IR 05000309/1993026
| ML20059A134 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 12/14/1993 |
| From: | Ihnen K, Meyer G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20059A128 | List: |
| References | |
| 50-309-93-26OL, NUDOCS 9312300026 | |
| Download: ML20059A134 (112) | |
Text
{{#Wiki_filter:-__ '. d U.S. NUCLEAR REGULATORY COMMISSION
REGION I
REPORT NO: 50-309/93-26 (OL) DOCKET NO: 50-309 LICENSEE: Maine Yankee Atomic Power Company 83 Edison Drive Augusta, Maine 04336 FACILITY: Maine Yankee DATES: November 15,1993 (requalification exam) November 16-18,1993 (initial exam) NRC EXAMINERS: K. Ihnen, Operations Engineer M. Jones, EG&G , I NE ' CHIEF EXAMINER: .- Kerry D. Ihnen, Operations Erg,meer Date PWR Section, Operations Branch Division of Reactor Safety APPROVED BY: e [4, de /A//9/98 .l , ' lenn W.' Meyer, Chief Dafe '/ WR Section, Operations Branch Division of Reactor Safety l l 9312300026 931215
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SUMMARY: Written examinations and operating tests were administered to three senior reactor operator (SRO) upgrades and one SRO instant. All individuals passed and were issued licenses. In addition, a requalification retake exam was administered to one crew.
The crew passed the exam. The candidates were well prepared and did very well on all portions of the examination. The training department provided good support for both the written examination and for the operating exams. When exiting the radiation area a radiation monitor was alarmed and radiation protection personnel did not respond to the alarm. This was noted as a weakness.
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. ! . DETAILS a 1.0 REQUALIFICATION EXAMINATION 1.1 Examination Results: l ' _ NRC RO SRO TOTAL CREW . Results Pass / Fail Pass / Fail Pass / Fail Pass / Fail Written N/A N/A N/A .N/A Simulator 2/0 2/0 4/0 1/0 Walk-through N/A N/A N/A N/A , Overall 2/0 2/0 4/0 1/0 Maine Yankee RO SRO TOTAL CREW Results Pass / Fail Pass / Fail Pass / Fail Pass / Fail Written N/A N/A N/A N/A- - Simulator 2/0 2/0 4/0 1/0
__ Walk-through N/A N/A N/A N/A
Overall 2/0 2/0 4/0 1/0 1.2 Programmatie Strengths . The simulator scenarios used for the examination were challenging and provided the
evaluators with ample opportunities for assessing the operators' skill.
l , The facility's management and evaluation crew did a good job of responding to some
unexpected problems that occurred during one of the simulator scenarios. In that scenario, a hardware related problem occurred, which caused several incorrect digital indications on the electrical control board. Maine Yankee and the NRC agreed that the indications were not adversely affecting the crew, and the scenario was completed.
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2.0 INITIAL EXAMINATION 2.1 Examination Results: NRC SRO Total Results Pass / Fail Pass / Fail Written 4/0 4/0
Simulator 4/0 4/0 I Walk-through 4/0 4/0 l Overall 4/0 4/0 2.2 Preexamination Activities The reference material requested for this examination was received in a timely manner by both the NRC and the NRC contractor. Prior to the administration of the written examination, Maine Yankee personnel reviewed the written examination in the Regional' Office on November 4,1993. The three reviewers signed security agreements prior to commencing the review.
During the week of the examination while the written examination was in progress, the hRC validated the simulator scenarios planned for the week. Verification was performed with the assistance of a simulator operator and training staff personnel who were also under security agreement.
2.3 Programmatic Strengths and Weaknesses: A.
Strengths Detailed, thorough preexam review conducted by Maine Yankee personnel improved
the examination and led to very few post-exam comments.
All four candidates appeared well prepared to take their initial examinations. Grades
on the written exam ranged from 87.9 to 91.9 percent. Communication skills exhibited during the scenarios were generally above average. Few weaknesses wem identified with the candidates during their plant walk-through.
The ease of entry into the security area and the radiation area were noted by the
examiners. The efforts by the security force and the health physics staffin quickly.
processing visitors helped to prevent un-due stress on the candidates.
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Weaknesses Due to problems encountered when frisking out of the radiation area, one of the i
radiation monitors alarmed. The radiation monitor continued to alarm for its full time ! cycle, and no members of the radiation protection staff responded to the alarm.
There was no one at the control point to inform of the problems encountered, and plant management was informed of the situation.
There were isolated instances where procedures were confusing and could be
enhanced from a human factors point of view. These areas for improvement were discussed with the Maine Yankee training staff. None of the areas in question caused , the candidates to make procedural errors.
The candidates displayed weaknesses with the following questions in the written
examination. Note: An identified weakness is when three or more candidates missed the same examination question. This information is being provided to assist in upgrading initial and requalification training programs. No response to these items is needed.
Ouestion # Tonic
The maximum allowed whole body dose in an emergency to protect . vital equipment t
Reactor trip circuit breaker combinations required for a reactor trip .
Process by which a nontripable control rod is prevented from inserting on a reactor trip
Condition that would cause the de-borating ion-exchanger to release boron back into the mactor coolant system (RCS)
Technical specifications basis for requiring one bank of proportional heaters to be operable whenever RCS average temperature is greater than 500 degrees F 2.4 Post-exam Comments and NRC Response Subsequent to the exit meeting on November 19,1993, Maine Yankee commented on additional questions in the written examination. Listed below is the resolution of these facility comments.
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Ouestion # NRC Resolution
The question asked how to makeup to the reactor coolant system (RCS) when on alternate letdown. Based on reference material supplied by the training department, the facility recommendation was accepted, and the answer key changed to accept a. or d.
The question asked about the design of the auxiliary feedwater system valve MS-T-63 and the affect of a loss of instrument air and was based on information contained in the system training manual. The design basis document supplied by the training department shows this valve to be a fail closed valve and, therefore, does not need an air accumulator for closing upon loss ofinstrument air. Maine Yankee will revise the system training manual at the earliest possible opportunity. - The facility recommendation to delete this question was accepted.
The question asked which of the emergency core cooling systems would be injecting into the core at an RCS pressure of 200 psia. Reference material provided by Maine Yankee show the shutoff head of the low pressure safety injection pumps to be 190 psig; thus, they will begin injecting at approximately 205 psia. This difference between psig and psia was the basis for changing the correct answer and the recommendation to change the answer to d. vice a. was accepted.
The question asked which of the reactor protection system (RPS) trips will protect the fuel from exceeding the specified acceptable fuel design limit (SAFDL) on fuel centerline melt. The final safety analysis report (FSAR) discusses that depending on core life; the variable overpower j trip (VOPT) may be needed to prevent exceeding the SAFDL. In addition to this information, the lesson plan showing the VOPT basis was provided by training. The facility recommendation to accept both ] answer c. or d. was accepted.
_ I ' 3.0 EXIT MEETING On November 19,1993, an exit meeting was held to summarize generic concerns and j observations the NRC examiners noted daring the week including the items mentioned above.
j In addition, the following item was notal: The training department provided good support during both the preexam review and
the exam week. This support included providing good comments on the written l examination, identifying and correcting potential problems with the simulator scenarios and job performance measures (JPMs), and prompt attention to the j simulator when problems developed.
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Maine Yankee Personnel Contacted: Robert Blackmore Plant Manager A. J. Cayia Manager, Operations Mike Evringham-Section head, Operations Training Jon Kirsch Supervisor, Operator Training Graham M. Leitch Vice President, Operations John Niles Assistant Manager, Operations Art Shoen Manager, Training Jim Weast Engineer, Licensing
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ATTACIIMENT 1
SRO WRI'ITEN EXAMINATION ANT r
ANSWER SIIEET , & f
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NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION 1
, CANDIDATE'S NAME: . FACILITY: H_aineYankee REACTOR TYPE: PWR-CE DATE ADMINISTERED: 93/11/15 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.
' CANDIDATE'S , TEST VALUE SCORE % - , 100.00 % TOTALS i FINAL GRADE All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature ' ! ) ) i i i
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- NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS l
. During the administration of this examination the following rules apply: 1.
Cheating on the examination means an automatic denial of your application j .and 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.
3.
Restroom trips are to be limited and only one applicant at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
4.
Use black ink or dark pencil ONLY to facilitate legible reproductions.
5.
Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer sheet.
, i 6.
Mark your answers on the answer sheet provided.
USE ONLY THE PAPER I PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
' ' 7.
Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writ.ing your answers on the examination question page.
8.
Use abbreviations only if they are commonly used in facility literature.- i Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer.
Write it out.
9.
The point value for each question is indicated in parentheses after the
question.
[ 10.
Show all calculations, methods, or assumptions used to obtain an answer to any short answer questions.
11.
Partial credit may be given except on_ multiple choice questions.
. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
12.
Proportional grading will be applied.
Any additional wrong information that is provided may count against you.
For example, if a question is worth one point and asks for four responses, 'each of which is worth 0.25-
points, and you give five responses, each of your responses will be worth j 0.20 points.
If one of your five responses is incorrect, 0.20 will be ! deducted and your total credit for that question will be 0.80 instead of i 1.00 even though you got the four correct answers.
i 13, Ii *he intent of a question is unclear, ask questions of the examiner only.
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When turning in your examination, assemble the completed examination with ! examination questions, examination aids and answer sheets.
In addition, j-turn in all scrap paper.
.15.
Ensure all information you wish to have evaluated as part of your answer l .is on your answer sheet.
Scrap paper will be disposed of immediately i following the examination.
l t 16.
To pass the examination, you must achieve a grade of 80% or greater.
l i - 17. There is a time limit of four (4) hours for completion of the examination..
18.
When you are done and have turned in your examination, leave the examination area (EXKMINER WILL DEFINE THE AREA).
If you are found in this area while the examination is still in progress, your license may be denied'or revoked.
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. SElfIOR REACTOR OPERATOR Page: 7 ' -.. QUESTION: 001 -(1.00) . !' Select the routine surveillance which has Technical Specification implications.
, a.
N2 and H2 Inventory Log i b.
Heat Tracing Checks
i c.
ECCS Lite Box Test d.
Transformer Log.
! f i . QUESTION: 002 (1.00) . -Operators are expected to perform all of the following with the [.' exception of: a.
replacement of fuses with "like for like" fuses.
b.
decon of boron crystals on pump.and valve glands.
, c.
checking that pipe caps are in place where required.
- d.
providing containers or hoses to floor drains for leak - collection.
, QUESTION: 003 (1.00) During refueling, with the SOS as the. Refueling Supervisor, a' fire occurs in the turbine building.
The fire brigade leader will be the: a.
Plant Shift Superintendent (PSS) b.
' Shift Operating Supervisor (SOS) c.
Control Room Operator (CRO)
d.
Secondary. Auxiliary Operator (SAO)- - -3
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] ' ,. QUESTION: 004 (1.00)- l All of the following fire protection equipment have Technical-Specification requirements with the exception of the: . a.
'cardox system.
' b.
smoke detectors.
c.
penetration fire barriers.
d.
portable fire extinguishers.
- . QUESTION: 005 (1.00) l ' At the SOS /PSS direction, the cable and clip device for a locked valve.
may be left in its free (unlocked) position if the valve is: ' a.
positioned by a procedure which does not specify locked.
b.
being used for system venting and draining.
c.
not in its required locked. position.
d.
inside a white tagged boundary.
t , QUESTION: 006 (1.00) < Select the Steam 3enerator parameter which specifically has a Technical Specification requirement associated with it.
. a.
pH b.
Oxygen l c.
Activity d.
Chloride 'i
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QUESTION: 007 (1.00) A tagging order cannot be cleared until all:
a.
portable grounding devices are removed from the system by Electrical Department personnel.
b.
surveillances are completed and the system is ready to be restored to service.
, QC blue hold tags are cleared from the system by QPD personnel.
c.
! d.
work has been completed and the work order has been closed out.
, QUESTION: 008 (1.00) r Select the temporary modification which requires yellow tagging.
a.
Installation of leak repair clamps.
, '! b.
Installation of stem blocking devices.
c.
Removal of fuses for white tag boundary.
' d.
Installation of jumpers for I&C calibration.
i QUESTION: 009 (1.00) { ' Any person removed from the Tagging and Switching List due to violation of the tagging procedure, cannot be reinstated without the approval of' the: , a.
Plant Manager.
. b.
individual's cognizant supervisor.
l r Manager of Operations Department.
[ c.
F d.
individual's cognizant department head.
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QUESTION: 010 (1.00) ! An area has been locked for. access control to protect individuals from exposure to radiation and radioactive materials.
i This area has been locked because:
a.
radiation levels in the area could result in an individual receiving a dose equivalent in excess of 1000 mrem in one hour { at 12 inches from the source.
b.
loose-surface activity is greater than or equal to 1000 dpm/100 cm2 (gross beta-gamma activity) or 20 dpm/100cm2 (gross alpha
' activity.
c.
airborne radioactive materials exist in concentrations such
that an individual present in the area without respiratory , ' protective equipment could exceed an intake of 0.6 percent-of the Annual Limit on Intake (ALI) or 12 DAC-hours.
, d.
radiation levels in the area could result in an individual receiving a dose equivalent in excess of 5 mrem in one hour at 12 inches from the source of radiation.
t , i ' QUESTION: 011 (1.00) A priority Radiation Work Permit (RWP) requires a pre-job briefing to be I conducted by the:
a.
Plant Shift Superintendent
I b.
Radiological Controls Technician l c.
Job Supervisor d.
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'Page 11 e ,e , QUESTION: 012 (1.00) { . The maximum dose to the extremities (including the whole body component) i that an individual can receive during an emergency to protect vital
' equipment is: I a.
75 rem - b.
100 rem c.
125 rem
d.
300 rem ! QUESTION: 013 (1.00) Select the LOWEST event classification which requires-plant staff with
emergency duties to report to their assigned emergency center and all other plant staff to exit the plant and assemble at the energy information center.
a.
Unusual Event b.
Alert .; ! ' c.
Site Area Emergency d.
General Emergency i QUESTION: 014 (1.00) Select the MINIMUM protective action recommendation developed for a General Emergency.
j a.
Evacuate Wiscasset and Westport.
Shelter Woolwich, Edgecomb, [ and Boothbay.
i b.
Evacuate Wiscasset and Westport.
Shelter towns downwind to 5 miles.
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Shelter Wiscasset, Westport, Woolwich, Edgecomb, and Boothbay.
d.
Shelter Wiscasset, Westport, and towns downwind to 5 miles.
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- SEMIOR' REACTOR OPERATOR Page.12 c; ' QUESTION: 015 (1. 0 0)
The Hot Line (white phone) is used by the PSS during emergency notifications to notify the:
a.
State-Police ' b.
Emergency Response Personnel , c.
Nuclear Regulatory Commission .' d.
