ML030430608
| ML030430608 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 08/05/2002 |
| From: | Ernstes M Operator Licensing and Human Performance Branch |
| To: | Scalice J Tennessee Valley Authority |
| References | |
| 50-327/02-301, 50-328/02-301 50-327/02-301, 50-328/02-301 | |
| Download: ML030430608 (382) | |
Text
Draft Submittal (Pinlk Pnna*r)
SEQUOYAH NUCLEAR PLANT EXAM 2002-301 50-327 & 50-328 DECEMBER 2 - 6, 2002 Reactor Operator Operator Written Exam
Draft Submittal (Pink Paper)
SEQUOYAH NUCLEAR PLANT EXAM 2002-301 50-327 & 50-328 DECEMBER 2 - 6, 2002 Senior Reactor Operator Written Exam
Sequoyah Nuclear Plant 12/2002 NRC EXAM 12(2002 NRC EXAM Written Exam
- Cross-Reference
- RO Questions
"*Cross-Reference
"*SRO Questions
"*RO Exam
"* RO Key
"*SRO Exam
"*SRO Key Tennessee Valley Authority FI%
( C RO/SROIKA Cross Reference I
RO/SRO M=Memory Type K/A ID Question S=SRO C=Comph A=As Is ID R=RO A=Analysis M=Modified B=BOTH N=New 001 AK1.03 AOP-C.01-B.1 002 B C A 001 K.301 001 K.301 B A N 001 K5.30 CONTROL*RODS 020 B M A
B A 003 A1.03 RCP-B.120602 M A 004 A3.07 K6.19 002 001 Bt 004 006 A3.07 ECCS-B.3 RHR-B.13.H n C N OOf355A1.01 A1.01 004 RCP-B.
RHR-B.13.H 0 A1.07 001 B M N 33Al1.07 K6.19 ECCS-B.3 002 B M A OOE K16.19 006 K6.19 B C N U08 K4.02 CCS-B.9.A 001 B A A 009 EK1.01 INPO3 955 B A An fin t .. - - .1 C U 10 2.0U2 010 A2.02 Bq M N I .... I U 12 K6.U0 012 K6.07 C N P
013 K4.01 013 K4.01 I -I-' N I - I M 015 AK2.08 AUO--H.05-B.5 001 I
R M A I .... ... .
U01 A3.U2 RCS-TEMP-B.2 008 P 016 K1.01 INCORE-B.1.B 002 B M A 017 K6.01 INCORE-B.1.D 003 B M A 022 K2.01 022 K2.01 B C N 024 AA1.26 ES-0.1-B.1 002 B C A 025 AK2.05 FR-Z.1-B.2 001 B C A 025 K5.02 CTMT-B.11 001 B C A 027 AK1.02 OPL271C353.4 001 B C A I
C 7O/SRO FR*O/SRO
,=SRO M=Memory A=As Is Type C=Comph K/A ID Question R=RO A=Analysis M=Modified ID 3=BOTH B N=New RPS-B.5.B 001 C A 029 EK2.06 B N 034 A1.02 034 A1.02 B A 034 K4.02 FH-B.12.B 001 M A B
035 K5.01 C N 035 K5.01 B AOP-R.01-B.2 003 A A 037 2.1.7 B 037 WA.11 AOP-R.01-B.2 004 M A B
038 EK3.08 INPO2 976 B FW-B.5 001 A A 039 K3.04 B 040 AK1.01 INPO2 231 B A 051 AK3.01 SDCS-B.12 023 B 058 AK3.01 AOP-P.02-B.4 002 B A M N 059 A4.01 059 A4.01 B C 059 K1.04 MFW 001 B C
M N 060 AK2.02 060 AK2.02 B A 061 AK2.01 M N 061 AK2.01 B M 061 -K2.01 A 061 K2.01 B M 061 K6.01 061 K6.01 B M N M
062 A4.03 INPO8 155 B M A 063 A2.01 063 A2.01 B M N 063 K4.04 FW-B.5.B 004 B C A 064 K2.02 064 K2.02 B M A 068 A3.02 RADWASTE-B.12 008 B C 069 AA1.03 CTMT-B.5 004 B A M
F6'71 Al-06l RMS-B.9 001 2
( (
RO/SRO M=Memory Type K/A ID Question S=SRO C=Comph A=As Is ID R=RO B=BOTH A=Analysis M=Modified N=New 072 A1.01 RMS-B.2 004 B M A 074 EK1.01 FR-C.1-B.2 003 B C A 074 EK2.02 FR-C.1-B.2 011 B A A 075 K1.01 RCW-B.9 001 B M A 076 AK3.05 076 AK3.05 B C N 078 K1.01 AIR-B.12 002 B C A 079 K4.01 AIR-B.5 014 B C A 103 A3.01 103 A3.01 B M N 2.1.24 2.1.24 B A N 2.2.11 CTMT-B.11 004 B M A 2.2.4 2.2.4 B A N 2.3.11 ODCM-B.5 001 B C A 2.3.9 CTMT PURGE-B.4 001 B M A 2.4.39 REP-B.1.D 002 B M A 2.4.4 AOP-R.02-B.2 001 B C A E01 EA1.1 ES-0.0-B.3 001 B C A E03 EA1.2 ES-1.2-B.2 006 B C A E03 EK1.3 ES-1.2-B.2 004 B M A E04 EK2.2 ECA-1.2-B.1 002 B M A E05 EA1.3 FR-H.1-B.3 004 B C A E08 EK3.2 FR-P.1 001 B C A E09 EA1.1 OPL271C382.4 001 B A A E10 EK3.1 ES-0.2-B.3 003 B C A E14 EK1.3 ECCS-B.2 002 B M A E15 EK1.1 FR-Z.2-B.2 001 B M A 3
C' C C RO/SRO M=Memory Type K/AID Question S=SRO C=Comph A=As Is R=RO A=Analysis M=Modified ID B=BOTH N=New 003 K4.02 T/S0304.02 003 R C A 005 AA2.01 005 AA2.01 R C N 008 A2.04 008 A2.04 R C N 012 K2.01 RDCNT-B.7.B 001 R C A 015 A2.03 FISSION*PROD*POISON 052 R M A 022 2.4.27 022 2.4.27 R A N 0252.2.13 0252.2.13 R M N 029 A1.02 P1-B.10 001 R M A 039 K5.05 RVINT-B.8 001 R M A 056 A2.04 056 A2.04 R C N 056 AK3.01 OPL271C368.3 001 R M A 056 K1.03 056 K1.03 R M N 059 AA2.05 059 AA2.05 R M N 064 A3.06 D/G-B.10 002 R A A 067 AA2.17 AOP-C.04 001 R M A 068 AK2.07 D/G-B.6 012 R M A 068 K4.01 RADWASTE-B. 12 006 R M A 071 A4.26 WGDS-B.12 001 R C A 078 K3.03 078 K3.03 R C N 0792.1.1 0792.1.1 R M N 086 K3.01 086 K3.01 R C N 2.1.16 2.1.16 R M N 2.1.31 PZR PRESS-B.9 006 R C A 2.2.26 REFUELING-B.1.G 001 R M A 2.3.10 2.3.10 R C N 4
C C RO/SRO M=Memory Type K/A ID Question S=SRO C=Comph A=As Is R=RO A=Analysis M=Modified ID B=BOTH N=New 2.3.2 RADIATION 002 S M A 2.3.3 AOP-C.04-B.5 008 S M A 2.4.33 OPDP-4 002 S M A 2.4.45 AOP-M.03-B.1 002 S C A E01 EA2.1 ES-0.0-B.5 002 S C A E04 EA2.1 ECA-1.2-B.2 001 S C A E09 EA2.1 ES-0.2-B.3 001 S C A E12 2.4.16 ECA-2.1-B.1 004 S A A E15 EA2.1 E15 EA2.1 S M N E16 2.4.41 E16 2.4.41 S C N 6
HLC12-02.BNK Page: 1 Wednesday, October 09,2002 @ 11:45 AM
- 1. 001 K3.01 001
"-' s functioning The unit is at 100% pow er w ith no Xenon transients and all system normally. When the following annunciators are received:
"TS-68-2P/Q Reac Cool Loops T Ref T Auct High-Low" "Rod Control System Urgent Failure" Tref is 40F above Tave.
Which one of the following is correct for this condition?
A. Raise turbine load at, borate.
B. Lower turbine load ad borate.
C. Raise turbine load ;borate.
,,D. Lower turbine load rr dilute.
load change OR A. Incorrect, per AOP-C.01 (Tref is > Tave). Step 1 RNO only allows change in RCS boron concentration. Increasing turbine load is not correct (Tref is > Tave). Boration is not correct (Tref is > Tave).
<> B. Incorrect, Step 1 RNO only allows load change OR change in RCS boron concentration. Boration is not correct (Tref is > Tave).
C. Incorrect, per AOP-C.01 (Tref is > Tave). change OR D. Correct, per AOP-C.01 (Tref is > Tave). Step 1 RNO only allows load is > Tave) OR change in RCS boron concentration. Either lower turbine load (Tref dilute (Tref is > Tave).
K/A[CFR]: 001 K3.01 [2.9/3.0] [41.7]
Reference:
1-AR-M5-A (C-6)
AOP-C.01 R8 section 2.1 step 2 RNO.
LP/Objective: OPL271RDCNT B.9 History: New question.
Level: Analysis Comments: FHW 12/02 001 K3.01
ROD CONTROL SYSTEM MALFUNCTIONS AOP-C.01 SON Rev. 8 STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Failure of a Control Bank to Move in AUTO
- 1. ENSURE rod control in MAN.
- 2. POSITION control rods to minimize PERFORM the following:
T-avg. - T-Ref deviation.
"* ADJUST turbine load OR
"* ADJUST RCS boron concentration.
- 3. MONITOR following parameters to ensure core power distribution is within normal limits:
"* PowerrangeNIS
"* T-avg. and AT channels
"* Incore T/Cs
"* Delta flux (Al)
Page 4 of 39
20 (C-6)
) Source Setpoint TS-68-2P/Q SER 2114, 2115 Tref +/- 3°F compared with TS-68-2P/Q, TS-68-2Q/P Tauct.
REAC COOL LOOPS T REF T AUCT HIGH-LOW Probable 1. Normal heatup, cooldown, power ascension or descension Causes inprogress with controls in MANUAL.
- 2. The reactor control system, steam dump system or S/G PORV's are not maintaining Tavg on program.
- 3. Inadvertent dilution or boration.
- 4. Channel testing and/or malfunctioning.
- 5. Secondary system leak.
Corrective [1] COMPARE plant indicators to verify validity of alarm.
Actions [2] IF controls are in AUTO when alarm occurs, THEN PLACE rod control system (1-HS-85-5110) in manual and match Tavg with Tref.
[3] IF rod control system is malfunctioning, THEN GO TO AOP-C.01, Rod Control System Malfunctions.
[4] IF a boron concentration change is suspected, THEN GO TO AOP-C.02, Uncontrolled RCS Boron Concentration Changes.
[5] IF a steam line or feedwater line break or leak is suspected, THEN GO TO AOP-S.05, Steam Line or FeedwaterLine Break/Leak.
[6] IF Tavg channel failed, THEN GO TO AOP-I.02, RCS Loop RTD Instrument Malfunction.
[7] IF Tref (P1-1-73) failed, THEN GO TO AOP-1.08, Turbine Impulse Pressure Instrument Malfunction.
[8] IF alarm is valid, Reactor is critical, and Tavg < 551 OF, THEN PERFORM O-SI-SXX-068-127.0, RCS and Pressurizer Temperature and Pressure Limits.
[9] EVALUATE Technical Specifications (3.3.1 and 3.3.2).
References 45B655-05A-0, 47B601-68-4
CVCS Student Handout Revision 0 Page 8 of 38 INTRODUCTION:
The chemical and volume control system is a major auxiliary system that functions to control RCS inventory, to maintain RCS chemistry, to provide the reactor coolant pumps with seal water, and to control the soluble poison concentration of the RCS. In addition, the CVCS supplies borated coolant to the RCS in the event of an accident.
pressurizer Major system interfaces include the pressurizer level control system, which varies charging flow based upon and corrosion inhibiting chemicals to be level. The reactor make-up system that allows boric acid, demineralized water, uses a added via the CVCS during normal at-power and shutdown conditions. The emergency core cooling systems accident portion of the system piping and the charging pumps to provide high-pressure injection of borated water during conditions.
Purposes
- 1. Maintains programmed pressurizer level;
- 2. Controls reactor coolant chemistry and activity;
- 3. Adjusts and controls reactor coolant boron concentration;
- 4. Provides reactor coolant make-up;
- 5. Supplies seal water to the reactor coolant pumps;
- 6. Portions used for emergency core cooling and emergency boration;
- 7. Used to fill and hydrostatically test the reactor coolant system.
Subsystems
- 1. Charging, letdown, and seal water system
- 2. Chemical control, purification, and make-up system Design Bases
- 1. Reactivity control
- 2. Boric acid used as a neutron absorber for following reactivity changes:
- 3. RCS temperature change between cold shutdown and hot standby.
- 4. Burnup of fuel and burnable poisons.
- 5. Xenon transients.
- 6. Buildup of fission products in the fuel.
Make-up control
- 1. System capable of boration of RCS from either of two paths and from either of two boric acid sources.
- 2. Amount of boric acid stored exceeds amount needed to:
- 3. Borate RCS to cold shutdown with highest worth rod stuck out.
- 4. Go to hot shutdown.
- 5. Compensate for subsequent xenon decay.
- 6. System capable of counteracting inadvertent positive reactivity due to maximum boron dilution accident.
RCS inventory regulation Pressurizer inventory maintained in allowable range for all operating modes as well as small RCS leaks.
The charging and letdown capacities permits heat-up and cool-down of RCS at designed rate with pressurizer level
"* normal.
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 4
- 4. 006 K6.19 001 This question has reference material attached.
Unit One is operating at 100% power. On January 10 at 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> the 1 A-A CCP failed to start on an attempted pump swap. The crew entered the appropriate LCO action. Maintenance can NOT get replacement parts until January 14.
Which one of the following is the correct Tech Spec action?
,/A. Be in Hot Standby on January 13 at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />.
B. Be in Hot Standby on January 13 at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />.
C. Be in Hot Shutdown on January 14 at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />.
D. Be in Hot Shutdown on January 14 at 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />.
A. Correct, the action is 72 hrs to operability or Hot Standby in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
January 10 0200 hrs. to January 13 0200 is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> plus 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is January 13 at 0800.
B. Incorrect per reference.
C. Incorrect per reference.
SD. Incorrect per reference.
K/A[CFR]: 006 K6.19 [3.7/3.9] [41.7]
Reference:
LP/Objective: OPL271C079 B.2 History: New question.
Level: Comprehension Comments: FHW 12/02 006 K6.19 Provide a copy of Tech Specs 3.5.2
EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.2 ECCS SUBSYSTEMS - T Greater Than or Equal to 350'F 9vg E
LIMITING CONDITION FOR OPERATIQN 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE centrifugal.charging pump,
- b. One OPERABLE safety injection pump,
- c. One OPERABLE residual heat removal heat exchanger,
- d. One OPERABLE residual heat removal pump, and
- e. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a REPORTABLE EVENT shall be prepared and submitted to the Commission pursuant to Specification 6.6.1. This report shall include a description of the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this report whenever its value exceeds 0.70.
SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves are in the indicated positions with power to the valve operators removed:
- UNIT 1 3/4 5-4 Amendment No.
MAY i [in'ýf 36, R144 140 t
EMERGENCY CORE COOLING SYSTEMS (ECCS)
SURVEILLANCE REQUIREMENTS (Conti nued)
Valve Number Valve Function Valve Position
- b. FCV-63-22 SIS Discharge to Common Piping
- b. At least once per 31 days by:
- 1. Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and
- 2. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:
- 1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and
- 2. Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.
- d. At least once per 18 months by:
- 1. Deleted.
R143
- 2. A visual inspection of the containment sump and verifying that by debris and the subsystem suction inlets are not restricted etc.) show no that the sump components (trash racks, screens, evidence of structural distress or corrosion.
- e. At least once per 18 months, during shutdown, by:
- 1. Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal and automatic switchover to containment sump test signal.
MAY 1 1990 Amendment No. 92,139, 140 R144 I SEQUOYAH - UNIT I 3/4 5-5 Correction letter of 05/1619o
EMERGENCY CORE COOLING SYSTEMS (ECCS)
SURVEILLANCE REQUIREMENTS (Continued)
- 2. Verifying that each of the following pumps start automatically upon receipt of a safety injection signal:
a) Centrifugal charging pump b) Safety injection pump c) Residual heat removal pump
- f. By verifying that each of the following pumps develops the indicated discharge pressure on recirculation flow when tested pursuant to Specification 4.0.5:
- 1. Centrifugal charging pump Greater than or equal to 2400 psig
- 2. Safety Injection pump Greater than or equal to 1407 psig
- 3. Residual heat removal pump Greater than or equal to 165 psig
- g. By verifying the correct position of each mechanical stop for the following Emergency Core Cooling System throttle valves:
- 1. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.
- 2. At least once per 18 months.
Charging R144 1 Pump Injection Safety Injection Cold Safety Injection Hot Throttle Valves Leg Throttle Valves Leg Throttle Valves Valve Number Valve Number Valve Number
- 1. 63 - 582 1. 63 - 550 1.63-542
- 2. 63 - 583 2. 63 - 552 2.63-544
- 3. 63 - 584 3. 63 - 554 3.63-546
- 4. 63 - 585 4. 63 - 556 4.63-548 SEQUOYAH - UNIT 1 3/4 5-6 Amendment No. 140 MAY 11. g990
EMERGENCY CORE COOLING SYSTEMS (ECCS)
SURVEILLANCE REQUIREMENTS (Continued)
- h. By performing a flow balance test during shutdown following completion of modifications to the ECCS subsystem that alter the subsystem flow characteristics and verifying the following flow rates:
- 1. For safety injection pump lines with a single pump running:
- a. The sum of the injection line flow rates, excluding the R1441 highest flow rate is greater than or equal to 443 gpm, and
- b. The total pump flow rate is less than or equal to 675 gpm.
- 2. For centrifugal charging pump lines with a single pump running:
- a. The sum of the injection line flow rates, excluding the highest flow rate is greater than or equal to 309 gpm, and R1441
- b. The total pump flow rate is less than or equal to 555 gpm.
- 3. For all four cold leg injection lines with a single RHR pump running a flow rate greater than or equal to 3931 gpm. R1441 SEQUOYAH - UNIT 1 3/4 5-7 Amendment No. 50, 140, MAY 11 1990
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 6
- 6. 010 A2.02 001 The unit is operating at 100% power when the controlling pressurizer pressure channel fails high.
Which one of the following is the correct system response and the required action to mitigate the event?
A. Only the pressurizer spray valve associated with the failed channel will open and the operators manually closes this valve.
,/B. Both pressurizer spray valves will open and the operator manually closes these two valves.
C. No pressurizer spray valves will open due to a pressurizer pressure channel bistable interlock and the operator places the spray valve controllers in manual.
D. Both pressurizer spray valves will open and the auxiliary pressurize ray valve will open if selected to the failed channel and the operator ma ly closes all three valves.
A. Incorrect the pressurizer pressure controller will open both spray valves.
B. Correct per references.
C. Incorrect the bistable is associated with the PORV.
D. Incorrect the pressurizer pressure control channels do not have input to the auxiliary spray valve.
K/A[CFR]: 010 A2.02 [3.9/3.9] [41.5 43.5]
Reference:
1-47W611-68-3 AOP-I.04 section 2.1.
LP/Objective: OPL271PZRPCS B.14 History: New question.
Level: Memory Comments: FHW 12/02 010 A2.02 1/2>
Revi.5 SON PRESSURIZER INSTRUMENT MALFUNCTION STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Pressurizer Pressure Instrument Malfunction NOTES:
- A failure of channel Ill (P-68-323) will affect the automatic actuation of PCV 68-334, PZR PORV, in the normal pressure control circuit. LTOPS operation of this PORV is unaffected by this failure.
- A failure of channel IV (P-68-322) will affect the automatic actuation of PCV 68-340A, PZR PORV, in the normal pressure control circuit. LTOPS operation of this PORV is unaffected by this failure.
- 1. MONITOR pressurizer pressure stable or RESTORE pressurizer pressure USING trending to desired pressure. manual control of the following:
"* PZR Spray controllers PIC-68-340D (Loop 1)
AND/OR PIC-68-340B (Loop 2)
" Pressurizer Heaters
<7 Page 4 of 57
SQN PRESSURIZER INSTRUMENT MALFUNCTION AOP-I.04 Rev. 5 Page 1 of 1 APPENDIX H PRESSURIZER PRESSURE CONTROL z
(i 71 cea ro Page 56 of 57
Wednesday,v October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 7 Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 7
- 7. 012 K6.07 001 Unit One is operating at 100% power. Instrument Mechanics are performing a surveillance which required removing RCS Loop 1 AT/Tavg instrument T-68-2 from service. The bistables associated with this loop have been placed in the tripped position.
Which one of the following is correct for this condition?
A. If loop 2 AT/Tavg instrument T-68-25 was removed from service and the bistables associated with this loop were tripped; Unit One would remain at power after a OTAT turbine runback.
,/B. Ifloop 2 AT/Tavg instrument T-68-25 was removed from service and the bistables associated with this loop were tripped; Unit One would trip.
C. Unit One will remain at power since the runback/trip logic has now changed to 2 of 3 remaining loops.
D. Tech specs will require Unit One to be in Hot Standby if the RCS Loop 1 AT/Tavg instrument T-68-2 is NOT declared operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
A. Incorrect, if all bistables associated with a second loop were tripped then the 1 of 3 trip logic would be made and the unit will trip.
B. Correct, if all bistables associated with a second loop were tripped then the 1 of 3 trip logic would be made and the unit will trip.
C. Incorrect, the runback/trip logic is changed to 1 of 3 loops.
D. Incorrect, per TS 3.1.1.1 Table 3.3-1. 7., action 6 the required action time is 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
K/A[CFR]: 012 K6.07 [2.9/3.2] [41.7]
Reference:
AOP-l.02 R1 Appendix A page 5 of 12.
Tech Specs 3.3.1.1 Table 3.3-1 and action 6.
LP/Objective: OPL271 RPS B.19 History: New question.
Level: Comprehension Comments: FHW 12/02 012 K6.07
3/4.3 INSTRUMENTATION 3/4.3.1 RTRIP SYSTM -NSTRU ATION LIMITING CONDITION FOR OPERATION and 3.3.1.1 As a minimum, the reactor trip system instrumentation channels IR194 interlocks of Table 3.3-1 shall be OPERABLE.
APPLICABILITY: As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock shall CHANNEL be demonstrated OPERABLE by the performance of the CHANNEL CHECK, at the TEST operations for the MODES and CALIBRATION and CHANNEL FUNCTIONAL frequencies shown in Table 4.3-1. Ri6 OPERABLE prior to 4.3.1.1.2 The logic for the interlocks shall be demonstrated92 days. The total each reactor startup unless performed during the preceeding per 18 months interlock function shall be demonstrated OPERABLE at least once during CHANNEL CALIBRATION testing of each channel affected by interlock operation.
function 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip limit at least once per 18 months. Neutron shall be verified to be within its detectors are exempt from response time testing. Each verification shall include at least one train such that both trains are verified at least once per R255 36 months and one channel per function such that all channels are verified at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3.1.
February 29, 2000 3/4 3-1 Amendment Nos. 12, 190, 251
\.*/ SEQUOYAH - UNIT 1
m R R TABLE 3.3-1 C) REACTOR TRIP SYSTEM INSTRUMENTATION FLHINIMUM UNIT TOTAL NO. CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACTION
- 1. Manual Reactor Trip 2 1 2 1, 2, and *
- 2. Power Range, Neutron Flux 4 2 3 1, 2 21
- 3. Power Range, Neutron Flux 4 2 3 1. 2 2 High Positive Rate
- 4. Power Range, Neutron Flux, 4 2 3 1, 2 2#
' High Negative Rate
-S. Intermediate Range, Neutron Flux 2 t
1 2 1, 2, and 3 N 6. Source Range, Neutron Flux A. Startup 2 B. 1 2 2## ,and 4 Shutdown 2 0 1 3, 4 and 5 5
- 7. Overtemperature Delta I Four Loop Operation 4 2 3 1, 2 6#
- 8. Overpower Delta T Four Loop Operation 4 2 3 1, 2 6#
b 9. Pressurizer Pressure-Low 4 2 3 1, 2 6#
S6#
R 10. Pressurizer Pressure--High. 4 2 fr- 11. Pressurizer Water Level--High 3 1, 2 3 2 2 1, 2 6# R145 C~o0 O.0
TABLE 3.3-1 (COntinued)
ACTION 3 - With the number of channels OPERABLE the Minimum Channels OPERABLE POWER level:
one less than required by requirement and with the THERMAL C
- a. Below the P-6 (Block of Source restore the Inoperable channel Range Reactor Trip) setpoint, to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.
b, Above the P-6 (Block of Source but below 5% of RATED THERMAL Range Reactor Trip) setpoint, channel to OPERABLE status POWER, restore the inoperable prior to increasing THERMAL above 5% of RATED THERMAL POWER. POWER
- c. Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.
- d. Above 10% of RATED THERMAL POWER, the provisions of Specification 3.0.3 are not applicable.
ACTION 4 - With the number of channels the Minimum Channels OPERABLEOPERABLE one less than required by requirement and with the THERMAL POWER leveli
- a. Below the P-6 (Block of Source restore the inoperable channel Range Reactor Trip) setpoint, Increasing THERMAL POWER above to OPERABLE status prior to the P-6 Setpoint.
1/2>
- b. Above the P-6 (Block of Source operation may continue.
Range Reactor Trip) setpoint, FA ACTION 5 With the number of channels OPERABLE one less than required the Minimum Channels OPERABLE requirement verify compliance by with the SHUTDOWN MARGIN requirements or 3.1.1.2, as applicable, of Specification within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least 3.1.1.1 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter. once per ACTION S - With the number of OPERABLE Number of Channels, STARTUP channels one less than the Total and/or provided the following conditions POWER OPERATION may proceed are satisfied:
- a. The inoperable channel is placed in the within S hours. tripped condition b.
IR51 The Minimum Channels OPERABLE the inoperable channel may requirement is met; however, for surveillance testing of be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> other channels per Specification 4.3.1.1.1.
ACTION 7 - Deleted.
R1451 May 16, 1990 3,:
SEQUOYAH - UNIT I 3/4 3-6 Amendment No. 47, 141
Rev. 1 SQN RCS LOOP RTD INSTRUMENT MALFUNCTION AOP-I.02 Page 5 of 12 APPENDIX A REMOVING RCS LOOP 1 AT/TAVG INSTRUMENT LOOP T-68-2 (T-411/412) FROM SERVICE A. Setup (cont'd)
- 5. (Continued)
ANNUNCIATORS [XA-55-4J CHECK ('
DARK LIT PROTECTION SET IIBYPASS [B-71 PROTECTION SET III BYPASS [C-7]
PROTECTION SET IV BYPASS [D-7]
- 6. VERIFY Steps 1 through 5 COMPLETE.
CAUTION: If any loop 2, 3, or 4 AT/T-avg trip status light in Step 5 is LIT, completion of this Appendix will initiate one or more of the following signals:
"*OPAT reactor trip
"*OTAT reactor trip
"*OPAT turbine runback/rod withdrawal block
"*OTAT turbine runback/rod withdrawal block
"*Feedwater isolation on low T-avg with reactor trip
"*Loss of P-12 permissive - unblock steam dumps
- 7. NOTIFY Unit Operator to VERIFY current plant status allows removing this AT/TAVG channel from service.
UO Initials Page 14 of 57
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 8
- 8. 013 K4.01 001 Which one of the following is correct concerning th94osot-ea Safety Injection signal?
A. Low Pressurizer pressure input to Safety Injection logic can be manually blocked at any time.
B. Hi Containment pressure input to Safety Injection logic can be manually blocked at any time.
C. Safety Injection can be reset immediately after automatic actuation.
,/D. If the Safety Injection has been reset and the reactor is tripped, all subsequent automatic Safety Injection actuation signals are blocked.
A. Incorrect per reference, manually blocked below P-1 1.
B. Incorrect per reference, no manual block.
C. Incorrect per reference, time delay pick up before reset logic can be made.
D. Correct per reference.
K/A[CFR]: 013 K4.01 [3.9/4.3] [41.7]
Reference:
47W611-63-1 LP/Objective: OPL271RPS B.16 History: New question.
Level: Memory Comments: FHW 12/02 013 K4.01
I II tD
'0 N
C, U,
-n to.
- C -v In N 6, Pm 6, z C
-V In In 6,
to -4 C In In C
6, m
r z
In
-v In 60 L0
-4
-c
-n In C)
C 0
C z-4 C
r 9=5V 10 Li)
-Ti m
c
-4
-v In H
233 3
- S N 202 2
i Ao H -'C 0
6, ri § S 2 77 z 2 g2 .0 S
zC I 2 3
\0 0
0 I
ri C-)
-I I A3 3p 2
No IN 0 C 0) 0-z N
A UP 4 0' Co 1 I,
en H
1 3 3 2
n
- I I II m I IWIrt C I m > [lihill
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 10 10.022 K2.01 001
"*~> Plant Conditions:
- The unit is at 100% power.
- 6.9 kv Shutdown Board "B" trips on a differential fault.
- All D/Gs functioned as designed.
Which one of the following is a result of this condition?
program.
A. Feedwater flow is increased due to more demand by the S/G level out.
B. Train "B" RHR pump will be available when the blackout sequencer times
,C. Available forced flow from the upper compartment coolers is reduced.
g.k -'r QCJ % Z.-
rucd 4L t-s,&s At to30% pvi.
1>."
D. Unit load isprocedurally flow is not A. Incorrect, AOP-P.05/06 requires manual reactor trip, therefore feedwater increased. S/G level is vital power and not load shed.
fault.
B. Incorrect, loads are not available from a board that had a differential BOX and BOY relays will not pick up due to dead bus.
reactor vent board C. Correct upper compartment fans that are fed from the associated are no longer available.
D. Incorrect, AOP-P.05/06 power reduction guidance not given.
K/A {CFR}: 022K2.01 [3.0/3.1] [41.7]
References:
1,2-15E500-1 1,2-45N724-1 1,2-45N755-3 AOP-P.05/06 App T.
LP/Objectives: OPN218E.012 B.4 History: New Question Level: Comprehension Comments: FHW 12/02 022K2.01
Page 1 of 2 APPENDIX T 480V REACTOR BUILDING VENT BOARD 1B-B LOAD LIST COMPT COMPONENT/ LOAD 10 ROP 2 Motor Heater 1D1 Incore Flux Drive Unit 1A 1D2 Incore Flux Drive Unit 1B 1E Incore Flux Drive Unit 1C 2A Containment Purge Air Exhaust Fan 1B 2C RCDT Pump 1B (Note 1) 3B H2 Recombiner 1B-B (Note 1) 3C Incore Instrument Room Purge Supply Fan 3D Incore Instrument Room Purge Exhaust Fan 3E RCP 2 Oil Lift Pump 4A Containment Instrument Room Unit Heater 1B 4B Containment Floor and Equipment Drain Sump Pump 1B (Note 1) 4C Ice Condenser Floor Cooling Pump I B (Note 1) 4D Ice Condenser AHUs (Note 1) 5A Reactor Upper Compartment Unit Heater 1B 5E RCP 4 Oil Lift Pump 6A Ice Condenser Bridge Crane 6B Spreading Room Exhaust Fan B 6C RCP 4 Motor Heater 6D Reactor Lower Compartment Unit Heater 1 B 6E Ice Condenser AHUs (Note 1) 7C Containment Purge Air Supply Fan 1B 7E1 Control Bay Sump Pump 1 (Note 1) 7E2 Reactor Building Jib Crane Note 1: Ifrequired, start redundant train equipment.
Note 2: Monitor temperatures locally.
P Page 74 of 83
Page 2 of 2 APPENDIX T 480V REACTOR BUILDING VENT BOARD 1B-B LOAD LIST COMPT COMPONENT/LOAD 8A Reactor Upper Compartment Cooler Fan 1B (Note 1) 8C Reactor Upper Compartment Cooler Fan 1D (Note 1) 8E1 Water Intake Structure Heat Trace Cabinet (Note 2) 8E2 Reactor Upper Compartment Unit Heater 1D 9A ERCW Cooling Tower Makeup Pump 9C RCC Change Hoist 9E1 Equipment Hatch Hoist 10B Ice Condenser End Wall Door 1 B 10C Ice Condenser Floor Defrost Heater 11B Page 75 of 83
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 12
- 12. 034 A1.02 001 This question has reference material attached.
Unit.ve is in mode 6 and currently performing 0-GO-13. The off oing shift advised yothfe upper and lower internals storage area was dry and is i rocess of filling the refuel cavity.
The Reactor Cavity AUO said the level is on ladder rung number 31.
How many gallons of water is required to reach ladder rung number 21?
A. Approximately 19,000 gallons.
B. Approximately 67,000 gallons.
v/C. Approximately 99,000 gallons.
D. Approximately 143,000 gallons.
A. Incorrect per reference.
B. Incorrect per reference.
C. Correct per reference, per appendix K page 2 each ladder rung is one foot.
Per Appendix K page 1 each foot above elevation 702 requires 9,889 gallons.
Per Appendix K page 2 ladder rung 31 is elevation 702. Rung 31 - 21 is 10 rungs.
10 rungs (10 feet) times 9,889 gallons is approximately 99,000 gallons.
D. Incorrect per reference.
K/A[CFR]: 034 A1.02 [2.9/3.7] [41.5]
Reference:
0-GO-13 R38 Appendix K pages 1 and 2.
LP/Objective: OPL271 C057 B.5 History: New question.
Level: Analysis Comments: FHW 12/02 034 A1.02 Provide a copy of O-GO-13 R38 Appendix K pgs 1 and 2
I SQN 0-GO-13 REACTOR COOLANT SYSTEM Rev: 38 DRAIN AND FILL OPERATIONS Page 16, 5 of 290 0,1,2 II APPENDIX K Page 1 of 2 REACTOR CAVITY/REFUELING CANAL DIAGRAM 733' 7 1/2" Floor Elev.
Top of Curb -733' 9 1/2" Vent Duct Opening M ME ".*-- 726-7" -' Mz Em 726' 1 1/2" Reactor Cavity Rx Vessel Flange 702' Refueling Canal Upper and Lower Internals 11' 3" I Storage Area 690' 10 1/2" I ý'
I Total gallons required to fill Upper and Lower Internals Storage Area from el. 690' 10 1/2" to Rx. Vessel Flange el. 702'. 56,782 gals.
Total gallons required to fill Rx Cavity from ele 702' to 703' 9,889 gals.
(1 ft ele = 9889 gals when Ž_ele 702')
Total gallons required to fill Rx. Cavity-from el. 703' 228,691 gals.
to 726" 1 1/2".
295,362 gals.
Total gallons required to fill Rx. Cavity (initial fill).
102,732 gals.
Total gallons required to fill Fuel Transfer Canal (initial fill).
0-GO-1 3
)N REACTOR COOLANT SYSTEM Rev: 38 F Sc DRAIN AND FILL OPERATIONS Page 166 of 290 0,1,2 APPENDIX K Page 2 of 2 REACTOR CAVITY LEVEL INDICATOR (Refueling Canal Ladder)
RUNG NO.
- EL 733'-9 1/2" (TOP of CURB)
NORMAL
- EL 725'-10 1/2" TECH SPEC WATER MIN. LEVEL LEVEL 725'-1 1/2" 726'-1 1/2"
- EL 717'-10 1/2"
- EL 713'-10 1/2" REACTOR VESSEL
- EL 701'-10 1/2" FLANGE 702'
- EL 691'-10 1/2" - REFUELING CANAL FLOOR 690'-10 1/2" REFUELING CANAL (EQUIPMENT PIT) LADDER 42 RUNGS SPACED AT 1V-0"
Wednesday, October 09,2002 @ 11:45 AM HLC12-02.BNK Page: 13
- 13. 035 K5.01 001
- Unit One is at 90% power when the S/G #4 safety fails open.
Which one of the following describes how the unit will initially respond?
A. Reactor power will remain the same because the S/G levels are on program.
,/B. The increased steam flow will insert positive reactivity.
C. Reactor power will increase due to a higher Tave.
D. Pressurizer level will increase due to a higher program level.
A. Incorrect, reactor power will change to due a lower Tave, positive reactivity.
B. Correct, increased steam flow will drop Tave which inserts positive reactivity.
this is a C. Incorrect, Tave will not increase; reactor power will not increase since negative reactivity effect.
D. Incorrect, pressure level will drop due to shrinkage when Tave drops and pressurizer level program inputs are auctioneered Tave vs. no-load Tave.
Since Tave drops then auctioneered Tave would initially drop.
K/A[CFR]: 035 K5.01 [3.4/3.9] [41.5]
Reference:
AOP-S.05 symptoms.
LP/Objective: OPL271MS B.9 History: New question.
Level: *emp f-
- Comments: FHW 12/02 035 K5.01
STEAM LINE OR FEEDWATER LINE BREAKILEAK RAOPeS.2 SQN 3/4.>
3.1 Symptoms C. The following automatic actions may occur due to a steam line or feedwater line break or leak:
"* Reactor trip on OTAT.
"* Automatic rod withdrawal due to dropping T-avg.
"* Automatic rod withdrawal stop from C-3 and C-4.
"* Turbine runback from OTAT or OPAT.
3.2 Entry Conditions None Page 11 of 12
Wednesday, October 09,2002 @ 11:45 AM HLC12-02.BNK Page: 16
- 16. 059 A4.01 001
" Unit One is operating at 75% power.
~
Instrument Mechanics are performing a test on the S/G level transmitters. They accidentlly make up a hi-hi level signal on channels II and Ill for S/G #1.
Which one of the following is correct for this condition?
A. Only 1-FCV-3-33 feedwater isolation valve to S/G #1 will close.
,B. "Main Feedwater Pump Turbine 1 B Abnormal" alarm will annunciate.
C. "Turbine Runback BOP" alarm will annunciate.
D. "PS-2-129 Low NPSH at MFP's" alarm will annunciate.
A. Incorrect, all 4 feedwater isolation valves will close, see 1,2-47W611-3-2.
B. Correct, 2/3 level transmitters made up for 1/4 S/Gs will create a P-14 and trip the MFP turbines (see 47W611-3-2). This annunciator will receive an input from the turbine tripped alarm SER.
C. Incorrect, this alarm will annunicate if MFP trip and load >80%,
see 1-AR-M2-A (B-i).
D. Incorrect, this alarm comes in if 100 psid decreasing between MFP intlet and #2 heater shell caused by increasing load or CBP trip. see 1-AR-M3-A (E-1).
KIA[CFR]: 059 A4.01 [3.1/3.1] [41.7]
Reference:
0-GO-5 R27 section 5.1 step [20]
1-AR-M3-B (B-i) 1-AR-M2-A (B-i) 1-AR-M3-A (E-1) 1,2-47W611-2-1 1,2-47W611-3-2 LP/Objective: OPL271COND.FW B.8 History: New question.
Level: Comprehension Comments: FHW 12/02 059 A4.01
SON O-GO-5 NORMAL POWER OPERATION Rev: 27 1 &2 Page 29 of 80 Unit STARTUP No. Date 5.1 Power Ascension From 30% to 100% (Continued)
[19] WHEN approximately 40% reactor power:
[a] VERIFY annunciator XA-55-4A, window E-7:
C-20 AMSAC ARMED is LIT.
El
[b] CLOSE the drains on the operating main feedwater pump turbine (N/A other pump).
MFPT DESCRIPTION HANDSWITCH POSITION INITIALS A DRAIN VALVES HS-46-14 CLOSED B DRAIN VALVES HS-46-41 CLOSED NOTE With verbal approval from the Operations Superintendent, placing the second main feed pump in service may be deferred until power is approximately 65%. Logic prevents opening the standby MFPT condenser isolation valves ifthe pump is not reset prior to exceeding 9 million lbs/hr flow on the running pump.
[20] WHEN approximately 40 to 45% turbine load, THEN PLACE the second MFPT in service by performing the following:
[a] IF the Operations Superintendent has approved one MFP operation during the power ascension, THEN
- 1. RECORD which MFPT is in service.
MFPT
- 2. MONITOR loading of the MFP in service as load is increased. El
SQN 0-GO-5 NORMAL POWER OPERATION Rev: 27 1 &2 Page 30 of 80 Unit STARTUP No. Date 5.1 Power Ascension From 30% to 100% (Continued)
[b] WHEN second MFPT is to be placed in service, THEN PLACE second MFPT in service in accordance with 1,2-SO-2/3-1. E]
[21] PERFORM the following as system parameters permit:
[a] VERIFY three (3) Hotwell pumps running. El
[b] VERIFY two (2) Condensate booster pumps running. 0
[c] VERIFY MFW pump(s) in service (only 1 required if approved by Operations Superintendent). El
[d] VERIFY two (2) #3 heater drain tank pumps El running.
[e] VERIFY (1) #7 Heater Drain Tank pump in one []
service.
[22] IF the second #7 heater drain tank pump has not been started, THEN START the second #7 heater drain tank pump in accordance with 1,2-SO-5-3.
8 (B-1)
Setpoint Source N/A MAIN FEtDWATER SER (Internal) PUMP TURBINE Windows A-2 thru A-7, B-2 thru B-7, 1B ABNORMAL C-2 thru C-7, and D-2 thru D-7 Probable 1. Thrust bearing wear detector.
Causes 2. Low bearing oil pressure (turbine and pump).
- 3. Main feedwater pump I B condenser isolation valve not fully open.
- 4. Low vacuum.
- 5. Low injection water pressure.
Corrective [I] REFER to source windows on this panel for corrective actions to Actions be taken.
12] IF source windows are not lit, THEN TEST annunciator panel to determine if bulbs good.
References 45N646-1, 45B655-031-0
8 (B-i)
Source Setpoint SER 83 1. #3 HDT bypass valve TURBINE
- 1. ZS-6-105A or 105B, LCV-8-IO5A or 105B open RUNBACK pS-47-13A, and turbine load > 85%, and and FS-6-107 FS-6-107 < 5500 gpm after BOP 10 sec. T.D.
tripped and PS-47-13B Probable 1. High level in #3 heater drain tank with turbine load > 85%.
Causes 2. Feedwater pump trip with turbine load > 80%.
Valve NOTE Runback is accomplished through the EHC System Position Limiter.
Corrective [1] PERFORM the following for a #3 HDT runback:
Actions [a] IF LCV-6-105A or 105B is open, THEN VERIFY turbine runback to 80%.
[b] IF any #3 HDT pump(s) trips, THEN GO TO AOP-S.04, Condensateor HeaterDrainsMalfunction WHILE continuing with this instruction.
flow obstructions NOTE Due to recent sparger redesigns and/or possible must be in the #3 HDTP bypass piping certain operator actions performed should a secondary side runback occur.
gpm,
[c] IF #3 HDT flow on R-6-10 is greater than 3000 THEN MONITOR FR-6-10 AND GO TO step Eel.
Drain Bypass to NOTE Position of 6-105A and 105B on XX-6-1, Heater determine if #3 Condenserlocated on panel M-2 may be used to step. Should 6-105A be HDT level has stablized in the following turbine load has FULL open and 6-105B be THROTTLED, then 6-105A and been reduced enough to allow stablization. Should
- 3 HDT 6-105B be FULL open (red light indication only), then require further turbine load level may NOT be stable and may reduction.
gpm, THEN
[d] IF #3 HDT flow on EES&1Q is less than 3000 a value REDUCE turbine load (use valve position limiter) to which allows #3 HDT level to stablize (load could be 60% or less)
AND GO TO AOP-S.04 if a Feedwater Heater String isolates.
CONTINUED
8 (B-I)
TURBINE RUNBACK CONTINUED BOP
[e] DISPATCH an operator to monitor #3 HDT level.
If] EVALUATE removal of one #3 HDT pump, and/or one Condensate Booster pump, and/or one or all Condensate Demineralizer Booster pumps to reduce condensate header pressure.
[2] PERFORM the following for a MFP tip:
[a] ENSURE operating MFP loads up and main rag valves respond to control SIG levels.
[b] ENSURE main feedwater controls are returning to a stable position for the present plant conditions.
[c] GO TO AOP-S.01, Loss ofNormalFeedwater.
[3] REFER TO O-GO-5, Norma/PowerOperations,to remove turbine control from governor valve limiter.
/I References 45N647-8, 45B655-02A-0, 45N657-26, 47B601-8-21, 471601-47-6
29 29 (E-d)
Setpoint Source SER 214 100 psid decreasing PS-2-129A (differential between MFP inlet and #2 heater shell)
Probable 1. Increasing load.
Causes 2. Condensate booster pump trip.
[1] ENSURE MFP inlet pressurePI-2-129]. greater than 320 psig by Corrective load EM-3, Actions reducing turbine
[2] ENSURE the following pumps are operating as required by O-GO-5 for the present unit load.
- a. Hotwell pumps
- b. Condensate DI booster pumps
- c. Condensate booster pumps
- d. #7 HDT pumps
- e. #3 HDT pump required to clear
[3] ADJUST condensate and feedwater pressure as alarm.
[4] IF MFP trips, THEN Feedwater.
GO TO AOP-S.01, Loss of Normal References 47B601-2-21 45B655-03A-2 47W611-2-2
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 18 18.060 AK2.02 001
"*'-~ Waste Gas Decay Tank "A" is in service and has developed a small external leak on
- t-5-t", k *,%*
'-,,,- A i-np-the gas analyzing sampling valve.
Which of the following are automatic actions tl-ý=( take place as result of this condition?
A-*wa-ste gas release,i-fw-pfogresstWill terminate.
B. Containment purge, 444-pf-es, will terminate.
vC. Auxiliary building ventilation will shut down.
D. Nitrogen to the waste gas system will isolate.
A. Incorrect, RE-1 18 is a closed pipe detector and not sensing this leak.
leak.
B. Incorrect, RM-130/131 has closed pipe sensing and not subject to this that will C. Correct, the spent fuel pool monitors or RM-90-101 will initate an ABI stop the normal ventilation.
D. Nitrogen has a manual isolation valve and no circuit with rad monitors.
and Student must understand the external leak will drift inside the auxiliary building initiate an ABI could make the spent fuel pit ARMs or RM-90-1 01 reach setpoint and that will stop the normal auxiliary building ventilation.
K/A[CFR]: 060 AK2.02 [2.7/3.1] [41.7]
Reference:
1,2-47W830-4 1,2-47W611-30-6 0-SO-90-2 0-SO-30-10 LP/Objective: OPL271RADMON B.9 History: New question.
Level: Comprehension Comments: FHW 12/02 060 AK2.02
AUXILIARY BUILDING 0-SO-30 10 SON Rev 19 1,2 VENTILATION SYSTEMS Page 10 of 158 3.0 PRECAUTIONS AND LIMITATIONS (Continued)
K. Ifan A train exhaust fan is going to be used in conjunction with a B train exhaust fan, the non-operating fan to be started needs to be manually set and left in manual for optimum operating performance.
L. Ensure that essential raw cooling water (ERCW) headers are charged up to the safety feature equipment coolers prior to making them operational.
M. The PASF will not be used as a normal sampling facility, but periodic sampling may be performed by the Radiochemical Laboratory. When this is done, radiation levels in the Auxiliary Building should be monitored.
N. After an ABI due to high radiation in the Auxiliary Building, restart of the Auxiliary Building supply and exhaust fans may cause a re-initiation of the ABI.
Inform the appropriate Operations personnel prior to restarting the Auxiliary Building Ventilation System.
- 0. An Auxiliary Building Ventilation Shutdown due to smoke detector cross zone is not an ABI or ESF. It only stops the Supply, Exhaust, and Fuel Handling Fans.
Auxiliary Building Ventilation fans will be placed and remain in the PULL-TO LOCK position unit [1-HS-30-102D1 has been RESET.
P. An Auxiliary Building isolation will be initiated from one of the following actuation signals:
- 1. Phase A containment isolation
- 2. Radiation detection in the refueling area
- 3. Auxiliary Building General exhaust vent high radiation
- 4. Auxiliary Building general supply fan suction high temperature
- 5. Manual
- 6. Manual with 1-HS-30-101A or 1-HS-30-101 B on 1-M-6.
- 7. Manual with 2-HS-30-101A or 2-HS-30-101 B on 2-M-6.
- 8. High radiation in the auxiliary building exhaust stack. (0-RM-90-101B, on 0-M-12)
- 9. Phase A containment isolation, either unit.
- 10. High radiation in refueling area, 0-RM-90-102 > 10mr/hr. (A Train)
- 11. High radiation in refueling area, O-RM-90-103 > 10mr/hr. (B Train)
- 12. High temperature in auxiliary building general supply fan suction, either unit, > 115'F: 1-TS-30-103 and 103A; 2-TS-30-104 and 104A.
/ j' GASEOUS PROCESS RADIATION 0-!L..40-2, Att. 6 MONITORING SYSTEM Date: 1 Aua 00 VALVE CHECKLIST 0-90-2.06 Page 2 of 2 ALIGNMENT OF (0-RE-90-118)
REQUIRED DESCRIPTION LOCATION VALVE NO. POSITION VERIFIED BY WGDT Discharge Isolation Upstream of 0-77-840D LOCKED 0-RE-90-118 OPEN 1st IV WGDT Discharge Isolation Downstream of 0-77-840C LOCKED - _
0-RE-90-118 OPEN 1st IV Vent Valve Downstream of O-VTIV-77-840G CLOSED 0-77-840D 1st IV Vent Valve Upstream of 0-VTIV-77-840H CLOSED 0-77-840C 1st IV Spare isolation Valve Upstream of 0-77-84OD/in 0-77-269 CLOSED valve gallery CAPPED 1st IV (1-47W600-106)
Spare isolation Valve Downstream of 0-77-840C/in 0-77-270 CLOSED valve gallery CAPPED 1st IV
___ ____ ___ ___ ___ (1 -47 W600- 106) _ _ _ _ _ _ _ _ _ _ _ _ _
Change 5
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 19 19.061 AK2.01 001 S> "O-RA-90-102A Fuel Pool Rad Monitor High Rad" alarm just annunciated.
Which one of the following describes the detector location and the automatic actions associated with this alarm?
A. East of the spent fuel pool on the penetration room wall elevation 749 and results in an ABI (fuel handling area isolation) train "A."
B. The transfer canal elevation 714 and results in ABI (fuel handling area isolation) trains "A" and "B."
C. West of the spent fuel pool on the penetration room wall elevation 749 and results in an ABI (fuel handling area isolation) train "B."
,/D. Near the spent fuel pool water elevation 734 and results in an ABI (fuel handling area isolation) train "A."
A. Incorrect per references.
B. Incorrect per references, this is 20 feet below floor elevation.
C. Incorrect per references.
D. Correct, per 0-AR-M12-B (B-3) and 0-RA-90-102A is associated with "A" train ABI per drawing. Where as 0-RA-90-103A is associated with "B"train.
K/A[CFR]: 061 AK2.01 [2.5/2.6] [41.7]
Reference:
0-AR-M12-B (B-3) 1,2-47W611-30-6 0-SO-30-10 R19 LP/Objective: OPL271C013 B.2 History: New question.
Level: Analysis Comments: FHW 12/02 061 AK2.01
10 (B-3)
Source Setpoint 0-RA 102A SER 760 50 mr/hr > 1 second FUEL FQOOL RAD 0-RE-90-102 MONITO R HIGH RAD Retransmitted to U-2 SER 2243 Probable 1. High radiation in spent fuel pit area elevation 734.
Causes Corrective [1] VERIFY the following:
Actions a. Auxiliary Building General Supply and Exhaust and Fuel Handling exhaust isolate (A-Train) (1-M-9).
- b. Auxiliary Building Gas Treatment System starts (1-M-9).
[2] CHECK O-RM-90-102 and O-RM-90-103 on 0-M-12 to verify alarm.
E3] IF high radiation alarm valid, THEN
[a] ANNOUNCE "High Radiation at spent Fuel Pool Area" over PA system.
[b] NOTIFY SM.
[c] NOTIFY RADCON.
[4] IF B-Train ABI has not actuated from a valid High Radiation condition, THEN INITIATE manually B-Train Auxiliary Building Ventilation Isolation via [I -HS-30-101 Bi or f2-HS-30-101 Bi (M-6).
[5] IF fuel handling in the Spent Fuel Pit is in progress, THEN REFER TO AOP-M.04, Refueling Malfunctions.
[6] REFER TO AOP-M.06, Loss of Spent Fuel Cooling.
[7] IF Auxiliary Building Ventilation Isolation resulted from an invalid ABI signal, THEN, REFER to 0-SO-30-10 Auxiliary Building Ventilation Systems to recover from ABI.
[8] EVALUATE Technical Specifications 3.3.3.1 and 3.9.12.
[9] INITIATE Corrective Actions.
[10] WHEN conditions return to normal, THEN RETURN Auxiliary Building Ventilation System to normal in accordance with 0-SO-30-10, Auxiliary Building Ventilation Systems.
References 45B655-12B-0, 47W610-90-1
Date.
8.5 Verification of Auxiliary Building Isolation Actuations (Continued)
N. Cask Loading Area supply isolation damper.
- 1. [0-FCO-30-1291 El
- 2. [0-FCO-30-1301 El
- 0. Cask Loading Area exhaust isolation damper.
- 1. [0-FCO-30-1221 El
- 2. [0-FCO-30-1231 El NOTE 1 The Auxiliary Building may not be accessible in the post-accident condition.
NOTE 2 In the following step, both dampers will be closed ifTrain A & B Auxiliary Building Isolation (ABI) signals were received.
[6] ENSURE one damper CLOSED for each ABI Train actuated, by local observation.
Auxiliary Building Exhaust Vent isolation damper.
(Verification by using lights on 1-HS-30-49 and 1-HS-30-55 located on South wall behind "B" Surge Tank, elevation 734)
- 1. [1-FCO-30-491 El
- 2. [1-FCO-30-551 El
[7] CHECK Auxiliary Building being maintained at 0.275-inch of water negative pressure: (Panel 1-M-9)
A. [0-PDI-30-1481 Aux Bldg Gas Trtmt Sys. El B. [0-PDI-30-1491 Aux Bldg Gas Trtmt Sys. El
[8] WHEN recovery from ABI is to be performed, THEN PERFORM Section 8.4, Recovery From Auxiliary Building
<v Isolation. El End of Section 8.5
Wednesday, October 09,2002 @ 11:45 AM HLC12-02.BNK Page: 20
- 20. 061 K2.01 001 Which one of the following is the correct power supply to the Auxiliary Feedwater 1-FCV-3-136B?
System ERCW Header isolation motor operated valve A. 480v Unit Bd 1A.
B. 480v Turbine MOV Bd 1 A.
D. 250v Battery Bd #1.
A. Incorrect per reference.
B. Incorrect per reference.
C. Correct per reference.
D. Incorrect per reference.
The student must understand this vital system is fed from a shutdown power source which is 480v Rx MOV Bd 1A2-A.
K/A[CFR]: 061 K2.01 [3.2/3.3] [41.7]
Reference:
1-SO-3-2, Att. 1, date 16 Nov 01, page 6 of 8.
1,2-45N779-10 LP/Objective: OPL271C035 B.4 History: New question.
Level: Memory Comments: FHW 12/02 061 K2.01
I
/
I K' Performance Date AUXILIARY FEEDWATER SYSTEM i-SO-3-2, Att. 1 Date: 16 Nov 01 / /
Page 6 of 8 POWER CHECKLIST 1-3-2.01 I I PRINT ~~TRANSFER IIIL EQUIPMENT I.D. PRINT BREAKER FUSES INSTALLED TINITIALS SWITCH REFERENCE POSITION ERCW Header 45N751-3 480V Rx MOV Four 1 amp 1-XS-3-136B Isol Valve 45N779-10 Bd 1A2-A/3B BUSS FRN 1 NORMAL 1-FCV-3-136B 1-BCTD-3-136B-A 1-FU4-3-136 B-A 1st IV CLOSED ERCW Header Isol 45N751-8 480V Rx MOV Four 1 amp 1-XS-3-179A Valve 1-FCV-3-179A 45N779-10 Bd 1B2-B/11E BUSS FRN1 NORMAL 1-BCTD-3-179A-B 1-FU4-3-179A-B _
1st IV CLOSED ERCW Header Isol 45N751-8 480V Rx MOV Four 1 amp 1-XS-3-179B Valve 1-FCV-3-179B 45N779-10 Bd 1B2-B/11B BUSS FRN1 NORMAL 1-BCTD-3-179B-B 1-FU4-3-179B-B 1st IV CLOSED TDAFW Pump 45N751-4 480V Rx MOV Four 1 amp 1-XS-1-15 Isol Valve 45N779-28 Bd 1A2-A/19A BUSS FRN 1 NORMAL 1-FCV-1-15 1-BCTD-1-15-A 1-FU4-1-15A _
1st IV CLOSED TDAFW Pump 45N751-4 480V Rx MOV Four 1 amp 1-XS-1-16 Isol Valve 45N779-28 Bd 1A2-A/17B BUSS FRN 1 NORMAL 1-FCV-1-16 1-BCTD-1-16-A 1-FU4-1-16-A 1st IV CLOSED TDAFW Pump dc Manual 45W646-6 125V dc Vital Three 6 amp Transfer Switch, T & T Valve Batt Bd III BUSS KWN6 1-FCV-1-51, and dc Vent 1-BKRC-3-KC/321-A 1-FU3-1-52-S (5) N/A Fan Normal Power Supply CLOSED 1st IV (5) Fuses located in JB 3043, elev 669 Chanae 8
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 21
- 21. 061 K6.01 001
"- Unit One is performing a surveillance test on the Terry Turbine when the following is experienced.
"Aux FWP Turbine 1A-S Mechanical Overspeed Trip" alarm is lit because 1-FIC-46-57 failed tn control numn outlet flow. -,rM& 4 c( ,,'%,,, -oN-Which one of the following is the correct operator actiokto place the pump back in service with the failed controller?
A. Place 1-HC-46-57-S in manual and go to open with 1-HS-1-51A-S.
B. Ensure the mechanical overspeed mechanism is latched and place 1-HC-46-57-S in manual and go to open with 1-HS-1-51A-S.
C. After the overspeed alarm is cleared place 1-HC-46-57-S in manual and go to open with 1-HS-1-51A-S.
,/D. After the overspeed alarm is cleared place 1-HC-46-57-S in manual and set the controller output to 20% and go to open with 1-HS-1-51A-S.
A. Incorrect, per reference.
B. Incorrect, per reference.
C. Incorrect, since the controller failed to control before then adjust the controller output to 20%.
D. Correct, the overspeed condition has been corrected and the controller output has been adjusted to 20% to prevent another overspeed.
