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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML20212E9031999-09-30030 September 1999 FPC Crystal River Unit 3 Plant Reference Simulator Four Year Simulator Certification Rept Sept 1995-Sept 1999 ML20211L1321999-08-31031 August 1999 EAL Basis Document ML20212C1501999-08-31031 August 1999 Non-proprietary Version of Rev 0 to Crystal River Unit 3 Enhanced Spent Fuel Storage Engineering Input to LAR Number 239 ML20209F5601999-07-31031 July 1999 EAL Basis Document, for Jul 1999 ML20155F4071998-10-31031 October 1998 Rev 2 to Pressure/Temp Limits Rept ML20236W6501998-07-31031 July 1998 Emergency Action Level Basis Document ML20236V8801998-07-30030 July 1998 Control Room Habitability Rept ML20217B1731998-04-16016 April 1998 FPC Crystal River,Unit 3,Tendon Surveillance Program Engineering Evaluation of Sixth Tendon Surveillance ML20203K1521998-02-28028 February 1998 Post-LOCA Boron Concentration Mgt for CR-3 ML20203K4991998-02-16016 February 1998 Boron Dilution by RCS Hot Leg Injection ML20202J4291998-02-13013 February 1998 Rev 2 to MPR-1887, Crystal River 3 Reactor Bldg Cooling Fan Logic Mod Failure Modes & Effects Analysis ML20198J8181998-01-10010 January 1998 Assessing Performance & Organizational Roles ML20198J8371998-01-10010 January 1998 Engineering ML20198J8621998-01-10010 January 1998 Maintenance ML20198J8821998-01-10010 January 1998 Licensing ML20198J7501998-01-10010 January 1998 Corrective Action Program Overview ML20198J7811998-01-10010 January 1998 Design & Licensing Basis ML20198J8041998-01-10010 January 1998 Integrated Verification Team Rept ML20198J7291998-01-10010 January 1998 Overall Restart Readiness ML20198J8571998-01-10010 January 1998 Operations ML20198J8881998-01-10010 January 1998 Training ML20198H5071997-12-31031 December 1997 Rev 5 to Justification for Continued Operation for CR Emergency Ventilation Sys & Control Complex Habitability Envelope ML20197B2911997-12-11011 December 1997 Generic Operability Evaluation for Large Bore Safety Related Piping at Crystal River-3 ML20198K4451997-11-30030 November 1997 Reactor Bldg Cooling Fan Logic Mod Failure Modes & Effects Analysis, Rev 0 ML20202D4161997-11-26026 November 1997 Mod Outage Integrated Verification Team Rept ML20199K4731997-11-24024 November 1997 Sys Readiness Review Summary Rept ML20212F4191997-10-31031 October 1997 Revised Boron Dilution by RCS Hot Leg Injection ML20199D5751997-10-30030 October 1997 Overview of Safety Related Large Bore Piping & Piping Support Design & Construction Currently Existing at Crystal River-3 Nuclear Power Plant, Rev 0 ML20198T1051997-10-25025 October 1997 Tracer Gas Air Inleakage Measurements within Crystal River Unit 3 Control Complex Habitability Envelope, Summary Rept ML20211N1771997-10-11011 October 1997 Rev 1 to Pressure/Temp Limits Rept ML20210T9421997-08-29029 August 1997 B Dilution by Hot Leg Injection ML20148K7961997-06-12012 June 1997 Safety Analysis Input to Startup Team Safety Assessment Rept ML20148J3821997-05-23023 May 1997 Rev 16 to PSA F Annunciator Response ML20141D9401997-04-16016 April 1997 Summary Document for Pressurizer NDE Development for FPC Crystal River Unit 3 ML20132F4861996-12-0303 December 1996 Restart Panel ML20133B2711996-10-31031 October 1996 Monthly Trend Rept for Oct 1996 ML20199D5431996-10-28028 October 1996 Evaluation of Piping & Support Documentation for Crystal River 3 ML20134K9581996-10-15015 October 1996 Independent Design Review Panel Rept on 961015 ML20117C6431996-07-25025 July 1996 FPC Crngp Unit 3 Graded Approach Methodology for Instrument Uncertainty ML20101F6081996-02-29029 