ML20134K172
| ML20134K172 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/07/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20134K160 | List: |
| References | |
| 50-245-96-12, NUDOCS 9702130167 | |
| Download: ML20134K172 (91) | |
See also: IR 05000245/1996012
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l U. S. NUCLEAR REGULATORY COMMISSION [
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REGION 1 l
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4 DOCKET / REPORT NO: 50-245/96-12 l
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lj LICENSEE: Northeast Nuclear Energy Company . e
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- FACILITY: Millstone Nuclear Power Station Unit 1 l
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{ DATES: December 2-6,1996
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j EXAMINERS: D. Florek, Sr. Operations Engineer :
i T. Walker, Sr. Operations Engineer !
! D. Prawdzik, NRC Contract Examiner !
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! APPROVED BY: Glenn W. Meyer, Chief
{ Operator Licensing and Human Performance Branch
{. Division of Reactor Safety
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9702130167 970207
PDR ADOCK 05000245
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EXAMINATION SUMMARY
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Examination Report 50-245/96-12 (OL)
Initial examir.3tions were administered to four senior reactor operator (SRO) instant j
applicants and three SRO upgrade applicants during the period of December 2 - 6,1996, at
the Millstone Unit 1.
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OPERATIONS l
Six of seven applicants failed the examination. One SRO instant applicant passed the
examination. The applicants were poorly prepared for the examinations. Significant
generic weaknesses were noted in both the written and operating test. The average grade
for the written was 75%. Significant weaknesses were identified during the operating test
related to operating reactivity controls, diagnosis of instrument failures, and diagnosis of l
ECCS injection status. f
An unresolved item was identified related to inconsistencies between the EOPs, the EOP
bases document, the plant specific technical guidelines (PSTGs), and the Boiling Water
Reactors Owners Group (BWROG) Emergency Procedure Guidelines (EPGs) for terminating
and preventing RPV injection when all control rods are not inserted and it has not been
determined that the reactor will remain shutdown under all conditions without boron.
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DETAILS
05.1 Operator Initial Examinations
a. Scope
The examiners administered initial examinations to four instant SRO applicants and
three upgrade SRO applicants in accordance with NUREG-1021, " Examiner
Standards," Revision 7.
b. Observations and Findinas
The results of the initial examinations are summarized below:
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PASS / Fall
Written 2/5
Operating 2/5
Overall 1/6
The Northeast Utilities (NU) staff reviewed the written examination and assisted in
the validation of the operating examination during the week of November 18,1996.
The NU staff provided comments on the examination that significantly improved the
examination. The NU staff, who were involved with the examination review, signed
security agreements to ensure that the initial examinations were not comproniised.
In a letter, dated December 13,1996 (see Attachment 2), NU provided ten
comments on the written examination. The NRC accepted six of the ten comments.
As a result, the correct answer was changed in four questions and two correct
answers were accepted in two questions. The NRC resolution of facility comments
is summarized in Attachment 3.
Because the examination contained a number of newly developed higher cognitive
level questions, the examiners were monitoring the progress of the applicants during
the administration of the written examination. Based on the progress at the three
hour point the examiners concluded that most of the applicants would not have had
sufficient time to complete the examination within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The examiners then
announced to the applicants that an additional 30 minutes would be added to the
examination. All applicants completed the examination with the additional time
allowed.
The performance on the written examination was poor. Based on the grading of the
written examination the overall average examination grade was 74.7%. The
administrative question area average was 58%, the systems question area average
was 79%, and the events question area average was 77%, indicating that average
performance in these areas was unacceptable. In particular, the examiners noted
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that poor performance in the administrative area had been identified in Millstone
exams a few years ago and addressed within the training program. Thirty percent
of the questions on the written examination were missed by at least three of the
- applicants.
The following summarizes the questions that were missed by at least four
applicants, indicating a weakness in the understanding of the subject.
SRO 1: Ability to determine the posting requirements for an airborne
radioactivity area
SRO 6: Knowledge of inservico surveillance testing requirements
SRO 7: Knowledge of surveillance periodicity and tolerances
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SRO 10: Knowledge of confined space permit requirements l
SRO 11: Knowledge of acceptable methods for performing a continuous
SRO 12: Understanding of the requirements for a continuous use
procedure ;
La015: Understanding of overtime limitations
SRO 23: Ability to determine the cause of a downscale rod block
monitor
SRO 35: Ability to determine Technical Specification requirements for
failure of an isolation condenser flow detector
SRO 41: Understanding of containment spray valve operation l
SRO 48: Understanding of shutdown cooling temperature indications
and limitations
SRO 50: Understanding of ATWS recirc pump trip operation and recirc
pump controller indications
SRO 80: Understanding of reactor feed pump high level trip and
initiation signals
SRO 84: Ability to recognize a failure of primary containment and
determine the correct emergency action level
SRO 85: Ability to select the correct reactor pressure vessel alternate
injection source with the condensate transfer pump house
inaccessible
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SRO 88: Understanding of xenon reactivity effects
SRO 91: Knowledge of the rear,un for LPCI heat exchanger flow
limitations
SRO 95: Knowledge of the effects of containment spray on containment
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differential pressures
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SRO 98: Ability to implement the secondary containment control
emergency operating procedure
SRO 100: Understanding of stack gas monitor operation
During the operating test, at least two applicants performed poorly in each of the
following areas: j
Operating controls that affect reactivity.
Recognizing and diagnosing instrument failures and taking action appropriate ,
to the failure.
Using the crew to diagnose ECCS (emergency core cooling system)
availability to inject into the vessel.
Performing SRO crew briefs with effective timing, sufficient content,
sufficient look ahead, and identification of priority.
Reporting back to the SRO on the status of equipment that failed to work
properly.
Performing the Emergency Director responsibilities related to site
evacuations.
Recognizing the appropriate actions to take when in single loop operation
and the plant is below the 80% rod line.
Understanding the recirculation pump response to a tachometer failure.
Applying the relationship between APRMs and IRMs and effectively using the
technical specifications.
Understanding the effect on power operation if 24 VDC power is lost.
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The above test items represent areas of weak understanding or performance and are
provided to enable improvement of the training program.
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] During the dynamic simulator test, the following item was significant and a
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Kr.owledge and understanding of the emergency operating procedures
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I During the development and administration of the examination, the examiners noted
the following items for further NU consideration of possible procedure ;
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improvements. ;
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1 Facility procedure SP 696.1 " Control Operators Log," could result in -
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exceeding the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> power average of 2011 MWth. If power is greater [
than 2011 MWth and then is at 2011 MWth for the next seven hours, then !
- the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> power average limit could be exceeded.
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1 OP 349D " Nuclear instrumentation (LPRM Channels)," did not contain a i
caution regarding minimum number of LPRMs for each ARPM consistent with
j OP 349C " Nuclear instrumentation (APRM Channels)," even though it
contained the same procedural steps.
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1 OP-301 " Reactor Recirculation System," was confusing as to when to check ;
j for reverse flow following shutdown of a recirculation pump. t
Guidance did not exist restricting the use of the emergency diesel or gas .
turbine generation under peak load generation conditions. !
Alarm response procedures CRAB 905 A-3,8-1 and 8-2 were unclear that
only one SDV at the valid scram SDV water level was needed to initiate a full
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OP-525A " Degraded Fire in Control Room of Cable Vault," section 2.7.4 and
1 2.7.5 did not specify the need for local operation and the use of a key to
] complete these steps.
l SP 668.2 " Gas Turbine Emergency Fast Start," did not provide clear
l guidance on exit of TS 3.5.F.3 to meet conditions of precaution 3.2.
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! The applicants were poorly prepared for the examination, and as a result,6 of 7
i applicants failed the examination. One SRO instant applicant passed the
l examination by a very small margin. Significant reneric weaknesses were noted in
both the written and operating test, and four of the seven applicants had written
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scores of 72 or below. Significant weaknesses were identified during the operating
- test related to operating reactivity controls, diagnosis of instrument failures, and
diagnosis of ECCS injection status.
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03 inconsistencies in Emergency Operating Procedures
a. Scope
During examination preparation and review activities, the examiners noted some
inconsistencies within the EOPs as compared to other sites. Subsequent followup
reviews were then performed,
b. Findinas
The examiners identified inconsistencies between the EOPs, the EOP bases
document, the plant specific technical guidelines (PSTGs), and the Boiling Water
Reactors Owners Group (BWROG) Emergency Procedure Guidelines (EPGs) for
terminating and preventing RPV injection when all control rods are not inserted and
it has not been determined that the reactor will remain shutdown under all
conditions without boron.
The BWROG EPGs specify that allinjection into the RPV (reactor pressure vessel)
except SLC, CRD, and RCIC should to be terminated and prevented prior to initiation
of emergency RPV depressurization. The MP1 PSTGs do not indicate that the
actions to terminate and prevent injection have to be completed prior to initiation of
emergency depressurization. The justification for this deviation in the MP1 PSTGs
states that it is not an intent change and that injection must be terminated prior to
opening SRVs or bypass valves to emergency depressurize. Nonetheless, the EOP
bases document also states that action to depressurize the RPV waits until it is
confirmed that injection into the RPV is terminated and prevented. However, the
EOP for emergency depressurization, which is consistent with the PSTG, does not
require that the actions to terminate and prevent injection are completed prior to
opening SRVs or bypass valves.
During administration of the examinations, the applicants did not wait for the
actions to terminate and prevent injection to be completed prior to initiating actions
to emergency depressurize. In one case reactor pressure dropped below the shutoff
head of the core spray (CS) system before actions were taken to prevent injection
and CS injected. During followup questioning, the applicant that was acting as the
SRO in the scenario stated tnat it was not necessary or appropriate to wait until
injection had been terminated and prevented to open SRVs or bypass valves. The
applicant's actions and response were consistent with the direction in the EOP flow
chart, but were not consistent with the EOP bases.
Additionally, the specific actions that are taken to terminate and prevent injection
from the low pressure coolant injection (LPCI) system, do not assure that injection
is prevented. No action is taken until the LPCIinjection valve strokes full open at
350 psig RPV pressure. If RPV pressure is dropping rapidly, pressure may drop
below the shutoff head of the LPCI pumps before the injection valve is closed,
allowing LPCI to inject.
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The applicable references are as follows:
EOP 560.3, " Emergency Operating Procedures Technical Guidelines," pg 34 i
EOP 560.4, " Deviation Justification Report," pgs 202 and 203 l
EOP 560.6, " Millstone Unit 1 Emergency Operating Procedure Failure to
l Scram," pg 68
l EOP 575, sheet 3, " Failure to Scram - Emergency Depressurization,"
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step DEP-4
l EOP 550.2, " Unit One Emergency Operating Procedure User's Guidelines," ;
pg l7
c. Conclusion
This item is considered unresolved pending facility actions to resolve the .
inconsistencies between the BWR EPGs, PSTG, EOP bases document and the EOPS
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regarding actions to terminate and prevent low pressure injection prior to emergency
depressurization. Unresolved item 50-245/96-12-01. :
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E.8 Review of UFSAR Commitments
A recent discovery of a licensee operating their facility in a manner contrary to the
updated final safety analysis report (UFSAR) description highlighted the need for a
special focused review that compares plant practices, procedures and/or parameters
to the UFSAR descriptions. While performing the examination activities discussed
in this report, the examiners reviewed portions of the UFSAR that related to the
selected examination activities, questions or topic areas. The particular sections
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reviewed were Section 7.4.3, and Section 7.6.1.4. The specific areas reviewed
were consistent with the UFSAR.
V. Manaaement Meetinas
X1 Exit Meeting Summary
At the conclusion of the examination, the examiners discussed their observations of the
examination process with members of NU management. NU acknowledged the examiners'
observations. The NU personnel present at the exit included the following:
H. Haynes, Director, Nuclear Training
W. Noll, Manager, Unit 1 Operations
K. Murphy, Supervisor - Operator Training, Unit 1
NRC Personnel
D. Florek, Sr. Operations Engineer
T. Walker, Sr. Operations Engineer
T. Easlick, Senior Resident inspector
J. Durr, DRP Section Chief
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Attachments: I
, 1. SRO Examination and Answer Key
2 2. Facility Comments on Written Examinations !
3 3. NRC Resolution of Facility Comments on the Written Examinations I
! 4. Simulation Facility Report
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ATTACHMENT 1
SRO EXAMINATION AND ANSWER KEY
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I U. S. NUCLEAR REGULATORY COMMISSION
SITE SPECIFIC EXAMINAT!ON
SENIOR OPERATOR LICENSE
REGION 1
CANDIDATEfS NAME: b b ,
FACILITY: Millstone 1
REACTOR TYPE: BWR-GE3
DATE ADMINISTERED: December 2.1996
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INSTRUCTIONS TO CANDIDATE:
Use the answer sheets provided to document your answers. Staple this cover sheet on
top of the answer sheets. Points for each question are indicated in parentheses after the
question. The passing grade requires a final grade of at least 80%. Examination papers
will be picked up four (4) hours after the examination starts.
TEST VALUE CANDIDATE'S SCORE FINAL GRADE
100.00
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All work done on this examination is my own. I have neither given nor received aid.
