IR 05000029/1990006
| ML20058L483 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/24/1990 |
| From: | Eselgroth P, Prell J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20058L480 | List: |
| References | |
| 50-029-90-06OL, 50-29-90-6OL, NUDOCS 9008070251 | |
| Download: ML20058L483 (127) | |
Text
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION REGION 1 OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 50-029/90-06 (OL) FACILITY DOCKET NO. 50-029 FACILITY LICENSE NO. DPR-3 LICENSEE: Yankee Atomic Electric Company FACILITY: Yankee Nuclear Power Station EXAMINATION DATES: June 19 - 21, 1990 t NRC EXAMINERS: J. Prell, Chief Examiner J. D' Antonio, Operations Engineer t.. CHIEF EXAMINER: 7ffd T (. Pre , Senior Operations Engineer Date h'_ _ _ //ar/90 APPROVED BY: l frP.W.Eselgroth, Chief,PWRSection ~ 0a'te SUMMARY: Written and operating examinations were administered to two Senior Reactor Operator (SRO)-Upgrade and one SRO-Instant candidates. All_three candidates passed their examinations and were-issued licenses.
0008070251 900725 PDR ADOCK 05000029 V PIIC
! DETAILS , TYPE OF EXAMINATION: Initial 1.
EXAMINATION RESULTS: ........................................... l INITIAL SRO l TOTAL l l EXAM Pass / Fail l Pass / Fail l ........................................... E E E E E WRITTEN l 3/0
3/0 l E E E E E I E E l' E SIMULATOR l NA l NA B ' E E E E E $ E E l WALK THROUGH E 3/0 l 3/0 l E E E E E E E E E OVERALL E 3/0 l 3/0 E E E E_ E
- Facility did not have a site specific control room simulator
, , 2.
PERSONS CONTACTED: Yankee Nuclear Power Station Personnel , i C. Russell Clark Manager, Training a
- T 9.tenderson Acting Plant Superintendent i
- K.r1 Jurentkuff Plant Operations Ma.iager
- Russell A. Mellor Technical Ofrector
- ,
- Norm St. Laurent Acting Manager of Operations
, David H. White Operations Training Supervisor
U. S. Nuclear Regulatory Commission !
- Tom Koshy Senior Resident Inspector L
Denotes those present at the exit meeting conducted on June 21, 1990.
- 3.
GENERIC WEAKNESSES: The following is a summary of generic weaknesses noted on the written examinations.
For purposes of this report, ceneric weaknesses are defined based on those questions, or portions of qurstions, which were missed by , - .. . m
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l t all three candidates.
This information is provided to aid the licensee in upgrading licensed operator training. No licensee response is required.
a.
Question number 36 asked the candidates to choose from a list of four different conditions the one which would trip.a battery charger AC.
j supply breaker.
Two candidates identified c.,"105v de critical.under l voltage or 150 v de critical over voltage" and the third candidate identified b., "105v de critical under voltage or 140v de over volt-l age." The correct answer was a., "100v de under voltage or 140v de over voltage."
' b.
Question' number 52 asked the candidates to list four of the five parameters identified in OP-3054, " Natural Circulation," which are
used to verify that natural circulation is taking place. All three i candidates listed as one of their four responses "MCS delta T less ' than full power delta T (< 42'F)."
This response was not listed in i either OP-3054 or ES-0.1 Attachment 2 as a parameter which can be , used to. verify natural circulation.
~ .c.
Question number 61 asked the candidates to identify two conditions . , 'under which the reactor must be manually scrammed following a dropped ! control rot incident at 100% power. All three candidates failed to , identify the condition where Tave would stabilize below the minimum
temperature for-criticality.
j 4.
SIMULATION FACILITY FIDELITY REPORT The licensee is preparing to install a site specific simulator in their new training center in July, 1990. Training of operators with the simu- ' lator will begin some time in October.
l 5.
STATUS REPORT ON OPEN ITEMS IDENTIFIED IN INSPECTION REPORT 50-029/89-80 t In November, 1989, the NRC conducted an inspection of the licensee's > Emergency Operating Procedures (EOPs) used by the operators at the Yankee
Rowe Power Station.
The inspection team identified problems with the, technical adequacy and usefulness of the Off-Normal Operating Procedures - (OP-3000 through OP-3259 series) which required a great deal of reliance
on operator memory for component location and equipment operation.
In [ L response to this inspection finding, the licensee committed to a three phase schedule for_ improving the quality of their operating procedures.
r (See letter to Mr. Robert M. Gallo, Chief Operations Branch, Division of l Reactor Safety from Mr. N.'N. St Laurent, Acting Manager of Operations, , Yankee Atomic Electric Company dated January 5,1990.) Prior to commenc- , ing their 1990 refueling outage the licensee _ committed to completing phases _one and two of the schedule.
Phase one was the revision of all l-emergency and abnormal operating procedures for technical adequacy and L operator usability.
Phase two was the revision of all normal operating l procedures including plant cooldown, plant heat-up and power operations.
_; i- . Phase three will be completed during the next operating cycle and will ' i ! , -
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- ~-~ . ( . !
! ! include a major review of all operating procedures for technical adequacy, ' . format, proper indexing and generation of a procedure generation package.
- The licensee briefed the NRC on the status of their schedule for improving the quality of their operating procedures. - Per the licensee, phases or:e and two were completed and phase three will be initiated following this . t refueling outage.
There were no major problems noted with the emergency, . abnormal or normal operating procedures used during the preparation and l , administration of the initial examination.
- .
6.
SUMMARY.0F NRC COMMENTS MADE AT EXIT INTERVIEW:
A summary of the week's activities was presented and discussed, including the item presented above. There were no weaknesses identified with the training department or operations department.
Prior to the exit meeting .i the licensee submitted his formal comments regarding the written examina-tion (See Attachment 2). All comments were reviewed and then accepted by.
~' _the NRC.
7.
SUMMARY.0F FACILITY COMMENTS AT THE EXIT INTERVIEW: i The facility expressed their. appreciation for having the opportunity to ! review the written examinations prior to their being administered.
Attach'ments: " 1.
Written Examination and Answer Key (SRO) 2.
Facility Comments on Written Examination
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, , u ATTACHMENT NO. 1 MASTER EXAM AND ANSWER KEY . Cb l .. , p . l j
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NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION REGION
FACILITY: Yankee-Rowe _ REACTOR TYPE: PWR-WEC4 < DATE ADMINISTERED: 90/06/18 , < .4 CANDIDATE: INSTRUCTIONS TO CANDIDATE: Points for each question are indicated in parentheses after the question.
To pncs this examination, you inust achieve an overall grade of at least 80%. EXCmination papers will be picked up fou~r nd One half (4 4ftt-hours after tho examination starts.
J I
I I I l NUMBER l TOTAL l CANDIDATE'S l C:.dDIDATE'S l , ' PC.NTS l POINTS l OVERALL-l l QUESTIONS l T l l l l GRADE (%) l l_____________l__________l_______________ _______________l l l l l l > , l.
l 100.00 l l l ! I I I I I I I I I I l l All work done on this examination is my own.
I have neither given nor received aid.
, Candidate's Signature ,
~ NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application-cnd 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 rcceived or given assistance in completing the examination.
This must be done after you complete the examination.
3. R stroom trips are to be limited and only one candidate 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. U00 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.
6. Fill in the date on the cover sheet of the examination (if necessary).
7. You may write your answers on the examination question page or on a ceparate sheet of paper.
USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON ,
THE BACK SIDE OF THE PAGE.
8.
If you write your answers on the examination question page and you need more space to answer a specific question, use a separate sheet of the p;per provided and insert it directly after the specific question.
DO NOT l WRITE ON THE-DACK SIDE OF THE EXAMINATION QUESTION PAGE.
, , 9.
Print your name-in the upper right-hand corner of the first page of answer cheets whether you use the examination question pages or separate sheets of paper.
Initial each of the following answer pages.
,
10. B3 fore you turn in your examination, consecutively number Nach answer ! cheet, including any additionel pages inserted when writing your answers on the examination question page.
l 11. If you are using separate sheets, number each answer and skip at least 3 lines between answers to allow space for grading.
! l 12, Write "Last Page" on the last answer sheet.
l l 13. Une abbreviations only if they are commonly used in facility literature.
Avoid using symbols such as < or > signs to avoid a simple transposition
orror resulting in an incorrect answer.
Write it out.
i ! .
t 14. The point value for each question is indicated in parentheses after the question.
The amount of blank space on an examination question page is NOT an indication of the depth of answer required.
10. Show all calculations, methods, or assumptions used to obtain an answer.
16. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
NOTE: partial credit will NOT be given on multiple choice questions.
17. 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 0.20 points.
If one of your five responses is incorrect, 0.20 will be deducted and your tctal credit for that question will be 0.80 instead of 1.00 even though you got the four correct answers.
18. If the intent of a question is unclear, ask questions of the examiner only.
19. When turning in your examination, assemble the completed examination with cxamination questions, examination aids and answer sheets.
In addition, turn in all scrap paper.
20. To pass the examination, you must achieve an overall grade of 80% or greater.
21. There is a time limit of (4 1/2) hours for completion of the examination.
(or some other time if less than the full examination is taken.)