February 1996 Iga & Wear Voltage Correlations & Uncertainty Analysis ML20100G4141996-02-29029 February 1996 Input to Items a & C of NRC Questions on Relief Request for Insp of Transition Piece to Bottom Head Weld at Crystal River Unit 3 ML20100G4281996-02-0101 February 1996 Flaw Acceptance Handbook for Crystal River Unit 3 RPV & Nozzle Weld Insps ML20095L1461996-01-0202 January 1996 Seismic Evaluation Rept for USI A-46, Rev 0 ML20098A4571995-09-19019 September 1995 Plant Ref Simulator Four Yr Simulator Certification Rept ML20101F6001995-06-30030 June 1995 Alternate Disposition Strategy for Low Vol OTSG Eddy Current Indications ML20092M2061995-06-27027 June 1995 Pyrolysis Gas Chromatography Analysis of 3 Thermo-Lag Fire Barrier Samples ML20084L6931995-05-31031 May 1995 Exam of Crystal River-3 Pulled SG Tubes Final Rept B51956, 8R/9R Bobbin Voltage (S/N) Growth Rate Calculations1995-05-31031 May 1995 8R/9R Bobbin Voltage (S/N) Growth Rate Calculations ML20092M2051995-05-11011 May 1995 Pyrolysis Gas Chromatography Analysis of 5 Thermo-Lag Fire Barrier Samples ML20080T4661995-02-27027 February 1995 Final Rept One & Three H Fire Endurance Tests & Hose Stream Tests Thermal Barrier Sys for Electrical Components 1999-09-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G0191999-10-15015 October 1999 Safety Evaluation Concluding That Licensee Followed Analytical Methods Provided in GL 90-05.Grants Relief Until Next Refueling Outage,Scheduled to Start on 991001.Temporary non-Code Repair Must Then Be Replaced with Code Repair 3F1099-19, Part 21 Rept Re Damage on safety-grade Cable Provided to FPC by Bicc Brand-Rex Co.Damage Was Created During Cabling Process While Combining Three Conducters.Corrective Action Program Precursor Card PC99-2868 Was Initiated1999-10-13013 October 1999 Part 21 Rept Re Damage on safety-grade Cable Provided to FPC by Bicc Brand-Rex Co.Damage Was Created During Cabling Process While Combining Three Conducters.Corrective Action Program Precursor Card PC99-2868 Was Initiated ML20217B0931999-10-0606 October 1999 Part 21 Rept Re Damaged Safety Grade Electrical Cabling Found in Supply on 990831.Damage Created During Cabling Process While Combining Three Conductors Just Prior to Closing.Vendor Notified of Reporting of Issue ML20212L0881999-10-0404 October 1999 SER Accepting Licensee Requests for Relief 98-012 to 98-018 Related to Implementation of Subsections IWE & Iwl of ASME Section XI for Containment Insp for Crystal River Unit 3 ML20212J8631999-10-0101 October 1999 Safety Evaluation Supporting Licensee Proposed Alternatives to Provide Reasonable Assurance of Structural Integrity of Subject Welds & Provide Acceptable Level of Quality & Safety.Relief Granted Per 10CFR50.55a(g)(6)(i) ML20212E9031999-09-30030 September 1999 FPC Crystal River Unit 3 Plant Reference Simulator Four Year Simulator Certification Rept Sept 1995-Sept 1999 3F1099-02, Monthly Operating Rept for Sept 1999 for Crystal River,Unit 3.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Crystal River,Unit 3.With ML20212E6911999-09-21021 September 1999 Safety Evaluation Supporting Proposed EALs Changes for Plant Unit 3.Changes Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 ML20211L1321999-08-31031 August 1999 EAL Basis Document 3F0999-02, Monthly Operating Rept for Aug 1999 for Crystal River,Unit 3.