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- 2 li l BUDGET 128006 1 la TECHNICAL DATA REPQ j NJ3 [ p c -ACTIVITYNO. PAGE' OF PROJECT: TMI-1 g- i t ii Safety Anal / Plant ::ont DEPgTjMENTISECTION DOCUMENT TITLE:
__DATE REVISION DATE 2[U SG Tube' Rupture Procedure Guidelines ORIGINATOR SIGNATURE DATE APPROVAL!S) SIGNATURE DATE l
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APPROVAL FOR EXTERNAL DISTRIBUTION l DATE
.:rr n p ~ u.t.n o DISTRIBUTION ABSTRACT: b R.' W. Bensel D. J. Boltz This document provides technical guidelines for dealing T. G. Broughton with single and multiple tube ruptures. A significani
- improvement in procedures will result from reduction of M. Campagna P. R. Clark the minimum subcooling margin and RC pump crip on loss
- of subcooling margin, walver of fuel-in-compression J. J. Colitz limits, and revised RCP NPSH limits. Other benefits can I. R. Finfrock be derived from revision of the RC pump restart criteria I. L. Gerhar and Irom additional guidance regarding OTSG steaming and R. J. Glavlano isolation. Fina11f, revised guidance is provided for
- R. W. Keaten preventing tube ' leak propagation. It is recommended O. Lehmann
- D. T. Leighton that the tube-to-shell delta T be limited to 70F0 during tube rupture events.
B. Leonard
- W. W. Love
- Revision 2 to this IDR includes the following J. G. Miller
- recommendations for procedural revisions, some T. Moran
- of which have already been incorporated in .
M. Nelson EP 1202-5.
- S. Newton
- M. J. Ross **
- 1. Isolate the DISG's on a measured or projected
- E. B. Shipman D. G. Slear dose rate of 50 mrem /hr whole boJr or 250 mrem /hr thyroid dose.
C. W. Smith
- M. J. Stromberg *
- 2. Stop the non-ES HP1 pump if the RCS is cooling R. J. Toole more than 100F/hr.
P. S. Walsh
- Dr. R. N. Whitesi i* 3. Priorities should be spelled out in EP-1202-5:
- R. E. Wilson a. Minimizing SCM has a priority over mini-mizing cooldown time.
DETJEN84-897 PDR
- COVER PAGE ONI.Y accetet0 7.il
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- c. TDR 406
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ABSIRACT (Cont' A)
- 4. Initiate the DER system at 300F under tube rupture conditions.
- 5. Trip the reactor if pressurizar level cannot be maintained above 200 inches with two HPI pumps on.
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- 6. Raise the unaffected OTSG 1evel to 95%:. before
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raising the affected OISG 1evel to 95%.unless .
incore temperatures are not decreasing and there is no OISG heat transfer.
- 7. If RCPs are not tripped within two minutes of a l loss of SCM, maintain 1 RCP in each loop running.
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SG ISSE 10PIURE PROCEDJ RE GJIDELIKE5 DR #406
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') 283e 4 of 74 IABLE OF CONIENTS
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- Page Iable of Contents . . . . . . . . . . . . . . . . . . 4
' List of Figures . . . . . . . . . . . . . . . . . . . 7 List of Iables ................. ..
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Summary of Changes .................
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- 1. 0 INIRODUCIION AND BACKGROUND . . . . . .-. . . . . . . 13 2.0 TECE FUNCTIONS SGTR PROCEDURE GUIDELINES DEVEIDPMENI PROGRAM . . . . . . . . . . . . . . . . . 14 l 2.1 Development of Design Basis Guidelines . . . . . 14 2.1.1 . Literature Search .......... 14 2.1.2 Limiting OISG Iube Stresses ..... 16 ,
2.1.3 Steaming.: Isolation and filling of ,
the Leaking OTSG .......... 17 l 2.1.3.1 Steaming, Isolation and Filling with- -
both OISG's Leaching . . . . . . . . . 19 2.1.4 Min h = Allowable Subcooling Margin . 19 2.1.5 Walve Puel in Compression Limits . . . 20 T. l. 6 Reactor Coolant Pump NPSE Limits . . . 20 2.1.7 Procedure Entry Point Condition ... 20 2.1.8 Simulator Experience . . . ... . . . . 21 2.1.9 -Emergency Limits for Decay Beat Sfstem Initiation ... ....... 21 2.1.10 Reactor Irip on Low Pressurizer Level. 21
- 2. 2 Development of Multiple Tube Rupture Procedure Guidelines . . . . . . . . . . . . . .. 22 2.2.1 Revision of RCP Irip and Restart ,
Criteria . . . . . . . . . . ..... 22 2.2.2 OISG Steaming and Level Control ... 23 2.2.3 Criteria for Feed and Bleed Cooling . 24 2.2.4 Cooldown/Depressurization ...... 24 ,
- 2.3 Additional Work Requirements . . . . . . . . . . 26
- 2.3.1 Analyses . . . . . . . . . . . . . . . 26
- - 2.3.2 Issue Resolution . . . . . . . . . . . 26
. 3.0 ' MAJOR REVISIONS 10 EXISTING PROCEDURE ........ 27 3.1 Sasic Plant State. . . . . . . . . . . . . . . . 27 3.1.1 Assumed Plant Conditions . . . . . . . 27 3.1.2 Tube Rupture Guidelines for Loss of Subcooling ............ 27 3.1.3 Revised Equipment Limits & Operating 1 Procedures . . . . . . . . . . . . . . 28 i-
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Rev. 2 f Page 3 of 74 IABLL OF CONIENI3 (Cont'd)
Pazd 3.2.' Discussion of Guidelines . . . . . . . . . . . . 29 3.2.1 Immediate Actions . . .. . . . . .. 29 3.2.2 Followup Actions - Subcooling Maintained and RCP'h Available . . . . 29 3.2.2.1 . Maintain a Minimum of 25F*
Subcooling Margin . . . . . . . .. . 29 '
3.2.2.2 Steaming / Isolation Criteria for the
.Affected 0TSG . .. . . . . . . . . . 29 3.2.2.3 Shell to Tube Delta I .. . . . . . . .. 30 3.2.3 Followup Actions (Automatic Ax Trip has Occurred) . . . . . . . . . . . . . 30 3.2.3.1 Trip with a Loss of Subcooling Margin. 30 3.2.4 Followup Actions for Loss of Subcooling . . . . . . . . . . . . . 30 4.0 *SIMULAIOR IRAINING EXPERIENCE . . . . . . . . . . .. 32 4.1 ' Introduction . . . . . . . . ... . . . .'. . . . 32
- 4.2 Results. . . . . . . . . . . . . . . . . . . . . 32 l.
- ' 4.2.1 January 1983 Training .. ... . . .. 32 .
- . 4.2.1.1 Comments . . . . ... . . . . . . . . . 33
- *- 4.2.2 June 1963 Training . . . . . . . . . . 33
.. 4.2.2.1 Control of RCS Cooldown Race . . . . . 33
- 4. 2. 2. 2 Plant Stabilization Before Cooldown . 34
- 4~. 2. 2. 3 RCP Restart Criteria . . . .-. . . . . 35
_4.2.2.4 Core Flood Tank Isolation . . . . .. 35
- 4.2.2.5 RCP Irip Criterion . . . . . . . . . . 35
- 5.0
- SINGLE AND MULIIPLE SGTR GUIDELINES . . . . . . . .. 36
- - 5.1 Scope. . . . . . . . . . . . . . . . . . . . . . 36
- 5.2 Guidelines and Limits. . . . . . . . . . . . . .. 36
- 5.2.1 Subcooling Margin Requirements . . . . 36
- 5.2.2 Reactor Coolant Pump Trip critericrn . 36
-* 5.2.3 -Reactor Coolant Pump. Restart Criteria. 36
- - 3. 2. 4 Reactor Coolant Pump NPSE for
-* Emergency Operations . . . . . . . . . 36
- 5.2.5 Eigh Pressure Injection Throttling Criteria . . . . . . . ... . . . . . .
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36
- 5. 2. 6
, OIS G Level . . . . . . . . . . . . . . 37
- 5.2.7 OISG Isolation / Steaming Criteria . .. 37 5.2.7.1. Pressure Control of an Isolated OTSG . 37
- .* 5.2.8 Cooldown Race During a Tube Leak
- _. Event . . .. . . . . . . . . . . . . 37
- 5.2.9 OISG Shell-to-Tube Differential
- Iemperature Limit
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- 5.2.10 Cooling Mode dhan Both OISG' h are Unavailable for RCS Haac Removal . . . 38
- - 5.2.11 Core flood Tank Isolation . . . . .. 38
- 5.2.12 Guideline Flow Chart . . . . . . . . .
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Rsv. 2 Page 6 of 71.
l TABLE OF CONIENTS (Cont'd), .
Page 6.0 -CONCLUSIONS AND AECOMMENDATIONS . . . . . . . . . . . 39
- 7. 0 REFERENCES ... . . . . . .. . . . . . . . . . .. 42
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Ap'pendix A: ' Comparison of Guidelines to INFO and ARC f i
l Recommendations l Appendix B: Procedure Change Safety Evaluations Appendix C: Guidelines Flow Chart Appendix D: . Simplified Event Tree Appendix E: Process Computer Output l
-There are a total of 74 pages in this report including Figures, Taoles and Appendices. '
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TDA 406 !
Las. 2 Rage 7 of 74 LIST CF IIGJRE3
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Title Fi2ure No.*
Steam Generator Tube Rupture Guideline Development Activit/ No: work . . . . . . . . . . . . . 1 Break Flow for Single Ruptured Iube . . . . . . . . . . . . 2 i
i Effect of A0 Pump Operation on
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Integrated Systes Leakage for Single Ruptured Tube . . . . . . . . . . . . . . . . . 3
- Mass and Energy Capabilities of EPI and PORV . . . . . . . . . . . . . . . . . . . . . 4 Time Behavior of Subcooling Margin for Spectrum of Ruptured Tubes . . . . . . . . . . . . . . 5 Emergency Reactor Coolant Pump NPSE Limits . . . . . . . . . . . . . . . . . . . . . . 6 Single and Multiple Iube Rupture '
, Guidelines. . . . . . . . . . . . . . . . . . . . . . . C-1 Simplified OTSG Event free . . . . . . . . . . . . . . . . . .
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. \.- Page LIST OF.IABLES .
Title Table No. -
. Tabular Values of ECP Emergency NPSE
. Requirtaants .-. . . . . . . . . . . . . . . . . . . . 1
- Pressurizar Spray Flow for Various Pump Combinations . . . . . . . . . . . . . . . . . . . . . 2
- Comparison of TD1 #406 to
.Other Sources. . . . . . . . . . . . . . . . . . . . .
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Shall Iharnocouple Substitution . . . . . . . . . . . . . . . . . . . . . E.1.4.1 +
Wide Range-I cold Input . . . . .............. E.1. 4. 2 +
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'Q p DOCUMENT NQ.
airLE SG Tube Rupture Procedure Guidelines, Rev. 2 Page"9 of 74 REV
SUMMARY
OF CHANGE APPROVAL DATE 2 Added detail to Table of Contents and reversee the order of. Sections 4.0 and 5.0. /N f-82.-U 2 Added Tables 1 in 2 which provide tabular daca on -
RCP NPSH requirements and on spraf flow for various RCP pump combinations.
IC J-/Eh 2 Section 2.1.3. Added recommandation for OTSG isolation if Lodine release race exceeds E ~O 250 mrem /hr or whole body dose rates 50 mrem /hr, correcting error in previous revision.
2 Section 2.1.6 and Figure 6. Revised emergenef NPSH limits to account for cal-culated instrument errors during LOCA .
g(f [-lL-[3 conditions. ; ;
2 Sections 2.1.8 and'4.2.2. Added dis-
- cussion on the experience gained from the y-r2.-73 June 1983 Lynchburg simulator sessiohs. [C/ .
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2 Section 2.1.9. Added recommendation to allow DER sfstem initiation at 300F instead of 275F.
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2 Section 2.1.10 Recommendation to trip I reactor if 200 inch pressurizer level can- NO) A- N 2 -D not be maintained with two HPI pumps i running.
1 2 Section 2.2.2. Clarified guidance on ;
1 solation criterion with leaks in both 0 d l Y3 l OISGs.
2 Section 2.2.2. Addressed raising OISG 1evel to 95%:without causing an overcooling. [O( F U
. 2 Section 2.2.2. Discussed why EFJ shoulf not be used to control OTSG pressure in an Nb S - 72.-45 1solated CISG (changing previous recom-mandation of Rev. 1).
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2 Nuclear 406 "M SG Iube Rupture Procedure Guidelines, Rev. 2 8
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- 10 OF 74 REV
SUMMARY
OF CHANGE APPROVAL D' ATE 4 5ection 4.4..$. 7.xpandes discussions on feed and bleed cooling to include ADV and r37 mass and energy relief capabilities g }I , f, Q p j, relative to decay heat and leak race.
2 Section 2.3. Discussed additional work g, 7-f1-E2 which will be addressed in a future revision to TDR 406.
2 Section 4.2.2. Added note regarding actions 7-f 2. - Yy to be taken if RCPs are not tripped within. %e, g 2 minutes of a loss of SCM.
2 Section 4.2.3. Added note regarding loss of SCM after RCPs are restarted.
NN 8 ~'1 ~83
- 2. Section 4.2.6. Revised guidance on raisind OISG 1evels to 95%,- %d[ . 7-/1.JJ 5 2 Section 4.2.7.
250 mrem /hr.
Added isolation criterion on [8.I y-iz.pg 2 Section 4.2.11. Added criterion for Core 76I g-s.2. -73 Flood Tanks Isolation.
2 Section 6.0. Added recommendations to isolate CFTs using criterion provided and initiation Mg 3 -t2. - D of DER system at 300F.
,2 Section 2.2.1. Clarified that emergency NPSH d* 'I ~d")
curves should be followed for both RC pump trip and restart.
2 Sections 3.2.4 and 4.2.3. Start one RCP per Y C- b I tb loop or both RCP's in the same loop.
2 Section 3.2.2. Revised to be consistent with g zy
. the steaming criterion in Section 2.1.3.
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NgOE TDR NO.
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406 TITLE SG Iube Rupture Guidelines, Rev. 1
- PAGE* 11 . OF 74 REV
SUMMARY
OF CHANGE APPROVAL D' ATE i Minor editorial changes and correction of /s/ 5/8/83 typographical errors on pages: 2,5,7,10,13,' '
19,20,22,A-1,A-3,A-4,B-1,A-2,B-2,6,18. !
1 Revised cover page to show shell-to-tube /s/ 5/8/83 delta T can be controlled below 100E. .
1 Added a List of Tables pp i & 111 /s/ 5/8/83 1 Included use of MEii as means to cool OISG . /s/ 5/8/83 shell. .p 5 1 Indicated that continuous steaming of OISG is /s/ 5/8/83 simplest means of meeting.0ISG 1evel, pressure and differential temp. considerations pp. 5,6,10,18' l
1 Eliminated reference to RAC for determining /s/ 5/8/83 when to isolate OISG based on radiological conditions. p6 1 Added Section 2.1.3.1 to discuss control when /s/ 5/8/83 both OISG's are isolated. (Also p 10).
