IR 05000410/1987018
| ML20235Q970 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/01/1987 |
| From: | Kolnauski L, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235Q933 | List: |
| References | |
| 50-410-87-18OL, NUDOCS 8710070754 | |
| Download: ML20235Q970 (88) | |
Text
{{#Wiki_filter:_ _ _ _ - _.. n .t U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING COMBINED REQUALIFICATION AND REPLACEMENT EXAMINATION REPORT REPORT NO.: 87-38 (OL) FACILITY DOCKET NO.: 50-410 FACILITY LICENSE NO. : NPF-69 LICENSEE: Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 FACILITY: Nine Mile Point Unit 2 EXAMINATION DATES: July 7-9, 1987 ' /o - /-17 CHIEF EXAMINER: ~ aus i, Reactor Engineer (Examiner) Date y ynn Kol 4p4f.
/8-/I7 ' APPROVED BY: - . ' David LAnge, ting Chiatt, Boiling Water Date Reactor (BWR) Section Division of Reactor Safety SUMMARY: Requalification written and operating examinations were administered to four (4) senior reactor operators (SR0s) and four (4) reactor operators (R0s).
Two (2) SR0s and three (3) R0s passed these examinations.
ES-601, " Administration of Requalification Program Evaluation," of NUREG 1021, the " Examiner Standards," establishes a minimum evaluation level of twelve (12) licensed operators in order to issue a final requalification program evaluation. Additional operators will be evaluated in the future; the NRC will issue a final requalification program evaluation at that time.
A replacement operating examination was administered to one (1) senior reactor operator candidate who had passed a previous NRC written examination.
The candidate passed the operating examination.
8710070754 871001 DR ADOCK 0500
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ . . DETAILS i TYPE DF EXAMINATIONS: Requalification ] EXAMINATION RESULTS:
1 I R0 l SR0 l l Pass / Fail l Pass / Fail [ l l l l l I l l Written I 3/1
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I I I Operating l 4/0
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I I I l Overall I 3/1 l 2 / 2 l l
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CHIEF EXAMINER AT SITE: L. Kolonauski (NRC) 2.
OTHER EXAMINERS: D. Jarrell (PNL) W. Cliff (PNL) D. Moon (PNL) 3.
The following is a summary of generic strengths or deficiencies noted on operatinj tests.
This information is being provided to aid the licensee in upgrading license and requalification training programs.
No licensee response is required.
STRENGTHS Most operators were very familiar with plant equipment locations.
DEFICIENCIES a.
Several operators were weak in providing effective communications during the simulator examinations, b.
Weaknesses were also noted in the areas of APRM calibration and new fuel inspection processes.
. c.
Several operators were unf amiliar with the applicability of various j fire extinguishers to the established fire classifications.
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-' I 4.
The following is a summary of generic strengths or deficiencies noted from .the grading of written examinations..This information is being provided to aid the licensee in. upgra' ding license and requalification training programs. No licensee response is required.
While individual strengths and deficiencies were noted from the grading of the written examinations, none were noted as generic to the group.
5.
Comments on availability of, and candidate familiarization with plant reference material available in the control room.
The applicable plant reference material was readily available in the main control room and the operators under evaluation were adequately familiar with its location and use.
6.
Simulation Facility Fidelity Report: During the conduct of the simulator portion of these operating examina-tions, the following performance and/or human facters discrepancies were observed: a.
The operating procedures generally refer to indications by gauge number; several simulator indications are not labeled with these identification numbers.
i ' b.
In order to allow the Standby Liquid Control system to inject, the High Pressure Core Spray pump must be in pull-to-lock.
c.
Inaccuracies were noted in the model for Suppression Pool level response.
d.
The operators noted other inaccuracies in plant indications during the simulator examinations. These discrepancies can lead to negative training; the operators may recognize the simulator fault and take - actions based on what they expect the questionable indication to be and in doing so, choose to ignore their indications.
7.
Personnel Present at Exit Interview: . l NRC Personnel ! Lynn Kolonauski, Reactor Engineer (Examiner) Charles Marschall, Resident Inspector NRC Contractor Personnel William Cliff, PNL i
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. Facility Personnel James Burton, NMP Unit 2 Operations Training Karl Zollitsch, Training Supervisor-Nuclear Randall Seifried, Assistant Training Supervisor-Nuclear Michael Jones, Superintendent of Operations-NMP 2 8.
Summary of NRC comments made at exit interview: The Chief Examiner noted that ES-601, " Administration of Requalification Program Evaluation," of NUREG 1021, the " Examiner Standards," establishes a minimum evaluation level of twelve (12) licensed operators in order to issue a final requalification program evaluation.
For this reason, the NRC will not issue a final requalification program evaluation until addi-tional operators are evaluated, i The comments listed in paragraphs 3 and 6 were discussed.
The examiners experienced delays in plant access and dosimetry issuance during the first day of operating examinations.
The examiners experienced delays in preparing for the examinations because the facility reference material was incomplete.
As a result, the examiners made four separate requests for additional reference material prior to the examinations.
Two (2) Senior Reactor Operators were identified as potential failures on the requalification operating examinations; their individual performance deficiencies were discussed. The Chief Examiner requested the licensee to take immediate actions in accordance with procedure NTP-11 which describes the Nine Mile Point 2 licensed operator fequalification program.
The Chief Examiner stated that the written examination results would be reported to the Training Manager by telephone once the regional review of the examination grading was completed.
The Chief Examiner stated that the NRC would conduct comparison grading of the written requalification examinations administered by the facility, and requested copies of the ungraded written examinations.
! 9.
Summary of facility comments and commitments made at exit interview: The NMP-2 Operator Training Department stated that efforts were ongoing to improve the simulator deficiencies.
The licensee apologized for the condition of the training material sent to the examiners and stated that in the future, more extensive reviews would be conducted prior to its issuance to the examiners.
The licensee indicated that the examiners conducted the evaluation in a professional manner.
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. 10. NRC Follow-up: The NRC requested a written report stating the corrective actions taken in accordance with procedure NTP-11 as a result of the unsatisfactory performances identified by the NRC examiners during the operating examinations.
j Attachments: 1.
Written Examination and Answer Key (RO) 2.
Written Examination and Answer Key (SRO) 3.
Facility Comments on Written Examinations after Facility Review 4.
NRC Response to Facility Comments i
_ _ - -. _ _ _ _ .. I NRC.Officiel Use Only Y ChWL%$_'\\ \\[7Q.-Q , . .- l ~
.. Nuclear Regulatory Commission Operator Licensing Examination . / i l
I i i ( This document is removed f rom Official Use Only rategory on
date of examination.
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! NRC Official Use On1y J , ! - - _ _. _-____-____A
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NUCLEAR-REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION , . - _ E MILE POINT 2 NI_N
FACILITY: REACTOR TYPE: _BWR-GES___________ _ __ DATE ADMINISTERED: _g7/97207 __________
EXAMINER: _MggN _g.
t CANDIDATE: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ INSTRUCllgNS_lg_CgNglggIE1 Read the attached instruction page carefully.
This examination replaces the current cycle facility administered requalification examination.
Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in.cach. category and a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
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-- 1.
PRINCIPLES OF NUCLEAR POWER 19t99__ _39199 _____ _____ ________ PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 19t99__ _Z9 99 ________ 2.
PLANT DESIGN INCLUDING SAFETY ___________ AND EMERGENCY SYSTEMS 19199__ _29 99 ________ 3.
INSTRUMENTS AND CONTROLS ___________ 19 99__ _29 99 ________ 4.
PROCEDURES - NORMAL, ABNORMAL, ___________ EMERGENCY AND RADIOLOGICAL CONTROL $9t99__ ________% Totals ___________ Final Grade All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature
1 _________________________J
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NRC RULES AND SUIDELINES FOR LICENSE EXAMINATIONS . During the administration of' this examination the f ollowing rules apply: L - 1.
Cheating on the examination means an automatic denial of your application ' and could result in.more severe penalties.
, l 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 gnly:ta facilitate legible reproductions.
, 4.
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).
E 6J Use onl y the paper provided f or answers.
7.
Print your name in the upper right-hand corner of the first page of each section of-the answer sheet.
l. 8.
Consecutivelygnumber each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne 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, 6.3.
10. Skip at least th ee lines between each answer.
t 11. Separate answer sheets from pad and place finished answer sheets face down on your cesk or table.
12.; Use abbreviations only if they are commonly used in f acility litetature.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
i.
l 14. Show all calculations, methods, or assumptions used to obtain an answer - to mathematical problems whether indicated in the question or not.
15. Partial credit may be gi ven.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE nn.' 'N9WER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
. 17. You must sign the statement on the cover sheet that indicates that the l work is your own and you have not received or been given assistance in ' . completing the examination.
This must be done after the examination has i ? been completed.
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, . id. When you comp 1Ete your examination) you shall: , c.
Assemble your. examination as follows: . a e (1) Exam questi;ns on top.
o < (2) Exam aids - figures, tables, etc.
(3) Answer pages including iigures 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'.; i d.
Leave the examination area, as defir*ed by the examiner.
If'after l eavi ng, you are f ound in thi s,,arep %while the examination is still in progress, your license may be denied or revoked.
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, _ _ -. -_ _ ._. _ _ _ _ _. .__ _ __ _ _ . _ _ !TD - P-l'.?' PRINCIPLES'DF NUCLEAR POWER PLANT OPERATION PAGEL
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THERMODYNAMICS _ HEAT; TRANSFER _AND_FtUID_FLOWL t
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,: GUESTION' 1.01 (1.00) i . ANSWER,each of the following TRUE-or FALSE.
c.
Control rod worth incr eases as moderator temperature (0.5) increapes from 150 to 500 degrees F.
L . b.1-Controi rod worth generally increases with core age.
.(O.5) .n s
' QUESTION 1.02 (1.00) - Reactor power increased'with a constant period from 1000 cps to-1500 cps in 30 seconds.
HOW much time is required for reactor ,e power to reach 3000 cps if the same constant period is maintained? (1.0)
- i RUETTION 1.03 ( 1. 50 ).
'The' reactor has. been operating 'at f ull power for one month when a reactor. scram occurs.
The pl ant remains shutdown. f or several e days.
'I DRAW a plot of xenon reactivity versus time.
Be sure to LABEL each axis and INCLUDE units and magnitude.
(1.5) ' QUESTION 1.04 (1.50) ANSWER each of the following TRUE or FALSE.
a.
Steady state critical power increases with increasing core 4, flow.
(0.5) b.
Steady state critical power decreases with increasing reactor pressure above 800 psig.
(0.5) ! n.
. Steady state critical power increases with increasing inlet , p/ c.
- subcooling.
(0.5) g u , - & (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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l;, 1,. 2 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION t L IHE8@QDyN8dlCS,_UE01_lBONSEEB_8ND_ELUID_ELQW ., - , . (- QUESTION 1.05 (2.00) L With respect to each of the three BWR thermal limits (LHGR, APLHGR and CPR): a.
ST ATE the limiting condition each limit is designed to prevent.
(1.5) b.
STATE the consequence (f ailure mechanism) of exceeding a thermal limit.
(0.5) GUESTION 1.06 .(2.00) The reactor is operating.at 75 percent power on the 100 percent rod line.
Reactor recirculation flow is subsequently increased by 5 percent.
DESCRIBE HOW and WHY the following parameters change during-the ensuing transient and subsequent return to steady state operations a.
core void fraction-(1.0) b.
core net reactivity (1.0) QUESTION 1.07 (2.00) a.
During a rapid power increase, very short reactor periods can-be maintained, yet for rapid power decreases, the stable reactor period is limited to -80 seconds.
EXPLAIN the reason for the difference.
(1.0) b.
During the first minute f ollowing a reactor scram f rom f ull power, actual thermal power decreases slower than indicated neutron power level.
WHAT are two (2) reasons for this difference? (1.0) (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) - _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
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.. . i QUESTION 1.08 (2.00) STATE HOW the reactor recirculation pump AVAILABLE NPSH changes (INCREASES, DECREASES or REMAINS THE SAME) for each of the following: c.
Reactor water level decreases from the normal level to just above the low level scram setpoint.
(0,5) b.
Feedwater heating is lost.
(0.5) c.
Reactor pressure increases during reactor startup.
(0.5) d.
Reactor coolant temperature increases from 100 degrees F to 200 degrees F during reactor startup.
(0.5) ' DUESTION 1.09 (2.00) STATE whether the following conditions would INCREASE, DECREASE, or NOT CHANGE overall plant thermodynamic efficiency.
ASSUME the plant is at 80 percent power.
a.
loss of one high pressure feedwater heater.
(0.5) b.
reduction in control rod drive cooling water flow by 50 percent.
(0,5) c.
a ten percent mass flow tube leak in the reactor water cleanup system regenerative heat exchanger.
(0.5) d.
operating with reactor water level above the high end of the normal operating band.
(0,5) (***** END OF CATEGORY 01
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1' l 2 _* PLANT DE@IGN_Itjtjt.UDING_gAFETY_AND_ EMERGENCY _@Y@TEMg PAGE
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. I . QUESTION 2.01 (2.00) While the plant is operating at full power, one SRV inadver-tently opens.
LIST f our (4) key observable PLANT PARAMETERS that would change as a result of the SRV being open.
Do NOT list valve position indication lights, acoustic monitoring instrumentation, SRV tailpipe temperature or computer alarms.