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Crystal River,Unit 3.With ML20212C1501999-08-31031 August 1999 Non-proprietary Version of Rev 0 to Crystal River Unit 3 Enhanced Spent Fuel Storage Engineering Input to LAR Number 239 ML20211B7291999-08-16016 August 1999 Rev 2 to Cycle 11 Colr ML20210P1111999-08-0505 August 1999 SER Accepting Evaluation of Third 10-year Interval Inservice Insp Program Requests for Relief for Plant,Unit 3 ML20210U5341999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Crystal River,Unit 3 ML20209F5601999-07-31031 July 1999 EAL Basis Document, for Jul 1999 3F0799-01, Monthly Operating Rept for June 1999 for Crystal River,Unit 3.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Crystal River,Unit 3.With ML20210U5411999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Crystal River,Unit 3 3F0699-07, Monthly Operating Rept for May 1999 for Crystal River,Unit 3.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Crystal River,Unit 3.With ML20210U5601999-05-31031 May 1999 Revised Monthly Operating Rept for May 1999 for Crystal River,Unit 3 ML20195C6271999-05-28028 May 1999 Non-proprietary Rev 0 to Addendum to Topical Rept BAW-2346P, CR-3 Plant Specific MSLB Leak Rates ML20196L2031999-05-19019 May 1999 Non-proprietary Rev 0 to BAW-2346NP, Alternate Repair Criteria for Tube End Cracking in Tube-to-Tubesheet Roll Joint of Once-Through Sgs 3F0599-04, Monthly Operating Rept for Apr 1999 for Crystal River Unit 3.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Crystal River Unit 3.With ML20210U5631999-04-30030 April 1999 Revised Monthly Operating Rept for Apr 1999 for Crystal River,Unit 3 3F0499-04, Monthly Operating Rept for Mar 1999 for Crystal River Unit 3.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Crystal River Unit 3.With ML20204D9661999-03-31031 March 1999 Non-proprietary Rev 1,Addendum a to BAW-2342, OTSG Repair Roll Qualification Rept 3F0399-04, Special Rept 99-01:on 990310,discovered Containment Tendons That Required Grease Addition in Excess of Prescribed Limits During Recent Insp Activites.Six Tendons Were Refilled with Appropriate Amount of Grease1999-03-10010 March 1999 Special Rept 99-01:on 990310,discovered Containment Tendons That Required Grease Addition in Excess of Prescribed Limits During Recent Insp Activites.Six Tendons Were Refilled with Appropriate Amount of Grease 3F0399-03, Monthly Operating Rept for Feb 1999 for Crystal River Unit 3.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Crystal River Unit 3.With ML20203A4381999-02-0303 February 1999 Safety Evaluation Supporting EAL Changes for License DPR-72, Per 10CFR50.47(b)(4) & App E to 10CFR50 ML20206E9891998-12-31031 December 1998 Kissimmee Utility Authority 1998 Annual Rept ML20206E9021998-12-31031 December 1998 Florida Progress Corp 1998 Annual Rept ML20206E9701998-12-31031 December 1998 Ouc 1998 Annual Rept. with Financial Statements from Seminole Electric Cooperative,Inc 3F0199-05, Monthly Operating Rept for Dec 1998 for Crystal River Unit 3.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Crystal River Unit 3.With ML20206E9261998-12-31031 December 1998 Gainesville Regional Utilities 1998 Annual Rept 3F1298-13, Monthly Operating Rept for Nov 1998 for Crystal River,Unit 3.With1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Crystal River,Unit 3.With 3F1198-05, Monthly Operating Rept for Oct 1998 for Crystal River,Unit 3.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Crystal River,Unit 3.With ML20155F4071998-10-31031 October 1998 Rev 2 to Pressure/Temp Limits Rept ML20155J2701998-10-28028 October 1998 Second Ten-Year Insp Interval Closeout Summary Rept 3F1098-06, Monthly Operating Rept for Sept 1998 for Crystal River Unit 3.