IR 05000321/1993300
| ML20056C167 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/19/1993 |
| From: | Bartley J, Lawyer L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20056C163 | List: |
| References | |
| 50-321-93-300, NUDOCS 9303300175 | |
| Download: ML20056C167 (114) | |
Text
. _. _ ._ . _ . _ _ _
UWTED STATES l
. ,o(c#"%g
'
NUCLEAR REGULATORY COMMISslON j 2 "
o REGION H ,
5 $ 101 MARIETTA STREET. ,
- i
, l t ATt.ANTA GEORGIA 30323
,., **, *
w/ ;
ENCLOSURE _1 y
' Report No.: 50-321/93-300 ;
Licensee: Georgia Power Company P. O. Box 1295 Birmingham, AL 35201 j Docket Nos.: 50-321 and 50-366 ;
License Nos.: DPR-57 and NPF-5 ;
Facility Name: E. I. Hatch Nuclear Plant
!
Inspection Conducted: February 18 - 25, 1993
,
rn Chief Examiner: A7Jfrodkw ~ 3/8/93 Date signed Jona 3 H. Bart1 Q Examiners: M. Daniels, Sonalysts, In l G. Harris, NRC !
!
Approved By: /' :
Lawrence L. Lawyer,' Chief Date Signed OperatorLicensingSectid(n/1 Operations Branch ;
Division of Reactor Safety ,
SUMMARY q Scope: Regular, announced initial examinations were conducted during the weeks of February 15 and 22,1993. The examinations were ,
administered under the guidelines of the Examiner Standards (ES), .
NUREG-1021, Revision 6. Written and operating examinations were adndnistered to seven Senior Reactor Operator (SRO) applicants. A requalification retake examination was administered on February-17, 1993, to one SRO who failed the requalification examinations given in September 199?.
Results: Seven of seven SR0s passed the initial examination for a 100 percent f pass rate. One of one SR0 passed the requalification retake !
examinatio ;
A' strength was noted in the area of crew communications (paragrap !
2. e) ..
!
- No violations or deviations were identifie '
9303300175 930322 l PDR ADOCK 05000321 ;
V PDR ;
,
. -_ ,
i
!
!
!
I t
REPORT DETAILS ;
t Persons Contacted l
Licensee Employees R. Belcher, Instructor, Operations Training i J. Betsill, Operations Unit Superintendent .
- S. Crosby, Operations Training Supervisor (Classroom) l
- 0. Fraser, Site Supervisor, SAER !
- G. Goode, Manager, Engineering Support !
- S. Grantham, Supervisor, Operations Training !
- Gunn, Plant Instructor, Operations Training
- J. Lewis, Operations Manager
- D. Read, AGM-PS
- L. Sumner, General Manager
- S. Tipps, Manager, Nuclear Safety and Compliance Other licensee employees contacted included instructors, operators, engineers, and office personne NRC Personnel
- E. Christnot, Resident Inspector i
- L. Lawyer, Section Chief, Region II ,
- T. McKernon, Examiner, Region IV i
- Attended exit interview Discussion Results
Seven SR0 candidates passed the initial examination. One SRO passed ,
a requalification retake examinatio ; Reference Material ,
The licensee submitted reference material as required by NUREG-102 The materials were well organized and easy to use. The training staff promptly informed the examiners when updates were made to procedures during examination development. This helped alleviate i last minute changes during the preparation and examination week I
! Examination Development l Representatives from the licensee training staff reviewed the written examination in the regional office on January 25 - 27, 199 The review was thorough and provided valuable input to improve the accuracy of the information contained in the written examinatio I
!
i l
I l '
Report Details 2 l
The facility provided the NRC with a proposed walkthrough l examination for the requalification retake examination. The l
examiners reviewed the proposed examination for content and 1 structure per the guidance of NUREG-1021. The examiners determined the examination to be adequate and administered it without any I modifications. Additionally, the examiners reviewed the operator's l remedial program and determined it to be adequate.
l ; Examination Administration I The examinations were administered without major problems. The i
simulator primary computer failed twice during the evaluation of one i l crew which caused a 1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> delay in the administration of the }
simulator examinations. Licensee support in running the simulator ;
and providing other assistance was good. Cooperation received from '
shift personnel was excellen :
The licensee provided an evaluator to assist in the administration j of the requalification retake examination. The evaluator did a good l job and no problems were encountere Candidate Performance :
l
'
l
'
Candidate performance during the examinations was excellen '
Written examination scores ranged from 90 percent to 97 percent, no i Job Performance Measures (JPMs) were failed, and overall performance during the simulator examinations was excellen During the simulator portion of the examination, the candidates' ;
communication skills were excellent. No instances of missed i i communications or mis-communications occurred. The SR0s gave crew briefings consistently and asked the crew for feedback at the end of ,
the briefings. Candidates in the R0 positions ensured the SR0 l understood and acknowledged their report Simulator Facility ;
The simulator worked well with the exception of downtime caused by I the simulator computer, a failed switch, and the inability to ;
override automatic actuation signals.
- (1) The simulator primary computer failed twice during the I evaluation of one crew. This caused a 1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> delay in the i I administration of the simulator examinations. The second l
failure of the primary computer required switching to the ,
backup computer which accounted for the majority of the l downtim i
! !
(2) The switch for the CS/RHR Diagonal Pump Room Cooler, 2T41-B002B, failed mechanically during the examinations and .
! was replace ;
i
!
l l
.
Report Details 3 i
(3) The simulator did not have the capability to override automatic actuation signals while still allowing manual actuation. This prevented the examiners from evaluating the candidates'
response to plausible events where systems fail to actuate automatically and require manual operation. The simulator :
operator could override the system or component switch to ;
off/close and remove the override to allow manual actuatio However, it was difficult for the simulator operator to monitor the candidate's actions so he knew when to remove the override to allow manual actuation. As aa example, the examiners wrote i a JPM for a manual initiation of Core Spray with a failure of *
the minimum flow valva to shut. To accomplish this, the simulator operator overrode the minimum flow valve switch to i open. The simulator operator removed the override when the '
candidate positioned the minimum flow valve switch to clos On back panel systems, such as the Standby Gas Treatment System, this could not be done at all because the simulator -
operator could not monitor the candidate's action Procedures ;
The examiners noted two procedural problems during the course of the examinatio (1) 31E0-E0P-106-2S Revision 1, " Restoration of RPV Water Level Following RPV Flooding," step 3.9 stated:
"0 PEN Torus Spray or Test Viv 2 Ell-F025A(B) by placing jumpers [
as follows:" '
This left the control switch in the close position which resulted in the valve repeatedly cycling full stroke. The facility training staff submitted a procedure change request to correct thi t (2) 34AR-601-406-2S Revision 4, "0/G AVG ANNUAL REL LIMIT WILL BE j EXCEEDED," step 5.2, directed the operator to check radiation levels on monitor 2011-K601 or 2D11-K602 at panel 2Hil-P60 These monitors, 2D11-K601 and 2011-K602, are NUMAC radiation ,
monitors which are located on panel 2Hil-P604. The monitors on 2Hil-P600, 2011-R601 and 2Dll-R602, are pen recorders that monitor the same points as 2Dll-K601 and 2Dil-K602. This 3 discrepancy caused confusion and resulted in the delay of l reporting radiation levels to the SR ' Plant Material Condition
.
On Tuesday of the exam week, while administering JPMs, an examiner noted numerous lights for the LPRM four rod display in the back of panel P603 were out. On Unit 1, there were 8 of 16 lights out, and >
,
I
Report Details 4 on Unit 2, there were 3 of 16 lights out. Thursday, the examiner determined the light bulbs provide the light for the illuminated pointer on the 16 LPRM modules and the meter is not functional if the light is out. The examiner went to the Control Room to '
follow-up with Control Room personnel and noted the lights had not i been replaced. The Control Board Operators (CB0) did not know the ;
lights were out until questioned. When asked if it was a problem, the CB0s stated they would only use them at power if there was a problem with power oscillations. The examiner informed the Shift Operations Supervisor who took prompt action to replace the bulb The label plate for the Containment Spray Viv Control switch did not include the alphanumeric designation, 2E11-S17A(B). This was inconsistent with the rest of the labeling in the Control Roo . Training Program Review f
The examiners reviewed shift manning practices and compared them to both Technical Specification requirements and requalification '
training crew size. The maximum Technical Specifications manning levels requires two SR0s, three R0s and a Shift Technical Advisor -
(STA) to be on shift. The requalification crews on the simulator are composed of two SR0s, three R0s and an STA for one unit. The facility assigns four SR0s, seven R0s and an STA to each shift to cover both units. The facility makes it a policy to maintain a ,
minimum of four SR0s, five R0s, and an STA on shift. Although this '
policy was not documented or committed to by facility management, a review of the shift manning records for the previous two months showed that the facility always exceeded these minimum manning ,
levels on shift. The examiners determined the facility shift l manning significantly exceeds the levels required by Technical Specifications. The examiners also determined, based on shift manning records, the crew size used during requalification training could be manned on both units by shift personnel during a dual unit casualt The examiners also reviewed simulator usage and availability for the period January 1992 to December 1992. The examiners determined the simulator availability was approximately 99 percent. During the time available, the simulator was used for operator training related functions approximately 32 percent of the time and was' secured for i maintenance approximately 7 percent of the time. The facility did l not maintain records to document simulator use of non-operator training functions. However, based on discussions with the Training Department staff, the examiners determined outside organizations were not affecting the simulator availability for the Training Departmen No violations or deviations were identifie ,
i
l l
Report Details 5 - Action on Previous Inspection Findings
!
i (Closed) IFI 92-301-01, review of facility implementation of procedures and operator training to assure that potential level indication errors will not result in improper operator actions. This item concerned the facility implementation of training and procedures in response to issues identified in NRC Generic Letter 92-04, " Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(F)." During this examination, the examiners reviewed the licensee's response to Generic Letter 92-04, incorporation of the concerns into initial license and requalification training, and procedure guidance for level instrumentation problems. Examiners found the issues l of the Generic Letter were incorporated into lesson plans and the l operators had adequate procedural guidanc Examiners considered the ;
licensee's corrective action to be adequate and this Inspector Follow-up !
Item is close . Exit Interview l At the conclusion of the site visit, the examiners met with l representatives of the plant staff listed in paragraph 1 to discuss the J results of the examinations. The licensee did not identify as
.
proprietary any material provided to, or reviewed by the examiner i l Dissenting comments were not received from the license I i
i
.
I l
l l
l
,
l
. .-
l
,
!
ENCLOSURE 2 ;
i SIMULATOR FACILITY REPORT !
t Facility Licensee: Georgia Power Company f Facility Docket Nos.: 50-321 and 50-366 .i t
Operating Tests Administered On: February 22 - 25, 1993 This form is to be used only to report observations. These observations, in :
and of themselves, do not constitute audit or inspection findings and are not, without further verification and review, indicative of non-compliance with 10 ,
CFR 55.45(b). These observations do not affect NRC certification or approval '
of the simulation facility other than to provide information which may be used j in future evaluations. No licensee action is required solely in response to these observation Item Comment l
Unable to override The simulator does not have the capability to l automatic actuation override automatic actuation signals while still allowing manual actuatio .
Simulator Primary Primary computer for the simulator failed twic computer failed The second failure was a hard failure that required shifting to the backup compute Pump Room Cooler The switch for the CS/RHR Diagonal Pump Room Cooler, switch failed 2T41-B002B, failed mechanically during the examinations and was replace i
e a
.l
!
!
!
- - - _ _ . . _ _ _ . _ _ - . - - . - _ _ _ _ _ . - _ _ _ _ _ _ _ .
(
m
@ W
,
/'kasbe~ $l0!c./{
NRC Official _Use Only 7J 3 CD lll3 - / 9/ Nuclear Regulatory Commission Operator Licensing Examination i
l This document is removed from l Official Use Only category on l date of examinatio l l
!
!
!
!
NRC Official Use Only l l
l l
l L.'
I O O- !
'l
!
i i
U. S. NUCLEAR REGULATORY COMMISSION :
SITE SPECIFIC EXAMINATION l SENIOR OPERATOR LICENSE ,
REGION 2 !
i t
"
CANDIDATE'S NAME: j
!
FACILITY: E. I. Hatch 1 & 2 j REACTOR TYPE: BWR-GE4' !
!
DATE ADMINISTERED: -93/02/19 i
!
INSTRUCTIONS TO CANDIDATE:
i Use the answer sheets provided to document your answers. ' Staple this cove j sheet on top of the answer sheets. Points for each question are indicated in- .i parentheses after the questio The passing grade requires a final grade of !
at least 80%. Examination papers wil] be picked up four (4) hours after the- i examination start !!
!
CANDIDATE'S I TEST VALUE SCORE % f
!
I 100.00 % TOTALS -i FINAL GRADE 'I
!
!
.
l
?
All work'done on this examination is my ow I have neither given nor received ai Candidate's Signature !
'l
!
!
!
~!
i
i
.i l
f I
o 8 i
&
i O O SENIOR REACTOR OPERATOR Page ~2 ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blan MULTIPLE CHOICE 018 a b c a b c d 019 a- b c d 002 a b c d 020 a b c d l
003 a b c d 021 a b c d l
.i 004 a b c d 022 a b c d l t
005 a b c d 023 a b c d !
i 006 a b c d 024 a b c d 007 MATCHING 025 a b c d l
}
a 026 a b c d .j
o 027 a b c d ]
c 028 a b c d i
d 029 a b c d' j
'
MULTIPLE CHOICE 030 a b c d j 008 a b c d 031 a b c d l l
009 h b c d 032 a b c d l
!
010 a b c d 033 a b c d .
j
'
011 a b c d 034 a b c d 012 a b c d 035 a b c d 013 a b c d 036 a b c d 014 a b c d 037 a b c d 015 a b c d 038 a b c a b c d 039 a b c d 017 a c d 040 a b c d
-
- _
.
o o 1-i
.t SENIOR REACTOR OPERATOR- Page 3 -l
!
ANSWER SHEET
i Multiple Choice (Circle or X your choice) !
If you change your answer, write your. selection'in-the blan :
i
';
041 a b c- d 064 b c d ,
i 042 a b c d 065 a b c~ d !
043 a b c d 066 a b c d 044 a b c d 067 MATCHING 045 a b c d a 046 a b c d b l 047 a b c d c l 048 a b c d d l
.;
'
049 a b c d MULTIPLE CHOICE t
050 a b c d 068 a b c d :
,
051 a b c d 069 a b c d ;
052 a b c d 07v a b c d 053 a b c d 071 MATCHING 054 a b c d a l
!
055 a b c d b i
!
056 a b c d c
[
057 a b c d d 058 a b c d MULTIPLE CHOICE 059 a b c d 072 a b c- d l 060 a b c d 073 a b c d !
061 a b c d 074 a b c d-0'62 a b c d 075 a b c- d 063 a b c d 076 a b c d I
!
i i
?
_
/7 &
V W SENIOR REACTOR OPERATOR Page 4 ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
077 a b c d 094 MATCHING 078 a b c d a 079 a b c d b 080 a b c d c 081 a b c d d 082 a b c d MULTIPLE CHOICE 083 a b c d 095 a b c d 084 a b c d 085 a b c d 086 a b c d 087 a b c d 088 a b c d 089 a b c d 090 a b c d 091 MATCHING a
b c
d MULTIPLE CHOICE 092 a b c d 093 a b c d (********** END OF EXAMINATION **********)
. . - . . . . . .
- -. -
.,
!
'
O O 3
i
"
Page 5 l
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS l During the administration of this examination the following rules apply: {
i Cheating on the exan.* nation means an automatic denial of your application l and could result in mure severe penaltie t
!
l
~
2. After the examination has been completed. 'cn2 must sign the statement on !'
the cover sheet indicating that the wo- .e your own and you have not d~
received or given assistance in complet.ng the examination. This must be !
done after you complete the examinatio l l
-- 3 . Restroom trips are to be limited and only one applicant at a time may l leave. You must avoid all contacts with anyone outside the examination j room to avoid even the appearance or possibility of cheatin *
l Use black ink or dark pencil ONLY to facilitate legible reproductions.
j li
! Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer sheet.
6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED !
AND DO NOT WRITE ON THE BACK SIDE OF THE PAG i 7. Before you turn in your examination, consecutively number each answer sheet, . ]
I including any additional pages inserted when writing your answers on the examination question pag . Use abbreviations only if they are commonly used in facility literatur Avoid using symbols such as c or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it .ou . The point value for each question is indicated in parentheses after the questio i 10. Show all calculations, methods, or assumptions used to obtain an answer to any short answer questions.
!
, 11. Partial credit may be given except on multiple choice questions. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
, 12. Proportional grading will be applied. Any additional wrong information
"
that is provided may count against you. For example, if a question is worth one point and asks for four responses, each of which is worth 0.25
, points, and you give five responses, each of your responses will be worth g 0.20 point If one of your five responses is incorrect, 0.20 will b !
deducted and your total credit for that question will be 0.80 instead of
-
1.00 even though you got the four correct answer . If the intent of a question is unclear, ask questions of the examiner only'.
l
l l
.- . , . _ . . , . _ _
O -
O !
