IR 05000277/1999302
| ML20217F701 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 10/13/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20217F690 | List: |
| References | |
| 50-277-99-302OL, 50-278-99-302OL, NUDOCS 9910210045 | |
| Download: ML20217F701 (200) | |
Text
~
L U. S, NUCLEAR REGULATORY COMMISSION l
REGION I
Docket Nos: 50-277,50-278
. Report Nos: 50-277/99-302,50-278/99-302 (OL) l License Nos: DPR-44 r,d DRP-56 Licensee: PECO Nuclear Facility: Peach Bottom Atomic Power Station j
Location: Delta, Pennsylvania i Dates: September 13-16,1999 (Administration)
September 20-24,1999 (Grading)_
Chief Examiner: J. Williams, Senior Operations Engineer / Examiner, Region l Examiners: C. Sisco, Operations Engineer / Examiner, Region i S. Dennis, Operations Engineer / Examiner, Region I Approved by: Richard J. Conte, Chief Human Performance and Emergency Preparedness Branch Divit. ion of Reactor Safety 9910210045 991013 PDR ADOCK 0S000277 t
V PDR L
EXECUTIVE SUMMARY Peach Bottom Atomic Power Station inspection Report Nos. 50-277 & 278 /99-302 (OL)
Operat!qog
.. Two reactor operator (RO) applicants and three senior reactor operator instant (SROI)
applicants were administered exams. All applicants passed all portions of the exa . Overall, the applicants were well prepared for the exa . There were no por t-exam facility comment .
The facility used an examination preparation team of experienced training department staff who assisted the NRC examiners in an excellent manner,
. There were several times during the exam when the simulator did not perform as expected. These lastances perturbed, but did not invalidate the examination . Examination security was well maintained during the week of the exa .
The responses to two violations associated with the licensed operator requalification training (LORT) program were reviewed by the examiners and found to be acceptabl The two open jtems were close !
i l
I i
!
L l
l
,ll-i
REPORT DETAILS 1. Operations 05 Operator Training and Qualifications 05.1 Reactor Operator (RO) and Senior Reactor Operator Instant (SROI) Initial Examinations Scope The NRC examiners reviewed the written and operating initial examinations prepared by the facility in accordance with the guidelines of the "Exami'1ation Standards for Power i Reactors,"(NUREG-1021, Revision 8). The review was conducted both in the Region I )
office and at the Peach Bottom facility. Un September 14-16,1999, the NRC examiners l administered the operating portion of the exam to all applicants. On September 13, 1999, the written examinations were administered by the facility, i Qbservations and Findinos Gradina and Results The results of the examinations are summarized below:
SROI Pass EaB EQ Pass Eai!
Written 3 0 2 0 l Operating 3 0 2 0 l Overall 3 0 2 0 Examination Preoaration and Quality
The written exams, job performance measures (JPMs) and simulator scenarios were developed by the facility using the guidelines of the examiner standards. The exam as <
submitted to the NRC met the guidelines of the examiner standards. There were no unacceptable test items. Enhancements were made to 12 of 126 written exam questions .
to improve clarity. There were no post-exam facility comments. All individuals with knowledge of the exam signed a security agreemen l The facility used an examination preparation team of experienced training department staff who assisted the NRC examiners in an excellent manne Written Examination Performance All applicants passed the written exam. The facility training department staff performed an analysis of the ten written exam questions missed by at least half the applicants for j generic weaknesses. The training department staff subsequently re-verified that the i questions met the guidance of NUREG 1021, and were technically accurate. The questions were reviewed with the applicants and will be reviewed for training program improvement. These actions were determined to be acceptabl !
I
2 L ;
Ooeratina Test Administration and Performance l Simulator and job performance measure (JPM) actions, by the applicants, were acceptabl During the operating test, there were several instances where the simulator did not respond as expected. Specific examples are noted in Attachment 1. These examples of simulator response perturbed, but did not invalidate the examination. The problems l
appeared to have resulted from a very recent simulator shutdown and startup and were not modeling discrepancies between the plant and simulator, They were not observed to be present during the exam validation proces l l The facility took the necessary precautions to maintain examination integrity during the
'
administration of the exa Conclusions l
Two reactor operator (RO) applicants and three senior reactor operator instant (SROI) !
l applicants were administered exams. All applicants passed all portions of the exa J Overall, the applicants were well prepared for tne exa There were no post-exam facility comment The facility used an examination preparation team of experienced training department
!
staff who assisted the NRC examiners in an excellent manne l
There were several times during the exam when the simulator did not perform as expected. These instances perturbed, but did not invalidate the examination Examination security was well maintained during the week of the exa Miscellaneous Operations issues 08.2 Open items insoected CLOSED NIO 50-277 & 278/98-04-02) The annual operating exams did not follow licensed operator requalification training (LORT) program procedures for JPM difference requirements from week to week. The NRC inspectors reviewed and verified the I
corrective actions described in PECO Energy Company letter dated June 17,1998, in response to the violation. The inspector also reviewed the annual operating tests given February and March 1999 to verify tha'. JPM difference guidelines were satisfied. Based upon these reviews, this item is closed.