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Crystal River Unit 3.With ML20206E9461998-09-30030 September 1998 Utilities Commission City of New Smyrna Beach,Fl Comprehensive Annual Financial Rept Sept 30,1998 & 1997 ML20206E9561998-09-30030 September 1998 City of Ocala Comprehensive Annual Financial Rept for Yr Ended 980930 ML20206E9101998-09-30030 September 1998 City of Bushnell Fl Comprehensive Annual Financial Rept for Fiscal Yr Ended 980930 ML20206E9811998-09-30030 September 1998 City of Tallahassee,Fl Comprehensive Annual Financial Rept for Yr Ended 980930 ML20195E3121998-09-30030 September 1998 Comprehensive Annual Financial Rept for City of Leesburg,Fl Fiscal Yr Ended 980930 3F0998-07, Monthly Operating Rept for Aug 1998 for Crystal River Unit 3.With1998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Crystal River Unit 3.With ML20236W6501998-07-31031 July 1998 Emergency Action Level Basis Document 3F0898-02, Monthly Operating Rept for Jul 1998 for Crystal River,Unit 11998-07-31031 July 1998 Monthly Operating Rept for Jul 1998 for Crystal River,Unit 1 ML20236V8801998-07-30030 July 1998 Control Room Habitability Rept 3F0798-01, Monthly Operating Rept for June 1998 for Crystal River Unit 31998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Crystal River Unit 3 ML20236Q4611998-06-30030 June 1998 SER for Crystal River Power Station,Unit 3,individual Plant Exam (Ipe).Concludes That Plant IPE Complete Re Info Requested by GL 88-20 & IPE Results Reasonable Given Plant Design,Operation & History 3F0698-02, Monthly Operating Rept for May 1998 for Crystal River Unit 31998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Crystal River Unit 3 1999-09-30
[Table view] |
Text
4 i
EVALUATION OF REACTIVITY RESPONSE FOR A STEAM LINE BREAK EVENT WITH UNTERMINATED EMCRGENCY FEEDWATER FLOW PREPARED FOR FLORIDA POWER CORPORATION CR-3 PROJECT CONTRACT #5B2-7087 TASK #184 MAY 6, 1981 l
B&W REFERENCE DOCUMENT: 86-1125549-00 PREPARED BY: //64.orArr/
REVIEWED BY: Mdad REVIEWED BY: b APPROVED BY: O. b 1 e
8109090164 810903 PDR ADOCK 05000302 P PDR
I CONTENTS g
, PAGE I. Background 1 II. Scope 2 III. Method 3 IV. Results. 4 l
V. Conclusions 5 PAGE ,
Table 1 Major Assumptions and Input Parameters 6-7 l
I Table 2 Steam Line Break Sequence of Events 8 Table 3 Steam Line Break Results 9 l
l l
l' i
Figure 1 Reactor Power /RC Temperature 10-11 Figure 2 Core Total Reactivity vs. Time 12 l
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. l 6
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ABSTRACT To reduce the patential for a loss-of-all-feedwater event, it has been
~
recommended that the Steam Line Rupture Matrix Signal which isolates emergency feedwater be eliminated. A previous report, " Evaluation of Steam Line Break Consequences Associated with Removal of Rupture Matrix I
Signals from Emergency Feedwater Valves" addressed the effects of this I change on the FSAR analysis, but did not specifically reanalyze the reactivity effects associated with a Steam Line Break (SLB). This report provides analyses, using current methods, which demonstrate that a return h
to criticality will not occur following a SLB and unterminated Emergency ;
Feedwater(EFW) flow. .
E e
d
I. Background _
Following the CR-3 transient in February,1980. Florida Power Corporation convened a Nuclea
- Safety Task Force to identify potential changes which could significantly improve plant safety and reduce the potential for major transients. One of the principal reconnendations of this task force (and NUREG-0667) was that the Steam Line Rupture Matrix System i (SLRMS) signals which close the emergency feedwater isolation valves be eliminated. Such a change would significantly reduce the potential for a loss-of-all-feedwater event and this signit'icantly improve the overall safety of the plant.
