ML100830067
| ML100830067 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 02/11/2010 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| References | |
| Download: ML100830067 (617) | |
Text
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 1. 007 EG2.4.21 001 Given the following conditions:
Unit 1 was operating at 100% power when Control Bank '0" started inserting with Rod Control in manual.
- A manual Reactor Trip is initiated from handswitches on1-M-4 ori1-M-4 and1-M-6.
Reactor Trip Breaker "A" indicates green.
Reactor Trip Breaker "B" indicates red.
Reactor power indicates 3% and decreasing.
- All CERPI indication is lost.
Pressurizer pressure is 1990 psig and slowly trending down.
- All SG Narrow Range levels are off scale low.
- The Unit Supervisor has entered E-O, "Reactor Trip or Safety Injection."
Which ONE of the following identifies the condition of the reactor and the appropriate action?
A. The Reactor is tripped. Continue in E-O; Actuate Safety Injection.
B~ The Reactor is tripped. Continue in E-O; Do B:t 00 not actuate Safety Injection.
C. The Reactor is not tripped. Transition to FR-S.1, "Response to Nuclear Power Generation/ATWS." Actuate Safety Injection.
o.
O. The Reactor is not tripped. Transition to FR-S.1, "Response to Nuclear Power-Generation/ATWS."
Power Generation/ATWS." Manually trip the turbine and start AFW pumps.
Page 1
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS:
A. Incorrect, Reactor is tripped. However Safety Injection is not required for the conditions. Plausible because. the reactor being tripped is correct and the applicant may conclude that the conditions require safety injection due to pressure dropping and all SG levels off scale low.
B. Correct. One of the Reactor Trip Breakers is opened and the flux is less thanthan 5%
and lowering, therefore, the Reactor is tripped. Safety Injection is not required.
C. Incorrect, the reactor is tripped, transition to FR-S.1 is not appropriate and Safety Injectionis Injection is not required for the conditions. Plausible because all the listed conditions of the reactor being>tripped being'tripped in E-O are not met and the applicant may conclude that the conditions require safety injection due to pressure dropping and all SG levels off scale low.
D. Incorrect, the reactor is tripped, the turbine would have automatically tripped due to the Train A reactor trip breaker opening and AFW pumps would have automatically started. Plausible because all the listed conditions of the reactor being tripped in E-O are not met and manually tripping the turbine and starting AFW are the immediate operator actions in FR-S. 1 Page 2
11/2009 Watts Bar RO NRC Exam., As submitted 10/2/2009 Question Number: 1 Tier: _1_ Group 1 KIA: 007 EG2.4.21 Reactor Trip Emergency Procedures / Plan Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
Importance Rating: 4.0/4.6 10 CFR Part 55: 41.7/43.5/45.12 10CFR55.43.b: Not applicable.
KIA Match: Applicant is required to assess the status of reactivity control safety function, and determine if the Safety Functions are challenged and the actions required from the conditions provided.
Technical
Reference:
E-O, "Reactor Trip or Safety Injection," Rev 27 FR-S.1, "Response to Nuclear Power Generation/ATWS," Rev. 19 Proposed references None to be provided:
Learning Objective: 3-0T-EOPOOOO
- 8. Analyze a given set of plant conditions and determine required procedural transitions per E-O.
Question Source:
. New Modified Bank x Bank Question History: Comanche Peak bank question Comments: Modified stem conditions, modified all distractors and correct answer, relocated correct answer.
Page 3
WBN REACTOR TRIP OR SAFETY INJECTION E-O Rev 27 IStep IIIIAction/Expected Response I IResponse Not Obtained NOTE 1 Steps 1 thru 4 are IMMEDIATE ACTION STEPS.
NOTE 2 Status Trees / SPDS should be monitored when transitioned to another instruction.
- 1. ENSURE reactor trip: Manually TRIP reactor.
Reactor trip and bypass breakers IF reactor will NOT trip, OPEN. THEN
- RPls at bottom of scale. ** GO TO FR-S.1, FR-S.1 ,
Nuclear Power Generation /
- Neutron flux DROPPING. ATWS.
- 2. ENSURE Turbine Trip: Manually TRIP turbine.
All turbine stop valves CLOSED. IF turbine will NOT trip,
( THEN THEN
- RUNBACK turbine manually OR
- CLOSE MSIVs and bypasses 3 of 28 30f28
WBN STATUS TREES FR-O Rev 13 SUBCRITICALITY FR-S TIME COLOR PROC NO GOTO POWER RANGE FR-S.1 LESS THAN 5%
5% YES ORANGE GOTO FR-S.1 INTERMEDIATE NO RANGE SUR ZERO YES ORNEGATNE OR NEGATIVE INTERMEDIATE RANGE SUR NO YELLOW 0 GO TO FR-S.2 MORE NEGATIVE SOURCE NO THAN -0.2
-O.2DPM DPM YES OK RANGE ENERGIZED YES SOURCE RANGE SUR ZERO NO YELLOW YELLOW 0:
0
'.GO TO GOTO FR-S.2 OR NEGATIVE GREEN YES ~i'=iFiJii£Si=ci3JiJi=~ OK 4 of 11
Commache Peak 2007 Examination Outline Cross- Level RO SRO reference:
Tier# 1 Group # 1 KIA # 007 EA2.06 Importance Rating 4.3 Ability to determine or interpret the following as they apply to a reactor trip: Occurrence of a reactor trip.
Proposed Question: Common 39 Given the following conditions:
- A manual Reactor Trip is initiated from CB-07 and CB-10.
- Reactor Trip Breaker "A" indicates green.
- Reactor Trip Breaker "B" indicates red.
- Reactor power indicates 3% and decreasing.
- All DRPI indication is lost.
Which ONE (1) of the following describes the condition of the reactor and the appropriate action?
(
A. The Reactor is tripped. Continue in EOP-O.OA, Reactor Trip or Safety Injection.
B. The Reactor is tripped. Actuate Safety Injection to meet Shutdown Margin criteria.
C. The Reactor is not tripped. Transition to FRS-O.1A, Response to Nuclear Power Generation/ATWT.
D. The Reactor is not tripped. Reattempt to manually trip the reactor and nottripped.
manually initiate Turbine trip.
Proposed Answer: A Explanation (Optional):
A. Correct. One Reactor Trip Breaker opened and flux is lowering, therefore, the Reactor is tripped.
B. Incorrect. Reactor Reactor is tripped, however, action not consistent with the EOP.
C. Incorrect. Action not consistent with the EOP-O.OA RNO.
D. Incorrect. Action not consistent with the EOP-O.OA RNO.
Technical Reference(s) EOP-O.OA, Step 1 && Bases (Attach if not previously
_ _ _ _ _ _ _ _ _ _ _ provided)
Proposed references to be provided to applicants during NONE examination:
Learning Objective: EOO.xG2.0B12 (As available)
Question Source: Bank#
Bank # -X-Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 1, 10 Comments:
Robinson Bank
--- -------~,--------.--------~--.--~--------~-.----~-------~----------~------------------
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 2. 009 EA 2.20002 2.20 002
\. Given the following:
Unit 1 is at 100% power.
Containment Purge Train A is in service to lower containment.
- A small break LOCA occurs.
- As the operating crew is responding to the break, a Containment Vent Isolation (CVI) occurs.
Four minutes later the crew initiates a manual Reactor Trip and Safety Injection.
Which ONE of the following identifies the position indicated for dampers, 1-FCV-30-40, LWR CNTMT PURGE EXH PRESS RLF valves, 1-FCV-30-37 and 1..,FCV-30-40, on the supply line to the Containment Vent Air Cleanup Units (CVACU) during the above period?
A'I Green lights would be lit on control switches before the LOCA developed A"!
but the dampers do receive a close signal due to the Containment Vent Isolation.
- 8. Red lights would be lit on the control switches before the LOCA developed B.
but the dampers would receive a close signal when the Containment Vent
. (' occurs..
Isolation occurs C. Red lights would be lit on the control switches before the LOCA developed and the dampers would remain open until the Phase A Containment Isolation occurs.
D. Red lights would be lit on the control switches before the LOCA developed and the dampers would remain open unless a Phase B Containment Isolation occurred.
Page 4
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS:
A.
A Correct, Green lights would be lit on control switches before the LOCA developed because the dampers are required to be closed when purge is in service and the dampers do receive a close command when a Containment Vent Isolation signal is generated.
B. Incorrect, the red lights would not be lit on the control switches before the LOCA developed because the dampers would be closed but the dampers would receive a close signal when the Containment Vent Isolation occurs.
Plausible because the applicant must recall the administrative requirement to have the dampers closed during purge operations and the dampers receiving a close signal from the CVI is correct.
C. Incorrect, the red lights would not be lit on the control switches before the LOCA developed because the dampers would be closed and the dampers would receive a close signal prior to the Safety Injection resulting in a Phase A A isolation. Plausible because other containment components remain open resulting in the red light being lit until a Phase A signal occurs.
D. Incorrect, the red lights would not be lit on the control switches before the LOCA developed because the dampers would be closed and the dampers would not remain open unless a Phase B isolation occurs. Plausible because other containment components remain open resulting in the red light being lit until a Phase B Signal signal occurs.
PageS Page 5
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 2 Tier: _11 _ Group 1 KIA: 009 EA 2.20 Small Break LOCA Ability to determine or interpret the following as they apply to a small break LOCA: Containment vent damper position indicator.
Importance Rating: 2.6/2.9 2.6 /2.9 10 CFR Part 55: 43.5 / 45.13 10CFR55.43.b: Not applicable.
KIA Match: Applicant is required to identify the position of dampers in the containment vent system during operating conditions, and why they would be expected to be in that position.
Technical
Reference:
SOI-30.02, "Containment Purge System," Rev 0054 SOI-88.01, "Containment Isolation," Rev 16 Drawing 1-47W611-30-1, Rev 12 Proposed references None to be provided:
Learning Objective: 3-0T-SYS088A 3-0T -SYS088A
- 9. Given a set of plant conditions, determine the correct response of the Containment Isolation System.
Question Source:
New x X
Modified Bank Bank Question History: New question Comments:
Page 6
WBN Containment Purge System 501-30.02 Unit 1 Rev.00S4 Page 9 of 74 of74 3.0 PRECAUTIONS AND LIMITATIONS (continued)
F. For lower compartment purge, the following sequence should be followed to prevent Upper-to-Lower Containment press imbalance from opening the Ice Condenser Doors:
- 1. . Upper compartment purge should be started first and operated for 5 to 1.*
10 minutes to permit air pressures to stabilize.
- 2. Stop upper compartment purge, close associated dampers and stop fans.
- 3. Start lower compartment purge.
G. Containment Vent Isol (CVI) or Aux Bldg Isol (ABI) should NOT be reset UNTIL all initiating signals are reset or blocked.
H. Smoke detected in U1 Aux Bldg air intake will SHUT DOWN Containment Purge Sys.
I. When outside air temp is 35°F or less, Purge Supply Preheat/Cooling Coils should be drained, or in service per SOI-44.01.
J. In Modes 1-4 only ONE set of supply and exhaust valves may be open to Containment at a time. .
00 K. In Mode 2, 3, or 4, minimum Containment temp is 60°F (62.500 F), Tech Spec 3.6.5 (and 1-SI-0-2-00 with instrument error). .
L. required Operable during Movement of Both Purge Air Cleanup Units are require.d Irradiated Fuel in Containment.
M. In Mode 6, with 1-ISV-78-600 (Wafer VI Vlv) v) Open, Containment Purge startup or shutdown can cause SFP/Rx Cavity overflow due to Containment press changes 11.
N. 1-FCV-30-54, 61, Operation of 1-FCV-30.;54, 61,62,213, 62, 213, or 216 creates a local PERSONNEL HAZARD while FCV is in motion due to the location of valve components. An Operator should be sent to ensure Safe Conditions during operation of of each FCV.
O. 1-SI-30-701, Containment Isolation Valve Local Leak Rate Test Purge Perform 1-SI.;30-701, Air, every 6 months or within 92 days after opening any Purge Air CIV.
P. raising Radiological implications (e.g.: causing an area to become airborne or raiSing the humidity in contaminated areas) should be considered any time Containment Purge is secured.
Q. 1-FCV-30-37 and 1-FCV-30-40 must be closed for all containment purge operations.
WBN Containment Purge System 501-30.02 Unit 1 Rev. 0054 Page 22 of 74 Date Date,_
__ __ __ INITIALS 5.2 Start Up Lower Containment Purge (continued)
[9] INDICATE which Train(s) will be used:
Train A Containment Purge o Train B Containment Purge o
Both Trains A and B Containment Purge (Modes 5 & 6 only)
'(see (see Precaution and Limitations 3.0S) 0o
[10] IF Wafer Vlv, 1-ISV-78-600, OPEN in Mode 6 WITH SFP & Rx Cavity Full, THEN MONITOR SFP and Rx Cavity levels during Purge startup, AND SHUT DOWN Purge Fans if Overflow is approached 1.
[11] IF aligning for 2-Train operation, THEN CHECK A Train ABGTS is NOT running.
IStart of Critical Step(s)
[12] CLOSE Containment vent filter flowpath:
PERF NOMENCLATURE LOCATION POSITION UNID INITIAL LWR CNTMT PURGE EXH 1-M-9 CLOSED 1-HS-30-37 PRESS RLF LWR CNTMT PURGE EXH 1-M-9 CLOSED 1-HS-30-40 PRESS RLF NOTE The following dampers must be closed when moving irradiated fuel with Containment open.
[13] PERFORM the following (N/A HS NOT used):
PERF NOMENCLATURE LOCATION POSITION UNID INITIAL PURGE EXH FAN A SUCT 1-M-9 OPEN 1-HS-30-61
WBN CONTAINMENT ISOLATION SYSTEM 501-88.01 1 Revision 16 Page 27 of 29 Date _ _ __ CHECKLIST 3 of 3 Page 2 of3 CVI Verification HANDSWITCH HAN DSWITCH OPER DEACT ALTERNATE NORMAL VALVE ID VALVEID PERF VERIF NOMENCLATURE ABLE &CLOSED VALVE POSITION INITIAL INITIAL CLOSED 1-M-9 1-HS-30-14, LWR CNTMT CLOSED 1-FCV-30-56 IV PURGE 1-FCV-30-14 && 56 1-HS-30-15, LWR CNTMT CLOSED 1-FCV-30-15 IV PURGE 1-FCV-30-15 && 57 CLOSED 1-FCV-30-57 IV 1-HS-30-16, LWR CNTMT 1-HS-30-16,LWR CLOSED 1-FCV-30-16 IV PURGE SUP 1-HS-30-17, LWR CNTMT CLOSED 1-FCV-30-17 1-FCV-30-17 - IV PURGE SUP 1-HS-30-19, INSTR RM CLOSED 1-FCV-30-19 IV PURGE 1-FCV-30-19 &&
58 CLOSED 1-FCV-30-58 1-FCV-30-58 IV 1-HS-30-20, INSTR RM CLOSED 1-FCV-30-20 IV PURGE 1-FCV-30-20 &&
59 CLOSED 1-FCV-30-59 IV 1-HS-30-37, LWR CNTMT CLOSED 1-FCV-30-37 1-FCV-30-37 IV PURGE EXH PRESS RLF 1-HS-30-40, LWR CNTMT CLOSED 1-FCV-30-40 IV PURGE EXH PRESS RLF 1-HS-30-54, ANNULUS CLOSED 1-FCV-30-54 IV PURGE EXH 1-HS-30-61, PURGE EXH CLOSED 1-FCV-30-61 IV FAN ASUCT 1-HS-30-62, PURGE EXH CLOSED 1-FCV-30-62 IV FAN B SUCT
L-m:- L L9MLt-L 2 3 4 5 11 12 A-AUTO
"-AUTO CONTAINMENT PHASE A CONTAINIAENT ISOLATION SIGNAL TRAIN A A 1-471611-88>-1 1-47'511-88-1 NOTES, NOTES*
FeV-30-50 VALVE FCV-3a-50 A AUX SLOG fULLY fUllY OPEN
- 1. faR FOR SYUBOLS S'!l.4BOLS SEE INSTRUNENTATION LATEST ISSUE.
INSTRUMENTATION AND ANO IDENTIFICATION IOENTIFICATION STANDARDS.
A ISOLATION C(),I;PLETE. INSTRUMENTATION AND COMPQHENT
- 2. FOR COMPLETE COMPONENT SEPARATION DESIGNATIONS 51 ART SIGNAL fROM rNCQRE VALVE FCV-lO-135 fCV-30-135 SEE CONTROL DIACRIJAS;DIAGRAMS, 1-471610-30 1-4"5tO-30 SERIES.
INSTR RM SUP?!.. Y FAN [-J CONTAINMENT VENT fULLY OPEN .3.
- 3. NOTE NorE DELETED.
- ISlN TRAIN A A-AUTO A-AUlD A-AUTO
- 4. ~DIGITAl "OIGITAL ANO AND ANALOG LOGICLOCIC SYMBOLS ARE USED ON LOGIC DIAGRAMS TO FUNCTIONALLY DESCRtBE DESCRIBE THE PROCESS CONTROL. REFER TO THE ASSOCIATED CONTAINMENT VENT WIRING SCHEMA.nC SCHEMATIC FOR fOR THE ELECTRICAL COMPONENTS USED TO IMPLEMENT START SIGNAL ISLN TRAIN B 8 (BUFFERED)
(BUFfERED) OPEN !/ HS 1/ HS CLOSE OPEN;;
OPEN !/ HS CLOSE THE CONTROL SCHEME. SCHEME.'~
fROM fAN 18, 30-7 30-8 5. DISCONNECT SWITCH ADDEO ADDED TO THE CONTROL CIRCUIT OF or FCV-30-7, -10,
-la, -14,
-14.
r------------"") -17. -20, ~40,
-17, -51. ~S2,
-40, -51, -51, 55 a.
/I:. -59 FOR APPENDIX APPENOIX R PURPOSES.
A-AUTO \ CONTROLS TYP OF I 6.* TRAIN A 6~ A AB!
ABI SIGNAL TRIPS FAN A MOTOR t.K}TOR STARTER, TRAIN B B ABI SIGNAL I fSV-30-t35 I TRIPS fAN FAN B J.lOTOR
~TOR STARTER.
~_~~~N_ B ______ J cvr IS USED AS AN INPUT SIGNAL ONLY DURING REFUELING
- 7. CVI RefUELING OUTAGES. USINC OUTAGE.S. USING ST"'RT Ii' HS ~ STOP THIS INPUT WILL MAINTAIN MAINTAIN. THE ABSCE IN THE EVENT OF A A HIGH RADIATION 30-1,10, PULL TO FSV
~~~~tbsD¥~J~~E:E~gE'rA~G ANNULUS IS OPEN TO THE A~~i~1n~N~uYr6I~(NTA::::NT SIGNAL DURING REFUELING OPERATIONS WHILE CO~ArNI.IENT AND/OR THE AUXILIARY BUILDING" = )
AND~R THE LOCK OUT 30-13-4 REFERENCE DRAWINGS:
CONTAINMENT VENT 8 ISLN 1-471611-88-1 .--.Ar,....-, 1-471"511-30 SERIES H
1-4 1-47'11511-0-\
LOG.IC DIAGRAM LOGIC DIAGRAM INDEX .t. S'!l.480LS 8
TRAIN A t~:RTUENTI--+-II---EJC=)---II-*----"'ANNULUS LOWER 1-47Ji610-30 .t. 31 SERIES 1-4 CONTROL DIAGRAM COt.lPARTI.4ENTI--+__r---C=)----II-+----.... ANNUlUS 478601-30 SERIES 4 INSTRUMENT TABULATION AUX BLDG t 1--4-7'111'866-1 I-I-
1-47W610-90-3 FLOW DIAGRAt.I RAD t.ION CONTROL DIAGRAM AIR SUPPLY 1-1-47Ji611-88-1 LOGIC DIAGRAI.I fROM OUTSIDE 1-45 t-45'11'600-30-8 THRU -12 SCHEMATIC DIAGRAM 1-45w760-30-t6
'~45 SCHEMATIC DIAGRAM
, 1-45.600-57-4.t. -7 1-45 SCHEMATIC DIAGRAM
,, I CONT AINIlAEHT -ANNULUS 6.
CONTAINMENT-ANNULUS t::. PRESSURE
,,:I ISOLATION VALVES I
TO FAN fAN A !'OR FOR FCV-30-IO/-S2 FCV-30-10/-52 rI PURGE AIR EXHAUST EXt-lAUST a. FCV-30-H/-56
.l FCV-30-14/-56 ONLY
_______________ J
_______________ J ONLY:I c c r------ --------,
I I I I I I I I I I I I I I r-"""il~~, I I I I I I I I I I I I I I FOR FSV-JO-37 .t. I IL _______________
-40 ONL Y ~
I o
D SHIELO SLOG o
D
\
CONTAINWiENT PURGE AIR SUPPLY iCONTROLS-TYP-OFi
- CPAS FAN lA :
L____ _ ____ .J ANNULUS LEAK-OFF f CCiNTROlS- TYP-OFi LL~X~L:~::-----J INCORE INSTR IN R~ EXHAUST FAN fAN E r----------I r----------, E 1I CONTROLS TYP 1I GAS I FCV-30-213 I
'-1____--\ :~-131GA~.,:"-L.~::-I~FCV ---~~~~:5T 30-54 G:~"W51l*L8~::-1~FGV---T:-u:n 1
--1 l-
~
-eEl ANNULUS RE BlDG 9~~'31B f~~~b~TBLDG EXHAUST
~f:
X
)f~
r 90-131 90-1318 30-216 EXHAUST INCORE INSTRUMENT ROI:'4 INcaRE SUPPL Y FAN ROOW FAN 18lB CONTROLLED ~
xX -SY -HS=--30-=-4A -AND ":fa
-BY-HS=30:-4A-A'No-:t.a-r----;-1------f x CON.TAINMENT CONTAINt.lENT PURGE EXHAUST FANS A A AND B 8 F F CONTAIN~ENT PURGE AI R SUP?!.. Y
_____________ ! ~~.! ~ ~~_~A~~E~ _C.9~.T~~L~ _______________ _
TYP OF fAN 1,10, l-13-08 PROJECT FACILITY POWERHOUSE G UNIT 1 TITLE ELECTRICAL TO INCORE tNCORE LOGIC DIAGRAM
- INSTRUMENT INSTRUIlAENT ROOM VENTILATION SYSTEM CONTAINMENT PRESSURE RELIEF'~
RELIEF ~
WATTS BAR NUCLEAR PLANT Q
CONTAINMENT PRESSURE RELIEf RELIEF ~"--------------'
~ ..- - - - - - - - - - - - - - - '
TENNESSEE VALLEY AUTHORITY ANNULUS DESIGN INITIAL ISSUE ENGINEERING LEAK-OFF LEAK-OfF APPROVAL DRAFTER DRAfTER CHECKER H RO ISSUE PER 1 D.P. PERRER
~~.~~~T~~ ~~~~~Yf~~~
DESIGNER REVIEWER EAI-3_
EAI-3.10 10 ..a. RIMS 22 S. LIN FSAR FIG. 9.4-29 ~:!.:....!~~I!._ ~:!::....~~E~T!R_
D.P. PERRER R.L. FORESTER T28 *. 92 0206 828 3 C.C. LYKE FOR MeB C.C lYKE ISSUED BY: DATE
-R *M. JOHNSON
-R.M. 2*27-92 2-27-92 85 1-47WSll-30-1 1-47W611-30-1 R12 PRtHTSRfQ'DR_ ~l F S~IN H 7 8 BR OR PRaJ AS CE EE tolE NE FE HE st TE BL 8F" 18 sa BII OR PRO~ 108 e<: EE WE NE FE HE SE TE 8L 8F 18 so CAD MAINTAINED DRAWING I (CONF (CONFIGURATION IGURATION CONTROL DRAWING)
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 3.011 EK1.01 003 During which ONE of the following LOCAevents does natural circulation and reflux cooling become important?
(Assume water source is available to S/Gs and S/Gs PORVs work correctly)
A. Any Large break LOCA with the break on a Cold Leg and RCS subcooling indicating O°F.
B. Only a Large break LOCA with the break on a Cold Leg and RVLlS indicating less than 100%.
C~ Any Small Break LOCA with break flow greater than the ECCS flow and RCS subcooling indicating O°F.
D. Only a Small Break LOCA in the pressurizer vapor space and RVLlS indicating less than 100%.
DISTRACTOR ANAL YSIS:
A. Incorrect, natural circulation and reflux cooling are not mentioned as cooling mechanisms in FSAR, WOG, or lesson plan for a LBLOCA.
B. Incorrect, natural circulation and reflux cooling are not mentioned as cooling mechanisms in FSAR, WOG, or lesson plan for a LBLOCA.
C. Correct, Natural circulation & reflux cooling are important when there is sufficient mass in primary to sustain density gradient and a condition when core boiling is in progress (for reflux)
D. Incorrect, Natural circulation & & reflux cooling are important when there is sufficient mass in primary to sustain density gradient and a condition when core boiling is in progress (for reflux) but it is not limited to a vapor space small break LOCA.
Page 7
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
~, Question Number: 3 Tier: _1_ Group
_1_ 1 KIA: 011 EK1.01 Large Break LOCA Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA :
Natural circulation and cooling, including reflux boiling.
Importance Rating: / 4.4 4.1 /4.4 10 CFR Part 55: 1.8/41.10/45.3 1.8 / 41.10 / 45.3 10CFR55.43.b: Not applicable.
KIA Match: The question tests the ability of the applicant to recognize the mechanism for core cooling during a large break LOCA and how natural circulation and reflux cooling can be used to provide cooling during the LOCA. The KA itself asks for an operational implication of natural circulation and reflux during a large break LOCA and the applicant must relate that these two cooling mechanisms are important during a small break LOCA only.
(
Technical
Reference:
WOG background documents Proposed references None to be provided:
Learning Objective: 3-0T-EOP0100, 3-0T-EOP01 00, "Loss of Reactor or Secondary Coolant," Obj. 12. Discuss the purpose of ES-1.2, "Post LOCA Cooldown and Depressurization."
Question Source:
New Modified Bank x X
Bank Question History: Seabrook exam question 2003 modified Comments:
Page 8
Breaks -1"
.Breaks - 1" < diameter<-
diameter <- 13-1/2" (1FT2)
(1 FT2)
\, For break sizes of one to two-inch in equivalent diameter, the ReS will rapidly depressurize early in the transient, and an automatic reactor trip and safety injection signal will be generated based on low pressurizer pressure. During the early stages of the depressurization, when the system is still full of of two-phase liquid, the break flow, which also will be mostly liquid, is not capable of removing all the decay heat.
Therefore, the early depressurization is limited by energy removal considerations, and the ReS pressure will temporarily hang up above the steam generator safety valve set pressure, assuming no steam dump is available. The ReS pressure stays at this level in order to provide a temperature difference from primary to secondary so that core heat may be removed by the steam generator. At this energy-balance controlled pressure, however, pumped safety injection flow is less than the break flow, and there is a net loss of mass in the ReS. Voiding throughout the primary side occurs and eventually the ReS begins to drain, starting from the top of the steam generator tubes.
The rate of ReS drain is determined determined by the net loss of liquid inventory, a function function of both SI flow and break size.
Prior to the occurrence of draining, heat is removed from the steam generator through continuous two-phase natural circulation, with two-phase mixture flowing over the top of
(
the steam generator tubes. As the draining continues, the natural circulation mode of heat removal as just defined ceases, and core heat is removed through condensation of steam in the steam generator. This method of heat removal is called reflux and is discussed in Reference 2.
The condensation mode of heat removal is almost as efficient as continuous two-phase natural circulation in removing removing heat. However, condensation heat transfer coefficients may be lower than continuous two-phase natural circulation heat transfer coefficients.
Thus, as the steam generator tubes drain, a slight increas,ein increase in primary system pressure occurs to give a greater AT LlT from primary to secondary in order to remove all the decay heat. The steam generator secondary side pressurizes to the safety valve set pressure E-1 Background 16 HP-Rev. 2, 4/30/2005 HElBG;doc HElBG.doc
05/30/2003
":M~r~;;iB},I"~1'~'
~.
1**011.EK1.01 1**011.EK1.01 New
~ IJ.Y
- Mark prin"t, ISeabrook Unit Unit 1, 1 Level~'
Exam Level ~' Ouestion 1;
- \'ouestion ~ Recor
.. Search During which of the following scenarios does natural circulation and reflux cooling become important?
Question SBLOCA withl~ak
)Any SSL()CAwith IAny such that SI flow leak rate suchthatSI th~~ breakflow flow is less than b~~akflo~~nd b~iling is core boiling a,nd,core is~ccurring.
oc~urring:
Answer:
Distracter designb~s~~~cidentLOCAcl~s~;ibed Distracter 1 Jonl)' LElLOCAs 10nly LBLOCAs with, with leal< than or equalto design leak, rates greater than base accident LOCA described in FSAR FSAR.* ..
1 Distracter 2 Any LBLOCA SLOCA less than the leak rate of design basis accident. Assume water source is available to S/Gs and S/Gs safety valves work correctly.
Distracter Distracter 3 3 :IAny SBLOCA'ihat does n;j includ'e 'vapor spa~~ accidents.
Distracter Analysis:
Answer: C - correct - Natural circulation & reflux cooling are important when there is sufficient mass in primaryary to sustain density gradient AND a
- ...,......**-.*....**..,-~-- condition when core boiling is in progress (for reflux).
Distracter 1: lA,S A,B - incorrect - natural circulation and reflux cooling are not mentioned as cooling mechanisms in FSAR, WOG, or lesson plan for kS ,.**L
-...O LBLOCA.
CA. .
D- incorrect Distracter 2: lA,S A,B - incorrect - natural circulation and reflux cooling are not mentioned as cooling mechanisms in FSAR, WOG, or lesson plan for
iLSLOCA. LBLOCA.
i D- incorrect Distracter 3: .IA,S A,B - incorrect - natural circulation and reflux cooling are not mentioned as cooling mechanisms in FSAR, WOG, or lesson plan for LSLOCA.
D- incorrect
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 4. 015 AK1.02 004 Given the following:
1 is at 41 % power.
Unit 1is Reactor Coolant Pump (RCP) #4 trips.
Which ONE of the following identifies how steam generator #4 level indications will be initially affected by the RCP trip and the effect the RCP trip has on the reactor status?
SG #4 Level Reactor Status A. Decrease RCP trip generates an automatic reactor trip.
B~ Decrease B:' Manual reactor trip is required due to RCP tripping.
C. Increase RCP trip generates an automatic reactor trip.
D. Increase Manual reactor trip is required due to RCP tripping.
Page 9
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS:
A. Incorrect, the level in the affected loop will drop but with power below P-8, the trip of one RCP will not cause an automatic reactor trip. Plausible because the steam generator level initially dropping is correct and if the power level had been above P-8, then an automatic reactor trip would occur due to the trip of the RCP.
B. Correct, when the reactor coolant pump trips, Tavg in the affected loop will rapidly drop. This causes a drop in the temperature in the steam generator which collapses a large percentage of the steam voids in the steam generator liquid, causing level to initially drop. With power below P-8 (48%), the reactor will not automatically trip but plant procedures require a manual trip of the reactor.
C. Incorrect, the level in the affected loop will not rise (it will drop) and with power below P-8, the trip of one RCP will not cause an automatic reactor trip. Plausible because the the steam generator will stop steaming resulting in feedwater flow higher than steam flow. This would make the level initially rise if not for the collapse of the steam voids. Also, if the power level had been above P-8, then an automatic reactor trip would occur due to the trip of the RCP.
D. Incorrect, the level in the affected loop will not rise (it will drop) but with power below P-8, a manual trip of the reactor will be required. Plausible because the steam generator will stop steaming resulting in feedwater flow higher than steam flow. This would make the level initially rise if not for the collapse of the steam voids. Also, the manual reactor being required due to the trip of the RCP is correct.
Page 10
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 4 Tier: _1_ Group 1 KIA: 015 AK1.02 Reactor Coolant Pump Malfunctions Knowledge of the operational implications of the following concepts as they apply to Reactor Coolant Pump Malfunctions (Loss of RC Flow):
Consequences of an RCPS failure Importance Rating: 3.7 1 4.1 10 CFR Part 55: 41.8/41.10/45.3 10CFR55.43.b: Not applicable.
KIA Match: Requires applicant to identify the consequences of an RCP failure on both the steam generator level on the same loop and the status of the reactor due to the trip of the RCP.
Technical
Reference:
AOI-24, RCP Malfunctions During Pump Operation, Rev 29 3-0T-AOI-2400, AOI-24, RCP Seal Malfunctions during Pump Operation, Rev 7 Proposed references None to be provided:
Learning Objective: 3-0T-AOI2400 10 .. Given a set of plant conditions, use AOI-24 to 10.
correctly:
- a. Recognize Entry Conditions.
- b. Identify Required Actions.
- c. Respond to Contingencies (RNO).
- d. Observe and Interpret Cautions and Notes.
Question Source:
New Modified Bank X Bank Question History: WBN bank question A010500.01 016 Comments:
Page 11
3-0T-AOI2400 3-0T -A0I2400 Revision 7 Page 7 of34 IX. INTRODUCTION INSTRUCTOR NOTES AOI-24 provides: Objective 1
- Instructions for the removal of a RCP due to a Use current AOI to emphasize condition which requires stopping the RCP as important actions to minimize soon as possible after trouble develops. event, & loss ofRCS of RCS inventory.
- Symptoms and Operator Actions for RCP Seal Malfunctions to decrease the potential for seal degradation and higher leakage.
- Actions to be taken if a RCP is tripped or shutdown with reactor in Mode 1.
Discussion related to RCP removal: Discussion Section of AOI-24 Unscheduled removal of an RCP is defined as a RCP trip or conditions requiring the RCP to be removed from service as soon as possible.
Removing an RCP from service at power causes an Temperature Increase and RCS increase in TT-hot
-hot for the active loops (Core outlet flow decrease reduces margin temperature closer to saturation) and the potential for toDNB DNB.
When an RCP is removed from service, a reverse flow Feedwater flow must be condition occurs in the affected loop. Compounding continued until the S/G level this effect is the effect of cold (~440 OF) feedwater to stabilizes, to ensure that the the preheat section of the S/G. The result is a low S/G reactor does not trip on 10wS/G significant reduction in the RCS hot and cold leg level, but should be minimized level, temperatures and cessation of steam generation from to limit the temperature the affected S/G. Previous experience in losing a RCP transient.
"at power" at WBN showed that Tavg in the affected loop went below Tcold in the active loops until feedwater was isolated and a thermal equilibrium was reached. In this instance, Tavg for the affected loop went below the minimum temperature for criticality.
Tavg returned to greater than 551°F in ~6 minutes following isolation of feedwater to the idle S/G
WBN RCP MALFUNCTIONS DURING 'PUMP AOI-24 Unit 1 OPERATION Rev. 0029 3.0 OPERATOR ACTION:
3.1 Diagnostics IF GO TO Subsection Rep tripped or shutdown required 3.2
- 1 sealleakoff flow HIGH, 3.3
- 1 seal sealleakoff leakoff flow LOW, 3.4 AND Standpipe level alarm DARK,
- 2 Seal Leakoff Flow HIGH 3.5
(#1 sealleakoff flow LOW, AND Standpipe level alarm LIT),
- 3 sealleakoff flow HIGH 3.6
(#1 seal leakoff flow NORMAL AND Standpipe level alarm LIT),
Page 5 of 27
WBN RCP MALFUNCTIONS DURING PUMP AOI-24 Unit 1 OPERATION Rev. 0029 IStep II Response~~
Action/Expected Response II Response Not Obtained 3.2 RCP Tripped Or Shutdown Required NOTE 1 Malfunctions addressed by this procedure require RCP shutdown as soon as possible.
NOTE 2 Exceeding any of the limits listed on Attachment 2 of this procedure will require immediate shutdown of the affected RCP.
NOTE 3 Malfunctions resulting in high #1 seal leakoff will require closing #1 seal return FCVfoliowing FCV following RCP coastdown
I~
- REFER TO ATTACHMENT 2, RCP
~ Immediate Shutdown Criteria.
~
~i"
- 1) IF RCP immediate shutdown
- ~'
required, THEN
- GO TO Step 2.
- 2) IF RCP immediate shutdown NOT required, THEN
- GO TO Step 9
- 2. CHECK unit in Mode 1 or 2 ** GO TO Step 4.
NOTE Control room staff should brief on on Steps 3 through 6 prior to tripping the reactor. This ensures that the affected RCP is stopped and that appropriate actions are taken when unit is removed from service.
Page 6 of 27
WBN RCP MALFUNCTIONS DURING PUMP AOI-24 Unit 1 OPERATION Rev. 0029 IStep II Action/Expected Response II Response Not Obtained 3.
TRIP the reactor, and GO TO E-O, Reactor Trip or Safety Injection, WHILE continuing with this instruction.
- 4. STOP and LOCK OUT affected RCP(s).
- 5. IF in Mode 3, **GO TO ES-O.2, Natural Circulation THEN Cooldown, WHILE continuing with this CHECK any RCP Running instruction CAUTION If the RCP seal return flow control valve (FCV) is NOT closed
( within 5 minutes of stopping the RCP with excessive leakoff, seal damage may occur.
- 6. MONITOR RCP seal leakoff less than WHEN the RCP has coasted down 6 gpm per pump: (between 3 and 5 minutes),
THEN
- 1-FR-62-24 [RCP 1 & 2]
- 1-FR-62-50 [RCP 3 & & 4] CLOSE affected RCP seal return FCV:
- 1-FCV-62-9 [RCP 1]
- 1~FCV-62-22 [RCP 2]
1-FCV-62-22
- 1-FCV-62-35 [RCP 3]
- 1-FCV-62-48 [RCP 4]
- 7. CHECK RCPs 1 and 2 running. CLOSE affected loop's pressurizer spray valve.
- 8. GO TO Step 15.
Page 7 of 27
WBN REACTOR TRIP RESPONSE II ES-O.1 ES.o.1 Rev 21 IStep II Action/Expected Response I IResponse Not Obtained NOTE Either Loop 1 or 2 pzr spray valve is effective for Loop 2 RCP in
& 4 Reps service or for Loops 1, 3, & RCPs in service.
- 20. CHECK RCP(s) RUNNING to provide ESTABLISH normal pzr spray, normal pzr spray. Loop 2 preferred:
REFER TO SOI-68.02, Reactor Coolant Pumps (Loop 2 OR Loops 1,3, 4).
1, 3, and 4).
WHEN RCP start conditions are established, THEN START Loop 2 RCP to provide normal pzr spray.
IF Loop 2 RCP can NOT be started, THEN START ALL other RCPs to establish normal pzr spray.
IF an RCP can NOT be started, THEN MONITOR natural circulation per the Foldout Page.
IF natural circulation is NOT established, THEN DUMP steam at a greater rate.
(
15 of 18
QUESTIONS REPORT for ILT Bank Current wo pw AOI0500.01 016 Given the following conditions:
- Reactor Power is 38%
- All control systems are in automatic
- Loop 2 RCP trips on overcurrent Which ONE of the following describes the expected plant response, and the correct action(s) to mitigate the event lAW AOI 5 "Unscheduled Removal of One RCP."
- a. #2 Steam generator steam flow rises rapidly, take manual control of feed regulating valve and increase feed rate to maintian S/G level.
maintianS/G b ..... # 2 Steam generator level will lower rapidly then begin to rise, take manual control of feed regulating valve and decrease feed rate to prevent overfill.
- c. RCS Pressure will decrease rapidly, take manual control of RCP 2 spray valve and close to minimize effects on DNB.
- d. RCS Pressure will increase, take manual control of RCP 2 spray valve and open to reduce RCS pressure.
Wednesday, September 30, 2009 20092:28:05 2:28:05 PM 1
11/2009 Watts Bar 11/2009 Watts Bar RO RO NRC NRC Exam Exam -- As As submitted submitted 10/2/2009 10/2/2009
- 5. 022 022 AA1.02 AA1.02 005 005 Given the Given the following:
following:
- Unit 1 at 100% reactor Unit reactor power.
- Charging is being supplied through 1-FCV-62-86, 1-FCV-62-B6, "Alternate Charging to to Loop 4."
Loop
- Letdown flow is 75 gpm.
Letdown If 1-HS-62-90A, "Charging Line Isolation" was placed to CLOSE by the MCR operator, which ONE of the following completes the statement below?
Annunciator 108-A, 10B-A, "CHARGING FLOW HIILO" HI/LO" would alarm when the charging flow dropped to (1) and when the valve reached full closed, the charging flow indicator would be _ __ _-"(2=.)L-_
- l. =(2::.L. )_ _ ____
A. (1 ) 47gpm (2) indicating '0' flow B. (1 ) 47gpm 47 gpm (2) indicating flow C. (1 ) 55gpm 55 gpm (2) indicating '0' flow D¥'
D~ (1) 5555gpm gpm (2) indicating flow Page 12 Page 12
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: YS/S:
A. Incorrect, The alarm will come in when the flow drops to 55 gpm (not 47 gpm) but the charging flow indication will not be reading '0' when the valve reaches full closed. Plausible because the alarm would have come in at 47 gpm if letdown flow had been 45 gpm (both 75 gpm orifices closed, 45 gpm orifice open) and if 1-FCV-62-93 (valve in the flow path) had been full closed then charging flow indication would be reading '0'.
B. Incorrect, The alarm will come in when the flow drops to 55 gpm (not 47 gpm) but the charging flow indication will be indicating the flow being supplied to the RCP seals when the valve reaches full closed. Plausible because the alarm would have come in at 47 gpm if letdown flow had been 45 gpm (both 75 gpm orifices closed, thus a 45 gpm orifice open) and the charging flow indication reading flow is correct.
C. Incorrect, The alarm will come in when the flow drops to 55 gpm but the charging flow indication will not be reading '0' when the valve reaches full closed. Plausible because the alarm coming in at 55 gpm is correct (a 75 gpm orifice opens) and if valve 1-FCV-62-93 (valve in the flow path) had been full closed then charging flow indication would be reading '0'.
D. Correct, With a 75 gpm orifice open, the low charging flow alarm will annunciate when flow drops to 55 gpm and when the charging line isolation valve reaches full closed the charging flow indication will still indicate approximately 32 gpm due to the flow being supplied to the RCP seals.
Page 13
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
( Question Number: 5 Tier: __ Group
_1_
_1 1 KIA: 022AA1.02 022 AA 1.02 Loss of Reactor Coolant Makeup Ability to operate and 1 / or monitor the following as they apply to the Loss of Reactor Coolant Makeup:
CVCS charging low flow alarm, sensor, and indicator Importance Rating: 3.0/2.9 3.0 /2.9 Importance Rating: 3.0/2.9 3.0 /2.9 10 CFR Part 55: 41.7/45.5/45.6 10CFR55.43.b: Not applicable.
KIA Match: A loss of reactor coolant makeup would result from closing the charging line isolation valve. The applicant is required to recall the setpoilJt setpoint of the low charging flow alarm and determine the indication on the charging flow indicator by relating the location of the flow sensor relative to the valve being closed.
Technical
Reference:
ARI-102-108, ARI-1 02-1 08, "HVAC & CVCS," Rev 25 Drawing 1-47W809-1, R56 Proposed references None to be provided:
Learning Objective: 3-0T-SYS062A
- 4. Using a block diagram of CVCS, explain system flow balance.
Question Source:
New X Modified Bank Bank Question History: New question Comments:
Page 14
D01M~1\)
Dt-S1 Mt1\) ~5~5 108-A SOURCE 1-FS-62-93-AlB SET POINT , /
SETPOINT 150gpm 150 gpm i/
~
6:.-
A~G PLANT COMPUTER 47 gpm with BOTH 75 gpm orifice valves CLOSED CHARGING OR FLOW "711
~ 55 gpm with EITHER 75 gpm HilLa Q~~~~tt..1 Q()~<<~'L1 ./
./' I orifice valve OPEN AN.~wt(L ANSwtfl Probable Cause: 1. System pipe break
- 2. Charging pump tripped
- 3. Malfunction of PZR level control system.
Corrective Action: [1] IF ALL the following conditions exist:
- Any RCP Thermal Barrier Out-Of-Service,
- In-Service Charging pump trips,
- RCP seal injection flow required, THEN IMMEDIATELY START available charging pump to restore seal flow.
( [2] CHECK 1-FI-62-93A [1-M-5] to determine if flow is high or low.
[3] CHECK PZR level indication on 1-M-4.
[4] IF PZR level control system malfunction, THEN GO TO AOI-20, MALFUNCTION OF PRESSURIZER LEVEL CONTROL CHANNEL.
[5] IF charging flow is low, THEN CHECK letdown temperature and CONSIDER increasing charging flow, or ISOLATE letdown.
[6] IF charging is lost, THEN IMMEDIATELY ISOLATE letdown.
[7] DETERMINE cause of problem and INITIATE corrective action.
[8] REFER TO SOI-62.01, CVCS - CHARGING AND LETDOWN.
References:
1-4 7W61 0-62-2 1-47W809-1 AOI-20 SOI-62.01 WBN Page 42 of 50 II ARI-1 02-1 08 1
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REV CHANGE REF APPROVED DATE SCALE: NTS EXCEPT AS NOTED PROJECT FACILITY FACILITY POWERHOUSE G UNIT 1 TITLE FLOW DIAGRAM CHEMICAL & ~ VOLUME CONTROL SYSTEM WATTS BAR NUCLEAR PLANT Q TENNESSEE VALLEY AUTHORITY INITIAL ISSUE ENGINEERING APPROVAL H RO ISSUE PER 1 JJ.R.IETH.R.WETH RC~ STANDPIPE L.S.
CALJBRATiONCONN.
W8EP 5.17 WEEP 5.17.t. ct. RIMS RIt.!S 826 '90 0514 aSH 376 J78 2 J.R.WETH J.R.IETH A. DeN M-01591-A I:. M-01S91-A
.3 L.W.SOYD L .W.BOYO FSAR FIG. (
9.3-15 SH 1 1-47W809-1 1-47wa09-1 .58 R58
( 2 3 4 5 8 "
e~ ORPROJ BR OR PRaJ AB lie C[ (( EE ~ t.IE N[
NE FE HE SE S[ TE BL 8L B'"IIB Bf liB so I CAD MAINTAINED DRAWING I (CONFIGURATION CONTROL DRAWING)
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11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 6. 025 AK2.05 006 6.025 Given the following:
Unit 1 at 100% power when a Reactor trip and Safety injection are initiated due to a LOCA.
RHR pump 1A-A is tagged with maintenance in progress.
RHR pump 1B-B trips when it attempts to start.
- The crew makes a transition from E-I, "Loss of Reactor or Secondary Coolant," to ECA-1.1, "Loss of RHR Sump Recirculation."
- The conditions at the transition are:
- RCS pressure is 480 psig.
- RWST level is 73% and dropping.
- Containment sump level is 17% and rising.
- Containment pressure had peaked and is now stable at 3.2 psid.
- STA reports an Orange path to FR-Z.1 ,"High Containment Pressure."
Assuming the plant conditions remain unchanged, which ONE of the following identifies ...
(1) the procedure controlling the operation of the containment spray pumps, and (2) if one of the spray pumps is later required for restoring the level in the RWST, how it will be aligned from the containment sump to the RWST?
(1 ) (2)
Procedure Alignment A. FR-Z.1 Through the containment spray test line.
B. FR-Z.1 Through the containment spray pump mini flow line.
C~ ECA-1.1 Through the containment spray test line.
D. ECA-1.1 Through the containment spray pump mini flow line.
Page 15
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS:
A. Incorrect. While FR-Z 1 normally controls the operation of the containment spray pump, if ECS-1.1 is in effect there is an exception. This is stated in step text in ECA-1.11 and in a Caution at the beginning of FR-Z.1.
ECA-1. FR-Z 1. If the containment spray pump was used to restore water to the Flow Test Line, isolation valves would be opened to establish the flow path. Plausible because FR-Z 1 normally controls the operation of the containment spray pumps and using the flow test line is correct.
B. Incorrect. While FR-Z 1 normally controls the operation of the containment spray pump, if ECS-1.1 is in effect there is an exception. This is stated in step text in ECA-1.1 and in a Caution at the beginning of FR-Z1. If the containment spray pump was used to restore water to the RWST from the containment sump, the Mini flow line would not be used (the Flow Test Line isolation valves would be opened to establish the flow path). Plausible because FR-Z1 FR-Z.1 normally controls the operation of the containment spray pumps and the mini flow line does return to the suction of the respective spray pump.
C. Correct. ECA-1.1 would control the operation of the containment spray pumps. This is identified in both step text in ECA-1.1 and in a Caution at the beginning of FR-Z 1. If the containment spray pump was used to restore water to the RWST from the containment sump, then the Flow Test Line isolation valves would be opened to establish the flow path D. Incorrect, ECA-1. 1 would control the operation of the containment spray pumps.
