ML20136H061
| ML20136H061 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 01/03/1986 |
| From: | Burdick T, Higgins R, Mcmillen J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20136H060 | List: |
| References | |
| 50-483-OLS-85, 50-483-OLS-85-0, NUDOCS 8601090209 | |
| Download: ML20136H061 (95) | |
Text
m-U. S. NUCLEAR REGULATORY COMMISSION REGION III Report No.- 50-483/0LS-85-01 Docket No. 50-483 License No. NPF-30 Licensee: Union Electric Company Facility Name: Callaway Examination Administered At: Callaway Examination Conducted: December 17, 18 and 19, 1985 Examiner:
igg
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-Approved By:-
M>1 perator Licensing Section Dite/
Examination Sumary Examination administered on December 17, 18 and 19, 1985 (Report No. 50-483/0LS-85-01 )
Administered to six instant senior reactor operator candidates and one reactor operator candidate.
Results: The six senior reactor operator candidates passed and the reactor operator candidate failed.
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REPORT DETAILS 1.
Examiners
- R. L. Higgins T. M. Burdick
- Chief Examiner 2.
Examination Review Meeting An examination review meeting is no longer conducted. Specific facility comments concerning the Reactor Operator examination, followed by the NRC response, are included in the following paragraphs.
Question 2.08 Facility Comment:
The PDP has a variable capacity and at its minimum speed pumps 47 gpm.
T61.06.11 S-11 CVCS Lesson Plan Page 13, Item 6.
NRC Response: Agree. The answer key was modified to award full credit for the response "PDP".
Question 2.11 Facility Comment:
The question is confusing because it implies a fault existed on the sequencer when no fault did exist. Condition only lasts 10 seconds.
NRC Response:
Disagree.
The scenario was based on an actual event.
Sufficient information was given to determine the probable cause.
Question 3.07 Facility Comment:
At Callaway, the circ pump runback reduces unit load to 75%. OTN-DA-00001 Precaution and Limitation 2.4.
NRC Response: Agree. The answer key was changed.
Question 3.12.a Facility Comment:
In addition to the key's answer, we use the trip bypass switches for testability at Callaway. See ISF-SE-0JN31 and 32 and ISF-SE-00N34 and 35.
NRC Response: Agree. The answer key was modified to also grant full credit for the response " allow instrument testing."
Question 4.04.b Facility Comment:
In addition, the 80P R0 also opens DC control power breaker to NB02 Bus.
OTO-ZZ-00001 attachment 2, page 2 of 2, step 1.7.3.
NRC Response: Agree. The answer key was modified to also award full credit for the response " Opens DC control power breaker to Bus NB02."
2
Specific facility comments concerning the Senior Reactor Operator examination, followed by the NRC response, are included in the following paragraphs.
Question 5.13 Facility Comment:
The rods inserted in the core alter the flux profile causing more power to be produced'near the edge of the core, thus increasing " Buckling."
Large Pressurized Water Core Control Pages 3-23 and 3-24.
NRC Response: Agree. The answer key was modified to award full credit for the response " Increases Buckling."
Question 6.02. Facility Comment:
The purpose of the relationship between SR NIS and CCP suction valves is to protect against inadvertent dilution accidents while shutdown.
Technical Specification 3/4 Page 3-2 and 3/4 Page 3-5.
NRC Response: Agree. The answer key was modified to also award full credit for the response " protect against inadvertent dilution accidents while shutdown."
Question 6.06 Facility Comment:
In ad t tion to the key, "INP0 Significant Operating Experience Report 82-3".should be added.
NRC Response: Since the examiner did not have a copy of "INP0 Significant Operating Experience Report 82-3," additional facility explanation was needed.
The additional explanation received was
' "overpress'urization of AFW suction piping." The answer key was modified to also award full credit for the response "overpressurization of AFW suction piping."
Question 6.10 Facility Comment:
There is a fifth type of detector used at Callaway.
T66.06.09F Fire Protection Lesson Plan, Pages 2, 3 and 4.
NRC Response:
Required additional information, since the examiner did not have the reference.
The additional information supplied by the facility was that the fifth type of detector was a Thermal type, which alarmed whan a preset temperature was reached. The answer key was modified to accept this response as correct.
The facility should update the lesson plari used for license operator training of fire detection instrumenta' tion to reflect this additional type of detector.
Question 6.17. Fa,cility Comment:
Reference detector numbers with noun names. T61.06.11 Process and Area Radiation Lesson Plan, page 7.
l NRC Response: Agree.
If 4e examinee referred to a radiation detector by its number, rather thac its name, the t xaminee was awarded full credit. The answer key was modified accordingly.
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Question 8.16 Facility Comment:
In addition to the key answer, ODP-ZZ-00002, Definition 2.2, offers additional information.
NRC Response: Agree. The answer k'ey was modified to also award credit for the following responses:-
1.
Systems covered by the supplemental QA program for fire protection, 2.
Systems covered by the supplemental-QA program for seismic Class II/I, 3.
Systems covered by the supplemental QA program, Group D augmented.
Question 8.17 Facility Comment:
In addition, the purpose of local control is to allow a person, other than the regular operator, to operate the equipment for testing, etc.
APA-ZZ-00310 Steps 2.1.3 and 3.5.2.1.
NRC Response: Agree. The answer key was modified to also award full credit for the response " allow a person other that the regular operator to operate the equipment for testing or maintenance."
3.
Exit Meeting A formal exit meeting was not conducted due to inclement weather. The following information was conveyed by the NRC examiners to Callaway Training Department personnel during the course of the examinations, a.
Facility representatives were informed that three SR0s and one R0 definitely passed the oral / simulator portion of the licensing exam, and three SR0s were considered marginal.
b.
The examination room is not very conducive to exam administration.
Wind noise was.very distracting, rest rooms are located outside in a
" port-a-potty," ventilation is very poor, there is no pencil sharpener, there is no phone in close proximity to the room, and the temperature of the room is very hard to control, The lesson plans are not nearly detailed enough for exam c.
preparation.
Detailed system descriptions should be provided to the examiners prior to preparing future examinations.
d.
Much-information was not included in the original packages of reference material and had to be specifically requested, such as the index.for the piping and instrument diagrams, and the external radiation exposure limits.
e.
Not all topics in the KSA catalog, NUREG 1122, are addressed in Callaway reference material.
Callaway is encouraged to generate a plant-specific KSA catalog and ensure all topics mentioned are addressed in reference material.
4
f.
An additional sign for the simulator's side door is needed to preclude instrument technicians from inadvertently entering the simulator during NRC exams.
g.
The simulator has a number of problems in properly simulating certain events. Specific instances were pointed out to the simulator operators during the course of the simulator examinations.
h.
The computer programs used by the operators to calculate certain parameters, such as the ECP and heat balance, are not user friendly.
Several examinees had difficulty using the computers, but were reluctant to calculate the values by hand.
i.
Several examinees ignored off-normal or improper meter readings, assuming they were the " usual abnormal indications." This habit is potentially troublesome, since it leads to operators ignoring indications. Several operators operated for their whole scenario with the turbine-driven AFW pump controller in the improper position, a Tech Spec violation, yet they took no action.
J.
During _an inadvertent SI on one train, both diesels start, but the ESW pumps on the other SI train do not get a start signal.
If an inadvertent SI occurred with no ESW pumps running, one of the diesel generators could be destroyed. The ESW pump start logic should be modified to prevent that problem.
k.
No procedures are written to address problems in cold shutdown.
Operators have to improvise with very little procedural guidance.
1.
Several examinees had difficulty using the large book of steam and compressed water tables.
m.
Personnel were observed in the plant without safety shoes, and in high noise areas without wearing sound attenuation devices.
Callaway Plant personnel were very cooperative and accommodating.
n.
Simulator operators John Dampf, Sam Henderson, and Paul Moody, were especially helpful.
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
CALLAWAY REACTOR TYPE:
PWR-WEC4 DATE ADMINISTERED: 85/12/17 EXAMINER:
BURDICK, T APPLICANT:
INSTRUCTIONS TO APPLICANT:
Ugo separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each qu2stion are indicated in parentheses after the question. The passins stode requires at least 70% in each category and a final Stade of at locet 80%.
Examination papers will be picked up six (6) hours after tho examination starts.
% OF CATEGORY
% OF APPLICANT'S CATECORY VALUE TOTAL SCORE VALUE CATEGORY
___I__0__ _'I_1_0_
________ 1.
PRINCIPLES OF NUCLEAR POWER 25 0 50 PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
___1_00____'I_S 00
________ 2.
PLANT DESIGN INCLUDING SAFETY 25 1__
AND EMERGENCY SYSTEMS
___I_00____'_I_0
________ 3.
INSTRUMENTS AND CONTROLS 25 50 2
___1__0 25 0
___1_0
________ 4.
PROCEDURES - NORMAL, ABNORMAL, 25 0 EMERGENCY AND RADIOLOGICAL CONTROL 100.00 100.00 TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I heve neither sivon not received aid.
EPPL5CEUT 5~555sETURE
~~~~~~~~~~~~~~
I
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
--- iAEER557WARiEs-AEAi iEAniFEs AR5 FEUi5 FE5E QUESTION 1.01 (2.00) 0.
For an operator taking data for a 1/M plot, how will the Shut-down margin (SDM) affect the time elapsed before a stable count rate can be obtained after withdrawing rods ?
(0.5)
- b. How will the initial count rate affect the count rate at crit-icality ?
(0.5) c.
If the speed of the control rods were to somehow increase,what would be the effect be on:
1.
Rod height at criticality ?
(0.5) 2.
Count rate at criticality ?
(0.5)
QUESTION 1.02 (1.50)
During natural circulation cooldown, you notice pressurizer level suddenly increase after the initiation of pressurizer spray. Explain what is occurring.
(1.5)
DUESTION 1.03 (2.00)
- o. Since fuel temperature cannot be measured, what power distribut-ion limit is observed at Callaway to prevent exceeding the fuel temperatuce limit ?
(0.5) b.
If fuel temperature limit is 4700 de3's and cladding limit is 2200 des's., what limit must be observed to prevent exceeding the clad limit when fuel temperature is above 2200 des's ?
(0.5)
- c. Why will the fuel rod surface temperature peak towards the top of the core rather than the location of peak actual heat flux ?
(1.0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE ***xx)
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
--- isEER55isARiCs-REEi isis 5FEs An5 FEUi5 FE55 QUESTION-1.04 (2.00)
- o. How do each of the following parameters change (increase, decrease or no change) if one main steam isolation valve closes with the plant at 50% load. Assume all controls are in automatic that no trip occurs.
1.
Affected loop steam Senerator level (INITIAL change only) 2.
Affected loop steam generator pressure 3.
Affected loop cold les temperature
- 4. Unaffected loop steam generator level (INITIAL change only)
- 5. Unaffected loop steam generator pressure
- 6. Unaffected loop cold les temperature (1.5)
- b. Which of the reactor protection system signals could be expected to cause a reactor trip ? (If more than one, list the one that would reach the trip point first.)
(0.5)
DUESTION 1.05 (1.50)
A.
Explain the effect on Shutdown Margin of a 25 ppm boron addition while operating at 50% power and all control systems in automatic.
(.75)
B.
List three (3) factors, other than RCS boron concentration, which effect Shutdown Margin (SDM) and are used in the SDM calculation.
(.75)
GUESTION 1.06 (2.00)
Will the Departure from Nuclear Boiling Ratio (DNBR) increase, d2 crease or remain the same if the fo11 cuing plant parameters increase during power operation?
Consider each parameter independently.
A.
Reactor Coolant System (RCS) Pressure.
B.
RCS Temperature.
C.
RCS Flow.
D.
Reactor Power.
CO.5 ea.3 (2.0)
(*xxxx CATEGORY 01 CONTINUED ON NEXT PAGE **xxx)
1.
' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, FAGE 4
--- iREER557REsiEs-REAi isissFEE AR5 FEUi5 FE5R a
QUESTION 1.07 (2.50)
- c. Name the two methods of Xenon production and removal.
- b. Name the method of Samarium production and removal.
- c. Compare Xenon and Samarium in regard to their variation in concentration following a power reduction from 100% to 50% and remaining at 50% for two weeks.
QUESTION 1.08 (2.00)
- o. Why does Power Defect increase over core life when Doppler decreases?
b.
Which reactivity affect is dominant in Power Defect at BOL and EOL?
QUESTION 1.09 (2.00)
Using Figure 1-1 produce a graph on answer paper representing the cpproximation of differential rod worth versus rod height. Be sure to Icbel the axis and assign values to the scales.
QUESTION 1.10 (2.50)
Port of the reactor thermal safety limit is based upon not allowing coturation conditions at the core hot le3 State the reasoning behind this basis.
QUESTION 1.11 (2.00)
Tho refueling load pattern is intended to achieve low neutron leakage.
- c. Why are used fuel assemblies placed on the periphery of the core?
b.
What two advantages does lower neutron leakage offer?
DUESTION 1.12 (1.50)
Th3re are several mechanisms by which fuel clad could potentially fail during operations. Name three
(*xxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)
1.
' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5
~~~~5EEE5 65 IUIU5I~5EIY~5EIE5EEE~IU6'ELU56~EL6U QUESTION' 1.13 (1.50)
Accume the Callaway reactor operates for one full cycle at 100% and rods full out..No rod motion other.than for exercising is performed and no trips cecur. How will the axial flux distribution behave over core life? Include rcocons for its behavior.
(xxxxx END OF CATEGORY 01 ***x*)
- 3. -
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY GYSTEMS PAGE 6
QUESTION 2.01 (1.50)
This past year an operator, alerted by an urgent failure alarm on the rod centrol system, tripped the reactor manually. Apparently the 480 VAC power cupply:to the MG sets had been lost.
t
- c. How are the MG sets designed to compensate for a momentary power j
interruption?
l b.
What kind of urgent failure was caused by the power failure?
c.
What would have happened if the RO did not trip the reactor manually?
00ESTION 2.02 (1.50)
Earlier this year the reactor operator on duty was alerted to a turbine trip caused by a HI-HI MSR level.
- o. Where does the moisture separator drain tank rormally drain to?
- b. Where does it alternately drain to?
c.
Why is there a turbine trip associated with a high MSR level?
DUfSTION 2.03 (2.00)
Under certain conditions followins a turbine trip the 13.8 KV buses will fast transfer to the startup transformer.
- e. What are the three conditions?
- b. What will happen if these conditions are not met?
DUESTION 2.04 (1.50)
An RO accidentally closed a MFW isolation valve resulting in a reactor trip.
a.
How does valve closure stroke time differ from this event and a FWIS actuation.
- b. What automatic signals will cause a FWIS?
(xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE *****)
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7
QUESTION 2.05 (2.00)
During a startup in February the RO was feeding the SG's manually while ornitoring feed flow on a recorder. Due to faulty indication the operator ovGrfed the steam senerators resulting in a HI-HI level. Part of the ceuse attributed to this event was the fact that the auxiliary boiler was cut of service. Explain this reasoning.