i
.. - .- -
CLOSED (VIO 50-277 & 278/98-04-03) The annual operating test did not sample items required by 10 CFR55.45 dealing with executing the emergency plan. The NRC inspector reviewed and verified the corrective actions described in PECO Energy Company letter dated June 17,1998, in response to the violation. Based upon this review, this item is close ;
,
V. Manaaement Meetinas l l '
XI Exit Meeting Summary On September 22,1999, NRC observations regarding the examination were dWcussed with members of the facility staff. The NRC expressed appreciation for the coopeidaon and assistance that was provided during both the preparation and examination week by licensed operator training personnel.
l l
l l
LIST OF ITEMS OPENED AND/OR CLOSED Closed NUMBER TYPE DESCRIPTION 50-277 & 278/98-04-02 VIO JPM different,e requirements on annual operating tests were not in accordance with LORT program procedure & 278/98-04-03 VIO Sampling items in 10 CFR50.45 on the annual operating tes
Attachments:
1. Facility Simulation Report 2. RO Written Examination with Answer Key 3. SRO Written Examination with Answer Key I
i
I i
s
.
!
Attachment 1 Facility Simulation Report Facility Licensee: Peach Bottom Atomic Power Station Facility Docket Nos.: 50-277 and 50-278 Operating Test Administered on: September 14-16.1999
.
This form is to be used only to report observations. These observations do not constitute audit or inspection findings and, without further verification and review, are not indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or 4 approval of the simulation facility other than to provide information that may be used in future evaluations. No licensee action is required in response to these observation While conducting the simulator portion of the operating tests, examiners observed the following l items: i
. The control rod position indication on the full core display showed one rod at minus eight when scrams o
REGION OF THE RPV
$ AfuR ATION CURVE AS DETERWINED IT THE LOCAL INSPECTION OR TR=2(3bl3-139 PT !!
g AN IN5IRUWENT EXHIIIf! AN UNEXPLAINEO TREND
. . - OR 0$CILLAi!0 LMig IMAT INSTRUWENT !$ UNAVAIL ASLE WIDE RANCE AND FUEL ZONE INSTS ONLT FOR E ACH OFRETHE IME INSTRUWENT A03 INSTRUWENTS IN THE T ASLE. f!ON THE WIN Ino!CA Lf VEL M INE TEWP ME A RX BLOS REFERENCE 400 VERICAL RUNS ARE THE WAI RUN TEMP As DETERWINED BT INSPECTION OR TR-I( 3kl3-139 PT !!.
INSTRUWEN! WIN INDICATED g WAX RUN TEWP IS DELOW LEVEL 13 A80VE WIDE RANCE -120 I T FUEle ZONE -305 I g SI E*F
. .
SRO - 70
~
_
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
.
ID Number: 99
-
71 ) A Designated Altemate (DA)is moving an old jet pump in the Unit 2 fuel
' pool when it falls off the auxiliary hoist. It is reported to the Control Room that a jet pump fell on an irradiated fuel bundle and damaged some fuel pin The Control Room also receives the following alarms and indications i
- Refueling Floor Vent Exhaust Hi Radiation (218 A-1)
- Reac. Bldg. Zone Vent Exhaust Hi Radiation (218 B-1)
- Reac. Bldg. Or Refueling Floor Vent Exh. Hi Rad Trip (218 D-4)
- Refueling Floor Radiation Trip Units A and D High lights are li Evaluate these conditions and determine the expected ventilation lireu A) Reactor Building Ventilation trip Refuel Floor Ventilation trip SBGT initiates and aligns to the entire Reactor Building / Refuel Floo B) Reactor Building Ventilation continues to ru Refuel Floor Ventilation trip SBGT initiates and aligns to the Refuel Floo C) Reactor Building Ventilation continues to ru Refuel Floor Ventilation continues to ru SBGT initiates and aligns to the Refuel Floo D) Reactor Building Ventilation continues to ru Refuel Floor Ventilation continues to ru SBGT remains in standb Page 71 of 100
.. . .
. _ _ _ - - _ _ - _
l l
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99 ID Number: 100 72 ) Unit 2 is in T-103, " Secondary Containment Control", due to high water level condition in Secondary Containment. The Reactor has been conservatively scrammed and the Group 11/111 isolations (from the level shrink) are complet The CRS is cun..;ntly attempting to determine whether a Primary System is discharging into the Reactor Building. Given the above conditions, evaluate the following and determine which constitutes a primary system discharging into the Reactor Buildin A) Leakage from a pipe flange on the discharge of the Reactor Water Cleanup Non-regenerative Heat Exchange B) Steam leakage from a rupture on the piping of the #2 Main Steam stop valve inle C) Leakage from a weld crack on the "A" RHR suction piping penetration to the Toru D) Steam leakage from the Standby Liquid Control Injection line just outboard of the drywell penetratio I l
l l
Page 72 of 100 l
_
_
$
Test ID: 1999 SRO- Exam Level: SRO Date: 9/13/99
'
ID Number: 101 73.) Unit 3 was operating in MODE 1 at 75% power when a fire was reported in the Reactor Building 135' elevation. The Crew has entered ON-114, the procedure for an " actual fire", and the CRS has directed that the Equipment Operator isolate the RPV Condensing Chamber Backfill Syste The basis for isolation of this system under these conditions is to prevent inaccurate level indication and unreliable automatic initiations due to:
A) Lowering Instrumentation Variable Leg densit B) Raising Instrumentation Variable Leg density, C) Lowering Instrumentation Reference Leg densit D) Rising Instrumentation Reference Leg density.