In support of this change, however, it was necessary to address the effects on the FSAR steam line break analysis which had been perfonned assuming automatic isolation of the EFW by the SLRMS. A report entitled, " Evaluation of Steam Line Break Consequences Associated with Removal of Rupture Matrix Signals from Emergency Feedwater Valves", dated May 23, 1980 and revised l
l on June 27, 1980 was prepared for that purpose. Using the same methods as i were originally used in the FSAR, the analysis described in that report
! demonstrated the acceptability of continued EFW flow from the standpoint of I containment response. However, the method originally used for reactivity response to an SLB (the SECRUP analog code) was no longer in use or available.
In lieu of the reanalysis, the report included an evaluation showing that the probability of a SLB accompanied by any stuck control rod could be conservatively estimated to be less than 2.75 x 10 -7 per reactor year.
However, the NRC staff rejected this evaluation, and, as a consequence, the proposed change has not been made.
9
~
2 Florida Power Corporation has subsequently inst.ituted a major program to upgrade the emergency feedwater system and associated controls. The new control system, the Emergency Feedwater Initiation and Control (EFIC) system will replace the existing SLRMS and provide emergency f eedwater
! isolation to only the steam ge:v rator in the affected loop. However, pending installation of the EFIL system, removal of the rupture matrix signals to the EFW valves is still a very desirable change.
1
- This report is intended to address the NRC concern associated with reactivity effects following a SLB with continued EFW flow. For this l
l analysis, current methods (specifically the TRAP code described in topical report BAW-10128) are used in lieu of the methods employed in the original FSAR analysis.
l l
l II. Scope l
The purpose of this analysis is to document the core total reactivity response to a large double-ended steam line break with unteminated l emergency feedwater flow to both steam generators. As a result of the l proposed removal of the Steam Line Break Rupture Matrix signals from the emergency feedwater valves, it is possible to have EFW flow to both the l
affected and unaffected steam generator loops. This situation requires
! operator action to recognize and isolate EFW from the unaffected steam generator loop. This analysis predicts the system response, specifically core reactivity, for the SLB event where this operator action has not been performed. Core reactivity during the transient is detemined by the insertion of control rods, the insertion of boron by the HPI system I
3 and the reactivity feedback due to fuel temperature and moderator density changes. Of special interest is the continued cooling provided by EFW. It is desired that core suberiticality be maintained throughout the transient.
III. Method The Double-Ended Steam Line Break event was analyzed using the TRAP 2 (version 6) digital computer code. Major assumptions and input parameters are provided in Table 1. The transient examined is a double-ended rupture of a single main steam line. The analysis assumptions were chosen to provide a conservative response with respect to core overcooling thus maximizing the core reactivity response and thus increasing the potential for recriticality. For this reason, a conservatively large secondary inventory, end of cycle kinetics parameters, and appropriate response
, delay times were used. No loss of offsite power was assumed. Main steam isolation is assumed to occur by closure of the main steam isolation valves rather than the turbine stop valves, thus allowing additional overcooling.
Control rod reactivity insertion is assumed to provide sufficient reactivity to account for power deficit and only a 1% shutdown margin at HZP conditions.
Shutdown reactivity provided by boron in the HPI system is assumed to be supplied by only 1 HPI pump. Positive reactivity feedback provided by the reactivity coefficients has been conservatively estimated as end of life values to maximize reactivity feedback response.
The double-ended rupture of a 22 inch (ID) main steam line results in a rapid increase in steam flow, secondary depressurization and an increase
4 in heat transfer across the steam generator. Overcooling of the primary pressure results in a low RC pressure trip at 1.1 seconds into the transient. Reactor trip initiates turbine trip and TSV closure, but this function was ignored in the analyses to provide conservative overcooling.