(2.0) l l ! QUESTION 2.02 (1.50) LIST three (3) automatic ini tiation signals f or the Standby Gas Treatment System.
Setpoints not required.
(1.5) QUESTION 2.03 (1.50) l The reactor is operating at 20 percent power.
An instrument technician performing a surveillance test, inserts a triple-low reactor water level signal in CSL logic channel A.
Inadvertently, a high drywell pressure signal is simultaneously inserted in CSL logic channel E.
a.
WILL the low pressure core spray system receive an auto-matic initiation signal? (0.5) b.
STATE the reason for your answer in part a.
(1.0) l QUESTION 2.04 (2.00) WHAT components in the plant design limit the consequences of j each of the following: j
a.
a rod drop accident (0.5) b.
over voltage on an RPS bus (0.5) c.
rupture of a main steam line outside the containment (0.5) ) d.
rupture of a reactor recirculation suction pipe close to the reactor vessel wall (0.5) (***** CATEGORY O2 CONTINUED ON NEXT PAGE
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t ' . ) . QUESTION 2.05 (1.50) The Residual Heat Removal System drywell spray valves (MOV-15A/D cnd NOV-25A/B) cannot be opened unless three conditions are met.
! LIST these three (3) conditions.
(1.5) i QUESTION 2.06 (2.50) A loss of all offsite power has just occurred.
Concerning the Division I standby diesel generator: a.
WHAT event initiates the load sequencer? (0.5) b.
WHAT are the first three (3) major loads sequenced to the bus? IDENTIFY the ORDER in which they are sequenced.
(1.0) .. c.
STATE two (2) conditions that will trip the diesel generator when a LOCA signal is present.
(Do not include manual actions.)
(1.0) l l QUESTION 2.07 (2.50) LIST six (6) system components that receive DIRECT signals and the TYPE of signal (open, close, start, stop) received, when the l High Pressure Core Spray system initiation logic is actuated.
(2.5) i
QUESTION 2.00 (1.50) DESCRIDE the function of the Reactor Protection System i backup scram valves during an RPS trip.
INCLUDE in your ' discussion the two (2) ways the valves provide a backup scram function.
(1.5) (***** END OF CATEGORY O2
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'13 _11NSTRUMENTS_AND CONTROLS
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a,. QUESTION'- 3.01-(1.50) , LIST all three (3) conditions that will cause the Source Range Monitoring System to initiate a rod block.
INCLUDE setpoints.
(1.5) DUESTION 3.02 (2. 00)- LIST the five (5) automatic reactor scram signals that are NEVER bypassed.
(2.0) l DUESTION 3.03.
(2.00) STATE the conditions that are necessary to cause a Standby l Liquid Control System automatic initiation with NO operator action.
INCLUDE setpoints.
Four (4) items required.
(2.0) 1' QUESTION 3.04 (1.00) l WHAT two (2) effects does high drywell pressure (1.68 psig) have l on the following reactor recirculation flow control system l components? (1.0) a.
Flow control valve b.
Loop flow controllers , ! l QUESTION 3.05 (1.00) ANSWER each of the f ollowing TRUE or FALSE.
a.
If ADS is manually initiated and all initiation condi-
tions are met, the ADS SRV opening will be delayed by l the 105-second timer to allow time for the CSH system s to try to recover vessel water level.
(0.5) i l \\ d l b.
If an ADS blowdown is in progress, with all the initiation signals present, repositioning the ADS Initiation Inhibit switches will reinitialize the 105-second timer and close all ADS SRV's.
(0.5) l I I l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) l ___
- _ _ _ - _ _ - - _ _. - '3_. __* IN_S__TRUMENTS AND CON _TR_OL_S_ PAGE B- _ _ _-__ -- - - - - - - - - - - __ ,. A F QUESTION '3.06- . ( 3. 00 ). The reactor is operating at. LOO percent power.
The EHC system status is as follows: Load Limit - 105 percent Load Set Selector - 1215 MW Maximum Combined Flow Limiter - 105 percent ~ Pressure regulator 'A' in control Using the attached EHC block diagram, EXPLAIN WHAT would happen to control valve position, bypass valve position, and reactor pressure in each of the following separate cases and DRIEFLY EXPLAIN the reason for your answer: a.
Load Limit potentiometer is reduced to 90 percent.
(0.75) I i b.
Maximum Combined Flow Limiter potentiometer is reduced to 90 percent.
(0.75) c.
Pressure regulator ' A' L f ail s law.
(0.75) d.
Both' turbine speed sensors fail.
(0.75) QUESTION 3.07 (1.50) ' EXPLAIN the response of indicated narrow range reactor vessel ' water level if containment air temperature increases from 130 to 200 degrees F, assuming no actual change in vessel water J level.
INCLUDE HOW and WHY it responds the way you indicate.
(1.5) ! QUESTION 3.08 (1.00) a.
WHICH water level instrument RANGE provides the triple-low water level trip inputs to the RPS logic? (0,5) b.
WHICH water level instrument RANGE provides the low water j level inputs for automatic initiation of ECCS systems? (0.5) ] l . I i (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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! .,. QUESTION 3.09 (2.00)- le.. uThe reactor is operating at 70 percent power'.. STATE whether-reactor water level will INCREASE, DECREASE, s i or REMAIN THE SAME during each of the following separate malfunctions.
1.
loss of feedwater flow signal (0.5) 2.. loss of. steam flow signal (0.5) 3.
reactor water-level signal fails high.
(0,5) b.
Wi th NO operator action, WHICH of the above malfunctions will result in a reacter scram? (0,5) ! - t I I' (***** END OF CATEGORY 03
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- DUESTION 4.01 (2.50)
l Use the figure containing a section of OP-OU, "Feedwater System" to
cnswer the f ollowing questions.
! J a.
Given the indicated alarm condition, with the plant operating at full power, what TWO (2) automatic responses should have occurred ! as a direct result of this alarm? (1.0) ) b.
What is the significance of having a computer point print out at ] both 210 psig and 190 pei g? (1.0) c.
If feedwater pump low suction pressure occurs AFTER a MAIN turbine trip, how will the Feedwater Level Control val ves (2FWS-LV 10A,B,C)
respond? (0.5) DUESTION 4.02 (1.00) FILL in the blanks to complete the f ollowing Saf ety Limits.
a.
Thermal power shall not exceed ______ percent of RATED ' THERMAL POWER with the reactor vessel steam dome pressure less than ______ psig or core flow less than ______ percent of rated core flow.
(0.75) b.
The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed ______ psig.
(0.25) DUESTICN 4.03 (1.00) Procedure N2-OP-lOIC, " Plant Shutdown" cautions that if Shutdown Cooling is not available during cold shutdown condition, reactor water level should be raised to between 227 and 243 inches on the shutdown range level indicator.
WHAT is the purpose of raising the water l evel ? (1.0) l DUESTION 4.04 (1.00) Under WHAT two (2) conditions does the CSO have the responsibility and authority to shutdown the reactor? (1.0)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE
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t 80DIDL991GOL_GQNIBOL .. . QUESTION 4.05 (1.50) Shortly bef ore shif t turnover you observe the following parameters end associated values: Outside ambient temperature 99.6 degrees F Drywell ambient air temperature 137.2 degrees F Service water inlet temperature 82.7 degrees F Suppression pool water level 198.6 ft Suppression pool water temperature 88.9 degrees F-a.
WHICH of the above parameters and values are entry condi-tions for at least one Emergency Operating Procedure? (0.5) b.
WHICH Emergency Operating Procedure (s) must be entered? (1.0) QUESTION 4.06 (2.00) L LIST the entry conditions f or N2-EOP-SCL, " Reactor Building Level Control."
(2.0) i I QUESTION 4.07 (1.00) i ' Several minutes after an instrument technician was given permission to perform troubleshooting in an RPG panel, the reactor operator observes a one quarter scram condition in { which several control rods become mispositioned.
WHAT two . ' (2) actions should the reactor operator take? (1.0) i l I DUESTION 4.08 (2.00) a.
WHAT are the IOCFR2O maximum allowable exposure limits
I (whole body, extremity, and skin) for a radiation worker who has an NRC Form 4 on file? (1.0) ,I b.
According to S-RP-1, " Access and Radiological Control," WHAT are the NMPC conservative guides f or maximum whole ! body radiation exposure in a calendar week AND in a j calendar year? (1.0) ] i l !
(***** CATEGORY 04 CONTINUED ON NEXT PAGE
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_ _ , > . QUESTION 4.09 (1.00) The reactor.is operating at 100 percent power.
According to N2-OP-0, "Feedwater_ Heaters'and Extraction Steam Systems," WHAT is the first_ action the reactor. operator must take on. lass of feedwater heating AND.WHY? -(1.0) DUESTION 4.10 (2.00) According to N2-OP-11, " Service Water System," WHAT are six (6) actions the reactor operator must take on loss of all service . water pumps? (2.0) i j i i l , i <
(***** END OF CATEGORY 04
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... . -- _ __ - _ _
$ f1 ., .. .._________..._____.. ______.___ __.__................_._________..___ _.. . EQUATION SHEET i .......__..._..._......___...__....._......_....._.____.__._.............. .!
. ' Where m3 = m2 (density)i(velocity)1(area); = (density)2(velocity)2(area)2 ...._________-_.................._-_____...... __________.... ______. __ KE="{2 PE = mgh PE + KE +P V where V = specific y 3 = PE +KE +P v22 y
2
volume P = Pressure .........__ _____._..._________________....___________._______._____..._.
. . , Q = mc (T-Tin) Q = UA (T-Tstm) Q = m(h -h ) ! p out ave 1 2 - ! __.__---...._......._-_..........__._..____....__.._-__........__..__....... j P = P 10(SUR)(t) P = P e /T SUR = 26.06 T = (B-p)t t
o o T p ......... __......._.._-__..______._______________._.__-_______...______.
delta K = (Keff-1) CR (1-Keff1) = CR (I-Keff2) CR = S/(1.Keff) ! i
l ' (1-Keff1) (1-Keff) x 100% M = (1 eff2) SDM = Keff ......._.._...._..................__.._._.-_.._..__.........._...._._.._.._.
decay constant = In (2) * 0.693 g = A e-(decay constant)x(t) A t t g 1/2 1/2
. ..._--_.............__....___.______._..__ _. -__..__....._...........____ Water Parameters Miscellaneous Conversions
1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 dps I gallon = 3.78 liters 1 kg = 2.21 lbs
3 1 ft = 7.48 gallons I hp = 2.54 x 10 Btu /hr
6 Density = 62.4 lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec ............______-.-___......_....._............._-.-_.......-.....--..... ' t i L
- . _ _ _._._. _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ D j . . ,
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. _ _ _ _ _ _ _ _ _ _ -. _ _ _ _
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From. OF- 0F, "Reidgsk * , , h.
I.
PROCEDURE FOR CORRECTING ALARM CONDITIONS (Cont.)
10.0 851539 Reactor feed Pump 1A/18/1C Suction Pressure Low / Low-Low l . REAC FEED PMP l l 1 A/18/1C , SUCTION PRESS i LOW /LO/LO { - -- - -- - - ' 851539 l ' l 851539
10.1 Computer Point Computer Printout Source CNMPC07 FW PMP 2fWS-PIA 2CNM-PS74A (210 psig) i ( - SUC PR $ .
. s.
, '
CNMPC08 FW PMP 2fWS-P18 2CNM-PS748 (210 psig) SUC PR l
CNMPC09 FW PMP 2FWS-PIC 2CNM-PS74C (210 psig) SUC PR I I CNMPC10 FW PMP 2FWS-PIA 2CNM-PS73A (190 psig) ! SUC PR . CNMPC11 FW PMP 2fWS-P18 2CNM-PS738 (190 psig) SUC PR I CNMPC12 FW PMP 2fWS-PIC 2CNM-PS73C (190 psig) i SUC PR )
l l l ,
-
i ! l , i
_ - -. __ _.
_ ili lC81M91EbEE 9E U9EbE06 E9dE8 EBON 1 9EE00119d1 Wi S R eaos is _ _ _ _ _ _ ! THERMODYNAMICS _ HEAT TRANSFER AND FLUID FLOW .. ANSWERS -- NINE MILE POINT 2-87/07/07-MOON, D.
l , ANSWER 1.01 (1.00) l-I a.
.True' b.
False .. 'E+0.53 each REFERENCE 1.
NMP-2: BWR Academics, Reactor Theory, pp. 5-13 and 5-15.
292OO5K109 ...(KA'S) ANSWER 1.02 (1.00) t/T P =Pe Where P = 1500 P = 1000 t = 30 seconds
0
O T = period' T = t/In (P /P )
O T = 74 seconds C+0.53 .,/4/U Then t' = T In (P /P ) Where P = 3000
1
t' = 51.3 seconds (accept 50.3 to 52.3) [+0.53 REFERENCE 1.
NMP-2: BWR Academic, Reactor Theory, p.
3-17.
292OO3K108 ...(KA'S) l I I i ! ! m____.__.____ __ _ _ _ j
- _ - - _ _ _ _ - - _ _ . - - _ _ .-. _ _.. _. ___.
I- ~1._ PRINCIPLES OF NUCLEAR POWEREPLANT DPERATIONr: PAGE: 1 <4 . LIUEBgggyN8MIC@t_UE@l_lBON@EEB_@Np_ELUID_ELQW , . ANSWERS -- NINE MILE PDINT 2-87/07/07-MOON, D.
., * - . _ ANSWER.
