IR 05000219/1988034
| ML20247A031 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/03/1989 |
| From: | Lange D, Walker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20247A029 | List: |
| References | |
| 50-219-88-34OL, NUDOCS 8907210111 | |
| Download: ML20247A031 (70) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT l
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EXAMINATION REPORT NO.
88-34(OL)
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FACILITY DOCKET NO.
50-219 l
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FACILITY LICENSE NO.
OPR-16 l
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LICENSEE:
GPU Nuclear Corporation
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P. O. Box 388 l
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Forked River, New Jersey 08371 FACILITY:
Oyster Creek Nuclear Generating Station EXAMINATION DATES:
December 19 - 22, 1988 CHIEF EXAMINER:
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. Valkst, Senior Operations Engineer Date
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APPRO ED BY:
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David J. Tange, Chief, BWR Section Date Operations Branch, Division of Reactor Safety SUMMARY: Written examinations and operating tests were administered to
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four (4) senior reactor operator (SRO) candidates. All of the (
candidates passed the written examination and the operating j
test.
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i U.S. NUCLEAR REGULATORY COMMISSION REGION I l
OPERATOR LICENSING EXAMINATION REPORT
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EXAMINATION REPORT NO.
88-34 (0L)
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FACILITY DOCKET NO.
50-219 FACILITY LICENSE NO.
OPR-16 LICENSEE:
GPU Nuclear Corporation P. O. Box 388 Forked River, New Jersey 08371 FACILITY:
Oyster Creek Nuclear Generating Station EXAMINATION DATES:
December 19 - 22, 1988 CHIEF EXAMINER:
T. Walker, Senior Operations Engineer Date APPR]VED BY:
David J. Lange, Chief, BER Section Date
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Operations Branch, Division of Reactor Safety l
SUMMARY: Written examinations and operating tests were administered to four (4) senior reactor operater (SRO) candidates. All of the candidates passed the written examination and the operating test.
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DETAILS
~ TYPE OF EXAMINATIONS:
Replacement EXAMINATION RESULTS:
R0 SR0
[ Pass / Fail [ Pass / Fail
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[ Written
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N/A
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4 / 0
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[ Operating
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'[0verall
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CHIEF EXAMINER AT SITE:
T. Walker, Senior Operations Engineer 2.
OTHER EXAMINERS:
C. Sisco, Operations Engineer (Examiner in Training)
T. Eas11ck, Operations Engineer (Examiner certification)
3.
The foll'. ing is a summary of generic strengths and deficiencies noted on the operating tests. This information is being provided to aid the licensee'in upgrading. license and requalification training programs. No licensee response is required.
STRENGTHS Awareness of contamination and radiation control in general, a.
b.
Good use of Emergency Plan Implementing Procedures DEFICIENCIES No generic deficiencies were noted 4.
The following 13 a summary of generic strengths and deficiencies noted from the gradisig of the Senior Reactor Operator written examination. This information is being provided to aid the licensee in upgrading license and requalification training' programs.
No licensee response is required.
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STRENGTHS
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a. Understanding of overall nuclear power plant theory - Section 5
b. Knowledae of Turbine Control System components'- Question 6.01 c. Knowledge'of Nuclear Instrumentation automatic response -
Question 6.05 d. Understanding of plant response during fill and pressurization of RWCU System - Question 7.03 e. Knowledge of " Reactor Scram", Procedure ABN-3200.01 - Question 7.07 f. Knowledge of "TBCCW Failure Response", F ocedure ABN-3200.20 -
Question 7.08 g. Knowledge and use of Technical Specifications - Question 8.01, 8.02, 8.10 and 8.11 DEFICIENCIES-a. Knowledge of air accumulators associated with the Isolation Condenser make up valve - Question 6.03 b Understanding the reason for bypassing the turbine trip signal to Reactor Protection System ~- Question 6.04 c. Understanding of why containment water level above it's limit is a concern - Question 7.01 d. Knowledge of differences between Category I and Category II events as defined in " Notification Events", Procedure
- 126 - Question 8.08 5.
Personnel Present at Exit Interview:
NRC Personnel T. Walker, Senior Operations Engineer C. Sisco, Operations Engineer T. Easlick, Operations Engineer D. Lew, Resident Inspector i
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Facility Personnel E. Fitzpatrick, Vice. President and Director, Oyster Creek R. Barrett, Director, Plant Operations J. Kowalski, Manager, ' Plant Training G. Cropper, Manager, Operator Training H. Tritt, Supervisor, Operator Training M. Heller, Oyster Creek Licensing Engineer J. Christian, lead Q. A. Auditor 6.
Summary of NRC Comments Made at Exit Interview:
On December 16, 1988 a pre-examination review was conducted at the NRC Region 1 Office. The Facility was informed that, although a post exam review was not conducted, comments on the written examination would be accepted. These comments should be sent to the NRC within five working days.
Access to the facility went smoothly with excellent support from Health Physics and Security.
The Operations staff was co> perative in maintaining a control room environment which was conducive to the administration of the operating test. The generic strengths noted on the operating test (listed in paragraph 3 ) were presented.
The Energy Center was used to conduc: the written examination.
Improvements in' exam security were di. cussed with emphasis on finding a testing area in which the rest rooms could be isolated, for-the exclusive use of the candidates. During the exam the examiners noted excessive noise in the lobby of the Energy Center i
and had to take corrective action.
The examiners noted a generic problem with the lack of pre-staged equipment necessary to perform the "in plant" steps of the E0P'S.
The examiners indicated that the current control room copy of
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Technical Specifications wa. missing page 3.1-17 and that TBCCW i
inlet valve to E M.G. Set (V-5-87) was improperly labeled.
l The results of the examinations would not be discussed at the exit
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meeting but would be contained in the Examination Report. Every effort would be made to send the candidate's results in approximately 30 working days.
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Summary of Facility Comments Made at Exit Interview:
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The facility pointed out that the comment on the pre-staging
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l of equipment for E0P'S is currently being addressed as part l
of an earlier NRC inspection. Immediate corrective action was taken to replace the missing page in T.S. and replacement of the valve I.D. for V-5-87.
The facility. training personnel commented that this was the first time the Energy Center was used for a written examination. Every effort would be made to find a more suitable testing location.
Attachments:
1.
Senior Reactor Operator Written Examination and Answer Key 2.
Facility Comments on Written Examination after Facility Review i
3.
NRC Response to Facility Comments
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Nuclear-Regulatory Comnission Operator Licensing
.1 Examination l
This document is removed from
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Official Use.Only category on
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date of examination.
l NRC Official Use only
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U.
S. NUCLEAR REGULATORY COMMISSION
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SENIOR REACTOR OPEEATOR LICENSE EXAMINATION
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FACILITY:
OYSTER CREEK REACTOR TYPE:
BWR-GE2 DATE ADMINISTERED:
88/12/19 EXAMINER:
_(C REGION I
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CANDIDATE IllETRUCTIONS TO CANDIDATE:
'Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at.least.70% in each category and a final grade of at i
least 80%.
Examination papers will be picked up six (6)
hours after the examination starts.
% OF CATEGORY E OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
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25.00 25.00-5.
THEORY OF NUCLEAR POWER PLANT OP!' RATION, FLUIDS AND THERMODYNAMICS 25.00 25.00 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_2L.RR_
25.00
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PR0rEDURES - NORMAL,' ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 25.00 25.00 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100.0
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Totals Final Grade All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature
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NRC RULES'AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
.ti Cheating on the examination means an automatic denial of your application-and could result in more severe penalties.
