ML20052H415
| ML20052H415 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/13/1982 |
| From: | Vissing G Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8205200359 | |
| Download: ML20052H415 (39) | |
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
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WASHINGTON, D. C. 20555 f
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MAY t 31982 o
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Docket No. 50-285 g
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LICENSEE: Omaha Public Power District (OPPD) j 4
FACILITY: Fort Calhoun w
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SUBJECT:
SUMMARY
OF MAY 6, 1982 MEETING WITH OPPD AND COMBUSTIO" ENGINEERING (CE) CONCERNING THE PRESSURIZED THERMAL SHOCK ISSUE (PTS) l Intmduction l
l j
{
This meeting was held in Bethesda, Maryland, May 6,1982, at the request of OPPD to discuss OPPD's response dated April 30, 1982, to our March 18,.
l
(
1982 request for information concerning the PTS issue as it relates to Ft. Calhoun. The meeting was also held to provide the staff with the latest information from CE concerning the PTS issue for CE plants.
i The meeting followed an agenda (Enclosure 1). The attendees of the meeting are identified in Enclosure 2 The material for CE presentations are included in Enclosure 3.
Summary of Discussions The CE discussion regarding our concerns included the following highlights:
l
.7egarding the Tripping of Reactor Coolant Pumps The Ft. Calhoun analysis assumed tripping the RCP within 30 seconds after the main steam line break (MSLB) accident. CE indicated this provided the worst transient and if the pumps were left on indefinitely the tempera-ture reduction at the downcomer would not be as great as for tripping the pumps at 30 seconds.
Regarding Timing of Operator Action For the SBLOCA & LORI CEN 189 identified two categories of operator actions:
1.
Open the PORVs 0 10 minutes or 2.
Restore feedwater at 30 minutes.
Restoring the feedwater at 30 minutes is considered the more severe transient.
Regarding Tennination of HPSI Flow in the MSLB Analysis The 150 day response assumed HPSI termination at 30 minutes.
Subsequent analysis which did not give credit for HPSI termination and assumed pumps are left on results in a 2-3 EFPY reduction of vessel life over the 150 day response results.
%[$h o
P
a PPD
' 2-Revi of Operating Experiences CE revi ed' operating experience of all CE operating plants -
identify overcool g events.
This ' review ~ represented 49 years of"r ctor' operating experienc The review consisted of a review of all'shu owns, LERs, abnormal
~
occurrence and operating ' ret orts.
The original data ase' consisted of hundreds of vents.
Sixteen (16) events met the 'scre ing' crit 4ria (see
~
Slide #12).
(2)' events" met the selection criter'.
The 16 events'are identified in e OPPD April 30 ' response.
Probability Anal is CE presented a comp hensive study in Slide 16 through 26 which provides ~a
~
base for further prob. ility risk analysis s it relates to PTS.
Reg. Guide 170 was used to define he frequency cat t ri es.
Slides 18,.2, 23 & 24 2
were edited 'to reflect ' e discussions.
The 'MSLB ' events are the mos t seie're
~
events and ~ range between' probability of 10-6 to 10-4 The higher proba-bility events are less sev than t MSLB' events.
Procedures and Training The Ft. Calhoun emergency proce es'have been modified'to address'the PTS
~
issue with emphasis on the pro r CS subcooling limits.
This includes caution statements concernin the oldown rate curve and instructions if the cooldown rate cu~rve is viol ed.
Al 0 PPD is a~dding discussions con-cerning the PTS issue to 't training ectures for operators.
Baltimore Gas and Electr c (BG&E) indic ed that procedure' changes have besn made at Calvert Cliffs to reflect more ncern for PTS.
These include preservation of press e temperature limit and instructions if'the reactor
~
is found to be on th wrong side of the P/T urve.
BC&E also i.n'dicated that operator training 11 address fracture mech' ics, background of PTS =and operating subcool' g limits.