This is identified in both step text in ECA-1.1 and in a Caution at the beginning of FR-Z1. If the containment spray pump was used to restore water to the RWST from the containment sump, the Mini flow line would not be used (the Flow Test Line isolation valves would be opened to establish the flow path. Plausible because ECA-1.1 controlling the operation of the pumps is correct and the mini flow line does return to the suction of the respective spray pump.
Page 16
11/2009 Watts Bar 11/2009 Watts Bar RO RO NRC NRC ExamExam -- As As submitted submitted 10/2/2009 10/2/2009 Question Number:
Question Number: 66 Tier: _1_
_1 Group _1_
__ Group 1 KIA: APE 025 AK2.05 APE of RHR Loss of RHR system, Knowledge of of the interrelations interrelations between the Loss of of Residual Heat Residual Heat Removal System and the the following: Reactor Building Building Sump Sump Importance Rating: 2.6 /12.6 2.6 10 CFR Part 55: 41.7 I 45.7 41.7/45.7 10CFR55.43.b: Not applicable.
KIA Match: Applicant is required to assess conditions that will result from the loss of RHR containment (reactor building) sump recirculation capability and if not resolved how the water from the containment sump will be returned to the RWST for continued core cooling capability following entry into ECA-1.1, due to a loss of Residual Heat Removal System.
Technical
Reference:
ECA-1.1, "Loss of RHR Sump Recirculation," Rev 11 E-1, "Loss of Reactor or Secondary Coolant," Rev 15 FR-Z.1, "High "Hig h Containment Pressure," Rev 10 Proposed references None to be provided:
Learning Objective: 3-0T-ECA0101
- 01. Identify and explain the major actions of ECA-1.1 and ECA-1.2.
Question Source:
New X Modified Bank Bank Question History: New question Comments:
Page 17 Page 17
WBN HIGH CONTAINMENT PRESSURE FR-Z.1 Rev 10 IStep IIIIAction/Expected Response II Response Not Obtained CAUTION If ECA-1.1, Loss of RHR Sump Recirculation, is in effect, the number of cntmt spray pumps to be operated is directed in ECA-1.1 rather than in this instruction.
NOTE Adverse containment setpoints [ADV] should be used where provided due to Phase B actuation.
- 1. ENSURE cntmt spray operation: ESTABLISH -at least one train of cntmt spray flow.
- a. Cntmt spray signal ACTUATED.
- b. Cntmt spray pumps RUNNING.
- c. Cntmt spray valves 1-FCV-72-2 and 1-FCV-72-39 OPEN.
(
- d. Cntmt spray pump suction valves OPEN:
- Valves from RWST:
- 1) 1-FCV-72-21 and
- 2) 1-FCV-72-22.
- Valves from cntmt sump:
- 1) 1-FCV-72-44 and
- 2) 1-FCV-72-45.
- e. Cntmt spray flow:
- 1-FI-72-34.
- 1-FI-72-13.
((,
3 of 7 30f7
WBN LOSS OF RHR SUMP RECIRCULATION ECA-1.1 Rev 11 IStep II Action/Expected Response I IResponse Not Obtained
- 4. DETERMINE cntmt spray pump alignment and operation:
- GO TO Step 6.
- b. MONITOR cntmt press, and DETERMINE number of spray pumps required:
CONTAINMENT PRESS SPRAY PUMPS REQUIRED Greater than 13.5 psig 2 2.0 psig to 13.5 psig 1 Less than 2.0 psig 0
(
- c. CHECK number of spray pumps c. STOP and PULL TO LOCK any running equal to number required. cntmt spray pump NOT required, AND CLOSE discharge valve(s) 1-FCV-72-2 and/or 1-FCV-72-39 for pump(s) stopped.
Manually OPERATE spray pumps as required.
- d. DO NOT OPERATE cntmt spray pumps as required by FR-Z.1, High Containment Pressure, UNTIL either of the following:
- Cntmt spray pump suction aligned to cntmt sump.
- RWST makeup sufficient to support cntmt spray pump operation.
4of34
WBN LOSS OF RHR SUMP RECIRCULATION ECA-1.1 Rev 11 IStep I IAction/Expected.
Action/Expected Response III IResponse Not Obtained
- 1-HS-63-720.
1-HS-63-72D.
- 1-HS-63-730.
1-HS-63-73D.
- 10. INITIATE makeup to RWST:
- a. NOTIFY RadprotiChemistry to evaluate radiation level of water in cntmt sump for potential transfer to RWST.
- Appendix C (ECA-1.1), Cntmt Spray Recirc to RWST Alignment.
- Spent fuel pit.
- Holdup tank.
WBN LOSS OF RHR SUMP RECIRCULATION ECA-1.1 Rev 11 APPENDIX C (ECA-1.1))
(ECA-1.1 Page 1 of2 CNTMT SPRAY RECIRC TO RWST ALIGNMENT (TRAIN A)
- 1. IF Train A cntmnt spray is to be aligned to recirc to RWST, THEN PERFORM the following:
- a. NOTIFY RADPROT of RWST alignment.
( b. RESET Cntmt Spray signal.
- c. PLACE cntmt spray pump A in PULL-TO-LOCK.
- e. CLOSE 1-FCV-72-39, CNTMT SPRAY HDR A TO CNTMT.
- g. OPEN 1-ISV-72-502, CNTMT SPRAY TEST LINE ISOLATION
[A5U/715 BIT rm].
[A5U1715
- h. OPEN 1-ISV-72-503, CNTMT SPRAY HDR A TEST LINE ISOLATION
[A5V/719 BIT rm].
- j. START cntmt spray pump A.
32 of 34
WBN LOSS OF RHR SUMP RECIRCULATION ECA-1.1 Rev 11 APPENDIX C
. (ECA-1.1)
(ECA-1.1 )
Page 2 of2 of 2 CNTMT SPRAY RECIRC TO RWST ALIGNMENT (TRAIN B) 8)
- 2. IF Train B cntmnt spray is to be aligned to recircto recirc to RWST, THEN PERFORM the following:
- a. NOTIFY RADPROT of RWST alignment.
( b. RESET Cntmt Spray signal.
- c. PLACE cntmt spray pump B in PULL-TO-LOCK.
- e. CLOSE 1-FCV-72-2, CNTMT SPRAY HDR B TO CNTMT.
- g. OPEN 1-ISV-72-502, CNTMT SPRAY TEST LINE ISOLATION
[A5U/715 BIT rm].
[A5U1715
- h. OPEN 1-ISV-72-504, CNTMT SPRAY HDR B TEST LINE ISOLATION
[A5V/719 BIT rm].
- j. START cntmt spray pump B.
33 of 34
22 3 4 5 6 7 8 9 10 11 12 CONT ON \-47.",811-COORD 0-11 153-374.4.63-371'" LT 63-51 A A UNIT 1 REFU£LlNG'IIATER STOR ....CE TANK CLASS e CAP .... CITy:
J80,OOO CAllONS
""C- :5/""
CONTAINMENT SPRAY HEADER-8 FLUSHING CONNECTOR B
8 OUTSIDE AUX BlOG BLOC 8B
,3"X 2" REDUCER
".1101 FLOW PUoIP RECIRCULATION LINE
'~ ,I
-l.!'
c &~&
RE~~VALV£-SET ~I~j AT 100 PSI ill r WOTES:
- 1. All PIPINC ON THIS DRAWING IS TVA CLASS e UNLESS c
RROD-72-514 PI I 72-508 OTHERWISE INDICATED.
1" CASING VENT (TYP) 2. All VALVES ARE THE SAIAE SIZE AS PH'INC. UNLE:SS CL ....SS C OTHERWISE NOTED.
::L 3. ALL INSTRUI,IENT VALVES ARE 3/4" CLOBE UNLESS OTHERWISE NOTED.
- 4. ALL L8EPR~DWlTH L8EPR~O"lTH T~ lA-A, Ere.
ETC.
- i. ALL CONNe:CT ON ARE 3/"'
CONNECT 3/4"...
- 6. FOR OR "L.O.",
- L.O.", REFER TO
- 7. T.V.
.,C
'"ORAIN
'" DRAIN.
>TC ..
8.
B. 1 ETC .* I
.. T~PERATURE EMPtRATL/HE OF Of ALL DRAIN VENTS S THRU HRU THE LAST ISOLAnON ISOLATION o
D II.
- 11. £ETC
&'ETC ** INDICATESINDICATES LINE SAIAE AS THE PROCESS LINE.
LINE NUI.If!ER UNE.
NUt.E!ER CORRESPONDING TO HYDROSTATIC HYDROSTATIC TEST PRESSURE D
CHART ON THIS SHEET. HYDROSTATIC HYtlROSTATIC T(STING n::STING SHALL BE BE: IN ACCORDANCE ACCORD"NCE "fITH WITH APPLICABLE CODE CASES WITH CASE BY CASE APPLICATION REQUIRING APPLICA8LE REQUIRING PRIOR APPROVAL 8Y ENGINEER INC. CS .. RHR SPRAY HEADERS AND CS PlAIP BY NUCLEAR ENGINEERING. PUUP TtST
&1I
&1I7l--I--ill 7 }--+---I1l lINE LINE I4AY MAY BE HYDROSTATICALLY TESTED IN ACCORDANCE WITH wITH CODE CODe: CASE C",SE N-237.
TEST
- 12. ~~~I~¥:I~~DO~~~~T~b~E~11~~
DRAINS. AND TEST CONNECTIONS \~°1tV~~A~~L~Eu~~E~'f~\HTh~~~t
- 12. ALL PIPING DOWNSTREAM OF THE ISOLATION VALVe: ON LOCAL VENTS.
IS TVA CLASS G UNl.ESS OTHERWISE NOTED.
13.
1.3. 0 I S CONTAINWENT cz:lIS CONTAINMENT SPRAY Pl1!.IPf'tJt.If' START UP STRAINER SYMBOL. SYldBOL. START UP
~RAINERS TO BE REI.IOVEO RE~VEO AFTER Ut-iIT UNIT 1 t CLEANUP AND iJSED USED fOR UNIT 2.
- 14. , t::r.: ~ INDICATES EXTENSION VALVE STEM OPERATOR 1!I. DES1 l:i. /SYSTEM DESCRIPTION IPTIOH REFERENCE DOCUMENTS
~~1E /
~~~E R~~~~1b~ ~NT~~L ~lL4~~~ +'Wi~ ~~~1~~ ~~~1rs ~1~i~rJ:~E.Wi~1lfoIED.
%1~i~~~E *~WiifGIED.
,m E
LINE MK LINE~!C HUMBER NUMBER HYDROSTATIC TEST PRESSURE (PSIC) (PSIG) 16. ~~!~~g~lUNIT i~I~ IS A UNIT :~~!:~N~i~~T~~:~C;E~i~i.
lJUNIT 2 INTE~;:c:E~i:i.S~:;:T~Y~i~~EO. S~~;:T;Yi~~~EO. <lID. E
"'" ~~n~~~g
~~n=~~g ~i~t~ ~g~~+~ t:l~t
~~+ ~~ t~~~g
[~~g iW~E~AN~O~;;l~[~T~E~~6~D.[illJ
- ~~E~;'N~O
- H~~r§T~E~~6:D.mJ *
".'"'00,
. ~ SH~~~O~~SPhl>i-~CtST~~a~~~~kk~ ~~~11R~N I!mJ SH~~~0\~Sp~11~c\ST~1:0~k~~~1~ ~~~b~R~N I%1bWiT~~I(~~~Al~~~E~S I~11:bmT~0'g;-I1~~A11~~E~S JOO 17.
'" "'8. TO PREVENT BONNET 90NNET OVERPRESSUIHZAnOM OVERPRESSURIZATION VALVES I-FCV-72-H l-FCV-72-+4 AND 4~ 4~
750 HAVE A RELIEf RELIEF LINE INSTALLED. SEE VENDOR VALVE DRA"fINC
- 19. TO PREVENT BONNET OVERPRESSURIZATION VALVES l-FCV-72-40 AND ...1,1 DRAWING; FOR DETAILS.
-41 HAVE A RELIEF UNE LINE INSTALLED. SEE VEN~OR VENDOR VALVE DRAWING fOR FOR DETAILS.
FOR PIPING ONLY, LIMn!NG LIMITING C(),IPONENTS COMPONENTS TO 8E BE DETERMINED 20. VALVES DENOTED ,l,S AS L.O. ARE lOCKED LOCKEO OPEN DURING NORMAL PLANT OPERATION.
BYFIELD BY FIElD VALVES OENOTEa DENOTED AS L.T. ARE LOCKED IN 11'4,1, POSITION DURING rtORIML A THROTTLED POSITION. NORMAL OPER,I,TION.
PLANT OPERATION.
HYDROSTATIC TEST PRESSURE H'rOROSTATIC 21. SEE SYSTEt.4 SYSTEM DESCRIPTION N3-72-4001 SECTION 3.2.4 3.2.+ FOR UNOBSTRUCTEO UNOBSTRUCTED HOZZlE NOZZLE DATA OAT ... IS 111STORIC"L HISTORICAL INFORIAATION INFORMATION QUANTITY QUANTIl'I' OESIGNDESIGN BASIS.
AND NO LONGER lONC£R MAINTAINED MAINTAINED REfERErtCE DRA'lHIiGS, OUTPUT.
AS DESIGN OUTPuT.
4711800-1 ---------------- F PLT SYS
-I-7"f610-72 -72 SERIES --------
47W611_ 72 SERIES --------
4-7W601--72 SERIES --------
47W4J7 SERIES ----------- IPH~.
F DESIGN PRESSURE .t.
DESICN TEMPERATURE DA1A A. l'E~PER"TURE DATA 47W&1 ~--100 -------------- ANAL YSIS S PROBlE\'(
PROBlEIA BOUNDARY F 308617 SERIES ----------- ,0.'10 IDENTIFICATION.
NTIFlCATlON.
S2QL072--S£RIES ----------
~6~E LINE p~fgN PRESSURE OESIGN PRESSURE O~iICN TEMPER DESIGN .... TURE TEMPERATURE 478437-7 -12Xl.-72X2 ------- MASTER VALVE STATUS REPORT.
NO. PSIG 'I' SO WAT BT 906-4 -------- WESTINGHOUSE SPECIFICATIONS. SPECIFICATIO~S.
1 50 190 l1JE&0 ----------------- WESTINGHOUSE e.s. C.S. FLOW OIAGRAIoI.
DIAGRAM.
UPPER COMPARTMENT 100 100 '"
150 190 THIS CONrtGURATtON CONTROL DRAWING SUPERSEDES UNIT 1 AS-CONSTRUCTED DRAWING 47W912-1 REVISION M.
LOWER COMPARTMENT 240 190 26 240
'0' 105 3/4' 3/..
- TV 150 150 '"
150 190 115 200 PREPARER CHECKER APPROVED DATE 600 400
<00 EXCEPT AS NOTED PROJECT FACILITY POWERHOUS G UNITS 1 TITLE FLOW DIAGRAM CONTAINMENT SPRAY SYSTEM WATTS BAR WATTS BAR NUCLEAR NUCLEAR PLANT PLANT Q TENNESSEE VALLEY AUTHORITY INITIAL ISSUE ENGINEERING APPROVAL H Ra RO ISSUE PER 1 R.E.KIRSCH 1 R.E.KIRSct-l weEP 5.17 ~.17 ... .t RIMS 22 R.E.KIRSCH R.E.Klfl:SOi 826 '90 0119 375 825 376
.3J L.W.eaYO L.W.BOYD ISSUED BY:
8'1', PATE FSAR FSAR FIG 6.2.2-1 R.M.JOHNSON FOR IISR R.!.A.JOHNSON ISR 7-23-90 1-47W812-1 R2S R26
, I 2 3 4 5 6 7 8 Bft Oft ~ftOJ AB CE £E WE ME rE HE SE TE at. 9F" ,a sa 11ft 0fiI
!III: ~!\OJ .0.8 Oft ~1\OJ CE EE !.IE AB C[ loIE liE n NE fE HE n lIE SE Tt at.Bt. er SF" ..
'B 8 sa II CAt) MAIl'JTAlNEO DRAWING 1 (CONFIGURATION CAO MMNTAINED DRAWINcj (CONF IGURAT ION CONTROL ORAWI-NC) DRAWING)
Watts Bar 11/2009 Watts 11/2009 Bar RO RO NRC NRC Exam Exam -- As As submitted submitted 10/2/2009 10/2/2009
- 7. 026 026 AK3.02 AK3.02 007 007 Given the Given the following:
following:
Unit 1 is at 100%
Unit 100% power power when when a LOCA occurs.
Which ONE of Which of the the following following identifies an an automatic automatic signal that that wi" will cause an an of the isolation of the CCS to to the the RCP RCP thermal barrier barrier cooling cooling during an an accident accident and reason for the reason for the isolation?
A. Phase A Isolation To isolate a potential containment release path.
B. Phase A Isolation To protect the CCS system from steam formation.
C!'
C~ Phase B Isolation To isolate a potential containment release path.
D. Phase B Isolation To protect the CCS system from steam formation.
(
Page 18 Page 18
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS:
A. Incorrect, the CCS Thermal Barrier flow path is not isolated on a Phase A containment isolation signal (the path is isolated on a Phase B signal) but the isolation is to prevent a potential containment release path to the environment.
Plausible because other containment penetrations isolating on a Phase A signal to prevent a potential containment release path to the environment is correct.
B. Incorrect, the CCS Thermal Barrier flow path is not isolated on a Phase A containment isolation signal (the path is isolated on a Phase B signal) and not to prevent the formation of steam in the CCS steam (it is to prevent a potential contaInment containment release path to the environment). Plausible because other containment penetrations do isolate on a Phase A signal and the CCS return path from the Thermal Barrier is isolated during a loss of shutdown power event to prevent the formation of steam in the CCS system which could be flushed into the CCS system.
C. Correct, the CCS Thermal Barrier flow path is isolated on a Phase B containment isolation signal to prevent a potential containment release path to the environment.
D. Incorrect, the CCS Thermal Barrier flow path is isolated on a Phase B containment isolation signal but not to prevent the formation of steam in the CCS steam (it is to prevent a potential containment release path to the environment). Plausible
( because the isolation signal is correct and the CCS return path from the Thermal Barrier is isolated during a loss of shutdown power event to prevent the formation of steam in the CCS system which could be flushed into the CCS system.
Page 19
11/2009 Watts 11/2009 Watts Bar Bar RORO NRC NRC Exam Exam -- As As submitted submitted 10/2/2009 10/2/2009 Question Number:
Question Number: 77 Tier:
Tier: _1_ Group
_1_ Group 22 KIA:
KIA: 026 AK3.02 026 AK3.02 Loss of Loss of Component Component Cooling Cooling Water Water (CCW)
(CCW)
Knowledge of Knowledge of the the reasons reasons forfor the the following following responses responses as as they they apply apply to to the Loss the Loss of of Component Component Cooling Cooling Water:
Water:
The automatic The automatic actions (alignments) within actions (alignments) within the the CCWS CCWS resulting resulting from from the the actuation ofof the ESFAS.
Importance Rating: 3.6 1 3.9 3.6/
10 CFR Part 55: 41.5,41.10/45.6/45.13 10CFR55.43.b: Not applicable.
KIA Match: The applicant must identify both a condition that would automatically cause an alignment change in the CCS system and the reason for the automatic action.
Technical
Reference:
N3-70-4002, System Description for Component Cooling Water, R15 Drawing 1-47W611-70-3, R4 WOG ECA-O.O Background Document, Rev 2 WBN FSAR, Amendment 7 Proposed references None to be provided:
Learning Objective: 3-0T-SYS070A
- 8. Describe the thermal barrier system; include purpose, pump capacity, and logic.
Question Question Source:
Source:
New X X
Modified Modified Bank Bank Bank Bank Question Question History:
History: New New question question Comments:
Comments:
Page 20 Page 20
WBNP-1 WBNP-l 6.2.4 CONTAINMENT ISOLA ISOLATION nON SYSTEMS The containment isolation systems provide the means of isolating fluid systems that pass through containment penetrations so as to confine to the containment any radioactivity that may be released in the containment following a design basis event. The containment isolation systems are required to function following any design basis event that initiates a Phase A or Phase B containment isolation signal or releases radioactive materials into containment to isolate nonsafety-related fluid systems penetrating the containment. The Watts Bar Nuclear Plant does not have a particular system for containment isolation, but isolation design is achieved by applying common criteria to penetrations in many different fluid systems and by using ESF signals to actuate appropriate valves.
6.2.4.1 Design Bases The main function of the containment isolation system is to provide containment integrity when needed. Containment integrity is defined to exist when:
1.
I. The nonautomatic containment isolation valves and blind flanges are closed as required.
- 2. The containment equipment hatch is properly closed.
- 3. At least one door in each containment personnel air lock is properly closed.
( 4. Automatic containment isolation valves are operable or are deactivated in the closed position or at least one valve in each line having an inoperable valve is closed.
- 5. Requirements of the Technical Specification with regard to containment leakage and test frequency are satisfied.
Containment integrity is required if there is fuel in the reactor which has been used for power operation, except when the reactor is in the cold shutdown condition with the reactor vessel head installed, or when the reactor is in the refueling shutdown condition with the reactor vessel head removed. Containment isolation is not essential for design basis events, such as HELBs, outside containment which do not release radioactive materials into containment. The failure of containment isolation valves for such an event would not result in the release of radioactive fluids from inside containment.
In general, the containment isolation system is designed to the requirements of General Design Criteria 54,55,56, and 57 of 10 CFR 50, Appendix A. The following are alternate containment isolation provisions for certain classes of lines:
6.2.4-1
WBNP-l 6.5.3 Fission Product Control Systems 6.5.3.1 Primary Containment The primary containment is designed to assure that an acceptable upper limit leakage of radioactive material is not exceeded under design basis accident conditions. For purposes of integrity, the primary containment is composed of both the free-standing steel shell containment vessel and the containment isolation system. This structure and system are directly relied upon to maintain containment integrity. The primary containment functional design is described in Section 6.2.1.
Containment isolation can be initiated by either of two signals:
Phase A signal is generated by either of the following:
- 1. Manual - either of two momentary controls.
- 2. Safety injection signal generated by one or more of the following:
- a. Low steamline pressure in any steamline.
- b. Low pressurizer pressure.
- c. High containment pressure.
- d. Manual - either of two momentary controls.
Manual-(
Phase B signal is generated by either of the following:
- 1. Manual- two sets (two switches per set) - actuation of both switches in either set is necessary for spray initiation.
- 2. High-high containment pressure signals.
Containment isolation Phase A exists if containment isolation Phase B exists; i.e., when the Phase B signal is initiated by automatic instrumentation. Phase A containment isolation does not occur when the Phase B signal is initiated manually. The instrumentation circuits that generate both Phase A and Phase B signals are described in Section 7.1.2.1.2.
Containment purge system isolation (containment purge lines only) can be initiated by either of two signals:
- 1. Manual - Phase A or B manual initiate
- SIS manual initiate
- 2. Automatic - SIS auto-initiate Purge exhaust high radiation (Train A or B sensor) 6.5-10
STEP DESCRIPTION TABLE FOR ECA-O.O Step _8_
\ STEP: Dispatch Personnel To Locally Close Valves To Isolate RCP Seals And Place Valve Switches In CLOSED Position PURPOSE: To isolate the RCP seals BASIS:
This step groups three actions, with different purposes, aimed at isolating the RCP seals. The actions are grouped since all require an auxiliary operator, dispatched from the control room, to locally close containment isolation valves (the reference plant utilizes motor operated valves for the RCP seal return, RCP thermal barrier CCW return lines and RCP seal injection lines). This grouping assumes that the subject valves are located in the same penetration room area and that they are accessible. Concurrent with dispatching the auxiliary operator, the control room operator should place the valve switches for the motor operated valves in the closed position so that the valves remain closed when ac power is restored.
Isolating the seal return line prevents seal leakage from filling the volume control tank (VCT) (via seal return relief valve outside containment) and subsequent transfer to other auxiliary building holdup tanks (via VCT relief
( valve) with the potential for radioactive release within the auxiliary building. Such a release, without auxiliary building ventilation available, could limit personnel access for local operations.
Isolating the RCP seal injection lines prepares the plant for recovery while protecting the RCPs from seal and shaft damage that may occur when a charging/SI pump is started as part of the recovery. With the RCP seal ECA-O.O Background 97 HP-Rev. 2, 4/30/2005 HECAOOBG.doc
STEP DESCRIPTION TABLE FOR ECA-O.O ECA-D.O Step _8_
injection lines isolated, a charging/SI pump can be started in the normal charging mode without concern for cold seal injection flow thermally shocking the RCPs. Seal injection can subsequently by established to the RCP consistent with appropriate plant specific procedures.
Isolating the RCP thermal barrier CCW return outside containment isolation valve prepares the plant for recovery while protecting the CCW system from steam formation due to RCP thermal barrier heating. Following the loss of all ac power, hot reactor coolant will gradually replace the normally cool seal injection water in the RCP seal area. As the hot reactor coolant leaks up the shaft, the water in the thermal barrier will heat up and potentially form steam in the thermal barrier and in the CCW lines adjacent to the thermal barrier. Subsequent automatic start of the CCW pump would deliver CCW flow to the thermal barrier, flushing the steam into the CCW system. If abnormal RCP seal leakage had developed in a pump, the abnormally high leakage rate could exceed the cooling capacity of the CCW flow to that pump thermal barrier and tend to generate more steam in the RCP thermal barrier CCW return lines.
Isolating these lines prevents the potential introduction of this steam into the main portion of the CCW system upon CCW pump start. This keeps the main portion of the CCW system available for cooling equipment necessary for recovering the plant when ac power is restored.
(
ACTIONS:
o Dispatch personnel to locally close valves o Place following switches in closed position:
- RCP seal return outside containment isolation valve
- RCP seal injection outside containment isolation valves
- RCP thermal barrier CCW return outside containment isolation valve INSTRUMENTATION:
Switch position indication for:
o RCP seal return outside containment isolation valve o RCP seal injection outside containment isolation valves o RCP thermal barrier CCW return outside containment isolation valve
(
ECA-O.O Background 98 HP-Rev. 2,
- 2. 4/30/2005 HECAOOBG.doc
Title:
COMPONENT COOLING SYSTEM N3-70-4002 2.2.10 Instrumentation and Control (I&C)
Adequate I&C shall be provided in the MCR and Auxiliary Control Room (ACR) to control and monitor portions of the CCS essential to safe shutdown. Analytical/operational limits for safety-related CCS instrumentation is in Ref. 7.4.7. Reference 7.1.1 provides CCS instruments setpoints and ranges. Reference 7.4.32 provides time delay requirement for the loading of CCS and TEE TBB pumps to the diesel generators.
Table 5 identifies instruments, their functions, and provides the basis for the setpoints.
Instruments with visual indication in the MCR shall be provided to indicate a decrease of the CCS flow in the return header piping of safety-related coolers. A low flow alarm shall be provided in the MCR.
(See Section section 3.3)
Surge tank level shall be used to monitor and control CCS inventory.
Level indication shall be displayed in the MCR and ACR. High and low surge tank level shall alarm in the MCR. (See Section 3.3.2 and Table
- 5)
- Pressure instruments shall be provided to monitor essential portions of the CCS. Local pump pressure indications shall be available for suction and discharge of all essential pumps in the system. Local pressure indication also shall be available for the main supply headers. Pressure in the three CCS pump discharge headers to the HXs
( shall be displayed in the MCR and ACR. A low discharge pressure alarm on headers 1A lA and 2A shall be provided in the MCR. A low discharge header pressure shall start the standby pump in that unit automatically. During LOOP low pressure auto start shall be bypassed.
A pump high discharge pressure alarm shall be provided in the MCR.
(Section 3.3, 3.3.1 and Table 5).
Temperature instruments shall be provided to monitor the essential portions of the CCS. Temperature monitoring instrumentation shall be provided at the outlet of each CCS HX, RHR HX, SFPCS HX, waste Gas Compressor, and Seal Water HX. Temperature monitoring shall also be provided for the Sample Chiller and HX Group, the RCP Upper and Lower Oil Coolers Group, the RCP Thermal Barrier HXs Group, and the discharge header of the Excess Letdown HX. Instrumentation for the above shall provide the MCR with temperature display and a high temperature alarm.
(See Table 5).
MCR instruments shall also display the temperature of the return headers to the CCS pumps. Provisions shall be made for local temperature measurement of other HXs. (Section 3.3)
Instruments shall be provided to allow the operator to identify malfunctioning equipment or leaks and isolate the problem to maintain system function. (Ref. Table 5) 6
Title:
COMPONENT COOLING SYSTEM N3-70-4002 Instrumentation shall be provided to alert the operator to loss of ccs flow to the RCPs (See section 3.4). ccs be, CCS flow to the RCPs will be automatically isolated on a Phase-B isolation signal. (Section 3.3.3)
Instrument and Control power shall be supplied per the criteria of Ref.
7.2.39. Instrumentation to monitor essential variables during post accident conditions shall be designed per Ref. 7.2.40.
IE controls, per Ref. 7.3.14, shall be used for the sample HX and Class 1E chiller isolation instrumentation on low train trcdn A surge tank level (see Section 3.3.5), and to provide for thermal barrier isolation on indication of a thermal barrier leak.
CCS pumps and TBBPs start/stop controls and air operated/motor operated valves open/close controls shall be provided inside the MCR. Components requiring auxiliary control are identified in Ref. 7.2.48. Thermal overloads and the torque switches for motor operated valves (MOVs) shall be bypassed per 'Refs. 7.2.4 and 7.4.16 (see Section 3.2.8).
2 .2.11 2.2.11 Deleted 2.2.12 Quality Assurance (QA)
QA must be applied in the design, procurement, and testing of the CCS and system components. These requirements are in Refs. 7.2.15, 7.2.25, and 7.2.26.
See Section 2.2.13 for requirements applicable to system/component design
( classification.
2.2.13 Codes and Standards and Safety Class The Code of Record (COR) for ASME Section III piping systems designed by TVA is the 1971 edition up to and including the Summer 1973 Addenda (Ref.
7.3.10) .
For those components ordered before the adoption of the Winter 1977 Addenda, the COR is that code in effect on the date the contract with the ultimate manufacturer was placed.
For those components ordered after the adoption of the Winter 1977 Addenda, the owner may establish a COR independent of the contract date and dating as early as three years prior to the date on which the construction permit application was filed. The COR will be established by the applicable ASME Section III design specification. The nonsafety-non safety-related portion of the CCS piping shall comply with the requirements of ANSI B31.1 (Ref. 7.3.11).
The piping and component safety classification for the CCS shall be as defined in WB-DC-40-36 (Ref. 7.2.11).
7
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TRAIN A. 1-47.611-88-1 A-AUTO A-AUTO CHTt.tT CNT",T IISUI SLN PHASE B OPE" /~ -HS~ CLOSE OPEN (~-HS\\
OPEN (~-HS~ CLOSE TRAIN 8 1-47.611-88-1 1-41.,611-88-1 70-140 70-140.4 .... 70-100A 70-10DA
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NOTES: B
- 1. FOR SYW80LS SYU80LS AND GENERAL NOTES SEE 1-47.611-70-1.
2.
2, SW ITCH LOCATED ON 480V SWITCH '80'0' I.iOV WOV BOARD, r------l J.
- 3. l-HS-70-898 1--115-70-898 HAS BEEN REIM)VED C~OITION ADVERSELY REWOVED TO PREVENT A CONDITION AFFECTltOG or AFFECTING THE SAFETY FUNCTION OF THE ASSOCIATED VALVE DURING AN
!: OPEN I CONTROLS T'f1l TVP or OF ACCIDENT.
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PUMP 18-B lB-8
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___ CONTROLS FOR P\M'P\.UP 1 IB-B_
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BARRIERS 2, 2. lAND 3 AND 4 3, AND 4
- 2. J, ELECTRICAL pulA? lA-A TYPICAL OF PUItP T'i'PICAL r------...,
LOGIC DIAGRAM THERMAL BARRIER BOOSTER CONTROLS TYP OF II 1-FCV-70-g0 1-FC\I-70-90 EXCEPT COMPONENT COOLING SYSTEM PUMPS 11A-A A-A & 1B-B 18-8 CHTM' USE CNT I SLN.
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'RAIN A CONT. WATTS BAR NUCLEAR PLANT Q I I TENNESSEE VALLEY AUTHOR I TV TY H
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-~.-r INIT!AL' ISSUE INIT!AL RO I SSUE PER ENG I NEER!
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, GARY BARNARD L-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~*.......
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3c.c. LYKE FOR MeB YeB EXCESS LETDOIIt4 LETDOWN HEAT EXCHANGER FSAR FIG.9.2-25 I SSUED BY:
ISSuED
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K4 PIIllIlSREQ'OR_
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'1 - -
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 8.027 AK2.03 008 Given the following plant conditions:
The plant is operating at 100% power steady state conditions with all controls in automatic.
1-XS-68-340D,PZR PRESS CONTROL CHANNEL SELECT, is in the PT 340 & 334 position Pressurizer backup heater C is energized.
Pressurizer sprays valves are throttling open.
The pressurizer pressure controller, 1-PIC-68-340, fails to 100%.
Which ONE of the following describes the immediate response of the pressurizer pressure control system to this failure?
A. Spray valves go full open, backup heater C de-energizes.
B~ Spray valves go full open, backup heater C remains energized.
B!'"
C. Backup heater banks A-A and B-B energize, spray valves go full closed.
D. Backup heater banks A-A and B-B energize, spray valves continue to throttle.
( DISTRACTOR ANAL YSIS:
A. Incorrect, The pressurizer spray valves will travel to full open but the C Backup heater bank will not de-energize (it will remain energized). Plausible because the spray valves opening is correct and the other backup heaters can de-energize due to increasing controller output.
B. Correct, The pressurizer spray valves will travel to full open and the C Backup heater bank will remain energized. A-A and B-B heaters can automatically de-energize on increasing controller output but the C Bank will not.
C. Incorrect, Backup heaters A-A and B-B would not energize and the spray valve would not go full closed due to the failure. Plausible Plausible that the backup heaters would come on and the spray valve would go full closed if the failure direction were reversed and the transmitter failed to 0% instead of 100%.
D. Incorrect, Backup heaters A-A and B-B would not energize and the spray valve would not continue to throttle due to the failure. Plausible that the backup heaters would come on if the failure direction was reversed and the transmitter failed to 0%
instead of 100% and the applicant concludes the spray continues to throttle due to the pressure increase from the heaters being energized.
Page 21
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 8 S
Tier: _1_
__ Group
_1 1 KIA: 027 AK2.03 Pressurizer Pressure Control System (PZR PCS) Malfunction Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners Importance Rating: 2.6 /2.S 2.6/2.8 10CFRPart55:
10 CFR Part 55: 41.7/45.7 10CFR55.43.b: Not applicable.
KIA Match: Applicant must identify the effect a controller malfunction has on the pressurizer control system.
Technical
Reference:
Drawing 1-47W611-68-3, 1-47W611-6S-3, R8RS AOI-1S, "Malfunction Of Pressurizer Pressure Control AOI-18, System," Rev 0022 Proposed references None
( to be provided:
Learning Objective: -SYS06SC 3-0T-SYS068C 8.
S. Describe the operation of the master pressure controller.
Question Source:
New Modified Bank xX Bank Question History: WBN bank question SYS068C.08 SYS06SC.OS 008OOS modified Comments:
Page 22
Page 1 of 1
(
(
file:IIC:\Documents and Settings\NRC4Test\Desktop\WBN Sim Photos\lM-05\lM05-35.J... 8/26/2009 Photos\lM-OS\lMOS-3S.J... 8/2612009
WBN MALFUNCTION OF PRESSURIZER AOI-18 Unit 1 PRESSURE CONTROL SYSTEM Rev. 0022 4.0 DISCUSSION 4.1 General This procedure covers failures within the pzr pressure control system which could result in abnormal pzr pressure. Since this is a system AOI, failures considered include sensing instrument failures, circuitry failures, controller failures and equipment failures such as stuck open valves. This procedure is designed to allow the operator to gain control of pzr pressure and return the unit to normal operation within Operator capabilities. An example of a condition outside operator capabilities would be the case of a stuck open pzr Safety valve.
4.2 Pressurizer Pressure Control Channels Selector switch 1-XS-68-340D is a 3 position switch which is used to select two of four pzr pressure instruments for control. One channel (normally 340) is used for control of pzr heaters, sprays and PORV 1-PCV-68-340A. The other channel (normally 334) is used for control of PORV 1-PCV-68-334.
Channel 323 may be substituted for channel 340 for control of pzr heaters, sprays and PORV 1-PCV-68-340A.
( Channel 322 may be substituted for channel 334 for control of PORV 1-PCV-68-334.
4.3 Instrument or Control Circuitry Failure Instrument failures can be quickly diagnosed by a channel check. Control responses occur only if a failed instrument is selected for control. Failure of any channel, however, affects the safeguards system capability. Failure of controllers or control circuitry can be compensated for by manual control of heaters and/or sprays or PORVs.
orPORVs.
4.4 Pressurizer PORV or Safety An inadvertently opened or stuck open pzr safety or PORV will result in a rapid drop in RCS pressure, yet pzr level will rise. A failed open PORV can be compensated for by closing the associated block valve. A pzr safety failed open or leaking to such an extent that this procedure is entered would result in a pzr pressure drop in an uncontrolled manner. In this case, a Rx trip and Safety Injection should be manually initiated. This is to prevent boiling in the core and hot legs. The events should be handled as a LOCA with the same criteria for SI operation and termination.
Page 13 of 15
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11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 9. 029 EK1.05 009 With the plant initially at full power, the following conditions are given:
- A spurious turbine trip occurs which resulted in the crew performing FR-S.1 FR-S.1,,
"Nuclear Power Generation 1 ATWS."
- The crew is at Step 20 for checking the reactor subcritical.
Power range channels indicate 2%. Intermediate range startup rate indicates slightly positive. NO charging pumps are available. Which ONE of the following describes: (1) an action which will address the step for reactor subcriticality AND (2) the time in core life when the action would have the greatest effect on reactivity? (1 ) (2) Required Action Time in Core Life A. Manually initiate SI to provide boration. BOL B. Manually initiate SI to provide boration. EOL C. Lower AFW flow to allow RCS to heatup. BOL D'!" Lower AFW flow to allow RCS to heatup. D!' EOL Page 23
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect. It is plausible for an applicant to reason that even though the turbine trip would cause RCS pressure to rise, the steam dumps actuation may lower pressure enough to make manual initiation of SI effective in delivering at least some boration. Further plausibility for this comes from the wording of the description of the Safety Injection Pumps' design, which says "these pumps are for emergency core cooling modes requiring relatively high head/low flow." An applicant may confuse the design shutoff head of the SIPs with the Centrifugal Charging Pumps and believe that initiation of the SI is effective in providing boron to the core. Further, it is plausible for an applicant to reverse the time in core life effect on reactivity change for a given boration, and incorrectly select BOL. They could reason that because of the negative moderator temperature coefficient (MTC), the cooling effect of the steam dumps (on the turbine trip) would have relatively less of an effect on reactivity at BOL vs. EOL, and therefore, addition of more boron from SI would result in more negative reactivity insertion overall at BOL than at EOL, where the negative MTC would have less of a negative reactivity effect. B. Incorrect. Plausible, since the time in core life effect is correct. It is also plausible for an applicant to reason that even though the turbine trip would cause RCS pressure to rise, the steam dumps actuation may lower pressure enough to make manual initiation of SI effective in deliverfng delivering at least some boration. Further (( plausibility for this comes from the wording of the description of the Safety Injection Pumps' design, which says "these pumps are for emergency core cooling modes requiring relatively high head/low flow." An applicant may confuse the design shutoff head of the SIPs with the Centrifugal Charging Pumps and believe that initiation of the SI is effective. C. Incorrect. The applicant is correct in believing that allowing the RCS to heatup is a viable and procedurally driven method of inserting negative reactivity into the core. The applicant has also correctly recognized conditions which indicate that NO boration flow is available by understanding that when the turbine trips, and the is reactor does not, RCS pressure rises quickly, and is not expected to lower to auto SI setpoint. However, the applicant reverses the effect of time in core life on the effectiveness of a boration. D. CORRECT. Allowing the RCS to heatup is a viable and procedurally driven method of inserting negative reactivity into the core. The applicant has also correctly recognized conditions which indicate that NO boration flow is available by understanding that when the turbine trips, and the reactor does not, RCS pressure rises quickly, and is not expected to lower to auto SI setpoint. Per standard reactor physics principles, at EOL the moderator temperature coefficient is more negative than at BOL. Page 24
11/2009 Watts 11/2009 Watts Bar Bar RO RO NRC NRC Exam Exam -- As As submitted submitted 10/2/2009 10/2/2009 Question Number: Question Number: 99 Tier: _1_ _1 __ Group 11 KIA: 029EK1.05 029EK1.05 ATWS ATWS Knowledge of Knowledge of the the operational implications of of the the following following concepts as they they apply to apply to the the ATWS: of negative temperature coefficient as applied to large PWR Definition of coolant systems. Importance Rating: 2.8 /I 3.2 10 CFR Part 55: 41.8 I 41.10 I 45.3 41.8/41.10/45.3 10CFR55.43.b: Not applicable. KIA Match: Applicant must apply the concept of the effect of a negative moderator temperature coefficient to a plant condition involving post-trip conditions and an ATWS, and also evaluate the operational implication of this condition; i.e., how to operate the plant to begin mitigating the condition. Technical
Reference:
FR-S.1, "Nuclear Power Generation I/ AATWS," TWS," step 20, Rev. 19 System Description for CVCS, and Safety Injection Proposed references None to be provided: Learning Objective: 3-0T-FRS0001, "Function Restoration Guidelines, FR-S.1 &S.2." FR-S.1&S.2." Obj. 9. Given a set of plant conditions, use FR-S.1, FR-S.2 and the Critical Safety Function Status Trees to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages Notes and Cautions. Question Source: New X Modified Bank Bank Question Question History: History: New question question Comments: Comments: Page 25 Page 25
The most severe plant conditions that could result from-a from a loss of load occur fo 11 owi ng a turbi following ne trip turbine tri p from full power when the turbi ne trip turbine tri pis is caused by a ( loss of condenser vacuum. Since the main feedwater pumps may be turbine driven with steam exhaust to the main condenser, loss of feedwater may also result from a loss of condenser vacuum. For this reason, the low feedwater flow coincident with low steam generator water level reactor trip and the low-low steam generator water level trip are included in the above listing. The pressurizer safety valves and steamsteam generator safety valves are sized to protect the ReS and steam generator against overpressure for all load losses without assuming the operation of the steam dump system, pressurizer spray, pressurizer power-operated relief valves, steam generator power-oper&ted power-operated relief valves, automatic rod control, or direct reactor trip on turbine trip. That is, the steam relief capacity of the pressurizer safety valves is selected to match the maximum pre~surizer pressurizer insurge following a turbine trip insurge~following without credit for the items mentioned above. The steam generator safety valve relief capacity is sized to remove the steam flow at the Engineered Safeguards Design rating (-105 percent of the steam flow at rated power) from the steam generator without exceeding 110 percent of the steam system design pressure. The pressurizer safety valve capacity is sized for a complete loss of heat sink with the plant initially operating at the maximum ~alculated calculated turbine load and with operation of the steam generator safety valves. The pressurizer safety valves are then able to maintain the ReS pressure to within 110 percent of the ReS Res design pressure without direct or immediate reactor trip action. Analysis Of Loss Of Electrical Load and/or Turbine Generator Trip Without Reactor Trip In this ATWS transient, plant behavior is evaluated for a turbine trip and loss of main feedwater occurring from full power with the assumption that the control rods fail to drop into the core following generation of a reactor trip signal. The evaluation shows the effectiveness of ReS Res pressure-relief devices and the extent of any approach to the core safety limits. The following assumptions are made in the analysis: FR-S.1 Background FR-S.l 10 HP-Rev. 2, 4/30/2005 HFRS1BG.doc HFRSIBG.doc
o Initial normal full power operation early in core life. Since the negative temperature coefficient of reactivity reduces core power as the coolant temperature rises, and the temperature coefficient becomes more negative with core life, the theATWS ATWS loss of load is less severe later l~ter in core 1life. ife. o Normal operation of the following control systems:
- 1) Pressurizer pressure control, including heaters, spray, and both the PORVs and the safety relief valves.
- 2) Turbine governor valves in impulse pressure control prior to trip, and valve closure on turbine trip.
o Loss of condenser vacuum at t = O. o No automatic reactor trip. o No automatic control rod insertion as reactor coolant temperature rises. ( o Main feedwater flow falls to zero in the first four seconds of the transient, with no main feedwater flow after that time. This is consistent with a steam driven main feedwater system and a loss of condenser vacuum. o Auxiliaryfeedwater Auxiliary feedwater flow begins at 60 seconds at a rate of 1760 gpm. o Auxiliary feedwater is injected into the feedwater pipe at a temperature of 130°F, upstream of the steam generator, such that the cooler water enters the steam generator after the feedwater header is purged of main feedwater (440°F). o Primary to secondary heat transfer area is reduced as the steam generator shell-side water inventory drops below the level necessary to wet the tubes. FR-S.l Background 11 HP-Rev. 2, 4/30/2005 HFRSIBG.doc
WBN NUCLEAR POWER GENERATION I A TWS FR-S.1 Rev 19 IStep I IActionlExpected Response I IResponse Not Obtained
- 19. CHECK incore TICs IF incore TICs are less than 1200°F. greater than 1200°F AND rising, THEN
'S-titM CJn\~atCl"f\\" Q(~
'S-t~ C§\'\~nt~ <l.A"~ ** GO TO SACRG-1, 1~ l\S
-t~ f\S ~9'\~
~8\~ bos 005" Severe Accident Control Room
~
CU"(. QJlJo..i\~Iq.
<lI.lo...i\~Iq,. , 'I \\ .* Guideline Initial Response.
- 20. CHECK reactor subcritical: CONTINUE to borate.
- a. Power range channels IF boration is NOT available, less than 5%.
THEN
- b. Intermediate range startup rate ALLOW RCS to heat up to insert NEGATIVE. negative reactivity from temperature coefficients.
IF red OR orange condition exists on ( other Status Trees, THEN PERFORM actions of other FR Procedures which do not cool down or otherwise add positive reactivity to the core.
** GO TO Step 4.
10 of 15
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 10. 038EA2.11 010 With the plant initially at 80% power, a single tube completely ruptures on #3 SG. The crew has entered E-3, "Steam Generator Tube Rupture" and is at the step for identifying which SG is ruptured.
Which method will be used for determining which SG is ruptured and why? A'I RADCON surveys of the main steam lines, since the Main Steam Line Post A"! Accident Monitors will not be effective due to lack of N16 response. B. RADCON surveys of the main steam lines, since the ability to perform a Chemistry sample is no longer available. C. Main Steam Line Radiation Monitors high range indication, since the low range has been disabled. D. Main Steam Line Radiation Monitors high range indication, since they have a high level of sensitivity to N16. DISTRACTOR ANAL YSIS: A. Correct. Since the crew has already entered E-3, then the reactor has already ( been tripped. As detailed in SOER 93-1, "Palo Verde Tube Rupture," the N16 response (short half-life) after a reactor trip renders a main steam line monitor indication ineffective in diagnosing which SG is affected. E-3 provides guidance in employing RADCON surveys of the main steam lines as aa method of diagnosing which SG is affected. For the given plant conditions, this method will be effective, and use of the Main Steam Line Radiation Monitors will NOT be effective. B. Incorrect. Plausible, since RADCON surveys is the correct method. It is also plausible for the applicant to be confused by believing that an isolation would have occurred that somehow has removed the capability to obtain a liquid sample of the affected S/G. C. Incorrect. Plausible, since part of the distractor is correct: the low ranges for the Main Steam Line Radiation Monitors have indeed been disabled. Also, plausible due to the title of the listed radiation monitor: "Main Steam Line Radiation Monitor. " However, the applicant fails to recognize plant conditions which render these monitors ineffective due to the response of N16 following a reactor trip. D. Incorrect. Plausible due to the title of the listed radiation monitor: "Main Steam Line Radiation Monitor." Also plausible, since the low ranges have indeed been disabled. However, the applicant fails to recognize plant conditions which render these monitors ineffective due to the response of N16 following a reactor trip. Page 26
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 10 Tier: _1_ Group 1 KIA: 038EA2.11 Steam Generator Tube Rupture (SGTR) Ability to determine or interpret the following as they apply to a SGTR: Local radiation reading on main steam lines. Importance Rating: 3.7 1 3.9 3.71 10 CFR Part 55: 43.5 I/ 45.13) 10CFR55.43.b: Not applicable. KIA Match: The applicant is tested on how radiation readings are used for determining which Steam Generator is ruptured. Technical
Reference:
FSAR 15.4.3, "Steam Generator Tube Rupture;" Operator Action 1. Lesson Plan, 3-0T-SYS-090A, "Radiation Monitoring System" SOER 93-1, "Palo Verde Tube Rupture" ( ARI 265-C, "SG1 DISCH RE-421 RAD HI," Rev. 6 E-3, "Steam Generator Tube Rupture," Rev. 22 Proposed references None to be provided: Learning Objective: 3-0T-EOP0300, Obj. 2. Given a set of plant conditions, the operator will be able to identify which SGs, if any, are ruptured by evaluating the symptoms of a ruptured SG. Question Source: New X Modified Bank Bank Question History: New question Comments: Page 27
WBN STEAM GENERA TOR TUBE RUPTURE E-3 Rev 22 IStep I IAction/Expected Response IIIIResponse Not Obtained I
- 3. IDENTIFY Ruptured S/G CONTINUE efforts to identify .
based on ANY of the following: Ruptured S/G.