I Page 6 {
i
!
i 14. When turning in your examination, assemble the completed examination with !
examination questions, examination aids and answer sheet In addition, !
turn in all scrap pape ls L
15. Ensure all information you wish to have evaluated as part of your answer is j en your answer sheet. Scrap paper will be disposed of immediately following :
the examinatio !
I 16. To pass the examination, you must achieve a grade of 80% or greate i i
i 17. There is a time limit of four (4) hours for completion of the examinatio .
.:
18. When you are done and have turned in your examination, leave the examination l l
area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while j the examination is still in progress, your license may be denied or revoke [
t i
!
!
l
!
t
!
i i
!
,
-t
,
.
-[
a
!
!
I k
!
!
.
i l
&
W SENIOR REACTOR OPERATCR Page 7 QUESTION: 001 (1.00)
A high radiation of 20 mr/hr in the Unit 1 Reactor Building Exhaust Ventilation has resulted in initiation of the Standby Gas Treatment System Unit 1 and Unit 2 Standby Gas Treatment System auto star Which ONE of the following describes the suction flow path for the two units? Unit 1 SBGT takes suction from only the Unit 1 Refueling floor and Unit 2 SBGT takes suction from the Unit 2 Refueling Floor and the Reactor Buildin b. Unit 1 SBGT takes suction from only the Unit 1 Reactor Building and Unit 2 SBGT takes suction from only the Unit 2 Reactor Buildin Unit 1 SBGT takes suction from the Unit 1 Reactor Building and Refueling Floor and Unit 2 SBGT takes suction from only the Unit 2 Refueling Floor, Unit 1 SBGT takes suction from the Unit 1 Reactor Building and Refueling Floor and Unit 2 SBGT takes suction from only the Unit 2 Reactor Building.
QUESTION: 002 (1.00)
DCR 91-171 was generated to install pushbuttons on Panels 1H11-P654 and P657 to reset automatic ESF actuatio Which ONE of the following describes ALL the additional initiating signals which can be reset by these pushbuttons? Unit 2 Reactor Building and Refueling Floor Ventilatio Unit 2 Reactor Building and Units 1 and 2 Refueling Floor Ventilation, c. Unit 1 Reactor Building and Refueling Floor Ventilatio Unit 1 Reactor Building and Units 1 and 2 Refueling Floor Ventilatio !
.- a . . - .~ a-. - - - - ,...-a . , . . ~ . . = . -
!
O i SENIOR REACTOR OPERATOR Page 8 .
!
i
!
'
- QUESTION: 003 (1.00) l
-
l Unit 2 is operating at rated power. Hydrogen Recombiner "A" has been j tagged out of service for the last 10 days due to a failed heater j sectio )
,
Maintenance has just determined that the RHR water supply valve to !
Hydrogen Recombiner "B" is in the closed position and can NOT be j operate I
,
Which ONE of the following is the required action? ;
a. Restore at least one of the Hydrogen Recombiners to operable
-
j status within 30 days or be in Hot Shutdown within the next 12 l hour *
I
'
b. Restore both of the Hydrogen Recombiners to operable status
.
within 20 days or be in Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
f Be in Cold Shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
J i
d. Be in Hot Shutdown within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> !
3 i i
e !
j l
'
!
QUESTION: 004 (1.00)
Unit 2 is operating at 78% power.- The outer Primary Containment Air
. Lock Door is found to be partially open and cannot be closed. The inside Air Lock Door is shut and locke l Which ONE of the following is the required action to be taken? [
!
! a. Be in Hot Shutdown in 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> j b. Repair the door in the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in Hot Shutdown in
- 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ,
- !
' Operation may continue until the next overall airlock leakage j tes d. Repair the door within 30 days or be in Hot Shutdown in 24
,
hour !
i d
-
l i !
l
i
<
l,
,
.- , ._ . _ - . A A T W l
SENIOR REACTOR OPERATOR Page 9 I
I l QUESTION: 005 (1.00)
Unit 2 is operating at 60% power when an event occurs resulting in the i following plant conditions:
!
"A" Feedwater Pump flow 16%
"B" Feedwater Pump flow 28%
i RPV Water Level 30 inches
Which ONE of the following describes the response of the Reactor Recirculation Pumps? Pumps run back to 22%, and reset automatically when "A" feedwater flow increases above 20%. Pumps run back to 44 % , and reset automatically when "A" feedwater flow increases above 20%. Pumps run back to 22%, but must be manually reset when "A" ,
'
feedwater flow increases above 20% and RPV level increases above 32 inche ;
I
' Pumps run back to 44%, but must be manually reset when "A" feedwater flow increases above 20% and RPV level increases above 32 inche ;
QUESTION: 006 (1.00)
Which ONE of the following describes the speed changes which are enforced by the Unit 2 recirculation speed controller?
a. Manual speed decreases using the master controller are limited to a MAXIMUM of 2.5% per secon ,
b. Manual speed increases using the master controller are limited l to 2.5% per secon !
l c. Manual speed increases using the individual M/A stations are ;
limited to 1.0% per secon )
l d. Manual speed decreases with the individual M/A stations are I limited to 2.5% per secon !
l
. , . . . . . - . . . . . .- . - .-- ._ -... .. .-
,
<
O O
'
l l
'
' SENIOR REACTOR OPERATOR Page 10
- - QUESTION
- 007 (2.00) j i
With regard to the RHR System, MATCH each of the valves in Column A with l its associated setpoint OR interlock from Column B. (Items in Column B j
'
may be used once, more than once, or not at al Only one answer may j occupy a space in Column A) (4 required at 0.5 each) ;
!
COLUIG A COLUMN B !
(Valves) (Setpoints/ Interlocks) l RHR HX bypass valve psig i .5 inches ! LPCI Outboard Injection Valve /3 core coverage l (F017) auto open (Unit 1) permissive l
! psig l c. LPCI Inboard Injection Valve Open for 3 minutes I (F015) auto open (unit 2) following a LOCA signal.
. .psig f
i d. Shutdown Cooling Valves close psig with LOCA j signal presen l Interlocked open for .,
. duration of LOCA i
, psig with LOCA -!
9 signal Presen l Interlocked open for .
. 10 minutes.
a
,
QUESTION: 008 (1.00) l A Technical Specification Quarterly Surveillance for a Unit 2 LPCI system was due July 1st. Because of an LCO on the system, the ,
surveillance was not completed until July 15t {
Which ONE of the following is the date when the next quarterly i surveillance becomes due? !
i ! October 1
b. October 16 1 i
c. October 23 f r
d. November 8
'
l
'
I
.
!
!
&
I i
- . _ . . . _ . , _ _ _ _ . . _ _ . ._ .- _ _ . , . _ _ . . _ , _ _ _ . . . . . -
O O SENIOR REACTOR OPERATOR Page 11 QUESTION: 009 (1.00)
Unit 2 RCIC is being used for level control following a level transient which resulted in a reactor scra RCIC turbine speed increases to 5150 RP Which ONE of the following describes the response of the RCIC turbine to the increased speed?
a. Continues to operate with increased flo b. Isolates but can be realigned when turbine speed decreases below 4950 RP c. Trips on mechanical overspeed and will have to be reset locally, d. Trips on electrical overspeed, but not mechanical overspeed.
QUESTION: 010 (1.00)
Unit 2 RCIC is injecting to control reactor vessel level. A large oil leak develops on the in-service RCIC oil filter which results in decreasing oil pressur Which ONE of the following describes the response of the RCIC system as the oil pressure decreases?
a. The Governor Valve will close and turbine speed will decrease _to-zero RP b. The Governor Valve will open and turbine speed will increase possibly resulting in a turbine tri c. The Auxiliary Oil Pump will start, the Trip Valve will close and turbine speed will decrease to zero RP d. The turbine will trip but the steam stop valve (F045) will fail open on low oil pressur O O SENIOR REACTOR OPERATOR Page 12
~ QUESTION: 011 (1.00)
The operating' procedure for the High Pressure Coolant Injection (HPCI)
system cautions against prolonged operation with turbine speed less than 2000 RPM. Which one of the following is the reason for caution?
a. The potential exists for exhaust check valve _ chatter and reduced oil supply to the turbine governor'and bearing b. The Booster Pump may not provide adequate net positive suction head to prevent cavitation of the Main Pum c. The rate of steam flow through the HPCI turbine may not be enough to prevent it from overheatin d. Cooling water flow from the Booster Pump to the lube oil system heat exchanger is inadequat QUESTION: 012 (1.00)
During an ATWS condition on Unit 2, the operator places the SBLC control switch to the "A" pump operate position and observes that pump "A" starts but neither Squib valve fires. He then places the switch in the
"B" pump operate position and the Squib valves fir Which one of the following describes the condition of the Standby Liquid Control system after placing the switch to the "B" Pump' position?
a. Both pumps running with 82 gpm injecting into the RP "B" Pump running with 41 gpm injecting into the RPV "A" Pump running with 41 gpm injecting into the RPV d. Both pumps tripped with "0" gpm flow rate indicate O O i
'!>
SENIOR REACTOR OPERATOR Page 13 i
?
!
QUESTION: 013 (1.00)
Which ONE of the following statements describes the method used for re-establishing the required SRM minimum count rate during Unit 2 core reload?
a. Spiral reload the core until 3 cps is establishe !
-l
. Increase the voltage of the SRM high voltage power supply to !
obtain the required count rat ;
i c. Prior to fuel reload, install neutron sources in the i source tubes to establish 3 cp j i
d. Load up to four fuel bundles into core positions next to each j SRM until 3 cps is establishe j
!
!
QUESTION: 014 (1.00)
i A transient at Unit 2 has occurred, resulting in the following plant j
'
conditions-I Reactor Water level -110 inches [
Drywell Pressure 1.52 psig I HPCI Equipment Room Temperature 150 deg F .j
.
RCIC Tbrbine Exhaust Diaphragm Pressure 5 psig -j RWCU Equipment Room Temperature 125 deg F !
!
Which ONE of the following groups of isolations should have occurred? l
! , 3, and 5 {
! , 5, and 6 !
I c. 2, 3, and 4 j
! , 3, and 5
]
I
,
>
- __ _
.- ._ .. .
. - _ . __ . _ . . . _ __ .__ _ .
,
!
O .O 1
!
!
SENIOR REACTOR OPERATOR Page 14 !
I
~
l i
QUESTION: 015 (1.00) i
!
"
A LOCA has occurred with a simultaneous loss of_125VDC bus 2A, j
Which ONE of the following describes the response _of the ADS system j
,
logic to these conditions? 'j a t
a Both "A" and "B" logics will swap to alternate powe .l
!
- "A" logic will remain deenergized and the "B" logic will swap _to i alternate powe !
I c. "B" logic will remain deenergized and the "A" logic will swap to f
alternate powe !
I i d. No effect ADS logic is not supplied by 125VDC bus 2A f
'
l I
-
?
-
- QUESTION
- 016 (1.00) !
4 i
- A LOCA has occurred on Unit The following plant conditions occurre i at the times indicated: l
~
RPV level 3 (12.3 inches) at 2048 RPV level 1 (-101 inches) at 2103 i
Drywell pressure constant at 1.1 psig i
'
Core Spray Pump Discharge pressure 163 psig at 2101 Which one of the following is the earliest ADS will initiate? (Assume ,
RPV level and Drywell pressures do not change) !
I*
a. 2050 i
- b. 2101 ;
- !
'
c. 2105 l
~
d. 2118 f i
t
<
s 1 !
i
. .
&
W SENIOR REACTOR OPERATOR Page 15 QUESTION: 017 (1.00)
Unit 2 is operating at 23% RT The following plant conditions exis Drywell pressure 0.6 psig ,
'
Unit 2 Drywell venting is in progress via valves 2T48-F319 and 2T48-F320 During this time the Unit 1 Refuel Floor Vent Exhaust Radiation Monitors, 1D11-D611 A through D reach their isolation setpoint Which ONE of the following describes the effect on the Unit 2 Drywell venting?
a. Venting would continue with Unit 2 SBGT taking suction on the Unit 2 Drywell onl b. Venting would continue with Unit 2 SBGT taking suction on the Unit 2 Drywell and Refuel Floo c. Venting would stop due to Unit 2 Drywell Vent and Purge valves ,
'
F319 and F320 closin d. Venting would stop due to the Unit 2 filter crains suction dampers realigning to take suction on the Refuel Floor.
.
I l
l
. - ~ _ _ _ - . . _ . . . - .. . -. .
O O I l
l SENIOR REACTOR OPERATOR Fage l'6 !
i
i
,
l QUESTION: 018 (1.00) i s
l Unit 2 has scrammed from 100% power due to a Main Turbine tri The following plant conditions exist:
All rods in l The-following valves have amber and green indicators illuminated: j F013B, F013C, F013D, F013F and F013G i
The Red indicator for F013B is lit l Plant pressure is 860 psig and slowly decreasing l
Which ONE of the following describes the status of the LLS system and i the SRVs? f l
a. LLS is NOT armed, F013B, F013C, F013D, F013F and F013G lifted :
and are now close j
a b. LLS is NOT armed, F013C, F013D, F013F and F013G lifted and have {
closed, F013B lifted and should have reclose ;
!
'
,
'
c. LLS is armed, F013C, F013D, F013W and F013G lifted, and are'now closed and F013B is open. d..LLS is armed, F013B, F013C, F013D, F013F and F013G are open and ,
j wil.'. close at 851 psi .j i
- i i l!
j i QUESTION: 019 (1.00) !
- !
Which ONE of the following is a reason for nornally operating with the l l Reactor Water Level Selector switch selected to "B"? !
i l j a. If "A" were selected a failure of the "B" reference leg would l
result in a turbine trip on high leve !
i b. With "B" selected and a rupture of "A" reference leg, reactor i water level control would prevent level from decreasing, j i If "A" were selected and the "B" reference leg fails low, l reactor level would increase until terminated by the operato i d
I d. With "B" selected if "A" reference leg fails low, the reactor j
.
will scram on low level.
i ,
i
'
J
'
I i
l f
- . - _ . -. . - - - . . . , -
_ _ _ _ - _ _ _ .
A A W W SENIOR REACTOR OPERATOR Page 17 QUESTION: 020 (1.00)
Given: Unit 2 in control of D/G "B" D/G "B" Mode Switch in TEST (Surveillance being performed)
Electrical distribution NORMAL (Full Power Lineup)
D/G "B" is at rated speed and voltage, but not synchronized, when power to 4160 volt Bus 2F from SUT 2D is los Which ONE of the following describes the system operation due to the power loss? Bus 2F can be powered by D/G "B" when the operator takes the Output Breaker Switch to CLOSE with the SYNC SCOPE activate Bus 2F will be powered by D/G "B" automatically, after 12 seconds; appropriate loads will be picked up sequentiall Bus 2F can not be powered by D/G "B" while it is in the TEST mode, given these condition Bus 2F can be powered by D/G "B" after the operator resets the Lockout Relay, activates the SYNC SCOPE, and takes the Output Breaker to CLOS QUESTION: 021 (1.00)
i I
Unit 2 is operating at rated power when an event occurs which results in the following plant conditions:
CRD Accumulator Low Press or High Level annunciator CRD Hyd High Temp annunciator Charging water header pressure indicating "0" psig CRD system flow indicating "0" gpm Khich ONE of the following is the event which caused the conditions?
. Cooling water pressure control valve failed close ,
l
'
b. Operating CRD Pump trippe I c. CRD flow control valve failed close Flow stabilizing valves failed close l
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _
O O ,
[
!
F SENIOR REACTOR OPERATOR Page 18 I t
!
I i
QUESTION: 022 (1.00) (
, f Which ONE of the following parameters is used by the Rod Worth Minimizer i
'
to determine the point at which the RWM stops enforcing rod blocks while INCREASING power? APRM indicated power or First Stage Turbine Pressure l i Feedwater Flow or Steam Flow (
- t First Stage Turbine Pressure and APRM indicated power
' f
l Feedvater Flow and Steam Flow l
,
t
l OUESTION: 023 (1.00) j
- Which CNE of the following is the. reason that continued plant operation with an inoperable (or failed) jet pump is restricted?
, Invalid APRM Flow Biased SCRAM setpoints due to the change in f
-
] flow through a failed jet pump
?
l Increased blowdown area during a Loss of Coolant Accident
!
' Unbalanced neutron flux across the core due to flow variations f I Physical core and fuel cladding damage from a loose piece of I l the damaged jet pump ;
!
.
!
~
W i- !
- I
-
r 1-
!
i s
I i
-
!
!
!
!
!