The secondary steam pressure continues to decrease actuating a Steam Line Rupture; Matrix signal (600 psia SG pressure) at 4.7 seconds into the transient. This signal initiates closure of the Main Steam Isolation valves and main feedwater isolation. EFW is subsequently initiated as a result of the loss of all main feedwater. The steam generator in the affected loop continues to depressurize and boil dry at a rate faster than the steam generator in the unaffected loop. Primary system pressure decreases to the HPI actuation signal at 5.3 seconds, thus starting HPI pumps and the addition of borated water to the primary system.
l Core reactivity becomes negative innediately following reactor trip
- providing at least a 1% suberitical margin. This margin is diminished by the positive reactivity feedback caused by the decrease in core average temperature due to the overcooling. Increased negative reactivity is l provided by the introduction of borated water by the HPI system. This
!~
negative reactivity competes with the positive reactivity caused by prolonged cooling by EFW injection. The borated water surolied by one HPI pump provides sufficient negative reactivity to overcome continued EFW flow, thus eventually resulting in increasing subcritical margin. A minimum suberitical margin of .10% ak/k occurs 16 seconds into the transient followed by an increasing j suberitical margin.
III. Results The resulting reactivity response shodn in Figure 2 indicates that the
- q j 5 core will remain subcritical ti.roughoyt the transient with a minimum suberitical margin of <10'; ak/k occurring at 16 seconds. Reactor power and reactor coolant _ pressure are shown in Figure 1. A sequence of events i is provided in Table 2 and sumary of pertinent results in Table 3.
l The analysis indicates that removal of the Steam Line Rupture Matrix signal to the EFW valves will not result in sufficient overcooling to cause l core recriticality with credit being assumed for only one HPI pump. Operator I action would normally be expectett to occur to terminate EFW flow to the affected steam generator loop, thus allowing it to boil completely dry and tenninate its contribution to the avercooling.
IV. Conclusions l Since subcriticality can be maintained even for the conservative assumptions presented in this analysis, it is concluded that removal of the SLBRM signals I, to the EFW valves does not result in an unacceptable core reactivity response.
l 4
8
6 TABLE 1 MAJOR ASSUMPTIONS AND INPUT PARAMETERS p
A. Thermal Hydraulics Parameter Value >
- 1. Power level (102%), MWt 2619
- 2. RC pump heat, MWt 18 8
- 3. Primary Flow rate,1bm/hr 1.3 x 10 7
- 4. Secondary flow rate, lbr.1/hr 1.06 x 10
- 5. SG outlet pressure psia 925
- 6. Steam generator inventory,1bm 46200 lbm 3
- 7. Initial pressurizer inventory ft 800
! 8. EFW flow rate 1
( affected loop gpm 880 l, unsffected-lo'op gpm 520 EFW temperatura, F 40 i
l l
l l B. Kinetics Parameters Parameter Value Doppler coefficient ak/k/F -1.3 x 10-5 Moderator coefficient ak/k/F -3.0 x 10 -5 Boron Worth ppm /% ak/k 108 Shutdown margin, % ak/k 1.0 f
,.y, .-.
. . -_ _ .=
1; 7
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l C. Trips a6d Setpoint Times
( Parameter Value
! 1. Low pressure RPS trip, psig 1800 l 2. HPI actuation setpoint, RC pressure psig 1500
- 3. SLB rupture matrix signal psig 600 t
- a. MSIV closure, delay, s. 2.5
[ stroki , s. 5.0
- b. Main feedwater isolation, s. 17
- 4. EFW actuation delay (sec.) 50 E-t
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J TABLE 2~
STEAM LINE BREAK SEQUENCE OF EVENTS
(
Event Time - (seconds) l l Rupture occurs 0.0 Reactor trip on low primary pressure 1.1 Control rods begin to fall 1.5 Low steam generator pressure occurs 4.7 HPI actuation signal reached 5.3 MSIV's closed 12.2 HPI flow established 15.3 Minimum subcritical margin reached 16.
l Main feedwater isolated 24.2 EFW flow establisheo 50.
9 l
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! 9 TABLE'3 STEAM LINE BREAK RESULTS l
i
( ' Minimum suberitical margin, %Ak/k .10 1
I Time of minimum suberitical margin, s. 16 Y
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