IR 07100027/2011006
| ML20149E969 | |
| Person / Time | |
|---|---|
| Site: | Clinton, 07100027  |
| Issue date: | 12/17/1987 |
| From: | Bishop M, Burdick T, Hanek J, Mcghee J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20149E965 | List: |
| References | |
| 50-461-OL-87-01, 50-461-OL-87-1, NUDOCS 8801140077 | |
| Download: ML20149E969 (134) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION III ;
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Report No. 50-461/0L-87-01
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Docket No. 50-461 L4ense No. NPF-62 <
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Licensee: Illinois Power Company 500 South 2th Street '
Decatur, Il 62525 !
Facility Name: Clinton !
i Examination Administered At: Clinton, Illinois
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Examination Conducted: October 27 through November 6, 1987 l I
Examiners: J. M. McGhee, EG&G
,t. Chief Examiner 5
/r_//7/97 Date / /
t J. F. Hanek, EG&G ,
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M. O. Bishop, EG&G
[ /E//7/# 7 Date '
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Approved By:
M.[ '
Tho'nkrs M. Burdick, Chief IDfIlh ;
Operating Licensing Section Datet i Examination Summary Examination administered on October 27 through November 6, 1987 (Report !
No. 50-461/0L 87-Ol(DRS))
Written and operating examinations were administered to 10 Senior Reactor !
Operator and 6 Reactor Operator candidate :
Results: All candidates passed the written examinations. One Senior Reactor :
Operator candidate failed the operating examinatio All other candidates :
passed the operating examinatio :
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PDR ADOCK 05000461 !
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. REPORT DETAILS : Examiners ,
"J. M. McGhee, EG&G EG&G O. Bishop,EG&G J. F.-Hanek,
- Chief Examiner Exit Meeting On November 6, 1987, an exit meeting was hel The following personnel were present at the meeting:
Illinois Power R. E. Wyatt, Director, Nuclear Training J. A. Miller, Manager, Scheduling & Outage Management E. J. Corrigan, Director, Qual. Engr. & Ve F. A. Spangenberg III, Manager, Licer.:ing & Safety J. S. Perry, Manager, Nuclear Program Coordination J. W. Wilson, Manager, Clinton Power Station R. D. Freeman, Manager, NSED M. W. Lyon Lead Instructor, Operations R.A.Schuitz, Director Planning & Programming D. Antonelli, Supervisor,, 0)erations Training
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K. A. Baker, Supervisor, I&E Interface R. F. Schelle, Assistant Manager, Plant Operations NRC P. L. Hiland, Senior Resident Ins)ector D. E. Hills, Operator Licensing, Region III The following topics were discussed in the meeting: The only generic area of weakness identified involved the use of
Emergency Operating Procedures (EOPs). Although the performance of the Senior Operator candidates during the simulator portion of the
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operating examinations indicated a generally good familiarity wit boththeprocedural-stepsandthe-basis,themajorityofthe a candidates did not open the E0Ps and use them to work through the casualties or to verify actions taken. The examiners believed that 1 more training emphasis should be placed on using procedures during and after events to verify corrective actio .
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a 4 During the operating examination, a candidate discovered the key contained in the Control Room uncontrolled key locker for the Standby Liquid Control handswitch (key #8) was not comaatible with the switch design. Licensee representatives noted that tie key had not been changed when the switch was modified. The correct key was to be obtaine The examiners noted with concern, the number of simulator / plant differences that are present in safety system simulation. These are autlined in detail on the Simulator Fidelity Report (Enclosure 4).
Licensee representatives outlined the program currently in place which will update the simulation using a performance comparison with actual plant dat . Examination Review The resolution of facility comments on the exams are included as Enclosure .
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J l f f U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CLINTON 1 REACTOR TYPE: BWR-GE6 DATE ADMINSTERED: 87/10/27 IXAMINER: MCGHEE. CANDIDATE INSTRUCTIONS TO CANDIDATE 1 l
Use separate paper for the answer Write answers on one side onl Staple question sheet on top of the answer sheet 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 six (6) hours after the examination start j
% OF I CATEGORY % OF CANDIDATE'S CATEGORY !
VALUE TOTAL SCORE VALUE CATEGORY ;
As',6 5 1 25.50 e6760 PRINCIPLES OF NUCLEAR POWER I PLANT OPERATION, THERMODYNAMICS, i
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HEAT TRANSFER AND FLUID FLOW ass 2Y.6% \
F6-GO &&WHF PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
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2 s*.13 25.00 2 5 . 00= INSTRUMENTS AND CiNTROLS as su 24.50 _E4.5^ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL .
CONTROL !
4 % Totals Final Grade All work done on this examination is my ow I have neither given nor received ai Candidate's Signature NASER C:PY
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. NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: I Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . Restroom trips are to be limited and only one candidate at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil only to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer . Print your name in the upper right-hand corner of the first page of each section of the answer shee . Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example, 1.4, . Skip at least three lines between each answe . Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl . Use abbreviations only if they are commonly used in facility literatur . The point value for each question is indicated in parentheses after the question and can be used as a gu!.de for the depth of answer require ,
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14. Show all calculations, methods, or assumptions used to obtain an answer j to mathematical problems whether indicated in the question or no i l
15. Partial credit may be give Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl . 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. Wnen you complete your examination, you shall: Assemble your examination as follows:
(1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are part of the answe Turn in your copy of the examination and all pages used to answer the examination question Turn in all scrap paper and the balance of the paper that you did not use for answering the question Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoke L F
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. 4 s. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pcgs 4
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW l
QUESTION 1.01 (0.75)
To permit reduction of reactor pressure to 100 PSIG during cooldown, the reactor vessel metal temperature must be reduced to WHAT temperature in order to prevent repressurization above 100 PSIG7 (Show all work for full credit)
QUESTIt .02 (2.00)
The reactor is taken to CRITICALITY from a cold condition and an 80 second POSITIVE period is attained: From control room nuclear instrumentation, HOW can the operator tell when the heating range has been reached?
(Rod position and recirculation flow are held constant.) (0.5) In WHICH of the following intervals was the heating range entered? (1.5)
(1) Interval 1 - reactor power increased by a factor of 6 in 143.3 second (2) Interval 2 - reactor power increased b> a factor of 3 iu 99.0 seconds (3) Interval 3 - reactor power increased by a factor of 5 in 128.8 second (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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.. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pcgs 5
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THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.03 (2.00)
During your Shift, an SRV inadvertently opens from 100% power and 1000 psi Use a Mollier Diagram or the Steam Tables to answer EACH of the following: (ASSUME A SATURATED SYSTEM AND INSTANTANEOUS HEAT TRANSFER) STATE the tailpipe temperature, assuming atmospheric pressure in the Suppression Pool and No Reactor Depressurizatio If the Suppression Pool Pressure were to INCREASE, STATE whether the Tailpipe Temperature would INCREASE, DECREASE, or REMAIN THE SAM If the sa3ctor starts to depressurize when the SRV opens, STATE whether the Tailpipe Temperature will initially INCREASE, DECREASE, or REMAIN THE SAME in relation to what it would have done if the pressure had not decreased, STATE the Reactor Pressure at which the Tailpipe Temperature would be at its HAXIMUM value (during the depressurization).
QUESTION 1.04 (1.50)
Reactor power is increased by control rod withdrawal. The void fraction increases 1.5% and the fuel temperature increases 40 degrees as the '
result of the rod withdrawal. CALCULATE the reactivity WORTH of the PORTION of the control rod that was withdraw SHOW ALL WORK AND STATE ALL ASSUMPTION (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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,. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Paga 6
. THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.05 (1.00)
Which ONE of the following thermal limits protects the fuel from clad rupture due to PLASTIC STRAIN (deformation)? APLHGR ; LHGR MCPR MAPRAT i
l QUESTION 1.06 (1.75) ;
EXPLAIN HOW the cooling lake functions to cool the circulating water as it passes throug Include in your discussion at least ;
TWO conditions which would limit the ability of the lake to cool !
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QUESTION 1.07 (1.50)
Briefly EXPLAIN the effect an increase in core age has on moderator temperature coefficient and WHY these changes occur. Include in the explanation, changes in the various core parameters affecting the temperature coefficien ,
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.. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pcgs 7
. THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.08 (2.75) Approximately WHAT percentage of neutrons from U-235 are born delayed? (0.5) The power generated by the reactor at the baginning of core life comes from U-235 thermal fission and U-238 fast fissio Later in core life, larger and larger fractions of power generation are produced by fission of what TWO isotopes? (1.0) HOW do delayed neutrons contribute to the control capability of a commercial reactor? (1.25 QUESTION 1.09 (2.50) What value of reactivity added to a core will cause a prompt critical condition? (0,5) List TWO methods of limiting reactivity insertion rates by control rods (to avoid Prompt Criticality) during NORMAL REACTOR OPERATIONS and LIST the system hardware designs which are intended to perform these function (2.0)
QUESTION 1.10 (2.00)
Given a constant fuel temperature, EXPLAIN HOW and WHY the Doppler Coefficient will change with an increasing void fractio (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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,. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Ptss 8
. THERMODYNAMICS, HEAT TRANSFER _AND FLUID FLOW QUESTION 1.11 (2.00) EXPLAIN HOW and WHY core flow will change when power is reduced from 100% to 854 by control rod insertion if the recirculation flow control valve position remains constan EXPLAIN HOW and WHY core flow will change when power is increased by control rod withdrawal during low power condition QUESTION 1.12 (2.00)
Exceeding thermal limits during a reactor transient can cause fuel damage via two failure mechanism List these TWO mechanisms and EXPLAIN what causes each to occu i QUESTION 1.13 (1.00)
MAPRAT is printed out on the Periodic NSS Core Performance Log (P1).
Answer the following questions concerning MAPRA What does MAPRAT represent? [ EXPRESS AS A RATIO] ;
1 What should the numerical value of MAPRAT be during normal operations? l
i QUESTION 1.14 (2.00)
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Reactor power is increased from 80% to 90% of full power by increasing recirculation flow. Did the average void fraction (steady state to steady state) INCREASE, DECREASE, or REMAIN THE SAME7 Briefly EXPLAIN your answe (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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I PRINCIPLES OF NUCLEAR __ POWER PLANT OPERATION,_ Pass 9
. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l
QUESTION 1.15 (0.75)
Choose the word in parenthesis which best completes the statement:
Increasing the amount of condensate depression from 8 degrees to 11 degrees will (increase or decrease) Clinton's thermodynamic cycle efficiency.
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- ~. l l PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pega 10 i SYSTEMS QUESTION 2.01 (2.75)
Concerning the Shutdown Service Water System (SX), answer EACH of the following: What are TWO potential sources of radioactive inleakage to the l SX system? (0.5) ,
, What loads from EACH division do not d.'scharge to the Ultimate Heat Sink thus, requiring caution during operation with low lake levels? (0.75) What THREE signals will automatically start the SX pumps? (SETPOINTS REQUIRED) (0.9) How is a flowpath for the A/B SX pumps ensured following an automatic pump start? (0.6)
QUESTION 2.02 (1.00)
What cre TWO methods of verifying HPCS is injecting into the vessel after it has automatically initiated?
QUESTION 2.03 (1.50)
The RWCU Drain Flow Regulator (F033) automatically closes at the two pressure signals listed below. For EACH of the TWO pressure signals (A. and B. below) answer the following questions:
1) WHERE is the pressure control signal sensed in relation to the regulating valve? (Upstream or Downstream of the valve) (0.5)
2) WHAT plant condition or situation is the auto closure designed to preverit? (1.0) PSIG decreasing pressure PSIG increasing pressure (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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.. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcas 11
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SYSTEMS QUESTION 2.04 (3.00)
List SIX of the eight automatic isolation signals which will activate a Group 1 isolation signal, in addition to the mancal initiation? [ INCLUDE SETPOINTS FOR FULL CREDIT)
QUESTION 2.05 (1.50)
What are THREE purposes of the Low Pressure Core Spray water leg pump?
QUESTION 2.06 (1.50) t What physically acts to CLOSE the extraction steam check valves ,
in the event of a turbine trip? (0.5) List TWO plant conditions, in addition to a turbine trip, which will close an extraction steam check valv (1,0)
QUESTION 2,07 (1.75) ,
During a loss of feed condition from low power, a control room operator MANUALLY armed and depressed the RCIC system initiation push-button on H13-P601 and left the panel to assist the reactor operato Moments later he returned to check system status and found that the ,
turbine was not running (System was in standby) and both the red and '
white lights were illuminate What are TWO possible explanations for the RCIC turbine status the operator found when he returned? (1.0) What does the RED light indicate to the operator? (0.25) What does the WHITE light indicate to the operator? (0.5) i
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.. PLANT DESIGN INCLUDING SAFETY AND EtiERGENCY Pega 12
~ SYSTEMS
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QUESTION 2,08 (2.00)
Concerning the HSIV Leakage Control Syste List THREE permissives which must be satisfied to initiate the MSIV Leakage Control System?
(1.0) How will the components of the MSIV Leakage Control system react if the system is initiated with a MSIV still OPEN? (0,5)
' What function do the pipe heaters serve? (0.5)
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QUESTION 2.09 (2.00) What is the SETPOINT for the trip of a Turbine Driven Reactor Feed Pump on RCIC initiation? (0,5) List FIVE signals, in addition to the manual and RCIC initiation ,
trips, which will trip a Turbine Driven Reactor Feed fum (Setpoints not required] (1.5)
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QUESTION 2.10 (2,50) i
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' Which "modus" of RHR can be controlled from the Remote Shutdown Panel? (1.0) List SIX systems (in addition to RHR) which have controls or indications on the Remote Shutdown Pane (1.5)
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.. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pago 13 SYSTEMS QUESTION 2.11 (2.25)
LIST and EXPLAIN HOW Control Rod Hydraulic system design features, components, and/or interlocks provide the following functions; Constant control rod speed / system flow during normal rod movemen (0,5) Prevent pump runout while a scram signal is presen (1.0) Prevent an excessive pressure difference across the drive piston during normal rod movement following a scra (0.5)
QUESTION 2.12 (1.25)
What are the FUNCTIONS of the 105 second AND 6 minute time delays in the ADS initiation logic?
QUESTION 2.13 (2.00)
List SIX of the nine RHR system valves which receive CLOSE signals when Reactor water level decreases to the Level 3 setpoint, +8.9".
[ Specific valve numbers are not required for full credit, but if *
numbers are not included, description of valve should be specific enough to identify the valv,e.] A l gr ks os a IMa cae +.~3 [A alve. dpstQ j, A a.i.(
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.. INSTRUMENTS AND CONTROLS Pcgo 14
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QUESTION 3.01 (2.00)
List all signals or combinations of signals which will cause a !
Recirculation System runback to occur and INCLUDE SETPOINTS7 l
QUESTION 3.02 (2.00) What is the FUNCTION of the Alternate Rod Insertion / Recirculation Pump Trip (ARI/RPT) Syste List TWO signals that will automatically actuate the ARI/RPT syste (SETPOINTS REQUIRED)
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QUESTION 3,03 (2.25) i l
What THREE differential temperature interlocks in the recirculation pump starting sequence must be satisfied prior to starting a recirc ,
pump? (Include setpoints) :
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QUESTION 3.04 (1.50) i Explain HOW and WHY indicated reactor water level would respond to the following; (Assume all actual plant parameters remain unchanged, i.e. actual level, steam flow, and pressure) ,
1 Small leak in level transmitter reference leg isolation valve ;
packing gland which is constantly made up for by the condensing chambe !
! Equalizing valve for level transmitter leaks b !
i Drywell temperature substantially higher than calibrated condition l
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.. INSTRUMENTS AND CONTROLS Pcg2 15
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QUESTION 3.05 (1.00)
One recorder on panel H13-P870 is used to indicate main turbine shaft eccentricity, turbine shaft speed, and control valve positio What plant equipment alignments determine when EACH of the above THREE parameters is displayed on the single pen recorder?
QUESTION 3.06 (2.25)
EXPLAIN HOW and WHY the Setpoint Setdown logic circuit changes the feedwater control system controlling setpoint if reactor vessel water level decreases to Level 3. [ Include any time delays associated with the changes and describe the effects through steady state with no operator action.]
QUESTION 3.07 (2.50)
List FIVE actuation / trip signals initiated at Level 8 (+52") by the reactor vessel level instrumentation and the associated analog trip modules (ATMs). (Assume normal plant configuration at 100% load.]
QUESTION 3.08 (3.00) What are THREE of the four conditions / signals which will cause an APRM INOP trip? (1.5) List THREE actuations or signals which MAY result when an APRM INOP trip signal is generate (1.5)
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.. INSTRUMENTS AND CONTROLS Pcga 16
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QUESTION 3.09 (1.50)
List THREE conditions when the CRO can fully retract a Source Range detector without receiving a rod block?
QUESTION 3.10 (1.00)
Concerning the APRM Gain Adjustment Factors (GAFs);
, Of the following APRM GAFs, which is the MOST CONSERVATIVE? .01 .00 .99 What does an APRM GAF of 1.00 mean?
QUESTION 3.11 (1.00)
STATE the effects of resetting the Containment Spray ten minute time delay if spray was stopped by closing the heat exchanger outlet valv .
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QUESTION 3.12 (2.00)
Briefly EXPLAIN the logic associated with generating an automatic Group 1 isolation signal from sensor to isolation signal. Include the signals necessary to operate each group of valves, i
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QUESTION 3.13 (1.50)
What are THREE methods available to the control room operator to stop Recirculation Flow Control movement if the valve starts to ramp open while operating in Flux Auto Control?