2.
Restroom 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.
'3.
.Use black ink or dark pencil only to facilitate legible reproductions.
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' Print your name in the' blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
' Print your name in the upper right-hand corner of the first page of each f
saction of the answer sheet, 8.
Consecutively number.each answer sheet, write "End of Category __" as appropriate, start each' category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
'9.
Number each answer as to category and number, for example, 1.4.
.3.
i 10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated'in parentheses after ne question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer j
to mathematical problems whether indicated in the question or not.
i 15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
'You must sign the statement on the cover sheet that indicates that the work is your own and you nave not received or been given assistance in coinpleting the examination.
This must be done after the examination has been' completed.
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'D 18, When you complete your examination, you shall:
a.
Assemble your examination as follows:
(1)
Exam: questions on top.
(2)
Exam aids - figuras. tables., etc.
(3)
. Answer pages including figures which are part of the answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions.
c.
Turn _in all scrap paper and the balance of the paper that you did not use for answering the questions.
d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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.- d THEORY OF NUCLE &R PbWER PLANT OPERAT10N'.
Page 2-FLUIDSd![D THER!iOIlYSMICS-
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,4-QUESTkON'
5'01:
(2.00),
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EXPLAIN the basis for the Linear Heat Generation Rate'(LHGR)
limit. Include.any limiting conditione and a description of i
the failure ~ mechanism associated with LHGR.
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l QUESTION 5.02 (3.00)-
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INDICATE TRUE or FALSE for each of the following statements concerning neutrons:
A.
The delayed neutron fraction of the core is dependent upon the type and enrichments of.the fuels present.
(0.75)
B. The value of the delayed neutron fraction will increase as the core ages.
(0.75)
C..On a down power transients the rate of power change is limited to the rate of decay of the shortest lived precursors.
(0.75)
D.
Delayed neutrons increase the average neutron lifetime in each generation.
(0.75)
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L :.QUESTI0ll'
5.03 (3.00)
k Concerning cor. trol rod movement:
A.
IDENTTPY the core power distribution (AXIAL, RADIAL. or BOTH)
that ~is.' predominately affected by each of the following control rod' groups:
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Shallow Rods (0.5)
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Intermediate Bods (0.5)
3. Deep Roda (0.5)
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BRIEFLY EXPLAIN the Reverse Power Effect. Include in yaur
answer which control rod group causes this effect.
(1.5)
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QUESTIOti 5.04-(3.00)
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~Concerning natural -4reulation:
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STATE three (3)
..ditions that must exist for natural
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circulation to occur in a~ CLOSED SYSTEM.
(1.5)
B. Shutdown Cooling System Operation Procedure #305 instructs the operator that "If during cold shutdown t.here are no reactor recirculation. pumps available and shutdown cooling is neede.d, w vessel water-level must be greater than 185" above TAF., E.WLAIN the purpose of this Caution.
(1.5)
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)i QUESTION 5.05 (3,00)
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l Certain plant evolutions could subject the reactor vessel to excensive. stresses tf operational limits were not observed.
LIST.FIVE (5) evolutions that are controlled by operational limits that could subject the reactor vessel to excessive
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stresses.
(3.-0)
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QUESTION 5.06 (3.00)
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For each of the events listed below.-state WHICH reactivity coefficient'will respond first WHETHER it adds positive or negative reactivity.and the reason WHY
'A.
SRV opening at 100% power (1.0)
f EB. Rod drop from 100%-power (1.0)
C.
Isolation of extraction steam to a feedwater heater string i
at-100% power.
.(1.0)
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~(3.00)
The plant had been at 100% power for Su days when a reactor scram occurred. The Shutdown Margin (SDM) is 1% delta K/K
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- 10 minates after the scram. Assume that temperature is maintalned at 540 deg F.
.A.
During each of the following periods after the scram, will the SDM INCREASE DECREASE or BEMAIN THE'SAME? EXPLAIN WHY.
1.'10 minutes to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (1 0)-
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30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> to 3 days (1.0)
B. Will the SDM be LESS THAN, EQUAL TO, or GREATER THAN 1%
delta K/K at the end of the 3 days period following the scram.
EXPLAIN WHY.
(1.0)
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-*-QUESTION.-
5.08.
(2.50)-
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Given.the following information:
APRM =.49.50%
'GAF =
1.02 MFLPD =
.50
A. Is the ratio FRP/MFLPD less than'1.0? SHOW ALL WORK.
( 1. 5 )
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B. WHAT action.is required if FRP/MFLPD is less than 1?
(1.0)
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QUESTION-'
5.09.
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-A, STATEitwo.(2)/ reasons that nubcooling'is. required in the-
.downcomer region of a.BWR during~ power. operations.
( 1. 0 ).
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B; Feedwater enters the reactor at 1000 psia and 400 deg F.
DETERMINE'the. amount of subcooling of the feedwater in BTU /lbm.
(1.5)
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f i 6L * Pl. ANT ' SYSTEMS DESIGN. CONTROL. ' AND INSTRUMENTATLG1 Page 11'
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I QUESTION:
6.01 (3.00L l
I For each of the Turbine Control System components'1'isted j
below (A thru D) SELECT the correct purpose of the component.
(1 thru 4)
A.
Pressure control unit (0.75)
B.
Acceleration relay (0.75)
C.
Speed / load control un.t (0.75)
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D.
Speed relay (0.75)
1 ' Controls turbine speed during start and synchronizing
as long as the generator output breakers are open. It controls turbine load when the generator is connected to the grid.
2. Develops the control signal which represents reactor steam flow demand.
3. Receives inputs from the speed load changer and load limit device to determine which will be the controlling signal.
4.
Initiate 9 a rapid closure of the turbine control valves to prevent an overspeed trip following a generator load reject.
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Concerning the reactor recirculation flow control system:
A. During your shift-you notice the loop C Individual Controller manual-auto switch is in the "BAL" posit, ion. WHICH controller
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'is controlling loop C?
(0.5)
B.
If the loop B-Individual Controller switch is in " MANUAL" WHAT_two.(2) signals are compared by the controller circuit to produce the signal that is sent to the Bailey (scoop tube)
positioner?
(1.0)
C. WHAT device provides a position signal to the scoop tube during the M.G_ Set start sequence?
(0.5')
D. WHAT in the input to tne Master Controller when the Master Controller is in " MANUAL'?
(0,5)
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- QUESTION 6.03'
(3.00)
An automatic initiation signal for the; Isolation Condenser System
'has been received and the condensers are in operation, when a steam line break occurs-BETWEEN the pressure vessel and the differential-pressure detectors for the "A" condenser.
A; Will the
"A" Isolation Condenser isolate?
BRIEFLY EXPLAIN your answer.
.(1.0)
10 On an isolation signal, WILL-a LOSS:of 480 volt AC power'
(1.0)
PREVENT the ISOLATION of the condenser? EXPLAIN.
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'C. ~ WILL a LOSS of Instrument and Service Air PREVENT makeup (1.0)
to the Isolation Condenser? EXPLAIN.
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' QUESTION 6.04 (3.'00),
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For the below listed Beactor ProteM son system trip signals.',
. EXPLAIN WHY each is bypassed as'shown.
