ML030830101
| ML030830101 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/04/2003 |
| From: | Lanksbury R NRC/RGN-III |
| To: | Skolds J Exelon Generation Co, Exelon Nuclear |
| References | |
| 50-254/02-301, 50-265/02-301 | |
| Download: ML030830101 (206) | |
Text
INITIAL SUBMITTAL OF THE WRITTEN EXAMINATION FOR THE QUAD CITIES EXAMINATION - DEC 2 00 2 (Par I o-2)
QUAD-MT'ES2-002N*-C-E-XM REFERENCE7S PROVIDED
- 7 - QCOA 1400-02
- 16 - QCOP 0201-11 SRO
- 78 (103) - T.S.3.3.6.1
- 81 (106)-SAF 1.6, 1.7 ( *Jh-
- 85 (110) - TS. 3.3.1.2
- 86 (111)-SAF 1.5
- 89 (114) - OP-AA-1 06-101, ATACHMENT A
- 90 (115)-T.S. 3.8.4 (116) -QGA 100
- ;:#91
- 92 (117) - QGA DETAIL K (DSI
- 93 (118) - T.S. 3.4.7 & 3.4.8
- 94 (119) - QGA 200
- 95 (120) - QGAý2 /O0
- 98 (123)- ODCM, FIGURE 1 II U
I I 4-s
-a VX r*
n =, E ho)
E
ES-401 BWR SRO Examination Outline Form ES-401 -1 Facility: /, 1/2%- Date of Exam: i,7-,-a*aExam Level: S L_2__
_____ IK K K K K A AA A G Total
- 1. 1 3 4 22 Emergency&& _._
Abnormal 2 33 17 Plant Evolutions Tier Totals / )3 43 1 _2 ;2_ 2 1 3 23
- 2. 2 Plant 2 / Il / 2 I l/i I 13 Systems 3 l o C) _I 4 Tier 3" L/. 3* L q** 40 Totals
- 3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4
_ _ __ _q_ -771 17 Note: 1. Ensure that at least two topics from every WA category are sampled within each tier (i.e., the "Tier Totals" in each K/A category shall not be less than two).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final exam must total 100 points.
- 3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
4 .. Systems/evolutions within each group are identified on the associated outline.
- 5. The shaded areas are not applicable to the category/tier.
6.* The generic K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.
- 7. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the SRO license level, and the point totals for each system and category. KfAs below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above.
13 of 46 NUREG-1021, Revision 8, Supplement 1 NRC COPY #1A
Facility:
( cities Quaa BWR SRO ( ,nination Outline Printed: 10/14/2(
ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Groun 1
]*
ull!
l~~V A*
01-1 E/APE # E/APE Name / Safetv Fuinction 01-1
/. . Functionc
....... .. . . . SName. Imp. Points 295003 Partial or Complete Loss of A.C. Power /6 X 2.1.14 - Knowledge of system status criteria which 3.3 1 require the notification of plant personnel.
295003 Partial or Complete Loss of A.C. Power! 6 X AK1.05 - Failsafe component design 2.7 1 295006 SCRAM / I X AA2.05 - Whether a reactor SCRAM has occurred 4.6* 1 295007 High Reactor Pressure / 3 X AA1.04 - Safety/relief valve operation: Plant-Specific 4.1
- 1 295009 Low Reactor Water Level / 2 X AA2.03 - Reactor water cleanup blowdown rate 2.9 1 295009 Low Reactor Water Level / 2 X AK2.03 - Recirculation system 3.2 1 (295010 High Drywell Pressure / 5 X AK1.03 - Temperature increases 3.4 1 (295010 High Drywell Pressure / 5 X AK3.05 - Temperature monitoring 3.4 1
- 95013 High Suppression Pool Temperature / 5 X AK1.01 - Pool stratification 2.6 1 950 13 High Suppression Pool Temperature / 5 X 2.4.4 - Ability to recognize abnormal indications for 4.3 1 system operating parameters which are entry-level conditions for emergency and abnormal operating
- -! -procedures.
295015 Incomplete SCRAM / I X AK2.01 - CRD hydraulics 3.9 1 295015 Incomplete SCRAM / I X 2.2.22 - Knowledge of limiting conditions for 4.1 1 operations and safety limits.
295016 Control Room Abandonment / 7 X AK2.02 - Local control stations: Plant-Specific 4.1* 1 295023 Refueling Accidents / 8 X AK1.03 - Inadvertent criticality 4.0 1 295023 Refueling Accidents / 8 X AA2.02 - Fuel pool level 3.7 1 I
BWR SRO.( .nination Outline Printed: 10/14/2(
Facility: QuadCities ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Groun 1 Fr* U1111 I*L..*-*
I I1 - -r - - -r -- -ba1II V,.,.r -
401-1 E/APE # E/APE Name / Safety Function I01-1
.r Imp. Points 295024 High Drywell Pressure / 5 K1 0K2 iK3 Al IA2 G I KA3T X EK3.02 - Suppression pool spray operation: 3.8 1 Plant-Specific 295025 High Reactor Pressure / 3 X EA1.04 - HPCI: Plant-Specific 3.9 1 295026 Suppression Pool High Water Temperature / 5 - X EK2.04 - SPDS/ERIS/CRIDS/GDS: Plant-Specific 2.8 1 295026 Suppression Pool High Water Temperature / 5 X EK3.01 - Emergency/normal depressurization 4.1 1 295030 Low Suppression Pool Water Level / 5 X 2.1.33 - Ability to recognize indications for system 4.0 1 operating parameters which are entry-level conditions for technical specifications.
a95030 Low Suppression Pool Water Level / 5 X EA2.04 - Drywell/ suppression chamber differential 0 3.7 1 pressure: Mark-I&II
-295031 Reactor Low Water Level / 2 X EK3.04 - Steam cooling 4.3* 1
- 95037 SCRAM Condition Present and Reactor Power X EA1.1 1- PCIS/NSSSS 3.6 1 Above APRM Downscale or Unknown / 1 295038 High Off-Site Release Rate / 9 X EA2.01 - tOff-site 4.3* 1 295038 High Off-Site Release Rate / 9 X EA1.03 - Process liquid radiation monitoring system 3.9 1 500000 High Containment Hydrogen Concentration /5 K.6 - Operation of wet well vent 3.7 1___
K/A Category Totals: 4 4 5 4 5 4 Group Point Total: 26 2
BWR SRO ( Aination Outline Printed: 10/14/; (
Facility: Quaac *ities ES - 401 Emerencv and Abnorm2l Pbrnt Evnluutmnng - Tfi. 1 I I
__,-_-_.................................. .... ., , ,orm ES-401-1 E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Imp. Points 295001 Partial or Complete Loss of Forced Core Flow X AK2.03 - Reactor water level 3.7 1 Circulation / I 295002 Loss of Main Condenser Vacuum / 3 X AA1.07 - Condenser circulating water system 2.9 295004 Partial or Complete Loss of D.C. Power / 6 X 2.2.22 - Knowledge of limiting conditions for 4.1 1 operations and safety limits.
295004 Partial or Complete Loss of D.C. Power/ 6 X AK3.01 - tLoad shedding: Plant-Specific 3.1 1 295005 Main Turbine Generator Trip / 3 X AK3.05 - Extraction steam/moisture separator isolations 2.6 1 J95012 High Drywell Temperature / 5 X AA2.02 - Drywell pressure 4.1 1 0
(95012 High Drywell Temperature / 5 X AK3.01 - Increased drywell cooling 3.6 1
-195020 Inadvertent Containment Isolation / 5 X AK2.12 - Instrument air/nitrogen: Plant-Specific 3.2 1
ý95021 Loss of Shutdown Cooling / 4 X 2.2.25 - Knowledge of bases in technical specifications 3.7 1 1for limiting conditions for operations and safety limits.
295021 Loss of Shutdown Cooling / 4 X AKI.04 - Natural circulation 3.7 1 295022 Loss of CRD Pumps / I X AK1.01 - Reactor pressure vs. rod insertion capability 3.4 1 295028 High Drywell Temperature / 5 X EA2.05 - Torus/suppression chamber pressure: 3.8 1 Plant-Specific 295028 High Drywell Temperature / 5 X EK1.02 - Equipment environmental qualification 3.1 1 295029 High Suppression Pool Water Level / 5 X EA2.03 - Drywell/containment water level 3.5 1 295032 High Secondary Containment Area Temperature / 5 X EK2.02 - Secondary containment ventilation 3.7 1 I
Facility:
(
Quad kLities BWR SRO ( iination Outline Printed: 10/14J, ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / CGrnn 2 S. .......... .... .. .... __r I--
][*^----
I UI IIi I i'*C'* Alfll 4 IU-I E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Imp. Points 295034 Secondary Containment Ventilation High Radiation / X EA2.02 - Cause of high radiation levels 4.2* 1 9
600000 Plant Fire On Site / 8 2.2 25-' nowledge of bases in technical specifications 3.7 limi
,__,.._.o g conditions for operations and safety limits.
K/A Category Totals: 3 3 3 1 4 3 Group Point Total:
-_ij 17 2
BWR SRO E ( aation Outline Printed: 10/1, (.
Facility: Quad Cities ES - 401 Plant Systems - Tier 2 / GrouD 1 Sys/Ev# System/ Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KATopic Imp. Points 203000 RHR/LPCI: Injection Mode (Plant X K2.01 - Pumps 3.5* 1 Specific) / 2 206000 High Pressure Coolant Injection X K2.04 - Turbine control circuits: BWR-2, 3, 4 2.7* 1 System / 2 209001 Low Pressure Core Spray System / 2 X A2.04 - D.C. failures 3.0 1 212000 Reactor Protection System / 7 X K5.02 - Specific logic arrangements 3.4 1 215004 Source Range Monitor (SRM) System X K4.02 - Reactor SCRAM signals 3.5 1 U/7
'215004 Source Range Monitor (SRM) System X K5.01 - Detector operation 2.6 D 1
/7 V215005 Average Power Range Monitor/Local X A3.06 - Maximum disagreement between flow 3.1 1
> Power Range Monitor System / 7 comparator channels: Plant-Specific 216000 Nuclear Boiler Instrumentation / 7 X K6.02 - D.C. electrical distribution 3.0 1 216000 Nuclear Boiler Instrumentation / 7 X A2.10 - Rapid vessel depressurizations 3.5 1 217000 Reactor Core Isolation Cooling X K3.01 - Reactor water level 3.7 System (RCIC) / 2 223001 Primary Containment System and X 2.1.33 - Ability to recognize indications for 4.0 1 Auxiliaries / 5 system operating parameters which are entry-level conditions for technical specifications.
I
BWR SRO E( ,nation Outline Printed: 10/J(.
Facility: Quad Cities ES - 401
________ ....--_ _._ - .- . . . viorm ES-401-1 Sys/Ev# System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KATopic Imp. Points 223001 Primary Containment System and X KI.09 - SBGT/FRVS: Plant-Specific 3.6 1 Auxiliaries / 5 223002 Primary Containment Isolation X 2.2.22 - Knowledge of limiting conditions for 4.1 1 System/Nuclear Steam Supply operations and safety limits.
Shut-Off/ 5 239002 Relief/Safety Valves / 3 X K3.02 - Reactor over pressurization 4.4* 1 239002 Relief/Safety Valves / 3 X A1.03 - Air supply: Plant-Specific 2.9 1 559002 Reactor Water Level Control System / X K 1.15 - Recirculation flow control system 3.2 1 2
'ý61000 Standby Gas Treatment System /9 X A1.01 - System flow 3.1 1 t62001 A.C. Electrical Distribution / 6 X 2.1.33 - Ability to recognize indications for 4.0 1
> system operating parameters which are entry-level conditions for technical specifications.
262001 A.C. Electrical Distribution / 6 X K1.04 - Uninterruptible power supply 3.4 1 264000 Emergency Generators (Diesel/Jet) / 6 X K4.06 - Governor control 2.7 1 264000 Emergency Generators (Diesel/Jet) / 6 X A4.02 - Synchroscope 3.4 1 290001 Secondary Containment / 5 X K6.04 - Primary containment system 4.1 1 290001 Secondary Containment / 5 X A3.02 - Normal building differential pressure: 3.5 1 1- 1Plant-Specific 2
( BWR SRO E.( nation Outline Printed: 10/1. (
Facility: Quad Cities ES - 401 Plant Systems - Tier
- I G*rnnn 1 1 T- -r r Ii . .rormi T ______________________________________ _____ r5-40u-j 1
Sys/Ev # System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points K/A Category Totals: 3 2 2 2 2 2 2 2 2 1 3 Group Poir t Total: 23
& ZL.....L.L...L....1....JIJL _______________________________________ _______
3
BWR SRO E_ ( ation Outline Printed: 10/1. (
Facility: Quad Cities ES- 401 Plant.S.. ste...
. . T.ie 2.I . n- if o rm E S$-4 0 1-1 Sys/Ev # System/ Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 201006 Rod Worth Minimizer System X A1.03 - Latched group indication: 3.0 1 (RWM) (Plant Specific) / 7 P-Spec(Not-BWR6) 201006 Rod Worth Minimizer System X A3.02 - Verification of proper functioning/ 3.4 1 (RWM) (Plant Specific) / 7 operability: P-Spec(Not-BWR6) 202001 Recirculation System / I X K2.02 - MG sets: Plant-Specific 3.3 1 214000 Rod Position Information System / 7 X K3.01 - RWM: Plant-Specific 3.2 1 214000 Rod Position Information System / 7 X A4.02 - Control rod position 3.8* 1 7
0}15002 Rod Block Monitor System / 7 X K6.05 - LPRM detectors: BWR-3, 4, 5 3.1 1
ý30000 RHR/LPCI: Torus/Suppression Pool X K5.06 - Heat exchanger operation 2.6 1 Spray Mode / 5
ý245000 Main Turbine Generator and X 2.1.33 - Ability to recognize indications for 4.0 Auxiliary Systems / 4 system operating parameters which are entry-level conditions for technical specifications.
286000 Fire Protection System / 8 X K5.02 - Effect of Halon on fires: Plant-Specific 2.6 1 290003 Control Room HVAC / 9 X 2.4.30 - Knowledge of which events related to 3.6 1 system operations/status should be reported to Ioutside agencies.
290003 Control Room HVAC / 9 X K1.05 - Component cooling water systems 3.0 1 I
( BWR SRO E ( aation Outline Printed: 10/1 (
Facility: Quad Cities ES - 401 Plant Sv~tem* - Tipr 2 I Ctruun S....... __....... .......... V- rorm ES-40U -1 Sys/Ev # System / Evolution Name K1 K2 K3 K4 KS K6 Al A2 A3 A4 G KA Topic Imp. Points 300000 Instrument Air System (IAS) / 8 X K4.03 - Securing of IAS upon loss of cooling 2.8 1 water 300000 Instrument Air System (JAS) / 8 X A2.01 - Air dryer and filter malfunctions 2.8 1 K/A Category Totals: 1 1 1 1 2 1 1 1 1 1 2 Group Point Total: 13 2
( BWR SRO E_ ( iation Outline Printed: 10/1, (
Facility: Quad Cities ES - 401 Plant.. m
................. rorm ES-40I-I Sys/Ev # System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 201003 Control Rod and Drive Mechanism / 1 X A1.01 - Reactor power 3.8 1 215001 Traversing In-Core Probe / 7 X 2.4.6 - Knowledge symptom based EOP 4.0 1 Smitigation strategies.
215001 Traversing In-Core Probe / 7 X K4.01 - Primary containment isolation: 3.5 1 Mark-I&II(Not-BWR1) 288000 Plant Ventilation Systems / 9 X K1.06 - Plant air systems 2.7 1 K/A Category Totals: 1 0 0 1 0 0 1 0 0 0 1 Group Point Total: 4 0n C) 0
-u
-K I
Generic Knowledge ap'
( ' bilities Outline (Tier 3)
Printed: 10/14/2002 BWR SRO Examination Outline Form ES-401-5 Facility: Quad Cities Generic Category KA KA Topic Imp. Points Conduct of Operations 2.1.25 Ability to obtain and interpret station reference materials such as graphs, monographs, 3.1 1 and tables which contain performance data.
2.1.22 Ability to determine Mode of Operation. 3.3 1 2.1.17 Ability to make accurate, clear and concise verbal reports. 3.6 1 2.1.27 Knowledge of system purpose and/or function. 2.9 1 Category Total: 4 Equipment Control 2.2.11 Knowledge of the process for controlling temporary changes. 3.4* 1 2.2.27 Knowledge of the refueling process.
3.5 1 2.2.1 Ability to perform pre-startup procedures for the facility, including operating those 3.6 1 controls associated with plant equipment that could affect reactivity.
2.2.3 (multi-unit) Knowledge of the design, procedural, and operational differences between 3.3 1 units.
Category Total: 4 Radiation Control 2.3.1 Knowledge of 10 CFR 20 and related facility radiation control requirements. 3.0 1 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against 3.3 1 personnel exposure.
2.3.11 Ability to control radiation releases. 3.2 1 2.3.9 Knowledge of the process for performing a containment purge. 3.4 1 Category Total: 4
Generic Knowledge at, \bilities Outline (Tier 3)
(. Printed:
10/14/2002(
BWR SRO Examination Outline Facility: Quad Cities Form ES-401-5 Generic Category KA KA Topic Imp. Points Emergency Plan 2.4.34 Knowledge of RO tasks performed outside the main control room during emergency 3.6 1 operations including system geography and system implications.
2.4.32 Knowledge of operator response to loss of all annunciators.
3.5 2.4.31 Knowledge of annunciators alarms and indications, and use of the response instructions.
3.4 2.4.25 Knowledge of fire protection procedures.
3.4 2.4.26 Knowledge of facility protection requirements including fire brigade and portable fire 3.3 fighting equipment usage.
Category Total: 5 Generic Total: 17
ES-401 BWR RO Examination Outline Form ES-401 -2 Facility: Ci,!hDate of Exam: h2-c_-ag.Exam Level:
eitJ,-j t O TierTirGruIGroup
- K/A Category 111Point PointsG Pon K K K KK A A A A Total 3 4 5 6 1 2K4 3 ,
- 1. 1 I *- *31i 13 Emergency& 2 Abnormal 2 19 Plant 3 4 Evolutions Tier 7736 Totals 1 13 23 -3 3 2 313 28 Pnt 2 _ A 0 - -_ 19 Systems 3 I Q/ _ 00 0 o o 4 Tier 5
Totals " 6$* 5 ,
- l5 3 51
- 3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4
_ _ _ _ _ _ _ _ _ _ 3 13 Note: 1. Ensure that at least two topics from every K/A category are sampled within each tier (i.e., the 'Tier Totals" in each K/A category shall not be less than two).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-0 from that specified in the table based on NRC revisions. The final exam must total 100 points.
- 3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
- 4. Systems/evolutions within each group are identified on the associated outline.
- 5. The shaded areas are not applicable to the category/tier.
6.* The generic K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.
- 7. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the SRO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above.
19 of 46 NUREG-1021, Revision 8, Supplement 1 NRC COPY #1A
( .lties Quau BWR RO , ilnation Outline Printed:
Facility: 10/14L(
ES - 401 Emerigency and Abnormal Plant Evoluitionn - Tipr 1 I ( .n... 1 Emergency
. ....................... - Tir...
..... 1 /9- 1 r orm EN-401-2 E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Imp. Points 295005 Main Turbine Generator Trip / 3 X AK3.05 - Extraction steam/moisture separator isolations 2.5 1 295006 SCRAM / I X AA2.05 - Whether a reactor SCRAM has occurred 4.6* 1 295007 High Reactor Pressure / 3 X AA1.04 - Safety/relief valve operation: Plant-Specific 3.9 I 295009 Low Reactor Water Level / 2 X AK2.03 - Recirculation system 3.1 1 295010 High Drywell Pressure / 5 X AK1.03 - Temperature increases 3.2 1 295010 High Drywell Pressure / 5 X AK3.05 - Temperature monitoring 3.5 1 t95015 Incomplete SCRAM/ 1 X AK2.01 - CRD hydraulics 3.8 1
,J95015 Incomplete SCRAM / 1 X 2.2.22 - Knowledge of limiting conditions for 3.4 1 operations and safety limits.
It95024 High Drywell Pressure / 5 X EK3.02 - Suppression pool spray operation: 3.5 1 Plant-Specific 295025 High Reactor Pressure / 3 X EAI.04 - HPCI: Plant-Specific 3.8 1 295031 Reactor Low Water Level / 2 X EK3.04 - Steam cooling 4.0 1 295037 SCRAM Condition Present and Reactor Power X EA1.11 - PCIS/NSSSS 3.5 1 Above APRM Downscale or Unknown / 1 500000 High Containment Hydrogen Concentration / 5 X EK3.06-Operation of wet well vent 3.1 K/A Category Totals: 1 2 5 3 11 Group Point Total: 13 I
BWR ROý( Aination Outline Printed:
Facility: Quai. ities 10/14L(
ES - 401 Emereencv and Abnormal Plant Evnlutinns - T;- 1 I/ 4..
I_____ - ----.- --. 'a.. '.lup Form ES-401-2 E/APE # E/APE Name / Safetv Function
~
Wi ~ AlAIC
~ ~ k A1C o Im . Points 295001 Partial or Complete Loss of Forced Core Flow X AK2.03 - Reactor water level Circulation / I 3.6 1 295002 Loss of Main Condenser Vacuum / 3 X AA1.07 - Condenser circulating water system 3.1 1 295003 Partial or Complete Loss of A.C. Power / 6 X AK1.05 - Failsafe component dei -a 2.6 1 I
295004 Partial or Complete Loss of D.C. Power / 6 X AA2.02 - Extent of partial or complete loss of D.C.
nx-ý~
3.5 1 X nu'F I
295004 Partial or Complete Loss of D.C. Power 6 X AK3.01 - tLoad shedding: Plant-Specific 2.6 1 e295008 High Reactor Water Level / 2 X AA1.05 - RCIC: Plant-Specific 3.3 1 (95012 High Drywell Temperature / 5 X AK3.01 - Increased drywell cooling 3.5 1
-295013 High Suppression Pool Temperature / 5 X AKI.01 - Pool stratification 2.5 1
'It
)295013 High Suppression Pool Temperature / 5 X 2.4.4 - Ability to recognize abnormal indications for 4.0 1 system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.
295016 Control Room Abandonment / 7 X AK2.02 - Local control stations: Plant-Specific 4.0* 1 295016 Control Room Abandonment / 7 X AA2.02 - Reactor water level 4.2* 1 295020 Inadvertent Containment Isolation / 5 X AK2.12 - Instrument air/nitrogen: Plant-Specific 3.1 1 295022 Loss of CRD Pumps / 1 X" I X I I I I A [1 Al -ieacAui D - pressure--. vs. rod insertion capability 1 3.3 I 1
I
Facility:
( BWR RO ( aination Outline Printed: 10/14/2(
ES - 401 I
Em r~n I
v ll IDIill tv=*E. VULI, ons - iier I Group L Form ES-401-2 E/APE # E/APE Name / Safety Function KA Topic Imp. Points 295026 Suppression Pool High Water Temperature / 5 X EK2.04 - SPDS/ERIS/CRIDS/GDS: Plant-Specific 2.5 1 295026 Suppression Pool High Water Temperature / 5 X EK3.01 - Emergency/normal depressurization 3.8 1 295028 High Drywell Temperature / 5 x EKI.02 - Equipment environmental qualification 2.9 I i i 1 i -i i 295034 Secondary Containment Ventilation High Radiation / x EA2.02 - Cause of high radiation levels 3.7 9 1 T
295038 High Off-Site Release Rate / 9 X EA1.03 - Process liquid radiation monitoring system 3.7 x 1 I
600000 Plant Fire On Site / 8 x AA2. 10 - Time limit of long-term-breathing air system z 2.9 1 for control room K/A Category Totals: 4 4 3 3 4 1 Group Point Total: 19 2
( BWR RO I( Aination Outline Printed:
Facility: Quda ,lties 10/14/2(
ES - 401 Emereencv and Abnnrmal Plant *Vt~lln _ Tim- 1 I * .... "
_.._. . . . . . . __ .. . X t I U .3j Fo irm ES-401-2 E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Im 295021 Loss of Shutdown Cooling / 4 x I AKl.04 - Natural circulation p.
3t.6 Points 1
295023 Refueling Accidents / 8 x 295032 High Secondary Containment Area Temperature / 5 X EK2.02 - Secondary containment ventilation 3.6 295035 Secondary Containment High Differential Pressure! X EA1.02 - SBGT/FRVS 3.8 1 5
K/A Category Totals: 2 1 0 1 0 0 Group Point Total: 4 I
( BWR RO E, ( iation Outline Printed: 10/I, (
Facility: Quad Cities ES - 401 Plant.... Sy .t.m... ... .. .orm E -401-2 Sys/Ev # System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KATopic Imp. Points 203000 RHR/LPCI: Injection Mode (Plant X K2.01 - Pumps 3.5* 1 Specific) / 2 203000 RHR/LPCI: Injection Mode (Plant X K3.03 - Automatic depressurization logic 4.2*
Specific) / 2 206000 High Pressure Coolant Injection X K2.04 - Turbine control circuits: BWR-2, 3, 4 2.5*
System / 2---
206000 High Pressure Coolant Injection X 2.1.32 - Ability to explain and apply system 3.4 System / 2 limits and precautions.
T209001 Low Pressure Core Spray System / 2 X A2.04 - D.C. failures 2.9 "209001 Low Pressure Core Spray System / 2 X A4.05 - Manual initiation controls 3.8 1 211000 Standby Liquid Control System / 1 X K6.03 - A.C. power 3.2 211000 Standby Liquid Control System / I X A2.08 - Failure to SCRAM 4.1*
212000 Reactor Protection System / 7 X K5.02 - Specific logic arrangements 3.3 215003 Intermediate Range Monitor (IRM) X K6.05 - Trip units 3.1 System / 7 215003 Intermediate Range Monitor (IRM) X 2.1.32 - Ability to explain and apply system 3.4 System / 7 limits and precautions.
215004 Source Range Monitor (SRM) System X K4.02 - Reactor SCRAM signals
/7 3.4
( BWR RO Ei (I iation Outline Printed: 10/1, (
Facility: Quad Cities ES - 401 Plant SvSte--* - T.,- I / l*_ I..
I_____ I-..~..
1.'. .7 AV~'U 1 Form ES-401-2 Sys/Ev # System / Evolution Name K1 K3 K2 K4 K*
KA Topic iI K2 K K61Al L7I IAI IAAII~ Imp. Points 215004 Source Range Monitor (SRM) System X K5 .01 - Detector operation 2.6 1
/7 1 1 1 1 1 1 1 1 1 1 1 19 215005 Average Power Range Monitor/Local X A3.06 - Maximum disagreement between flow 3.0 Power Range Monitor System / 7 1 I I I, Icmparator channels: trlant-Speclifc 216000 Nuclear Boiler Instrumentation / 7 X K6.02 - D.C. electrical distribution 2.8 216000 Nuclear Boiler Instrumentation / 7 X A2. 10 - Rapid vessel depressurizations 3.3 217000 Reactor Core Isolation Cooling X 3.7 K3.01 - Reactor water level System (RCIC) / 2 3.1 217000 Reactor Core Isolation Cooling X 2.1.32 - Ability to explain and apply system 3.4 1 System (RCIC) / 2 limits and precautions.
218000 Automatic Depressurization System / X K5.01 - ADS logic operation 3.8 1 3
218000 Automatic Depressurization System / X A3.01 - ADS valve operation 4.2* 1 3
223001 Primary Containment System and X KI.09 - SBGT/FRVS: Plant-Specific 3.4 1 Auxiliaries / 5 223002 Primary Containment Isolation X A4.04 - System indicating lights and alarms 3.5 System/Nuclear Steam Supply Shut-Off/ 5 239002 Relief/Safety Valves / 3 X K3.02 - Reactor over nressurization A')*
I K3.02 Reactor over nressun7t1An
" -- r ..............
A *
"T.z2, 239002 Relief/Safety Valves / 3 X A 1.03 - Air supply: Plant-Specific 2.8 2
( BWR RO Ex( iation Outline Printed: 10/ 1, (
Facility: Quad Cities ES - 401
-.------ ~- -. "r rorm ES-4U1-2 Sys/Ev # System / Evolution Name K1 IK2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 259002 Reactor Water Level Control System / X K1.15 - Recirculation flow control system 3.2 1 2
261000 Standby Gas Treatment System / 9 X A1.01 - System flow 2.9 1 264000 Emergency Generators (Diesel/Jet) / 6 X K4.06 - Governor control 2.6 1 264000 Emergency Generators (Diesel/Jet) / 6 X A4.02 - Synchroscope 3.4 1 K/A Category Totals: 2 2 3 2 3 3 2 3 2 3 3 Group Point Total: 28 3
( BWR RO Ex( iation Outline Printed: 10/1,(
Facility: Quad Cities ES - 401 Plant -**orm v- E*-401-2 I I 1 T - I - ~ Systems - Tier 2 / CGrnn
~_ _ _
2
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ viio rm E S-4 0 1-2 Sys/Ev # System / Evolution Name K1 K2 K74 KA I*g nts
- .. .. .. . .. , ,I-p c Im p. Poi nts 201003 Control Rod and Drive Mechanism / 1 X A 1.01 - Reactor power 3.7 1 201006 Rod Worth Minimizer System X Al1.03 - Latched group indication: 2.9 1 (RWM) (Plant Specific) / 7 P-Spec(Not-BWR6) 201006 Rod Worth Minimizer System X A3.02 - Verification of proper functioning/ 3.5 1 (RWM) (Plant Specific) / 7 operability: P-Spec(Not-BWR6) 202001 Recirculation System / 1 X I K2.02 - MG sets: Plant-Snecific 'A1)
.214000 Rod Position Informnation System /7 X K3.01 - RWM: Plant-Specific 3.0 1 b 14000 Rod Position Informnation System /7 X A4.02 - Control rod position 3.8* 1
,215002 Ro lc oio ytm/7X K6.05 - LPRM detectors: BWR-3, 4, 5 2.8 1
'220>RRLC:Cotimn pa X K3.02 - Containment/drywell/suppression 3.5 1 System Mode / 5 chamber temperature 230000 RHR/LPCI: Torus/Suppression Pool XK5.06 - Heat exchanger operation 2.5* 1 Spray Mode / 51 262001 A.C. Electrical Distribution / 6 X K1 .04 - Uninterruptible power supply 3.1 1 262001 A.C. Electrical Distribution / 6 X K2.01 - Off-site sources of power 3.3 1 263000 D.C. Electrical Distribution / 6 X(11 DA.2-atr otaeidctr .
anpecific 272000 Radiation Monitoring System / 7 1X K4.01 - Redundancy 2.7 1 I
( BWR RO Ex( iation Outline Printed: 10/4 Facility: Quad Cities ES - 401 Plant Systems - Tier 2 / G~rnnn "7 Plant Svstems - Tier 2 / Grou 2 Form ES-401-2 Svs/Ev # Svstem / Evolution Name
____ .. ...... ...... ___ _C K1 1(4 Ik' kt IAl IA IA2 IAA IC' IJA m,..
I Imp. Points 286000 Fire Protection System / 8 X K5.02 - Effect of Halon on fires: Plant-Specific 2.6 1 290001 Secondary Containment / 5 X K6.04 - Primary containment system 3.9 1 290001 Secondary Containment / 5 X A3.02 - Normal building differential pressure: 3.5 1 Plant-Specific 290003 Control Room HVAC / 9 X KI.05 - Component cooling water systems 2.8 1 300000 Instrument Air System (IAS) / 8 X K4.03 - Securing of IAS upon loss of cooling 2.8 1 water 00000 Instrument Air System (lAS) / 8 X A2.01 - Air dryer and filter malfunctions 2.9 1 K/A Category Totals: 2 2 2 2 2 2 2 1 2 2 0 Group Point Total: 19 2
( BWR RO Ex ( ation Outline Printed: 10/1. (
Facility: Quad Cities ES - 401 Plant Systems - Tier 2 / Groun 3 I I I- r 1 1______________________________ I ES-401-2 Sys/Ev # System / Evolution Name K1 K2 K3 1K4 K15 TI- - - - -I - I.~
4K5K6. 4~
IAl IA2IIA- .V.v lAA IG lI WA n..
L Imp. Points 215001 Traversing In-Core Probe / 7 x K4.01 - Primary containment isolation: 3.4 1 234000 Fuel Handling Equipment / 8 X K3.03 - tFuel handling operations 3.1 1 288000 Plant Ventilation Systems / 9 X K1.06 - Plant air systems 2.7 1 288000 Plant Ventilation Systems / 9 X A2.04 - High radiation: Plant-Specific 3.7 K/A Category Totals: 1 0 1 1 0 0 0 1 0 0 0 Group Point Total: 4 I
K Generic Knowledge ar( ' bilities Outline (Tier 3)
BWR RO Examination Outline Printed: 10/14/2002 (
Facility: Quad Cities Form ES-401-5 Generic Category KA KA Topic Iml *. Points Conduct of Operations 2.1.17 Ability to make accurate, clear and concise verbal reports.
5 1 2.1.27 Knowledge of system purpose and/or function. 2, 8 1
2.1.28 Knowledge of the purpose and function of major system components and controls.
3.2 1 Category Tcoral: 3 Equipment Control 2.2.1 Ability to perform pre-startup procedures for the facility, including operating those 3.7 1 controls associated with plant equipment that could affect reactivity.
2.2.3 (multi-unit) Knowledge of the design, procedural, and operational differences between 3.1 units.
z 2.2.34 Knowledge of the process for determining the internal and external effects O) on core 2.8 1 reactivity.
0 "0 Category Total: 3
-< Radiation Control 2.3.11 Ability to control radiation releases. 2.7 1 Knowledge of the process for performing a containment purge.
2.5 1 2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible 2.5 1 levels in excess of those authorized.
Category Total:
3 Emergency Plan 2.4.32 Knowledge of operator response to loss of all annunciators.
3.3 1 2.4.31 Knowledge of annunciators alarms and indications, and use of the response instructions. 3.3 1 2.4.25 Knowledge of fire protection procedures.