!
!
I l
i Page 73 of 100
~
Test ID! 1999 SRO- Exam Level: SRO Date: 9/13/99
ID Number 102 74 ) ON-114, for an Actual Reported Fire, has a note to inform the Operator that a loss of power to the Motor Driven Fire Pump for more than 8 seconds will defeat that pumps automatic start capabilit The basis for this feature is to prevent:
' A) A simultaneous start with the Diesel Driven Fire Pump and resultant water hamme B) A spurious start due to loss of power to the fire header pressure instrumentatio C) The pump from automatically starting with reduced bus voltag D) Overloading the diesel generators on a loss of off-site powe i l
l
,
..
Page 74 of 100
-Test ID:. 1999 SRO Exam Level: SRO Date: 9/13/99
,
,
' ID Number: 103 l
75) Given the following conditions:
- A loss of off-site power has occurre The E-1 and E-4 Diesel Generators (DG) are runnin The E-43 4KV bus has an overcurrent lockou No DG cooling water is availabl Drywell pressure is 3.6 psig and slowly risin Why are jumpers, installed in the Control Room, the PREFERRED -
method for shutting down the two diesel generators?
A) This bypasses the 10 minute timer on the MCA signal enabling the DG Control Switch " Pull-To-Lock" positio B) . Local methods of DG shutdown are disabled for these condition C) The DG shutdown actions need to be completed as quickly as possible, D) Use of the DG Control Switch " Pull-To-Lock" position will not allow a restart should cooling be restored.
1
!
,
-l l
!
Page 75 of 100
F'
,
Test ID: 1999 SRO Exam' Level: SRO Date: 9/13/99 l *
l ID Number: 104 l '76) Unit 2'was operating in MODE 1 at'40% power when it experienced a
!
i loss of 20Y050. All required control room actions have been completed.
l-Under these conditions, operator actions will be impscted by a loss of
- power to
A) The RBCCW backup of DWCW which will require manual ta isfe B) The lighting in vital areas which will require the use of flashlights.
l C) The Fire Alarm Panel which will require continuous roving fire ,
I
! watches.
l D) The Control Room radios which will require the use of alternate communication l l
I
Page 76 of 100
!
!
r.-
[ +
t l
y Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
!
l lD Number: 105 77 ) Following a reactor scram, the Unit Reactor Operator reported that all
!
APRMs are downscale. Later, the Control Room Supervisor (CRS)
l
'
directed all control rods be verified to be inserted to or beyond Notch
,,02 ".
Which of the following describes why the CRS needs this information?
The CRS:
A) Will direct boron injection (Standby Liquid Control) it this is not tru ) Is assured the reactor is shutdown and will remain shutdown during the ensuing cooldow C) Will exit T-101, "RPV Control" and enter T-117, " Level / Power Control", if this is not tru D) is assured the Heat Capacity Temperature Limit will not be exceeded, l
!
l
!
[
I Paget 77 of 100 L.
!1
!
!
- Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
'
ID Numbe ) Unit 2 is operating in MODE 1 at 100% power when the following occurs:
i
"
- REACTOR HI PRESS" alarm 210 G-2 annunciate Reactor Pressure indicates 1075 psig and rising slowly, i
In accordan'ce with OT-102 " Reactor High Pressure" which of the following is an appropriate immediate operator action?
A) Control reactor pressure by raising the Bypass Jack setting.
i B) Control reactor pressure by lowering the Max Combined Flow Limit Po C) Control reactor pressure by lowering reactor power.
l D) Control reactor pressure by raising the Max Combined Flow Limit i Pot.
I l
I l
l l
L l
l
!
I
!
i
!
i Page 78 of 100 l
-
l
'
.
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
'
I ID Numbe ) Unit 2 is operating at 87% power when the "A" Condensate pump shaft coupling shears. The Condensate pump continues to run at low motor amp Given that all three Reactor Feedpumps (RFPs) remain in service and no Operator action is taken, what is the expected plant response to this event?
A) A Recirc runback to 45% speed will occur immediatel B) A Recirc runback to 45% speed will occur when levelis less than
+17".
C) A Recire runback to 30% speed will occur immediatel D) A Reactor scram will occur when level is less than +1".
l-l .
l Page 79 cf ivo
,
!
l I i
- !