(3.0)
Reactor Protection Trip System
. Bypassed
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1. Low-Condenser Vacuum
< 600'psig and not-in RUN i
2. Turbine Trip:
Turbine steam flow (40%
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3.
fiode Switch to Shutdown After a 20 second time delay
.
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- EQUESTION 6.05 (2.50)
For the following list of conditions DETERMINE whether a SCRAM, HALF-SCRAM, ROD BLOCK, or~NO ACTION is generated. For'a' condition that could produce more 'than one action STATE the more liniiting action only.(ie: half-scram is more limiting than a rod block)
A.:APRM Channel 1 downscale Beactor Mode switch in Startup/ Hot-Standoy.
(0.5)
B.
IBM' Channel 8 on range 3, downscale, Reactor Mode switch in Startup/ Hot Standby.
(0.5)
C.
APRM Channel'6 with 4' operable LPRM inputs, Beactor Mode switch'in Run.
(0.5)
D. APEM Channel 4 has "APRM HI" alarm, Reactor Mode switch in Run.
(0.5)-
E.
IRM Channel 5. fails upscale, Beactor Mode switch in Startup/ Hot Standby.
(0.5)
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l4 ESTION 6.06
'( 2 ; o0 ).
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Concerning the Containment Spray System:
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A.. LIST the algnalf bi: required for an automatic initiation of the
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( 1,- 0 )
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,B. EXPLAIN the purpose of the 201second, flow dependent, time delay associated-with the 5% valve. INCLUDE in your explanat' ion ~the-
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reeulta.of.not having this time delay.
(1,0)
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?.07
- (2~.00).
QUESTIDN
1 EFor:EACH of the applications or' characteristics from' COLUMN-A'
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CHOOSE the' appropriate. radiation detector.in COLUMN B.
(2.0)
COLUMN A-COLUMN B A.
Used primarily as.a device for-precise 1.
Ion Chamber measurements or where high sensitivity is required.
2. Scintillation
'Ih Air. Ejector Off Gas Rad. Monitor 3. Geiger-Mueler LC. The sise of the electron pulse is
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proportional to'the original incoming-radiation. energy.
D.
Isolation Condenser Vent Rad. Monitor.
R. The same size pulse is generated regardless of-the. type'and specific ionization characteristics of the' radiation.
F.
BBCCW Rad. Monitor.
G.
Uses a photocathode to convert light into free electrons.
K. Main Steam'Line' Rad. Monitors.
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. QUESTION-6.08'
(2.00)
lNDICATE if the'following. Reactor'Feedwater valves FAIL SHUT,
' FAIL OPEN or FAIL JM3 IS on a 'completa loss of Instrument / Service.
Air.
(2.0)
L
,
JA.~ minimum flow valve L
.
B. feedwater flow control valves C.
feedwater flow-control, bypass valves (low flow valves)
D. high1 pressure feedwater heater normal drain valves I
(*****
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EMRVfSyLINCLUDE theicontrolfawitch. position required for
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? UA.'LCOMPLETElthe1following: statement with~ON'I or MORElof'the
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Ei 4 B. STATE.the purpose'ofothe check:".11ves ( V-18-26 & 28 ).: in the-
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](I jTheJdugment'ed: Fuel. Pool; Cooling. System supplies additional
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'i 7 'PROCEDUREE
~ NORMAL, ABNORMAL. EMERGENCY Page121
.
'AND-RADIOLOGICAL CONTROL
.
l
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l QUESTION-7.01 (3.00)~
I Concerning the RPV Flooding procedure EMO-3200.08:
A.
WHAT is the purpose of'the. procedure?
( 1. 0 ).
B.
Conditional statements in RPV Flooding instruct the operator to terminate injection into the RPV from sources external to primary containment to limit containment water level, irrespective if whether adequate core cooling is assured.
STATE two (2) concerns with the water level in the containment increasing above it's limits.
( 1. 0 )
C.
EXPLAIN'what is meant by MINIMUM RPV FLOODING PRESSURE.
(1.0)
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. QUESTION 7.02 (3.00)'
A.
WHAT Emergency Operating Procedure (s) would be entered with Drywell Pressure above 3.0 psig?
(0.75)
B; (TRUE/ FALSE) A torus water temperature of 85 F would be an entry' condition'for " Primary Containment Control" Procedure EMO-3200.02.
(0.76)
C.
" Secondary' Containment Control" Procedure EMO-3200.11-has an entry condition which requires a Reactor Building Ventilation exhaust radiation. level above (0.75)
.
D.
(TRUE/ FALSE)"RPV Control" Procedure EMO-3200.01 would be entered if RPV pressure had increased to 1060 psig.
(0.75)
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.iOUESTION 7 03-(2.00)
A-caution in the " Reactor Water Cleanup System" Procedure #303; concerning the' system inlet valve IV-16-14) states: "V-16-14 should be slowly bumped open when filling and pressurizing'the Reactor
. Cleanup:syntem".
A.
EXPLAIN the two (2) adverse effects on Reactor water level which'could result from opening V-16-14 too quickly.
(1.5)
1B.
In addition to the Rx vessel, WHAT system could be 'ised to fill the cleanup system per this procedure?
(0.5).
l l
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QUESTION.
7.04 (P.50)-.
-
Concerning "RPV Control".. Procedure EMG-3200.01:
A.' IDENTIFY three.(3) systems utilized to augment RPV level' control (1.5)
.
-( Alternate Injection Subsya tems)
B. BRIEFLY EXPLAIN'when these systems are required to be utilized.
Il.0)
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W QUESTION 7.05 (2.00)
In accordance with-Oyster Creek. Emergency' Plan Implementing Procedures IMP-1300.01 and IMP-1300.02, MATCH the individuals with their responsibilities listed below.
lA.-Group Shift Supervisor.
L( 0. 5')
B.
Emergency Director (0.5)
C.
Emergency-Support Director (0.5)
D. Operations Coordinator (0.5).
1. Overall responsibility for managment of'the response to the accident and recovery operations from the EOF.
2.
Responsible for the initial evaluation of abnormal or emergency-site conditions and for directing immediate Emergency' Plan Implementing Procedures' emergency actions.
3.
Responsible for completing and maintaining IMP-1300.02, Exhibit 3: Equipment Operator Dispatch From Control Room
'
Check List.
4. Directing onsite evacuation at the Alert or lower level emergency classification based on potential hazard to non-assigned personal.
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QUESTION:
7.06.
(3,00)
-
In-accordanc'e with " Radiation Work Permit" (RWP), Procedure ADM-4110,04:
A.
LIST five (5) conditions which require the use~of a RWP.
(2.b)
'
'B.
FILL in the blank:
RWP'n with a termination time of greater-than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> snall'
'be reviewed on a daily basis.by the-(0.5)
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".. QUESTION'
7.07 (2.00'L)
Thel" Reactor Scram". Procedure ABN-3200.'01.. states that RPV water level should be maintained.between 138" and l'75" TAF. A caution in that procedure provides guidance to the operaLot if~RPV water-
~1 eve'l exceeds 180"'TAF.
A. WHAT immediate action must be taken if this water level-is exceeded?
(1,o)
B. EXPLAIN the concern with water level greater than 180"TAF (1.0)
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.! QUESTION
.7.08
.(2.25)
.1
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.In accordancelwith "TBCCW Failure Responce". Procedure'ABN-3200.20, l:1
'under WHAT three (3) conditions should the immediate actions of ll
'scraming the reactor and tripping all operating recirculation
..
pumps 'oe performed?
