IR 05000317/1985023
| ML20133F570 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 09/20/1985 |
| From: | Dudley N, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20133F568 | List: |
| References | |
| 50-317-85-23, 50-318-85-23, NUDOCS 8510110112 | |
| Download: ML20133F570 (100) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATIN REPORT EXAMINATION REPORT NO. 85-23 FACILITY DOCKET NOS. 50-317/50-318 FACILITY LICENSE N05. DPR-53, DPR-69 LICENSEE: Baltimore Gas and Electric Company Post Office Box 1475 Baltimore, Maryland 21203 FACILITY: Calvert Cliffs Unit I and 2 EXAMINATION DATES: August 20 to 23, 1985 CHIEF EXAMINER: Ied w y f-/ / - F 6 Lead Reactor En eer (. Examiner) Date REVIEWED BY: bf du/ T// N I Chi f P oj cts Section IC Date'
APPROVED BY:
Chief, FroJ1H t Branch No. 1 Date JO [
SUMMARY: Licensing examinations were administered to one R0 and four SRO candidates. One RO and three SRO licenses will be issued. There was one SR0 written examination failure. Generic weaknesses were identified in the use of Technical Specifications and the Emergency Response Plan. Generic strengths were noted in the use of Main Control Room Reference Material.
8510110112 850925 PDR ADOCK 05000317 G PDR
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REPORT DETAILS EXAM RESULTS:
I R0 l SR0 l l Pass / Fail l Pass / Fail l l l l l l 1 I l Written Exam i 1/0 1 3/1 l l l l l l l l l l Oral Exam i 1/0 1 4/0 l l l 1 I l l l l l Overall l 1/0 1 3/1 l l l l l Chief Examiner at Site: N. Dudley Summary of generic strengths or deficiencies noted on oral exams:
- All candidates were familiar with Technical Specifications and action statements; however, some candidates had difficulty applying Technical Specifications to specific situation * Some candidates incorrectly classified a steam generator tube rupture accompanied by a failed open atmospheric dump valve as an alert due to a lack of understanding of the definition of fission product boundarie * Most candidates do not understand the difference between installed sources and source neutron * All candidates examined on the normal Shutdown Cooling lineup pro-vided accurate responses without hesitatio Summary of generic strengths or deficiencies noted from grading of written exams:
- Most candidates were weak in basic electrical theory, i
e All candidates were confused as to the affect changing plant con-ditions have on the TM/LP trip set point * Most candidates did not know the temperature on the Unit 2 turbine-generator main journal bearing at which the reactor should be trippe Baltimore Gas and Electric C . Comments on availability of, and candidate familiarization with, plant reference material in the control room:
All candidates made good use of available reference material throughout the oral examination . Comments on availability of, and candidate familiarization with, design, procedure, and T. S. changes, and with LERs and recent significant event All SR0 candidates were knowledgable of the requirements for stationing an SRO in the containment during head left, the normal shutdown cooling lineup, the control requirements for the Steam Generator Isolation System block key during dooldown, and the appropriate instrument for reading Tc with the reactor vessel head removed. These items were noted as defi-ciencies in Inspection Reports Nos. 50-317/85-09 and 50-318/85-09. In-spection Reports Nos. 50-317/85-09, 50-318/85-0 . Personnel Present at Exit Interview:
NRC Personnel N. Dudley, Lead Reactor Engineer (Examiner)
Facility Personnel J. Hill, Supervisor of Operations Training R. Heibel, General Supervisor - Operation Summary of NRC Comments made at exit interview:
The number and type of examinations conducted were summarized. Weaknesses noted during the oral examinations in the application of the Technical Specifications and the Emergency Plan were described. The familiarity of candidates with recent plant problems was noted. Arrangements for the up-coming NRC requalification program evaluation were discussed. The facility was informed that the waiver of up to one month of the three month on-shift training prior to the examination was applicable on a case by case basis, however, would not be granted for entire licensing classe . Summary of Facility questions and NRC responses at exit interview:
The facility asked how may hours a candidate must spend in training in the control room to meet the three month on-shift eligibility requiremen The NRC responded that there was no specific number of hours; however, i the facility was responsible for maintaining records to support three months of on-shift training for each candidate, a>
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Baltimore Gas and Electric C In January 1986, following the Unit 2 outage, new symptom based Emergency Operating Procedures (EOP) will be implemented. These procedures will replace the present event based E0P's. The facility requested that, as part of the NRC requalification program evaluation, the licensed op-erators be held responsible only for the new symptom based E0P's since ,
they were the E0P'r taught during requalification trainin '
The NRC stated that licensed operators would be examined only new symptom based E0P's. CHANGES MADE TO WRITTEN EXAM DURING EXAMINATION REVIEW:
Question N Change Reason 1.12b and Add "or fluctuating Question does not 5.06 power if voiding occurs specify where in react-at pump causing cavita- tor vessel voiding tion." occur .02 Change "DC" to "AC". Provides correct source of backup power to in-verter .04b Delete answer b. Saltwater valves are not considered part of com-ponent cooling water syste .08a Change to " Ambient air Provides answer for RCP which is cooled by Motor instead of RCP service water in the Motor bearin CTMT air coolers".
2.10a Change " leakoff Provides correct nomen-return" to "recirc." clature for facilit .07 Add "c. RAS cooling; Provides third means of verify proper RAS cooling core during a lineup and LPIS or LOC HPIS flow".
4.08a Add "or baron" after Boron may also be used
"CEA's". to control reactor power between 5% and 10%.
4.12c and Change to "No trip; Provides information in-7.06 trip if not in hot cluded in T.S. action standby in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />." statemen _ ___.- _ _ . _ __ _ _ _ _ _ _ . _ - - _ _ - . _
Baltimore Gas and Electric C ,
5.01 Add "or Q:m cpAT" and Provides alternate
"or using correct method of calculating value for Cp". AT powe .03 Add "or 845/2700= Uses rated thermal power 31.3%". to calculate efficienc .04 Change "0.85" to "0.41". Corrects math erro Change "68" to "33".
Change "132" to "167".
6.04c Change to "None". Training material in-correctly includes an
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automatic action on Spent-Fuel Storage-Pool area radiation monito .03 Add "9. High or low Includes additional in-steam flow, dications in accordance 10. SGIS initiation with E0P-4, 11. AFAS block signal".
7.04 Change to a. Underload Corrects answers in and cable slack; b. Dil- accordance with defuel-lon cell reading and ing procedur high load interlock."
7.04 reference Change to "0.I. - 25c". Provides procedure used for fuel handlin .09 Change " rapid" to "RCS". Improves wording of answe Attchments: Written Examination and Answer Key (RO) Written Examination and Answer Key (SRO) Facil-;e Co.v.ments on Written Examinations i
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/11 JmC// /11tn i 4 MASTE2 U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CALVERT CLIFFS
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REACTOR TYPE: PWR-CE
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DATE ADMINISTERED: 85/08/20
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EXAMINER: DUDLEY, _________________________
APPLICANT: _________________________
INSTRUCTIONS TO APPLICANT:
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U'ao separ a te paper for the answer Write answers on one side onl Stcple question sheet on top of the answer sheet Points for each qu2stion are indicated in parentheses after the question. The passing grcde requires at least 70% in each category and a final grade of at Icest 80%. Examination papers will be picked up six (6) hours after th? examination start .
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
________ ______ ___________ __-_____ __________________-_-___-___---_--_
_ 1 __ _ 1 _ ;_ {L_U___ ________ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 24.00 25.00 O, PLANT DESIGN INCLUDING SAFETY
________ ______ _ _ _ _ _I_n_ _t f_ _ _ ________ yg , 9,y gyg,y,,
_ I____ _ I_ _b__bMd4-__ ________ INSTRUMENTS AND CONTROLS
_ 1 __ _ ! _{t_ ![f{W1 ________ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 95.99 100.00 TOTALS
________ ______ ___________ _____--_
FINAL GRADE _________________%
All cork done on this examination is my own. I have neither Sivan nor received ai ~~~~~~~~~~~~~~
EPEL5CAsiT5~55GUhTURE
, PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
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QUESTION 1.01 ( .50)
A centrifugal pump operatin3 at 1800 rpm is pumping 400 spm at a discharge head of 20 psi which requires power of 45 Kw. If the pump speed is increased to 2700 rpm, which one of the following will be true?
A. The discharge head will increase to 45 ps B. The power requirements will increase to 151 K C. The flow rate will increase to 800 gp D. The pressure drop in the pumpin3 system will increase to 30 psi.
QUESTION 1.02 (3.00)
What affect will the following events have on the Departure from Nucleate Boiling Ratio (DNBR)? E::p l a i n . Dropping a rod at 100% powe Reactor Coolant average temperature is reduced by 5 F while maintaining 100% power.
QUESTION 1.03 (3.00)
Compare the CALCULATED Estimated Critical Position (ECP) for a etartup to be performed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a trip from 100% power, to the ACTUAL control rod position if the following events / conditions cecurred. Consider each independently. Limit your answer to HIGHER than, LOWER than. or SAME as the EC c. One reactor coolant pump is stopped two minutes prior to criticalit (0.6) The startup is delayed until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the tri (0.6) The steam dump pressure setpoint is increased to a value just below the Steam Generator saftey valve setpoin (0.6)
d. Condenser vacuum is reduced by 4 inches of Mercur (0.6)
o. All Steam Generator levels are being raised by 5% as the ECP is reache (0.6)
(xxxxx CATEGORY 01 CONTINUED ON HEXT PAGE **xxx) PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
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QUESTION 1.04 (3.00)
Compairing a slightly suberitical reactor [shudown margin = 1%3 to a greatly soberitical reactor [ shutdown margin = 5%], explain h2w an addition of 0.5% of positive reactivity will affect the followin a. THE CHANGE IN THE COUNT RATE THE TIME TO REACH A STAE:LE COUNT RATE QUESTION 1.05 (3.00)
The reactor has been operating at 50% for several day CEAs are in manual control. An EH system malfunction causes the turbine control valves to open slightly, causing an increase in steam flow, which increases electrical output by 8 M With no CPerator action explain HOW and WHY each of the following parameters will chang Steam generator pressure Primary Tave GUESTION 1.06 ( .50)
If during accident conditions after a trip, the pressuriner pressure is 1600 psi, Th is 580 F, and Te is 540 F, what CPPr o:<imate pr es sur e reading should be seen on the subcooled cater? psi E: . 275 psi C. 465 pst D. 635 psi (zuxxx CATEGORY 01 CONTINUED ON NEXT PAGE *xxxx)
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QUESTION 1.07 ( .50)
Haw long should it take to reach 1% power from an initial critical pswer level of 10 ^-4% if a constant 0.5 DPH startup rate is maintained?