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99 i
ID Numoer: 108 80) Unit 2 is at 100% power when Drywell pressure begins to ris In accordance with OT-101 "HIGH DRYWELL PRESSURE" follow up
!
actions the following parameters and alarms are note i
"
- A RECIRC PUMP SEAL STAGE 2 HI FLOW' alarm 214 A-1
- PI-2-02-2-033A " Seal 1 Inner" 1056 psig
- PI-2-02-2-032A " Seal 2 Outer" 1043 psig Evaluate these indications, using the attached drawing, and select the appropriate statement belo A) The 1st stage seal has failed but it is NOT the source of high j drywell pressur i B) The 2nd stage seal has failed but it is NOT the source of high )
drywell pressur l C) The 1st stage seal has failed and is the snurce of high drywell 1 pressur D) The 2nd stage ses! has failed and is the source of high drywell pressur l l
l l
l l
l
!
!
l
!
Page 80 of 100 j i
!
!
,
i ,
_
_
-
_
o
_
.
-
P
.
v
-
I e
.
.
" U '" R Q^
-
x/
.
.
-
-
_
E G
tea O
T o -
_
_
-
-
-
L A A
ES A
L G -
N I
.i i *!hs P I
P
3 L g
P A G
N a
8 ]% A E I
G W ' > ) 2 S A
TO P3 -
SL L
A F ;s ' l13 !5:
M0
$
. E S
U30
.
t P0 -
O i NT OLO i
l T
.
' N
,I V^ E V
Y L
P M I
TT
-
P E
.'.
,I I
US T A SY L I' , f X n ES U P GD C
N x W RR I
4UI A QR E
Uc P R I
D O C T E A
<I i A
R i
3 i p 3 P 2 3 4 3 b9%< > a
7 WA N
I RR CD
- yLJ P PH
' DM C S RUI yL; a cPD
-
O2O g o
i
'
.
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
'
10 Number: 109 81 ) Unit 3 was operating at 70% power when it experienced a rising drywell pressure. Using OT-101, High Drywell Pressure, the source of the leak has been determined to be the "A" Recirculation pump seals. The CRS has directed you to trip and isolate the "A" Recirculation pum Given these conditions, what is the proper sequence for isolating the i recirculation pump and why?
A) Shut the suction valve first since it can close against a higher d B) Shut the discharge valve first since it can close against a higher d C) Shut the suction valve first since it is limited to closing against a l lower d D) Shut the discharge valve first since it is limited to closing against a lower dP.
I i
l
,
Page 81 of 100
-
.,
'
Test ID: 1999 S'RO Exam Level: SRO Date: 9/13/99
'
ID Number: 110
~
82) Unit 2 is at 100% powert Which of the following events would require power to be reduced or maintained in accordance with OT-104, " Positive Reactivity insertion"?
A) ' "A" Reactor Feedpump min flow valve fails ope B) EHC pressure set setpoint drops 10 ps C) Condensate pump tri D) Loss of RBCCW to RWCU Non-regen Heat Exchange ,
.i l
l
,
i I
i
<
.
Page 82 of 100
,
,-
.
.
' Test ID:
-
1999 SRO - . Exam Level: SRO Date: 9/13/99
-
ID Number: 111 83) The following conditions. exist following the receipt of an automatic scram signal:
- Reactor power: < 1,00 E 0%
- RPV pressure: 950 psig AND dropping
- RPV level: +25 inches AND steady
- Drywell pressure: .5 psig AND steady
- Scram Air Header pressure: 0 psig
- Control Rod 34-27 is at position 48
- All other Control Rods are fully inserte '
- Boron has NOT been injected to the RPV Which one of the following procedures will provide direction for the successfulinsertion of Control Rod 34-27 in this situation?
A) GP-3, " Normal Plant Shutdown" B) T-100, " Scram" C) T-101, "RPV Control" D) T-117, " Level / Power Control" l
l Page 83 of 100
.
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99 j
ID Number: 112 1
84 ) Which of the following is tre reason why the Main Steam Isolation Valves (MSIV) are closed prior to evacuating the Main Control Room in accordance with SE-1," Plant Shutdown from the Remote Shutdown Panel"? !
I A) With MSIVs closed, all reactor inventory and pressure control may take place at the Remote Shutdown Pane B) Since plant release points cannot be monitored at the Remote Shutdown Panel, closing the MSIVs precludes any concern for off-site release C)_ The MSIV closure outside the Main Control Room requires I access to plant areas that may not be accessible during an evacuatio D) If the MSIVs are closed from outside the Main Control Room, there is no method for verification of complete closur )
- . 1 i
i l'
l
l Page 84 of 100
!
I I
.
~
Test'ID: 1999 SRO Exam Level: SRO Date: 9/13/99 l
ID Number: 113 85) Unit 2 Reactor Operator-is controlling reactor level using HPCI at the Unit 2 Alternate Shutdown Panel following Control Room Abandonment. Indicated reactor level on LI-2-2-3-112 is currently 20" and reactor pressure is 500 psig. Using SE-10 Attachment 9, provided,
, determine the current reactor level and the expected HPCI response if I
an actual high level condition occur A) Actual level is > 40", HPCI will automatically trip on high level condition.