QUESTION 3.14 (1.50)
Concerning the low-low set functions of the SRV l STATE the purpos Describe HOW it functions to perform its purpos Name the PARAMETER and SETPOINT that initiates low-low set, t
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. .. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Paco 18 AND RADIOLOGICAL CONTROL QUESTION 4.01 (2.25)
Answer EACH of the following concerning the Safety Tagging Procedure, CPS Procedure No. 1014.01; Who is the tagging authority? (1.0) What is a Safeguards Tagout? (0,5)
e Other than an Emergency Release by the Shift Supervisor, who may release a tagout? (0,5) What is the maximum number of holders allowed on a tagout which will still allow a temporary tag lift to be preformed? (0.25)
QUESTION 4.02 (2.25) ,
Concerning Off-Normal Procedure 4007.02, Inadvertent Red Movement; fill in EACH of the blanks below with the correct word or phras When the reactor is below the Low Power-Set point (LPSP) of (a.) %,
a (b.) will occur if a rod is selected and/or moved out of sequenc When operating between the LPSP and the High Power Set Point (HPSP) of (c.) %, rod withdrawal movement is limiten to (d.)
notche When operating above the HPSP, rod withdrawal movement is limited to (e.) notche If a rod is inadvertently inserted more than (f.) notches, the Nuclear Engineer should be consulted prior to restoring the rod to its proper position. The concern is that there'
are (g.) present in the area prior to withdrawal to insure that (h.) requirements are maintaine l l
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. PROCEDURES - NORMAL, ABNORMAL, EHERGENCY Pcg2 19 AND RADIOLOGICAL CONTROL QUESTION 4.03 (1.00)
Concerning Off-Normal Procedure 4100.01, Reactor Scra WHY does this procedure contain a step necessary to close the Charging Header Isolation Valve, 1011-F034, when attempting a manual Control Rod insertion WITHOUT RPS Logic reset?
QUESTION 4.04 (1.00)
CPS Procedure 3110.01 cautions the operator against TWO conditions which will cause a Turbine Generator runback to 25% generator load (7361 stator amps). LIST the TWO conditions.(Setpoints not required)
QUESTION 4.05 (1.50)
List THREE conditions specified in CPS Procedure 4004.01, immediate operator actions, which require a Manual Reactor Scram during a rapid depressurization of the Instrument Air Syste (INCLUDE APPLICABLE SETPOINTS)
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.. PROCEDURES - NORMAL, ABNORMAL,_ EMERGENCY Pcgo 20
AND RADIOLOGICAL CONTROL QUESTION 4.06 (1.50)
With the plant in a normal full power configuristion, it becomes necessary to evacuate the Main Control Room and Off-normal Procedure 4003.01, Remote Shutdown is implemente In addition to manually scramming the reactor, what ACTIONS should be accomplished prior to evacuating the Control Room? (1.0) The operator at the remote shutdown panel is cautioned to transfer control for only the systems required to be operated at the Shutdown pane WHY7 (0,5)
QUESTION 4.07 (3.00)
LIST the immediate operator actions for a Loss of Feedwater Heating, per CPS Procedure 4005.01. [Specify the order in which the steps would be performed]
QUESTION 4.08 (1.00) :
What are the TWO ENTRY CONDITIONS stated in CPS Procedure No. 4404.01, REACTIVITY CONTROL - EMERGENCY?
QUESTION 4.09 (0.50)
What does an orange dot on a control room annunciator indicate to the Control Room Operator?
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pega 21
- AND RADIOLOGICAL CONTROL QUESTION 4.10 (1.50)
Match the following MCR journal margin symbols (A through E below) with the appropriate definition (1 through 5).
SYMBOL DEFINITION Red T Indicates non-Technical Specification equipment Red I removed from servic Red E Indicates equipment that is inoperable for Red arrow reasons other than removal from service for Blue C maintenanc . Indicates that an earlier symbol is no longer
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vali . Indicates equipment removed from service for maintenance which will require retest per Technical Specifications before being considered operabl . Indicates an abnormality that requires special attention by those using the journal.
QUESTION 4.11 (1.00)
While executing a surveillance procedure, a step is encountered with a @ (circle "R") in the margin next to it. EXPLAIN WHAT this symbol signifies and WHO is authorized to initial this step?
QUESTION 4.12 (2.50)
According to CPS No. 1016.01, CPS Condition Reports, what are FIVE types of events or conditions which require an individual to submit a Condition Report?
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. PROCEDURES - NORMAL, ABNORMAL,_EKERGENCY Pega 22
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QUESTION 4.13 (1.00)
According to CPS Procedure 3309.01, what would the presence of steam or abnormally hot water, discovered in the HPCS discharge piping during system venting, indicate?
QUESTION 4.14 (2.00)
According to the CPS Procedure No. 1001.05, Authorities and Respon,1bilities of Reactor Operators for Safe Operation and Shutdown, the operator "at-the-controls" will manually initiate a reactor scram whenever one of two general conditions or situations exist. List these TWO conditions or situations?
QUESTION 4.15 (1.00)
CPS No. 3105.01, TURBINE, states it is possible to cause a rector scram during main turbine shell warming. EXPLAIN HOW shell warming could cause a reactor scram?
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pcgo 23
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AND RADIOLOGICAL CONTROL QUESTION 4.16 (1.50)
Concerning High Radiation Areas and the associated controls, answer Each of the following TRUE or FALS Initial entry into areas where dose rates have not been established may be made with a radiation dose rate conitoring device or a radiation monitoring device which integrates the dose and alarms when a preset integrated dose is receive High Radiation Areas which have no provision for locking shall be roped off, conspicuously posted and a flashing light activated as a warning device or continuously monitore In lieu of an RWP, continuous surveillance, direct or remote (such as closed circuit TV cameras) may be used to provide positive exposure control over activities in an area.
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EQUATION SHEET s
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f = ma v = s/t Cycle efficiency = "g* y'
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v = at s = v,c + lsat
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E 3C a = (vg - v,)/t 4g Kg = I gmy 2 y A = AN A = A,e g,y g4 ,g PE = mgh w = e/t A = in 2/tg = 0.693/tg ,
W = v4P' .
, (t.)(ts)
i .
AE = 931am .
(g + g)
6=AC,AT -rx
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,
I = I,e q = Uut
', I = I,e -Vx , ,
'Pvr = Wg$ I=I
-
10 */ M
,
g
.
p=p to SUR(t). TV1. = 1.3/u y=y o ,t/T HVI. * 0.693/u
'
'SUR = 26.06/T ~
T = 1.44 DT SCR = S/(1 - K,gg)
fA oh SUR = 26 CR = S/(1 - K,gg,)
6 o x 1(
~
aff)1 I eff)2 T = '(t*/o ) + [(i ' o)/ A,g,og
- ,
T = 1*/ (o - T) M * I/II ~ Eeff) = CR /CR g O I " II ~ 8)I Aeff' '
8"I M = (1 - Keff)0/II - Eeff)1 aff -1)I aff = AK,gg/K,gg '
3D3 , Cl ,K gg)jK,gg
~
p= [1*/TK,'gg .] + [I/(1 + A,ggT )] 1* = 1 x 10 ' seconds I
,
,
P = I4V/(3 x 10 0) 1,ggA? 0.1 seconds
~
E = No -
.
Idgg=1d22 WATER PARAMETERS Id =Idg2 g
1 gal. = 8.345 lbm 2 R/hr - (0.5 CE)/d (meters) i 1 gal. = 3.78 liters '
R/hr = 6 CE/d (feet) -
1 ft = 7.48 ga MtSCEI.I.ANEOUS CONVERSIONS ,
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Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps 10
Density = 1 gm/cm 1 kg = 2.21 1ha Heat of vst orizationi = 970 teu/lbm 1 hp = 2.54 x 10 ETU/hr !
Hese of fusica = 144 Btu /lbm 1 N = 3.41 x 10 Btu /hr
1 Atm = 14,7 psi = 29.9 in. I' Etu = 778 f t-lbf '
1 ft. H 2O = 0.4333 lbf/in 1' inch = 2.54 cm F = 9/5'c + 32
"C = 5/9 ('T - 32)
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pcga 24
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWER 1.01 (0.75)
100 PSIG + 14.7 PSIA = 114.7 PSIA (0.25)
Saturation temp for 114.7 PSIA is 338 deg. f i de (0.5)
REFERENCE CPS: Nuclear Power Plant Thermal Sciences, LO 3.1. NPPTS PP. 3-2 & 3-3 ANSWER 1.02 (2.00) Operator can notice that period has become longer (0.25)
and that power change on IRMs, SRMs is leveling off (turning around due to power overshoot). (0.25)
b. (2) (1.5) (From P = Poe(t/T) --> T = t/in (P/Po), in Interval 2 the period has lengthened from 80 seconds. The other intervals have 80 second periods)
REFERENCE '
CPS Introduction to Nuclear Reactor Operations, LO 4.1. '
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INRO PP. 4-17 & 4-18 !
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' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pego 25 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWER 1.03 (2.00) des F (+/- 15 des F) Increase Increase psia (+/- 50 psia)
[4 @ 0.5 ea.] (2.0)
REFERENCE Steam Tables /Mollier Diagram CPS Nuclear Power Plant Thermal Science, LO 3.1. BSEP Lesson Plan - Heat Transfer Chapter 4. Lesson Objective No. 3 and 5 from bottom of page 4-1(no number assigned)
,
ANSWER 1.04 (1.50)
Worth due to voids = (1 X 10E-3 dK/K/%V) (1.5%V) (0,5)
= 1.5 X 10E-3 dK/K Worth due to fuel temp. = (1.0X10-5 dK/K/F) (40 F) (0.5) l
'
= 0.4X10E-3 dK/K ROD WORTH = VOIDS + FUEL TEMP. = 1.9 X 10E-3 dK/K (0.5)
REFERENCE l
CPS Introduction to Nuclear Reactor Operations, LO 5.1. Reactor Theory Sec. 1 Pg. 16,14 and BSEP Lesson Plan 2A, Reactor Theory, Chapt. 14 pp. 172 & 181. Lesson Objective No. 5 I I
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.. PRINCIPLES CF NUCLEAR _ POWER PLANT OPERATION, Pcgo 26
- THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWER 1.05 (1.00)
b. LHGR (1.0)
REFERENCE CPS Nuclear Power Plant Thermal Sciences, LO 12.1. NPPTS P. 12-2 ANSWER 1.06 (1.75)
The process of removing heat energy from the bulk of the we.ter by evaporation cools the remaining circulating water (0.75).
Evaporative cooling would be impeded by the following; (any 2 @ 0.5 each) High relative humidity High outside air temperature Temperature inversion 4- Abnormal wind conditions REFERENCE r
CPS Nuclear Power Plant Thermal Sciences. LO 6.1. NPPTS PP. 6-29 THEU 6-32 ANSWER 1.07 (1.50)
i I
As the core ages, control rods are withdrawn for fuel burnup, and the result of this action is an increase in core size , a decrease in the negative effects of leakage (0,5). A decrease in the number of fuel nuclei as the core ages (0.25) and an increase in moderator-to-fuel ;
ratic (0.25) causes a positive trend to the total moderator temperature l
'
coefficient. (0,5)
s Ma is 7Xea.,J udkdda kemen se l deenase)%<paa ocecp/ ave(o.s)
. s eap w allowJe whileaerpeae:
lea ese.ye a siasi n{ b H n'.1e #, /*ye " slowig,' towa fe9fs, y%e ed mena'ej ,* gt' #'
effret a
deeAert s dicey'e i;#^
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,. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pego 27
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THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW REFERENCE CPS Introduction to Nuclear Reactor Operations, LO 6.1. INRO P. 6-17 GG, Reactor Theory, Ch. 4, P. 11 ANSWER 1.08 (2.75) .65% (also accept 0.64%) (0,5) Pu-239 and Pu-241 (1.0) Delayed neutrons increase the average neutron generation time (1.0)
{by a factor of more than 1000) Increasing the contrel timej of the reactor. (0.25) (1.25 REFERENCE CPS Introduction to Nuclear Reactor Operations, LO 3.1.1.2/3.1. INRO PP. 3-11 & 3-31 Grand Gulf Rx Physics pg. 31-34 Perry - Perry Introduction to Nuclear Resotor Operation Chapter 3. Pages 3-11 and 3-3 *
ANSWER 1.09 (2.50) Reactivity equal to Beta (0.5)
also accept : Beta. e m e.twe (.oe(or 0.75%))
o.n 7., . Rod withdrawal rates are limited (0.5) hydraulically by the throttle (needle) valves in the CRD system. (0.5) Rod worths are linited (0.5) by the Rod Pattern Controller and Preselected rod sequencing (0,5).
B. & holln i v3 a ge. receff aUe. Lesfemec S N E lhol w d R.a h oc h4f43 :
A c.4I s d ia e c. U n ( c a o b / -h a e a s b,a e.%.t ua%I ussesh % meMu3 valves
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
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Pcga 28 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW REFERENCE ,
CPS Introduction to Nuclear Ranctor Operations, LO 4.1.1.1/4.1. Student Handbook, Vol 10, LP 74034, LO INRO P. 4-31 & LP 74034, P. 14 ANSWER 1.10 (2.00)
Doppler Coefficient will become more negative (0.5). As voids increase, slowing down length increases (0.5) and the neutrons spend more time in the resonance energy band (0.5). More neutrons will be resonantly captured (0.5).
REFERENCE CPS Introduction to Nuclear Reactor Operations, LO 6.1. INR0 P. 6-40 RO EXAM BANK, Q 1.51 ANSWER 1.11 (2.00) The flow will increase (0.5) due to less 2-phase flow resistanc [0.5] (1.0) ,
i The flow will increase (0.5] due to increased natural circulation driving hea [0.5] (1.0)
REFERENCE
CPS Nuclear Power Plant Thermal Sciences. LO 9.1.1.1 P. 9-14 !
Standard Thermodynamic Principles, Reactor Operating Hap, 7681 l RO EXAM BANK, Q 1.37 l
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,. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pcso CD
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWER 1.12 (2.00)
(%wsiN bdel.uj) Severe overheating of clad (0,5) causedbyinadequatecooling/(0.5) Fracture of the fuel cladding (0.5) caused by the expansion of the pellet inside the clad (r.5).
REFERENCE CPS Nuclear Power Plant Thermal Sciences, LO. 10.1. NFPTS P. 10-10 ANSWER 1.13 (1.00)
APLHGR_ _A PL AG P~
AA%rts,%/
) MAPRAT = MAPHLGR/MAPHLGR limi MeuGR.3 M Less than REFERENCE CPS Student Handbook, Volume 6 LP 74011, LO LP 74011, P. 32 ANSWER 1.14 (2.00)
Decrease (0,5). In order to compensate for negative reactivity from doppler as fuel temperature increased (0.5) and from the moderator as subcooling decreased (0.5), void fraction must decrease to add positive reactivity to bring net reactivity to zer (0.5)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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I PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pass 30
- THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW l
l REFERENCE CPS Nuclear Power Plant Thermal Sciences, LO 9.1. NPPTS P. 9-8 .
ANSWER 1.15 (0.75)
Decrease (0.75)
REFERENCE CPS Nuclear Power Plant Thermal Sciences, LO 9.1. *
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcgs 31
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SYSTEMS ANSWER 2.01 (2.75)
& of ske leIlmht3e 0.25 e A. Anb RH heat exchangers -[&B&]*ene fuel poog8*5 ooling and cleanup heatexchangers.,d?.25]s c #M4 JkA/ (0.5) DIV I: Drywell chillers [0.15] and breathing air compresso [0.15]
DIV II: Drywell chillers [0.15] and breathing air compresso [0.15]
DIV III Pass sample pane [0.15]
[5 @ 0.15 ea.] (0.75) High drywell pressure (0.15] 1.68 psig [0,15]
RPV level 2 [0.15] -45.5" [0.15]
109 (lo Service water (WS) low pressure [0.15] -79'psig [0.15]
[6 @ 0.15 ea.] (0.9) The RHR heat exchanger bypass valve (ISX-173 A/B) [0.15] opens (0.15] if either of the RHR heat exchanger inlet or outlet valves (E12-F014 A/B or E12-F068 A/B) [0.15] respectively is off it open sea [0.15]
[4 @ 0.15 ea.] (0.6)
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REFERENCE '
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CPS: L. P. 72013, PP. 5, 6, 7, 9, 10, 11, 12 & 13. Enabling Objectives 1.4, 2.1 & i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paga 32
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SYSTEMS ANSWER 2.02 (1.00)
Any 2 of the following @ 0.5 each; Verifying pump flow is proper for indicated pump discharge pressur . RPV level is increasin . Injection check valve opens as pump discharge pressure reaches RPV pressure.
