ML20080P501
| ML20080P501 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/12/1983 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20080P499 | List: |
| References | |
| TDR-406, NUDOCS 8310070082 | |
| Download: ML20080P501 (75) | |
Text
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BUDGET 12d006 1
la TECHNICAL DATA REPC FQ A[f] l p g ACTIVITY NO.
PAGE OF PROJECT:
IMI-1 dlh i i IU 3afet/ Anal / Plant Cont Ul
__DEPAf3fjMENT/SECTION 4
DATE REVISION DATE
- b DOCUMENT TITLE:
SG rube Rupture Procedure Guidelines ORIGINATOR SIGNATURE DATE APPROVAL (S) SIGNATURE DATE f /
.g c. ?-
l-n Is @ 4 #. A K a%/o i
kw k om 2-t 2 -E13 APPROVAL FOR EXTERNAL DISTRIBUTION DATE T%%M~
5.,z. s s o
DISTRIBUTION ABSTRACT:
R. W. Bensel this document provides technical guidelines for dealing D. J. Boltz with single and multiple tube ruptures.
A significant I. G. Broud ton improvement in procedures will result from reduction of h
4 M. Campagna the minimum subcooling margin and RC pump trip on loss P. R. Clark of subcoollag margin, walver of fuel-in-compression J. J. Colltz limits, and revised RCP NP3H limits. Other benefits can I. R. Finf rock be derived from revision of the RC pump restart celteria I. L. Gerber and from additional duidance regarding OISG steaming and R. J. Glaviano isolation. Fina11, revised guidance is provided for f
R. W. Keaten preventing tube leak propagation.
It is recommended G. Lehmann that the tube-to-shell delta r be limited to 70F0 D. T. Leighton during tube rupture events.
B. Leonard W. W. Lowe Revision 2 to this TOR includes the following J. G. Miller recommendations for procedural revisions, some T. Moran of which hava alreadf been incorporated in M. Nelson EP 1202-5.
S. Newton M. J. Ross
- 1. Isolate the DISO's on a measured or projected H. B. Shipman dose rate of 50 mrem /hr whole bodf or 250 mrem /hr D. G. Slear th rold dose.
f O. W. Smfth M. J. Stromberg
- 2. Stop the non-83 HP1 pump if the RCS is cooling R. J. Toole more than 1002/hr.
P. S. Walsh Dr. R. N. Whites: 1*
- 3. Priorities should be spelled out in EP-1202-5:
R. 2. Wilson a.
Minimizing 3CM has a priority over mini-mizing cooldown time, b.
Keeping OrSO level below 95t:is less im-portant than control of the RCS cooldown rate.
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Company Document No.
Sheet Rev.
Title QCL G PJ W TOR L40 6 1Y 2-
"S G TAe. A & ' A"'d"e "lA Gv,k Ik es
7if pu.,e3 SpecialInstructions References (b7M LS-TDC.-
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i TDR 406 Rev. 2 Page 2 of 74 ABSIRACI (Cont' 3)
- 4. Initiate the DER system at 300P under tube rupture conditions.
- 5. Trip the reactor if pressurizer leval cannot be maintained above 200 inches with two HP1 pumps on.
- 6. Raise the unaffected OTSG 1evel to 954 before raising the affected OI3G 1evel to 954 unless incore temperatures are not decreasing and there is no OfSG heat transfer.
- 7. If RCPs are not telpped within two minutes of a loss of SCM, maintain 1 RCP in each loop running.
+ Rev. 1
- Rev. 2
T0R 406 Rev. 2 Page 3 of 74 E
E SG IJBE RJPIURZ PROCE0JAS GJIDELINES IJR f/406
+ Rev. L
- Rev. 2 i
TJR 406 Rev. 2 Page 4 of 74 i
IABLE 02 CONfENTS Page fable of Contents.........
4 List of Figures........
7 List of Iables 8
Summary of Changes 9
- 1. 0 INTR 00JCfl0N AND BACKGROJND.
13 2.0 TECH FJNCTIONS SGIR PROCEPORE GJIDELINES DEVELOPdENT PROGRAM..
14 2.1 Development of Design Basis Guidelines..
14 2.1.1 LLterature Search 14 2.1.2 LLmiting OTSG rube Stresses 16 2.1. 3 Steaming, Isolation and filling of the Leaking OTSG 17 2.1.3.1 Steamlng, Isolation and Filling wLth both OfSG's Leaching...
19 2.1.4 Minimum Allowable Subcooling Margin 19 2.1.5 Waive Fuel in Compression Limits..
20 2.1.6 Reactor Coolant Pump NPSH Limits..
20 2.1.7 Procedure Entry Point Condition 20 2.L.8 Simulator Experience.
21 2.1.9 Emergenef Limits for Decay Heat 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 Criterla for Feed and Bleed Coo 1 Lng 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 TO EXISIING PROCE00RE 27 3.1 Sasic Plant State..
27 3.1.1 Assumed Plant Conditions.
27 3.1.2 Tube Rupture Guidelines for Loss of Subcoolind 27 3.1.3 Revised Equipment Limits & Operating Procedures...
28
+ Rev. 1
- Rev. 2
IJR 406 Rev. 2 Page 5 of 74 s
B IABLE OF CONTENf3 (Cont'd)
Pagd 3.2 DLscussion of Guidelines.
29 3.2.1 Immediate Actions 29 3.2.2 Followup ActLons - Subcooling Maintained and RCP'h Avallable..
29 3.2.2.1 Maintain a Minimum of 25F*
Subcooling Mard n 29 i
3.2.2.2 Steaming / Isolation Criteria for the Affected OTSG 29 3.2.2.3 Shell to Tube Delta I 30 3.2.3 Followup Actions (Automatic Rx Trip has Occurred).
30 3.2.3.1 Irlp with a Loss of Subcooling Margin.
30 3.2.4 Followup Actions for Loss of Subcooling...
30 4.0 *31MULAIOR IRAINING EXPERIENCE 32 4.1 IntroductLon...
32 4.2 Results.......
32 4.2.1 Januarf 1983 Iraining 32 4.2.1.1 Comments.
33 4.2.2 June 1983 Training........
33 4.2.2.1 Control of RCS Cooldown Rate.
33 4.2.2.2 Plant Stab 11tzation 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 Celterion.
35 5.0
- SINGLE AND MJLIIPLE SGTR GUIDELINES 36 36 5.1 Scope.
3.2 Guideltnes and Limits.
36 5.2.1 Subcooling Margin Requirements....
36 5.2.2 Reactor Coolant Pump Irip Criterion 36 5.2.3 Reactor Coolant Pump Restart Celteria.
36 5.2.4 Reactor Coolant Pump NPSH for Emergenef Operations.
36 5.2.5 High Pressure Injection Throttling l
Criteria.
36 5.2.6 O IS G Le ve l......
37 5.2.7 OISG Isolation / Steaming Criteria.
37 5.2.7.1 Pressure Control of an Isolated 0TSG.
37 5.2.8 Cooldown Rate During a Tube Leak Event 37 5.2.9 OISG Shell-to-Tube dif ferential Temperature Limit 38 5.2.10 Cooling Mode Jhen Both OISG's are Unavailable for RCS Heat Removal...
38 5.2.11 Core Flood Tank Isolation 38 5.2.12 Guideline Flow Chart.
33
+ Rev. 1
- Rev. 2 l
IJR 406 8
Rev. 2 Page 6 of 74 TABLE OF CONfENTS (Cont'd)
Page
- 6. 0 CONCLUSIONS AND RECOMMENDATIONS.
39
- 7. 0 REFERENCES 42 AppendLx A:
Comparison of Guidelines to INP3 and dRC Recommendations Appendix B: Procedure Change Safety Zvaluations Appendix C: Guide 1Lnes Flow Chart Appendix D: Simp 1Lfled Event Tree Appendix Z:
Process Computer Output
- There are a total of 74 pages in this report in:1uding Figures, Teoles and Appendices.
1
+ Rev. 1
- Rev. 2 e
-r
..,-.,-r-e v._ ~ ~ -..
,.~._-_,-_-,,.,--,----,--7-,
IDR 406 a
dev. 2 Pade 7 of 74 LIST OF FIGJRE3 fitle Figure No.
Steam Generator Tube Rupture Guide 1Lne Development Activltf Network....
1 Break Flow For Single Ruptured Iube........
2 Effect of R0 Pump Operation on Integrated S stem Leakage for f
Single Ruptured Tube........
3 Mass and Energ/ Capabilities of HPI and FORV...
4 Time Behavior of Subcooling Margin for Spectrum of Ruptured Tubes.
5 Emergency Reactor Coolant Pump NPSH LimLts....
6 Single and Multiple Iube Rupture Guidelines......
C-1 Simp 1Lfied Or3G Event free.
0-1 i
l
}
+ Rev. 1
- Rev. 2
. ~..
_. ~.. _. _ _ _.
J IDA 406 Rev. 2 Page 8 of 74 LIST OF IABLES a
Ittle Table No.
1 1
Pressurizer Spraf Flow for Various Pump l
Combinations.
2 Comparison of IDR #406 to Other Sources.
A-1
+
4 Shell thermocouple Substitution.....
E.1. 4.1
+
Wide Range I cold Input S. I. 4. 2
+
4 a
i s
t i
r 9
t i
+ dev. 1
- Rev. 2
e o
LE DOCUMENT NO.
iUClear 403 TITLE SG rube Rupture Procedure Guidelines, Rev. 2 Page'9 of'74 REV
SUMMARY
OF CHANGE APPROVAL DATE 2
Added detall to Table of Contents and reversed IfI L'l'O the order of Sections 4.0 and 5.0.
2 Added Tables 1 & 2 which provide tabular daca on RCP NPSH requirements and on spraf flos for
[L J-/1 various RCP pump combinations.
2 Section 2.1.3.
Added recommandation for OTSG E
lI'81'I3 lsolation if todine release rate exceeds 250 mrem /hr or whole bodf dose rates 50 mrem /hr, correcting error in previous revision.
2 Section 2.1.6 and Figure 6.
Revised f cp
[-ft-O emergenef NPSH limits to account for cal-culated instrument errors during LOCA conditions.
2 Sections 2.1.8 and 4.2.2.
Added dis-cussLon on the experience gained from the
[C/
y-rl-p)
June 1933 Lfachburg simulator sessiohs.
2 Section 2.1.9.
Added recommendation to allow DHR sfstem initiation at 300F instead
[C/
g -s 2 - p3 of 275F.
2 Sei,tton 2.1.10.
Recom:nendation to trip reactor if 200 inch pressurizer level can-
[C yn2-D not be maintained with two HPI pumps running.
2 Saction 2.2.2.
Clarifled guldance on O
$1W isolation criterion with leaks in both OISGs.
2 Section 2.2.2.
Addressed raising OISG 1evel to 95% without causing an overcooling.
[C/
F Y3 2
Section 2.2.2.
Discussed why EFA should*not be used to control OISG pressure in an Nb ef J.S isolated OISG (changing preelous recom-mendation of Rev. 1).
A0000036 7 82
E E
406 TITLE SC fube Rupture Procedure Guidelines, Rev. 2 PAGE ' 10 OF 74 REV
SUMMARY
OF CHANGE APPROVAL D' ATE 4
5ectLon 2.2.3.
Expanded discussions on feed and bleed cooling t.o include A3V and g "g) f Q,p 3 fBV mass and energ/ relief capabilities I
relative to deca / heat and leak rate.
gg g-r 2. -f_?
2 Section 2.3.
Discussed additional work which will be addressed in a future revision to IDR 406.
2 Section 4.2.2.
Added note regarding actions
%dg T-12.-y3 to be taken if RCPs are not tripped within 2 minutes of a loss of SCM.
2 Section 4.2.3.
Added note regarding loss of
%Q g -i2 -y3 SCM af ter RCPs are restarted.
2 Section 4.2.6.
Revised guidance on raising ggg 7_j y _g OrSC levels to 957..
[6[
y-s1.g 2
Section 4.2.1.
Added isolation criterion on 250 mrem /hr.
2 Section 4.2.11.
Added critecton for Core
'[ & I y - s2 -yj Flood Tanks Isolation.
2 Section 6.0.
Added recommendations to isolate g
g,y.g CFIs using critecton provided and initiation of DHR sfstem at 3002.
2 Section 2.2.1.
Clarified that emergencf NPSH
[
d-8E"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
[C b ll 3
loop or both RCP's in the saac loop.