1 Provided discussion and Elgure for RCP NPSE /s/ 5/8/83 limits. Section 2.1.6 and Elgure 6. Ref 25,26. .
& Sections 4.2.1 and 4.2.4.
1 Revised explanation of ADV & IBV flow capability /s/ 5/8/83 relative to CISG flooding (incorrect in Rev 0) pp 11,16 1 Added Reference to B&J guidance which allows /s/ 5/8/83 cooldown at 1001/hr during Iube Ruptures -
without a soak time even if cooldown race is j exceeded. p 11,& Ref 24 l' i Section 4.2.3 revised to account for inab111tf /s/ 5/8/83
. to start either ACE in the A loop. ,
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i 1 Added Section 4.2.7.1 and revised 4.2.7 to /s/ 5/8/83 i
l address' Steaming isolation of GISG considering the continuous steaming philosophy.
1 Simplified Section 4.2.8 on cooldown race. /s/ 5/8/83 l' Added Section 4.2.9 on contro1Iing OISG shell-to /s/ 5/8/83 cube differential temperature. -
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406 si LE SG Iube Aupture Guidelines, Rev.1 '
PAGE ' 12 - OF
, 74 REV-
SUMMARY
OF CHANGE APPROVAL D' ATE 1 Added Section B.7,B.8, and B.9, which were lefc /s/ 5/8/83 out of Rev. 0 inadvertentif.
1 Deleted redundant Section of Part 4 (guidelines) /s/ 5/8/83 1 Added Section C.2 through C.6 to discuss the. /s/ 5/8/83 Guidelines Flow Diagram (Eigure C-1) in words.
1 Rewrote Appendix E on Process Computer output /s/ 5/8/83 and Aleras. '
1 Revised Figure 4 to show decay heat levels /s/ 5/8/83 as a function of time.
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1.0 ' INTRODUCTION AND BACKGROUND '
In November 1981, primary to secondary side leaks were discovered in the tubes of both of the TMI-1 Once Through Steam Generators (OTSG). There are 15,531 tubes in each OTSG. The plant design basis for a steam generator tube rupture (SGTR) acciden't is the double ended offset severence of a single tube. Since extensive circumferential cracking was discovered in approximately 1200 of the-31,000 tubes, it became clear that a revised set of procedures for dealing with both single and multiple SGTRs.should be developed.
This report describes a program which has been formulated to improve existing.
procedures and operator training by providing improved operator guidelines for dealing with tube leakage and tube rupture events. The guidelines development program will be described in detail, and the major revisions to the existing procedures which have been identified as part of the program will be discussed. The proposed guidelines will then be presented in terms of their overall scope, with a step by step discussion of required operator actions. The analytical evaluations which are the basis for the recommendations, consist of a series of simulations which are ongoing and will be documented in detail in a subsequent report. The guidelines in this TDR were tested at the B&W simulator training cycle beginning in January, 1983. The results of this training i experience are discussed. Finally, the overall conclusions and major reconsnendations of the guidelines development program are documented.
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. Page 14 of 74 2.0 TECH FUNCTIONS SGTR GUIDELINES DEVELOPMENT PROGRAM Figure 1 shows the execution of the steam generator e ~ : rupture guideline development program. The plan has three main paths: Path 1 is _ the development of design basis tube rupture guidelines. Path 2 is'the development of multiple tube rupture guidelines; and, Path 3, is a benchmark effort to compare the RETRAN and RELAP 5 computer codes. This last effort.also includes an evaluation of the B&W ATOG analysis of a single tube rupture using MINITRAP. The purpose of this TDR is to explain paths 1 & 2. .The benchmarking and comparison efforts are' discussed in a separate TDR describing all of the tube rupture analysis work. None of the computer analysis of Path 3 has been used to justify the recommendations of this report. The C
analyses were an aid in conceptualizing the physical processes during a tube rupture.
. 2.1 Development of Desian Basis Guidelines (Path 1) -
,. The major activities involved in developing this part of the ,
guideline were to:
- 1. Search existing industry events and procedures for lessons to be
- 1 earned about handling tube ruptures.
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- 2. Define. allowable steam generator stresses during cooldown
~ (either as cooldown rate or as tube /shell delta T).
- 3. Determine when OTSC's should be isolated and when they should be steamed.
4.- Revise the minimum allowable subcooling margin.
- 5. Waive fuel in compression limits.
- 6. Develop emergency RCP NPSR limits.
- 7. Redefine entry point conditi'ons.
- -8. Factor in experience from use of the guidelines on the B&W simulator.
, Each of these items are discussed in detail in the following sections.
2.1.1 Literature Search Several tube rupture leaks have occurred at various operating reactors within the last four years. The experience gained from these events has offered us an opportunity to improve tube rupture 4
guidelines. The major lessons learned from these events have been t
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suummarized' in various documents from the NRC', INPO, and plant
- procedures and . included in the B&W ATOG tube rupture guidelines (References 1-10). The lessons ~ include the following:
- 1. Subcooling margin should be minimized to minimize primary to secondary leakage. Subcooling is maintained by keeping the RCS temperature below the saturation temperature with OTSG cooling.
Since the OTSG is in a saturated condition, it is always lower ,
in pressure than the RCS if subcooling is maintained.
Therefore, keeping subcooling margin at or near its minimum acceptable value reduces leakage.
- In order to maintain the minimum subcooling margin, several plant limits have to be violated: fuel pin-in-compression limits and RCP NPSR limits. The former is acceptable to violate during emergency conditions, while the latter has been reevaluated to determine acceptable emergency operation of the pump.
- 2. RCP's should be maintained running for'several reasons. Pump i
trip on loss of subcooling margin allows the operator to '
- maintain-forced flow for a leak size of up to several tubes
- while 1600 psig ESAS is much more restrictive.- Porced RC flow provides several benefits during a tube rupture. First, they
- " - assure that steam voids do not form in the hot les U bends or upper vessel head. Steam voids in these locations can interrupt
-natural circulation or prevent RCS depressurisation. Second, RCP operation results in a lower primary to secondary differential pressure for a given subcooling margin (since core !
delta T is smaller with the RCP's running). Finally, with RCP's !
running, pressuriser spray is available and RCS pressure control is not dependent on the FORY or pressuriser vent. ;
l Main feedwater can be used if RCP's are' running; with pumps off. I emergency feedwater must be used, which is less effective in cooling the OTSC shell, thereby increasing tube to shell delta T. (i.e., tube tensile loads).
- 3. RCS pressure should be maintained low enough to prevent secondary side safety valves from lifting. HPI flow was not throttled sufficiently in the Ginna event of January 25, 1982
, and the steam generator filled with water. Since the RCS
' - pressure was above the SG safety valve setpoint, the safeties opened resulting in an atmospheric release of radioactivity. . _ _ . . _ , _
Moreover, the safeties were forced to pass liquid, which might cause the open failure of the valves.
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coolant pumps at low pressures. Shutting the pumps allowed noncondensible gases to collect in ,the top of the staan
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O O TDR 406 Rev. 2 Page 16 Of 74-generator tube U bends. These trapped ' ases g prevent'ed "RCS depressurization for many hours. An analogous situation might occur at the hot, leg U beads. The TMI-1 design.has always had capability of venting noncondensable gases from the U bends, however, which can be used if RCP's are not available.
.5. BWST Inventory-The Oconee' tube leak of September 18, 1981 resulted in a sustained (17. hour) leakage from the RCS to the OTSG's. This leakage caused the generator to fill. In order to prevent steam -
line filling, the operators at Oconee transferred water out of the OTSG's. In effect there was a once through cooling path from the BWST through the. core and out of the OTSG's. This experience illustrated the need to assure adequate BWST Inventory for core cooling. Second, it highlighted the need to
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store radioactive water in the plant during a prolonged RCS cooldown.
- 6. 'Shall-to-Tube Delta T A tube leak at Rancho Seco in May 1981 yiel'ded evidence of the '
importance of controlling OTSG tube /shall delta T. The existing limits and precautions at THI-1 is 100F'. However, before
- ' ~ tube /shell delta T exceeded 100F', the leaking tube was placed under tensile stress and the tube was pulled into a circumferential ta'ar. Maintaining tube /shell delta T limits are important during tube rupture and are discussed in more detail below.
2.1.2 - Limitina OTSG Tube Stresses Steam generator tube stresses are generated as a result of tensile loads placed on the tubes. These tensile loads come from two load components. The first is the temperature differential between the tube and steam generator shell. As the RCS temperature decreases, tube temperature decreases. At some point the difference in i' temperature between the colder tubes and warmer shall is sufficient to result in tensile stresses that pull apart a leaking tube. This topic has been the subject of extensive analyses within GPUN in conjunction with BI,W, EPRI, and MPR and the subject of a separate report (see Ref. 15).
The second load component is from OTSG pressure loading on the tubesheet which causes elongation of the shell. Isolation of the OTSG causes the tube /shell difference to increase while adding a tensile load on the tubes by alongating the shell via pressure loading. Structurally there are compensating effects involved in -
mitigating these two load contributors. Rapid depressurization l eliminates the pressure induced stress but aggravates the delta T 1
. + Rev. 1
- Rev. 2
. y . _ . s ._ . . . , . - . . . . . . . . . . . _ . . _ . . . . _ . . . = - , ._ .
L_. --
- ~. - . -. -
- .-. . . - _ . - _ . . . .~- -
, IDA 405 p p. Asv. 2
(
t Pago 17 of 74 Induced stresses. The optinua OISG cooldown/depressurization rate has not been determined. However, it is known that isolating the
- OTSG 'at 1000 psig is not the best means of ' reducing stress. -
~Cooldown/depressurisation.is the preferred method.
There are three 11 ales for tube /shall delta I that presently apply to' THI-1. Plant "Liaits and Precautions" (Aaf. 22) limit delta,T to
.60F' during heatup and to 100F' during cooldown with one DISG isolated. This value of 100F' assumed that tubes had no more than 405:through-wall cracks. In reference 23, AW established 142F' for '
.a cooldown using both OISG. The 70F' value in this IDA is proposed as a guide in determining an acceptable cooldown rate. If delta T
.can be maintained at or below 70F', the operator has optimised the plant cooldown rate. The 70F' limit more conservatively assumes that tubes in the OISG are leaking below a detectable limit. A 70F' value limits propagation of these cracks.
+. Control ofinhell to tube d' elta T is accomplished in several ways.
+- First by cooling the OISG liquid (steaming) to allow the metal shall
+ to. cool. Second, by providing cool, main feedwater into the
+ -downcomer. If neither of these methods works, the RCS cooldown must
+,
_ be decreased until the OTSG shall cools sufficiently. If reducing
+ the cooldown doesn't work, then the cooldown must be terminated.
2.1.3 Steasins Isolation and Fillina of the Leakina OTSG
~ Isolation of the leaking DISG can result .in the overfilling of that s'anerator. It is preferable to prevent overfilling, however, to allow plant cooldown in an expeditious manner. If the OTSG fills, it becomes a large pressuriser (as evidenced by the Ginna event). The cine it took to cool' down this mass of hot water greatly extended 'the cooldown of the plant. Steaming also maintains some natural circulation flow in the hot. leg. Ihis flow cools the hot leg 0 band ._
and decreases the chances of steam void formation.
+ As discussed in Section 2.1.2, steaming and depressuriza' tion of the 0TSG also reduces DISG tube stresses. However, depressurisation of the OTSG elso increases leakage race. As discussed in Section 2.2.2, the OISG pressure should be below ACS pressure to promote flow through the hot leg. The optimum OT3G control results in
- 1) depressurization of the CISG without causing large delta T's;
- 2) sinimum RCS lastage; 3) promotion of natural circulation flow in the hot leg; and 4) positive leakage from the ACS into the OISG to assure hot-leg cooling in the absence of natural circulation. Iba opelmum pressure control scheme to meet this criteria has not been determined analytically.
e e
+ 1ev. 1
.
- Aav. 2
a.,... '
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-~~~~~r -- w-~-
, - - ~
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' (~' TDR 406 Rev. 2 Page 1g of 74
+
+
-Meeting all four of these criteria will probably result in a'nearly continuous steaming of the affected DTSG. Moreover, intermittent
+
+
steaming of the OTSC's will result in release of all the noble gases transported into the OTSG from the RCS. Therefore, the TDR
+ recommends continuous steaming of the OTSG's.
+
The advantages of continuously steaming the affected OTSG's are:
+ 1. All of the above OTSG control conditions are met.
+ 2. The operator follows his normal cooldown procedures.
+ 3. Plant response is symmetric.
+ 4. Cooldown at low pressure / temperature can be accomplished more quickly, allowing DR system operation sooner.
Continuous steaming should result in a more rapid cooldown than intermittent steaming because of tube /shell delta T limitations.
Cooldown at 100 F/hr using the unaffected OTSG will result in a 70 F delta T limit in 1-2 hours. From this time on, the OTSG would have ;
to be staaned. Similarly, the OTSG would have to be steamed to e maintain natural circulation. '
Although it is highly desirable to prevent steam line filling, there
"*~ are certain circumstances which dictate that the OTSG should be fi1Ted. The Engineering Mechanics Section of GPUNC has established
-the capability of the steam lines to sustain the water hammer and dead load effects of flooding the steam lines (Raf 11). This analysis shows that the loading is acceptable without pinning (except for the dead load effects during a design basis earthquake). Since this combination of events is extremely remote, the procedures have been modified to allow filling of the OTSG. ,
The guidelines have the operator fill the OTSG's only under two ,
circumstances. The first condition is that BWST level decreases )
below 21 ft. At this leve'1, there is still sufficient inventory to flood both steam lines and put about 30,000 gallons of water into the 1 containment building (Ref 12). This amount of water is sufficient to provide adequate NPSR in an LPI to HPI " piggyback" mode of core injection from the reactor building sump (Ref. 13).
. + A second reason to fill the OTSG is for radiological considerations.
+
The OTSG should be isolated if offsite doses are approaching levels which would require declaration of a Site emergency. It should be noted that a Site Emergency may already have been declared based on
+ OTSG 1eakage rate. Nevertheless, this level provides a rationale for deciding that release rates are high enough to warrant OTSG isolation.
+ Rev. 1
- Rev. 2
_. . . . - . _ _ , . . . . . . . .. .- = . . = -
7 ~%
y ,
.__ , ._ _ __.usc evo ~
1sv. 2 Pads 19 of 74
+ Consideration was given to defining isolation conditions nased on RCS
-+ activity levels, meteorology and steam line radiation levels.- RCS '
activity level cannot be correlated to offsite releases, siace ,
offsite dose will be affected by the location of the tube leak in the OISC, availability of.the condenser and plateout and decontamination factors. It is also undesirable to isolate the CISG based on assumed, meteorological conditions. The most desirable approach is to isolate based on actual releases occurring during the event.
- The existing site emergency limits are 50 mRea/hr whole body and
-* 250 mRea/hr-thyroid dose measure or projected at the site "ooundary i (Ref. 33). Section 2.1.9. discusses the length of time required to
- cool the plant down to DER systes conditions. This length of time
- defines the integrated does allowed by the guidelines (i.e. release
- rate for the specified period of time).