Choose the most correct answe min B. 4 min C. 6 min D. 8 min GUESTION 1.08 (1.50)
Explain when AND why a plot of axial flux distribution for the Colvert Cliffs cores assumes a unique double-humped shap (1.5)
DUESTION 1.09 (1.50)
o. What is the primary reason for having a start up neutron source for a new core ? (0.75)
b. Explain why a start up source is rapidly depleted after attain-ins high power condition (0.75)
GUESTION 1.10 (2.00)
What is the most significant type of heat transfer (conduction, convection, or radiation) taking place under each of the following conditions? Consider each condition separatel Nucleate boilin Accident condition in which coolant is boiled and converted to steam in the reactor vesse Heat from fission thru the fuel pelle Decay heat removal by natural circulatio (xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)
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GUESTION 1.11 (2.50)
c. If a trip.from 100% power occurs with Xenon at equilibruim, what is the approximate time interval after trip that the operator should be concerned about Shutdown Margin decreasing below the SDH which existed when the trip occurred ? (0.75)
b. How would this approximate time interval compare if the trip occurred from 50% equilibruin conditions ? (0.75)
c. When a reactor is returned to 100% power from peak Xenon condit-ions,why does Xenon reactivity undershoot the equilibruim value? (1.0)
GUESTION 1.12 (3.00)
c. What will happen to the current drawn by a reactor coolant pump as the system 2s heated from 200 des F to 564 des F? Explain your answe What would happen to the current drawn by a reactor coolant pump if a void forms in the reactor vessel during accident conditions?
Explain your answe (xxxxx END OF CATEGORY 01 xxxxx)
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OUESTION 2.01 (1.50)
o. What is the purpose of the TWO high range samma monitors located in the containment? (0.5) Where are these high range samma monitors located in the containment? (1.0)
OUESTION 2.02 (1.50)
Explain how power supply reliability is achieved for the 120 VAC instrument power buse (Disregard computer power.)
GUESTION 2.03 (2.00)
In addition to cooling RCS water during shutdown cooling operations, what THREE other systems use the shutdown cooling heat exchangers to provide a cooling function? Include when these would be used.
QUESTION 2.04 (2.50)
Dascribe what automatically happens in each of the following systems upon receiving a SIAS signa c. Chemical and Volume Control system (1.5) Service water syste (1.0)
GUESTION 2.05 (2.00)
W h t' is running the diesel at less than normal operating speed not desired. Under what conditions may the diesel be run at less than normal operating speed for extended periods of time?
QUESTION 2.06 (1.50)
What will occur in the blowdown system if service water flow is lost to the blowdown heat exchanger during nors.al blowdown operations? Explain why?
(mmmmm CATEGORY 02 CONTIi4UED ON NEXT PAGE mmmmm)
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DUESTION 2.07 (3.00)
c. What will cause AND what action will result from a Recireviation Actuation Signal (RAS) ? (1.5)
b. It may be necessary to flush the core during long term cooling followin3 a loss of coolant. Describe one flow path for performing this flus (1.5)
GUESTION 2.08 (2.00)
Explazn how each of the following components of a Reactor Coolant Pump (RCP) are coole RCP motor Thermal barrier area Mechanical seals GUESTION 2.09 (2.50) Would containment design pressure and temperature limits be exceeded if the total Safety Injection System failed and all other Engineered Safety Features Systems fuctioned normally following a Loss of Coolant Incident (LOCI)? (0.5) What two other systems would reduce containment pressure and temperature during a LOCI? Include how the systems perform their function (2.0)
GUESTI0ld 2.10 (2.50)
c. For each of the indicated lines, A through E, on the attached figure, provide a description of where the line provides flo (1.5) Can the LPSI pump 11 be used to provide water to the Unit 2 Reactor Coolant System durin3 shutdown cooling operations at Unit 2? Explai (1.0)
(***x* CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8
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GUESTION 2.11 (3.00)
o. What are the ma::imum allowable heatop and cooldown rates for the pressurizer? (1.2)
b. Durang a heatupe how of ter is the pressurizer temperature required.to be verified? (0.6) If during a cooldown the cooldown limit is exceeded and
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pressuri er temperature is 50 F below the limit, how can temperature be brought within limits? (1.2)
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DUESTION 3.01 ( .33)
As containment temperature increases, what will be the relationship b3 tween actual and indicated pressurizer level? Select the best cnswe A. Indicated level reads higher than actual leve B. Indicated level increases reguardless of actual leve C.. Indicated level and actual level are the sam D. Indicated level reads lower than actual leve QUESTION 3.02 ( .33)
Which of the following inputs will cause both a SIAS and CIAS initiation?
A. Low pressuri:et pressures and high containsient pressur B. HiSh pressurizer pressure and high high containment pressur C. High pressuricer pressure and low steam generator leve D.1 Low pressurizer pressure arid var iable overpower tri QUESTION 3.03 ( .33)
The pressuri:er level progras. is developed frosi '
(Select the best answer)
A. The T ref signal B. The T ave signal C. The reactor power signal D. The pressurizer pressure signa QUESTION 3.04 (2.00)
If after e:: tended steady state operations at 100 % power the foed flow input signal to the feedwater control system CFIC-11113 foils low, how will the plant respond. Assusie no operator action, follow tr a ns i e rit to a stable plant coindition and EXPLAIN the reason for the plant respons (m**m* CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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QUESTION 3.05 (3.00)
c. Explain the logic that provides an Auxiliary Feed Actuating Sys-tem (AFAS) signal and how it determines which pump (s) will star b. Explain how the AFAS can indicate that a rupture has occurred either in a stean. line or an au>:iliary feed lin QUESTION 3.06 (2.25)
o. What type of detector (s) is/are used in the following monitors?
1. Main vent APDs 2. Main vent gaseous monitors 3. Containment APDs (0.75) Which of the following nionitor channels have automatic actions (other than indication and alarm) associated with them? Briefly describe the auton.atic action . Main vent APDs 2. Waste gas discharge monitor 3. Liquid waste discharge nio n i t o r Component cooling radiation monitor (1.5)
GUESTION 3.07 (3.00)
c. How does an increase in the Thermal Margin / Low Pressure (TM/LP)
trip setpoint affect the allowable DNBR ? (0.75)
b. List the three parameters that are measured and analyzed on the primary and secondary side of the Stean, Generators to provide Reactor Protection signals. Also state what specific protection is provide (2.25)
QUESTION 3.08 (3.25)
o. Describe the two methods of CEA position detectio (1.5)
b. What is the purpose of the CEA Group Deviation system (s) ? (0.75)
l c. Describe how the system (s) function (s) to inhibit CEA motio (1.0)
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GUESTION '3.09 (3.50)
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Ccncerning the axial power distribution reactor tripi o..What THREE' inputs are used to generate the APD signal and where is each input received from? (1.5)
b. What initiates an APD channel trip signal? (1.0)
c. 'In addition to generating the APD signal, what other functions does the AFD calculator provide? (1.0)
.0UESTION 3.10 (2.50)
Explain how and why the pressurizer pressure control system would rospond to a rapid insurge into the pressurizer which raised level 30 inches above the setpoint. Assume operating at 50% power and on initial pressure of 2250 psi.
! QUESTION 3.11 (3.50)
o. What type of detectors n.akeup a wide ranSe lo9arithmic nuclear instrument detector assembly? (0.75)
b. What type of detectors n.ateop the auxiliary wide range logarithmic instron.entation channel? (0.75)
c. Explain the function of the log count rate and campbell signal processing circuits for the wide range nuclear instrumentatio (2.0)
(***** END OF CATEGORY 03 maamm)
._ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12
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QUESTION 4.01 ( .50)
Lcrge overcooling transients can be characteri:ed by:
(Select the most correct answer.)
A. Decreasing primary system flo B. Increasing primary system boron concentratio C. Decreasing primary system average temperatur D. Increasing presurizer leve QUESTION 4.02 ( .50)
When performing shell warming of the main turbine, which of the following set of valves will be fully open: (Select the most correct cnswer.) Stop valves B. Control valves C. Intercept valves D. Intermediate stop valve QUESTION 4.03 (1.50)
c. What are the Calvert Cliffs administrative limits concerning weekly, quarterly, and yearly whole body radiation dose? (0.9)
6. Whose approval is necessary prior to exceeding the weekly, the quarterly, or the yearly whole body radiation dose? (0.6)
GUESTION 4.04 (1.50)
In accordance with E0P-3 (Loss of main feedwater) each of the following situations requires reactor power to be reduced. What
'is the maximuni allowed power level? (Assume 100% initial power)
o. One condensate pump available (0.5)
b. One condensate booster pump available (0.5)
c. One heater drain pump available (0.5)
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QUESTION 4.05 (1.50)
Explair. what will happers and what operator action is required for the unit 1 and unit 2 turbine upon experiencing a loss of pcwer to the il DC bu (1.5)
GUESTION 4.06 (1.50)
What affect would failure to equalize boron concentration between the Reactor Coolant System arid the pressurizer have on normal plant power operations if Reactor Coolant System boron concentration in decreased 60 pps?
QUESTION 4.07 (3.00)
The loss of reactor coolant emergency procedure discusses TWO ocdes of cooling the core depending on the size of the brea Briefly discuss how these two cooling siodes would be verified operable by the operato QUESTION 4.08 (2.50)
The plant shutdown procedure, OP-4, contains the following NOTE (Reactor between 5% to 10% power)* 'With the Turbirie Bypass Volve in automatic operation, the CEA's do not control primary systes, average t e ni p e r a t u r e . '
e. What is controlling Tave and reactor power at this power level? Explain how this control is performe (1.5) What controls Tave and reactor power at 50% power level?