REFERENCE CPS Student Handbook, Volume 8, LP 74026, LO 1.10 LP 74026, P. 22 ANSUER 2.03 (1.50) Upstream (0.25) prevents draining an isolated RWCU system t Radwaste or condenser (0.5) Downstream (0.25) Prevents overpressurizing the Radwaste system (0,5)
REFERENCE CPS Student Handbook, Vol 11, LP 74039, LO LP 74039 SD (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pego 33
- SYSTEMS ANSWER 2.04 (3.00)
Any 6 of the following 9 0.25 for signal and 0.25 for setpoint; RPV Level 1 -145.5" MSL Radiation High or Inop 3 x FPB MSL Flow High 170 PSI (or 54" H2O) MSL Low Pressure 849 PSIG Main Condenser Low Vacuum 8.5 "Hg Vac MSL Ambient Temp High 165 deg. F MSL Vent Diff Temp High 54.5 deg. F Turb. Eldg. MSL Area Temp High 131.2 deg. F REFERENCE CP9 Student Handbook, Vol. 8, LP 74024, LO LP 74024, P. 22
<
ANSWER 2 05 (1.50)
Aq lbue d /Ae -fellnaig @ 05 wl, 1) Verifying system piping integrit ( 0. 5 )t-2) Shorten time between receiving the initiation signal and water entering the reactor pressure vessel, (0.5)"-
3) Prevent water hammer in the LPCS piping when the LPCS pump is starte (0.5) o 4) AHeevah iwjachs s3sb ft4 Ve5sel Moleap dudiNJ aedidewf cedibs.
REFERENCE ,
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CPS Student Handbook LP 74029, LO LP 74029, P. 9 0.ps 440I.ol , Leal Cock (~-Couenge l
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcge 34
- SYSTEMS ANSWER 2.06 (1.50) (Turbine Air Relay)
vents the air supplyDump to thevalve actuator on the turbine (0.25) front standargdir)i allowing spring 1 y gnadh pressure to close the valves (0.25). Any 2 of the following @ 0.5 each; 1. High-high level in the associated feedwater heate . Loss of instrument ai . Manual GWieor operation healex sfnary <of airoA-Wet solenoidedle t for testin valve s. 2c + full y de verse -f4w (ex 20 clean f/ov).
REFERENCE CPS Student Handbook, Vol 7, LP 73011, LO 1.4/ LP 73011, SD PP. 4 & 13 Ec2. - c699 sbes+ 9 , ao 2.-es99 SAce f t
.
ANSWER 2,07 ( 1. 7 5 ) ^ -(/,16) . - The crer=+~ did not hcid the button depressed until the "- '
b hetien culvo-had begun ;,o etrche spen (0.5). :-
F The system started, but automatically shutdown when RPV level reached Level (0.5) Red light indicated an initiation signal was present. (0.25) White light indicates the initiation signal is no longer present (0.25) and the red light may be reset (0.25).
REFERENCE CPS Student Handbook, Vol 7, LP 74018, LO LP 74018, P. 19 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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l PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 35 SYSTEMS ANSWER 2.08 (2.00)
6., 4Associated G!l_ ,9 e C.?? ^ N G=-
=a:( $c MSIVs - . must be closed (>90%), and RPV pressure must be less than or equal to 20 psig, and MSL pressure must be less than or equal to 20 psi $-
( 0 0 0 . 0 0 ) e.- (3 @ o39 Blower will tri (0.5) Evaporate any condensate prior to discharge to the SBGT system (0.5).
REFERENCE CPS Student Handbook. Vol 7, LP 74019, LO LP 74019. SD P. 10 ANSWER 2.09 (2.00) , Injection valve (E51-F013) not fully closed (0.25) and system flow greater than 120 gpm (0.25) Any 5 of the following @ 0.3 each; Low - Low NPSH (Low suction) Overspee . RPV High Water Level (Level 8). Low Vacuum Turbine bearing oil pressure lo . High thrust bearing wea . Pump bearing oil pressure lo (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pago 36 ;
- SYSTEMS REFERENCE CPS Student Handbook, Vol 3, LP 73012, LO LP 73012, SD PP. 15, 28, 29, & 155 ANSWER 2.10 (2.50) . LPCI 2. Shutdown Cooling 3. Suppression Pool Cooling 4. %(wyTGm 0.33 9d'3 each, 1.0 total) Any 6 of the following @ 0.25 each; RCIC Main Steam (SRVs) Shutdown Service Water Suppression Pool Diesel Generator Nuclear Boiler Instrumentation Containment Monitoring D/G Room HVAC SSW pump room Ventilation 10. Essential Switchgear heat removal 11. ECCS equipment room HVAC 12 4/4 %et Di\ %sfen s3c+ w REFERENCE ,
CPS Student Handbook, Vol 6, LP 74006, LO l LP 74006, PP. 4,%13 & 14 , 7'444E a j
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcgs 37
- SYSTEMS ANSWER 2.11 (2.25) Stabilizing valves (0.25) maintain constant flow through the pressure control valve (0.25), thus maintaining RPV/ drive water differential pressure constant (0.25). '(0.75) ) Restricting orifice-er.d throttic valvev(0.25) in the charging water header limits flow to less than 200 gpm (0.25). (0,5)
2) Flow element for Flow Control Valves is located betwson the pumps and the charging water header (0.25) so a high charging flow closes the FCVs (0.25). (0.5) Equalizing valves (0.25) repressurize the exhaust water header after a reactor scram (0.25). (0.5)
REFERENCE CPS Student Handbook, Vol 6 LP 74005, LO LP 74005, P. 7,8, & 13 ANSWER 2.12 (1.25) The 105 second timer allows time for HPCS [0.25] to reflood the Reactor Vesse [0.25] For transients and accidents which do not directly produce a high Drywell pressure signal [0.25] and are degraded by a loss of all high pressure injection systems [0.25] adequate automatic core cooling [0.25] is assured by actuation of the 6 minute time (Exact wording is not required) (1.25)
REFERENCE CPS: L. P. 74018 P.22, Enabling Objectiva l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 38
- SYSTEMS ANSWER 2.13 (2.00)
Any 6 of the following 9 0.33 each; Shutdown Cooling Upper Pool Isolation Valves (1E12-F037A(B)) Shutdown Cooling Injection Throttle Valve (1E12-F053A(B)) Shutdown Cooling Outboard Isol. Suction Valve (1E12-F008) Shutdown Cooling Inboard Isol. Suction Valve (lE12-F009) RHR Discharge to Radwaste Isolation Valve (1E12-F049) RHR Discharge to Radwaste Throttle Valve (1E12-F040) RHR Head Spray Injection Valve (1E12-F023) RHR Sample Valve (1E12-F060A) RHR Sample Valve (1E12-F075A)
REFERENCE CPS Student Handbook, Vol 9, LP 74030, LO LP 74030, PP. 45 - 55 I
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~ INSTRUMENTS AND CONTROLS Pega 39
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ANSWER 3.01 (2.00) Turbine driven RFP/F trip (0,5) and RPV level decreases to Level 4 (0.5) Trip of one cirewater pump (0,5) and low condenser vacuum of 7.5"Hg (0.5)
REFERENCE CPS Student Handbook, Vol 10, LP 74035, LO LP 74035, P. 23 ANSWER 3.02 (2.00) To prevent (0.25) and to mitigate [0.25] the consequences of an Anticipated Transient Without Scram (ATWS). [0.5) (1.0) High RPV Pressure [0.25) 1125 psi [0.25]
RPV low Water Level [0.25) (2) -45.5". [0.25) [4 @ 0.25 ea.] (1.0)
REFERENCE
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CPS: L. P. 74005, P. 14 Enabling Objective 1. l l
ANSWER 3.03 (2.25)
/ess o ,e. s Delta-T between loops (0.5) must be gr: tc/#than er cku+/ 4ael tvJ"-
50 deg. F (0.25) A,,, Delta-Tbetypen 50 essel dome and either loop (0.5) must be -greate?**
than E'r*?ihe_
b tween . bottom deg. F (0.25)
drain and dome (0,5) must be-greetc? %han A. ,, l l Dplgf1
<e e m_
4 5100 deg. F (0.25)
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~ INSTRUMENTS AND CONTROLS Pcgo 40 REFERENCE CPS Student Handbook, Vol 10, LP 74035, LO LP 74035, P. 13 , TS. 3,e/,/,/
ANSWER 3.04 (1.50) Indicated level would be higher than actual (0.25) because reference leg density will decrease as the water temperature rises and DP will decrease. (0.25) Indicated level would be higher than actual (0.25) because DP across the detector will decrease as the legs equalize. (0.25) Indicated level would be higher than actual (0.25). Both reference and variable legs will heat up, but reference leg is longer so temperature change would decrease DP across detector. (0.25)
REFERENCE CPS Student Handbook, Vol 7, LP 74014, LO 1.2/ LP 70414, PP. 4 & 16 ANSWEG 3.05 (1.00)
. Shaft eccentricity - when turbine is on the turning gea . Shaft speed - when turbine is off the turning gear and the generator output breaker is ope . Control valve position - when the generator is on-line.
REFERENCE CPS Student Handbook, Vol 4, LP 73016, LO LP 73016, SD P. 24 & 25 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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. INSTRUMENTS AND CONTROLS Pega 41
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ANSWER 3.06 (2.25) The setpoint is immediately raised to +40 inches (0,5) to counter the level decrease due to void collapse and to restore normal level (0.5). After 10 seconds (0.25), the setpoint is lowered to approximately
+18 inches (0.5) to prevent overfilling the vessel and reaching Level 8 (0.5).
REFERENCE CPS Student Handbook, Vol 4, LP 75013, SD P. 11 ANSWER 3.07 (2.50) Main turbine tri . Reactor scra . Feed pumps tri . Close RCIC steam supply valves. (E5I F09F and 651-FMr) Close signal to HPCS injection valve. (Eat -Foo4)
(5 @ 0.5 each)
REFERENCE CPS Student Handbook, Vol 7, LP 74014, LO LP 74014, P. 13 l
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,. INSTRUMENTS AND CONTROLS Pcgo 42
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ANSWER 3.08 (3.00) l Any 3 of the following @ 0.5 each; Less than 16 LPRM inputs to the channe . APRM mode switch nct in "Operate". Flow channel mode switch not in "Operate". Any inte W ,rnal podulp(gPRM,gr f1,oy,changel),upplugge ,
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,_ ,,,,..x,. -, :. Any 3"of the fTfow1Yd7@ W.T'eEcW;""7 ""*' 4' Reactor scram (2 or more channels) Rod Block Annunciator on P6SO 3.'
M- Saf*
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m kn.Ybwf'da
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e
"Sysf* k/e Y1*d L e.fdf ^D'& AAsUas fa / Cpfy Nm Ahr O N.& S'*l k
RFW.RENC[ "t "" M"" A l-
i CPS Student Handbook, Vol 12, LP 74046, LO 1.6/ l LP 74046, PP. 9 & 10 )
L.f TA C W , S l> 1.0 3I ANSWER 3.09 (1.50)
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Any 3 of the following at 0.5 each; IRMs on Range 3 or abov . Selected SRM channel is bypasse . Reactor mode switch is in RU . Channel count rate remains greater than 100 cps and less than 10E5 cp REFERENCE CPS Student Handbook, Vol 11, LP 74040, LO LP 74040, P. 13 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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, INSTRUMENTS AND CONTROLS Paga 43 ANSWER 3.10 (1.00) (0.99) (0.5) Reactor power indicated on the APRM channel is the same as that calculatedbytheheatbalanc (0,5)
(coe.e dewc( (pewer)
REFERENCE CPS Student Handbook, Vol 6, LP 74011, LO LP 74011, P. 12 RO EXAM BANK Q 2.16 ANSWER 3.11 (1.00)
The heat exchanger bypass valve (F048) will open and reestablish flow to the spray headers.
REFERENCE CPS Student Handbook, Vol 9, LP 74030, LO 1.29, P. 6 LP 74030, P. 67 l
l ANSWER 3.12 (2.00)
Four sensor channels feed 2/4 logic in each of the 4 division (0,5) A 2/4 trip from the sensor logic will trip a division. (0.5) A division 1 and 4 trip will isolate outboard valves (0,5), and a division 2 and 3 trip will isolate inboard valve (0.5)
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. INSTRUMENTS AND CONTROLS Pega 44 REFERENCE CPS Student Handbook, Vol. 8, LP 74024, P. 6 ANSWER 3.13 (1.50)
Any 3 of the following 9 0.5 each; Take manual control of valve (either Flux or loop manual) Shutdown HPU using manual pushbuttons at H13-P68 . Shutdown HPU at H13-P614 by placing both subloops in Maintenanc . Have auxiliary operator open supply breakers to HPU pumps and fan motors.
REFERENCE CPS Student Handbook, Vol. 10, LP 74035, LO 1.27 LP 74035, P. 24 ANSWER 3.14 (1.50) The low-low set system reduces the number of SRVs cycling following'
any overpressure transient to reduce the duty on the containment.(0.5) ,
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' By lowering the opening and closing setpoints of two SRVs (0.26) and reclosing setpoints of three more SRV' (0.25) ;
//03 Armed from RPV pressure (0.25) at 4446'PSIG. (0.25)
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, INSTRUMENTS AND CONTROLS Pcga 45
REFERENCE CPS Student Handbook, Vol. 4, LP 73015, LO LP 73015, SD P. 14, 47, & 48 CPS St ol, e t
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 46
- AND RADIOLOGICAL CONTROL ANSWER 4.01 (2.25) The on duty Shift / Assistant Shift Supervisor [0.5) and/or the Duty Radwaste Superviso [0,5) (1.0) Tagouts which disclose the type or function of equipment which have a direct / indirect impact on the integrity of the CPS security syste (Exact wording not required) (0.5) The individual job supervisor to whom a tagout is issue (0.5) Two (0.25)
REFERENCE CPS: A. P. 1014.01, Sections 2.2, 6.4 & PP. 4, 6 & ANSWER 4.02 (2.25) Rod Block (fo) No [0.25] Voids [0.25] Preconditioning [9 @ 0.25 ea.] (2.25)
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4 .' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pego 47
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AND RADIOLOGICAL CONTROL REFERENCE CPS: ONP 4007.02 Suud fla ce,e fo 3 o,os ANSWER 4.03 (1.00)
It is necessary to isolate this flow path to develop a drive differential pressur (or) Following the Scram the CRD flow control valve will be shut directing full flow to the HCU' (1.0)
l REFERENCE CPS: ONP 4100.01, P.15 ANSWER 4.04 (1.00) Stator coolant outlet high temperatur . Low stator cooling water flow (low pressure). [2 @ 0.5 ea.] (1.0) l REFERENCE j
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CPS: OP 3110.01 P.1 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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4l PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Paga 48
- AND RADIOLOGICAL CONTROL ANSWER 4.05 (1.50) If the air pressure drops to 30 psig [0.25] and cannot be restore [0.25] (0.5) The SDV level increases [0.25] to the Rod Block Setpoin [0.25]
(0,5) The Control Rods begin to drif (0.5)
REFERENCE CPS: ONP 4004.01 ANSWER 4.06 (1.50) . Verify all rods are fully inserte . Sound Containment Evaluation alar (2 @ 0.5 each, 1.0 total) To preserve as many automatic functions (0.25) and interlocks as possible. (0.25)
REFERENCE CPS Procedure No. 4003.01, Remote S/D (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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4.~ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pego 49
- AND RADIOLOGICAL CONTROL ANSWER 4.07 (3.00) If Recire Pumps are at fast speed, Recirculation flow is decreased until power is reduced by 20% below pretransient power level (or minimum flow control valve position is reached) (1,0) Insert control rods to reduce power (to provide scram margin to APRM setpoints) (1.0) Verify automatic actions occur as require (0.5)
Steps 1 and 2 must be performed in that orde (0.5)
REFERENCE CPS Procedure 4005.01, P. 2&3 RO EXAM BANK Q. 4.05 ANSWER 4.08 (1.00) Any condition exists which requires a reactor scram and reactor power is greater than or equal to 3%. (0.5) Any condition exists which require a reactor scram and reactor '
power cannot be determined. (0.5)
REFERENCE CPS Procedure No. 4404.01, P. 2 l
l ANSWER 4.09 (0.50)
The annunciator is disabled (also accept; Out of Service / Disabled)
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4 .' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Peso 50
' AND RADIOLOGICAL CONTROL REFERENCE C7/S Procedure No. 1406.01, P. 5 ANSWER 4.10 (1.50) ( 5 0 0.3 each)
REFERENCE CPS Procedure No. 1401.01, P. 19 ANSWER 4.11 (1.00)
The step is a Radiation Protection Hold Point (for ALARA) (0.5)
and can only be signed off by qualified Radiation Protection personne (0.5) ,
REFERENCE CPS Procedure No. 1024.65, P. 10 l-1 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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4 .' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pcgo 51
- AND RADIOLOGICAL CONTROL ANSWER 4.12 (2.50)
Any 5 of the following @ 0.5 each; Technical Specification violation Safety-related procedure violation Quality-related procedure violation Emergency Preparedness procedure violation Violation of any regulations, codes, or standards Discovery of unauthorized modifications Inadvertent trip occurs Design change is indicated upon completion of a MWR Significant Radiological occurrences 1 Significant Security occurrences REFERENCE CPS Procedure No. 1016.01, P. 6 ANSWER 4.13 (1.00)
Possible back leakage from the RP REFERENCE ,
CPS Student Handbook, Vol. 8, LP 74026, LO SOM, procedure #3309.01, P. ANSWER 4.14 (2.00) The safety of the reactor is in jeopardy and scramming would mitigate this condition, or 1' A scram setpoint is exceeded and the automatic action did not occur.