2.9 1 2.4.26 Knowledge of facility protection requirements including fire brigade and portable fire 2.9 1 fighting equipment usage.
Category Total: 4 Generic Total: 13
( ( (
ES-401 Record of Rejected K/As Form ES-401 -10 Tier / Group Randomly Selected K/A Reason for Rejection SRO 1 / 1 500000 2.1.33 Not applicable at Quad Cities. Randomly replaced by 295023 A2.02 BOTH 1 / 2 295033 EA1.06 Performed by Radiation Protection at Quad Cities. Randomly replaced by 295034 A2.02 RO 2 / 3 233000 K2.02 Double Jeopardy with 203000 K2.01. Randomly replaced by 234000 K3.03 SRO 1 / 2 600 00 2.2.2-**
5 Will use Administrative Technical Requirements (ATRs) in ad of Technical Specifications (TS) since requirements were moved from TS to ATR., , K4*'t I 1-t I i i 1 i I i i i I i I t I i i i t i S.. L7
-i--
NUREG-1021, Revision 8, Supplement 1 LV
EXAMINATION ANSWER KEY 2002 Quad Cities NRC Exam, 1 11 . ID: SR-0302-K21 Points: 1.00 Reactor power is increased from 20 to 100%.
The CRD Flow Control Valve AO 1(2)-0302-06A is in manual.
4
Ž ,
In odrder tm-inta he-O-q 4Presssure.
RDcooJ inagwPiogw
_CJ 0 t, the NSO will have to manually
,con Flow Control Valve (AO 1(2)-0301-06A) vhic;h will
" A.<<
" CRD Drive Water
_ -- .- )
A. CLOSE; AFFECT
- ? .o .. e B. CLOSE; E e e iA C. OPEN; T D. OPEN; AFFEGT 6?
Answer: D
- Question I Detail's Question Type
- Multiple Choice Topic: Question #1 (ROISRO)
System ID: 6204 User ID: SR-0302-K21 Status: Active Must Appear: No Difficulty: 4.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LF-0302 R3 pg 15, fig 4 User Text: 201003 A1.01 /
User Number 1: 3.70 User Number 2: 3.80 Comment: New question Comprehensio ith the FCV in manual, it will] eo be manually opened as Rx pressure increases to maintain the same Cooling water flow. As the FCV is opened, it will affect CRD drive water pressure because it taps off downstream of the FCV.
OPERATIONS of 130 10/15/02
-3 V 2
Content/Skills Activities/Notes Content/Skills V Activities/Notes check valve prevents bleeding down accumulator pressure on a loss of a CRD pump.
- 5. A charging water vent valve is located in the Reactor Building for the purpose of filling the system during an initial start.
G. Flow Control Station
- 1. The flow control station is provided to automatically control SR-0302-K14 the system flow (normally 60 gpm). Two flow control SR-0302-K15 valves are arranged in parallel with manual isolation valves. SR-0302-K16 One flow control valve in service while the other is in **SR-0302-K21 standby. Limit switches mounted on each flow control valve are connected to red and green position indicating lights on the 90X-5 panel.
- 2. The flow control valves are air operated valves which open against spring pressure by air from the instrument air system.
The pneumatic control signal to the valve is selected by a three-way valve which is controlled locally at the flow control station panel. The three way valve can be positioned so the control signal will come from local or remote.
- a. LOCAL - In local the control pneumatic signal is controlled by a manual pressure regulator mounted on the flow control station panel.
- b. REMOTE - In remote the pneumatic signal comes from an E/P converter. The converter receives an electrical signal from the flow controller on the 90X-5 panel.
,4. The flow indicating controller located on the 90X-5 panel SR-0302-K14 has two modes of operation. SR-0302-K15 SR-0302-K16
- a. Automatic **SR-0302-K21
- 1) Selected by pushing the "A" button just below the bargraph, the automatic mode maintains pump flow at a pre-selected setpoint, normally 60 gpm.
- 2) When the CRD FIC is in the automatic mode, the FIC monitors system flow, using a venturi installed in the system upstream of the charging header, and compares it to what the FIC setpoint is adjusted to, then sends a signal to E/P converter to open or close the FCV.
q:\topslp\If-0302.doc Page 14 of 50 NRC COPY #1
Content/Skills Activities/Notes ContentlSkills Activities/Notes
- 3) The automatic flow control setpoint is displayed on the controller and can be changed by pushing the N
setpoint up/down pushbuttons.
- b. Manual Mode
- 1) Selected by pushing the "M" button just below the bargraph.
-2) -When-the-PRJ"FIC is in the manual mode, the PIC---
maintains the same demand signal that was being maintained in the automatic mode when switching from Automatic to Manual mode.
- 3) In order to vary pump flow while in the manual mode, a slide lever is moved to the right (0) to increase flow or the left (C) to decrease flow. Two rate of change positions are available for both increasing and decreasing.
- c. Fail Light A red fail light in the upper right corner of the flow controller indicates when the controller has failed.
- d. Alarm (ALM) Light SR-0302-K14 SR-0302-K15 An amber alarm light is provided below the fail light to SR-0302-K16 indicate one of the following: **SR-0302-K21
- 1) The high or low limit alarm has actuated, or
- 2) The input/output signal line is open, or
- 3) The voltage of the internal data protect battery is low.
- e. Additional Controls Pushbuttons C and PF have no function in this application. Depressing them can confuse the controller's programming and should not be used.
- 4. Local Manual
- a. In this mode the 3-way selector valve is in the LOCAL position.
q:\trtopslp\lf-0302-doc Page 15 of 50 NRC COPY #1
IN 0
CI z
2iii 24~ID: SR-O207-K2o Points: 1.00 LRod step 20 has control rods H-10, F-8, H-6 and K-8 with a rod limit from position 08 to 12.
Control rod H-10 is withdrawn to position 12. vFa e-1; Control rod F-8 is withdrawn to position 10, .... .. /i "
The NSO then selects control rod H-6, which is currently at position 08.
On the RWM display, control rod H-6 will indicate:
A. green.
B. white. ' i[J
, 1" C. red.
D. cyan.
V/
Answer: A Question 2 Details Question Type: Multiple Choice Topic: Question #2 (RO/SRO)
System ID: 9714 User ID: SR-0207-K20 Status: Active Must Appear: No Difficulty: 1.00 Time to Complete: 0 Point Value:
Cross
Reference:
1.00 QCOP 0207-01, R.9, pg. 5 W'A uwt"Y
/ V User Text: 201006 A1.03 User Number 1: 2.90 User Number 2: 3.00 Comment: New question. i er. Control rods are in red when they are out of equer~ce, green when they are in the current latched .lp or selected for rod exercising, white if they are not the rod selected for exercising or not part of the in-sequence step. H-6 should remajp green the enire time.
L - 'U
(/W
~k 0 1~V~vYA?
/
()A,,
OPERATIONS OPERAIONSPa" of 130 10/15/02 I M-5 Co00py
i, QCOP 0207-01 UNIT 1(2)
REVISION 9 k R F.4.a. (cont'd)
(3) The third line contains the correct position for the selected rod according to the current latched step of the RWM. If the selected rod is in the current latched step, a range of allowable positions will be displayed.
- b. Use the lower left area of the Main Display Screen for the following information about current latched step in the sequence: the (1) The first line contains the current latched step in the sequence.
(2) The second line contains the array designation that is to be moved in this step.
(3) The third line contains the range of mQvement allowed for the rods in this step.
- c. Use the center portion of the Main Display Screen ,
for the full core display showing all current rod positions.
(1) Rods that appear green are in the-current-,
latched step. .-- ---
(2) Rods that appear red are rods that have withdraw errors.
(3) Rods that appear magenta are rods that have insert errors.
(4) Rods that appear light blue are rods that have been taken out of service with the RWM.
(5) Rods that appear yellow are rods that have had a substitute value entered for their position.
(6) All other rods will appear white on the Full Core Display.
- 7) The selected rod will be 'displayed in inverse vi:eo to highlight it.
NRC C6PY #1
3 ~ A ID: SR-'0207-1<26 ý0'1'nts: Ab All RWM blocks are enabled.
The NSO is performing QCGP 1-1, NORMAL UNIT STARTUP.
Rod step one contains control rods H-i, F-i, D-2, B-4, A-6, A-8, A-10, B-12, D-14, F-15, H-15, K 15, M-14, P-12, R-10, R-8, R-6, P-4, M-2, K-1.
Control rods H-I, F-1 and D-2 are fully withdrawn.
How would the RWM respond if B-5 pushbutton was depressed and attempted to be withdrawn?
A. RWM select block would prevent rod motion.
B. RWM would allow the rod to be moved until low power setpoint was reached.
C. RWM would prevent the rod from being selected.
D. RWM withdrawal block would prevent rod motion when the control rod reached position 02.
Answer: A Question Type: Multiple Choice Topic: Question #3 (RO/SRO)
System ID: 9715 User ID: SR-0207-K20 Status: Active Must Appear: No Difficulty: 4.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCGP 1-1, R 43, pg. 38-40 User Text: 201006A3.02 User Number 1: 3.50 User Number 2: 3.40 Comment: New question. Higher. When an incorrect rod is selected, the RWM will issue a select block. A withdrawal block will be issued when the OOS rod is withdrawn I notch. The Mode Switch cannot be in SHUTDOWN and moving rods. When in REFUEL, a rod block is issued when the 2nd rod is selected with a rod withdrawn.
OPERATIONS i ' 17/ 10/15/02 aPar-- of 130
QCGP I-1 UNIT 1(2)
REVISION 43
F. PROCEDURE
NOTE I_ using this procedure for Turbine start up only, THEN proceed to step F.7. Mark steps F.1. through F.6 "N/A".
INITIALS F.1. Begin Reactor start up as follows:
- a. Prior to placing Mode Switch in START/HOT STBY, verify all applicable surveillances on Attachment D for MODE 3 or MODE 4 to MODE 2 Transition are current or within the 25% grace period. Unit Supervisor must EVALUATE all previous incomplete steps for Mode Change impact.
NOTE It is permissible to enter MODE 2 to withdraw selected control rods for the purposes of determining the OPERABILITY of the RWM. The RWM must be determined OPERABLE within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn at < 10% RTP. The following steps test the RWM by verifying proper indication of the selection error of at least one out-of-sequence control rod and by verifying the rod block function. This Channel Functional Test (F.l.c - F.l.g) need not be performed if it has been completed within the last 92 days. (H.l.n)
- b. Enter MODE 2 by placing RX MODE SELECT switch in START/HOT STBY.
- c. Verify the following for RWM testing:
(1) RWM Mode Switch in NORMAL position.
(2) NO Control Rod drifts present.
NRC COPY #1 38
QCGP 1-1
'2 UNIT 1 (2)
REVISION 43 F.1. (cont'd) INITIALS
- d. Select the first Control Rod in selected Control Rod sequence.
(1) Record if the READY light is present for the respective RWM computer:
- e. IF "A" RWM READY light is lit, THEN:
(1) Place selection switch to "A".
(2) Depress INITIALIZE button.
(3) Verify insert and withdraw blocks cleared.
(4) Verify No alarm messages are present.
(5) Depress DIAGNOSTIC button.
(6) Verify diagnostic is completed with NO errors.
(7) Verify screen indicates BELOW 20% POWER with blocks enabled.
- f. IF "B" RWM READY light is lit, THEN:
(1) Place selection switch to "B".
(2) Depress INITIALIZE button.
(3) Verify insert and withdraw blocks cleared.
(4) Verify NO alarm messages are present.
(5) Depress DIAGNOSTIC button.
(6) Verify diagnostic is completed with NO errors.
(7) Verify screen indicates BELOW 20% POWER with blocks enabled.
NRC CsqPY #1
QCGP 1-1 UNIT 1 (2)
REVISION 43 F.1. (cont'd) INITIALS
- g. Select an out-of-sequence Control Rod and verify that a Rod Block occurs:
(1) Disable RWM select blocks.
(2) Verify Rod Block resets.
(3) Notch withdraw the out-of sequence Control Rod and verify ROD BLOCK occurs at position 02.
(4) Return Control Rod to position 00.
(5) Select an in-sequence Control Rod.
(6) Enable RWM select blocks.
(7) Verify blocks enabled to 100% power.
(1) Unit Supervisor.
(2) Qualified Nuclear Engineer.
F.2. Verify Nuclear Instrumentation and computer setup as follows:
(1) Verify all operable SRMs are fully inserted in the core.
(a) The above required SRM channels may be reduced to three operable SRM channels with the concurrence of Operations Manager. (H.8.q)
Operations Manager (2) Verify all operable SRMs are indicating
> 3.0 cps OR 0.7 cps with a signal to noise ratio > 20:1 in MODE 2 with IRMs on range 2 or below.
NRC COPY #1
4 D: SRO 22O -K19 0Porniýs: 1i.00' Unit 2 is operating at 100% power in a normal electrical line-up when the reactor auxiliary power transfer fails.
Which of the following components are de-energized?
I A. 2A Recirculation Motor Generator Set B. 2A Condensate/Condensate Booster Pump C. 2B Recirculation Motor Generator Set D. 2B Condensate/Condensate Booster Pump Answer: A V
- uestion 4 Details Question Type: Multiple Choice Topic: Question #4 (RO/SRO)
System ID: 9716 User ID: SR-0202-K1 9 Status: Active Must Appear: No Difficulty: 2.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QOA 6500-01, r. 6, pg 1,2 User Text: 202001 K2.02 User Number 1: 3.20 User Number 2: 3.30 Comment: New question. Higher. T-21 su.pplies Busses 21 and 24, which auto xfer to T-22 on Gen trip. 2A Recirc MG is power from bus 21, 2B from bus 22. 2A and 2B Cond/Cond Booster pumps are powered from Bus 2,3.
OPERATIONS IN [ --- , PaC of 30 l A1 10/15/02
QOA 6500-01 Revision 6 Continuous Use 4 KV BUS 11(21) FAILURE A. SYMPTOMS
- 1. Alarms.
- a. 4 KV MAIN FEED BREAKER TRIP.
- b. 4 KV BUS 11(21) MAIN/RES BKR AUTO CLOSE.
B. AUTOMATIC ACTIONS
- 1. The following equipment will auto-trip if the bus is'
\* de-energized:
- a. Reactor feed pump lA(2A).
- b. Reactor feed pump lC(2C) if it is being fed from Bus 11(21).
- c. Recirculation MG set 1A6
- 2. Unit 2 Only: A Reactor recirc runback to approximately 70% of rated flow will occur if steam flow is greater than approximately 85% of rated and either of the following conditions are present:
- a. An RFP auto-trips after three RFP operations have been established AND Reactor water level drops below the low level alarm setpoint within 45 seconds after the RFP auto-trip.
- b. Less than four Condensate/Condensate Booster Pumps are running AND Total Feedwater Flow is greater than approximately 90% of rated.
C. IMMEDIATE OPERATOR ACTIONS
- 1. None.
D. SUBSEQUENT OPERATOR ACTIONS
- 1. Verify automatic bus transfer to opposite transformer occurred:
- a. IF automatic bus transfer did NOT occur, THEN attempt to close the reserve feed breaker manually unless alarm 901(2)-8 F-3, 4KV BUS OVERCURRENT, is up.
- b. IF breaker tripped, THEN determine the cause and reset trip at Bus 11(21).
NRC C9PY #1
QOA 6500-04 "Revision 11
- d. For Unit 2 only: Check STATOR OUTPUT VOLT at Panel 902-8 AND verify nominal voltage of 18KV (green band).
CAUTION Verifying main generator voltage is essential as ELMS study states 4KV buses will fall below 4000V during single transformer operation during full load conditions.
(1) IF main generator voltage is 17KV or less, THEN notify the Load Dispatcher to increase system voltage and concurrently, with US permission:
(a) Decrease main generator load and/or (b) Increase main generator excitation by increasing VARS and/or (c) Secure any large 4KV loads (Circulating Water Pump, RFP, etc.) as operation will allow.
- 3. Notify Shift Manager to classify the event as a possible E-Plan condition and initiate E-Plan as necessary.
- 4. IF Bus 14(24) is lost, THEN refer to Technical Specification 3.8.1 or 3.8.2 for loss of ability to supply Bus 14-1(24-1) from T12(22).
E. DISCUSSION
- 1. The normal feed for Bus 14a(24 hl--
is Transformer - ..1)4 Transformer 12(22).
- 2. WHEN the bus is de-energized, THEN the following equipment is affected:
- a. Control Rod Drive Pump lB(2B).
- b. Residual Heat Removal System Service Water Pump lC(2C).
- c. Residual Heat Removal System Service Water Pump lD(2D).
- d. Condensate and Condensate Booster Pump lC(2C).
- e. Condensate and Condensate Booster Pump lD(2D).
NRC C(pPY #1
5 ID: SIR-1O-0K-1 P s<19n:
. Points. 1.00 Unit 1 was operating at full power when a plant casualty occurred.
Unit 1 scrammed as a result of the transient.
The Unit NSO noted that UI HPCI started automatically while Ul RCIC remained in a standby lineup as expected.
Both Unit 1 and the 1/2 Emergency Diesel Generators (EDGs) started automatically but the Unit 1 EDG TRIPPED on an overspeed condition.
1)
- 9-,- ' Assuming all equipment was in a normal operating configuration prior to the transient, and that the remaining auto actions ocurred, what is the expected status of Unit 1 RHR pumps? l A&B C&D I'
A. RUNNING OFF 4 B. RUNNING RUNNING C. OFF OFF D. OFF RUNNING Answer: B Qiuestion 5 Details Question Type: Multiple Choice Topic: Question #5 (RO/SRO)
System ID: 9496 User ID: S/R-1000-K19 Status: Active Must Appear: No Difficulty: 4.00 Time to Complete: 2 Point Value: 1.00/
LF-1000, R.6, pg 62 Cross
Reference:
User Text: '203000 K2.01 7 User Number 1:
User Number 2: 3.50 "'
Commertt.., Bank question. Coio . lA andX RHR pumps are powýd-rom Bus 13-1, which never lost power from Transformer 12. The EDGs auto started on 2.5 psig, but would not have loaded to busses 13-1 and 14-1. 1C and ID RHR pumps are powered from bus 14-1, which never lost p6Wer from Transf~rmer 11 then 12.
Content/Skills Activities/Notes.
- 2. The RHR system supports the Primary Containment system by:
- a. Maintaining drywell temperature and pressure within design limits, during LOCA conditions, through the use of drywell and torus sprays.
- b. Maintaining the torus water temperature within its heat capacity limit.
C. Power Supplies **SR-1000-K19 N
- 1. Bus 13-1 (23-1) sunDlies nower to RHR Pumns A and P. SR-1000-K17
- 2. Bus 14-1 (24-1) supplies power to RHR Pumps C and D.
Q: With this arrangement of
/ power supplies, which EDG will
- 3. Bus 13 (23) supplies power to RHR SW Pumps A and B. be feeding the 1A RHR pump during a loss of off-site power?
A: 1/2 EDG.
- 5. MCC 18-lB (28-1B) supplies power to:
- a. Valves MO- 1001-7A and B (Torus Suction)
- b. Valves MO-1001-43A and B (Shutdown Cooling Suction)
C. Valve MO- 1001-1 6A (RHR Hx Bypass)
- d. Valve MO-1001-19A (RHR Loop Crosstie)
- e. Valve MO-1001-50 (Shutdown Cooling Suction) f Valve MO- 1001 -23A (Outboard Drywell Spray)
- g. Valve MO- 1001 -26A (Inboard Drywell Spray)
- h. Valve MO-1001-34A (Torus Cooling & Test Line Isolation)
- i. Valve MO-1001-36A (Torus Cooling)
- j. Valve MO- 1001-3 7A (Torus Spray)
- 6. MCC 19-4 (29-4) supplies power to:
- a. Valves MO-1001-7C and D (Torus Pool Suction)
- b. Valves MO-1001-43C and D (Shutdown Cooling Suction)
- c. Valves MO-1001-16B (RHR Hx Bypass)
- d. Valves MO- 1001-19B (RHR Loop Crosstie)
- e. Valve MO- 1001-50 (Shutdown Cooling Suction)
- f. Valve MO-1001-23B (Outboard Drywell Spray)
Q:\TRNOPSLP\LNF- 1000.doc Page 62 of 81 NRC COPY #1
1D: SR-23Ob-KI9 froints: IoT The HPCI Flow Controller is powered from:
A. 125 VDC.
B. Instrument Bus.
C. Essential Service.
D. 250 VDC.
Answer: C Question 6 Detais ~ I Question Type: Multiple Choice Topic: Question #6 (RO/SRO)
System ID: 9717 User ID: SR-2300-K1 9 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: "1.00 Cross
Reference:
LN-2300, R. 10, pg. 56 User Text: 206000 K2.04 User Number 1: 2.50 User Number 2: 2.70 Comment: New question. Memory. 125 VDC powers the initiation logic, 250 VDC powers the valves, Instrument Bus powers nothing on HPCI.
OPERATIONS ]1 o 130 10/15/02 C Pap f nn
Content/Skills Activities/Notes Activities/Notes unit's HPCI is inop due to not having power to the valves and motors for the support pumps.
"Thenew 2399-40 MOV receives 480V AC power from MCC 19-1(29-1).
- 5. The flow controller is fed from the Essential Service Bus.
On a loss of essential service you lose power to the flow **SR-2300-K23d indicating controller. HPCI is inoperable with no power to **N-2300-K23d the flow indicating controller. HPCI can still be used if speed is controlled manually at HPCI with the local handwheels or can be controlled from the Control Room I
"-,usingthe MSC. -
- 8. The gland seal cooling water pump is powered by MCC 19-1(29-1).
- 10. Slave trip units from the Analog Trip System provide level initiation instruments. These transmitters provide control DCP 9900377 Mechanical Yarways are being room indicators and computer inputs.
replaced by new slave units in
Q:\ KRNOkSLP\LN-2300.doc Page 56 of 61 NRC COPY #1
7r~I;S- 0-2 ID: SR-1400-K26 Points: 1.00 Annunciator 902-3 D-5, CORE SPRAY SYS 2 BUS/LOGIC PWR FAILURE is up on Unit 2.
A casualty occurs on Unit 2 resulting in the following conditions:
RPV water -150 inches and lowering.
Reactor pressure 300 psig and lowering.
Drywell pressure 8 psig and rising.
At this point in this event, predict how the Unit 2 Core Spray system has responded and describe any actions required to restore it.
A. "B" loop will auto-initiate and inject, while "A" loop will NOT auto-initiate, but may be manually started locally.
Manually start the Unit 2 Diesel Generator, verify it energizes Bus 24-1, manually initiate Core Spray Subsystem 2B and restore Core Spray Subsystem 2B 125 VDC control power.
B. "A" loop will auto-initiate and inject, while "B" loop will NOT auto-initiate, and can not be manually started from the Control Room or locally.
Manually start the Unit 2 Diesel Generator, verify it energizes Bus 24-1, manually initiate Core Spray Subsystem 2B and restore Core Spray Subsystem 2B 125 VDC control power.
C. "A" loop will auto-initiate and inject, while "B" loop will NOT auto-initiate, but may be manually started locally.
Manually initiate Core Spray Subsystem 2B and restore Core Spray Subsystem 2B 125 VDC control power.
D. "B" loop will auto-initiate and inject, while "A" loop will NOT auto-initiate, but may be manually started locally.
Manually initiate Core Spray Subsystem 2A and restore Core Spray Subsystem 2B 125 VDC control power.
Answer: C bv bt ly IV r~thl (6C) ({
OPERATIONS Pa z~ of130 J0 I D N11J
Question 71eal Question Type: Multiple Choice Topic: Question #7 (RO/SRO)
System ID: 9718 User ID: SR-1400-K26 Status: Active Must Appear: No Difficulty: 3.50 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOA 1400-02, R. 7 User Text: 209001 A2.04 User Number 1: 2.90 User Number 2: 3.00 Comment: Modified question. Higher. With "2B" 125 VDC out, initiation logic for "B" loop is out but still has power to the pumps and components. "A" loop is unaffected.
The EDG should auto start and load on the loss of Bus 24-1, so it should not need to be manually started.
OPERATIONS Pa of 130 10115/02 1152
QCOA 1400-02 r, XUNIT 1(2)
REVISION 7 Continuous Use CORE SPRAY LOSS OF 125 VDC AUTO INITIATION CONTROL POWER A. SYMPTOMS A.1. Possible alarms:
- a. 901 (2) -3 Panel:
(1) C-5, CORE SPRAY SYS 1 BUS/LOGIC PWR FAILURE (2) D-5, CORE SPRAY SYS 2 BUS/LOGIC PWR FAILURE B. AUTOMATIC ACTIONS B.1. IF loss of 125 VDC Core Spray Subsystem 1(2)A THEN:
- a. Core Spray Subsystem 1(2)A automatic initiation logic is inoperative and Core Spray Subsystem 1(2)A will NOT auto actuate regardless of the condition of Core Spray Subsystem 1(2)B initiation logic.
- b. 1/2 Diesel Generator LOCA auto start signal for Unit 1(2) and output breaker auto transfer signal due to a LOCA signal on Unit 1 (2) is inoperative.
- c. ADS Channel 1(2)A logic and timer is inoperative.
- e. Core Spray Subsystem 1(2)A to CAM Division 1(2)A LOCA isolation signal is inoperative.
- f. RCIC Channel 1(2)A initiation logic is inoperative.
- g. Drywell Coolers 1(2)A, 1(2)B, 1(2)F and Drywell 1(2) Booster Fan LOCA auto trip signal is inoperative.
NRC COPY #1
QCOA 1400-02 UNIT 1(2)
REVISION 7 D. SUBSEQUENT OPERATOR ACTIONS NOTE Restore power to the respective essential bus prior to restoration of control power to the affected core spray subsystem.
IF only Alarm C-5, CORE SPRAY SYS 1 BUS/LOGIC PWR FAILURE, is lit, THEN only Core Spray Subsystem 1(2)A automatic initiation logic is inoperable.
IF only Alarm D-5, CORE SPRAY SYS 2 BUS/LOGIC PWR FAILURE, is lit, THEN only Core Spray Subsystem 1(2)B automatic initiation logic is inoperable.
CAUTION During loss of 125 VDC control power to Core Spray Subsystem 1(2)A/B, manual initiation of either subsystem is possible, however all valve protective interlocks except MO 1(2)-1402-38A/B, CS PMP MIN FLOW VLV, are inoperable.
D.1. IF Core Spray Subsystem I(2)A/B is required for adequate core cooling, THEN manually initiate Core Spray Subsystem l(2)A/B per QCOP 1400-02.
D.2. IF Alarm C-5, CORE SPRAY SYS 1 BUS/LOGIC PWR FAILURE, is lit AND power to Bus 13-1 (23-1) has been restored, THEN restore Core Spray Subsystem 1(2)A 125 VDC control power as follows:
- a. At 125 VDC Turbine Building Dist. Panel 1(2)A-1, verify closed ckt. 05.
- b. For Unit 2 only, at Bus 23-1 verify knife Test Switch (TS) located under black removable covers below protective relays on front of Bus 23-1 Cubicle 09, Component TS-SJ, left box, Test Switch H closed.
NRC COPY #1
QCOA 1400-02 UNIT 1(2)
REVISION 7 D.3. (cont'd)
NOTE Alarm 901(2)-3 D-5, CORE SPRAY SYS 2 BUS/LOGIC PWR FAILURE, will NOT clear unless power to Bus 14-1 (24-1) AND 125 VDC control power to Core Spray Subsystem 1(2)B have been restored.
- d. WHEN 125 VDC control power to Core Spray Subsystem 1 (2)B has been restored, THEN verify Alarm 901(2)-3 D-5, CORE SPRAY SYS 2 BUS/LOGIC PWR FAILURE, clears.
D.4. IF Core Spray Subsystem 1 (2)A/B can NOT be restored, THEN:
- a. Notify Shift Manager to classify event as a possible GSEP condition and initiate GSEP as necessary.
- b. Perform QCAP 0230-19.
- c. Initiate corrective action.
E. DISCUSSION This procedure covers the abnormal condition for loss of VDC power to Core Spray Subsystem 1(2)A and Core System 125 Subsystem 1(2)B. The main entry condition are the Alarms 901(2)-3 C-5, CORE SPRAY SYS 1 BUS/LOGIC PWR FAILURE, indicate loss of Core Spray Subsystem 1(2)A or D-5, to SYS 2 BUS/LOGIC PWR FAILURE, for Core Spray Subsystem CORE SPRAY 1(2)B.
However, these alarms can also be entry conditions loss of Bus 13-1 (23-1) or Bus 14-1 (24-1) respectively. for the the intent of this procedure to restore power to It is respective essential bus first the (major abnormal condition) per QOA 6500-5/6, THEN restore control power to the affected Core Spray Subsystem (less severe abnormal condition).
Loss of 125 VDC control power to Core Spray Subsystem renders automatic initiation logic inoperative. Manual 1(2)A/B initiation and control of Core Spray Subsystem 1(2)A/B is still possible, however WITHOUT the protective valve interlocks except MO 1(2)-1402-38A/B, CS PMP MIN FLOW VLV.
NRC COPY #1
QCOA 1400-02 UNIT 1(2)
REVISION 7
- _27-- P&lDs:
- a. M-36 (M-78), Diagram of Core Spray Piping.
G.3. Drawings:
- a. 4E-1345 (4E-2345), Schematic Control Diagram 4160 V. Bus 13-1 Stand By Diesel 1/2 Feed Bkrs Unit 1 (Unit 2).
- b. 4E-1346 (4E-2346), Schematic Control Diagram 4160 V. Bus 14-1 Stand By Diesel 1(2) Feed Bkrs Unit 1 (Unit 2).
- c. 4E-1430 sh 1 (4E-2430 sh 1), Schematic Diag Core Spray System I & II Unit 1 (Unit 2).
- d. 4E-1430 sh 2 (4E-2430 sh 2), Schematic Diag Core Spray System I & II Unit 1 (Unit 2).
- 9. 4E-1655D (4E-2655K), 4160 V. Switchgear Bus 13-1 (23-1) Cubicle 2 (9) Internal Schematic & Device Location Diag.
4160 V. Switchgear Bus 14-1
(24-1) Cubicle 9 (2) Internal Schematic & Device Location Diag.
G.4. Manuals:
None.
G. 5. Procedures:
- a. QCOP 1400-02, Core Spray Manual Initiation.
- b. QCAP 0230-19, Equipment Operability.
NRC COPY #1
(ell lb:. 81~81 oint ri Unit I has experienced a loss of 125 VDC bus 1B.
A casualty occurs on Unit 1 resulting in:
Reactor water is +19" and lowering.
Reactor pressure is 850 psig and lowering.
Drywell pressure is 4 psig and rising.
At this point in this event, predict how, if at all, the Unit 1 Core Spray system has responded.
A. "B" loop will auto-initiate and run on minimum flow, while "A" loop will not auto initiate, but may be manually started locally.
B. "A" loop will auto-initiate and run on minimum flow, while "B" loop will not auto initiate, but may be manually started locally.
C. "A" loop will auto-initiate and run on minimum flow, while "B" loop will not auto initiate, and can not be manually started from the Control Room or locally.
D. "B" loop will auto-initiate and run on minimum flow, while "A" loop will not auto initiate, and can not be manually started from the Control Room or locally.
Answer: B OPERATIONS Page: 1 of 2 09/24/02 NRC COPY #1
"/1 Associated objective(s):
SR-1400-K18 (Freq: LIC=I)
LIST the plant systems which are supported by the following systems and DESCRIBE the nature of support:
- a. Core Spray 125vdc logic circuit
- b. ECCS Keep Fill System SR-1400-K23 (Freq: LIC=B)
Given a Core Spray System operating mode and various plant conditions, PREDICT how the Core Spray System will respond to the following support system failures:
- a. Loss of 125vdc
- b. Loss of 4160vac
- c. Loss of 480vac
- d. ECCS suction strainer clogging SR-1400-K26 (Freq: LIC=B)
EVALUATE given Core Spray System key parameter indications and/or responses depicting a system specific abnormality/failure and DETERMINE a course of action to correct or mitigate the following abnormal conditions:
- a. Loss of 125vdc power to Core Spray logic
- b. 4 and 25 valves open simultaneously
- c. Core Spray pump trip
- d. Abnormal Core Spray discharge header pressure
- e. Core Spray System fails to start automatically
- f. Inadvertent Core Spray System start
- g. ECCS suction strainer clogging Question Type: Multiple Choice Topic: ILT. 11683 System ID: 6090 User ID: 81181 Status: Active Must Appear: No Difficulty: 0.00 Time to Complete: 2 Point Value: 1.00 Cross
Reference:
User Text: 209001 K4.08 User Number 1: 3.80 User Number 2: 4.00 Comment: With "l B" 125 VDC out, initiation logic for "B" loop is out but still has power to the pumps and components. "A" loop is uneffected.
OPERATIONS Page: 2of2 OPERATIONS Page: 2 of 2 09/24/02 NRqC COPY #1
t,.
16ý liil* Poiflts$:ý'1.00 Unit 1 has experienced a loss of 125 VDC bus lB.
A casualty occurs on Unit 1 resulting in:
Reactor water is +19" and lowering.
Reactor pressure is 850 psig and lowering.
Drywell pressure is 4 psig and rising.
At this point in this event, predict how, if at all, the Unit I Core Spray system has responded.
A. "B"loop will auto-initiate and ru n on minimum flow, while "A" loop will not auto initiate, but may be manually started locally.
B. "A" loop will auto-initiate and ru n on minimum flow, while "B"loop will not auto initiate, but may be manually started locally.
C. "A" loop will auto-initiate and run on minimum flow, while "B"loop will not auto initiate, and can not be manually started from the Control Room or locally.
D. "B" loop will auto-initiate and run on minimum flow, while "A'loop will not auto initiate, and can not be manually started from the Control Room or locally.
A~n-swer_:- B r)Dr:DATInKI0
("lPl= I* B.TI '*KI* ....