- 2.1.3.1 Steaming, Isolation and Filling with Both OTSG's Leaking
+- Isolation and steaming of the OISG's must be addressed for leaks in
~+ both OISG's. Once 108 temperature is below 540 F, a choice has to be
+. ande regarding OISG isolation. Both OTSG's should be steamed unless
+* either the 3WSI level or offsite release criteria is reached. If
+* both OISG's are steamed, then all steam loads from both OTSG's should *
+ be isolated except for the TBV's/ADV's. All other steaming, l
+ isolation ~and filling criteria should be follound. .
- If the BWST level reaches 21 feet, then both OISGs must be isolated.
- If the dose criteria is reached, one OTSG should be isolated and the
--
- doses reevaluated. If the dose criteria still cannot be met, then the second CISG should be isolated.
2.1.4 Minlaus Alloweble Subcooling Margin
, A primary goal during a tube rupture is to minimize offsite dose.
Minialzing laakage from the RCS is the first line of defense. ,
Leakage from the primary to secondary is determined by the size of '
the leak, and by the differential pressure between the RCS and OTSG. l
, Primary to secondary differential pressure is controlled by fixing
- + the degree of RCS subcooling. Once secondary pressure is fixed, cold 1eg temperature is determined. For any time, decay heat is fixed.
RCS flow (which is determined by 0TSG 1evel or RCP operability) then i determines hot leg-temperature. Reactor coolant pressure or HPI flow then fixes the degree of subcooling. Since the operator controls OTSG and RCS pressure and EPI flow, he is in control of the
', subcooling margin, hence primary to secondary delta P. figure 2, illustrates the effect of subcoolitig margin on primary to secondary
. leakage. 1 j
Figure 3 illustrates the relative effects of a cooldown with RCP's off using 50F* and 25F' subcooling. Even at the maximum cooldown of 100F'/hr, the integrated leakage differs by a factor of two. '
l 1
+ Rev. 1 4
I
- Rev. 2
. =e m *
- e - *
- ee*
- een.= +=n e.ee mee . -mee e .w es, on e .eap e a=. e so.e.en em e en e=== es, aan
, rw -
' ~ I. - ',
- - y ,y,- -
Rev. 2 Fege' 20 of 74
{
2.1.5- Walve Fuel in Compression Limits
' ~
Fuel pin-in-compression lim 1ts are sgacifia*4 to assure tihat fuel pina are always in compression above 425F in order to prevent detrimental orientation (i.e., radial orientation of hydrides) (Ref. 14). These limits require a high subcooling margin for ECS pressures ranging fros 1350 psi to 550 psi. In correspondence dated January 20, 1983,
- (Raf. 14) AF4 confirmed that violation of these lialts during tube rupture events is acceptable. When these limits are violated it is laportant that the pressure and temperature versus' time be recorded '
so the effects on cladding can be evaluated. The evaluation must determine whether elad ballooning or incipient' cracking has been ,
induced. .
2.1.6 ' Reactor coolant Pump NPSB Limits l
+ RCP NPSE requirements place limitations on the ainlaus subcooling !
+ margin. At low RCS pressures RC pump NPSE 11alts approaca 100F' of i
+ subcooling. However, general centrifugal pump test data have shown
+ that NPSE requirements are substantially reduced at water tempera-
+ tures above 250F'. A review of TMI-1 test data on the subject
+- reactor coolant pumps indicates a single loop flow of 98,500 spa with l
+ two loops in operation with one pump per loop. The pumps' manu-
.+ facturer (Westingbouse) has provided required NP3B at various pump -
+ suction temperatures (Reference 25) for the flow associated with two
- . pump operation. The NPSE available, as indicated by the saturation margin monitor in the hot leg, is then calculated by considering the
-- +' total pressure drop from the hot leg to the pump's suction (Reference
+ 26). ,The resulting NP3R requirements for 2 pump operation (one per
- loop) are shown in Table'I and Figure 6. Also shown is tha 4 pump
+ operation NPSE curve which has considered the changed flow
+ distribution in the coolant loops. In. addition, the normal NP3R
+ eurve and the 25F' subcooling curve.are shown for comparison purposes.
RCP operation below 300 psig is not permitted. This assures adequate pressure differencial across the No. 1 pump seal (Ref. 32).
+ The usergency NPSE 11mits are incanded for operation of ECP's during
+ abnormal and emergency conditions such as small break LOCA, SG tube i rupture, station blackout and secondarf side upsat events. Pump limits and precautions must be adhered to while following the emergency NPSE liales '(e.g., the pump should be :: ripped on high vibration).
, 2.1.7 Procedure Entry Point Condition The use of an emergency tube rupt.ura procedure should be limited to situations where normal limits (e.g. fuel pin-in-compression and 107 :
NPSE) are being waived. The guidelines' entry point condition is chosen as 50 spa. A leak race of this magnitude would be expected froa che complace separation of one tube (as opposed to 385 spa for a double-ended off set of one tube). Less likelf, (but more serious)
+ 1ev. 1
- Rev. 2
-.--.....v.-. . . . . . . . . . . _ -
. = - - _
R::v. 2
(?
y; Fags 21 of 71.
would be leakage of this extent from a number of tubes. Both situations warrant-entering the emer'genef procedure. Below this limit, plant cooldown should be achieved witnin normal 11mits unless additional equipment failures occur.
2.1.8
- Slaulator Azpariance Steam generator tube rupture procedures were exercised during the January and June 1983 simulator sessions. Ihe experienca gained froa these two sessions has been factored into this IJR. The principal 1essons learned were that:
- 1.
- - Controlling plant cooldown rate with 2 or 3 HPI pumps running
'Is very difficult at best. Raising OT3G 1evel to 95% during this plant condition maf not be possible. -
- 2.
Prioritization of plant control parameters was not obvious to the operator in certain situations. Ihe two situations which were encountered were:
- a.
Minialzing subcooling margin has priority over mini-alzing the cooldown time, and;
- b.
- Steaming to control OISG 1evel is less laportant l
, chan RCS cooldown race. '
- 3.
- Plant response after RCPs are restarted was unexpected to the operation. Second pump restarts may be required before sun-cooling margin stays above 23F*.
4 Criteria for isolating core flood-is required. Core flood canks should be isolated in a subcooled system before chef initiate.
- 5.
Additional guidance is required if the RCPs are not tripped within two minutes of a loss of subcooling margin.
~*
These items are discussed in more deca 11 in Section 4.0.
2.1. 9
- Emergencf Limits for Decay Heat System Initiation Plant experience indicates that a large portion of time during cooldown occurs below the temperature of 3507. Slaple analyses, assuming onif one ADV for a loss of offsite power, using the CSMF computer code (Ref 31) indicate that the RC3 can be cooled, , _ ,
down below 300F in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and coo 1Jown to 275F can be
- accomplished in about ten hours. 275F is the normal DERS initiation temperature. GPUiG has evaluated the capability of the DER system to operate at a temperature of 300F (Ref 32) and concluded that it is within the design capabilities of the systaa. Therefore we recommend that the tube rupture procedure allow inLtiation of the DER sfstem at 300F instead of 275F.
2.1.10
- Reactor Irlp on Low Pressuriser Level
- Preventing a loss of subcoolin2 margin'has aanf advantages in
- concolling the plant. sefore the spring of 1983, EP 1202-5 required
- the plant to be tripped if level could not be maintained above 100
+ Aev. 1 i
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- Rev. 2 .
woe .
. . " ~ ~ ~ ~ ~
~IDi #UD'
' RGv. 2 0s
(~;
t.
Pri3 22 of 7L inches with two HPI pumps running. This may not be a sufficient 1aval to prevent voiding of the pres.surizar after a reactor trLp.
Emptying of the pressurizar causas a loss of subcooling margin and the subsequent tripping of the RC pumps. In order to prevent this situation, the reactor should be tripped if laval cannot be maintained above 150 inches or higher. This is suffielent voluma (about 600 cubic feet) to prevent pressurizar voiding.
There is a disadvantage to this recommandation since the safetf valves will lift after the reactor trips. However, this situation is '
considered acceptable when valghed against the plant control advantages of having RCP's running. Also, only a certain window of break sizes will result in reaching 150 inches and not 100 inches with full HP1 flow. Outside of this window, both levels would be reached.
2.2 Development of Multiple Tube Rupture Procedura Guidelines (Path 2)
Ina treatment of multiple tube ruptures railed on savaral sources of information. The Ginna tube laak exceeded the single tube flow for a B&W plant and also resultad in a loss of subcooling. Therefora, that ,'
event legitimately represented a multiple tube ruptura. Iha Oconea '
tube leak with a delay in getting onto deca / heat removal, prompted -
analysis of aster inventories required to assure a source of water
- for EPI cooling.
Besides plant operating experience, this TDR investigated the following aspects of multiple tube ruptures:
- 1. Revision of the RCP trip and restart criteria.
o
- 2. OISG sesaming and laval control. .
- 3. Establishmento ' f criteria for going on feed and bleed coolin3
- 4. Cooldown/depressurization.
2.2.1 RevisionofRhPTripandRestartCriteria Based on initial small break LOCA analfses recalved from FWA vendors in 1979, NRC concipdad in NUREG 0623 that delafed trip of reactor coolant pumps during a raa11 break LOCA can lead to predicted fuel
, cladding camparatures in excess of current licensing limits. At tha time of RC pump trip the liquid that was previously dispersed around
,. the primary system through pumping action now collapsed down to low points of the primary system such as the bottom of the vessel and steam generators. This separation results in significant uncovery of the reactor cora if system voiding is high enough, due to an insufficient amount of 11guld being available to provida acceptable core cooling. Unacceptable consequences would result from delayed reactor coolant pump crip only for a range of small breaks LOCA (.025 to 0.25 ft 2) and a range of trip delaf times after accident initiation. Based on these findings, a maecing of utility vendors and owners was held with NRC in September 1979. At this maating it was agreed that the 1600 psig ESAS signal provided timely Control Room indication for manual action to prevent possible voiding scenarios.
+ Rev. 1
- Rev. 2
1-p..%.
-.. ., TDR 406 '
. .o p _ Rev. 2
- k. ~ '
Page 23 of 7/.
GPU had B W reevaluate these LOCA scenarios assuming RCP's were tripped on loss of subcooling margin.- Tne, conclusion of that ,
reanalysis was that loss of subcooling was an acceptable alternative to pump trip on 1600 psig ESAS. In March 1983, tha NRC Staff required Utilities to reevaluate their pump trip achemes (Ref.17).
GPUNC provided an evaluation of the pump trip criterion and a schedule for implementing this criterion by May 1,1983.
The advantage of maintaining RCP't is that during Steam Generator i Tube Rupturea in whl=h minlaus subcooling margin is maintained, continuous RC pump operation assures expeditious cooldown with a miniana primary to secondary diffarancial pressure. This change in criteria for RCP trip will allow RCP'k to be operated for a greater spectrum of tube ruptures (including ruptures beyond the design
- basis) and to reduce the offsite doses for those events. Reducing the allowable subecoling margin is not intended to reduce RCP equipment protection. RCP'n should be tripped if emergency NPSE requirements are not met, and should not be started until NPSE requirements are re-established. If applicable NPSE pump vibration 11alts are exceeded, then the RCP's should be tripped. Pumps should
- be restarted as indicated in the TNI-1 Small Break LOCA Procedure (EP 1202-68, Attachment 1). As noted in Section 2.3, j bumping criterion requires additional clarification._ .
Figure 3 illustrates the reduced leakaga possible with RCP'h on.
Similarily, restart of RCP's has a great advantage. During tube ruptures, primary to secondary differential pressure decreases rapidly since OTSG pressure is high. Leakage flow is exceeded by EPI l flow and subcooling margin should normally be restored within 20-60 minutes after larger tube ruptures. Restarting RCP's provides pressuriser spray, and prevents void formation in the hot legs D bends and reactor vessel head.
, i 2.2.2 OISG Steaming and Level Control \
- The guidelines for OrSG steaming are nearly the same when either one I or both DISGs are affected. Tne OT3G pressure should be controlled l
~
to prevent lif ting of safety valves (i.e. stay below 1000 psig). I Level should be maintained below 95E;on the operate range. The re are *"-~ ' ~ -
1 saveral other issues to be considered for multiple tube ruptures, however. First, large tube ruptures may result in RCP trip. The OTSG's should be steamed to maintain natural circulation in the affected loop. Natural circulation flow will minlaize the l* + -chances of drawing a bubble in the hot leg U band. Continuous
+ steaming of tha OTSG allows all of these considerations to be
+ acconnodated. '
It is important to recognize that a large tube rupture with loss of subcooling is a LOCA condition. Therefore, it is required to raise OTSG 1evel to 95% to assure that liquid level in the tube region is
+ Rev. 1
- Rev. 2 '
~~~r:m:?v~7---- ~:-~ = m -*: e - ~ ~ ~ ~ ~ -
-- A
t - -
,___a .. _ .
n . Rev. 2
( ( Pags 24 of 74
^
high enough to allow water to flow into the core during boiler
- condenser cooling. If level is not raised to 954, then I.EJ. flow must
- .be at a high enough flow rate to penetrate f.he tube bundle *
- sufficiently to provide adequate heat transfer. A flow rate of 450 spa total (225 sps/0TSG) has been verified as acceptable
- by BM (Ref 29). --This flow is the nintana available after a seismic
'* event and worst case single f ailure, coincident with a small break l
- 'IACA in which boiler condenser cooling is required. It is important l co recognize that with two RFI's available, boiler condenser cooling ,
. i
- is not required. Procedures should therefore allow the operator to l
- raise DISG 1evel to 95Z* tempered with the need to control the ACS
'* 'cooldown rate. During tube ruptura events with both RPI pumps ,
- . available, , the unaffected OISG 1evel should be raised first vn11e the affected CISG 1evel should be prevented from boiling dry (maintain a minimum level of 30"). The operator can control
- 1 OISG instead of trying to raise level in both OTSG'4 1
- simultaneously. For the case with only one RPI pump, if incore j temperatures are not decreasing and the DISG is not removing RCS
- heat, then there will not be a coollown rate control consideration;
- noreovar, the plant asy be in a condition that requires boiler
~
- condenser cooling. Therefore, OTSG 1evels must be raised to the 95.%-
- . level slaultaneously in this situation (Ref. 30).
Section 2.1.3.1 discusses steam generator isolation, steaming and
+ filling critarla when both OTSG'i are leaking. This discussion also
+'
applies when the RCS subcooling margin has been lost.
~~
2.2.3 ~CEteria for Feed and Sleed Coolina Analyses of multiple tube ruptures indicate that existing plant procedures for establishing feed and bleed coollag are correct. Feed and niend cooling should os iniciated when the OTSG heat sink is not available. If both steam generators are isolated during a tube rupture, the FORV should be opened with full 171 turned on. An additional complication for tube ruptures, however,:is the potential - -
to flood the OIAG'n and force open the safety valves under this condition. If ACS pressure is below 1000 psig, the PGAV is capable
- ,of removing decay boat, even with liquid relief within two hours of
- reactor trip assuming that there is no energy relief out of the
- ruptured tube (see Figure 4). Therefore, the operator can control
- ACS pressure by throttling EPI. Moreover, with ACS pressure below 1000 psig the OTSG safety valves will not lift.
If ACS pressure stays above 1000 psig, however, the operator must
- cake action to prevent safety valve lifts. A situation with ACS pressure above 1000 psig and neither OISG avallanie requires the
- opening of the IBV or ADV'& to control OISG pressure below 1000 psig
- and level below the upper tube sheet, dither the ADV or IAV have .