'
B) Actuallevelis > 40", HPCI must be manually tripped on high level I conditio ]
C) Actual level is between 0" and 40", HPCI will automatically trip on a high level conditio ,
D) Actual level is between 0" and 40", HPCI must be manually l tripped on a high level conditio l
!
l j
l l
Page 85 of 100 l
l
l SE-10 ATTACHMENT 9 Rev. O Page 2 of 2 SE-10 Attachment 9 Figure 1 SE-10 Attachment 9 Figure 1 ACTUAL RX LEVEL AS A RX PRESSURE FUNCTION OF RX PRESS AND l PSIG INDICATED LEVEL
'1100 - ^
""
i l
1000 -
I
- J-- ACT U AL L E V E L + 4 0*
900 -
?
C ACTU AL LEVEL O'
8 -
+
5"^' " ' ' "
0- ACTU AL LEVEL i
iS >0' AN D 44 O'
.
!
600 $
ACTU AL LEVEL e4 0'
500 -
400- W .
'
300 - ?
. . _
.;... .;
200 ACTU AL LEVE L 'O' '--
' :M:-
'
N .
. .. .. a.:. . .
- . . , ,:
'
N-30 -20 -10 0 1b 20 30 4O 50 60 INDICATED RX WATER LEVEL LI 2(3)- 2 - 3 -112
. . SRO - 85
!
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
'
ID Number: 114 86) Unit 2 was operating at full power in MODE 1 when a positive reactivity event occurred due to a control rod drifting out. The CRS has directed
. you to monitor for evidence of fuel damag ;
)
Which of the following indications would be the first indication of a small J fuel pin leak from this transient?
- A) Main Steam Line Radiation Recorder B) Air Ejector Discharge Log Monitor Recorder C) Off-Gas Adsorber Outlet Radiation Indicatio D) Main Stack Gas Recorder.
I Page 86 of 100 f..
I L
iTest ID: 1999 SRO Exam Level:' SRO Date: 9/13/99
"
ID Number: 115 87) Given the following cond'itions:
- Unit 2 has experienced a loss of coolant accident with confirmed fuel failure Drywell and Torus pressures reached 29 psig and sprays were initiate Sprays were NOT manually secured when pressure reached 2.0 psi Sprays did NOT automatically isolate at 1 psi Which of the following is the expected impact on the plant for these conditions?
A) The drywell oxygen concentration may ris ) Torus water level indication will be unavailabl C) The running Residual Heat Removal Pumps may cavitate.
'
D) Failure of the Reactor Building - Torus Vacuum Breakers will make the Reactor Building a High Radiation Area.
l l
l
!-
l
l Page 87 of 100
FT l Test ID:' 1999 SRO ' Exam Level: SRO Date: 9/13/99 i
- ID Number: 116 88) Following a LOCA on Unit 2 the CRS directs restoration of Drywell Cooling, using T-223, "Drywell Cooler Fan Bypass" for Drywell pressure control. The Unit Reactor Operator reports that the Drywell Cooler fans cannot be placed inservice without an engineering evaluation due to plant conditions falling on the UNSAFE side of T-223 Figure 1, "Drywell Chilled Water (DWCW) Saturation Curv Which of the following describes the basis for restricting Drywell Fan restoration when on the UNSAFE side of the curve?
A) Water hammer and rupture of piping inboard of DWCW lsolation valves when flow is restore B) Inadvertent lifting of overpressure relief valves inboard of the DWCW lsolation valves when flow is restore l
'
C) Overcurrent trips of the Drywell Cooler Fans if restarted with a ,
LOCA conditio l l
D) Overpressurization and rupture of piping inboard of the closed l DWCW lsolation valves with a LOCA conditio !
I l
l L i Page 88 of 100 l
!
l L
.
-Test ID: 1999 SRO Esam Level: SRO Date: 9/13/99
-
ID Number: 117 89) Unit 2 was operating at 100% power when a totalloss of Instrument Air occurred resulting in a plant scram. T-101, "RPV Control" was entered on high reactor pressure at the time of the scram Normal scram actions have been completed, no other actions have been performe In accordance with T-101, RPV pressure controlleg, which of the following is the correct method for pressure control under these conditions? .,
A) Manual operation of SRVs between 950 psig and 1050 psi B) . Automatic operation of the EHC system at 920 psi C) Manual operation of' ADS SRVs to stabilize pressure below 1050 i psi ,
D) Automatic operation of SRVs at their setpoin i l
l l
l Page 89 of 100
.
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
-
ID Number: 118
,
90) Unit 3 has experienced a reactor scram following a steam leak in the
~ Drywell. The CRS directs restoration of Drywell Instrument Nitrogen from T-101, RPV Control, to permit manual reactor pressure contro Restoring instrument Nitrogen to the Drywell in accordance with GP-8E,
" Primary Containment isolation Bypass":
A) ' May contribute to a flammable environment in the Drywel B) Will only. supply nitrogen to the "B" Instrument Nitrogen Heade C) May deplete CAD nitrogen tank inventor D) Will only be permitted if Instrument Air Header pressure is greater than Drywell pressure.
!
l l
l l l
'
)
l l
!
i i
i l
Page 90 of 100 l
I
L i l
l
'
- .