Oi MM Guy S. Viss g, Project Manager Operating Re ctors Branch '#4 Division of L censing Enclos es:
1.
A nda 2.
tendance List 3.
lide Presentation cc / enclosures :
S > next page t
"r
~
- a s.
OPPD Review of Operating Experiences CE reviewed operating experience of all CE operating plants to identify overcooling events. This review represented 49 years of reactor operating experience. The review consisted of a review of all shutdowns, LERs, abnormal occurrences, and operating reports. The original data base consisted of hundreds of events. Sixteen (16) events met the screening criteria (see Slide #12). Two (2) events met the selectioncriteria. The 16 events are identified in the OPPD April 30 response, i
Probability Analysis CE presented a comprehensive study in Slides 16 through 26 which provides a base for further probability risk analysis as it relates to PTS. Reg. Guide 1
170 was used to define the frequency categories.
Slides 18, 22, 23 & 24 were edited to reflect the discussions. The MSLB eve ts are the most severe events and range between a probability of 10-6 to 10. The higher prcba-bility events are less severe than the MSLB events.
Procedures and Training The Ft. Calhoun emergency procedures have been modified to address the PTS issue with emphasis on the proper RCS subcooling limits. This includes caution statements concerning the cooldown rate curve and instructions if the cooldown rate curve is violated. Also OPPD is adding discussions con-cerning the PTS issue to the training lecture:; for operators.
Baltimore Gas and Electric (BG&E) indicated that procedure changes have been made at Calvert Cliffs 1 to reflect more concern for PTS. These include preservation of pressure temperature limits, and instructions if the reactor is found to be on the wrong side of the P/T curve.
BG&E also indicated that operator training will address fracture mechanics, background of PTS and operating subcooling limits.
CHSal signed ti?
Guy S. Vissing, Project Manager Operating Reactors Branch #4 Division of Licensing i
Enclosures:
1.
Agenda 2.
Attendance List 3.
Slide Presentation i
cc w/ enclosures:
See next page l
l ORB #4:0L omce>
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j suame >
......./ 82..........
cmy nnc ronu sia tisso3 nacu oa4o OFFICIAL RECORD COPY uso m i.. m m
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ORB #4:DL MEETING
SUMMARY
DISTRIBUTION Licensee:
- Copies also sent to those people on service (cc) list for subject plant (s).
H. Denton/E. Case D. Eisenhut G. Lainas Docket File NRC PDR W. Hazelton L PDR R. Mattson ORB #4 Rdg T. Speis TNovak T. Murley JStolz H. Thompson Project Manager -GVissing L censing Assistar.t-RIngram PKreutzer D. Basdekas L. Shao OELD R. Bernero AE00 IE E. Igne SShowe (PWR) or CThayer (BWR), IE T. Marsh Meeting Summary File-ORB #4 J. Austin RFraley, ACRS-10 J. Buzy Program Support Branch.
B. D. Liaw M. Vagins ORAB, Rm. 542 D. Ziemann BGrimes, DEP SSchwartz, DEP C. Johnson SRamos, EPDB FPagano, EPLB E. Abbott Meeting Participants Fm. NRC:
R. Johnson E. Goodwin SHanauer RKlecker EThrom MVi rgilio N. Randall FBlitton RWoods G. Zech Llois J. Roe RAClark C. Morris EGTourigny C. Serpan JClifford L. Shotkin W. Johnston A. Spano T. Dunning C. Rossi J. Strosnider
~
S. J. Bhatt
r-AGENDA FOR MEETING WITH OPPD AND CE CONCERNING 1
PRESSURIZED THERMAL SH0CK ISSUE MAY 6, 1982 e
e CE Owners Group Program Overview e Response to NRC Questions j
CEN 189 Impact of RCP Trip Times l
Overcooling Transient Experience
. ~ PTS Scenario Probabilities Additional Materials Cata CE Mixing Verification Results e 8G&E/0 PPD Procedures & Training i
4
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ATTENDANCE LIST FOR MEETING WITH OMAHA PUBLIC POWER DISTRICT CONCERNING PTS ISSUE MAY 6, 1982 F-NRC
_CE i
S. H. Hanauer Jim Pfeifer R. W. Klecker Dan Peck E. D. Throm Jake Westhoven M. Virgilio John J. Herbst F. B. Litton Dave Earles R. Woods Dave Ayres L. Lois i
R. A. Clark E. G. Tourigny OPPD G. S. Vissing J. Clifford Ken Morris j
Joe Gayer f
Maine Yankee Atomic Power Co.