- suf ficient stese capacity at high OTSG pressure to remove decay
- heat. The IAV'h also have suff Leient capacity to prevent DISG
- flooding. However, if the leak race is large enough, the steaming
- rate required to control level in the CISG maf result in an
- unacceptable ACS cooldown rate. In this case, cooldown rate must be
- controlled and the OISG allowed to flood. As discussed in Section
+ Aav. 1
- Aav. 2
.-w....y.- . . - _ ,
. . . w .p. .y. m ., . , , , , , , , . , , , , ,
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i '
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Fase 25 of 74 l
- 4.2.2.1, this situation does not seen likely (at least at high
- - j decay heat' levels). As decay heat decreases, steaming can.be
- - terminated when RCS pressure goes below 100'O psig and is controlled
- by'the FORY and EFI.
The steaming capacity of an A3Y at 1000 psig exceeds decay heat' 1evels within several minutes after reactor trip. RFI capacity
.escoeds the capacity of one ADY. Therefore, the RCS pressure can be
- controlled at 1000 psig in this mode without lifting safety valves.
Subcooling margin can be regained and the plant coolad down until an
- OTSG heat sink can be restored or until the plant can be put on deca / l
- heat removal.
Datil OI3G 1evel is above the upper tube sheet, pressure in the OISG will reaala below 1000 psig, since the RCS temperature is below
- 540F*. . With level less than 600 inches, however, the operator still
- aust steam to keep pressure below 1000 psig; therefore he should not 4-have to steam to control level on the affected OISG. When level goes
- above 600 inches, pressure in the OTSG is determined by the steam
- pressure in the steam line. If the lines are leak tight, then
- compression of the steam bubble can cause a pressure increase above .
. *- 1000 psig. In this case, the operator.would steam the OTSG to reduce 1
- pressure. However, if there are steam leaks in the afstem (e.g., '
3
- through steam craps.) then the 11 ass coul'd Illi with water before
' ** OISG pressure increased. Therefore to prevent this situation the
- Of3G's must be steamed to preclude this possibility.
~
2.2.4 Eldown/De$ressurisation - ' " ^ -
Analyses of multiple tube ruptures demonstrated that subcooling margin should be regained in 20-60 minutes (see Figure 5). ACP's can be started and a forced flow cooldown instituted. Even if RCF's are not available, the cooldown during a multiple tube rupture can be accomplished within the single tube rupture guidelines. If equipmect
- failures prevent a normal natural circulation cooldown, then the plant would be cooled down with fsed and bleed cooling. This saneuver would probabif require initiation of feed and bleed coollag in the EFI/LPI " piggyback" mode. Existing plant procedures give correct guidance about when to initiate this mode (8437 level below 3 ft.). .
+ Guidance from 3*.W on FTS/3rittle Fracture liales requires a " soak
+ time" to allow the vessel wall to reach the RCS temperature.
+ '
Bowever, &&W has also recommended that the " soak tina" is not
+. required during tube rupture events in vnich a rapid cooldown is
+ necessarf (Reference 24).
Steam releases during multiple tube rupture events can be minimised bf judicious use of the EFW, EFI and ISV's. Full HFI flow, in
+ conjunction with throttled EFW flow allows a 100F'/hr cocidown without having to steam either OISG.
(
+ Rev. 1 .
- Rev. 2
, . . . . =
w.
IJR 406
', Asv. 2
( ., Pcge 26 cf 74 2.3 Additional Work Requirements 2.3.1
- Analyses ' '
As noted in Section 1.0, there is a program of ongoing computer.
analysis work simulating single and multiple steam generator tube
- ruptures. The effort includes the plant states listed in Sections
- 3.1.1 and 3.1.2. Inis list does not reflect two specfte detailed
- analysis efforts which are being undertaken as part of the tube
- rupture quantitative development effort. The first analysis is a '
- simulation of the vessel head region during natural circulation
- cooldown. This analysis effort will help in evaluating the effect of
- a vessel head bubble on the RC pressure response. It will also aid
- in evaluating the benefit of the reactor vessel haad vent.
The second analysis effort is being perforred in conjunction with Babcock and Wilcox Company. Detailed analysas are being performed to
- provide improved guidelines for OTSG filling af ter a loss of subcooling margin, with one, two and three HPI pumps available. The intent of the analyses is to assure that the OTSG's are filled
- without violating cooldown or tube to shall differential temperature limits, while still meeting core coolabilitf requirements. This effort was considered after the January 1983 simulator training A
session and furchar defined after the June 1983 simulator session.
2.3.2
- Issue Resolution A number of issues were ideutified which require further effort
- to resolvu. These are the following items:
- A. Operator guidance for identifying two phase natural cire-
- ulation cooling 'boller-condenser).
- 3. Acceptability of axcessive cooldown rates for, vary short time
- Intervals.
- C. Importance and technical basis of Fuel-in-Compression limits.
- D. Viability of DER system initiation at temperatures above
- 3000F.
- . Identification of pump vibration limits for various pump
- combinations.
- Iha bumping criterion would allow running pumps without
- adequate subcooling or NPSH margin as long as a steam generator
- heac sink is available. Determine whacher the AIOC criterion
- anticipates that NPSH will be reestablished since the heat
- sisk is available. The existing bumping criterloc in IMI-1
- emergency does not require that a heat sink be established
- limits.
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' TDR 406 Rev. 2 Page 27 of 74 3.0 DISCUSSION OF MAJOR REVISIONS TO EXISTING PR6CEDURES The development of the design basis guidelines discussed in Section 2.1 identified s. number of areas which were investigated to determine where specific changes should be incorporated into the new guidelines. This section further explains what areas of the guidelines should be revised. -
3.1 Basic Plant State 3.1.1
- Assumed Plant Conditions The following assumptions apply to the development of guidelines as they apply single tube leak / ruptures.
- 1. Subcooling margin (SCM) is maintained.
- 2. Only one OTSC is affected.'
- 3. Condenser is available.
4 Reactor Coolant Pumps (RCP's) remain on. *
- 5. Decay heat is removed by the intact OTSG until the Decay Heat
,_ __ Removal (DH) system can take over.
- 6. The affected CyrSG can be steamed to maintain less than 95% level (Operating Range) and less than 1000 psig.
In addition the revised guidelines will have provisions to deal with the following circumstances:
1.. RCP's not available. '
- 2. Condenser not available.
- 3. High radiation releases offaite.
- 5. Steam lines associated with leaking OTSG flood.
- The consideration of items 1 and 2 are equivalent to an assumption of loss of offsite power.
3.1.2 Tube Rupture Guidelines For Loss of Subcooling Tube leaks in this category generally go beyond the licensing basis, or are otherwise remarkable due to plant conditions (aside from the tube leak) or equipment malfunction.
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TDR 406 R:v. 2 .-
Page 2g of 74 The following conditions were assumed in developing guidelines for
~
c this category of. tube rupture event _.
9
- 1. .More.than one tube leak.
- 2. SCM is lost.
- 3. RCP's are unavailable..
- 4. Pilot-operated relief valve (PORV) and Reactor Coolant System (RCS) high point vents are available.
5.- . Unaffected CrrSG can be steamed.
Continrencies 4:
' The revised guideline will have provisions to deal with the following additional circumstances:
- 1. Both OTSG's are affected..
I-t
removal and either a)' the FORV is unavailable or b) RCS pressure stays above the main steam safety valve setpoint due to void
.. formation in the RCS.
N
~ 3. Neither OTSG is capable of removing decay heat, and either a) the PORY is available, or b) the PORV is unavailable.
3.1.3 . Revised touipment Limits & Operating Practices
~
During the course of the analyses leading to the guidelines provided in Section 4.0. It became apparent that certain normal equipment limits and operating. practices should be adjusted to effectively deal with a tube leak / rupture. -These changes will help accomplish the following:
- 1._ Mitigate or prevent further OTSG damage.
4
- 2. Maximize the cooldown rate to cold shutdown.
. 3. Minimize SCM (thus minimizing primary to secondary leakage).
- 4. Maximize RCS pressure control options. :
An Event Tree showing the various possible developments of an OTSG tube leak appears as Appendix D to this report.
O I
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IDR 405
{ p Aav. 2-Pege 29 of 74 3.2 Discussion of Guidelines Appendix C provides a logie diagram of the tube rupture guidelines (with a written discussion of those guidelines). Ibis section of the report describes the guidelines shown in that diagram. Ihe symptoms of' the tube rupture procedure define the entry point conditions when the emergency procedure is used. This procedure need only be entered for situations where a rapid depressurization of the plant is warranted. Wnen such conditions warrant, than the plant should be shut down and cooled down as expeditiously as possible and cereal,n normal plant limits (RCP NPSE, normal tube /shall delta T and fuel in compression 11mics) are waived.
3.2.1 Immediate Actions The tube leak in question may not be large enough to cause a reactor trip. In such a case, the operator begins a load reduction as - - -
rapidly as possible without causing a reactor trip (104/ min.).
Avoiding a reactor trip prevents lif ting of the OI3G safety valves.
3.2.2 Followup Actions - Subcoo11n2 Maintained and ACP's Available Once the load reduction is initiated, the operator has several aajor goals to achieve while bringing the plant to a cold shutdown condition. First, he must prevent lifting of the OI3G safety valves; second, mir'mize primary to secondary leakage by minimizing primary to. secondary differential pressure; and, third, minimize stresses on the DISG cubes by limiting tube /shall delta T. Finally, the operator will minimize offsite dose by allowing the leaking OISG to flood if offsite doses are large enough (approaching levels at which a Site Emargency would be declared).
The major differences between the existing plant procadura and the proposad new procedure would be the following:
3.2.2.1 Maintain a Minimum of 25f' Subcooling Minimizing subcooling margin means that primary to secondary differential pressure is also minialzad, whLeh reduces leakage and offsite doses making the event more manageable.
3.2.2.2 Steaming / Isolation Criteria for the Affected OI3G l
The present procedure allows the operator to let the OISG fill
, anytime that RCS pressure is below 1000 psig. The revised procedure has the operator steam the DI3G for these purposes: Eirst, to prevent lifting of the OISG safety valves. Second, to prevent the generator from filling.
~
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- -... . IDA 406 Aav. 2 h h Pags 30- of 74 3.*2.*2.3 Shell-to-Tube Delta I
- Plant 1imits and precautions require mainta'ining the OISG tube
~
temperature within 100E' of the shell cesperature. A shall to tube delta I-of 70F' limits stresses and minialzas the enances of increasing the leak size.
i
- 3. 2.' 3 Pollowup Actions (Automatic Reactor Irip has occurred)
All of the f'ollowup actions discussed above stL11 apply when tha tube leak is large enough to cause an automatic reactor trip. In l addition, the f allowing procedure changes would apply. i I
3.2.3.1 ACP Trip With a Loss of Subcoolina Marzin s
Present plant procedures require RCP trip on initiation of 1600 psig ESAS. Rupture of one or a fav OISG tubes will likelf result in ACS depressurization to the RPI setpoint, but saf not result in a loss of Scti. .
I 3.2.4 Follovup Actions for Loss of Subcoolina The third section of the tube rupture procedure is entered when RCS '
subcooling is~1ost. Here, the operator must treat LOCA, as well as ~
tube rupture symptoms. He is than able to pursue the followup tube rupture actions. All of the guidance for followup actions without loss of.subeoollag appif.
The objective in this portion of the procedere is to maintain natural circulation (if possible), reestablish subcooling margin, restart one reactor coolant pump per loop (if possible), and return to the
- section of the procedure for. forced flow cooldown.- If one pump can not be started per loop, then start both RCP's in one loop. This will maximize the pressurizar spray flow for the given ACP
- arallability. If the affected OISG canact be steamed for either radiological or equipment reasons,.then EEsi is used to control OISG pressure. Essentia11yi EFW is used as a pressurizar spray to keep- - - - - -
the leaking generator slightly lower in pressure than the RCS. Ihe benefits in controlling pressure are:
- 1. safeties will not lift.
- 2. the steam generator will not control RCS pressure. -
- 3. there will'not be backleakage into the ACS of water or steam f rom the OISG.
- 4. leakage from the RCS to the DISG will be small since differential pressure will be saali.
- 5. the small flow through the hot leg aaf prevent void formation in the hot leg.
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- are started and the operator continues the cooldown. The desired RCP
- configuration is to start one pump in each loop. If the operator is
The reasons for restarting RCP's are similar to the reasons for not tripping them on low RCS pressure. If subcooliing margin is lost '
immediately af ter RCP restart, it implies that the increased RCS flow has caused voids in the system to collapse, thus dropping RCS pressure. Allowing 2 minutes for SCM recovery prevents incessant
" pump bumping," but keeps the RCS out of the fuel damage region,.
If subcooling cannot be restored, the operator cools the plant down on natural circulation unless the OTSG heat sink is lost (for example, due to-loss of natural circulation in the unaffected loop).
With no steam generator heat sink, the operator must put the plant in a' feed and bleed cooling mode. Feed and bleed cooling is initiated.
by isolating the OTSG's, assuring full HPI is on and opening the PORV. With RCS pressure below 1000 psig, water relief out of the PORV is sufficient to keep the core cooled (See Figure 4) after about
- 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. If the OTSG heat sink is restored, the feed and bleed mode '
is terminated and a natural circulation cooldown is reinitiated.
~ If RCS pressure stays above 1000 psig during feed and bleed cooling (e.T. .- the head bubble prevents depressurization or the PORV fails closed) then the secondary side safety valves have to be protected from challenge.. The operator controls OTSG pressure with whatever
- means are available (turbine bypass, or Atmospheric' Dump Valves.).
When the OTSG tube region is filled with water, the operator opens the ADV and leaves it open. This action minimizes the chances that safety valves will be forced to relieve water-and/or steam and fail open.
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TDR 406 Rev. 2 Page 32 of 74 4.0 SIMULATOR TRAINING EXPERIENCE * ' "
4.1 Introduction
- Nost of the guidelines proposed in this TDR were incorporated into a lesson plan for the annual requalification training of the TMI-1 licensed operators at the BW simuistor. A draft revision to TMI-l's
. EP 1202-5, incorporating the guidelines, was also prepared.
- These documents were then used to inform the licensed operators of the changes contemplated for EP 1202-5, and to demonstrace the combined .
effects these changes would have. During the classroom session, each guideline was described and the reasoning behind it was explained.
During the simulator session, their combined effect was illustrated by running a large tube leak scenario twice.
For the first simulator run, the then-existing revision of EP 1202-5 was used to deal with the leak. For the second run, the draft version was employed. It became apparent that the new guidelines made plant control easier. ,
-* As indicated in Section 4.2.2, the January 1983 training cycle was not ,
- effective in training operators on the basic concepts for treating
- tube ruptures. The June training cycle was successful in
-* consunicating concepts.
4.2 Results
- 4.2.1 January 1983 Traininz
- Of all the guidelines proposed in this TDR at that time, the two changes most useful (and obvious) to the operators were:
- 1. Reactor Coolant Pump (RCP) trip as a followup to low subcooling margin (SCM) rather than following automatic HPI from a low RCS pressure ESAS signal;
- 2. HPI throttling when SCM requirements are met and pressuriser level is back on scale, rather than waiting fer pressurizar level to reach 100".