22. When you are done and have turned in your examination, leave the examin-ation area as defined by the examiner.
If you are found in this area while the examination is still in progress, your license may be denied or revoke F , SENIOR REACTOR OPERATOR Page
QUESTION: 001 (1.00) Which of the following best describes the main coolant loop check valves? G)-They are installed in the hot legs, and are designed to remain 10 degrees open to allow 100 gpm backflow in an idle loop.
b) They are installed in the cold legs, and have a hole drilled through the disc to allow 100 gpm backflow in an idle loop.
c) They are installed in the hot legs, and have a hole drilled through the disc to allow 100 gpm backflow through an idle loop.
d) They are installed in the cold legs, and have a slotted seat to allow 100 gpm backflow through an idle loop.
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i QUESTION: 002.(1.00) , l ' ~Wh t induces normal pressurizer spray flow? c) Thermal driving head resulting from the density difference between'. hot leg l end cold leg water.
-i b) Loop.2 MC pump differential pressure and velocity head.
! c) Reactor Vessel differential pressure and loop 2 velocity head.. .d) Backpressure from loop 2 MC check valve.
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i i I QUESTION: 003 (2.00) A LOCA is in progress. Complete the following statements concerning the Safety I ' Injection system.
c) A SIAS signal is generated when main coolant pressure decreases to psig or vapor container pressure increases to.
psig.
b) The.HP safety injection pumps will start to deliver water to the I-!c system I at about psig Mc pressure.
I c) The LP safety injection pumps will start to deliver water to the Mc system at about psig Mc pressure.
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l QUESTION: 004 (1.00) Th3 Control Room Operators request your assistance in diagnosing a transient.
, Ycu observe the following on recorders and indicators: - Wide range pressurizer level has rapidly increased to 130 inches, but is now-trending down.
- Narrow range pressurizer. level is 115 inches, trending down.
, - MC system pressure is 1950 psig, slowly trending down.
- The running charging pump is starting to increase speed.
' Which of the following is the most likely cause? c) Leak on wide range sensing leg , b) Leak on wide range reference leg c) Leak on narrow range sensing leg d) Leak on narrow range reference leg , ! l l l l I l
%. SENIOR REACTOR OPERATOR Page
. QUESTION: 005 (1.00) .A spurious reactor trip occurs at 100% power.
The 20RS reactor scram auxiliary relay fails to operate.
Which of the following best describes plant response, a) Both scram breakers will open, but the turbine will not trip.
b) BK-1 will not open; 20RSA will open BK-2 and trip the turbine.
c) 20RSA will trip the turbine after either one or both of the scram breakers open.
d) Scram breaker BK-2 will not open; BK-1 will open and initiate a turbine trip through 20RSA.
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QUESTION:1006 (1.00) -Tho Supervisory Control Room Operator is permitted to take' reasonable t..tions
that depart from license conditions or Technical Specifications when such- ' cetions.are required to protect public health and. safety and no action consistent with license conditions and technical specifications are adequate, provided that: c). The approval of the Shift Supervisor'is obtained.
b)"The-approval of the Plant Superintendent is obtained.
! c) The approval of the Operations Manager is obtained.
- d) The-NRC is notified within 1 hour.
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. ..... . , SENIOR REACTOR OPERATO Page_10 , , ' QUESTIONi' 007' (1.50) ' - ' Three' conditions-which will automatically transfer the rotating exciter to ~ .? manual voltage control WITHOUT causing a plant trip are , and . ,
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1.00) . ( QUESTION: 008 , ' gjigff j'/t oc 2 9'tJ 'A mixture of hydrogen and air is considered explosive between % to - % hydrogen concentration'by volume.
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p; SENIOR REACTOR OPERATOR Page 12 l ' QUESTION: 009-(1.00) Which of the following BEST describes the reason why explosive hydrogen gas is Luced'for generator cooling rather than an inert gas.
G) Hydrogen is the least expensive. cooling medium with the necessary ' thermodynamic properties.
b) The generator casing is designed to contain the maximum hydrogen-air l explosion considered credible.
c) Hydrogen gas has the Jost desiraable properties as a cooling medium.
I d) Other available cooling gasses may-cause excessive insulation erosion due to their more irregular molecular structure.
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L E' SENIOR REACTOR OPERATOR Page 13 I QUESTIONi'010- (1.00)- I i ~ -A running' service water pump trips.
Another pump will start'if header
pressure decreases to psig if.its control switch is .. > I i .f '
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- - - - -. -. . _ _ _ _ _ _. SENIOR REACTOR OPERATOR, Page 14 . QUESTION: 011 (2.00)
- . Technical Specification 3.7.2 " Steam Generator.Pressur3/ Temperature-Limitation" states four Liniting Conditions for Operation.
a) List any THREE ' of -these limits (1.5 pts). b) State the time allowed for the specified corrective action if any of the LCO requirements are not met (.5 pts).
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........ SENIOR REACTOR OPERATOR Page 15 ' QUESTION: 012 (2.50) -Q) The Main Coolant System Pressure Safety Limit-is psig (.5 ). "b) State the required action for each Operational Mode if.this safety Limit is . exceeded (1.0).
c)-State the actions required within 1 hour for any Safety Limit violation in accordance with the Technical Specifications section 6.
(1.0)
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Page 16 i . QUESTION: 013 (2.00) l Mmtch'the electrical loads in-Column A wit'h the power sources in Column B.
Answers may be used more than once or not at all.
Column A Column-B
- G) C-1 service air compressor 1) 2400v section 1 b)LP-35-1 fire pump 2) 2400v section 2 c)?MCC-2 normal feed 3) 2400v section 3 i-
' d)- P-23 LPST. cooling pump 4) 480v BS 4-1 L
- o)' P-1-3 boiler feed' pump 5) 480v BS 5-2 l
f) P-79-1 emergency feed pump 6) 480v BS 6-3
9; deatic exciter 7) 480v emergency bus 1 ? L h):MCC-2 omergency feed 8) 480v emergency bus 2 9) 480v emergency bus 3 j c) r l b)
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p.: SENIOR' REACTOR OPERATOR Page 17 . QUESTION:=014-(l'. 0 0 ) ! l ' -Thi plant is operating normally at 100 % when a fault occurs on the #1480V cm;rgency-bus.. Which of the following is the MOST LIMITING,of the applicable = T0chnical Specifications.
G) 3~.8.2.1 "AC DISTRIBUTION-OPERATING" .;
b)-_3.8.2.3 "DC DISTRIBUTION-OPERATING" j i c) 3.6.2 " CONTAINMENT ISOLATION VALVES" it) 3.'4.3 " SAFETY VALVES-OPERATING" < l l ! !
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, _ ! i LSENIOR' REACTOR OPERATOR Page'18 . QUESTION:~015 (1.50) . ? =-Lict the THREE valves that will i: lose automatically to terminate or prevent a - ralease if their associated proc.nss monitor. fails.
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SENIOR REACTOR' OPERATOR Page 19- . . QUESTION: 016 (2.00) Plant Conditions: - 100% power
1 - 532-deg T-ave '- all rods.out . equilibrium xenon - core.burnup':-4,600 GWD/MTU One-of-the low setpoint secondary safety valves fails open.
Assuming no operator action, answer the following.
Data Reference Manual figures are o provided; show your work for partial credit.
_ - a) safety valve setpoint .psig b)-safety valve capacity- % (or lbm/hr) - c) final MCS T-ave deg F d) final reactor power % i . ' a . b $
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p.- , , , t ' SENIOR REACTOR OPERATOR Page 20-. - , , , QUESTION: 017 (1.00)- ! -During'an accident, the. Shift Supervisor may leave'the Control Roorn at his ow6-diccretion, without a qualified-relief, under what.two conditions? ,
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c:- i , i; SENIOR REACTOR OPERATOR Page 21 l - QUESTION: 018 -(2.00)
AP-2001 " Responsibilities and Authorities of Operations Department Personnel"? includes approximately'l 1/2 pages on the subject of communicationsi List any FOUR communications practices either recommended or required by this ~
. procedure.
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- - _ _ _. .. SENIOR REACTOR OPERATOR Page 22 . QUESTION: 019 (1.50) Before performing any procedural step, there are three actions, checks or verifications an operator must perform. These are , , and .- y-;, i' -
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-l-SENIOR REACTOR OPERATOR' Page_23' , l , . ., I QUESTION: 020' (1.50)- ) c). Indicate on the attached diagram the areas considered to be' Control Room Zones 1,2 and 3 (.75 pts).
' .b)' BRIEFLY EXPLAIN any REQUIREMENTS or RESTRICTIONS on occupancy of these , zones in modes 1-4 (.75 pts).
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- CENIOR REACTOR, OPERATOR Page 24
' QUESTION: 021.
(l'. 00 ) Which of the'following personnel is authorized to perform an independent .vsrification of switching and tagging of equipment _ requiring Quality Assurance?' - a).Any' individual listed on OP-Memo-2EEE-3 " Local' Control Tagging' List."
.b) Any. person qualified to perform the initial-switching and tagging.. , Lc) The person requesting the switching and tagging.
d):An RO or SRO licensed person listed on OP-Memo-2EEE-3.
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SENIOR REACTOR OPERATOR Page_25 QUESTION: 022 (l'. 0 0) When performing independent verifications of component positions, physical ccparation of the initial person and the verifier is usually recommended.to .cyoid errors.
However, there is one type of component for_which it is ) prsferable'to have the verifier observe the actions of the initial positioner ! rather than performing a second-manipulation of the component.
These components are: . ! ! ! ! !