(2.25)'
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. : QUESTION 7.09 (2.75)
STATE the reason (s) for the following precautions in the
" Main Turbine", Procedure #315.1.
A. With the reactor " hot", do act close the vacuum breaker l
until the mechanical vacuum pump is started. (2 required)
(1.0)
l
'B.
If a turbine trip occurs while the HP turbine is being warmed up and reactor pressure is 140T being controlled by the MPR' the operator should immediately operate
,
the bypass jack to close bypase valves. (2 required)
(1.0)
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C, In no case should vacuum be pulled or steam admitted to seals if the rotor is not turning.
(0.75)
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i QTIESTION.
7.10'
'( 2. 50 )
.During refuel operations l you are. the SBO on > the bridge when.
Ja? fuel assembly, over the core, becomes: inadvertently, disengaged
,
and drops from therfuel' handling equipment. Irr accordance with
' '
" Reactor. Refueling" Procedure #205.0.:
A. WHAT'immediate ACTION is required.
'(0.5)
l:
B.
IDENTIFY two'(2) individuals required to be notified
!
immediately.
(1,0)
C.
IDENTIFY two (2) conditions in which-immediate evacuation of the area is required.
(1.0)
,
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8.
'ADNTNISTEAJIVE PROCEDURES. CONDITIONS.
Page'31
,",
MLD LIMITATIONS
.
' QUESTION 8.01
- ( 1. 5 01 -
In the course of returning the "A" Isolation Condenser system to Standby Readiness after an inadvertent system initiation, the operator observed that the iso]ation condenser. vent. valve V-14-5 did not close when the system initiated.
All other valves operated properly and the isolationicondencer was placed in standby. readiness.
WHAT. Limiting Condition (s) for Operation-is (are) in effect in accordance with. Technical Specifications and WHAT ACTION (S),
if any, are required.
NOTE: Use tihe attached sections of the Technical Specifications to answer the question.
FULLY REFERENCE all applicable sections of the Technical Specifications in your answer.
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QUESTION 8.02 (3.00)
.
The plant is. operating at 100% power, you are the GSS on shift.
During the performance of the monthly operability teut on the 'A" Standby Liquid Control (SLC) pump it is determinud that the pump is inoperable due to excessive vibration.
I NOTE:
Use the attached section of the Technical Specification L
to answer the questions.
FULLY REFERENCE all applicable sections of the Technical' Specifications in yo*2r answer.
l A..
WHAT Limiting Condition for Operation is in effect and l
WHAT ACTION (S) are required?
(1.5)
l'
B.
Later it. the shift the Chemistry tech informs you that the l
concentration of bcron in the SLC tank is below the Technical l
Specification requirements.
WHAT additional Limiting Condition for Operations and/or ACTION (S), if any, are required?
(1.5)
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[9UESTION 8.03-(2..'00)
.
For the following plant. conditions listed below,. STATE:the
,. MINIMUM' number.of RO and SRO~ licensed.' personnel that shall-be.in-the control-room in accordance with " Conduct of Operations". Procedure #106.
(0.0)
A.
Refuel mode, core: alterations'ARE NOT in progress.
B Reactor startup is in progress.
C.
100% power operations. NO control rod manipulations in progress.
D. Plant.is in' Cold Shutdown.
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..-QUESTION.
8.04 (3.001'-
DEFINE or EXPLAIN the following: terms as.they appear in Technical Specifications (3.0)
A.
Core Alteration
'B, COLD Shutdown C.
Identified Leakage D.
Power operations
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QUESTION..
8.05 (2.50)
.
A, WHAT is a Non-Substantive change to'a procedure'in accordance with " Procedure Control", Procedure #1077 (0.5)
II. WH0_can temporarily approve a Non-Substantive change?
(0.5)
C.
A temporary change (other than temporary Non-Substantive change)
that cannot be. delayed for normal review and approval can be approved by two (2) members of the GPUN management staff.
STATE two (2) requirements these staff members must meet.
(1.0)
D. TRUE or FALSE All outstanding temporary changes in a procedure shall'
automatically expire 60 days after the most recent-temporary change entered in a procedure; (0.5)
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'V
.-QUESTION 8.06 (~2.50)
t Concerning " Equipment Control", Procedure #108. INDICATE if the following statements are TRUE or FALSE.
(2.5)
l A, The Yellow Tag is an electrical equipment caution tag and is to be used as a warning of conditions other than normal.
l l
B. A Red and a Blue Tag may be attached to the same switch for
testing purposes.
C A Supplemental Log Sheet is used when new work is to be done on a system / equipment that is already isolated.
,
D.
A mechanical jumper consisting of temporary piping or a' hose connection in a piping system is defined as a temporary variation.
E.
The GSS ;uay approve the intentional defeat of an alarm circuit in the main control room and at local alarm panels.
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..,.(.
QUESTION-
-8.07.
(2.50)
.
Review the proposed licensed operator work schedule below-and IDENTIFYifive (5) deviations from "Oonduct of' Operations".
Procedure #106.
(2.5).
SAT OFF
- OFF D'AY.1 SUN 0700~~ SUN 2400 DAY 2 MON-0800 - MON 2400 DAY 3 TUES 0800 - TUES 1600 LDAY 4'
WED'
0800 - WED 1600 DAY 5 WED 2200 - THUR 0800 DAY 6 ERI 0800 - FRI -2000 DAY 7 SAT 0800 - SAT 2000 DAY 8 0FF
- 'OFF
.
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.QUFSTION 8.08 (2.00):
1 A.,
DESCRIBE the, difference between Category I: and Category'II events as defined in " Notification of a Station Event",
$
. Procedure #126.
. ( 1~.' 5 )
B; WHO is responsible for determining'.the category of an event in accordance with Procedure #126..
( 0. 5 )'
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?l- - '.' QUESTION
. S. tJ 9 -
~(l'.50)
During your review of-monthly curveillances on.the
"A" Core Spray pump,.you note-the surveillance completion dates.were.
.as follows:
i July
!~
June 10'
.May
April
. IDENTIFY.the Technical Specification'n surveillance. time interval requirement (s).that may have been exceeded.
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LQUESTION.
8.10 (3.00)
..
- I'y A.
STATE whether a SAFETY LIMIT would be violated for each of the following operating conditions. (YES or NO)
1. Reactor pressure is 750 psig, core flow is 2b% and core thermal power is 675 MWt.
(0.5)
2. Reactor pressure is 900 psig, core flow is 30% and MCPR is 1.10.
(0.5)
3. Mode switch is in SHUTDOWN and the reactor water level is 5 ft. 0-in. above top of active fuel, (0,5)
B.
DESCRIBE the three (3) ACTIONS required if a Safety Limit is violated.
INCLUDE any TIME requirements.
(1.5)
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.: QUESTION-8 11~
(1;50)
The plant-is op3 rating at 100% ateady state-power when the GSS is informed by an-operator of a. slowly increasing hydrogen concentration on the discharge of the Augmented off Gas System.
The concentration is 4.1% a t this time.
STATE the Limiting i.
l Condition for operation in effect and any actions required.
_
'
NOTE:
Use the attached-sections of the Technical Specifications to answer the question.