- Unexpected rise in S/G NR level. NOTIFY RADPROT and Chemistry.
OR WHEN Ruptured S/G identified, S/G discharge monitor high THEN radiation. PERFORM Steps 4 thru 8. OR
** GO TO Step 9.
RADPROT Survey. OR
- Chemistry sample.
- 4. ENSURE RupturedS/G PORV aligned:
- a. ENSURE controller in AUTO
. set at 90%.
(
- b. ENSURE HS in P-AUTO.
- c. WHEN Ruptured S/G pressure less than 1130 psig.
THEN
- 1) ENSURE Ruptured S/G PORV CLOSED, OR
- 2) OBTAIN RADPROT support and Locally CLOSE Ruptured S/G isolation valve:
- Loop 1, 1-ISV-1-619
[South Valve Room].
- Loop 2, 1-ISV-1-620
[North Valve Room].
- Loop 3, 1-ISV-1-621
[North Valve Room].
- Loop 4, 1-ISV-1-622
[South Valve Room]. 4of4?
265-C SOURCE SETPOINT 1-RE-90-421 3.5 x 1022 mRlhr . SG SG11 DISCH RE-421 RAD HI Probable Cause: 1. Steam Generator tube leak or rupture with high RCS activity or potential fuel failure
- 2. Loss of power to microprocessor Corrective Action: [1] CHECK 1-RM-90-119, 1-RM-90-120 and 1-RM-90-121.
[2] NOTIFY RADIOLOGICAL PROTECTION to investigate alarm. [3] NOTIFY Chemistry to initiate sampling. [4] REFER TO AOI-33, STEAM GENERATOR TUBE LEAK.
References:
1-45W600-55-46 1-47W61 0-90-5 AOI-33 ARI-265-268 WBN Page 6 of 17 Rev. 6 (
WBNP-O Operator actions are described below. I. Identify the ruptured steam generator. High secondary side activity, as indicated by the condenser vacuum exhaust radiation monitor, steam generator blowdown liquid monitor, or main steam line radiation monitor, typically will provide the first indication of an SGTR event. The ruptured steam generator can be identified by an unexpected increase in steam generator narrow range level, a RADCON survey, or a chemistry laboratory sample. For an SGTR that results in a reactor trip at high power, the steam generator water level as indicated on the narrow range scale will decrease significantly for all of the steam generators. The auxiliary feedwater flow will begin to refill the steam generators, distributing flow to each of the steam generators. Since primary to secondary leakage adds additional liquid inventory, to the ruptured steam generator, the water level will increase more rapidly than normally expected in that steam generator. This response, as displayed by the steam generator water level instrumentation, provides confirmation of an SGTR event and also identifies the ruptured steam generator.
- 2. Isolate the ruptured steam generator from the intact steam generators and isolate feedwater to the ruptured steam generator.
Once the steam generator with a tube rupture has been identified, recovery ( actions begin by isolating steam flow from and stopping feedwater flow to the ruptured steam generator. In addition to minimizing radiological releases, this also reduces the possibility of overfilling the ruptured steam generator with water by I) minimizing the accumulation of feedwater flow and 2) enabling the operator to establish a pressure differential between the ruptured and intact steam generators as a necessary step toward terminating primary to secondary leakage.
- 3. Cool down the RCS using the intact steam generators.
After isolation of the ruptured steam generator, the RCS is cooled as rapidly as possible to less than the saturation temperature corresponding to the ruptured steam generator pressure by dumping steam from only the intact steam generators. This ensures adequate subcooling will exist in the RCS after depressurization of the RCS to the ruptured steam generator pressure in subsequent actions. If offsite power is available, the normal steam dump system to the condenser can be used to perform this cooldown. However, if offsite power is lost, the RCS is cooled using the steam generator power operated relief valves to release steam from the intact steam generators. l 15.4-46
3-0T-SYS090A Revision 13 Page 29 of74 I \
- x. LESSON BODY INSTRUCTOR NOTES
- e. SG Tube Leak/Rupture indication-Although not designed to monitor Monitors are designed for post gamma from N 16, the monitors can accident monitoring of fission detect primary-to-secondary leaks at fragments which would indicate power, however not considered reliable. cladding failure.
Since N16 has a short half-life (7.13 80 16 16 en, N16 (n, p) 77N16 ~ sec), the SG Disch Monitors won't indicate tube leaks after reactor With ~isablin~the e disabling 0 the low shutdown. The following event (f.) range detec ors, the S/G detectors, demonstrates the problem with having Discharge monitors cannot be the SG Disch monitors as the only considered a reliable source for radiation indication following Rx Trip detection of a Steam Generator and diagnosing a SGTR in E-O. Tube Leak.
- f. 3114/93, Palo Verde Tube Rupture, 3/14/93, SOER 93-1 Operating at 98% power.
(1) The operating crew manually Auto S1 resulted shortly tripped Rx due to decreasing thereafter. PZR level from a suspected SG Tube Leak. (2) The only radiation indication Reading was 5-7 times low due was from a SG Disch monitor in to discriminator circuit failure. alarm. The Condenser Vacuum exh monitor was thought to be inop, and the SG Blowdown Blowdown isolated after S1 also. monitor did not indicate hi rad because the rupture was high in the tube bundle, causing a long delay for the contaminated water
- Dfst\o..ckr
?\~s;b~l
?\~s~b~l Dtst\o..dor
;1j fo-r C .
to reach the blowdown. (3) The crew did not initially diagnose a Tube Rupture in E-O since the SG Disch monitor had cleared alarm. The Diagnostics required Hi rad for transition. (4) Not until a Function Restoration Procedure directed the crew to un-isolate the blowdown rad monitors, was proper diagnosis made. This caused a 53 min delay before the proper procedure was implemented.
3-0T-SYS090A Revision 13 Page 30 of74 ( x. X. LESSON BODY INSTRUCTOR NOTES (5) At WBN, E-O also uses radiation See RM-23 Readout Module SeeRM-23 monitors as the only means of below for indications after a SG diagnosing a SGTR. It is Discharge Rad High alarm clears.
-Discharge important to note that after Rx shutdown, the SG Disch monitors will go out of alarm due to N 16 decay.
- 2. RM-80 Microprocessor - The database for 4 of these, 1 per detector this System.
- a. Functions:
(1) Store alarm setpoints. (2) Retain memory of previous readings. (3) Determines problems in system. (4) Receives data from detector and sends it to the RM-23 in MCR.
- b. Location- RM-80s on EI 737' Aux Bldg:
(1) 2 near exit from Additional Equip Bldg. (2) 2 near entrance to South Vault Rm. (3) Each RM-80 has local indicating light that is normally lit indicating that the RM RM-80-80 is operable.
- 3. RM Readout Module OTT.:.
OTT-88 Objective 15
- a. Location- MCR on 1-M-30. Each SG Discharge Monitor is recorded M-31.
M -31.
- b. Purpose- receives output ofRM-80 and Purpose':' The auto selector has been set to displays digital reading (10 1 to 107 zero which means it will only mRlhr when in MRlHR Auto or High indicate the high range.
Range).
- c. Hi-rad alarm at 350 mrlhr (Red alarm light lit).
- d. Common malfunction alarm for all d.*
1-M-30 monitors on 0-M-12 window 176E, PAS MaNMON Pnl M30/M31 Instr Malf.
11/2009 Watts Bar 11/2009 Watts Bar RO RO NRC NRC Exam Exam -- As As submitted submitted 10/2/2009 10/2/2009
- 11. 040
- 11. 040 AA1.23 AA1.23 011 011 With the With the plant plant initially initially at at full full power power aa steam steam line line break break occurred.
occurred. The The crew crew is is currently currently at Step at Step 3 of of E-2, E-2, "Faulted "Faulted Steam Steam Generator Generator Isolation" Isolation" and and has has determined determined that that SG SG 1 and 2 areare intact. intact. For For SGs SGs 3 and and 4, the the crew crew is evaluating the following following parameters: 1-PI-1-20A (SG 3) 1-PI-1-20A 3) indicates 470 psig and lowering. 1-PI-1~27A (SG 4) 1-PI-1-27A 4) indicates indicates 440 psig andand lowering.
- #3
#3 RCS Thot Thot = = 454°F
- #4 RCS Thot = 456°F
- All Core Exit Thermocouples are 460°F
- #3 RCS Tcold = 453°F.
- #4 RCS Tcold = 410°F.
- All Reactor Coolant Pumps are in service.
Which ONE of the following describes the status of Steam Generator 3 and 4? Intact ? SG 3 Intact? SG 4 Intact ? A'! A"! YES NO B. YES YES C. NO NO ( D. NO YES Page 28 Page 28
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. CORRECT. E-2, Step 3 guidance for determining if the SG is intact is for the SG pressure to be greater than P-sat for RCS incore temperature. RCS incore temperature is given as 460 00 F. P-sat for 460 00 F is 467 psia. Therefore, SG 3 is intact since its pressure is 470 psig (485 psia), but SG 4 is NOT intact since its pressure is 440 psig (455 psia) , less than P-sat for RCS incore temperature. B. Incorrect. Plausible, if applicant uses the incorrect RCS temperature value, in this case - use of the associated loop 4 Thot value of 45600 F. The P-sat for this temperature is 448 psia. With SG 4 pressure at 440 psig (455 psia), it is plausible, but incorrect, for the applicant to conclude that SG 4 is also intact. C. Incorrect. Plausible, if applicant believes that a lowering pressure is indicative of a faulted SG, in this case, believing that both SG 3 and 4 are not intact. D. Incorrect. Plausible, if applicant misinterprets the given parameters, including a failure to understand the implications of the Core Exit Thermocouple temperatures in relation to Steam Generator (SG) parameters. In this case, the applicant looks at these parameters (temperature and pressure), uses the Steam Tables to calculate subcooling/saturation sub cooling/saturation conditions for SG 3, as compared to SG 4, and ( misapplies those results to make a determination on which SG is intact. If the applicant incorrectly uses Thots to make this determination, it will result in determining that with Loop 3 Thot at 454°F, 454 0 F, and SG 3 pressure at 485 psia, the difference between SG 3's pressure and its associated loop saturation pressure of 448 psia) is 37 psi. Using a similar methodology for SG 4, the applicant determines that with Loop 4 Thot at 456°F, and SG 4 pressure at 455 psia, the difference between SG 4's pressure and its associated loop saturation pressure (450 psia) is only 5 psi. The applicant misuses these calculations and mistakenly concludes that SG 4 is intact and SG 3 is not intact. Page 29
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 11 Tier: _1_ _1 __ Group 1 KIA: 040AA 1.23 Steam Line Rupture Ability to operate and / or monitor the following as they apply to the Steam Line Rupture: All pressure gauges per steam generator (for pressure drop) Importance Rating: 3.6 / 3.5 3.6/3.5 10 CFR Part 55: 41.7/45.5/45.6 10CFR55.43.b: Not applicable. KIA Match: Applicant is presented with an array of pressure indication values for two of the steam generators and must then interpret these values to determine the status of the steam generators in the context of a Steam Line Rupture event. Technical
Reference:
Steam Tables E-2, "Faulted Steam Generator Isolation," Step 3 (Determine intact SGs.) Rev. 11 ( Proposed references None to be provided: Learning Objective: 3-0T-EOP0200, "Faulted Steam Generator Isolation"
- 5. Given a set of plant conditions, use procedure E-2 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes and Cautions.
Question Source: New x X Modified Bank Bank Question History: New question Comments:. Comments: ( Page 30
WBN GENERA TOR ISOLATION FAULTED STEAM GENERATOR E-2 Rev 11 IStep I IAction/Expected Response I IResponse Not Obtained CAUTION for RCS cooldown, it should If a faulted SIG is NOT needed forRCS remain isolated during subsequent recovery actions.
- 1. ENSURE all MSIVs and Manually CLOSE valves.
MSIV bypasses CLOSED. IF valves can NOT be closed, THEN Locally REMOVE power to valves:
- DISPATCH NAUO to perform Attachment 1 (E-2).
NOTE If it is known that a steam leak exists in the Turbine building, the following step should not be performed until the affected steam header is depressurized.
- 2. steam dump controls OFF:
PLACE steam
- 1-HS-1-103A, STEAM DUMP FSV"A".
- 1-HS-1-1038, STEAM DUMP
. FSV"8".
- 3. CHECK for at least one Intact S/G: IF pressure in all four S/Gs dropping uncontrolled,
- Any S/G pressure THEN controlled or rising, ** GO TO ECA-2.1, ECA-2.1 ,
Uncontrolled Depressurization of OR All Steam Generators.
- Any S/G pressure greater than P-sat for RCS incore temperature.
( 3 of 12
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 12. 054AK3.03 012 Due to the loss of the running MFPT, Unit 1 is operating at 12% power with the Standby Main Feedwater Pump ~ervice.
IN IIJ The following occurs:
- The Standby Main Feedwater Pump trips.
- The crew implements AOI-16, "Loss of Normal Feedwater," and initiates action to stabilize the plant within AFW pump capability.
- Steam Generator Narrow Range levels are slowly trending up and are currently:
#1 - 27%, #2 - 25%, #3 - 28%, #4 - 31 %.
In accordance with AOI-16, which ONE of the following set of conditions would direct the operators to manually control AFW flow and the reason why? Conditions Reason A. Tavg 554°F and rising with To restore minimum heat sink AFW flow 100 gpm to each SG. conditions. B. Tavg 554°F and rising with To terminate the RCS heatup. AFW flow 120 gpm to each SG. ( C. Tavg 555°F and dropping with To restore minimum heat sink AFW flow 100 gpm to each SG. conditions. D~ Tavg 555°F and dropping with D!' To minimize the RCS cooldown AFW flow is 120 gpm to each SG. Page 31
11/2009 Watts Bar RO NRC Exam - As submitted 101212009 10/2/2009 \ DISTRACTOR ANAL YSIS: A. Incorrect, with Tavg 554°F 554 of and rising, the temperature is still below the AOI-16 required 557°F so continued heatup is needed and minimum heat sink is not a concern for the current conditions even though the AFW flow is below the 410 gpm because one steam generator is above the minimum level. Plausible because if the steam generator level had been below the minimum then minimum heat sink would have been a concern and the flow could be increased B. Incorrect, with Tavg 554°F 554 of and rising, the temperature is still below the AOI-16 required 557°F 55JDF so continued heatup is needed. Do not need to throttle AFW flow to stop the heatup. Plausible because the flow would be throttled if the temperature had been higher. C. Incorrect, with Tavg 555°F and dropping, AOI-16 directs AFWflow be controlled to minimize the cooldown (not raised to restore minimum heat sink conditions). Minimum heat sink conditions are present due to one SG level greater than 29%. Plausible because the temperature lower than 557°F 55JDF requiring the manual control of AFW to stop the cooldown is correct but restoration of the minimum heat sink is not required. D. Correct, if Tavg is not trending to greater than 557°F, 55JDF, AOI-16 directs AFWAFWflow flow be controlled to minimize the cooldown. Page 32
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 12 Tier: _1_ Group 1 KIA: 054 AK3.03 Loss of Main Feedwater (MFW) Knowledge of the reasons for the following responses as they apply to the Loss of Main Feedwater (MFW): Manual control of AFW flow control valves Importance Rating: 3.8 /4.1 3.8/4.1 10 CFR Part 55: 41.5,41.10/45.6/45.13 10CFR55.43.b: Not applicable. KIA Match: Applicant is required to identify from stated conditions when the AFW level control valves would be placed to manual and the reason for placing the valves in manual. Technical
Reference:
AOI-16, "Loss of Normal Feedwater," Rev 30 ( Proposed references None \ to be provided: Learning Objective: 3-0T-AOI1600 3-0T-AOJ1600
- 7. Demonstrate ability/knowledge of AOI, AOJ, by correctly:
- a. Recognizing Entry conditions
- b. Responding to Action steps
- c. Responding to Contingencies (RNO)
Question Source: New X Modified Bank Bank Question History: New question Comments: Page 33
AOI-16 WBN LOSS OF NORMAL FEEDWATER Revision 30 of 27 Page 6 of27 3.2 Loss of Standby MFWP Without Main Turbine In Service ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 1. IF loss of S/G level is imminent, THEN TRIP reactor, and
** GO TO E-O, Reactor Trip or Safety Injection.
- 2. CHECK at least one MFW pump ** GO TO Step 5.
in service.
- 3. ENSURE adequate feed flow Manually CONTROL MFWPT for existing conditions: speed.
- Feed flow greater than or equal to steam flow.
- S/G levels returning to program.
- 4. ** GO TO Step 10.
- 5. START AFW pumps:
a.* MO AFW pumps. Both MD
- b.
- TO TD AFW pump.
CAUTION Care should be taken to avoid excessive rod insertion. If resulting Reactor power is less than desired, rods SHALL NOT be withdrawn in an attempt to recover Reactor power.
- 6. INSERT control rods to reduce reactor power to within AFW capabilities (less than 4%).
- 7. MAINTAIN zero startup rate.
AOI-16 WBN LOSS OF NORMAL FEEDWATER Revision 30 Page 7of27 7 of 27 3.2 Loss of Standby MFWP Without Main Turbine In Service (Continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 8. CHECK AFW pumps LCVs Manually CONTROL AFW controlling S/G levels to 38%. LCVs to maintain S/G levels at 38%.
- 9. STABILIZE RCS Tave greater than ENSURE steam dumps and S/G or equal to 557°F. PORVs CLOSED.
IF cooldown continues, THEN CONTROL AFW flow to minimize cooldown.
- 10. WHEN S/G levels controlled, THEN CONSULT plant staff for recovery actions.
- 11. REFER TO appropriate instruction:
- a.
- SPP-3.5, Regulatory Reporting Requirements.
- b.
- GO-3, Unit Startup From Less Than 4% Reactor Power To Less Than 30% Reactor Power.
- c.
- GO-5, Unit Shutdown From Less Than 30% Reactor Power To Hot Standby.
- END OF SUBSECTION -
11/2009 Watts Bar RO NRC Exam - As submitted 11/2009 10/2/2009 10/2/2009
- 13. 055 EA2.06 013 Given the following:
In response to a station blackout event the decision was made to restore power by backfeeding from Unit 2 in accordance with AOI-40, "Station Blackout." During alignment the operators determine that 6.9kV Unit Board 1B has overcurrent relays that operated when the blackout occurred. Which ONE of the following identifies the system power supply that will be aligned to establish the backfeed from Unit 2 and the action required to reset 6.9kV Unit Board 1B overcurrent lockout? Power Supply Lockout Relay ... A. 161 kV system must be manually reset locally at the unit board. B. 161 kV system can be reset by placing the boards normal and alternate feeder control switches on 1-M-1 to TRIP. C!" C~ 500kV system must be manually reset locally at the unit board. D. 500kV system can be reset by placing the boards normal and alternate feeder control switches on 1-M-1 to TRIP. Page 34
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, When backfeeding from Unit 2, the transformers are not connected to the 161 kV switchyard to provide the backfeed but the overcurrent relay lockout must be reset locally at the 6.9kV Unit Board. Plausible because the 161kV system is the normal source of offsite power and local manual action being required to reset the lockout relay is correct. B. Incorrect, When backfeeding from Unit 2, the transformers are not connected to the 161 kV switchyard to provide the backfeed and the overcurrent relay lockout can not be reset by placing the normal and alternate feeder breaker control switches to the trip position (the overcurrent causes the actuation of a lockout relay that must be manually reset locally at the 6.9kV Unit Board.) Plausible because the 161kV system is the normal source of offsite power and placing control switches to trip does reset the white light and some switches have reset functions associated with being positioned. C. Correct, The Unit 2 transformers are connected to the 500kV switchyard to provide the backfeed and the overcurrent relay causes the actuation of a lockout relay that must be manually reset locally at the 6.9kV Unit Board. D. Incorrect, The Unit 2 transformers being connected to the 500kV switchyard to provide the backfeed is correct but the overcurrent relay lockout was not reset by placing the normal and alternate feeder breaker control switches to the trip position (the overcurrent causes the actuation of a lockout relay that must be manually reset locally at the 6.9kV Unit Board. Plausible because the correct transformers are identified and placing control switches to trip does reset the white light and some switches have reset functions associated with being positioned. Page 35
11/2009 Watts Bar RO NRC Exam - As submitted 11/2009 10/2/2009 10/2/2009 Question Number: 13 Tier: _1_ Group 1 KIA: 055 EA2.06 Loss of Offsite and Onsite Power (Station Blackout) Ability to determine or interpret the following as they apply to a Station Blackout: Faults and lockouts that must be cleared prior to re-energizing buses Importance Rating: 3.71I 4.1 3.7 10 CFR Part 55: 43.5 I 45.13 10CFR55.43.b: Not applicable. KIA Match: Given a condition where the operator had taken an action related to the event, the applicant is required to identify the status of an overcurrent lockout prior to re-energizing the board that had been tripped by a fault. Technical
Reference:
AOI-40, "Station Blackout," Rev 10 Drawing 1-45W721-1, R14 Proposed references None to be provided: Learning Objective: 3-0T -SYS040A 3-0T-SYS040A
- 4. Define in sequence the steps required to back feed the 6.9kV SID boards from Unit 2 500kV switchyard.
Question Source: New xX Modified Bank Bank Question History: New question Comments: Page 36
AOI-40 WBN STATION BLACKOUT Revision 10 Page 9 of 33 4.0 DISCUSSION Analysis of the vital battery capacity, for the duration of the postulated 4 hour Station Black Out (SBO) event, shows that the batteries are capable of powering all loads necessary to obtain and maintain the unit in the Safe Shutdown condition throughout the entire event, provided that the nonessential loads are shed from the emergency DC lighting cabinets to extend the useful life of the batteries. Nonessential loads are shed from the 250V Station Batteries as soon as possible after the event in order to ensure sufficient capacity to support breaker operation for restoration of AC power to the Shutdown Boards from the 500Kv system. The nonessential loads are listed on Appendices A and B B to this instruction. The instruction and associated appendices are referenced from ECA-O.O, Loss Of Shutdown Power, or may be performed at the discretion of the operator when plant conditions indicate its use. A SBO is considered to exist on the applicable unit when a Loss Of Offsite Power (LOOP) has occurred with subsequent failure of both trains of onsite Emergency AC Power (failure of both EDGs associated with a single unit). ( Accountability for personnel involved in the performance of this procedure should be maintained by the Shift Manager or the Unit SRO in order to allow completion of the procedure to continue uninterrupted if the accountability siren sounds. Steps to initiate backfeed of the Unit 1 Shutdown Boards from the 500 Kv system via the Unit 2 Main Bank and Unit Station Service Transformers are intended to be carried out in sufficient time to complete the alignment within the first 4 hours and when it has been determined that the DIG or offsite power supplies cannot be restored within this 4 hour period. Power restoration efforts should not be limited to securing only one source but should be pursued concurrently. This method contributes to the reduction in the probability of core damage during a major event. All personnel required to respond to this instruction are either stationed onsite on a 24 hr. basis or can report onsite in time to complete performance of this instruction within the coping period for a SBO event (four hours).
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- a. FREQUENT AUTO RESTARTS WHICH COULD DAMAGE t.<<:lTOR ARE PREVENTED BY TEIoIP
- 8. TEt.AP SENSORS IN THE WINDINGS or Of THE CONDENSATE SOOSTER PUMPS AND HEATER DRAIN PUMPS' I S. AUXILIARY OIL all PUt.\PS PUMPS ARE STARTtO BY HEATER DRAIN PUMP START ST"RT SIGNAL.
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- 3. 8US DIFFERENTIAL ,!,NO
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- 4. THE LEVEL lEVEL SWITCH IN THE HorrELL HOTWEll ANNUNCIATES IN THE CONTROL ROOM ON LOW 11. BREMERS 1114 II, BREAKERS I1H .. '112 HAVE INTERLOCKING RELAYS ON A. 1112 ffi LEVEL so SO THAT THE OPERATOR MAY SAFELY SHUTDOWN THE HOrwELL HOTWELl PUMP.PUMP, DESIGNATES MECH,I,NICAL tr.IECHANICAL AND ELECTRICAL APPROVAL 7.5Nt504THROUGH75N15D9.
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11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 14. 057 AG2.1.23 014 Given the following plant conditions:
- Plant is operating at 100% power.
- A loss of 120V AC Vital Instrument Power Board 1-1 occurred.
- The crew implements AOI-25.01, "Loss of 120V AC Vital Instrument Power Boards 1-1 and 2-1."
Which ONE of the following identifies the Main Feedwater Reg Valve controller(s) that will not function in automatic control and why the AFW pump 1A-A is placed in PULL-TO-LOCK during performance of the procedure? Main Feed Reg. AFW pump 1A-A is placed Valve controller(s) in PULL-TO-LOCK ... A. Only #1 to prevent injection of cold water into SG #1 and # 2. B:' Only #1 because the Pressure Control Valve is failed closed. C. Both #1 & #2 to prevent injection of cold water into SG #1 and # 2. D. Both #1 & #2 because Pressure Control Valve is failed closed. Page 37
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, Main feedwater regulating valve #1 is the only regulating valve that will not function in automatic but the AFW Pump 1A-A 1A-A is not placed in P- T-L to prevent injecting cold water into the steam generator. It is because the pressure control valve on its discharge would be failed closed. Plausible because only the #1 Reg not functioning in auto is correct and the LCVs for the motor driven pump do fail open and cold water would be injected if not for the pressure control valve. B. Correct, As identified in A AOI-25.01 01-25. 01 the only main feedwater regulating valve that will not function in automatic is the #1 main feedwater regulating valve and the AFW Pump 1A-A is place in P- T-L because the pressure control valve on its discharge P-T-L would be failed closed. C. Incorrect, Main feedwater regulating valve #2 would not be prevented form automatic function due a loss of the 125v AC Vital Power Board #1 and the AFW 1A-A is not placed in P-Pump 1A-A P-T-L T-L to prevent injecting cold water into the steam generator. It is because the pressure control valve on its discharge would be failed closed. Plausible because the AFW controllers for both #1 and #2 SG level controller are affected and if the pump started, cold water would be injected if not for the pressure control valve. D. Incorrect, Main feedwater regulating valve #2 would not be prevented from ( automatic function due a loss of the 125v AC Vital Power Board #1 but the pressure control valve on the discharge would fail closed. Plausible because the AFW controllers for both #1 and #2 SG level controller are affected and the closed.... pressure control valve on its discharge does fail closed Page 38
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 14 Tier: _1_ Group _1_ _1_ 1 KIA: 057 AG2.1.23 Loss of Vital AC Electrical Instrument Bus Ability to perform specific system and integrated plant procedures during all modes of plant operation. Importance Rating: 4.3 / 4.4 4.3/4.4 10 CFR Part 55: 41.10/43.5/45.2/45.6 10CFR55.43.b: Not applicable. KIA Match: Applicant is presented with a plant conditiion involving a loss of a vital AC electrical instrument bus, and must demonstrate knowledge of the effects of this failure, and recognizing a procedural action used for mitigation. Technical
Reference:
AOI-25.01, "Loss of 120V AC Vital Instrument Power Boards 1-1 and 2-1," Rev 27 Proposed references None to be provided: Learning Objective: 3-0T-AOI2500 3-0T -AO 12500
- 4. Demonstrate ability/knowledge of AOI, by:
- a. Recognizing Entry conditions
- b. Responding to Actions
- c. Responding to Contingencies (RNO)
- d. Responding to Notes/Cautions Question Source:
New Modified Bank x X Bank Question History: WBN bank question AOI-2500.01 017 modified Comments: Both parts of the stem Question Statement replaced with alternate questions. All choices changed. Correct answer relocated. Page 39
AOI-2S.01 AOI-25.01 WBN LOSS OF 120V AC VITAL INSTRUMENT Revision 27 POWER BOARDS 1-1 AND 2-1 Page 6 of 24 3.2 OPERATOR ACTIONS (RHR letdown NOT in service) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION 1 Channel I feedwater and steam flow inputs to feedwater reg valve controls are lost. All reg valve controllers should be monitored and placed in MANUAL if required until power is restored or conditions stabilize to allow transfer to alternate instrument inputs. Loop 1 controllers will not function in AUTO until Channel I power is restored. 44 CAUTION 2 Borated water will be injected into the RCS due to 1-LCV-62-132 VCT outlet valve valve closing and 1-LCV-62-135 1-LCV-62-13S and 136 opening from the RWST due to separation relay de-energizing. NOTE Control board VCT level indication, 1-Ll-62-129, is lost. Use computer point, L0112A, for 1-U-62-130. 1-Ll-62-130. (
- 3. DISPATCH personnel to 120V AC Vital Instrument Power Bd 1-1 to restore power to board USING SOI-235.01, 120V AC Vital Power System 1-1.
- 4. ISOLATE RCS charging and letdown:
- CLOSE 1-FCV-62-90.
- CLOSE 1-FCV-62-91.
- CLOSE letdown orifice(s).
- CLOSE 1-FCV-62-69.
- CLOSE 1-FCV-62-70.
(
AOI-2S.01 WBN LOSS OF 120V AC VITAL INSTRUMENT Revision 27 POWER BOARDS 1-1 AND 2-1 Page 7 of 24 3.2 OPERATOR ACTIONS (RHR letdown NOT in service continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE NOTE Loop 2Pzr 2 Pzr Spray valve is inoperable in MAN and AUTO. Loop 1 Spray valve is available in MAN only.
- 5. CHECK RCS pressure stable. ENSURE Pzr heaters are OFF.
IF RCS pressure continues to rise, THEN MONITOR PORV for actuation.
- 6. MONITOR VCT level and manually makeup as required.
( NOTE T-ref signal is lost, impulse pressure can be used for matching T-avg.
- 7. MONITOR T-avg and T-ref. ADJUST Turbine load within 3°F. to match T-avg.
- 8. PLACE 1A-A MD AFW pump control switch in PULL-TO-LOCK.
- 9. MONITOR Unit in stable condition. TRIP Rx.
- RCS Pressure stable. ** GO TO E-O, Reactor Trip or Safety Injection.
- RCS Temperature stable.
- 10. ALIGN SG feedwater and steam flow to operable channels
- a. PLACE all MFW reg and bypass reg valves in MANUAL.
- b. PLACE all SG feedwater and steam flow selector HSs to B train inputs.
AOI:25.01 AOI.:25.01 WBN LOSS OF 120V AC VITAL INSTRUMENT Revision 27 ( POWER BOARDS 1-1 AND 2-1 Page 12 of 24 4.0 DISCUSSION This instruction covers the loss of 120V AC Vital Instrument Power Boards 1-land 1-1 and 2-1, assuming Unit 1 is at power operation with the other vital instrument power boards remaining operable throughout the event. Loss of the 120V AC Vital Board 1-1 results results in: 1) loss of T T-ref
-ref signal due to loss of 1- PT-1-73 causing rods to insert if in AUTO, 2) loss offeedwater auto control (Channell feedwater and steam flow inputs) complicated by false alarm failures,
- 3) erroneous computer outputs, 4) loss of trend recorders, 5) lossef loss of Channell Channel I Main Control Room indicators (pressure, flow and temperature), 6) loss of power to all ChannellI SSPS bistables, and 7) Low flow trip of running A-Train (Unit 1) 480V Board Channel Room Pressurizing Fans and start of standby Battery Room Exhaust Fan ..
Loss of the 120V AC Vital Board 2-1 results in: 1) CCS Surge Tank (auxiliary) auto makeup due to loss of level transmitter, and 2) Low flow trip of running A-Train (Unit 2) 480V Board Room Pressurizing Fans. OT ~ T or S/G LO-LO Level may occur due to failure of automatic Reactor Trip due to OT.il function of feedwater regulating and bypass regulating valve controllers being supplied power from Channel I. Manual control may be taken in the main control room to regulate flow. Ultimate reactor safety is provided by the RPS initiation due to the S/G Lo-Lo Level condition and the subsequent AFW System initiation with the turbine driven AFW pump feeding S/G 1 and 2, and B-B Motor Driven AFW pump feeding S/G 3 and 4. Loss of 120V AC Vital Board 1-1 causes VCT outlet isolation valve 1-LCV-62-132 to close and RWST valves 1-LCV-62-135 and 136 to open. This will insert borated water to the RCS from the RWST and result in a negative reactivity insertion. If in service, Train B MCR, EBR, and SDBR Chillers will swap to Train A, and auto start of Train B is disabled. Upon restoration of power to 120V AC Vital Board 1-1, MD AFW pump 1A-A master level controllers for SGs 1 and 2, and EGTS Train A CNTMT Annulus .ilP ~P controller will be in MANUAL, and will have to be reset to AUTO. Also, a Train A CRI, and CVI will occur, and will need to be addressed per the applicable SOL If the ABSCE refueling logic switch, 1-HS-90-410-A [1-R-73], is in the REFUEL position, then an "A" Train ABI (partial ABI associated with high rad in refuel area) will also occur when the CVI is initiated. No operator action is is required to put the Reactor in a safe condition and no safety limit is exceeded. As with any loss of an AC Vital Power Board, early diagnosis and recovery is of greatest assistance toward unit restoration.
AOI-2S.01 WBN LOSS OF 120V AC VITAL INSTRUMENT Revision 27 ( POWER BOARDS 1-1 AND 2-1 Page 14 of 24 APPENDIX A Page 1of4 1 of 4 120V AC VITAL INSTRUMENT POWER BOARD 1-1 LOADS MISCELLANEOUS:
** SSPS (A) Ch. I Input and Train A Output Relays in 1-R-52. ** SSPS (B) Ch. I Input Relays in 1-R-49. ** SSPS Aux Relays in 1-R-73. ** Process Protection Set I in 1-R-1. ** Process Control Group I in 1-R-14. ** NSSS Aux Relay Rack Bus in 1-R-54. ** Aux Relay Rack Separation Separation and Aux Relays in 1-R-73.
it BOP Process Inst Control Rack A Bus in 1-R-127.
** Panel 1-M-4 Instrument Bus. ** PaneI1-M-5 Panel 1-M-5 Plug mold Instrument Bus. ** Panel 1-M-6 Plug mold Instrument Bus. ** Panel*O-M-27B Panel O-M-27B Instrument Instrument Bus.
(
- Panel 1-L-11A Aux Control Room Relay Bus and Instrument Bus A.
SYSTEM 1 - Main Steam:
- 1-PIC-1-6A & & -31A, S/G 1 && 4 PORV PICs are NOT operable. Auto high pressure opening is in service for S/Gs 2 and 4 with H.S. selected to auto.
S/Gs 1 and 3 PORVs will not open.
- Steam dumps will not open (Arm Interlock, "A" solenoid de-energized).
- Steam Flow indicators for Channel I fails low (selecting Channel II may be required if power is not restored to Channel I).
SYSTEM 3 - AFW:
- 1-PDIC-3-122A, AFW Pump A-A Discharge PCV is NOT operable [1-M-4].
- 1-LCV-3-164 & 156, MDAFW LCVs to S/G 1 & 2 auto control is NOT operable.
- Aux Control Room controllers for S/G 1 & 2 swap to Manual.
AOI-25.01 WBN LOSS OF 120V AC VITAL INSTRUMENT lOSS Revision 27 (., POWER BOARDS 1-1 AND 2-1 Page 15 of 24. 24 APPENDIX A Page 2 of4 of 4 120V ACVITAllNSTRUMENT ACVITAL INSTRUMENT POWER BOARD 1-1 lOADS LOADS (continued) SYSTEM 3 - Main FW Reg Valves:
- Feedwater and steam flow inputs to reg valve controllers (Channel I normal).
- 1-LCV-3-35, Main Feedwater Reg valve Loop 1, auto function NOT operable.
- 1-LCV-3-35A, Bypass Reg valve Loop 1, auto function NOT NOToperable.
operable. SYSTEM 26 - HP Fire Protection:
- HPFP Pump 2A-A Separation Relays in JB 5064.
SYSTEM 30 - HVAC:
- 0-FCO-30-149, ABGTS Fan A-A Pressure Control [0-L-429].
- 1-FCV-30-164, AB General Exhaust Fan 1-A Flow Control [0-L-426].
- 0-FCO-30-169, Fuel Handling Fan A Flow Control [0-L-427].
** 2-FCO-30-273, AB General Exhaust Fan 2-A Flow Control [0-L-427].
(
** Redundant Vital Battery Room Exhaust Fan may start on loss of power to either Vital instrument power board 1-1 or 2-1 (refer to annunciator 150)
SYSTEM 31 - HVAC:
,.* System 31 Train A Flow Switches [JB 4313]. If Train B MCR, EBR, & SDBR chillers were in service at time of event, low flow caused auto start of Train A chillers, and disabled auto start of Train B.
- 0-FCO-31-10 & -17, Toilet & Locker Room Exhaust and Spreading Room Supply Dampers CLOSE.
- Train A CRI will occur upon restoration of board.
(
QUESTIONS REPORT for RO-14 bank Questions A0I2500.01 017 A0I2500.01017 Given the following plant conditions: ( \ Plant is operating at 100% power. A loss of 120V AC Vital Instrument Power Board 1-1 occurred. The crew implements AOI-25.01, "Loss of 120V AC Vital Instrument Power Board 1-1". Which ONE of the following identifies the train selector switches that must be selected for automatic control of feedwater and which loop 1 feedwaterlsteam feedwater/steam flow indications are available? FeedwaterlSteam Feedwater/Steam Flow FeedwaterlSteam Feedwater/Steam Flow Selector Switches Indications
- a. A train Channell
- b. A train Channell!
- c. B train Channell d ..... B train Channel II The correct answer is D.
(
- a. Incorrect - ~' train inputs for #1 S/G feedwater control is lost therefore the
'B'train controls and indications must be selected in order to provide auto control. Channel II indications only are available due to the loss of channel I power. Examinee must be able to correlate Board 1-1 to channell Channell and B train. Channelll(B train) provides the backup controls and indications.
- b. Incorrect - ~' train inputs for #1 S/G feedwater control is lost therefore the
'B' train controls and indications must be selected in order to provide auto
'B'train control. Channelll(B train) provides the backup controls and indications.
- c. Incorrect - ~' train inputs for #1 S/G feedwater control is lost therefore the
'B' train controls and indications must be selected in order to provide auto
'B'train control. Channelll(B train) provides the backup controls and indications.
- d. Correct - ~' train inputs for #1 S/G feedwater control is lost therefore the d, the
'B'train controls and indications must be selected in order to provide auto control. Examinee must be able to correlate Board 1-1 to Channell and B train. Channelll(B train) provides the backup controls and indications.
( Thursday, August 27,20097:17:34 AM 1
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 15. 058 AG2.3.34 015 Given the following plant conditions:
power-
- The Unit was at 80% power.
- A loss of 125v DC Vital Battery Board III occurred.
- AOI-21.03, "Loss of 125V DC DC' Vital Battery Bd III," Appendix A, "Transfer of 125V DC Buses," is in progress.
Depressing and holding the BO-RESET switch while 6.9kV Shutdown Board 2A-A 125V DC control bus is reenergized will prevent which ONE of the following? A. Inadvertent start of ALL four Emergency Diesel Generators. B. Inadvertent start of only the Emergency Diesel Generator 2A-A. C. Unwanted trip of components supplied by 6.9kV Shutdown Board 2A-A. D~ Unwanted start of components supplied by 6.9kV Shutdown Board 2A-A. DISTRACTOR ANAL YSIS: A. BO-Reset switch does not prevent the start of all Diesel Incorrect, Holding the BD-Reset Generators. They all started when the control power was lost as identified in AOI-23.03. Plausible because the diesel generators are affected by the loss of DC control power. B. Incorrect, Holding the BD-Reset BO-Reset switch does not prevent the start of Diesel Generator 2A-A. The DG started when the control power was lost as identified in AOI-23.03. Plausible because the diesel generator is affected by the loss of DC control power. C. Incorrect, Holding the BO-Reset switch does not prevent the unwanted trip of components; it prevents the unwanted start of the components. Plausible if the applicant reverses the equipment operation during to the transfer. D. Correct, a NOTE preceding Step 1 of Appendix A identifies that holding the BO-Reset switch after the switch is depressed will prevent the unwanted start of equipment when the BO relays energize. The switch sWitch is released after the control power is transferred. Question Number: 15 Tier: _1_ Group _1_ 1 Page 40
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 KIA: 058 AG2.3.34 Loss of DC Power Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. Importance Rating: 4.2 / 4.1 4.2/4.1 10 CFR Part 55: 41.10/43.5/45.13 10CFR55.43.b: Not applicable. KIA Match: The applicant is required to identify the resultant operational effects of holding a reset switch located in the blackout relay logic cabinet at the shutdown boards will have when 125v DC control power is transferred. Technical
Reference:
AOI-21.03, "Loss of 125v DC Vital Battery BD III," Rev 19 Proposed references None to be provided: Learning Objective: 3-0T-AOI2100 3-0T-AOI21 00
- 9. Discuss transfer of 125V DC buses using AOI-21.01 or AOI-21.02, Appendix B for 125V DC Battery Board I or II respectively.
- AOI-21.03 transfer is the same as AOI-21.01 and AOI-21.02 just a different Appendix and the DC Channel and 2A-A shutdown boards are required for Unit 1 operation.
Question Source: New Modified Bank Bank xx Question History: WBN bank question AOl21 00.09 001 Comments: Question stem and choices reworded, correct answer relocated, one distractor changed. Stem only minor changes such as power level and Board I to Board III because Board III has fewer major effects on the unit and does not require a reactor trip. Question statement reworded to eliminate the question asking for a response of something not done. ( Page 41
AOI-21.03 WBN LOSS OF 125V DC VITAL BATTERY.BD BATTERY BD III Revision 19 ( Page 3 of 15 1.0 PURPOSE This instruction provides actions to respond to a loss of 125V DC Vital Battery Board III. No major equipment will be lost which would cause a Reactor or Turbine trip... trip Various indications and alarms, as well as inoperable equipment, will be evident on Unit 2. 2.0 SYMPTOMS* SYMPTOMS 2.1 Alarms A. 125 DC VITAL CHGR/BATT III ABNORMAL [19-A]. B. 125 DC VITAL BATT BD BO III ABNORMAL [19-B]. C. 6.9 SDBD SO BO 1A-A UV/OV/CONTROL PWR FAILURE [12-B]. o. D. 480 SDSO BD BO 1A1-A/1A2-A FAILURE/ABN [10-D]. [10-0]. E. 480 AUX BLDG COMM BD BO OV, UV, TIE CLOSED [501-0]. [501-D]. 2.2 Indications ( A. Voltmeter (1-EI-57-96) when 1-XS-57-96 is selected to Bd III and ammeter (1-EI-57-94) for 125V Battery Bd III, on 1-M-1, indicates ZERO. B. Breaker indicati.ng indicating lights for equipment feeding from 6.9kV SDSO BD BO 2A-A and 480V SO BDs 480VSD BOs 2A1-A and 2A2-A will be dark (ACBs electrically inoperable). C. Position indicating lights for AFW Pump 1A-S T&T valve will be out on 1-M-4 1-M-4... 2.3 Automatic Actions A. Turbine Driven Auxiliary Feedwater Pump 1A-S control power is lost. AuxiliaryFeedwater
~~:
SO Bds. All four DIGs start but do not tie onto SD Auxiliary Building H2 supply 0-FCV-77-240 fails closed ca~sing causing loss of H2 to VCT. D. O. Cask loading area vent and exhaust 0-FCO-30-122 and 0.:.FCO-30-129 fail closed.
AOI-21.03 WBN LOSS OF 125V DC VITAL BATTERY BD III Revision 19 ( Page 5 of 15 3.0 Loss Of 125V DC Vital Battery Bd III (continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 6. PLACE Handswitch 43TL, in 6.9KV Logic Panel 2A-A, in the TEST position.
NOTE Placing the 43TL handswitch for DG 2A-A to the NORMAL position in the following SOls, should be noted as N/A when 125V DC Vital Battery Board III is deenergized, and only performed when the vital battery board has been reenergized.
- 7. STOP the following DIG's, and PLACE in standby:
- a. 1A-A USING SOI-82.01, Diesel Generator (DG) 1A-A.
1A..,A.
- b. 1B-B USING SOI-82.02,
(( Diesel Generator (DG) 1B-B.
- c. 2B-B USING SOI-82.04, Diesel Generator (DG) 2B-B.
- 8. REFER TO Tech Specs:
- 3.8.4, DC Sources-Operating.
- 3.8.5, DC Sources-Shutdown.
- 3.8.7, Inverters-Operating.
- 3.8.8, Inverters-Shutdown.
- 3.8.9, Distribution System-Operating.
- 3.8.10, Distribution System-Shutdown.
(
AOI-21.03 WBN LOSS OF 125V DC VITAL BATTERY BD III Revision 19 Page 12 of 15 1S APPENDIX A Page 1 of 1 TRANSFER OF 125V DC BUSES NOTE Holding the BO-RESET SO-RESET Switch prevents unwanted start of Train A equipment, such as CCPs, AFW pumps, ERCW pumps, CCS pumps, and PZR Htrs when BOSO relays energize, and requires two operators to perform.
- 1. DEPRESS and HOLD the 6.9kV SO Bd Sd 2A-A BO-RESET SO-RESET switch
[Logic Panel 2A-A].
- 2. WHEN DC bus is energized (125V(12SV DC CONTROL BUS SUS ENERGIZED red light ON, panel 2), THEN RELEASE the 6.9kV SO Bd Sd 2A-A BO-RESET SO-RESET switch
[Logic Panel 2A-A].
- 3. PLACE 6.9kV SO Bd Sd 2A-A 125V 12SV DC SUPPLY TRANSFER SWITCH NORMAL BUS SUS in the alternate position [panel 2].
- 4. PLACE 480V SO Bd Sd 2A 1-A 12SV 125V DC control power normal bus transfer switch in the alternate position [compt SA].
5A]. ( S.
- 5. PLACE 480V SO Bd Sd 2A2-A 125V DC control power normal bus transfer switch in the alternate position [compt SA].
5A].
- 6. DEPRESS and HOLD the 6.9kV SO Bd Sd 1A-A BO-RESET SO-RESET switch
[Logic Panel 1A-A].
- 7. WHEN DC bus is energized (125V(12SV DC CONTROL BUS SUS ENERGIZED red light ON, panel 17), THEN RELEASE the 6.9kV SO Bd Sd 1A-A BO-RESET SO-RESET switch
[Logic Panel 1A-A].
- 8. PLACE 6.9kV SO Bd 125V DC SUPPLY TRANSFER SWITCH BACKUP Sd 1A-A 12SV SACKUP SUS in the alternate position [panel 17].
BUS
- 9. PLACE 480V SO Bd Sd 1A1-A 1A 1-A 12SV 125V DC control power backup bus transfer switch i~ the alternate position [compt 6A].
in
- 10. PLACE 480V SO Bd Sd 1 A2-A 12SV 1A2-A 125V DC control power backup bus transfer switch in the alternate position [compt 6A].
- 11. IF CSST C switchgear NORMAL control power feed is available, THEN ENSURE CSST C switchgear on NORMAL control power feed.
((,
AOI-21.03 WBN WBN* LOSS OF 125V DC VITAL BATTERY BD III Revision 19 Page 13 of 15 APPENDIXB Page 1 of 1 RESTORATION OF 125V DC BUSES NOTE Holding the BO-RESET Switch prevents unwanted start of Train A equipment, such as CCPs, AFW pumps, ERCW pumps, CCS pumps, and PZR Htrs when BOBO relays energize, and requires two operators to perform.
- 1. DEPRESS and HOLD the 6.9kV SO Bd 2A-A BO~RESET SD Bd2A-A BO-RESET switch
[Logic PaneI2A-A]. Panel 2A-A].
- 2. WHEN DC bus is energized (125V DC CONTROL BUS' BUS ENERGIZED red light ON, panel 2), .
THEN RELEASE the 6.9kV SD SO Bd 2A-A BO-RESET switch [Logic Panel 2A-A].
- 3. PLACE 6.9kV SD SO Bd 2A-A 125V DC SUPPLY TRANSFER SWITCH NORMAL BUS in the normal position [panel 2].
- 4. PLACE 480V SO SO Bd 2A1-A125V 2A 1-A 125V DC control power normal bus transfer switch
( in the normal position [compt 5A].
- 5. PLACE 480V SO SD Bd 2A2-A 125V DC control power normal bus transfer switch in the normal position [compt 5A].
- 6. DEPRESS and HOLD the 6.9kV SO SD Bd 1A-A BO-RESET switch
[Logic Panel 1A-A]. 1A-A].
