ML033510582
ML033510582 | |
Person / Time | |
---|---|
Site: | Catawba |
Issue date: | 05/15/2003 |
From: | Ernstes M Operator Licensing and Human Performance Branch |
To: | Gordon Peterson Duke Energy Corp |
References | |
50-413/03-301, 50-414/03-301 50-413/03-301, 50-414/03-301 | |
Download: ML033510582 (246) | |
See also: IR 05000413/2003301
Text
Draft Submittal
(Pink Paper)
CAPAWBA APRIL 2003 EXAM
50-413 & 50-414/2003-301
MARCH 31 -APRIL 4 &
APRIL 30, 2003
AFT COMBINED RO/SRO WRITTEN
(PART 3 OF 3 )
Bank Question: I000 Answer: C
1 Pt(s) Unit 1 is operating at 100%power. Given the following events and
conditions:
0 CF REG valves drift close
e Reactor trip occurs on S/G Lo-Lo level
e E§-0.1 (Reactor Trip Response) has been entered
CA functions normally
Steam dumps do not operate
e S/G PORVs do not open in AUTO
e S/G pressure in all S/Gs is approximately 1175 psig
e NC system pressure 2345 increasing
0 NC PORVs do not operate
Which one of the following procedures should be implemented?
A. FR-H.1 (Loss OfSecondary Heat Sink)
B. FR-H.4 (Response To Loss OfNormal Beurn Release Capabilities)
C AP-22 (Loss of Instrunlent Air)
D. AP-29 (Loss of Vital or Aux Cfintrol Power)
Distracter Analysis: The loss of VI causes CF REG valves to close, s t e m
dumps to fail to opcn in auto and NC PORVs to fail closed.
A. Incorrect: CA flow functioned normally
Plausible: §/G Lo-Lo reactor trip occurred
B. Incorrect: S/G pressure is iess than 1175 psig
Plausible: steam dumps and S/G PORVs not functioning
C. Correct: spurious component operation is entry condition for AP-22
D. Incorrect: vitat or aux control power not lost
Plausible: components are not functioning
KA: APE 065 G2.4.4 (4.0/4.3)
Lesson Plan Objective: APFAM-22 Obj: 4
Source: New
Level of knowledge: analysis
References:
1. QP-CN-§§-VI page 20
2. AP-22 page 1
3. F-0 page 5
-
.
.
,r
.
DUKE .POWER .. - 111_
CATAWBA
..........,OPERATIONS TRAINING
......... ..... ........._.....
.% .% ~..~
____..%
Objective
State the purpose of AP/OIN5500/22 (Loss of Instrument Air)
Summarize major actions in APIOIN5500122 (Loss of Instrument Air)
Given a set of specific plant conditions and APIOIN5500122 (Loss of
Instrument Air) use the rules of usage and outstanding PPRBs to identify
the correct procedure flowpath.
Determine if entry into APIO/A/5500/22 (Loss of Instrument Air) is
required ushg avaiCable control room indications.
Evaluate and execute each step in AP/O/A/5500/22 (Loss of Instrument
Air) using available control room indications and controls:
0 For all steps contained in the ACTIONEXPECTED RESPONSE
column
For all steps contained in the RESPONSE NOT OBTAINED
column
Evaluate applicability of all NOTES and CAUTIONS in APIOIN5500122
(toss of Instrument Ais) using available control room indications.
Time: 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />
OP-CN-APFAM-22 FOR TRAINING PURPOSES ONLY REV. 08
Page 3 of 7
-_
DUKE POWER
L IIII
CATAWBA OPERATIONS...................................
TRAINING
--
c) Aux Feedwater (CA) System Response to a Loss of Instrument Air
With a boss of Instrument Air, the flow control valves for all SIG's
will no longer fail open. A Mod has been completed O R both
units which added air receiver tanks to the CA Row control valves
with enough air that will allow closure of these valves for 60
minutes after a loss of VL This will also preclude SIG overfill on a
SGTR with a loss of AC Power and a subsequent loss of VI.
d) Transient
1) Air will be lost to:
a Main feedwater control
Steam dump control
S/G PORVs in AUTO
a NC PORVs and PZR sprays
2) Reactor trip will occur on 'S/GLO LO Level due to CF control
valves failed closed.
3) Steam dumps will not operate. S/G PORVs will not open in
'AUTO'. Heat removal will be via S/G safeties.
4) NC system pressure may increase beyond PZR PORV setpoint.
N2 may have to be aligned to PZR PORVs.
2.2 Station Air System (VS)
A. System Purpose (Obj. #13,14)
1. Supplies ais to miscellaneous tools and equipment throughout the piant.
2. Backup air supply for the Instrument Air system.
3. Interconnections with other systems.
a) Recirculated cooling water (KR) cools the compressor
1) Eftluent cooling water temperature controls the KR flow rate
2) Inlet isolation valves closes automatically when the associated
compressor is shutdown.
b) Low Pressure Setvice Water (RL) cools the after coolers
c) Instrument Air (Obj. #8, 17)
1) Backup supply for VI
2) IVS-78 will automatically open if VI1 pressure drops to 76 psig.
3) Via two oil removal filters
d) Logic from the Engineered Safety Features Actuation System doses
the containment isolation valves upon receiving a Phase A (S,)
containment isolation signal. (VS-54B) (Obj. #17)
OP-CN-SS- W FOR TRAINING PURPOSES ONLY REV. 26
Page 20 of36
LOSS OF INSTRUMENT AIR
API0IN5500122
A. PurrJose
To provide guidance for responding to a loss of instrument air.
IAB-8, A13 "VI COMPRESSOR D TROUBLE" - LIT
IAD-8, 813 "VI COMPRESSOR E TROUBLE LET -
1AD-8, C13 "VI COMPRESSOR F TROUBLE" - LIT
1AD-8, D13 "VI DRYER E TROUBLE" - LIT
1AD-8, E13 "VI DRYER F TROUBLE" - LIT
1AD-8, Fi3 "VI DRYER BYPASS VbV OPEN'a- LIT
VI pressure - DECREASING
IAD-8, A!? "VI LO P R E S S - LIT
-
Plant instrumentation or control ERRATIC.
CRITICAL SAFETY FUNCTLON STATUS TREES
EP/I/N5000/F-O Heat Sink - Page 1 of 1
TOTAL PEEDWATZR
FLOW TO Si(;*
GREATER THAN
450 GPM
SIR LEVELLY AT
LEAST ONE $10
GRFATtR THAh
I
0
0
.~~~ -0 ,
PREssiiw: IN ALL
S G s LESS THAN
1230PSiG
i:
i t 0
1 ~
NIR LEVEL IN ALL
SiGs LRSS THAN
R3%
~~~~~
NIR LEVEL IN A
SlGs OREATCR T
Bank Question: 989 Answer: C
1 Pt(s) Unit 1 was operating at 100%power when the foliowing containment floor
and equipment sump indications were note$:
Time sum^ A level indication sum^ B level indication
8200 4.1 5.9
0215 4.2 6.0
0230 4.3 6.1
0245 4.4 6.2
0300 6.5 6.9
A chemistry sample f-roni the sumps shows the presence of boric acid in the
water.
If the OAC is unavailable, which one of the following statements correctly
describes the required actions by Tech Specs?
REFEMhTCESPRQlTDED: Tech Spec 3.4.13 and PT/l/M4600/009
A. Shutdown to mode 3 required no later than 0815
5. Shutdown to mode 3 required no later than 0900
C. Shutdown to mode 3 required no later than 1300
D. Operations may continue indefinitely while conducting
containment entry to determine source of Leak and to reduce
leakage rate.
Dirtracter Analysis: The foliowing reflects PT/l/A/4600/009 Encl 13.4 and
13.5 for leak rate determination.
Time Level Volume Level Volume Leakrate
200 4.1 131.9 5.9 227.6 (gpm)
21 5 4.2 137.3 6 232.8 0.7'1
230 4.3 142.6 6.1 238.1 0.91
245 4.4 148 6.2 243.3 0.71
300 6.5 259.1 6.3 280.1 9.86
First - must determine if the leakage at 0300 (9.86 gpm) is cmnsidered
unidentified leakage or pressure boundary leakage. The proper classification is
unidentified leakage because there is no indication that the leak is corning from a
pressure boundary.
A. Incorrect: Must be in mode 3 by 1300 per action A and B.
Plausible: Ifthe candidate thinks that the lea!age is pressure
boundary leakage.
B. Incorrect: Must be in mode 3 by 1300 per action A and ID.
Plausible: math error in Ieak rate calculation such as failing to divide
leak rate by 15 minutes or if candidate considers leakage to be
I
C. Correct: The leakage rate at 0300 is 9.8 gpm > 1 gpm unidentified
leakage - Tech Spec 3.4.13 action A and B required - 4+6=10 hours
to reach mode 3.
D. Incorrect: Must be in mode 3 by 1300 per action A and B.
Plausible: Easy to make a math error such as not dividing by 15
minutes - or if candidate confuses identified leakage spec (10 gpm)
with unidentified leakage spec (1 gprn).
Level: SRQ Only 10CFR55.43(b)5
KA: EPE009AA2.28 (2.813.1)
Lesson Plan Objective: none
Source: New
Level of Knowledge: analysis
References:
2. PT/l/A/4600/009
3.4.13
3.4 REACTOR COOLANT SYSTEM (RCS)
4.4.23 RCS Operational LEAKAGE
LCO 3.4.43 RCS operational LEAKAGE shall be limited to:
a. No pressure boundary LEAKAGE;
b. 1 gpm unidentified LEAKAGE.;
c. 10 gpm identified LEAKAGE;
d. 576 gallons per day total primary to secondary LEAKAGE through all
steam generators (SGs); and
e. 150 gallons per day primary to secondary LEAKAGE through any
one SG.
APPLICABILITY: MODES 2 , 2,3, and 4.
CONDITION COMPLETION TIME
A. RCS LEAKAGE not A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
within limits for reasons within limits.
other than pressure
boundary LEAKAGE.
B. Required Action and B.l Be In MODE 3.
associated Completion
Time of Condition A not AND
met.
8.2 Be in MODE 5.
-
Pressure boundary
LEAKAGE exists.
~
Catawba Units I' and 2 3.4.13-1 Amendment Nos. 1731165
3.4.13
SURVEILLANCE REQUIREMENTS
Verify RCS Operational LEAKAGE within limits by 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
performance of RCS water inventory balance.
SR 3.4.13.2 Verify steam generator tube integrity is in accordance In accordance with
with the Steam Generator Tube Surveillance Program.
Generator Tube
Catawba Units f and 2 3.4.$3-2 Amendment Nos. 173/%65
Duke Power Company 'rocedurc No.
Catawba Nuclear Station PT/1/Af460O/OU9
levision No.
067
Loss of Operator Aid Computer
~
Electronic Reference No.
- * * * * * * * * * UNCONTRQLLEDFORPRINT * * * * * * * * * *
(ISSUED) - PDF Format
PT/~/A/~GCIO/OW
Page 2 of 8
Loss of Operator Aid Computer
1. Purpose
To document Technical Specifications requirements normally perfonned by the Operator Aid
Computer in the event that the unit Operator Aid Computer is out of service.
2. Reference
2.1 OP/1/N6700/003 (Operation with the Operator Aid Computer Out of Service)
2.2 Catawba TS and SLC Requirements:
2.2.1 TS3.1.4
2.2.2 TS 3.1.6
2.2.3 TS 3.2.3
2.2.4 TS 3.2.4
2.2.5 TS 3.3.1
2.2.6 TS 3.4.2
2.2.7 TS 3.4.5
2.2.8 TS 3.4.13
2.2.9 TS 3.4.15
2.2.10 SR3.1.4.1
2.2.11 SR 3.1.6.2
2.2.12 SR 3.2.3.1
2.2.13 SR 3.2.4.1
2.2.14 SR3.4.2.1
2.2.15 SLC 16.5-7
2.2.16 TS 3.6.3
2.2.17 TS 3.1.3
PT/!'1/A/4600/009
Page 3 of 8
3. Time Required
3.1 Manpower - One Operator
3.2 Time - Until the Operator Aid Computer is restored to service.
3.3 Frequency When the Operator Aid Computer is out of service.
4. Prerequisite Tests
None
'5. Test Equipment
None
6. Limits and Precautions
6.1 If an acceptance criteria is NOT met, the Operations Shift Manager and the Operator at
the Controls should be notified immediately.
6.2 If the unit status or system condition prevents the performance of a surveillance item, the
item should be noted on the affected data sheet with an explanation and the Operations
Shift Manager and the Operator at the Controls should be notified immediately.
7. Required Unit Status
None
8. Prerequisite System Condition
__- Verify the Operator Aid Computer is out of service.
9. Test Method
A visual inspection of various system instrumentation wiil be made until the computer is
returned to service.
10. Data Required
Complete Enclosures as required.
11. Acceptance Criteria
No data taken shall exceed limits listed on the Enclosures.
PT/l/A/4600/009
Page 4 of 8
12. Procedure
12.1 IF in Modes 5
_. 6, EVERY 15 MINUTES document the critical core parameters listed
on Enclosure 13.1 (Critical Core Parameters Sheet) (Reference OEP).
12.2 -IF Start Up OfND System During Plant Cooldown (OP/I/A/6200/004) is in progress
~AND KCHX Maximized Cooling Temperature Monitoring is being performed, within
15 minutes and every 15 minutes thereafter record parameters on Enclosure 13.2 (KCHX
Maximized Cooling Temperature Monitoring).
12.3 EVERY 15 MINUTES record on Enclosure 13.3 (Auxiliary Building Ventilation Supply
Unit Status) the status of the Auxiliary Building Ventilation System supply units.
12.4 IF in Modes 1-4, within 30 minutes of Loss of OAC and once per hour thereafter, verify
and record on Enclosure 13.4 (Ventilation Unit Condensate Drain Tank Input Rate
Determination) that the rate of increase in VUCDT level is 1% per hour. (TS 3.4.13
and 3.4.15)
12.5 -
IF in Modes 1-4, within 30 minutes of Loss ofOAC, begin performing Enclosure 13.5
(Conbinment Floor and Equipment Sumps Input Rate Determination) to verify input to
the Containment Floor and Equipment Sump is less than 1 gpm. (TS 3.4.13 and 3.4.15)
12.6 -
IF in Modes 1-4, within 30 minutes of Loss of OAC and once per hour thereafter, verify
and record on Enclosure 13.6 (IEMF-38 Delta Count Rate Determination) that the
change in count rate on 1EMF-38 is < 750 cpm in one hour. (TS 3.4.13 and 3.4.15)
12.7 -
IF in Modes 1-4, within 30 minutes of Loss of QAC and once per hour thereafter, verify
and record on Enclosure 13.4 (1EMF-39 Delta Count Rate Determination) that the
change in count rate on IEMF-39 is < 6700 cpm in one hour. (TS 3.4.13 and 3.4.15)
12.8 -
IF ALL the following conditions exist (Reference SK 3.4.2.1):
Reactor Critical
T~~(i<561'F
T ~ - TAU^
F H i h Alarm Present, Annunciator 1AD2 A/4
EVERY 30 MbWTES verify Reactor Coolant loops TAVG ? 55 1"F by completing
Enclosure 13.8 (TAVC; Data Sheet).
NOTE: The Y C Operable But Degraded Condition is normally active during the winter months
based on Lake Wylie and SNSWP temperatures.
12.9 the YC Operable But Degraded Condition is active, perform Enclosure 13.9 (YC
Operable But Degraded Temperature Monitoring).
PT/1iAi4600/009
Page 5 of 8
12.10 E both trains of the plasma display monitor are inoperable in Modes 1-6, EVERY 60
MINUTES or after 10% change in power, complete Enclosure 13.10 (Subcooling Data
Sheet) to monitor subcooling margin.
12.1I Unit 1 net generation CANNOT be obtained from the Unit 1 operator aid computer,
perform the following:
12.1 1.1 At the top of the first hour during loss of OAC, notify SOC that they will not
be getting station or unit MWH hourly values from both CNS units.
Person notified
12.11.2 EVERY HOUR on the HOUR complete Enclosure 13.1 1 (Eiectrical Data
Sheet).
NOTE: I. If pressure (primary and secondary) are verified < 200 psig, then temperatures are
-
NOT required to be taken nor recorded.
2. Use a calibrated pyrometer to obtain SiG shell temperatures.
12.12 NC Tc is > 80°F a NC pump is operating, then the secondary side temperature
is > 80°F and documentation of shell temps is NOT necessary. E in Modes 5 , 6 No
Mode, EVERY 60 MINUTES complete Enclosure 13.12 (Steam Generator Data Sheet)
(Reference SLC 16.5-7).
12.I3 in Mode I less than 50% rated power, prior to exceeding 50% rated power and
every 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> thereafter, with the AFD monitor alarm inoperable, monitor and log the
indicated Axial Flux Difference for each operable excore channel on Enclosure 13.13
(Axial Flux Difference (Yd Flux) Following Loss of AFD Monitor A l m ) . (Reference
SR 3.2.3.1 and TS 3.2.3).
12.14 E in Mode 1 AND 2 50% rated power, once within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> thereafter
with the AFD monitor alarm inoperable, monitor and log the indicated Axial Flux
Difference for cach operable excore channel on Enclosure 14.13 (Axial Flux Difference
(%A Flux) Following Loss of AFD Monitor Alarm). (Reference SR 3.2.3.1 and TS
3.2.3).
12.15 E in Modes I 2, EVERY 4 HOURS verify by signing off on Enclosure 13.14 (Rod
Verification Checklist) that the Digital Rod Position indication for all rods are within
f 12 steps of their group step counter denland position and operable (Reference SK
3.1.4.1).
12.16 in Mode 1 2 E D KEFF2 1.0, EVERY 4 HOURS verify and record on Enclosure
13.15 (Rod Insertion Limit Checksheet) that each coetrol bank of rods is above the rod
insertion limit (Reference SR 3.1.6.2).
PT/1/A/4600/009
Page 6 of 8
12.17 E in Modes 1,2,3, Mode 4, when steam generators are being used for heat removal,
EVERY 4 HOURS record CA suction source temperatures measured locally using a
calibrated Keithley 872 digital thermometer, Type J or its equivalent, as required, per
Enclosure 13.16 (CA Suction Source Temperature Monitoring Data)
12.18 E in Modes 1-4, within 4 HOURS and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter, monitor the CF
containment isolation valves N2 accumulator pressures on Enclosure 13.17 (CF
Containment Isolation Valve N2 Accumulator Pressure Monitoring).
12.19 in Modes 1-4, EVERY 6 HOURS, document data needed for primary to secondary
leakage calculation on Enclosure 13.18 (Primary to Secondary Leakage Calculation
Data) and provide data to Chemistry. Notify Secondary Chemistry to perform
PT/l/B/4600/028 (Determination Of Steam Generator Tube Leak Rate For Unit 1).
12.20 E Auxiliary Spray is being used for pressurizer pressure control, EVERY 12 HOURS
complete Enclosure 13.19 (Pressurizer Spray AT Data Sheet).
12.21 E in Mode 1 AND above 50% ratedpower, once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
thereafter, document Quadrant Power Tilt Ratio, as calculated by PT/Q/A/4600/08B
(Man. Cal. of Quad. Tilt), in Enclosure 13.1 of PT/l/A/4600/002A (Mode 1 Periodic
Surveillance Items). (Reference SR 3.2.4.1)
12.22 in Modes 1-3, within 12 HOURS of the Loss of OAC and every 12 ~ O U F Sthereafter,
monitor the CA piping surface temperatures. Perform OP!l/A/6250/002, Enclosure 4.12
(Checking Pipe Surface Temperatures).
12.23 E in Modes 1-2, within 12 HOURS of the Loss of OAC and every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thercafter,
monitor the Overteniperature Delta T parameters and record on Enclosure 13.20
(Overtemperature Delta T Setpoint Channel Check). (Reference SK 3.3.1.1)
12.24 in Modes 1-4, EVERY 24 HOURS perform a manual leakage calculation of the NC
System in accordance with PT/I/A/4150/0011 (NC Manual Leakage Calculation).
(Reference TS 3.4.15, Required Action A. 1).
12.25 Update Enclosure 13.21 (Chemistry Data Sheet) as information becomes available from
Chemistry.
12.26 WHEN the OAC is returned tu service, notify Shift Work Manager to coordinate with
Lccal IT and Reactor Group Duty Engineer to ensure OAC is updating properly.
12.26.1 Notify SOC that MWM data should be valid at the top of the next hour.
Person notified
12.26.2 Give a copy of Enclosure 13.1 1 to the SSA to assist them in editing the switch
board logs.
PT/1/A/4600/009
Page 7 Of 8
12.27 Evaluate the acceptance criteria by performing one of the following:
_ I 12.27.1 Verify the acceptance criteria specified in Section 11 is met.
~ 12.27.2 @the acceptance criteria is met, perform the following:
CS Notify the Unit/WCC SRO that the acceptance criteria is NOT met.
--J.--.
Unit/WCC SRO Contacted Date Time
61 Initiate a PIP to document the test failure.
0 Document all issues on a procedure discrepancy sheet.
12.28 E any discrepancy is noted during the performance of this test that does NOT keep the
test from meeting the acceptance criteria, it shall be given to the Unit/WCC SRO for
evaluation via a discrepancy sheet.
- 12.29 Submit PT/I/A/4600/009 (Loss of Operator Aid Computer) to the Unit'WCC SRO.
13. Enclosures
13.1 Critical Core Parameters Sheet
13.2 KCHX Maximized Cooling Temperature Monitoring
13.3 Auxiliary Building Ventilation Supply Unit Status
13.4 Ventilation Unit Condensate Drain Tank Input Rate Determination
13.5 Containment Floor and Equipment Sumps Input Rate Determination
13.6 1EMF-38 Delta Count Rate Determination
13.7 IEMP-39 Delta Count Rate Determination
13.8 TAFrG Data Sheet
13.4 YC Operable But Degraded Temperature Monitorhg
PT/1/A/4600:009
Page 8 of 8
13.10 Subcooling Data Sheet
13.1 1 Electrical Data Sheet
13.12 Steam Generator Data Sheet
13.13 Axial Flux Difference (%A Flux) Following Loss of AFD Monitor Alarm
13.14 Rod Verification Checklist
13.15 Rod Insertion Limit Checksheet
13.16 CA Suction Source Temperature Monitoring Data
13.14 CF Containment Isolation Valve N2 Accumulator Pressure Monitoring
13.18 Primaq to Secondary Leakage Calculation Data
13. I9 Pressurizer Spray AT Data Sheet
13.20 Overtemperature Deita T Setpoint Channel Check
13.21 Chemistry Data Sheet
Enclosure 13.1 PT/llA/4600/009
Critical Core Parameters Sheet Page 1 of 1
NOTE: Make additional copies of this sheet as necessary.
Enclosure 13.2 PT/1/A/4hOO/009
KCHX Maximized Cooling Temperature Page 1 of 1
Monitoring
to be > 4YF. KCHX outlet piping temperature measurements are required only when RN
Essential header temperature is less than 4 S T . A pyrometer is required to obtain these
readings.
2. Ifany ofthe following occur, terminate maximized cooling to the KCHXs per
(OP/ 1/A/6200/004)
VCT Outlet temperature decreases to 65°F.
Enclosure 13.3 PTll/Ai4600!009
Auxiliary Building Ventilation Supply Unit Page 1 of 1
Status
for Control Room D/P verification and OP/O/A/6450lOO3
Enclosure 13.4
PT!1 /A/4600/009
Ventilation Unit Condensate Drain Tank Page 1 of I
Input Rate Determination
Acceptance Criteria - Rate of increase must be l%ihour.
NOTE: 1. Either of the following instruments may be used to perform this surveillance, however,
the same instrument should be used for the duration of time the procedure is in effect:
"UNIT 1 VUCDT LEVEL" on Auxiliary Waste Processing Control Panel
(1ELCC0013) (AB-543, MM, 53-54) .
- 1WLP5771 (AB-543, BB-CC, 50) CA Pump Room (Next to WCDT)
2. If the rate of increase is L l%/hour, the W C D T input rate is 1 gpm. Refer to TS 3.4.13
and TS 3.4.15 and determine if NC System leakage is > 1 gpm.
3. Coordinate with Radwaste Chemistry as required when the VUCDT needs to be pumped
I I I
Enclosure 13.5 PTil/A;4600/009
Containment Floor and Equipment Sumps Page 1 of 4
Input Rate Determination
1. Procedure
I I
NOTE: If any containment floor and equipment sump pump starts during the 10 minute test period
the test results will NOT be valid. The test should be repeated until valid results are
- 1.1 Stop the following sump pumps and place in "Manual":
"Pump 1Al Cont Floor & Equip Sump"
- "Pump 1A2 Cont Floor & Equip Sump"
"Pump 1Bl Cont Floor & Equip Sump"
"Pump 1B2 Cont Floor & Equip Sump"
"Manual" position.
1.2 at any time during the performance ofthis test the sump level reaches 2 15", perform the
following:
1.2.1 Place the following sump pumps in "AUTO":
"Pump 1Al Cont Floor & Equip Sump"
"Pump 1A2 Cont Floor & Equip Sump"
"Pump 1B1 Cont Floor & Equip Sunip"
"Pump 1B2 Cont Floor & Equip Sump"
I
NOTE: A level less than 4" is below the calibration range of the Containment Floor and Equipment
Sump level instrumentation, therefore the Leakage Detection Systems must be declared
1.2.2 Verify the affected sump level is lowered to 10" as indicated on 1WLP5740 (Cont
Floor and Equipment Sump A Level) or lWLP5750 (Cont Floor and Equipment
Sump B Level).
1.2.3 Return the following sump pumps to "Manual" and stopped
"Pump IAl Cont Floor & Equip Sump"
"Pump 1A2 Cont Floor & Equip Sump"
"Pump 1B1 Cont Floor & Equip Sump"
"Pump 1B2 Cont Floor I% Equip Sump"
Enclosure 13.5 PT/1/M4600/009
Containment Floor and Equipment Sumps Page 2 of 4
Input Rate Determination
- 1.3 Record initial sump readings on the "Containment Floor and Equipment Sump Inleakage
Rate Log Sheet".
1.4 Once per hour, record sump level readings on the "Containment Floor and Equipment Sump
Inleakage Rate Log Sheet".
1.5 Calculate the leakage rate using the "Sump Volume vs. Level Indication Table".
1.6 Verify leakage is < 1 gpm.
1.7 the input to the Containment Floor and Equipment Sumps is > 1 gpm, perform the
following:
RefertoTS3.4.13 andTS3.4.15.
e Determine if NC System leakage is > 1 gpm.
~ 1.8 WHEN the OAC is returned to service, place the following sump pumps in "AUTO":
"Pump I A1 Cant Floor L?L Equip Sump"
- "Pump 1A2 Cont Fioor & Equip Sump"
- "Pump 1B1 Cont Floor & Equip Sump"
- "Pump 1B2 Cont Floor & Equip Sump"
Enclosure 13.5 PT/1/A/4600/009
Containment Floor and Equipment Simps Page 3 of 4
Input Rate Determination
A Containment Floor and Equipment Sump Inleakage Rate Log Sheet
Enclosure 13.5 PT/1/A/4600/009
Containment Floor and Equipment Sumps Page 4 of 4
Input Rate Determination
NOTE: 1. TI is the data from the previous reading.
Enclosure 13.6
PT/1/A/4600/009
1EMP-38 Delta Count Rate Determination Page 1 of 1
I I I
Acceptance Criteria - Change in count rate < 750 cpm per hour.
NOTE: 1. If the change in count rate per hour is 2 750 cpm, refer to TS 3.4.13 and TS 3.4.15 and
determine if NC System leakage is > 1 gpm.
Enclosure 13.7
PT/1/A/46OO/O09
1EMF-39 Delta Count Rate Determination Page 1 of 1
Ledcage Acceptable
Counts Rate/Hour
Acceptance Criteria - Change in count rate < 6700 cpm per hour.
Enclosure 13.8
PT/hM4600/009
TAW;Data Sheet Page 1 of2
ACCEPTANCE CRITERIA: With the Reactor Critical and TA\,G< 561°F and TWF- TAU^^
W i h a l m present, verify NC System loops TAVG> 551°F. (SR 3.4.2.1)
Loop TAVG
Enclosure 13.8 PT/I /A/4600/009
TAVG Data Sheet Page 2 of 2
Enclosure 13.9 PT/1 /Ai4600/009
YC Operable But Degraded Temperature Page 1 of 2
Monitoring
1. Procedure
- 1.i Record the temperature limits as documented on the NSD 203 Operability Notification
Form for the YC Operable But Degraded Condition.
SNSWP RN Essential Header
Hi Temperature Limit
%-Hi Temperature Limit
1.2 Record the required temperatures on the the "YC Operabie But Degraded Temperature
Monitoring" data sheet.
1.a Determine the monitoring frequency as follows:
1.3.1 any temperature is within 5 O F of either Hi Limit, the temperatures are to be
recorded every hour.
1.32 all of the temperatures are > 5 OF below both Hi Limits, the temperatures are
to be recorded every four hours.
1.4 -
IF any ofthe temperatures exceed either Hi Limit determined in step 1.1, contact the YC
System Engineer for reanalysis of the YC Operable But Degraded Condition.
1.5 -
IF any of the temperatures exceed either Hi-Mi Limit determined in step I. I , the YC
chillers should be considered inoperable unless the YC Operable But Degraded
Conditions are reanalyzed.
1.6 Repeat steps 1.2 through 1.5 at the required frequency determined in step 1.3
Enclosure 13.9 PT/ 1/A/46OO/OO9
YC Operable But Degraded Temperature Page 2 o f 2
Mnnitnring
(1) IfORNP8130 (SNSWP Temp) gauge is inoperable, SNSWP temperature reading may be obtained
per PT/O/A/4400/024 (SNSWP Temperature Monitoring).
Enclosure 13.10
PT/1/A/4600/009
Subcooling Data Sheet Page I of 1
CALCULATION SHEET FOR NC SYSTEM DEGREES SUBCOOLED
ACCEPTANCE CRITERIA:
Subcool limit is IO'F while at power.
30'F while shutdown.
INFORMATION:
NC Pressure - Record lowest indicated system pressure.
-
T-SAT Using NC pressure, determine saturation temperature from the Unit One Revised Data
Book Figure 57 or Figure 58.
Nighest NC Temp Determine the highest NC Temp:
- In Modes I and 2, use Loop Tmr.
kn Modes 4-6:
Compare the average of the 5 highest reading operable core exit
T/CSto Loop Trior.
- Use the operating train(s) of ND inlet temperature, Loop T H ~ T
and/or the operable core exit T/Cs.
"F Subcooled - Calculate by subtracting "HIGHEST NC TEMP" from "F-SAT".
Enclosure 13.12
Steam Generator Data Sheet Page 1 o f 1
ACCEPTANCE CRITERLA: If S/G shell temperature is < 80"F, SIG pfirnary or secondary pressure must NOT exceed 200 PSIG.
I NOTE: If pressure (primary and secondary) are verified < 200 pig, then temperatures are NOT required to be taken nor recorded I
Enclosure 13.13 PTll/A/4600/009
Axial Flux Difference (YOA Flux) Following Page 1 of 1
Loss of AFD Monitor Alarm
( I ) Record the current Axial Flux Difference limit.
ACCEPTANCE CRITERIA Each channel within the listed limit. Reference the COLR for each
channels limit
Enclosure 13.14 PT/1/k/46OO/OO9
Rod Verification Checklist Page I of 1
ACCEPTANCE CRITERIA - Ali rods are positioned k 12 steps uftheir step demand counters and
E ~ c ~ Q 13.15
su~~
Rod Insertion Limit Checksheet Page 1 of 1
Enclosure 13.16 PT/ 1/A/4600/009
CA Suction Source Teniperature Monitoring Page 1 of 4
Data
I. Procedure
1.1 Every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, record the following temperatures on the table of this enclosure:
- UST IA at 1CSTT5990 (TB-640, ID-30)
- UST 1B at 1CSTT6000 (TB-640,1D-29)
Hotwell Femperture at 1CMTT7260 (TB-577, 15-26)
1.2 Every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, measure CST temperature at 1CSTX5020 (TB-573, IC-33) by using a
Caiibrated Kiethly 872 Digital Thermometer, Model "J" or equivalent, and record on the
table of this enclosure.
1.3 -
IF UST temperature is 2 134 OF,
1.3.1 Notify CRSRO
1.3.2 Verify 1CM-127 (CM-CF Cleanup Flow Ctrl) is:
A. Isolated
B. E in Modes 3 OR 4 1CM-I27 is open, immediately close
1CM- 127.
1.3.3 Ensure Aux Steam is isolated to the UST by verifying the following valves
closed
1AS- 15 (AS to UST BTR Isol) (TB 609, 1 C-29)
m IAS-22 (AS to 1A USF HTR Ctri Inlet) (TB-640, 1C-30)
IAS-25 (AS to 1A UST HTR Ctrl BYP) (TB-640,1C-30)
1.3.4 Verify 1CM-33 (Hotwell High Levei Control) (TB-581, 1 6 2 5 ) is isolated and
-
NOT leaking past seat by checking downstream pipe temperature 5 120 OF.
1.3.5 Notify MSE.
1.3.6 temperature of 1A OR 1B UST is 2 I36'F, declare CA System inoperable
per TS 3.7.5.
Enclosure 13.16
PT/1/A/4600/009
CA Suction Source Temperature Monitoring Page 2 Of4
Data
1.4 E CST is 2 134 "F
1.4.1 Notify CRSRO
1.4.2 -
IF UST is overflowing, as determined by UST level 2 130%, measure
overflow pipe temperature (TB-594, IC/ID-29) using Calibrated Kiethly 872
Digital Thermometer, Model "J" or equivalent and record on table of this
enclosure.
1.4.3 -
IF overflow pipe temperature 2 134 OF, refer to step 1.3.
1.4.4 Verify Manual Loader for Condensate Storage Tank (IASML0170, CST
Temp Ctrl) (TB 573, 1E-32) is set per OP/O/Bi6250/007A (Auxiliary Steam
System Alignment).
1.4.5 -
IF Aux Steam Manual Loader for CST is malfunctioning, then:
NotifyMSE.
1.4.6 -
IF temperature of CST is 5 136"F, secure both CST pumps and
A. UST is overflowing to the CST as indicated by UST level 2 130%,
measure temp on UST overflow line (TB-594, IWD-29).
B. E temp is 5 136'F, declare CA System inoperable.
C. E the UST is NOT overflowing to the CST as indicated by UST level
S 130%, measure UST temperature at 1CSTT5990 (TB-640, 1I)-30)
and lCSTT6000 (TB-640, ID-29). E temp is 2 136 "F, declare CA
system inoperable.
NOTE: Step 1.5 is only applicable if ICM-33 is open in Modes 2 and 3 and Mode 4 when SiGs
1.5 E Hotwell pump discharge temperahire reaches > 136°F Main Condenser Vacuum
decreases to <27" Hg vac, immediately close 1CM-33 (Hotwell High Level Control).
Enclosure 13.16 PT/l/A/4600/009
CA Suction Source Temperature Monitoring Page 3 of4
Data
NOTE: ICM-33 is required to be isolated when CA is aligned for Standby Readiness. Refer to
OP/l/A/6250/002 (Auxiliary Feedwater System).
1.6 Hotwell teniperature reaches 2 134"F, then:
1.6.1 E ICM-33 is isolated, then no further action is required.
1.6.2 E Unit 1 S/Gs are NOT relied upon for heat removal, then no hrther action is
required.