QUESTION 2.06 (2.00)
Collaway was on RHR at 180 degrees and 360 psis with one RCP running. The oporators were restoring train A RHR from a pump operability test. Upon opening crossover valve 8716B the operator noticed B RHR flow so to zero end turned the B RHR pump off. 15 seconds later he noticed =ero preseure in the RCS and turned the RCP off. It was later determined that no RHR reliefs had lifted and no leaks had occured. The problem was attributed to a valving sequence error in the test recovery procedure.
- o. Why did openins 8716 B cause a loss of RHR flow?
- b. Why did the RCS depressurize immediatly afterward?
- c. What affect, did the event have on the RCP that was in operation at the time?
OUESTION 2.07 (2.50)
- c. State the norsial status / position during standby for RHR:
1.
suction 2.
heat exchanger flow control valves 3.
CCW flow 4.
train crosstie isolation
- 5. cold les return
- b. Which valves must be manually opened when initiating recirevlation flow for post accident core cooling?
(***** CATEGORY 02 CONTINUED ON NEXT PAGE
- )
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2.
~ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8
QUESTION 2.08 (1.50)
An RCS leak is defined as being within the capacity of the CVCS system.
- c. Which CVCS pump has the lowest flow capacity?
- b. Which ECCS pump has the highest flow capacity?
- c. Which ECCS pump has the highest head capacity?
QUESTION 2.09 (1.50)
- c. What two conditions will cause automatic diversion of flow around the letdown deminerali=ers?
- b. What condition requires manual diversion of flow around the letdown demineralizers? Eexcluding failure of any automatic functions 3 QUESTION 2.10 (1.50)
.Tha AFW system check valves at some nuclear plants have failed to seat prcperly allowing main feed water to back flow through the AFW pumps.
- o. How does this compromise the ability of AFW to function?
b.
What kind of LOCA condition requires the AFW for core cooling?
GUESTION 2.11 (3.00)
Th2 Callaway plant was operating at full power when a faulted startup transformer resulted in a plant trip. During the subsequent transient ctaam dump failed to reseat and an SI was initiated. While performins ths immediate actions for RT/TT and SI the RO noted that 1A RHR pump had not started and manually started it. If 18 RHR started explain why 1A had nut.
~
(***** CATEGORY 02 CONTINUED ON NEXT PAGE wxxxx)
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2.
PLANT. DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9
QUESTION 2.12 (2.00) c..
If;the RCS is designed for 2485 psis then why is there a requirement for an' overpressure protection system when the plant is at low
, temperature and pressure?
b_t_The pressurizer safeties are designed for a specific transient. Describe it.
huESTION 2.13 (2.50)
- o. Describe two possible flowpaths from the boric acid storage tanks to the chargins Pumps.
-b. The reactor makeup water isolation valve EV-1783 to the blending tee has an orficed bypass line around it. What is the purpose of this line?
When is it used? Why is it necessary?
.5 l
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(*****
END OF CATEGORY 02 *****)
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3.
INSTRUMENTS AND CONTROLS PAGE 10 GUESTION 3.01 (3.00)
An operator at Callaway was inattentive to his instrumentation and allowed o TS violation of the axial flux limits to exist for an extensive period of
- t i c.7.
)
- c. What two specific symptoms were present which identified the deviation?
- b. What two specific improvements have been made to help operators be more ottentive to axial flux?
QUESTION 3.02 (2.00)
The plant is being cooled down using the atmospheric steam dumps. Steam proosure is 700 psis. You proceed to cool down further.
- o. What control is adjusted to produce further cooldown?
- b. What precaution must be observed while adjusting the control?
c.
Why is this precaution necessary?
DUESTION 3.03 (2.00)
Th2 instrument technicians were performing a test on a SG pressure channel wh2n the associated steam generator level dropped to the LO-LO setpoint cnd tripped the reactor. Why did testing a pressure channel have such an affact on level?
QUESTION 3.04 (2.50)
As the last of the four power range NI channels was being adjusted the lavols in all four SG's began decreasing. In the effort to restore level tha plant experienced a Turbine Trip and FWIS but the reactor did not trip.
-o.
Why would adjusting of the last power range channel cause SG 1evels to decrease?
- b. What caused the TT and FWIS?
- c. Why didn't the reactor trip?
(***** CATEGORY 03 CONTINUED ON NEXT PAGE
- )
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3.' INSTRUMENTS AND CONTROLS PAGE 11 QUESTION 3.05 (1.00) s.,
t, Lict the four autnmatic start signals fog Component Cooling Water Pumps.
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GUESIION 3.06 (1.50)
Lict ALL automatic start signals for the motor driven and turbine driven AFW pumps.
t OUESTION 3.07 (1.50)
L, Noce' ALL plant conditions that will override tde turb!ne controls to either rcduce or stop turbine loadin3+
QUESTION 3.08 (1.50)
Ncce all the switches / controls / components that are manually operated
., incidental to realigrdus a control rod.
6 L
QUESTION 3.09 (2.00)
I
- c. What are the input signals for the RVLIS?
0 f
- b. What is the low point tap for RVLIS and how does this relate to the vessel elevation?
g, QUESTION 3.10 (2.00)
None'the four types of, leak detection used to monitor for RCS leakage and which are required to be operable by TS.
QUESTION 3.11 (2.00)
- s. At what pressure is SI blocked during a controlled plant cooldown?
)
s
- b. How do'es the operator know the block permiss;ive is activated?
l
- c. How does the operator block SI?
(xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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. 3.' INSTRUMENTS AND CONTROLS PAGE 12 QUESTION 3.12 (2.00)
Tha source and intermediate ranse nuclear instrumentation have both trip bypass and trip block features.
.o.
Why are the trip bypass controls needed?
- b. When do trip block features reset automatically?
GUESTION 3.13 (2.00)
- c. Name the six effluent flowpaths that are monitored by the RMS which provide automatic flow isolation features.
- b. What is the reason for heat tracing sample lines on some radiation monitorin3 channels?
4 (xxxxx END OF CATEGORY 03 xxxxx)
4.' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 13
~~~~ A5i5E55fEAE E5UTR6E-~~~~~~~~~~~~~~~~~~~~~~~
R QUESTION 4.01 (1.50)
Your crew is preparing food in the control room pantry which is causins cpurious alarms on the fire detectioin system. What administrative roquirements are necessary to silence this nuisance alarm?
QUESTION 4.02 (3 00)
As the RO in the CR during refueling operations you are alerted to an increasing containment instrument sump level.
- c. What CR instrumentation is used to determine level in the refueling pool?
- b. What component would be the highest possible source of radiation if the refueling pool were draining?
- c. What are eight the immediate operator actions for a decreasing RP level?
QUESTION 4.03 (1.50)
As.the RO, you are alerted to a loss of CCW flow and a decreasing surge tenk level,
- c. How much time do you have to restore CCW before a reactor trip must be initiated?
- b. Why must the reactor be tripped within this time frame?
QUESTION 4.04 (2.50)
Wh2n a fire results in CR evacuation:
- b. He is also directed to the DC switchgear rooms. Why?
- c. The RO is directed tu NB02. What for?
- d. When starting pumps locally how do you verify the pump is running?
G.
Why should train B equipment be used for plant shutdown?
(***** CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx)
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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 14
~~~~
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QUESTION 4.05 (1.50)
-Yco have indication that instrument air pressure is dropping and is at 100 psis but the service air header isolation valve is still open and the bcckup air compressors have not started.
- o. What is the cause of the low instrument air pressure?
- b. What actions are you required to perform if plant parameters begin to deviate from their normal operating band?
QUESTION 4.06 (1.50)
Th2 plant undergoes a turbine trip from 100% power and the RO notes that on automatic trip did not occur nor does it trip manually.
- c. What procedures are referred to in this situation?
- b. What is the next alternative following the failure of a manual reactor trip initiation?
QUESTION 4.07 (2.00)
Note five of the seven symptoms of a misaligned control rod per the precedure, OTO-SF-00004.
QUESTION 4.08 (2.00)
Nece four symptoms for a failed number one RCP seal.
QUESTION 4.09
('. 25 )
- c. Who controls the use of orange Hold Off Tags and white Hold Off Tass?
b.
Where are tags hung for manual and motor operated valves?
- c. Who can release a taSout?
(xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxx**)
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. 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15
~~~~ I65 EU55CEE~EUNTRUE~~~~~~~~~~~~~~~~~~~~~~~~
R QUESTION 4.10 (1.50)
- c. How often must a radiation worker review the requirements for entry into the RCA?
- b. How does the radiation worker document that review?
c.
How often is the radiation worker required to document the review?
QUESTION 4.11 (1.75)
Wh2n relieving the off going RO:
- o. How far back are you required to review the RO logs?
- b. What six additional actions must be performed to satisfy the turnover checklist?
QUESTION 4.12 (1.00)
Noce two exceptions to the requirement for independent verification on octety related systems.
QUESTION 4.13 (2.00)
Assume it is 0300 on 12/10/85 and the reactor is presently at 45% power.
Censidering the Delta-I penalty history listed below, when will you be 011 owed to increase power above 50%?
DATE TIME OUT TIME IN POWER (%)
12/09/85 0300 0318 85 12/09/85 1557 1633 65 12/10/85 0138 0300 45 QUESTION 4.14 (1.00)
A precaution in the Power Operation Procedure, OTG-ZZ-00004, states that cn rapid load decrease, the control rods should be promptly returned half way back to their original position. State why this is necessary.
(*****
END OF CATEGORY 04
- )
(************* END OF EXAMINATION ***************)
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.___..-__._-Q-
3 y'/
i
_,',- y L.
a.
I.'
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 16
~~~~ UERs66 U5U5C57 55dT TRIY5FER dU6 FLU 56 FLUE
~
T ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T
-ANSWER 1.01 (2.00)
- c. The closer to criticality, (less SDM) the longer time required to reach a stable count rate.
(0.5)
- b. A higher initial count rate will result in a higher count rate at criticality.
(0.5) c.
1., Critical rod height is not affected.
(0.5) 2.
Critical count rate will be lower.
(0.5)
REFERENCE FUNDAMENTALS OF NUCLEAR REACTOR PHYSICS CHAPTER 8 ANSWER 1.02 (1.50)
- c. Due to the decrease in pressurizer temperature / pressure [0.53 the system is voiding somewhere else [0.53 and forcing coolant into the pressurizer. CO.53 (1.5)
REFERENCE THERMAL HYDRAULIC PRINCIPLES AND APPLICATION TO THE PWR II PAGE 14-11 ANSWER 1.03 (2.00)
- 0.. Local power density-KW/FT.
st 4/
(0.5)
Fuel surface temperature is a function of heat flux and moder-ator temperature. E0.53 Moderature temperature is higher at the top of the core. CO.53 (1.0)
REFERENCE THERMAL HYDRAULIC PRINCIPLES AND THE AFPLICATION TO THE PWR II CHAPTER 13
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 17
~~~~ 55R5667U555C5, EEIT TRI 5555'5 6~FL0i6 fL6E T
~
ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T ANSWER 1 04 (2.00)
O.
1.
Decrease
-2.
Increase
- 3. Increase 4.
Increase
- 5. Decrease 6.
Decrease EO.25 each]
b.-Lo-Lo S/G Level E.53
' REFERENCE THERMAL HYDRAULIC PRINCIPLES AND APPLICATION TO THE PWR CHAPTER 12 ANSWER 1.05 (1.50)
A.
SDM is increased E.253, with power remaining constant, rod position will be higher (and boron concentration will increase)
[0.5].
(Since SDM is the instantaneous amount of reactivity by which the reactor is, or would be suberitical from its present condition.)
(.75)
B.
1.
Control rod position.
4.
Xenon concentration.
- 2. RCS average temperature.
(Time since shutdown.)
3.
Fuel burnup.
- 5. Power level.
- 6. Sanarium E3 9.25 each3 REFERENCE
. REACTOR CONTROL FOR LARGE PWR'S CHAPTER 7-13 ANSWER 1.06 (2.00)
A.
Increase.
.B.
.Dectease.
C.
Increase.
D.
Decrease.
[0.5 es.]
(2.0)
REFERENCE THERMAL HYDRAULIC' PRINCIPLES AND APPLICATION TO THE PWR CHAPTER 13-41 O
~1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 18
--- isEss55isAsiE5-REEi isinsFEE As5 FE5i5 FE5s ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T CNSWER 1.07 (2.50)
- c. Xenon is produced by fission E.253 and Iodine decay E.253 Xenon is removed by neutron absorption E.253 and decay E.253
- b. Samarium is produced by Promethium decay E.253 and removed by neutron capture E.25]
- c. Xenon will peak E.253 and then decrease to a new equilibrium below the initial value E.253 Samarium will peak E.253 and then return to the initial value E.253 REFERENCE REACTOR CORE CONTROL FOR LARGE PWR'S CHAP 4, PAGE 3 ANSWER 1.08 (2.00)
- o. As Doppl,er decreases the MTC is MORE NEGATIVE to a greater degree. E1.03
- b. Doppler dominates both at BOL and EOL. E1.03 REFERENCE REACTOR CORE CONTROL FOR LARGE PWR'S CHAPTER 3 ANSWER 1 09 2
y REF ERENCE REACTOR CORE CONTROL FOR LARGE PWR'S PAGE 6-18 ANSWER 1.10 (2.50)
If caturation conditions were allowed to exist at the hot les then further increases in core heat output would be undetected by the hot les RTD and protection would be de3raded.
' REFERENCE THERMAL HYDRAULIC PRINCIPLES AND APPLICATIONS TO THE PWR II, PAGE 13-53
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 19
--- iAEER55iRARiEE-REAi isAn5 FEE AA5 FEUi5 FE5E ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T J.NSWER 1.11 (2.00)
- a. Used fuel will produce a lower neutron flux at the core periphery.E1.0]
b.
1.
increased net reactivity Elonger cycle]
2.
reduced neutron embrittlement of the vessel E.5 each3 REFERENCE LARGE PWR CORE CONTROL 1-27 AND 28 ANSWER 1.12 (1.50) 3.
burst or rupture due to internal forces
- b. Zr-H2O reaction
- c. crrosion d.
corrosion E3 at
.5 each3
- c. celting REFERENCE LARGE PWR REACTOR CORE CONTROL, PAGES 1-22 THRU 25 ANSWER 1.13 (1.50)
Packs below the centerline initially E.253 due to the HTC and lower core inlot t emperatur e E.25]. Moves upward over core life E.25] due to lower coro fuel depletion E.253 but tends to flatten out E.25] because MTC increases its affect E.25]
REFERENCE LARGE PWR REACTOR CORE CONTROL 8-19 THRU 22
s
__________-DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 2.
-PLANT PAGE 20 ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T ANSWER 2.01 (1.50)
- o. Flywheels are used for storing energy if power is lost momentarily.