- Another useful.(but less obvious) change is the RCP restart criterion based on regaining SCM rather than various combinations of primary and secondary pressures. This and Item 1 above may be considered under the same general heading of increased RCP availability.
The exercise of the draft EP 1202-5 was useful in critiquing the
+ contemplated changes. Simulator experience also showed that it is not
+ possible to raise OTSG 1evel to 95% with full RPI on, while steaming the OTSC and maintaining a 100F'/hr cooldown. The difficulty was
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I created by the steaming of ~ the OTSG's in this situation. dPI and l throttled EFW flow can provide a plant cooldown at near 100F'/hr if the DISG's are not steamed.. e l
Durlug the sf=ntator session, B&W revised the simulator to allow l
leakage of more than-2 tubes and to allow leakage in both OTSG's.
4.2.1.1 Comments *. {
l This material was presented to two of seven groups by Tech. -
Functions personnaL The remaining five groups received it from )
B&W training personnel who taught the material using the same lesson outline. S&W did not endorse the material. Comments from ,
trainees indicate that the training was of dubious value. It
~
i vill be necessary to repeat the training for all personnel.
J :
4.2.2 June 1983 Training *
.A number of items were identified during the B&W operator training l simulator sessions held Irom June 6 to June 29, 1983. The experience
. gained from using a revised tube rupture procedure EP 1202-5. These I
. items will be discussed below. '
i
- 4.2.2.1 Control of RCS Cool $own Rate I l 4
i
- Section 5.2.1 discussed the difficulties in controlling !
' cooldown rate while raising OTSG 1evel to 95%; At the time, l
~J * -
it was the author's belief that steaming of the OISG'is was
- -* causing the excessive cooldown rate. However, furchar ;
discussion with B&W (Ref 29) indicated a different )
explanation. The B&W BPI model calculates flow by {
'iteratively-solving two equations of the form: " - - -
- I Pd = 2840 - K (1)-
N and l
- W= (Pd - Pges) A 1/2 (2) i
- where:
Pd .= Peng discharge pressure
- PRCS = RCS pressure )
j
- W = flow, Iba/sec '
- N = number of HPI pumps running
- A = coefflclant to account for number of HPI valves open.
- ' The EPI flow is overpredicted for. IMI-1 with three HPI pumps running and/or low RCS pressure. The difference results from the
- . physical arrangement at IMI-1, in which two pumps discharge into
- a common header. The cavitating venturies at Unit 1 also reduce the maximum flow of the HPI pumps compared to the valve
- calculated by the simulator.
As a result of this understanding,, subsequent slaulator drills were run with only two EPI pumps avaliable and control of
- cooldown rate was improved.
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~*' The RPI-initiation rule has been reemphasized to the operator, '
- 'namely that "Iun" EPI means the . fun flow from two . EPI pumps.
- It is acceptable to secure cbe third HFI pump when the RCS is saturated, and the DISG beat s.sk is available or if cooldown rate is 100P'/hr. or more.
- ~ .A second. consideration in controlling cooldown rate was in
- ' _ raising ~the OISG level increase to 95% after a~1oss of subcooling
- margin. Operations believes that it is an unnecessary burden on the operator to control cooldown race while raising level on both '
OTSG's slaultaneously. Therefore, the leasing generator level
- win not be ratsad until the unaffected CISG has been raised to 95% unless incore thermocouple temperatures are not decreasing -
and the OISGs are removing decay' heat. The 95% level is
- -important in establishing boiler-condensor cooling during saan LOCA's in which only one BPI is avanable. Howeyer, the RCS 4 cooldown is only a concern if both EPI pumps are running. .- -
Therafore the two concerns are autually exclusive.
Operator training and EP 1202 5 have been revised to have the
- - operator control one 0TSG 1evel at a clae as long as incore ,
temperatures are decreasing. If RCS temperatures are not affected by the secondary side cooldown (i.e., no secondary side
- heat removal) than both OISG's should be raised to the 95% level-slaultaneously. ;
- During the slaulator session of June n , 1983 the operators were
~ " - *
- faced with a large (about 1400 gpa) cube rupture. They attempted
- to control OISG 1evel below 95% on the affected ganarator.
However, the cooldown race was too high even with the unaffected
- - OTSG isolated. The simulator response to the event was partially responsible, but the procedura siso needed to be more explicit.
~*'
The simulator ^ leak model currently uses-the orifice equacLon to predict lesa flow (Ref. 28).- This model would initia11f overpredict the break flow and would account for the verf rapid flooding of the DISG's compared to rc suits of the REIRAN and REMP5 computer codes. REISAN and REL&P5 (Ref. 35) analfsas to -
date do not_ predict such a response. Nevertheless, the operator
- needs to recognize that cooldows rate control takes preesdence e.
over OTSG 1evel control and EP 1202-5 was subsequentif revised to prevent this conflict in plant control requirements.
-4.2.2.2 Plant Stabilization Before Cooldown
- The procedure used during the training session had the operator initiate plant cooldown and then establish minimus subcooling margin. The simulator sessions showed that the RCS could not be depressurized fast enough to maintain a ainiana subcooling margin. Therafore, training material was revised to emphasize the need to stabilize the plant and reach tne alalaus allowable subcooling margin. Plant
~*
cooldcies should then be initiated. The procedure was
- aodiLLsd so that all four RCf's can be lef t on until 5001 Instead of 540F. Based on the TMI-1 AIOG (Ref. 28), this change provides,a difference of about 47%
in the spray flow (see Table 2). Ihus pressuriser spray flow.is maxialzad for as long as possible. Finally, the
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- operator is given the option of using the pressurizar vent if he is still unable to reduce pressure sufficiently to maintain a minimum subcooling margin. .
4.2.2.3 RCP Restart Criteria
- Sectica 5.2.1 observed that the RCP restart criteria on 25F subcooiing margin (SCM) was verf usaful to the operator.
-
- Several areas required clarificatloa or expansion, however.
First, RCP restart should not be attempted unless pump
- emergency NPSH limits are met. The only exception is chac NP5H requirements are waived, with pump restart allowed
- during certain inadequate core cooling conditions as specified in plant procedures. Pump " bumps," however, should be attempted even if NP3H requirements are not met. Second, loss of subcooling may occur after the RCP's are restarted. Collapse of steam volds in the RCS may cause voiding of the pressurizer,'resulting in a loss of SCM. RCS temperatures may be hotter in the tube region than in the core if natural circulation has been lost. Mixing of
, this hotter water with the cooler core water will cause a decrease in the SCM. Several pump starts may be required before subcooling margin stabilizes above 25F'. Allowing two ;
- minutes of RCS flow is halpful in eliminating both staan
- l volds and temperature gradients so that successive restarts will be successful.
- 4.2.2.4 Core Flood Tank Isolation
- In several simulator runs, the operations were faced with
- tube rupture or small break LOCA conditions in which the RCS was subcooled, but core flood tanks initiated, proriding
- cooling water which was not required, since SCM was maintained. Most significant was that CEI initiation
- delayed RC3 depressurization. Naither tha LOCA nor tube rupture procedure provides any guidance about 1 solation of the core flood tanks. Therefore, this IDA has bean revised to provide guidance about when to isolate the core flood tanks (see Section 5.2.11).
- 4.2.2.5 R,CP Trip Criterion During the June 1983 simulator session, questions arose re- - - -
garding the actions to be taken if RCPs were not tripped within 2 minutes of a loss subcooling margin. Clarification
- was provided using the guldance of the EfI-1 " Abnormal Transient Operating Cuidelines" (AIOG) (Ref. 28). AIOC re-quires the operator to keep the RCPs in each loop running if the two minute time limit is exceeded. If the RCPs are tripped af ter this time, the RCS may be at a sufficient high void fraction to uncover the core. The RCPs should be run
- for at least 7000 seconds in this situation to assure that the core will not uncover (basad on Appendix K assumptions). If pump damage may occur, then one pump in each loop should be tripped. If either of the running pumps fails, the 'crioped pump in that loop should be started. This strategy maxlmizes the likelihood of maintaining adequate core cooling.
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{ Page 35 of 74
-operator is given the option of using ene pressurizar vent if
- ~
'he is.scill unable to reduce pressure sufficiently to maintain a minimum subcooling margin. .- .
4.2.2.3 RCP Restart Criteria
- i Section 5.2.1 observed that the RCP restart criteria on 257 l
.subcooling margin (SCM) was verf useful to the operator. l
- - Several . areas required clarification or expansion, however.
-* First, RCP restart should not be attempted unless pump +
- emergenef NPSE limits are met. The only exception is that
.NPSE requiremants are waived, with pump restart allowed 4
- during certain inadequate core cooling conditions as specified in plant procedures. Pump " bumps," however, should be attempted even if NPSE requirements are not mat. Second, loss of subcooling may occur after the RCP's are 1
1
- restarted. Collapse of steam volds in the RCS may cause'
- i voiding of the pressurizar, resulting in a loss of '
SCK. RCS temperatures may be botter in the tube region than L*
in the core if naceral circulation has been lost. Mixing of this hotter water with the cooler core water v111 cause a decrease in the SCK. Several pump starts may be required before subcooling margin stabiu zas above 25t*. Allowing two l
- ainutes of RCS flow is helpful in ellainating both steam
)
voids and temperature gradients so that successive restarts l will be successful.
< ~-
- 4.2.2.4 Core Flood Tank Isolation l
- l In several slaulator runs, the operations were faced with tube rupture or small break LOCA conditions in which the RCS was subcooled, but core flood tanks initiated, praviding ;
-* cooling water which was not required, since SCM was
- maintaines. Most significant was that Cfr initiation '
- - tube rupture procedure provides any guidance about isolation of the core flood tanks. Therefore, this T3L has been revised to provide guidasco about when to isclate the core flood tanks (see Section 5.2.11).
i
- were not tripped within 2 minutes of a loss subcooling margin. Clarification was provided using the guidance of the l E. IMI-l ATOG (Ref. 28) which requires the operator to keep the
' RCPs in each loop running if the two minute time limit is
- : voided enough to uncover the core. ECPs should be run for at least 7000 seconds to assure that the core will not uncover (based on Appendix K assumptions). If pump damage
- may occur, then-one pump in each loop should be
- . tripped. If either of the running pumps fails, the tripped pump in that loop should be started. For slaplicity, the guidelines in this TDR call fot 1 RCP to be run in.sach loop.
This provides suf ficient flow to cool the core (Raf. 28)~.
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, TDR 406 Rev. 2 Page 36 of 74 5.0 TUBE LEAK / RUPTURE GUIDELINES *
- 5.1 Scope
- The guidelines will deal with tube leaks in excess of 50 gpm.
Primary-to-secondary tube leak rates less than 50 gpm will be handled in accordance with " Guidelines for Plant Operations with Steam Generator Tube Leakage," TDR 400 (Ref. 16). i 5.2 Guidelines & Limits
- This section provides plant specific technical guidelines for tube rupture events which can be used to generate plant Emergency Procedures.
5.2.1 Subcooling Margin Requirements
- Control Reactor Coolant System (RCS)'subcooling margin (SCM) between
+ 25F* and 50F*. Maintain SCM as close to 25F' as possible consistent
+ with the RCP NPSE curve of Figure 6 and while waiving fuel ,'
+ pin-in-compression limits.
This will minimize primary to secondary differential pressure, thus minimizing the leak rate.
5.2.2 Reactier Cociant Pump Trip Criterion * -
Trip Reactor Coolant Pumps (RCP's) when SCM is lost.
Note:
- If RCP's are not tripped within 2 minu.es of loss of 25F' SCM, then a
- run 1 RCP in each loop.
5.2.3 Beactor Coolant Pump Restert Criteria
- When the required subcooling margin (25F') has been established,
- both RCP's in the opposite loop.
Note: If subcooling margin is lost immediately after pump restart and does not return within 2 minutes, the RCP's must be tripped again and not
- restarted until SCM is regained. Subcooling may be lost several times
- before the pumps can be left running.
5.2.4 Reactor Coolant Pump NPSH for Emergency Operations *
+ The attached curve (Figure 6) depicts the RCP NPSH limit to be used
+ during a cooldown with a tube leak.
5.2.5 High Pressure Iniection Ihrottling Criteria
- Throttle HPI when SCM requirements are met and pressurizer level comes back on scale. (Note that the other HPI throttling criceria remain -
unchanged.)
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TDR 606 Rev. 2 Page 37 of 74 5.2.6 OTSC Level *
- If SCM is lost:
- a)-
Raise, level on the unaffected OTSG to 95%'while leaving level control'on the unaffected OTSG at 30 inches.
- b) Raise level on the affected OTSG to 95%.
,**NOTE:' If incore thermocouple temperatures are not decreasing and there is
-no heat transfer _ to the OTSG's, then both OTSG 1evels must be raised .
to 95% sismultaneously.
5.2.7' OTSG Isolation /Steamint criteria *
. When the leaking OTSG is identif'ied, close all steam valves except the ADV's and TBV's. ,
Note: Do not close MS-VID until an alternate source of gland steam is available. '
When RCS Thot is less than 540F' the affected OTSC must be isolated
- if:
-(a) _ Borated Water Storage Tank level is below 21 ft., or (b) Offsite dose projections approach the level requiring a Site Emergency (50 mrem /hr whole body or 250 mrem /hr thyroid).
Nota: 'If both CT5G's are leaking and isolation is required based on offsitE dose projections, first isolate the OTSO with the higher leshage. If such a distinction caanot be made, isolate one OTSG and reevaluate offsita dose projections.
5.2.7.1 Pressure Control of an Isolated OTSC
- Steam the affected OTSG(s) only:
1.. To keep OTSC pressure below 1000 pois.
- 2. If the plant is on feed and bleed cooling. OTSG 1evel must be controlled below 600 inches on the vide range indication.
If the OTSG aust be steamed open the Turbine Bypass Valves or
- ' Atmospheric Dump Valve on the affacted OTSG. ~
5.2.8 Cooldown Rate Durint s Tube Leak Event
- The cooldown rate shall be limited to a maximum of 1.67F'/ min (100F'/hr) whether on forced.or natural circulation.
Note: Steaming of the OTSC's may not be required if OTSG 1evel is being increased using EFW.
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' TDR 406 Rev. 2 Page 38 of 74 5.2.9 OTSG Shell-to-Tube Differential Temperature Limit +* ~
+ Maintain OTSG differential temperature less than 70F*.
+ If this limit is approached, then:
+ 1. Reduce the cooldown rate in half.
+- 2. Continue steaming on the affe:ted OTSG.
+ 3.
Supply MFW thru the startup control valve at about 6
. .05 x 101bm/hr (if MFW is net being used).
+
+
If the differential temperature approaches 100F*, stop the cooldomt
+ and maintain RCS temperature constant. Remove decay heat by steaming
+
the OTSG(s) with the high differential temperature. Resume the cooldown when the differential temperature drops below 70F*.
5.2.10 Cooling Mode When Both OTSC's are Unavailable for RCS Rest Removal *'
Use HPI " feed and bleed" to cool the'RCS when" both OTSG's are
+
unavailable. Open the Pilot Operated Relief Valve (PORV), RC-RV-2, to provide a cooling water flow path to the Reactor Building Sump. ;
5.2.11 Core Flood Tank Isolation
- Isolate the Core Flood Tanks if:
~ '
- 1. Subcooling margin can be maintained above 25F*, and
- 2. RCS pressure is below 700 psig.