K/A[CFR]: 061 K6.01 [2.5/2.8] [41.7]
Reference:
1-AR-M3-C (A-4) 1-SO-3-2 R27 section 8.5 and section 5.3.
LP/Objective: OPL271C035 B.6 History: New question.
Level: Memory Comments: FHW 12/02 061 K6.01
Date 5.3 Startup and Control of TDAFW Pump from the UCR NOTE 1 Possible cause for TDAFW Pump Lot manually starting is tripping of thermal overloads for the trip and throttle valve at ITE box next to the local controls in the Terry Turbine Room.
NOTE 2 AFW should be maintained at a constant flow rate to reduce the potential for S/G nozzle cracking. The goal should be to establish the lowest possible constant flow rate.
[1] ENSURE AFW system is in Standby per Section 5.1 Startup/Standby Readiness of this Instruction.
[2] NOTIFY the appropriate operator to locally inspect the TDAFW pump to ensure it is ready for operation.
ENSURE the following level control valves are
[31 CLOSED:
INITIALS VALVE NUMBER DESCRIPTION 1-LCV-3-172 S/G 3 LCV 1-LCV-3-173 S/G 2 LCV 1-LCV-3-174 S/G 1 LCV 1-LCV-3-175 S/G 4 LCV
Date.
5.3 Startup and Control of TDAFW Pump from the UCR (Continued)
[4] IF desiring to control Terry Turbine speed manually, THEN PLACE [1-HC-46-57-S1 in MANUAL and set controller as desired to control pump speed.
[5] START the Terry Turbine by PLACING
[1-HS-1-51A-S in the OPEN position.
[6] VERIFY pump comes up to operating speed of - 3970 (or desired speed) rpm as seen on [1-SI-46-56A-S1.
[7] CONTROL S/G level as necessary by PLACING the applicable level control valve handswitch in the OPEN position until desired S/G level is achieved, THEN PLACE the applicable level control valve handswitch in the CLOSED position.
NOTE 1 To increase the pressure of [1-PCV-3-1831, turn the valve hand wheel in a clockwise direction. To decrease pressure Turn valve hand wheel in a counter-clockwise direction.
NOTE 2 Due to the varied inlet steam pressures encountered with TDAFW pump operation, [1-PCV-3-1831 setpoint must be adjusted to prevent steam seal leakage. A setting of - 120 psig corresponds to a 1000 psig inlet steam pressure.
[8] IF TDAFWP inlet steam supply pressure is - 1000 psig, THEN VERIFY [1-PCV-3-1831, Ejector Pressure Control Valve, is controlling pressure at approximately 120 psig as indicated on [1-PI-3-1841.
[9] PERFORM visual inspection to determine if steam is leaking down the turbine shaft through the gland steam seals. El
Date .
5.3 Startup and Control of TDAFW Pump from the UCR (Continued)
[10] IF there is steam leaking through the steam seals, THEN adjust [1-PCV-3-183, Ejector Pressure Control Valve to prevent leakage by performing the following:
[a] REQUEST MIG to install a temporary pressure gauge (- 1000 psig range) at PI-3-184 location.
El
[b] ADJUST [1-PCV-3-1 831 to obtain a pressure of
- 500 psig. (This will clear the PCV of any debris). El
[c] IF additional pressure control is required, THEN THROTTLE [1-VLV-3-920] WHILE adjusting
[1-PCV-3-1831. El
[d] IF TDAFWP inlet steam supply pressure is
- 1000 psig, THEN ADJUST [1-PCV-3-1831 to obtain a pressure of
- 120 psig. El
[e] IF TDAFWP inlet steam supply pressure is NOT
- 1000 psig, THEN ADJUST [1-PCV-3-1831 to prevent leakage of steam seals down the turbine shaft. El
[f] REQUEST MIG to remove the temporary pressure gauge and to reinstall [P-3-18
[g] ENSURE [1-VLV-3-920l is OPEN. /
1st IV END OF TEXT
SQN AUXILIARY FEEDWATER SYSTEM Rev: 27 1 Page 36 of 57 Date 8.5 Operation of TDAFW Pump 1-FCV-1-51 (Trip and Throttle) Valve if it Closes Unintentionally or Will Not Open
[1] IF TDAFW pump Trip and Throttle valve will not electrically open or closes unintentionally, THEN ENSURE the following:
[a] ENSURE Trip and Throttle Valve is run down to full close position.
[b] ENSURE mechanical overspeed mechanism is LATCHED (refer to local placard).
[c] ENSURE Terry Turbine overspeed alarm in UCR is clear.
[2] IF an overspeed trip has occurred due to 1-FIC-46-57 failing to control pump outlet flow automatically, THEN PLACE [1-HC-46-57-S1 in MANUAL and set controller output to 20 percent as seen on
[1-FI-46-57-S1 on M-3 prior to attempting restart of TDAFW pump.
[3] IF needing to continue Terry Turbine operation, THEN GO TO applicable startup section.
END OF TEXT
4 (A-4)
Source Setpoint AUX FWP SER 264 TURBINE 1A.-S SS-46-53 4900 rpm (125%) MECHANICA RL OVERSPEED 1iRIP Probable 1. Governor malfunction Causes a. Leaky governor or leak near governor.
- b. Governor responds slowly due to worn parts or sticking.
- c. Linkage out of adjustment.
- d. Low governor valve hydraulic fluid.
Corrective E1] IF auxiliary feedwater pump turbine was in service, THEN Actions ENSURE feedwater pump turbine tripped.
[2] IF motor driven auxiliary feedwater pumps are not in service and are available, THEN START additional auxiliary feedwater pumps as needed.
NOTE To restart the auxiliary feedwater pump turbine from a mechanical overspeed the mechanical trip linkage will have to be reset locally.
[3] REFER TO 1-SO-3-2, Auxiliary Feedwater System, for further guidance on reset and restart.
References Terry Steam Turbine Instruction Manual VTD DR04 0010 (Cont. # VTM-1075-0250) 45B655-03C-0, 45N646-6, 47W611-3-4
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 22
- 22. 063 A2.01 001 S- "125V DC Vital Bat Bd 1 Abnormal" alarm is lit.
Which one of the following is the correct cause and action for this condition?
A. An undervoltage condition exists and dispatch operations personnel to adjust the char er output voltage.
'B. Clearing the alarm requires adjustment of the red flag (indicator) ground setpoint by V
dispatching operations personnel to the local board.
condition exists and dispatch operations personnel to adjust the C. An overvoltage voltage.
cha ger output
't. Clearing the alarm will require a local push button reset and Svadditional IM support in the control room at the annunciator panel.
A. Incorrect, an undervoltage does not have input to this alarm. Also the charger output voltage is adjusting by maintenance section.
B. Correct per reference.
C. Incorrect, a overvoltage does not have input to this alarm. Also the charger output voltage is adjusting by maintenance section.
D. Incorrect, the reset is not a push button and IM support is not required.
KIA[CFR]: 063 A2.01 [2.5/3.2] [41.5 43.5]
Reference:
1-AR-Mi-C (A-5)
LP/Objective: OPN218E.007 B.17 History: New question.
Level: .jReMery ooa*f Comments: FHW 12/02 063 A2.01
5 (A-5)
Setpoint Source 125V DC VITAL SER 40 N/A BAT BD I
- 1. Fuses and breaker provided with alarm ABNORMAL contacts which close when fuse opens or BKR trips. Contacts wired to common column alarm bus for annunciation.
+/-80V 2.- Ground indicator alarm Probable 1. Any battery board breaker tripped.
Causes 2. Any fuse blown on fuse column A, B, C or D.
- 3. Positive or negative ground on DC system-above setpoint
(+/- 80V).
CHECK Corrective [1] DISPATCH Personnel to 125V DC Vital Battery Bd I AND following :
Actions on 125V DC ground
[a) Ground present greater than setpoint detector.
[b] Any breaker(s) TRIP on battery board.
D (located in back
[c] Fuse blown on fuse column A, B, C, or of panel D).
Battery Bd, THEN
[2) IF ground present on 125V DC Vital INITIATE maintenance.
alarm.
[3] ADJUST red flag setpoint to clear reclosing any tripped breaker.
NOTE SRO approval must be provided prior to Bd TRIP, THEN
[4] IF any feeder breaker on 125V Dc Vital-Battery
[a] NOTIFY Unit SRO of breaker trip.
breaker.
[b] EVALUATE effects of reclosing tripped THEN
[c] IF permission obtained from SRO, ATTEMPT to CLOSE tripped Breaker.
again THEN
[5] IF breaker will not CLOSE or TRIPS INITIATE maintenance.
CONTINUED
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 23
- 23. 064 K2.02 001
"~ The Diesel Generator 1 A-A Day Tank Fuel Oil Transfer Pump #1 was running in automatic due to a low level in a 1 A-A Day Tank caused when maintenance pumped out the day tank. The D/G 1 A-A is NOT running at this time.
6.9 kv Unit Bd 1B tripped on a ground fault while the D/G 1 A-A Day Tank Fuel Oil Pump #1 was running.
Which one of the following is correct*fthis condition?
A. The Day Tank Fuel Oil Pump will continue to run since it is DC powered from the D/G battery.
",/B. The Day Tank Fuel Oil Pump wjjl stop and after the 1 A-A D/G ties to the 1 A-A 6.9 kv Shutdown Bd9,,'vay Tank Fuel Oil Pump will start.
C. The Day Tank Fuel Oil Pump will stop and require manual restart after 6.9 kv Shutdown Bd 1 A-A is energized.
D. The Day Tank Fuel Oil Pump will continue to run since it is powered from "B" train shutdown power.
A. Incorrect, the pump's power supply is Diesel Aux Bd 1A1-A, not DC powered.
B. Correct, the pump's power supply is Diesel Aux Bd 1A1-A which will be energized when the D/G 1 A-A ties on to 6.9 kv Shutdown Bd. 1 A-A.
C. Incorrect, the pump was running in auto when the 6.9 unit bd tripped therefore the switch position remains the same and the transfer pump will start again since the day tank level was not increased to normal level.
D. Incorrect, the pump's power supply is Diesel Aux Bd 1A1-A which is fed from the 480v Shutdown Bd 1A1-A.
K/A[CFR]: 064 K2.02 [2.8/3.1] [41.7]
Reference:
45N771-4 15N500 LP/Objective: OPL271 D/G B.10 History: New question.
Level: Comprehension Comments: FHW 12/02 064 K2.02
Wednesday, October 09,2002 @ 1,0145 AM HLC12-02.BNK Page: 25
- 25. 076 AK3.05 001k Unit Two is currently at 100% power. Chem Lab reports the RCS activity for Iodine 131 is 480 pLci/gm.
Information: 2-RM-90-277 (RCDT).
2-RM 278 (RCDT).
2-RM-90-400A (Shield Bldg Vent).
2-RM 400B (Shield Bldg Vent).
Which one of the following symptoms and procedural actions are correct for Iodine removal?
,/A. Monitor 2-RM-90-277, 2-RM 278, and place mixed beds and cation bed demineralizers in service.
B. Monitor 2-RM-90-400A, 2-RM-90-400B, and divert letdown to the hold up tank.
C. Monitor 2-RM-90-277, 2-RM 278, and isolate normal letdown and place excess letdown in service.
D. Monitor 2-RM-90-400A, 2-RM 400B, and place mixed beds and cation bed demineralizers in service.
A. Correct, RM-90-277/278 monitors for high RCS activity; demin beds reduce RCS activity per references.
B. Incorrect, RM-90-400 monitors only during a gas release; hold up tank does not remove RCS activity.
C. Incorrect, placing excess letdown in service will by pass the demineralizers.
D. Incorrect, RM-90-400 monitors only during a gas release.
K/A[CFR]: 076AK3.05 [2.9/3.6] [41.5 41.10]
Reference:
1-47W809-1 1-AR-M30-A (A-3) 1-AR-M30-A (B-3)
AOP-R.06 LP/Objective: OPL271C370 B.3 History: New question.
Level: Comprehension Comments: FHW 12/02 076 AK3.05 Written by Scott Poteet.
3 (A-3)
Source Setpoint 1-RA-278A SER 1731 1-RE-90-278 316mR/Hr RCDT HI RAD Probable 1. Fuel cladding failure.
Causes 2. High activity in reactor coolant.
Corrective [1] VERIFY r1-FCV-77-10r RCDT pump to TDCT isolation valve Actions closes.
[2] CHECK r1-RM-90-2781 for high radiation on panel 1-M-30.
[3] IF need to pump RCDT to FDCT with this high radiation condition present, THEN REQUEST permission from SRO to place the following hand switches in block:
- a. [1 HS-77-9B1
- b. [1-HS-77-10B1
[4] WHEN ready to pump RCDT, THEN OPEN r1-FCV-77-10o and [1-FCV-77-91.
[5] REQUEST radiochemical laboratory sample to verify activity.
[6] IF high activity verified, THEN GO TO AOP-R.06, High RCS Activity.
References 471601-55-75, 47W610-90-4
11 (B-3)
Source Setpoint 1-RA-277A SER 1739 RCDT 1-RE-90-277 316 mR/Hr HI RAD Probable 1. Fuel cladding failure.
Causes 2. High activity in reactor coolant.
Corrective [1] VERIFY [1-FCV-77-91 RCDT pump to TDCT isolation valve Actions closes.
[2] CHECK [1-RM-90-2771 for high radiation on panel 1-M-30.
[3] IF need to pump RCDT to FDCT with this high radiation condition present, THEN REQUEST permission from SRO to place the following hand switches in block:
- a. rl-HS-77-9B1
- b. rl-HS-77-10B1
[4] WHEN ready to pump RCDT, THEN OPEN [1-FCV-77-1i] and .rl-FCV-77-9i.
[5] REQUEST radiochemical laboratory sample to verify activity.
[6] IF high activity verified, THEN GO TO AOP-R.06, High RCS Activity.
References 47B601-55-75, 47W610-90-4
K -/
STEP I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.0 OPERATOR ACTIONS (cont'd)
- 5. MONITOR Auxiliary Building area NOTIFY RADCON of abnormal radiation radiation monitors NORMAL. levels.
- 1 (2)-RR-90-1 A and 1B USE PA system to evacuate affected areas.
ICS Area Rad Monitor Displays
- 6. PERFORM the following based on Chem Lab recommendations:
- a. PLACE mixed beds and cation bed demineralizer in service USING 1,2-SO-62-9, CVCS Purification System.
- b. ADJUST letdown flow.
- 7. GO TO appropriate plant procedure.
END OF SECTION Page 4 of 7
HLC12-02.BNK Page: 29 Wednesday, October 09, 2002 @ 11:45 AM
-/
- 29. 103 A3.01 001 B containment Which one of the following is correct concerning an automatic phase isolation?
,/A. Panel 6E will be lit and both te n spu"F*m B. Panel 6E will be lit outside the outlined are AI' are th c"ntainm Ltspry p p C. Panel 6E will be lit inside the outlined
-NO-T-beuny.-
will he running D. Panel 6E will be dark andsUoth-containment spray pumrps A. Correct per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A[CFR]: 103 A3.01 [3.9/4.2] [41.7]
Reference:
E-0 R23 step 9 LP/Objective: OPL271C379 B.3 History: New question.
Level: Memory Comments: FHW 12/02 103 A3.01
SON REACTOR TRIP OR SAFETY INJECTION E-0 Rev. 23
"-S-TEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 9. MONITOR containment spray PERFORM the following:
NOT required: a. ENSURE containment spray
"* Phase B NOT ACTUATED INITIATED:
AND 1) Containment spray pumps RUNNING.
less than 2.81pressure
"* Containment psid.2)Cnametsryhdr 2) Containment spray header isolation valves FCV-72-39 and FCV-72-2 OPEN.
- 3) Containment spray recirculation valves to RWST FCV-72-34 and FCV-72-13 CLOSED.
- 4) Containment spray header flow greater than 4750 gpm per train.
- 5) Panel 6E LIT.
- b. ENSURE Phase B valves CLOSED:
"* Panel 6K PHASE B GREEN.
"* Panel 6L PHASE B GREEN.
- c. STOP RCPs.
- d. MONITOR containment air return fans:
- 1) RECORD present time.
- 2) WHEN 10 minutes have elapsed, THEN ENSURE containment air return fans are running.
10 of 25
Wednesday, October 09, 2002 @ 11:45 AM 11LC12-02.BNK Page: 31 31.2.1.24 001
<~ This question has reference material attached.
idle Using..0drawirIJthe attached drawing which one of the following will cause an Hotwell Pump Motor to start?
A. HS-2-33A placed in start, HS-2-33B in RESET andpofrMFP 'A" "B" condenser isolation valves closed.
in the full B. HS-2-33A in P-Auto and both MFP "A"and "B" condenser isolation valves open position.
'C.loth MFP "A"4aid "B" condenser isolation valves in the full open position, all electrical faults reset, and HS-2-33B placed in the TEST position.
D. Reset of electrical fault relays while both MFP "A" and "B" condenser isolation mid position.
valves are open, HS-2-33B in reset and HS-2-33A in spring return to A. Incorrect per reference. No flow path if the valves are closed, interlock.
B. Incorrect per reference. No P-Auto position.
CP. Correct per reference.
D. Incorrect per reference switch was never in start postion.
K/A[CFR]: 2.1.24 [2.8/3.1] [41.10 43.5]
References:
1,2-47W611-2-1 LP/Objective: OPN218SS.002 B.6 History: New Question Level: Analysis Comments: FHW 12/02 2.1.24 Written by Jim Kearney (10/3/02).
Must provide a copy of drawing 1,2-47W611-2-1
Wednesday, October 09, 2902 @ 11:45 AM HLC12-02.BNK Page: 32 32.2.2.4 001 /
- This question has reference material attached.
Given the following plant conditions:
- Unit One and Unit Two are both in Mode 5.
- Unit One RCS level is being maintained at elevation 708 ft.
- Unit Two RCS level is being maintained at elevation 706 ft.
Which one (1) of the following describes the corresponding RVLIS upper range indications for both units?
Unit One Unit Two
,/A. 99% 94.5%
B. 94.5% 94.5%
C. 94.5% 99%
D. 99% 99%
A. Correct per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A[CFR]: 2.2.4 [2.8/3.0] [45.1/13]
Reference:
0-GO-13 R23 Appendix J.
LP/Objective: OPL271RVLIS B.6 History: New question.
Level: Analysis Comments: FHW 12/02 2.2.4 Written by Scott Poteet (9/11/02)
Provide a copy of O-GO-13 R23 Appendix J 1/2>
APPENDIX Page 1 of J2 RVLIS UPPER RANGE VS RCS ELEVATION and reactor head vented.
D 80 C>75 65 710 695 "REACTOR COOLANT SYSTEM ELEVATION (ft) 2'
APPENDIX Page 2 of J2 RVLIS UPPER RANGE VS RCS ELEVATION
105 100
-.95 W
0D90 z
CE85 10~
U) ar75 65 699 700 701 702 703 704 705 706 707 708 709 710 695 696 697 698 2"
REACTOR COOLANT SYSTEM ELEVATION (ft)
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 34
- 34. AIR-B.12 002
">*- Given the following:
- Unit 1 and 2 are at 100% power
- A leak develops on the Control Air System
- Air pressure is at 73 psig Which ONE/of the following describes the system response?
vA. A and B Auxiliary Air Compressors automatically started and loaded.
B. A and B Train Auxiliary Air automatically isolated from Control Air.
C. A and B Train Auxiliary Air automatically isolated to Containment.
D. A and B Control and Service Air Compressors started and are operating half loaded.
A. Correct, Aux Air Cmpr will start <77 psig and load <83.5 psig.
B. Incorrect this setpoint is <69 psig.
C. Incorrect this setpoint is <50 psig.
D. Incorrect C&S air cmprs will operate at full load at <94 psig.
K/A: 078 K4.01 [2.7/2.9]
078 K1.01 [2.8/2.7] [41.2-9]
Reference:
AOP-M.02, Control Air system description.
Objective: OPL271CSA, B.12 History: System bank Level: Comprehension Note: FHW 12/02 078 K1.01 ,1 ,
100 Q2
3.1 Symptoms (cont'd) may indicate a A. Deviations or unexpected indications on any of the following loss of Control Air System pressure:
"* Aux Control Air Header pressure dropping:
"* O-PI-32-104A, Aux Control Air Hdr A Press [M-15].
"* 0-PI-32-105A, Aux Control Air Hdr B Press [M-15].
"* Control and Service (C&S) air compressors tripped.
"* Valves or dampers moving to failed position(s).
"* S/G anomalies due to MFW Reg Valves failing closed.
"* MSIVs fail closed.
Air System B. Any of the following automatic actions may indicate a loss of Control pressure:
RANGE (psig) INITIATING EVENT SETPOINT (psig) min max EVENT Normal control C&S air compressors auto start with N/A -
backup control circuit circuit failure Service Air isolates from Control Air 88 86 90 Control Air Receiver (0-PCV-33-4) dropping C&S air compressors load to 50%
86 84 88 Control Air Receiver C&S air compressors load to 100%
pressure dropping 77 74.5 79.5 Aux Air Receiver Aux Air Compressors start pressure dropping 69 66.5 71.5 Control Air header Aux Air isolates from Control Air pressure dropping (0-FCV-32-82 and 85)
Aux Air Receiver Aux Air to containment valves fail 50 .. .. pressure dropping closed. (1-FCV-32-80, -102,
-110, 2-FCV-32-81, -103, -111) 3.2 Entry Conditions None Page 43 of 58
tdnesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 35
'35. AIR-B.5 014 Given the following plant conditions:
- Control Air receiver #1 pressure is 96 psi and steady
- Control Air receiver #2 pressure is 99 psi and steady
- Service air receiver pressure is 78 psi and decreasing Which ONE the following d' th auseftdecreasing service air receiver pressure? X A. Loss of compressors A & B sequencer power.
B. Stuck open blowdown valve on air dryer A tower # 1.
C. Pressure control valve 0-PCV-33-4 failed open.
",/D. Loss of power to pressure control valve 0-PCV-33-4.
A. Control air receivers 1 and 2 pressure is normal thus ruling out an air compressor problem. the difference between control air receiver pressure is calibration tolerances.
B. The air dyers are on the control air header and the capacity of the blowdown valve is within the capacity of the compressors. A stuck open blowdown valve would have to decrease control air receiver pressures down to < 78 psig for this to affect the service air system via PCV-33-4.
C. Pressure control valve 0-PCV-33-4 fails closed on loss of air or loss of power. If the valve could fail open it will not cause the service air reciever to depressurize.
D. O-PCV-33-4 is the crosstie valve that supplies service air from the control air receivers 1 and 2. This valve fails closed on loss of power and closure of this valve will cause service air receiver pressure to decrease while control air pressure remains normal.
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 36
- 35. AIR-B.5 014 S KIA(CFR}: 079 K1.01 [3.0/3.11 {41.2, 41.3, 41.4, 41.5, 41.6, 41.7, 41.8, 41.9}
079 K2.01 [2.3/2.3] {41.2, 41.3, 41.4, 41.5, 41.6, 41.7, 41.8, 41.9}
079 K4.01 [2.9/3.2] (41.2, 41.3, 41.4, 41.5, 41.6, 41.7, 41.8, 41.9}
079 A2.01 [2.9/3.2] 141.2, 41.3, 41.4, 41.5, 41.6, 41.7, 41.8, 41.9}
078 K1.02 [2.7/2.8] {41.2, 41.3, 41.4, 41.5, 41.6, 41.7, 41.8, 41.9)
References:
AOP-M.02 1,2-47W611-32-1 45N632 45N779-6 45N779-18 LP/Objectives: OPL271C038 Obj. B.2, B.5 History: System bank Level: Comprehension Comments: LP-5/2000. C&S Air 002.; FHW 12/02 079 K4.01
3.1 Symptoms (cont'd)
A. Deviations or unexpected indications on any of the following may indicate a loss of Control Air System pressure:
"* Aux Control Air Header pressure dropping:
"* 0-PI-32-104A, Aux Control Air Hdr A Press [M-15].
"* 0-PI-32-105A, Aux Control Air Hdr B Press [M-15].
"* Control and Service (C&S) air compressors tripped.
"* Valves or dampers moving to failed position(s).
"* S/G anomalies due to MFW Reg Valves failing closed.
"* MSIVs fail closed.
Control Air System B. Any of the following automatic actions may indicate a loss of pressure:
RANGE si INITIATING EVENT SETPOINT si min max EVENT N/A .. .. Normal control C&S air compressors auto start with circuit failure backup control circuit Service Air isolates from Control Air 88 86 90 Control Air Receiver (0-PCV-33-4) dropping C&S air compressors load to 50%
86 84 88 Control Air Receiver C&S air compressors load to 100%
pressure dropping 77 74.5 79.5 Aux Air Receiver Aux Air Compressors start pressure dropping 66.5 71.5 Control Air header Aux Air isolates from Control Air 69 0-FCV-32-82 and 85) pressure dropping Aux Air Receiver Aux Air to containment valves fail 50 .. .. pressure dropping closed. (1-FCV-32-80, -102,
-110, 2-FCV-32-81, -103, -111) 3.2 Entry Conditions None Page 43 of 58
HLC12-02.BNK Page: 37 Wednesday, October 09, 2002 @ 11:45 AM
- 36. AOP-C.01-B.1 002
" > Given the following plant conditions:
- The unit is initially operating at 75% power.
- The following symptoms occur:
- Tavg greater than Tref
- Pressurizer spray valves partially OPEN
- Pressurizer level INCREASING Which ONE of the following would cause the above symptoms to occur?
A. Turbine Runback
,/B. Uncontrolled rod withdrawal C. S/G Safety Valve failed OPEN D. Power Range Channel N-43 failed high decreasing if a A. Incorrect, pressurizer level would be above program but runback occured.
to increasing B. Correct, pressure increase and water expansion due RCS temperature from 75% power.
C. Incorrect, S/G safety failing open would decrease Tave.
D. Incorrect, no inputs to pressure control.
KIA: 001 AK1.03 [3.9/4.0] {41.8 41.10)
Reference:
AOP-C.01, Symptoms and Entry Conditions LP/Objective:OPL271AOPCO1, B.1 Level: Comprehension History: Procedure Bank Comments: FHW 12/02 001 AK1.03
3.0 SYMPTOMS AND ENTRY CONDITIONS 3.1 Symptoms A. Any of the following annunciators may indicate a rod control system malfunction:
PANEL XA-55-4B, NISIROD CONTROL A-7 ZB-412A ROD CONTROL BANKS LIMIT LOW A-6 ROD CONTROL SYSTEM URGENT FAILURE B-3 NIS POWER RANGE UPPER DETECTOR HI FLUX DEVN OR AUTO DEFEAT B-7 ZB-412B ROD CONTROL BANKS LIMIT LOW-LOW C-3 NIS POWER RANGE LOWER DETECTOR HI FLUX DEVN OR AUTO DEFEAT C-7 ZB-412C BANK D ROD WITHDRAWAL LIMIT HIGH D-4 COMPUTER ALARM ROD DEV AND SEQ NIS PWR RANGE TILTS D-7 FULL LENGTH RODS AT BOTTOM E-3 NC-46B NIS POWER RANGE CHANNEL DEVIATION I PANEL XA-55-5A, REACTOR COOLANT - STM - FW C-6 TS-68-2P/Q REAC COOL LOOPS T REF T AUCT HIGH-LOW I PANEL XA-55-6A, REACTOR PROTECTION AND SAFEGUARDS B-1 I NC-41U/NC-41K NIS POWER RANGE HIGH NEUTRON FLUX RATE B. Any of the following parameters may indicate rod control system failure:
- 1. Unwarranted rod motion as indicated by RPI or step counters.
- 2. Steam header pressure deviations without corresponding turbine load change.
- 3. T-avg. - T-Ref deviations.
3.2 Entry Conditions None S.
S.-
Page 34 of 39
Wednesday, October 09, 2002 @ 11:45 A HLC12-02.BNK Page: 45
- 44. AOP-R.01-B.2 003
< Given the following plant conditions:
The operating crew has identified a S/G tube leak.
AOP-R.01 has been implemented.
Letdown flow = 75 gpm.
1 CCP is in service with charging valves FCV-62-93 and 89 full open.
Charging flow = 160 gpm.
Pressurizer level is stable at 58%.
All other parameters are normal Which ONE of the following best estimates the total primary to secondary leak rate?
A. 55 gpm.
,/B. 75 gpm.
C. 85 gpm.
D. 150 gpm.
- a. Incorrect - Nearest if Total letdown plus seal return are used. (160 20=53)
- b. Correct Nearest if correct numbers are used (160 - 87 = 73.)
C. Incorrect - Nearest if only letdown number is included. (160 - 75 = 85) ue. (16 - Nearest if letdown flow is neglected and seal return flow is vrI Inn~nrre used. (160. - 12 = 148)
K/A {CFR}: APE 037 2.1.7 [3.7/4.4] [43.5]
References:
OPL271 CVCS LP/Objectives: OPL271 C366 B.2 OPL271 C022 B.3 History: Procedure Bank Level: Analysis Comments: FHW 12/02 037 2.1.7 Provide a calculator
CVCS Student Handout Revision 0 Page 9 of 38 Reactor coolant purification is used to remove the following from the RCS:
- 1. fission and activation products (in ionic or particulate form);
- 2. corrosion agents (02, chlorides, and fluorides);
- 3. excess lithium for pH control.
The bases for the purification portion of CVCS design are:
I. protection of RCS integrity and components from chemical attack;
- 2. permits access to lines carrying reactor coolant (for chemical monitoring);
- 3. activity release reduction due to leaks.
Various chemicals are added for corrosion control. Chemical addition purposes are:
- 1. Control pH during startup (Lithium);
- 2. Scavenge oxygen during startup (Hydrazine);
The system is capable of maintaining all reactor coolant chemistry limits.
Seal water injection supplies filtered seal water to reactor coolant pumps as specified by their design. (See the Rector Coolant Pump lesson material for specific operational information.)
Hydrostatic testing - capable of supplying water at maximum test pressure to verify RCS integrity.
Emergency Core Cooling System interrelation The centrifugal charging pumps and associated valves serve as part of emergency core cooling. The remainder of system is isolated during accident. Depending on the function being lined up for operation or the function being isolated, the SI signal or the OA isolation signal may align the CVCS system components.
System and Component description Letdown Charging Seal Water, Chemical Control Purification and Make-up General Description This portion of the CVCS maintains programmed pressurizer level by balancing letdown flow with charging and seal water flow, provides filtered seal water to reactor coolant pump seals and controls reactor coolant system pressure during solid plant pressure control operations. The capability of maintaining system chemistry limits and make-up to the RCS is also provided by the following components and flowpaths.
Flow balance (normal operation):
- 1. Letdown flow = 75 gpm
- 2. Seal return flow = 12 gpm
- 3. Charging flow = 55 gpm
- 4. Seal injection flow = 32 gpm Therefore, 75 gpm letdown flow + 12 gpm seal return flow = 87 gpm, 55 gpm charging flow + 32 gpm seal injection flow = 87 gpm.
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 46
- 45. AOP-R.01-6.2 004
"-> Given the following plant conditions:
- The plant is operating at 90% power.
- A tube leak has developed on SG #1.
- PRESSURIZER LEVEL HIGH-LOW annunciator has just alarmed.
- Pressurizer level is below program, decreasing slowly.
in accordance Which ONE 9*of the following actions is required for these conditions with AOP-R.01, "Steam Generator Tube Leak"?
A. Manually trip the reactor and actuate SI.
B. Isolate letdown and evaluate pressurizer level.
",/C.Maximize charging for pressurizer level evaluation.
swapover level.
D. Maximize VCT blended makeup to prevent approach to RWST A. Incorrect, not procedurally required with initial conditions.
B. Incorrect, not procedurally required with initial conditions.
C. Correct per procedure.
D. Incorrect, not procedurally required with initial conditions.
K/A {CFR}: 037 2.4.4911 [4.0/4.0] {41.7) 037 AA1. [3.4/3.3]
2.4.4 [4.0/4.3] 41.10, 43.2 2.4.11 [3.4/3.6] 41.10, 43.5
References:
AOP-R.01 LP/Objectives: 0PL271C366 B.1 History: Procedure Bank Level: Memory Comments: Reviewed by J. Epperson; FHW 12/02 037 AA1.1 1
STEP ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED 2.1 Rising Secondary Radiation and Low Pressurizer Level NOTE: This section should be used if tube leakage has an observable effect upon charging flow and/or Pressurizer level OR if directed by Section 2.2 due to leak rate exceeding 75 gallons per day with an increasing rate of leakage greater than 15 gpd/30 minutes.
- 1. CONTROL charging flow as necessary to maintain Pressurizer level on program.
MONITOR Pressurizer level STABLE or IF loss of Pressurizer level is imminent,
- 2. THEN RISING.
PERFORM the following:
- a. TRIP the reactor.
- b. INITIATE Safety Injection.
- c. GO TO E-0, Reactor Trip or Safety Injection.
MAINTAIN VCT level greater than 13% IF VCT level CANNOT be maintained,
- 3. THEN using automatic or manual makeup.
ENSURE CCP suction aligned to RWST.
Page 4 of 48
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 47
- 46. AOP-R.02-B.2 001 S> Unit 1 is at 100% RTP when a small reactor coolant system leak develops. The operators are responding to the event attempting to locate the leak. The following parameters are observed:
"RCSpressure at approximately 2205 psig and decreasing
- All ice condenser doors open Containment pressure at 1.5 psid and increasing
-Pressurizer level at approximately 55% and decreasing (with Maximum charging)
Which ONE /of the following actions should the operating crew perform?
A. Initiate Phase A Containment Isolation.
B. Initiate Phase B Containment Isolation.
C. Initiate rapid load decrease per AOP-C.03.
v/D. Trip the reactor and initiate Safety Injection.
A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference, this is a management expectation. Ice doors are open and containment pressure is approaching SI setpoint of 1.54 psid.
K/A:[CFR]: 2.4.4 [4.0/4.3] [41.10 43.2]
Reference:
EPM-4 R12 3.4 B. 2 (Management expectation)
LP/Objective: OPL271C266 B.14 History: Procedure bank, old Bank Number B-0383A Level: Comprehension Comments: FHW 12/02 2.4.4 C
3.4 Management Expectations A. When taking prudent operator actions, the operator is expected to:
- 1. Stabilize the plant.
- 2. Utilize supporting procedures as appropriate.
B. When a parameter is approaching a protective action setpoint in an uncontrolled manner, the operator is expected to:
- 1. Evaluate the parameter magnitude and trend.
- 2. Take actions as necessary to place the plant in a safe condition without relying solely on automatic actions.
C. Crew briefs are expected to meet the following guidelines:
- 1. Crew briefs should occur at each major procedure transition, except when implementing RED or ORANGE path FRPs. If implementing a RED or ORANGE path FRP, no brief should be conducted; the SM or procedure reader should merely state that a RED or ORANGE path condition exists and identify the procedure being implemented. If a RED or ORANGE path condition is identified during a brief, the brief should immediately be terminated and the appropriate FRP implemented
- 2. Provided that no RED or ORANGE path condition exists, briefs may also be conducted at other times (determined by the SM) if necessary to ensure all crew members are cognizant of current plant status or mitigative actions.
1
- 3. The brief should consist of all mitigating crew( ) members, except for the brief held at the first transition from E-0. When the step is reached to transition from E-0, the person selected to monitor status trees will begin monitoring status trees and will NOT attend in the brief.
- 4. Briefs may be deferred or waived by the SM depending on plant conditions and whether the operating crew is aware of the current situation.
D. When two or more procedures must be performed concurrently, the SM is expected to determine the highest priority procedure and ensure the mitigating crew is focused on that procedure.
(1) Defined in Section 5.0, Definitions.
Page: 50 Wtdnesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK
- 49. AOP-R.05-B.5 001 Given the following plant conditions:
- Unit 1 is operating at 100% power.
stable.
- CCS surge tank level was increasing but is now Alarm is LIT.
..1 MA Qn.orA CCS LIQ EFF MON HIGH RAD"
- Surge tank vent valve is closed.
control b "
Which ONEJl"of the following describes the consistent with the above conditions?
inlet and outlet valves:
The thermal barrier containment isolation affected RCP close (the non-affected RCPS isolation valves remain open),
A. to the trip.
and the thermal barrier booster pumps affected RCP close (the non-affected RCPs isolation valves remain open),
B. to the oil coolers close, and the thermal barrier outlet CCS valves to the RCP the inlet and valves open.
booster pumps continue to run with miniflow close, the inlet and outlet CCS valves to the RCP oil coolers remain I'C. to all 4 RCPs pumps trip.
open, and the thermal barrier booster NOT 4 RCPs close, the inlet and outlet CCS valves to the RCP oil coolers are D. to all valves pumps continue to run with miniflow affected, and the thermal barrier booster open.
RCP thermal barriers do not have pump specific containment isolation valves.
A. barrier pumps will trip.
ClVs to all RCPs will close and thermal but ClVs for RCP oil coolers will not B. ClVs for RCP thermal barriers will close, trip--they do not have miniflow valves.
close. The thermal barrier pumps will C. Correct:
thermal barrier pumps will trip--they do D. The first two parts are correct, however not have miniflow valves.
J '*
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 51
- 49. AOP-R.05-B.5 001 S K/A{CFR}: 008 K1.04 [30./3.3] {41.2, 41.3, 41.4, 41.5, 41.6, 41.7, 41.8, 41.9}
008 K6.04 [2.1/2.3] {41.7}
008 A3.02 [3.2/3.2] {41.7}
009 EK3.01 [3.1/3.6] {41.5, 41.10}
009 EK3.05 [3.4/3.81 {41.5, 41.10) 009 EK3.09 [3.1/3.4] {41.5, 41.10}
[2.6/2.6] {41.7}
015 AK2.08
References:
0-AR-M12-A (B-i) 47W611-70-3 LP/Objectives: OPL271C026 Obj. 5 OPL271C026 Obj. 9 OPL271 C026 Obj. 12 OPL271 C367 Obj. 5 History: Modified SYS.bnk Q# CCS 007 stem and 2 distractors Not modified for 12/02 Level: Memory Comments: LP-5/2000. CCS 001 FHW 12/02 015AK2.08 2- 16C _
tz'" 6Ký.
Itk01
8 (B-i)
Source Setpoint I -RA-90-123A
<.X SER 718 CCS LIQ EFF MON 1-RM-90-123A Refer to TI-18 for HIGH RAD CCS Hx outlet activity setpoint calculation monitor 1Al or 1A2 Probable 1. RCS, CVCS, RHR or Primary Sampling system leakage to Causes component cooling water system. [C.3]
- 2. Performing test on monitor.
Corrective [1] CHECK instruments on 0-M-12 (1-RM-90-123A and Actions 1-RR-90-123) to determine if alarm is due to high radiation. [C.3]
E2] IF alarm is due to high radiation, THEN
[a] ENSURE 1-FCV-70-66 and 2-FCV-70-66 Component Cooling Water Surge Tank Vents are CLOSED. EC.3]
[b] IF CCS to RCP thermal barrier cooling coils has isolated, THEN GO TO AOP-M.03, Loss of Component Cooling Water, for loss of Component Cooling Thermal Barrier Booster Pump(s).
[C.3]
[c] IF source of in-leakage is unknown, THEN REFER TO AOP-R.05, RCS Leak and Leak Source Identification to locate, AND ISOLATE source of leak. [C.3]
[d] NOTIFY Radiochemical Laboratory to sample component cooling water for radioactivity. [C.3]
[e] NOTIFY RADCON to survey for radiological hazards. [C.3]
[f] CONSIDER system "feed and bleed" in accordance with 1-SO-70-1, Component Cooling Water System Train A, to reduce radiation levels. [C.3]
References 45B655-12A-0 47B601-90-38 47W610-90-2
Wednesday, October 09,2002 @ 11:45 AM HLC12-02.BNK Page: 52
- 50. CCS-B.9.A 001
- Given the following plant conditions:
- Unit #1 is operating at 100%.
- All systems aligned normal.
- Loss of ERCW Supply header 2A occurs due to a rupture in the yard.
Which ONEof the following describes indications the Unit 1 operator would see in the main control room in this event? (Assume no operator actions).
A. Ice condenser chillers trip.
B. "Diesel Gen 1 A-A Jacket Water Temp High-Low Engine 1 or 2" alarm lit on the non running 1 A-A D/G.
C. General ventilation chillers trip.
VD. CCS surge tank level increasing with auto makeup valve closed.
A. Incorrect - these chillers use RCS even though located in the Aux. Building.
B. Incorrect - ERCW is isolated on a non running D/G.
C. Incorrect - Aux Building chillers are also RCW.
D. Correct - loss of cooling to the A CCS heat exchanger would cause a heatup of A CCS train and expansion into the surge tank.
K/A {CFR}: APE 062 AA1.0513.1/3.11 {41.7, 45.5, 45.6}
008 K1.01 3.1/3.1 008 K4.02 [2.9/2.7] [41.7]
References:
0-47W859-1-4 CCS Flow 0-47W845-2 ERCW Flow 0-AR-M26-A (C-4)
AOP-M.01 App. C LP/Objectives: OPL271CCS, B.9.a, B.12 History: System bank Level: Analysis Comments: FHW 12/02 008 K4.02
F AOP-M.01 Rev.2 WATER SON LOSS OF ESSENTIAL RAW COOLING Page 1 of 1 APPENDIX C AFFECTED EQUIPMENT LIST (HEADER 2A)
- 1. Centrifugal Charging Pump 2A Oil Cooler (Note 1)
- 2. Safety Injection Pump 2A Oil Cooler (Note 1)
- 3. CCS Heat Exchangers 1A1/1A2 and 2A1/2A2 (Note 2)
- 4. Lower Containment Coolers 2A and 2C (Note 3 & 4)
- 5. Upper Containment Coolers 2A and 2C (Note 3 & 4)
- 6. Control Rod Drive Vent Coolers 2A and 2C (Note 3 & 4)
- 7. Instrument Room Cooler 2A (Note 3 & 4)
- 8. Space and room cooling equipment (Note 4)
- 9. Containment Spray Heat Exchanger 2A (Note 5)
- 12. Unit 2 RCP Motor Coolers 1 and 3 (Note 6)
Note 1: Damage is imminent unless equipment is stopped.
Note 2: Monitor RCPs for TRIP criteria. Multiple systems affected.
REFER TO AOP-M.03, Loss of Component Cooling Water.
Note 3: Containment parameters may increase due to inadequate cooling.
Note 4: Start redundant equipment and secure affected components.
Note 5: ERCW not normally required.
if alarming.
Note 6: REFER TO AOP-R.04, Reactor Coolant Pump Malfunctions, Page 62 of 69
I 18 (C-4)
Source Setpoint SER 918 DIESEL GEN 1A-A
- 1. Engine 1A1 JACKET WATER 0-TS-82-5006/1 Low water jacket temp switch 100-F (+/-3°F) TEMP HIGH-LOW 0-TS-82-5005/1 High water jacket temp switch 186°F (+/-40 F)
- 2. Engine 1A2 ENGINE 1 OR 2 0-TS-82-50Q3/1 Low water jacket temp switch 100 -F (+/-3'F) 0-TS-82-5002/1 High water jacket temp switch 186°F (+/-40 F)
Probable 1. Malfunction of the thermostatic bypass valve on the cooling system.
Causes 2. Malfunction-of immersion heaters (shutdown situation).
- 3. Insufficient heat removal due to dirty cooler, loss of ERCW supply.
Corrective [1] DISPATCH personnel to D/G Bldg to verify alarm and to monitor Actions DIG water jacket (TI-82-5006/1 and 5003/1) and lube oil temperatures (TI-82-5010/1 and 5008/1).
[2] IF D/G running, THEN
[a] ENSURE r1-FCV-67-68] or f1-FCV-67-68] OPEN, and supplying approximately 650 gpm (1-FI-67-69 or 1-FI-67-277).
[b] REDUCE the D/G 4.4MW rating by .044MW 0 for every 3.5 0F the outside ambient air temperature is >90 F.
0
[c] IFjackejt water temperature exceeds 195 F, THEN REDUCE D/G load (MVV) until jacket water temperature stabilizes below 1951F.
NOTE Diesel Generator jacket water temperature from engine shutdown limit is >205°F
[d] EVALUATE need to emergency stop the D/G AND impact on Technical Specification 3.8.1.1 or 3.8.1.2.
[3] IF D/G aligned for standby operation, THEN CHECK water jacket temperatures 115-125°F and lube oil temperatures 90.°-120°F.
[4] IF D/G in standby and water jacket temperature less than 1000F or 0
lube oil temperature less than 90 F, THEN
[a] ENSURE 11-FCV-67-661 and [1-FCV-67-68] CLOSED, (ERCW to heat exchangers).
[b] CHECK engine water immersion heaters operable by ensuring immersion heaters aligned in accordance with 0-SO-82-1, Diesel Generator IA-A, Power Availability Checklist
[5] IF lube oil temperature decreases to 851F, THEN PERFORM local D/G run in accordance with 0-SO-82-1 to increase oil temperature.
0
[6] IF lube oil temperature less than 85 F, THEN DECLARE DIG inoperable.
[7] EVALUATE impact on Technical Specifications (3.8.1.1 or 3.8.1.2).
References 45N767-2, 45N767-6, 45N767-9, 45N767-10, 45B655-26A, A950F02501 0-AR-M26-A Page 21 of 39 1&2 Rev. 17 F SQN
(CA-a44 ,,-N Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 53
- 51. CONTROL*RODS 020 At beginning of Core life, the differential control rod worth is larger near the center of the core compared to the top and bottom of the core.
Which 19 of the following explains this condition?
A. boron concentration.
B. xenon concentration.
v'C. neutron flux distribution.
D. reactor coolant temperature.
A. Incorrect per reference.
B. Incorrect per reference.
C. Correct per reference, neutron flux density is greater at the core center.
D. Incorrect per reference.
K/A[CFR]: 001 K5.30 [2.9/3.1] [41.5]
Reference:
Rx Theory- Control Rods - LP1 - PR051r2 LP/Objective: Rx Theory- Control Rods - LP1- PR051r2 Obj. 5 History: Genfunme bank Level: Memory Comments: FHW 12/02 001 K5.30 - &0"4"'" u4 -, ,,,'"b (SA ýorýo O6
U IINSTRUCTOR GUIDE KEY POINTS, AIDS, i QUESTIONSIANSWERS Example 5-2 / TP 5-17 through TP 5-18 A control rod is positioned in reactor with following neutron flux parameters:
Core average thermal neutron flux = I x 1012 neutrons/cm 2 sec.
Control rod tip neutron flux = 5 x 1012 neutrons/cm 2 sec.
If control rod is slightly withdrawn such that tip of control rod is located in a neutron flux of 1013 neutrons/cm 2 sec, what effect will this have on differential control rod worth? (Assume average flux is constant.)
Solution:
( 2 DRW @tip
- avg)
The reactivity worth of tip of control rod can be assumed to be proportional to square of neutron flux that it is in The increase in neutron flux at tip from 5 x 10'2 to 1 x 1013, which is an increase by a factor of two, produces a DRW increase by a factor of four.
- 8. As control rod moves in reactor core, Objective 8 differential worth of rod changes
- a. The neutron flux in bare Figure 5-5 / TP 5-19 homogeneous core is greatest near midplane of core. Figure 5-5 shows this axial flux variation cone~ . ..................
.do MOPLJNEP0O'eI0 I/
- b. Differential rod worth will be largest near core midplane, and lowest near top and bottom of core a
PWR / REACTOR THEORY! 25 of 49 © 1999 GENERAL PHYSICS CORPORATION CHAPTER 5
/ CONTROL RODS REV 2
I IINSTRUCTOR GUIDE KEY POINTS, AIDS, QUESTIONS/ANSWERS
- c. Anything affecting this flux distribution affects differential worth of rods
- 9. Movement of rods changes flux shape and values of differential rod worths
- a. The flux is depressed in region with rods inserted and is relatively higher in region without rods
- b. The flux distribution shifts as rods Figure 5-6 / TP 5-20 move into or out of core. ""Rl/Nai Figure 5-6 shows flux shift from core midplane to core bottom as rods are inserted
.0A.1#
- 10. When control rods are near bottom, maximum flux shifts back to core midplane
- a. Because of this flux shift, highest differential rod worth occurs at rod height below core midplane
- b. A graph of differential rod worth Figure 5-7 / TP 5-21 versus rod height in typical reactor a
with banked control rods is given in z 20 Figure 5-7 C / ",
25 50 75 100 125 150 CONTROL ROD BANK HEIGHT (INCHES)
- c. A rod bank is group of control rods which move together
,,> PWR / REACTOR THEORY I 26 of 49 © 1999 GENERAL PHYSICS CORPORATION CHAPTER 5
/ CONTROL RODS REV 2
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 54
- 52. CTMT PURGE-B.4 001
~ 0-SO-30-3 contains the following precaution: "IF operating the containment purge system in MODE 5 or 6 WHILE the other unit is in MODES 1 through 4 THEN, an operator shall be available to stop the containment purge system in the event of an ABI."
Which ONE §4 of the following describes the basis for this precaution?
A. During containment purge system operation with the unit in MODE 5 or 6 its isolation function on a high rad condition is blocked and an operator is required to isolate the system.
B. The volume of air, that could pass from the unit being purged to the ABSCE via the containment purge fans and the open blast doors, would exceed the capacity and design basis of the EGTS.
'/C. The volume of air, that could pass from the unit being purged to the ABSCE via the containment purge fans and the open blast doors, would exceed the capacity and design basis of the ABGTS.
D. During MODE 5)the automatjp Safety lnjectir functionzslQckedwghigh inhibits the ABI signal. Apt,' ow4, f-r S tv" ,,4ttrh- S .- ,.
- a. Incorrect - Not a consideration in 0-SO-30-3
- b. Incorrect - EGTS is not considered.
- c. Correct - per reference material LER 50-327/88007 R5
- d. Incorrect - blocking auto SI does not inhibit ABI signal, see 1,2-47W611-30-6.
K/A [CFR]: 2.3.9 [2.5/3.4] [43.4]
References:
0-SO-30-3 "Containment Purge Systen Operation" LER 50-327/88007 R5 1,2-47W611-30-6 LP/Objectives: OPL271 CONTPURGE B.7 History: Systems bank Level: Memory Comments: FHW 12/02 2.3.9
3.0 PRECAUTIONS AND LIMITATIONS (Continued)
D. The ice condenser doors must be monitored closely during all manipulations of the containment purge system. A slight imbalance in pressure between the upper and lower compartments will cause the doors to open. The Instructions must be followed exactly to minimize the probability of opening the ice doors.
E. During outage conditions, a flow imbalance can be created when a lower/upper compartment purge is inservice and changes are made to the upper and/or lower airlock doors (i.e. breaching or closing doors after purge is started). This situation may cause the Ice Condenser lower inlet doors to go open.
F. Operation of containment purge in Mode 5 or 6 with the equipment hatch closed and airlock doors breached may result in pressurization of the Reactor Building access rooms. This occurs due to thermal expansion of the colder outside air entering containment. Closing FCV-30-16 and FCV-30-17 and/or opening FCV-30-37 and FCV-30-40 in accordance with the applicable steps in this procedure will reduce the pressure in containment and reduce the risk to personnel entering and exiting the access rooms. This should also be used when lifting the head and moving fuel to reduce the spread of contamination.
G. When a containment ventilation isolation is initiated by a containment purge air exhaust radiation level signal, whether it is known to be spurious or not, this signal should be cleared from the circuit before resetting containment ventilation isolation.
H. IF operating the containment purge system in MODE 5 or 6 WHILE the other unit in MODES 1 through 4 THEN, an operator shall be available to stop the containment purge system in the event of an ABI. This action can be performed with minimum Tech Spec shift crew without impeding mitigation of the event. [C.1]
I. During MODE 6 operation with the wafer valve open, containment purge startup or shutdown may result in over flow of the SFP or reactor cavity due to pressure changes in the containment building.[C.2]
J. When air temperature entering the Auxiliary Building heating/cooling coils is 5:350 F, flow must be maintained through the coils unless they are isolated and drained.
SOURCE NOTES Page 1 of 1 SOURCE IMPLEMENTING REQUIREMENTS STATEMENT STATEMENT DOCUMENT LER 88007 R2 C.1 TVA will revise SOI-30.2 (SO-30-3) to include the S53 880916 892 appropriate CM that must be used when operating the containment purge system in modes 5 and 6 with the blast doors open while the other unit is in modes 1 through 4.
Deleted blast doors open requirement due to DCN M01443.
C.2 NER 91 0067001 Overflow of the spent fuel pit and INPO SER 91-001 the reactor cavity has occurred at several plants due to pressure changes in either building during operations which change the configuration of the ventilation systems.
8 0 910531-1-839 Tennessee 37379 Tennessee Valley Authority. Post ofice Box 2000. Soddy-Daisy.
Jack L. Wilson Vice President, Sequoyah Nuclear Plant May 30, 1991 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
PLANT UNITS I AND 2 TENNESSEE VALLEY AUTHORITY - SEQUOYAH NUCLEAR LICENSES DPR-77 AND 79 DOCKET NOS. 50-327 AND 50-328 - FACILITY OPERATING 5
- LICENSEE EVENT REPORT (LER) 50-327/88007, REVISION actions to ensure The enclosed LER has revised the long-term corrective design system can perform its that the auxiliary building gas treatment function during various modes of 2-unit operation.
modification to interlock As a result of a further review of the proposed the auxiliary building Units I and 2 containment purge systems with to this LER, TVA has isolation signal as presented in Revision 4 solution would be the determined that the most feasible long-term This currently in place.
continued use of the compensatory measures the existing decision considered the acceptability of maintaining in the last two outages controls in the long-term, the performance issues involved, and the utilizing the existing controls, the technical cost and/or benefit of the alternatives.
with 10 CFR 50.73, This event was originally reported in accordance paragraph (a)(2)(i)(b), on February 23, 1988.
Very truly yours, TENNESSEE VALLEY AUTHORITY Wilson Enclosure cc: See page 2
2 U.S. Nuclear Regulatory Commission May 30, 1991 JLW:MAC:CHW:KRR Enclosure cc (Enclosure):
INPO Records Center Institute of Nuclear Power Operations 1100 Circle 75 Parkway, Suite 1500 Atlanta, Georgia 30339 Mr. D. E. LaBarge, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 Mr. B. A. Wilson, Project Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 RIMS, MR 2F-C R. J. Beecken, POB 2B-SQN J. R. Bynum, LP 3B-C W. R. Cobean, Jr., LP 3B-C D. L. Conner, STC 2H-SQN M. A. Cooper, OPS 4C-SQN (Attn: J. S. Smith)
J. H. Garrity, FSB IA-WBN R. L. Lumpkin, Jr., SB lC-SQN R. W. Martin, OPS 4B-SQN (Attn: I. J. Hollomon)
V. D. McAdams, SB 2B-SQN T. J. McGrath, LP 3B-C M. 0. Medford, LP 3B-C W. J. Museler, OSA 2A-BLN D. A. Nauman, LP 3B-C Nuclear Experience Review Files, OPS 4D-SQN P. G. Trudel, DSE lA-SQN E. G. Wallace, LP 5B-C
- 0. J. Zeringue, PAB E-BFN 1442h
LICENSING TRANSMITTAL TO NRC DATE TRANSMITTED
SUMMARY
AND CONCURRENCE SHEET TO MJR ASSURE THE ACCURACY AND THE PURPOSE OF THIS CONCURRENCE SHEET IS TO COMPLETENESS OF TVA SUBMITTALS TO THE NRC.
DATE DATE DUE NRC ACTION NO.
FEES REQUIRED YES NO X SUBMITTAL PREPARED BY C. H. Whittemore Sequoyah Nuclear Plant (SQN) - Licensee Event Report PROJECT/DOCUMENT I.D.
(LER) 50-327/88007, Revision 5 to ensure that PURPOSE/
SUMMARY
To revise the long term corrective actions (ABGTS) can perform its design the auxiliary building gas treatment system function during all modes of 2-unit operation.
RESPONDS TO N/A (RIMS NO.) COMPLETE RESPONSE YES X NO PROBLEM OR DEFICIENCY DESCRIPTION Operability of ABGTS could not be assured enclosure was not always because the auxiliary building secondary containment during technical specification testing maintained within the configuration set used to verify ABGTS operability.
and evaluated long term CORRECTIVE ACTION/COMMITMENT TVA has reviewed solution that the most feasible long term alternatives and has determined the measures currently in use.
to the problem is to continue to implement N/A DATE INDEPENDENT REVIEW signatory has assured that the A concurrence signature reflects that the TVA policy, applicable submittal is appropriate and consistent with and supporting documentation for commitments are approved for implementation,
-2" submittal completeness and accuracy has been prepared.
CONCURRENCE SIGNATURE DATE NAME ORGANIZATION R. J. Beecken SQN Plant Manager PORC Chairman,.
'xA/
_M. A. Cooper SQNSite LicMgr
-DATE______________I_
APPROVED NLRA4A 1442h
Sequoyah Site Licensing Concurrence Sheet DUE TO NTD DATE DUE NRC PROJECT/DOCUMENT I.D. Sequoyah Nuclear Plant (SQN) - Licensee Event Report (LER) 50-327/88007, Revision 5 CONCURRENCE SIGNATURE OR ORGANIZATION LET E E E DATE NAME C. H. Whittemore SQN Licensing Engineer - S .*'
73 AJ. W. Proffitt SQN Compliance Lic Mgr J. S. Smith SQN Site Lic R. W. Martin SQN Site Controller 4-O CommtT#EA-Tt )
t/MA J. K. Gates SQN Technical Support Mgr 4.7?,
NRC response or approval required? __ Yes X No Licensing upon receipt.*4*
- NOTE: This sheet should be removed by Corporate 1442h
Approved 0MB No. 3150-0104 Approved OMB No. 3150-0104 NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION Expires 4/30/92 (6-89)
LICENSEE EVENT REPORT (LER)
T N 1IDOCKET NUMBER (2) I PAGE (3) 0 equovah Nuclear Plant. Unit the Boundary Set for in Containment Envelope Outside TITLE (4) Opening of Unit I Containment Results Surveillance Testing of Auxiliary B il in Tr eament System - INVOLV D LER NUMBER (6) 1 REPORT DATE 7) OTHER FACILITIES IOE EVENT OAt (5) 1 DOCKET NUMBER(S)
I I ISEQUENTIAL I IREVISIONI I I I FACILITY NAMES I I 2 101slo1o10131218 I I NUMBER IIMONTHII DAY IYEAR 1I Seuoyah, Unit MONTHI DAY IYEAR IYEAR I I I I I NUMBER I-..- I_~.I I 9 1 1 lOls ololol I 1 01 11 21 41 81 81 81 81 10 10 7 I I 0 1 5 10 15 13 I0 2 OF 10 CFR §:
OPERATING I ITHIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS MODE I I (Check one or more of the followinq)(ll) I -173.71(b)
I--120.405(c) I - 50.73(a)(2)(iv)
(9I I1l 120.402(b)
I_150.36(c)(1) I_j50.73(a)(2)(v) I-1I73.71(c)
POWER I L..120.405(a)(1)(i)
I__SO.73(a)(2)(vii) L_.OTHER (Specify in LEVEL 1 IJIZO.4O5(a)(1)(ii) I._150.36(c)(2)
I I50.73(a)(2)(viii)(A) I Abstract below and in (10) I Q1 01 01 120.405(a)(1)(iii) lIXXX 50.73(a)(2)(i)
I- 120.4O5(a)(1)(iv) I_150.73(a)(2)(ii) I I50.73(a)(2)(viii)(B) I Text, NRC Form 366A)
I IS".73(a)(2)liii ) x I I 120.405(a)(1Hv).