Explai (1.0)
DUESTION 4.09 (2.50)
c. In accordance with AOP-8 (Excessive Reactor Coolant Leakage)
what is the difference between a MAJOR leak and a MINOR leak? (1.0)
b. What is your required response for a major leak? (1.5)
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QUESTION 4.10 (2.50)
The followin3 concern a loss of AC power (EOP 15):
a. The first automatic action states 'all full length CEA's should be fully inserted'. What are you required to do if this condition is not met? (1.0)
b. In addition to the full insertion of the CEA's, what other automatic actions should occur? (1.5)
OUESTION 4.11 (3.00)
c. While operating at 80% reactor power you have indications of a steam line rupture from the 412 S/G. What immediate actions must you perform? (2.0)
b. Assume that in addition to the steam leak a safety injection actuation signal is received. What additional immediate cctions must be performed? (1.0)
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QUESTION 4.12 (3.50)
For each of the situations below indicate whether the plant should bo tripped immediatel For situations which do not require an iCaediate trip explain at what point a reactor trip, if any, is roquired assuming conditions continue to deteriorate. Assume plant hos been operating for one week at 90% power and consider each cituatiore seperatel c. A rupture occurs in the Service Water subsyste The motor or the operating component cooling Pump f ail It is discovered that containsent integrity has been breached when a blind flange is found improperly secure Ar. unexplained dilutiori raises power by 5%.
o. Instrument air pressure drops to 75 psis, The main journal bearing wetal temperature is 230 F (5 F above the alarm set point) for the U r. i t 1 turbin The main journal bearing metal temperature is 225 F (5 F above the alarm set poirit) for the Ureit 2 turbin .
(mmman END OF CATEGORY 04 zumma)
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CNSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, ANSWER 1.01 ( .50)
1.12 a. Head increases to 45 ps ANSWER 1 02 (3.00)
c. The DNBR will decrease. [0.53 The dropped rod will reduce the flux in the local area [0.53 and will cause power to increase in other areas of the core. [0.53 b. DNBR will increase. [0.53 Reduced tes.perature will increase subcooling [0.53 causin3 steam bubbles to quench more rapidly. [0.53 REFERENCE CE Thermal Hydraulics, p 14 ANSWER 1.03 (3.00)
c. SAME, HIGHER c. HIGHER , SAME o. LOWER [0.6 each] (3 0)
REFERENCE C-E Reactor Theory Pgs. 166,204-206 ANSWER 1.04 (3 00) The slightly (greatly) soberitical reactor will have a larger (smaller) increase in count rat (1.5) The slightly (greatly) soberitical reactor will take a longer (shorter) time to reach a stable count rat (1 5)
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1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 17
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~~~~T sfR U6Y 55C5~~UE T TRIY5f5R I 6~fEU56~fE6U
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ANSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, REFERENCE C-E Reactor Theory Pgs. 147-148 ANSWER 1 05 (3.00)
c. SG pressure decreases [0.5]
SG pressure decreases as load is increased due to the removal of more energy. [1.0] Teve will decrease. [0.53 More energy is bein3 removed fros. SG which will lower Tc [0.53 Lower Tave will add positive reactivity to i ncrease reactor power to match turbine load. L).53 ANSWER 1.06 ( .50) psi REFERENCE Steam. Tables ANSWER 1.07 ( .50)
D. 8 min REFERENCE CE Nuclear Physics, Reactor Theory, and Core Operating Characteristics, p 148 ANSWER 1.08 (1.50)
This will occur at the end of core life at high power condition [0.53 The peak in the upper portion of the core is due to the shift in power density to the top of the core caused by fuel depletion in the bottom.CO.53 The peak at the lower elevation is caused by the higher moderator density at the core inlet.[0.5] (1.5)
REFERENCE CE Training Center Theory, P. 199 l
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ANSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, CNSWER 1.09 (1.50)
O. To provide an initial count rate for the nuclear instrument-atio (0.75)
c. A high neutron flux will result in the rapid burnout of the 94Pu-23 (0.75)
REFERENCE CE Training Center Theory, Pp. 138,139 ANSWER 1.10 (2.00) . Convection Radiation / convection (large Delta T) Conduction Convection (natural) (2.0)
REFERENCE General Physics pgs.99-115 ANSWER 1.11 (2.50)
o. ~ 24 Hrs. (accept 20 - 30 hrs.) (0.75)
b. Time would be shorte (< 20 hrs.) (0.75)
c. Due to the t i nie delay in Xenon production from the decay of Iodin (1.0)
REFERENCE WJE 164 CE Training Center Theory,Pp. 204,206
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c- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 19
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f.NSWERS -- CALVERT CLIFFS -85/08/20-0UDLEY, ANSWER 1.12 (3.00)
o. Power decreases CO.53 The pump is pumping the same volumetric flow rate but the density of the water decreases [0.53 which casuses a lower mass flow and less pump power. CO.53 Power decreases. [0.53 The pump as pumping against a lower discharge head due to the reduction of head losses in the area of the void. C1.03 CR r s a t t u <? r. ... c ., :e [c G ;i e::c.v.. cc s Am M r ~ *' , s a r " ""'
v:.. e en . D.c 1
. - . - _ = - __ --- - - -. - _ . - - _ . . _. . . . _ - - . - - _ _ - -
, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 20
_______________________________________________________
ANSWERS -- CALVERT CLIFFS -85/08/20-DUDLEYe ANSWER 2.01 (1.50)
o. Provides the ability to check radiation levels in containment under accident condition (0.5)
b. Located ore the 73 foot level 00.23, one near the steam generator 42 [0.43 and one near the pressuri:er CO.4 (1 0)
i REFERENCE Radiation monitoring system description pg. 13 CNSWER 2.02 (1.50)
There are 4 120 VAC busses per unit that are directly supplie by inverters. [0.253 Each of the inverters can be supplied by 125 VDC CO.53 or 120 VAC regulated power. [0.53 Two of the inverters on each unit are supplied by [gk power from the other unit.[0.253 REFERENCE 4'
'
SD 4 54, Fig's 54-1,54-2 ANSWER 2.03 (2.00)
1. Containment spray CO.43 during containment spray ops CO.26 . Spent fuel pool CO.43 when the complete core is removed and stored in the spent fuel pool CO.26 . Cooling water to HPSI suction [0.43 when HPSI is cavitating or during RAS [0.263 REFERENCE Safety Injection system dectription pgs. 10, 23 & 41
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CNSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, N.
ANSWER 2.04 (2.50) . Boric acid pumps start CO.253 2. Charging pumps start CO.253 3. Boric acid storage tank is lined up to inject boric acid
[0.253 VCT makeup stop valve CO.253 and outlet valve shut CO.253 5. Letdown line loop isolation valve shuts [0.253 (1.5) . Two service water pumps start 3 g, ,5; ;; n. 5 _ . . . u ma _m,u., ..i...... g . . ._ p . ,, g . . _
_
m_ ,,.._ . ,. . _ e_ n_ _ _9_ 9 9_ 2
,7 The turbine building SRW isolation valve shuts uv.337 (1.0)
REFERENCE ESFAS system description pg. 9 CNSWER 2.05 (2.00)
Operating at lower than norsa1 operating speeds can damage the exciter-regulator CO.53 and the generator field [0.53. E >:c i t a t i o n cost be removed from the generator field if operated at lower than normal speeds for u.aintenance C1.0 (2.0) 4 REFERENCE 01-21 pg. 1 ANSWER 2 06 (1.50)
Blowdown recovery flow is automatically diverted around the blowdown ion exchanger. C1.03 protects against overheating resin. [0.53 REFERENCE Blowdown system description pg. 4-6 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22
_________________.._____________________________________
CNSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, ANSWER 2.07 (3.00)
by) RWT level decreasing below approximately 30'. . (0.5)
Caused (Av' er . Action that results [0.33 ea.]
o Containment sump isolation valves open o Both LPSI pumps stop o Mini flow recire. Isolation valves receive a shut si (1.0) . Containment sunip > LPSI Pun.p > recirculation line > SDC return header > Hot les CO.3 each]
or 2. HPSI Pump > Au::. HPSI header > CVCS > P:r. A u >: . Spray >
P:r. > Surge line > Hot les CO.21 each] (1 5)
REFERENCE SD 4 7a8, Pp. 65,67,68 ANSWER 2.08 (2.00)
o. Upper end 1: err til etervei CO.?] cre cooled by :::=penent'-
ecoling system. [0. 3h A N'H f $1 M 6 I6 e l **Jo' h #"# *
(co N 07 J N " " " ## '"'
d ' # ' 'Y' '*# -
b. Cooled directly by component cooling. CO.663 c. 1 spm reactor coolant is passed throv3h the pressure breakdown capillarie [0.663 REFERENCE SD 5; RCS, p 16-19 ANSWER 2.09 (2.50)
o. No. [0.53 (temp and press would resain within limits)
b. CCS [0.43 condenses steam in CTMT removing heat and reducing and reduces pressure. [0.63 CTMT Air Recire System [0.53 removes heat by use of fans and cooling coils and thereby drop pressure.CO.53 REFERENCE SD 13 Containment System, p 38
>
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ANSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, CNSWER 2.10 (2.50)
o. A. SIT 1.RtctCC.k.fA reture,line B. HX outlet C. RCS loops or SIT D. RCS loops or SIT E. HX return header or RCS hot leg letdown [0.3 each3 b. No. [0.53 each unit has a seperate independent Safety Injection Syste [0.53 REFERENCE SD 7 and 8; SI and CS Systes.se p 2 Fig A-2
,
ANSWER 2.11 (3.00) heatup at 100 F/hr [0.63 cooldown at 200 F/hr [0.63 b. every 30 seinutes [0.63 c. Stop cooldow [0.63 Turn on heaters [0.23 Minimize spray [0.23 Do not allow level to increase [0.23 REFERENCE SD 51 Reactor Coolant Systeme p 25
,. .-. . . . . -. ._ _ . . - . _
.- INSTRUMENTS AND CONTROLS PAGE 24
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CNSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, ANSWER 3 01 ( .33)
"" rd, Indicated higher ANSWER 3.02 ( .33)
A. Low p2R press - high CTMT pres ANSWER 3.03 ( .33)
B. T ave signal CNSWER 3.04 (2.00)
Roactor will trip on TG trip which trips on S/G Hi level CO.83 Foodwater control valve will open CO.03 and level dominant function will be too slow to prevent trip. CO.43 REFERENCE SD 17, SG System, p 18 SD 32, MF System, p 17-19 E_________________ . _ _ _
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TNSTRUMENTS AND C TROLS PAGE 25
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ANSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, ANSWER 3.05 (3.00)
o. 1. Start signal is provided when 2/4 level indicating channels (0.75)
so below setpoin . The motor driver, pump and the steam-driven pump that is valved in will star (0.75)
b. Each of 4 S/G pressure channels is compared to it's correspon-dans channel on the other S/G. [0.253 When 2/4 channels reach the differential setpoint, a block signal is generated [0.253 and a steam line rupture alarm is actuated.[0.253 (0.75)
(The feed line break logic monitors 2/4 S/G pressure channels and 2 channels of AFW pressure to each S/G. Two channels
! of AFW flow to each S/G are also monitored. A pipe rupt-ure signal is generated when):
AFW feed flow > 100GPMi[0.43 AND S/G to f eed lirie dif f er eritial pressure > 175-200 psid [0.353 (0.75)
l REFERENCE
! SD 4 34. Pp.5,6 ANSWER 3.06 (2.25)
I o. 1. Scintillation detector 2. Ge19er-Nueller tube 3. Scintillation detector [0.25 each] (0.75)
b. 1. none CO.253 l 2. High alarm closes redundant waste gas discharge isolation j valves [0 53 l 3. High alarm c1cses two discharge isolation valves to stop
! discharge flow [0.53 none [0.253 (1.5)
l REFERENCE Rcdiation Monitoring system description pgs. 20,24,39,43,45&48
,
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l INSTRUMENTS AND CONTROLS PAGE 26
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CNSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, N.
