ML20138B711
| ML20138B711 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/18/1986 |
| From: | Mcmillen J, Morgan T, Spencer K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20138B667 | List: |
| References | |
| 50-237-OL-86-01, 50-237-OL-86-1, NUDOCS 8603250136 | |
| Download: ML20138B711 (100) | |
Text
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U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-237/0L-86-01 Docket Nos.:
50-237; 50-249 Licenses No.:
Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name:
Dresden Nuclear Power Station Examination Administered At:
Dresden 2/3 Nuclear Power Station near Morris, IL j
Examination Conducted:
January 28-30, 1986 h
/
3//fr/E Examiners:
T.
Mor n Date s
K. A. Spencer J//8/sc Date~
N k
Approved By: /
I. McMillen, Chief 8//f/7/s perating Licensing Section DaKe /
Examination Summary Examination administered on January 28-30, 1986, (Report No. 50-237/0L-86-01)
Examinations were administered to two senior operator candidates and four reactor operator candidates. A section 8 only, written examination was administered to one additional senior operator candidate.
Results:
Two senior operator candidates and four reactor operator candidates passed the written examination and three reactor operator and all senior operator candidates passed the oral / operating examination.
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REPORT DETAILS 1.
Examiners T. L. Morgan, INEL K. A. Spencer, INEL 2.
Examination Review Meeting At the conclusion of the written examination, the questions and answers were given to the facility training staff for review and comment.
At the exit meeting the facility supplied the examiners with their comments to the Reactor Operator (RO) and Senior Reactor Operator (SRO) written examination.
On the following pages are the facility comments and the examiner's resolution to each.
3.
Exit Meeting On January 30, 1986, an exit.ma'. ting was held.
The following personnel were present at this meeting:
B. Zank, Training Supervisor R. Christensen, Unit 1 Operating Engineer D. Scott, Station Manager S. Stiles, Training Instructor R. Flessner, Services Superintendent T. Morgan, INEL K. Spencer, INEL The facility was informed of those individuals who clearly passed the oral and/or simulator examinations.
In addition, the examiners raade the following observations:
a.
Areas of common weakness were found in the use of procedures, especially during supplementary actions.
b.
Concerns were expressed because several emergency procedures had been changed in October and the outdated material was supplied to the examiners in December, as reference material used to prepare the written and oral examinations.
c.
During one of the examinations it was noted that the notification checklist, EPI-300-c-1, that was in the control room file drawer was different than the form in the controlled copy of the Emergency Plan Implementing Procedures.
d.
The simulator reference material supplied to the examiners would not be sufficient for future examinations because of the soon to be published requirements for simulator examinations.
The deficiency j
in the simulator material is that effects are not given for any of 2
the malfunctions listed. This is necessary information for the examiners to prepare the type of simulator scenarios that will be required in the future.
e.
The G.E. simulator Instructor, who assisted the examiners during the examining process, was very helpful in ensuring the smooth running and realism of the simulator scenarios.
Attachment:
Facility Examination Comments and Examiner Resolution e
3
FACILITY EXAMINATION COMMENTS AND EXAMINER RESOLUTION 1.
FACILITY COMMENT QUESTION 1.05 and 5.04 Additional answer:
" pressure in the steam dome" Reference - GE Thermodynamics, Heat Transfer and Fluid Flow, page 7-93 and 94.
EXAMINER RESOLUTION Above answer and reference added to answer keys.
2.
FACILITY COMMENT QUESTION 1.09 Although this information is found in the GE lesson plan, the information in the lesson plan is incorrect as per the station nuclear engineers.
We consider this to be beyond the scope of knowledge required by a Reactor Operator.
EXAMINER RESOLUTION It is agreed that the "why" portion of the question is not as important as the concept, therefore, the question value has been reduced from 1.0 to +0.5.
The total point value of Section 1 has been reduced by 0.5 points.
3.
FACILITY COMMENT QUESTION 1.10 Additional acceptable answer:
" Trip on low suction pressure" Reference - Lesson Plan Book 2, Chapter 6, page 14.
EXAMINER RESOLUTION Facility comment is referenced to "Feedwater Level Control System."
The question applies to general theory and not to specific systems.
The answer key has not been changed.
4.
FACILITY COMMENT QUESTION 1.12 Dresden considers this question to be obscure and beyond the scope of knowledge required by a Reactor Operator.
EXAMINER RESOLUTION Operator Licensing Examiner Standards (NUREG-1021) ES-202 B.
Scope states, "This category shall contain questions...
measured as resultant characteristics shall be included in this category."
The answer key has not been changed.
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5.
FACILITY COMMENT QUESTION 2.01 and 6.01 Answer key is incorrect:
Should be changed to "
AC and DC solenoids are deenergized and the test solenoid valve is energized Reference Lesson Plan Book 3, Chapter 3, page 13.
EXAMINER RESOLUTION Answer key modified to reflect AC and DC solenoids deenergize and the test solenoid energizes.
6.
FACILITY COMMENT QUESTION 2.04 Normal readings for this instrument should not be expected knowledge.
This reading is not available in the control room and is only reported as being in specification to the control room operator by a non-control room operator who is responsible for monitoring this instrument.
EXAMINER RESOLUTION This question is in the " Plant Design, Including Safety and Emergency Systems." The lesson plan describes in detail the operation of this system including the instrument values.
Therefore, the values become an important, integral part for understanding system operation.
i The answer key remains unchanged.
7.
FACILITY COMMENT QUESTION 3.04 Additional acceptable answer:
" Place manual inhibit switch to inhibit." Reference - Lesson Plan Book 3, Chapter 11, page 13 and Fig. 7.
I EXAMINER RESOLUTION Above answer and reference added to answer key as a potential answer.
8.
FACILITY COMMENT QUESTION 3.12 and 6.10 Additional acceptable answer:
" Sensing point of reactor pressure for CRD valves MO-8 and M0-10.
Reference - P & I D.
EXAMINER RESOLUTION Above possible answer is unacceptable due to the reference material provided with comment did not support the facility comment.
Arswer key has not been changed.
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9.
FACILITY COMMENT QUESTION 4.01 and 7.01 Question asks for the eight (8) operator immediate action steps.
The answer key references three (3) procedures.
No single procedure lists eight (8) immediate action steps.
Considering all three (3) reference procedures, there are more than eight (8) possible correct immediate action steps.
Reference - 00A 6000-1, 5600-1, DGP 2-3 as listed in the answer key.
EXAMINER RESOLUTION During the exam, the candidates indicated there are only seven (7) immediate steps following a scram.
The exam was changed on the spot to require seven (7) operator steps for full credit.
DGP 2-3 lists seven (7) steps and two (2) substeps following a scram.
Answer key modified to list seven (7) possible steps.
Point value of question modified to 0.44 each.
Reference modified to DGP 2-3 only.
Possible answers are:
1.
Press both scram buttons and place the mode switch to SHUTDOWN.
2.
a.
VERIFY all rods are inserted to or beyond 04.
b.
Insert any rod not already at 00.
Record all rods not at 00 and notify the Shift Supervisor.
3.
a.
Attempt to maintain level between +8 and ++40 inches by multiple indication.
Control feedwater in AUTO unless control failure occurs.
b.
If necessary, STOP the feed pumps to prevent excessive reactor water level.