State Nuclear Safety Inspector f ' QUESTION: 016 (1. 0 0) > Select the procedure type for which the Temporary Procedure Change ' (TPC) process may be used.
'
a.
"0" Series Procedures b.
Emergency Procedures < c.
Class A Procedures d.
Vendor Procedures ' , QUESTION: 017 (1.00) Select the drawing type for which a DCR would NOT have to receive written concurrence from the PSS/ SOS.
a.
ESK b.
FE one-line ! c.
High number FM (70A through 104) d.
Electrical Distribution Book (EDB)_ ' ! , I , .
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l . -. h QUESTION: 018 (1.00)' With the LPSI/CTMT Spray Pump S (P-61S) aligned to replace CTMT Spray Pump A (P-61A), which one of the following conditions would result in P-61S tripping upon receipt of a Recirculation Actuation Signal (RAS)? a.
LPSI/RHR Pump A (P-12A) breaker racked down.
, b.
LPSI/RHR Pump A (P-12A) control switch in " Pull-to-Lock".
l c.
CTMT Spray Pump A (P-61A) breaker racked in.
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CTMT Spray Pump A (P-61A) control switch in mid position.
QUESTION: 019 (1.00) l
With the plant at full power, the CRO reports that the CEA mimic indication for one CEA is white.
This white indication indicates that the CEA is:
a.
dropped to the bottom of the core.
l . b.
at the lower electrical limit.
l .. c.
between the upper and lower electrical limit.
,
d.
at-the upper electrical limit.
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QUESTION: 020 (1.00) l Select the Reactor Trip Circuit Breaker combination which if open would result in a Reactor trip with all trippable rods on the bottom.
a.
1, 2, 3,
b.
1, 2, 5,
c.
2, 3, 6,
d.
3, 4, 7,
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. QUESTION: 021 (1. 00) The method by which a non-trippable CEA is prevented from inserting is that a Reactor' trip: a.
does not de-energize the upper gripper.
b.
energizes the lower gripper at the same time the upper gripper is de-energized.
c.
energizes the non-trip coil (in place of the anti-ejection > device) at the same time the upper gripper is de-energized.
' d.
does not energize the non-trip coil (in place of the anti-- ejection device).
QUESTION: 022 ( 1. 0 0) With the Reactor at full power, a CEA Withdrawal Prohibit (CWP) is generated by 2 out of 4 Reactor Protection System channels reaching the , pretrip setpoint on: ' a.
High Pressurizer Pressure, Variable Overpower, or High Startup Rate.
b.
Variable overpower, High Pressurizer Pressure, or Symmetric Offset.
j I c.
High Startup Rate, Thermal Margin / Low Pressure, cnr Symmetric Offset.
! d.
Variable Overpower, High Startup Rate, or Thermal Margin / Low
Pressure.
i QUESTION: 023 (1.00) I Select the parameter which would be used to differentiate between a Reactor Coolant Pump (RCP) SHEARED rotor and a LOCKED rotor.
! a.
pump amps , b.
loop flow
c.
pump eccentricity , .i d.
loop differential pressure I
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The major problem with operating Reactor Coolant Pumps (RCPs) without- ' seal injection flow is-that.
a.
hot water enters the seal cavity causing the seal cartridge.O-rings to breakdown and fail.
t i b.
seal differential pressures are not adequate causing seals to remain un-staged.
c.
corrosion products enter the seal cavities causing the seal to , degrade or fail.
j , d.
bleedoff flow is not adequate causing the seal to degrade or fail.
.i QUESTION: 025 (1.00) . Select the condition which would result in the de-borating demineralizer releasing boron back into the coolant (i.e. inlet boron concentration less than outlet boron concentration).
a.
Demineralizer saturated with boron.
, ' b.
High Letdown temperature.
i ' c.
Demineralizer resin decomposition.
d.
High Letdown flow.
. QUESTION: 026 (1.00) In the event of a Control Room Evacuation, the' Charging Pumps.are controlled from the: i a.
PAB Emergency Panel b.
Charging Pump Breaker '! c.
Alternate Shutdown Panel i , d.
Boron Recovery System Panel ' i j
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... QUESTION: 027. (1. 0 0) When transferring from Normal Charging to Alternate Charging, Letdown ! ' shall be transferred from Normal to Alternate prior to securing Normal charging.
. Select the heat exchanger which forms the basis for this requirement.
. a.
Regenerative Heat Exchanger b.
High Pressure Drain Cooler c.
Seal Water Heat Exchanger ' d.
Letdown Heat Exchanger QUESTION: 028 (1. 00) . When performing a boration of the Reactor Coolant System, BA-A-80 is to ' , be used only at the direction of the PSS/ SOS because: a.
any primary water remaining in the piping from previous dilutions would be added resulting in a dilution.
, b.
the boric acid "'dition rate would be to high and therefore hard to contre.
t c.
the relatively colder water would thermally shock the Reactor Coolant Pump seals.
i d.
any automatic actuations (normal or emergency) are bypassed by this flowpath.
. . QUESTION: 029 (1.00) With Normal Charging and Alternate Letdown in service, what action would be taken to makeup to the Reactor Coolant System? , a.
Reduce Alternate Letdown flow.
, b.
Establish maximum blended makeup flowrate.
c.
Place Alternate Charging in service.
> ' d.
Direct Alternate Letdown to the Quench Tank.
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. ; ] i Q ESTION: 030 (1.00) The Main' Steam-isolation (MS-219) to the Turbine Driven Main Feed Pump (P-2C) is maintained shut below 50% power.
, This is to: ,
prevent Main Steam flow through P-2C back to the. Main Turbine.
a.
b.
ensure P-2C does not inadvertently start prior to placing it in' , service at 55% power.
, prevent overheating of P-2C high pressure steam chest while c.
idle.
d.
ensure P-2C starts up on Extraction Steam from the Main Turbine.
.. QUESTION: 031 (1.00)' , With the plant at 75% power, if the Turbine Driven Main Feed Pump'(P-2C) tripped,: (ASSUME NO OPERATOR ACTIONS.)
the P-2C trip relays 20TT1 and 20TT2 will start the Motor a.
Driven Main Feed Pumps.
b.
the Main rurbine will trip, if neither Motor Driven Main Feed Pump is operating.
, the standby Motor Driven Main Feed Pump will start on low Steam c.
Generator level.
d.
Steam Generator level will decrease resulting in a Reactor Trip.
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QUESTION: 032 (1.00) .Given the following plant conditions:
- 1 Steam Generator pressure = 350 psig level = 200" (WR)
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- 2 Steam Generator pressure = 450 psig level = 30%
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- 3 Steam Generator pressure = 450 psig level = 40%
} - - Pressurizer pressure = 1700 psig - Containment pressure = 25 psia . Select the response of the Emergency Feedwater (EFW) System control and , isolation valves.
a.
The EFW control and isolation valves for all 3 Steam Generators will be shut.
b.
The EFW control and isolation valves to #1 and #3 Steam Generators will be shut, and the EFW control and isolation , valves to #2 Steam Generator will be open.
c.
The EFW control and isolation valves for all 3 Steam Generators will be open.
d.
The EFW control and isolation valves to #1 Steam Generator will.
be shut, and the EFW control and isolation valves to #2 and #3 Steam Generators will be open.
F i QUESTION: 033 (1.00) The Reserve Air Receiver (TK-25) provides backup air to' he: 'igry Feed Pump Turbine Steam Suppl Isolation Valve (MS-T-a.
1631 forsplosing on a Safety Injection Actuation Signal or a - Containmen'tsI' solation Signal.
s
s s b.
Emergency Feedwaterszsolat-ion Valves (AFW-A-338, 339, 340) for ~ closing on a low Steam Generator pressure of 400 psig.
- ss s c.
Auxiliary Feed Pump' urbine Steam Supply Pressure Control' Valve
s (MS-P-168) for up'to 12 hours of' continued feed pump steam-I pressure regulation.
d.
Emergency,Feedwater Control Valves (EFW-A-101 s201, 301) for
closing'on a low Steam Generator pressure of 400'psig.
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- -QUESTION: 034 (1.00) ! (D/G) Start Select the power supplies for the Diesel. Generators Circuitry.
D/G 1A - DC Bus 1 l a.
D/G 1B - DC Bus 2 b.
D/G 1A - DC Bus 3 ,- D/G 1B - DC Bus 4 D/G 1A - DC Bus 1
c.
D/G 1B - DC Bus 3
d.
D/G 1A - DC Bus 2 D/G 1B - DC Bus 4 , QUESTION: 035 (1.00) ! If'the core was uncovered, what would the Wide Range Logarithmic Channel - ' indication do? " Increase due to the reduction of boron concentration in the a.
Core.
Decrease due to less moderation of neutrons.. ! b.
Decrease due to the fuel temperature' coefficient and heatup'of c.
the fuel.
. , Increase due to less reflection of neutrons.
d.
QUESTION: 036 (1.00) . in that Select the bistable which generates a Symmetric Offset trip, channel, if actuated above 15% power.
, Wide Range Logarithmic Channel Level 1 bistable.
a.
l Power Range Safety Channel Rod Drop bistable b.
a Wide Rarage Logarithmic Channel Level 2 bistable c.
Power Range Safety Channel High Voltage bistable
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! -QUESTION: 037 (1. 0 0) l When Motor Control Centers (MCCs) are. cross-tied for purposes.of ground- , isolation, an operator is stationed at the micro-versa trip device to- { cnsure-immediate detection of an overload condition to preclude a trip ! of'the feeder breaker which would result in a loss of both MCCs.
The' micro-versa trip device would indicate such an overload' condition by displaying: a.
a red overcurrent trip flag.
j b.
a blinking digital current indication.
, c.
a steady bright red light.
! d.
an overcurrent alarm message.
j , QUESTION: 038 (1.00) The Anticipated Transient Without Scram / Diverse Scram System'(ATWS/ DSS) trips the Reactor by: sensing Pressurizer pressure and opening the motor-generator ' a.
output AC contactors.
b.
providing a redundant trip signal from the Reactor Protection ' System to the motor-generator output circuit breakers.
independently sensing Pressurizer pressure and opening the c.
Reactor Trip Circuit Breakers.
- d.
providing a-redundant trip signal.from the Reactor Protection-System to the motor generator input AC interrupters.
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y . SENIOR REACTOR OPERATOR Page 21 . QUESTION: 039-(1.00) Salect the condition which would prevent the Primary Inventory Trend System (PITS) reading from becoming unreliable due to reference leg.
flashing.
Containment pressure less than 5 psig.
j a.
b.
Reactor Coolant System pressure greater than atmospheric.
. c.
Reactor Coolant Pumps secured.
! d.
Containment Ventilation System operating.
. QUESTION: 040 (1.00) Select the condition which would prevent a Reactor Coolant System.(RCS) Loop Cold Leg Isolation Valve from OPENING.
a.
Reactor Coolant Pump operating.
b.
RCS Loop Bypass Isolation Valve closed.
c.
RCS Loop Hot Leg Isolation Valve closed.
d.
Delta T between loop Tcold's greater than 35 degrees F.. QUESTION: 041 (1.00) Following a Loss of Coolant Accident (LOCA), RCS pressure is at 200 psia and' dropping.
Which ONE of the following statements describes the expected status of High Pressure Safety Injection (HPSI), Low Pressure Safety Injection . .(LPSI), and Safety Injection Tank (SIT) flow? a.
HPSI and SIT flow exist with no LPSI flow.. o b.
HPSI and LPSI flow exist with no SIT flow.
i c.
HPSI flow exists with no SIT or-LPSI flow.
- d.
HPSI, SIT, and LPSI flow exist.
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SENIOR REACTOR OPERATOR Page 22 ' t QUESTION: 042 {1.00) Select the flowpath that is used for Hot Leg Injection.
' Refueling Water Storage Tank to Containment Spray Pumps to High a.
Pressure Safety Injection Pumps to Residual Heat Removal Suction piping.
b.
Refueling Water Storage Tank to High-Pressure Safety Injection Pumps to Reactor Coolant System Fill piping.
Containment Spray Sump to Containment Spray Pumps.to High c.
Pressure Safety Injection Pumps to Reactor Coolant System Fill P ping.
i d.
Containment Spray Sump to High Pressure Safety Injection Pumps to Residual Heat Removal Suction piping.
' i QUESTION: 043 (1.00)' The High Pressure Safety Injection Pump subsystem is the only Emergency Core Cooling System subsystem which has Technical Specification operability requirements based on which one of the following: a.
reactor coolant temperature.
I b.
reactor coolant boron concentration.
? c.
reactor coolant pressure.
d.
reactor operating condition.
-i QUESTION: 044 (1.00) i If Diesel Generator DG-1A failed to start coincident-with a Loss Of offsite Power, the affect on the Power Operated Relief Valves and their z associated Isolation Valves would be that there would be no electrical
power to: , a.
both Relief Valves PR-S-14 and PR-S-15.
, b.
Relief Valve PR-S-14 and Isolation Valve PR-M-16.
- c.
both' Isolation Valves PR-M-16 and PR-M-17.
d.
Relief Valve PR-S-15 and Isslation Valve PR-M-17.
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.. ! QUESTION: 045 (1.00)
The consequences of leaving the Pressurizer Spray Valve manual bypass valves closed is that:
a.
Pressurizer pressure would be maintained higher due to normally i energized Pressurizer heaters.
, b.
the Pressarizer spray line could brittle fracture if auxiliary spray were used due to the relatively colder water.
c.
the Pressurizer spray' head would heat up and subsequently , themal shock would occur when spray flow was initiated.
d.
Water hammer would damage the nozzle head when spray flow is l initiated.
- t i QUESTION: 046 (1.00) , Select the Pressurizer Level Control action which is a function of actual Pressurizer level (NOT level deviation from setpoint).
a.
De-energize all Pressurizer heaters.
b.
Energize all Pressurizer backup heaters.
> c.
HI PRESSURIZER LEVEL alarm.
! d.
LO PRESSURIZER LEVEL alarm.
. QUESTION: 047 (1.00) > ! The basis for programmed Pressurizer level being 34% at 0% power is to: l t a.
prevent emptying the Pressurizer on a Reactor trip.
l t b.
ensure the heaters remain covered on a Reactor trip.
., , c.
prevent going solid plant on a complete loss of Feedwater.
d.
ensure Reactor maneuvering will be acceptable at, design rates.
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' SENIOR REACTOR OPERATOR Page 24 . , QUESTION: 048 (1.00) ~Salect the Reactor Protection System trip which is intended to protect .the core from exceeding the Specified Acceptable Fuel Design Limit on ' fuel centerline melt.
' a.
. Low Reactor Coolant Flow b.
Thermal Margin / Low Pressure c.
Variable Nuclear Overpower f d.
Symmetric Axial Flux Offset ? QUESTION: 049 (1.00) Given the following Reactor Protection System (RPS) conditions: - RPS Channel B High containment Pressure in Channel Bypass.