2 Section 3.2.2.
Revised to be consistent with g
7g the steaming critecton in Section 2.1.3.
e A000 0017 9 80
N OE 406 TITLE SG fube Rupture Guidelines, Rev. 1 PAGE' 11 OF 74 REV
SUMMARY
OF CHANGE APPROVAL DATE 1
Minor editorial changes and correction of
/s/
5/8/83 tfpographical errors on pages:
2,5,7,10,13, 19,20,22,A-1,A-3,A-4,B-1,A-2,3-2,6,18.
1 Revised cover page to shoe shell-to-tube
/s/
5/d/83 delta I can be controlled below 1002.
1 Added a List of fables pp 1 & 111
/s/
5/d/83 1
Included use of MEW as means to cool Of3G
/s/
5/8/83
- shell, p5 1
Indicated that continuous steaming-of Of3G is
/s/
5/8/83 simplest means of meeting Of3G 1evel, pressure and differential temp. considerations pp. 5,6,10,18 1
Eliminated reference to RAC for determining
/s/
5/8/83 when to isolate Of3G based on radiological conditions.
p6 i
Added Section 2.1.3.1 to discuss control when
/s/
5/8/83 both Of3G's are isolated.
(Also p 10).
I leovided discussion and Figure for RCP NPSH
/s/
5/8/83 limits.
Section 2.1.6 and Figure 6.
Aef 25,26.
& Sections 4.2.1 and 4.2.4.
1 Revised explanation of ADV & f3V flos capabilit/
/s/
5/8/83 relative to Of3G flooding (incorrect in dev 0) pp 11,16 1
Added deference to B&4 guidance which allows
/s/
5/8/83 coollown at 1002/hr during fube duptures without a soak time even if cooldown rate is exceeded.
p 11.& Ref 24 1
Section 4.2.3 revised to account for inabilltf
/s/
5/8/83 to start either ACE in the A loop.
I 1
Added Section 4.2.7.1 and revised 4.2.7 to
/s/
5/8/83 address Steaming Isolation of Of3G considering the continuous steaming philosophf.
1 Simplified Section 4.2.8 on cooldown rate.
/s/
5/8/83 1
Added Section 4.2.9 on contro1I1ng Of3G shell-to
/s/
5/8/d3 l
tube differential temperature.
A000 0017 9 80
O.
406
@ gg TITLE SG Iube Rupture Guidelines, Rev. L PAGE~
OF 13 74 REV
SUMMARY
OF CHANGE APPROVAL D' ATE 1
Added Section B.7,B.8, and B.9, which were left
/s/
5/8/83 out of Rev. O Inadvertentif.
1 Deleted redundant Section of Part 4 (guldelines)
/s/
5/8/83 1
Added Section C.2 through C.6 to discuss the
/s/
5/8/83 Guldelines Flow Diagram (ilgure C-1) in words.
L Rewrote Appendix E on Process Computer Output
/s/
5/8/83 and Alarms.
L Revised Figure 4 to show deca / heat levels
/s/
5/8/83 as a function of time.
I i
l l
l l
l i
1 l
t A000 0017 9 80
TDR 406 Rev. 2 Page 13 Of 74
1.0 INTRODUCTION AND BACKGROUND
In November 1981, primary to secondary side leaks were discovered in the tubes of both of the TMI-l Once Through Steam Generators (OTSG). There are 15,531 tubes in each OTSG. The plant design basis for a steam generator tube rupture (SGTR) accident is the double ended of fset 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 i
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 B6W simulator training cycle beginning in January, 1983. The results of this training experience are discussed. Finally, the overall conclusions and major recommendations of the guidelines development program are documented.
i
+ Rev. 1
- Rev. 2
TDR 606 Rev. 2 Page 14 of 74 2.0 TECH FUNCTIONS SGTR GUIDELINES DEVELOPMENT PROGRAM Figure I shows the execution of the steam generator tube 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 ef fort 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 analyses were an aid in conceptualizing the physical processes during a tube rupture.
2.1 Development of Design 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 learned about handling tube ruptures.
2.
Define allowable steam generator stresses during cooldown (either as cooldown rate or as tube /shell delta T).
3.
Determine when OTSG'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 NPSH limits.
7.
Redefine entry point conditions.
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 guidelines. The major lessons learned from these events have been
+ Rev. 1
- Rev. 2
=.
TDR 606 Rev. 2 Page 15 of 74 summarized in various documents from the NRC, INPO, and plant procedures and included in the B&W ATOG tube rupture guidelines (Re ferences 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 NPSH 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 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. Forced RC flow provides several benefits during a tube rupture. First, they assure that steam voids do not form in the hot leg U bends or upper vessel head. Steam voids in these locations can interrupt natural circulation or prevent RCS depressucization. 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, pressurizer spray is available and RCS pressure control is not dependent on the PORV or pressurizer vent.
Main feedwater can be used if RCP's are running; with pumps off, emergency feedwater must be used, which is less effecti~ve in cooling the OTSG 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.
4.
RCS Degassing RCP NPSH limits at Ginna required shutting down of the reactor coolant pumps at low pressures.
Shutting the pumps allowed noncondensible gases to collect in the top of the steam
+ Rev. 1
- Rev. 2
TDR 406 Rev. 2 Page 16 Of 74 generator tube U bends. These trapped gases prevent'ed'RCS depressurization for many hours. An analogous situation might occur at the hot leg U bends. 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 hour1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />) 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, i* ' ' Slighted the need to store radioactive water in the plant d
...6 a prolonged RCS cooldown.
6.
Shell-to-Tube Delta T A tube leak at Rancho Seco in May 1981 yielded evidence of the importance of controlling OTSG tube /shell delta T.
The existing limits and precautions at TMI-1 is 100F*.
However, before J
tube /shell delta T exceeded 100F*, the leaking tube was placed under tensile stress and the tube was pulled into a circumferential tear. Maintaining tube /shell delta T limits are important during tube rupture and are discussed in more detail below.
2.1.2 Limiting OTSG Tube Stresses Steam generator tube stresses are generated as a result of ter.sile 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 temperature between the colder tubes and warmer shell 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 B&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 dif ference to increase wIille adding a tensile load on the tubes by elongating the shell via pressure loading. Structurally there are compensating effects involved in mitigating these two load contributors. Rapid depressurization eliminates the pressure induced stress but aggravates the delta T
+ Rev. 1
- Rev. 2
TOR 406 Aev. 2 Page 17 of 74 induced stresses. The optimum OISG cooldown/depressurization rate has not been determined.
However, it is known that isolating the OISG at 1000 psta Ls not the best means of reducing stress.
Cooldown/depressurization is the preferred method.
There are three limits for tube /shell delta T that presentif appif to IMI-1.
Plant " Limits and Precautions" (Ref. 22) limit delta T to 602' during heatup and to 100F' during cooldown with one OISG isolated. This value of 1002' assumed that tubes had no more than 400 through-wall cracks.
In reference 23, 35W established 142F' for a cooldown using both OTSG.
The 70l* value la this IJA is proposed as a guide in determLning an acceptable coollown rate.
If delta I can be maintained at or below 70F*, the operator has optimized the plant cooldown rate.
The 70F' ilmLt more conservative 1f assumes that tubes in the OISG are leaking below a detectable limit. A 70F' value limtes propagation of these cracks.
+
Control of shell to tube delta T ts accomplished in several ways.
First bf cooling the OISG Liquid (steaming) to allow the metal shell
+
to cool. Second, bf providing cool, main feedwater into the
+
downcomer.
If netcher of these methods works, the ACS cooldown must
+
be decreased until the OISG shell cools sufficient 1.
If reducing f
+
the cooldown doesn' t work, then the cooldown must be terminated.
2.1.3 S tea.n t ng. Isolation and Filling of the Lea'cing OI3G Isolation of the leaking OI3G can result in the overfliling of that generator.
It is preferable to prevent overfilling, however, to allow plant cooldown in an expedLttous manner.
If the OI3G fills, Lt becomes a large pressurizer (as evidenced bf the Ginna event).
The time it took to cool down this mass of hot water greatif extended the cooldown of the plant.
Steaming also maintains some natural circulation flow in the hot leg.
Ihts flor cools the hot leg 0 bend and decreases the chances of steam void fot ation.
+
As discussed in Section 2.1.2, steaming and depressurization of the OISG also reduces OISG tube stresses.
However, depressurization of the OTSG also increases leakage rate. As discussed in Section 2.2.2, the OI3G pressure should be below RCS pressure to promote flow through the hot leg.
The optimum OT3G control recults in
- 1) depressurization of the OISG without causlog large delta I'h;
- 2) minimum RC3 leakage; 3) promotion of natural circulation flow in the hot leg; and 4) posttive leakage from the ACS Into the JI3G to assure hot leg cooling in the absence of natural circulation.
The optimum pressure control scheme to meet this criteria has not been determined analyticall.
f
+ Rev. 1
- Rev. 2
TDR 406 Rev. 2 Page lg of 74 Meeting all four of these criteria will probably result in a'nearly
+
continuous steaming of the affected OTSG. Moreover, intermittent steaming of the OTSG'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 DH 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 steamed. Similarly, tne OTSG would have to be steamed to maintain natural circulation.
Although it is highly desirable to prevent steam line filling, there are certain circumstances which dictate that the OTSG should be filled. 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 (Ref 11). This analysis shows that the loading is acceptable without pinning (except for the dead load ef fects 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 level, there is still sufficient inventory to flood both steam lines and put about 30,000 gallons of water into the containment building (Ref 12). This amount of water is suf ficient to provide adequate NPSH in an LPI to HPI " piggyback" mode of core 1
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 Emerbency may already have been declared based on
+
OTSG leakage rate. Nevertheless, this level provides a rationale for deciding that release rates are high enough to warrant OTSG isolation.
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TJR 406 Revo 2 Pade 19 of 74
+
Consideration was given to defining isolation conditions based on R03
+
activLtf levels, meteorolog/ and steam line radLation levels. RCS activity level cannot be correlated to offsite releases, sinca offsite dose will be affected bf the location of the tuba leak in the Of3G, availabilltf of the condenser and plateout and decontamination factors.
It is also undesirable to isolate the OISG based on assumed, meteorological conditions.
The most desirable approach is to isolate based on actual releases occurring during the event.
The existing site emergenef 11mtts are 50 alem/hr whole bodf and 250 mrem /hr thyroid dose measure or projected at the site boundarf (Ref. 33). Section 2.1.9 discusses the length of time required to cool the plant down to DHR sfstem conditions.
This length of time defines the integrated does allowed by the guidelines (i.e. releasa rate for the specified perLod of time).
2.1. 3.1 Steaming, Isolation and Filling with Both OrSG's Leaking
+
Isolation and steaming of the Or3G's must be addressed for leaks in
+
both OfSG's. Once ROS temperature is balow 540 F, a cholce has to be
+
made regarding OfSG isolation. Both OT3G's should be steamed unless
+*
either the 34Sr level or of f stte release criteria is reached.
If
+*
both OISG's are steamed, then all steam loads from both OI3G's should
+
be isolated except for the ISV's/A3V's.
All other steaming,
+
1 solation and Ellling criteria should be followed.
If the BASI level reaches 21 feet, then both Or3Gs must be isolated.
If the dose criteria is reached, one OTSG should be isolated and the doses reevaluated.
If the dosa criteria still cannot be met, then the second OISG should be isolated.
2.1.4 Minimum Allowable Subcooling Margin A primarf goal during a tuba rupture is to mintmLze offsite dose.
Minimiziag laakage from the RCS is the first line of defense.
Leakage from the primarf to secondarf is determined by the size of the leak, and bf the differential pressure between the RCS and GTSG.
Primarf to secondarf differential pressure is controllad bf fixing
+
the degree of RC3 subcooling. Once secondarf pressure is fixed, cold leg temperature is determined. For anf time, decaf heat la fixed.
RCS flow (which is determined bf OT3G 1evel or RCP operabilltf) then determines hot le3 temperature. Reactor coolant pressure or HPI flow i
then fixes the dedree of subcooling.
Since the operator controls OT3G and RCS pressure and HPI flow, he is in control of the subcooling margin, henca primarf to secondarf delta P.
Figure 2, illustrates the effect of subcooling margin on prLaarf to secondarf leakage.
Figure 3 illustrates the relative ef fects of a cooldown with RCP's i
off using 50F* and 25/* subcooling. Even at the maximum cooldown of 100F"/hr, the integrated leakage dLf fers by a factor of two.
i
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T IOR 406 Rev.,2 Page 20 of 74 2.1. 5 Walve Fuel in Compression Limits Fuel pin-in-compression limits are specified to assure that fuel pina are alwa/s in compression above 4252' in order to prevent detriment.al orientation (i.e., radial orientation of h/ eldes) (Ref.14). These d
limits require a high subcooling margin for RCS pressures ranging from 1350 psi to 550 psi.