VERIFY that more than one type of level indication has high level before stopping pumps.
4.
IF ALL rods inserted to or beyond position 04, THEN VERIFY that the turbine and generator have tripped.
5.
VERIFY that recirculation pumps A and B run back to the minimum pump speed.
6.
VERIFY that the auxiliary power has transferred to the reserve auxiliary transformer.
7.
INSERT the SRM/IRM's.
Maintain the IRM's on scale to monitor shutdown.
10.
FACILITY COMMENT QUESTION 4.02 and 7.02 New emergency operating procedures give further guidance concerning injection of SBLC.
Reference DEOP 100-3, page 1 and 3 and DEOP 400-2, page 2.
3
EXAMINER RESOLUTION During the exam, the candidates indicated the ATWS procedure had been deleted.
Therefore, part A. was immediately deleted from the exam and the point value of this question decreased by 1.25.
The answer to part B. of the questions has been changed to read:
1.
All control rods are not inserted to or beyond Position 04 (0.5) AND Torus temperature cannot be maintained below 110 F (0.5).
- OR ******************
2.
If the reactor cannot be shutdown BEFORE Torus temperature reaches 110 F.
Shutdown = All rods are at or beyond Position 04, or as determined by a Qualified Nuclear Engineer.
Either answer 1 or answer 2 with at least one definition of shutdown is required for full credit.
The reference to questions has been changed to DEOP 400-2, Revision 0, page 2 and DEOP 100-3, Revision 0, page 3.
11.
FACILITY COMMENT QUESTION 4.06 and 7.04 The answer to this question is no longer referenced to DGA-18 but rather to DEOP 110-3. Also, a specific "nine ways" is no longer mentioned.
EXAMINER RESOLUTION During the exam, the candidates indicated DGA-18 had been deleted.
The question was changed to read, "... identified in DEOP 100-3 after you..
Answer key modified to list those means to alternately insert rods as described in DEOP 100-3 as follows:
1.
Verify that all scram valves are open as indicated by the blue scram lights on the full core display, if not remove the fuses behind the panels (902(3) - 15 and 17)).
2.
Arm the Alternate Rod Insertion (ARI) pushbuttons and depress the pushbuttons to open the ARI valves, hold for at least 20 seconds.
3.
If not yet scrammed, replace the fuses and close the ARI valves, then reset the scram and initiate a manual scram.
4.
Reset the scram and individually open the scram test switch for each rod at >04.
4
5.
If scram cannot be. reset, start both CR0 pumps, close the charging water valve, and try inserting using the emergency Rod In Control Switch, l
6.
Insert any not at 04 or beyond by venting the drive over piston area thru a hose to the floor drain for each rod.
Reference has been changed to DEOP 100-3, step 6, page 3, 4, 5, 6 and 7.
1 12.
FACILITY COMMENT QUESTION 5.08c and 1.11c i
Dresden operates in single element control.
Therefore, the correct answer should state, " Decrease due to level increase" with no mention of steam flow decrease.
EXAMINER RESOLUTION Answer key modified to remove "the steam flow decrease" and the point values were adjusted accordingly.
13.
FACILITY COMMENT QUESTION 6.09 Question is ambiguous.
"10 LPRM inputs to APRM Channel 2, mode switch in STARTUP." Other portions of the same question stated "downscale," " upscale" and " upscale." This portion; however, did not.
If 10 inputs failed, 11 would remain and no actions would be I
generated.
However, if 10 inputs remained,11 have failed and a half scram would have resulted.
EXAMINER RESOLUTION Assume facility comment is intended to Question 6.07.
Answer key modified to remove part "b" and point value of question decreased by i
0.5 points.
14.
FACILITY COMMENT QUESTION 7.09 Candidates are expected to describe general objectives and methods used in normal and abnormal procedures.
Therefore, the question is valid.
However, the answer is to specific as it lists the exact items as stated in the subsequent operator actions.
Reference:
NUREG-1021, ES-402, pages 3 and 4.
EXAMINER RESOLUTION The answer to the question is the "Immediate Operator Actions."
The ES-402 A, 3, Category 7 states, "In general, a candidate must demonstrate complete knowledge and understanding of the symptoms, automatic actions, and immediate action steps specified by offnormal or emergency operating procedures.
The candidate should be able to i
5
describe generally the objectives and methods used in the normal, offnormal, and emergency operating. procedures and -the methods used to perform the verifications." The answer falls in the category of
" complete knowledge and understanding of immediate action steps."
Answer key has not been changed.
15.
FACILITY COMMENT QUESTION 8.10 Asking for specific items from an operating order is not in accordance with NUREG-1021, ES-402, pages 3 and 4.
EXAMINER RESOLUTION The ES-402, A, 4, Category 8 states, "This category contains question on administrative, procedural, and regulatory items that affect safe operation of the facility." Question 8.10 falls into this category.
Answer key has not been changed.
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[-
l' 1 EXAMINER COMMENTS 1.
During the exam, Question 6.02a was clarified as to the fluid coupler stability.
2.
During the exam, Questions 3.07b and 6.09b were changed to remove " local"'.
control panel and insert " control room" control panel.
3.
During exam grading, the comment to Question 5.08c was discovered to apply to Question 1.11c.
Appropriate changes as described above were made to answer key.
4.
During exam grading, the point value assigned to Question 5.11 was found to be in error.
The total point value for Section 5 was reduced 1.0 points.
5.
During exam grading, the comment to Question 1.05 was discovered to apply to Question 5.04.
Appropriate changes as described above were made to answer key.
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6.
The EHC portion of answers 1.03 and 5.02 were removed because question
/
did not elicit this. The point value for the questions remained tre same.
7.
During exam grading, an additional answer to Question 8.09 was added, the point value remained unchanged, and reference added.
Additional answer:
Check APRM Gain Adjustment Factor on the computer.
If any AGAF greater than or equal to 1.02, adjust per DIS 700-17.
(How the AGAF is checked or DIS No.
are not required for full credit.)
Reference:
DGP 2-1, Revision 13, page 4 8.
Review of EPIP 200-20 required two changes to Question 7.05 as listed
- below, i
1.
Part 8 - Changed " Command' Center" t'o " Technical Support Center" 2.
Part C - Proceoure lists only one means to verify operation.
Items 1 and 3 were removed from answer key /.
The point value of question remained the same.
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I U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
DRESDEN 2&3 EC[ N/[
REACTO.R TYPE:
BWR-GE3 h//M/dE4 M/
DATE ADMINISTERED: 86/01/28 Cuan~Ts #g EXAMINER:
KING. M.
j7-)fh I
J APPLICANT:
gg
- g. d g INSTRUCTIONS TO APPLICANT:
Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 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 25.50 25.89 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 24.25 24.62 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.50 25.89 3.
INSTRUMENTS AND CONTROLG 23.25 23.60 4.
PROCEDURES - NORMAL, ABNORMAL.
EMERGENCY AND RADIOLOGICAL CONTROL 38.50 100.00 TOTALS FINAL GRADE All werk done on this examination is my cwn. I have neither given nor received aid.
APPLICANT'S SIGNATURE NAR 1419gg 1
F F:
Q
.'s
_1.
PRINCIPLES OF NUCLE POWER PLANT OPERATION.
's PAGE 2
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW t
QUESTION 1.01 (1.00)
The reactor is' started up after.a ' refueling outage.