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I Candidate's Signature
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SENIOR REACTOR OPERATOR Page 2
l ANSWER SHEET
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Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
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MULTIPLE CHOICE O23 a b c d _
001 a b c d 024 a b c d
002 a b c d 025 a b c d
003 a b c d 026 a b c d
004 a b c d 027 a b c d
Od5 a b c d 028 a b c d
006 a b c d 029 a b c d
007 a b c d 030 a b c d '
008 a b c d 031 a b c d
009 a b c d 032 a b c d
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 d
016 a b c d 039 a b c d
017 a b c d 040 a b c d
018 a b c d 041 a b c d
019 a b c d 042 a b c d
020 a b c d 043 a b c d
021 a b c d 044 a b c d
022 a b c d 045 a b c d
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SENIOR REACTOR OPERATOR Paga 3
ANSWER SHEET
Multiple Choice (Circle or X your choice)
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If you change your answer, write your selection in the blank. ;
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046 a b c d 069 a b c d
047 a b c d 070 a b c d
048 a b c d 071 a b c d
049 a b c d 072 a b c d
050 a b c d 073 a b c d
05'1 a b c d 074 a b c d
052 a b c d 075 a b c d
053 a b c d 076 a b c d
054 a b c d 077 a b c d
055 a b c d 078 a b c d
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056 a b c d 079 a b c d
057 a b c d 080 a b c d
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058 a b c d 081 a b d
059 a b c d 082 a b c d
060 a b c d 083 a b c d
061 a b c d 084 a b c d
062 e b c d 085 a b c d
063 a b c d 086 a b c d
064 a b c d 087 a b c d
065 a b c d 088 a b c d
066 a b c d 089 a b c d
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067 a b c d 090 a b c d i
068 a b c d 091 a b c d
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SENIOR REACTOR OPERATOR Page 4
ANSWER SHEET
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Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
092 a b c d _
093 a b c d i
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094 a b c d
095 a b c d ]
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096 a b c d
097 a b c d l
098 a b c d _
099 a b c d _
100 a b c d
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( * * ' ' ' " END OF EXAMINATION * * " "' ")
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS '
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During the administration of this examination the following rules apply:
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1. Cheating on the examination means an automatic denial of your application and
i could result in more severe penalties.
2. After the examination has been completed, you must sign the statement on the ;
cover sheet indicating that the work is your own and you have not received or given
assistance in completing the examination. This must be done after you complete
the examination.
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3. Restroom trips are to be limited and only one applicant at a time may leave. You j
must avoid all contacts with anyone outside the examination room to avoid even ;
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the appearance or possibility of cheating.
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4.. Use black ink or dark pencil ONLY to facilitate legible reproductions. i
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5. Print your name in the blank provided in the upper right-hand corner of the -
examination cover sheet and each answer sheet.
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6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER
l PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
7. The point value for each question is indicated in parentheses after the question.
8. If the intent of a question is unclear, ask questions of the examiner only.
I 9. When turning in your examination, assemble the completed examination with
examination questions, examination aids and answer sheets, in addition, turn in all
scrap paper.
10. Ensure allinformation you wish to have evaluated as part of your answer is on your
answer sheet. Scrap paper will be disposed of immediately following the
examination.
11. To pass the examination, you must achieve a grade of 80% or greater.
12. There is a time limit of four (4) hours for completion of the examination.
13. When you are done and have turned in your examination, leave the examination
area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the
examination is stillin progress, your license may be denied or revoked.
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QUESTION: 001 (1.00) ;
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Conditions in a recently surveyed area are:
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I 50 dpm/100 cm2 alpha loose surf ace
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0.50 DAC airborne radioactivity
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WHICH ONE of the following describes the posting requirements for the
area?
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a. Contaminated area - required
Airborne radioactivity area - required
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b. Contaminated area - required
Airborne radioactivity area - not required
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c. Contaminated area - not required
Airborne radioactivity area - required
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d. Contaminated area - not required
Airborne radioactivity area - not required
OUESTION: 002 (1.00)
A 29 year old Northeast Utility radiation worker with a complete and
current exposure history needs to perform work in a high radiation area.
The expected job exposure is 1500 mrern. Thts worker has a lifetime Total
Effective Dose Equivalent (TEDE) of 30.5 Rem and 750 mrem exposure for
this year. ;
WHICH ONE of the following describes the worker's ability to perform the
job?
a. The worker cannot perform the job. I
b. The worker can perform the job without any additional exposure
limits.
c. The worker can perform the job with an additional exposure limit
approved by RSP-Services only.
d. The worker can perform the job with an additional exposure limit
approved by RSP-Services and the Health Physics Manager.
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SENIOR REACTOR OPERATOR Paga 8
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QUESTION: 003 (1.00) ;
The position of an open valve in a safety related system needs to be [
verified following maintenance on the system.' Radiation levels in the l
area are 75 mr/hr and it is expected to take 10 minutes to verify the l
position of the valve.
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WHICH ONE of the following describes the requirements for verification
of the valve's position?
a. Valve position must be verified using normal verification f
methods. Independent verification is required. i
. b. Valve position must be verified using normal verification l
methods. Dual verification is required.
c. An alternate verification method may be used in place of normal .l
verification methods. Independent verification is required. !
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d. An alternate verification method may be used in place of normal
verification methods. Dual verification is required.
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QUESTION: 004 (1.00)
A flow transmitter is being returned to service. l
WHICH ONE of the following describes the correct sequence to ensure i
system disturbances are not created? l
a. Open the low pressure side isolation valve. I
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Close the equalizer valve.
Open the high pressure side isolation valve.
b. Open the low pressure side isolation valve.
Open the high pressure side isolation valve.
Close the equalizer valve.
c. Close the equalizer valve.
Open the high pressure side isolation valve.
Open the low pressure side isolation valve.
d. Open the high pressure side isolation valve.
Open the low pressure side isolation valve.
Close the equalizer valve.
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- QUESTION: 005 (1.00)
WHICH ONE of the following is allowed when hanging tags in accordance
- with WC-2, " Tagging?" i
a. Two blue tags may be hung on an open drain valve if the same
contact person has responsibility for both tags,
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b. A red tag may be hung on a closed breaker inside an energized
electrical cabinet. The tag can be attached to the breaker with ;
string and a small section of tygon hose.
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c. A red tag may be hung for a removed breaker. The tag can be
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attached to the breaker cubicle door with an adhesive hanger.
d. The tagging authority may correct a typographical error by ;
initialing and dating the handwritten correction on a yellow I
panel tag. l
QUESTION: 006 (1.00)
On 12/2/96, the plant was shutdown and cooled down for an unplanned
outage. The outage is expected to last approximately five days. Prior
to the unplanned shutdown, the plant had been operating continuously
since 10/2/96. IST cold shutdown surveillance testing was last
completed on 8/28/96.
WHICH ONE of the following describes the requirements for performance of
IST cold shutdown surveillance testing during the unplanned outage?
a. IST cold shutdown testing must begin within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of reaching
cold shutdown and allIST cold shutdown testing must be
completed prior to startup.
b. IST cold shutdown testing must begin within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of reaching
cold shutdown, but startup does not have to be delayed in order
to complete the testing.
c. IST cold shutdown testing is not required because the shutdown
is not expected to last more than 90 days. i
d. IST cold shutdown testing is not required because the plant has
been in continuous operation for less than 90 days.
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QUESTION: 007 (1.00)
WHICH ONE of the following lists the tolerance and the station
administrative periodicity for a surveillance with a "Q" frequency?
a. Tolerance 31 days -
Periodicity 92 days
b. Tolerance 31 days
Periodicity 84 days
c. Tolerance 10 days
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Periodicity 92 days
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d. Tolerance 10 days -
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Periodicity 84 days '
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QUESTION: 008 (1.00)
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WHICH ONE of the following has the responsibility for ens 2 ring an -
Automated Work Order (AWO) package is complete and ready to performance? I
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a. Tagging Authority
b. Work Planning and Outage Management
c. Shift Management / Unit Supervision
d. Work Planning Quality Assurance Services
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SENIOR REACTOR OPERATOR Pags 11
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QUESTION: 009 (1.00)
A battery charge was in progress when the " BATTERY ROOM HYDROGEN" alarm
was received on CRP 903. The operator dispatched to the battery room
- reports that:
- the hydrogen monitor is reading approximately 40%
- the battery room exhaust fan is not running
- the exhaust fan switch on the main control panel is in AUTO
- the exnaust fan switch on MCC F-5 is in AUTO
WHICH ONE of the following describes the conditions in the battery room
and the required actions? (The ARP is provided.)
a. A warning condition exists and the battery room ventilation has
malfunctioned. Verify detector power supply UP-1H, breaker 14
and circuit breaker located in detector cabinet are closed.
b. A warning condition exists. Manually start the battery room
exhaust fan and monitor hydrogen readings closely.
c. A danger condition exists. Secure the battery charge, manually
start the battery room exhaust fan, and rig temporary
ventilation.
d. A danger condition exists and the battery room ventilation has
malfunctioned. Secure the battery charge and rig temporary
ventilation.
~ ..- - . .. - .. . - ... . -.. - . . _ -
. - - - . ... ..- . - . - - - - . -
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SENIOR REACTOR OPERATOR Page 12
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j QUESTION: 010 (1.00)
- Work needs to be performed in a confined space with the following i
-
potential atmospheric conditions: !
I
i - Oxygen levels of 20%
- Flammable gas levels at 30% of the lower flammability limit '
- - Toxic gas 16 vest at 40% of the OSHA Permissible Exposure Limit
!
(PEL).
1 i
4
'
Non-atmospheric /non-energy or energy hazards DO NOT exist in the space. '
a !
l WHICH ONE of the following describes the permit requirements for the I
i sp. ace?
l
1
a. No permit is required for the given potential atmospheric l
- conditions. t
- !
,
!
j b. No permit is required if the potential atmospheric hazards can
be isolated or eliminated with a tagout.
I c. A permit is required, but the space can be temporarily ;
downgraded to a non-permit required confined space if the
atmospheric hazards can be controlled with forced ventilation. i
d. A per.1 nit is required, but the atmospheric hazards must be
contrailed with forced ventilation.
l
!
l
I
!
_ - - _ - - - - -
_ - ,
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SF.NIOR REACTOR OPERATOR Page 13
1
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!
,
QUESTION: 011 (1.00)
l
A continuous fire watch is required due to inoperable fire protection
equipment.
WHICH ONE of the following is NOT an acceptable method for performing a
<
l a. One of the three licensed ROs on-shift acts as a continuous fire ;
watch in the control room while performing his normal duties,
b. A qualified HP technician acts as a continuous fire watch while
monitoring a maintenance activity in the reactor building.
m
c. A non-licensed operator acts as a continuous fire watch in two
'
adjacent areas of the turbine building by moving from cne area
to the other every fifteen minutes.
d. A qualified security guard acts as a continuous fire watch in
the screen house and also performs five minute security patrols ,
of a nearby area twice an hour.
QUESTION: n12 (1.00)
WHICH ONE of the following describes the requirements for performance of
a " Continuous Level of Use" procedure?
a. The procedure steps must be read by the operator performing the
task. The procedure steps cannot be read by another person.
b. The procedure steps may be read by another operator, but both
operators must be at the job site.
c. The procedure steps may be read by another operator at a
different location using telecommunications as long as a copy of
tne procedure is also located at the job site.
d. The procedure steps may be read by another operator at a
different location using tel3 communications. It is not
necessary to have a copy of the procedure at the job site as
long as a copy of the procedure is in proximity to the job site.
,
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SENIOR REACTOR OPERATOR Page 14
QUESTION: 013 (1.00)
WHICH ONE of the following describes equipment specifically associated
with the Emergency Operating Procedures?
a. Green annunciator windows.
b. Instruments with green bordered labels.
c. Annunciator windows with yellow and black borders.
d. Components with yellow and black bordered labels.
.
QUESTION: 014 (1.00)
WHICH ONE of the following describes how safety system out of service
time is normally calculated for the purpose of safety system performance
indicator monitoring?
a. Out of service time is the difference between the time that the
LCO was entered and the time that the LCO was exited.
b. Out of service time is the difference between tl:e time that the
LCO was entered and the time that the tags were cleared,
c. Out of service time is the difference between the time that the
tags were hung and the time that the LCO was exited.
d. Out of service time is the difference between the time that the
tags were hung and the time that the tags were cleared.
.
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SENIOR REACTOR OPERATOR Page 15
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l
QUESTION: 015 (1.00)
I A licensed SRO is needed to assist with special control rod testing.
l The test is scheduled to begin at 0600 on Sunday morning and should be
l completed by 1600 the same day. The SRO has worked the following hours
on-shift over the last week:
Sunday - 0600 to 1600
Monday - 0600 to 1600
Tuesday - Off
Wednesday - 0600 to 1800
Thursday - 0800 to 2000
Friday - 0800 to 2000
. Saturday - 1200 to 0000
WHICH ONE of the following describes the approval that is required for
the SRO to assist with the special test?
a. No special approval is necessary because the overtime limits
will not be exceeded.
b. No special approval is necessary because the SRO will not be
performing licensed duties.
c. Operations Manager approval is necessary prior to exceeding the
overtime limits.
d. Operations Director approval is necessary prior to exceeding the ;
overtime limits.
i
_ _______._.__.__....__..__________.-_m. . _ _ _ . _ _ . _
- Page 16
SENIOR REACTOR OPERATOR
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,
QUESTION: 016 (1.00)
0630 on 12/2/96 Inoperable equipment was identified that required the plant to be in
Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with Technical I
Specifications (T/S).
'
f
0800 on 12/2/96 Attempts to repair the inoperable equipment were unsuccessful and
l .