- RPS Channel C High Containment Pressure in Trip.
If RPS Channel A High Pressurizer Pressure were to trip, the results on the plant would be that the: a.
plant would trip since the logic was 1 out of 2 prior to the .i RPS High Pressurizer Pressure trip channel tripping.
' b.
In order to trip the plant, additional channels. required to trip would be 2 out of 2 for High Containment Pressure and 2 out of 3 for High Pressurizer. Pressure.
c.
plant would trip since there are now two channels in a tripped condition.
d.
In order to trip the plant, additional channels required to trip would be 1 out of 2 for High Containment Pressure and 1 out of.
3 for High Pressurizer Pressure.
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. QUESTION: 050 (1.00) .The Loose Parts Monitoring System monitors the: a.
Reactor Coolant Pumps [ r b.
Hot and Cold Legs
. c.
Reactor Vessel i d.
Pressurizer ! . QUESTION: 051 (1.00) ' . The keylock NORMAL / EMERGENCY switch (69-DG-LA) ~for Diesel Generator DG-1A in the NORMAL position,:
- >
a.
enables the Bus 3 to Bus 5 tie breaker (3TS) synchronizing interlocks.
] > b.
opens the Bus 3 to Bus 5 tie breaker' (3T5) on a Safety - Injection Actuation Signal.
c.
allows DG-1A to supply Bus 3 from Bus 5.
,
d.
enables the load shedding sequence on a Safety Injection' Actuation Signal.
i-QUESTION: 052 (1. 0 0) C A Diesel Generator START FAILURE alarm will result if the: _
- j a.
engine speed is not at 900 rpm in 10 seconds.
b.
generator voltage is not at 4160 volts in 10 seconds, a c.
Main Control Board Diesel Generator control switch not in auto.- d.. engine start relay is not picked up within 1 second of an auto start signal.
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, QUESTION: 053 (1. 0 0) 'The Technical Specification definition of CONTAINMENT INTEGRITY includes-requirements for Containment: a.
air temperature.
b.
leakage rates.
c.
internal pressure.
. d.
weight of air.- QUESTION: 054 ( 1. 0 0) The Reactor Protection System senses a Main Turbine trip by: a.
2 out of 4 turbine stop valves shut.
, r b.
2 out of 4 turbine governor valves shut.
c.
2 out of 2 auto stop oil pressure switches actuated at less than 45 psig.
t d.
Electro Hydraulic Control (EHC) pressure less than 2000'psig at the EHC accumulators.
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. QUESTION: 055 (1.00) , Given the following plant conditions: - Diesel Generator supplying Bus 6 following an undervoltage condition.
'
-t - CRO has placed ACB 4U in PULL-TO-LOCK and reset Bus 6 undervoltage relay 27Y6 after diesel load shed sequence.is ! complete.
, If prior to the. event the configuration of the Secondary Component Cooling pumps were P-10B operating and P-10A in auto, what is.their . status now? a.
P-10B-in' auto, P-10A operating.
b.
P-10A in auto, P-10B operating.
c.
P-10B secured, P-10A operating.
, d.
P-10A and P-10B operating.
.. -QUESTION: 056 (1.00) Select the parameter that is used to generate a Steam Dump and Turbine j Bypass System " Quick Open" signal, a.
Pressurizer pressure ,
b.
Turbine trip signal , c.
Steam Generator pressure d.
Reactor trip signal ! ! i ? I , 'I , I
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,Page 28 ' , e ! QUESTION: 057 (1.00) r The Turbine Bypass System Mode Select switch in the COOLDOWN position:- l allows only three bypass valves to be operated.
a.
. b.
allows only two bypass valves to be operated.
! bypasses the steam header pressure interlock.
e c.
d.
bypasses the condenser interlock.
l t QUESTION: 058 (1.00)
ES-1.4, " Establishing Hot Leg Injection", requires hot leg injection to , be established no later than 19 hours after Safety Injection initiation.
This maximum time limit is based on: t a.
Refueling Water Storage Tank level.
' b.
boron precipitation.
c.
Steam Generator availability.
d.
core decay heat.
, QUESTION: 059 (1.00) When the brass handle key operated switch on the 480V bus for each of [ the Pressurizer Backup Heaters is placed in the EMERGENCY position,.it: i a.
transfers the heater control from the Main control. Board to the f Alternate Shutdown Panel.
b.
permits.the heaters to only be operated by local pushbuttons on
' the heater breaker.
c.
transfers the heater control to the PAB Emergency Panel.
d.
overrides all heater cutout signals.
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. . I E QUESTION: 060 (1.00).
i ' Select the condition which, during normal operations, would be E considered a Control Element Assembly (CEA) misalignment.
a.
A CEA in group.4 is out of position from the remainder'of the . group by 12 steps.
b.
A CEA in subgroup SA is out of position from subgroup SB by 12.
.; steps.
j c.
Group 4 CEA positions differ from Group 5 CEA positions by 12 steps.
. d.
Subgroup 5A CEA positions differ from subgroup SB CEA positions by 12 steps.
i QUESTION: 061 (1.00) With TWO CEAs in the same group inoperable, continued operation is ! permitted: provided boron concentration is adjusted in accordance with a.
. ' inoperable CEA curves within 2 hours.
! b.
provided the remaining CEAs are' aligned to the inoperable CEAs < within 4 hours.
, I c.
only if low power physics-testing is in progress.
d.
only if reactor power is below 50%. . r a b ) .
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. , QUESTION: 062 (1.00) i Prior to aligning a dropped.CEA using the "Long Term Slow Recovery" , procedure, Reactor power is reduced to less than 50%. Reactor power is reduced to 50% to ensure that: , a.
radial tilts resulting from the dropped CEA will not cause power levels associated with the different loops.to be unequal.
! ' b.
peaking induced by CEA withdrawal will not cause fuel pre- , conditioning limits on pellet clad interaction to be exceeded.
c.
radial tilts resulting from the dropped CEA will not cause subsequent xenon oscillations.
d.
peaking induced by CEA withdrawal will not cause a Symmetric Off Set Reactor Trip.
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QUESTION: 063 (1.00) , The initial action performed for a dropped CEA is to reduce Turbine load , to match: , e a.
steam flow and feed flow and stabilize Steam Generator level.
] > b.
steam flow and feed flow and stabilize Reactor power.
{
c.
Tave and Pref and stabilize Pressurizer pressure.
l d.
Reactor power and stabilize Tcold, i . I h i
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' ! - , , QUESTION: 064 (1.00) ( Given the following plant conditions: Reactor power at 20% with power ascension in progress.
-I ' - Group 5 CEAs at 140 steps.
I - .i' - Shortly after assuming the shift, you notice that the CEAs start stepping out.
The required action you would direct the CRO to take is to: ] a.
increase Turbine load to match Tref with Tave.
l ! b.
initiate emergency boration to compensate for the reactivity ! addition.
c.
trip the Reactor and carry out E-0, " Reactor Trip Or Safety Injection".
j i d.
depress the OFF pushbutton on the Control Element Drive System
(CEDS) panel.
, , QUESTION: 065 (1.00) g Select the indication which could be used to determine that a Loss Of Coolant Accident was occurring instead of an Excess Steam Demand Event? j a.
Pressurizer level b.
Containment radiation ! fi c.
Pressurizer pressure d.
Containment pressure j f .[ f . ? . t s % ! ! p .
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, QUESTION: 066 (1.00) l Given the following plant conditions: i A Loss Of Coolant Accident (LOCA) has occurred.
- Core region subcooling is 25 degrees F.
- t Reactor Coolant Pumps (RCPs) have been secured.
- , The RCPs are secured under these conditions to: a.
decrease the amount of water mass inventory lost through the . ' break, therefore enhancing efforts to keep the core covered.
! b.
increase the flow of steam (instead of two-phase mixture) from the break, therefore enhancing heat removal from the core.
, decrease the-cold leg pressure head, therefore enhancing Safety c.
, Injection System performance at higher flow rates.
, d.
increase flow stability of the RCS by allowing for development f of natural circulation, therefore enhancing slow, controlled heat removal.
. QUESTION: 067 (1.00) Following a Reactor trip, the Turbine fails to trip both automatically and manually.
, , Select the contingency action required to be performed immediately.
a.
Close EFCVs and NRVs.
b.
Secure EHC HP fluid pumps and place in PULL-TO-LOCK.
l l c.
Locally trip the Turbine, d.
Manually initiate Containment Isolation Signal (CIS).
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. fSENIOR= REACTOR OPERATOR' Page 33- [ .. ! QUESTION: 068 (1.00) , S31ect the event which creates the'most potential for Pressurized ' Thermal Shock (PTS)? a.
Loss Of Coolant Accident ' b.
Steam Generator Tube Rupture - c.
Excess Sceam Demand
i d.
Loss Of All Feedwater QUESTION: 069 (1. 0 0) If all three seals fail on a Reactor Coolant Pump (RCP), the Reactor and ' then the affected RCP are tripped.
' Select the indication which would indicate a failure of all three seals , on a RCP.
. a.
Red Pen (Middle Seal Pressure) = 2200 Blue Pen (Upper Seal Pressure) = 2000
Green Pen (Low Pressure Seal Pressure) =0 i i b.
Red Pen (Middle Seal Pressure) = 1200 Blue Pen (Upper Seal Pressure) = 200 Green Pen (Low Pressure Seal Pressure) =0 , c.
Red Pen (Middle Seal Pressure) = 200 Blue Pen (Upper Seal Pressure) = 200 Green Pen (Low Pressure Seal Pressure) = 200 .,i d.
Red Pen (Middle Seal Pressure) =0 Blue Pen (Upper Seal Pressure) =0 t Green Pen (Low Pressure Seal Pressure) =0 > 'l i
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QUESTION: 070 (1.00) , i All of the following Reactor Coolant Pump (RCP) indications have trip criteria associated with them except for: ! a.
decreasing Primary Component Cooling flow.
b.
decreasing seal return flow.
-r c.
increasing RCP motor winding temperature.
. d.
increasing seal temperature.
QUESTION: 071 (1.00) In the event of a loss of Primary Component Cooling, Diesel Generator DG-1A is provided backup cooling from:
a.
Fire Water.
b.
Secondary Component Cooling.
c.
Service Water.
d.
Auxiliary Charging Pump P-7.
e i QUESTION: 072 (1.00) ,{ ! When attempting to mitigate an Anticipated Transient Without Scram , -(ATWS), Emergency Boration is initiated from the Boria Acid Storage Tank l to the Charging Pumps.
If response is not obtained, the preferred order for alternate Emergency Boration sources is.
t a.
1) Boric Acid Storage Tank to Auxiliary Charging Pump , 2) Refueling Water Storage Tank to Auxiliary Charging Pump .l b.
1) Refueling Watar Storage Tank to Charging Pumps 2) Refueling Water Storage Tank to Auxiliary: Charging Pump t c.
1) Refueling Water. Storage Tank to Auxiliary Charging Pump.
- 2) Boric Acid Storage Tank to Auxiliary Charging Pump d.
1) Refueling Water Storage Tank to Charging Pumps 2) Boric Acid Storage Tank to Auxiliary Charging Pump ] . > > b . - . . - - .
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.- , ' QUESTION: 073 (1.00) - i While' implementing PR-C.2, " Degraded Core Cooling" (orange path), a loss . of.all AC power. occurs followed by a red path'in heat sink.-
Solect the appropriate action to be taken.
a.
Go to ECA-0.0, " Loss Of All AC Power".
b.
Go to FR-H.1, " Loss Of Secondary Heat Sink" (red path).
c.
Go to E-0, " Emergency Shutdown From Power Or Safety Injection".
d.
Continue with FR-C.2, " Degraded Core Cooling" (orange path)'. , .i .i QUESTION: 074 (1.00) . During Emergency Operating Procedure implementation, harsh Containment.
instrument values should be used whenever one of two criteria are met.
One of the criterin is based on Containment Isolation Signal 1(CIS)
. actuation and the other is based on: a.
Containment Spray Actuation Signal actuation.
b.
Safety Injection Actuation Signal actuation.
c.
Containment pressure and temperature, d.
Containment radiation levels.
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. . . _- - . .~ __ .!, '& ' SENIOR REACTOR OPERATOR Page'36 i J ., . QUESTION: 075 (1.00) Given the following. Critical Safety Function Status Tree conditions: l 1) Containment- - RED 2) Core Cooling - RED
3) Heat Sink - YELLOW
4) Integrity - ORANGE ! 5) Inventory - YELLOW ! 6) Suberiticality - YELLOW j , Select the order in which the applicable Functional Restoration-
Procedures should be implemented.
I a.
1, 2, 4, 6, 5,
.} ! ~ b.
2, 1, 6, 4, 5,
t c.
2, 1, 4, 6, 3,
i . d.
1, 2, 6, 4, 3,
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I ~l ! QUESTION: 076 (1.00) .; A manual Reactor trip is required to be initiated if Condenser-differential pressure increases to greater than: , a.
3.5 in. HgA dP.
b.
5 in. HgA dP.
, c.
10 in. HgA dP.
i d.
11 in. Hgh dP.
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k QUESTION: 077- (1.00).
i The initia1' action to be performed in the event of a LOSS OF INSTRUMENT ' BUS DP/IAC alarm is to: depress tne.ATWS DSS. trip pushbuttons due to the loss.of a.
Control Element Assembly position indication.- j b.
commence rapid power reduction until the Heater Drain Tank high level dump can maintain level, c.
shift to the alternate power supply, MCC-7A, inside Main Control Board section B.
. d.
start Diesel Generator DG-1B and re-energize the bus from MCC- ' 8A.
QUESTION: 078 (1. 0 0)
Several of the Plant Shutdown For Fire Abnormal Operating Procedures authorize violations of Technical Specifications.
, These violations are authorized because the event occurring could:
a.
result in loss of vital plant equipment necessary for plant shutdown and cooldown.
b.
immediately threaten the health and safety of the public.
c.
result in uncontrolled radiological releases to the , environment.
i' d.
immediately jeopardize fire fighting efforts and safety.
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' QUESTION: 079 (1.00) E The Plant; Shutdown For Fire Abnormal Operating Procedure requires implementation of E-0, Emergency Shutdown From~ Power Or Safety
Injection, for fires in all the following areas except the: ' a.
Containment.
b.
Turbine Hall'. c.
Emergency Feed Pump Room.
d.
Control Room (habitable).
- ' i QUESTION: 080 (1.00) ,
Select the action that is performed at the Primary Auxiliary Building Emergency Panel during a Control Room Evacuation.
a.
Maintaining Reactor Coolant System temperature.
b Operating Emergency Feedwater Pumps.
c.
Maintaining Pressurizer level.
. d.
Operating Reactor Coolant Pumps.
, QUESTION: 081 (1.00) Select the shift position that is responsible for locally tripping the Turbine during a Control Room Evacuation for reasons other than a fire.
e a.