Rods are pulled to the 100% line and power is then increased to 100% with recirculation flow.
After approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, reactor power has decreased to about 98% with no operator action.
Explain the primary cause for this reduction in power.
(1.0)
QUESTION 1.02 (2.00)
For the following transients, indicate which COEFFICIENT of reactivity; alpha T, alpha D, or alpha V tends to change reactor power FIRST and in what DIRECTION.
A. Fast closure'o2 one MSIV.
(0.5)
B.
Isolation of a feedwater heater string.
(0.5)
C. A control rod drop.
(0.5)
D. Relief valve lifting.
(0.5)
QUESTION 1.03 (2.50)
Explain. WET reactor power INCREASES 'as recirculation flow is INCREASED.
Ensure your exlaination continues to a STABLE CONDITION after the f. low ~ change.
QUESTION 1.04 (2.00)
Briefly explain which time in core life (BOL, Mid-of-life, or EOL) requires the least amount of positive reactivity addition to achieve
" prompt critical".
GUESTION 1.05 (1.50)
What are three (3) of the design or operational factors that insure adequate Net Positive Suction Head (NPSH) for the recirculation pumps?
(3 @ 0.5 ea.)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
m
',1.
PRINCIPLES OF NUCLE-POWER PLANT OPERATION.
's PAGE 3
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.06 (2.00)
While performing a reactor startup'and heatup per DGP 1-1, Unit 2/3 Normal Unit Startup, the MODE SWITCH has been placed in Run per procedure step 3-w.
ONE OF THE SRV's(Target Rock) OPEN.
DESCRIBE what will occur during this transient.
Assume no operator action and the valve remains open for ~5 minutes then closes.
(Include in your description the effects on plant power, pressure, vessel level, if any trip occurs, and the final steady state conditions.)
QUESTION 1.07
-(1.00)
The reactor is exactly critical LOW in the intermediate range.
A control rod is withdrawn one notch.
Describe what happens to indicated neutron level AND why?
(Continue description until a steady state condition is reached.
Assume no futher operator action other than ranging the IRM meters.)
(1.0)
QUESTION 1.08 (2.50)
With the plant operating at 100% power, a total loss of feedwater flow occurs.
Answer the following using the attached transient information (next page of exam),:
A.
WHY does reactor POWER initially decrease [ Area 1] AND subsequently decrease more rapidly [ Area 2]?
(1.0) 3.
WHAT is causing the reactor LEVEL increase several minutes after the loss of feed [ Area 3]?
(0.5)
C.
WHAT caused core FLOW decrease initially [ Area 4] AND subsequently [ Area 5]?
(1.0) s
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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PAGE 4
1.
PRINCIPLES OF NUCLEwft POWER PLANT OPERATION.
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.09 (1.50)
Unit 2 presently uses the CORE CON' TROL CELL concept.
A. What is a CONTROL CELL?
(1.0)
B. Why is this operating strategy used?
(0.5)
QUESTION 1.10 (3.00)
Give ONE undesirable result for each of the following.(Be more specific than " pump failure"):
1 A. Operating a motor driven centrifugal pump for extended periods of time with the discharge valve shut.
(1.0)
B. Starting a motor driven centrifugal pump with the discharge valve full open.
(1.0)
C. Operating a motor driven centrifugal pump under " PUMP RUNOUT" conditions.
(1.0)
QUESTION 1.11 (3.00)
Assume the reactor is operating at 100% power and one recirculation pump trips. Indicate how each listed indicated parameter would first change (Increase or Decrease.) and briefly explain why the change occurs.
A.
reactor power (1.0)
B. reactor water level (1.0)
C.
feedwater flow (1.0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
1.
PRINCIPLES OF NUCLE POWER PLANT OPERATION.
PAGE 5
THERMODYNAMICS. MAT TRANSFER AND FLUID FLOW QUESTION 1.12 (1.00)
~
The 8x8 fuel has a thermal time constant of approximately 5 to 6 seconds. This means that in 5 to 6 seconds following a sudden power increase: (choose ONE answer below)
(1.0)
- a. The fuel centerline temperature will reach its maximum (final)
- value,
- b. Clad surface temperature will reach its final value.
c.
Fuel centerline temperature will reach approximately 2/3 of its final value.
- d. Fuel centerline, clad and coolant temperature have reached their final values.
e.
Clad surface temperature will reach approximately 63% of its final value.
QUESTION 1.13 (2.50)
A reactor startup is in progress (actual control rod movement).
How would each of the following conditions or events affect the ACTUAL critical rod position?
(more rod withdrawal, less rod withdrawal, or no significant change)
A. One recirculation pump is ssopped (with no change in heat losses)
(0.5)
B. Xenon is changing due to extended power operation terminated 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> earlier.
(0.5)
C. Shutdown cooling (Decay Heat Removal) is secured 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after extended power operation was terminated.
(0.5)
D. Moderator temperature is gradually decreasing.
(0.5)
E. Reactor vessel water level is raised 12 inches.
(0.5)
(***** END OF CATEGORY 01 *****)
(Q m
.2.
PLANT DESIGN INCLUDn4G SAFETY AND EMERGENCY SYmEMS PAGE 6
QUESTION 2.01 (3.00)
In the performance of a test closu.re on a MSIV, DESCRIEE what takes place when the test pushbutton is momentarily depressed AND when it is held in.
(Include in your discussion any differences between normal and test operation of the MSIV's i.e.
speed, distance traveled etc.)
QUESTION 2.02 (2.00)
A reactor feed pump trips, list four (4) conditions required for another reactor feed pump to auto-start (Include setpoints where applicable) ?
E QUESTION 2.03 (3.50) a.
List five (5) conditions that cause an automatic closure of the isolation valves in the Reactor Water Cleanup System?
(Setpoints are not required)
(5 @ 0.5 ea.)
b.
Valve 7 (System return isolation valve) closes on which TWO auto isolation signals ?
(1.0)
QUESTION 2.04 (2.50)
How does the core spray syst.em., piping break detection system comfirm the integrity of the core spray piping?
Include in your answer what section of piping is monitored, Sensing points, normal readings, AND how the readings change when a break is detected.
QUESTION 2.05 (3.00)
A. What are three (3) sources of makeup water to the shell side of the Isolation Condenser?
(1.5)
B. What is the relationship of the Isolation Condenser System to the following electrical supplies?
1.
250 VDC Distribution
(.75)
- 2. 480 VAC Bus 28 & 29
(.75)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
PLANT DESIGN INCLOD]YdG SAFETY AND EMERGENCY SYLiEM i
G PAGE 7
.2.
QUESTION 2.06 (3.00)
INDICATE if the following statemen.ts regarding the Fire Protection System are TRUE or FALSE.
If False, briefly explain why?
A. The ten wall-mounted hose-reel CARDOX assemblies located throughout the plant each have a seperate CO2 storage tank.
(0.5)
B. Although the Service Water Tie Line Valve (M02-3906) is locked in the closed position, pressure in the Fire Protection Water System is normally maintained by the Service Water System.
(0.5)
C.
In the automatic mode of operation of the Halon Suppression l
System, an activation signal turns on the evacuation lights, sounds a siren and immediately commences a 3 minute injection of Halon.
(0.5)
D. Once the Halon system begins to inject in the Aux. Electric Room, it is possible to secure the injection with an abort switch located adjacent to the door to the Aux. Electric Room.