- 7. WHEN DC bus is energized (125V DC CONTROL BUS ENERGIZED red light ON, panel 17),
THEN RELEASE the 6.9kV SD SO Bd 1A-A BO-RESET switch [Logic Panel 1A-A].
- 8. PLACE 6.9kV SO SD Bd 1A-A 125V DC SUPPLY TRANSFER SWITCH BACKUP BUS in the normal position [panel 17].
- 9. PLACE 480V SO SD Bd 1A1-A125V 1A 1-A 125V DC control power backup bus transfer switch in the normal position [compt 6A].
- 10. PLACE 480V SO SD Bd1A2-A Bd 1A2-A 125V DC control power backup bus transfer switch in the normal position [compt 6A].
(
QUESTIONS REPORT for ro-15 bankQuestion w6f'1 G/tN)L WoN 6/tN)L A0I2100.09001 Q\A~ 11GtJ Q\AE:Sl/Q;J Given the following plant conditions:
- The Unit was at 50% 50% power.
- A loss of 125v DC Vital Battery Board I occurred.
- AOI-21.01, Loss of 125v DC Vital Battery Bd I, App B, Transfer of 125v DC Busses, is in progress.
- Contrary to step [1], the BO-RESET switch was NOT depressed when the DC control bus was reenergized.
Which of the following conditions will occur as a result of this error?
- a. Running components on 6.9kv Shutdown Board 1A-A 1A-A will load shed.
- b. Load shedding and sequencing would not occur if a loss of offsite power were to occur after the control bus was re-energized.
c." c ...... Unwanted starts of components supplied by 6.9kv Shutdown Board 1A-A.
- d. An unwanted start of the 1A-A EmergencyEmergency Diesel Generator.
The correct answer is C ( Thursday, August'27, 2009 7:27:12 AM August 27,20097:27:12 AM 1
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 16. 065 AK3.04 016 Which ONE of the following is the reason for the backup nitrogen supply for the Turbine Driven AFW Pump LCVs and the action/condition required to align the backup supply?
A. Allows the LCVs to be OPENED duri'ng durihg a Station Blackout event and automatically supplied when air pressure drops below regulator setpoint. B~ Allows the LCVs to be OPENED during a Station Blackout event and will require manual alignment locally when needed. C. Allows the LCVs to be CLOSED during a Station Blackout event and automatically supplied when air pressure drops below regulator setpoint. D. Allows the LCVs to be CLOSED during a Station Blackout event and will require manual alignment locally when needed. DISTRACTOR ANAL YSIS: A. Incorrect, the backup supply is to allow the valves to be opened during a station blackout where the normal air is lost and while there is a regulator, the backup supply is not functional until it is manually aligned. Plausible because other AFW LCVs do fail closed and there are regulators to maintain pressure to the LCVs . ( when using the backup supply. B. CORRECT, the backup supply is from high pressure nitrogen cylinders that allow the valves to be opened a limited number of times after the normal air pressure is lost during a station blackout and its use requires manual valve alignment in accordance with SOI-3.02, Auxiliary Feedwater System. C. Incorrect, the backup supply is to allow the valves to be opened, not closed, during a station blackout where the normal air is lost but it requires a manual valve to be opened to enable its use. Plausible because other AFW LCVs do fail open and there are regulators to maintain pressure to the LCVs when using the backup supply. D. Incorrect, the backup supply is to allow the valves to be opened, not closed, during a station blackout where the normal air is lost. The manual alignment of the supply is correct. Plausible because other AFW LCVs do fail open and the manual alignment is required to use the backup supply. Page 42
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 16 Tier: _1_ Group 1 KIA: 065 AK3.04 Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: Cross-over to backup air supplies. Importance Rating: 3.0 /3.2 3.0/3.2 10 CFR Part 55: 41.5,41.10/45.6/45.13 41.5, 41.10 / 45.6 / 45.13 10CFR55.43.b: Not applicable. KIA Match: The applicant must determine how the backup nitrogen bottle supplies would be used to operate the TD TO AFW level control valves, an event that is a basis for the backup supply of nitrogen, and whether the swapover to the backup on loss of air is automatic or manual. Technical
Reference:
47W848-12 R22 FSAR FSAR,, Section10.4.9.3, Amendment 7 Drawing 1-47W611-3-4, R17 SOI-3.02, "Auxiliary Feedwater System," Rev 0046 Proposed references None to be provided: Learning Objective: 3-0T-SYS0038
- 23. Using plant drawings, determine the effect of a loss of instrument air/control power on the following valves/components:
- a. MDAFWP regulating valve (main and bypass)
- b. TDAFWP regulating valve
- c. AFW pumps.
Question Source: New Modified Bank Bank x X Question History: SON 2009 exam Comments: I. Page 43
- WBNP-1 The AFW system is designed to deliver 40°F to 120°F water to the secondary side of the steam generators at a pressure ranging from the RHR system cut-in point of 350°F in the RCS (equivalent toto110 110 psig in the steam generators) to the lowest MSSV set pressure, plus 3% set error, plus 3% accumulation pressure (1257 psig).
Separate ESF power subsystems and control air subsystems serve each motor driven AFW pump the turbine-driven pump are served by one and its associated valves. The valves associated with the of the two electric and control air subsystems. The turbine driven pump valves associated with SG 1 and 2 are served by Train B and the valves associated with SG 3 and 4 are served by Train A. The turbine-driven pump receives control power from a third dc electric channel that is distinct fi'om from the channels serving the motor driven pumps. 10.4.9.3 Safety Evaluation For the design bases considerations discussed in Section 10.4.9.1, sufficient feedwater can be provided over the required required pressure range even if the failure of a feedwater line is the initiating event, anyone AFW pump fails to start, and no operator action is taken for 12 minutes. The 'Loss of Normal Feedwater' and loss of offsite power analyses in Chapter 15 demonstrate that the auxiliary feedwater system satisfies the design bases described in this section. In the event that loss of offsite power (LOOP) occurs, 41410Ogpm gpm of AFW delivered to two steam ( generators within one minute will prevent relief of reactor coolant via the pressurizer safety valves. Water levels in the steam generators will remain above the required minimum tube sheet coverage. The AFW system meets these requirements even when the single failure criterion is applied. In the event of a feedwater line break, essentially the same requirements are imposed and act as for the LOOP case. Other cases discussed in Section 10.4.9.1 impose less stringent conditions. Following a loss-of-coolant accident (LOCA) for a small break with offsite power, the RCS pressure and temperature decrease at a relatively slow rate. The AFW system provides sufficient flow to the steam generators so that RCS cooldown can proceed. In this case, the AFW system function is similar to its function following other events described in Section 10.4.9.1, such as loss of offsiteelectrical offsite electrical power. In contrast, the AFW system serves a distinctly different function during a large break LOCA, where steam generator tube leaks may be present. A large LOCA causes a rapid depressurization of the RCS so that the secondary side pressure and temperature exceed primary pressure and temperature, and consequently any fission products in the RCS cannot escape to the secondary side. Subsequent cooling of the secondary side fluid could eventually reduce the secondary side pressure to atmospheric, permitting pem1itting any fission products in the RCS to escape into the secondary system. The AFW system is used to maintain sufficient water level on the steam generator secondary side so that static head prevents primary-to-secondary tube leakage and prevents the escape of any fission products. 10.4-35
--~-~-~-~-------------------------'------ WBNP-I WBNP-l Whenever a flood above plant grade is anticipated, an orderly shutdown to hot shutdown and a cooldown to cold shutdown will be initiated immediately. In a little more than 5 hours hoUrs after reactor shutdown, the secondary system pressure will bereduced be reduced to approximately 100 psig. Within 27 hours after reactor shutdown, the fIre-protection ftre-protection system piping will be connected to the auxiliary feedwater discharge piping by means of special spool pieces not normally installed, and the secondary system system pressure will be maintained :::;;125
- 125 psia. Operator action is taken, during a flood transient at 17 hours, to reduce SG operating pressure to
- ;;80
- 80 psig, to facilitate the supply of cooling water from the HPFP system in the event of the failure ofO-PCV-26-18 to
~c1ose.
close. Valve 0-PCV-26-18 is mounted in the HPFP pump recirculation line to the river and is removal. When the flood exceeds non-safety grade. This pressure is sufficient for decay heat removaL plant grade, the auxiliary feedwater pumps will be inoperable, and the ftre fIre protection pumps, which are located above the maximum possible flood elevation, will supply feedwater. Appropriate portions of the HPFP system are designed to function under normal conditions as well as for the maximum possible flood with the coincident or subsequent loss of the upstream and/or downstream dams. The HPFP pumps are located In andlor in the intake station above the flood line and are arranged alTanged to supply water directly to the steam generators in the event the auxiliary feedwater pumps are flooded. feed water to the steam generators is The portion of the HPFP system which supplies auxiliary feedwater ASME Section III, Class 3, Seismic Category I with the exception of the fIre pumps discharge relief valve. These valves are replaced by ASME Section III, Class 3 blind flanges during flood ( mode preparations to ensure the integrity of the ASME Section III, Class 3 auxiliary feedwater supply piping during flood mode operation. Tl~e The AFW system is required to be available in the event of an ATWS event. The most severe ATWS scenarios have been determined ATWS detennined to be those in which there is a complete loss of normal feedwater (Reference WCAP-I0858). WCAP-10858). The design basis events for the AMSAC are Loss of Normal Feedwater/ATWS and Loss ofLoadlATWS. of Load/ATWS. Since there is a complete loss of normal feedwater during both of these transients, the accident analysis of both transients (Chapter 15) assumed AFW reaches full "flow within 60 seconds after the initiating event for long tenn term reactor protection. Also, the Loss of Normal Feedwater transient assumed a turbine trip within 30 se~onds seconds after the initiating event to maintain short term tenn pressures below ASME Service Level C Nomlally these features will be actuated by the RPS. However, if a common pressure limits. Normally mode failure to the RPS incapacitates AFW initiation andlor and/or turbine trip in addition to altemate method of providing AFW flow and a turbine trip is prohibiting a scram, then an alternate required to maintain RCS pressure below ASME Service Level C pressure limits. These two functions, turbine trip and AFW flow actuation, are provided via the AMSAC. The AFW piping system layout has been optimized to prevent water hammer OCCUlTences occurrences induced by the piping system. I ( 10.4-36
WBNP-5 Minimum flow rate requirements are met by the system for the design transients or accident conditions shown in Table 10.4-6, even if the worst case single active failure occurs simultaneously. The single active failures considered in the table are listed below: A. Alternating current train failure B. Turbine-driven pump (TDP) (TOP) failure C. Motor-driven pump (MDP) (MOP) failure D. O. Pressure control valve failure (MDP-runout (MOP-runout protection) E. Level control valve failure for TDP TOP F. Level control valve failure for MDP MOP G.Pressure G. Pressure switch failures H. AFW System check valve failure (failure to close on reverse flow) I.
- 1. Flow controller failure (TDP-runout (TOP-runout protection)
Operator intervention within 10 minutes is required in order to meet the maximum flow requirements for the main steamline break inside containment, or within 12 minutes (based on recent safety reanalysis) to meet the minimum flow requirements for the feedline rupture. However, the original 10 minute requirement is maintained whichwhich is conservative. In addition to using high quality components and materials, the AFW system provides complete redundancy in pump capacity and water supply for all cases for which the system is required. Under all credible accident conditions, at least one AFW pump is available to supply two steam generators not affected by the accident with the required feedwater. Only two steam generators are required to be usable for any credible accident condition. ( Redundant electrical power and air supplies assure reliable system initiation and operation. The electric motor-driven pumps and associated controls, valves, and other supporting systems are powered by offsite or onsite ac sources. The exceptions are the MOPMDP level control valves and the control circuits supplied by trained dc power. The TOPTDP and associated valves, controls, and other supporting systems are powered by steam and dc electric power, except for four 480V ac power operated valves in the steam supply lines to the pump turbine which are included to satisfy pipe rupture criteria. Stealnfor Steam for the TDP TOP is provided from either of two of the four main steam loops as controlled by two of these valves. They assure that only one steam source is available at a time, time,and, and, in the event of a pipe rupture, prevent two steam generators from blowing down simultaneously. The other two valves simply isolate in the event of a pipe rupture mpture (due to high temperatures in TOP TDP room). Four of these are motor operated, fail as is, and all but isolating the standby steam source are normally open. This assures that a steam supply the one isol&ting to the turbine will be available for design basis events and a LOOP event (see Table 10.4-4, Items 2, 4, and 5). Failure modes and effects analyses for thetheAFW AFW System are presented in Tables 10.4-3, 10.4-4, and 10.4-5 ( 10.4-37
WBNP-l ( During certain postulated 10 CRF 50, Appendix R fires, the electric cables for the electro-pneumatic controllers of the turbine-driven steam generator level control valves (LCV) could be destroyed. A backup source of compressed nitrogen has been provided to allow manual control of the LCVs for both the station blackout and Appendix R fire. No electric power is required for the operation of this backup nitrogen source. Valves associated with this backup source must be
)llanually operated to align the high pressure nitrogen bottles and associated piping to take local .manually manual control of the LCVs. The LCV position is adjusted upon orders from either the Control Room or the Auxiliary Control Room, based on the observed water level in the steam generators being controlled.
10.4.9.4 Inspection and Testing Requirements A comprehensive test program is followed for the AFW system. The program consists of periodic tests of the activation circuitry and mechanical components to assure reliable performance throughout the life of the plant. During plant operation, the system can be tested by pumping condensate storage water to the condensate storage tank.tank:. BRCW ERCW and HPFP water will not be fed to the steam generators during tests, but separate tests on the BRCW ERCW and HPFP system will assure the availability of the alternate water supplies. Test requirements are given in Chapter 14. Surveillance test requirements are given in Chapter 16. Inservice inspection requirements are ( given in Chapter 3. 10.4.9.5 Instrumentation Requirements The three pumps start automatically on a loss of o ffsite power, trip of both main feedwater offsite pumps, a safety-injection signal, or a AMSAC AM SAC signal. The electric-motor-driven pumps also start automatically on a two-out-of-three low-low level signal from any steam generator, and the turbine driven pump starts automatically on a two-out-of-three low-low level signal from any two steam generators. All pumps can be started remote-manually. A modulating level control valve (which is norn1ally normally closed) between each pump and each steam generator fed by the pump receives an an opening signal on a low-low water level in the steam generator. For the motor driven pumps, two modulating level control valves, a 4-inch and a parallel2-inch, parallel 2-inch, (which are normally closed) between each pump and each steam generator fed by the pump receives an opening signal on a low-low water level in the the steam generator. For the motor driven pumps these valves will continue to modulate and automatically maintain steam generator water level. After the steam generator decreases in pressure to a certain setpoint the 4-inch valve will close to protect it from cavitation damage. The 2-inch LCV is designed for for extended operation at low flows and high pressure drops. The system may be controlled manually. (
\
10.4-38
WBN Auxiliary Feedwater System 501-3.02 Unit 1 Rev. 0046 { Page 8 of 92 \ 3.0 PRECAUTIONS AND LIMITATIONS (continued) K. After each AFWP start, 6.9kV ACB closing spring must be checked to ensure it is charged. 1o 1o L. An operator with no other duties will be assigned to initiate AFW any time the auto initiation circuits are inoperable. Engineering or Maintenance personnel must notify the Shift Manager (SM) if this condition exists. ss M. Excessive RCS cooldown is possible when using AFW System. N. Any time Turbine Driven or Motor Driven AFW Pumps are running, oil level and temperature should be checked frequently. Pumps must be TRIPPED if pump bearing oil temperature exceeds 165°F and cause of overheating determined, and corrected prior to pump resuming operation. O. Turbine Bearing Oil pressure should be above 15 psig, and Turbine Bearing Oil temperature below 180°F. TO AFW Pump should be TRIPPED if Turbine Bearing Oil Temperature exceeds 200°F. P. When SGs are above 212°F, backleakage to the AFW system can lead to pump steam binding. AUO rounds require periodic checking of the pumps for this ( condition. Venting is required until the cause is found and corrected whenever this condition occurs. 1,2.4.6.7 1,2,4,6,7 Q. 4" LCVs from M-D AFWPs auto close when Feedwater header pressure drops below 500 psig to prevent cavitation damage to LCV. R. Low flow operation of both motor and turbine driven AFW pumps must be minimized to prevent possible degradation of pump impeller. Main Feedwater System should be utilized for lowflow low flow conditions if available. S. The minimum pressure that the backup nitrogen supply bottles can reach and still supply the required volume to cycle one train of LCVs five times, is 1085 psig. Bottles should be changed out when pressure lowers to 1200 psig . or below. T. With the additional recirculation line in service, use of the Motor-drivenpump(s) (if available), is preferred over the use of the Turbine-driven pump during low flow conditions when the AFW demand is within the capability of the Motor-driven pump(s). U. SGBD isolation will initially raise indicated Calorimetric Power associated with the resulting feedwater flow transient. It may take several minutes for Calorimetric Power and Feedwater Flow to stabilize at their new lower values. 11 11 V. The turbine lube oil level in the sight glass is expected to drop out of the marked fill band while the turbine is operating, but oil level should remain visible in the sight glass at all times.
WBN Auxiliary Feedwater System SOI-3.02 501-3.02 Unit 1 Rev. 0046 Page 48 of92 of. 92 Oate_ __ Initials 8.6 Nitrogen Alignments to TO AFWPump AFW Pump SG LCVs Source Note 9 applies to this entire section CAUTIONS
- 1) Radio use in this area may affect SG 1 & 4 Main Steam Line Rad Monitors (1-RM-90-421 (1-RM-90-421B B & 424B).
- 2) Press regulators (1-PREG-3-172C, 173C, 174C, and 175C) are preset, and do NOT require adjustment for operation. Inadvertent adjustment of these regulators may damage the LCV's diaphragm.
NOTE Coordination with MCR Operator, the AUO at the TO AFWP, the AUO(s) operating the SG PORV(s), and TO AFW LCVs at the N2 2 Station(s) may be required to control plant cooldown. ( [1] ENSURE communications with UO established BEFORE local AFW LCV manipulation. [2] IF using SG 1 & 2 LCVs, THEN PERFORM the following [AB e1737, el 737, Outside SVR]: NOTE Only ONE N2 cylinder should be valved into service at a time. [2.1] ENSURE ONE of the following N2 cylinder valves OPEN (N/A cylinder/valve NOT used):
- valve (no UNIO) on 1-TANK-3-400A
.** valve (no UNIO) on 1-TANK-3-400B
WBN Auxiliary Feedwater System 501-3.02 Unit 1 Rev. 0046 Page 49 of92 of 92 Oate, Date_ __ Initials 8.6 Nitrogen Alignments to TO TD AFW Pump SG LCVs (continued) [2.2] VERIFY cylinder pressure greater than 1650 psig for N2 cylinder used in Step 8.6[2.1]: 1-PI-3-400A (1-TANK-3-400A) OR 1-PI-3-400B (1-TANK-3-400B). [2.3] CLOSE the following Vent isolation valves: [2.3.1] [2.3.1 ] 1-VTIV-3-1740 for SG 1 (1-LCV-3-174) [2.3.2] 1-VTIV-3-1730 for SG 2 (1-LCV-3-173) [2.4] OPEN N2 supply valve for N2 cylinder used in Step 8.6[2.1)): 8.6[2.1 )): 1-ISIV-3-400A (1-TANK-3-400A) ( OR 1-ISIV-3-400B (1-TANK-3-400B). [2.5] ENSURE 1-SPV-3-400, Pressure Regulator bottom selector knob, is pointing to N2 cylinder selected in Step 8.6[2.1] [2.6] ENSURE, 1-PREG-3-400, Pressure Regulator top selector knob is set to provide 85 to 95 psig on 1-P 1-3-4000. 1-PI-3-4000. [2.7] ENSURE adjustment screw on top of 1-PREG-3-173B and 1-PREG-3-174B has zero spring tension (finger tight). [2.8] OPEN the following N2 supply isol valves: [2.8.1J [2.8.1] 1-ISIV-3-400F1 for SG 1 (1-LCV-3-174)
,[2.8.2]
[2.8.2] '1-ISIV-3-400EI 1-ISIV-3-400EI for SG 2 (1-LCV-3-173) [2.8.3] 1-ISIV-3-400F2 for SG 1 (1-LCV-3-174) ( [2.8.4] 1-ISIV:*3-400E2 1-ISIV-3-400E2 for SG 2 (1-LCV-3-173)
WBN Auxiliary Feedwater System 501-3.02 Unit 1 Rev. 0046 Page 50 of92 of 92 Date. Date_ __ Initials 8.6 Nitrogen Alignments to TD TO AFW Pump SG LCVs (continued) NOTE LCVs may be manipulated locally by turning PREG adjustment screw. Clockwise raises N2 pressure to OPEN the LCV. Counter-clockwise (CCW) lowers N2 pressure to CLOSE the LCV. [2.9] ADJUST the following PREGs as required to position selected AFW LCVs per Unit Operator's instruction:
- 1-PREG-3-174B, for SG 1 (1-LCV-3-174),
(obseNe 1-PI-3-174). (observe
- 1-PREG-3-173B, for SG 2 (1-LCV-3-173),
(obseNe 1-PI-3-173). (observe [3] IF using SG 3 & 4 LCVs, THEN PERFORM the following [AB el 737, Outside SVR]: ( NOTE Only ONE N2 cylinder should be valved into service seNice at a time. [3.1] ENSURE ONE of the following N2 cylinder valves OPEN (N/A cylinder/valve NOT used):
- valve (no UNIO) on 1-TANK-3-401A
- on 1-TANK-3-401 B valve (no UNIO) on1-TANK-3-401B
[3.2] VERIFY cylinder pressure greater than 1650 psig for N2 N2 cylinder used in Step 8.6[3.1] 1-PI-3-401A (1-TANK-3-401A) OR 1-PI-3-401 B (1-TAN (1-TANK-3-401 K';3-40 1B). [3.3] CLOSE the following Vent isol valves: [3.3.1] 1-VTIV-3-1720 for SG 3 (1-LCV-3-172) ( [3.3.2] 1-VTIV-3-1750 for SG 4 (1~LCV-3-175) (1-LCV-3-175)
WBN Auxiliary Feedwater System 501-3.02 Unit 1 Rev. 0046 Page 51 of 92 of92 Date_ __ Initials 8.6 Nitrogen Alignments to TO AFW Pump SG LCVs (continued) [3.4] OPEN N2 supply valve for N2 cylinder used in Step 8.6[3.1]: ISIV-3-401A (1-TANK-3-401A), OR 1-ISIV-3-401 B (1-TANK-3-401 B). [3.5] ENSURE 1-SPV-3-401, Pressure Regulator bottom selector knob, is pointing to N2 cylinder selected in Step 8.6[3.1]. [3.6] ENSURE, 1-PREG-3-401, Pressure Regulator top knob is set to provide 85 to 95 psig on 1-PI-3-401D. 1-PI-3-401 D. [3.7] ENSURE adjustment screw on top of 1-PREG-3-172B and 1-PREG-3-175B has zero spring tension (finger ( tight). [3.8] OPEN the following N2 supply isol valves: [3.8.1] 1-ISIV-3-401EI for SG 3 (1-LCV-3-172) [3.8.2] 1-ISIV-3-401 1-ISIV-3-401F1F1 for SG 4 (1-LCV-3-175) [3.8.3] 1-ISIV-3-401 1-IS IV-3-40 1E2 for SG 3 (1-LCV-3-172) [3.8.4] 1-ISIV-3-401F2 for SG 4 (1-LCV-3-175) NOTE LCVs may be manipulated locally by turning PREG adjustment screw. Clockwise raises N2 pressure to OPEN the LCV. Counter-clockwise (CCW) lowers N2 pressure to CLOSE the LCV. [3.9] ADJUST the following PREGs as required to position selected AFW LCVs per Unit Operator's instruction:
- 1-PREG-3-172B for SG 3 (1-LCV-3-172),
(observe 1-PI-3-172).
- 1-PREG-3-175B for SG 4 (1-LCV-3-175),
(observe 1-PI-3-175).
WBN Auxiliary Feedwater System 501-3.02 Unit 1 Rev. 0046 ( Page 52 of92 Date_ __ Initials 8.6 Nitrogen Alignments to TD AFW Pump SG LCVs (continued) [4] IF in-service N2 Cylinder pressure fails LOW, THEN PERFORM Appendix A to place a spare cylinder in service. [5] WHEN use of N2 for local AFW LCV control is complete, THEN [5.1] [5.1 ] ENSURE the following valves are CLOSED:
- 1-ISIV-3-400A 1-IS IV-3-400A IV
- 1-ISIV-3-400B IV
- 1-ISIV-3-401A IV
(
- 1-ISIV-3-401 B IV
[5.2] VERIFY the following Nitrogen cylinder valves CLOSED:
- valve (no UNIO) on 1-TANK-3-400A IV
- valve (no UNIO) on 1-TANK-3-400B IV
- valve (no UNIO) on 1-TANK-3-401A IV
- valve (no UNIO) on 1-TANK-3-401 B IV
WBN Auxiliary Feedwater System 501-3.02 Unit 1 Rev. 0046 Page 53 of92 of 92 Date Initials 8.6 Nitrogen Alignments to TD AFW Pump SG LCVs (continued) [5.3] ENSURE N2 supply isolation valves are CLOSED: (Unused portion may be N/A'd)
- 1-ISIV-3-400E1 IV
- 1-ISIV-3-400F1 IV
- 1-ISIV-3-401 E 1 IV
- 1-ISIV-3-401 F1 IV
(
- 1-ISIV-3-400E2 IV
- 1-ISIV-3-400F2 IV
- 1-ISIV-3-401 E2 IV
- 1-ISIV-3-401 F2 IV
WBN Auxiliary Feedwater System 501-3.02 Unit 1 Rev. 0046 Page 54 of92 Date Initials 8.6 Nitrogen Alignments to TO AFW Pump SG LCVs (continued) [5.4] [504] BLEED pressure off line of associated LCV(s) by turning adjustment screw CCW, and using regulator petcock to VENT N2:
- 1-PREG-3-174B for SG 1 (1-LCV-3-174), (observe 1-PI-3-174)
- 1-PREG-3-173B for SG 2 (1-LCV-3-173), (observe 1-P 1-3-173) 1-PI-3-173)
- 1-PREG-3-172B for SG 3 (1-LCV-3-172), (observe 1-PI-3-172).
- 1-PREG-3-175B for SG 4 (1-LCV-3-175), (observe 1-PI-3-175).
[5.5] OPEN LCV N2 vent(s) Closed in Steps 8.6[2] or 8.6[3]:
- 1-VTIV-3-1740, 1-VTIV-3-174D, SG 1 LCV N2 Vent
( IV
- 1-VTI~-3-173D, SG 2 LCV N2 Vent 1-VTIV-3-1730, IV
- 1-VTIV-3-172D, SG 3 LCV N2 Vent 1-VTIV-3-1720, IV
- 1-VTIV-3-1750, 1-VTIV-3-175D, SG 4 LCV N2 Vent IV
[5.6] NOTIFY UO when complete. [5.7] IF cylinder pressure is below 1650 psig, THEN INITIATE WO to fill used N2 cylinder.
OL-BvBMLv-L 5 6 7 8 9 10 11 (V211 ) O-FCV-65-28A-S l-/')oc'h:-,.""~-,,=,,, CRANE WALL A
~ A TRAIN A FROM CNTIAT fROM eNTIAT ISOL VALVE UNIT 1 1-47W848-1 COORD C-9 1-+7W848-1 '"
2" OTE -+ 4 1-32-431 ('11079) (Vl07S) l-L-366 (V666) (V6GG) l-PCV-68-3400 TO REACTOR BLDG UNIT 2 B 47"8+8-1 COORD (-7 B NOTES, NOTES;
- 1. SEE SHEET 1 FOR GENERAL NOTES.
1/2" ,. 2. 3. LINE SIZES OFF OFf HEADER ARE 1/2" UNLESS NOTED.
- 4. ROUTE'"
ROUTE I" SS PIPE TO BOTTOM OF PRESSURIZER. FROM AUX COMP TR-8 EL 757 'L- 5.
~
..1.:.. 0-32-409 0-32-4091 I TO REACTOR SLOGat DG UNIT UN IT 1 6. THIS VALVE rILL WILL BE USEO FOR fOR HOR HDR SLOWDOWN.
BLO'OOWN. 1-47.848-1 COORD E-7 1-+7W848-1 1-+71'1848-1 COORD 0-7 CONTROL BLDG RADIATION FILTER fILTER TO BE INSTALLED INSTALLED.!:. a. SUPPORTED
~L7J7 I REQUIRE'-4ENTS OF TVA SAFETY CLASS C. VENDOR USING REQUIREMENTS VENOOR WILL BUILD FILTER flL TER TO ASME SECTION 3ZJJl :JZIII REaUIREMENTS REIJUIRE"!ENTS (TVA CLASS G).
EL 713'~ REFER TO 1-47W848-1 1-47111'848-1 NOTE 13 FOR VCSH DEFINITION. DEfINITION. 1-32-3834 8. ALL PIPING IS CLASS C. THIS CLASSIFICATION CLASSIfICATION EXTENDS UP TO (V88S) (V885) AND INCLUOING THE FEATURE ISOLATION VALVE PIPING AND TUBING 1-FCV-90-114 l-FeV-SO-114 DOWNSTREAM OF THE FEATURE ISOLATION VALYE IS CLASSIFIED AS (V1JI6) El713' "INSTRUMENT".
"INSTRUI.4ENT". SEE ENGINEERING REQUIREMENTS REQUIRE'-4ENTS SPECIFICATION SPECIfICATION (ERS) l-PCV-J-1J2 (V408)
N3E-934 FOR INSTALLATION AND INSPECTION REQUIREMENTS. c (V967) f'/2" l-LCy-,J-173 1-LCV-J-173 TRAIN B FILTER IS VENDOR SUPPLIED UNDER REO NO. 831283. c L-530 (vsa7) [F~R~~~AU~X~'~L~~U~N~IT~'>_~r_JL~~J~~~~R~LH~V:A~LV:E~2~"=r-{~~~~~~--{] [F~RtOM~AU~X~8~LDC~U~N~IT~'t>_"iT..J~~~~~80~R~LH":V:AL~V::E....;!'~"=r-t*"~l.n:.~'=" ..2'F-'~'j-'~':"5_-o 10. 1(;), CIID ANOD..:IIJANOo:::IIJ DENOTE SAFETY RElATED RELAlED AND NON-SAFETY RELATED UNIT 1/2 3739 (V1162) (VI162) (V406) (V4D6) 1-lCV-J-1H 1-LCV-.3-1H I-FCV-90-11S 1-47'8<8-' 1-47W848-1 COORD COORD 0-8 0-8 '"1
'" j ~
fl. NOTE 6 \~;g~'-'2 i~~t~~'-12 INTERfACE POINTS. SAFETY RELATED INTERFACE POINTS WILL BE a..OSED WITH THE METHOD OF AC1UATION REMOVED OR CLOSED RELATED INTERFACE POINTS WIll 8E LOCKED CLOSED OR CLOSED WITH 8E LOCKED CLOSED. REt.10VED OR DISABLED. ClOSED, DISABlfD. NON-SAFETY HANOWH(ElS HANOWHEELS REMOVED. TCV-.31-142 TCV-JI-142 3783 3783 .1LX
.Ji1." W SHIELD SHIELD 1-32-444 1-32-*H4 11. ITEM AND ASSOCIATED COMPONENTS ARE REQUIRED TO REMAIN RE'-4AIN lGURATION AS SHOWN TO PR THE ABSCE PRESSURE BOUNDARY.
~~ L-_-t:37:o8:J7 3740 WALL MPONENTS. INCLUDING SUPP REQUIRED FOR OR SEI~lIC SEISMIC
'--C"O'j'--D (V1169)
(VI169) 1-32-1480 PRESSURE BOUNDARY 3745 WALL (Vl 083)J
-fl{Vl0B3 ARE LABELED IN THE FIELD.
TT-Jl-54 TT-.31-54 MAY INCLUDE COMPONENTS ON E ANNULUS LUS AND AUXILIARY (V435) 1-lCV-3-148 l-LCV-3-148 (V882) (Y882) l-FCV-90-110 1-FeY-90-11 0 1/2"l 1/2" 1 1-L-421 l-L-421 ISMIC THE SHIELD WALL THE SUP? QUALIFICATION ARE IOENTIF ED OM DCAs IC OUAL! REQUIRED IRED TO MAINTAIN ABSCE CAs 52283-268 &. & -269
-2S9.*
~
.3784 3784
('1709) (V709) 12
- 12. LABELING IS NOT APPLICABLE APPL.ICABLE TO SUPPORTS AND AIR SUPPLY LINE FROM MANUAL MANUAL.
0-32-400 0-32-+00 3746 (:ri~jl-.3.37 1l--!' -n(~:~i) (~:~i) (V699) l-PCV-6S-B3 l-PCY-65-83 ISOLATION VALVE TO HOST Fev. Fey. l-LCV-3-148A (V435) (V436) (1i!j,J-3J7 :x7x.441-,-t:37:o4:1'-II Q-l':x':(Hh-C"o'j' 1-peY-65-87 l-PCV-6S-87 1---_ _--'-.:::-'-_=-_C 1-_ -'-::::-_ _ _ _C"O' -n\~~~~~) 37,,8:Jj66-{] \~L ~~~) l-LCV-3-171 TC-31-J37 TC-.31-.3.37
!.fl..:..
~
0-32-+02 0-.32-+02 !fl;.. FCO-31-3.37 FCO-31-337 l~g~65-89 (V99J) (V993) D 3748 (V702) FCO-JI-31 FCO-.31-31 ~ CRANE o D l-LCV-J-171A (V980) WALL WALL*.
REFERENCE:
REfERENCE DRAWINGS;
...!..':.-I
...!.:.-.I fE E lL 73 7' 7.37' El'-l.3'~
El713'~ V: EL f/:EL 692~_32_404 69~_32_4D4 TT-Jl-J38 TT-.31-.3.38 TC-31-.338 TC-31-338
- l
>"1 0
TRAIN B FROM CNTMT
~
+7W5DO-210.
47'1600-210.211.216. 211, 216. 217. 218. 219------PHVSICAl219------PHYSICAL PIPING REACTOR BLDG r-FCO-,31-.3.38 FCO-31-338 '5 c YALVE UNIT 1 rSOL VALVE 1*-32-440 2" +7W600-203-------------------------------PHYSICAL 47W6QO-203-------------------------------PHYSICAL PIPING -AUX 47W600-20~-------------------------------P ... YSICAL PIPING AUX BLDG EL 757'
.oI-7W600-20+-------------------------------pHYSICAl AUX BLOC BLDG ElEL. 737'
~-I 1-47W848-1 COORD 0-11 1-+7W848-1 I +7W600-207-------------------------------PHYSICAl 47W600-207-------------------------------PHYSICAL PIPING AUX BLDG LEl757' NOTE ~ 47W848-110---':..---------------------------I.4ECHANICAl 47'11'84B-l10---:..---------------------------MECHANICAL STRESS ANAL YS!S PROBLEM ANALYSIS PROBLE"!
1-32-Hl BOUNDARY CONTROL AIR THIS CONFIGURATION CONTROL DRAWING SUPERSEDES UNIT lINIT 1 AS-CONSTRUCTED DRAWING 47W8+8-1047111'848-10 REYISIONREVISION S. E E (Vl168) (VI168) TT-.31-47 TT-Jl-47 (VI159) (V1159) TCV-JI-108 TCV-31-10B F F 22 3-lo-oq 3-/0-09 TO REACTOR BLDG UNIT 1 0-32-.379 ~ EL 71.3' 0-32-381 COMPANION DRAWINGS: FACILITy (V+.37) 1-47W848-1 THRU -9, 1-of.7W8+8-1 -S, -11. :-12---FLOW
--FLOW DIAGRAMS PROJECT FACILITY l-lCV-J- 1 56' (V407) POWERHOUSE G l-LCY-.3-172 1-LCY-J-172 UNITS 1 &. 2 TITLE TITL.E TRAIN A MECHANICAL
[F~R~O"~AU~X~8~LO~G~U~N~'T~'~~~17~~~~~~~~L-____r--C~~~~~-{] 1-4711848-1 CDORD C-B (VS83) l-PCV-1-23 FLOW DIAGRAM W:~~~336 (Vl.315) (VI315) 1-PCV-J-122 1-?GV-J-122 rC-31-.336 TC-31-336 FCO-3\-336 FCO-31-336 (V992) 1-32-.38.30 1-32-3830 1-32-43+ CONTROL AIR (V438j (V4.38) FCO-31-30 \~n~~90-107* D-[><J-r---------~
\~~t~~9(]-107* D-I><J-,------------'
l-LCV-.3-16+ l-LCV-J-164 1-32-3631 1-32-.3831 1/2"! 3801 (Vl082) WATTS BAR NUCLEAR PLANT
~~rg~~3-164A
\~rg&~3-164A (V883)
(V863) l-FCV-90-111 1-FCV-90-111
~~g~~65-81 l-PCV-65-86 l-L-420 TENNESSEE VALLEY AUTHORITY Q (V989)
TT-31-335 TT-Jl-3J5 DESIGN INITIAL ISSUE ENGINEERING TC-.31-J.35 TC-31-335 APPROVAL FCO-31-JJ5 FCO-31-335 (V888) DRAFTER CHECKER H (VB88) l-FCV-90-117 l-FCV-90-1 "7 RO ISSUE PER 1 H.R. PERSINGER
...-_ _ _ _ _ _ _ _.J EL 692~
692~, 0~¥g:10-149 o~¥t6:10-1 +9 g ~ =- E!...A.!~ gJ ..N.E.Il.L_ M:.!'.:....
~!...A.!!.N.E.Il.L_ ~:.!':... IT!.tt.4QN IT!i.MQ.N__
~--------------~ I I
(V853) O-FCO-JO-2BO O-FCO-30-280 DESIGNER REVIEWER WBEP 5.17 cl 826 'SO
'90 010S
<l RIMS 0109 .376 376 2 H.R. PERSINGER l!1~=-QttM'MA1L !!.£,--b!J@ __ _ 3 L.W. BOYD 80'1'0 I
y-EL 737' I I 0-J2-375 SYSTEM TRAINS A ~&. B ISSUED BY: DATE I~CONTROL BLDG _..'!:....~!A~O.:..G..!.~~~~,~R~ _.:!~~!A~O_G...!.~~~~~R~ __ _ 1-26-S0 1-26-90 85 M 1-47W848-10 R22 PRINTS REQ'O REO'D R _ s~~~ , ~~~E
~RIZ~ H H
3 4 5 8R eR OR. PRO~ A8 CE EE ME WE NEN£ FE HE SE TE IE 8L BL 9F 18 we sasa BR OR PROJ .0.8 AS CE C£ EE 1£ ME NEHE FE HE SE Sf TE BLBl BF we so SF W8 CAD MAINTAINED DRAWING (CONFIGURATION CONTROL DRAWING)
t-£- L L9MLt- L 2 3 4 5 6 7 8 9 10 11 12 NOTES:
- 1. F FOR SYMBOLS LS AND GENERAL NOT NOTE SHEET 1. I.
- 2. ID TRIP OPERATES AT o o AND RESETS AUTOMAT ICALL Y. I.4ECHANICAL MECHANICAL
< EO OPERATES AT 125% MUST BE RESET I.4ANUALLY.
IotUST I.1ANUALL Y. J.
- 3. t.I TL WHEN IN AUX CHTL AlJX 'eN CN ISS IICCOMPLISHED ACCOMPLISHED AT THE CONTROLLER.
~
- 4. Ace ACC NT SIGNAL TO TRAIN A VAL ISS TRAIN-B TRIIIN-8 BUFFERED TRAJN-8 VALVES WILL BE 1R TRAI B WITHITH TRAIN-A BUffERED.
BUFFERED.
~
A AUX
~ijtl 6. UCjFIC-X-XXX-A IN I.4ANUAl UC/FIC-X-XXX-ll MANUAL MODE WILL CAUSE OUTPUT OF LCjFC-X-XXX-8 o TACK OUTPUT OF LCjFC-X-XXX-A.
A
~
LC/fC-X-XXX-A.
~ 7.0~C A C 'HEN C IS PRESENT, INPUT A IS ALLOW[Q ALLOWED TO PROCEED (8-A). (B>-A). WHEN C IS
~~
A T NOT PRESENT INPUT 0 IS ALLOWED TO PROCEED (6=0, (8=0, IIA IS BLOCKED). o 0 0~ in u
~
B.
- 8. TRANSFER FROM AUTO TO MAN. OR t.lMi MAN TO IIUTO AUTO IS BUI.4PLESS.
BUt.lPLESS. TRIINSFER IS WHEN TRANSFER. MADE FR0L4 FROM AUTO TO I.4AN.MAN. LAST OUTPUT IS MAINTAINED *UNTIL IT IS MANUALLY t.!ANUALLY *,
~
u Z w
~
%. CHANGEo.
CHANGED. WHEN TRANSfER IS MADE FROM TO IIUTO AUTO OUTPUT. FROt.! MAN TO AUTO. OUTPUT CHANGES SLOWLY
-~- 9. SEE CONTRACT NO. 89NLC-755HA 89NLC-75514A VENDOR DOCUMENTS FOR INFO. ON LIC'S, LC'S, LC*S, FIC'S FIC'S.l.4. FC'S.
- 10. WHEN TRANSFER SWITCH IS OPEN, OPEN. 1I/!.1 AIM CARD IS TRANSFERRED TO AUTO MODE.
r~~--- -~-----~---------~--------- ---------------, I
+=;::F!~~~ffi=J~
I I CLOSE~ FCV-J-191. FCV-3-191. 192* 192 I B I I 47W61'-3-6 47'611-3-6 B I I I I I I I I TRIIINA TRAIN A I
. L-b/I L_~V/I FM I
I I I START SIGNAL TO MOTOR DRIVEN AUX I I I
~y18
<4-5-578 I I ~~~Dr~n:6~v~~-~A-A
~~~D~~n:6~V~~_~A-A :l :
I I I I r--~~~~;~~~~------------- i--~~~~;~;~~------------ I A-7 TR A I I I I I ~!1 GQCl 11 I I
~---1..1
~---l..l c !II !:+
I I I I GQCl ~I I I c I -Q1-____ s~~g09 --~! I I
~ L-.,---aJ------{T I L-...,. __ -{T I ~--J: I I I II If{t)_~
If(t)_~ I I AUXf'l( AUX FYI I L--~i VII I o I
- : : RESET I TR B I 1 I I I I
~
S I I I VII r---..J r---...l I I I I LM I I
'I . ~~
L_~ 6 IL __ l L 3-16"rC 3-16-'rC I I I FROM 2/3 I LO-LO LEVEL 66 I TO BLOWDO~N SLOWOOy(N I I SEE 1-47W611-3-8 ISEE 1-47W611-3-8 ir~~:~~T~~:-------- --l L.. _ _ _ -, L--- I POTENTIO'-lETER POTENTIOMETER fOR OVER SPEED +I I ISOLATION VALVES SH 1-47WS11-1-J 1-47W611-1-3 I FROM L__ __..J 1 i II LIC-J-IH LIC-'3-IH
~~2~~ ____ --~t-------l I I -{'
'{' I I
I I I I I I I I I IL ________ _ I L_ L_ I I II o D I I I I r-------..I r-------...1 o D I I I I I FROM STEAM STEA!.I GEN 2 I : (SAt.I[ (SAME AS GEN 1) I) I
~--~~~~~~~~~=t::::::::::::::::::::::::::::::::::::::::::::::::::::;::::T_C=>::~{>~--~-f"r-CX:J-~rE~::~~~----~!
1-:::-....!:=====f===E==============:;;=====;:====::;:=::JG=;::;:::T-r::><l--l~f-f'-J--£><J-....lr~~:;::::~;-::-_..J! FR~STEAMGEN3 I I I I L- FROM START UP FROM (SAME STEAM I (SAIotE AS GEN I) GEN 3 1) 1J I r----- FEEDWATER SEE 1-47"11'611-3-2 II I I I I I I 1II r-------L1----------------h I1 I1 I1 l r~Z:E S:~A~E~Ef)
! r~;::E Sl~A~E~E~) 4+
IL __________________________ ~I
~
I I1 LT II I I* I LT 3-156 STEAt.! STEAM GENGEN 2 2 II I II I1 II I II I1 START SIGNAL TO STIIRT II I I II I I ~~Jg~A~~~v~~M~U~e-B Wlg:A~~~V~~M~U~8-8 l_ ....... r-------, E
- II ,I L ___________________
L---- fI --------------,
~ :II I SEE 1-+7'1611-3-3.
1-+7W61'-3-3, E J! E 9 j ! e J II E H[ H:L_______J I I LT I TRAIN 8 I I I 3-173 I (SAt.!E (SAME AS JI
- I I :I I START SIGNAL TO TRAIN A)
TRIIIN [: I I I I I I I I ~~f~i~~~~~~~~~
~~i~~~~~~~~~~~' ~~\ ~~:S L _______
r----J I I I L-------------~,----@l, L-------------~---_©l I L---,_---' L....-,_--' 3e-SOA :----' I
©,
I FRO~ MAIN A ~~~~~~~IN I I
.--!-----+--!>'<l-------------+----()l<:J_---"-....,...'
r-t-'-----+I---c>'<r------------+---C'<l------'--i H-t ~nA~WG I FROM START UP TRIP '" TRIP o!t. 1-47W611-1-2 F2 1-<7'611-1-2 F2 I F£[DWATER SEE HEDWATER THROTTLE VALVE I 1-"r7W611-J-2 1--'r7W61'-3-2 MECHANICAL I
~2A OVERSPEED TRIP LINKAGE rsAA(EIAS~1 REfERENCE REFERENCE DRIIWINGS, DRAWINGS:
"rJWS10-3-J----------CONTROL
+7W610-3-3----------CONTROl DIAGRAM DIAGRAt.! AUX FEEDWATER ILLCV 164 I
____ ...1 "r7W610-"r6-1---------CONTROl 47W610-+6-1---------CONTROl DIAGRAM FEEDWATER 45N600-3-2.3 4. 45N600-3-2.J &. 5----WIRING DIAGRAM MAIN .l. &. AUX FEEDWATER FEEOWIITER
+5W600-+6-6---------WIRING DIAGRAMS "rSW600-46-6---------WIRING DIAGRAI.4S FEEDWATER PUMP 4. &.
F TURBINE SCHEI.AATICS SCHEMATICS F THIS CONFIGURATION CONTROL DRAWING SUPERSEDES UNIT 1 I AS-CONSTRUCTED DRAWI~G -'r7W611-3-+ REVISION o. DRAWr~G HW61'-3-4 O. 1/-/7-0~ 11-/1-o(P AUXILIARY FEEOWATER fEEDy/ATER STEAl.! STEAM SUPPLY FRO,,", FROM CONDENSATE TRANSFER STORAGE TANK CIRCUIT SEE 1-47111'611-3-,3 1-47W611-3-3 1-47W611-1-1(T-2) 1-47W61 1 1 (T -2) PROJECT fACILITY POWERHOUSE G AUXILIARY UNITS 1 11 a 2 FEEDWIITER FROM FEEDWATER TITLE MOTOR DRIVEN AUX FW lA-A FYI PUMP 1 A-A SEE ELECTRICAL 1-47'(611-3-3 1-47W61'-3-3 LOGIC DIAGRAM AUXILIARY FEEDWATER SYSTEM WATTS BAR NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY Q DESIGN INITIAL ISSUE ENGINEERING APPROVAL DRAFTER CHECKER H FSAR G.J.BARNFlELD G.L8ARNFIELO C.E.THOMPSON C.L THOMPSON RO ISSUE PE PERR 1 D_P.PE:RRER D.P.PERRER FIG. 10.4-20 oi. a: DESIGNER REVIEWER W8EP 5.17 "A:. RIMS WBEP 826 '90 0,329 RJMS 0329 ,377 377 2 R. T.GRAHAI.4 l.GRAHAM FIG. 7.3-3 SH 2 M_L.CHAPMAN M.L.CHAPMAN O.F D.F .FAULKNER 3 LF. TORTORAjMCB J -'
- TORTORA,lMC9
( ISSUED BY: By, R. M _JOHNSON FOR WSR R.I.A.JOHNSON rSR DATE 7-23-90 85 1-47W611-3-4 R17
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~I~~T SIZE H SR OR PROJ AS CE EE tolE NE FE BROOPROJABCEEEME n; HE SE TE BLBFweSQ ilL IIF III SQ CAD MAINTAINED DRAWING I (CONFIGURATION CONTROL DRAWING)
~~~~-~----
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 17. 17.077 077 AK2.03 017 Given the following plant conditions:
- Unit 1 is at 100% power with all equipment aligned and functioning normally.
- Generator reactive load indicates 70 MVAR IN as indicated on 1-EI-57-8, MEGAVARS.
- A grid disturbance occurs which results in a gradually lowering grid frequency.