- b. Power cabinet urgent failure.
- c. The reactor should trip.
[.5 each3 REFERENCE LER 483/85-11 AND LESSON PLAN PAGES 25-27
. ANSWER 2.02 (1.50)
- o. heater drain tank
- b. condenser
- c. prevent dama3e due to moisture carryover to LP turbines REFERENCE LP FW HTR EXT PAGE 12 AND LER 483/85-39
-ANSWER 2.03 (2.00) c.
- 1. CS in normal 2.
MG output breaker open 3.
synchro check relay satisfied E.5 each]
b.
The buses will dead bus transfer to the statrup transformer a few seconds later.
[.53 SERVICE ELECTRICAL LESSON PLAN PAGES 2-7 7
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 21 ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T ANSWER 2.04 (1.50)
- c. 5 minutes E.253 vs. 5 seconds for FWIS'E.253 b.
- 1. SI
- 2. HI-HI SG LEVEL
- 3. LO-LO SG LEVEL 4.
LOW TAVE E.25 each]
REFERENCE MFW LESSON PLAN PAGES 25-27 ANSWER 2.05 (2.00)
Tha boiler preheats the feedwater during plant startup E13. Without prcheating feed water will undergo considerable expansion in the SG as it hacts up resulting in level swell that is uncontrollable [1].
REFERENCE LER/85-012 ANSWER 2.06 (2.00)
- c. All flow was diverted through the opened valve to the RWST.[1.03
- c. The seal was damaged.
CO.53 REFERENCE LER 483/84-16 AND RHR LESSON PLAN ANSWER 2.07 (2.50) c.
1.
open to RWST 2.
open for full flow
- 3. CCW isolated 4.
crossties open
- 5. cold les returns open E.3 each3 b.
1.
6
~
,2.-
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22 CNSWERS -- CALLAWAY
-85/12/17-BURDICK, T N PLAN PAGES11-13j OW-M-) J' d* P N O LES ANSWER 2.08 (1.50)
- c. boric acid transfer pump
- b. RHR pump c.
CCP E.5 each]
REFERENCE LESSON-PLANS FOR CVCS, RHR AND SI ANSWER 2.09 (1.50) o.
- 1. Hi temp from the LDHX E.53
- 2. Hi temp from BTRS RHHX E.53 b'.
Hydrazine in the RCS E.53 gp j,/d REFERENCE CVCS LESSON PLAN PAGES 7-8 ANSWER 2.10 (1.50)
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T ANSWER 2.11 (3.00)
-Th3 faulted startup transformer resulted in a loss of power to 1A RHR pu p as well as the plant trip. E1.03 Since the diesel. takes ten seconds to ready for loading and the RHR pump 00quences on ten seconds after the DG output breaker closes the 1A RHR pu p start is delayed. E1.03 1B RHR started immediately on SI actuation since it is powered from cnother source.
[1.03 REFERENCE ESF POWER LESSON PLAN PAGES 3-1 THRU 17 CNSWER 2.12 (2.00)
- c. Because the RCS is subject to failure at much lower pressures when the temperature is reduced.
[1.03
- b. A TT E.253 without RT from 100% E.253 with no relief protection E.253 or SD actuation. [.253 REFERENCE RCS LESSON PLAN ANSWER 2.13 (2.50) a.
- 1. BAST to BATP then to blender then to inlet / outlet of VCT.
2.
BAST to BATP to charging pump suction via HN-8104.
- 3. BAST to BATP to FCV 110A to manual valve V-177.
E2 at.5 each3
- b. The bypass allows re'duced flow when the isolation is shut E.53 It is used in Mode 5.
E.53 To restrict flow and prevent a dilution accident. E.53 REFERENCE CVCS LESSON PLAN AND NOTE 14 DWG 22BG05 9
O
\\
,3.
INSTRUMENTS AND CONTROLS PAGE 24
' ANSWERS -- CALLAWAY.
-85/12/17-BURDICK, T ANSWER 3.01 (3.00)
.c.
- 1. CRT-display of cumulative violation time E.753
~
R I/t D _L-G s.o-- >l g3
- 2. Alarm printer output E.753 b.
- 1. CRT display changes color E.753
- 2. Audible alarm window C.753 REFERENCE LER 483/85-037 ANSWER 3.02 (2.00)
- c. pressure setpoint control E.53 b.
adjust very slowly E.53
- c. rapid pressure adjustment will cause safety injection E.53 due to rate sensitive sensors E.53
. REFERENCE LER 483/85-009 ANSWER 3.03 (2.00)
Tha channel being tested was the controlling channel E.53. Changins the chcnnels output signal during the test affected the associated steam flow chcnnel E.53, A flow error was indicated by the level control system C.53.
'The resultant adjustment to feed flow caused a low SG level E.53 REFERENCE LER 483/85-031
n 3.
INSTRUMENTS AND CONTROLS PAGE 25 ANSWERS -- CALLAWAY
-85/12/17-BURDICI(, T ANSWER 3.04 (2.50)
These valves receive an input from the auctioneered high PRNI E.53.
As the last channel was adjusted down the valve control shut the valves
[.53
- b. c high SG 1evel E.53
- c. The reactor trip with a turbine trip does not occur at lower power.E.53 REFERENCE INCIDENT REPORT 84-814 ANSWER 3.05 (1.00)
- c. LOCA sequencer
- d. chutdown sequencer E.25 each3 REFERENCE CCW LESSON PLAN PAGES 9-10 l
)
ANSWER 3.06 (1.50)
Octor driven:
a.
loss of both MFW pumps c.
LOCA sequencer d.
shutdown sequencer f
turbine driven:
- b. undervoltase on NB01 or NB02 E.25 each3 REFERENCE AFW LESSON PLAN.PAGES 5-7
v P
4 3.
INSTRUMENTS AND CONTROLS PAGE 26 ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T CNSWER 3.07 (1.50) b.
- c. C-16 (low temp return to power) d.
loss of CW pump >.MHt 757. M
- o. high stator witer temperature f.
Iow stator wi r pressure
~b REFERENCE
~
RPS LESSON PLAP PAGES 9-10 D EH LESSON PLAN PAGES 5-27 ANSWER 3.09 (1.50)
- c. disconnect switches for affected bank be aute manual switch at P-A converter
- c. Bank Selector Switch
- d. In-Hold-Out switch
- o. sroup step counters for affected bank f.
alarm reset for urgent failure E.25 each3 REFERENCE CRD LESSON PLAN PAGE 25 AND OTO-SF-00004 PAGES 3-4 ANSWER
' 3.09 (2.00)
- c. 1. wide ranse hot les temperature E.53
- 2. reactor vessel differential pressure E.53
- b. seal table E.53 which is same elevation as vessel flange E.53 REFERENCE RCS INST LESSON PLAN PAGE 11 AND RV LESSON PLAN PAGE 22 ANSWER 3.10 (2.00)
- c. particulate rad monitor
- b. saseous rad monitor
- c. sump level d.
air cooler drain flow E.5 each]
l
't
3.
INSTRUMENTS AND CONTROLS PAGE 27 ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T REFERENCE TS 3.4.6.1 ANSWER 3.11 (2.00)
- c. 1970 psis C.53
- b. P-11 status light illuminates C.53
- c. Place DOTH A and B train switches for low pressurizer pressure SI to BLOCK [.53 and BOTH switches for low SG pressure SI to BLOCK C.53.
REFERENCE RIS LESSOM PLAN PAGE 22; SI LESSON PLAN PAGE 7; OTG-ZZ-000 6 P r 11
. ANSWER 3.12 (2.00)
,[
7 45 rg[),)
O.
To allow removal of a failed channe from service. E1.03 b.'IR block resets below P-10 at 10% E.53 SR block resets below P-6 at 5x10E-11 E.53 REFERENCE NIS LESSON PLAN PAGE 29 AND OTG-ZZ-00005 PAGES 4 AND 8 ANSWER 3.13 (2.00) c.
1.
liquid radwaste discharge 2.
SG blowdown
- 3. turbine buildins drains 14. secondary liquid waste
- 5. containment purse 6.
radwaste building vent
[.25 each]
- b. To prevent condensation in the sample lines E.253 which would remove Tritius, from the process sample flow stream. E.253 REFERENCE TS PAGES 3/4.3-63 TO 75 AND OTA-RL-RK061 WINDOW 61F (TCN 85-407) l l
4.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 28
~~~~ I656L66I6AL 66NTR6L
~
~~~~~~~~~~~~~~~~~~~~~~~~
R ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T C.NSWER 4.01 (1.50)
C.
TS review and SRO approval E.53
- b. Establishment of fire watches within one hour E.53 c.
Watch tours must be conducted at least hourly E.53 REFERENCE
-TS 3.3.3 7.bf 0DP-ZZ-00001, 3.3.3 PAGE 2 ANSWER 4.02 (3.00)
O.
pressurizer level E.53
- b. spent fuel in the RCC change fixture [.53 c.
- 1. Evacuate unnecessary fuel handlers from containment.
- 2. Verify / increase air pressure in seal.
- 3. Transfer any fuel assembly in upender or transfer to SFP side.
4.
Close transfer tube isolation valve.
- 5. Transfer any. fuel in RP to the vessel.
6.
Evacuate the remaining fuel handlers.
- 7. Notify the CR of refueling pool status.
8.
Sound evacuation alarm.
[8 at.25 each3 REFERENCE OTO-KE-00001 ANSWER 4.03 (1.50)
O.
2 minutes E.53
- b. The RCP'S must be tripped if CCW is lost for 2 minutes or more and the reactor must be tripped first.
E1.03 REFERENCE OTO-EG-00001 l
l
4 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29
~~~~ A5i5E55iEAL E5RiRUL'~~~~~~~~~~~~~~~~~~~~~~~
R ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T ANSWER 4.04 (2.50) go t, {2f)
- a. To trip RCP's
- b. deenersi=e PORV s
- c. verify loss of offsite power and energize NB02 from DG d.
amp meters on breaker cubicles I
- o. it can be electrically isolated from the CR E.5 each]
REFERENCE OTO-ZZ-00001 PAGES 1 AND 75 ATTACHMENTS 2, PAGE li 3, PAGE 1 ANSWER 4.05 (1.50)
- c. blocked air dryers E.53 b.
Take appropriate action to restore parameters to NOB [.53 If they cannot be controlled or approach a trip setpoint then manually trip the reactor. E.53 REFERENCE OTO-KA-00001 PAGE 1-2 ANSWER 4.06 (1.50) i
- o. E-0, Reactor Trip or SI E.253 and FR-S.1, Response to Nuclear Power Generation E.253 eft (UT
- b. Deenergize the load centers PG19 E.53 and 20 E.53 REFERENCE E-0 AND FR-S.1 6
4.
PROCEDURES
' NORMAL, ABNORMAL, EMERGENCY AND PAGE 30
~~~~
RA656LUU5EAL E5nTR L'~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T ANSWER 4.07 (2.00)
- c. RPI rod deviation annunciator E80C3
- b. RPI deviation or PR tilt E7903
- c. PR lower detector flux deviad. ion E78C3 d.
upper E7003
- f. One or more RPI not in agreament with associated group step counters demand height by plus or minus 12 steps.
- g. Abnormal incore thermocouple or flux map readings.
E5 at
.4 each]
REFERENCE OTO-SF-00004 ANSWER 4.08 (2.00)
- c. 41 seal leakoff flow hfin b.
41 seal leakoff temperature increasing
- c. CCW thermal barrier discharge temperature increasing
- d. 41 seal leakoff flow low
- o. 41 seal DP low E4 at
.5 each3 REF ERENCE OTO-BB-00002 ANSWER 4.09 (2.25)
- c. Plant Operations controls orange tags E.25] and white are controlled by load dispatcher E.25] and power dispatcher E.25]
- b. manual valve handwheel E.253 motor valve handwheel E.253, supply breaker E.25] and control switch E.25]
- c. The original holder E.25] or the Emergency Duty Officer E.25]
REFERENCE APA-ZZ-00310 PAGES 2, 11, 18
r 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 31 RA65 L66E6hE~66NTRUE~~~~~~~~~~~~~~~~~~~~~~~~
~~~~
CNSWERS -- CALLAWAY
-85/12/17-BURDICK, T ANSWER 4.10 (1.50)
C.
Prior to each entry E.53
- b. cigns RWP sign in sheet E.53
- c. first entry each day E.53 REFERENCE-APA-ZZ-00141 PAGE 6-7 ANSWER 4.11 (1.75)
O.
Last three days or last watch whichever is shorter.
E.253
- b. 1. Review Standing and Night Orders.
- 2. Test control room annunciators.
- 3. Review annunciator defeat log.
4.
Discuss significant operations or maintenance in progress.
- 5. Perform control board walkdown.
- 6. Review incident reports.
E.25 each3 I
REFERENCE-DDP-ZZ-00003 PAGES 3 AND 4 AND ATTACHMENT ANSWER 4.22 (1 00)
- o. Indirect indication depicting actual status E.53
- b. When the concept of ALARA would be violated E.53
- c. Post work functional test performed proves all equipment is correctly aligned. E.53
- d. Outage related work, system checklist completed prior to req'mt for cperability. [.53 REFERENCE APA-ZZ- 00310 PAGES 16-17 L
l l
I
I p.
l 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32 I' ~~~~E5656L66 EdE~E6NTRUL
~~~~~~~~~~~~~~~~~~~~~~~~
j ANSWERS -- CALLAWAY
-85/12/17-BURDICK, T l
i i
CNSWER 4.13 (2.00) 1614 ON 12/10/85 l
1 REFERENCE TS 3/4.2.1 ANSWER 4.14 (1.00)
To dampen the resultant Xenon oscillation.
REFERENCE DTG-ZZ-00004 PAGE 2
\\
.. - _ _ ~ _. _
I
~
thdrawn positten and becomes mere negative as the rods are p erted.
Figure FND-RF-47 shows graphs of integral red worth responding to the differential red werth curve in Figure
-44, traph A has the reference at the bottom of the core and B 'is drawn with the reference at the top of the cere.
In 1
ther case, the reactivity change resulting from any red motion is:
1:
I as=
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S. NUCLEAR NEGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
CALLAWAY REACTOR TYPE:
PWR-WEC4 DATE ADMINISTERED: 85/12/17 EXAMINER:
- HIGGINS, R.
APPLICANT:
INSTRUCTIONS TO APPLICANT:
Use separate paper for the answers.
Write answers on one side only.
.Stople question sheet on top of the answer sheets.
Points for each quantion are indicated in parentheses after the question. The passing
.Stade requires at least 70% in each category and a final grade of at loast 80%.
Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY VALUE
_____- _ _-TOTAL SCORE VALUE CATEGORY
_'_'5 00.I.____25 00
________ 5.
THEORY OF NUCLEAR POWER PLANT
__I__
OPERATION, FLUIDS, AND THERMODYNAMICS
___I.00___ ___I__0
________ 6.
PLANT SYSTEMS DESIGN, CONTROL, 25 25 0 AND INSTRUMENTATION
___I_00--- ___I_00
..._____ 7.