LICENSEE CONTACT FOR THIS LER (12)
T'LEPHON NUMBER NAME SAREA CODE I I 1 I 5 18 14 1 3 - 7 I2 I C. H. Whittemore. Compliance Licensing Engineer REPORT (13)
MPONENT FAILURE DESCRIB IN THIS COMPLETE ONE LIN FOR EACH IREPORTABLE1 IREPORTABLEI I I I I I 1 I
I
- -AUSEISYSTEMI ICOMPONENT IIMANUFACTURERII TO NPRDOS I B I I I I ICAUSEISYSTEMI I COMPONENT IIMANUFACTURERII TO NPRDS I I I I I I II V I Al HI Ul II El 31 21 2 1I Y I I II II I II I I I I I B I I I iI lI llIII 11111111111I I I I 1 EXPECTED IMONTHI DAY I YEAR SUPPLEMENTAL REPORT EXPECTED (14) I I I- I SUBMISSION I I DATE (15) I I I I I YES (If yes. complete EXPECTED SUBMISSION DATE) I X I NO fifteen single-space typewritten lines) (16)
ABSTRACT (Limit to 1400 spaces, i.e., approximately revised to describe the results of an evaluation to ensure that the This LER has been 2-unit auxiliary building gas treatment system (ABGTS) remains operable during 2 in Mode 5 (cold shutdown), it was operation. On January 24, 1988, with Units I and containment enclosure was not being discovered that the auxiliary building secondary the technical specification (TS) maintained within the configuration set during On August 24, 1988, with Unit 1 operability.
surveillance testing used to verify ABGTS purge determined that the Unit I containment in Mode 5 anri Unit 2 in Mode 1, it was properly compensatory measures being system was in operation without the required These conditions were caused by (1) the lack of adequate controls to documented.
within the condition set by surveillance ensure the ABSCE boundary was maintained made during plant construction on how testing, (2) an inappropriate design assumption an incomplete compensatory measures (CM)
ABSCE breaches would be controlled, and (3) I blast door and tagging the program. Corrective Actions included closing the Unit ABSCE service, revising the procedure governing Unit 1 containment purge system out of of the Following subsequent leak testing breaches, and upgrading the CM program.
reopened. As long-term corrective action, Unit 1 annulus, the Unit 1 blast door was through the CM program to ensure TS TVA has instituted administrative controls conditions.
requirements are satisfied under the subject NRC Form 366(6-8g1
U.S. NUCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104 NRC Form 366A Expi res 4/30/92 (6-89)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION IDOCKET NUMBER (2) I LER NUMBER (6) 1 PAGE (3) kCILITY NAME (1)
I I I ISEQUENTIAL I IREVISIONI I I I I 1 18 ERI I I NUMBER I I NUMBER I I I I I Sequoyah Nuclear Plant Unit I 1Q1510101013 IZ Is.9 i I-I I F4 I- I I 2 OF 1o 01 more space is required, use additional NRC Form 366A's) (17)
TEXT (If of the evaluation considering This LER has been revised to describe the results the ABGTS can perform its design function during various alternatives to ensure that modes of 2-unit operation.
DESCRIPTION OF CONDITION (0 percent power, 4 psig, 121 degrees On January 24, 1988, with units 1 and 2 in mode 5 F, respectively), a potential deficiency F and 0 percent power, 310 psig, 118 degrees enclosure (ABSCE) (EBlS Code WF) was in the Auxiliary Building secondary containment and subsequent discussions with test discovered during a tour of the refueling area ABSCE in accordance with personnel. The plant configuration used when testing the (SR) 4.7.8.d.3 was not consistent Technical Specification (TS) Surveillance Requirement As a modes of two unit operation.
with allowable plant configurations during various to Quality be assured, and Condition Adverse result, operability of the ABGTS could not Report (CAQR) SQP 880090 was issued.
I and 2, which share a common Auxiliary The ABGTS and the ABSCE are common to units before Building (ETIS Code NF). Both trains of the ABGTS are required to be operable The ABGTS maintains negative either unit can enter mode 4 from a mode 5 condition.
filters the ABSCE air before it is released to the pressure in the ABSCE and fuel handling environment. One ABGTS train is required to be operable for unrestricted fuel pool (although the ABGTS is not operations while irradiated fuel is in the spent in modes the ABSCE during plant operations required to maintain a negative pressure in 5 and 6).
ABGTS can maintain the spent fuel ,>
TS SR 4.7.8.d.3 requires verification that the the ABSCE at a (ESF) pump rooms within storage area and the engineered safety feature 1/4-inch water gage (wg) while pressure equal to or more negative than minus relief flow rate greater than 2000 cubic feet per minute (cfm) and maintaining a vacuum This SR is satisfied by the a total system flow rate of 9000 cfm + 10 percent.
"Auxiliary Building Gas Treatment performance of Surveillance Instruction (SI)-149, unit 2 blast unit I and System Vacuum Test." Past performances of SI-149 had both the on the 734 feet elevation) in the doors (refueling floor to containment annulus doors shut down.
purge on both units Reactor Building shield walls closed, and containment it is normal for that unit to have it!
During plant operation in modes 5 or 6, however, Opening the blast door increases the ABSCE, blast door and/or equipment hatch open.
If the equipment hatch or personnel access boundary by the addition of the annulus.
increased further by the addition of the doors are also open, the ABSCE boundary is into the ABSCE primary containment. The increased boundary causes additional leakage performances of SI-149.
that was not accounted for during the previous blast door/equipment hatch open, and the Thus, if one unit is in mode 5 or 6 with the an operational mode that requires the opposite unit is in modes 1, 2, 3, or 4 (i.e.,
would not be the same as the ABGTS to be operable), the actual plant configuration of SI-149.
configuration that was tested during the performance NPC rn'- 191 PoI
U.S. NUCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104 NRC Form 366A Expires 4/30/92 (6-89)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION I LER NUMBER (6) I PAGE (3)
ACILITY NAME (1) IDOCKET NUMBER (2) 1 I 1 ISEQUENTIAL IREVISIONI I I 1 1 1 IYEAR I I NUMBER I NUMBER I I I I I Sequoyah Nuclear Plant Unit 1 IQ1510I 01013 12 aIL I1 I _--I I0 4 I--I 0 I s I ol 310F I0 7 TEXT (If more space is required, use additional NRC Form 366A's) (17)
DESCRIPTION OF CONDITION (continued) affecting the performance of A second concern that has been identified as potentially of the containment purge system the ABGTS during an accident relates to the operation hatch open. The containment purge system, on a unit with the blast door and equipment of air into the Reactor Building (EIIS when it is operating, provides a large amount system was not accounted for Code NH). Air contributed from the containment purge status was not being controlled during the performance of SI-149, and its operational Thus, there was no hatch.
with the opening of the blast doors and the equipment if the blast door and equipment hatch assurance that TS SR 4.7.8.d.3 could be satisfied operation.
that unit was in were open, and the containment purge system for occurred on February 6, 1988), TVA In order to allow unit 2 to enter mode 4 (which the unit 1 containment purge system administratively prohibited the operation of open by implementing the provisions of whenever the equipment hatch and blast door were This TACF, which was approved on temporary alteration change form (TACF) 1-88-02-030.
the unit I containment purge fans, January 28, 1988, placed hold order 1-88-240 on thereby preventing their operation. In addition to implementing the TACF, TVA Test," to measure the leakage into tCe performed SI-264,."EGTS Annulus Vacuum Draw Down previously measured unit 1 annulus. This leakage was then conservatively added to the its intended function with the ABSCE leakage to verify that the ABGTS could perform blast door open.
it was determined that there was a Following further investigation into this event, purge system in a unit that had need to demonstrate that operation of the containment established containment integrity would not have an adverse effect1-minute on the ability-.S to minus 1/4-inch wg within the time interval the ABGTS to draw down the ABSCE (FSAR). That is, even with containment specified in the Final Safety Analysis Report purge system duct work in containment integrity established, it was postulated that the ABGTS from performing its design the Auxiliary Building could leak and prevent the function.
duct work, TVA performed smoke tests and To verify the integrity of the purge system in accordance with SI-506.7, "Containment visual inspections of the subject duct work Purge Air Exhaust Filter Train Test." However, performance of this test required had been tagged out of service by TACF operation of the containment purge system which measure was approved 1-88-02-030. In order to operate the purge system, a compensatory action was taken within four to allow operation of the system as long as operator to shutdown the signal (EIIS Code JE) minutes of an Auxiliary Building Isolation (ABI) ICF 88-0977 were and permanent system. Temporary Instruction Change Form (ICF)88-890 measure into SI-506.7.
subsequently approved to incorporate this compensatory S01-30.2, "Containment Purge System A similar (but temporary) ICF was written against system to reduce an unexpected Operation, to allow a one-time operation of the purge on July 25, 1988.
increase in the containment airborne radiation level NRC Form 366(6-89)
U.S. NUCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104 NRC Form 366A Expi res 4/30/92 (6-89)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION 1 LER NUMBER (6) I I PAGE (3)
"*LITYNAME (1) IDOCKET NUMBER (2) 1 1 I ISEQUENTIAL I IREVISIONI I I I I 1 IYEAR I I NUMBER I I NUMBER I I I I I Sequoyah Nuclear Plant Unit 1 Ig151 10101O 3 12 Is 119i I--I I 0 I 4 I--I 0 I 5 1I 410F1 01 TEXT (If more space is required, use additional NRC Form 366A's) (17)
DESCRIPTION OF CONDITION (continued) was presented to the Shift On August 24, 1988, a revision to TACF 1-88-02-030 This revision changed the tagging Operations Supervisor (SOS) for implementation.
isolation valves to only one train of boundary from both trains of containment purge with Safety Injection - D/G Containment valves to allow SI-26, "Loss of Offsite Power Isolation Test," to be performed.
the SOS realized that the unit I Upon receiving the revision to TACF 1-88-02-030, time (in accordance with S01-30.2) to containment purge system was being run at that However, since the ICF to reduce the temperature inside the unit I containmenz.
the appropriate compensatory measure SOI-30.2 had expired, the subject SO did not have for purge system operation. The SOS immediately suspended purge system operation, an investigation be initiated.
reissued the hold order on the system, and requested were aware of the This investigation revealed that, although most operators unit I purge system, these measures compensatory measures necessary for operating the were they formally communicated to had not been adequately documented in SOI-30.2, nor would have Operations personnel. Thus, there was no assurance that plant operators an ABI signal, Sshut and as a result, down the unit i containment purge system following able to perform its design there was no assurance that the ABGTS would have been function.
CAUSE OF CONDITION to ensure the ABSCE configuration The immediate cause of this condition was the failure during surveillance testing ol>
was maintained in the same configuration that was set the ABGTS in accordance with SI-149. TS 3.6.1.1 requires primary containment integrity only for a unit that is in modes 1 through 4. TS 3.7.8 requires the ABGTS to be However, operability of the operable whenever either unit is in modes 1 through 4.
Breaches of the ABSCE are ABGTS was verified only with the blast doors closed.
the Shield Building, ABSCE, or controlled by Technical Instruction (TI)-77, "Breaching properly evaluate the condition wher Control Room Boundaries." However, this TI did not (through an open blast (i) the Shield Building boundary becomes part of the ABSCE of the ABSCE (if the equipment hatch door), (2) the primary conta'.ment becomes part system is in operation.
and blast door are open), or (3) the containment purge assumptions that were made during the The root cause of this event was improper design in the ABSCE. The need for an interir period of plant construction to address breaches one unit was in operation and the ABSCE was recognized (and provided) during the time other unit was still under construction. At that time, it was also recognized that times when the need to breach the ABSCE upon completion of both units, there would be breaches would be would exist. However, it was believed at that time that most ABSCE low probability of an accident of short duration and could be justified based on the during that time.
NRC Form 366(6-89)
U.S. NUCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104 NRC Form 366A (6-89) Expires 4/30/92 LICENSEE EVENT REPORT (LER)
"TEXT CONTINUATION IOOCKET NUMBER (2) I LER NUMBER (6) I I PAGE (3)
,CILITY NAME (1) 1 1 1 I J ISEQUENTIAL I IREVISIONI I 1
I IYEAR I I NUMBER I I NUMBER I I I I Sequoyah Nuclear Plant Unit I I i S5OFr SoU1511OIO13 12 Ia 1I 1I I--I 0 I 4 -- I I 0i 7 more space is required, use additional NRC Form 366A's) (17)
TEXT (If CAUSE OF CONDITION (continued) modifications would be It was expected that long duration breaches for major similar to that established during compensated for by establishing an interim ABSCE at that time because construction. However, this design philosophy was not documented type of documentation.
no formal procedure existed that required this formal compensatory measures established Running the containment purge system without (CM) program instituted by was caused by an incomplete compensatory measures A of Compensatory Measures."
Administrative Instruction (AI)-49, "Control and Tracking that, although the program review of the compensatory measures program has shown for tracking and evaluating the effectiveness of CMs once appears to be appropriate that require CMs to be considered.
they are identified, there are no specific guidelines for (i) performing safety evaluations, Specifically, a review of implementing documents performing temporary facility changes (TACFs)
(2) performing procedure changes, and (3) these changes for nec ssary CMs.
failed to identify any requirements for evaluating a CM has been deemed appropriate,
<2 Further review of the CM program revealed that, once the CM program manager to ensure that there is only one step in AI-49 which requires CM. Although this step is certainly the implementing organization is aware of the to be accomplished. Specifically, appropriate, there was no clear method for it to shift operating crews concerning administrative measures to disseminate information were inadequate. In addition, there was CMs were not standardized, and consequently, existing CM information to be passed no administrative control in place that required on during shift turnover.
ANALYSIS OF CONDITION CFR 50.73, paragraph a.2.i.b, as a This condition was originally reported under 10 condition prohibited by TS.
verifiation that the ABGTS is operable and TS SR 4.7.8.d.3 is performed as a partial design function. Since the actual plant configuration was capable of performing its used when testing the ABGTS in nonconservatively different from the configuration assurance that the ABGTS would have accordance with TS SR 4.7.8.d.3, there was no satisfied its design function.
considered to have had a significant The condition as discovered, however, was not the public because units I and 2 were in safety consequence to the health and safety of plant during cold shutdown, and the ABGTS was not required to satisfy TS SR 4.7.8.d.3 no fuel handling operations were in progress operation in modes 5 or 6. In addition, in the spent fuel pool area.
while the opposite there have been occasions when a blast door has been open S>However, unit was not in modes 5 or 6.
NRC Form 366(6-89)
U.S. NUCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104 NRC Form 366A Expires 4/30/92 (6-89)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION
.CILITY NAME (1) IDOCKET NUMBER (2) 1 LER NUMBER (6) I I PAGE (3) 1I I SEQUENTIAL I IREVISIONI I I I I IYEAR I I NUMBER 1 [ NUMBER I I I I I Sequoyah Nuclear Plant Unit I I0151ololo1 3 12 I8 1g 11 I--J 0 1 Q I 4 -- i 0I 5 1 01 6lOFI 01 TEXT (If more space is required, use additional NRC Form 366A's) (17)
CORRECTIVE ACTIONS 4, fission products could If a LOCA had occurred while a unit was in modes 1, 2, 3, or If the fission products were released to the ABSCE have been released to the ABSCE.
(and that unit was operating its while the blast door and equipment hatch were open assurance that all radioactive materials containment purge system), there would be no be containment into the ABSCE would leaking from the ESF equipment or from primary event the environment. This postulated filtered by the ABGTS filters before reaching the offsite dose calculations for would then be outside the assumptions made in accident analysis. However, the ABGTS filters were available for filtration of air are used to filter air released from the ABSCE, and containment exhaust fil:ers containment purge system is operating.
released from the primary containment when the consisted of closing the As described previously, the short-term corrective action purge system out of service before unit 1 blast door and tagging the unit 1 containment on February 6, 1988). To allow unit 2 entered operational mode 4 (which occurred 6 while the opposite unit is in modes I, opening the blast door of a unit in modes 5 or with ICF 88-0191. Incorporation of the 2, 3, or 4, TI-77 was changed in accordance controls to ensure that the provisions of this ICF establishes administrative door and/or requirements of TS SR 4.7.8.d.3 are satisfied when one unit's blast in modes 1, 2, 3, o: 4. To account for equipment hatch is open and the other unit is and annulus become part of the the additional leakage when the primary containment area was calculated and subtracted from the ABSCE, the maximum expected leakage of this to satisfy TS SR 4.7.8.d.3 was tolerance by which the ABGTS flowrate required area that exceeded. The remaining tolerance was then used to determine the cumulative can be breached and still satisfy TS SR 4.7.8.d.3.
Test data from The maximum expected leakage was based on the FSAR value of 500 cfm.
verified that the leakage into the annulus was the most recent performance of SI-264 In addition, the majority of this leakage is from the well within the 500 cfm limit.
classified as ABSCE leakage when a blast door is Auxiliary Building which would not be open.
CMs, TVA has To ensure adequate consideration is given to establishing necessary AI-4, "Preparation, Review, Approval, and reviewed appropriate plant procedures (e.g.,
AI-9, "Control of Temporary Alterations the Use of Site Procedures/Instructions;"
Control Program;" and Order;" AI-19, Part VI: "Modifications; Permanent Design Change if the subject procedures should be revise SQA-119, "Safety Evaluations") to determine to de-termine if compensatory measures are to require personnel using these procedures Building gas treatment system can perform its involved. To ensure that the Auxiliary unit operations, TVA has enhanced S0I-30.2 design function during various modes of two Reactor Coolant System Leak," AOI-31, Abnormal Operating Instruction (AOI)-6, "Small Instruction E-O "Abnormal Release of Radioactive Materials," and Emergency Operating "Reactor Trip or Safety Injection," such that a TACF will not be required to from service during the unit 2 cycle continuously remove the containment purge system
U.S. NUCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104 NRC Form 366A Expires 4/30/92 (6-89)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION IDOCKET NUMBER (2) I LER UIMBER (6) I I PAGE (3)
,CILITY NAME (1)
II ISEQUENTIAL I IREVISIONI I I I IYEAR I I NUMBER I NUMBER I I I I I Sequoyah Nuclear Plant Unit I 1Q1510101013 I Is 19 I I--I 0 QI 1 4 I--I 0 I 5 I Q 710FI 01 7 TEXT (If more space is required, use additional NRC Form 366A's) (17)
CORRECTIVE ACTIONS (continued) to ensure refueling outage. These enhancements ensure that adequate CMs will be taken and allows for operation that the ABGTS will perform its designed function if required In addition to the above described of the containment purge system when required.
TVA has established requirements for a technical review of all procedure changes, active CMs on a periodic basis. This review will verify that all the assumptions that were originally used to justify a particular CM remain valid.
almost all CMs, TVA has Since Operations personnel are responsible for implementing In contains all active CMs.
established a CM log book in the main control room that revised to require appropriate addition, AI-5, "Shift Relief and Turnover," has been they assume shift.
Operations shift personnel to review the active CMs before TVA has implemented design To prevent recurrence of this type of event in the future, and communication quality information control procedures that requires documentation of Nuclear on site. Specifically, between design organizations and/or Operations groups for Engineering Procedure (NEP)-5.3, "External Interface Control," establishes controls the interactions between organizations outside Nuclear Engineering (NE) to ensure engineering, design and "appropriate transfer of information necessary to accomplish ensures that reviews done related services for TVA. In addition, NEP-5.2, "Review,"
appropriate Operation and Maintenance data review.
within NE include an ADDITIONAL INFORMATION TS SR.4 `
There has been one previous occurrence reported in the ABGTS failing to meet because of improper ABSCE boundary control - SQRO-50-327/84053.
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 55
- 53. CTMT-B.11 001 Which ONE Clof the features stated below is designed to provide more efficient operation of the Ice Condenser and Containment Spray for heat removal after blowdown resulting from a LOCA?
A. The Divider Barrier Seal between upper and lower containment will open to enhance mixing of the two environments.
B. Lower containment Ventilation Fans are designed to mix the air and provide additional cooling during a loss-of-coolant accident.
VC. The containment Air return fans provide flow to return air from upper containment to lower containment during a LOCA.
D. Pressure-operated Ice-Condenser doors will open to allow upper containment air to flow to lower containment.
A. Incorrect, the divider barrier seal must be operable to enhance flow through the ice condenser.
B. Incorrect per TS bases, only considered for a non LOCA accident.
C. Correct per TS bases.
D. Incorrect the doors were opened for flow from the lower to upper.
Student must comprehend containment conditions, equipment status during and after blowdown.
KIA[CFR]: 025 K5.02 [2.6/2.8] [41.5]
Reference:
TS bases for 3.6.5.6 and 3.6.5.9 LP/Objective: OPL271C024, b.11 History: Old Bank Number PL-0227. Modified distractor A to ensure one correct answer, FHW.
Level: Co fension. A*7eoty Comments: FHW 12/02 025 K5.02
CONTAINMENT SYSTEMS BASES event that observed sublimation rates are equal to or lower than design predic tions after three years of operation, the minimum ice baskets weight may be adjusted downward. In addition, the number of ice baskets required to be weighed each 9 months may be reduced after 3 years of operation if such a reduction is supported by observed sublimation data.
3/4.6.5.2 ICE BED TEMPERATURE MONITORING SYSTEM The OPERABILITY of the ice bed temperature monitoring system ensures that the capability is available for monitoring the ice temperature. In the event the monitoring system is inoperable, the ACTION requirements provide assurance that the ice bed heat removal capacity will be retained within the specified time limits.
3/4.6.5.3 ICE CONDENSER DOORS The OPERABILITY of the ice condenser doors ensures that these doors will open because of the differential pressure between upper and lower containment resulting from the blowdown of reactor coolant during a LOCA and that the blow down will be diverted through the ice condenser bays for heat removal and thus containment pressure control. The requirement that the doors be maintained closed during normal operation ensures that excessive sublimation of the ice BR will not occur because of warm air intrusion from the lower containment.
If an ice condenser inlet door is physically restrained from opening, the system function is degraded, and immediate action must be taken to restore the opening capability of the inlet door. Being physically restrained from opening is defined as those conditions in which an inlet door is physically blocked from opening by installation of a blocking device or by an obstruction from temporary or permanently installed equipment or is otherwise inhibited from opening such as may result from ice, frost, debris, R165 or increased inlet door opening torque beyond the values specified in Surveillance Requirement 4.6.5.3.1.
3/4.6.5.4 INLET DOOR POSITION MONITORING SYSTEM The OPERABILITY of the inlet door position monitoring system ensures that the capability is available for monitoring the individual inlet door position.
In the event the monitoring system is inoperable, the ACTION requirements provide assurance that the ice bed heat removal capacity will be retained within the specified time limits.
3/4.6.5.5 DIVIDER BARRIER PERSONNEL ACCESS DOORS AND EQUIPMENT HATCHES The requirements for the divider barrier personnel access doors and equipment hatches being closed and OPERABLE ensure that a minimum bypass steam flow will occur from the lower to the upper containment compartments during a LOCA. This condition ensures a diversion of the steam through the ice condenser bays that is consistent with the LOCA analyses.
3/4.6.5.6 CONTAINMENT AIR RETURN FANS The OPERABILITY of the containment air return fans ensures that following a LOCA 1) the containment atmosphere is circulated for cooling by the spray system and 2) the accumulation of hydrogen in localized portions of the contain ment structure is minimized.
SEQUOYAH - UNIT 1 B 3/4 6-5 Amendment No. 161 August 10, 1992
CONTAINMENT SYSTEMS BASES 3/4.6.5.7 and 3/4.6.5.8 FLOOR AND REFUELING CANAL DRAINS The OPERABILITY of the ice condenser floor and refueling and canal drains ensures that following a LOCA, the water from the melted ice containment spray system has access for drainage back to the containment lower compartment and subsequently to the sump. This condition ensures the availability of the water for long term cooling of the reactor during the post accident phase.
1/4.6.5.9 DIVIDER'BARRIER SEAL The requirement for the divider barrier seal to be OPERABLE ensures that a minimum bypass steam flow will occur from the lower to the upper containment compartments during a LOCA. This condition ensures a diversion of steam through the ice condenser bays that is consistent with the LOCA analyses.
3/4.6.6 VACUUM RELIEF LINES The OPERABILITY of three primary containment vacuum relief lines ensures that the containment internal pressure does not become more negative than 0.1 psid. This condition is necessary to prevent exceeding the containment design limit for internal vacuum of 0.5 psid. A vacuum relief line consists ofvalve, a self-actuating vacuum relief valve, a pneumatically operated isolation associated piping, and instrumentation and controls.
April 28, 1995 B 3/4 6-6 Amendment No. 197 SEQUOYAH - UNIT 1
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 56
- 54. CTMT-B. 11 004 During outages the CRDM motor power supply may be temporarily realigned to supply receptacle power in the lower containment.
Which ONE/4of the following describes how this condition is controlled?
With the motor breaker racked out:
A. the motor leads are lifted and re-landed on the receptacle power pack. A TACF is placed on the breaker to identify this temporary condition.
B. the motor leads are lifted and re-landed on the receptacle power pack. A Hold Order is placed on the breaker to identify this temporary condition.
C. a transfer switch is aligned to supply power to the receptacle power pack. A TACF is placed on the breaker to identify the breaker is under receptacle load.
,/D. a transfer switch is aligned to supply power to the receptacle power pack. A Caution Order is placed on the MCR handswitch to identify the breaker is under receptacle load.
A. A transfer switch is used so the leads will not have to be lifted and re landed.
A transfer switch is used so the leads will not haveSB. to be lifted and re landed.
C. A transfer switch is used , however a CO is used rather than a TACF.
D. Correct per reference.
K/A{CFR): SYS 022 2.2.11 [2.5/3.4] {41.10, 43.3}
022 K2.01 [3.0/3.1] {41.7}
2.2.11 [2.5/3.4] [41.10 43.3]
h
References:
0-SO-30-6 R17 section 8.2 [9]
LP/Objectives: OPL271CONTCOOLING B.6 History: System bank Level: Memory Comments: LP-5/2000. CV 001; FHW 12/02 2.2.11, modified selected correct answer to conform with 0-SO-30-6 as to "handswitch" not "breaker."
This is a unit difference question. ".
Sýsj<,
o, 44a L',
SON CONTROL ROD DRIVE O-SO-30-6 COOLING UNITS Rev: 17 1 &2 Page 22 of 37 Unit 1 Date 8.2 Transfer of CRDM Cooling Fan Motor Power Supply from Fan to 480v Receptacle Inside Lower Containment During Outages for Unit 1 ONLY.
(Continued)
NOTE The handswitches for the CRDM cooling fans are spring returned to mid-position.
[8] PLACE applicable handswitch for CRDM Cooling Fan in START position.
A. [1-HS-30-83A11A CRDM Cooling Fan B. [1-HS-30-92A11 B CRDM Cooling Fan C. [1-HS-30-88A]1C CRDM Cooling Fan D. [1-HS-30-80A1 D CRDM Cooling Fan
[9] PLACE Caution Order on applicable handswitch that warns NOT to manipulate handswitch due to breaker being under receptacle load.
A. [1-HS-30-83A 1A CRDM Cooling Fan B. [1-HS-30-92A 1B CRDM Cooling Fan C. [1-HS-30-88A 1C CRDM Cooling Fan D. [1-HS-30-80A 1D CRDM Cooling Fan END OF TEXT
<-I
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 57
- 55. CTMT-B.5 004 This question has reference material attached.
Unit 1 is operating at 1000 power. Monthly containment entry inspections are in progress. Which ONE,(I) of the following would result in a loss of Primary Containment Integrity?
A. An AUO makes an entry into the Unit 1 Annulus and leaves the El 690 Annulus Entry Door unlatched.
B. Primary Containment internal pressure is discovered to be at .75 psig relative to Annulus pressure.
vC. The DI water system was discovered unisolated and unattended to lower containment service outlets.
D. The Upper Containment Personnel Airlock door operating mechanism is broken and the door is locked closed.
A. Incorrect, the annulus door is not a containment air lock door.
B. Incorrect, outside the limits of LCO 3.6.1.4.
C. Correct per LCO 3.6.1.1.
D. Incorrect, actions of LCO 3.6.1.3 have been met.
K/A: 103 K3.02 [3.0 / 4.2]
069 AA1.03 [2.8/3.0] {41.7}
Reference:
Tech Spec LCO 3.6.1.1 Tech Spec LCO 3.6.1.4 Tech Spec LCO 3.6.3 1-SI-OPS-088-014.0 R1 3 Objective: OPL271C083, B.3 History: Systems Bank Level: Comprehension Comments: Provide tech specs 3.6.3, 3.6.1.1, 3.6.1.4 FHW 12/02 069 AA1.03
(
VERIFICATION OF 1-SI-OP,S-088-014.0 CONTAINMENT INTEGRITY Rev: 13 SQ Page 13 of 36 APPENDIX A Page 4 of 13 NOTE Prior to closing the following valves, US/SRO or WCC/SRO approval is required to verify that closure will not affect in-progress work.
Approval for closure (check appropriate valve(s)): El 1-VLV-59-522 LI 1-VLV-33-740 US/SRO or WCC/SRO (circle one) 690 PENETRATION ROOM EQUIPMENT ID REQUIRED LOCATION PENETRATION PRINT INITIALS POSITION REFERENCE 1-59-522 LOCKED Behind RCP seal flow indicators X-77 47W856-1 DI water to cntmt CLOSED south wall 1st IV 1-33-740 LOCKED Behind RCP seal flow indicators Service Air to CLOSED south wall X-76 47W846-2 Cntmt I I 1st IV
Date 6.0 PERFORMANCE NOTE 1 Independent verification is only required for initial performance and following valve, blind flange, or plug manipulation. Otherwise, the independent verification may be N/A'd.
NOTE 2 Containment Boundary Tags (App H of 0-GO-15) may be picked up during the performance of this instruction provided the RCS is no longer in Reduced inventory/Midloop condition.
[1] ENSURE that Prerequisites in Section 4.0 are met.
NOTE 1 Appendixes A & B are to be completed by Operations. Appendixes C and D are to be completed by Instrument Maintenance Group.
NOTE 2 Refer to Technical Instruction 0-TI-OPS-088-001.0 as needed.
NOTE 3 The following prints may be useful in identifying and locating the penetrations: 47W470-2, 47W253-2, 48N405, 48N406.
NOTE 4 Appendixes not required for this performance may be N/A'd.
[2] IF Appendix A is scheduled to be performed, THEN COMPLETE Appendix A as follows:
[a] LIST any open Tech Spec 3.6.3 LCO action items and associated method of isolation on Table B-1 of Appendix A.
[b] VERIFY the method of isolation for each item listed on Table B-1.
[c] VERIFY all portions of Appendix A are completed.
[3] IF Appendix C is scheduled to be performed, THEN COMPLETE Appendix C.
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT
- ' CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
R258
- a. Deleted.
- b. By verifying that each containment air lock is in compliance with R134 the requirements of Specification 3.6.1.3.
- c. Perform required visual examinations and leakage rate testing in R221 accordance with the Containment Leakage Rate Testing Program.
R258 March 29, 2000 "EQUOYAH - UNIT 1 3/4 6-1 Amendment Nos. 12, 130, 176, 191, 203, 217, 254
CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between
-0.1 and 0.3 psig relative to the annulus pressure.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to within the limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SEQUOYAH - UNIT 1 3/4 6-9
CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES
<-<" LIMITING CONDITION FOR OPERATION 3.6.3 Each containment isolation valve shall be OPERABLE.*
1, 2, 3 and 4.
APPLICABILITY: MODES ACTION:
- a. With one or more penetration flow paths with one containment isolation valve inoperable; except for containment vacuum relief isolation valves(s), isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed deactivated automatic valve, closed manual valve, blind flange, or check valve## with flow through the valve secured; and, verify# the affected penetration flow path is isolated once per 31 days R258 for isolation devices outside containment, and prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment.
- b. With one or more penetration flow paths with two containment isolation valves inoperable; except for containment vacuum relief isolation valves(s), isolate each affected penetration within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> by use of at least one closed deactivated automatic valve, closed manual valve, or blind flange and verify# the affected penetration flow path is isolated once per 31 days.
- c. With one or more containment vacuum relief isolation valve(s) inoperable, the valve(s) 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 />.
- d. With any of the above ACTIONS not met, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- e. The provisions of Specification 3.0.4 do not apply.
SURVEILLANCE REQUIREMENTS 4.6.3.1 Deleted R207
"*1. Penetration flow path(s) may be unisolated intermittently under administrative controls.
- 2. Enter the ACTION of LCO 3.6.1.1, "Primary Containment" when containment isolation valve leakage results in exceeding the overall containment leakage rate acceptance criteria.
- 3. Isolation devices in high radiation areas may be verified by use of administrative means.
R258
- 4. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.
- 5. A check valve with flow through the valve secured is only applicable to penetration flow paths with two containment isolation valves.
March 29, 2000
,EQUOYAH - UNIT 1 3/4 6-17 Amendment No. 12, 197, 203, 217, 254
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each automatic containment isolation valve shall be demonstrated OPERABLE at least once per 18 months by:
IR258
- a. Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.
- b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.
- c. Verifying that on a Containment Ventilation isolation test signal, each Containment Ventilation Isolation valve actuates to its isolation position.
- d. Verifying that on a high containment pressure isolation test signal, each Containment Vacuum Relief Valve actuates to its isolation position.
- e. Verifying that on a Safety Injection test signal that the Normal Charging Isolation valve actuates to its isolation position.
IR105 4.6.3.3 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5. IR207 4.6.3.4 At least once per 31 days, verify that all penetrations* not
)f being closed by OPERABLE containment automatic isolation valves and capable required
ý,.o be closed during accident conditions are closed by valves, blind deactivated automatic valves secured in their positions, except for flanges, or valves that are open under administrative control.
R258
- Except valves, blind flanges and deactivated automatic valves which are located inside the annulus or containment or the main steam valve vaults are locked, sealed or otherwise secured in the closed position. and These penetrations shall be verified closed during each COLD SHUTDOWN except such verification need not be performed more often than once per 92 that days.
\__.EQUOYAH - UNIT 1 March 29, 2000 3/4 6-18 Amendment No. 12, 81, 101, 120, 203, 254
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 66
- 64. ECA-1.2-B.1 002 Given the following plant conditions:
- Unit 2 is operating at 100% power when a LOCA~eu~sdq..,oejnmenti&
Fex'~
er~n e.,g see4P
- The operators initiated safety injection.
The crew has determined that SI cannot be terminated.
- SI has been reset.
- RCS pressure is 500 psig and dropping slowly.
Which ONE of the following best describes the procedure methodology to mitigate this condition?
A. Ifthe LOCA can NOT be isolated then transition to E-1.
,/B. Ifthe LOCA can NOT be isolated then transition to the Loss of RHR Sump Recirculation procedure.
C. When the RWST level is <27%, the operator should transfer ECCS pumps to the containment sump.
D. When the RWST level is <8%, the operator should transfer the pumps to the containment sump.
A. Incorrect - ECA-1.2 will not transition you to E-1.
B. Correct - per EPM-3 "Basis Document" C. Incorrect - ECA-1.2 makes NO provisions for this evolution.
D. Incorrect - ECA-1.2 makes NO provisions for this evolution.
K/A CFR: W/E04 2.4.4 [4.0/4.3] {41.10, 43.2}
E04 EK2.2 [3.8/4.0] 141.71
References:
ECA-1.2 LP/Objectives: OPL271C418 B.1 History: Bank Level: Memory Comments: FHW 12/02 E04 EK2.2
I STEP IACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 5. DETERMINE if LOCA isolated:
- b. GO TO E-1, Loss of Reactor or Secondary Coolant.
END Page 6 of 6
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 69
- 67. ECCS-B.2 002 2 LS.
>*-> Which ONEJ,- "ofthe following- bet sr a safety injection initiation signal?
A. High containment pressure of 2.81 psid on 2/4 pressure transmitters.
B. High containment pressure of 2.81 psid on 2/3 pressure transmitters with P-1 1 blocked.
C. High containment pressure of 1.54 psid on 2/4 pressure transmitters with P-1 1 blocked.
VD. High containment pressure of 1.54 psid on 2/3 pressure transmitters.
A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
KIA[CFR]: 00013 K1.01 (4.2-4.4) (41.2 -9)
E14 EK1.3 [3.3/3.6] {41.8 41.10}
Reference:
TI-28 History: Systems bank old Bank Number PL-1468 LP/Objective: OPL271 RPS b.10 Level: Memory Comments: FHW 12/02 E14 EK1.3
,'Y3j' 57
ýc JA-J"A 1005
ý"ý W 5tO0
NITI-28 Att. 9 SQN UNIT 1 & 2 CYCLE DATA SHEET Effective Date 04.2./.1.
Page 7 of 16 (FO(FOR INFORMATION ONLY)
Logic Block/Permissive Signal Setpoint SAFETY INJECTION SIGNALS SLR),
- 1. Containment Press Hi 1.54 psid
- 2. Pressurizer Press. Low 1870 psig
- 3. Steamline Press Low 600 psig
- 4. Manual N/A SECOND TIMER TO RESET. THEN THE SI RESET PB SI RESET - AFTER SI INITIATION, MUST WAIT FOR 60 BLOCK ANY AUTOMATIC SI ACTUATION FOR EACH TRAIN MUST BE ACTUATED. THIS WILL AUTO SI BLOCK, THE RX TRIP SIGNAL BUT MANUAL $1IS NOT BLOCKED. TO REMOVE REMOVE THE P-4 SEAL-IN SIGNAL.
BREAKERS MUST BE CYCLE TO
'ION SIGNALS (CIS) Phase, Any Signal Phase A/CVI HS CONTAINMENT VENT ISOLATION SIGNALS (CVI)
- 1. RM-90-130 & 131 High Rad Signal
- 2. SIS Any Signal
- 3. Manual Phase B Phase B Handswitch
- 4. Manual Phase A Phase A Handswitch MAIN STEAM.LINE ISOLATION SIGNALS
- 1. Containment Press Hi-Hi 2.81 psid 2/4 PTs
- 2. Steamline Press Low 600 psig 2/3 PTs on 1/4 loops Manual Below P-11 Rate Press Negative 50 psig a100 decreasing second time in 2/3 PTs on 1/4 loops Steamlineonly Enabled when Press SI
- 3. Steamline constant signal blocked.
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 70
- 68. ECCS-B.3 002 The unit is experiencing a large break LOCA. Which ONEPI~of the following best describes the order of ECCS component injection?
--Highct p ea~ Lu luwest iJess,-ur--,.
A. CCPs, SI Pumps, RHR Pumps, Cold Leg Accumulators.
,/B. CCPs, SI Pumps, Cold Leg Accumulators, RHR Pumps.
C. SI Pumps, CCPs, Cold Leg Accumulators, RHR Pumps.
D. SI Pumps, CCPs, RHR Pumps, Cold Leg Accumulators.
A. Incorrect per reference.
B. Correct per reference.
C. Incorrect per reference.
D. Incorrect per reference.
Student must comprehend the order of injection based on the pump/accumlator discharge head.
K/A [CFR]: 006 A1.07 [3.3/3.6] [41.5]
006 K1.03 (4.2 - 4.3) 006 K5.06 (3.5 - 3.9) 006 A3.01 (4.0 - 3.9) 006 A3.02 (4.1 - 4.1)
Reference:
ECCS System description section 3.
LP/Objective: OPL271ECCS B.13 History: Systems bank Old Bank Number PL-1530 Level: Memory Comments: FHW 12/02 006 A1.07L/
System Operations Emergency Operations, (Continued)
ECCS process The table below describes the ECCS process during a Large Break Loss of Coolant Obj 1O Accident (LBLOCA) in the RCS.
.P..s....... ....... .
PresUre C
"*A Safety Injection occurs. ECGS performs the following automatic actions:
>- Both CCPs start (one CCP is normally running with RCS level at the RCP seals).
> Both SIPs start.
>-Both RHR Pumps start.
> LCV-62-132, 133 close (CCP VCT suction).
> FCV-63-39, 40 open (CCP cold leg discharge).
- FCV-63-25, 26 open (CCP cold leg discharge).
< 1870 psig > FCV-62-90 and 62-91 close. (Regenerative Heat Exchanger valves).
> FCV-63-118, 98, 80, 67 receive an open signal(should already be open) (Accumulator discharge).
"*The CCPs inject water into the RCS Cold Legs, since RCS pressure is less than the shutoff head of the CCPs (2250 psig).
"*The SIPs flow recirculates back to the RWST through valve FCV-63-3, since the shutoff head of the SIPs is 1500 psig.
"*The RHR pumps flow recirculates through valve FCV-74-12, 24, since the shutoff head of the RHR pumps is 180 psig.
<1500 psig The SIPs begin to inject water into the RCSGold Legs.
The Accumulators begin to inject water into the RCS Cold Legs,
< 640 psig since the Accumulators nitrogen gas pressure is now greater than RCS pressure.
< 180 psig The RHR pumps begin to inject into the RCS Cold Legs.
page next on Continued Continued on next page 063.doc 3-7 Rev 1
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 72
- 70. ES-0.0-B.3 001 Given the following plant conditions:
- Reactor trip and SI have occurred 10 minutes ago
- RCS Tcold 535°F
- RCS Pressure 1810 psig
- Pressurizer level 15%
- S/G NR Levels 20% each
- S/G #1 Pressure 875 psig
- Other S/G Pressures 890 psig
- AFW Flow is 150 gpm to each S/G Which ONE pi/of the following identifies the type of accident that is in progress?
,A. Small break LOCA B. Steam line break C. Feed line break D. S/G tube rupture A. Correct for initial conditions.
B. Incorrect since all S/G pressure are approx. the same after 10 minutes.
C. Incorrect since all feed flows are the same.
D. Incorrect since all S/G levels are approx. 20% after 10 minutes.
K/A: 000009A202 [3.5/3.8]
E01 EA1.1 [3.7/3.7] (41.7)
Reference:
ES-0.0, pages 4 &5 Objective: OPL271 C380, b.3 History: Procedure Bank Level: Comprehension Comments: FHW 12/02 E01 EA1.1l
STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 1. DETERMINE procedure applicability: IF SFrequired, THEN
"* CHECK anySI pump RUNNING. PERFORM the following:
"* CHECK CCPIT flow INDICATED. a. ACTUATE Si.
"* CHECK E-0, Reactor Trip or Safety b. GO TO E-0, Reactor Trip or Safety Injection, previously COMPLETED. Injection.
IF SI NOT required, THEN RETURN TO procedure and step in effect.
- 2. CHECK S/G secondary pressure IF controlled cooldown in progress, boundary integrity: THEN GO TO Step 3.
- Any S/G pressure stable or rising.
IF controlled cooldown NOT in progress, THEN PERFORM one of the following:
"*IF main steamlines NOT isolated, THEN GO TO E-2, Faulted Steam Generator Isolation.
"*IF main steamlines isolated, THEN GO TO ECA-2.1, Uncontrolled Depressurization of All Steam Generators.
Page 3 of 6
ISTEP ACTION/EXPECTED RESPONSE T RESPONSE NOT OBTAINED
- 3. CHECK S/G secondary pressure VERIFY all Faulted S/G(s) ISOLATED:
boundary integrity:
"*MSIVs and bypasses CLOSED
"* S/G pressures controlled or rising "*AFW ISOLATED
- MFW ISOLATED
"* S/G pressures
- Atmospheric relief CLOSED greater than 140 psig.
- S/G blowdown valves CLOSED
- Steam supply to TD AFW pump ISOLATED (S/G 1 or 4).
IF any Faulted S/G NOT isolated, THEN GO TO E-2, Faulted Steam Generator Isolation.
Page 4 of 6
I STEP ACTIONEXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 4. CHECK S/G tube integrity: PERFORM the following:
- a. CHECK the following indications of a 1) VERIFY procedure in effect is one of the S/G tube rupture, including available following E-1 or ECA-1 series trends prior to isolation: procedures:
- Any S/G level rising in an uncontrolled manner. 0 E-1, Loss of Reactor or Secondary Coolant.
"* Main steamline high radiation.
"* ES-1.2, Post LOCA Cooldown OR and Depressurization.
"* Condenser exhaust high "* ES-1.3, Transfer to RHR radiation. Containment Sump.
OR "* ES-1.4, Transfer to Hot Leg Recirculation.
- S/G blowdown recorder RR-90-120, pen #1 and pen #2 "* ECA-1.1, Loss of RHR Sump high radiation. Recirculation.
OR "* ECA-1.2, LOCA Outside Containment.
- Post-Accident Area Radiation 2)
Monitor recorder RR-90-268B, IF E-1 or ECA-1 series procedure points 3 (blue), 4 (violet), in effect, 5 (black), or 6 (brown) THEN high radiation. RETURN TO procedure and step
[M-31 (back of M-30)] in effect.
- 3) IF E-1 or ECA-1 series procedure NOT in effect, THEN GO TO E-1, Loss of Reactor or Secondary Coolant.
Page 5 of 6
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 79
- 77. FH-B.12.B 001
"-< Which ONE$of the following describes the function of the SFP bridge crane hoist "lower" limit gear operated switch?
VA. Stops downward travel before the crane hook enters the water.
B. Stops downward travel before the fuel assembly reaches the bottom of the fuel cell.
C. Automatically changes hoist speed from fast to slow while the fuel assembly is in the fuel cell boundary.
D. Prevents bridge travel when the fuel assembly is within the fuel cell boundary.
A. Correct. This prevents contamination of the hook and possible contamination of SFP water.
B. Fuel handling operator uses a load cell to monitor and stop downward travel when the FA reaches the bottom of the fuel cell.
C. There is no fast/slow speed on the hoist. It only has a normal speed with a jog button D. Bridge travel is interlocked with the full up limit switch.
K/A{CFR}: 036 AK3.02 [2.9-3.6] {41.5, 41.10}
034 K4.02 [2.5/3.3] {41.7}
034 A3.01 [2.5/3.1] {14.7}
References:
FHI-3 Sect. D. IV.A LP/Objectives: System Description 079a Obj. 12.b History: Refuel.bnk Q# OPL274A069.5 004. Revised stem and D distractor.
Level: Memory Comments: LP-5/2000.; FH 001; FHW 12/02 034 K4.02
SON FHI-3 MOVEMENT OF FUEL Rev: 36 1,2 Page 44 of 97 SECTION D Page 2 of 9 Unit- Date IV. INTERLOCKS AND BYPASS SWITCHES A. SFP Bridge Hoist Interlocks/Limit Switches:
- 1. The hoist is interlocked with the bridge crane such that the hoist will not move up or down if the bridge is moving in either direction.
- 2. The bridge is interlocked with the hoist such that the bridge can not move in normal speed if the hoist is not full up (jog would be available).
- 3. The lower limit switch is a gear operated switch that will activate to stop hoist travel before the hook enters the water.
- 4. The upper limit is also a gear operated switch that will stop upward travel before the hook reaches the hoist drum block assembly.
- 5. The "hoist overtravel rod" is a backup mechanical-electrical stop that stops upward travel when the hook hits the rod just below the hoist drum block assembly.
B. Transfer Cart/Upender Interlocks:
- 1. Valve Open - Wafer valve has to be fully open before any traverse can begin - cannot be bypassed with a switch.
- 2. Frame Down - Both frames have to be down before traverse can begin can be bypassed with a switch.
- 3. Conveyor at Pit - Conveyor has to be at the SFP side before upending can begin on the SFP side - can be bypassed with a switch.
- 4. Conveyor at Rx - Conveyor has to be at the Rx side before upending can begin on the Rx side - can be bypassed with a switch.
- 5. Manipulator Crane - The manipulator crane has to be over the core or the gripper tube is fully up - cannot be bypassed with a switch.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 78
- 76. ES-1.2-B.2 006
\<> During the performance of ES-1.2, "Post-LOCA Cooldown and Depressurization," it is desirable to have only one RCP running.
Which ONE,,of the following describes the reason for having only one RCP in service?
A. One RCP provides the DELTA-P required to provide letdown. Additional RCPs would add unnecessary heat load.
,/B. One RCP is desired for spray and RCS heat transport to the SGs. Additional RCPs would add unnecessary heat load.
C. One RCP is needed for RCS heat transport to the SGs. Additional RCPs could overload the electrical power supply.
D. One RCP is desired for spray and RCS mixing. Additional RCPs would strain the plant electrical power supply in the post-LOCA condition.
A. Incorrect - RCS pressure provides force for letdown.
B. Correct - per EPM-3 "Basis Document" C. Incorrect - RCPs are fed from independent 6.9 Kv Unit Boards and will not overload them.
D. Incorrect - RCPs are fed from independent 6.9 Kv Unit Boards and will not overload them.
K/A {CFR}: W/E03 EK3.1 E3.3/3.7] {41.5, 46.13}
E03 EA1.2 [3.7/3.9] {41.7}
References:
EPM-3-ES-1.2. Step 14 LP/Objectives: OPL271C387 B.2 History: Procedure Bank Level: Comprehension Comments: FHW 12/02 E03 EA1.2 rjt7(< 3, ý
SQN EOI BASIS DOCUMENT FOR ES-1.2 EPM-3-ES-1.2 PROGRAM POST LOCA COOLDOWN AND Rev. 3 MANUAL DEPRESSURIZATION Page 29 of 81 EOP Step Number: 14 DETERMINE if one RCP should be started:
ERG Step Number: 12 Check IfAn RCP Should Be Started:
Purpose:
To establish forced circulation flow from one RCP.
ERG Basis:
Forced coolant flow is the preferred mode of operation to allow for normal RCS cooldown and provide pressurizer spray. If RCPs had not been tripped, all but one are now stopped to minimize heat input to the RCS. The RCP started or left running should be one that can provide normal pressurizer spray (see preceding note). If no RCP is running, RCS subcooling, pressurizer level, and certain plant specific conditions are required before starting an RCP.
Depressurization of the RCS may generate a steam bubble in the upper head region of the reactor vessel if no RCP is running. This bubble could rapidly condense during pump startup, drawing liquid from the pressurizer and reducing reactor coolant subcooling. If pressurizer inventory is not sufficient, level may decrease off span. In addition, local flashing of reactor coolant could occur if RCS subcooling is not adequate. These conditions would require SI reinitiation and may confuse the operator if such behavior was unexpected.
Knowledge:
Plant-specific procedures for starting an RCP may require a steam bubble to be present in the pressurizer. RCP restart should be permitted if an RCS leak path is certain since the leak ensures that there will not be a significant pressure surge when the RCP is started.
EOP Basis:
Same.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 77
- 75. ES-1.2-B.2 004 S' Given the following plant conditions:
A LOCA has just occurred on Unit 1 and appropriate responses have been taken.
Only #2 RCP is running.
The crew is commencing cooldown and depressurization.
post-LOCA epressu-cst E£-!.2, O~qwn an aio:ntncctos rcavt oRGS vuidiyg.
Which ONE/of the following is an indication that voiding exists in the RCS?
A. Rapidly increasing RVLIS Upper Plenum level if auxiliary spray is initiated.
B. Rapidly decreasing safety injection flow rate when RCS pressure is decreased.
"VC. Rapidly increasing pressurizer level when normal spray valves are opened.
D. Rapidly decreasing RCS pressure if high head injection is realigned through the CCPIT.
A. Incorrect, auxiliary spray would cause pressurizer level to increase.
B, Incorrect, decreasing RCS pressure would allow increased SI flow.
C. Correct per ES-1.2.
D. Incorrect, SI flow throught the CCPIT would increase RCS pressure.
K/A {CFR): 000008K301 [3.7/4.4]
E03 EK1.3 [3.5/3.8] {41.8 41.10)
References:
ES-1.2, caution at step 13.
LP/Objectives: OPL271C387, B.2 History: HLC 9809 Audit Exam, HLC 12/02 Level: Memory Comments: FHW 12/02 E03 EK1.3
I STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 11. CHECK if ECCS in service: GO TO Step 20.
"* Any SI pump RUNNING.
"* CCPIT flow INDICATED.
"* Any RHR pump RUNNING in SI mode.
- 12. TURN OFF pressurizer heaters.
CAUTION Upper head voiding due to RCS depressurization may result in rapidly rising pressurizer level.
- 13. DEPRESSURIZE RCS to refill pressurizer:
- a. USE normal pressurizer spray. a. USE one pressurizer PORV.
IF pressurizer PORV NOT available, THEN USE auxiliary spray USING EA-62-4, Establishing Auxiliary Spray.
- b. INITIATE RCS depressurization.
- c. WHEN pressurizer level greater than 20% [35% ADV],
THEN STOP RCS depressurization.
Page 7 of 27
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 74
- 72. ES-0.1-B.1 002 A reactor trip following a spurious turbine runback occurred on Unit 1. The following plant conditions are observed:
- RCS Tavg = 5470 F
- RCS pressure = 2198 psig and slowly increasing
- CCP 1 A-A tripped on overcurrent at the time of the trip
- Control Rods are fully inserted except Rods H-8 and C-4 are at 24 and 16 steps respectfully.
Which ONEPIOf the following describes the appropriate action for the OATC to take?
A. Start CCP 1B-B, open FCV-62-132 and FCV-62-133 (VCT Outlet), close FCV-62-135 and FCV-62-136 (RWST Supply to CCP Suction), verify at least 80 gpm charging flow through the normal charging flowpath.
,/B. Start CCP 1B-B, shift the BAT pumps to fast, open FCV-62-138 (Emergency Borate), verify at least 35 gpm boric acid flow.
C. Shift the BAT pumps to fast, open FCV-62-138 (Emergency Borate), verify at least 35 gpm boric acid flow.
D. Shift the BAT pumps to fast, open FCV-62-140 (Normal Boration to VCT Outlet) and FCV-62-128, verify at least 30 gpm boric acid flow.
A. incorrect this is normal boration path.
B. correct per EA-68-4 requires emergency boration.
C. incorrect a CCP is required for emergency boration.
D. incorrect this is not an emergency boration path.
KIA (CFR): 024 AK3.02 024 AA1.26 [3.3/3.3] (41.7}
References:
EA-68-4 ES-0. 1 LP/Objectives: OPL271C381, b.1 History: Procedure Bank Level: Comprehension Comments: FHW 12/02 024 AA- 07
4.1 Section Applicability (Continued)
- 3. IF entering this instruction from ES-0.1 due to two or more control rods indicating greater than 12 steps, THEN PERFORM the following:
- a. IF using BAT as boration source, THEN GO TO Section 4.2, Emergency Boration from BAT. E
- 4. IF Emergency Boration to be terminated, THEN GO TO Section 4.4. El
- 5. RETURN TO procedure and step in effect. El a
END OF SECTION
4.2 Emergency Boration from the BAT
- 1. PLACE boric acid transfer pumps in fast speed. LI
- 2. ADJUST emergency borate valve [FCV-62-1381 to obtain boric acid flow between 35 gpm and 150 gpm on [FI-62-137A1. IL
- 3. MONITOR emergency boration flow:
- a. CHECK emergency boration flow established on [FI-62-137A1..I
- b. IF boric acid flow less than 35 gpm, THEN CLOSE recirculation valve for the BAT aligned to the blender: El
"* 1-FCV-62-237 for BAT A.
"* 0-FCV-62-241 for BAT C.
"* 2-FCV-62-237 for BAT B.
- 4. IF emergency boration flow NOT established, THEN ALIGN normal boration path:
- a. VERIFY VCT outlet valves LCV-62-132 and LCV-62-133 OPEN. IL
- b. ALIGN normal boration to VCT outlet:
"* OPEN FCV-62-140. LI
"* OPEN FCV-62-144. LI
- c. CHECK boration flow greater than 35 gpm on FI-62-139. []
I STEP JACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 4. g. Total feed flow to S/Gs g. ENSURE AFW operation greater than 440 gpm. USING EA-3-8, Manual Control of AFW Flow.
IF AFW can NOT be established, THEN ESTABLISH MFW pump operation and MFW flow control on MFW regulating bypass valves.
NOTE A control rod is considered fully inserted if it is less than or equal to 12 steps from bottom.
- 5. VERIFY all control rods fully inserted. two or more RPIs indicate reater than 12 steps,
"* Rod bottom lights LIT.. HEN I- "* Rod position indicators MERGENCY BORATE USING EA-68-4, less than or equal to 12 steps. mergency Boration, to compensate for loss f shutdown margin due to stuck control
)d(s).
- 6. ANNOUNCE reactor trip USING PA system.
Page 6 of 16
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 81
- 79. FR-C.1-B.2 003
"-> Given the following plant conditions:
- The Unit has tripped from 100% power with a LOCA in progress.
- Pressurizer pressure indicates 900 psig.
- Reactor Coolant Pumps have been tripped.
Core exit thermocouples are at 720 F.
- RVLJS lower plenum indicates 36%. 0 Which ONE of the following describes the conditions existing in the core as applicable t6 the Emergency Operating Procedures?
and v'A. Superheated conditions, which present an imminent challenge to the fuel matrix fuel cladding.
B. Superheated conditions, which do NOT present a challenge to the fuel matrix and fuel cladding as long as hot leg temperatures are at saturation conditions.
C. Saturated conditions, which present Whallenge to the fuel matrix and fuel cladding.
D.*-atwated conditions, which do NOT present a challenge to the fuel matrix and fuel cladding.
A. Correct, per steam table. RCPs off and RVLIS < 40 % equates to core water level < 3.5 feet from bottom of the fuel, therefore, imminent challenge is correct.
B. Incorrect, water level is < 3.5 feet from bottom of the fuel.
C. Incorrect, conditions are superheat.
D. Incorrect, conditions are superheat and water level is < 3.5 feet from bottom of the fuel.
K/A {CFR): 074 EK1.01 [4.3/4.7] {41.8 41.10}
References:
FR-C.1 EPM-3-FR-C.1 LP/Objectives: OPL271C398, B.2 History: HLC 9809 Audit Exam HLC 12/02 Level: Comprehension 7
Comments: Provide steam tables. FHW 12/02 074 EK1.01g
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED STEP
- 9. CHECK RVLIS lower range indication:
- 9. P~~~a IF risingI
- a. CHECK indication greater than 40%. a. IF THENrising, GO TO Step 2.