(0.5)
EXPLANATIONS of FALSE STATEMENTS (1.0) d QUESTION 2.07 (3.00)
For each of the following components of the OFF-GAS System:
- 1. Catalytic Recombiner
- 2. Charcoal Adsorber A. Briefly explain the purpose of each.
(2.0)
B. Indicate whi.ch OTHER component in the OFF-GAS train, that IMMEDIATELY PRECEDES each?
(1.0)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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.2. ' PLANT DESIGN INCLUDrdG SAFETY AND EMERGENCY S E EMS PAGE 8
QUESTION 2.08 (2.25)
A. During normal HPCI oper ttion (with the system running) what provides turbine lube oil cooli'ng water pressure?
(.5)
B. During overspeed testing of the turbine, what provides turbine lube oil cooling water pressure?
(.5)
C. How many lube oil pumps are associated with the HPCI system what is the motive force for each (type of power; AC, DC, pneumatic, etc)?
(1.25) 1 QDESTION 2.09 (1.00) 3 List four (4) reactivity effects the standby liquid control system must (4 9 0.25 ea.)
overcome.
QUESTION 2.10' (1.00)
What core design component prevents " flow starving" the higher power fuel bundels ?
(1,g) 4 1
I
(***** END OF CATEGORY 02 *****)
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p PAGE 9
3.
- INSTRUMENTS AND CONW.0LS QUESTION 3.01 (2.00)
Describe the control action of the recirculation flow control system, if both recirculation pumps were o'perating at 60% when an operator closes the discharge valve on th
'B' recirculation pump.
(Assume normal system line up and operation)
QUESTION 3.02 (3.00)
Describe how the EHC Pressure Control and Logic System would respond if while operating at 100% power, an operator slowly reduces the Load Limit setting to zero.
Take your discussion to a final steady state condition.
(Assume no other operator action and all systems are in their normal full power lineup.)
QUESTION 3.03 (3.50)
Answer the following question in regard to LPCI loop select logic:
A. HOW does the logic determine how many recire pumps are running? (NOTE: Include setpoints where applicable)
(1.5)
B. HOW does the logic determine which is the UNDAMAGED recire locp?
(1.5)
If the logic determine't'at neither loop is damaged, WHICH h
C.
LOOP WILL IT SELECT for LPCI injection?
(.5)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
,3. ~ INSTRUMENTS AND CON OLS PAGE 10 QUESTION 3.04 (2.00)
A. JLn operator takes the control switch of an Electromatic Relief Valve to the " MANUAL" position.~. Select the correct statement below.
(1.0)
- 1. The 250 VDC solenoid assembly energizes and ports instrument air to open the relief valve through the pilot valve assembly.
- 2. The 125 VDC solenoid assembly de-energizes allowing the pilot valve to assist in opening the main relief valve.
- 3. The 250 VDC solenoid assembly de-energizes which actuates a plunger which positions the relief valve to open.
- 4. The 125 VDC. solenoid assembly energices and actuates a plunger which positions the pilot operating lever on the pilot valve assembly.
B. How may a safety relief valve be shut once it has opened on an ADS initiation? (give two (2) situations / conditions).
(1.0)
QUESTION 3.05 (2.50) a.
The Emergency Diesel Generator (EDG) control switch on the 902-8 panel is in AUTO.
List the PARAMETERS AND SETPOINTS which will automatically start the EDG, (3 REQUIRED)
(1.5)
- b. Energizing the " Fast Start Relay" prevents what EDG AUTO-SEUTDOWNS?
(4 REQUIRED)
~ '
(1.0)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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.3. ' INSTRUMENTS AND COhcROLS J
PAGE 11 QUESTION 3.06 (2.00)
Given the following conditions and the attached EHC logic diagram:
~
Reactor power = 30%
Generator output = 30%
Reactor pressure = 934 psig Throttle pressure = 929 psig Pressure regulator setpoint = 920 psig Recirc pumps in MANUAL at 28% speed Load set = 30 Load Limit 100%, Maximum combined flow = 105%
Turbine speed = 1800 rpm 4
The generator output circuit breakers (OCB's) are suddenly opened manually.
A. Has a LOAD REJECTION occurred? Justify your answer.
(1.0)
B. Which of the below subsystems of the EHC Pressure Control and Logic System is the first to produce an error signal when the OCB's open? EXPLAIN.
(1.0)
- 1. Pressure Control Unit
- 2. Bypass Control Unit
- 3. Load Control Unit
- 4. Speed and Acceleration Control Unit
- 5. Valve Flow Control Unit QUESTION 3.07 (1.00)
- a. How are the 4 KV electrical circuit breakers tripped on a loss of 125 VDC control power ?
- b. The red / blue lights on the 4 KV circuit breaker control room control panel indicates WHAT two conditions exists ?
QUESTION 3.08 (2.00)
List four conditions (or signals) and SETPOINTS that cause an automatic closure of the MSIV's. (Assume the MODE switch is in "STARTUP")
I (v**** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
e ',
".3. ' INSTRUMENTS AND CObdOLS PAGE 12 QUESTION 3.09 (1.00)
When are all SRM rod blocks automa~tically bypassed (2 required)?
(1.0)
QUESTION 3.10 (1.00)
The Unit 2 and Unit 3 control rod drive water pressure are not equal by procedure.
What is the delta-P BETWEEN UNITS ?
QUESTION 3.11 (1.00)
TRUE or FALSE:
All IRM Scram functions are bypassed if the MODE switch is placed in "RUN".
QUESTION 3.12 (1.50) 4 List three (3) instruments or indications that use the Standby Liquid i
Control injection sparger line as an instrument tap.
(3 @ 0.5 ea.)
QUESTION 3.13 (3.00) a.
The Rod Worth Minimi::cr System uses what inputs for reactor power?
(1.0) b.
What are the requirements to automatically bypass the RWM blocks? (1.0) c.
What is the significance of the TRANSITION ZONE?
(Include power levels and RWM functionc)
(1.0)
(***** END OF CATEGORY 03 '*****)
p 3
.4.
PROCEDURES - NORMAL ABNORMAL. EMERGENCY AND V
PAGE 13 RADIOLOGICAL CONTROL QUESTION 4.01 (4.00)
The Unit 2 plant is operating at-8'5% power, for some reason the Main Generator field breakers trip open which results in a turbine trip and a reactor scram.
WHAT are the 7 required operator immediate action steps.
(Be specific)
(An action step may have more than one action item.)
QUESTION 4.02 (2.25)
A. Under what conditions can the NSO inject SBLC without
^
authorisation from a supervisor?
(1.0) 1 B. After the Standby Liquid Control System is initiated, WHAT are the five (5) indications that the system is operating?
(1.25)
QUESTION 4.03 (2.50)
When paralleling electrical sources the synchroscope should be rotating A
in the B
direction and C
voltage should be slightly D
than the E
voltage.
(Fill in the Blanks)
QUESTION 4.04 (3.00)
In accordance with the Radiation Protection Standards, WHAT are six (6) cf the eight (8) conditions / situations that would require a worker to leave a controlled area as quickly as possible, t
consistent with safety?
OUESTION 4.05 (1.50)
In accordance with the DFP-800-1, How many SRM's are required AND where are they required to be inserted for a fuel movement?