PROCEDURES - NORMAL, ABNORMAL, 25 25 EMERGENCY AND RADIOLOGICAL CONTROL I
__ _ _Ibb
_______. 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100.00 100.00 TOTALS FINAL GRADE _________________%
'All work done on this examination is my own. I have neither given nor received aid.
y A P P L i ~C A U ~T S S i ~G U ~A i 'U R E
~~
~
I~~~
~~~~~~~~~~~~~~~~
i 8
k 1
E
r t,
'.O 5.
THEORY OF NUCLEAR POWER PLANT.0PERATION, FLUIDS, AND PAGE 2
1
____7gggggg7gggggg______________________________________
i QUESTION 5.01 (1.50)
~
- 9 Assumenthat RCS temperature is 199 F at BOL, boron concentration is 1000 ppo, and shutdown margin is the minimum required by Technical Specifica-tions under these conditions.
What is the minimum number of sallons of 4%
boric acid which must be added in order to raise RCS temperature above 200 F?
Use Figures 5-la and 5-ib.
Show your work.
QUESTION 5.02 (1 00)
How is adequate shutdown margin verified during a reactor startup prior to criticality?
00ESTION 5.03 (1.00)
Why does the critical boron concentration decrease at a much more rapid rote at 9000 MWD /MTU than it does at 1000 MWD /MTU?
Refer to Figure 5-3.
QUESTION 5.04 (2.00) i
- a. What are the three bases for establishin3 control rod INSERTION (1.5) limits?
- b. What is the reason for establishing control rod WITHDRAWAL limits? (.5)
GUESTION 5.05 (1.00)
Tho reference text LARGE PWR CORE CONTROL states.that a xenon instability (occillation) is not a nuclear hazard.
What problem does a xenon uncillation pose?
QUESTION 5.06 (1.00)
What is the basis for establishins limtts on axial flux difference?
I
(*****
CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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i (mdd/ mod) 837/07 *ytaoM uoJos (953UBJ8)})O a'
I.
4 1
~
FIGURE 5-lb RUCTOR MAKEUP CONTROL SYSTEM NOM 0 GRAPHS _
F_ I BORON ADDITION 9000 -
g
~ 700 3
f7000-Ct 400 M
I I
Vs
- 3.33 I8 (7000-C s 6000 -
i 500 4500 -
40 2500 -
- =
300 E 3000 -
250 a
2400 o
M ~-
0 o
1800 -
1505 20M -
- g. 1500 -
W m
5 1200 -
2 m
C 3
100 a
t, 90 $
g 900 -
80
_o 750 -
3
-~~
E *l500 -
g 70 g 5
d 600 ~
60
- E a
50 I*
i_
450 _
WO M
=
350
~
E. 1000 -
300 go i
e 250 25 200 -
20 i
500 -
/, h _h l/Xll3
~
Da:e
~ SuperintenTent, Engineering Refer to Table 7-1 for correction factors.
NOTE:
k
~
1400 l I
l i
I 1208 ppm 1200 Burnuh C
i (MWD /MTU)
(p,sm)
No Xenon 8
0 1208 S
150 911 n
x=
1 1000 -
1000 895
,=.
O 2000 893
,g
[p 4000 825.
e gr-3 6000 716
-g 8000 591 U
800 pg 10000 450
- 8 cz E
12000 295
,M ncg c
14000 137 l y ggg O
u 15150 46 ggg w
,,p e
600 i
c l
3H o
,x, y
m C
O 1-
.i. i.
%-);
g 3
400l 1
. ['.
a n vi to o-a 5
g u
enog x,
I 200
- 4/1 3
. 6
- L.
0 0
3000 6000 9000
- 12000, 15000 18000 i
M h
w
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3
QUESTION 5.07.
(2.00)
Answer the following True/ False questions *
.a.
The equilibrium xenon reactivity at 100% power is more than twice the equilibrium xenon reactivity at 50% power.
(.5)
-b.
Thes peak xenon reactivity insertion (addition) followins a trip from 100% equilibrium power is more than twice the peak xenon reactivity insertion (addition) followins a trip from 50% equilibrium power.
(.5)
- c. The equilibrium samarium reactivity at 100% power is the same as the equilibrium samarium reactivity at 50% power.
(.5) d.
The samerium reactivity added followins a trip rrum 100% equilibrium power is the same as the samarium reactivity added following a trip from 50% equilibrium power.
(.5)
QUESTION 5.08 (1.00)
Why is the delayed neutron importance function, I,
less than one?
QUESTION 5.09 (1.00)
Why does the effective delayed neutron fraction decrease from BOL to EOL?
00ESTION. 5.10
(.50)
If the axial flux difference is out of the band to the left (too negative),
chould the operator borate or dilute?
(Choose the correct response.)
--QUESTION 5.11 (1.00)
Which two factors cause the target axial flux difference at EOL to differ from that at BOL?
(***** CATEGORY 05 CONTINUED ON NEXT PAGE
- )
O
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4
i
~ THERMODYNAMICS QUESTION 5.12 (1.00)
Assumins all rods fully insert after a reactor trip from 100% power, approximately how lons will it take for power to decrease tot
- e. 5%7
(.5)
-b. the level at which the source ranse detectors will reenserize?
(.5)
QUESTION 5.13 (1.00)
For a given boron concentration, why is MTC more negative with Banks C and D inserted in the core than with Bank D alone?
Refer to Figure 5-13.
QUESTION 5.14
(.75)
What pressure, in psis, must be maintained in the steam senerators in order to obtain a 100 F subcooling margin when RCS pressure is 1500 psis?
QUESTION 5.15
(.75)
What would be the tail pipe temperature downstream of an open pressuriner
'PORV if the pressuriner pressure is 2100 psis and the PRT pressure is 25 psis?
GUESTION 5.16 (1.00)
Naae two mechanisms by which hydrogen can be introduced into the contain-osnt atmosphere following a LOCA.
QUESTION 5.17 (1.50)
(.5) rebombin'ers may be placed into service?
- b.. What is the basis for this precaution?
(1.0)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE
- )
4
~.,
n.--
v.,
FIGURE 5-13 RODDED MODERATOR TEMPERATURE COEFFICIENTS AT HZP VERSUS BORON CONCENTRATION BOL, NO XENON CYCLE 1 Os N ;-..
Y NYD Superintendeut. Engineering Date
=.....
7 7
l 1
1 L
L 1
1.
L L
1 1
L..
E 1
i L
E 1
1 p
L k
1 1
L L
I
<...mm imapq
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i'manunana>
<,.e,..
.mm><>
i 1
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(
. ; g..
. g...
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em..
. r, au.mana.
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6
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p ad.>E.
M
==p.
s pq. O ma-suuma
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... s c........q.
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Y Y
9 9
o~
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i N
l
..- m
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.=
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5
QUESTION 5.18 (1.50)
With the plant at 100% power, how much higher above its nominal 100% power value must Tave be raised before the steam generator safety valve with the lowest setpoint opens?
Nominal 100% power steam generator pressure is 1000 psis.
Show your work.
QUESTION 5.19 (2.50)
For,the following questions refer to Figure 5-19.
- o. What is the basis for curve A?
(.5)
.b..
Why.does curve A have.a r.e g a t i v a slopa (decrease with increasing
(.5) 3
. power)?
c.
On Figure-5-19, draw the curve representing the OTAT setpoint, (1.0) using the.following equation:
OTAT setpoint =JVr [1.10 -.0137/ F (T - T')3 o
AT.
indicated AT at rated thermal power
=
T'
= nominal Tave at rated thermal power d.1What design feature prevents Tave from exceeding the reactor core
(.5) safety limit when reactor power is less than 30%?
QUESTION 5.20 (1.00)
Why is the initial reference transition nil ductility temperature less than the reference transition nil ductility temperature after 7 EFPY?
- GUESTION' 5.21 (1.00)
Why-is a centrifugal pump normally started-with its discharge valve shut?
T
(***** END OF CATEGORY 05
- )
REvI8/CN
' FIGURE 5-19 y
g 800
-s
_... =,.
l
. iW L._
.-~
l
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UNACCEPTABLE l
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=
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)
EOUATIONS REACTOR THEORY
, RADIATION FI.UIDS/TilERHO/IIEAT TRANSFER
~
r
..*ec P = P,e /t = P,10 N=Ne m = Alp 1V1 = A2p2V2 t
SUR*t
-At
+
A = AN 0~D
=0~0 Q = A1V1 = A2V2 r=-8'+
or t p
Ap Ap
-x/
g,g,-px = I 10 E,=E
+0 g
cut stored k-l k
-k 2
i p=
- A k
kak R
ATis = 0.693 E = KE + PE + U + pV + Q + W R/hr @ d feet =
d I Z 2
2 Point source E
cp I1-k 1d1 = 12d2 2
1 1d2 - line source
" g" g
lidt
=
2
-=1-k M
R/hr x time - R reduced for - turbine. SG pump. nozzle.
. cps
- Rad x QF = Rem rifice. C ndenser. Pipe. Rx H
- cps, flow a /dp Tig Bio Rad
=
egg
- " 'L ya O
"0# doppler + Dmod + Dvoid Tbgg, + Us nst Rad D 2g c head loss a op
+DXe
- PSm
- DPu +
P"h + Pambient
=k F = pA E
D Orod + Pfuel +
c Boron AP2p% " AP1 phase x K O
"' "" I I Poisons Pump laws speed a flow ka - kt + 5k (speed)2a pressure (speed)3a power 6k=k-1 MATil Q = kAaT = hAAT = UAaT SUR =
y" = b 0 " * *p 0
t Q = mah E
E4V y
Q = cot
"'8 C d p, 3.1 x 1010 log x = c lug x all = m c. AT P
E = No E
E*~
EY
""v log xy = log x + log y 11 - U + pV AS = S Defect = Coeff x A Parameter T
pV = nRT P1VI,M T1 T2 C Vi + C2V2 - Ci(Vi + V2)
I
=
~
t Os Table 1.
Saturated Steam: Temperature Table y
ACs Press 50ccific Wolume Estnalpy feme it per
$41
$8t.
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$41.
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bead hap Wapor bead Evap Vapor bead Evep % spor fahr Temp t
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364 0 160 903 0 01816 2 83C2 2 8144 336 5 8586 tilS 2 0 5212 0424 1 % 37 3H e 368 169113 0 01421 2H91 2 687?
3408 4%I 3195 9 0 $263 0337 1 5596 368 8 872 0 177648 0 018?6 2 54$1 2MM 345 0 851 6 1196 7 0 $314 3 0243 8$%4 3 72 I 376 8 106 517 0 01831 2 4279 2 4462 349 3 848 8 1197 4 0 5365 18144 14513 378 8 300 0 195 729 0 01836 2 3170 2 3353 353 6 844 5 1998 0 C S416 1 6057 1 5473 set s 3e4 0 205294 0 01842 2 2120 22304 3579 840 8 1198 7 0 5466 0 99o6 I S432 Sea t 3es t 215 220 0 01447 2 11?6 2 1311 362 2 8372 8199 3 0 5S16 0 9876 1 5392 3e8 8 4,..
3t2 8 225 516 0 01853 2 0184 2 0369 366 5 833 4 1199 9 05%7 0 9786 I S352 312 8 396 0 236193 0 01858 1 9291 1 9477 370 8 829 7 1200 4 0 % 17 0 96 % 1 6333 196 8 40s t 247 259 0 01864 3 6444 3 8630 375 1 825 9 1701 0 0 % 67 0 9607 1 5274 est s
..._ 7
- ~
au g 258 72%
0 01870 1 7640 1 7:27 3798 822 0 1201 5 0 5717 0 9S14 1 5234 sea t 400 0 270600 0 01876
,6877 17%4 383 8 818 2 1201 9 0 5766 0 9429 I S19S 488 8
"- 412 e 23; 894 0 01881
,6152 - I 6340 388 1 81a 2 17024 0 $816 09341 1 $157 812 8 ell e 293 617 0 01447
. 6463 1 %$1 392 5 5102 1202 8 0 $8H 09253 1.5118 til t
- 42..
m 780 00iHe ime i.,97 m9 m2 am O Sm Oml i=0 47 0 424 0 322 391 0 01930 t alk 1 4374 401 3 8022 l203 5 0 $9ds 0 9077 1 5042 424 8 476 0 336 a63 0 01906 1 3S91 1 3782 405 7 798 0 203 7 06N4 08990 I S06R 479 8 432 8 351 00 0 01913 1307H 1 32179 4101 7939 l204 0 0 6063 0 8903 149H 432 8 eas t 3H Q3 0 01919 IJ4887 1 16836 414 6 7897 l204 2 0 6112 0 8816 149f 8 436 8 eas t 381 54 0 01926 1 89761 1 21687 419 0 7854 1704 4 0 6161 0 8729. 4493 448 8 444 9 397 %
0 01933 1 14874 1 16836 4235 781 1 120s 6 0 6210 0 H43 ' 4853 444 0 eas t 414 09 0 01940 1 10212 1 12162 428 0 776 7 1204 7 0 6259 08SS7
.4815 eas t 412 8 43114 0 01947 10$7W 1 07711 432 S 772 3 1204 8 0 6308 0 s471 1 4778 4528 496 8 448 73 0 01954 1 01518 303472 4370 767J 1204 8 0 63 % 0 8385 1 4741 aus e
- i. n3 !,-Q.
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E Table 1. Saturated Steam: Temperature Table-Cor72/mted ADS Preis 5pecilec volume intnalpy talropy Teme 16 pee sat
Sat 5 81 Sat Sat set Temo fahr 5g in it0uid
[vap Vapor b0ued
[vap Vapor bquid [vap Vjper f ahr t
p e,
wir 88 he b eg h,
8, 8,,
.8, t
484 8 eM 87 0 01961 0 97463 0 99424 441 5 763 2 1204 8 0 6405 0 8299 1 4704 ass e 2
464 8 485 %
0 01 % 9 0 93588 0 95557 4461 758 6 1704 7 0 W$4 0 8213 Iauf een a 460 0 504 83 0 01976 0 89885 0 91862 4507 754 0 1204 6 0 6S02 08 27 1 4629 ass e
- -, 472 8 S24 67 0 01984 0 86345 0 88329 455 2 7493 1204 4
&&%I 0 8042 1 4392 472 8 t
olla
$4511 0 41992 0129S8 0J4960 4S9 t 744 3 1204 3 0 6S99 0 79 % 1 4565 476 0 a.
ass e S66 15 0 02000 0 79716 0 81717 464 5 739 6.3204 I O ua8 0 7871. I 4518,,,, 404 l.,,.