5.2.12 cuideline Flow Chart " .
+
+
Appendix C includes a flow chere and explanatory text showing the logic path of the tube rupture guidelines.
O l
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- Rev. 2
~
g~, ~ .w TJA 405
.. % g Rev. 2
's - Page 39 of 74
- 6. 0 CONCLU3 IONS AND RECOMMENDAIIONS The ability of the plant to handle beyond design basis events'ca:1 be substantially increased and the RC3 leakage can be reduced for design basis leaks with the adoption of the following changes /adi.itions to tube ruptura procedures.
- 1. Reduce minimum subcooling margin to 25F"
- 2. Replace the existing RCP trip criteria with trip on loss of '
subcooling.
- 3. Adopt the steam generator isolation and pressure / level control .
guidelinas of this guideline.
4 Provide the RCP NPSE limits of Pigure 6 for use during emergency conditions.
- 5. Waive fuel pin-in-compression limits.
- 6. Control plant cooldown to limit the tube /shall delta T to 70F*.
I
- 7. Revise procedure entry point conditions to be leakage greater a chan 50 spa.
- 8. Incorporate criteria for initiation of feed and bleed cooling
,__into the tube ruptura procedure.
- 9. Adopt criteria for opening TS7's/AD7's if RC3 pressure stays above 1000 psig during feed and bleed cooling.
- 10. HP1 throttling should be allowed when subcooling is regained and ~ ~ ~ ~~
pressurizar level is on scale.
- 11. Core flood isolation criteria be incorporated into amargsney
- procedures dealing with LOCA, tube rupture and stesa line breaks .
- and in operating procedures deallag with dorced and natural
- circulation cooldown.
- 12. Decay heat removal system initiation under emergency conditions
- be allowed at 300F.
It is further recommended that these changes be implemented prior to restart of IMI Unit 1.
+ Rev. 1
- Rev. 2
l 1s _ _ .
.. m.- ..____..=w2.
- .g __ _ _ _ _ _. .
4 Rsv. 2
,' h'- , Pass 39 of 74 7ggg y(2)
Tabular Values of RCP Emergency NFSH Requirements Fot-2 1CP FER-LOOP OPERATION INDICATED TEMP. ALLOWA8LE INDICAIED PRESS. '
(F)- (FSIG) 94.4 300.0 (1)
'194.4' 300.0 (1) 294.4 300.0 (1) 344.4 310.8
- 3 94. 4 ~413.9 .
444.4 567.4 544.4 _ 1187.1 FOR 1 RCP FER Ih0F OPERATION
-INDICATED TEMP. ALLO 2ABLE INDICAIED PRESS.
(F) ,
(PSIC) 94.4 300.0 (1) l 194.4 300.0 (1)
' l 294.4 300.0 (1) 344.4 .
302.8 394.4 405.9 444.4 559.4 544.4 1178.6 NOTE 1.' Limitations based on the #1 seal differential pressure requirement.
- 2. --Entire table was added in Rev. 2. .
+ Rev. 1 l
. A Rev. 2 l l
' r h . m
t._;~
- " ~ -
.......-.._..2ag,_ gg , ,
,,. asy. 2 pw. Pegs 41 cf 74 IABLE 2(
Pressurizar Spray Flow for Various P' mp u Combinations NUMBER OF RC PDMPS RUNNING SPRAY FLOW ,
SPRAY LOOP , OPPOSITE LOOP, (I FULL PIEW) ,
2 2 100 2 , 1 92 2 0 84 1 (Spray line next to' running pump) 2 60 1 (Spray lina next to running pump) 1 53 1 (Spray line next to idle pump) 2 50 1 (Spray line next to running pump) 0 41 1 (Spray lina next to idle pump) 1 38 1 (Spray lina next to idle pump) 0 26 0 2 30 0 1 0 As a rule of thumb tripping one pump in each loop will provida a good balance between the spray flow rate and the heater capacity. It viu also provide good forced circulation for cooldown.
NOTE: The fo Mowing table will give ganaral guidance for the affects of running various pumps. This table was calculated Ior normal operating conditions.
NOTES:
- 1. Reproduced from THI-1 ATOG (Ref. 28), "Sast Methods for Equipment Operation *,
Table 6.
- 2. Entire table was added with Rev. 2.
e
+ Rev. i
- Kev. 2
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Page 42 of 74
7.0 REFERENCES
e 1.
U.S. Nuclear Regulatory Commission. NRC Report on the January
~25, 1982 Steam Generator Tube Rupture at R. E. Ginna Nuclear Power Plant. NUREG-0909. U.S. NRC.
2.: U.'S. Nuclear Regulatory Commission. Safety Evaluation Report
_Related to the Restart of R. E. Ginna Nuclear Power Plant.
- Docket No. 50-244 NUREG-0916. May 1982. U.S. NRC.
- 3. C..Y. Cheng. ~ Steam Generator Tube Experience. NUREG-0886. U.S. .
NRC.
, 4. - Rochester Gas & Electric Corp. " Procedure E-1.4. S/G Tube Rupture." June 23, 1982. Ginna Station.
. 5. " Procedure No. E1.0. Safety Injection Initiation." February 23, 1982. Frarie Island Nuclear Generating Station.-
- 6. " Procedures E-1.3.and E-1.4 S/G Tube Rupture." June 3, 1982.
- Praria Island Nuclear Generating Station.
l
- 7. Duke' Power Co. "EP/0/A/1800/17. . Steam Generator Tube Ru'ture." p Oconee Nuclear Station..
.~- . 8. ~INP0.~ "SOER 82-16 Draft." January 4, 1983.
- 9. INFO. " Analysis of Steam Generator Tube Ruptures at Ocones and Ginna" INPO 82-030. November 1983. ,
- 10. . Babcock & Wilcox " Abnormal Trassier.t operating Guidelines. Three Mile Island Nuclear Station - Unit One." June 1961.
- 11. GPU Nuclest Corp. " Acceptability of Intentionally Loading Main Steam Lines During an OTSG Tube Rupture Event." Document No.
1101X-5320-A18. July 30, 1982 - Engineering Mechanics section.
I
- 12. L. C. Lanese "TMI-1 OTSG Tube Rupture Basis.for Design Leak Race." August 31, 1982. SAPC/118.
- 13. M. Sanford "NPSH Requirements for Piggyback Safety Injection
, Operation - TMI-1." December 27, 1982. MSS-82-584.
- 14. J. Veenstra. " Fuel Pin Compression Limits During an SGTR Event." TMI-83-009. January 20, 1983.
- 15. S. D. Leshabff. " Mechanical Integrity Analysis of TMI-1 OTSG Unplugged Tubes." TDR 388. GPUNC. March 9, 1983. l i
l l
+ Rev. 1 4
- Rav. 2 e $a 7 *4,. 6 y w,a em em4 - -.g gues>me ,gumemm..di N --- N _
" w .~ .
,-v=
3-= - - ==.
Aav. 2 (s
.-(_ .
j Prie !.3 of 74
- 16. 'P. S. Walsh.' "TDR 400 Draf t Guidelines for Plant Operation with
~
Steam Generator Tube Leakage." PA-893. February 15, 1983.
- 17. Darrell G. Eisenhut. to E. 'D. Eukin. March 4,1983. Docket No.
50-289. ' March 4, 1983. U.S. Nuclear Regulatory Commission.
- 18. N. E. Savani, Dunn, 5. M. , Jones, R. C. , Response to GPU letter dated August 28, 1980 (TMI-1/E 1203). Document Identifier 12-11217.8-00. September 26, 1980. Babcock & Wilcox Co.,
Lynchburg, Virginia. '
+ 19. E. D. Bukill (GPUN) to D. G. Eisenhut (NRC). "RCP Trip on 25F*
.Subcooling Margin." March 31, 1983. 52n-83-017 ,
- .20s. W. Dranda n. " Saturation Margin Monitor Inaccuracy (Non-Accident Conditions)". Cale. No. C-n 01-665-5350-002.
June 14, 1983.-
20b.-G. J. Sadanskas. "IMI-1 Saturation Margin Monitor Loop Error Analysis." Cale..No. 11011-3228-009 Rev. 2. June 14, 1983..
GPUNC.
- 21. G. L. Lehmann. "0TSG Leakage and operating Limits." TDR 417. 8 GPONC. March 1983. -
- 22. TMI Unit 1. " Plant-Limits and Precautions. OP 1101-1. Rev.
14." February 26, 1982. GPONC.
- 23. " Determination of Minimus Required Tube Wall Thickness for 171-PA
- Once Through Steam Generators." 3A4 -10146. October 1580.
+ 24 J. Veenstra to L. G. Slaar. Thermal Shock Clarification For Small Areak Operating Gu14alinas." Septambar 9, 1982.
TMI-82-071.
+ 23. M.E. Kostrey, Meno to L.C. Lanase "IMI-1 Reactor Coolant Pumps NPSE limits Westinghouse Pumps" MC-1752, March 30,1983
+ 26. L.C. Pwu. GPUK Calculation # H 01x-5450-013, key. 2 " Minimum Subcoollag Margin for RCP Operation." August 5, 1983.
1
- 27. L.C. Pwu. GPUN Calculation i n 011-5450-015, " Saturation Margin Monitor Adjustaant on Elevation Difference Between the Pressure Tap and the Tops of the Hot Leg." July 29, 1983.
1
- 28. R&W Abnormal Iransient Operating Guidelines (AIOG).
74-n24158-00. Lynchburg, VA, April 24, 1983. ,
- 29. Telecon with Ralph Rosser (B&d) by L. C. Lanese. June 13, 1983.
- 30. -Babcock & Wilcox Co. " Evaluation of SS LOCA operating Procedures and Effectiveness of Energency Peedwater Spray for B&W - Designed Operating NSSS". B&W Doc. ID 77-1141270-00. Lynchburg, VA. !
j February, 1983. ,
+ Rev. 1 .
- Rev. 2 '
T ~ 'm . _ . . - _ -
-~~~"Da 406'
~
u ,. 2 Piga 44 of 74
- - 31.. L.'C.-Pwu. GPJ Cale, No. 11011-5450-014. "TMI-1 Cooldown Aate using Atmospheric Dump Valves", August 3,1983. ,
- 32. J. P.: Logetto. "TKI-1 Decay Heat Removal System".
June 7, 1983. M35-83-504.
- .' 33.~ N. G. Trikouros to 1. J. Tools. . "Tnfroid Dose Limit Ior OT3G Isolation during SG Tube Rupture", SAPC #140, dated July 13, 1983. ..
- 34. TWI Unit 1 0P-1103-6,'Aav. 24. " Reactor Coolant Pump Operation" . '
' Step 3.3.2.
- 35. J. 1. White. RELAPS Analysis of Steam Generator Tube Rupture Transients in a-Generic Lower Loop Planc. 1.P2420-4 fluclear Safety Analysis Center. Palo Alto, CA. 1983._
e 8
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, Page 46 of 74
- 1 Break Flow for Single Ruptured Tube 1850 , , , , ,
345 -
I II III B40 -
4 E 735 -
a.
N 830 m
en E 525 - -
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>=
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I: 25'F SC, PUMPS DN -
~
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lit: SO'F SC, PUMPS OFF -
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T01: 400 nEv. 7 Page 46 of 74 FIGURE 4 -
Mass and Energy Capabilities of HPIand PORV -
880 40
(.55)
/
800 - TWO HPI PUMPS #
f -
35 f c.s )
/
.,2 0 -
PORV ENERGY e
REMOVAL t* ****' ~ ' ** .. y ,f -
30 -
.* /
/ '** ..,
(1.3) ,'
f 640 -
/. **
/ - '
25 m
-/-. **, (2.s) R
.n ., a
= / . m g 550 -
/ PORV LIGUID REllEF I
- 20 o / (5.s) E d d / E '
/ ' =
~
480 -
l'
/ 15 ONE HPI PUMP / r
/ T
/ ~
400 - / -
10 j _
/ '
/
/
320 . 1_ -~'~
4 - 5 240 I I I I O
O 500 1000 1500 2000 2500 '
RC SYSTEM PRESSURE (psig)
- Energy relief is product liquid relief' capacity and *1.0 ANS entholpy of 100F subcooled water. 2535 Mw (t) i
, , . . , . ,n.,n...-. -- - -. a
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.-....,....a..m----- . - - a ;. u,,,,, _
-~ -
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F88URE 5 TDR # 406 Rev 2 Page 49 of 7t.
Time Behavior of Subcooling Margin for a Spectrum of Ruptured Tubes too
/ .
~
./ - - -
80 - W 1 TUBE I
lL -
i !
5 at 60 -
./
g _./ -
as E s g 40 3, ,
/
'S as i -
5 TUBES e'
~
g 5 TUBES /
R f -
2E -!
g p'
25 TUBES ,,"" 26 TUBES R-e e u- _, .
g , , , ,
8 12 1g
, 24 3R 33 42 48 : 54 60 t .
l TIME (MINUTES) i i
)
l l
l l
l l
93.
'i. ..
,N f TDR404 Rev.2
' 4- O' a ce 30 e -
.RCP NPSH Curves 1500 Minimes NPSH for RCP Normel 1500 Operations OP 1101 Figure 1.0-5.5
- ---- Ex:-.- :i NPSH for 4 RCP operation !
or 2 RCP per leep. !
1400
- - - Emergoesy NPSH for 2 RCP operation !
(one perleep) l 1300 -- - ~~ 25'F Subseeling Mergia Cerve.
[
Notes f!
1200 12*F and 10 psig instrument errors have been incorporated in the normal
/ ,
/
NPSM serves, while 5.5'F and 94.0 psig .
1100 errors were for the emergency NPSH serves. !
t i ,
I1000 s ! !
.-s .
E 'l 2 900 /,j
..... s ..
n 300
= /.,
s /// ._
/v g .700
/!!
///
500 i
/.ti 500
///--
/.f:,f f
l 400 -t*./ l -*
/ / i
. ,,0 ----- ---------- u .
i
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200 ,j
,00 0
100 200 300 400 500 500 Indicated RC Temperature ('F)
by e. . _. . .
.. . .. . . . . . :--:~.=~-~_'- ~'
~~
- TDR 406 )
Rev. 2 i Page 51 of 74 l o
APPENDIX A.
9 i
THI-1 SGTR PROCEDURE GUIDELINE ANALYSIS e 4
T
+ Rev. 1
- Rev. 2
, h
f'
. TDR 406 Rev. 2 Page 52 of 74 COMPARISON OF GUIDELINES AND EP 1:02-5 iREV. - 16 TO -
TIC REQUIREMENTS OF VARIOUS SOURCE DOCUMENTS A'. 0 SCOPE The purpose of this Appendix is to, compare the guidelines of this TDR to the current revision (16) of EP 1202-5, OTSG Tube Leak / Rupture, and' '
guidelines, requirements, couaniements or recommendations from various sources. The sources reviewed were the TMI-1 Anticipated Transient Operating Guidelines (drsft of 15 May 1981; hereinafter referred to as
'ATOC), " Clarification of TNI Action Plan Requirements" (referred to as
, NUREG 0737), aid the Safety Evaluation Report related to restart of t
Ginna (Raf 1) (referred to as NUREG 0916) and the INPO draft Significant Operating Event Report of.04 January 1983 concerning steam generator tube leaks (referred to as SOER).