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QUESTION: 023' (1.50) ' Provide the followint] plant administrative exposure _ limits from'AP-0803.
-CExternal Radiation Exposure Control."
Whole-Body (with NRC form 4).
(.75), Whole' Body (without-NRC 4) _ (. 25)- Extremity (.25) Skin (.25) ,
, SENIOR REACTOR OPERATOR .Page 27-QUESTION: 024 ( 1. 50 ). - i - l - AP-0804 " Internal. Radiation Exposure-Control" states that there are three, main:. j p;thways by which radioactive material can enter the body.
These are: l
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- QUESTION
- 025 (2.00)
G) The minimum Fire Brigade consists of a Brigade Leader and how many additional brigade' members? (.5) ' b) Who'is the Brigade' Leader? (.5)- s c) Who are the Brigade Members? (1.0)
1
' , , SENIOR REACTOR OPERATOR-Page 29 L QUESTION: 026 (1.50) ( (Thi THREE conditions which will automatically close the Lleed line isolation --valve'CH-LCV-222, and their setpoints, are: c).
b) . C) r.
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' SENIOR REACTOR OPERATOR Page 31 . QUESTION: 028 (1.00) The Source Lange detectors'are located-near the bottom half of the coro.
Why.
cro they in that position rather than at core centerline? 'a) The thimble tubes are angled in toward the core, so this location places the detectors closer to the flux they are to detect.
b). Neutron flux is greater in the bottom half of the core during a startup.
c) To allow a spare detector to be installed in the upper half of the thimble should the' installed detector fail.
d) This position provides optimal cooling for the detector by natural circulation of air in the thimbl w.
> c. SENIOR' REACTOR OPERATOR Page 32~ , ' , QUESTION: 029 (1.00) o {During a reactor startup, you note that with both source range channels.at b x ,10^4.' cps the' intermediate range channel 3 indicates 5 x 10^-10' amps and chnnnel 4 indicates.5 x'10^-11 amps.
Which of the following is the most-probable cause'of this discrepancy? a a) IR channel 3 is overcompensated = ! b)' IR channel 3 is undercompensated - 'I c)-IR channel 4 is undercompensated ' l d)-IR channel'4 is overcompensated .Il i , , i' - j ,' ! '
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. .- SENIOR' REACTOR OPERATOR' Page 33 --QUESTION: 030 (1.00)- OP-2113 " Operation of the Reactor Control. Rod System" lists as a prerequisite
- a ninimum Main-Coolant pressure.
This pressure is:- ' a -! n) 50 psig-- 'b).:200 psig l ' c)f325 psig-d)'500'psig - --
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. . ' QUESTION: 031 (2.00) . i OP-2113 " Operation of the Reactor Rod Control System" lists ten precautions.
Lict any FOUR of these precautions.
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[ '! l! QUESTION: 032 (1.50).
i i_ .i l Lict'THREE of the four bases.for the Technical Specification limits on t, s
inovcable control rods position and. alignment.
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u i ! l 5-QUESTIONtf033 (2.00) On the attached diagram, IDENTITY the following components.
If there is more ! i -- th n one of the component, you only need to identify one.
(.25 each) l f
- O)lheating and cooling fans
! ,if-b) post accident recirc fans . 19) heating coils for heating and cooling fans , '^~ d) cooling coils'for heating and cooling fans ! .o) gravity dampers , f); ring duct " g) purge supply duct h) gate valves to shield tank cavity
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, , j QUESTION: 034: (1.50) , ,
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"!4 l L ~ QUESTION: 035 (1.00) l
I LHOW LONG can the station batteries supply their safety loads-in.the event of a f Ic23 of all AC power? id "c) 30 minutes-j l < b)'l hour.
] . , /c) 2 h'ours
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- QUESTION: 036 (1.00)
i. -- Which of the following conditions trip a battery charger AC supply breaker? c) 100v de undervoltage or 140v du overvoltage " b)..105v'dc critical undervoltage or 140v de overvoltage -: l1. c) 105v de critical undervoltage or 150v de critical overvoltage d) 100v dc undervoltage or 150v de critical overvoltage .g - i -
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. , li SENIOR REACTOR OPERATOR' Page 40 l- . LQUESTION: 037--. (1. 00) - 1g A 1%' excess of steam demand greater than reactor power will cause.T-ave to d;;rease at approximately dog f/ minute.
Pressurizer level will d: crease at approximately-inches /r.inute.
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SENIOR REACTOR OPERATOR Page 41 l t ! i QUESTION: 038 (1.00) Th3 power requirement for a Main Coolant Pump with the plant HOT is about , '1380 kw; with the plant COLD it is about 1600 kw.
Which of the following best 'drcribes the reason for this difference?
i O) The increased flowrate at low temperatures creates greater headloss.
-b).The greater mass flowrate at low temperature requires more power.
.c) The mechanical efficiency of the pump impeller is reduced at low temperature.
. d).The lower-viscosity of hot water reduces pumping power needed.
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_ SENIOR REACTOR OPERATOR Page 42 ' i QUESTION: 039 (1.00) Th3 plant ir operating at 100% power.
A leak occurs in the sensing line for MS-PT-1000 causing it to sense 100 psig low.
Neglecting any possible trip or cercm, what effect does this have on feedwater flow control and steam g:ncrator level? 0) Feed flow will match indicated steam flow; SG level wi.'l decrease , continuously.
b) Feed flow will initially mat',h indicated steam flow, but will increase to match actual steam flow in response to level error; SG 1evel will stabilize at a level below setpoint.
, c) Feed flow will match indicated steam flow; SG level will increase continuously, d) Feed flow will initially match indicated steam flow, then will oscillate due to conflicting flow and level error signals; SG level will oscillate around an average level.
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,i .i ,, -m ! .i , q c +e . (1.00) ? (QUESTIOWs1040 t
i . . . . tLict FOUR conditions other than-ranual which will' trip a boiler feed pump.
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, i -- .Lict the FOUR criteria which must be satisfied in' order to terminate safety-
injection;in accordance.with E-0 * Reactor Scrarc or Safety Injection."
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_ _______ - , SENIOR REACTOR OPERATOR Page 45 , ' QUESTION: 042 (1.00) Tho pressurizer spray line breaks off at the pressurizer.
Which of the-following best describes pressurizer level response? c) Initial rapid level drop due to inventory loss, then increase when SI actuates.
b) Initial increase due to boiling and swell in the pressurizer, with further increase from SI injection flow.
c) Initial level decrease due to inventory loss, then increase faster than SI fill rate when pressure drops below T-sat for vessel head temperature.
d) Erratic indication due to turbulence in instrument lines, followed by increase as cold injection flow quenches boiling and fills the pressurizer.
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QUESTION:.043 -(1.00) }- Which of the following best describes the basis for the MCP stop criteria of E-1. " Loss of Reactor or Secondary Coolant"? j a c) Prevent pump damege frem cavitation.
'.5 ) Prevent overpressurizing VC since pumping fluid out the break can exceed the design leakrate.
.c) Keeping the MCP's in service may cause core thermal limits to be exceededu l if they. trip at the wrong time'in a small break LOCA.
d)~ Running MC pumpr. are an unnecessary heat load which increase the' severity of a LOCA which reaches the trip criteria.
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QUESTION: 044 (1.00) In the event of a SGTR with inability to isolate the affected main coolant ' loop, E-3 "Stenn Generator Tube Rupture" directs the operator to cool down the MCS by dumping steam from the intact Steam Generators.
Why doesn't the ruptured SG depressurize as the MCS cools down?
a) Hot mein coolant flowing out the broken tube flashes and keeps the SG
pressurized.
b) Rising level in the ruptured SG compresses the steam volume, holding pressure up.
c) Ruptured SG steam volume is partially insulated from cold water around tubes by hotter water above tubes, thus slowing depressurization.
i , l l d).The ruptured steam generator is isolated, thus there is no energy removal from the ruptured SG.
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- QUESTION
- 045 (1.00)
, Fallowing a reactor scram, E-0 " Reactor Scram or Safety Injection" requires 'th-t a minimum feed flow be established to the steam generators.
What is this minimum required flow? G) 45 gpm to each steam generator b) 45-gpm to all intact steam generators a) 87.5 gpm to each individual steam generator d)' 87.5 gpm total to all steam generators
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Which of the following best describes the reason for this minimum flow? G) Ensure adequate SG inventory to prevent uncovering the tubes.
b) Ensure adequate secondary heat sink.
c)-Ensure no water hammer in feed lines d) Ensure sufficient condensate flow for the air ejector condenser.
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-QUESTION: 048 (1.00) Wh'.:n the size of a primary to secondary leak has been quantified, when would y;u be required to scram the reactor and initiate safety injection? c) Leakrate greater than or equal to 5 gpm.
'b) Leakrate greater than the capacity of one charging pump.
c.
c) Leakrate greater than or equal to 50 gpm.
d) leakrate greater than the capacity of two charging pumps, f
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- _ _ . _.. _ _ _ _ _ _ _ _ _ l , SENIOR REACTOR OPERATOR Page 52 l l l QUESTIONS.049 (1.00): i
While operating at 100% power,~you observe the following indications.
Pressurizer'high pressure alarm - Pressurizer' low pressure alarm i - Pressurizer < heaters =off
E-PR-MOV-1911open
- -MCS= pressure decreasing rapidly , Which of.the following is the most likely cause? 0) PR-P-6 pressurizer pressure failed high.
b) Loss of. power to PR-P-6 pressurizer pressure channel.