1.6.3 E Unit 1 is in modes 1,2,3, 4 with CA required operable ICM-33 is
-
NOT isolated, then isolate ICM-33.
1.I Rotweli temperature reaches 2 136'F, then:
1.7.1 1CM-33 is isolated, then no further action is required.
1.7.2 E Unit 1 S/Gs are relied upon for heat removal, then no further action is
required.
1.7.3 E Unit 1 is in modes 1,2,3, 4 with CA required operable 1CM-33 is
-
NOT isolated, then declare auxiliary feedwater system inoperable and apply
I---
Enclosure 13.17 PT/1/A/4600/009
CF Containment Isolation Valve N2 Page 1 o f 4
Accumulator Pressure Monitoring
1. Procedure
1.1 Every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, monitor the CF isolation valves N2 accumulator pressures on the table of this
enclosure for the following:
1CP-33 (1A S/G Feedwater Containment Isol), 1CFPT5710, (DH, 597, EE-43)
- ICF-42 (1B S/G Feedwater Containment Isol), ICFPT5420, (DH, 597, EE-52)
1CF-60 (ID S/G Feedwater Containment Isol), 1CFPT5740, (SH, 597, EE-44)
1.2 N2 pressure is c: I285 psig for any ofthe CF containment isolation valves, then perform
the following:
1.2.1 Notify the Control Room SRO that the applicable CF containment isolation valve
is inoperable per Technical Specifications 3.6.3 and 3.7.3.
1.2.2 Notify the Shift Work Manager
1.2.3 Issue the following Model W/O for the applicable CF containment isolation valve
for WE to recharge the N2 accumulator:
1CF-33, Model W/O #91003804
- ICF-42, Model W/O #91003805
1CF-51, Model W/O #91003806
- 1CF-60, Model W/O #91003807
Enclosure 13.17 PT/ 1/A/46OO/OO9
CF Containment Isolation Valve N2 Page 2 of 4
Accumulator Pressure Monitoring
1.3 @ N2 pressure is 2 1285 psig < 1625 psig for any of the CF containment isolation
valve, then perfom the following:
1.3.1 Notify the Cont~olRoom SRO.
1.3.2 Notify the Shift Work Manager
1.3.3 Issue the following Model WiO for the applicable CF containment isolation valve
for IAE to recharge the N2 accumulator:
- 1CF-33, Model W/Q #91003804
ICF-42, Model W/O #91003805
- 1CF-51; Model W/O #91003806
ICF-60, Model W/O #91003807
1.4 @ N2 pressure is > 2425 psig AND 5 2415 psig, then perfom the following:
e Notify the Control Room SRO.
failure due to overpressurization.
Dispatch operator to the applicable doghouse to investigate if a local steam leak in the
area could be the cause of the high pressure.
Notify the Shift Work Manager that the applicable CF N2 accumulator needs to be
vented.
Enclosure 13.17 $Ti I lN46OOM)OY
CF Containment Isolation Valve N2 Fage 3 of 4
Accumulator Pressure Monitoring
1.5 N2 pressure is > 2445 psig, then perform the following:
I NOTE: The CF containment isolation valve is inoperable when N2 pressure is > 2475 psig. I
Notify the Control Room SRO that the applicable CF containment isolation valve is
inoperable per Technical Specification 3.6.3 and 3.7.3.
NOTE: Due to system design limits, there is no immediate concern with catastrophic equipment
failure due to overpressurization. I
- Dispatch operator to the applicable doghouse to investigate i f a local steam leak in the
area o f the applicable CF
- Notify the Shift Work Manager that the applicable CF isolation valve N2 accumulator
needs to be vented.
Enclosure 13.17 PT/1!N4600/009
CF Containment Isolation Valve 112 Page 4 o f 4
Accumulator Pressure Monitoring
NOTE: The normal operating range for the CF containment isolation valves N2 accumulator pressure
is approximately 2200-2300 psig.
_______
(1) With N2 pressure < 1285 p i g , the CF containment isolation valve is inoperable.
(2) With N2 presume > 2475 psig, the CF containment isolation valve is inoperable.
Enclosure 13.18
Primary to Secondary Leakage Calculation Page 1 of 1
Data
Enclosure 13.19 PT/l/N4600/009
Pressurizer Spray AT Data Sheet Page 1 of 1
ACCEPTANCE CRITERIA - Differential temperature between the pressurizer and auxiliary spray
water must be < 260'F.
NOTE: When using Residual Heat Removal Pump 1A (ND Pump 1A), use Residual Heat
Removal Heat Exchanger "A" Inlet Temperature on 1MC7. When using Residual Heat
Removal Pump 1B (ND Pump lB), use Residual Heat Removal Heat Exchanger "B" Inlet
temperature on IMC7. When using Chemical Volume and Control System, use
Regenerative Heat Exchanger charging temperature on 1MC5.
Enclosure 13.20
Overtemperature Delta T Setpoint Channel Page 1 of 1
Check
1. Procedure
1.1 Gauges to be used:
NCkoopA 1NCP5422
- NCLoopB lNCP5462
- NCLoopC 1NCP5502
0 N C h o p D 1NCP5542
NOTE: At lower power levels (<;. 85%), the instrumentation will be overranged (> 150%). Ifthe
instrumentation is overranged, use a value of 150% when performing the Channel Check.
1.2 Verify that the difference between the highest and lowest reading loop is less than or equal
to 10% and record on table below.
1.3 the difference is > IOOh, notify Reactor Group Duty Engineer to perform a qualitative
assessment of channels to determine operability.
Enclosure 13.21 PTlliA/4600/009
Chemistry Data Sheet Page 1 o f 4
Enclosure 13.21 PT/1lA/4600/009
Chemistry Data Sheet Page 2 of 4
!
Analysis CONC DATE/TIME j
I I PPB I
NC Chloride PPB
NC Fluoride
INC Dose Equiv
1-131
Enclosure 13.21
PT/1/A/4600/009
Chemistry Data Sheet Page 3 of 4
CF Cation
BE Sodium
BB Cation
(Conductivity)
Enclosure 13.21
Chemistry Data Sheet
Analysis DateITinie
i i
Seal Inj. Filter 1A
Seal Inj.Filter 1B
NC Filter IA
MC Filter 1I3
NV Mixed Bed Demin. 1A
NV Mixed Bed Demin. 1B
NV Cation Bed Demin.
Comments:
DUKE POWER CATAWBA OPERATIONS TMINlNC
c) Aux Feedwater (CA) System Response to a Loss of Instrument Air
With a Loss of Instrument Air, the flow control valves for all SIG's
will no longer fail open. A Mod has been completed on both
units which added air receiver tanks to the CA flow control valves
with enough air that will allow closure of these valves for 6Q
minutes after a loss of VI. This will also preclude SIG overfill on a
SGTR with a loss of AC Power and a subsequent loss of VI.
d) Transient
1) Air will be lost to:
Main feedwater control
Steam dump control
e SIG PORVs in AUTO
e NC PQRVs and PZR sprays
2) Reactor trip will occur on 'SIG LO LO bevel due to CF control
valves failed closed.
3) Steam dumps will not operate. SIG POWVs will not open in
'AUTO. Heat removal will be via SIG safeties.
4) NC system pressure may increase beyond PZR P O W setpoint.
N2 may have to be aligned to PZR PORVs.
2.2 Station Air System (VS)
A. System Purpose (Obj. #13, 14)
I. Supplies air to miscellaneous tools and equipment throughout the plant.
2. Backup air supply for the Instrument Air system.
3. Interconnections with other systems.
a) Recirculated cooling water (KR)cools the compressor
1) Effluent coding water temperature controls the KR flow rate
2) Inlet isolation valves closes automatically when the associated
compressor is shutdown.
b) Low Pressure Sewice Water (RL) cools the after coolers
c) instrument Air (Obj. #8, 14)
1) Backup supply for VI
2) IVS-78 will automatically open if VI pressure drops to 76 p i g .
3) Via two oil removal filters
d) Logic from the Engineered Safety Features Actuation System closes
the containment isolation valves upon receiving a Phase A (ST)
containment isolation signal. (VS-54B) (Ob]. #ti')
OP-CN-SS-VI FOR TRAiNING PURPOSES ONLY REV. 26
Page 20 of 36
Bank Question: 988 Answer: B
1 Pi(s) Unit 1 was operating at 100% powcr when the pressurizer spray line
developed a leak. Which one of the following conditions would cause the
safety injection system to be inoperable during this accident?
A. 1NI-118A and 1NI-150B are open with power applied
B. 1NI-162A is closed with powcr applied
C. 1NI-121A and lNI-152B are closed with power removed
D. INI-100B is open with power removed
Distracter Analysis:
A. Incorrect: This is a nonnal alignment for the NI cold leg injection
vaives.
Plausible: If the candidate thinks that these valves open on a safety
injection signal. These valves tic the two trains of safety injection
together so the candidate may think that cross-tying the two coid leg
injection paths constitutes a loss of train separation
B. Correct: INI-162 is the common isolation line for NI. The proper
position would be for the valve to be open with power removed.
C. Incorrect: INI-121A and 1NI-152B are the NI hot leg isolation lines
-the proper position is closed with power removed as listed.
Plausible: If the candidate thinks that power should not be
disconnected or the hot leg injection path should be open
D. Incorrect: 1NI-100B is the FWST supply iSOkdtiOII line. Normal
position is open with power removed as listed.
Plausible: Ifthe candidate conkses 1NI-100B with 1NI-103A --
which is a different FWST suppiy isolation valve. INI-103A is
normally open with power removed.
Levcl: SRO Only IWFR55.43@)5
KA: APE 008 AA2.10(3.6/3.6)
Lesson Plan Objective: NI Obj: 8,13
Source: New
Level of knowledge: comprehension
References:
1. OP-CN-ECCS-NI page 8
DUK PO WEB CATAWBA OPERATiQNS TRAINING ~
OBJECTIVES
I
Objective S
S
State the pu~poseand explain the operation of the NI Svstem. / X
State the ournose of the safeh, iniection D U ~ D Sand the FWST. Ix
Draw a block diagram of the ECCS system from the FWST to the NC X
Svstem per the lesson plan simplified drawina.
3escribe the indications to enter and actions taken for Reactor Trip or
SI Actuation durina plant shutdown.
State from memory all B.S.actions for the applicable systems,
subsystems and components, which require remedial action to be
!aken in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
OP-CN-EGCS-NI 6 0 R TRAINING PURPOSES ONLY REV. 35
Page 3 of 20
__-_
DUKE.. POWER ...
CATAWBA OPERATIONS - -- .
TRAINING
3) 'This recirculation flow path must be isolated during CoId/Hot leg
recirculation to prevent putting sump water back into the FWST.
The possibility exists for removing water from the sump and
contaminating the FWST, which would result in an uncontroliec!
radioactive release via FWS'T vent.
6 . The following is the power arrangement for the NI and other ECCS
pumps; all are powered from ETAIB. The sequencer loads are:
a) LOCA SEQUENCER: ND, NV, and NI Pumps
b) Blackout SEQUENCER: NV Pumps
B. The purpose of Refueling Water Storage Tank (FWS'I) is to provide the
initial source of borated water to the NI pumps, other ECCS components,
and the containment spray pumps. Limited in volume, this tank will
eventually be isolated. The continued suction source of the ECCS pumps
will be the containment sump
C. NI system motor operated valves (Unit l&2-the same)
1. NI-1006 FWST S ~ ~ p pIsolation,
ly open with power rernoved.
2. NI-I03/Ni..I 35 A&B pump's FWST suction isolation valves, normaliy
open with power available.
3. Discharge Isolations
a) Cold Leg injection isolation vaives, NI-I 18A and 150B are norrnally
open, tying the two trains of pimp discharge togethei.
b) NI-l62A, common isolation to the cold legs, is required open with
cmtrol board switches disconnected.
c) Globe throttle valves are locked in a position, determined by the
ECCS FLOW BALANCE TEST, to prevent puntp runout and to
equalize the flow to the loops.
d) M i Leg injection valves, NI-121A arid 1528, are riomially closed
with control board switches disconnected until going into Hot leg
Recirculation
2.2 Standby Alignment and procedure Limits a!id Precautions, refer to latest copy of
OP~l(2)/A!6200/006 (Obj. #5)
A. The system is placed in standby when al! NC cold legs teniperatures w e
greater than 3OC"F Tavg is less than 350°F.
6 . Norrnnl Standby Condition
1. An injection path exists from the FWST (NI-100B) to the cold legs (NI-
128A, 2508, 162A). Valves NI-1005 and NI-162A have a"5isconnect"
switch used to meet operability per T.S. surveillance. In the "disconnect"
position, the breaker control circuit is shut off and the control board
switch will not funccioc.
OP-CN-ECCS-NI FOR TRAINING PURPOSES ONLY REV. 35
Page 8 of 20
ECCS-Operating
3.5.2
3.5 EMERGENCY CORE CQOLBNG SYSTEMS (ECCS)
3.5.2 ECCS-Operating
K O 3.5.2 Two ECCS trains shall be OPERABLE*.
APPLICABILITY: MODES I;2,and 3.
ACTIONS
CONDlTlON I REQUIRED ACTION COMPLETION TIME
A. One or mere trains A.1 Restore train@)to 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s*
inoperable. OPERABLE status
&NJ
At le& 100% of the
ECCS flow equivalent to
a single OPERABLE
ECCS train available.
8. Required Action and B.l Be in MODE 3 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
associated Completion
Time not met.
8.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
'For each ECCS train on Unit 2. the Completion Time that one ECCS train can be inoperable as specified by
Required Action A . l may be extended beyond th6 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> up to 2R8 hours as part of the NSWS system upgrades.
System upgrades include maintenance and modification activities associated with cleaning of NSWS piping. vaI\es
ard branch lines, necessary repairs andior replacement, valve repair andlor replacement, and replacement of
portions of the NSWS piping to the AFW system. Upon completion of the cleaning. upgrades. and system restoratton
in refueling outage Z EOC12, this footnote is no longer applicable.
Catawba Units 1 and 2 3.5.2-1 Amendment Nos. 1891182
ECCS - Oaeratino
. 3.5.2
SURVEILLANCE REQUIREMENTS e _ _ _ _
SURVEILLANCE FREQUENCY
SR 3.5.2.1 Verify the following valves are in the listed position with 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
power to the valve operator removed.
Number Position Function
N1162A Open SI Cold beg
Injection
N1121A Closed SI Hot Leg
Injection
N1152B Closed SI Hot beg
Injection
Nil838 ClQSed RHR Hot Leg
Injection
N1173A Open RHR Cold Leg
lnjection
Nil788 Open RHR Cold Leg
Injection
N1100B Open SI Pump Suction
from RWST
N1147B Open SI Pump
Mini-Flow
SR 3.5.2.2 Verify each ECCS manual, power operated, and 31 days
automatic valve in the flow path, that is not locked,
sealed, or otherwise secured in position, is in the correct
PQskbl.
SI? 3.5.2.3 Verify ECCS piping is full of water. 32 days
SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with
flow point Is greater than or equal to the required the Inservice
developed head. Testing Program
(continued)
Catawba Units 1 and 2 3.5.2-2 Amendment Nos. 1931165
ECCS -Operating
3.5.2
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY
SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is 18 months
not locked, sealed, or otheiwise secured in position,
actuates to the correct position on an actual or simulated
actuation signal.
SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual 18 months
or simulated actuation signal.
SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each 18 months
position stop is in the correct position.
Centrifugal Charging Safety Injection
Pump Injection Throttle Pump Throttle
Valve Number Valve Number
N114 NI164
N116 N1166
N118 MI168
N120 N1170
SR 3.5.2.8 Verify, by visual inspection, each ECCS train containmeni 18 months
sump suction inlet is not restricted by debris and the
suction inlet trash racks and screens show no evidence
of structural distress or abnormal corrosion.
Catawba Units Iand 2 3.5.2-3 Amendment Nos. 173/165
ECCS-Operating
B 3.5.2
B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
B 3.5.2 ECCS-Operating
BASES
BACKGROUND The function of the ECCS is to provide core cooling and negative
reactivity to ensure that the reactor core is protected after ~ R ofY the
following accidents:
a. boss of coolant accident (LOCA), coolant leakage greater than the
capability of the normal charging system;
b. Rod ejection accident;
c. Loss of secondary coolant accident, including uncontrolled steam
or feedwater release; and
d. Steam generator tube rupture (SGTR).
The addition of negative reaetivlty is designed primarily for the loss of
secondary coolant accident where primary cooldown could add enough
positive reactbity to achieve criticality and return to significant power.
There are three phases of ECCS operation: injection, cold leg
recirculation, and hot leg recirculation. In the injection phase, water is
taken from the refueling water storage tank (RWST) and injected into the
Reactor Coolant System (RCS) through the cold legs. When sufficient
water is removed from the RWST to ensure that enough boron has been
added to maintain the reactor subcritical and the containment sumps
have enough water to supply the required net positive suction head to the
ECCS pumps, suction is switched to the containment sump for cold leg
recirculation. When the core decay heat has decreased to a level low
enough to be successfully removed without direct RHR pump injection
flow, the RHR cold leg injection path is realigned to discharge to the
auxiliary containment spray header. After approximately 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. part of
the ECCS flow is shifted to the hot leg recirculation phase to provide a
backflush which. for a cold leg break, would reduce the boiling in the top
of the core and prevent excessive boron concentration.
The ECCS consists of three separate subsystems: centrifugal charging
(high head), safety injection (Si) (intermediate head), and residual heat
removal (RHR) (low head). Each subsystem consists of two redundant,
100% capacity trains. The ECCS accumulators and the RWST are also
part of the ECCS, but are not considered part of an ECCS flow path as
described by this K O .
Catawba Units -I
and 2 B 4.5.2-1 Revision No. 0
ECCS - Operating
B 3.5.2
BASES
BACKGROUND (continued)
The ECCS flow paths consist of piping, valves, heat exchangers, and
pumps such that water from the RWST can be injected into the RCS
following the accidents described in this LCO. The major components of
each subsystem are the centrifugal charging pumps, the RHR pumps,
heat exchangers, and the SI pumps. Each of the three subsystems
consists of two 100% capacity trains that are interconnected and
redundant such that either train is capable of supplying 100% of the flow
required to mitigate the accident consequences. This interconnecting and
redundant subsystem design provides the operators with the ability to
utilize components from opposite trains to achieve the required 100% flow
to the core.
During the injection phase of LOCA recovery, a suction header supplies
water from the RWST to the ECCS pumps. Mostly separate piping
supplies each subsystem and each train within the subsystem. The
discharge from the centrifugal charging pumps combines, then divides
again into four supply lines, each of which feeds the injection line to one
RCS cold leg. The discharge from the SI and RHR pumps divides and
feeds an injection fine to each of the RCS cold legs. Throttle valves in the
SI lines are set to balance the flow to the RCS. This balance ensures
sufficient flow to the core to meet the analysis assumptions following a
LOCA in one of the RCS cdd legs. The flow split from the RHR lines
cannot be adjusted. Although much of the two ECCS trains are
composed of completely separate piping, certain areas are shared
between trains. The most important of these are 1) where both trains flow
through a single physical pipe, and 2) at the injection connections to the
RCS cold legs. Since each train must suppiy sufficient flow to the RCS to
be considered 100% capacity, credit is taken in the safety analyses for
flow to three intact cold legs. Any configuration which, when combined
with a single active failure, prevents the flow from either ECCS pump in a
given train from reaching all four cold legs injection points on that train is
unanalyzed and might render both trains of that ECCS subsystem
For LOCAs that are too small to depressurize the RCS below the shutoff
head of the SI pumps, the centrifugal charging pumps supply water until
the RCS pressure decreases below the SI pump shutoff head. During
this period, the steam generators are used to provide part of the core
cooling function.
During the recirculation phase of LOCA recovery, RHR pump suction is
transferred to the containment sump. The RHR pumps then supply the
other ECCS pumps. Initially, recirculation is through the same paths as
the injection phase. Subsequently, for large LOCAs, the recirculation
phase includes injection into both the hot and cold legs.
Catawba Units 1 and 2 B 3.5.2-2 Revision No. 0
ECCS - Operating
B 3.5.2
BASES
BACKGROUND (continued)
The high and intermediate head subsystems of the ECCS also functions
to supply borated water to the reactor core following increased heat
removal events, such as a main steam line break (MSLB). The limiting
design conditions occur when the moderator temperature coefficient is
highly negative, such as at the end of each cycle.
During low temperature conditions in the RCS, limitations are placed on
the maximum number of ECCS pumps that may be OPERABLE. Refer to
the Bases for LCO 3.4.12, "Low Temperature Overpressure Protection
(LTBP) System," for the basis of these requirements.
The ECCS subsystems are actuated upon receipt of an St signal. The
actuation of safeguard loads is accompbished in a programmed time
sequence. If offsite power is available, the safeguard loads start
immediately in the programmed sequence. If offsite power is not
availabie, the Engineered Safety Feature (ESFJbuses shed normal
operating loads and are connected to the emergency diesel generators
(EDGs). Safeguard loads are then actuated in the programmed time
sequence. The time delay associated with diesel starting, sequenced
loading, and pump starting determines the time required before pumped
flow is available to the core following a safety injection actuation.
The active ECCS components, along with the passive accumulators and
the RWST covered in LCO 3.5.2, "Accumulators," and LCO 3.5.4,
"Refueling Water Storage Tank (RWST)," provide the cooling water
necessary to meet GDC 35 (Ref. I).
APPLICABLE The LCO helps to ensure that the following acceptance criteria for the
SAFETY ANALYSES ECCS, established by 10 CFR 50.46 (Ref. 21,will be met following a
small break LOCA and there is a high level of probability that the criteria
are met following a large break LOCA:
a. Maximum fuel element cladding temperature is 5 2200°F;
b. Maximum cladding oxidation is 5 0.17 times the total cladding
thickness before oxidation;
c. Maximum hydrogen generation from a zirconium water reaction is
m< 0.01 times the hypothetical amount generated if all of the metal in
the cladding cylinders surrounding the fuel, excluding the cladding
surrounding the plenum volume, were to react;
Catawba Units 1 and 2 B 3.5.2-3 Revision No. 9
ECCS - Operating
B 3.5.2
APPLICABLE SAFETY ANALYSES (continued)
d. Core is maintained in a coolable geometry; and
e. Adequate long term core cooling capability is maintained.
The LCQ also limits the potential for a post trip return to power foilowing
an MSLB event and ensures that containment pressure and temperature
limits are met.
Each ECCS subsystem is taken credit for in a large break LQCA event at
full power (Refs. 3 and 4). This event has the greatest potential to
challenge the limits on runout flow set by the manufacturer of the ECCS
pumps. It also sets the maximum response time for their actuation. Direct
flow from the centrifugal charging pumps and SI pumps is credited in a
small break LQCA event. The RHR pumps are also credited, for larger
small break LQCAs, as the means of supplying suction to these higher
head ECCS pumps after the switch to sump recirculation. This event
establishes the flow and discharge head &.the design point for the
centrifugal charging pumps. The MSbB analysis also credits the SI and
centrifugal charging pumps. Although some ECCS flow is necessary to
mitigate a SGTR event, a single failure disabling one ECCS train is not
the limiting single failure for this transient. The SGTR analysis primary to
secondary break flow is increased by the avaiiabiiity of both centrifugal
charging and SI trains. Therefore, the SGTR analysis is penalized by
assuming both ECCS trains are operable as required by the LCQ. The
QPEBABILITY requirements for the ECCS are based on the following
LQCA analysis assumptions:
a. A large break LQCA event, with loss of offsite power and a single
failure disabling One ECCS train; and
b. A small break LBCA event, with a loss of offsite power and a single
failure disabling one ECCS train.
During the blowdown stage of a LQCA. the RCS depressurizes as
primaly coolant is ejected through the break into the containment. The
nuclear reaction is terminated either by moderator voiding during large
breaks or control rod insertion for small breaks. Following
depressurization, emergency cooling water is injected into the cold legs,
flows into the downcomer, fills the lower plenum, and refloods the core.
The effects on containment mass and energy releases are accounted for
in appropriate analyses (Ref. 3). The LCO ensures that an ECCS train
will deliver sufficient water to match boiloff rates soon enough to minimize
the consequences of the core being uncovered following a large LQCA.
~
Catawba Units 1 and 2 B 3.5.2-4 Revision No. 0
ECCS - Operating
B 3.5.2
BASES
APPLICABLE SAFETY ANALYSES (continued)
It also ensures that the centrifugal charging and SI pumps will deliver
sufficient water and boron during a small LOCA to maintain core
subuiticality. For smaller LOCAs, the centrifugal charging pump delivers
sufficient fluid to maintain RCS inventory. For a small break LOCA, the
steam generators continue to serve as the heat sink, providing part of the
required core cooling.
The ECCS trains satisfy Criterion 3 of 10 CFR 50.36 (Ref. 5).
LCO In MODES 1,2, and 3, two independent (and redundant) ECCS trains are
required to ensure that sufficient ECCS flow is available, assuming a
single failure affecting either train. Additionally, individual components
within the ECCS trains may be called upon to mitigate the consequences
of other transients and accidents.
In MODES I, 2, and 3, an ECCS train consists of a centrifugal charging
subsystem, an SI subsystem, and an RHR subsystem. Each train
includes the piping, instruments, and controls to ensure an OPERABLE
flow path capable of taking suction from the RWST upon an SI signal and
automatically transferring suction to the containment sump.
During an event requlring ECCS actuation, a flow path is required to
provide an abundant supply of water from the RWST to the RCS via the
ECCS pumps and their respective supply headers to each of the four cold
leg injection nozzles. In the long term, this flow path may be switched to
take its supply from the containment sump and to supply its flow to the
RCS hot and cold iegs. The flow path for each train must maintain its
designed independence to ensure that no single failure can disable both
ECCS trains.
APPLICABILITY In MODES 1,2, and 3, the ECCS OPERABILITY requirements for the
limiting Design Basis Accident, a large break LOCA, are based on dulI
power operation. Although reduced power would not require the same
level of performance, the accident analysis does not provide for reduced
cooling requirements in the lower MODES. The centrifugal charging
pump performance is based on a small break LOCA, which establishes
the pump performance curve and has less dependence on power. The SI
pump performance requirements are based on a small break LOCA. For
both of these types of pumps, the large break LOCA analysis depends
only on the flow value at containment pressure, not on the shape of the
flow versus pressure curve at higher pressures. MODE 2 and MODE 3
requirements are bounded by the MODE 1 analysis.
Catawba Units 1 and 2 B 3.5.2-5 Revision No. 0
ECCS - Operating
B 3.5.2
BASES
APPLICABILITY (continued)
This LCO is only applicable in MQDE 3 and above. Below MODE 3, the
SI signal setpoint is manually bypassed by operator control, and system
functional requirements are relaxed as described in LCO 3.5.3,"ECCS-
Shutdown."
As indicated in the Note, the flow path may be isolated for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in
MODE 3, under controlled conditions, to perform pressure isolation valve
tesfhg per SR 3.4.14.1. The flow path is readily restorable from the
control room.
In MODES 5 and 6, plant conditions are such that the probability of an
event requiring ECCS injection is extremely low. Core cooling
requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops-
MODE 5, Loops Filled," and LCQ 3.4.8, "RCS Loops-MQDE 5, hoops
Not Filled." MODE 6 core cooling requirements are addressed by
LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-
High Water Level," and LCO 3.9.5, "Residual Heat Removal (RHR) and
Coolant Circulation-Low Water Level."
ACTIONS
With one or more trains inoperable and at least 100% of the ECCS flow
equivalent to a single OPERABLE ECCS train available, the inoperable
components must be returned to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The
72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on an NRC reliability evaluation
(Ref. 6) and is a reasonable time for repair of many ECCS components.
An ECCS train is inoperable if it is not capable of delivering design flow to
the RCS. individual components are inoperable if they are not capable of
performing their design function or supporting systems are not available.
The LCO requires the OPERABILITY of a number of independent
subsystems Due to the redundancy of trains and the diversity of
subsystems, the inoperability of one component in a train does not render
the ECCS incapable of performing its function. Neither does the
inoperability of two different components, each in a different train,
necessarily result in a loss of function for the ECCS. The intent of this
Condition is to maintain a combination of equipment such that 100% of
the ECCS flow equivalent to a single OPERABLE ECCS train remains
available. This allows increased flexibility in plant operations under
circumstances when components in opposite trains are inoperable.
Catawba Units 1 and 2 0 3.5.2-6 Revision No. 0
ECCS - Operating
3 3.5.2
BASES
ACTIQNS (continued)
An event accompanied by a loss of offsite power and the failure of an
EBG can disable one ECCS train until power is restored. A reliability
analysis (Ref. 6) has shown that the impact of having one full ECCS train
inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Reference 7 describes situations in which one component, such as an
RHR crossover valve, can disable both ECCS trains. With one or more
compenent(s) inoperable such that 100% of the flow equivalent to a
single OQEfWBLE ECCS train is not available, the facility is in a condition
outside the accident analysis. Therefore, LCO 3.0.3 must be immediately
entered.
B.1 and 8.2
If the inoperable trains cannot be returned to OPERABLE status within
the associated Completion Time, the plant must be brought to a MODE in
which the LCQ does not apply. To achieve this status, the plant must be
brought to MQDE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The
allowed Compietion Times are reasonable, based on operating
experience, to reach the required plant conditions from full power
conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.5.2.1
REQUIREMENTS
Verification of proper valve position ensures that the flow path from the
ECCS pumps to the RCS is maintained. Misalignment of these valves
could render both ECCS trains inoperable. Securing these valves using
the power disconnect switches in the correct position ensures that they
cannot change position as a result of an zctive failure or be inadvertently
misaligned. These valves are of the type, described in Reference 7,that
can disable the function of both ECCS trains and invalidate the accident
analyses. A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered reasonable in view of other
administrative controls that will ensure a mispositioned valve is unlikely.
Verifying the correct alignment for manual, power operated, and
automatic vaives in the ECCS flow paths provides assurance that the
proper flow paths will exist for ECCS operation. This SR does not apply
to valves that are locked, sealed, or otherwise secured in position, since
these were verified to be in the correct position prior to locking, sealing,
~
Catawba Units 1 and 2 B 3.5.2-7 Revision No. 0
ECCS - Operating
0 3.5.2
BASES
SURVEILLANCE REQUIREMENTS (continued)
or securing. A valve that receives an actuation signal is allowed to be in a
nonaccident position provided the valve will automatically reposition
within the proper stroke time. This Surveilrance does not require any
testing or valve manipulation. Rather, it involves verification that those
valves capable of being mispositioned are in the correct position. The
32 day Frequency is appropriate because the valves are operated under
administrative control.
This Frequency has been shown to be acceptable through operating
experience.
With the exception of the operating centrifugal charging pump, the ECCS
pumps are normally in a standby, nonoperating mode. As such, flow path
piping has the potential to develop voids and pockets of entrained gases.
Maintaining the piping from the ECCS pumps to the RCS full of water by
venting the ECCS pump casings and accessible discharge piping high
points ensures that the system will perform properly, injecting its full
capacity into the RCS upon demand. This will also prevent water
hammer, pump cavitation, and pumping of noncondensible gas (e.g.. air,
nitrogen, or hydrogen) into the reactor vessel foliowing an SI signal or
during shutdown cooling. The 31 day Frequency takes Into consideration
the gradual nature of gas accumulation in the ECCS piping and the
procedural controls governing system operation.
- -
Periodic surveillance testing of ECCS pumps to detect gross degradation
caused by impeller structural damage or other hydraulic component
problems is required by Section XI of the ASME Code. This type of
testing may be accomplished by measuring the pump developed head at
only one point of the pump characteristic curve. This verifies both that the
measured performance is within an acceptable tolerance of the original
pump baseline performance and that the performance at the test flow is
greater than or equal to the performance assumed in the plant safety
analysis. SRs are specified in the Inservice Testing Program, which
encompassesSection XI of the ASME Code.Section XI of the ASME
Code provides the activities and Frequencies necessary to satisfy the
requirements.
Catawba Units 1 and 2 B 3.5.2-8 Revision No. 0
ECCS - Operating
B 3.5.2
BASES
SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.25 and S R 3.5.2.6
These Surveillances demonstrate that each automatic ECCS valve
actuates to the required position on an actual or simulated SI and
Containment Sump Recirculation signal and that each ECCS pump starts
on receipt of an actual or simulated SI signal. This Surveillance is not
required for valves that are locked, sealed, or otherwise secured in the
required position under administrative controls. The 18 month Frequency
is based on the need to perform these Surveillances under the Conditions
that apply during a plant outage and the potential for unplanned plant
transients if the Surveillances were performed with the reactor at power.
The 48 month Frequency is also acceptable based on consideration of
the design reliability (and confirming operating experience) of the
equipment. The actuation logic is tested as part of ESF Actuation System
testing, and equipment performance is monitored as part of the Inservice
Testing Program.
The position of throttle valves in the flow path on an SI signal is
necessary for proper ECCS performance. These valves have mechanical
iocks to ensure proper positioning for restricted flow to a ruptured cold
leg, ensuring that the other cold legs receive at least the required
mlnimum flow. The 18 month Frequency is based on the same reasons
as those stated in SR 3.5.2.5 and SR 3.5.2.6.
Periodic inspections of the containment sump suction inlet ensure that it
is unrestricted and stays in proper operating cundition. The 18 month
Frequency is based on the need to perform this Surveillance under the
conditions that apply during a plant outage and on the need to have
access to the location. This Frequency has been found to be sufficient to
detect abnormal degradation and is confirmed by operating experience.
~~ ~
Catawba Units 1 and 2 B 3.5.2-9 Revision No. 1
ECCS - Operating
B 3.5.2
BASES
REFERENCES 1. 10 CFR 50, Appendix A, GDC 35.
2. 10 CFR 50.46.
3. UFSAR, Section 6.2.4.
4. UFSAR, Chapter 15.
5. 10 CFR 50.36, Technical Specifications, (cj(2)(ii).
6. NRC Memorandum to V. Stello, Jr., from R.L. Baer,
"Recommended Interim Revisions to LCOs for ECCS
Components," December 1, 1975.
7. IE Information Notice No. 87-01.
Catawba Units 1 and 2 B 3.5.2-10 Revision No. 0
Rank Question: 969 Answer: C
1 Pt(s) Unit 1 is operating at 65% power following a turbine runback caused by a feed
pump failure. Given the following events and conditions:
e The RO reports that rod control bank D has inserted past the rod insertion
limit (FUL)
b Annunciator 1AD-2, FV9, Control Rod BankLo-Lo Limit, has not alarmed.
0 IAE has reported that a failed anrimciator card must be replaced but the part
will not be available until next week.
Which one of the following acti0nr is required in response to this failed alarm?
A. The shift work manager must initiate a temporary modification to
change the ControfItodBank Lo Limit (1AD-2, A/9) annunciator
setpoint to the ContPulRodBankLo-Lo rod insertion limit.
B. The unit supenisor must initiate a control panel information tag for
the 1AD-2, Bl9.
C. The operations s h i t manager shall ensure that aiternate indications are
monitored to duplicate the function of the failed annunciator.
D. The reactor operator must enter the requirement to verify RIL limits
manually during transients in the shift turnover log.
____s__________________s__s_____s__s____---------------------------------------------------
Distracter Analysis:
A. Incorrect: An increased surveillance sheet must be initiated.
Plausible: If the candidate does not know the requirement, this is a
logical alternative.
J3. Incorrect: An increased surveillance sheet must be initiated.