FULLY REFERENCE all applicable sections.
of the Technical Specifications ir. your answer.
!
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(********** END OF EXAMINATION ***'*******)
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'5.i THEOPY OF NUCLEAR POWER FLANT OPERATION.
Page'42 l~
4;
. FLUIDS AND THEEKQDYNAMICS A
l h
,.
L IANSWER 5.01-(2.00)
Due to the thermal expansion during heatup the fuel pellet could come into contact with the cladding and put stress on it.(0.75) If the stress exceeds tho yield stress of the cladding material the cladding will crack.(0.75) 1% plastic. strain on the clad.(0.5)(is the' limiting condition of concern)
REFERENCE
'LP: THERMO.HT.~and FF CH9 pg.75 OBJ: TCR 852.09 LO.N KA': 293009 K1.07 (2.8/3.6)
293009K107
..(KA's)
ANSWER 5.02 (3.00)
A. TRUE (0.75)
B. FALSE (0.75)
C. FALSE (0.75)
D. TRUE (0.75)
REFERENCE LP: REACTOR THEORY CHil pg. 5.7,9,10
.OBJ: TCR 842.11 LO. A,C,D.E KA: 2920D3 K1.06 (3.7/3.7) 292003 K1.04 (2.5/2.5)
292003K104 292003K106
.(KA's)
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'gSi o THEORY ~ OF NUCLEAR POWEF_f,1M17 OPERATION.
Page 43
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FLUIDS atJD TH,EEDODYNAMICS
,,,
.
L ANSWER l 5.03-
. 3. 0 0 ')
(
l 1.
AXIAL power distribution (0.5-)
i4I 2. BOTH (0.5)
,w 13.
RADIAL power distribution (0,5)
,
B. Results-from a one.or two notch withdrawal of a shallow rod (0.75)
where the void fraction in the top of the bundle will generally decrease the power in the region.(0.75)
OR Void related power decrease being dominant over the shallow rod withdrawal.(1.5)
REFERENCE LP: BWR Operating Charact.0PM-00 pg.
9.
OBJ: TCR 823.02 LO. B KA: 292005 K1.04 (3.5/3.5) 292005 K1.12 (2.6/2.9)
292005K112 292005K104
.(KA's)
ANSWER 5.04 (3.00)
A.
1. A difference in density in one portion of the fluid relative to the other.
(0.5)
2.
A difference in height between the regions of different density. (0.5)
3. A communication path for mass transfer between the two regions.
(0.5)
B.
Ensure water level in sufficient to cover the separator to provide a flow path between the core region and the downcomer region (0.75)
to prevent temperature stratification. (0.75)
REFERENCE LP: SDC Eys. Oper. Pro #305 pg. 9 OBJ: TCR 819.01 LO. D,E KA: 293008 K1.34 (2.9/3.1) 293008 K1.37 (3.2/3.4)
293008K137 233008K134
.(KA's)
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THE0EY OF tlU.QLEAR -POWER.2LAliL 0PJRATION.
Page 44 ELUIDS. AND TJ3EBMol>lile, MICE
. _,.
<.
I'.
l ANSWER 5 P5 (3.00)
.
1.
Vessel head bolts tensioned
'- I ' #" d d A ' f "'f3 "'N *" r er 2. Pressurizing the reactor sessel
' #' ' ^'"" e -s o T* Aornnr
<
vessel heatup
.,.
3,%.,3ct,,,, l,, 4,,,
4. Vessel cooldown 5 (L(fk raea-u too.
5.
Recirculation pump start ( h 0.6 each)
g a.2 9.s g
REFERENCE LP: THEMO.HT.and FF. CH10 pg.10-15.10-16 OBJ: TCR 861.07 LO) C.D KA: 293010 K1.04 (2.9/3.2)
293010K104
.(KA's)
ANSWER 5.06 (3.00)
(0.25 for coefficient, 0.25 for + or - reactivity, and 0.5 for each reason)
A. Void coefficient would add negative reactivity first. The decrease in pressure caused by the SRV opening would cause void production to increase.
B. Fuel temperature coefficient would add negative reactivity first.
The rapid addition of positive reactivity due to rod removal causes a rapid increase in power and fuel temperature.
C.
Moderator temperature coefficient will add positive reactivity first. The loss of feed heating will increase inlet subcooling.
REFERENCE LP: Reactor Theory LP #300.08 pg. 28,32.47 OBJ: TCR 84.2.08 LO. A KA: 292004 hl.01 (3.2/3.2) 292004 K1.05 (2.9/2.9)
292004 K1.10 (3.2/3.2)
292004K110 292004K105 292004K101
.(KA's)
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S'. 4 THEORY 5F NUCLEAR POWER F.LANT OPERATION.
Page 45 l
,'
FLUIDS.A!TD THERMODYNAMICS
'
ANSWER 5.07 (3.00)
A.
1.
INCREASE (0.5)
due to xenon buildup (0.5)
2.
DECREASE (0,5)
due to xenon decay (0.5)
l B. LESS THAN (0.5) After 3 days, the reactor is essentially xenon i
free, whereas at 10 minutes after the ceram xenon in sti]1 present and contributing negative reactivity (increasing the SDM)
(0.5)
REFERENCE LP: Reactor Theory LP #300.10 pg. 13-15 OBJ: TCR 842.10 LO. B KA: 292006 K1.07 (3.2/3.2)
292006K107
.(KA's)
ANSWER 5.08 (2.50)
A.
NO (.5) FRP=GAF*APRM (FRP=1.02+.495)=,5049 (.5)
.5049/.5:1.01 (.5)
B. Adjust the APRM gain (ratio of FRP/MFLPD) (1,0)
on
.
nomr erAm sc um eun 4.
at.c,e scen,o c g
g g
,,
,
LP:
Tech Spec pg 2.3-2 OBJ: TCR 850.90 LO..i KA:
293009K109 (3.1/3.7)
293009K109
..(KA'c)
,
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4 THEOEY OF NUCLEAR POWER PLANT nPERATION.
Page 46 ELUIDF'.. AND THERMODYNAMIC
,
.
.
ANSWER 5.09 (2.50)
A.
1.
To prevent cavitation of the recirculation pumps (0.5)
2. To prevent excessive steam formation in the lower portion of the core which could lead to inadequate coo. ling of the upper core. e s.
F uws ti. eer raea ore ( 0. 5 )
(v.nanu r. ca o ~e.aren s t n. u. or B.
h of sat. liquid at 1000 paia is 542.6 BTU /lbm (frcm steam table) (0.E)
h of liquid at 400 deg F and 1000 paia is 375.81 BTU /lbm (0.5)
542.6 - 375.81 = 166.79 BTU /lbm (+
or - 5 BTU /lbm)
(0.5)
REFERENCE LP: THMO. HT.and FF. CH4 pg. 4-23,4-24 Hot Lic Op Annual Exam Bank Item Code I-74 OBJ: TCR 852.04 LO. C,H KA: 293003 K1. 'c3 (2.8/3.1) 293008 Kl.19 -(2.6/2.8)
293008K19 293003K123
. (KA's)
l i
i l
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6. * PLANT SYSTEM 3 DESIGN. CONTROIa AND INSTRUMENTATION Page 47
.,
.
.
ANSWER 6.01 (3.00)
A.
B.
C.
D.