, r ~c) PR-MOV-191 failed open.
- i d) Loss,of. power to vital bus #1.
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. -. SENIOR REACTOR OPERATOR Page 53 i QUESTION: 050 (1.00) A reactor startup is in progress.
The reactor operator informs you that the rCictor is critical.
A few minutes later, ha informs you that one of the ocurce range channels has failed low.
Which of the following is the 'i cppropriate action? c) Sinco the reactor is critical, plant startup may continue.
b) Stabilize power; restore the inoperable "hannel to operable status prior to increasing reactivity.
c) Stabilize power; verify compliance with the Technical Specification shutdown margin requirements within one hour.
d) Stabilize power; restore the inoperable channel within one hour or be in hot standby with the scram breakers open within 6 hours.
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SENIOR REACTOR OPERATOR Page 54 QUESTION: 051 (1.00) A reactor startup is in progress.
The reactor operator informs you that he d:cs not have proper overlap on either intermediate range channel.
Which of tha following is the appropriate action? 0) Reduce power below the intermediate range; restore at least one channel to operable status within one hour or be in hot standby within 6 hours.
b). Place the plant in hot standby with the scram breakers open within one
hourt remain in mode 3.
c) Restore at least one channel to operable status within one hour or immediately open the reactor trip breakers.
d) Place the plant in at least hot standby within one hour and in cold shutdown within the following 30 hours unless at least one channel is restored to operable status.
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! . . QUESTION: 052 (2.00)
-
L OP-3054L" Natural circulation" lists five parameters,to be monitored to verify i natural' circulation.. List any FOUR, including value or trend as appropriate.- [
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, SENIOR REACTOR OPERATOR Page 56 , QUESTION: 053 (1.00)
- which of the following functions are disabled as a result of a loss of #2125
) VDC bus? l a) Pressurizer PORV will not open.
b) Loss of control power for #1 2400V bus.
c) Turbine trip will not occur.
. d) Scram breakers will not open.
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i .!
QUESTION: 054 (2.00)
Th3 plant is operating at.100% power when control air. pressure begins -; ~d creasing rapidly.
The reactor should be scrammed under the following two
C;nditions or . The Main coolant !
- purps should be tripped if_either or
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. _ - _ _ _ - _ _ _ _ _ _ _ _ ,. SENIOR' REACTOR OPERATOR Page 58 j QUESTION: 055 (2.00) You are the refueling SRO.
Fuel novement is in progress.
You observe that shield tank cavity level has started decreasing at a visible rate.
Whitt are the EIGHT immediate operator actions required by OP-3117 " Refueling Accidents"? ,p t > a l rf tT5 4 g/t 9 5 #
- -> SENIOR REACTOR OPERATOR Page 59
QUESTION: 056-(1.00) . Tho plant is in Mode 5, with the main coolant system drained down to the top of the loops for maintenance.
If there is a hot leg opening in the McS and a 1c03 of the shutdown cooling pumps occurs, the core could start to become unctvered in approximately what amount of time? , i c) Less than 15 minutes i b) Less than I hour , c) 4 to 6 hours d) Not for at least a hours ! -t l l l
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SENIOR REACTOR OPERATOR Page 60 QUEJTION 057 (1.00) FR-C.1 " Response to Inadequate Coro Cooling" step 9 directs the operator to dsprecsurizc al) intact steam generators (after earlier steps attempt to achieve HPSI flow).
What is the purpose of this step? n) To cool down and depressurize the RCS and obtain LPSI and accumulator-injection.
b) To dump energy from the steam generators so they don't act as a heat source.
-c) To allow maximum possiblo feedwater injection.
d) To provide the maximum possible thermal driving head for natural circulation cooling and condensation of core boiloff in the U-tube ' SENIOR REACTOR OPERATOR Page 61 i l- < QUESTION: 058 (1.00) i While performing FR-C.1 " Response to Inadequate Core Cooling" a loss of all , fo;d capability occurs.
Which of the following is the required action? 0) Continue with actions of FR-C.1.
) b) Transition to FR-H.1 " Response to Loss of Secondary Heat Sink."
c) Transition to ES-0.0 "Rediagnosis."
l d) Perform either a or b based on the Shift Suparvisor's judgement of the best course of action per 10CFR50.54X.
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. - - .= - . SENIOR REACTOR OPERATOR Page 62 QUESTION: 059 (1.00)
The plant is at 100%. The turbine driven EBFP.is out of service for maintenance.
A loss of offsite power occurs, but the diesels fail to start.
The reactor operator informs you that the scram breakers have failed to open.
Which of the following is your course of action? a) Enter E-0 " Reactor Scram or Safety Injection", then transition to FR-S.1 " Response to Nuclear Power Generation /ATWAS."
' b) Perform the immediate actions of ECA-0.0 " Loss of All AC Power", transition i to FR-S.1,.then transition.to FR-H.1 " Response'to Loss of Secondary Heat Sink."
c) Enter E-0, then transition to ECA-0.0 from the step 6 RNO column.
, d) Enter ECA-0.0 directly, make no transitions unless specifically directed by , that procedure.
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... .. .. _ _ _ _ _ _ - _ _ _ _ _ _ SENIOR REACTOR OPERATOR Page 63 QUESTION:: 060 (2.00) What FIVE actions are required to check the MCS isolated in accordance with tho~immediate actions of ECA-0.0 " Loss of All AC Power"? -
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SENIOR REACTOR OPERATOR: Page t34 . QUESTION: 061 (1.00) L .During normal-100%-operation, a Group A control rod drops.
If an automatic . ' , .rcram does not occur, the reactor must be manually scrammed if ' ' or- . , , . l
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SENIOR REACTOR OPERATOR Page.65 QUESTION: 062 (1.00)
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A opurious reactor scram occurs.
Two control rods fail to fully insert.. What action is required? a) Transition to FR-S.1'" Response to Nuclear Power Generation /ATWAS."
b) Borate 500 gallons from the BAMT.
c) Verify adequate SDM within or,e hour, borate if required.
d) Borate 1000 gallons from BAMT.. , ) I
e.
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Page 66 . . QUESTION: 063; (1.00)-
- A large LOCA'is in progress.
SI pumps running.should be reduced to two trains 'if SIT level is less than feet; transfer to VC recirculation should - be-initiated when SIT level drops to feet.
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, SENIOR REACTOR OPERATOR Page 67 .. . QUESTION: 064 (1.00) -Which.of the following best describes the reason for SI trimback prior to-rOnching the criteria for switchover to VC recirculation? c) To-ensure there is adequate time to achieve VC recirculation.
.b)L Full SI flow is no longer needed due to the initial rapid drop in core thermal power in the first few mine+^7 after a scram, t c)LTo unload the diesels if a loss 'e power has occurred.
- Id) To maintain adequate NPSH for th Ps_ u..J LPS1 pumps.
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i l ~ SENIOR REACTOR OPERATOR Page 68 l l i ' QUESTION: 065 (1.00) WhOt is the purpose of transfer to rot leg injection 18 hours after a large ] i-l LLOCA?
'c)' To ensure cooling of the reactor vessel head in order to reduce thermal c l ctress.and collapse voids.
. .b) To; prevent boron precipitation on the upper core _ region which could I interfere with adequate core cooling.
i ! L c) In the worst case, two' phase cooling of the upper core cannot prevent ' heatup of the upper fuel elements; hot leg injcetion is necessary to prevent exceeding design basis clad temperature.
l Hot leg injection is necessary by 18 hours after the accident to flush-the l d) . hydrogen bubble out of the head to preclude ignition by contact with hot fuel elements.
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< p ' , I.. : SENIOR REACTOR OPERATOR = Page 69 L p , -' QUESTION:~ 066 (2.'50) -G)'The FOU2~~immediate actions of FR-S.1 " Response to Nuclear Power Generation / ' ATWAS are and -- , , , . ., b) What is the RNO. action for the first immediate action of FR-S.1? ' A tt 8# gg3 (Lh ' "A g( ,- l
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- - . . SENIOR. REACTOR' OPERATOR Page 70 - ,. . - QUESTION: 067.(1.00) . . . i Two conditions in which emergency boration is required are and' ! ' c .. l , I ( '. ai .j ts ! j i - l l } l r ! . tl' i l l l t' L
.i- . SENIOR-REACTOR OPERATOR Page 71 QUESTION: 068 (1.00) .-Why is Pressurized Thermal Shock a critical safety function? a) Repressurizing a-cold plant can cause safeties to fail, resulting in a IDCA.
b)' Thermal shock reduces allowed plant lifetime due to excessive cyclic stresses.
c)? Thermal shock can cause fuel clad failure and damage core support components.
'd) Repressurizing a cold, thermally stressed plant can cause failure of the reactor vessel, an event beyond the plant design basis.
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SENIOR-REACTOR OPERATOR.
Pags;72 - QUESTION: 069 (2.50) 'OP-3017." Fire Energency" lists five conditions requiring that OP-3018 " Plant Shutdown with SSS" be initiated. -OP-3018 lists four fire-locations which may , generate-those conditions.
Listiany FIVE of these conditions.or fire ilocations.
Credit-is only.given once-for answers that are common'to both ' < procedures.