ANSWER 3.07 (3.00)
o. A setpoint increase means that allowable DNBR has decrease (0.75) . Coolant flowiC0 43 Prevent fuel damage should a loss of flow ccur. CO.353 (0.75)
C. 3 5 2. S/G water leveli[0.43 Ensure heat removal capability.[4+43 (0.75)
3. S/C pressoreiCO.43 Protect against excessive heat removal rate caused by steam rupture.[0.353 (0.75)
REFERENCE WJE 166 General Physics HTaFF, Thermal Hydraulics sect. P.46 Sys. description 4 59, Pp.14,29,30,31 ANSWER 3.00 (3.25)
o. 1. The primary indication utili es up-down signals generated in the coil power programmers. The pulses are trensmitted to an up-down counter in the plant cos.pute (0.75)
2. Secondary indication and CEA mimic display use two different sets of reed switche (0.75)
b. To prevent power peaking in the core as a result of CEA devia-tion within a grou (0.75)
c. The system associated with the secondary indacating system auctioneers out the highest and lowest CEA'S within a grou CO.253These signals are sent to a comparator and when a 4' dev-iation is sensedeCO.53 a bistable is tripped to actuate a motion inhibit interlock.CO.253 (1 0)
REFERENCE WJE 169 Syotem Description 460, Pp.25,26,29
. _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - - - -
i l INSTR,dPENTS AND CONTROLS PAGE 27
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C.NSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, N.
ANSWER 3.09 (3.50)
o. 1. Reactor power [0.253-generated by the TM/LP calculator [0.253 2. ASI [0.253-from the nuclear instruments [0.253 3. CEA Function CO.253-fined input from the safety analysis
[0.253 (1.5)
b. A channel trip occurs if the axial shape index (YI) exceeds a positive or negative limiting value [0.53 of the power-dependent insertion limits (YP or YN) [0 53 (1.0)
c. Also generates the allowable positive and negative ASI limits
[0.53 and the a::ial of f set used in the TM/LP caleviator [0.5 (1.0)
REFERENCE Roactor protection system description pg. 21,23 and fig. A-8 ANSWER 3.10 (2.50)
Pcwer to proportional heaters so to zer [0.63 Spray valve ran.ps open until pressure is restored to 2300 psi. [0.63 Power to proportional heaters decrease as pressure decreases below 2275 ps1. [0.53 Insurge will con. press stean. space raising pressure until spray condenses the stean, to lower the pressure. [0.83 REFERENCE SD Reactor Coolant Systen. Instrus.entation, Figure 62-11 ANSWER 3.11 (3.50)
o. (2) proportional [0.43 and (1) fission chamber [0.353 (2) fission chambers [0.753 c. Loc count rate produces a st3nal proportional to count rate upto 2n to ^5 eps [0.03 l Cambelling circuit produces a signal proportional to power level above 6 n 10 ^2% power [0 83 and is combined with count rate signal to produce the power level signal. [0.43 REFERENCE SD 571 Nuclear Instrumentation System Description, p 4, 12, 19-21
_ _ _ - . _ _ _ - _ _ _ _ - _ _ - . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 28
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CNSWERS -- CALVERT CLIFFS -85/08/20-0UDLEY, N.
CNSWER 4.01 ( .50)
C. Decreasing primary system temperature.
ANSWER 4.02 ( .50) Control valves CNSWER 4.03 (1.50)
c. Weekly 300 mre Ovarterly 2 0 Rem Yearly 4.0 Re (0.9)
b. heediate t e r+ " i s e t -
E::.er:! reperviser e r:d Ge :ee +1 s e r e e " t e c e - e = d 4 = + 4 n e- ==f= tug Gerie t a l supervisor arid Gerier al super visor-radiat t ori saf ety (0.6)
REFERENCE CCI-800A pas. 9-10 ANSWER 4.04 (1.50)
a. 5 0 ?. (0.5)
b. 70% (0.5)
c. 80% (0.5)
REFERENCE E0P-3 pg. 2
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. PRO:EDURES - NORMAL, AE: NORMAL, ENERCENCY AND PAGE 29
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RADIOLOGICAL CONTROL
ANSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, '
ANSWER 4.05 (1.50)
! The unit 1 turbine will automatically trip CO.5]. The unit 2
, turbine will not automatically trip but all remote and automatic olectrical trips will be lost CO.5]. An operator must be stationed
Ot the turbiree front standard in direct communication with the control roon. to allow for manual tripping of the turbine CO.5] (1.5)
REFERENCE 1 GSO Standing Instructions
!
j ANSWER 4.06 (1.50)
- RCS temperature would decrease when an outsurge from the pressuri
- er occurre REFERENCE I
OP-4, p 1
ANSWER 4.07 (3.00)
a. Natural circulation CO.53 verified by the observation of a j teniperature rise across the core, by the stabili:ation and decrease of Th and ability to charige RCS temper ature by f eed/
steas, rates C1.53.
l b. Core boiling CO.53 evidenced by the observation of saturation gonditions t re the RCS and are empty pressuri er C1.0 Q7f, '{- v r [c S] VCKr F ? PWck RM travr A an w; r e t nrn ni,v D.c]
E0P-5 pg. 2 1 ANSWER 4.08 (2.50)
i c. Tave controlled g*y, bypass valves set e 9004 which is saturation for 532 F. CEA's control reactor power with reactivity
! addition or ren. ova (1.5)
i .
b. Reactor power controlled by steam deman Tave controlled by rod movement or boron concentratio (1.0)
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< . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30
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RA65ULU55 CAL CUNiRUL'~~~~~~~~~~~~~~~~~~~~~~~
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ANSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, REFERENCE OP-4 pg. 3 ANSWER 4.09 (2.50)
o. If pressurizer level can be maintained with one charging pump you have minor leakage CO.53, if'more than one charging pump is required, you have major leakage CO.5]. (1.0)
b. 1. Manually initiate or verify initiation of back-up charging pumps CO.75 .. Trip the reactor if all charging pumps are running and pressur i er level'is decreasing CO.75 (Refer to Tech. Specs. for leakage limits)CO.253 (1.50)
REFERENCE AOP-8 pgs. 184 f.NSWER 4.10 (2.50)
c. Increase RCS boron concentration by 200 ppm for each full length CEA that is not fully inserte (1.0)
b. 1. Turbine has tripped and the generator output, and exciter field breakers have tripped C1.23 2. Diesel generators have started CO.33 (1.5)
REFERENCE E0P-15 pg. 2 ANSWER 4 11 (3 00) . Trip reactor and turbine Insure AFAS start, AFAS block and AFW pump starts Isolate S/G Emergency borate CO.5 each] (2.0)
b. 1. Stop all RCP's 5 seconds after the reactor is tripped Implement E0P-f2 CO.5 each] (1.0)
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CNSWERS -- CALVERT CLIFFS -85/08/20-DUDLEY, REFERENCE E0P-4 pgs. 3-4 ANSWER 4.12 (3.50)
o. Trip reacto CO.53 b. Trip if not restored in 10 min. [0.3] or alarm received on RCP thrust temperature (
e. No trip. CO.23 be a rl#*
a i ngJf J'i t i'Er?' iTf,A19 5 F.) .fi'ifif4*fj0.23E0.33
.