Page: 1 otf 09/24/02 NRC COPY #1
'-1 Associated objective(s):
SR-1400-K18 (Freq: LIC=I)
LIST the plant systems which are supported by the following systems and DESCRIBE the nature of support:
- a. Core Spray 125vdc logic circuit
- b. ECCS Keep Fill System SR-1400-K23 (Freq: LIC=B)
Given a Core Spray System operating mode and various plant conditions, PREDICT how the Core Spray System will respond to the following support system failures:
- a. Loss of 125vdc
- b. Loss of 4160vac
- c. _.Loss of 480vac
- d. ECCS suction strainer clogging SR-1400-K26 (Freq: LIC=B)
EVALUATE given Core Spray System key parameter indications and/or responses depicting a system specific abnormality/failure and DETERMINE a course of action to correct or mitigate the following abnormal conditions:
- a. Loss of 125vdc power to Core Spray logic
- b. 4 and 25 valves open simultaneously
- - Core spray pump trip
\-t - - d. Abnormal Core Spray discharge header pressure
- e. Core Spray System fails to start automatically
- f. Inadvertent Core Spray System start
- g. ECCS suction strainer clogging uestionIDeal Question Type: Multiple Choice Topic: ILT.1 1683 System ID: 6090 User ID: 81181 Status: Active Must Appear: No Difficulty: 0.00 Time to Complete: 2 Point Value: 1.00 Cross
Reference:
User Text: 209001 K4.08 User Number 1: 3.80 User Number 2: 4.00 Comment: With "IB" 125 VDC out, initiation logic for "B" loop is out but still has power to the pumps and components. "A" loop is uneffected.
OPERATIONS - Page: 2 of 2 09/24/02 NRC COPY #1
_ ID: SR-05OO-K67 ~Points: ' 1.00 Ifthe reactor mode switch is in RUN, which ONE of the following conditions will cause either a half scram or a full scram?
A. Reactor power is 45%, Main Steam Isolation Valves IA & 1D are both closed.
B. Reactor power is 45%, Turbine Stop Valves 2 & 3 are both closed.
C. Reactor power is 10%, Main Steam Isolation Valves 1C & 2D are both closed.
D. Reactor power is 10% , Turbine Stop Valves 3 & 4 are both closed.
Answer: C Question 8-Ptails Question Type: Multiple Choice Topic: Question #8 (RO/SRO)
System ID: 230 User ID: SR-0500-K07 Status: Active Must Appear: No Difficulty: 3.25 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LF-0500, R. 6, pg 44-51 User Text: 212000 K5.02 User Number 1: 3.30 User Number 2: 3.4 Comment: ank quetio Application. MSIVs A and D and TSV 2 n 1 and 4 meet the "5" alive requirement.
MSIVs C and D do not meet this, so a 1/2 scram would result. At 10% power, Turbine Stop Valves would not cause a 1/2 scram.
OPERATIONS 10/15/02 r-\Ln C PaC" of 13
7-Content/Skills Activities/Notes During a scram condition or a condition where an alarm is received for a scram condition but the scram did not occur, the perform a visual inspection of Panels 901(2)-15 and 17 to determine the status of each RPS Trip Channel or Trip Logic relay. During normal operation all of the Trip Channel relays are energized. When the relays are energized, all of the relay contacts (fingers) are made up, or pulled in away from the operator, when facing the relay. When a trip condition exists, the relays are deenergized and the contacts will be released (separated). Some of the relays have "b" contacts that are open when the relay is energized and closed when the relay is deenergized. These will be identifiable because there is a contact point in front of the relay contact finger. Typically the "b" contacts control the associated annunciator and computer point. When an abnormal condition exists such as an expected annunciator or specific plant response is not observed, the operator can walkdown the panels and find the relay contact(s) that may be faulty.
A. Automatic RPS Operation
- 1. Turbine Stop Valve (10% Closure) **SR-0500-K15
- L-0500-K15 (Figure 0500-8, 0500-9, E.S. 4E-1465) SR-0500-K04 SR-0500-K07 a: Closure of the turbine main stop valves (MSVs) and SR-0500-K08 subsequent failure of the bypass valves to open with the L-0500-K08 reactor at power can result in a significant addition of **S-0500-K25 positive reactivity to the core as the pressure rise collapses steam voids. The MSV closure initiates a Show Figures 0500-8 and 9.
scram earlier than either the neutron flux or high reactor Refer to print 4E-1465 as pressure scrams to provide a satisfactory margin below necessary to discuss the MSV the core thermal hydraulic safety limit. This is trip circuitry.
considered an anticipatory scram. This means that this scram is to protect the reactor whenever it is sensed that Q: What is the purpose of the its link to the heat sink is in the process of being MSV closure scram?
removed A: To anticipate the pressure, neutron flux, and fuel cladding
- b. Each of the four MSVs are equipped with two position surface heat flux increase caused limit switches for a total of eight switches. There are by a rapid closure of the MSVs two different MSV limit switches assigned to each Trip and failure of the bypass valves Logic (Al, A2, B1, & B2) The valve limit switches are to open.
set at < 10% closure from full open, instead of < 100% of stroke, to provide for the effects of thermal expansion changing the valve/switch orientation. This setting maximizes the reliability of the valve closure signal.
q:\tmopslp\lf-0500.doc Page 44 of 91 NRC COPY #1
Content/Skills Activities/Notes ContentlSkills ActivitieslNotes
- f. An easy way to remember this logic is to substitute the Another way to remember the letters A, B, C, D, for MSV numbers. A slight logic is "Add up to 5 and stay re-arrangement of the logic (physically, not functionally) alive (i.e., 1 & 4, 2 & 3) then spells BADC & CADB from valve letter BA-DC &
CA-DB, in Trip Logics A-2, A-i; B-2, B-1. Closing Q: When is the MSV closure only one MSV in any Trip Channel pair of valves could scram bypassed?
not result in a half-scram. A: < 45% power (325 psig first stage pressure).
- g. This scram is bypassed when Reactor power is less than 45%, which corresponds to a pressure of 325 psig as sensed by the turbine first stage pressure. A closure at power levels less than this does not constitute a threat to the integrity of any barrier to the release of fission products and also allows system startup.
- h. This is accomplished by a pressure switch (PS) off four pressure detectors for first stage pressure. PS-504A feeds Trip Logic A1, 504C feeds A2, 504B feeds B1, and 504D feeds B2. When first stage pressure reaches 325 psig decreasing, PS-504A opens a contact in RPS Trip Channel Al which deenergizes relay 123A. This shuts two series contacts in the A l Trip Logic that bypass the I
"MSVclosure contacts and the Turbine Control Valve Fast Closure contacts, preventing Trip Actuators 108A and C from deenergizing upon closure of these valves.
This prevents a half-scram signal and ultimately a full scram signal from being initiated from valve closure.
- 2. Turbine Control Valve (CV) Fast Closure (Generator-Load **SR-0500-K15 Reject) (Figure 0500-8 & E.S. 4E-1465) **L-0500-K15 SR-0500-K04
- a. This scram anticipates the rapid increase in pressure and SR-0500-K07 neutron flux which results from a fast closure of the CVs SR-0500-KO8 due to a load rejection and postulated subsequent failure L-0500-K08 of the bypass valves to open. Again it also serves to **S-0500-K25 protect the reactor from an imminent loss of its heat sink.
- b. Load Rejection is the condition where a greater than 40%
mismatch exists between the generator stator amps and the turbine crossover pressure. This indicates a difference between the work energy into the turbine and the work energy removed from the generator.
- ~ -,-.I-M -. --
q:\trnopslp\if-0500.doc Page 46 of 91 NRC COPY #1
Content/Skills Activities/Notes Content/Skills Activities/Notes 3 SDV High Level (40 Gallons) (E.S. 4E-1464, Sh. 1, 1465 & **SR-0500-K15 1467, Sh. 3) (Figure 0500-8) **L-0500-K15
-- ' - -- e ...... 7 ---
SR-0500-K04
- a. As discussed i Section III, the SDIV inates a scram SR-0500-K07 while an adequate volume is available to receive the SR-0500-K08 scram discharge water, so as to ensure that all operable L-0500-K08 CRDs can be fully inserted. Each SDIV has six level **S-0500-K25 sensors on it and two sensors that detect a failed instrument. Show Figure 0500-8. Refer to prints 4E-1464, Sh. 1, 1465, and
- 1) One thermal switch for hi level alarm. 1467, Sh. 3, as necessary to discuss SDV High-High trip and
- 2) One thermal switch for rod block.
Bypass function.
- 3) Two thermal switches that detect a failed high level instrument.
- 4) Two thermal switches for scram.
- 5) Two Barton D/P cells for scram.
A thermal switch for a failed high instrument, a thermal switch for high level scram, and a D/P cell for high level scram feed each RPS Trip Channel leg.
- b. Trip System operation is as follows:
Assume a high level condition is detected by thermal switch (SWPT2) 82AX2. This will open its associated contact in the Al Trip Channel, deenergizing relay 100A which will open its associated contact in the A l Trip Logic. This will deenergize the Al Trip Actuators 108A and 108C.
q:\trnopslp\lf-0500.doc Page 48 of 91 NRC COPY #1
/0 Content/Skills Activities/Notes Content/Skills Activities/Notes
- 4. Condenser Low Vacuum (21-inch Hg)
(E.S. 4E-1464, Sh. 1, and 1465) (Figure 0500-8)
- a. Loss of con enser vacuum occurs when the condenser **SR-0500-K15 can no longer handle the heat input or a leak occurs. **L-0500-K15 Loss of condenser vacuum initiates a closure of the SR-0500-K04 MSVs and turbine bypass valves which eliminates the SR-0500-K07 heat input to the condenser. This results in a rapid SR-0500-K08 pressure transient on the reactor vessel, neutron flux rise L-0500-K08 and an increase in surface heat flux will occur. To **S-0500-K25 prevent the cladding safety limit (MCPR) from being exceeded the low condenser vacuum scram was added to Show Figure 0500-8. Refer to anticipate the MSV closure. prints 4E-1464, Sh. 1, and 1465 as necessary to discuss the low
- b. With the reactor mode switch in the run position, the condenser vacuum trip.
scram will occur at 21" Hg vacuum, the MSV closure at 20" Hg vacuum, and the turbine bypass valves close at 7" Hg vacuum.
- c. Trip System operation is as follows: Q: When will the main turbine trip on low vacuum?
Four vacuum switches are provided to monitor the A: 20" Hg vacuum.
condenser vacuum. Sections A and C of the condenser each have one switch. Two switches are provided for Q: Why is the scram setpoint Section B. RPS Trip System A uses switches from higher than the turbine trip
'-, - -'t. condenser sections A and B. RPS Trip System B is setpoint?
controlled by the other switch on the B condenser section A: To avoid the pressure and and the switch on Section C. Trip Channel Al is fed by power transient that may occur switch 503A, A2 from 503C, B1 from 503B, B2 from when the turbine trips.
503D. If a low vacuum condition is sensed by vacuum switch 503A, it opens a contact in the Al Trip Channel which deenergizes relay 101A. The associated contact in the Al Trip Logic opens, deenergizing Trip Actuators 108A and 108C.
q:\tmopsip\lf-0500.doc Page 50 of 91 NRC COPY #1
SID: 12434 Points: 1.00, Which conditions will cause a Half Scram OR a Full Scram to be generated by t Reactor Protection System?
A. Mode switch is in STARTUP; Condenser Vacuum is 20. 5nches Hg; and CHANNEL A/B CONDENSER LOW VACUUM annun .tor is ALARMED.
B. Mode switch in STARTUP; SDV High Water Leve ypass Keylock Switch is in BYPASS; and SDV HI-HI LEVEL annunciator i LARMED.
C. Mode switch is in RUN; Main Steam Isolati Valves 1A and 2D are CLOSED; and reactor power is 20%
D. Mode switch is in RUN; Stop valve # and #3 are CLOSED; andLr actor power is 30%.
Answer: B
'Qistion 5 Details Question Type: Multiple Choice Topic: LORTB System ID: 7821 User ID: / 124434 Status: Active Must Appear: No Difficulty: 0.00 Time to Complete" 6 Point Value: / 1.00 Cross
Reference:
User Text: // 212000 A2.19 User Number 1: 3.80 User Num 6 er 2: 3.90 Cýpy N OPERATIONS 10/15/02 IFD
A reactor scram occured on Unit 2 approximately 1 minute ago.
The scram has NOT been reset.
The NSO can verify all rods in by noting that individual rod position is indicating:
A. an orange double dash.
B. a green double dash.
C. a green 00.
D. an orange 00.
Answer: B Que~stion 9"lefa`i ls Question Type: Multiple Choice Topic: Question #9 (RO/SRO)
System ID: 9719 User ID: SR-0280-K20 Status: Active Must Appear: No Difficulty: 2.25
-Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LIC-0280, R. 6, pg. 7 User Text: 214000 A4.02 User Number 1: 3.80 User Number 2: 3.80 Comment: New Question. Memory. Post scram indication is a green double dash until the scram is reset, at which time the indication turns to an orange double 00 and then to a green 00.
OPERATIONS OPERAIONS PaaNill
" of 130 10/15/02 AL
ýýOpy
Content/Skills Activities/Notes
- 2. Reed switches numbered S49 to S52 are used to provide
-information as follows:
- a. Reed switch S49 closes simultaneously with S48 to illuminate the red "full out" digital indication for a rod at the 48" position.
- b. Reed switch S50 is associated with the rod "overtravel" out annunciator. This switch should not be closed unless the, control rod and CRD become uncoupled.
- c. Reed switch S51 picks up after a scram to illuminate ihe green "full in" digital indication for a rod at the overtravel-in-position.
- d. Reed switch S52 closes nearly simultaneously with S00 to illuminate the green "full in" digital indication for a rod at the "00" position.
- 3. Colored Digital Indications **SR-0280-K14 Red (full out) and green (full in/overtravel in) digital indication show up on the rod position full core digital displays. Colored indications are not present on the four rod display, only on the full core display.
- 4. Translation Electronics **SR-0280-K22
- SR-0280-K05
- a. Each position probe (i.e., each CRD) has a printed circuit card in the 90X-27 back panel located in the auxiliary electric room. This circuit card translates the reed switch closures to a numeric readout.
If an electronic malfunction is detected, an "RPIS INOP" trip is generated, indicating that the RPIS data may not be correct. An INOP is caused by any of the following:
(1) Invalid probe data; (2) Card pulled; (3) Loss of power supply; (4) Internal logic stall.
An "RPIS INOP" trip causes a select block and an annunciator alarm on the 90X-5 panel.
Q \TR,(*)PS.P*tIC-028O dx. Page 7 of 32 NRC COPY #1
The operator is withdrawing a control rod which is part of the current latched step.
The limits of the step and the bounds of the control rod being withdrawn is 00 - 48.
The operator withdraws the rod one notch and notices that the selected rod indicates ?? on the RWM display.
Which of the following statements best describes the RWM system condition as it stands right now?
A. The RWM system will immediately block all movement of the rod that indicates
?? until a substitute position is entered. No other rods are effected by this event.
B. The RWM immediately declares the rod OOS and allows the operator to continue with rod movement on the next rod in the sequence.
C. The RWM will immediately initiate a full core scan and if proper position information is not obtained on the next scan, the RWM will consider itself failed and block all rod movements.
D. The rod is treated just like a withdraw error. Insert and withdrawal blocks are applied to all other rods and a withdrawal block is applied to the selected rod once it reaches a known position.
Answer: D
,Question 16betails Question Type: Multiple Choice Topic: Question #10 (ROISRO)
System ID: 9720 User ID: SR-0280-K22 Status: Active Must Appear: No Difficulty: 3.25 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LIC 0207-01, R.6, pg. 31 User Text: 214000 K3.01 User Number 1: 3.00 User Number 2: 3.20 Comment: Bank question. Application. Dist #1 & 2 - You are still allowed to move the affected rod. Dist #3 - The RWM does not automatically put rods OOS. With a loss of RPIS for a given rod position, the RWM will display a OPERATIONS 10/15/02 Nr5 CPa of 130
Content/Skills .. I '..
Activities/Notes
"(1) Rods in the latched group are colored green.
All other rods will be colored white. All ro'ds in the array are expected to be driven to the 00 or full in position prior to proceeding to any rod in the next array. A withdraw block will be issued if the operator attempts to withdraw a rod unless that rod is at an unknown position, indicated by'??'. A rod at unknown position will be allowed to be inserted or withdrawn until a valid position is reached, at which time the withdraw block will be issued.
If the operator selects a rod in the next valid array, its rods will become the latched group. /
If any rods of prior arrays are at positions other than 00, they will be indicated as withdraw errors and colored red, but rod blocks will not be applied. If a rod, not in the ,
.\next array, is selected, it will be treated as a nelect error and its movement blocked b IVoth "inert and withdraw blocks.
(2) When all rods in the current array have been SR-0207-K14 position to 00, the function will change the
.atch-edgroup to the next array. When all of the arrays have been addressed by the operator the function will not allow any further rod movement of rods. The operator must then exit the function and resume operation using to normal sequence to reposition the rods to normal operating conditions. The operator may exit the Power Reduction mode at any time to return to normal operations.
To exit from power reducing, select the "Exit Function" box and the primary menu will reappear.
Q:\TRNOPSLPLIC-0207.doc Page 31 of 46 NRC COPY #1
ID S -0704-K1 Points: 1.6 The plant is operating at 100% power and a Traversing In-Core Probe (TIP) tr~ce is in progress.
A spurious reactor scram occurs and reactor water level decreases t -10 c: and then recovers.
IDENTIFY the response of the TIP system. V7 A. The TIP system automatically withdraws and the ball valve shuts.
B. The TIP system automatically withdraws and the shear valve fires if the ball valve fails to shut.
C. The shear valve automatically fires.
D. The TIP system will continue the trace without interruption.
Answer: A Question 11 Details Question Type: Multiple Choice Topic: Question #11 (RO/SRO)
System ID: 910 User ID: SR-0704-Kl 2 Status: Active Must Appear: No Difficulty: 2.75 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LIC-0704B, R. 6, pg. 14 User Text: 215001 K4.01 User Number 1: 3.40 User Number 2: 3.50 Comment: ILT.01885 (75987) Bank question. Hig TIPS auto withdraw and the Ball valve valves do not auto fire. N /
4/
10/15/02 OPERATIONS OPERATIONS 10/15/02
-\\C aýobpy
Content/Skills Activities/Notes
- 2. Control Room (cont)
.Instrument/Location Sensing Point/Device Description/Function
- 5. Purge indicating light Red indicating light iilluminates when the purge system is activated.
- 6. Cont. Isol. Light White indicating light Illuminated when Gp. II Isolation bus fuse is intact.
B. Automatic Functions
- 1. Initiation NO auto initiation.
- 2. Trips and Isolations Purpose Device/Setpoint/ Bypass/Reset Response SR-0704-K11 SR-0704-K12 Isolate TIP ball valve PCIS isolation logic, Reset when GP 2 Shifts drive into to complete primary SR-0704-K13 I out of two taken PCIS isolation is reverse and retracts containment twice, RX Jo water reset detector into shield isolation (GP 2) level (+8"), High Show Figure 0704B-14 chamber. The ball Drywell pressure valve will attempt to (2.5 psig) or High close and will ride Drywell Radiation on the cable until the (l00R/hr) detector is I withdrawn past the ball valve. It will then close completely.
- 3. Interlocks Purpos Device/Setpoint/ Bypass/Reset Response LoLg_
Prevents damage to Limit switch on ball Open ball valve If ball valve is not detector cable valve prevents drive fill open by position mechanism from 0020, drive inserting detector mechanism is unless ball valve is deenergized.
open.
Prevents more than Indexer release limit None Ensures only one one indexer from switch on indexer indexer can align being in position 10 mechanism, with channel 10 at at the same time. any given time.
Preven*Is 10 i t-omputer stops None If in-shield limit insertion with bad in detector insertion switch does not pick shield limit switch. just outside Rx end up at position 0011 +
of shield chamber. I drive mechanism is deenergized.
I q:\tmopslp\lic-0704b.doc Page 14 of 32 NRC COPY # 1
I*R0* 5-.2 ....... 1.**Points:
1.6 With Unit One at 50% power, the NSO selects rod D-9 for withdrawal.
The following indications are observed on the 4 Rod Display:
Two bypass lights are lit for "A" level selected LPRMs.
Two bypass lights are lit for "B" level selected LPRMs One bypass light is lit for "C" level selected LPRMs.
Three bypass lights are lit for "D" level selected LPRMs.
Will the operator be able to withdraw control rod D-9 with the present plant conditions?
A. No, RBM 7 is INOP due to less than 50% of it's assigned inputs.
B. Yes, RBM 7 is automatically bypassed due to too few inputs.
C. No, RBM 8 is INOP due to less than 50% of it's assigned inputs.
D. Yes, RBM 8 is automatically bypassed due to too few inputs.
Answer: C Question 12 Details~
Question Type: Multiple Choice Topic: Question #12 (RO/SRO)
System ID: 7386 User ID: SR-0705-K21 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LIC-0700-5, R. 4, pg. 20 User Text: 215002K6.05 User Nu.mber 1: 2.80 User Number 2: 3.10 Comment: LWQ.00082 (82500) Modified bank question. Higher.
50% of LPRM inputs to a RBM inop will inop the RBM.
A and C LPRMs feed RBM 7, B and D LPRMs feed RBM 8. RBMs are not auto bypassed due to too few inputs, must be manually bypassed.
OPERATIONS Pa of 130 _*
A 10/15/02 YRCaC~SPy #Y
Content/Skills Activities/Nota-z The input signals to the RBM averaging circuits come from the LPRM's surrounding the selected rod.
If the count circuit detects that less than half of the assigned B. Supported Systems SR-0705-K18 The RBM system also sends signals to the RMCS to generate rod blocks and initiate a rod withdrawal inhibit when needed.
C. Power Supplies...
- 3. The alarm lights on the desk section of the 90X-5 panel ar
- powered from the Instrument Bus.
\,,4. The recorders on the 90X-5 panel are powered from the)
"' Essential Service Bus.
q:\tmopslp\lic-0700-5.doc Page 20 of 26 NRC COPY #1
1 14 Points. 1.06 How many LPRM inputs are automatically routed to EACH RBM channel when a Rod with FOUR adjacent LPRM strings is selected? (ASSUME NO LPRM's are bypassed)
A. 8 B. 16 C. 2 D. 4 Answer: A
- Detailsj
__sio Question Type: Multiple Choice Topic: ILT.01224 : LPRM strings to RBM System ID: 605 User ID: 75677 Status: Active Must Appear: No Difficulty: 0.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
User Text: 215002K1.02 User Number 1: 3.20 User Number 2: 3.10 Comment: Complete Rev OPERATIONS P 4 of 5 10/15/02 C CoY RJ[---Dý
I D: 0701K22Points: 1.00 A plant startup is in progress with all IRMs on Range 1 and the Mode Switch is in the Startup/Hot STBY position.
Which ONE of the following describes the operation of the SRM instruments with all of the shorting links removed?
A FULL Reactor Scram will occur if SRM...
A. 23 is WITHDRAWN from the core.
B. 22 goes less than 100 CPS.
C. 21 and 23 BOTH reach 1 X 10E5 CPS.
D. 24 reaches 5 X I0E5 CPS.
Answer: D Question I etil Question Type: Multiple Choice Topic: Question #13 (RO/SRO)
System ID: 928 User ID: SR-0701-K22 Status: Active Must Appear: No Difficulty: 2.75 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LIC-0701, R. 5, pg. 16 User Text: 215004K4.02 User Number 1: 3.40 User Number 2: 3.50 Comment: ILT.01906 (76006) Bank question. Memory. The only scram signal for SRMs is one or more SRMs at 5 X 1OE5 CPS with shorting links removed. The other distractors give SRM rod blocks.
OPERATIONS C Pa "Pof R0 7 *"*E-* *10/15/02
Content/Skills . 0,1 Activities/Notes B. Automatic Functions
- 1. Trips and Interloc KS Device/Setpoint Device/Setpoint/ Bypass/Reset I
Response
Logic Retract Permit - If LCR output is The trip is If the trip prevents 100 cps and the bypassed by any of conditions are met SR-0701-K09 withdrawing SRM detector is the following: SR-0701-K13 for any SRM, a control rods with not fully inserted, - SRM detector rod block is neutron levels too a trip is produced. fully inserted, generated. The NOTE: TRM requires that the low (maintains Logic is arranged - All IRM range trip can be cleared reliable such that any one switches in the setpoint for: Retract Permit by inserting the indication), trip will produce a SRM's trip channel SRM to the "Full shall be > 100 cps. SRM HI rod block, are on/above In" position. shall be < 2.8 x 10W cps.
range 3.
- The Mode Switch is in the "RUN" position.
SRM HI - prevents If LCR output is The trip is If the trip withdrawing 10 cps, a trip is bypassed by either conditions are met control rods with produced. Logic js of the following:
neutron levels too for any SRM, a arranged in the - All IRM range rod block is high (approaching same manner as switches in the generated.
meter off scale the retract permit SRM's trip channel high). Maintains logic, are on/above range reliable indication. 8.
- The Mode Switch is in the
\.. INOP - prevents "RUN" psition.
If one of the The trip is If the trip withdrawing following exist: bypassed by either conditions are met control rods with - High voltage of the following: for any SRM, a the SRM circuitry becomes low - All IRM range rod block is malfunctioning. - An internal switches in the generated.
module is SRM's trip channel unplugged are on/above
- The function range 8.
switch is not in - The Mode "OPERATE". Switch is in the a trip signal is "RUN" position.
initiated. Logic is arranged in the same manner as
--- the retract permit SRM HI-HI- IfLCR output 5 x The trip is If any one of the prevents excessive 10 5 cps, a trip is bypassed when SRM trips are power during produced. Logic is shorting links are received, a 1/2 scram initial fuel loading arranged such that installed is received with 2 and startup. if any SRM trip is (the normal shorting links are received, a full condition). removed or a full reactor scram is /
reactor scram received with all 4 /
occurs. shorting are
"'" removed.
.Cc ia'ors - /
q:\trnopslp\Iic-0701 .doc Page 16 of 34 NRC COPY #1
ID:SR-0701-IK15 . 11Poin~ts. .
A Unit 1 startup is in progress.
SRM's are fully inserted and reading approximately 10,000 cps when annunciator 901-5, "SRM HIGH OR INOP", alarms and the associated rod block occurs.
The NSO observes that SRM 21 is now reading approximately 5,000 cps, while SRM's 22, 23 and 24 are still indicating 10,000 cps.
Which of the following operations / malfunctions could explain the observed indications?
A. SRM 21 is automatically withdrawing from the core.
B. -Hi'gh voltage power supplyAs low. ./
C. 100t48 gn Bt A
C.'*I448 VDC Bus A voltage is low.
D. X\ 'INOP INHIBIT" pushbutton on th / el is depressed.
Answer: b L4AVILI it~E 4 7 Que-stion,141 Details Question Type: Multiple Choice Topic: Question #14 (ROISRO)
System ID: 9822 User ID: SR-0701-K15 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 1 Point Value: 1.00 Cross
Reference:
LIC-0701, R. 5, pg. 3-5 User Text: 215004K6.04 User Number 1: 2.90 User Number 2: 2.90 Comment: ILT.11627 : 81138 Bank question. Low High Voltage power supply voltage to an SRM will cause erratic operation. A loss of 24/48 VDC A will cause SRMs 21 and 22 to fail downscale. SRMs do not auto withdraw.
INOP INHIBIT pushbutton bypasses the INOP trip while testing.
, -*&*-** -*` -,*-*
- :;/G A':`!! -/*
- -`* * :T U*.*-`i:-*` i*.,y* *!-*`* *;;*-:* ` :`-* ...
- G -; i*¸;7T :
OPERATIONS 10/15/02
[m D Pa?"o 3
Content/Skills 'V t i e Activities/Notes 1I. COMPONENT DESCRIPTION A. In-Core Detector SR-0701-K14 SR-0701-K15 The in-core detector is a "Fission Chamber" type detector that generates an electrical signal proportional to the neutron flux level in the core, for use by the electronics, to present a display of that flux level in the control room.
There are two types of fission chamber:
- Operational Chamber Show Figure 0700-1-3 This chamber fits in the gaps between the fuel bundles NOTE: The fuel loading on the opposite corner from the control rod. chamber was formerly used for (Figure 0700-1-3) (The LPRM presentation defines the "gap" terminology.) refueling outages. It is no longer normally required due to less stringent Tech. Spec.
Fuel Loading (Dunking) Chamber requirements and exposed fuel This chamber is fitted into a stainless steel cylinder providing adequate source with a nitrogen purge on it to keep water out. The neutrons.
Dunking chamber has more U-235 than the operational
/-
chamberfor added sensitivity.
The surface of the outer electrode is coated with uranium oxide SR-0701-K14 (Figure 0700-1-4). When a neutron penetrates the coating, it can cause a U-235 atom to fission. The fission products recoil into Show Figure 0700-1-4 the chamber and strip the electrons from the argon gas in the chamber, causing the gas to ionize. The bias voltage on the electrodes (furnished by a high voltage power supply) causes the electrons and ions to be collected at the inner and outer electrodes, respectively. As the electrons strike the inner electrode, they cause a current pulse to be generated, as do the "argonions when they strike the outer electrode. Since the electrons are lighter and travel faster, they are collected faster.
This makes the current pulse caused by electron collection faster and higher than the pulse caused by argon ion collection.
The IRM and LPRM detectors operate in Region-B of the Gas Conductivity curve, shown in Figure 0700-1-5. The SRM and Dunkers operate in Region C because higher voltage is required for their greater sensitivity.
S. . . . . .. . .. . . . .. . . . . .. . . . .. . . .. . . . .. , , .. . J q:\trnopslp\lic-0701 .doc Page 3 of 34 NRC COPY #1
/1 (C(
Content/Skills Activities/Notes Region A is characterized by a voltage difference (applied to the detector) too weak to attract the ion-pairs formed by the passage of the fission products. The ion pairs recombine; therefore, this region is termed the recombination region.
Region B (operating range of the IRM and LPRM detectors) is Show Figure 0700-1-5 characterized by a relatively constant current output over a wide range of detector voltage values. Basically, the curve flattens in Q: How does detector accuracy region B because the ion-pairs are moving so fast that and sensitivity vary as you recombination does not occur. Therefore, essentially every progress along the Gas charged particle produced by the nuclear reaction reaches the Amplification Curve?
electrode.
A: As applied voltage increases, in regiin C, (operatifiuiange of the SRM operational and fuel sensitivity increases while loading chambers) the higher detector voltage causes the ions accuracy decreases. At higher and electrons, traveling towards the electrodes, to gain enough oltages, the detector can energy to ionize additional argon atoms. This phenomenon is produce an output from a single
--kn wn as secondary ionization. incident particle because it accelerates faster at the higher Secondary ionfi-zation-increasesthe ouULfro e detector and the current output correspondingly increases with a proportional voltage. At the same time, the secondary ionizations and increase in detector voltage (Figure 0700-1-5). In region C, any ionization initially produced results in a proportional amount of eventual detector tube flooding secondary ionization, hence the name. The ratio of initial results in lower accuracies due w- - Fnizations to total ioniza-tions is called the gas amplification to the dead time created.
factor. The gas amplification factor increases with increasing voltage. This is why the SRM's are more sensitive than the IRM's and LPRM's (other reasons include higher gas pressure and a thicker U-oxide coating).
Gammas are generated from both fission product decay and from SR-0701-K14 the decay of materials that were activated by the neutron flux.
A gamma entering the fission chamber can directly ionize the argon gas in the space between the electrodes. The argon ions and the electrons are collected and the pulses are generated, just as in the neutron event. The specific ionization (ions produced per unit track length) of the gamma is much less than for the fission products due to the large comparative size and charge of the fission product. Thus fewer ions are produced by the gamma than by the neutron event. Also, the gamma pulses are much smaller than the neutron pulses. This difference in pulse size allows the discriminator to eliminate the gamma pulses, while passing the neutron pulses.
q:\trnopslp\Jic-0701 .doc Page 4 of 34 NRC COPY #1
Content/Skills K, Activities/Notes Operational Chamber Specifications (Figure 0700-1-6)
NOTE These specifications are nominal values, which may vary slightly from different model parts.
- 1. Dimensions: Show Figure 0700-1-6
- a. 1.6 in. overall length
- b. 1 in. sensitive length
- c. 0.160 in. diameter
- 2. The active coating is composed of 3.3 mg of uranium oxide.
The uranium is 95% enriched in U-235. This coating is applied only to the inner surface of the outer electrode.
- 3. The chamber is argon filled to 14.5 atmospheres (213 psia).
- 4. The following sensitivities assume a properly calibrated discriminator, as they are actually quoted for the output from the discriminator rather than from the detector itself.
.amNeutrons: 1.0 x 10-3 cps/nv.
- b. Gammas: Zero caunts-in-5-x--0-R/hr. field.
The chamber operates at 600 VDC. This voltage is set high Q: How do the IRM detectors to provide optimum detector sensitivity. differ from the SRM's?
A: IRM detector voltage is
/ lower because less sensitivity is required.
The thickness of the IRM detector Uranium coating is lower due to the higher flux levels experienced by the IRM's.
Less Uranium is needed for detector operation.
The Argon gas pressure in the IRM's is lower which makes them less sensitive.
q:\tmopslp'lic-0701 .doc Page 5 of 34 NRC COPY #1
15 ID:Rk'073-1<09~ Points 1.b Which of the following would constitute the MAXIMUM disagreement between APRM flow converter channels that would still allow control rod withdrawal?
A. 17%
B. 11%
C. 9%
D. 5%
Answer: C Question Type: Multiple Choice Topic: Question #15 (RO/SRO)
System ID: 286 User ID: SR-0703-K09 Status: Active Must Appear: No Difficulty: 3.00 gwTime to Complete: 0 Point Value: 1.00 Cross
Reference:
QCAN 901-5 D-6, R. 2 User Text: 215005A3.06 User Number 1: 3.00 User Number 2: 3.10 Comment: ILT.00877 (75358) New question. Lower. Alarm comes in at 10%, so 9% is the highest you can have without getting the rod block.