In correspondence dated January 20, 1983, (Ref. 14) B&W confirmed that violation of these ilmits during tube rupture events is acceptable. When these limits are violated it is important that the pressure and temperature versus time be recorded so the effects on cladding can be evaluated. The evaluation must determine whether clad ballooning or incipient cracking has been induced.
2.1.6 Reactor Coolant Pump NPSH Limits
+
RCP NPSH requirements place limitations on the minimum subcooling
+
margin. At low RCS pressures RC pump NP3R limits approacn 100l* of
+
subcooling.
However, general centrifugal pump test data have shown
+
that NP3H requirements are substantially reduced at water tempera-F tures above 250P*.
A review of TMI-1 test data on the subject
+
reactor coolant pumps indicates a single loop flow of 93,500 gpm with
+
two toops in operation with one pump per loop. The pamps' manu-
+
facturer (Westinghouse) has provided required NP3H at various pump
+
suction temperatures (Reference 25) for the flow associated with twa pump operation.
The NP3R available, as indicated bf the saturation margin monitor in the hot Leg, is then calculated bf considering the
+
total pressure drop f rom the hot leg to the pump's suction (Reference
+
26). The resulting NP3d requirements for 2 pump operation (one per loop) are shown in Table 1 and Figure 6.
Also shown is the 4 pump
+
operation NP3H curve which has considered the changed flow
+
distribution in the coolant. loops.
In addition, the normal NP3H
+
curve and the 25F* subcooling curve are shown for comparison purposes.
RCP operation below 300 psig is not permitted.
This assures adequate pressure differential across the No.1 pump seal (Ref. 32).
+
The emergenef NP3H limits are intended for operation of RCP's during
+
abnormal and emergenef conditions such as small break LOCA, SG tube rupture, station blackout and secondarf side upset events.
Pump llalts and precautions must be adhered to while following the emergency NP3H limits (e.g., the pump should be tripped on high vibration).
I 2.1.7 Procedure Entef Point Condition Ihe use of an emergenef tube ruptura procedure should be limited to situations where normal limits (e.g. fuel pin-in-compression and RCP NPSH) are being waived. The guidelines' encef point condition is chosen as 50 gps.
A leak rate of this magnitude would be expected from the complete separation of one tube (as opposed to 385 gpm for a Less likel, (but more serious) double-ended offset of one tube).
f I
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EDR 400 Rev. 2 Page 21 of 74 4
would be leakage of this extent f rom a number of tubes.
Both situations warrant entering the emergenef procedure.
Below this limit, plant cooldown should be achieved within normal limits unless additional equipment failures occur.
2.1. 8
- Simulator 2xperience Steam generator tube rupture procedures were exercised during the Januarf and June 1983 simulator sessions, the experienca gained from these two sessions has been factored into this f]R.
The principal 1essons learned were that:
1.
Controlling plant cooldown rate witn 2 or 3 HPI pumps running is verf difficult at best.
Raising OTSG 1evel to 95% during this plant condition maf not De possible.
2.
Prioritiza:Lon of plant control parameters was not obvious to the operator in certain situations.
The two situations which were encountered were:
a.
Minimizing subcooling margin has priorttf o7er mint-mizing the cooldown time, and; b.
Steaming to control OISG level is less important than RCS cooldown rate.
3.
Plant response after RCPs are restarted was unexpected to the operation. Second pump restarts may be required before sub-cooling margin staf s above 2520 4
Criteria for isolating core flood is required.
Core flood tanks should be tsolated in a subcooled system before thef 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 detall in Section 4.0.
2.1. 9
- Emergencf Limits for Decay Heat Sfstem Initiation Plant experience indicates that a larde portion of time during cooldown occurs below the temperature of 3502.
Simple anal ses,
/
assuming onif one ADV for a loss of of fsite power, using the CSMP computer code (Ref 31) Indicate that the RCS can be cooled Jown 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 coollown to 2/52 can ba accomplished in about ten hours.
215F ls the normal DHRS initiation temperature. GPJd0 has evaluated the capability of the DHA system to operate at a temperature of 300F (Ref 32) and concluded that it is within the design capabilities of the sfstem.
Therefore we recommend that the tube rupture procedure allow inLtlation of the DHR system at 300F instead of 275F.
2.1.10
- Reactor Irlp on Low Pressurizer Level Preventing a loss of subcooling margin has manf advantages in conto 11 Lng the plant. Before the spring of 1983, EP 1202-5 required the plant to be tripped if leiel could not be maintained above 100
+ dev. 1 IL
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IDA 406 Rev. 2 Page 22 of 74 i
Inches with two HPI pumps running.
ThLs may not be a sufficient 1evel to prevent voiding of the pressurizer after a reactor trip.
Emptfing of the pressurizer causes a loss of subcooling margin and the subsequent trippind of the RC pumps.
In order to prevent this sLtuation, the raactor should be tripped Lf level cannot be maintained above 150 inches or higher.
This is sufficient volume (about 600 cubic feet) to prevent pressurizer voiding.
t, There is a disadvantade to this recommendation since the safetf valves will 1Lft after the reactor trips. However, this situation is considered acceptable when weighed against the plant control advantages of having RCP's running. Also, onif a certain window of break sLzes will result in reaching 150 inches and not 100 inches with f ull HPi flow. Outside of this window, both levels would be reached.
I 2.2 Development of Multiple Tube Rupture Procedure Guldelines (Path 2)
Tne treatment of multiple tube ruptures relied on several sources of information.
The Ginna tube leak exceeded the single tube flow for a 3F4 plant and also resulted in a loss of subcooling.
Therefore, that event legitimatelf represented a multiple tube rupture.
Ihe Oconee tube leak with a delaf in getting onto deca / heat removal, prompted anal sis of water inventories required to assure a source of water f
1 for HPI cooling.
Besides plant operatind eXPerlence, this TDR investigated the following aspects of multiple tube ruptures:
1.
Revision of the RCP trip and restart celteria.
2.
DISO steaming and level control.
3.
Establishment of criteria for going on feed and bleed cooling.
)
4.
Cooldown/depressurization.
2.2.1 Revision of RCP Trip and Restart Celteria Based on initial small break LOCA analyses received from P4A vendors l
in 1979, NRC concluded in NGREG 0623 that delafed trip of reactor coolant pumps during a small break LOCA can lead to predicted fuel cladding temperatures in excess of current licensing limits. At tha time of RC pump trip the ILquid that was previously dispersed around the primarf sfstem through pumping action now collapsed down to low points of the primary sfstem such as the bottom of the vessel and steam generators. -This separation results in significant uncoverf of the reactor core if sfatem voiding is high enough, due to an insufflcient amount of 11guld being available to provide acceptable core cooling. Unacceptable consequences would result from delafed reactor coolant pump trip only for a range of small breaks LOCA (.025 2
to 0.25 ft ) and a range of trip delaf times after accident initiation. Based on these findings, a meeting of utility vendors and owners was held with NRC in September 1979 At this meeting it was adreed that the 1600 psig ESAS signal provided timely Control Room indication for manual action to prevent possible voidind scenarios.
+ Rev. 1
- Rev. 2
IOR 406 Rev. 2 Page 23 of 74 GPG had B54 reevaluate these LOCA scenarios assuming RCP's were tripped on loss of subcooling margin.
Ine conclusion of that reanalysis was that loss of subcooling was an acceptable alternative to pump trip on 1600 pstg ESA3.
In March 1983, tha NRC 3taf f required Utilities to reevaluate their pump trip schemes (Ref.17).
GPJNC provided an evaluation of the pump trip criterion and a schedule for Implementing this criterion bf May 1, 1983.
The advantage of maintaining RCP's is that during Steam Generator Tube Ruptures in which minimum subcooling mard n is maintained, i
continuous RC pump operation assures expeditious cooldown with a minimum primary to secondarf differential pressure.
This change in criteria for RCP trip will allow RCP's to be operated for a greater spectrum of tube ruptures (including ruptures befond the design basis) and to reduce the offstte doses for those events.
Reducing the allowabla subecoling margin is not intended to reduce RCP equipment protection. RCP's should be tripped if emergenef NPSH requirements are not met, and should not be started until NP3H requirements are re-established.
If applicable NPSH pump vibration limits are exceeded, then the RCP's should be tripped.
Pumps should be restartad as indicated in the T31-1 Small Break LOCA Procedure (EP 1202-68, Attachment 1).
As noted in Section 2.3, bumpind criterion requires additional clarification.
Figure 3 tilustrates the reduced Icakaga possible with RCP's on.
Similarti, restart of RCP's has a great advantage. During tube f
ruptures, primarf to secondarf differential pressure decreases rapidly since OT3G pressure is high. Leakage flow is exceeded b/ HPI flow and subcooling margin should norma 11f be restored within 20-60 minutes after larger tube ruptures.
Restarting RCP's provides pressurtzer spray, and prevents void formation in the hot legs U bends and reactor vessel head.
2.2.2 OISG Steaming and Level Control The guidelines for OISG steaming are nearif the same when either one or both OISGs are affected. Tne OTSG pressure should be controlled to prevent lifting of safetf valves (i.e. staf below 1000 psig).
Level should be maintained below 954;on the operate range.
Thera are several other issues to be considered for multiple tube ruptures, however. First, large tube ruptures maf result in RCP trip.
The OTSG's should be steamed to maintain natural circulation in the affected loop. Natural circulation flow will minimize the
+
chances of drawing a bubble in the hot leg U bend.
Continuous
+
steaming of tha OTSG allows all of thase considerations to be
+
accommodated.
It is important to recognize that a large tube rupture with loss of subcooling is a LOCA condition.
Therefore, it is required to raise i
OISG 1evel to 95% to assure that 11guld level in the tube reglon is
+ Rev. 1
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IJA 406 Rev. 2 Page 24 of 74 hLgh enoudh to allow water to flow into the core during boiler condenser cooling.
If level is not raised to 950, then EF4 flow must be at a high enough flow rate to penetrate the tube bundle suf ficLently to provide adequate heat transfer. A flow rate of 450 gpm total (225 gpm/0I3G) has been verifled as acceptable by B&W (Ref 29).
Ihts flow is the minimum avallable after a seismic event and worst case single f ailure, coincLdent with a small break LOCA in which boiler condenser cooling is required.
It is important to recognize that with two HPI'h available, boiler condenser cooling is not required.
Procedures should therefore allow the operator to raise OrSG level to 95% tempered wLth the need to con. col the ACS cooldown rate. During tube rupture events with both HPI pumps available, the unaffectad Or3G 1evel should be raised first wntle the affected OI3G level should be prevented from bolling def.
(maintain a minimum level of 30").
The operator can control 1 OI3G Instead of trf ng to raise level in both OT3G'h i
simultaneousif.
For the case wLth only ona HPI pump, if Incore temperatures are not decreasing and the OI3G is not resofing ACS heat, then there will not be a coollown rate control consideratlon; moreover, the plant maf be in a condition that requires boller condenser cooling.
Therefore, DISG levels must be raised to the 954 level slaultaneousif in this situatLon (Aef. 30).
3ection 2.1.3.1 discusses steam generator tsolation, steaming and
+
fll1 Lng criteria when both OISG'h are leaking.
Ihts discussLon also
+
applies when the RC3 subcooling margin has haan lost.
2.2.3 Criteria for Feed and 31eed Cooling Analyses of multiple tube ruptures indicate that existing plant procedures for establishing feed and bleed cooling are correct.
Feed and bleed cooling should oe initiated when the Of3G heat sink is not available.
If both steam generators are isolated during a tube rupture, the PGAV should be opened with full HPI turned on.
An addittonal complication for tube ruptures, however, is the potentLal to flood the OI3G's and force open the safetf valves under this condition.
If RC3 pressure is below 1000 psig, the P0dV is capable of removing decaf heat, even with ILquid 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 RCS pressure bf throttling HPI. Moreover, wLth RCS pressure below 1000 pstg the OISG safetf valves will not lift.
If RCS pressure stafs above 1000 psig, however, the operator must take action to prevent safetf valve lifts.
A situa:Lon with AC3 pressure above 1000 psig and neither Or3G available requires' the opening of the IBV or ADV' h to control Of3G pressure below 1000 psig A
and level below the upper tube sheet.
dither the AJV or ISV have sufficlent steam capacity at high OI3G pressure to remove deca /
heat.
The ISV' h also have suf fielent capacLtf to prevent Of3G flooding. However, LE the leak rate is large enough, the steaming rate required to control level in the Of3G maf result in an unacceptable RCS cooldown rate.