.-
t
'i
. _ ..- - . _ - , -
- - - -
i
-
O O SENIOR REACTOR OPERATOR Page 19 QUESTION: 024 (1.00)
Unit 2 is operating at 800 power when an event results in the following conditions:
Reactor water level -12 inches RWCU Room ambient temperature 128 deg F RWCU Nonregenerative Heat Exchanger Outlet Tem deg F Which ONE of the following describes the response of the RWCU system?
a. The operating RWCU Pump trips with no isolatio Inboard Isolation valve F001 close c. Outboard isolation valve F004 closes d. Both inboard and outboard isolation valves clos QUESTION: 025 (1.00)
Unit 1 is in operation at 80% RTP. Unit 2 is in refueling and preparations are being made to remove the Unit 2 Drywell hea Which ONE of the following defines the conditions which must be maintained while removing the Unit 2 Drywell head?
a. At least one train of Unit 2 Standby Gas Treatment must be in operation and one train of Unit 1 SBGT must be operabl b. The Unit 2 Drywell must be maintained " air tight" until the Drywell Bulkhead Manways are closed and seale c. The Unit 2 Drywell bulkhead manways must be closed and sealed until the Unit 2 Drywell equipment hatch is. opene If the Unit 2 bulkhead manways are_open, the Drywell hatch may_
only be opened if the Unit 2 SBGT system is operatin . . _ . . ._ _ _ . .. . __
,
!
.
O o l l
i
SENIOR REACTOR OPERATOR Page 20- !
{
i
!
i QUESTION: 026 (1.00) !
)
Which ONE of the following sets of plant conditions will require the {
operator to manually initiate the Main Steam Isolation Valve Leakage !
Control System (MSIV-LCS) with a suspected fuel element failure and a !
<
LOCA signal present? j
'
a. MSIV leakage is suspected, Main Steam Line High High Radiation- l is present, Reactor pressure is less than 31.5 psig and Main !
Steam Line pressure is less than 35 psi lt d. 10 minutes have elapsed since the Loss of Coolant Accident l (;OCA),
L Reactor Pressure is less than 31.5 psig and, Main Steam i Line Pressure has bled down to 0 psi !
l I c. MSIV leakage is suspected, 10 minutes have elapsed since the !
Loss of Coolant Accident (LOCA), and both Main Steam line j q
pressure and reactor pressure are less than 31.5 psi j b. No steam line High or High High Radiation, Reactor pressure has }'
-
been less than 31.5 psig for a minimum of 10 minutes and, Main j Steam Line pressure is less than 35 psi )'
'
l
!
! !
j QUESTION: 027 (1.00) [
l t Which ONE of the following sets of parameters will cause a main turbine )
i trip following a turbine runback due to a loss of stator water cooling? ;
j (ASSUME a 18095 amp load prior to failure of stator water cooling)
f r
i TIME SINCE LOAD i
- FAILURE l
i .8 minutes 5290 amps j
<
i
!
3.2 minutes 4675 amps
' j
'
I
, .6 minutes 15230 amps j
! .3 minutes 16340 amps i
!
'
<
x
5
i f
I
- !
i
_ ._ ._ _. . _ _ -. . _ - . - - .
_.,
i O O !
l
SENIOR REACTOR OPERATOR Page 21- !
i
i
.
i !
(1.00)
'
QUESTION: 028 {
i Unit 2 is operating at 90% RTP when the following alarns are received: }
,
"APRM Upscale" {
- " Heater Trouble" i
"4th Stage HTR B007A Level High" i Which ONE of the following is the IMMEDIATE action which must be taken? .
l (Assume no reactor scram). !
i Reduce thermal power to 72% by decreasing recirculation flo !
I b. Reduce thermal power to 80% by decreasing recirculation flo ;{
!
c. Reduce thermal power to 45% with recirc flow or until the heater !
trouble alarm clear i
!
d. Manually scram the reactor if the APRM Upscale alarm has not l
'
cleared within two (2) minute l t
!
!
QUESTION: 029 (1.00) I f
p
,
Which ONE of the following sets of plant parameters would result'in a j
- trip of the "B" Reactor Feed Pump Turbine on Unit 2? (Assume all time *
delays have timed out.)
'
a. Reactor Water Level +9 inches t
) Bearing Oil Pressure 3 psig i Pump Suction press 175 psig -;
, Condenser Vacuum 24' inches Hg j l !
l b. Reactor Water Level +38 inches l Bearing Oil Pressure 5 psig Pump Suction press' 220 psig Condenser Vacuum 27 inches Hg l
!
c. Reactor Water Level '47
+ inches :
Bearing Oil Pressure 5 psig I
, Pump Suction press 230 psig
- Condenser Vacuum 23 inches Hg
!-
- d. Reactor. Water Level -3 inches Bearing Oil Pressure 6 psig i Pump Suction press 235 psig cl Condenser Vacuum- 28 inches Hg l
t
)
.
.
. _ _
-_
i
O gr SENIOR REACTOR OPERATOR Page 22
,
QUESTION: 030 (1.00)
Unit 2 is in HOT SHUTDOWN with a reactor pressure of 805 psig and MSIVs ,
are closed. Operability Surveillances are performed on all of the MSL Radiation Monitoring System Channels 2D11-K603A through D. Channels A and D test UNSAT, while Channels B and C test SA Maintenance has no estinate of repair time and will not be able to commence troubleshooting and repairs for 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> Which ONE of the following actions describe the allowances and/or limitations imposed by the Technicr.1 Specifications in this instance?
a. No action required; function not required to be OPERABL Place one Trip System in the tripped condition within one hour; no additional action is require '
c. De in Cold Shutdown within the next 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> Place one Trip System in the tripped condition within one hour; be in Cold Shutdown within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
QUESTION: 031 (1.00)
From the drawing provided as Figure 1, which ONE of the following sets of contacts are closed under NORMAL conditions for the Vital AC Static Invertor?
a. B1, B2, and B3 B2 c. B2 and B3 B3 and B4
!
__ _ _ _ _ __ _ _ - _ __
I O O :
i i
'
SENIOR REACTOR OPERATOR Page 23
i !
-
QUESTION: 032 (1.00) :
i A complete loss of Unit 1 Service Air has occurre l 1- 'Which one of the following describes the effect this loss will have on [
the Puel Pool Transfer Canal inflatable seals? c
,. *
'
" Pneumatic pressure will be immediately lost to'only the inner gate seal i
b. Pneumatic pressure will be immediately lost to only the outer
. gate seal Pressure to neither seal will be immediately lost due to the air
- receivers that are available to supply air pressure to the
-
seal I Pressure to the seals will not be lost due to a backup nitrogen -
,
a bottle automatically supplying pressur l
!
i
!
QUESTION: 033 (1.00)
Which ONE of the following REOUIRED actions should be taken if feedwater ;
- conductivity cannot be maintained less than 0.2 umhos/cm?
i ; Operation may continue for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while efforts are ;
being made to return the conductivity to norma ,
..
' Perform a normal shutdown within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> if conductivity cannot ;
be restored to less than 0.1 umhos/c i i
,
c. Scram the Reactor and terminate feedwater injection if HPCI or ;
,
RCIC are availabl ;
-
i
'
d. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and COLD SHUTDOWN l within the next 24' hour :
!
-l i
'!
!
.;
i l
l i
!
j b
J l
-. -.
,
O O-
.
i SENIOR REACTOR OPERATOR Page 24' !
i l
l QUESTION: 034 (1.00) l
The Control Rod Reed Switches provide input to the Rod Position f Indicating System for use in the rod control system !
Which ONE of the following DOES NOT receive input from the Reed l Switches? i
!
a. Rod Worth Minimizer :
!
b. Reactor Manual Control System i
c. Process Computer d. Rod Block Monitor System
'
i
!
i i
QUESTION: 035 (1.00)
fi
.
Unit 2 is at rated power when the following indications occur:
[ Fluid Drive "A" Scoop Tube Lock annunciator alarns
> Speed Control "A" Signal Failure annunciator alarms j Feedwater flow decreases PSW Div I pressure indication is zero f Inverter 2R44-S002 Trouble annunciator alarms ;
- Which ONE of the following conditions would result in the listed i
- indications? -
i
.
a. A loss of Vital A b. A trip of the "A" Recirc Pump.
,
f
!
c. A recirc system runback to the #2 speed limi l
d. A loss of Instrument Bus "A" i f
!
!
d !
_
r i
I
.
h
- .- . - - ~ ~ . ._ . , . - . . -
'!'
-
O O i
!
!
SENIOR REACTOR OPERATOR Page 25
,
l
!
QUESTION: 036 (1.00) 1
!
WHICH ONE of the following conditions would be a Technical Specification Safety Limit violation?
a. Unit 2 is operating at 100% power. Reactor steam dome pressure increases to a maximum of 1100 psig without a scra The operator manually scrams and restores pressure to 920 ;
psi I i
b. Unit 1 is in Cold Shutdown with the RHR system operating in 3 Shutdown Cooling. The reactcr steam dome pressure increases ;
to 155 psig without occurrence of an isolatio ;
,
'
c. Unit 1 is operating at 22% pcwer end the EHC pressure ;
regulator fails. The reactor preosure drops to 820 psig before the reactor scrams on M31V closure.
d. Unit 2 is in Refueling. RPV level is lost due to an !
unisolable leak. Level drops to -167 inches before it is l
- restored to -15 inches with emergency system {
i :
i i
-
!
, QUESTION: 037 (1.00) !
'
l
-
The EOPs direct emergency depressurization of the RPV prior to entering i
[ the UNSAFE region of the HCLL due to low Suppression Pool leve j
- Which ONE of the following is the reason for this action?
4 i
'
a. Ensures that the RPV is depressurized prior to evacuating 99% of j
-
.
the non-condensibles to the Suppression Pool.
-
' .
b. Ensures the downcomers will not become uncovered causing loss of l
, pressure suppression on a LOC '
-!
c. Ensures that unstable steam condensation (chugging).does not occur in the downcomer during a LOC !
d. Ensures that excessive hydraulic stress is not-applied to the f SRV tailpipes and supports on a LOC t
,
l
!
I t
.
-
,. . _ . . _. . . . . _ . . . . . . . _ ,
I
&
w SENIOR REACTOR OPERATOR Page 26 QUESTION: 038 (1.00)
Which ONE of the following is the bases for maintaining the MINIMUM water level in the Spent Fuel Storage Pool per Unit 2 Technical Specifications? To keep the fuel below a Keff of 0.95 delta K/ To provide adequate shielding to personnel working in the are To remove iodine gap activity f ollowing a fuel rupture acciden To insure adequate heat removal from irradiated fuel.
QUESTION: 039 (1.00)
Which ONE of the following is the advantage of venting the primary containment through the suppression pool? Condenses any steam from the primary containmen Prevents excessive drywell-to-torus differential pressur c. Minimizes amount of radioactivity release d. Dilutes the explosive hydrogen concentration . - - .- . -. . _ - -.
O O i i
SENIOR REACTOR OPERATOR Page 27
,
f QUESTION: 040 (1.00) f; During testing of the Unit 2 Standby Gas Treatment' System (SBGT) !
Train "A", the following plant conditions exist: !
~f Reactor power 80% !
SBGT "A" control switch In STANDBY l
"
SBGT "A" Discharge _ Isolation Valve 2T48-F002A OPEN j SBGT "A" Suction Isolation Valve F076 CLOSED l Train "A" CR1 heater outlet temperature 305 deg F- !
!
Which ONE of the following is the response of the SBGT Train "A" !
following a high drywell pressure initiation? I
,
a. Will not start due to the control switch being in STANDB {
!
,
b. Will not start due to high heater outlet temperature.
j c. Will start after the Discharge Isolation Valve realign ,
q d. Will start and the Suction Isolation Valve will realig !
I i
"
1 I i
(1.00)
'
,
QUESTION: 041 During an ATWS, reactor power is 35%.
Which ONE of the following actions would be taken to MINIMIZE the heat addition to the suppression pool? .
l
. Boron injection is commenced if temperature approaches 110 F }
to ensure'that the Technical Specifications limits are not !
exceede If power incres.ses to greater than 45% then immediate tripping j of the recircu3ation pumps from their present speed is >
necessary to reduce heat added to the suppression poo If, due to a failure of the SLC system, injection of boron is .
delayed, then lowering of= level should be delayed until boron l is injected regardless of heat addition to the poo i
! Level should be reduced until EITHER level is at the top of i active fuel (TAF) OR power is less than 3% with suppression l
. pool _ temperature less than 110-F, AND all SRV's close !
.I i
, ;
.
.
- . , - - - - < =
. . .. - . _- . . - .
,
O O I
'
i, l
SENIOR REACTOR OPERATOR Page 28 I l
l
'
QUESTION: 042 (1.00) !
i Select the Emergency Classification that completes the following l statemen i
!
When Events are in progress or have occurred which involve actual OR l imminent substantial core degradation OR melting with the potential for ;
loss of containment integrity, the Emergency Director would declare i a(an) . l Unusual Event !
l Alert !
f f
' Site Area Emergency !
!
l General Emergency l i
d
- !
I i
4 !
a l QUESTION: 043 (1.00) j i !
In preparation for plant startup, the operator lining-up the Standby _ l
- Diesel Service Water (PSW) system informs the Shift Supervisor that a ;
valve is in the correct position but is listed incorrectly on the_ valve i checkoff shee ]
'
'
Which ONE of the following describes the action to be taken to make a ;
temporary change to the procedure? j i Only the Shift Supervisor needs to authorize this change, since it does not change the intent of the procedur ,
"
I b. Two members of management nmst approve the change, since this is ;
i a safety related procedur !
'
c. The change must be reviewed by PRB (can be done by phone) and approved by the applicable manager prior to implementatio ;
'
d. The valve should be repositioned with an * beside the desired position and a note sent to Document Control to initiate a i i
procedure chang j i
i
!
.
'
i 4 f
'
l r
!
!
F
- . _ _ _ _
. -- .-. ~ - _ - . - . . . - . .
!
O O :
!
!
!
SENIOR REACTOR OPERATOR Page 29 i-t
.
-
!
QUESTION: 044 (1.00) !
'
Which ONE of the following. Plant Hatch personnel makes the determination that Urgent Maintenance is required to maintain a schedule or plant j safety? l
'
!
a. Operation Supervisor on Shif e
,
b. Plant Support Manage i
'
!
c. Plant Manage !
d. Shift Supervisor on Duty -f i
i k
h
. C
!
. :
I e
I l
!
[
, i
!
f
!
!
!
,
,
!
i
!
!
,
-
-
i i
.!
l t
F
'
, t
& 4 v w SENIOR REACTOR OPERATOR Page 30 QUESTION: 045 (1.00)
During an entry into a HIGH RADIATION AREA, an operator will receive an estimated 70 mrem whole body dose. The following is the radiation exposure information available on the operators who are available for the task. Time constraints will not permit authorization of an increase in Plant Hatch administrative limit NRC Form 4s are on file unless otherwise indicate Operator 1 Operator 2 Operator 3 Operator 4 Sex Male Male Female Male Age 29 30 24 20 Wk/ Exposure 200 mrem 0 mrem 5 mrem 98 mrem Otr/ Exposure 1190 mrem 1960 mrem 435 mrem 420 mrem Lif e/ Exposure -
55370 mrem 2735 mrem 8970 mrem Remarks History NONE 3 months NONE Unavailable Pregnant-Signed Prenatal Document on File Which ONE of the following personnel should be selected for the job?
a. Operator 1 b. Operator 2 c. Operator 3 d. Operator 4
. . - . . _ ~ . - - - . . .~ . - o o l
!
SENIOR REACTOR OPERATOR Page 31 i
.
QUESTION: 046 (1.00) l l
Which ONE of the following procedure changes WOULD alter the intent of :
the procedure?
l a. The procedure is changed to indicate an authorized temporary ;
test gauge will be in place for three weeks while the normal '
pressure instrument is in the shop for repair i t b. The procedure is changed so that when the APRMs are !
recalibrated, the neutron trip setpoint would be lower than that i required by Technical Specification j c The procedure is changed so that shielding is removed in the f vicinity of the RWCU Cleanup Filter /Demins which will result in i a higher radiation dose to the PEO stationed in the are d. The procedure is changed such that a trip setpoint listed in the !
l procedure is altered to agree with the vendor's technical ;
- manua j i
!
QUESTION: 047 (1.00)
- Which ONE of the following IS NOT a valid reason for exiting an EOP flow chart?
I j_ a. The EOP flow charts have directed exiting to a normal operating [
procedur '
J M
'
b. All entry conditions for that flow chart have cleared and an ;
emergency no longer exist ;
i
'
i c. An entry condition to another EOP chart with a higher priority
has occurre d. An entry condition exists, but-the Shift Supervisor has determined that an emergency no longer exist !
r
!
l
i
!
n
l t
r
!