Plant Shift Superintantent (PSS) f b.
Shift Operating Supervisor (SOS) -i c.
Control Room Operator #1 (CRO1) , d.
Secondary Auxiliary Operator (SAO) , , i ?
~ - , '*' SENIOR' REACTOR OPERATOR Page 39 j f ,,- ( QUESTION: 082 (1.00)' [ i If ONE train of Safety Injection (SI) actuates at full power, what is the minimum course and sequence of actions that are to be taken? , , ' a.
Trip the reactor, implement the Emergency Operating Procedures, and manually fire the other SI train.
[ b.
Trip the plant and implement the Emergency Operating Procedures.
, , Manually fire the other SI train and reset both trains of SI in c.
accordance with the Emergency Operating Procedures.- , d.
Reset the one train of SI in accordance with the Emergency [ Operating Procedures.
i QUESTION: 083 (1.00)
If a Pressurizer code safety were leaking, the downstream safety relief temperature would indicate approximately the same temperature as the:
a.
saturation temperature for the Pressurizer.
! I b.
Pressurizer vapor space.
t c.
saturation temperature for the Quench' Tank.
d.
Quench Tank temperature.
i r QUESTION: 084 (1.00) Bullets (o) are used in the Emergency Operating Procedures to indicate i steps that: , a.
are immediate actions.
! b, can be completed in any order.
, c.
require continuous performance.
! d.
transition to other procedures.
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JSENIORl REACTOR OPERATOR-Page 40
, i e' r p-l QUESTION:'085 (1. 0 0 ). During natural circulation cooldown and depressurization following a . Loss Of Coolant Accident (LOCA), upper head voiding may occur.
'I The expected indication of upper' head voiding is rapidly: a.
increasing Thot.
i b.
increasing Pressurizer level.
t c.
decreasing Pressurizer pressure.
d.
decreasing natural circulation flow.
, , ! QUESTION: 086 (1.00) ES-1.2, Post LOCA Cooldown And Depressurization, provides a caution about failure to isolate SITS when Steam Generator (SG) pressure is .; reduced to less than 80 psig.
. The SITS are isolated to prevent: i a.
unnecessary depletion of SIT inventory due to Reactor Coolant System cooldown.
! l b.
a loss of heat sin ~k due to nitrogen blanketing of the SG.U- ' tubes.
c.
Reactor Coolant flow into the SIT on the loop with low-SG pressure, d.
void formation in the Reactor Vessel head.
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.- ~'~~SEH10R REACTOR OPERATOR 'Page'41- , ..5 i: i ) QUESTION: 087 (1.00) The minimum steam generator tube leak that requires implementation of E- ,
0,.. Emergency Shutdown From Power Or Safety Injection,-and manual initiation of safety injection is:
. the capacity of the Charging Pumps.
a.
> t b.
100 gpm.
c.
50 gpm.
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, d.
Technical Specification limits.
! QUESTION: 088 (1.00) During recovery from a Steam Generator (SG) tube leak IAW AOP 2-49, the Reactor Coolant System Loop Stop Valves are shut based on , a.
SG level.
L b.
Pressurizer level.
c.
leakage rate and time.
d.
RCS pressure and temperature.
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QUESTION: 089 (1.00) It is important to maintain the ruptured Steam Generator (SG) level ! above the top of tne SG U-tubes during a SG Tube Rupture recovery.
If the SG U-tubes uncover: , SG pressure may decrease causing the leakage rate from the RCS a.
into the SG to increase.
, , b.
the affected SG will not have any other means'of heat removal ' if the RCS loop stops are closed to isolate the affected SG.
.! they would heat up unevenly possibly resulting in further SG c.
tube failures.
l t d.
the affected SG will pressurize and the SG safeties could lift' ' releasing radioactivity to the atmosphere.
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QUESTION: 090 (1.00) > The major concern with using backfill during a Post Steam Generator Tube- 'l ' ! Rupture Cooldown is that the-i , Reactor Coolant System chemistry ends up out of specification.
a.
q b.
activity released to the environment is greater.
Pressurizer pressure control restoration ends up being. delayed.
'l c.
' d.
shutdown margin is reduced due to boron dilution.
? QUESTION: 091 (1.00) , During post Steam Generator (SG) tube rupture cooldown using backfill, j the operator places the Charging flow restrictor in service prior to > depressurizing the Reactor Coolant System (RCS) below 375 psig.
l This is done to: , l prevent overpressurization of the RCS.
a.
b.
ensure adequate Residual Heat Removal flow.
, prevent runout of the Charging Pumps.
c.
, d.
ensure pressurizer does not fill solid.
l .c I i QUESTION: 092 (1.00) . The most likely cause of-an " air bound" Low Pressure Safety Injection , i pump while running for Residual Heat Removal (RHR) is: a.
throttled flow.
i b.
rising suction temperature.
! i ' c.
lowering suction level.
l d.
venting the RHR heat exchanger.
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. QUESTION: 093 (1.00) . Following a loss of Residual-Heat Remoral, it is decided to let the ' R: actor Coolant System heat up to 212 degrees F'and use the Steam G:nerators.
This' mode of heat removal will require: i a.
the Low Pressure Safety Injection Pumps.to be secured.
b.
Containment Integrity to be established.
c.
a Reactor Coolant Pump to be started.
d.
the Pressurizer to be filled solid.
, QUESTION: 094 (1.00) " Technical Specifications requires at least one bank of Pressurizer ! proportional heaters to be operable whenever the Reactor Coolant System l (RCS) Tavg is greater than 500 degrees F.
' Pressurizer heaters are required to maintain: a.
minimum subcooling margin for Reactor Coolant Pump operation.
' b.
RCS overpressure with a primary leakage in progress.
' c.
RCS and Pressurizer boron concentrations equalized.
d.
conditions necessary for natural circulation flow.
t ? QUESTION: 095 (1.00) Feedwater is not restored rapidly to a Steam' Generator (SG) with narrow range levels less than 56% to preclude: a.
the possibility of causing a SG tube rupture.
b.
-water hammer damage to the feedwater piping and feed ring.
, .c.
-excessive Reactor Coolant System cooldown.
' d.
' shrinking SG level below the top of the SG U-tubes.-
, .
L' SENIOR' REACTOR OPERATOR' Page 44
( :.. , , -QUESTION: 096 (1. 0 0) , In addition to being controlled by the Offsite Dose Calculation Manual (ODCM),.there is Technical Specification operational requirements for-the radiation monitor (s) monitoring the:- , , a.
Waste' Gas Effluent.
b.
Steam Generator Blowdown.
c.
Primary Vent Stack.
. k d.
Service Water Effluent.
. > QUESTION: 097 (1.00) Select the valves which will fail OPEN on a loss of Instrument Air.
i a.
AS-P-3 and AS-P-653 - b.
CH-A-32 and CH-A-33 , c.
CS-A-55 and CS-A-56 d.
PCC-T-20 and SCC-T-23 , ^ QUESTION: 098 (1. 0 0) . Pressurizer level can indicate HIGHER than actual level due to: l a.
a leak in the reference leg of the Pressurizer level transmitter, since_this causes the transmitter differential pressure to be HIGHER for a given actual. level.
b.
a leak in the variable leg of the Pressurizer level transmitter, since this causes the transmitter differential pressure to be LOWER for a_given actual level.
c.
low Pressarizer pressure and temperature, since this causes'the Pressurizer level transmitter differential pressure to be' LOWER , for a given actual level.
d.
high Containment pressure and temperature, since this causes the Pressurizer level transmitter differential pressure'to be HIGHER for a~given actual level.
, I ,
. e*- SENIOR REACTOR OPERATOR.
Page.45-f _ -a 'n'
7 QUESTION: 099 (1.00).
_v . . i-lFollowing a Reactor trip with loss of offsite power, AOP 2-46, Loss Of i Offsite Power While Shutdown Or Post Trip, is implemented.
As required =by th'is procedure,- the Steam Generator Non-Return Valves ~(NRVs) are.
shut.
The NRVs are shut to prevent:
a.
damage to the Main Condenser.
) b.
loss of inventory from the Secondary.
c.
an unmonitored release to the environment.. !- d.
excessive Reactor Coolant System cooldown.
..i ! QUESTION: 100 (1.00) ! d Select the radiation monitor which should automatically isolate- ! Containment Purge (refueling mode) if a Refueling Accident were to ' occur.
] a.
Containment High Range Radiation Area Monitor , ! b.
Containment Particulate / Gas Monitor j c.
Primary Vent Stack Particulate / Gas Monitor j d.
Manipulator Crane Area Monitor . l t .! l
'l t ! l (**********.END OF EXAMINATION **********) - i .
y.-
- }
,- ... SENIOR' REACTOR OPERATOR Page 46 _j - l - e
[ ANSWER: .001 _(1.00)
'b.
i ! ~ REFERENCE:
Procedure No. 1-200-4, Operations Department Routing Schedule, pg.
1.' " , SRO-L-10.8, ELO# SNO-0303-13 l l [2.7/3.9) , 194001A109 ..(KA's) ANSWER: 002 (1.00) - , ! a.. REFERENCE: Procedure No. 1-200-10, Conduct Of Operations, pg. 7 & 8.
SRO-L-10.2, ELO# SNO-0302-4 [3.1/4.1] >
194001A112 ..(KA's) ANSWER: 003 (1.00) b.
, REFERENCE: Procedure No. 1-200-10, Conduct Of Operations, pg.
9.
.l SRO-L-10.2, ELOf SNO-0302-1
' [3.5/4.2] 194001K116 ..(KA's)
. ANSWER: 004.
(1.00) d.
!: , .3
s ,
. ' SENIOR REACTOR OPERATOR Page 47 -' ,
..-
- REFERENCE
Technical Specification 3.23, pg. 3.23-1 through 3.23-3.
, SRO-L-10.10,>ELO#1SNO-0303-13 l . -[3.5/4.2] .194001K116 ..(KA's) , ANSWER: 005 -(1.00) , , d.
REFERENCE: , Procedure No. 1-200-10, Conduct Of Operations, pg.
7.
, NO FACILITY ELO FOUND [3.6/3.7] 194001K101 ..(KA's) . ANSWER: 006 (1.00) , c.
REFERENCE: Technical Specification 3.24, pg. 3.24-1.
SRO-L-10.10, ELOf SNO-0303-13 [2.5/2.9) , 194001A114 ..(KA's) ANSWER: 007 (1.00) c.
' . , l l - - - - - - _ _ - _ _ _ _-
i i SENIOR-REACTOR OPERATOR Page 48
, +- _ REFERENCE: l Procedure No. 0-14-1, White Tagging Procedure, pg. 17.
'SRO-L-10.1, ELO# M-0302-12 i [3.7/4.1) . 194001K102 ..(KA's) . . ' ANSWER: 008 ( 1. 0 0) b.
s REFERENCE:
= Procedure No. 0-14-2, Temporary Modification Control, pg. 1 through 5.. i SRO-L-10.1, ELO# M-0302-5 . ! [3.7/4.1] , 194001K102 ..(KA's) t ' ANSWER: 009 ( 1. 0 0) a.
. REFERENCE: Procedure No. 0-14-1, White Tagging Procedure, pg. 13.
SRO-L-10.1, ELO# M-0302-7 . [3.7/4.1) , 194001K102 ..(KA's) ANSWER: 010 (1.00)
B.
, , i
9 i n f ' -. .
'l ' ' %1 ' - :*' SENIOR-REACTOR; OPERATOR .Page 49 ' REFERENCE: ' , -Procedure:No.'9-2-100,_ Access Control And Radiation Work' Permit Program, . pg.
4.
RO-L-9.1, ELO/_AD-0103-6- ., l .; [2.8/3.4) . 194001K103 ..(KA's) , -ANSWER: 011 (1.00) b.
REFERENCE: , Procedure No. 9-2-100, Access Control And Radiation Work Permit Program, l pg.-25.
- RO-L-9.1, ELO# AD-0103-8
[2.8/3.4] 194001K103 ..(KA's) , LANSWER:- 012 (1.00) b.
t REFERENCE: { Procedure No. 2-50-12,' Radiological Controls Coordination, pg. 10.
'RO-L-9.1, ELO# AD-0103-3 ,
[2.8/3.4]. 194 001F1.03 ..(KA's)
ANSWER: 013 (1.00)
'b.
> , e i t \\
P , '
- LSENIOR REACTOR OPERATOR Page 50 a
L I: REFERENCE: , i i
- Procedure No. 2-50-2, Alert, pg.
2.
' RO-L-10.1, ELO/ AD-0102-11 , i Facility Exam Bank # RO-AD-02-011-10-01-04 [3.1/4.4] 194001A116 ..(KA's) , b ANSWER: 014 (1.00) , , c.
i , REFERENCE: Procedure No. 2-50-16, Protective Action Recommendations, pg.
6.
l SRO-L-10.6, ELO# SEO-0301-6 t > [3.1/4.4] 194001A116 ..(KA's) j r ' ANSWER: 015 (1. 0 0) a.
! REFERENCE:
Procedure 2-50-1, Unusual Event, pg.
2.
SRO-L-10.6, ELO# SEO-0301-1 [3.1/4.4] .
-194001A116 ..(KA's) ) ? ANSWER: 016 (1.00) r F C.
t -
l i r s W
. * SENIOR REACTOR OPERATOR Page Sin +. REFERENCE: Procedure.No.s0-06-2,-Administrative: Controls For Procedures And ~ Procedure' Changes, pg. 17.
i SRO-L-10. 3, J ELOf G A-03 02-1 i - [3. 3 / 3.4 ]' , '194001A101 ..(KA's) J ,
ANSWER: 017 (1.00) b.
, LREFERENCE: .P
Procedure No. 0-06-3, Drawing Contro), pg.
4.
SRO-L-10.3, ELO# GA-0303-2 . [2.5/3.2] , 194001A107 ..(KA's)
' ANSWER: 018 (1.00) I ' a.
k REFERENCE: OP MEMO 9-E-11, Actions To Prevent HPSI Pump Damage On A RAS,-rev 2.
RO-L-2.6, ELO# CS-0902-2 & ECCS-0101-11-i [4.2/4.2]
r 026000K101 ..(KA's) .; ! . ANSWER: 019 (1. 00)
C.
, , F
- l o
.,. _ I SENIOR REACTOR OPERATOR Page 52
. ' REFERENCE: Systems Training Manual Chapter 12, Control Element Drive System, Figure: .; .12-21.
RO-L-4.4, ELOf CEDS-0102-9 -[2.6/2.9)
014000K404 ..(KA's) ANSWER: 020 (1.00)
l C.
REFERENCE: -, Systems Training Manual Chapter 11, Reactor Protection' System, Figure , 11-16a.
.; RO-L-4.4, ELO# CEDS-0101-1 [3.6/3.7)
001000K202 ..(KA's)' i
. ANSWER: 021 (1.00) d.