BG&E J
Howard F. Jones D. W. Latham M. D. Patterson j
L. E. Titland I
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s AGEf;DA C-E Ownert '.'r ou;/NE,'2Lef/3Fi, P75 %
- r.5 kay 6, 1982 C-I Owners Group Program Gverview Response to NRC Questions
. CEN-189
. Imoact of RCP Trip Times
. Overcealing Tr?<.sie.r.t Experior.cc
. PTS Scenaric 3racabilities
. Additional Materials 6ata
. C-E Mixir.g Model Verification Results BG&E/0 PPD
. Procedures & Trainir.g e
e r
NRC SUBMITTALS ON PTS DATE SUBMITTAL SUBJECT 12/31/81 CEN-189 REPORT II.K.2.13, SBLOCA+LOFW 1/21/82 150-DAY LETTERS MSLB, A00 SCOPING STUD 11/30/82 LETTERS ADDITIONAL INFORMATION
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SUMMARY
OF PTS EVALUATIONS -
TRANSIENT SCENARIO PLANT II.K.2.13 MSLB
~ A00-CALVERT CLIFFS-1 E0L
+21 E0L FORT CALHOUN E0L EOL E0L MAINE YANKEE EOL EOL EOL PALISADES EOL E0L EOL MILLSTONE-2 E0L EOL E0L ST. LUCIE-1 E0L EOL EOL CALVERT CLIFFS-2 EOL l
ARTWISAS-2 E0L ST. LUCIE-2 EOL 3
WATERFORD-3 E0L I
SAN ONOFRE-2&3 E0L PALO VERDE-1,283 EOL "
l
y CEN-189 OPERATOR ACTIONS (SBLOCA + L-0FW)
I.
TRIP RCP 30 SECONDS AFTER SIAS II.
REC 0VERFEEDWATERAT30MkNUTES OR OPEN PORV'S AT 10 MINUTES III.
CONTROL HPSI + CHARGING TO TERMINATE EVENT L
6
--=-+>f
--'s--a w-
-e ww-w
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-_ _ _ __ y
3 i
i e
l Figure 1.
Sensitivity of reactor vessel downcomer water temperature to time of RCP trip for large SLB at zero power.
i 600 1
i 500 4 j
u-uI
!E l
r E
400 l
5 F-RCPs 0FF AT 300 SEC L
xN t
L 2
l 5
300 6
M i
5 o
d RCPs 0FF AT 30 SEC m0 200 i
e 1
0 W
i u
100 l
l i
1 1
0 i
O 500 1000 1500 l
TIME, SEC i
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i g-FIGURE 4-22 CASE 5 l
REACTOR VESSEL PRESSURE i
2800.0 i
2400.0 r
1 2000.0 3$
- i, d 1600.0 5
20E i
1200.0 BREAK SIZE O FT BREAK LOCATION N/A 800.0 HPI MAXIMUM OPERATOR ACTION RESTORE Fw 30 MIN 400.0 O.
- 1200, 2400.
3600.
4800.
6000.
TIME, SECONLS 4-65
7 FIGURE 4-23 CASE 5 COLD LEG add DOWNC0fiER FLUID TEMPERATURES 700.0 BULK FLUID TEMP., COLD LEG (FLASH)
FLUID TEMP. AT VESSEL WALL (MIX-UP) 600.0 MID-DISTANCE, CORE AND COLD LEG TOP OF CORE
- - MIDDLE OF CORE w'
L..._...