With respect to NUREG 0737 and 0916,'only those requirements or commitnants directly related to a tube leak emergency procedure were considered. With respect to ATOG, only the followup guidance for tube .,
leaks was considered, and then only if it differed from the guidance '
+ in the latest approved tube rupture procedure (EP 1202-5 Rev. 16). .
+ With respect to the SOER, only the recommendations related to procedures were considered.
.- + The'results of this comparison work are summarized in Table A-1.
e e
4 9
4 e
s
' A-1 .
6
+ Rev. 1
- Rev. 2 e
. n . - -- . - _- w ._g_-___~_ _-..n _ _ _ - . . - - - - - - - - - - - -
. . _ - . - - . . . .. .= - - -
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- {- -
TDL 406 Rev. 2 Page 53 of - 74 TABLE A-1 .
Comparison of TDR 406 Guidelines to Other Sources Source Addressed By Requirement ATOG 0737 0916 SOER 1202-5 406 Comeents Run RCP's with -X X X X low RCS pressure i
RCP restart I X X X X No spe.:ific guid-ance provided for '
RCP restart with a " solid" pressuriser.
Subcooling-marsiu I I I I HPI throttling I X- I X
, Steam line X X X X ATOG does not '
flooding recognise THI-1 8 capability to .
flood steam lines without daname.
- Cooldown of - X X X Guidelines provide .
damaged OTSC for continued steaming of afa facted OTSG for cooling except when OTSG isola-tion is reevired.
Specify entry I X X- I Symptoms, thre shold Method for plant X I X X cooldown following SGTR Plant cooldown X X X ATOG refers to following SGTR excessive heat with stuck open transfer section
. SG safety valve EP 1202-5 and guidelines pro-vide means for minimizing prob-ability of lifting a SG relief valve.
A-2 -
+ Rev. 1
- Rev. 2
.= _ _ __
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~.
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TDR 406 Rev. 2 Pa8* 54 of 74
-+1 .~ TABLE A-1 ,
E (continued) .
Source Addressed 3v Requirement ATOG- 0737 0916 SOER 1202-5 406 Coimeent s . _ . . .
L
'Affected SG X X X
- pressure ccatrol
- l- OTSG tube'to X X l
shell differen-
'tial temperature , i 1
l
+; Criteria for X
'using ADV's in _ Guidelines h e ~ ~~ ~ ~ ~ '
same that- , , _ _ , _ _ _ , _ _ , . _,,,I preforence to- steaming to main condenner
- condenser is_. , , _
always preferable to steaming to atmosebers. l
}: .. .
! Consider multiple X X Guidelines pre- '
',, tube ruptures-- , pared in con- _ , , , .
sideration of
. . - -Consider tube leaks X X these two cases.
in both OTSG's
+ EPI.en inadequate X X .Not specifically
+ SCM stated in TDR
, , , , _ 406, but it is an implicit re-quirement of the HPI throtti,ing,_ .
criteria.
Consider excessive X X Overfeeding con-
,. primary to secon . '
sidered by __ _,,_
dary heat transfer . , _ , . _ _ ,_.
EP 1202-5.
Consider loss of X~ X Guidelines make no offsite power particular dis-tinction between offsite power available or
. unavailable, but
, c. . they do provide j .
guidance if the
,, , , , equipment dis-abled by LOOP is unavailable.
17 A-3
+ Rev. 1
- Rev. 2
W-. .
(O TDR 406 Rev. 2 Page 55 of 74
+ TABLE A-1 ,
(continued)
Source Addressed By Requirement ATOG 0737 0916 SOER 1202-5 406 Consnent s .. _ .
1-OTSG 1evel X X X control Primary pressure X X control with-out pressurizar -
sprsy -
Isolation of X X X affacted OTSG
. 4
. .g 4
. m m. .
QO eM"
- @m e eeDen me ,
. . e N 6e 6 me O
eummemme e une . .
A-4 *
+ Rev. 1
- Rev. 2
L- ',
7.*7 ' _.
d , . . . . -
- DA 406
.ry new. 2
,' k Page 56 of 74 A.1- Interpretation
- A.1.1 "Requiresent" Coluan .
These are. paraphrased descriptions of guidelines, requirements, commitments, or recommendations from source documents.
' A.1. 2 " Source" Columns These columns define the origin of the requirement considered 16 this - - ^
comparison. '
A.1.3 " Addressed 3/" Columns e
These. columns defina che document which answers the requirements. If '
a mark appears in the 1202-5 column without a corresponding mark _in the TDE 406 column, it means that the guidance-in EP 1202-5, Rev. 16
'should be retained in the revision that incorporates tne guidelines of - - -
Section 4 of this TDA. If a mark appears in both columns, it generally means that the guidelines in this DR supercede the guidance in EP 1202-5, Rev. 16. If a mark appears in the TOR 406 column alone, it denotes a new guideline.to be incorporated into the revised EP 1202-5.
A.1. 4 Comments
~ This column provides additional information if necessary. ~The guidelines in this EE supercede the guidance in EP 1202-3, Rev. 16.
If a mark appears in the TDA 406 column alona, it denotes a new guideline to be incorporated into the revised EP 1202-5.
e W
A-5
+ Rev. 1
- Re v. 2
,v._-.-..,.
. - ~ - - -
- . _ _ __n _m.n n__ . _._
i* ,
s,. ,
. ( (m TDR 406.
Rev. 2 Page 57 of 74 e
4 o
9 APPENDIX 3 PROCEDURE CHANGE SAFETY EVALUATIONS t
e a
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+ Rev. 1
- Rev. 2 OO_A . J.*
- 9 D8 OW ps '
, Igg 4o3 x (J Aav. 2
-fage 58 of 4
~ 3. 0 PAOCEDURE CHANGE 3MEIf EVALdAIIONS '
The purpose of this Appendix is to address ene safetf implications of the key changes required co'laplement the tube rupture guidelines described in this TDA.-
- 1. . ACP trip on loss of subcooling margin (SCM) 1 2.- Change in SCM
- -3. Shell/ tube delta I of 70t* during emergencies
. 4. Revised ACS NP3M curve -
5.. Relaxation of fuel pin in compression limits
'6. OI3G isolation criteria
- 7. ECP restart criteria
- 8. EPI throttling at 0" instead of 100 inches.
- 9. Leaving ADV open when no OISG heat sinks and ACS is above 1000 ,
psig.
- 10. Isolation of core flood tanks -
Each of these items is addressed below:
3.1. ECP Irip on Loss of Subcoolina Martin
+ In a letter dated March 4, 1983 to E. D. Eukill (Rev 17), the ARC
.supercaded the actions required in IE Bulletins79-05C and 79-06C.
The staff has instead concluded that "the need for ACP trip following a transient or accident should be determined by each ape l l**t1** **
- case-by-case basis considering the Owner's Group input. For several years, the B&W Owner's Group has supported the concept of. ACP trip on loss of subcooling margin. In Reference 19, GPdAC informad the NAC of their reasons for revising the trip criterion to loss of subcooling
.,_ margin. . The safet/ evaluation for this change has been transmitted separately to the IMI plant staff for review and approval. ,
B. 2. Chanae in Subcoolina Marain GP3NC has evaluated the' instrument error associated with the subcooling margin monitor and'alars.. Under normal containment
- conditions, the loop arror is + 10.1pc (Ref. 20a & b). Onder the t temperature and radiation envi7onment of a small break LOCA, this error is no worse than - 21.7E*. .Iha basis for the original SCM was 3-1
+ Aav. 1
- Aav. 2 l
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C Aav.- 2 Page 59 of 74
. *- 5F' geometry correction plus 45F' string : inaccuracy. Ascent
- calculations (Asf. 27) have shown enat onif 1.30F geosacrf -
- . correction to the top of the hot leg is required. Inerefore safec/
margins are not decreased by this change.
Acompletesafschevaluationhasbeenpreparedandtransmittedtothe site separately for review and approval.
- 3. 3. .
Chanae in Shell to Tube Delta T The existing emergency limit for tube /shall delta I at TMI-1 is 100t'. The 11 alt is being-revised to reduce censile stress on lear,ing CISG tubes. Previously the limit of 100E' was based on stresses to
' intact tubes.-.Indeed, the limit has been increased to 150F* by .l Babcock & W11cos; and, it is valid in the absence of degraded OTSG l tubes. . l Ine more restrictive '70t' limit increases plant safetf liales bf !
reducing the. likelihood of propogating a crack. Inis analftical work is documented in Reference 15. This change can be made under the provisions of 10 CFA 50.59 becausa it does not affect technical.
specifications. Since tube stresses are reduced, plant safety liales I
- are increased and the two additional criteria of 10 CIA 50.59 are also met. Namelf, there are no new accidents introduced into the plant that have not been previously analfzed. Since shall to tube delta T has not been explicitif addressed in the FSAA, existing plant safet/
margins have not'been decreased. In fact, plant safety margins have been increased since the allowable delta I has been decreased.
3.4. ACP NPSE Limits Reduced NPSE limits bring the pump closer to a point of cavitation. - - - - - - ^
Eowever, NPSE requirements have been reduced for lower temperatures as
+ determined by the pump manufacturer Westinghouse in Aeference 25.
, Margins have been modified based on safety margins identified by the pump aanufacturer, therefore, the probability of pump cavitation has
~
not been increased and plant safety margins are protected. Neither are technical specifications affected. The operation of reactor coolant pumps at low AC3 pressures does not introduce any new accident or transient other than those already analysed in the ISAA. Pump operation is allowed at these lower RCS pressures but at a higher subcooling margin. Since real plant subcooling mardin is still being .
maintained as discussed in iten 2, there is no reduction in plant safety. Operation of the reactor coolant pumps increases plant safety
- margins with respect to thermal shock, increased 3dB ratios, and -- - -
Laproved capabilities for dagassing the reactor coolant system under tube rupture conditions when miniana subcooling margins are being maintained.
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TDR 406 Rev. 2 Page 60 of 74 5.5. Fuel Pin and Compression Limits " *
+.' As addressed in Reference 14 B&W has recommended that fuel pin'in compression, limits be waived during certain plant transient conditions including steam generator tube rupture events. Yuel pin in
-compression limits have been established in order to maintain cladding integrity. Weiser of these limits does not reduce pin integrity although reanalysis by 5t,W may be required when fuel compression limits have been waived. Since cladding integrity will have to be addressed each time these limits are violated, the demonstration of ,
acceptable clad integrity will be made. No new accidents or transients will be introduced then have been previously analysed in the FSAR. Similarly, plant safety margins will not be reduced, namely, cladding integrity will not be challenged.
J B. 6. - UTSG Isolation Criteria -
The existing steam generator tube rupture pro'cedure. EP-1202-5, allows
.the operator to isciate the affected steam generatcr anytime RCS
. pressure is below 1000 psi. The revised criteria would allow steaming of the OTSG until BWST level is 21 ft. or radiation limits approach site emergency limits. Steaming of the OTSG introduces the potential for increasing offsite radiation doses; however, these limits will be maintained within the requirements of 10 CFR Part 20, It should be noted that the isolation of the steam generator on h.sh radiation is keyed toward's maintaining Part 20 limits. Steaming of the generator when possible increases the. chances of preventing major offsite releases since flooding of an OTSG can result in liquid relief out of the- steam safety valves with the possibility of safety valve failure. _
The value of BWST level at 21 ft. is sufficient to assure a source of . ..
water for the ECCS pumps. The value of 21 ft. allows sufficient inventory to flood both steam lines and allow the plant to be placed on feed and bleed cooling in the recirculation mode from the RB building sump (Ref. 12, 13). It should be further noted that the doses associated with a steam generator tube rupture were increased when the requirement for maintaining subcooling margin was introduced -
into the plant procedures following the TMI-2 accident. At that time the issue was addressed in writing to the NRC staff (Ref.,21) justification for the change was that Part 20 limits were being maintained. This criterion is still being maintained with the change in OTSG isolation criteria. ~
These changes can be made under 10CFR50.59 because safety margins are not decreased. Technical Specifications are not affe:ted by this change. No new accidents or transients are introduced which have not-been previously analyzed since this guidance is intended to deal with events which are beyond the design basis of the plant (i.e., tube rupture without ' condenser and RCP's).
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- 5. 7 . - + RCP Restart Criteria
+' Ibe RCf Aestar't Criterion assures that the pumps are not restarted
+ until the core is-adequately subcooled. (Note that there are other
-+
+ RCP restart unrelated to this criterion). Reference 19 and Sections 5.1 and 8.2 demonstrate that the core is adequately subcooled with
+ - 1CF' n running and a 25F' 'subcoo11ag margin. No Technical
+= Specifications are affected. No new accidents.or cransients are
.+ introduced into the plant; no safety margins are decreased and no
+
.+ accident consequences are increased. Allowing an earlier pump restart
+
gives .the operator greater control over the plant since forced flow is preferable to natural circulation cooling. Ibis change can therefore t
+' be made under the provisions of 10CFR50.59.
B.8 + epi ~Ihrottlinz at 0 inches Indicated Level
+-
.+ The safety aspects of throccling EPI on 25f' subcooling margin are
-addressed in section g.2. Cora coolability is not dependent on the' "
_+ pressuriser level at which EPI is throttled 1.e., core cooling is only
+ dependent on an ladication that the core coolant is subcooled. The
+ basis for requiring pressuriser level is so that the existing
'+ pressuriser heaters are covered with water so that they can be
+ energized. Energising the heaters before they are covered causes them
.+ co burn out. On the other hand, there is no need to refill the
+ pressuriser to the 100 inca level at full EP1 flow. In fact, this
+ . flow race is undesireable for two reasons. Aapid filling of the
. + pressuriser causes an RCS pressurisation during conditions when
+ pressuriser sprays are unavailable. Insurges to the pressuriser
+ compress the steam space. -Pressuriser pressure aust be reduced either
+ 'by sprays (if available) or pressuriser venting (vent line or PGAV).
+ Controlling the EPI flow minimizes the insurge rate, and hence, the
+ pressurisation. This reduced pressurization provides more margin to
.+ the 1001' subcooling curve thereby minimizing challenges to the
+ thermal shoct/ brittle fracture limit. .
+ Ihis change does not represent an unreviewed safety question because
+ 1. No change to tha Technical Specifications is required .
+ 2. No new accidents are introduced to the plant (the operator is still required to cover the pressuriser heaters before energizing them), and -
+ 3. The consequences of previously analyzed accidents / transients is '
not increased. It is less 11kaly that the operator will violate the 100f' subcooling margin. Core coolability is not dependent on established pressuriser level, but only an adequate subcoollag margin.
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(C '
(-} TDR 406 Rev. 2 Page 62 of 74 ;
l B.9 - + ADv's Doen when RCS is above 1000 osin with no OTSG-Heat Sink?s
+ This TDR provides guidance for certain situations well beyond the
+
design basis.. One such situation is the case where the plant is one
+ feed and bleed cooling, but RCS pressure is above 1000 psig. This
+
condition can result in liquid relief out of the OTSG safety valves.
+ Opening the ADV's is the preferred course of action because it
+ minimises the chance of an uncontrolled blowdown though the OTSG
+ safety valves. This condition is well beyond the plant design basis.