ir d. No trip. CO.23 only if dilution raises power to RPS high power trip. CO.33 o. No trip. CO.23 trip when pressure reaches 50 psis. CO.3]
f. No trip. CO.2] trip at 250 F. CO.3]
g. Trip reactor. CO.53 REFERENCE AOP 3, p3 AOP 4, p 1 AOP 6, p 1, 2 CDP 7, p4 AOP 70, p2 AOP 7E Unit le p3
'
AOP 7E Unit 2, p3 l
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! U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION
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Facility: Calvert Cliffs Reactor Type: PWR-CE Date Administered: August 20, 1985 Examiner: J. D. Smith, J. W. Upton, J Candidate: Answer Key i
,
INSTRUCTIONS TO APPLICANT:
Print your name on the line above marked " Candidate." The grade points available for each question are indicated within parentheses af ter each ques-tion. The passing grade is at least 70% in each of the four (4) categories and is at least 80% for the total grade. Use separate paper for your answers and 1 write on only one (1) side of the paper, unless a specific question instructs you otherwis Staple this question package to your answer sheet The exami-nation questions and answers will be picked up six (6) hours after the examina-tion was starte Read the statement at the bottom of this page. When you have finished this examination, af firm the statement by signing your nam Category % of Applicant's % of Value Total Score Cat. Value Category 25 25 5. Theory of Nuclear Power Plant
,
Operation, Fluids and
. Thermodynamics 25 25 6. Plant System Design, Control and Instrumentation 25 25 7. Procedures--Normal, Abnonnal ,
Emergency and Radiological Control 25 25 8. Administrative Procedures, Conditions and Limitations 100 TOTALS Final Grade %
l All work done on this examination is my own; I have neither given nor received J aid, i
Candidate's Signature
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1 CALVERT CLIFFS August 20, 1985
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5.0 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS AND THERMODYNM11CS (25)
The following statements apply to Questions 5.01 through 5.0 The Calvert Cliffs Unit 1 power plant has been operating continuously at a steady 100% of full power for 14 days. All control rods (CEAs)
are fully withdrawn from the nuclear reactor core ( ARO). All con-trolled parameters are equal to their respective programmed value The fuel-hurnup status is that the core has reached 10,000 MWD /MTU in cycle VII. The present boron concentration is 200 ppm and the inverse boron worth is 80 ppm /% ap. With four (4) RCPs running, the total pri-mary coolant flowrate is 134 _M_ lbm/hr. The main generator is producing 845 MWe with a pf of 0.95 lagging. Use any of the provided figures and tables. Show your work and your procedures and indicate your assumption Points Available OHESTION 5.01 What is the "aT-power" as determined from the operating values in the primary loop? Give your answer in Rtu/hr and in M (1.5) If the flowrate of primary coolant through the core was reduced (without causing a trip), explain the effect that this would have on the aT across the cor (1.0)
-
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2 CALVERT CLIFFS
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August 20, 1985 l
.
Points Available ANSWER 5.01 a. For 100% of full power, Tg = 599.4*F C T ave = 572.5*F T
c = 548*F L C. I]
h(599.4*F) = 617 Btu /lbm '
h(548*F) = -547 Btu /lbm L en c cri' net r ref ' uc- Re? ah = 7G Btu /lbm go.g} ,
(+t:S=for-the-procedure}L Q = inah Ok $1 'h C. r 4 T E C 0S
= (134 x 10 lbm/hr)(70 Btu /lbm) Cod
= 9380 x 10 Btu /hr
'
'
(+0,76- f o r-t he-proc ed u r e- us i n g-t he-v al ues-abov e}- , 9380 x 10 Rtu 3600 sec
=
9380 3600 10W W Dd
(9380)(1055) #
3600
= 2748 MW
.
4AG.15_for the-conyersiound-for-the-conect-order-of-magnituder)
t (01.5 % h J
- Section 5 continued on next page -
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3 CALVERT CLIFFS
- . . August 20, 1985
. ,i .
I Points Available ANSWER 5.01 (contd)
b. If the flowrate of the primary coolant is reduced, nore heat will be added to the coolant as it passes through the cor This will raise the core aT (+1.0).
l Reference (s) -
1. .Calvert Clif fs: Systems Description No. 5, " Reactor Coolant
, Systen," Figure A-31 and p. . Generic: Nuclear Power Plant Operator Training Program,
" Thermodynamics, Fluid Flow and Heat Transfer for Nuclear Power Plants," Duke Power Company, pp. 56-5 . Generic: Steam Tables, C-E Power System i
^ l,
- Section 5 continued on next page -
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4 CALVERT CLIFFS
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August 203 1985 Points Available OttESTION 5.0? What is the expected axial neutron-flux shape; i.e., sketch (qualitatively) the thermal neutron flux as a function of I
axial distance? Explain the rationale for the shape of your sketc (2.5)
)
l If the flowrate of the primary coolant through the core was reduced, explain the effect this would have on the axial neutron-flux shap (1.0)
I
.
I
I l
,
l
- Section 5 continued on next page -
. . - , . .- ._, . . _ . . _ . . . _ _ .. _ __ . CALVERT CLIFFS
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August 20, 1985 Points Available ANSHER 5.02 core height h
neutron core midplaae flux (+1.0 for shape)
-
With a fuel-burnup status of 10,000 MWD /MTU, the core is at the End-of-Li fe (EOL) (+0.5). At E0L, the U-235 fuel has been disproportionally burned-up in the center of the core (+0.5),
which has depressed the ROL peak in the center of the cor l, Additionally, the fuel is disproportionally burned in the bottom half of the core and this shift of the relative fuel concentration to the top half of the core is responsible for the peak in the top half of the core (+0.25). At full power, the moderator density decreases with core height. The higher moderator density in the bottom half of the core is responsible for the thermal-neutron peak in the bottom half of the core (+0.25). If the flowrate of the primary coolant was reduced, the tem-peratures of the reactor core would rise. The increase in temperature would be greater in the top half of the core than in the bottom half. This would cause the neutron flux distribution to shift to the bottom of the cor (+1.0)
Deference (s) C-E Training Center: " Reactor Theory," Flux Distribution, Section 6.3, Figures 6-24 through 6-29 and Figure 6-38.
.
- Section 5 continued on next page -
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6 CALVERT CLIFFS
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August 200 1985 Points Available OllESTION 5.03 What fraction of the thermal power generated in the nuclear-reactor core is converted to electrical output power; i.e.,
what is the efficiency of the power plant? (1.0) If the power plant is not 100% efficient, some of the thermal energy that is produced by the nuclear reaction is not con-verted into electrical output. What happens to this lost or unused energy? (1.0) Determine the " apparent power" produced by the main generato (1.0) If the "MVAR loading" of the main generator was increased and the "real power" maintained constant, what would be the changes (qualitatively) in the pf (power factor) and in the " apparent power?" (1.0)
- Section 5 continued on next page -
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7 CALVERT CLIFFS
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August 20, 1985
.
Points Available ANSHER 5.03 a. 845/2750 = .307 gg 'Y2W
' 2 3 /. 3 %
= 30.7% efficient (+1.0 for the numerical procedure and for the correct
-
order-of-magnitude)
h. The other 70". is " lost" or unused heat. Primarily, this heat is removed from the secondary fluid in the condenser by the water of the Circulating Water System (+0.5). The heat absorbed by this cooling water is then transferred to the Bay (+0.5). The heat that is not removed via the condenser cooling
, is dissipated by maintaining the temperature of the plant piping and components (+0.5).
(+1.0 max)
c. Real power = (apparent power) (pf) (+0.5)
Apparent power = 845 M/0.95
= 889 M_VA (+0.5)
d. The pf would decrease. (+0.5)
The apparent power would increas (+0.5)
Reference (s) Calvert Cliffs: System Description No. 35, " Circulating Water System, pp. 1-3 and Figures 35-1, 2 and . C-E Training Center: " Pressurized Water Reactor Training Manual," Simulator Training Manual, Turbine-Generator Section, pp. 30-3 ,
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8 CALVERT CLIFFS
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August 20, 1985 Points Available OVESTION 5.04 If the control rods (CEAs) were inserted while in the manual sequential-(MS) mode until Group 5 reached 50 in., and if the boron concentration were adjusted to maintain the present power level of 100%; what would be the new concentration of boron required for steady operation? Neglect any effect from changes in the xenon or samarium concentratio (1.5) If the same maneuver was executed with the fuel burnup equal to 2000 MWD /MTV on cycle VII, the change in the boron concen-tration would be different. Explain what has changed and wh (1.5)
ANSWER 5.04 Using Figure 1-11.B.6, 1 c. 4/
ao rods = M (+ c.y/ (0.95t}(80 ppm /%) = 6BSp)pm-boron (+0.5)
200-p=IQ2 ppm (+0.5)
ss IG7 The boron worth (and inverse boron worth) would be different at BOL (+0.25). In particular, the inverse boron worth is larger so that the change in the boron concentration for a 0.85% change in reactivity would be greater (+0.25).
g A core at BOL, as contrasted to E0L, has more U-235 fuel and correspondingly more boron. With more fuel and more boron i
in the core, a change in the boron concentration would have a smaller effect on the neutron population. Hence, the boron worth would be less and the inverse boron worth would be larger (+1.0).
Reference (s) Calvert Cliffs: NE0G-7, Rev. 10, Figure 1-II. . SONGS 2 and 3: " Reactor Theory Review," Student Handout, pp. 40-41.
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9 CALVERT CLIFFS August 20, 1985
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Points Available OUEST10N 5.05 The Unit 1 power plant described prior to Question 5.01 (100% power, AR0, etc.) is to be taken from 100% to 80%
of full power. Assume T ave = Tref at both 100% and at 80%
powe What would be the magnitude of the change in reactivity (in % ap) due to the change in power in the power plan (0,5) Assuming that this maneuver takes 20 minutes or less, make a sketch on the provided graph (Figure 5.05 Question) of the xenon worth as a function of time. Show time from the point in time of starting the down-power maneuver and for the next 50 hour5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> (1.5) What five (5) factors must be considered in order to deter-mine whether the reactor plant can be returned to 100% power
, immediately after a trip from 1,00%? (1.5)
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10 CALVERT CLIFFS August 20, 1985 Points Available ANSWER 5.05 a. Using the power defect curve, Figure 1-II.C.1, reactivity change = 0.25% a (+0.5)
b. See the attached figure, Figure 5.05 Answe (+1.5)
I c. * CEA worth e boron e denon o samarium e power defect (+0.3 each)
Reference (s) Calvert Cliffs: NE0G-7, Rev.10, Figure 1-II. I Generic: " Reactor Theory Review Student Handout," C-E l
. Power Systems, Nuclear training. pp. 24-2 . Calvert Cliffs: NEOG 6, Attachment 6- Calvert Cliffs: NE0G-7, Rev.10, Figures 1-II.D.1 and 1-II.D. . Generic: " Reactor Theory Review Student Handout," C-E Power Systems, Nuclear Training, pp. 32-3 I
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11 CALVERT CLIFFS
, , August 20, 1985
- 5. 0-L5-LQ x
s - w
"
, -3. 0
- '
'
.l
,
- t
- LO 2D 30 41 50 60 O 10 HOURS
'
FIGtlRE 5.05 (00EST10N)
.a
- Section 5 continued on next page -
i e
,- , , , _ . , . . . , , _ _ _ . _,w v -,-_.-.,,.m,,.--, . . . . , _ , , . _ , . _ . , , . , . . , , _ ,~,7 ,,r..-,.. . ,-,c, - ,. v ,,-- ,_,,
.. _-
12 CALVERT CLIFFS
- -
August 20, 1985
,4 4 7. -+ $%
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- LO O 10 2D 9 HOURS FIGURE 5.05 (AMSWER)
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13 CALVERT CLIFFS August 20, 1985 Points Availab OtlESTION 5.06 What will happen to the current drawn by the motor of a Reactor Coolant Pump as the system is heated from 200 F to 564*F?