OPERATIONS 10115/02
- RC Pa***P L-
I OCAN 901-5 D-6 U) UNIT 1*
REVISION 2
___ ___ .Continuous Use "FLOWCONVERTER REFERENCE OFF NORMAL ROD BLOCK DESCRIPTION SETPOINT Actual: 1. 10%_mismatch between channels. NEUTRON MON
- --2§-T10- (increasing) channel output.
2--.TI-0FT~icreaFLOW FLOW UNIT UN
- 3. Flow converter inoperable. OFF NORMAL
- a. Flow converter mode switch NOT in OP (operate).
- b. Loss of + 15 volt power supply.
Tech Specs: None SENSOR Relay 756-K1 or 756-K3 in Flow Unit.
A. AUTOMATIC ACTIONS
- 1. Rod out block in all modes of operation.
- 2. Half-scram when channel fails downscale.
B. OPERATOR ACTIONS
- 1. Stop all power changes in progress.
- 3. Contact QNE for assistance.
- 4. In Panel 901-37:
- a. Monitor flow units to determine if UPSCL/INOP or COMPARATOR lights are lit.
- b. Check APRM flow bias signal by placing the Flow Converter Power Supply in the CONVERTER OUTPUT position.
(1) IF flow bias signal fails downscale or low (conservative), THEN a half-scram may occur in the corresponding RPS channel.
NRC C6PY #1
16I:SR-22 Poi'nts:' i.00 Given:
- RVLIS backfill has been secured for 18 days.
- The RPV has rapidly depressurized from 1003 psig due to a steam leak in the drywell.
- Drywell temperature is 235 degrees F.
- RPV pressure is 275 psig and slowly lowering.
- Pressure corrected lower wide range instruments indicate -10 inches and lowering.
- Narrow range instruments indicate +10 inches and steady.
What is the status of Rx level instrumentation and which of the following conditions can be used to determine RPV water level is > -68 inches if the recirc pumps are off? Rt tieue ( , ,V A. Will become inaccurate when pressure drops below 250 psig; read level directly on narrow range instruments to determine level > -68 and steadv.
B. Became inaccurate when pressure dropped belo p indicated level is lowering on the narrow range instruments.
C. Became inaccurate when pressure dropped below 450 psig; indicated level is rising on the lower wide range irlstper*is.
D. Will become inaccurate when pressure drops below 250 psig; determine level > -68 inches by indicated level rising on the upper instrument.
Answer: B gAP Q~4~Jc 1/,1 17 R C PaC**f 3 ,! *- 10/15/02
Question 16 Details Question Type: Multiple Choice Topic: Question #16 (RO/SRO)
System ID: 7530 User ID: SR-0263K22 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOP 0201-11, R.4 pg. 1-3 User Text: 216000A2.10 User Number 1: 3.30 User Number 2: 3.50 Comment: L.00637 (82648) New question. Higher Pressure at which gassing occurs is 450 psig not 250. Level is proved above -68" by a decreasing trend only, not increasing.
OPERATIONS RC (
Pa? C of 30 rF~zýý 7-Il 10/15/02
QCOP 0201-11 UNIT 1(2)
REVISION 4 SAFETY- RELATED
- - ---- - -~-~ -ON-HAND DETERMINING'RPV LEVEL DURING RAPID DEPRESSURIZATION BELOW 450 PSIG SHIFT OPERATIONS SUPVSR q-_ __-_ _
APPROVAL SIGNATgRE TITLE EFFECTIVE DATE A. PURPOSE This procedure contains the direction on how to determine RPV water level during rapid depressurization of the RPV when RPV pressure is below 450 psig. It is implemented alone or concurrently with the QGAs and does NOT alter the RPV level instrument direction provided in the QGAs. This procedure is only applicable when RVLIS Backfill system flow has been outside the acceptable range for >14 days.
B. DISCUSSION B.1. RPV level indication is affected by various plant parameters. The majority of these are defined in the QGA procedures and related specifically to post accident plant conditions. This procedure does NOT alter the QGA direction but does clarify the use of RPV level instruments for plant conditions that may or may NOT involve entry conditions into the QGAs.
B.2.' The specific industry concern that created the need for this procedure is non-condensable gas accumulation in the RPV level instrument reference legs. During rapid RPV depressurization (i.e. > 100 OF/hr.), gas in the reference legs can come out of solution and cause the instrument to indicate an RPV level that is higher than the actual RPV level. Therefore, there is a need to be able to determine the actual RPV level or else the QGA procedures will direct performance of RPV flooding.
RPV flooding will insure that the core is adequately cooled but is a severe transient on the RPV. Use of this procedure provides an alternative to stating the RPV level is unknown for every rapid RPV depressurization.
Continuous Use 1 I
QCOP 0201-11 UNIT 1(2)
REVISION 4 "E. LIMITATIONS AND ACTIONS E.l. Since this procedure will be implemented transient conditions, the SCRE is fulfilling during role. the STA Due to this and other time constraints, verification of calculations is NOT documented as being performed but the evolution will be overviewed by the SCRE.
E.2. Attachments A and B are provided as aids to the crew to facilitate performance of repetitive, calculations by the crew. ongoing These Attachments may be used by the crew as deemed necessary.
E.3. IE Control Room Narrow Range instruments are NOT available, THEN local instrumentation may be used provided the low end of their range is NOT below -60".
F. PROCEDURE
NOTE This procedure provides two methods of verifying that RPV
,/ .A*
1level is "known".
Step F.1. verifies level is > -68", the level of the lower tap for the Narrow Range Level instruments GeMac, by verifying the Level instruments AND the Upper 400 are trending.
Step F.2. provides direction for calculating RPV level when the Narrow Range Level instruments AND the Upper 400 GeMac are NOT showing a trend.
F.1. Verify RPV level to be > -68", by using either of the following two methods:
- a. Verify a decreasing trend on Panel 901(2)-5 Narrow Range level instrument OR on Panel 901(2)-4 Upper 400 GeMac.
- b. IF RPV pressure is greater than or equal to the lowest value it reached, THEN verify an increasing OR decreasing trend on Panel 901 (2)-5
- .. level instrument OR on Panel 901(2)-4 Narrow Range Upper
_4 0 0 GeMac.
NRC COPY #1
Unit One is operating at full power when a loss of Bus 18 occurs.
Shortly afterwords, a loss of the 250 VDC system occurs.
Predict the effect on the 901-5 panel reactor water level instrumentation.
A. Only Wide range level instrumentation will be available.
B. All Medium range level insturmentation will be downscale.
C. All level instruments will still be available.
D. All Narrow range level instrumentation will be downscale.
Answer: C Iu6estilon 17 is Question Type: Multiple Choice Topic: Question #17 (RO/SRO)
System ID: 9777 User ID: SRN-6800-K23 Status: Active Must Appear: No Difficulty: 3.25 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LN-6800 pg. 1, fig. 2 User Text: 216000K6.02 User Number 1: 2.80 User Number 2: 3.00 Comment: New. Higher. Loss of Bus 18 removes 2 of the 4 power sources from the ESS system. Loss of 250 VDC removes one more source. ESS still receives power from the static switch without power interruption that keeps instrumentation on line without change.
OPERATIONS 10/15/02 C PaCof 0 1
Conte/nt/Skills I.
I. INTRODUCTION - BRIEF DESCRIPTION
- , A. Purpose The main function of the ESS UPS and Instrument Bus 120-V systems is to provide a reliable source of 120-V, 60 Hertz, single phase power for plant controls and instrumentation.
B. Basic System Operation
- 1. Instrument Bus System **N-6800-K15 SR-6800-K15 The instrument bus system supplies 120 VAC power for various control circuits, relays, solenoids, and instruments. The normal power supply to the instrument bus is from MCC 18-2 Present as a classroom lecture (28-2) via a 480 to 120/240V step down transformer located at with overhead transparencies and slide show.
MCC X8-2. The alternate power supply is from MCC 15-2 (25-2) which has its own step down transformer. On loss of theý normal power supply, an automatic bus transfer (ABT) switch swaps power to MCC 15-2 and transfers it back upon restoration of normal power (normal seeking ABT). Note that ABT operation will cause a momentary loss of power. The instrument bus distribution panel is located in the Auxiliary Electric Room (Panel 901(2)-50). Bus voltage indication is available at the bus cabinet in this room.
....-.- f
-/ 120 VAC EssentialaService Bus ancfUninterruptable Power SOER 83-3, Inverter Failures Supply System Show Figure 1 The ESS bus supplies 120 VAC power to essential instruments and control circuits. Normal power to the ESS bus is supplied from the static (solid-state) uninterruptable power supply (UPS). The UPS transfers between its power sources without interruption in power to the load. The reserve power supply for the ESS bus is MCC 18-2 (28-2) via an ABT. Note that when power supplies are swapped via an ABT, momentary power loss is experienced. Power to the UPS is supplied, in order of preference, by Bus 18 (28), 250 V battery via 250 VDC MCC 1(2), and Bus 17 (26). The UPS is physically separated and electrically isolated by use of circuit breakers. The UPS and the essential service distribution panels are located in the Auxiliary Electric Room (UPS: 901(2)-63, ESS: 901(2)-49).
The power supply from MCC X8-2 originates at the same step /
down transformer that feeds the instrument bus. The /
transformer at MCC X8-2 is supplied from compartment D4.
The out put from the transformer is split and sent to Scompartment C6 to feed the instrument bus and to compartment b5to feed the ESS service bus.
°*i:
7 Q:\tmopslp\LN-6800.doc Page 1 of 27 NRC COPY #1
250 VDC MCC-1 SWGR 18 (28) SvVGR 17 (26) MCC 18-2 (28-2)
UPS DC )UPSNY S-F NORMAL SUPPLY ) SUPPLY UPS ALT ~.INSTIBUS NORMl AI ESS. BUS RESERVE CUBICLE I)2 Q CUBICLE D4 CUBICLE D6 CUBICLE D4 I
)BATT. D, INPUT 6Z
)RECTIFIER eb INPUT 63B I 480 VAC
.1 wU..
I ER DC 63 A I ESS RESERVE SUPPLY 0 ) CUBICLE D5 I.
I I
IL IX "UPS PANEL 90X-63 ASCO 120 VAC SWITCH ABT RESERVE POWER SUPPLY ESS BUS FIGURE 6800-02 REV. 0 ESSENTIAL SERVICE POWER SUPPLIES NRC COPY #1
18 RCIC automatically started and is maintaining reactor water level at -40 inches.
Annunciator 901-4 F-15 "RCIC TURBINE BEARING OIL PRESSURE LOW" is alarming.
The Unit One NLO reports that RCIC lube oil pressure is 3 psig decreasing despite efforts to 'a;1 t $
restore pressure.
Oil levels are all normal.
Continued operation of RCIC in this condition will result in reactor water level:
A. maintaining due to theemergency oil pump auto starting. , t F
C. maintaining due to all RCIC trips being bypassed on an autostaq..
D. decreasing due to ip on overspeed from the governor valve failing open.
Answer: D 6 6 e s i16-n, 1-8 De-t 'a'iIs Question Type: Multiple Choice Topic: Question #18 (ROISRO)
System ID: 9801 User ID: SR-1 300-K22 Status: Active Must Appear: No Difficulty: 4.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOA 1300-04 R8 User Text: 217000K3.01 User Number 1: 3.70 User Number 2: 3.70 Comment: Modified question. Higher Answer is correct due to governor valve failing open as oil pressure decreases will cause a turbine trip on overspeed. i never bypassed. Turbine bearing high temperature is not a trp.UnIe HPCI, RCIC does not have an emergency oil pump.
References QCAN 901-4 F-15 rev 1, QCOA 1300-04 rev 8.
N - -.- -. ----.------ - -
OPERATIONS 10/15/02 EDC Pa o)30
QCOA 1300-04 UNIT 1(2)
J REVISION 8 Continuous Use RCHICTURBINEl BL LOW PRESSURE OR OIL HIGH TEMPERATURE A. SYMPTOMS
- 1. Possible RCIC Trouble alarms:
- a. Panel 901(2)-4 (1) D-15, RCIC TURBINE TRIP.
(2) E-15, RCIC GOVERNOR END BEARING HIGH TEMP.
(3) F-14, RCIC COUPLER END BEARING HIGH TEMP.
(4) F-1S, RCIC TURBINE BEARING OIL LOW PRESSURE.
B. DISCUSSION The RCIC Turbine is equipped with an oil ring on both the outboard and the inboard bearings. These oil rings allow operation of the turbine even on loss of lube oil pressure IF the oil level in the bearing is maintained. However, the loss of oil pressure may make the governor valve inoperable.
IF this occurs, THEN the RCIC Turbine may be manually operated by throttling the turbine trip and throttle valve.
.TheSUBSEQUENT OPERATOR ACTIONS of Section E assume that the RCIC System is required for continued operation to support either adequate Core Cooling OR Reactor pressure control.
Recovery steps for a high lube oil temperature condition include checking operation of several relief valves.
Adjustment to the lube oil cooling water relief valves MUST be done with care to prevent over pressurizing the lube oil cooler, Barometric Condenser and/or low pressure piping as well as prevent personnel injury.
C. AUTOMATIC ACTIONS None.
D. IMMEDIATE OPERATOR ACTIONS None.
1 NRC COPY #1
QCOA 1300-04 UNIT 1(2)
REVISION 8
[]
NOTE High RCIC Barometric operating temperature or pressure is indicative of inadequate cooling water flow and/or RCIC exhaust high back pressure.
- 8. IF lube oil temperature is high (i.e., > 180 0 F) as indicated by Alarm 901(2)-4 E-15, RCIC GOVERNOR END BEARING HIGH TEMP, or Alarm 901(2)-4 F-14, RCIC COUPLER END BEARING HIGH TEMP, THEN direct Operator in attendance to:
- a. Confirm bearing oil temperature > 180OF on local oil temperature indicators (outboard end of turbine).
- b. Check oil level/flow within RCIC Turbine Lube Oil System sight glasses (inboard end and outboard end of turbine).
(1) IF oil level is low, THEN add oil to RCIC lube oil reservoir.
- c. Check RCIC Barometric Condenser pressure approximately 10 in. hg. on PI 1(2)-1360-8203 and temperature approximately 160OF on TI 1(2)-1360-8204.
- d. Check operation of RCIC Barometric Condenser Condensate and Vacuum Pumps.
- f. Check for cooling water flow through RV 1(2)1301-42, U-1(2) RCIC PMP TO LUBE OIL CLR RV, on FG 1(2)1301-72.
- 9. IF lube oil pressure is low (i.e., < 3 psi) as indicated by Alarm 901(2)-4 F-15, RCIC TURBINE BEARING OIL LOW PRESSURE, THEN direct Operator in attendance to attempt to restore lube oil pressure by:
- a. Confirm bearing oil pressure < 3 psig on local oil pressure indicators (outboard end of turbine).
3 NRC COPY #1
QCOA 1300-04 UNIT 1(2)
REVISION 8 F. REFERENCES
- 1. TS 3.5.3, Reactor Core Isolation Cooling (RCIC) System.
I
- 2. M-50 (M-89), Diagram of RCIC Piping.
. RCIC System Part 6.
- 11. C00396 (GEK-9546), Operation and Maintenance Instructions, Reactor Core Isolation Cooling System.
- 12. C00471 (GEK-27820A), Quad Cities 1/2 Process Instrument Subsystem of the Reactor Core Isolation Cooling System.
- 13. QCOA 1300-06, RCIC System Trouble Following an Auto-Start.
- 14. QCOP 1300-05, RCIC System Shutdown.
- 15. QCOP 1300-09, RCIC Local Manual Operation.
- 16. QCAP 0230-19, Equipment Operability.
- 17. UFSAR Section 5.4.6, Reactor Core Isolation Cooling System.
- 18. QCNPS Procedure Writer's Guide, Rev. 1, dated 1-31-90.
(final) 5 NRC COPY #1
QCAN 901(2)-4 F-15 UNIT 1(2)
REVISION I SAFETY RELATED SHIFT OPERATIONS SUPVSR APPROVAL SIGNATURE TITLE EFFECTIVE DATE DESCRIPTION RCIC TURBINE BEARING OIL PRESSURE LOW SETPOINT Actual: I. RCIC Turbine bearing oil RCIC TURBINE
-pressure low; 3 psig. BRG OIL Tech Specs: None LOW PRESSURE SENSOR 1. 1(2)-1303-PS1 A. AUTOMATIC ACTIONS None.
B. OPERATOR ACTION NOTE IF lube oil pressure is lost, AND oil is still reservoirs as indicated by oil sight glass, available in bearing THEN RCIC Turbine operation can continue.
On loss of lube oil pressure, the Governor of RCIC Turbine can be performed by manuallyValve fails open and control throttling the Trip Throttle Valve.
- 1. IF RCIC System is required for adequate Core pressure control-- THEN continue to operate Cooling OR Reactor RCIC System AND attempt to correct cause of low bearing oil pressure as follows:
- a. Dispatch operator to perform the following:
(1) Confirm bearing oil pressure < 3 psig at local pressure indicator.1(2)-1303-PI oil (outboard end of turbine).
,(2) lerify proper oil level/flow within RCIC Turbine Lube
.Oil System sightglasses.
NR C COPY I #1
RCIC is running at full flow for a surveillance. Annunciator 901(2)-4 F-15, "RCIC Turbine Bearing Oil Lo Pressure," alarms. If the oil pressure continues to decrease slowly, and no operator actions are taken, what will happen to the running RCIC system?
A. The Turbine Steam Supply Valve, MO 1(2)-1301-61, will automatically close.
B. The Governor Valve, HO 1(2)-1303A, will drift closed.
C. The RCIC turbine will trip when Bearing Oil Pressure decreases below 1.0 psig.
D. The Governor Valve, HO 1(2)-1303A, will drift open and the turbine will trip on overspeed.
Answer: D Queston 3Details Question Type: Multiple Choice Topic: ILT.01916: NO TOPIC System ID: 938 User ID: 76016 Status: Active Must Appear: No Difficulty: 0.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
User Text: 217000A2.07 User Number 1: 3.10 User Number 2: 3.00 Comment:
OPERATIONS MEC Cp3#f5-J/ýIy 10/15/02
19 .':SR-1602-Kif '~ Z7 ~Po1nks: I2600 A Group II isolation will occur if the Unit One Drywell reaches , aInd this may be bypassed to allow opening the 2" vent valve to SBGTS by_
A. 1.55 psig; ayf*o*ck-- sý-on the 9015nep B. 2.5 psig; a keylock-switch on the .912-1 panel '2 C. 1.55 psig; a keyleckswitch-on tre 912-1 panel D. 2.5 psig; a kIeyloek-swftch on the.01-5 pan Answer: D Question 1-f9- betails Question Type: Multiple Choice Topic: Question #19 (RO/SRO)
System ID: 2036 User ID: SR-1602-Kll Status: Active Must Appear: No Difficulty: 2.75 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
901(2)-5 D-11, R. 7 pg. 1 User Text: 223001 K1.09 User Number 1: 3.40 User Number 2: 3.60 Comment: ILT.04298 (77115) Bank question. Memory. These are Gp 2 valves, the Gp 2 comes in at 2.5 psig. The keylock switch is on th 901-5 panel.
10/15/02 OPERATIONS P 9,1 of 130 10/15/02 N IM C a7
ýPy J1/Z
QCAN 901(2)-5 D-11 UNIT 1(2)
. . . . . .. .. .. . . . . . . . REVYI S I ON 7 .
Continuous Use DESCRIPTION HIGH DRYWELL PRESSURE SETPOINT Actual: High Drywell Pressure: 2.32 psig. PRIMARY CNMT Tech Spec: < 2.43 psig. HIGH PRESS SENSOR PS 1(2)-1001-88A/B/C/D.
A. AUTOMATIC ACTIONS
- 1. Reactor scrams, Group II isolation occurs, Control Room Vents isolate, Reactor Bldg Vents isolate, and SBGT initiates.
- 3. IF DW high pressure is also sensed by PS 1(2)-1001-90A/B/C/D, THEN LPCI and Core Spray initiate, DGs auto start, DW Coolers trip, DW Vent Booster Fan trips, RBCCW Pumps trip, Recirc MG Set Vent Fan i pý= and B East Turb Bldg Supply Fans trip, C Turb Bldg Exhaust Fan trips, and Fuel Pool Cooling Pumps trip.
B. OPERATOR ACTIONS
- 1. IF High Drywell Pressure exists, THEN:
- a. Verify Reactor scram and enter QCGP 2-3.
- b. Refer to applicable QGAs.
- 2. IF a half-scram has occurred AND a high Drywell Pressure does NOT exist, THEN determine and correct cause of half scram:
- a. Determine if testing is being performed on Drywell Pressure channels OR Reactor Protection System.
- b. Monitor the following Unit 1 (Unit 2) computer points to determine which RPS channel is in trip condition:
. (1) Point ID W512 (W612), CONTAINMENT HIGH PRESSURE A.
(2) Point ID W513 (W613), CONTAINMENT HIGH PRESSURE B.
(3). Point ID'W514 (W614), CONTAINMENT HIGH PRESSURE C.
(4) Point ID W515 (W615), CONTAINMENT HIGH PRESSURE D.
NRC C6PY #1
20 ~ID:R-1O06M-K22, Points: 1.00.
Torus sprays are being tested on Unit One when a recirc system leak results in a Rx Scram and entry into the QGAs.
The ANSO has started Torus Sprays, Torus Cooling and RHR Service Water.
The MO-1-1001-16A, RHR Hx Bypass Valve is fully closed.
The NSO also notes that the maximum RHR service water flow with the MO-1-1001-5A, RHR Hx SW Disch Valve, full open 1 2500 gpm at a discharge pressure of 275 psig.
t aoA What action(s) should be taken.
I A. Secure Torus sprays.
B. Cross connect the "A" and "B" RHR Service Water loops. ,
C. Start a 2nd RHR Service Water Pump. C .
- 0. Stop the RHR Service Water pump and reverse heat exchanger flow.
Answer: D Q'uetion 20 Details Question Type: Multiple Choice Topic: Question #20 (RO/SRO)
System ID: 9760 User ID: S/R-1 000-K22 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 2 Point Value: 1.00 Cross
Reference:
QCOP 100-04, R. 14, D.1 User Text: 230000K5.06 User Number 1: 2.50 User Number 2: 2.60 Comment: New question. Higher. Indications of HX fouling.
RHRSW pump discharge pressure is higher than TS minimum, so do not suspect RHRSW pump failure. If Torus sprays are on the "A" loop, start them on the "B" loop. Cannot crossconnect "A" and "B" RHR SW loops.
OPERATIONS FD Pa of 3 10 10/15/02
QCOP 1000-04 UNIT 1(2)
REVISION 14 Continuous Use RHR SERVICE WATER SYSTEM OPERATION A. PURPOSE The purpose of this procedure is to provide the necessary steps for RHR Service Water System operation.
B. DISCUSSION B.l. This procedure is divided into several procedure sections: different
- a. Step F.l is for operation of A Loop pumps.
- b. Step F.2 is for operation of B Loop pumps.
- c. Step F.3 is for HX Valve operation.
C. PREREQUISITES None.
D.. I* F there appears to be inadequate heat transfer acrosA the RHR Heat Exchanger or inadequate RHR Service Water low (< 3500 gpm and < 198 psig discharge press refer t-Techn-iea cigjiation 3.7.1J2and-Te-nical Requirements Manual (TRM) section 3.7.a.), THEN consider taking the RHR Service Water Pumps off and reversing flow in the HX to flush it out (this is a possible sign of biofouling in the heat exchanger).
(H.8.b.)
D.2. Motor Operated Valve Guidelines: (H.8.a.)
- a. A maximum of five starts within a one minute period, followed by a 30 minute cooling off time.
- b. The valve is operable during the cooling off period.
- c. WHEN throttle valves are required to adjust flow or pressure, THEN it may be necessary to wait a few seconds to abide by this guideline.
NRC COPY #1
A transient occured on Unit 1 resulting in a reactor scram and a Group 2 isolation.
The Inboard MSIVs are closed.
Drywell pneumati receiver pressure is 75 psig.
The ANSQý c* the Target Rock Relief Valve Control Switch to "MANUAL" Which of the following supplies will provide motive force for Target Rock Relief Valve operation?
- 1. Drywell pneumatic compressors A - #
- 3. Re~lieff Valve Aiccumulator
- 4. Nitrogen akeup System - /
A. 2 and 4 B. 2, 3, and 4@tqt*
C. 3and4 M4J D. 1, 2, and 3..",* -
Answer: C QUestion 21 Details Question Type: Multiple Choice Topic: Question #21 (RO/SRO)
System ID: 9769 User ID: SR-0203-K19a Status: Active Must Appear: No Difficulty: 3.50 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LF-46/4700,pg, 56, fig 15 User Text: 239002A1.03 User Number 1: 2.80 User Number 2: 2.90 Comment: New question. Higher. The DW Pnuematic compressors will be isolated on the Gp 2. The N2 reciever makeup valve opens at 82 psig, which will cause the recievers to be out of the picture because of the higher pressure and the check valve. The accumulator will still be effective.
OPERATIONS 10115/02 OPERATIONS 10/15/02 K RC PaC of 30- --Y L
Content/Skills Activities/Notes I
B. Automatic Functions
- 1. Initiation **S/R-4700-EK020 Device/Setpoint Logc-- Bypass/Reset RespOnses Valve PCV-1(2)- DW pneumatic air When header When PCV-1(2) 4723 will header pressure pressure increases 4723 opens it automatically open switch to above 82 psig, allows an alternate when DW (PS-1(2)-4741-12) the solenoid valve source (the N 2 or pneumatic air sends a signal to will reposition, Instrument Air pressure lowers to open SO-I (2)4723 venting air off of System) to supply 82 psig. which ports air to the PCV which the system loads.,'
the PCV, opening causes it to close.
it.
- 2. Trips and Isolations S/R-4700-EK006b
- S/R-4700-EK011 Purvoo Device/Setpoint/ Bypass/Reset Response **S/R-4700-EK012 Logic Compressors I & 2 The following trips To restart the When the trip to prevent open contacts in the compressor, the trip compressor trips, damage to the compressors condition must be the backup supply compressor and the operating circuitry: cleared, and the valve system. - High vacuum in local "Reset PCV-I(2)-4723 will the suction line @ Button" light must open @ 82 psig to 7.5" Hg be pushed. supply the system PS-1(2)4741-16. loads from the
- High separator Instrument Air level System or the N LS- 1(2)-4741 - System.
8222 DW Pneuiiatiý . The GP-2 isolation When the condition-- When thessuction suctio? valvd-* signals are a One- has cleared, and the valves close on"i*
PCV- 1(2)4720/472 Out-Of-Two-Twice PCIS has been PC IS GP-2, the 1 will shut upon logic and isolate on reset, the valves suction line will receiving a PCIS the following will open provided quickly lower to the GP-2 isolation in signals: their switch is not 7.5" Hg compressor order to help 2.5 psig DW press in the "Close" trip setpoint and the /
maintain primary 100 R/hr in the position. compressor will containment DW trip.
integrity. . +8 " RPV water level Dryer refrigeration If refrigerant When temperature When the compress&i and fan temperature lowers raises above 29" F. . refrhg'eration unit Q: Why does the refrigerant will trip to prevent to 29F, a con-ffrt.... the-doffit wil--- trips, air compressor shut off at 29°F?
freezing of the in the operating close and the temperature will A: Any moisture in the air refrigerant and logic will open to compressor/fan will increase until the stream would start to freeze.
damaging the dryer. de-energize the start. unit re-starts.
refrigeration compressor and Ian.
Q:\TRNOPS LP\46004700.doc Page 56 of 78 NRC COPY #1
12 VAC rip Iri FIUR 60470CA Diyl PnuatcCSse
- 'FIGURE ° 1"RV
'14 46001'4 201 -* " ,,*+
4123::
Aliot NV *M 11)
") . ? .PC
i.:.*.S*R-O'2..-K2 Points:
Unit 2 has experienced a Group I isolation and reactor scram.
The ANSO reports that ALL relief valve indicating lights on the 902-3 panel are EXTINGUISHED.
Without operator action, Reactor pressure will increase until the: Q, A. -iifsvtUp Safety Valves open at 1250 psig. i B. "wjet4Rqk Relief valve opens at 1135 psig. -.
C. "irst-two Safety Valves open at 1240 psig.
D. Ttrget-tk Relief valve opens at 1115 psig.
Answer: B Question Type: Multiple Choice Topic: Question #22 (RO/SRO)
System ID: 9734 User ID: SR-0203-K23 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 3 Point Value: 1.00 Cross
Reference:
LIC 0203, pg, 3 & 8.
User Text: 239002K3.02 User Number 1: 4.20 User Number 2: 4.40 Comment: New question. Higher) TEarget rock safety feature is set mm at1135psig.
16 N
N.
OPERATIONS
~P 40opf130~ 10/15/02
Content/Skills Activities/Notes Activities/Notes I1. COMPONENT DESCRIPTION 7
A. Electromatic Relief/Power Operated Relief Valves (PORVs on U-2)
I. The electromatic relief valves/PORVs are designed to prevent over-pressurizing the vessel or lifting the safety valves. They are also designed to relieve pressure rapidly to the pressure reset value or to allow the Low Pressure Coolant Injection (LPCI) System and the Core Spray System to function.
- 2. The relief valves are sized to prevent lifting the safety valves Q: The ERV's/PORVs are siz during a specific transient. The transient the relief valves are to prevent lifting the safety designed to protect against is: valves during a specific
- a. The turbine trips from full power, and transient. What transient is this?
- b. The bypass valves fail to operate, and A: The turbine trips from full
- c. The reactor scrams from a closure of the turbine stop power, and the bypass valves f valves. to operate, and the reactor scrams from the closure of the 2
turbine stop valves.
- 3. Four electromatic relief valves/PORVs are located in the drywell, one each on Main Steam Lines C and D, and two on S/R-0203-EK014 Main Steam Line B, upstream of the flow restrictors. The Q: Where are the relief valves valves are actuated by energizing a 125 vdc solenoid physically located?
assembly. Three methods of actuation are used:
A: The 3B and 3E reliefs are located on MSL B, the 3C relie is located on MSL C, and the 3 relief is located on MSL D.
- a. Pressure switches (2201(2)-5 rack):
- S/R-0203-EK007a Valve B C D E Opening Setpoint 1115 1115 1135 1135 Closing Setpoint 1070 1070 1090 1090
- b. A manual demand (keylock switch).
- c. An ADS initiation signal.
- 1) High Drywell Pressure (2.5 psig), and i
- S/R-0203-EK007b
- 2) Low-Low RWL (-59"), and
- \ KRNU0PrLP\LIC-0203 doc NRC COPY #1 Page 3 of 37
Content/Skill Content/Skills Activities/Notes
.. The valve is self-actuated in the safety mode and air (pilot) Activities/Notes
- S/R-0203-EK007a actuated in the relief mode.
Q: What is the high reactor
- a. As a safety valve, the valve is set to open at 1135 psig. pressure setpoint for the targ rock?
A: 1135 psig.
- b. In the relief mode, the pilot solenoid is actuated by the Show Figure 0203-3.
following (Figure 0203-3):
- 1) High pressure setpoint (1135 psig).
- 2) A manual demand (keylock switch).
- 3) An ADS initiation signal.
- 4. Target Rock Valve Operation (Safety Mode) (Figure 0203-4)
S/R-0203-EK015
- a. The steam pressure is sensed at the pilot sensing port (2).
Show Figure 0203-4.
- b. The bellows (6) is forced to the right if the setpoint of 1135 psig is reached. Q: If the target rock bellows ruptures, what actions must bi
- c. This opens the pilot valve disc (3). taken per Tech Specs?2
- d. This allows the pressure to be transferred to the second A: Commence an orderly stage piston (8), which is forced shutdown and reduce reactor down.
coolant pressure and
- e. This vents the pressure from the top of the main valve temperature below 90 psig and piston (12). 3200 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- f. Pressure is then vented via the second stage disc (10) and out of the main valve piston vent (15).
- g. This unbalances pressure across the main valve piston (12).
- h. A pressure differential is created because of the small size of the main valve piston orifice (13) compared with the main valve piston vent (15). I-
- i. The reactor steam pressure then lifts the main valve piston (12) and the main valve disc (14).
- j. The steam flows out and is piped to the suppression pool.
- k. When the steam pressure is approximately 45 to 50 psig below the setpoint, the pilot setpoint adjust spring (4) he forces the, pilotva-ve disc (3) closed.
Y:\ KjNuo-SLP-'UjC-3 doc NRC COPY Page 8 of 37
- 1
'ID SRO020-K23 Points: 1.00
) Unit 2 is operating at rated conditions. (.A
) Z q0 set,'I. ,W11 t* +44A K?, V I 0'!*pratn I An operating RFP trips. *I/A i *L. £1 w1.*44
__V%
M#'* Which of the following describes the plant response?., ,o zt t SA. ches-26,4nches-,within 45-se-w nds, the recirc pumps will runback to 70%.
B. irc-pumps w I runback to minimum immediately.
C. pwL-ruD k. t7%_imjm~ediately_,
D. a ls 26 inchesiý thibjn445secGi4&, the recirc pumps will runback to minimum.
Answer: A dusio 3Detais7 Question Type: Multiple Choice Topic: Question #23 (RO/SRO)
I - System ID: 9823 User ID: SR-0202-K23 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCAN 902-4 F-7 R1 User Text: 259002K1.15 User Number 1: 3.20 User Number 2: 3.20 Comment: New question. Higher. The FWLC system will runback to 70% rated core flow if < 3 RFPs are running due to an auto-trip if RPV water level < 26 inches within 45 seconds and steam flow > 85%. At rated conditions, 3 RFPs are required and steam flow > 85%.
QCAN 902-4 F-7 rev 1, QCOP 0600-21 rev 1 OPERATIONS 10/15/02 F0-)C Pa of' 130
QCAN 902-4 F-7 UNIT 2 REVISION 1
continuous Use DESCRIPTION RECIRCULATION LOOP FLOWS LIMITED BY FEEDWATER FLOW AND REACTOR VESSEL LEVEL SETPOINT RECIRC LOOPS Actual: 1. Steam flow is greater than LIMITED BY FW approximately 85% of rated AND FLOW/RX LVL either of the following:
- a. Less than four Condensate/
Condensate Booster Pumps running AND total Feedwater flow greater than approximately 90% of rated.