4D4 9 58781 0 02009' 0 7H13 0 78622 4693 734 7 1703 8 OM% 07785 1 4481 eed s eas t 610 10 0 02017 07M47 0 7MSt 4738 779 7 1203 5 0 6745 0 7700 1 4444 das s es2 3 61303 0 02026 0 70794 0 ??820 478 5 724 6 12031 0 6793 0 7614 1 4407 es!8 496 8 6M 61 0 02034 0 68065 0 70100
&&32 719 3 12C2 7 0 6842 0 7528 1 4370 at4 I e% '.r Ise t 680 86 0 02043 06S44a 0 67492 4879 714 3 1202 2 0 6890 0 7443 1 4333 taa s tes t 70578 0 02053 0 62938 0 64991 4927 709 0 1701 7 0 6939 0 7357 1 4296 tes t tes t 73340 0 02062 0 60530 0 62592 4975 7037 1703 8 0 6987 07271 3 4258 0e8 0 517 8 757 72 0 02C72 0 58218 0 602t9 SC2 3 6962 1200 5 0 70M 0 7185 1 4221 812 8 lisJ FM 76 0 02081 0 5S997 058079 50? !
692 7 litit 0 7085 0 7099 14183 8163 528 9 812 53 0 02091 0 53864 0 559 %
S12 0 687 0 1199 0 0 7133 J 7013 I 4146 179 8 874 8 841 04 0 02102 0 $1814 0 53916 516 9 681 3 1198 2 0 7182 0 6926 1 4108 124 8
- --m%9 628 0 870 31 0 02112 0 49eal O Sf9%
S21 8 676 S 18973 0 7231 0 6&39 1 4070 878 0 532 9 900 34 0 02123 041947 0 $0070 526 8 M96 Il96 4 0 7280. 0 6752 3 4032 132 e 138 0 93117 0 02134 0 46123 0 46257 S31 7 M36 1195 4 0 7329 06%$ IJM3 Salt Het 962 79 0 02146 0 44367 0 46513 S36 8 657 5 1194 3 0 7378 06577 _ 13954 Set e
,, f ',.-
844 8 995 22 00?)$7 0 42677 0 448 M Sel 8 651 3 1193 1 0 7427 0 64 9 IJ115 ka8.
WIS 1028 49 0 02165 0 41048 0 41217 S46 9 645 0 1191 9 0 7476 0 6400 IJ876 W18 152 8 10t2 59 0 021!! O 34474 0 4 twi
$5?O
&3n %
11 % 6 0 7525 il 6311 1 3837 H23 M60 1097 55 0 C2164 0379% 0 40160 5572 632 0 liti2 0 7575 0 6222 13797
- 986 8 368 8 1133 38 0 02207 0 36507 0 '8714 562 4 6?53 1187 7 0 76M 0 6132 1 3757 M4o M48 3170 10 0 02221 0 3SOM 037320
%76 618 5 1886 1 0 7674 0 6041 IJ716 SM S Me8 1207 72 O C22 M C 3374l O M9 ?$
S72 9 611 5 lik 5 0 7725 0 5950 13675 868 8 872 8 1246 26 0 02249 0 32429 0 34678 578 3 604 5 1182 7 0 7775 0 $859 3 36M 3728 576 8 1285 74 0 02264 0 31162 0 33426 S&37 6972 !!IO S 0 7825 0 57 % 3 3592 8788 M8' 1326 17 0 02279 0 29937 03 216 St91 589 9 : 179 0 0 7876 0 % 73 135$0 Set t
'? 4 3367 7 0 02295 0 28753 0 31045 H46 582 4
' 176 9 0 7927 0 5580 1 3507 Mas set t 1410 0 0 02311 027608 0 29919 600 1 574 7 174 8 0 7978 0 5485 1 3464 saa 8 5s78 1453 3 0 02328 0 26499 026527 605 7 M60 172 6 0 8030 0 5390 1 34 ?0 M23 9e6 8 14978 0 02345 0254M 027770 6.14 558 8 1870 2 0 8082 0 S293 13376 tes t 808 8 15432 0 02364 0 24384 026747 ~
617 1 550 6
- 167 7 0 8134 0 5196 3 3330 800 t Sei t 15F9 7 O C2342 0 23374 0 25757 622 9 S42 2 il651 0 8187 0 5057 13284 te4 8 608 8 437 3 0 C2402 0 22394 0 24796 628 0 533 6 11624 0 8740 04s97 1 3238 est t 812 8 6%I C C2422 0 21442 0 23865 634 8 524 7 11$9 5 0 82M 04896 1 3190 812 e L-818 6 1359 0 02444 020516 022960 640 8 S15 6 11%4 0 U48 0 4794 1J141 616J i
873 8 17869 0 024 % 01961$ 022081 646 9 506 3 1153 2 O NO3 04689 3J092 828 9 874 8 18390 O C2494 0 18737 021226 653 1 4*66 1149 8 0 8450 04583 13041 824 8 878 8 19924 0 0M14 0 17880 02:!s:
659 5 4867 1146 1 0 8514 0 4474 12988 828 8 632 0 19470 0 0?S39 017b44 0 19562 HS9 4764 1142 2 0 8571 04364 IJ934 832 1 638 8 2002 8 0 02 % 6 0 16226 0 18792 6724 465 7 1838 1 0 8628 0 4261 IJ879 836 B M88 2059 9 0 0?S95 0 15427 0 18021 6'l 1 454 6 1133 7 0 SE86 0 4134 3 2821 ke8 844 0 21183 0 0?625 0 14644 0 17269 68$9 4431 1129 0 0 8746 04015 12761 844 0 848 8 2178 )
D C2657 013876 O l6H4 692 9 431 1 3124 0 0 8836 03393 12695 848 8 657 8 22392 0 02691 0 13124 0 15416 700 0 4187 1138 7 0 8868 03767 12634 552 0 Gle t 23017 0 02728 0 12387 0 15115 7074 4057 1113 1 0 8931 0J637 IJM7 Ob48
~~
860 8 2N57 0 02768 0 11663 0 14431 714 9 3921 1107 0 0 9995 03502 12498 Ste s 844 8 2431 1 0 02811 010947 0 13757 722 9 3777 110c 6 0 9064 0 3361 12425 084 8 648 8 2498 )
0 028S8 01C229 0 13087 731 5 3621 1093 5 0 9137 0J210 1 2347 054 0 872 8 2%66 0 02911 0 09S14 0 12424 7402 MS7 1081 9 0 9212 03054 IJ266 472 0 475 8 2636 8 042970 0 08799 0 11769 7492 328 5 1077.6 0 9287 02892 12179 lis t Sal t 27086 0 03C37 0 08080 0 11117 758 5 3101 1068 5 0 9365 02720 12086 tes t 884 8 2782) 0 03114 0 07349 0 10463 768 2 2902 10584 0 M47 02537 1 1984 esa e 688 0 28574 0 03204 0 06595 0 09799 778 8 268 2 1047 0 0 9535 02337 4 1872 Set 8 892 8 2934 5 0 03313 0 05'97 0 09110 790 5 243 1 1033 6 0%M 02130 1 8744 est e 006 8 30134 0034% 0 04916 0 08371 004 4 212 8 10172 0 9749 01M1 11$91 086 8 M..
188 0 3094 3 0 0 % 62 0 03857 0 0?$19 822 4 172 7 995 2 0 9931 01490 1 1390 fee t 782 8 3135 5 0 03824 0 03173 0 06*S7 135 0 844 7 979 7 3 0006 0 1246 1 1252 182 8 764 8 3177 2 0 04108 0 02112 0 06330 854 2 1C20 9M 2 1 0169 0 0876 1 1046 led e let t 3198 3 0 04427 0 01304 0 05730 873 0 614 Suo 1 0329 0 0527 1 00 %
195 8 7th 47' 32062 005C78 0 00000 00$078 906 0 00 906 0 1.0612 0 0000 1 0612 385 47'
.~
g.
- Cnt cal temperature
- 9. :.. -
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a.,.w.-:.=.. =...s. - n. w.,.,...
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h Table 2: 'S'aturated Steam: Pressure Table C
4 Spetsf4 Volume Lathalpy
[ntropy M
-Abs hess Temp Sat
$at
$41.
$at
$41 Sat Abt htSs
]tbl$q In.
f8hi bound Ivap Vapor 1 squid (vsp Vapor Lmuid L.ap Vapor 7 lb/$0 in p
I og t,g o
hg h,g li 8,
8,g 8,,
p g
g 88ea61 32 018 0 016027 3302 4 33024 0 0003 10755 l075 5 0 0000 21872 21872 08s8u
- On S9 323
. 0 014032 12n S 12355 2738?
10608
.0874 4c542 2 04M 2 0967
-4 76--
M OM 19 586 0 016071 H15
-His 47623 10486
(**6 3 0 09?S 1 9446 2 0370 8 58 m.
88 101 74 0 016136 333 $9 333 60 6973 1036 I
.1068 0 1326 3 8455 9781 I8 90 16224 0 016407 73516 73 632 13020 10009 1131 1 02349 1 6094 M43 88 18 8 19321 0 016S$?
38404 38420 16126 982 1 1143 3 0 2836 1 5043 7879 10 0 te n$
21200 0 016719 M 782 M 799 18037 970 }
1150 5 0 3121 1 4447 7%t 14 6 %
18 3 213.03 00lH26 26274 M190 18123 969 s 1150 9 0 3137 1 4415 1 7552 ut 20 0 227 96 0 016834 20 070 20 087 19627 E0I l156 3 0 3358 13%2 3 7320 38 8 as t 250 H 0 017009 13 7266 13 7436 218 9 945 2 19 1 0 3682 13313 1 69 %
3el 48 0 M725 0 017151 10 4794 104965 236 1 933 6 169 8 0 3921 11844 1 6765 48 8 M8 281 02 0 017274 84967 8 5140 2$02 923 9 lI741 0 4112 12474 165tt H8 40 0 292 /1 0 017383 7 1662 717M 262 2. 915 4
- 1776 0 4273 12167-1 6440 es t
- M 18 8 30293 0 017482 61876 6 2050 2727 9078 3806 0 4411 11906 1 6316 78 8 m,
v s6mtsa144G 00 0 312 04 f 037573 54S% * $ 4711 282 1 900 9 L1831 0 4534 1 1675 1 6206 Sal es 8 320 28 0 0176$9 4 8729 4 8953 29C 7 094 6
- 1185 3 0 4643 1 1470 1 6113 ts 8 fee e 32782 0 017740 44133 4 4310 298 5 888 6 11872 0 4743 I 1244 I 6027 tot 8
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III 8 33479 0 01782 4 C306 4044
- 3058 883 1 1188 9 0atM 1 1115 1 S450 til8 128 8 34127 0 01789 3 7097 3 7275 312 6 8778 1190 4 0 4919 1 0960 13879 1288 7 1 l 130 8 34733 0 017 %
3 4364 3 454a 319 0 872 8 1191 7 0 4998 1 0815
. 1 5813 tale 1st 9 353 04 0 01803 32010 3 2130 325 0 E60 liv 3 p 0 3071 1 0681
- I 5752 - 148 8.
H8 3S843 0 01809 2 9958 3 0139 330 6 8634 IIM I 0 5141 10%4 IM%-
tH e 168 8 363 %
0 01816 2 8155 k t336 336 1.
8%0 1195 1 0 5206 104M 3 %4!
188 0 178 8
%542 0 01821 2 65 %
2 6738 341 2 854 8 11 % 0 052H 1 0322 I S$91 lis t ISI e 373 08 0 01827 2 5129 2 5312 346 2 SSC 7 11 % 9 0 5328 3 0215 1 $$43 188 0 IM S 37753 0 01833 2 3847 2 4030 350 9 846 7 11976 0 5384 1.0113 1.5498 199 8 a
231 8 381 82 0 01839 2 2689 2 2873 3%$
842 8 1198 3 0 5438 1 0016 S4Sa pees FIS O 385 91 00144 2 16373 2 18217 3%9 839 1 1199 0 0 6490 0 9923
$413 118 8 278 3 379 84 0 018SO 2 06779 2 08629 364 2 Sn e 3199 6 0 % 40 0 9834 5374 228 8 23t I M3 70 0 018 %
8 97991 I 95646
%83 831 8 1200 1 0 5588 0 9748
$336 238 0 34I t 39739 0 0186C 3 29909 1 91769 3723 828 4 1200 6 0%M 0 9665
. $299 248 1 PS8 0 400 97 0 01865 1 82452 1 84317 3761 825 0 1201 1 0 % 79 0 %e5 152M 1H 8 2ss e 40444 0 01670 11%s8 ! ?7418 379 9 821 6 1201 5 0 5722 0 950f
$230 268 8 378 8 40780 0 01875 1 69137 I 71013 383 6 818 3 120: 9 0 5764 0 94?3 S197 278 I tes t 4I107 0 01880 1 63169 3 6 % 49 3871 8151 1202 3 0 5805 09hl Slu 7s88 298 8 41415 0 01865 1 5 7597 I SW82 390 6 812 0 1202 6 0 $844 0 9291
. 5136 798 0 3sc e 417 35 0 01899 I S2384 1 54??4 394 0 800 9 17029 0 5882 0 9223 i 5105 3es e 35 3 431 73 0 01912 1 30042 1 32 %4 40'8 754 2 12o4 0 0 6059 08%9 l asts 358 egg e es4 60 0 01934 1 84162 1 n09S 424 2 780 4 1204 6 O W17 0 8630 34H7 448 8 4SII 45E 28 0 01954 1 01224 1 03179 4373 767 5 120a 8 0 6360 0 8378 l 4738 4HI Su e 46701 0 0:175 0 10787 0927C 449 6 7%)
2047 0 6490 0 8148 l 4EH Has 8 81 476 94 0 019M 0 82143 0 84177 460 9 743 3 204 3 0 6611 0 79M 14547 SH g 888 6 486 20 0 0?cl3 0 74962 0 76975 4717 732 C 203 7 0 6723 0 7736 i4461 set t
$588 49489 0 02032 0 68411 0 70643 481 9 720 9 202 8 C &828 0 7552 1 4351 stal 380 8
$03 08 0 020$0 0 63605 062 %
191 6 710 2 12018 C 6928 0 7377 1 4304 788 8 TH I S10 84 0 02069 0 $8880 0 60949 500 9 699 8 1200 7 a 7022 0 7210 1 4232 758 e tes t Sie ?!
0 02c47 O Sa8?9 0%89!
SO9 8 689 6 1994 v 7111 0 70$1 14163 08 I ISO 8 52$ 24 0 02105 0 $1167 0 53302 618 4 679 5 198 0 0 7197 0 6899 I 4096 834 4 Det 8
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0 02141 0 45064 0 47205 S3a 7 ECC 194 7 0 7358 0 6612 IJ970 958 8 1888 1 S44 S0 0 021$9 0 424X 0 44S96 542 6 650 4 192 9 0 7434 0 6476 1391C tes8 8 10S8 8
% 053 0 02177 040047 0 42224 5S01 640 9 1910 0 7507 06344 1 3851 1858 8
.; ;.:t."ithM gp 1180 8 556 28 0 0?!95 0 37863 0400SI
$$7 5 631 5 I99 1 0 757 0 6216 1 3794 1100 8 1880
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- 1674 0 8142 0 5182 1 3324 HS8 8 1800 0 M4 8' 0 02387 0 23159 0 2 % 46 624 2 5403
.lH S 0 8199 0 5076 1 3274 1688 8 18$$ 8 609 0S 0 02:07 0 22143 0 24 % I 6304 531 3 l1616 0 8254 0a973 1.32M 1858 0 17e8 s 63333 0 02428 021178 013607 636 5 522 2 158 6 0 8309 0 4867 1 3176 lise l V...