I
an orderly shutdown was initiated.
I
0630 on 12/3/96 Reactor coolant temperature was 212*F and the mode switch was
place in SHUTDOWN.
1
0700 on 12/3/96 Reactor head vents were opened.
1
!
WHICH ONE of the following describes the required notifications to the state and unusual -
l
event declaration for this situation?
a. The state must be notified of the T/S shutdown by 0730 on
12/2/96. No unusual event declaration is required,
b. The state must be notified of the T/S shutdown by 0900 on
12/2/96. No unusual event declaration is required.
c. The state must be notified of the T/S shutdown by 0730 on
12/2/96. An unusual event must be declared on 12/3/96.
d. The state must be notified of the T/S shutdown by 0900 on
12/2/96. An unusual event must be declared on 12/3/96.
i
n
l
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._ _ , , _ . _ _ _ _ _ . _ - . _ - _ . _ . ,
. - - - . _ . - . . ~. - . - . . . . . . - - _ . . . - - - ~ - . - . - _ - ---. . . - . ...
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SENIOR REACTOR OPERATOR Page 17 l
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+
OUESTION: 017 (1.00) l
A LOCA has occurred and a General Emergency - Bravo has just been )
declared. No release is in progress and it cannot be determined if and l
when a release wil: occur. The meteorological specialist has determined i
a projected plume direction. No constraining conditions exist. l
The Manager of Radiological Dose Assessment (MRDA) has recommended !
evacuation in an approximate 2-mile radius. !
WHICH ONE of the following describes the assessment that the DSEO should
make when reviewing the Protective Action Recommendation (PAR)
- recommended by the MRDA? ,
l
l . a. The recommended PAR is appropriate for these conditions.
b. The recommended PAR is not appropriate for these conditions '
because sheltering is preferred when a release is not in
progress or imminent.
c. The recommended PAR is not appropriate for these conditions
because evacuation should not be recommended unless a General ,
Emergency - Alpha has been declared. j
d. The recommended PAR is not appropriate for these conditions
because the keyhole approach should have been used to recommend
actions for the downwind sectors rather than for the entire
ring.
l
l
(
l
,
- , , - . - , - ,, 4.a y .p- g - ,,-+,__-# -
w s,w- -
.,....9 .e + y
. . . _ _ _ _ _ . ._ _. _ _. _ .. _ _ ____ _ _. _ _ _ _ _.
_ _ _.. .. _ . . . . . .
.
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SENIOR REACTOR OPERATOR Page 18
i. l
,
i QUESTION: 018 (1.00)
l
! The plant is operating at 100% power. RWCU was operating in the normal
lineup with RWCU pump B running when the MCC-F3 feeder breaker from 12F !
tripped open due to a fault. ,
WHICH ONE of the following describes the expected RWCU systern response?
l
l
a. Fl.'/CU pump B continues to run. No system isolation signal is ,
l generated. l
l
'
b. RWCU pump B trips. No system isolation signal is generated. ,
I
! , c. RWCU pump B trips. A system isolation signal is generated, but
l
all system isolation valves do not close.
r .
)
l
d. RWCU pump B trips. A system isolation sign::! is generated and
! all system isolation valves close.
i
!
QUESTION: 019 (1.00) -
l
SBLC pump A is out of service for maintenance with the electrical supply
breaker open and the control power fuses removed. An ATWS has occurred
and SBLC was initiated ten minutes ago by placing the SBLC initiation
keylock switch to the SYS 1 position and then to the SYS 2 pnsition.
Initial fm.C tank level was 60% Reactor pressure is 1100 psig and
reactor pwer is 25%
WHICH ONE of the following conditions would indicate IMPROPER SBLC
system operation?
a. Annunciator "SLC TANK LOW LEVEL" is illuminated.
J
b. SBLC pump discharge pressure indicates 1400 psig.
c. Annunciator "SQUlB VALVE FIRED OR LOSS OF POWER" is illuminated.
d. Amber light for squib valve A is not lit.
l
. --
- - - . - _
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SENIOR REACTOR OPERATOR Page 19
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l
I
OUESTION: 020 (1.00) '
)
l
The plant is operating at 85% power. The following are the indications
'
received when the APRM meter function switches on Panel CRP 937/938 are
placed in the AVERAGE, COUNT, and FLOW positions:
AVERAGE COUNT FLOW
APRM 1 94 % 70 % 71 %
APRM 2 74 % 55 % 71 % '
APRM 3 86 % 80 % 71 %
APRM 4 99 % 55 % 63%
APRM 5 83 % 65 % 63 %
l
APRM 6 75% 70 % 63 %
WHICH ONE of the following describes the expected plant response for
these conditions?
a. Rod block only ,
b. Half scram only
c. Rod block and half scram
d. Rod block and full scram
,
1
___._. -._ . _ _ _ . _ . _ _ _ _._ _ . _ _.___ . _. _ _ . . . _ . . _ . . _ _ _ _ . _ . .__
.._.7._______
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SENIOR REACTOR OPERATOR Page 20 !
!
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i
QUESTION: 021 (1.00)
APRM 4 is reading 82% with the following plant conditions:
!
Core Thermal Power = 85%
Total Recirc Flow = 70 %
Total Core Flow = 75 %
i
! MFLPD = 0.92
l FRP = 0.85 i
'
! MFLCPR = 0.68
WHICH ONE of the following is the correct action concerning APRM 47
a. APRM 4 readings should be lowered. ,
"
!
l b. APRM 4 readings should be raised.
l
c. APRM 4 readings should not be adjusted at the current power
level.
l
1
d. APRM 4 readings should not be adjusted because the readings are
within 10% of actual power. ;
i
i
QUESTION: 022 (1.00)
A plant startup is in progress with the mode switch in STARTUP/ HOT
STANDBY. IRM 12 is bypassed for maintenance. While preparing to place l
the mode switch in RUN, IRM 17 failed downscale and no automatic actions i
occurred,
WHICH ONE of the following describes the required procedural and/or
technical specification actions for these conditions?
a. Bypass IRM 17. The startup may continue,
b. Insert a rod block. The startup cannot continue until IRM 17 is
repaired.
c. Insert a rod block and a half scram on the affected RPS channel.
The startup cannot continue until IRM 17 is repaired.
,
d. Initiate insertion of operable rods and complete insertion of
l all operable rods within four hours.
!
. - .
.. __. . .. . . .
- - . - - . -.. - - - - - - _ .. - - .- . - - .. .-. ~.-_- -.. .-. ..
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SENIOR REACTOR OPERATOR Page 21
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>
i
QUESTION: 023 (1.00) l
!
The plant is in a refueling outage. SRMs 21 and 23 are inoperab!e.
! WHICH ONE of the following activities CANNOT be performed? Special l
!
dunking type detectors are NOT going to be used. A core map is l
provided. l
t
a. Removal of control rod 30-19. j
b. Loading of a new fuel assembly into core location 15-20. [
\ !
l. c. Removal of a spent fuel assembly from core location 47-44. !
1
L
. d. Removal of LPRM atring 20-33. !
t
.
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!
I
i
QUESTION: 024 (1.00)
During control rod surveillance testing at 100% power, the 'A'
Recirculation pump tripped. All of the appropriate actions were taken !
'
in response to the recirc pump trip. The following annunciators are
received on CRP 905: l
"RBM DOWNSCALE" !
" ROD OUT BLOCK" !
WHICH ONE of the following could be the cause of these alarms?
a. A voltage transient during the recire pump trip caused the
output of the averaging circuit for one of the RBMs to
momentarily lower below 3/125 of scale.
b. Local power around the selected rod decreased causing the RBM
averaging circuit output signal to decrease below 22% of the
reference APRM signal.
c. Local power around the selected rod decreased causing the count
circuit to determine that more than half of the LPRM inputs to
the RBM were downscale.
,
d. Flow Converter 1 generated a downscale signal to RBM 8.
. _ _ - - - - - . .-
.- __.
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SENIOR REACTOR OPERATOR Page 22
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QUESTION: 025 (1.00)
During a plant shutdown, feed flow is 18% and steam flow is 21%. The
following information is displayed on the RWM message display:
SR 26-07:16 SE IB WB
WE 14-19:08 A 2 - 316
i
'
l IE 06-15:08
.
WHICH ONE of the following describes actions that can be taken by the
operator due to the RWM?
I
a. Any rod can be inserted or withdrawn because the plant is in the RWM +
transition zone
b. RWM is enforcing the control rod sequence rod 30-03 can be selected and
inserted.
c. RWM is enforcing the control rod sequenca, rod 26-07 can be selected and
inserted. i
d. RWM is enforcing the control rod sequence, rod 14-19 can be selected and l
inserted. !
l
_ . _ _ . . _ _ _ _ . . _
- _. ._ _ ._ .__ __ ... _ _ _ __
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SENIOR REACTOR OPERATOR Page 23 .
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!
'
QUESTION: 026 (1.00)
During a plant shutdown with the reactor at 15% power, the CRP 905
operator is inserting rods using single notch rod insertion. As rod
'
10-35 is being inserted, an RPIS failure occurs sending an INOP signal
to the Reactor Manual Control System.
WHICH ONE of the following describes the expected response of rod 10-35 ,
and the operator's ability to continue inserting rods?
a. Rod 10-35 continues to insert and settles at the next notch.
The operator can continue to insert rod 10-35 using the Rod i
Movement Control Switch or the EMERGENCY IN switch. The
operator cannot select any other rods for insertion.
b. Rod 10-35 continues to insert and settles at the next notch. - l
The operator can continue to insert rod 10-35 using only the
EMERGENCY IN switch, but cannot select an" other rods for
insertion.
c. The insert cycle is halted and rod 10-35 scttles at the original
notch. The operator cannot insert rod 10-35 or any other rods
any further using RMCS. ,
i
d. Rod 10-35 is immediately deselected and settles at the original I
notch. The operator cannot insert rod 10-35 or any other rods
any further using RMCS.
QUESTION: 027 (1.00)
The CRD system was operating normally with the flow control valve in
auto when the flow control valve mechanically bound.
WHICH ONE of the following describes the expected effect on CRD system
response if the CRD FCV is mechanically bound?
a. During RPV depressurization, there could be difficulty with rod
motion, affecting rod speeds and unlatching.
b. During RPV depressurization, there could be excessive rod speeds
and rod drifts outward.
c. During a scram, CRD pump runout could occur.
d. During a scram, there could be slow scram times.
. . . . - . . . - . - -- ._ - . . . = . - - . - . . - . - . . . . . . -
. , . . - - . -
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SENIOR REACTOR OPERATOR Page 24 i
>
!
!
QUESTION: 028 (1.00)
!
On the initiation of a reactor scram from 100% power, the insert line of
CRDM 22-19 breaks completely off at the insert port weld. l
WHICH ONE of the following describes the expected response of rod 22-19 l
'
to the scram?
a. Will immediately start to insert and will fully insert with !
reactor pressure only.
b. Willimmediately start to insert and will fully insert with
accumulator pressure only.
c. Will not insert until reactor and accumulator pressures equalize
and willinsert with reactor pressure only. -
.
. .
.
I
i
!
u
F
l
1
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___ . . _ . - . _ _ _ . _ _ . . . . _ . _ _ _ _ . _ _ . _ _ _ _ . _ _ . _ . . _ . . - . _ _ _ . _ _ _ . _ .. ._.
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SENIOR REACTOR OPERATOR Page 25
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l QUESTION: 029 (1.00)
'l
)
, Scram time testing has just been completed. The following data was j
,
collected for the 90% insertion times for the following 2x2 arrays: l
ROD SCRAM TIME ROD SCRAM TIME :
i i
14-11 4.075 sec 38-43 3.255 sec j
{ 14-15 3.348 sec 38-47 7.524 sec j
1
18-11 3.462 sec 42-43 3.186 sec '
- 18-15 5.121 sec 42-47 3.437 sec l
i l
5 AVERAGE SCRAM '
l
l TIME FOR 3
l FASTEST IN -
l
4
ARRAY 3.628 sec 3.292 see -
- l
l WHICH ONE of the following describes the minimum actions required to
,
continue reactor operation? (Assume all other control rods are operable
l and have scram insertion times within limits.) I
a. No action is required to continue reactor operation.
?
4
b. Rod 38-47 must be fully inserted and electrically disarmed.
c. Rods 18-15 and 38-47 must be fully inserted and electrically
disarmed.
d. Rods 14-11,18-15, and 38-47 must be fully inserted and
electrically disarmed.
. . , . - . - .- . - . - - - -_ - . -- - . _ - _ _ -
', ' SENIOR REACTOR OPERATOR Page 26
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1
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QUESTION: 030 (1.00) ;
! WHICH ONE of the following correctly describes reactor core flow
orificing?
$ a. Orificing is used to provide increased coolant flow in the lower i
power fuel bundles which experience a higher resistance to flow i
j than the higher power fuel bundles. t
l b. Orificing is used to provide increased coolant flow in the [
l
higher power fuel bundles which experience a lower resistance to j
i flow than the lower power fuel bundles. ,
4 , c. Orificing is used to prevent decreased coolant flow in the l
4 central fuel bundles which experience a higher resistance to I
flow than the peripheral fuel bundles.
'
d. Orificing is used to prevent decreased coolant flow in the
- peripheral fuel bundles which experience a lower resistance to
! flow than the central fuel bundles.