Pý IF THENstable or dropping, GO TO Step 10.
r
- b. RETURN TO procedure and step in effect.
Page 8 of 19
SQN EOX BASIS DOCUMENT FOR FR-C. EPM-3-FR-C.1 PROGRAM INADEQUATE CORE COOLING Rev. 2 MANUAL IPage 16 of 51 EOP Step Number: 9 CHECK RVLIS lower range indication:
ERG Step Number: 6 Check RVLIS Full Range Indication:
pu-ose:
To check if an RCS inventory condition symptomatic of an inadequate core cooling condition still exists.
ERG Basis:
The trend in RVLIS full range indication is used to check the effectiveness of safety injection in restoring RCS inventory. If increasing, then no further action may be necessary. The operator is instructed to return to Step 1 and repeat the initial guideline steps until the RVLIS full range indication is greater than (3).
If the RVLIS full range indication is greater than (3), then safety injection has been successful in restoring RCS inventory and core cooling. This step will transfer the operator to the guideline and step in effect.
If the RVLIS full range indication is not increasing, then the operator is instructed to check core exit TCs in the next step to determine if an inadequate core cooling condition still exists.
EOP Basis:
Same.
Deviation:
RVLIS full range changed to RVLIS lower range.
Justification:
AT SQN, RVLIS lower range monitors vessel level in the active fuel region.
Setpoint:
Identifier: <JO1>
==
Description:==
RVLIS value which is 3.5 feet above the bottom of active fuel in core with zero void fraction - normal accuracy.
HLC12-02.BNK Page: 82 Wednesday, October 09,2002 @ 11:46 AM
- 80. FR-C.1-B.2 011
"' Given the following plant conditions:
Core Cooling."
- The operating crew entered procedure FR-C.1, "Inadequatewere unsuccessful.
flow
- All attempts to establish high pressure Safety Injection
- RVLIS lower range level is 28% and dropping slowly.
F and slowly increasing.
- Core Exit Thermocouples are reading 820 degrees
- Reactor Coolant pumps have been secured.
NEXT step in mitigating the core Which ONE P$of the following methods would be the cooling challenge?
(SAMGs).
A. Enter the Severe Accident Management Guidelines allow RCS B. Prior to RCP restart open available pressurizer PORVs to pressures.
depressurization to the SI accumulator and SI injection
",/C.Depressurize all intact steam generators using Steam dumps or ARVs to allow RCS pressures.
depressurization to the Si accumulator and SI injection one RCP in an idle loop to D. When RCP support condtions are available restart provide forced two-phase flow through the core.
- a. Incorrect - This event does not require entry into cooling is not restored after the
- b. Incorrect - If RCPs cannot be started, or if core vent paths are not RCPs are started, then RCS vent paths are opened. RCS considered prior to restart of RCP.
is not established
- c. Correct -This is first evolution attempted if normal ECCS in reducing core not successful
- d. Incorrect - If the secondary depressurization is are not required to temperatures, then the RCPs are started. Support conditions start a RCP.
yt /4671 /
'I-
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 83
- 80. FR-C.1-B.2 011
> K/A {CFR}: EPE 074 EK2.05 [3.9/4.1] {41.7, 45.7) 074 EK2.02 [3.9/4.0] {41.7)
References:
FR-C.1 "Inadequate Core Cooling" step 14 LP/Objectives: OPL271 C398 B.2 History: Bank; Modified distractor B to include "prior to RCP restart" to avoid confusion as to assumed procedure step progress. Modified distractor D to include "when RCP support conditions are available" to avoid confusion as to assumed procedure step progress. FHW 12/02 Level: Analysis Comments: FHW 12/02 074 EK2.02
STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINEDI I
NOTES
- Blocking low steamline pressure SI as soon as pressurizer pressure is less than 1920 psig will prevent an inadvertent MSIV closure and keep the condenser available for steam dump.
"* After the low steamline pressure SI signal is blocked, main steamline isolation will occur if the high steam pressure rate is exceeded.
"* S/G depressurization at the maximum rate may cause S/G narrow range levels to drop to less than 10% [25% ADV]. This is acceptable and expected for this inadequate core cooling condition.
- 14. DEPRESSURIZE Intact S/Gs to reduce RCS pressure to less than 125 psig:
- a. WHEN RCS pressure less than 1920 psig, THEN PERFORM the following:
- 1) BLOCK low steamline pressure SI.
- 2) CHECK STEAMLINE PRESS ISOL/SI BLOCK RATE ISOL ENABLE permissive LIT.
[M-4A, A4]
- b. DUMP steam to condenser b. DUMP steam at maximum rate at maximum rate. USING Intact S/G atmospheric relief(s).
IF local control of atmospheric relief(s) is necessary, THEN DISPATCH personnel to dump steam USING EA-1-2, Local Control of S/G PORVs.
(Step continued on next page.)
Page 15 of 22
I STEP I ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 14. c. CHECK RCS pressure c. IF dropping, less than 125 psig. THEN GO TO Caution prior to Step 12.
IF stable or rising, THEN GO TO Note prior to Step 21.
- d. STOP S/G depressurization.
- 15. MONITOR if CLAs should be isolated:
- a. CHECK RCS pressure a. GO TO Note prior to Step 21.
less than 125 psig.
w
- b. RESET SI and CHECK the following:
0 AUTO S.I. BLOCKED permissive LIT. [M-4A, c4]
- S.I. ACTUATED permissive DARK. [M-4A, D4]
- c. CLOSE CLA isolation valves. c. VENT any unisolated CLA USING EA-63-1, Venting Unisolated Cold Leg Accumulator.
Page 16 of 22
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 84
- 81. FR-H.1-B.3 004
"- The emergency procedures have been implemented and the following plant conditions exist:
- 1. Performance of FR-H.1, "Loss of Secondary Heat Sink", is in progress.
- 2. Neither Feedwater or Condensate flow can be established.
Which ONES1)lof the followinga'"f' Qind b',or* methods will be used?
A. Feed makeup from the CVCS to the Reactor Coolant System and use the Pressurizer PORV's to bleed off the decay-heat energy.
,/B. Feed ECCS to the Reactor Coolant System and use the Pressurizer PORV's to bleed off the decay-heat energy.
C. Feed makeup from the CVCS to the RCS and use normal Letdown to bleed the RCS.
D. Feed ECCS to the RCS and use the Steam-Generator ARV's to bleed steam to the atmosphere.
A. Incorrect, procedure requires actuation of SI to start ECCS pumps to feed.
B. Correct per FR-H.1.
C. Incorrect, procedure requires actuation of SI and bleeds using PORVs.
D. Incorrect, procedure requires using PORVs for a bleed path.
K/A[CFR]: E05EK2.2 (3.9-4.2)
E05 EA1.3 [3.8/4.2] {41.7)
Reference:
FR-H.1 LP/Objective: 0PL271C401, b.3 History: Procedure bank old Bank Number PL-0900 Level: Co_ ension ,4zne-'-",
Comments: FHW 12/02 E05 EA1.3 Vt e 6 2 13 1Pj
~ (tl
I STEP I ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 17. VERIFY RCS feed path: START pumps and ALIGN valves as necessary.
- a. CHECK ECCS pump status:
"* At least one CCP RUNNING IF a feed path CANNOT be established, THEN CONTINUE attempts to establish feed flow.
OR GO TO Step 10.
"* At least one SI pump RUNNING.
- b. CHECK ECCS valves ALIGNED as appropriate:
"* REFER TO EA-63-5, ECCS Injection Mode Alignment.
"* REFER TO ES-1.3, Transfer to RHR Containment Sump.
- REFER TO E8-1.4, Transfer to Hot Leg Recirculation.
- 18. ESTABLISH RCS bleed path:
- a. CHECK power to pressurizer PORV a. DISPATCH personnel to restore power to block valves AVAILABLE. block valves USING EA-201-1, 480 V Board Room Breaker Alignments.
- b. CHECK pressurizer PORV block b. OPEN block valves.
valves OPEN.
- c. OPEN pressurizer PORVs.
Page 18 of 30
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 85
- 82. FR-P.1 001 FR-P.1, Response to Pressurized Thermal Shock, has the operator check that RHR flow is greater than 1500 gpm if RCS pressure is less than 180 psig.
This step is based on:
A. ensuring adequate mixing in the cold leg downcomer region during natural circulation conditions.
B. preventing core exit temperatures from exceeding the required temperature to place RHR in service.
C. ensuring adequate Low Head Safety Injection cooling prior to Accumulator isolation.
,ID. preventing implementation of Pressurized Thermal Shock (PTS) actions if a Large Break LOCA has occurred.
A. Incorrect, RHR does not mix the CL region; RCS is uncoupled.
B. Incorrect, CET has no input to RHR C. Incorrect, Isol CLA in P.1 D. Correct per EPM-3-FR-P.1 K/A: E08 EK3.2 [3.6/4.0] {41.5 41.10)
Reference:
EPM-3-FR-P.1 Objective: OPL271 C406 b.3 History: Procedure Bank Level: Comprehension Comments: FHW 12/02 E08 EK3.2V/-
SQN EOI BASIS DOCUMENT FOR FR-P.1 EPM-3-FR-P.1 PROGRAM PRESSURIZED THERMAL SHOCK Rev. 2 MANUAL Page 8 of 61 EOP Step Number: 3 CHECK RCS pressure greater than <B08> psig.
ERG Step Number: N/A
Purpose:
To determine if the entry into FR-P.1 was due to a large-break LOCA.
ERG Basis:
For transients where RCS pressure is less than the RHR pump shutoff head and flow from the RHR pumps has been verified, the operator should return to the guideline and step in effect since these symptoms are indicative of a large-break LOCA. In this instance, the actions in FR-P.1 should not be performed since pressurized thermal shock is not a serious concern for a large-break LOCA.
A large-break LOCA will depressurize the RCS to containment pressure in a matter of minutes. RCPs will be tripped due to low pressure and natural circulation will stop as fluid levels in the S/G tubes drop below the U-bends. A safety injection signal will be actuated by low pressurizer pressure. Large volumes of water will be injected into the loop cold legs, leading directly to the severe cooling of the vessel downcomer. Most of the water will come from the RWST, which is at the ambient temperature of its location; this may be either indoors or outdoors, so water temperature could range from 70 0 F down to near freezing. Another source of water injected into the cold legs comes from pressurized accumulators located inside containment. Depending on containment design and power history, ambient temperature could range from 700 F to 120 0 F. This relatively cold safety injection water produces significant thermal shock in the downcomer region, but for the large-break LOCA, pressure is not a concern. Most of the water initially in the RCS, and most of the colder water added by injection, will flow out the break. This break flow mixes together in the containment sump to result in a somewhat higher temperature for the sump water than that purely from injection. During the recirculation phase of recovery, this warmer fluid is pumped back into the RCS, somewhat offsetting the initial thermal shock.
EPM-3-FR-P.1
[SQN EOI BASIS DOCUMENT FOR FR-P.1 Rev. 2 PROGRAM PRESSURIZED THERMAL SHOCK Page 9 of 61 MANUAL I' -
As discussed in the Pressurized Thermal Shock (PTS) Training Program documentation, temperature stress alone is normally not sufficient to threaten the integrity of the pressure vessel. The PTS scenario also assumes that the RCS is pressurized and a critical flaw which is the correct size, shape, and orientation exists at the inner surface of the vessel wall. In addition, the warm prestressing effect, which results in an apparent increase in fracture toughness of the vessel material, is applicable to this situation since repressurization of the RCS is virtually impossible following a large-break LOCA. Any material flaw would be prevented from reinitiating and growing beyond 75% of the vessel wall thickness. Therefore, for a large-break LOCA, PTS is not a serious concern and Guideline FR-P.1 does not need to be implemented. (Reference DW-92-032).
EOP Basis:
Same. The value for <B07> and <BO8> (corresponding to the ERG normal and adverse containment setpoints) has been calculated to be the same. Therefore, only one value is displayed for these setpoints. Refer to EPM-6, EOI Setpoints Document, for additional information.
Deviation:
None.
Justification:
N/A.
Setvoint:
Identifier: <B07>
Description:
Shutoff pressure for RHR pumps - normal accuracy.
Identifier: <BO8>
Description:
Shutoff pressure for RHR pumps - adverse containment accuracy.
Identifier: <S03>
Description:
Minimum RHR pump flow to indicate injection into the RCS.
HLC12-02.BNK Page: 86 Wednesday, October 09,2002 @ 11:46 AM
- 83. PR-Z.1-B.2 001 FR-Z.1, "High Containment Pressure", directs that if ECA-1 .1, "Loss of RHR Sump steps in Recirculation", is in effect then operate containment spray using applicable ECA-1.1.
to ECA-1.1?
Which ONE.qf the following describes the basis for giving priority A. ECA-1.1 operates the containment spray pumps to maintain level in the containment sump for the RHR pumps.
B. ECA-1.1 operates the containment spray puppsZtoen o nu sur "' pS.,A,,-"- S >
. 7 tpump swapover of C. ECA-1.1 operates the containment spray pumps to prevent automatic the spray pumps to the containment sump.
the level in the
,/D. ECA-1.1 operates the containment spray pumps in order to conserve RWST.
A. Incorrect, not addressed in EPM-3-ECA-1.1.
B. Incorrect, not addressed in EPM-3-ECA-1.1.
- c. Incorrect, not addressed in EPM-3-ECA-1.1.
D. Correct, the RHR sump is not available, therefore the crew must conserve water for injection from the RWST.
K/A: 000069SG1 1 [3.8 / 4.2) 025 AK2.05 [2.6/2.6] {41.7}
Reference:
EPM-3-ECA-1.1 step 12 LP/Objective:OPL271 C417, B.2 History: Procedure Bank Level: C ensicýný "'
Comments: FHW 12/02 025 AK2.05 AK30gT
SON EOI BASIS DOCUMENT FOR ECA-1.1 EPM-3-ECA-1.1 PROGRAM LOSS OF RHR SUMP RECIRCULATION Rev. 2 Page 20 of 74 MANUAL EOP Sten Number: 12 MONITOR RWST level greater than <U02>%.
ERG Step Number: 8 Check RWST Level - GREATER THAN (1)
Purpose:
To determine course of action based upon amount of water available in RWST.
ERG Basis:
Ifthe RWST is not empty, the operator proceeds with Steps 9 through 26, which are concerned with minimizing the RWST outflow and, therefore, extending the time that fluid for core cooling is provided by the RWST. This is accomplished by stopping the containment spray pumps and decreasing the SI pump flowrate. However, if the RWST is empty, the operator is instructed to skip to Step 27. In Step 27, the operator stops all pumps taking suction from the empty RWST.
EOP Basis:
Same.
Deviation:
The EOP step is changed to a continuous action step.
Justification:
In accordance with DW-92-056.
Setooint:
Identifier: <U02>
Description:
RWST Lo-Lo alarm (empty).
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 87
- 84. FR-Z.2-B.2 001 Which ONE of the following is addressed in FR-Z.2, "Containment Flooding", as the potential siurce of excessively high containment sump levels?
A. Condensed steam from a steam break B. RWST C. Accumulators
,/D. CCS A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
Student must comprehend the RWST and Accumulators are bounded [limited] sources and the CCS has potential to continue flowing. Also the engineering evaluation for the sump level transmitter setpoint [ not high level ] includes the volume of the RWST and the accumulators.
K/A{CFR}: E15 EK1.1 [2.8/3.0] [41.8 41.10]
Reference:
FR-Z.2 LP/Objective: OPL271C409, b.2 History: Old Bank Number PL-0014 Level: Memory Comments: FHW 12/02 E15 EK1.1J
ISTEPI ACTION/EXPECTED RESPONSEI RESPONSE NOT OBTAINED I
- 1. ATTEMPT to identify and isolate unexpected source of water to containment sump:
- High pressure fire protection
- Primary water
- Demineralized water
- SFP cooling water.
- 2. CHECK containment sump activity level:
GO TO Step 3.
r
- b. NOTIFY chem lab to sample containment sump for activity.
Page 3 of 4
Wednesday, October 09, 2002 @ l11:46 AM HLC12-02.BNK Page: 92
- 88. INCORE-B.1 .B 002 The plant Exosensor System calculates RCS subcooling by using which ONE/() of the following instrument inputs?
A. Pressurizer pressure and RCS WR Thot.
B. Pressurizer pressure and core exit thermocouples.
C. RCS WR pressure and RCS WR Tcold VD. RCS WR pressure and core exit thermocouples.
A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
K/A {CFR}: 016 K1.01 [3.4/3.4] [41.2-9]
References:
GOI-6 section P LP/Objectives: OPL271 RCSTEMP, B.9 History: System bank Level: Memory Comments: FHW 12/02 016 K1.01 xosensor also uses non nuclear inputs for RCS subcooling calculation, i.e. RCS WR pressure.
GOI-6 SON APPARATUS OPERATIONS Rev: 87 Page 50 of 74 SECTION P Page 1 of 5 INCORE THERMOCOUPLE TEMPERATURE AND SUBCOOLING MARGIN MONITORING SYSTEM (EXOSENSOR TEMPERATURE MONITOR)
1.0 INTRODUCTION
The Incore Thermocouple Temperature and Subcooling Margin Monitoring System provides fuel assembly outlet temperatures at selected core locations and calculates saturation margin. The system consist of two (2) 1E channels which are safety-related, designed, and qualified for post-accident monitoring (PAM). The principle function is to provide information to the operator during an accident situation for assessing plant post-accident status and verify core-cooling and subcooling conditions, as well as detection of potential fuel-clad breech following an accident.
The system is used for acquiring data only, and performs no operational plant control.
The operator has the capability to select thermocouples that are to be monitored by the Incore Thermocouple Temperature and Subcooling Margin Monitoring System.
The Incore Thermocouple Temperature and Subcooling Margin Monitoring System uses the hottest incore thermocouple, the hottest wide-range hot leg RTD, average of
-> all thermocouples, and RCS wide-range pressure for calculations.
2.0 SYSTEM DESCRIPTION A. Microprocessor Inputs
- 1. Channel 1 Microprocessor receives analog voltage signals from 34 thermocouples, 3 cold-junction reference RTD's, 2 wide-range hot-leg RTD's, and 1 wide-range pressure transmitter. Power supply is from the Vital Inst. Pwr. Bd. 1-1 (Unit 1) and 2-1 (Unit 2).
- 2. Channel 2 Microprocessor receives analog voltage signals from 31 thermocouples, 3 cold-junction reference RTD's, 2 wide-range hot-leg RTD's, and 1 wide-range pressure transmitter. Power supply is from the Vital Inst. Pwr. Bd. 1-11 (Unit 1) and 2-11 (Unit 2).
B. Key Pad layout:
TC TC REF DELETED SELECT ALARM TEMPS LIMIT RTD RTD/TC MARGIN TC ESC 1 2 3 4 5 RCS QUAD 1 QUAD2 QUAD3 QUAD 4 NEXT HELP 6 7 8 9 0 ENTER
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 93
- 89. INCORE-B.1 .D 003
~ The post-accident monitoring instrumentation for core-exit temperature and saturation margin is located on M4 Control Panel. One of the Main Menu displays on the Channel-I plasma display monitor has an asterisk (*)notation.
Which ONE .of the following describes the current alarm at4te b04t004-S
.mou4e9r, which is associated with the asterisk?
or
,/A. "Any (core-exit) Thermocouple-Above Physical Limit." A thermocouple is above
,elowits limit setpoint. %k
."liideyut Satur>C~t T(a- ctMay'"k~'lbcCdgv~
from C. "Monitor Mode is in Effect." The channel has been taken off line for calibration the Auxiliary Instrument Room panel.
has been D. "Diagnostic Channel Error." All capability for calculating Saturation Margin disabled.
A. Correct per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A: 017 K6.01 [2.7-3.0] [41.7]
Reference:
Incore Temperature Monitoring System Description section 2 "Plasma Display Unit." GOI-6 section P LP/Objective: Incore Temperature Monitoring System Description Objective 3.
History: Old Bank Number PL-1 126 Level: Memory Comments: FHW 12/02 017 K6.01
\tb
SECTION P Page 2 of 5 C. Key description:
KEY DISPLAY TITLE DISPLAY DESCRIPTION LABEL Thermocouple (TC) Displays average, hottest, and current user TC Temps selection thermocouple and trend parameters.
Temperatures TC Limits TC Limits Gives upper and lower limits for each TC'number.
RTD Temperature Summarizes RTD reference-junction temperatures Ref RTD and physical limits (low and high).
and Physical Limits I Deleted Deleted TC and RTD Shows by Tag number TC's and RTD's that have RTD/TC Sensors by Tag No. been deleted.
Summarizes and displays all six calculations for Margin Margin to Saturation margin to saturation.
Select Crnt/Quad Lists a menu of sub displays for current TC, TC Select TC Quadrants, and Recorders.
TC, Trend Recorders Allows user to quit what they are doing and go ESC Escape back to the menu/display.
Displays current alarm status plus all accumulated Alarm Alarm Status I alarms.
RCS Temperatures Gives RCS Temperatures and Pressures.
RCS and Pressures Quad #1 Quad #1 Lists Temperature of TC's in Quadrant #1 fli,r4 4*0 0usad Quad #2#2 Lists Temperature of TC's in Quadrant #2 Q"nrl #2 9 No Display Not Connected Quad #3 Quad #3 Lists Temperature of TO's in Quadrant #3 Quad #4 Quad #4 Lists Temperature of TC's in Quadrant #4 Next Next Allows user to inspect/modify or step through the next item in a long list.
HelpHelp Tells operator what action to take. Can only exit by Enter depressing the ESC key.
D. Plasma Display Unit/Indicator (PDU)
The PDU provides an in'dication of monitored plant parameters status.
- 1. Display Notations:
Denotes TC outside of physical limits Denotes TC outside of electrical limits XXX Denotes a deleted TC
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 95
- 91. INPO2 976 Given the following conditions:
- A Reactor trip and Safety Injection have occurred due to a SGTR.
- E-3, "Steam Generator Tubg Rupture," is in effect wjth a cooldown started at maximum rate. 4o -Avc--*-- eA wpe .
- The highest S/G pressure is 900 psig.
- RCS pressure is 1000 psig and dropping.
- High Head Safety Injection flow is approximately 300 gpm.
Which of the following describes what should be done with the Reactor Coolant Pumps?
A. RCP'&s lde tripped because RCP trip criteria is currently met.
v/B. RCP's s6l NOT be tripped because RCP trip criteria does NOT apply once an operator initiated RCS cooldown is commenced.
C. RCP's of2ý16e tripped because RCP trip criteria applies after an operator initiated RCS depressurization is commenced.
,, u "lf" D. ROP's NOT be tripped because RCP trip criteria does NOT apply until the operator-initiated RCS cooldown is completed.
A. Incorrect per procedure.
B. Correct per procedure.
C. Incorrect per procedure.
D. Incorrect per procedure.
KIA [CFR]: 038 EKi.08 [4.1/4.2] {41.5 41.10)
References:
E-3 LP/Objective: OPL271C391 B.3 History: INPO2 976 Level: Comprohn'sion Comments: FHW 12/02 038 EK3.08
SQN STEAM GENERATOR TUBE RUPTURE E-3 Rev. 13 STEP IACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE # Blocking low steamline pressure SI as soon as pressurizer pressure is less than 1920 psig will prevent an inadvertent MSIV closure and keep the condenser available for steam dump.
"* After the low steamline pressure SI signal is blocked, main steamline isolation will occur if the high steam pressure rate setpoint is exceeded.
"* The 1250 psig RCP trip criterion is NOT applicable after RCS cooldown is initiated in the following step.
- 9. INITIATE RCS cooldown:
- a. DETERMINE target core exit T/C temperature based on Ruptured S/G pressure:
Lowest Ruptured Target Core Exit S/G pressure (psig) TIC Temp (0 F) 1100 515 1000 500 900 490 800 475 700 460 600 445 500 425 400 400 380 390 (Step continued on next page.)
11 of 36
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 96
- 92. INPO3 955 Given the following plawconditions:
- A small breakiOCA has occurred.
- SI pumps fail to start.
- RCS Hot Legs and the Reactor Vessel Head have voided.
- RCS Pressure is 775 psig.
- RCP's are tripped in accordance with the EOP network.
- Assume that all other ECCS equipment operates as required.
Which ONEX"of the following describes the current method of cooling the core?
A. Break flow is the only core cooling method available.
B. Natural Circulation is the principle means of core cooling.
C. No core cooling mechanism exists at the present time.
,/D. Break flow and refl1iiow are providing core cooling.
A. Incorrect, the initial conditions due not exclude a S/G heat sink.
B. Incorrect, voiding in the hot legs and vessel head does not allow natural circulation.
C. Incorrect, a heat sink and break flow cooling still exists.
D. Correct, initial conditions woujaallow break and reflux flow.
K/A [CFR]: 009 E41C01 [4.2/4.7] {41.8 41.10}
References:
EPM-3-ES-0.3 "Check RVLIS upper plenum range."
LP/Objectives: OPL271 C383 b.4 History: INPO3 955 Level: Compr ension Comments: FHW 12/02 009 EK1.01
I EPM-3-ES-0.3 SQN EOI PROGRAM MANUAL BASIS DOCUMENT FOR ES-0.3 NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL (WITH RVLIS)
Rev. 0 Page 19 of 35 I
EOP Step Number: 7 CHECK RVLIS upper plenum range greater than <102>%.
ERG Step Number: 5 Check RVLIS Upper Range Indication - GREATER THAN (5).
Purpose:
natural circulation.
To ensure that the void does not enter the hot legs and disrupt ERG Basis:
subcooled hot leg fluid.
If steam enters the hot legs, it would most likely be condensed by the legs and reach the top of the S/G However, there may be the potential for some steam to enter the hot By monitoring RVLIS and limiting U-tubes, thereby disrupting the natural circulation flow circuit.
if necessary), the potential for the void growth to the top of the hot legs (repressurizing the RCS is applied to the reading to introducing voids into the S/G U-tubes is minimized. An uncertainty assuming worst case channel ensure that the void will actually be above the hot legs, even when inaccuracies.
LOP Basis:
Same.
Deviation:
None.
Justification:
N/A Setnoi~l Identifier: <102>
accuracy.
Description:
RVLIS value corresponding to the top of the hot legs - normal K.__
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 97
- 93. INPO8 155 When synchronizing the Main Generator, procedures direct the operator to adjust the speed of the generator until the synchroscope is rotating slowly in the fast direction.
The two parameters indicated by the synchroscope are:
A. Current and voltage differences.
,/B. Frequency and phase differences.
C. Frequency and current differences.
D. Current and phase differences.
A. Incorrect per reference.
B. Correct per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A[CFR]: 062 A4.03 [2.8/2.9] [41.7]
References:
OPL271 MTGC LP/Objective: OPL271 MTGC B.9 History: INPO8 155 Level: M ory Comments: FHW 12/02 062 A4.03; modified inpo question for SQN MTGC.
MT/G Cont Revision 0 Page 19 of 31 are nearly equal and the electrical speeds (60 hertz) are nearly equal. The only difference would be generator frequency slightly higher than system frequency in order to have a sync. scope turning slow in the fast direction.
- When a generating unit is to be paralleled to a running power system, its voltage, speed and synchronous phase position must be balanced to those of the system. If speed or phase position is not matched, excessive current flow could occur as the generator is tied to the system. Generator damage could result from the high current flow through the windings.
- To parallel a generator to an operating system, four conditions must be satisfied:
- Phase rotations must be the same (fixed phase connections).
- Speeds must be the same, to ensure frequency of the alternating cycles are the same.
- Voltage levels must be the same, to prevent reactive power flow when the generator breakers are closed.
- Voltage phase positions must be the same (in-phase) to prevent heavy current flow through the generator windings during acceleration or deceleration of the generator speed that would be required to match system voltage.
- Speed and voltage of the incoming generator can be varied until the system frequency and voltage are matched.
The revolving speed of the synchroscope is proportional to the difference in speed between the generator and in the the system. When the two speeds are matched, the sync' needle will be stationary or moving very slowly fast direction.
- The synchroscope dial indicates the angular phase difference between the running and incoming voltages.
be When the needle points to zero phase difference and is stationary, the two voltages are in phase and can paralleled.
to 12:00.
NOTE: 12:00 position is zero phase difference. Startup procedure parallels at 5 minutes prior
- If the incoming generator is tied-on in phase but at less than synchronous speed, it will be motored by the system to accelerate it to synchronous speed. Generator winding damage could occur if excessive current flow is required to develop the motor torque required for acceleration.
- If the incoming generator is in-phase but at a speed higher than system frequency, the generator will supply a high current flow to the system when the tie breaker is closed to decelerate the generator to synchronous speed.
Generator damage could occur if excessive current flow is required to retard the speed.
- If frequency is at synchronous speed but the incoming and running voltages are OUT OF PHASE, heavy current flow will exist while the generator is accelerated or decelerated to bring its voltage to synchronous phase position. Once again, generator damage could be the consequence of the improper operation.
- When paralleling a generator to a system, frequency and phase position of the two voltages must be matched.
Proper synchronizing of the two voltages will ensure that little or no transfer of energy (true or reactive power) occurs between the incoming and running equipment when the tie breaker is closed.
- Generator Capability
- Power consumed (useful work) and dissipated (wasted work) in a pure resistive circuit depends on the (12 R losses) is the "generatedvoltage and current and the resistance of the circuit. Typical dissipated power heat loss from a conductor while current is flowing. Power in a resistive circuit is always positive, outgoing power which is completely consumed and dissipated in the circuit. The total power generated is the algebraic
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 99
- 95. ODCM-B.5 001 This question has reference material attached.
A lightning strike has disabled 0-LS-27-225, Cooling Tower Blowdown Effluent Line and causing both unit's blowdown to isolate. Repairs to the level switch will take about 45 days from today.
Unit 1 needs to restore blowdown to cooling tower blowdown today. Which ONE of the following lists the actions required to COMPLETELY address this situation?
C A. Have the lineup for blowdown to cooling tower blowdown double verified and have two samples from each steam generator analyzed independently. Jumper the circuitry so that the blowdown release may be initiated. Initiate blowdown to cooling tower blowdown.
B. Jumper the circuitry so that the blowdown release may be initiated. Initiate Sblowdown to cooling tower blowdown. If RCS dose equivalent 1-131 is less than 0.01 microcuries per gram, sample the steam generators every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Notify the Chemistry Supervisor when blowdown has been in service for 30 days without 0-LS-27-225 for inclusion in the Annual Radioactive Effluent Release Report.
C. Estimate the cooling tower blowdown flow locally. Jumper the circuitry so that the blowdown release may be initiated. Initiate blowdown to cooling tower blowdown.
Estimate cooling tower blowdown flow locally every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during blowdown. No further notifications are required.
v/D. Estimate the cooling tower blowdown flow locally. Jumper the circuitry so that the blowdown release may be initiated. Initiate blowdown to cooling tower blowdown.
Estimate cooling tower blowdown flow locally every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during blowdown. Notify the Chemistry Supervisor when blowdown has been in service for 30 days without 0-LS-27-225 for inclusion in the Annual Radioactive Effluent Release Report.
A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
Z, . -A /l L e-
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 100
- 95. ODCM-B3.5 001 K/A[CFR]: 2.3.11 [2.7/3.2] [41.11]
References:
ODCM Control 1.1.1 action b, Table 1.1-1 action 33 LP/Oblective: OPL271LRW B.3 History: Procedure bank Level: Comprehsion /41 Comments: FHW 12/02 2.3.11 Provide copy of ODCM R46 page 13,14,15
SQN ODCM Revision 46 Page 13 of 168 1/2 CONTROLS AND SURVEILLANCE REQUIREMENTS "1/2.1 INSTRUMENTATION 1/2.1.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION CONTROLS 1.1.1 In accordance with SQN Technical Specification 6.8.4.f.1, the radioactive liquid effluent monitoring instrumentation channels shown in Table 1.1-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of ODCM Control 1.2.1.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the methodology and parameters in ODCM Section 6.2.
APPLICABILITY: This requirement is applicable during all releases via these pathways.
ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required above, without delay suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable, or change the setpoint so that it is acceptably conservative.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the action shown in Table 1.1-1. Exert best effort to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability could not be corrected within 30 days.
- c. The provisions of Controls 1.0.3 and 1.0.4 are not applicable. Report all deviations in the Annual Radioactive Effluent Release Report.
SURVEILLANCE REQUIREMENTS 2.1.1 Each radioactive liquid effluent monitoring channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE/SENSOR CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 2.1-1.
SQN ODCM Revision 46 Page 14 of 168 Table 1.1-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION (Page 1 of 2)
Minimum Instrument Channels Action OPERABLE
- 1. Gross Radioactivity Monitors Providing Automatic Termination of Release
- a. Liquid Radwaste Effluent Line (0-RM-90-122) 1 30
- b. Steam Generator Blowdown Effluent Line 1 31 (1,2-RM-90-120A,121A)
- c. Condensate Demineralizer Effluent Line 1 30 (0-RM-90-225A)
- 2. Gross Radioactivity Monitors
- a. Essential Raw Cooling Water Effluent Header** 1 32 (0-RM-90-133A,-134A,-140A,-141A)
- b. Turbine Building Sump Effluent Line 1 31 (0-RM-90-212A)
- 3. Flow Rate Measurement Devices
- a. Liquid Radwaste Effluent Line 1 33 (0-FI-77-42)
- b. Steam Generator Blowdown Effluent Line 1 33 (1,2-F1-1 5-44, 1,2-F-15-43)
- c. Condensate Demineralizer Effluent Line 1 33
- (0-_FR-14-456, 0-F-14-185, 0-F-14-192)
- d. ooling Tower Blowdown Effluent Line 1 33 0P(-LS-27-225)
- 4. Tank Level Indicating Devices
- a. Condensate Storage Tank (0-L-2-230, 0-L-2-233) 34
- 5. Continuous Composite Sampler and Sample Flow Monitor
- a. Condensate Demineralizer Regenerant Effluent Line 1 35 (0-FI-14-466)
"**Requires minimum of 1 Channel/Header to be OPERABLE.
SQN ODCM Revision 46 Page 15 of 168 Table 1.1-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION (Page 2 of 2) TABLE NOTATION ACTION 30 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:
- a. At least two independent samples are analyzed in accordance with ODCM Control 1.2.1.1, and
- b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for principal gamma emitters in accordance with SR 2.2.1.1.1 and 2.2.1.1.2.
- a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than or equal to 0.01 gCi/g DOSE EQUIVALENT 1-131.
- b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than 0.01 LCi/g DOSE EQUIVALENT 1-131.
ACTION 32 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for principal gamma emitters at a limit of detection of at least ACTI? 33 -
-*.xl 0-7 xCi/ml.
ith the number of channels OPERABLE less than required by the Minimum CChannels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curves may be used to estimate flow.
ACTION 34 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continued provided the tank liquid level is estimated during all liquid additions to the tank.
ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided representative batch samples of each tank to be released are taken prior to release and composited for analysis according to Table 2.2-1, footnote g.
Thursday, October 10, 2002 @ 09:42 AM HLC12-02.BNK Page: 1
- 1. OPL271C353.4 091 t 2
,/ , C WHICH ONE."of the following will cause actual pressurizer pressure to increase?
A. Pressurizer Pressure channel fails high.
B. Steam line leak.
vC. ;, ib ease in RO tmporatur& 72A-ir -t% av- ,,..4,.--s_
A. incorrect a pressure channel failing high will not inreas pressure.
B. incorrect a steam line break will reduce Tave and pressure.
C. correct an increase in RCS temperature will cause expansion of liquid and pressure increase.
D. incorrect a boration will reduce Tave.
K/A[CFR]: 027AK1.02 [2.8/3.1] (41.8 41.10}
,rd ,t c ,,44'../66 . Ž1-f
Reference:
Thermo Basic Energy Concepts-SH 1-Pr0Srdoc LP/Objectives: 0PL271C353 B.4 2 History: Procedure Bank Level: Comprehension Comments: FHW 12/02 027AK1.02 S/¢ 1VX O,5C'-.m.n,-, al.Lb(
temperatures, the specific heat of liquid water is in some applications. This will be covered in somewhat greater than 1.0 Btu/lbr 0F. more detail in the next chapter.
"* Specific heat is used to calculate the addition of The value of specific heat for materials is heat to a system from observed temperature tabulated in various engineering handbooks. An changes. If the specific heat is known, the heat average value of specific heat will be given over addition (or subtraction) may be calculated by an applicable range of temperature and pressure.
rearranging Equation 2-15: It is up to the individual user to ensure that the specific heat value selected is applicable to the q = cPAT conditions of interest.
Where: The most widely used specific heat is that of water at one atmosphere of pressure. The q = heat transferred per unit mass specific heat of water is used to define the British (Btu/Ilbm) thermal unit (Btu). The Btu is the amount of heat energy that must be added to raise the Cp = specific heat capacity temperature of one pound mass of water one (Btu/lbm OF) degree Fahrenheit, at one atmosphere of pressure. Strictly speaking, the Btu is 1/180th of AT = temperature change (OF) the heat needed to raise water temperature from 320F to 2120F. It is important to note that the Equation 2-17 value of a Btu is an average taken from 32°F to 212WF. Above 212°F, the specific heat of water This equation assumes that c. is an average increases considerably, due to the large amount specific heat over the range of temperature of heat required to change a liquid to a gas.
change.
Most steam tables do not list values of properties The specific heat of a solid does not change for subcooled liquids. Using values for specific appreciably with pressure or expansion, but does heat capacity, subcooled liquid property values change slightly with temperature. The specific can be calculated.
heat of a liquid does not change appreciably with expansion, but does change with pressure and temperature. Generally, an increase in pressure reduces in the specific heat of a liquid.
Conversely, an increase in liquid temperature raises the specific heat of a liquid. Therefore, it is important not to use a single, average value of specific heat over an excessively wide range of pressure and temperature.
Direct use of specific heat should be avoided in vapor and gas calculations. The values of specific heat for substances in this phase vary widely with conditions. However, specific heat for gases undergoing certain limiting processes can be used PWR / THERMODYNAMICS / 19 of 24 ©1999 GENERAL PHYSICS CORPORATION CHAPTER 2
/ BASIC ENERGY CONCEPTS REV 2
SON PRESSURIZER INSTRUMENT MALFUNCTION AOP-I.04 Rev. 5 I
I1 Page 1 of 1 APPENDIX H I PRESSURIZER PRESSURE CONTROL C-)
C,,
U,.#
Page 56 of 57
Wednesday,v October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 94 Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 94
- 90. INPO2 231 Following a large steam line rupture, monitoring of Critical Safety Function Status Trees indicates a RED path for FR-P.1, "Pressurized Thermal Shock." Which one of the statements below correctly identifies the major component and reason for concern that Pressurized Thermal Shock conditions may result in brittle failure?
A. The RCS piping due to ,, ....d t .. r.. f*
.rou..t..n. ...... u f a*'urd' erprps..z o\.*-- .... . ,al. lo t ...... ur . Uce.. <0
- g...." --" . "OA* ,4
ýý/V* qe C. Tb sta eeaos~t nr~o too e"tn mnrpd J
,/D. The reactor vessel due to increased stresses resulting from a cooldown of unexpeeted-6verity or an overpressure condition at low temperature.
- a. incorrect per EPM-3-FR-P.1
- b. incorrect per EPM-3-FR-P.1
- c. incorrect per EPM-3-FR-P.1
- d. correct per EPM-3-FR-P.1 "The thermal stress due to a rapid cooldown and the pressure stress are additive in the vessel wall."
K/A {CFR}: 040 AK1.01 [4.1/4.4] [41.8 41.10]
References:
EPM-3-FR-P.1 LP/Objectives: OPL271C406 b.3 History: INPO2 231 Bank Level: Comprension Comments: FHW 12/02 040 AK1.01
SON EQI BASIS DOCUMENT FOR FR-P.1 EPM-3-FR-P.1 PROGRAM PRESSURIZED THERMAL SHOCK Rev. 2 MANUAL Page 4 of 61 INTRODUCTION This procedure provides actions to respond to reactor vessel thermal shock, pressurized thermal shock, and low temperature overpressure conditions.
The major action categories performed in this procedure are:
Stop RCS cooldown.
Terminate ECCS if criteria satisfied.
Depressurize RCS to minimize pressure stress.
Establish normal operating conditions and stable RCS conditions.
Soak if necessary prior to further restricted cooldown.
A detailed discussion of the high level action steps, including any applicable notes or cautions, is presented on the following pages.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 110 103. OPL271C382.4 001 Which ONE of the following indicates natural circulatiorq0. =
,/A. Core exit thermocouples decreasing; T-HOT decreasing; RCS subcooling increasing.
B. Core exit thermocouples decreasing; T-HOT decreasing; RCS subcooling decreasing.
C. T-HOT increasing slowly; T-COLD decreasing; RCS subcooling increasing.
D. Core exit thermocouples increasing slowly; SG pressure decreasing; RCS subcooling decreasing.
A. Correct per ES-0.2.
B. Incorrect due to subcooling decreasing.
C. Incorrect due to T-COLD decreasing.
D. Incorrect due to RCS subcooling decreasing.
K/A [CFR]: E09 EAI.1 [3.5/3.5] {41.7},1 4
4 c-e_ 4 'f12 9P ,
Reference:
EA-68-6 f
/ 1ý-1-4'-e7ý:2 C/-a LP/Objectives: OPL271C382 b.4 History: Old Bank Number PL-0904 Level: Analysis Comments: FHW 12/02 E09 EA1.1
4.2 Verification of Natural Circulation
- 1. MONITOR the following indications of natural circulation at 15- to 20-minute intervals:
"* RCS subcooling greater than 40'F. El
"* S/G press stable or dropping. El
"* T-hot stable or dropping. El
"* Core exit T/C stable or dropping. El
- T-cold at saturation temperature for S/G pressure. El
- 2. DETERMINE parameter trends between monitoring intervals and EVALUATE if natural circulation is occurring. El
- 3. IF natural circulation NOT verified, THEN NOTIFY ASOS. El
- 4. GO TO Section 4.1, step in effect. H END OF TEXT
Wednesday, October 09, 2002 @ l11:46 AM HLC12-02.BNK Page: 76
- 74. ES-0.2-B.3 003
<- Given the following plant conditions:
- All reactor coolant pumps are deenergized.
The core exit thermocouples are reading 440 F and decreasing.
Upper head thermocouples are reading 605 F.
0; The RCS cold leg temperatures are 400 F and decreasing.
i? t"* The RCS pressure is 1585 psig and slowly decreasing.
The pressurizer level has rapidly increased from 35% to 48%.
Which ONE P4of the following is the reason for the change in pressurizer level?
A. RCS pressure has decreased to the point where the safety injection pumps have begun injecting into the RCS.
B. Anticipated response due to uneven loop cooling and flows during natural circulation.
VC. RCS depressurization has caused saturated conditions and a steam void in the reactor vessel head area.
D. Pressurizer level swings are expected due to steam entering the pressurizer and surge line.
A. Incorrect SI pumps start injecting at 1500 psig.
B. Incorrect by design there should be no uneven loop flow.
C. Correct per steam tables. Sat temp for 1600 psia is 604.870 F.
D. Incorrect, the RCS temperature is subcooled.
K/A{CFR}: ElO EK3.1 [3.4/3.7] {41.5 41.10) 7-A! 6e 0.2 '< 42CC4 kt /k A
References:
ES-0.2 LP/Objectives: OPL271C382, b.3 History: Procedure Bank Level: Comprehension Comments: FHW 12/02 E10 EK3.1 Provide steam tables
/'4 'ýV_ ZeZ41 J-eý-Z ,
SON NATURAL CIRCULATION COOLDOWN ES-0.2 Rev. 12 STEP IACTION/EXPECTED RESPONSE FRESPONSE NOT OBTAINED NOTE "Unexpected pressurizer level changes indicative of vessel voiding" is defined as an unexplained level rise when reducing RCS pressure OR an unexplained level drop when raising RCS pressure.
- 16. MONITOR for steam voids in reactor vessel:
- a. CHECK for the following indications of a. GO TO Step 17.
steam voids:
"* unexpected pressurizer level changes indicative of voiding OR
"* RVLIS upper plenum range less than 98%.
- b. REPRESSURIZE RCS within limits of Tech Spec temperature/pressure limit curves to collapse voids in system and CONTINUE cooldown.
IF repressurization is NOT desirable OR NOT possible, THEN PERFORM one of the following:
"* GO TO ES-0.3, Natural Circulation Cooldown With Steam Void in Vessel (With RVLIS).
" GO TO ES-0.4, Natural Circulation Cooldown With Steam Void in Vessel (Without RVLIS).
11 of 19
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 121 113. RCP-B.12 002 Which ONE ,,Itof the following is the reason that the RCP's have an anti-reverse rotation devicp installed on the pump rotor?
, Preventereveese flow through a tripped RCP when other RCP's are running.
B. Prevent overheating of pump bearings due to a tripped RCP rotating without normal internal cooling flow.
C. Prevent damage to the pump thrust bearing due to operation in the reverse direction.
VD. Prevent stator winding damage due to excessive pump starting current.
A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
K/A [CFR] 003 A1.03 [2.6/2.61 E1.5] (41 0 4
Reference:
RCPsystem description section 2.0 LP/Objective: RCP System Description Objective 12.
History: System bank Level: Memory Comments: FHW 12/02 003 A1.03 9'( /
- 2 4V
Reactor Coolant Pump RCP Flywheel Purpose The RCP flywheel has the following purposes:
"*Ensures continuation of forced flow for a short time following a reactor trip and loss of power to the RCPs. The flywheel increases the rotational inertia of the motor so that the pump coast down time becomes longer.
Prevents a deener rotatin bwards when another pump is running and causin9 reverse fow thi-ouqUg ina*ctýive lo. The yrh" s
-equipped with an anti-reve-rserotation device consisting of pawls mounted on the flywheel that engage the motor frame when the pump starts to turn backwards.
Starting the motor while rotating backwards would draw excessive starting currents which would overheat the motor.
Basic Block The RCP flywheel and anti-rotation device are depicted below.
Diagram Pawl Pawl Stop Flywheel RCP-14 Continuedon next page 068d.doc 2 -24 Rev 2
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 122 114. RCS-TEMP-B.2 008 Which ONE.J(1 of the following temperature elements is used to supply a temperature compensation signal to correct for the Reactor Coolant liquid-density and/or steam-density changes for the Reactor Vessel Level Indicating System (RVLIS) level transmitter outputs?
A. RCS wide-range cold-leg temperature elements.
",/B. RCS wide-range hot-leg temperature elements.
C. Hot-leg RVLIS instrument-line temperature elements.
D. Cold-leg RVLIS instrument-line temperature elements.
A. Incorrect per reference.
B. Correct per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A [CFR]: 016 A3.02 [2.9/2.9] [41.7]
Reference:
1-47W610-68-7 RCS temperature system description page 2-4.
LP/Objective: OPL271 RCSTEMP B.11 History: System bank, old Bank Number PL-1 148.
Level: Mery Comments: FHW 12/02 016 A3.02
Major Components Piping, Continued Instruments The following drawing and tables describe instruments associated with the RCS:
RCS-04 Narrow Ra eotL RTDs 3 RTDs, placed in Each loop hot leg TE-68-2A,B,D Provides input into: piping has three
- loop AT circuit thermowell, 1200 TE-68-25A,B,D thermowells after
- Tavg circuit apart to minimize TE-68-44A,B,D the piping has
- control and protection temperature TE-68-67A,B,D exited the cavity functions streaming effects wall and prior to the S/Gs I
toflf 'flue" JfirtA&
C*ontinu~edl on ne*xtpage 2-3 REV 0 068a.DOC
Reactor Coolant System Piping, Continued Instruments (continued)
Narrow Range Cold Le RTDs Each loop cold leg TE-68-14A,B Provides input into: 2 RTDs, placed in TE-68-37A,B
- loop AT circuit thermowell, 45* piping has two thermowells TE-68-56A,B
- Tavg circuit apart. downstream of RCP TB-68-79 AB - control and nrotection discharge and prior functions to the piping entering the cavity wall.
Continued on next page 2-4 REV 0 068a.DOC
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 126 S18. REP-B.1.D 002
<> Select the LOWEST emergency classification from the list below that requires activation of the Technical Support Center. Do NOT consider a security event.
A. Notification of Unusual Event
,/B. Alert C. Site Area Emergency D. General Emergency A. Incorrect per reference.
B. Correct per reference.
C. Incorrect per reference.
D. Incorrect per reference. 0ýý 2.4-/ [3.3-K2.4[MF[2.62-44.0[2.
K/AECFR]: 3.1] 443.2]
[45.11]
Reference:
EPIP-6 R34 section 3.2.3.
LP/Objective: 0PL271C198, B.1.
History: Procedure bank, old Bank Number PL-0308 Level: Memory Comments: FHW 12/02 2.4.39
r I CI r - -I ACTIVATION AND UPIERA I OUNI Or IfHE Rev. 344 SON TECHNICAL SUPPORT CENTER Page 6Qof 51 3.2 Activation of the TSC 3.2.1 Shift Manager (SM)
The SM will activate the TSC and OSC by announcing the emergency condition by one or more of the following methods.
A. Plant Public Address (PA) announcement.
B. The SM or Operations Clerk will normally activate the Emergency Paging System (EPS) or, contact the persons designated on the call list. If the EPS cannot be activated from the site, contact the Operations Duty Specialist (ODS) on the ringdown line or 5-751-1700 and have the EPS activated from the CECC.
C. The SM may activate the onsite emergency sirens at an "Alert" and shall activate the sirens at a "Site Area Emergency" or 'General Emergency."
3.2.2 Call List The Emergency Preparedness Manager (EPM) shall maintain a call list listing all TSC personnel by name, plant and home telephone numbers. The REP by the EPM or designee with input Call List will be updated at least quarterly provided to the by the appropriate section/group supervisors. The list will be SM and placed in the TSC.
<,> 3.2.3 Response should report to the TSC, or Personnel performing the following REP functions Appendix B Figure B-3 for the assigned TSC support locations (see NP-REP or higher emergency TSC Layout), upon announcement of an "ALERT" classification or at the direction of the SED.
A. Site Vice President B. Site Emergency Director C. Operations Manager D. Technical Assessment Manager E. Operations Advisor, TAT F. Site Security Manager G. Radiological Control Manager (RCM)
H. Chemistry Manager
- 1. NRC Coordinator J. Control Room Communicator (affected Unit Control Room)
K. EP Manager be called)
L. TSC Clerical/Logkeeper Staff (Clerical will M. Maintenance Manager N. Technical Assessment Team
- 0. Operations Communicator P. Other Plant staff the SED determines to be necessary to support TSC functions will be called.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 129 121. RMS-B.2 004 annunciators SWhich ONkof the following Common Radiation Monitor 0-XA-55-12B in alarm indicates an area high radiation condition?
A. (A-5) o-RA-90-132A SERVICE BLDG VENT MON HIGH RAD B. (B-i) 0-RA-90-101A AUX BLDG VENT MONITOR HI RAD
,/C. (B-3) 0-RA-90-102A FUEL POOL RAD MONITOR HIGH RAD D. (C-7) O-RA-90-125A MAIN CNTRL RM INTAKE MON HIGH RAD A. Incorrect per reference.
B. Incorrect per reference.
C. Correct per reference.
D. Incorrect per reference.
K/A: 072 f,-"Cf3.4/3.6] [41.5]
Reference:
0-XA-55-12B (B-3)
LP/Objective:OPL271 CO13, B.2 Level: Memt'Wj History System bank (Developed 7/23/98)
Comments: FHW 12/02 072 A1.01
10 (B-3)
Seipoiht Source 0-RA-90-102A SER 760 50 mr/hr > I second FUEL POOL RAD 0-RE-90-1 02 MONITOR HIGH RAD Retransmitted to U-2 SER 2243 Probable I. High radiation in spent fuel pit area elevation 734.
Causes Corrective [1] VERIFY the following:
Actions a. Auxiliary Building General Supply and Exhaust and Fuel Handling exhaust isolate (A-Train) (I-M-9).
- b. Auxiliary Building Gas Treatment System starts (1-M-9).
E2] CHECK 0-RM-90-102 and 0-RM-90-103 on 0-M-12 to verify alarm.
[3] IF high radiation alarm valid, THEN
[a] ANNOUNCE "High Radiation at spent Fuel Pool Area" over PA system.
[b] NOTIFY SM.
[c] NOTIFY RADCON.
[4] IF B-Train ABI has not actuated from a valid High Radiation condition, THEN INITIATE manually B-Train Auxiliary Building Ventilation Isolation via [1-HS-30-1 B or 2-HS-30-101B1 (M-6).
[5] IF fuel handling in the Spent Fuel Pit is in progress, THEN REFER TO AOP-M.04, Refueling Malfunctions.
E6] REFER TO AOP-M.06, Loss of Spent Fuel Cooling.
[7] IF Auxiliary Building Ventilation Isolation resulted from an invalid ABI signal, THEN, REFER to 0-SO-30-10 Auxiliary Building Ventilation Systems to recover from ABI.
[8] EVALUATE Technical Specifications 3.3.3.1 and 3.9.12.
[9] INITIATE Corrective Actions.
[10J WHEN conditions return to normal, THEN RETURN Auxiliary Building Ventilation System to normal in accordance with 0-SO-30-10, Auxiliary Building Ventilation Systems.
References 45B655-12B-0, 47W610-90-1
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 130 122. RMS-B.9 001 A Waste Gas Decay Tank is being vented to atmosphere using the normal gaseous ONEP of the waste discharge path. If high gaseous activity exists in this tank, which following radiation monitors would generate an alarm?
A. Auxiliary Building Ventilation Monitor [RE-90-101].
B. Containment Purge Air Exhaust Monitor [RE-90-130].
C. Service Building Ventilation Monitor [RE-90-132].
,/D. Shield Building Ventilation Monitor [RE-90-400].
A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
K/A {CFR}: 071 A1.06 12.5/2.8] [41.5]
- 0 073 K1.01 (3.6 - 3.9) ,
073 A4.01 (3.9 - 3.9)9)
Reference:
1,2-47W611-77-4 0-SO-77-15 LP/Objective: OPL271GRW B.11 History: System bank Level: Memory Comments: FHW 12/02 071 A1.06
Date 6.0 NORMAL OPERATION (Continued)
NOTE Radiation Control Valve O-RCV-77-119 is interlocked with 1-FT-30-150 and 2-FT-30-165, and will terminate the release if the fan stops running.
[9] VERIFY the applicable ABGTS fan is running. UO
[10] IF [0-RE-90-1181 is operable, THEN ENSURE O-RM-90-118 is in service. UO
[11] IF [0-RE-90-118I is INOPERABLE, THEN ENSURE Compliance with actions of ODCM 1.1.2, AND WO initiated for corrective action. UO NOTE If the selected unit shield bldg. radiation monitor is inoperable, then the following step may be N/A.
[12] VERIFY radiation monitor RM-90-400 is in service and operable for the Shield Building vent indicated in step [3]. UO
[13] IF the shield bldg vent release path selected in step [3],
radiation monitor RM-90-400 is inoperable, THEN NOTIFY the U-I SRO to take appropriate actions in accordance with Section 1.1.2 of the ODCM, AND WO initiated for corrective action. UO
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 44
- 43. AOP-P.02-B.4 002 A loss of 125V DC Vital Battery Board I occurs during a surveillance test on 1A-A Unit Board diesel generator. The diesel generator has been paralleled to the 1B 6.9kV result in which and is sharing the load. The loss of 125V DC Vital Battery Board I will ONE Q(,of the following?
A. diesel trip due to underfrequency.
B. loss of control to the diesel generator output breaker ONLY.
with the ability to
,/C. loss of speed and voltage control from the control room along shutdown the diesel from the control room.
ability to start the D. no effect - 125V DC Vital Battery Board I power only affects the diesel.
A. Incorrect per reference.
B. Incorrect per reference.
C. Correct per reference.
D. Incorrect per reference.
K/A {CFR): 65'58 AK3.01 [3.4/3.7] 41.5, 41.10 4tt4 -- 063 K3.02 [3.5/3.7] 41.7
-&464058 AA2.03 [3.5/3.9] 43.5
References:
45N767-5 LP/Objectives: OPL271C364, b.4 History: Procedure Bank Level: Compronsion.eQ Comments: FHW 12/02 058 AK3.0l
HLC12-02.BNK Page: 131 Wednesday, October 09, 2002 @ l11:46 AM J23. RPS-B.5.B 001
-> The Reactor Protection System manual handswitches in the Control Room are operated in an attempt to trip the Reactor during an ATWS event.
Which ONEŽ,5 of the following describes how the Reactor Trip Breaker's trip attachments should respond to operation of a manual Reactor Trip handswitch?
A. The SSPS undervoltage trip attachments and the 125V-DC trip attachments must
- both deenergize to trip the Reactor.
,/B. The SSPS undervoltage trip coils must deenergize, while the 125V-DC trip coils g
must energize to trip the Reactor.
The SSPS undervoltage coils will d= e and the 125V-DC trip coils are meDr The 125V-DC trip coils will energize, while the SSPS undervoltage coils will be unaffected.
A. Incorrect, 125V-DC trip coil will energize.
B. Correct, per reference.
C. Incorrect, 125V-DC trip coil will energize.
D. Incorrect, the undervoltage coil will deenergize.
K/A[CFR]: 012 A3.07 (4.0 / 4.0) 1 0 1rt j$ 0/C, 029 EK2.06 [2.9/3.1] {41.7}5
Reference:
0-47W611-99-1 LP/Objective: OPL271 RPS, b.9 History: System Bank old Bank Number PL-0811 Level: Comprehension Comments: FHW 12/02 029 EK2.06
HLC12-02.BNK Page: 98 Wednesday, October 09, 2002 @ 11:46 AM
- 94. MFW 001 input signals from:
The Main Feedwater Regulating Bypass Valves receive level A. Associated SG's Motor Driven Auxiliary Feedwater Level transmitters.
,es--,S r"
,/B. Associated SG's Turbine Driven Auxiliary Feedwater Level transmitterst SG control circuit. ' ,,Z,'
C. Narrow range level signal from the median selector on each ,OatS gvWj~'#l- t SG.
D. The non-PAM narrow range level transmitter from the associated A. Incorrect per reference.
B. Correct per reference.
C. Incorrect per reference.
D. Incorrect per reference KA[CFR]: 059 K,.!O4 [3.4/3.4] [41.2-9]
References:
1,2-47W611-3-2 2
LP/Objective: Steam Generator System description section 2. Objective History: System bank Level: Memory Comments: FHW 12/02 059 K1.04
HLC12-02.BNK Page: 1 Thursday, October 10, 2002 @ 10:38 AM
- 1. FW-B.5 001 Given the followin plant conditions and information:
o;' o
- Uni aO000 rated thermal power
- Feedwater Master Controller and Feedwater Pump Speed Controllers are in AUTOMATIC.