REFERENCE: " Systems Training Manual Chapter 12, Control Element Drive System, pg.- , ' 12-16.
r RO-L-4.4, ELO# CEDS-0101-8 .) [3.2/3.4] ,, l 001000K408 ..(KA's) ! 't ANSWER: 022 (1.00)
d.
, l , t ,
SENIOR, REACTOR OPERATOR Page 53 , < ~ , -- REFERENCE: .i Systems Training-Manual Chapter 12, Control Element Drive System, pg.
-12-44.
RO-L-4.4,-ELO# CEDS-0101-9 M -[4.5/4.4]- 001000K105 ..(KA's) - ANSWER: 023 (1.00)
a.
REFERENCE: , Facility Exam Bank # RO-RCP-01-014-02-01-05 (MODIFIED) RO-L-2.1, ELOf RCP-0101-14
[2.7/3.1] 003000A203 ..(KA's) ANSWER: 024 (1.00) c.
REFERENCE:
Systems Training Manual Chapter 2, Reactor Coolant System,_pg. 2-81.
RO-L-2.1, ELO# RCP-0902-3 [2.7/3.1] 003000K602 ..(KA's)
! ANSWER: 025 (1. 00) i b.
, ! .
i &
i
'~" SENIOR REACTOR OPER ATOR-Page 54~ - -s
- REFERENCE:
-Fccility Exam Bank # RO-CVCS-02-021-02-05-01-&. RO-CVCS-02-021-02-05-07 (HODIFIED) RO-L-2.5, ELO# CVCS-0102-21 ~[2.5/2.8) 004020K503 ..(KA's) ANSWER: 026 (1.00) b.
REFERENCE: l Facility Exam Bank ( RO-CVCS-02-009-02-05-01 RO-L-2.5, ELO# CVCS-0102-10
. [3.6/3.4] I 004000G009 ..(KA's) t ! ANSWER: 027 (1.00) a.
7 REFERENCE: ' Procedure No. 1-11-6, Chemical And Volume Control System Operation (CVCS), pg. 24.
! RO-L-2.5, ELOf CVCS-0102-7 [3.7/3.3] , 004020A402 ..(KA's) .;
ANSWER: 028 (1.00) { c.
! > F . h ,
. ,- > ' SENIOR REACTORLOPERATOR' _Page 55 b
.s
- ' REFERENCE:
- Procedure No. 1-11-5, RCS Boron. Control And' Chemical Addition, pg.
2.
RO-L-2.5, ELOf CVCS-0103-15 ! , -[3.1/3.4] 0040000010 ..(KA's) - ANSWER: 029 (1.00)
- f b
.a.
REFERENCE: ' RO-L-2.5, Chemical And Volume Control System (CVCS), pg. 80 & 81.
RO-L-2.5, ELO# CVCS-0103-16 [3.9/3.7] [ 004010A403 ..(KA's) ANSWER: 030 (1.00) , a.
- > REFERENCE:
Procedure No. 1-104-5, Turbine Driven Feed Pump Operation, pg.
1.
, RO-L-3.7, ELO# MFW-0102-3 [2.7/2.9] , 059000A103 ..(KA's)
- i
ANSWER: 031 (1.00)
b.
i .t ? !
, f i f i
I: > - . i .' ' ' SENIOR' REACTOR OPERATOR Page 56 l{ - - i -! ., rREFERENCE: I Procedure No. 1-104-5, Turbine' Driven Feed-Pump Operation, pg.
1.
! ! RO-L-3.7, MainLFeed System-(FW), pg. 84.
'RO-L-3.7, ELO#'MFW-0102-17 l
[3.0/3.3] 059000A207 ..(KA's) . ANSWER: 032 (1.00)
i r d.
REFERENCE: Systems Training Manual Chapter 28, Emergency & Auxiliary Feedwater ' System, pg. 28-24.
Facility Exam Bank # RO-AFW-01-014-03-13-02 (MODIFIED) i RO-L-3.13, ELO# AFW-0101-14 .[
[4.2/4.2] '
- '
061000A301 ..(KA's) i ANh ER: 033 (i. 00) / 'i
> a.
,( \\ / M~ LEE -
REFERENC / . Systems, ining Manual Chapter 28, Emergency & Auxiliary Feedwater- .; System,jpg.\\ging11550-FM-81G 28-12 through 28-17.
~ Facility Dra ! RO-L-3 '.13, ELO. AFW-0101-10 /. - i / [3.2/3.6] p ' '061000A202 ..(KA's) ANSWER: 034 (1.00) C.-
i-I
. SENIOR REACTOR OPERATOR Page 57 ' . ! . REFERENCE:
Systems Training Manual Chapter 33, Low voltage Electrical Systems, pg, i 98 & 100.
, RO-L-5.4, ELO# DG-0101-7 l ' [ [3.7/4.1] 063000K301 ..(KA's) ,
, I i ANSWER: 035 (1. 0 0) j d.
, REFERENCE: RO-L-4.2,-Excore Nuclear Instrumentation, pg. 45.
RO-L-4.2, ELOf NI-0101-12 , [2.6/3.1]
015000K504 ..(KA's) ! !
ANSWER: 036 (1.00) , d.
, REFERENCE: ! Excore Nuclear Instrumentation .: Systems Training Manual Chapter 14,' [ System, pg. 14-20.
RO-L-4.2, ELOf NI-0102-5 ! [3.9/4.3] o 015000K301 ..(KA's) -)
-1 ANSWER: 037 (1.00) c.
l .
i l l
. __ . - - - . .. . 'l p' l f
- f
- - SENIOR REACTOR OPERATOR Page 58~
i t - ! I.T- . i - 1 REFERENCE: { OP MEMO 9-M-1,-Lessons Learned From'A.C. Ground Isolation, rev.
0..
I , 'SRO-L-10.2,.ELO# SNO-0302-9- , } [3.0/3.1] !
- 062000A301
..(KA's) i l i ANSWER: 038 ( 1. 0 0) l --, -a.
, REFERENCE: i Systems Training Manual Chapter 11, Reactor Protection System, pg. 11-45.
RO-L-4.5, ELOf RPS-0902-3 l [3.9/4.0] f 012000K301 ..(KA's) !
? ANSWER: 039 (1.00) . 'd.
REFERENCE:
! Systems Training Manual Chapter 2, Reactor Coolant. System, pg. 2-58 & 2-60.
RO-L-2.1, ELOf RCS-0102-11
! [3.1/3.6] ! 002000K603 ..(KA's)
I-ANSWER: 040 (1. 0 0) - a d.
',
E , ! l .! )
i ~
. ' * ' SENIOR T REACTOR OP'ER ATOR Page'59' .i
l REFERENCE: ,r ' Systems Training Manual Chapter 2, Reactor Coolant System,~pg. 2-51- 'through 2-54.
j RO-L-2.1, ELO/.RCS-0101-6 [3.3/3.5] , . ! 002000A107 ..(KA's) , l ANSWER: 041 (1.00) < .b % REFERENCE: ! RO-L-2.6, Emergency Core Cooling System, pg.
6,-17, & 20.
, RO-L-2.6, ELO# ECCS-0101-14 [4.2/4.3] > 006030A102 ..(KA's) , i ANSWER: 042 (1.00) , t C.
REFERENCE: c) Systems Training Manual Chapter 6, Emergency Core Cooling System, pg.
6-l-11.
' RO-L-2.6, ELO# ECCS-0101-1 [4.2/4.3] , 006000K103 ..(KA's) ! P ANSWER: 043 (1.00) b.
!
. .!
i ' .
- SENIOR REACTOR: OPERATOR Page 60"
- . REFERENCE: { T chnical Specification 3.6, Emergency Core Cooling And Containment i Spray Systems, pg.
3.6-1.
' SRO-L-10.8, ELO# SNO-0303-13 [3.5/4.2] +
006000G005 ..(KA's) ANSWER: 044 (1.00) b.
REFERENCE:
Systems Training Manual Chapter 33, Low Voltage Electrical Systems, PGS-- 18-1 Electrical Power Distribution 8/92.
i-Systems Training Manual Chapter 3, Pressurizer And Pressure Relief I System, pg. 3-17.
Facility Exam Bank # RO-PPC-01-011-02-04-02 RO-L-2.4, ELO# PPC-0101-11 i [2.8/3.0) 010000K203 ..(KA's) ! ANSWER: 045 (1.00) l - C.
REFERENCE: Systems Training Manual Chapter 3, Pressurizer And Pressure Relief' I System, pg. 3-12.
.RO-L-2.4, ELOf PPC-0902-2 [3.2/3.6)
010000K603 ..(KA's) { ANSWER: 046 (1.00) a.
. T t i
! ,
. . ... - . . "' SENIOR REACTOR OPERATOR .Page 61 ' ,. REFERENCE: Systems-Training Manual Chapter 3, Pressurizer And Pressure Relief { System,.pg. 3-44 & 3-45.
RO-L-2.4, Pressurizer And Pressure Relief System,_pg. 21.
? RO-L-2.4, ELO# PLC-0101-8 [3.3/3.7] .
l 011000K401 ..(KA's)
ANSWER: 047 (1. 00)
, { a.
< REFERENCE: ! RO-L-2.4, Pressurizer And Pressure Relief System, pg. 48.
l RO-L-2.4,-,!LO# PLC-0101-1 l [3.2/3.4] 011000K513 ..(KA's) ANSWER: 048 (1. 00) .d.
Cl[ () REFERENCE: ! Technical Specification 2.1, Limiting Safety System Setting - Reactor . Protection System, pg.
2.1-2.
'RO-L-4.5, ELO# RPS-0101-20 ., ! [4.4/4.7] 012000A206 ..(KA's)
ANSWER: 049 (1. 0 0) d.
, 'l .P !
.,.
- SENIOR REACTOR OPERATOR-Page 62
- -.
.- REFERENCE: , Systems Training Manual Chapter 11, Reactor Protection System, pg. 11-55.
RO-L-4.5, ELO# RPS-0101-2 & RPS-0902-3 [3.6/3.6] 012000A403 ..(KA's) , I ANSWER: 050 (1.00) c.
REFERENCE: RO-L-4.5, Reactor Protection System, pg. 73.
Facility Exam Bankf RO-LPM-01-002-04-05-02 RO-L-4.5, ELOf LPM-0101-2 [3.4/3.4] 016000K101 ..(KA's) ANSWER: 051 (1.00) . , a.
REFERENCE: RO-L-5.4, Diesel Generators, pg. 51.
RO-L-5.4, ELO# DG-0102-5 [3.1/3.3) 064000A209 ..(KA's) 1 ANSWER: 052 (1.00) d d.
i >
.
' * ; SENIOR;. REACTOR OPER ATOR - Page 63 - ! ' -REFERENCE: Systems Training Manual Chapter 34, Diesel Generators And Appendix "R" Diesel,And Electrical Distribution System AS-12, pg.152.
Fccility Exam Bank # RO-DG-41-009-05-04-01 (MODIFIED) i NO FACILITY OBJECTIVE FOUND , l [2.8/3.2] > 064000K409 ..(KA's)
, iE ANSWER: 053 (1.00)
b.
! REFERENCE: -Technical Specification 3.11, Containment, pg. 3.11-1.
RO-L-2.12, ELO# C-0101-3 [3.3/4.1] 103000G005 ..(KA's) ! t ANSWER: 054 (1.00) , c.
. REFERENCE: Systems Training Manual Chapter 11, Reactor Protection System, pg. 11-27.
, Facility Exam Bank # RO-RPS-01-002-04-05-08 l RO-L-4.5, ELO/ RPS-0101-2 & RPS-0101-4 (3.3/3.5) 045050A301 ..(KA's) ANSWER: 055 (1.00) .. b.
, l
l l ~! ! .
SENIOR REACTOR OPERATOR Page 64 ~* . REFERENCE: ' Systems Training Manual Chapter 34, Diesel Generators And Appendix "R" - Diesel And Electrical Distribution System AS-12, pg. 60 & 61.
+ Facility Exam Bankt RO-DG-02-004-05-04-03-RO-L-5.4, ELOf DG-0102-4 RO-L-2.3, ELO# CC-0101-5 [3.1/3.3] 008000K401 ..(KA's) , -ANSWER: 056 (1.00) b.
.; REFERENCE: Systems Training Manual Chapter 25, Steam Dump And Turbine Bypass System, pg. 25-13.
RO-L-3.4, ELOf SD-0101-4 , [2.8/3.0] 041000G007 ..(KA's) t ANSWER: 057 (1.00)
- C.
REFERENCE: Systems Training Manual Chapter 25, Steam Dump And Turbine Bypass System, pg. 25-21.
- RO-L-3.4, ELOf SD-0101-5
[3.1/3.3] j 041020A405 ..(KA's) '
ANSWER: 058 (1.00) b.
! , h !
SENIOR REACTOR OPERATOR Page 65 - ' ' REFERENCE: -
- Systems Training Manual chapter 6, Emergency Core Cooling System, pg.
6- ' 45.
RO-L-2.6, ELO# ECCS-0101-15 [3.8/4.2] 000011K313 ..(KA's) ANSWER: 059 (1.00) b.
REFERENCE: Systems Training Manual Chapter 3, Pressurirer And Pressure Relief-System, pg. 3-32.
Facility Exam Bank ( RO-PPC-02-010-02-04-02 RO-L-2.4, ELOf PPC-C102-10 [4.1/4.2) 000068A107 ..(KA's) ANSWER: 060 (1.00) a.
REFERENCE: Procedure No. AOP 2-21, Misaligned (Dropped) CEA, pg.
1.
Facility Exam Bank ( RO-EAS-50-002-08-10-03 (MODIFIED) RO-L-8.10, ELO# EAS-0150-1 [2.9/3.8) 000005x106 ..(KA's) ANSWER: 061 (1.00) c.
.
m 'I a. SENIOR REACTOR OPERATOR ' Page-66 a REFERENCE: , ..
- Procedure No. AOP.2-23', Inoperable CEA, pg.
3.
! .RO-L-8.10,.ELO# EAS-0903-5 i . [3.5/4.4) i I 000005A203.
..(KA's) .; - - . ANSWER: 062 (1. 0 0) b.
REFERENCE: , Procedure No. AOP 2-21, Misaligned (Dropped) CEA, pg.
2.
RO-L-8.10, ELO# EAS-0150-4 [3.4/4.1) l 000003K305 ..(KA's) 'i . I ANSWER: 063 (1.00) d.
_ REFERENCE:- , Procedure No. AOP 2-21, Misaligned (Dropped) CEA, pg.
3.
RO-L-8.10, ELOf EAS-0150-4 & EAS-0903-5 [4.1/4.1) ] ' 000003A105 ..(KA's) ,
i
ANSWER: 064 (1.00) d.
' -; I , t . .! .i
' -+
r SENIOR REACTOR OPERATOR Page : 67- "
- ..