J
i
i
QUESTION: 031 (1.00)
'
The plant is operating at 100% power with the 'A' side instruments
j selected for RPV level control.
WHICH ONE of the following describes the plant response, without
.
operator action, to a break in the reference leg that supplies the 'A'
I
side NR GEMAC, WR Yarway, and Fuel Zone levelinstruments?
i
i a. Feedwater FRVs open. RPV water level increases and the feed
- pumps and turbine trip on actual high water level. The reactor
i scrams due to the turbine trip.
2
3 b. Feedwater FRVs are not affected. The reactor scrams due to
indicated level below the scram setpoint.
c. Feedwater FRVs close. RPV water level decreases and the reactor
4
4
scrams on actual low water level. ,
! d. The feed pumps and turbine trip due to indicated level above the
i trip setpoint. The reactor scrams due to the turbine trip.
,
!
!
_ . _ - - _ . _ _ _ _ . . _ . _ _ . . _ . - . _ . - _ _ _ _ _ _ . _ _ _ . _ . _ _ . _ _ _ . _ _ _ . _ _ . . _ _ _ _
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SENIOR REACTOR OPERATOR Page 27 :
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QUESTION: 032 (1.00)
'
!
t Feedwater pump suction pressure is decreasing following a FWCl ;
! initiation. WHICH ONE of the following describes the earliest l
!
,
point in the transient that the feedwater pump will trip?
,
,
1
j a. Feedwater pump suction pressure is less than 335 psig for 10 i
i seconds.
!
- b. Immediately when feedwater pump suction pressure is less than ,
i' 335 psig. '
!
! c. Feedwater pump suction pressuro less than 85 psig for 10
i seconds.
i
j d. Immediately when feedwater pump suction pressure is less than 85 ;
1 psig.
I
!
4 QUESTION: 033 (1.00)
i
I The Feedwater sye+em is being secured during a plant shutdown.
!
- WHICH ONE of the following lineups will satisfy FWCl requirements?
i
'
a. String A selected for FWCl. FRV-5A in automatic and FRV-5B in
balance. FRV-4A shut and FRV-4B open.
b. String A selected for FWCI. FRV-5A in balance and FRV-5B in
autome. tic. FRV-4A shut and FRV-4B shut,
c. String B selected for FWCl. FRV-5A in manual and FRV-5B in
automatic. FRV-4A open and FRV-4B shut.
d. String B selected for FWCl. FRV SA in manual and FRV-5B in
manual. FRV-4A open and FRV 5B open.
,
!
. . . _ _ _ _ . . _ . _ _ - _ _ _ . . . . _ . _ _.__ .. _.._.m . _. _ _ . _
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SENIOR REACTOR OPERATOR Page 28 f
I
QUESTION: 034 (1.00)
The isolation condenser automatically initiated following an inadvertent
MSIV closure. Later the isolation condenser automatically isolated and
the control switches for all of the IC isolation valves were placad in
l the CLOSE position. The IC initiation signal has cleared. [
1 ;
WHICH ONE of the following correctly describes the operation of the i
i isolation condenser valves under these conditions?
5
.
a. If the switches for valves IC-1, IC-2, IC-3, IC-4, and IC-6/7 !
are placed in AUTO or AUTO-OPEN as applicable, the !
initiation signal can be reset. -i
.
!
b. If the control switch for IC-4 is taken to the AUTO position
after the initiation and isolation signals have been reset, IC-4 *
.I
will not open.
c. If the control switch for IC-6/7 is taken to the AUTO-OPEN i
position before the initiation and isolation signals have been ,
reset, IC-6/7 will open.
j
d. If the switches for valves IC-1, IC-2, IC-3, and IC-4 are taken
to the OPEN position before the isolation signal has been
reset, the valves will open.
!
QUESTION: 035 (1.00)
i
WHICH ONE of the following describes the technical specification !
required actions if one IC steam line flow differential pressure detector l
has failed downscale.
a. No action is required,
b. Place the affected Group IV isolation channel in the tripped
condition until the instrument is repaired.
c. Close the IC isolation valves immediately. Reduce reactor power
below 40% within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the instrument cannot be repaired.
,
d. Close the IC isolation valves immediately. Initiate an orderly
shutdown and reduce reactor coolant temperature below 330*F
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the instrument cannot be repaired.
. _
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SENIOR REACTOR OPERATOR Page 29
QUESTION: 036 (1.00)
A smallloss of coolant accident has occurred coincident with a loss of
normal power. At T = 0 sec, the APR 120 second timer initiated on a valid
signal.
In WHICH ONE of the following situations will the APR valves open at
120 seconds?
a. The diesel generator failed to start. All other automatic
actions occurred as expected due to the LOCA signal. The " Auto
Blowdown Sys Fuse Monitor" annunciator came in at T = 30 seconds
and is stillilluminated.
.
b. All automatic actions occurred as expected due to the LOCA
signal. At T = 30 seconds drywell pressure decreased to 1.2 '
psig and the operator placed the Drywell High Pressure Reset
Switch to the RESET position.
c. The gas turbine failed to start. All other automatic actions
occurred as expected due to the LOCA signal. At T = 60 seconds
the operator placed the APR System Bypass Switch in the BYPASS
position momentarily and then returned the switch to the NORMAL
position. :
l
d. All automatic actions occurred as expected due to the LOCA
signal. The diesel generator and the gas turbine tripped at T = 1
60 seconds.
l
!
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_. _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ . __ __ _ _ _ _ _
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!
OUESTION: 037 (1.00) ;
I
A transient occurred which caused RPV water level to decrease. RPV !
water level dropped below the low-low level setpoint and the Core Spray l
pumps automatically started. The following conditions existed as the j
transient progressed: i
'
l i
l T= 10 min The Core Spray pump control switches were taken to the STOP position
- RPV water level was -75 inches !
RPV pressure was D00 psig j
!
l
T= 15 min Drywell pressure increased above the high pressure setpoint {
j RPV water level was -100 inches i
!
RPV pressure was 800 psig
WHICH ONE of the following describes the expected response of the Core !
l Spray pumps to the transient? (
,
a. The Core Spray pumps continue to run throughout the transient.
b. The Core Spray pumps stop momentarily at T = 10 minutes and i
immediately restart when the operator releases the control ;
switch.
l
c. The Core Spray pumps stop at T= 10 minutes and restart at T= 15
minutes.
d. The Core Spray pumps stop at T= 10 minutes and do not restart.
QUESTION: 038 (1.00)
A LOCA is in progress and all of the low pressure ECCS pumps are
operating and cavitating. RPV level is less than TAF.
l WHICH ONE of the following correctly describes the first action that
should be taken to maximize ECCS pump performance?
a. Core Spray flow should be decreased until cavitation stops.
l
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b. Decrease Core Spray pump flow to 3600 per pump. l
c. LPCI flow should be decreased to 5000 gpm total flow.
- d. LPCI flow should be decreased until cavitation stops or
l until LPCI flow is 1000 gpm/ train.
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. SENIOR REACTOR OPERATOR Page 31 I
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] QUESTION: 039 (1.00) l
- .
! A valid LOCA initiation signal was received 5 minutes ago.
!
- Feedwater, condensate and core spray systems are NOT available.
All other automatic actions occurred as expected. ,
1
- Jet pump A riser manifold pressure is lower than jet pump B riser !
j manifold pressure.
4
- RPV level is -60" and decreasing.
- - Reactor pressure is 800 psig and decreasing.
!-
2 Section ALC of EOP 570 has been entered and the SRO needs to determine
if at least 2 injection subsystems can be lined up with pump (s) running
in accordance with step ALC-4.
, WHICH ONE of the following describes the status of the LPCI subsystems?
1
i l
l a. Only LPCI A can be considered lined up for injection due to LPCI l
- loop selection. l
!
b. Only LPCI B can be considered lined up for injection due to LPCI
] loop selection.
c. LPCI A and B can be considered lined up for injection, because
- the non-selected injection valve and crosstie valves can be
j manipulated to separate the subsystems after loop select
interlocks have timed out.
d. Neither LPCI A and B can be considered lined up for injection,
[ because reactor pressure is above the LPCI shutoff head. )
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SENIOR REACTOR OPERATOR Page 32
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QUESTION: 040 (1.00)
With the plant operating at 100% power, an operator performing a valve
lineup inadvertently shuts LPCI system cross-tie valve, LP-8A and it
remains uncorrected.
WHICH ONE of the following describes the effect, if any, of shutting
LP-8A on LPCI system operation if an initiation signal is received?
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a. Shutting LP-8A could result in damage to LPCI A due to water
hammer, because system keepfull is isolated from LPCI A.
b. Shutting LP-8A could result in damage to LPCI B due to water
hammer, because system keepfull is isolated from LPCI B.
c. Shutting LP-8A will not affect LPCI system operation, because it
is designed to shut automatically in response to LPCIloop
selection.
d. Shutting LP-8A will not affect LPCI system operation, because it
is designed to open automatically on an initiation signal.
QUESTION: 041 (1.00)
Following a LOCA, drywell sprays and torus sprays were in service on
both LPCI subsystems. When directed to secure containment sprays, the
operator placed the control switches for the following valves
momentarily in the CLOSE position: l
LP-13A - Outboard Torus Spray Valve !
LP-148 - Inboard Torus Spray Valve
LP-15A - Outboard Drywell Spray Valve
LP-168 - Inboard Drywell Spray Valve
WHICH ONE of the following lists the containment sprays that are still
in service?
a. No containment sprays.
b. Torus spray only.
c. Drywell spray only.
d. Torus spray and drywell spray.
._ _ ._. . __ _ . . _ . _ - . _ . _ _ . _ . _ _ . _ _ . _ . _ _ ._ _ _
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. SENIOR REACTOR OPERATOR Page 33
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QUESTION: 042 (1.00)
1
The plant is operating at 100% power. l&C has reported that the
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following torus water level transmitters failed their monthly instrument
,
checks: !
'
Ll 1602A on CRP 925 (-10" to + 10")
i LT-1630A PAM recorder on CRP 925 (2.2' to 22.2').
!
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The remaining torus water level transmitters and the drywell water level
- transmitter have been demonstrated OPERABLE.
I
j WHICH ONE of the following describes the most limiting technical
specification required actions?
a. Torus water levelinstrumentation calibration frequency must be
I increased to once per month. Plant operations can continue
-
indefinitely because both transmitters supply the A channel.
l
b. LT-1630A must be restored within 7 days or the plant must be in
4 at least 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 />.
1
i c. LT-1630A or the transmitter that supplies Ll-1602A must be
restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or the plant must be in at least HOT
j SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i
d. The transmitter that supplies LI-1602A must be restored within 6
hours or a shutdown must be initiated and the plant must be in
l COLD SHUTDOWN or REFUEL within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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OUESTION: 043-(1.00) 5
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An event has occurred and the following conditions are noted: l
The reactor building is at atmospheric pressure
Torus pressure is 0.5 psig negative l
Drywell pressure is 0.3 psig. ;
l
WHICH ONE of the following describes the expected status of I
l
DRYWELL VAC RELIEF VLV 1-A OPEN (CRP 903 A2 window 1-3) and -
TORUS VACUUM BREAKER OFF SEAT (CRP 903 A3 window 4-7)? .
1
(The applicable ARPs are provided.)
a. DRYWELL VAC REllEF VLV 1-A OPEN -lit l
TORUS VACUUM BREAKER OFF SEAT -lit
b. DRYWELL VAC RELIEF VLV 1-A OPEN -lit
TORUS VACUUM BREAKER OFF SEAT - not fit
c. DRYWELL VAC RELIEF VLV 1-A OPEN - not lit
TORUS VACUUM BREAKER OFF SEAT -lit. ]
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d. DRYWELL VAC RELIEF VLV 1-A OPEN - not lit i
TORUS VACUUM BREAKER OFF SEAT - not lit.
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SENIOR REACTOR OPERATOR Page 35
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QUESTION: 044 (1.00)
WHICH ONE of the following correctly describes the interlock between
containment vent to SBGT, valve AC-10, and the SBGT system?
a. On an automatic SBGT start, valves SG-1 A and SG-1B will only
'
open if the AC-10 control switch is in the CLOSE position.
b. On an autometic SBGT start, AC-10 will close, if open, as soon
as valve SG-1 A or SG-1B leaves its fully closed seat.
c. Following an automatic start, with SBGT running, AC-10 will open
if the keylock switch on CRP 905 is in the TORUS or DRYWELL
position and the AC-10 control switch is taken to AUTO /OPEN.
'
d. Following an automatic start, with SBGT running, AC-10 xvill l
auto open when either SG-1 A or SG-1B are open but will
auto close if both SG-1 A and SG-1B are open. I
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QUESTION: 045 (1.00)
The Reactor Building "HVAC ISOLATIONS-RX BLDG" keylock switches (color
turquoise) have been placed in " BYPASS" on CRP 902 as directed by EOP
590.17.
WHICH ONE of the following describes the function provided when these
switches are placed in the " BYPASS" position.
a. Allows operation of the reactor building transfer fans with any
isolation signal present,
b. Allows operation of the reactor building transfer fans if the
exhaust fans are secured.
c. Allows operation of the reactor building supply fans with high
drywell pressure and low RPV water level. j
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d. Allows operation of the reactor building supply fans with high ,
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reactor building ventilation exhaust radiation.