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4 .' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 52 l
, AND RADIOLOGICAL CONTROL REFERENCE CPS Procedure No. 1001.05 RO EXAM BANK, Q. 4.68 ANSWER 4.15 (1.00)
A reactor scram will occur if first stage shell pressure reaches approximately 175 psig (0.5) with the stop valves closed. (0,5)
REFERENCE CPS Procedure No. 3105.01, P. 6 ANSWER 4.16 (1.50) FALSE RLT - false FALSE
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[3 @ 0.5 ea.] (1.5)
REFERENCE CPS Procedure No. 1024.25, P. 5 Procedure No. 1905.10, PP. 4 & 5 Procedure No. 1905.20, P. 5 (***** END OF CATEGORY 4 *****)
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TEST CROSS REFERENCE Paga 1
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QUESTION VALUE REFERENCE 1.01 0.75 ZZZ0000001 1.02 2.00 ZZZ0000002 1.03 2.00 ZZZ0000003 1.04 1.50 ZZZ0000004 1.05 1.00 ZZZ0000005 1.06 1.75 ZZZ0000006 1.07 1.50 ZZZ0000007 1.08 2.75 ZZZ0000008 1.09 2.50 ZZZ0000009 1.10 2.00 ZZZ0000010 1.11 2.00 ZZZ0000011 1.12 2.00 ZZZ0000012 1.13 1.00 ZZZ0000013 1.14 2.00 ZZZ0000014 1.15 0.75 ZZZ0000015
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25.50 2.01 2.75 ZZZ0000016 2.02 1.00 ZZZ0000017 2.03 1.50 ZZZ0000018
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2.04 3.00 ZZZ0000019 2.05 1.50 ZZZ0000020 2.06 1.50 ZZZ0000021 2.07 1.75 ZZZ0000022 2.08 2.00 ZZZ0000023 2.09 2.00 ZZZ0000024 2.10 2.50 ZZZ0000025 2.11 2.25 ZZZ0000026 2.12 1.25 ZZZ0000027 2.13 2.00 ZZZ0000028
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25.00 3.01 2.00 ZZZ0000029 ,
3.02 2.00 ZZZ0000030 1 3.03 2.25 ZZZ0000031 3.04 1 50 ZZZ0000032 3,05 1.00 ZZZ0000033 3.06 2.25 ZZZ0000034 3.07 2.50 ZZZ0000035 3.08 3.00 ZZZ0000036 3.09 1.50 ZZZ0000037 3.10 1.00 ZZZ0000038 3.11 1.00 ZZZ0000039 3.12 2.00 ZZZ0000040 3.13 1.50 ZZZ0000041 3.14 1.50 ZZZ0000042
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25.00 4.01 2.25 ZZZ0000043 4.02 2.25 ZZZ0000044 4.03 1.00 ZZZ0000045 4.04 1.00 ZZZ0000046
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TEST CROSS REFERENCE Pago 2
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QUESTION VALUE REFERENCE 4.05 1.50 ZZZ0000047 4.06 1.50 ZZZ0000048 4.07 3.00 ZZZ0000049 4.08 1.00 ZZZ0000050 4.09 0.50 ZZZ0000051 4.10 1.50 ZZZ0000052 4.11 1.00 ZZZ0000053 4.12 2.50 ZZZ0000054 4.13 1.00 ZZIO000055 4.14 2.00 ZZZ0000056 4.15 1.00 ZZZ0000057 4.16 1.50 ZZZ0000058
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U. S. NUCLEAR REGULATORY COMMISSION I
', SENIOR REACTOR OPERATOR LICENSE EXAMINATION l FACILITY: CLINTON REACTOR TYPE: BWR-GE6 1"l I=
l F)l W< DATE ADMINSTERED: 87/10/27 _ l EXAMINER: HANEK. CANDIDATE INSTRUCTI'"S TO CANDIDATE:
Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheet 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 six (6) hours after the examination start I
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL , SCORE _VALUE CATEGORY
? 6. 2 25.25 15.00 3 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,AND THERMODYNAMICS 2 4.60 & z+.4 +
f -2 5. 00 s _- 2 4. 8 E4 PLANT BYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION d 24.9&
25.00 -24.81 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL se LS 44 25.50 35.31 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS l
/co. "r0 % Totals Final Grade l I
All work done on this examination is my ow I have neither given i I
nor received ai Candidate's Signature MASER CPY I
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS Du' ring the administration of this examination the following rules apply: Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . Restroom trips are to be limited and only one candidate at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil only to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer . Print your name in the upper right-hand corner of the first page of each section of the answer shee Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example, 1.4, . Skip at least three lines between each answe . Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl . Use abbreviations only if they are commonly used in facility literatur . 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 require . Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl . 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: Assemble your examination as follows:
(1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are part o.1 the answe Turn in your copy of the examination and all pages used to answer the examination questions, Turn in all scrap paper and the balance of the paper that you did not use for answering the question Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoke .
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5'1 THEORY OF NUCLEAR POWER PLANT OPERATIO Paco 4
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, ELUIDS.AND THERMODYNAMICS
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QUESTION 5.01 (2.00)
The reactor is taken to CRITICALITY from a cold condition and an 80 second POSITIVE period is attained, From control room nuclear instrumentation, HOW can the operator tell when the heating range has been reached?
(Rod position and recirculation flow are held constant.) (0.5) In WHICH of the following intervals was the hesting range entered? (1.5)
(1) Interval 1 - reactor power increased by a factor of 6 in 143.3 second (2) Interval 2 - reactor power increased by a factor of 3 in 99.0 second ( T, ) Interval 3 - reactor power increased by a facter of 5 in 128.8 second \
QUESTION 5.02 (2.00)
During your Shife, an SRV inadvertently opens from 100% power and 1000 i psi Use a Mo;11er Diagram or the Steam Tables to answer EACH of the following: ( ASSUME A SATURA' FED SYSTEM AND INSTAN'fANEOUS HEAT TRAN"FER) STATE the tailpipe temperature, assuming atmospheric pressure in the Suppression Poc1 and No Reactor Depressurizatio If the Suppression Pool Pressure were to INCREASE, STATE whether the l Tailpipe Temperature would INCREASE, DECRE,\SE, or REllAIN 4.HE )
SAM If the reactor starts tc depressuiize when the SRV cpens, STATE whether the Tailpipe Temperature will initially INCREAS DECREASE, or REMA1N THE SAME in relation to what it would have done if the pressure had not decrease STATE the Reactor Pressure at which the Tailpipe Temperature
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would be at its MAXIMUM value (during the depressurization).
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. . THE0BY_QF NUCLEAR __ POWER PLANT OPERATIO Pega 5
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' ELHIDS. AHD_THEBMODYNAtilGE
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QUESTION S.03 (1.50)
Reactor power is increased by control red withdruus1. The veia fraction increases 1.5% and the fuel temperature increases 40 dearses as the result of the rod withdrawal. CALCULATE the reactictty WORTH of the PORTION of the control rod that was withdraw SHOW ALL WORK AND STATE ALL ASSUMPTION QUESTION 5.04 (3.00) For EACH of the following categories (a. through d. ) below, SELECT one item (1, 2 or 3) for each category that describes the basis of the Thermal-Hydraulic limit, MCP . For EACH of the following categrries (a. through d. ) below, SELECT one ltem (1, 2 or 3) for each category that describes the basis of the Thermal-Hydraulic limit, APtHG . For EACH of the following categories (a. through d. ) below, SELECT one item (1, 2 or 3) for each category that describes the basis of the Thermal-Hydraulic limit, LHG NOTE: EACH ITEM (1, 2 or 3) BELOW IS USED ONLY ONC CATEGORIES ITEMS
__________ _____ FAILURE MECHANISM: (1) Gross clad failure due to lack of cooling (2) Fuel clad cracking due to lack of cooling (3) Fuel clad cracking due to high stress CAUSE OF FAILURE: (1) Fuel pellet expansion l (2) Loss of nucleate boiling around cladding I (3) Decay heat and stored heat following LOCA ) LIMITING CONDITION: (1) Clad temperature of 2200 F (2) Boiling Trancition (3) 1% plastic strain on cladding ITEM MEASURED: (1) Total fuel bundle power (2) Local fuel pin power in node
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(3) Average fuel pin power in node (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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. THEQBY.0F_HEGLEAR POWER PLANT OPERATIO Pace 6
' ' FLUIDS.AND THERMODYNAMICS
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QUESTION 5.05 (1.00)
The Process Computer contains a program for calculating MAPRA Express (in terms of a ratio) what MAPRAT equals, During operation should MAPRAT be [ <1.0 ) [ =1.0 ) [ >1.0 ]?
(CHOOSE ONE OF THE CHOICES INSIDE THE BRACKETS [ ])
QUESTION 5.06 (1.50)
Briefly EXPL.*.IN the effect an increase in core age has en moderator temperature coefficient and WHY these changes occur. Include in the explanation, changes in the various core parameters affecting the temperature coefficien QUESTION 5.07 (2.50)
Concerning delayed neutrons, anewer EACH of the following: Approximately WHAT percentage of neutrons from '1-235 are born delayed? (0,5) The power generated by the reactor at the beginning of core life comes from U-235 thermal fission and U-238 fast fissio Later in core life, larger and larger fractions of power generation are produced by fission of what TWO '
isotopes? (1.0) HOW do delayed neutrons contribute to the control capability of a commercial reactor? (1.0)
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. . THEORY OF NUCLEAR POWER PLANT OPERATIO Peca 7 4 ' FLUIDS. AND THERMODYNAMICS
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QUESTION 5.08 (2,50) What VALUE of reactivity added to a core will cause a prompt critical condition? (0.5) List TWO methods of limiting Reactivity insertion rates by control rods (to avoid Prompt Criticality) during NORMAL REACTOR OPERATIONS and LIST the system hardware designs which are intended to perform these function (2.0)
QUESTION 5.09 (2.00)
Given a constant fuel temperature, EXPLAIN HOW and WHY the Doppler Coefficient will change with an increasing void fractio QUESTION 5.10 (2.00)
a .- EXPLAIN HOW and WHY core flow will change when power is reduced from 100% to 85% by control rod insertion, if the recirculation flow control valve position remains constan EXPLAIN HOW and WHY core flow will change when power is increased by control rod withdrawal during low power condition QUESTION 5.11 (2.00)
Reactor power is increased from 80% to 90% of full pc;:er by increasing I reciToulation flow. Did the average void fraction (steady state to steady ;
state) INCREASE, DECREASE, or REMAIN THE SAME7 Briefly EXPLAIN your answe i
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. THEQRY OF NUCLEAR POWER PLANT OPERATIO Paga 8
- ' FLUIDS. AND THERMODYNAMICS
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QUESTION 5.12 (0.75)
To permit reduction of reactor pressure to 100 PSIG during cocidown, the reactor vessel metal temperature must be reduced to WHAT temperature in order to prevent repressurization above 100 PSIG7 (Show all work for full credit)
QUESTION 5.13 (1.00)
A reactor heat balance was performed (by hand) during the 00-08 shift due to the Process Computer being 00 The GAF's were computed, but the APRM GAIN ADJUSTMENTS HAVE NOT BEEN MAD Which ONE of the following statements is TRUE concerning reactor power? If the feedwater flow rate used in the heat balance calculation was LOWER than the actual feedwater flow rate, then the actual power is HIGHER than the currently calculated powe If the reactor recirculation pump heat input used in the heat balance calculation was OMITTED, then the actual power is HIGHER than the currently calculated powe If the steam flow used in the heat balance calculation was LOWER than the actual steam flow, then the actual power is HIGHER than the currently calculated powe If the RWCU return temperature used in the heat balance calculation was LOWER than the actual RWCU return temperature, then the actual power is HIGHER than the currently calculated -
powe (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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5 THEQBY OF NUCLEAR POWER PLANT OPERATION, Pasa 9 ,
- FLUIDS,AND THERMODYNAMICS ,
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QUESTION 5.14 (1.50)
Consider the TWO Reactor Plant conditions listed belo . Low Power and Low Flow (<10%)
OR High Power and High Flow (>85%).
Answer EACH of the following: During which condition is the REQUIRED NPSH for a recirculation pump greater? (0.50) During which condition is AVAILABLE NPSH for a recirculation pump greater and WHY is it greater? (1.0)
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6.. PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION Pcga 10 t
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QUESTION 6.01 (2.75)
Concerning the Shutdown Service Water System (SX) answer EACH of the following: What are TWO potential sources of radioactive inleakage to the SX system? (0.5) What loads from EACH division do not discharge to the Ultimate Heat Sink, thus requiring caution during operation with low lake levela? (0.75) What THREE signals will automatically start the SX pumps?
(SETPOINTS REQUIRED) (0.9) How is a FLOWPATH for the A/B SX pumps ensured following an automatic pump start? (0.6)
QUESTION 6.02 (2.00)
Concerning the Main Control Room Halon Fire Protection System, answer EACH of the following TRUE or FALSE, The control room portable Halon fire extinguishers are rated for use on class A, B and C fires, Halon 1301 is harmless to personnel, in concentrations L:ed in systems at CPS, which are accessible to personnel, The control room Halon system is divided into four zones with two
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circuits (A and B) in each zon The second Halon bottle in a circuit will automatically dump 10 minutes after the first bottle is initiate QUESTION 6.03 (0.75)
WHAT is the function of the Control Rod Drive Housing Support Network?
(INCLUDE THE LIMIT IT ENFORCES)
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, . PLANT SYSTEMS DESIGN. CONTROL. AND INEIRUMENTATION Pacs 11
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QUESTION 6.04 (1.00)
Concerning the EHC system valve closure sequence for a turbine overspeed condition, FILL in blanks a. through d. in the following statemen Assume 100% steam flow and load set signal set at 100%
The control valves will throttle to limit overspeed during the first
__a.__ %, and the intercept valves will throttle closed f rom __b.__ %
to __c.__ % turbine speed. If turbine speed continues to increase to
__d.__ % the stop valves will shut, thus tripping the turbin QUESTION 6.05 (2.00)
During operation at 100% power with normal vessel level and 3 element control selected, one steam flow transmitter's output signal fails lo FILL in EACH of the following concerning the expected final steady state plant parameters. (Assume no corrective operator action.) Feedwater Flow % Total Steam Flow (indicated) % Steam Flow (actual) % Vessel Level controlling at (LOWER, HIGHER or NORMAL) leve QUESTION 6.06 (2.00) '
List EIGHT signals that will initiate a Group 1 MSIV automatic isolatio (Setpoints not Required)
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. Pk&NT SYSTEMS DEElqH. CONTROL. AND INSTRUMESTATION Pass 12
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QUESTION 6.07 (1.50)
There are FOUR pressure switches which sense Main Turbine first stage pressure, and provide an input to the RP List TWO functions these switches provide in the RP (1.0) WHEN are these switches closed (Armed)? (0.5)
(l.50)
QUESTION 6.08 (T:12M2r)
Concerning the Charcoal Adsorbers in the Off-Gas systems, answer EACH of the following TRUE or FALS Nbbie saae5 KiryLon and X uen ar; held "r hut n^+ ad=^rhad in the D&feTo The low operating temperature is primarily a fire prevention method against self-ignition and does not enhance noble gas removal, Absorber vault inlet valve, F133, and outlet valve, F053, automatically close upon receipt of a High Temperature Alarm in the adsorber The adsorber train bypass valve, F045, will close on a Post Treatment High Radiation Condition if the TREAT-AUTO-BYPASS Control Switch is the BYPASS positio .
QUESTION 6.09 (2.00) What is the FUNCTION of the Alternate Rod Insartion/ Recirculation Pump Trip (ARI/RPT) System? List TWO signals that will automatically actuate the ARI/RPT syste (SETPOINTS REQUIRED)
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i QUESTION 6.10 (1.00)
What effect would EACH of the following conditions have (INCREASE or DECREASE) on indicated Reactor Vessel level indication?
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' Seat leakage on a level transmitter equalizing valv Increase in Drywell temperatur Reference les leakage greater than the capacity of the steam )
condensing po j Recirculation loop operation on wide range instrumentatio l l
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How is leakage from a broken instrument sensing line minimized in EACH of the following situations? Sensing line break inside containment, outside the drywel ;
I Sensing line break outside containmen QUESTION 6.12 (1.75)
List the signals and time delays necessary for the Division I ECCS Logic to AUTOMATICALLY shift from the LPCI injection mode to Containment Spra (INCLUDE APPLICABLE SETPOINTS)
QUESTION 6.13 (1.25) l
1 What is the FUNCTION of the 105 second AND 6 minute time delays in the ADS initiation logic?
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QUESTION 6.14 (2.00)
Concerning the neutron monitoring system (NHS) answer EACH of the following TRUE or FALS Removing the "shorting links" will place all NHS scram signals in a coincidence mode, All IRM trips are bypassed when the mode switch is in run, The APRM flow biased scram is "clamped" at 118% regardless of recirculation flo The APRM flow biased scram is conditioned through a six second time dela QUESTION 6.15 (2.00)
List FOUR RADIATION signals which will automatically initiate the Standby Gas Treatment System. (INCLUDE APPLICABLE SETPOINTS OF EACH)
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7. . PROCEDURES - NORMAL. ABNORM &L. EMERGE 8QY Poco 15
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I QUESTION 7.01 (4.00)
Concerning Off-Normal Procedure 4100.01, Reactor Scram, Symptoms; LIST EIGHT of the nine automatic Reactor Scram signals that are NOT initiated by the Neutron Monitoring System (INCLUDE APPLICABLE SETPOINTS VALUES)
QUESTION 7.02 (1.00)
Concerning Off-Normal Procedure 4100.01, Reactor Scram, WHY does this procedure contain a step to close the Charging Header Isolation Valve, 1011-F034, when attempting a manual Control Rod insertion WITHOUT RPS Logic reset?