- Main Feedwater Regulating Valves are in AUTOMATIC.
- All four main feedwater flows start increasing with level in all four steam generators trending upwards.
For information:
- PT-1-33 is a Main steam header pressure transmitter
- PT-3-1 is a Main feedwater header pressure transmitter failures that could have Which ONE (1) of the following describes the instrument caused this transient?
A. PT-1-33 has failed LOW and PT-3-1 has failed LOW.
LOW.
,/B. PT-1-33 has failed HIGH and PT-3-1 has failed HIGH.
C. PT-1-33 has failed HIGH and PT-3-1 has failed D. PT-1-33 has failed LOW and PT-3-1 has failed HIGH.
Thursday, October 10, 2002 @ 10:38 AM HLC12-02.BNK Page: 2
- 1. FW-B.5 001 A. PT-1-33 failing low would cause a high delta-P and the MFPT control system would reduce MFPT speed causing low feed flow and decreasing SG levels. The results of PT-3-1 failing low are described in "B" below.
B. Correct. Feedwater header pressure is normally higher than steam header pressure by a programmed value of 80 psid at 0% totalized steam flow and 195 psid at 100% totalized steam flow. PT- 1-33 failing high OR PT-3-1 failing low would indicate a lower than normal delta-P between steam header pressure and FW header pressure. This would cause the MFPT control system to increase speed in an attempt to restore programmed delta-P. This increased delta-P would increase FW flow and cause SG level to trend upward. The FW regulating valves would be closing in an attempt to reduce SG level back to programmed value.
C. The results of PT-1-33 failing high are described in "B" above. PT-3-1 failing high would cause a high delta-P and the MFPT control system would reduce MFPT speed causing low feed flow and decreasing SG levels.
D. PT-1-33 failing low would cause a high delta-P and the MFPT control system would reduce MFPT speed causing low feed flow and decreasing SG levels. PT-3-1 failing high would cause a high delta-P and the MFPT control system would reduce MFPT speed causing low feed flow and decreasing SG levels.
K/A{CFR}: 439 K3.04 [2.5/2.6] [41.7]
- 059 K1.04 [3.4/3.4] {41.2, 41.3, 41.4, 41.5, 41.6, 41.7, 41.8, 41.9)
/,-r059 K6.03 [1.9/2.1] {41.7}
ftO59 K6.09 [2.4/2.6] {41.7}
+ 059 A2.11 [3.0/3.3] (41.5, 43.5)
References:
47W611-3-2 OPL271 CCON D/FW LP/Objectives: OPL271C034 Obj. B.5 History: WBNOPS--~ .bnk Q# SYS003D. 14 001 (Modified stem)
Level: Ans~is Comments: LP-5/2000;. FW 004; FHW 12/02 039 K3.04
COND/FW Student Handout Revision 0 Page 10 of 22
<*_> Intermediate Pressure Feedwater Heaters
- Condensate from the intermediate pressure heater strings is then routed to the main feed pumps Main feed pumps
"* Pumps: -65% capacity each; 20,000 gpm at 1680' TD
"* Pump controls
- Handswitch located on M-3
- RPM, suction flow,
"* Speed control and the S/Gs.
The main feed pumps maintain a programmed AP across the #1 heaters Load index is totalized steam flow. The AP is held Program setpoints: 80 psid at no load, to 195 psid at full load.
100% load.
level @ 80 psid from 0-20% load, then is linear from 20% to
"* Controllers: discuss startup with one in auto using the master controller.
"* One for both (master controller)
- This controller is susceptible to Reset Windup.
SOER 94-001 describes reset windup in the following manner:
signal and the setpoint of an "Reset Windup" occurs when there is an error between the process (reset) causes the output to controller for an extended period of time. During this time Integration result, when the process As a continue to change until high or low output saturation is reached.
there may be a significant time changes such that the controller is driven back out of saturation, and begins to change in the right delay before the controller output actually comes out of saturation Windup" condition is observed, direction. This time delay can be several minutes. When a "Reset required adjustments and place the the operator should take manual control of the controller , make controller back in auto when the condition has cleared.
+ reverse);
Automatic Pump Trip: Excessive thrust - Turbine 5 mils (forward
- Pump 5 mils (forward + reverse);
Low bearing oil pressure 10 psig; Electrical Fault.;
Suction valve not fully open; Low injection water pressure; this);
Low condenser vacuum (feed inlet and outlet valves closed will cause Feedwater isolation signal; Overspeed (mechanical):
Manual.
are shown on NOTE: High Vibration: vibration trips on both the pump and turbine have been disconnected These the electrical prints and referenced to the drawing notes indicating such.
HLC12-02.BNK Page: 119 Wednesday, October 09,2002 @ 11:46 AM 111. RADWASTE-B.12 008 Co-t" -rc What TWO conditions willIQ eentl* cause automatic closure of Liquid Radwaste Release Valve, 0-RCV-77-43? 4 ( ,4, ", -' 4-*_ 4r 6# 7.)
in the release header.
,/A. Low cooling tower blowdown flow, and high radiation sensed in the cooling tower B. Low cooling tower blowdown flow, and high radiation sensed blowdown flow.
in the release header.
C. High release header flow, and high radiation sensed in the cooling tower blowdown D. High release header flow, and high radiation sensed flow.
A. Correct per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A {CFR}: 068 A3.&2*3.6/3.6] [41.7]
References:
1,2-47W611-77-2 LP/Objectives: OPL271LRW B.12 History: System bank Level: M 4ory Comments: Reviewed by T Jetton, FHW 12/02 068 A3.02
HLC12-02.BNK Page: 90 Wednesday, October 09, 2002 @ 11:46 AM
- 86. FW-B.5.B 004 FW Pump is provide pro ection devices that will trii param T ._ S rovides some inputs for that input to the MFW Pump trip Which ONE ()of the following describes the SSPS circuitry? x valve which dumps trip oil A. The SSPS provides a signal to a 120v AC trip solenoid thus tripping the pump.
plunger which dumps trip oil thus B. The SSPS provides a signal to the overspeed trip tripping the pump.
valve which dumps trip oil C. The SSPS provides a signal to a 48v DC trip solenoid thus tripping the pump.
trip solenoid valves which dump trip oil
,/D. The SSPS provides signals to two 1 v DC thus tripping the pump. -
- A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per referenAe. dt KA[CFR]: ,06,3 K4.04 [2.6/2.9] [41.7]
.- 59 K4.16 (3.1-3.2)
References:
1,2-47W611-99-4 LP/Objective: OPL271COND.FW B.12 History: System bank; 9/2/97 Makeup Audit Exam.
Level: M emW9 Comments: FHW 12/02 063 K4.04
<7
HLC12-02.BNK Page: 123 Wednesday, October 09,2002 @ 11:46 AM 115. RCW-B.9 001 Water Which ONE ( the following is the source of water for the Raw Cooling
'of System? !X44, ,,z %,,S A. Emergency Raw Cooling Water System B. Raw Service Water System
",/C.Condenser Circulating Water System D. High Pressure Fire Protection System A. Incorrect per reference.
B. Incorrect per reference.
C. Correct per reference.
D. Incorrect per reference. Fire protection is used for a backup to the CTLP bearing lube water.
K/A[CFRI: 075 01 [2.5/2.5] [41.2-9]
References:
1-47W844-1 LP/Objective: OPL271CCW B.9 History: System bank Level: Memy Comments: FHW 12/02 075 K11.01
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 133
,25. SDCS-B.12 023 Unit One is operating at 100% power whon4 a llh n m
v~c"", b-ir window i .
Which ONE (1) of the following conditions is required for operation of the Steam Dumps?
,/A. The Steam Dump System will be enabled only after Main Condenser pressure is 3.4 psia or less and one (or more) Condenser Circulating Water Pump is running.
B. The Steam Dump System is.t46W available for operation,> 4 1r 2.7 C. The Steam Dump System will be enabled only after Main Condenser pressure is psia or less and one (or more) Condenser Circulating Water Pump is running.
D. The Steam Dump System will be enabled only after Main Condenser pressure is 1.72 psia and one (or more) Condenser Circulating Water Pump is running.
A. Correct per 1-AR-M4-A (E-6), 6 to 7 inches Hg abs or approx. 3.4 psia.
B. Incorrect C-9 is lit when available, see 1-AR-M4-A (E-6).
C. Incorrect outside PS setpoint.
D. Incorrect outside PS setpoint, this value is the manual trip setpoint per AOP-S.02.
Student must understand when the C-9 interlock window is lit the condenser is available. l V' "
C-,.,,A*-, &Y-*;,, r- dA91,,_ H K/A[CFR]: -. x041 A4.08 (3.0 - 3.1) "
..- ,vo051 AK3.01 [2.8/3.1] (41.5 41.10} 4 0c, A1"t
Reference:
1-AR-M4-A (E-6)
TI-28 LP/Objective: OPL271C030, b.10 History: System bank, old Bank Number PL-1440 Level: CompJnsion Comment: FHW 12/02 051 AK3.01
34 (E-6)
Source Setpoint C-9 SER 1934 2/2 PS 6 to 7" Hg abs (approx. 3.4 psia)
CONDENSER Condenser vacuum PS-2-1 B AND INTERLOCK and PS-2-7C one condenser CCWP "A"-6.9 UNIT Bd "A", circulating water Comp 9 pump breaker closed.
CCWP "B"-6.9 UNIT Bd. "B",
Comp 5 (Normally Lit at Full Power)
CCWP "C"-6.9 UNIT Bd "D",
Comp 9
- 1. Main condenser has sufficient vacuum to support steam Probable dump operation with Ž- one CCWP running.
Causes NOTE The presence of this alarm allows steam dumps to operate.
pressure less than 3.4 psia.
Corrective [1] VERIFY condenser Actions [2] VERIFY at least one condenser circulating water pump operation.
during normal operation, THEN
[31 IF C-9 goes dark REFER to AOP-S.02, Loss of CondenserVacuum.
References 45N601-1, 45B655-04A-0, 47W611-2
CONTROLS TAVG PZR LEVEL 24.7%0 - 60% 0 547 OF - 578.2OF (547 F - 578.2 F Tavg [Auict. Hi Tavg])
(0 -100% power)
FEEDPUMP SPEED CONTROL S/G LEVEL 80 -195 psid, (0 - 20 %, constant 80 psig) 33 - 44% (0 - 20% turbine load) 44% (20 - 100% turbine load)
(20 100% total steam flow)
FIRST-STAGE IMPULSE PRESSURE ROD CONTROL Auto: 8 spm; 1.5°F -), 3.0°F error 0 - 628 psia 0 0 - 100% power 8 spm - 72 spm; 3°F -->5 F error Manual: 48 spin Bank Select: 48 spin control rods 64 spm shutdown rods FP Rod Insertion Limit: 182 steps on "D"Bank STEAM DUMPS Blocking:
Steam Dump Bypass Interlock Switches (M-4) in OFF Condenser not available (absence of C-9) be bypassed for 3 cooldown valves using Lo-Lo Tavg (P-12) locks out all steam dumps; interlock can the two bypass interlock switches on M-4.
Arming:
by PT-1-72 (C-7).
Load Rejection: 10% load decrease in a 2 minute time constant as sensed Controller I/S.)
Reactor Trip: P-4 from 'A' Tr. Rx Trip Breakers ('B' Tr. P-4 places Rx Trip Mode Selector Switch (M-4) in STEAM PRESSURE Opening:
Load Reject: Tave - Tref (PT-1-73)
(20F -4 1F = 0 -- 100% open)
Tave Mode = Trip Open - 1/2 @ 1O°F; 1/2 @ 18'F Reactor Trip: Tave - 552 0 F (Fixed Reference Signal)
(00F -4 50°F = 0 --> 100% open)
/ Trip Open - 1/2 @ 257F; 1/2 @ 50'F Pressure Mode Auto = steam pressure - setpoint Manual = Operator Controlled
HLC12-02.BNK Page: 128 Wednesday, October 09,2002 @ 11:46 AM 120. RHR-B.13.H 001 Given the following plant conditions:
- Plant cooldown in progress using two trains of RHR.
"* The flow is 2500 gpm per train.
"* The RCS cooldown rate is too high.
to _ECREA$E Which ONE (1) of the following operator actions is required flow?
the RCS cooldown rate while maintaining constant RHR valves on the RHR A. 4-T--Q-o~e)f the Component Cooling water outlet heat exchangers to slow the cooldown rate. /v",ra *,*lu 4 i, ,
B. Throttle open the RHR heat exchanger bypass valve.
closed to C. Throttle the outlet valves on the RHR heat exchangers decrease the cooldown rate.
and throttle closed v/D. Throttle open the RHR heat exchanger bypass valve the RHR heat exchanger outlet valves.
A. Incorrect per reference.
B. Incorrect per reference.
lead to vortex.
C. Incorrect per reference and decreased flow could D. Correct per refereace.
K/A [CFR]: co06 A1.01 [3.5/3.6] [41.5]
Reference:
0-SO-74-1 R39 section 5.6 LP/Objective: OPL271RHR B.12 History: Systems bank Level: Me ri"eyY Comments: FHW 12/02 005 Al1.01
SQN RESIDUAL HEAT 0-SO-74-1 REMOVAL SYSTEM Rev: 39 1,2 Page 83 of 185 Unit Date 5.6 Startup of RHR System for Normal Cooldown Mode (Continued)
CAUTION Flow through FCV-74-16, FCV-74-28, and FCV-74-32 must be coordinated to achieve RCS cooldown rate of 50°F per hour (1OOF per hour absolute maximum).
The valve for the RHR train in service may be throttled to obtain NOTE 1 desired cooling rate.
NOTE 2 Steps [29] and [30] should be performed concurrently.
[29] SLOWLY THROTTLE [FCV-74-32] RHR Hx bypass to avoid thermal shock.
[30] THROTTLE the RHR Hx valve for the RHR train being placed in service. (NIA other valve)
VALVE NO. FUNCTION INITIALS FCV-74-16 RHR Hx A Outlet FCV-74-28 RHR Hx B Outlet CAUTION RHR must be capable of 4 Loop Cold Leg injection while in Mode 4.
[31] IF required, THEN CLOSE the RHR cross-tie valve for the train not in service (N/A other valve):
VALVE NO. FUNCTION INITIALS FCV-74-33 RHR Hx A Outlet FCV-74-35 RHR Hx B Outlet
[32] IF two TRAIN RHR cooldown is desired, THEN GO TO Section 6.2.
END OF TEXT
C C ROISRO/KA Cross Reference RO/SRO M=Memory Type K/AID Question S=SRO C=Comph A=As Is R=RO A=Analysis M=Modified ID B=BOTH N=New AOP-C.01-B.1 002 B C A 001 AK1.03 001 K.301 001 K.301 B A N CONTROL*RODS 020 B M A 001 K5.30 RCP-B.12 002 B M A 003 A1.03 004 A3.07 004 A3.07 B C N 005 A1.01 RHR-B.13.H 001 B M N 006 A1.07 ECCS-B.3 002 B M A 006 K6.19 006 K6.19 B C N 008 K4.02 CCS-B.9.A 001 B A A 009 EK1.01 INPO3 955 B C A 010 A2.02 010 A2.02 B M N 012 K6.07 012 K6.07 B C N 013 K4.01 B M N 013 K4.01 AOP-R.05-B.5 001 B M A 015 AK2.08 RCS-TEMP-B.2 008 B M A 016 A3.02 INCORE-B.1.B 002 B M A 016 K1.01 017 K6.01 INCORE-B.l.D 003 B M A 022 K2.01 B C N 022 K2.01 024 AA1.26 ES-0.1-B.1 002 B C A 025 AK2.05 FR-Z.1-B.2 001 B C A 025 K5.02 CTMT-B.11 001 B C A 027 AK1.02 OPL271C353.4 001 B C A
- 1
C C RO/SRO M=Memory Type K/A ID Question S=SRO C=Comph A=As Is R=RO A=Analysis M=Modified ID B=BOTH N=New 029 EK2.06 RPS-B.5.B 001 B C A 034 A1.02 034 A1.02 B A N 034 K4.02 FH-B.12.B 001 B M A 035 K5.01 035 K5.01 B C N 037 2.1.7 AOP-R.01-B.2 003 B A A 037 AA1.11 AOP-R.01-B.2 004 B M A 038 EK3.08 INPO2 976 B C A 039 K3.04 FW-B.5 001 B A A 040 AK1.01 INPO2 231 B C A 051 AK3.01 SDCS-B.12 023 B C A 058 AK3.01 AOP-P.02-B.4 002 B C A 059 A4.01 059 A4.01 B C N 059 K1.04 MFW 001 B M A 060 AK2.02 060 AK2.02 B C N 061 AK2.01 061 AK2.01 B A N 061 K2.01 061 K2.01 B M N 061 K6.01 061 K6.01 B M N 062 A4.03 INPO8 155 B M A 063 A2.01 063 A2.01 B M N 063 K4.04 FW-B.5.B 004 B M A 064 K2.02 064 K2.02 B C N 068 A3.02 RADWASTE-B.12 008 B M A 069 AA1.03 CTMT-B.5 004 B C A 071 A1.06 RMS-B.9 001 B M A 2
C ( (
d/sRo M=Memory Type RO/SRO S=SRO M=Memory C=Comph A=As Is Type Question R=RO A=Analysis M=Modif ied KJA ID B=BOTH N=New ID B M A 072 A-1.01 RMS-B.2 004 B C A 074 EK1.0 FR-C.1-B.2 003 B A A 074 EK2.02 FR-C.1.-2.2 01 B M A 075 K1.01_ RCW-B.9 001 B C N 076 AK3.05 076 AK3.O5 B C A 078 K1.01 AOR-B.12 002 B C A 079 K4.01 AIR-CR.5 014 B M N 10O3 A3.01 103 A3.01 B A N 2.1.24 2.1.24 B M A 2.2.11 CTMT-B.111 004 B A N 2.2.4 2.2.4 B C A 2.3.11 ODCM-B.5 001 B M A 2.3.9 CTMT PURGE-B3.4 001 MA A B
2.4.39 REP-B.I.D 002 B C A 2.4.4 AOP-R.02-B.2 001 B C A E01 EAI.1 ES-0.0-B.3 001 B C A E03 EA1.2 ES-1.2-B.2 006 B M A E03 EK1.3 ES-1.2-B.2 004 B M A E04 EK2.2 ECA-1.2-B.1 002 B C A E05 EA1.3 FR-H.1-B.3 004 B C E08 EK3.2 FR-P.1 001 A B
E09 EAI.1 0PL271C382.4 001 A A E10 EK3.1 ES-0.2-B.3 003 B C A E14 EK1.3 ECCS-B.2 002 B6 M A M A E15R EK1 _1 FR-Z.2-B.2 001 B 3
$ 43 RO/SRO M=Memory Type Question S=SRO C=Comph A=As Is K/A ID R=RO A=Analysis M=Modified ID B=BOTH N=New 003 K4.02 T/S0304.02 003 R C A 005 AA2.01 005 AA2.01 R C N 008 A2.04 008 A2.04 R C N 012 K2.01 RDCNT-B.7.B 001 R C A 015 A2.03 FISSION*PROD*POISON 052 R M A 022 2.4.27 022 2.4.27 R A N 025 2.2.13 025 2.2.13 R M N 029 A-1.02 PI-B.10 001 R M A 039 K5.05 RVINT-B.8 001 R M A 056 A2.04 056 A2.04 R C N 056 AK3.01 OPL271 C368.3 001 R M A 056 K1.03 056 K1.03 R M N 059 AA2.05 059 AA2.05 R M N 064 A3.06 D/G-B.10 002 R A A 067 AA2.17 AOP-C.04 001 R M A 068 AK2.07 D/G-B.6 012 R M A 068 K4.01 RADWASTE-B. 12 006 R M A 071 A4.26 WGDS-B.12 001 R C A 078 K3.03 078 K3.03 R C N 0792.1.1 0792.1.1 R M N 086 K3.01 086 K3.01 R C N 2.1.16 2.1.16 R M N 2.1.31 PZR PRESS-B.9 006 R C A 2.2.26 REFUELING-B.1.G 001 R M A 2.3.10 2.3.10 R C N 4
C (
RO/SRO M=Memory Type K/AID Question S=SRO C=Comph A=As Is R=RO A=Analysis M=Modified ID B=BOTH N=New 2.4.20 AOP-M.01-B.3 002 R M A EPM-4-B.7 001 R C A 2.4.5 E01 2.2.25 E01 2.2.25 R C N E15 EK2.2 E15 EK2.2 R C N 003 DROP ROD AOP-C.01-B.5 011 S M A 2.4.4 004 2.4.1 OPL271C367.1 002 S C A 0072.1.10 PRT-B.7 004 S M A 010 2.4.47 AOP-1.04-B.2 001 S A A 011 2.1.6 OPL271C367.1 003 S C A 013 2.4.47 E-0-B.6 003 S M A 015 AA2.07 AOP-R.04-B.5 001 S A A 022 AA2.01 AOP-R.05-B.2 001 S A A 025 AA2.06 RHR-B.12.A 003 S M A 027 AA2.04 T.S-2.1 -B.2 001 S A A 040 AA2.01 OPL271C379.3 001 S A A 056 AA2.18 E-0-B.3.A 007 S C A 0692.1.14 0692.1.14 S C N 2.1.10 OPL271C458.1 001 S M A 2.1.22 GO-2-B.1 003 S M A 2.1.7 PZR LEVEL-B.14 003 S A A OPL271C1 80.3 001 S M A 2.2.25 2.2.9 SPP-9.5 001 S M A 2.3.1 RCI-15 001 S M A 5
(K (I (
RO/SHO M=Memory Type RO/SRO S=SRO M=MemoryI C=Comph A=As IsSType Question R=RO A=Analysis M=Modified K/A ID N=New ID B=BOTH S M A 2.3.2 RADIATION 002 S M A 2.3.3 AOP-C.04-B.5 008 S M A 2.4.33 OPDP-4 002 S C AOP-M.03-B.1 002 A 2.4.45 S C A E01 EA2.1 ES-0.0-B.5 002 S C A E04 EA2.1 ECA-1.2-B.2 001 S C A E09 EA2.1 ES-0.2-B.3 001 A- A E12 2.4.16 ECA-2.1-B.1 004 S M A S
E15 EA2.1 E15 EA2.1 N S c N El6 2.4.41 E16 2.4.41 6
Wednesday, October 09,2002 @ 11y AM HLC12-02.BNK Page: 3
- 3. 005 AA2.01 001 /
" > Rod Cluster Control Assembly F8 is greater than 12 steps from it's group step counter met demand position. Tech Spec actions for continued power operation have been except for verifing FQ (z) and FN AH are within their limits.
Which one of the following will Reactor Engineering use for evaluation of FQ (z) and FNAH per Tech Specs for this situation?
A. delta I indications.
/B. incore detector flux map.
C. core exit thermocouples.
D. NIS indications.
A. Incorrect, delta I does not determine rod power distribution.
B. Correct method to determine rod power distribution.
C. Incorrect, core exit T/Cs do not determine rod power distribution.
D. Incorrect, NIS does not determine rod power distribution.
K KIA[CFR]: 005 AA2.01 [3.3/4.1] [43.5]
Reference:
TS 3.1.3.1 action c.3.c).
LP/Objective: OPL271C071 B.6 History: New question.
Level: Comprehension.
Comments: FHW 12/02 005 AA2.01V'lt is doubtful that a student can recall an action method in the second sub-paragraph of an LCO, therefore he must understand is by that the only method of obtaining the heat flux along the "z" direction of a fuel rod an incore probe.
REACTIVITY CONTROL SYSTEMS
\<> 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within +/- 12 steps (indicated position) of their group step counter demand position.
APPLICABILITY: MODES 1* and 2*
ACTION:
- a. With one or more full length rods untrippable, determine that the R219 SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. With more than one full length rod misaligned from the group step R219 counter demand position by more than +/- 12 steps (indicated position),
be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- c. With one full length rod misaligned from its group step counter R219 demand height by more than +/- 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:
- 1. The rod is restored within the above alignment requirements, or R219
- 2. The remainder of the rods in the group with the misaligned rod R21 0 are aligned to within +/- 12 steps of the misaligned rod while maintaining the rod sequence and insertion limit of specification 3.1.3.6. The THERMAL POWER level shall be R1.
restricted pursuant to Specification 3.1.3.6 during subsequent operation, or
- 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:
N
- See Special Test Exceptions 3.10.2 and 3.10.3.
November 21, 1995 SEQUOYAH - UNIT 1 3/4 1-14 Amendment No. 114, 155, 215
<2
REACTIVITY CONTROL SYSTEMS ACTION: (Continued) of Table 3.1-1 is a) A reevaluatien of each accident analysis shall confirm reevaluation performed within 5 days; this results of these accidents that the previously analyzed operation under these remain valid for the duration of conditions.
of Specification 3.1.1.1 is FP b) The SHUTDOWN MARGIN requirement determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
from the movable c) A power distribution map is obtained verified to be incore detectors and FQ(Z) and FN AH are within their limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
to less than or equal to d) The THERMAL POWER level is reduced one hour and within the 75% of RATED THERMAL POWER within trip setpoint is reduced next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the high neutron flux POWER.
of RATED THERMAL to less than or equal to 85%
SURVEILLANCE REQUIREMENTS length rod shall be determined to beat 4.1.3.1.1 The position of eachby full verifying the individual rod positions within the group demand limit time intervals when the Rod Positiononce least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during verify the group positions at least Deviation Monitor is inoperable, then per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
be by movement core shall movement of IR219 trippable by rodverifying freedom inof the not fullyrodinserted 4.1.3.1.2 to determined Eachbe full-length per 92 days.
a10 steps in either direction at least once November 21, 1995 Amendment No. 215
- TIPJT 1 3/4 1-15 CEflTQVL
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 5
- 5. 008 A2.04 001 "0-RA-90-123A CCS Liq Eft Mon High Rad" alarm is lit.
Which one of the following is the correct cause and action for this condition?
A. A leak in the Containment Spray Heat Exchanger and ensure vents are closed.
B. The emergency su.ply valve to the CCS surge tank is not fully seated and dispatch -§*C) operations personnel to ensure the valve is closed and surge tank vents are closed<
",/C.A leak in the RHR heat exchanger and ensure the CCS surge tank vents are closed. A4 D. A leak exists in the CCP gear drive reservoir heat exchanger an surge tank vents are closed.
A. Incorrect, the cooling medium is ERCW.
B. Incorrect, the make up supply is ERCW and the supply has double isolation valves with a spool piece.
C. Correct per reference.
'- D. Incorrect, the cooling medium is ERCW.
K/A[CFR]: 008 A2.04 [3.3/3.5] [41.5 43.5]
Reference:
0-AR-M12-B (C-5) 1,2-47W845-2 V 2-47W845-4 ,t, C LPlObjective: OPL271C0S 8.9 History: New question.
Level: Comprehension U*t , I Comments: FHW 12/02 008 A2.04 -
djwý
19 (C-5)
Source Setpoint SER 769 Refer to TI-1 8 for setpoint calculation O-RA-90-123A O-RE-90-123 CCS LIQ EFF MON monitors outlet CCS Hx 081 and 082 HIGH RAD Retransmitted to U-2 SER 2248 Probable 1. RHR to CCS leakage in RHR Heat Exchanger 1B or2B. [C.1J Causes 2. Performing test on monitor.
Corrective [1] CHECK instruments on 0-M-12 (0-RM-90-123 and 0-RR-90-123)
Actions to determine if alarm is due to high radiation. [C.1]
[2] IF alarm is due to high radiation, THEN
[a] ENSURE 1-FCV-70-66 and 2-FCV-70-66 Component Cooling Water Surge Tank Vents are CLOSED. [C.1]
[b] REFER TO AOP-R.05, RCS Leak and Leak Source Identification to locate, AND ISOLATE the source of the leak. [C.1]
[c] NOTIFY Radiochemical Laboratory to sample component cooling water for radioactivity. [C.1]
[d] NOTIFY RADCON to survey for radiological hazards. [C.1]
[e] CONSIDER system "feed and bleed" in accordance with 0-SO-70-1, Component Cooling Water System "B" Train.
[C.1J References 45B655-128-0, 47B601-90-34, 47W610-90-2
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 9
- 9. 022 2.4.27 001 SUnit in progress and outside air being used for One is in mode 6 with fuel movement containment cooling. An outside spontaneous fire ignited near the Unit One Auxiliary Building air intake. The outside AUO reports there is a lot of smoke.
Which one of the following will prevent smoke from getting into Unit One containment?
A. Containment Purge air fans will automatically shut down.
B. Unit One Auxiliary Building supply fans will automatically shut down. V "
'I
"" - .nn . .
mC"era. , lfl..
'/ .Manual C ontainm ent vent isolation D. Upper and Lower compartment coolers have HEPA filters.
A. Incorrect, even though the purge fans have a suction source in the auxiary building air intake plenum, the fans will not automatically shut down, reference print 1,2-47W611-30-1.
B. Incorrect, smoke detection in the auxiliary building supply will automatically trip the auxiliary building supply fans, but the purge fans take suction from the intake plenum just as the auxiliary building supply fans, therefore smoke can still be drawn through S the air intake plenum, see reference print 1,2-47W611-30-5. 1,2-47W866-2.
C. Correct, even though the containment vent isolation will close the containment dampers as well as stop the purge fans; a smoke detection signal is not part of the logic, so a manual action is required, see reference print 1,2-47W611-30-1.
D. Incorrect, the upper and lower compartment coolers' suction are located inside containment and have cooling coils, see reference 1-47W866-1.
K/A[CFR]: 022 2.4.27 [3.0/3.5] [41.10 43.5]
Reference:
0-SO-30-3 V 1-47W866-1 flt* ¢ 1,2-47W866-2 1 ,2-47W611-30-1 1,2-47W611-30-5 TS 3.9.9.
LP/Objective: OPL217C426 B.3 History: New question.
Analysis SLevel:
Comments: FHW 12/02 022 2.4.27
Unit Date 8.6 Simultaneous "A" Train" Purging of Lower Containment and Upper Containment (MODES 5 & 6 ONLY)
CAUTION This procedure may only be performed in MODES 5 or 6.
[1] INITIATE 0-SI-CEM-30-410.2, unless not required during Mode 5 & 6, as determined by the Unit Supervisor/SRO and RCL Supervision.
[2] VERIFY there is not a containment vent isolation signal.
CAUTION During MODE 6 operation with the wafer valve open, containment purge startup may result in over flow of the SFP or reactor cavity due to pressure changes in the containment building. [C.2]
[3] IF in MODE 6 with the wafer valve open, THEN
[a] COMPARE SFP level to reactor cavity level.
[C.2]
[b] MONITOR locally SFP level or reactor cavity level during purge startup. [C.2]
[4] CHECK Containment Purge Train A filter assembly drain loop seal sight glass level equal to or greater than 50%
(AB, El 669, Penetration Room).
[5] IF the loop seal sight glass level is not equal to or greater than 50%, THEN FILL loop seal in accordance with 0-SO-30-3, Section 8.5.
NOTE If Rad Monitor 400 is inoperable, it may be N/A provided the ODCM requirements are met.
[6] ENSURE the Shield Building Vent Monitor RM-90-400A, B, C radiation monitor is in service.
SON CONTAINMENT PURGE SYSTEM 0-SO-30-3 OPERATION Rev: 23 1,2 Page 84 of 115
,/.
Unit Date 8.6 Simultaneous "A" Train" Purging of Lower Containment and Upper Containment (MODES 5 & 6 ONLY) (Continued)
[7] ENSURE Containment Purge Exhaust monitor RM-90-130 and/or RM-90-131 operable and aligned to the train of purge fans being used in accordance with 1(2)-SO-90-2.
[8] ENSURE containment pressure between -0.1 and 0.23 psid.
NOTE Opening damper [FCV-30-21 last ensures a positive pressure will be maintained in the upper compartment with respect to the lower compartment during the startup transient, and will ensure that the lower ice doors remain closed during startup when out-of-balance flow situation may exist.
[9] VERIFY Power Checklist 1(2)-30-3.01 complete.
[10] VERIFY Valve Checklist 1(2)-30-3.02 complete.
[11] IF Ice condenser doors are NOT blocked closed, THEN MONITOR the Ice Condenser Lower Inlet Door Open alarm on M-6.
[12] IF Ice Condenser Lower Inlet Door Open alarm is inoperable, THEN MONITOR the ice condenser door status lights on M-10 once per hour.
[13] VERIFY all containment purge fans OFF.
[14] IF the unit to be purged is in MODE 5 or 6 AND the opposite unit is in MODE 1, 2, 3, or 4, THEN VERIFY an operator is available and briefed on his duties related to shutdown containment purge system in accordance with 0-SO-30-3 in the event of an ABI.
(This action can be performed with minimum Tech Spec shift crew without impeding mitigation of the event.) [C.1]
SON CONTAINMENT PURGE SYSTEM 0-SO-30-3 OPERATION Rev: 23 1,2 Page 85 of 115 Unit Date 8.6 Simultaneous "A" Train" Purging of Lower Containment and Upper Containment (MODES 5 & 6 ONLY) (Continued)
[15] OPEN the exhaust fan discharge damper [FCV-30-2131 with rHS-30-2131.
[16] START the "A"Train purge/exhaust fan with [HS-30-1Al.
[17] VERIFY dampers OPEN.
"A" Train Initials FCO-30-1 A FCO-30-1 B FCO-30-295 FCO-30-294
[18] OPEN exhaust fan suction damper [FCV-30-611 with
[HS-30-611.
[19] OPEN lower containment valves [FCV-30-141 (supply) and
[FCV-30-561 (exhaust) with [HS-30-141.
[20] OPEN lower containment valves [FCV-30-151 (supply) and
[FCV-30-571 (exhaust) with [HS-30-151.
SON CONTAINMENT PURGE SYSTEM 0-SO-30-3 OPERATION Rev: 23 1,2 Page 86 of 115 Unit Date.
8.6 Simultaneous "A" Train" Purging of Lower Containment and Upper Containment (MODES 5 & 6 ONLY) (Continued)
CAUTION During outage conditions a flow imbalance can be created when a lower/upper compartment purge is inservice and changes are made to the upper and/or lower airlock doors (i.e. breaching or closing doors after purge is started). This situation may cause the Ice Condenser lower inlet doors to go open.
NOTE In modes 5 or 6 only, FCV-30-16 & 17, and FCV-30-37 & 40 may be opened. However, opening them could cause an increase in containment pressure.
[21] IF in modes 5 or 6 AND additional purge flow is desired, THEN OPEN the following:
A. 'FCV-30-161 with [HS-30-16 B. 'FCV-30-171 with [HS-30-17 C. "FCV-30-371 with [HS-30-371 D. rFCV-30-401 with [HS-30-401
[22] OPEN supply fan discharge damper [FCV-30-21 with
[HS-30-2].
CAUTION During fuel handling activities inside containment, purge flow shall not exceed 16,000 cfm. Containment purge flow may be monitored using FI-90-400 or using computer point Y2210A.
(
Reference:
FSAR 15.5.6)
[23] IF fuel handling is in progress or scheduled during the period this procedure section will be in effect, THEN MONITOR [FI-90-4001 OR [Y2210A1 to ensure flow does not exceed 16,000 CFM.
SON CONTAINMENT PURGE SYSTEM 0-SO-30-3 OPERATION Rev: 23 1,2 Page 87 of 115 Unit Date 8.6 Simultaneous "A"Train" Purging of Lower Containment and Upper Containment (MODES 5 & 6 ONLY) (Continued)
NOTE Lower containment purge is now in service the remaining portions of this procedure will place the Upper containment purge in service.
[24] OPEN one set of containment dampers with the associated handswitch. (NIA those not opened)
Initials Initials A. Handswitch HS-30-7 A. Handswitch HS-30-9 (Supply) FCV-30-7 (Supply) FCV-30-9 (Exhaust) FCV-30-51 (Exhaust) FCV-30-53 AND OR AND B. Handswitch HS-30-8 B. Handswitch HS-30-10 (Supply) FCV-30-8 (Supply) FCV-30-1 0 (Exhaust) FCV-30-50 (Exhaust) FCV-30-52
Unit Date.
8.6 Simultaneous "A" Train" Purging of Lower Containment and Upper Containment (MODES 5 & 6 ONLY) (Continued)
CAUTION During outage conditions a flow imbalance can be created when a lower/upper compartment purge is inservice and changes are made to the upper and/or lower airlock doors (i.e. breaching or closing doors after purge is started). This situation may cause the Ice Condenser lower inlet doors to go open.
NOTE In modes 5 or 6 only, both sets of containment dampers may be opened. However, opening them could cause an increase in containment pressure.
[25] IF in modes 5 or 6, and additional purge flow is desired, THEN OPEN Initials Initials A. Handswitch HS-30-7 A. Handswitch HS-30-9 (Supply) FCV-30-7 (Supply) FCV-30-9 (Exhaust) FCV-30-51 (Exhaust) FCV-30-53 AND AND AND B. Handswitch HS-30-8 B. Handswitch HS-30-10 (Supply) FCV-30-8 (Supply) FCV-30-10 (Exhaust) FCV-30-50 (Exhaust) FCV-30-52
[26] ENSURE Lower Ice Condenser door are CLOSED by either of the following:
[a] Absence of Ice Condenser Lower Inlet Door Open alarm on M-6.
[b] Monitoring the ice condenser door status lights on M-10
Unit Date 8.6 Simultaneous "A" Train" Purging of Lower Containment and Upper Containment (MODES 5 & 6 ONLY) (Continued)
NOTE Due to changes in the specific volume of air during purging, pressure problems may arise when equipment hatch is closed and air lock doors are breached.
Additional exhaust capacity will aid in reducing Rx. Bldg. Access room pressurization.
[27] IF Unit is in Mode 5 or 6 and Rx. Bldg. Access room(s) are being pressurized by purge, THEN PERFORM the following:
[a] ALIGN purge dampers as follows:
Damper Position Initials FCV-30-16 CLOSED FCV-30-17 CLOSED FCV-30-37 OPEN FCV-30-40 OPEN
[b] WHEN pressurization of Rx. Bldg. Access room(s) no longer exists (i.e. containment equipment hatch open or air locks no longer breached), THEN PERFORM the following:
[i] IF additional purge flow was desired in step [21],
THEN REALIGN purge dampers as follows:
Damper I Position Initials FCV-30-16 OPEN FCV-30-17 OPEN
[ii] IF step [21] was N/A, THEN ENSURE the following are CLOSED:
Damper Position Initials FCV-30-16 CLOSED FCV-30-17 CLOSED FCV-30-37 CLOSED FCV-30-40 CLOSED END OF TEXT
REFUELING OPERATIONS 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Ventilation isolation system shall be OPERABLE.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.
ACTION:
With the Containment Ventilation isolation system inoperable, close each of the Ventilation penetrations providing direct access from the containment atmosphere to the outside atmosphere. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.9 The Containment Ventilation isolation system shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of and at least during CORE ALTERATIONS by verifying that Containment Ventilationonce per 7 days occurs on manual initiation and on a high radiation test signal isolation from each of the containment radiation monitoring instrumentation channels.
SEQUOYAH - UNIT 1 3/4 9-9
Wednesday, October 09, 2002 @ 1P:45 AM HLC12-02.BNK Page: 11
- 11. 025 2.2.13 0017/
~ Unit One is presently in mode 4 and preparing for a refueling outage. In order to stay ahead of the schedule you are directed to tag out the glycol expansion tank level switches.
Which one of the following devices will you use?
,/A. Plastic tag with stainless wire. /
B. Paper/cloth tag with waxed twine.
C. Plastic tag with waxed twine.
D. Paper/cloth tag with stainless wire.
A. Correct, the glycol tank is inside containment. Mode 4 is greater than 200°F therefore the procedure requires a plastic tag with stainless wire if the RCS temperature is greater than 2000 F.
B. Incorrect per reference.
C. Incorrect per reference.
D. Incorrect per reference.
. The student must understand the glycol tank is inside containment and that mode 4 is greater than 200OF which requires a unique card and hanging device (plastic with stainless wire).
KIA[CFR]: 025 2.2.13 [3.6/3.8] [41.10]
Reference:
SPP-10.2, Tech Spec definations Table 1.1 LP/Objective: OPL271 C528 B.9 History: New question.
Level: Memory (V00(ý'
Comments: FHW 12/02 025 2.2.13 tsJ
TVAN STANDARD SPP-10.2 PROGRAMS AND CLEARANCE PROGRAM Rev. 3S1 PROCESSES Page 11 of 47 3.4 Clearance and Tag TVpes A. TVAN clearance tags shall be used for SM controlled clearances.
B. At an RCS/Moderator temperature of less than or equal to 2000 F, tags (paper/cloth or plastic) that are placed inside containment may be attached to devices with either waxed twine or stainless steel wire. With an RCS/Moderator temperature in excess of 2000 F, tags (plastic) that are placed inside containment shall be attached with stainless steel wire. Before exceeding an RCS/Moderator temperature of 2000 F, an audit of all clearances in effect will be conducted for the purpose of identifying all tags in containment. Those tags shall be visually inspected and any tags that are paper/cloth will be replaced with plastic and any twine used to attach the tags will be replaced with stainless steel wire.
C. Tagouts may contain multiple Clearance Sections.
D. Each Clearance Section of a Tagout provides a safe clearance boundary for work on one or more discreet plant components, e.g., a pump.
E. A Clearance Section may list multiple tag types, e.g., Danger Tags, Caution Tags and Operating Permit Tags, as necessary to provide a safe clearance boundary.
F. A Clearance Section may list devices as required to support hanging the Section or to ensure restoration to normal configuration upon Section removal. Such devices do not require placement of an actual clearance tag.
G. TVAN Danger tags (TVA Form Series 19631 A) must be installed on all control points that isolate equipment from sources of energy and shall bear a device identifier similar to that on the device.
H. For Caution tags, the following apply:
- 1. Caution tags shall be attached to equipment, switches or controls where hazardous or abnormal operating conditions exist.
- 2. TVAN Caution tags (TVA Form Series 19629) may be issued by the SM to indicate abnormal operating conditions.
- 3. Caution tags should also be used as necessary to document the need for post maintenance/modification testing (PMT) after a clearance boundary has been released.
- 4. A position held is not specified for Caution tags listed on a Section. The operator's signature for tag placement documents placement of the tag only. If the device must be placed in a specific position to support the placement of the clearance, specify a position in the Tag Notes for the tag.
- 1. For Operating Permit tags the following apply:
- 1. Components may be held by only one Operating Permit tag at a time.
- 2. Only one person at a time may sign-on as a holder of Operating Permit tags.
- 3. Devices held by Operating Permit tags may not be held by Danger tags at the same time.
TABLE 1.1 OPERATIONAL MODES REACTIVITY % RATED AVERAGE COOLANT CONDITION. Kff THERMAL POWER* TEMPERATURE MODE R215
- 1. POWER OPERATION Ž0.99 >5% Ž350OF
- 2. STARTUP Ž0.99 Ž350"F R205
<0.99 0 Ž350"F I
- 3. HOT STANDBY
- 4. HOT SHUTDOWN <0.99 0 350"F > TV
>200 *F
- 5. COLD SHUTDOWN <0.99 0 *2000F R205
- 0.95 0 <140"F I
- 6. REFUELING**
- Excluding decay heat.
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
June 1, 1995 SEQUOYAH - UNIT I 1-8 Amendment No. 71, 201
Wednesday, October 09, 2002 @ ,14:45 AM HLC12-02.BNK Page: 14 Wednesday, October 09, 2002 @ jA:45 AM HLCI2-02.BNK Page: 14
- 14. 056 A2.04 001 v
"*J Unit One is operating at 65% power with all systems aligned normal. The crew had just
~~ placedj the second Main Feedwater Pump 1 B-B in service. A /0C pý SA
- * *ndthe pdS-oný common injection water supply line the MFPs.
Which one of the following is the correct result and proper action for this condition?
A. The turbine will run back and the crew should use AOP-S.01, "Loss of Normal Feedwater."
The unit will trip and the crew should use E-0 "Reactor Trip or Safety Injection."
C. Only the 1 B-B Main Feedwater Pump will trip and the crew should use the Annunciator Response Procedure for MFW pump "Tripped."
aen n dter Pump trips and the crew should use 0-GO-5, "Normal Power Uperation." " " * "-Ek7 A. Incorrect, per TI-28 attachment 9.
B. Correct per 1-AR-M3-B (A-2). the trip pressure switches are on the common supply line which will cause a trip signal to both MFPs.
C. Incorrect per 1-AR-M3-B (A-2), both MFPs will trip.
D. Incorrect per 1-AR-M3-B (A-2), both MFPs will trip.
K/A[CFR]: 056 A2.04 [2.6/2.8] [41.5 43.5]
Reference:
1,2-47W803-1 1-45N657-18 1-45N646-1 1-AR-M3-B (A-2)
TI-28 attachment 9 LP/Objective: OPL271COND.FW B.11 History: New question.
Level: Comprehension Comments: FHW 12/02 056 A2.04 7
to4 bý, ý,
2 (A-2)
Source Setpoint SER 1404 N/A R/TT-1, energized MFPT-1A tripped; SER 1406 RrrT-1, energized MFPT-1 Btripped.
Probable 1. Thrust bearing failure.
Causes 2. Low turbine bearing oil pressure.
- 3. Low feedwater bearing oil pressure.
- 5. MFWP condenser vacuum low.
- 6. Low injection water pressure.
- 7. Turbine overspeed.
Corrective E1] IF unit > 80% load, THEN Actions VERIFY all auxiliary feedwater pumps start.
12] IF unit > 80% load and one MFP trips, THEN GO TO AOP-S.01, Loss of Normal Feedwater.
[3] IF unit < 80% load and one MFP trips, THEN
[a] REDUCE turbine load until feedwater flow is greater than steam flow.
[b] IF operating MFP will NOT respond in AUTO, THEN PLACE MFP master controller to MANUAL and adjust feedwater flow to greater than steam flow.
[c] ENSURE affected MFPT condenser isolation valves CLOSED.
[d] DISPATCH personnel to investigate trip of MFP.
[e] ENSURE unit is returning to stable conditions.
1f] INITIATE repairs on the affected MFP.
[g] RETURN MFP controller to AUTO if step [b] was performed.
[h] EVALUATE removal of secondary plant equipment based on condensate pressure.
[i] RESET steam dumps using appropriate plant instructions, if C-7 (AR-4A, window E-5) is illuminated.
w] GO TO O-GO-5, Appendix B for restoration of turbine controls.
[4] IF both MFPs trip and reactor power > 50%, THEN VERIFY reactor tripped, AND GO TO E-0, Reactor Tnp or Safety Injection.
[5] IF both MFPs trip, THEN VERIFY main turbine tripped.
References 45B655-036-0, 45N646-1
\,-I.
-- - - TI-28 Att. 9 SON UNIT 1 & 2 CYCLE DATA SHEET Effective Date 04/25/01.
(FOR INFORMATION ONLY)} Page 11 of 16 Signal Setpoint Permissive MAIN FEEDWATER PUMP TURBINE ( FPT) TRIPS
- 1. Manual M-3 or Local
- 2. Low Turbine Bearing Oil Pressure < 10 psig
- 3. Low Pump Bearing Oil Pressure < 10 psig
- 4. Thrust Bearing Wear Excessive Variable (changes each time we Check with Instrument Dept.
come up in power)
- 5. MFP Suction Isolation Valve NOT Full Open
- 6. MFP Turbine Condenser Vacuum > 12.2 psia Low
- 7. Low Injection Water Pressure < 220 psig 20 second time delay
- 8. Mechanical Overspeed 6600 rpm.
Any 1 of 3 signals
- 9. Feedwater Isolation Signal
TI-28 Aft. 9 SON UNIT 1 & 2 CYCLE DATA SHEET Effective Date 04./.* .
(FOR INFORMATION ONLYV Page 12 of 16 Signal Setpoint Logic Permissive TURBINE TRIPS-
- 1. Reactor Trip From P-4 Contacts 1/2 Trains
- 2. Safety Injection SIS (K621 input) 1/2 Trains
- 3. S/G Hi Hi Level (P14) > 81% 2/3 LTs on 1/4 SGs
- 4. Manual (M-2 or Local)
- 5. Auto Stop Oil Pressure < 45 psig 2/3
- 7. Mechanical Overspeed 110% of Rated Speed 1/1
+ 10 rpm (1950- 1975 RPM)
- 8. Low Vacuum 3.9 - 5.4 psia 1/1
- 9. Thrust Bearing Wear > 60 psig 1/2
- 10. Low Bearing Oil Pressure < 7 psig 1/1
- 11. Stator Cooling System 45 sec. time delay
- Low AP across Stator 12 psid less than normal 2/3 Gen. amps >15%
- Stator Outlet Temp. > 90C 2/3 Gen. amps >15%
- 12. High Vibration > 14 mils (Alarm @ > 7 mils) 1/11 Not Cutout
- 13. MFPT A & B Tripped 2/2
- 14. Main Generator Input Trips
- Reverse Power (32X)*
- 30 sec. time delay
- Gen. Backup (86GBX)
- Neg. Phase Sequence (46X)
- Overcurrent (51 GX)
- System Ground (51/59X)
- Differential (87X)
- 15. Main Transformer Input Trips
- Differential (87TX) 1/3
- Low Voltage Bushing Oil 1/3 (on 1/3 Unit 1 only.
Flow Low transformers)
- 16. USS Transformer A & B Input Trips
- Differential (A87SX)
- Differential (B87SX)
- 17. Loss of + 48 V DC 2/2
- 18. Loss of +/- 15 V DC 2/2
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 15
- 15. 056 K1.03 001 Unit One is operating at 100% power and all systems are aligned normal. The crew notices the "PS-2-129 Low NPSH at MFP's" alarm just annunciated.
Which one of the following could bring in this alarm?
A. Condensate low pressure heater strings "A"and "B" isolated on hi-hi #7A and #7B heater levels.
B. High pressure feedwater heater #1 A discharge valve was inadvertently closed.
VC. Trip of a condensate booster pump.
D. Condensate intermediate heaters #3B and #3C are above program level.
A. Incorrect, there is no string isolation logic associated with the #7 heaters.
B. Incorrect, the isolated #1 heater discharge valve is downstream of the MFP.
C. Correct, per reference.
D. Incorrect, the #3 heaters do not have a program level.
K/A[CFR]: 056 K1.03 [2.6/2.6] [41.2-9]
S
Reference:
1-AR-M3-A (E-1) 1,2-47W611-2-1 1,2-47W611-2-2 1,2-47W803-1 LP/Objective: OPN218SS.006A B.3 History: New question.
Level: Memory Comments: FHW 12/02 056 K1.03
29 (E-1)
Source Setpoint I PS-2-129 I SER 214 LOW NP-M 100 psid decreasing PS-2-129A (differential between MFP inlet and #2 heater AT MFP'S i shell)
Probable 1. Increasing load.
Causes 2. Condensate booster pump trip.
Corrective [1] ENSURE MFP inlet pressure greater than 320 psig by Actions reducing turbine load [M-3, P1-2-1293.
as required E2] ENSURE the following pumps are operating by O-GO-5 for the present unit load.
- a. Hotwell pumps
- b. Condensate DI booster pumps
- c. Condensate booster pumps
<---V
- d. #7 HDT pumps
- e. #3 HDT pump to clear
[3) ADJUST condensate and feedwater pressure as required alarm.
E4] IF MFP trips, THEN GO TO AOP-S.01, Loss of Normal Feedwater.
References 47B601-2-21 45B655-03A-2 47W611-2-2
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 17
- 17. 059 AA2.05 001
"- Given the following plant conditions and information:
- Unit 1 is at 21 00 F and is being cooled down after discovery of a fuel clad failure.
- RCDT pump A is being used to lower level in the RCDT.
- Panel 1-XA-55-30 window "1-RA-277A RCDT Hi Rad" is in alarm.
9(inside containment isolation valve).
FCV-77-10 (outside containment isolation valve).
The SRO directs the Operator to pump the contents of the RCDT to the TDCT. Which of the following actions should the Operator take to pump down the RCDT?
A. Place 1-HS-77-9B and 1-HS-77-1OB to "Normal" then open 1-FCV-77-9A and 1-FCV-77-1 OA.
,/B. Place 1-HS-77-9B and 1-HS-77-1OB to "Block" then open 1-FCV-77-9A and 1-FCV-77-1 OA.
C. Place 1-HS 9A and 1-HS-77-1 OA to "Open."
D. Cyclel-HS-77-9A andl-HS-77-1OA back to "A-Auto."
A. Incorrect per reference.
B. Correct per reference, block position will override a high rad signal.
C. Incorrect per reference, will not override a high rad signal.
D. Incorrect per reference.
K/A[CFR]: 059 AA2.05 [3.6/3.9] [43.5]
References:
47W611-77-1 1-AR-M30-A (B-3) .
LP/Objectives: OPL271 RADMON B.8 History: New question K ,
Level: Memory Comments: FHW 12/02 059 AA2.05 ..fra O0 .,,P ¢
11 (B-3)
Source Setpoint 1-RA-277A SER 1739 1-RE-90-277 RCDT 316 mR/Hr HI RAD Probable 1. Fuel cladding failure.
Causes 2. High activity in reactor coolant.
Corrective [1] VERIFY [1-FCV-77-9I RCDT pump to TDCT isolation valve Actions closes.
[2] CHECK rl-RM-90-2771 for high radiation on panel 1-M-30.
[3] IF need to pump RCDT to FDCT with this high radiation condition present, THEN REQUEST permission from SRO to place the following hand switches in block:
- a. [1-HS-77-9B1
- b. [1-HS-77-10B1
[4] WHEN ready to pump RCDT, THEN OPEN rl-FCV-77-10] and .l1-FCV-77-91.
[5] REQUEST radiochemical laboratory sample to verify activity.
[6] IF high activity verified, THEN GO TO AOP-R.06, High RCSActivity.
References 471601-55-75, 47W610-90-4
Wednesday, October 09, 2002 @!Z1:45 AM HLC12-02.BNK Page: 26
- 26. 078 K3.03 001o
"~ Plant Conditions:
- Unit One is operating at 100% power.
- Unit Two is operating at 100% power.
- Auxiliary Control Air Compressor A-A power supply breaker on 480V C&A Vent Bd 2A1-A is tagged.
A pipe break in the control air header has allowed the control air header pressure to drop to 65 psig.
The Auxiliary Control Air Compressor B-B breaker on 480V C&A Vent BD 2B1-B has tripped.
Which one of the following is correct for these conditions.
A. Unit Two equipment using "B" train instrument ai ffected.
B. Unit Two equipment 4 sing both trains of instrument air are 'ffected.
C. Unit One equipmenirfungg both trains of instrument air areVfected.
,/D. Unit One and Unit Two equipment using both trains of instrument air areyffected.
A. Incorrect, control air compressors to auxiliarly air compressors isolate when control air pressure reaches 69 psig and neither aux air compressors are available therefore both trains of air are effected.
B. Incorrect, control air compressors to auxiliarly air compressors isolate when control air pressure reaches 69 psig and neither aux air compressors are available therefore both trains of air are effected which also supply Unit one.
C. Incorrect, control air compressors to auxiliarly air compressors isolate when control air pressure reaches 69 psig and neither aux air compressors are available therefore both trains of air are effected.
D. Correct, the auxiliary control air compressors supply both units.
K/A[CFR]: 078 K3.03 [3.0/3.4] [41.7]
Reference:
AOP-M.02 1,2-47W848-1 1,2-47W848-12 LP/Objective: OPL271 CSA B.11 History: New question.
Level: Comprehension Comments: FHW 12/02 078 K3.03
3.1 Symptoms (cont'd)
A. Deviations or unexpected indications on any of the following may indicate a loss of Control Air System pressure:
"* Aux Control Air Header pressure dropping:
"* 0-PI-32-104A, Aux Control Air Hdr A Press [M-15].
"* 0-PI-32-105A, Aux Control Air Hdr B Press [M-15].
"* Control and Service (C&S) air compressors tripped.
"* Valves or dampers moving to failed position(s).
"* S/G anomalies due to MFW Reg Valves failing closed.
"* MSIVs fail closed.
B. Any of the following automatic actions may indicate a loss of Control Air System pressure:
RANGE SETPOINT (psig) INITIATING EVENT (psig) min max EVENT N/A -- -- Normal control C&S air compressors auto start with circuit failure backup control circuit Service Air isolates from Control Air 88 86 90 Control Air Receiver (0-PCV-33-4) dropping C&S air compressors load to 50%
86 84 88 Control Air Receiver C&S air compressors load to 100%
pressure dropping 77 74.5 79.5 Aux Air Receiver Aux Air Compressors start pressure dropping 69 66.5 71.5 Control Air header Aux Air isolates from Control Air pressure dropping (0-FCV-32-82 and 85)
Aux Air Receiver Aux Air to containment valves fail 50 -- -- pressure dropping closed. (1-FCV-32-80, -102,
-110, 2-FCV-32-81, -103, -111) 3.2 Entry Conditions None Page 43 of 58
Wednesday, October 09, 2002 @)1:45 AM HLC12-02.BNK Page: 27 27.079 2.1.1 001/
- / The "Control and Service Air Compessor C or D Trouble/Shutdown" alarm just annunciated.
Which of the following is correct according to conduct of operations requirements.
A. Use the PA to dispatch operations personnel to the air compressors. "
- 'O,,
B. Use two way communications since there is no urgency.
C. Acknowledge the alarm and assume it was a pre-planned activity.
,/D. Acknowledge the alarm and perform actions in the annunciator response.
A. Incorrect per reference.
B. Inorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
K/A[CFR]: 079 2.1.1 [3.7/3.8] [41.10]
Reference:
OPDP-1 R1 LP/Objective: OPL271C209 B.6 and 8 History: New question.
Level: Memory Comments: FHW 12/02 079 2.1.1 1Dc't
TVAN STANDARD OPDP-1 DEPARTMENT CONDUCT OF OPERATIONS Rev. 1 PROCEDURE Page 23 of 31 3.10 Communications 3.10.1 Public Address System Use of the plant public address system (page) should be administratively controlled to ensure it retains its effectiveness in contacting plant personnel.
When using the PA system:
A. Speak slowly and deliberately in a normal tone of voice, using normally accepted equipment nomenclature and phonetic alphabet designations where needed to preclude potential confusion.
B. When announcements of abnormal or emergency conditions are made, they shall be made at least twice.
C. When announcing drill exercises, precede and follow all communications with, "This is a drill."
The following activities require a PA announcement:
A. Planned starting or tripping of large or loud equipment.
B. Change of plant status, i.e., unit scram/trip.
C. To summons operators or other plant personnel as needed.
D. Ventilation alignment changes which may temporarily affect access and egress to plant areas.
3.10.2 Radios Radio usage shall not be allowed in areas where electronic interference with plant equipment may result. Areas where radio use is prohibited shall be posted.