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.. . SENIOR' REACTOR OPERATOR - Page 73 .i , .. .' QUESTION :., 070 (1.00) ' Two. conditions in which T,-0 " Reactor Scram or Safety Injection" is 'not entered - .inLresponse to a reactor scram are and' , .
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Mb . +fSENIOR REACTOR OPERATOR Page 74- ~ \\;. - , ~ QUESTION: 071- (1.00) o> .Tochnical Specification 3.4.7 "MCS Specific Activity". allows primary coolant ac tivity to exceed -the LCO for 48 hours within the limits of a figure provided - otsallowed activity vs power.
Why is this temporary increase in the-LCO J ac ceptable?. 'e) ToLallow time'for oxidation to seal minor fuel-defects.- b)'LTo allow; time for the purification system to be'placed in service and ~ cleanup minor fuel leaks.
'c)-To! accommodate-iodine spiking from power changes.
-d) To accommodate crud bursts.
x
m .... ... ... . ..... -. ... .. .. - - - - - - - - - _ _ ' . -. Page 75 . .SENIORLREACTOR OPERATOR QUESTION: 072 (1.50) The' plant is-operating normally at 100%. The bleed line radiation monitor alarms.: The channel appears to be operating satisfactorily by channel' check'. Three means of verifying a potential clad failure are , and . _,
,, 1.
SENIOR REACTOR OPERAT M-Page 76 , l I QUESTION: 073- (1.00) .i i p-In-the event of a loss of CCW to the Main Coolant Pumps, the. normal bearing t:mperature trip criteria is raised.
Why is this done? q G) Because the cause of the temperature rise is known not to be-due to bearing
- damage.
i b)2To allow time for' restoration of CCW flow.
. c) To allow MCS flow for two minutes after a scram.
,i d) Heat capacity of the CCW water in the cooling coil is adequate to prevent-heat damage to the bearing up to the increased temperature limit.
! , i l~ l' l-i l' l . - - - - - - - - - - - - - - - -. - -
, b h .I e s .' he< SENIOR-REACTOR' OPERATOR 'Page 7.7: , t , t . QUESTION:' 074- (1.50) - (Lict any.THREE conditions which could cause a loss of condenser vacuum.
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_ SENIOR-REACTOR OPERATOR.
'Page 78 ' QUESTION * 075 (1.00) 'OP-3018 " Plant' Shutdown with the SSS" has been initiated.
The decision has b cn made to cool down to mode 5.
The average cooldown rate should be 'anintained at deg f/ hour. (.5) .( ' 'BRIEFLY EXPLAIN how the cooldown is continued when steaming is no longer- , .Gffactive. (.5) .)- . .
l i SENIOR REACTOR' OPERATOR Page 79 QUESTION: 076- (1.00) 'Ths plant is in Mode 5, with the main coolant system drained down to the top of the loops for maintenance.
A hot leg isolation valve is removed. In the-cVent'of a loss of the' shutdown cooling pumps, what action is required if the-(core exit thermocouples reach 170 dog f increasing? la) ! Place the LPST cooling. pump in operation, b) Inject water from the SI tank at 40 gpm.. c); Establish makeup via the charging system.
d) Unisolate at least two MC loops and fill the associated steam generators to 10 feet for reflux coolin ri:
- 1
' f; ' i ~~ W-SENIOR REACTOR OPERATOR Page 80 , i QUESTION:c077 (1.50).
i For the following examples of Main Coolant System Operational Leakage, state ! J tho CLASSIFICATION of the leakage and the Technical Specification LIMIT on that type'of leakage.
i a)' Steam Generator U-tube leakage.
t b).~MCS piping through wall crack.
c)LReactor Vessel flange leakof.4?.
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L l-(********** END OF EXAMINATION **********)
- - - - - - - - -
_ _ _ ,,. t < SENIOR REACTOR OPERATOR-Page 81-ANSWER:- 001 (1.00) d' REFERENCE: STM ch.13; study question 5 002000K4.09 (3.2) .002000K4.0 ..(KA's) ANSWER: 002 (1.00) . REFERENCE: - STM'14-3, study question 4 010000K1.03 (3.7) 010000K1.0 ..(KA's) ANSWER: 003 (2.00) 1650 + or - 50 psig, 5 psig 1525 +'or - 50 psig A4 7 dd ${5// J iOO 650 + orL-50 psig - REFERENCE: stm;19-2; study question 3 006020A3.01 (4.3) 006020A3.0 ..(KA's)
, (SE'NIOR REACTOR OPERATOR; .Page 82-ANSWER:1-004 '(1.00) , b , REFERENCE: Otr ch 15= 011000A2.10-(3.4/3.6) , 011000A2.1 ..(KA's) , . ANSWER: 005- (1.00) i C
, < REFERENCE: ctm'33; study question 4 012000A2.01 (3.1/3.6) l-012000A2.0 ..(KA's) ! 1 ANSWER: 006 (1.00) ' . I a x REFERENCE: AP-2001',pg. 5,6 294001A1.11 (4.1) L, =294001A1.1 ..(KA's) .. - - - - - - - - .. -
, SENIOR' REACTOR OPERATOR Page'83 ' -ANSWER: 007 (1;50) supply breaker overload-loss of bias to boost / buck magamps t u(>>J J // (ac / t vi )
- ~operationof60CPvoltagebalancer[elay REFERENCE:
- stm 36 study question 13 045000G15 045000G15
..(KA's) . ANSWER: 008 (1.00)' .5% to 70% (4% to 75%) $ '[ F f3 " '" # " '# O f"'
= REFERENCE: stm 36-2 045000K5.01 045000K5.0 ..(KA's) ANSWER: 009' (1. 00 ). .c REFERENCE:- stm ch 36-2; study question 3 045000KS.01 (2.8/3.2) s45000K5.0 ..(KA's) .
.,.,, _ _ _ _ _. .. - . -.
- - - - - - - - -- -- - - - - - - -- - --
TSENIOR' REACTOR OPERATOR; Page 84-ANSWER:: 0101 (1.00) 45,[in auto-REFERENCE: -stmfl0-8 1076000K4.02 (2.9/3.2) '076000K4.0 ..'(KA's) ANSWER: 011- (2.00) as any three of the following: .tubesheet delta T < 100 deg <-200# if < 70 deg secondary <=500# if.< 70 deg primary < or = 1001deg/hr heatup or cooldown b: 30 minutes REFERENCE: T.S.
3.7.2 035000G5 (3.2/3.8) 035000G5 ..(KA's) I l .
. . . - . . . . . ... . !' SENIOR REACTOR OPERATOR.
Page 85 ANSWER: 012 (2.50), a)'2735 # b). restore in 1 hour in mode 1,2
restore in 5 minutes-in mode 3,4,5 c)-IAW T.S.
section 6.7-hot standby in-1 hour-report to NRC in 1 hour-(safety limit violation report) , -(submit.to NRC within 14 days) REFERENCE: T.S.L6.7 K/A 002000G5.(3.6/4.1).
002000G5 .. ( v1. 's ) ANSWER: 013 (2.00) A)' 4: B); 5 .C)-9 ~D)-6 E) 2 F) 3 , (G) 1-H) 8 REFERENCE: stm.37,39 062000K2.01 062000K2.0 ..(KA's) ...... _ _ _. _ _ _ _ _ _
- ' < SENIOR REACTOR OPERATOR Page 86 oANSWER:' 014- (1.00) . .i 'd .,9 v 4 p' j !
i .. ' , REFERENCE: , -l p l T sS-3.4'. 3, STM 39-9: PR-MOV-512 powered from EMCC-1 +
- 062000G5 (3.1/3.8)
Ej ! (; 062000G5 ..(KA's) . ! .i ' ANSWER:
- 015 (1.50)
l L l VD-LCV-408 blowdown tank Isvel control l L WD-FCV-310 liquid waste test tank effluent flow control i CC-RCV-210 CCW surge tank vent valve ! l-l ' ' REFERENCE: l l
t STM-35-12,14,16 ! 073000A2.02 (2.7/3.2) _073000A2.0 ..(KA's) ! '
i , b " t i
-SENIOR REACTOR l OPERATOR Page 87 ANSWER: 016 (2.00) ) ' i c)J935 # j b) 3.4% or 80,872 lbm/hr(3-4%; + or - 5000#/hr) c)'from' curve c-1: Cb = 1000 ppm from curve c-3: MTC = -13 pcm/deg f from curve c-6:' power coef = -8.25 pcm/% including -2.21 pcm/% from MTC . final power: 103.4% .dalta.t-ave: (3. 4 ) (8. 25-2. 21)/13 = 1. 58 deg-j finalit-ave 532-1.58 = 530.42 deg j ! partial credit .1 each figure; .1 recognize MTC in power defect; .1 i calculation ,d) 103.4:%.
, { . REFERENCE: STM ch 3; calculation.above J-D2ta Reference Manual section C 039000A2.05 039000A2.0- ..(KA's) , l ANSWER: 017.(1.00) j , to mitigate or limit massive damage to protect'the safety of the shift or the public , REFERENCE: AP-2001 pg 4
-294001A1.0- ..(KA's).
- - - - - - - - - -
! $ bhhfbfb bb t k d $[ $ N . M # "" si M $[ (N N' WSd$$ Myj hf
';a.y iThesuNrvisory contro1Noom operator-shall be> inforMdY'TQ [ '~ ' ' ' of all plant operations by the control room operator.- e! g f x b.' The supervisory control room operator'(SCRO)'shall bel ' o~~ responsible for proper communications on his shift.< In? performing this responsibility the~SCR0 is responsible for-ensuring that all aspects of good communication on shif t-is established and eff'ectively maintained for the specific - , >s.
evolution as required.