- - - . . . - - - _ . - _ _ _ . _ _ . _ _ _ - - _
.. . _
. . - . . _ . _ _ m - _. _ . _
O O :
SENIOR REACTOR OPERATOR Page 3 QUESTION: 048 (1.00) ,
FILL'IN THE BLANKS The Fire Brigade must consist of a minimum of qualified ,
persons. A minimum of of these persons one must have competent !
knowledge of safety related systems and component , 2
- , 5 l t , 3 i
! , 4 I
,
I t
QUESTION: 049 (1.00) !
!
Following a core damage accident, gaseous effluent is released at the .
plant. At the site boundary the whole body radiation dose is measured at l 620 mR/hr for 45 seconds and then drops rapidly to 45 mR/hr. By the end I'
of the following 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> the dose rate drops to background level i Which ONE of the following is the lowest level Emergency Classification for this event? (73EP-EIP-001-OS is enclosed as attachment 1 for i reference) ! Unusual Event i Alert t I
3 Site Area Emergency i
! General Emergency I
N a
i n
!
,
j
'
!
i
-,
.. . , . - , c
.. _ _ _ __ _ _ . _ _ __-_ . ._ _ ._._. ___ _
,
!
O O i
<
t SENIOR REACTOR OPERATOR Page 33 }
i :
!
!
i QUESTION: 050 (1.00) [
!
"
i A reactor startup is being performed on Unit 2 with the Reactor Mode -!
Switch in START / HOT STANDBY and Reactor Pressure at 700 psi .
A review of the ADS LSFT 42SV-B21-003-2S surveillance data package i reveals the "E" ADS valve is NOT acceptable. HPCI CSS and LPCI j surveillances are current and not require ,
Which ONE of the following actions should be taken by the Shift ;
Supervisor? ;
a. Enter a 14 day LCO and direct the Unit 2 Startup to be stopped !
prior to entering the RUN mod ,
! Enter a 14 day LCO and continue the Unit 2 Startup to rated 3 powe '
j
- Direct the Unit 2 Startup to be continued because NO LCO needs
to be entered for present plant condition Initiate a reactor shutdown to HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> j i
i
_ QUESTION: 051 (1.00) ,
,
, :
A valve has been tagged closed for-a clearance on a safety related f,
,
. syste Independent verification of the valve position is require j
~ Health Physics has determined that there is a high radiation level in- !
the area of the valv Which ONE of the following describes the disposition of the independent ,
' verification of the valve? 3 i-i
, a. Verification can be waived by the Shift Supervisor, bu i
- alternate means of verification should be considere !
! !!
l b. Verification can be waived by the Shift Supervisor, but must be i visually checked prior to commencing wor I c. Verification can be waived only if approved by.the Operations l Department Manager when visual exposure results in excessive j exposur ;
l d. Verification can be waived only by the Operations Department l Manager if'an. alternate means of verification is availabl .! 4 I
o i
.. . . . .
S &
(v w SENIOR REACTOR OPERATOR Page 34 QUESTION: 052 (1.00)
An event has occurred on unit 2 resulting in an entry into PC-1, Primary Containment Control. The following plant conditions exist:
,
Drywell temperature 122 Deg F Suppression Pool Water Level 148 inches Suppression Pool temperature 105 Deg F Drywell pressure 1.1 psig Which ONE of the following is the required IMMEDIATE action if the Suppression Pool water level increases to 151 inches? Return to that key parameter, readdress, and continue from ther Ignore the change since subsequent actions have insured plant safet , Return to top of flow chart and begin to follow all flow paths agai Go to End Path Manual and follow procedure for that paramete QUESTION: 053 (1.00)
A reduction in recirculation pump speed has occurred while operating at 100% power causing the Region of Potential Instabilities to be entere Which ONE of the following actions would be taken based on the given nuclear instrumentation respor.se? (Consider each case independently.) IF three LPRMs in one quadrant are oscillating at an 8%
bandwidth, rods should be inserted in that quadrant to suppress the oscillation If one APRM is oscillating at a 12% bandwidth with the others ;
oscillating at an 8% bandwidth, recirculation flow should be !
increased to suppress oscillation j If two LPRMs in opposite quadrants are oscillating at an 8%
bandwidth, the reactor should be scramme l If all observed LPRMS are oscillating at a 12% bandwidth, the reactor should be scramme .- - . - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l
& A w w l
SENIOR REACTOR OPERATOR Page 35 l
QUESTION: 054 (1.00)
An ATWS has occurred on Unit 2 with a steam line isolation. To reopen the MSIVs, EOP 31EOP-111-2S, Emergency Opening of MSIVs, directs the operator to override the high flow isolation on 2P70-F004 and 2P70-F00 Which ONE of the following is the purpose f or this override?
a. Restores operating pressure to the outboard MSIVs to allow opening the valve b. Permits resetting the main steam line flow isolation to allow opening the MSIV c. Restores operating pressure to the inboard MSIVs to allow opening the valve d. Permits resetting the RPV low water level isolation to allow opening the MSIVs.
QUESTION: 055 (1.00)
Unit 2 is at power when a complete loss of the CRD system occurs. The operator has been unable to restore charging water pressur Which ONE of the following describes the conditions and actions which should be taken in accordance with 34AB-C11-001-2S, Loss of CRD System? Commence a FAST REACTOR SHUTDOWN if RPV pressure is LESS than 800 psig and more than four accumulator trouble lights are illuminate Commence a FAST REACTOR SHUTDOWN if RPV pressure is GREATER than 800 psig and four accumulator trouble lights are illuminate c. Manually SCRAM the reactor if RPV pressure is LESS than 800 psig and four accumulator trouble lights are illuminate d. Manually SCRAM the reactor if RPV pressure is GREATER than 800 psig and more than four accumulator trouble lights are illuminate &
w nw SENIOR REACTOR OPERATOR Page 36 QUESTION: 056 (I.00)
Which ONE of the following describes the significance of a Primary Containment level of 83.9 feet during Primary Containment Flooding? The top of active fuel in the RP The structural load limit of the containment wall The highest containment vent capable of rejecting all core decay hea The bottom of the main steam lines.
QUESTION: 057 (1.00)
After a transient Unit 2 RPV level is being maintained at -150 inches with RCIC as the only high pressure injection syste Which ONE of the following statements describes RCIC use for pressure control? RCIC MAY NOT be used because defeating the suction swap is not permitte RCIC MAY NOT be used because there is an initiation signal presen RCIC MAY be used for pressure control after realigning the suction to the CS d. RCIC MAY be used for pressure control after overriding the CST test isolation valve interlock _ . _ , . _ _ __ -- . . _ _ _ - _ _ . __. . .=
J O O !
a
'
SENIOR REACTOR OPERATOR Page 37 l
!
QUESTION: 058 (1.00) [
'
Unit 2 Reactor Building Closed Cooling Water (RBCCW) System is operating t with pumps 2A and 2B in service and pump 2C in standby. The feeder ,
breaker to 600V bus 2D fails causing the bus to deenergiz !
Which ONE of the following will be the (RBCCW) system response? ?
, Pump 2A will trip off, 2C will fail to start and pump 2B will i continue ru !
! Pump 2B will trip off and pump 2C will start on low header !
pressur l i . i
- Pump 2A and 2B will trip off and pump 2C will start on low l
'
header pressur !
' Pump 2A will trip off and pump 2C will be locked out on an l undervoltage tri t
.
!
l i
,
QUESTION: 059 (1.00)
.
Unit 2 is at rated power when a loss of Instrument and Service air ,
occur ;
I Which ONE of the following indications will require a reactor scram after the loss of air?
l E l
, Indication of 1 control rod drifting in coincident with a CRD !
1 HYD HIGH TEMP annunciator, i Indication of 3 control rods drifting i i
j Scram pilot air pressure of 46 psi ,
t
- SCRAM VLV PILOT AIR HEADER PRES 6URE HIGH/ LOW annunciator is !
receive l
i I
F
!
j j i 4 !
!
i
- - . .- _
. . _ _ _ _ _ _ _ _ _ . _ - - - _ _ _ _ _ - _ - - _ _ _ _ - _ _ _ _ _ - _ _ - - _ _ - -
.- . . . - - .- . - -
,
!
O O l i
.
l l SENIOR REACTOR OPERATOR Page 38 f
t
, ;
i
"
QUESTION: 060 (1.00) i
!
Which ONE of the following is the definition of " Minimum Alternate RPV j Flooding Pressure" (MARPVFP)? j
- The lowest pressure at which steam flow through the SRVs will }
remove the decay heat generated'10 minutes after a scram from full powe b. The lowest pressure at which steam generation needed for steam I cooling will not be suppresse j l c. The lowest pressure at which steam flow through the SRVs will f
,
preclude clad temperature from exceeding 2200 deg j d. The lowest pressure at which steam flow through the SRVs will
,
preclude clad temperature from exceeding 1500 deg :
i l
!
QUESTION: 061 (1.00) i
!
Which ONE of the following level instruments is reliable for the given l drywell temperatur !
>
Instrument Reading RTD 2T46-N001K RTD 2T47-N001A
, ;
L B21-R610 -295" 288 deg F 286 deg F :
i 1 C32-R606C +2" 280 deg F 271 deg F !
!
- B21-R605' +37" 255 deg F 260 deg F .f
.
' B21-R604B -144 198 deg F 207 deg F l i
! l-
-I
,
< t a
i
!
. i
, t 4 ,
?
E !
b
!
i
!
i
'
f
,.
b v
O w
SENIOR REACTOR OPERATOR Page 39 QUESTION: 062 (1.00)
A transient has occurred in which the reactor scrammed at 2237 hours0.0259 days <br />0.621 hours <br />0.0037 weeks <br />8.511785e-4 months <br /> and all RPV level indication became erroneous. RPV Flooding has been initiated. At 2258 a 50 psid differential was established between RPV pressure and Suppression Chamber pressure with 5 SRVs ope Which ONE of the following is the earliest time that core coverage can be assured? hours hours hours hours
,
QUESTION: 063 (1. 0 0)
While restoring RPV water level following RPV Flooding, jumpers are installed for the RHR Torus Spray Valves (2E11-F02 8A/B) .
Which ONE of the following is the reason that the jumpers are installed? To establish torus spray capabilit To establish suppression pool coolin To override the LOCA isolation signa To drain the RPV into the suppression poo ,
f l
!
l l
I
. _ _ . . - .- . _ . - .~ . . -
i i
O O !
l; I
SENIOR REACTOR OPERATOR Page 40
!
t i
i i QUESTION: 064 (1.00) ;
l Which ONE of the following statements explains the reason for inhibiting l ADS if boron is injecting into the RPV? ;
!
'
a. ADS actuation could slow-the rate of boron injection into the core due to the additional two phase flow resistanc !
t
- b. ADS actuation results in void formation which reduces control ,
'
rod worth and lowers shutdown margin.
' '
,
c. ADS actuation results in boron dilution by carry over into the i Torus causing a power excursio ;
r L
d. ADS actuation could result in the rapid addition of-cold l unborated water to the reactor by low pressure injectio j J i
I QUESTION: 065 (1.00) j 1 Which ONE of the following is a reason that drywell sprays may be !
initiated IRRESPECTIVE of adequate core cooling per "PC - Primary [
Containment Control"?
l To prevent excessive drywell-to-torus differential pressure t i j
,
l prevent f ailure of the drywell floo i To reduce drywell pressure when above 13 psig and suppression pool water level is above the vacuum breaker l I To prevent tripping tlue RCIC and HPCI turbines due to elevated :
'
Suppression Chamber pressure.
'
! To dilute drywell hydrogen and oxygen concentrations to prevent a containment explosio j i
!
i
6
' ;
i i
!
!
)
. , - - ,-
- . . . . - . . - . . . ~ . . - - - - _ . - _ - . . . . . _ . . . . . . - . . . - . .
[
o o
i
! ' SENIOR REACTOR OPERATOR Page 41
!'
!
i i
l l QUESTION: 066 (1.00)
-An event has occurred on Unit 2 resulting in the following plant
!- conditions: !
'
! !
3 Reactor pressure 1020 psig
'
Main Steam Tunnel temperature 160 deg F Suppression Chamber air temperature 132 deg F i RWCU Equipment Room temperature 122 deg F j Drywell Pressure 1.95 psig
!' Which ONE of the following sets of valves should have isolated?
I i Reactor Water Sample Line Valves l i
l Suppression Chamber Vent Valves I
1
! HPCI Steam Supply Valves
! .
f RWCU Suction Supply Valves I
!
,
y
!
l -
i
.
!
I -
t l. -
M i -
s
!
I I
s l .'
e
l l
.
!.
<
I
- - - .-,.--,.. --. . ..._.-. - -
_ . __ _ - . _ . _ _ . _ _ _ . . - - _ . _ _ _ _ _ . . . - . - _
l i O '
SENIOR REACTOR OPERATOR Page 42 I
,
,
QUESTION: 067 (2.00)
'
All Unit 2 Service Air Compressors have tripped and system air pressure !
,
is decreasin For the actions listed in Column A SELECT the I appropriate air pressure set point from Column (Items in Column B i
. may be used once, more than once, or not at all. Only one answer may l l occupy a space in Column A) (4 required at !
0.5 each) !
!
'
.,
CCLUMN A COLUMN B (Action) (Setpoint)
,
a. When Turbine Building Instrument Air psig -j
,
pressure drops to , the standby t
prefilter and afterfilter are psig !
,,
'
automatically placed in servic l psig j b. When Control Air Pressure drops to , Startup LCV, 2N21-F111 locks psig up in its existing positio , psig !
'
c. When Non-interruptable air pressure ;
drops to , Reactor Building psig i j Instrument Nitrogen to Non-Interruptable l 1 Air Isolation Valve 2PS2-F565 open . 60 psig ;
- 6 i i
'
d. When Service Air Pressure drops psig !
to , Service Air Isolation Valve !
'
2P51-F017 close . 50 psig d
,
i QUESTION: 068 (1.00) .;
{
i
'
l The plant is operating normally at power when Recirculation Pump A Seal
4 ;'
Stage Flow alarms low and the operator notes No. 2 Recirc Pump seal
' pressure decreasing towards zero with No. 1 seal pressure stable at 1000 psi Which ONE of the following failures would cause these indications?
(No other alarns are present. Figure 2 is enclosed for reference.). .
t I
a. Failure of No. 1 seal ;
!
b. Failure of No. 2 seal !
l l 1 Plugging of the No. 1 internal restricting / breakdown orifice
. Plugging of the No. 2 internal restricting / breakdown orifice .,
l
!
!
-- ._ . . ___ . , - . -.
. . _ _ . _ _ .. _ . __ _ _- . - _ . _
l O O 1
.
t
, SENIOR REACTOR OPERATOR Page 43 l l
i !
J :
'
.
l QUESTION: 069
(1.00) {
I i
,i A Unit 2 reactor startup is in progress with the Plant Operator ;
withdrawing Control Rods in Rod Sequence Step 9 which has the following j limits: I
!
. >
, Insert 08 }
- ) Withdraw 12 i i
!
Several Rod Worth Minimizer (;RWM) Group 9 Control Rods are still at l l position 08 when the operator inadvertently withdraws a Group 9 rod to ;
i position 18.
Which ONE of the following describes the operational status of the Roc l
'
- Worth Minimizer System and the corrective action to be taken? (No RWM i lights or alarms are present.) ;
1 .
a. The RWM is INOPERABLE and a second licensed operator or I l
1 qualified member of the tech staff should be stationed at the
! P603 panel to double verify any further rod movemen l
"
, b. The RWM is INOPERABLE; all Control Rod movement should be l
'
terminated and the reactor manually scrammed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> unless the RWM is made operabl j c. The RWM is INOPERABLE and the RWM Group 9 Control Rod withdrawn :
j to position 18 should be immediately inserted back to within its j
- withdrawal limit before any other Control Rod movement is !
i attempte {
\ l l The RWM is INOPERABLE and the RWM Group 9 Control Rods should be l inserted to their insert limit before any other Control Rod j movement is attempte !
!
I
'
t H t
>
I
- l
l s
' ,
, !
- -i
,
I (.
__ _ . . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ .
O
-
SENIOR REACTOR OPERATOR Page 44 QUESTION: 070 (1.00)
DCN 90-066 installed a 3/8 inch bypass line around the EGR actuator for the HPCI turbine sper-d contro Which CNE of the following is the reason for installing the bypass line? Reduces the peak speed of the HPCI Turbine on startup to prevent tripping on overspee Keeps the Governor Valve closed until the Stop Valve is fully opened to reduce startup vibratio , Removes the Ramp Generator from the speed control during manual operation to allow faster speed increas d. Maintains the Governor Valve at approximately 50% open until the stop valve begins opening to prevent Ramp Generator oscillations.
_ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ . _ _ _
. .
.
e
-
A w
SENIOR REACTOR OPERATOR Page 45 QUESTION: 071 (2.00)
For the components of the Control Rod Drive (CRD) system labeled a through d on Figure 3 make the proper identification in Column A from the list of components in Column B. (Components listed in Column B will only be used once and only one answer may occupy a space on Column A.)
