by Inadequate Procedures & Operator Decision.Sop Used for Starting Primary Coolant Pump Enhanced ML18066A3181998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Palisades Nuclear Plant ML18066A2901998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Palisades Nuclear Power Plant.With 980903 Ltr ML18066A3191998-08-31031 August 1998 Revised Monthly Operating Rept Data for Aug 1998 for Palisades Nuclear Plant ML18066A2831998-08-18018 August 1998 LER 98-010-00:on 980721,reactor Manually Tripped.Caused by Failure of Coupling Which Drives Feedwater Pump Main Lube Oil Pump.Main Lube Oil Pump Coupling & Associated Components Replaced & Satisfactorily Tested ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency ML20237E0301998-07-31031 July 1998 ISI Rept 3-3 ML18066A2701998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Palisades Nuclear Plant.W/980803 Ltr ML18066A2311998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Palisades Nuclear Plant ML18066A2261998-06-30030 June 1998 LER 98-009-00:on 980531,small Pinhole Leak Found on One of Welds,During Leak Test Following Replacement of Pcs Sample Isolation Valves.Caused by Welder Error.Leaking Welds Repaired ML18066A3061998-06-18018 June 1998 SG Tube Inservice Insp. ML20249C4951998-06-17017 June 1998 Rev 1 to EA-GEJ-98-01, Palisades Cycle 14 Disposition of Events Review ML18066A1781998-06-0909 June 1998 LER 98-008-00:on 980511,noted That Procedure Did Not Fully Satisfy Requirement to Test High Startup Rate Trip Function. Caused by Misunderstanding of Testing Requirements.Revised TS Surveillance Test Procedure & Reviewed Other Procedures ML18066A1711998-06-0101 June 1998 Part 21 Rept Re Impact of RELAP4 Excessive Variability on Palisades Large Break LOCA ECCS Results.Change in PCT Between Cycle 13 & Cycle 14 Does Not Constitute Significant Change Per 10CFR50.46 ML18066A1741998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Palisades Nuclear Plant.W/980601 Ltr ML18066A2321998-05-31031 May 1998 Revised MOR for May 1998 for Palisades Nuclear Plant ML18068A4701998-05-31031 May 1998 Annual Rept of Changes in ECCS Models Per 10CFR50.46. ML18065B2451998-05-13013 May 1998 LER 98-007-00:on 980413,HPIS Sys Was Noted Inoperable During TS Surveillance Test.Caused by Performance of Flawed Procedure.Operators & Engineers Will Be Trained to Improve Operational Decision Making Through Resources & Knowledge ML18066A2331998-04-30030 April 1998 Revised MOR for Apr 1998 for Palisades Nuclear Plant ML18068A3461998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Palisades Nuclear Plant.W/980501 Ltr ML18066A3411998-04-22022 April 1998 Rev 0 to EMF-98-013, Palisades Cycle 14:Disposition & Analysis of SRP Chapter 15 Events. ML18065B2071998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Palisades Nuclear Plant.W/980403 Ltr ML20217C2741998-03-31031 March 1998 Independent Review - Is Consumers Energy Method (W Method) of Determining Palisades Nuclear Plant Best Estimate Fluence by Combining Transport Calculation & Dosimetry Measurements Technically Sound & Does It Meet Intent of Pts ML18066A2341998-03-31031 March 1998 Revised MOR for Mar 1998 for Palisades Nuclear Plant ML18068A3041998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Palisades Nuclear Plant.W/980302 Ltr ML18066A2351998-02-28028 February 1998 Revised MOR for Feb 1998 for Palisades Nuclear Plant ML18065B1641998-02-0505 February 1998 Rev 0 to Regression Analysis for Containment Prestressing Sys at 25th Year Surveillance. ML18067A8211998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Palisades Nuclear Plant.W/980203 Ltr 1999-09-30
[Table view] |
Text
Attachment B ATTACHMENT B Consumers Power Company Palisades Plant Docket 50-255 DIFFERENTIAL PRESSURE TEST BASIS January 15, 1988 0001200025 e~g&l,~ 55 PDR ADOCK 0 PDR Pages 8 G
MI0687-2096D-MA05
Attachment B
- DIFFERENTIAL PRESSURE TEST BASIS AUXILIARY FEEDWATER SYSTEM Equipment ID M0-0743, M0-0748, M0-0753, M0-0760 Test Basis The two worst case conditions for operation of the above MOV's are opening the valve after inadvertent closure or closing the valve after a line break downstream. The maximum pressure drop across the valve is developed with
- P-8A/B on minimum flow and 0 psig in the steam generators. _It is important to note that .the steam generators 'are normally pressurized under accident conditions, greatly minimizing the pressure drop across these valves.