1.03 (1.50) ! Peak 7 to 11 hours 4- ' Xenon i Reactivity
'(-% delta k/k)
2-i 1-l _ _ _ _ _ -- . - - - _ _ _ _ _ _ _ _ _. _________ O
70 Time (hours) Starts at'-3% [+0.53 Peaks at.-47, after'7 to 11 hrs .[+0.53 Decays to 0% after:70 hrs [+0.53 REFERENCE 1.. NMP-2: BWR Academics, Reactor Theory, p.
6-11, 292OO6K107 ...(KA'S) -ANSWER 1.04-(1.50) a.
True [+0.53-b.. True [+0.53 c.
True [+0.53 REFERENCE 1.
NMP-2: BWR Academics, Heat Transfer and Fluid Flow, pp. 9-26 through 9-29.
293OO9K122 293OO9K123 293OO9K124 ...(KA'S) _______-_______ _-__ ____ _ - _
-_ ____ ___ - _
- alz_1EBINglELEG_gF_NygLE88_EgWEB_EL@NI_gEEggIlgN PAGE
t IUEBdgDyNOd1CS _HEQI_IB8NSEEB_OND_ELUID_ELOW t .. ANSWERS -- NINE MILE POINT 2-87/07/07-MOON, D.
. ) gII
ANSWER 1.05 (2.
c.
LHGR - one percen 3 plastic strain on cladding [+0759 APLHGR - clad temperature of 2200 degrees F [+0.53 j CPR - boiling transition E+0.53 j l b.
fuel cladding cracking E+0.53 (also accept cladding failure)
l REFERENCE
l
1.
NMP-2: HWR Academics, Heat Transfer and Fluid Flow, p.
9-15.
! 293OOOK120 293OO9K10B 293OO9K112 ...(KA'S) l i ANSWER 1.06 (2.00) a.
Void fraction initially decreases [+0.25] as increased core l flow sweeps voids out of the core [+0.253.
As power increases, I the rate of boiling increases [+0.253 causing the void fraction l to return (increase) to its original value [+0.253.
y 0@w7h/i7 wax b.
Initially, core net reactivity (is zero and) becomes positive l (increases) as void content decreases E+0.253.
As the increase I in power level begins to take effect, with void fraction beginning_to increase [+0.253, net reactivity begins to decrease (become less positive) E+0.253.
(The power increase will also cause doppler to add negative reactivity.)
Net reactivity returns to zero [+0.25] (as the transient settles out, with the core at a higher power level).
REFERENCE 1.
NMP-2: BWR Academics, Reactor Theory, pp. 7-10 and 7-19.
292OOOK120 ...(KA'S) l l L-------_-____---
. - - _ _ _ _ -- . 1 _2EBINCIPLESz_OF NUCLEAR _PgWER_ PLANT _gPERATIgNe PAGE. 16 lME8DgDyN@51CS _Ug@I_IB@NSEE6=AND FLUID _FLgWe t ,.. 1 ANSWERS -- NINE MILE. POINT 1 2.
-87/07/07-MOON, D.
. ' ANSWER.
1.07 (2.00) a.. The reactor period during power increases is governed by how.
quickly the neutron. population can increase [+0.53.
.The same holds true on a power decrease however, the neutron population j ' is dominated by the longest lived delayed neutron precursor E+0.53.
(This decays with -80.second period.)- l b.. 1.
-decay heat 2.
the time-delay in heat transfer from the fuel to the coolant [+0.53 each REFERENCE 1.
NMP-2: BWR Academics, Reactor Theory, pp. 3-45 and 7-22.
'292OO3K106 292OOBK125-292OOBK130 ...(KA'S) i ANSWER 1.08 (2.00) a.
Decreases b.
Increases-c.
Increases d.
Decreases ! -[+0.53 each ! ' REFERENCE 1.
NMP-2: BWR Academics, Heat Transfer and Fluid Flow, pp. 6-76 through 6-81.
293OO6K110 ...(KA'S) < ! { , .---------_-__J
_ - - _ _ _ _ _.. -- _ _ .___ ,1.
- PR'INCIPLES'DF NUCLEAR POWER PLANT OPERATIDN PAGE
t l-ISEB50gyNOMICgz_UE81_IB8N@EgB_8NQ_E691p_E60H - ANSWERS - - NINE MILE POINT 2-87/07/07-MOON, D.
4:
- ...
1l-ANLWER 1.09 (2.00) l l i' ' m.
Decrease l b.
Increase l c.
-Increase d.
Decrease ([ C+0.53 each
. REFERENCE
1.
NMP-2: BWR Academics, Heat Transfer and Fluid Flow, pp. 5-12, 5-14, 5-41, 5-49, 5-57, and 5-64.
293OOSK103 29305K105 ...(KA'S) l
1 ' l
-i
l . i i l
- - - - _ - - - - - -- - --
-, _ _ _ _ _ _ _ _. - _ _ _ _ _ _ , _ -. - ____ - - - _ _ _ .. _ - -. _. _, - t - ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 18- - . '2.
,
p .iANSWERS - <NINE MILE POINT 2-87/07/07-MOON, D.
1, l.
\\. .. ANSWER ~ 2.01 ' (2. 00). 1.. ' reduction in electrical output 2.
steam flow to f eedflow mismatch j.
3.- Px vessel level. change '4.- ' suppression pool temperature' increase L < - (other. answers are acceptable-if supported)
-l C+0.53 each .i ! REFERENCE 1.
NMP-2: Operating Procedure N2-OP-34, Section B, " System Description," p.
9.
KAs-239002A100 239002A105 239002A107 239002A109 A'NSWER-2.02 (1.50) .1. : reactor building above/below refueling floor exhaust radiation high
- 2.
reactor building above/below refueling floor ventilation. air flow low l 3.
high D/W pressure 4.
Iow-law Rx water level (Accept LOCA'for 3'and 4) Any three (3) [+0.53 each, +1.5 maximum.
REFERENCE "1. .NMP-2: Operating Procedure N2-OP-61D, Section D, " System Description," p.
3.
- 2.
NMP-2: Operations Lesson Plan, Standby Gas Treatment, p.
14.
. KAn ' 261000K401 s ll
i _--_ -_-__- -
- _ _ .2 _;[g@yl_gggigy_lyggyplyg_ggggly_gyg_gdggggggy_gySTEMS PAGE
7 i . ANSWERS -- NINE MILE PDINT 2-87/07/07-MOON, D.
- Y S
' oM ANSWER 2.03 (1.50) ggAT c.
yes [+0.53 0F k#' f p b.
The (LOCA) initiation logic 2s one-out-of-two taken twice J+0.53 k k b */ lessart & Lwleve$ clo n) w(#xe/tryka tr'ple-law level [+0.53.
for high r ell press re and/ r Qky O$c54 Lov mNoM a % f a. inf{tasa.
REFERE 'E ' t//7 1.
NMP-2: Operations Lesson Plan, Low Pressure Core Spray, pp. 10 and 14.
KAs 209001K408 ANSWER 2.04 (2.00) a.
control rod velocity 1imiter b.
electrical protection assemblies (EPA's) c.
main steam line flow elements (or MSIV's) gp d.
core shroud (provides two-thirds core coverage) a [d ~ mg l [+0.53 each REFERENCE 1.
NMP-2: Operations Lesson Plan, Control Rod Drive Mechanism, p.
19.
2.
NMP-2: Operations Lesson Plan, Reactor Protection System, p.
7.
j 3.
NMP-2: Operating Procedure N2-OP-1, Section B, " System Descripti on," p.
1.
4.
NMP-2: Operations Lesson Plan, Rx Vessel & Internals, p.
13.
i _ -_______-____- - __ -
... - _ - _ _ 22a_jPLONI_pEglgN_INCLyglNg_@@ Eely _@Np_EME8gENCY_@y@lEMg- .PAGE-20 ' sANSWERS--- NINE' MILE POINT 2-87/07/07-MOON, D.
l , ' ANSWER: 2.05 '(1.50) l' i I l '.. LPCI initiation signal (high drywell pressure and/or triple-law reactor water-level) is present J2.
high-drywell pressure-(1.68 psig) exists 3.
associated LPCI injection valve is shut j [+0.53 each-REFERENCE .1.1 NMP-2: Operations Lesson Plan, Residual Heat Removal System, l pp. 20, 21, and 25.
KAsn 226001A105 hb //gy LM
ANSWER 2.06 '2.50)- // I a.
load sequencing starts'when the. diesel generator output breaker shuts [+0.53 DE9' fS1 \\, keeut. M0r haq hD,61 .PCI pump A) b.
1.
S mp A ( 2.. CS ump (or L CS pump) 3. ' S v ce Water P p ~ i e-t+Gr253-emetri p1us-f+Or253-f or correeb-or-der.h/7/g7
c.
1.
engine overspeed C+0.53 2.
generator differential [+0.53 .a. ) REFERENCE 1.
NMP-2: Operations Lesson Plan, Standby Diesel Generators,
p.
6.
2. ' NMP-2: Operations Lesson Plan, Emergency AC Power Systems, p.
6.
KAs 264000K402 264000K405 I l i ! J__ _ _ - _ - _ _ - _ _ -. -
_ _ _ _ -. 32 _LPLgNTiDESIGN_ INCLUDING _gAFETY AND__EMER@ENCY_QYgTEMS PAGE-21
- ANSWERS -. NINE MILE-POINT 2
--87/07/07-MOON, D.
' .< ANSWER' 2.07 (2.50) -1.
CSH diesel start signal i 2.. CSH pump start signal-3.
.(CSH) injection valve open signal (MOV-107)- 4.
CST suction valve (MOV-101) open signal (if suppression pool suction valve, MOV-118, is closed) 5.
suppression pool test close signal return valve (MOV-111) 6.
- CST test return valve close signal (MOV-110)
7.
CGT test return valve close signal ,yg 2" . (7 ( "'~) f/Jf f" (MOV-112) of Syd(udd or/ 9.
csH wp (W #Wt f [+0.33 each component, +1.0 maximum.
[+0.123 each signal, +0.7 maximum.
REFERENCE 1.
NMP-2: Operations Lesson Plan, High Pressure Core Spray, p.
13.
KAs 209002A201 ANSWER 2.08 (1.50) i When both RPS trip systems trip, the backup scram valve solenoids j energize [40.253 shifting the two backup scram valves to block the instrument air supply to the exhaust and scram valves [+0.253.
They also provide a redundant means of venting air from the 105 scram pilot valves [+0.53 and the SDV vent and drain valves.
The pilot air header bleeds down and the scram valves open, causing insertion of all control rods [+0.53.
REFERENCE 1.
NMP-2: Operations Lesson Plan, Reactor Protection System, p.
11.
2.
NMP-2: Operations Lesson Plan, Control Rod Drive Hydraulics, - - _ ___
-- . _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ 2_. * PLA_NT - DES _I G_N__I N_CL_U_D_I N_G SA_F_E_TY_AN_D_E_ME_RGEN_C_Y_ S_Y_S_T_E_M_S PAGE
' _
- - -- -- - _ - -_- _
- ANSWERS -- NINE MILE POINT. 2-87/07/07-MOON, D.
- p.
14, e 201001K404 212OOOK115- ...(KA*S)
i i
I l i i ! l l I _ _ _ _ - -. _ _
____ -- - _ -_ -- --_ - _ _____ _ _ 3.
INSTRUMENTS AND CONTROLS.
PAGE.;23 ] . _ . tb ANSWERS -- NINE' MILE POINT 2-87/07/07-MOON, D.
i .I i CANSWER: 3. <01 (1.50) I 1.. SRM downscale <3 cps C+0.53 1:2.' SRM upscale or INOP >1 x 10**5 OR module unplugged 'E+0.53 . OR low'high voltage, OR i ' SRM select! switch not in operate-3.
SRM detector not <100 cps and detectors not' fully.
C+O.53 ' ' full in inserted REFERENCE ~ i . . " 1. : NMP-2: Operations Lesson Plan, Source Range Monitoring System, . p.
16. : KAs 215004K401 . . ANSWER !3.02 (2.00) 1.- high drywell' pressure .1 ':2. - -high reactor vessel. pressure 3.
low reactor water level 4.
high main. steam'line radiation ~ 5.
neutron monitoring system (individual inputs may be bypassed, but there is always some type of NMS scram) [+0.43 each REFERENCE .1.
NMP-2: Operations Lesson Plan, Reactor Protection System, pp.
16,-17, and 18.
KAs .212OOOK412 I V l u . L .________-___-________a
_ _ _ _ _ .. _ _ _ - _ ___ _.
__ - _ - _ _ ._ _ __ . _ - _ _ _ _ _ -, 3tiflNSTRUMENTS___AND_CONTRgLS PAGE
'a ANSWERS -- NINE MILE POINT 2-87/07/07-MOON, D.
. ! p/E7 ANSWER 3.03 (2.00) eTdod High reactor pr =ssure 1050 psig [+0.53 OR low-low reactor level (level 2) '100.
inches [+0.51 concurrent with APRM power greater than 4 percent [+0.53 after a 98 second time delay [+0.53.
(Initiation occurs if the initiation conditions are still present after 90 seconds.)
REFERENCE 1.
NMP-2: Operations Lesson Plan, Standby Liquid Control System, p.
15.
2.
NMP-2: Operations Lesson Plan, Redundant Reactivity Control System, p.
10.