. - 1719 9 61717 0 02450 0 20?63 0 72713 642 5
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494 6 189 0 0 8470 04Mi 1 3030 18H e
~ftes t 628 %
0 07517 0 17761 0 20215 M04 48S2 18S 6 0 8$22 0 4459 1 2981 Itse s inte t 63222 0 0h41 0 16999 O lMo M63 4758 142 0 0 8574 04M8 1 29317 1951 s
- 30se g 6JS 80 0 02 % S 016?M 0 18831 6 72 )
4H 2 138 3 O M25 04?%
3 2881-78es t 5 7100 0 H2 76 0 0261$ 014stS 017$0:
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,122 2 0 8828 0 3848 1 2676-27st t 2300 0 6%99 0 02727 0 124 % 0 15133 707 2 406 0 113 2 0 8929 0 Ma0 12% 5-230s e 3 set e M211 0 02790 0 11287 0 14076 719 0 384 8 303 7 0 9031 0 34Jo i 2460 3ese 8 3980 0 E811 002t59 0 10209 0 13068 731 7 M16 1093 3 0 9139 0 3206 12345 F900 0 M88 0 673 91 0 02138 0 09172 0 12130 74s 5 3376 1082 0 0 9247 0 2977 12225 2nse s 37 30 67953 0 03029 0 08165 011I9s 2S73 312 3 1069 7 0 93 %
.0 2741 1 2097 ties t Het t 684 %
0 0 JIM 007171 01030S 770 7 286 4 30 % 8 09468 0 2491 31%8 2900 0 fees t 69022 0 03252 0 06154 00Sa2C 785 1 254 7 10398 0 9588 0 2215 1 1803 rees t Sees t 69533 0 03428 0 05073 0 08 SOC 801 0 218 4 1020 3 0 9728 01891 1 1619 sett e Blot t 70028 0 0M81 0 03771 0 0741?
824 0 169 3 993 3 0 9914 0 1460 1 1373 8188 8 37e0 0 705 08 0 04472 0 01191 00%63 8755 HI 9316 103Si 00482 1 0832 32 eel 3308 2*
70S 47 00$078 0 00000 00$078 906 0 00 906 0 1 0612 0 0000 1 0612 82s8 2*
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 6
QUESTION 6.01 (1.00)
Why do the B and C SI accumulators have two vent valves?
QUESTION 6.02 (2.00)
- c. Explain the relationship between the source range nuclear (1.0) instruments and the CVCS.
b.
What is the basis for this relationship?
(1.0)
QUESTION 6.03 (1.00)
Which power range reactor trip (s) can be blocked?
QUESTION 6.04 (2.00)
What are the two open interlocks provided for the RHR discharge valves to the charging and SI pumps, 8804 A and Br and what is the basis for each interlock?
GUESTION 6.05 (2.00) a.
What four signals will automatically start the motor driven AFW (1.0) pumps?
Include setpoints and coi.acidence.
b.
What signal (s) will cause the automatic switch of AFW pump suction (1.0) from the CST to ESW?
Include setpoints and coincidence.
QUESTION 6.06 (1.00)
Whet AFW pump problem would be caused by back leakage of main feedwater into the AFW pump suction piping?
QUESTION 6.07 (1.00)
What air system design feature is installed to limit the rate of high pressure air loss during a major air line rupture?
l (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxx**)
f
6.
AND INSTRUMENTATION PAGE 7
____ PLANT SYSTEMS DESIGN, CONTROL, QUESTION 6.08 (1.00)
- a. How will the steam senerator atmospheric relief valves fail
(.5)
(open, closed, as-is) if there is a complete loss of pneumatic pressure to the valve?
- b. What design feature is employed to reduce the likelihood of a
(.5) complete loss of pneumatic pressure to the steam senerator atmospheric relief valves?
QUESTION 6.09 (1.00)
- 0. Y
=a.
What signal (s) will cause the automatic closure of the CCW
(+ vet containment isolation valves?
Include setpoints and coincidence.
CLS -
- b.-Why are bypass valves installed around the CCW containment
( tTC$
isolation valves?
QUESTION 6.10 (1.00)
Describe how two of the four types of fire. detectors used at Callaway datect the presence of fires.
QUESTION 6.11 (1.00)
Name two of the three interlocks which must be satisfied in order to move the fuel handling system's transfer car.
QUESTION 6.~12 (1.00)
What is the source of. normal and emergency makeup to the spent fuel system?
QUESTION 6.13 (3.00) ca. Name four of the five control circuits which receive an (2.0) auctioneered high Tave signal.
- b. Describe how the auctioneered low Tave signal is used.
Include (1.0) setpoints.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
9 4
m
f a
-6.
- PLANT ~ SYSTEMS-DESIGN, CONTROL, AND INSTRUMENTATION PAGE 8
QUESTION 6.14 (1.00)
Neme two loads which'are started by the LOCA sequencer which are not started by the. LOP sequencer.
QUESTION
'6.15 (1.00)
Why;is it inadvisable to take more than one power range instrument at a tine out of service during refuelins?
' QUESTION 6.16-(1.00)
Why arez orifices. installed in-the reactor vessel head vent lines?
_" QUESTION 6.17 (1.00)
Nace four. conditions which will cause a Control-Room Ventilation Isolation.
Sstpoints'are not required.
QUESTION 6.18 (1.00)
A high radiation alarm is received on a particulate monitor.
The filter push -button'-is. depressed, advancing the filter..Immediately.after being returned to service the particulate monitor high radiation alarm actuates
'osain.
Is the alarm valid?.
Explain.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTf4 TION PAGE 9
. QUESTION 6.19 (1.00)
Why is the reactor makeup water valve V178 required to be closed when the reactor is in Mode 5?
Refer to Figure 6-19.
-b4--4'f-- W v60s vdes REACTOR To VCT I>G N,
7 MAKE-UP vn78 FCV WATER is A a
FROM FCv TO VCT-Figure 6-19 103 ouTi.ET GUESTION 6.20 (1.00)
Why are anti-reverse rotation devices installed on the reactor coolant puaps-to prevent their reverse rotation?
(xxxxx END OF CATEGORY 06
- )
.i.
['
i.
i 7',
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 10
~~~~ 5656[665CEE"66NTR6[~~~~~~~~~~~~~~~~~~~~~~~~
R
' QUESTION 7~.01 (1.00)
What'four indications are used to verify a reactor trip per E-0?
'GUESTION 7.02 (1.50)
- e. What-is ADVERSE CONTAINMENT?
(.5)
- b. Why-are RCS wide range RTDs used instead of core exit TCs to (1.0) determine RCS subcoolin3 during ADVERSE CONTAINMENT?
GUESTION 7.03 (1.00) b~
- a. Main steam lines should be isolated following an SI if steam
(.25) generator pressure drops below _____ psis,
(.25) manually restarted.
- c. Alternate water sources for the AFW pumps will be necessary if
(.25)
CST-level-decreases below
- d. Manually reir.itiate SI if pressure =er level drops below
(.25) during ADVERSE CONTAINMENT.
QUESTION 7.04 (1.50) b What four conditions specified in E-0 require one of more RCPs to be tripped?
OUESTION 7.05 (1.00)
What action must -tne taken if two or more rods do not fully. insert after a reactor trip?
(xxxxx CATEGORY 07 CONTINUED-ON NEXT PAGE xxxxx)
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 11
~~~~ 5656L665CdL 66NTR6L'~~~~~~~~~~~~~~~~~~~~~~~
R
~
QUESTION 7.06 (1.00)
If a pressurizer spray valve. fails open and can not be closed, its associated RCP must be stopped.
If pressurizer pressure continues to
-dacrease, procedure ES-01 requires the D RCP to be stopped.
Explain why.
QUESTION 7.07 (1.00)
LWhan, following a failed RCP start attempt in which the RCP failed to achieve full speed, can a restart be attempted?
QUESTION 7.00 (1.50)
Whet five indications, per ES-0.4, are used to verify natural circulation?
QUESTION 7.09 (1.00)
Why are the seal injection throttle valves opened VERY SLOWLY when restoring seal injection flow to the RCPs?
GUESTION 7.10 (1.00)
Naae the four ways specified in the steam senerator tube rupture procedure to identify the ruptured steam senerator.
QUESTION 7.11 (1.00)
In additionEto dumping steam to the condenser or out the atmospheric relief valves, what'two methods are available for post steam senerator tube rupture cooldown?
QUESTION. 7.12 (1.00)
What would be the indication of voiding in the RCS during depressurization?
(***** CATEGORY 07 CONTINUED ON NLXT PAGE
- )
r l.
I.
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12
~~~~ 56E6L66iC5E 66sTRUL
~~~~~~~~~~~~~~~~~~~~~~~~
R
. QUESTION 7.13 (2.00) c.,When, after a LOCA, must ECCS be transferred from injection to
(.5) cold les recirculation?
b.
When, after a LOCA, must ECS be transferred from cold les to hot
(.5) les. recirculation?
c.'What are the two reasons for transferrins to hot les (1.0) recirculation?
QUESTION 7.14 (1.50) e. 7 Wiie t 'is the~ pressurizer auxiliary spray water differential
(.5) temperature limit?
b..What action must be taken if this differential temperature limit (1.0) is exceeded?
QUESTION. 7.15 (2.00)
All RCPs and RHR pumps may be shut off for a certain period of time during cold shutdown provided two precautions are observed.
=a.
How lons is that period of time?
(0.5) b.
What are those two precautions?
(1.5)
QUESTION 7.16 (1.00)
Any plant changes which cause a sudden change in RCS temperature during an Epproach to criticality must be avoided.
Name two plant changes which could affect RCS temperature.
QUESTION 7.17 (1.00)
What action must be taken if the reactor is not critical by the time the control rods reach the maximum allowable withdrawal limit?
(*xxxx CATEGORY 07 CONTINUED ON NEXT PAGE *****)
l
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 13
~~~~ 5656E66fCEE~66NTR6E-~~~~~~~~~~~~~~~~~~~~~~~
R OUESTION 7.18 (1.00)
What action must be taken following a thermal power change in excess of 15%
of rated thermal power?
QUESTION 7.19 (1.00)
Why should the turbine not be operated at load with a back pressure in excess of 5 inci'es of mercury absolute?
QUESTION 7.20 (1.00)
When n.ust the PORV Cold Overpressure MitigaLiun System be placed into ecrvice?
QUESTION 7.21 (1.00)
What are the immediate operator actions in case of a continuous rod withdrawal accident?
(*xxxx END OF CATEGORY 07 xxxxx)
O
n T.
ADMINISTR'ATIVE PROCEDURES, CONDITIONS, AND' LIMITATIONS PAGE 14 8
-QUESTION 8.01
(.50)
- ?True:or False.- The shift compliment may be one less than.the minimum
-requirement for-a maximum'of two hours in order to accomodate on-comins Eshift members-who arrive late.
UUESTION. 8.02
(.50)
Who h'as1thefauthority to. clear the Control Room of nones'sential personnel?
.0UESTION 8.03 (1.00)
.If an. individual performing an activity can not follow the procedure as
' written, Wha't must'he do?
-~ QUESTION 8.04
(.50)
'True or False.
If a temporary procedure change.is beins written, approved, and implemented-in the field, its implementation is permitted prior to corp stion of the Nuclear Safety Evaluation Checklist.
1 QUESTION 8.05
(-.50)
- If-a. temporary. change to a procedure is siven to the SS/OS on a backshift,
-to whom does he send the original copy of this temporary change?
' QUESTION 8'.06
(.50)
Who can' approve deviations from the overtime restrictions?
QUESTION 8.07.
(.50)
'True or False.'
Routine Reactor Building entries require completion of a JVary High Radiation Area Access Request Form (CA 420) prior to the entry
.if.the-Reactor Building is designated a Very High Radiation Area.
(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 15 GUESTION 8 08
(.50)
A aaximum of _____ people are allowed inside containment at any one time while the reactor is in mode 1.
QUESTION 8.09 (1.00)
What action must be taken if the 2047 level Personnel Hatch becomes inoperable while personnel are inside containment when the reactor is in code 1?
GUcSTION 8 10
(.50)
Which individual has the authority to waive tha pre-entry radiation survey requirement when conditions necessitate immediate access to the Reactor Buildins?
QUESTION 8.11
(.50)
True or False.
The protective clothing requirements for an emergency R;cetor Building entry are not applicable to fire brisace members who are dressed in bunker sear.
QUESTION 8.12
(.50)
True or False.
Personnel entering a suspected atmosphere of unknown herards in an emersency situation shall use an air purifyins respirator (filter mask) until the atmosphere can be tested.
QUESTION 8.13
(.50)
True or False.
Personnel who have accessed the RCA under a GRWP are not parmitted to enter a room which has a SRWP in effect for any portion of that room.
(**xxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)
v l*
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 16 QUESTION 8.14 (1.00)
- o. Who has' jurisdictional authority over the main senerator disconnect (.5) switches?
b.
Who has functional authority over the main generator disconnect
(.5) switches?
QUESTION 8.15 (2.00)
Name two general situations in which tass placed on safety related systems do not1 require physical independent verification to be perforned.
nt!ESTION 8.16-(liOO)
Nane two of the three seneral categories of Safety Related systems.
QUESTION 8.17 (1.50)
- a. What is the purpose of a Local Control Tas?
(1.0)
- b. How many Local Control Tass may be issued for a component at one
(.5) time?
QUESTION 8.18
(.50)
If the individual holding the Workman's Protection can not be reached, who osy assume responsibilty and authorize release?
OUESTION 8.19
(.50)
True or False.
Priority E maintenance may start without an approved work request.
-QUESTION 8.20
(.50)
Trueoor False.
Personnel Protection requirements are unnecessary and need not be assigned by the Safety Department prior to entry into a confined spece if the atmosphere is classified as Class A.
(*****-CATEGORY 08 CONTINUED ON NEXT PAGE
- )
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 GUESTION 8.26
(.50)
True or False.
When one component addressed in a surveillance is in the Equipment Out-of-Service Los, the surveillance can not be signed as Ecceptable until the out-of-service component is returned to service and tessted.
QUESTION 8.27
(.50)
True or False.
Equipment Out-of-Service Los entries are required to be aade for all Technical Specification equipment taken out of service, even if the reactor is in mode 4 and the equipment is only required to be operable when the reactor is in mode 1.