When using radios:
A. Identify yourself and your watch station.
B. Utilize three-way communication technique, normally accepted equipment nomenclature, and phonetic alphabet designations where appropriate.
C. Use professional language.
D. If signal breakup is experienced, relocate or use alternate method of communication.
E. When announcing drill exercises, precede and follow all communications with, "This is a drill."
3.10.3 Telephone Telephones are the preferred method of remote communication. When using telephones:
A. Identify yourself and your watch station.
B. Utilize three-way communication technique, where appropriate.
C. Use professional language.
D. When announcing drill exercises, precede and follow all communications with, 'This is a drill."
K-->
Wednesday, October 09, 2002 @ 1445 AM HLC12-02.BNK Page: 28
- 28. 086 K3.01 001,
" >The fire protection system failed to actuate allowing a spreader room fire to burn control wiring. This initiated a control room abandonment requiring the Operators to complete AOP-C.04 "Control Room Inaccessibility" checklists.
Which one of the following is correct for this condition? 6,#,c a.,*40 a"#44J*
A. The 1 A-A MDAFW pump can be started from the Unit 1 Auxiliary Control Room.
B. The 2 B-B RHR pump can be started from the Unit 2 Auxiliary Control Room.
C. The 1 B-B Safety Injection pump can be started from the 480v Shutdown Bd.1 61-B.
,/D. The 2 B-B RHR pump can be started from the 6.9 kv Shutdown Bd. 2 B-B.
A. Incorrect, the transfer of control was completed when the AOP-C.04 checklists were finished B. Incorrect, electrical devices are controlled from their switchgear.
C. Incorrect, the pump is supplied from the 6.9 kv shutdown bd.
D. Correct, the electrical devices are controlled from their switchgear.
KIA[CFR]: 086 K3.01 [2.7/3.2] [41.7]
Reference:
AOP-C.04 LP/Objective: 0PL271C423 B.5 History: New question.-'
Level: Comprehension Comments: FHW 12/02 086 K3.01 l1 04 Q0ýýA
SHUTDOWN FROM AUXILIARY CONTROL ROOM AOP-C.04 SQN Rev. 5 Page 4 of 7 CHECKLIST 3 6.9-kV SHUTDOWN BOARD 1B-B NOTE To ensure one ERCW pump will sequence on following a blackout, its auxiliary control switch must be placed in START. If an ERCW pump is running (breaker closed), it should be selected.
[7] PLACE auxiliary control switch for one ERCW pump in START momentarily.
[Compt 8 or 9] El
[8] ENSURE transfer switches in AUX position:
EQUIPMENT NAME COMPT TRANSFER SWITCH SWITCH POSITION EMERG SUPPLY BRKR FROM D/G 1B-B 6 1-XS-57-73 El AUX BREAKER 1914 ESSENTIAL RAW CLG WATER PUMP L-B 8 O-XS-67-440 El AUX ESSENTIAL RAW CLG WATER PUMP N-B 9 0-XS-67-452 El AUX AUX FEEDWATER PUMP 1B-B 10 1-XS-3-128 0l AUX NORM SUPPLY BRKR FROM 6.9KV UNIT BD 1C 11 1-XS-57-68 El AUX BREAKER 1726 CONTAINMENT SPRAY PUMP 1B-B 13 1-XS-72-10 El AUX RESID HEAT REMOVAL PUMP 1B-B 14 1-XS-74-20 El AUX SAFETY INJECTION PUMP 1B-B 15 1-XS-63-15 El AUX ALT SUPPLY BRKR FROM 6.9KV UNIT BD 1D 16 1-XS-57-71 El AUX BREAKER 1728 CENTRIFUGAL CHARGING PUMP 1B-B 18 1-XS-62-104 El AUX PZR HEATERS BACKUP GROUP 1B-B 20 1-XS-68-341D El AUX PZR HEATER BACKUP GROUP 1C 21 1-XS-68-341H El AUX
[9] ENSURE [O-XS-82-1221 GEN 1B-B TRANSFER switch in AUX position
[Shutdown Bd 1B-B Logic Panel 4] El Page 69 of 174
SHUTDOWN FROM AUXILIARY CONTROL ROOM AOP-C.04 SON Rev. 5 Page 3 of 7 CHECKLIST 3 6.9-kV SHUTDOWN BOARD IA-A NOTE To ensure one ERCW pump will sequence on following a blackout, its auxiliary control switch must be placed in START. If an ERCW pump is running (breaker closed), it should be selected.
[4] PLACE aux control switch for one Train A ERCW pump in START momentarily.
[Compt 8 or 9] El
[5] ENSURE transfer switches in AUX position:
EQUIPMENT NAME COMPT TRANSFER AUX SWITCH POSITION EMERGENCY SUPPLY BREAKER FROM DSL 6 1-XS-57-46 EB AUX GENERATOR 1A-A BREAKER 1912 ESSENTIAL RAW CLG WATER PUMP J-A 8 O-XS-67-432 El AUX ESSENTIAL RAW CLG WATER PUMP Q-A 9 O-XS-67-460 El AUX AUXILIARY FEEDWATER PUMP 1A-A 10 1-XS-3-118 LI AUX NORMAL SUPPLY BREAKER FROM 6900 V UNIT 11 1-XS-57-44 El AUX BOARD 1B BREAKER 1718 CONTAINMENT SPRAY PUMP 1A-A 13 1-XS-72-27 LI AUX RESIDUAL HEAT REMOVAL PMP 1A-A 14 1-XS-74-10 El AUX SAFETY INJECTION PUMP 1A-A 15 1-XS-63-10 E- AUX ALTERNATE SUPPLY BREAKER FROM 6900 V 16 1-XS-57-41 El AUX UNIT BOARD 1A BREAKER 1716 CENTRIFUGAL CHARGING PUMP lA-A 18 1-XS-62-108 El AUX PZR HEATERS BACKUP GROUP 1A-A 20 1-XS-68-341A LI AUX PZR HEATERS GROUP 1D 21 1-XS-68-341F EL AUX
[6] ENSURE [O-XS-82-1211 GEN 1A-A TRANSFER switch in AUX position
[Shutdown Bd 1A-A Logic Relay Panel 4] E]
Page 68 of 174
AOP-C.04 SQN SHUTDOWN FROM AUXILIARY CONTROL ROOM Rev. 5 Page 4 of 7 CHECKLIST 4 6.9-kV SHUTDOWN BOARD 2B-B NOTE To ensure one ERCW pump will sequence on following a blackout, its auxiliary control switch must be placed in START. If an ERCW pump is running (breaker closed), it should be selected.
[7] PLACE auxiliary control switch for one ERCW pump in START momentarily.
[Compt 8 or 9] l
[8] ENSURE transfer switches in AUX position:
EQUIPMENT NAME COMPT TRANSFER SWITCH SWITCH POSITION EMERGENCY SUPPLY BREAKER FROM DSL 6 2-XS-57-73 El AUX GENERATOR 2B-B BREAKER 1924 ESSENTIAL RAW CLG WATER PUMP P-B 8 0-XS-67-456 El AUX 3/4> 0-XS-67-444 El AUX ESSENTIAL RAW CLG WATER PUMP M-B 9 AUX FEEDWATER PUMP 20-B 10 2-XS-3-128 I- AUX NORMAL SUPPLY BREAKER FROM 11 2-XS-57-68 El AUX 6900 V UNIT BOARD 2C BREAKER 1826 CONTAINMENT SPRAY PUMP 2B-B 13 2-XS-72-10 L- AUX RESID HEAT REMOVAL PUMP 2B-B 14 2-XS-74-20 El AUX SAFETY INJECTION PUMP 2B-B 15 2-XS-63-15 El AUX ALTERNATE SUPPLY BREAKER FROM 16 2-XS-57-71 El AUX 6900 V UNIT BOARD 2D BREAKER 1828 CENTRIFUGAL CHARGING PUMP 2B-B 18 2-XS-62-104 El AUX PRESSURIZER HEATERS BACKUP GROUP 2B-B 20 2-XS-68-341 D El AUX PRESSURIZER HEATER BACKUP GROUP 2C 21 2-XS-68-341 H 0l AUX
[9] ENSURE [o-XS-82-1241 GEN 2B-B TRANSFER switch in AUX position []
[Shutdown Bd 26-B Logic Panel 4]
Page 76 of 174
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 30
- 30. 2.1.16 001 The Unit One operator notices the service air receiver is isolated and control air pressure is dropping.
Which of the following is correct for announcements of this abnormal condition?
A. Announce over the PA system one timepand phone the AUO field office.
v/B. Announce at least twice over the PA system.
C. Announce over the PA system one time. &J'lý D. .A-j'e--a-dde'd .ac ,
A. Incorrect per reference.
B. Correct per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A[CFR]: 2.1.16 [3.7/3.8] [41.10]
Reference:
OPDP-1 RI section 3.10.1 SLP/Objective: OPL271C209 B.6 History: New question.
Level: Memory Comments: FHW 12/02 2.1.16
TVAN STANDARD OPDP-1 DEPARTMENT CONDUCT OF OPERATIONS Rev. 1 PROCEDURE Page 23 of 31 3.10 Communications 3.10.1 Public Address System Use of the plant public address system (page) should be administratively controlled to ensure it retains its effectiveness in contacting plant personnel.
When using the PA system:
A. Speak slowly and deliberately in a normal tone of voice, using normally accepted equipment nomenclature and phonetic alphabet designations where needed to preclude potential confusion.
B. When announcements of abnormal or emergency conditions are made, they shall be made at least twice.
C. When announcing drill exercises, precede and follow all communications with, "This is a drill."
The following activities require a PA announcement:
A. Planned starting or tripping of large or loud equipment.
B. Change of plant status, i.e., unit scram/trip.
C. To summons operators or other plant personnel as needed.
D. Ventilation alignment changes which may temporarily affect access and egress to plant areas.
3.10.2 Radios Radio usage shall not be allowed in areas where electronic interference with plant equipment may result. Areas where radio use is prohibited shall be posted.
When using radios:
A. Identify yourself and your watch station.
B. Utilize three-way communication technique, normally accepted equipment nomenclature, and phonetic alphabet designations where appropriate.
C. Use professional language.
D. If signal breakup is experienced, relocate or use alternate method of communication.
E. When announcing drill exercises, precede and follow all communications with, "This is a drill."
3.10.3 Telephone Telephones are the preferred method of remote communication. When using telephones:
A. Identify yourself and your watch station.
B. Utilize three-way communication technique, where appropriate.
C. Use professional language.
D. When announcing drill exercises, precede and follow all communications with, "This is a drill."
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 33
- 33. 2.3.10 001 u-" A Control Room Isolation was automatically initiated during normal power operations.
Plant Conditions:
- Control Bldg Emergency Air Cleanup Fan A-A is running.
- Control Bldg Emergency Air Cleanup Fan B-B is running.
- Control Bldg Emergency Pressurization Fan A-A is running.
- Control Bldg Emergency Pressurization Fan B-B is NOT running.
- Spreading Room Supply Fan is NOT running.
- Spreading Room Exhaust Fans A and B are NOT running.
- Locker Room Exhaust Fan is NOT running.
FCV-311-105A (MCR fresh air inlet) is open.
FCV-311-106A (MCR fresh air inlet) is open.
Which one of the following is correct for this conditin?
,/A. Close 0-FCV-311-105A and 106A to limitwpersonnel exposuret4e-o5--e wholue body. '
B. 0-FCV-311-105A and 106A should-be-Gpen to ensure control room will remain habitable.
C. Spreader Room Supply, Spreader Room Exhaust Fans, and the Locker Room exhaust fans should be running to ensure control room will remain habitable.
D. Stop Control Bldg Emergency Pressurization Fan A-A since Control Bldg Emergency Air Cleanup Fan A-A is running maintaining Control Room environment.
A. Correct per references.
B. Incorrect, MCR fresh air inlets should~e closed.
C. Incorrect, these fans should be stopped.
D. Incorrect, at least one emergency air pressurizing fan should be running.
K/A[CFR]: 2.3.10 [2.9/3.3] [43.4] C9/,t6 e
Reference:
0-SO-30-2 R9 Tech Specs bases B3/4..7.7. 1._& CQ .*L
'J 'le4 LP/Objective: OPL271CBVENT B.11 History: New question.
Level: Comprehension Comments: FHW 12/02 2.3.10*
PLANT SYSTEMS BASES 3/4.7.5 ULTIMATE HEAT SINK (UHS)
The limitations on UHS water level and temperature ensure that sufficient cooling capacity is available to either 1) provide normal cooldown of the IR83 facility, or 2) to mitigate the effects of accident conditions within acceptable limits.
The limitations on the maximum temperature are based on providing a 30 R12 day cooling water supply to safety related equipment without exceeding their design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Plants", March 1974.
The limitations on minimum water level are based on providing sufficient flow to the ERCW serviced heat loads after a postulated event assuming a time dependent drawdown of reservoir level. Flow to the major transient heat loads R83 (CCS and CS heat exchangers) is balanced assuming a reservoir level of elevation 670. The time-independent heat loads (ESF room coolers, etc.) are balanced assuming a reservoir level of elevation 639. IBR-10 3/4.7.6 FLOOD PROTECTION This specification is deleted.
3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for
. continuous duty rating for the equipment and instrumentation cooled by this "f system and 2) the control room will remain habitable for operations personnel luring and following all credible accident conditions. The OPERABILITY of this
- , system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50. ANSI N510-1975 will be used as a procedural guide for surveillance testing.
October 6, 1999
)EQUOYAH UNIT 1 B 3/4 7-4 Amendment No. 8, 79, 247
SON 0-SO-30-2 CONTROL ROOM ISOLATION Rev: 9 0 Page 7 of 20 Date 8.1 Emergency Mode Control Room Isolation (Continued)
[7] ENSURE either Control Building Emergency Air Cleanup fan RUNNING and associated fan inlet OPEN:
CONTROL BLDG EMERGENCY AIR RUNNING FAN INLET OPEN CLEANUP FAN 4 4 A 0 0-FCO-311-9 El B El 0-FCO-311-11 El
[8] ENSURE at least one Emergency Air Pressurizing Fan RUNNING and associated fan inlet OPEN:
CONTROL BLDG RUNNING FAN INLET OPEN EMERGENCY 4 4 PRESSURIZING FAN A El 0-FCO-311-108 El B El 0-FCO-311-109 El
[9] ENSURE MCR and Spreading Room Fresh Air Fans STOPPED:
[a] Spreading Room Supply Fan. El
[b] Spreading Room Exhaust Fan A.
El
[c] Spreading Room Exhaust Fan B.
El
SON O-SO-30-2 CONTROL ROOM ISOLATION Rev: 9 0 Page 8 of 20
/
Date 8.1 Emergency Mode Control Room Isolation (Continued)
[10] ENSURE MCR and Spreading Room Fresh Air Dampers CLOSED:
CLOSED C S DAMPER DESCRIPTION 0-FCV-311-105A MCR fresh air Dl 0-FCV-311-106A MCR fresh air El 0-FCV-311-105B Spreading room fresh air LI 0-FCV-311-106B Spreading room fresh air El 0-FCO-311-79 Spreading Room Exhaust Fan A outlet 0-FCO-311-80 Spreading Room Exhaust Fan B Li outlet 0-FCO-311-17 Spreading room supply discharge El 0-FCO-311-102 Spreading room supply discharge LI
[11] ENSURE Locker Room Exhaust Fan STOPPED. El
[12] ENSURE Locker Room Exhaust Dampers CLOSED:
[a] [0-FCO-311-1031,T oilet and Locker Room Exhaust Fan Discharge. El
[b] [0-FCO-311-1041,Toilet and Locker Room Exhaust Fan Discharge. El NOTE Battery Room Exhaust Fans are started and stopped, via their respective breakers on the 480V C&A Vent Boards.
[13] IF one Electrical Board Room AHU in service, THEN ENSURE one of the following Battery Room Exhaust Fans RUNNING:
[a] Battery Room Exhaust Fan A. [C&A Vent Board 1A1-A/l12A] 0
[b] Battery Room Exhaust Fan B. [C&A Vent Board 1B1-B 11 E]
E]
[c] Battery Room Exhaust Fan C. [C&A Vent Board 261-B 11E]
Wednesday, October 09,2002 @ 11:45 AM HLC12-02.BNK Page: 39
- 38. AOP-C.04 001 S- Given the following plant conditions:
-A fire occurs in the cable spreading room while both units are at 100% power.
- The operating crew places HS-13-204 and 205, Train A and B MOV Shunt Trip, to the TRIP position.
Which ONE /) of the following explains why this action is performed?
VA. Remove power from cgrtain critical valves to prevent inadvertent operation.
B. Initiate the fire suppression equipmbnt (CO2) to the cable spreading room. -_-LP C. Remove control power from certain critical valves to prevent operation either from the main control room or from the switchgear.
D. Trip the spreading room ventilation equipment to compartmentalize the fire and to keep it from spreading.
A. Correct per reference.
B. Incorrect per reference.
C. Incorrect per references.
- D. Incorrect per reference.
2-47W611-74-1 shows that the shunt trip breaker will "stop" valve motor when in "normal" but will not when in "auxiliary," therefore distractor "C" is incorrect.
KIA[CFR]: 067 AA2.17 [3.5/4.3] [43.5]
Reference:
AOP-C.04, page 5; 1-47W611-74-1 Objective: OPL271 C423, B.1 Level: Memory History: Procedure bank Comments: FHW 12/02 067 AA2.171,4 "4
AOP-C.04 SQN SHUTDOWN FROM AUXILIARY CONTROL ROOM Rev. 5 STEP I ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Control Room Abandonment (cont'd)
NOTE The following step trips shunt trip breakers for thermal barrier isolation valves on both units and various ERCW and CCS valves.
- 6. ENSURE the following handswitches placed in TRIP: [1-M-15]
O-HS-13-204 0
- 7. ANNOUNCE"Unit_ Reactor trip, abandoning the Main Control Room" USING PA System.
NOTE Radio use is allowed in ACR during an emergency.
- 8. ENSURE the following items are taken to Auxiliary Control Room when Main Control Room is evacuated:
- flow prints
- radios
- 9. EVACUATE Main Control Room on affected unit(s).
Page 6 of 174
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 42 Wednesday, October 09, 2002 @111:45 AM HLC12-02.BNK Page: 42
- 41. AOP-M.01-B.3 002
- With a rupture of the ERCW return header in the Auxiliary Building, during normal full power operation, how long may the CCP and SI pumps operate without experiencing bearing failure after the loss of ERCW cooling?
A. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. 45 minutes
,/D. 10 minutes A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
K/A[CFR]: 2.4.20 [3.3/4.0] [41.10]
Reference:
AOP-M.01 section 2.3.
LP/Objective: OPL271AOPM01, B.3 History: Procedure bank, old Bank Number B-1090 Level: Memory Comment: FHW 12/02 2.4.20 AV? fl4t~
e olqo f/Awe 1.D50 0ý
AOP-M.O6 SON LOSS OF ESSENTIAL RAW COOLING WATER Rev. 6 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.3 ERCW Supply Header 1B Failure to Auxiliary Building CAUTION: During operation, CCP and SI Pumps may experience bearing failure 10 minutes after loss of ERCW cooling.
- 1. DISPATCH personnel to locate failure.
- 2. DISPATCH operators with radios to PERFORM:
Appendix F, Rx MOV Board Appendix R ERCW Valves
[Aux Bldg, 749' elev, Rx MOV Boards].
- 3. ENSURE 1A CCP RUNNING.
- 4. STOP and LOCK OUT the following:
" 1BBCCP
"* IBSIPump
- 5. START additional Lower Compartment Cooling Fans and CRDM Fans as required to maintain containment temperature.
Page 11 of 130
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 59 Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 59
- 57. D/G-B.6 012
> Diesel Generator 1A-A was emergency started from the backup control room during an evacuation of the main control room. Which ONE (lof the following conditions would result in a trip of the Diesel Generator 1A-A?
VA. Diesel Generator 1A-Atube oil pressure an B. Diesel Generator lA-A phase imbalance relay actuates.
C. Diesel Generator 1A-A reverse power relay actuates.
D. Diesel Generator 1A-A neutral overcurrent relay actuates.
A Correct while emergency start switches on O-M-26 and O-L-4 are held in emergency start position, all engine trips are defeated except for overspeed.
When the switches are released to normal, the emergency-start relay is reenergized and all engine protection is restored.
B. Incorrect, the only generator protection is differential overcurrent.
C. Incorrect, the only generator protection is differential overcurrent.
D. Incorrect, the only generator protection is differential overcurrent.
K/A[CFR]: 064 K4.02 (3.9-4.2) {41.7) 068 AK2.07 [3.3/3.4] (41.7)
Reference:
0-SO-82-1 0-45N767-2 LP/Objective: OPL271C065, b.6 History: System bank old Bank Number PL-1426 Level: Memory FHW 12/02 068AK2.07V' M4' Comment:
/Iwlý 44WWý
SON DIESEL GENERATOR 1A-A 0-SO-82-1 0 Rev 18 Page 74 of 83 APPENDIX A Page 1 of 1 STANDBY MODE PARAMETERS LIMITS PARAMETER INSTRUMENT MIN MAX INITIALS ENGINE lAl1 Lube Oil Temperature to Engine 0-TI-82-5010/1 85°F 125°F Lube Oil Crankcase Level Dipstick Low Mark High Mark Lube Oil Circulating Pump -- Running Circulating Lube Oil Pressure 0-PI-82-5016/1 10 psig -
Disconnect Jacket Water Immersion Heater Switch Heater ON (1)
Jacket Water Temperature to Engine 0-TI-82-5006/1 100°F 125 0 F Cooling Water Expansion Tank Level 0-LI-82-5004/1 Min Stolp Max Stop Mark Mark I Woodward Governor Oil Level Sightglass Low Mark -
-1 Day Tank #1 Level 0-LI-1i8-61/1 250 gals (2) 500 gals ENGINE 1A2 Lube Oil Temperature to Engine 0-TI-82-5008/1 85°F 125OF Lube Oil Crankcase Level Dipstick Low Mark High Mark Lube Oil Circulating Pump -- Running Circulating Lube Oil Pressure O-PI-82-5015/1 10 psig -
Disconnect Jacket Water Immersion Heater Switch Heater ON (I)
Jacket Water Temperature to Engine 0-TI-82-5003/1 100°F 125 0F Min Stop Max Stop Cooling Water Expansion Tank Level 0-L-82-5001/1 Mark Mark Woodward Governor Oil Level Sightglass Low Mark -
Day Tank #2 Level 0-L1-18-76/1 250 gals (2) 500 gals Seven Day Tank Level 0-LI-18-38 4.7 ft. (2) 5.1 ft.
(1) Not required for operability if lube oil temperature is greater than 85°F.
(2) Tech Spec limit. Notify Unit SRO if level is approaching or is below the limit.
K
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 62
- 60. E01 2.2.25 001 Keep * ? - c4 9 se
- Plant Conditions:
- CCP "A" is tagged for motor replacement.
- A small RCS leak occurs with RCS pressure dropping to 1950 psig.
The procedure reader s 'eents4whiereading E-0 step four and4o ccnzornzd
.-abeo e only one train of ECCS4u4§-&k-. ,,v,,..aL The procedure reader used his own judgement and transitioned to ES-0.0.
Which one of the following is correct concerning his actionsaand concerns?
bases Ei..
. ,"r/4 B/A.His transition was incorrect &6TS bases for ECCS is based on a LOCA.
B. His transition was correct kid-S bases for ECCS is based on a LOCA.
D. His transition was incorrect. TS bases for EGGS is based on a SG tube rupture. /'
D. His transition was correctip~dTS bases for ECCS is based on a SG tube rupture. ,,
A. Correct, he should not go to ES-0.0 since there was no SI (RCS pressure dropped to 1950 psig), and TS bases for ECCS is a LOCA.
B. Incorrect, he should not go to ES-0.0 since there was no SI, and TS bases for ECCS is a LOCA.
C. Incorrect, he should not go to ES-0.0 since there was no SI, and TS bases for ECCS is a LOCA.
D. Incorrect, he should not go to ES-0.0 since there was no SI, and TS bases for ECCS is a LOCA.
K/A[CFR]: E01 2.2.25 2.5/3.71 [43.2)
Reference:
EPM-4 section 3.11.5 and TS bases 3/4.5.2 and 3/4.5.3.
LP/Objective: OPL271 C266 B.3 History: New question.
Level: Comprehension Comments: FHW 12/02 EO1 2.2.25/
vI CLle-,4
3.11.5 Use of ES-0.0, Rediagnosis A. ES-0.0, Rediagnosis, is unique among the EOPs in that it has no specific transition into it. It is entered strictly based on operator judgment and is applicable only if SI is in progress and E-0 has already been performed.
B. ES-0.0 should be used when the operator has any concern that he may not be in the right EOP based on plant conditions. This is most likely to happen if multiple accidents occur either simultaneously or sequentially.
C. Once entered, ES-0.0 will either transition the operator to ECA-2.1, E-1, E-2, or E-3, or will return him to the procedure and step in effect, depending on diagnostics done within the procedure.
D. If ES-0.0 determines that an operator should be in a certain series of procedures (e.g., E-1 or ECA-1 series), and he is, then he simply returns to the procedure and step in effect.
E. If ES-0.0 determines that an operator should be in a certain series of procedures (e.g., E-3 or ECA-3 series), and he is NOT, then he is sent to either E-1 (if he should be in E-1 or ECA-1 series) or E-3 (if he should be in E-3 or ECA-3 series) to enter the appropriate series at the beginning and work his way through the series normally from that point on.
/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS ensures that a [R144 The OPERABILITY of each cold leg injection accumulator forced into the reactor sufficient volume of borated water will be immediately the specified pressure of core in the event that the RCS pressure falls below this condition the accumulators. For the cold leg injection accumulators, IR144 occurs in the event of a large or small rupture.
and pressure ensure The limits on accumulator volume, boron concentration safety analysis are injection in the that the assumptions used for accumulator nitrogen cover pressure met. The limits in the specification for accumulator The cold leg BR include instrument uncertainty.
are analysis limits and do not condition for operation, 13 accumulator volume (level) values in the limiting limits bound the TS 3/4.5.1, are the operating limits. The analysis limits with instrument uncertainty applied. The minimum boron operational R159 remain subcritical during the concentration ensures that the reactor core will phase based upon the cold post-LOCA (loss of coolant accident) recirculation sump mixture concentration.
leg accumulators' contribution to the post-LOCA are considered to be The accumulator power operated isolation valves which requires that Std. 279-1971, "operating bypasses" in the context of IEEE whenever permissive be removed automatically bypasses of a protective function accumulator isolation valves conditions are not met. In addition, as these criteria, removal of power to the valves is fail to meet single failure required.
for any reason 7 The limits for operation with an accumulator inoperable exposure of the R196 not within limits minimizes the time
- /z*xcept boron concentration with failure of an additional plant to a LOCA event occurring concurrent cladding temperatures. Under accumulator which may result in unacceptable peak is not available and these conditions, the full capability of one accumulator to place the reactor in a mode where this capability prompt action is required not is not required. For an accumulator inoperable due to boron concentration boron limits for operation allow 72 hours to return within limits, the concentration to within limits. This is based on the availability of ECCS water not being affected and an insignificant effect on core subcriticality during reflood because boiling of ECCS water in the core concentrates boron in the saturated liquid.
3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS ensures that The OPERABILITY of two independent ECCS subsystems available in the event of capability will be sufficient emergency core cooling any single failure a LOCA assuming the loss of one subsystem through the accumulators consideration. Either subsystem operating in conjunction with cladding to limit the peak is capable of supplying sufficient core coolingpostulated break sizes ranging temperatures within acceptable limits for all cold leg pipe downward. In from the double ended break of the largest RCS core cooling capability in the addition, each ECCS subsystem provides long term period.
recirculation mode during the accident recovery January 25, 1999 B 3/4 5-1 Amendment No. 155, 140, 192 YAv - TUNTT
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 64
- 62. E15 EK2.2 001
'~ Plant Conditions:
- The unit is recovering from a LOCA.
- Containment pressure peaked atr2# sid. 7
- RWST level is 3%.
- Containment sump level is 60%.
All actions due to initial con ito have been completed. Which one of the following best describes the core heatt moval?
A. Core heat is removed through the break and by ECCS injection from the RWST.
,/B. Core heat is removed throug tho*broaand the RHR heat exchangers.
C. Core heat is removed through the break and containment spray heat exchangers.
D. Core heat is removed through the RHR and containment spray heat exchangers.
A. Incorrect, the RWST is 3% and sump level is 60% therefore sump swapover should have occurred.
B. Correct, heat removal is from the break and from the initial conditions the operators should have performed ES-1.3 therefore RHR heat exchangers are removing core heat.
C. Incorrect, the containment pressure never reached 2.81 psid which requires containment sprays.
D. Incorrect, the containment pressure never reached 2.81 psid which requires containment sprays.
K/A[CFR]: E15 EK2.2 [2.7/2.9] [41.7]z
Reference:
ES-1.3 -st%
EPM-3-ES-1.3 step 5 LP/Obeective: 0PL271C388 B.2"" $
History: New question.
Level: ComprehensioniC2o Comments: FHW 12/02 E15 EK2.2 eon-,
~ ~ 404 ~\Ž r~vý
',ftA AAA 4w L~ atr 5 fo
TRANSFER TO RHR CONTAINMENT SUMP ES-1.3 SQN Rev. 9 1.0 PURPOSE This procedure provides the necessary instructions for transferring the safety injection system and containment spray system to the recirculation mode.
2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 ENTRY CONDITIONS E-1 Loss of Reactor or Secondary Coolant:
- RWST level less than 27%.
ECA-2.1 Uncontrolled Depressurization of All Steam Generators:
- RWST level less than 27%.
Other EOPs:
- When RWST level reaches the switchover setpoint.
3.0 OPERATOR ACTIONS 2 of 15
SQN EOI BASIS DOCUMENT FOR ES-1.3 EPM-3-ES-1.3 PROGRAM TRANSFER TO RHR CONTAINMENT SUMP Rev. 3 MANUAL Page 12 of 39 EOP Step Number: 5 ESTABLISH CCS to RHR heat exchangers:
ERG Step Number: 2 Verify CCW Flow To RHR Heat Exchangers:
Purpose:
To ensure CCS flow to cool the recirculation fluid.
ERG Basis:
This step assumes that the RHR heat exchangers are used for heat removal during the post-accident recirculation phase and that either CCS flow has been automatically provided to the heat exchangers or the operator has manually established CCS flow prior to the switchover alarm. If CCS flow had not previously been established, then it should be established at this time.
Knowledge:
~'* If CCS cannot be established to one heat exchanger, the remaining guideline can be performed as listed provided that the uncooled recirculation fluid temperature and pressure do not exceed equipment design conditions.
EOP Basis:
Same.
Deviation:
Made high level step ESTABLISH versus the ERG's AER:VERIFY/RNO:ESTABLISH arrangement.
Justification:
Improves readabililty and is more direct when implementing plant-specific details.
Setnoint:
Identifier: <S07>
Description:
Component Cooling System rate for flow to RHR heat exchanger.
1__ý
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 71
- 69. EPM-4-B.7 001
<-> Given the following events and conditions:
Unit 1 was conducting control rod drop tests during a plant startup at 2% reactor power when a complete loss of 'A'Train CCS occurred.
- Control room operators enter AOP-M.03 (Loss of Component Cooling Water)
- RCP Thrust Bearing temperature annunciator actuates.
- The operators manually trip the reactor but the trip breakers fail to open..
- Reactor power has increased to 5%
- Pressurizer pressure = 1930 psig Which ONE of the following statements correctly describes the proper procedural flow path fo rthese conditions?
A. Remain in AOP-M.03, trip all RCPs and commence a reactor shutdown.
B. Implement FR-S.1 (Nuclear Power Generation/ATWS) concurrently with AOP-M.03.
C. Terminate AOP-M.03, enter E-0 (Reactor Trip or Safety Injection) and immediately transition to FR-S.1.
/D. Enter E-0 and immediately transition to FR-S.1 while continuing on in AOP-M.03 as time and conditions permit.
A. Incorrect, per reference.
B. Incorrect, per reference.
C. Incorrect, per reference.
D. Correct, per reference.
K/A [CFR]: 2.4.5 [2.9/3.6] [41.10 43.5]
References:
EPM-4 LP/Objectives: OPL271C266, b.7 History: Procedure bank Level: Comprehension Comments: FHW 12/02 2.4.5/
3.11.6 Use of AOPs Within the EOP Network A. EOPs have priority over AOPs at all times, except when a reactor trip or safety injection has occurred in conjunction with an Appendix R fire (AOP N.08), Control Room abandonment (AOP-C.04), or Loss of all ERCW capability (AOP-M.01).
B. AOP performance while in the EOP network is allowable under the following two circumstances: [0.1]
- 2. AOP performance is deemed necessary by the SM or US to address abnormal plant conditions NOT directly addressed by the EOPs but which have a significant impact on the ability of the EOPs to perform their function (e.g., loss of ERCW, CCS, off-site power, vital instrument power board, etc.) In this case, the following guidelines should be followed:
- a. Concurrent performance of the EOPs and the AOP should enhance, NOT degrade, the performance of EOPs in progress.
- c. The AOP should be performed using the "reader-doer" method so the procedure reader remains dedicated to the EOPs in progress, which are mitigative in nature. The SM may elect to deviate from this requirement when in ES-0.1.
- d. Certain AOPs may be required to be performed concurrently with the EOPs in order for the EOPs to function as intended; for example, loss of CCS, loss of ERCW, loss of air or vital power to equipment important to safety-- any of these could have a significant impact on the ability of the EOPs to achieve their goals.
- e. Upon transition to ES-0.1, the SM will designate the mitigating crew responsibilities as appropriate, based on the events in progress.
For example, the procedure reader and OATC might perform an AOP while the CRO performs ES-0.1 as a reader-doer.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 80
- 78. FISSION*PROD*POISON 052 ,
SWhich of the following enemnees can cause Feaetorpower to fluctuate between the top and bottom ofte core when steam demand is constant? AA> ( ,,-- ,*A4ou A. steam generator level oscillations. /
B. iodine spiking./
,/C. xenon oscillations./
D. inadvertent boron dilution.
A. Incorrect per reference.
B. Incorrect per reference.
C. Correct per reference.
D. Incorrect per reference.
K/A [CFR]: 015 A2.03 [3.2/3.5] [41.5 43.5]
References:
OPL271 C228 LP/Objective: OPL271 C228 B.38 History: General Fundamental Bank Level: Memory Comments: FHW 12/02 015 A2.03
- C
OPL271C228 Revision 0 Page 51 of 146 X. LESSON BODY: INSTRUCTOR NOTES M. Xenon Redistribution Following a Step Load Decrease Objective B.38 from 100% Power to 50% Power
- 1. Prior to the power reduction, the axial Xenon Show CP-68 distribution is relatively flat due to the magnitude of CP-69 the flux and the saturation of equilibrium Xenon.
- 2. When the step load decrease occurs, control rods move into the core to decrease power. This causes an overall reduction in the magnitude of the flux and also causes the axial power distribution to shift to the bottom of the core.
- 3. Initially there is no change in the axial Xenon or iodine distributions. However, as Xenon starts to buildin following the power reduction, the control rods start to step out to compensate for the negative reactivity insertion due to Xenon. This causes the axial power distribution to shift to the top of the core.
- 4. After 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, control rods are almost fully withdrawn due to Xenon buildin. The axial Xenon distribution has not changed significantly but the absolute magnitude is increasing. The axial iodine distribution has not changed significantly but the absolute magnitude has decreased. This occurs because the axial power distribution has been constantly changing since the power reduction (shifting toward the top of the core due to rod motion).
- 5. After 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the overall Xenon concentration is decreasing causing control rods to insert to compensate. Flux is high in the top of the core for
> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> which is now causing Xenon to burnout in the top of the core and buildin in the bottom. Iodine is beginning to buildin in the top and decay out in the bottom.
- 6. After 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the iodine increase in the top of the core is causing Xenon to buildin in the top and the iodine decrease in the bottom is causing Xenon to decay out in the bottom. This coupled with rod insertion due to overall Xenon reduction causes the power distribution to begin to shift to the bottom of the core.
OPL271C228 Revision 0 Page 52 of 146 X. LESSON BODY: INSTRUCTOR NOTES
- 7. After 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, the Xenon changes in the top and bottom of the core and the rod insertion cause flux to peak in the bottom of the core.
- 8. After 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, flux has been relatively stable and peaked in the bottom of the core to cause iodine to begin to buildin in the bottom and decay out in the top. Xenon is now building in in the top and burning out in the bottom. This is causing flux to peak even higher in the bottom of the core.
- 9. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, iodine is peaking in the bottom of the core while Xenon is peaking in the top. This is causing flux to peak even higher in the bottom of the core.
- 10. Beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, rod motion is not significant because Xenon has built in to > 75% of its equilibrium value. Flux will begin to shift back to the top of the core as Xenon and iodine once again begin to build and burn in opposite ends of the core.
The result is an axial Xenon oscillation that may or may not converge.
N. Xenon Transients
- 1. Rod motion is the most significant contributor to axial Objective B.39 Xenon oscillations although temperature changes also contribute.
- 2. Rod insertion for whatever reason (runback or rod motion at steady state) will cause flux to shift to the Show CP-70 bottom of the core (provided rods are initially above the midplane of the core).
- 3. If rods are allowed to stay inserted for a significant period of time (>_2 hours), Xenon will start building up in the top and burning out in the bottom.
- 4. Xenon buildup in the top and burn out in the bottom will cause axial flux to shift even further to the bottom of the core. This could potentially cause hot channel factor and DNB problems in certain areas of the core.
If rod motion is occurring at the same time, this will cause the flux to be even higher in the bottom of the core.
1/2>
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 108 101. OPL271C368.3 001 Unit One is taking actions to mitigate a loss of offsite power. Fifteen (15) seconds after 1A-A 6.9kv shutdown board was energized by the 1A-A D/G the Operator noticed the 1A-A CCP was NOT running and the 1A-A AFW pump was NOT running.
What should be the Operator's next actionr /A. tiA4v11
,/A. Manually start the 1A-A CCP, then monitor the 1A-A AFW pump for auto start.
B. Manually start the 1A-A CCP and the 1A-A AFW pump.
C. Nothing, the 1A-A CCP and the 1A-A AFW pump load sequencers have NOT timed out for starting.
D. Nothing, the 1A-A CCP and lA-A AFW pump sequencers function only with a black out and SI signal combined.
A. Correct, the CCP should have auto started and the AFW pump sequencer has not timed out for auto start.
B. Incorrect, the AFW pump sequencer has not timed out for auto start.
C. Incorrect, the CCP sequencer should have auto started.
C. Incorrect, the CCP and AFW sequencer will function for a black out only as well as for a black out and SI combined.
K/'A{CFR) 056AK3.01 [3.5/3.9] {41.5 41.10}
064K4.10 [3.5-4.0] 141.7)
References:
AOP-P.1 appendix B LP/Objective: OPL271C368 B.4 (Q ,,{
History: Developi Bank H t1/ 4'o,"
Level: Mem6d:i. 4tf 0$
Comments; FHW 12/02 056 AK3.01/
Page 1 of 1 APPENDIX B LOSS OF OFFSITE POWER DIESEL GENERATOR LOAD SEQUENCE [C.2]
NOTE Diesel generator is rated at 4400kW continuous or 4800kW for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in 24-hour period.
EQUIPMENT NAME LOADED TIME IN SECONDS (1)
Miscellaneous loads BO or SI with BO 0 BO or SI with BO 2 CCP SI with BO only 5 SI Pumps SI with BO only 10 RHR Pumps BO or S1 with BO 15 ERCW Pumps BO or SI with BO 20 AFW Pumps BO or SI with BO 20 Thermal Barrier Booster Pumps BO or SI with BO 30 CCS Pumps Pressurizer Heaters BO only 90 (A-A & B-B only)
CS Pump Phase B with BO 180 Main Control Room AHU Phase B with BO(3) 220(')
Electric Board Room AHU Phase B with BO(3) 240(')
(1) Time is measured from the time of closing the breaker connecting the D/G to the power train.
(3) Time delay only if phase B isolation is present.
Page 50 of 86
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 112 105. PI-B.10 001
-> Which ONEXPof the following radiation detectors will initiate a Containment Ventilation Isolation Signal on high radiation level?
A. Lower Containment Particulate Detector.
,/B. Containment Purge Radiation Detector.
C. Containment High-Range Area Detector.
D. Spent Fuel Pit Continuous Air Monitor Detector.
A. Incorrect per reference.
B. Correct per reference.
C. Incorrect per reference.
D. Incorrect per reference.
KIA[CFR]: 029 A1.02 [3.4/3.4] [41.5]
Reference:
0-SO-30-3 LP/Objective: OPL271 C013 B.4 History: System bank, old Bank Number PL-0280 Level: Memory Comments: FHW 12/02 029 A1.02j,/
SON CONTAINMENT PURGE SYSTEM O-SO-30-3 OPERATION Rev: 23 1,2 Page 73 of 115 Unit. Date.
8.3 Containment Ventilation System Isolation Actuation NOTE 1 If the containment vent system is not in service and any of the dampers associated with the system are open, a containment vent isolation signal will close them.
NOTE 2 The following signals will actuate a containment ventilation isolation:
A. Safety Injection Signal B. Phase "A"Containment Isolation C. Phase "B" Containment Isolation D. High radiation detected by containment purge air exhaust radiation monitor RM-90-130 or RM-90-131.
[1] WHEN a CVI actuation has occurred by one of the above signals, THEN DISPATCH an operator locally to stop the pumps on RM-90-106 and RM-90-112 to prevent the pumps from possibly burning up while running with the isolation valves closed.
-r Page: 115 Wednesday, October 09, 2002 @ 11:46 AM>>>
HLC1202.BNK 108. PZR.PRESS-B.9 006
< .. *~ the following plant conditions:
Unit 2 is operating at 100% power.
manual and closed.
PZR level at 60% and both PZR spray valves in MIG is investigating erratic responses.
reduction.
A main turbine control failure results in a rapid load rise rapidly.
RCS temperature, PZR level, and PZR pressure The RO stabilizes RCS pressure at 2300 psig by manually opening one spray valve.
period of time.
Pressure is held constant at 2300 psig for an extended
- The RO then observes that PZR level is 68%.
to
- Pressurizer pressure master pressure controller output has increased 85%.
- PCV-68-340 and PCV-68-334 are closed.
- The "Pressurizer Level High / Low" alarm is NOT lit.
- Backup heaters are on.
Pressure &
Which ONE kKof the following describes the status of the Pressurizer Level Control systems?
VA. Functioning properly.-4w ý'
be open.
B. Malfunctioning because PCV-68-334 and 340 should should be in alarm.
C. Malfunctioning because "Pressurizer Level High\Low" be de-energized.
D. Malfunctioning because the backup heaters should A. Correct per reference.
B. Incorrect the 2335 psig bistable logic is not made.
5 percent of span deviation C. Incorrect the alarm setpoint is 70% span increasing, but less than 70%.
below program level. 68% level is above program level PZR level is more than 5%
D. Incorrect, the backup heaters should be on because program level is 60%.
of span above level program. For 100% power PZR
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 116 108. PZR PRESS-B.9 006 K/A {CFR}: 027 G 2.1.7 2.1.31 [4.2/3.9] [45.12]
References:
AOP-I.04 and 2-AR-M5-A (C-3), (E-4)
LP/Objectives: OPL271PZRLCS B.9 OPL271PZRPCS B.9 History: System bank Level: Comprehension Comments: FHW 12/02 2.1.31/
17 (C-3)
Source Setpoint LS-68-335D/E SER 1216 70% span increasing PRESSURIZER LS-68-335D/E - High 5 percent of span LEVEL SER 1217 deviation below level HIGH-LOW LS-68-339F/E - Low program Probable 1. RCS leak exceeding charging capacity.
Causes 2. Load transient or RCS temperature transient condition.
- 3. Charging and/or letdown flow mismatch.
- 4. Instrument malfunction for level or Tavg.
Corrective [1] CHECK pressurizer level (2-LI-68-339A, 335A, 320)
Actions [2] IF level is high, THEN ENSURE backup heaters ON.
[3] ENSURE level control system is attempting to return level to program with letdown and charging.
[4] IF level channel failed, THEN GO TO AOP-1.04, Pressurizer Instrument Malfunction.
[5] IF RCS leakage is suspected, THEN GO TO AOP-R.05, RCS Leak and Leak Source Identification.
[6] IF in MODE 4 or MODE 5 and a LOCA is identified, THEN GO TO AOP-R.02, Shutdown LOCA (MODE 4, or 5).
[7] EVALUATE Technical Specifications 3.3.1, 3.3.2 and 3.4.6.2 as applicable.
References 45B655-05A, 47B601-68-45
32 (E-4)
Source Setpoint LS-68-339E/F SER 367 5% of span above level PRESSURIZER 2-LS-68-339E/F program LEVEL HIGH BACKUP HTRS ON Probable 1. Charging and/or letdown flow mismatch.
Causes 2. Instrument malfunction of level or Tavg.
- 3. Load transient condition.
Corrective [1] CONFIRM instrumentation by CHANNEL CHECK.
Actions [2] IF instrument has failed, THEN GO TO AOP-I.04, Pressurizer Instrument Malfunction.
[3] IF instrument has not failed, THEN ENSURE level is returning to program 2-LR-68-339 with appropriate charging and letdown.
[4] IF RCS pressure Ž>2265 psig, THEN DEENERGIZE backup heater 2C. [C.1]
[5] EVALUATE Technical Specifications (3.3.1 and 3.3.2).
References 45N657-15, 451655-05A, 47B601-68-45
PRESSURIZER INSTRUMENT MALFUNCTION AOP-R.04 SQN Rev. 5 Page 1 of 1 APPENDIX H PRESSURIZER PRESSURE CONTROL r'3 C) con r*
cfl m
I 0,9 o
z ý II k 42 99 c'J 03 Z01 2
I.
(UJ(
=
C, Page 56 of 57
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 118 110. RADWASTE-B.12 006 Which ONE (,) of the following explains why the Liquid Radwaste Processing System Monitor Tank is recirculated for one hour before being sampled?
A. Eliminates thermal stratification.
v/B. Provides homogeneous mixture for sampling.
C. Minimizes accumulation of radioactive material in tank.
D. Reduces deposition of chemical precipitates (boron) in tank.
A. Incorrect per reference.
B. Correct per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A [CFR]: 068 K4.01 [3.4/4.1] [41.7]
Reference:
ODCM section 6.1.1 LP/Objective: OPL271LRW B.12 History: System bank Level: Memory*
Comments: FHW 12/02 068 K4.01---
A bfl,9.Z,&
SQN ODCM Revision 46 Page 26 of 168 Table 2.2-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM (Page 3 of 3) TABLE NOTATION d A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, by the method described in ODCM Section 6.1.1, to assure representative sampling.
e A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume or system that has an input flow during the continuous release.
The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141.
Ce-144 shall also be measured with an LLD of 5x10"6 . This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
g Releases from these tanks are continuously composited during releases. With the composite sampler or the sampler flow monitor inoperable, the sampling frequency shall be changed to require representative batch samples from each tank to be released to be taken prior to release and manually composited for these analyses.
h Applicable only during periods of primary to secondary leakage or the release of radioactivity as detected by the effluent radiation monitor provided the radiation monitor setpoint is set to alarm if activity in the stream exceeds a routine normal background, or compensatory requirements associated with applicable inoperable monitors are met.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 125 117. REFUELING-B. .G 001 Which ONE of the following identifies the requirement for RHR cooling loops while in the refueling mode with no core alterations near the hot legs?
VA. When > 23 feet in cavity only one RHR loop is required to be operable and in operation.
B. When < 23 feet in cavity only one RHR loop is required to be operable.
C. When < 23 feet in cavity two RHR loops are required to be operating.
D. When > 23 feet in cavity two RHR loops are required to be operable and one in operation.
A. Correct per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A: 2.2.26 [2.5/3.7] [43.5]
Reference:
Tech Spec 3.9.8.2 Tech Spec 3.9.8.1 Objective: OPL271 c269, b. 1.g Level: Memory History: Procedure bank Comments: FHW 12/02 2.2.26/7'
REFUELING OPERATIONS ux 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be in operation.
APPLICABILITY: MODE 6.
ACTION:
- a. With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- b. The residual heat removal loop may be removed from operation for up to 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 during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one residual heat removal loop shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 2000 gpm at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. R138 SEQUOYAH - UNIT 1 3/4 9-8 Amendment No. 13 4 April 2, 1990
REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION R16 3.9.8.2 Two independent Residual Heat Removal (RHR) loops shall be OPERABLE.*
AEZLLC fl: MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet. R217 ACTION:
- a. With less than the required RHR loops OPERABLE, imedlately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined OPERABLE per Specification 4.0.5.
Ri6
- The normal or emergency power source may be inoperable for each RHR loop.
R217 October 4, 1995 Amendment No. 12, 213 SEQUOYAH - UNIT 1 3/4 9-8a
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 132 124. RVINT-B.8 001
"*-> Technical Specification 3.4.9, Pressure/Temperature Limits curves are for limiting Reactor Coolant System cooldown rates.
Which ONEkW'of the following describes the technical specification basis for using the composite curves for limiting reactor vessel cooldown?
A. The thermal gradients produced during cooldown produce compressive stresses at the inside of the reactor vessel wall.
B. The thermal gradients produced during cooldown produce tensile stresses at the outside of the reactor vessel wall.
,/C. The cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the reactor vessel temperature at the tip of the assumed flaw.
D. The cooldown procedure is based on measurement of reactor coolant pressure, whereas the limiting temperature is actually dependent on the reactor vessel stress at the tip of the assumed flaw.
A. Incorrect because the thermal gradients produced during cooldown produce tensile stresses at the inside of the reactor vessel wall.
B. Incorrect because the thermal gradients produced during cooldown produce compressive stresses at the outside of the reactor vessel wall.
C. Correct because there is no way to measure the reactor vessel temperature at the tip of the assumed flaw.
D. Incorrect because the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the reactor vessel temperature at the tip of the assumed flaw.
K/A: 2.2.25 039 K5.05 [2.7/3.1] [41.5]
Reference:
Technical Specifications, page B 3/4 4-12 LP/Objective: OPL271 RVINT, B.5 History: New question (Developed 7/17/98).
Level: Memory ,
Comments: FHW 12/02 039 K5.05
REACTOR COOLANT SYSTEM BASES stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the l/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation.
It follows that at any given reactor coolant temperature, the delta T developed during cooldown results in a higher value of KIR at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in KIR exceeds Kit, the calculated allowable pressure durinc cooldown will he greater than the steady-state value.
The above procedures are needed because there is no direct control on heetemperature if the violated at the 1/4T of cooling therefore, rate location; is decreasedallowable pressures at various may unknowingly intervals along a D cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.
HEATUP Three separate calculations are reouired to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a I/4T defect at the inside of the vessel wall. The thermal cradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KIR for the 1/4T crack during heatuD is lower than the KIR for the 1/4T crack during steady-state conditions at the same coolant temperature. Puring heatup, especially at the end of the transient, conditions may exist such that the effects of compressive SEOUOYAH - UNIT 1 B 3/4 A-12
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 136
- 28. T/S0304.02 003 Unit One is starting up from Cold Shutdown. The 0-SI-OPS-000-004.0 procedure has acceptance criteria that requires, the primary and secondary Steam Generator metal temperatures shall be verified greater than 70°F on an hourly basis when RCS or Steam Generator pressures is greater than 200 psig and no RCP is in service.
This acceptance criteria will prevent A. rapid depressurization of the RCS and subsequent injection of non-condensible gases upon RCP start.
B. subsequent reactivity excursion on RCP start.
C. pressurized thermal shock of the reactor vessel.
,/D. pressurized thermal shock of Steam Generators.
A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
K/A [CFR]: 003 K4.02 [2.5/2.7] [41.7]
Reference:
0-GO-1 section 3.2.K 0-SI-OPS-000-004.0 Appendix A Table 1 acceptance criteria.
TS 3.7.2 and TS bases 3/4.7.2.
LP/Objective: OPL271C181 B.2 History: Develop one bank.
Level: Comprehension Comments: Modified the question to SQN LCO 3.7.2 which has different criteria.
FHW 12/02 003 K4.02 4VOV14ý '4 'ý-k/tcf dŽJ0&A Ac , ,T ,' I , n.
abeA r- A/-czý awv;ýý-
/010ýwl 07ý
LANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION R16 3.1.2 The temperatures of both the primary and secondary coolants in th I steam generators shall be greater than 70oF when the pressure of eiLher coolant in the steam generator is greater than 200 psig.
APPLICABILITY: At all times.
ACTION:
With the requirements of the above specification not satisfied:
- a. Reduce the steam generator pressure of the applicable side to less than or equal to 200 psig within 30 minutes, and
- b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural inte(Irity of the -,team g-neraitor.
Determine that the steam generator remains acceptable tor continued operation prior to increasing its temperatures above 200'F.
SURVEILLANCE REQUIREMENTS 4.7.2 The pressure in each side of the steam generator shall be determined to be less than 200 psig at least once per hour when the temperature of either, RI6 the primary or secondary coolant is less than 700F.
MAR 25.82 3/4 7-11 Amendment No. 12 SEQUOYAH - UNIT 1
PLANT SYSTEMS BASES 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure thatof the 10 CFR resultant off-site radiation dose will be limited to a small fraction includes Part i00 limits in the event of a steam line rupture. This dose also leak in the steam
-,the effects of a coincident 1.0 GPM primary to secondary tube with the generator of the affected steam line. These values are consistent assumptions used in the accident analyses.
3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.
314.7.1.6 MA.IN FEEDWATER ISOLATION, REGULATING, ANTD BYPASS VALVES Isolation of the main feedwater (MFW) system is provided when required to mitigate the consequences of a steam line break, feedwater line break, excessive feedwater flow, and loss of normal feedwater (and station blackout) accident. Redundant isolation capability is provided on each feedwater line consisting of the feedwater isolation valve (MFIV) and the main feedwater regulating valve (MFRV) and its associated bypass valve. The safety function of these valves is fulfilled when closed or isolated by a closed manual isolation valve. Therefore, the feedwater isolation function may be considered R236 OPERABLE if its respective valves are OPERABLE, if they are maintained in a The closed and deactivated position, or if isolated by a closed manual valve.
72-hour completion time to either restore, close, or isolate an inoperable valve takes into account the redundancy afforded by the remaining OPERABLE valves and the low probability of an event occurring that would require isolation of the MFW flow paths during this time period. The 8-hour completion time for two inoperable valves in one flow path takes into account the potential for no redundant system to perform the required safety function and a reasonable duration to close or isolate the flow path. Although the steam generator can be isolated with the failure of two valves in parallel, the double failure could be an indication of a common mode failure and should be to treated the same as the loss of the isolation function. The 7-day frequency verify that an inoperable valve is closed or isolated is reasonable based on valve status indications available in the control room, and other administrative controls to ensure the valves are closed or isolated.
3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed0 the maximum JBR allowable fracture toughness stress limits. The limitations of 70 F and 200 psig are based on a steam generator RT of 25 0 F and are sufficient to prevent brittle fracture.
June 8, 1998 crnrTnvr - UNT 1 fl 3/4 7-3 Amendment No. 232
SON SURVEILLANCE REQUIREMENTS 0-SI-OPS-000-004.0 PERFORMED ON INCREASED FREQUENCY Rev: 22 0 OR WITH NO SPECIFIC FREQUENCY Page 13 of 23 Unit. Date APPENDIX A Page 3 of 3
[2] IF primary or secondary side pressures are greater than 200 psig, THEN RECORD information in Table 1 once each hour until all S/G primary or secondary pressures are indicating less than 200 psig or a RCP is running. El NOTE P1-1-33 should be used only for secondary pressure indication when the MSIV's are open, otherwise, any one indicator may be used and logged in the table. This Table may be copied as needed for additional hourly record.
TABLE 1 I I I T T TIME Pressure Indicator Pressure (psig)
SECONDARY SIG#1 S&G#1 S/G#1 SIG#1 SIG#1 S/G#1 Appendix 'E' Data SIG#2 SIG#2 SIG#2 S/G#2 S/G#2 SIG#2 SG#3 S/G#3 SIG#3 SG#3 SG#3 SG#3 SIG#4 S/G#4 SIG4 S/G#4 S/G#4 S/G#4 Pressure Indicator _____
Pressure PRIMARY (psig) ____
Temperature Indicator Temperature
___________ (OF) _ _ _ _ _ _ _ _ _ _ _ _ _
INITIALS ___ ___ ___ ___
ACCEPTANCE CRITERIA: The primary and secondary Steam Generator metal temperatures shall be verified >70°F on an hourly basis when RCS or Steam Generator pressures is
>200 psig and no RCP is in service.
SON UNIT STARTUP FROM COLD 0-GO-1 SHUTDOWN TO HOT STANDBY Rev: 27 1 &2 Page 13 of 97 3.2 LIMITATIONS (Continued)
K. Prior to exceeding 200 psig in either the RCS or steam generator secondary side, perform 0-SI-OPS-000-004.0 on an hourly basis to verify temperatures greater than 70 0 F. Continue 0-SI-OPS-000-004.0 until pressure is less than 200 psig and/or an RCP has been started.
L. Greater than or equal to 200 psid is the #1 seal AP limit for RCP operation.
However, when instrument inaccuracy is taken into account the AP limit has to be increased to 220 psid to maintain the required 200 psid.
M. When the RCS temperature is less than 350 0 F, both safety injection pumps and one centrifugal charging pump shall be made incapable of automatic injection into the RCS. This is necessary to minimize the potential for and the severity of low temperature overpressurization events (TS 3.4.12).
N. Whenever the RCS temperature is above 350 0 F, at least one train of control rod drive mechanism cooling fans must be in service to maintain shroud temperature less than 164 0 F.
- 0. Plant Computer is updating if the time indication on monitor is changing. [C.4]
P. After any significant change in charging flow, the RCP seal injection flows should be checked, and adjusted if necessary, to maintain a minimum of 6 gpm to each reactor coolant pump.
Q. The MSIVs should not be stroked more often than once per 10 minutes when the valve internals are dry (no steam in pipe).
R. The applicability of SR 4.4.6.3.c needs to be evaluated anytime an RCS pressure isolation valve listed in table 3.4-1 of Tech Specs has been actuated due to automatic or manual action or flow through the valve.
S. Credit can be taken for filled steam generators only if the RCS remains pressurized above 50 psig (TS 3.4.1.4). [C.15]
Thursday, October 10, 2002 @ 05:54 AM HLC12-02.BNK Page: I
- 1. WGDS-B.12 001 Given the following:
- A waste gas decay tank release was in progress.
- RM-90-118, Waste Gas Discharge Radiation Monitor, is in service.
- 0-RA-90-1 1BA WDS GAS EFF MON HIGH RAD annunicator alarms.