KAn 211000A308 I ANSWER 3.04 (1.00) n.
inhibits flow control valve motion b.
transfers loop flow controllers from automatic to manual [+0.51 each REFERENCE 1.
NMP-2: Operations Lesson Plan, Re. actor Recirculation Flow Control System, p.
15.
KAs 202OO2A301 ANSWER 3.05 (1.00) a.
False [+0.53 b.
False [+0.53 REFERENCE 1.
NMP-2: Operating Procedure N2-OP-34, p.
9.
2.
NMP-2: Operations Lesson Plan, Automatic Depressurization ' System, p.
16.
! KAs 21BOOOK501 _
_ _ _ ___ _ 3, ' INSTRUMENTS AND CONTROLS PAGE
,
.
-
ANSWERS -- NINE MILE POINT 2-87/07/07-MDON, D.
' l r- . ! .. t' " y ' ANSWER 3.06 (3.00) ' e a.
, Control valves close to limit turbine load to 90 percent C+0.253
Bypass valves open [+0.253 to maintain r eactor pressure constant I 4' Reactor pressure remains constant [+0.253 ,, a b.
Control valves close to limit turbine load to 90 percent [40.253 Dypass valves remain closed L40.2SJ $ Reactor pressure increases C+0.253 (and reactor scrams on high pressure) g - c.
(Pressure / regul a t or 'D' takes over.)
Control valve demand / decreases slightly (due to offset bias between regulator 'A' i; J /and 'B') then reopen to 1007. demand [40.253 ,> a , 'l ' Dypass valves remain closed C40.25] Ia Reactor pressure is maintair'ed /,+ssentially constant (at approximately,950 psig) [+0.253 d.
(Loss of speed signals causes a turbine trip and a subsequent reactor scram.)
Control val vem cl ose.
[40.253 . Bypass valves open C+0.25] (to control reactor pressure).
' Reactor pressure is maintained at approximately 920 psig C+0.253 / > REFERENCE 's 1.
NMP-2: Operations Lesson Plan, Turbine EHC System, Learning Objectives E06 and EO9.
2.
NMP-2: Operating Procedure N2-OP-21, Main Turbine, pp. 22 and 23.
i KAs 24100K406 241000A107 241000A10B < i , Y -.__.__..______________________m
. _ - - _ - _ _ 3 __INylBUMENIS_AND_CGNIBOLg /<c PAGE
' t / \\ ' '. ANSWERS -- NINE MILE POINT 2-87/07/07-MOON, D,! \\ l . ' ' ANSWER 3.07 (1.50) Indicated vessel water level would be higher than aciual level E40.53.
As containment temperature irecreases, the densi ty '.sf water in the , reference leg will decrease E40.53.
This will result in a lower > d/p between the reference leg end the variable ~ leg [+0.53 (which gives a higher ind'icated level).
(* REFERENCE (, 1.
NMP-2: Operations Lesson Plan, Reactor Vesnel Instrumentation, Learning Objective EO-9, pp. 8 and 9.
, KAs 216000K507
- -
ANSWER 3.00 (1.00) r a.
Narrow range E40.53 , b.
Wide range [+0.53 REFEREi!CE 1.
NMP-2: Operations Lesson Plan, Reactor Vessel Instrumentation, p.
10.
j KAs 2160COK101 21LOOOK104 216000K106 i .] i f (2.00) ANSWER 3.09 increase [b53 a.
1.
2.
decrease bo.53 d6'kN / 3.
decrease Do.5.l b.
3" [+0.53 t REFERENCE 1.
NMP-2: Operations Lesson Plan, Feedwater Control System, pp. 6 ,_ ' and 10.
KAu 259001K607 ' > i I .2 f.l'
o/o .4.
' PROCEDURES - NORMAL _ ABNORMAL _ EMERGENCY _AND PAGE
t t i - RADIOtJ1GICAL CONTROL '* l t'..,
- ,
, 'N dhSWERS -- NING MILE POINT 2-87/07/07-MOON, D.
. ~ s ' 5 % s ! ANSh7.H 4.01 (2.50) I 9I.
1.
Reactor feedpump 2FWS-P1A (B,C) trips - 45 sec/18 sec TD (0,5) 2.
Auto start of standby condensate booster pump (0.5) '" j (if FW pump suction pressure drop to 210 psig) b.
1.
The feedwater' pump will trip after a 45 second time delay (0.5) if the suction pressure is between 190 and 210 psig.
2. The feed.(ah r ump will trip after an 10 second time delay (O.5) if the suction pressure is below 190 psig.
> c.
If feedwater pump suction low pressure occurs after a main (0.5) !. turbine trip,'the FW 1evel control valves 2FWS-LV10A (B,C) A will be automatically shut to 48% open.
, REFERENCE < . ,, - 'NMP-2 OP-OD, feeduater System ! 259001A201 '/ 5$9001A310 259001S008 ...(KA'S) l ANSWER 4.02 (1.00) a.
25, 705, 10 b.
1325 i ,N [+0.25] each blank j REFERENCE.
F 1.
NMP-2: Technical Specifications, p.
2-1.
239001SGO3 ...(KA'S) ANSWER 4.03 (1.00) ] l establish a natural circulation flow path [+1.03 ]
REFERENCE
1.
NMP-2: Operating Procedure, N2-OP-101C, " Plant Shutdown," I t p.
R.
\\ 295022AK30 ...(KA'S) !
! .
1 L _ _______ _ l
- - - _ _ - l d._' PRgCEDURES - NgRMAl ;AgNQRMAL _ EMERGENCY _ANg PAGE 28-l _ t t 80919LgGlCQL_CQNIBQk ..., ANSWERS - 'NINE MILE POINT 2-87/07/07-MOON, D.
. L
l l ' ANSWER 4.04 (1.00)
.1.
whenever the' safety of the reactor is in immediate jeopardy- . when operating parameters exceed any reactor protection setpoints 2.
and automatic shutdown should, but does not occur.
L l' \\ [+0.53 each REFERENCE 1. - NMP-2: Administrative Procedure AP-4.0, Administration of . Operations, p.
2.
295006SG10 ...(KA'S) ANSWER 4.05 (1.50) a.
suppression pool water level (at 198.6 ft) [+0.53 b.
1.
drywell temperature control (DWT) 2.
suppression pool level control (SPL) 3.
suppression pool temperature control (SPT) 4.
primary containment pressure control (PCP) l [+0.253 each REFERENCE .1.
NMP-2: Emergency Operating Procedure, N2-EOP-DWT, "Drywell Temperature Control," p.
4.
295030SG11 ...(KA'S) !
1
l _ _ _ _ _ _ _ _
- _ _ _ .. 4. PROCEDURES - NORMAL _ABNORMAlt_ EMERGENCY _AND PAGE
t E 68D106001G86_GgNIGgL l
ANSWERS -- NINE MILE POINT 2-87/07/07-MOON, D.
l
. l
l ANSWER 4.06 (2.00) 1.
differential pressure Rx building-to-outside greater than or equal to 0 inches of water 2.
any high area temperature which causes an isolation 3.
any unexpected high radiation alarm on a Rx building area radiation monitor 4.- any Rx building HVAC exhaust radiation which causes an isolation S.
any Rx building floor drain sump high-high level [+0.43 each REFERENCE 1.
NMP-2: Emergency Operating Procedure, N2-EOP-SCL, " Reactor Huiloing Level Control," p.
1.
295036SG11 ...(KA'S) ANSWER 4.07 (1.00) 1.
notify the senior shift supervisor 2.
initiate a manual reactor scram C+0.53 each REFERENCE 1.
NMP-2: Operating Procedure, N2-OP-101C, " Plant Shutdown," p.
8.
29501SSG10 ...(KA'S) i I - - - _ _ - _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _
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' PROCEDURES _ _NgRMAl _ADNgRMAl _ EMERGENCY _AND PAGE
t t RADIOLOGICAL CONTROL e ANSWERS -- NINE MILE POINT 2-87/07/07-MOON, D.
. ANSWER 4.08 (2.00) m.
Whole body - 3.0 rem / quarter E40.253, not to exceed 5 (N-18), where N is age C+0.253.
Extremity - 18.75 rem / quarter C+0.253 Skin - 7.5 rem / quarter C+0.253
ylof87 b.
100 mrem / calendar week C + 0. 259
4000 mrem / calendar year [ +0. 25-3 REFERENCE 1.
10CFR20.101, 2.
NMP-2: Radiation Protection Procedure, S-RP-1, " Access and Radiological Control," pp.
1, 2, 17, and 18, 294001K103 ...(KA'S) ANSWER 4.09 (1.00) Reduce recirculation flow to reduce reactor power C+0.53 to 75 percent [+0.253, or by 20 percent, whichever change is larger C+0.253.
REFERENCE 1.
NMP-2: Operating Procedure, N2-OP-8, p.
10.
259001SG14 ...(KA*S) ! l l l l l j - _ _
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1 t t . RADIOLOGICAL CONTROL ' , ANSWERS -- NINE MILE POINT 2-87/07/07-MOON, D.
l
l ANSWER 4.10 (2.00) l 1.
verify (all service water pumps have tripped and) that no service water pumps can be restarted 2.
reduce recirculation flow to minimum 3.
scram the reactor (f ollow scram procedure N2-OP-101C) 4.
trip the main turbine
S.
shut the MSIV's (manually initiate RCIC) 6.
trip the Rx recirc pumps I 7.
trip the RWCU pumps O.
notify the SSS
C+0.33] each, +2.0 maximum REFERENCE
1.
NMP-2: Operating Procedure, N2-OP-11, " Service Water System," p.
13.
29501GSG10 ...(KA'S) l l ) I
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' ' * , GGESTION VALUE REFERENCE ________.______ ____ = '01.01 1.00 BANOOOO777 01.02 1.00 BANOOOO770 01.03 1.50 BANOOOO779 01.04 1.50 BANOOOO780 01.05 2.00 BANOOOO781 01.06 2.00 BANOOOO782 01.07 2.00 BANOOOO7B3 01.08 2.00 BANOOOO784 01.09 2.00 BANOOOO785 ______ 15.00 02.01 2.00 BANOOOO795 02.02 1.50 BANOOOO796 02.03-1.50 BANOOOO797 02.04 2.00 BANOOOO798 1/161.50 BANOOOO799
02.05 O2.06 2 #_2,40-BANOOOOOOO O2.07 2.50 BANOOOOB01 02.08 1.50 BANOOOOB02 ~~; E,go ant 7 03.01 1.50 BANOOOO803 03.02 2.00 BANOOOOBO4 03.03 2.00 BANOOOOBOS 03.04 1.00 BANOOOOB06 03.05 1.00 BANOOOOBO7 03.06 3.00 BANOOOOOOB 03.07 1.50 BANOOOOBO9 03.08 1.00 BANOOOO810 03.09 2.00 BANOOOO811 ______ 15.00 04.01 2.50 B/,.,0000748 04.02 1.00 BANOOOO786 04.03 1.00 BANOOOO787 04.04 1.00 BANOOOO788 04.05 1.50 BANOOOO789 04.06 2.00 BANOOOO790 04.07 1.00 BANOOOO791 04.00 2.00 BANOOOO792 04.09 1.00 BANOOOO793 04.10 2.00 BANOOOO794 ______ 15.00 ___ ______ 60.00 _ _ _ _
. - _ - _. - - - - , l ' RC fficial Use Only . . n 1, \\ Nuclear Regulatory Commission }.1 i, - Operator Licensing }j' Examination Li This document is removed from Official Use Only category on date of examination.
l NRC Official Use Only __ ____________a
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s .' . O . U.
S.
NUCLEAR REGULATORY COMMISSION - , SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION .
FACILITY: NINE MILE POINT 2 -_ ,
REACTOR TYPE: __BWR-GES _ _____ ____ DATE ADMINISTERED: _gZfgZ[O7 _________ i _gARRELL _D._____________ EXAMINER: t CANDIDATE: _ _ _ _ _ _____________ INgIBygligNg_lg_gBNglgBIEL Read the attached instruction page carefully.
This examination replaces the current cycle facility administered requalification examination.
Retraining requirements for failure of this examination are the same as for failure of a requalification en: amination prepared and administered by your training staff.
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 least 80%. Examination papers will be picked up four (4) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY __Y0LgE_ _lgIOL ____ggggE___ _y@LgE__ ______________gQlEggBY____ ___ 15199__ _35199 ________ 5.
THEORY OF NUCLEAR POWER PLANT ___________ OPERATION, FLUIDS, AND THERMODYNAMICS _19199__ _3gzgg _ _ _ ____ ________ 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 19199_ _29199 ________ 7.
PROCEDURES - NORMAL, ABNORMAL,
______
EMERGENCY AND RADIOLOGICAL CONTROL ~~~~~~~~ ~~~~~~ ~~~~~~~~~~~ ~~~~~~~~ B.
ADMINISTRATIVE PROCEDURES, 15.00 25.00 CONDITIONS, AND LIMITATIONS l 60.00 % Totals ' Final Grade All work done on this examination is my own.
I have neither given nor received aid.
___-------------------------- ----- Candidate's Signature - _ _ _ _ - _ _ - _ _ -
_-- -_ _.
- V ,: l .s . . . NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the f ollowing rules apply: l 1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
,
1 1, 2.
Restroom trips are to be limited and only one candidate at a time may
l leave.
You must avoid all contacts with anyone outside the examination ! [ room to avoid even the appearance or possibility of cheating.
l l 3.
Use black ink or dark pencil gnly to facilitate legible reproductions.
j l ' 4.