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. SENIOR REACTOR OPERATOR Page 36
QUESTION: 046 (1.00)
Given the following conditions:
- The reactor mode switch is in REFUEL.
- One rod is withdrawn.
- The refueling platform is approaching the vessel and the following
I
lights are lit on the operator's console:
J
" GRAPPLE NORMAL UP"
,
" GRAPPLE ENGAGED"
" HOIST LOADED"
WHICH ONE of the following describes the available operation of the main
hoist when the refuel platform interlock travellimit switches are
tripped?
a. The main hoist can be raised or lowered.
i
b. The main hoist can be raised only. !
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c. The main hoist can be lowered only.
- d. The main hoist cannot be raised or lowered.
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. SENIOR REACTOR OPERATOR Page 37
QUESTION: 047 (1.00)
During a shutdown for refueling, rod 06-03 was stuck at position 12.
The control rod was electrically disarmed. All other rods were fully
inserted at 1000 on 11/04/96 and the RE determined that the reactor
would remain shutdown under all conditions without boron with rod 06-03
stuck out. Troubleshooting was performed and rod 06-03 was fully
inserted at 1400 on 11/08/96.
Fuel movement is in progress. RBCCW outlet temperature is 70 F. TBCCW
outlet temperature is 65 F.
WHICH ONE of the following is the earliest time that SDC can be secured?
(Attachment 1 of OP 305 is provided.)
a. 698 hours0.00808 days <br />0.194 hours <br />0.00115 weeks <br />2.65589e-4 months <br /> from 1000 on 11/04/96.
b. 863 hours0.00999 days <br />0.24 hours <br />0.00143 weeks <br />3.283715e-4 months <br /> from 1000 on 11/04/96.
c. 698 hours0.00808 days <br />0.194 hours <br />0.00115 weeks <br />2.65589e-4 months <br /> from 1400 on 11/08/96.
d. 863 hours0.00999 days <br />0.24 hours <br />0.00143 weeks <br />3.283715e-4 months <br /> from 1400 on 11/08/96.
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QUESTION: 048 (1.00)
, Shutdown cooling has just been placed in service with both SDC pumps
running. Recirc pump B is running and RPV water level is +40 inches.
System temperatures are as follows:
Recirculation loop A temperature: 265 F
Recirculation loop B temperature: 260 F ;
SDC heat exchanger tube side outlet temperature: 200 F
RBCCW OUT S/D CLNG HX: 145 F
WHICH ONE of the following correctly describes the status of system
temperatures?
a. Shutdown cooling inlet temperature is indicative of reactor core
temperatures.
RBCCW OUT S/D CLNG HX temperature is within limits.-
b. Shutdown cooling inlet temperature is indicative of reactor core
temperatures.
RBCCW OUT S/D CLNG HX temperature is not within liraits.
c. Shutdown cooling inlet temperature is not indicative of reactor
t
core temperatures.
RBCCW OUT S/D CLNG HX temperature is within limits.
d. Shutdown coolir.g inlet temperature is not indicative of reactor
core temperatures.
RBCCW OUT S/D CLNG HX temperature is not within limits.
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QUESTION: 049 (1.00)
The plant is in single loop operation at power.
For WHICH ONE of the following conditions can the idle recirculation
pump be started? (Consider administrativo and functional limitations.)
a. The idle recirculation MG set generator field breaker is closed.
MG set lockout relays have been reset.
b. The idle recirculation MG set lube oil pump discharge pressure
is 55 psig. Lube oil temperature is 90 F.
c. The idle pump was secured and isolated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ago. The pump
was unisolated one hour ago. The idle pump suction temperature
is 408 F. The operating pump suction temperature is 435 F.
d. The idle recirculation pump suction and discharge valves are
fully open. Differential pressure across the pump is 1.7 psid.
The scoop tube lockup light is off and speed control is in
manual.
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QUESTION: 050 (1.00)
i
Recirc pump "B" M/A transfer station was in MAN with the output meter 1
l reading 50%. The vertical percent speed meter on the M/A station was ,
l reading 67% and the pump was running at approximately 67% speed.
I
WHICH ONE of the following describes the expected indication on the i
recire pump "B" M/A transfer station output meter, vertical percent
speed meter, and actual pump speed if the operator simultaneously
depressed the ATWS "A" and "C" pushbuttons?
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a. The output meter is reading 50%. ;
The vertical percent speed meter is reading 67%. j
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- The pump is running at 67% speed. i
b. The output meter is reading 50%. *
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The vertical percent speed meter is reading approximately 80%. !
The pump is not running.
c. The output meter is reading 50%.
The vertical percent speed meter is reading offscale low.
The pump is not running. I
d. The output meter is reading offscale low.
The vertical percent speed meter is reading offscale low.
The pump is not running,
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QUESTION: 051 (1.00)
WHICH ONE of the following correctly describes the purpose of the upper
and townr limits established by the recirc speed dual limiter?
a. The upper limit prevents exceeding MCPR limits. The lower limit
I prevents damage to the recirc pumps and motors caused by axial
! thrust.
b. The upper limit prevents darnage to the recirc pumps and motors
caused by axial thrust. The lower limit prevents two pump ;
operation in the thermal hydraulic instability region of the
power to flow map.
c. The upper limit insures stability of the hydraulic cciupling.
'
The lower limit reduces the effects of vessel bottom head
thermal stress due to stagnation,
d. The upper limit prevents damage to the recirc pumps due to
excessive flow and vibration. The lower limit insures stability
of the hydraulic coupling.
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QUESTION: 052 (1.00)
While in the Gas Turbine Building on a plant tour, you discover a small
fire in the area of the generator lube oil pumps. You are a qualified
fire fighter. ;
WHICH ONE of the following describes the actions that you should take in '
response to the fire?
I
a. Leave the area immediately then report the fire to the Control
Room. The carbon dioxide system will automatically initiate to
extinguish the fire.
b. Leave the area immediately then report the fire to the Control
Room. The deluge system will automatically initiate to
extinguish the fire.
c. Report the fire to the Control Room then wait for the fire
brigade to arrive before attempting to fight the fire.
d. Report the fire to the Control Room then attempt to fight the
fire as safely as possible.
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SENIOR REACTOR OPERATOR Pago 42
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QUESTION: 053 (1.00)
,
, A LOCA has occurred concurrent with a loss of normal power. The gas
turbine is currently loaded to 11.2 MWe. Gas turbine exhaust
,
temperature is increasing. If gas turbine exhaust temperature continues
to increase WHICH ONE of the following actions will occur first?
- a. The gas turbine will automatically runback,
b. The 14A-G tie breaker will trip open.
c. The gas turbine will trip,
d. The running FWCl feedwater pump supply breaker will trip open.
QUrdTION: 054 (1.00)
The following events occurred:
T= 0 Power was lost to the Vital AC MG set DC motor.
T = 5 min Power was lost to the Vital AC MG set AC motor. The MG set
stopped and the Vital AC bus automatically transferred to the
alternate power supply.
T = 10 min Power was restored to the Vital AC MG set DC motor.
T= 15 min Power was restored to the Vita! AC MG set AC motor.
WHICH ONE of the following describes the effoct of these events on the
Vital AC bus? Assume only operator actions were the restoration of
power to the AC and DC motors,
a. The MG set automatically restarted at T = 10 min and the Vital AC
bus automatically transferred back to the MG set.
b. The MG set automatically restarted at T= 10 min and the Vital AC
bus remained on the alternate power supply.
c. The MG set automatically restarted at T = 15 min and the Vital AC
bus autornatically transferred back to the MG set.
d. The MG set automatically restarted at T = 15 min and the Vital AC
bus remained on the alternate power supply. 1
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QUESTION: 055 (1.00) ;
t
RPS Bus B is being supplied by transformer IRP-1 while maintenance is t
being performed on the normal supply 480 VAC motor control center.
l
During performance of a core spray surveillance, a fault caused Bus 14E '
to drop to 1500 volts for 3 seconds when the core spray pump was ;
started. '
i
WHICH ONE of the following describes the expected effect on the RPS !
busses? j
i
a. Both RPS busses remain energized. !
. b. RPS Bus A remains energized. RPS Bus B deenergizes resulting in f
a half scram. i
c. RPS Bus B remains energized. RPS Bus A deenergizes resulting in
a half scram.
d. Both RPS busses deenergize resulting in a full scram. j
.
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OUESTION: 056 (1.00)
The plant was operating at 100% power when a generator load reject ;
occurred. All TCVs and turbine BPVs responded as expected with the ,
exception of BPV #4 which stuck fully closed. !
WHICH ONE of the following describes the expected automatic actions as a
direct result of the transient?
a. Select rod insert only,
b. Half scram only.
c. Select rod insert and half scram.
d. Full scram.
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- QUESTION
- 057 (1.00) !
i
The plant was operating at 100% power with the EPR in control of reactor
- l
pressure when the Bypass Opening Jack (BPOJ) was taken to the LOWER i
position. >
1 I
WHICH ONE of the following describes the expected plant response if the
!
!
BPOJ switch is held in the LOWER position? )
a. The bypass valves will open fully, then the control valves will I
be closed by the EPR. Reactor pressure will stabilize at the l
EPR setpoint. , l
.
, b. The bypass valves will open fully, then the control valves will ;
open until the control valve limit stop is reached. Reactor ;
pressure will decrease and stabilize at a lower value than !
before the transient.
c. The control valves will open until the control valve limit stop i
is reached, then the bypass valves will open fully. Reactor
pressure will decrease until the MSIVs close,
l
d. The control valves will open until the control valve limit stop
is reached, then the bypass valves will open until the reactor l
flow limit is reached. Reactor pressure will decrease until the l
MSIVs close. !
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. SENIOR REACTOR OPERATOR Page 45
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QUESTION: 058 (1.00)
i
The mode switch is in STARTUP/ HOT STANDBY during a reactor startup. '
Containment High Range Radiation Monitor (CHRRM) RIT 1826 has failed j
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upscale,
j
WHICH ONE of the following describes the availability of the primary l
containment atmosphere control functions in this condition?
l
a. The H2/02 analyzer and containment vent and purge are still
i available because both CHRRMs would have to fail to cause a !
l Group VI isolation. l
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b. The H2/02 analyzer and containment vent and purge isolate. Both !
functions can be restored using the RIT-1826 Group 6 Bypass !
Switch. l
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( c. The H2/02 analyzer and containmant vent and purge isolate. Only f
the H2/02 analyzer can be restored using the RIT-1826 Group 6 j
Bypass Switch. ;
.
d. The H2/02 analyzer and containment vent and purge isolate. Only
the containment vent and purge can be restored using the RIT- ,
1826 Group 6 Bypass Switch. l
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, SENIOR REACTOR OPERATOR Page 46 l
QUESTION: 059 (1.00)
The control room operator records the following sump integrator data,4 ;
hours after the previous data was taken: j
Drywell floor drain sump = 545 gallons (since last reading) ,
Drywell equipment drain sump = 5480 gallons (since last reading) !
l
WHICH ONE of the following describes the required technical
specification action? l
, .
l a. A shutdown is not required because all coolant leakage is within !
l specifications. !
I
b. A shutdown is not required because unidentified leakage is -
within specifications. '
c. A shutdown is required because neither identified nor
unidentified coolant leakage is within specifications.
d. A shutdown is required because reactor coolant leakage is not
within specifications.
QUESTION: 060 (1.00)
The plant was operating at 100% power when all service water was lost
due to an unisolable pipe break.
WHICH ONE of the following describes actions that can be taken to
restore closed cooling water systems?
i
a. Restore TBSCCW from Emergency Service Water, then crosstie l
TBSCCW and RBCCW to restore RBCCW.
b. Restore TBCCW from Fire Protection, then crosstie TBCCW and
TBSCCW to restore TBSCCW.
c. Restore RBCCW from Emergency Service Water, then crosstie RBCCW
and TBCCW to restore TBCCW.
d. Restore TBSCCW from Fire Protection, then crosstie TBSCCW and
TBCCW to restore TBCCW. ;
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QUESTION: 061 (1.00) l
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All RBCCW was lost. The reactor was scrammed, but a number of rods I
failed to insert. Boron injection was not initially required. Then RPV i
pressure was reduced to 500 psig for one hour. Current plant conditions :
are as follows.
i
RPV Pressure: 500 psig, steady l
RPV Water Level: + 30", steady j
Reactor Power: 40% on IRM Range 8, steady
{
Drywell Pressure: 7.5 psig, increasing :
Drywell Temperature: 255 F, increasing ;
WHICH ONE of the following describes how RPV pressure should be
.
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controlled in accordance with ONP 524C and the EOPs now that RPV :
i pressure has been at 500 psig for one hour? -
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a. Maximize reactor cooldown rate. Cooldown rate may exceed !
100*F/hr, but reactor temperature should be maintained greater '
than 25 F above drywell temperature.
l b. Maximize reactor cooldown rate, not to exceed 100 F/hr.
l Maintain reactor temperature greater than 25 F above drywell -
temperature.
.
c. Cooldown the reactor at a rate not to exceed 60 F/hr. Maintain
reactor temperature greater than 25 F above drywell temperature.
'
d. Hold reactor pressure at 500 psig. Do not commence cooldown at
l this time
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SENIOR REACTOR OPERATOR Page 48
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QUESTION: 062 (1.00)
l
! The plant was operating at 100% power when the in-service CRD pump
tripped. The other CRD pump is available, but there will be a delay in
starting it while tags are cleared.