+ Plant Tech Specs are not affected by this procedural step. Therefore
+ the change can be made under the provisions of 10CFR50.59.
+ Beyond the consideration of whether this change can be made under the
+ provisions, of 10CFR50.59, it is_ believed that opening the TBV's/ADV's
+ is prudent and reduces the risk of an uncontrolled release to the
+ environment. -
3.10
- Criteria for Core Flood Tank Isolation -
1 The purpose of the core flood tanks is to assure core cooling for LOCA's.in which: 1) RCS pressure is below 600 psig, 2) EPI cannot provide core cooling, and 3) RCS pressure is too high for the LPI system to operate. The only situations when these conditions occur
- are: 1) design basis LOCA's, which EPI does not initiate before the
- core begins to uncover and:2) core flood line break accidents with an HPI failure, and 3) small break LOCA's in which the break is just 1arge enough to remove decay heat, but not to depressurize the RCS.
For the large break LOCA situation, CFT isolation is not a principal concern. The operator should isolate to prevent nitrogen introduction into the RCS once the tank is empty.
- For small break LOCA conditions, a subcooled RCS means that there is sufficient heat removal. In the pressure ranges in which core flood tank isolation is of interest, one HPI pump supplies sufficient flow to keep the core covered (500 gym). If the RCS is 25F subcooled with :
- the RCS below 700 psig, then the CFT can be isolated. j The core flood tanks also function in one non-LOCA situatiott - steam line break accident. For large steam line break accidents, the CFT's provide shutdown margin assuming the most reactive rod is stuck out. -
Therefore, CFT isolation cannot occur until either HPI is operating and providing a source of borated water to the core or until all rods ,
7* have inserted. If both these conditions are met, then a plant procedural change can be made without introducing an unreviewed safety question.
I i
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% y 9) e Q TDR 406 ,
Rev. 2 -
Page 63 of 74 eemme APPENDIX C GUIDELINES FLOW CHART G
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- TDR 406 '
Rev. 2 Page 64 of 74 C.0- GUIDELINES FLOW CHART -
The flow chart in this section shows the major milestones and decision points on the path from operation at full power through the development of an OTSG tube leak / rupture to inspection and repair of '
the damage.- The flowchart is not meant to be an exhaustive treatment
+. of all actions required to reach cold shutdown, rather it is the
+ framework upon'which a procedure can be constructed.
C.1 INTERPRETATION .
Diamond boxes are decision points. The path taken out of a diamond '
depends on the answer to the question posed in the diamond. Boxes i enclosed by a single line represent steps that take seconds or minutes to execute. Boxes enclosed by double lines represent tasks that may require minutes to hours to accomplish. For the sake of simplicity, certain steps that will be required in the procedure have been omitted (e.g., confirming reactor trip, or making radiation surveys of the secondary plant).
The decision points inmediately following a double-line box are meant to force the operator into a " thought-loop" so that if conditions
-cha' age, the operator may select an alternate, more appropriate
~ '
cooldown path. For instance, while cooling down on forced flow with a tube leak in excess of 50 syn, the operator should continually inquire as to whether the Reactor Coolant System pressure and temperature are within the capability of the Decay Beat Removal System. If so, when the operater should obviously change the RCS heat removal mode from steaming the CISC's to using the DERS. If not, then the operator should continue to ask whether the RCS conditions are suitable for forced flow coolius vi's OTSC's, i.e., is subcooling inadequate, are the OTSC's available/0K for use, are the RC pumps available. If the ,
answers to these questions always no, yes, and yes, then continued '
forced flow cooldown is acceptable. If any of the answers change, then the thought flow breaks out of the loop and presents the operator
- . with new criteria for selecting an alternate cooldown mode.
+ This thought loop" philosophy should be incorporated into the
. procedure revision.
C.2 + PROCEDURAL OBJECTIVE
+ The objective of the tube leak procedure is to expeditiously cool down
+ and depressurise the plant so as to minimise primary to secondary
+ leakse and thus, it is hoped, offsite doses. The process involves
+ recognition of the event, shutting down the plant, and cooling down
+ the plant to the point where the Decay Heat Removal System can remove
+ core heat.
e e C-1 s
+ Rev. 1 i
- Rev. 2
. ' ' ~ ~T . %_ .. .. . _ _ . . . . . ~ _ . . _ . . . . . ..
T3R 403 O 1.2 Page 65 of 74' C.3 + ENTRY POINT .
+ The procedure will be entered when a primary to secondary leak is t
+- en ountered that requires the plant to be shut down. The symptoms of !
+ a tube leak requiring shutdown are described in TOR 400 (Ref. 16).
C.4 + FLANI SEUTDO~dN .
+
The rate of plant shutdown from 100% power will be determined in part j
+ by the magnitude of the RCS depressurization due to the leak. If the t
+ leak is small (the Makeup system is able to keep up withit) .then the l
+ plant can be shutdown at a rate commensurate with equipment
+ .
capabilities and, to a certain excent, the leak race. When the
+ reactor and turbine are off line, the plant is ready to enter the
+ cooldown phase.
+ Eowever, if the leak results in RCS depressurization to the trip.
+ setpoint, the reactor and turbine will be off line lamediately. The
+ ensuing transient will have to be dealt with and the plant status will
+ have to be evaluated prior to the cooldown.
C.4.1 + Preparation for Cooldown
+ If the shutdown transient results in a loss of subcooling margin, RFI
+ sust be lamediately actuated and the Reactor Coolant Pumps (RCF's)
+ aust be immediately tripped. The CISG's must then be evaluated for :
+. suitability as heat sinks for the 1C8.
+ If the shutdown transient does not result in a loss of subcooling
+ aargin, the DISG's must still be evaluated for suitability as RCS heat
+ einka.
+- If neither OT33'can be used because of high offsite doses or low B'JSI
+ level, then the cooldown will proceed directly using the EFI " feed and
& bleed" asthod. -
+ For the balance of tne discussion in this section, assume that RFI
+ " feed and bleed" is unnecessary.
+ If the RCF's are off, Emergency Feedwater flow to the OTSG's must be
+ confirmed. The ICS will automatically control OISG 1evel at 501 on
+ the Operating Rande if the RCF'h are off. If subcooling margin is *
+ 25F, the operator must manually raise the inval to 95% to promote .
+ two phase natural circulation in the 105. - ' ' " - " "
+ -
Since a forced circulation cooldown is the most preferred aode, the
+ RCS conditions should be evaluated for RCP restart. If subcooling
'+ margin is regained and the RCP NPSE limits are met, 2 RCF't should be .
+ restarted. If the pumps cannot be restarted, the cooldown must
+ proceed by natural circulation.
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TJA 406
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. Page 66 of 74 C. 5 + PLANr C00LD3'fW
- I I
+ During the cooldown, RCS conditions must be continuously evaluated to
+ ensure that the cooldown mode is appropriate and to determine whether
+ conditions are suitable for the Decay Beat temoval System. {
+ Eagardless of cooldown mode, the following items, may be encountered
+ while coollag down. -
C.5.1
- HPI Throttlinz-
- The existing EPI throttling criteria are unchanged with the fo110wL4
- exception RPI may be throttled when suecooling is regained and
- pressuriser level comes on scale.
C.5.2 + orSG Steamina
+ The affected OTSG may be staaned for RCS heat removal purposes, but ic
+ sust be steamed to avoid lifting the Main Steam safety valves, prevent
+ premature Steam line flooding, keep orSG pressure less than RCS-
+ pressure, and control OTSG tube to shall differential temperature.
C.5.3 + orSG Shell to Tube Dif farential Temperature -
It is necessary to minimize shall to tube differential temperature to-
+ minimise tensile stresses on the OTSG tubes. As noted above, steaming is one way to accomplish this; another is to decrease the cooldown
+ rate; a third is to use main Feedwater to cool the lower downcomer.
C.5.4 + orSG Pressure Control when RCS Pressure is Greater Than 1000 psis
+ During a natural circulation cooldown or an RPI feed and 'o' lead
+ cooldown, RCS pressure asy stay high. Kaargency feedwater, can be used
+- to quench the steam space; if the OTSG is flooded, inventory can be
+: relieved via the Turbine Bypass Valves or the Atmospheric Dump Valves.
C.5.5 + Cooldown Race
+ The cooldown rate should be lialted to less enan 1.6 F/hr to avoid
+ reactor vessel brittle fractura concerns. It may not alvs/s be
+ possible to observe this limit due to the offacts HPI cooling and the
+ occassional necessity to staae the damaged OTdG. .
C.6 + 11IT POINr. . .
+ The operators exit the procedure when the RCS heat sink becomes the
+ Jacay Heat memoval System.
6 C-3
+ Rev. 1
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't
! h
. .'t testa n
- cluta -w-. e
- litillRL C I N P=temS ,*
3 SINGLE & MUt.TIPLE TUBE RUPTURE GUIDELINES ' '
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k i TDR 406 Rev. 2
.Page 68 0f 74 APPENDIX D SIMPLIFIED EVENT TREE s
e W .~-
+ Rev. 1
- Rev. 2
.cv .. - -
'6- .
p s L' TDR 406 Rev. 2 Page 69 of. 74 D.0 SIMPLITIED EVENT TREE N event tree on the following page shows possible combinations of circumstances that were considered that resulted in the guidelines
. presented in this TDR.
The guidelines explicitly stated in section 4, when incorporated into a revised OTSG Tube Leak / Rupture Roergency Procedure, will enhance the
- capability of THI-1 to deal with an OTSG tube leak. The purpose of this section is to describe the features of the revised procedure.
The discussion which follows assumes that the logic presented by the
-flowchart depicted in Appendix D is adopted for the revised procedure.
9 am 9
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e 1
4
. D-1 ,
+ Rev. 1
- Rev. 2
1 ==-"; _
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1
+,.f.
TIGURf D-1 TDR 406 EVENT
,r- .
(,R:v.2
-k '
Tnst RCPS Page 70 of 74 NO.
SIMPLIFIED OTSG ***"
CONDENSER '
EVENT TREE - -
AvAstaeLa
- 1 LSAsuseg '
0734 .
CONDENSER AVAILABLE 2
NO RCP's 4 .
ACS 3 SUSCOOLED CONDENSER AvAILAsts A
RCP"S Avan Am e SOTH I 0780*S N
. A80Vt CONDENSER - 1060 4 AVA3LA8LE ggLow g 1060PS HO ACP's A80VE 1050 Pat BELOW 7 10s0 Pat Tutt AWPTURE
.. CONDENem' AVAILAELE 8
RCP'S AvAALAeLE i
. I 1LSAsuNG OTse A40VE I
CONDENSER avalLAOLE '
BILOW 10 4 C40 PSI NO MCP's Agoyg 1050 PSI SUSC00UNG RELOW m 1060 PBt 11
, LOST CONDENSER
- AVAsLAeLE 12 .
RCP"5 AVAILAaLE GOTN 07801 13 LEA 8C A80vt CONOENSgg 1060 N
- BELOW 14 1050 PSI NO ACP's agoyg 1060 Pt' M SELOW 18 CAPACITY 1050 REQUdtEO
-- w,-,,< n,- e,., -ww-w,- -----w,w,-,we,-,-een.e---r---e - ~~ - - --e-w- - -., ,..,w~~-e -- ,,--,v ,
p_ . 7 ,s _ _m.m m ._
._ ..-._.. ~ - .--_ ... . . . . . . .
s'
..!= .
. g
+~ (' >
TDR 406 Rev. 2 Page 7} of 74 AFFENDIX E
, FROCESS COMPUTER OUTPUT 0
. .*e .
+ Rev. 1
-- . -- _
- Rev. 2
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t !. o
{'.
C f .7!. ~ '
TDR 406 i Rev. 2 Page 72 of 74
+ PROCESS COMPUTER OUTPUT AND ALARMS
+ Scope The process computer will have the following information available with alarms as noted:
Subcooling margin OTSG Tube to Shell Differential Temperature
+ Subcoolina Martin Alarm Subcooling margin will be computed for each hot leg and the average of the five highest incore thermocouples. The process computer should trigger an alarm state ift SCM 25F' OTSC Tube to Shell Differential T e rature Calculate shell temperature as follows for each OTSG if all shell thermocouples are operable:
Tshe11 = 0.242 Ti + 0.176 T2 + 0.201 T3 + 0.143 T4 + 0.238 T5
'~ ~
Tables E.1.4.1 and E'.1.4.2 define acceptable substitutes for various failed thermocouples and combinations thereof.
Limiting the alarm state to conditions when T eold is <53:; 3.nhibits the alarm during normal operations.
k a !
\
4 E-2 .
. 1
.2 .
+ Rev. 1
. . . , , . . . . . . ~ . . . . . . . . . . . . . -
- Rev. 2
.. - - . . - . . - - .- - - -- ~ ~ ~ --- - -- -
n -- - m .a. .u
, _ .. . .~.
i
.c .J.
=
.g p gag gn3 Aav. 2 tage 73 of 74
+. Iable E.1.4.1 Shell Thermocouple Substitution .
Failed l Substitute l T/c l r/c i i
I5 i T4 I I I4 T5 T3 I2 I I
T2 1 0.5 (Il + I )l l 3l T1 l I2 l l
I4&I3 l No Cale i I I '
I3&I2 -l T1 l I
. T3&T1 T2 l I2 &T1 T3 i
II &I2 & T3 i Ao calc
+ liide Aange Tcold should be used in determining DISG tube to shall
+ differential temperatures. Normally, use the wide range input f rom
+ IL-1-5A&E and Id-3-5A&s, although II 959 and I2 961 can be used in
+ certain cases. Iable E.1.4.2 defines the data sources.
+ For'each Loop, Calculate Shell to Tube I as follows:
+ TI -5 " Ishall '" Iceld
+ Tr.3 should trigger an alarm state if
+ Tcold 5351' g Tr.g 70f'
\
. 2-3 S
+ Aav. 1
- Aav. 2 n __ _ _ - .
. _ , . . , . _ . . . - . . - - . . . . . . , . ~ . . . . . . . - - - . - -
l 1 , 'd . ,
8 TDR 406 Rev. 2 Page 74 of 74 Table E.1.4.2 Wide Range Teoid Input
- RC-P-1 Teoid A B C D A Loop 3 Loop 0 0'0 0 Avg A Avg B 0 0 0 X Avg A TE 4 5B 0 0 X 0 Avg A TE 2 5B -
0 0 X X Avg A
- Avg 3 0 X 0 0 TE 4 5A Avg 3 0 X 0 X- TE 4 5A . TE 4 55 0 X X 0 TE 4 5A TE 2 5B 0 X X X TE 4 5A Avg 3 ,
I O O 0 TE 2 5A Avg 3
,. X 0 0 1 TE 2 5A TE 4 55 K 0 X 0 TE 2 5A TE 2 55 -
X 0 X X TE 2 5A Avg B X X 0 0 Avg A Avg B .
X X 0 X Avg A TE 4 53 .
X X X 0 Avg A TE 2 5B -
X X X X Avg A Avg 3 .
0 =. Pump Running X= Pump Off Avg A = (TE 4 5A + TE 2 5A)/2 Avg B = (TE 4 5A + TE 2 5A)/2 TE 959 May be substituted for TE 2 5A TE 961 May be substituted for TE 4 5B E-4 + Rev.'l
- Rev. 2
__ .