(4 answers required, 0.5 each)
COLUMN A COLUMN B (Components) (Component Identification) . Drive Water Pressure Control Valv l I . Stabilizing Valv . Flow Control Valv l Return Line Pressure Control Valv . Cooling Water Pressure Control Valve j Suction Filte l Drive Water Filte . Flow Elemen I i
. -- . . _ - _ . _ -- .. . . .
I O O !
!
.l I
' SENIOR REACTOR OPERATOR Page 46' l
-
l i
I QUESTION: 072 (1.00)
. Both RFPTs have tripped with a failure of HPCI to start resulting.in a
~
significant reactor water level' transien The following plant conditions currently exist:
The reactor has scrammed All rods in i Reactor water level -95 inches decreasing l Reactor pressure 987 psig decreasing j Drywell pressure 1.67 psig steady j Torus water temperature 96 deg F j
An operator observes the Core Spray System is runnin Which ONE of the following describes the correct status and corrective action to be taken for the Core Spray System?
. I a. The Core Spray System should have initiated on high-drywell i pressure and should be allowed to run
-i b. The Core Spray System should have initiated on low reactor water ;
'
level and should be allowed to run '!
I c..The Core Spray System should NOT have initiated but can be i allowed to run until water level is restore s
!
'd. The Core Spray System should NOT have initiated but can be used !
for pressure contro >
l l
QUESTION: 073 (1.00) !
!
A caution in the procedure for cooling in the steam condensing mode with the RHR Heat Exchangers states "A pressure of 75 psig on the.RCIC pump '
i suction must not be exceeded during the steam condensing mode of l operatio !
.
'Which ONE of the following is the reason for maintaining this pressur !
'
below 75 psig?
a. Could cause failure of the RCIC suction seal l b. Could result in exceeding the RCIC design pressure discharge limi ~:
i c. Could result in over pressurization of the RCIC Lube Oil Cooling
. System d. Could result in decreasing the RHR heat exchanger level too-rapidl ;
l l-l
_ . _ _ . ,_ _ _ _ _ _ . . . _ - - - _ . _ . = _ . - .. _ .. . .. ,
,
!
O O l
!
?
SENIOR REACTOR OPERATOR Page 47_ l
l
'l
QUESTION: 074 (1.00) {
i A transient has occurred on Unit 2, resulting in the following 'j condition i Reactor Power All rods in RPV Pressure 950 psig i
'
Reactor Water Level +30 inches and stable l Drywell Pressure 1.2 psig j Drywell Temperature 160 deg F (Bulk temperature) l Torus Level +46 inches !
!
4160 VAC buses "A" and "B" are deenergize !
'
MSIV's are close The Primary System is discharging into the reactor building with r the following radiation levels: ;
)
,
158' level are mR/hr l 130' N.E. working area 1300 mR/hr l
!
Which ONE of the following methods would the SS direct to Emergency Depressurize the reactor?
i I
! a. Opens 7 SRV' Place HPCI in full flow tes !
!
c. Open the steam.line drain ,
3 d. Equalize around the MSIV's and use the MSIV's with the Bypass l Jac f I
l
,
)
l l t
l t
'
.
]
.,.
r i
i f
n v
a
-
SENIOR REACTOR OPERATOR Page 48 QUESTION: 075 (1.00)
Unit 2 has experienced a loss of Main Condenser Vacuum resulting in MSIV closure and a reactor scram. The following plant conditions exist:
Control rods Fully inserte HPCI system Inoperabl RPV water level -10 inches and slowly increasin Reactor pressure 905 psi '
While operating RCIC to restore RPV water level, a steam leak develops on the Trip and Throttle Valve, but RCIC continues to operat RCIC diagonal ambient 195 deg F temperature Area radiation 55 mr/hr Which ONE of the following is the action the SS would direct? Operate RCIC area coolers until area temperature or radiation levels exceed Max Safe operating value, then isolate RCIC and use Core Spray to restore RPV water leve Continue operating RCIC to restore RPV water level until water level is greater than +32 inche Immediately isolate RCIC and use CRD to restore RPV water level to between +12.5 and +51.5 inche Immediately isolate RCIC and depressurize the plant to use low pressare systems to restore water level per CP-1 Alternate Level Contro l l
!
l l
,
I l
i
.. - .. . -- . -_ _ .. - .- _ -
!
O O lt
!
SENIOR REACTOR OPERATOR Page 49 ]
f i
!
t QUESTION: 076 (1.00) ] l CP-3 ATWS Level Control has been implemented to the point of " SLOWLY I raise injection to restore and maintain RPV water level above - 185 l inches." !
Which ONE of the following describes the action (s) to be taken to SLOWLY
,
raise injection?
a. Start one low pressure pump at minimum flow and gradually l increase flow until RPV level begins to increas i b. Start one low pressure pump, immediately increase flow to rated !
} and start additional pumps sequentially at rated flow until {
n level begins to increas p I
c. Start two low pressure pumps at maximum flow until level begins .;
. to increase, then reduce flow to less than 5000 gp j
!
,
d. Start two low pressure pumps at minimum flow, increase flow to !
] 5000 gpm per pump and start additional-pumps if required.- (
l :
J !
,
q QUESTION: 077 (1.00) l t
-
'
Restarting Turbine Building HVAC during a rad release above an ALERT- !
Emergency insures that: l
-
J -
a. Any release to the Turbine Building is exhausted through SBG [
,
>
b. Accessibility to the Turbine Building is preserve I
- i j c. The Turbine Building Blow Out Panels do not blow ou ;
i i d. Release to the Turbine Building is contained in the Turbine 1
- Buildin l
I a f i
!
!
l
" i l
'
t
-
i i
,
. - - - . _ _ _ _ _ _ _ _ - - - - - - _ _ _ . - - _ _ . _ _ - . _ . _ _ - - _ _ _ - _ _ - _ - _
. . . - . . .-. - - -. . .- --. ~.- . .
!
O O !
t
'[
!
SENIOR REACTOR OPERATOR Page 5 l ( '
t i
I QUESTION: 078 (1.00) l
,
. i The following conditions exist after a Unit 2 scram from 100% power: {
Highest Drywell temperature 210 deg F Drywell pressure 1.3 psig !
Reactor pressure 920 psig ;
'
-i
,
The following water level instruments read as indicated: l i
Floodup Range, 2B21-R605 +20 inches f Narrow Range, 2C32-R606A +6 inches i
Wide Range, 2B21-R604A +5 inches [
Fuel Zone, 2B21-R610 -20-inches I l
Which ONE of the following Reactor Water Level indicators WOULD NOT be !
reliable for level trend information in accordance with the EOPs? {
I a. Floodup Range, 2B21-R605 !
I b. Narrow Range, 2C32-R606A j
!
c. Wide Range, 2B21-R604A l
r
.
- d. Fuel Zone, 2N21-R610 .j
!
I i-i QUESTION: 079 (1.00) i i
Unit 2 is operating at 75% RTP when SRV 2B21-F013A inadvertently open !
,'
Suppression Pool temperature increases to 115 deg F before the operator !
pulls'the fuses for the valve and it close :
'
Which ONE of the following is the appropriate IMMEDIATE action to be-
"
taken in accordance with plant procedures and the Unit 2 Technical !
Specifications? !
'
!
a. Place the Mode Switch to SHUTDOW !
b. Reduce Suppression Pool Temperature to less than 100 deg ;
within the next four hours I
c. Initiate a fast reactor shutdow I
!
? d. Place the plant in the HOT SHUTDOWN within 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> and l COLD SHUTDOWN within the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> !
i
[
l
. _ -
- - - - - - . _-
o o 1I i
i SENIOR REACTOR OPERATOR Page 51 ;
'i OUESTION: 080 (1.00) !
A loss of the 2C and 2D 4160 VAC switchgear occurs'following a main j turbine tri .
Which ONE of the following describes the operation of the' Unit 2 Reactor j Recirculation MG set Lube Oil system?
a. The DC oil pump for each Reactor Recirculation MG set will auto- j start, supply oil to the motor and the generato ,
b. The running AC oil pump for each Reactor Recirculation MG set j
will continue running from its invertor's backup power suppl ;
c. The A3 and B3 AC Reactor Recirculation MG set oil pumps will ft auto-start supplying oil to the fluid drive bearings.
. d. The DC oil pump for each Reactor Recirculation MG set will auto- l
'
start and supply oil to the fluid drive bearing ;
I
i
- . f QUESTION
- 081 (1.00) j
' !
A pipe break in the Unit 2 Reactor Building resulting in steam tunnel :
temperatures greater than Max Normal and an automatic shutdown of the i Reactor Building HVAC system. The leak has now been isolate !
-Which CNE of-the following is a condition'that has to be met prior to {
restarting the HVAC system? :
i i t
a. Drywell pressure nmst be below 1.85 psi l
b. The reactor building potential contamination area ventilation exhaust radiation level must be below 9.5 mr/h I c. RPV water level must be greater than 0 inche '
. d. Not more than one area above the Maximum Nornal Operating ,
temperatur j
!
!
f r
h
i i
!
r h
!
t l
!
'
&
w SENIOR REACTOR OPERATOR Page 52 QUESTION: 082 (1.00)
An event has occurred on Unit 2 with the plant operating at 97% RT The following conditions exist:
RE SE Diagonal Area Floor Drain Sump High Alarm RB SE Diagonal Area Instrument Sump High Alarm Southeast Diagonal Area Radiation Level 9 mr/hr Southeast Diagonal Area Ambient Temperature 92 deg F Which ONE of the following describes the action (s) which should be taken by the crew and the reason f or the action (s) ? Entry into the SC flow chart is NOT required because no entry conditions exis b. Entry into the SC flow chart is required because sump levels arc abnorma c. Entry into RC and SC flow charts is required because area radiation levels are above max normal operating leve Entry into RC and SC flow charts is required because SE Reactor Building diagonal water level is above max normal operating leve ,
& /
W SENIOR REACTOR OPERATOR Page 53 QUESTION: 083 (1.00)
An event has occurred on Unit 2 with the plant operating at 68% RT The following plant conditions exist:
Steam Tunnel Area temperature 145 deg F 130' Northwest Area Radiation 1100 mr/hr 130' Southwest Area Radiation 820 mr/hr It has been reported that steam is leaking from one of the main steam line Which ONE of the following is the action that would be taken based on the plant conditions?
a. The reactor must be shutdown per 34GO-OPS-013-2S, Normal Shutdow b. The reactor must be shutdown per 34GO-OPS-14-2S, Fast Reactor Shutdow c. The reactor must be scramme d. The reactor must be scrammed and Emergency Depressurization initiated.
QUESTION: 084 (1.00)
At the completion of an " EMERGENCY IN" rod movement, a " ROD DRIFT" alarm is receive Which ONE of the following describes the reason for this alarm?
a. The rod is not at an even reed switch and none of the selected relay busses are energized (insert, withdraw, or settle).
b. The rod is not at an odd reed switch and none of the selected relay busses are energized (insert, withdraw, or settle).
c. The sequence timer is bypassed causing an insert and withdraw signal at the same tim d. " EMERGENCY IN" bypasses the Rod Position Indication Syste O O .
!
i SENIOR REACTOR OPERATOR Page-54 QUESTION: 085 (1.00) >
The " Reactor Recirc Loop B Out of Service" annunciator on P602 failed to reset after the operator completed the startup of Recirc Pump 2B. The annunciator has been in the alarm condition for the last two (2) shift ,
An MWO has been initiated and approved to repair the annunciato !
Which ONE of the following describes the status of the annunciator?
a. The annunciator IS CONSIDERED a Problem Annunciator because it has been lit (cannot be reset) for more than one shift, b. The annunciator IS CONSIDERED a Problem Annunciator because an Engineering Evaluation has not been complete The annunciator IS NOT CONSIDERED a Problem Annunciator because it has not been deactivate ,
i d. The annunciator IS NOT CONSIDERED a Problem Annunciator because '
it cannot bend deactivated (reset) and is therefore no distraction to the operator
-
.
!
t QUESTION: 086 (1.00)
Which ONE of the following is the reason for preventing Primary [
Containment water level from exceeding 103.5 feet during Primary i
'
Containment flooding?
!
'
a. This level is equal to the top of active tuel in the reactor vesse b. This level exerts pressure equal to the structural design load ,
limit of the containment wall !
!
c. This level is the elevation of the highest containment vent l'
capable of rejecting all core decay hea d. This level is the elevation of SRV pilot valve solenoid i operator l
4
_-- -
_ _ -
_ ..
.. - _ _ _ . _ - _ . . _ _ . _ _ _ . . . _ -
- - _ _-. . ._
O O !
O'
SENIOR REACTOR. OPERATO Page 55 j i
l i
QUESTION: 087 (1.00) !
!
e Fuel movemer is'in progress on Unit 2.with the following plant !
conditions: l
!
MODE COOLANT REACTOR i SWITCH POSITION TtMPERATURE POWER !
.I Unit 1 RUN 545 deg F 80% I Unit 2 REFUEL 128 deg F 0% !
!
Which ONE of the following is the MINIMUM. shift staffing requ1 red by the .;
Unit 2 Technical Specifications? l I SROs 2 + 1 for Fuel Handling Ros 3 :
NLOs 3 -;
STA 1 !
i b. SROs 1 + 1 for Fuel Handling ;
,
.STA 1 c. SROs 1+ 1 for Fuel Handling f ROs 2 i
'
NLos 3 STA NONE
'
t d. SROs 2+1 for Fuel Handling (
ROs 2 4 NL0s 2 l
'
STA NONE !
,
'
i t
!
I t
!
!
!
't i
!
!
!
';
i l
- _ _ - . - _ . . - . .
!
&
w (O v
SENIOR REACTOR OPERATOR Page 56 QUESTION: 088 (1.00)
A Temporary Release (TR) to a clearance has been written and pldced in the TR boo Subsequently a Maintenance Foreman requests an additional subclearance on this clearance from the Shift Superviso Which ONE of the following actions should be taken by the Shift Supervisor?
a. Direct Foreman to obtain the current subclearance holders permission and then proceed with his subclearance, b. Authorize the subclearance after obtaining the approval of all other subclearance holder Direct the Foreman to work his job under another supervisors subclearanc Deny the Foreman a subclearance until the clearance is removed from the TR book.
QUESTION: 089 (1.00)
Which ONE of the following IS NOT in conformance with 30AC-OPS-001-OS, Control of Equipment Clearances and Tags? When pulling fuses to deenergize power supplies, the DANGER tags will be attached to the blockout fuse When danger taggfay- an MOV as a clearance boundary, tagging the breaker and local aperator will satisfy the tagging requirement It is permissible to repack a MOV with the valve on its backseat and a DANGER tag on its control switch, breaker, and local operato When danger tagging an MOV which requires dependent verification it is permissible for the in(2 pendent verifier to use the valve position indication lights to determine valve positio _ _ _ - - - . , . . .. - - , -
!
O O -
i
.
SENIOR REACTOR OPERATOR Page 57 l
! !
t QUESTION: 090 (1.00) [
,
Which ONE of the following describes the method for performing a valve l lineup check of a locked throttle valve? *
I a. Unlock the valve, turn the handwheel 1/4 turn in the closed i e direction, then 1/4 turn in the open direction, and relock the (
valv j b. Unlock the valve, turn the handwheel 1/4 turn in the open
, direction, then 1/4 turn in the closed direction, and relock the valv c. Verify the position of the valve by checking the valve position ,
indicato '
d. Verify the operability of the attached throttle valve locking ;
devic i
! )
"
!
. i i !
i !
(2.00)
"
QUESTION: 091 t
.i For the components of the OFF GAS SYSTEM labeled "a" through' "d" on Figure 4 l make the proper identification in Column A from the list of components in l Column B. (Components listed in Column B will only be used once and only_one j answer may occupy a space in Column A) (4 answers required, 0.5'each) .l t
I COLUMN A COLUMN B (Items) (Components) l . Charcoal Absorbers j i . Off-Gas Condenser l c 3 Intercondenser !
!
'
t
, . Water Separator i
- Cooler Condenser ;
i Holdup Pipe f
- Catalytic Recombiner l t Off-Gas Preheater f i Moisture Separator j
,
!
!
I t
!
.-,
!
.. . .. - -.- . . - . .. -_ . -. . . _
..,
i O O !
!
,
'
- SENIOR REACTOR OPERATOR Page_58-i I
!
QUESTION: 092 (1.00) l
Jul ATWS requiring boron injection has occurred-on Unit 2. An-unisolable steam leak into the RCIC diagonal area has resulted in RCIC isolation ,
and Secondary Containment temperature above Max Safe in more than one !
are The Shift Supervisor has determined that Emergency -!
Depressurization is require !
Which ONE of the following actions should be directed prior to emergency l depressurization? ;
a. Terminate and prevent injection from all injection system ,
including CR !
b. Terminate and prevent injection from Core Spray onl f f
c. Terminate and prevent injection from RHR and Core Spray onl j
'
d. Terminate and prevent injection from'all injection systems except boron, CRD and RCIC ;
e
!