Plausible: Ifthe candidate does not know the requirement, this is a
logical alternative.
C. Correct:
D. Incorrect: An increased surveillance sheet must be initiated.
Plausible: If the candidate does not h o w the requirement, this is a
logical alternative.
Level: SRO Only 10CFW55.43@)5
K A G 2.4.33 (2.4D.8)
Lesson Plan Objective: ADM-NSOS Obj: 29
Source: New
Level of howledge: memory
References:
I . OQ-CN-ADM-NSOSpage 11
2. OMP 2-3 1 page 2
DUKE BOWER CATAWBA QPERATlONS TRAlNlNG
I N L L P
Objective S b P P T
S O R S R
-
0 0 0
12 rocess for approval and extension of Technical
I
13
14 Describe the Work List Priority scheme and how priority conflicts x x
- are resolved (OMP 2-5)
15 Describe how the Work List is completed (OMP 2-5)
- 1 jx x
16 Explain peBonnel responsibilities for key control (OMP2-9)
- 1x1 I x l x l x
27 Describe restrictionslrequirementsfor the use of file keys.
- (OMP 2-9)
18 Describe restrictions/requirementsfor long term key removal.
l x I lxlxlx
19
-
20
-
21
-
22
I
23
control (OMP 2-16, SD 3.1.20,).
24 Describe the expectations for conduct of personnel while in the
Control Room (OMP 2-16)
25 Explain the process of maintaining the Unified Logbook in
accordance with OMP 2-17 (as appropriate to job level).
-
26
before, during and after the formal transfer of shift responsibilities.
(OMP 2-22)
27 Explain the process of maintaining Control Room Instrumentation
Status Logs and actions required for failed annunciators.
- (QMP 2-32)
! 1 lXIXlX
28 Describe how to initiate a corrective work request. (WFM 401)
- IX I 1X I I X
OP-CN-ADM-NSO5 FQR TRAINING PURPOSES ONLY Rev11
Page 5 of 12
I-
DUKE POWER
_
CATAWBA OPERATIONS TRAlNlNG
5. Alarm response during execution of EFs/AFs
6. Alarm response during Crew Briefs.
7. Exceptions to normal alarm response protocol.
2.6 Unit Unified Logbooks
A. Using OMF 2-17, (Unit Unified Logbook Maintenance), review responsibilities
and processes associated with the Unit Supervisor and Control Room
subiogs. (Obj. #25):
1. Personnel responsible for maintaining Control Room and Unit Supewisor
and FT sublogs.
2. General instructions and entry procedures.
3. Definition of QFEN ITEM and how OPEN ITEMS are documented.
4. Logbook required entries.
5. Test Logbook (PT Sublogs)
El. Using OMP 2-34, (Control Room Instrumentation Status), review the process
involved in maintaining the Control Room Annunciator Status Cog, audits and
increased surveillance. (Qbj. #27)
2.7 Key Control
A. Using OMP 2-9, (Administration and Control of Keys), review personnel
responsibilities for control of keys. (Obj. #16)
6. Using OMF 2-3,(Administration and Control of Keys), describe
requirements/restrictiorssfor:
1. Use of File Keys (Obj. # I T )
2. Long Term Key Removal (Qbj. #18)
3. Interlock Keys (Obj. #19)
2.8 Work Requests (Obj. #28)
A. Refer to WFM 401 (Problem Communication Guidelines)
B. Problem Communication
1. Normally problems with plant equipment are communicated to the Work
Control Single Point Of Contact (SFOC) via the WMS Problem identifier
Screen (RQQ5 or Rat 0).
2. For emergency work, SPOC should be notified of problems by telephone.
SPOC will complete work request documentation as required.
3. Problem complexity may require that the problem be communicated in
person in the Work Control Center.
QP-CN-ADM-NSOS FOR TRAINING PURPOSES ONLY Rev71
Page l f of 12
Operations Management Procedure 2-3 1 Page 2 of 3
4. Responsibilities
4.1. The Reactor Operators (ROs) are responsible for review and knowledge of
Control Room annunciator status.
4.2. The Unit Lead is responsible for Increased Surveillance Items.
A. The Unit Lead should work with Engineering to ensure there are no
problems with using the Increased Surveillance.
H. The IJnit Lead should work with Work Control to ensure timely
completion of any work order creating the need for an Increased
Surveillance.
4.3. 'The Operations Shift Manager (OSM) is responsible for actions on failed
instrumentshannunciators.
5. Annunciator Status Log Instructions
5.1. Control Room annunciator status shall k reviewed at turnover per
CbMP 2-22 (Shift Turnover').
5.2. "Strapped" Bistable Inputs
0 Bistable with "strapped" inputs shall be reviewed at turnover per
OMP 2-22 (Shift 'turnover).
0 As directed by IAE procedures, distinctive red flags shall be placed on
"strapped bistable windows.
5.3. Annunciators with inputs disabled by Temporan1 Modifications ( t M ) shaall be
recorded on Attachment 8.1 (TMs Affecting Control Room Annunciators).
e Attac.hnient shall be filed in Ops Shift Routine Logbook.
6. Action on Failed ~nstrumentation/Annunciators
6.1. The OSM shall ensurc that alternate indications are monitored to duplicate the
function of the failed instruinentation/annuncia~or.
0 Alternate indications to be monitored shall be recorded on an Increased
Surveillance Sheet (Attachment 8.2) per Section 7.
6.2. The Shift Work Manager shall be notified of any instrumentation determined
to be failed.
PANEL: 1AD-2
CONTROL ROD BANK LO-LO LIRIIl B/9
SETPOPNT: Variable, ealcuiated based on auctioneered high D:T.
ORIGIN: Kod insertion limit computer.
PROBABLE 1. Boron concentration too low for power andjor Xenon conditions.
CAUSE: 2. Reactor Shutdown.
3. Rod insertion limit computer or instnlment malfunction.
4. Major load rejection or transient.
AUTOMATIC hone
ACTIONS:
IMMEDIATE 1. Stop any dilution in progress.
ACTIONS: 2. Refer to AI/1/.45500/013 (Boron Dilution).
3. Detennine affected control bank(s) by referring to lRFM17 in
Cable Spading Room (AUX 5 4 7 ) and acknowledge the a l m .
SUPPLEMENTARY 1. Kefer to TS 3.1.6 (Contra1 Bank Insertion Limits).
ACTIONS: 2. Refer to TS 3.1, I (Shutdown Margin (SDM)).
REFERENCES: I. TS 3.1.I (Shutdown Margin (SDM)) and
TS 3.1.6 (Control Bank Insertion Limits).
2. FSAR, Figure 7-2 (Part 9 of 16)
OPiI iRi6100iOlOC
PANEL: 1AD-2 Page 12 of 68
CONTROL ROD BANK LO LIMIT
SETPOINT: Vanable calculated based on auctioneered high DiT
ORIGIN: Rod insertion limit computer.
PROBABLE 1. Boron concentration too low for power andior Xenon conditions.
CAUSE: 2. Rod insertion limit computer or instrument malfunction.
3. Major load rejection or transient.
AUTOMATIC None
ACTIONS:
IMMEDIATE i. Stop any dilution in progress.
ACTIONS: 2. Determine affected control bank(s) by referring to 1RFM16 in
Cable Spreading Room (AtJX 577) and acknowledge the alarm.
SUPPLEMENTARY 1. Borate coolant system as necessary until alami clears per
ACTIONS: OPil/A%150i009 (Boron Concentration Control).
2. -
IF due to temperature instrument malfunction, defeat affected
instrument using "T-AVG DEITAT" and "ATEMP DEFEAT"
switch.
3. Refer to TS 3.1.6 (Control Bank Insertion Limits).
4. -
IF due to instniment failure, have IAE trip bistable on affected
channel per Tech Spec Action.
REFERENCES: 1. TS 3.1.1 (Shutdown Margin (SDM) and
Ths 3.1.6 (Control Rank Insertion Limits)
2. FSAR, Figure 1-2 (Part 9 of 16)
Bank Question: 962 Answer: D
1 Pt(s) Unit 1 is conducting a containment ais release at 85% power. Given the
following events and conditions:
The last containment atmosphere sample at 0600 was consistent with
the current gaseous waste release (GWR) pennit.
- The latest air release was started at 1400.
The release flow indicator on l h K - 5 reads 200 CFM.
1EMF-36 (UnitVent Gas Monitor) is at the trip 1 setpoint.
The VQ flow totalizer is out of service.
Which of the following conditions reported at 1700 requires isolation of &e
air release until a new GWR permit is approved?
A. The VQ air release flow indicator on 1MC-5 fails high.
B. IEMF-39 (Containment Gas Monitor) increases to Trip 2 alarm
due to a power supply fluctuation.
C. The 1800 containment atmosphere sample must be delayed for 4
hours due to sample valve repairs.
D. IEm-37 (UnitVent Iodine Monitor) increases to Trip 2 aIarm
due to a fuel pin failure.
--
Distracter Analysis:
A. Incorrect: The flow indicator is not required to conducVcontinue an
air release.
Plausible: If the candidate believes that release flow is critical with
the totalizer out of service.
B. Incorrect The release can be reset and continued on the current
GWR.
Plausible: If the candidate does not know that spurious nips can be
reset twice.
C. Incorrect: The release can continue on the current GWR as long as
the sample is analyzed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the previous sample.
Plausible: Ifthe candidate thinks the sample frequency (12 hrs) is the
GWR validation requirement.
D. Correct:
Level: SRO Only lOCFR55.43@)4
KA: SYS 029 G 2.3.8 (233.2)
Lesson Plan Objective: CNT-VQ Obj: 4
Source: New
Level of knowledge: comprehension
References:
1. OP-CN-CNT-VQ pages 9
2.OP/I/A/64.50/17
3. OP/I/A/6450/17 EncI 4.2
4.OP/UA/6450/17 Enci 4.3
I
Objective S
s
State the P L I ~ C JofS the
~ VQ system Ix
Describe the sources of pressure fluctuations in containment during X
normal operations
Describe the simal that will auto stop a VQ fan I
Given appropriate plant conditions, apply Limits and Precautions
assodated with related station procedures.
Exalain how an ais addition from containment is ammDlished I
Describe the effect a S, signal has on VQ
Given a set of specific plant conditions and access to reference materials, I
determine the actions necessary
...
to comply
....
with._
Tech Spets/SI.Cs
...... ..
..... ...
'li
TIME: 1.O Hour
OP-CN-CNT-VQ FOR TRAINING PURPOSES ONLY R N . 20
Page 3 of If
DUKE BO WEW CATAWEA OPERATIONS TRAINING
2. Air Release Mode is performed if containment pressure is greater than
0.09 psig and a release is desired. Refer to OP/1/A16450/017 Enclosure
4.2 (OBJ. 6)
UNIT
VENl
I
UPPER CONTAINMENT
3. Initiating and terminating a GWR form refer to OP/l/AI6450iO17
Enclosure 4.3
4. Signals or conditions that terminate a release or addition (OBJ. 7)
a) At any time by manual operator action from the control room.
b) Automatically when containment pressure returns to 0 psig by closing
VQ-to or VQ-23.
c) The Containment Ais Release Fans automatically shut off when low
flow is detected at fan discharge.
d) EMF-35,36,37 trig 2 will auto dose VQ-10.
OP-CM-CMT-VQ KIR TRAINING PURPOSES ONLY REV. 20
Page 9 of 1f
OP/1/N6450/0 17
Page 2 of 3
Containment Air Release and Addition System
1. Purpose
To outline the proper operation of the Containment Air Release and Addition System.
2. Limits and Precautions
2.1 Do NOT exceed Containment Pressure Limits of -0.08 psig and +0.25 pig. Tech Spec
Containment Pressure Limits are -0.1 p i g to +0.3 p i g .
2.2 When manually operating any motor operated valve, minimize the torque applied to the
handwheels.
2.3 ARer manual operation, maintenance or packing adjustment of any motor operated
Safety Related valve, it shali be cycled electrically to ensure reliable automatic operation.
2.4 Pressure switches for valve operation should m T be manuaily overridden since ice
condenser doors are very sensitive to over or under pressure conditions.
2.5 When Containment Air Release Filter unit pre-filter or absolute filter differential
pressure reaches 2.5 inches H20,the standby fan should be placed in service and action
initiated to replace the dirty filter(s).
2.6 A new Gaseous Waste Release (GWR) sample is required if:
E 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has elapsed since the last sample.
e VQ release is automatically terminated due to a valid controlling EMF actuation. If
actuation is due to an EMF spike, the release may be re-attempted twice before a new
sample is required.
2.1 A VP; VQ or Unit Vent Sample is required if:
e Rx Trip or Startup occurs.
E Rated Themial Power change of 2 15% in one hour occurs followed by a Thermal
Power Stabilization bower level constant at desired power level).
2.8 If 1EMF-37 or 1EMF-40 has reached the Trip 1 setpoint, RP should be notified to
change the c'artridge before a release is attempted.
3. Procedures
Refer to Section 4 (Enclosures).
OP/llA/645O/O17
Page 3 of 3
4. Enclosures
4.1 Air Addition Mode
4.2 Air Release Mode
4.3 lnitiation and Termination of a GWR Permit Report
4.4 Auxiliary Building Valve Checklist
4.5 Reactor Building Valve Checklist
4.6 Auxiliary Building Separate Verification Valvc Checklist
4.7 Reactor Building Separate Verification Valve Checklist
4.8 Manual Air Addition Mode
Enclosure 4.2 OP/1/N6450/0 17
Air Release Mode Page 1 of 6
1. Initial Conditions
1.1 Review the Limits and Precautions.
1.2 Verify Conkainment Pressure > 0.09 psig.
__ 1.3 Verify CR SRO has signed and dated the appropriate sheet of the Gaseous Waste
Release (GWR) Record authorizing releases:
c3 "VQ release monitored by EMF 39(L)"
0 "VQ release monitored by EMF 36(L)"
__ 1.4 Verify Containment pressure increase is NOT due to a LOCA or steam line break.
1.5 Review the "SPECIAL INSTRUCTIONS FOR RELEASE" section on the GWR Permit
Report.
2. Procedure
__ 2.1 Ensure the following enclosures are complete:
Enclosure 4.4 (Auxiliary Building VaIve Checklist)
0 Enclosure 4.5 (Reactor Building Valve Checklist)
0 Enclosure 4.6 (Auxiliary Building Separate Verification Valve Checklist)
0 Enclosure 4.7 (Reactor Building Separate Verification Valve Checklist)
2.2 Perform the following to sign off the "EMF Operable and Source Checked" blank on the
appropriate Gaseous Waste Release (GWR) Record
2.2.1 1EMF-39 (low range) is operable, perform the following:
2.2.1 .I Verify EMF-39 is specified for use on the GWR Permit Report.
2.2.1.2 Verify 1EMF-39 is operable per SLC 16.11-7 using
OP/O/N6500/080 (EMF W86A Output Modules).
2.2.1.3 Set 1EMF-39 (low range) setpoints to the value specified on the
GWR Permit Report using OP/O/N6500/080 (EMF RP86A
Output Modules).
2.2.1.4 Verify trip setpoints are set to the values as specified on the GWR
Permit Report using OPI0IN65001080 (EMF RP86A Output
Modules).
Enclosure 4.2 OP/hA/6450/0 17
Air Release Mode Page 2 of 6
__ 2.2.2 IEMF-39 (low range) is inoperable &Q 1EMF-36 (low rangc) is to be
used to monitor this release, perform the following:
CI 2.2.2.1 Verify 1EMF-36 is specified for use on the GWR Permit Report.
0 2.2.2.2 Verify 1EMF-36 is operable using OP/O/N6500~080(EMF
RP86A Output Modules).
NOTE: 1EMF-36 (low range) trip setpoints are pre-established for offsite dose.
E! 2.2.2.3 Verify trip setpoints are set to the values as specified on the GWR
Permit Report using OP/O/N6500/080 (EMF RE'86A Output
Modules).
2.2.2.4 N/A the "Wblank on the "VQ release monitored by EMF 36(L)"
sheet of the Gaseous Waste Release (GWR) Record.
_ I 2.2.3 -
IF 1EMF-39 1EMF-36 are both inoperable, perform the following:
Kl 2.2.3.1 Verify EMF-39 and EMF-36 are both N/Aed on the GWR Permit
Report.
0 2.2.3.2 Notify RP to take grab samples per HP/O/B/1004/005
(Containment Air Release and Addition (VQ) And Containment
Purge Ventilation (VP) System Release).
Person notified
El 2.2.3.3 N/A the "EMF Operable and Source Checked" blanks on the
Gaseous Waste Release (GWR) Record.
l
l 2.2.3.4 N/A Step 2.3.
-2.3 Set u- EMF Chart re Jrder as follows:
2.3.1 Ensure the paper drive is on for the applicable EMF chart recorder:
- 1MICR6640if IEMF-39 (L) is used
- 1MICR6650 if 1EMF-36 (L) is used
2.3.2 Stamp and record the following on the chart paper:
Date
Time
GWW#
Initials
Enclosure 4.2
OPi~lN64501017
Air Release Mode Page 3 of 6
~
sv - - .. 2.4 Adjust "IVQ-10 VQ Fans Disch To Unit Vent" manual loader (IMC5) to I the
"Recommended Release Rate (cfm)" on the GWR Permit Report.
2.5 Record the following on the appropriate Gaseous Waste Release (GWR) Record.
2.5.1 IF the tobilker is operable, reset it and enter "0" in the "Initial Integrator
Reading" blank
2.5.2 the totalizer is inoperable, N/A the "Initial Integrator Reading" blank.
2.6 Open the following valves (1MC5):
Record date/time f k t valve is opened 1
0 IVQ-2A (VQ Fan Suct From Cont Isol)
0 IVQ-3B (VQ Fan Suct From Cont Isol)
2.7 Place one VQ train in service as foilows (1MC5):
-2.7.1 To place A train in service, perform the following:
0 2.7.1.1 Place "VQ Filt Htr A" in the "AUTO" position.
E l 2.7.1.2 Start "Cont Air Re1 Fan 1A".
-2.7.2 To place B train in service, perfonn the following:
63 2.7.2.1 Place "VQ Filt Htr B" in the "AUTO" position.
0 2.7.2.2 Start "Cont Air Rei Fan IB".
2.8 Notify Rp that the VQ release has been started and give start time as recorded in
Step 2.6.
Person notified
m _ 2.9 Record the VQ start dateltime on the following:
3 Appropriate Gaseous Waste Release (GWR) Record
0 Control Room Autolog
Enclosure 4.2 OP/I/A/6450/OI7
Air Release Mode Page 4 of 6
-2.10 the OAC OR Computer Point C I P l l l 2 (Average Containment Pressure, best) is out
of service, record containment pressure as read on 1VQP5040 (Containment Pressure) on
BMCS every 30 minutes in the Control Room Log for the duration of the VQ Release.
{PIP 93-0074)
__ 2. I 1 the VQ fan does NOT automatically shutdown at approximately 0 psig, perfom the
following:
- N/A Step 2.12.
Perform Step 2.13.
__ 2.12 WHEN Containment pressure decreases to approximately 0 psig, verify that
"lVQ-IO VQ Fans Disch To Unit Vent" closes, by no flow indicated on the manual
loader (black needle).
2.13 Revet "IVQ-10 VQ Fans Disch To Unit Vent" by adjusting the manual loader fully
counterclockwise.
2.14 Secure the VQ train placed in service in Step 2.7 as follows:
__ 2.14.1 To secure A train, perform the following:
5 2.14.1.1 Ensure "Cont Air Re1 Fan 1A" has stopped.
0 2.14.1.2 Place "VQ Filt Htr A" in the "OFF" position
2.14.2 To secure B h i n , perform the following:
c3 2.i4.2.1 Ensure "Cont Air Re1 Fan 1B" has stopped.
2.14.2.2 Place "VQ Filt Btr B" in the ""OFF" position
2.15 Close the following valves:
Record ddte/time both valves are closed i
IVQ-2A (VQ Fan Suct From Cont Isol)
[3 IVQ-3B (VQ Fan Suct From Cont Isol)
Enclosure 4.2 OP/1/A/6450/017
Air Release Mode Page 5 of 6
-2.16 Stamp and record the following on the chart paper:
Date
- Time
- GWRX
Initials
-2.17 Notify RP that the VQ release has been terminated and give termination time as recorded
inStep2.15.
Person notified
2.18 Record the VQ terminate datehime on the following:
I3 Appropriate Gaseous Waste Release (GWR) Record
0 Control Room Autolog
2.19 Record the following on the appropriate Gaseous Waste Retease (GWR) Record
2.19.1 -
IF the totalizer is operable, perform the following:
2.19.1.1 Record totalizer reading
2.19.1.2 Record totalizer value in "Final Integrator Reading" blank of the
appropriate Gaseous Waste Release (GWR) Record sheet.
2.19.1.3 Enter the volume releawd in the "Volume" blank.
- Volume = final integrator reading X 10.
2.19.2 -
IF the totalizer is inoperable, perform the following:
2.19.2.1 NiA the "Finai Integrator Reading" blank.
2.19.2.2 Enter the volume released in the "Volume" blank.
- Volume = 350 CPM X Release Time (in minutes).
e Release Time = DateiTime initiated DateRime terminated.
2.19.3 Enter the "Highest EMF Reading" during the release as read on the chart
recorder.
2.19.4 Sign the "Control Room Operator" blank.
Enclosure 4.2 o P i l i ~ 6 4 5 o i 17
o
Air Release Mode Page 6 of 6
e nearest 100 prior to
2.20 IEMF-39 was used for this release, reset IEMF-39 (low range) trip setpoints using
OP/O/A/6500/080 (EMF RP86A Output Modules):
__ 2.20.1 in Mode 5 or 6, the trip setpoints shall be as follows:
- Trip 2 = 17,400 cpm + Existing reading, Rounded down to the
nearest 100 cpm
Trip 1 = Trip2X.70
___ 2.20.2 in Mode 1 , 2 , 3 or 4, the trip setpoints shall be set as follows:
Trip 2 = 3 X Containment Atmosphere Activity (sampled at all three
locations) as indicated by EMF allowing about I S minutes for
indication to stabilize.
- Trip 1 = Trip 2 X .70.
2.20.3 Signoff "EMF Setpoints Reset" blank on the "VQ release monitored by EMF
39(L)" sheet ofthe Gaseous Waste Release (GWR) Record.
2.20.4 Verify trip setpoints are reset as described in Step 2.20.1 or 2.20.2 using
OP/O/N6500/080(EMF W86A Output Modules).
2.2 1 1EMF-36 was used for this release, NIA the "EMF Setpoints Reset" blank on the
"VQ release monitored by EMF 36(L)" sheet of the Gaseous Waste Release (GWR)
Record.
2.22 Do NOT file this enclosure in the Control Copy folder of this procedure.
Enclosure 4.3 OP/1/N645O/O 14
Initiation and Termination of a GWR Page 1 of 1
Permit Report
Information Use
1. Initial Conditions
1.1 Review the Limits and Precautions.
1.2 Verify Radiation Protection has taken a sample and has issued a new GWR Permit
Report.
2. Procedure
2.1 Request the CR SRO sign, date and time the appropriate sheet o f the Gaseous Waste
Release (GWR) Record authorizing releases.
be started and stopped as often as desired as long as the GWR Permit Report remains
valid. Subsequent steps in this section will terminate the GWR Permit Report when
2.2 The following steps are to aid the operator in temlinating the GWR Permit Repoa:
2.2.1 Add all values in the "VQ Volume" column of the appropriate sheet o f the
Gaseous Waste Release (GWR) Record. Record total in "Total Volume
Released" blank.
2.2.2 Ensure the CR SRO performs the following:
2.2.2.1 Sign, date and time the "Termination o f GWR Release
Acknowledged by Shift Supervisor or Designee" blank of the
appropriate sheet of the Gaseous Waste Release (GWR) Record.
2.2.2.2 Place the completed VQ GWR Permit Report in the completed
release box.
Bank Question: $36.1 Answer: A
1 Pt(s) Unit 1recently shut down. Given the following events and conditions:
NCS temperature is 230°F.
Both trains of ND are operable with A train in senrice
KC loops A and B are operating.
CAPT is tagged for maintsnance.
- CA pump 1B and CA pump 1A are isolated for merentia1 pressure
testing.
- Engineering has reported that a test on the 1B KC heat exchanger showed
that fouling had reduced its heat transfer capability beiow the minimum
design value.
The plant is expected to remain in mode 4 for another 7 days fer special
testing.
What is the EARLIEST action required for the operators to take?
REFPRENCESSPROWDED: Tech Spec's 3.4.6,3.7.5,3.7.7 and Bases
A. Return at least one motor driven CA pump t~ operable status or
be in mode 5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B. Return at least one motor driven CA pump to operable status or
be in mode 5 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
C. Clean the 1B KC heat exchanger within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in mode 5
within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
D. Clean the 1B KC heat exchanger within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> o r be in mode 5
W h i n the newt 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />.
__
Distracter Analysis: One loop of KC is operable therefore one loop of ND
is operable. With both MDCA pumps isolated, the S/Gs do not have
a source of feedwater. Tech Spec 3.4.6 requires CA to be operable
for the NC loops to be operable. Action statement B of Tech S p c
3.4.6 applies.
A. Correct answer: Without m operable CA pump to supply S/Gs, the
NC loops are not operable in mode 4 (TS 3.4.6 bases). Action
statement B applies. Tech Spec 3.7.5 action D (do not change
modes) does not apply because the plant is not in modes 1-3.
B. Incorrect: This does not restore CA and meet Tech Spec 3.4.6
Plausible: Meets the Tech Spec 3.7.5 action B for one train of CA
Ques-936.1 .doc
C. Incorrect: must be in mode 5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Plausible: This is the correct answer for TS 3.7.7 action R.
D. Incorrect: must be in mode 5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Plausible: if the candidate adds the time allowed to reach mode 3 (6
hrs) to the mode 5 action time (36 hrs) in TS 3.7.7 action B.
KA: G2.1.33 (3.4'4.0)
Lesson Plan Objective: PSS-KC Obj: 13
Source: Mod Ques-936 NRC McGuire 2002
Level of knowledge: comprehension
References:
1. Tech Spec 3.4.6 and bases - PROVIDED
2.Tech Spec3.7.5 andBases-PROVIDED
3. Tech Spec 3.7.9 and bases - PROVIDED
- ... DUKE POWER .. .... - ....-...... - CATAWBA
._ -_. .....
OPERATIONS TRAlNlNG
...
I
Objective s
1
1 State the Durgose of the KC Svstem. l x
2 I Describe how the KC Svstem is cooled. I x
1
3 Describe the normal flowpath of the KC System, including
each header and the t w e of loads serviced by each.
X
- Safety Injection (Ss)
- Phase A Containment Isolation (St)
Phase B Containment Isolation (Sp)
I
6 State the tvpical values of the KC pump discharge I x
I pressure, KC Hx outlet temperature and KC pump flow.
of the KC system per the KC
concernincs the KC svstem.
System and what is indicated by a high level radiation
OB-Cff-PSS-KC FOR TRAINING PURPOSES ONLY REV. 42
Page 3 of 26
RCS LOOPS- MODES 4
3.4.6
3.4 REACTOR COOLANT SYSTEM (RCS)
3.4.6 RCS LOOPS -MODE 4
LCO 3.4.6 Two loops consisting of any combination of RCS loops and residual heat
removal (RHR) loops shall be OPERABLE, and one loop shall be in
operation.
NOTES-------------------------~~-----------~----------
1. All reactor coolant pumps (RCPs) and RHR pumps may be
de-energized for 5 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period provided:
a. No operations are permitted that would came reduction of the
b. Core outlet temperature is maintained at least 10°F below
saturation temperature.
2. No RCP shall be started with any RCS cold leg temperature 5 285°F
unless the secondary side water temperature of each steam
generator (SG) is 5 50°F above each of the RCS cold leg
temperatures.
_s_______________s____________ss__s_____--~----------------------~~--~--------------------~-
APPLICABILITY: MODE 4
ACTIONS
CONDITION REQUIRED ACTION COMPLETION TIME
A. One RCS loop A.1 Initiate action to restore a Immediately
OPERABLE. second loop to
OPERABLE status.
ANB
Two RHR loops
. 1
(continued)
Catawba Units 4 and 2 3.4.6-1 Amendment Nos. 173165
RCS LOOPS - MODES 4
3.4.6
ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME
B. One RHR loop 3.1 Be in MODE 5 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
&NJ
ALL RCS loops
I
C. Both required RCS or C.l Suspend all operations lmmediateiy
RHR loops inoperable. involving a reduction of
-
QR
operation. C.2 Initiate action to restore immediately
one loop to OPERABLE
status and operation.
_e
SR 3 4 6 2 Verify SG secondary side water levels are 2 12% narrow 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
range for required RCS loops
SR 3.4.6.3 Verify correct breaker alignment and indicated power are 7 days
available to the required pump that is not in operation.
Catawba Units 1 and 2 3.4.6-2 Amendment Nos. 173/165
RCS LooPs-MQDE 4
B 3.4.6
B 3.4 REACTOR COOLANT SYSTEM (RCS)
E3 3.4.6 RCS LooPs-MODE 4
BASES
BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of
decay heat and the transfer of this heat to either the steam generator
(SG)secondary side coolant or the component cooling water via the
residual heal removal (RHR] heat exchangers. The secondary function of
the reactor coolant is to act as a carrier for soluble neutron poison, boric
acid.
The reactor coolant is circulated through four RCS loops connected in
parallel to the reactor vessel, each loop containing an SG,a reactor
coolant pump (RCP), and appropriate flow, pressure, level, and
temperature instrumentation for control, protection, and indication. The
RCPs circulate the coolant through the reactor vessel and SGs at a
sufficient rate to ensure proper heat transfer and to prevent boric acid
stratification,
In MODE 4, either RCPs or RHR loops can be used to provide forced
clrculation. The intent of this LCO is to provide forced flow from at least
one RCP or one RHR loop for decay heat removal and transport. The
flow provided by one RCP loop or RHR loop is adequate for decay heat
removal. The other intent of this LCO is to require that two paths be
available to provide redundancy for decay heat removal.
APPLICABLE In MODE 4, RCS circulation is considered in the determination of the
SAFETY ANALYSES time available for mitigation of the accidental boron dilution event. The
RCS and RHR loops provide this circuiation.
RCS Loops-MODE 4 satisfy Criterion 4 of 10 CFR 50.36 (Ref. 1)
LCO The purpose of this LCO is to require that at least two loops be
OPERABLE in MODE 4 and that one of these loops be in operation. The
LCO allows the two loops that are required to be OPERABLE to consist of
any combination of RCS loops and RHR loops. Any one loop in operation
provides enough flow to remove the decay heat from the core with forced
circulation. An additional loop is required to be OPERABLE to provide
redundancy for heat removal.
Note 1 permits all RCPs or RHR pumps to be deenergized for 5 'Ihour
per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period. The purpose of the Note is to permit tests that are
designed to validate various accident analyses values. One of the tests
Catawba Units 1 and 2 3 3.4.6-1 Revision No. 0
RCS Loops - MODE 4
B 4.4.6
BASES
LCO (continued)
performed during the startup testing program is the validation of rod drop
times during cold conditions, both with and without flow. The no flow test
may be performed in MODE 3 , 4 , or 5 and requires that the pumps be
stopped for a short period of time. The Note permits the de-energizing of
the pumps in order to perform this test and validate the assumed analysis
values. If changes are made to the RCS that wouM cause a change to
the flow characteristics of the RCS, the input values must be revalidated
by conducting the test again. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time period is adequate to
perform the test, and operating experience has shown that boron
stratification is not a problem during this short period with no forced flow.
Utilization of Note i is permitted provided the following conditions are met
along with any other conditions imposed by initial startup test procedures:
a. No operations are permitted that would dilute the RCS boron
concentration, therefore maintaining the margin to criticality. Boron
reduction is prohibited because a uniform concentration distribution
throughout the RCS cannot be ensured when in natural circulation;
and
b. Core outlet temperature is maintained at least I O T below
saturation temperature, so that no vapor bubble may form and
possibly cause a natural circulation flow obstruction.
Note 2 requires that the secondary side water temperature of each SG be
s 50°F above each of the RCS cold leg temperatures before the start of
an RGP with any RCS cold leg temperature s 285°F. This restraint is to
prevent a low temperature overpressure event due to a thermal transient
when an RCP is started.
An OPERABLE RCS loop comprises an OPERABLE RCP and an
OPERABLE SG in accordance with the Steam Generator Tube
Surveillance Program, which has the minimum water level specified in
SR 3.4.6.2.The water levei is maintained by an OPERABLE AFW train in
accordance with LCO 3.7.5,"Auxiliary Feedwater System."
Similarly for the RHR System, an OPERABLE RHR loop comprises an
OPERABLE RHR pump capable of providing forced flow to an
OPERABLE RHR heat exchanger. RCPs and RHR pumps are
OPERABLE if they are capable of being powered and are able to provide
forced flow if required.
Catawba Units 1 and 2 B 3.4.6-2 Revision No. 0
RCS LOOPS - MODE 4
B 3.4.6
BASES
APPLICABILITY In MODE 4, this LCO ensures forced circulation of the reactor coolant to
remove decay heat from the core and to provide proper boron mixing.
One loop of either RCS or RHR provides sufficient circulation for these
purposes. However, two loops consisting of any combination of RCS and
RHR loops are required to be OPERABLE to meet single failure
considerations.
Operation in other MODES is covered by:
LCO 3.4.4, "RCS Loops-MODES Z and 2";
LCO 3.4.5, "RCS Loops-MODE 3";
LCO 3.4.7, "RCS Loops-MQDE 5,hoops Filled";
LCO 3.4.8, "RCS Loops-MODE 5,Loops Not Filled";
LCO 3.4.1 7, "RCS Loops-Test Exceptions";
LCO 3.9.4, "Residual Heat Removal (RHR)and Coolant
Circulation-High Water Level" (MODE 6); and
LCO 3.9.5. "Residual Heat Removal (RHR) and Coolant
Circulation-Low Water Level" (MODE6).
ACTIONS A.1
If only one RCS loop is OPEMBLE and two RHR loops are inoperable,
redundancy for heat removal is lost. Action must be initiated to restofe a
second RCS OF RHR loop to OPERABLE status. The immediate
Completion Time reflects the importance of maintaining the availability of
two paths for heat removal.
If only one RHR loop is QPERABLE and in operation and there are no
RCS loops OPERABLE, an inoperable RCS or RHR loop must be
restored to OPERABLE status to provide a redundant means for decay
heat removai.
If the parameters that are outside the limits cannot be restored, the unit
must be brought to MODE 5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Bringing the unit to
MODE 5 is a conservative action with regard to decay heat removal. With
only one RHR loop OPERABLE, redundancy for decay heat removal is
lost and, in the event of a loss of the remaining RHR loop, it would be
safer to initiate that loss from MODE 5 (< 200°F) rather than MODE 4
(200 to < 350OF). The Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is a reasonable time,
based on operating experience, to reach MODE 5 from MODE 4 in an
orderly manner and without challenging plant systems.
Catawba Units 1 and 2 B 3.4.6-3 Revision No. 0
RCS LOOPS- MOB 4
E 3.4.6
BASES
ACTIONS (continued)
C.l and C.2
If no loop is OPERABLE or in operation, except during conditions
permitted by Note 1 in the LCO section, all operations invoiving a
reduction of RCS boron concentration must be suspended and action to
restore one WCS or RHR loop to OPERABLE status and operation must
be initiated. RCP seal injection flow is not considered to be an operation
involving a reduction in RCS boron concentration. Boron dilution requires
forced circulation for proper mixing, and the margin to criticality must not
be reduced in this type of operation. The immediate Completion Times
reflect the importance of maintaining operation for decay heat removal.