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99 i ID Number: 119 91) Unit 3 has experienced a transient and the following is observed:
- Torus pressure: 9 psig
- Torus temperature: 200 degrees F
- Torus level: 14 feet
- Reactor pressure: 1000 psig !
- RHR "A" Loop Flow: 23,000 gpm
- Core Spray "B" Loop Flow: 7500 gpm
- All other low pressure ECCS pump are NOT in servic Use the attached T-102 Sheet 3 curves to determine if Net Positive !
Suction Head (NPSH) requirements are being me !
A) There is sufficient NPSH for the "A" Loop of the RHR ONL i B) There is sufficient NPSH for the "B" Loop of Core Spray ONLY, C) There is sufficient NPSH for both the "A" Loop of RHR and the "B" Loop of Core Spra l D) There is NOT sufficient NPSH for either the "A" Loop of RHR or i the "B" Loop of Core Spra .
I i
Page 91 of 100 l
l l
s Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99 i
ID Number: 120
'
92) A full power ATWS occuned on Unit 2 which caused excessive heat input to the Torus and a Torus leak. The following conditions currently
- exist:
- Main Condenser is availabl Six rods are stuck full out, all other rods are fully inserte Reactor pressure: 950 psig
- Torus temperature: 175 degrees F. and steady
- Torus level 14 ft. and dropping Use the attached portion of T-102 to determine which of the following actions are required as Torus level drops from 14 ft. to 12 f A) Perform an Emergency Blowdown using T-112.'
B) Perform an Emergency Blowdown using Bypass valve C) Depressurize to 900 psi D) Depressurize to 750 psi !
l
!
l l
!
,
!
l Page 92 of 100 i
L ..
F-
.
.
a
-- Ce .
3 NE A1 CA,sFG ACIff ifarI/f-l Ltutt s
e n. nu.. :: ute. tu e... l
'
g,...........]
in ,
.,, , u n...
8 #EGAa0Lt38 0F as poser CA C00L0ceu AAIE ; ",g'=~ -
~
S -g g (#f #f INTO T-115 OR If E AM C00t!NG N gg g,ggga ,ggg g
sECr10N Os I-tis is 31 at0utat0 I rm - -
g g LOWEA T HE N~*" RPt PR St TO WAINfAIN AN =
- '" r
'
- -2 8 '"' '8 8 8 see te sesJ rsta OPERAf!NC PC NT 84 LOW CURVE 1/f-l w - - --= %
g 3 seg # fM f6 0
,u.90.J.AP.I.!. 8 g
,-
= ,e - -
g u i,. ,.
-
. i ,.
I ..
, i
., - ,
, -
8 - 150 /
' / 184 Il 11 18 14 il 18 t1A l I
j
.... u,it . n !
.. ,- . .
I CAN L ITESS
'
s APv Pai,st,ANO l
. .... i g,u,i,,,i, ,, p SELOW CURVE t it .g I/I-l l ( NO) L m... I"III L PERFORW AN EWERGENCY 9Lov00sN Ut!NQ I.lly
,
L T/f =10 I
-
SRO - 92
I l
.
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
ID Number: 121 93 ) For a lowering suppression pool level T-102, " Torus Level". directs that if Torus level cannot be maintained above 9.5' secure HPCI. It does not l direct that RCIC be secured until < 6'.'
What is the basis for securing HPCI but not RCIC at 9.5'?
A) HPCI turbine exhaust becomes uncovered at 9.5', RCIC turbine exhaust becomes uncovered at 6'.
B)' HPCl turbine exhaust becomes uncovered at 9.5', RCIC turbine exhaust is an insignificant containment inpu C) HPCI NPSH becomes a concern at 9.5', RCIC turbine exhaust becomes uncovered at 6'.
D) _ HPCI NPSH becomes a concern at 9.5', RCIC turbine exhaust is a insignificant containment inpu I
. .
l l
I'
!
l l
l
,
i l
'
Page 93 of 100
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
'
ID Number: 122 94 )' T-111, " Level Restoration" was entered on Unit 3 following a loss of all off site power and a failure of all diesel generators to start. Current plant conditions are as follows:
- Reactor pressure is 800 psi Reactor level-195" and dropping slowl HPCI tripped on a loss of lube oi RCIC is blocked out of servic Evaluate these plant conditions and determine the status of Adequate
- Core Cooling (ACC).
A) ACC exists untillevelis below -200" B) ACC exists untillevelis below -210" C) ACC does NOT exist, since levelis below -172".
'
D) ACC does NOT exist, since injection is not presen I i
l
I l
Page 94 of 100
,
-
ID Number: 123 95) Level recorder LR-2-02-3-110A blue pen is fed by LT-2-02-3-072C
' Wide Range" and LT-2-02-3-073C " Fuel Zone" level transmitter If level transmitter' LT-73C failed upscale and then actual reactor level dropped to -172", what would be the impact on vessel level indications and ECCS initiation from reactor level?