REFERENCE LP: OPM Mod 51 pg.51-6,51-9,51-11.51-12 GBJ: TCR 828.51 LO. C KA: 241000 K4.01 (3.8/3.8) 241000 K4.03 (3.0/3.1)
241000K403 241000K401
.(KA's)
ANSWER.
6.02 (2.50)
A.
the Master Controller (0.5)
B. Speed signal (set by the operator) (0.5) and the feedback speed signal (from the M.G. Set (Generator) tachometer).(0.5)
C. The low limiter portion of the dual limiter. (0.5)
D. T'
r"d f rdhch 1g.w l fi c-c. '5c M.C S c4,.c ( 0. 5 )
l mfW n F /t em AHEosmr f / A - 2.T-d!i'
REFERE~NCE LP:0PM Mod 38B pg 38B-10thur19 Fig 38B-5B OBJ: TCR 828,40 LO. E KA: 202002 K4.02 (3.0/3.0) 202002 '30.07 (3.6/3.0)
202002G007 202002K402
.(nA's)
l l
l l
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. 6. * PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Page 48
.
.
ANSWER 6.03 (3.00)
A.
YES. (0.5) The differential. pressure detectors respond to a differential l
L pressure (with steam flow passing in either direction)
.0.5)
B.
NO. (0.S) Each steam cupply and condensate return line has two (2)
MOV isolation valves in ceries. One valve is 480 voltc AC and the L
other is.125 volt DC. On loss of 480 volt AC power the DC valves l-will'still close. (0.5)
C. NO. (0,5) An air accumulator provides air to operate the makeup valves at least six (6) times with complete loss of Instrument and Service Air.. (0.5)
REFERENCE LP: OPM Med. 23 pg.9.10,13 OBJ: TCR 828.23 LO. I KA: 27000 K2.01 (3.8) 27000 K4.01 (4.5)
27000 K6,04 (3.3) 27000 K6.07 (3.2)
207000K607 207000K604 207000K401 207000K201
..(KA's)
ANSWER 6,04 (3.00)
1.
Permits plant operation to generate adequate steam and pressure to establi_n turbine seals and condenser vacuum at relative-low power.
(1.0)-
2.
Sufficient protection is provided by other scrams below 45%
(As detailed analysis have shown)
(1.0)
3.
Provides the ability to reset the scram (0,5) thereby-relieving scram pressure from the contral rod drives which will increase their expected lifetime.(0.5)
REFERENCE LP:
Tech Spec 3.1-3 3.1-4 OBJ: TCR 828.37 LO. A KA:
212000 K4.12 (3.9/4.1)
212000K412
..(KA's)
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$
- 36!= PLANT SYSTEMS DESIGM. CONTROL. AND INSTRUMENTATION Pago 49
.
.
ANSWER 6.05 (2.50)
A.
NO ACTION (0 5 EACH)
B.
ROD BLOCK C.
HALF SCRAM D.
ROD BLOCK E.
HALF SCRAM REFERENCE
!
LP: OPM Mod 29E pg. 29E-24,25,26, OPM Mod 29C pg
<
E.
J.
.104: 215005 K4.01 (3.7/3.7) 215005 K4.02 (4.1/4.2)
215003 K4.02 (4.0/4.0)
2iS003K402 215005K492 215005K401
..(KA's)
ANSWER 6,06 (2.00)
1y ea.u.a A.
1. High Drywell pressure (0.25)
3.5 psig (0.25)
2.
Low-Low reactor water level (0.25)
86" TAF (0.25)
.B.
To prevent short cycle of the blow down flow path upon initiation of containment spray. (0.5) If it were to open immediately upon containment spray initiation a flow path would exist from the drywell through the 5% valve to the air space of the torus, since flow had not yet developed.(.25) This would pressurice the torus and could prevent complete blow down. (0.25)
REFERENCE LP: OPM Mod 09 pg. 09-12, 09-24 OBJ: TRC 828.09 LO.
E, F
KA: 226001 K4.09 (3.2/3.4) 226001 G007 (3.5/3.5'
226001K409 226001G007
..(KA's)
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- J $L* -PLANT SYSTEMS ~ DESIGN. CONTROL. AMD I&cLTRUMENTATION Page 50
,.
.
' ANSWER 6.07 (2.00)
(0.25 each)
A.
E.
B.
F. 2 l
C.
G.
D. 3 H. 1 REFERENCE LPi OPM Mod 33 pg. 33-14,38.46,49,48
.OBJ: TCR'828.33ELO. B KA: 272000 K1.01 (3.6/3.8) 272000 K1.07 (3.0/3.2)'
272000K107 272000K101
..(KA's)
!
' ANSWER 6.08 (2.00)
'
A.
FAIL OPEN (0.5)
i B. FAIL AS IS (0.5)
C.
FAIL SHUT (0.5)
D. FAIL OPEN (0.5)
REFuRENCE LP: OPM Mod 17 pg 17-13, 17-24 OBJ: TCR 828.17 LO. E KA: 259001 K6.01 (3.0/3.0)
259001K601
..(KA's)
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6.+
PTANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTAIlON Page bl
,
.
' ANSWER 6.09 (2.60)
A.
1.
(remote manual) control awitch to the manual position (0,5)
2.
(remote manual) control switch in auto if the relief set point is reached (0.5)
3. (remote manual) control switch in auto or off if ADS requirements are met (0.5)-
B.
time delay to stagger opening of EMRV'S.which minimizes the probability of steam vent clearing damage (0.5)
C.
seal-in to either assure positive actuation of ADS or prevent EMRV'S closure-reopening event (0.5).
REFERENCE LP: OPM Mod 05 pg. 05-15. 16.. 19 OBJ:TCR 828.95 LO. C KA: 239002 K4.08 (3.6/3.7) 239002 A4.01 (4.4/4.4)
239002A401 239002K408
.(KA's)
ANSWER 6.10 (2.50)
A.
2 (0.5), 4 (0.5)
B. prevents siphoning of water from the fuel pool (0,5) in the event of a system line break below the level of the fuel pool (0.5)
C. FALSE (0.5)
REFERENCE LP: OPM Mod 20 gp. 20-9,20-12,20-14 OBJ: TCR 828.20 LO.
H.
D KA: 233000 K3,02 (3.1/3.2) 233000 A1.02 (2.9/3.1)
233000A102 233000K302
.(KA's)
l (*****
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,i7.*
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 52 L
AND R6DIOLOGICAL CONTROL
,
,
ANSWER 7.01 (3.00)
A.
To assure adeqate core cooling under plant conditions where RPV water level cannot be determined.
(1.0)
B.
1.
Water level covering containment vents and subsequent loss of preesure control (0.5)
2.
Structural strength of the torus under the weight of the water (0.5)
C.
The lowest differential pressure between the RPV and the suppreccion chamber (0.5) at which steam flow through the EMRV'S is cufficient to remove all decay heat generated within the core (0.5)
REFERENCE LP:
EMO-3200.08 Handout 87.17 pg.4 OBJ: TCR 845.18 LO A,B KA:
295031G011-(4.4/4.6) 295029K101 (3.4/3.7)
295029K101 295031G011
.(KA's)
ANSWER 7.02 (3.00)
A.