The action to restore must be continued untll one loop is restored to
OPERABLE status and operation.
SURVEILLANCE SR 3.4.61
REQUIREMENTS
This SR requires verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that one RCS or RHR loop is
in operation. Verification includes flow rate, temperature, or pump status
monitoring, which help ensure that forced flow 1s providing heat removal.
The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient considering other indications and
alarms available to the operator in the control room to monitor RCS and
RHR loop performance.
SW 3.4.6.2
SR 3.4.6.2requires verification of SG OPEBAEILITY. SG OPERABILITY
is verified by ensuring that the secondary side narrow range water level is
2 12%. If the SG secondary side narrow range water level is c 12%, the
tubes may become uncovered and the associated loop may not be
capable of providing the heat sink necessary for removal of decay heat.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered adequate in view of other
indications available in the control room to alert the operator to the loss of
SG level.
Verification that the required pump is OPERABLE ensures that an
additional RCS or RHR pump can be placed in operation, if needed, to
maintain decay heat removal and reactor coolant circulation. Verification
is performed by verifying proper breaker alignment and power available to
the required pump. The Frequency of 7 days is considered reasonable in
view of other administrative controls available and has been shown to be
Catawba Units I and 2 B 3.4.6-4 Revision No. 0
RCS LOOPS - MODE 4
B 3.4.6
BASES
acceptable by operating experience.
Catawba Units 1 and 2 B 3.4.6-5 Revision No. 0
RCS LOOPS - MODE. 4
B 3.4.6
REFERENCES 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.4.6-6 Revision No. 0
AFW System
3.7.5
3.7 PLANT SYSTEMS
3.7.5 Auxiliary Fedwater (AFW) System
APPLICABILITY: MQDES I,2, and 3,
MODE 4 when steam generator is relied upon for heat removal.
ACTIONS
CONDI%[QN REQUIRED ACTION COMPLETIQM TIME
A. Qne steam supply to A.l Restore steam supply to 7 days
turbine driven AFW QPERA%LEstatus.
pump inoperable.
10 days from
discovery of
failure to
meet the LCO
B. One AFW train B.l Restore AFW train to 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s: I
inoperable in MODE 1 , 2 OPERABLE status.
or 3 for reasons other
than Condition A.
10 days: from I
discovery of
failure to
meet the LCO
(conthued)
-each A F \ a a i n on Unit 2. the Com&tion Time that one A F \ n r t r a i n m bempetahleas3PeGifiedpy &Lluir.&'
'For ~ ~
~
Action Aa I.y be extended beyond the "72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 10 davs froAdis!z!!Leryofailureto m~eAthe&O:mo 288
hours a a a t d h e N4!S m t e n u p g r a d e s . S y s t e m . u ~ ~ d e s ~ c l u d e m . a ~ n t e n a n . c emdificati0.n~
and advitles
associatedwithae NSWS e i ~ ~ n g , v a l v e ~ a n d ~ n ~ ~ i n e ~ n . e c e ~ a ~ ~ ~ a i r I&acement ~~~d!o~r@~ia~e~
ofeonions of the NSWS Oiping to&e~AFW s v s t e m , ~ o n x m p M o n o f h e c&nine_upgra&s,and SYSW
~- nolongemx?k&b!e
r e s t o r a m in refuel~imoutw I~ . C l . 2 h i ~ s f o o t n n ~ i.s
Catawba Units 1 and 2 3.7.5-1 Amendment Nos. 1891182
AFW System
3.7.5
ACTIONS (continued)
CONDiTlON REQUIRED ACTION COMPLETION TIME
6. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
associated Completion
Time for Condition A y
d
J
or E! not met.
C.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
-
Two AFW trains
inoperable in MODE 1,
2, or 3.
D. Three AFW trains D. ------------NOTE-------------
inoperable in MODE 1, LCO 3.0.3 and all other
2, or 3. LCO Required Actions
requiring MODE changes
are suspended until
one AFW train is restored
to OPERABLE status.
..................................
Initiate action to restore Immediately
one AFW train to
OPERABLE status.
E. Required AFW train =.IInitiate action to restore Immediately
inoperable in MODE 4. AFW train to OPERABLE
status.
Catawba Units 1 and 2 3.7.5-2 Amendment Nos. 1731165
AFW System
3.9.5
SURVEILLANCE REQUIREMENTS
SL6RVElbLANCE FREQUENCY
SR 3.7.5.1 ____________~ __s______l__ _______
s__-_ NOTE____----__ ---__
_s__s_ s__s~--
Not applicable to automatic valves when THERMAL
POWER is 5 10% RTP.
_____________________________s__s_______~--~~----------------------------
yr
Verify each AFW manual, power operated, and automatic 31 days
valve in each water flow path, and in both steam supply
flow paths tu the steam turbine driven pump, that is not
locked, sealed, or otherwise secured in position, is in the
correct position.
SR 3.7.5.2 ------------------------------------NOTE--~~---------~--------------
Not required to be performed for the turbine driven AFW
pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after 2 600 p i g in the steam
generator.
Verify the developed head of each AFW pump at the flow In accordance
test point is greater than or equal to the required with the Inservice
deveioped head. Besting Program
98 months
(continued)
Catawba Units 1 and 2 3.9.5-3 Amendment Nos. 1731165
AFW System
3.7.5
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY
SB 3.7.5.4 ____- s__~___s____ NOTES--_____--____---
_-_-_
s_---
3. Not required to be performed for the turbine
driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after Z 600 psig
in the steam generator.
2. Not applicable in MODE 4 when steam generator
is relied upon for heat removal.
Verify each AFW pump starts automatically on an actual
or simulated actuation signal. 18 months
SR 3.7.5.5 Verify proper aiignment of the required AFW flow paths Prior to entering
by verifying flow from the condensate storage system to MODE 2,
each steam generator. whenever unit
has been in
MODE 5 or 6 for
> 30 days
Catawba Units 1 and 2 3.7.5-4 Amendment NOS. 1941165
AFW System
3 3.7.5
B 3.7 PLANT SYSTEMS
B 3.7.5 Auxiliary Feedwater (AFW) System
BASES
BACKGROUND The AFW System automatically supplies feedwater to the steam
generators to remove decay heat from the Reactor Coolant System upon
the loss of normal feedwater supply. The AFW pumps take suction
through suction lines from the condensate storage system (CSS)
(LCO 3.7.6) and pump to the steam generator secondary side. The
normal supply of water to the AFW pumps is from the condensate
system. The supply valves are open with power removed from the valve
operator. The assured source of water to the AFW System is supplied by
the Nuclear Service Water System. The turbine and motor driven pump
discharge lines to each individual steam generator join into a single line
outside containment. These individual lines penetrate the containment
and enter each steam generator through the auxiliary feedwater nozzle.
The steam generators function as a heat sink for core decay heat. The
heat load is dissipated by releasing steam to the atmosphere from the
steam generators via the main steam safety valves (MSSVs) (kCO 3.7.1)
or SG PQRVs (LCO 3.7.4). If the main condenser is available, steam
may be released via the steam dump valves and recirculated to the
hotwell.
The AFW System consists of two motor driven AFW pumps and one
steam turbine driven pump configured into three trains. Each of the motor
driven pumps supply 155% of the flow requirements to two steam
generators, although each pump has the capability fo be realigned to feed
other steam generators. The turbine driven pump provides 250% of the
flow requirements and supplies water to all four steam generators. Travel
stops are set on the steam generator flow control valves such that the
pumps can supply the minimum flow required without exceeding the
maximum flow allowed. The pumps are equipped with independent
recirculation lines to prevent pump operation against a closed system.
Each motor driven AFW pump is powered from an independent Class 1E
power supply. The steam turbine driven AFW pump receives steam from
two main steam lines upstream of the main steam isolation valves. Each
of the steam feed lines will suppiy 100% of the requirements of the
turbine driven AFW pump.
~~
Catawba Units 1 and 2 B 3.7.5-1 Revision No. 0
AFW System
B 3.7.5
BACKGROUND (continued)
The AFW System is capable of supplying feedwater to the steam
generators during normal unit startup, shutdown, and hot standby
conditions. One turbine driven pump at full flow is sufficient to remove
decay heat and cool the unit to residual heat removal (RHR) entry
conditions. During unit cooldown, SG pressures and Main Steam
pressures decrease simultaneously. Thus, the turbine driven AFW pump
with a reduced steam supply pressure remains fully capable of providing
flow to all SGs. Thus, the requirement for diversity in motive power
sources for the AFW System is met.
The AFW System is designed to supply sufficient water to the steam
generalor(s) to remove decay heat with steam generator pressure at the
iowest setpoint of the MSSVs plus 3% accumulation. Subsequently, the
AFW System supplies sufficient water to cool the unit to RHR entry
conditions, with steam released through the SG PORVs or MSSVs.
The motor driven AFW pumps actuate automatically on steam generator
water level low-low in 1 out of 4 steam generators by the ESFAS
( K O 3.3.2).The motor driven pumps also actuate on loss of offsite
power, safety injection, and trip of all MFW pumps. The turbine driven
AFW pump actuates automatically on steam generator water level iow-
low in 2 out of 4 steam generators and on loss of offsite power.
The AFW System is discussed in the UFSAR, Section 10.4.9 (Ref. I].
APPLICABLE The AFW System mitigates the consequences of any event with loss
SAFETY ANALYSES of normal feedwater.
The design basis of the AFW System is to supply water to the SkaM
generator to remove decay heat and other residual heat by delivering at
least the minlmurn required flow rate to the steam generators at
pressures corresponding to the lowest steam generator safety vaive set
pressure plus 3%.
In addition, the AFW System must supply enQUghmakeup water to
replace steam generator secondary inventory lost as the unit coois to
MODE 4 conditions. Sufficient AFW flow must also be available to
account for flow losses such as pump recirculation valve leakage and line
breaks.
The limiting Design Basis Accidents (DBAs) and transients fer the AFW
System are as follows:
Catawba Units 1 and 2 B 3.7.5-2 Revision No. 0
AFW System
B 3.7.5
APPLICABLE SAFETY ANALYSES (continued)
a. Feedwater Line Break (FWLB); and
b. Loss of MFW.
In addition, the minimum available AFW flow and system characteristics
are Considered in the analysis of a Small break loss of coolant accident
(LOCA) and events that could lead to steam generator tube bundle
uncovery for dose considerations.
A range of AFW flows is considered for the analyzed accidents, with the
Main Steam Line Break being the most limiting for the maximum AFW
flowrate.
The AFW System design is such that it can perform its function following
a FWLB between the steam generator and the downstream check valve,
combined with a loss of offsite power following turbine trip, and a single
active failure of the steam turbine driven AFW pump. in such a case, one
motor driven AFW pump would deliver to the broken MFW header at the
pump runout flew until the problem was detected, and flow terminated by
the operator. Sufficient flow would be delivered to the intact steam
generators by the redundant AFW pump.
The ESFAS automatically actuates the AFW turbine driven pump and
associated power operated valves and controls when required to ensure
an adequate feedwater supply to the steam generators during loss of
offsite power.
The AFW System satisfies the requirements of Criterion 3 of 20 CFR
50.36 (Ref. 2).
LCO This LCO provides assurance that the AFW System will perform its
design safety function to mitigate the consequences of accidents that
could result in overpressurization of the reactor coolant pressure
boundary. Three independent AFW pumps in three diverse trains are
required to be OPERABLE to ensure the availability of RHR capability far
all events accompanied by a loss of offsite power and a single failure.
This is accomplished by powering two of the pumps from independent
emergency buses. The third AFW pump is powered by a different means,
a steam driven turbine supplied with steam from a source that is not
isolated by closure of the MSBVs.
Catawba Units 1 and 2 B 3.7.5-3 Revision No. 0
AFW System
B 3.7.5
BASES
K O (continued)
The A W System is configured into three trains. The AFW System is
considered OPEFPABLE when the components and flow paths required to
provide redundant AFW flow to the steam generators are OPERABLE.
This requires that the two motor driven AFW pumps be OPERABLE in
two diverse paths, each supplying AFW to separate steam generators.
The turbine driven AFW pump is required to be OPERABLE with
redundant steam supplies from two main steam lines upstream of the
MSIVs, and shall be capable of supplying AFW to any of the steam
generators. The piping, valves, instrumentation, and controls in the
required flow paths also are required to be OPERABLE. T M S A S
a s s u r e d m d !&laktsupply~iscMfiguredwvi- etubine
driu?nApumP recei\LesM!&Sfrnm hatb_trainsnthlSmh&rK
the&ss.afmetminafassuredsame r&ersnuIylmeAEWtPain
n~aPERARLEassurd
i r S a p e r a h L T h r e m a'njng~hlslnLstcainpra\Lidesa
saurcEFhthe~atbEiMnotmdmenpump xdth&rhinedriwt~puq~
The L e 0 is modified by a Note indicating that one AFW train, which
includes a motor driven pump, is required to be OPERABLE in MODE 4.
This is because of the reduced heat removal requirements and short
period of time in MODE 4 during which the AFW is required and the
insufficient steam available in MODE 4 to power the turbine driven AFW
pump.
APPLICABILITY In MODES 1, 2, and 3, the AFW System is required to be OPERABLE in
the event that it is called upon to function when the MFW is lost. In
addition, the AFW System is required to supply enough makeup water to
repiace the steam generator SeCQRdaiy inventory, lost as the unit cools to
MODE 4 conditions.
In MODE 4 the AFW System may be used for heat removal via the steam
generators.
In MODE 5 or 6, the steam generators are not normally used for heat
removal, and the AFW System is not required.
ACTiONS &1
If one of the two steam supplies to the turbine driven AFW train is
inoperable, action must be taken to restore OPERABLE status within
7 days. The 7 day Completion Time is reasonable, based on the
following reasons:
Catawba Units 1 and 2 B 3.7.5-4 Revision No. i
AFW System
B 3.7.5
BASES
ACTIONS (continued)
a. The redundant OPERABLE steam supply to the turbine driven AFW
pump;
b. The availability of redundant OPERABLE motor driven AFW
pumps; and
c. The low probability of an event occurring that requires the
inoperable steam supply to the turbine driven AFW pump.
The second Completion Time for Required Action A.1 establishes a limit
on the maximum time allowed for any combination of Conditions to be
inoperable during any continuous failure to meet this LCO.
The 10 day Completion Time provides a limitation time allowed in this
specified Condition after discovery of failure to meet the LCO. This limit
is considered reasonable for situations in which Conditions A and B are
entered concurrently. The Bhill connector between 7 days and 10 days
dictates that both Completion Times apply simultaneously, and the more
restrictive must be met.
&1
With one of the required AFW trains (pump or flow path) inoperable in
MODE 1, 2, or 3 for reasons ether than Condition A, action must be taken
to restore OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This Condition includes the
loss of two steam supply lines to the turbine driven AFW pump. The
72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on redundant capabilities
afforded by the AFW System, time needed for repairs, and the low
probability of a DBA occurring during this time period.
The second Completion Time for Required Action B.l establishes a limit
on the maximum time allowed for any combination of Conditions to be
inoperable during any continuous failure to meet this LCO.
The 10 day Completion Time provides a iimitation time allowed in this
specified Condition after discovery of failure to meet the LCO. This limit
is considered reasonable for situations in which Conditions A and B are
entered concurrently. The AND connector between 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 10 days
dictates that both Completion Times apply simultaneously, and the more
restrictive must be met.
Catawba Units 1 and 2 B 3.7.5-5 Revision No. 0
AFW System
B 3.7.5
BASES
ACTIONS (continued)
c l
a n d u
When Required Action A.1 or B.Z cannot be completed withln the
required Completion Time, or if two AFW trains are inoperable in
MODE 1, 2, or 3, the unit must be placed in a MODE in which the LCO
does not apply. To achieve this status, the unit must be placed in at least
MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The allowed Completion Times are reasonable, based on operating
experience, to reach the required unit conditions from full power
conditions in an orderly manner and without challenging unit systems.
In MODE 4 with two AFW trains inoperable, operation is allowed to
continue because only one motor driven pump AFW train is required in
accordance with the Note that modifies the LCO. Although not required,
the unit may continue to cool down and initiate RHR.
1Li
If all three AFW trains are inoperable in MODE 1, 2, or 3, the unit is in a
seriously degraded condition with no safety related means for conducting
a cooldown, and only limited means for conducting a coddown with
nonsafely related equipment. In such a condition, the unit should not be
perturbed by any action, including a power change, that might result in a
trip. The seriousness of this condition requires that action be started
immediateiy to restore one AFW train to OPERABLE status.
Required Action B.1 is modified by a Note indicating that all required
MODE changes or power reductions are suspended until one AFW train
is restored to OPERABLE status. In this case, LCO 3.0.3 is not
applicable because it could force the unit into a less safe condition.
&1
In MODE 4, either the reactor coolant pumps or the RHR loops can be
used to provide forced circulation. This is addressed in LCO 3.4.6, "RCS
Loops-MODE 4." With one required AFW train inoperable, action must
be taken to immediately restore the inoperable train to OPERABLE
status. The immediate Completion Time is consistent with LCO 3.4.6.
Catawba Units 1 and 2 B 3.9.5-6 Revision No. 0
AFW System
B 3.7.5
SURVEILLANCE SEZ-KEd
WEQULREMENTS
Verifying the correct alignment for manual, power operated, and
automatic valves in the AFW System water and steam supply flow paths
provides assurance that the proper flow paths will exist for AFW
operation. This SR does not apply to valves that are locked, sealed, or
otherwise secured in posltion, since they are verified to be in the correct
position prior to locking, sealing, or securing. This SR also does not
apply to valves that cannot be inadvertently misaligned, such as check
valves. This Surveillance does not require any testing or valve
manipulation; rather, it involves verification that those valves capabie of
being mispositioned are in the correct position. The St3 is also modified
by a note that excludes automatic valves when THERMAL POWER is
5 10% RTP. Some automatic valves may be In a throttled position to
support low power operation.
The 31 day Frequency is based on engineering judgment, is consistent
with the procedural controls governing valve operation, and ensures
correct valve positions.
x5RLU52
Verifying that each AFW pump's developed head at the flow test point is
greater than or equal to the required developed head ensures that AFW
pump performance has not degraded during the cycle. Flow and
differential head are normal tests of centrifugal pump performance
required by Section XI of the ASME Code (Ref. 3). Because it is
undesirable to introduce cold AFW into the steam generators while they
are operating, this testing is performed on recirculation flow. This test
confirms one point on the pump design curve and is indicative of overall
performance. Such inservice tests confirm component OPERABILITY.
trend performance, and detect incipient failures by indicating abnormal
performance. Performance of Inservice testing discussed in the ASME
Code,Section XI (Ref. 3) (only required at 3 month intervals) satisfies this
requirement.
This SR is modified by a Note indicating that the SR should be deferred
until suitable test conditions are established. This deferral is required
because there is insufficlent steam pressure to perform the test.
Catawba Units 1 and 2 B 3.7.5-7 Revision No. 0
AFW System
B 3.7.5
SURVEILLANCE REQUIREMENTS (continued)
sL3.759
Phis SR verifies that AFW can be delivered to the appropriate steam
generator in the event of any accident or transient that generates an
ESFAS, by demonstrating that each automatic valve in the flow path
actuates to its correct position on an actual or simulated actuation signal.
This Surveillance is not required for valves that are locked, sealed, or
othewise secured in the required position under administrative controls.
The 18 month Frequency is based on the need to perform this
Surveillance under the conditions that apply during a unit outage and the
potential for an unplanned transient if the Surveillance were performed
with the reactor at power. The 18 month Frequency is acceptable based
on operating experience and the design reliability of the equipment.
Phis SR is modified by a Note that states the SF? is not required in MODE
4. In MODE 4, the required A f W train may already be aligned and
operating.
SaEu5.4
This SR verifies that the AFW pumps will start in the event of any
accident or transient that generates an ESFAS by demonstrating that
each AFW pump starts automatically on an actual or simulated actuation
signal in MBBES 2 , 2, and 3. In MODE 4, the required pump may
already be operating and the autostart function is not required. The
18 month Frequency is based on the need to perform this Surveillance
under the conditions that apply during a unit outage and the potential for
an unplanned transient if the Surveillance were performed with the
reactor at power.
This SR is modified by two Notes. Note 1 indicates that the SR can be
deferred until suitable test conditions are established. This deferral is
required because there is insufficient steam pressure to perform the test.
Note 2 states that the SR is not required in MODE 4. In MODE 4, the
required pump may already be operating and the autostart function is not
required. In MODE 4, the heat removal requirements would be less
providing more time for operator action to manually start the required
AFW PUMP if It were not in operation.
Catawba Units 1 and 2 B 3.7.5-8 Revision No. 0
AFW System
0 3.7.5
SURVEILLANCE REQUIREMENTS (continued)
sE32235
This SR verifies that the AFW is properly aligned by verifying the flow
paths from the CSS to each steam generator prior to entering MODE 2
after more than 30 days in MODE 5 or 6. OQEfWBlblTY of AFW flow
paths must be verified before sufficient core heat is generated that would
require the operation of the AFW System during a subsequent shutdown.
The Frequency is reasonable, based on engineering judgement and other
administrative controls that ensure that flow paths remain OPERABLE.
To further ensure AFW System alignment, flow path OPERABILITY is
verified following extended outages to determine ne misalignment of
valves has occurred. This SR ensures that the flow path from the CSS to
the steam generators is properly aligned.
~ ~
REFERENCES 1. UFSAR, Section 10.4'9.
2. 10 CFR 50.36, Technicai Specifications, (c)(2)(ii).
3. ASME, Boiler and Pressure Vessel Code,Section XI.
Catawba Units 1 and 2 B 3.7.5-9 Revision No. 0
cew System
3.7.7
3.7 PLANT SYSTEMS
3.7.7 Component Cooling Water (CCW) System
LCO 3.7.7 Two CCW trains shall be OPERBBLE.:
APPLICABILITY: MODES 1. 2, 3, and 4
ACTIONS
CONDITION REQUIRED ACTION 1 COMPLETION TIME
A. One CCW train A.1 _______
________-----NOTE------
inooerable. Enter applicable
Conditions and Requit-ed
Actions of LCO 3.4.6,
"RCS LOOPS-MODE 4,"
loops made inoperable by
ccw.
Restore CCW train to 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s:
OPERABLE status.
B. Required Action and 3.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
associated Completion
Time of Condition A not AND
met.
8.2 Be in MODE 5. i 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
Catawba Units 1 and 2 3.7.7-1 Amendment Nos. 1891282
CCW System
B 3.7.7
B 3.7 PLANT SYSTEMS
B 3.7.7 Component Cooling Water (CCW) System
BASES
- -
BACKGROUND The CCW System provides a heat sink for the removal of process and
operating heat from safety related components during a Design Basis
Accident (BBA) or transient. During normal operation, the CCW System
also provides this function for various nonessential components, as well
as the spent fuel storage pool. The CCW System serves as a barrier to
the release of radioactive byproducts between potentially radioactive
systems and the Nuclear Service Water System (NSWS), and thus to the
environment.
The CCW System is arranged as two independent, full capacity cooling
loops, and has isolatable nonsafety related components. Each safety
related train includes two 50% capacity pumps, surge tank, heat
exchanger, piping, valves: and instrumentation. Each safety related train
is powered from a separate bus. An open surge tank in the system
provides sufficient inventory to protect the pumps from a [ack of net
positive suction head available (NPSHA) due to a moderate energy line
break. The pumps have sufficient NPSHA with the surge tank empty
provided the piping up to the tank is filled. The pumps on each train are
automatically started on receipt of a safety injection signal, and all
nonessential components are isolated.
Additional information on the design and operation of the system, along
with a list of the components served, is presented in the UFSAR,
Section 9.2 (Ref. 1). The principal safety related function of the CCW
System is the removal of decay heat from the reactor via the Residual
Heat Removal (RHW) System. This may be during a normal or post
accident cooldown and shutdown.
APPLICABLE The safety related design basis function of the CCW System is to remove
SAFETY ANALYSES waste heat from various components essential in mitigating design basis
events which require Emergency Core Cooling System (ECCS)
operation. The CCW System is also used to support normal operation.
The normal temperature of the CCW is 9QF, and, during unit cooldown to
MODE 5 (TmIde 20O0F),a maximum temperature of 128°F is
Catawba Units 1 and 2 B 3.7.7-1 Revision No. 0
CCW System
B 3.7.7
BASES
APPLICABLE SAFETY ANALYSES (continued)
assumed (Ref. I). This 120°F limit is to prevent thermal degradation of
the large pump motors supplied with cooling water from the CCW
System.
The CCW System is designed to perform its function with a single failure
of any active component, assuming a loss ob offsite power.
The CCW System also functions to cool the unit from RHR entry
conditions (Teald 350°F), to MODE 5 (Icod l a 2OO0F),during normal and
post accident operations. The time required to cool from 350°F to 200.F
is a function of the number of CCW and RHR trains operating. One CCW
train is sufficient to remove decay heat during subsequent operations with
TWlaa 200°F. This assumes a maximum service water temperature of
100°F occurring simultaneousiy with the maximum heat loads on the
system.
The CCW System satisfies Criterion 3 of 10 CFR 50.36 (Ref. 2).
LCO The CCW trains are independent of each other to the degree that each
has separate controls and power supplies and the operation of one does
not depend on the other. In the event of a DBA. one CCW train is
required to provide the minimum heat removal capability assumed in the
safety anabysis for the systems to which it supplies cooling water. To
ensure this requirement is met, two trains of CCW must be OPEFWBLE.
At least one CCW train will operate assuming the worst case single active
failure occurs coincident with a loss of offsite power.
A CCW train is considered OPERABLE when:
a. Both pumps and associated surge tank are OPERABLE; and
b. The associated piping, valves, heat exchanger, and instrumentation
and controls required to perform the safety related function are
The isolation of CCW from other components or systems not required for
safety may render those components or systems inopsrable but does not
affect the OPERABILITY of the CCW System.
Catawba Units 1 and 2 B 3.7.7-2 Revision No. 0
CCW System
B 3.7.7
BASES
APPLICAB ILlTY In MODES 1,2,3, and 4,the CCW System is a normally operating
system, which must be prepared to perform its post accident safety
functions, primarily RCS heat removal, which is achieved by cooling the
RHR heat exchanger.
in MODE 5 or 6, the requirements of the CCW System are determined by
the systems it supports.
Required Action A.l is modified by a Note indicating that the applicable
Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4,"
be entered if an inoperable CCW train results in an inoperable RHR loop.
This is an exception to LCO 3.0.6 and ensures the proper actions are
taken for these components.
If one CCW train is inoperable, action must be taken to restore
OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining
OPERABLE CCW train is adequate to perform the heat removal function.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on the redundant
capabilities afforded by the OPERABLE train, and the low probability of a
DBA occurring during this period.
BLamiR2
If the CCW train cannot be restored to OPERABLE status within the
associated Completion Time, the unit must be placed in a MODE in which
the LCO does not apply. To achieve this status, the unit must be placed
in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The
allowed Completion Times are reasonable, based an operating
experience, to reach the required unit conditions from full power
conditions in an orderBy manner and without challenging unit systems.
SURVEILLANCE SR2ZZ.l
REQUIREMENTS
This SR is modified by a Note indicating that the isolation of the CCW
flow to individual components may render those components inoperable
but does not affect the OPERABILITY of the CCW System.
Verifying the correct alignment for manual, power operated, and
automatic valves in the CCW flow path to safety related equipment
provides assurance that the proper flow paths exist for CCW operation.
Catawba Units 1 and 2 B 3.7.7-3 Revision No. 0
CCW System
r3 3.7.7
BASES
SURVEILLANCE REQUIREMENTS (continued)
This SR does not apply to valves that are locked, sealed, or otherwise
secured in position, since these valves are verified to be in the correct
position prior to iocking, sealing, or securing. This SR also does not
apply to valves that cannot be inadvertently misaligned, such as check
valves. This Surveillance does not require any testing or valve
manipulation; rather, it involves verification that those valves capabie of
being mispositioned are in the correct position.
The 31 day Frequency is based on engineering judgment, is consistent
with the procedural controls governing valve operation, and ensures
correct valve positions.
sB3LL2
This SR verifies proper automatic operation of the CCW valves on an
actual or simulated actuation safety injection, Phase A isolation, or
Phase 5Isolation signal. The CCW System is a normally operating
system that cannot be fully actuated as part of routine testing during
normal operation. This Surveillance is not required for valves that are
locked, sealed. or othewise secured in the required position under
administrative controls. The 18 month Frequency is based on the need to
perform this Surveillance under the conditions that apply during a unit
outage and the potential for an unplanned transient if the Surveillance
were performed with the reactor at power. Operating experience has
shown that these components usually pass the Surveillance when
performed at the 18 month Frequency. Therefore, the Frequency is
acceptable from a reliability standpoint.
sR?Lz23
Phis SR verifies proper automatic operation of the CCW pumps on an
actual or simulated actuation signal. The CCW System is a normally
operating system that cannot be fully actuated as part of routine testing
during normal operation. The 48 month Frequency is based on the need
to perform this Surveillance under the conditions that apply during a unit
outage and the potential for an unplanned transient if the Surveillance
were performed with the reactor at power. Operating experience has
shown that these components usually pass the Surveillance when
performed at the 18 month Frequency. Therefore, the Frequency is
acceptable from a reliability standpoint.
Catawba Units 1 and 2 B 3.7.7-4 Revision No. 0
CCW System
B 3.7.7
BASES
REFERENCES 1. UFSAR, Section 9.2.
2. 10 CFW 50.36, Technical Specifications, (c)(a)(ii).
Catawba Units 1 and 2 B 3.7.7-5 Revision No. 0
1 Pt(s) Unit 1 is shutting down, in mode 4. Engineering reported that a recent test on
the 1B KC heat exchanger showed that fouling had reduced its heat transfer
capability below the minimum design value. Given the following events and
conditions:
NCS tempertture is 260 O F .
ND loops A and B are operating.
CAPT is tagged for maintenance.
CA pump 1B and CA pump 1A are isolated for differential pressure
testing.
"he plant is expected to remain in mode 4 for another 7 days.
What is the EARLIEST action required for the operators to tnke?
REFEREArCESPROF7DED: Tech Spec's 3.4-6,3.%6
A. Be in mode 5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B. Return at least one CA pump to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
C. Clean the 1B KC heat exchanger within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> o r be in mode 5
within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
D. Clean the 1B KC heat exchanger within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in mode 5
within 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />.
Distracter Analysis:
A. Correct answer: TS 3.7.6 cascades to TS 3.4.6 (see note).
B. Incorrect: Any action on the CA system would have immediate
response required.
Plausible: Restoring a CA pump couki resolve the TS 3.4.6 action
but not the TS 3.7.6 action.
C. Incorrect: must be in mode 5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Plausible: This is the correct answer for TS 3.7.6, disrcganling the
cascading note.
D. Incorrect: must be in mode 5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Plausible: if the candidate adds the time allowed to reach mode 3 to
the mode 5 action time in TS 3.7.6.
Level: RO Only
KA: ADM G 2.2.22 (3.5 / 4.1)
Lesson Plan Objective: PSS-KC SEQ 15
Source: New
Level of knowledge: comprehension
References:
1. Tech Spec's 3.4.6 and 3.7.6 - PROVIDED
. . ,. ,. .. . .
.......
Bank Question: 718 Answer: B
I Pt(s) Unit 2 is conducting a plant shutdown fkom 100% power. Given the following
events and conditions:
- Reactor power is 6%
All manual actions have been taken as required in the procedures
Intemiediate Range channel N-36 fails HIGH.
Which ofthe following statements correctly describes how this failure affects the
reactor shutdown and subsequent operation of the Nuclear Instrumentation System?
A. The reactor will trip; the source range detectors will reenergize when N-35
decreases to the proper setpoint.
B. The reactor will trip; the source range detectors will have to be manually
reenergized.
C. The reactor will not trip; the source ranges will reenerghe when N-35
decreases to the proper setpoint.
D. The reactor WiIl not trip; the source ranges will have to be manually
reenergized.
Distracter Analysis:
A. Incorrect: The source range instruments will not automatically reenergize
Plausible: If the operator believes the remaining IR energizes the Source
Ranges, but knows the reactor trips.
B. Correct: The IR trip will occur when either IR channel increases to > 25%
equivalent. However this trip is blocked manually when P-10 is satisfied.
Since reactor power is given as 6%, when N-36 fails high, the IR high flux
trip will occur.
The Source Range instmnents will automatically reenergize when:
1. P-10 is not satisfied, 3/4 NIS PW 10% and
2. P-6, both IR < 10." amps
Othenvise they will need to be MANUALLY reenergized. With a high
failure of IR N-36, they will not automatically reenergize.
C. Incorrect The reactor will trip and the source range instruments will not
automatically reenergize
Plausible: Operator believes the IR trip is blocked and only one IR is
necessary bo energize the SR
D. Incorrect: The reactor will trip and the source range instruments will not
automatically reenergize
Plausible: Operator believes the IR trip is blocked hut knows both LR are
necessary to energize the SR.
KA: APE 033 AA2.08 (3.3/3.4)
Lesson Plan Objective: ENB Qbj: 9
Source: bank
Level of knowledge: analysis
References:
1. OP-CN-IC-ENB page 10,11,12,13
Ques-718.doc
DUKE BOWER CATAWBA OPERATlONS TRAiNlhlG
OBJECTIVES
-
L
Objective P
S
I I
5 Explain the function of each portion of the individual ranges when
given a block diag%rarnof each range.
1 6 Explain the function of all indications and controls associated with
(ENB.
-.+I 1
9 Describe the plant response to a given detector or instrtirnent
failure.
-
X
OP-CN-IGENB MIR TRAINING PURPOSES ONLY REV. 29
Page 3 of 41
DUKE BOWER
= . - ..... - .-_ . ...
CA TAWBA OPERATONS TRAlNlNG
. . . . , I - _
-
.
.A .
.
- ..
15. Level Adiust Potentiometer - Adji!stable test signal into level amp.
Enables adjustment of the trip level of various bistables.
16. &ghJFlux at Shutdowrr Switch - Two position switch.
a) Normal -allows circuit to provide High Flux at Shdtdown and
Containment Evacuation alarm when setpoint is exceeded.
b) Block-used during startup - Blocks High Flux at Shutdown Alarm and
Containment Evacuation Alarm.
c) With the Coritrol Power fuses biown or rernoved the containment
evacuation alarm wil! alarm and cannot be BLOCKED. The OFF
switch on the main control board must be depressed and held until
the fkses can be replaced or reinstalied.
E. Source Range interlocks
1. During reactor startup as power increases into the Intermediate Range
and increases above amps (1 out of 2 channels) perrnissive P-6 is
energized ailtorriatically. This permissive allows the operator to block the
Sf3 High Flux Reactor Trip ( I O 5 cps increasing).
2. When reactor power is reduced to less than P-6 the Si3 detectors are
autoniaticaily energized and the SR tiigh Flux Reactor Trip is placed
back in service for both channels.
3. When reactor power is increased greater than P-10 (2/4 PR charinels
greater than 10% reactor power) the SR detector power is further
assured to be de-energized by b!ocking voltage to the source range
detectors. This blocking of the voltage to the source range detectors is
automatically removed when reactor power is reduced below the P-10
setpoint and P-10is de-energized. Should P-10not de-energize (either
due to 3 out cf 4 PH chanriel P-20 bistabies not clearing or the 2 out of 4
and gate not goirig to the NOT P-10 state) then the SR detectors can
not be re-energized either automatically or manually. iAE intervention is
required should this event occur.