As a result of the above conditions, how will the identified generator parameter indication be affected? Generator Volts 1-EI-57-15 1-EI-57-15 Exciter Amps 1-EI-57-7 A. Rising Rising B. Lowering Lowering C. Rising Lowering D~ D!' Lowering Rising Page 44
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect. Plausible, since the exciter amps response is correct. It is also plausible that an applicant could believe that attempting to raise generator voltage would be the correct automatic response, since there is a parameter on the grid lower than what it should be. Raising generator terminal voltage, in theory, would tend to raise grid frequency, but this is not what the generator automatic control circuit is designed to do. It controls terminal voltage at a setpoint by varying exciter amps. The applicant could apply the correct exciter amps response (i.e., rising) and believe that this would result in a higher generator terminal voltage (which it normally would if grid conditions were stable and within limits), however, because of this grid disturbance the generator control circuitry is attempting to maintain terminal voltage at a setpoint, and due to the nature of the grid disturbance (lowering frequency), it does this by raising exciter amps. B. Incorrect. Plausible, since the generator control circuitry will wi/l attempt to maintain terminal voltage at the setpoint. Since no single generator connected to a grid is capable of maintaining a "stable" grid / terminal voltage, the actual terminal voltage will be lowering, if the grid voltage is wi/l is lowering (per the conditions given in the stem). A lowering exciter amps response is incorrect, but plausible, since the applicant A may believe the lowering grid frequency is predominant and also causes exciter amps to lower. C. Incorrect. Plausible, if the applicant confuses generator amps with voltage; i.e., if ( grid frequency starts to lower and generator terminal voltage sags along with it, then amps would rise. Applicant may confuse this amps response with voltage and select this answer. D. CORRECT. As grid frequency starts to lower, generator terminal voltage wi/l will tend to lower also. The voltage regulator is in automatic, and will attempt to control terminal voltage at the setpoint, by raising or lowering exciter current. In this case, grid frequency lowers, terminal voltage lowers, and the voltage regulator will raise exciter current in an attempt to maintain terminal voltage stable at the setpoint. Page 45
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 17 Tier: _1_ _1__ Group 1 KIA: 077AK2.03 077 AK2.03 Generator Voltage and Electric Grid Disturbances Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Sensors, detectors, indicators. Importance Rating: 3.0 / 3.1 10 CFR Part 55: 41.4,41.5,41.7,41.10/45.8) 10CFR55.43.b: Not applicable. KIA Match: Applicant must analyze conditions involving a lowering grid frequency, and apply knowledge of generator voltage control to make a conclusion about the expected system response. Technical
Reference:
- 1. SOI-47.02, "Turbo-Generator Startup Operation,"
Rev. Rev. 63
- 2. Lesson Plan 3-0T-SYS057A, 3-0T-SYS057 A, " Generator Excitation, Voltage Regulation and Ground Detection" Proposed references None to be provided:
Learning Objective: 3-0T-SYS057A, 3-0T-SYS057A, "Generator
" Generator Excitation, Voltage Regulation and Ground Detection," Obj. 15 - Discuss how generator excitation affects reactive load.
Question Source: New x X Modified Bank Bank Question History: New question Comments: Page 46
WBN Turbo-Generator Startup Operation SOI-47.02 Unit 1 Rev. 0063 Page 35 of 98 Date _ __ INITIALS 5.4 Placing Generator in Service and Synchronizing with the Exciter Regulator On (continued) NOTE Step 5.4[16] matches the "Auto" Voltage Adjust output to the "Base" Adjust output previously established, and allows the Base Adjust to follow the auto adjust output after the regulator is in auto. IStart of Critical Step(s) [16] PERFORM the following: [16.1] ADJUST 1-HS-57-22, EXCITER VOLTAGE ADJUSTER, to obtain "0" on 1-EI-57-12, V REG XFER BAL VOLTS. [16.2] PLACE 1-HS-57-20, EXCITER REGULATOR, to ON. IEnd End of Critical Step(s) [17] NOTIFY Northeast Area Dispatcher (NEAD) that that generator is ( READY to SYNCHRONIZE. [18] ENSURE VALVE POSITION LIMIT set 15% OPEN. [19] COORDINATE Turbine operation with Operator on feedwater and reactor controls. [20] PUSH TURBINE MODES lOOPER AUTO SYNC" button. [21] ADJUST 1-HS-57-21, GENERATOR SPEED OPER AUTO SYNC [1-M-1], to obtain SLOWLY moving SYNC-SCOPE (1-XI-57-1 or SYNC PNL 1 on ECB 8 if using PCB 5064 to sync) in the FAST direction (one revolution every 10- 15 seconds). [22] ENSURE 1-EI-57-2, INCOMING VOLTAGE, and 1-EI-57-3, RUNNING VOLTAGE, are EQUAL.
WBN Turbo-Generator Startup Operation SOI-47.02 501-47.02 Unit 1 Rev. 0063
- Page 72 of 98 Date_ __ INITIALS 8.15 the Voltage Regulator Into Service Switching the
[1] INFORM NEAD that the Main Generator Voltage Regulator in AUTOMATIC AND RECORD notification in operator log. [2] ENSURE 1-HS-57-20, EXCITER REGULATOR [1-M-1], in TEST, AND* AND
- CHECK Amber light LIT
- CHECK Green Light OUT NOTE To prevent a change in machine voltage, the regulator is balanced or zeroed by the operation of the voltage adjuster (not the base adjuster). This adjustment, with 1-HS 20, EXCITER REGULATOR, in TEST position, has the effect of nulling the regulator to balance the output meter.
[3] ADJUST 1-HS-57-22, EXCITER VOLTAGE ADJUSTER, to obtain "0" on 1-EI-57-12, V REG XFER BAL VOLTS. [4] PLACE 1-HS-57-20, EXCITER REGULATOR, to ON AND
- CHECK Amber light OUT
- CHECK Red Light LIT
3-0T-SYS057A Rev. 1 Page 13 of35
- x. LESSON BonyBODY Instructor Notes Normal Operation
- 4. NonnalOperation Att. #2 SOI-47.02 Appendix E
- a. Operation within the capability curve limits the Obj.14 following conditions; (1) Stator differential expansion.
(2) Hot spot temperatures in the stator and rotor windings. (3) Hot spot temperatures in the stator core. (4) Temperature differential across the winding insulation.
- b. The capability curve is divided into three regions: Obj. 14 (1) Operation between 0 to 0.900 pf (overexcited),
is limited by rotor winding temperatures. This produces a lagging pf, or outgoing V ARs. (2) From 0.900 pf (overexcited) to 0.950 pf Obj. 14 & & Obj. 15 ( (underexcited), operation is limited by stator In this region the field will vary winding temperature. Operation in this portion with load & pf and will always of the curve corresponds to constant stator be less than the maximum amperes. V ARs could be incoming or outgoing allowable value. in this region. (3) Operation from 0.950 pf (underexcited). to 0 pf Obj. 14 & & Obj. 15 is limited by stator core temperature. This produces a leading pf, or incoming VARs. B. VOLTAGE REGULATOR SYSTEM
- 1. Purposes Obj.l
- a. Maintain a stable generator output by controlling Rotation speed is fixed by the firing circuits of the power amplifiers. System frequency, and the number of conductors is fixed (1) The generator output is directly proportional to at thetime the time of manufacture, the strength of the rotating magnetic field, the therefore, the only variable is speed of rotation, and the number of conductors the strength of the magnetic cutting the field. field.
(2) The strength of the generator field is dependent Obj.9 on the excitation current applied to the generator rotor windings. Therefore, the generator output directly follows exCiter field
3-0T-SYS057A 3-0T -SYS057A Rev. 1 Page 170f35 17 of35 ( x. X. LESSON BODY Instructor Notes (2). VoltslH~rtz Limiter (2) Theory of Operation - The Volts/Hertz Obj.15 receives a three phase voltage input from the voltage regulating PT on the machine output. A See related information in main voltmeter on the front of the logic drawer gives single line lesson plan. an indication of this potential. This voltage VoltslHz Volts/Hz limiting only in effect input is processed to yield a voltslhertz limiter when voltage regulator is in the output signal if required, which is sent to the "ON" position. signal mixer. This limiter output is intended to correct the low frequency, high voltage condition before it reaches protective magnitude.
- e. Minimum Excitation Limiter Obj. 15 Obj.15 (1) Purpose - By establishing a minimum 45W573-2, G2 excitation level, the effects of incoming reactive (inductive), load which tends to cause stator heating through current lag, can be reduced. Extensive overheating could result in insulation deterioration.
(2) Theory of Operation - The Minimum Excitation Limiter compares the three phase voltage and 'B" phase current inputs. This is compared to a minimum set point, and the difference is sent to the signal mixer for control. The magnitude of the error signal can be observed on a meter located on the face of the Logic drawer.
- f. Maximum Excitation Limiter.
Limiter Obj. 15 Obj.15 (1) Purpose - Maximum excitation is also based on FIG F/G 3/4 reactive load. Increasing excitation means more current to the generator field, which will decrease stator heating. A point is soon reached where the heat buildup in the rotor greatly offsets the advantage of increased outgoing reactive. It is this parameter which was considered in the upper portion of the generator capability curve.
r
, ~)
( 0 .i_ _ __ Control MVAR Control; ~ Operation of voltage regulator in MANUAL or BASE ~Operation ADJUST is undesirable.
~ It will not respond to changing generator output voltage or MVAR load in this mode
~ All the automatic 'regulator regulator protections and controls are removed in this mode and must be performed by the operator. ~ Procedures require notification of the NETO if prolonge-d prolonged operation in manual is required so he is prepared for possible manual compensatory action in the event of a transient.
~
! \
~------------------------------ -------------------------------
Load, Load ConJtrol Control I
~ Generator load is controlled using the turbine controls to adjust steam flow, flow to the turbine, increased steam
. results in increased MWoutput.
MIVAR MVAR Control
~MVARs ~ MVARs are controlled by generator field strength. ~ Increasing field strength, Voltage Adjuster to RAISE, will cause generator to pick up lagging reactive load. ~ Going to LOWER will cause it to reduce those outgoing MVARs. Continuing to LOWER will result in incoming MVARs. ~ Reactive Current Compensator automatically adjusts field as load is changed to maintain MVARs constant at the selected setpoint in normal automatic control.
.---- --- , - I - t-tt
~~OY\ Qtruc'\
~SOY\ ~\CU'\ -to" \\~~uatorl
\\~~ucrl-orl r<\crtct) r<\cr\ct) -\~-10~-~-~~~57S CU\d -l~--\M~_~~~57S Rev. 0 Page 140f31 X. LESSON BODY Instructor Notes (2) The stator and field have class "B" insulation with the following temperature limits:
(a) Stator 96° C (b) Field 11 0° C (3) The generator has a conventional oil sealed H2 Covered in a separate lesson cooling system. Normal H2 pressure is 75 psig. (4) Stator bars are water cooled. Covered in a separate lesson (5) Excitation is provided by a shaft driven Covered in a separate lesson brushless excitation system complete with a solid state voltage regulator.
- 2. Paralleling the Generator to the Grid Obj.3
- a. When paralleling the generator to the grid, three parameters must be verified. The generator output and the grid must have:
(1) the same voltage ( (2) the generator frequency should be slightly higher than the grid (3) the voltages in phase (synchroscope)
- b. Recall that a resistor represents the power OTT -3 OTT-4 OTT-30TT-4 dissipation or consumption in watts. Both the inductors and capacitors are, ideally, energy storing devices and do not dissipate or consume any energy. The power that is transmitted to these elements is called V ARs or Volts-Amps Reactive.
- c. When a generator is supplying a load it must supply all the Watts and VARs VARs required by the load. Watts are true power, V ARs are reactive VARs power (energy swapped between generator and the load).
3-0T-SYS057S Rev. 0 Page 15 of31 X. LESSON BODY Instructor Notes The diesel generator is
- d. The total power output by the generator is a supplying 708amps at combination of Watts and VARs. This total power 6900volts. How much apparent is called apparent power with the units VA. The power is being produced?
main generator at WBN is rated at 1,356,200 KV KVA A True power if pf=0.9? and 0.9 power factor. What is the ratedKw output? P=708a*6900v 4885KVA P= 4885KV A *.9 = 4397Kw P'FPA*pf = 1,220,580Kw
- e. The voltage between the grid and the generator OTT -7: House curves OTT-7:
should be equal, to ensure no VARs will be supplied from the generator and no V ARs ARswill will be Obj. 4.a picked up by the generator. Changing generator output voltage (excitation) will change the VAR loading when it is connected to the grid. (1) Ifthe generator output voltage is lower than grid voltage; then the generator voltage will be increased to grid voltage and the generator will become a VAR load to grid. (2) Ifthe generator voltage is higher than grid voltage; then the generator will pickup VARs from the grid and would try to increase the grid voltage.
- f. Generator frequency will be slightly higher than the OTT-7 grid frequency when paralleled, to ensure that the of the Mwatt load from generator picks up some ofthe Obj.4.b the grid. (Syncscope (Syncs cope rotating slowly in the 'Fast' direction).
(1) If the generator frequency is lower than the grid, then the generator will become "motorized". If the difference in frequency is large enough a reverse power condition will occur and cause the generator to trip. (2) Frequency depends on the speed of the turbine, Refer to SOI-47.02 App D&E and turbine speed is changed by increasing or for frequency and excitation decreasing steam flow. Once the generator is limits Att #2 connected to the grid, changing steam flow will not change the speed but increase or decrease the Mwatt output of the generator.
3-0T-SYS057S Rev. 0 Page 170f31 17 of31 ( x. X. LESSON BODY Instructor Notes
- c. In order to change the power being supplied by the OTT-7 OTT-7 main generator, two things are changed.
(1) (1) To change the MWatt output, (watts are real Real power in and real power power out), then more steam (real power in), out. must be put on the turbine. As frequency increases load (current) increases, the cemf Lenz's Law increases and returns ~ 1800 rpm with a returns to -1800 higher steam flow to the turbine and a higher current flow from the generator. VAR V AR (Reactive Power) also (2) To change the VAR output, the excitation on called the imaginary power the main generator is increased or decreased. since no work is done. AlC Motor Operation
- 4. NC
- a. The induction motor is the most popular AC motor Synchronous motors will not be because of its simple construction, good operating discussed.
characteristics, and reliability.
- b. The induction motor consist of two parts: the stator OTT-6 oftwo
( and the rotor. The stator is connected to the 39 38 AlC supply. The rotor is not electrically connected, NC but its current is induced from a rotating field developed by the stator.
- c. The stator is wound such that with 38 39 power, the fields appear to rotate, at a speed magnetic fields dependent on frequency of the supply and the number of poles in the stator. This rotating magnetic field induces currents on the rotor. These currents flowing through the rotor cause a magnetic field, which forces the rotor to rotate.
- d. There are two types of rotors used in induction motors. One is a squirrel cage rotor and the other is a wound rotor, squirrel cage rotors are used in motors up to approximately 50 HP.
- e. The squirrel cage rotor is the most popular. The rotor of a squirrel cage construction is a laminated core with conductors placed parallel or approximately parallel on the surface of the rotor, rotor.
At each end of the rotor, the rotor conductors are short circuited with end rings. The rotor with end short rings looks similar to a squirrel cage in appearance.
3-0T-SYS057A Rev. 1 13 of35 Page 13of35 x. X. LESSON BODY Instructor Notes
- 4. NonnalOperation N onnal Operation Att. #2 SOI-47.02 Appendix E
- a. Operation within the capability curve limits the Obj.14 following conditions; (1) Stator differential expansion.
(2) Hot spot temperatures in the stator and rotor windings. (3) Hot spot temperatures in the stator core. (4) Temperature differential across the winding insulation.
- b. The capability curve is divided intointo three regions: Obj. 14 Obj.14 (1) Operation between 0 to 0.900 pf (overexcited),
is limited by rotor winding temperatures. This produces a lagging pf, or outgoing V ARs. 0.950 pf (2) From 0.900 pf (overexcited) to 0.950pf Obj. 14 & Obj. 15 (underexcited), operation is limited by stator In this region the field will vary winding temperature. Operation in this portion with load & pf and will always of the curve corresponds to constant stator be less than the maximum amperes. V ARs could be incoming or outgoing allowable value. in this region. (3) Operation from 0.950 pf (underexcited). to 0 pf Obj. 14 & Obj. 15 is limited by stator core temperature. This produces a leading pf, or incoming VARs.V ARs. B. VOLTAGE REGULATOR SYSTEM
- 1. Purposes Obj.l
- a. Maintain a stable generator output by controlling Rotation speed is fixed by cif the power amplifiers.
the firing circuits of System frequency, and the number of conductors is fixed (1) The generator output is directly proportional to at the time of manufacture, the strength of the rotating magnetic field, the therefore, the only variable is speed of rotation, and the number of conductors the strength of the magnetic cutting the field. fi~ld. fidd. (2) The strength of ofthe generator field is dependent Obj.9 the generator on the excitation current applied to the generator rotor windings. Therefore, the generator gep.erator output directly follows exciter field
3-0T-SYS057A 3-0T-SYS057A Rev. 1 Page 14 of35 x. X. LESSON BODY Instructor Notes current changes.
- b. Monitor exciter operation, generator output, and Obj.l Obj. 1 provide inputs to protect the generator/exciter.
By changing the output potential as conditions require, tripping the exciter field breaker if needed, and providing alarms alanns for abnormal abnonnal conditions.
- 2. Component Descriptions/Operation
- a. Exciter Cubicle -located
- located on elevation 729, houses OTT-9 the components of the WTA regulator and exciter:
(1) Base Adjust Assembly (2) Regulator Logic Drawer (3) Voltage Adjust Assembly (4) Protective Drawer Breaker, power amps, firing (5) Exciter Field Breaker.(Device Breaker (Device 41) circuit and shunts discussed in (6) Power Amplifiers (3 drawers) exciter section. (7) Firing Circuit Drawers (2) ( (8) Shunts (2) for Measuring Exciter Field Current (9) Fuses from PMG Output to Firing Circuits
- b. Base Adjust Assembly (70B) Obj. S.e (1) Purpose - controls the circuitry which adjusts 1-45W573-1, E2 the SCR firing durations.
3-0T -SYS057A 3-0T-SYS057A Rev. 1 Page 18 of35 ( X. LESSON BODY Instructor Notes (2) Theory of Operation - This module monitors exciter field current through an isolation transducer. It compares the magnitude of this signal to a setpoint, and gives an error signal proportional to the amount by which the setpoint is exceeded. The error signal is fed to the signal mixer, and a meter on the face of the panel.
- g. Voltage Error Detector - This module compares a 45W573-2,03 45W573-2, G3 voltage signal from "A" phase of the voltage regulator PT to the outputs of the reactive current compensator. This gives a feedback signal to the error detector, modified by any reactive load on the system, either inductive or capacitive. This feedback signal is compared with the reference signal supplied by the voltage adjust. This error signal is then sent to the signal mixer, the output of which goes to the firing circuits to automatically adjust excitation to compensate for the changing reactive load (i.e. any change in actual generator voltage as compared to the reference voltage from the voltage adjuster).
- h. Reactive Current Compensator H2/3 (1 ) The reactive current compensator is a set of (1) mutual coupled inductors.
(2) The machine current from the "B" phase CT, is passed through the primaries of both inductors in series. III senes. (3) The current derived from the Voltage Reg PT from the "B" and "C" phase voltages, is
. supplied through the secondaries. The result of the interaction between the two currents
.(difference)
(difference) is taken off offthrough through variable taps s~nt to the voltage error detector as an and sent reactive load... indication of changing reactiveJoad (4) This system reacts to either an inductive or capacitive reactive load.
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 18. WlE04 W/E04 EA1.3 018 Given the following:
- The crew has entered ECA-1.2, "LOCA Outside Containment" from E-O, "Reactor Trip or Safety Injection" based upon high radiation levels in the Auxiliary Building.
- After closing 1-FCV-74-33, 'RHR crosstie valve,' and 1-FCV-63-93, 'RHR Train A cold leg injection valve,' the following conditions exist:
- RWST level continuing to drop.
- RCS pressure slowly increasing.
Which ONE of the following indicates the status of the LOCA and the subsequent action that would be taken? A'I The LOCA is isolated; Ar:' Stop RHR 1 1A-A A-A pump and close 1-FCV-74-3, RHR Pump 1A-A suction valve. B. The LOCA is isolated; Leave RHR 1A-A pump running and injecting through the RHR crosstie valves, 1-FCV-74-33 and 1-FCV-74-35. C. The LOCA is NOT isolated; ( Re-open 1-FCV-63-93, and close 1-FCV-63-94, RHR Train B cold leg injection valve. D. The LOCA is NOT isolated; Leave 1-FCV-63-93 closed and close 1-FCV-63-94, RHR Train B cold leg injection valve. Page 47
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 ( DISTRACTOR ANAL YSIS: A. Correct, The LOGA Gorrect, LOCA is isolated and the 1A-A RHR pump is directed to be 1-FCV-74-3, closed to stop the loss of stopped and its suction valve, 1-FGV-74-3, inventory from the RWST. B. LOCA is isolated but the 1A-A RHR pump is to be stopped Incorrect, The LOGA and its suction valve closed not left running and injecting through the crosstie valves. These valves 1-FCV-74-33 and 1-FGV-74-35 1-FCV-74-35 are closed in the procedure prior to closing the cold leg injection valve. Plausible because the LOGA LOCA status is correct and the RHR pumps are normally aligned with the cross-tie valves open in order that one pump can meet the EGGS Tech Spec requirement to inject into all four cold legs. ECCS G. C. LOCA is isolated. 1-FGV-63-93 Incorrect, The LOGA 1-FCV-63-93 should not be reopened and the Train B cold leg injection valve, 1-FGV-63-94, 1-FCV-63-94, closed. Plausible because the RWST level is still dropping but it is not due to the LOCA LOGA not being isolated and if the LOGA LOCA were not isolated, then 1-FGV-63-93 1-FCV-63-93 would be reopened and the Train B cold leg injection valve, 1-FCV-63-94, 1-FGV-63-94, would be closed. D. LOCA is isolated. 1-FCV-63-93 Incorrect, The LOGA 1-FGV-63-93 should be left closed but the Train B cold leg injection valve, 1-FGV-63-94, 1-FCV-63-94, should not be closed. ( Plausible because the RWST level is still dropping but it is not due to the LOCA not being isolated and it the LOGA LOGA LOCA were not isolated, then the Train B cold leg injection valve would be closed. Page 48
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 18 Tier: ~1_ _1_ Group _1_ 1 KIA: W/E04 EA1.3 LOCA Outside Containment Ability to operate and I or monitor the following as they apply to the (LOCA Outside Containment): Desired operating results during abnormal and emergency situations. Importance Rating: 3.8 I 4.0 10 CFR Part 55: 41.7 I 45.5 I 45.6 10CFR55.43.b: Not applicable. KIA Match: Applicant is required to evaluate the results of operating actions taken to determine the status of the LOCA and based on that determination, identify if the desired results were achieved and the subsequent actions that are needed to be taken. Technical
Reference:
ECA-1.2, "LOCA Outside Containment," Rev 4 E-O, "Reactor Trip or Safety Injection," Rev 27 Proposed references None to be provided: Learning Objective: 3-0T-ECAO 10 1 3-0T-ECA0101 Material covered but no objective identified Question Source: New Modified Bank X Bank Question History: Seabrook bank question modified Comments: Seabrook 2007 exam ( Page 49
TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT EMERGENCY OPERATING INSTRUCTIONS ECA-1.2 LOCA OUTSIDE CONTAINMENT Revision 4 Unit 1 ( QUALITY RELATED REQUESTED BY: S. M. Baker SPONSORING ORGANIZATION: OPERATIONS APPROVED BY: A.K A. K. Keefer EFFECTIVE DATE: 11/22/06 LEVEL OF USE: CONTINUOUS
WBN LOCA OUTSIDE CONTAINMENT ECA-1.2 ( Rev 4 1.0 PURPOSE This Instruction provides actions to identify and isolate a LOCA outside containment. 2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 Indications Abnormal Auxiliary Building radiation due to LOCA outside containment. 2.2 Transitions Trans itions A. E-O, Reactor Trip or Safety Injection. B. E-1, Loss Of Reactor Or Secondary Coolant. 3.0 OPERATOR ACTIONS ( 2of5
WBN LOCA OUTSIDE CONTAINMENT ECA-1.2 Rev 4 IIStep Step II II Action/Expected Response II II Response Not Obtained
- 1. ENSURE RHR suction from RCS CLOSED:
- 1-FCV-74-1 and 1-FCV-74-9.
AND
- 1-FCV-74-2 and 1-FCV-74-8.
- 2. ENSURE SI pumps hot leg injection 1-FCV-63-156 and 1-FCV-63-157 CLOSED.
- 3. ENSURE RCS letdown isolated:
- Letdown isolation 1-FCV-62-69 1-FCV-62-69
( and 1-FCV-62-70 CLOSED.
- Excess letdown isolation 1-FCV-62-54 and 1-FCV-62-55 CLOSED.
- 4. ENSURE RHR hot leg injection 1-FCV-63-172 CLOSED.
- 5. CHECK RCS press ** GO TO Step 14.
DROPPING or stable.
- 6. CLOSE RHR crosstie valve 1-FCV-74-33 or 1-FCV-74-35.
3 of 5
WBN LOCA OUTSIDE CONTAINMENT ECA-1.2 Rev 4 IStep II Action/Expected Response II Response Not Obtained
- 7. CLOSE RHR Train A cold leg injection IF 1-FCV-63-93 failed OPEN, valve 1-FCV-63-93. THEN
** GO TO Step 10.
151, 8. CHECK LOCA isolated: )J f OPEN 1-FCV-63-9 p~1
- RCS press rising. ,~ I'N
,----,->> JJ~{ ( \-el**
SO Vic I so'lie
\£'** GO TO Step 10.
10'.
?
f
,~ Jr- i.~(l
~Jr- i.~ (llA f (~
LA '1 /
/
- 9. ISOLATE RHR Train A:
- a. STOP RHR pump A-A, and PLACE in PULL TO TO LOCK.
- b. CLOSE RHR suction valve 1-FCV-74-3.
(
- c. ** GO TO Step 15. /"
/1/
/'
//
'/
- 10. CLOSE RHR Train B B cold leg';
led;:' IF1-FCV-63-94 IF 1-FCV-63-94 failed OPEN, injection valve 1-FCV-63-94. 1-FCV-63-94. THEN
** GO TO Step 13.
- 11. CHECK LOCA isolated: OPEN 1-FCV-63-94.
- RCS press rising. ** GO TO Step 13.
4of5
WBN LOCA OUTSIDE CONTAINMENT ECA-1.2 Rev 4 IStep II Action/Expected Response II Response Not Obtained
- 12. ISOLATE RHR Train B:
- a. STOP RHR pump B-B, and PLACE in PULL TO LOCK.
- b. CLOSE RHR suction valve 1-FCV-74-21.
- c. ** GO TO Step 15.
- 13. ENSURE RHR crosstie valves 1-FCV-74-33 and 1-FCV-74-35 OPEN.
- 14. IDENTIFY break location: NOTIFY TSC of failure to identify break location.
- RADPROT surveys.
(
- RHR pipe break lights [M-6].
- ECCS pump flows.
- Aux bldg flood alarms [M-15; light panel, Aux Bldg 757].
- Radiation area monitor recorders 1-RR-90-1 and O-RR-90-12A.
- 15. DETERMINE appropriate Instruction: NOTIFY TSC of failure to isolate break.
- IF LOCA outside cntmt isolated, THEN ** GO TO ECA-1.1,
** GO TO E-1, Loss of RHR Sump Recirculation.
Loss of Reactor or Secondary Coolant.
- End-(
5 of 5
Examination Outline Cross-reference: . Level RO SRO Tier # Question # 17 Group # K/A# E04EK1.3 Importance Rating 3.5 3.9 Proposed Question: The crew has entered ECA-1.2, 'LOCA OUTSIDE CONTAINMENT' from E-O based upon RDMS indication of high radiation levels in the RHR vaults. After closing RH-V14, RH-VI4, 'RHR Train A discharge to the RCS' and RH-V22, 'RHR Train A cross-connect' and placing the 'A' train RHR and CBS pumps in pull-to-Iock, the following conditions exist:
- ECCS flow is decreasing
- RCS pressure is 1100 psig and slowly increasing Which of the following indicates the status of the LOCA and the subsequent procedural action that should be taken?
A. The LOCA is isolated. The crew should transition to E-l, LOSS OF REACTOR OR OR SECONDARY COOLANT, step 1. B. The LOCA is not isolated. The crew should transition to ECA-l.l, ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, step 1. C. The LOCA is isolated. The crew should transition to ES-I.l,ES-1.1, SI TERMINATION, step 1. D. The LOCA is not isolated. The crew should continue with actions in ECA-l.2,ECA-1.2, LOCA OUTSIDE CONTAINMENT. Proposed Answer: A- -
-,-::-A==--::--
A is correct. Per ECA-l.2, ECA-1.2, step #4 if ifRCS RCS pressure is increasing due to successful leak isolation the crew should transition to E-1. . B is incorrect but plausible. The crew would transition to ECA-1.l ECA-1.1 at step #4 of ofECA-1.2 ECA-1.2 ifRCS pressure is increasing due to leak isolation. C is incorrect but plausible. A transition to ES-l.l may ultimately be made to terminate SI based upon isolation of the LOCA , however this transition will most likely be made from E-l not directly from ECA-l.2. D is incorrect but plausible. If the initial actions in ECA-l.2 are not successful in isolating the LOCA additional actions may be taken to isolate valves in the other train. Technical Reference(s): ECA-1.2, LOCA OUTSIDE CONTAINMENT. Proposed references to be provided to applicants during examination: None _ _::"""""_----::_----::~-- _N_o_n_e
----------~-------------
K/A K/ A E04EK1.3 E04 EK1.3 Knowledge of the operational implications of the following concepts as they apply to the Topic: LOCA Outside Containment: Annunciators and conditions indicating signals, and remedial actions associated with the LOCA Outside Containment. Question Source: New Question Cognitive Level: Application 10 CFR Part 55 Content: CFR 41.8/41.10/45.3 Learning Objective: Ll209I, Objective L1209I04RO Lesson L1209I, Ll209I04RO
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 19. 024 AA2.02 019 Given thethe following:
- Following a turbine runback, AOI-34, "Immediate Boration" is entered to restore control rods above the LO-LO insertion limit.
- The normal boration is unavailable due to "VCT Makeup Mode" control handswitch being broken and the Emergency Borate valve 1-FCV-62-138 .
valve .. could not be opened from the MCR or locally at the valve. Which ONE of the following identifies the boric acid flow path AOI-34 would next alignment{s) required when using ~- direct to be established and additional valve alignment(s) the flow path? Flow Path Valve alignment A'!' Open 1-ISV-62-929, A'I Open 1-FCV-62-140, Manual Borate Valve VCT Makeup Control B. Open 1-ISV-62-929, Close 1-FCV-62-128, Manual Borate Valve Makeup to the VCT Inlet C. Open 1-FCV-62-135 & 136 Open 1-FCV-62-132 &1 33 CCP suction from RWST CCP suction from VCT ( D. Open 1-FCV-62-135 & 136, Close 1-FCV-62-132 & 133 CCP suction from RWST CCP suction from VCT Page 50
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Correct, the next path directed is to establish flow through 1-ISV-62-929 and the 1-FCV-62-140 valve is directed to be opened to provide a flow path to the CCP suction. B. Incorrect, the next path directed is to establish flow through 1-ISV-62-929 but the 1-FCV-62-128 valve is not directed to be closed. The 1-FCV-62-140 valve is directed to be open. Plausible because establishing the flow path through the 1-ISV-62-929 valve is the path to be attempted next and the desired path is to the CCP suction so closing 1-FCV-62-128 to the VCTwould be appropriate. C. Incorrect, Swapping the CCP suction to the RWST by opening 1-FCV-62-135 & 136 is a path to get boration but it is not the next path to be aligned and 1-FCV-62-132 & 133 being open is proper alignment for the conditions identified in the question. Plausible because aligning CCP suction to the RWST is the method identified to be aligned if the 1-ISV-62-929 path is not successful and 1-FCV-62-132 & 133 being open is a correct valve alignment for the conditions in the stem. D. Incorrect, Swapping the CCP suction to the RWST by opening 1-FCV-62-135 & 136 is a path to get boration but it is not the next path ( to be aligned and 1-FCV-62-132 & 133 being closed is the proper alignment after the suction is swapped to the RWST. Plausible because aligning CCP suction to the RWST is the method identified if the 1-ISV-62-929 path is not successful and 1-FCV-62-132 & 133 being would be closed following the opening of the CCP suction valves from the RWST. { Page 51
11/2009 Watts Bar 11/2009 Watts Bar RO RO NRCNRC Exam Exam -- As As submitted submitted 10/2/2009 10/2/2009 Question Number: Question Number: 19 19 Tier: _1_ Group _1_ 22 KIA: 024 AA2.02 024 Boration Emergency Boration interpret the following to determine and interpret Ability to following as they apply to to the Emergency Boration: When use of manual boration valve is needed. Importance Rating: 3.9 I 4.4 3.9/4.4 10 CFR Part 55: 43.5 I 45.13 43.5/45.13 10CFR55.43.b: Not applicable. KIA Match: Applicant must determine alignment of the system that would allow flow 'When use of manual boration valve is needed' and an additional alignment(s) while using the flow path. Technical
Reference:
AOI-34, Immediate Boration, Rev 23 Drawing 1-47W809-2, R22 Proposed references None ( to be provided: Learning Objective: 3-0T-AOI3400
- 3. Identify 4 methods of Immediate Boration 3-0T -SYS062A 3-0T-SYS062A
- 29. Discuss how to perform manual immediate boration, include when this is necessary.
Question Source: New xX Modified Bank Bank Questioh Question History: New question Comments: Page 52 Page 52
AOI-34 WBN IMMEDIATE BORATION Revision 23 Page 5 of 22 Page.5 3.2 Boration of RCS with CVCS in Service ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Boric acid addition should be noted to assist in determination of reactivity changes.
- 1. INITIATE normal boration to change CB as necessary:
- a. PLACE BA flow controller, 1-FC-62-139, to desired flow rate.
- b. ADJUST BA batch counter 1-FQ-62-139 to ensure boration continues.
- c. PLACE mode selector 1-HS-62-140B to BOR.
- d. PLACE VCT makeup control 1-HS-62-140A, to START.
- e. VERIFY boric acid flow indicated on 1-FI-62-139.
- 2. ENSURE PW to blender isol 1-FCV-62-143, CLOSED.
AOI-34 WBN BORATION IMMEDIATE BORA TION Revision 23 Revision.23 Page 6 of 22 3.2 Boration of RCS with CVCS in Service (Continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 3. CHECK PW to blender DISPATCH operator to CLOSE flow 1-FI-62-142, indicating PW to blender isol ZERO. 1-ISV-62-933, [A4V/713].
NOTE A delay of 15 to 20 minutes may be expected before effects of negative reactivity insertion are observed.
- 4. MONITOR for negative reactivity IF normal normal boration NOT inserting Insertion: negative reactivity, THEN
(
- Neutron flux dropping. **GO TO Step 6.
- T AVG dropping.
- Control rod bank position rising (if in AUTO).
AOI-34 WBN IMMEDIATE BORATION Revision 23 Page 7 of 22 3.2 Boration of RCS with CVCS in Service (Continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 5. IF normal boration effective, THEN
**GO TO Step 9.
- 6. ESTABLISH required emergency boration flow:
- a. PLACE both BA pumps in FAST speed.
- b. ADJUST emergency b. Locally ADJUST borate valve 1-FCV-62-138 1-FCV-62-138 to to obtain required flow. obtain required flow.
( c. CHECK emergency borate Locally OPEN manual, flow on 1-FI-62-137 1-FI-62-137A. A. boration valve 1-ISV-62-929 [Blender Station/713].
~
V ENSURE BA flow control 1-FCV-62-140 OPEN, ENSURE BA to Blender, 1-FI-62-139, indicating flow.
AOI-34 WBN IMMEDIATE BORATION BORATION Revision 23 Page 8 of 22 3.2 Boration of RCS with CVCS in Service (Continued) ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7. IF emergency boration flow established, THEN
** GO TO Step 9.
- 8. ALIGN RWST to CCP suction:
- a. OPEN RWST outlet valves 1-LCV-62-135 and -136.
- b. CLOSE VCT outlet valves 1-LCV-62-132 1-LCV-62-132 and -133.
- 9. REFER TO the following Tech Spec:
- 3.1.1, Shutdown Margin (SDM)---Tave > 200°F.
- 3.1.2, Shutdown Margin (SDM)---Tave ~ :s; 200°F.
- 3.1.6, Shutdown Bank Insertion Limits Limits..
- 3.1.7, Control Bank Insertion Limits.
- 3.4.2, RCS Minimum Temperature for C riti cality .
Criticality.
- 3.9.1, Boron Concentration.
I Z-60BIALt-1 Z-6DBMLV-1 1I 22 I 3 ./t,
/
/ I r 4 I1 5 I1 6 J1 7 1
8 1 9 10 1 11 1 12
,-§~~~!6~c-------,'---j~""H2XQ'--r "Y; @, ~ ~r - - - - - - , NOTES:
-;~'-,Ii<i~)I"'I-F+-~If~-1~'"¥'. l-'~- I
-\~~'
~: ;~E~~OjTl~"~R~n6.
~~E~gDFi~"~R~n6. fa~ fO~ tfrhl-,,
ff1rhl'
~
ANO FOR UNIT 2 11TH "NO WITH THE plJAP ___ 2-{..;j-_2CX+'-c>Y;q.;.:'2X '&"*l_C
- Jo'. . 2X;f-t' lX' SH' BORIC ACtO TANK I J.
- 3.
EXCEPTIOHSNOTED. EXCEPTIONS HOTED. PIPING ON SHEETS 2 THRU 6 IS TVA
".CLASSCHAHGESARESI*O'NINTHETYPICALWA.NNER~
TV" CLASS Il, a, C, 0, C, G, OR t..
....s..rc-L.
~ook ~:391I--i<l-r---<
- 4. CL"SS CHANGES ARE SHOWN IN THE TYPICAL w"NNER COOf! F'-'O A
A E-l0 1,_62_929 1-62-928 1-62-927 5. PIPINCSliO'NINTHISI.IANNER~IS[LECTRICALLYHEATTRACEO. PIPING SHOWN IN THIS MANNER ..o..:,-IS ELECTRICALLY HEAT TRACED. A A H H 1--i<3-"'-r---< f--i<:Jo"'-r---< SH' BORIC ACID TANK 6. ALL TVA PIPl~ PIPI~ SHOIN SHOtN Olt ON SHEETS 2 THRU 6 IS DESIGNED fOR FOR 150 PSIG AND
,~~ '"' ZOO'F ON 2" OJAloIETER 200'F DIAMETER ,l,NO .... NO SloIALLER.
St.I .... LLER. "'NO AND FOIl FOR 210 PSIG AND JOO'f ON J'
... NO .300'F DIAMETER
- 3" OIAMETER rs
~2-62-927 COOR CooR f-lF-l AND
"'NO LARGER. EXCEPT AS SHOWN BY 0, 0. SEE LINE NO. itO. TABLE T.... BLE ON SHEET 1. DESIGN (:f
'<Z,p12-62-927 SKID l£IUNTED IClUNTED PIPING FOR FOft THE CAS GAS STRIPPER BORIC ACID EVAPORATOR EVI\POR"TOR PACKAGE 6~~ IS BY THE VENDOR.
--J
-J VEltOOR.
- 7. DESIGN PRESSURE AND TElotPERATURE TEIolPERATURE fOR FOR INDIVIDUAL INDIVIOUAL TANKS 1""1(5 "NO AND C04POIotENTS CQ,f'Ot.IENTS Of THE HEhI.
'~-;t SEE NOTES 20 .. .t. 22 IS CI'Inl GIVEN IN THE VICINITY rs ITEt.I.
SEE DETAIL A2 FOR FDR UNIT Nffi'Es
- TRANSFERED EQUIPMENT SEE OOrES 20./l 20 .t. 22 8.
B. THE OPERATING 16~'F
- 1SS'f.
OPER"'TING TEMPERATURE TEt.lPERATURE HIR F'OO ALL .... LL HEAT TRACED PIPING AND EQUIPI.IENT EQUIPMENT IS 1-62-9:30 e, 9. VALVES SHOWN IN THIS "'ANNEIi ~ARE t.lANNER -clr-ARE PROVIDED PROVIOED lITH 11TH EXTENSION STEW STEt.I
" 'I"I VALVE OPERATORS.
II ' L L(l...~ cl.\<- \eo.,
\eo. ;:;
;" ~ JX2
'?' I 10 10.....ARROW RRQ. BLOCKS INDICATE LOC LOCATION... TION (:f rs INTERFACES ON THIS SERIES Of lr'r--fll
... ' ICItARCINGPUIoIP DRAWINGS AND ON OTHER REL.a.TED OJ-Gl,.I)1 0
I Y\ ~._~ I--t:+--....J-i I L-1--1<l----II~:fpON 1--1)>/----'-1
~I--i<l---I SUCTION I
CH .... RGING PUMP
~k_
DETAIL A2 11. OR""INCS DELETED OELETED RELATED OR DRAWINGS.
.... INGS.
B ,--- - COOR f-IO 1"-'0 B
~ ~~ ~ ~ 1-O'*H"
~ _- - J JX2 '-62-H" 4-62-4<" I.E
.~
.~~~
~t,... J i2~\!~-~"
lX2
~
J - rr
'/'jj ~
",-.:~~'
6T2-~a18 ,1 '~ ~'9
~;:~.
13. 1:3. ALL CLASS 'C' 'C" PIPING IS SEIShlICALLY SEISt.lICALLY SUPPORTED AKO ANO NON 0'" UNLESS
- ....1.4 l-BLDR-1-O'-"'-@c 1-62-9.31-.A OTHER'Ist OTHERWISE ItOTED TESTED .ITHIN NOTED . *HERE NOTED, WITHIN THE Q'"
NOTEO. CLASS 'C-QA PROCRAloI. PROGRAM.
'G" PIPINC IS TO BE HYDROSTATICALLT HYDROSTATICALLY I-
~i ~'C 14 ....
- 14. ALL LL 'ELDING WELOING TO "'NO AND INSPECTION AND OOClJlr4NTATION OOCUNENT ... TION OF THE VENT a
-fCY N "-m 1-FCV CONDENSER, CONDENSER. DISTILLATE OISTlLL ... TE COOLER AND EVAP. EV ... P. CONDENSER SHALL BE IN IH
~-14'3 rw ~
~_" ACCORO.vtCE .ITH
"'CCOROJ\HCE lITH THE REOUIRE"'ENTS REOUIRE~ENTS fOR TVA CLASS 'C'. "C'.
I-
'~:t~,. fill""-'" ~ ~,"" '-~~;'"
2-144
----------, NOTE 18 1-62-932 62-142 15 15..
- HEIiE THIS NOTE IS REFERENCED, ALL FIELD WHERE rIELO FABRICATION.
F .... BRlCATION ..... ASSEIELY. SSEMElLY, r'-CH-AR-G-'HG-PU-'-P--;1I1 'T':,.....-[:;I<J-----=,.:.....----,r---.JL.-------l L.... 1-62-934 [ill 162933 I EXAWINATION EXAioIIN .... TION AND rs
......10 TESTS fOIl FOR PIPING SYSTEWS SYSTEIoIS PERFOR"'EO PERFORt.lED AfTER ArTER ISSUANCE ISSUM~E OF rs J
R-l0 OF THIS HOI' FLOW DIAGRAI.IDIAGRAM AND .... ND ALL IIORK WORK REOUIRED REQUIRED BY ECN 3815 SHALL .... Ll BE
~~~Tr6~ ~
HOT' " " '-"-939 .3815 SH
~~CpON ,. L..--'-L..
L....... '-'!-"-'-t..
-,...-l..->i--I _-Pl(j...__=2*:...._-,,---.JL-_ _--!1 1--=='", t-="-I~[)Z<'i'-I--<
. ..,-I>I<-I--{ PRIMARY 'ATER WATER WiTHIN WITHIN THE REQUIREIEHTS REQUIREt.£NTS or or CENERAL GENERAL CONSTRUCTION SPECIFICATION NO. G-69. C-69. I-
~c C PUIoIP 471819-1 47.a19-1 0-TANK-62-117 ~ _____ __ r----
.r-------, C
~c COCR COOR E-5 DOWNGRAOE TO NON NUCLEAR SSAFETY
- 16. DO.HCRAOE .... F'[TY CLASS AT THIS POINT IS BASED BASEO ON AN .... N I-C E-11 COO, H '
coo, 1 J/4" BORIC ACID BORIC ACID sj.£G
*0 r-' I LL~
,J EXISTING STATIC LEG IN THE PIPINC PIPING CONFlGURATION.
CONFIGUR .... TION. SEE EN DES CALes C.... LCS eRS BRS
~['~LE:NDER
,~LE;NDER f~NN;L F~NN;L '-52-448All~~~ ~
~
DOWNGRADE OOWNGRMlE (NE8840426-700) (NEB840+26-700) PRIOR TO 'ANY ANY PIPING CONFICURATION CONf'IGURATION REVISION. 1-62-939 lEI ED - ~ 1-62-9.36 1-62-9,36 " VENT VENT 1-62-448,1, ~
.... ~ 1-62-447,1.
1-62-4m PRESS- .... DESIGN PRESS-An" n.t RESIN
~ ~I
,'I~! PRESS-l~O ~j ~"
- 17. L.O.-LOCKED OPEN. CONCURRENT CLOSURE OF l-ISV-62 1-1SV-62 903 95:3 AND 2-ISV-62-953 2-ISV-62-95.3 TEI.IP-200' DESIGN TEIoIP-200"
-;:;r r-__ ._-;;f"_u"
... OESICN DESIGN PRESS-l:10 PSIC DESICN TEJ.lP-250'F DESIGN TEIoIP-250'F PSIG
~_rIT I-FIT "
fILL FILL IS NOT PERIoIITTED. CONCURRENT CLOSUER Of NOT PERJ.lITTED. PERl.IITTED. 0-ISV-62-1080 CLOSURE IS PERJ.lITTED OF 1-15'.1-62-957 1-ISV-52-957 AND 2-ISV-62-957 IS PERl.IITTEO ONLY WHILE UNIT 2 IS
~~ - TANK c~~~LbttCONN
~
UNDER CONSTRUCTION .1TH WITH UNIT 1 Of>ERATIONAL. OPER .... TION ...L. ... AFTER FTER UNIT UNiT 2 BECOUES SAIAPLE CONN CONT ON N N 1-62-942 1-62-9+2
- (E 1-62-9+1 [ill rc-,-62-94,
~\"
U!J 62-142 l@52-142 I-PI OPERATIONAL 0-15'.1-62-1080 O-ISV-62-1080 IS TO REJ.lAIN REl.I .... IN LOCKED LOCKEO OPEN UNLESS PRIOR BECOMES
- 1-47.625-2!..f....£ 1-47'625-2 ~ 1-62-937
'-62-9:37
- ...1*1 1 62940 1 62940
- w~1---1 l-f'T 2-112 APPROVAL APPROV .... L IS OBTAINED FRot.I FROt.I NUCLEAR ENGINEERING.
~J ~tl-~~_-I-~~w-:
- 18. VALVES REFERENCINC REF'ERENCING THIS TNIS HOTE NOTE 'ILL WILL HAVE THE V"'LV[ VALVE 'DIAPHRACIoi'DIAPHRAG~ SUPPORT SHEET" SHElT' f?N:~~:~T SPRAY I COORD F'-1 COORD F-l
~
.. 1 2'
-1.:
[ji]:...1 Iii! ~ I-~I:-__"""'__...J 1-FT Z-I+2 2-142 REI.t:IVED. REI.IOVED. THESE VALVES ARE LIJ.l1TED TEIIPERATURE TEIoPER .... TURE OF 225 PSIG AND LIt.lITED TO A WAXIt.l1J,4 ANO 200"f 2DD"F OR LESS FOR IoIAXIt.lUM OPERATING F'OR .3" OPERATINC PRESSURE AND 3' AND .... ND 4' 4" VALVES AND .... "10
=------s1=~:oJ. __,1 7'l~OR-"-J '--------.:.00-_' ~1-62-J6:1A COORD C-' 1 1-62-~u' 1-62-9~
k-f-h m 1l!L __<+----> g~i~i~E~~~~:26E~lSIG LQRIFICE
~ 1-6Z-J65A 150 PSIG ...
OUALIFIED fOR OUALlf'IEO HOURS PER RIWS a26 AND ND 200"F200'F OR LESS fOR 2' FOR DESIGN PRESSURE ANtI 926 88 0105 957). NO OPERATINC 2" AND AND TEWPERATURE BELOW. VALVES J'
.... ND BELO'.
TEIoIPERATURE ON LIW!TEO OPERATING CONDITIONS EXCEED 3' AND ABOVE ARE LIt.lITEO TIlE TIME BASIS (48 EXCEEO THESE t- Lil ~f ~ --,:m I L.!.J OESICN DESIGN PRESS-150 PS1G
~--------, 'j -1 I TEMPORARY TEt.lPORARY DESIGN PRESS OESIGN TEt.lP-200'!'"