In this case, cooldown rate must be controlled and the OISG allowed to flood. As discussed in SectLon
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- Rev. 2
-,-r,--
--e
TOR 406 Rev. 2 Page 25 of 74 3
4 4.2.2.1, this situation does not seem likelf at least at high decay heat levels). As decaf heat decreases, steaming can.be terminated when RC3 pressure goes below 1000 psig and is controlled b/ the PORV and HPI.
Ihe steaming capacit/ of an AJV at 1000 paig exceeds decay heat levels within several minutes after reactor trip.
HPI capacity exceeds the capacity of one ADV.
Therefore, the RCS pressure can be controlled at 1000 psig in this made without lifting safetf valves.
Subcooling margin can be regained and the plant cooled down until an
~
Of3G heat sink can be restored or until the plant can be put on deca /
heat removal.
Ontil Or3G level is above the upper tube shaet, pressure in the Or3G will remain below 1000 paig, since the RCS temperature is below 540F*.
With level less than 600 inches, however, the operator still must steam to keep pressure below 1000 psig; therefore he should not 4
have to steam to control level on the affected Of3G. 4 hen level goes above 600 inches, pressure in the OISG is determined bf 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 OT3G to reduce pressure.
However, if there are steam leaks in the sfstem (e.g.,
through steam traps.) then the lines could fill with water before
{
OI3G pressure increased, therefore to prevent this situation the OI3G's must be steamed to preclude this possibility.
2.2.4 Cooldown/Depressurization Analyses of multiple tube ruptares demonstrated that subcooling margin should be regained in 20-60 minutes (see Figure 5).
RCP's can be started and a forced flow coo 13own instituted.
Even if RCP's are not available, the cooldown during a multiple tube rupture can be accomplished within the single tube rupture guidalines.
If equipment failures prevent a normal natural circulation cooldown, then the plant would be cooled down with feed and bleed cooling.
This maneuver would probablf require initiation of feed and bleed cooling in the HPI/LPI " piggyback" mode. Existing plant procedures give correct guidance about when to initiate this mode (343r level below 3
. ft.).
+
$Guldan'ce from B&W on PTS / Brittle Fracture limits requires a " soak
+
time" co allow the vessel wall to reach the RC3 temperature.
+
However, 854 has also recommended that the " soak time" is not
+
requLred during tube rupture events in which a rapid cooldown is
+
necessarf (Reference 24).
Steam releases during multiple tube rupture events can be minimized bf judicious usa of the EF4, HPI and ISV's.
Full HPI flow, in
+
conjunction with throttled EFW flow allows a 100F'/hr cooldown without having to steam either GISG.
4
+ Rev. 1 ts.
- Rev. 2 v
IOR 406 Rev. 2 Page 26 of 74 2.3 Additional Work Requirements 2.3.1
- Anal ses f
As noted in Section 1.0, there is a program of ongoing computer analysis work simulating single and multiple stea= generator tube ruptures.
The effort includes the plant states listed in Sections 3.1.1 and 3.1.2.
Ints list does not reflect two specfte detailed analysts efforts which are being undertaken as part of the tube rupture quantitative development effort.
The first analysts is a simulation of the vessel head region during natural circulation cooldown.
I:ls analysis effort will help in evaluating the effect of a vessel head bubble on the RC pressure response.
It will also aLd in evaluating the benefit of the reactor vessel head vent.
The second anal sis effort is being performed in conjunction witn f
Babcock and Wilcox Company.
Detailed anal ses are being performed to f
provide improved guidelines for OTSG filling af ter a loss of subcoollag margin, with one, two and three HP1 pumps available.
The intent of the analyses is to assure that the OISG's are f Liled wLthout violating cooldown or tube to shell dif ferential temperature limits, while still meeting core coolabilitf requirements.
This t
effort was considered after the-Januarf 1983 simulator training session and further defined after the June 1983 simulator session.
2.3.2
- Issue Resolution A number of issues were identified which require further effort to resolve.
These are che following items:
A.
Operator guidance for identif ing two phase natural circ-f ulation cooling (boller condenser).
B.
Acceptability of excessive cooldown rates for ver/ short time Intervals.
C.
Importance and technical basis of Fuel-in-CompressLon limits.
0.
Viabilltf of DRR system Initiation at temperatures above 3000F.
E.
Identification of pump vibration limits for various pump combinations.
Determine viability of ATJG RCP " Pump Bump" criterion.
Ihe bumping criterion would allow running pumps without adequate subcooling or NPSH margin as long as a steam generator heat sink is available.
Determine whether the AI00 criterion anticipates that NPSH will be reestablished since the heat sink is available.
The existing bumping criterton in E4I-1 emergency does not require that a heat sink be established 4
and would allow continuous RCP operation in violation of NPSH limits.
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TDR 406 Rev. 2 Page 27 of 74 3.0 DISCUSSION OF MAJOR REVISIONS TO EXISTING PROCEDURES The development of the design basis guidelines discussed in Section 2.1 identified a 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 OTSG is af fected.
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 OTSG 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 offsite.
4.
Tube leaks in both OTSG's (but one OTSG remains capable of removing decay heat).
5.
Steam lines associated with leaking OTSG flood.
The consideration of items 1 and 2 are equivalent to an assumption of loss of of fsite 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.
+ Rev. 1
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TDR 406 Rev. 2
=
Page 2g of 74 The following conditions were assumed in developing guidelines for this category of tube rupture event.
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 OTSG can be steamed.
Contingencies The revised guideline will have provisions to deal with the following additional circumstances:
1.
Both OTSG's are affected.
2.
Both OTSG's af fected, but one OTSG remains capable of RCS heat removal and either a) the PORV is unavailable or b) RCS pressure stays above the main steam safety valve setpoint due to void formation in the RCS.
3.
Neither OTSG is capable of removing decay heat, and either a) the PORV is available, or b) the FORV is unavailable.
3.1.3 Revised Equipment 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 j
following:
1.
Mitigate or prevent further OTSG damage.
2.
Maximize the cooldown rate to cold shutdown.
3.
Minimize SCM (thus minimizing primary to secondary leakage).
l 4.
Maximize RCS pressure control options.
l An Event Tree showing the various possible developments of an OTSG tube leak appears as Appendix D to this report.
l l
+ Rev. 1
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IDA 406 Aev. 2 Page 29 of 74 3.2 Discussion of Guidelines Appendix C provides a logic dLagram of the tube rupture guidelines (with a written dLscussion of those Suldelines).
This section of the report describes tne guidelines shown in that diagram.
The sfsptoms of the tube rupture procedure define the entef point conditions when the emergencf procedure is used.
This procedure need onif be entered for situatLons where a rapid depressurization of the plant is warranted. Wnen such condLttons warrant, then the plant should be shut down and cooled down as expeditiousl/ as possible and certain normal plant limits (RCP NPSH, normal tube /shell delta I and fuel in compression ilmits) are walved.
3.2.1 Immediate ActLons The tube leak in question maf not be large enough to cause a reactor trip.
In such a case, the operator begins a load reduction as capidly as possible without causing a reactor trip (10C/ min.).
Avolding a reactor trip prevents lifting of the OI3G safetf valves.
3.2.2 loilowup ActLons - Subcooling Maintained and ACP's Available Once the load reduction is initiated, the operator has several major goals to achieve while brLagLng the plant to a cold shutdown conditLon.
First, he must prevent lifting of the OI3G safetf valves; second, minimize primarf to secondar/ leakage bf minimizing primarf to secondarf differential pressure; and, third, slaimize stresses on the OI3G tubes bf 11mLting tube /shell delta I.
Finally, the operator will alntmLze offstte dose by allowing the leaking OISG to flood if offstte doses are large enough (approaching levels at which a Site Emergenef would be declared).
The major dLfferences between the existing plant procedure and the proposed new procedure would be the following:
3.2.2.1 Maintain a Minimum of 25/* Subcooling MinimLzing subcooling margLn means that primarf to secondarf dif ferential pressure is also mLntmized, which reduces leakage and offsite doses makLng the event more manageable.
3.2.2.2 Steaming / Isolation Criteria for the Af fected OISG Ihe present procedure allows the operator to let the OI30 fill anf time that RCS pressure is below 1000 pstg.
The revised procedure has the operator steam the OI3G for these purposes: liest, to prevent lifting of the GI3G safety valves.
Second, to prevent the generator from fL11 Lng.
+ aev. 1
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I IJA 405 Aev. 2 Page 30 of 74 3.' 2.' 2. 3 Shell-to-rube Delta I Plant limits and precautions require maintaining the Of3G tube -
temperature wLthin 100F' of tha shell temperature. A shell to tube delta I of 70F* llatts stresses and minimLzes the chances of increasing the leak size.
3.2.3 Pollowup ActLons (Automatte Reactor Irly has Occurred)
All of the followup actions discussed above stL11 apply when tha tube leak is large enough to cause an automatic reactor trip.
In addition, the following procedure changes would apply.
3.2.3.1 RCP Trip alth a Loss of Subcooling Margin Present plant procedures require RCP trip on initiation of 1600 psig E3AS. Rupture of one or a few OI3G tubes will llnelf result In AC3 depressurization to tha HPI setpoint, but maf not result in a loss of 3 C4.
3.2.4 Followup Actions for Loss of Subcooling The third section of the tube rupture procedure is entered when RC3 subcooling is lost.
Here, the operator must treat LOCA, as well as tube rupture symptoms. He Ls then able to pursue the followup tube rupture actLons. All of the guidance for followup actions without loss of subcooling appi.
f The objective in this portion of the procedure is to maintain natural cLeculation (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 coollown.
If one pump can not be started per loop, then start both RCP's in one loop.
This will maximize the pressurizer spraf flow for the given RCP availablllty.
If the affected OI3G cannot be steamed for either radiolo3 eal or equipment reasons, then Eid is used to control OI30 L
pressure.
Essenttally, EFW ls used as a pressurizer spraf to keep the leaking generator silghtlf lower in pressure than the AC3.
The 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 from the OISG.
4.
leakage from the RCS to the Of3G will be small since dLfferentLal pressure will be small.
5.
the small flow through the hot leg maf prevent vold formation in the hot leg.
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a TDR 406 Rev. 2
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Page 31 of 74 If subcooling is regained in the RCS, then HPI is throttled,"RCP's are started and the operator continues the cooldown. The desired RCP configuration is to start one pump'in each loop.
If the operator is unable to start an RCP in each loop then he should start both RCP's in one loop.
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 after RCP restart, it implies that the increased RCS flow has caused voids in the system to collapse, thus dropping RCS I
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 unaf fected 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.g., 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.
I i
i
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TDR 406 Rev. 2 Page 32 of 74 e
4.0 SIMULATOR TRAINING EXPERIENCE
- 4.1 Introduction
- Most of the guidelines proposed in this TDR were incorporated into a lesson plan for the annual requalification training of the TMI-l licensed operators at the B&W simulator. 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 demonstrate 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 ef fect 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 ef fective in training operators on the basic concepts for treating tube ruptures. The June training cycle was successful in communicating concepts.
4.2 Results
- 4.2.1 January 1983 Training
- 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 pressurizer level is back on scale, rather than waiting for pressurizer 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 HPI on, while steaming
+
the OTSG and maintaining a 100F'/hr cooldown.
The difficulty was
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TDR 406 Rev. 2 Page 33 of 74 created by the steaming of the OI3G's in this situation.
HPI and throttled EFW flow can provide a plant cooldown at near 100F'/hr if the DISG's are not steamed.
Durlag the simulator session, B&W revised the simulator to allow leakage of more than 2 tubes and to allow leakage in both OISG's.
4.2.1.1 Comments This material was presented to two of seven groups by Tech.
Functions personnel.
The remaining five groups received it from S&W training personnel who taught the material using the same lesson outline. B&W did not endorse the material.
Comments from trainees indicate that the training was of dublous value.
It will be necessarf to repeat the training for all personnel.
4.2.2 June 1983 Training
- A number of items were identifled during the B&W operator training simulator sessions held from June 6 to June 29, 1983.
The experience gained f rom using a revised tube rupture procedure EP 1202-5.
These Items will be discussed below.
4.2.2.1 Control of RCS Coollown Rate Section 5.2.1 discussed the difficultles in controlling cooldown rate while raising OT3G level to 95%.
At the time, it was the author's belief that steaming of the OTSG's was causing the excessive cooldown rate.
However, furthar discussion with B&W (Ref 29) indicated a different explanation.