WHICH ONE of the following CRP 905 A-2 alarms would NOT be expected as a
result of the loss of CRD flow?
i a. "CRD HIGH TEMP"
b. "CRD CHARGING WATER LOW PRESSURE"
i
c. "CRD ACCUMULATOR TROUBLE" - The light on the local panel for the l
accumulator in alarm goes out when the button is pushed.
d. " ROD OUT BLOCK"
QUESTION: 063 (1.00)
The plant was operating at 95% steady state power when the following
indications were observed:
- Reactor power increased slightly
- Reactor pressure decreased j
- Reactor water level remained relatively steady !
-Indicated steam flow decreased
- Generator output decreased
WHICH ONE of the following events has occurred?
i
a. A recirculation pump increased speed slightly.
b. An SRV has failed open.
c. One set of SJAEs has tripped off and decreased turbine vacuum.
d. RPV pressure control has shifted from the EPR to the MPR.
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{ QUESTION: 064 (1.00)
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l Given the following conditions:
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- - Unit 1 had been operating at 75% power
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l - A loss of Vital AC has just occurred t
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j - A reactor scram has also occurred l
WHICH ONE of the following reactor water level instruments is the ,
j operator directed to monitor prior to securing the reactor feedwater i
j pumps after the scran.? '
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l a. Wide Range GEMAC Floodup Level j
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b. Wide Range Fuel Zone Level l
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c. Narrow Range GEMAC Level :
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f d. ATWS Narrow Range Level !
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QUESTION: 065 (1.00)
i
The plant was operating at 100% power when an electrical fault caused a
loss of 480V bus 12D.
WHICH ONE of the following would be a concern and the corresponding
required action if 12D cannot be restored?
a. The reactor building may isolate due to loss of plant heating.
SBGT should be started if the reactor building isolates to
maintain negative pressure.
b. Airborne releases may occur in the recombiner building and the
Xe/Kr building due to foss of ventilation. The MSIVs should be
closed to limit airborne releases.
c. Overheating of the reactor feed pumps may occur due to a loss of
area coolers. An orderly shutdown and plant cooldown should be I
performed so that the pumps can be secured.
d. Condenser vacuum may be lost due to loss of the condenser vacuum ;
pump. A reactor scram should be initiated if condenser vacuum '
cannot be maintained.
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l QUESTION: 066 (1.00) )
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! The plant is operating at 100% power with the diesel generator tagged
out for maintenance. A reactor scram occurs with a failure to fast '
transfer to the RSST.
WHICH ONE of the following describes the 125 VDC battery chargers, if
any, that are supplying their respective buses immediately after the
gas turbine generator energizes its respective buses. !
a. No battery chargers are supplying the DC buses.
b. "A" or "C" battery charger only.
{
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c. "B" or "D" battery charger only. !
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d. "A" and "B" battery chargers.
,
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QUESTION: 067 (1.00) ;
A complete loss of 24 VDC power occurred with the plant shutdown.
WHICH ONE of the following describes the source range monitoring
equipment that are still energized or available?
a. SRM high voltage power supplies and recorders on CRP 905,
b. SRM electronic circuitry and detector drive power.
c. SRM high voltage power supplies and electronic circuitry,
d. SRM detector drive power and recorders on CRP 905,
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SENIOR REACTOR OPERATOR Page 51
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QUESTION: 068 (1.00)
The plant was in cold shutdown with both recire pumps and both SDC pumps
operating when both recirc pumps tripped. Operators were NOT able to
close the discharge or suction valves for either pump. Reactor water
level is + 55 inches.
WHICH ONE of the following describes the status of SDC flow in this
condition?
a. Shutdown cooling flow will not bypass the core,
b. Shutdown cooling flow will bypass the core through the A recirc
pump only.
c. Shutdown cooling flow will bypass the core throuch the B recirc -
pump only.
d. Shutdown cooling flow will bypass the core through the A and B I
recirc pumps,
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SENIOR REACTOR OPERATOR Page 52
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QUESTION: 069 (1.00)
, The plant was shutdown in preparation for a refueling outage at 1200 i
l hours on 11/1. The reactor cavity was flooded. The following events
l occurred. :
i
1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> on 11/4 - A total loss of decay heat removal capability occurred. The
only source of injection available was one CRD pump.
1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> on 11/6 - A condensate transfer pump becomes available I
i
1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> on 11/7 - Decay heat removal capability can be restored )
WHICH ONE of the following describes the expected effect on reactor
water level? (ONP 531 is provided.)
.
a. Boiloff will not occur and the core will remain covered,
b. Boiloff will occur, but the core will remain covered because
makeup capacity will be sufficient.
c. Boiloff will occur, but even though there is insufficient makeup l
capacity, decay heat removal capability will be restored prior
to core uncovery.
d. Boiloff will occur and the core will be uncovered.
QUESTION: 070 (1.00)
During refueling, WHICH ONE of the following correctly describes how the
fuel pool gates should be positioned for a leak in the fuel pool and the
reason for the action?
a. The fuel pool gates should be closed to ensure that the steam
dryer and steam separator are not uncovered. !
b. The fuel pool gates should be opened to ensure that the fuel
pool cooling system remains operable.
c. The fuel pool gates should be closed to minimize the loss of
water in the reactor cavity.
d. The fuel pool gates should be closed to ensure fuel pool level
remains above the skimmer weirs. l
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QUESTION: 071 (1.00)
During refueling operations, 35 days after plant shutdown, water level
in the reactor cavity began to decrease due to an unisolable leak. A !
spent fuel bundle was being transferred to the spent fuel pool. When
the operator attempted to lower the bundle back into the core, the hoist !
Jammed and the bundle could not be lowered. When the water level
dropped below the top of the bundle, the refuel floor area radiation !
monitor alarmed.
,
l
WHICH ONE of the following is the minimum required emergency action
level for the current conditions? !
I
a. Unusual Event
b. Alert -
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c. Site Area Emergency
i
d. General Emergency '
.
QUESTION: 072 (1.00) I
i
Millstone Units 1 and 2 were both operating at 100% power. Millstone
Unit 3 was shutdown for refueling. Transmission line 310 was out of j
service for corrective maintenance. A storm has just caused the loss of l
transmission lines 348 and 383.
WHICH ONE of the describes the current status of Units 1 and 27
a. Unit 1 and Unit 2 were both load-rejected.
b. Unit 1 was load-rejected. Unit 2 is still on-line.
c. Unit 1 is still on-line. Unit 2 was load rejected.
d. Unit 1 and Unit 2 are both on-line.
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OUESTION: 073 (1.00) ,
A turbine trip occurred while the plant was operating at 100% power.
WHICH ONE of the following describes the response of the Extraction ,
Steam nonreturn valves and the reason for the response? +
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a. The nonreturn valves remain open to provide feedwater preheating j
'
which will ensure cold water is not injected into the RPV. l
l '
b. The nonreturn valves remain open to provide steam to drive the
SJAEs and maintain condenser vacuum.
,
.c. The nonreturn valves close to prevent water from the tube side i
I
of the heaters from backing up into the turbine, : : sntially
damaging turbine blading. -
d. The nonreturn valves close to prevent the feedwater heaters from
acting as steam sources due to flashing, potentially causing
turbine overspeed.
QUESTION: 074 (1.00)
.
The plant was operating at 100% power when condenser vacuum began to
decrease.
WHICH ONE of the following describes actions that would prevent a low
condenser vacuum scram?
a. If a loss of instrument air to the condenser air removal system
occurred, then condenser vacuum could be maintained by
transferring control of SJAE inlet pressure to manual.
b. If a loss of instrument air to the AOG recombiner system
occurred, then condenser vacuum could be maintained by shifting
offgas flow to the 30 minute delay line, ,
i
c. If a loss of Vital AC power occurred, then condenser vacuum
could be maintained with the mechanical vacuum pump.
d. If a loss of Instrument AC power occurred, then condenser vacuum i
could be maintained by maintaining turbine steam seal pressure
with the steam seal regulator bypass and unloading valves.
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SENIOR REACTOR OPERATOR Pags 55
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l QUESTION: 075 (1.00)
WHICH ONE of the following correctly describes the bases for the reactor
scrams that are designed to occur on a loss of condenser vacuum?
a. The turbine stop valve closure scram prevents exceeding the
reactor pressure safety limit. .
The low condenser vacuum scram is a backup that makes the
resulting transient less severe.
b. The turbine stop valve closure scram prevents exceeding the MCPR
safety limit.
The low condenser vacuum scram is a backup that makes the
resulting transient less severe.
c. The low condenser vacuum scram prevents exceeding the reactor
pressure safety limit.
The turbine stop valve closure scram is a backup that makes the
resulting transient less severe.
1
d. The low condenser vacuum scram prevents exceeding the MCPR
safety limit. i
The turbine stop valve closure scram is a backup that makes the ;
'
resulting transient less severe.
l
QUESTION: 076 (1.00)
While resetting a reactor scram, the operator neglected to place the
scram discharge volume isolation test switch to "lSOLATE." He performed
the remainder of the procedure correctly.
WHICH ONE of the following describes the response of the SDV vent and I
drain valves?
a. All SDV vent and drain valves open as soon as the scram reset l
switch is placed in the "GR 2 & 3" position.
b. The vent and drain valves for one of the SDVs open as soon as
the scram reset switch is placed in the "GR 2 & 3" position.
c. One valve in each SDV vent and drain line open when the scram
reset switch is placed in the "GR 2 & 3" position.
,
d. All SDV vent and drain valves open when the scram reset switch
'
is placed in the "GR 1 & 4" position.
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SENIOR REACTOR OPERATOR Paga 56 ;
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QUESTION: 077 (1.00)
A fire in the cable vault has caused all of the SRVs to open. The red :
indicator lights on CRP 903 are lit for all of the SRVs. Reactor
pressure is 930 psig. The operator leaving the Control Room places the i
Appendix R CRIS switches in the ISOLATE position.
WHICH ONE of the following describes the immeuiate effect on the SRVs?
a. All SRVs will remain open. i
b. All SRVs will close. All SRVs will reopen on high pressure !
only.
c. All SRVs will close. All SRVs will reopen on high pressure, ,
APR valves will reopen on an APR initiation.
!
d. All SRVs will close. No SRVs will reopen on high pressure. APR ,
valves will reopen on an APR initiation. i
QUESTION: 078 (1.00)
!
The control room has been evacuated due to a fire in the cable vault, j
lt is necessary to shut down the diesel generator because no service
water pumps are available.
WHICH ONE of the following actions should be completed prior to shutting
down the diesel generator?
a. Ensure CRD pump A is running in the self-coo!ing mode.
Establish backfeed from Unit 2 if necessary.
1
b. Initiate drywell cooling via LPCI keep-full or via Fire
Protection fire hose. l
1
c. Verify 1-IC-1, steam isolation valve, and 1-IC-4, condensate l
return isolation valve, open.
d. Ensure manual control of 1 IC-10, mak. op isolation valve, has
been established to provide IC makeup. ;
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SENIOR REACTOR OPERATOR Paga 57
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l
QUESTION: 079 (1.00) I
WHICH ONE of the following conditions would be a safety limit violation?
a. While operating at 30% power, the turbine bypass valves fail '
open. Reactor pressure drops to 800 psig before the MSIVs close l
l and the reactor scrams. ,
I I
b. While operating at 75% power, a common mode failure in the !
r recirculation flow controllers causes the speed of both recirc [
l pumps to increase to the high speed stop. The Minimum Critical
l Power Ratio is 1.08. ;
1
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c. While operating at 90% power, a turbine trip occurs. The l
reactor fails to scram and the ATWS system logic trips the i
recirculation pumps and initiates ARI. .
!
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d. While operating at 100% power, an inadvertent MSIV closure -
causes reactor pressure to increase. The reactor scrams and
reactor pressure increases until 5 SRVs lift. ;
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QUESTION: 080 (1.00) ,
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l The plant was operating at 100% power with RFPs A and B running. RFP C l
}'
was in standby. Feedwater string A was selected for FWCl. The following ;
events occurred:
i
{ - Reactor water level increased rapidly due to a level control system !
]- malfunction. The RFPs and turbine tripped and the reactor
j scrammed.
>
l - Reactor water level began to decrease and RFP B was manually
started for RPV level control,
h *
4 - A high drywell pressure signal was received and RFP B tripped on
i overcurrent.
-
!
- WHICH ONE of the following describes the automatic response of RFPs A
i and C to the high drywell pressure and RFP B trip? Assume RPV level is i
j greater than -48 inches.
1
l a. Neither RFP A nor C starts. ,
4
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- b. Only RFP A starts.
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c. Only RFP C starts. ;
1 d. Both RFP A and C start.
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QUESTION: 081 (1.00)
l
The plant was operating at 100% power with RFPs A and B running.
'
Feedwater string B was selected for FWCl. A LNP occurred and power was
'
restored to all the 4160V busses with the exception of 14A which has an l
electrical fault. l
1
WHICH ONE of the following describes the expected response of the (
- condensate system when RPV water level drops below the low low level ;
setpoint and RPV pressure drops below 600 psig?