- REFERENCE: -Procedure'No. AOP 2-22, Uncontrolled Reactivity Addition, pg.
3.
Facility Exam'Bankf RO-EAS-50-005-08-10-05 RO-L-8.10, ELOf=EAS-0150-4 & EAS-0903-5 [3.9/4.0] 000001G010 ..(KA's) . ANSWER:' 065 (1.00) b.
REFERENCE: .RO-L-8.27, E-1, ECA-1.2; Loss'Of Primary Or Secondary Coolant, pg.:2.
RO-L-8.27, ELO# EAS-0903-6 [3.7/3.7] 000011A213 ..(KA's) ANSWER: 066 (1.00) a.
, REFERENCE: Procedure No.
E-0, Emergency Shutdown From Power Or Safety Injection, pg.
5.
Facility Exam Bankf RO-EAS-01-007-08-27-11 RO-L-8.27, EAS-0101-7 [4.1/4.2] 000011K314 ..(KA's) ANSWER: 067 (1.00) b.
.. ...-.. .. 4 SENIOR-REACTOR OPERATOR Page 68 ' . . s :.
. REFERENCE: Procedure No.
E-0, Emergency Shutdown From Power Or Safety Injection, , pg.
4.
, RO-L-8.31, ELO# EAS-0903-3
, ,[4.1/4.2] ! ' -000040G010 ..(KA's)
. ANSWER: 068 (1.00) C.
, , REFERENCE: Procedure No. ES-1.1, SI Termination, pg.
7.
RO-L-8.31, ELO# EAS-0903-12
[4.1/4.4) t 000040K101 ..(KA's)
ANSWER: 069 (1.00) C.
, , REFERENCE: Procedure No. AOP 2-50, Reactor Coolant Pump Abnormalities, pg. 12.
> RO-L-8.8, ELOf RCP-0101-20
[4.0/4.2] .j 000015A122 ..(KA's) .i , ANSWER: 070 (1.00) b.
. Y , ' i i l
'l . . ....
h &
~ } b
- J
.h ~ -d cSENIOR ' REACTOR OPER ATOR .Puse 69. -r . . .
- '
REFERENCE: Procedure No 'AOP 2-50, Reactor Coolant Pump Abnormalities, pg. 3 & 5.
! .RO-L-8.8, ELO# EAS-0903-11 & RCP-0101-19 i [3.7]/3.7] f 000015A210 (KA's) .. i ANSWER: 071 (1.00)
a.
REFERENCE: Procedure No. AOP 2-33,-Degraded Or Loss Of Primary Component. Cooling, pg. 3 & 5.
. RO-L-8.15, ELO# EAS-0903-11 , [3.2/3.3) 000026A102 (KA's) .. , i ANSWER: 072 (1.00) d.
REFERENCE: Procedure No. FR-S.1, Nuclear Power Generation /ATWS, pg.
4.
l RO-L-8.34, ELO# EAS-0903-12 ,
[4.2/4.3) ' 000029K311 (KA's) ..
- I
' f ANSWER: 073 (1.00) I a.. ~ -i f P R 'i & I t .
L 1* SENIOR' REACTOR OPERATOR Page 70, I . REFERENCE: ' Procedure No. 1-200-16, Rules Of Usage For Emergency Operating Procedures And Critical Safety _ Function Status Trees, pg.
6.
Facility Exam Bank / RO-EAS-09-003-08-23-03 'RO-L-8.23, ELO# EAS-0903-3 [3.9/4.0) 000055G012 ..(KA's) ANSWER: 074 (1.00) d.
REFERENCE: , Procedure No. 1-200-16, Rules Of Usage For Emergency Operating Procedures And Critical Safety Function Status Trees, pg.
4.
Facility Exam Bankt RO-EAS-01-009-08-23-06 RO-L-8.23, ELO# EAS-0903-3 [3.8/4.2] 000069K301 ..(KA's) ANSWER: 075 (1.00) c.
REFERENCE: Procedure No. 1-200-16, Rules Of Usage For Emergency Operating _ Procedures And Critical Safety Function Status Trees, pg.
6.
Facility Exam Bank ( RO-EAS-01-002-08-23-04 RO-L-B.23, ELOf EAS-0101-2 & EAS-0101-3 [4.3/4.4] 000074G012 ..(KA's) ANSWER: 076 (1.00) a.
.
SENIOR REACTOR OPERATOR Page 71
L ,
- .
REFERENCE: , Procedure-No. AOP 2-2, Degraded Or Loss Of Condenser Vacuum, pg.-3.
l RO-L-8.1, ELO# EAS-0903-11 & EAS-0150-4
[3.9/4.1] 000051A202 ..(KA's) , B ANSWER: 077- (1.00) c.
REFERENCE: Procedure No. AOP 2-37, Non-Safeguards Annunciators, pg. 73.
RO-L-8.5, ELO# EAS-0903-11 '
[3.6/3.7] 000057G010 ..(KA's) ANSWER: 078 (1.00) , -i b.
, -REFERENCE: . -r ' Procedure No. AOP 2-90-4, Plant Shutdown For Fire In PAB Elevation 21', pg. 2.
i RO-L-8.21, ELO# EAS-0150-8 & EAS-0903-12 [3.3/4.1] _, 000067K304 ..(KA's) . ANSWER: 079 _(1.00) i , b.
< i - k l a
' .; ~ '- SENIOR ~ REACTOR OPERATOR Pagn 72
.
- ., REFERENCE: Procedure No. AOP 2-90-1, Plant Shutdown Plan For Fire In: Control ! Room..., pg.
4.
' Procedure No. AOP.2-90-2, Pla-t Shutdown Plan For Fire In I Containment...- pg.
2.
. Plant Shutdown. Plan For Fire In Emergency Feed i , . Procedure No. AOP 2-90-6, Pump Room, pg.
2.
Procedure No. AOP 2-90-7, Plant Shutdown Plan For Fire In Turbine ' Hall..., pg.
2.
RO-L-8.19, ELOf EAS-0903-11 RO-L-8.21, ELOf EAS-0903-12 i [3.3/4.4] l
000067A213 ..(KA's) 'i
ANSWER: 080 (1.00) c.
REFERENCE: Procedure No. AOP 2-44, Control Room Evacuation, pg.
3, 6 ', 10, &1l.
RO-L-8.20, ELO# EAS-0150-4 & EAS-0903-11 [3.9/4.0] 000068K201 ..(KA's)
- j t
ANSWER: 081 (1.00) - a.- ., ) REFERENCE: , Procedure No. AOP 2-44, Control Room Evacuation, pg.
8.. , RO-L-8.20, ELO/ EAS-0903-11 [3.8/4.O] l 000068G012 .. (KA's) q ANSWER: 082 (1.00) j a.
i . .
. SENIOR L REACTOR OPER ATOR' Page 73
+ r iREFERENCE: ' OP FU2(O 9-E-12, Safety Injection Actuation System (SIAS), rev.
2.
RO-L-8.28, ELOf EAS-0903-3 , , [4.3/4.6] 000007A202 ..(KA's) ANSWER: 083 (1.00) c.
REFERENCE: , . Systems Training Manual Chapter 3, Pressurizer And Pressure Relief i System, pg. 3-14.
RO-L-2.4, ELO# PPC-0902-2 [3.2/3.7) , 000008K101 ..(KA's) ANSWER: 084 (1.00) b.
. REFERENCE: Procedure No. 1-200-16, Rules For Usage For Emergency Operating , Procedures And Critical Safety Function Status Trees, pg.
4.
RO-L-8.23, ELOf EAS-0101-4 . [3.8/3.9) 000007G012 ..(KA's) . ' ANSWER:. 085 (1.00)
- b.
-. - - - -.
., . . -. .. . .
~ M ':-SENIOR REACTOR OPERATOR' Page 74 ~ .., .l , ., . REFERENCE: , Procedure No..ES-1.2, Post LOCA Cooldown And Depressurization, pg.
18.. ' RO-L-8.29,.ELO# EAS-0101-7 & EAS-0903-6-i [4.2/4.7] )
000009K101 ..(KA's) ,
- ANSWER:
086 (1.00) 't b.
REFERENCE:
, Procedure No. ES-1.2, Post LOCA Cooldown And Depressurization, pg.-21.- RO-L-8.29, ELO/-EAS-0101-7
-[3.0/3.3] 000009K203 ..(KA's) . ANSWER: 087 ( 1. 0 0). C.
> REFERENCE: , Procedure No. AOP 2-49, Steam Generator Tube Leak -(CRS-1)'. (CRS-2),. pg. - 1.
. ~RO-L-8.42, ELO/ EAS-0150-8 & EAS-0903-8
[4.2/4.4] 000037K307- ..(KA's)
. ' ANSWER: 088 (1.00)
-; C.
? I .5 i f +
. ... -
_ -- - , ' SENIOR REACTOR OPERATOR Paga 75 - * ' < JREFERENCE: , Procedure No. AOP 2-49, Steam Generator Tube Leak (CRS-1) (CRS-2), pg.. 6.
RO-L-8.42, ELO/ EAS-0150-8 & EAS-0903-14 [3.0/2.9] 000037A103 ..(KA's) .,
> j' ANSWER: 089 (1.00) ' .a.
REFERENCE: , Facility. Exam Bank # RO-EAS-01-007-08-32-27- - RO-L-8.32, ELO# EAS-0101-007 . [4.2/4.5] ! 000038K306 ..(KA's) , t ANSWER: 090 (1.00) , d.
REFERENCE: Procedure No. ES-3.1, Post-SGTR Cooldown Using SG Backfill, pg. 2 & 3.
RO-L-8.32, ELO# EAS-0101-7 & EAS-0903-12- [3.6/3.8) 000038G007- ..(KA's) . f ANSWER:- 091 (1.00) -l a.
i ! t a l l
, ,
,;. . ! -.. Page 76 iSENIOR. REACTOR OPERATOR s-REFERENCE: 'Fccility Exam Bank / RO-EAS-01-007-08-32-28 , RO-L-8.32, ELO# EAS-0101-007'
.s (4.0/3.8] 000038A130 ..(KA's)
ANSWER: 092 (1.00) < C.
REFERENCE: Procedure No. AOP 2-34, Loss of Core Decay Heat Removal Capability While.
, Shutdown, pg.
2.
RO-L-8.16, ELOf EAS-0150-4 & EAS-0903-5 , [3.2/3.2] j i 000025K202 ..(KA's) . ANSWER: 093 (1.00) b.
' REFERENCE: Procedure No. AOP 2-34, Loss Of Core Decay Heat Removal Capability While .; Shutdown, pg. 3 & 4.
J RO-L-8.16, ELO# EAS-0150-4 & EAS-0903-11 -; i [3.2/3.4] i
000025A206.
..(KA's) . , ANSWER: 094 (1.00)
I d.
[ . i
i + ' i i - -
__ ; SENIOR. REACTOR OPERATOR Page 77- -
L -
' REFERENCE: Technical Specification 3.3, Reactor Coolant System Operational Components, pg.
3.3-1.
L RO-L-2.4, ELO/'PPC-0902-1 [2.4/3.4] 000027G004 ..(KA's)
ANSWER: 095 (1.00) , b.
REFERENCE: Procedure No. ES-0.1, Reactor Trip Response, pg.
4.
RO-L-8.25, ELO# EAS-0101-7 [3.6/4.2] 000054K102 ..(KA's) ~ ANSWER: 096 (1.00) C.
REFERENCE: Systems Training Manual Chapter 20, Radiation Monitoring System, Ing. 20-28 & 20-29.
SRO-L-10.9, ELO# SNO-0303-13 [3.2/3.9] 000060G003 ..(KA's) ANSWER: 097 (1.00) c.
- u
l :- -...*. SENIOR REACTOR OPERATOR Page 78: .. ~ . 1 REFERENCE: l' ' Procedure _ No'. AOP 2-28, Degraded Or Loss Of Control Air Pressure,: pg.
2.- . . .' R O - L - 8. 1 3, ELO# CA-0902-1 & CA-0902-3- ? ~[2.8/3.3] '000065A208 ..(KA's) ' ANSWER: 098 (1.09) c.
REFERENCE: f RO-L-2.4, ELOf PLC-0101-6 & PLC-0902-3 [3.4/3.8) 000028A202 ..(KA's) ANSWER: 099 (1.00) .a.
REFERENCE: Procedure'No. AOP 2-46, Loss Of Offsite Power While Shutdown Or Post Trip, pg.
4.
RO-L-8.7, ELO# EAS-0150-4 [4.4/4.7]. 000056K302 ..(KA's) ANSWER: 100 (1.00) 'd.
, e
_,. _ . SENIOR REACTOR OPERATOR Page 79 > , . REFERENCE: Procedure No. AOP:2-35, Refueling Accidents, pg.
2.
Systems Training Manual Chapter 20, Radiation Monitoring System, Table 20-1 & Table 20-2.
' RO-L-8.16, EL0# EAS-0903-9 , RO-L-9.5, ELO# RMS-0101-7 , l [3.4/3.9] 000036K202 ..(KA's) ! .
.! ! !
! ! , , ! .l -
i ! r i
' i .i l l ) (********** END OF EXAMINATION **********) l J
m TEST CROSS REFERENCE-Page
- *
~ ~ PWR .R e a c t o r-SRO Exam Organized by Question N. umber
QUESTION VALUE REFERENCE ' ! 001 1.03 9000218 002 1.00 9000219 003 1.00 9000220 004 1.00 9000221 > 005 1.00 9000222 006 1.00 9000223 , 007 1.00 9000224 008 1.00 9000225 , ' 009 1.00-9000226 010 1.00 9000227 011 1.00 9000228 012 1.00 9000229 [ 013 1.00 9000230 014 1.00 9000231 015 1.00 9000232 l 016 1.00 9000233 017 1.00 9000234 ' 018 1.00 9000235 019 1.00 9000236 i 020 1.00 9000237 021 1.00 9000238
022 1.00 9000239 J 023 1.00 9000240 j 024 1.00 9000241 .; 025 1.00 9000242 026 1.00 9000243
- l 027 1.00 9000244 028 1.00 9000245 029 1.00 9000246-030 1.00 9000247 j
031 1.00 9000248 032 1.00 9000249 R 033 1.00 9000250 034 1.00 9000251 035 1.00 9000252 036 1.00 9000253 037 1.00 9000254 038 1.00 '9000255-l ' 039 1.00 9000256 040 1.00 9000257 041 1.00 9000258 042 1.00 9000259-043 1.00 9000260 044 1.00 9000261 045 1.00 9000262 046 1.00 9000263-047 1.00 9000264 048 1.00 9000265 049 1.00 9000266 l J
. . ._ _ ...
'TESTJCROSS REFERENCE Pahe 2.
I ' .. SRO Exam P W.R R e~a c t o r '; t Organize'd by Ques. tion Number .
& QUESTION-VALUE REFERENCE , P 050 1.00 9000267 .l-051 1.00 9000268 '! 052 1.00 9000269 053 1.00 9000270 054 1.00.