' T 500.0 i
L._:
..: A % s N. N ul N
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300.0 l
2
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1 200.0 100.0 O.
1200.
2400.
3600.
4800.
6000.
TIME, SECONDS a.na
y; FIGURE 4-1E l
CASE 2 l
REACTOR VESSEL PRESSURE 2800.0 i
BREAK SIZE 0 FT2 BREAK LOCATION N/A 2400.0 HP1 MAXIMUM OPERATOR ACTION 2 PORVS AT 10 MIN i
2000.0 s
E Q
1600.0 5
20 f
1200.0 800.0 400.0 O.
600.
1200.
1800.
2400.
3000.
TIME, SECONDS 4-59 W
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FIGURE 4-17 CASE 2 COLD LEG AiiD DOW14 COMER FLUID TEMPERATURES 700.0 BULK FLUID TEMP., COLD LEG (FLASH)
FLt1T D TEMP. AT VESSEL WAt.t- (MIX-UP) 600*0
-- -- MID-DISTANCE, CORE AND COLD LEG
- TOP OF CORE MIDDLE OF CORE 4
i' 500.0 -r"~~\\
L.
e19*.
\\\\.
I\\W 400.0 g
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300.0
'(s
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3
[-
200.0
-g,
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'j 100.0 O.
600.
1200.
1800.
2400.
3000.
l l
TIME, SECONDS 1
i
/6 Figure 2.
Sensitivity of RCS Pressure to time of RCP trip for large SLB at zero power.
2500
/
RCPs OFF AT 300 SEC
/
/
2000 RCPs OFF AT 30 SEC y
/
/
/
/
G
/
a-1500
/
a
/
5:
w-E JANUARY 1982 0
SCOPING TRANSIENT
.y 1000
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y, 500
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500 1000 1500 TIME SEC l
r
OVERC00 LING EVENTS AT OPERATING PLANTS WITH C-E NSSS PROCEDURE 1)
DEVELOP OVERC00 LING CRITERIA (SCREENING & SELECTION) 2)
REVIEW OPERATING EXPERIENCE TO SCREEN EVENTS 3)
CONFIRM EVENT DETAILS 4)
IDENTIFY OVERC00 LING EVENTS
OVERC00 LING EVENT CRITERIA e
SCREENING CRITERIA A)
EXCESS STEAM FLOW EVENT: OR B)
EXCESS FEEDWATER FLOW EVENT: OR c)
RCS PRESSURE DECREASE EVENT FOLLOWED BY ACTUAL SAFETY INJECTION FLOW.
e SELECTION CRITERIA
- 1) >100 F/ HOUR C00LDOWN RATE: AND U
- 2) > 100 F' TOTAL C00LDOWN; AND
- 3) > 10 MINUTES DURATION TO ALLOW FOR REACTOR VESSEL RESPONSE.
=
b 5
OVERC00 LING EVENTS SCREENING AND SELECTION RESULTS NUl1BER OF NUMBER OF UTILITY POTENTIAL EVENTS SELECTED EVENTS BG8E 4
OPPD 2
1 I
FP&L 0
CONSUMERS POWER 4
NORTHEAST UTILITIES 3
AP&L 3
1 TOTAL 16 2
SUMMARY
e
. ACTUAL OVERC00 LING EVENTS MUCH LESS SEVERE THAN EVENT ANALYZED, e
NO UNCONTROLLED REPRESSURIZATION ON EITHER EVENT, l
l N
9 9
POTENTIAL EVENT SEQUENCES FOR PRESSURIZED THERMAL SH0CK 1
l J. J. HERBST l
\\
MAY 6, 1982 g
SUMMARY
OF PROCEDURE
- 1. IDENTIFY INITIATING EVENT TYPES WITH POTENTIAL FOR COOLING OF VESSEL
- 2. DEFIllE FOUR PLANT STATES BASED ON EQUIPMENT STATUS
- 3. IDENTIFY SPECIFIC INITIATING EVENTS FOR EACH PLANT STATE
- 14. SELECT SEVERAL EVENTS FOR FURTHER ANALYSIS l
- 5. DETERMINE EXPECTED SEQUENCE OF EVENTS FOR EACH EVENT
- 6. CONSTRUCT EVENT TREES - QUANTIFY SCENARIO FREQUENCIES
- 7. CATEGORIZE SCENARIOS - MODERATE FREQUENCY, INFREQUENT, LIMITING FAULT
- 8. SELECT MOST SEVERE SEQUENCES l
l l
ThMS S
l l
y Frequency Categories of Initiating Events INITIATING EVENTS 3
1.