QUESTION P-28
(.50)
Which of the followins los books is NOT required to be reviewed by the on-comins shift supervisor PRIOR to watch relief?
- o. Workman's Protection Assurance and Caution Tassing Los b.
U.
R.
O.
Los
- c. Temporary Modification Los d.
Equipment Out-of-Service Los GUESTION B.29 (1.50)
- e. Which Technical Specification Safety Limit is applicable when the
(.5) reactor is in mode 5?
- b. What two actions must be taken within one hour if this Safety (1.0)
Limit is violated while the reactor is in mode 5?
GUESTION 8.30 (1.00)
How many centrifugal charsins pumps are required by Technical Specifications to be operable when the reactor is in mode
- a.
2?
b.
4?
(**xxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)
a
-D8.;JADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19
~ QUESTION 8.31-(.
50)
Whose permission is:needed prior to changing the position of any locked conponent?'
QUESTION 8.32 (1.00)
How is the' position of a locked'open valve verified-if an installed locking-device prevents handwheel movement?
(***** END OF CATEGORY 08 *****)-
(************* END OF EXAMINATION ***************)
I e
5.
FLUIDS, AND PAGE 20
____ THEORY OF NUCLEAR POWER PLANT OPERATION, ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER 5.01 (1.50)
SDM in mode 5 is 1.0% Ak/k
(.2)
SDM in mode 4 is 1.3% Ak/k
(.2) increase in SDM is.3% Ak/k, which equals 300 pcm
(.3) fron,Fi ure 5-la, boron worth at 1000 pcm and 200 F is 13 pcm/ ppm
(.3) 3 chcnse in boron concentration is 300/13 = 23 ppm
(.2) from Figure 5-1b, 220 gallons of boric acid must be added
(.3)
REFERENCE Tach. Specs 3.1.1.1, 3.1.1.2. Table 12 KSA 3.1 001 010 K5.35; 3.6 ANSWER 5.02 (1.00)
Critical rod position is predicted to be greater than the zero power rod inaertion limit (46 steps on bank C or 161 steps on bank B).
REFERENCE s'
Tsch Specs 4.1.1.1.1 KSA 3 1 001 010 K5.35; 3.6 ANSWER 5.03 (1.00)
At 1000 MWD /HTU the poison rods are burning out, affsetting fuel depletion.w s
At 9000 MWD /MTU almost all of the poison rods have burned out, requiring boron dilution to offset fuel depletion.
REFERENCE Lcree PWR Core Control, p 2-12 KSA 3.1 001 010 K5.21; 3.9 e
l
3 --
^)
{
.b 2
i 5.
THEORY OF NUCLEAR POWER PLANT' OPERATION, FLUIDS, AND PAGE 21
-ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
~
h 1 ANSWER
,5.04 (2.00) s.
- a. Any three-o( the followins (.5 each);
.1.
Maintain' acceptable power distribution limits.
'2. Maintain minimum shutdown marsin.
- 3. Limit potential effects of rod misalignment.
'4.
Minimize effects of an ejected rod.
~
~
b.-jMaintain a negative MTC.
+
/
w
..k 1
3 x
REFERENCE is, Tech Spec Bases 3.'/( 1r3 KSA 3.1-2 K5.04; 4.7{
t
- b. Curve Book Figure'2-13 5 ^'
-q.
ANSWER
'5.05 (1.00)
Possibility of localized damage from some of the fuel elements becoming ovorheated.
OR Exceeding power peaking factors.
REFERENCE Lerse PWR Core Control, p 4-29 s
KSA 3 1-3 K5.38; 4.1 ANSWER 5.06 (1.00)
Encure that the axial peaking factor for Fe(Z) is not exceeded durins oither normal operation or xenon redistribution.
REFERENCE s
Tech Spec Bases 3/4 2.1
\\
KSA.3.1-4 K5;52; 3.6 g
fi
,U '
1 s
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22 ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER 5.07 (2.00)
- e. False
(.5)
- b. True
(.5) c.
True
(.5) d.
False
(.5)
REFERENCE
-Curve Book-Fisures 4-1, 4-2 and 4-3 i
Large PWR Control, p 4-31 KSA 3.1-2 K5.13; 4.0 ANSWER 5.08 (1.00)
Dalayed neutrons are born at energies below the threshold energy for fast fission of uranium 238-( ^.ich i:
- e,13nificant ':- c 1 r 3 ; ; - 1 r - e-i ehetl
-eir2y md n e+
- 1 ; ; '.. ; 3.)
ga + %: -
i-
'ha d:x. _^ ; e n h a b i 1 M y f o i t
REFERENCE Fundamentals of Nuclear Reactor Physics, p 7-34 3.1 001 000 K5.47; 3.4 ANSWER 5.09 (1.00)
Production of plutonium isotopes.
REFERENCE Fundamentals of Nuclear Reactor Physics, p 7-33 KSA 3.1-3 K5.47; 3.4 ANSWER 5 10
(.50)
Borate REFERENCE Large PWR Core Control, p 8-32
-KSA 3.1-2 K5.06; 4.1
ggj -
~-
T
=:.
.. N.
,s 4.;
t 4
-s
~ 5 i' ' THEORY OF NUCLEAR POWER PLANT OPERATION,-FLUIDS, AND PAGE 23
' ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER ~
'5.11 (1.00)
Herefnegative MTC at EOL.-
(.5)
Greater fuel depletion at the bottom of the core at EOL.
(.5)-
REFERENCE y-Large~PWR Core Control, p 8-20
. /KSA:3.1-38 EK1.21; 3.2 ANSWER 5.12 (1.00)
-y
. 4
- 3. g 4-og M to 15 seconds %
y
(.5) g
-b.\\12-tox16 mindtes
(.5) e %-
.g
. C
's J 4-REFERENCE ~
' ' ~
' Fundamentals of Nuclear Reactor Physics, p 7-70 KSA 3 1-43 EK1'.04; 3.9 h,
. ANSWER 5.13 (1.00)
' Increasing moderator temperature ~ increases the neutron migration. length,
. increasing the probability of neutron leakage.
With more rods inserted into.the core, the! significance of n'eutron leakage-becomes greater for a j-
,given temperature.
Increasing neutron leakage.causes MTC to become more j
n2gatjv
' REFERENCE Large-PWR Core Control, p.3-22
[c cKSA 3.1-35 EK1.17; 3.7 2
9 lp-a
,. 3 i;
b e
d
15.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND.
PAGE 24
. ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
. ANSWER.
5.14
(.75) 1500 psis = 1515 psia 1515 psia corresponds to a saturation temperature of 598 F 598 100'= 498 F 498 F corresponds to a saturation pressure of 675 psia 675 psia = 660 psis
(.75)
(650.to 670 psis will be awarded full credit)
REFERENCE-
' Steam Tables KSA'3.1-4 K5.56; 4.6 ANSWER 5.15
(.75) 2100'psis = 2115 psia 25 psis = 40 psia Isanthalpic expansion.of saturated steam from 2115 psia to 40 psia gives
.c.caturated mixture of liquid and steam.
'Th21 saturation temperature at 40 psia is 267 F.
(.75)
.(260 F to 280 F.will be awarded full credit.)
REFERENCE Steam Tables Mollier Diagram KSA 3.3-2 A1.09; 3.7 ANSWER 5.16.
(1.00)
..Any two of'the following:
- 1. Radiolytic-decomposition of water
(.5) 2.-Zirconium-water reactions
(.5)
- 3. Metal corrosion
(.5) 4.' Release of hydrogen dissolved in the RCS
(.5)
REFERENCE Containment Ventilation Lesson Plan, p.20 KSA 3.6-23 K5.03; 3.6
a 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 25 ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER 5.17 (1.50)
- o. 6%
(.5)
- b. Prevent an explosion.
(1.0)
REFERENCE Containment Ventilation Lesson Plan, p 21
-KSA 3.1-16 K5.02; 3.9 ANSWER 5.18 (1.50)
[d?;;i.cl 1^^% pm.ur iave is dau.o t.J Grev
\\
LLawest opening steam senerator safety valve setpoint is 1185 psis.
(.2)v 1185 psis = 1200 psia
(.2)y Saturation temperature corresponding to 1200 psia is 567 F.
(.2)5 Stoam senerator pressure at 100% power is 1000 psig.
1000 psis = 1015 psia
(. 2 )4 Seturation temperature corresponding to 1015 psia is 547 F.
(.2)v 567 F -~547 F = 20 F
(. 7) o5 REFERENCE
-Toch Spec 3/4, p 7-3 Rod Control Lesson Plan, p6 KSA 3.2-5 K5.08; 4.1 ANSWER 5.19 (2.50)
- c. Prevent Thot from reachins saturation.
(.5)
'b.
As power increases, AT increases, so Tave must decrease to keep
(.5)
That from reaching saturation.
_f;,
_-W
% % -- ' ~
See attached Figure 3-19. & c stifoky.
(1.0) c.
- d. Steam Generator Safety Valves
(.5)
REFERENCE Thermal Hydraulics, p 13-53 Tech Specs, p 2-1 System Generics 2-1, 45 O
FIGURE 5-19 s REVISIcy 3
g g
M N
.. - _g
- 9% L..-
C&SS. k-UNACCEPTABLE OPERATION 000 p*g
_--i ly w
=_
- .=.-
_:P.
r___De/g 040 U 1
- A
'_Q
_h y = y g 88/g -h
-_ =_
7_
- _-=_
C W
w P, 78g #el g.
L a
~T
_p = g
_ g:
1
-l.
_ _ peg 9 ;
_- =
l 2:..
m p: --
UE 800
^
e
_ =_
- w
\\
.. :_r.
-SSS E
- =
i 1_
550 i;"
ACCEPTABLE y
OPERATION
=
500 _'~
r.=:
F-+.9 *.9 0
0.2 0.4 0.5 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION 1
../
CALLAWAY - UNIT 1 2-2 e,, - - -,-- - - -, -,, - - - -
c
r
.5s THEORY OF NUCLEAR' POWER' PLANT OPERATION, FLUIDS, AND PAGE 26-
' ANSWERS -- CALLAWAY,
- -85/12/17-HIGGINS, R.
KSA 3.9-2 K5.01; 3.8 ANSWER.
5.20 (1.00)
Nsutron embrittlement
' REFERENCE Toch Specs: Figure 3.4-3 Tharmal Hydraulics, p 13-60 KSA 3.1-2.K5.15; 4.0
' ANSWER' 5.21 (1.00)
So the' pump reaches rated speed. faster, thus limiting the amount of time otorting current is applied to the pump.
REFERENCE LThormal Hydraulics, p 10-43
-KSA.p A-9, 412
.,..-..,n
,,e
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 27 ANSWERS---'CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER 6.01 (1.00)
IT'ha motor-operated gate valves for B and C accumulators may become inoper-eble after a LOCA because of their location 4 These accumulators have two vant valves to provide redundant means for venting nitrogen and preventing it:from being introduced into the RCS.
' REFERENCE-AccumulatorLLesson Plan, p 4
-KSA-3.2-13-K6.03; 3.2 ANSWER 6.02 (2.00)
- o. Boron dilution protection system: if counts received by the (1.0) source ranSe: instruments in one minute exceed twice the number of counts received in any one of the last ten minutes, charsing pump suction shifts from the VCT to the RWST.
b.:Re3ain' adequate shutdown margin.
(1.0)
REFERENCE-NIS Lesson Plan, p 18 ANSWER 6.03 (1.00) power ranse low power trip REFERENCE
'NIS Lesson Plan,=p 27 ANSWER 6.04 (2.00)
(.5) closed.
Overpressurization protection for the SI system.
(.5)
-2.
Either both of the SI pump recire valves (8814 A/B) or the common
(.5) header isolation valve (8813) must be closed.
Prevents pumping of radioactive fluid back to the RWST.
(.5)
I
r 6..~ PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 28 ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
REFERENCE RHR Lesson Plan,.p 6 ANSWER 6.05 (2.00) s.' 1. 2/4'1o-1o steam senerator levels (.05)'of 23.5% (.1) on 1/4 steam senerators-(.1)
'2.
loss of both MFW pumps
(.25) 3.
LOCA sequencer
(.25) 4.
LOP sequencer
(.25) b+ 2/3 AFW suction pressure cf 21.71 psia or less (.5) in conjunction with an AFW actuation si3nal
(.5).'
REFERENCE AFW Lesson Plan, p 15 ANSWER 6.06 (1.00)
.Possible pump failure due to steam binding. Of
~k NCk E
O IE-Dulletin No 85-01 KSA 3.5-93 A2.
i 3.0
&+y 11~3 ANSWER 6.07 (1.00)
Flow restricting orifices.
REFERENCE Sarvice and' Instrument Air Lesson Plan, p 11 ANSWER 6.08 (1.00)
- e. Closed
(.5)
- b. Nitrosen Accumulator
(.5)
REFERENCE Main Steam System P+ID, OP-M-02AB01(G)
Service and Instrument Air Lesson Plan, p 10
n-6.'__ PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 29 ANSWERS ---CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER
'6.09 (1.00) 08
- c. CIS 'B'i~2/4 containmer.t pressure signals 27 psis or greater.
(tv44 b.
Restore CCW flow to the RCPs during a CIS
'B' condition
($Ibh
-REFERENCE.
CCW Lesson Plan, p 19 and 22 Containment Spray Lesson Plan, p 11 ANSWER 6.10 (1.00)
Any two'of the followins (.5.each):
- 1. Ionization! detects. ions produced during combustion.
- 2. Photo-electrict detects visible smoke particles.
- 3. Infrared: senses light produced by open flames.
- 4. Thermal:
e of temperature rise. vid REFERENCE Pyrotronics Lesson Plan,-p 7 ANSWER 6.11 (1.00)
~
Two of the following:
1.
Permissive switch on the containment control console is in the
(.5) permissive position.
- 2..Both opender lifting arms are in the down position.
(.5)
- 3..The. fuel transfer tube sate valve is fully open.
(.5)
REFERENCE.
Fuel Handlins System Lesson Plan, p9 ANSWER.
6'.12 (1.00)
Normal - CVCS
(.5)
Enersency - ESW
(.5)
REFERENCE Fual Pool Cooling Lesson Plan, p 10
i.
6..
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 30 ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
= ANSWER 6.13 (3.00)
- a. Four.of the followins (.5 each):
- 1. Rod Control' System
- 2. Steam Dump Control System 3.
Pressurizer Level Control System
.4.-Reactivity Computer for HTC
- 5. Rod Insertion Limit b.
C-16,. the automatic turbine loading stop signal (.25) which inhibits
. turbine loading'if Tave is less than 553 F (.25) or 20 F below Tref
-(.25), whichever is lower (.25).