! '.
1.)
When communicating alpha-numeric information,.the sender < 4-and receiver should use a standard phonetic alphabet._ An-t c exception to this practice is the use of approved standard.
4-c abbreviations like SI for safety injection', or: LPSI-for low . Fe ipressure sa'fety injection pump.
> < a I ',.
,. ... 2)- The~ completion of control room ordered actions should be . . ' - hgQi_ reported back to the person in the command and control function. This communication should be acknowledged by - " the person in the control room command and control
- function so that the sender knows he was heard.
j 3) Verbal instructions should not be contained in a verbal ' order where control or coordination concerns exist.
It , is preferable to report back for additional instructions, A verbal message and its ackr ' 21edgecent should follow a > 4) standard format, i.e. " Joe, swact #2 main coolant pump,"
"I understand start #2 main coolant pump."
5) A verbal _ instruction shobid contain pre-cautions necessary > to successfully achieve' the desired results, i.e. " Joe, ,
- when you start #2 main coolant pump make sure that - the PA0 has at least 135 gpm component cooling flow established to #2 main coolant pump " The chain of command shall be followed: y c.
1) The shift supervisor directs SCR0 activities.
2) The supervisory control room operator directs the - control room activities through the control room operators.
, 3) The control room operators direc,A0's for specific activities as assigned by the st ?ervisory control room operator. The SCR0 or the SS may give instructions -- directly to the AO's, but should let the CR0's know i that he has done so.
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N b 1, KM!W ' C. ! - Y d[[Communnati on s e sfialT be et nd uc ted t i nia' prrJe s si ona1[ c nd - ' - !- (1;-i! consistent?manne( -
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. ' . . . R' 'i1 1)- Clear lines of' communications must be established prior
to-conducting evolutions, tests, surveillances, and other - y~ plant operations.
, 2) The n.essage sender must assure that the attention of the receiver har Seen gained by. identifying him by name, prior to sending the message, (i.e. " Joe, line up-#2 condensate pump for operation"). , 3) Messages must be clear, concise,. specific, and presented using proper terminology (i.e. " Start condensate pump #2," not " Start the pump"). 4) Instructions or information involving the operation.of plant equipment shall be repeated back by the receiver , ! to confirm-proper understanding, (i.e. " Mark, you want ' me to start #2 condensate pump"). l , < s 5) Instructions or 1'nformation involving equipmeot identifica-9, tion or 'oth'er numbrical information shall be repeated back T " l by the' receiver t'o' confirm proper understanding (i.e. " Mark, lL ' you wan't me to op'eh CS-MOV-532 SI recirculation valve.") "' ! . .i ' , t !~ 6) The radio shall be used for business only.
&W'
7) Limit radio, telephone or gaitronics use to the minimum possible.
g 8) Sound powered phones shall be utilized for any special
evolutions when it is necessary to maintain constant , communication between an operator at the location of a I significant evolution and the operator controlling the evolution. The operator who is controlling will usually q be loc.ated..in the " Controls Area".
Examples of such L evolutions include but are not limited to the following: ' I:- a) Turbine overspeed testing.
' b) Turbine roll off with steam (0P-2102).
J c) NRV testing.
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_ _ _ _ _ -. -. - - -. ,, . s e SENIOR REACTOR OPERATOR Page 88 ANSWER: 018 (2.00) See AP-2001 B.
4.
" communications" excerpt attached REFERENCE: ' AP-2001 294001A1.03 (3.4) ' 294001A1.0 ..(KA's) ANRWER: 019 (1.50) manipulating the correct component or device instrumentation identified plant response anticipated REFERENCE: 'AP-2001 pg 14 294001A1.0 ..(KA's) ANSWER: 020 (1.50) 1: continuously occupied by a L.O.
2: don't enter unless there is an L.O. in zone 1 (unless emergency) 3: don't enter unless relieved or emergency REFERENCE: APF-2009.1 294001A1.0 ..(KA's)
.3 ++=+ g y e > . k a N $.! R . p 8-p h.
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L. g i !l ' SENIOR REACTOR. OPERATOR Page 59' L , j ANSWER: 021 (1.00) ] , d l REFERENCE:
l
A'P-0017.
J - -_ 2 94 001A1. 03 (3.4) (KA's) 294001K1.0 .. - ANSWER:- -022 (1.00) , tnrottledvalves[et/Co'as VJ br > - REFERENCE: ! AP-0207 pg 2 > 294001K1.01-(3.7) - , 294001K1.0 (KA's) ( .. ! t
ANSWER: 023 (1.50) 2.5 r/qtr NTE 5 r/yr or 5(n-18) lifetimo 11.0 r/qtr 18.0 r/qtr 7.0 r/qtr > REFERENCE: - AP-0803
- i -294001K1.03 (3.4)
! ' ~294001K1.0 (KA's) .. . l-d - ~ , _ _. - - , .,,. -... - -.. . _ - . - -
, , ,,, . . . . .. . .. .-- l LSENIOR REACTOR OPERATOR Page 90 ANSWERt' 024 (1.50) . inhale, ingest (eat ;r Ur""d), absorbthroughskin[-u-o) REFERENCE AP-0804 294001K1.04 (3.5) 294001K1.0 ..(KA's) ' t ANSWER: 025 (2.00)- at 4 additional members b SCRO or an equivalently trained individual c: 2 aux operators shift RP tech security officer . REFZRENCEl-AP-050 294001K1.16 (4.2) 294001K1.1 ..(KA's) ANSWER: 026 (1.50) VC isolation 5 psig L IM' i lowMCpressure1650psig(ft act-o r N ) low pzr level 50 inches
_ _ - _ - _ _ _ _ _ _ _ _ _ _ _ ___.. _ ,
' , . . SENIOR REACTOR OPERATOR Page 91 , ,u
s .. l l
- REFERENCE
, .STM:15-3,4,51-sq 5
- 004000Kl.01-(3;6/4.0)
j i 004000Kl.0- ..(KA's) l .i '
..; fANSWER: 027 .(1.00) i
i- , \\
lO )
i '
- , REFERENCE:
, ! STM:15-33 l' ', 004000K4.04 (3.2/3.1);
J 004000K4.0- ..(KA's) ' -: ANSWER:. 028.(1.00) , 1; l i.. t ,b' 1.
- 1:
.? REFERENCE: l STM.3;1-5,6 study question 2 $ ~~ i 015000K6.01 (2.9/3.2) , u
! i 015000K6.0 ..(KA's) 1' i-l, ANSWER:- 029 (1.00) , ' ld- . 1'. ! .i..... ..... _..-.__.-, ... ... _ _. _ _. _._ _ ,. ~., '
- , - SENIOR REACTOR OPERATOR Page 92 - REFERENCE:
stm 31-27 015000A2.02 (3.1/3.5) 015000A2.0 ..(KA's) ! ANSWER: 030 (1.00) b REFERENCE: _ OP-2113 001000G10 (3.3/3.5) 001000G10 ..(KA's) 7-ANSWER: 031 (2.00) - - IAW OP-2113 - REFERENCE: - 001000G1 (3.7/3.8) _ .OP-2113 _:, . 001000G1 ..(KA's) ANSWER: 032 (1.50) -shutdown margin ' rod ejection accident analysis assumptions on rod worth and hot channel factors l power distribution limits - .- - .-. - _ - -. _ _ _. - _ _ -. _. - _. _ - _ _ _ _ _ - _. _ - _ _ _ _ - _ _ _ _ _
.. _ PRECAUTIONS
,a: .c .l kt > O - 1.. During1plantcperatiencdt 1 iss (hall B21 be move d i ally b i except forJ physics ztestingscontrol'e rod exercise, or; to; correct! ; ' l individual rod position within a group.
't t 4. 2.
All control rod primary and secondary position indi:ator chani.els
shall be operable in agreement within i 8 inches while in Modes 1 a'nd 2.
[6, 3.1.3.2] 3.
All control rods shall be operable and positioned within 18 inches ~ of every other rod in their group while in Modes 1 and 2.
[6.
3.1.3.1.] 4.
Tave should not be allowed to exceed the pertinent core guidelines.
Maintain rods and power level in accordance with Ref. 5.
5.
The rod operation selector switch should be in AUTO position unless the operator is at the rod control section of the control board.
6.
Do not move the rod group selector switch to another position until the rod group stops moving.
Continuous rod withdrawal could occur if the rod group selector switch is positioned to another group while rod motion is taking place.
NOTE: When at power, this first step should be in the 'lH' direction.
7.
After moving the rod group selector switch to another group, check that the group selector lights are out on all other groups. Hove the new group one step and et k that only its odometer moves.
B.
Do not use the ' All Rods In' control except in an emergency.
- - _ _ _ _ When moving control rods continuously, monitor the primary position 9.
indicating lights and the VC sound system so that a rod hangup or other malfunction may be detected as soon as possible.
10.
If any control rod cannot be manually withdrawn or inserted, discontinue further rod motions, THEN initiate AP-2016. ' Event Administrative Actions * and AP-0205, 'Haintenance Request,* if applicable.