1. RPV Control (EMO-3200.01) (.375)
2. Primary Containment Control (EMO-3200.02)(.375)
B) FALSE (0.75)
C) 13 mr/ hour (0.75)
D) TRUE (0.75)
REFERENCE LP:
EMO-3200.01(.02)(.03) pg 3 OBJ: HLPTC 845.03 LO.G 845.06 LO.F 845.11 LO.F FA:
295034G011( 4. '/4. 3 ) 295024G011(4.3/4.5) 295026G011(4.4/4.6)
295025G011( 4. 2 /4. 3 )
295025G011 2950L6G011 295024G011 r.95034G011 (KA's)
ANSWER 7.03 (2.00)
A.
1. Scram (.25) on low Rx water level (.25)
2. Feed pumps could go to runout (.5) causing a feedwater transient that yields a high level turbine trip and subsequent scram (.5)
B. Condensate Transfer system (0.5)
i (*****
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PROCEDURES'- NORMAL. ABilQHMAL. EVERGENCY Pase 53
.'
6ND RADIOLOGICAL CONTRQL
.
REFERENCE LP:
Annual exam bank item code 4-56 RWCU Demin system op 303 pg 12.18'
OBJ: HLTCH 828.39 LO.0 KA:
204000K302(3.1/3.1) 204000G013(3.3/3.1)
204000G015 204000K302
..(KA's)
ANSWER 7.04 (2.50)
A.
1.
Fire Water 2. Core Spray Keep Fill-(Condensate Transfer)
3.
SLC (test ~ tank or poison tank) (.5 each).
- ]B. If RPV water level cannot be restored and maintained in the normal band. (1.0) oA i r v. m y ecn.%
e m o s va re m
.~s t.,Lc o w r
'
fon o oske n.% w o +u s, usasr one t"mr l
REFERENCE go g gle s.
g
,1,yp LP:
EMG 3200.01 step RC/L-4 OBJ:
KA:
295031EA108 (3.9)
295031A108
..(KA's)
ANSWER 7.05 (2.00)
A.
(0,5)
B.
(0,5)
C.
(0.5)
D.
(0,5)
. REFERENCE LP: IMP-1300.01 pg.
4.
IMP-1300.02 pg.3, 5 OBJ:
KA: 294001 A1.16 (4.7)
294001A116
.(KA's)
..
l l
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7 PROCEDURES - NORMAL. AHti&RMAL. EtiERGENCY Page 54
,
- AMD RADIOLOGICAL COliTFOL
.-
.
' ANSWER 7.06 (3.00)
(any five (5) at 0.5 each)
A.
1.-entering a radiation area 2. entering a high radiation area 3. enterimg a contaminated area 4. entering an airborne radioactivity area
-
5. unknown radiological conditions in an area, equipment or system to be entered or opened 6. handling exposed sources in excess of 10 times the quantity specified in Appendix C of-10 CFR 20 7.
transfer of radioactive material out side of the protected area B. GRCS (Group Radiological Control Supervicor)
(0.5)
REFERENCE LP: RWP ADM-4110.04 pg.
2, 3,
OBJ: TCR.838.14 LO. D KA: 294001 K 1.03 (3.8)
294001K103
..(KA's)
ANSWER 7.07.
(2.00)
A.
Place the control switches for the Condensate Return Isolation valves (V-14-34 and 35) in the closed position (0.5)
(unitl reactor water level is below 180"TAF.)L11.t9.A (0.5)
B.
(Severe) Water hammer of the isolation condenser may occur.
(1.0)
REFERENCE LP: ABN-3200.01 pg.4 OBJ: TCR 801.01 LO.A.I. C KA: 20700 A1.06 (3.5/3.7) 295006 A2.03 (4.0/4.2)
295018K101
..(KA's)
ANSWER 7.08 (2.25)
1. no TBCCW pump is running or can be started (0 75)
2.
t major TBCCW pipe break cannot be isolated (0.75)
3.
less than four (4) recirculation pumps are running (0.75)
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, 7 H PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page'55-
.'
..
AND RADIOLOGICAL _SQL{ TROL
4-s REFERENCE LP: ABN 3000.20 pg 31
'
OBJ: TCR 801.01 LO.
A.8.
KA: 295018 K1.'01 (3.5/3.6) 295018 K2.'02 (3.4/3.6)
295018K202 295018K101
..(KA's)
ANSWER 7.09:
(2.75)
A.
Positive pressure in'the condenser (0.5) and rupture of the turbine relief diaphragms.may result.(0.5)
B. This will prevent. inadvertent depressurization of the Reactor.(!.5) and possibly a high reactor water level
condition (due to swell)-(0.5)
C.
This will renult-in a bowed rotor.
(0.75)
REFERENCE LP:
Procedure 315.1 pg 7.
8,
OBJ: TCR 828.50 LO. N-KA:
245000 KS.02 (2.8/3.1) 241000 K3.02 (4.2/4.3)
245000K502 241000K302
.(KA's)
ANSWER 7.10 (2.50)
A.
Immediately cease all refuel operations (0.5).
B. Notify the Group Shift Supervisor (0.5) and Manager Plant Operations.(0,5)
C.
If radiation levels begin to increase (0.5) or' inadvertent criticality is suspected (0.5)
REFERENCE LP: Pro #205.0 pg.21 OBJ: TCR 812.02 KA: 295023 K2.03 (3.4/3.6) 295023 K3.01 3.6/4.3)
. 295023K301 295023K203 (KA's)
L
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.,' 87* ADMJNISTEATIVE EROCEDURES. CONDITIONS Page 56 AND LIMITATIQ!iS.
.
.
ANSWER'
8.01 (1.50)
The valve is located on Table 3.5.2 Containment Isolation Valves'(.25)
T/S 3.5.3.a With one or more of the containment isolation valves shown in Table 3.5.2 inoperable ( 25)
Maintain at least one icolation valve operable in each affected penetration that is.open(.25) and within-4 hours either; 1)
Restore the inoperable valve (s) to operable status (.25)
OR 2)
Isolate each affected penetration by use of at least ene deactivated automatic valve secured in the isolation position (.25)
OR 3)
Isolate each affected penetration by use of at least one closed manual valve or blind flange (.25)
BEFEEENCE LP:
T/S pg. 3.5-2,3,13 and 3.8-1 OBJ: TCR 850.90 LO.K KA:
207000G011(3.7/4.2)
207000G011
.(KA's)
ANSWER 8.02 (3.00)
A.
T/SR3.2.C.3 If one SLC system pumping circuit becomes inoperable during the run mode and specification 3.2.A is met the reactor may remain in operation for a period not to exceed 7 days (0.75) provided the pump in the other circuit is demonstrated daily to be operable (0.75)
B..T/S 3.2.C.2 The SLC solution shall be maintained within the volume-concentration requirement area of Fig 3.2-1...when the SLC system is required to be operable (0.5).
Since it is not within the limits, T/S 3.0.A applies. (0.5) The unit shall be placed in Cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective actions are completed (0.5)
REFERENCE LP:
T/S pg 3.2-2,3 OBJ: TCR 850.90 LO.J KA:
211000G011(3.4/4.1)
211000G011
..(KA's)
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i I
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g 8?*'ADMUllSTRATIVE PFOCEDURES. CnNDITIONS.
Page 57 a-tdl.D L I M I T A T I O N S.
,.
,
Ai;SWER 8.03 (2.'00)
,
A.
1 RO (0,5)
B.
C.
1 SRO 1 80 (0,5)
D.