LETOO.N LElDOWN HEAT HEAT 1-62-900 3' .3X2 0-62-924 PARAIoI:TERS P .... RAMETERS PER OPERATING "13-62-5,1,. NJ-62-5A. OPER .... TINC 1oI00E I.t:IDE CALCUL CALCULATION ... TION (826 (B25 850724014)85 0724 014) PROBLEM D FX;~ HOlDUP eves HOLDUP
~~ '-10 1ilJ 8-10
~TI Q I CONNECTION 150 PSIG 150PSIC D£SIGH DESICN I'"
T.... MK-62-JA TANIC-62-3A CHEMICAL MIXING
'--!oo_:c_!_'~_-:_:__
EXCHANGER SH' CooR "'-12 J r -.--------,...-J----'-----'=--'~--,-.----------------,-.--;--~i--':----.--~---,-. ---1>1---, '~~~,~::. ~~T H
- 19. ALL PIPING DOINSTREAM VENTS ANeI THIS NOTE APPL DOWNSTREAM OF THE LAST ISOLATIOI'I AND TEST CONNECTIONS IS TVA CLASS APPLIES IES TO SHEETS 2 THROUGH 6 ONLY.
ISOLATION V CL"'SS C VALVE... LVE ON LOCAL DRAINS. G UNLESS OTHERWISE NOTED. OR"'INS, D b T,"P
'00"'
200" TANK 20. <!])
<i])INTERFACE INTERFACE POINT DESIGNATION OESIGN .... TION APPLIES ONL ONLY Y TO UNIT 2 VALVES V.... LVES
~ EHT ~'
2-452-927 AND 2-fCV-62-138. 1 -6~:-~44 2-fCV-62-1:38. 1-~"'4 I '..'----r---' .3/4" ':...~943
*lillm "----y--" 1 62 943
~ :?loL1_~_N~_0_-'____ ~HUj ~I
§~~A~G CHARGINC PUMP
~l:fpON PUlo(P I-----.~'+'~-=='::.
1 I
' -_ _. ..... _-'-I_-='=/"-.*-j:.~~-~I-_
..-==------==-I~I_-
RESIN FILL SH 2 <!])
<1],) .t.1Iillu:m - E~~ (-10 II' ~c !IT!
TRITIATED
~~n~""TED r.r_____ '_'S-1."1
-:.'=r.": rc.. J CCOR G-2 ~
.Jill <L T liP 21.
IIOUNDARY
.t. DENOTE UNIT 1 ..
BO!JNO .... RY IS THE FIRST ANCHORED EQUIPIENT
.t. UNIT 2 INTERfACE INTERFACE POINTS. THE STRUCTURAL EOUlf'tEMT OR PIPE PIpE ANCHOR ON THE UNIT UMIT 2 I- -
~
CooRE-l0 DRAIN mJ w W j7 rr..1lf--
----;;r ~"""'I!If
~ I.lr ~\,,_~ ~ ~:r-IE ~
SIDE OF"
'ILL WILL H OF THE INTERFACE POINT. BECAUSE Of" TEES, HAVE.... VE I.t:IRE IClRE THAN ONE STRUCTURAL TERMINATION.
rs TEES. SQ.IE INTERFACE POINTS TERt.lINATION. ( 1519 1-62-904:( 4-62-90" ill . 1-62-901 T 1-62-911 m
- 22. UNIT l/UNIT l!UNIT 2 INTERFACE POINTS ARE INDICATED BY
[]lJ []l] - ern> -
<ill> SAFETY S .... F'ETY RELATED L!*.'--:~;::u:==:sl:==:L~1-'62-903 m".-'--";:frlf==:sl::'=*,='-='=""L..2.:'-62-9D3 1-62-910..1 1-62-910.1 1-62-90B 1-62-908 AND J<<)H-SAf"[TY NON-SAF'ETY REL RELATED. ... TED. 'lHEN WHEN ""'NUAL UANUAL VALVES ARE USED, NON-SAFETY USEO, NON-S"'fETY RELATED INDICATES THE HAND.HEELS ARE REL()VED REUOVED AND SAFETY SAfETY RELATED INDICATES
~ THE VALVE IS LOCKED CLOSED. WHEN A I.IOTOR MOTOR OPERATED VALVE V.... LVE IS USED, THE VALVE E i;;~
IS ClOSED, t.I CLOSED. ELECTRICAL t.OTIVE
.... NU ....L OPERATOR IS DISABLED. WHEN "'N t,jANiJAL IClTIVE POWER IS DISCONNECTED "'NO AN AIR OPERATED VALVE AND THE V"'LVE IS USED. USED, THE VALVE E
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11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 20.028 AIG.05 AK3.05 020 Given the following plant conditions: Unit 1 is at 100% power.
- The Pressurizer Level Master Controller setpoint is failed at at the current value.
Pressurizer Level Control is selected to U-68-339.
- A load reduction to 60% is initiated to remove a Main Feedwater Pump from service.
Which ONE of the following describes the action that will be taken, and the reason for those actions? A. Change the pressurizer level channel input to the pressurizer level master controller and restore Letdown because the backup level channel response to actual level will cause a letdown isolation. B~ Place the charging flow Control Valve in MANUAL and reduce charging flow B:' because actual pressurizer level will be higher than program level for the actual power level as load is decreased. C. Place the charging flow Control Valve in MANUAL and raise Charging flow because actual pressurizer level will be lower than program level for the ( actual power level as load is decreased. D. Change the pressurizer level channel input to the pressurizer level master controller and restore Letdown because the controlling level channel failure will cause a letdown isolation. Page 53
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect. Changing input will not affect the controller because the setpoint signal has failed, not the input signal. Plausible because actual pressurizer level would drop and cause a letdown isolation if the failure had been on the transmitter inputting to the master controller instead of on the setpoint input. A A backup channel would have caused a letdown isolation. B. Correct, If the setpoint fails as is and power is then reduced, the pressurizer level setpoint will wiff be maintained higher then normal for the reduced power. Charging flow would have to be placed in Manual and reduced to lower the pressurizer level to the proper setpoint. C. Incorrect. Charging flow would not be raised; it must be reduced because level will wiff be maintained higher than the proper value. Plausible if the applicant reverses the effect of the failure in the pressurizer level control system. D. Incorrect. Changing input willwiff not affect the controller because the setpoint has failed. Letdown will wiff not isolate because actual level will wiff indicate the same on all channels. Plausible because actual pressurizer level would drop and cause a letdown isolation if the failure had been on the transmitter inputting to the master controller instead of on the setpoint input. ( Page 54
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 20 Tier: _1__ Group _1_ 2 KIA: 028 AK3.05 Knowledge of the reasons for the following responses as they apply to the Pressurizer Level Control Malfunctions: Actions contained in EOP for PZR level malfunction Importance Rating: 3.7 / 4.1 3.7/ 10 CFR Part 55: 41.5,41.10/45.6/45.13 10CFR55.43.b: Not applicable. KIA Match: Applicant is required to determine the impact of the malfunction in the pressurizer level control system and identify the procedural action required due to the failure. Technical
Reference:
AOI-18, Malfunction of the Pressurizer Level Control System, Rev 22 3-0T-SYS068C, PZR and PRT, Rev 9 ( None Proposed references to be provided: Learning Objective: 3-0T-SYS068C
- 15. Describe the response to a deviation from pressurizer level program.
Question Source: New Modified Bank Bank x Question History: SON 2007 exam Comments: Page 55
AOI-20 WBN MALFUNCTION OF PRESSURIZER Revision 31 LEVEL CONTROL SYSTEM Page 5 of 17 3.0 OPERATOR ACTIONS ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
.RESPONSE CAUTION Charging and letdown must be in service together. If letdown isolates or charging is lost, the other must be isolated.
- 1. CHECK pzr level program PERFORM the following: .
signal NORMAL:
- a. PLACE charging valve controller
- 1-LR-68-339 1-HIC-62-93A in MAN, MAN, and (green pen). RESTORE level to normal.
- b. IF letdown in service, service; THEN
** GO TO Step 10.
- c. IF letdown NOT in service, THEN
** GO TO Step 6.
NOTE control level to program and one 1-XS-68-339E selects one channel to controllevelto backup channel for control interlocks.
- 2. CHECK if 1-XS-68-339E is IF pzr level is low OR selected to FAILED channel dropping, (control or backup): THEN
** GO TO Step 12.
- LI-68-339, IF pzr level is high OR OR rising, THEN
- LI-68-320, ** GO TO Step 14.
OR IF non-selected channel failure,
- LI-68-335. THEN
** GO TO Step 8.
AOI-20 WBN MALFUNCTION OF PRESSURIZER Revision 31 LEVEL CONTROL SYSTEM Page 14 of 17 4.0 DISCUSSION (Continued) 4.3 Instrument or Control Circuitry Failure failures can be quickly diagnosed by a channel check. Pzr program Instrument failures signal failure can be quickly diagnosed by checking recorder 1-LR-68-339. Failure low of a pzr level instrument channel will result in letdown isolation resulting in actual level rising. This level rise will be greater if Channel 339 fails since it also would result in a large error signal to raise charging. In this case (lOW (low failure), an operable channel can be selected and letdown quickly restored. Failure high of a pzr level instrument will result in a reduction of charging to the minimum 55 gpm. Instrument and program level signal failure can be compensated for by taking manual control of charging. Signal failures up to the individual controllers for the charging valve can be compensated for by manual control of the individual controller. If necessary, local manual control of the charging valve can be taken. taken .
~:. -\'-.
4.4 Loss of Charging Loss of charging can occur due to loss of Charging Pump or closing of a valve in the ( charging line. Since all air operated valves in the charging line fail open, the most likely cause of a loss of charging would be due to an instrument failure causing letdown isolation or trip of the Charging Pump. Instrument failures have already been addressed. If a Charging Pump were to trip, letdown should be manually isolated to conserve RCS inventory and to prevent damage to the regenerative heat exchanger. Another Charging Pump can be started and letdown reestablished. that Operator response to a tripped Charging Pump would be to It should be noted that immediately isolate letdown. Entry into this procedure from this point would result in basically reverification of actions already taken with additional actions to restore charging and letdown.
Pressurizer Level Program ~ Normal Programmed level increases linearly from 25% 25% to 60% 600/0 as Tavg increases from no load (557°F) to full load (586.2°F) value (as shown by the blue shaded area). ~ Pressurizer level setpoint is limited to 25% 25% for Tavg S 557°F and
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WBN Containment Purge System 501-30.02 Unit 1 Rev. 0054 Rev.00S4 Page 8 of 74 3.0 PRECAUTIONS AND LIMITATIONS A. In Modes 1-4, the containment purge system is operated only when the Reactor Building equipment and personnel access hatches are CLOSED. B. During refueling operation when ABI/CVI crosstie is required, the following conditions apply: (Refer to Tech Spec Bases 3.3.6, 3.3.8, 3.7.12, & 3.9.8.)
- 1. 1-HS-90-410-A & 415-B are placed in the REFUEL position prior to the beginning of refuel operations (Mode 5) when the Contaiment and/or Annulus is open to the Auxiliary Bldg ABSCE spaces and returned to NORMAL postion prior to entering Mode 4 from Mode 5 unless repositioned by another approved configuration document (eg. 18 Month GO Test, DG Blackout Test, etc.). When moving irradiated fuel inside Containment, both trains of the Containment Purge System must be operable and at least one train must be operating, or Containment must be isolated.
- 2. If 1-HS-90-41 O-A & 415-B are required to be placed in the NORMAL position during movement of irradiated fuel in containment or the auxiliary bldg, then fuel movement must be stopped or containment must be isolated.
(
- 3. When moving irradiated fuel in the Auxiliary Bldg with the Containment open to the Auxiliary Bldg ABSCE spaces, Containment Purge System may be operated, but all Containment ventilation isolation valves and associated instrumentation must remain operable.
- 4. Both trains of ABGTS must remain operable if Contaiment and/or Annulus is open to the Auxiliary Bldg ABSCE spaces during movement of irradiated fuel inside Containment.
- 5. SSPS train must be maintained operable for whatever train is supporting the ABI/CVI functions.
C. Max Containment Purge HEPA and Charcoal filter combined L\P ContainmentPurge ~P is 4.7 in. H20. D. Containment Purge & Vent with normal purge dampers is NOT to be done when Containment press is above atmospheric. With Annulus L\P ~P 5.4 in. H20, Containment press above 0.2 psi is above atmospheric (SOI-30.03 is used reduce Containment press). E. Ice Condenser Door positions should be monitored when purging Containment. A slight upper-to-Iower Containment press imbalance will open the doors. (
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 21. 036 AG2.4.2 021 021 Given the following:
- Unit 1 is in Mode 6.
1-HS-90-41S-B [back of 1-R-78] are 1-HS-90-410-A [back of 1-R-73] and 1-HS-90-415-B in the REFUEL position. Compare the effects of the following 2 events where an irradiated Fuel Assembly was dropped causing entryentry-into into AOI-29, "Dropped or Damaged Fuel or Refueling Cavity Seal Failure." (1) An assembly is dropped in the Spent Fuel Pit resulting in radiation monitors 1-RM-90-102 and -103 High Radiation relay actuation. (2) An assembly is dropped inside containment resulting in radiation monitors 1-RM-90-130 and -131 High Radiation relay actuation. Which ONE of the following identifies how the automatic initiation of Containment Vent Isolation (CVI) and isolation of the Auxiliary Building are affected due to the positioning of the switches? A'I Dropping either assembly would result in initiation of CVI and an isolation of A'! the Auxiliary Building. B. Dropping the assembly inside containment would result in an isolation of the Auxiliary Building but dropping the assembly in the Spent Fuel Pit would NOT result in a CVI. C. Dropping the assembly in in the Spent Fuel Pit would result in a CVI but dropping the assembly inside containment would NOT result in an isolation of the Auxiliary Building. D. Dropping the assembly inside containment would NOT result in an isolation of the Auxiliary Building NOR would dropping the assembly in the Spent Fuel Pit result in a CVI. I Page 56
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Correct, With the switches in the REFUEL position, if either the Containment Purge radiation monitors or the Spent fuel Pit Radiation Monitors detect high radiation, the Auxiliary Building will isolate and a CVI will wiff be initiated. B. Incorrect, the fuel assembly dropped in the Spent Fuel Pit would result in a CVI signal. Plausible if the applicant correctly concludes the two handswitches are for separate trains and the reasoning was to isolate the auxiliary building if an accident occurs in containment during fuel movement and not to isolate containment due to an accident in the Auxiliary Building since the equipment hatch is required to be closed for fuel movement. C. Incorrect, the fuel assembly dropped in the Containment would result in an isolation of the Auxiliary Building. Plausible if the applicant correctly concludes the two handswitches are for separate trains and the reasoning was to isolate containment if an accident occurs in the Auxiliary building and not to isolate the auxiliary Building due to an accident inside Containment since the equipment hatch is required to be closed for fuel movement. D. Incorrect, Both of the isolations listed would occur (either not occurring is incorrect). Plausible for neither to occur because each of the two would be correct conditions if 1 -HS-90-410-A and 1-HS-90-415-B had been identified as being placed in the ( NORMAL position in the stem oftheof the question. ( Page 57
11/2009 Watts 11/2009 Watts BarBar RO RO NRC NRC Exam Exam -- As As submitted submitted 10/2/2009 10/2/2009 Question Number: Question Number: 21 21 Tier: _1__ Group _1 Group 22 KIA: 036 AG2.4.2 AG2.4.2 Fuel Handling Incidents Knowledge ofof system set points, interlocks and and automatic actions associated with EOP entry conditions. Importance Rating: 4.5 /1 4.6 10 CFR Part 55: 41.7/45.7/45.8 41.7/45.7 / 45.8 10CFR55.43.b: Not applicable. KIA Match: Applicant must determine automatic actions that occur associated with conditions requiring entry into the Abnormal Procedure for fuel handling incidents. Technical
Reference:
Drawing 1-45W600-30-4, R25 Drawing 1-45W600-57-19, R19 Drawing 1-47W611-88-1, R23 Drawing 1-47W611-30-7, R23 ( SOI-30.02, "Containment Purge System," Rev 0054 AOI-29, "Dropped or Damaged Fuel or Refueling Cavity Seal Failure," Rev 20 N3-30AB-4001, "Auxiliary Building Heating Ventilation, Air Conditioning System" (30,31,44), Rev 0028 Proposed references None to be provided: Learning Objective: 3-0T -AOI-2900 3-0T-AOI-2900
- 4. Identify possible Auto Actions from:
- a. Dropped fuel assembly in SFP area.
- b. Dropped fuel assembly in Cntmt.
Question Source: New X Modified Bank Bank Bank Bank Question History: New question Comments: Comments: Page 58 Page 58
AOI-29 WBN DROPPED OR DAMAGED FUEL OR Revision 20 ./\ REFUELING CAVITY SEAL FAILURE Page 3 of 29 1.0 PURPOSE To provide a response to a refueling cavity seal failure or a dropped or damaged fuel assembly. 2.0 SYMPTOMS 2.1 Alarms A. SFP LEVEL HI/LO [128-A]. B. LWR CNTMT AIR 1-RM-106 RAD HI [173-B]. C. UPR CNTMT AIR 1-RM-112 RAD HI [173-A]. D. CNTMT PURGE EXH 1-RM-130/131 RAD HI [174-A]. E. 1-RR-90-1 AREA RAD HI [174-B]. F. SFP 0-RM-1 02/1 03 RAD HI [184-B]. G. AB VENT 0-RM-101 RAD HI [183-C]. H. CONTAINMENT PIT SUMP ALARM [Incore Inst Rm]. 2.2 Automatic Actions A. Auxiliary Bldg Isol from: 0-RM-90-102, Spent Fuel Pit. 0~RM-90-1 03, Spent Fuel Pit. Containment Vent Isol. (if in refueling outage with 1-HS-90-410-A and 1-HS-90-415-B in REFUEL position) B. Containment Vent Isol from: 1-RM-90-130, Tr A Purge Air Exh. 1-RM-90-131, Tr B Purge Air Exh. Auxiliary Bldg Isol (if in refueling outage with 1-HS-90-410-A and 1-HS-90-415-B in REFUEL position')
NPG System AUXILIARY BUILDING HEATING, N3-30AB-4001 Description VENTI LATION, AIR CONDITIONING VENTILATION, Rev. 0028 Document SYSTEM (30, 31, 44) Page 60 of 195 ( 2.2.4 Radiation Release Requirements The AB HVAC System shall be designed to limit the radioactivity released to the environment to levels* levels below 10 CFR 20 Appendix B (Ref 7.5.4) during normal operation and below the requirements of 10 CFR 100 during an accident. The system shall also be designed to maintain concentrations of airborne radioactivity in normally occupied areas within the requirements of 10 CFR 50 Appendix I (Ref 7.5.5). To control airborne activity, ventilation air shall be supplied to clean areas, then routed to areas of progressively greater contamination potential. Upon detection of high radiation in the fuel pool area, high temperature in the Unit 1 air intake, or receipt of an isolation signal from the reactor unit, the AB supply and exhaust fans and the FH area exhaust fans shall automatically stop, the ABSCE isolation dampers shall close to isolate the ABSCE and the ABGTS shall automatically start. The reduced exhaust shall be diverted through the air cleanup units and discharged via the Shield Building vent. A minimumnegative minimum negative pressure of 1/4 in.w.g. shall be maintained by the ABGTS in the ABSCE to limit outleakage. See Ref 7.2.11 for additional radiation protection (ALARA) guidelines. When moving irradiated fuel in the Auxiliary Building and/or Containment with the Containment and/or the annulus open to the Auxiliary Building ABSCE spaces, a signal from the spent fuel pool radiation monitors 0-RE-90-102 0-RE-90-1 02 and -103 shall initiate a Containment Ventilation Isolation (CVI) in addition to their normal function, and a signal from the containment purge radiation monitors 1-RE-90-130, and -131 shall initiate a CVI signal, which shall subsequently initiate that portion of the ABI normally initiated by the spent fuel pool radiation monitors. However, the containment equipment hatch cannot be open when ( moving irradiated fuel inside containment per Technical Specification 3.9.4. See Section 4.8 for additional details. The AB HVAC system shall be designed to limit radiological release as defined in Ref 7.2.22. 2.2.5 Pipe Rupture Requirements The AB HVAC System shall be evaluated for the effects of high and/or moderate energy pipe rupture in accordance with Ref. 7.2.16 and Ref. 7.2.17. Primary safety-related components of the AB HVAC System shall be protected against the effects of missiles, pipe whip, flooding, and jet impingement that may result from piping or equipment failures. 2.2.6 Environmental Qualification Requirements The AB HVAC System and components that measure, monitor and control the system operation shall be designed to operate in environments associated with normal and accident conditions for which the system is required. Specific environmental conditions are given in Ref. 7.2.2. The following criteria shall be used for the design and procurementof equipment: A. WB-DC-40-42, "Environmental Design" (Ref. 7.2.2) B. WB-DC-40-54, "Environmental Qualification to 10 CFR 50.49" (Ref. 7.2.8)
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- 2. FUSE NUMBERS SHOWN THIS ORAWINC WITH THE BOARD PREFIX 1-45"600-57-3.
DRAWING SHOULD BE CONarNEO CONBINEO PREF'lX LISTED BELOW TO FORMULATE COt.IPLETE UNIQUE IDENTIFICATION NUMBERS: UN!QUE COMPLETE A
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~ 91 81 91 B1
" (1-L-11" B1 o
15 11 ADMIN J1 ESJ 1J JEW I1 ~ 110/.;,.5/00-JIO/..2 5 /Dbf-- REVISED BY ADMIN TO CORRECT ERRORS IDENTIFIED BY PER 112679 o (T71 (T7 1 061025 80.3). 803).
< REV I CHANGE REF I PREPARER JI CHECKER 1l APPROVED 1I DATE
- a SCALE: NTS SCALE; PROJECT FACILITY POWERHOUSE EXCEPT AS NOTED G-G UNITS 1 '" & 2
<~
~o TITLE
~~
<~ WIRING DIAGRAM
~; ~ p~ ,~lDlJ,l SEPARATION & MISC AUX RELAYS
- SCHEMATIC DIAGRAMS (1-L-l1B)
( l-L-l1B)
; ~ (1~), Ie, tlli 1 1I DESIGN WATTS BAR NUCLEAR PLANT TENNESSEE VALLEY AUTHOR AUTHORITY I NIT I AL ISSUE INITIAL ITY ENGINEERING H 1246VG A 1H"22 (CONTACTS ARE
'C')
FORM "C') 12-46VG AA>
-!Ill-1'H4 1 H4 2 1303VG i '"
-lI1l-
---:-,"11.";;';"H-'='- -
(CONTACTS ARE FORM "C') C(),iPANION CWPANION DRAWING, 1-45\1600-57-2 1-45\1500-57-2 DRAWING: DRAfTER DRAF TER C.It.CLA80UGH C.N.CLABOUGH CHECKER C.A.H,I,MILTON C.A.HAM[LTON RO ISSUE PER W8EP 5.17 cI:. & RHAS RIMS APPROVAL 1 Q.P.FERRER D.P.PERRER DESIGNER REVIEWER 626 826 '90 *90 0802 377 2 R. T .GRAHAM T.GRAHAM PRESSURIZER HEATER BACKUP GROUPS 1A-A & ~ 1B-B 18-B SEPARATION RELAYS M.L.CHAPMAN M.L.CHAPM,l..N D.F O.F .fAULKNER
.f"AUlKN£R J3 C.C.LYKE FOR 1.C8 ~8
{INTERLOCKS (INTERLOCKS FROM fROt.t l-LS-6B-325C l-LS-68-o325C 8/A, B/A, l-LS-6B-J26C A/8 ... l-PS-68-337C B/Al l-LS-68-326C ,1../8 &. l-PS-66-337C 8/,1..) ISSUED BY, IISSUED BY: J.EOWARD GIBBS J.EOII'ARD DATE 9-22.-90 9-22-90 1851 11-45W600-57-7
-45W600-57-7 R15
- 22 I I I 9 E
II I I I 3 4 I 5 I 6 I I 77 I 8 I 9 I I I I II CAD MAINTAINED DRAWING I (CONFIGURATION CONTROL DRAWING) ii
4 5 6 7 8 9 11 12 H-690 4,1,-690 ",1,-680
-4A-68D 4A-670 4,1,-670 4A-660
+
4' + 4' +4' + 4' A ~ A ACTUATE AlB TRAIN A/B SIS ~~~2AL
~~~8Al
\-47W611-63-1
\-47'611-63-\
B 8B cC cC CONTAINf.4ENT CONTAINMENT SPRAY DISCHARGE VALVE CONTAINt.4ENT SPR-'Y l-FCV-72-2 l-fCV-72-2 PUMP 1-+7'611-72-1 1-47W611-72-1 1-47W611-72-\ t.4AIN STEM.I LINE SYSTEM SHEET ISOL"TION VALVES 1-47W61'-1-1 1-47W611-30
\-47W61'-30 1-47W611-as 1-47W611-1!8
'.5 1,:;
1 SYSTEI.4 SYSTEM SHEET SYSTEM SHEET 1-4-7W61'-30 1--1-7111611-30 1. 6 1-47W611-30 3. 4-4 NOTES: 1-"7'01'611-61 1-"7111611-61 2 1-47'11611-70 1-47W611-70 .3J 1. FOR TRAIN B. 1-RE-SO-IJI. \-HS-90-415, AND D-HS-90-IJ6A2 l-RE-90-131. l-HS-90-415, O-HS-90-136A2 ARE USED. 1-47W611-62 1-"'111611-62 1 O-HS-90-IJ6A\ a. O-HS-90-136Al .t 1:56A2 136,1,2 ARE MULTI SELECT 5W'S 5"5 USED TO TEST ONE 1-47W61'-53 1-47111611-63 J, 3, 5, 6, 8 1-47W611-32 CHANNEL AT A 111.4E. TIME. D 1-47'01'611-65 1-47W611-65 1. 1, 2, J3 1-+71611-67 1-47W611-67 D l-H'II61'-68 1-47'11611-68 J.
.3. ALL ISOLATION SIGNALS AND CIRCUITRY THAT ARE PART OF ENGINEERED 1-*oHW61'-70 1-47W611-70 SAFEGUARDS ARE REDUNDANT.
- 4. NOTE DELETED.
1-47W611-77 1-47'11611-81
'.5
, 5. EACH TRAIN A RESET HAS AN ASSOCIATED TRAIN B RESET. SEE TVA DRAWING l-FW610-30-1 1-{7W610-30-1 FOR TRAIN B "RESET" HANDSWITCH NUMBERS.
1-47'11611-1 3 6. VALVES l-FCV-43-22 AND I-FCV-4J-23 MAY "NO l-FCV-43-23 M"Y BE NORM"LLY NORMALLY OPEN WITH 1-+7'11611-31 l-FSV-43-23 ENERGIZED TO ALLor now ALLOW CONTINUOUS RCS FLOW DUE TO THERMAL 1-+7111'611-88 1-4711'511-88 OCL ING CONCERNS (REF CYCLING {REF. EOC-511 34) . EDC-511341. 1-47111611-26 1-47'11611-26 l-RE-90-130 It
- 8. l-RE-90-1J0 .to 131, WHICH HOUSE GAS RADIATION DETECTORS, MONITOR THE CONTAINf.4ENT CONTAINMENT PURGE EXHAUST.
E 9. DIGITAL AND ANALOG LOGIC TO FUNCTIONAlL fUNCTlONALL Y DESCRI8E lOGIC SYMBOLS ARE USED ON lOGIC DIAGRAMS OESCRIBE THE PROCESS CONTROL. REFER TO THE E ASSOCIATED WIRING SCHEMATIC SCHE!.t.-.TIC FOR THE ElECTRICAL ELECTRICAL COMPONENTS USED TO IMPLEMENT THE CONTROL SCHEME. 11 FOR
- 11. fOR SYMBOLS SEE INSTRUMENT ION ItIDENTIfiCATION
.l.IOENTIFICATION STANDARDS.
STANDARDS, LATEST lATEST ISSUE. REFERENCE ORAWINGS; DRAWINGS; 1-47'11611-0-1 1-47W611-0-1 ---------- LOGIC INDEX" lOGIC INDEX It SYMBOLS 47B501-SERIES ------- 47BSD1-SERIES T INSTRUMENT TABULATION nON 1-45'11'600-43-2 --------- 1-45'11600-43-2 SAMPLING MiDAND WATER au R QUALITY SYSTEM SCHEMATIC or"GRAM 1-451500-90-1 1-45W600-90-1 --------- RADIATION RADIA I.IONITORING ING SYSTEM SCHEMATIC DIAGRAM 1-45W600-57 -------- 1-45160D-57-B SEPAR SEPARATION .t. MISC ,a.ux AUX RELAYS SCHEII1ATlC DHGRAMS 1-471/610-30-1 1-47'11'610-30-1 THRU 6 -- CONTAINMENT CONTA VENTILATION lAn SYSTEM 1-47W610-90-1 1-47",610-90-1 THRU J3 -- MONITORING SYSTEM CONTROL RADIATION OONITORING TROl DIAGRAM 1-47W610-43-3, -5, 1-4"610-43-3, SAMPLING .l. QUAL It WATER QUALITY DIAGRAM SYSTEM CONTROL OI"GRAM 1-47'611-1-1 1-47W611-1-1 ---------- MAIN STEAM SYSTEM LOGIC lOGIC DIAGRAM 1=!j:~11 =~~=l ========= ~~~iUN~~~~C~~~~yS~~~WML~ob1cD61~~~~M"---
=!~:~ll =~t: ~~~nIN~~~~C~~~~yS~~g~M L~YcD6~~~~~M'---
F CONTAINMENT ISOLATION SIGNAL F PHASE A l-FCV-90-107 FULl. Y OPEN 23 2-13-08 CONTAINMENT CONTAINMENT } VENT ISOLATION TRAIN B 8 COORD C-l1 l-HS-4J-202A G d I-HS-43-207A 1-HS-4.3-208A TYPICAL FOR NOTE 6 POWERHOUSE UNITS 1 f ACIL ITY PROJECT FACILITY
& 2 dd TITLE l-HS-43-202B
'-HS-+3-Z02B l-HS-43-2078 l-HS-4J-207B l-HS-43-20B8 l-HS-43-2088 X
ISOl.ATION VA.LVES TYPICAL SAMPLING SYSTEM ISOLATION VALVES ELECTRICAL
.-="--1-;";-':;';;';-;;-"'='----1 r-="-l7;'~;;-"'='-I g~ICAL n;~ICAL FOR REACTOR COOLANT, PRESSURIZER LIQ 1-47"'0-43-'.
1-<7>61 0-<3-1, -3. L 10 It AND ACCUM TANKS. (PWR TRAIN AS NOTED ON
-5 * ,-""00-43-2).
-J * -5.
A. GAS, 11'-FCV_90_'08 LOGIC DIAGRAM 1-45'600-<3-2), 1 -,cv-,0_,08 l-FCV-90-109 l-FCV-SO-l09 I-FS l-FS CONTAINMENT ISOLATION dd l-FCV-90-110 S l-FCV-4J-4.H-
'_FCV-43-434 TYPICAL l-FCV-90-ll1 90-107
,1-FCV-90-111 l-FCV-43-435 1-FCV-4J-4J5 FOR l=~g:~g=ll~
l-FCV-90-11J WATTS BAR NUCLEAR PLANT l-FCV-4J-436 l-FCV-43-436 FOR l-FCV-90-114 1-FCV-90-115 l-FCV-90-115 l-FCV-90-116 l-FCV-SO-116 I-Fe TENNESSEE VALLEY AUTHORITY Q Q TYPICAL l-FCV-90-117 FOR
'OR DESIGN INITIAL ISSUE ENGINEERING APPROVAL H X CONTAINMENT BLDG UPPER It TYPICAL CONTAiNM[Nl lWR A. LWR DRAFTER CHECKER RO ISSUE PER 1
FOR REACTOR SA~PLING SYSlEM TYPICAL SAMPLING SYSTEM ISOLATION VALVES RE,I,CTOR COOLANT, PRESSURIZER LIC a:. GAS. LID It COMPARTt.lENT COMPARTt,jENT AIR MONITOR ISOL (PWR TRAIN AS NOTED O~ {PWR VLV ISOl "lV ON 1-47"'-610-90-3 1-47'/(610-90-3 It.l. FSAR FIG 6.2.4-21 &. & G.W. JOHNSO~ JOHNSON GARY BARNARD EAI 3,' 3,10 a Itol RIMS 2 C.A. ATKINS ROBERT O. AND ACCUM TMKS. TANKS. (PWR TRAIN AS NOTED ON 1-47W610--i3-1, -3 1-47'/1610-43-1, -3.t a. 1-45'11600-+3-2). It -5 It 1-+5W600-+3-2). 1-45\11600-90-1 ) 1-4SW600-90-1} FSAR FIG 7.3-3 SH 4 DESIGNER REVIEWER T28 .'92 T2B 92 0227 860 B60 D. MURR
.33 C.C. LYKE lYKE FOR !.1GB MeB OFF OfF CQNTAHIMENT H2 SAMPLE SYSTEM CONTAINI.4ENT ISSUED BY: DATE (P.,,-R (PWR TRAIN AS NOTED ON 1-471610-43-6 1-47W610-43-6 It .t
! _ _ _ _ _ _ _ _ ~.~._J.9~~~
R.~. JOHNSON __ 3/5/92 85 1-47W611-88-1 R23 1-4SW600-43-2) 1-45W600-43-2J 1----------------- 1 I 2 I 3 I I I I I I iI 4 I 5 I 6 I 7 I 8 I 9 I CAD MAINTAINED DRAWING (CONFIGURATION CONTROL DRAWING)
NOTES; NOTES: ADD JlU"E:R 1L ~ F'al cE:IiERAL IOTES f'OO CrliERAl IIOlES A~D AtlO RHUEIrICE RHE:itEIiICE DRA.'fINCS DRA.'lI'INCS SE£ 1-451'60o-Jo 1-451'60o-Jo-l.
, 2 2 ..... StiOIiliN ON nus
_ .... l l FUSES SliDILIN THIS DRAnM; DRAnr<<> H ...VI A aJNTACT HAVE aJRTACT I11I01 IHlt:H nOSES Q.OS[S FortFOR AN FOR InIle ( OPEN FUSE 45~1688-2 J03__ HANlS'll1CH F!JSE A!Il 45~lS88-2 AIIO All) WILL (J>E:RAIE A~ -tSN169!3-2. HAIIlS'lITCH DEVEl.CI"YE:Nl IlIQ,UNG 1-.. DRAU"';:
<FE:RAIE AN ALARM.
"~N16!3-2_
DEVEl.o'I.lE:Nl lABlE l-.. !ii~6(JO-90~I_ 5.6()O-90~1_ tABLE rCM
"'lARM. fOR IUIIe: 'CEU[LS 1-HS.-'9o--tl0 fOFl: l-HS.~'9o-tl0
'DETAltS SEE
- a. ...... 1-5i 15 SH SHOIiN CI\I IS 'SInIiN C'I\I A +. @ AIID AIIID Iill1
!ill] DEJlDTES lINIT 1 AND UNIT DEflOTES lJNI1 UIiIT 2 INTERFACE mINts. mlNrS. A
- 5. E.~p OEVICE .1\5 '11'01\5 DEPOIERE~ BY B't DCN ......05736-D.
O:l-7lfi-D_
- ~~~~~~~ ~~~D~~~~~ ~~~!fil ~~-r~~1S"II(JT~E"["S"'RE"'F"'["'RE",",,='~""
- EHO DEYICf IS ootslDE: THE U~n "J BOVIIJA'RY. """Y"~=""'="""
- ISOL.~TION OC:CORS ~ p~rtEL 1~t.I-fi ~ERE lHIS WOT.E 15 REFERENCED.
B B B o c ~ 0 (:) c
~ ~~
l-FU-r~~~~f4 TRAIN A AUXILIARY BUILOING ISOLATION A HIGH RADIATION IN REFUELING AREA LOGIC BUS (1-Ft-7a] ('ton: 2) 1_FlJ_27S_fl78,jY.l'] 12+~n I~~wrn~'~----------~~----~~CC~----~----------T-~~--------------------------~----------~=----- ______________________________~r_----------~------~v~.~~------------------~--------------~------. VKIIl o D VKBI D v.<B~-6 D ( E E o o F o ~ F VleOIl l-FlJ-VS-R78/IC16 l--fIJ-.275-R;78/ICla
'1-11;-78)
'1-f;l-78) TRAIN BB AUXILIARY BUILDING ISOLATION ~
A HIGH RADIATION IN REFUELING' REFUELING" AREA LOGIC BUS (NOTE (HOTE 2) 4-9-09 COt.IFlGtlR~nO" CDNTROL THIS GOt.IFIGlJRAHOM CCNTROl D~.IJ<5 E-7 IlCilcc
/-LCi/cc or tritium is
>1 E-5llCilcc E-5/-LCilcc THEN, DIVERT SGBD SGSD to the Hotwell.
References:
55W1381-7 1-45W760-27-4,-5 501-27.03 SOI-27.03 1-SI-0-2-00, Data Sheet 38 WBN WSN Page 27 of 48 ARI-159~165 ARI-159-165 Rev 34
---------------------~-----------~------~--
SOURCE SETPOINT 0-RM-90-122 VARIABLE above background WASTE LIQUID MONITOR RADIATION HI RADIATION
&2 UNITS 1 & 2 23 Note 0-RCV-77-43, CNDS FLOW TO CCW FLOW CONTROL isolates on high radiation.
Probable Cause: 1. Hi radiation in release effluents Corrective Action: [1] ENSURE release is terminated. [2] REQUEST RADCON to perform a survey of the area to determine if background level increased. ( [3] NOTIFY SRO of the alarm and survey results.
References:
1-4SN600-77-1 1-45 N600-77 -1 1-47W61 0-77-1 1-47W610-77-1 1-47W61 0-90-2 WBN I Page 26 of 36 ARI-O-L-2D Rev 7
WBN OFFSITE DOSE CALCULATION MANUAL Revision 22 ( o (ODCM) Page 65 of 195 RELEASE POINTS Liquid Radwaste Treatment System The lRTS processes liquid from the Reactor Building and Auxiliary Building Floor Drains, the CVCS laundry/HotShower, Holdup Tanks, the laundry/Hot Shower, the Waste Condensate Tanks and Chemical Drain Tanks. Figure 6.2 provides a schematic of the Liquid Radwaste System, showing the flow pathways and flow rates. The lRTS has individual release points for each of the tanks. The routine release points for liquid radwaste are the Monitor Tank and the Cask Decontamination Collector Tank (CDCT). The Monitor Tank has a capacity of 20,000 gal and can be released at a maximum design flow rate of 150 gpm. The CDCT has a capacity of 15,000 gal and can be released at a maximum design flow rate of 100 gpm. The Monitor Tank and CDCT discharge to the Cooling Tower Blowdown line as a batch release and are monitored by radiation monitor O-RE-90-122. There is a flow interlock on the CTBD which suspends the release if the flow is less than 20,000 gpm. Condensate Demineralizer System The Condensate Demineralizer System liquid wastes are released from the High Crud Tanks (HCT-A and -B), the Neutralization Tank, and the Non-Reclaimable Waste Tank (NRWT). Figure 6.3 provides a schematic ofof the the Condensate Demineralizer System, showing the flow pathways and flow rates. The HCTs have a capacity of 20,000 gal and a maximum design discharge flow rate of 150 gpm. The Neutralization Tank has a capacity of 20,000 gal and a maximum design discharge flow rate of 100 gpm. The NRWT has a capacity of 11,000 gal and a maximum design discharge flow rate of 30 gpm. Each of these tanks is defined as a release point for the system. When tank contents are required to be permitted for radioactivity, the Condensate Demineralizer System waste is released to the CTBD line and is monitored by radiation monitor 0-RE-90-225. There is a flow interlock on the CTBD which suspends the release if the flow is less than 20,000 gpm. The low flow interlock may be bypassed and releases made from the Condensate Demineralizer System with less than 20,000 gpm CTBD flow provided the sum of the ratios of the ECls for all release points is s; 10. Turbine Building Sump The TBS normally releases to the low Volume Waste Treatment Pond, but can be released to either the Metal Waste Cleaning Ponds or the 35 acre Yard Holding Pond. The TBS has a capacity of 57,783 gal and a design discharge release rate of 2,000 gpm per pump. The TBS is monitored by radiation monitor 0-RE-90-212. Steam Generator Blowdown The SGBD is processed in the Steam Generator Blowdown Flash Tanks or SGBD Heat Exchangers. Figure 6.3 provides a schematic of the SGBD System, showing the flow pathways and flow rates. The SGBD discharge has a maximum design flow rate of 65.5 gpm per steam generator when discharging to the Cooling Tower Blowdown. SGBD is recycled or is released to the CTBD line and monitored by radiation monitors 1 ,2-RE-90-120A and 1,2-RE-90-121. There is a flow interlock on the CTBD which suspends the release if the flow is less than 20,000 gpm. The low flow interlock may be bypassed and releases made from the Steam Generator Blowdown System with less than 20,000 gpm CTBD flow provided the sum of the ratios of the ECls for all release points is s; 10. Groundwater Sump The Groundwater Sump (GWS) normally releases to the 35 acre Yard Holding Pond via the storm drains. The GWS has a design discharge release rate of 300 gpm per pump. The GWS is not monitored by a radiation monitor.
WBN OFFSITE DOSE CALCULATION MANUAL Revision 22 ( o (ODCM) Page 98 of 195 Figure 6.1 - LIQUID EFFLUENT RELEASE POINTS
) Turbine Building Sump Condensate Demineralizer System -
O-RE-~-225 O-RE-2?-225 I~ O-RE-90-212 .
~f ~
Steam Generator Blowdown Slowdown Unit 1 - 1-RE-90-120A,121 Metal Cleaning Waste Ponds Low Volume Waste low Treatment Pond Steam Generator 2-RE-90-120A,121
'2-RE-90-120A,121 Blowdown Unit2 Unit 2 Ir ERCW 1I JO-RE-90-133,140 lo-RE:O-133,140 t
Holding Pond Liquid O-RE-90-134,141 o-RE-90-134,141 O-RE-90-122 Radwaste System (/ (j D I I lL U T I o N F Blowdown Cooling Tower Slowdown lL o W W
--- At At- < 3t:oo 3~ efi ~ r\
"'1 \lU' VU"' f(ou)
-How lfoJ\lQ.,..
\lQ\VI2....
ck~s ckJ" tUQd~$ ~ol\~ CVr\ d. ruoat-es Glr\d. ~Cl II ~ 'Tower
'TC:Hj.)~
8Io\J,)do~ 8iowck\l-.)r') ..p~ ~~ -thIO\)~-thro\J~h ~m.cU\~ h ctQ <5TI\oJ\QX"'"' h~tr fQ) -tQ) ~ f{,lclo~ Hol~h~ Pend, Pond,
- Co~(\~
a,c(\~ ""TO\J..:Qr
-rO\J.:2r BlowcbU)n Blo\pcb~n flouj f\ovJ die;
<~51~C>
-<~5,roQ) [jfYY1 8f1Yl tLSC>\cd-~
'lSc\qj~b~ h~u~d U'K~ r<J.UXlSt-e.
re~~" re,~<a" '5c;z. DrOJ,.\..)"1j J-
<See. l:>1'O.>>.J'llj J- L\1~<ct{ -7'7r2..
LIt,(WCal( -77 . . 2,.
~O(" ~\()Xlod-\~
{O\' ~ \(}.XIct\- \m .....
4 5 6 7 8 9 AUX "UX BLDG I"L A [OPT f'L .. EOPT OR PUUP PUMP NOTES: I. FOR SYMBOLS SYUBOLS AND CENERAL GENERAL NOTES, NOTES. SEE A NON TRITIATED rOPT DRAIN H["DER~;"'..,. HEADER~;""..,. _ _ _ _- - , 1-471611-77-1. A
~~~~~O~~O'ER
~~~~~O~~O'[R _-
TO LAUNDRY AND HOT RB a. [CPT RS FL ... DR SUIoCP SUIM' EOPT { SHOIER SHOWER PU.... PUMP COlLECTOR YANI( FLOOR DRAIN COLLECTOR TANK WOS LAUN.DRY LAUNDRY TANK PUMP lANI( PUI.4P B B PLANT VENT (< SP ADDITIONAL [OPT EOPT BLDG
- a. [OPT ORAl FLOOR ... DRAINIt StIWP (UI.
SUt.lP (UI, U2)U21 LAUNDRY LAUNDRY VENT CHEMICAL CHEl.HCAL AND HOT AND HOT DRAIN SHOWER SHOWER SHorER TANK YANK TANK A TANK 8 YANK B r--------------,I I CONTROL TYPICAL c ~~~~~~¥tS~L
!I ~~~~;~¥~S~L PUI.4P PUUP I O-HS-77-29 A .. B AND I It .t.
! c TO t.tOBILE OEI.4INERAL!ZER
~BILE OEI.4INERALlZER L______
I O-XS-77-29_______ J I TRITt." TED DRAIN. TRITtATED DRAIN COllECTOR TANK 1-471611-77-3 (F (1-"-4)
-4)
TO /r,()NITOR TANK FLOOR flOOR DRAIN COLLECTOR YANK TANK DISCH OISCH PUWP PUt.tP IIA a. HOl LAUNDRY ... HOT SHOliER SHOWER TANK PUI#' PUlE' D D L---!><l---f,J-...J----r---------------------------f'-..J----t:<I------------------------..L..,.-t:I<J---f,l---~~2~I~lgg~lECTOR L---!><l---f,J-...J----r---------------------------f'-..J----t:<I---------------*---------..L..,.-t:I<J---f,I---~~2~I~Lgg~LECTOR TANK TANK CASK DECONTAMINATION TANK CooLi NG TO'ER COOLING TOWER BLOWDOIN BLOWDO'N ( FLOOR DRAIN COLLECTOR COllECTOR TANK DISCH PlU' PUt.4P B E E BUILDING EXHAUST F PUMP F LEVEL < LO La LEVEL < LO La PROJECT fACILITY LEVEL < LO POWERHOUSE G UNITS 1 II:. a:. 2 PI..It.IP A TITLE COOLING TO COOLINC
} TOWER BLOWOOWN TO.ER BLOtDO'"
} LINE ELECTRICAL LOGIC DIAGRAM CONDENSATE DEIAIM
'ASTE EIIAP FEED PUt.f' COlE BLDG CASK DECONTAMINATION DECONTAIAINATION COllECTOR TANK COLLECTOR 1-471611-77-3 (B-12) 1-47'11611-77-3 (9-12)
WASTE DISPOSAL SYSTEM WATTS BAR NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY Q Q DESIGN INITIAL ISSUE ENGINEERING APPROVAL DRAHER H DRAF'TER G.'II. G.W. JOHNSON CHECKER C.A. ATKINS RO ISSUE PER GARY BARNARD CARY DESIGNER REVIEWER EAI 3.10" 3.10 .. RIMS ROBERT O. D. IoIURR t.lURR WASTE CONDENSATE PUMPS T28
- 92S2 0225 859 85S J lYKE fOR C.C. LYICE FOR MeB orr OFF ISSUED BY: BY; DATE
________ !.~._J..9ti..!4~~__
!.~._J2H~~~__ 3/5/92 J/5/92 85 1-47W611-77-2 R5 RS 2 3 4 5 7 PRINTS REQ'O PIIIMrs MQ'O R - r,,~.~"'ROO-I,
~"C;;ROO-I,.."c."""",,,,", *.-',..""""'
"0'0",,",,,<'..;.-';;
'-.<', "'-,,,"-,"','0..,..,..,
""'"",",,0"-,,,",' ,-'-;;;.."""'........
..., '0$0;'-'-'
50;'-'--' PRQJ Alii CE BR OR PROJ Ct [E tE WE Nt n: H[ N[ r[ HE st SE TE lit. 8l. IIf so IF llil SQ CAD MAINTAINED DRAWING (CONFIGURATION CONTROL DRAWING)
AOI-22 WBN BREAK OF DOWNSTREAM DAM Revision 18 Page 3 of 12 . 1.0 PURPOSE This Instruction provides actions to be taken in the event of a break in the downstream dam. 2.0 SYMPTOMS 2.1 Alarms A. OBE aBE SPECTRA EXCEEDED [1660] [166D] B. SEISMIC RECORDING INITIATED [166E] 2.2 Indications A. Notification of downstream dam break is received from Systems Operations Center (Load Coordinator) or River Systems Operations. 2.3 Automatic Actions A. IF flow through Watts Bar Dam drops to less than 3500 cfs, the following actions should occur:
- a. 0-FCV-27-100, 54" DIFFUSER A SUP, CLOSES
- b. 0-FCV-27-1 01,48" 0-FCV-27-101, 48" DIFFUSER B SUP, CLOSES
- c. 0-FCV-27-97, HOLDING POND ISOL, OPENS B. IF cooling tower blowdown flow drops to less than 25,000 gpm, the following valves will close:
- a. 1-FCV-15-8, SG BLOWDOWN DISCH TO CTBD
- b. 1-FCV-15-44, SG BLOWDOWN DISCH TO CTBD, unless bypassed in accordance with SOI-15.01.