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 , section of the answer sheet.
8.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne side of the paper, and write "Last Page" on the last answer sheet.
l l 9.
Number each answer as to category and number, for example, 1.4, 6.3.
l l 10. Skip at least gjree lines between each answer.
i l 11. Separate answer sheets f rom pad and place finished answer sheets f ace 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 the 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 to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER DLANK.
16. If parts of the examination are not clear as to intent, ask questions of , the examiner only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has been completed.
_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ ~
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.>. , . . > 18,,.. When you complete your examination, you shall: a.
Assemble your examination as f ollows: (1) Exam questions on top.
(2) Exam aids - figures, 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 f or answering the questions.
d.
l. eave the examination area, as defined by the examiner.
IfLafter 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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QUESTION 5.01 (3.00) Assume that the reactor is being started up with the bulk coolant temperature less than the saturation temperature and the core is at BOL.
Wi th the reactor cri tical, a malfunction in the rod drive system causes several control rods to drive out and the reactor to increase in power level on a short period.
a.
Of the Void,. Doppler, and Moderator Temperature coefficients WHICH would come into effect first, second and third to lessen the rate of power increase? (1.0) b.
EXPLAIN your choices of part a.
Assuming no operator action and that a scram does not occur.
(2.0) QUESTION 5.02 (1.50) ANSWER each of the f ollowing TRUE or FALSE.
a.
Steady state critical power increases with increasing core flow.
(0.5) 6.
Steady state critical power decreases with increasing reactor pressure above 800 psig.
(0.5)
c.
Steady state critical power increases with increasing inlet subcooling.
(0.5) , I I QUESTION 5.03 (1.50) Five (5) minutes following a reactor scram from 1007. power, ) reactor power is 15 on IRM Range 4 and decreasing.
I
WHAT is the minimum IRM Range that you could go to two (2) ] minutes later without violating any administrative operational ! limits? P SHOW calculation and EXPLAIN any assumptions made.
(1.5)
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t IHERMODYNOMIC@ ,2 A^ }} f
' f$goisf,,of $& QUESTION 5.04 (2.50) a.
GIVE two (2) reasons WHY f eedwater heating improven power plant efficiency.
(1.0) b.
If the highest pressure feed heater is removed f rom service (extraction steam isolated) with no operator action, (1) WHAT would happen to Megawatt output power of the generator and WHY; (2) WHAT would happen to reactor output power; and (3) WHICH output changes more? (1.5) QUESTION 5.05 (2.50) With respect to clad failure mechanisms, complete the following ttble (replace the Xs with short answers): (nC' (2.5) 4.p t.T ' '.W,ejs Limiting,' l' Reason for ! ,, 7 Parameter Name Clad Failure im't Therg 1} l$7 / a.
1% clad plastic strain X clad-fuel pellet interaction b.
clad temperature X X < 2200 deg F c.
boiling transition X X temperature QUESTION 5.06 (2.50) A centrifugal pump i s operating at rated speed (3600 RPM) with an output head of 240 psig.
If the pump speed is reduced to 900 RPM, a.
WHAT is the new output head? (1.5) b.
By WHAT factor is the power consumption reduced? (1.0) SHOW all work.
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IUEggggyN8MICg.
f..' -QUESTION 5.07 (1.50)- The reactor.is operating'at 90 percent power when a parti.al.. lass- .of feedwater heating occurs.
According to the Reactor Analyst, the'frr9 M d emperature decreased by 5 degrees F and voids-decreased by percent.
c.., ~ m det.a ar.
1iff7 s Assuming the reactor re urnsf t a 90 percent power condition i .'with NO rod movement or recirculation flow changes, WHAT would be the corresponding fuel temperature change? (1.5) ' ! l ! <
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pc _PLONI_SYSlEMg_pE@l@Nt_ggNIBQ6t_6Mp_INSTRUMENTATlgN I .. I ! DUESTION 6.01 (3.00) Ttue reactor has been operating at 100 percent power f or 3 months.
! i - The EHC Load Limit is set at 105 percent and the Load Set Selector is set at 1215 MW.
- The Maximum Combined Flow Limiter is set at 105 percent.
- Pressure regulator 'A' is in control.
Using the attached EHC block diagram, EXPLAIN WHAT would happen to control valve position, bypass valve position, and reactor pressure in each of the following separate cases (GIVE percent of valve change and steady state pressure f ollowing transient): a.
Load Limit potentiometer is reduced to 90 percent.
(0.75) b.
Maximum Combined Flow Limiter potentiometer is reduced to 90 percent.
(0.75) f ai ls l ow. [ pre ssur e fra w s dheer 5 /fu [ (0.75) c.
Pressure regulator 'A' 1" S $>*D ) kl> ?l7/97 (0.75) d.
Both turbine speed sensors fail.
DUESTION 6.02 (2.40) LIST six (6) interlocks with setpoints that will cause recircu-lation pump speed to automatically shi f t f rom high to low.
(2.4) DUESTION 6.03 (2.50) During normal operation at 50% power the f ollowing alarms energize in rapid succession in the order listed.
1.
CONDENSATE STORAG TK 1A/1D LEVEL LOW 2.
CONDENSATE STORAG TK 1A/1D LEVEL LO-LO 3.
CNST XFER PMP 1A/ID AUTO TRIP / FAIL TO START a.
In a normal value lineup, WOULD the apparent low level , condition apply to one or both CSTs? WHY? (O.5) b.
WHAT effect does this have on ECC valve positions? (2.0) (***** CATEGORY 06 CONTINUED ON NEXT PAGE
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t t qP ' pp z.co f GUESTION 6.04 02r907 g* g CO A loss of all offsite power has just occurred.
Concerning the Division I standby diesel generator: a.
WHAT event initiates the load sequencer? (0.5) r e sto r ed bak 1oadhsequenced[tothe 9[f7 o,g b.
WHAT ar-e-the first-thice '3 ajor
4-h-tD bus? 1s c.
STATE two (2) conditions that will trip the diesel generator when a LOCA signal is present.
(Do not include manual acti ons. ) (1.0) QUESTION 6.05 (2.00) While the plant is operating at full power, one SRV inadver-tentl y opens.
LIST four (4) key plant parameters which the operator could observe that would change as a result of the SRV being open.
Do NOT list valve position indication lights, acoustic monitoring instrumentation, SRV tailpipe temperature or computer alarms.
(2.0) QUESTION 6.06 (1.60) WILL a loss of INSTRUMENT AIR have any effect on the ability of the Standby Liquid Control system to inject baron solution into the reactor vessel? EXPLAIN WHY.
(1.6) < (***** CATEGORY 06 CONTINUED ON NEXT PAGE
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. QUESTION '6.07 '(l.00)' ANSWER each'of the f al1owing.TRUE or FALSE.
a.
If ADS is manually initiated and all' initiation conditions are met, the ADS SRV. opening will be delayed by the 105-second timer to allow time for the CSH system to try to recover vessel. water level.
(0.5) b.
If'an ADS blowdown is in progress, with all the initiation j signals present,' repositioning the ADS Initiation Inhibit
switches will reinitialize the IO5-second timer and close all ADS SRV's.
(0.5) l i , ,
!
(***** END OF CATEGORY 06
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t
B001gLgg1 COL _CQNIBgL , QUESTION. 7.01 (2.50) s idG Use the figure containing a section of OP-08, "Feedwater System" to answer the following questions.
a.
Given the indicated alarm condition, with the plant operating at full power, what TWO (2) automatic responses should have occurred i as a direct result of this alarm? (1.0) b.
What is the significance of having a computer point print out at both 210 psig and 190 psig? (1.0) c.
If feedwater pump low suction pressure occurs AFTER a MAIN turbine trip, how will the Feedwater Level Control valves (2FWS-LV 10A,B,C) respond? (0.5) QUESTION 7.02 (2.50) The reactor is at 25 percent power steady state conditions.
No testing has taken place during the previous 24 hours.
Shortly bef ore shif t turnover at the end of day shift you observe the following parameter values (no alarms are energized): Outside Ambient Temp - 99.6 degrees F Drywell Temp - 147.2 degrees F Suppression Pool Level - 200.6 ft Suppression Pool Temp - 93.1 degrees F l Service Water Inlet Temp - 82.2 degrees F a.
CAN you proceed with shift turnover? Justify your answer.
(0.5) I b.
WHICH setpoint(s) (i f any) are exceeded? (0.5) c.
WHICH procedure (s) (if any) MUST be entered? (1.0) j d.
WHAT conditions must be met to exit any entered procedure (s).
(0.5) { i (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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QUESTION 7.03 (1.00) STATE two (2) reasons for manually opening SRVs until RPV pressure drops below the lowest SRV setpoint in procedure NP-EOP-RP (RPV pressure control).
(1.0) QUESTION 7.04 (3.00) LIST six (6) conditions that require the use of a Radiation Work Permit.
(3.0) QUESTION 7.05 (2.00) According to N2-OP-11, " Service Water System," WHAT are six (6) actions the reactor operator must take on loss of all service water pumps? (2.0) QUESTION 7.06 (2.00) Upon a valid initiation signal for the LPCI mode of RHS: a.
WHAT is the position of the heat exchanger bypass valve? (0.5) b.
HOW long is this interlock valid? (0.5) c.
WHAT is the reason for the interlock and WHAT is the basis g,,);r ,; // L,,,,5;)Lyf (1.0) that makes it possible? pur 9 r * b "'] f1l7l27 'r QUESTION 7.07 (2.00) Following a trip of the Reactor Core Isolation Cooling System Turbine (other than a mechanical trip), HOW is the trip throttle (2.0) valve reset? l f (***** END OF CATEGORY 07
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,9:__890lNiglggllVE_PBQCEQUBEgz_CQNQlligN@t_QNQ_LldlI@llgNg i ' f QUESTION 8.01 (3.00) While perf orming a routine protective system surveillance at ! 75 percent power, the APRM "C" upscale trip setpoint was deter-mined to be 122 percent of rated thermal power.
You are then informed that APRM "A" is inoperable and that 4 hours will be required to readjust the APRM "C" trip function to its proper level.
WHAT actions must be taken, WHY, and with WHAT time restrictions? (Tech Spec sections are provided.)
IDENTIFY all sections of Tech Specs used in arriving at your answer.
(3.0) QUESTION O.02 (2.00) STATE whether the following would constitute a core alteration.
Answer YES or NO.
a.
Removal of an irradiated NDT sample located inside the core shroud with fuel in the reactor vessel.
(0.5) , b.
Relocating a f uel loading chamber in the core during (0.5) refueling.
c.
Replacement of a defective SRM detector with the reactor vessel head removed and fuel in the vessel.
(0.5) d.
Removal of a control rod during refueling with fuel in the vessel, but with no fuel in the 5x5 array centered on the control rod being removed.
(0.5) QUESTION B.03 (2.00) With the plant at 1007. power and the Division 1 diesel oper-ating for a surveillance test, a filter in the air receiver cross connect line fails.
With both air compressors running, the air receiver pressure can be maintained at 200 psig.
Under these conditions, IS the Division 1 diesel generator OPERABLE according to Technical Specifications? EXPLAIN.
(Tech Spec sections are provided.)
GIVE sections of any (2.0) applicable LCO.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE
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> e QUESTION O.04 (1.00) ! J WHAT is the basis for each of the Technical Specifications regarding suppression pool level and temperature? (1.0) QUESTION B.OS (2.00) During 100% power operation, the RCIC steam supply outside isolation valve (2ICS*MOV121) is found to be failed in the open position.
Using the attached Technical Specifications, STATE whether the RCIC, as found, is OPERABLE or INOPERABLE and GIVE ANY Tech. Spec. acti on statement (s) required.
(2.0) QUESTION O.06 (2.00) Technical Specifications require that all jet pumps must be operable or be in at least hot shutdown within 12 hours.
LIST two (2) checks for operability of the jet pumps and GIVE two (2) reasons WHY jet pump operability is a concern.
(2.0) ! QUESTION 8.07 (1.00) During an emergency where normal RWP processing would result in an unacceptable delay in response, WHO may authorize RWP processing and WHAT form of i mmediate documentation is required? (1.0) l , ! (***** CATEGORY 08 CONTINUED ON NEXT PAGE
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l GUESTION O.00 (2.00) You are performing a plant startup following a refueling shutdown.
RPV pressure is approximately 175 psig and increasing.
Due to a faulty transfer switch, the RCIC transfer capability to the Remote Shutdown Pannel will not function.
If you reduce RPV pressure to below 150 psig, is: 1.
RCIC INOP? jpke-(0.5) 2.
the action statement of Tech Spec section 3.7.4 d.
(0. 5 ) 3.
DO ANY OTHER Tech Spec LCOs apply? I/ (1.0) i i l i
3 i (***** END OF CATEGORY 08
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(************* END OF EXAMINATION
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t t IHEBMODYN8MICS , ANSWERS -- NINE MILE POINT 2-87/07/07-JARRELL, D.
i ANSWER S.01 (3.00) a.- Doppler, Moderator Temperature, Void j C+0.33] each, +1.0 maximum.
j b.
As the rods are withdrawn, power level increases causing fuel temp to increase.
The heat generated in the fuel is then conducted into the coolant.
~(It takes about 3 time constants or 30 seconds f or the total increase in heat genercted to be transported to the coolant.)
So, the fuel temperature rises first, causing the doppler to be the first effect.