The B&W HPI model calculates flow by lteratively solving two equations of the form:
Pd = 2840 -
3 4
(1)
N and W=
(Pd - Pgcg) A 1/2 (2) where:
Pd = pump discharge pressure PRCS = RCS pressure W = flow, ibm /sec N = number of HPI pumps running A = coefficient to account for number of HPI valves open.
The HPI flow is overpredicted for IMI-1 with three HPI pumps running and/or low RCS pressure.
Ihe difference results from the j
l physical arrangement at TMI-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 simulator drills were run with only two HPI pumps available and control of cooldown rate was improved.
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~
=
fDR 406 Rev. 2 Page 34 of 74 The HPI initiation rule has been reemphasized to the operator, namely that " full" HPI means the full flow from two.HPI pumps.
It is acceptable to secure the third HPI pump when the RC31s saturated, and the OfSG heat sink is available or if cooldown rate is 100F'/hr. or more.
A second consideration in controlling cooldown rate was in raising the OTSG 1evel increase to 95! af ter a loss of subcooling i
mard n.
Operations believes that it is an unnecessary burden on the operator to control cooldown rate while raising level on both OT3G's simultaneously.
Therefore, the leasing generator level will not be raised until the unaffected Of3G has been raised to j
95% unless incore thermocouple temperatures are not decreasing
~*
and the OfSGs are removing decaf heat.
The 954 level is Important in establishing boiler-condensor cooling during small LOCA's in which onif one HPI is available.
However, the RCS coo 13own is onif a concern if both HPI pumps are running.
Therefore the two concerns are mutually exclusive.
Operator training and EP 1202-5 have been revised to have the operator control one of3G 1evel at a time as long as incore i
temperatures are decreasing.
If RCS temperatures are not j
affected by the secondary side cooldown (i.e., no secondarf side heat removal) then both OfSG's should be raised to the 95% level simultaneously.
During the simulator session of June 11, 1983 the operators were faced with a large (about 1400 gpm) tube rupture.
Thef attempted to control OfSG 1evel below 95/. on the affected generator.
However, the cooldown rate was too high even with the unaffected OTSG isolated.
The simulator response to the event was partially responsible, but the procedure also needed to be more explicit.
the simulator leak model currently uses the orifice equation to A
predict leak flow (Ref. 28). This model would initia11f overpredict the break flow and would account for the verf rapid flooding of the OfSG's compared to results of the REIRAN and RELAP5 computer codes.
REIRAN and RELAP5 (Ref. 35) analyses to date do not predict such a response. Nevertheless, the operator needs to recognize that cooldown rate control takes precedence r
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 j
operator initiate plant cooldown and then establish min'imum subcooling margin. The simulator sessions showed that the RC3 could not be depressurized fast enough to maintain a i
minimum subcooling margin. Therefore, training material was revised to emphasize the need to stabilize the plant and reach the minimum allowable subcooling margin.
Plant cooldown should then be. initiated. -The procedure was modified so that all four RCP's can be lef t on until 500F instead of 540F. Based on the TMI-1 Af03 (Ref. 28), this change provides a difference of about 47%
in the spraf flow (see Table 2).
Thus pressurizer spraf flow is maximized for as long as possible.
Finally, the
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f]R 400 Rev. 2
. e Fade 35 of 74 operator is given the option of using the pressuelzer vent if he is still unable to reduce pressure sufficientif to maintain a minimum subcooling margin.
4.2.2.3 RCP Restart Criterla Section 5.2.1 observed that the RCP restart criteria on 25F subcooling margin (SCM) was very usaful to the operator.
Several areas required clartficatioa or expansion, howeier.
First, RCP restart should not be attempted unless pump emergenef NPSH limits are met.
The only exception is that NPSH requirements are walved, with pump restart allowed during certain inadequate core cooling conditions as specified in plant procedures.
Pump " bumps," however, should be attempted even Li NP3R requirements are not mat.
- Second, loss of subcooling may occur af ter the RCP's are restarted.
Collapsa of steam volds in the RCS maf cause voiding of the pressurtzer, resulting in a loss of 30M. RC3 temperatures may be hatter Ln the tube ragion than in the core if natural circulation has bean lost. X1xing of this hotter water with the cooler core water will cause a decrease in the SCM. Several pump starts =ay be required before subcooling margin stabilizes above 25F*.
Allowing two minutes of RCS flow is helpful in eliminating both staam volds and temperature dradients so that successive restarts alli 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 RC3 was subcooled, but core flood tanks initiated, providing cooling water which was not required, since SCM was maintained. Most significant was that CFI initiation delayed RC3 depressurization. Neither tha LOCA nor tube rupture procedure provides anf guidance about 1 solation of the core flood tanks.
Therefore, this IDR has been revised to provide guidance about when to isolate the core flood tanks (see Section 5.2.11).
4.2.2.5 RCP TrLp Criterion During the June 1983 simulator session, questions arosa re-garding the actions to be taken Lf RCPs were not tripped within 2 minutea of a loss subcooling margin.
Clarlftcation was provided using the guldance of the fMI-1 " Abnormal Transient Operating Guidelines" (Af0G) (Ref. 28).
Af0G re-quLees the operator to keep the RCPs in each loop running if the two minute time limit is exceeded.
If the RCPs are tripped after this time, the RC3 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 R 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 ba started.
this strategf maximizes the likelihood of maintaining adequate core cooling.
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Tal 400 Rev. 2 Page 35 of 74 b
operator is given the option of usin; the pressurizar vent if he is selli unable to reduce pressure sufficiently to maintain a minimum subcooling margin.
4.2.2.3 RCP Restart Criteria Section 5.2.1 observed that the RCP restart criteria on 25F subcooling mard n (SCM) was varf useful to the oparator.
i 3everal areas required clarification or expansion, however.
First, RCP restart should not be attempted unless pump emergenc/ NP3d limits are met.
The onif exception is that NP3H requirements are walved, with pump restart allowed during certain inadequate core cooling conditions as specified in plant procedures.
Pump " bumps," however, should be attempted even if NPSH requirements are not mat.
- 3econd, loss of subcooling may occur after the RCP's are A
restarted.
Collapse of steam volds in the RC3 may cause voiding of the pressurtzer, resulting in a loss of SCM. RCS temperatures may ba hotter in the tube region than in the core if natural circulation has baen lost. Mixing of this hotter water with the cooler core water will causa a decrease in the SCM.
Several pump starts maf be required before subcooling margin stabilizes above 252*.
Allowing two minutes of RCS flow is helpful in eliminating both steam volds and temperature gradients so that successive restarts will be successful.
4.2.2.4 Core Flood rank 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, providing cooling water which was not required, since SOM was si ntficant was that CFI initiation maintained. Most d
delayed RC3 depressurization. Neither the LOCA nor tube rupture procedure provides any guidance about isolation of the core flood tanks. Therefore, this IDR has been revised to provide guldance about when to isolate the core flood tanks (sea Section 5.2.11).
I 4.2.2.5 RCP Trip Criterion QuestLons arose regarding tne actions to be taken if RCPs were not tripped within 2 minutes of a loss subcooling margin.
Clarification was provided using the guidance of the IMI-1 ATOG (Ref. 28) which requires the operator to keep the RCPs in each loop running if the two minute time limit is exceeded.
If RCPs are subsequently tripped, the RCS maf be voided enough to uncover the core. RCPs should De run for at least 7000 seconds to assure that the core will not uncover (based on Appendix K assumptions).
If pump damage maf occur, then one pump in each loop should be tripped.
If either of the running pumps fails, the telpped pump in that loop should be started.
For simplicity, the guidelines in this TDR call for 1 RCP to be run in each loop.
This provides sufficient flow to cool the core (Raf. 28).
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TDR 406 Rev. 2 1
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.
i 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).
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 NPSH curve of Figure 6 and while waiving fuel
+
pin-in-compression limits.
+
This will m imize primary to secondary differential pressure, thus minimizing the leak rate.
5.2.2 Reactor Coolant Pump Trip Criterion
- Trip Reactor Coolant Pumps (RCP's) when SCM is lost.
If RCP's are not tripped within 2 minutes of loss of 25F' SCM, then Note:
- i
- run 1 RCP in each loop.
5.2.3 Reactor Coolant Pump Restart Criteria
- When the required subcooling margin (25F") has been established, restart 1 RCP per loop.
If unable to start an RCP in one loop, start 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 lef t running.
5.2.4 Reactor Coolant Pump NPSH for Emergency Operations
+
+
during a cooldown with a tube leak.
5.2.5 High Pressure In_iection Throttling Criteria
- Throttle MPI when SCM requirements are met and pressurizer level comes back on scale.
(Note that the other HPI throttling criteria remain unchanged.)
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TDR 406 Rev. 2 Page 37 f
74 5.2.6 OTSG 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% simmultaneously.
5.2.7 OTSG Isolation / Steaming Criteria
Note:
Do not close MS-V1D until an alternate source of gland steam is available.
When RCS Thot is less than 540F' the af fected OTSG 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).
Note:
If both OTSG's are leaking and isolation is required based on of fsite dose projections, first isolate the OTSG with the higher le_kage.
If such a distinction cannot be made, isolate one OTSG and reevaluate offsite dose projections.
5.2.7.1 Pressure Control of an Isolated OTSG
- Steam the affected OTSG(s) only:
1 1.
To keep OTSG pressure below 1000 psig, 2.
If the plant is on feed and bleed cooling. OTSG level must be controlled below 600 inches on the wide range indication.
If the OTSG must be steamed open the Turbine Bypass Valves or Atmospheric Dump Valve on the affected OTSG.
5.2.8 Cooldown Rate During a Tube Leak Event
- The cooldown rate shall be simited to a maximum of 1.67F'/ min (100F'/hr) whether on forced or natural circulation.
Note:
Steaming of the OTSG's may net be required if OTSG level 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 af fected OTSG.
+
3.
Supply MFW thru the startup control valve at about 6
.05 x 101bm/hr (if MFW is not being used).
+
If the differential temperature approaches 100F*, stop the cooldown
+
+
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 OTSG's are Unavailable for RCS Heat 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 Guideline Flow Chart
- Appendix C includes a flow chart and explanatory text showing the
+
+
logic path of the tube rupture guidelines.
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E3R 406 Rev. 2 Page 39 of 74
- 6. 0 CONCLG310N3 AND RECOMM2NJAfl0NS The ability of the plant to handle befond design basis events can be substantially increased and the RC3 leakage can be reduced for design basis leaks with the adoption of the following changes / additions to tube rupture 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 guidelines of this guideline.
4.
Provide the RCP NPSH limits of Figure 6 for use during emergency conditions.
5.
Walve fuel pin-in-compression limits.
6.
Control plant cooldown to limit the tube /shell delta T to 70F*.
7.
Revise procedure entry point conditions to be leakage greater than 50 gpm.
8.
Incorporate criteria for initiation of feed and bleed cooling into the tube rupture procedure.
9.
Adopt criteria for opening T8V's/A3V's if RCS pressure stafs above 1000 pstg during feed and bleed cooling.
10.
HP1 throttling should be allowed when subcooling is regained and pressurizer level is on scale.
11.
Core flood isolation criteria be incorporated into emerdency procedures dealing with LOCA, tube rupture and steam line breaks and in operating procedures dealing with dorced and natural circulation cooldown.
12 Decaf heat removal system initiation under energencf conditions be allowed at 300F.
It is further recommended that these changes be implemented prior to restart of fMI Unit 1.
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TOR 405 Ree. 2 Pago 39 of 74 TABLE 1(2)
Tabular Values of RCP Emergency NPSH Requirements FOR 2 RCP PER LOOP OPERATION INDICATED TEMP.
ALLO'JASLE INDICATED PRE 3S.
(F)
(P31G) 94.4 300.0 (1) 194.4 300.0 (1) 294.4 300.0 (1) 344.4 310.8 394.4 413.9 444.4 567.4 544.4 1187.1 FOR 1 RCP PER LOOP OPERATION INDICATED TEMP.
ALLO 4ABLE INDICAIED PRESS.
(F)
(PSIG) 94.4 300.0 (1) 194.4 300.0 (1) 294.4 300.0 (1) 344.4 302.8 394.4 405.9 444.4 559.4 544.4 1178.6 NOTE:
1.
LLattations based on the #1 seal differential pressure requirement.
l 2.
Entire table was added in Rev. 2.
+ Rev. 1
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i
IDA 406 Rev. 2 Page 41 of 74 TABLE 2(1)
Pressurizer Spray Flow for Various Pump Combinations NUMBER OF RC PUMP 3 RUNNING SPRAY FLO'J SPRAY LOOP OPPOSITE LOOP
(%-FULL FLO4) 2 2
100 2
1 92 2
0 84 1 (Spray line next to running pump) 2 60 1 (Spray line 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 line next to idle pump) 1 38 1 (Spray line next to idle pump) 0 26 0
2 20 j
1 O
1 0
i As a rule of thumb tripping one pump in each loop will provide a good balance I
between the spray flow rate and the heater capacity.