F. Intermediate Range Detector (OM#2)
-
1. Detector Compensated Ion Chamber - (gamma wmpensatiop required
on!y beiow 105amps).
a) Two Volume Detector
I) Outer volume Boron lined, sensitive to neiitroris and gammas.
2) Inner volume sensitive only to gamma - not boron lined.
3) Each volume generates current output
b) Inner volume current due to gamma interactions with N? gas.
c) Outer volume current generdted from neutron interactions with Boron
10 lining and gamma interactions with N2 Qas.
QP-CN-IGENB FOR TRAINING PURPOSES ONLY REV. 29
Page I O of41
_____
DUKE POWER CA TAWBA OPERA- TlONS TRAINING
-
IR COMPENSATED ION CHAMBER ~
I
NITROGEN FlLLE
GROMO
d) A high DC voltage is applied to the detector to provide for the
collection of all charged particles for each ionizing event. A voltage in
opposition to this voltage is applied to the compensation electrode.
Compensation is necessary because after sustained full power
operation of the reactor, there is an appreciable amount of residual
gamma flux. The compensation voltage is adjusted to cancel out the
signal due to gamma flux leaving an output from the detector which is
proportional to neutron flux only. Due to the high rate of neutron
pulses detected in the Intermediate Range the output from the
compensated ion chamber detector is direct current, and is coupled
directly to log current amplifies. (Obj. #7)
2. Under Compensation (OB4 #8)
a) Higher power indication than actual'power level.
b) Can result in attaining an extremely high SUR without seeing it on
SUR meters.
c) Can possibly prevent P-6 clearing and prevent automatic re-
energizing of SR detectors following a Reactor Trip.
d) Will energize P-6 earlier than expected during Reactor startup.
3. Over compensation (OBJ #8)
a) indicated power Iowes than true power, possibly pegged low.
b) Indicates a much higher SUR than actual SUR.
6) Will clear $4at higher actual flux level than normal.
OP-CN-IGENB FOR TRAINING PURPOSES ONLY REV. 29
Pagellof41
DUKE POWER CATAWBA OPEUATlONS TRAlNlNG
G. Intermediate Range Channels (IR) N35 and N36 (OBJ #5)
INTERMEDIATE RANGE
MGH VOLTKGE
CMAPENSKTWQ
VOLTKGE
MnER
P-6
LOW POWER ROD STOP
1. Range of indication IO-2 to I O 9 amps. (Startup to full power.)
2. Bistable Circuits
a) P-6 (112 SI? greater than 10-a amps) - allows the operator to block
the SW High bevel Reactor Trip.
b) bow Power Rod Stop (amps equivalent to 20% power) - prevents any
outward rod motion.
e) Reactor Trip (amps equivalent to 25% power).
3. Isolation Amp Feeds:
a) SUR Circuit
b) CIB Recorder - NIR 45
c) CIB Meter
4. Local Meter - On BR Panel Drawer
H. If?Circuits Outputs (OW #4)
1. Loa Level Arndifier
a) Receives De signal from detector.
b) Provides logarithm voltage output proportional to linear input current.
c) Output voltage to:
1) CIB and local meter calibrated in amps (IO- to I O 3 Amps)
OP-CN-IC-ENB FOR TRAINING PURPOSES ONLY REV. 29
Page f 2 of 4f
DUKE POWR CATAWEA OPERATIONS TRAINING
2) Isolation Amp
3) Bistable Relay Drivers
2. Bistable Relay Drivers
a ) P-6 (I/% IR greater than 10 -" amps)
b) bow Power Rod Stop
1) Current equivalent to 20% full power (1/2 channels).
2) Rod withdrawal stop in manual or automatic.
3) Blockable at P-10 (2/4 PR greater than 10% power).
c) Reactor Trip
1) Current equivalent to 25% full power (f/2 channels)
2) Blockable at P-I0
3. Isolation Amplifier
a) isolates kR channel from remote equipment.
b) Provides output for following:
1) SUR Circuit
(a) Converts rate of change of power level to SUR in DPM.
(b) Reads out on C/B.
2) C/B Indication - Meter calibrated in amps (1Q-I to IO3).
3) C/B Recorder
(a) 1NR-45 two pen recorder.
(b) Records IR level in amps when selected.
I. la Drawer Panei(0BJ #6)
1. Ampere Neutron Level Meter
a) Indicates current output of detector
b) Indicates in amps - Eight decades (IOmf1to amps)
2. Instrument Power "ON" Lam0 - 118 voits AC instrument power applied to
drawer.
3. Control Power "ON" Lame - 118 voits AC control power applied to driver
assembly controi circuits.
4. - Indicates OPERATION SELECTOR switch is in
a position other than "NORMAL".
OP-CMiGENB FOR TaAlNING PURPOSES ONLY REV. 29
Page73of41
Bank Quesstion: 502.1 Answer: A
1 Pt(s) A LOCA occurred on Unit 1 at 2:00 AM. Given the following events and
conditions:
0201 A containment air release was in progress and was immediately
terniinated.
0205 The control room operators are responding to the events in E- 1.
02 I0 The OSM has assumed the role ofthe Emergency Coordinator,
0215 A Site Area Emergency is declared
0230 Completed initial notification of the State and Local authorities for
declaration of a site area emergency
0240 The OSM notes the following plant conditions and determines the
appropriate classification for the following indications:
- Containment pressure indicates 18 psig
Containment Hydrogen concentration is 1.5%
- Containment valve VQ-2B failed to isolate on phase A signal.
Core exit T/Cs indicate temperatures of 1 3 W T
RVLIS lower range level indicates 90%
- IEMF-53AB indicate 85 R/hr
- The Operators enter FR-C. 1 (Response to Loss ofcore Cooling)
If the BSM determines the appropriate classification when the conditions are
first indicated (do not assume it takes 15 minutes for the OSM to classify the
events), which one of the following statements correctly describes the first
notification that is reauired?
REFERENCES PRO WDED: W/Q/M5000/01 (Class$cution of
EmergenqJ
A. Notify the State and Local authorities that the plant has declared
a general emergency no Later than 0255.
B. Notify the NRC Headquarters Operations Center of plant
conditions no later than 0300.
C. Notify the State and Local authorities that the plant has declared
a general emergency no later than 0310.
D. Provide the first follow-up notification for a site area emergency
to the State and Local authorities no later than 0315.
Dlstraeter Analysis: Comparison to RP/O/A/5000/001 EALS shows:
- Cladding - 5 pts - T/Cs > I200 'F = core cooling red path (note:however,
EMF-53AW do not meet criteria for clad failure -must be >114 WIT)
Ques-SOt. 1.doc
- NC integrity - 5 pts - LOCA has occurred
- Containment integrity - 1 point for potential loss
- 18 psig - criteria is I5 psig for potential loss - YES
- 1.5% If2 - criteria is 9% H2 far potential loss - NO
- In FR-C. 1Red Path for 15 minutes -Not until 0255
The OSM meets the criteria for a GE at 0240
A. Correct: Meets criteria for a GE at 0240 .The OSM has 15 minutes
to notify the State and Local Autholdtics - must be completed by
0255.
B. Incorrect: NRC is required to be notified within one hour of the
declaration of an emergency clawification (event time), no Later than
03 15 - not from when the event first occurs.
Plausible: Notification o f m C is a high priority communication. If
the candidate thinks that NRC notification must occur within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
of the time that the LOCA occurred at 0200.
B. Incorrect: The events are classified at the General Emergency level
at 0240. The OSM has 15 minutes to notify the State and Local
Authorities - must be completed by 0255.
Plausible: The core cooling red path must be in for 15 minutes if the
escalation tu S E is made on the basis of the core cooling red path -
this wodd be correct if the candidate does not recognize that the high
containment pressure is suficient to declare a GE at 0240.
D. Incorrect: Follow-up reports for lesser classifications are not made
Plausible: One hour from notification of State and Local authorities
for the S A E - this is correct if the candidate does not recognize the
classification of the general emergency.
Level: SRO Only 10CFR55.43(b)5
KA: APE WE03 64.35 (2.213.6)
Lesson Plan Objective: SEP Obj: 2,16
Source: Mod Ques-502, Ques-109
Level of knowledge: comprehension
References:
1. OP-CN-EP-SEP pages I I Bi 12
2. RP/O/A/5000/001 page 1-3
3. Rp/O/A/5000/OOI End 4.1pages 1-5
4.lZP/O/B/5000/013 Encl 4.1 page 2
DUKE POWER CATAWBA OPERATlONS TRAINING
Objective
List the four levels of emergency classification at Catawba Nuclear
Station.
When given a set of plant conditions and access to reference
materials, correctly classify an event using RP/O/N5QOO/OO1.
Explain personnel responsibilities during an emergency at Catawba
Nuclear Station.
0 List the persons (by title) who may assume the role of emergency
coordinator.
e State who is responsible for activation of the emergency warning
system (sirens).
e State who is responsible for making recommendations to offsite
agencies.
State your assembly point.
State the emergency tevels that require various actions to be taken.
OSC activation.
0 TSC activation
EQF activation
0 Site Assembly
e Mandatory site evacuation
Name the locations for emergency evacuation sites.
Explain how to determine which emergency evacuation site to use in
case of an emergency.
List the offsite agencies which should be notified during an emergency.
Explain the meaning of emergency sirens sounding.
State the size of the Emergency Planning Zone (EPZ).
Describe how to conduct site assembly/evacuation according to
RP/O/N5000/010 and NSD 114.
OP-CN-EP-SEP FOR TRAlNlNG PURPOSES ONLY REV. 22
Page 3 of 16
DUKE POWER - CA TAWBA OPERA TIONS TRAINING
Objective
Describe the symptoms that may or will require a site assembly or
evacuation Der RPIQ/A/5000/010 and NSB 114.
Describe the procedure for securing from a site assembly or
evacuation.
Explain the purpose of the NSB for Site Assernbly/Evacuation per NSD
114.
Describe the procedure for site assembly during the following per NSD
1q4:
inability to reach assembly point
working in RCA
Summarize the procedure for site assembly per NSD 114 including:
who may enter the plant during a site assembly
- the procedure lo account for personnel
State the time frames in which immediate and follow-up notifications
are to be made to various ofkite agencies.
Prepare and evaluate Emergency Notification Forms for both initial and
follow-up notification for any given accident scenario.
When given a copy of RP/QIN5500/002,apply the Immediate Actions
required for a Notification of Unusual Event.
Summarize the subsequent actions required for the Notification of
Unusual Event procedure per RP/O/N5000/002:
Describe the procedure to terminate the emergency.
Describe the procedure to give a follow-up message.
Z . For events lasting greater than one hour.
2. Significant change in the situation.
3. Escalation to a higher classification.
When given a copy of RP/0/A/5(300/003, apply the Immediate Actions
required for an Alert.
OP-CN-EP-SEP FOR TRAINING PURPOSES ONLY REV. 22
Page 4 of 16
DUKE POWER CA TAWBA OPERATIONS TRAINING
4. Immediate Actions Burinq an Emerqency
a) Compare actual plant conditions to the Emergency Action bevel(s)
listed in Enc. 4.1, RPIO/Af5000/001 then declare the appropriate
Emergency Class. (Obj. #2)
NOTE: Using the appropriate forms, have students classify an event and
fill out an Emergency Notification Form (ENF). (Qbj. # 2 ) (Obj.
- I 7)
b j Make notifications to state and county agencies within 15 minutes
of declaring the event. Notify the NRC immediately after the states
and counties, not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Notify all others as soon as
possible after the states, counties, and NRC are notified. (Qbj.
- 16) (Obj. #7)
1) NC and SC
2) York, Gaston, Mecklenburg counties
3) NRC (ENS phone)
c) Refer to the applicable Emergency Response Procedure (RP) for
the classification found in Enc. 4.1, RP/O/N5000/001:
Notification of Unusual Event, WP/O/A/50Q0/002(Obj. 18)
Alert, RP/O/Ai5000/003 (Obj. 28)
Site Area Emergency, RP/8/A/5000/004 (Obj. 22)
General Emergency, RP/0/8/5000/005 (Obj. 24)
d) In the event of an "Urgent Condition" problem that has radiological
implications, but does not require EP/RP implementation,
Radiation Protection Shift Supervisor shall be contacted to respond
with Protective clothing and equipment to accommodate 1 RP
Technician and 3 workers in an "Urgent Response Kit".
5. Subsequent Actkms Durina an Emeruency
a) Refer to the applicable Emergency Response Procedure (KP) for
the classification:
1) Notification of Unusual Event, RP/O/A/5000/002 (Obj. 19)
2) Alert, RP/O/N5000/083 (Obj. 21)
3) Site Area Emergency, RP/O/A/5000/004 (Obj. 23)
4) General Emergency, RPIO/N5000/005 (Obj. 25)
b) Make follow-up notifications to states and counties (Obj. 16), (Obj.
26)
I)Refer to the most recent revision of KP/O/A/5500/006A and B.
OP-CN-EP-SEP FOR TRAlNl" PURPOSES ONLY REV. 22
Page 11 of 76
_. DUKE POWER.......~ ..._.-- CATAWEA
.......... -OPERATIONS TRAINING
__. ....
.....
2) Notification of Unusual Event, Alert, Site/Area Emergency and
General Emergency
(a) Hourly until emergency closed out
or
(b) If any significant change in situation
or
(c) As agreed upon with each agency (not to exceed 4
hours).
c) Assess plant conditions to determine the need to change
emergency classifications :/Qbj. 27)
1) Raise classification by declaration of Alert, Site Area or
General Emergency. Make initial notification within 15 minutes
of declaration.
2 ) Lower classification by declaration of an Alert or NOUE per
Encl. 4.3 of RP/O/A/500Q/004,Site Area Emergency or
RP/O/A/5008/003, Alert respectively. Make initial notification
within 15 minutes of declaration.
3) Termination of an emergency varies with the classification from
which the emergency is being terminated. Refer to
RF/0/,4/5000/002 through 006A.
d) Brief plant personnel as to status of the emergency via the plant
page.
e) Make Protective Action recommendations to states and counties,
as appropriate.
f) Ensure emergency worker doses do not exceed the limits of
RPfOfA/50O0/018, Emergency Worker Dose Extension, as
appropriate.
g) Provide turnover to the TSC Emergency Coordinator, as
appropriate.
C. Personnel Responsibilities (Qbj. #3)
1. Initially the OPS Shift Manager is the individual who assumes the role
of Emergency Coordinator.
2. Subsequently the Station Manager or his designee will assume the role
of Emergency Coordinator. After EQF activation, the EOF Director is
responsible for overall emergency management.
OP-CKEPSEP FOR TRAINING PURPOSES ONLY REV. 22
Page 12 of 16
Procedure No.
Duke Power Company
Catawba Nuclear S k i o n
015
Classification of Emergency
(ISSUED) .. PDF Format
Classification of Emergency
1. Symptoms
1.1 Notification of Unusual Event
1.1.1 Events are in process or have occurred which indicate a potential degradation
of the level of safety of the plant.
1.1.2 No releases of radioactive material requiring offsite response or monitoring
are expected unless further degradation of safety occurs.
1.2 Alert
1.2.1 Events are in process or have occurred which involve an actual or potential
substantial degradation of the level of safety of the plant.
1.2.2 Any releases are expected to be Iiniited to small fractions of the EPA
Protective Action Guideline exposure levels.
1.3 Site Area Emergency
1.3.1 Events are in process or have occurred which involve actual or likely major
failures of plant functions needed for protection of the public.
1.3.2 Any releases are not expected to exceed EPA Protective Action Guideline
exposure levels except near the site boundary.
1.4 General Emergency
I .4.1 Events are in process or have occurred which involve actual or imminent
substantial core degradation or melting with potential for loss of containment
integrity.
1.4.2 Releases can be reasonably expected to exceed EPA Protective Action
Guidelines exposure levels offsite for more than the immediate site area.
2. Immediate Actions
2.1 Determine operating mode that existed at the time the event occurred prior to any
protection system or operator action initiated in respome of the event.
2.2 the plant was in Mode 1-4 and a valid condition affects fission product barriers,
proceed to Enclosure 4.1.
RP/~/k~5000/001
Page 3 of 3
__ 2.3 -
IF a General Emergency is NOT declared in Step 2.2 the condition does not affect
fission product barriers, review the listing of enclosures to determine if the event is
applicable to one the categories shown.
__ 2.4 Compare achal plant conditions to the Emergency Action Levels listed, then declare the
appropriate Emergency Class as indicated.
m _ 2.5 Implement the applicable Emergency Response Procedure (RP) for that classification
and continue with subsequent steps of this procedure.
Notification of Unusual Event RP/O/N5000/002
Alert RP/O!N5000/003
Site Area Emergency RFVO~A/5000/004
General Emergency RP/O/N5000/005
3. Subsequent Actions
3.1 To escalate, de-escalate, or terminate the Emergency, compare piant conditions to the
Initiating Conditions of Enclosures 4.1 through 4.7.
- 3.2 Refer to enclosure 4.9, Emergency Declaration Guidelines, as needed.
4. Enclosures
4.1 Fission Product Barrier Matrix
4.2 System Malfunctions
4.3 Abnormal Rad Levels/Radiological Effluent
4.4 Loss of Shutdown Functions
4.5 Loss of Power
4.6 Fires/Explosions and Security Events
4.7 Natural Disasters, Hazards and Other conditions Affecting Plant Safety
4.8 DefinitiondAcronyms
4.9 Emergency Declaration Guidelines
4.10 Radiation Monitor Reading for Enclosure 4.3 EALs
Enclosure 4.1
RP/O/NSOOO/OOI
Fission Barrier Matrix Page 1 of 5
Use E.4Ls to determine Fission Product Barrier status (Intact, Potential Loss, or Loss). Add points for all 3 barriers. Classify according to the table below.
Note 1: This table is only applicable in Modes 1-4
Note 2: Also, an event (or multiple events) could QCCW which results in the conclusion that exceeding the Loss or Potential. Loss tkresholds is I&IMINENT
(ie., within 1-3 hours). In this IMMINENT LOSS situation, use judgement and classify as if the thresholds are exceeded.
Note 3: When determining Fission Product Barrier status, the Fuel. Clad Barrier should be considered to be lost or potentially lost if the conditions for the
Fuel Clad Barrier loss or potential loss EALs were met previously during the event, even if the conditions do not currently exist.
Note 4: Critical Safety Function (CSF) indications are not meant to include transient alarm conditions which may appear during the start-up of engineered
safeguards equipment. A CSF condition is satisfied when the alarmed state is valid and sustained. The STA should be consulted to affirm that a CSF has
been validated and the appropriate functional restoration procedure has been implemented prior to the CSF being used as a basis to classify an emergency.
-- - -- - _.
EAL # Unusual Event E m# Alert EAL# Site Area Emergency EAL # General Emergency
- - -
4.1.11.1 Potential Loss of 4.1 A . 1 JASS Potential Loss 4.13.1 LOSS =Potential LOSS 1.1.G. 1 Loss of All Three Barriers
Containment of 0fBoth
Nuclear Coolant System Nuclear Coolant System
Fuel Clad
-- -- -
4.1.U.2 Loss of Containment 4.1.A.2 Loss Potential Loss 4.1.S.2 Loss 4.1 .G.2 Loss of Any Two Barriers
of An,
Fuel Clad Potential Loss Potential Loss of the Third
Combinations of ~0th
Nuclear Coolant System
ANI)
Fuel Clad
__ _ _
4.1.A.3 Potential Loss of 4.1.s.3 Loss o f Containment
Containment -
AND
-
AND Ims Potential Loss
Loss 3Potential Loss o f h y Other Barrier
of Any Other Barrier
Enclosure 4.1 RP/@A/5000/001
~ F ~ SBarrier
S ~ O QMatrix Page 2 of 5
NOTE: If a barrier is affectec it has a single point value based on a potential 10s: 3r a 10s: Not Applicable is
F
Barrier
Containment
I
included in the table as a place holder only, and has no point value assigned.
Points (1-5) I potential Loss (X)
...
I
Loss (XI
3
Classification
I
Fuel Glad 5 7- 10 Site Area
Emergency
I I
1. Compare plant conditions against the Fission Barrier Matrix on pages 3 through 6 of 6.
2. Determine the potential loss or loss status for each barrier (Containment, NCS and Fuel Clad) based on the EAL symptom description.
3. For each barrier, write the highest single point value applicable for the banier in the Points column and mark the appropriate loss column.
4. Add the points in the Points column and record the sum as Total Points.
5 . Determine the classification level based on the number ofTota1 Points.
6 . In the table on page 1 of 6, under the classification column. select the event number (e& 4.1 .A.1 for Loss ofNuclear Coolant System) that best
fits the loss of barrier descriptions.
7. Using the number (e.g. 4.1 .A.I ) select the preprinted notification form and complete the required information for Emergency Coordinator
approval and transmittal.
Enclosure 4.1
Fission Barrier Matrix
4.1.C CONTAINMENT BARRIER 4.1.N NCS BARRIER 1.1.F FUEL CLAD BARRIER
I
POTENTIAL LOSS
(1 Point)
-
I LOSS -
(3 Points)
POTENTIAL LOSS
(4 Points)
- 1,OSS -
(5 Points)
POTENTIAL. LOSS
(4 Points)
I LOSS -
( 5 Points)
1. -Safee Function Status 1. Critical Safehi Function Status 1. Critical Safetv Function Stam
- Containment-RED Not applicable NCS Integrity-Red Not applicable . Core Coolig-
Orange
Core Cooling-Red
Core cooling-RED
Path is indicated
for ,I 5 minutes
Heat Sink-Red
. Heat Sink-Red
2. Containment Conditions 2. NCS Le& Rate 2. PhZmarv Coolant Aetivitv Level
Containment
Pressure > 15 PSIG
Rapid unexplained
decrease in
Unisolable leak
exceeding the
. GREATERTHAN
available makeup
rn 9 o t applicahle Coolant Activity
GREATER THAN
containment capacity o f one capacity as 300 pCwkc Dose
H2 concenmtion > pressure following charging pump in indicated by a loss Equivalent Iodine
9% initial increase the normal of NCS subcooling. (DEI) 1-131
- Containment
pressure greater than
Containment
pressure or sump
charging mode
with letdown
isolated.
3 psiig with less than level response not
one full train ofNS consistent with
and a VX-CARF LOCA conditions.
operating.
CONTINIJED CONTINUED CONTINUED
Enclosure 4.1 RP~olA/5OOO/QO1
Fission Barrier Matrix Page 4 of 5
4.1.C CONTAINMENT B U E R
POTENTIAL I,OSS - I.OSS - POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS -
( I Point) (3 Points) (4 Points) (5 Points) (4 Points) (5 Points)
I
3. Containment Isolation Vy!.v Status After . SG Tribe Ruoture
I I
. Containment Raai?.ti.pMunitoring
Containment Isolation Actuation
Not applicable Containment Primary-to- Indication that a Not applicable Containment
isolation is Secondary leak SG is Ruptured ani radiation monitor
incomplete and a rate exceeds the has a Non-IsoIable 53 A or53 B
release path from capacity of one sccondary line faul reading >11.7 R h
containment exists charging pump in
the normal Indication that a
charging mode SG is ruptured and
with letdow a prolonged releas,
isolated. of contaminated
Secondary coolant
is occurring from
the affected SG to
the environment
4. SG Secondarv Side Release With Primarv-to- I. Containment Radiation Monitoring .Emergencv CoordinatormOF Director
Secondarv Leakage
Not applicable Releaseof
secondary side to
Not applicable . Not applicable
Judeement
h y condition, including inability to monitor
the barrier, that intbe opinion of the
the environment Emergency CoordinatorlEOF Director
with primary to indicates LOSS or POTENTIAL LOSS of
secondary leakage the fuel. clad barrier.
GREATER THAN
Tech Spec
allowable -
END
CONTINUED CONTINUED
?
a
Enclosure 4.1 RF/O/B/5000/013
Events Requiring IMMEDIATE NRC Notification Page 2 of 2
irements for the following events as soon a5 practical after the o( urrence becomes known to the licensee
Event Description Reporting Requirement
B Declared emergency classification as specified in Notify the NRC Operations Center immediately after
Rp/0/A/5000/00 1, "Classification of Emergency". notification ofthe appropriate state or local agencies
Emergency and not later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the time one of the
3assification B Change from one emergency classification to another emergency classes is declared.
Yotifications
w Termination of an emergency classification Activate the Emergency Response Data System
(ERDS) as soon as possible but not later than one
- Any further degradation in the level of safety of the plant or hour after declaring an Alert or higher emergency
other worsening plant conditions, including those that classification.
require the declaration of any of the emergency classes, if
such a declaration has not been previously made
The results of ensuing evaluations or assessments of plant
conditions
The effectiveness of response or protective measures taken.
Information related to plant behavior that is not understood
- As a courtesy in situations deemed necessarq..
CNS CRITICAL SAFETY FUNCTION STATUS TREES PAGE NO.
EP/l/N/5OOO/F-O Core Cooling - Page 1 of 1
FK-L . I
GO TO
FR-C .I
1
' KtACTOR VESSSI.
LOWER RANOE LEVF.1.
GREATER I R A N 41%
N y
YES
GO TO
FR-C .2
. . C0KEFXITl:Ca
LESS THAN 1200°F
GO TO
FR-C .2
RFACIOR VESSEL
LOWER RANGE LEVEL
GRFATtRTtlAN 4116
L - -
"u'
AT LEAST ONE
NC PUMP ON
- -
I
NC' SUBCOO1.ING
RASEU ON CORE
EXIT TICS GREATER
THAN 0'F
-
No '
YES
-.
I
,'.-~.-
? CSFSAT
.. i
.~*
Bank Question: $09 Answer: C
1 Pt(s) Unit 1 was operating at 100% power when a large break LOCA occurred at
0200. The control room operators are responding to the events in E-I (Loss
of Reactor or Secondary Coolant). The OSM has assumed the role of the
Emergency Coordinator.
-
Given the following conditions and events at 0245:
Containment pressure indicates 16 psig
- Containment Hydrogen concentration is 8.0%
Core exit T/Cs indisate temperatures of 1 100 "F
RVLIS lower range level indicates 30%
EMF 51a and b indicate 85 R/hr
Subcooling margin indicates -25T
Assuming that all required EQP actions were taken, what is the sorrecl
classification for the emergency event?
REFERENCES PROI'IDED
RP/0/A/5700/000
F-0
A. Notification of an Unusual Event
B. Alert
C. Site Area Emergency
D. General Emergency
~~~
Distracter Analysis: Classification as follows using N W C EALs:
Containment RED path = 1 EAL point (potentid failure)
NGS Barrier failure = 5 EAL points
Core cooling RED path = 4 EAL point (potential failure)
A. Incorrect: the correct classification is an SAE
Plausible: if the candidate does not recognize the orange path or the
NCS barrier failw - with the red path for containment - this results
in 1 EALpoint=NOUE
B. Incorrect: the correct classification is an SAE
Plausible: based an LOCA table only 4 EAL points from orange path
in core cooling
C. Correct Answer:
Ques-109.doc
D. Incorrect: the correct classification is an SAIJ
Plausible: based on reasonable guess and misunderstanding EAL
philosophy - conditions we very bad
Bank Question: 491.1 Answer: C
I Pt(s) Unit 1 experienced a LOCA with a breach of containment at 0200. The
OSM a~sumedthe duties of the Emergency Coordinator and declared a
general emergency at 02 10. The initial recommended protective actions at
0225 were as foilows:
- Shelter zones A2,A3, B2, C2, D2, E2, F2, F3
Evacuate zones AO; Ai, B1, CI, D1, El, FI
At 0235, the initial dose projection information was presented to the OSM.
Given the following conditions at 0245:
- Wind direction = 450"
- Windspeed-4MPH
e Projected dose at the site boundary
- TEDE = 250 mrem
- CDE thyroid = 400 mrem
- Containment radiation levels
1EMF53A = 950 R/hr
1EMF53B = 955 R/hr
Which one of the following protective action recommendations arc correct in
accordance with W/O/A/5000/05?
REFERENCES PROPTDED - RP/0/A/5000/005
A. Change the protective action recommendation in zones A1 and
B1 from evacuate to shelter.
B. Extend the evacuation to zones A2, A3, B2,62
C. Extend the evacuation to zones D2, E2, F2
D. The initial set of protective action rccommendations remain in
effect with no changes required.
-
Distracter Analysis:
A. Incorrect: - wrong zones for evacuation - never reduce actions
Plausible: - will get this answer if the candidate enters the wrong
table - the one that has wind speed > 5mpb
i%. Incorrect: - wrong zones for evacuation
Plausible: will get this answer if the candidate enters the PAR table
with a reciprocal wind direction
C. Correct answer
D. Incorrect: - wrong zones for evacuation
Plausible: - if the candidate thinks that no changes are required to be
made until the first 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> update - or if the candidate thinks that gap
activity has not been exceeded.
Level: SRO Only lOCFR55.43@)5
EL4: ADM G 2.4.44(2.1 / 4.0)
Lesson Plan Objective: SEP Obj: 24
Source: Bank
Level of knowledge: analysis
References:
I. OF-CN-EF-SEPpages 12,15, 16
2. RP/O/A/5800/005 -pages 1-6
3. RPi)~A/SOOO/QOSEncl4.2 and 4.3
Ques-91.1 .doc
DUKE POWER CATAWBA OPERATlONS TRAINING
Objective
. ..... . .i. '.
I
.
.,,..,,; . . :..
.:... ' .
State the exposure limits for emergency workers. .,., .~'.. :.... . .., .
- .>..,
When given a copy of RPIOIA/5000/005, apply the Immediate Actions
required for a General Emergency.
Expiain the EPA Protective Action Guides and Recommendations.
Explain the on-site and off-site protective actions.
per RP/OIA/5080/005:
Shew how to use Protective Action Zone Determination Tables.
Summarize the procedure for follow-up notifications per
RP/OIAl5000/006Aand 8:
OP-GN-EP-SEP FOR TRAlNlNG PURPOSES ONLY REV. 22
5 Of f 6
DUKE POWEff CA TAWBA OPERA TlONS TRAlNlffG
2) Notification of Unusual Event, Alert, Site/Area Emergency and
General Emergency
(a) Hourly until emergency closed out
or
(b) If any significant change in situation
Of
(6) As agreed upon with each agency (not to exceed 4
hours).
c) Assess plant conditions to determine the need to change
emergency classifications :(Obj. 27)
1) Raise classification by declaration of Alert, Site Area or
General Emergency. Make initial notification within 15 minutes
of declaration.
2) Lower classification by declaration of an Alert or NOUE per
End. 4.3 of RF/O/A/5000/004, Site Area Emergency or
RP/O/N5000/003, Alert respectively. Make initial notification
within 15 minutes of declaration.
3) Termination of an emergency varies with the classification from
which the emergency is being terminated. Refer to
WF/O/Af5000/002 through 006A.
d) Brief plant personnel as to status of the emergency via the plant
page.
e) Make Protective Action recommendations to states and counties,
as appropriate.
f) Ensure emergency worker doses do not exceed the limits of
RPIOIN50001018, Emergency Worker Dose Extension, as
appropriate.
g) Provide turnover to the TSC Emergency Coordinator, as
appropriate.
C. Personnel Responsibilities (Obj. #3)
1. Initially the OPS Shift Manager is the individual who assumes the role
of Emergency Coordinator.
2. Subsequently the Station Manager or his designee will assume the role
of Emergency Coordinator. After OF activation, the OF Director is
responsible for overall emergency management.
f
OB-CNdP-SEP FOR TRAINING PURPOSES ONLY REV. 22
Page I 2 of f 6
DUKE POWER CATAWBA OPERATIONS TRAINING
-
3. Site Evacuation (Obj. #4) May occur for Site Area Emergency and
always occurs for General Emergency, must be Dreceded bv a Site
Assembly. All non-essential site personnel should proceed to one of
two Evacuation Sites based on site selection criteria.
a) Site Newport Newport Tie Station on Mt. Gallant Road near SC
~
Hi way I 6 1 in York County (5 miles SW). (Obj. #5)
b) Site Allen - Allen Steam Plant off Southpoint Road in Gaston
County (12 miles N). (Obj. #5)
4. Evacuation Site Selection Crtieria: (Obi. #6)
a) Site Allen will be sekected if the wind speed is less than 5 mph.
b) If wind speed is greater than or equal to 5 mph
1) Select Site Allen if the wind is from 0 to 144.9 degrees or
255.1 to 360 degrees.
2 ) Select Site Newport if the wind is from 145 tu 255 degrees.
5. Evacuation of Publie - Ordered by states or counties, Duke Power can
only make a recommendation. (Qbj. #3)
6. Securing from a Site Assembly andhr a Site Evacuation (Obj. #12)
a) The decision to secure from a site assembly will be made by the
QSM or the TSC Emergency Coordinator, as appropriate.
b) The decision to secure from a Site Evacuation will be made by the
TSC Emergency Coordinatur.
F. Offsite Emergency Planning
-
I. Emeraencv Planninq Zones Evacuation plans are prepared for the
public in and about a 10 mile radius from the plant called the IO mile
Exposure Pathway Zone (EPZ), or Plume Exposure Zone. Other
planning efforts are taken in the 50 mile radius from the plant called the
50 mile Ingestion Pathway Zone (IPZ). (Obj. #9)
-
2. PromDt Alertins Svstem In order to alert the public to a problem at the
station, an outdoor warning system of sirens has been installed in the
10 mile EPZ These sirens can be activated only by the counties
warning point for either a radiological emergency or a civil emergency
(tornado, flood or nuclear attack). Sirens alert public to tune N o r
radio to AS (Emergency Alert System) for further information or
directions. (Obj. #8)
3. Notification Svstem - After hearing a siren, the public is to turn on a
radio or TV and listen for an Emergency Alert System (AS) message.
The message could be: (Obi. #8)
a) Information only about the emergency.
b) An instruction to evacuate homes, offices, schools, factories, etc.
OP-CN-EP-SEP FOR TRAINING PURPOSES ONLY REV. 22
DUKE POWER CATAWBA OPERATIONS TRAlNlNG
c) An instruction to remain indoors for sheker until further instructions
are available.
-
4. Sheltering The states and counties have designated shelter space
available (located outside the 10 mike EPZ in the host counties) for
everyone in the 10 mile EPZ (EPZ counties) with food, water, and a
place to sleep until they are instructed to return home.
2.2 Conducting a Site Assembly or Preparing the Site for an Evacuation
A. Referencing to RP/O/N5000/010 and NSD 114, discuss the foliowing:
1. Symptoms of entry. (Obj. #I 1)
2. Immediate and subsequent actions for Site Assembly and Evacuation.
(Obj. # I O )
3. Locations for Site Evacuation and basis for selection. (Obj. #5) (Obj.
- 6)
4. Actions to take if an individual cannot reach their assembly point. (Obj.
- I 4)
5. Actions to take if working in the RCA/RCZ and wearing protective
clothing. (Obj. #$4)
6. Who may enter the protected area during a Site Assembly? (Obj. #15)
2.3 NRC Notification Requirements (Obj. #7)
A. Referencing to RPIO/N5000/013, discuss the actions required in
communicating to the NRC. Refer to enclosures to clarify notification
requirements.
2.4 Distribution of Potassium Iodide Tablets in the Event of a Radioiodine Release
A. The Radiation Protection Manager in conjunction with available medical
advice shall control the distribution of KI tablets.