Explain your answe (1.0) What would happen to the current drawn by the motor of a Reactor Coolant Pump if a void forms in the reactor vessel during accident conditions? Explain your answe (1,0)
ANSWER 5.06 Power decreases (+0.33)
The pump is pumping the same volumetric flow rate but the density of the water decreases (+0.34) which causes a lower mass flow and less pump power. (+0.33) Power decreases (+0.34)
The pump is pumping against a lower discharge head due to the reduction of head losses in the area of the ' voi (+0.66)
CR Reference (s) p, ,, t em , v M r e s [0, NJ ,0 M Pc '
,>
' #*' ## ' ' Calvert Cliffs: CA c a c ; rr> a ci:' [ r. L'-]
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14 CALVERT CLIFFS
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August 20, 1985 Points Available 00ESTION 5.07 Consider a centrifugal pump which is operating with the following values for the parameters power, flowrate and head:
power = 100 hp l flowrate = 1800 gpm
) head = 150 f If you doubled the speed of the pump, what would be the corresponding values for power, flowrate and head? (1.5) In the situation where alternate core cooling is required (i.e., there is a SIAS), the reactor-coolant pressure may nevertheless exceed the LPSI pump discharge head. Describe the design feature of the Safety Injection System that exists to prevent the problem of running these centrifugal pumps with no flowrat (1.0)
ANSWER 5.07 flow 2 * IIU"1 x 2- = 3600 gpm head 2 = headyx2 2 = 600 ft power 2 = p wery x 2 3= 800 hp (+0.5 each) Each LPSI pump has a 40 gpm recirculation line that takes water upstream of the LPSI shutoff valve and returns the water to the RW (+1.0)
Reference (s) Generic: Nuclear Power Plant Operator Training Program,
" Thermodynamics, Fluid Flow and Heat Transfer for Nuclear Power Plants," Duke Power Company, p.15 . Calvert Cliffs: System Description Nos. 7 and 8, " Safety Injection and Containment Spray Systems," July 1983, p.18
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15 CALVERT CLIFFS August 20, 1985 00EST10N 5.08 What effect will the following events have on the Departure from Nucleate Boiling Ratio (DNBR)? Explai Dropping a rod at 100% powe (1.0) Reactor coolant average temperature is reduced by 5*F while maintaining 100% powe (1.0)
ANSWER 5.08 a. The DNBR will decreas (+0.33)
The dropped rod will reduce the flux in the local area (+0.34)
and will cause power to increase in other areas of the cor (+0.33)
b. DNRR will increase. (+0.33)
Reduced temperature will increase subcooling (+0.34) causing steam bubbles to quench more rapidly. (+0.33)
Reference (s) Calvert Cliffs: C-E Thermal Hydraulics, p.14
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16 CALVERT CLIFFS
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August 20, 1985
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I Points Available
! -OUESTION 5.09 Comparing a slightly subcritical reactor (shutdown margin = 17,)
to a greatly subcritical reactor (shutdown margin = 5%), explain how an addition of 0.5% of positive reactivity will affect the following:
. The change in the count rate (1.0) The time to reach a stable count rat (1,0)
ANSWER 5.09 The slightly (greatly) subcritical reactor will have a larger (smaller) increase in count rat (+1.0) The slightly (greatly) subcritical reactor will take a longer (shorter) ume to reach a stable count rat (+1.0)
~1 Refer'nce(s)
1 Calvert Cliffs: C-E Reactor Theory, pp. 147-14 End of Section 5 -
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17 CALVERT CLIFFS
. August 20, 1985 l !- 6.0 PLANT SYSTEtt DESIGil, C0t4 TROL ANO INSTRUMENTATION (25)
-; Points t Available OllESr10N 6.01 i
.
- This question refers to the operatinn of the Auxiliary Feedwater
,
(AFW) Syste Explain the response of the AFW System to a continuously dropping level in S/G 11: include information on setpoints, the valves whose positions are affected and the pumps which
_l
.) are turned on or of (1.5) Explain how the Auxiliary Feedwater System actuation signal
'
is developed; include sensors logic and outpu (1.5)
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. . 18 CALVERT CLIFFS August 20, 1985
.
Points Available ANSWER 6.01 a. e starts the motor-driven pump (+0.5)
e opens the two steam-supply valves 4070,1 to the AFW System turbine pumps (+0.5)
e after a 60-second time delay, sends a signal to shut
,
both main feedwater isolation valves and to run-back both Steam-Generator feed-pump turbines to a predetermined minimum speed setting (+0.5). (ut.'.'r 1 th 4 5'ha aos /m ferf c. rd)
b. The AFS monitors four (4) channels of wide range level indication for each S/G (+0.5). Each low-level bistable module converts the analog level signal to a digital signal I
and compares it to a setpoint level of -170 inches (0.5).
These bistable modules feed four logic matrices, two per
.
S/G, which implement 2/4 logic. The output signals from the
! four low-level logic matrices enter two AFS START LOGIC MATRICES, each of which is an OR gate receiving one signal from a S/G 11 low-level matrix and one from S/G 12. In I
this case, both channels of AFS START would be actuated J
(+0.5).
Reference (s) Calvert Clif fs: System Description No. 34', " Auxiliary
'
Feedwater System," Figures 34-1, 34-3 and pp. 45-47, a 63-64
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19 CALVERT CLIFFS August 20, 1985 Points
-< Avail abl e 00ESTION 6.02 Calvert Cliffs Unit I has been operating at 40% of full power for the last 20 days when a rea-tor trip occurs, Explain in detail the expected automatic response of the six (6) Steam-Dump / Turbine Bypass valve (1.5)
. Identify and explain two (2) indications (other than valve-position indications or limit lights) in the control room that could be used to verify that the Steam Dump / Turbine Bypass valves had opened excessivel (1.5)
.
ANSWER 6.02 Both steam-dumps ramp OPEN to lower Tave to 535*F (+0.6).
Some bypass valves open sequentially to lower steam-header pressure to 900 psia or T ave to 535*F (+0.9). Excessive cooldown would reduce T ave and could reduce T ave I below the no-load value. The 3/G pressure would be low due
' to the excessive cooldown. And the lowering of Tave would produce a low pressure in the Pressurize (+0.75 each,
&1.5 max).
,
Reference (s)
l Calvert Cliffs: System Description No.19, " Main Steam and MSIV System," pp. 16-2 I-I J
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21 CALVERT CLIFFS August 20, 1985 Points Available OilESTION 6.04 If the high-level alarm setpoint is exceeded, what automatic action (s), if any, should occur for each of the following radia-tion instruments? Do not include the initiation of alarm CVCS-letdown radiation monitor (0.66) Waste-Gas System discharge radiation monitor (0.67) Spent-Fuel Storage-Pool area radiation monitor (0.67)
ANSWER 6.04
, There is no automatic actio (+0.66) The Waste-Gas System discharge isolation valves (discharge isolation valve and redundant isolation valve) should clos (+0.67)
c. -The-Sper.t %e' St-erage "cel-Vent-i-let4en-System-should-eete-srtert . 3.67)
NQilE (C. L 'O(:
Reference (s) Calvert Cliffs: System Description No. 14A, " Waste-Gas System," June 1981, pp. 1 . Calvert Cliffs: System Description No.10, " Spent Fuel l Pool and Spent Fuel Pool Cooling and Purification Systems,"
i July 1983, p. 34 t
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. . 22 CALVERT' CLIFFS August 20, 1985 Points Available OUESTION 6.05 Explain how power supply reliability is achieved for t'4e 120 VAC instrument power buse (Disregard computer power.) (1.5)
ANSWER 6.05 There are four (4) 120 VAC busses per unit that are directly supplied by inverters. (+0.25) Each of the inverters can be supplied by 125' VDC (+0.5) or 120 VAC regulated powe (+0.5)
Two of the inverters on each unit are supplied by DC power from the other unit (+0.25).
Reference (s) Calvert Cliffs: System Description No. 54, Figures 54-1 and 54- I
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23 CALVERT CLIFFS August 20, 1985 Points Available OUESTION 6.06 The following questions refer to the Component Cooling Syste The temperature of the water of the Component Cooling System must be maintained low enough to adequately carry the loa At what point in the System is the temperature monitored for control? To what value is this temperature controlled? How can this temperature be controlled? (2.0)
b. What is the expected response of the System to a SIAS at Unit I with the component cooling pump 13 selected to the 480 volt bus 11. Assume that there has not been an initiation of the CSAS and the CI (1.5)
ANSWER 6.06 It is the outlet water of the System heat exchanger that is controlled (+0.5). It is controlled to 95*F (+0.5). This control is accomplished by the automatic positioning of the temperature-controlled heat-exchanger bypass valve (+0.5) and/
or by the manual control of the heat exchanger's salt-water
, outlet valve (+0.5).
b. lipon the initiation of SIAS:
e component cooling pumps 11 and 12 start e if the breaker for pump 11 does not close within 1 second, pump 13 starts e the shutdown-cooling heat-exchanger outlet valves are opened e the component-cooling heat-exchanger salt-water inlet and outlet valves are close (+0.5 each, +1.5 max)
'
Reference (s) Calvert Cliffs: System Description No. 40, " Component
"
Cooling System," pp. 5,17, 6 Section 6 continued on next page -
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. . 24 CALVERT CLIFFS August 20, 1985 Points Available OtlESTI0t1 6.07 Figure 6.07 (Question) on the next page shows the one-line electrical diagram for the Power Range Safety instrumentatio i Two (2) detectors are shown. What type of detectors are
} these and how are they constructed? (1.5) What is the relationship between the output of amplifier-1 and its inputs? (What is the output of Amp I?) (0.5)
c. What is the relationship between the output of amplifier-2 and its inputs? (What is the output of Amp 2?) (0.5) What is the relationship between the output of amplifier-3 and its inputs? (What is the output of Amp 3?) (0.5)
, The Internal Vibration Monitoring System (IVMS) receives information from the power range safety channels. What information is received and briefly describe what is done
, with the information to obtain an indication of vibration? (1.0)
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. . 25 CALVERT CLIFFS August 20, 1985
,I a o a a o o a %), a n a , ~,( )- ,( ),- 1, a 7), % {-})( i, d!