- b. RPV low level alarm within 45 seconds of auto trip of an RFP after three pumps initially running.
STech Spec: None.
SENSOR FWLC System Relay 2-0202-60-197/198.
A. AUTOMATIC ACTIONS 1 IF core flow is > 70%, THEN both Reactor Recirc Pumps will run back to a value equivalent to 70% rated core flow.
B. OPERATOR ACTIONS
- 1. IF required to control Reactor vessel level, THEN perform I QCAN 902-5 E-8 or QCAN 902-5 F-8, as applicable, concurrently with this procedure.
- 2. WHEN the condition requiring Recirc pump runback has cleared.
THEN:
- b. Place the following in MAN:
(1) 2-262-22, RECIRC MASTER FLOW CONTLR.
NRC CdPY #1
QCOP 0600-21 UNIT 2 REVISION 1 ATTACHMENT A (Page 7 of 12)
OPERATIONAL CHARACTERISTICS 4.c. (cont'd)
(4) The level is >15" and the two scram signals have been active for >20 seconds.
(5) The level is >15" and the two scram signals are NOT active. This allows the FWLC to restore reactor water level in the event of a false scram signal that is quickly reset by the operator.
- 5. Automatic Reactor Recirculation Runback
- a. A Reactor Recirculation pump runback will occur if feedwater flow drops to less than 20% of rated to protect the purmp fromc4 vnnttio:QA less severe
..--runb-a may be initiated to prevent a Reactor Low Water Level Scram upon loss of either an RFP OR condensate boos~ter_ pump,_when oper-11ngat high powr leve ls.fJ_ e FWLC System will initiate a 70% recirc runba-ck ft-less than three RFPs are running due to an auto-trip of a running feed pump, Reactor water level is less than approximately 26", and steam flow is > 85% - OR - if less than four condensate booster pumps are running, steam flow is > 85%, and feedwater flow is greater than 90% of rated.
- 6. RFP Low Suction Pressure Trips
- a. Upon detection of a RFP low suction pressure condition, the control system will first attempt to restore suction pressure using a staggerred pump trip approach but will trip all pumps if suction pressure drops too low. The staggered pump trip logic will trip Pump C if suction pressure is less than 125 psig for 3-5 seconds, pump B if less than 125 psig for 7-12 seconds, and pump A only if suction pressure is less than 90 psig for 5 seconds. All pumps will trip if suction pressure is less than 90 psig for greater than 5 seconds. The 90 psig setpoint for tripping all pumps protects the pump from running any significant length of time near min.
NPSH, and the time delay avoids unnecessary trips due to pressure spikes. Any Feed Pump trip on low suction pressure is considered a major and is annunciated as FWLC system trouble alarm on the 902-6 panel in addition to other annunciation for RFP auto-trip.
NRC CQPY #1
24i, ,**i*iiii *
`. ýý ~:SR M - Pints:1.0 Given the following conditions:
- 1/2B SBGT SELECT switch is in PRIM
- 1/2A SBGT SELECT switch is in STBY
- SBGT has received an initiation signal.
Which of the following conditions would result in 1/2A SBGT train flow increasing?
A. The inlet to B SBGT Train (1/2-7505B) fails to open.
B. A loss of Instrument Air to the flow control damper has occurred.
C. The SBGT failed to maintain Reactor Building to Outside DP more negative than
-0.25 inches.
D. A failure of the heater for the 1/2B SBGT to start.
Answer: A
ý-desti~on 2Details Question Type: Multiple Choice Topic: Question #24 (RO/SRO)
System ID: 9746 User ID: SR-7500-K21 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOA 7500-01, R. 14 User Text: 261000A1.01 User Number 1: 2.90 User Number 2: 3.10 Comment: New question. Higher. With the failure of the Primary Train "B" inlet to open, the "B" SBGT train cannot develop adequate flow, so the "A" SBGT train will start.
A loss of IA will cause the flow control damper to fail open, not closed. Reactor Building Delta-P is not directly controlled by SBGT. Heater ops do not affect SBGT flow.
OPERATIONS N.] CD Pa" of 130 10/15/02
QCOA 7500-01 UNIT 1(2)
REVISION 14 Continuous Use
-. /STANDBY GAS TREATMENT SYSTEM AUTO START A. SYMPTOMS
- 1. Possible alarms at panel 901(2)-3:
- a. G-3, RX BLDG VENT CHANNEL A HI HI RADIATION.
- b. H-3, RX BLDG VENT CHANNEL B HI HI RADIATION.
- c. E-3, RX BLDG VENT CHANNEL A DOWNSCALE.
- d. F-3, RX BLDG VENT CHANNEL B DOWNSCALE.
- e. G-16, FUEL POOL CHANNEL "A" HI RADIATION.
- f. H-16, FUEL POOL CHANNEL "B" HI RADIATION.
- g. C-16, FUEL POOL CHANNEL A DOWNSCALE.
- h. D-16, FUEL POOL CHANNEL B DOWNSCALE.
- 2. Possible alarms at panel 901(2)-5:
- a. A-8, GROUP 2 ISOL CH TRIP.
- b. B-13, CHANNEL A/B REACTOR LOW LEVEL.
- c. D-11, PRIMARY CONTAINMENT HIGH PRESSURE.
- 3. Possible alarm at panel 901(2)-55/56:
- a. A-l, DRYWELL HIGH RAD CONC.
B. AUTOMATIC ACTIONS CAUTION To ensure system performance, the 1/2 B SBGTS TRAIN MODE SELECTOR SWITCH should NOT be placed in the STBY position.
(F-17) I I I
- 1. For SBGTS selected to PRIM:
- a. IF the SBGTS initiation signal is from one unit only, THEN that unit's U-1(2)-7503 RB INLET DMPR TO SBGTS will 6pen while the other unit inlet damper will close.
- b. 1/2-7504A(B), TURB BLDG CLG AIR DMPR is CLOSED.
NRC CbPY #1
IM' S' 6ý00-K14 Olo~ins: 10 If the Unit Two ESS UPS fails an operator would verify that the ESS ASCO ABT has switched to A. MCC 28-2 B. MCC 25-2 C. Bus 28 D. Bus 27 Answer: A "b1uestio 2ýbta Question Type: Multiple Choice Topic: Question #25 (ROISRO)
System ID: 2582 User ID: SR-6800-K14 Status: Active Must Appear: No Difficulty: 2.75 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QOA 6800-03, R. 26 User Text: 262001 K1.04 User Number 1: 3.10 User Number 2: 3.40 Comment: ILT.05479 (77664) Bank question. Lower. MCC 28-2 is the backup for a failure of the UPS. Bus 28 is the normal power supply to the UPS. Bus 17 is the backup to the static switch, which is part of the UPS, so they may choose Bus 27, but on unit 2 it is Bus 26. MCC 25-2 is the backup to the Instrument Bus.
OPERATIONS R--ý C Pa" of 13 nn3171 10/15/02
QOA 6800-03 Revision 26 Continuous Use 120/240 VAC ESSENTIAL SERVICE BUS FAILURE A. SYMPTOMS
- 1. Alarms.
- c. 901(2)-8 F-8, ESS SERV UPS TROUBLE.
- d. 901(2)-8 E-9, ESS SERV BUS ON EMERG SPLY.
- e. 901(2)-5 B-16, CHANNEL B MAIN STM LINE HI HI RADIATION.
- f. 902-5 B-7, GROUP I ISOL CH TRIP.
- h. 902-5 D-15, CHANNEL B REACTOR SCRAM.
- i. 901(2)-4 A-1, RECIRC MG A SPEED SIGNAL FAILURE.
- j. 901(2)-4 A-5, RECIRC MG B SPEED SIGNAL FAILURE.
- k. Unit 2 only, 902-6 E-10, FW LEVEL CONT SYS TROUBLE.
- 2. Failure of the FW level control system. (FRy lockup)
- 3. Loss of Essential Service will disable RPIS and rod select. CRD flow control will be lost as will the recorders Panel 901(2)-5.
- 4. Unit 2 only, loss of Essential Service will disable Reactor Recirc Runback circuitry.
B. AUTOMATIC ACTIONS
- 1. 1/2 Group I Primary Containment isolation.
- 5. RWCU isolation will occur on a momentary or complete loss of Essential Service.
NRC COPY #1
/
250 VDC SWGR 18(28) SWGR 17 (26)
MCC-1 S R_8() MCC 18-2 (28-2)
Il
)
UPS NORMAL UPS ALT SUPPLY CUBICLE D4 II SUPPLY CUBICLE D6 ) ESS. BUS RESERVE INST.BUS NORMAL CUBICLE D4 RECTIFIER INPUT 63B 480 VAC VAC 63A 0
) CUBICLE D5 ESS RESERVE SUPPLY INVE REGULATOR AC.
INPUT 63 C SW) 3R -0 0r-:
UPS PANEL 90X-63 ASCO 120 VAC SWITCH ABT RESERVE POWER SUPPLY ESS BUS FIGURE 6800-02 REV. 0 ESSENTIAL SERVICE POWER SUPPLIES NRC COPY #1
ints: 1.00 Prior to closing the DtE- -EN-T-tBUS 24-1 breaker while synchronizing the Diesel Generator to Bus 24-1, the operator is to verify that the Diesel and Bus meet the requirementF synchronization.
This is done by verifying the synchroscope is:
A. rotating slowly in the slow direction with the synchroscope approaching o'clock position.
B. rotating slowly in the fast direction with the synchroscope approaching the 12 o'clock position.
ti C. rotating slowly in the fast direction with the synchroscope approaching the l o'clock position.
D. rotating slowly in the slow direction with the synchroscope approaching the 12 o'clock position.
Answer: f -P
-Question __b etilis Question Type: Multiple Choice Topic: Question #26 (RO/SRO)
System ID: 9733 User ID: SR-6600-K21 Status: Active Must Appear: No Difficulty: 2.25 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOP 6600-02, R.21, pg 6 User Text: 264000A4.02 User Number 1: 3.40 User Number 2: 3.40 Comment: New question. Lower. Per the procedure, synchronization is to occur when the synchroscope is rotating slowly in the fast direction and approaching the 12 O'clock position.
OPERATIONS 10/15/02 P] " o 130
QCOP 6600-02 UNIT 1(2)
REVISION 21 F.2. (cont'd)
- h. Adjust DG voltage to approximately 4160 volts with 1(2) DG VOLT REGULATOR (VARS) switch at Panel 901(2)-8.
... i. Synchronize across DIESEL GEN TO BUS 14-1(24-1) ACB:
(1) Turn on SYNCHROSCOPE for the 1(2) DG.
(2) Adjust 1(2) DG GOVERNOR AND VOLT REGULATOR (VARS) until the synchroscope is moving slowly in the FAST (clockwise) direction and INCOMING VOLTS is slightly higher than RUNNING VOLTS.
(3) WHEN synchroscope approaches twelve oIclock, THEN close DIESEL GEN TO BUS 14-1(24-1) GCB AND apply a small initial load of approximately 200 to 300 kw with DG GOVERNOR control switch.
(4) Turn off SYNCHROSCOPE switch.
- j. Gradually load 1(2) DG over 2 to 4 minutes to > 500 and < 2600 KW with DG GOVERNOR control switch and maintain outgoing VARS approximately one-half the DG KW value using VOLT REGULATOR (VARS) control.
CAUTION The following steps are provided to operate the 1(2) DG unloaded and with the Speed Droop set at 0 for circumstances that would require this type of operation (i.e., Tech Spec operability demonstration or other emergency situations as deemed necessary by the US).
F.3. IF DG will be manually started AND NOT loaded to the Electrical System, THEN:
S-a. Dispatch two Operators to DG Room to:
(1) Verify Governor SPEED DROOP set at 0.
(upper left knob)
NRC COPY #1
How is the amount of fuel regulated to the cylinders for the diesel when it is at speed?
A. The load limit control automatically controls the fuel rack position which controls the amount of fuel injected into the cylinders which controls the speed of the engine.
B. As speed changes on the diesel the governor changes the speed of the fuel pump to send the proper amount of fuel.
C. The fuel injectors are set at a predetermined value which will maintain the amount of fuel constant therefore maintaining speed constant.
D. The governor positions the fuel racks which controls the amount of fuel injected into the cylinders which controls the speed of the diesel as load is added or removed.
Answer: D IQuestion ý7 betaiis
.. .Question Type: Multiple Choice Topic: Question #27 (RO/SRO)
System ID: 595 User ID: SRN-6600-Kl 5 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LN-6600, pg. 11 User Text: 264000K4.06 User Number 1: 2.60 User Number 2: 2.70 Comment: LN.01212 (75667) ILT.01212 replaced redundant NLO.00118 Bank question. Lower. The governor postions the fuel racks to control the amount of fuel of fuel injected.
OPERATIONS 10/15/02 C Pa"o30
Content/Skills Activities/Notes Content/Skills Activities/Notes
- 7. Governor/Speed Control **SRN-6600-K14(h)
- a. Purpose'.Y a.Prps **NRN-6600-KI5(h)
The formula:
f = NP/120 where: f = freq.
P = # of poles N = speed (RPM)
Shows that by changing the speed of a generator, its frequency changes also. Therefore, the diesel speed control is also the generator frequency control.
All of the emergency diesel generators at Quad Cities (including the security diesel) use a Woodward UG-8 Speed Governor to control diesel engine speed and, consequently, generator frequency.
The UG-8 Governor is a hydraulically operated unit and uses an oil booster pump to charge the governor accumulators on a start. It normally will maintain engine speed regardless of the generator load.
The governor controls engine speed by regulating the *"
amount of fuel supplied to each of the engine cylinders "bymoving the fuel rack. There is a fuel rack on each sidej of the diesel. This rack is connected to each of the ten /
\ fuel injector pumps on the same side of the engine.
Lateral motion of the fuel racks rotates the plungers, which are located in the center of the fuel injectors.
Rotation of the plungers controls the amount of fuel oil injected into each cylinder during each stroke.
- b. Controls The synchronizer is a manual speed adjustment, used to set the desired engine speed when paralleling the diesel generator to its bus. The synchronizer is also used to set the desired load after paralleling. Synchronizing is normally accomplished remotely using the governor switch, which energizes a synchronizer motor (mounted on top of the governor).
It is possible to energize the synchronizer motor when the diesel is idle; however, this would change the position of the governor and on subsequent starts could cause the diesel to start up and run at low RPM (< 800 Q:trnopslp\LN-6600.DOC Page 11 of 56 NRC COPY #1
281: SR-41T00KlT 5~i Pohints: 1.0b An automatic actuatio of t fire p tection system for the New Computer Room has occurred.
Which of the following d crib the operational implications?
Both air conditioning units trip closing the intake damper, the room exhaust damper:
A. closes and the process computer is susceptible to errors in data processing and calculations at 80 degrees F.
B. remains open and the process computer will automatically trip.
C. remains open and the process computer is susceptible to errors in data processing and calculations at 80 degrees F.
D. closes and the process computer will automatically trip.
Answer: A Question Type: Multiple Choice oopic: Question #28 (RO/SRO)
System ID: 9778 User ID: SR-4100-K15 Status: Active Must Appear: No Difficulty: 3.25 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LN-5751, QOA 5750-12, R.6 User Text: 286000K5.02 User Number 1: 2.60 User Number 2: 2.60 Comment: New. Higher. The room exhaust damper closes on a fire and the process computer is susceptible to errors in daa processing and calculations at 80 degrees F.
h 'P0 OPEATONS~J
't t44f W~
c~Co S AA 10/15/02*
QOA 5750-12 Revision 6 Continuous Use LOSS.OF COMPUTER ROOM VENTILATION A. SYMPTOMS
- 1. Alarm.
- 2. Indicating light on local HVAC panel 2251-92, in main computer room indicates which unit has tripped.
B. AUTOMATIC ACTIONS
- 1. Standby HVAC unit will auto-start.
C. IMMEDIATE OPERATOR ACTIONS
- 1. None.
D. SUBSEQUENT OPERATOR ACTIONS
- 1. Verify standby HVAC unit auto-start.
- 2. WHEN problem is corrected, THEN push RESET button on the local control panel. The tripped unit will restart and standby unit will stop and be restored to standby mode.
- 3. IF ventilation can NOT be restored, THEN prior to reaching 85 0 F, shutdown the process computer and refer to QOA 9900-02, Loss of Plant Process Computer.
- 4. IF the process computer must be shutdown due to excessive room temperature, THEN turn off power to the process computer using the power switch on the front panel AND power supply toggle switches inside the cabinets.
E. DISCUSSION
- 1. Receipt of MAIN COMPTR RM HVAC SYSTEM TROUBLE alarm indicates that room temperature has reached 78 0 F. This high temperature is indicative of the operating unit malfunctioning. At this point, the standby unit will start and room temperature should decrease thereby turning off the alarm on the main control panel within 20 minutes. However, the indicating light at the local panel will remain on until the unit reset button is pressed.
NRC C 6 PY #1
Content/Skills Activities/Notes XXXI. COMPONENT DESCRIPTION
'. -- A. Air Handling Units S-5751-K14
- N-5751-K14
- 1. There are two air conditioning units located on the north side of the Service Building outside Trackway One. Each A/C S-5751-K15 unit has its own heating and cooling units. Each A/C unit **N-5751-K15 has a fan which is interlocked with the heater and condensing units. Normally the heating units do not operate even in the Show Figure 6, Computer Roomn winter due to the heat load generated by the computers in the HVAC Control Panel.
room. As a result if the heaters do become energized either manually or automatically a trip of the halon system may Q: What would be the result result due to accumulated dust on the heaters.
of a failure of the running
A: S/B should auto start @
- 1) and 25-2 (Unit #2).
78 0 F.
- 3. Normally one A/C unit will be running with the other in standby. The running A/C unit is selected using a priority switch located in the new computer room. If the selected A/C unit fails the standby unit will auto start when room temperature reaches approximately 78°F to control temperature.
- 4. Room temperature is controlled by thermostats located in the computer room. There is one thermostat for each A/C unit.
The thermostat are dual function and control heating and cooling.
- 5. The A/C units trip on an initiation of the Halon fire (r protection system.
B. Humidifier S-5751-K14
- N-5751-K14
- 1. The Armstrong Humidifier has an enclosed steam generator and blower to Control Roorrm humidity. S-5751-K15
- N-5751-K15
- 2. The humidifier is located along the north wall of the new computer roo0mi. --. .. ..
- 3. The humidifier is controlled by a humidistat to maintain the humidity within the desired band.
- 4. Power to the humidifier is supplied by 480 VAC MCC 15-1 and the humidifier blower is powered from the 120 VAC distribution panel on MCC 15-1.
Q:\tmopslp\LN-5751 .DOC Page 55 of 75 NRC COPY #1
Content/Skills
,a it Activities/Notes C. Dampers S-5751-K14
- N-5751-K14
- 1. Dampers are located in the intake and exhaust ducts to allow air flow in only one direction. S-5751-K15
- N-5751-K15
- 2. The room exhaust damper is interlocked to isolate the computer room in the event of an initiation of the Halon fire 'I protection system. The intake damper will close when the A/C unit trips off.
Q:\tmopslp\LN-5751 .DOC Page 56 of 75 NRC COPY #1
Points: 1.00 Both units are operating at full power with the plant in a normal configuration.
On a complete loss of instrument air, the emergency isolation dampers will fail and the fan dampers will fail A. OPEN; OPEN B. CLOSED; CLOSED C. CLOSED; OPEN D. OPEN; CLOSED Answer: C
_ uestion 2 Dtails~
Question Type: Multiple Choice Topic: Question #29 (RO/SRO)
System ID: 4576 User ID: SR-5750-K23 Status: Active Must Appear: No Difficulty: 3.50 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LNF-5750 User Text: 288000K1.06 User Number 1: 2.70 User Number 2: 2.70 Comment: LN.08733 (79666) ILT.08733 replaced redundant NLO.02245 On a loss of IA, the Emergency dampers close due to an accumulator. The fan dampers fail open on a loss of IA. Lo Wu c OP R TI. Il 5/..02 .10..
OPERATIONS 10115/02 Ký, F0?)C PaC~ of13 ý
Content/Skills A i e Activities/Notes
- 2. Cooling can be accomplished through the use of evaporative actvte/oe N-5750-Kl4a.(3) coolers, located downstream from the steam coil. SR-5750-Kl4a.(3)
- a. Spray pumps associated with the system located in the The evaporative coolers work on duct work spray water for collecting trays into the air the principle of evaporation. A stream to provide cooling. Water not evaporated falls small pump sprays water into back into the collecting trays and is recycled.
the ventilation air flow
- b. Makeup water for the collecting trays is provided by humidifying & cooling the air in the ducts. The water that does clean demin.
not evaporate falls back onto a The evaporative cooling system is not currently used due collection tray and flows back to to system inefficiency, but the equipment is still in the pump suction. This system is place. designed for about a 15'F dT intake to the building. Makeup water to the collection tray is automatically provided from the clean demin system.
C. Reactor Building Supply Fans N-5750-Kl4a.(4)
- 1. Located on the Turbine Building 658' 10" level above Buses SR-5750-Kl4a.(4)
X8 and X9. The supply system has three fans each with a capacity of 47,500 cfm. Normally two fans are operated for fll flow of 95,000 cfm, with the third fan in standby.
fua
- 2. -The uppy fans are direct connected, axial flow fans.
- 3. The supply fans utilize a constant volume control system to N-5750-K15c.
regulate the air flow from the supply fans. SR-5750-Kl5a.
- a. Vanes on the intake side of the fans are controlled by air operators to adjust the air intake opening.
- b. This controls the volume of air entering the system. The vanes are adjusted automatically by a pitot tube sensing downstream air velocity and static pressure.
- c. The controllers are in manual set at full flow due to supply and discharge fans "fighting" each other.
- 4. An air operated damper in the supply fan discharge is operated in conjunction with the fan to open when the fan is and close when the fan is shut off.
- a. The discharge dampers are energize-to-close. When the supply fan breaker is closed, a "b" contact opens to de energize the air solenoid allowing the damper to fail open.
Q:\tmopslp\Lnf-5750.doc Page 7 of 110 NRC COPY #1
Content/Skills Contnt/SillsActivities/Notes
- 5. Power supply to the supply fans are: Refer to M-4-1(2)-85-47 (OTR 89-58). This modification
- b. lB (2B) - Bus 18 (28) buttons on the 912-1 panel.
- c. 1C(2C)-Bus 18(28) Prior to this mod, these controls were only available on the D. Emergency Dampers 2251(2)-24X panel.
- 1. There are four emergency air operated dampers that shut N-5750-Kl4a.(5) during emergency conditions, to prevent the release of SR-5750-Kl4a.(5) contaminants to the environment. There are two dampers on the supply fn discharge duct and two on the exhaust fan N-5750-K15f.(2) inlet duct. ýThe dai *stItf6-close on any one of the" SR-5750-K15d.(2)
(followmiigconditions:
Refer to DVR 4-2-88-061, "Rx
- a. High drywell pressure (+2.5 psig). Bldg Vents Started Without Starting a Sample Pump" (OTR
- c. High drywell radiation (100 R/hr). Building exhaust fan without having the particulate sampler
- d. High refuel floor radiation (100 mR/hr).
turned on. The Unit NSO did not recognize the significance of
- e. High radiation level in the Reactor Building vent exhaust duct (10 mR/hr). having the "Rx Bldg Stack Monitor Low Flow" alarm up.
- f. Rx Bldg vent exhaust or Refuel floor radiation detectors This alarm is annunciated on the downscale. 90X-3 panel.
g( Low instrument air pressure at the damper (65 psig).j Prior to barrier fuel (when there were leakers). A typical release
- 2. These four (4) emergency air operated dampers may also be rate was 600 micro ci/sec, now a secured in the closed position by use of a manual handwheel typical release rate for Unit One operator. The Hand wheels are located below the individual is 1 micro ci/sec and U-2 is not dampers on the Turbine Building 658'10" level on each side detecting any release.
of the supply duct.
Effluent air is also sampled to check for
- 3. The isolation dampers are energized to open, and require aiýr tritium. The air sample must be frozen to.open andajir to close. and the frozen condensation is sampled for tritium. This must be done (frozen) because tritium emits a very low
- a. A 4-way solenoid valve will port air to the top of the air energy beta which would otherwise be operator to open the damper, when the solenoid is undetectable.
energized, and allow the underside of the air operator to vent to atmosphere. Halogens are sampled weekly.
. .... .... .. . .. . . . . . . . . .. . . .. .I1. .. .. . .. . . ..: S -- .. . . . . - .'T' _'.
- z_-
Q:\tmopslp\Lnf-5750.doc Page 9 of 110 NRC COPY #1
Content/Skills S* ........ v~es/Not, *VV B. Automatic Functions
- 1. Initiation Device/Setpoint Response Standby Supply Fan Auto Start. / Low On sensed low air flow the standby air flow and control switch in normal Reactor Building supply fan will auto after close or trip and all Rx Bldg start if the limits are met.
isolation dampers are open, and Rx Bldg atmosphere is less than .6" H20 above atmospheric pressure and the maintenance switch is in normal.
Reactor Building vents interlock. / Rx On an isolation of the Rx Bldg vents the Bldg vents system isolated. SBGT system auto starts.
- 2. Trips and Isolations Purpose Device/Setpoint Bypass/Reset Response Protect the fan Flow switch / Low Bypassed on On low flow the from overheating.
N-5750-K15f.(1) air flow as sensed startup by holding fans associated Dp by a Dp cell control switch in SR-5750-K12 switch 5703 closes between the fan closed position. / and energizes a LF-5750-K12 suction and Automatically control relay to SR-5750-Kl5d(1) downstream of the resets, close a contact in outlet damper. the fan breaker trip circuitry. On a fan start the C/S must be held closed for at least 5 seconds I) tto prevent a trip due to low flow.
Protect the fan Undervoltage relay No bypass. / Relay closes a from overheating / 80% normal Manually reset at contact in the fan due to sustained voltage. the fan breaker, breaker trip higher than normal circuitry.
current levels. Following reset, fan can be manually restarted or will start on low air flow provided C/S is in auto.
Q:\tmopslp\Lnf-5750.doc Page 21 of 110 NRC COPY #1
Content/Skills Activities/Notes Activities/Notes V. SYSTEM INTERRELATIONS
. A. Support Systems
- 1. 480 VAC N-5750-K19a.
- a. Power supplies to the supply fans are: SR-5750-K19
- 1) 1A (2A) - Bus 19 (29) SRN-5750-K23a., b., c.
- 2) 1B (2B) - Bus 18 (28)
- 3) 1C (2C) - Bus 18 (28)
- b. Power supplies the exhaust fans are:
- 1) 1A (2A) - Bus 18 (28)
- 2) 1B (2B) - Bus 19 (29)
- 3) 1C (2C) - Bus 19 (29)
- 2. 125 VDC
- a. Supplies control power to the fan breakers for remote
,start and tripfunctions.
- b. Supplies power to the fan damper and the isolation damper air operator solenoids. On a loss of 125 VDC, the fan dampers will fail open while the isolation dampers will fail closed.
- 3. Instrument Air .
Q: Should instrument air fail, Provides the motive force to operate fan and emergency \\,
N how could you close the Rx isolation dampers. On a loss of instrument air, the discharge building ventilation isolation "damperswill fail open and the isolation dampers will close dampers?
, due to an accumulator.
A: Use the manual hand wheel
- 4. Process Radiation Monitoring System to close them.
- a. Chimney Gas Radiation Monitoring The Chimney Gas Radiation Monitoring System alarms when the radiation level of the effluent gases being discharged exceeds the prescribed limit.
- b. Reactor Building Ventilation Radiation Monitoring System Q:\tmopslp\Lnf-5750.doc Page 31 of 110 NRlC COPY #1
30 ID: SR5750KF20 Points: 1.00 A storm front is approaching causing atmospheric pressure to drop.
How will this be indicated in the Control Room and what is the expected system response?
Reactor Building Delta-P will _and Rx Building Exhaust Fan Vortex dampers will __
further.
A. close B.) close C. i (ee4,,,) open D. deefeaas* bGo&el ative open Answer: D Question Type: Multiple Choice Topic: Question #30 (RO/SRO)
System ID: 9770 User ID: SR-5750-K20 Status: Active Must Appear: No Difficulty: 3.75 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LNF-5750, pg 12 & 37 User Text: 290001A3.02 User Number 1: 3.50 User Number 2: 3.50 Comment: New question. Higher. A drop in atmospheric pressure will cause the Delta-P to drop since the RX BLDG is at a vacuum. This will cause the exhaust dampers to open to restore the DP.
OPERATIONS WC mrJ Paof 130 10/15/02
Content/Skills Activities/Notes
- 9) High Drywell radiation (IOOR/hr).
- 10) Low reactor water level (0 inches).
- 11) Refuel floor or Rx Bldg radiation detectors downscale.
- 2. There are air operated adjustable vanes in the fan intake that operate to control flow by receiving inputs from a flow element downstream of the fan and repositioning the vanes accordingly.
- 3. There are air operated dampers in the fan discharge that provide isolation of the fan when it is not running.
- a. The dampers are interlocked with the fan to open when the fan starts and close when the fan stops.
- b. The damper air solenoid valves normally receive power from the fan 125 VDC control power circuitry.
- c. On an OOS when 125 VDC power is removed from the fan, a maintenance switch can be used to supply an alternate source of power to the damper to maintain it closed while the fan is being worked on.
C. Reactor Building Exhaust Ventilation
- 1. There are three 50% capacity fans located on the Turbine Building 658' 10" level above Buses X8 and X9.
- a. Normally two fans are running with one in standby.
- b. The standby fan will auto start on a low flow condition under the same conditions as a standby supply fan.
- c. The exhaust fans have all the same trips as the supply fans with the exception of the + 1.0" H 2 0 supply fan trip.
An extreme negative differential pressure, greater than 1.0" H 2 0 will cause the Reactor Building exhaust fans to trptrip.
- 2. There are air operated adjustable vanes in the exhaust fan intake that are repositional according to inputs from a Dp cell to maintain the Reactor Building to outside differential pressure.
- 3. Each fan has an air operated damper in the fan intake that provides isolation of the fan when it is not running.
Q:\tmopstp\Lnf-5750.doc Page 37 of 110 NRC COPY #1
Content/Skills Ar-fij-Mn4--1K1rf Nl
- b. This vacuum is maintained by a building differential N-5750-K15d.
pressure control system. (Shown on M- 1531)
SR-5750-Kl5b.
/ 1) Differential Pressure Transmitters DPT- 1-5741-2A, 22B, 2C & 2D which measure atmospheric pressure Q: How does Rx building with respect to Refueling Floor pressure pass their ventilation maintain a negative lowest differential pressure thru selectors C-1 A, B, pressure in the Rx building?
C & D to Receiver Controller T3 which modulates both Units 1 & 2 Exhaust Fan Vortex Dampers to A: Supply fans run at full flow maintain a minimum building differential pressure. (not restricted). Exhaust fans, which are larger capacity,
- 2) This is with the Unit I Controller in control. If throttle flow based on building necessary, the Unit 2 Controller can be valved in and pressure as sensed from refuel control AP. If both controllers are inoperative, they floor.
can be bypassed and D/P controlled manually via a PCV.
- 3) Differential pressure sensors (4) are located on each Show Figure 3, Rx Building wall of the refuel floor (690' el.). The D/P Differential Pressure Control controller, bypass valves, and pressure control valves System.
are located at their respective local panel 2251(2) 24X (658'10" el.) near the exhaust fans. The D/P controller(s) may be bypassed and isolated and a manual PCV 1-5999-607 may be adjusted to vary the
""- air supply pressure to the exhaust fan vortex dampers. The vortex damper operation controls Reactor building D/P.
- c. Assisting in the effort for pressure control are the exhaust N-5750-Kl4a.(4) fans themselves. The fans are identical to the supply SR-5750-Kl4a.(4) fans but have a larger capacity than the supply fans. The exhaust system, has three fans, each with a capacity of N-5750-K15f.(1) 55,000 cfm. Normally two fans are operating for a full SR-5750-K15d.(1) flow exhaust of 110,000 cfm, with the third fan in standby.
- 1) The standby exhaust fan will auto start on a low flow condition under the same conditions as a standby supply fan.
- 2) The exhaust fans have all the same trips as the supply fans with the exception of the +1.0" H20 supply fan trip. An extreme negative differential pressure, greater than -1.0" H20 will cause the Reactor Building exhaust fans to trip.
- d. Power supply to the exhaust fans are:
- 1) lA (2A) - Bus 18 (28)
Q:\tmopslp\Lnt-f750.doc Page 12 of 110 NRC COPY #1
'31 11'YSN161&fj
~Points: i'0 Both units are operating at full .
The Unit one Torus develops lekn the ECCS suction header.a If no operator action is taken Rx building basement Torus area water levels:
A. and reactor building atmosphere will be unaffected. CAmoý B. will increase,, reactor building atmosphere will be unaffected.~J14 C. i not be affecte ,but the reactor building will be uninhabitable due to oxygen de icient atmosph ere.
- 0. will increase, followed by the reactor building being uninhabitabl due to oxyge n deficient atmosphere.
'T r ,d Answer: D A
Question Type: Multiple Choice GW-- ~ Topic: Question #31 (ROISRO)
System ID: 9820 /Y User ID: SRN-1601-K-2 v4d (
Status: Active Must Appear: No Difficulty: 2.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOA 1600-05 R6 User Text: 290001 K6.04 User Number 1: 3.90 User Number 2: 4.10 Comment: New question. Higher. Answer is correct due to ECCS suction header location near the bottom of the-t~ru an Iduring normal plant operation the torus is iner~te~di PtI I OPERATIONS 10/15/02 L~C~Cýpy JO
QCOA 1600-05 UNIT 1(2)
REVISION 6
.................... .................... _ -- -- C o n t i nuous Us e A. SYMPTOMS A.1. For leak above Torus water line:
- a. Unable to maintain Drywell/Torus Differential Pressure per the applicable procedure:
(1) QCOP 1600-14.
(2) QCOP 1600-21.
- b. Loss of Torus Atmosphere if inerted, Alarm 912-7 A-4 (B-4), DRYWELL 1(2) 02 CONTENT HI.