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e , . t L i i ., i i[.'8ENIOR REACTOR' OPERATOR Page 93 i p , IREFERENCE: I
, -T.S. basis 001000G6 (2.9/3 8)
'001000G6 ..(KA's) ' f i ci . ANSWER - 033 (2.00)- F-
iM-Gttached ! ! ' -; REFERENCE: [
' ' t
- cta-figure 27-5
, 022000G9 (3.3/3.3) i
022000G9 ..(KA's) , W ! t ANSWFR: ~034 (1.50) ' , o r-4 L _MCC1 (bus 1)~ . "MCCl-(bus.2) MCC2-(bus 2)- ,
L-t-- REFERENCE: h, . k otm'41
1063000Kl.02 ~ (2.7/3.2) .; o 1- ' -063000K1.0- ..(KA's) - - !1.
- ANSWER
- ;
035 ( 1 ~. 00 ) b y > , I C-
-, - . - - ~ + e -, - -. + - ., - -, -. - - -.. ~,,,.. -,,. .,, - - - , ,r +
_ __ VAPOR CONTAINER ATMOSPHERE CONTROL l AN C OL () i AN f' it AllNG COllS COOllNG COILS x, /[ GRAVITY , DAMP (R /y-COOLING COILS , / e-G AT [ VALVES , PRI $$Unitt R -._, '
fg] ( , N COOLING FAN N
e o\\ / RING DUCT - ~ %- _/ g *Q % s GATE VALVES % 'N ' N 4 ' Hl^
PURGE SUPPLY DUCT f'. CO % [ } k k GRAVITY DAMPER }gggy j ~ ~C N O ( DAMPER HEATING ColLS- -, , ROD
NO.1 HEATING - C [y g . l
- .
AND COOLING F AN%. {C %, COOLING COILS - . '
, ' \\
GATE LRING VALVES DUCT e k l ONT A NER TO PURGE EXHAUST FAN * l
F N 11 N i FILTERS GRAVITY DAMPER POST ACCIDENT ~" HEATI COILS - L FAN N % .C. PURGE AIR OPERA"D+ V NO. 4 BO ST Ft F AN F N 22
- ~ PERSONNEL ACCESS HATCH ot bHEATING rIGURE 27-5 COILS 27 19 Re >
I ' , c i-i ih -SENIOR REACTOR OPERATOR-Page 94 .l , i .. REFERENCE: .-; ! jct 3 41.
<:063000A1.01 (2.5/3.3) ! 063000A1.0 ...(KA's) j , ' ANSWER:- 036 (1.00) l
G.
i REFERENCE: .- Ctm 41 '- 063000K4.04 (2.6/2.9)
-063000K4.0 ..(KA's) P ANSWER: 037 (1.00) "I ' fl deg/ min; 4-5 in/ min h*
.. t REFERENCE: ' , Ctm:19-1 ,# 013000A2.02 (4.3/4.5) < . .
013000A2.0 ..(KA's)
ANSWER: 038 (1.00) i i , . i .t l.
i . .. .... - _ _,,. _. _.. - . . .- . -
SENIOR REACTOR OPERATOR Page 95 REFERENCE: otm 13 003000K5.08 (2.2/2.4) 003000K5.0 ..(KA's) ANSWER: 039 (1.00) b REFERENCE: stm 3 059000A2.11 (3.0/3.3) 059000A2.1 ..(KA's) ANSWER: 0.10 (1.00) .2400v undervoltage overcurrent scram breaker open stuck breaker protection-low suction pressure REFERENCE: stm 3 059000K4.16 059000K4.1 ..(KA's)
Il n ,
. SENIOR REACTOR OPERATOR-Page 96 ANSWER: 041 (1.00) , >25 deg subcooling by CET ->87;5Lgpm: feed or > 14 feet L sg MCS pressure >1650# stable _or increasing >30 inches L pzr REFERENCE: , E-0 007000G11 ('4.1/ 4. 3 ) 007000G11- ..(KA's) = ANSWER: 042 (1.00) / b ' REFERENCE: K/A 000008EK3.01 (3.7/4.4) 000008EK3.
..(KA's) ANSWER:- 043, (1.00) , c , REFERENCE: ERG: generic background document 000009EK3.23 ( 4. 2/ 4.-3 ) - '000009EK3.
..(KA's) .i '
. .. ..... . _ .- -..
_ _ _ _, ... - - - - - - -SENIOR REACTOR OPERATOR Page 97 ANSWEP.: 044 (1.00) l^C-REFERENCE: E-3 background document 000038EK3.06 (4.2/4.5) 000038EK3.
..(KA's) ANSWER: 045 (1.00) d REFERENCE: E-0 K/A 000007EK3.01 (4.0/4.6) 000007EK3.
..(KA's) ANSWER: 046 (1.00) b-REFERENCE 2 E-0; stop 16 background 000007EK3.01 (4.0/4.6) 000007EK3.- ..(KA's) . - -. - -. _. -.. -.. - -. - - -
. _ -._ _ ___ . s- ) e i
, I SENIOR REACTOR OPERATOR Page.98 ,
, -
- t ANSWER: 047.
(1.50) - l
l MS monitor lblcwdownimonitor _
y, . ( (.,,,, ' . ch:mistry analysis ' . a t f.W/ /
' REFERENCE: ~ OP-3107;-87*4/')
- 000037EA2.13;(4.1/4.3)
] 000037EA2.
..(KA's) ' ' j i-ANSWER: 048 (1.00) >- )
- C,
, ' REFERENCE:: e ' E .OP 3109 000037G11 (3.9/4.1) j l- .- l 000037G11 ..(KA's) o
' ' ANSWER: 049 (1.00) ' ' ,. 'b !
- REFERENCE
OP-3111 027000G11 (3.6/3.7) 027000G11 ..(KA's) r ,
-... - . , , + . ..-# r--- -- -....-., .,.. . ~,-
- -
,, ___ -, -. - _ _ -. _ _ - - - 3- - ; , .l
- 8ENICR' REACTOR OPERATOR
Page 99 ' ., d . ANSWER:' 050 (1.00) , ., 'y, & .!
.g
REFERENCE: e i .T.S.; OP-3120.
! - 000032G3 (2;6/3.3) _ 000032G3 ..(KA's) 'l . ANSWER ' 051 (1.00)
- , ]
b 01v- . .
- REFERENCE
, =T.S. 3.0.37.OP-3120 - , 000033EK3.02 (3,6/3.9) 000033EK3.
..(KA's) ' - . , i'
- ANSWER:~
052 (2.00) ^ cubcooling >.40'deg f i s /eg g,f,F ' ) ' SG pressure stable or decreasing [dk f.. p.l. y /e <./ os , T-hot stable or decreasing , CET stable or decreasing ! T-cold near T-s( for P-sg.
' -[ ,
- REFERENCE-
-
- OP-3054 000056EK1.01 (3.7/4.2)
'000056EKl.
..(KA's) . i e ' '! , , ., , . .-.
._ r CENIOR REACTOR OPERATOR Page100 ANSWER: 053 (1.00) -b Ed u4 HEFERENCE: OP-3257 000058FA2.03 (3.5/3.9) 000058EA2.
..(KA's) ' ANSWER ' 054 (2.00) couro*u Aca fit d IS u rid completelossofpria.uypreuuureccntrol[t 2 c3aplete loss of CC to MCPS, can't rapidly rectore. ((c - r v-7 *//'2 * f .lo s of CC > 3 ninutes 200 deg bearing temperature REFERENCE: OP-3002 000065EK3.03 (2.9/3.4) 090065EK3.
..(KA's) ANSWER: 055 (2.00) I con attached REFERENCE 3
, OP-3117 f;00036G10 (3.7/3.9) 000036G10 ..(KA's) l l !
m-
_ _....... . ML-11-OP-3117 Rev. 15 ey ^ . s S Case F.
UY6xnlained Decreasing Shield Tank Cavity level e {
SYMPTOMS 1.
Low Level Alarm from the Shield Tank Cavity ($TC).
2.
Visible decreasing STC level.
' IMMEDIATE OPERATOR ACTION 1.
Return any irradiated fuel assemblies in the STC to the core or SFP expeditiously or get fuel below reactor nozzle by laying on top of core.
2.
Return any control rods stored in the STC to the reactor core if possible or lay them on the STC floor horizontally if necessary.
3.
Place fuel chute carriage above lower lock valve'and Close the lower lock valve.
4.
Visibly check the STC fuel chute clapper valve Closed.
5.
Evacuato the VC of all non-essential personnel.
6.
Clear the manipulator crane and polar crane of personnel when necessary actions are complete or when radiation levCs are ' excessive.
7.
Notify the Control P.oom, Plant Emergency Director (operating shift supervisor) and Radiation Protection Department.
8.
Initiate OP-3300, " Classification of Emergencies."
SUBSEQllENT OPERATOA ACTION 1.
Check the STC bladder seal pressure, normal pressure is 30 PSIG.
2.
Close the tell tale drain valves.
3.
Verify that the MCS loop stop and drain valves are in the proper position.
4.
Check icop areas and interconnected systems for signs of
volume changes.
5.
Check PAB cubicle area for Shutdown Cooling (SC) integrity; if necessary, then refer to OP-3113, " Loss of Shutdown Cooling Case V."
6.
If necessary to restore Shield Tank Cavity level from the VC sump, then utilize the safety injection pumps as directed the , ( Shift Supervisor.
7.
If high radiation becomes a problem, then go to Case B of this procedure.
! t ' - - - .