Decrease in Feedwater Enthalpy a) MFW Heater System Failure I
b) AFW Delivery LF1 2.
Excess Feedwater Flow a) MFW Flow Increases Above That for a Particular Fower Level I
b) AFW Flow Increases Above That for a Particular Power Level LF1
-)
Reactor Power Decreases and MFW Flow Fails to Adjust I
J) Reactor Power Decreases and AFW Flow Fails to Adjust LF1 e) Turbine Trip and MFW Fails to Rampbact M + Add. Fail. = LF1 3.
Excess Steam Flow e Steam Flow Increases Above That Req'd for a Particular Power Level a) Steam Line Break LF2 b) MFW Line Break d/s of Check Valve LF2 c) ADY Inadvertently Opens LF1 d) TBV Inadvertently Opens I
e) MSSV Inadvertently Opens LF1 f) Excess Steam Flow Through Turbine I
Reactor Power Decreases and Steam Flow Fails to Adjust e
g) ADY is Open and Fails to Close LF1 h) TBV is Open and Fails to Close I
- 1) Turbine Fails to Decrease Steam Flow I
j) Reactor Trips and Turbine Fails to Trip M + Add. Fail. = LF3 k) Reactor Trips Turbine Trips, and TBV Fails to Close After Quick Open or During Modulation M + Add. Fail. = I l) Hi Pressure Transient, MSSV Opens and Fails to Close I + Add. Fail. = LF1 4.
Large LOCA a) Large Pipe Break LF2 5.
Small LOCA a) Non-Isolable Pipe Break I
b) Isolable Pipe Break (Letdown Line)
LF1 c) RCP Seal Failure I
d) PORY Inadvertently Opens LF1 e
RCS Overpressure Scenario and One PORV/PSV Fails to Reclose I + Add. Fail. = LF1 f
SG Tube Rupture 6.
Pressurizer Pressure Control Failures a) Spurious Main Spray Actuation LF1 b) Spurious Aux. Spray Actuation, c) Pressure Transient Actuates Main Spray, Spray Fails to Decrease M + Add. Fail. = I d) Excess Main Spray During a Controlled Depressurization or Boron Mixing LF1 e) Excess Aux. Spray During a Controlled Depressurization LF1 7.
Inadvertent SIAS When Below Shutoff Head a) Failure to Block SIAS Setpoint I
b) Spurious SIAS I
8.
Decrease in Charging Enthalpy a) PLCS Failure (Max. Charging, Loss of Letdown)
LF1 9.
Maximum Shutdown Cooling LF1 Spurious auxiliary spray actuation is not considered plausible due
+
to complexity of actuation procedure.