1 RO (0.5)
REFERENCE LP:
Procedure 106 OBJ: TCR 830.05 LO. D.D.
KA:
294001A109(4.2)
.294001A109
..(KA's)
ANSWER 8.04 (3.00)
A. A core alteration is the addition. removal, relocation or oth'er manual movement ofLfuel or controls in the reactor core. ( 75)
B. The reactor is at cold shutdown when the mode saitch is in the shutdown mode position (.25) (there is fuel in the reactor vessel) all operable control rods a fully inserted (.25) and the reactor coolant system maintained at less than 212 F and vented (.25)
C.
Identified Leakage is that leakage which is collected in the primary containment equipment drain tank (. 5) (and eventually transferred to radwaste for processing.) (.25)
I a.4 gg D.
Power operations is any operation when the reactor is in startup mode (.25) or run mode (.25) except when primary containment integrity is-not required (.25)
REFERENCE LP:
T/S pg 1.0-1,2,4 OBJ: TCR 850.90 LO B KA:
212000A216(4.0/4.1)
212000A216
.(KA's)
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8.*
ADMINISTRATIVE PROCEDUREF. CONDITIONS Page 58
'.-
-AND LIMITATIONS
.
ANSWER 8.05 (2.50)
A. clerical changes, typographical error and other similar items (0.5)
B.
A person holding an RO or SRO license for OCNGS (0.5)
C.
1. Qualified responsible Technical Reviewer 2. At least one of these individuals shall be the on-shift GSS or SRO licensed GOS 3.
Must be knowledgeable in the area or category affected by the procedure.
(Any two (2) at
.5 each)
D.
TRUE (0.5)
REFERENCE LP:
Procedure 107 pg 24,26,27 OBJ: TCR 850.05 LO. MM.HN KA:
294001A103(3.7)
294001A103
.(KA's)
ANSWER 8.06 (2.50)
A. TRUE (0.5)
B. FALSE (0.5)
C. FALSE (0.5)
D.
TRUE (0,5)
E. TRUE (0.5)
REFERENCE LP:
Procedure 108 pg. 21,26,50.58,88 OBJ: TCR 830.05 KA:
294001K102(4.5)
294001K102
.(KA's)
ANSWER 8.07 (2.50)
1.
DAY 1 > 16 Hrs straight
?..
DAY 1 > 16 Hrs in any 24 Hr period 3.
DAY 1&2 > 24 Hrs in a 48 Hr period 4.
DAY 4&S < 8 Hr between work periods 5. > 72 Hrs in a 7 day period Ay46 @
.5 each)
(,. 0A y 4+f > /c, #d.s sk Aay A4 97t.
g6tt,oo 7.
Day 3 otoo ro wwt * *** > x < /ys no 2 no e sa v is** ra s nw/"!"4 i
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8. ADMINT STRATIVE PJLOCEDUPJS. ' CONDITIONS.
Page 59~
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t AND LIMITATIONS
.
,
L REFERENCE i
LP:
Procedure 106 pg 68 OBJ: TCR 830.5 LO FF KA:
294001A103 (3,7)
294001A103-
.. (' K A ' s )
ANSWER 8.08 (2.00)
A.
Cat I includes all events which result in declaration of an unusual event, alert, site area emergency or'a general emergency under the emergency plan.
(.75)
Cat II includes events not classified a Cat I which require notification of the NRC and others within one (1) hour of (, z g,, m y 7;,,,, y,,;r, m (.75)
occurence.
munkso )
B.
Group Shift Supervinor (0.5)
REFERENCE LP:
Procedure 126 pg 4,8 OBJ: TCR 830.06 LO. XX,YY i
KA:
294001A103 (3.7)
l 294001A103
.(KA's)
ANSWER 8.09 (1.50)
The maximum allowable extension has exceeded 25% (June)( 75)
The total maximum combined interval has exceeded 3.25 times the specified surveillance interval. (May, June. July) (.75)
REFERENCE LP:
T/S 1.24 pg. 1.0-5 OBJ: TCR 850.9 LO.K
'KA:
294001A103 (3.7)
294001A103
.(KA's)
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, 8. * ADMINISTRATIVE PROCEDURES. CONDITIONS.
Page'60 i
AND LIMITATIONS
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a l
l ANSWER 8.10'
(3.00)
A.
1. YES (0.5)
2.
NO (0.5)
3.
NO
~( 0. 5 )
B.
1.
Immediately shutdown the reactor (0.5)
2.
Report the violation to the NBC (.25) and Director Oyster Creek (.25)
3.
Prepare a Safety Limit Violation Report (.25) and submit to the
!
VP and Director Oyster Creek and to the NRC within 10 days (.25)
REFERENCE LP:
T/S pg 2.3-5 thru 7 OBJ: TCR 850.90, LO D.
KA:
212000SG6(3.4/4.3)
212000SG6
..(KA's)
l ANSWER 8.11 (1.50)
T/S 3.6.F.2 Condenser Off Gas Hydrogen Concentration (0.5)
In the event the H2 concentration exceeds 4% by volume, the concentration shall be reduced to less than 4% with 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (0.5.) OB be in at leant SHUTDOWN CONDITIONS OR within the limit within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.(0.5)
REFERENCE LP:
T/S 3.6.F
!
OBJ:
TCR 850.90 LO.K l
KA:
271000G011 (2.9/3.9)
271000G011
.(KA's)
l I
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END OF EXAMINATION
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TEST CROSS. REFERENCE
'Page. 'l
,
.
<QUESTIONL VALUE.
REFERENCE A
5.01-2.00
'ZZ Z0000001 '
5.02, 3.00.
ZZZ0000002
'
.5.03.
3.00 ZZZ0000003-5.04'
3.' 0 0 -
ZZZ0000004 5.05'
3.'00 ZZZ0000005 5.06 3.00~
ZZZ0000006 5.07
.3.00 2720000007 5.08-2.50 ZZZ0000008 5.09 2.50.
ZZZ0000009
______
25.00 6.01 3.00 ZZZ0000010 6.02 2.50 ZZZ0000011 6.03-3;.00 ZZZ0000012 6.04 3.00 ZZZ0000013 6.05 2.50 ZZZ0000014 6.06 2.00 ZZZ0000015 6.07-2.00 ZZZ0000016 6.08 2.00 ZZZ0000017 6.09:
.2.50
.ZZZ0000018 6.10 2.50 ZZZ0000019
______
25.00 7.01'
3.00 ZZZ0000020 7.02 3.00 ZZZ0000021'
~7.03 2.00 ZZZ0000022 7.04 2.50 ZZZ0000023 7.05 2.00 ZZZ0000024 7.06 3.00'
ZZZ0000025 7.07 2.00 ZZZ0000026 7.08 2.25 ZZZ0000027 7.09-2.75 ZZZ0000028 7.10 2.50 ZZZ0000029
______
25.00 8.01 1.50 ZZZ0000030 8.02 3.00 ZZZ0000031 8.03 2.00 ZZZ0000032
~8.04 3.00 ZZZ0000033 8.05 2.50 ZZZ0000034 8.06 2.50 ZZZ0000035 8.07 2.50 ZZZ0000036 8.08 2.00 ZZZ0000037 8.09:
1.50.
ZZZ0000038
.
8.10 3.00 ZZZ0000039 8.11-1.50 ZZZ0000040
______
25.00
______
______
100.0 (
i l1 -.