A) LR-110A blue pen input would swap at -100", low level ECCS
' initiations would NOT be ' impacte B). LR-110A blue pen input would swap at -100", low level ECCS initiations would be impacte C) LR-110A blue pen input would NOT swap at -100", low level ECCS initiations would NOT be impacte D) LR-110A blue pen input would NOT swap at -100" low level ECCS initiations would be impacud.
i l
l Page 95 of 100 l
L
. .. , .
.
l Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
-
ID Number: 124 96) Following an ATWS and Group 1 isolation on Unit 2, the following conditions exist:
- Reactor power: 30 %
- Reactor level: -100"
- Torus temperature: 115 degrees SRV's A, B, C, G open T-117 level power control directs RPV injection be terminated and prevented using T-24 For the conditions listed above, which of the following concerns is the basis for performing T-240?
A) Uncontrolled injection of large amounts of cold wate ) Power generation which is a threat to primary containmen C) Neutron flux oscillations which challenge fuel clad integrit D) Power excursions while establishing minimum alternative RPV flooding pressur Page 96 of 100
-
- _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ .
,
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99 ID Number 125 97) Unit 2 was operating at 100% power when a Reactor high pressure scram condition occurred due to a totalloss ofinstrument air. Control rods failed to insert, reactor pressure peaked at 1180 psi The following plant conditions currently exist:
- Reactor power: 35%
- Reactor level: +23"
- Reactor pressure: 1140 psig
- Full core display blue lights lit
- A & B Air Header pressure: 0 psig Determine which of the following TRIP procedures willinsert the control l
'
rod A) T-213 " Scram Solenoid De-Energization" B) T-214," Isolating and Venting the Scram Air Header"
)
C) T-215, " Control Rod Insertion by Withdraw Line Venhng" D) T-216 " Control Rod insertion by Manual Scram or individual l
Scram Test Switches" I
l Page 97 of 100 f
l
[
,
Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
1
ID Number: 126 98) A steam leak exists in the Unit 3 Turbine Building. T-104," Radioactivity Release", has been entered due to high ventilation stack radiation alarms. The Equipment Operator (EO) then reports that Turbine Building Ventilation is trippe Under these conditions, determine the appropriate response to the EO's report that Turbine Building Ventilation is trippe A) Restart ventilation to monitor the releas . B) Restart ventilation to lower the radioactive releas C) Maintain ventilation tripped to prevent an unmonitored releas 'D) Maintain ventilation tripped to lower the radioactive releas l
!
i Page 98 of 100
.
,
L--
[?
l l
i' Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99
'
l ID Number: 127 99) For which of the followirig conditions would direction be given to initiate l Drywell Sprays regardless of whether Adequate Core Cooling is assured?
l
'. A) To prevent exceeding the Pressure Suppression Pressure Limi B) To maintain Drywell pressure below the Drywell Spray Initiation i Limi C) To mitigate the consequence of a H2 deflagratio D) To mitigate the consequences of containment overpressurization.
l l
l
!
l I
l I
I i
l
. l I
l
.
l
!
l l
Page 99 of 100 l
--
,.. . . . . .. . .
' Test ID: 1999 SRO Exam Level: SRO Date: 9/13/99 ID Number: 128 100) Unit 2 has experienced a large LOCA during which the core was uncovered and fuel cladding oxidation occurred. Adequate core cooling is now assured, containment pressure is 22 psig. Chemistry reports the following containment parameters:
- Torus 02: 4%
- Torus H2: 3%
- Drywell 02: 6%
- Drywell H2: 2%
Using this data and T-102 Sh. 2, " Primary Containment Control" tables (PC/G-1, PC/G-2), provided, select the appropriate TRIP Legs to be entered from those listed belo A) DW/G-1 and T/G-1 B) DW/G-2 and T/G-1 C) DW/G-3 and T/G-2 D) DW/G-2 and T/G-2
.
'
<
t Page 100 of 100
- -
-
- -
- -
. . . . _ . _ . . . . . . _ _ . . _ _ _ _
T ABLE PC/G-1 DW COMBUSTIBLE DAS COWTROL ORYWELL OIYDEN LEVEL AT LEAST SI OR UNKNOWN BELOW SI TORUS HYOR00EN LEVEL BELOW 0.5I 0.5I TO 5.991 AORLUNKNOWN
^8 S NO ACTION NO ACTION BELOW 0.5I REQUIRED REQUIRED su og DW/O-2
- a: * 0.5I to 5.99I DWM-3 g AT LEAST 61 DW/0-1 OR UNKNOWN
>
T ABLE PC/G-2 TORUS COWBUSTIBLE DAS CONTROL TORUS DIYDEN LEVEL AT LEAST SI OR UNKNOWN BELOW SI DRYWELL HYOR00EN LEVEL
.
A L S BELOW 0.5I 0.51 T O 5.9 9 5 OR UNK WN NO ACTION NO ACTION BELOW 0.55 REQUIRED REQUIRED
= w T/0-2 EO 0.51 to 5.99I
~ #~
AT LEAST 65 OR UNKNOWN SRO - 100
-
,\
= J c x: t-a
,~
-
gI a =
-m c .o D - W 1 o < OO I m
a
=
I o>
G o- -
m- d_ - O- < mO I O
e I J > ,os N t C C -
m< d_ , O_ o I 0O c I a
- m_ , o . gp m a O
uI A> -
c TI O g < NC I u
m A
I s U> o _ gy m- N O
N ,N- o
-
e, -
I
~-
O_ ? o@ I 2 M U
I O> * c o , gy -
mu S A -d_
.