The next effecc would be the moderator temperature, as the coolant is heate.d to saturation.
Finally comes the effects of voids, as
the heat generated in the fuel boils the water flowing through the core.
E+2.O] REFERENCE 1.
NMP-2: BWR Academic Series, Reactor Theory, pp. 4-42, 4-44, and 4-48.
292OO4K114 ...(KA'S) ANSWER 5.02 (1.S0) a.
True b.
True to True [40.S] each REFERENCE 1.
NMP-2: BWR Academics, Heat Transfer and Fluid Flow, pp. 9-26 through 9-29.
293OO9K122 293OO9K123 293OO9K124 ...(KA'S) l
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~ THEgRY_QF_ NUCLEAR POWER _ PLANT _gPERATIgN _FLUIQS _ANQ PAGE
5 t t
! THERMODYNAMICS
ANSWERS -- NINE MILE POINT 2-87/07/07-JARRELL, D.
, I ANSWER 5.03 (1.50) , ' Using P = Po e**(pt/T) C+0.53 15 e**(-120/80) = 3.35 on Range 4 E+0.53 = 33.5 on Range 2 = Therefore, Range 2 is the lowest Range C+0.53 Assumptions: On a down power transient, with large negative reacti vi ty i nsertions, the stable decay period is determined by the longest lived half-life.
For this example, it is assumed to be -80 seconds.
(Administrative assumption is to remain between 25
and 75*/. of scale on instrument.)
REFERENCE 1.
NMP-2: OP # N2-OP-92, Section E.2.7.
2.
NMP-2: BWR Academics; Reactor Theory, p.
3-15.
292OO3K108 ...(KA'S) ANSWER 5.04 (2.50) g ' , e ),,p*p(,, @ ' a.
The energy recovered in feed heating would otherwise be lost , to the main condenser C+0.53 and less heat is required from D 'g, @*,9 I,,r* pe the reactor to reach the desired conditions (more efficient
a# process) [+0.53.
,g 1s* r - ' b.
(1) Megawatt output from the generator would increase E+0.253 tr#e because steam that was formerly being extracted now passes fgk through the turbine to the condenser C+0.253.
(2) Reactor power output increases [+0.5J.
(3) Reactor power [+0.53.
REFERENCE 1.
NMP-2: UWR Academic Series, Thermodynamic Cycles, pp. 5-48.
293OOSK105 ...(KA'S) i j
g___-,___ , -
. . .- . THEORY OF NUCLEAR POWER PLANT OPERATION _FL_UIDS_t_AND_ PAGE
g *'pjEfjMODYNAMICS t ANSWERS -- NINE MILE POINT 2-87/07/07-JARRELL, D.
, , I ANSWER 5.05 (2.50) Limiting Reason for Thermal Limit Parameter Name Clad Failure - - -. - --------------- _ = _ ..-------- - -. - --- a.
17. clad plastic strain LHGR [+0.53 clad-fuel pellet (Linear Heat interaction Generation Rate) b.
clad temp APLHGR (Average (gross) clad melt < 2200 deg F Planner LHGR) (LOCA) [+0.53 E+0.53 s c.
boiling transition CPR (critical local inability temperature power ratio) to transfer heat to the coolant [+0.53 C+0.53 REFERENCE 1.
NMP-2: BWR Academics, Heat Transfer and Fluid Flow, p.
9-15.
293OO9K108 293OO9K112 293OO9K120 ...(KA'S) --___ -_ _ _. - _-___ >
- - _ - _ , . . .. 31' THEgRy_gF_ NUCLEAR _PgWER PLANT _gPERSTIgN _ FLUIDS _ANQ PAGE
- t t "TUEgdggyN8dICS ANSWERS -- NINE MILE POINT 2-87/07/07-JARRELL, D.
, l l.
I ANSWER 5.06 (2.50) ' a.
Centrifugal pump head-speed relationships can be approximated by: Head is nearly equal to pump speed squared --OR-- E40.53 l
L Head Speed
1 _____ = ___ Head
2 Speed
2 240 3600 [+0.53 --- = ---- l x
l 900 240
(4)
= --- = n x = 240/16 = 15 psig [40.53 b.
The horsepower-speed function can be approximated by: hp is nearly equal to speed cubed [+0.53 so the hp would be reduced by a factor of 900 3 1 3
[+0.53 ) (-) (----> = -- = 3600
64 d i.e., only 1/64 of the f ull speed horsepower would be required to produce a 15 psig output.
REFERENCE 1.
NMP-2: Heat Transfer and Fluid Flow, p.
6-96.
i l 291004K105 ...(KA'S) _ _ _ _ '
,. _ - - - - - - - - - - - - - -, , l . 5 __IUEQBY_QE_NgGLEOB_EgWEB_ELONI_QEEBOIlgN,._ELylDGr_OND PAGE
l
IUEBBgDYNOdlG@ ' ANSWERS -- NINE MILE POINT 2-87/07/07-JARRELL, D.
, ,WS 49a'a to(pf ANSWER 5.07 (1.50) -
-aD* delta Ti = aM* delta Tm + aV* delta %V E+0.253 T AcC
av = -1 x 10**-3 delta.k/k/ delta %V [+0.2M3 ' 7[[g7 gk II . f r,,f-aM = -1 x 10**-4 delta k/k/Tm C+0.253 af aD = -1 x 10**-5 delta k/k/ delta Tf [+0.23J s ,}
- ,,, r E(-1 x 10* *-4 ) (-5) + (-1 x 10**-3)(-1)3/(-1 x 10**-5)
4~' delta Tf = 150 degrees F ., delta Ti = i or average fuel temperature increases [+0.253 150 degrees F C+0.253.
REFERENCE 1.
NMP-2: BWR Academics, Reactor Theory, pp.
4-0, 4-19, and 4-37.
292OO4K114 ...(KA'S)
l , , f J !
- ____ _ _ _ _ _ _ ._
N . pg__ PLANT _ SYSTEMS DESIGN _ CONTROL AND INSTRUMENTATION PAGE
t t ANSWERS -- NINE MILE POINT 2-07/07/07-JARRELL, D.
i e \\ . . l ANSWER 6.,dl / (3.00) ! ' a.
1.
Cf3n tt-Ol valves cldse to limit turbine load to 90 percent 2.
bypass valves open to pass 10 percent of full load (approximately 40 percent of bypass valve capacity) 3.
reactor pressure remains constant,,e letM45e5 sf;9 b to 6Pt/ eloWy bias.
G gap {s7 [40.253 each b.
1.
control valves clay.e to limit turbine load to 90 percent
2.
bypass valves remain closed
.,[oe.I'l & K 1/O j/ N 3.
reactor pr 'sure incVeases rand reactor scrams on high ,j pressure bypass valves then maintain reactor pressur/> at ' approximately y 920 psig [40,15] j 4e ~ [+0.253 each ,, . c.
1.
(pressure regulator 'B' takes over) Control valve position - decreases slightly (due to offset bias between regulator ' ' 'A' and 'D') - depositions to 100 percent demand l ' < 2.
UPV's remain closed ',, ' - , t
< 3.
Reactor pressure increases slightly (due-to bias offset) [40.253 each d.
.1.
(Loss of speed signals c.tuses a turbine trip arid a subsequent reactor scram.7 Control valves close.
2.
BPV's open to control reector pressure.
( 3.
BPV's control reactor pressure at nearly equal to 920 psig.
" [40.253 each REFERENCE < 1.
NMP-2-Operations Lesson Plan, Turbine EHC System, Learning Objectives E06 and EO9.
, 2.
NMP-2: Ope-ating Procedure N2-OP-21, Main Turbine, pp. 22
i and 23.
i , (KA'S) 24100K4Q6 241000A107 241000A108 t.. s , - L \\ x.. - _ _ _. - _ _ -. _ _ '.. _ - _ _ _ - _ _ _}
W ' I,. [[ ' - {h, , c V. <, y - t .,, . , ,
- .
. . . ... . INSTRUMENTATION - PAGE' 19 + j6 __P!:AN1_@y@ led @_DE@l@dt" CONTROL _AND
1 ANSWERS -- NINE'.' MILE POINT 2'- - -rJ.7 /07 /07-J ANRELL,1 5.
J -
< - .n.
, !
,.( l
.. h ' ,.(2.40F
' ANSWER. '6.02 .Y ). Ag delta T bebs}een steam dome and recirc ' loop suction < < 10.7 W-l 1.. . ), degrees F for 15 seconds % ^] g S ' 2.
-feed flow < 30% rated for 15 seconds 3.
RPV level' level 3 . .
f, 4.
EOC-RPT signal present ,.,, RRCS high rea, tor pressur u signal' present (1050 psig) 5.. c _ _ _ ll _ , '6.
' during ' a. low speed. start-t hen pump speed reaches 95% (exact N ,
.setpoints not required) fp .y,%(porkees cM <j U + 7.
MT stop.' valve closure (<90% open) or control valve closure (EHC < 530 psi) with Rx power > 30%. . . Any'six' (6) C+0.41, +2.4 maximum.
REFERENCE ,e 1.
N2-OP-29, p.
6.
g.
-, 202OO2K402 ...(KA'S) ..e ANSWER 6.03 (2.50) a.
Both CSTs; nor.eial valve lineup has the tank cross-corinect valve in the open position C+0.53.
b.
1.
The HPCS suction valve to the CST (B) would shut (2CSH*MOV101)fL+0.53 and the HPCS suction to the suppression pool would openj(2CSH*NOV118) [40.53.
ould (21CB*NOVPZ9) hl6 2.
The RC W suctipn-valve - CST ( Aj) RCI uctipn dilve M the [40,53 f'al1pW1ng'th pening of i , g/, suppressitbn poO3 21CS*yOVI36) .53.
fg SPC5 )[, er* I Is REFERENCE gud k) 1.
NMP-2: Lesson Plan, RPS, pp. 7 and 8, EO-3.
212OOOK201 212OOOK601 ...(KA'S) - g 9 g/ i
- + , ______m___ _ _ _ _. _. _ _
- - _ - _ _.. _ _ %, )x$,e - . i y. 6.
" PLANT SYSTEMS' DESIGN _CONTR0 6 AND,I_NSTRUMENTAT10N PAGE 20- ,.
. ANSWERS --.NINE MILE POINT.'2' -87/07/07-JARRELL, D.
, . t*; 2. '00 ' . ANSWER
- 6.04 02<e07'
,. p, /d. ' Load sequencing starts when the diesel generator output breaker .< YY - shuts [+0.53- ,g T-' < . ' b. 'ii' RHS pump A (oriLPCI pump A) . L CSL pump _(or LPCI pump) . . Service Water Pump.
f+p,60),g 7 7[Q @_ .. . [.~ -E-+0c25 ] c2ch r-phss-EWE 3 - f or correc t crdcr. "- ! l c.
l '. engine overspeed C+0.53-p !2.
generator differential [+0.53 , Afa . n li REFEiENCE ' 1.. NMP-2:-Operations Lesson Plan, Standby Diesel Generators, p.
20.
2.
NMP-2: Operations Lesson Plan, Emergency AC Power Systems, p.
6.
, 264OOOK402 264000K405 ...(KA'5) l., 1!. l . ANSWER ~ ~ 6.05 (2.00) l'. reduction in electrical output ' *i.. steam f low to f eedf low mismatch u l 3. - Rx vessel level' change '4.
suppression pool temperature increase ' E40.53 each REFERENCE-l 1.
NMP-2: Operating Procedure N2-OP-34, Section 8, " System Description," p.
9.
239002A105 239002A107 239002A108 239002A109 ...(KA'S) \\ .,. ANSWER 6.06 (1.60) l l . i :0.Mf The indicated level would be zero (due to no back I- [ M r c gSLC pumps.a.eee # trip on tart ' -- "^ ' pressure) C+0.553 h.4 f 6 eg 'l.stunal C+0.553.
3en 'Y cau.s from (re dwA*N) h P c ell drae sheets.h '
sji' '_ji' _ _. . - _.. _ - _ _ _ - _ _ _ _ _ _ -
- -_ . . .. ht__PLONI_gyglEdg_QEgigN _CgNIBgL _@NQ_lNglgydENIOllgN PAGE
t t ANSWERS -- NINE MILE POINT 2-07/07/07-JARRELL, D.
.. REFERENCE 1.
NMP-2: N2-OP-36A, p.
2.
211000K103 211000K503 211000KSO6 211000K601 ...(KA'S) ANSWER 6.07 (1.00) a.
False U.
Tr u= * FALS&~q 7 'y q) [+0.53 each REFERENCE 1.
NMP-2: N2-OP-34, p.
2.
NMP-2: Lesson Plan, p.
16.
210000K501 ...(KA'S) i ) . lt _ . _ _. _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _
_ __ . - _ -.. f . J.
PRgCEggBEg_;_NgBM8(t_OgNgBd@(t_EDEBgENCy_ONg PAGE
B091gLQQlC@L_QgNIBQL l , ANSWERS -- NINE MILE POINT 2-87/07/07-JARRELL, D.
ANSWER 7.01 (2.50) a.
1.
Reactor feedpump 2FWS-P1A (B,C) trips - 45 sec/18 sec TD (0.5) l ! 2.-Auto start of standby condensate booster pump (0.5) (if FW pump suction pressure drop to 210 psig) b.
1.
The feedwater pump will trip after a 45 second time delay (0.5) if the suction pressure is between 190 and 210 psig.
2.