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-
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. 26 CALVERT CLIFFS August 20, 1985 Points l ANSWER 6.07 a. The power range _ safety channels utilize UICs (+0.5). Each UIC detector assembly consists of two independent UICs installed end-to-end in a common cylindrical housing. Each UIC contains a high voltage electrode and a signal electrode which connect to respective high voltage and signal cable Hitrogen filler gas occupies the space between the two cylindrical electrodes. The VICs operate in the ionization region. The neutron-sensitive material used in the UIC is a coating on the electrode surfaces consisting of baron enriched with baron-10. (+1.0)
b. U-L (+0.5)
c. A+B/2 (+0.5)
d. sum U+L/8 (+0.5)
e. The A+B/2 output signal from each PR safety channel supplies an input to the IVMS (+0.5). The system monitors reactor core and core-barrel motion by comparing the differential between input signal pairs and a frequency analysis of the signals is performed (+0.5).
I Reference (s)
{' Calvert Cliffs: System Description No. 57, " Nuclear Instrumentation, Revision 1, June 1984, pp. 35-36, Figure 57-14, pp. 39-42, 49-5 l
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- - 27 CALVERT CLIFFS August 20, 1985 Points Avail able 00ESTION 6.08 What will cause AND what action will result from a Recircula-tion Actuation Signal (RAS)? (1.5) It may be necessary to flush the core during long-term cooling following a loss of coolant. Describe one flow path for per-forming this flus (1.5)
'
ANSWER 6.08 . Caused by RWT level decreasing below approximately 30"'
(+0.5). Action that results:
- Containment sump isolation valves open ,
- Both LPSI pumps stop e Mini flow recirculation isolation valves receive a shut signa (+0.33 each)
, . Containment . sump > LPSI Pump > recirculation line > SDC g return header > hot leg (+0.3 each), g HPSI pump > Aux. HPSI beader > CVCS > Pzr aux. spray >
Pzr > surge line > hot leg (+0.21 each).
Reference (s) Calvert Cliffs: SD #7 and 8, pp. 65,67,68.
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, , 28 CALVERT CLIFFS August 20, 1985 Points Available 00ESTION 6.09 Would containment design pressure and temperature limits be exceeded if'the total Safety Injection System failed and all other Engineered Safety Features Systems functioned normally following a Loss of Coolant Incident (LOCI)? (0.5) What two (2) other systems would reduce containment pressure and temperature during a LOCI? (0.5)
ANSWER 6.09 No (temperature and pressure would remain within limits).
(+0.5) CC (+0.25)
CTMT Air Recirculation Syste (+0.25)
Reference (s)
' Calvert Cliffs: SD #1; Containment System, p. 3 J - End of Section 6 -
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. . 29 CALVERT CLIFFS-August 20, 1985 l
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7.0 PROCEDURES--NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL (25)
Points Available OUESTION 7.01 Unit 1 experiences a reactor trip from 100% powe The first step of the Immediate Operator Actions is to depress the reactor trip buttons and to verify that two (2) results 3 have taken plac List those two (2) response (1.0)
I If the RPS fails to trip the reactor and de-energizing the CEDM MG sets still results in two (2) CEAs failing to fully insert, then what action should the operators take? (1.0) What four (4) responses of the Turbine / Generator should be verified as part of the Immediate Operator Actions? (1.0) If the Generator output breakers have not tripped and the Turbine stop valves cannot be confirmed to have shut, then what action (s) should the operators take? (0.5)
- Section 7 continued on next page -
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. . 30 CALVERT CLIFFS August 200 1985 Points Available ANSWER 7.01 a. * CEAs are fully inserte * Reactor power is decreasin (+0.5 each)
' " Increase the boron concentration (+0.5) by 200 ppm for each CEA that fails to fully insert." Hence, increase the boron concentration by 400 ppm (+0.5).
c. e Turbine is tripped, o Generator output breakers are ope * Generator field breaker is ope e Exciter field breaker is ope (+0.25 each) Shut the MSIV (+0.5)
.
Reference (s) Calvert Cliffs: E0P-1 (Reactor Trip), Rev. 7, pp. 2- ,
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. . 31 CALVERT CLlFFS August 200 1985 Points Available 00EST10N 7.02 Explain what will happen and what operator action is required for the unit I and unit 2 turbine upon experiencing a loss of power to the 11 DC bu (1.5)
i ANSWER 7.02
.
The unit 1 turbine will automatically trip (+0.5). The unit 2 turbine will not automatically trip but all remote and automatic electrical trips will be lost (+0.5). An operator must be stationed at the turbine front standard in direct conmunication with the control room to allow for manual tripping of the turbine (+0.5).
Reference (s)
1. Calvert Cliffs: GS0 Standing Instructions.
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. 32 CALVERT CLIFFS August 200 1985 Points Available 00ESTION 7.03 List six (6) control room indications of a steam line rupture on Unit (3.0)
ANSWER 7.03 Rapid decrease in steam generator pressure Reactor trip on low steam generator pressure
' Rapid drop in reactor coolant temperature 4 Rapid drop in reactor coolant pressure Loud noise and poor inplant visibility depending on the location of the rupture 6.- Low pressurizer level Safety injection System Actuation Signal (SIAS) Containment Isolation Signal (CIS) and Containment Spray Actuation Signal (CSAS) if location of rupture is inside I containment y, j,.;. .. : t rn F. e /'
'
(+0.5 each, +3.0 max) , g73 f f , ; ; f, - fc,f II, liFAS ih CC K $IGWAL Calvert Cliffs: E0P-4 (Steam Line Rupture), Rev. 9,1, p. 2.
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33 CALVERT CLIFFS August 20, 1985 Points Avail able OilESTION 7.04 What indication does a fuel handling operator have that a fuel bundle binds 1 in, from the bottom while loading fuel? (0.75) What indication does a fuel handling operator have that a fuel bundle is mechanically bound during fuel removal? (0.75)
ANSWER 7.04 n, u t,0 c is u ut n M.'n Lente S L A t k'
b.44 The Dillon cell reading (+0.35) and high load interloc (+0.4) The tape-measure-attashed-to-the gripper will act read propef4 (+0r7&)-MV-sur-ve4 Manse--after-loading-tac core.)
Reference (s)
! Calvert Cliffs: -GA M O. I. -2iC
1 i . I i
)
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- 34 CALVERT CLIFFS August 20, 1985 i
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i Points Available l OUESTION 7.05 List the weekly administrative external dose limits for individuals 18 years of age or olde (2.5)
~
ANSWER 7.05 g
External Dose:
The weekly administrative exposure limits, for individuals 18 years of age or older, where a week is defined as 2400 h Friday through 2400 h Friday, are: Dose to the whole body, head, trunk, blood forming organs,
'
lens of eyes or gonads shall be limited to 300 mrem / week until the quarterly dose accumulation reaches 900 mrem (alert point and quarterly limit for 18-year olds) and 150 mrem / week for the balance of the quarter, Dose to the skin of the whole body shall be limited to 2000 mrem / wee Dose to the hands and forearms, feet and ankles shall be limited to 4500 mrem / wee (+0.5 for each underlined limit)
Reference (s) Calvert Cliffs: CCI-800A, Attachment (1), p. 9.
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. . 35 CALVERT CLIFFS August 200 1985
!
Points Available 00EST10N 7.06 For each of the situations below indicate whether the plant should be tripped immediately. For situations which do not require an immediate trip explain at what point a reactor trip, if any, is required assuming conditions continue to deteriorate. Assume plant has been operating for 1 week at 90% power and consider each situation separatel I A rupture occurs in the Service Water subsyste (0.5) The motor on the operating component cooling pump fail (0.5) It is discovered that containment integrity has been breached when a blind flange is found improperly secure (0.5) An unexplained dilution raises power by 5%. (0.5) Instrument air pressure drops to 75 psi (0.5)
. The main journal bearing metal temperature is 230*F (5*F above the alann set point) for the Unit I turbin (0.5)
I i The main journal bearing metal temperature is 225*F (5*F above the alarm set point) for the Unit 2 turbin (0.5)
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. . 36 CALVERT CLIFFS August 20, 1985
.
Points Available ANSWER 7.06 a. Trip reactor (+0.5).
b. Trip if not restored in 10 min (+0.3) or alarm received on RCP thrust bearing temperature (>195'F) (+0.2)
rer>> t5- i.2 : w ncr s rn.Mr n [ H. .i:s c. No trip, (+0.2) no-t-ond440n will rcquire tri (;0.3)'
d. No trip, (+0.2) only if dilution raises power to RPS high power trip (+0.3).
I e. No trip, (+0.2) trip when pressure reaches 50 psig (+0.3).
, f. No trip, (+0.2) trip at 250*F (+0.3).
Me, 4 g. Trip reactor (+0.5).
Reference (s)
, Calvert Cliffs: A0P 3, p. 3 j Calvert Cliffs: A0P 4, p. . Calvert Cliffs: A0P 6, pp.1-2, 7"E 3 6 /< /
4 Calvert Cliffs: A09 7, p. 4 Calvert Cliffs: A0P 70, p. . Calvert Cliffs: A0P 7E Unit 1, p. . Calvert Cliffs: A0P 7E Unit 2, p. 3.