B. KI tablets should be given to:
1. Persons suspected of having been in the affected area.
2. Persons present in the affected area.
3. Persons who will enter the area while a significant amount of
radioiodine is present.
3. SUMMARY
3.1 Review Lesson Plan Objectives and answer student questions.
OP-CN-EP-SEP FOR TRAlNlNG PURPOSES ONLY REV. 22
Page f6 of f 6
Duke Power Company Procedure No.
Catawba Nuclear Station RPA)/A/5000/005
Revision No.
041
General Emergency
~ ~
Electronic Reference No.
- * * * * * * * * * UNCONTROLLEDFORPRINT * * * * * * * * * *
(ISSUED) - PDF Formiit
RP/O/A/5000/005
Page 2 of 7
General Emergency
1. Symptoms
1.1 Events are in process or have occurred which involve or imminent substantial core
degradation or melting with potential for loss of containment integrity.
2. Immediate Actions
NOTE: 1. Lines in left margin are for place keeping. Immediate actions may be perfornied
simultaneously.
2. Security events may require the suspension of access to and movement about the site.
Staffing and activation of the on-site emergency response facilities could complicate
or interfere with security operations resulting in unwarranted casualties.
__- IF a security event exists, discuss the feasibility of conducting a site assembly and activating the
TSC/OSC with the Security Shift Supervisor at 5765 or 5766.
__ site assembly and activation ofthe TSC/QSC are not feasible, refer to the following
procedure enclosures for guidance and NIA the associated steps in this procedure under
Immediate Actions concerning site assembly and ERO activation:
__ wP/0IBI5000/026, "Site Response to Security Events," Enclosure 4.3 - Step 5
that evaluates taking protective action
__ RP/OIB/5000/026,"Site Response to Security Events," Enclosure 4.4 ~
Activation of ERO during an Imminent Securiw Event
- IF- thc security event involves an insider threat, implement 2-person rule for access to all
vital areas.
- Consider delaying other actions in this procedure that could endanger site personnel until
the security threat is contained.
IF TSC, OSC and EOF have NOT been previously activated, notify the ERO to staff emergency
response facilities by performing the following steps (A and B):
____ A. Notify site personnel to activate the TSC and OSC by making the following
announcement twice over public address system:
"This is the Operations Shshifr Munager. A General Emergency has been declared.
Unit(s) __ is (are) affected. Activate the TSC OSC, and EOF. )'
__ B. Activate Emergency Response Organization by completing Enclosure 4.1 of this
procedure.
RF/O/A/jOoo/oos
Page 3 of 7
~ Make an immediate PROTECTIVE ACTION RECOMMENDATION (PAR) to be entered on
of the Emergency Notification Form. Determine PAR based on current lower tower
wind speed (use upper tower wind speed if lower tower wind speed is not available) as below:
Evacuate zones: AO, Al, B1, C1, B1, E l , F1
-WIND SPEEIGGREATER THAN 5 MPH
Evacuate two mile radius AND all affected zones 5 miles
downwind AND shelter in place remaining 10 mile EPZ as
shown on Enclosure 4.2, page 2 of 2.
__ Notify off-site agencies within 15 minutes of Emergency declaration time using an Emergency
Notification Form. Refer to one of the following procedures for instructions:
RP/O/A/5000/006A, Notifications to States and Counties from the Control Room
e RP/0/A/50QQ/006B9 Notifications to States and Counties from the Technical Support
Center
0 SWO/B/2000/004, Notificationsto States and Counties from the Emergency Operations
Facility
__- IF there is an indication of a radioactive release the TSC is not activated, contact RF shifi
to perfom off-site dose assessment per HP/O/B/1009/26.
__- IF a radioactive release or hazardous material spill is occurring or has occurred the TSC is
not activated, contact Environmental Management (EM), ext. 3333, for assistance in reporting
to state, local or federal authorities. After hours, contact the Environmental Duty person by
phone or pager. no answer, page 8-777-3333 which wil1 page all Environmental
Management personnel.
~ Conduct a Site Assembly using RP/O/A/5000/010, Conducting a Site Assembly or Preparing
the Site for an Evacuation.
~ Conduct a Site Evacuation using RP/O/A/jO00/015, Conducting a Site Assembly or Preparing
the Site for an Evacuation.
RP/o/A/sooo/oos
Page 4 of 7
-Notify the NRC using RP/OIB/5000/013, NRC Notification Requircments. This notification
should be made as quickly as possible but shall be made within one hour of the emergency
declaration time.
.~ -
IF Emergency Response Data System (ERDS) transmission has not been initiated (Alert or SAE
classification), initiate ERLPS within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of initial Alert or higher declaration by perforniing
the following:
-Type ERDS or select Main, then General,then ERDS on a Control Room
OAC workstation connected to the affected units OAC.
-Initiate ERDS transmission by depressing Fd or clicking Activate.
-- IF ERDS transmission will not connect to the NRC, inform the NRC using E.NS. The
TSC Data Coordinator will troubleshoot and initiate ERDS transmission upon arrival in
the TSC.
3. Subsequent Actions
__- IF a security event has occurred, perform the following to account for site personnei:
- A. WHE.N Security notifies the OSM that the security threat has been terminated,
make the following announcement over the public address system:
Thisis the Operratians SJiift Manager. The securip event has been terminated.
The security event has been terminated. I
- B. Conduct a site assembly per RP/O/A/5000/10, Conducting a Site Assembly or
Preparing the Site for an Evacuation.
~ Ensure WP has dispatched On-Site and Off-Site Field Monitoring Teams with associated
communications equipment per HP/0IB/1009/009, Guidelines for Accident and Emergency
Response.
Evaluate specific plant conditions, off-site dose projections, field monitoring team data, and
assess need to update Protective Action Recommendations made to states and counties in
previous notification. Refer to:
Enclosure 4.3, page 1 of 3, Guidance for Subsequent Protective Actions, Subsequent
Protective Action Recommendation Flowchart
e Enclosure 4.4, Evacuation Time Estimates for Catawba Plume Exposure EPZ
RP/~/A/5000/005
Page 5 of 9
-Make follow-up notifications to state and county authorities:
Every &until the emergency is teiminated
-
If there is any significant change to the situation
-
As agreed upon with an Emergency Management official from individual agency
~ RP/O/A/5000/018, Emergency Worker Dose Extension, shall be used to authorize emergency
worker doses expected to exceed normal occupational exposure limits during a declared
emergency event or exceed blanket dose extension h i t s authorized by the Radiation
Protection Manager.
__ Augment shift resources to assess and respond to the emergency situation as needed.
~ Announce over the plant public address system the current emergency classification level and
summary of plant status.
Assess the emergency conditions and the corresponding emergency classification. See
RP/O/A/5000/00 1, Classification of Emergency, then:
Remain in a General Emergency
-
Terminate the emergency (Refer to RP/O/A/50001020 or SlUO/F3/2OOO/OO3 for
Termination Criteria).
Announce any emergency classification ievel changes over the plant public address
system including a summary of plant status.
NOTE: Turnover of command and control to the TSC or EOF relieves the OShVEmergency
Coordinator of classification notification and Protective Action Recommendation (PAR)
m _ Turnover the responsibility of command and control for the emergency as follows:
__ Provide turnover to the TSC Emergency Coordinator per Enclosure 4.5.
__ - IF the emergency situation prevents activation of the TSC within 75 minutes of
declaration, contact the EOF Director and perform a turnover. Refer to EOF Director
Turnover Form in RP/0/A/5000/020, Technical Support Center (TSC) Activation,
Enclosure 4.1.
RP/@N5000/005
Page 6 of 7
__ - IF neither facility can take M o v e r , maintain command and controi until one of the
facilities is capable of accepting turnover.
__ In the event that a workers behavior or actions contributed to an actual or potential substantial
degradation of the level of safety of the plant (incidents resulting in an Alert or higher
emergency declaration), the supervisor must consider and establish whether or not a for cause
druy/alcohol screen is required. The FFD Program Administrator is available to discusdassist
with the incident.
EOF Director will terminate the emergency and reconmend entry into Recovery by briefing the
off-site authorities at the Emergency Operations Facility or if necessary by phone. Document
the termination briefing using Enclosure 4.6.
The EOF Director shall assign an individual to provide a written report within thirty days. This
report could be an LER or a written report if an LER is not required.
Person Assigned Responsibility
Enclosure 4.2 RPi0!A/S000/005
10 Mile Emergency Planning Zone (EPZ) Map Page 2 of 2
and Protective Action Zone Determination Tables
U s this table to determine the recommended zones for evacuation within the:
281.26 -303.75
Enclosure 4.3 RP/O:A/S000/005
Guidance for Subsequent Protective Actions Page 1 of 3
Subsequent Protective Action
Recommendation Flowchart
9 START
LAROE FlSSlON PROD 7
IhWNTORY OREATER MILE RADIUS AND 10 MILES
O M ACTNITY IN W W N W M D AND RCOMMEhW N-
CONTAINMENT? PLACE SHE.l.TwI FOR ZONES NOT
(ENCI..l.J,PAGE 2 o F 3 ) EVACUATED
(ENCL 4.3 PAGE 3 OF 3)
W S E S PROJECTED I
T o BE:
RECOMMEND EVACUATION OF
AND&* 5 REM coe I N ANY IDENTIFIED ZONE~S)wrnt WSE
ZONE'SHAT tIAS NO? BEEN PROJEfTlONS L I REM TEDE
PRT\'IOUSLY EVACUATED? AND/OR t 5 W.M CDE
NOTE: CHANGES IN WIND SPEED AND/OR WIND DIREC7ION MAY REQIJIRE TIlAT ADDITIONAL ZONES BE RECOMhlENDED FOR EVACUATION. 'THTSI
ADDITIONAL. RECOMMENDATIONS ARE BASED ON THE FOLLOWM(I:
. a W N D SPEED IS OREATER THAN 5 MPH bplE LAROE FISSION PRODUCT INVENTORY IS LESS THAN O M ACTIVITY IN C O N T A N F - W , USE
ENCWSllRE 4.2 PAGE Z OF 2 TO DETERMINE IF EVACUATION OF ADDITIONAL ZONES SHOULD BE RECCMMEhWED
LAROE FISSION PRODUCT INVENTORY IS OK-ATER THAN GM ACTIVSN IN COhTArNMEW USE 7 TO
DITBRMINE IF EVACUATION OF ADDITIONAL ZONS SHOUlD BE RECLHIMEWDED
Enclosure 4 3 RP/~/A/5000/005
Guidance for Subsequent Protective Actions Page 2 of3
Guidance for Determination of Gap Activity
Fission product inventory inside Containment is greater than gap activity if the containment
radiation level exceeds the levels in the table below:
I ~ TIME AFTER 1 HIGH RANGE CONTAINMENT MONITOR READING -
SHUTDOWV
(HOURS)
100 % Gap ActiviQ Release
&nclosure 4.3 RP/o/Af5000i005
Guidance for Subsequent Protective Actions Page 3 of 3
This Table Only Used For Large Fission Product Inventory Greater Than Gap Activity In Containment.
Use this table to determine the recommended zones for evacuation within the:
5 mile radius and 10 miles downwind for any windspeed.
NOTE 1, Upper tower wind direclion ispEfwred. gnat available, use lower tower wind direction. Use wind directionfront
Nutional Weaker Setvice $site meteorological informalion is not availablr. NWS: Primary: i-800-268- 7785
Backup: 864-67Y-IO85
2. Wind direction indicator in Control Room has a scde of 0 to 540 degrees. Both 0 and 360 degretv indicate North.
3. Subtracf 360from wind direction indicalions flearer :ha" 360 degrees to arrive at wind direction for table below.
Wind Direction
(Degrees from North) 5 Mile Radius 10 miles Downwind Remainder of EPZ
Bank Question: 485.2 Answer: D
1 Pt(s) Unit 1 was shutdown in mode 6 . Engineering reported that a recent test on
the l B KC heat exchanger showed that fouling had reduced its heat transfer
capability.
0 The heat exchanger was now incapable of meeting the design A
temperature with a maximum (FSAR) lake water temperature of 90 O F .
The design A temperature couid be met with lake water with a maximum
temperature of 80 OF.
0 Current lake temperature is 68 OF.
0 Lake temperature is not expected to reach 80 "F far another 30 days.
The plant is expected to remain in mode 6 for another 30 days.
Which of the following statements correctly characterizes the operability of
the KC system?
REFERENCES PROYIDED: Tech Spec 3.7.7 & Bases
A. The B train of KC is inoperable because it is unable to maintain
KC system temperatures below 120'F during a plant shutdown.
B. The B train of KC is inoperable because it is unable to maintain
normal KC temperatures below 90 OF.
C. The B train of KC is operable but degraded as long as Lake Wylie
temperature remains below 90 OF.
D. The B train of KC is operable but degraded as long as Lake Wylie
temperature remains below 80 OF.
Distracter Analysis:
A. Incorrect: It is operable but in a degraded condition because it
cannot perform its intended safety function - within the FSAR limits.
Plausible: The B train does not meet FSAR requirements for
perfomance. 120°F is the maximum KG temperature for shutdown
B. Incorrect: The FSAR function can be performed for the existing
-
situation can temporarily satisfy the FSAR requirements
Plausible: The B train does not meet FSAR requirements for
performance. 90°F is the maximum KC temperature for shutdown.
C. Incorrect: The intended F§AR bctions cannot be performed above
80 OF - so it cannot be operable above 80 O F
Piausibie: B train of KC can perform FSAR functions until lake
water temp exceeds 80 OF. The FSAR design iiniit for Lake water
temperature is 90 OF - and 50me candidates may become confused
between these temperatures
D. Correct: B train KC meets the requirements ofNSD 203.7 - the
intended FSAR function can be satisfied by the presence of certain
temporary conditions (Le. lake water remaining below 80 O F )
K A APE 026 G2.2.25 (2.513.7)
Lesson Plan Objective: KC Obj: 13
Source: Mod Catawba NRC 1999
Level of knowledge: comprehension
References:
1.OP-CN-PSS-kc page 13
2. Tech Spec 4.7.7 PROVIDED
~
3. Tech Spec Bases 3.7.7 -PROVIDED
DUKE POWER .....
..... .-., = .,..-I.- ...CA TA
I.. WBA OPERA TlONS TRAlNlNG
.-.......
z
P
Objective T
R
Q
State the purpose of the KC System.
Describe how the KC System is cooled.
Describe the normal flowpath of the KC System, including X
each header and the type of loads serviced by each.
Explain what happens in the KC System during: X
- Safety Injection (Ss)
~ Phase A Containment Isolation (Stj
- Phase 4 Containment Isolation (Sp)
- Blackout
- Low Low KC Surge Tank Level
_s
Given appropriate pfant conditions, apply limits and X
precautions associated with OP/l(2)/A64OO/QQ5
~ComoonentCooling Water System)
State the typical values of the KC pump discharge X
pressure, KC Hx outlet temperature and KC pump flow.
State the basic actions required of an NLO for a loss of
Component Cooiina Water and why.
Describe KC system makeup.
Draw a block diagram of the KC system per the KC System
Simplified Drawing.
Explain when the Chemistry group is to be notified X
concernina the KC svstem.
Describe the purpose of the EMF'Sassociated with the KC X
System and what is indicated by a high level radiation
alarm.
List the instrumentation available in the control room for the
KC System.
When given a set of plant conditions and access to X
reference materials, determine the actions necessary to
comply with Tech SpedSLC's.
Discuss the supplementary actions for the loss of KC AP. X
__
QP-CN-BSS-KC FOR TRAINING PURPOSES ONLY REV. 42
Page 3 of 26
DUKE BOW%!? CATAWEA OPERATlONS TRAINING
b) NCDT and Excess Letdown Hxs
I ) Flow controlled
2 ) Containment isolation for excess letdown is controlled ft-01~1 the
NV board.
H. KC Drain header and Drain Sump
1. Containment drain isolations will ciose on St
a) Located outside containment downstream of drain header
containment penetration.
b) Allows drain header to be aligned during all modes of operation.
3. One 500 gal. steel lined covered sump per unit.
-
a) 2 pumps per sump Aux. Bldg 522
b) Able to discharge to:
1) NR Chiller Surge Tank
2) Other Units Sump
3) KC Surge Tank
4) Mixing and Seffling Tank
2.2 Operation
A. Technical Specifications and Selected Licensee Commitments (8BJ. #13)
1. Refer to Technical Specification 3.7.7(Component Cooling Water (CCW)
System) and Bass.
2. Refer to Selected Licensee Commitments 16.7-10 (Radiation Monitoring
For Plant Operations)
B. KC System Limits and Precautions and Special Lineups.
I Review Limits and Precautions per OP/I/A/6400/05 (OBJ. #5j
~
2. KC System Alignment fer KC Heat Exchanger Cleaning (O-C95-?69)
a) Alignment is used to maintain the availability of all essential heat
loads associated with the KC Train having its heat exchanger
cleaned.
b) The KC Train containing the Heat Exchanger which is not being
cleaned supplies all Train A and B component loads.
QP-CN-PSS-KC FOR TRAINIMG PURPQSES ONLY R N . 42
Page 13 of 26
CCW System
I3 3.7.7
B 3.7 PLANT SYSTEMS
5 3.7.7 Component Cooling Water (CCW) System
BASES
BACKGROUND The CCW System provides a heat sink for the removal of process and
operating heat from safety related components during a Design Basis
Accident (DBA) or transient. During normal operation, the CCW System
also provides this function for various nonessential components, as well
as the spent fuel storage poo!. The CCW System serves as a barrier to
the release of radioactive byproducts between potentially radioactive
systems and the Nuclear Service Water System (NSWS), and thus to the
environment.
The CCW System is arranged as two independent, full capacity cooling
loops, and has isolatable nonsafety related components. Each safety
related train includes two 50% capacity pumps, surge tank, heat
exchanger, piping, valves, and instrumentation. Each safety related train
is powered from a separate bus. An open surge tank in the system
provides sufficient inventory to protect the pumps from a lack of net
positive suction head available (NPSHA) due to a moderate energy line
break. The pumps have sufficient NPSHA with the surge tank empty
provided the piping up to the tank is filled. The pumps on each train are
automatically started on receipt of a safety injection signal, and all
nonessential components are isolated.
Additional information on the design and operation of the system, along
with a list of the components served, is presented in the UFSAR,
Section 9.2 (Ref. I). The principal safety related function of the CCW
System is the removal of decay heat from the reactor via the Residual
Heat Removal (RHR) System. This may be during a normal or post
accident cooldown and shutdown.
APPLICABLE The safety related design basis function of the CCW System is to remove
SAFETY ANALYSES waste heat from various components essential in mitigating design basis
events which require Emergency Core Cooling System ( E M S )
operation. The CCW System is also used to support normal operation.
The normal temperature of the CCW is 9 0 T , and, during unit cooldown to
MODE 5 (Tdd -= 200°F). a maximum temperature of 120T is
Catawba Units 1 and 2 B 3.7.7-4 Revision No. 0
CCW System
B 3.7.7
BASES
APPLICABLE SAFETY ANALYSES (continued)
assumed (Ref. 1). This 1 2 0 F limit is to prevent thermal degradation of
the large pump motors supplied with cooling water from the CCW
System.
The CCW System is designed to perform its function with a single failure
of any active component, assuming a loss of offsite power.
The CCW System also functions to cool the unit from RHR entry
conditions (Tcold6 35OoF),to MODE 5 (Tmu c 2OO0F), during normal and
post accident operations. The time required to cool from 35QVto 20PF
is a function of the number of CCW and RHR trains operating. One CCW
train is sufficient to remove decay heat during subsequent operations with
TWlde 200°F. This assumes a maximum service water temperature of
10B°F occurring simultaneously with the maximum heat loads on the
system.
The CCW System satisfies Criterion 3 of 10 CFR 50.36 (Ref. 2).
LCQ The CCW trains are independent of each other to the degree that each
has separate controls and power supplies and the operation of one does
not depend on the other. In the event of a DBA, one CCW train is
required to provide the minimum heat removal capabllity assumed in the
safety analysis for the systems to which it supplies cooling water. To
ensure this requirement is met, two trains of CCW must be OPERABLE.
At least one CCW train will operate assuming the worst case single active
failure occurs coincident with a loss of offsite power.
A CCW train Is considered OPERABLE when:
a. Both pumps and associated surge tank are OPERABLE; and
b. The associated piping, valves, heat exchanger, and instrumentation
and controls required to perform the safety related function are
The isoiation of CCW from other components or systems not required for
safety may render those components or systems inoperable but does not
affect the OPERABILITY of the C6W System.
Catawba Units 1 and 2 E) 3.7.7-2 Revision No. 0
CCW System
B 3.7.7
APPLICABILITY In MODES 1,2, 3, and 4,the CCW System is a normally operating
system, which must be prepared to perform its post accident safety
functions, primarily RCS heat removal, which is achieved by cooling the
RHR he& exchanger.
In MODE 5 or 6, the requirem6nts of the CCW System are determined by
the systems it supports.
ACTIONS a_a
Required Action A.Z is modified by a Note indicating that the applicable
Conditions and Required Actions of LCO 3.4.6, "RCS L ~ Q ~ s - M B D E 4,"
be entered if an inoperable CCW train results in an inoperable RHR loop.
This is an exception to LCO 3.0.6and ensures the proper actions are
taken for these components.
If one CCW train is inoperabie, action must be taken to restore
OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining
OPERABLE CCW train is adequate to perform the heat removal function.
The 2 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Completion Time is reasonable, based on the redundant
'
capabilities afforded by the OPERABLE train, and the low probability of a
DBA occurring during this period.
R.landR.2
If the CCW train cannot be restored to OPERABLE status within the
associated Completion Time, the unit must be placed in a MODE in which
the LCO does not apply. To achieve this status, the unit must be placed
in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The
allowed Completion Times are reasonable, based on operating
experience, to reach the required unit Conditions from full power
conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE 5B2JI-l
REQUIREMENTS
This S R is modified by a Note indicating that the isolation of the CCW
flow to individual components may render those components inoperable
but does not affect the QPERABILITY of the CCW System.
Verifying the correct alignment for manual, power operated, and
automatic valves in the CCW flow path to safety related equipment
provides assurance that the proper flow paths exist for CCW operation.
Catawba Units 1 and 2 B 3.7.7-3 Revision No. 0
CCW System
R 3.7.7
BASES
SURVEILLANCE REQUIREMENTS (continued)
Phis SR does not apply to valves that are locked, sealed, or otherwise
secured in position, since these valves are verified to be in the correct
position prior to locking, seaiing, or securing. This SR also does not
apply to valves that cannot be inadvertently misaligned, such as check
valves. This Surveillance does not require any testing or valve
manipulation; rather, it involves verification that those valves capable of
being mispositioned are in the correct position.
The 31 day Frequency is based on engineering judgment, is consistent
with the procedural controls governing valve operation, and ensures
correct valve positions.
2x3Lz.2
This SR verifies proper automatic operation of the CGW valves on an
actual or simulated actuation safety injection, Phase A Isolation, or
Phase B Isolation signal. The CCW System is a normaily operating
system that cannot be fully actuated as part of routine testing during
normal operation. This Surveillance is not required for valves that are
locked, sealed, or otherwise secured in the required position under
administrative controls. The 18 month Frequency is based on the need to
perform this Surveillance under the conditions that apply during a unit
outage and the potential for an unplanned transient if the Surveillance
were performed with the reactor at power. Operating experience has
shown that these components usually pass the Surveillance when
performed at the 18 month Frequency. Therefore, the Frequency is
acceptable from a reliability standpoint.
sR-3LZ2
This SR verities proper automatic operation of the CCW pumps on an
actual or simulated actuation signal. The CCW System is a normally
operating system that cannot be fully actuated as part of routine testing
during normal operation. The 18 month Frequency is based on the need
to perform this Surveillance under the conditions that apply during a unit
outage and the potential for an unplanned transient if the Surveillance
were performed with the reactor at power. Operating experience has
shown that these components usually pass the Surveiilance when
performed at the 18 month Frequency. Therefore, the Frequency is
acceptable from a reliability standpoint.
Gatawba Units 1 and 2 8 3.7.7-4 Revision No. 0
CCW System
B 3.7.7
REFERENCES 1. UFSAR, Section 9.2.
2. 10 CFR 50.36, Bechnlcal Specifications, (c)(2)(ii)
Catawba Units 1 and 2 E3 3.7.7-5 Revision No. 0
CCW System
3.7.7
3.7 PLANT SYSTEMS
3.7.7Component Cooling Water (CCW) System
LCO 3.7.7 Two CCW trains shall be OPERABLE*.
APPLICABILITY: MODES 1 2, 3,and 4.
I
ACTIONS -
CONDITION REQUIRED ACTION COMPLETION TIME
A. One CCW train A.1 __ss __
_____-- NOTE--------------
inoperable. Enter applicable
Conditions and Required
Actions of LCO 3.4.6,
"RCS LooPs-MODE 4,"
loops made inoperable by
ccw.
Restore CCW train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />'
OPERABLE status.
B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
associated Completion
Time of Condition A not NJ
.p
l
met.
8.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
- For each CCW :rain on Unit 2 , the Completion Time that one CCW train can be inoperable as specified by Required
Action A.1 may he extended beyond the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> up to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> as part of the NSWS system upgrades. System
upgrades include maintenance and modification activities associated with the NSWS piping. valves, arid branch 1hnos.
necessary repairs and!or replacement, and replacement of portions of the NSWS piping to the AFW syster.. Upon
completion of the cleaning. upgrades, and system restoratior in refueling outage 1 EOC12. this footnote is nci longer
applicable.
Catawba Unlts 1 and 2 3.7.7-1 Amendment Nos. 189/182
CCW System
3.7.7
SURVEILLANCE REQUIREMENTS
SURVEILLANCE
Verify each CCW manual, power operated, and 32 days
automatic valve in the flow path servicing safety related
equipment, that is not locked, sealed, or otherwise
secured in position, is in the correct position.
SR 3.7.7.2 Verify each CCW automatic valve in the flow path 18 months
servicing safety related equipment that is not locked,
sealed, or otherwise secured in position, actuates to the
correct position on an actual or simulated actuation
signal.
SI? 3.9.9.3 Verify each CCW pump starts automatically on an actual 18 months
or simulated actuation signal.
Catawba Units Iand 2 3.7.7-2 Amendment Nos. 1731165
1 Pt(s) Unit 1 was shutdown in mode 6. Engineering reported that a recent test on
the w1T heat exchanger showed that fouling had reduced its heat transfcr
capability.
- The heat exchanger was now incapable of meeting the design delta
temperature with a maximum (FSAR) lake water temperature of 90 O F .
The design delta temperature could be met with lake water with a
maximum temperature of 80 O F .
- Current lake temperature is 68 "P
- Lake temperature is not expected to reach 80 "F for another 30 days.
The plant is expected to remain in mode 6 for another 30 days
Which of the following statements correctly characterizes the operability of
the RN heat exchanger?
REFEREACES PROVIDED: Tech Spec 3.7.8 & Bases
A. Operable for 30 days
B. Operable but degraded as long as Lake Wylie temperature
remains below 80 OF.
C. OperabIe but degraded as long as Lake Wylie temperature
remains below 90 O F .
D. Inoperable
Distracter Analysis:
A. Incorrect: It is operable BUT in a degraded condition because it
cannot perform its intended safety function - withim the WAR limits.
Plausible: If lake temperature will not exceed it's delta temp limits
for 30 days, it could be considered operable until temp reaches the
limit. In addition, the RN system is not required to be operable in
mode 6 so some may think that they can call it operable if it is not
required in the mode that they are in.
B. Correct answer meets the requirements of NSD 203.7 - the intended
S A R function can be satisfied by the presence of certain temporary
conditions @.e.lake water remaining below 80 O F )
C. Incorrect: The intended S A R functions cannot be performed above
80 "F - so it cannot be operable above 80 O F
Plausible: Can perfom S A R functions until lake water temp
exceeds 80 OF. The FSAR design limit for Lake water temperature is
90 "F - and some candidates may become confused between these
temperatures
D. Incorrect: The SAR function can be performed for the existing
situation - can temporarily satisfy the SAR requirements
Plausible: if the candidate does not know the definition of
"operable" and "operable but degraded".
Bank Question: 483.f Answer: D
1 Pt(s) Unit 1 is in mode 6 and refueling operations are in progcss. Given the
following Conditions and events:
The Fuel Handling Manipulator Crane Operator (FHMCO) has indexed
the mast over the location where fuel assembly N-8 \vi11 be inserted.
- All conditions and indications on the Fuel Handling Manipulator Crane
are satisfied for inserting the fuel assenibly.
Which one of the fallowing statements describes the responsibility of the
Fuel Handling SRO associated with inserting the fuel assembIy?
A. Must give his permission to the OAT6 prior to inserting the
assembly. Operates the Latch/Unlatch switch. Verifies the fuel
handlers are documenting satisfactory completion of each rod
latch.
B. Must obtain permission from the OATC prior to inserting the
assembly. Qperates the Latchflinlatch switch. Documents
satisfactory completion of each assembly insertion.
C. Supervises reactivity management and verifies maintenance
technicians are continuously monitoring the load cell. Must
obtain permission from the OATC prior to inserting the
assembly.
D. Supervises reactivity management and personally monitors the
fuel insertion operation. Must give his permission to the
FHM60 prior to unlatching the assembly.
__--
Distrarter Analysis:
A. Incorrect: The OATC is not in the approval chain - FHMCO
operates the latching mechanism.
Plausible: matches with elements in other distractew and sounds
supervisory.
R. Incorrect: The OATC is not in the approval chain - FIIMCO
operates the latching mechanism.
Plausible: if the candidate feels that more control is needed.
C. Incorrect: The OATC is not in the approval chain
Plausible: If the candidate thinks that the CR is in charge.
D. Correct:
Level: SRO Only lOCFR55.43@)6
KA: C 2.2.27 (2.6 / 3.5)
Lesson Plan Objective: FL 0b.j: 5
Source: Bank
Level of howledge: memory
References:
1. OP-CN-FH-FL pages IO, I4
Qua-483.1 .doc
-~
DUKE POWER --. CA TA..WBA OPERATIONS TRAINING
-__ i__ _ .--.
.
. _ . ~._m____
Objective
Zxplain the purpose of the Containment Purge System.
Iescribe the normal flowpath of the VP System and the refueling
lowpath.
Sxplain the importance of a proper flow balance during VP System
3perations.
Zxplain the purpose and use of local controls.
Explain the functions and locations of the VP System controls and
ndications.
Iescribe the startup, monitoring, and shutdown of the VP System per
.he OP.
lescribe the automatic actions that occur to the VP System in the
?vent of an SH signal, an alarm on EMF-39, or fan trig signal.
3ven the appropriate plant conditions, apply Limits and Precautions
3ssociated with related station procedures.
-
sxplain how to complete a purge release form after a purge is
- omplete.
lescribe stattup, monitoring and shutdown of the lncore Instrument
Soom PurQeSystem per the OP.
3ven a set of plant conditions and access to reference materials,
letermine the actions necessary to comply with Tech SpecdSLCs.
State the system designator and nomenclature for major components
OP-WCNT-VP FOR TRAINING PURPOSES ONLY REV. 23
sag% 3 of 29
DUKE BOWER CATAWBA OPEaATIONS TRAINING
D. The rheostats associated with the supply and exhaust dampers will be
positioned fully clockwise to ensure the suppiy and exhaust dampers are
closed and recirc dampers are open (refer to Figure 9, 10).
E. Set EMF-39 setpoints to the specified values and setup the associated
chart recorder.
F. Enter release initiation information on the Release Record
1. DateTTime release initiated
2. Initial integrator reading
6. Start VP supply and exhaust fans. (refer to Figure 6 )
H. Verify containment isolation valves open.
I. Balance supply and exhaust flows to prevent pressurization or vacuum
inside containment (refer to Figure 5).
J. Notify RP that Containment Purge has been initiated
3.3 Shutdown of Containment Purge (Obj. #6)
A. Verify Initial Conditions.
B. Notify appropriate personnel that that the VP release will be terminated.
C. Place the Containment Purge Fan Units Control Switch to "OFF" (refer to
Figure 6).
1 . Verify all fans stop and all containment isolation valves close (refer to
Figure 6).
D. The rheostats associated with the supply and exhaust dampers will be
positioned fully clockwise to ensure the supply and exhaust dampers are
closed and recirc dampers are open (refer to Figures 9, 10).
E. Ensure mode selector switch is in the "NORM" position prior to the reactor
vessel missile shieid being put into place. This will prevent
overpressurizing upper containment (refer to Figure 5).
F. Place the key operated valves "Enable" switches to the "BLK CLSD"
position (refer to Figures 7, 8).
G. Position "Enable" switches for valves and fans to "BLOCK' (refer to
Figures 7, 8).
H. Stamp EMF chart recorder.
I. Notify RP and enter release termination information on the release record.
(Obj. #9)
1. DateTTime release terminated
2. Final integrator reading and volume released
3. Highest EMF reading
OP-CN-CNT-VP FOR TRAlNl" PURPOSES ONLY REV. 23
Page 15 of 29
O P i k B i 6 100!010X
PANEL: lRAD-l Page 4 of 37
1EMF-39 CONTAINMENT GAS HI RAD A/2
SETPOINT: Per IIP/O/B/1000/010 (Determination of Radiation Monitor Setpoints).
ORIGIN: 1EMF-39 beta scintiilation detector (low range).
PROBABLE Radioactive spill/leak inside containment,
CAUSE:
AUTOMATIC I. -
IF below P-6. the containment evacuation alarm is actuated.
ACTIONS: 2. -
IF at least one train of SSPS is NOT in test, the containment
ventilation isolation signal is actuated (SF,).
3. The Containment Purge System (VP) is isolated.
IMMEDIATE: 1. Verify that the Containment Purge System (VP) has isolated.
ACTIONS: 2. -
IF at least one train of SSPS is NOT in test, verify that the
Containment Air Reiease and Addition System has isolated.
3. Ensure all personnel are evacuated from containment.
4. LJse "SAMPLE FLOW SELECT' module to determine if alarm is
from upper containment. lower containment or incore instrument
room.
5. Refer to .4Pil!Ai5500ilO (Reactor Coolant Leak).
SlJPPLEMENTAHY 1. Notify Radiation Protection personnel of this alarm
ACTIONS:
I
_-_
-. -
1
~
NOTE: Ifthe EMF is reset prior to secnrmg VP per the following step, the system will restart
automatically.
2. Secure VP per OP/I!A/6450iOlS (Containment Purge System).
3. E actuated, reset the containment ventilation isolation signal when
this alarm clears.
NOTE: If annunciator alarm is due to an actual hi rad signal, a new GWR will be required hefore
reinitiating VP or VQ. I
4. Manually reinitiate the Containment Purge System (VP) or
Containment Air Release and Addition System (VQ) as needed
5. Refer to Tech Specs 3.4.13,3.4.14 and 3.4.15.
CONTINUED ON THE NEXT PAGE
Bank Question:479.3 Answer: B
1 Pt(s) Unit 2 is conducting a containment purge in accordance with
OP/2/A/6450/01S(C0ntainment Purge System). Given the following
conditions provided on the GWR permit:
Most restrictive release rate = 16000 CFM
0 Recommended release rate = 16000 CFM
2EMF-39(Lj trip 1 setpoint = 1.0E5 CPM
2EMF-39(L) trip 2 = 2.OE5 CPM
1EMF-36(L) is in service
Time 0215 0230 &?g
Release rate (CFM) 15750 16500 17500 18500
EMF-39 (CPM) 1.8E5 2.2E5 2.1E5 3.2E5
If the operators restart the VP purge whenever allowed by procedure, what is
the earliest time (if any) that the operators are reauired to terminate the
gaseous release and obtain a revised GWW?