!
!
QUESTION: 093 (1.00) l
Which ONE of the following conditions would require Emergency Depressurization of the reactor in accordance with the Emergency Operating Procedure? !
,
a. Ruptured instrument leg is releasing in excess of 1500 mr/hr to l
'
the Primary Containment and the Primary Containment cannot be isolated. SGBT is operatin ?
b. Offsite release rate is 1200 mr/hr and a main steam line break !
in the Tbrbine Building cannot be isolate Turbine building ,
ventilation is operatin .;
!
c. Offsite release rate is 1000 mr/hr and an unisolable RBCCW j
!
break is discharging outside the primary and secondary '
containmen :
d. All the 130 foot elevation radiation levels are in excess of l 500 mr/hr I
!
I l
i !
'k
,
)
l i
-
!
!
.
, - . ,
_-- . _ _ _
.__
O O SENIOR REACTOR OPERATOR Page 59 1
,
l
!
QUESTION: 094 (2.00) I
For the components of the ARI/ SCRAM AIR HEADER system labeled "a" '
through "d" on Figure 5 make the proper identification in Column A from j the list of components in Column B. (Components listed in Column B will ,
only be used once and only one answer may occupy a space in Column A) (4 answers required, 0.5 each)
l COLUMN A COLUMN B !
(Components) (Component Identification)
I . Scram Outlet Valve i i
" . ARI Valve . Scram Inlet Valve ,
i . Discharge Volume Drain Valve .
! Discharge Volume Vent Valve Backup Scram Valve j Scram Air Header Vent Valve
,
QUESTION: 095 (1.00)
Which ONE of the following describes the results of a loss of a :
Feedwater Heater Train? ;
, Core inlet subcooling decreases ;
'
Actual bundle power decreases Critical power ratio decreases Core inlet subcooling increases i Actual bundle power decreases Critical power ratio increases !
l Core inlet subcooling decreases 7 Actual bundle power increases .
Critical power ratio increases r
i Core inlet subcooling increases !
Actual bundle power increases Critical power ratio decreases
!
!
(********** END OF EXAMINATION **********)
.
., .- . - - . . . . . . _ . - ~ , . - .. .
,
O O l t
SENIOR REACTOR OPERATOR Page 60 ;
l
!
.
ANSWER: 001 (1.00) L[
!
L ?
t i
REFERENCE: .[
f EIH Student Text LT-ST-03001-01 Page 18 Learning Objective EO-1 r
290001K104 [3.7/3.9) ,
l 290001K104 ..(KA's) ;
i
ANSWER: 002 (1.00) i
? ;
REFERENCE:
l EIH Student Text LT-ST-03001-01 DCN 91-171 page 22 ;
Learning Objective EO-13 ,
261000G007 [3.5/3.7) '
!
P
.261000G007 ..(KA's) f i
,
'
-l ANSWER: 003 (1.00)
i
' '
REFERENCE:
EIH: LT-IH-04801, l Learning Objective E.O. ,
!
EIH: Unit 2 Technical Specification 3.6. G005 [ 3 . 3 / 4 .1]
!
223001G005 ..(KA's) !
!
!
l
->
I
- ~-
_ . . . . . . _ - __. - , _ -. .
!
O !
!
- )
!
SENIOR REACTOR OPERATOR ~
'
Page 61 .l t
!
l
<
' ANSWER: 004' (1 00)
t I REFERENCE: r
.
EIH Student Text LT-ST-01301-00 Learning Objective EO 27 EIH Technical Specification 3.6. J-l 223001G005 [ 3 . 3 /4 .1] [
i 223001G005 ..(KA's)
i
.
ANSWER: 005 (1.00) .,
t !
!
REFERENCE:
l
EIH: LT-IH-00401-00, Page 27 j Learning Objective EO 13 !
>
202002K402 [3. 0/3. 0) -'
202002K402 ..(KA's)
!
I ANSWER: 006 (1.00) s
! l t
i REFERENCE: [
'!
EIH Student Notes'LT-ST-00401-00 DCR 90-043 Page 16 !
Learning Objective EO 7 i i
202002A407 [3.3/3.2] i I
i 202002A407 ..(KA's)
i
!
-i l
l
?
-.-
.- - . - . - _... .
- ..- . .. . .. ,
.
!
O O !
I
'
,
-i SENIOR REACTOR OPERATOR Page 62 l i
I
i ANSWER: 007 (2.00) l
!
i a.: 5 b. 8 !
c. 7 !
d. 4 {4 required at 0.5 each)
] !
i REFERENCE- 6
!
EIH LT-ST-OO701-01 Pages 21-23 l Learning Objectives EO 11 & _20 {
i
-
203000K410 [3 . 9 /4 .1] !
!
i 203000K410 ...(KA's) l
!
!
!
ANSWER: 008 (1.00) l
!
a, j
.. ;
REFERENCE: !
!
EIH Unit 2 Technical Specifications 3. f EIH Student Text LT-ST-30005-00 I Learning Objective TO 300.01 !
!
203000G005 '[3.5/4.4] l, y
203000G005 ..(KA's) I
!
?
.
ANSWER: 009 (1.00) !
i {
I t
t
..',
!
l
'l
-
t
. . -
_.- ._ . .
__ _ . . _ __ _ __ .
,
r d O
.
!
!
!
I
' SENIOR REACTOR OPERATOR Page 63 -l
.
REFERENCE:
EIH Student Text LT-ST-02901-01 Table 03901-2 ,
Learning Objective TO 039.01 '
217000A202 [3.8/3.7] _
!
,
t 217000A202 . . (KA's)
i ANSWER: 010 (1.00) :
l REFERENCE:
EIH Student Text LT-ST-03901-01 Page 17 !
'
Learning Objective EO 10
-!
217000A210 [3.1/3.1] j
!
~
a 217000A210 . . (KA's)
l ANSWER: 011 (1.00) ;
i l l
REFERENCE: 1
!
EIH Student Text LT-ST-00501-03 Page 14 l
- Learning Objective EO 6 :
206000K501 [3.3/3.4] j
I 206000K501 . . (KA's) ,
!
!
ANSWER: 012 (1.00) f
i ;
P l
,
-r- r , y . _ , _ _ _ _ _ _ _ _*
_ . , .__ .. _ _ . _ _ _ . _ . _ . . _
i O .O '
i
.
SENIOR REACTOR OPERATOR Page 64:
REFERENCE: -l EIH Student Text'LT-ST-01101-01 Page 13 !
Learning Objective EO 7 l t
'K/A 211000K408 [4.2/4.2] i
!
I 211000K408 ..(KA's)
,
t
!
ANSWER: 013 (1.00) f ;
i REFERENCE:
l EIH Unit 2 Technical Specification 4.9.2. ;
EIH Student Text LT-ST-04502-00 Page 28 l Learning Objective EO 8 l 215004G005 [3.2/3.9) [
l 215004G005 ..(KA's)- j;
.;
i
'!
ANSWER: 014 (1.00)
, ,
!
REFERENCE: l t
EIH Student Text LT-ST-01301-00 Table 3 I Learning Objective EO 21 l i
216000K406 [3.8/4.0) ;
"
t i
216000K406 ..(KA's)
!
ANSWER: 015 (1.00) i
! !
l l
l
!
,;
i
!
, -. _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
- .. . - . . _ . _ ~ - .
,
, O l 1 !
SENIOR REACTOR' OPERATOR Page'65 -[
t
!
REFERENCE: !
EIH Student Text LT-ST-03801-04 Page 7 .j
_
Learning Objective EO 2 ;
> !
i REFERENCE: !
!
EIH LT-ST-02702-00 Page 12 -l Learning Objective EO 5 i i
KA 262001K401 [3.0/3.4] !
-i I 262001K401 ..(KA's) j
-
!
>
I ANSWER: 021- (1.00) !
.; :!
'$
}
'l
)
t
.
'I-
.~ .. , . - . - . . - . . = . . .- . . ..
,
f, O O t
,
SENIOR REACTOR. OPERATOR Page 67 !
.l
?
f REFERENCE:
EIH Student Text LT-ST-00101-00 Page 26 Learning Objective EO 11 ,
!
'201001A201 [3.2/3.3] .!
201001A201 ..(KA's)
!
l ANSWER: 022 (1.00) !
! l I
REFERENCE:
EIH Student Text LT-ST-05403-00 Page 35 i Learning Objective EO 10 {
i 201006A102 [3.4/3.5]- {
i 201006A102 ..(KA's) i r
.
!
i ANSWER: 023 (1.00) i
- !
REFERENCE: f
,
. 'EIH Student Text LT-ST-00401-00 Page 16 f Learning Objective EO-1 EIH Unit 2 Technical Specification Bases 3.4. !
K/A 202001G006 [3. 0/4.1]
r
!
202001G006 ..(KA's) l
!
!
!
!
ANSWER: 024 (1.00) :
r ;
i
.-
J
-!
.
.. _ _ - -,,
&
w
&
w i
SENIOR REACTOR OPERATOR ,
Page 68 REFERENCE:
EIH Student Text LT-ST-00301-01 Page 13 Learning Objective EO Sa 204000K404 [3.5/3.6)
204000K404 ..(KA's)
ANSWER: 025 (1.00)
b.
REFERENCE:
EIH Unit 2 Technical Specification 3.9.5.1 Clarification 0-88-1 EIH Student Text LT-ST-30005-00 Learning Objective TO 30000 K101 [3.7/3.9]
223001K101 ..(KA's)
ANSWER: 026 (1.00)
c.
REFERENCE:
EIH Student Text LT-ST-04901-00 Fage 17 Learning Objective EO #6 EIH Abnormal Procedure 34AB-J11-002-2S Page 2 EIH Operating Procedure 34SO-E32-001-2S Page 4 239001K607 [2.8/3.2)
239001K607 ..(KA's)
ANSWER.: 027 (1.00) ,
O O SENIOR REACTO_R OPERATOR Page 69.
REFERENCE:
EIH Student Text.LT-ST-02301-01 Page 16 ,
-Learning Objective EO 5 j-i K/A 245000G007 [3.5/3.6]
245000G007 ..(KA's) f I
i
!
ANSWER: 028 (1.00) l r
i !
-l '
REFERENCE EIH Abnormal Procedure 34AB-N21-001-2S Page 1 (Caution), and EIH Instructor Handbook LT-IH-20201-02 Learning Objective EO 2 ,
259001A202 [3.1/3.3] !
t 259001A202 ..(KA's) !
I
?
ANSWER: 029 (1.00) i j
i j I
i REFERENCE: .
!
EIH Student Text LT-ST-00201-00 Page 28 l Learning Objective EO 3 l
[
K/A 259001A310 [3.4/3.4] l
-;
k 259001A310 ..(KA's) {
i i
!
ANSWER: 030 (1.00) -)
t !
I
[
.
.-
!
. _ . . _ . _ . _ _ _ _ _ _ _ _ _ .
. . . . . . . .. . . . ~ __- - . .. _ - - . _ - - - . -
O O l I
SENIOR REACTOR OPERATO Page 70- l
REFERENCE:
EIH Unit 2 Technical Specifications 3.3.1 & 3. i EIH Student Text LT-ST-30005-00 l Learning Objective TO 300010 G011 [3.1/4.2]
, 272000G011 .(KA's) {
i
..
'
i ANSWER: 031 (1.00) {
.
,
j ,
j REFERENC .;
l Student Text LT-ST-02705 Figure 1 j Learning Objective EO 1
'
262002G007 [2.9/3.2)
.
{
4 :
262002G007 ..(KA's) !
'
i i
i ANSWER: 032 (1.00) !
,
- j i
, REFERENCE: l
'
!
EIH Student Text LT-ST-04501-00 Page 11 l Learning Objective 8 ~
l t
l !
.
233000K406 [2.9/3.2] l I
233000K406 ..(KA's)
.I
!
I ANSWER: 033 (1.00) 1
,1 I
'
l i
'
'l l
_ _ , _ _ . _ . _ _ _ _ _ _ . . .. . _ _ _ . _ . - _
i i
O O !
!
I i
SENIOR REACTOR OPERATOR .Page 71 ]
i
- , REFERENCE
"'
i EIH Student Text LT-ST-00201-00 !
EIH' Instructor Handbook LT-IH-20201-02 Learning Objective EO 2 EIH 34AB-N61-001-2S Page 2 ;
KA 256000G005 [2.9/3.6] l
!
256000G005 ..(KA's)
i- !
l I l ANSWER: 034 (1.00) l
'l ~
. ,
'
REFERENCE: ,
,
EIH Student Text LT-ST-00102-00 Page 21 l c Learning Objective EO 2 ;
'l
201003K103 [3 .1/ 3.1] i t
I
.
t 201003K103 ..(KA's) j i
i 1 !
ANSWER: 035 (1.00) .
! l,
!
REFERENCE: !
EIH Student Text LT-ST-20201-01 f Learning Objective EO 2 l EIH 34AB-R25-001-2S page ;
i K/A 295003A201 [3.4/3.7] .!
I i
295003A201 ..(KA's)
!
!
ANSW7: 036 (1.00) j t
(
,
i
!
- - - _..__..________________________.______2
- - . . . . _ _ . . . - . - - .
. - . . - -
~
O O ;
!
!
-i-SENIOR REACTOR OPERATOR Page 72~ l e
'ti REFERENCE: l EIH Unit 2 Technical Specifications, 2. }
EIH LT-ST-30002-02, E.O. ;
K/A 295009G003 [3.4/4.2] -!
I l
295009G003 ..(KA's) l
'!
!
>
'5 ANSWER: 037 (1.00) !
. :
!
REFERENCE: -[
4 i EIH Student Handout LR-ST-20310-00 Page 19, and graph 6 j Learning Objectives PCL-5-K1,1 and PCL-6-K !
295030K301 [3. 8 /4.1] i
- I 295030K301 ..(KA's) ,
!
i ANSWER: 038 (1.00) :i
.
REFERENCE:
.
EIH: Unit 2 Technical Specification Bases 3/4.9.1 i l EIH Student Text LT-ST-04502-00 Page 34 -
Learning Objective EO 6 K/A 295023A202 [3.4/3.7]
295023A202 ..(KA's)
!
!
!
ANSWER: 039 (1. 0 0)
- .
f f
i f
i
!
t
- . , ,
'!
. , . ,. - - , _ . , . _ ,
.
.
l 't E
^
O O
!
!
SENIOR REACTOR OPERATOR Page 73 f l
REFERENCE: ,
EIH Student Text LT- IH-20310-01 page 70 Learning Objective EO PCP-5-K K/A 295024K307 [3.5/4.0]
'l
!
t 295024K307 ..(KA's) ,
!
!
i
,
i ANSWER: 040 (1.00)
- REFERENCE
- i
"
EIH Student Text LT-ST-03001-01.p. 16, f
Learning Objective EO f K/A 295024A120 [3.5/ !
,
295024A120 ..(KA's) [
!
ANSWER: 041 (1.00) !
l . REFERENCE: l
!
i EIH Student Text LR-ST-20328-02 Learning Objective TO'201.06 i 295026K304 [3. 7/4.1) i
t
~I
,
295026K304 ..(KA's)
l
!
>
ANSWER: 042 (1.00) i
! j
!
'i i
l
.
. . _ . . .. . . - _ - - . . ~ . - - - - - - - - . - . _ - . - . . . - . - . - _ - . _ . .
'l l O O !
i
,
l
! ,
.,
j -SENIOR REACTOR OPERATOR Page 74
- .
i I
'
REFERENCE: i j- !
j EIH Emergency Procedure 73EP-EIP-001-OS Page 4 W
295030G011 [4.3/4.5]
l 295030G011 ..(KA's) !
!
!
!- !
!
!
[ ANSWER: 043 (1.00)
l l !
-: - 1
, ,
REFERENCE: l
! !
EIH Student Text LT-ST-30004.00 Page 12
]'j Learning Objective EO 8
{
'
l 294001K101 [3. 7/3. 7]
!. I i- i
- 294001K101 ..(KA's) .j
. ;
) i
,-
J
,
ANSWER: 044 (1.00)
)
- ,
.
!
I l REFERENCE:
i l EIH Student Text LT-ST-30004'Page 51 1 Learning Objective EO 33 i l- i 294001A10 [3.4/3.6] i
.
.i
. l a
j 294001A106 ..(KA's) -l i
l
'
i ANSWER: 045 (1. 00) .!
i ..
e
'
d.
J.
e i
! !
i-1 !
>
i j ,.
.-
'
.
} .._- - . - . =- . ,= -. .
. . ~- . -. . . . - _ _ . . -_ .. - _
o o !
SENIOR REACTOR OPERATOR Page 75 l
'
.