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37 CALVERT CLIFFS August 20, 1985
Points Available 00ESTION 7.07 Specify the requirements that an individual must satisfy to serve as an escort in a controlled are (1.0)
ANSWER 7.07 The escort must be Contro'iled Area qualified and assigned to the plant for a minimum of 6 months or authorized by the SRC to act as a Controlled Area Escort. (+1.0)
Reference (s) Calvert Cliffs: CCI-800A, Attachment (1), D, p. 3 ESTION 7.08 What requirements are imposed on the Pressurizer if RCS boron concentration is to be changed by 50 ppm or greater? (1.0)
ANSWER 7.08
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Whenever boron concentration is changed by 50 ppm or greater, the Pressurizer spray valves shall be operated, consistent with pressure requirements, until the boron concentration of the Pressurizer is within 10 ppm of RCS concentration. (+1.0)
Reference (s) Calvert Cliffs: OP-4 (Plant Shutdown from Power),
Rev. 7, D, p. ..
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38 CALVERT CLIFFS August 200 1985 Points Available OUESTION 7.09 a. What four (4) indications are used to continually verify natural circulation? (1.6)
b. What operator actions can be used to affect each of the fellowing conditions which are required to accomplish natural circulation? Maintain RCS inventory (+0.A) Maintain adequate Steam Generator water inventory (+0.8) Maintain adequate RCS subcooling (+0.7)
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ANSWER 7.09 a. AT is less than full power AT T is constant or decreasing c
1 Th is stable No abnormal difference between Th and core exit thermocouples (+0.4 each) . Proper operation of CVCS (+0.4) Feeding the SG with main or auxiliary feedwater (+0.4)
Discharging steam using Turbine Bypass and/or Atm. dumps (+0.4) Rapid cooldown rate (+0.4) Pzr press, maintained high (+0.3)
IfCS j Reference (s)
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t Calvert Cliffs: E0P-12, Att (1), pp. 1-2.
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. . 39 CALVERT CLIFFS August 20, 1985
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e Points Avail able
[ OUESTION 7.10 During operation at power a piping rupture occurs on the main header of the instrument air system. List the sequence of events and their set points that automatically take place for this even (2.0)
ANSWER 7.10 The standby instrument air compressor starts at 90 psi ' The automatic plant air to instrument air cross connect valve opens at 85 psig (instrument air pressure), The plant air header automatic isolation valve closes at 85 psig (plant air header pressure) causing the plant air compressor to discharge to the instrument air system only, The other unit's plant air compressor automatically starts
.
at 90 psig and supplies air to the affected units through normally open cross-connect valve {
.(+0.5 each)
f Reference (s) Calvert Cliffs: A0P-70 (Loss of Instrument Air), Rev. 7, Discussion, p. ~
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. . 40 CALVERT CLIFFS August 20, 1985 l
Points Available OUESTION 7.11 According to the loss of load procedure, a sudden large reduction in power demand would likely be due to one (1) of four (4) mal-
,.
functions or system changes. List these four (4) occurrence (2.0)
ANSWER 7.11 Spurious closure of a main stream isolation valve Spurious closure of one (1) or more governor valves, without
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a reactor trip Spurious closure of two (2) or more turbine stop valves, intercept valves or intermediate stop valves without a
,
4 Separation from the interconnected system (loss of lines 5051 and 5052)
(+0.5 each) ,
Reference (s) Calvert Cliffs: (Loss of Load), Rev. 7, Discussion, p. I
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41 CALVERT CLIFFS August 200 1985 8.0 ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS (25)
Points Available OllESTION 8.01 While operating at power, it is determined that the shutdown margin is 4.1% ak/k. What actions, if any, are required by Tech-Specs.? (2.0)
I ANSWER 8.01 i
With the SHUTDOWN MARGIN <4.3% ak/k, immediately (+0.5) initiate and continue boration at 140 gpm (+0.5) of 2300 ppm boric acid solution or equivalent (+0.5) until the required SHUTDOWN MARGIN is restored (+0.5).
Reference (s) Calvert Cliffs: Technical Specification, Section 3.1. ;
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42 CALVERT CLIFFS August 200 1985 Points 00ESTION 8.02 Complete the following table to indicate the minimum shift crew composition in the applicable modes for Unit 2 with Unit 1 in mode 5 or (2.0)
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I MINIMilM SHIFT CREW COMPOSITION #
LICENSE CATEGORY APPLICABLE MODES
. _
1, 2, 3, & 4 5A6 SQL OL Non-Licensed Shift Technical Advisor If the minimum shift-crew composition cannot be met, what action (s), if any, must be taken by the shift supervisor?
Specify any time requirements associated with the action (s)? (1.0)
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43 CALVERT CLIFFS
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Points Available ANSWER 8.02 LICENSE CATEGORY APPLICABLE MODES
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1, 2, 3, & 4 5&6 SOL 2 1 OL 3 2 Non-Licensed 3 3 Shift Technical Advisor 1 0 (+0.25 each) The shift supervisor must immediately act to bring the composition to the required minimum (+0.5). He has 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (+0.5).
Reference (s) Calvert Cliffs: Technical Specification Administrative Controls, Table 6.2- !.
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44 CALVERT CLIFFS August 200 1985 Points Available 00EST10N 8.03 What are the four (4) conditions that must be met for the Refuel-ing Water Tank to be considered operable in modes 1, 2, 3 and 4? (2.0)
ANSWER 8.03
The Refueling Water Tank shall be OPERABLE with; A minimum contained borated water volume of 400,000 gallons A boron concentration of between 2300 and 2700 ppm A minimum water temperature of 40"F A maximum solution temperature of 100 F in mode (+0.5 each)
Reference (s) Calvert Cliffs: Technical Specification, Section 3. a
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-, . 45 CALVERT CLIFFS August 200 1985 Points l OUESTION 8.04 For the leakage conditions shown below, indicate whether you would CONTINUE TO OPERATE indefinitely or SHUTDOWN under specific time requirements. Assume no other leakage than that listed. Consider each item separately, .5 gpm each, from five different valve packing gland (0.5) .2 gpm from a Th loop RTD weld (0.5) .3 gpm unknown leakage (0.5) gpm seat leakage on a Pressurizer safety valve (0.5)
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ANSWER 8.04 continue to operate shutdown , shutdown continue to operate (+0.5 each)
Reference ('s) Calvert Cliffs: Technical Specifications, pp. 3/4, 4-1 .,
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- 00ESTION 8.05 l
l Under what two (2) conditions can a temporary electrical
!
jumper be installed without meeting the logging requirements l CCI-117D, the instruction that addresses temporary electrical jumpers? (1.25) What are two (2) methods of verifying the proper removal of a temporary electrical jumper? (1.25)
ANSWER 8.05 . When approved procedures include instructions for installa-tion and lifting of the-jumper. (+0.6) When installation and lifting is covered in a maintenance reques (+0.6) . A second individual verifie (+0.6) Verify removal with a functional tes (+0.6)
Reference (s)
, . Calvert Cliffs: WJE 189, Calvert Cliffs: CCI-117D, pp. 2,3, I J
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- ' 47 CALVERT CLIFFS August 20, 1985 Points Available OtlESTION 8.06 What provisions are required in order to make a temporary change to a procedure? (3.5)
ANSWER 8.06 Temporary changes to procedures may be made provided: The intent of the original precedure is not altere The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected, The change is documented, reviewed by the POSRC and approved by
...
the Plant Superintendent within 14 days of implementation, i
(+0.5 for each underlined provision)
Reference (s) Calvert Cliffs: Tecnnical Specification Administrative Control, Section 6.8.3.
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. . - - _ _ - __ , . 48 CALVERV CLIFFS August 20, 1985 Points Available OilESTION 8.07 What are the four (4) administrative actions that shall be taken if RCS pressure reaches 2800 psia? (4.0)
ANSWER 8.07 The following actions shall be taken in the event a safety limit is violate The facility shall be placed in at least HOT STANDBY within I hour. (+1.0) The NRC Operations Center shall be nc'ified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Manager, Nuclear Power Department and the OSSRC, shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (+1.0) A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the POSRC. This report shall describe (1) applicable c cumstances preceding the violation, (2) effects i of the violation upon facility components, systems and structures, and (3) corrective action taken to prevent recurrence. (+1.0) The Safety Limit Violation Report shall be submitted to the Commission; the OSSRC; and the Manager, Nuclear Power Department within 14 days of the violatio (+1.0)
i Reference (s)
1 Calvert Clif fs: Technical Specification Administrative Control, Section 6.f ?
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Points Available
! OUEST10N 8.08 List the four (4) conditions which must be met in order that each Safety Injection Tank will be considered operable when in mode (2.0)
ANSWER 8.08 Each reactor coolant system Safety Injection Tank shall be OPERABLE with; The isolation valve open A contained borated water volume of between 1113 and 1179 cubic feet of borated water (equivalent to tank levels of between 187 and 199 inches, respectively) A boron concentration of between 2300 and 2700 ppm A nitrogen cover-pressure of between 200 and 250 psig (+0.5 each)
Reference (s) ,
j Calvert Cliffs: Technical Specification, Section 3. .
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50 CALVERT CLIFFS
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August 20, 1985
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Points Available i OtIESTION 8.09
.I When should the Technical Support Center and the Operational Support Center be activated? (0.9) What is the function of:
1. Technical Support Center? (0.8) Operational Support Center? (0.8)
ANSWER 8.09 When an alert or higher emergency action level is declare (+0.9) . To analyze current and projected plant status and provide advice and assistance during an acciden (+0.8) An assembly point for off-shift operators, HP personnel, and other emergency response personne (+0.8)
Reference (s) Calvert Cliffs: WJE 19 , Calvert Cliffs: ERPIP Sections 4.1.2 and 4.1.3.
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August 20, 1985 Points Avail abl e 00EST10N 8.10 l
What are the Tech-Specs bases for the requirement that the l reactor coolant system must have a minimum flow rate of 3000 gpm? (1.5)
ANSWER 8.10 A minimum flow rate of at least 3000 gpm provides adequate mixing, prevents stratification, and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 gpm will circulate an equivalent Reactor Coolant System volume of 9601 cubic feet in approximately 24 minutes. The reactivity change rate associated with baron concentration reductions will therefore be within the capability of operator recognition and contro (+0.5 for each underlined provision)
Reference (s) Calvert Cliffs: Technical Specification Bases, Section 3/4.1.1.3 BORON DILUTIO ,
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" END OF EXAMINATION
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