l POWER e un,-
. n
l System Status - vs. Plant Operating States l
SYSTEM RELATED PLANT OPERATING STATES SYSTEMS STATE 1
STATE 2
STATE 3 STATE ll TURBINE A
B D
D TURBINE BYPASS C
B A OR B B
ATMOSPHERIC DUMP VALVES E
B B
B CHARGING A
A A
A LETDOWN A
A A
A MAIN PRESSURIZER SPRAYS A
A A
A OR B AUXILIARY PRESSURIZER SPRAYS D
D D
B OR D ENGINEERED SAFETY FEATURES ACTUATION SYSTEM C
C C
C*
C*
C REACTOR REGULATING SYSTEM C INSERT E INSERT E
F l
E WITHDRAWAL B WITHDRAWAL E
F MAIN FEEDWATER A
B D
S MAIN FEEDWATER BYPASS C
B D
B B
B SHUTDOWN COOLING SYSTEM F
F F
B MAIN STEAM ISOLATION VALVES OPEN OPEN OPEN OPEN MAIN FEEDWATER ISOLATION VALVES OPEN OPEN/ CLOSED CLOSED CLOSED GLOSSARY: A OPERATING UNDER AUT0. CONTROL-gg7 y,/g
- PORV B BLOCKED B
OPERATING UNDER MANUAL CONTROL gj j, f/af./
a C - STANDBY - CONTROL / ACTUATION SYSTEM AUTOMATIC gf f _ //, / f/,ygf D
STANDBY - COMPLEX MANUAL PROCEDURE TO ACTUATE g fg/dwa POWER l
E STANDBY - SIMPLE MANUAL ACTION TO ACTUATE SYSTEMS s
F OUT-0F-SERVICE l
l
Initiating Events vs. Plant Operating States PLANT OPERATING STATE INITIATING EVENTS 1
7 3
4 1.
Decrease in Feedwater Enthalpy a) MFW Heater System Failure X
X b) AFW Delivery X
2.
Excess Feedwater Flow a) MFW Flow Increases Above That for a Particular Power Level X
X' b) AFW Flow Increases Above That for a Particular Power Level X
X X
c) Reactor Power Decrcases and MFW Flow Fails to Adjust X
X d) Reactor Power Decreases and AFW Flow Fails to Adjust X
e) Turbine Trip and MFW Fails to Rampback (D
3.
Excess Steam Flow e Steam Flow Increases Above That Reg'd for a Particular Power Level a) Steam Line Break X
QD X
X b) MFW Line Break d/s of Check Valve X
X X
X c) ADV Inadvertently Opens X
X X
d) TBV Inadvertently Opens X
X X
e) MSSV Inadvertently Opens X
X X
X f) Excess Steam Flow Through Turbine X
X e Reactor Power Decreases and Steam Flow Fails to Adjust g) ADV is Open and Fails to Close X
X X
h) TBV is Open and Fails to Close X
X X
i) Turbine Fails to Decrease Steam Flow X
X j) Reactor Trips and Turbine Fails to Trip (S)
X l
k) Reactor Trips, Turbine Trips, and TBV Fails to Close After X
X l
Quick Open or During Modulation
(D X
POWER SYSTEMS
__s
Initiating Events vs. Plant Operating States PLANT OPERATING STATE INITIATING EVENTS 1
7 3
4 4.
Large LOCA a.
Large Pipe Break X
X X
X 5.
Small LOCA a) Non-Isolable Pipe Break X
X X
X b)
Isolable Pipe Break (Letdown Line)
X X
X X
c) RCP Seal Failure X
X X
X d)
PORV Inadvertently Opens X
X X
e) RCS Overpressure Scenario and One PORV/PSV Fails to P.eclose X
X X
X f) SG Tube Rupture X
X X
X 6.
Pressurizer Pressure Control Failures
- a) Spurious Main Spray Actuation X
X X
X b) Spurious Aux. Spray Actuation +
c) Pressure Transient Actuates Main Spray, Spray Fails to Decrease X
X X
X d) Excess Main Spray During a Controlled Depressurization or X
X X
X Boron Mixing e) Excess Aux. Spray During a Controlled Depressurization X
7.
Inadvertent SIAS When Below Shutoff Head a) Failure to Block SIAS Setpoint X
b) Spurious SIAS X
o Spurious auxiliary spray actuation is not considered plausible due to complexity of actuation procedure.
O Pressurizer control system failures which result in over-pressure transients are included in Category 5.
This category refers to RCS de-pressurization events.
p.