It will also provide good j
forced circulation for cooldown.
i NOTE: The following table will give general guidance for the effects of running various pumps. This table was calculated for normal operating conditions.
NOTES:
1 Reproduced from TMI-1 AIOG (Ref. 28), "Best Methods for Equipment Operation",
Table 6.
l 2.
Entire table was added with Rev. 2.
i
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TDR 406 Rev. 2 Page 42 of 74
7.0 REFERENCES
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, 1)82. Ginna Station.
5.
" Procedure No. El.0.
Safety Injection Initiation." February 23, 1982. Prarie Island Nuclear Generating Station.
6.
" Procedures E-1.3 and E-1.4 S/G Tube Rupture." June 3, 1982.
Prarie Island Nuclear Generating Station.
7.
Duke Power Co. "EP/0/A/1800/17.
Steam Generator Tube Rupture."
Oconee Nuclear Station.
8.
INPO.
"SOER 82-16 Draft."
January 4, 1983.
9.
INPO.
" Analysis of Steam Generator Tube Ruptures at Oconee and Ginna" INP0 82-030.
November 1983.
10.
Babcock & 'w'ilcox " Abnormal Transient Operating Guidelines. Three Mile Island Nuclear Station - Unit One."
June 1981.
11.
GPU Nuclear Corp. " Acceptability of Intentionally Loading Main Steam Lines During an OTSG Tube Rupture Event."
Document No.
Il0lX-5320-A18. July 30, 1982 - Engineering Mechanics section.
l 12.
L. C. Lanese "TMI-l OTSG Tube Rupture Basis.for Design Leak i
Rate."
August 31, 1982.
SAPC/118.
13.
M. Sanford "NPSH Requirements for Piggyback Safety Injection Operation - TMI-1."
December 27, 1982. MSS-82-584.
I 14.
J. Veenstra.
" Fuel Pin Compression Limits During an SGTR I
Event." TMI-83-009.
January 20, 1983.
l 15.
S. D. Leshnoff.
" Mechanical Integrity Analysis of Thl-1 OTSG l
Unplugged Tubes." TDR 388. GPUNC. March 9, 1983.
l
+ Rev. 1
(
- Rev. 2
IJR 406 Rev. 2 Paga 43 of 74 16.
P. S. Walsh. "IDR 400 Draf t Guidelines for Plant Operation with Steam Generator Tube Leakage."
PA-893. Februaef 15, 1983.
17.
Darrell G. Eisenhut, to H. D. Hukill. March 4,1983.
Docket No.
50-289. March 4, 1983.
U.S. Nuclear Regulatorf Commlssion.
18.
N. K. Savant, Dunn, B. M., Jones, R. C., Response to GPJ letter dated August 28, 1980 (TMI-1/E 1203).
Document Identifier 12-1121718-00. September 26, 1930.
Babcock & Wilcox Co.,
Lynchburg, Virginia.
+
19.
H. D. Hukill (GPJN) to D. G. Eisenhut (NRC).
"RCP Irip on 25F*
Subcooling Margin." March 31, 1983.
3211-83-017 20a. W. Drendall.
" Saturation MargLn Monitor Inaccuracf (Non-Accident Conditions)".
Calc. No. C-1101-665-5350-002.
June 14, 1983.
20b. G. J. Sadauskas.
"IMI-1 Saturation Margin Monitor Loop Error Analysis." Calc. No. 1101X-322B-009 Rev. 2.
June 14, 1983..
GPJNC.
21.
G. L. Lehmann.
"0TSG Leakage and Operating Limits."
IJR 417.
GPJNC. March 1983.
22.
TMI Unit 1.
" Plant Limits and Precautions. OF 1101-1.
Rev.
14."
Februarf 26, 1982. GPJNC.
23.
"Detarmination of Minimum Required Tube Wall Thickness for 177-FA Once Through Steam Generators." BAW-10146. October 1980.
+
24.
J. Veenstra to D. G. Slear.
"Ihermal Shock Clartiteation For Small Break Operating Guidelines." September 9, 1982.
EdI-82-071.
+
25.
M.H. Kostrey, Memo to L.C. Lanese "IMI-1 Reactor Coolant Pumps NP3H 11mits Westinghouse Pumps" MC-1752, March 30,1983
+
26.
L.C. Pwu.
GPJN Calculation # 1101x-5450-013, Rev. 2 "MLntmum Subcooling Margin for RCP Operation." August 5, 1983.
27.
L.C. Pwu.
GPUN Calculation # 1101X-5450-015, " Saturation Margin Monitor Adjustment on Elevation Difference Between the Pressure Tap and the Tops of the Hot Leg."
Juif 29, 1983.
i 28.
B&W Abnormal Iransient Operating Guidelines (AIOG).
74-1124158-00 Lfnchburg, VA, April 24, 1933.
29.
Telecon with Ralph Rosser (B54) bf L. C. Lanese. June 13, 1983.
30.
Babcock & Wilcox Co.
" Evaluation of S3 LOCA Operating Procedures and Effectiveness of Emergenef Feedwater Spraf for B54 - Designed Operating NSSS".
B&W Doc. ID 77-1141270-00. Lynchburg, VA.
February, 1983.
l l
+ Rev. 1
- Rev. 2
IDR 406 Rev. 2 Page 44 of 74 a
31.
L. C. Pwu. GPJ Calc. No. 1101X-5450-014.
"I41-1 Cooldown Rate using Atmospheric Dump Valves", August 3,1983.
32.
J. P. Logatto.
"IMI-1 Deca / Heat Removal Sfstem".
June 7, 1933. MSS-33-504 33.
N. G. Ir1kouros to R. J. Ioole. "Infroid Dose Limit for OTSG Isolation during SG Tube Rupture", SAPC #140, dated Juif 13, 1983.
34.
TML Gntt 1 OP-1103-6, Rev. 24.
" Reactor Coolant Pump Operation".
Step 3.3.2.
A 35.
J. R. White.
RSLAP5 Analysts of Steam Generator rube Rupture Transients in a Generic Lower Loop Plant.
R.P2420-4.
Nuclear Saf et/ Analysis Center.
Palo Alto, CA, 1983.
I k
+ Rev. 1
- Rev. 2
FIGURE 1 STEAM GENERATOR TUBE RUPTURE GUIDELINE DEVELOPMENT-ACTIVITY NETWORK
~
ISSul ORAf f IP 1202 5 FORflMt1ATOR CH!CROUT LlilRAT URE I 3 I3 is 42 R
//
',/g sirs,0 //
m23 82 i 2 +,82 i20s 82 i i23082 n iin,82, n 83
,2,,,s,HG
,,,,,8,
,,,,, M A,,,
,8 2,
0Ib* N Alt 0W A8t t ICENSI l
ISRG PLA41 Olt Of lNif 0RAf f CRITtRIA 00L DOWN RATE APPR0 val 3
APPROVAL
/ROCfDURES Gul0llINE S IOR 50R $TI AMlMet ASID ON YUSE Of $1Rs I
Of $lRs Qf / 4 / O PATH 1 fyfjjj i
"" Z2
/Jims i y Baw FOR At0G
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _0 4_8 2 _ _ _ _ _ _ _ _ _ _ _ _ _ _W 15/ 8 3_ _ y _ _ _ _ _ _ _ _ _ _ y _ _ _ _ _Y_ _
1
/^
,0
/
""5'""*"
"' ^
a' o"^"
""^t PATH 2 Of M00f t lN80 PROG REV ROG Rf R[ ADV iDR R
R(PORT g
REPORT T0 tPRi tTI Mt MIG A ALV$ts g
I'10/4 i' 22 82}
8 f
I
- 83 0 83 Il g
i g
rr7 a
e e6 1
FORMul AT E A M stNC I
BIGIN MUlf COMPltTE Mulf SG10 PROJE C T.
SWGli SGTR SGill ANAL SGTR ANAL
]
PL AN IIM t
PATH 8 i
PATH 8 M 1182 11 15 82 g
1
--8 1
11/30.82 10t82 3
I ZilNE MUlffPil I
g
$ Gill ANALV515
/
lN A Pnip 11182 II'Is 82 l
g g
10182 115 RE TRAN 0t(It p gf gjg,, f gg,$,GiR DR g
94 f
4fIy '
/
g l
"u/
jo 8 82 RivieW )
COMPlt TI Mulf i
PATH 3 i2 83
< CoMPti n o SG'R =l i
l PATH 14 11,15 82 d
I-_---_-_
H/30 82 12 23 82 g
o, e o, e, e g
5 GIGIN SINGil I
3 OIS'N INOLI DMPLIll SINGL i
SGTR WITH N41 l -
COMP,t,i fI,$1,4,Gl,i SGIR WUN N'I I
SGTR WITH isat OlviL0P UPPE R ggy g
RC COOLDOWN H{ A0 Mil 0it
/
CIRC C00l00Wg CIRC COOLDOWN g,IRC CM,g gy00WN, C
- f f.
gggggggg g ggy l ANALV$1$ (2 SG)
" 2 8t,GI*g
$ "G,,,
~
dis 82
"'3' 82 M
sii -
"" ' v20 82 Hf A0 M00tl 80TH 5G 12 83 FATH I4 I
!]f//
l
?YM to s 82 - _ yyg y ':
BIGIN ANAL O
l MULI SGTR D M C' Sf4Gli SG PATH 10 C'
O 10882 m
12 82 b
e
TDR 406 FIGURE 2 Rev. 2 Page 46 of 74 Break Flow for Single Ruptured Tube 1050 i
i i
i 945 1
il 111 840 E 735 a.
N 630 5
l0 525 se CL
$ 420 ai 315 210 1: 25*F SC, PUMPS ON II: 50*F SC, PUMPS ON 105 111: 50 F SC, PUMPS OFF I
I 0
O 10 20 30 FLOW LBM/SEC W
e l
OGURE 3 Effect of RC Pump Operation on Integrated System Leakage for Single Ruptured Tube 300 I
I I
I I
I I
I I
i i
n h
160 25 F SCM, RCP'S ON
--- 50 F SCM, RCP'S ON
=
E 50 F SCM, RCP'S OFF 140 S
a
's * -
m o
/
1 cc
/
/*/**".
120
/
j E
/
C:3
- /.**
/
g
/
u.
g 100
- ./
/
/**.
/
aa:
3gll
)
att
/
m 80
/
Q
.+*./*.- /
l C3 m
/
S 60
/ / *.*.
/
/
/
/
1
\\
r
/
/
a l
,/*/ /./
40 -
/*/'/
/h*./
p
/
/
20
.4/
- y g g e
9.< =-
l l
l l
l l
1 l
l l
l 0"
g _ _.
l 20 40 60 80 100 120 R
l TIME IN MIN E
1011400 ilEV. ?
Page 48 of 74 FIGUllE 4 Mass and Energy Capabilities of HPI and PORV 880 40
(.55)
/
/
800 - TWO HPI PUMPS f
- 35 f
t.s >
/
PORV ENERGY 720 R E M OVA L t.. **'"""""""""
..y,/
30
/ '....
o.e>
/
/
'E 640
/
/
(25 m{
2.0) f Y.x
/
m PORV LIQUID REllEF 500
/
20 E
/
(5.8) E d
/
G
/
480
/
15 I
ONE HPI PUMP
/
5-
/
E
/
'~
400
/
10 j
/
/
1
/
320
/
5 1
240 I
I I
l 0
0 500 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)
FIGURE 5 Rev 2 Page 49 of 74 Time Behavior of Subcooling Margin for a Spectrum of Ruptured Tubes 100 f
j 61 TUBE 80 I
m f
ai Q
60 -.I e
/
E r'
E i
3 40 3
/
E l
/
g 6 TUBES M
. g/
6 TUBES
<n
\\
N, '
m 20 - 4
,s, 1
s'/
l
~ l 26 TUBES 26 TUBES Y
u" a
<a 0
6 12 18 24 30 36 42 48 54 60 TIME (MINUTES)
,-wn---w,-.,._-,,----,,,e,
,.,-,-,,.._p.
,,--,.--r-
-,,n,
--,._p,...
aw,n,
..-p,y w_,.m-,,.-,-,----.m.-,
TDR406 R:v.2 50 M 7 FIGURE 5 RCP NPSH Curves 1600 j
Minimum NPSH for RCP Normal Operations OP 1101 Figure 1.0-5.6 1500
Emergency NPSH for 4 RCP operation or 2 RCP per loop.