QUESTION 7.03 (2.75)
Concerning Off-Normal Procedure 4401.01, Level Control - Emergency, LIST the FIVE entry conditions stated in the symptons sectio (INCLUDE APPLICABLE SETPOINTS)
QUESTION 7.04 (2.75) List THREE immediate operator actions taken if it is necessary to evacuate the Main Control Room and initiate Off-Normal Procedure 4003.01, Remote Shutdown Pane (1.75) WHY is it preferred to leave the mode switch in run prior to evacuating the Main Control Room? (0.5)
e NHY is the operator directed to operate only those Remote Transfer Switches for systems that will need to be operated when reporting to the Remote Shutdown Panel? (0.5)
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. . PBOCEDURES - NORMAL. ABNORMAL EMERGENCY Paga 16
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QUESTION 7.05 (1.50)
Concerning Off-Normal Procedure 4004.01 Instrument Air Loss, .
LIST THREE conditions which when reached, will requi.*e a Manual Reactor Scram during a rapid depressurization of the snatrument Air Syste (INCLUDE APPLICABLE SETPOINTS)
QUESTION 7.06 (2.25)
Concerning Off-Normal Procedure 4007.02, Inadvertent Rod Movement; fill in EACH of the blanks below with the correct word or phras When the reactor is below the Low Power-Set point (LPSP) of (a.) %,
a (b.)___ will occur if a rod is selected and/or moved out of sequenc When operating between the LPSP and the High Power Set Point (HPSP) of (c.) %, rod withdrawal movement is limited to (d.)
notche When operating above the HPSP, rod withdrawal movement is limited to (e.) notche If a rod is inadvertently inserted more than (f.) notches, the Nuclear Engineer should be consulted prior to restoring the rod to its proper position. The concern is that there are (g.) present in the area prior to withdrawal, to insure that (h.) requirements are maintaine QUESTION 7.07 (1.00)
l Concerning Operating Procedure 3106.01, Meisture Separator '
Reheater, (MSR) anscer EACH of the following TRUE or FALS i
. Opcrativu vf the MSR uith tho .eh;; ting :t;;r. eu ryly .. cur:d i=;;;co restricti:n: en+=hha v retien, de /e7e f MSR's should be placed in and out of service as a pair to prevent uneven heating of the LP Turbine casings.
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. . . PROCEDURES - NORMAL. ABNORMAL EMERGEEQX Pega 17 AND RADIOLOGICAL CONTROL
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QUESTION 7.08 (1.00)
Concerning Operating Procedure 3110.01, Generator Stator Cooling Precautions; LIST TWO conditions which will cause a Turbine Generator runback to 25% generator load (7361 stator amps). (Setpoints not required)
QUESTION 7.09 (1.00)
Concerning Technical Procedure 2203.01, Return Of Out Of Sequence Rods, ;
Provide the definition of the following terms:
4 Deep Rods Shallow Rods )
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QUESTION 7.10 (1.00)
Concerning the Approach to Critical Checklist, CPS 3001.01C001 answer each of the following: What is the maximum time interval allowed between completion of the Approach to Critical Checklist, CPS 3001.01C001 and commencing the reactor start-up? What action is required if this time limit is exceeded?
QUESTION 7.11 (3.50)
LIST SIX conditions defined in Operating Procedure 3020.01, Primary and/or Secondary Containment Integrity Verification, which must be satisfied in order for Primary Containment Integrity to exis (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) . . .- -
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1 PROCEDUBES - HQBMAL. ABNORMAL. EMERGENCY Pega 18
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QUESTION 7.12 (1.25)
Concerning Off-Normal Procedure 4401.01, Level Control - Emergency, LIST the conditici.e which must exist before an Operator is allowed to override automatic ECCS initiation signal QUESTION 7.13 (1.00) What is the Technical Specification Limit for Linear Heat Generation Rate? If any fuel rod is found to be exceeding the LHGR limit, WHEN must corrective action be initiated?
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QUESTION 7.14 (1.00)
Concerning Radiation Work Permits (RWP) answer EACH of the following:
, WHO is responsible for initiating the RWP request? l WHO is responsible for processing the RWP request? 7 c WHO is responsible for approval / disapproval of the RWP7
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. ADMINISTRATIVE PROCEDURES. CONDITIONSm Poca 15 !
. AND LIMITATIONS
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QUESTION 8.01 (2.50)
List the Reactor Coolant System Leakage Limits as defined in Technical Specification QUESTION 8.02 (1.50)
LIST the requirements as defined in Technical Specifications, Section 6.8.3, which are required, when making a temporary change to an Operating Procedure on a backshif (Consider only the requirements that must be met on shift)
QUESTION 8.03 (2.00)
LIST FOUR of the six conditions addressed in Technical Specifications which require a Control Rod to be considered inoperable in operational condition QUESTION 8.04 (2.50)
During Refueling Conditions indicate (Yes or NO) whether EACH of the following is considered a "CORE ALTERATION" as defined in Technical Specification , Removal of an LPRM assembly for replacemen ' Removal of an uncoupled control rod for replacemen c Withdrawal of and insertion of an SRM detector to check the drive motor operatio ,
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' Control rod withdrawal and insertion to test the position indicator l
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prob Removal of a control rod position indicator probe for repai l l
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CATEGORY 8 CONTINUED ON NEXT PAGE
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.8 . ADMINISTRATIVE PROCEDURES CONDITION Pcga 20
'AND LIMITATIONS
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QUESTION 8.05 (1.50)
Concerning 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, state WHICH THREE of the following conditions requires a ONE hour notification to the NRC: The initiation of any nuclear plant shutdown required by the plant's Technical Specification Any event requiring the transport of a radioactively contaminated person to an off-site medical facility for treatment, Any event or condition that results in manual or automatic actuation of any Engineered Safety feature (ESF), including the Reactor Protection Syste Any event or condition during operation that results in the Nuclear Power Plant being in a condition not covered by the plant's operating and emergency procedures, t Any airborne radioactive release that exceeds 2 times the applicable concentrations of the limits specified in Appendix B, Tablo II of 10 CFR 20 in unrestricted areas when averaged over a time period of one hou ; Any event or condition during operation that results in the Nuclear Power Plant being in a condition that is outside the design basis of the plan ,
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.. ADMINISTRATIVE PROCEDURES. CONDITION Pcga 21
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QUESTION 8.06 (2.00)
Concerning Administrative Procedure 1001.05, Authorities and Responsibilities of Reactor Operators for Safe Operation and Shutdown, answer EACH of the following TRUE or FALSE concerning SRO supervisory requirements during fuel handling, An SRO or SRO Limited to Fuel Handling shall directly supervise all core alterations and shall have no other concurrent dutie An SRO or SRO Limited to Fuel Handling is not required in the Fuel Building when moving irradiated fuel in non-refueling situation When fuel is being moved between the Fuel Building and Containment refuel floor an SRO or SRO Limited to Fuel Handling is required at both locations, An SRO or SRO Limited to Fuel Handling shall direct all activities which may adversely affect the Fire Protection System in the fuel handling areas during fuel handling activitie l QUESTION 8.07 (3.00)
l Concerning Administrative Procedure 1001.06, Clinton Power Station Fire l Brigade, Answer EACH of the following: l
! WHAT are the required Shift Supervisor DUTIES and ACTIONS upon 1 notification of a fire involving radioactive material? (2.5) If the fire is not extinguished in TEN minutes, WHAT additional action is required? (0.5) l l
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D.u__ ADMINISTRATIVE PROCEDURES. CONDITION Pcco 22
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AED LIMITATIONS
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a QUESTION 8.08 (1.50)
FILL in EACH of the blanks (a. through c. ) in the following statement as defined in Technical Specification Section 4. Surveillance requirements shall be performed within the Technical Specification required time limitations with a maximum extension not to exceed __(a.)__ % of the surveillance interval, but the combined time interval for any __(b.)__ surveillance intervals shall not exceed
__(c.)__ times the specified surveillance interva QUESTION 8.09 (3.25)
Concerning Administrative Procedure 1014.01, Safety Tagging, answer EACH of the following: WHO is the tagging authority? (1.0) WHAT is a Safeguards Tagout? (0,5)
e Other than an Emergency Release by the Shift Supervisor, WHO may release a tagout? (0.5) WHAT is the maximum number of holders allowed on a tagout which will still allow a temporary tag lift to be preformed? (0.25) WHAT additional notifications, reviews, and documentation are required if the Shift Supervisor preforms an Emergency Release?
(1.0)
QUESTION 8.10 (0.75)
Concerning Administrative Procedure 1014.03, Temporary Modification, answer EACH of the following concerning Temporary Modification Permit Approva I IS the Shif t Supervisor authorized to make the decision that ,
determines if a Safety Evaluation (10 CFR 50.59) is/is not required? WHO is authorized to perform the Safety Evaluation if required?
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. . . , ADMINISTRATIVE PROCEDURES. CONDITION Pcga 23
- AND. LIMITATIONS
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QUESTION 8.11 (2.00)
Concerning Administrative Procedure 1041.01, Post Trip Review, answer EACH of the following: List the TWO parts of the Post Trip Review Process completed by the operating shift and the individual and/or position which is responsible, (or authorized) for completion of EAC (1.75) Is The Shift Supervisor authorized to obtain verbal concurrence of the Plant Manager for ALL restarts? (0.25)
QUESTION 8.12 (2.00)
Concerning the Technical Specification requirements for minimum shift crew composition, answer EACH of the following TRUE or FALS The shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> upon shift change due to a crew member being late, A Radiation Protection Technician shall be on site at all times when fuel is in the reactor but the position is allowed to be vacant for up to two hours to accommodate unexpected absence, The Shift Technical Advisor may be assigned to the fire brigade to fill an unexpected absence without taking further action on that shift to fill the required positio In the absence of the shift supervisor during refueling operations',
an individual with a valid reactor operator license may be designated
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to assume the control room command functio QUESTION 8.13 (1.00)
What constitutes non-compliance with a Technical Specification?
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EQUATION SHEET
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t f = ma v = s/t '
2 Cycle efficiency = "g**, *
v = ss s = v,t + lsat E = aC -
a = (vg - v,)/t 2 4g ;
KE = mv A = AN y
g,y, +g A = A,e j PE = agh w = 6/t A " 18 2/Ch = 0.693/tg .
W = VAP' (t,.)(td
, ,
AE = 931Am .
( + )
Q=$ CAT *IX p ,
I = I,e
, , k=UAAT I = I ge-ux ,
Pwr = W g a , I=I lo g */ M
. to SUR(t). TVL = 1.3/u y=y HVL a 0.693/u o t/T ~
SUR = 26.06/T T = 1.44 DT SCR = S/(1 - K,gg)
/A * o)
SUR = 26 CR, = S/(1 - K,gg,)
-
T = '(t*/o ) + [(i-' o)/A,ggo ] I C
eff}1 " 2(1 "Ke f f)'2
~~
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T,= 1*/ (o - D M " I/(1 - Kaff) = CR /CR0 g T = (I - o)/ A,gg o g , (g ,g.g,) f(t ,K,gg)
8*5 eff'III aff " AEefflEeff SDM = (1 - K,gg)/g,g, p= [1*/TKygg .] + [I/(1 + A,ggT )] t* = 1 x 10' seconds
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P = I4V/(3 x 1010) A,gg = 0.1 seconds'I E = No
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Idgg=Id22 WATER PARAMETERS Id g =Idg2 1 gal. = 8.345 lba 2 R/hr = (0.5 CE)/d (,,,,,,)
1 sal. - 3.78 liters R/hr = 6 CE/d2 (g,,C) ,
1 ft3 = 7.48 ga MISCEt.L\NEOUS CONVERSIONS ,
Density = 62.4 lbm/f t 1 Curie = 3.7 x 10 dps 10
Density = L ge/cm 1 kg = 2.21 lba Heat of vstorization = 970 teu/lbm 1 hp = 2.54 x 103 RTU/hr Hest of fusien = 144 Beu/lbm 1 Hw = 3.41 x 10 Stu/hr
1 Atm = 14.7 psi = 29.9 in. I Stu = 778 ft-lbf i 1 ft. H 2O = 0.4333 lbf/in 1' inch = 2.54 cm !
T = 9/5"C + 32
- C = 5/9 (*r - 32) l
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E -. THEQRY OF NUCLEAR POWER PLANT OPERATIO Paca 24
- FLUIDS. AND THEBtiQDlH6til91
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ANSWER 5.01 (2.00) Operator can notice that period has become longer [0.25)
and that power change on IRMs. SRMs is leveling off (turning around due to power overshoot). [0.25] (0.5) (From P = Poe(t/T) --> T = t/in (P/Po), in Interval 2 the period has lengthened from 80 seconds. The other intervals have 80 second periods.) (1.5)
REFERENCE CPS: Introduction to Nuclear Reactor Operations, LO 4.1.1.2, P.4-17 & 18 292008K112 292008K113 ..(KA's)
ANSWER 5.02 (2.00) deg F (+/- 15 deg F) Increase Increase psia (+/- 50 psia)
[4 0 0.5 ea.] (2.0)
REFERENCE CPS: Nuclear Power Plant Thermal Science, LO 3.1. BSEP Lesson Plan - Heat Transfer Chapter 4. Lesson Objective No. 3 and 5 from bottom of page 4-1(no number assigned)
Steam Tables /Mollier Diagram 218000A101 ..(KA's) l l
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ANSWER 5.03 (1.50)
Worth due to voids = (1 X 10E-3 dK/K/%V) (1.5%V) (0.5) l
= 1.5 X 10E-3 dK/K ;
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Worth due to fuel temp. = (1.0X10-5 dK/K/F) (40 F) (0.5)
= 0.4X10E-3 dK/K ROD WORTH = VOIDS + FUEL TEMP. = 1.9 X 10E-3 dK/K (0.5)
REFERENCE CPS: Introduction to Nuclear Reactor Operations, LO 5.1. Reactor Theory Sec. 1 Pg. 16,14,and BSEP Lesson Plan 2A, Reactor Theory, Chapt. 14 pp. 172 & 18 Lesson Objective No. 58, 201003K506 ..(KA's)
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52 .._IHEORY
OF_ NUCLEAR POWER PLANT OPERATIO Pcso 26
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FLUIDS.AND THERMODYNAMICS
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ANSWER 5.04 (3.00) MCPR a. =2 b. = 2 c. = 2 d. = l [4 0 0.25 ea.] (1.0) APLHGR a. = 1 b. = 3 c. =1
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d. = 3 [4 0 0.25 ea.) (1.0) LHGR a. =3 b. = 1
' =3 d. = 2 (4 6 0.25 ea.] (1.0)
REFERENCE CPS: Nuclear Power Thermal Sciences Chapter 10, Objective GE Thermodynamics Heat Transfer and Fluid Flow, Chapter 9, ,
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293009K112 293009K111 293009K120 293009K119 ..(KA's) !
ANSWER 5.05 (1.00) HAPRAT = HAPLHGR/MAPLHGR Limit oe A8ZM4 or AN MSb (0,5)
I?/4MdA G 54L//6K 114l < (0.5)
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Pag REFEREtiCE CPS; Nuclear Power Thermal Sciences P 293009K113 10.16. Objective 2
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ANSWER 5.06 (1.50)
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As the core ages result of effects this ac,oftion leakis an increase icontrol rods are withdrawn for
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negative rnup and the nuclei as the core ages (age (0.5].