POWER
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b{ 'Y Event Tree Branch Descriptions EVENT DESCRIPTION CODE OCCURS DURING MAIN STEAM LINE gREAK OUTSIDE CONTAINMENT, UPSTREAM OF MSIV B.
MS PLANT CONDITION 4.
PORY INADVERTENTLY OPENS DURING PLANT CONDITION 1.
PV ONE A HIGH PRESSURE TRANSIENT ON THE SECONDARY SIDE CAUSES THE MS VS p
d MV MSSV ON SG B FAILS TO RESEAT. OCCURS DURING PLANT CONDITION REACTOR / TURBINE TRIP AND MFW FAILS TO RAMPBACK DURING PLANT C
]
MF THE REACTOR TRIPS AND THE TURBINE FAILS TO TRIP DURING PLANT CON
( _ TT A
THE MSIS IS NOT GENERATED DUE TO AN ACTUATION LOGIC FAILURE.
B MSIV A FAILS TO CLOSE AUTOMATICALLY.
C MSIV B FAILS TO CLOSE AUTOMATICALLY.
THE AFW SYSTEM IS IN MANUAL AND THE OPERATOR FAILS TO ISOLATE AFW FLOW TO D
RUPTURED SG WITHIN > MINUTES AFTER ThE INITIATING EVENT.
THE AFW SYSTEM IS IN MANUAL AND THE OPERATOR FAILS TO TERMINATE AFW FLOW TO E
INTACT SG WITHIN lb MINUTES AFTER THE INITIATING EVENT.
THE OPERATOR FAILS TO THROTTLE HPSI FLOW WITHIN 10 MINUTES AFTER THE INITIATIN F
EVENT.
AFW FLOW IS LOST TO SG A BETWEEN 10 MINUTES AND 1 HOUR AFTER THE INITIATING G
EVENT.
H THE TURBIHE FAILS TO TRIP ON REACTOR TRIP. MSIS GENERATED.
I MFW FAILS TO RAMPBACK ON TURBINE TRIP.
ONE MFW PUMP DISCHARGE VALVE FAILS TO CLOSE AUTOMATICALLY ON HI SG LEVEL.
J K
ONE MFW PUMP FAILS TO TRIP ON HI-HI SG LEVEL.
THE OPERATOR FAILS TO CLOSE THE PORV BLOCK VALVE WITHIN 30 MINUTES AFTER THE L
PORV INADVERTENTLY OPENS.
THE OPERATOR FAILS TO THROTTLE HPSI FLOW WITHIN 30 MINUTES TO AN HOUR AFTER THE M
INITIATING EVENT.
MFW BYPASS FLOW IS LOST BETWEEN 10 MINUTES AND 1 HOUR AFTER THE INITIATING EVENT N
0 FAILURE TO AUTOMATICALLY DELIVER AFW FOLLOWING LOSS OF MFW BYPASS.
P LOSS OF AFW FLOW FOLLOWING SAFETY INJECTION UP TO ONE HOUR AFTER SAFETY INJECTION.
OPERATOR FAILS TO DECREASE AFW FLOW WITHIN 30 MINUTES AFTER THE REACTOR TRIPS.
Q R
OPERATOR FAILS TO ISOLATE THE LETDOWN LINE BREAK.
S MFIV A FAILS TO CLOSE AUTOMATICALLY T
MFIV B FAILS TO CLOSE AUTOMATICALLY U
AFAS A FAILS TO PREVENT AFW DELIVERY TO RUPTURED SG A DUE TO AN ACTUATION LOGIC FAILURE V
AFAS B FAILS TO PREVENT AFW DELIVERY TO RUPTURED SG B DUE TO AN ACTUATION LOGIC FAILURE.