(1.5)
\\
l
\\
l
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
.---,.m
--m
',4.
PROCEDURES - NORMAL ABNORMAL. EMERGENCY AND PAGE 14 RADIOLOGICAL CONTROL QUESTION 4.06 (3.00)
Briefly discribe six (6) of the ni'ne (9) different means to initiate control rod movement in the event a scram did not occur identified in DEOP-100-3 after you depressed both manual scram push buttons and placed the Mode Switch in the Shutdown position.
(6 @ 0.5 ea.)
QUESTION 4.07 (3.00)
For the following parameters indicate what you would expect the indications to be at high power operations during your Normal Control Room inspection.
(Assume 95% Rated Power with all systems in their normal lineup.)
A. Panel 902(3)-3:
- 1. Torus water level (Give a range).
(0.5)
- 2. LPCI & Core Spray discharge pressure.
(0.5)
B. Panel 902(3)-6:
- 1. Contaminated Condensate Storage Tanks (Give a minimum level) (0.5)
- 2. Feedwater Heater operating vents (Give a position: OPEN, CLOSED, THROTTLED)
(0.5)
C. Panel 902(3)-7:
- 1. Steam Seal Header pressure (0.5)
- 2. Exhaust Hood Spray Valve (Give a position: OPEN, CLOSED, THROTTLED)
(0.5)
QUESTION 4.08 (1.00)
Operating Abnormal Procedure RECIRCULATION PUMP TRIP (DOA-202-1) reduces the speed of the operating recirc pump to what value for each Unit?
QUESTION 4.09 (1.50)
List 3 of the 4 indications or auto actions DOP-2300-1 describes as cccuring upon an AUTO trip of the HPCI pump / turbine when in operation.
(1.0)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
^g c
'4. PROCEDURES - NORMADr ABNORMAL. EMERGENCY AND J
PAGE 15 RADIOLOGICAL CONTROL QUESTION 4.10 (1.50)
In accordance with Operatin Order '#6-85,
Unit 2 and 3: Drywell Equipment Floor Drain Pumping and Leakage Limitations., it states When an equipment or floor drain sump high level alarm occurs, determine radiological conditions, using the criteria below prior to pumping down the sumps.
MATCH the plant conditions in part
'A' with the conditions that must exist prior to pumping, in part
'B',
i.e. 4,d PART
'A' 1.
With the Reactor Mode Switch in Run or Startup with steam flow to the turbine or condenser i.e.
generator on line or bypass valve open:
i 2.
When in Refuel or Shutdown when reactor pressure is above atmospheric:
3.
When in Startup or Hot Standby with no steam flow to the condenser:
PART
'B' n.
Reactor water level above -59" when irradiated fuel is in the vessel or check Drywell Cam for normal activity.
b.
Verify reactor water level, Drywell pressure and main steam line Rad Monitors indicate normal condition or check Drywell Cam for normal activity.
c.
Reactor water level above -59",
Drywell pressure normal, i.e.
less than 2 psig or check Drywell Cam for normal activity.
(3 @ 0.5 ea.)
1
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
11.
PRIkCIPLES OF NUCL POWER PLANT OPERATION, PAGE 16 THEFh0 DYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 1.01 (1.00)
As the reactor operates at power, Xenon builds in to equilibrium, adding negative reactivity, causing power to decrease.
(1.0)
REFERENCE DRESDEN - General Physics BWR RX Theory ANSWER 1.02 (2.00)
A. Alpha V increases power B. Alpha T increases power C. Alpha D decreases power D. Alpha V decreases power (4 @ 0.5 ea)
REFERENCE DRESDEN - GE BWR Transient Analysis ANSWER 1.03 (2.50)
As flow is increased coolant flow past the fuel elements tends to sweep away the voids more rapidly than they are formed.
This causes a positive reactivity addition by the void fraction of reactivity.
[0.5]
Power level begins to rise immediately, which increases the fuel element temperatures.
This implies more heat transfer to the coolant, thus increasing the steam generation rate.
[0.5]
With the power increase, the void fraction increases.
[0.5]
4 Reactor power increases until the negative reactivity insertion rate due to increasing void fraction and Doppler coefficient overcomes the positive reactivity still present in the core.
[0.5]
The power level steadies out with net reactivity equal to zero, and void fraction near its original value.
[0.5]
REFERENCE DRESDEN -
Lesson Plan Book 4, Chapter 12, page 52 (rev. 10)
O.
/]-
PAGE 17 cl.. PRINCIPLES OF NUCLEd POWER PLANT OPERATION.
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- DRESDEN 2&3
-86/01/28-KING M..
ANSWER 1.04 (2.00)
EOL (0.5)
Because B eff is at its minimum value.
(0.5)
Prompt criticality is achieved when added reactivity exceeds B eff.
(1.0)
REFERENCE Dresden Reactor Theory ANSWER 1.05 (1.50)
- 1. They are physically located as far below the normal water line as possible to provide the greatest static head.
- 2. With feed flow less than 20% they are kept on minimum speed.
4.
Low Reactor '.'essel water level trip, cavitation interlock.
- 5. Suction valve closed trip, cavitation interlock.
- 6. Pressure in the steam dome.
(3 9 0.5 ea)
REFERENCE DRESDEN - Recire System Lesson Plan pg 16 & 18 GE Thermodynamics, Heat transfer & Fluid Flow, page 7-93 & 94 ANSWER 1.06 (2.00)
When the SRV opens plant pressure will decrease [0.20). Power starts decreasing [0.20] due to increased voiding [0.20], MSIV's close on low header pressure [0.20] resulting in a Rx Scram [0.20].
Vessel level will fluctuate up due to increased voiding then down after scram due to void collapse [0.20]. Plant continues a rapid cooldown[0.20] until SRV closes [0.20].
Steady state condition will be, Rx S/D with press & temp steadily increasing due to decay heat [0.2].
Feed to the reactor from RFP's with level at normal [0 20].
(2.0)
REFERENCE Dresden DGP 1-1 Unit 2/3 Normal Unit Startup; Reactor Theory
1.'
PRINCIPLES OF NUCLE POWER PLANT OPERATION, PAGE 18 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 1.07 (1.00)
Neutron level would start and continue to increase until the point of adding heat is reached.
As the coolant heats up, negative reactivity is added and power turns.
Power would stablise at the point of adding heat.
(1.0)
REFERENCE DRESDEN NUS Theory 1
ANSWER 1.08 (2.50)
A.
Area 1 -- Decrease in core inlet subcooling due.to loss of feed flow to vessel.
Area 2 -- Reactor scram on low level.
B.
Area 3 -- HPCI injection.
C.
Area 4 -- Recire pump runback on low feed flow.
Area 5 -- Recire pump trip on low-low level (5 @ 0.5 ea)
REFERENCE 1
BWR Transients EXY-6.
ANSWER 1.09 (1.50)
A. CONTROL CELL: one control'rdd and four surrounding fuel bundles.
(1.0)
B. This strategy eliminates the need for periodic control rod
.-s pattern changes during the operating cycle.
}( 1. 21),)
REFERENCE CONTROL ROD BLADE AND DRIVE MECHANISM LP, pg 4
1.
PRINCIPLES OF NUCL POWER PLANT OPERATION, PAGE 19 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 1.10 (3.00)
A. The pump will eventually add a sufficient amount of heat to the fluid to cause cavitation. (Also: Will accept overheating of the (1.0) pump.