- c. 0-RCV-77-43, COOLING TOWER BLDN LN RADIATION RELEASE CNTL
((
TVA
Title:
STEAM GENERATOR BLOWDOWN SYSTEM N3-1S-4002 2.2.3 Single Failure Requirements The safety-related portions of the SGB System, i.e., the piping out to and including the containment isolation valves shall be designed to accomplish the system's safety function in the presence of any single failure. A containment isolation signal is provided to close the containment isolation valves and the blowdown isolation valves. The safety function will be accomplished in the event of failure of one power train. The system shall be designed in accordance with single failure criteria as defined in WB-DC-40-64, "Design Basis Events Design Criteria" (Ref. (Re f. 7.2.31).
- 7. 2 . 31) .
2.2.4 Radiation Release or Hazardous Material Control Requirements The SGB system shall be designed to limit radiological release as defined in WB-DC-40-64, "Design Basis Events Design Criteria" (Ref. 7.2.31). Radiation monitors shall continuously monitor the blowdown lines downstream of the heat exchangers and the flash tank. The radiation monitors shall automatically divert the blowdown from the CTB destination destination to the CDSV destination whenever the radioactivity concentration reaches a level determined in accordance with procedures established in the Offsite Dose Calculation Manual (ODCM), (Ref. 7.2.22). An alarm shall be set up to alert an operator of increasing radioactivity level at a setpoint also determined per the ODCM. Instrumentation shall be such that setpoints are adjustable to allow for various CTB operating conditions. This design feature shall prevent the release of unacceptable amounts of radioactive water to the river in the event of a primary-to-secondary leak in a steam generator. The sample system which samples each steam generator may be used to determine which steam generator is leaking. Since the Class B SGB piping penetrates the containment, it shall be designed to comply with the requirements WB-DC-40-34, "Containment Isolation" (Ref. 7.2.7). 7.2.7) . 2.2.S Environmental Qualification Requirements The safety-related components of the SGB system which are located in a harsh environment shall be environmentally qualified to accommodate the effects of the environmental conditions associated with normal operation, maintenance, testing and postulated accidents which require the SGB system to function. The criteria listed below shall be used for the design and procurement of new equipment for plant additions, and for the procurement of replacement equipment: WB-DC-40-42 Environmental Design (Ref. 7.2.6) 10CFRSO Appendix A - GDC No. 4 - Environmental and Missiles Design Basis (Ref. 7.S.3) 10CFRSO.49 Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants (Re f. 7. S . 6) S
TVA
Title:
LIQUID RADWASTE PROCESSING SYSTEM N3-77C-4001 If results of analysis indicate radioactivity levels and chemistry characteristics are permissible for discharge, the contents are discharged and diluted with CTB flow and discharged to the reservoir. If results of the analysis indicate radioactive limits are higher than discharge limits, then the contents are recycled through the MWDS. During this mode the recirculation valve should be throttled such that the flow to be recycled to the MWDS is maintained (Ref 7.1.5.29). 2.2.13.15 Mobile Waste Demineralizer System The mobile waste demineralizer system (MWDS) shall remove soluble and suspended radioactive materials from the waste stream via ion exchange and filtration. The MWDS shall process liquids from the tritiated drain collector tank and the floor drain collector tank at a feed rate of 40 gal/min. During peak flow rates, the MWDS may process higher flow rates by throttling the recirculation valves for the FDCT and TDCT based on MWDS vendor recommendation. The MWDS shall also have the capability to receive liquids from the CVCS holdup tanks. MWDS vendor shall sluice spent resin to a liner or MDSRDC, i.e., a Rad-Vault with a HIC, located in the Auxiliary Building Railroad Bay. The Rad-Vault is a concrete storage container that provides process shielding. After waste is dewatered, the liner/HIC is loaded into a transport cask within the Auxiliary Building Railroad Bay. Note: Precautions must be taken during operation to prevent the flow of process water to the demineralizer system. 2.2.13.16 Deleted 2.2.13.17 Releases of Waste Release of radioactive liquid from the liquid radwaste processing system shall be from the CDCT, monitor tank, chemical drain tank, and the laundry and hot shower tanks. These tanks empty into the theCTB CTB line where their contents are diluted with the CTB before being discharged to the river (Chickamauga Lake) . Release of the radioactive radioac-ti ve liquids from the liquid waste system shall be made only after laboratory analysis of the tank contents. Once the fluids are sampled, analyzed and verified to meet discharge criteria, hd~\c:;&... ~U"\G~ S~ they shall be pumped to the discharge pipe which is routed to the CTB. There shall be no discharge when the Tennessee River flow is less than fOO ~ d.-5 f OCJ j ~'{'i\-
~'{'i\ - 3,500 cubic feet per second or the cooling tower blowdown flow is less than 20,000 GPM. Positive interlocks with the Watts Bar Hydroelectric Plant shall be provided to assure compliance with the Tennessee River flow requirement (Ref 7.1.2.7). The radwaste discharge pipe shall be provided with a remotely operated control valve interlocked with a radiation monitor and a flow transmitter in the CTB to assure that sufficient dilution flow (20,000 GPM) is available for the discharge of radioactive liquids from the site. Referen~es ReferenCes 7.1.1.10 and 7.1.1.14 have been issued by the NRC to provide regulations and guidelines for releases of radioactive liquids and shall be consulted prior to final system design.
14
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 23. 067 AK1.01 023 For a fire in the charcoal filter unit of the CREVS system, which ONE of the following describes the classification of the fire and what type of extinguishing agent is used t9 put out the fire?
Fire Classification Extinguishing Agent A'I A A Water B. D Foam c. C. A Foam D. D Water DISTRACTOR ANAL YSIS: A. CORRECT. Since charcoal is a derivative, or product of wood, it is a Class A fire. Class A fires are generally extinguished with water. B. Incorrect. Plausible, if applicant mistakenly reasons that charcoal is not any of the materials listed for a Class A fire (paper, cloth, wood, etc.), and believes that ( charcoal is a form of "mineral" (or similar reasoning) and then misapplies the definition of Class D 0 fires. The use of foam as an extinguishing agent is plausible, since it will provide some extinguishment of a fire, but it is not the type of agent specified for this fire, and is not part of plant design for this location. C. Incorrect. Plausible, since the class of fire is correct, and because foam can be used on Class A fires. However, the Control Room HVA C system charcoal filter units do not have foam extinguishing capability. D. Incorrect. Plausible, since the charcoal filter units do have water spray capability for extinguishing a fire. However, the applicant applies the wrong classification to the type of fire for these charcoal units. Page 62
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 23 Tier: _1_ Group 2 KIA: 067AK1.01 Plant Fire On-site Knowledge of the operational implications of the following concepts as they apply to Plant Fire on Site: Fire classifications by type. Importance Rating: 2.9 / 3.9 2.9/3.9 10 CFRPart 55: 10CFRPart55: 41.8/41.10/45.3 10CFR55.43.b: Not applicable. KIA Match: Applicant must recognize conditions involving a fire, and determine the classification of the fire, and which extinguishing agent is to be used for fighting this fire. Technical
Reference:
FPT-312.000, "Watts Bar Fire Protection Training, Duties and Responsibilities, Section VI., Classes of Fire." SOI-13.01, Fire Detection System, Rev 22 N3-30CB-4002, Control Building Heating, Ventilating, Air ( Conditioning, And Air Cleanup System, Rev.15 Part X X - NFPA Code Evaluation, Rev 10 Proposed references None to be provided: Learning Objective: None identified. Question Source: New X X Modified Bank Bank Question History: New question Comments: Page 63
~~--~--~----~-.--- FPT -112.000 FPT-112.000 Revision 4 of IS Page 16 of18
- x. Lesson Body X. Instructor Notes
- 3. Disadvantages
- a. No protection from heat for the user.
- b. High concentrations may be harmful or fatal to humans, if inhaled for a short period.
- c. Very expensive.
- d. NOTE: Whenever you must enter a Halon or C02 protected area where the system has dumped, you must wear SCBA and deal with limited visibility.
VI. Classes of Fire The basic chemistry of fire requires heat, oxygen, fuel and a chemical Remind everyone that a reaction. Which is true of all fires. Some unique aspects of a fire will fire can have result in the fire being characterized into one of four classes which characteristics of tends to dictate the optimum extinguishing agent. multiple fire classes. A. Class A Fire ( 1. Deep seated - charring occurs.
- 2. Examples of material
- a. PVC cable insulation
- b. Wood and paper
- 3. Best extinguishing agent - water spray - cools, penetrates, has a relatively long reach and affords protection for personnel.
B. Class B Fires
- 1. Flammable and combustible liquids and gases.
- 2. Examples of material
- a. Gasoline and fuel oil
- b. Hydrogen and acetylene
- 3. Extinguishing method - isolate supply valve and extinguish remaining fire with Dry Chemical, Halon or Foam C. Class C Fires
- 1. Relates to fire involving energized equipment. Pertains to an existing condition not necessarily to what is burning.
- 2. Almost all class C fires involve multiple classes:
- a. Class C and A Fires, i.e., energized cable insulation burning, 6.9 kV, 4160 V and 480 V Boards.
- b. Class C and B Fires, i.e., energized lube oil pump leaking and burning.
- 3. Best Extinguishing Agent - water spray - relatively long reach, provides personnel protection and won't conduct electricity when used properly (Fire Brigade Only).
- 4. Best extinguishing method - de-energize and handle as a Class A and/or a Class B Fire.
FPT-112.000 FPT -112.000 Revision 4 Page 17 of18 X. Lesson Body Instructor Notes D. Class D Fires
- 1. Involves combustible metals
- 2. Specialized extinguishing agents are required.
- 3. Water and Dry Chemical are ineffective.
VII. Incident Command System - Mitigating Actions The roles and responsibilities of the Incident Commander and Fire Brigade Leader will change, should a major event occur in which the incident is beyond the capabilities of the site emergency response tern. In order to effective manage the emergency scene, the OPS IC will become the Incident Manager. The BFN Emergency response Organization (Fire Ops) will provide an Incident Commander. VIII. Overview of the Emergency Response Organization - Mitigating Actions A. Shift Manager (SM), Ops. ( 1. Assumes overall responsibilities of the plant.
- 2. In case of multiple fire events occurring simultaneously, the SM has the authority to direct fire fighting resources to the fire emergency which presents the greatest threat to public safety.
- 3. Receives technical information from the Incident Manager that may be required to safely shut down an operating reactor if a fire ocCurs occurs and damages safety - related equipment or components.
B. Operations Incident Manager (OIM) - Senior Reactor Operator (SRO) or Unit Supervisor (US), Ops. (In smaller fires this person would be the IC.) The OIM position will be responsible for coordinating and communication with the operations department through the Main Control Room (MCR) the Operations Support Center (OSC) Operations positions. This position should be filled initially by the on shift Operations Incident Commander C. Incident Commander (IC) (lC) - Assigned by the BFN Emergency response Organization (Fire Ops). (In smaller fires this person
WEN SOI-13.01 0 FIRE DETECTION SYSTEM Revision 22 Page 43 of 441 APPENDIX F Page 2 of 24 OPERATOR ACTIONS 11 ,5 Control Building Page 1 of 6 ROOM ZONES DAMPERS FANS TO STOP DRAWINGS Mech Eqpt Rm - EL 755 214, 215 0-31-3932 MCR AHU A-A 45W760-31-11 Room C1 0-31-3933 MCR AHU B-B 45W760-31-12 0-31-3934 Women's Toilet - EL 755 227,228 0-31-4614 Spreading Room Supply Fan 45W760-31-7 Room C2 0-31-5004 Note: Zone 214 will stop 45W600-26-15 0-31-5006 fan. Corridor - EL 755 227,228 0-31-5007 45W760-31-13 Room C3 0-31-5008 Elec Bd Rm.AHU A-A, B-A 45W760-31-14 0-31-5009 Elec Bd Rm AHU C-B, D-B Main Control Room - EL 755 220 0-31-5012 45W760-31-8 Room C12 0-31-5013 Batt Rm El 692 Exh Fan A-A 0-31-5016 Batt Rm El 692 Exh Fan B-B 0-31-5017 0-31-5018 0-31-5019 0-31-5020 0-31-5021 0-31-5022 0-31-5023 0-31-5024 0-31-5025 0-31-5026 0-XFD-31-83 0-XFD-31-86 Note: Zone 220 will close 0-XFD-31-153 O-XFD-31-86. Control Bldg Ener Air Supply 218, 219 0-31-3935 Control Bldg Emer Air 1-45W760-31-9 ( Cleanup Charcoal Filter Train A Cleanup Fan A-A (Mech Equip Rm EL 755) Control Bldg Ener Air Supply 216, 217 0-31-3936 Control Bldg Emer Air 1-45W760-31-9 Cleanup Charcoal Filter Cleanup Fan B-B Train B (Mech Equip Rm EL 755) soi13
TVA
Title:
CONTROL BUILDING HEATING, VENTILATING, AIR N3-30CB-4002 CONDITIONING, AND AIR CLEANUP SYSTEM 2.2.8 Separation, Fire Protection, protection, or Intrazonal Protection The primary safety-related portions of the Control Building HVACAC system shall be designed to provide two independent redundant trains. The two independent trains of equipment shall be sufficiently separated by fire barriers and/or distance from each other and from other safety-related equipment as required by Ref. 7.2.28 and to eliminate the possibility of a single event destroying the capability for the Control Building HVACAC system to remain functional during a safe shutdown event. Fire protection shall be provided for the charcoal adsorbers within the MCR emergency air cleanup units (ACUs) by an automatic closed-head water sptay spray system. The fire protection water sprayed on the charcoal adsorbers shall be removed by drainage piping designed to withstand a design basis seismic event to ensure that the ACU housing pressure boundary is maintained through the drain traps. Separate trains of emergency power and control power shall be used to drive and control the Control Building HVACAC system. The following criteria shall govern the separation of redundant systems: WB-DC-30 Separation/Isolation (Ref. 7.2.1). WB-DC-40-31.50 - Evaluating the Effects of a Pipe Failure Inside and Outside Containment (Ref. 7.2.25). WB-DC-40-31.51 - Evaluating the Effects of Flooding Due to Moderate Energy Pipe Failure Inside and Outside of Containment (Ref. 7.2.24). WB-DC-40 Fire Protection (Ref. 7.2.26). ( 2.2.9 Electrical Power Trained and nontrained electrical power shall be supplied to the Control Building HVACAC system in accordance with design criteria WB-DC-30-28 (Ref. 7.2.2). Trained and nontrained control and instrumentation power shall be supplied in accordance with design criteria WB-DC-30-27 (Ref. 7.2.11). Grounding of electrical components shall be in accordance with design criteria WB-DC-30-32 ((Re Re f. 7.2.
- 7. 2 . 16) .
For power supplies to major components, see Table 9.2. 2.2.10 Instrumentation and Control Safety-related HVACAC subsystems shall be operable from the MCR. Safety-related system instrumentation and controls shall meet applicable design requirements including single failure criteria, separation criteria, and environmental qualification requirements. These instruments and controls shall also meet the requirements of Ref. 7.5.1. Required Instrumentation and Controls for the Control Building HVACAC System are identified in Section 3.3. Instrumentation shall be provided to monitor MCRHZ pressure and damper status during post accident conditions as required by WB-DC-( Re f 7.2. 3 0 - 7 (Re 30-7 7. 2 . 31) 3 1) . 2.2.11 System Interfaces For Control Building HVACAC system interfaces, see Section 3.6. 6
Rev. 10 PART X - NFPA CODE EVALUATION 3.3.1.1 Reactor Coolant Pumps (RCP) Each of the four RCPs are protected by a closed head, automatic pre-action water spray fixed system. Listed nozzles are located around the top of the RCP motor along a ring header. The systems are hydraulically designed in accordance with NFPA 13 & 15 and produce a minimum design density of 0.25 gpmlft2. gpm/fe. This density corresponds to NFPA 15 requirement for transformer protection (oil hazard). The pre-action system is automatically actuated upon initiation of cross zoned thermal detectors located above the motor. 3.3.1.2 Charcoal Filters Closed head, automatic pre-action, water spray fixed systems are provided for the Control Building Emergency Air Supply Unit (CBEASU), Containment Purge Air Exhaust System (CPAES), Post Accident Sampling Filter (PASF), Emergency Gas Treatment System (EGTS), and the Auxiliary Building Gas Treatment System (ABGTS). The use of water spray fixed systems for protection of charcoal filters is the only type of suppression system recognized by NFPA 803-1983 (Table 10-1.2). The systems are hydraulically designed in accordance with NFPA 13 & 15 and produce a minimum 2 gpmlft2 across the surface of the exposed filters. design density of 0.25 gpm/ft In accordance with NFPA 15, Section 4-4.1.2, a design density was determined for this unique application based on analysis of the combustibles. The flow density of 0.25 2 gpmlft gprnlfe corresponds to the flow density specified in Section 5-4.6.1 of NFPA 850, Recommended Practice for Fire Protection for Fossil Fueled Steam* Steam Electric Generating Plants" for sprinklers provided for coal handling structures subject to accumulations of coal or coal dust. Charcoal filters have a lower heat energy value (Le., (i.e., Btu/volume) than coal (based on lesser density and che-mistry chemistry of the material). The flow density of 2 gpmlft2 over the plan area for bag-type dust collectors (in coal handling facilities) is 0.20 gpm/ft adequate for the hazard. Therefore the use of 0.25 gpm/ft gpm/ff2 for the charcoal is conservative. The temperatures for the fusible elements in the spray nozzles was determined by correlating the maximum temperature expected in the filter units to the recommendations of NFPA 13, Table 2-2.3.1, "Temperature Ratings, Classifications and Color Codings". In the event of a charcoal fire in the ABGTS, EGTS, CBEASC or CPAES filter units, the smoke detectors will annunciate the alarm and trip the spray system's deluge valve, allowing water up to the closed nozzles. After shutdown of the fan and closure of the downstream damper, the heat will build inside the unit. Because the unit is constructed of heavy gauge metal, it will essentially act as an oven. This will ensure operation of the thermal elements in the nozzles if they have not previously been activated by the fire. The PASF unit's suppression system is is similar to the other units but is not provided with a dedicated flow control valve (FCV). The system is supplied by the general area preaction sprinkler system. A manual isolation valve located at the unit must be opened upon confirmation of a fire to allow water into the unit.
--r xX-7
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 24. 069 AAl.OI 024 Given the following:
- A Steam Line break occurs on Unit 1.
Containment pressure is 3.7 psig. Control Air pressure is 97 psig. During the performance of E-O, "Reactor Trip or Safety Injection" the Unit Supervisor directs the CRO perform E-O Appendices A and B to evaluate the support systems. Containment Isolation Status Panel 1-XX-55-6E window 13 "FCV-32-11 0" indicates RED. Which ONE of the following identifies the correct CRO response to the window indication? A. The valve is in the correct position for the current conditions; continue with the support system evaluation. B. Verify the penetration is isolated by verifying the Train B valve is closed in the flow path. C'!'" If the FCV can NOT be closed, direct an AUO to close manual isolation valve C!' in the penetration flow path. ( D. If the FCV can NOT be closed, direct an AUO to stop all Control & & Station Air Compressors. Page 64
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, the FCV is not in the correct position, the window should be indicating GREEN because a Phase B isolation has occurred. Plausible because if the containment pressure had been less than Phase B setpoint, the window being RED would be correct and because even after the isolation, a red window could be normal as in the case with the window for the containment air return fan which will start after a time delay. B. Incorrect, Verifying the other Train has the penetration isolated is not correct because the inside containment isolation valve for this penetration is a check valve. Plausible because on most containment penetrations there are both Train A and Train B isolation valves that have indications on the Containment Isolation Status Panels that can be checked. C. Correct, E-O, Appendix A directs the FCV be closed and if it cannot be closed Appendix B directs an AUO be dispatched to perform Attachment B1 which will 0-ISV-32-1013 - CONTROL AIR EL 713 AB HDR ISOL direct the AUO to "CLOSE O-ISV-32-1013 [A6/S EL. 713J (chain operated - behind Fuel and Waste Handling Bd. A)." D. Incorrect, If the FCV could not be closed shutting down the air compressors is how the action required. E-O, Appendix A directs the FCV be closed and if it cannot be closed Appendix B directs an AUO be dispatched to perform Attachment B1 which ( will direct the AUO to close the manual valve O-ISV-32-1013. 0-ISV-32-1013. Plausible because shutting down the air compressors and removing their breakers is the correct AUO action if the manual valve could not be closed. Page 65
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 24 Tier: _1__ Group 2 KIA: 069 AA1.01 Loss of Containment Integrity Ability to operate and / or monitor the following as they apply to the Loss of Containment Integrity: Isolation valves, dampers, and electropneumatic devices Importance Rating: 3.5.// 3.7 3.5 10 CFR Part 55: 41.7/45.5/45.6 10CFR55.43.b: Not applicable. KIA Match: The applicant is required to determine how a failed valve affects the status of containment integrity and the required action related to the failure. ' Technical
Reference:
E-O, "Reactor Trip or Safety Injection," Rev 27 ( Proposed references None
.to be provided:
Learning Objective: 3-0T -EOPOOOO 3-0T-EOPOOOO
- 15. Explain the purpose for and basis of each step in E-O, ES-O.O, ES-0.1, ES-0.2, ES-0.3, and ES-OA.
3-0T -SYS088A 3-0T-SYS088A
- 9. Given a set of plant conditions, determine the correct response of the Containment Isolation System.
Question Source: New X . Modified Bank Bank Question History: New question Comments: Page 66
WBN REACTOR TRIP OR SAFETY INJECTION E-O Rev 27 IStep I IAction/Expected Response I IResponse Not Obtained
- 5. EVALUATE support systems:
- REFER TO Appendixes A and B B (E-G),
(E-O), Equipment Verification pages 15-28.
- 6. ANNOUNCE reactor trip and safety injection over PA system.
- 7. ENSURE secondary heat sink ** GO TO FR-H.1, available with either: Loss of Secondary Heat Sink.
- Total AFW flow greater than 410 41 G gpm, OR At least one S/G N R level greater than 29% [39% ADV].
( 5 of 28
WBN REACTOR TRIP OR SAFETY INJECTION E-O ( Rev 27 I1Step I1 I1Action/Expected Response II
.1 1Response Not Obtained APPENDIXA (E-O)
Page 4 of 10 EQUIPMENT VERIFICATION
- 6. CHECK cntmt pressure: PERFORM the following:
- B DARK [MISSP].
Phase B 1) ENSURE Phase B actuated.
- Cntmt Spray DARK [MISSP]. 2) ENSURE Cntmt Spray
- Cntmt press less than 2.8 psig. actuated.
- 3) ENSURE cntmt spray pumps running.
- 4) ENSURE cntmt spray flow.
- 5) ENSURE Phase B isolation:
- Train A GREEN.
(
- Train B GREEN
- Manually CLOSE valves and dampers as necessary.
- 6) STOP all RCPs.
- 7) ENSURE MSIVs and bypasses CLOSED.
- 8) PLACE steam dump controls OFF.
- 9) WHEN 10 minutes has elapsed B actuated, since Phase B THEN ENSURE air return fans start.
- 10) USE adverse cntmt [ADV]
setpoints where provided. 17 of 28
WBN REACTOR TRIP OR SAFETY INJECTION E-O Rev 27 IStep I IAction/Expected Response I IRespon~e Not Obtained APPENDIXA (E-O) Page 10 of 10 EQUIPMENT VERIFICATION
- 23. ENSURE Control Building fans Manually STOP fans.
STOPPED and dampers CLOSED:
- SPREADING ROOM SUPPLY and NOTIFY TSCif TSC if any EXH FANS AND dampers.
dampers . damper NOT CLOSED.
- TOILET & LKR RM EXHAUST FAN AND dampers.
- 24. INITIATE Appendix B.
( 23 of 28 230f28
WBN REACTOR TRIP OR SAFETY INJECTION E-O ( Rev 27 IStep I IAction/Expected Response I IResponse Not Obtained APPENDIX B (E-O) Page 1 of 1 PHASE B PIPE BREAK CONTINGENCIES
- 1. CHECK PHASE B 8 actuated. WHEN PHASE B 8 actuation occurs; (MISSP XX-55-6C, -60) THEN GO TO step 2.
- 2. ENSURE 1-FCV-32-110 CLOSED. DISPATCH AUO to perform (CISP XX-55-6E) ATTACHMENT B1.81.
[A-train, window 13] ( 3. ENSURE 1-FCV-67-107 CLOSED. DISPATCH AUO to perform (CISP XX-55-6E) ATTACHMENT B2.82. [A -train, window 43]
- 4. ENSURE 1-FCV-70-92 CLOSED. DISPATCH AUO to perform (CISP XX-55-6E) ATTACHMENT B3.83.
[A -train, window 73]
- 5. ENSURE 1-FCV-70-140 CLOSED. DISPATCH AUO to perform (CISP XX-55-6F) ATTACHMENT B4.84.
[B -train, window 74] [8 (
\
24 of 28
WBN REACTOR TRIP OR SAFETY INJECTION E-O ( Rev 27 ATTACHMENT B1 (E-O) Page 1 of 1 CONTROL AIR ISOLATION
- 1. CLOSE O-ISV-32-1013 - CONTROL AIR EL 713 AB HOR ISOL
[A6/S EL. 713] (chain operated - behind Fuel and Waste Handling Bd. A). II'
/ / 2. IF O-ISV-32-1013 CANNOT BECLOSEO, 2.
THEN: BE CLOSED, OPEN and DISCONNECT C&SS air compressor breakers: a) O-BKR-32 480V SO BO 1A2-A, 1A2-A, C/30 b) O-BKR-32 480V SO BO 1B1-B, 1B1-B, C/30 c) O-BKR-32 480V AUX BLDG COM BO, C/6C d) O-BKR-32-4900A - 480V TURB BLDG COM COM BO, C/6C I~
~ .
25 of 28
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 25. W/EOI EA1.3 025 Given the following:
Unit 1 experienced an inadvertent Safety Injection.
- The operating crew entered E-O, " Reactor Trip or Safety Injection" and has reached the step for checking if SI termination criteria are met.
Due to equipment failures and conflicting information the crew is uncertain as to the correct course of action and is contemplating use of ES-O.O, "Rediagnosis."
"Red iagnosis."
Current 'conditions are: SG#1 SG#2 SG#3 SG#4 Levels 34%NR 22% NR 29%NR 38%NR AFW flow 40gpm 260 gpm 130 gpm o gpm RCS Subcooling 66°F RCS Pressure 1630 psig and rising Pressurizer Level 11 % and rising Which ONE of the following identifies ... (1) if current conditions allow the termination of safety injection, and (2) an additional requirement that must be met prior to implementation of ES-O.O?" A. (1) SI Termination criteria are currently met (2) Status Trees must be implemented. B. (1) SI Termination criteria are currently met (2) Transition from E-O must be made. C. (1) SI Termination criteria are NOT currently met (2) Status Trees must be implemented. D~ (1) SI Termination criteria are NOT currently met D'f (2) Transition from E-O must be made. Page 67
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, Pressurizer level is not high enough to meet SI termination criteria (but all other criteria is met) and there is no requirement to be monitoring Status Trees prior to implementing ES-O. O. Plausible because all criteria except the pressurizer level is met and Status Tree implementation would check all critical safety functions for any challenges .. B. Incorrect, Pressurizer level is not high enough to meet SI termination criteria (but all other criteria is met) and ES-O.O cannot be implemented until a transition from E-O has been made. Plausible because all criteria except the pressurizer level is met and the requirement to have made a transition from E-O is correct. C. Incorrect, SI termination criteria is not currently met because the pressurizer level is below 15%.but there is no requirement to be monitoring Status Trees prior to implementing ES-O. O. Plausible because SI termination criteria is not currently met is correct and Status Tree implementation would check all critical safety functions for any challenges. D. Correct, SI termination criteria is not currently met because the pressurizer level is ES-O.O0 cannot be implemented until a transition from E-O has been below 15% and ES-O. made. ( ( Page 68
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 25 Tier: _2_ Group 2 KIA: W/E01 EA1.3 Rediagnosis Ability to operate and Ilor or monitor the following as they apply to the (Reactor Trip or Safety Injection/Rediagnosis): Desired operating results during abnormal and emergency situations. Importance Rating: 3.4 I 3.8 10 CFR Part 55: 41.7 I 45.5, 45.6 10CFR55.43.b: Not applicable. KIA Match: Applicant is required to recall conditions (desired operating results) that must exist to allow termination of the safety injection and conditional restrictions placed on the implementation of ES-O.O "Rediagnosis." Technical
Reference:
ES-O.O , "Rediagnosis," Rev 2 E-O, "Reactor Trip or Safety Injection," Rev 27 ( TI-12.04, "User's Guide For Abnormal and Emergency Operating Instructions," Rev. 0007 Proposed references None to be provided: Learning Objective: 3-0T-EOPOOOO
- 15. Explain the purpose for and basis of each step in E-O, ES-O.O, ES-0.1, ES-0.2, ES-0.3, and ES-O.4.
3-0T-TI12.04
- 30. List 2 limitations which apply to the use of procedure ES-O.O, Rediagnosis.
Question Source: New X X Modified Bank Bank Question History: New question Comments: Page 69
WBN REACTOR TRIP OR SAFETY INJECTION E-O Rev 27 IStep II Action/Expected Response . IIIIResponse Not Obtained
- 16. CHECK cntmt conditions: ** GOGOTOE-1, TO E-1, Loss of Reactor or Secondary
- Cntmt pressure NORMAL.
Coolant.
- Radiation NORMAL from Appendix A.
- Cntmt sump level NORMAL.
- Cntmt temp ann window DARK
. [104-8].
- 17. CHECK SI termination criteria:
CHECK RCS subcooling a. ** GO TO Step 18. greater than 65°F.
- b. CHECK secondary heat sink
- b. ** GO TO Step 18.
available with either: (
- Total feed flow greater than 410 gpm, OR
- At least one S/G NR level greater than 29%.
- c. CHECK RCS pressure c. ** GO TO Step 18.
stable or rising.
~d.
.~d. CHECK pzr level d. RESTORE pzr level:
greater than 15%. 1 1)) OPERATE pzr sprays to stabilize RCS pressure.
'0 K-\.<v~~'V{\-
"0~-\" . 2) ** GO TO Substep 17a.
-2) 17a .
(;
.;; . ~,~
'<~
. ~\~
~\\:i e. ** GO TO ES-1.1,
\~ t.-==;
t--==; SI TERMINATION.
- 18. MONITOR Status Trees I/ SPDS.
90f28 9 of 28
WBN User's Guide For Abnormal TI-12.04 Unit 1 And Emergency Rev. 0007 Operating Instructions Page 28 of 5464 2.4.4 Status Tree Rules of Usage (continued) F. Once implemented because of any RED or ORANGE path, that Function Restoration Instruction will be performed to completion or to a point of transition UNLESS a higher priority condition develops.
- 1. As a Function Restoration Instruction is performed, the status of that tree may change. This change does NOT change the priority of an instruction in progress.
- 2. If a higher priority condition develops, the instruction in effect should be suspended and the higher priority condition addressed.
G. When no RED or ORANGE path exists, a YELLOW path Function Restoration Instruction can be implemented at the operator's discretion. 2.6 2.5 EOI Network The EOI set is intended to be used as an integral set of instructions in order to to a wide variety of potential plant conditions. respond toa 2.6.1 2.5.1 Rediagnosis ( ES-O.O, Rediagnosis, is a unique Instruction in the EOI network that can be used to respond to conflicting symptoms during an event. This Instruction is intended to be used purely as an operator aid and is entered based solely on operator judgment. It is applicable after transition from E-O, but only if SI has been actuated. ES-O.O provides reassurance that the operator is in the correct instruction series, or provides the necessary transition to the correct instruction series for the existing plant symptoms. 2.6.2 2.5.2 Conflicts in Rules of Priority Some instructions contain action steps which can conflict with actions contained in previously performed instructions. (
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 26. W/E13 EG2.1.31 026 Following an ATWS event that resulted in the MSIVs being closed, the crew has:
- Tripped the reactor.
reactor.
. Transitioned to and implemented ES-O.1, "Reactor Trip Response."
The crew determines the following conditions exist:
- Reactor trip breakers are open.
- Tavg has dropped from a high of 591°F to 566°F.
- All SG levels are between 39% and 41 % NR.
1-HS-3-45, MFW Mode Switch, is in the 'LONG CYCLE RECIRC' position. PORVs on S/G # 1, # 2 and # 4 are throttling open to maintain SG pressure approximately 1025 psig.
- SG #3 pressure is 1223 psig.
- PORV on S/G #3 is closed and cannot be opened from the controller or by placing the PORV control switch to OPEN.
- The crew transitions to FR-H.3 "Steam Generator Overpressure."
Which ONE of the following* following' identifies the expected status of the Feedwater Isolation valves and a procedurally directed action to lower SG #3 pressure? FW Isolation Valves Action to lower SG SG #3 pressure A. Open Cooldown RCS using INTACT S/Gs. ( B~ B:t Closed Cooldown RCS using INTACT S/Gs. C. Open Supply AFW into the steam space. D. Closed Supply AFW into the steam space. Page 70
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect, Feedwater Isolation Valves would not be open. They would be closed because the MFW Mode Switch is in the 'LONG CYCLE RECIRC' position. The procedure direction is provided to cooldown the RCS using INTACT S/Gs. Plausible because no FWI signal has been generated and the applicant may not relate the FW mode switch position closing the valves and cooling down the RCS using INTACT S/Gs is correct. B. Correct, Feedwater Isolation Valves would be closed because the MFW Mode Switch, is in the 'LONG CYCLE RECIRC' position and with Tavg as stated in the stem, the procedure direction is provided to cooldown the RCS using INTACT S/Gs. C. Incorrect, Feedwater Isolation Valves would not be open. They would be closed because the MFW Mode Switch is in the 'LONG CYCLE RECIRC' position. With S/G level elevated, the procedure direction is provided to isolate AFW flow to the affected S/Gs and check AFW flow at zero gpm. Plausible because no FWI signal has been generated and the applicant may not relate the FW mode switch position closing the valves and AFW does enter into the steam space which could cause a pressure drop. D. Incorrect" Feedwater Isolation Valves would be closed because the MFW Mode ( Switch, is in the 'LONG CYCLE RECIRC' position but with level elevated, the procedure direction is provided to isolate AFW flow to the affected S/Gs and check AFW flow at zero gpm. Plausible because FWI valves being closed is correct and AFW does enter into the steam space which could cause a pressure drop. ( Page 71
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 26 Tier: _"_1 _1_ _ Group _2_ 2 KIA: W/E13 EG2.1.31 Steam Generator Overpressure Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup. Importance Rating: 4.6 I 4.3 10 CFR Part 55: 41.10 145.12 I 45.12 10CFR55.43.b: Not applicable. KIA Match: Applicant must determine the expected status of plant components from stated indications and conditions, and also the needed plant lineup to address correcting the condition of overpressure. Technical
Reference:
FR-H.2, "Steam Generator Overpressure," Rev 5 FR-S.1, "Nuclear Power Generation I ATWS," ATWS," Rev 1919" 1-47W611-3-6 R18 ( Proposed references None to be provided: Learning Objective: 3-0T-FRH0001
- 18. Identify the 2 major action categories of procedure FR-H.2 and explain the basis for each major action step.
- 22. Explain the purpose for and basis of each step in FR-H.1, FR-H.2, FR-H.3, FR-HA and FR-H.5.
Question Source: New xX Modified Bank Bank Question History: New question Comments: ( Page 72
WBN NUCLEAR POWER GENERATION I ATWS FR-S.1 Rev 19 IStep IIIIAction/Expected Response I IResponse Not Obtained
- 5. CHECK pzr pressure DEPRESSURIZE RCS:
less than 2335 psig.
- a. ENSURE pzr PORVs and block valves OPEN.
- b. WHEN RCS pressure is psig, less than 2135 pSig, THEN ENSURE pzr PORV or associated block valve CLOSED.
- 6. VERIFY Cntmt Vent Isolation: ACTUATE Phase A and Cntmt Vent Signal, Isolation signal,
- Train A GREEN.
OR
- Train B GREEN. Manually CLOSE valves and dampers as necessary.
(
- 7. IF AFW flow established, THEN
- a. PLACE 1-HS-3-45 to LONG CYCLE RECIRC.
- b. PLACE MFW Bypass Reg Valves in AUTO.
- 8. IF SI actuated OR required, THEN PERFORM Steps 1 through 6 of E-O, Reactor Trip or Safety Injection, as time allows.
5 of 15
9-,- L L9MLt- L 9-£-ll9IALt-l 2 3 4 5 6 7 8 9 10 11 12 TRAINA TRAIN A T~AIN TftAIN 8 FWISLN FW 15LH 1"11' lSlN FW 15lH 25-J-359 (VALVE POS < SP) A A I B TO I I B LOOPS 2,3,4 -v-).,
~5--@
"- , CONT OUTPUT
< SP DCA 5/724-607-0, CVCLE LONG CYCLE DCA 5/939-25-0 ~ 0 RECIRCULATION c
C c C A*P*AUTO A*P-AUTO OPEN~S~CLOSE OPEN ~s ~ CLOSE
- 6. FINAL TIME DELAY SETTINGS WILL BE O[TERI.IIN[O DETERtoIINEO BASED B'!'SED ON TEST OATA DATA OBTAINED FROUFRI:lM F'RE-OP PRE-Of TEST NO. NCS-ZJ8. NCS-2JB.
RO~VED TO PREVENT A CONDITION t-HS-J-JJ8 HAS BEEN REMOVED
- 7. '-HS-J-JJ8 ADV[RSEL'( AFFECTiNG THE SAFETY ADVERSELY SAf[H FUNCTION Of THE ASSOCIATED D VALVE DURING AN ACCIDENT. D S. LO-LO La-La LEVEL CLOSURE OF reV_J_191, 192.
Of' FCV_3_191, 192, 193.193, A ot. 19+ 194 HAS BEEN OEF[ATED DEFEATED BY DISABLING RELAYS K63+A !'iSJ4,r.. .to A K634B, K6J4B.
~Sl~~~f~E~
~SRPP~~f~E~ -~N~6FU.l~Tlo=~77.
-~N~6FU.ltrlot77. REFER 1-+~'760-J-a REfER TO 1-4!!1W760-J-8
( I I I E II I SPV-J-J,!;'" E I I I LL______ ~~~!~~ ________ J
~~~~~~ ________
F
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ORAWHIG SUPERSEDES UNIT 1\ CONFIGURATION CONTROL DRAWING THIS CONrIr;URATlON AS*CONSTRUCTED AS-CONSTRUCTEO ORAWINC H'611-3-6 REVlSION ORA'INC; 471/S11-3-6 REVISION O. LOOP 2.J O~ + LOW FLO. 18 1/-I7-Ofp
/1-/7-1:&
REVISED PER DCA'S 51724--507-0, -581-0, 51939-25-0 AND 0. REV CHANGE REF' PREPARER CHECKER APPROVED DATE SCALE: NTS EXCEPT AS NOTED PROJECT F",CIL ACIl ITV IlY POWERHOUSE G UNIT 1 TITLE TiTLE r ., ELECTRICAL rIOPEN I STOP CLOSE " I I LOGIC DIAGRAM L TYF' I TYP t.<<)V r.4Ov CONTROL.,' CONTROL J FEEDWATER SYSTEM "r--------l r--------T WATTS BAR NUCLEAR PLANT
£ l+
+--.- .-
--- -----::----"- "rT --((T TOO CONDENSER TENNESSEE VALLEY AUTHORITY Q
Q E~~r---~------""T----------~----------------------------------~~------J E~~~-~----r------'------------------'~--~ DESIGN INITIAL ISSUE ENGINEERING APPROVAL DRAFTER H Fev FeY CNC/SRH RO ISSUE PER 1 J.IoI.RUTHERFORD 1 J.I.4.RUTHERFORD 3-\ ~ WBEP WeEP 5.17 5. 17 ~ RI~S ol. RIMS 2 R.T.GRAHAIoI R. T .CR"H"t.t FROM LP , FROM "OIl LP' OPEN CLOSE 8Y: ISSUED BY: DATE 826 '90 00329 B26 0329 377 0377 J03 J.F. J.F.TORTORA/MCB TOATORA/MC8 R.M.JOHNSON FOR WSR R.t.C.JOHNSON 7-23-90 1-47W611-3-6 1-47WS11-3-6 R18 2 3 4 5 6 7 8 8R OR PitOJ AS Ct E£ I.E NE H HE SE TE BL 9f USO " ' 9FI 011 BlI ~ROJ .I.B OR PROJ AS C£ r. ME I'E CE EE IE ~E SE TE
~E ME T£ 8L BL Sf so WI! SO SF '11 CAD MAINTAINED DRAWING I (CONFIGURATION CONTROL DRAWING)
TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT EMERGENCY OPERATING INSTRUCTIONS FR-H.2 STEAM GENERATOR OVERPRESSURE Revision 5 Unit 1 ( QUALITY RELATED REQUESTED BY: S. M. Baker SPONSORING ORGANIZATION: OPERATIONS APPROVED BY: A. K. Keefer 11/22/06 EFFECTIVE DATE: _---..:.1....:..:...1/2=2::....:./0....:..:...6_
----'--'-'-~-"--
LEVEL OF USE: CONTINUOUS
WBN STEAM GENERATOR OVERPRESSURE FR-H.2 Rev 5 RevS 1.0 PURPOSE This Instruction provides actions for an overpressure condition affecting any steam generator where pressure has risen above the highest steamline safety valve setpoint. 2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 Indications S/G pressure greater than or equal to 1220 psig. 2.2 Transitions FR-O, Status Trees, FR-H in YELLOW condition. 3.0 OPERATOR ACTIONS ( ( 20f6 2of6
WBN GENERA TOR OVERPRESSURE STEAM GENERATOR FR-H.2 Rev 5 IStep I IAction/Expected Response I IResponse Not Obtained
- 1. IDENTIFY affected S/G(s):
- a. Any S/G pressure greater a. IF press in all S/Gs than or equal to 1220 psig.
pSig. less than 1220 psig, THEN RETURN TO Instruction in effect.
- 2. ENSURE MFW isolated to Manually CLOSE valves, and affected S/Gs: STOP pumps, as necessary.
- S/G MFW isolation and bypass IF valves can NOT be closed, isolation valves CLOSED.
THEN
- S/G MFW reg and bypass reg CLOSE #1 heater outlet valves.
valves CLOSED. MFP A and B TRIPPED. Standby MFP STOPPED. (
- Cond demin pumps TRIPPED.
Cond booster pumps TRIPPED.
- 3. CHECK affected S/Gs NR level ** GO TO FR-H.3, less than 93% [85% ADV]. STEAM GENERATOR HIGH LEVEL.
30f6
WBN STEAM GENERATOR OVERPRESSURE FR-H.2 Rev 5 IStep II Action/Expected Response II Response Not Obtained
- 4. DEPRESSURIZE affected S/Gs:
- OPEN S/G PORVs, OR
- OPEN S/G MSIV bypass valves, OR
- OPEN S/G steam supply to TD TO AFWpump, OR
- OPEN S/G blowdown valves.
- 5. ESTABLISH affected S/Gs press control:
- a. CHECK S/G press dropping. a. IF affected S/G press rising or
( stable, THEN
** GO TO Caution prior to Step 6.
- b. CHECK S/G press b. IF affected S/G press greater less than 1220 psig. than or equal eq ual to 1220 psig, THEN
** GO TO Step 3.
- c. MAINTAIN S/G press less than 1220 psig.
- d. RETURN to instruction in effect.
4of6
WBN STEAM GENERATOR OVERPRESSURE FR-H.2 Rev 5 IStep IIIIAction/Expected Response CAUTION AFW flow should remain isolated to affected SIGs UNTIL a steam release path is established.
- 6. ISOLATE AFW flow to affected S/Gs:
- a. CLOSE S/G MOMD AFW LCVs. a. IF LCVs will NOT close, THEN Locally CLOSE affected manual isolation valves.
- b. CLOSE S/G TOTD AFW LCVs. b. IF LCVs will NOT close, THEN Locally CLOSE affected manual isolation valves.
- c. CHECK S/Gs AFW flow zero gpm.
(
- 7. CHECK RCS T-hot less than 545°F. COOLDOWN RCS T -hot to less than 545°F by dumping steam from Intact S/Gs.
( 50f6
WBN STEAM GENERA TOR OVERPRESSURE FR-H.2 Rev 5 IStep I IAction/Expected Response II Response Not Obtained
- 8. CONTINUE attempt to manually or locally depressurize affected S/Gs:
OPEN S/G PORVs, OR
- OPEN S/G MSIV bypass valves, OR
- OPEN S/G steam supply to TO AFWpump, AFW pump, OR
- OPEN S/G blowdown valves.
- 9. RETURN TO Instruction in effect.
( - End-( 6of6
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 27. W/E15 EK2.2 027 Given the following:
- Unit 1 has experienced a large break LOCA.
- Containment Sump level has increased to 83% and is rising.
- RWST level has reached sump swapover setpoint.
Which ONE of the following identifies ... (1) a source of water, in addition to the expected sources, entering containment that could result in the containment sump level and trend, and (2) how the ECCS systems will be aligned in accordance with ES-1.3, "Transfer to Containment Sump?" A. (1) High Pressure Fire Protection (2) Only one train of RHR will be aligned to the containment sump. B~ (1) High Pressure Fire Protection B!'" (2) Both trains of RHR will be aligned to the containment sump. C. (1) Condensate Storage Tank (2) Only one train of RHR will be aligned to the containment sump. ( D. (1) Condensate Storage Tank (2) Both trains of RHR will be aligned to the containment sump. ( Page 73
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect" Water from the High Pressure Fire Protection System is not included in the expected sources of water being supplied to containment and after ES-1.3, Transfer to Containment Sump, is completed the Cold Leg Recirculation would not be limited to only one train in operation. Plausible because the unexpected water source is HPFP and other abnormal containment sump conditions (sump blockage) can result in only one train of ECCS being in service. Applicant could relate higher level to more potential for blockage. B. Correct, Water from the High Pressure Fire Protection System is not included in the expected sources of water being supplied to containment and after ES-1.3, Transfer to Containment Sump, is completed both trains of Cold Leg Recirculation would be in operation. C. Incorrect, The Condensate Storage Tank contents is not an unexpected source. The CST volume is included in the expected water sources and with the level at 83% and rising, the level is above the expected range. After ES-1.3, Transfer to Containment Sump, is completed the Cold Leg Recirculation would not be limited to only one train in operation. Plausible because the entire contents of the CST is not typically thought of as being put into containment (although the contents are accounted for in the ( calculation) and other abnormal containment sump conditions (sump blockage) can result in only one train of ECCS being in service. Applicant could relate higher level to more potential for blockage. D. Incorrect, The Condensate Storage Tank contents is not an unexpected source. The CST volume is included in the expected water sources and with the level at 83% and rising, the level is above the expected range. Both trains of Cold Leg Recirculation would be in operation. Plausible because the entire contents of the CST is not typically thought of as being put into containment (although the contents are accounted for in the calculation) and both trains of Cold Leg Recirculation being in operation is correct. Page 74
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 27 Tier: _1_ Group _2_ _1_ 2 KIA: W/E15 EK2.2 Containment Flooding Knowledge of the interrelations between the (Containment Flooding) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility. Importance Rating: 2.7 2.7/2.9 I 2.9 10 CFR Part 55: 41.7 I 45.7 10CFR55.43.b: Not applicable. KIA Match: Applicant must decide which of two sources of water being put into containment is from a source not accounted for in the design and the effect the higher containment level would have on the operation of ECCS alignments. ( . Technical
Reference:
FR-Z.2, "Containment Flooding," Rev. 6 WOG Background Documents, Rev 2 ES-1.3, "Transfer to the Containment Sump," Sum p," Rev 17 Proposed references None to be provided: Learning Objective: 3-0T -FRZOOOO 1 3-0T-FRZ00001
- 7. Identify all sources of water to the containment which might cause containment flooding.