- RCV-77-119, Gas Release Header Flow Control Valve, has tripped closed.
- The Chemlab has performed a backup sample analysis on the waste gas decay tank.
- A new release permit has been issued, RM-90-118 setpoint has been adjusted, and approval has been granted to resume the release.
Which ONE of the following is (are) the MINIMUM action(s) to take for RCV-77-119 (Gas Release Header Flow Control Valve) in order to continue the waste gas release?
A. Reset the valve locally and allow it to automatically re-open.
,/B. Take the valve controller to ZERO (0), then re-open the valve.
C. Purge the release line with nitrogen until RM-90-118 returns to background, reset the valve locally, then re-open the valve.
D. Place the PROCESS RADMON SYS BLOCK SW TRAIN "A" to the RM-90-118 position, take the controller to ZERO (0), then re-open the valve.
A. Incorrect per reference.
B. Correct per reference.
C. Incorrect per reference.
D. Incorrect per reference.
Student must comprehend this situation requires a new procedure for the second release and begins at step one.....-
K/A[CFR]: 071 A4.26 [3.1/3.9] [41.7]
Reference:
0-SO-77-15 section 6.0 [15] and [22].
Objective: OPL271GRW, B.12 History: Modified from 97 Summer NRC Exam Level: Comprehension Comments: FHW 12/02 071 A4.26
SON WASTE GAS DECAY 0-SO-77-15 TANK RELEASE Rev: 11 0 Page 9 of 16 Date 6.0 NORMAL OPERATION (Continued)
NOTE The control switch for 0-FCV-77-245 is 0-HS-77-245, and is located in the Unit 1 690 pipe chase, approximately 40 feet east of the BIT approximately 5' from the floor.
[14] ENSURE [0-FCV-77-2451 waste gas vent flow control valve is OPEN to the Shield Building vent indicated in step [3].
[15] ADJUST [0-FIC-77-1191 (0-FIC-77-119, on panel O-L-2A) release header flow control valve controller on panel O-L-2A to the zero setpoint.
[16] OPEN [0-77-7501 release header filter outlet isolation valve.
[17] OPEN [0-77-7491 release header filter inlet isolation valve.
[18] RECORD below the initial pressure of the gas decay tank to be released.
Tank ID (Letter) Tank Pressure I I PSIG
[19] ENSURE the tank to be released is aligned to the "Release to Vent Header Alignment" in accordance with Attachment 3 Valve Checklist 0-77-15.03.
3/4
SON WASTE GAS DECAY 0-SO-77-15 TANK RELEASE Rev: 11 0 Page 10 of 16 Date 6.0 NORMAL OPERATION (Continued)
[20] ENSURE the following interlocks satisfied for 0-FCV-77-119:
[a] Selected unit's Shield Bldg. Exhaust flow
> 3000 cfm (1, 2Y2210A). El
[b] High radiation on 0-RE-90-118 clear. El
[c] Instrument Malfunction on 0-RM-90-118 clear. El
[d] FIC-77-119 output signal at "ZERO". El
[21] IF any interlock(s) NOT satisfied, THEN TAKE action to satisfy the interlock OR INTIATE WO for corrective action. WO#
NOTE 1 When the gas release is initiated, the data requested on 0-SI-CEM-077-410.4, APP. B, should be recorded at initiation "ofthe release and at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> intervals.
NOTE 2 The following two steps will require two operators, one at panel O-L-2A, and one at panel 0-L-473.
NOTE 3 The radiation control valve 0-RCV-77-119 will close on high radioactivity as indicated on 0-RE-90-1 18. Ifthe high radiation alarm is present, then appropriate actions must be taken to clear the high radiation condition. The control signal must be zeroed before the valve will reopen.
[22] OPEN SLOWLY [RCV-77-1191 (0-FIC-77-119 on panel 0-L-2A) gas release header flow control valve, AND OBTAIN the release flow rate approved in 0-SI-CEM-077-410.4, as indicated on FE-77-230 flow indicator on panel 0-L-473 (inches H20).
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 58
- 56. D/G-B. 10 002 Given the following plant conditions:
Unit 1 & 2 are steady-state at 100% power 125V DC Vital Battery Board III is inadvertently deenergized Which ONEof the following describes the effect this has on the diesel generators?
A. All diesel generators except Diesel Generator 1A-A would auto start.
B. All diesel generators would auto start but all engine and generator trips on Diesel Generator 2A-A would be disabled.
C. Diesel Generator lA-A would auto start and could only be shutdown using the EMERGENCY STOP pushbutton on 0-M-26 until control power was restored.
VD. Diesel Generator 2A-A would auto start and could only be shutdown using the Local EMERGENCY STOP pushbutton until control power was restored.
A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
K/A {CFR}: 064000K203 [3.2/3.6]
064 A3.06 [3.3/3.4] [41.7]
References:
1,2-45N767-2 & 5 LP/Objectives: OPL271 D/G, B.9 History: HLC 9809 Audit Exam Level: Analysis Comments: FHW 12/02 064 A3.06,
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 124
- 16. RDCNT-B.7.B 001 Unit 1 was operating at 60% power. Given the following events and conditions:
-( 3TA 44r ,
- Pressurizer pressure decreased to 1940 psig.
- The SSPS train "A" low PZR pressure trip logic relay failed to actuate.
What effect would this failure have on the function of the reactor protection system?
A. The reactor would NOT trip because the Train A logic relay would not remove power from the UV coil for-RTA. %'fA-.
B. The reactor would NOT trip because the Train B logic relay would not remove power from the UV coil for-RiT iYA-.
,/C. The reactor would trip because the Train B logic relay would remove power from the UV coil for--R-B. 1YA-.
D. The reactor would trip because the Train B logic relay would remove power from the UV coil for-RTA. L1Yfý.
A. Incorrect per reference.
B. Incorrect per reference.
C. Correct per reference.
D. Incorrect per reference.
K/A {CFR): 012 K2.01 [3.3/3.7] [41.7]
References:
1,2-47W611-99-1 LP/Objectives: OPL271 RPS, b. 11 History: Systems bank Level: Comprehension Comments: FHW 12/02 012,K20TU"
&rz toY
RO/SRO/KA Cross Reference RO/SRO M=Memory Type K/AID Question S=SRO C=Comph A=As Is R=RO A=Analysis M=Modified ID B=BOTH N=New 001 AK1.03 AOP-C.01-B.1 002 B C A 001 K.301 001 K.301 B A N 001 K5.30 CONTROL*RODS 020 B M A 003 A1.03 RCP-B.12 002 B M A 004 A3.07 004 A3.07 B C N 005 A1.01 RHR-B.13.H 001 B M N 006 A1.07 ECCS-B.3 002 B M A 006 K6.19 006 K6.19 B C N 008 K4.02 CCS-B.9.A 001 B A A 009 EK1.01 INPO3 955 B C A 010 A2.02 010 A2.02 B M N 012 K6.07 012 K6.07 B C N 013 K4.01 013 K4.01 B M N 015 AK2.08 AOP-R.05-B.5 001 B M A 016 A3.02 RCS-TEMP-B.2 008 B M A 016 K1.01 INCORE-B.1.B 002 B M A 017 K6.01 INCORE-B.1.D 003 B M A 022 K2.01 022 K2.01 B C N 024 AA1.26 ES-0.1-B.1 002 B C A 025 AK2.05 FR-Z. 1-B.2 001 B C A 025 K5.02 CTMT-B.1 1 001 B C A 027 AK1.02 OPL271C353.4 001 B C A 1
RO/SRO M=Memory Type Question S=SRO C=Comph A=AsIs R=RO A=Analysis M=Modified ID B=BOTH N=New 029 EK2.06 RPS-B.5.B 001 B C A 034 A1.02 034 A1.02 B A N 034 K4.02 FH-B.12.B 001 B M A 035 K5.01 035 K5.01 B C N 037 2.1.7 AOP-R.01-B.2 003 B A A 037 AA1.11 AOP-R.01-B.2 004 B M A 038 EK3.08 INPO2 976 B C A 039 K3.04 FW-B.5 001 B A A 040 AK1.01 INPO2 231 B C A 051 AK3.01 SDCS-B.12 023 B C A 058 AK3.01 AOP-P.02-B.4 002 B C A 059 A4.01 059 A4.01 B C N 059 K1.04 MFW 001 B M A 060 AK2.02 060 AK2.02 B C N 061 AK2.01 061 AK2.01 B A N 061 K2.01 061 K2.01 B M N 061 K6.01 061 K6.01 B M N 062 A4.03 INPO8 155 B M A 063 A2.01 063 A2.01 B M N 063 K4.04 FW-B.5.B 004 B M A 064 K2.02 064 K2.02 B C N 068 A3.02 RADWASTE-B.12 008 B M A 069 AA1.03 CTMT-B.5 004 B C A B M A 071 A1.06 RMS-B.9 001 2,
( (
RO/SRO M=Memory Type Question S=SRO C=Comph A=As Is K/A ID R=RO A=Analysis M=Modified ID B=BOTH N=New 072 A1.01 RMS-B.2 004 B M A 074 EK1.01 FR-C.1-B.2 003 B C A 074 EK2.02 FR-C.1 -B.2 011 B A A 075 K1.01 RCW-B.9 001 B M A 076 AK3.05 076 AK3.05 B C N 078 K1.01 AIR-B.12 002 B C A 079 K4.01 AIR-B.5 014 B C A 103 A3.01 103 A3.01 B M N 2.1.24 2.1.24 B A N 2.2.11 CTMT-B.11 004 B M A 2.2.4 2.2.4 B A N 2.3.11 ODCM-B.5 001 B C A 2.3.9 CTMT PURGE-B.4 001 B M A 2.4.39 REP-B.1.D 002 B M A 2.4.4 AOP-R.02-B.2 001 B C A E01 EA1.1 ES-0.0-B.3 001 B C A E03 EA1.2 ES-1.2-B.2 006 B C A E03 EK1.3 ES-1.2-B.2 004 B M A E04 EK2.2 ECA-1.2-B.1 002 B M A E05 EA1.3 FR-H.1-B.3 004 B C A E08 EK3.2 FR-P.1 001 B C A E09 EA1.1 OPL271C382.4 001 B A A El0 EK3.1 ES-0.2-B.3 003 B C A E14 EK1.3 ECCS-B.2 002 B M A E15 EK1.1 FR-Z.2-B.2 001 B M A 3
C K KIA ID Question ID 003 K4.02 T/S0304.02 003 005 AA2.01 005 AA2.01 008 A2.04 008 A2.04 012 K2.01 RDCNT-B.7.B 001 015 A2.03 FISSION*PROD*POISON 052 022 2.4.27 022 2.4.27 0252.2.13 0252.2.13 029 A1.02 PI-B.10 001 039 K5.05 RVINT-B.8 001 056 A2.04 056 A2.04 056 AK3.01 OPL271C368.3 001 056 K1.03 056 K1.03 059 AA2.05 059 AA2.05 064 A3.06 D/G-B. 10 002 067 AA2.17 AOP-C.04 001 068 AK2.07 D/G-B.6 012 068 K4.01 RADWASTE-B. 12 006 071 A4.26 WGDS-B.12 001 078 K3.03 078 K3.03 0792.1.1 0792.1.1 086 K3.01 086 K3.01 2.1.16 2.1.16 2.1.31 PZR PRESS-B.9 006 2.2.26 REFUELING-B.1.G 001 2.3.10 2.3.10 4
(
RO/SRO M=Memory Type K/AID Question S=SRO C=Comph A=As Is R=RO A=Analysis M=Modified ID B=BOTH N=New 2.4.20 AOP-M.01-B.3 002 R M A 2.4.5 EPM-4-B.7 001 R C A EO1 2.2.25 E01 2.2.25 R C N E15 EK2.2 E15 EK2.2 R C N 003 DROP ROD AOP-C.01-B.5 011 S M A 2.4.4 004 2.4.1 OPL271C367.1 002 S C A 007 2.1.10 PRT-B.7 004 S M A 010 2.4.47 AOP-1.04-B.2 001 S A A 011 2.1.6 OPL271C367.1 003 S C A 013 2.4.47 E-0-B.6 003 S M A 015 AA2.07 AOP-R.04-B.5 001 S A A 022 AA2.01 AOP-R.05-B.2 001 S A A 025 AA2.06 RHR-B.12.A 003 S M A 027 AA2.04 T.S-2.1-B.2 001 S A A 040 AA2.01 OPL271C379.3 001 S A A 056 AA2.18 E-0-B.3.A 007 S C A 0692.1.14 0692.1.14 S C N 2.1.10 OPL271C458.1 001 S M A 2.1.22 GO-2-B.1 003 S M A 2.1.7 PZR LEVEL-B.14 003 S A A 2.2.25 OPL271C180.3 001 S M A 2.2.9 SPP-9.5 001 S M A 2.3.1 RCI-15 001 S M A 5
K ci (
RO/SRO M=Memory Type K/AID Question S=SRO C=Comph A=As Is R=RO A=Analysis M=Modified ID B=BOTH N=New 2.3.2 RADIATION 002 S M A 2.3.3 AOP-C.04-B.5 008 S M A 2.4.33 OPDP-4 002 S M A 2.4.45 AOP-M.03-B.1 002 S C A E01 EA2.1 ES-0.0-B.5 002 S C A E04 EA2.1 ECA-1.2-B.2 001 S C A E09 EA2.1 ES-0.2-B.3 001 S C A E12 2.4.16 ECA-2.1-B.1 004 S A A E15 EA2.1 E15 EA2.1 S M N E16 2.4.41 E16 2.4.41 S C N 6
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 135
- 27. T.S-2.1-B.2 001
"*-' This question has reference material attached.
Unit 2 was operating at 60% power when an event occurred. All 3 pressurizer safety valves lifted. T-ave peaked at 680°F and the RCS pressure transient reached 2675 psig.
The .. t4t-was-muanalyipped and RCS temperature and pressure rapidly decreased.
Which ONE (1) of the following describes the Tech Spec action applicable to this transient (prior to rods tripping)?
A. Reduce the RCS pressure to within its limit within 5 minutes.
B. Reduce the RCS pressure to within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
C. No action is required because RCS pressure never exceeded 2735 psig.
,/D. Be in Hot Standby within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
A. Incorrect because initial conditions are mode 1.
B. Incorrect because RCS pressure never exceeded 2735 psig.
C. Incorrect TS 2.1.1 was exceeded.
- d. Correct per references.
K/A {CFR}: 027 AA2.04 [3.7/4.3] [43.5]
References:
Tech Spec Safety Limit 2.1.1 and Figure 2.1-1.
LP/Objectives: OPLI271C075, B.2 History: Tech Spec bank Level: Analysis Comments:
FHW 12/02 027 AA2.04 ; Modified the stem and distractors for analysis. Provide safety limits figure 2.1-1. DO NOT provide safety limit 2.1.1 since this is a required memory one hour action.
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
- /
2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1.
IR45 -
APPLICABILITY: MODES I and 2.
ACTION:
Whenever the point defined by the combination of average temperature and THERMAL POWER has exceededthe highest operating loop the appropriate pressurizer pressure line, be in HOT STANDBY within I hour.
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.
APPLICABILITY: MODES 1, 2, 3, 4 and 5.
r ACTION:
MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.
SEOUOYAH - HI..IT I M1-1 September 3, 1985 Amendment No. 41
.Ht 4J 7[fl- a-Aa- *rIrht4.
7 .4- ---- --
-a". H rm Revision 23 E 4j) -4 IHrn £..-.ll.F - iT A PERk 121 rYYT7iTntr
~25O~t 00t u r-
- I S U 4.-"-
+t. I-- I .i m
-t r'
I
- I --- I -- I fl= i-a 2
AGCE¶ABLE; r 2flPERATIcflS
--- a ERT
=F----
M~r-,- TT
= 57 r._-
- . .. . FRCTIO-op-RA E-THERMAALPOW
- ,-~- ~,- December 23, 1982 2*SEQUOYAB UNIT 1 2- ~ . Amendment 19 __
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Pago: 134
- 26. SPP-9.5 001 Which ONE (1) of the following states responsibilities of the designated Senior Reactor Operator for normal installation of Temporary Alterations (TA) on quality related equipment?
A. Determine the need for environmental evaluation by the Environmental Section and reactivity management evaluation by Reactor Engineering.
B. Verify all required reviews and documentation have been completed, prepare he TA Tags and place the TA tags after the TA is installed. 4*,
ý*L' V/C. Review the TACF to ensure required reviews/ýnd documentation have been completed and that the TA can be performed'Without adverse effect on plant operation.
D. Perform a 10CFR50.59 evaluation on the TA to ensure it won't have an adverse affect on plant operation and review the TA to determine if training is needed prior to installation.
A. Incorrect - This is done by the Environmental Section B. Incorrect - Performed by System or Design Engineer C. Correct - per SPP 9.5 "Temporary Alterations"
, D. Incorrect - Done by System or Design Engineer K/A {CFR}: 2.2.9 [2.0/3.3] [43.3]
References:
SPP-9.5 LP/Objectives: OPL271OPSMGMTL B.11 History: Procedure bank Level: Memory Comments: FHW 12/02 2.2.9
TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 9 of 21
- 1. If a TACF is no longer needed, prior to installation, the requester voids the TACF number in the site TACF Log Book.
- 2. The requester provides the original TACF to the TACF Coordinator for canceling.
3.3 Installation NOTE Refer to Appendix C for material requirements.
A. Precautions and Limitations
- 1. Ensure lifted leads are suitably insulated and jumpers securely attached.
- 2. Verify the affected component is properly identified and in the correct circuit.
- 3. De-energize electrical circuits prior to the installation of jumpers or the removal of leads, whenever possible.
- 4. Ensure Category 1 drawings are updated and in the Control Room.
- 5. Ensure secondary drawings are listed and referenced in Curator.
- 6. Ensure materials (cables, jumpers, etc.) are equal to or better than the quality engineered into the host component or system.
- 7. If the original TACF is removed, a copy of the TACF should replace the original TACF while it is removed.
B. Concerning installation, the SM or designee is responsible for the following:
- 1. Review the TACF to ensure required reviews and documentation have been completed and that the TA can be performed without adverse effect on plant operation.
- 2. Authorize installation of the TA and maintain control of the approved TACF. Sign and record date.
C. The System Engineer shall perform the following:
- 1. Ensure the WID number is recorded.
- 2. Verify all post installation testing has been successfully completed, if required. Sign and record date.
- 3. Verify TA installation and TACF tags, if required, have been properly placed. Sign and record date.
- 4. Forward a copy of the TACF to the TACF Coordinator. Sign and record date.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 127
- 19. RHR-B.12.A 003 Which ONE PI-of the following correctly describes the operation of the RCS to RHR System Supply Valves FCV-74-1 and FCV-74-2?
A. RCS pressure > 687 psig causes auto-closure.
B. RWST suction valve 63-1 SHUT prevents opening V'C. RCS pressure ->380 psig prevents opening D. RCS temperature of > 350°F causes auto-closure A. Incorrect per reference.
B. Incorrect per reference.
C. Correct per reference.
D. Incorrect per reference.
K/A[CFR]: 025 AA2.06 [3.2/3.4] [43.5]
Reference:
0-SO-74-1 LP/Objective: OPL271 RHR B.9 History: System bank.
Level: Memory Comments: FHW 12/02 025 AA2.06 All itp
>MVw
SON RESIDUAL HEAT 0-SO-74-1 REMOVAL SYSTEM Rev: 39 1,2 Page 78 of 185 Unit Date 5.6 Startup of RHR System for Normal Cooldown Mode (Continued)
NOTE Refer to Tech Spec LCO 3.5.3.
[8] PLACE both RHR pumps in PULL TO LOCK.
[9] CLOSE the following:
VALVE NO. FUNCTION INITIALS FCV-63-1 RHR Suction from RWST RHR Pump A-A Discharge to FCV-63-93 Loops 2 and 3 Cold Leg RHR Pump B-B Discharge to FCV-63-94 Loops 1 and 4 Cold Leg
[10] ENSURE the following are CLOSED:
VALVE NO. FUNCTION INITIALS FCV-63-8 RHR Hx A to CVCS Chg Pumps FCV-63-11 RHR Hx B to SIS Pumps HCV-74-34 RHR to RWST Return Isol CAUTION Opening FCV-74-1 and 2 with the pressurizer solid may cause a large drop in RCS pressure. Do not continue this section if the pressurizer is solid with RCPs running.
NOTE 1 Pressure interlocks prevent FCV-74-1 and FCV-74-2 from being opened until RCS pressure is < 380 psig.
NOTE 2 FCV-74-1 and FCV-74-2 are administratively controlled at motor control panel. An M-5 key will be needed for unlocking breakers.
[11] UNLOCK and CLOSE breaker for [FCV-74-11 in Compt 6C2 on 480V Rx MOV Bd Al-A.
[12] UNLOCK and CLOSE breaker for [FCV-74-21 in Compt 14B on 480V Rx MOV Bd B1-B.
Wednesday, October 09,2002 @ 11:46 AM HLC12-02.BNK Page: 120 112. RCI-15 001 "Apipe elbow on the spent resin transfer line in the El 690 Pipe Chase is producing a 500 mr/hr field at 30 centimeters from the elbow.
Which ONE (1) of the following identifies the proper posting requirement for the area surrounding the pipe elbow?
,/A. CAUTION, HIGH RADIATION AREA B. DANGER, AIRBORNE RADIATION AREA C. DANGER,RLOIENTIALW-RY HIGH RADIATION AREA D. GRAVE DANGER, VERY HIGH RADIATION AREA A. Correct per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A {CFR}: 2.3.1 [2.6/3.0] [41.12/43.4]
References:
RCI-15 R12 section 6.8.
LP/Objectives: OPL271 C260 B.3 History: HLC 9809 Audit Exam Level: Memory Comments: FHW 12/02 2.3.1; modified stem for a correct answer.
RCI-15 SQN ESTABLISHING AND UPDATING Revision 12 RADIOLOGICAL SIGNPOSTINGS Page 6 of 13 6.0 REQUIREMENTS (Continued) 6.7 Radiation Area A. A Radiation Area is an area, accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess of 0.005 rem (5 mrem) in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates.
B. Each Radiation Area shall be posted with a conspicuous sign or signs bearing the radiation symbol and the words CAUTION RADIATION AREA.
6.8 High Radiation Area A. A High Radiation Area is an area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 0.1 rem (100 mrem) in one hour at 30 centimeters from the radiation source or 30 centimeters from any surface that the radiation penetrates.
B. Each High Radiation Area shall be posted with a conspicuous sign or signs bearing the radiation symbol and the words CAUTION HIGH RADIATION AREA or DANGER - HIGH RADIATION AREA.
C. In lieu of the control devices or alarm signals required by 10CFR20, each High Radiation Area in which the intensity of radiation is _Ž100 mrem/hr at 30 cm, but < 1,000 mrem/hr at 30 cm, shall be barricaded and conspicuously posted as a High Radiation Area and entrance shall be controlled by requiring issuance and use of an RWP. Any individual or group of individuals permitted entrance to such areas shall be provided with or accompanied by one or more of the following: [C.2]
- 1. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- 2. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 117 109. RADIATION 002 Which ONEfr,*f the following statements is correct concerning the SQN ALARA program?
,/A. The SQN Plant Manager must approve all lower containment entries inside the polar crane wall when the unit is in Mode 1.
B. The SQN Site Vice President must approve all lower containment entries inside the polar crane wall when the unit is in Mode 1 or 2.
C. The SQN Plant Manager must approve all lower containment entries inside the polar crane wall when the unit is in Mode 1 or 2.
D. The SQN Site Vice President must approve all lower containment entries inside the polar crane wall when the unit is in Mode 1.
A. Correct - As described in RCI-10 ALARA Program B. Incorrect - Plant Manager must approve entry only in MODE 1.
C. Incorrect - Plant Manager must approve entry only in MODE 1.
D. Incorrect- - Plant Manager must approve entry only in MODE 1.
K/A {CFR}: 2.3.2 [2.5/2.9] [41.12 43.4]
References:
RCI-10 R27 section 5.0.
LP/Objective s: OPL271C260 B.9 History: Admin bank Level: Memory-Comments: FHW 12/02 2.3.2 aJýx -ý 60 N
RCI-10 SON ALARA PROGRAM Revision 27 Page 8 of 26 5.0 GENERAL (Continued)
M. All identified rework items shall be documented in the post job review of the ALARA Planning Report and clearly indicated as a rework item.
Rework is defined as the repeat performance of a work activity because of improper performance of work, which could include but is not limited to, poor workmanship, re-machining, reassembling, poor engineering design or inadequate planning. [C.2]
N. The Plant Manager shall approve all Lower Containment entries inside the polar crane wall during Mode 1 operations. Approval will be documented on a form similar to that provided in Attachment 05, Plant Manager's Approval for Lower Containment Entries Inside the Polar Crane Wall at Power.
- 0. The RAD/CHEM Manager, or designee, shall approve all Containment Building entries during periods which are outside the pre-determined Containment Building entry schedule. Approvals shall be documented on an Attachment 06, Containment Building Entry Request/Authorization.
P. Individuals preparing work planning documents will ensure that the proposed work is evaluated for potential radiological impacts. The Radiological Impact Evaluator reviews the control document to determine if such impacts exist using Appendix C. Specific RADCON steps should be included in the applicable work plans or work orders for the job.
6.0 REQUIREMENTS 6.1 ALARA Pre-job Planning Criteria Pre-job planning shall be carried out to identify and to minimize the hazards associated with jobs involving potential significant radiological hazards identified for each job where any of the following conditions occur:
A. The whole body (30 cm) dose rate is _Ž1 rem/hr, and exposures are estimated to be greater than 50 mrem to any individual.
B. The collective total effective dose equivalent (TEDE) is expected to exceed 1 person-rem.
C. The collective extremity dose is expected to exceed 10 rem.
D. All entries into areas posted as a Very High Radiation Area.
E. At the discretion of the ALARA Coordinator.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 114 107. PZR LEVEL-B.14 003' Given the following plant conditions:
- Reactor power is 60%
- Pressurizer pressure is 2240 psig
- Charging flow is being controlled in MANUAL
- All other controls are in AUTOMATIC
- The backup heaters just ENERGIZED Which ONE (1) of the following would be the indicated pressurizer level?
A. 41%
B. 43%
C. 46%
,YD. 51%
46% is the programmed PZR level for 60% power, +5% above program is the level which energizes the backup heaters.
K/A: 2.1.7 [3.7/4.4] [43.5]
Reference:
TI-28 1-AR-M5-A (E-4)
Objective: OPL271C019, B.6 History: System bank Level: Analysis Comments: FHW 12/02 2.1.7
SYTI-28 Aft. 9 UNIT 1 & 2 CYCLE DATA SHEET Effective Date 04/,..'1.
(FOR INFORMATION ONLY) Page 14 of 16 CONTROLS TAVG PZR LEVEL 547 OF - 578.2 OF 24.7% - 60%
(0 - 100% power) (547 OF - 578.2 OF Tavg [Auct. Hi Tavg])
FEEDPUMP SPEED CONTROL S/G LEVEL 80 - 195 psid, (0 - 20 %,constant 80 psig) 33 - 44% (0 - 20% turbine load)
(20 - 100% total steam flow) 44% (20 - 100% turbine load)
ROD CONTROL FIRST-STAGE IMPULSE PRESSURE Auto: 8 spin; 1.5°F -->3.0 0F error 0 - 628 psia 8 spm -72 spin; 30F -->5°F error 0 - 100% power Manual: 48 spin Bank Select: 48 spm control rods 64 spin shutdown rods FP Rod Insertion Limit: 182 steps on "D"Bank STEAM DUMPS Blocking:
Steam Dump Bypass Interlock Switches (M-4) in OFF Condenser not available (absence of C-9)
Lo-Lo Tavg (P-12) locks out all steam dumps; interlock can be bypassed for 3 cooldown valves using the two bypass interlock switches on M-4.
Arming:
Load Rejection: 10% load decrease in a 2 minute time constant as sensed by PT-1 -72 (0-7).
Reactor Trip: P-4 from 'A' Tr. Rx Trip Breakers ('B' Tr. P-4 places Rx Trip Controller I/S.)
Mode Selector Switch (M-4) in STEAM PRESSURE Opening:
Load Reject: Tave - Tref (PT-1-73)
(20F -4 18tF = 0 -) 100% open)
Tave Mode
- Trip Open - 1/2 @ 10F; 1/2 @ 18°F Reactor Trip: Tave - 552tF (Fixed Reference Signal)
(0°F --- 50'F = 0 -- 100% open)
Trip Open - A1@ 25tF; 12 @ 50sF Pressure Mode Auto = steam pressure - setpoint Manual = Operator Controlled
32 (E-4)
Setpoint Source 5% of span above level LS-68-339E/F SER 367 program 1-LS-68-339E/F PRESSURIZER LEVEL HIGH BACKUP HTRS ON Probable 1. Charging and/or letdown flow mismatch.
Causes 2. Instrument malfunction of level or Tavg.
- 3. Load transient condition.
Corrective [1] CONFIRM instrumentation by CHANNEL CHECK Actions [2] IF instrument has failed, THEN GO TO AOP-1.04, Pressurizer Instrument Malfunction.
13] IF instrument has not failed, THEN ENSURE level is returning to program 1-LR-68-339 with appropriate charging and letdown.
[4] IF RCS pressure Ž 2265 psig, THEN DEENERGIZE backup heater 1C. [C.1]
[5] EVALUATE Technical Specifications (3.3.1 and 3.3.2).
References 45B655-05A-0, 45N657-15, 47B601-68-45
<7
Wednesday, October 09, 2002 @ 11X6 AM HLC12-02.BNK Page: 113 Wednesday, October 09,2002 @ 1JA'6 AM HLC12-02.BNK Page: 113 106. PRT-B.7 004 X' Unit 2 was operating at 90% after a start-up from a refueling outage. A PORV is found to be leaking and the PORV block valve was shut. Given the following PRT conditions:
- Level - 80%
- Pressure - 6 psig
- Temperature - 145°F What action is required to restpre normal operating conditions to the PRT?
A. Increase-eRT pressure to 8 psig.
B. Vent/purge the PRT to the waste gas system.
,/C. Initiate cooling of the PRT.
D. Lower the PRT level.
A. Incorrect, increasing to 8 psig would bring in the high pressure alarm per reference.
B. Incorrect, venting is not required at 6 psig per reference.
C. Correct, alarm setpoint is 1320 F per reference.
D. Incorrect, the level has not reached high alarm level of 88% per reference.
K/A[CFR]: 0072.1.10 [2.7/3.9] [43.1]
References:
2-SO-68-5 2-AR-M5-A, (C-1)
LP/Objectives: OPL271PRT B.8 History: System bank Level: Memory Comments: FHW 12/02 007 2.1.10; modified "A" distractor to remove similar conditions as "B" distractor.
K->
15 (C-1)
K. /
Source Setpoint TS-68-309 SER 1213 132 0 F increasing PRESSURIZER 2-TS-68-309 RELIEF TANK TEMP HIGH Probable 1. Pressurizer safety valve or power relief valve open or leaking Causes through.
- 2. Other relief valves (CVCS and RHR) or valve packing leakoff lines connected to PRT header have flow into the PRT.
- 3. Rx vessel head vent system leaking through.
- 4. Instrument malfunction.
Corrective [1] CHECK pressurizer relief tank temperature (2-TI-68-309).
Actions [2] CHECK temperature on safety valve and power relief valve line.
[3] VERIFY vessel head vent isolation to PRT closed (2-HS-68-394 and 2-HS-68-395).
[4] ADJUST pressurizer relief tank temperature using 2-SO-68-5, PressurizerRelief Tank.
[5] IF in MODE 4 or MODE 5 and a LOCA is identified, THEN GO TO AOP-R.02, Shutdown LOCA (MODE 4, or 5).
[6] IF a small RCS leak is indicated, THEN GO TO AOP-R.05, RCS Leak and Leak Source Identification.
[7] EVALUATE TS 3.4.6.2.
References 45B655-05A, 47B601-68-34,
N.-
2.2 Developmental References (Continued)
D. TVA Drawing
- 1. 47W813-1
- 2. 47W819-1
- 3. 47W830-1
- 1. Section 5.5 3.0 PRECAUTIONS AND LIMITATIONS A. During normal operation, PRT water temperature should not exceed 1200F.
B. During normal operation, maintaining approximately 3 to 6 psig N2 gas blanket on the PRT will prevent the formation of explosive hydrogen-oxygen mixtures.
C. The PRT concentration of oxygen shall be limited to less than of equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.
D. Over filling the PRT to solid water condition during oxygen reduction per Section 8.8 may result in failure of the PRT rupture disc.
E. The PRT pressure should be maintained < 7.5 psig during normal operation. (Except during the performance of section 8.9.)
F. The PRT rupture discs are rated at 85 psig.
G. During normal operation, the level in the PRT should be maintained at 70%. If the level increases to 88% then decreasing level to 70% is necessary. If the level decreases to 55% then increasing level to 70% is needed when the PRT is required to be operable.
H. Completely draining the PRT may result in gas binding the RCDT pumps.
I. Water intrusion into the waste gas vent header is possible during PRT venting operations with PRT level high. This could affect RCP seal leakoff flows and the vent capability of tanks which vent to waste gas vent header.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 111 104. OPL271C458.1 001
-'* Per SPP-10.0, Plant Operatons, under which one of the following conditions can an individual without a license manipulate controls that directly affect reactivity.?
A. ýtrainee under the direction and in presence of an inactiveiicensed Senior Reactor Operator.
res-ANs B. When the individual has taken the NRC License Examination and is awaitinjfor confirmation.
I I(V'(& efflMX'r
,/C. Pjrainee under the direction and in presence of an active licensed Reactor uOperator.
D. When the licensed Reactor Operator requires assistance and asks for any individual's help.
A. Incorrect per reference.
B. Incorrect per reference.
C. Correct per reference.
- d. Incorrect per reference.
K/A[CFR]: 2.1.10 [2.7/3.9] [43.1]
Reference:
SPP-10.0 R2
-I LP/Objectives: OPL271 OPSMGMTL B.21 History: Procedure bank, old Bank Number B-0743 Level: Memory Comments: FHW 12/02 2.1.10.'" odified stem and distractors to conform to requirements of SPP-10.0.
TVAN STANDARD SPP-10.0 PROGRAMS AND PLANT OPERATIONS Rev. 2 PROCESSES Page 8 of 20 3.2 Surveillance Areas for Operators The UO (unit operator) at the controls shall not leave the control room surveillance area (as shown in Appendix C) for any nonemergency reason without obtaining a qualified relief. In the event of an emergency affecting the safety of operations, the UO at the controls may momentarily be absent from the surveillance area in order to verify the receipt of an annunciator alarm or initiate corrective action provided the operator remains within the confines of the Control Room. The restrictions set forth for a Unit's surveillance area are not applicable for that unit when the reactor vessel is defueled.
3.3 Personnel Recall The SM has the authority to call out all required personnel, regardless of discipline.
3.4 Authority to Operate Equipment The overall operation of the plant shall be directed by the SM. The UO, Unit Supervisor (US), and the SM should be aware of all activities affecting plant equipment.
During emergencies, operators may take necessary immediate actions required to ensure personnel and plant safety, without prior approval; however, appropriate supervisors should be promptly informed of these actions.
Only licensed operators or individuals authorized by approved procedures should operate control room equipment. Only licensed operators are permitted to manipulate the controls that directly affect the reactivity or power level of a reactor except for training purposes. A trainee may manipulate controls that directly affect the reactivity or power level of a reactor only under the direction and in the presence of an active licensed operator. Trainees operating other plant equipment not affecting reactivity or power level shall be under the supervision of an individual qualified to perform the operations.
3.5 Potentially Distractive Reading Material and Devices Reading material that does not relate to plant operation and entertainment devices such as radios, televisions, tape players, and computer games are prohibited from use by on duty Operations personnel.
Unauthorized written material and entertainment devices shall not be brought to work stations. Unauthorized written material is that which is not job related or has not been issued by TVA. Judgment should be used to ensure the plant personnel's' primary duties are not compromised.
3.6 Control Room Activities A. Potentially distracting activities in the control room and other watchstations are prohibited (e.g., radios, TV, games, horseplay, and hobbies).
B. Necessary plant-related technical/administrative control room business must be conducted at a location and in such a manner that neither licensed Unit Operator attentiveness nor the professional atmosphere will be compromised.
C. Access to the "surveillance area" (reference Appendix C) will be restricted and limited to official business only. Access for nonwork-related reasons will be prohibited. Permission to enter the "surveillance area" for personnel other than the Operations shift complement must be obtained from the UO, US or the SM.
Wednesday, October 09, 2002 @ 11:46 AIM HLC12-02.BNK Page: 106 100. OPL271C367.1 003 /'
Unit 1 is at 100% RTP when a small RCS leak develops. The operating crew takes the necessary actions to stabilize PZR level. Subsequently to these actions the crew attempts to determine the leakage source. After isolating letdown, the crew observes the following parameters/indications:
Letdown flow 0 gpm Cntmt pressure approximately 0.2 and decreasing RCS pressure approximately 2235 Pzr level approximately 65% and increasing Charging flow 40 gpm Cntmt radiation monitors decreasing Which ONE of the following actions should the SRO direct the operating crew perform next?
A. Proceed to Cold Shutdown per AOP-C.03.
,/B. Place excess letdown inservice and adjust charging flow.
C. Immediately trip the reactor.
D. Decrease turbine load to approximately 50% at 2%/min.
A. Incorrect, AOP-C.03 Emergency Shutdown is entered for conditions that require a rapid shutdown without a reactor trip. Since pressurizer level is stabilized and the other initial conditions indicated the RCS leak has been isolated a rapid shutdown is not required per AOP-R.05.
B. Correct, the RCS leak was determined to be on normal letdown. Therefore to maintain pressurizer level and RCS chemistry (chemical feed and boration) excess letdown is required.
C. Incorrect, the initial conditions do not require a reactor trip.
D. Incorrect, a load reduction of 2%/o/min is within the guidelines of AOP-C.03, Emergency Shutdown which is not required since the RCS leak is isolated.
The student must understand from the initial conditions the RCS leak was on the letdown line and has been isolated. In order to maintain pressurizer level and chemistry excess letdown is required.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 107 1000. OPL2710C367.1 003 KIA[CFR]: 011 2.1.6 [2.1/4.3] [43.5]
Reference:
AOP-R.05 EA-62-3 LP/Objectives: OPL271 C367 B.2 History: Procedure bank, old Bank Number B-0383D Level: Comprehension Comments: FHW 12/02 011 2.1.6
RCS LEAK AND LEAK SOURCE IDENTIFICATION AOP-R.05 SON Rev. 7 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Small Reactor Coolant System Leak (cont'd)
- 10. DETERMINE whether leak stopped RESTORE CVCS charging and letdown because of isolation of charging and USING EA-62-5, Establishing Normal letdown. Charging and Letdown.
- 11. EVALUATE placing excess letdown in service using EA-62-3, Establishing Excess Letdown.
- 12. INITIATE leak repairs.
- 13. GO TO appropriate plant procedure.
END OF SECTION Page 10 of 45
4.0 OPERATOR ACTIONS 4.1 Section Applicability
- 1. PERFORM Section 4.2.
[]
4.2 Placing Excess Letdown in Service
- 1. IF excess letdown is only letdown flowpath, THEN pressurizer CONTROL charging flow as necessary to prevent high El level.
r FCV-70-143 1
- 2. ENSURE CCS inlet to excess letdown heat exchanger []
OPEN.
- 3. ENSURE CCS outlet to excess letdown heat exchanger r FCV-70-85
[]
OPEN.
H]
El
- 4. VERIFY CCS flow to excess letdown heat exchanger greater than 230 gpm, as indicated on F-70--84.
H]
in NORMAL.
- 5. ENSURE excess letdown divert valve FCV.62-59 H]
- 6. OPEN excess letdown isolation valve FFFCV-62-54-4*
- 7. OPEN excess letdown isolation valve IFCV-62-55 as
- 8. ADJUST excess letdown flow control valve FFCV-62-568 exchanger outlet necessary to control flow WHILE maintaining heat El 0 F, as indicated on T*-62-58 temperature less than 200
Thursday, October 10, 2002 @ 09:48 AM HLC12-02.BNK Page: 1
- 1. OPL271C379.3 001 x The plant is operating and the following conditions exists:
Reactor power = 58% slowly increasing RCS pressure = 2210 psig slowly decreasing Auctioneered T-avg = 560°F slowly decreasing Turbine power = 595 MW (steady - no change)
S/G levels = 44% stable Steam pressure = 900 psig slowly decreasing Containment pressure = approximately 0.5 psig slowly increasing Based on the indications listed above, the most likely event in progress is which ONE of the following?
A. LOCA inside containment.
,/B. Steamline break inside containment.
C. Steamline break outside containment.
D. LOCA outside containment.
A. Incorrect, RCS pressure is dropping slowly, reactor power in increasing.
B. Correct with initial conditions.
C. Incorrect, containment pressure is slowly increasing.
D. Incorrect, RCS pressure is dropping slowly, reactor power in increasing, containment pressure is slowly increafii"g K/A[CFR]: 040 AA2.01 [4.2/4.7
Reference:
AOP-S.05 symptoms LP/Objective: OPL271C390 B.2 History: Part B, old Bank Number B-0185 Level: Analysis Comments: FHW 12/02 040 AA2.01 Comment: Reactor power is slowly increasing due to T-Avg slowly decreasing.
Containment pressure is above TS limits and slowly increasing. The cause for these symptoms is an accident in containment which drops T-Avg. The correct answer is steamline break inside containment.
AOP-S.05 Rev.02 SON STEAM LINE OR FEEDWATER LINE BREAK/LEAK Rev. 2 3.0 SYMPTOMS AND ENTRY CONDITIONS 3.1 Symptoms A. Any of the following annunciators may indicate a steamline or feedwater line break or leak:
PANEL XA-55-2C, HEATER DRAINS AND CONDENSATE C-7 LS-2-3A CONDENSER HOTWELL LEVEL ABNORMAL D-7 LS-2-9A CONDENSER HOTWELL LEVEL ABNORMAL E-7 LS-2-12A CONDENSER HOTWELL LEVEL ABNORMAL PANEL XA-55-3C, FEEDWATER AND SG LEVELS E-2 STM GEN LEVEL ADVERSE SETPOINT PANEL XA-55-5A, REACTOR COOLANT - STM - FW A-6 TS-68-2M/N RC LOOPS T AVG/AUCT T AVG DEVN HIGH-LOW A-7 FS-3-35A STEAM GEN FEEDWATER FLOW HIGH B-6 TS-68-2A/B REACTOR COOLANT LOOPS AT DEVN HIGH-LOW C-6 TS-68-2P/Q REAC COOL LOOPS T REF T AUCT HIGH-LOW PANEL XA-55-5C, VENTILATION B-1 TS-30-31 LOWER COMPT TEMP HIGH B-3 MS-30-241 LOWER COMPT MOISTURE HI PANEL XA-55-6A, REACTOR PROTECTION AND SAFEGUARDS C-2 TS-68-2E OVERTEMP AT AUTO TURB RNBK BLK C-3 ROD WTD D-2 TS-68-2F OVERPOWER AT AUTO TURB RNBK BLK C-4 ROD WTD Page 9 of 12
STEAM LINE OR FEEDWATER LINE BREAKILEAK AOP-S.05 SON Rev. 2 3.1 Symptoms PANEL XA-55-6B, REACTOR PROTECTION AND SAFEGUARDS A-1 LS-3-39D STM GEN LOOP 1 LOW FW FLOW LOW WATER LEVEL A-7 FS-3-35B STM GEN LOOP 1 STEAM/FEEDWATER FLOW MISMATCH B-1 LS-3-52D STM GEN LOOP 2 LOW FW FLOW LOW WATER LEVEL B-7 FS-3-48B STM GEN LOOP 2 STEAM/FEEDWATER FLOW MISMATCH C-1 LS-3-94D STM GEN LOOP 3 LOW FW FLOW LOW WATER LEVEL C-7 FS-3-90B STM GEN LOOP 3 STEAM/FEEDWATER FLOW MISMATCH D-1 LS-3-107D STM GEN LOOP 4 LOW FW FLOW LOW WATER LEVEL D-7 FS-3-103B STM GEN LOOP 4 STEAM/FEEDWATER FLOW MISMATCH B. Deviations or unexpected indications on any of the following may indicate a steam line or feedwater line break or leak:
"* Steam flow higher on one or more channels.
"* Increase in feedwater flow.
"* Deviations on feedwater regulating valves.
"* Main feedwater pump speed increasing.
- Increasing reactor power, decreasing T-avg with automatic rod withdrawal.
"* Main steam header pressure dropping.
"* Increasing containment pressure, temperature, humidity, and sump level.
"* Steam generator level dropping.
Page 10 of 12
STEAM LINE OR FEEDWATER LINE BREAKILEAK AOP-S.05 SON Rev. 2 3.1 Symptoms C. The following automatic actions may occur due to a steam line or feedwater line break or leak:
"* Reactor trip on OTAT.
"* Automatic rod withdrawal due to dropping T-avg.
"* Automatic rod withdrawal stop from C-3 and C-4.
"* Turbine runback from OTAT or OPAT.
3.2 Entry Conditions None Page 11 of 12
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 104
- 99. OPL271C367.1 002 2 Unit 1 is operating at 100% RTP when the operating crew observes the following:
All ice condensers doors open.
RM-90-106 & 112 radiation increasing.
Pzr level at approximately 58% and decreasing slowing.
1A-A charging pump is running.
FCV-62-93 fully open.
RCS pressure 2225 psig and decreasing slowly.
Cntmt pressure at approximately 0.2 psig and increasing slowly.
Which ONE of the following actions should the operating crew perform?
A. I d . ase#.
,/B. IltJ0iady start the 1 B-B charging pump.
C. I t start a load decrease.
D. Iveýtiate a reactor Trip.
A. Incorrect, decreasing charging flow will not bring pressurizer level back to program
-< in order to determine if charging can keep up with the RCS leak.
B. Correct, starting an additional charging pump will increase charging flow and aid in determining if the pressurizer level can be stabilized.
C. Incorrect, a load decrease will drop the pressurizer level program and make it more difficult to determine if the pressurizer level can be maintained with charging flow.
D. Incorrect, even though a reactor trip is required if the pressurizer level can not be maintained greater than 10%; that determination has yet to be made since no effect has been made to stablize the pressurizer level.
The student may recall the AOP requires controling charging flow; he must understand that an additional charging pump will increase flow as well as taking manual control of FCV-62-93. He must also understand that determining the stability of the pressurizer level is very important and could allow a RCS leak hunt while at power. A reactor trip is not a solution until the stability of the pressurizer level can be determined unable to be maintained greater than 10%.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 105
- 99. 0PL271C367.1 002 K/A[CFR]: 004 2.4.1 [4.3/4.6] [41.10 43.5]
Reference:
AOP-R.05 LP/Objectives: OPL271 C367 B.2 History: Procedure bank, old Bank Number B-0383C Level: Comprehension Comment: FHW 12/02 004 2.4.1
RCS LEAK AND LEAK SOURCE IDENTIFICATION AOP-R.05 SON Rev. 7 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Small Reactor Coolant System Leak
- 1. CONTROL charging flow as necessary to IF pressurizer level is less than 10% or loss maintain pressurizer level greater than of pressurizer level is imminent, 10%. THEN PERFORM the following:
I IF in MODE 1,2 or 3, THEN TRIP the reactor, INITIATE Safety Injection, and GO TO E-0, Reactor Trip or Safety Injection.
IF in Reduced Inventory OR Midloop, THEN GO TO AOP-R.03, RHR System Malfunction.
IF in MODE 4 or 5, THEN GO TO AOP-R.02, Shutdown LOCA.
- 2. MAINTAIN VCT level greater than 13 % IF VCT level can NOT be maintained using automatic or manual makeup. THEN ENSURE CCP suction swaps to RWST.
\K Page 4 of 45
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 102
- 97. OPL271C180.3 001 Which ONE f thwe foIwing assures the operator that the heat flux hot channel factor Upper bu n(r1 zed-wietpeaking factor remains within limits in the event of r bounorm Xenon redistribution following power changes?
A. Maintaining all control rods at the full out position for the core cycle.
B. Maintaining control rods in specified sequence and overlap.
C. Maintaining specified rod insertion limits.
,/D. Maintaining axial flux difference within specified limits.
A. Incorrect per reference. This requirement reduces rod fretting. see TI-28 att 6 B. Incorrect per reference.
C. Incorrect per reference.
D. Correct per reference.
K/A[CFR]: 2.2.25 [2.5/3.7] Ct434
Reference:
TS Bases 3/4.2.1 LP/Objective: OPL271C180 B.3 History: Part B bank, old Bank Number B-0040 Level: Memory Comments: FHW 12/02 2.2.25
3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the calculated DNBR in the core at or above design during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for0 the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.
The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
FQ(XY,Z) Heat Flux Hot Channel Factor, is defined as the maximum local 1R227 heat flux on the surface of a fuel rod at core elevation Z diyided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.
Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio of 1R227 FH(X,Y) the integral of linear power along the rod with the highest integrated power to the average rod power.
3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)
The limits on AXIAL FLUX DIFFERENCE assure that the FQ(X,Y,Z) upper bound R2271 in the COLR times the normalized axial peaking R15 envelope of FQ limit specified R144 or in the event of xenon factor is not exceeded during either normal operation redistribution following power changes.
from Provisions for monitoring the AFD on an automatic basis are derived the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector for at least 2 of outputs and provides an alarm message immediately if the AFD excore channels are outside the allowed AI-Power operating 4 or 2 of 3 OPERABLE THERMAL POWER.
space and the THERMAL POWER is greater than 50 percent of RATED FACTORS JR227 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY RISE HOT CHANNEL The limits on the heat flux hot channel factor and the nuclear enthalpy design limits on peak local power R142 rise hot channel factor ensure that 1) the and 2) in the event of a LOCA the density and minimum DNBR are not exceeded 0 fuel clad temperature will not exceed the 2200 F ECCS acceptance criteria peak limit. The peaking limits are specified in the COLR per Specification 6.9.1.14.
April 21, 1997 B 3/4 2-1 Amendment No. 19, 138, 140, SEQUOYAH - UNIT 1 155, 223 00
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 101
- 96. OPDP-4 002
-> A main control room annunciator, HIGH PRESS INAUX BLDG, has sounded on the 1-M-5 panel repeatedly over the past several hours. You eventually determine that it is a nuisance alarm. All compensatory actions have been taken. Which of the following actions should you take to remove this short term nuisance alarm from service?
A. Follow procedure SPP-10.2, "Clearance Procedure", and have the annunciator's input leads lifted and tagged.
B. Follow procedure S s", issue a TACF to disable the annunciator by liftin the associa t e transmi. r.
,/C. Complete Form OPDP-4-1, Disabled Alarm Checklist, including SM signature, and disable the alarm from the MCR.
D. Initiate a maintenance work request to have the annunciator disabled in the
- mucctions room._.
A. Incorrect per reference. VtA B. Incorrect per reference.
C. Correct per reference.
D. Incorrect per reference.
K/A {CFR}: 2.4.33 [2.4/2.8] [41.10 43.5]
References:
OPDP-4 LP/Objectives: OPL271OPSMGMTL B.12 History: Procedure bank Level: Memory Comments: FHW 12/02 2.4.33; revised stem to remove similar wording with the correct answer.
"TVAN STANDARD OPDP-4 DEPARTMENT ANNUNCIATOR DISABLEMENT Rev. 1S1 PROCEDURE Page 4 of 16 1.0 PURPOSE This procedure establishes the requirements for disabling/enabling alarms, tracking disabled alarms, establishing compensatory monitoring requirements, and the use of disabled alarm identifiers on annunciator windows. This instruction also establishes the requirements for disabling of individual inputs to annunciator windows. Attachments to this instruction provide the required approvals, compensatory monitoring, instructional steps, tracking mechanism, and restoration steps for strict control of this activity.
2.0 SCOPE This procedure provides administrative instructions to control and identify the status of the control room annunciators. This instruction should not be used to circumvent the permanent design change process. It is intended to apply to alarms or alarm inputs for reasons such as the following:
"* to avoid short-term nuisance alarm conditions.
"* to obtain a dark board and to restore alarm functions from other inputs when an alarm input is invalid or cannot be promptly cleared.
"* to support maintenance or testing activities.
3.0 INSTRUCTIONS 3.1 GENERAL REQUIREMENTS A. Before an annunciator can be disabled, the action must be reviewed to ensure that it will not result in an unsafe condition for equipment, personnel, or the public. Any Annunciator disablement requiring a 10CFR50.59 review or a Technical Evaluation (as specified in Appendix A) shall have necessary paperwork completed before disabling the alarm.
B. Each alarm input to be disabled will be reviewed to determine its impact on Technical Specifications, TRM, ODCM, FSAR, EOls, Radiological Monitoring, and Environmental Evaluation equipment.
C. Independent verification is required for removal and replacement of alarm points associated with safety systems.
D. An alarm point may be temporarily placed in service to determine if the condition has cleared or ifcorrective maintenance was sufficient to correct the deficiency.
E. Numerous alarms receive input from multiple points, any of which may cause the alarm to annunciate. If the alarm does not have "ref lash" (i.e., the alarm contacts are "Daisy Chained" in the field), other alarm contacts may be masked while the one alarm contact is in. If possible, the cause of the alarm should be determined, and the individual point removed from scan, or leads disconnected in the field to restore the alarm function from other inputs.
TVAN STANDARD OPDP-4 DEPARTMENT ANNUNCIATOR DISABLEMENT Rev. 1S1 PROCEDURE Page 6 of 16 F. Form OPDP-4-1, "Disabled Alarm Checklist" shall be maintained in the Disabled Annunciator Book with a copy of the 50.59 review and Technical Evaluation (if applicable).
G. The Shift Manager/Unit Supervisor shall approve any compensatory monitoring required for annunciators to be disabled.
H. If an annunciator is found to be in a failed state or inoperable, the Unit Supervisor or Shift Manager should refer to the appropriate Technical Specification and/or FSAR sections to evaluate system operability.
3.2 Disabling an Alarm Employees Disabling Alarms A. Initiate Form OPDP-4-1 for each alarm to be disabled.
B. Determine if alarm point can be disabled by Operations:
C. Ifalarm cannot be disabled by operations, then contact Site Engineering/System Engineer to determine a method for disabling the alarm point.
D. Complete Form OPDP-4-1, and submit to SM/US for review and approval.
STA/Site Engineering E. Perform Technical Evaluation (Form OPDP-4-2), ifrequired.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 91
- 87. GO-2-B.1 003 S' When restarting the unit after a trip, a NOTE in O-GO-2 gives specific guidance about when the operations staff is to declare mode 2.
Which ONE4) of the following is the correct time to declare MODE 2 per this procedure?
The unit enters mode 2 administratively when:
,/A. the control banks are first withdrawn.
B. the shutdown banks are first withdrawn.
C. the Keff is > .99 and RCS Temperature is > 350°F with Reactor Power less than 5%.
0 D. reactor power is greater than or equal to 1% and RCS Temperature is >.,-40 F.
- a. Correct - per NOTE in O-GO-2. This is an administrativerequirement.
- b. Incorrect - per NOTE in O-GO-2. When control banks are first withdrawn
- c. Incorrect - per NOTE in O-GO-2. Mode 2 >= 0.99 <= 5% >= 350°F 0
- d. Incorrect - per NOTE in O-GO-2. Mode 2 >= 0.99 <=5% >= 350 F K/A {CFR}: 2.1.22 [2.8/3.3] [43.5]
References:
0-GO-2 R14 section 5.2 [11] note.
LP/Objectives: OPL271 C50 B.1 History: Procedure Bank Level: Memory Comments: FHW 12/02 2.1.22/
UNIT STARTUP FROM HOT 0-GO-2 STANDBY TO REACTOR Rev: 14 CRITICAL Page 20 of 52 Unit STARTUP No. Date 5.2 Reactor Startup after a refueling outage (Continued)
[10] ENSURE all shutdown rods are fully withdrawn in accordance with 0-SI-OPS-000-004.0, Surveillance Requirements Performed on IncreasedFrequency with no Specific Frequency,within 15 minutes prior to withdrawing control rods.
_____ / /_ _
Initials Time Date
[11] IF 0-SI-SXX-068-127.0 is not in progress, THEN INITIATE applicable sections of O-SI-SXX-068-127.0 PRIOR to achieving reactor criticality to satisfy SR 4.1.1.4.a.,
Minimum Temperature ForCriticality.
NOTE The unit enters Mode 2 when the control banks are first withdrawn.
[12] PERFORM 0-RT-NUC-000-003.0, Low Power Physics Testing to control approach to criticality while continuing with this instruction. El
[13] WHEN annunciator XA-55-4A, window D-2 P-6 INTERMEDIATE RANGE PERMISSIVE is LIT, THEN [C.2]
[a] RECORD both source range readings.
[b] RECORD both intermediate range readings.
N-35 __ % RTP N-36 -% RTP Initials
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 75
- 73. ES-0.2-B.3 001
'- Unit 1 has had a loss of offsite power and is cooling down using ES-0.2, Natural Circulation Cooldown. RCS temperature is 500 OF. Power has just been restored to the CRDM fans.
Which ONE (g of the following describes the effect that starting all CRDM fans will have on the cooldown?
A. The fans will aid significantly in removing heat from the upper head region. A GREATER amount of subcooling is procedurally required for cooldown.
,/B. The fans will aid significantly in removing heat from the upper head region. A SMALLER amount of subcooling is procedurally required for cooldown.
C. The fans will NOT significantly contribute to the overall upper head cooldown rate.
A GREATER amount of subcooling is procedurally required for cooldown.
D. The fans will NOT significantly contribute to the overall upper head cooldown rate.
A SMALLER amount of subcooling is procedurally required for cooldown.
A. Incorrect, with the CRDM fans inservice more ambient heat is removed from the vessel head which allows for a smaller subcooling margin.
B. Correct, with the CRDM fans inservice more ambient heat is removed from the vessel head which allows for a smaller subcooling margin.
C. Incorrect, the CRDM fans will remove a significant amount of heat as discussed in Westinghouse ERG ES-0.2.
D. Incorrect, the CRDM fans will remove a significant amount of heat as discussed in Westinghouse ERG ES-0.2.
The student must understand the significance of additional cooling on the vessel head and it's contribution to a smaller required subcooling margin.
K/A [CFR]: E09 EA2.1 [3.1/3.8] [43.5]
References:
EPM-3-ES-0.2 LP/Objectives: OPL271C382, B.3 History: Procedure bank "Level: Comprehension Comments: FHW 12/02 E09 EA2.1
SON EOI BASIS DOCUMENT FOR ES-0.2 EPM-3-ES-0.2 PROGRAM NATURAL CIRCULATION COOLDOWN Rev. 2 MANUAL Page 32 of 57 EOP Step Number: 14 INITIATE RCS depressurization:
ERG Step Number: 12 Initiate RCS Depressurization:
Purpose:
To initiate depressurization of the RCS while maintaining required subcooling.