POWER SYSTEMS l
Scenario Occurrence Frequency'Categortration State 1 Events I IN AUG VERY LOW CATEGORY MODERATE INFREQUENT FREQUENCY 7
z MAY OCCUR MAY OCCUR LOW PROBABILITY VERY LOW PROBABILITY EXCEEDINGLY LOW FREQUENCY DURING A DURING A 0F OCCURRING OF OCCURRING PROBABILITY OF DURING A DURING A OCCURRING DURING A CALENDAR YEAR PLANT LIFETIME PLANT LIFETIME PLANT LIFETIME PLANT LIFETIME PV PV-Q PV-MQ PV-P PV-M PV-LQ PV-N MF PV-PV-LM PV-NQ PV-LMQ PV-MN PV-I PV-LN MF-M TT PV-IQ MF-Q TT-Q PV-IM TT-M PV-IMQ
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SEQUENCES MF-J
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SUMMARY
THERE ARE RELATIVELY FEW EVENTS THAT ARE MORE PROBABLE e
THAN THE EVENT ANALYZED.
THOSE HIGHER PROBABILITY EVENTS ARE JUDGED TO BE LESS e
SEVERE THAN THE EVENT ANALYZED.
EVENTS OF SIMILAR PROBABILITY TO THE EVENT ANALYZED e
HAVE SIMILAR CONSEQUENCES.
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ft ADDITIONAL MATERIALS DATA VS.
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1.
HELDS -30 F BETTER THAN PLATES 2.
LOWER BOUND PLATE DATA A
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KIC = 33.2 + 2.800 EXP [.02 (T - RTNDT + 100 F)]
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27 C-E MIXIflG CALCULATIONS t
CEN-189 f1ETHODOLOGY t
MIXING ONLY IN VERTICAL FLOW REGI0flS EPRI/CREARE TESTS PHASE I i
SLOPED PIPE VENT VALVE NO LOOP FLOW PHASE II HOP.IZONTAL PIPE NO VEllT VALVES LOOP FLOW r
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INJECTION N0ZZLE
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COLD LEG MIXING REGIONS
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FALLING JET (PLUttE) - GOOD MIXING 2.
Il0RIZONTAL G,RAVITY CURRENT - P00R tilXING' d
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VERTICAL GRAVITY CURRENT - GOOD MIXING
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Figure G.10.
Schematic of C-E's Mixing Model fj
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TEST NO. 42
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= 2.14 GPM HPI O
= 20 GPM LOOP T100% MIXED = 145.8 F T
= 60 F gpg 2.14 GPM
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20 40 60 80 TIME (sec)
Figure G.3.
Fluid Temperature Distribution for CREARE Mixing Test 42.
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BOTTOM OF OLD LEG 8
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24 PREDICTI0ft l
28 a:
T24 60 80 100 120 140 160 FLUID TEMPERATURE, F Figure G.12.
Comparison of Predicted Average Plume Temperature to CREARE Mixing Data (Test 42).
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- 32..'
FORT CALHOUN EMERGENCY PROCEDURES e
Explicit HPSI and Charging Pump Termination Criteria Achieve 500F Subcooling Establish Minimum Level in the Pressurizer e
Caution Statements Concerning Cooldown Rate Curve e
Instructions for Depressurization if Cooldown Rate Curve Is Violated e
Explicit Instructions for Depressurization and Cooldown Following a SGTR e
Restart of RCP's in a SLB and SGTR Under Consideration
o o
JJ FORT CALHOUN TRAINING G
LECTURES Elements of PTS Probable PTS event initiators Sources of pressure Emergency Procedure Guidance Fort Calhoun System Response Review of overcooling experience Core cooling vs. PTS requirements Effect of RCS head void on pressure control Quiz O
SIMULATOR
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,s FLUENCE REDUCTION l'
I e
NEXT CYCLE Fuel management reduces fluence to critical longitudinal welds by a factor of two j
Peak pin power increases t
Full power capability maintained using CE l
setpoint and safety analysis methodology 1
i e
FUTURE Optimize fuel management to reduce vessel fluence and minimize fuel cycle costs 1
1 Maintain full power capability j
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