B. Cause excessively long starting currents or water hammer if the downstream piping was not filled.
(1.0)
C. Causes excessive motor amps to be drawn and the high current could cause damage to the motor windings.
(1.0)
REFERENCE DRESDEN - GE THERMO HT & FF pg 7-123, 124 ANSWER 1.11 (3.00)
A. Decrease (.5) due to increased void content in the core as flow decreases (.5).
(1,0)
B. Increase (.5) due to increased voiding in the core (.25) and recire pump no longer taking a suction on the annulus (.25)
(1.0)
C. Decrease (.5) due to level incerase (.5)
(1.0)
REFERENCE BWR TRANSIENT ANALYSIS ANSWER 1.12 (1.00)
(1.0) e.
REFERENCE GE THERMO HTX & FF, pg 9-102
1l PRINCIPLES OF NUCLE POWER PLANT OPERATION.
PAGE 20 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 1.13 (2.50)
A. No significant effect B. Less rod withdrawal C. More rod withdrawal D. Less rod withdrawal E. No significant effect (5 @ 0.5 ea)
REFERENCE DRESEN - Lesson Plan Book 4, Chapter 12 i
G
2.'
PLANT DESIGN INCLUDr G SAFETY AND EMERGENCY SY EMS PAGE 21 ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 2.01 (3.00)
If the test pushbutton is momentarily depressed, the AC & DC solenoid valves de-energize and the test solenoid valve is energized to supply air to close the valve. The MSIV then slowly closes until the 90% open limit switch opens.'
The solenoid valves then reposition and air is supplied to open the MSIV.
(1.5)
If the test pushbutton is held in, the same occurs (above) except the 1
90% open limit switch is bypassed and the MSIV will continue to close and be closed in 45-60 seconds. (Closing time not required for credit.)
(1.5)
REFERENCE DRESDEN - Lesson Plan Book 3, Chapter 3, page 13 ANSWER 2.02 (2.00)
- 1. Standby RFP must be selected
- 2. Suction press must be greater [0.25] than 120 (+/-12) psig[0.25]
- 3. Vent fan must be on
- 4. Oil press [0.25] must be greater than 20 (+/-2) psig[0.25]
- 5. Water level [0.25] <55 inches [0.25]
(**** ANSWERS 2,4, & 5 REQUIRE THE SETPOINT FOR FULL CREDIT ****)
l (4 @ 0.5 EA.)
REFERENCE j
DRESDEN - Lesson Plan Book 2, Ch 4 (rev 7) i ANSWER 2.03 (3.50) a.
- 1. Auxiliary pump high cooling water outlet temp
- 2. High temp downstream cf the Non-Regen heat exchanger
- 3. High press downstream of the press reducing station 4.
Low flow through the filter in service (normally bypassed)
- 5. Low reactor water level
- 6. SBLC activated (5 @ 0.5 ea)
- b. 1. Water Level [0.5:
- 2. SBLC activation [0.5]
(1.0) d 4
-,,4
- m
O a
.2.
PLANT DESIGN INCLUD'roG SAFETY AND EMERGENCY SYS_dMS PAGE 22 ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
REFERENCE DRESDEN - Lesson Plan Book 3, Chapter 4, pages 21 & 22 (rev. 10)
ANSWER 2.04 (2.50)
The piping MONITOURED is that between the reactor vessel penetration and the shroud [0.5].
Pressure is sensed above the core plate [0.5] and at the spray sparger [0.5].
The indication is normally 3.2 PSID(@_100% pwr)[0,5].
When a break occurs, the dp will go to -2.7 psid (@ 100% pwr)[0.5].
(2.5)
REFERENCE DRESDEN - Lesson Plan Book 3, Chapter 13, pages 8 & 9 (rev. 4)
ANSWER 2.05 (3.00)
A.
- 1. Clean demineralized water system
- 2. Condensate transfer system
- 3. Service Water or Fire Protection (3 @ 0.50)
B.
- 1. 250 VDC-powers motor operated outboard isolation valves (1301-2i-3) [0.75]
- 2. 480 VAC-powers motor operated inboard isolation valves (1301-1,-4) [0.75]
(1.5)
REFERENCE DRESDEN - Lesson Plan Book 3, Chapter 12, pages 3 & 5 (rev. 10)
ANSWER 2.06 (3.00)
A. False [0.5].
The entire CARDOX system shares a common CO2 storage tank [0.5].
(1.0)
B. True (0.5)
C. False [0.5].
Activation signal starts a 60 second predischarge timer [0.25].
When the timer is timed out, the injection starts (0.25]
(1,0)
D. True (0.5)
q q
a
.2.
PLANT DESIGN INCLUDrdG SAFETY AND EMERGENCY SYb.dMS PAGE 23 ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
i REFERENCE Fire Protection L.P.,
Rev.7, pg. 3,7,12,16,17 ANSWER 2.07 (3.00)
A.1. Purpose is to recombine Hydrogen and Oxygen gases into a water vapor at a high temperature [0.5] to reduce Hydrogen concentra-tion to < 1% by volume [0.5].
(1.0)
(1.0)
B.
1.
Preheaters (0.5)
- 2. Prefilters (0.5)
REFERENCE DRESDEN - Lesson Plan Book 2, Chapter 16, pages 5 & 8 (rev. 10)
ANSWER 2.08 (2.25)
A. HPCI booster pump
'(.5)
B. aux cooling water pump
(.5)
C. 3[.25] one shaft driven [0.33]
two DC powered [0.66)
(1.25)
REFERENCE DRESDEN - Lesson Plan Book 3',' Chapter 10, pages 9-13 (rev. 7)
ANSWER 2.09 (1.00)
Elimination of steam voids Moderator temperature change (hot to 125 F)
Reduced doppler effect Decreased control rod worth Xenon decay Maintain SEM of 3% (overcome excess reacitivty)
(4 @ 0.25 ea.)
REFERENCE DRESDEN Lesson Plan Book 3, Ch 3, rev 11, pg 10 & 11
o i
.2.'
PLANT DESIGN INCLUD'nfG SAFETY AND EMERGENCY SYS 2MS PAGE 24 l
l ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 2.10 (1.00)
The fuel. bundle (fuel support plece) orfice
'(1.0)
REFERENCE DPESDEN - Lessen Plan Book 1, Ch 1, page 23 (rev 12)
.3.
INSTRUMENTS AND CONhdOLS
/
PAGE 25 ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
l ANSWER 3.01 (2.00)
- 1. The limiter would try to reduce pump
'B' speed to 28%.
- 2. The mismatch circuit would stop the scoop tube insertion at 40% (10% mismatch).
- 3. The
'B' pump would be running at 40% with the discharge valve shut (pump damage could result).
- 4. The mismatch circuitry trips the low speed pump when the discharge valve goes shut [<90% open].
(4 @ 0.5 ea) s REFERENCE l
Dresden Recire Flow Cntrl Sys Lessen Plan pg 13 l
ANSWER 3.02 (3.00)
As the load limit decreases the control valves will start to close and the bypass valves will start to open, after the snall close bias signal is overcome.
At approximately 60% turbine load and bypass valves at 100% open, 40% steam flow, the header pressure will start to increase.[1.0]
As the load limit is futher reduced reactor pressure and power will start to increase.