A.2. For leak below Torus water line:
- a. Decrease in Torus Level, Alarm 901(2)-3 A-14, TORUS HIGH/LOW LEVEL.
- b. Reactor Building Sump Pump Flow abnormally high:
(I) Alarm 901(2)-4 C-18, RX BLD FLOOR DRAIN SUMP A HIGH LEVEL.
(2) Alarm 901(2)-4 D-18, RX BLD FLOOR DRAIN SUMP B HIGH LEVEL.
- c. Reactor Building Basement ARM reached trip setpoint, Alarm 901(2)-3 A-l, RX BLDG HI RADIATION.
- d. Auto Start-of SBGT System if Reactor Building Vent Radiation > 3 mr/hr.
- e. Visual observation of a leak.
B. AUTOMATIC ACTIONS None.
C. IMMEDIATE OPERATOR ACTIONS None.
ou". .. I . . 1 "' l : I ..
NRC CbPY #1
During a loss of Service Water, which ONE of the following systems can supply cooling water to the CR HVAC "B" AHU air conditioning unit?
A. Circulating Water (CW)
B. Turbine Building Closed Cooling Water (TBCCW)
C. Reactor Building Closed Cooling Water (RBCCW)
D. Residual Heat Removal Service Water (RHRSW)
Answer: D
,Question 32 Details Question Type: Multiple Choice Topic: Question #32 (ROISRO)
System ID: 6686 User ID: SR-5752-K23 Status: Active Must Appear: No Difficulty: 2.25 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOP 5750-9,R.26, pg. 6,7 User Text: 290003K1.05 User Number 1: 2.80 User Number 2: 3.00 Comment: LWQ.00260 (81783) Bank question. Lower. RHRSW is the backup. TBCCW, RBCCW and Circ Water are not.
-- Maw AM - -'. -
OPERATIONS
[LI]
RC Pa of 130 iib 1f 10/15/02
ýy QCOP 5750-09 UNIT 1/2 REVISION 26
- :-.i*¸=. ... " '. _.* REVISION . '*26*
.F.3. IF Control Room B Train HVAC is to be started, THEN:
Steps F.3.A., F.3.b., and F.3.c. ensure the Air Handling Unit (AHU) will auto-start and remain operating during a degraded voltage condition, and further will NOT cause an undetectable voltage condition on Bus 18 or associated MCCs.
(H.8.a.)
S--------------------------------------------------------------------------------------------------------------------------------------------- ----------------
- a. At Panel 1/2-9400-105, verify:
(1) AIR HANDLING UNIT "B" in AUTO.
(2) A/C UNIT "B" COMPRESSOR in AUTO.
- b. At Panel 1/2-9400-102, verify the STOP/RESET-OFF-AUTO switch in OFF.
- c. JF alarms 901(2)-5 B-10, CHANNEL A REACTOR LOW LOW LEVEL, and 901(2)-5 B-15, CHANNEL B REACTOR LOW LOW LEVEL, Q. . 901(2)-5 D-11, HIGH DRYWELL PRESSURE (a valid LOCA signal is present), THEN open breaker Bus 18 cubicle 5D, TURB &
REACTOR BLDG LIGHTING lB.
> -191", THEN open the following breakers: (H.8.d)
(1) Bus 19 cubicle 4D, REACTOR BUILDING LIGHTING lB.
(2) Bus 28 cubicle 5D, TURB & REACTOR U
BLDG LIGHTING 2B. U (3) Bus 29 cubicle 4D, REACTOR BLDG LIGHTING 2.
- e. IF Service Water is to be utilized for cooling water to the RCU, THEN at Panel \
1/2-9400-105, verify the A/C UNIT "B" COOLING WATER SUPPLY SELECTOR switch in I
=77ý NORMAL.
NRC CQPY #1 -U.
QCOP 5750-09 UNIT 1/2 REVISION 26 F.3. (cont'd)
- f. jF RHR Service Water System is to be utilized for cooling water to the RCU, THEN:
(1) Verify RCU is NOT operating.
(a) At Panel 1/2-9400-105, place A/C UNIT "B" COOLING WATER SUPPLY SELECTOR switch in EMERG.
The valves listed in following 2 steps are located on the CRD level.
(2) jF 1(2)A _OR 1(2)B RHR Service Water Pump is to be run, J1N verify:
- (a) 1(2)-5799-385, CR HVAC TRAIN B
__.... RCU RHRSW SPLY FR PMPS 1(2) 1001-65A & 65B OUTBD SV, open.
(b) 1(2)-5799-384, CR HVAC TRAIN B RCU RHRSW SPLY FR PMPS
( 2 )-1001-65C & 65D OUTBD SV, closed.
(c) 1(2)-5799-406, CR HVAC TRAIN B RCU RHRSW SPLY FR PMPS 1(2)-1001-65C & 65D INBD SV, closed.
(d) IF 2A OR 2B RHR Service Water Pump is to be run, THEN verify 2-5799-407, CR HVAC TRAIN B RCU RHRSW SPLY FR PMPS 1(2)-1001-65A & 65B INBD SV, open.
(3) IF 1(2)C OR I(2)D RHR Service Water Pump is to be run, THEN verify:
(a) 1(2)-5799-385, CR HVAC TRAIN B RCU RHRSW SPLY FR PMPS
..(2)-1601-65A & 65B OUTBD SV, closed.
NRC COPY #1
Which of the following examples illustrates the proper format for 3-way communications using the plant radios per Operations expectations?
A. "Unit I EOAYnit I NSO$erform prestart checks on the 1Bo4b CRD pump".
"Unit I NSO Unit 1 EO`1 Understand perform prestart checks on the'lBfiW CRD pump".
"Unit I EOUnit I NSO"Aat is correct."
tr- "U pB.
Erestart checks on the 1 Bravo CRD Pump"
" UOO" Perfor mstart checks on the 1 Bravo CRD pump"
/
C. "Unit 1 EOIunit 1 NSOfjrform prestart checks on the 1Brdv; pump".
"Unit I NSO, Understand perform prestart checks on the 1Bray D mp" "That's correct" D. "Unit I EO,UnitNSO rform prestartchecks on the u "Unit 1 NSO, Unit I EO; Understand perform prestart k" "Unit 1 EO, Unit 1 NSOVz4/at is correct". ,e tlr Answer:
Questi on 33 Details Question Type: Multiple Choice Topic: Question #33 (ROISRO)
System ID:
.JJser ID: SR-CROP-K04 Status: Active Must Appear: No 1.00 \
Difficulty:
0 Time to Complete: 1.00 ( ,,,
Point Value: CREW-COM R. 4, pg. 8 Cross
Reference:
User Text: G 2.1.17 User Number 1: 3.50 User Number 2: 3.60 Comment: LN.00064 (82488) New question. Lower. Answ r is correct due to requirement of sing-directed-3-v co°munie°aions°itfi°brepeat backs and use of phonetic
..Distractofo
-d-6-not contain all of the-seýee ms:-
OPERATIONS Pa" f3 10/15/02
Content/Skills Activities/Notes
- d. When using the radio or party-line communication S/R/A/B/FHIL/W
- -.- -.-... . systems, the sender and receiver identification should be included in each message.
- e. Orders should be simple and not contain more than two actions. It is preferable to have the individual call back to receive additional instructions.
- f. Complex instructions, may require written guidance to ensure important information is not forgotten. DO NOT rely on memory. When a complicated procedure is to be performed by a number of personnel at different locations, each working group should have the procedure/applicable checklist available to ensure proper work interaction, sequencing and communications take place.
- 2. Repeat-Backs
- a. Oral communications shall be repeated back to the extent n to allow the sender to ensure the order is corrpadly-tundersto6d. Complete ver-hi r(wrfo wordT)repeat-backs normally shouldI t-e necessary.
For repetitivehvu--tio g,th*-e'--6fuV-s 6"tim --e-st
" - s--
hall not be required, provided the details of
- -,5XV etfd communications are discussed or written out in advance.
,/-Forexample:
ORDER: Unit 2 EA, Unit 2 NSO. Reset the Unit 2 Main generator 86 device.
REPEAT-BACK: Unit 2 NSO, Unit 2 EA.
Understand, reset the Unit 2 Main Generator 86 device.
/ CONFIRMATION: Unit 2 EA, Unit 2 NSO. That is correct.
- b. It is important.to accomplish the confirmatory step because a lack of response by the originator will be interpreted as silent confirmation that the repeat-back was correct.
- c. The person originating the order shall insure the receiver has properly understood the order. If the repeat-back
- .nalcatesthe-receiver incorrectly understands the order,
- '*-J--~-~
. .the sender shall state "Wrong" and repeat the order.
TRNOPSLPRCREW-COM.R04 Page 8 of 41 IN,H L_5 ý: U I--` Y *F 1
- J4 ID
- LIC-ECCS ~Points:1.00i Part of the overall ECCS design bases is to:
A. prevent fuel cladding melting for any mechanical failure of the primary system up to and including a break area equivalent to the largest primary system pipe.
B. prevent fuel cladding melting for any mechanical failure of the primary system with at least one source of offsite power.
C. provide a barrier which in the event of a loss of coolant accident will control the release of fission products to the secondary containment and limit the release of radioactive materials to the environment.
D. provide a means of alternate core cooling following a shutdown from 100% rated thermal power when the reactor is isolated from the condenser and shutdown mode of RHR is unavailable.
Answer: A usion ea Question Type: Multiple Choice
,Topic: Question #34 (ROISRO)
System ID: 9779 User ID: LIC-ECCS Status: Active Must Appear: No Difficulty: 3.25
,- ;,:Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LIC-ECCS pg. 2 User Text: 2.1.27 User Number 1: 2.80 User Number 2: 2.90 Comment: New. Lower. ECCS is designed to prevent fuel clad melting on the largest primary system pipe break.
OPERATIONS 10/1of/02 10/15/02 OPEATOr P
Page 2 ot t
A. Content/Skills / *Sugge instructor
- 1. BRIEF DESCRIPTION A. Purpose The Emergency Core Cooling System (ECCS) is made up of the High Pressure Coolant Injection (HPCI) system, the Automatic Depressurization System (ADS) the Core Spray Sysitem, and the Low Pressure Coolant Injection (LPCI) System.]
The arrangement of these ECCS subsystems protects the reactor core against fue.l cladding damage across the entire spectrum of line break accidents. Additionalli the ECCS, in conjunction with the primary and secondary containments, limits release of radioactive materials to the environs following a loss-of-coolant accide (LOCA), so that resulting radiation exposures are kept to a practical minimum.
are within the guideline values given in published regulations.
B. Design Bases The ECCS is designed to provide protection against the postulated LOCA caused ruptures in the primary system piping. During normal operation, when normal auxiliary power is available, heat is removed from the core through the steam-turbine-condenser cycle or through the use of the shutdown cooling mode the RHR system. These are the normal provisions for core cooling. When the "reactoris isolated from the main condenser and the shutdown mode of the RHR system is unavailable, or when electrical power is unavailable to pump cooling water to the condenser and heat exchangers, the core is cooled by relief valve action followed by use of the Reactor Core Isolation Cooling (RCIC) system. It assumed that there is no loss of coolant from the primary system.
However, during postulated accident conditions, other means are needed to provic continuity of core cooling where it is assumed that mechanical failures occur in t primary system and coolant is either partially or completely lost from the reactor vessel, and (1) normal auxiliary power is unavailable to drive the feedwater pump or (2) the loss of coolant occurs at a rate beyond the capability of the feedwater I
system. Under these circumstances, core cooling is accomplished by means of thd ECCS.I Each of the ECCS subsystems was designed to specific design'bases; however the 1
overall ECCS design bases are: .-----
- 1. C he ECCS is designed to prevent fuel cladding melting for any mechanical' failure of the primary system up to and including a break area equivalent to \.
the largest primary system pipe.
TRNILTsY\.ECcs.R02 LEFT HAND SIDE NRC COPY #1
I I ., Iý-ý-sý 35 ID: SR-0202-K28 Po~7ints:1.00 Plant conditions are as follows:
- Unit Two is recovering from a scram.
- Preparations are underway to start-up the 2B recirc. pump.
- 2A recirculation pump is running at 32% speed.
- Reactor vessel dome pressure = 980 psig.
- A recirc loop temperature = 540 degrees F.
- B recirc loop temperature = 500 degrees F.
- Bottom head coolant temperature = 390 degrees F.
Which of the following describes the limitations, if any, imposed on starting the 2B recirc pump?
A. The pump should NOT be started because bottom head coolant temperature is too low.
B. The pump can be started immediately.
C. The pump should NOT be started because the 2A recirc pump is running too fast.
D. The pump should NOT be started because the loop differential temperature is too
- .. . . . . . . . .h ig h .
Answer: A
'dues'6ion 3 5ti Question Type: Multiple Choice Topic: Question #35 (RO/SRO)
System ID: 9818 User ID: SR-0202-K28 Status: Active Must Appear: No Difficulty: 3.50 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOP 0202-02, R.23, E.7.a User Text: G.2.2.1 User Number 1: 3.70 User Number 2: 3.60 Comment: Modified from INPO bank # 14486 Higher. Exceed 145 degrees Delta-T between bottom head and reactor coolant temperature. Loop delta-T is 40, limit is 50.
vJIFR\ýC OPERATIONS Pa of 13 n n /1 10/15/02
(§4
iY QCOP 0202-02 UNIT 1(2)
REVISION 23 E.3. Unit 2 only: ©EIF Total Core Flow is less than 49 Mlb/hr, AND Flow Control Line is greater than 59.4%, THEN refer to QCOA 0400-02.© (H.7.a.)
E.4. WHEN Reactor core is defueled, THEN only one Recirc Pump should be operated. This is to prevent damage nuclear instrumentation due to flow induced vibration. to (H.8.e)
E.5. WHEN both recirculation loops are in operation, THEN jet pump loop flow mismatch shall be maintained within
_:10% of rated core flow when operating at < 70% of rated core flow and < 5% of rated core flow when operating at > 70% of rated core flow. (H.l.a.)
NOTE Steps E.6 and E.7 only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup. (H.l.c.)
- ::~~~~~~~~~~~~~~~~~---
- =-------* :------* ---- ..
E.6. WHEN starting the first recirculation pump, once
- j o - within 15 minutes prior to each startup, THEN:
- a. Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is < 145OF AND verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is < 50 0 F.
(H.l.c.)
E.7. WHEN starting the second recirculation pump, once within 15 minutes prior to each startup, (H.1.c.) THEN:
- a. Verify the difference between the bottom head coolant temperature and the reactor pressure S- . vessel (RPV) coolant temperature is < 145 0 F.
- b. Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is < 50°F.,
- c. Speed of operating recirculation pump must be less
_,.......___ . . . ..than 45%.
NRC C6PY #1
The illuminated red light above the Relief Valve Control Switches indicates the (1)- is activated on Unit One and the _(2) is activated on Unit Two.
44 NM>
(1) valve solenoid open limit switch; (2) valve position reed switch (1) valve solenoid open limit switch; (2) valve solenoid open limit switch C. (1) valve position reed switch; (2) valve position reed switch D. (1) valve position reed switch; (2) valve solenoid open limit switch Answer: /A bý uesti&6ý Details Question Type: Multiple Choice
.- Topic: Question #36 (RO/SRO)
.=-:M System ID: 9787 User ID: SR-0203-K20 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LIC-0203 R. 9, pg. 11 User Text: G.2--.
User Number 1: 3.70 User Number 2: 3.60 Comment: New question. Lower. Unit 1 valve indication is from a limit switch inside the solenoid. Unit 2 valve indication is from reed s OPERATIONS 10/15/02 LY C PaC of 3 0P
/~ z>
ý Content/Skills At-fiuifiaQ/KJ#%*g%4ý I
- 3. T-quencher To reduce the hydraulic loading on the torus, each ADS tailpipe is equipped with a T-quencher which acts to distribute the flow over a larger area.
D. Electromatic/PORV/Target Rock Relief Valve Position Indication Three means to determine the position of the electromatic/Target Rock relief valves are available:
I. Above the control switch on the 901(2)-3 panel, red and S/R-0203-EK005 green lights indicate open and closed respectively. **S/R-0203-EK020 (Figure 0203-5)
- a. A limit switch inside the solenoid, indicating only Show Figure 0203-5 whether it is picked up or dropped out, provides position indication for the electromatic valves on Unit 1. Q: What does the ERV open and close lights on the 901-3
- b. The PORVs have Reed switches that are actuated by the panel indicate?
valve pilot rod. This ensures actual valve position indication. A: Electromatics - Indicates whether the solenoids are pick
- c. Pressure switches in the air supply line act as the position up.
indicators for the Target Rock valve.
Target Rock - Indicates wheth, Below 40 psig: valve closed light is energized. or not air pressure is availabiP Above 50 psig: valve open light is energized.
- d. Because the 901-3 panel indications are not actually dependent on the position of the relief valve itself on U-1, they cannot always be counted on as accurate position indication.
- 2. An acoustic monitor is provided on panel 901(2)-21 to Show Figure 0203-6 monitor the status of the electromatic relief valves/PORVs, the Target Rock safety/relief valve and the safety valves S/R-0203-EK005 (Figure 0203-6). The acoustic monitors have the following S/R-0203-EK016 indications and controls: S/R-0203-EK020
- S/R-0203-EK021
- a. A green closed LED light is provided for each valve. It indicates the valve is closed.
- b. A red open LED light is also provided for each valve. It comes on when the acoustic monitor output reaches the trip setpoint, which indicates flow through the valve.
r.Vrfl I\iJXIflflO1fl,.tA.t..,
LrL&.-UUj UOC V:,\ IIM*raLI'\x1-1k_02U3 (dOc NRC COPY #1 Page 11 of 37
Points. 11.0Y Which one of the following is a prerequisite to Purging/Deinerting the Primary Containment through SBGT?
A. Both divisions of Rx Bldg Vent rad monitoring must be verified operable within four hours pr-* - * . ;'*-*n-rti-g.
B. The drywell and torus pressure must be equalized within one hourpfie comn*rgpur~ge/*einerti-preto.
C. Torus must be vented for four hours p.i ..
D. Both the drywell and torus must be sampled within eight days p purjk~g/d~lrWM.l Answer: D di~iqesiiion 37 Details Question Type: Multiple Choice Topic: Question #37 (RO/SRO)
N -
System ID: 9740 User ID: SR-1602-K28 Status: Active Must Appear: No Difficulty: 3.50 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOP 1600-07 R. 16 C.5 User Text: G 2.3.9 User Number 1: 2.50 User Number 2: 3.40 Comment: Bank question. Lower. Drywell has been sampled within 8 days per step C.5 of QCOP 1600-07.
OPERATIONS I C a" of10f.J
~
10/15/02
9/3 7 QCOP UNIT 1600-07 1(2)
/ jREVISION 16 Continuous Use DEIEwN RIAYCNANENT WITH SBGTS A. PURPOSE The purpose of this procedure is to provide the necessary steps for de-inerting and purging the Primary Containment with the SBGTS as well as steps necessary to transfer de-inerting line-up from SBGT to Reactor Building Ventilation System.
B. DISCUSSION B.1. The Chemistry Department will provide a venting path recommendation on a weekly basis based on routing Drywell/Torus air samples. Changes in vent path recommendations will be provided by the Chemistry Department based on indications of significant fuel degradation as determined by Chemistry Department interpretation of increased activity on these routing samples. Non-fission product sources of activity in the Drywell/Torus should NOT change the vent path recommendation even when HPCI, RCIC, or Main Steam Relief Valves are operated.
- a. Rad Protection Department is to be contacted when sampling the containment for de-inerting purposes.
Chemistry Department is to be contacted when sampling the containment for purposes of weekly surveillances while inerted. The reason is that Chemistry owns equipment accurate in the low ranges of oxygen while inerted and RP owns the equipment that is accurate in the higher ranges of oxygen that are expected when de-inerting.
B.2. It may be desired to transfer from SBGT purge and vent line-up to Reactor Building Ventilation to ensure adequate dilution and/or flow, or to accomplish de-inerting and purging with greater flow. This procedure contains steps to transfer suction from SBGT System to the Reactor Building Ventilation.
C. PREREQUISITES C.l. IF the SJAE monitor is available AND the Offgas activity has increased by > 50% within the last four hours after factoring out increases due to power, THEN notify the ni Supervisor to request the Chemistry
. .partment.to provlde.. a new venting path recommendation.
N~, H'- Ut. U,.t.) I-'- Y V 1
Why is it NOT permissible to run the Mechanical Vacuum Pump when the reactor mode switch is in the RUN position?
A. Because the SJAE's are required to be on when the mode switch is in RUN and they both use the same suction path.
B. Because the Mechanical Vacuum Pump would trip on high temperature once steam was being dumped to the condenser through the bypass valves.
C. Because this would bypass the Low Condenser Vacuum scram with the mode switch in RUN.
D. Because this would provide an unfiltered release pathway to the Main Chimney.
Answer: D uestion 38.'Deails Question Type: Multiple Choice Topic: Question #38 (ROISRO)
System ID: 2277 User ID: SR-5400-K28 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
TS 3.3.7.2 Bases User Text: G. 2.3.11 User Number 1: 2.70 User Number 2: 3.20 Comment: LN.05154 (77356) Modified question. Lower. The Mechanical Vacuum Pump bypasses the off-gas train.
OPERATIONS 10/15/02
- R C Pa"
- f* 0 p -
Content/Skills 1ý H Cten..S.Activities/Note
- d. Turbine Building Ventilation.
- e. Offgas Building Ventilation.
f Offgas Recombiner Rooms Ventilation.
- g. Gland Steam Condenser Exhauster.
- h. Mechanical Vacuum Pump.
- i. Standby Gas Treatment Trains.
- j. Offgas System.
- 2. Sources of Activity
- a. The main source of activity leaving the chimney is Xe133 with a TI/ 2 of 5.3 days.
- b. The major source of long-lived activity is Kr8' with a TI/2 of 10.5 years. Since the fission yield of KR8 5 is only 0.285% it is not a major gas source for the offgas system.
Q. Mechanical Vacuum Pump SR-5400-K14.
SR-5400-KI5
... y mechanical vacuum pump is provided in each unit to
- N-5400-K02 evacuate air from the turbine and Main Condenser and
- N-5400-K14 establish a sufficient vacuum during plant startup. One
- N-5400-K15 vacuum pump suction line services each Main Condenser
- N-5400-K19 shell via two air release valves in each shell.
- N-5400-K05 The pump discharges to the base of the chimney via the **N-5400-K22 gland seal holdup line.
Show Figure 5400-06,
- 3. The pump is designed for a flow rate of 2320 scfrn at 15" Condenser Vacuum Hg. Pump/Gland Exhaust Arrangement
- 4. The pump requires 3 to 4 gpm of clean demin water to lubricate the pump and hold the discharge air temperature below 200 F. Clean demin flow as well as discharge temperature indication is available outside the SJAE room.
- 5. By Administrative Procedure, the pump cannot be operated
( with the Mode Switch in the RUN position.
Q: TRNOPS LP.LN-5400.DO(K Page 20 of 58 NRC COPY #1
q z
0 0 4-0-u SJAE 2B ____________*
FIGURE - 7 REV. 0 SJAE CONDZNBSR VACUUM PUMP/
GLAND HZXRAUBT ARRANGEMNTW I I LGLAND EXHAUST ARRANGEMENT 2B
Mechanical Vacuum Pump Trip Instrumentation B 3.3.7.2 BASES (continued)
APPLICABILITY The mechanical vacuum pump trip is required to be OPERABLE in MODES I and 2, when any mechanical vacuum pump is in service (i.e., taking a suction on the main condenser) and any main steam line not isolated, to mitigate the consequences of a postulated CRDA. In this condition fission products released during a CRDA could be discharged directly to the environment. Therefore, the mechanical trip is necessary to assure conformance with the radiological evaluation of the CRDA. In MODE 3, 4 or 5 the consequences of a control rod drop are insignificant, and are expected to result in any fuel damage or fission not product releases. When the mechanical vacuum pump is not in service or the main steam lines are isolated, fission product "releases via this pathway would not occur.
ACTIONS A Note has been provided to modify the ACTIONS related to Mechanical Vacuum Pump Trip Instrumentation channels.
Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable Mechanical Vacuum Pump Trip Instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable Mechanical Vacuum Pump Trip Instrumentation channel.
A.I and A.2 With one or more channels inoperable, but with mechanical vacuum pump trip capability maintained (refer to Required Action B.1 Bases), the Mechanical Vacuum Pump Trip Instrumentation is capable of performing the intended function. However, the reliability and redundancy of the Mechanical Vacuum Pump Trip Instrumentation is reduced, such that a single failure in one of the remaining channels could (continued)
Quad Cities 1 and 2 B 3.3.7.2-3 Revision 0 NRC COPY #1
3 _§ R41 1,Kf Po'int:i.6tO The purpose of the Pre-Fire Plans is to provide A. direction to the crew for initiating fire actions from the control room.
B. the fire brigade leader with guidance for fighting a fire in a specific area of the plant.
C. the Shift Manager guidance concerning personnel accountability during a fire (assembly).
D. identify actions to the Off-Site Fire Department to egress into the protected area.
Answer: B uestion 39betails Question Type: Multiple Choice Topic: Question #39 (RO/SRO)
System ID: 6824 User ID: SR-4101-KOl
-- Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
OP-AA-201-008, R. 1 User Text: G 2.4.25 User Number 1: 2.90 User Number 2: 3.40 Comment: LWQ.00451 (81923) Bank question. Lower. Per OP AA-210-;JQ6e*ection 1.2.
16l1WO-L-OPERATIONS F\Jq1P o3 10/15/02
OP-AA-20 1-008 Exe ,n. Revision 1
.. . Nuclear Page 1 of 3 Level 3- Information Use PRE-FIRE PLANS
- 1. PURPOSE 1.1. The purpose of this procedure is to describe the provision and control of Pre-Fire Plans.
1.2. Pre-Fire Plans are established for use as a "guide" to aid and assist station fire fighting personnel during a fire event. The plans are designed to provide as much useful information as possible in a short amount of time, yet still have the flexibility of a workable and easily accessible reference document. Fire pre-plans do not delineate "how to" fight a fire in any specific plant area, rather they provide useful information for quickly determining emergency response strategies based on hazards and equipment in the area.
- 2. TERMS AND DEFINITIONS - None
- 3. RESPONSIBILITIES
",, 3.1. The Fire Marshal is responsible for ensuring the Pre-Fire Plans are current, including the performance of periodic reviews and updates, as necessary.
3.2. Enqineerinq is responsible for submitting proposed Pre-Fire Plan changes that result from plant design changes.
- 4. MAIN BODY 4.1. Limitations 4.1.1. A Pre-Fire Plan shall be established for all safety related areas, areas representing a hazard to safety-related equipment, and insured buildings.
4.1.2. A station predefine (or equivalent formal tracking mechanism) shall be established to ensure Pre-Fire Plans are reviewed (and updated as necessary) annually.
4.1.3. As a minimum, controlled copies of the Pre-Fire Plans should be kept in the Main Control Room and in each Fire Brigade Equipment Cage (or equivalent location).
4.1.4. If the station specific Pre-Fire Plans are controlled in accordance with other site procedures or processes (e.g., procedure control process, etc.), then Sections 4.1.2,
.. - . 4.2 and 4.3 of this procedure are not applicable.
NRC COPY #1
AD':NrE PIo"ints: 1.06 Which readily available hand held fire extinguisher should be your first choice to extinquish a small electrical fire on the 902-5 panel in the control room?
A. Carbon Dioxide B. Pressurized water C. AFFF Foam D. Dry Chemical Answer: A
,gusti 740- l~fa~ils Question Type: Multiple Choice Topic: Question #40 (RO/SRO)
System ID: 9742 User ID: NGET Status: Active Must Appear: No Difficulty: 2.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
ERG fire brigade LP fbpl0 User Text: G. 2.4.26 User Number 1: 2.90 User Number 2: 3.30 Comment: Bank question. Lower. Foam and water extinquishers are not available in the control room. Use of dry chemical is not recommended due to the amount of residue left. C02 is the best choice.
Emergency Response Fire Brigade Training lesson plan fbpl0 rev 5 OPERATIONS NM C~8 3 py Pa"EfD3
~
10/15/02
Content/Skills <*0 Activities/Notes Section Heading I. C.ass.D Material Burning (Class B) Preferred Extinguishing Second Choice Agent All Oils - Lube, Hydraulic, Dry Chemical preferred CO 2 may be used for Transformer, Diesel Fuel for small fires larger small fires in enclosed or Oils, #2 Fuel Oils, Grease, fires may require water close proximity areas.
Battery Casing (Acrylic fog stream for cooling Plastic) effects, along with dry chemical.
Foam may be used for smaller fires, however, is preferred for larger open spill type files. Water god may be desired prior to the application of foam for cooling, but must not allow the dilution of the applied form.
- 3. ClassC..
7..
.y Material Burning (Class C) Preferred Extinguishing Second Choice Agent Energized electrical CO 2 Dry Chemical equipment (motors, MCC's, etc.) Note: De-energized electrical equipment will become Class A or B material.
Energized cabling (cable Water fogstream pans) Dry Chemical /
/
/
/
/
I _____________________________________________________________________________
1: Idocs lerVirebrigl'bplO. doc Page 11 of 13 NRC COPY #1
Points: 1.0 ,0' A Control Room annunciator on the 902-5 panel has annlJnc~i~for wAindo.l, "This annunciatorzW4t(,( * ..
- C A. ,taerrassociated equiprpunt deficiency.
B. / QGA entry C. has been taken OS.
D. window is operable with problem inputs bypassed.
Answer: A
-Qustin 4 Deails, Question Type: Multiple Choice Topic: Question #41 (RO/SRO)
System ID: 9767 User ID: SR-PGTM-K3 Status: Active Must Appear: No Difficulty: 2.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
OP-AA-108-105 RO SI I 11 I I I I J Use Text: G.2.4.31 UserNq' ber 1: 3.30
.erer 2: 3.40 New question. Lower. Answer is correct utilizing referenced procedure. QGA annunicators have yellow border. Disabled or otherwise inoperable annuciators are recorded in round sheets.
OP-AA-108-105 rev 0 It.
/ ft W) "* u"T - -
V4) 4Wl't0 Lk.
cCOJ I
(' )2f r1v OPERATIONS C Pac" of 130 c7UEw1 10/15/02
QCAN 901(2)-3 A-14 UNIT 1(2)
REVISION 5 Continuous Use DESCRIPTION TORUS HIGH/LOW LEVEL SETPOINT Actual: Water Level in Torus + 1.25" or - 0.5" Tech Spec: > 14'1" and < 14'-5".
I TORUS SENSOR LS 1(2)-1602-6. HIGH/LOW LEVEL A. AUTOMATIC ACTIONS None.
B. OPERATOR ACTIONS
- 1. Observe level on LI 1(2)-1602-3. TORUS LVL (Narrow Range Indictor) to verify alarm.
- 2. Maintain Torus Level within operating range per QCOP 1600-12.
- 3. IF Torus level is > 14'5" or < 14'1", THEN refer to applicable QGA's.
4.
Observe level as indicated on PR/LR 1(2)-1602-7, TORUS PRESS AND LVL Recorder.
- 5. IF necessary, dispatch operator (with required S-Key) to East wall of RB Basement to determine Torus level as indicated by local Torus Level Sightglass:
- a. Unlock and open 1(2)-1601-100, TORUS LG-1(2)-1602-10 UPPER SV.
- b. Unlock and open 1(2)-1601-101, TORUS LG-1(2)-1602-10 LOWER SV.
- c. Determine Torus level as indicated by Sightglass.
- d. Close and lock i(2)-1601-100, TORUS LG-1(2)-1602-10 UPPER SV.
- e. Close and lock 1(2)-1601-101, TORUS LG-1(2)-1602-10 LOWER SV.
- 6. IF Torus Level is high, THEN:
- a. Check for any plant evolution that may have caused an increase in Torus Level:
(1) Any ECCS System operation.
(2) Minimum Flow Valves for HPCI/RCIC NOT fully closed.
(3) Condensate Transfer lined up for ECCS Keep Fill.
1
"OP-AA-108-105 Revision 0 Page 3 of 6
- 4.2.5. Attachment 1, "Deficiency Evaluation Checklist", may be used as an aid when evaluating deficiencies.
4.3. Follow-up actions may include:
4.3.1. Evaluate the problem with regard to its effect on equipment and system OPERABILITY.
4.3.2. Take action to comply with applicable Technical Specification Actions or ATRITRM/ODCM Compensatory Measures. The timeliness of complying with the Actions/Compensatory Measures shall be in accordance with the stated Completion Times.
4.3.3. Perform an "unavailability review" if the deficient equipment or system becomes unavailable for service.
4.3.4. Determine reportability requirements in accordance with the Exelon Reportability Reference Manual.
4.3.5. Evaluate the deficiency's impact on ability to implement emergency operating procedures.
4.3.6. Assess other deficient conditions impacting related or redundant equipment.
4.3.7. Determine if equipment/system can still be used with special restrictions.
4.3.8. Identify if any compensatory actions or additional monitoring required.
4.3.9. Request additional evaluation by Engineering or support personnel as required.
4.3.10. Evaluate deficiency for generic implications (i.e. is it highly probable that a similar deficiency could exist or occur on other systems or equipment) and, if generic implications are found, is testing of systems or equipment required?
4.4. Provide appropriate status controls for the deficiency, that will:
4.4.1. Identify the condition.
4.4.2. Initiate compensatory actions, if required.
4.4.3. Provide special restrictions or instructions for continued operation.
4.5. Identify the equipment deficiency within the main control room, if applicable, as a MCR deficiency.
4.5.1. Deficiency information should be posted using equipment deficiency tags or stickers.
They shall be placed in such a manner so as to preclude interference with operation of.equ ipment NRC COPY #1
12ID: S-9900WK26 Points. 1.0' Unit 2 has experienced a total loss of annunciators due to a loss of the normal power supply.