I 1400
- - *- Emergency NPSH for 2 RCP operation
/
(one per loop) l 1300
- 25 F Subcooling Margin Curve.
I Note:
I 12 F and 10 psig instrument errors 1200 have been incorporated in the normal NPSH curves, while 5.6 F and 94.9 psig errors were for the emergency NPSH curves.
1100 l
5 a
I.
E 1000 S
/!
I
/,l l
N 900 E
ll u
5 800 e
/.:
6 lll
/'*
5 700
~
I.jll
/tl f *,.E 600
/.j!
///
00 f./ :!
f~
//l f*l l
400
/ *7 l
//
l u
300 f
l----
200
/
/
100
- ,..**.r 0
100 200 300 400 500 600 Indicated RC Temperature ( F)
TDR 406 Rev. 2 Page 51 of 74 4
APPENDIX A TMI-1 SGTR PROCEDURE GUIDELINE ANALYSIS I
+ Rev. 1
- Rev. 2
0 0
4 TDR 406 Rev. 2 Page 52 of 74 COMPARISON OF GUIDELINES AND EP 1202-5 REV. 16 TO THE 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, commitments or recommendations from varicus sources. The sources reviewed were the TMI-1 Anticipated Transient Operating Guidelines (draf t of 15 May 1981; hereinaf ter referred to as
,,' ATOG), " Clarification of TMI Action Plan Requirements" (referred to as NUREG 07)7), and the Safety Evaluation Report related to restart of Ginna (Ref. 2) (referred to as NUREG 0916) and the INPO draf t
(
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 commitments directly related to a tube leak emergency procedure were s
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.
l i
A-1 m
+ Rev. 1
- Rev. 2
.),
TDR 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 Comments Run RCP's with X
X X
X low RCS pressure RCP restart X
X X
X X
No specific guid-ance provided for RCP restart with a " solid" presserizer.
Subcooling margin X
X X
X HPI throttling X
X X
X l
Steam line X
X X
X ATOG does not flooding recognize TMI-1 capability to I
flood steam lines l
without damage.
Cooldown of X
X X
Guidelines provide damaged OTSG for continued steaming of af-fected OTSG for cooling except when OTSG isola-tion is required.
Specify entry X
X X
X Symptoms.
threshold Method for plant X
X 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 CG relief valve.
s A-2
+ Rev. 1
- Rev. 2
TDR 406 i
Rev. 2 Page 54 of 74 0
TABLE A-1 (continued)
Source Addressed By Requirement ATOG 0737 0916 SOER 1202-5 406 Comments Affected SG X
X X
pressure control OTSC tube to X
X shell differen-tial temperature
+
Criteria for X
Guidelines pre l
using ADV's in sume that preference to steaming to main condenser condenser is always preferable to steaming to atmosphere.
Consider multiple X
X Guidelines pre-tube ruptures pared in con-sideration of Consider tube leaks X
X these two cases.
in both OTSG'e l
0 HPI on inadequate X
X Not specifically 0
SCM stated in TDR 406, but it is an
)
implicit re-i quirement of the HPI throttling 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 of fsite power available or unavailable, but they do provide guidance if the equipment dis-abled by LOOP is unavailable.
A-3
+ Rev. 1
- Rev. 2
TDR 406 Rev. 2 Page 55 of 74 i
TABLE A-1 i
o (continued)
Source Addressed By Requirement ATOG 0737 0916 SOER 1202-5 406 Comments OTSG 1evel X
X X
control Primary pressure X
X control with-j out pressurizer I
4 spray Isolation of X
X X
i af fected OTSG 1
i i
I J
l s
t A-4
+ Rev. 1
- Rev. 2
f0R 406 Rev. 2 Page 56 of 74 A.1 Interpretation A.1.1
" Requirement" Coluan These are paraphrased descriptions of guidelines, requirements, commitments, or recommendations f rom source documents.
A.1. 2
" Source" Columns These columns define the origin of the requirement considered in this comparison.
A.1. 3
" Addressed 3/" Columns these columns define the document which answers the requirements.
If a mark appears in the 1202-5 column without a corresponding mark in the TOR 406 column, it means that the guidance in EP 1202-5, Rev. 16 should be retained in the revision that incorporates One duidelines of Section 4 of this TDR.
If a mark appears in both columns, it generallf means that the guidelines in this T3R supercede the guidance in EP 1202-5, Rev. 16.
If a mark appears in the f0R 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 informatten if necessarf.
The duidelines in this T]R supercede the guidance in EP 1202-5, Rev. 16.
If a mark appears in the IDR 406 column alone, it denotes a ned guideline to be incorporated into the revised EP 1202-5.
k k
k l
i A-5
+ Rev. 1
- Rev. 2 l
TDR 406 Rev. 2 Page 57 of 74 APPENDIX B PROCEDURE CHANGE SAFETY EVALUATIONS l
4
+ Rev. 1
IDA 406 a
4 Re v. 2 Fada SS of 74 B. O PROCEDJdE CHANGE 3AFErl EVALJAfl0A3 The purpose of this Appendix is to address the safetf implications of the kef changes required to implement the tube rupture guldelines described in this TDR.
1.
RCP trip on loss of subcooling mard n (SCM) i 2.
Change in SCM 3.
Shell/ tube delta I of 70i' during emergencies 4.
Revised ACS NP3H curve 5.
Relaxatloa of fuel pin in compression limits 6.
Or3G tsolation criterta 7.
RCP restart criteria 8.
HPI throttling at 0" instead of 100 inches.
9.
Leaving ADV open wneq no Or3G heat sinks and ACS is above 1000 psig.
- 10. Isolation of core flood tanks 22ch of these Ltems is addressed belos:
3.1.
RCP felp on Loss of Subcooling Margin
+
In a letter dated March 4, 1983 to H. D. Hakill (Rev 17), the NAC superceded the actions required in IE Bulletins79-05C and 79-06C.
Tne staff has instead concluded that "the need for ACP trip folloaing a transient or accident should be determined by each application on a case-bf-case basis considering the Danet's Group input."
For several fears, the B&J Owner's Group has supported the concept of ACP trip on loss of subcooling margin.
In Reference 19, GPJdC informed the NRC of thele reasons for revising the trip criterion to loss of subcooling mard n.
The safetf evaluation for this change has been transmitted i
separate 1/ to the EMI plant staff for revies and approval.
B. 2.
Change in Subcooling Margin GPJNC nas evaluated the instrument error associated with the subcooling margLn monitor and alarm.
Jnder normal containment conditions, the loop error is + 10.120 (Ref. 20a & b).
Onder the temperature and radiation environment of a small break LOCA, this error is no worse than - 21.71*.
The basis for the original 3CM was B-1
+ dev. 1
- dev. 2
~_
IJA 405 Rev. 2 lade 59 of 74 5/* deomettf correction plus 452' string inaccuracf.
decent calculations (Ref. 27) have shown tnat onif 1.302 geomettf correction to the top of the not leg is required.
Inerefore safet/
mergins are not decreased bf this change.
A complete safetf evaluation has been prepared and transmitted to the site separatelf for review and approval.
I
- 8. 3.
Chande in Shell to rube Delta r The existing emergenef limit for tube /shell delta I at TMI-1 1s 100l*.
The ilmlt is being revised to reduce tensile stress on leacLag OfSG tubes.
Previous 1/ the limit of 100l* was based on stresses to intact tubes.
Indeed, the 11mt: has been increased to 150/* bf Babcock & Wilcox; and, it is valid in the absence of degraded OTSG tubas.
fne more restrictive 102* llatt increases plant safet/ 11mits b/
reducing the likelihood of propogating a crack.
Inis analftical work is documented in Reference 15.
This change can be made under the provtsions of 10 CFA 50.59 because it does not affect technical specifications. Since tube stresses are reduced, plant safet/ llatts are lacreased and the two additional criteria of 10 CIA 50.53 are also met.
Namelf, there are no new accidents introduced into the plant that have not been previous 1/ analfzed.
Since shell to tube delta T has not been explicitif addressed in the FSAA, existing plant safet/
margins have not been decreased.
In fact, plant safet/ margins have been increased since the allowable delta I has been decreased.
B.4.
ACP NPSH Limits Reduced NPSd limits bring the pump closer to a point of cavitation.
However, NPSR requirements have been reduced for lower temperaturas as
+
3etermined by the pump manufacturer Westinghouse in deference 25.
Margins have been modified based on safet/ margins Ldentified by the pump manufacturer, therefore, the probab111tf of pump cavitation has not been increased and plant safetf margins are protected. Neither are technical specifications affected.
The operation of reactor coolant pumps at low AC3 pressures does not introduce anf new accLdent or transient other than those alreadf anal zed in the ISAA.
Pump f
operation is allowed at tuese lower ACS pressures but at a higher subcooling margin.
Since real plant subcooling margin is still being maintained as discussed in Ltem 2, there is no reduction in plant safetf. Operation of the reactor coolant pumps increases plant safety margins with respect to thermal shock, increased JNB ratios, and improved capabilittes for degassing the reactor coolant afstem under tube rupture conditions when mlnimum subcooling margins are being maintained.
B-2
+ Aev. 1
- Rev. 2
TDR 406 Rev. 2 Page 60 of 74 B.S.
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.
Fuel pin in compression limits have been established in order to maintain cladding integrity. Waiver of these limits does not reduce pin integrity although reanalysis by B&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 analyzed in the FSAR. Similarly, plant safety margins will not be reduced, namely, cladding integrity will not be challenged.
B.6.
OTSG Isolation Criteria The existing steam generator tube rupture procedure, EP-1202-5, allows the operator to isolate the affected steam generator anytime RCS pressure is below 1000 psi. The revised criteria would allow steaming of the OTSG until BWST level is 21 f t. 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 high radiation is keyed towards 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 staf f (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 affected 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).
B-3
+ Rev. 1
- Rev. 2
TJA 406 Ac t.
2 Page 61 of 74 B.7 + R0P Restart Criteria
+
Ihe RCP Restart Criterion assures that the pumps are not restarted
+
until the core is adequatelf subcooled.
(Note that there are other
+
RCP restart unrelated to this criterion).
Reference 19, and Sections
+
3.1 and 8.2 demonstrate that the core is adequate 1/ subcooled with
+
RCP's running and a 25/* subcooling margin.
No Techalcal
+
SpecifLcations are affected.
No new accidents or transtents are
+
Introduced into the plant; no safet/ margins are decreased and no
+
accident consequences are increased. Allowing an earlier pump restart d
+
gives the operator greater control over the plant since forced flow is
+
preferable to natural circulation cooling.
This change can therefore
+
be made under the provisions of 10C/R50.59.
B. 8 + HPI Throttling at 0 inches Indicated Level
+
Ihe safety aspects of throttilng HPI on 252* subcooling margin are
+
addressed in section B.2.
Core coolabllltf is not dependent on the
+
pressurizer level at whlch HPI is throttled L.e.,
core cooling is onif
+
dependent on an indleation that the core coolant is subcooled.
The
+
basis for requiring pressurizer level is so that the existing
+
pressurtzer heaters are covered with water so that thef can be
+
energized.
Energizing the heaters before thef are covered causes thes
+
to born out.
On the other hand, there is no need to refill the
+
pressurizer to the 100 inch level at full HPI flow.
In fact, this
+
flow rate is undesireable for two reasons. Rapid filling of the pressurizer causes an RCS pressurization during conditions when
+
pressurizer sprafs are unavailable.
Insurges to the pressurizer
+
compress the steam space.
Pressurizer pressure must be reduced cither
+
bf sprafs (if available) or pressurizer ventlng (vent 11ne or PJAV).
+
Controlling the HPI flow mLntsizes the insurge rate, and hence, the
+
pressurization.
ThLs reduced pressurization provides more margin to
+
the 100/* subcooling curve thereby minimizing challenges to the
+
thermal shock /brLttle fracture limit.
+
Ihis change does not represent an unreviewed safety question because:
+
1.
do change to the rechnical dpecifications is required
+
2.
No new accidents are introduced to the plant (the operator is stL11 required to cover the pressurizer heaters before energizing them), and
+
3.
The consequences of previousif analyzed accidents / transients Ls not increased.
It is less likely that the operator will violate the 100/* subcooling margin.
Core coolability is not dependent on established pressurizer level, but onif an adequate subcooling margin.
3-4
+ Rev. 1
- Aev. 2
TDR 406 Rev. 2 Page 62 of 74 ADV's Open when RCS is above 1000 psig with no OTSG Heat Sink's B.9
+
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 minimizes 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.