The differential pressure is performed with Auxiliary Feedpump P-8A due to the cold shutdown conditions required for test performance. Steam is required to drive Auxiliary Feedpump P-88, and steam is not available with the plant in cold shutdown. If P-8B is used, the presence of steam in the steam generator will limit the pressure drop across the motor operated valves, which would make the test nonconservative.
Auxiliary Feedpwnp P-8A- develops between 1500 psig and 1540 psig at minimum flow. Relief valve RV-0783 is set at 1565 +/- 15 psig. The test is performed with P-8A on minimum flow, the auxiliary feedwater piping full, and the steam generator at 0 psig. Steam generator level will vary throughout the test, and the effects of steam generator level above the auxiliary feedwater nozzles are ignored as negligible. The maximum steam generator level is 6 feet above the nozzles *
2 This test does not meet the CE design pressure drop criteria for these valves, but it does demonstrate operability under the most severe accident conditions.
A hydropump will develop a higher initial pressure drop, but the pressure drop will quickly disappear when the valve begins to open, and very little flow is developed. Using Auxiliary Feedpump P-8A will develop a slightly lower initial
~P, but it will subject the MOV to the continued stresses of stroking under service conditions.
Testing the MOV's with P-8A on minimum flow is representative of the actual accident conditions the valves may see. Draining the discharge piping to increase the initial ,pressure drop across the valve is undesirable due to water hammer considerations. Condensate tank level is maintained at 95 to 100% to maximize Auxiliary Feedpump discharge pressure. Although higher pump discharge pressures may be attained by valving fire protection into the P-8A suction, this is not desirable due to Chemistry concerns. Valving fire protection in is a very cons~ervati ve action that will only be taken in an extreme emergency.
The MOV's are required to close under high flow conditions due to a line break downstream. This is simulated by placing the line in service with full flow through the MOV. While it is not feasible to simulate a line break at the MOV, the test is performed with the steam generator pressure at 0 psig. The valves are cycled one at a time.
Equipment ID M0-0748, M0-0755, M0-0754, M0-0759 Test Basis The two worst case conditions for operation o*f the above MOV' s are opening the valve after inadvertent closure, or closing the valve after a line break downstream. The maximum pressure drop across the valve is developed with P-8C on minimum flow and 0 psig in the steam generators. It is important to note that the steam generators are normally pressurized under accident conditions.
This will greatly minimize the pressure drop across the MOV's.
Auxiliary Feedpump P-8C develops approximately 1275 psig on minimum flow. The test is performed with P-8C on minimum flow, the auxiliary feedwater piping full, and the steam generator at 0 psig. Steam generator level will vary throughout the test, and the effects of steam generator level above the auxiliary feedwater nozzles are ignored as negligible. The maximum steam generator level is 6 feet above the nozzles.
This test does not meet the CE design pressure drop criteria for these valves, but it dows demonstrate operability under the most severe accident conditions.
A hydropump will develop a higher initial pressure drop, but the pressure drop will quickly disappear when the valve begins to open, and very little flow is developed. Using Auxiliary Feedpump P-8C will develop a slightly lower initial pressure drop, but it will subject the MOV to the continued stresses of stroking under service conditions.
MI0687-2096D-MA05
3
- Testing the MOV's with P-8C on minimum flow is representative of the actual conditions the valves may see. Draining the discharge piping to increase the initial pressure drop .across the valve is undesirable due to water hanmer considerations. Condensate ta.nk level is maintained at 95 to 100% to maximize Auxiliary Feedpump discharge pressure. Although higher pump discharge pressures may be, attained by valving service water into the P-8C suction, this is not desirable due to Chemistry concerns. It is also a very conservative action.