The reactor j
will scram on high pressure or high power to flow.
The isolation condenser may or may not initiate.[1.0]
The final steady state conditions will be the reactor shutdown with pressure being main-i tained with the bypass valves and makeup with main feed.[1.0]
(3.0)
REFERENCE Dresden RFS, Recire Cntrl & EHd Logic Lesson Plans i
l ANSWER 3.03 (3.50)
A. By monitoring the differential pressure (.5) across each recirc pump (.5) for >2 psid, indicating the pump is running (.5)
(1.5)
B. By comparing the pressure in the riser pipes on one recire loop with the pressure in the riser pipes of the other loop. The undamaged loop will have a higher pressure than the damaged loop.
(1.5)
C. Loop B
(.5) l l
1 1
I
_ - _. ~ _
4
.3 i INSTRUMENTS AND CONTROLS PAGE 26 ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
REFERENCE DRESDEN - Lesson Plan Book 3, Ch 14, pg 15 (rev 5)
ANSWER 3.04 (2.00)
A. 4 (1.0)
B.
- 1. Depress timer reset button f
- 2. Steam pressure < 150 psig so it is over come by valve spring pressure
- 3. Place the manual inhibit switch to " INHIBIT" (any 2 9 0.5 ea = 1.0) i REFERENCE DRESDEN - ADS LP, pg 3,10 Lesson Plan Book 3, Chapter 11, page 13 and fig. 7 I
I ANSWER 3,05 (2.50)
- o. Drywell Pressure [0.25] 9 2.0 psig [0.25]
Reactor Water Level [0.25] 9 -59" [0.25]
4 Respective bus undervoltage [0.25] 9 (CAF) [0.25]
(1.5) i i
- b. Low Oil Pressure Low Water Pressure High Water Temperature Positive Crankcase Pressure.,
(4 9 0.25 ea.)
l REFERENCE DRESDEN - Lesson Plan Book 3, Chapter 15, page 9 (rev. 15) 1 ANSWER 3.06 (2.00)
]
A. NO(.5) Load rejection is caused by a 40% mismatch between the first stage pressure and stator amps (.5).
(1.0)
B. 4(.5) When the OCB's open the turbine is virtually unloaded and j
begins to overspeed. The speed and acceleration unit senses the i
overspeed and produces a speed error signal (.5)
(1.0) i I
(
l 4
m.
3.'
INSTRUMENTS AND CO T OLS PAGE 27 ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
REFERENCE EHC LOGIC LP, REV 7, pg 25 ANSWER 3.07 (1.00) a.
Local, at the breaker with a manual pushbutton, b.1 The breaker is closed [0.25].
.2 There is continuity thru the trip circuit (coil) [0.25].
REFERENCE DRESDEN Lesson Plan Book 2,.ch 14, rev 10, pg 11 & 13.
ANSWER 3.08 (2.00)
Low-low reactor water level [0.25] @ -59 (+/- 2) inches [0.25]
(0.5)
Main Steam Line high radiation [0.25] @ 3 * (normal 100% lvl) [0.25]
(0.5)
Steam tunnel high temperature [0.25] @ 200 (+/- 10)F [0.25]
(0.5)
High Steam Flow in any main steam line [0.25] @ 120 (+/- 5)% [0.25]
(0.5)
REFERENCE DRESDEN Lesson Plan Book 2,.Ch,3, rev 14, Main Steam, pg 15.
ANSWER 3.09 (1.00)
- 1. All IRM range switchs are on or above range 8 [0.5].
- 2. Mode switch is in "RUN" [0.5].
(1.0)
REFERENCE DRE3 DEN - Lesson Plan Book 1, Chapter 3, page 26 (rev. 13)
.3 i INSTRUMENTS AND CONT OLS PAGE 28 ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 3.10 (1.00)
~
Il
- a. Unit 2
= 250 paid
~
Unit 3
280 psid Delta-P
30 psid (1.3)
REFERENCE DRESDEN DOP-300-1, rev 7, pg 3 ANSWER 3.11 (1.00)
False INFORMATION ONLY (IRM upscale / Comp. APRM downscale trip is active.)
REFERENCE 1
DRESDEN Lesson Plan Book 1, Ch 10, rev 12.
{.
ANSWER 3.12 (1.50)
+
Core plate differential pres's. measurement Core spray line break detection Jet pump diff. press. tap (3 @ 0.50 ea.)
REFERENCE DRESEDEN Lesson Plan Book 3, Ch 3, Rev 11, pg 3
. ~ _
INSTRUMENTSANDCON)ROLS
')
(
PAGE 29 T
3:
ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 3.13 (3.00) a.
total steam flow and total feed flow (1.0) b.
Steam flow > 20% [0.33] and feed flow > 10% [0.33] for at least 60 seconds [0.33]
(1.0) c.
The TRANSITION ZONE is the operating region between the LPSP and the LPAP.
Expressed in % power, it corresponds to power levels between 20% and 35% [0.5]
4 When operating in the transition zone all RWM rod block actions are removed, but the displays remain active [0.5]
(1.0) f REFERENCE DRESDEN - Lesson Plan Book 1, Chapter 13, page 7 (rev. 11) i 4
1 i
+
.4 ?
PROCEDURES - NORMA ABNORMAL. EMERGENCY AND PAGE 30 RADIOLOGICAL CONTROL ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 4.01 (4.00)
(Generator trips the Turbine, Turbine scrams the Reactor when greater than 48% load.)
1.
Press both scram buttons and place the mode switch to SHUTDOWN.
2.
A.
VERIFY all rods are inserted to or beyond 04.
I Record all rods not at 00 B.b nsert any rod not already at 00.
Y#
and notify the Shift Supervisor.
l 3.
A.
Attempt to maintain level between +8 and ~+40 inches by j
multiple indication.
l Control feedwater in AUTO unless control failure occurs.
B.
If necessary, STOP the feed pumps to pervert excessive reactor water level.
positih04,thenverifythatthe 4.
IF all rods inserted to or eyo turbine and generator have t pped.
5.
VERIFY that recirculation pumps A & B run back to the minimum pump speed.
VERIFYthaft the auxiliary power has transfered to the reserve 6.
auxiliary wransformer.
7.
Insert SRM/IRM's.
Maintain the IRM's on scale to monitor shutdown.
(7 9 0.44 ea)
[ge/
REFERENCE DRESDEN - DGP 2-3 (gf4fCl l
y a?f fk ad
PROCEDURES - NORMAL'T) ABNORMAL. EMERGENCY AND
(
.4 i PAGE 31 RADIOLOGICAL CONTROL i
ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M ANSWER 4.02 (2.25)
A.
1.
All control rods are not iserted to or beyond position 04 [.05]
and torus temperature cannot be maintained below 110F.[0.5]
l
- or **
2.
If the reactor cannot be shutdown before the torus temperature reaches 110 F.
Shutdown = All rods are at or beyond position 04, or as determined by a qualified nuclear eng.
EITHTER ANSWER 1 OR ANSWER 2 WITH AT LEAST ONE DEFINITION OF SHUTDOWN IS REQUIRED FOR FULL CREDIT.
(1.0)
- 3. 1 Amber pilot light of squib firing continuity circuit not lit.