REFERENCE:
EIH APC 60AC-HPX-001-0 Page 6 l 10 CFR 20 i EIH: Student Text LT-ST-30003-00R-301 Page 6 l-Learning Objective EO 1 j
294001K103 [3. 3/3. 8] ;
I
!
i 294001K103 ..(KA's)
.,
!.
ANSWER: 046 (1. 00) ;
!
. h
!
REFERENCE: j i
EIH Student Text LT-ST-30004-00 Page 11 .l'
Learning Objective EO 7 E1H Procedure 10AC-MGR-003-0S
KA 294001A106 [3.4/3.6] ,
) 294001A106 ..(KA's) ,
i l i
ANSWER: 047 (1.00) ]
' l i
i REFERENCE:
"
.I
, i EIH Student Text LR-ST-20307-02 Page 10 Learning Objective EO 3 ;
i 3
~
294001A103 12.7/3.7] l
!
.!
!
294001A103 ..(KA's) l
- l
.
ANSWER: 048 {1.00)
c.
.
. .;
. - _ . _ _ _ .
_. - . _ _ _ , ~ . - . . _ _ _ __. _ .._._. _ . .. _ _. -_-- . . _-._ --_ _. m___ .__ _.-
i i
'
O O i i
- s
!
SENIOR REACTOR OPERATOR Page 76 !
,
)
I
<
!
- REFERENCE
- ;
! I
'
EIH Student Text LT-ST-30004-00 Page 51 j l Learning Objective EO 35 i
. 1
! l a 1 j 294001K116 [3.5/3.8] '
t I
i g
294001K116 ..(KA's)
'
ANSWER: 049 (1.00) l
-
-!i i i I !
i l
'
!
.
REFERENCE:
i EIH Procedure 73EP-EIP-001-OS, . ;
!
a K/A 294001A116 [2.9/4.7) !
i !
[ i l 294001A116 ..(KA's)
- l l- !
i'
i l ANSWER: 050 (1.00) !
$ !
^-
! l
!
[ REFERENCE: f r -
3 EIH Technical Specification 3.5. EIH LT-ST-30005-00 l
' EIH Student Text LT-ST-03801-04 !
j: Learning Objective EO
i j' ~K/A 294001A103 [2.7/3.7)
i
i 294001A102 ..(KA's) i j
l l
.
ARSWER: 051 (1.00) -
,
i i l j l
.
s
.
'
. .- _ . . . _ . __ .a._.~ ~ ._ _. _ _- ~. 2
& A W W SENIOR REACTOR OPERATOR Page 77 REFERENCE:
EIH Student Text LT-ST-30004-00 Page 11 Learning Objective EO 3 294001K102 [3.9/4.5]
294001K102 ..(KA's)
ANSWER: 052 (1.00)
c.
REFERENCE:
EIH EOP Flow Charts EIH Student Text LT-ST-20307-02 Page 7 Learning Objective EO 2 295038G012 [3.9/4.5]
295038G012 ..(KA's)
,
ANSWER: 053 (1.00) d.
REFERENCE:
EIH Abnormal Procedure 34AB-C51-001-2S page EIH Student Text LR-ST-20307-02 Page 7 Learning Objective EO 2 295001G010 [3.8/3.7]
295001G010 ..(KA's)
ANSWER: 054 (1.00)
, <
,
. - - _ _ _ _ _ _ . - _ - - _ _ - _ _ _ ______.___-_______-____._.___m_
, _ _ . . . _ . . . _ _ _ _ _ _ - - -_
- _ _ -. - _ . . .. h O ;O i
!
!
SENIOR REACTOR OPERATOR Page 78
!
REFERENCE: ,
EiH Student ~ Text LR-ST-20322-02 Page 9 Learning Objective EO 7a .}
'
295037K306 [3.8/4.1]
,
i 295037K306 ..(KA's) ;
.
ANSWER: 055 (1.00) ;
- '
i
. REFSRENCE: ~
i i EIH Student Text LT-ST-00101-00 i Learning Objective TO 001 .01 !
-
34AB-C11-001-25 Page 2 i
?
s :
'
295022K207 [3.4/3.6]
I l
295022K207 ..(KA's) l
!
!
i ANSWER: 056 (1.00) ! l i i REFERENCE: j
!
EIH Student Text LR-ST-20309-02 Page_70 -!
Learning Objective EO 4 '!
l
'
K/A~295031K302 [4 . 4 /4 . 7] l j
'"
i 295031K302 ..(KA's) {
l
!
, !
,
ANSWER: 057 (1.00) l
! {
f l
-
i
!
-- . _ .
.- . _. .- - . - -
O O !
!
!
SENIOR REACTOR OPERATOR Page 75- ;
i REFERENCE: l EIH Student Text LR-ST-20308-01 Page 32 Learning Objective EO 19 ;
i 295025A105 [3.7/3.7] .!
i 295025A105 ..(KA's)
)
.i i
'
ANSWER: 058 (1.00)
, ;
i REFERENCE:
l EIH Student Text LT-ST-00901-01 Pages 9, 10 and 18 f Learning Objectives EO 1, 4, 8 and 9 -j i
K/A 295018K304 [3.3/3.3] j
!
!
295018K304 ..(KA's) [
-
]
i
-
(1 00)
'
ANSWER: 059 I REFERENCE.
'
EIH Student Text LT-ST-03501-01 Page 50' l
-
34AB-P51-001-25 Learning Objective EO 31-l K/A 295019G010 [3.7/3.4] l t
i 295019G010 ..(KA's)
-,
!
-
ANSWER: '060- (1. 00 ) - 't i
e f
- i
-
--
i
!
. . . .- .- .
. . . __ _ _ . ~ . . . _ _ . .- _ _
i O O ;
i
!
SENIOR REACTOR OPERATOR Page 80 f REFERENCE: i
!
EIH Student Text LR-ST-20306-02 Page 51 ~
Learning Objective EO 20 ;
K/A 295031K304 [4.0/4.3] l l
295031K304 ..(KA's) !
!
' ANSWER: 061 (1. 00 ) !
! l f
REFERENCE: l
'
EIH Student Text LR-IH-20305, , E.O. K/A 295028A203 [3.7/3.9) *
i .
i 295028A20 ..(KA's) !
i
!
ANSWER: -062 (1.00) .
i REFERENCE:
i i
EIH Student Text LR-IH-20306-02, ;
E.O. 29 and 3 f Flow chart CP-2 I t
'
i l
K/A 295028K302 [3.5/3.8) j
__
295028K302 ..(KA's)
ANSWER: 063 (1. 0 0)
' !
!
l
i
- , _ . _ , _ _ __
!
. - _ _ . - _ _ _ . _ .
_ . _ _ _ . _ . . . _ _ _ . _ . . _ . . _ _ _ . _ _ _ . _ _ . _ _ _ _ _ _ . _ _ _ _ . . . . . _ . . . . _ . . _ . _
)
O O 1 l
L l -SENIOR REACTOR OPERATOR Page 81
,
i'
REFERENCE:
EIH Student Text'LR-ST-20317-03 Page 9 ,
Learning Objective EO 5 i i
i
- .-
- .
K/A 295008A109 [3.3/3.3] l.
.
- 295008A109 ..(KA's)
!
i
.
jL l
.
l ANSWER: 064 (1.00)
i i d
!
i .. REFERENCE:
! E1H Student Text LR-ST-20327-02 Page 18 l- Learning Objective EO 4
i i
j 295006G012 [3.8/4.4]
l l -i l 295006G012 ..(KA's) l
,
! !
! ;
i :
I ANSWER: 065 (1.00) l
- i a
i l- .i-i
!- ;
i REFERENC ,
I ;
'
j j EIH Student Text LR-ST-20310-00 Page 76 -l l Learning Objective EO PCP-6-K :
I l
!
l K/A 295024A117 [3.9/3.9] l
! ;
295024A117 ..(KA's)
!
!
l l- ANSWER: 066- (1.00)
lx :
<
l i .
I i
!< {
- ;
i .-
i-
-
l- .
... . . . _ . _ _ _ _
-
__.,,._.__,_..___...--_.,__.,_,..._......,,:._,-.._.;,,.._.,,..- --
_ _ . _ _ . .. _ - __ _ _ _
t
.O !
t l
t SENIOR REACTOR OPERATOR Page 82
.
l
!
'
REFERENCE:
'
!
EIH Student Text LT-ST-01301-00 Page 35, i Learning Objective EO 25b {
?
K/A 295024K207 [3.9/3.9] :
!
i 295024K207 ..(KA's) .
!
!
l ANSWER: 067 (2. 00) . f
'
i a. 4
,
b. 9 c. 3 l
! (4 required at 0.50 each)
REFERENCE: .
'I
EIH 34AB-P51-001-2S Page 1 EIH Student Text LT-ST-03501-01 Pages 24, 25,
- 26 and 27
- Learning Objective EO 2
!
295019K303 [3.2/3.2] .!
'l i
, 295019K303 ..(KA's) !
i
!
"
!
ANSWER: 068 (1.00)- !
i N REFERENCE:
t EIH LT-ST-00401 Page 18 !
Learning Objective EO 5 K/A '202001A109 [3.3/3.3]
o ,
-202001A109 ..(KA's)
,
.;
I i
.. {
!
~
!
i
$
4 -
, ._., ,
.- . . . - - . . _. . - _ - - . - _ _ . - . - - . . . . . . . - . _ . _ -. _ ,
!
.
.
l'
.
.
O O^ 1
- . SENIOR REACTOR OPERATO Page 83- ;
'
j
'
l I !
ANSWER: 069 (1.00) !
.
<
'
i '
'
i
- i l: . REFERENCE:
f i
j EIH Procedure 34GO-OPS-065-2S, Control Rod Movement Section 7. ;
1 EIH Student Text LT-ST-05403-00 i Learning Objective #1 l I
201006K512 {3.5/3.5] t
.
!
l 201006K512 ..(KA's) l I
1 !
L !
! i ANSWER: 070 (1.00) !
)
,
a.
! REFERENCE: )
( i
{ EIH Student Text LT-ST-00501-03 DCN 90-066 Page 18 i l Learning Objective EO 14 s
e l 206000K411 [3.4/3.5]
!
! !
206000K411 ..(KA's)
l !
L t i i
! l l ANSWER: 071 (2.00) -!
-
e
!' 1 ,
-
-i'
[ . Lj l-
- !
~ !
..
! ,
l !
'
i I.
!
l ! !
s- ,
j ?
.
- -
&
i L -
, , - . . . . , .._ , . . , _ . . . - _ . . . . , _ . _ . . . . - . . - . . - . - - - . -
- _. . . . _ - , - . . . _ _ _ _ . . . . ~ - . _ _ . _ _ - _ _ . . . _ - . _ _ . .. _ _._ . _ _ _ _ _ _ _ _ _
I' .j-f
, O O f i' ! >
l SENIOR REACTOR OPERATOR Page 84 ,
l :
-
!- REFERENCE: !
i I
!. EIH Student Text LT-ST-00101-00 Figure 00101-1 l
[ ' Learning Objective EO 3
'l l 201001G004 [3.7/3.7]
i !
i
-
201001G004 ..(KA's) !
i
!
'
' 'l i
l ANSWER: 072 ( l'. 0 0 ) !
i i !
,
REFERENCE: !
! .
'
. EIH Student Text LT-ST-00801 Table 00801-1 j Learning Objective EO 7 j 206000A217 [3.9/4.3] i l
I i !
l 206000A217 ..(KA's) ANSWER: 073 (1.00)
l--
<
,
c.
!
l . REFERENCE: .l l I j EIH Student Text LR-ST-20318 Page 17 'l
[ Learning Objective EO 11
_
295031K204 [4 . 0 / 4 .1] !
!
!
295031K204 ..(KA's)
ANSWER: 074 (1.00)
l l
C.
i'
'
!
!
!
i
'
l
!
1-
- . . . . _ -_ . . - . ~ _ - ..
!
O O l
!
I t
' SENIOR REACTOR OPERATOR Page 85 >
REFERENCE: l
!
EIH Student Text LT-ST-20319-01 l Learning Objective 14.01 i 31EO-EOP-108-25 i
!
295025G012 [3.9/4.5] !
c I
295025G012 ..(KA's) l i
a
!
r ANSWER: 075 (1. 00 ) ,
! ;
I REFERENCE:
{
l
'
EIH Student Notes LR-ST-20326-01 Page 15 Learning Objective EO 5 i i
'
295002G012 [2.4/4.1] !
l 295002G012 ..(KA's) i i
!
!
ANSWER: 076 (1.00) I
j l REFERENCE: l l
EIH Student Text LR-ST-20326-01 Page 21 l Learning Objective EO 11 i EIH Letter CN 92-060' j EIH EDI Flow Chart CP-2 and CP-3 1,
'295031A103 [4.4/4.4] !
'
i
295031A103 ..(KA's) f
-- t
!
-!
- ANSWER: 077 (1.00) !
'l
- - l
}
l
!
!
!
!
. , _ . _ . ._ . _ _ _ _ _ _ . _ . _ _ _ _ __ ___________.__J
--
i
&
O w
'
SENIOR REACTOR OPERATOR Page 86
,
REFERENCE:
EIH Student Text LR-ST-20325-02 Page 50 ,
Learning Objective EO 24 295017A103 [3. 4 /3. 4]
295017A103 ..(KA's)
ANSWER: 078 (1.00)
a.
REFERENCE:
EIH Student Text LR-ST-20305-03 Page 12 Learning Objective EO 7 295027K102 [3.0/3.2]
295027K102 ..(KA's)
ANSWER: 079 (1.00)
a.
REFERENCE:
EIH Instructor Handbook LT-IH-20201-02 Page 7 Learning Objective 5 34AB-B21-003-2S 34AB-T23-003-2S 295013G003 [3.3/4.2]
295013G003 ..(KA's)
'
,
ANSWER: 080 fl.00) ,
.- - ~ . - ~ . _ _ _ ... . - - . . ..
-
O O !
!
I
SENIOR REACTOR OPERATOR Page 8 I i
REFERENCE: [
t Student Text LT-ST-00401-00 Page 30 l Learning Objective EO 3 l l
295005K203 [3.2/3.3] [
!
295005K203 ..(KA's) f i
i ANSWER: 081 (1.00) I t .
i
,
REFERENCE: i
!
.EIH Student Text LR-ST-20325-02 Page 20 !
Learning Objective EO 6 j i
295034K203 [3.7/3.9] 4 i
t 295034K203 ..(KA's) .f
!
&
ANSWER: 082 (1.00) !
i .
!
REFERENCE: I
!
EIH Student Text LR-ST-20325-02 Page 10 ,
Learning Objective EO 1 l
!
295036G011 -[3. B/4.1] !
!
!
295036G011 ..(KA's)
.
ANSWER: 083 (1.00) !
! ;
I
<
h
.
t
!
- ~ . . .
-
. . . , . _
- . -. .
i O O I
Y SENIOR REACTOR OPERATOR Page 88
!
REFERENCE: l I
EIH Student Text LR-ST-20325-02 Page 10 {
Learning Objective EO 1 295033G011 [4. 0/4. 5] ,
295033G011 ..(KA's) l
.
ANSWER: 084 (1.00)
!
- :
!
REFERENCE: !
!
IEH Student Text LR-ST-05401-00 Page 9 [
Learning Objective EO 4 !
!
295015A103 [3.5/3.5] l l
295015A103 ..(KA's) _{
!
ANSWER: 085 (1.00) {
. ;
REFERENCE:
'!
EIH Administrative Procedure 30AC-OPS-009-OS Page 5 t EIH Student Text LT-ST-30004-00 Page 49 Learning Objective EO 32 i i
294001K107 - [3.3/3.6) .i i
f
.294001K107
'
..(KA's)
i
'i
.
ANSWER: 086 (1.00) ,
t
! :
!
i i
i f
i l
.
- - -
. _ + .-. - - , - - - - - - - --v -r- , -~
l O O
.
SENIOR REACTOR OPERATOR Page 89 I
'i REFERENCE:
EIH Student Text LR-ST-20309-02 Page 71 l Learning Objective EO 45
.
295029A203 [3.4/3.6]
295029A203 ..(KA's)
s ANSWER: 087 (1.00) , ;
REFERENCE: ;
,
EIH Student Text LT-ST-30003-00 Page 5 [
. Learning Objective EO '
EIH Technical Specification 6. Table 6.2.2-1 294001A103 [2. 7/3 . 7] :
294001A103 ..(KA's) ;
>
ANSWER: 088 (1.00) l l i i
REFERENCE: ;
EIH Student Text LT-ST-30004-00 Page 51 Learning Objective EO 34 i EIH Procedure 30AC-OPS-001-OS
- 1
'
294001K102 {3.9/4.5]
294001K102 ..(KA's) .f i
t ANSWER: 089 (1.00) :
a I !
.
.
N .Nl
.- - - _ , - - - .-_ .- - -- -.
.
!
'
.O O i
i SENIOR REACTOR OPERATOR Page 90 l
- REFERENCE: f EIH 30AC-OPS-001-OS Page 9 y
, EIH Student Text LT-ST-30004-00 Page 29 '
Learning Objective EO 3 [
!