A. 0200
B. 0215
C. 0230
D. 0245
~~~~~~~~~~~~~~~~~~~~~~~~ ~ ~
Distracter Analysis: OP/l/A/64501015 allows the operators to reinitiate a
containment release IF 1EMF-39(L) spikes. This question does not
indicate that there is a spike - but rather that the maximum release
rate has been exceeded due to high flow rates through the
containment purge line.
A. Incorrect: - no reason to terminate at 0200.
Plausible: - exceeds trip I on IEMF-39 and close to recommended
release rate.
B. Correct answer - exceeded recommended release rate, most
restrictive release rate and reachcd trip 2 on lEMF-39(L). This
clearly exceeds the release limits and there is no indication that the
EMF spiked.
C. Incorrect: -already should have terminated.
Plausible: - Exceeds recommended release rate - must terminate.
EMF-39 tripped VP for the 2"*time - if the candidate thinks you can
reset once and continue as allowed for a spike in the EMF
D. Incorrect: - should already be terminated.
. ......
. .. . .
.................... .
Plausible: - This is the answer is the release rate was not exceeded.
Level: SRO Only 10CF;R55.43(b)4
KA: G 2.3.9 (2.5b.4)
Lesson Plan Objective: CNT-VP Obj:7/9
Source: Mod Ques-479. I McGuirc 22002
Level of knowledge: comprehension
i . OP-CN-ChTT-VP page 15
2.0P/2/M6450//15page 2
3,OP/I/A/6450 End 4.1 page 4
3.OPlliI3/6100/010X I U D - 1 Ai2
4. HP/O/B/1004/005 pages 1-5
5 . WO/B/1004/034 End 5.2 (GWR)
Ques-479.3.doc
DUKE POWER
a
.<
.- ......-
...... CA TA.....WBA
... .._-.OPERATlONS TRAINING
Objective
Explain the purpose of the Containment Purge System.
Describe the normal flowpath of the VP System and the refueling
flowpath.
Explain the importance of a proper flow balance during VP System
Operations.
Explain the purpose and use of local controls.
Explain the functions and locations of the VP System controls and
indications.
Describe the startup, monitoring, and shutdown of the VP System per
the OQ.
Describe the automatic actions that occur to the VP System in the
event of an SH signal, an alarm on EMF-39, or fan trip signal.
Given the appropriate
. . . plant conditions, apply Limits and Precautions
associated with related station procedures.
Explain how to complete a purge release form after a purge is
comolete.
Describe startup, monitoring and shutdown of the Incore Instrument
Room Purge System per the OP.
Given a set of plant conditions and access to reference materials,
determine the
.....
actions necessary to comply
.. .................. .........
with Tech Specs/SLCs.
State the system designator and nomenclature for major components
09-CN-CNT-VP FOR TRAlNlNG PURPOSES ONLY REV. 23
Page 3 of 29
OP/1/N6450/015
Page 2 of 3
Containment Purge System
1. Purpose
The purpose of this procedure is to outline the operation of the Containment Purge System (VP).
2. Limits and Precautions
2.1 After refueling, ensure the "FUEL-NORM" switch is returned to "NORM" prior to
putting the reactor vessel missile shield into place. This prevents pressurizing upper
containment.
2.2 A new Gaseous Waste Release (GWR) sample is required if:
- 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has elapsed since the last sample.
NOTE: If actuation is due to an EMF spike, the release may be re-attempted twice before a new
sample is required. I
2.3 Any time initial entry into lower containment is desired, the incore instrument room shall
be sampled by Radiation Protection (RP) and purged unless the Operations Shift
Manager (OSM) deems purging U M ~ C C ~ S X Y .
2.4 Do reset containment ventilation isolation until spurious containment radiation
level signals OK any associated EMF alarms are properly cleared.
2.5 Any sudden increase or decrease in pressure across any filter bank shall be investigatcd
immediately.
2.6 For two train operation, the VP Pre-filters shall be replaced when the combined Pre-filter
and upstream HEPA differential pressure reaches 2" HzQ.
2.1 For single train operation, the VP Pre-filters shall be replaced when the combined Fre-
filter and upstream HEPA differential pressure reaches 3" H20.
2.8 If IEMF-37 or IEMF-40 has reached the Trip 1 setpoint, RP shall be notified to change
the CdItridge before a release is attempted.
2.9 Do NQT initiate purge of containment with 1EMF-39 inoperable. Once initiated, VP can
continue with IEMF-39L inoperable under the limitations presented in Enclosure 4.9
(Actions for EMF-39 Inoperability With VP in Service).
Enclosure 4.1 OPll/A/6450/0 I5
containment Purge System Startup Page 4 of 4
-2.9.4 Verify the following valves located on IRE%-ECP-3(AH-594, MM-52) open:
61 lVPOOlB (Upper Cont Purge Supply Outside Hsol)
0 1VP003B (Upper Cont Purge Supply Outside isol)
0 1VP006B (Lower Cont Purge Supply Outside Isol)
e3 IVP008B (Lower Cont Purge Supply Outside Isol)
66 iVPOl1B (Upper Cont Purge Exhaust Outside Isol)
lVP013B (Upper Cont Purge Exhaust Outside Isol)
0 lVP016B (Lower Cont Purge Exhaust Outside Isol)
CAUTION: *
_ _ - 2.10
e
Supply and exhaust air flow rates should be identical to prevent pressurization OF
vacuum inside containment.
Bo NOT exceed the Recommended Release Rate on the GWR Permit Report.
Establish Containment Purge supply and exhaust air flow at IRE%-CP-1(AB-534, LL52, Rm
1
500) at a rate greater than 10,000 CFM and Less than or equal to the "Recommended Release
Rate (cfm)" specified on the GWR Permit Report as follows:
2.10.1 Maintain supply and exhaust air flow rates q u a i as indicated on the following
gauges while adjusting flow in the foliowing step:
- "lVPP5150 (CPS-AFMD-1) C O W . PURGE SUPPLY AIR FLOW"
- "1VPP5200 (CPE-AFMD-I) CONT. PURGE EXHAUST AIR FLOW"
2.10.2 Simultaneously adjust the following minimum position switches in the
"DECREASE" (counter-clockwise) direction to increase supply and exhaust air
flow rates to achieve the desired flow rate while maintaining air flow rates
balanced:
"CONTAINMENT AREA PURGE SUPPLY"
- "CONTAINMENT AREA FILTER EXHAUST"
2.1 1 Not@ RP shift personnel that containment purge has been initiated.
Person notified
2.12 File this enclosure in the Control Copy folder ofthis procedure.
O P i l i S 6 I oo/o lox
PANEL: 1RAD-1 Page 4 of 37
1EMF-39 CONTAINMENT GAS HI RAD AI2
SETPOINT: Per HP!O~Hi1000!010 (Determination of Radiation Monitor Setpoints).
ORIGIN: IEMF-39 beta scintillation detector (low range).
PROBABLE Radioactive spillileak inside containment.
CAUSE:
AUTOMATIC 1. E below P-6, the containment evacuation alarm is actuated.
ACTIONS: 2. -
IF at least one train of SSPS is in test, the containment
ventilation isolation signal is actuated (SH).
3. The Containment Purge System (VP) is isolated.
IMMEDIATE 1, Verify that the Containment Purge System (W)has isolated.
ACTIONS: 2. at least one train of SSPS is NOT in test, verify that the
Containment Air KeIease and Addition System has isolated.
3. Ensure all personnel are evacuated from containment.
4. Use "SAMPLE FL.OW SE.LECT" module to determine if alann is
from upper containment, lower containment or incore instrument
room.
5. Refer to APiUN5500110 (Reactor Coolant Leak).
SUPPLEMENTARY I. Notify Radiation Protection personnel of this alarm.
ACTIONS:
NOTE:
automatically.
2. Secure VP per OP!l/A/6450/015 (Containment Purge System).
A
If the EMF is reset prior to securing VP per the following step, the system will restart
3. E actuated, reset the containment ventilation isolation signal when
this alarm clears.
~~
NOTE: If annunciator alarm is due to an actual hi rad signal, a new GWR will be required before
reinitiating VP or VQ.
Enclosure 5.1
Gaseous Waste Release ( W R ) Record
Shift Supervisor or Designee authoriring GUX release (signahlre required) - Datflime I GWR#_
V Q release monitored by EMF 39(L) This Copy has been compared with the Control Copy and Verified Correct Initial Date Time
(I' DateiTime Initial Final ("EMF39L Operable EW39L
VQ Release Integrator Integrator Checked and setpoints
Initiated Reading Reading Verification (1.V.) reset to
WQtify w) EMF Trip Setpoints non-releasc
release value
Total volume released ft3
Note 1 Notify RP Compliance of each VQ release start and stop time. Ensure EMF chart recorder is stamped at start of release and at completion of each release.
Note 2 E consecutive VQ releases are made ensure GWR record is updated for each release. EMF 39(L) is removed from service and EMF 36(L) is used to monitor the
release, request a new GWR from Rp Compliance. Not Applicable (N/A) may be used on this GWR record.
Note 3 Reset EMF 39 (I-) Trip 1 and 2 ~ ~ t p o i ntot snon-release setpoints when VQ System is NOT in service. Update Control Room EMF Setpoint Log.
Note 4 Volume for each VQ release =z Final Integator Reading x 10
Termination of GWR release acknowledged by Shift Supenisor or Designee (signature required) __ DateEime I
Enclosure 5.1 a~/Qi~/1oo4/oos
Gaseous Waste Release (CWR) Record Page 2 of 6
Shift Supervisor or Designee authorizing GWR release (signature required) - DakfTime / GWR #
EMF 39(L.) was inoperable in TSAIL on: Date Time (OH)Initial
VQ release monitored by EMF 3 6 0 y been compared with the Control Copy and Verified Correct Initial
This C ~ p has Date Tie
(I' Datefhne Initial Final u1EMF36L Operable and Source Highe# EMF36L Datemime (" VQ votme = (2) Control ROO^ operator
VQ Rclease Integrator integrator Checked and Independent EMF36L Setpoints Release Final Integrator (signature required)
Initiated Reading Reading Verification (1.Q.) required for reading per OPS Supended OK reading x 10
(Notify RP) l3W Trip Setpoints during Setpoint Terminated
release (Notify RP)
(LV.)
l1.V.)
(LV.)
1I.V.)
(1.V.)
0.V.)
Total volme releawk -
Note 1 Notify RP Compliance of each V Q release start and stop time. Ensure EMF chart recorder is stamped at start of release and at M n a t i o n of each release.
Note 2 E consecutive releases arc made ensure GWR record is updated for each release. E EMF 36(L) is removed from service. request a new GWR froni W Compliance
Enme EMF inoperable date and time is consistent with TSAIL. Not Applicable (N/A) m y be used on this GWR record.
Note 3 EMF 3 6 0 T ~ 1pand 2 setpoints are per OPS Setpoint Log. During normal VQ operation, EMF 36&) setpoints are NOT required to be reset.
Note 4 Volume for each VQ release = Final Integrator Reading x 10
Termination of GWR releax acknowledged by Shift Supervisor or Designee (si-maturc required) DateRime i
Enclosure 5.1 HP/QE3/1004/005
Gaseous Waste Release (GWB) Record Page 4 of 6
Shift Supervisor or Designee authorizing GWR release (signature required) DateiTime i GWRii
EMF 39&) was inoperable in TSAIL on: Date - Time (QPS) Initial
\T release madtared by E M F 36(L) This Copy has been compared With the Control Copy and Verified Correct Initial Date Time
VPDatemime
Release
2Control Room Operator
(signature required)
Initiated
I I I J I I I I E
Total volume released , A
Note 1 Noti& Rp Compliance of each VP release start and stop time. Ensure EMF chart recorder is stamped at start of release and at termination of each release.
Note 2 E consecutive releases are made, ensme the GWR record is updated for each release. @EMF 39(L) is removed fromservice and EMF 36(L) is used to monitor the
release, ensure EMF inoperable date and time on GWR record is consistent with TSAIUI,. Not Applicable (NIA) may be used on this GWR record.
Note 3 EMF 36(L) Trip 1 and 2 setpoints are per OPS Setpoint Log. During normal VP operation, EMF 36(L)setpoints are NOT required to be reset.
Note 4 Volume for each VP release = Final Integrator Reading x 1000
Termination of GWR release acknowledged by Shift Supervisor or Designee (signature required) - Date,Time !
Enclosure 5.1
Gaseous Waste Release (GWR)Record
Shift Supervisor or Designee authorizing G W R release (signature requuired) __ Datflime I---.- GWR #
IP release monitored by EMF 39(Jd) This Copy has been compared With the Control Copy and Verified Correct Initial Date __Time
(I) DateEime "EMFvIF39L Operable and Source Highhe@ EMF39L DatcKi (4) IP vol~me = "'Control Room Operator
IP Release Checked and Independent EMF39L setpoints Release Final Integrator (signature required)
Initiated Verification (LV.) required for reading resetto Suspended or reading x 100
(Notify RP) EMF Trip Sgoints during non-release Terminated
release value (Now w
Total. volume released: ft3
Note 1 Notify $3Compliance of each IP release start and stop time. E n m e EMF chart recorder is stamped at start of release and at completion of each release.
Note 2 E consecutive releases are on the same GVIU ensure record is updated for each release. E EMF 39&) is removed from service and EMF 3 6 0 is used to monitor the
release, ensure RP Compliance is notified. Not Applicable @#A) may be used on this record.
-
Note 3 Reset EMF 3%) Trip 1 and 2 setpoints to non-release setpoints when 1P System is NOT in service. Update Control Room EMF Setpoint Log.
Note 4 Volume for each IP release Final Integrator Reading x 100
Termination of GWR release acknowledgedby Shift Supenisor or Designee (signature required) _I
.- DateKime I
Enclosure 5.1
Gaseous Waste Release (GWQ Record
ShiA Supervisor or Designee authorizing GWK release (signamre required) ~
.DateiTime I ._ GWR #
E M F 3 9 0 was Inoperable in TSAIL on: Date Time (OPS) Initial
IF' release monitored lay EMF 3qL) This Copy has been compared wth the Control Cmpy and Verified Correct. Initial Date Time
(')Datet%ne Initial Final si EMF36L Operable and Source Highesp) EW36L DaterTim IP Volume = (2iControlRoom operator I
Ii
IP Release Integrator Integrator Checked and Independent EMF36L Setpoints Release Final Integrator (signature required)
Initiated Reading Reading Verification (1.V.) required for reading per 0PS Suspended or reading x 100
o\lotify Rp) E.W Trip Setpoints during Setpoint Terminated
I
release Log wo'rotifyw
Total volume released: ff
Note 1 Notify RP Cmmpliance of each IP release start and stop time. Ensure BMMF c h i recorder is stamped at start of release and at completion of each release.
Note 2 E consecutive relaqes are on the same GWR, ensure record is updated for each release. E EMF 3 9 0 is removed from service and EMF 36(L) is used to monitor the
release, ensure EMF inoperable date and time on CiWR record is consistent with TSAIL. Not Applicable (N/A) may be used on this word.
Note 3 EMF 3 6 0 Trip 1 and 2 setpoints are per OPS Setpoint Log. During n o m 1 IP operation EMF 36&) setpoints are NOT required to be reset.
Note 4 Volume for each IP release * Find Integrator Reading x 100
Tennination of GWR release aclohowledged by Shift Supervisor or Designee (signature required) Date/Time /
Bank Question: 479. I Answer: B
1 Pt(s) Unit 1 is in the process of making a radioactive gaseous waste release h m
the waste gas decay tank in accordance with OP/O/A/6200/18 (Waste Gm
Qperation). Given the following conditions:
- MRIRR-31CFM
- MOSRR=40CFM
EMF-50 (WASTE GASDZSCH) trip 1 setpoint = 2.OE5 CPM
IEMF-50 trip 2 = 3.0E5 CPM
1EW-34 @INKT VENTGAS) is in scrvicc
Time 0215 0230 0245
Release rate (CFM) 30 32 41 27
EMF-50 (CPM) 2.8E5 3.2E5 3.1E5 4.255
If the operators reset 1EMF-50 whenever allowed by procedure, what is the
earliest time that the operators are rewired to terminate (and not
immediately restart) the gaseous release?
A. 0200
B. 0215
c. 0230
D. 0245
Distracter Analysis:
A. Incorrect: - neither Trip 2 nor MRIRR exceeded at 0200.
Plausible: If candidate thinks Trip 1 is sufficient to terminate.
B. Correct: the release rate (32 CFM) MRIRR (3 I CFM) (most
restrictive instantaneous release rate)
C. Incorrect: - exceeded MRIRR at 0215.
Plausible: exceeded MOSRR (maximum observed system release
~
rate) - if the candidate thinks he/& can reset EMF-50 once before
being required to terminate the release - this is the 2"' time EMF-50
has reached trip 2.
D. Incorrect: - exceeded MRIWR at 0215
Plausible: exceeded trip 2 on EMF-50 for the 3rdtime - allowed to
~
reset this trip 2 times before terminating release
Lev& SRO Only; 10CFR55.43(B)(4)
KA: SYS 073A4.01(3.9/3.9)
Lesson Pian Objective: WE-KGR SEQ 5
Source: Mod; Ques-479, McGuire NRC 2000
Level of knowledge: analysis
References:
1. QP-MC-WE-RGR page 15
Ques-479.1.doc
Enclosure 5.2 HP/O/B/1004/034
Sample of RETDAS Gaseous Waste Release Page 2 of 2
Permit Reports
RETDAS <DPCCNS R.v.O.O> VSST
GASEOUS PRE-RELEASE P E W I T REPORT
- - - _ - _ _ _ - - _ - - _ s _ _ s s - _ _ _ _ _ _ s s _ _ _ _ _
QWR Number:
Releaae ID: Waste Gas Decay Tank "C"
-PI RECOKMENDED FLOWRPITES (cfm) I ~ ~ ~ ~ ~ ~ ~ ~ P P - = ~ P ~ - ~ - ~ I I I I P I / I I D P I I . c I P ~ ~ ~ ~ ~ ~ ~ - ~ - -
1 Units 2 Unite
Releasing Releasing
2 1 2 Station 1 / 2 Staticn
Limit Limit
(U.1) (U-2)
_ _ _ - _ - s s _ _ _ _ _ s _ _ _ _ _ _ _ _
Total body dose release rate (cfm) ................... 4.52B+05 2.26E+05
Skin and Gamma air dose release rate (cfm)..... ....... 1.12E905 5.58E+04
Food. Qround, Inhalation dose release rata (cfm).... ..5.00E+01 2.50E+01
- - - - - - _ - - - - s - - _ - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ s s - ~ ~ - - - - - - - ~ ~ - - ~ - - - - - - - - - - - - - - - - ~ - - - - - - ~ ~
Float restrictiva releaee rate (cfm) .................................. 1.50Ec01
Recommended release rate ............................................. 2.50E+01
==- l6L5TIPLE RELE?.SB CALCULATION s s i i ~ i i i - n n n / ~ i i - i ~ i ~ ~ - ~ ~ - = ~ ~ ~ ~
Sum o f calculated release rats divided by
allowable releaeo rate for a l l ccncurrent r e l e a e s e . . . . . .............. 6.3SB-03
I-D SETPOINT DATA I I I D P I I I I D I P P P _ P ~ D ~ ~ ~ ~ - ~ ~ ~ ~ ~ ~ ~ ~ - - - ~ - ~ ~ ~ ~ ~ - -
KMFSOL monitor Operable? ............................................ Yes
BIPSOL Entered Background ( c p )...................................... 4.00Ec03
-SOL Expected (cpm)................................................ 3.12E+04
W 3 6 L Monitor Operable. ............................................. NA
KUF3BL Enterad Background (cpm) ...................................... NA
m 3 S L Expectad (cpm) ................................................ NA
X e - 1 3 3 Bquivalonce (uCi/cc) ......................................... 3.12E-04
T r i p 1 Satpoint (Cpm) ................................................ 9.13E+04
T r i p 2 Setpoint (epm)................................................ 1.30B+05
Duke Power Company 'rocedurc No.
Catawba Nuclear Station tIPfiIB/I004/034
Levision No.
007
RADIOACTIVE WASTE GAS ( W S ) SYSTEM
RELEASE
HP/O/B/L004/034
Page 2 of 6
Radioactive Waste Gas (WG) System Relgase
-
1. Purpose
To provide a method of preparing a gaseous waste release package and to describe the
method of calculating the gaseous waste release rate to be used when making a Gaseous
Waste Release (GWR) from the WG system during normal plant conditions.
0 To provide actions for EMFSOL upon notification of inoperability/operability.
2. References
2.1 HP/OiE/lOOO/OIO - Determination of Radiation Monitor Setpoints
2.2 HP/OW/1001/01S - RP Compliance Sampling
2.3 SN/O/B/2001/004Investigation of Unusual Radiological Occurrences
2.4 Catawba Nuclear Station, Liquid and Gaseous Radwaste Computer Program User
Documentation
3. Limits and Precautions
None.
4. Procedure
4.1 Use of Procedure
NOTE: Staff Support Scientist or designated Qualified Reviewer may authorize operation
outside the scope and acceptance criteria stated in this procedure provided the
technical basis and impact to existing procedure 10CFRS0.59 evaluation is clearly
documented an applicable paperwork.
ation is affected, another evalu
4. I. 1 Notify Operations Shift Manager immediately of any inadvertent
(uncontrolled) gaseous effluent release.
4.1.2 Refer to SH/0/B/2001/004 (Investigation of Unusual Radiological
Occurrences) for notification process following any inadvertent (uncontrolled)
gaseous waste release.
HPIOIBI I0041034
Page 3 of 6
4.2 Sampling of WGDT-C
4.2.1 WHEN notified by Chemistry, collect samples per HP/O/W!1001!018 (W
Compliance Sampling).
4.2.2 Review completed WGDT-C sample results
4.2.3 Maintain results for use with GWR process.
4.3 Determination of Release Monitor for GWR
4.3.1 E EMF 50L is operable, obtain EMF 50L background reading (existing
reading).
- Ensure background reading (existing reading) is less than 20,000 cpm.
4.3.2 E monitor background (existing reading) is 2 20,000 cpm, declare EMFSOL
0 Generate work request for IAE IO decontaminate EMF.
Notify Radwaste Chemistry personnel of actions taken.
- Complete Section 4.7
4.3.3 E EMFSQLis inoperabie, use lEhaF36L as controlling monitor for release.
A. Ensure 1EMF36L background reading (existing reading) is Less than I000
cpm.
-
IF 1EMF36L background (existing reading) is 2 1000 cpni,
discontinue release process.
/
B. Obtain IEIvlF36L Trip 2 and Trip 1 setpoints.
4.4 Generate GWR Permit Report (Pre-Release)
4.4.1 Start RETDAS Launcher for data entry.
4.4.2 Verify GWR number assigned by RETDAS is next consecutive number.
4.4.3 Input assigned GWR number in LWWGWR Logbook.
4.4.4 Compiete remaining required entries.
- Refer as nccessary to RETDAS Computer P K O ~ EUser U ~ Documentation.
HP&B/l 004!034
Page 4 of 6
NOTE: "Multiple Release Calculation" (MRC) is performed by RETDAS to account for multiple
releases in progress and to ensure no release limits are exceeded. MRC uses data from all
open GWR Pre-Release packages. MRC sums the "recommended release rate" for each
open GWR and ratios to the allowable release rate of the GWR Pre-Release currently
being created. An information "box" opens to notify the user when the ratio is >0.9.
4.4.5 -
IF MRC is > 0.9, notify RP Staff Support and/or RP Supervision before
continuing GWR Pre-Release process.
4.4.6 Select "Knstmctions" on RETDAS to print information automatically on GWR
Pre-Release Permit report.
4.4.1 Ensure release data is correct on screen
4.4.8 Print report.
4.4.9 Compiete "Performed by" signature and "Date" blocks.
4.4.10 Perform verification of information by another qualified technician.
Complete "Verified by" signa(ure/date block.
4.4.1 1 Make copy of GWR Pre-Release Report.
Maintain with original Count Room sample analysis results in Compliance
Lab while GWR is active.
4.4.12 Ensure GWR package contains the following:
Sample analysis results
m Working copy of Chemistry procedure
- Reports forms generated by RETDAS
0 Additional information as needed for release
4.4.13 Deliver GWR package to Control Room.
4.4.14 WHEN notified by Chemistry, record initiation, suspension and/or
termination of WGDT GWR in RP Shift Gomplknce Logbook.
IIP/O/B/lO04/034
Page 5 of 6
4.5 RP Response to Trip 3 Alarms
4.5.1 notified of Trip 2 Alarm, ensure setpoints are correct.
4.5.2 Notify RP Supervision for approval to re-start release.
4.5.3 release is not re-started, go to Section 4.6.
4.6 G\YR Ciosed Permit
I
~ ~~
NOTE: Performance of RETDAS Closed Permit shail be accomplished by RP Staff Support or
qualified RP Compliance Technicians.
4.6.1 Review GWR package.
A. Verify isotope narne and concentration of computer inputs.
B. Verify background value is acceptable per Section 4.3.
C. Trip 2 alann(s) occurred during release, determine actual duration
time:
1. Add total time between initiations and suspensions of release.
2. Determine calculated stop time of release by adding actual duration
time of release to initial start time of release:
Example: 9-1-99/2i03 plus 80 minutes gives
calculated stop time of 9-1-99/2223.
3. Enter calculated stop on RETDAS Closed Permit screen.
4.6.2 Perform find review of GWR package.
4.6.3 File GWR in RP Satellite master file.
4.6.4 Discard duplicate copy of GWK maintained in RP Compliance Lab.
4. RF' Ac --Ins for Inoperabie EMF50L
4.7.1 WHEN EMF5OL is declared inoperabie, complete applicable sections of
EMF 50L Inoperable Status Log Sheet (Enclosure 5.1).
e Place Jnoperabie Status Log Sheet in EMF Status Logbook.
HP/Q/B/1004/034
Page 6 of 6
4.7.2 -EMF 50L is declared operable, perform the following:
A. Detemine/verify EMF SOL setpoints per HP/O/B/1000;010
(Determination of Radiation Monitor Setpoints).
B. Document setpoints on Enclosure 5.1.
C. Complete "Notified Operable" section.
4.8 Record Retention
4.8.1 Maintain the following documents in RP Satellite Master File:
0 GWR Release Papenvork
Enclosure 5.1
5. Enclosures
5.1 EMF SOL Inoperable Status Log Sheet
5.2 Sample of RE'TDAS Gaseous Waste Release Permit Reports
Enclosure 5.1 IIPIoIB/10041034
EMF 5QLInoperable Status Log Sheet Page 1 of 1
UNIT #: 1 EMF #: 50L Waste Gas Dischk NOTIFIED INOPERABLE PER
LOCATION m..594. Col.JJ-KK,.49-50 DATE/TIME:
WORK REQUEST #: RIP TECHNICIAN NOTIFIED:
REASON FOR INOPERABLITY:
NOTIFIED OPERABLE PER: DATEXIME I
DETERMINED SETPOINTS (CPM): TRIP 2 = TRIPl=
SETPOINTS PROVIDED TO: GROUPISECTION:
COMPLETED BY: DATEEIME: I
RP TECKNICIAN
Enclosure 5.2 NP/OiB/ 1OO4/034
Sample of RETDAS Gaseous Waste Release Page i of 2
Permit Reports
RETDAS cDPCCNS Rev.0.0r VBST
GASEOUS PRE-RELEASE PERMIT REPORT
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ s s _ _ _ _ _ _ _ _ _ _ _ s _ _
(IIYR Number:
Release ID: Waste Gas Decay Tank " C =
Release nods: Batch
Pernit Statnm: P - Bre-Rslease
Bank Question: 4?8. I Answer: A
1 Pt(s) Unit 1 is responding to a faulted steam generator from 100% power. Given
the following events at their respective times:
- 0200 - a reactor trip occurred, the crew entered E-0, (Reactor Trip or
Sufeg Injection)
- 0220- tmnsitioned to E-3, (Steam Generutor Tube Rupture)
- 0215 - entered FR-Z. 1 (Responseto High Contuimient Pressure), on a
valid orange path
- 0220- safety injection was actuated manually after a failure of auto S/I
- 0225 - completed FR-Z. 1
If the SRO determined that they were responding to the event in the wrong
procedure, which one of the following statements is correct regarding the
time the SRQ could transition to ES-0.0, (Rediagnosis)?
A. The SRO can enter ES-0.0 at 0211.
B. The SRO can enter ES-0.0 at 0216.
C. The SRO can enter ESO.0 at 0221.
D. The SRO can enter ES-0.0 at 0226.
Distracter Analysis: A faulted S/G has caused containment pressure to rise
to 3 psig. Safety injection has failed to x h t e in auto. The SRQ
implements E-3 instead of E-2. He recognizes that E-3 is not the
right procedure - so he wants to go to ES-0.0 (Rediugnosis). ES-0.0
does not apply if safety injection has not actuated -but
administratively, you can enter ES-0.0. Step 2 will verify that SI has
actuated and kick you out of ES-0.0 if SI has not yet actmted.
A. Correct: ES-0.0 can be enteEd after completion of E-0.
B. Incorrect: - Can enter ES-0.0 after completion of E-0 -also cannot
enter ES-0.0 until FR-Z. 1 has been completed.
Plausible: - if candidate notes SI has not actuated and thinks that
hdshe ES-0.0 cannot be entered until SI has actuated.
C. Incorrect: - ES-0.0 can be entered any time after completion of E-0.
Plausible: - if candidate notes that ES-0.0 is applicable only after SI
actuation and does not know that FR-2. 2 must be completed.
D. Incorrect: - ES-0.0 can be entered any time after completion of E-0.
Plausible: - if candidate thinks that ES-0.0 is not applicable until
after SI has actuated and cannot leave FR-Z. 1 until completed.
Level: SRQ Only 10CFR55.43@)5
KA. WEOt EA2.qA3.2/4.0)
Lesson Plan Objective: CSF Obj: 1, EP1 Obj: 19
Source: Mod Catawba Audit Exam 2000
Level of knowledge: comprehension
Referenccs:
1. OP-CN-EP-EPI page 6
2. ES-0.0 Background Document step 2 page 2
3. O W 1-7 page 11
4. ES-0.0 step 2 page 2
DUKE POWER CATAWBA OPERA TlONS TRAlNI ~
OBJECTIVES
Objective
correct procedure flowpath and necessary actions
DYKE POWER CATAWBA OPERATIONS TRAINING
Objective
(Natural Circulation Cooldown with Steam Void in Vessel (wlth
Given a set of specific plant conditions and all required
OP-CN-P-Pl FOR TRAINING PURPOSES ONLY REV. 7
Page 4 of 8
DUKE POWER CATAWBA OPERATlONS TRAlNlNG
I.EP/I/N5000/E-O (Reactor Trip or Safety Injection)
1.I Major Action Step Summary
A. Verify automatic actions as initiated by the Protection and Safeguards
Systems - Following entry into this procedure, the operators verify reactor
trip, turbine trip, and essential power available. If S/I was not required, a
transfer is made to E§-0.1. If §/I was required, the operators proceed to
verify proper equipment alignments. Additional safeguards equipment is
verified as the system setpoint is reached.
B. identify appropriate optimal recovery guideline - This begins the first
attempt to identify an appropriate recovery guideline.
C. Shutdown unnecessary equipment and continue trying to identify
appropriate recovery guideline - If SI1 cannot be terminated and diagnosis
has not been determined, the operators continue evaluating plant conditions.
1.2 Use the most current retype to explain the purpose, symptoms, and immediate
actions of E-0.
1.3 Use the "Enhanced Background Document" to explain the bases for all steps,
notes, cautions, and Enclosure 1 actions.
2. EP/I/N5000/ES-0.0 (Rediagnosis)
2.1 Major Action Step Summary
A. Determine if any SlGs are not faulted -Attempts to diagnose a faulted S/G.
3. Determine if any S/G is faulted and if it was isolated - Identify any faulted
S/G and see if it's isolated.
C. Determine if there is a SGTR - If faulted or faulted and isolated, attempt
to identify any S/G ruptured. If not, it must be a LQCA.
2.2 Use the most current retype to explain the purpose and symptoms E.S-0.0.
2.3 Use the "Enhanced Background Document" to explain the bases for all steps,
notes, cautions, and Enclosure 1 actions.
3. E$/1/N5000/ES-0.1 (Reactor Trip Response)
3.1 Major Action Step Summary
A. Ensure the primary system stabilizes at no-load conditions The ~
operators verify NC temperature returns to no-load via steam dump
operation. Verify, and if necessary, establish adequate shutdown margin.
Check Pnr level and pressure responding correctly.
B. Ensure the secondary system stabilizes at no-load conditions -Verifies
the operation of the steam dump and feed systems. S/Gs are returned to no-
load level values.
OP-CN-EP-EPI FOR TRAINING PURPOSES ONLY REV. 7
Page 6 of 8
STEP 2: V e r i f y S / I - HAS ACTUATED
PURPOSE :
To remind t h e operator t h a t E S - 0 . O only app'lies when S/I i s i n service o r i s
requi red.
APPLICASLE ERG BASIS:
The p a r t i c b l a r sequence o f steps i n t h i s procedure was based on the assumption
t h a t S/I i s i n service o r skolild be i n service. Therefore. t h i s procedure
should be used only i f S / i s i n service o r is required and E - 0 (Reactor T r i p
Or Safety I n j e c t i o n ) has been completed.
PLANT SPECIFIC INFORTATION:
KNDWLEDGE/ABILITY:
Page 2 o f 8 Revision 4
Operations Management Procedure 1-7 Page 11 of 25
0 If a valid orange path is encountered, the
operator is expected to scan all of the remaining
trees, and then, if no red path is encountered, to
promptly implement the corresponding EP. If
during the performance of an orange path
procedure, any red condition or higher priority
orange condition arises, then the red or higher
priority orange condition shall be addressed
first, and the original orange path procedure
suspended.
e Once a procedure is entered due to a valid red or
orange condition, that procedure shall be
performed to completion unless preempted by
some higher priority condition. It is expected
that the actions in the procedure will clear the
red or orange condition before all the operator
actions are complete. However, these
procedures shall be performed to the point of
the defined transition to a specific procedure.
At this point, any lower priority red or orange
paths currently indicating or previously started
but completed shall be addressed.
e If a CSF procedure directs the operator to return
to the procedure and step in effect and the
Corresponding status tree continues to display
the off normal condition, then the corresponding
CSF procedure does havc to be implemented
again since all recovery actions have already
bcen completed. However, if the same status
tree subsequently changes to a valid higher
priority condition, then the corresponding CSF
procedure shall be implemented as required by
its priority.
0 Certain CSF procedures are used to address both
orange and red path conditions for the same
parameters. If the procedure is already in
progress due to the orange path condition, it is
required to return to the first step if the
condition becomes red. Also, at the completion
of the procedure, the procedure does have to
be implemented again, since all recovery
actions have already been implemented.
REDIAGNOSIS
EP/l/N5000/ES-0.0
ACTIONlEXPECTED RESPONSE I RESPONSE NOT OBTAINED I
C. Operator A c m
- 1. Monitor Enclosure I (Foldout Page).
- 2. -
Verify SI1 HAS ACTUATED. - RETURN TO procedure and step in
effect.