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ratio (0.25) (0.5] causes a positive trendA decrease t increaseininthe number o coefficien .25) and an moderato r-to-fuel o the total moderator temperature REFERENCE (1.5)
I l: CPS INROIntroduction P. 6-17 to Nuclear Reacto r Operations, LO 6.1.1.E
" GG, Reactor Theory, Ch. 4, P. 11 292004K102
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ANSWER 5.07 (2.50)
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, .65%
(also accept 0.64%) Pu-239 [0.5) and Pu-241 [0.5) (0.5)
s/ [0.5] (by a factor of more thaDelayed neutrons age neutron generation time n 1000),
(1.0) incr S '. o & time of the reactor. (0.5) (je eJfest 7/-einer asing the control i i
/b I ' (1.0)
4A 4 6ft 4!A ' Thermal gjdf 4dt r hydrogen absorptionutilization increases due to d (0.5) ecreased slowing down lengthsresonance escape probability dwhil decrease in Keff (0.S). ecrease due to lor ,
less negative) and a posi ti ve trendThe net effect in !
coef f i ci er,t . (0.5)" (becomes It in the moderator tem perature (***** CATEGORY 5 CONTINUED ON NEXT PAGE
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, THEQBY_OF NUCLEAR POWER PLANT OPERATIO P:go 28 l
' FLUIDS.AHD_THERMODYHAMICS
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REFERENCE CPS: Introduction to Nuclear Reactor Operations, LO 3.1.1.2/3.1. INRO PP. 3-11 & 3-31 Grand Gulf Rx Physics pg. 31-34 Perry - Perry Introduction to Nuclear Reactor Operations, Chapter 3, Pages 3-11 and 3-31, 292003K106 ..(KA's)
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ANSWER 5.08 (2,50) Reactivity equal to Bet (or 0.75%) (0,5) . Rod withdrawal rates are limited (0,5) hydraulically by the throttle (needle) valves in the CRD syste [4r6t ANm/' c aw - Rod worths are limited [0,5) by the Rod Pattern Controller and preselected rod sequencin [0.5] (2.0)
(Th A c t.1s glr edion a l conlul Ti~ ra s~L/W cfa m ,(r,;,
REFERENCE # E d# # '~ '
""b!###'d h [0 D CPS: Introduction to Nuclear Reactor Operations, LO 4.1.1.1/4.1. Student Handbook, Vol 10 LP 74034, LO INRO P. 4-31 & LP 74034. P. 14 201002G007 201001G007 292003K107 292008K102 ..(KA's)
ANSWER 5.09 (2.00) ,
Doppler Coefficient will become more negativ [0,5) As voids increase, slowing down length increases (0,5) snd the neutrons spend more time in the resonance energy ban [0.5) More neutrons will be resonantly capture [0.5] (2.0)
REFERENCE '
CPS: Introduction to Nuclear Reactor Operations, LO 6.1. INRO P. 6-40 SRO EXAM BANK, Q 5.80 292004K103 292004K107 ..(KA's)
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5.... THEQRY OE_HUCLEAR POWEB_ELAHT OPERATION, Pcga 29
- ELMIDS, AND THEBMQQYNAMICS
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ANSWER 5.10 (2.00) The flow will increase (0.5] due to less 2-phase flow resistanc [0.5] (1.0) The flow will increase (0.5] due to increased natural circulation driving hea (0.5] (1.0)
REFERENCE CPS: Nuclear Power Plant Thermal Sciences, LO 9.1.1.1, P. 9-14 SRO EXAM BANK, Q 1.37 Standard Thermodynamic Principles, Reactor Operating Map, 76810, 293008K129 ..(KA's)
ANSWER 5.11 (2.00)
Decrease (0.5). In order to compensate for negative reactivity from doppler as fual temperature increased (0,5) and from the moderator as subcooling decreased (0,5), void fraction must decrease to add positive reactivity to bring net reactivity to zer (0,5)
REFERENCE CPS: Nuclear Power Plant Thermal Sciences, LO 9.1. NPPTS P. 9-8 292008K120 292004K111 ..(KA's) t ANSWER 5.12 (0.75)
100 PSIG + 14.7 PSIA = 114.7 PSIA (0.25)
Saturation temp for 114.7 PSIA is 338 deg. F. pC1 deg. (0,5)
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l ' THEORY OF NUCLEAB_EQYiR_ELANT OPERATIO Paco 30 '
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REFERENCE CPS: Nuclear Power Plant Thermal Sciences, LO 3.1. NPPTS PP. 3-2 & 3-3 29300K123 ..(KA's)
ANSWER 5.13 (1.00) (1.0)
REFERENCE CPS: Nuclear Power Thermal Sc'ences, , 8 & 9. Objective 11.1. ,
293007K111 ..(KA's)
ANSWER 5.14 (1.50) [2] High flow, High powe (0.5) High flow, High power (0.50), due to the increased inlet subcooling from the increased feedwater flo [0.50) (1.0)
REFERENCE CPS: Nuclear Power Plant Thermal Sciences Chapter 17, SRO Exam Bank 5.46, Lesson Plan 74035 P.19, Lesson Objective 74035. K402 ..(KA's) '
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Q.u__ PLANT SYSTEMS _ DESIGN. COMTROL, AND INSTRUMEF.IATION Pogo 31
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ANSWER 6.01 (2.75) RHR heat exchangers [0.0b] and fuel poo cooling and cleanup heat exchanger [0.25j /N4544Lcaep4?A#3 (0.5)
Any .t o.4 a Jr es. , DIV I: Drywell chillers [0.15] and breathing air compresso [0,15)
DIV II: Drywell chillers [0.15] and breething air compresso [0.15]
DIV III Pass sample pane ;0.15]
. [5 @ 0.15 ea.] (0.75) High drywell pressure (0.15] 1.68 psi [0.15]
RPV level 2 [0.15] -45.5". [0.15]
pg (M1')
Service water (WS) low pressure [0.15] 49 psi [0.15]
[6 0 0.15 ea.] (0.9) The RHR heat exchanger bypass valve (1SX-173 A/B) [0,15] opens
[0.15] if either of the RHR heat exchanger inlet or outlet valves (E12-F014 A/B or E12-F068 A/B) [0.15] respectively is off it's open sea [0.15] [4 @ 0.15 ea.] (0.6)
REFEEENCE CPS: L. P. 72013, PP. 5, 6, 7, 9, 10, 11, 12 & 13 Enabling Objectives 1.4, 2.1 & K104 ..(KA's)
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ANSWER 6.02 (2.00) True N C*^"Y'"w-True fJfn7 9acusW, ^ False False [4 @ 0.5 ea.] (1.5)
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REFERENCE CPS: L. P. 72006, PP. 30, 50 & 51, Enabling Objective, 1.2.a. & K402 286000A304 286000K404 ..(KA's)
ANSWER 6.03 (0.75)
Prevents the ra d ejection of a control rod [0.5] greater than 3 inchesy [0-961 or AJ r no ret ), go . ;g} (g,75)
REFERENCE l
CPS: L. P. 74023 P.8, Enabling Objective G006 201003G007 ..(KA's) j ANSWER 6.04 (1.00)
. % (90 RPM)
i % % % (1980) [4 @ 0.25 ea.] (1.0)
REFERENCE CPS: L. P. 73010, PP. 5 & 9. Enablina Objective 73010.1. .
241000K418 241000K403 ..(KA's)
ANSWER 6.05 (2.00) % % % Lower [4 9 0.5 ea.] (2.0)
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l REFERENCE CPS: L. P. 73013, P.7, Enabling Objective 73013.1. K603 ..(KA's)
l ANSWER 6.06 (2.00) ; Reactor Water Level (1) Low l Main Steam Line High Radiation
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' Main Steam Line Flow High Main Steam Line Turbine Building Temperature High !
J Main Steam Line Tunnel Temperature High Main Steam Line Tunnel Differential Temperature High I Condenser Vacuum Low [8 9 0.25 ea.] (2.0)
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REFERENCE
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l CPS: L. P7'73015, PP. 12 & 13. Enabling Objective, 73015. j
239001K606 ..(KA's)
s ANSWER 6.07 (1.50) ' Arm (or bypass) [0.25] the TCV fast closure [0.25] and stop valve ,
closure [0.25] scram [0.25] (1.0) I First stage pressure corresponding to-[CAF, ioference list 30%, 3 4 % -
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9end- 4 0% - - - - + -^ i -4 ' --- ' ----- - (0.5) l l
REFERENCE !
l CPS: L. P. 74021, PP. 10, 11 & 19 Enabling Objective 3. : s s, /
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/. .fo ANSWER 6.08 ( 0-0&)
.. Tm oeleted False True fa lf '-
.7 /. s Falce [4'O 0.5 ea.] (M)
REFiRENCE CPS: L. F. 73022 P. 10, Enabling Objective 73022. & A303 271000A204 2?1000K508 271000K407 ..(KA's)
ANSWER 6.09 (2.00) To prevent [0.25] and to mitigate [0.25] the consequences of an Anticipated Trausient Without Scram (ATWS). [0.5] (1.0) High RPV Pressure [0.25] 1125 psi [0.25]
RPV low Water Level [0.25] (2) -45.5". [0.25] [4 @ 0.25 ea.] (1.0)
REFERENCE CPS: L. P. 74005, P. 14, Enabling Objective 1. '
202001K414 ..(KA's)
ANSWER 6.10 (1.00) Increase
, Increase Increase Increase dec/ rah [49 0.25 ea.] (1.0)
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REFERENCE CPS: L. P. 74104 PP. 8 & 19. Enabling Objective K509 216000K507 ..(KA's)
ANSWER 6.11 (1.00) Each instrument line has a restricting orifice installed in it just inside the drywell to limit blowdow (0.5) Excess flow check valves are installed in each line penetrating containmen (0.5)
REFERENCE CPS: L. P. 74104 P.18 Enabling Objective 2.3, L. P. 74106 P. 5, Eaabling Objective G007 ..(KA's)
ANSWER 6.12 (1.75)
LPCI mode initiated for 10 (.17) minutes [0.25] AND [0.25] high drywell pressure [0.25] 1.68 psig [0.25] AND [0.25] high containment pressure
[0.25] 23.0 ps4 [0.25] [7 @ 0.25 ea.] (1.75)
PJIA .
REFERENCE CPS: L. P. 74018, P.10, Enabling Objective 1. '
226001K409 ..(KA's)
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. Ei . PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pago 36
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ANSWER 6.13 (1.25) j l The 105 second timer allows time for HPCS [0.25] to reflood the '
Reactor Vesse [0.25]
l For transients and accidents which do not directly produce a high !
Drywell pressure signal [0.25] and are degraded by a loss of all )
high pressure injection systems [0.25] adequate automatic core i cooling [0.25] is assured by actuation of the 6 minute time (Exact wording is not required) (1.25) '
REFERENCE CPS: L. P. 74016 P.22, Enabling Objective K501 ..(KA's)
ANSWER 6.14 (2.00) Me Fah ** True False True REFERENCE CPS: L. P. 74040 P.15, Enabling Objective L. P. 74041 '
Enabling Objective L. P. 74046 PP. 10, Enabling Objective K411 215005K109 215004A103 215004K406 ..(KA's)
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a PLANT SYSTEMS DESIGN. CONTROL AND INSTRUMENTATION Page 37
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ANSWER 6.15 (2.00) Containment Building Refueling Pool exhaust duct [0.25] 100mR/h [0.25) Containment Building Main exhaust duct [0.25) 100mR/h [0.25; Continuous Containment Purge exhaust duct (0.25) 100mR/h [0.25) Fuel Building exhaust [0.25) 10mR/h [0.25] [8 @ 0.25 ea.] (2.0)
REFERENCE CPS: L. P. 74052 P.8, Enabling Objective l 261000K401 ..(KA's)
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.7. . PROCEDURES - NORMAL. ABNORMAL. ENERGENCY Paga 38
- AND RADIOLOGICAL CONTROL
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ANSWER 7.01 (4,00) Reactor Vessel High Pressure [0.25] >1065 psi [0.26] Reactor Vessel Low Level [0.25) <8.9 inche _
[0.25) Main Steam Line Isolation Valve Closure [0.25) > 8% close _
[0.25) Main Steam Line Radiation High [0.25) >3.0X normal or ino _
[0.25] Drywell Pressure High [0.25) >1.68 psi _
[0.25] Scram Discharge Volume Water Level High [0.25] >762 f _
[0.25] Turbine Stop Valve Closure [0.25] >5% close _
[0.25) Turbine Control Valve Closure [0.25] <530 psi _
[0.25) Reactor Vessel Water Level High [0.25) >52 inche _
[0.25)
[any 8 scrams 'l 0. 25 ea, with setpoints 0 0.25 ea.] (4.0)
REFERENCE CPS: ONP 4100.01 PP.2 & K407 ..(KA's)
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ANSWER 7.02 (1.00) l l
It is necessary to isolate this flow path to develop a drive differential pressur (or) Following the Scram the CRD flow control valve will be shut directing full flow to the HCU' (1.0)
l REFERENCE l CPS: ONP 4100.01, P.15 295037E308 201001A204 ..(KA's) 4 l
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. . PROCEDURES _- NORMAL. ABHORMAL EMERGENCY Pego 39
- AND RADIOLOGICAL CONTROL
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ANSWER 7.03 (2.75) j Reactor Water Level [0.25] below level (+8.9 inches) [0.25] '
(0.5) l Drywell Pressure [0.25] above 1,68 psi [0.25] (0,5) , RPV Pressure [0.25] above 1064.7 psi [0.25] (0.5) i A condition which requires MSIV isolatio [0.25] (0.25) ! A condition exists which requires a Reactor Scram [0.25] AND Reactor power [0.25] is >3% [0.25] or cannot be determine _
[0.25] (1.0)
REFERENCE CPS: ONP-4401.01, G008 ..(KA's)
ANSWER 7.04 (2.75) . Manually scram the reactor; [0.5] observe all control rods fully inserted. [0.25] (0.75) , Sound the Containment Evacuation Alar (0.5) Initiate an Aler (0.b) To ensure the MSIV's close (at 850 psig). (0.5) To preserve as many automatic functions (0.25] and interlocks as possibl [0.25] (0.5)
CPS: ONP 4003.01 PP.3 & 5 l
l REFERENCE i 295016A107 295016A301 ..(KA's) l l
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- . PBQQEDURES - HQ_RMALJNORMAL. EMERGEHQY Pago 40
- AND RADIOLOGICAL CQNTROL
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ANSWER 7.05 (1.50) If the air pressure drops to 60 psig [0.25] and cannot be restore [0.25] (0.5) The GDV level increases [0.25] to the Rod Block Setpoin [0.25]
(0.5) The Control Rods begin to drif (0.5)
REFERENCE 1 CPS: ONP 4004.01 K115 212000K306 ..(KA's) i l
l ANSWER 7.06 (2.25) Rod Block ,
" (SD) IJ C~b' "'"l * l No [0.25] Voids [0.25] Preconditioning [9 9 0.25 ea.] (2.25)
REFERENCE CPS: ONP 4007.02 K404 201005K403 ..(KA's)
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. PE20EDBES - NORMAL. ABNORMAL. EMERGENCY Pego 41
- AND RADIOLOGICAL CONTROL 1
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ANSWER 7.07 (1.00) ?:1 e d . / 4 7.e l ; f6 M r.5 R.M /6pec r6 l True [2 @ 0.5 ea.] (1,0)
REFERENCE CPS: OP 3106.01, K301 ..(KA's)
ANSWER 7.08 (1.00) Stator coolant outlet high temperatur . Low stator cooling water flow (low pressure). [2 @ 0.5 ea.] (1.0)
REFERENCE CPS: OP 3110.01 P.1 K605 ..(KA's)
ANSWER 7.09 (1.00) Control Rods positioned between positions 00 and 24. B47<4eh9% '
AAs W4G 00 TO M/'ack. /8 Control Rods positioned between positions 26 and 46. 44re u # N pp.S we/t Jo ;ro pg, [2 @ 0.5 ea.] (1.0)
REFERENCE
CPS: TP 2203.01, K111 ..(KA's)
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7. . PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pago 42
- AHD RADIOLOGICAL CONTROL
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ANSWER 7.10 (1.00) Hours (0.5) The Shift Supervisor or Assistant Shift Supervisor should review to ensure its validit (0.5)
REFERENCE CPS: OP 3001.01, K101 ..(KA's)
ANSWER 7.11 (3.50) All containment equipment hatches are closed [0.25] and seale i
[0.25] (0.5) Each containment airlock [0.25] is operabl [0.25] (0.5)
' The Suppression Pool [0.25] is operabl [0.25] (0.5) The Containment leakage rates [0.25] are within Technical Specification Limit [0.25] (0.5) The sealing mechanism associated with each Containment Penetration [0.25] is operabl [0.25] (0.5)
csp ulk of b '.9 All Containment Penetrations [0,1] requiredftobeclosed[0.13 during accident conditions [0.1) are eitherklosed [0.1] by an operable [0.1] containment automatic isolation system (0.1] or closed by at least one manual valve, [0.1] blind flange [0.1]
or deactivated automatic valve [0.1] in its closed position. [0.1] [10 0 0.1 ea.] (1.0)
REFERENCE CPS: OP 3020.01, P.4, SRO Exam Bank 7.9 G005 .(KA's)
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,? . PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pago 43
AND RADIOLOGICAL CONTROL
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ANSWER 7.12 (1.25) Confirm by at least two independent indications [0.25] that misoperation in automatic is confirmed [0.25] or adequate core cooling is assure [0.25] (0.75) Directed to do so [0.25] by the Level Control-Emergency Procedur [0.25] (0.5)
REFERENCE CPS: ONP 4401.01, P.49, SRO Exam Bank, 7.13 295031G009 ..(KA's)
ANSWER 7.13 (1.00) .4 kW/ft Within 15 Minutes [2 @ 0.5 ea.] (1.0)
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REFERENCE CPS: Technical Specifications, 3.2.4, P.2-10 293009K107 ..(KA's)
ANSWER 7.14 (1.00)
.{4MMf7 dC 48/'/d* h/ ,emuSNb J^k'dIu Al [0A3
- The job planne Radiation Protection grou [W d' #AAI /#f da l
/ 6/P/dd'x hy go,1{ Shift Supervisor [0.25] Assistant Shift Superviso [0.25]
[4 9 0.25 ea.] (1.0)
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REFERENCE CPS: SRO Exam Bank 7.50, AP 1905 PP. 6 & K103 ..(KA's)
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ADMINISTBATIVE PROCEDURES. CONDITION Pago 45 AND LIMITATION .