1
a. FWCl does not initiate. A condensate pump and condensate
, booster pump must be manually started to restore RPV level. The
ECT pump must be manually started to restore hotwell level. l
b. Condensate pump B and condensate booster pump B automatically
start due to the FWCl initiation and inject to restore RPV
level. The ECT pump automatically starts due to the FWCl
initiation.
c. Condensate pump B and condensate booster pump B automatically
start due to the FWCl initiation and inject to restore RPV
level. The ECT pump automatically starts when hotwell level
drops below 12 inches.
d. Condensate pump B and condensate booster pump B automatically
start due to the FWCl initiation and inject to restore RPV )
'
level. The ECT pump must be manually started to restore hotwell
level.
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a
, SENIOR REACTOR OPERATOR Pagn 60
- QUESTION
- 082 (1.00)
The plant was operating at 100% power when a break occurred on the
instrument air header. The reactor scrammed on low scram pilot header
pressure and EOP 570 was entered.
, WHICH ONE of the following describes methods that can be used for RPV
level and pressure control?
a. Feedwater for RPV level control. SRVs for RPV pressure control.
b. Feedwater for RPV level control. RWCU Reject Mode for RPV
pressure control.
c. Maximize CRD for RPV level control. Main steam line drains for
RPV pressure control.
d. Maximize CRD for RPV level control. Isolation Condenser for RPV
pressure control.
.
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, , SENIOR REACTOR OPERATOR Page 61
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QUESTION: 083 (1.00)
- The plant was operating at 100% power when the plant scrammed due to a !
loss of feedwater. A loss of normal power occurred due to a failure of l
~
l .the RSST. All rods inserted and the diesel generator (DG) and gas
l turbine (GT) started and are loaded as follows.
. ;
DG CURRENT LOADING GT CURRENT LOADING !
l
. 1800 KW 9200 KW :
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s LPCI pump C is the only ECCS pump running. RPV levelis decreasing and !
l
torus temperature is increasing.
I
i
l WHICH ONE of the following describes sources of RPV injection and torus ;
j cooling that can be lined up under these conditions? Figures 7.2 and I
j 7.3 from ONP-503B are attached.
1 <
1 a. Core Spray pump A for RPV injection. l
LPCI B subsystem for torus cooling,
l
i
j b. LPCI pump A for RPV injection.
"
LPCI B subsystem for torus cooling,
i
j c. Core Spray pump B for RPV injection.
j LPCI A subsystem for torus cooling,
j
! d. LPCI pump B for RPV injection.
i LPCI A subsystem for torus cooling.
,
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SENIOR REACTOR OPERATOR Pagn 62
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QUESTION: 084 (1.00)
A LOCA occurred 5 minutes ago that caused RPV water level and pressure
to decrease rapidly. The reactor failed to automatically scram. All
control rods inserted on the manual scram. No injection subsystems are
available. Efforts are in progress to lineup alternate injection
subsystems. The following conditions exist:
- RPV pressure is 50 psig and steady
- RPV level is -175" and decreasing slowly
- Drywell pressure is 1.4 psig and steady. '
WHICH ONE of the followin, uie minimum required emergency action
le. vel for the current conditior'
a. Alert - Charlie One -
b. Site Area Emergency - Charlie Two
c. General Emergency - Bravo
d. General Emergency - Alpha
QUESTION: 085 (1.00)
Following a major loss of coolant accident and LNP, RPV level cannot be
maintained above TAF. The operating crew determines that an alternate
injection system needs to be aligned in order to recover RPV level.
Reports from the field indicate that the condensate transfer pump house
is inaccessible. RPV pressure is 80 psig.
WHICH ONE of the following lineups should be used to make up to the
vessel given the current plant conditions?
a. Fire Water via Condensate Transfer and Core Spray
b. Fire Water via SLC injection drain line
c. Emergency Service Water
d. ECCS Keep-Full
,
, SENIOR REACTOR OPERATOR Page 63
QUESTION: 086 (1.00)
During WHICH ONE of the following conditions would it be allowable to
use Core Spray with suction from the CST in accordance with the EOPs?
a. During performance of Alternate Level Control with two
condensate pumps lined up and running.
b. During performance of ATWS - RPV Flooding with LPCI pump A lined
up and running, 4 SRVs open and RPV pressure at 225 psig.
c. During performance of Primary Containment Flooding with one
condensate pump available.
d. During performance of ATWS - Primary Containment Flooding with
LPCI pump B available. -
QUESTION: 087 (1.00)
A reactor scram has occurred and approximately 1/2 of the control rods
failed to insert. The following indications are available:
- SCRAM DISCH VOLUME HI LEVEL SCRAM annunciators lit.
- Approximately 1/2 of the individual white lights on CRP 905 are
illuminated.
- 4 of the CRD Scram Solenoid Group lights on CRP 905 are
illuminated.
WHICH ONE of the following describes the most likely cause of the
f ailure of the rods to insert and the first action that should be taken
to attempt to insert the remaining rods?
a. The problem is most likely electrical. The RPS circuit breakers
should be opened,
b. The problem is most likely electrical. The ATWS system should
be deenergized and RPS should be bypassed.
c. The problem is most likely mechanical. The RPS circuit breakers
should be opened,
d. The problem is most likely mechanical. The ATWS system should
be deenergized and RPS should be bypassed.
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, SENIOR REACTOR OPERATOR Page 64
QUESTION: 088 (1.00)
The plant was operating at 100% power when an ATWS occurred. No rod
motion occurred. 24% of SLC tank level was injected and the boron has
been mixed. RPV pressure is being maintained at 1000 psig and RPV level
is in the normal band.
,
WHICH ONE of the following describes the expected current status of the
l reactor and the expected status of the reactor 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> later? Assume
( RPV pressure and level are held constant, no additional boron is
injected and no control rods are inserted.
a. The reactor is currently subcritical and will be more suberitical 8
l
, hours later.
b. The reactor is currently subcritical and will be less subcritical 8 -
hours later,
c. The reactor is currently critical and will be more critical 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
later,
d. The reactor is currently critical, but will be suberitical 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
later.
QUESTION: 089 (1.00)
EOP 575, Sheet 1," Failure-to-Scram," is being implemented. Six rods are
stuck out. Reactor power is 50% on IRM range 6 and decreasing slowly.
SBLC is injecting. Initial tank level was 64%. Current tank level is
32%.
WHICH ONE of the following describes the appropriate actions that should
be taken if RPV water level cannot be determined?
a. Exit section FSL. Go to EOP 575, Sheet 4, "FS - RPV Flooding."
Cooldown may commence in accordance with section FSP.
b. Exit sections FSL and FSP. Go to EOP 575, Sheet 4, "FS - RPV
Flooding."
c. Exit EOP 575. Go to EOP 570, Sheet 1. Exit section RL. Go to
EOP 570, Sheet 4, "RPV Flooding." Cooldown may commence in
accordance with section RP.
d. Exit EOP 575. Go to EOP 570, Sheet 1. Exit sections RL and RP.
Go to EOP 570, Sheet 4, "RPV Flooding,"
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QUESTION: 090 (1.00)
!.
l Following a loss of coolant accident the reactor is shutdown and the
f
recirculation pumps are tripped. Reactor pressure is 75 psig. Drywell
- temperature indications are
I .
Tl 1602-6A: 315 F TR 1602-6, Point 1: 340 F
-
Tl 1602-6B: 305 F TR 1602-6, Point 9: 310 F
'
WHICH ONE of the following is an accurate RPV level indication?
a. -50" on WR LI 263-115 on CRP 903
l ,
l b. + 2" on Floodup LI 263-101 on CRP 904
c. -45" on NR Yarway B on CRP 905
(
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d. -33" on NR ATWS B on CRP 980
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QUESTION: 091 (1.00)
i
During torus cooling, LPCI flow through the heat exchanger is limited to I
5000 gpm to prevent damage to the heat exchanger.
WHICH ONE of the following is the reason that only one LPCI pump may be
operating with the heat exchanger bypass valve closed (rather than
running 2 pumps and throttling LPCI flow)?
a. One pump operation reduces the damage to the test return valve l
due to flow induced vibration. l
b. Higher ESW flow can be established with only one LPCI pump
running, which results in better heat exchanger performance.
c. Operation of only one pump increases system reliability,
l allowing one pump to remain in standby readiness for RPV
injection.
l d. Operation of only one pump allows the pump to operate with
additionsi margin to pump cavitation.
l
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SENIOR REACTOR OPERATOR Page 66
QUESTION: 092 (1.00) .
An ATWS occurred and power was lost to bus 14E. All rods have been ,
inserted and power has just been restored to 14E. Current conditions j
are as follows:
RPV Level: -110" and stable
LPCI A is injecting 5000 gpm with both pumps running
Torus Water Level: 12'
Torus Temperature: 174 F
Torus Bottom Pressure: 3 psig
WHICH ONE of the following is the highest containment cooling flow that
can be established now that Core Spray and LPCI B subsystems are
available and all rods are inserted?
a. 3000 gpm.
,
i b. 5000 gpm.
c. 6000 gpm.
d.10000 gpm.
4
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SENIOR REACTOR OPERATOR Pagn 67
QUESTION: 093 (1.00)
The plant was operating at 100% power when the reactor scrammed on low
water level. Both SBGT trains automatically started. Current plant
conditions are as follows:
- RPV water level is -5" and decreasing slowly
- Drywell pressure is 1.2 psig and increasing slowly
- Drywell temperature is 160 F and increasing slowly
WHICH ONE of the following describes the actions that should be taken to
vent the primary containment in accordance with OP 3117
a. The containment cannot be vented because it is not permissible
to bypass the Group 11 isolation signal under these conditions.
.
b. The containment can be vented. Both SBGT trains can be left in
service.
t
c. The containment can be vented. One SBGT train must be locked
out.
d. The containment can be vented. Both SBGT trains must be locked
'
out.
QUESTION: 094 (1.00)
When executing EOP 580, " Primary Containment Control," emergency I
depressurization is required if torus water level cannot be maintained I
above 10.5 feet.
WHICH ONE of the following becomes uncovered at 10.5 feet?
a. Drywell downcomers,
l
b. SRV tailpipes.
c. ECCS suction strainers.
d. Reactor building to torus vacuum breakers.
_ , - . _ _ _ _ . - _ _ _ _ _ . _ . . - _ . . - _ _ . - _ _ . _ . _ . . . . _ _ .
. SENIOR REACTOR OPERATOR Paga 68
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QUESTION: 095 (1.00)
l
WHICH ONE of the following correctly describes the effects of spraying
t the containment with torus water level above 24 feet?
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t
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a. The drywell to torus differential pressure limit could be
j exceeded. Torus to reactor building differentian pressure is
j not a concern.
b. The torus to reactor building differential pressure limit could
i be exceeded. Drywell to torus differential pressure is not a
concern.
- ;
4
c. Both drywell to torus and torus to reactor building differential
j pressure limits could be exceeded. i
.
!
l d. Differential pressure limits are not a concern, but there is no !
l benefit to spraying the torus with the torus spray nozzles >
. covered.
i
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i
i QUESTION: 096 (1.00)
1 ,
,
A LOCA has occurred. Plant conditions are as follows:
i
'
l - Drywell pressure is 25 psig and increasing
!
- Drywell temperature is 350 F and increasing 6
- Torus water level is 16 feet and steady
,l - RPV water level is -60" and steady
4 - RPV pressure is 700 psig and steady
I
'
WHICH ONE of the following describes the appropriate actions in
accordance with EOP 580, " Primary Containment Control?"
a. Initiate both drywell and torus sprays.
4
b. Initiate drywell sprays only.
!
c. Initiate torus sprays only.
l d. Do not initiate containment sprays.
,
.
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SENIOR REACTOR OPERATOR Pege 69
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QUESTION: 097 (1.00)
{
A LOCA has occurred and containment sprays are unavailable. The !
, following conditions exist in primary containment: !
l !
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- Drywell pressure is 35 psig and increasing l
r
- Torus bottom pressure is 40 psig and increasing i
l - Torus water level is 16 feet and steady l
l - Drywell hydrogen concentration is 8%
- Drywell oxygen concentration is 4%
'
- Torus hydrogen concentration is 5%
-Torus oxygen concentration is 6%
i
WHICH ONE of the following describes the appropriate actions in
accordance with EOP 580, " Primary Containment Control?"
'
a. Vent the torus via the hardened vent and simultaneously purge
the drywell until hydrogen and oxygen concentrations are within :
limits and torus bottom pressure drops below 35 psig. !
b. Vent the torus via the hardened vent until torus bottom pressure
drops below 35 psig, then secure the vent. Purge the drywell !
until hydrogen and oxygen concentrations are within limits. I
c. Vent the torus via the hardened vent until torus bottom pressure j
drops below 35 psig, then secure the vent. Drywell purge is not ;
required. ]
I
d. Do not vent the torus until torus bottom pressure exceeds the !
Pressure Suppression Pressure (PSP) limit or approaches 71 psig. I
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_ _ _ . . _ _ _ . _ _ _ _ _ _ . _ _ _ . _ .
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, SENIOR REACTOR OPERATOR Pags 70 l
QUESTION: 098 (1.00)
!
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The plant was operating at 100% power, when the reactor building PEO l
reported that water was spraying from the running RWCU pump. When
attempting to isolate the system, system inlet valve 1-CU-2 cottd not be
)
1
closed. All other RWCU isolation valves were verified closed. Current
conditions are as follows:
Temperature Switch:
1290-26A: 160 F 1290-26E: 190 *F i
1290-26B: 170 F 1290-26F: 185 F l
1290-26C: 155 *F 1290-26G: 200 *F
1290-26D: 205 *F 1290-26H: 180 F j
Radiation Monitor:
ARM 7: 100 mR/hr
ARM 8: 100 mR/hr
ARM 9: 1000 mR/hr
WHICH ONE of the following describes the required EOP actions for the
current conditions?
a. Power operations may continue while attempting to shut 1-CU-2.
b. Initiate a shutdown.
c. Scram the reactor.