3. E any of the following procedures are in
effect, THEN RETURN Ip procedure and
step in effect:
0 EP/I/N5000/E-0 (Reactor Trip Or
Safety Injection)
- 0 EP/l/N5000/ECA-0.0 (LOSS Of All AC
Power)
- 0 P/l/N5000/ECA-0.1 (LOSS Of All AC
Power Recovery Without §/I Required)
- e EP/l/N5000/ECA-0.2 (LOSSOf All AC
Power Recovery With SI1 Required)
1
Bank Question: 478 Answer: D
1 Pt(s) Unit 1 is rcsponding to a reactor trip from 100% power. Given the following
events and conditions:
0200 - a reactor trip oecurred, SRO entered E-0
0210- transition4 to ES-0.1, Reactor Trip Recovery
- 0215- cntered FR-Z.2, Containment Flooding, on a valid orange path
- 0216 - safety injection actuated
- 0220 - completed FR-Z.2
If the SRO determined that thcy were responding to the event in the wrong
procedure, which one afthe following statements is correct regarding the
appropriate action to take to transition to ES-0.0, Rediagnnsis?
A. The SRO can enter to ES0.O at any time during the event based
solely upon SRO judgement
B. The SRO can enter ES-0.0 between 0210 and 0215 or after 0220
C. The SRO can enter ES-0.0 after 0216
D. The SRO can enter ES-O.0 after 0220
Distracter Analysis:
A. Incorrect: - cannot enter ES-0.0 while in E-0 or while in FR-2.2.
ES-0.0 is onLy applicable after SI actuates
Plausible: - if candidate does not know restrictions and applicability
of ES-0.0
B. Incorrect: - cannot enter ES-0.0 while in ES-0.1 because ES-0.0 is
not applicable until after SI actuation
Plausible: - if candidate does not h o w that ES-0.0 is applicable
only after SI actuation
C. Incorrect: - cannot enter ES-0.0 while in PR-Z.2
Plausible: - if candidate notes SI has actuated and does not know
that he/she must complete FR-Z.2 before ES-0.0 can be entered
D. Correct answer
Bank Question: 357 Answer: C
iiu'd#7S& 1
1 Pt(s) &&+is respondinggte a main steam line break i&w containment, iq t'~ A.
A
Given the following events and conditions:
The operators completed E-0 (Reactor Tn'p and Sa&@ Injection) arid
transitioned to E-2 (Faulted Steam GeneratorIsolation)
- A RED PATH on Containment Integrity occurred and the opentors
transitioned to FR-Z. 1 (Raponse to High Containment Pressure) at step 8
of E-2.
- A RED PATH on NC Integrity occurred and the operators transitioned to
FR-P. 1 (RRfsponseto Imminent Pressurized Thermal Shock Condition)
fiornstep4ofFR-2.1.
The operators performed ail required actions in FR-P. 1
Upon completion of FR-P.l, the STA reports that all CSFs are now
GREEN (including Containment Integrity).
Which one of the following describes the correct procedure flow path?
A. Return to E-2 step 1and continue.
B. Return to E-2 step 8 and continue.
C. Return to FR-Z.l step 4 and complete the procedure, then return
to E-2 step 8.
D. Enter ES-0.0 (Rediagnosis) and rediagnose the situation.
Distracter Analysis:
A. Incorrect: Must address FR-2.1 first
Plausible: since all CSFs are green, the candidate may feel a r e m to
step 1 is appropriate
B. Incorrect: Must address FR-2.1 first
Plausible: this would be correct if Z. 1 had been completed.
6. Correct: per OMP 1-4, once a procedure is entered due to a valid red
or orange condition, that procedure shall be performed M completion
unless preempted by some higher priority condition. It is expected that
the actions in the procedure will clear the red or orange condition
before all the operator actions are complete. However, these
prokedures shall be performed to the point of the defined transition to a
specific procedure. At this point, any lower priority red or orange paths
currently indicating or previously started but not completed shall bs
addressed.
D. Incorrect: Have specific guidance to return to FR-Z. 1
Plausible: the operator can use rediagnosis at any time, hut the CSP
would not be addrcssed as required.
Level: SRO Only lOCFR55.43@)2
KA: APE 069 (32.4.4(4.CV4.3)
Lesson Plan Objective: FRZ Ohj: 5
Source: Bank
Levei of knowledge: comprehension
References:
1. O m 1-7 page 13
Ques-357.doc
DUK POWER CATAWBA OPERAPlONS TRAlNl"
OBJECTIVES
Objective
I I
3 Explain the Bases for the Major Actions of each of the Function
Restoration ~rocedures:EP/l/A/5000/FR-ZSeries - Containment
I I
4 Explain the Bases for all steps in each of the Function Restoration
procedures: EP/1/A/5OQQ/FR-2 Series Containment
~
5 Given a set of specific plant conditions and required procedures,
apply the rules of usage and outstanding PPRBs tQidentify the
correct procedure flowpath and necessaw actions
OP-CN-EP-FRZ FOR TRAINING PURPOSES ONLY REV. QI
Page 3 of 8
Operations !v4atiagcmentProcedure 1-7 Page 13 of 25
0 Yellow path procedures are to be performed
concurrent with the non-critical safety function
EP in effect when the yellow path is
implemented. While performing the actions of
the yellow path, continuous actions or foldout
page items of the non-critical safety function EP
in effect are still applicable and shall be
monitored by the operator. (DW-95-043)
0 If a red or orange condition indicates and then
clsars prior to implementation of the
corresponding procedure, the procedure shall
not be performed. The CSF procedure is
considered to be implenicnted when the
procedure reader reads the first step to the crew.
0 The STA shall keep the Operations Shift
Manager informed of all off normal CSFs. The
Operations Shift Manager shall ensure the crew
is updated as appropriate, typically by ailocating
time during updates for the STA. (SOER 94- 1)
F. Normally, the condition of the CSF Status Trees is
continuously displayed by SPBS on the OAC. Control room
indications shall be used to validate any off normal alami and
to determine which procedure to implement. Once status tree
monitoring is initiated, the STA should periodically monitor
the status trees and compare against control board indications
to ensure SPDS is functioning properly. Status tree monitoring
shall be continuous if an orange or red condition exists.
Otherwise, monitoring frequency shall be every 10 to 20
minutes. (SOER 94-1)
Bank Question: 339.2 Answer: D
I Pt(s) 5 3 (Stem Generator Tube Rupture) step 18 reads as follows:
WHEN P-IT PZR S/I BLOCK
PEMISSWE status light (ISI-18) is lit,
E N :
__a. Depress ECCS Steam Pressure ~ ~ L 0 C K p u ~ ~ h b u t t o i i ~
Pressurizer pressure is 1985 psig when the SRQ !mmitions to ECA-3.1.
After the transition into ECA-3.1, which one ofthe following statements is
correct with regard to this step?
A. The step is applicable only while in E-3. The SRO shall not
return to the step when notified that permissive status light is
LIT.
B. The step is applicable only while in E-3. The RO shall
immediately perform the actions of sub step 18a, only after
reentering E-3 upon completion of ECA-3.1.
C. The step is applicable while in E-3 and after transition to
ECA-3.1 until alternative gnidance is provided. The RO shall
immediately perform the actions of sub step 18a when the
permissive status light is LIT and report this action to the SRO.
D. The step is applicable while in E-3 and after transition to
ECA-3.1 until alternative guidance is provided. The RQ shall
notify the SRO that the permissive status light is LIT and the
SRO will return to step 18a and direct the action.
Distracter Analysis:
A. Incorrect: the step is applicable in E-3 atid after transition out of E-3
until alternative guidance is provided
Plausible: partially correct - page-keeping requirement is correct.
B. Incorrect: the step is applicable in E-3 and aftcr transition out of E-3
until alternative guidance is provided - RO must notify the SRO to
return to the step and receive direction to implement sub-step 18a.
Plausible: Ifthe candidate does not know the process for conditional
steps, this is reasonable and similar to foldout page actions.
C. t : notify the SRQ to retain to the step and receive
I n c ~ ~ rR~8 must
direction to implement sub-step 18a.
Ques-333.2.dos
Plausible: this is reasonable and similar to foldout page actions.
D. Correct Answer:
Level: SRO 0nly tOCFRSS.43@)5
MA:APE 038 G 2.4.6 (3.U4.0)
Lesson Plan Objective: EP-INTRO Obj: 2,3
Source: Mo& Qucs-339, Catawba NRC 2000
Level of howledge: memory
References:
1. OP-CN-EF-INTRO page 8
2.0MP 1-7 page 9-8
3. E-3 step 18 page 17
DUKE POWER - -.....................................
CATAWBA OPERATlONS
.. TRAINING
OBJECTIVES
=
L
0 bjective P
R
1 I
4 State when ACC conditions are used.
1 5 I Explain why ACC values are used.
1 6 1 Explain the purpose of the red film on the monitor light panel
I 7 I Explain how Enclosure 1 (F~IC-IOU~page) is used
1 1
13 Explain the General Statements of Philosophy.
State when a unit trip must be investigated and any exceptions to
NSD 505.
I 5 State who is responsible for notifying station supervision of a
1 16 I State the requirements that must be met prior to plant restart
for completion of time critical operator
actions. ............ .-
actions identify those that are time
critical in nature.
OP-CN-EP-iff TRO FOR TRAlfflffG PURPOSES ONLY REV. 02
Page 3 of 10
DUKE BOWER CATAWBA OPERA 7YQNS TRAiNlNG
4. EPIAP IMPLEMENTATION
4.1 OMP 1-7 (EIneFgenCy/AbnQrmalProcedure Implementation Guidelines) provides
the rules for use and implementation for EPs and AQs.
4.2 PPRBs:
A. PPRBs may also provide general and specific guidance for use and
implementation of EPs and APs.
B. The PPRB process is described in OMP 4-10 (Validation Process for
Operations Procedures).
5. REACTOR TRIP INVESTIGATION
Refer to current revision of NSD 505 (Reactor Trip Investigation).
6. TIME CRITICAL OPERATOR ACTIONS
6.1 Refer to PT/0/A/4700/061 (Time Critical Operator Action Review).
6.2 Expectations for performance of time critical actions (Obj. #17)
A. Operators are expected to perform time critical tasks in accordance with all
established procedures and work practices. The time critical nature of the
task does not preclude the need to perform the task properly and safely.
B. Inability to complete these time critical tasks within the time allotted may
indicate a need to evaluate procedures and work practices or to reanalyze
the event to include the actual performance time.
C. See the following table for the specific NLO tasks and consequences.
~
OP-CN-EP-INTRO FOR TRAINING PURPOSES ONLY REV. 02
Psge 8 Qf 70
Operations Management Procedure 1-7 Page 7 of 25
6. If desired. conditional steps (E ..., THEN...)can be
evaluated by the procedure reader for applicability prior
to giving a verbal command. The command should
then be constructed from the statement(s) after the word
T".
Example:
Step: IF Containment pressure has exceeded 3 psig,
TIIEN verify Phase B Isolation actuated.
Evaluation: Based on a previous response. or a
request for verification of containment
pressure status, the procedure reader
determines that the step is applicable.
Command: Verify Phase B Isolation actuated.
7. Konscquential steps (WHEN ..., THEN ...) require that
an action be performed after a particular event has
occurred. These steps should be handled as follows:
a. The procedure reader gives the RO a command
to notify lrini when the particular event occurs.
b. The procedure reader can mark the page at the
nonsequential step with a "tape flag" and
continue in the procedure.
C. Upon observing the event. the KO notifies the
procedure reader.
d. The procedure reader returns to the flagged step
and gives the command. The command should
be constructed from the statement(s) after the
word "THEN".
e. 'Ihe procedure reader removes the "tape flag"
and checks off the step.
Prior to branching to another procedure, the procedure
in effect should be reviewed for remaining "tape flags"
to ensure all required actions have been taken. Those
not completed should be reviewed to dctemine whether
_ I
they are progressing satisfactorily or if they are even
pertinent to the remainder of the recoven. effort.
Operations Management Procedure 1-7 Page 8 of 25
J. The following rules of usage shall be applied to continuous
action steps:
1. "IF AT ANY TIME" steps that require returning to
and performing a particular step of a prcxedure shall
only be applicable while that procedure is in progress.
2. "IF A 1 ANY TIME " steps that do E t require
performance of a particular step shall apply until the
current or subsequent procedures provide alternate
guidance.
3. "WHEN" steps shall apply continuously unless they
are determined to not be pertinent to the recovery effort.
4. All other continuous action steps (Le.*control, monitor,
maintain, etc.) shall apply until the current or
subsequent procedures provide alternate guidance.
K. Parameter values listed in parentheses with the acronym
"AGC." (Advwx. Containment Conditions) shall be used when
containment pressure has exceeded 3.0 PSIG. Once
containment pressure has exceeded 3.0 PSIG, ACX values shall
be used during the remainder of the accident even if pressure
subsequently decreases below 3.0 PSIG. This second setpoint
is used to account for the additional error in the setpoint due to
the containment environment following a high energy line
break.
L. The monitor light panels are used in the EPs to verify safety
components in their proper position for various phases of an
accident. Certain indicating lights are misplaced and do fit
the rules of use. These lights are marked with red film to
indicate an exception to the rules of use 'and shall be evaluated
to determine if the component is in the proper position.
M. Enclosure 1 (Foldout Page) shall be delegated to at least one
RO for the purpose of monitoring and reacting to various plant
parameters independent of the procedure reader's guidance.
Additional copies should be distributed to other crew members
to aid in the monitoring.
N. General statements of philosophy for EP/AIPusage are listed on
Attachment 1 1. 1 (General Statements of Philosophy).
Operations Management Procedure 1-7 Page 9 of 25
7.2. Optimal Recovery Procedures
A. This group of event related emergency procedures (EPs) covers
the diagnostic, mitigating and recoven actions for the
following accidents:
0 Loss of coolant accident (LOCK)
Steam/feed line break
0 Steam generator tube rupture (SGTR)
0 Loss of all AC power
R. Entrance into this EP series is limited to the following two
procedures:
0 E-0 (Reactor Trip or Safety Injection) shall be entered
when:
0 A reactor trip occurs or is required above P-1 1.
8 A safely injection occurs or is required above
P-l I , or below P-11 with autornatic. S/I
blocked.
- A safety injection occurs below P-l 1 with
automatic SI blocked, and entry is dictated by
applicable procedure.
8 ECA-0.0 (1.0~sof All AC Iower) shall be entered if a
complete loss of power on both emergency buses
occurs. This includes any time during the performance
of any other emergency procedulr.
I CNS STEAM GENERATOR TUBE RUPTURE PAGE NO.
EP/l/N5000/E-3 17of84
Revision 23
RESPONSE NOT OBTAINED
16. Verify the following valves on all Perform the following:
ruptured S/G(s) CLOSED:
~
a. Verify the following valves on at least
- 0 MSlV one intact S/G - CLOSED:
-0 MSlV bypass valves.
- MSlV
- MSlV bypass valve.
- b. E at least one intact §E cannot be
isolated from all ruptured S/G(sj, THEN
_
GO _TO EP/l/N5000/ECA-3.1 ( S G F
With boss Of Reactor Coolant -
Subcooled Recovery Desired).
- 17. Verify at least one NC pump ON. -
CAUTION NC T-Cold indication in
the ruptured loop may
cause an invalid Integrity
Status Tree condition.
- Disregard NC T-Cold indication in the
ruptured loop, until directed by this EP
or until this EP is exited.
18. "P-I 1 BZR Sll BLOCK
PERMISSIVE" status light (IS-18) is lit,
m:
a. Depress ECCS steam pressure
"BLOCK" pUshbUttons.
- b. Verify main steam isolation blocked
status lights (1SI-13) - LIT.
c. Maintain NC pressure less than
1955 PSlG using one of the following:
- Pzr spray
- PzrPORV.
I Pt(s) E-3, (Steam Generator Tube Rupture), step #21.b reads as follows:
IFAT ANY TIME ruptured S/G(s) pressure is decreasing...,
E N p e v f o r m Step 21.
Which onc of the following statements is correct with regards to this step?
A. The step is applicable continuously unless it is determined not to
be pertinent to the recovery effort.
B. The step Is applicable while in E 3 and after transition to
subsequent procedures until alternative guidance is provided.
C. The step is only applicable until another continuous action step is
reached in E-3.
D. The step is only applicable while in E-3.
Distracter Analysis:
A. Incorrect: step applicable only inE-4
Plausible: this is the construct for When ... then actions
B. Incorrect: step applicable only id!-3
Plausible: this is an alternate construct for If at any time actions
C. Ineorrect: step applicable only inE-3
Plausible: this is the construct for generalized continuous action
steps.
D. Correct:
Level: RO&SRB
KA: 62.4.19 (2.7 13.7)
Lesson Plan Objective: ADM-OP SEQ 21
Source: NRC Catawba Exam 97 Ques-339
Level of knowledge: meniory
References:
1. OP-CN-ADM-QP page 10
2. Qh4P 1-7 page 7
3. EP/l/A/SOOO/E-3 page 22
Bank QW?St/OR: 323.1 Answer: 19
1 Pt(s) Unit 1 is shutdown, in mode 5 , following a S/G tube rupture event.
Maintenance is dewatering the main condenser into a portable holdup tank
outside the turbine building (for disposal offsite). Upon completion of the
dewatering and piping flushes, the tank is 3f4 hull. Given the following
radiochemistry analysis of the tank contents:
e Total tank activity = 18 Ci with a combined half life of 8 days
0 Tritium activity = 1 Ci with a half life of 12.6 years
Noble gas activity = 6 Ci with a half life of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />
Which one of the following action(s) (if any) meets plant requirements for
these conditions, with the minimum risk of further contarnination?
REFERENCESBROVIDED: SLC16-11.17
A. Immediately move the tank into the turbine building.
B. Stop all additions of radioactive niaterial into the tank and allow
the contents to decay to within limits.
6. Reduce the tank contents by transferring radioactive material
back to the condenser.
D. No action is required at this time.
~~~~~~~~~~~~~~~~~~~
Distractcr Analysis: SEC 16-11.17 requires a maximum of 10 Ci in an
outside tank, exclusive of tritium and noble gases. This tank has 11 Ci, but
the non-tritiudnoble gas contribution is decaying at a rate of more than
1Ci/24 brs. Thus after 48 h o w , the total activity = 18-1-7= 11-2=9 Ci
A. Incorrect: Moving the tank does not provide spill/ovefflow
protection and risks further contamination by spillage.
Plausible: Moving the tank indoors technically meets the SLC
requirement.
B. Correct: Within 48 hrs the tank will be within the SLC limit.
C. Incorrect: Transferring water back to the condenser would
recontaminate it and risk spillage.
Plausible: This is a possible option to meet the SLC requirement, and
could be chosen if the candidate cannot detemiine the decay rate of
the tank.
D. Incorrect: Must at least suspend additions to the tank.
Plausible: If the candidate miscalculates the non-tritiwdnoble gas
contribution.
Level: SRO Only 10CFR55.43(b)4
d K A : G2.3.3 (1.8/2.9)
Lesson Plan Objective: WE-WL. Obj: 16
Source: Bank
Level of knowledge: analysis
References:
1. OP-CN-WE-WL page 22
2.SLC 16.11-17pages 1-2-PROVpDED
-. DUKE POWER
...-. .. .......-.. CATAWBA OPERATIONS TRAINING
. ._._- . ......
OP-CN-WE-WL FOR TRAINING PURPOSES ONLY REV. I I
Page 4 Of 23
16.iz RADlObOGlCAL EFFLUENTS CONTROLS
16.31-17 LIQUID HOLDUP TANKS
COMMITMENT:
The quantity of radioactive material contained in each temporary unprotected
outdoor tank shall be limited to less than or equal to 10 Curies, excluding tritium and
dissolved or entrained noble gases.
APPLICABILITY:
At all times.
REMEDIAL ACTION:
With the quantity of radioactive material in any of the above tanks exceeding the
above limit, immediately suspend all additions of radioactive material to the tank,
within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit, and describe the events
leading to this condition in the next Radioactive Effluent Release Report. pursuant to
Technical Specification 5.6.3.
TESTING REQUIREMENTS:
The quantity of radioactive material contained in each of the above tanks shall be
determined to be within the above limit by analyzing a representative sample of the
tank's contents at least once per 7 days when radioactive materials are being added
to the tank.
1, Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical
Specifications Amendments for Catawba, September 30, 1998.
2. Technical Specification 5 5 4 2 7Explosive Gas and Storage Tank
Radioactivity Monitoring Program.
BASES:
The tanks induded in this COMMITMENT are all those outdoor radwaste tanks that
are not surrounded by liners, dikes or walls capable of holding the tank contents and
Chapter 16.11-17 Page 1 of 2 01/16/99
-
16.11 RADIOLOGICAL EFFLUENTS CONTROLS
16.11-17 LlQUlD HOLDUP TANKS
COMMITMENT:
The quantity of radioactive material contained in each temporary unprotected
outdoor tank shall be limited to less than or equal to 10 Curies, excluding tritium and
dissolved or entrained noble gases.
APPLICABILITY:
At all times.
REMEDIAL ACTION:
With the quantity of radioactive material in any of the above tanks exceeding the
above limit, immediately suspend all additions of radioactive material to the tank,
within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit, and describe the events
leading to this condition In the next Radioactive Efnuent Release Report, pursuant to
Technical Specification 5.6.3.
TESTING REQUIREMENTS:
The quantity of radioactive material contained in each of the above tanks shall be
determined to be within the above limit by analyzing a representative sample of the
tanks contents at least once per 7 days when radioactive materials are being added
to the tank.
REFERENCES:
1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical
Specifications Amendments for Catawba, September 30, 1998.
2. Technical Specification 5.5.12, Explosive Gas and Storage Tank
Radioactivity Monitoring Program.
BASES:
The tanks included in this COMMITMENT are all those outdoor radwaste tanks that
are not surrounded by liners, dikes or walls capable of holding the tank contents and
Chapter 18.13-17 Page 1 of 2 01/16/99
BASES (con't)
that do not have tank overflows and surrounding area drains connected to the Liquid
Radwaste Treatment System.
Restricting the quantity of radioactive material contained in the specified tanks
provides assurance that in the event of an uncontrolled release of the tank's
contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix 5, Table II, Column 2, at the nearest potable water supply and the
nearest surface water supply in an UNRESTRICTED AREA.
Chapter 16.$1-57 Page 2 o f 2 O i l 1 6/99
-- DUKE POWER CATAWBA OPERATIONS Tf?A!Mh!G
D. S.L.6. 46.11-2- Radioactive liquid effluent monitoring instrumentation.
E. S.L.C. 16.11-3 Dose - Limits quarterly and yearly dose to members of the
Public at all times.
1. Quarterly
a) Less than or equal to 1.5 mrem whole body
b) Less than or equal to 5 mrem any organ
2. Yearly
a) Less than or equal to 3 mrem whole body
b) Less than or equal to 10 mrem any organ
F. S.L.6. 16.11-4 Liquid Radwaste Treatment System shall be operable when
the liquid effluent, from each unit, to unrestricted areas would exceed 0.06
mrem whole body or 0.2 mrem to any organ in a 31 day period.
G. S.L.C. 16.11-7 Radioactive Gaseous Effluent Monitoring Instrumentation.
Table 16.11-5.
H. S.L.C. 16.11-17 Liquid Holdup Tanks -The quantity of radio-active material
contained in each temporary unprotected outdoor tank shall be limited to
less than or equal to 10 Curies, excluding tritium and dissolved or entrained
noble gases at all times.
I. Design Basis Specifications - Wb
1. NDINS Sump Pumps and level switches are safety related, and are
required to be operable. At least 2 pumps shall be operable at all times.
2. CAPT Sump Pumps (?Aand 2A) are also safety related. If either of
these sump pumps are inoperable, then the CAPT is declared
inoperable and the SSF is placed in degrade.
2.6 Annunciator Response (Obi. # ) - Refer to actual Annunciator Response for
II
current and complete actions and information.
2.7 Review Limits and Precautions (Obj. #17).
A. 0P/0/B/6500/014 (Operations Liquid Waste Release)
B. OP/?/N6500/009(Miscelianeous Equipment Brains to WL System)
C. OP/l/Ai6500/014 (Operations Controlled Liquid Waste Systems)
3. Summary
3.1 Review Objectives
OP-CN-WE-WL FOR TRAINING PURPOSES ONLY REV. f l
Page 22 of 23
Bank Question: 096. f Answer: B
1 Pt(s) Unit 1 is operating in mode 3 preparing for a reactor startup following a
refueling outage. Given the following events and conditions:
- NC Pump IC is running.
Reactor trip breakers are tagged open.
Maintenance determincs that the MOV test data from the outage
indicates that the torque switches for 1ND-6SB (NO TR" I B IIOT
LEG INJISQL) have been set too low.
Which one of the following statements correctly describes the operating
restrictions and implications of tagging closed 1ND-6SB?
REFERENCES PROWDED: - Tech Spec's w/Bases 3.4.5,3.4.6,3.5.2
A. 1ND-65B may be tagged closed for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, if the steam
generator in the running NC loop is operable.
B. 1ND-65B may not be tagged closed because this would make both
trains of MD inoperable.
C. 1ND-65B may not be tagged closed, unless two NCPs are running
with operable steam generators.
D. 1ND-65B may be tagged closed, if 1ND65B is restored to
operation prior to transitioning to mode 2.
Distracter Analysis:
A. Incorrect: Both trains of hl)will be inoperable.
Plausible: If the candidate assumes 1 S i 6 and the A ND loop.
B. Correct: ND-65 prevents ND flow to all 4 loops.
6. Incorrect: Both trains of ND will be inoperable.
Plausible: If the candidate focuses only on decay heat removal.
D. Incorrect: Both trains of ND will be inoperable.
Plausible: Ifthe candidate assumes that one ND train is sufficient In
mode 3.
Level: SRO Only IOCFR55.43@)2
KA: G 2.2.24(2.6/3.8)
Lesson Plan Objective: PS-NB SEQ 11
Source: Bank Catawba NRC 2000
Level of ktiowledge: comprehension
References:
1. OP-CN-PS-ND pages 12.18
2. Tech Spec & Bases 3.4.5 -PROVIDED
3. Tech Spec & Bases 3.4.6 -PROVIDED
4. Tech Spec &Bases 3.5.2 -PROVIDED
DUKE POWER..........................
-....... CATAWBA OPERATIONS
.........
TRAINING
....... ..
-
P
Objective T
R
Q
I
X
Describe ND system startup
Describe NB system operation in parallel mode
Describe establishing pressurizer spray from the ND system
-
X
X
in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
-
Time: 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />
QP-CN-PS-ND FOR TRAINING PURPOSES ONLY REV.29
Page 5 of I 9
DUKE POWER _c_ - CATAWBA OPERATIONS TRAINING
c) Local Temperature indication downstream of ND Hx (NDT 5020,
5030)
5. Flow Indications (Obj. #8)
a) Downstream of ND Hx's (NDFT 5180.5190)
1) Indication in CIR (MC-11) for Cold legs flow A&B (C&B)
2) Used for auto control of NB Hx Bypass valves
I. NV System letdown Isolation ND-24A f58B)
1. Motor-operated, normally closed
2. Supplies letdown from NB to NV
3. Operated from MC-11 (Obj. #8)
J. Discharge Paths
1. ND Injection to the Cold legs. A & 8 (C & D)
a) Train cross ties. ND-3% (65B) (Obj. #$)
1) Motor operated gate valves
2) Normally open, to ensure flow to all four cold legs (FSAR
Commitment) even if one train fails
3) Close during recirculation phases foBlowing a LOCA.
b) Containment Isolation valves NI-173A (1788) with 3 position select
switch. (Obj. %)
1) Normally opened with power removed to ensure injection Row;
DISCONNECT positron.
2) Throttle position was added for use as a means of controlling
ND flow in the event normal flow control is lost (loss of VI)
during partial drain operations which would cause ND-26 & SO
to fail open resulting in vortexing.
3) In the ENABLE position the valves will open or close without
throttle capability.
c) Flow limiting orifices in discharge line to ea& Cold Leg to balance
Row during injection.
d) Flow indication provided in Cia.
e) Annunciator, AD-9, for "ND TKN A(B) to NC C-LEGS LOOPS C-
D(A-B) LO FLOW set at 1050 gpm and ND purn'p running (time
delay keeps alarm functional two (2) minutes following breaker trip.
9 Discharge relief set at 600 psig relieves to the Recycle Holdup Tank.
1) To protect against backleakage from the NC system through the
discharge flowpath check valves.
OP-CN-FS-ND FOR TRAINING FURFOSES ONLY REV.29
Page ?2of 19
DUKE POWER
.............................. ......_ _ .....
CATAWBA
. .
OPERAT/ONS TRAINING
. nxmm,irrr
2) Design analysis has shown that ND flow through the hot leg
injection header is not required to dilute the boron concentration
at the top of the core. One NI pump injecting to its respective
hot legs provides sufficient flow to perform this task.
3) The shift to hot leg recirculation for the NI train will ensure that
heat removal from this upper area of the core is accomplished
and the boron concentration is diluted...
4) The ND system will be aligned for hot leg recirculation only if
flow cannot be established from at least one NI train.
5) The NV pumps continue to discharge to the cold legs to ensure
suficient core cooling in the event of a hot leg break.
b) Hot leg recirc. transfer is made with the NI or ND pumps running
c) Procedure Summary
1) Nl Pumps are aligned for Hot Leg Recirculation
2) If flow is verified from at least one train of the NI system the
operators are directed to return to the procedure and step in
effect.
3) If flow is not verified from at least one train of the NI system, the
ND pumps will be aligned for Hot Leg Recirculation as follows:
(a) Verify the ND loop suction isolations and the crossover
isolations to the cold legs are closed.
(b) Open the hot leg injection isolation
(c) For any ND train not aligned for auxiliaty containment
spray, dose the cold leg isolation and open the
associated crossover isolation valve.
(d) Ensure monitor light panel in correct alignment for hot leg
recirculation and return to procedure and step in effect.
d) Hot leg recirculation flow is to NC loops " B and "C" from the ND
system.
e) The ND pumps will continue to supply the NV and NI pump suction.
2.4 Technical Specifications (Obj. #I 1 & #12)
A. 3.4.6 RCS Loops - Mode 4
B. 3.4.7 RCS Loops - Mode 5, Loops Filled
C. 3.4.8 RCS Loops - Mode 5, Loops Not Filled
D. 3.5.2 ECCS-Operating
E. 3.5.3 ECCS-Shutdown
QP-CN-PS-ND FQR TRAINING PURPOSES ONLY REV.29
Page 18 of 19
ECCS .-Operating
3.5.2
3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
3.5.2 ECCS-Operating
LCO 3.5.2 Two ECCS trains shall be OPERABLE'.
APPLICABILITY MODES 1 ,2, and 3,
In MODE 3, both safety injection (SI) pump flow paths may be isolated by
closing the isolation valves for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform pressure isolation
valve testing per SR 3.4.14.1
ACTIONS
CONDITION REQUIRED ACTION COMtXETION TIME
A. One or more trains A.l Restore train($ to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />' I
inoperable. OPERABLE status.
At least iOO% of the
ECCS how equivalent to
a single OPERABLE
ECCS tnin available.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
1 6.2 Be in MODE 4 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
'For eaee ECCS train on Unit 2. the Completion Time that one ECCS Uain can be inoperable a8 specikd by
Required Actian A.1 may be extended beyond the 72 hour8 up io 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> as pati Of ihe NSWS system>Upgrades.
System upgrades include maintenance and modification activilles 89MCialed wilh cleaning OF NSWS poing. va!ves
and branch line?, neressav repairs a~,bior r s p t a c m t , vahe repair andior repla.lacement.and replacement ai
pCltBn~of the NSWS pip!ng io the AFW sy8tenl. Upon completior of the cleaning. up~radees.and sys!em
restoration in refueling oiltage 1 EOCIZ. this footnote is no longer applicable.
Catawba Units 1 and 2 3.5.2-1 Amendment Nos. 189!182
RCS Loops - MODES 4
3.4.6
3.4 REACTOR COOLANT SYSTEM (RCS)
3.4.6 RCS LOODS-MODE 4
LCO 3 4 6 Two loops consisting of any combination of RCS loops and residual heat
removal (RHR) loops shall be OPERABLE, and one loop shall be in
operation
1. All reactor coolant pumps (RCPs) and RHR pumps may be
de-energized f o r 5 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period provided:
a. No operations are permitted that would cause reduction of the
b. Core outlet temperature is maintained st least 10°F below
saturation temperature.
2. No RCP shall be started with any RCS cold leg tsmperature 5 285°F
unless the secondary side water temperature of each stearn
generator (SG) is 5 50°F above each of the RCS cold leg
temperatures.
APPLICABILITY MODE 4
ACTIONS
CONDITION REQUIRED ACTION COMPLETION TIME
A. OneRCSloop A.l Initiate action to restore a Immediately
OPERABLE. second loop to
OPERABLE status.
A Q
Two RHR [oops
C a t a d a Units 1 and 2 3.4.6-1 Amendment Nos. 1731165
RCS LoODs - MODES 4
3.4.6
COMPLETION TIME
8. One RHR loop B1 Be in MODE 5 . 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
-
AND
ALL RCS loops
ineperable.
C Both required RCS or C1 Suspend all operations Immediately
RHR loops inoperable involving a reductlon of
KCS boron eollcentratlon
w
-
No RCS or RHK loop in
weratin C.2 lnniate action to restore Immediately
I one bop io OPERABLE
status and operation.
S U R V E I U N C E REQUIREMENTS
I
SURVEILLANCE FREQUENCY
SR 3.4.6 1 Verify one RHR or RCS loop is in operation. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
range for required RCS loops.
Catawba Units I and 2 3.4.6-2 Amendment Nos. 1931165
RCS Loops - MODES 3
3.4.5
3.4 REACTOR COOLANT SYSTEM (RCS)
3.4.5 RCS Loops - MQDE 3
LCO 3.4.5 Three RCS loops shall be OPERABLE. and either:
a. Three RCS loops shall be in operation when the Rod Control System
is capable of rod withdrawal; or
b. One RCS Imp shalt be in operation when the Rod Control System is
not capable of rod withdrawal.
All reactor coolant pumps may be de-energized for 5 Ihour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
period provided:
a. No operations are permitted that w w l d cause reduction of the RCS
boron concentration; and
b. Core outlet temperature is maintained at least 10°F below saturation
temperature.
APPLICABILITY: MQDE 3
ACTIONS
CONDITION
A. Qne w two required A.I Restore required RCS 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
RCS loop(s) inoperable. loop($) to OPERABLE
status.
8. Required Action and B.1 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Campletion
Time of Condition A not
met.
E (continued)
Catawba Units 1 and 2 3.4.5-1 Amendtnent Nos. 173/165
RCS Loom - MODES 3
3.4.5
ACTIONS pontinued)
CONDITION REQUIRED ACTION COMPLETION TiME
C. One or two required 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
RCS loop($) not in loop(s) to operation.
operation and Rod
Control System capable
of rod withdrawal.
C.2 De-energize all control rod 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
drive mechanisms
(CRDMs).
D. Three required RCS 8.1 De-energize a11 CRDMs Immediately
IOODSinooerable.
No RCS loop in
operation
IAN0
D.2 SusDend all Operations
invoking a reduction of
Immediately
@
0.3 hitiat@action to restore lmmediatelv
one RCS loop io
OPERABLE status and
operation.
Catawba Units 1 and 2 3.4.5-2 Amendment NUS. 1731165