B.10
- Criteria for Core Flood Tank Isolation 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) HPI 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 HPI 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 large 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 suf ficient 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 gpm).
If the RCS is 25F subcooled with the RCS below 700 psig, then the CFT can be isolated.
The core flood tanks also function in one non-LOCA situation - steam line break accident. For large steam line bretk 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 have inserted.
If both these conditions are met, then a plant procedural change can be made without introducing an unreviewed safety question.
B-5
+ Rev. 1
- Rev. 2
TDR 406 Rev. 2 Page 63 of 74 APPENDIX C GUIDELINES FLOW CHART
+ Rev. 1
- Rev. 2
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 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., confinning reactor trip, or making radiation surveys of the a
secondary plant).
The decision points immediately following a double-line box are meant to force the operator into a " thought-loop" so that if conditions change, 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 gpm, the operator should continually inquire as to whether the Reactor Coolant System pressure and temperature are within the capability of the Decay Heat Removal System.
If so, when the operator should obviously change the RCS heat removal mode from steaming the OTSG's to using the DHRS. If not, then the operator should continue to ask whether the RCS conditions are suitable for forced flow cooling via OTSG's, i.e.,
is subcooling inadequate, are the OTSG'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 depressurize the plant so as to minimize primary to secondary
+
+
leakge and thus, it is hoped, offsite doses. The process involves
+
recognition of the event, shutting down the plant, and cealing down the plant to the point where the Decay Heat Removal System can remove
+
+
core heat.
C-1 4
+ Rev. 1
- Rev. 2
~ _ _
TDR 406 Rev. 2 Page 65 of 74 C.3 +
ENTRY POINI t
the procedure will ba entered when a primarf to secondarf leak is
+
encountered that requires the plant to be shut down.
The sfmptoms of
+
a tube leak requiring shutdown are described in IJR 400 (Res.16).
C.4 +
PLANf SHJIDOWN
+
Ihe rate of plant shutdown f rom 100% power wl11 be determined in part
+
bf the nagnitude of the RC3 depressurization due to the leck.
If the
+
leak is small (the Makeup Sfstem is able to keap up withit), then the
+
plant can be shutdown at a rate commensurate with equipment
+
capablitties and, to a certaLn extent, the leak rate. When the
+
reactor and turbine are off line, the plant is ready to enter the
+
cooldown phase.
+
Howaver, if the leak results in RC3 depressurization to the trip
+
setpoint, the reactor and turbine will be off line immediatel.
The f
+
ensulng translent 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, HPI
+
+
must be lamedLatelf actuated and the Reactor Coolant Pumps (RCP's)
+
must be immediatelf trippad. The GI3G's must then be evaluated for
+
suttability as heat sinks for the RCS.
l t
If the shutdown transtent does not result in a loss of subcooling
+
margin, the OISG's must still be evaluated for suitability as RCS heat sLnks.
+
If neither DISG can be used because of high offsite doses or low BW3r
+
level, then the cooldown will proceed directly using the HPI " feed and bleed" method.
+
For the balance of tne discussion in this saction, assume that HPI
" feed and bleed" is unnecessarf.
+
If the RCP's are off, Emergenc/ Feedwater flow to the OTSG's must be
+
confirmed.
The ICS will automatically control OrSG 1evel at 50%: on 1
l
+
the Operating Range if the RCP's are off.
If subcooling margin is
+
25F, the. operator must manually ratse the level to 95% to promote
+
two-phase natural circulation in the RCS.
+
Since a forced circulation cooldown is the most preferred moda, the
+
RCS conditions should be evaluated for RCP restart.
If subcooling margin is regained and the RCP NP3H limits are met, 2 RCP's should be
'+
rastarted.
If the pumps cannot be restarted, the cooldown must
+
proceed by natural circulation.
C-2
+ 2ev. 1
- Rev. 2
IJA 406 Rev. 2 Page 66 of 74 C. 5 + PLANf C00L33WN
+
During the coollown, RCS conditions must be continuously evaluated to
+
ensure that the cooldown mode is appropriate and to determine whether
+
conditions are suitable for the Deca / Heat Removal Sfstem.
+
Re3ardless of coollown mode, the following items, may be encountered
+
while cooling down.
C.5.1
- HPI Throttling The existing H?I throttilng criteria are unchanged with the followind exception: HPI maf be throttled when suocooling is regained and pressurizer level comes on scale.
C.5.2 + OrSG Steaming
+
Ihe affected Of3G may be steamed for RCS heat removal purposes, but it
+
must be steamed to avoid lifting the Main Steam safetf valves, prevent
+
premature Steam line flooding, keep OISG pressure less than RCS
+
pressure, and control OISG tube to shell differential temperature.
C.5.3 + OfSG Shell to Tube Dif ferential Temperature It is necessarf to minimize shell to tube differential temperature to
+
minimize tensile stresses on the OT3G tubes. As noted above, steaming is one waf to accomplish this; another is to decrease the cooldown
+
rate; a enled is to use main Feedwatet to cool the lower dowacomer.
C.5.4 + OISO Pressure Control When RCS Pressure is Greater Ihan 1000 psig
+
During a natural circulation cooldown or an HPI fe2d and oleed
+
cooldowa, RC3 pressure may staf nigh.
2mergeacf Feedwater can be used
+
to quench the steam space; if the OfSJ is flooded, inventorf can be
+
relieved via the Turbine afpass Valves or the Atmospheric Dump Valves.
C.5.5 + Cooldown Rate
+
The cooldown rate should be limited to less tnan 1.6 F/hr to avoid
+
reactor vessel brittic fracture concerns.
It may not alwafs be
+
possible to observe this limit due to the effects HPI cooling and the
+
occassional necessity to steam the damaged OISG.
C.6 + EXIT POINr.
+
Ine operators exit the procedure when the RCS heat sink becomes the
+
Jeca/ Heat Removal Sfstem.
C-3
+ dev. 1
- Rev. 2
tritR n
ilGuilt C I 01Pf2
+N PROCiOURis P
SINGLE & MULTIPLE TUBE RUPTURE GUIDELINES e
\\
~ e*
MONif 04 8 OPERATE 4
A NO UNINit mTION AL IOII q
TU61 V E Ril VI y
$U8C00tlNG s
REAC1,0R f MARGIN gg5 ypg
%p
,gigiggg L E,A R g
i l
9,g/
III NO N0 h
e JL JL l
e l
e l
e
- Mui OOM SMtfi
{
N0 g,$g 3 yi r,,,
,,,p 00WN e
A V A'l'8II#
O AC $
P T
Ato O e
fin 10 8
b t
U$t a
.A s
e VES 8
vis j
a READY a
N0 f0R g
g C00l00WN OTSG S 7g 40 COOL 00WN
> 50 T0 ust C P,M W~
,ggg gggge s
Yt$
S
?!5 l
Al NO 8
8 RCP'S Jk A VAtl ABil NO C00t 00WN ON j
RC$
l e
FORCE 0fl0W e
P.i W. IN NC
[
USE NORMAL DMR PROC E DUPA S CAP e
yg g l
\\*
VE Rif Y,1NiilAIE Jk
/
g EMERGINCV Ef W vis a
J RCS E
NO
/ P,1 W/tg 8
OMR g
lf CA,P e
a E
RCP N Yl$
R($1ARI vis a
/
OR POSS e
2 4
e v
- a NO PUI RCS ON e
CICAY ME AT l
PIM0 VAL e
C00t00WN C00t 00WN COULING e
CN ON R
T%d INSPECT 8 RE PAIR e
Of5G S 8
R R 5 (D
NO P,T W <th 40 P T W,1N
&W O e
CHR DMR N
m CAP
- se I
.t,
8 k
g e
e TDR 406 Rev. 2 Page 68 of 74 APPENDIX D SIMPLIFIED EVENT TREE
+ Rev. 1
- Rev. 2
e TDR 406 Rev. 2 Page 69 of 74 D.0 SIMPLIFIED EVENT TREE The event tree on the following page shows possible enmhinations of circumstances that were considered that resulted in the guidelines presented in this TDR.
The guidelines explicitly stated in section 4, when incorporated into a revised OTSG Tube Leak / Rupture Emergency Procedure, will enhance the capability of THI-1 to deal with an OTSG tube leak. The purpose of this section is to describe the features of che revised procedure.
The discussion which follows assumes that the logic presented by the flowchart depicted in Appendix D is adopted for the revised procedure.
1 i
4 D-1
+ Rev. 1
- Rev. 2
. - - ~ -
- - - - = - - - - - - - -
FIGURE D-1 TDR 406 EVENT Rev. 2 TREE NO, RCP'S Page 70 Of 74
^ ## # 8
SIMPLIFIED OTSG i
CONDENSER EVENT TREE AVAILABLE 1 LEAKING OTSG CONDENSER AV A:1. A BLE NO RCP's 3
RCS SU8 COOLED CONDENSER AVAILABLE 4
RCP'S AVAILABLE BOTH OTSG'S LEAK ABOVE 1050 %
CONDENSER AVAILABLE BELOW 6
1050 PSI NO RCP's ABOVE 1050 PSI BELOW 7
1050 PSt TUBE RUPTURE CONDENSER AVAILABLE 8
RCP'S AVAILABLE 4
9 1 LEAKING OTSG ABOVE 1050 PSI CONDENSER AVAILABLE BELOW to 1050 PSI j
NO RCP's ABOVE 1050 PSI SUBCOOLING BELOW 11 MARGIN 1050 PSI LOST CONDENSER AVAILABLE 12 RCP'S AVAILABLE BOTH 13 OTSG'S LEAK ABOVE CONDENSER 1050 PSI AVAILABLE BELOW 14 1050 PSI NO RCP's ABOVE 1050 PS?
BELOW l$
MORE ingo CAPACITY REQUIRED I
e TDR 406 Rev. 2 Page 71 of 74 APPENDIX E PROCESS COMPUTER OUTPUT
+ Rev. 1
- Rev. 2
r o
e TDR 406 Rev. 2 Page 72 of 74 E.0 + PROCESS COMPUTER OUTPUT AND ALARMS E.1
+ Scope The process compater will have the following information available
+
+
with alarms as noted:
+
Subcooling margin OTSG Tube to Shell Differential Temperature
+
E.1.1 + Subcooling Margin 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 if:
+
SCM 25F*
E.1.2 + OTSG Tube to Shell Differential Temperature Calculate shell temperature as follows for each OTSG if all shell
+
+
thermocouples are operable:
+
Tshell = 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 Tcold is <535 inhibits
+
the alarm during normal operations.
+
E-2 i
+ Rev. 1
[
- Bawn R
(~
e o
IJA 406 e
i Aev. 2 Page 73 of 74
+
rable 6.1.4.1 Shell rhermocouple Substitution Falled i Sibstltute l r/c l
r/c 1
1 I
r5 l
r4 I
I I
r4 f5 I
I r3 l
r2 I
I I
l r2 I o.5 ;rt + r3)I l
r1 l
r2 l
i I
I4&I5 l
No Cale l
i I
r3 & r2 i
r1 1
I I
r3 &f1 l
I2 I
I r2 &T1 l
r3 I
It &T2&T3i No Calc
+
Wide dange Icold should be used in determining Or3G tube to shell
+
differeutLal temperatures.
Normally, use the wide range input from
+
rd-1-5AiB and rd-3-5ASB, although r2 959 and r2 961 can be used in
+
certain cases.
rable E.1.4.2 defines the data sources.
+
For each Loop, Calculate Snell to Tube I as follows:
+
Tr_3 = Ishell - Icold
+
Tr_3 should trigger an alarm state Lf
+
reo13 5352' and Ir_3 70i' e
2-3
+ Rev. 1
- dev. 2
o e p
TDR 406 a%
Rev. 2 Page 74 of 74 Table E.1.4.2 Wide Range Tcold Input RC-P-1 Tcold A B C D A Loop B 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 SB 0 0 X X Avg A Avg B 0 X 0 0 TE 4 5A Avg B 0 X 0 X TE 4 5A TE 4 5B 0 X X 0 TE 4 5A TE 2 SB 0 X X X TE 4 5A Avg B X 0 0 0 TE 2 SA Avg B X 0 0 X TE 2 SA TE 4 5B X 0 X 0 TE 2 SA TE 2 5B 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 5B X X X 0 Avg A TE 2 5B j
X X X X Avg A Avg B 0=
Pump Running X=
Pump Off Avg A = (TE 4 5A + TE 2 5A)/2 Avg B = (TE 4 5A + TE 2 5A)/2 I
TE.959 May be substituted for TE 2 5A TE 961 May be substituted for TE 4 SB
+ Rev. 1 E-4
- Rev. 2