1
d. Scram and emergency depressurize the reactor.
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i SENIOR REACTOR OPERATOR Page 71
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QUESTION: 099 (1.00) l
!
A steam leak area high temperature alarm was received on CRP 903-A2. An :
operator has verified that the alarm is valid. -
!
WHICH ONE of the following temperature conditions would require
emergency depressurization?
a. Temperature Switch: 260-18A: 200 *F l
260-18B: 210 *F
260-18C: 205 *F ;
260-18D: 145 *F ;
!
b. Temperature Switch: 1040-8A: 230 F l
1040-8B: 200 *F
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1040-8C: 270 *F . !
1040-8D: 250 *F !
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c. Temperature Switch: 1040-8E: 160 *F
1040-8F: 195 F !
1040-8G: 185 *F l
1040-8H: 200 F ;
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d. Temperature Switch: 1340-7A: 160 *F ;
1340-7B: 120 *F
1340-7C: 200 *F
1340-7D: 200 F
QUESTION: 100 (1.00)
A fuel cladding failure has occurred. The following alarms are in on
CRP 903:
"
- KAMAN STACK GAS HI RAD" 903 A4 (1-3)
-
" STACK GAS MONITOR IN DISABLE POSITION" 903 A3 (5-7)
WHICH ONE of the following describes the stack gas monitor status?
a. Both the Kaman and the GE stack gas monitors are in service.
b. Only the Kaman stack gas monitor is in service.
c. Only the GE stack gas monitor is in service.
d. Neither the Kaman or the GE stack monitor is in service.
(* * * * * * * * * * END OF EXAMINATION * * * * * * * * * *)
.
, SENIOR REACTOR OPERATOR Pega 1 l
~
ANSWER KEY i
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MULTIPLE CHOICE O23 c j
001 ,# CL 024 a
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002 a 025 d ,
003 c 026 d
004 a 027 c
005 c 028 a
006 b 029 b
007 b 030 c
!
008 b 031 c
009 b 032 c
010 d 033 a
011 c 034 d
012 c 035 b
013 c 036 a l
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014 g a. 037 d
015 d 038 c
016 h 04 039 c j
017 a 040 a
018 c 041 b
019 b 042 b
020 c 043 b
021 b 044 b
022 a 045 c
, i
SENIOR REACTOR OPERATOR P gs 2 j
1
ANSWER KEY !
046 d 069 b f
047 b 070 b
048 g d., 071 b
049 b 072 b I
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050 b 073 d
051 b 074 b
\
!
052 d 075 b
05'3 b 076 d
!
054 af M 077 b
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055 d 078 c
056 c 079 c
057 d 080 d
058 # b 081 a
059 d 082 a
060 c 083 d
061 d 084 d ,
)
062 c 085 d
063 b 086 a
064 d 087 a
065 a 088 a
066 a 089 b
067 d 090 a
068 c 091 b
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.c _ _ _
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. SENIOR REACTOR OPERATOR Paga 3 ]
ANSWER KEY
4
- 092 c
093 e a< A I
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094 a )
J
095 c l
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'
096 c
,
097 b
, 098 a
099 d 1
a
100 a '
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( * * * * * * * * * * END OF EXAMIN ATION * * * * * * * * * * )
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ATTACHMENT 2
FACILITY COMMENTS ON WRITTEN EXAMINATIONS
, _ _ . _ _ _ _. .__ _ _ _ ._ . ______ _ _ . _ . _ _ _ _ _ . - _ _ _
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- N:rthe:st Mill tone orrice. . nope rerry na., reierrora, cr l
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'
j$ Utilities System r.o. no, es
!
Waterrord. CT 06385-0128
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(203) 447 1791
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December 13,1996 I
4
Docket No. 50-245
i B16087
1
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? Mr. H. J. Miller
! U.S. Nuclear Regulatory Commission
Region 1
475 Allendale Road
King of Prussia, PA 19406-1415
'
Millstone Nuclear Power Station, Unit No.1
i NRC Exam Comments
The purpose of this letter is to forward comments regarding the written NRC license
examination administered at Millstone Unit No.1 on December 2,1996. .
.'
'
These comments result from a review of the examination conducted by members of the
Millstone Unit No.1 operating and training staff. References are provided to substantiate
these comments. l
.
4
Please contact Mr. Kevin Murphy, Supervisor, Operator Training, Millstone Unit No.1 with
! any questions concerning this submittal.
Very truly yours,
NORTHEAST NUCLEAR ENERGY COMPANY
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w N, 'O ,~
V
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' P. D. Hinnenkamp
Director, Unit Operations
!
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cc: S. Dembek, NRC Project Manager, Millstone Unit No.1
T. A. Eastick, Senior Resident inspector, Millstone Unit No.1
,
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0S3422 9 REV.195
.. , _ _
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Docket No. 50-245
B16004
i
Attachment 1 ,
1
Millstone Nuclear Power Station, Unit No.1
NRC Exam Comments .
1
December 1996
~ ~ ~ ' ~ ' ~' ~~ ^ ^
. f U.S. Nuct::ar Regulatory Commission'
, B16087\ Attachment 1\ Paga 1
.
NRC EXAM COMMENTS
i
The following changes to the answer key are recommended based on changes performed
during the validation: ,
,
Question 58:
Incorrect answer in key. Answer was not changed during validation. ;
Recommend changing key to Answer"B".
l
The following changes to the answer key are recommended based on additional technical
justification:
Question 14:
!
WC-1, pg. 20, Rev. 2 provides the guidance for determining the safety system
out-of-service time. The out-of-service time is calculated as the difference between the l
time that the LCO was entered and the time the LCO was exited. Answer "A" reflects safety
system out-of-service time.
Recommend accepting answer "A" as the only correct answer.
Question 48:
OP 305 Section 3.10 states "Without forced circulation (reactor recirc pump or SDC pump
running), shutdown cooling inlet temperature (reactor recirc loop 'A') is not indicative of
reactor temperatures." The question stem includes both reactor recirc pump "B" and both
SDC pumps. Therefore, shutdown cooling inlet temperature is indicative of reactor core
temperatures. Answer "A" reflects this condition.
Recommend accepting answer "A".
l
The following changes to the answer key are recommended based on a difference in the
interpretation of the question by the students:
Question 16:
During normal plant shutdowns, the plant enters COLD SHUTDOWN with the MSIVs open,
vented to the main condenser. Therefore it is very reasonable to believe that COLD
SHUTDOWN was achieved in time on 12/3/96 and no UNUSUAL EVENT report was
required. This would make answer "B" correct.
^
'
U.S. Nucl:iar Hegulatory Commission
l, ,. B16087\Attachm:nt 1\ Paga 2
.
If the candidate assumes, as the question had intended, that the vessel was isolated and
COLD SHUTDOWN was not achieved until the head vents were opened, an Unusual
Event report would be required and choice "D" is correct.
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Recommend accepting either "B" or D".
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Question 54:
l The question stem states " Assume only operator actions were the restoration of power to
the AC and DC motors." The stem does not include the status of field flashing. After
restoring power to the motors, OP 343, Section 5.1.7, requires the operator to " Momentarily
PRESS pushbutton " GEN FIELD FLASH" and OBSERVE voltage buildup to 235 - 245
volts."
If the candidate assumed the " GEN FIELD FLASH" pushbutton still needs to be depressed
in order to flash the generator field then answer "D" would be correct. If the student
assumed the " GEN FIELD FLASH" pushbutton has been depressed, then "C" would be
correct.
Recommend accepting either Answer "C" or "D"
Question 85:
The stem of the question states the " Reports from the field indicate that the condensate
transfer pump house is inacessible." Without giving any additional information on why
exactly the condensate transfer pump house is not accessible, it may be assumed that the
condensate transfer pumps are not available for ECCS keep full. Since no other injection
source is mentioned in the question stem, fire water via SBLC injection drain line may be
used.
Recommend accepting either answer "B" or "D".
Question 93:
The question attempts to determine if the primary containment can be vented in
accordance with OP 311. OP 311, Section 4.10.3a, specifies the conditions necessary to
vent in the given conditions. "Drywell pressure is expected to remain less than 2 psig."
The stem of the question indicates drywell pressure is 1.2 psig and drywell temp is 160 F
both increasing slowly. Therefore, if the candidate assumes drywell pressure will not
remain below 2 psig, answer "A" is correct. If the candidate assumes drywell pressure will
remain below 2 psig, answer "C" is correct. The stem of the question does not provide the
level of detail to make this specifically clear.
l
Recommend accepting either answer "A" or "C".
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~ o.i nudfar Meguisioryi,oisimission
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,, B16087\Attachm:nt 1\ Paga 3
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, Question 98:
The temperature and radiation readings provided do not include any trend information. If
the isolation was successful, the temperature and radiation readings would begin to trend
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downward. If the candidate assumed the readings were constant and the leak was not fully :
1 isolated, EOP 585, step SC-15, would direct the initiation of a reactor scram. This is a
reasonable assumption based on the elevated temperatures and radiation without a trend if
it is assumed that an isolation valve is ieaking by. This action is reflected in answer "C". ,
,
Additionally, a reactor scram in this condition would be a conservative and prudent action. !
If the candidate assumes the leak is isolated, answer "A" remains correct.
]
Recommend accepting either answer "A" or "C".
Question 100: l
The question stem does not provide a specific sequence of events for the fuel cladding i
I
failure and the associated annunciators. This could result in some confusion on the status ' i
! of the GE stack monitor. Additionally, with the STACK GAS MONITOR IN DISABLE
j POSITION and a fuel cladding failure condition, the GE monitars would saturate, and
therefore, would not provide any valid information and could be considered out of service.
j Annunciator Alarm Response procedure 903A3 (5-7) provides information that with the
monitor in DISABLE, the monitor will not purge on a high radiation condition. If the
i candidate assumed a saturated GE monitor was not in service, answer "B" would be
correct. If the candidate assumed the GE monitor was stillin service, however, the
information was incorrect, and answer "A" would be correct.
! ,
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Recommend accepting either answer "A" or "B". !
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i The following changes to the answer key are recommended based on a lack of operational
- orientation and difficulty in interpretation
l
1 Question 6- l
This question does not represent a fundamental level of memory that is appropriate for an
SRO applicant. In Service Testing cold shutdown surveillance testing requirements would
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- not be expected to be recalled from memory.
Recommend deleting the question.
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ATTACHMENT 3 ;
<
NRC RESOLUTION OF FACILITY COMMENTS ON WRITTEN EXAMINATION !
,
SRO 6: Comment not accepted. The question was designed to test general i
- knowledge of inservice surveillance testing administrative requirements and i
i purpose. Detailed knowledge of testing requirements was not needed to
answer the question. The facility did not indicate that this question was not
appropriate during the pre-examination review. No change was made to the
i answer key. i
s SRO 14: Comment accepted. The correct answer was changed to answer "a" based i
on the revised procedure reference provided by the facility. '
.
'
SRO 16: Comment accepted. Answer "b" was accepted as an additional correct ,
y answer. !
i c
SRO 48: Comment accepted. The correct answer was changed to answer "a". .
i ;
i SRO 54: Comment noted. The question stem states that " assume only operator i
,
actions were the restoration oV power to the ac and dc" motors. Therefore, !
-
the generator field would not have been flashed, if the generator field was ,
'
not flashed, answer "c" would be incorrect. The correct answer was r
changed to answer "d".
SRO 58: Comment accepted. The correct answer was changed to answer "b". This ,
error was recognized by the NRC staff during examination administration. ,
SRO 85: Comment not accepted. The condensate transfer pumps can be operated
from the control room; therefore, access to the pump house is not required
for ECCS keep full. There was no reason to assume that the pumps were
not available because the pump house was not accessible. This question ,
was discussed during the pre-examination review with the facility and l
'
revised to address this concern. No change was made to the answer key.
i
SRO 93: Comment accepted. Answer "a" was accepted as an additional correct
answer.
SRO 98: Comment not accepted. The question was intended to test the applicant's
ability to recognize that the system would be isolated even if 1-CU-2 could
not be closed. The stem specifically states that the other isolation valves
were verified closed; therefore, it was not reasonable to assume that the leak
was not fully isolated. The question asks for the " required" actions. A
scram is not required if the system is isolated; therefore, answer "c" is not
correct. No change was made to the answer key.
i
SRO 100: Comment not accepted. With the monitor in DISABLE, the GE monitor 1
would not isolate on a high radiation condition; therefore, it would still be "in
service." The question does not ask if the monitor will provide valid 4
information. This question was discussed during the pre-examination review I
'
with the facility and revised to address this concern. No change was made
to the answer key.
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ATTACHMENT 4 ;
SIMULATION FACILITY REPORT ;
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Facility License: DPR-21 B
l Facility Docket No: 50-245 j
l
l Operating Test Administration: December 2-6,1996 ,
This form is to be used only to report observations. These observations do not constitute j
audit or inspection findings and are not, without further verification and review, indicative '
of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC
certification or approval of the simulation facility other than to provide information that ,
may be used in future evaluations. No licensee action is required in response to these .
l observations. i
ITEM DESCRIPTION
None
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