9000271 a 055 1.00 9000272 056 1.00 9000273
057 1.00 9000274
! 058 1.00' 9000275 - [ 059 1.00 9000276' 060 1.00 9000277 061 1.00 9000278 ! 062 1.00 9000279 j 063 1.00 9000280 , 064 1.00 9000281-065-1.00 9000282 j ! 066 1.00 9000283 067 1.00 9000284-068 1.00
- 9000285
- 069-1.00 9000286 '! - 070 1.00 9000287
071 1.00 9000288
072' 1.00 9000289 j i 073 1.00 ~9000290 ! 074 1.00 9000291 075 1.00 9000292 .l 076-1.00 9000293 l 077 1.00 9000294'
078 1.00 9000295 I 079 1.00 9000296 _ ' 080 1.00 9000297 081 1.00-9000298.
.; 082 1.00 9000299.
083 1.00 9000300 084 1.00 9000301 ' 085 1.00 9000302 .1
086 1.00 9000303 .) 087 1.00 9000304 , 088 1.00 9000305 ] 089 1.00 9000306 090 1.00 9000307
091 1.00 9000308-092 1.00 9000309 093 1.00 9000310
094 1.00 9000311 q 095 1.00 9000312 i 096 1.00-9000313-l , 097 1.00 9000314 l 098 1.00 9000315 )
. . .
I TEST CROSS REFERENCE Page .3 '* - ' t .... SRO Exam PWR Reactor- . . . - , .. , -Organized by Question Nunber ,
, QUESTION.
VALUE REFERENCE > 099 1.00 9000316 100 1.00 9000317 ! ______ 100.00 ______ ~l ______ . 100.00 'f i
. I
, .- ! .; , h !
!
I ' -
! .-i .I ,-), l . j
,
- -
TEST CROSS REFERENCE Page
' S R O.
Exam PWR R.e a c t o r Organized by KA Group _ PLANT WIDE GENERICS { QU ESTION VALUE KA __ 016 1.00 194001A101 + 017 1.00 194001A107 , 001 1.00 194001A109 ! 002 1.00 194001A112 , 006 1.00 194001A114 015 1.00 194001A116 013 1.00 194001A116 014 1.00 194001A116 005 1.00 194001K101 009 1.00 194001K102 007 1.00 194001K102 008 1.00 194001K102 010 1.00 194001K103 012 1.00 194001K103 011 1.00 194001K103 004 1.00 194001K116 003 1.00 194001K116 , ______ PWG Total 17.00 PLANT SYSTEMS ,
Group I QUESTION VALUE KA , f 022 1.00 001000K105
020 1.00 001000K202 ' 021 1.00 001000K408 023 1.00 003000A203 024 1.00 003000K602 026 1.00 004000G009 028 1.00 004000G010 029 1.00 004010A403 027 1.00 004020A402
025 1.00 004020K503 019 1.00-014000K404 036 1.00 015000K301 , 035 1.00 015000K504 018 1.00 026000K101 030 1.00 059000A103 031 1.00.
059000A207 033 1.00 061000A202 i 032 1.00 063000A301 034 1.00 063000K301 i ______
' TEST CROSS REFERENCE Page
> 'S R O Exam PWR Reactor O r g a n.i z-e d by KA Group , PLANT SYSTEMS Group I ' QUESTION VALUE KA PS-I Total 19.00
Group II QUESTION VALUE KA __ 040 1.00 002000A107 039 1.00 002000K603 l 043 1.00 006000G005 042 1.00 006000K103 041 1.00 006030A102 044 1.00 010000K203 045 1.00 010000K603 ' 046 1.00 011000K401 047 1.00 011000K513-048 1.00 012000A206 049 1.00 012000A403 038 1.00 012000K301 ' 050 1.00 016000K101 037 1.00 062000A301 051 1.00 064000A209 052 1.00 064000K409 053 1.00 103000G005 ______ $ PS-II Total 17.00 Group III , QUESTION VALUE KA .; 055 1.00 008000K401 056 1.00 041000G007 , 057 1.00 041020A405
054 1.00-045050A301 !
______ PS-III rotal 4.00 ______ ! ______ PS Total 40.00 . EMERGENCY PLANT EVOLUTIONS Group I ,
TEST CROSS REFERENCE-Page
- S R O Exam PWR Reactor Orga'nized by KA Group
_ EMERGENCY PLANT EV3LUTIONS Group I QUESTION VALUE KA 064 1.00 000001G010 063 1.00 000003A105-062 1.00 000003K305 061 1.00 000005A203.
, 060 1.00 000005K106 065 1.00 000011A213 058 1.00 000011K313 066 1.00 000011K314 069 1.00 000015A122 070 1.00 000015A210 071 1.00 000026A102 072 1.00 000029K311 067 1.00 000040G010 068 1.00 000040K101 076 1.00 OOOO51A202 073 1.00 000055G012 077 1.00 000057G010 079 1.00 000067A213 078 1.00 000067K304 059 1.00 000068A107 081 1.00 000068G012 080 1.00 000068K201 074 1.00 000069K301 075 1.00 000074G012 ______ EPE-I Total 24.00 Group II QUESTION VALUE KA 082 1.00 000007A202 .084 1.00 .000007G012 083 1.00 000008K101-085 1.00 000009K101 086 1.00 000009K203 093 1.00 000025A206 092 1.00 000025K202 094 1.00 000027G004 088 1.00 000037A103 087 1.00 000037K307 091 1.00 000038A130 090 1.00 000038G007 089 1.00 000038K306 095 1.00 000054K102
TEST. CROSS REFERENCE Page
- j
SRO Exam PWR Reactor -; Organiz'ed by KA Group , EMERGENCY PLANT EVOLUTIONS i Group II QUESTION VALUE KA , 096 1.00 000060G003 097 1.00 000065A208 ______ , EPE-II Total 16.00 Group III QUESTION VALUE KA
098 1.00 000028A202 ' 100 1.00 000036K202 099 1.00 000056K302 , ______ EPE-III Total 3.00
______ ______ EPE Total 43.00 ______ . -t ______ , Test Total 100.00
- )
, , ! , t . -i
1 ' .
- - - . i a A'ITACIIMENT 2 Simulation Facility Report f' - Facility Licensee: Maine Yankee
Facility Docket Nos: 50-309 ' Operating Tests Administered from: November 15-18. 1993
This form is used only to report simulator observations. These observations do not constitute r audit or inspection findings and are not, without further verification and ieview, 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 observations.
The simulator exhibited minor problems during the examination week. During the requalification exam several incorrect digital indications occurred on the electrical control , board. Troubleshooting conducted after the exam was not conclusive, although one 5 volt power supply soltage was found low and was readjusted. During the administration of the ' initial examination operating test, oscillations of several analog indications on the electrical control board occurred. Troubleshooting conducted after the scenarios were complete , revealed a bad circuit board in the distributed I/O system. The board was replaced, and the problem did not reoccur. In addition, a potential modeling deficiency was identified during performance of the JPMs. During manual transfer to recirculation cooling, the high pressure ' safety injection pumps (HPSI) were observed to cavitate at a higher refueling water storage tank level than was expected. The level at which HPSI pump cavitation would be expected is being investigated by Maine Yankee. The effect on the candidates was minimal; all individuals responded as if the indications were real and the problems did not adversely affect the administration of the examination.
. - I I l i I
.. . , . . , . 4-ATTACHMENT 3 4-MaineYankee- !: '. ' kueou racmenum m we sinct $972
- ,
PO BOX 408. WISCASSET, MAINE 04578. (207) 882 6321 l November 18,'1993 ) MN-93-102-j ! Mr. Lee Bettenhausen
Operations Branch, DRS !
UNITED STATES NUCLEAR REGULATORY COMMISSION L 475 Allendale Road King of Prussia, PA 19406 l Reference: (a) License No. DPR-36 (Docket No. 50-309)
Dear Mr. Bettenhausen,
Enclosed are Maine Yankee's comments and supporting reference materials covering !
- Senior Reactor Operator written examination administered at our facility on
-{ November 16, 1993.
Overall, we feel-this was a very reliable and valid' examination.
Our pre-examination review of this exam revealed relatively few technical inaccuracies which i were responsively corrected.
We would like to express our' gratitude for the effort applied in creating'a high-
quality written examination and for the professional manner in which all aspects of , the examining process-were conducted.
!
Sincerely, M E) E E.-. .j / - ' A - 9, Robert W. Blackmore. , Plant Manager Enclosure c: Kerry Ihnen, Chief Examiner , t .i ! , l
e
. Novemb:r 18, 1993 MN-93-102 . Page 2 of 2
9_UESTION #
There are two correct answers to this question. The question asks how to " makeup" to the RCS when using normal charging and alternate letdown. As indicated in the question reference, Lesson Plan R0-L-2.5, this may be accomplished by decreasing alternate letdown flow to the VCT (Answer a.)
by either throttling close valve DR-A-6 or by diverting alternate letdown flow from the VCT to the Quench Tank (Answer d.). Both answers A and D should be accepted as enrrect.
Reference: Maine Yankee Lesson Plan R0-L-2.5 - Chemical and Volume Control System (CVCS), pg. 81 ,
This question has no correct answer. The question reference, Maine Yankee Systems Training Manual Chapter 28 - Emergency / Auxiliary Feedwater System is incorrect in stating that "MS-T-163 is an air to close valve."
The Maine Yankee Design Basis Summary Document for the Instrument Air ' System is correct in stating that valve MS-T-163 faih shut on loss of air. Therefore, the Reserve Air Receiver (TK-25) is not needed to provide backup air for closing MS-T-163 on a Safety Injection Signal or a Containment Isolation Signal. This question should be removed from consideration on this exam. Maine Yankee will revise the Systems Training Manual at the earliest possible opportunity.
Reference: Maine Yankee Design Basis Summary Document - Instrument Air Pages I-8 and I-9. (attached)
The correct answer for this question should be (d) vice (a). Since the shutoff head of a LPSI pump is 190 psig (Systems Training Manual Chapter 6 - Emergency Core Cooling), the LPS2 pumps will begin injection at a RCS pressure of approximately 205 psia (190 psig).
Reference: Maine Yankee Systems Training Manual - Chapter 6 - Emergency Core Cooling System, pg. 6-17
There are two correct answers to this question. As stated in the question' reference (Technical Specification 2.1, Limiting Safety System Setting - Reactor Protection System, pg 2.1-2) the symmetric offset trip is provided to prevent the fuel from exceeding the SAFDL on fuel centerline melt.
Also, the Maine Yankee Final Safety Analysis Report states'that the principle concern in the CEA withdrawal analysis is protection of the fuel from exceeding the SAFDLs on minimum DNBR and fuel centerline melt. The analysis report goes on to state that the variable overpower trip (V0PT) may be needed to protect the core from thermal damage depending on core life.
Based on this information, discussion with the Maine Yankee Reactor Engineering Section, and discussion with the Yankee Atomic Electric Company's Transient and Accident Analysis Group the Maine Yankee Training Department trains it's operators that the V0PT is necessary to protect the core from both fuel centerline melt and minimum DNBR SAFDLs (see reference - Maine Yankee Lesson Plan SRO-L-10.7 - Technical Specifications 1.1 - 2.3 - page 26). Both answers C and D should be accepted as correct.
Reference: Maine Yankee Lesson Plan SRO-L-10.7 - Technical Specifications lil - 2.3 - page 26
3. SENIOR' REACTOR OPERATOR .P gs 16' L- !
QUESTION: 027 (1.00)' ~
When transferring from Normal Charging to Alternate Charging, Letdown ^ chall be transferred from Normal to Alternate prior to securing Normal ! ~ Charging.
t Salect the heat exchanger which forms the basis for this requirement.
, a.
Regenerative Heat Exchanger b.
High Pressure Drain Cooler L c.
. Seal Water Heat Exchanger d.
. Letdown Heat Exchanger i , QUESTION: 028 (1.00) When performing a boration of the Reactor Coolant System, BA-A-80 is to be used only at the direction of the FSS/ SOS because: , any primary water remaining in the piping from previous a.
dilutions would be added resulting in a dilution.
b.
the boric acid addition rate would be to high and therefore hard to control.
. the relatively colder water would thermally shock the Reactor c.
Coolant Pump seals.
d.
any automatic actuations (normal or emergency) are bypassed by this flowpath.
i jQUESTION: s 029, J(1. 00) ,/ With Normal. Charging and Alternate Letdown in service, what action would b3 taken to' makeup ~to the Reactor Coolant System? a.
Reduce Afternate Letdown flow.
b.
Establish maximum blended makeup flowrate, c.
Place Alternate Charging in service.
. d.
Direct Alternate Letdown to the Quench Tank.
i ? '. i
'ELO OUTLINE OF INSTRUCTION ' INSTRUCTOR ACTIVITY REFERENCE c.
To get the Boric Acid /Watet addition to the charging pump suction,-water must be taken from VCT.
' 1) Done by unbalancing charging / alternate . letdown flowrates, , a) Increase charging flow Open CH-F-38
b) Decrease alternate letdown flow Close DR-A-6
OB Divert DR-A-10 to Quench Tank
2) Flow from VCT can be verified by: a) VC1 level decrease Constant Temperature
b) Pressurizer level increase Constant temperature -* l
LESSON R0-L-2.5 PAGE 81 0F 84 .- . . , --.; y-a . -- c-+ -, -, -
kSENI'ORREACTOROPERATOR Page 18 .. QUESTION: 032 (1.00) Given the following plant conditions: '
- 1 Steam Generator pressure = 350 psig level = 200" (WR)
-
- 2 Steam Generator pressure = 450 psig level = 30%
- '
- 3 Steam Generator pressure = 450 psig level = 40%
- Pressurizar pressure = 1700 psig - Containment pressure = 25 psia - Select the response of the Emergency Feedwater (EFW) System control and
isolation valves.
! a.
The EFW control and isolation valves for all 3 Steam Generators will be shut.
b.
The EFW control and isolation valves to #1 and #3 Steam Generators will be shut, and the EFW control and isolation valves to #2 Steam Generator will be open.
c.
The EFW control and isolation valves for all 3 Steam Generators ' will be open.
d.
The EFW control and isolation valves to #1 Steam Generator will l be shut, and the EFW control and isolation valves to #2 and #3 Steam Generators will be open.
QUESTION: 033 (l'. 0 0) The Reserve Air Receiver (TK-25) provides backup air to the: a.
Auxiliary Feed Pump Turbine Steam Supply Isolation Valve-(MS-T-163) for closing on a Safety Injection Actuation Signal or a Containment Isolation Signal.
b.
Emergency Feedwater Isolation Valves (AFW-A-338, 339, 340) for- ' closing on a low Steam Generator pressure of 400 psig.
c.
Auxiliary Feed Pump Turbine Steam Supply Pressure Control Valve (MS-P-168) for up to 12 hours of continued feed pump steam pressure regulation.