Question Source: New Modified Bank x X Bank Question History: WBN bank questions FRZ0001.07 01111 modified Comments: Added second part to question. Relocated correct answer. Reworded stem and distractors. Page 75
- 2. DE SCRIPHOt~
DESCRIPTION Guideline Guide1 i ne FR .. Z.2. RESPONSE FR-Z.2:. RESP'OtiSE TO CONfAUtlErtT CONTP.INMEtH FlOOIHrtG FLOODING.~ provfprovicies des aetiactions ons to r"espond when the (;onta1t1111ent respond cOrlita] n~Tlent 1level eve) is greater than design flood leve1. level. Thi s level is significant sign1 cant since the critical systems s.ys.tems and components. which are necessar"Y to ensure an orderly necessary or"derly safe pI pl ant allt shutdown shLltdowr~ andalli1 provide feedback feedbact. to the operat or regarding operator regardi ng plant. plant oonditions. cOlldH'; OilS. are nomally rlormall y 1located oca ted above the design des; gn flood level. Therefore. the guideHne guideline FR-Z.2 is entered enter-ed from fr{8 the Contaillment theContairullent St atus Tree on an ORAnGE Status pri or! ty 'when ORf~t~GE priority when this tni s design flood fl cod level his exceeded. The primary purpose of the oontah'ul!ent contaillment sump area; area 15s to coliect call ectthe the 'Ylater" water injected into irato th.ethe containment or or' spi spt 11 ned ed fr{8 fr"Or1l the reactor coolant system foll accider;t. The owl fig an B.,ccident. follo'wir~g Tile water" water collected irI in the conta1 containment SUr1lp is then t1II1ent soop ava; lable for 10f'l9i available long termter"cn core con~ and/or conta'inent containment cooling via the emergency core coolitllg eooH cor;taimnent spr'ay f'I9i or containment r-ecirculation systems. spray reelreu1at10n systems.. In addition. the' the* containment sLimp coll ects the SL.Irlnp collects tbe;injected njeded or spill spilled ed water into areas ar-eas such sLlch that
',ri ta 1 systems or vital Or" ccmponents cropanetits will wi 11 not not. be flooded and thus rendered i noperabl e.
n~ndered inoperable. rnax.imUr1l 1level The rnaXhiLm evel of 'water containment water in the conta*j follo'wing ment fonO\..; ng a major acci accider-It dent genera 11y is based upon the entire ent i re wat water er contents cont ents. of the t he reactor react or cool ant arit ( system. refueling refueH ng water storage tank, comiensate storage tarl~:. condensate tank. and SI st.orage tanto accumul ators. Thi accullul Th; ss. water
'water vol ume ue approx lmates the maximum 'water approximates '.01 urne water volue i ntr"or:luced into introduced i nt 0 the containment contal rlment fo 11 ow;0'1'11 f'ng I9i a lOCA plus LOClI, p llls a ss. teaml; teamH ne or feedH feedl i ne ins i de CM ta i nment .
break lnslde.conta1nment. i ndi cated 'Water An indicated 1m water- le'"el level in the containent contai nment greater greater' thorl maxlrrll.ltfl expeded than the maximLm expected volume (design(des.;grl basis flood floor.J level) is anlt\dication an indication tr;at water volumes that 'Water volumes. other those represented than tl'lose repr-esented by the above. above noted vol Lmes Ur1les have been 1introduced ntroduced into the catHa i nmerH. Also. tnecontail'lf:llfnt. /I,ho. the h1gh h; gl1 water wat'T' 1level eve 1 provides. prov; des an 1i ndi ca, ti on that nd; cation potential potent h 1 flooding flood1 f'I9i of criti critical cal systems sys tells and components needed needer:l for plant recovery may occur. actiorss in The actions 1n ttl;U~1s s guideH guideli ne aHernpt attempt to identi identify fy any unexpected sour'ce source of wa ter- and isolate
'Water i so 1 ate it jt if poss; bl e. Beyond that the plant engineer; If possible. eng; nee-ri (191 staffs to. ff 1S is consulted to determine determi ne- if transfer transfer' of conta1nf:llfnt conta lmrie-rillt SUllpsump 'water
'Water to other tanks is appropri a.ppr'opr i ate.
FR-Z.2 gac~g~nund Dackground 2 HP-Re liP-Rev..... 2, 4/30/2005 HfRZ2B6.rln<: IIfRZZDG .doc
TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT EMERGENCY OPERATING INSTRUCTIONS FR-Z.2 CONTAINMENT FLOODING Revision 6 Unit 1 ( QUALITY RELATED REQUESTED BY: S. M. Baker SPONSORING ORGANIZATION: OPERATIONS APPROVED BY: A. K. Keefer _ _11/22/06 EFFECTIVE DATE: ----'-'--~- 1:.....;,,1;,.=/2=:.,2/..::....;06=---_ LEVEL OF USE: CONTINUOUS
( WBN CONTAINMENT FLOODING IFR-Z.2 Rev 6 1.0 PURPOSE This Instruction provides actions to respond to containment flooding. 2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 Indications Cntmt Sump level greater than 83% (14.4 ft). 2.2 Transitions FR-O, Status Trees, FR-Z in ORANGE condition. 3.0 OPERATOR ACTIONS ( 2of4
( WBN CONTAINMENT FLOODING I FR-Z.2 Rev 6 IStep I IAction/Expected Response . IIIIResponse Not Obtained
- 1. IDENTIFY and ISOLATE unexpected source of water:
- a. ERCW.
- b. CCS.
- c. High pressure fire fire protection.
- d. Primary water.
- e. 01 water.
- f. SFP cooling water.
- 2. CHECK cntmt sump activity and
( chemistry:
- a. CHECK RHR suction aligned from a. IF RHR suction is aligned to the RWST. cntmt sump, THEN:
- 1) NOTIFY Chemistry to sample from RHR system.
- 2) ** GO TO Step 3.
- b. NOTIFY Chemistry to sample cntmt sump.
3 of 4 30f4
WBN CONTAINMENT FLOODING IFR-Z.2 Rev 6 IStep I IAction/Expected Response I IResponse Not Obtained
- 3. NOTIFY TSC to evaluate the following:
- Stopping leakage into cntmt.
- Operational problems with equipment located below water level.
- Water transfer from cntmt sump to Aux Bldg.;e.g., Passive Sump, HUT, Waste Tanks, etc.
- Water transfer to RWST, PWST, etc.
- 4. RETURN TO Instruction in effect.
~End-
-End-(
( 4of4 40f4
TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT EMERGENCY OPERATING INSTRUCTIONS ES-1.3 TRANSFER TO CONTAINMENT SUMP Revision 17 Unit 1 ( QUALITY RELATED REQUESTED BY: S. M. Baker SPONSORING ORGANIZATION: OPERATIONS APPROVED BY: A. K. Keefer EFFECTIVE DATE: 11/5/07 LEVEL OF USE: CONTINUOUS
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 1.0 PURPOSE This Instruction provides actions for transferring ECCS and containment spray suction to the containment sump for recirculation mode. 2.0 SYMPTOMS AND ENTRY CONDITIONS 2.1 Indications RWST level below 34%. 2.2 Transitions A. E-1, Loss Of Reactor Or Secondary Coolant. B. ES-1.2, Post LOCA Cooldown. C. FR-C.1, Inadequate Core Cooling. D. FR-C.2, Saturated Core Cooling. E. FR-H.1, Loss Of Secondary Heat Sink. Sink. F. ECA-2.1, Uncontrolled Depressurization Of All Steam Generators. Steam.Generators. G. ECA-3.1, SGTR and LOCA - Subcooled Recovery. H. ECA-3.2, SGTR and LOCA - Saturated Recovery. 3.0 OPERATOR ACTIONS l( 2of22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 IStep IIIIAction/Expected Response I IResponse Not Obtained CAUTION
- ECCS flow to RCS must be maintained at all times to ensure adequate core cooling.
- Transfer to recirculation mode may cause high radiation in the Auxiliary Building.
NOTE
- Performance of this Instruction is a higher priority than performance of the FRs because it maintains ECCS pump suction.
- The transfer sequence should be performed without delay.
Implementation of FRs is delayed UNTIL transfer sequence is completed or transitioned from.
- 1. ENSURE both RHR pumps IF NO RHR pumps can be started, RUNNING. / THEN
** GOTOECA-1.1,LossofRHR GO TO ECA-1.1, Loss of RHR
( Sump Recirculation.
- 2. ESTABLISH CCS to RHR heat exchangers [M-27B]:
- a. ENSURE RHR heat exchanger outlet valves 1-FCV-70-153 and 1-FCV-70-156 OPEN.
- b. CLOSE SFP heat exchanger A CCS supply 0-FCV-70-197.
- c. ENSURE CCS flow to c. WHEN transfer to cntmt sump ESF supply header and greater completed, than 5000 gpm. THEN
- Train A: 1-FI-70-159 REFER TO Appendix B (ES-1.3),
CCS Operation, to adjust CCS
- Train B: 1-FI-70-165 flows as necessary.
- d. MONITOR level in CCS surge tanks.
30f22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 IStep IIIIAction/Expected Response II Response Not Obtained
- 3. CHECK RWST level DO NOT CONTINUE this Instruction less than 34%. UNTIL RWST level less than 34%.
- 4. CHECK cntmt sump level IF RWST level less than 34%
greater than or equal to 16.1 %. AND cntmt sump level less than or equal to 16.1 %, THEN
** GO TO ECA-1.1, Loss of RHR Recirculation .
Sump Recirculation.
.,5.
- 5. ENSURE automatic switchover complete:
- a. ENSURE cntmt sump valves IF at least one flow path from cntmt 1-FCV-63-72 and 1-FCV-63-73 sump to the RCS can NOT be
( OPEN. established or maintained, THEN
** GO TO ECA-1.1, Loss of RHR Sump Recirculation.
- b. ENSURE RWST to RHR suction valves 1-FCV-74-3 and 1-FCV-74-21 CLOSED.
- c. INITIATE power restoration to 1-FCV-63-1 USING Appendix A (ES-1.3), 1-FCV-63-1 Breaker Operation.
4of22 40f22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 IStep IIIIAction/Expected Response II Response Not Obtained CAUTION If RWST level drops to 8%, then any charging, SI or cntmt spray pump taking suction from the RWST must be stopped.
- 6. MONITOR RWST level STOP and PULL TO LOCK pumps greater than 8%. taking suction from the RWST.
- 7. DETERMINE if cntmt spray should be stopped:
- a. MONITOR cntmt press a. WHEN cntmt press less less than 2.0 psig. than 2.0 psig, ..
THEN PERFORM Substeps 7b thru e.
** GO TO Step 8, 8.
- b. CHECK at least one cntmt spray b. IF both spray pumps stopped, pump RUNNING. THEN
** GO TO Step 9.
- c. RESET cntmt spray signal.
- d. STOP cntmt spray pumps and PLACE in A-AUTO.
- e. CLOSE cntmt spray discharge 1-FCV-72~2 and valves 1-FCV-72-2 1-FCV-72-39.
- f. ** GO TO Step 9.
l 50f22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3
. Rev 17 IStep I IAction/Expected Response I IResponse Not Obtained
- 8. DETERMINE if ONE train cntmt spray should be stopped:
- a. CHECK BOTH trains cntmt spray a.
- 8. ** GO TO Step 9.
delivering flow.
- b. RESET cntmt spray signal.
- c. STOP ONE cntmt spray pump and PLACE in A-AUTO.
- d. CLOSE spray discharge valve for stopped pump.
- 9. IF RCS press rises to greater than 1350 psig, THEN STOP both SI pumps.
CAUTION If a valve fails during the transfer sequence, any corrective action should be postponed UNTIL transfer is complete, EXCEPT as required to satisfy each step. NOTE Each transfer sequence action is identified by a number on the control board (e.g. *#1).
#1).
- 10. (#1) ISOLATE SI pump miniflow: ENSURE either:
- CLOSE 1-FCV-63-3. a. 1-FCV-63-3 CLOSED,
- CLOSE 1-FCV-63-175. OR
- CLOSE 1-FCV-63-4. b. 1-FCV-63-4 and 1-FCV-63-175 CLOSED.
l 60f22 6of22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 IStep II Action/Expected Response IIIIResponse Not Obtained
- 11. (#2) ISOLATE RHR crossties: ENSURE either 1-FCV-74-33 or 1-FCV-74-35 CLOSED.
- CLOSE 1-FCV-74-33.
- CLOSE 1-FCV-74-35.
- 12. (#3) ALIGN charging pump and SI ENSURE either 1-FCV-63-6 or pump supply from RHR: 1-FCV-63-7 OPEN.
- OPEN 1-FCV-63-6.
- OPEN 1-FCV-63-7.
- ENSURE 1-FCV-63-177 OPEN.
NOTE 1-FCV-63-8 and 1-FCV-63-11 are interlocked with the SI pump miniflows being full closed. (
- 13. (#4) ALIGN RHR discharge to charging pump and SI pump suction:
- a. OPEN 1-FCV-63-8. a. ENSURE Train B RHR operation:
- B RHR pump RUNNING.
Train B
- 1-FCV-63-11 OPEN.
- Either 1-FCV-63-6 or 1-FCV-63-7 OPEN.
- b. OPEN 1-FCV-63-11. b. ENSURE Train A RHR operation:
- Train A RHR pump RUNNING.
- 1-FCV-63-8 OPEN.
- Either 1-FCV-63-6 or 1-FCV-63-7 OPEN.
7 of 22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 IStep IIIIAction/Expected Response I IResponse Not Obtained .
- 25. ENSURE cntmt spray pumps in A-AUTO.
CAUTION Continued ECCS or Containment Spray pump operation following loss of suction will result in pump damage. Loss of suction to RHR pumps will require stopping all ECCS. ECCS.pumps. pumps.
- 26. MONITOR for containment sump blockage. NO No;" t> \ t.A No b I '" t.A- 11 '0 f~
(~
- a. INTIATE Appendix D 0 (ES-1.3), of Su.~~
Su.IYly/ Monitoring for Containment Sump <3l0~\~-(.. Gl()L\L~<- Blockage.
- b. CHECK for indications of cavitation b. ** GO TO Note prior to Step 27.
ECCS or Containment Spray. on EGCS
- c. IF sump blockage results in loss of suction to ECCS pumps, THEN STOP CCPs, SI pumps and RHR pumps, PLACE in PULL TO LOCK, and NOTIFY TSC.
- d. IF sump blockage results in loss of suction to Cntmt Spray pumps, THEN STOP Cntmt Spray pumps, PLACE in PULL TO LOCK and NOTIFY TSC.
Cntm~ Spray flow
- e. IF ECCS or Cntmt lost due to sump blockage, THEN
** GO TO ECA-1.1, Loss of RHR Sump Recirculation.
12 of 22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 IStep IIIIAction/Expected Response III IResponse Not Obtained NOTE Time since initiation of event and guidance for transfer to hot leg recirculation is defined in E-1, LosS'of Loss of Reactor or Secondary Coolant. . 27.
- 27. VERIFY that this instruction was CONSULT CONSUL TSC for guidance on TTSC entered from E-1. when to transfer to hot leg recirculation (ES-1.4).
- 28. RETURN TO Instruction in effect.
- End-(
13 of 22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 APPENDIX D D (ES-1.3) Page 1 of6 of 6 Monitoring. for Containment Sump Blockage NOTE The following step records initial baseline readings after cold leg recirculation is established to aid in monitoring for sump blockage. This step should be repeated if parameters change due to operator action.
- 1. RECORD the following parameters: --------~/----------
Date Time INDICATOR READING CIRCLE ONE RCS Pressure (ICCM plasma display) 1-M-4 OR 1-M-6 Containment Sump Level (1-Ll-63-180 or 181) 1800R181 RHR Pump A amps (1-EI-74-5A) Pump B amps (1-EI-74-17A) ( Flow to CL 2 & 3 [1-FI-63-91A (NR) or 91 B (WR)] NRORWR Flow to CL 1 & 4 [1-FI-63-92A (NR) or 92B (WR)] NRORWR Pump A Discharge Press (1-PI-74-13) Pump B Discharge Press (1-PI-74-26) CCP Pump A amps (1-EI-62-108A) Pump A amps (1-EI-62-104A) Charging Flow (1-FI-62-93A) Charging Header Press (1-PI-62-92A) BIT flow (1-FI-63-170) Pump A flow (1-FI-63-151)V<~:~'!~~~!~,I1~ Pump A amp' (1 c EI-72-26A) I' ,;1i!i:~" £1 :s,;, Pump A flow (1-FI-72-34) .;~T~1~~:, r~ eM
~~----~----~---------------4----------~
Pump B amps (1-EI-72-12A)~.~Jf~ F~;! (I ;!' Pump B flow (1-FI-72-13) 17 of 22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 APPENDIX D (ES-1.3) Page 2 of 6 Monitoring for Containment Sump Blockage NOTE The following step records baseline readings after RHR spray has been initiated. If RHR spray is initiated the data recorded in Step 2 should be used in place of the RHR data previously recorded in Step 1 for monitoring for containment sump blockage.
- 2. IF RHR spray is placed in service, THEN RECORD the following parameters: ----------------~/-----------------
--------~/---------
Date Time INDICATOR READING CIRCLE ONE RHR Pump A amps (1-EI-74-5A) Pump B 8 amps (1-EI-74-17A) Pump flow to CL 2 & 3 NRORWR [1-FI-63-91A (NR) or 918 91 B (WR)] Pump flow to CL 1 & 4 NRORWR [1-FI-63-92A (NR) or 928 92B (WR)] Pump A Discharge Press (1-PI-74-13) Pump 8B Discharge Press (1-PI-74-26) 18 of 22
---~--~---- -----------~~- --~- ----_._--- -------- ~---.- ----~--------
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 APPENDIX D (ES-1.3) Page 3 of6of 6 Monitoring for Containment Sump Blockage NOTE TSC assistance should be requested, if necessary, in evaluating the effect of RCS pressure rises on ECCS flow.
- 3. MONITOR for containment sump blockage:
- a. CHECK for any of the following indications.
indications. ( ECCS pump flow!/ amps/discharge flow amps! discharge press RHR / CCP / SIP Erratic or gradually dropping (flow/amps/discharge press) (unexplained) Containment Spray flow! amps flow/amps CS Pump Erratic or gradually dropping
- b. IF any indications of sump blockage are observed, THEN NOTIFY TSC to evaluate indications.
19 of 22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 APPENDIX D (ES-1.3) Page4of6 Page 4 of 6 Monitoring for Containment Sump Blockage NOTE Containment sump level dropping indicates potential loss of sump inventory due to leakage or clogging of drain paths inside containment.
- 4. MONITOR containment sump level:
- a. IF containment sump level is dropping, THEN NOTIFY TSC to evaluate the following.
- 1) Indications of dropping containment sump level,
- 2) Need for RWST RWST refill.
l 20 of 22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 APPENDIX D (ES-1.3) Page 5 af6 of 6 Monitoring for Containment Sump Blockage NOTE Reducing Containment Spray flow and ECCS flow to single train operation may reduce the rate of debris accumulation on the sump screen and will reduce pressure drop across screens. Pumps should only be stopped in the following step if TSC approval has been obtained.
- 5. IF indications of sump blockage continue to worsen, THEN PERFORM the following:
(
- a. IF both trains of Containment Spray running, THEN EVALUATE stopping one train of Containment Spray.
- b. IF both trains of ECCS pumps running, THEN EVALUATE stopping ECCS pumps on one train.
- c. INITIATE makeup to RWST:
- REFER TO SOI-62.02, Boron Concentration Control.
- d. NOTIFY TSC to evaluate transferring water to RWST from the following, following,
- 1) Spent fuel pit
- 2) Holdup tank L
21 of 22
WBN TRANSFER TO CONTAINMENT SUMP ES-1.3 Rev 17 APPENDIX 0 (ES-1.3) Page 6 of 6 Monitoring for Containment Sump Blockage NOTE The following step records baseline readings after hot leg recirculation is established. If hot leg recirculation is initiated data recorded in Step 6 should be used in place of data previously recorded in Step 1.
- 6. WHEN ES-1.4, Hot Leg Recirculation COMPLETE, THEN RECORD the following parameters: --------~/----------
Date Time INDICATOR READING CIRCLE ONE RCS Pressure (ICCM plasma display) 1-M-4 OR 1-M-6 Containment Sump Level (1-Ll-63-180 or 181) 180 OR 181 RHR Pump A amps (1-EI-74-5A) Pump B amps (1-EI-74-17A) Flow to HL 1 & 3 (1-FI-63-173A) <* *~.~.*.. -c, Pump A Discharge Press (1-PI-74-13) Pump B Discharge Press (1-PI-74-26) CCP Pump A amps (1-EI-62-10SA) Pump A amps (1-EI-62-104A) Charging Flow (1-FI-62-93A) Charging Header Press (1-PI-62-92A) BIT flow (1-FI-63-170) SIP Pump A flow (1-FI-63-151) Pump B flow (1-FI-63-20) Pump A amps (1-EI-63-12A) Pump B amps (1-EI-63-16A) Pump A Discharge Press (1-PI-63-150) Pump B Discharge Press (1-PI-63-19) Containment Spray Pump A amps (1-EI-72-26A) Pump A flow (1-FI-72-34) Pump B amps (1-EI-72-12A) Pump B flow (1-FI-72-13) 22 of 22
QUESTIONS REPORT for ILT Bank Current wo pw FRZOOOl.07011 Given the following conditions: ( Unit has tripped and Safety Injected from 100% due to a Main Steamline Break. Crew has transitioned to E-2. . STA has determined an orange path for FR-Z.2 exists. STA Containment sump level is 85%. FR-Z.2, "Containment Flooding" requires the operator to identify the source of water causing the flooding condition. Which ONE of the following during this accident is an EXPECTED source of water (that is, not expected to cause excessive flooding)?
- a. Primary Water Storage Tank.
- b. Component Cooling Water Surge Tank.
c ..... Condensate Storage Tank.
- d. Essential Raw Cooling Water.
( Sunday, June 28, 200912:11:02 2009 12:11 :02 PM 2
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 28. 003 Al.09 028 Given the following plant conditions:
Plant is in Mode 5. A heatup is in progress. RCS pressure is 320 psig. RCS temperature is 190°F. As the pressurizer pressure is raised to 400 psig , which ONE of the following correctly describes ... (1) the response of #1 seal leakoff flow from the RCPs and (2) how the pressure indicated on the #1 seal differential pressure indicators on 1-M-5 will respond to the change in pressurizer pressure? (assume seal injection flow, VCT temp, and VCT press remain constant)
#1 seal leakoff flow Indicated #1 Seal Differential Pressure A'I rises rises
( B. rises remains the same C. drops rises D. drops remains the same Page 76
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Correct. Flow is proportional to the square root of the differential pressure. Actual sealleakoff flow will rise because the differential pressure between the RCS and the VCT rises for the conditions given in the stem of the question. Indicated differential pressure will rise during the pressurization of the RCS and will be indicated on 1-M-5 differential pressure indicator until the indicator reaches its upper limit of 400 psid. B. Incorrect. Plausible, since the applicant must recall that flow is proportional to the square root of the differential pressure. Actual sealleakoff flow will rise for the conditions given in the stem of the question due to increase in differential pressure but the indicated differential pressure will not remain the same. Plausible because the flow rising is correct and the indicated differential pressure would remain the same after the pressure was increased further because the No. 1 seal differential pressure indicator "pegs out" high at 400 psi. C. Incorrect. Actual flow will rise for the conditions given in the stem of the question, the opposite of the effect described in the distractor. Plausible if the effect is reversed and the indicated differential pressure rising is correct for the conditions .. D. Incorrect. Actual flow will rise for the conditions given in the stem of the question, the opposite of the effect described in the distractor. Plausible if the effect is ( reversed and the indicated differential pressure remaining the same would be correct if the pressure was increased further because the No. 1 seal differential
pegs out" high at 400 psi.
pressure indicator "pegs ( Page 77
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 Question Number: 28 Tier: _2_ Group _1_ 1
.KIA: 003 A 1.09 Ability to predict andlor monitor changes in parameters (to A1.09 prevent exceeding design limits) associated with operating the RCPS controls including: Seal flow and DIP.
Importance Rating: 2.8/2.8 10 CFR Part 55: 41.5 I 45.5 10CFR55.43.b: Not applicable. KIA Match: Question addresses both seal flow and differential pressure indication response during a plant heatup. Technical
Reference:
SOI-68.02, "Reactor Coolant Pumps," Rev. 33, Appendix B, No.1 Seal Normal Operating Range. 47W809-1, Rev. 56. Proposed references None to be provided: ( Learning Objective: 3-0T-SYS068B
- 5. Describe the RCPs' Seal Injection System, including
- a. Flowpath/Components,
- b. Flowrate,
- c. Purpose.;
- 6. List the RCP Seal #1 normal DIP and required minimum DIP.
Question Source: New Modified Bank X X Bank Question History: WBN bank question Comments: Changed stem of question to better match the KIA statement. Page 78
WBN Reactor Coolant Pumps 501-68.02 SOI-68.02 Unit 1 Rev. 0033 Page 35 of 39 Appendix B (Page 1 of 1) No.1 Seal Normal Operating Range 6 I I I I I I 55 v~
~
I I I
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I I I I I 44 ~ I
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~OO 600 800 1000 1200 1400 1600 1800 2000 220C, I
2400 3Gipllt 3~3piJ, I 2250 2500
#1 SEAL DIFFERENTIAL PRESSURE (PSI)
(Reactor Coolant System Press (-) VCT Press) 3AO PSi3" JAG psi~ . . Ir~~ psij pSij ~~ 303 303 1 (:)P 11JtSEJ 4OQ OD ('>~;~ -
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~:J~~,~~~DESICN BIoSlS sao EVENT IS DOCUl.lE:NTED 1M CAl.CUL~TION APPROVED DATE SCALE: NTS EXCEPT AS NOTED PROJECT FACILITY POWERHOUSE G UNIT 1 TITLE FLOW DIAGRAM CHEMICAL & VOLUME CONTROL SYSTEM WATTS BAR NUCLEAR PLANT TENNESSEE VALLEY AUTHORITY Q Q
INITIAL ISSUE ENGINEERING APPROVAL H RO ISSUE PER J.R .... ETH 1 J.R."ETH WBEP 5.17 WeEP a. RIMS 826 '90 0514 J78 378 2J.R.WETH 2 J.R.WETH
- a. DeN M-01S91-A 3 L.W.BOYD JL.".BO'l'O FSAR FIG.
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11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009
- 29. 004 K5.36 029 Given the following conditions:
- The plant is at 100% power.
TE-62-78A, Letdown Heat Exchanger Temperature, fails HIGH. Which ONE of the following describes (1) the impact on the plant and (2) what effect placing 1-M-6 controller 1-HIC-62-78A, LETDOWN HX OUTLET TEMP TCV-70-192 CNTL, to manual and adjusting in the closed direction would have on letdown temperature? (1 ) (2) Impact on the plant Letdown Temperature response A. Negative reactivity addition; Letdown temperature would rise because valve would throttle closed. B~ Positive reactivity addition; Letdown temperature would rise because valve would throttle closed. C. Negative reactivity addition; Letdown temperature will NOT change ( because the high temperature input prevents controller input to the valve control. D. Positive reactivity addition; Letdown temperature will NOT change because the high temperature input prevents controller input to the valve control. Page 79
WBN CVCS-Charging and Letdown SOI~62.01 SOI-62.01 Unit 1 Rev.00S7 ( Page 8 of 116 3.0 PRECAUTIONS AND LIMITATIONS NOTE During power changes, letdown should be maximized maoximized when possible toto reduce Crud Induced Power Shift, also know as, Axial Offset Anomaly. A. Maximum letdown flow is 120 gpm. B. If Letdown Heat Exchanger (LDHX) outlet temperature reaches 140°F, demineralizer resin damage could occur. C. During summer months, the lowest achievable letdown heat exchanger outlet temperature is limited by the CCS temperature, which in turn is limited by river temperature (via ERCW cooling to the CCS heat exchanger). The lowest achievable letdown heat exchanger outlet temperature is approximately 6 degrees higher than the river temperature. Attempting to operate with a letdown temperature less than 6 degrees higher than the river temperature will force 1-TCV-70-192 full open rendering 1-HIC-62-78A ineffective. Letdown ( temperature will then vary slightly as river temperature varies throughout the day. This causes slight changes in reactor power as boron affinity changes in the demin beds as the letdown temperature changes. Therefore, 1-HIC-62-78A, LETDOWN HX OUTLET TEMP TCV-70-192 CNTL should not be allowed to control 1-TCV-70-192 at its full open position. controI1-TCV-70-192 D. If Reactor Coolant (RC) filter is bypassed, flow through demins should be secured or diverted to the HUT to prevent resin entering RCS if the deminoresin demin resin screen fails. 22 E. Charging and Letdown are in service together. If Letdown isolates or Charging is lost, the other must be isolated (see exception below). If Charging is lost and Letdown remains in service, flashing or lifting of the letdown relief could occur. If Letdown is lost and Charging remains in service, thermal shock or a positive reactivity insertion could result from cold water entering RCS. EXCEPTION: If all the following conditions exist:
- 1. ° Any RCP Thermal Barrier Out Of Service
- 2. In service Charging Pump trips
- 3. RCP seal flow is required.
Then the Operator may immediately start an available charging pump. F. RCP seal damage can occur if VCT press is below 15 psig with RCPs running.
Cation exchange resins begin to undergo TEMPERATURE thermal decomposition at temperatures above 250°F by the reaction: Ion exchange resins are very sensitive to elevated temperatures. Although the inert resin bead structure is stable up to fairly high temperatures (300°F), the active exchange sites Equation 4-9 are not. The anion resin begins to decompose slowly at about 140°F and the decomposition This reaction destroys all exchange capacity of becomes rapid above 180°F. The cation resin is the cation bed and produces sulfuric acid. stable up to 250°F. Therefore, with purification systems that handle high temperature water, the The boron affinity of a resin bed is affected by influent to the demineralizers must be cooled to the temperature of the coolant passed through acceptable temperatures before it reaches the ion the bed. At lower temperatures, the borate ion exchange resins. bonding to the exchange site contains three boron atoms. At higher temperatures, the borate The anion resin (OK (OIr form) decomposes by ion contains only one boron atom. The result of either of two mechanisms with approximately this characteristic is. is that at lower temperatures equal probability as shown in Equation 4-7 and the resins are more efficient at removing boron Equation 4-8. from the coolant than at higher temperatures. A saturated resin bed will actually release boron as Alcohol Resin temperature is increased. RCH 2 N(CH 3 )3 + OH- ---; RCH 2 0H + N(CH 3 )3 Trimethylamine The temperature of a deborating demineralizer (TMA) increases 75°F. What effect will this have on demineralizer operation? Equation 4-7 RCH RCHzN(CH})} +ow ---;RCH 2 N(CH 3 )3 +OW ~RCH2N(CH})2 2 N(CH 3 )2 +CHjOH
+CHPH Amine Resin MethylAlcohol Equation 4-8 Examp/e4-5 Example 4-5 The alcohol form of the resin has no exchange In systems where it is possible to subject the ability. Trimethylamine (TMA) is a weak base. demineralizer resin to high temperatures, the The amine resin formed in the second reaction is demineralizers have automatic features to an inferior ion exchange resin with exchange protect against temperature damage. This is capabilities much less than the original form of usually accomplished by automatic closure of the resin. Thus, both reactions lead to partial or . the demineralizer inlet valves to isolate the complete loss of exchange ability and if the demineralizer from the high temperature liquid,*
liquid, temperature is sufficiently high, or if a lower when a high temperature at the inlet to the temperature (l80°F) (180°F) is sustained for a long demineralizer is sensed. These systems are period oftime, the resin will be unfitfor unfit for use. typically equipped with a bypass valve that can divert flow around the demineralizer until normal system temperature is restored. PWR/CONWONENTS PWR! COMPONENTS!I CHAPTER 4 140f18 14 of 18 © 2007 GENERAL PHYSICS CORPORATION
!/ DEMINERALIZERS AND ION EXCHANGERS REV 4 GF@gpworldwide.com GF@gpworidwide.com www.gpworldwide.com
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 DISTRACTOR ANAL YSIS: A. Incorrect. The failure described will result in a drop in letdown temperature. The drop in temperature will cause more boron to be removed by the demineralizers, resulting in a positive reactivity effect, the opposite of the effect given in the dis tra ctor. Plausible because the reactivity change or its effect could be reversed by distractor. the applicant and the letdown temperature increasing if 1-HIC-62-78 is placed to manual and adjusted to close 1-TCV-70-192 1- TCV-70-192 is correct. B. Correct, the high temperature input to 1- TC-62-78 will cause a signal to be sent to fully open 1- TCV-70-192, resulting in dropping letdown temperature. This drop in letdown temperature will result in an increased affinity for ion exchange in the de mineralizers, removing more boron. The relative drop in boron will result in a demineralizers, positive reactivity addition. To regain control of letdown temperature, the operator will take manual control of 1-HIC-62-78 and close 1- TCV-70-192 to return letdown temperature to normal. C. Incorrect. The failure described will result in a drop in letdown temperature. The drop in temperature will cause more boron to be removed by the demineralizers, resulting in a positive reactivity effect, the opposite of the effect given in the distractor. Plausible because the reactivity change or its effect could be reversed by the applicant and if the failure had been 1- TE-62-79 then control from 1-HIC-62-78 would have been prevented from controlling the valve. ( D. 1- TC-62-78 will cause a signal to be sent to Incorrect, the high temperature input to 1-TC-62-78 fully open 1- TCV-70-192, TC V-70-192, resulting in dropping letdown temperature. This drop in letdown temperature will result in an increased affinity for ion exchange in the demineralizers, removing more boron. Placing 1-HIC-62-78 in manual will allow control of the 1-TCV-70-192 to restore temperature. Plausible because the reactivity change is correct and if the failure had been 1- TE-62-79 then control from 1-HIC-62-78 would have been prevented from controlling the valve. Page 80
11/2009 Watts Bar RO NRC Exam - As submitted 10/2/2009 ( Question Number: 29 Tier: _2_ Group _1_ 1 KIA: 004 K5.36 Knowledge of the operational implications of the following concepts as they apply to the CVCS: Solubility of boron in water; temperature effect. Importance Rating: 2.5/2.8 2.5/2.B 10 CFR Part 55: 41.5/45.7 10CFR55.43.b: Not applicable KIA Match: Applicant is required to determine the effect the CVCS temperature change will have on the solubility of boron in water and how the change will affect reactivity as well as the affect manual operator actions taken would have on the condition. Technical
Reference:
SOI-62.01, "Chemical and Volume Control System," Rev. 57, Precaution C. 0-SI-67-901-A, "Essential Raw Cooling Water Pump A-A and Pump C-A Performance Test," Rev. ( 28, Precaution I. 2B, 1-47W611-62-1, Rev. 6 1-47W611-62-3, Rev. 9 1-47W611-70-2, Rev. 11 SSD-1-LPT-62-7B, Scaling and Setpoint Document, Rev SSD-1-LPT-62-78, OB. 08. Proposed references None to be provided: Learning Objective: 3-0T-AOI0300
- 4. List 3 ways Inadvertent Dilution could occur.
Question Source: New x Modified Bank Bank Question History: New question Comments: ( Page 81
WBN CVCs-Charging and Letdown 501-62.01 Unit 1 Rev. 0057 P~ge 8 of 116 Page 3.0 PRECAUTIONS AND LIMITATIONS NOTE During power changes, letdown should be maximized when possible to reduce Crud Induced Power Shift, also know as, Axial Offset Anomaly. A. Maximum letdown flow is 120 gpm. B. If Letdown Heat Exchanger (LDHX) outlet temperature reaches 140°F, demineralizer resin damage could occur. C. During summer months, the lowest achievable letdown heat exchanger outlet temperature is limited by the CCS temperature, which in turn is limited by river temperature (via ERCW cooling to the CCS heat exchanger). The lowest achievable letdown heat exchanger outlet temperature is approximately 6 degrees higher than the river temperature. Attempting to operate with a letdown temperature less than 6 degrees higher than the river temperature will force 1-TCV-70-192 full open open rendering 1-HIC-62-78A ineffective. Letdown temperature will then vary slightly as river temperature varies throughout the day. This causes slight changes in reactor power as boron affinity changes in the demin beds as the letdown temperature changes. Therefore, 1-HIC-62-78A, LETDOWN HX OUTLET TEMP TCV-70-192 CNTL should not be allowed to controI1-TCV-70-192 at its full open position. D. If Reactor Coolant (RC) filter is bypassed, flow through demins should be secured or diverted to the HUT to prevent resin entering RCS if the demin resin screen fails.2 E. Charging and Letdown are in service together. If Letdown isolates or Charging is lost, the other must be isolated (see exception below). If Charging is lost and Letdown remains in service, flashing or lifting of the letdown relief could occur. If Letdown is lost and Charging remains in service, thermal shock or a positive reactivity insertion could result from cold water entering RCS. EXCEPTION: If all the following conditions exist:
- 1. Any RCP Thermal Barrier Out Of Service
- 2. In service Charging Pump trips
- 3. RCP seal flow is required.
requited. Then the Operator may immediately start an available charging pump. F. RCP seal damage can occur ifVCT if VCT press is below 15 psig with RCPs running.
WBN Essential Raw Cooling Water 0-51-67 -901-A O-SI-67-901-A Unit 0 Pump A-A and Pump C-A Rev. 0028 Performance Test Page 10 of 78 3.0 PRECAUTIONS AND LIMITATIONS A. This SI is written with the Unit in normal operation. Due to the cross-tying of headers during this SI, this SI can not be performed during any accident operations. Performance of this SI during accident operations may cause Limiting Condition for Operation (LCO) 3.0.3 to be applicable due to header cross-ties during accident operations being a non-analyzed condition. Therefore, Operators are stationed at those valves to reposition them, if needed to prevent entering LCO 3.0.3. B. The ERCW header cross-tie will be required to be restored to normal in the event of an accident or emergency as determined by the Main Control Room Operator. Assigned Operators are dedicated in Step 4.4[4] to perform the emergency restoration actions for the duration of the cross-tie alignment. The individual restoration actions are contained in Section 6.4, Steps 6.4[2], 6.4[3], and 6.4[4]. Note that 1-FCV-67-22 and 1-ISV-67-505A must be opened BEFORE closing 1-FCV-67-147. C. For operation of a single pump at a flow rate above 13,000 gpm, the river level must be at EL 673.33 or higher. D. If the reading on 0-U-67-490, 0-Ll-67-490, INTAKE PUMP STATION RIVER LEVEL, is ( <10 psig, a M&TE pressure test gauge must be installed due to 0-Ll-67-490 0-U-67-490 reading in the lower third of its scale. E. Pumps that are operated at either shutoff or run-out conditions tend to vibrate and should not be run at either condition. F. Backwashing or flushing ERCW strainers during this SI will invalidate the data taken. G. 3.7.8, 3.8.1, and other LCOs as determined by the Safety Function LCOs 3.7.8,3.8.1, Determination Program (SFDP) will be applicable when this SI is performed in Modes 1, 2, 3, or 4. H. Component Cooling System (CCS) water temperature needs to be monitored during test. It is recommended that testing be suspended and another ERCW pump started if CCS temperature exceeds 95°F (110°F during unit shutdown). I. Performance of this instruction could could lead to a thermal power increase of 10 to 15 MW due to an increase in mixed bed demineralizer boron absorption efficiency (as a result of additional CVCS letdown cooling). Due to this increase, it is preferable to direct flow through CCS Heat Exchanger B rather than CCS Heat Exchanger A. J. Low flow rates may cause a timed overcurrent trip of the ERCW Pump being tested.
B~ p--.~ of 1 11 of1 1-I+lt-~2'lgll file:IIC:\Documents and Settings\NRC4Test\Desktop\WBN Sim Photos\lM-06Left\lM06-57.JPG 8/2812009 8/28/2009
WEN WBN SSD-l-LPT-62-78 PAGE 2 OF 35 REVISION 08 SCALING AND SETPOINT DOCUMENT LOOP DRAWING LOOP NAME: LETDOWN HTX OUTLET TEMP
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l-HIC 62-78 -_ OUT ..' o Reviewed by: D. Echols Approved by: J. S. Woods Date: 03/07/2005
WBN WEN SSD-1-LPT-62-78 PAGE 8 OF 35 REVISION 08 SCALING AND SETPOINT DOCUMENT PROCESS CONTROL VALVES AND CONTROL DEVICES Instrument No.: 1-TCV-70-192 Mfrr: Masoneilan Model: 37-20721 Vendor ID No: N/A Contract No: N/A Function: CV Valve Action: Air to Close Positioner Mfrr: Masoneilan Model No: 8012 Cam Type: N/A Calibrated Ranges: Input: 10-50 mAdc Document: MFRR Output: N/A Document: N/A Stroke: 100-0% Open Document: MIG Accuracy: +/- 5% Document: MIG Associated Interlock: N/A Associated Drawings: Manual ID WBN-VTM-M120-0010, WEN-VTM-M120-0010, 45N600-70 Remarks: (1) Masoneilan 8012 is a electropneumatic positioner. (2) DCN M-33221-A states that mechanical blocking device be adjusted so that 1-TCV-70-192 will fully close. DCN M-33221-A also added switch contacts from 1-TIS-62-79 that will ensure that 1-TCV-70-192 will fail open when setpoint is exceeded. - '_- S~fVA.~ ('. L
~vygRUAI)
( 0ir1M'l~A () h
):r1JlItl Il
<3 ~ 0 Function: Letdown Htx Outlet Temp
( Location: EL 737, COL A6-U Reviewed by: D. Echols Approved by: J. S. Woods Date: 03/07/2005
WBN SSD-1-LPT-62-78 PAGE 6 OF 35 REVISION 08 ( SCALING AND SETPOINT DOCUMENT CONTROL DEVICES Instrument No: 1-HIC-62-78C Mfrr: GE Model: 542 Vendor ID No: N/A Contract No: 92784 Function: Indicat Controller Action: Direct Calibrated Ranges: Input: N/A Document: N/A Indicator Output: 0-100% Document: MFRR Accuracv: +/- 5% Document: MIG Control Output: 10-50 mAdc Document: MFRR Accuracy: N/A Document: N/A Settings: Setpoint: 50% Document: (2) Proportional Band: N/A Document: N/A Reset: N/A Document: N/A Rate: N/A Document: N/A Other: N/A Document: N/A Matching Controller: N/A Load Required: 0-660 ohms Governing Equation: Output [mAdc] = 40
- Dial Settinq/100 + 10 +/- 2 Associated Drawings: 45W600-62, 45N1637-2 & 5, Manual ID WBN-VTM-B045-0010 Remarks: (1) Manual loading station, valve is closed with 100% output.
( (2) Controller setting per TI-57.002. Function: Letdown Htx Outlet Temp ( Location: 1-L-10 Reviewed by: D. Echols Approved by: J. S. Woods Date: 03/07/2005
WBN SSD-1-LPT-62-78 PAGE 4 OF 35 REVISION 08 SCALING AND SET POINT DOCUMENT SETPOINT SIGNAL PATH SHEET Function: Letdown Htx Outlet Temp Signal Path # 3 Signal Source: 1-TE-62-78 Signal Termination: 1-TI-62-78 Input: 50-150 Deg F Output: 50-150 Deg F Setpoint: N/A Transfer Function: N/A Tolerance: TE + TM + TI
+/- 0.5% +/- 0.5% +/- 1.5% = +/- 2.5%
Signal Path # 4 Signal Source: 1-TE-62-78 Signal Termination: 1-TCV-70-192 (Auto) Input: 50-150 Deg F Output: 0-100% Open ( Setpoint: 127 Deg F Transfer Function: N/A Tolerance: TE + TM + TC + HIC + TCV
+/- 0.5% +/- 0.5% +/- 0.5% +/- 0.5% +/- 5.0% = +/- 7%
Remarks: (1) 1-TCV-70-192 fails open when 1-TIS-62-79 is actuated. SL\fPOILtS. t3 {o I _ Reviewed by: D. Echols Approved by: J. S. Woods Date: 03/07/2005
Z-OL- L L9MLt- L Z-OL-119MLt-1 2 3 4 7 11 12 CI,.OSE CI.OSE RADIATION r----------j r----------, r----------j r----------, r----------j r----------, DETECTION DE:TECTION TO
- CONTROLS TYPICAL TYP ICAI. : : CONTROLS CONTROL_S TTPICAL TYPICAl. : : CONTROLS TYPICAL :
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l-FCV-70-66 1-FCV-70-66 I OF Of 1-FCV-70-2S 1-FCV-70-25 I 1-47"11611-70-'
'-47W6"-7D-' L C ~ TO UNIT 1 A COORD A-2 A
.T CCS PUMP 1,1, 1A INTAKE HEADER (1-47'1611-70-1)
(l-47WS1,-70-1) 70-2 r-____ __ __ r----::--.... TO INTAKE HEADER OF CCS
~~------,~I~--------'---~-----------------r~~----~-------------------------------------
~~------,~I~--------~--'------------------r~~--~~-------------------------------------
o o PUMP lA-A FOI!. UNIT I1 FOR.
<> <> (1-47W611-7D-l (1-47'111311-70-1 )
ccs CCS EXCHAM(;ER 14. HEAT EXCHANGER A B 8 B 8 o<> r---- -----j c : I CONTROLS TYPICAL or Of l-FCV-70-2S l-fCV-70-25
.J I
L _ _ _ _ _ _ _ _ _ ...l c C f~~,,~rf-70-1,-------fl""o,f----------I i~~lV~rf-70-1'-'------fl"'<>,I-----------; o<> r----- --- -- -, j lA 1A ENCINEERING ENGINEERING SAFEGUARD 0,-H~CLOSE
-H~
RADIATION DETECTION TO 1-FCV-70-66 1-47'11611-70-1
\-47'f1'6"-70-'
I c CONTROLS CONTROI.S TYPICAL : OF' l-FCV-70-2S or 1-FCV-70-2S I OPEN 0, CLOSE HEAT EXCHANCERS EXCHANGERS i ~~~fv!j~:~~L -- ~
) EXCEPT NO AUXILIARY I
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~J TO UNlT 1 CCS o
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NOTE J 1 l 8-11:*05 REVISED PER DCA 51918-05-0. 5191S-0S-0. PROJECT F FACILITY ACIL ITY POWERHOUSE G 70-194 70-"4 <> 0 UNITS 1 &. 2 TITLE TO FLOOR ELECTRICAL
" " ~J-L NOTES: ORAIN CRAIN
.. LOGIC DIAGRAM
- 1. FOR S'I'IoISOLS GENERAl. NOTES SEE SHEET 1.
SYMBOLS AND GENERAL
- 2. TYPICAL FOR:
...~ L L 1::m,~\~n~,~NIT COMPONENT COOLING SYSTEM
~ 1::m,~i~%~,~'IT L
l-FS-70-155 1-FS-70-1S5 (27C-246D) (27C-2460) AND l-FCV-7D-15J-B l-FCV-70-153-8 ...... iCI TO TO CCS
'OS PUMP PUMP >A-A
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~:~~:~g:l~i gjg:~::~j l:~~:jg:l~~ ~~~g:~::~l ecs SPENT FUEL PIT WATTS BAR NUCLEAR PLANT J. TYPICAL FOR fOR VALVE l-FCII-70-156-" {27C-245D) AND FLOW SWITCH l-F5-70-158127C-2450) 1-FeV-70-ISS-A (27C-2450)
CCS HEAT EXCHANGER B HEAT EXCHANGER B TENNESSEE VALLEY AUTHORITY Q Q
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l-FCI/-70-168-A 1270-256.11) AND FLOW FLDI'j fLOW SWITCH 1\-1"5-70-170
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27C-24JA) ,V.~,l,VR'FS'C',"p'ORE,.~,CS!N.
- 7. ,V,A,L,V,',SA"P'O''S. G'H',Hy'SH,NV',TE G'H',Hy'SHA'VO,TEB,A,'.~ C"l"",',",""HH"C',CN',lROl
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O-FCV-70*194-B (27C-252E) "NO 0-FCV-70-194-8 SWITCH 2-1"5-70-170 AND FLOW SWITCH 0-1"5-70-6 (27C-252E) 270-25614.) 270-2561\) (27C-252E). AND OVERLOAD FUSES REJ.ClVED BETWEEN UNIT 1 AND UNIT 2. S .. OS REMOVED TO PRECLUDE ADVERSE INTERACTION
- ; - - - - Hn~\~M~~N~H
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~n~TE~~~~N~l~ A ------+----a- iWT~~~ ~~~6d\j~IT
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H~,~4~oo~'~'~'~EL~TE~O~~?v:~~~~V:vj~g:_1~~l~:~~I~~vj~~:~_l~~6~:v:_~8~~~t~~v~_:vH~g~~~?:~~~~:~j~g:~ij2vl~~7~7~~:~~l~§6~: (1-47'196',-70-1) (1-47W6l 1-70-' ) DRAF'TER oRAF'TER RO ISSUE PER 1 CARY BARNARD
- s. VALVES O-FCV-70-12. -22, 1-I"CII-70-2.
5. ADMINISTRATIVEl'" l-FCV-70-2. -3, -4. -8, -25. AND 2-FC"'-70-2 ADI.1INISTRATIVELY LOCKED IN THE OPEN POSITION (WITH BREAKER 2-FCV-70-2 AND -4 ARE BREMER OPEN)(A??F;NDIX OPENj{APPENDIX R). 1-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ iP_~~~J,f-¥g~Jl'TENT MIKE 8ALES
---------------- EAI-J. 10 .. RIMS NO.
EAI-3.10" T28 920121 805 DeN M-OS454-A}}