ERG Basis:
The pressurizer pressure should periodically be decreased to maintain the reactor coolant and pressurizer pressure-temperature relationship in accordance with the Technical Specifications and the figures shown in the ERG Appendix to this section (described below).
The depressurization should be accomplished using pressurizer auxiliary spray or pressurizer PORVs, depending upon whether letdown is in service.
To prevent possible void formation in the upper head, the plant-specific minimum RCS subcooling based on core exit TCs, as described in the Appendix, should be maintained.
<> With the availability of CRDM cooling fans, the total upper head cooldown rate for T-cold plants varies from a maximum of 54°F/hr to about 450F/hr when the upper head temperature is cooled to 350°F (34 0F/hr from the natural circulation cooldown rate of 50°F/hr plus 20 0 F/hr 0
from the CRDM fans when the upper head temperature is at its highest, 572 F, to 11°F/hr when the upper head temperature is 350'F). For T-hot plants, the total upper head cooldown rate due to both the natural circulation cooldown rate of 25 0F/hr (upper head cooldown 0F/hr at rate 0
of 1O F/hr) and the availability of CRDM fans (upper head cooldown rate from 21 600°F to 11°0F/hr at 350°F) varies from 31 F/hr initially to about 21 F/hr when the upper head temperature is cooled to 350°F. With CRDM fans available, a subcooling margin of at least 5O°F should be maintained for both types of plants during depressurization. Without the availability of CRDM fans, T-cold plants must maintain a minimum subcooling of 100°F during depressurization. For T-hot plants without CRDM fans available, the appropriate precautions, as outlined in Appendix, should be taken to ensure that a minimum subcooling of 200PF is maintained at all times during depressurization. However, if problems arise from a more restrictive Technical Specification limit, they will have to be resolved on a plant specific basis (e.g., lower minimum subcooling requirement will result in a longer upper head cool-off time period).
See letter from R. W. Jurgensen to P. S. Check, St. Lucie Cooldown Event Report, OG-57, April 20, 1981, for more detail on the determination of these limits for natural circulation cooldown.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 73
- 71. ES-0.0-B.5 002 Given the following plant conditions:
- Reactor trip and SI have occurred.
- S/G level in #2 S/G was increasing uncontrollably.
- Crew transitioned to E-3.
- Chemistry and Rad Con report NO activity in ANY S/G.
- Turbine Building AUO reports #2 S/G MDAFW LCV is leaking by.
- The indicator for AFW flow to #2 S/G is determined to be failed low.
- Isolation of #2 S/G MDAFW LCV stops level increase in #2 S/G.
- #2 S/G pressure is stable at 1010 psig.
- E-3 cooldown is in progress.
Which ONE $A(fthe following is the correct procedural action to take in response to the above conditions?
A. Continue in E-3 and terminate SI in E-3.
B. Transition directly to ES-l.1 and terminate SI.
C. Transition to E-2. Transition from E-2 to E-1. Transition from E-1 to ES-1.1.
vD. Transition to ES-0.0. Transition from ES-0.0 to E-1. Transition from E-1 to ES-1.1.
A. Incorrect, E-3 provides steps to mitigate a SGTR.
B. Incorrect, ES-I.1 has entry conditions from E-0, E-1, and FR-H.1.
C. Incorrect, if in E-3 the any S/G pressure dropping in an uncontrolled manner, any S/G pressure less than 140 psig, or any faulted S/G not isolated would be a transition to E-2. These conditions are not listed in the initial conditions.
D. Correct, per EPM-4 ES-0.0 is applicable only if SI is in progress and E-0 has been performed and entered based on Operator judgment.
K/A[CFR]: E01 EA2.1 [3.2/4.0] [43.5]
Reference:
ES-0.0, ES-I.1 entry conditions, E-3, EPM-4.
LP/Objective: OPL271C380 B.2 History: Procedure bank Level: Comprehension Commets: FHW 12/02 E01 EA2.1
3.11.5 Use of ES-0.0, Rediagnosis A. ES-0.0, Rediagnosis, is unique among the EOPs in that it has no specific transition into it. It is entered strictly based on operator judgment and is applicable only if SI is in progress and E-0 has already been performed.
B. ES-0.0 should be used when the operator has any concern that he may not be in the right EOP based on plant conditions. This is most likely to happen if multiple accidents occur either simultaneously or sequentially.
C. Once entered, ES-0.0 will either transition the operator to ECA-2.1, E-1, E-2, or E-3, or will return him to the procedure and step in effect, depending on diagnostics done within the procedure.
D. If ES-0.0 determines that an operator should be in a certain series of procedures (e.g., E-1 or ECA-1 series), and he is, then he simply returns to the procedure and step in effect.
E. If ES-0.0 determines that an operator should be in a certain series of procedures (e.g., E-3 or ECA-3 series), and he is NOT, then he is sent to either E-1 (if he should be in E-1 or ECA-1 series) or E-3 (if he should be in E-3 or ECA-3 series) to enter the appropriate series at the beginning and work his way through the series normally from that point on.
'a 1.0 PURPOSE This procedure provides operator actions to terminate safety injection and stabilize plant conditions.
2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 ENTRY CONDITIONS E-0 Reactor Trip or Safety Injection:
- Specified termination criteria satisfied.
E-1 Loss of Reactor or Secondary Coolant:
- Specified termination criteria satisfied.
FR-H.1 Loss of Secondary Heat Sink:
- Secondary heat sink reestablished and SI terminated.
3.0 OPERATOR ACTIONS Page 2 of 17
ISTEP ACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED I
- 1. DETERMINE procedure applicability: IF SI required, THEN
"* CHECK any SI pump RUNNING. PERFORM the following:
"* CHECK CCPIT flow INDICATED. a. ACTUATE SI.
"* CHECK E-0, Reactor Trip or Safety b. GO TO E-0, Reactor Trip or Safety Injection, previously COMPLETED. Injection.
IF SI NOT required, THEN RETURN TO procedure and step in effect.
- 2. CHECK S/G secondary pressure IF controlled cooldown in progress, boundary integrity: THEN GO TO Step 3.
- Any S/G pressure stable or rising.
IF controlled cooldown NOT in progress, THEN PERFORM one of the following:
"*IF main steamlines NOT isolated, THEN GO TO E-2, Faulted Steam Generator Isolation.
"*IF main steamlines isolated, THEN GO TO ECA-2.1, Uncontrolled Depressurization of All Steam Generators.
Page 3 of 6
I STEP IACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I
- 3. CHECK S/G secondary pressure VERIFY all Faulted S/G(s) ISOLATED:
boundary integrity:
"*MSIVs and bypasses CLOSED
"* S/G pressures controlled or rising "*AFW ISOLATED
- MFW ISOLATED
"* S/G pressures
- Atmospheric relief CLOSED greater than 140 psig.
- S/G blowdown valves CLOSED
- Steam supply to TD AFW pump ISOLATED (S/G 1 or 4).
IF any Faulted S/G NOT isolated, THEN GO TO E-2, Faulted Steam Generator Isolation.
Page 4 of 6
ISTEP ACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED I
- 4. CHECK S/G tube integrity: PERFORM the following:
- a. CHECK the following indications of a 1) VERIFY procedure in effect is one of the S/G tube rupture, including available following E-1 or ECA-1 series trends prior to isolation: procedures:
- Any S/G level rising in an uncontrolled manner. E-1, Loss of Reactor or Secondary Coolant.
- Main steamline high radiation.
OR " ES-1.4, Transfer to Hot Leg Recirculation.
"* S/G blowdown recorder RR-90-120, pen #1 and pen #2 " ECA-1.1, Loss of RHR Sump high radiation. Recirculation.
OR " ECA-1.2, LOCA Outside Containment.
"* Post-Accident Area Radiation
- 2) IF E-1 or ECA-1 series procedure Monitor recorder RR-90-268B, points 3 (blue), 4 (violet), in effect, 5 (black), or 6 (brown) THEN high radiation. RETURN TO procedure and step
[M-31 (back of M-30)] in effect.
- 3) IF E-1 or ECA-1 series procedure NOT in effect, THEN GO TO E-1, Loss of Reactor or Secondary Coolant.
Page 5 of 6
E-3¸ Ev3 SON STEAM GENERATOR TUBE RUPTURE Rev. 11 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 ENTRY CONDmIONS (Continued)
ECA-3.1 SGTR and LOCA - Subcooled Recovery:
0 SIG level rising in an uncontrolled manner.
ECA-3.2 SGTR and LOCA - Saturated Recovery:
- SIG level rising in an uncontrolled manner.
ECA-3.3 SGTR Without Pressurizer Pressure Control:
- S/G level rising in an uncontrolled manner.
- Pressurizer pressure control restored.
FR-H.3 Steam Generator High Level:
0 Secondary Radiation.
3.0 OPERATOR ACTIONS Page 3 of 35
SEP ACTION/EXPECTED RESPONSE ESPONSE NOT OBTAINED 1.0 PURPOSE This procedure provides actions to terminate leakage of reactor coolant into the secondary system following a steam generator tube rupture.
2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 ENTRY CONDITIONS E-0 Reactor Trip or Safety Injection:
- Secondary radiation.
- S/G level rising in an uncontrolled manner.
E-1 Series Foldout Page
- S/G level rising in an uncontrolled manner.
E-1 Loss of Reactor or Secondary Coolant:
- Secondary radiation.
- S/G level rising in an uncontrolled manner.
ES-1.2 Post LOCA Cooldown and Depressurization:
- S/G level rising in an uncontrolled manner.
E-2 Faulted Steam Generator Isolation:
- Secondary radiation.
ES-3.1 Post - SGTR Cooldown Using Backfill:
0 S/G level rising in an uncontrolled manner.
ES-3.2 Post - SGTR Cooldown Using Blowdown:
- S/G level rising in an uncontrolled manner.
ES-3.3 Post - SGTR Cooldown Using Steam Dump:
- S/G level rising in an uncontrolled manner.
ECA-2.1 Uncontrolled Depressurization of All Steam Generators:
- Secondary radiation.
(Step continued on next page.)
Page 2 of 35
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 68
- 66. ECA-2.1-B.1 004 ECA-2.1, Uncontrolled depressurization of all Steam Generators is in effect. The following conditions occur:
- TDAFW pump trips on overspeed and can NOT be reset.
- "A" MDAFW pump is tagged out.
- "B" MDAFW pump trips on overcurrent.
- Offsite power is lost, with both D/G's re-energizing their respective 6.9 kv AC bus.
- All SG levels are below 0% NR.
The operating team should:
A. Remain in ECA-2.1.
B. Transition to E-0.
C. Transition to E-2.
,/D. Transition to FR-H.1.
A. Incorrect per reference.
B. Incorrect per reference.
C. Incorrect per reference.
D. Correct, the reduction in AFW flow was not by procedure direction therefore a transition to FR-H.1 is required.
KIA[CFR]: E122.4.16 [3.0/4.0] [41.10 43.5]
Reference:
ECA-2.1 note on page 3.
LP/Objectives: OPL271C419, b.1 History: Procedure bank, old Bank Number PL-0798 Level: Analysis Comments: FHW 12/02 E12 2.4.16
SON UNCONTROLLED DEPRESSURIZATION OF ECA-2.1 ALL STEAM GENERATORS Rev. 8 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION Isolating both steam supplies to the TD AFW pump when it is the only source of feed flow will result in loss of secondary heat sink.
NOTE Reducing total feed flow to less than 440 gpm, as directed in this procedure, does NOT require implementation of FR-H.1, Loss of Secondary Heat Sink, as long as a total feed flow capability of 440 gpm is available.
- 1. CHECK secondary pressure boundary:
- a. CHECK the following: a. CLOSE valves.
"* MSIVs and MSIV bypass valves IF valves can NOT be closed manually, CLOSED THEN DISPATCH personnel to close valves
"* MFW regulating valves and locally, one loop at a time.
regulating bypass valves CLOSED IF MFW flow inrlinate.
"* MFW isolation valves CLOSED THEN CLOSE additional feedwater or condensate MOVs as necessary.
"* Atmospheric reliefs CLOSED
"* S/G blowdown valves CLOSED
"* MFW flow indication at ZERO.
IF either MD AFW pump can NOT be started, THEN GO TO Step 2.
1
- c. CLOSE TD AFW pump steam supply valves FCV-1 -17 or FCV-1 -18.
Page 3 of 26
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 67
- 65. ECA-1.2-B.2 001 Unit 2 is responding to a LOCA into the Auxiliary Building in ECA-1.2 (LOCA Outside Containment). upon completion of ECA-1.2, RCS pressure continues to decrease.
Which ONE,1of the following statements correctly describes the correct mitigating strategy to assure continued removal of decay heat under these conditions?
A. Transition back to E-1 (Loss of Reactor or Secondary Coolant).
,/B. Transition to ECA-1.1 (Loss of RHR eol*a4sua12tSump Recirculation).
C. Transition to ES-1.2 (Post LOCA Cooldown and Depressurization)
D. Transition to ES-1.3 (Transition to RHR Containment Sump).
A. Incorrect per reference 4 -
B. Correct per reference.
C. Incorrect per reference.
D. Incorrect per reference.
K/A {CFR}: E04 EA2.1 [3.4/4.3] [43.5]
References:
ECA-1.2, p. 7 LP/Objectives: OPL271c418, b.2 History: Procedure bank.
Level: Comprehension Comments: FHW 12/02 E04 EA2.1.
biL
STEP IACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
- 5. DETERMINE if LOCA isolated:
- b. GO TO E-1, Loss of Reactor or Secondary Coolant.
END Page 6 of 6
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 63
- 61. E15 EA2.1 001 The operating crew is at step two (2) of FR-Z.2, "Containment Flooding."
Which one of the following is the correct criteria necessary to exit FR-Z.2?
A. RHR suction must be aligned to the RWST.
,/B. Containment pressure increased to 3.0 psid.
C. Chem Lab analysis is complete on the RHR system sample.
D. Containment sump was pumped down to < 55%.
A. Incorrect, per reference the only action verb associated with RHR is "check."
B. Correct, per reference. transistion to FR-Z.1.
C. Incorrect, per reference the chem lab is only notified to take a sample.
D. Incorrect, there is no level requirement in the reference.
K/A[CFR]: E15 EA2.1 [2.7/3.2] [43.5]
Reference:
FR-0 EPM-4 LP/Objective: OPL271C409 B.1 History: New question.
Level: Memory Comments: FHW 12/02 E15 EA2.1Z
3.10.5 Status Tree Rules of Usage
- 3. Status trees are designed to monitor for the most severe challenges first to shorten response time in addressing those conditions. Therefore, typically, RED paths will be at the top of a status tree, following by ORANGE, YELLOW, and GREEN as the status tree branches downward.
- 4. If any RED challenge is detected, the person monitoring status trees informs the procedure reader immediately before continuing with monitoring any subsequent status trees. Since they are monitored in order of importance, the first RED challenge encountered will be the highest priority RED and therefore, the highest priority challenge.
- 5. If any ORANGE challenge is encountered, the person monitoring status trees continues monitoring until all six status trees have been evaluated.
This is necessary because a subsequent RED challenge has priority over any ORANGE challenge. If any RED is encountered, then Rule 3.10.5.D.4 applies. Otherwise, once it is determined that no RED challenges exist, then the person monitoring status trees informs the procedure reader of the highest priority ORANGE challenge.
- 6. RED or ORANGE challenges must be addressed immediately by implementing appropriate FRPs in order of priority and per the rules of usage. When the person monitoring status trees informs the procedure reader that a RED or ORANGE challenge exists, the procedure reader immediately suspends the ORP (or lower priority FRP) in progress and implements the appropriate FRP, as indicated at the terminus point of the CSF under challenge.
- 7. YELLOW challenges may be addressed by implementing appropriate FRPs if desired, but do not require immediate operator action.
Addressing YELLOW challenges is optional since these are usually temporary, off-normal conditions that will be restored to normal status by actions already in progress. In other cases, the YELLOW path might provide an early indication of a developing RED or ORANGE condition.
Following FRP implementation, a YELLOW might indicate a residual off normal condition. When the person monitoring status trees informs the procedure reader that a YELLOW challenge exists, the procedure reader should evaluate if the YELLOW challenge FRP should be implemented.
This decision will be based on the following:
- Whether the procedures in effect will address the challenge as a matter of course.
"* Whether the procedures in effect are more important at that time based upon available time and current plant conditions.
"* Whether the challenge is of a nature that it will likely develop into an ORANGE or RED condition if action is not taken early.
\ I
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 65
- 63. E16 2.4.41 001 question has reference material attached. S-This Plant Conditions.-,,.,
- Chem lab reports RCS activity 280 pCi/gm dose equivalent 1-131.
RM-90-271 and 27 arendicating 370 Rem/Hr.
Make NO assumptions.
Classify this condition per the Emergency Plan Implementing Procedure.
A. Unusual Event.
B. Alert.
/A4 . Site Area Emergency.
f"*. General Emergency.
C. correct, EAL 1.1.5.L a 1.3.5.P. Site Area Emergency is a "Loss or Potential Loss of any two barriers" K/A[CFR]: E162.4.41 [2.3/4.1] [43.5]
Reference:
EPIP-1 LP/Objective: 0PL271C198 B.3 History: New question.
Level: Comprehension Comments: FHW 12/02 E16 2.4.41 Provide a copy of EPIP-1 A R(I~_I, - e4 -
MODES 1,2,3,4 EPIP-1 SQN Rev 33 FISSION PRODUCT BARRIER MATRIX Page 8 of 52
- 1. Fu Cla el Bare
- 1. R S -
I1. Critical Safetv Function Status 1. Critical Safety Function Status J
m LSS Potential LOSS
- Potential LOSS Core Cooling Red Core Cooling Orange Not Applicable. Pressurized Thermal (FR-C.1) (FR-C.2) Shock Red (FR-P.1).
OR OR Heat Sink Red (RHR SD Heat Sink Red (RHR SD cooling not in service) cooling not in service)
(FR-H.1). (FR-H.1).
-OR- -OR rrimary Coolant Activity Level D A m D* ID
- 2. 2. RCS LeakagleILOCA Potential LOSS
- Potential LOSS RCS sample activity is Not Applicable. RCS leak results in Non isolatable RCS leak greater than 300 gCi/gm subcooling < 40 IF as exceeding the capacity of dose equivalent Iodine-131 indicated on XI-94-1 01 OR one charging pump in the 102 (EXOSENSOR). normal charging
-OR alignment.
-eater than 1200 OF on Gre*ter than or equal to Entry Into E-1.
)4-101 OR 102 700 IF on Xl-94-101 or
'K ptXOSENSOR). 102 (EXOSENSOR).
-OR
-OR 3. Steam Generator Tube Rupture
- 4. Reactor Vessel Water Level *. Potential LOSS
- Potential LOSS SGTR that results in a Not Applicable.
Not Aplicble.VALID RVLIS level < 40%
safety injection actuation.
on LI-68-368 or 371 with OR no RCP running.
Entry into E-3.
-OR
- 5. Containment Radiation Monitors
- , Potential LOSS -OR VALID reading of Greater Not Applicable. 4. Reactor Vessel Water Level Than: *- Potential LOSS VALID RVLIS level < 40% Not Applicable.
2.8E + 01 Rern/hr On on LI-68-368 or 371 with RM-90-271 and 272. no RCP running.
OR 2.9E + 01 Rem/hr On RM-90-273 and 274.
-OR- -OR Site Emergency Director Judgment 5. Site Emergency Director Judgment y condition that, in the judgment of the SM or SED, Any condition that, in the judgment of the SM or SED, icates loss or potential loss of the Fuel Clad Barrier indicates loss or potential loss of the RCS Barrier ccomparable to the conditions listed above. comparable to the conditions listed above.
EMERGENCY PLAN EPIP-1 SON CLASSIFICATION MATRIX Rev 33 Page 9 of 52 INSTRUCTIONS
- 1. Critical Safet Function Status NOTE: A condition is considered to be MET if, in the S,.Potential LOSS judgment of the Site Emergency Director, the Not Applicable. Containment Red condition will be MET imminently (i.e., within (FR-Z.1) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). The classification shall be made as OR soon as this determination is made.
Actions of FR-C.1 (Red Path) are INEFFECTIVE (i.e.: core TC's trending up). 1. In the matrix to the left, REVIEW the Initiating Conditions in all three barrier columns and CIRCLE the Conditions that are Met.
-OR
- 2. Containment Pressure/Hydrogen 2. In each of the three barriers columns, IDENTIFY if any
- Potential LOSS Loss or Potential Loss Initiating Conditions have been Rapid unexplained pressure Containment hydrogen Met.
decrease following initial increases to > 4% by volume increase on Pdl-30-44 or 45 on H21-43-200 or 210. 3. COMPARE the number of barrier Losses and Potential OR OR Losses to the Criteria below and make the appropriate Containment pressure or Pressure > 2.81 PSID (Phase declaration.
sump level not increasing on B) with no containment spray LI-63-178 or 179 with a LOCA operating when required NOTE: MONITOR the respective status tree criteria if a CSF in progress. (FR-Z.1). is listed as an InitiatingCondition.
-OR-
- 3. Containment Isolation Status III LOSS6 Potential LOSS EMERGENCY CLASS CRITERIA e.tainment isolation, when Not Applicable.
quired, is incomplete and a elease path to the GENERAL EMERGENCY environment exists.
LOSS of any two barriers and Potential
-OR- LOSS of third barrier.
- 4. Containment Bypass S°"f/
Potential LOSS Secondary side release Unexpected VALID increase in SITE AREA EMERGENCY outside containment from a area or ventilation RAD RUPTURED S/G that cannot monitors adjacent to LOSS or Potential LOSS of any two be terminated in < 15 minutes containment (with LOCA in (E-2 and E-3). progress). barriers.
> 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> secondary side ALERT release outside containment from a SIG with a S/G tube Any LOSS p[ Potential LOSS of Fuel Clad leak > T/S limits (AOP-R.01, App A). barrier.
-OR
- 5. Significant Radioactivity in Containment LOSS Any LOSS or Potential LOSS of RCS s _Potential Not Applicable. VALID Reading of greater than: barrier.
3.6 E + 02 Rem/hr on RM-90-271 and RM-90-272.
OR UNUSUAL EVENT 2.8 E + 02 Rem/hr on RM-90-273 and RM-90-274. LOSS Q1 Potential LOSS of Containment
-n'R- Barrier.
ý'*.Site Emergency Director Judgment Any condition that, in the judgment of the SM or SED, indicates loss or potential loss of the CNTMT Barrier comparable to the conditions listed above.
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 61
- 59. E-0-B.6 003 /
SWhich ONE of the following symptoms would require the initiation of a manual reactor trip AND safety injection if neither had occurred automatically?
A. Containment pressure = 1.0 psig B. General Warning alarm on the Solid State Protection System B VIC. Pressurizer pressure = 1850 psig, pressurizer level = 40%
D. Power = 33% and loss of flow in one loop A. Incorrect per reference.
B. Incorrect per reference, general warning requires 2/2 logic.
C. Correct per reference pressurizer pressure < 1970 psig for reactor trip and < 1870 psig for safety injection.
D. Incorrect per E-0, however AOP-R.04 does require a manual reactor trip if greater then P-10 (10%) but it does not require a safety injection.
K/A[CFR]: 013 2.4.47 [3.4/3.7] [41.10 43.5]
2.4.1 (4.3 - 4.6) 2.4.2 (3.9 -4.1) 2.4.4 (4.0 - 4.3)
Reference:
E-0 LP/Objective: 0PL271C379 B.1 History: Procedure bank Level: Memory Comments: FHW 12/02 013,2.4-7 {-oA) 962 I 0-Ho-ti
- 1 W
1<1 /A
SON REACTOR TRIP OR SAFETY INJECTION E-0 Rev. 23 2.2 SYMPTOMS OF REACTOR TRIP A. Any valid reactor trip signal [status panel M-5 or M-6].
B. Any reactor trip alarm lit [M-4D].
C. Rapid drop in neutron level indicated by nuclear instrumentation.
D. Shutdown and control rods inserted.
E. Rod bottom lights lit.
F. Rod position indicators at zero.
2.3 SYMPTOMS REQUIRING SAFETY INJECTION A. Pressurizer pressure less than 1870 psig (blockable below P-11).
B. S/G pressure less than 600 psig (lead/lag) (blockable below P-11).
C. Containment pressure greater than 1.54 psid.
2.4 SYMPTOMS OF SAFETY INJECTION A. Any valid SI signal [status panel M-6].
B. Any SI alarm lit [M-4D].
C. ECCS pumps running.
3.0 OPERATOR ACTIONS 3 of 25
Wednesday, October 09, 2002 @ 11:46 AM HLC12-02.BNK Page: 60
- 58. E-0-B.3.A 007 Given the following plant conditions:
- A reactor trip has occurred due to loss of offsite power.
- SI did NOT actuate.
- 6.9kV Shutdown Boards 1A-A and 1B-B are energized by the D/Gs.
- The crew has just ensured that all Control Rods are fully inserted.
- The following plant parameters exist:
- PZR Level is 15% and increasing very slowly.
- PZR Pressure is 1980 psig and decreasing.
- AFW Flow is -580 gpm.
- All S/G levels are approximately 5%
- The crew ensures that the PZR PORVs are closed and the spray valves are closed.
Which ONE ( of the following following describes the MINIMUM actions that must be performed t stabilize plant conditions?
A. Reset the Blackout Relays on both 6.9kV Shutdown Boards and allow the PZR Pressure Control System to automatically stabilize conditions.
B. Manually actuate SI and return to E-0, "Reactor Trip or Safety Injection".
C. Secure AFW flow to the S/Gs and place the PZR backup heaters control switch to CLOSE.
v/D. Increase PZR level, then allow the PZR Pressure Control System to automatically stabilize conditions.
A. Incorrect, blackout relays can not be reset until offsite power is restored.
B. Incorrect, initial conditions do not require a SI.
C. Incorrect, AFW flow is needed for a heat sink, S/G level >10% per reference.
D. Correct, control charging/letdown so PZR level trending to 25% per reference.
K/A[CFR]: 056 AA2.18 [3.8/4.0] [43.5]
Reference:
ES-0.1 Objective: OPL271C381 B.1 History: Procedure bank Level: Comprehension Comments: FHW 12/02 056 AA2.18
SQN REACTOR TRIP RESPONSE ES-0.1 Rev. 26 STEP IACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7. MONITOR pressurizer level control:
- a. CHECK pressurizer level a. PERFORM the following:
greater than 17%.
- 1) ENSURE letdown ISOLATED.
- 2) ENSURE pressurizer heaters OFF.
- 3) CONTROL charging to restore pressurizer level greater than 17%.
- 4) WHEN pressurizer level greater than 17%,
THEN OPERATE pressurizer heaters as necessary.
- b. VERIFY charging IN SERVICE. b. ESTABLISH charging USING EA-62-5, Establishing Normal Charging and Letdown.
- c. VERIFY letdown IN SERVICE. c. WHEN charging established AND pressurizer level greater than 17%,
THEN ESTABLISH letdown USING EA-62-5, Establishing Normal Charging and Letdown.
- d. CHECK pressurizer level d. CONTROL charging and letdown to trending to 25% maintain pressurizer level at 25%
(normal range 20% to 30%). (normal range 20% to 30%).
7 of 16
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 49
- 48. AOP-R.05-B.2 001
"-> Given the following plant conditions:
- Reactor power is STABLE at 90%
- Pressurizer level is DECREASING
- VCT level is DECREASING
- The following annunicators are actuated:
"* PRESSURIZER LEVEL HIGH-LOW
"* LTDN HX OUTLET TO DEMIN TEMP HIGH
"* REGENERATIVE HX LETDOWN LINE TEMP HIGH Which ONEJIof the following events would most likely cause these indications?
A. Pressurizer level control valve malfunction.
B. Isolation of CVCS letdown.
D. Letdown header rupture.
A. Incorrect, both pressurizer and VCT level would not be decreasing.
B. Incorrect, "Regenerative Hx Letdown Line temp High" would not alarm.
C. Correct, with a loss of charging due to a pipe break the pzr would still lower in level due to letdown. Since there is no flow balance VCT level would decrease. Without charging acting as a cooling medium on the regen ht exchgr the temp alarms would annunciate.
D. Incorrect, letdown header rupture would cause a loss of letdown therefore the high temp alarm would not annunciate.
K/A[CFR]: 22 AA2.01 [3.2/3.8]
Reference:
1-AR-M6-C(A-4) 1-AR-M6-C(D-4) 1-AR-M5-A(C-3)
AOP-R.05 Objective: OPL271C367, B.2 History: Procedure bank Level: Analysis Comments: FHW 12/02 22 AA2.01
17 (C-3)
Source Setpoint LS-68-335D/E SER2112 70% span increasing PRESSURIZER LS-68-335D/E - High 5 percent of span deviation LEVEL SER 2113 below level program LS-68-339F/E - Low HIGH-LOW Probable I. RCS leak exceeding charging capacity.
Causes 2. Load transient or RCS temperature transient condition.
- 3. Charging and/or letdown flow mismatch.
- 4. Instrument malfunction for level or Tavg.
Corrective [1] CHECK pressurizer level (1-LI-68-339A, 335A, 320)
Actions E2] IF level is high, THEN ENSURE backup heaters ON.
[3] ENSURE level control system is attempting to return level to program with letdown and charging.
[4] IF level channel failed, THEN GO TO AOP-1.04, Pressurizer Instrument Malfunction.
15] IF RCS leakage is suspected, THEN GO TO AOP-R.05, RCS Leak and Leak Source Identification.
E6] IF in MODE 4 or MODE 5 and a LOCA is identified, THEN GO TO AOP-R.02, Shutdown LOCA (MODE 4, or 5).
E7] EVALUATE Technical Specifications 3.3.1, 3.3.2 and 3.4.6.2 as applicable.
References 45B655-05A-0, 47B601-68-45
25 (D-4)
Source Setpoint TS-62-71 SER 2172 REGENERATE HX 1-TS-62-71 3970 F increasing LETDOWN LINE TEMP HIGH Probable 1. Letdown flow is too high.
Causes 2. Charging flow is too low.
Corrective [1] CHECK letdown temperature on [1-TI-62-711.
Actions [2] CONSIDER the following operations to reduce letdown temperature:
[a] REDUCE letdown flow.
[b] MAINTAIN letdown pressure at -340 psig.
[c] INCREASE charging flow.
[3] MONITOR reactor coolant pump seal flow.
[4] MONITOR pressurizer level by observing
[1 -LI-68-335A1.
References 45B655-06C-0, 47B601-62-13, 47W610-62-2, 47W809-1
4 (A-4)
Source Setpoint TS-62-78 SER 2160 LTDN HX OUTLET 1-TS-62-78 1320F TO DEMIN TEMP HIGH Probable ,. Letdown flow is greater than charging flow.
Causes 2. Inadequate component cooling H20 flow through heat exchanger.
- 3. Failure of 1-TCV-70-192 to respond in AUTO.
Corrective [1] CHECK letdown temperature on [1-TI-62-781.
Actions [2] ENSURE letdown demins bypassed at Ž 137 0F to prevent resin damage.
[3] ENSURE proper valve alignment in accordance with 1-SO-62-1, Chemical and Volume Control System.
NOTE Automatic positioning of 1-TCV-70-192 is controlled by input from HIC-62-78 (in Auto) and TS-62-78 (In Auto or Manual at >1241F).
[4] IF rl-HIC-62-781 not operating properly in AUTO, THEN PLACE in MANUAL, AND ATTEMPT temperature control.
[5] VERIFY low pressure letdown valve maintaining letdown pressure at - 340 psig.
[6] IF Letdown heat exchanger temperature cannot be controlled, THEN
-EVALUATE switching to excess letdown in accordance with 1-SO-62-6, Excess Letdown.
References 45B655-06C-0, 47W610-62-2
SON I FRCS LEAK AND LEAK SOURCE IDENTIFICATION AOP-R.05 Rev. 7 3.0 SYMPTOMS AND ENTRY CONDITIONS 3.1 Symptoms A. Any of the following annunciators may indicate a RCS leak:
PANEL XA-55-4B, NIS/ROD CONTROL E-7 TS-68-398 REACTOR VESSEL VENT TEMP HI PANEL XA-55-5A, REACTOR COOLANT - STM - FW A-1 TS-68-21 REACTOR VESSEL FLANGE LEAKOFF TEMP HIGH B-1 LS-68-30OA/B PRESSURIZER RELIEF TANK LEVEL HI-LOW B-5 LS-77-125E CNTMT FLOOR & EQUIP DRAIN SUMP HI-HI-HI C-1 TS-68-309 PRESSURIZER RELIEF TANK TEMP HIGH C-3 LS-68-335D/E PRESSURIZER LEVEL HIGH-LOW C-5 LS-77-41 OA REACTOR BLDG AUX FL & EQ DRAIN SUMP HI D-2 TS-68-328 PRESSURIZER SAFETY VALVE LINES TEMP HIGH D-4 PS-68-340G/F PRESSURIZER PRESSURE LOW BACKUP HTRS ON E-2 TS-68-331 PRESSURIZER POWER RELIEF LINE TEMP HIGH E-3 LS-68-335E/D PRZR LVL LOW HEATER OFF & LETDOWN SECURED PANEL XA-55-5C, VENTILATION B-1 TS-30-31 LOWER COMPT TEMP HIGH B-3 TS-30-241 LOWER COMPT MOISTURE HI B-4 TS-30-240 LOWER COMPT MOISTURE HI B-6 XS-68-363 PRESSURIZER RELIEF VALVE OPEN Page 15 of 45
RCS LEAK AND LEAK SOURCE IDENTIFICATION AOP-R.05 SON Rev. 7
- /
3.1 Symptoms (cont'd)
PANEL XA-55-6C, CVCS-HEAT TRACE-UHI A-3 LS-62-129A/B VOLUME CONTROL TANK LEVEL HI-LOW C-5 CONTAINMENT VENTILATON ISOLATION TRAIN A C-6 CONTAINMENT VENTILATON ISOLATION TRAIN B E-3 LS-63-104 CONTAINMENT SUMP FULL E-4 LS-63-176 CNTMT LEVEL HI RHR RECIRC E-7 FCV-74-1/2 TROUBLE OR RHR PRESS HI PANEL 1(2)-XA-55-6D, AUXILIARY SYSTEMS A-4 (2-)PS-74-13 RHR PUMPS DISCH PRESS HI OR MINIFLOW CONDITION E-1 AUX BLDG HIGH ENERGY LINE BREAK PANEL XA-55-30, POST ACCIDENT RADIATION MONITORING A-1 RA-271A UPPR INCNTMT HI RAD A-2 RA-273A LWR IN CNTMT HI RAD B-1 RA-272A UPPR INCNTMT HI RAD B-2 RA-274A LWR IN CNTMT HI RAD PANEL O-XA-55-12A, UNIT 1 AND COMMON RADIATION MONITOR A-4 1-RA-90-106A CNTMT BLDG LWR COMPT AIR MON HIGH RAD B-1 1-RA-123A CCS LIQ EFF MON HIGH RAD PANEL O-XA-55-12B, COMMON RADIATION MONITOR A-7 1-RA-90-59A RX BLDG AREA RAD MON HIGH RAD C-5 O-RA-123A CCS LIQ EFF MON HIGH RAD PANEL O-XA-55-12D, UNIT 2 AND PLNT LIQ DISCH RADMON A-4 2-RA-90-106A CNTMT BLDG LWR COMPT AIR MON HIGH RAD B-1 2-RA-123A CCS LIQ EFF MON HIGH RAD B-3 2-RA-90-59A RX BLDG AREA RAD MON HIGH RAD Page 16 of 45
SON I RCS LEAK AND LEAK SOURCE IDENTIFICATION AOPRR.05 Rev. 7 3.1 Symptoms (cont'd)
B. Deviations or unexpected indication on any of the following may indicate an RCS leak:
- 1. Rising charging flow to maintain pressurizer level.
- 2. Rising makeup flow to VCT.
- 3. Rising containment pressure.
- 4. Dropping pressurizer level.
- 5. Rising pressurizer level (leak in vapor space).
- 6. Rising containment moisture.
- 7. Rising containment radiation.
- 8. Rising HELB recorder indications.
C. Any of the following automatic actions may indicate an RCS leak:
- 1. CCS high radiation will cause the surge tank vent FCV-70-66 to close.
- 3. RCP thermal barrier cooling coil leak will cause CCS flow to isolate on high differential flow.
- 4. Ice condenser door(s) opening.
3.2 Entry Conditions None END OF SECTION Page 17 of 45
-5 HLC12-02.BNK Page: 48 Wednesday, October 09,2002 @ 11:45 AM
- 47. AOP-R.04-B.5 001
".47. AOP-R.04-B.5 001 1" Unit Two is operating at 30 % RTP when Loop 3 RCP trips.
Which ONE X(of the following describes the initial unit response? n (Ad1 A 'here will be no flow in loop 3 , 7 ' r B. **-eacto- tý-eeewf-i-t4.oo3 flow will decrease. #M*4 ' 4 t r 40 C. A--;~ac trip wanthere will be no flow in loop 3.
,vD. ' :-- loop 3 flow will decreasep dt# z','4C" le Ccd'v-L. Of9 ctalL ocnO4V(ý eova< wn r Z44u&;
A. Incorrect, the reayXor will not trip as initial unit response, the unit is belowP-8 logic.
AOP-R.04 will reFise a manual reactor trip. p/A B. Incorrect, the reaotvr will not trip as initial unit response, the unit is below P-8 logic.
AOP-R.04 will ree a manual reactor trip.
X'ZO1/
s5t C. Inrrrect, the reactqr i/below P-8 auto trip logic and the initial unit response is NOT a reactor trip. )4ffected loop flow decreases to zero and then reverses to 20-25% of nominprflw caused by the delta-P across the reactor vessel.
<2 Therefore no is,inorrect.
D. Correct, affected loop flow decreases and then reverses to 20-25% of nominal flow caused by the delta-P across the reactor vessel.
K/A[CFR]: 015 A [2.1/2.9] [43.5]
Reference:
AOP-R.04, RCP system description, abnormal operations.
LP/Objective: OPL271RCP B.13 History: Procedure bank Level: Any sis Comments: FHW 12/02 015 AA2.07
- 47. Unit Two is operating at 30% RTP when Loop 3RCP trips. Which ONE of the following describes the initial unit response?
A. There will be no flow in loop 3; Core DP will decrease causing operating loops flow to increase.
B. Loop 3 flow will decrease to approximately 20%; Core DP will increase slightly causing operating loops flow to decrease slightly.
C. There will be negative flow in loop 3; Core DP will increase slightly causing operating loops flow to increase slightly.
D. Loop 3 flow decrease to approximately 20%; Core DP will decrease slightly causing operating loops flow to increase slightly.
Author: rfa
System Operation Abnormal Operations Loss of a RCP below P-8 Upon a loss of a single RCP, plant parameters will change as described in the table below. The loss of two or more RCPs above P-7 and below P-8 results in a reactor trip.
Turbine Power Reduces slightly as Pstm lowers.
power Reactor Power reduces. slightly as turbine Reduces Core AT Increases due to decreased mass flow through the reactor.
Tavg (affected loop) Decreases to Tc due to backflow in the loop caused by the delta-P across the reactor vessel.
Tavg (unaffected loops) Remains the same.
Steam Generator Pressure (affected Decreases due to decrease in affected loop) loop Tavg and hence TSTM .
Steam Generator Pressure (unaffected Decreases since each unaffected loop loops) has to produce more power. Therefore, (Tavg - TSTM) increases. Since Tavg remains the same; TSTM must decrease.
If TSTM decreases; PSTM decreases, AT (affected loop) Decreases due to reduction in steam generation in the loop.
AT (unaffected loops) Increases due to each unaffected loop producing significantly more power.
Loop Flow (affected loop) Affected loop flow decreases to zero and then reverses to 20-25% of nominal flow caused by the delta-P across the reactor vessel.
SG Feedwater Flow (affected loop) Decreases as steam flow from the affected SG decreases.
SG Feedwater Flow (unaffected loops) Increases as steam flow from the un affected SGs increase.
Continued on next page 068d.doc 3-7 Rev 2
6k/
Thursday, October 10, 2002 @ 08:51 AM HLC12-02.BNK Page: 1
- 1. AOP-M.03-B.1 002 Given the following plant conditions:
- Unit Two in MODE 3 for maintenance.
- Panel 0-XA-55-27B-D Annunciator A-4,"MISC EQUIP SUP HDR FLOW LOW", starts alarming.
- Panel O-XA-55-27B-D Annunciator A-6, "LETDOWN HX OUTLET FLOW/TEMP ABNORMAL", starts alarming.
Which ONlý,ýf the following events could cause both alarms to actuate?
",/A. CCS supply header rupture.
B. Letdown HX tube rupture.
C. Loss of seal injection.
D. Loss of charging flow.
A. Correct because CCS supply header rupture causes a loss of cooling to the letdown heat exchanger.
B. Incorrect because the combination of annunciators A-4 and A -6 indicate a CCS failure instead of a letdown heat exchanger failure.
C. Incorrect because the combination of annunciators A-4 and A-6 indicate a CCS failure instead of a loss of seal injection.
D. Incorrect because the combination of annunciators A-4 and A-6 indicate a CCS failure instead of a loss of charging flow.
K!A {CFR}: APE 026 AA2.01 [2.9/3.5] [43.5]? 130, 4& &oA 2.4.45 [3.3/3.6] [43.5],4
References:
AOP-M.03 "Loss of Component Cooling Water' 0-AR-M27-B-D (A-4) 0-AR-M27-B-D (A-6)
LP/Objectives: OPL271 C425, B.1 History: 98 NRC Exam Level: CoWrphensio C Comments: FHW 12/0. 2.4.45
SON LOSS OF COMPONENT COOLING WATER AOP-M.03 Rev. 6 3.0 SYMPTOMS AND ENTRY CONDITIONS 3.1 Symptoms A. Any of the following annunciators may indicate a CCS failure:
I PANEL 0-XA-55-12A, UNIT 1 AND COMMON RADIATION MONITOR B-1i 1-RA-90-123A CCS LIQ EFF MON HIGH RAD I I PANEL O-XA-55-12D, UNIT 2 AND PLT LIQ DISCH RADMON B-1 2-RA-90-123A CCS LIQ EFF MON HIGH RAD Page 48 of 62
4 (A-4)
Source Setpoint SR 1204 MISC EQUIP 2-FS-70-164A 300 gpm decreasing SUP HDR FLOW Retransmitted to U-2 LOW SER 2179 Probable 1. Component cooling pump tripped.
Causes 2. Inadvertent MOV closure.
- 3. Manual valve misaligned.
- 4. Pipe Break.
Corrective [1] CHECK miscellaneous equipment supply header flow by Actions observing 12-FI-70-1164A.
[2] CHECK indicating lights to verify CCS pump running.
[3] VERIFY proper valve alignment in accordance with 2-SO-70-1, Component Cooling Water System Train A.
[4] IF component cooling water loss is indicated, THEN GO TO AOP-M.03, Loss of Component Cooling Water.
NOTE The 50.59 safety assessment requirement for disablement of this annunciator is not applicable during unit outage. (SA/SE requirement of 0-PI-OPS-055-001.0 Appendix A.
[5J IF component cooling water low flow due to Unit outage, THEN DISABLE this annunciator in accordance with 0-PI-OPS-055-001.0 Appendix A.
References 45B655-27BD-0, 471601-70-43, 47W610-70-2, 47W659-1 SQN O-AR-M27-B-D Page 6 of 38 0 2 Rev. 11
6 (A-6)
Source Setpoint SER 1230 LETDOV N HX 2-FS-70-1 90 100 gpm decreasing OUTL_ET SER 1231 132°F increasing FLOW/M rEMP 2-TS-70-1 91 Retransmitted to U-2 ABNOF tMAL SER 2201 and 2202 Probable 1. CCS pump off.
Causes 2. Heat exchanger isolation valves closed or misaligned.
- 3. Low CCS water pressure.
- 4. 2-TCV-70-192 failed closed.
- 5. Pipe break.
Corrective [1] CHECK letdown heat exchanger outlet flow by observing Actions [2-FI-70-190] and outlet temperature by observing r2-TI-70-191 .
[2] MONITOR CVCS letdown temperature on [2-TI-62-781.
[3] IF r2-TCV-70-1921 not operating properly in AUTO, THEN PLACE in MANUAL, AND ATTEMPT temperature control.
[4] VERIFY low pressure letdown valve maintaining letdown pressure at - 340 psig.
[5] IF Letdown heat exchanger temperature cannot be controlled, THEN EVALUATE switching to excess letdown in accordance with 2-SO-62-6, Excess Letdown.
[6] VERIFY proper valve alignment in accordance with 2-SO-70-1, Component Cooling Water System Train A.
NOTE 2-TCV-70-192 fails open on loss of air~or high-high temperature of 124 0F on 2-TIS 62-79, but could fail closed on a circuit failure of loop 2-T-62-78.
[7) IF this annunciator clears and re-alarms, AND [2-TCV-62-791 has diverted to the VCT, THEN NOTIFY Instrument Maintenance to troubleshoot instrument loop 2-T-62-78.
[8] IF pipe break is suspected, THEN GO TO AOP-M.03, Loss of Component Cooling Water.
[9] START additional CCS pumps in accordance with 2-SO-70-1, as necessary.
[10] IF maintenance is required, THEN NOTIFY Work Control Group for support.
[11] IF alarm in due to Letdown HX being out of service, THEN Continued operation with alarm in is acceptable.
References 45B655-27BD-0, 47B601-70-49, 47W610-70-2, 47W859-3.
SoN -AR-M27-B-D Page 8 of 38 0, 2 Rev. 11
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 41
- 40. AOP-l.04-B.2 001 During Mode 1 operation, with the Pressure Control Selector Switch selected to the NORMAL position (340/334), which ONE-Kof the following statements describes the effects of Pressurizer Pressure Control Channel I failing LOW? Assume no operator action is taken.
VA. All heaters on; LOW PRESSURE alarm; PORV-340A blocked; RCS pressure rises; PORV-334 openat 2335 psig.
B. Sprays full on; PORV-340A opens; Low Pressure Reactor Trip at 1970 psig; Low Pressure SI at 1870 psig.
C. All heaters on; LOW PRESSURE alarm; PORV-340A setup for open signal; RCS pressure increases to 2335 psig.
D. Sprays full on; PORV-340A blocked; all heaters on at 2210 psig; heaters modulate to maintain RCS pressure between 2210 and 2218 psig.
A. Correct per reference.
B. Incorrect, PZR press channel failing low will not fully open the sprays.
C. Incorrect, channel I failing low does not set up PORV 340A for opening.
D. Incorrect, PZR press channel failing low will fully open the sprays.
eo -K/A[CFR]: 010 2.4.47 [3.4/3.7] [41.10 43.5]
dI.027 AK2.03 (2.6 - 2.8) 027 AK2-.33(.62)
Off 027 AA1.01 (4.0 - 3.9) e#_t027 AA2.16 (3.6 - 3.9)
AA10 A2 1030
....169 (4.9439)
Reference:
AOP-I.04 LP/Objective: OPL271 PZRPCS B.11 History: Procedure bank. old Bank Number PL-0914 Level: An<sis L" Comments: FHW 12/02 010 2.4.47.
44
AOP-I.04 PRESSURIZER INSTRUMENT MALFUNCTION Rev.05 SON Rev. 5 Page 1 of 1 APPENDIX H PRESSURIZER PRESSURE CONTROL
.3T, C=,*w..
4,
(.4
- 0. 0 V C a' (U 0 C) 0 (U (U C C III 0I 0 o
'C
+
Page 56 of 57
Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 40
- 39. AOP-C.04-B.5 008 Due to conditions causing Control Room Inaccessibility, the main control room has been abandoned and all checklists are complete. Hot Standby conditions are being maintained from the Auxiliary Control Room when 2B-B 6.9-kV S/D Bd. experiences a loss of voltage.
Which ONE X of the following is the expected response by the operating staff for this condition?
A. Check Diesel Generators running and all auto-connected to the 6.9-kV S/D Boards.
/B. Ensure Diel Generators running and dispatch perso0manually close the Bf connected to 2B-B b'AB-B 6.9-kV S/S oard Emergency Feeder Bkr. Ver 6.U-kV S/ID d.
C. Verify Diesel Generators running and dispatch personnel to manually close all 6.9-kV S/D Bd. Emergency Feeder Breakers. Verify all D/Gs connected to the 6.9-kV S/D Boards.
D. Verify D/Gs running and 2B-B D/G auto-connected to the 2B-B 6.9-kV S/D Board.
A. Incorrect - Diesels will not auto connect after checklist have been completed.
Switches are in auxiliary position.
B. Correct - Diesels will not auto connect after checklist have been completed.
Switches are in auxiliary position and breaker must be closed locally.
C. Incorrect - Only ON E SD BD has lost voltage.
D. Incorrect - Diesels will not auto connect after checklist have been completed.
Switches are in auxiliary position.
K/A {CFR): [1.8/2.9] {43.4}
References:
AOP-C.04 section 2.1 [13] note LP/Objectives: OPL271C423 B.5 History: Procedure bank Level: Memory Comments: FHW 12/02 2.3.3; since the SRO is the procedure reader it is his responsibility to ensure actions are completed.
SHUTDOWN FROM AUXILIARY CONTROL ROOM AOP-C.04 SQN Rev. 5 STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.1 Control Room Abandonment (cont'd)
NOTE Operations Duty Specialist phone number is 751-1700.
- 12. NOTIFY Shift Manager to activate REP USING EPIP-1, Emergency Plan Classification Matrix.
NOTE 1 D/G breakers will not automatically close when transfer switches are in AUXILIARY. Ifoff-site power is lost, D/Gs should automatically start but D/G breakers must be manually closed to energize shutdown boards.
NOTE 2 Breaker position indication lights inside ACR are not available until associated transfer switches are placed in AUX position.
- 13. MONITOR shutdown boards ENERGIZED a. ENSURE affected shutdown from start busses. [0-L-4] board(s) energized from D/G as follows:
- 1) ENSURE D/Gs running. [0-L-4]
- 2) NOTIFY operator in 6.9KV Shutdown Board Rm to ensure affected shutdown board(s) energized from D/G USING Checklist 3 (Unit 1) or Checklist 4 (Unit 2).
- 3) NOTIFY operator at D/G Bldg to ensure D/G ERCW valves OPEN.
- b. WHEN personnel available, THEN REFER TO AOP-P.01, Loss of Offsite Power.
Page 9 of 174
Wednesday, October 09,2002 @ 11:45 AM HLC12-02.BNK Page: 38
- 37. AOP-C.01-B.5 011 Given the following:
- Unit 1 was operating at 83% power.
- A control rod dropped into the core
- One negative rate trip status light came on
- One minute later, a second rod dropped into the core
- NO additional rate trip status lights are lit Which ONE of the following actions is required?
A. Recover the dropped rods using AOP-C.01, Rod Control System Malfunctions.
B. Shutdown the reactor using AOP-C.03, Emergency Shutdown.
",'C.Trip the reactor and go to E-0, Reactor Trip or Safety Injection.
D. Reduce power and enter AOP-I.01, Nuclear Instrument Malfunction. S A. Incorrect per reference. "
B. Incorrect per reference.
C. Correct per reference.
D. Incorrect per reference.
K/A: 2.4.11 (3.4/3.6 {41.10, 43.5}
w'4.4 (4.0/4.3) {41.10, 43.2}
003 AK3.04 (3.8/4.1) {41.5, 41.101 003 AA2.03 (3.6/3.8) {43.5}
001 K1.05 (4.5/4.4) {41.2-9}
Reference:
AOP-C.01, Section 2.3, page 13 Objective: OPL271AOPCO1, b.5 Level: Memory History: Procedure bank Comments: FHW 12/02 003 .4.4.
AOP-C.O1 Rev. 8 SON ROD CONTROL SYSTEM MALFUNCTIONS
'-I STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 2.3 Dropped Rod(s)
- 1. DETERMINE number of dropped rods TRIP the reactor and LESS THAN TWO. GO TO E-0, Reactor Trip or Safety Injection. [C.5]
- 2. PLACE rod control in MAN.
NOTE: Reducing turbine load to match T-Ref to T-avg. is the preferred action.
Rod withdrawal is allowed.
- 3. ESTABLISH T-avg. at T-Ref by either of the following methods:
"* REDUCE turbine load OR
"* WITHDRAW control rods NOTE: If the reactor is subcritical, retrieval of the dropped rod is NOT the conservative action to take. Opening the reactor trip breakers is preferred.
- 4. CHECK reactor CRITICAL. TRIP the reactor and
[C.2] [C.5] GO TO E-0, Reactor Trip or Safety Injection.
Page 13 of 39
dL Wednesday, October 09, 2002 @ 11:45 AM HLC12-02.BNK Page: 24
- 24. 069 2.1.14 001 question has reference material attached. SThis Unit 1 is at 100% RTP. 0-SI-SLT-000-1 60.0, "Primary Containment Total Leak Rate,"
has just been completed for Unit 1. System Engineering reports to the Unit 1 US that the combined bypass leakage rate is 65 scfh. La is 225.17 scfh.
Which one of the following should be notified of the surveillance results?
A. No additional notifications are required
,/B. Plant Manager, Senior Vice President Nuclear Operations, Duty Plant Manager C. Site Vice President Nuclear Operations, Duty Plant Manager, Operations Duty Specialist D. Plant Manager, Site Vice President Nuclear Operations, Operations Duty Specialist.
A. Incorrect. Analysis of the data (65/225.17 > .25) reveals that Unit 1 should enter LCO 3.6.1.2. This is a four hour LCO. Therefore, SPP-3.5 Appendix C requires site notifications for an unplanned entry into a LCO with a time duration of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or less.
B. Correct. Condition describes entry into a LCO with duration of less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Therefore, the Plant Manager, Senior Vice President Nuclear Operations since LCO less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and the Duty Plant Manager are required to be notified. A TS safety limit was not exceeded; therefore, the ODS is not required to be notified. Per the
- note, the Plant Manager notifies the Site Vice President.
C. Incorrect. Justification per "B" above.
D. Incorrect. Justification per 'B" above.
K/A[CFR]: 069 2. l 4 [2.5/3.3] [43.5]
References:
TS 3.6.1.2 and SPP-3.5 Rev. 10, Appendix C LP/Objective: OPL271C168 B.3 History: New Question Level: Comprehension.
Comments: FHW 12/02 069 2.1.14; Written by Jim Kearney (9/11/02).
Provide TS 3.6.1.2 and SPP-3.5 Rev 10, Appendix C.
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT BYPASS LEAKAGE
> LIMITING CONDITION FOR OPERATION 3.6.1.2 Secondary Containment bypass leakage rates shall be limited to a R207 combined bypass leakage rate of less than or equal to .0.25 percent L for all penetrations that are secondary containment BYPASS LEAKAGE PATHS TO THE IR221 AUXILIARY BUILDING when pressurized to P..
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION: R1 80 With the combined bypass leakage rate exceeding 0.25 L for BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING, restore the combined bypass leakage rate from BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING to less than or equal to 0.25 L within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
"Enter the ACTION of LCO 3.6.1.1, "Primary Containment" when Secondary Containment Bypass Leakage results in exceeding the overall containment leakagc rate acceptance crittia.
i JR221 February 5, 1996 SEQUOYAH - UNIT 1 3/4 6-2 Amendment No. 12, 71, 176, 203, 217
CONTAINMENT SYSTEMS 1R180 SECONDARY CONTAINMENT BYPASS LEAKAGE SURVEILLANCE REQUIREMENTS RIS0 rates shall be demonstrated:
4.6.1.2 The secondary containment bypass leakage to the auxiliary building shall be
- a. The combined bypass leakage rateequal to 0.25 L, by applicable Type B determined to be less than or and C tests in accordance with thewhich Containment Leakage Rate Test for penetrations are not individually testable; program, except shall be determined to have no penetrations not individually testable soap bubbles while the detectable leakage when tested with (12 psig) during each Type A test.
containment is pressurized to P, are sealed with fluid from a seal
- b. Leakage from isolation valves that to the provisions of Appendix J, system may be excluded, subject combined leakage rate provided Section III.C.3, when determining the are pressurized to at least 1.10 P. (13.2 the seal system and valves is adequate to maintain system psig) and the seal system capacity spray system and RHR pressure (or fluid head for the containment 48A, 48B, 49A and 49B) for at spray system valves at penetrations least 30 days.
are not applicable.
- c. The provisions of Specification 4.0.2 R221 February 5, 1996 12, 71, 101, 102, Amendment No.
3/4 6-3 127, 130, 176, 217 SEQUOYAH - UNIT 1
(ý ;TANDARD REGULATORY REPORTING REuUIREMENTS SPP-3.5 Rev. 10 PiuGRAMS AND Page 28 of 45 PROCESSES APPENDIX C Page 1 of 1 SITE EVENT NOTIFICATION MATRIX Notification Requirements Event/Condition Plant Manager SVP, Nuclear Duty Plant Ops. Duty Spec.
Operations Manager (ODS)
Reactor/Turbogenerator trip, unscheduled unit power reduction, Yese Yes Yes* Yes or nonscheduled unit shutdown; and when unit is restored to full service.
Unplanned entry into a Limiting Condition for Operation with time Yes* Only for Yes* Only when TS duration of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or less. duration of 24 safety limits hours or less exceeded Classification of the Radiological Emergency Plan (REP) at any Yes* Yes Yes t Yes entry level.
Personnel injuries that are potential lost-time injuries, serious Yes* Yes Yes* Yes recordable injuries, and injuries requiring hospital admittance or transport to an offsite medical facility.
Death of any person as a result of injuries received on site or due Yes* Yes Yes* Yes to medical problems occurring while onsite.
Release of oil or hazardous materials to the discharge canal, No No Yes Yes ponds or river and violations of the NPDES permit.
Any event which may be newsworthy to the public. (1) Yes* Yes Yes* Yes NRC 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> phone calls. Yes* Yes Yes* Yes Any unusual radiation exposure to personnel. Yest No Yes* Yes Accidental, unplanned or uncontrolled radioactive release. Yest No Yes* Yes Any reasonable threat to generation. Yes Yes Yes No Outage critical path extensions exceeding 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Yes Yes Yes No Any reactivity event or unplanned reactivity change. Yes Yes Yes No NOTE:
(1) Consider items specified in Appendix D step 2.2 of this procedure.
Plant Manager should ensure the Site Vice President (VP) is notified. If Site VP can not be contacted within 30 minutes, the SVP, Nuclear Operations should be notified. Ifthe SVP, Nuclear Operations can not be reached, the Chief Nuclear Officer should be notified.