294001K102 [3.9/4.5] {
i
294001K102 ..(KA's) !
i ANSWER: 090 (1.00) .
! [
!
REFERENCE: l l
EIH Student Text LT-ST-30004-00 Page 38 ;
Learning Objective EO 20 l
i 294001K101 [3.7/3.7] l i
!
294001K101 ..(KA's) f
!
!
\
,
. ANSWER: 091 (2.00) i
.i I
a. 3 !
. c. 1 l
d. 9 l
REFERENCE:
'l EIH Student Text LT-ST-03101-00 ~
Page 10 l Learning Objective EO 4 )
I
'
271000G007 [3.3/3.4] l ll
,
271000G007 ..(KA's) -l
I
.-- - _ - -.
_ _ _ _ _ _ - - . - _ - _ _ _ -
O O l
SENIOR REACTOR OPERATOR Page 91 )
l
!
l ANSWER: 092 (1.00) j l
REFERENCE: 1-j EIH Student Text LR-ST-20326-01 Page 15 i Learning Objective EO 4 I l
l 295032K301 [3.5/3.8)
i 295032K301 ..(KA's) ;
ANSWER: 093 (1.00)
b.
REFERENCE:
EIH Student Text LR-ST-20325-02 Page 55 Learning Objective EO 30 EIH EOP Flow Chart SC/RR 295038K304 I3.6/3.9]
295038K304 ..(KA's)
ANSWER: 094 (2.00) c. 3 l
i
>
l i
l
, - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . . . . _ - - - _ _ . - _ ,
_ . _ . .
& & r w w *
SENIOR REACTOR OPERATOR Page 92 REFERENCE:
EIH Student Text LT-ST-00101-00 Pages 14-17 Learning Objective 3 201001G009 [3.7/3.4) -
201001G009 ..(KA's)
j
ANSWER: 095 (1.00)
i P REFERENCE:
j EIH LT-IH-01501-00 Page 31 Learning Objective EO 19 295014K202 [3.7/4.2)
295014K202 ..(KA's)
>
i
,
!
.
,
,
i
,
s (********** END OF EXAMINATION **********)
'
!
O &
b W SENIOR REACTOR OPERATOR Page 1 r ANSWER KEY
,
MULTIPLE CHOICE 018 c t 001 c 019 b 002 d 020 c 003 d 021 b 004 c 022 d 005 d 023 b 006 c 024 c 007 MATCHING 025 b a 5 026 c b 8 027 a c 7 028 a i
d 4 029 a MULTIPLE CHOICE 030 b 008 a 031 c i l
009 d 032 c 010 t 033 c 011 a 034 d I 012 b 035 d 013 d 036 d 014 b 037 b l l
015 b 038 c 016 d 039 c 017 b 040 b l
,
- - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - - - - _ _ - - _ _ - _ _ _ _ - - - - - _ _ _ _ _ - - - _ - - _ - - - - . - _ . - - - _ - - - - _ _ - _ _
. - . ~ . .--- . _ _ . - . ~ = _ -- .
!
O O L
i SENIOR REACTOR OPERATOR Page 2 i
ANSWER KEY I
,
i f
,
.
041 a 064 d :
042 d 065 d ;
..
>
b 066
~
043 b i
044 a 067 MATCHING i 045 d a 4 I
046 c b 9 047 c c 3 048 c d 5 -l 049 b MULTIPLE' CHOICE 050- a 068 c '
,
!
a 051 a 069 a l 052 c 070 a
!
'
053 d 071 MATCHING ,
t 054 c a 3 !
!
055 c b 1
]
056 a c 7 [
!
057 b d 8 j-058 b MULTIPLE CHOICE f 059 c 072 c 1
060 d- 073 c f i
061 a 074 c' l
?
062 d 075 c l 063 d 076 b l i
-?
!
!
. . _ . . __ . _ _ _ __ - _ _ . _ _ _ _ . _ = .
?
. .. . - _ _ - - . . . - . . - . - _- .. . . . . . . .
,
i
.
O O l I
SENIOR REACTOR OPERATOR Page 3 !
ANSWER KEY I
l l
l
)
i i
!
077 b 094 MATCHING !
!
078 a a 6 -I 079 a b 2
080 d c 3 j d
081 b d
082 a MULTIPLE CHOICE l
,
083 c 095 d l l
, 084 a -
.
l 085 a i !
086 e i
!
I 087 a ;
i 088 d i
)
089- b -
t 090 d j 091 MATCHING t
a 3 l
>
b 7 l
c 1 d 9 !
!
MULTIPLE CHOICE -!
, 092 d ;
- i
5 093 b i l !
-
i d
t (.********** **********)
.
.
END OF EXAMINATION i
-
i l
l
_ . - . , .. - ._ _ ._ ,'
___ . . . . . __ _ __ . . . .
,
!
O O i i
TEST CROSS REFERENCE Page 1 j
!
!
!
Oraanized by Ouest ion Number j i
.l OUESTION VALUE REFERENCE 001 1.00 9000001 002 1.00 9000002 003 1.00 9000003
004 1.00 9000004 f 005 1.00 9000005 !
006 1.00 9000006 ,
007 2.00 9000007 l
'008 1.00 9000008 l 009~ 1.00 9000009 ;
010 1.00 9000010 -
011 1.00 9000011 !
012 1.00 9000012 013 1.00 9000013 i 014 1.00 9000014 l
'
015 1.00 9000015 l 016 1.00 9000016 !
017 3.00 9000017 i 018 1.00 9000018 l 019 1.00 9000019 020 1.00 9000020 021 1.00 9000021 j 022 1.0 .00 9000023 ;
024 1.00 9000024 1 025 1.00 9000025 l 026 1.00 9000026 i 027 1.00 9000027 ;
028 1.00 9000028 j 029 1.00 9000029 !
030 1.00 9000030 !
031 1.00 .9000031 l 032 1.00 9000032 !
033 1.00 9000033 Lj 034 1.00 9000034 035 1.00 9000035 036 1.00 9000036 ;
037 1.00 9000037 !
038 1.00 9000038 I 03 .00 9000039-040 1.00 9000040 041 1.00 9000041 c 042 1.00 9000042 043 1.00 9000043 044 1.00 .9000044 l
- 045 1.00 9000045 046 1.00 9000046 .i 047 1.00 9000047 l 048 1.00 9000048 !
049 1.00 9000049 ,
,
-
. , , -,
'
, _ _ - =. . _ _ . - - _ _ ._ . . _ . . _
!
!'
O TEST CROSS REFERENCE O Page :
l f
!
f Orcanized by Ouest ion Number OUESTION VALUE REFERENCE l 050 1.00 9000050 t 051 1.00 9000051 ,
052 1.00 9000052 I 053 1.00 9000053 ;
054 1.00 9000054 .i
..
'
055 1.00 9000055 :
056 1.00 9000056 j 057 1.00 9000057 i 058 1.00 9000058 .;
059 1.00 9000059 !
060 1.00 9000060 i
'
061 1.00 9000061 062 1.00 9000062 .!
063 1.00 9000063 i 064 1.00 9000064 l 065 1.00 9000065 ;
066 1.00 9000066 067 2.00 9000067 [
,
068 1.00 9000068 ;
069 1.00 9000069 l 070 1.00 9000070 j 071 2.00 9000071 j 072 1.00 9000072 [
073 1.00 9000073 i 074 1.00 9000074 ;
075 1.00 9000075 076 1.00 9000076 l
"
077 1.00 9000077 i 078 1.00 9000078 079 1.00 9000079 !
080 1.00 9000080 [
081 1.00 9000081 ;
082 1.00- 9000082 083 1.00 9000083 084 1.00 9000084-085 1.00 9000085 086 1.00 9000086 087 1.00 9000087 I
, 088 1.00 9000088 l 089 1.00 9000089 ;
090 1.00 9000090 !
091 2.00 9000091 1
'
092 1.00- 9000092 !
093 1.00 9000093
-
l 094 2.00 9000094 r 095 1.00 9000095 i
!
...___ l
,
100.00 ~
______
______
100.00 ~
^t
.
. . .
-
.. - . - -- _ . _ _ - . -- .. - ... .
!
O TEST CROSS REFERENCE O Page 4
!
'
e
!
Oraanized by KA Group i
!
PLANT WIDE GENERICS !
!
OUESTION VALUE KA i 050 1.00 294001A102 !
047 1.00 294001A103 $
087 1.00 294001A103 i 046 1.00 294001A106 l I 044 1.00 294001A106 ;
049 1.00 294001A116 !
043 1.00 294001K101- !
.
090 1.00 294001K101 :
051 1.00 294001K102 l 089 1.00 294001K102 l
,
'
088 1.00 294001K102 l 045 1.00 294001K103 ,
l- 085 1.00 294001K107 1 048 1.00 294001K116 f
______
PWG Total 14.00 !
!
I I
. PLANT SYSTEMS i
'
- !
>
,
Group I j i i
- OUESTION VALUE KA f I
'
006 1.00 202002A407 l 005 1.00 202002K402 i 008 1.00 203000G005 {
.
007 2.00 203000K410 :
! 072 1.00 206000A217 f
) 070 1.00 206000K411 l
011 1.00 206000K501 l l 012 1.00 211000K408 !
013 1.00 215004G005 f .00 216000K406 }
'
009 1.00 217000A202
,
010 1.00 217000A210 {
016 1.00 218000A405 i j 015 1.00 218000K606 [
003 223001G005
'
1.00 !
004 1.00 223001G005 .
025 1.00 223001K101 !
017 1.00 223002K112 018 1.00 239002A309 019 1.00 259002K414 i 002 1.00 261000G007 j 020 1.00 262001K401 ;
- 001 1.00 290001K104 [
<
f
'
!
L !
_ _ . . -. . . . . . _ _ . . _ _ . . . .- --~ _ -
!
. . - . . . . . . . . . - - - - . . - . . - - . - . - . . . - = _ . . - ~ ~ . ~ . . - . - . . . - . . ~ . _ - - - . ~ I L
i O TEST CROSS REFERENCE O Page 5 i
!
i i i i
> '
l SRO E x.a m BWR React cr Oroanized bv KA Groun
! I
!
!
l PLAITI' SYSTEMS i l
Group I OUESTION VALUE KA l
- ____ _
l; PS-I Total 24.00 ,
.
"
.
l Group II !
} !
!
OUESTIO VALU KA -!
021 1.00 201001A201 l j 071 2.00 201001G004 !
! 094 2.00 201001G009 022 1.00 201006A102
.
069 1.00 201006K512 l f
068 1.00 202001A109 l 023 1.00 202001G006 l l 024 1.00 204000K404 027 1.00 245000G007 028 1.00 259001A202 j 029 1.00 259001A310 3 l 031 1.00 262002G007
{ 091 2.00 271000G007 !
030 1.00 272000G011 l
.__ j i PS-II Total 17.00 l
'
l l
- Group III .!
OUESTION- VALUE KA l 034 1.00 201003K103 l 032 1.00 233000K406
026 1.00 239001K607
) 033 1.00 256000G005
......
l PS-III Total 4.00 g _____.
j ......
j
! PS Total 45.00 -!
!
!
!
i i
! i
- ~
i
$
1 i
!
l 1 i
! !
i !;
<
k.
a !
l .
.. !
O
'"
&
TEST CROSS REFERENCE Page 6 !
Oroani zed by KA Group EMERGENCY PLANT EVOLUTIONS Group I OUESTION VALUE KA 035 1.00 295003A201 064 1.00 295006G012 036 1.00 295009G003 079 1.00 295013G003 095 1.00 295014K202 '
084 1.00 295015A103 077 1.00 295017A103 038 1.00 295023A202 065 1.00 295024A117 040 1.00 295024A120 066 1.00 295024K207 039 1.00 295024K307 057 1.00 295025A105 074 1.00 295025G012 041 1.00 295026K304 078 1.00 295027K102 042 1.00 295030G011 037 1.00 295030K301 076 1.00 295031A103 073 1.00 295031K204 056 1.00 295031K302 060 1.00 295031K304 054 1.00 295037K306 052 1.00 295038G012 093 1.00 295038K.304 .
._____
EPE-I Total 25.00 Group II OUESTION VALUE KA 053 1.00 295001G010 075 1.00 295002G012 080 1.00 295005K203 063 1.00 295008A109 '
058 1.00 295018K304 059 1.00 295019G010 067 2.00 295019K303 055 1.00 295022K207 061 1.00 295028A203 ,
062 1.00 295028K302 1 086 1.00 295029A203 ,
092 1.00 295032K301 083 1.00 295033G011
-
a a TEST CROSS REFERENCE Page 7 SRO Exam BWR React or Oroani zed bv KA Group EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE KA ,
081 1.00 295034K203 082 1.00 295036G011 ,
'
.. ..-
EPE-II Total 16.00
___ ......
EPE Total 41.00
. . . . . .
___ ......
Test Total 100.00
.---_ . . _ . - ._ _ . _ _ _ , _ . , - _ . _ _ . . _ . _ -
_ . . . . . . . _ . _ _ . - ._ . . . . _ . . ._. _ _ . _ ..
ABNO3NEAL W DLE * *CE DC VOLTAGE W ^
t RC R () -
(NVERTER hit N I )R =
l l RATTfRY
[ =
INVERTER O2 g IDAD NORMAL SOtJRCE p) 1EST g I
600 VAC 2D _
_
AL'8 1.J N I u l RETURNktODE sd A%il( ti l i
=
JD O l
AtTO htAN I
I ss Nc I (1RCTil'
I O "4 ' "! "'
______g,__,________m___
Bi PASS nWilC11 ALTERNATE SOURCE 600 VAC 2C VITAL AC STATIC INVERTER Figure 1
.
_ . . . _ _ _ . . . . . _ . _ . . _ . . _ . _ . _ . . _ _ . . . _ ~ . . _ . . _ _ _ . _ _ _ _ . _ . . . _ _ _ _ _ . . __ _ _ . . . _.._ .._ ._... .. _, ..._ _. __..._._ ._-_ . _ . . _ _ .
H
'
A H
'
' FS - A
,/
A MOTOR L FS j
~
$ lkM B I
i l
'O (CONTROLLED LEAKAGE)
awens ln' RO
..
SECOND SEAL No. 2 ~
V
- - 2 o
'
CONTROLLED PRESSURE SEAL PURGE .
BREAKDOWN FROM CRD SYSTEM g g-(INTERNAL AT SEAL) ll
' BREAKDOWN BUSHil4G o !!
J <
PUMP
.
Figure 2 1 .
-
e v- , m v
i .,j # .
'
.
0; o .
,-
g ~l g '
t i :
X '
-O **
'
i x o$, o 00 i oof ;
,
-- I a
x X X l
--
3 1
is Ts I
n
e -
(XM-
'O '
-b<H>4 y m n () > .
, 3 l <D lf :
i '
w '
$ 4V @ %
a E
--
~~ w$O C}-M)
@ **
l O
bdAI)
~~
- X --
.
g g _
b @ e F eI- !;# **
.
0 W- T-T 0 W-O @ = :
no 4* :
,
- i o
l -
'
.
k o O
-
C O
N D
E NN S
E R
"
S TM
" EA
'.
AIN M
'l h 3-V
@
'
u .
1 Am l S
'ih T M EA
& ' AI N M
O A
F F-G
~ <
~
p" li * m S TM i
A s S i, F
0 EA S
Y
&
n n
5 AIN M
S n d
T t, F E g
i g M
',
y u
r F
L sa a !
e O --
4 W -
-
~
D " * ti-!
AI G C b R O A N M UD I
I An N TI I R T O N
I
N t G l j
I_
"
' f
,K T R~
E S T
AG T AN A MSD E
N B T Y
- <' , 2 , ' , t'
e .
L . i
.
.
.
O t
O g .
_
.
.
.
.
_
.
.
_
.
..
.
_
_
.
_
.
.
_
.
_
.
v $dee>- O _
b>
4$h C e _
.
A _
SR
.
.
.
I I c .
_
.
M S/ .
.
_
P e .
L C I R r
r><
r
.-
$f
.. F IMA
..
.
_
E a _
DAI
..
.
.
vw DR m'
-
.
R .
AH .
WAE I
ND .
.
.
GE R +M b
.
.
.
R .
P S yl .
F B" TA _
.
i "
g R .
.
u P _
r e
S T { _
b
.
A" ~
-
_
5 " _
7, .
3 .
7 _
.RM
_
P .
.
S 8
.
.
.
_
A" B"" _
,
.
- 9" l
..
?b
.
-
._
.
.
.
.
,
h _
-
.
"1 l R
M
__
S I
_
.
1 C
.
.
O N
[.
h T R _
.
O
, V
, L _
, /
.
.
_
R R O P O .
M _
S
, _
, A" .
.
,
, " _
.
..
.
.
._
.
.
.
_
..
_
? , - *