The MOV's are required to close under high flow conditions due to a line break downstream. This is simulated by placing the line in service with full flow through the MOV. While it is not feasible to simulate a line break at the MOV, the test is performed with the steam generators at 0 psig. The valves are cycled one at a time.
HIGH PRESSURE SAFETY INJECTION SYSTEM Equipment ID:
M0-3007, 3009, 3011, 3013, 3062, 3064, 3066, 3068 TEST BASIS:
The two worst case situations for operation of the above listed MOV's would be either when a valve is required to be reopened after inadvertent closure or closed under high flow conditions due to a break downstream. In the case where the valve is required to be reopened, the maximum differential pressure would be with the pump operating on miniflow and all HPSI MOV's closed. This would give approximately 1235 psig upstream of the valve, HPSI shutoff head minus elevation difference. On the downstream side of the valve the pressure would be the head of water between the valve (572' elevation) and the PCS piping.
This head of water would exist in a post-accident situation or the piping would not be intact and the valve would not be reopened.
When the PCS is on shutdown cooling with the hotleg drained to 617' 6" and LTOP requirements met we have the same water level conditions we would have in a post-accident situation. In fact the test conditions exceed post-accident conditions because the PCS would be at some pressure higher than atmosphere reducing the t.P. In the above referenced letter it was proposed to test these valves using a charging pump to a pressure of 1245 psig. Since T-249 tests the valve under a.ctual post-accident conditions there is no need to test at higher pressures. Also this test will simulate the actual conditions of opening through the entire stroke where as with the charging pumps, do to their limited flow capacity, would offer no resistance to opening once the valve is cracked open. The charging pump also does not allow for testing the valve in its design direction to isolated flow to one. of the PCS cold legs in the event the break is there. Test T-249 will test the valves under worst case closing conditions to verify proper operation in both the open and closed directions.
MI0687-2096D-MA05
4
M0-3082 and M0-3083 TEST BASIS:
M0-3082 and M0-3083 would operate approximately five hours after an accident.
These valves are interlocked with M0-3080 and 3081 respectively and would only be operated with one HPSI (RHPSI) open. Therefore test T-249, with the system conditions specified, will test these valve under conditions at least as severe as would exist in a post-accident situation. With one HPSI or PHPSI valve open the pump discharge pressure should be approximately 950 psig. Downstream of the valve to the PCS piping. This head of water would exist in a post-accident situation or the piping would not be intact and the valve ~ould not be opened.
When the PCS is on shutdown cooling with the hotleg drained to 617 1 6" and LTOP requirements met we have the same water level conditions we would have in a post-accident situation. In fact the test conditions exceed post-accident conditions because the PCS would be at some pressure higher than atmosphere reducing the ~P. In the above referenced letter it was proposed to test these valves using a charging pump to a pressure of 1245 psig. Since T-249 tests the valve under actual post-accident conditions there is no need to test at higher pressures. Also this test will simulate the actual conditions of opening through the entire stroke where as with the charging pumps, due to their limited flow capacity, would offer no resistance to opening once the valve is cracked open. This test has the added advantage of testing the ability of these valves to close under full flow condition in the event the valve is required to close.
Eguipment ID:
M0-3080, M0-3081 TEST BASIS:
M0-3080 and M0-3081 would operate approximately five hours after an accident.
These valves are interlocked with M0-3082 and 3083 respectively and would only be operated with one HPSi (RHPSI) open. Therefore test *T-249 with the system conditions specified will test these valves under conditions at least as severe as would exist in a post-accident situation. As indicated in WJA86*007 the ~p that these *valves would see in either the open or closed direction is small and therefore should not require testing. Since the valves are interlocked with M0-3082 and M0-3083 and will be moved during testing they.will be checked for proper operation in the open and closed direction *
RGURE 1 TYPICAL STEM THRUST AND CONTROL SWITCH ACTUATION SIGNATURES
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