' d.
Emergency Feedwater Control Valves (EFW-A-101, 201, 301) for closing on a low Steam Generator pressure of 400 psig.
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Water Treatment Service Air System , The Water Treatment Service Air System is a separate self- ' contained compressed air system installed to support the water c treatment facility.
Piping cross-ties were installed for emergency back-up to plant service instrument air. The water l- ' treatment compressor (C-10) was sized to be approximately ! , equivalent to a plant service air compressor (C-1A B. C) in i > capacity (300 SCFM).
The compressor (C-10) is an air cooled unit capable of automatically loading / unloading according to the' periodic demands of the water treatment plant. An air receiver (TK-141) was installed with a capacity of 32 cubic feet in order to prevent the compressor from short-cycling. Filters (FL-106 and 107) were installed downstream of the air receiver to ensure that the compressed service air is clean, dry, and relatively oil free.
The Water Treatment Service Air System loads consist primarily: -200 SCFM: TK-70A and B Mixed Bed Demineralizer-50 SCFM: TK-40 Waste Regenerate Neutralization Tank-30 SCFM: P-185A, B. C: P-186A. B: P-187A. B: P-188A B Caustic and Acid Pumps These loads do not occur concurrently. They occur'during the various steps of the water treatment regeneration cycle.
_.
The use of Compressor C-1D to supply the Instrument Air System, is restricted to emergency situations only. The compressor is not powered from an emergency bus. A check valve (SA-410) was installed in the cross-tie piping to prevent depressur_ization of the instrument air header should an upstream failure occur.
C.
ACCIDENT MITIGATION FUNCTION , The function of the instrument air system is not required for a safe plant shutdown or for the operation of the engineered safeguards.
i Air-operated valves that are required to function during accident conditions will fail to the accident mode on loss of instrument air.
Air-operated valves which are required to be operable during an accident ' have separate air supplies. The following is an outline of these air accumulators and their design bases:
Accumul a tor YTK-25 gMS-T-163T(fa W closed)P_T Auxiliary Feed Pump Steam' Isolation MS-P-168 (fail open) > . Auxiliary' Feed Pump Steam Supply . R69\\19 I-8 , e ?
q c i i DESIGN FUNCTION: To supply air for 4 hours to hold MS-T-163 open and to control MS-P-168.
The auxiliary feedwater pump must be capable of operating for at least 4 hours. (Reference C81) There is also a temporary hose connection to the Emergency Feedwater Control Valves EFW-A-101 201. 301. This provides the capability to operate these valves if instrument air is lost and the EFCV compressors are available.
(References 02 and H3) TK-83A,B.C.D VP-A-1,2 (Fail-as-is) Containment Purge Air Supply VP-A 3,5 (Fail-as-is) Containment Purge Air Supply DESIGN FUNCTION: To close one time to provide containment isolation.
TK-110 SCC-A-460.461 (Fail open) SCCW trip valves DESIGN FUNCTION: To close valves ore time and hold the valves closed for 24 hours.
The SCCW trip valves must be closed to isolate the non-essential loads. The air supply must maintain the valves closed for 24 hours.to give the operators sufficient time to manu61y block the valves closed with their handwheels.
(Reference C75) TK-111 EFW-A-101.201,301 (Fail open) Emergency Feedwater Control Valves j DESIGN FUNCTION: To close valves one time and ' hold the valves closed for one hour.
l These valves are required to clo:.e as part of ' the Main Steam Line Break Mitigation System.
This action is assumed in the FSAR Chapter 14 Safety Analysis. The valves must be maintained closed for one hour to give the operators sufficient time to close the manual isolation valves.
(Reference C48) TK-123 EFW-A-338.339.340 (Fail open) Emergency Feedwater Isolation Valves R69\\19 I-9 l
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. QUESTION: 039- _ (1. 0 0) - Solect the condition which would prevent the.-Primary' Inventory Trend .Syctem (PITS) reading from becoming unreliable due to' reference leg fleshing.. Containment pressure less than 5 psig.
a.
b.
Reactor Coolant System pressure greater than atmospheric.
c.
Reactor Coolant Pumps secured.
d.
Containment Ventilation System operating.
QUESTION: 040 (1.00) , Select the condition which would prevent a Reactor Coolant System (RCS) -l Loop Cold Leg Isolation Valve from OPENING.
i a.
Reactor Coolant Pump operating.
b.
RCS Loop Bypass Isolation Valve closed.
c.
RCS Loop Hot Leg Isolation ~ Valve closed.
d.
Delta T between loop Tcold's greater than 35 degrees F.
' QUESTION: 041 (1.00) . Following a Loss of Coolant Accident (LOCA), RCS pressure'is at_200-psia and dropping.
j Which ONE of the following statements describes the expected status of
High Pressure Safety Injection (HPSI), Low Pressure Safety Injection (LPSI), and Safety Injection Tank (SIT) flow? } a.
HPSI and SIT flow exist with no LPSI flow.
b.
HPSI and LPSI flow exist with no SIT flow.
c.
HPSI flow exists with no SIT or LPSI flow.
, d.
HPSI, SIT, and LPSI flow exist.
, . i l > _ _
p -
r Emergency Core Cooling System Section 2.0 x Major Components a i i 2.1.3 Associated Piping and Valves During normal plant operation, the ! LPSI pumps are aligned to take
Piping used throughout the HPSI suction from the RWST. The pumps - . ! System is made of stainless steel for - start automatically on SIAS and the-corrosion resistance. The piping is pump discharge is aligned through arrangedin parallel trains for motor operated valves to the reactor reliability. Cross-connection lines are coolant loop safety injection lines.
provided in the HPSI pump suction The pumps trip on RAS, to ensure a and discharge headers. Isolation NPSH for the Containment Spray valves on these lines are normally pumps while suction is being aligned closed to maintain separation of to the Safeguards sump.
> trains. Motor operated valves in the , HPSI pump discharge lines may be The LPSI pump discharge is routed to ! opened from the Control Room if the same 10-inch safety injection- - necessary. Major valves associated lines to the reactor coolant loops that -
with the HPSI System are listed in the HPSI pumps supply. The shutoff Table 6-1 and shown in Figure 6-5.
head of the LPSI pumps is 190 psig.
' No wateris delivered by the LPSI 2.2 Low Pressure Safety Iniection pumps to the RCS until system System (LPSI) - pressure decreases below 190 psig.
. The capacity of the pumps is 5250 The components of the LPSI System, gpm each, at 100 psig discharge shown in Figure 6-6, are: pressure.
a. LPSIpumps (2) The LPSI pumps are vertical,2-stage, centrifugal pumps, located in the b. SI tanks (3) Spray Building (refer to Figure 6-7).
They are each driven by a 400
c. Associated piping and valves horsepower,4160 volt motor through' a flanged coupling. Pump minimum-recirculation is provided through a
recirculation header to the RWST.
2.2.1 LPSIPumps The recirculation header also serves
the Containment Spray pumps.
' The LPSI pumps provide low head, Primary Component Cooling water is , high volume safety injection flow to supplied for gland cooh,ng of P-12A . the RCS from the RWST following and Secondary Component Cooling i SIAS. The LPSIpumps also provide the head for Residual Heat Removal supphes gland coolmg for P-12B.
- flow when the reactor is in the cold I
shutdown configuration.
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C..; ~~ h ;[SENIOft REACTOR OPERATOR ' ' Pag'e'24-i '; . ~ QUESTION: 048 - (1. 0 0 ). -Select the Reactor Protection System trip which is intended to protect > the core from exceeding the Specified Acceptable Fuel Design Limit on fuel centerline melt.
, a.
Low Reactor Coolant Flow b.
Thermal Margi:./ Low Pressure ' c.
Variable Nuclear Overpower d.
Symmetric Axial Flux Offset . T QUESTION: 049 (1.00) Given the following Reactor Protection System (RPS) conditions:
RPS Channel B High Containment Pressure in Channel Bypass.
- RPS Channel C High Containment Pressure in Trip.
- If RPS Channel A High Pressurizer Pressure were to trip, the results on the plant would be that the: , plant would trip since the logic was 1 out of 2 prior to the a.
RPS High Pressurizer Pressure trip channel tripping.
b.
In order to trip the plant, additiona_ channels required to trip would be 2 out of'2 for High Containment Pressure and-2 out of 3 for High Pressurizer Precuure.
plant would trip since there are now two channels in a-tripped c.
condition.
' '. t-d.
In order to trip the plant, additional channels required to trip , would be 1 out of 2-for High Containment Pressure and 1 out of 3 for High Pressurizer Pressure.
T L $ i L AI-
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w.
- . ELO OUTLINE OF INSTRUCTION INSTRUCTOR ACTIVITY REFERENCE.
. . SN0-303-1 B.
THE'RPS trip setting limits and bypass are as SN0-303-2 follows: SNO-303-3 Review all LSSS'S per TP-TS-3.
TP-TS-3
1.
Core Protection ' a '. Variable Nuclear Overpower 1) Trip setpoint Less than or equal to Q + 10, or Q.
State the basis for each
106.5 (whichever is smaller) for Q LSSS.
greater than or equal to 10 and less than or equal to 100.
Less than or equal to 20 for Q less Q.
Which LSSS's may be
than or equal to 10.
bypassed? At which condition may they be bypassed? Q = percent thermal or NI power,
whichever is greater.
2) Discuss fact that V0P calculator bottoms out at 10% power, therefore Q(10%) + 10% - 20% trip setpoint for Q 1ess than or equal to '10.
3) No bypass setpoint . 4) Basis - V0P prevents damage to the fuel cladding from reactivity transients too rapid to result in a high pressure or TM/LP trip.
Safety limit concern - T.,,, and DNB.
e' LESSON SRO-L-10.7'PAGE 26 0F 33 . - - - - - - . . ...-. . -. .-. . .-. .. . - .
, . . , , ..
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-10 MYAPC ' A Once the shutdown groups have been withdrawn, the control' groups'are withdrawn'as necessary to achieve criticality and to meet power demand.
i Control groups are withdrawn in a programmed sequence as discussed in Se ction 7. 4.1.3.
I I Reactivity addition by withdrawal of CEA groups is dependent on the initial position of the groups prior to the withdrawal and on the integral.
worth of these groups. The regulating groups are withdrawn in a specified
' sequence having.40% group overlap, and the position of the groups under-steady-state conditions is a function of power level.
The principal concern of the CEA withdrawal analysis, because of the ' increasing power during the event, is the protection of the fuel from exceeding the Specified Acceptable Fuel Design Limits (SAFDLs) for minimum 'f DNBR and fuel. centerline melt.
Primary system overpressurization is also
of concern in the - analysis. , t Protection of the core from thermal damage and/or the reactor coolant-system (RCS) from overpressurization for this incident may be provided by , ' one or more of thelfollowing reactor trip signals at various times in core-(,,d life and for various rod worths and reactivity addition rates : the high' rate of change of power trip, high power level trip, variable overpower - , , trip (VOPT), high pressurizer pressure trip,. thermal margin / low pressure (TM/LP) trip, and low steam genetator water level trip.
. -l In addition to' the high'pressutizer pressure trip at 2400 ps a, protection r ) against overpressurization of the RCS is provided by pressurizer power-operated relief valves (POKVs ), set to become operative at 2400 psia, and by ASME code design safety valves set at 2500 psia.
14.2.2 Method of Analysis The CEA withdrawal incident was analyzed using the GEMINI-II digital - computer model (References 1 and 2) simulation of the Maine Yankee NSSS to l ' determine the temperature, pressure, power, and core average heat flux responses.
This simulation includes a point kinetics core model with~ moderator and fuel reactivity feedback, reactivity control and shutdown , 14-17 Rev. 1, 4,
.
( - %- , l ! l t .
re ? ATTACHMENT:4 L , MaineYankee
. REllABLE E LECTRtCITY FoR MMNE LINCE 19] i , PO BOX 408. WISCASSET, MAINE 04578 (207) 882-6321 i November 18, 1993 MN-93-103 Mr. Lee Bettenhausen Operations Branch, DRS UNITED STATES NUCLEAR REGULATORY COMMISSION 475 Allendale E3ad
-King of Prussia, PA 19406 . Reference: (a) License No. DPR-36 (Docket No. 50-309)
Dear Mr. Bettenhausen,
This report concerning plant simulator operation during the operator licensing examinations conducted on November 15 and November 17, 1993 is being provided as requested by Mr. Kerry Ihnen, Chief Examiner.
, During the licensed operator requalification simulator re-examination and initial licensing simulator examinations conducted on November 15 and 17, 1993, the Maine Yankee plant simulator performed as expected with the following exceptions; o During the requalification examination, a hardware related problem occurred which caused several incorrect digital indications on the Electrical Control Board. Both the Maine Yankee and NRC examiners agreed that the operators being examined were not adversely affected by the problem. Troubleshooting conducted after the scenario was not conclusive although one 5 volt power supply voltage was found low and was subsequently readjusted.
, o During the initial licensing examinations conducted on November 17, 1993, a hardware related problem occurred which_ caused oscillation of several analog indications on the Electrical Control Board. Again, the Maine' Yankee Training Department and the NRC Chief Examiner agreed that the candidates being examined were not adversely affected by the problem. Troubleshooting conducted after the scenarios were complete revealed z. bad circuit board in the Distributed I/O system.
Once this component was replaced, the problem did not recur during the remainder of the examination process.
> ' o During the conduct of a Job Performance Measure involving manual transfer to Recirculation Cooling, the High Pressure Safety Injection Pumps (HPSI) began to cavitate at a low Refueling Water Storage Tank level The level at which HPSI pump cavitation would be expected is currently under investigation.
I s A review of our simulator work order system indicates an extremely low frequency of ' these type problems. Although certainly untimely, Maine Yankee considers these events to be isolated incidents with no trend requiring further investigation.
Sincerely, i l MA ** IVgE lb ' obert W. Bacm5r ' Plant Manager c: Kerry Ihnen, Chief Examiner
P' ATTACHMENT 5
MaineYankee - ' OELIABLE EttCTRICITY FOR MAINE SINCEM - ' Po. Box 408. WIScASSET MAINE 04578. (207) 882 6321 November 18, 1993 MN-93-107 Mr. Lee Bettenhausen > Operations Branch, DRS UNITED STATES NUCLEAR REGULATORY COMMISSION 475 Allendale Road King of Prussia, PA 19406 Reference: (a) License No. DPR-36 (Docket No. 50-309)
Dear Mr. Bettenhausen:
This letter is forwarded per HUREG 1021 - Operator Licensing Examiner Standards - ES-601 to inform you that Maine Yankee examiners have passed the Operations Crew D licensees which received a requalification simulator re-examination on November 15, 1993.
Sincerely, MAINE Y KEE - / / Ro ert a Plant Manager c: Kerry Ihnen, Chief Examiner t }}