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ANSWER 8.01 (2.50) No [0.25] pressure boundary leakag [0.25] (0.5) Unidentified leakage [0.25] Sgp [0.25] (0.5) Identified leakage [0.25] (averaged over any 24-hour period)
25 gp [0.25] (0.5) .5 gpm [0.25] leakage per nominal inch of valve size (0.25]
up to a maximum of 5 gpm [0.25] from any valve specified in referenced Technical Specification tabl [0.25] (1.0)
REFERENCE CPS: T.S. 3.4.3.2, OP 4001.01, G008 ..(KA's)
l ANSWER 8.02 (1.50) The intent of the original procedure is not altere (0.5) The change is approved by two members of the unit management staff,
[0.5] at least one of whom holds a SRO license on the affected uni [0.5] (1.0)
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REFERENCE ,
CPS: Technical Specifications Section 6.8.3, P.6-1 A101 ..(KA's)
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.S . ADMINISTRATIVE PROCEDURES. CONDITION Pago 46 i
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ANSWER 8.03 (2.00) Immovable l Untrippable l 1 Failure to meet scram insertion time specifications i Scram accumulator inoperable Rod uncoupled Rod Position indicators inoperable [any 4 @ 0.5 ea.] (2.0)
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REFERENCE CPS: Technical Specifications, Section 3.1.3, PP. 1-3 through 1-1 G005 ..(KA's)
ANSWER 8.04 (2.50) Yes Yes No Yes [5 9 0.5 ea.) (2.5) No REFERENCE CPS: Technical Specification, Section 1.0, P.1- G005 ..(KA's)
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l ANSWER 8.05 (1.50) l One Hour 1 One Hour One Hour [3 0 0.5 ea.] (1.5)
4. A c c.e lo / e ,7n e t p & f p ' l d m esp e n c y p e , 3 p g , l REFERENCE CPS: SRO Exam Bank, 10 CFR 50.72, PP. 509 & 51 A116 ..(KA's)
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ANSWER 8.06 (2.00) I True False False False [4 6 0.5 ea.] (2.0)
REFERENCE CPS: A. P. 1001.05, Section 8.14, P.9, A. P. 1001.06, Section 3.3.8, A110 ..(KA's)
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. . ADMINISTRATIVE PROCEDURES. CONDITION Paga 48
AND LIMITATIONS
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s ANSWER 8.07 (3.00) . The Shift Supervisor acting as the interim station emergency director [0.5] should determine if an emergency action level has been exceede [0.5] (1.0) Inform the RP Shift Supervisor as soon as possibl (0,5) Monitor the Fire Brigade communications to determine if additional assistance is require (0.5) Inform Plant Management and Fire Protection Supervision of the nature and status of the fir (0.5) Declare an unusual even (0.5)
REFERENCE CPS: A. .06, Section 8.2, PP. 13 & 1 K116 ..(KA's) ,
l ANSWER 8.08 (1.50) % Three ,
l .25 [3 @ 0.5 ea.] (1.5) l REFERENCE CPS: A. P. 1011.00, Section 8.13, A106 ..(KA's)
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. ADMINISTRATIVE PROCEDUREL. CONDITION Pego 49
AER_LlHITATIONS
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J ANSWER 8.09 (3.25) The on duty Shift / Assistant Shift Supervisor [0.5] and/or the Duty Radwaste Superviso [0.5] (1.0) Tagouts which disclose the type or function of equipment which have a direct / indirect impact on the integrity of the CPS security syste (Exact wording not required) (0.5) The individual job supervisor to whom a tagout is issue (0.5) Two (0.25)
l . Reviewed by the Plant Manage (0.25) Documented in the CRO Lo (0.25) Notify the department involved if personnel are available.(0.25) ; Cause a note to be affixed to the badge of the individual who held the tagou (0.25)
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REFERENCE I CPS: A. P. 1014.01, Sections 2.2, 6.4 & PP. 4, 6 & K102 ..(KA's) I ANSWER 8.10 (0.75) .
a . --Ntr [#./ (0.25) STA [0.25) or Technical Department Member [0.25] (0.5) ;
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REFERENCE CPS: A. P. 1014.03 Section 8.2, P A102 ..(KA's)
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. , ADMINISTRATIVE PROCEDURES. CONDITION Peco 50
, * AND LIMITATIONS
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ANSWER 8.11 (2.00)
l . Data Collection [0.25] STA [0.25] or any Licensed Operato I
[0.25] (0.75) I l Post Trip Investigation [0.25] Shift Supervisor [0.25] or another SRO [0.25] and/or ST [0.25] (1.0) No (0.25)
REFERENCE l
CPS: A. P. 1041.01, Section 3.0, A106 ..(KA's)
ANSWER 8.12 (2.00) False True
- False ' True (4 9 0.5 ea.] (2.0)
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REFERENCE ,
l CPS: Technical Specifications, Section 6.2.2.c & Table 6.2.2-1, PP. 6-1 & 6- A103 ..(KA's)
ANSWER 8.13 (1.00)
l Non-compliance shall exist when an LCO [0.25] and its associated action i statement (0.25] are not met (0.25] within the specified time interval [0.25] (1.0) '
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AND.LiljITATIONS
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l REFERENCE j
CPS: Technical Specifications Section 3.02, P. 01* l 294001A102 ..(KA's)
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TEST CROSS REFERENCE Pago 1 A
QUESTION VALUE REFERENCE o
5.01 2.00 ZZZ0000001 5.02 2.00 ZZZ0000002 5.03 1.50 ZZZ0000003 5.04 3.00 ZZZ0000004 5.05 1.00 ZZZ0000005 5.06 1.50 ZZZ0000006 5.07 2.50 ZZZ0000007 5.08 2.50 ZZZ0000008 5.09 2.00 ZZZ0000009 5.10 2.00 ZZZ0000010 5.11 2.00 ZZZ0000011 5.12 0.75 ZZZ0000012 5.13 1.00 ZZZ0000013 5.14 1.50 ZZZ0000014
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25.25 6.01 2.75 ZZZ0000015 6.02 2.00 ZZZ0000016 6.03 0.75 ZZZ0000017 6.04 1.00 ZZZ0000018 6.05 2.00 ZZZ0000019 6.06 2.00 ZZZ0000020 6.07 1.50 ZZZ0000021 6.08 2.00 ZZZ0000022 6.09 2.00 ZZZ0000023 6.10 1.00 ZZZ0000024
6.11 1.00 ZZZ0000025 i 6.12 1.75 ZZZ0000026 1 6.13 1.25 ZZZ0000027 6.14 2.00 ZZZ0000028 6.15 2.00 ZZZ0000029
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25.00 7.01 4.00 22Z0000030
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7.02 1.00 ZZZ0000031 I 7.03 2.75 ZZZ0000032 7.04 2.75 ZZZ0000033 7.05 1.50 ZZZ0000034 l 7.06 2.25 ZZZ0000035 7.07 1.00 ZZZ0000036 7.08 1.00 ZZZ0000037 7.09 1.00 ZZZ0000038 7.10 1.00 ZZZ0000039 7.11 3.50 ZZZ0000040 7.12 1.25 ZZZ0000041 7.13 1.00 ZZZ0000042 7.14 1.00 ZZZ0000043
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25.00 8.01 2.50 ZZZ0000044 8.02 1.50 ZZZ0000045 8.03 2.00 ZZZ0000046
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TEST CROSS REFERENCE Pago 2 t
gs @ESTION VALUE BEFERENCE 8.04 2.50 ZZZ0000047 8.05 1.50 ZZZ0000048 8.06 2.00 ZZZ0000049 8.07 3.00 ZZZ0000050 8.08 1.50 ZZZ0000051 8.09 3.25 ZZZ0000052 8.10 0.75 ZZZ0000053 ,
8.11 2.00 ZZZ0000054 l 8.12 2.00 ZZZ0000055 8.13 1.00 ZZZ0000056 l
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25.50
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ENCLOSURE 3 CLINTON POWER' STATION FACILITY COMMENTS REACTOR AND SENIOR REACTOR OPERATOR EXAMS OCTOBER 27, 1986 l
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P CLINTON POWER-STATION FACILITY COMMENTS AND RESOLUTIONS FOR REACTOR OPERATOR EXAMINATION OCTOBER 27, 1987 Generic Comment: Words in parenthesis should not be required for' full credi NRC Resolution: Comment note Words in parenthesis on the answer key are included as alternate acceptable responses or are ,
intended to further clarify the answer. These words are !
not required to receive any credit for the questio .01 Facility Comment: Should be +/- 1 degree Fahrenhei Appears to be a typographical erro ,
NRC Resolution: Comment noted, answer key has'been changed to include
+/- 1 degree Fahrenhei '
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1.02a .
Facility Comment: Operators may also indicate that the Intermediate Range !
Monitors will be on Range 7 or above to indicate the ,
heating rang .
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NRC Resolution: This response is not included in the facility material and no supporting documentation was provided to support the ,
commen No credit will be given for this response in !
lieu of those required by the answer key. Experience !
indicates this response may indeed be true and points will !
notbededucted(foradditionalinformation)ifthis !
response is give j 1.03d l Facility Comment: By ins)ection of the Mollier Diagram, 400 aounds per square !
inch a) solute to 550 pounds per square inc1 absolute i should be considered due to the flatness of the curve at- i this temperature ban i NRC Resolution: Inspection of the Mollier Diagram indicates the peak of the saturation curve lies between 500 and 400 psia. The ;
curveisclearlydroppng(slopeisbecomingJositive)at !
550 psia and thus coul not be mistaken for tie curve pea l The range provided in the answer key is considered adequate i and will not be change l 1.06
' Facility Comment: Rain should be considered also as an acceptable answe l
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NRC Resolution: No supporting documentation was provided by the utility regarding tMs comment. Rain would affect both the air and lake, and any limit-on evaporative cooling would be difficult to determine. The answer key will not be change .07
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Facility Comment: Discussion of control rod effect on thermal absorp' tion and therefore thermal utilization instead of "leakage should be considere Also a response including multiplication constant parameters such as thermal utilizatiors exposure, multiplicationconstant(K)andeffectivemultiplication constant (Keff.)etceterashouldbeacceptableinregard to "core parameters".
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NRC Resolution: The following has been included in the answer key as an alternative acceptable response:
"Thermal utilization increases due to decreased hydrogen absoration(0.5)whileleakageandresonanceescape roba)ility decrease due to longer slowing down lengths p(0. 5). The net effect is a decrease in Keff and a positivetrend(becomeslessnegative)inthemoderator temperature coefficien (0.5)
1.08c Facility Comment: The term "control time" is not a standard nuclear term and
, should not be required for full credit, response, response time, or this concept should be~ allowe .
NRC Resolution: Comment noted, terms are interchangeable and "response time" will be added in parenthesis to .the answer ke .
1.09b Facility Comment: In addition, Rod Blocks from the Neutron Monitoring ,
System, stabilizing valves, directional control valve timers,anddrivewaterheaderreliefvalve(preventing
excessive drive )ressure and therefore excessive drive speeds), should 3e considered as an acceptable answer for methods of limiting reactivity insertio NRC Resolution: This guestion specifically addresses limiting HOW FAST reactivity is inserted HOWMUCHreactivityis(inserted.reactivityinsertionrates)not Neutron Monitoring System Rod Blocks limit vice rate of insertion. quantity of valves reactivity insertion
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Stabilizing themselves do not limit flow in the CRD system (as do the attached needle /thrcttle valves) and therefore do not limit insertion rates. -The RC&IS directional control reactivity timers and /or flow metering valves in the directional
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g.._.- ,,~, . e+ Ap --y 4 -- g--- g-p- - ,- * * ' ' - " " - - *-*-"'
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control assembly limit reactivity insertion rates per facility LP 74005 pa0e 11, and these responses have been added to the answer key as alternate acceptable responses. The facility did not provide any documentation to indicate the drive water header relief valve was provided to limit reactivity insertion rates and this response will not receive grading credi .12 Facility Comment: In addition, "transition Mlin instead of "inadequate cooling"g" should be accepted
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NRC Resolution: Comment noted and answer key has been changed to include this respons .13 Facility Comment: All of the following are equal to MAPRAT depending on whether the candidate refers to terminology given in Technical Specifications, lesson plans; he may giveP-1 anyprintouts,isted of these l below. procedures or MAPLHGR - Maximum Average Planar Linear Heat Generation Rate APLHGR - Average Planar Linear Heat Generation Rate LIMLHGR - Limiting Linear Heat Generation Rate MAPRAT - Maximum Average Power Ratio MAPLHGR or APLHGR or APLHGR MAPLHGR Limit MAPLHGR APLHGR Limit or APLHGR or APLHGR or MAPLHGR MAPLHGR Limit LIMLHGR EIMLHGR
NRC Resolution: The following responses will be accepted as alternative responses to this questio APLHGR or APLHGR APLHGR Limit MAPLHUR Limit These answers can be supported using Technical Specification references. No documentation or justificationwasprovidedfortheadditionalresponses and these ratios differ significantly from the Technical Specification ratio. Only the additional responses indicated above are acceptable for credi . .
1.14 Facility Comment: Because of the minimal change in subcooling between 80-90%
power, doppler is the predominate effect; therefore, the point value for doppler should be greater than the point value for subcoolin NRC Resolution: Comment noted, no change will be made to the answer ke I
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r SECTION 2 2.01a Facility Comment: RHR Seal Coolers should also be an acceptable answe NRC Resolution: Residual Heat Renoval Seal Coolers has been added to the r swer key. Reference Byron Jackson drawing 1C-3640,
,covided by the facilit .01c Facility Ccmment: A pump start pressure of 109(109.5) pounds per square inch gauge should also be considered. References CPS 5064.02 t and 5064.0 NRC Resolution: The answer key has been changed to 109(109.5)psi The
facility supplied reference, CPS 5064.02 is a controlled annunciator response procedur The setpoint stated'in the uncontrolled lesson plan, 72013 will not be accepte .02 Facility Coment: The additional response of the minimum flow valve closing
! or is closed shouid also be acceptabl NRC Resolution: Minimum flow valve position can be used to verify that the but HPCS this is system is diverting not a direct indication flow to thethe that Suppression system isPool,is or notinjectingintothevessel. This response will not be added to the answer ke .05 Facility Comment: The Low Pressure Core water leg pump is listed as an alternate iniection system in CPS 4401.01, Level Control E::,ergenc Thisshouldbeanacceptableanswe ,
NRC Resolution: Comment note Answer key and reference have been changed to reflect this answer as an additional correct respons .06a Facility Comment: In addition to the spring, gravity assists in the closure of the extraction steam check valves, this should be i
! considered in addition to response in the examination ke ) NRC Resolution: Comment note The answer key has been changed to reflect
gravity assist and )oint values have been redistributed to correspond to the clange.
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- Facility Comment
- In addition, the extraction steam check valves will close ;
if either heater string inlet or outlet valves are not .
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fully open (E02-CB99 Sheet 9 to'E02-ES99 Sheet 8). Also reverse flow or a no flow conditio l
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NRC Resolution: The.following responses have been added to the answer key;
"4.. Either heater string inlet or outlet valves not full ope . Reverse flow (or no steam flow)."
The drawing numbers have been added to the reference '
2.07a Facility Comment: Momentarily depressing initiation )ush button will open F095 leaving bypass steam through :095 available to idle the turbine, as such it will not be in the standby condition described, students may not respond with A.1. in ;
the answer ke E02-1RI99 Sheet 6 and NRC Resolution: Comment noted.- Answer A.1 has been deleted from the answer ,
, key. Answer A point value has been changed to 0.5 and the total question value has been changed to 1.25 point .08a Facility Comment: This question does not differentiate between inocard and outboard systems. Only the inboard system requires the the outboard Main system Steam Isolatio does not requireValves to beIsolat Main Steam closed, ion Valves to be closed. Additionally, setpoints are not asked fo NRC Resolution: There are only three setpoints associated with system actuation and the candidate is not required to differentiate between inboard and outboard system *
Setpoints are the permissives and therefore are required for full credi The answer key has not been change '
2.10a Facility Comment: Containment Spray should also be considered as a correct Respons NRC Resolution: Comment note The valves indicated from LP 74006, '
Table 2, do provide control of Containment Spray. The answer and reference keys have been changed to reflect Containment Spray as an acceptable respons .10b Facility Comment: The Diesel Generator Fuel Oil Transfer pump system also has indications on the Remote Shutdown Pane NRC Resolution: The answer key has been changed to list the Diesel Generator Fuel Oil Transfer system as an acceptable response and PP. 6 has been added to the reference ;
2.11b Facility Comment: There is a throttle valve in the charging water header;- ,
however, it is not used to affect the pressure dro '
NRC Resolution: Comment note The throttle valve has been deleted from the required response, i
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SECTION 3 3.0 Facility Comment: Only one Turbine Driven Reactor Feed Pu'mp trip'is necessary in coincidence with Level 4, not both Turbine Driven Reactor Feed Pump NRC Resolution- Comment note Plural has been deleted.
J 3.0 Facility Comment: All three Circulating Water pumps must be running-initially to arm logic such that a trip of Circulating Water pump in coincidence with a low vacuum will cause a turbine trip, however a temporary modification (87-125) is in place which removes all Circulating Water runbacks, students mi'
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not include Circulating Water for'this reason. Either response should be considered correc *
NRC Resolution: The facility could not. provide any documentation indicating the Temporary Modification was to be a permanent design chang The answer key will not be modifie .02
Facility Comment: Instead of "prevent" or "mitigate", the candidate may respond with a description of how the Alternate Rod '
Insertion / Recirculation Pump Trip system functions since the question deliberately emphasizes FUNCTION vice purpose; i this should be taken in consideration when evaluating student respons NRC Resolution: Comment note A verbatim response is not required. If a
- candidate provides an in-depth discussion of the end of life requirements for the system, full credit will be give j 3.03 :
Facility Comment: Answer key should be less than or equal to vice less tha NRC Resolution: Comment note Answer key has been changed to "less than or equal to" and the reference will be changed to include T.S. 3.4. .04
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Facility Comment: The question does not ask for the effects on the transmitter differential pressure. . Explaining the change in indicated level, by explaining the change in reference
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leg density for example, should be sufficient for the answer. At Clintea, the vertical length has been
. intentionally minimized inside the drywell so some j
{ students may respond with an answer of; minimal increase !
or no chang ,
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"NRC Resolution: The question asks for HOW and WHY indicated level change Although a word for word response is not expected, a
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complete response should indicate why the detectors
! respond in thc expected manner. The response should mentior. a change-in differential pressure across the .