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- 8ENIOR REACTOR OPERATOR Page101 j
! ANSWER: 056 (1.00) b.l
REFERENCE: l , CP-3121, OP-3113 ! .000025EK1.01 (3.9/4.3) ' , 000025EK1.
..(KA's)- ] ' ! ! ANSWER: 057 (1.00)
,
. . ~ REFERENCE: l - . 'FR-C.1 background-document 000074EK1'.03 ( 4. 5/ 4 ~. 9 ) < 000074EK1.
..(KA's) f -! ANSWER: 058~ (1.00) i ! e
- -
p--REFERENCE: .t u FR-C.1; user's guide
1000074G12 - (4. 3/4. 4 )- ' 000074G12 ..(KA's)- t > L . - r p - - - 2 ) - _- . .
-
-( SENIOR REACTOR OPERATOR Page102 ., ANSWER: 059 (1.00) d I REFERENCE: .urcr's guide, background document 000055EK3.02 (4.3/4.6) 000055EK3.
..(KA's)
ANSWER: 060 (2.00) . ch ck PORV - X ch ck CH-LCV-222 - X Icip drain - X MCS vent - X MCS sample - X REFERENCE: ECA-0.0 000055G10 000055G10 ..(KA's) ANSWER: 061 (1.00) T-cVe stabilizes below MTC i l c2n't realign the rod within one hour L l .
.. _ _ _ _ _. _ __. - _. , i ' SENIOR REACTOR OPERATOR Page103 REFERENCE: . CP-3118 . . 003000G11 (3.6/3.8)- , 003000G11 ..-(KA's) - , i ANSWER:. 062 (1.00) . ! . i - d:
- REl'ERENCE:
i . .
~ '4S - 0.1 - + - 000005EK3. 01 - (4. 0/4. 3). ,. l 000005EK3.
..(KA's)
- ANSWER: 063 (1.00) i ? I o 20',-10' -f ! , REFERENCE: .; s-E-l' - 000011EK3.15 (4.3/4.4)
- 000011EK3.
..(KA's).
- l " l: . '. ANSWER: 064 (1.00) .> > d- !
i < ,
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- . . . SENIOR REACTOR OPERATOR Page104 . REFERENCE: E-1 step 17 background 000011EK3.12 (4.4/4.6) 000011EK3.
..(KA's) ANSWER: 065 (1.00) b REFERENCE: E-1~ background 000011EK3.13 (3.8/4.2) 000011EK3.
..(KA's) ANSWER: 066 (2.50) verify scram verify turbine trip start EBF pump smergency borate-manually scram; if not, then ALL RODS IN ' REFERENCE: FR-S.1 000029G10 (4.5/4.5) 000029G10 ..(KA's) ANSWER: 067 (1.00) FR-S.1; >2200 ppm Cb needed for adequate SDM C N i c A t. t r y A M / D E N f,' ,0ffMMAleATMN /f 4u. t/ S% . . -._ _
T-SENIOR REACTOR OPERATOR Page105 REFERENCE: OP-4708 SDM FR-S.1 000024EK3.01-(4.1/4.4) 000024EK3.
..(KA's) ANSWER: 068 (1.00) d REFERENCE: II/A-000000EK1.01 (4.1/4.4) g 0000402K1.
..(KA's) ANSWER: 069 (2.50) loca primary pressure control loza secondary. heat removal
1cca secondary makeup 1000 all AC > 30 min i loca control. room-for any. reason (control room good for l firo in control-room 1 answer only) L fira in bus room ' fira in SI/EDG bldg ., firo in turbine bldg pump room ! I I l. REFERENNE: OP-3017,3018 i 000067G11 (3.8/4.0) l
000067G11 ..(KA's) ! l i
f I ' SENIOR REACTOR OPERATOR' Page106
-ANSWER: 070 (1.00) ceram due to OP-3018 1C:s all AC -ECA-0.0 Sp (h! fe,N.y-REFERENCE: ! e CP-3018, user's guide 000068G11'(4.0/4.11) ' 000068G11 ..(KA's) . ANSWER: 071 (1.00) ' e . REFERENCE: T.S. basis 000076G4 (2.1/3.7) 000076G4 ..(KA's) i ANSWER: -072 (1.50) valve room monitor ocmple room monitor PAB, chg cubicle monitor 1ccal survey ch:m. sample r30ctor instrumentation .fV) ML-$LAff -. .. - . - - . .. . .
f l SENIOR REACTOR OPERATOR Page107 .. REFERENCE: -CP-3109 000076EA1.04 (3.2/3.4) 000076EA1.
..(KA's) . ANSWER: 073 (1.00) e REFERENCE: OP-3115 _l 000026EK3.03 (4.0/4.2)
000026EK3.
..(KA's) ANSWER: 074 (1.50) l cir inleakage l ' 1ecs cire water i lors GS steam i lo2s air ejectors REFERENCE: -! OP-3781 l 000051EA2.01 (2.4/3.7) !
' 000051EA2.
..(KA's)
.
ANSWER: 075 (1.00) 10 deg/hr > fill-SG water solid, maintain 10 deg/hr with secondary makeup
i i t t .
' SENIOR REACTOR OPERATOR Page108 REFERENCE: OP-3018 000068EK3.12 (4.12/4.5) 000068EK3.
..(KA's) ANSWER: 076 (1.00) 'b REFERENCE: OP-3121, OP-3113 000025EK1.01 (3.9/4.3) 000025EK1.
..(KA's) ANSWER: 077 (1. E 1) a: primary to secondary leakage; 1 gpm total b: pressure boundary leakage; none c identified leakage; 4 gpm-REFERENCE: Tech Spec 3.4.5.2 002000GS_ ..(KA's) (********** END OF EXAMINATION **********)
_ _ _ _. _ _ _ 054 2.00
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, TEST CROSS REFERENCE Page
QUESTION VALUE REFERENCE _ . 001 1.00
002 1.00
003 2.00
004 1.00
005 1.00
006 1.00
007 1.50
008 1.00
009 1.00
010 1.00
011 2.00
012 2.50
013 2.00
014 1.00
015 1.50
016 2.00
017 1.00
018 2.00
019 1.50
020 1.50
021 1.00
022 1.00
023 1.50
024 1.50
025 2.00
-026 1.50
027 1.00
028 1.00
029 1.00
030 1.00
031 2.00
032 1.50
033 2.00
034 1.50
035 1.00
036 1.00
037 1.00
038 1.00
039 1.00
040 1.00
041 1.00
042 1.00
043 1.00
044 1.00
045 1.00
046 1.00
047 1.50
048 1.00
049 1.00
050 1.00
051 1.00
, 052 2.00
' 053 1.00
_ _ _ _ _____ ___ __ _. ___ _ I TEST CROSS REFERENCE Page
QUESTION VALUE REFERENCE - 055 2.00
056 1.00
057 1.00
058 1.00
059 1.00
060 2.00
061 1.00
062 1.00
063 1.00
064 1.00
065 1.00
066 2.50 G6 057 1.00
068 1.00
069 2.50
070 1.00
071 1.00
072 1.50
073 1.00
074 1.50
075 1.00
076 1.00
077 1.50
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1 ATTACHMENT NO. 2 FACILITY *S WRITTEN COMMENTb i
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- . .*: , , , MEHQEAlfDUM To: Regional Admin.
USNRC-June 21, 1990 From: .D. White YAEC-Rowe TR-90-230.5.2 Subject: SRO EXAM COMMENTS ! The following are comments for the NRC SRO Exam dated June 18, 1990, submitted to Jim Prell on June 21, 1990: Qpestion Enz Commento Referencgg 002 Answer should be C STM 14-3 ' QO3 B 850-900 for HPSI, 1500 is STM 19 only if piggybacked
007 Buck & boonc also known as OP-3674 EXCESSIVE REACTIVE 008 H%: differs in Mitigating Core LP 4003 Damage course.
We will update Systems Training Manual 014-Should be a.
Due to loss of TS 3.8.2.1 non-redundant bus, 1 hr shutdown required, d only makes valve inoperable 024 Wounds belong with absorbing OP-0804 through skin 037 Double jeopardy - if answer is 1 degree and 5" should accept 10" if puts 2 degrees - this still shows 1 degree = 5" 042 Taught that level increases, Power answer b Safety film P 514 j 044 Typo answer should be C 047 Leakage is also verified by OP-9105 sampling for boron, fission products, tritium 052 Steam dump adjustnonts change OP-3054 steau pressure 053 Accept B - OP-3257 list both OP-3257, b&d Case 2 - -- -
.... _. .__ , 'i ,< a . zo .
- SRO EXAM COMMENTS - 6/18/90-(Continued)
Page 2-054 Typo - should be loss-of OP-3002 control Air vice loss primary e pressure control as per OP-3002, Loss of component cooling.to t MCPS is commonly referred to as TV-205, 208 . 067 Also accept Criticality OP-3117, ___ Accident, Depress, of All Case A = 4 S/Gs ECA 2.1 Step 4 0,70 Also; accept Physics Testing EoP Users Guide - s 072 Also accept PVS, Main ~ Coolant past plant System Leak Air Particulate performance' - Monitor; chemical analysis noble gas,' iodine, fission products (question did not ask for a " g - specific procedure) . _ 'If-you have any questions, please contact David White at , (413) 625-5140, ext 312.
_ - Submitted by: _ F // _S2_;d2d_.5.8 David H. White - operations Training Supervisor - - L _ =_ - - E.
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