-
Os o. t oO I
I N> a t ,
- m-
,
G _ X E e -d_ - O- < VO I t
y L O, S-u s
i D>
C a., - ma /
R I ta n iWm - Oa 1 < M@ I v H g B s I
C = n. c ,o: , - m2- P T d- N@
o v N S n n
N I
u i -
s a. O , > a .' mc *
c o,. ga - ma O I
M O ,du wOu .o-
.
5 -@ I ui w a ,. > e cn , ,om g5 - ma T A R K ,Wa e rO qos l OC I
'0
.
T O R 3 I aN , > c ,oa 9 m s,. S F O .
W - , f mW I J W VI eC J n>a ,ou , ,ma ,da ,O . g f 0G L c I I
aA,>a , o ,. , 9 emu L E A C nda r Ov q *
0 NW I i
aU ,> a
%cnu ,9 ,mo Z; N ,We A ,Oa yo 4 CM I am ,> o ,ou ,9 nma E m ts t
e I
U TS Q S I ,da
.
, m@
I I
l aN n>u noo ,9 amo D S ,du qOa , ,
i vD I E A T
i I
aC o n>o aC ,> s 7uu
,n, c ,o,. gW , mu 9 , ms W -
C D E E R V I ndu
,Wa
- r tOa oo-r Oa go i MW t NW I
I I
MO , > nm ,ou gW , mu R E e c O C nW-- tO ._ g 5 "Q I I qnc 9 ,mu C E
'
CQa a >a E R nda 7Oa 6 9 1 OO I
,oo - H R I
CN Q ,>a 9 , mo T O ndu o- 5 ON I I
QJ n>u GC ov 9 ,m D N nda 1O r b oNt I N N I
MA ,> u - ot m ,9
, , m ,- A E nda -roe o 4 NN h T S V I
I I
CQtU 1 c .. ou ,9 .mv
R G ,d-s o. 4 oN I I
I Q) n>a Gc MN , > i
,oa
,oa ,9 9 .ma, 3 WR
E B
L L
O E
D N
U R
E H
,du
,
,4Oo { 4 mN I
.
.
m's ts
^ la idu o s vN
% T -
I M R I I OD CC m n ,. ,oa ,9 , m ,, U Y M D E nda POv s MN N N W A i I I O n>a N P A I
GO ,oo m9 , m :.
Y E nd ,. o . b NN T O D V I
C E AH I
I J
CO,>a ,nc u 1ou ,ou
,9 ,mu
,ma R
A E E LT D I R
U E
P MW I
,du moo q l eN I I
J *-
(
Jon>o CG ,cu n t ,9
,9
,
,mu L
C T I C E
S
/
Y 6VEE WN I
, ,do
,do 1Oo g P
'tOa q o ;.
4 Ok
$ oC I
U E N S N R O I
I JJ n>u (C 1mu r c n9 ,mu T N
L A
A P E ndc tOa r g .
% oC I O R EI R I C M V Y t I J CA,>o , too 9 ,mu C U MP A M I Jt CU,>si n c, ,
9 ,mu E O P C A
N O
S O
C EH D S ndu qOa g
,du qOa i, o 5 NC I
- r I I t o@ I I o , o ,. ,9 go
- ,to ma sdu s m@ I I 3 Pm oo g5 ,mo E
LN n > 2 c u ,da 2 G qOu ,- o .
4 vD I I 3D ,>o ,oo ,9 ,ma E N 0C r A nda .Oo 3 ,
5 MC I
.
J 2 H I
(C,>. , mo c
,
.ma D C ndu -o s
% NO I c
-. gm ,mo O FO > o * n ,o
I c, .
- T ,da ,Ou 3o t "O I
.-
I
&a ,to o noo ,9 ,ma *= Y L ,du s o- 5 O@ I 4 I A o , > n, ou gw , m >-
3 E G c 8 T ,
Nu r Oa {
4 oD I I J ,>a ". to . ,mo L t E AC ,9 I
- L , du , Oa ) 4 oD C I C A P I
AA,>a Ono c o,. g5 mu N c M ,
du iOe
-
ou < ND I E :
O I
A U , > t
'nm o gy ,mo P E
""c 2 L C , da t o- % OD I P E
"oa E #
I M AO k
- t ou 9 ,mo M S , de tO; qou
-
t WD I W E A A I
AN n> a * n- oa g5 ,mo t
M S X R ,du q0a q i vD I S
U E E
c I
mD,>e eC *c n 1 u 1 ,9 ,mu * * * 3 da q0a q f MD I
l
I AO w >a ou ,9 ,ma ,Wa - .
f ND I y t
I UO a > a 1c * oa l ,9 ,mu ,du n0u -u % "C I E
t bm i l ,. n i tEc
-
l