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detector for full credit. A minimal increase is an increase. No change is an incorrect response. The answer key will not be changed, i- 3.06 Facility Co' mment: Students may respond with 54-5b inches vice 40 inches or 18-19 inches vice 18 inches because of material in E02-1FW99 Sheet 106 C NRC Resolution: The reference provided by the facility were inconclusive and could not be verified. The answer key will not be change .07 Facility Comment: Students may respond with valve numbers for Reactor Core Isolation Cooling (E51-F095 and F045) and High Pressurd ;
Core Spray (E22-F004). In addition "annunciators" should
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be considered an acceptable response as they were in 3.0 '
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NRC Resolution: Valve numbers are an acceptable alternative to the system :
noun name for the valves. Credit will not be given for both a trip signal and annunciator triggered from the same electrical signa '
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3.08b
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Facility Comment: Should consider adding I 5) Average Power Range Monitor _ status light on Panel P680 !
6) Safety Parameter Display System Reactivity Control {
entry bo ;
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NRC Resolution: Both of these responses'have been added to the an:wer key as acceptable response .0 i Facility Conment: Source Range Monitor High setpoint of 10E5 counts per !
second should not be in the answer - this does not effect j the ability to retract the Source Range Monitor !
NRC Resciution: Comment Noted. Answer key was changed to delete 10E5 counts per'second. Change made to suit exam bank validity :
t only since dispositica did not affect exam result i i
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3.10b Facility Comment: "AveragePowerRangeMonitorindicatedpoweristhesameas core thermal power should be considered an acceptable answer as wel NRC Resolution: This response is the same as the answer key. The words
"core thermal power" will be added to the answer key in parenthesis to alleviate facility confusio .12 ..
Facility Comment: Students may interpret "the signal" in the second part of-the question to be actuation parameters and may not address divisional operation of the inboard and outboard valve NRC Resolution: The question is specific requiring'the logic for each group of valves. Credit is not subtracted for additional ccrrect information such as actuation signals and setpoint These signals themselves are not sufficient to receive credit for the questio .14c l Facility Comment: Setpoint value should be 1103 pounds per square inch gauge ,
(CPS 3101.01 Revision 4 page 10 of 15). l
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NRC Resolution: Comment noted and answer key has been revised.
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SECTION 4-4.D2c Facility Comment: Ths High Power Set Point given in 4007.02 in the discussion section differs from the Surveillance Data Sheet 9030.010002 RCP HPSP FUNCTIONAL TEST. (50.56to50.98%).
For this reason either 70% or 50% should b accertabl NRC Resolution: ecific In referencing Because 4007.02, and theitsquestion HPSP of was 70 sp% cred't will be given,orocedur for 70%.
This was pointed out during the exam review as a p ocedure error. Credit will also be giver, if a candidate g.ves a response of 50%, the correct setaoint as stated in the Surveillance Test Procedur .06a Facility Comment: Students may respond by including all immediate actions:
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Leave the Mode Switch in RUN to ensure the hin Steam '
Isolation Valve (MSIV's) close at 850 psi mein steam-line pressur ,
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report to tie Remote Shutdown panel (RSP). One Evacuate qual the Main Control Room;if
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preferably the A area operator, will assume and maintain control at the RSP until properly relieved or control is returned to the main control room in an
orderly fashio Initiate an ALERT 3er Emergency Plan Implementing j Procedure EC-02, EiERGENCY CLASSIFICATIO NRC Resolution: Actions taken after leavino the Control Room do not address conditions stated in the question. No credit will be added or subtracted for information which is correct but does not answer the questio .12 Facility Comment- Procedure 1016.01 was revise No longer does it list the types of conditions which require an individual to submit
- a Condition Report rather it says 8.1.1 "Any person upon becoming aware of a condition believed to be adverse to quality at the Clinton Power Station shall report such conditions to their su Either answer should be i considered acceptable.parvisor."
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NRC Resolution: The facility did not provide enough information to allow verification of this comment. Additional information was requested, but did not arrive withi.' the available time
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constraints. The answer key addresses conditions which are adverse to quality. Though the procedure has been revised, the examiners could not determine that the ;
appropriate sections had been change The answer key i will not be change ,
4.15 Facility Comment: Setpoint was not asked for in the question and should not be required for full credi NRC Resolution: Exact pressure is not required for full credi Any statement indicating the 40% power bypass of the stop valve closure scram will be-deactivated will receive full credit. A note has been added to the answer key to reflect this grading criteri .16b
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Facility Comment: This answer should be "false", the conditions described are for a Restricted Radiation Area (CPS 1905.20 R3 ,
Pay 6 of 9).
NRC Resolution: The answer key has been revised to indicate FALSE for this statemen NOTE: In addition to the Facility comments noted above, the examiner discovered one additional discrepancy in the answer key during grading of the exams. The following response has been listed as an a:ceptable response to Question 3.08.b; 7) Rate of change transfer to manual of RR flow control.
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CLINTON .
SENIOR REACTOR OPERATOR EXAMINATION OCTOBER 27, 1987 FACILITY COMMENTS AND NRC RESOLUTION r
5.02 a -
Facility Comment: By insaection of the Hollier Diagram, 400 )ounds per square inch a> solute to 550 pounds per square inci absolute i should be considered due to the flatness of the curve at .
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this temperature ban .
NRC Resolution: Inspection of the Mollier Diagram indicates the peak of the i saturation curve lies between 500 and 400 psi. The curve is clearly dro) ping -(slope is becoming positive) at 550 psia and tius could not be mistaken for the curve peak. The range provided in the answer key is considered i adequate and will not be-change .05 a Facility Comment: All of the following are equal.to MAPRAT depending on whether the candidate refers to terminology given in Technical Specifications,'P-1 printouts procedures, or lessonplans;hemaygiveanyoftheseIistedbelo '
MAPLHGR - Maximum Average Planar Linear Heat Generation Rate
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APLHGR - Average Planar Linear Heat Generation Rate
l LIMLHGR - Limiting Linear Heat Generation Rate
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MAPRAT - Maximum Average Power Ratio MAPLHGR or APLHGP, or APLHGR MAPLHGR Limit MAPLHGR 'APLHGR Limit or APLHGR or APLHGR or MAPLHGR MAPLHGR Limil LIMLHGR LIMLHGR NRC Resolution: The following responses will be accepted as alternative i responses to this question.~
Or i APLHGR Limit MAPLHGR Limit l i
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These answers can be supported using Technical Specification references. No documentation or justificationwasprovidedfortheadditionalresponses and these ratios differ significantly from the Tec;:nical Specification ration. Only the additional responses
indicated above will be accepted for credit.
i 5.06 i Facility Comment: Discussion of control rod effect on thermal absorp' tion and i therefore thermal utilization instead of "leakage should be considere Also a response including parameters such as thermal utilization, exposure, multiplication constant (K) and effective multiplication constant (K eff.) et cetera should be acceptable in regard to "core parameters".
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NRC Resolution: The following will be included in the answer key as an alternative acceptable response:
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"Thermal utilization increases due to decreased hydrogen absor> tion (0.5) while leakage and resonance escape proba>ility decrease due to longer slowing down lengths (0,5). The net effect is a decrease in Keff and a '
positive trend (becomes less negative) in the moderator temperature coefficient. (0.5)
5.07 c !
Facility Comment: The term "control time" is not a standard nuclear term and :
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should not be required for full credit, response, response time, or this concept should be allowed.
P NRC Resolution: Comment noted, terms are interchangeable and "res?onse time" will be added in parenthesis to the answer (ey.
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5.08 b o
- Facility Comment: In addition, Rod Blocks from the Neutron Monitoring System, stabilizing valves, direct!onal control valve timers, and ,
drive water header relief valve (preventing excessive !
drive pressure and therefore excessive drive speeds), !
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should be considered as an-acceptable answer for methods l
of limiting reactivity insertio !
NRC Resolution: This question specifically addresses limiting HOW FAST reactivity is inserted (reactivity insertion rates) not :
HOW MUCH reactivity is inserte Neutron Monitoring System .
Rod Blocks limit quantity of reactivity insertion vice l
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rate of insertion. Stabilizin i limit. flow in the CR0 systemas(gdovalves the attached themselves needle do
/ not l throttle valves) and therefore do not limit reactivity insertion rates. The RC&IS directional control timers .
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and/or flow metering valves in the directional control assembly limit reactivity insertion rates per facility ,
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l LP 74005 page 11, and these responses will be added to the answer key as alteraate acceptable responses. The facility did not provide any documentation to indicate the drive water header relief valve was provided te limit reactivity insertion rates and this response will not receive grading credi .11 Facility Comment: Because of the minimal change in subcooling betweea 80-90%
power, failure to mention subcooling effects should not reduce points equal to that of not considering the effects of dopple NRC Resolution: Comment noted, no change will be made to the answer ke .12 Facility Comment: Should be +/- 1 degree Fahrenhei Appears to be a
.I typographical erro NRC Resolution: Comment noted, answer key has been changed to include
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+/- 1 degree Fahrenheit.
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O .I SECTION 6 6.01 a Facility Comment: Residual Heat Removal Seal Coolers should also be an acceptable answe NRC Resolution: Residual Heat Removal Seal Coolers have been added to the answer ke Reference Byron Jackson drawing 10-3640, provided by the facilit .01 c Facility Comment: A pump start pressure of 109(109.5) pounds per square inch gauge should also be considered. (ReferenceCPS 5064.02).
NRC Resolution: The answer key has been changed to 109 (109.5) psi The facility supplied reference, CPS 5064.02 is a controlled annunciator response procedure. The setpoint stated in the uncontrolled lesson plan, 72013 will not be accepte .02 b Facility Comment: This could be false also as CPS 3313.01 section 4.6 says Halon 1301 is slightly toxi NRC Resolution: The ques ion was specific in addressing areas which are accessible to personnel. False will be accepted only if the candidate includes a reference to the > N concentration limit specified in CPS 3313.01 section .03 Facility Comment: Half a notch is also an acceptable limit (same as 3 inches).
NRC Resolution: Answer key has been changed to accept 3 inches or half a notc .08 a Facility Comment: This is technically false:
The Van Der Waal forces that hold up Krypton and Xenon are described as physical absorption vice chemical absorption for Iodine Either true or false should be acceptabl (Off Gas System Description Page 10, Lesson Plan number 73022)
NRC Resolution: The facility supplied reference, Lesson Plan 73022 supports either true of false as a correct response depending on which paragraph is utilized. The question has been deleted from the examination. The section and total point values have been reduced by 0.50 point c >
6.08 c Facility Comment: Lesson Plan is incorrect, E02-10G99 sheet 312 shows no temperature inputs. The response should be "false".
NRC Resolution: Facility supplied reference supports a correct response of false, and the answer key has been change ,09 a Facility Comment: Instead of "prevent or mitigate" the candidate may res with a description of how the Alternate Rod Insertion / pond Recirculation Pump Trip system functions. Since the question deliberately emphasizes FUNCTION vice purpose, this should be taken in consideration when evaluating student respons NRC Resolution: Comment noted, a verbatim response is not require If a candidate provides a discussion of the end of life requirements for the system, full credit will be give .10 d Facility Comment: Should be decrease, flow down the annulus past the variable leg tap will create a low pressure which will cause indicated level to be lowe NRC Resolution: Answer key has been changed to indicate a correct response of decreas .12 Facility Comment: Setpoint should be 23 pounds per square inch absolute not 23 pounds per square inch gaug (Technical Specifications 3.3.9)
NRC Resolution: Answer key has been changed to 23.0 psi .14 a Facility Comment: False, the removal of the shorting links places the Reactor Protection System in a non-coincidence mode with regard to the Neutron Monituring System signal NRC Resolution: Answer key has been changed to indicate a correct response of fals .14 d Facility Comment: The students have been taught the difference between a Time Constant and Time Dela The average Power Range Monitoring System flow biased scram is conditioned by a time constant. This question may confuse the studen NRC Resolution: Comment note Lesson Plan 74046 describes the signal conditioning as a "six second time constant circuit".
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SECTION 7 7.04 a Facility Comment: "Evacuate the Main Control Room; report to the Remote Shutdown Panel. One qualified operator, preferably the A area operator, will assume and maintain control at the Remote Shutdown Panel until properly relieved or control is returned to the main control room in an orderly fashion" may also be included as an immediate actio NRC Resolution: Although this is listed as an immediate operator action, the question gave initial conditions of control room evacuation and initiating the Remote Shutdown Panel Procedure, credit will not be give for this response as an immediate action in lieu of the responses required by the answer ke .06 c Facility Comment: The High Power Setpoint given in 4007.02 in the discussion section differs from the Surveillance Data Sheet 9031.010002 RPC HPSP FUNCTIONAL TEST (50.56 to 50.98%) for this reason either 70% or 50% should be accepte NRC Resolution: Because the question was specific in referencing pror.edure 4007.02, and its HPSP of 70% credit will be given for 70%
This was pointed out during the exam review as a procedure error. Credit will also be given if a candidate gives a response of 50% the correct setpoint as stated in the Surveillance Test Procedur .07 a Facility Comment: Section 6.6 of 3106.01 states:
Operation of the Moisture Separator Reheater with the reheating steam supply secured imposes no restriction on the turbines, however if operating at high power levels for extended periods of time, erosion of the low pressure turbine blading will increase and the low pressure turbine's efficiency will decreas In addition CPS 3105.01, step 6.2, states the turbine should never be run with steam isolated from one Moisture Separator Reheate An answer of true or false should be accepte NRC Resolution: Comment note Part a delete Points assigned to Part .09 Facility Comment: Definition similar to these should also be accepted:
Deep' Control Rod - a control rod interted more than 2/3 into the core. A normal range is from position 00 to approximately position 16 or 18. (The boundaries are not exactly defined). Deep control rods are refereed to as power rod Shallow Control Rods - a control rod inserted less than 1/3 into the core (i.e., position 30 or 32 to 48, fully withdrawn). Shallow control rods are often refereed to as shaping rod (Reference CPS 2202.01; 2.2.5 and 1.21.12)
NRC Resolution: Facility comment is valid; the additional reference referred to by the facility, CPS 2202.01 does contain the definitions included in the facility comment. Credit will be gi'/en for the alternate responses. Answer key change .11 6 Facility Co:nment: "Ccpable of being" is missing between "either" and
"closed" in the answer ke NRC Resolution: "Capable of being" will be added to the answer key between
"either" and "closed". Question point value and point breakdown will not be change .14 a Facility Comment: "Designated Responsible Individual" should also be acceptable. Either answer should be acceptabl (CPS 1095.10 Revision 4, Step 3.3)
NRC Resolution: "Designated Responsible Individual" will be added to the answer key and accepted in lieu of "Job Planner".
7.14 b Facility Comment: In addition "Radiological Operations Group" should be accepted us a correct respons ,
NRC Resolution: "Radiological Operations Group" will be added to the answer key and accepted in lieu of "Radiation Protection Group".
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7-r D SECTION 8 8.05 b Facility Comment: Per 10 CFR this is correct as a four hour notification, however some students may respond that this is a one hour notification as the transportation of a contaminated injured man is an Emergency Action Level entry condition, which is a one hour notificatio Either response should be considered correc NRC Resolution: Because the question was specific in referencing 10 CFR 50.72, credit will be given for a response of one hour only if the candidate includes a reference to the requirements as an Emergency Action Level Entry Conditio .10 a Facility Comment: The answer should also allow yes as a correct respons If the shift supervisor performs a review per Step 8. of CPS 1014.03 Revision 11 he may then make the decision to not perform a safety evaluation based on his revie NRC Resolution: The answer key has been changed to allov yes as the only correct answer. Facility supplied reference, CPS 1014.03 Revision 11 supersedes Revision 10 used in exam preparatio ,
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ES-104-1 ENCLOSURE 4-SIMULATION FACILITY FIDELITY REPORT
. Facility Licensee: Illinois Power Company Facility Licenseee Docket No.: 50-461 Facility Licensee No.:
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NPF-62 Operating Tests Administered At: Clinton Power Station Operating Tests Given On: 10/28/87 thru 11/5/87
During the conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies ,
were observe l When condensate booster pumps are aligned to the reactor vessel by bypassing the Reactor Feed Pumps (F024 through F004), reactor pressure has to c'ecrease to 300 psig before injection occurs. Since pump discharge pressure is approximately 700 psig, injection should begin when +
reactor pressure decreases to that poin . With a fuel failure at power (increasing in size), Main Steam Line radiation increased until it reached 32 mr/hr and started to decreas l As the leak increased in size, radiation levels should have continued to increas . During small LOCA, Equipment and Floor Drain Chart recorders in the simulator did not simulate leakage increases. Leakage should increas ' Simulated logic for High Pressure Core Spray Containment Outboard Isolation Valve, F004 (injection valve) will not allow the level 8 - !
closure if a high drywell pressure signal is present. The valve should ;
close on level 8 regardless of the initiation signa . Current system configuration for the Reactor Core Isolation Cooling (RCIC) system does not conform to the plant design modifications. The Steam Supply Bypass Valve, F095, is not simulated and system responds to +
an injection signal much faster than the plant wil , Feedwater header flows displayed by CRTs are actually feedwater A and B discharge flow. During periods where unbalanced flow conditions exist, these indications are misleading to the operato ,
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ES-104-1 7. The Alternate Rod Insertion (ARI) function of the ARI/RPT system is not simulate . There are no Automatic Depressurization Switches (ADS) inhibit switches in the simulator. This modification is used during portions of emergency procedures.
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