The feedwater pump will trip after an 18 second time delay (0.5) if the suction pressure is below 190 psig.
c.
If feedwater pump suction low pressure occurs after a main (0.5) turbine trip, the FW level control valves 2FWS-LV10A (b,C) will be automatically shut to 48% open.
REFERENCE NMP-2 OP-OB, Feedwater System I 259001A201 259001A310 2590015008 ...(KA'S) ANSWER 7.02 (2.50) a.
No, an EDP has just been entered [40.53 (safe plant operation).
b.
suppression pool temp (T > 90 degrees F) [40.53 c.
DWT, SPT, PCP., and SPL (EOPs) C+0.253 each, +1.0 maximum ente 7f1hl d.
all eterb anditions (f or all four procedures) must be cleared [+0.53 REFERENCE 1.
NMP-2: N2-EOP-DWT, Drywell Temperature Control; Lesson Plan, p.
4.
295027SG11 ...(KA'S) i l
l _ _ _ _
_____ _ ___ - __ - _ -. ,Zt__EB99 Egg 8ES_ _NgBdO6t_8pNgSM86t_EDEB@ENgy_QNp PAGE
B09196991Geb_G9N1896 , ANSWERS -- NINE MILE POINT 2-07/07/07-JARRELL, D.
ANSWER 7.03 (1.00) 1.
Minimize dynamic loading on containment structures (f rom multiple ! SRV actuations).
2.
Reduce the possibility of a stuck open SRV.
3.
Equalization of suppression pool heating Any two (2) [40.5] each, +1.0 maximum.
REFERENCE 1.
NMP-2: N2-EOP-RP; Lesson Plan, pp. 10 and 11.
295025SGO7 ...(KA'S) ANSWER 7.04 (3.00) 1.
contamination levels greater than 10,000 dpm/cm**2 2, airborne radioactivity requiring the use of respiratory equipment 3.
neutron radiation exposure 4.
use of vacuum cleaners or portable HEPA in the Restricted Area 5.
High Radiation Area Entry 6.
unknown conditions in the area to be entered 7.
maintenance in a Radiation area Any six (6) [40.5] each, +3.0 maximum.
REFERENCE 1.
NMP-2: SRO Requal Exam, May 1986, Exam H1, O 7.2.
l ! 2940001K10 ...(KA'S) ! I l l i
' l
1
l l l l l l l \\ l I l l - --_---_- a
- -___-_-_-- .. 's,
' ' . PAGE
,Z._ PRQCEDUREg_ _NgRMAL _AgNQRMAl _ EMERGENCY _AND t t RODI9LggICOL_CgNiggL ANSWERS -- NINE MILE POINT 2-87/07/07-JARRELL, D.
, ANSWER 7.05 (2.00) 1.
verify (all service water pumps have tripped and) that no service water pumps can be restarted .2.
reduce recirculation flow to minimum 3.
scram the reactor (f ollow scram procedure N2-OP-101C) 4.
trip the main turbine 5.
shut the MSIV's (manually initiate RCIC) 6.
trip the Rx recirc pumps 7.
trip the RWCU pumps D.
notify the SSS [40.333 each, +2.0 maximum.
REFERENCE 1.
NMP-2: Operating Procedure, N2-OP-11, " Service Water System," p.
13.
295010SG10 ...(KA'S) ANSWER 7.06 (2.00) a.
open [+0.53 b.
10 min.
[+0.53 c.
Provides maximum water flow to the vessel [+0.53 uses the high heat capacity of the suppression pool for cooling [+0.53 (during the interlock period).
REFERENCE 1.
NMP-2: N2-OP-31, Ressdual Heat Removal System, p.
2.
223OO1SGOS ...(KA*S)
l l I _ _ - _ _ - - _ _ _ _. _
_. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -. _ __ -. _ _ _ _ _ -. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _. _ _ _. _ " - .. . Ll .,Z __PBggEggBEg_ _NggdS6t_QgNggd@6t_EDEggENgy_@NQ PAGE
00919LgglgG6_ggNIgg6 , ANSWERS -- NINE MILE POINT 2-87/07/07-JARRELL, D.
..
- ANSWER 7.07 (2.00)
Run the turbine trip throttle valve to the f ully closed position (green light) [+1.03 and then run the valve to the fully open position (red light) ensuring the valve position. indication (red light) is ) $*hhs'J.
also lit [+1.03.
, , je REFERENCE pS g 1.
NMP-2: N2-OP-35, Reactor Core Isolation Cooling, Section G.5.
217000A402 ...(KA*S) ! _ _ _ _ _ _ _ _ - _ _
_ _, ' ',,
.
. . G_ AQMINiglg0llyE_P60CEDygE@t_ CONDITIONS AND LIMITATIONS PAGE
i r u ! ANSWERS -- NINE MILE POINT 2-07/07/07-JARRELL, D.
l' .
j ' I , l t 2. 5' o ANSWER 8.01 63rOO) ) The upscale trip of 122g. renders APRM "C" inoperable per Table 2.2.1-1 (item 2.tn"2) [+0.53 and, with Channel "A" also I declared inop, the RPS minimum operable chanc.el criteria of (z.c) Oj l p the surveillance [+0.33h / tng / Table 3.3.1-1 are not met C+1 g Per Table 3.3.1-1 note (a) / (channel may De placed INOP 4or 2 hrs during .s-I
f unen it must be placed in a tripped condition within i hrTiDr a.a.1 yg action a.
set a u.ra 1 and-4-em60w app 4-imabh. Au i. i ca-4-in p
, mor.t-restr-ie 4-ve-4be-in-GTARTUP-wi-thin-6-hr-M--t4.-5JA c,&, 6 [+ l. o ] 1,g7 REFERENCE 1.
NMP-2: Technical Specifications, Tables 2.2.1-1 and 3.3.1-1.
215005SG11 ...(KA'S) ANSWER G.02 (2.00) a.
No b.
No c.
Yes d.
Yes [+0.53 each REFERENCE 1.
NMP-2: Technical Specifications, Section 1.7.
234000SGO1 ...(KA'S) ANSWER B.03 (2.00) No [40.53.
Survei l l ance requi remen t 4. 8.1.1. 2. a. 7 0+0.53 cannot be met [+1.03 (air start receiver pressure is greater than or equal to 225 psigi.
REFERENCE 1.
NMP-2: Technical Speci f i ca ti ons, Sections 1.27 and 4.0.1.1.2.a.7.
264000SGO4 264000SCOS ...(KA'S) - .___ -_-____ -
c- - - - - _ - - _ -. _ _ _. - _ - - _ - - - - _ . w 9:__0DDINISIB@llyE_PSOCEDUBES._CgNDillgNgt_QND_Lldll@llgNS PAGE
3 ANSWERS -- NINE MILE POINT 2-87/07/07-JARRELL, D.
t ANSWER 8.04 (1.00) Minimum level assures sufficient volume of water to absorb maximum LOCA energy release E40.53.
Maximum temperature limit provides complete quenching of blowdown steam C+0.253 and provides ECC pump NPSH C+0.253.
REFERENCE 1.
NMP-2: Technical Specifications, Sections 3.6.2.1.a.1 and D3/4 6-3, 6-4.
295026SGO4 295030SGO4 ...(KA'S) ANSWER 8.05 (2.00) In the as found condition, the RCIC system is OPERADLE [+0.53.
Action Step 3.6.3.a applies so the val ve must be closed within 4 hours C+O.53.
This action causes RCIC to be INOP C+0.5] and (assuming that HPCS is operable) a (L4" day) LCO is in ef f ect per Section 3.7.4 acUpo b.
g C+0.53.
3 ~ if REFERENCE 1.
NMP-2: Technical Specifications, Section 3/4.6.3, pp. 21 through 36.
217000SG11 273OO2SG11 ...(KA'S)
!
l l l l l , _.______ _ ___ _ _ - - -
o ! . . . ' i.
l - . PAGE
,0 __ ADMIN 1gTRATIVE_PRQC_EQUREgt_CgNgITIgNgt_ANg_LIMITATIgNQ l
l-l ANSWERS -- NINE MILE POINT 2-87/07/07-JARRELL, D.
.. ' ANSWER O.06 (2.00) Operability - 1.
either recirculation loop flow differs (by > 10%) from given speed / flow characteristic 2.
total core flow differs (by > 10%) from measured norm 3.
diffuser-to-lower plenum delta P of any individual jet pump differs from norm (by > 10%) i Any-two (2) [+0.53 each, maximum +1.0.
Bases -
1.
Increases blowdown area during LOCA C+0.53 2.
reduces capability of reflooding core following a LOCA E+0.53 REFERENCE 1.
NMP-2: Technical Specifications, Sections 3.4.1.2, 4.4.1.2, and B 3/4 4-2.
202OO1SGOS 202OO1SGO6 202OO1SG11 ...(KA*S) ANSWER O.07 (1.00) ! Should the situation warrant, the Station Shif t Supervisor (SSS) l C+0.53 may authorize the RWP process via verbal command only [+0.53 (i. e., no immediate documentation is needed).
REFERENCE 1.
NMP-2: S-RP-2, Radiation Work Permit Procedure.
294001K103 ...(KA'S) i
I
1 l l l l l __-__----__ _ _ _ b
_ _ _ _ _ - - -. _ _ _ - - - - .. _ ., '.. . s ' . - .
- Bi__6pMIN1@l8611VE PRQCEQURE@t_CgNDITIQNQ,_ANp_LIMITATIQNg.
PAGE
i ' ANSWERS -- NINE MILE POINT 2-07/07/07-JARRELL, D.
. l l ANSWER O.00 (2.00) 1.
Yes [+0.53, RCIC in INOP.
2.
Yes 0+0.53-(dropping below 150 psig does meet the action statement for RCIC operability requirements).
I i 3.
Yes [+0.53 (other TS sections. apply).
Section 3 3.7.4 action !
b (per Table 3.3.7.4-2) places the plant in a 7 day, LCO then requires being in hot shutdown within 12 hours [+0.53.
REFEHENCE 1.
NMP-2: Technical Speci f i cati ons.
209002SGO5' ...(KA'S) !
l l l l l <
__- 7.- q .' ' ? '
, l - -. TEST CROSS REFERENCE PAGE
'*! GUESTION' VALUE REFERENCE
- -
I 05.01 3.00 DANOOOO749 ' 05.02 1.50 DANOOOO750 05.03 1.50 DANOOOO751 l 05.04 2.50 DANOOOO752 05.05 2.50 UANOOOO753 05.06 2.50 DANOOOO754 05.07 1.50 UANOOOO755
15.00 i 06.01 3.00 DANOOOO756 06.02 2.40 BANOOOO757 06.03 2.50 UANOOOO750 06.04 2.50 DANOOOO759 06.05 2.00 DANOOOO760 06.06 1.60 UANOOOO761 , 06.07 1.00 DANOOOO762 __ 15.00 07.01 2.50 DANOOOO747 07.02 2.50 UANOOOO763 ! 07.03 1.00 BANOOOO764 107.04 3.00 DANOOOO765 07.05-2.00 DANOOOO766 07.06 2.00 DANOOOO767 07.07 2.00 DANOOOO760 - -_-- 15.00 00.01 3.00 DANOOOO769 00.02 2.00 DANOOOO770 08.03 2.00 DANOOOO771 08.04 1.00 DANOOOO772 08.05 2.00 DANOOOO773 08.06 2.00 DANOOOO774 08.07 1.00 DANOOOO775 08.08 2.00 DANOOOO776
15.00 - ----
GY-60.00
_ _ _ _ _ _ - -
l PAQ C0in%dT OM- .., , 89 put , 1.03 Tolerance band on 1004 equilibrium Xenon and peak Xeno-following scram should be allowed.
Previous texts used at this i , i facility have used -2.2% for 100% equilibrium and -4.7% for peM
l Xenon following a scram from 1001 power.
These values were used by some students and should be acceptable.
i y= 7 g 7.a 1.07b Stored energy (heat) contained in metal mass of the Reactor,
I should be acceptable as another reason for why thermal power does not decrease as rapidly as neutron power.
I 1.08 b and d.
Most students considered the change in efficiency to be negligible as compared to part a and selected No Change as the answer.
This should be accepted.
Since No Change was a
choice given in the question.
2.02 Answers a and b are grouped together as above/below refuel floor for low flow and high radiation.
Since the low flow or high _ radiation signal from either location (i.e.
above or below refuel floor) will initiate t andby Gas Treatment, they should be acceptable as separate answers.
This would make six separate answers.
2.03 Part a would require a logic print to answer correctly.
The logic for this system may or may not initiate on one high drywell and one low-low level signal depending on which channels initiated.
Students are not required to memorize which channels are required for this.
So some students may not know this answer.
This answer was changed to include a Not Necessary answer with support in Part b.
2.04 Since the question states component not system, the student must draw a conclusion on the difference and the intent of the question.
This would cause a range of answers for certain parts.
Part a.
Many included Rod Worth Minimizer and Rod Sequence Control System as components designed to limit the consequences of a rod drop accident.
Part d.
Emergency Core Cooling Systems are designed for this accident and should be acceptable as a correct response.
Key also corrected to include jet pumps.
t Co.
t -.a 2.06 See SRO response 6.04 - _ _ _ _ - _ _ _
- _ - _ _ - _. . . l 2/7 For response #4 MOV-ll8 ne+ fully open should be acceptable in , lieu of closed.
Key corrected to add minimum flow valve.