The operators should align reserve power supply from:
A. 250 VDC B bus.
B. the essential service bus.
C. 125 VDC B bus.
D. the instrument bus.
Answer: C "Question '42 Details Question Type: Multiple Choice Topic: Question #42 (RO/SRO)
System ID: 9745 User ID: SR-9900-K26 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOA 0900-01, r. 8, pg. 2 User Text: G 2.4.32 User Number 1: 3.30 User Number 2: 3.50 Comment: New question. Lower. 125VDC from division 1 (A bus) is the power source for the annunciators with division 2 (B bus) as the reserve supply. Others are not a potential source of power.
QCOA 0900-01 rev 8 OPERATIONS ]RCPaC* f 130 10115/02
QCOA 0900-01 UNIT 1(2)
REVISION 8
!Continuous Use LOSS OF ANNUNCIATORS
- [ ~ ~~OPERATIONS MANAGER '-"q APPROVAL SIGNATURE{ TITLE EFFECTIVE DATE A. SYMPTOMS A.1. ALARM POTENTIAL FAILURE or ANNUNCIATOR DC POWER FAILURE alarms on one or more panels.
A.2. Failure of Annunciator Test.
A.3. Loss of Annunciator Horn.
SA.4 .... Loss of Sequence Of Events Recorder Monitor.
B, AUTOMATICACTIONS None.
C. IMMEDIATE OPERATOR ACTIONS None.
D. SUBSEQUENT OPERATOR ACTIONS D.1. IF there is a loss of annunciators on 901(2)-3, 901(2)-5, or 901(2)-8 Panels, THEN notify the Shift Manager to consider classifying the event as a possible GSEP Condition and initiate GSEP as necessary.
NRC COPY #1
ID,4-:,ii S R"02 ! i ,,P o in ts :11 .0 0 Unit 2 is operating at 100% power on the 95% Flow Control Line when a trip of the 2B Recirc Pump occurs.
RPV water level will:
A. decrease to the low level scram setpoint.
B. increase to the RFP high level trip setpoint.
C. increase first and return to normal.
D. decrease first and return to normal.
Answer: C QUestioni'3 Details Question Type: Multiple Choice Topic: Question #43 (RO/SRO)
System ID: 9757 User ID: SR-0202-K22 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOP 0202-21, R. 3, pg. 4 User Text: 295001AK2.03 irn User Number 1: 3.60 User Number 2: 3.70 Comment: New question. Lower. Reactor water level increases on a recirc pump trip and then returns to normal.
- _qAMP uyt,-
OPERATIONS P "ofPaC130 10/15102 IR
QCOP 0202-21 UNIT 1 REVISION 3
F. PROCEDURE
F.1. Verify 1-262-25A/B, PMP A/B SPEED CONTROLLER for both Pumps in INDIVIDUAL MAN mode per QCOP 0202-03.
NOTE Power changes are to be performed in concurrence with QCGP 3-1.
F.2. ©IF time permits, THEN insert control rods in-sequence to be less than or equal to 70% FCL.© (H.7.a)
F.3. Concurrently perform the following:
- a. Adjust 1-262-25A/B, PMP A/B SPEED CONTROLLER potentiometer for Pump to be shut down so that Pump speed is less than or equal to 32%6' speed.
- b. Verify jet pump loop flow mismatch with both recirculation loops in operation is:
(1) < 10% of rated core flow when operating at < 70% of rated core flow; (2) <_ 5% of rated core flow when operating at > 70% of rated core flow. (H.l.aJ . . .
~~---------
NOTE Recirc Pump trip will result in a large level (swell). transient level to It less may be necessary than 30 inches. to lower Reactor vessel water
/
~------S--
F.4. Trip desired Recirc MG Set by closing
__.......... MO
- -202-SA/B, PMP DISCH VLV, for desired Recirc MG Set.
NRC COPY #1
44 ID: EýR' 0'0-K26 11Points:
Unit 2 is operating at 100% power.
Condenser backpressure is 3".
Main Condenser Flow Reversal is in progress from the Control Room.
The NSO notes that Condenser Backpressure is 4.5" and rising .25 inches every five seconds..
All valves are stroking normally.
The NSO should:
A. dispatch an operator to complete the flow reversal manually.
B. have the operator stationed at the Local Panel (2252-71) take Local Control and complete the flow reversal.
C. stop the reversing operation and return the valves to their original position.
D. have the operator stationed at MCC 27-2 attempt to reset the breaker and thermals for any valve that tripped to complete the flow reversal.
Answer: C ues on etall "W"J"Po"alm -
Question Type: Multiple Choice Question #44 (RO/SRO)
System ID: 2252 User ID: SR-4400-K26 Status: Active Must Appear: No Difficulty: 3.25 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOP 4400-09, R. 11 User Text: 295002AA1.07 User Number 1: 3.10 User Number 2: 2.90 Comment: ILT.05128 (77331) Bank question. Lower. Per limitations and actions step E.1.
OPERATIONS 10/15/02 V Pa"o30 1071
. 1[ QCOP 4400-09 UNIT 1(2)
S.. . .... . . .. . . . .. . . . .. . .. .. . . . .. .. . . ... . . .. . . . .. . . . . . . . . ... . . .. . . .. R E V I S I ON Ii In-order to keep a 2" margin to the Condenser Low Vacuum SCRAM set point (which has a lower limit 23.3" Hg in the normal band for Unit 1 and 22.2" Hg in the normal band for Unit 2), IF condenser backpressure is above 3.5" Hg prior to reversing condenser flow, data from recent flow reversals should be reviewed to determine the increase in backpressure expected during the upcoming flow reversal. Items to consider are beginning backpressure, beginning Circ Water Flow direction and increase in backpressure during previous reversals for the same direction. Based on the review of this data, IF it is expected that condenser backpressure will exceed 4.7" Hg during the flow reversal, reduce generator load to achieve the required beginning backpressure necessary so the maximum backpressure during the flow reversal is NOT expected to exceed 4.7" Hg.
A flow reversal should NOT be attempted with beginning backpressure of greater than 4" Hg.
This step does NOT apply in cases where flow reversals are being performed during a transient condition in an attempt to recover cndenser vacuum.
D. PRECAUTIONS D.I. Monitor condenser backpressure closely LF vacuum is established during the reversing operation.
D.2. When vacuum is established in the condenser, do NOT leave the circulating water flow selector switch in the OFF position because this closes all SJAE suction valves.
D.3. Consideration should be given to reducing load to reduce backpressure to < 3" Hg for the flow reversal.
D.4. Local operation is NOT normal and serious damage can result if an improper valve lineup is performed through local control.
E. LIMITATIONS AND ACTIONS E. 1 IF vacuum starts to d ecrease rapi dly, THEN:
- a. Stop the reversi ng operation AND return the valves K i- n f~- Inv = IPV f
- a. g. "a"s L n /1
~ load if' necessary.
E.2. IF the CONDENSER PIT HIGH LEVEL annunciator alarms,
...... THEN trip the Circulating Water Pumps.
NRC CQPY #1
1ý l:§k-"000K26 Polnts: 1.00 Unit One was operating at full power with all systems in their normal lineup when both feed breakers to 480 vac MCC's 18-2 and 19-2 simultaneously trip.
What is the operational impact of failsafe design associated with this loss of AC power?
A. The alternate feed breakers automatically close to restore power to essential loads.
B. A half scram and half Groups II and III Isolations occur due to lost loads.
C. A full reactor scram and full Groups II and III Isolations occur due to lost loads.
D. The alternate feed breakers automatically close maintaining all power and loads.
Answer: C 1 e :s.ti,on'45'b tails Question Type: Multiple Choice Topic: Question #45 (RO/SRO)
System ID: 9784 User ID: SR-0500-K26 .
~Status:
Must Appear: No Difficulty: 2.75 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QOA 7000-01, R. 26, pg. 1 User Text: 295003AK1.05 User Number 1: 2.60 User Number 2: 2.70 Comment: New. Higher. Ref QOA 7000-01. Full scram and Gp 2 and 3 isolations due to loss of RPS. No auto transfer of power for RPS.
OPERATIONS P f101 0/1 5/02
~RC~a~b~ooE
QOA 7000-01 Revision 26 Continuous Use 1=20 VAC RE ACTOR PROTECTION BUS FAILURE (ONE OR BOTH BUSES)
A. SYMPTOMS
- 1. Alarms.
- a. 901(2)-5 A-10/A-15, CHANNEL A/B MANUAL SCRAM.
- b. 901(2)-5 D-10/D-15, CHANNEL A/B REACTOR SCRAM.
- c. 901(2)-5 B-7, GROUP I ISOL CH. TRIP.
- d. 901(2)-5 A-8, GROUP II ISOL CH. TRIP.
- e. 901(2)-5 B-8, GROUP III ISOL CH. TRIP.
- f. Alarms associated with RPS trips 901(2)-5.
- g. Alarms associated with SJAE suction valve closure 902-7 (Unit 2 only).
- h. 901(2)-5 A-14, CHANNEL A/B DISCH VOL HIGH LEVEL.
B. AUTOMATIC ACTIONS
- 1. 'Loss of BOTH RPS buses:
- a. Full Reactor scram.
- b. Group II isolation.
- c. Group III isolation.
- d. Reactor Building Vent, Drywell and Torus Purge Fans trip.
- e. Reactor Building Vent isolation.
- f. Control Room Vent isolation.
- g. Standby Gas Treatment System auto-start.
- h. Condenser Mechanical Vacuum Pump trips.
- i. SJAE suction valves close (Unit 2 only).
- k. Channel A Reactor Building Vent and Fuel Pool
_Floor Rad Monitors fail downscale.
- 11. Reactor Building Floor Drain Sumps and Equipment Drain Tank pump trip off.
NRC COPY #1
Content/Skills Activities/Notes
--IV. INTERCONNECTIONS WITH OTHER SYSTEMS
- -'-A. Supporting Systems SR-0500-K19
- SR-.0500-K(23
- 1. Electrical Distribution
- a. 480 VAC power from MCC 18-2(19-2) is supplied to the RPS MG Sets which supply normal power to the RPS buses.
- 1) Loss of power to the MG Sets will result in a loss of that RPS Bus which gives you in a half scram. The flywheel inertia will maintain power temporarily if the power loss is of a short duration. As the MG Sets go away, the EPAs will provide protection to RPS Bus and its associated loads as the voltage and frequency go out of tolerance.
- b. 480 VAC power from MCC 15-2(25-2) is also supplied through a reserve power transformer to the RPS buses (one at a time) if an MG Set is inoperable for some reason.
c*. Power to the backup scram valve solenoids is 125 vdc Turbine Building Bus 1A1, circuit 8, Essential Division I
`ý_OM9~ Turbine Building Bus 1BL1, circuit 12, Essential Division I1for 0302-19B.
- 1) If DC power is lost to one Division, that Backup valve will not open upon receipt of a scram signal, but the other one should energize to bleed down the scram air header.
- 2) If DC power is lost to both backup valves they will not reposition upon receipt of a scram signal and thus will not provide backup to the scram pilot solenoid valves.
- d. ARI Valves 25A, 181A, and 182A valve solenoids receive power from the Division 1125 VDC @ 4160 switchgear Bus 11 (21). The 25B, 181B and 182B valve solenoids receive power from the Division 11125 VDC
@ 4160 switchgear Bus 12 (22).
- 2. If power is lost to one division of 125 vdc, that division of SR-0500-K19 ARI vales will not open upon receipt of the ATWS/ARI
'ai signals, but the other division should respond as necessary to bleed down the scram air header.
q:\trnopslp\lf-0500.doc Page 37 of 91 NRC COPY #1
Why is the Emergency Seal Oil Pump required to be tripped within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of a Unit One blackout?
A. The battery sizing calculations assumed that specific loads are shed from the bus during the analyzed four-hour period.
B. There would be no need for the Hydrogen Seal Oil pump since the generator would be no longer rotating after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
C. To extend the battery capability beyond the analyzed four-hour design period.
D. To ensure that Unit One RCIC remains available for the four-hour design period.
Answer: A u~estion 4Details Question Type: Multiple Choice Topic: Question #46 (RO/SRO)
System ID: 9783 User ID: SRN-6900-K01 Status: Active Must Appear: No Difficulty: 3.75' Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOA 6900-05,R. 7, pg 2,3 User Text: 295004AK3.01 User Number 1: 2.60 User Number 2: 3.10 Comment: New. Lower. Battery sizing calculations assume specific loads are shed during the analyzed 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period.
OPERATIONS 10/15/02 Pa of 1Pa30of130
QCOA 6900-05 UNIT 1(2)
REVISION 7
"(1?..D._1. (cont'd
- b. Unit 2 - MCC 29-2 compartment B4, #2 Battery Charger.
C. Unit 1/2 - MCC 18-2 compartment D3, or MCC 28-:2 compartment D4, 4i/2 Battery Charger.
D.2. IF AC power is available, THEN re-energize affected charger pper QOP 69007O1'.
D. 3. IF AC power is NOT available OR affected charger can N"OT be re-energized, THEN place the spare 250V DC Battery Chaergr in operation per QOP 6900-01.
D.4. I a 250V battery charger is NOT available to supply "the load, THENpei"6ii 'tLhe following*steps within the indicat*d time' period':
- a. Wit.,7{i33 c0 einut'er,pe** .circ MG Set Emergency ne**
Ltb~&O. PI ups 1(2)A and12Bb placing local contirol switches to STOP.
- b. Emergency Hydrogen Seal Oil Pump:
S(1-) IF AC power is available to MCC 18-2 (28-2), I (2-) THE PMP.
(a) Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, trip EMERG H2 SEAL OIL (b) Notify Radiation Protection to monitor H2 concentration on the Main Turbine Floor.
(c) Purge Hydrogen from Main Generator in accordance with QCOP 5320-03.
- c. IF at any time AC power becomes available, THEN verify that a charger is energized and feeding the battery and DC loads.
b-5W.- Perform a review of QCAP 1500-02 for possible Safe Shutdown system inoperability.
NRC COPY #1
QCOA 6900-05 UNIT 1 (2) 7 REVISION E.I. Each 250V DC battery is sized to start and carry normal DC loads required for safe shutdown on the one and operations required to limit the consequences unit design-basis-event on the other unit for a period of a four hours without p-ower from the chargers. of This period is deemed adequate to safeguard the plant time n :--"" -er a~re r~estogrr until no-sures ~fpoer are restor~~~_T e aaery sizing calculation assumes that the 250V DC addressed in this procedure are shed from theloads Cduri*ng the four hour evolution, at the specified bus time..
E.2. 250 VDC loads may fail to operate if battery terminal voltage drops below 210 VDC.
F. ATTACHMEN ..
None.
G. R EFERENCES.
G. 1. Technical Specifications:
- a. t8 f 8.41., D Sources - Operating.
G.2. P&IDs:
None.
G.3. Drawings:
G.4. Manuals:
None.
G.5. Procedures:
- a. QCAP 1500-02, Administrative Requirements for Inoperable Safe Shutdown Equipment.
J b. QOP 6900-01, 250 VDC Electrical System.
- c. QCOP 5320-03, Main Generator Hydrogen Removal.
NRC COPY #1
\-I -
47 R3ýdkI ~rQPoints: 1,0 Unit 2 is operating at 100% power when a reactor scram occurs.
Instrument Air is the extraction steam non-return check valves in order to prevent A. applied to; condenser overpressurization B. vented off; condenser overpressurization C. vented off; turbine overspeeding D. applied to; turbine overspeeding Answer: C 7 f etaits Question Type: Multiple Choice Topic: Question #47 (RO/SRO)
System ID: 9802 User ID: SR-3500-K15 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
LN-3500 R.2, pg. 18 User Text: 295005AK3.05 User Number 1: 2.50 User Number 2: 2.60 Comment: Modified question. Lower. IA is vented off on a turbine trip to prevent steam from the FW heaters causing the turbine to overspeed.
z2m OPERATIONS r-K p P a 3 o 0 L 10/15/02
"~ "Page
~~' 18 of 102 SExtraction Steam Non-Return Check Valves.(Pig
..-.. .ur 'sn d 3500-10) there is one non-return check valve on each extraction steam line from theLP 6 ie section to turin the 13, C, an dDfeedwater heatrs. Tese che~ck valves protect the turbine.* .om ov.iiedrscig`. 1iiis might occur w.hen .t. t e is pp.d
... and the~sub~sequent lowern of pressure in the turbine and heaters (due to the vacuum in teciisr)results in' fTiU ig- of 'the moisture in the heaters to steam and the passage of tifili steam back thi~ug the extraction steam lines into the turbine, thr *uh ifiteblý g, and on into the condenser.
Even though this may only be a relatively small amount of steam, the turbine can generator d6e to' 'the trip.
The extraction non-return check valves are swing check valves placed in the 6 i steam lines between th be andth feedwater heaters. Each valve is equipped with an air operated sp.in.. lo' csinej0g cyhinder. The closing cylinder .
is mounted on, the side of the valves. Each V fe is' also equiped with a ud lev'er andp weighit whic~h is ujii~t to give a sligh~t closing tendency (bias) to the
/T i e.igiink h cylinder i"61amd i 'the "valve open position by insmentair SIn the valve o. position a piston is pushed. up compressing the closing spring.
Anopen limiwitch-is a6`&ictuiated, which ilurmnates the red open indication located next to the test 'joysicks.~I thils position te valve disk is free to swing open or closed due to extracion steam flow, independent of the closing cylinder.
The piston inside the closming cyin'der seats against a neoprene washer in the valve open position preventing air leakage. On a loss of 'air the spring inthe:closing cylinder is released pulling the check valve disc min the closed- direction. Whnthe turbine trips the extraction non-retun check valves close on the loss of instrument air created by the operation of the turbine extraction dump valve. The reverse flow in the extraction lines caused by the flashing "of the heate drains will also close the extraction non-return check valves.
There are local position indications and test "joysticks' for each of the non-return check valves. 'he joysticks'an'dposition indication are located on two separate racks. Rack 225X-I6Ecninsth jysticksad i tion frte "B" and "C" heater extraction non-return check valves. This rack is located down on the CRD level (Turbine Building, 5726'" tlev.). Rack 22BX-A4B contains the joysticks and indication for the "D" heater extraction non-return check valves. The rack is located on the 595' level across from the APP loo-ms.
TRNOPSLIXLN-3500.R02 LEFT HAND SIDE NRC COPY #1
IP; 75539 ~~ oints:
P 1.00 What component in the Feedwater Heating System prevents the Main Turbine from overspeeding when the turbine is tripped?
A. Low Pressure Feedwater Heater Level Control Valves B. Extraction Steam Stop Valves C. Low Pressure Feedwater Heater Flash Tank D. Extraction Steam Non-Return Check Valves Answer: D
,.ustjon 2 Deails i Question Type: Multiple Choice Topic: ILT.01080: NO TOPIC System ID: 467 User ID: 75539 Status: Active Must Appear: No Difficulty: 0.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
User Text: 245000SG.7 User Number 1: 3.50 User Number 2: 3.60 Comment:
OPERATIONS r\ -I C2of5 _l P~ ~
10/15/02
48 ID: SR-0500-K20 Ploints. 1.00 Unit 2 is in RUN.
The scram discharge volume DISCH VOL HI WTR BYP keylock switch is in BYPASS.
Both scram discharge volumes have increased to 50 gallons.
The blue SCRAM lights on the full core display will be and the Scram Solenoid Group lights will be A. energized; energized B. de-energized; energized C. energized; de-energized D. de-energized; de-energized Answer: C Question 48 Details Question Type: Multiple Choice
--- Topic: Question #48 (RO/SRO)
System ID: 1978 User ID: SR-0500-K20 Status: Active Must Appear: No Difficulty: 2.50
,Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOP 0300-28,QCOA 0500-02 User Text: 295006AA2.05 User Number 1: 4.60 User Number 2: 4.60 Comment: ILT.04238 (77057) Bank question. When the scram valves are opened, the blue scram lights are energized. The Scram Solenoid Group lights are deenergized when RPS is tripped.
OPERATIONS OPERATIONS Nfl R'\\C_ý ý(Dpy IN*
pParofl ED Pa" 3
O of 130 1011 5/02 10/15/02
QCAN 901(2)-5 B-1 UNIT 1(2)
REVISION 2 Continuous Use DESCRIPTION SCRAM DISCHARGE VOLUME HIGH LEVEL BYPASS SETPOINT SDV HIGH LVL SCRAM Actual: DISCH VOL HI WTR BYP Switch in BYPASS BYPASSED Position.
Tech Spec: None.
SENSOR Relay 1(2)-590-111A/B/C/D.
A. AUTOMATIC ACTIONS
- 1. IF Reactor Mode Switch is in SHUTDOWN OR REFUEL, THEN a Control Rod block will occur.
B. OPERATOR ACTIONS
.......... I "NOTE DISCH VOL HI WTR BYP Switch is only active when RX MODE SELECT Switch is in SHUTDOWN OR REFUEL.
- 1. Verify DISCH VOL HI WTR BYP Switch is in BYPASS Position.
- 2. IF DISCH VOL HI WTR BYP Switch is NOT required to be in BYPASS Position, THEN:
- a. Verify alarm on Panel 901(2)-5 Window A-14, CHANNEL A/B DISCH VOLUME HIGH LEVEL is NOT up.
- b. Return DISCH VOL HI'WTR BYP Switch to NORMAL.
- 3. IF alarm is in with DISCH VOL HI WTR BYP Switch in NORMAL, THEN check for faulty circuit OR relay.
1
Content/Skills Activities/Notes
- b. L.a. Scrams (continued)
I Name/Purpose Device/Setpoint/ Bypass/Reset Response I
Logic -
APRM INOP Output from the Bypassed with by Output from APRM APRM High-High joystick. Trip Unit opens Trip Unit. contact in its Setpoints: APRM assigned Trip mode switch out Channel which of OPERATE, deenergizes one of Module the 107 relays which unplugged, < 50% opens a contact in its of assigned associated RPS Trip LPRM inputs in Logic.
OPERATE.
.Logic is I-out-of 2 taken twice
-1I + -
/ SDV HIGH Output from one Bypassed when When a hig I LEVEL or thermal level Mode Switch in condition or failed FAILED switch, one D/P "SHUTDOWN" or instrument exists, its INSTRUMEN level switch and REFUEL with SDV associated contact in T one thermal High Level Bypass the assigned Trip switch feed each Keylock Switch in Channel opens Trip Channel leg. "BYPASS" and both deenergizing one of Setpoint: 40 gal. RPS Buses energized. four 590-100 relays Logic is 1-out-of which opens a 2 taken twice. contact in its associated RPS Trip Logic.
L *1-
'K1FA.L.k 1IU K Output from one No Bypass When a low level LOW WATER of four Level condition exists a LEVEL Transmitters. contact opens in its
[ I(2)-263-57A/B assigned Trip
& 58A/B] Channel Setpoint: +8 deenergizing one of inches Logic is 1 four 590-105 relays out-of-2 taken which opens a twice. contact in its associated RPS Trip Logic.
REACTOR I Output from one t 4 No Bypass When a high SR-0500-K07 HIGH of four Pressure pressure condition PRESSURE Transmitters.
SR-0500-K08 exists in the RPV, a
[1(2)-263-55A-D] contact opens in its L-0500-K08 Setpoint: 1060 assigned Trip SR-0500-K13 psig Channel Logic is 1-out-of deenergizing one of Show Figure 0500-12.
2 taken twice. four 590-104 relays which opens a contact in its associated RPS Trip Logic.
q:\tmopslp\lf-0500.doc Page 25 of 91
QCOS 0500-02 UNIT 1(2)
REVISION 12
"-F. LIMITATION4S AND ACTro}
F.1. At the completion of the surveillance the US must immediately review the results of this test for compliance to Technical Specifications requirements.
F.2. JE a Control Rod ina'dvertently scrams, THEN refer to QCOA 0300-04 and QCOA 0300-11.
F.3. JE the logic fails to operate properly, THENnotify Unit Supervisor.
F.4. E two og more control rods start to drift, AM all RPS scram solenoid lights lit, THE-Nscram the Reactor.
F.5. IE four H more control rods start to drift, THEN scram the Reactor.'
F.6. iE partial testing is required, THEN the Unit Supervisor will document in the PREREOUISITES, E.....
the steps to be performed DR the steps to be disregarded, AM any special instructions required for the performance of the partial test.
G.. REfFORM~fANCE AQCCPTANCE CR!1E RIA G1.1. WHENj RX SCRAM CH{ A £~RX SCRAM CII B manual scram pushbutton is depressed, THEN the following actions oc-c
unciator 901 2 -- A-15, CHANNEL A or B MANUAL SCRAM alarms.
H. PROCEDURiE H.1. Verify proper operation of RPS Channel A manual scram instrumentation:
- a. Depress RX SCRAM CH A manual scram pushbutton.
2 NRC COPY #1
QCOP 0300-28 UNIT 1(2)
REVISION 19 W. . .2. l_.-rep -tive steps are in progress that insert individual control rods (e.g., manual insertion using RMC, individual rod scramming, etc.) AND the method is NOT successful, THEN that method may be stopped following 2 or 3 attempts on rods associated with each CRD bank.
E.3. Documentation of jumper or fuse manipulation may be performed any time after the appropriate step is completed and will typically be done when plant conditions are stable enough to allow sufficient time.
NOTE The next three steps of this procedure are to be implemented
.. .cncurrently. Completion of the next three steps is NOT required*prior to implementing subsequent procedure steps.
- '-The next three steps are:
De-energizing scram solenoids.
Manual control rod insertion.
F.2. IF ALL Scram Valves are NOT oen, as indicated by the blue scram lights NOT being lit on full core display, THEN remove the following fuses to de-energize the scram solenoids:
- a. At Panel 901(2)-15, Terminal Board "C":
(1) 590-715A, F5.
(2) 590-715C, F6.
(3) 590-715E, F7.
(4) 590-715G, F8.
.- "-ON NRC COPY #1
IDt: SR-02O3-K28 Points:
Unit 2 was operating at 100% power when an inadvertent Group 1 occured. / A Relief valves are cycling on their auto setpoints. < .
Reactor pressure is 1116 psi .tg I .,,
The "B" relief valve closed yve secovdi agzo. .,
The "B" relief valve is expected to automatically open22 A. r reessure reach ("
B. immediately C. intfve
,****."C¢
%seconds, D. .in ine seconds.-*
- Answer: D Q u-si'on 9'e tails .....
Question Type: Multiple Choice Topic: Question #49 (RO/SRO)
-System ID: 9786 S*User ID: SR-0203-K28 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0
-,'Point Value: 1.00 Cross
Reference:
LIC-0203, pg. 4 User Text: 295007AA1.04
... User Number 1: 3.90
ý7÷--User Number 2: 4.10 Comment: New question. Higher. The ADS valves have a timer that prevents auto reopening for 14 seconds.
I - -
OPERATIONS PPRTIN of0/15/02f 10/15/02
Content/Skills Activities/Notes Activities/Notes
.3) 110 second timer timed out, and I
CS-r" HRpump -running> -I 0,0-psi-g I
-OR
- 1) Low-Low RWL (-59"), and
- 3) 8.5 minutes timer timed out.
- 4. Electromatic Relief Valve Control (Figure 0203-1)
Show Figure 0203-1
- a. Each relief valve has a control switch with 3 positions:
- S/R-0203-EK007a Auto: An ADS signal or exceeding high pressure **S/R-0203-EK021 setpoint will actuate the valve.
Q: What signal will cause the Off: The valve will actuate on an ADS signal only. ERV's to open with their contr switches in OFF?
Man: This position opens the relief by directly A: ADS signal only.
energizing the valve solenoid.
- b. Whe e 3-3B and 203-3C open autonitieally due to S/R-0203-EK013 eD.
DS signal or high pressure, a 14.5 seconds delay is SJR-0203-EK028
" , / activated which prevents the valves from automatically **S/R-0203-EK020 reopening for 14.5 seconds after closing.
/ This time delay allows the tailpipe vacuum breakers to Q: The 3C ERV is cycled open cycle, ensuring water hasn't been drawn up into the line then closed and the green and as a vacuum is being formed when the steam in the line ambe~r lights come on. What condenses. The subsequent reopening of a relief valve dbes the amber light indicate?
when the tailpipe is partially filled with water could A* 10 second inhibit timer over-pressurize the relief line and/or result in structural acliuated.
damage to the suppression pool when this slug of water is blown into the suppression pool.
/I During these 14.5 seconds, manual actuation is O/What is the purpose of this physically possible; however, procedure cautions direct interlock?
the operator not to actuate the valves for 14.5 seconds A: Prevent possible damage to after closing. A light labeled INHIBIT is illuminated during these 14.5 seconds to warn the operator of the .the tail pipe/suppression pool b]
warning operator not to open 14.5 second limitation.
RV for 14.5 seconds.
The inhibit time delay was changed to 14.5 seconds, from 10 seconds, to account for valve stroke time from Q: Can the relief valve be full-open to full closed. opened manually?
-- -A0iI M A: Yes.
REF. ISC 96-OOIE yAi I\iNurLrLILO2Uj fl.vrn ,at\nC. V ,-. .-.-. - - doc QA*I RNOU",LP\LIU..-02.03 doe NRC COPY #1 Page 4 of 37
SR-0W22-KO9, ID~: pois 1.00 Which of the following statements correctly describes the operation of the Reactor Recirculation MG sets with RPV level at -59" and RPV pressure 800 PSIG?
A. LPCI loop select logic causes the drive motor breakers to trip and the ARI system causes the field breakers to trip after a 9-second time delay.
B. PCIS logic causes both drive motor breakers to trip and the ARI system trips the field breakers immediately.
C. The ARI system causes the drive motor breakers to trip and the field breakers to trip after a 9-second time delay.
D. The ARI system causes the field breakers to trip and the drive motor breakers do NOT trip.
Answier: A uestion 50 Detils Question Type: Multiple Choice
, " ... ,Topic: Question #50 (RO/SRO)
System ID: 9736 "UserID: SR-0202-K09 Status: Active Must Appear: No Difficulty: 3.50 Time to Complete: 4 Point Value: 1.00 Cross
Reference:
QCAN 901-4 H-3 & A-9 User Text: 295037 EK2.03 User Number 1: 4.10 User Number 2: 4.20 Comment: NEW 124902 Lower. LPCI loop select logic causes the drive motor breakers to trip and the ARI system causes the field breakers to trip after a 9-second time delay.
OPERATIONS F\ 1l a"o 3 /7 R7 10/15/02 Fr-D)
QCAN 901(2)-4 A-9 UNIT 1(2)
REVISION 2 Continuous Use DESCRIPTION RECIRCULATION MG SET DRIVE MOTOR AUTOMATIC TRIP SETPOINT RECIRC SETS Actual: 1. < -59 inches Reactor Vessel level. DRIVE MOTOR
- 2. Recirc Pump suction valve < 90% open. AUTO TRIP/
- 3. Recirc Pump discharge valve < 90% open.
- 4. 4kV Bus undervoltage.
- 5. Fluid drive oil temperature > 210 0 F.
6 ,--:..u.be T-presu-re < 25-psig- r 5 seconds.
(7. LPCn Loop Select logic trip.
Emerg Pushbutton.
"St**JQet Tech Spec: None.
SENSOR 1. 52a contact at SG 11(21), Compt 4 for Pump 1(2)-202-51A.
- 2. ..... 52b contact at SG 12(22), Compt 11 for Pump 1(2)-202-51B.
A. -AUTOMATIC ACTIONS
- 1. Recirculation Pump A/B drive motor breaker opens.
B. OPERATOR ACTIONS
- 1. Verify MG Set A/B tripped.
------------------------------------------ I -
NOTE The maximum current before overloading a Reactor Recirculation Pump motor is 724 amps.
The'maximum speed during single loop operation to prevent riser brace failure is 78%.
- 2. Reduce speedon operating Recirculation Pump to less than 78% and maintain Pump motor current less than 724 amps as indicated at 1(2)-202-730A/B, PMP CUR.
NRC C6PY #1
QCAN 901-4 H-3 UNIT 1 REVISION 0 SAFETY-RELATED STATION MANAGER APIFROVAL SIGNATURE TITLE EFFECTIVE DATE DESCRIPTION RECIRCULATION MG SET A FIELD BREAKER OPEN f---*
SETPOINT Actual: 1. Generator lockout relay '
tripped. A RECIRC MG SET
- 2. Generator exciter field FIELD BREAKER overvoltage. OPEN
- 3. ATWS- Reactor pressure Ad 1250 ig-
- ,-4-.--ATWS- Reactor low-low level * -only
"-59 inches with 9 second time delay.
Tech Spec: None.
SENSOR 1-202-60-IIIA A. AUTOMATIC ACTIONS "1.. Recrculation MG Set A trips:
- a. Generator drive motor tripped and locked out.
- b. Generator field breaker tripped and locked out.
B. OPERATOR ACTIONS
- 1. Verify MG Set A tripped.
- 2. IF unit is operating in EGC, THEN:
- a. Trip EGC.
- b. Transfer Recirculation flow control from MASTER AUTO to INDIVIDUAL MANUAL.
. ------- - IIL .. ....---
L L*L ..............
NOTE The maximum current before overloading a Reactor Recirculation Pump motor is 724 amps.
-The maximum speed during single loop operation to prevent riser brace
-,---falure is 78%.
- Ju-_______-_-_-_-_'_'.'___'_-_______-*_____*
NRC CODY #1
ID: 6O-370124~ Po ints: 1.00 S n t e is still inerted.
A reactor startup is in progress nce wth QCGP 1-1, Normal Unit Startup.,V While placing the first FRV in serv ce, e REACTOR VESSEL HIGH LEVEL annunciator ALARMS he NSO takes action to reduce vessel level to normal by increasing RWCU system blowdown fro 101 GPto 200 PM.
A. RWCU system demins will isolate on high post strainer DP.
SReactor pressure will decrease and a Groupysolation will be received.
C. RWCU system demins will isolate on high post strainer temperature.,
D. Drywell temperature would increase, p causing the QGAs to be initially entered on High Drywell Pressure.
Answer: D
"=ues ion51bii's Question Type: Multiple Choice Topic: Question #51 (RO/SRO)
System ID: 9775 User ID: SR-3700-K24 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCGP 1-1, R43 pg. 32 D.7 User Text: 295010AK1.03 User Number 1: 3.20 User Number 2: 3.40 Comment: Bank question. Higher. Due to the reject flow, you have less cooling flow returning to the RHX, which in turn places a greater heat load on the NRHX and RBCCW, which in turn causes overheating the Drywell coolers.