REFERENCE RCS Instrumentation Lesson Plan, p 18 ANSWER ^
6.14 (1.00)
Any two.of the following (.5 each):
- 1. SI pumps 2.-RHR pumps
- 3. Containment Spray Pumps
. REFERENCE Sefeguards Power Lesson Pl.n, p 3-15, 16 ANSWER 6.15 (1.00)
If two or more power ranse instruments exceed P-10, both source range datectors vould be disabled.
REFERENCE NI Lesson Plan, p ?3 ANSWER 6.16 (1.00)
Limit flow to the' capacity of one centrifusal charging pump in the event of~a. vent'line rupture.
REFERENCE Recetor Vessel.and Core Construction Lesson Plan, p 21 9
i.
6.-
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 31
~ ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
-ANSWER 6.17 (1.00)
Any four of the following:
- 1. Containment High Radiation RE 21 13a-(.25)
- 2. Control-Room Air _ Supply High Radiation R E O V
- 05~
(.25)
- 3. Containment Purge Exhaust High Radiation KE 22 +33
(.25)
- 4. Containment Isolation ~ Phase A
(.25) 5.-Safety Injection
(.25) 6.
Fuel Building High Radiation (FBIS)
(.25)
REFERENCE Vcntilation System Lesson Plan, p 11 ANSWER 6.18 (1.00)
Tho alarm is not valid. (.5) It takes several minutes for particulates to collect on the filter paper and be transported to the detector. (.5)
REFERENCE Process Radiation and Area Radiation Monitoring System Lesson Plan, p 9 and 18 ANSWER 6.19 (1.00)
Minimize the severity of a dilution accident.
REFERENCE CVCS P+ID, OP-22BG05(0), Note 14 ANSWER 6.20 (1.00)
R2 verse rotation during pump starts causes excessive starting currents.
REFERENCE Reactor Coolant Lesson Plan, p 10
<y l
7'.
PROCEDURES - NORMAL, ABNORMAL,. EMERGENCY AND PAGE-32
~~~~ d656L6656dL"66 TREL
~~~~~~~~~~~~~~~~~~~~~~~~
R ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER
-7.01 (1.00) 1.'All-rod bottom lights illuminated
(.25)
- 2. Reactor-trip and bypass breakers open
(.25)
- 3. NR-45 recorder'- decreasins flux
(.25)
- 4..NIS indicators - decreasing flux ~
(.25)
REFERENCE E-0, p2 ANSWER
-7.02 (1.50)
-c.
160 F
(.5)
- b. Core exit TCs are not reliable in the post-LOCA containment (1.0) environment.
. REFERENCE E-0, p 2 and 13 ANSWER 7.03 (1.00)
O.-615
(.25)
- b. 320
(.25) c.
15%
(.25) d.
28%
(.25)
REFERENCE
'E-Or.p 6, 8 and 14, and the foldout
o -
k.
- 7.: ' PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 33
~~~~ E6 EEL 65EEdL EUNTRUL
~
~~~~~~~~~~~~~~~~~~~~~~~~
R s.____________________
ANSWERS -- CALLAWAY-
-85/12/17-HIGGINS, R.
"ANSWEO-7.04.
(1.50)
- 1. At11 east-one' charging pump or SI pump running
(.6) and RCS pressure less than 1400 psis.
-(.3)
- 2. CCW. flow to the RCPLmotor is lost for more than two minutes.
-(.3)
-3.
Upper.bearin3 temperature reaches 195 F.
(.15) a4. Lower ~ bearing temperature reaches 195 F.-
(.15)
REFERENCE.
E-0 foldout.
l ANSWER' 7.05
- ( 1. 0 0 )
Ecorgency borate 100 ppm for each control rod not fully _ inserted.
REFERENCE ES-0.1, p. 3 '
ANSWER-7.06' (1 00)
The spray 1 driving head is the differential pressure between the surge line
.and spray.line.
W'ch D RCP runnin39 the static pressure-at the surge line will be less thar. the static pressure in the loop ~with the idle RCP,
.thareby-seneratin3 Spray flow.
REFERENCE Reactor. Coolant. Lesson plan, p 22
' ANSWER 7.07 (1 00)
~l:..After the motor has been allowed to cool by standing idle for at least
30-minutes.
a,,
- REFERENCE ES-0.1,: attachment 4, p 1 f
w
p 7o PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34
~~~~
RA656L6GEUEE~E6NTR6L'~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS - 'CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER-7.08 (1.50) 68 10 ~RCS subcoolins - more than instrument error ( e @- -
e= ydqF )
d
(.3) 20 Steam pressure stable-(.3) 3a RCS hot les_ temperature - stable or slowly decreasins
(.3) 40 RCS cold les temperature - near saturation temperature for steam
(.3) pressure 5.
Core exit TCs - stable or slowly decreasins
(.3)
REFERENCE-ES-0.4, attachment 2 ANSWER 7.09 (1.00)
Avoid thermally shocking the RCP bearings and seals.
REFERENCE ES-1.1, attachment 4, p2 ANSWER 7.1 44rw %(1.00) W
(
e4 M
- 1. tine @l4,c.
fW J-w i e s
ect increap in the st am senerator narrow ranse level (725T 20 High activity in the steam s e n e r a t,o r sample
?"i 30 High radiation in the steam senerator steam line kvEEF 40 High radiation in the steam senerator blowdown line
(.25)
Gr(Re2e"A & V " Y W " &
WK E-3, p4 ANSWER 7.11 (1.00)
- 1. Backfill'
(.5)
- 2. Blowdown
(.5)
REFERENCE ES-3.1, 3.2, 3.3
7.
-PROCEDURES NORMAL, ABNORMAL, EMERGENCY AND PAGE 35
~~~~ d656L66E6AL'666TR6L
~~~~~~~~~~~~~~~~~~~~~~~~
R ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER 7.12
(
.00).
n Repidly increasin-pressu zer level.
REFERENCE ES-1.2,~p 17 ANSWER 7.13 (2.00)
- c. When RWST_ level drops to 36%
(.5) b.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiation of the LOCA
(.5) c.
1.
Flush baron from the upper regions of the core
(.5) 2.
Quench any steam bubble in the top of the head
(.5)
REFERENCE SI Lesson Plan, pC E-1, p 12 and 13 ANSWER 7.14 (1.50)
- n. 320 F
(.5)
- b. Initiate an incident report for accountability (1.0)
REFERENCE OTG-ZZ-00001, step 2 6.1.1 AISWER 7.15 (2.00) 8.
One hour
(.5) b.
1.
No operations which could cause RCS dilution are permitted.
(.75)
- 2. Core' outlet temperature is maintained at least 10 F below
(.75) saturation.
REFERENCE OTG-ZZ-00001, step 2.9.1
+-.-g,--
e
,..,,.e,. - -. _ _, -.,
..y,.-,-.,,.-__,,-.,-.,--._.-e mw.
t 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36
~~~~ d656L65565L 66NTR6L""~~~~~~~~~~~~~~~~~~~~~~
R
~
ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER
'7 16 (1.00)
.1.
Feeding.the steam generators
(.5)
- 2. Withdrawing steam from the steam generators
(.5)
REFERENCE OTG-ZZ-00002, step 2.8 ANSWER 7.17 (1.00)
Fully insert Control Bank D REFERENCE OTG-ZZ-00002, step 4.2.11 ANSWER 7.18 (1.00)
Sreple the unit vent and containment purge for
& Q y u N, principle,s mma emitters.
OR RcsErze um
- M REFERENCE M/
y fiF cM / reg OTG-ZZ-00003,~ step 2.6
- ONi, 2 7 aed.;;t../8 ANSWER 7.17 (1.00)
A sencrator or electrical trip could overspeed the turbine.
REFERENCE OTG-ZZ-00003, step 2.7 ANSWER 7.20 (1.00)
Prior to any RCS cold les temperature reaching 368 F H644 jwou grr:11 r
'O^
F;i j-4v64 5
wiun y r c a m v.
4-c.
REFERENCE OTC-ZZ-00006, step 4.1.23 m
0 L_
f 27.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 37
~~~~ d6E6L6556dL"66UTE6L
~~~~~~~~~~~~~~~~~~~~~~~~
R
' ANSWERS -- CALLAWAY-
-85/12/17-HIGGINS, R.
ANSWER.
7.21" (1.00) 1.
Place -the rod bank selector switch in manual and attempt to insert
(.5) the control rods.
- 2. If the rods continue withdrawing, trip the reactor.
-(.5) f" REFERENCE OTO-SF-00002,'p 2 Il' L-l l
l
8.
-ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 38
' ANSWERS.-.CALLAWAY
-85/12/17-HIGGINS, R.
ANSWER 8.01
(.50)
~Folse.
. REFERENCE-APA-ZZ-00010 step 6.6 ANSWER 8.02
(.50)
SS/OS REFERENCE APA-ZZ-00010 step 6.8.3
' ANSWER 8.03 (1.00)
Pisce the system / component in a stable and safe condition (.5) and notify tha responsible supervisor. (.5)
REFERENCE
~APA-ZZ-00100 step 2.2.4.1.1
-ANSWER 8.04
(. 50)
True REFERENCE.
APA-ZZ-00101 step 5.1.5.1 ANSWER 8.05
(.50)
RMS Supervisor (Document Control)
REFERENCE APA-ZZ-00101 step 5.1.11
8.'__ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 39 ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
-ANSWER 8.06
( -.50)
Plant Manager'or EDO. -
' REFERENCE APA-ZZ-00130 step 2.2 ANSWER 8.07
(.
50)
' False.
REFERENCE APA-ZZ-00160 step 4.4.4.1.3 i
ANSWER 8.08
(.50) 20 REFERENCE APA-ZZ-00160 step 4.4.4.3.4 ANSWER 8.09 (1.00)
Roactor Building access-should be discontinued and all personnel should i
oxit the Reactor Building.
REFERENCE APA-ZZ-00160 step 4.4.4.3.5 ANSWER 8.10
(.50)
Shift Supervisor REFERENCE APA-ZZ-00160 step 4.4.5.1 l
l --
I i
1
- t..
V..
L 8.___' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 40 7
_ ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
j' ANSWER-
'8.11
(.50)
-True.-
REFERENCE
.APA-ZZ-00160Lstep'4.4.5.1.4 ANSWER 8.12
(
.50)-
False
' REFERENCE APA-ZZ-00160 steps 4.13.6.4 and 4.13.7 ANSWER 8.13
-(
.50)
False REFERENCE.
APA-ZZ-00161-step 3.5
. ANSWER 8.14 (1.00) 0.; Power. Dispatcher b.-SS/OS~
REFERENCE APA-ZZ-00310 step 3.4.2 ANSWER-8.15 (2.00)
- 1. When independent 1 indication unequivocally depicts the status of the component.
- 2. When-the concept of ALARA would be violated.
REFERENCE APA-ZZ-00310 step 4.2.2.31
L
[
8.-
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 41 ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
i ANSWER 8.16 (1.00)
Two of the following:
1.
systems which assure the integrity of the RCS boundary.
2.' systems which assure the capability to shutdown the reactor and maintain it safely shutdown.
3._ systems wh'ch mitigat off-site exposures.
/[
bI REFE E
~p APA-ZZ-00101 step 2.8 00f-D'Wa f M ANSWER 8.17 (1.50)
- o. Protection of the individual who must operate a component during (1.0) the c e of maintentance testing.
- b. Or
[
(.5)
REFERENCE APA-ZZ-00310 steps 3.5.2.1 and 3.5.2.5 ANSWER 8.18
(.50)
EDO REFERENCE APA-ZZ-00310 step 4.3.2 ANSWER 8.19
(.50)
True REFERENCE APA-ZZ-00320 step 2.9.1 ANSWER 8.20
(.50)
False
V l
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 42 ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
REFERENCE APA-Z7-00372" step 4.4.1.6 ANSWER 8.21 (1.00) 0+: Plant Manager or. EDO
- b. SS
. REFERENCE
- APA-ZZ-00380 steps 5.1.5.1.1 and 5.1.5.2 ANSWER 8.22-(.50)
Folse REFERENCE APA-ZZ-00742 step 5.2.2-ANSWER 8.23
(.50)
OS REFERENCE APA-ZZ-00743 step 2.2 ANSWER 8.24 (2.00) 0.-Site Emergencies b.' Alert
'c.
Public Relations
- d. Recovery Manager
' REFERENCE EIP-ZZ-00102 steps 4.3, 4.11.1, 4.10.4 and 4.12 O
8.
-ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 17 QUESTION 8.21 (1.00)
Who is the approval authority for installing jumpers on:
- o. Safety Related equipment?
(.5) be non-Safety Related equipment?
(.5)
DUESTION 8 22
(.50)
Trua or False.
When several hot work jobs are to be performed in a plant croa where hot work permits are required, only one hot work permit needs to bJ issued.
QUESTION 8.23
(.50)
Who is the Fire Brigade Leader?
QUESTION 8.24 (2.00)
Co plete the following statements.
a.
Evacuation is mandatory for emergencies classified as _____ or
(.5) higher.
b.
The OSC shall be activated for emergencies classified as _____ or
(.5) higher.
c.
The emergency coordinator or EDO shall contact _____ to give
(.5) approved information for release to the news media.
- d. At Site or General Emergency classification levels, the _____ is
(.5) notified and asked to activate the Corporate Emergency Organization.
QUESTION 8.25 (1.00)
Undst what circumstances is it permissible to use recycled (tritiated) water to provide makeup to the RMWST?
(*x***
CATEGORY 08 CONTINUED ON NEXT PAGE *****)
r i
?
8..
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 43 ANSHERS -- CALLAWAY ~
-85/12/17-HIGGINS, R.
- ANSWER
~8.25 (1.00)
LTo avoid load. reduction when normal makeup water is not available from the DWST or.RMWST.
REFERENCE Standing Order.85-025 ANSWER 8.26
(.50)
Folse REFERENCE
'Stending Order 84-24
= ANSWER 8 27
-(
.50)
True REFERENCE
.00P-ZZ-00002 step 4.1.2 s
ANSWER-8.28
(.50).
'b.
. REFERENCE-ODP-ZZ-00003 steps 4.2.4 and 4 2.7.2
' ANSWER 8.29 (1.50)
- o. RCS pressure shall not exceed 2735 psis.
(.5) b.
- 1. Reduce.RCS-pressure to within its limit within 5 minutes.
(.5)
- 2. Notify the NRC Operations Center by telephone as soon as
(.5) possible.
REFERENCE Toch Spec 2.1.2 and 6.7.1.a L
f.
I D-
. 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 44 ANSWERS -- CALLAWAY
-85/12/17-HIGGINS, R.
~
- ANSWER 8.30 (1.00)
- c. 2 b.
1 REFERENCE
- Tech Spec 3.5.2 and 3.5 3 AN'SWER-8.31
(.50)
SS/OS-
-REFERENCE' ODP-ZZ-00004 step 5.4.2
- ANSWER 8.32 (1.00) l Rbnove the-locking device and attempt to move the handwheel in the closed direction only enough to verify valve movement.
REFERENCE ODP-ZZ-00004 step 6.1.2 e
b h
_