. su c-a ,.-s iv - L 3.01 SRM upscale and inoperative are separate trips as per Technical Specification 3/4 3-60 and 3/4 3-62.
They should be considered separately on key making four possible answers of which only three are required.
3.02 For response #5 in lieu of Neutron Monitoring System, accept either APRM thermal trip or APRM neutron trip.
3.06 a.
See SRO response 6.01 for part a.
c.
Some assumed setpoint failure so answer would be opposite.
Should be acceptable if technically correct; i.e.
Control valves and bypass valves open to limit of maximum combined flow and reactor depressurized to MSIV closure (low pressure, mode switch in run).
3.08 a.
Question is confusing since there are no triple low inputs to RPS.
If student assumed the key wanted range for triple low inputs to ECCS and PCIS then wide rage is correct response.
If student assumed level inputs to RPS were asked for (i.e.
level 3 scram called low level or single low) then narrow range is correct response.
Either range should be acceptable.
b.
Low water level inputs for ECCS initiation are double low and triple low from wide range for pump starts.
! Low level from narrow range used in ADS initiation as confirmatory low.
Either range should be acceptable.
4.v.
L CO.
't
^,
m 'i "
) I i f ___ - - - - _ - - - J
. _ _ _ - - - -. _ _ _ _ _ _ _ _ _ _ _ _ _ , . NRC EXAM COMMENTS SRO EXAM L.v Z: Ji '
- , g; comy n e u
- 5.03 The key answer is based on the procedure for normal ranging on a start up (or shut down) which follows standard guidelines to maintain indication between 25% and 75% of full scale.
The question, however, is based on a post scram condition.
Under this condition, it is allowed to just avoid another scram signal.
J Recommendation: Accept answer that states the lowest range could be range 1.
Reference: N2-OP-92, Neutron Monitoring, Rev.
1, P. 7 5.04 Part b.
The key answer states that generator load increase is due to steam being redirected to the turbine.
Another reason is that reactor power is increasing.
The response of the EHC system will be to open the TCV's, increasing load.
Recommendation: Accept as an answer that generator load increases due to an increase in reactor power and/or increased steam flow to the turbine.
5.C5 % C= = t.
l The key answer concerning reactor pressure states that
- 6.01 Part a.
it will remain constant.
The supplied diagram of EHC shows a small (3%) bias on the bypass valve control circuit.
This would cause pressure to increase slightly.
Recommendation: Accept for full credit that final pressure will be slightly higher due to the bias on the bypass control . ' circuit.
If it is assumed that this is negligible, then accept pressure remains constant.
Part b.
The key answer carries the discussion of the pressure control to af ter the scram.
Several examinees' addressed the effect on pressure up to the point of causing the reactor scram to occur.
It should be noted that no direction was given as to what extent to carry the discussion out to.
Recommendation: Accept for full credit that " pressure increases causing the reactor to scram (on high pressure or high flux).
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- 6.03 Part b.
(1) The key answer has the. auto swapping of the HPCS (and RCIC) pump suctions on CST low level.
This would occur if , level continued to decrease below the listed alarm setpoints.
The two setpoints in the question are 20 feet and 18 feet 7 inches, respectively, which the swap over occurs at 12.5 feet.
If the examinee limited his answer to these annunciators, a swap over would not occur yet.
(2) RCIC is not an ECC System, so should not be required.
Recommendation: If examinee limited his answer to response on the listed annunciators, accept that it would not affect valve positions.
If examinee assumed level continued to decrease, accept HPCS pump suction swap over.
Also, should accept an answer of "HPCS pump suction will shif t to the suppression pool" vice the key answer of naming valves (same concept).
Reference: N2-0P-4, Condensate Transfer & Storage, Rev. 1, P. 15, 19.
- 6.04 Part a.
Some examinees responded to this question by addressing the root cause initiation of load sequences - loss of or degraded bus voltage.
Part b.
This caused confusion among the examinees because three loads were specified in a condition where only one load would sequence on.
Some examinees assumed a LOCA was occurring due to the number of loads being sequenced on.
l Recommendation: Accept for full credit Service Water (and it's try at 3 pumps if necessary) or, the original answer key items due to several people assuming a LOCA must have been occurring.
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- 6.06 The answer key is wrong.
Instrument air supplies a back i indication o_rLy, the low level trip l pressure detector for level is from level transmitters independent of this.
I Recommendation: Accept that it will have no effect.
' Reference: N2-0P-36A, Standby Liquid Control, Rev. 2, P. 17, 23 and P&ID.
- 6.07 Part b.
Answer key is wrong, it should be false (corrected).
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There is an alternate answer to this part.
If an I . e>;aminee assumes that the low feed pump suction has tripped the
feed pumps, level will go to level 3 and a "setpoint setdown" occurs.
(This affects the level controller /LV valves).
, Recommendation: To part b, accept alternate answer as discussed.
Reference: N2-0P-3, Condensate and Feedwater, Rev. O, P. 6.
- 7.02 Part a.
The key answer states that turnover would not be allowed by our station procedures, an E0P entry condition does ng prevent shift turnover.
Recommendation: Accept answer that turnover can proceed (noting that caution is being exercised due to being in E0Ps).
Part b.
There is another answer to this question.
The stated service water temperature is 82.2*F which is above the tech spec limit of 76*F, forcing an entry into an action statement.
Several people knew this even though the appropriate tech spec was not supplied.
If this appears in the answer, being true, it should not get marked wrong.
Lack of it, however, should not count as a loss of credit since the spec was not supplied.
Recommendations: If Service Water is mentioned, realize it is true.
IF it is not, no loss of credit (since spec not given).
Part c & d.
Per part b.
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- 8.01 Answer key is wrong.
Per tech specs, the answer should be "per spec 3. 3.1 action a, place the inop channel (s) and/or that trip system in the tripped condition within 1 hour".
(corrected) " Recommendation: In lieu of " place that trip system.
., Accept " insert a half scram" (this is the same thing).
I Reference: NMP2 T.S., LCO 3.3.1 action A
8.02 Part b.
The movement of a special moveable detector, since it requires lifting of the detector, is considered a core alteration.
Also, the terminology confused at least one person who thought the component in question was something else.
Recommendation: Change answer to YES.
8.04 The answer key does not address maximum water level, as specified in the question.
Credit would not be lost for answering the bases for maximum level.
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- 8.05 The second half of this answer key is wrong.
The' limiting specification would be on remote shutdown instrumentation, '# ytec 3.3.7.4, which would require a plant shutdown in 7 days.
(corrected except for specification number.)
, 8.07 An al ternate.. answer is possible.
If the examinee assumes.the plant is in the emergency plan (" emergency" was stated in the question) then the Site Emergency Director can authorize the processing.
Recommendation: Accept for full credit either the shift supervisor or the Site Emergency Director.
Reference: NMPNS - EPP's EPP-15, Rev. 7, P. 5
- 8.08 Part 3.
If list the other LCO's, do not specifically have to say Yes to get full credit (it is implied by listing the LCO's).
Recommendation: Accept answers as explained above.
NOTE: Items with an asterisk (*) were discussed during the post exam review with the examiners.
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FACILITY COMMENT RESOLUTION ON NINE MILE POINT-2 R0 Exam '1.03 Question Resolution Facility comment has been reviewed.
The reference material provided prior to the exam stated the correct values for equilibrium xenon reactivity at 100% power and peak xenon reactivity following a reactor scram. The grader's tolerance band included -2.4% for equilibrium, and -4.7% for peak xenon. However, errors in previous texts do not justify acceptable answers.
The facility should understand that previous negative training is not justification for changing answer keys.
1.07b Question Resolution Facility comment accepted.
No change to the answer key is required since " decay heat" includes the heat capacity of core components.
1.09 Question Resolution Facility comment not accepted.
The facility comment refers to question 1.09, not 1.08.
The question did not ask the student to compare the magnitude of the change in parts "b," "c," and "d" to part "a."
Also, the magnitude of the change for part "c" is on the same order as part "b," yet none of the students selected "no change" and three of the four students answered " decrease" instead of the correct answer " increase." The "no change" responses are considered to be incorrect guesses.
2.02 Question Resolution Facility comment not accepted.
The reference materCal provided by the facility grouped "Above/below" as one item. The question only asked for three out of four initiation signals.
Revising the answer to three out of six initiation signals l substantially dilutes the original questions' difficulty.
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__ -. _ _ _ _ - _ _ . _ _ _ _ _ _ _. tU ., ., ~' 2.03 Question Resolution Facility comment accepted.
This comment was previously' discussed at the facility and the answer key was previously changed to include "not necessarily" for part "a" if that answer was support in part "b."
2'04 Question P.esolution . Facility comment not accepted.
The question specifically asked for a component.
" Emergency core cooling systems" is not considered to be a component.
It should be noted that ' " components" were listed by all four students for all four parts of the question, except one student listed "ECC systems" for part "d."
2.06 Question Resolution Facility comment not accepted.
Part "a": The question states a loss of offsite power has just occurred and asks what event initiates the load sequencer. -The answer is clearly not " loss of or degraded bus voltage." Obviously on a LOOP where the EDG did not start, the load sequencer would not initiate (even though a loss of or degrade bus voltage occurred).
Part "b": Based on previous discussion with the facility the answer key.
was changed to reflect the fact that only the service water pump starts on a LOOP.
Full credit would be given if the student stated his assumption that a LOCA was also occurring simultaneously. However, no student stated that assumption.
2.07 Question Resolution
Facility comment accepted.
This comment was previously discussed at the facility and the answer key was changed accordingly.
3.01 Question Resolution Facility comment accepted.
Although the question asks for "ALL" the conditions to be listed, and i.
, l the training material provided by the facility groups these conditions j ! into a set of three, full credit was given when only three of the four < { conditions were listed.
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.. l o ( , 3.02 Question Resolution j Facility comment accepted.
! No answer key change is required since the APRM thermal and neutron trips { are considered to be part of the neutron monitoring system.
" 3.06 Question Resolution Facility comment accepted.
i Part "a": The slight increase in pressure that would occur due to the bypass valve control circuit bias was assumed to be negligible and " pressure remains constant" was accepted for full credit.
Part "e": "Setpoint failure" instead of pressure regulator failure was accepted with a technically correct discussion of EHC system response.
3.08 Question Resolution Facility comment not accepted.
Part "a": The question asked about the reactor vessel water level instrument RANGE which provides inputs to RPS. Narrow range is the only range that provides inputs to RPS.
If the student had assumed the word " triple" was the key word and then inferred the inputs to ECCS were being I asked for (instead of RPS as stated), the student would have then been further puzzled when he read "b" which would then be a redundant question.
At this point the student should have realized part "a" is asking for RPS inputs or, if not, asked the examiner for clarification, as stated in rule #16 on the exam coversheet.
This question was not raised during the examination.
Part "b": " Low water level inputs for ECCS initiation" clearly means the low water level setpoints that initiate ECCS. The fact that those setpoints are " double low and triple low" or " low low and low low low," or " level 2 and level 3" or "108.8 inches and 17.8 inches" is not relevant to the question, and certainly not confusing.
Even though ADS initiation logic has a confirmatory Level 3 input from narrow range, no justification is provided for accepting " narrow range."
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, SR0' Exam [ J 5.03 Question Resolution ' i Facility comment not accepted.
l i The' comment cites reference material ~which makes no mention of action l during a scram situation.
N2-0P-1010, p.10, does require full insertion of SRMs and IRMs, and the selecting of IRM'on the recorder immediately _ following a scram. While no direct mention of reranging to-follow flux level is made, the reactor operator is charged with verifying normal scram behavior.to ensure that 1) the reactor is responding-to rod insertion (backup to rod position displays) and 2) that instrumentation is-functioning properly by observing IRM/SRM flux overlap.
. t 5.04b Question Resolution Facility comment accepted.- 6.01 Question Resolution-Part "a": Facility comment accepted.
Part "b": Facility comment reviewed.
The question clearly states "GIVE... steady state pressure following transient." The discussion must therefore continue until a steady state condition is reached. Some credit (+0.1) will be given for high pressure scram since this information is essential to arriving at the final condition.
- 6.03 Question Resolution Facility comment accepted.
! i 6.04 Question Resolution ' Facility comment reviewed.
This question was reduced in point value due to the confusion noted.
Credit was given for part b if service water pump was stated either separately or in conjunction with LOCA sequencing.
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i 6.06 Question Resolution Facility comment accepted.
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Facility comment accepted.
.7.01c Question Resolution
- Facility comment ~not accepted.
Review of;the question indicates a "feedwater pump low suction pressure" condition. A feed pump trip has not occurred, therefore the alternate answer.is not: valid.
7.02 Question Resolution 'Part "a".
Facility comment reviewed.
, Until the cause is established and corrective action is determined not to require immediate: action, turnover would not be allowed. Additionally per AP-4.0-10,-Section 7.2.1, proper notification of management and their concurrence is required.
If immediate action and proper. notification are described, credit will be given for an affirmative answer.
Part "b".
Facility comment' accepted.
'8.01 Question Resolution Facility comment accepted.
8.02 Question Resolution Facility comment not accepted.
Technical Specification definition 1.7, CORE ALTERATION, states that normal movement of special movable detectors is NOT considered a core alteration.
Fuel loading chambers are only moved by lifting the detector and therefore that is considered " normal movement." Therefore, by Technical Specifications, the answer key is correct.
8.04 Question Resolution Facility comment accepted.
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8.08 Question' Resolution Facility coment. accepted.
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