3 OPERATIONS r . ~ C Y~1/50
QCGP 1-1 UNIT 1 (2)
REVISION 43 D.6. Reactor operation at or below the point of adding heat is inherently different than when the Reactor is operated above the point of adding heat; fuel temperature reactivity effects are negligible below the point of adding heat and small reactivity changes can result in large changes in power before any temperature
-- f-Afedback- occuras.
D. Excessive reject of Reactor water can overload the RBCCW System via the RWCU Non-Regenerative Heat Exchangers and cause an increase in Drywell temperature resulting in high Drywell pressure Scram/Group 2 Isolation and/or RWCU Isolation due to high Non-Regenerative Heat Exchanger outlet temperature 0F
"ýf-1--40_
D.8. Reactor water temperature should be maintained < 284 0 F until Reactor water dissolved oxygen is < 300 ppb.
Reactor water dissolved oxygen needs to be minimized in order to reduce the rate of IGSCC and reduce the probability of initiating new cracks.
D.9. IF Reactor water temperature is > 284 0 F AND Reactor water dissolved oxygen is > 300 ppb, THEN Reactor
- heat-up should remain on hold until dissolved oxygen drops below 300 ppb.
D.10. During normal heat-up, do NOT exceed Reactor coolant heat-up rate of 100OF/hr when averaged over a one hour period. (H.l.ac)
D.11. The following Pressure/Temperature Limit surveillances shall be observed:
- a. During system heatup, the reactor coolant system temperature and pressure shall be determined to be within the required heatup limits and to the right of the limit lines of Figure 3.4.9-3 at least once per 30 minutes and refer to QCOS 0201-02. (H.l.ac)
D.12. Opening the Turbine Vacuum Breaker at high rpm (i.e.,
greater than 1200 rpm) imposes excessive loads on the Turbine LP Rotor last stage buckets.
D.13. IF the 1/2 250 VDC Battery Charger is feeding the Unit battery, THEN when a large load is placed on the battery, nuclear instrumentation may spike due to induced currents in the 24/48 VDC System. (H.8.e)
NRC CG2@Y #1
ID: SR-0001-K22 Pqints: 1.00 A LOCA on Unit 2 has caused high Drywell pressure.
Drywell temperature is required to be monitore ri tsrayingthe Drywell iorder to verify:
A.
B.
Drywell parameters are within the DSIL curve.
. ký"ý 90 B.
C.
rp*ratu'c DrwUQ*
, bc',ov 2G0 degr-ee..-.
~W % e.
po
ýVj (A CL'2-,,
D. Drywell parameters are within the PSP curve. IL'j "0 Answer: A Q u9,§i 6on2 ~ iDt Is Question Type: Multiple Choice Topic: Question #52 (RO/SRO)
System ID: 9772 User ID: SR-0001-K22
--- Status: Active
'MustAppear: No Difficulty: 4.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QGA Details, pg. 26-31 User Text: 29501 OAK3.05 User Number 1: 3.50 User Number 2: 3.40 Comment: New question. Recall. Required to check DSIL prior to initaiting sprays. PSP looks at Torus pressure and temperature. 180 and 260 degrees F by themselves do not prevent spray initiation.
QGA 200 rev 8 0~(LAPI 10115/02 OPERATIONS Pa of 130 of 130 10/15102 ED R--ý\c QU-) nrll PY
'J0
'4' Content/Skills %F Wo Ar-fivrifiný-/K[ #
_ION IM ITIIIIII I "11I. FIG.. .KDRYWELL
.. SPRAY I~l I ITIAT 11 FIG K, DRYWELL SPRAY INITIATION LIMIT.
A. Definition and general description SR-0001-K09
- 1. The DWSIL was previously defined as the highest drywell S-0001-Kll temperature at which initiation of drywell sprays will not result S-0001-K12 in an evaporative cooling pressure drop to below either:
"* The drywell-below-wetwell differential pressure capability, or
"* The high drywell pressure scram setpoint.
- 2. No credit was taken for vacuum breaker operation. It was conservatively assumed that the worst-case evaporative cooling pressure drop could occur more rapidly than pressure could be equalized through the vacuum breakers.
- 3. New analyses indicate that the evaporative cooling transient will occur much more slowly than assumed, even Vacuum breakers are now assumed to function effectively.
if the drywell atmosphere is initially dry.
- 4. The curve defines a maximum drywell temperature as a function of pressure.
- 5. The DWSIL is used to avoid containment failure or deinertion following initiation of drywell sprays.
- 6. Evaporative cooling
- a. Occurs when water is sprayed into a dry or superheated atmosphere.
- b. The spray droplets absorb heat and flash to steam until the surrounding atmosphere saturates.
- c. Results in a large, rapid pressure reduction. The rate can be faster than the capacity of the vacuum breakers.
- d. Unrestricted spray initiation could result in a negative drywell-to-torus differential pressure large enough to damage the primary containment.
- e. Evaporative cooling continues until the atmosphere is saturated. Higher initial temperatures and pressures result in greater pressure drops since the atmosphere can absorb a greater mass of water vapor before saturating.
Q:\TRNOPSLP\L-QGADET.doc Page 26 of 82 NRC COPY #1
QGA Step
/U(U.
I 500C 400 E
(D 0-.
0 0
/
100- _00
/
20&
0 10 20 30 40 50 60 Drywell Pressure (psig)
Figure 3: Drywell Spray InitiationLimit
-' MI . .. *"1 ' ' 't I1 .. [I U
Page 27 of 82
- AGE Error! Bookinark not NRC COPY #1
Content/Skills Activities/Notes B. Locations I. The DWSIL is used in the following QGAs:
"* QGA 200, Primary Containment Control
"* QGA 200-5, Hydrogen Control C. Derivation (Figure 3)
- 1. The DWSIL is determined by heat balance between the injected spray water and the drywell atmosphere.
- 2. Line G)defines the drywell temperature and pressure from which evaporative cooling will reduce drywell pressure to the scram setpoint.
- a. Limiting the final pressure to a positive value permits time to take manual action to terminate sprays before pressure drops below atmospheric.
- b. Maintaining a positive pressure avoids opening the vacuum breakers and deinerting the containment.
- c. The scram setpoint is used since it is a predefined, easily recognized, relatively low value.
- d. The allowable temperature increases with pressure since a higher initial pressure permits a greater pressure drop.
- e. Initiating sprays to the left of the line could result in a final pressure below atmospheric, causing the vacuum breakers to open.
- 3. Line 0 is no longer used. New analyses indicate that the evaporative cooling transient will occur much more slowly than assumed, even if the drywell atmosphere is initially dry.
- a. The rate is limited by physical processes:
- 1) Evaporative cooling will slow as the local atmosphere surrounding each spray droplet saturates.
- 2) The water normally stored in the spray headers is preheated by the drywell atmosphere. The first water discharged when sprays are initiated is therefore expected to be warmer than previously assumed.
- 3) Full spray flow is not reached instantaneously "whi-en sprays are initiated.,
Q:\TRNOPSLP\L-QGADET.doc Page 28 of 82 NRC COPY #1
QGA Step LKJ Drywell Spray Initiation Limit C
E (D
Drywell Pressure (psig) 41 Page 29 of 82
- AGE Error! Bookmark not NRC COPY #1
Content/Skills
- Contet/S illsActivities/Notes
- b. Analyses indicate that the torus-to-drywell vacuum breakers will be effective in limiting the differential pressure across the drywell-torus boundary.
- c. The right side of the DWSIL has therefore been eliminated. The new limit corresponds to the left side of the old limit.
- 4. Line ( is no longer used.
&19
/7 I
AjP'&'
D. Assumptions------- -------_f* ,
- 1. CDywell spray water temperature is 32'
- 2. Drywell humidity is 0% when drywell sprays are initiated.
- 3. The torus and drywell are at the same pressure when drywell sprays are initiated.
- 4. Vacuum breakers deal with the evaporative cooling transient.
Q:\TRNOPSLP\L-QGADET.doc Page 30 of 82 NRC COPY #1
QGA Step - ý ýr III Pressure Suppression Pressure 0) 0 0) 10 0) 0 0
I-
"8 :' 1: * ' 1' 1' 1' ' ' 18 1 19 2 Wide Range Torus Water Level (ft) 18.5 N
Page 31 of 82
- AGE Error! Bookmark not NRC COPY #1
e s tarting additional DrywellCoolers to prevent jeopardizing 2 on tb A. Recirc Pump Seal N h
/
B. Reactor Vessel Head C. Primary Containment D. RPV Level Instrument Answer: C Question 3 etais . 3' Question Type: Multiple Choice Topic: Question #53 (RO/SRO)
System ID: 9774 User ID: SR-0001-K20 Status: Active Must Appear: No Difficulty: 2.50 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QGA 200 LP, pg. 5, 37, 39 User Text: 295012AK3.01 User Number 1: 3.50 User Number 2: 3.60 Comment: Bank question. Lower. QGA 200 is concerned with Primary Containment integrity. The distractors are concerned with reactor integrity.
'S. .. . .. .. .'--;--7 KPa"or10CAo OPERATIONS 10/15/02
Content/Skills Activities/Nofa-z II. ENTRY CONDITIONS A. General SR-0001-K20 SR-0001-K21
- 1. The entry conditions are symptomatic of emergencies or events that may degrade into emergencies.
- 2. The entry conditions are related to the key parameters controlled by QGA 200.
- 3. QGA 200 must be entered whenever any entry condition occurs. If the flowchart is already in use, it must be reentered if another entry condition occurs or an entry condition clears and later reoccurs.
- 4. The entry conditions setpoints were chosen to be:
- a. Operationally significant.
- b. Unambiguous.
- c. Easy to detect.
- d. Familiar to operators.
B. Drywell pressure above 2.5 psig
- 1. Corresponds to the high drywell pressure scram setpoint.
- 2. High drywell pressure is a symptom of a break in the drywell.
- 3. Also requires entry of QGA 100.
C. Drywell temperature above 180F
- 1. Corresponds to the maximum normal operating temperature.
- 2. High drywell temperature is a symptom of events that may jeopardize primary containment integrity and equipment I operability:
Loss of coolant
- Loss of drywell cooling
- ADS valve actuation
\./. .
Q:\TRNOPSLP\QGA 200a.doc Page 5 of 71 NRC COPY #1
Content/Skills Ate Activities/Notes IV. DRYWELL TEMPEATURE A. General SR-0001-K22 SR-0001-K23
- 1. The initial step uses the normal method of drywell S-0001-K24 temperature control--drywell cooling.
- 2. If temperature cannot be held below 1807F, later steps use drywell sprays.
- 3. If temperature cannot be restored and held below the drywell design temperature, an RPV blowdown is performed.
- 4. Significant action levels include:
- a. 1807F (maximum normal operating temperature)
- b. 280'F (drywell design temperature)
B. Drywell cooling
- 1. Hold drywell temperature below 180'F
- a. The initial control method is the same as that used during normal planioertion-operate drywell cooling.
- b. Provides a transition from normal operating procedures.
- c. 180'F is the maximum normal operating temperature the highest temperature expected to occur during normal plant operation.
- d. No further action is required as long as temperature can be held below 180'F.
- 2. If temperature cannot be held below 180'F, go to #18.
- a. Provides more detailed guidance on contingency actions (sprays, scram, blowdown).
- b. The "cannot hold" decision can be made before temperature actually reaches 180'F. (Refer to the definition of "cannot hold.")
a;L Q:\TRNOPSLP\QGA 200a.doc Page 37 of 71 NRC COPY #1
Content/Skills Wýý
-I Contnt/SillsActivities/Notes 7 3. Start all available drywell cooling.
-a Provides exp cit direction to maximize drywell cooling.
- b. Ensures that drywell cooling has been used in preference /
"I%. to less desirable actions.
- 4. Detail A
- a. High drywell temperatures may affect RPV water level indications.
- b. Detail A identifies conditions under which RPV water level instruments may be unreliable or must be considered invalid.
- c. If drywell temperature is above the RPV Saturation Temperature, water in the instrument runs may start to boil, resulting in unreliable level indication.
- d. If the criteria in Table C are not satisfied for an instrument, the actual RPV water level may be below the instrument variable leg tap. Under these conditions, the
......... *instrument
_ will not, respond to changes in actual level and cannot be used.
- e. Inaccuracies due to out-of-calibration conditions are not addressed. The intent of the detail is to define conditions under which neither the displayed value nor the indicated trend of an instrument can be relied upon.
- f. The derivation of Detail A is discussed in the Calculationslesson plan.
- g. References to Detail A are also included in flowchart branches controlling RPV water level.
Q:\TRNOPSLP\QGA 200a.doc Page 39 of 71 NRC COPY #1
Given the following plant conditions:
- RPV level 10 inches
- Drywell pressure 3 psig
- RPV pressure 1050 psig
- Drywell temperature 170 OF
- Reactor power 2%
WHICH ONE of the following correctly states the QGA procedures that initially should be entered based on the above information ONLY?
A. QGA 101 and QGA 200-5.
B. QGA 100 and QGA 200.
C. QGA 101 and QGA 200.
D. QGA 100 and QGA 200-5.
Answer: B Question Type: Multiple Choice Topic: Question #54 (RO/SRO)
System ID: 9751 User ID: SR-0001-K21 Status: Active Must Appear: No Difficulty: 2.75 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QGA 100, 200 User Text: 295013G2.4.4 User Number 1: 4.00 User Number 2: 4.30 Comment: Bank question. Lower. Drywell pressure of 3 psig requires entry into QGA 100 and 200. Reactor power is
< 3%, so no QGA 101 entry. No indications of H2, so no requirement to enter QGA 200-5.
OPERATIONS 10/15/02 Cl-aCPy 01
QGA Steps
,4ý Q:\TRNOPSLP\QGA 100.doc Page 4 of 76 NR~C COPY #1
QGA Step Q:\TRNOPSLP\QGA 200a.doc Page 4 of 71 NRC COPY #1
Unit 2 had an ADS valve leaking for several days that is still operable.
A plant cooldown is in progress on Unit 2, Reactor pressure is currently 700 psig.
The RHR system was JUST started in the Torus Cooling Mode and the ANSO reports Torus temperature rapidly rising.
The rapid rise in Torus temperature is due to If indicated Torus temperature exceeds __ _ degrees F, a Reactor scram is required.
ADS-valve leakage impirihj1-dFridtly on the temperature-senso#j 105 B. initial stratification of water in the Torus; 105 valve leakage impingigrdirectly on the temperature senso // 10 D. initial stratification of water in the Torus; 110 Answer: D I
uesio In1515D.ea Question Type: Multiple Choice Topic: Question #55 (ROISRO)
System ID: 9768 User ID: SR-I 000-KX&
Status: Active Must Appear: No Difficulty: 3.50 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
OE-4679 User Text: 295013AK1.01 User Number 1: 2.50 User Number 2: 2.60 Comment: New question. Higher. With a leaking Relief valve and no Torus Cooling flow, you can get high Torus temps when the flow is initiated.
OPERATIONS 10/15/02 NnCPCy CF"Dfl3 nP N
From: "rRoot, Clarence M.(TVA) cRootCM@inponn.org>
<ki
Subject:
OE4679 - SUPPRESSION POOL TEMPERATURE EXCEEDED TECH SPEC LIMIT DUE TO STRATIF ICATION - BROWNS FERRY NUCLEAR PLANT UNIT 2 Date: Tuesday, July 02, 1991 2:34 PM
SUBJECT:
BROWNS FERRY UNIT 2 RESTART EXPERIENCE - NOTIFICATION OF UNUSUAL EVENT (NOUE) - SUPPRESSION POOL TEMPERATURE EXCEEDED TECHNICAL SPECIFICATION LIMIT DUE TO STRATIFICATION ON JUNE 29, 1991, WITH UNIT 2 IN HOT STANDBY AT LESS THAN ONE PERCENT REACTOR POWER AND THE REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM IN SERVICE FOR SEVERAL HOURS TO CONTROL REACTOR PRESSURE (OUTBOARD MAIN STEAM ISOLATION VALVES CLOSED), HEAT ADDED TO THE SUPPRESSION POOL BY THE RCIC TURBINE EXHAUST CAUSED A STRATIFIED LAYER OF WARMER WATER AT THE POOL SURFACE. SUPPRESSION POOL BULK WATER TEMPERATURE INDICATION WAS WITHIN LIMITS AND STABLE UNTIL A RESIDUAL HEAT REMOVAL (RHR) PUMP WAS STARTED AND THE RHR SYSTEM PLACED IN THE SUPPRESSION POOL COOLING MODE FOR PERFORMANCE OF THE QUARTERLY RHR SYSTEM RATED FLOW TEST. AFTER THE RHR PUMP WAS STARTED, MIXING OF THE SUPPRESSION POOL WATER RESULTED IN A RAPID INCREASE OF INDICATED TEMPERATURE, AND ALL AVAILABLE SUPPRESSION POOL COOLING WAS PLACED IN SERVICE IN ACCORDANCE WITH EMERGENCY OPERATING INSTRUCTIONS. WITHIN A FEW MINUTES, SUPPRESSION POOL WATER TEMPERATURE EXCEEDED 110 DEGREES FARENHEIT. THE REACTOR WAS MANUALLY SCRAMMED AND A NOUE DECLARED.
THE NOUE WAS TERMINATED WHEN SUPPRESSION POOL WATER TEMPERATURE DECREASED TO 103 DEGREES.
INVESTIGATION OF THE EVENT IS IN PROGRESS. THE ROOT CAUSE APPEARS TO BE INADEQUATE PROCEDURAL CONTROLS TO COMPENSATE FOR THE DESIGNED LOCATION OF THE SUPPRESSION POOL TEMPERATURE ELEMENTS. A STICKING SUPPRESSION POOL ATMOSPHERE TEMPERATURE RECORDER PEN MAY HAVE HINDERED TIMELY IDENTIFICATION OF INCREASING POOL WATER TEMPERATURE.
ACTIONS TAKEN THE DAY OF THE EVENT INCLUDED A WALKDOWN OF THE SUPPRESSION POOL AREA TO ENSURE DAMAGE HAD NOT OCCURRED AND ISSUANCE OF AN OPERATIONS STANDING ORDER REQUIRING INITIATION OF SUPPRESSION POOL COOLING WHENEVER SIGNIFICANT HEAT IS ADDED TO THE SUPPRESSION POOL. (A SURVEY OF OTHER UTILITIES DID NOT INDICATE THAT THIS WAS A UNIVERSAL PRACTICE.)
THE FINAL EVENT REPORT IS EXPECTED TO IDENTIFY ADDITIONAL ACTIONS, SUCH AS REVISING PROCEDURES TO ADDRESS SUPPRESSION POOL COOLING REQUIREMENTS AND A METHOD TO ENSURE CONTROL ROOM RECORDERS ARE FUNCTIONING PROPERLY.
OE4679 - SUPPRESSION POOL TEMPERATURE EXCEEDED TECH SPEC LIMIT DUE TO STRATIFICATION BROWNS FERRY NUCLEAR PLANT UNIT 2 Information
Contact:
RAY SWAFFORD, OPERATIONS, (205) 729-3338
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NRC CO PY #1
'56 .1.ID: SkO800-K2 Points:
ont:I1.0b b
An ATWS has occured. Reactor power is 3% and steady. Reactor pressure is 920 psig and being controlled by turbine bypass valves. Reactor water level has been lowered to -145 inches lAW QGA 101, RPV Control (ATWS).
Which one of the following describes the status of core cooling and safety limits?
Adequate core cooling _(1)_ assured and _(2)_ safety limit has been violated.
A. (1) IS NOT (2) A B. (1)IS (2) A C. (1) IS NOT (2) NO D. (1)IS (2) NO Answer: B duestion betails Question Type: Multiple Choice Topic: Question #56 (RO/SRO)
System ID: 9780 User ID: SR-0800-K28 Status: Active Must Appear: No Difficulty: 3.75 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
L-QGA101 LP, pg. 35 User Text: 295015G2.2.22 User Number 1: 3.40 User Number 2: 4.10 Comment: New. Higher. Adequate coore cooling is assured by maintaining RPV level above the miniumum steam cooling level (-166 inches). Level above the top of active fuel (-142 inches) safety limit is violated.
Ref TS 2.1 OPERATIONS
[m((cPby 113 10/15/02
Content/Skills Activities/Notes MW'NOM 3. If the conditions in the second row of the override exist, lo__"6-0wer level-ltoreduce reactor power.
- a. The combination of conditions indicates that heat is being added to the torus faster than it can be removed.
- 1) Power above 3% (the APRM downscale setpoint) indicates that significant power is being generated.
a) At lower power levels, lowering RPV water level would be of little benefit since decay heat would still be produced.
b) If power is unknown, it must be assumed to be above 3%.
- 2) An ADS valve open or drywell pressure above 2.5 psig (the scram setpoint) indicates that heat is being added to the torus.
a) Heat can be added through the ADS valve discharges or the drywell vents.
b) The high drywell pressure scram setpoint is specified since it is a relatively low pressure, is well-known, and is easily recognized.
- 3) A torus temperature above 11 0°F indicates that torus heatup is occurring.
a) The action level is the temperature at which Tech Specs requires a scram.
b) The Tech Spec LCO is specified because:
"* It is relatively low but significantly above normal.
"* A scram is not required until temperature reaches this value.
a) Level must be held above -166 in. to prov adequate core cooling.
b) Further reduction would be of little benefit.
Q:\trnopslp\L-QGA101.doc Page 35 of 124 NRC COPY #1
SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:
THERMAL POWER shall be _<25% RTP.
2.1.1.2 With the reactor steam dome pressure _Ž785 psig and core flow _>10% rated core flow:
For Unit 1, with Cycle 17A exposure less than or equal to 4,000 MWD/MT, for two recirculation loop operation, MCPR shall be _Ž1.11, or for single recirculation loop operation, MCPR shall be _>1.12. For Unit 1, with Cycle 17A exposure greater than 4,000 MWD/MTU, for two recirculation loop operation MCPR shall be > 1.15, or for single recirculation loop operation MCPR shall be > 1.16.
For Unit 2, MCPR shall be 2!1.11 for two recirulation
..loop operation, or for single recirculation loop
- MCPR'shall be _>1.12.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. )
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be
- 1345 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.
Quad Cities 1 and 2 2.0-1 Amendment No. 207/202 NRC COPY #1
/ 11I~-302R~1 C ints 1.00 Both Units are operating at full power when Unit 1 experiences a scram from full power. Plant conditions on Unit 1 are as follows:
-Half of the control rods are still at positions greater than 04.
-Reactor power indicates approximately 8%.
-Reactor water level is between +8" and +48" and stable.
-RPV pressure is less than 1040# and is being controlled with bypass valves.
-The rods DO NOT move inward when scrammed with reactor pressure.
-The running CRD pump TRIPS and CANNOT be restarted.
-The other CRD pump also TRIPS when it is started and WILL NOT restart.
What is the next action taken to insert control rods?
A. Open CRD crosstie and use opposite unit pump to insert control rods.
B. Shut the 1 301-25 CRD Charging Header Isolation and drive Control Rods.
C. Open SDV vents to relieve the hydraulic lock.
D. Locally vent the overpiston area of each control rod that IS NOT inserted.
Answer: A
_66 slon'5'DtiW Question Type: Multiple Choice Topic: Question #57 (RO/SRO)
System ID: 7816 User ID: SR-0302-K1 5 Status: Active Must Appear: No Difficulty: 3.00 Time to Complete: 2 Point Value: 1.00 Cross
Reference:
QCOA 0300-1, QCOP 0300-19 User Text: 295015AK2.01 User Number 1: 3.80 User Number 2: 3.90 Comment: LORT 124429 Bank question. Higher. With both CRD pumps tripped, you must crosstie to get CRD flow. It does no good to close the 0303-25 valve with no CRD pumps running. There is no direction to open SDV vents or vent CRD overpistion areas.
777777ý_
WtN OPERATIONS 10/15/02 HR PaC bof13 l
_ýl QCOA 0300-01 UNIT 1(2)
REVISION 12 CAUTION The cross-tie line may be used if sufficient time is available to perform the in-plant lineups. Loss of both CRD Pumps on a Unit will result in possible accumulator trouble lights within a few minutes. Use Step D.2 for scram criteria.
- 6. IF neither Unit CRD Pump is available,' THEN perform QCOP 0300-19 or QCOP 0300-33 as applicable.
I
- 7. IF standby CRD Pump was started in Step D.1, THEN:
- b. Dispatch an operator to verify proper operation of the running pump.
C. Close 1(2)-301-254A/B, l(2)A/B CRD PUMP MIN FLOW ISOLATION VLV on the tripped pump.
E. DISCUSSION
- 1. For Unit 2 only, with NO CRD Pump in operation, Reactor Recirc Pump Seal Injection flow is NOT available.
- 2. Technical Specification 3.1.5 requires that the reactor mode switch by placed in the shutdown position if charging water header pressure can NOT be restored to
Ž 940 psig within 20 minutes. An accumulator is inoperable if the local pressure is < 940 psig. There is NO method directly available to determine accumulator operability in the Control Room; conservative action is directly based on the presence of accumulator trouble alarms. (F-10)
A cross-tie line exists between the discharges of th5e CRD pumps of the two Units. This cross-tie may be utilized in situations where ample time exists to valve in a pump from the other Unit. Discretion should be used based on the possible imminent failure of a CRD Pump with the standby CRD Pump inoperable. In these situations, both Unit CRD Pumps may be inoperable provided the same CRD Pump is NOT supplying both Units.
Loss of both CRD Pumps on a Unit will result in numerous accumulator trouble lights within a few minutes.
NRC COPY #1
QCOP 0300-19 UNIT 1 REVISION 6 Continuous Use CRD PUMP CROSS-TIE OPERATION USING Ul CRD PUMPS A. PURPOSE The purpose of this procedure is to provide the necessary steps to use the CRD Cross-Tie Line between Units using Ul CRD pumps.
B. DISCUSSION B.1. This Procedure contains four sections. The first section is for operating with both CRD Suction Filters.
The next two sections are for Cross-Tie Operations which contain sub-sections for the Unit 1 CRD Pump cross-tie combinations. The last section is for returning the CRD System to normal:
- a. Section step titles are as follows:
(1) Step F.1, IF operation with both CRD Suction Filters is required, THEN:
(2) Step F.2, IF Unit 1 CRD Pump is to supply Unit 2 CRD System, THEN:
(3) Step F.3, IF Unit 1 CRD Pump is to supply both Units CRD System, THEN:
(4) Step F.4, WHEN CRD System Cross-Tie Operation is NO longer required, THEN:
C. PREREQUISITES C.I. This procedure will be executed when one of the following criteria is satisfied:
- b. Discretion based on the possible imminent failure of a CRD Pump with the standby CRD Pump inoperable.
C.2. CRD System required to provide cross-tie support is in operation with either the A OR B CRD Pump B' "W per QCOP 0300-01.
NRC C6PY #1
158 llj~~~(t:
WRP-1(O4
- -ý0-its .0 An uncontrolled fire in the Control Room necessitates evacuation of the Control Room before the safe shutdown equipment can be obtained.
Where can the operators go to acquire the necessary equipment?
To the QCARP locker in the:
A. OSC.
B. Unit 1 Turbine Building Trackway.
C. Work Execution/Communications Center.
D. Unit 2 Turbine Building Trackway.
Answer: C Question Type: Multiple Choice Topic: Question #58 (RO/SRO)
, *o . System ID: 103 "AUser- D: SRN-ARP-K04 Status: Active Must Appear: No Difficulty: 2.75 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOS 0010-03, R. 14 User Text: 295016AK2.02 User Number 1: 4.00 User Number 2: 4.10 Comment: ILT.00671 75175 Bank question. Lower. Comm Center is location of QCARP locker outside of Control Room.
OPERATIONS OPEATON "0of5130 10/15102
QCOS 0010-03 UNIT 1(2)
REVISION 14 Continuous Use SAFE SHUTDOWN EQUIPMENT INSPECTION A. PURPOSE To outline the method of routinely verifying that tools and components are available to support the Appendix R Safe Shutdown Procedures.
B. DISCUSSION None.
C. EQUIPMENT REQUIRED C.1. V-Key to open QCARP Lockers in the Control Room and Communication Center.
D. PREREQUISITES D.1. The Unit Supervisor has completed the following:
- a. Unit 1(2)
- b. Reason for test (check appropriate item):
Normal Surveillance ( )
Post-Maintenance ( )
Partial for ( )
Other (
____/____ )
start test:
to Permission c.
C. Permission to start test:/
US Signature Date/Time E. PRECAUTIONS None.
F. LIMITATIONS AND ACTIONS F.I. IF any item is found to be defective OR missing, THEN
____2_qnotify the Unit Supervisor, document in comments section and initiate action to correct the deficiency.
NRC C6PY #1
ID F,5 700-K24 Roints: 1.0' The reactor has been scrammed from full power and the Mode Switch taken to S/D in response to an instrument air header rupture that has resulted in a loss of Instrument Air on Unit 2.
Which one of the following describes how the operation of the MSIVs will be affected by this condition?
A. The inboard MSIVs would close when their accumulators discharged; the outboard MSIVs would remain open.
B. All MSIVs would remain open since the drywell pneumatic system will automatically align to supply the MSIVs.
C. All MSIVs would remain open since the MSIV Instrument Air Crosstie will automatically open.
D. The inboard MSIVs would remain open; the outboard MSIVs would close.
Answer: D Q~tqstlion 59 Details Question Type: Multiple Choice Topic: Question #59 (RO/SRO)
System ID: 966 User ID: SR-4700-K24 Status: Active Must Appear: No Difficulty: 2.75 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QOA 4700-06, R. 12, User Text: 295020AK2.12 User Number 1: 3.10 User Number 2: 3.20 Comment: ILT.01951 (76044) Bank question. Higher.
The Outboard MSIVs are supplied from IA. The Inboard MSIVs are supplied from DW pnuematics.
OPERATIONS NRFCConpy Pa of 130 10/15/02
QOA 4700-06
. -Revision 12 Instrument Air provides air pressure to air operated valves throughout the plant. When air pressure is lost, these valves may fail in one of three modes; i.e. fail open (FO),
fail closed (FC), and fail as is. Some valves have air accumulators that are designed to maintain the valve in position following a loss of air supply. However, air systems cannot be made completely leak tight, and these valves may begin drifting to their fail position shortly after a loss of supply air. The accumulators on the MSIVs provide a motive force to assist the springs in closing the MSIVs on loss of supply air. On a slow loss of instrument air, the valves may fail in a random sequence, causing )
operational difficulties.
It is important to document all actions taken to recover from this type of event to aid in restoration. Once the plant is stable a plan should be developed to restore the unit to normal status. This restoration plan should include repair and inspections of affected components. Prior to restoring instrument air, ensure positive control of all air operated components.
Following restoration of Instrument Air, include walkdowns of all air operated components to ensure they are in their expected configuration. If a scram occurred, include this walkdown in the Master Outage Checklist QCGP 1-5 to be completed prior to startup. (SOER 88-1)
NRC COPY #1
QOA 4700-06 Revision 12 A list of selected valves and fail positions is orovided blno :
VALVE FAIL MODE NOTES Scram Inlet and Outlet Valves FO SDV Vent and Drain Valves FC Low _.Fqa~wate Regulator Valve LockuP . MayupDiftOpen-.-.-.--
Outboard MSIVS FC Air Accumulator Ensures Closure '4 t --
LAIU I-Lv---------------
SJAE Suction Valves FC Air Accumulators Chimney Isolation Valve FC Air Accumulator PCI Group II Valves FC Air Accumulators Condensate Normal Reject FO Condensate Emergency Reject FO Condensate Normal Makeup FC Condensate Emergency Makeup FC RFP Minimum Flow FO Unit 1 Condensate Pump Minimum Flow FO Unit 2 Condensate Pump Minimum Flow FC Steam to SJAE and Booster SJAE FC Hood Spray Valve FC Torus to Reactor Building Vacuum Breaker FO r-Outboard Recirc Sample Valve FC Off-Gas Line Drain Valves FC SBGTS FCV FO Extraction Steam Bypass Valve FO Heater Normal Drains FC Heater Emergency Drains FO Desuperheater Valve FC Reactor Water Cleanup Reject FCV FC Gland Water LCV FC TBCCW LCV FC RBCCW LCV FC Service Water Load TCVs FO RHR Head Spray FCV FO Off-Gas Charcoal Absorber Bypass Valve FO Bypasses the Absorbers HPCI Drain Pot Drain Valves FC HPCI Drain Pot Bypass Valve FO RCIC Steam Line Drain Valve FC RWCU Demin Inlet Isolation Valve FC ACAD Valves FC SBLC 90X-5 Tank Level Ind. Fails Downscale NRC :'Y #1
ID: HTFF K8.35 Points: 1.06-Initial conditions are as follows:
I
- Unit One in mode 4.
- Reactor Water level is 30 inches.
- Shutdown Cooling is in operation.
A spurious High Drywell Pressure signal is recieved and will NOT reset.
Reactor pressure is slowly increasing.
Reactor Shell and Flange temperatures are also slowly increasing.
The correct operator action is to:
A. open safety relief valves.
B. secure Reactor Water Clean Up reject flow.
C. raise reactor water level to between 90 and 100 inches.
D. monitor running recirc pump parameters.
- --- ND Answer: C Queýstiion' '66;beaills~
Question Type: Multiple Choice Topic: Question #60 (RO/SRO)
System ID: 5269 User ID: HTFF-K8.35 Status: Active Must Appear: No Difficulty: 4.00 Time to Complete: 0 Point Value: 1.00 Cross
Reference:
QCOA 1000-02 R12 User Text: 295021AK1.04 User Number 1: 3.60 User Number 2: 3.70 Comment: New question. Higher. Answer is correct due to no recirc pumps running due to the spurious high drywell pressure signal and signs of thermal statification from pressure and metal temps increasing require enhancing natural circulation. Distractors would add to the stratification.
QCOA 1000-02 rev 12 OPERATIONS 10/15/02 a-fl-M-C Pa o130