2 Flow indication pilot light lit 3 Water clean-up system isolation.
4 Decreasing level of SBLC storage tank.
5 SBLC valve circuit fail announciator light lit.
6 Pump discharge pressure increases.
(5 9 0.25ea)
REFERENCE DRESDEN -
DEOP - 100-3, page 3, rev. O.
DEOP - 400-2, page 2, rev. O.
SBLC Lesson Plan pg 14 ANSWER 4.03 (2.50)
A. Slow B. Fast (clockwise)
C.
Incoming D. Higher E. Running (5 @ 0.5 ea)
REFERENCE Dresden DGP 1-1 pg 19 of 26 steps 5.1 & 5.m i
1 l
l
9 q
4 '.
PRO _CEDURES - NORMAL'. ABNORMAL. EMERGENCY AND PAGE 32 RADIOLOGICAL CONTROL ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 4.04 (3.00)
- 1. When instructed or signaled to do so by the Rad-Chem department.
- 2. Failure or suspected failure of personal protective equipment.
- 3. Unexpected deterioration of radiological conditions.
- 4. In the event that the workers current accumulated dose equivalent status becomes uncertain for any reason or dose equivalent is equal to the exposure authorised for the job.
- 5. " Assembly" sirens sounds (practice or actual)
- 6. Completion of work assignment.
7.
Injury.
- 8. Unexpected area radiation monitor alarm and the area dose rate is unknown.
(6 @ 0.5 ca)
REFERENCE Dresden Radiation Protection Standards pg 8 of 69 ANSWER 4.05 (1.50)
I 2 SRM's, [0.5] one in the core quadrant where fuel or control rods are being moved [0.5] and one in an adjacant quadrant. [0.5]
(1.5)
REFERENCE DRESDEN - DFP-800-1, page 4, step 4 (rev.11) l
n n
d
'. )
4:
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 33 RADIOLOGICAL CONTROL ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 4.06 (3.00) 1.
Verify that all scram valves are open as indicated by blue scram lights on the full core display, if not remove the fuses behind panels [902(3) - 15 % 17]
2.
Arm the ARI pushbuttons and depress the bushbuttons to open the ARI valves, hold for 20 seconds.
3.
If not yet scrammed replace the fuses and close the ARI valves, then reset the scram and initiate a manual scram.
4.
Reset the scram and individually open the scram test switch for each rod > 04.
5.
If scram cannot be reset, start both CRD pumps, close the charging water valve, and try inserting using the emergency rod in control switch.
6.
Insert any not at 04 or beyond by venting the drive over piston area thru a hose to the floor drain for each rod.
REFERENCE DRESDEN -
DEOP 100-3, step 6, page 3, 4,
5, 6, & 7.
ANSWER 4.07 (3.00)
A.
1.
-1.5 to -5.5 inches (0.5)
- 2. ~85 psig (0,5)
B.
1.
>9 feet (0.5)
- 2. Open (0,5)
C.
1.
4 psig (0.5)
- 2. Closed (0.5)
REFERENCE Dresden DGP 3-2, Normal Control Room Inspection, Rev 2 pg 1-3 ANSWER 4.08 (1.00)
UNIT 2 43%
UNIT 3 EO%
(2 @ 0.5 EA.)
REFERENCE DRESDEN DCA-202, PAGES 1 AND 4
PROCEDURES-NORMAL'). ABNORMAL.EMERGENCYAND
']
(
o
~
4?
PAGE 34 RAD'IOLOGICAL CONTROL ANSWERS -- DRESDEN 2&3
-86/01/28-KING, M.
ANSWER 4.09 (1.50)
Pump discahrge valve closes (MO-2301-8).
Motor speed changer is positioned to low speed stop.
Turbine control valve closes.
Turbine reset light ON (indicating ability to reset stop valve)
(3 @ 0.5 ea.)
REFERENCE DRESDEN - DOP-2300-1 ANSWER 4.10 (1.50) 1 b
2 a
3 c
(3 @ 0.5 ea.)
REFERENCE DRESDEN Operating Order #6-85 pg 1 of 3 July 1, 1985 O
1 O
Q'
\\
TEST CROSS REFERENCE PAGE 1
{
l QUESTION VALUE REFERENCE 01.01 1.00 KZS0000691 01.02 2.00 KZS0000692 01.03 2.50 KZS0000693 01.04 2.00 KZS0000702 01.05 1.50 KZS0000703
~.
01.06 2.00 KZS0000704 01.07 1.00 KZS0000705 01.08 2.50 KZS0000706 01.09 1.50 KZS0000710 01.10 3.00 KZS0000711 01.11 3.00 KZS0000712 01.12 1.00 KZS0000717 01.13 2.50 KZS0000719 25.50 02.01 3.00 KZS0000698 02.02 2.00 KZS0000699 02.03 3.50 KZS0000700 02.04 2.50 KZS0000701 02.05 3.00 KZS0000714 02.06 3.00 KZS0000715 02.07 3.00 KZS0000716 02.08 2.25 KZS0000724 02.09 1.00 KZS0000733 02.10 1.00 KZS0000735 24.25 03.01 2.00 KZS0000708 03.02 3.00 KZS0000709 03.03 3.50 KZS0000713 03.04 2.00 KZS0000718 03.05 2.50 KZS0000720 03.06 2.00 KZS0000721 03.07 1.00 KZS0000727 03.08 2.00 KZS0000728 03.09 1.00 KZS0000729 03.10 1.00 KZS0000730 03.11 1.00 KZS0000731 03.12 1.50 KZS0000732
(
03.13 3.00 KZS0000734 l
25.50 l
04.01 4.00 KZS0000694 "'
5 04.02 2.25 KZS0000695 04.03 2.50 KZS0000696 04.04 3.00 KZS0000697 04.05 1.50 KZS0000707 l
l
d o
e'.
TEST CROSS REFERENCE PAGE 2
QUESTION VALUE REFERENCE 04.06 3.00 KZS0000722
~
04.07 3.00 KZS0000723 04.08 1.00 KZS0000726 04.09 1.50 KZS0000736 04.10 1.50 KZS0000737 23.25 98.50 i
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NUCLEAR REGULATORY COMMISSION SENIOR REACTOR CPERATOR LICENSE CXAMINAT!ON Fr.CIL ITY :
D r., E 9 D E. M_ M ~.
C
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95/0t/C CATE ADMINISTERED:
--.ING
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EXAMINER:
K 1
APPLICANT:
.= :._ =..u = _:-= _ _ a -- -: -_ _ 22_:u.8,,6 J.:,c M _..:.
T T p-c : 'r
., c Tr t
I t
l f
U se :cpArcte paper for the answers.
Write answere en one side only.
S t.2p l a questien abeet en tcp cf the answer chaetc.
Points'for each q.;.cction arc i nd i.c at e d in parenthcces after the qu es t i or.. The pancir.g
)
.; race require:
- 3. t least 7r2% in cach category and a final grade cf at leact 20%.
Examinaticn papers will be picked up cix (6) heurt after the c:: amination starts.
F
. a 7TEC-OR v
. OF 4Fo' ! CANT' S CATEGCRY
--- UE_. _
_-___________ CAT.EGCr_.y
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Um__Hf10EY_2E_MUCLEGE_ErdEE_ELQMI_.GEEEGIICM&_ELUlm /_GMD PAGE 2
T_ u F A M_ O D Y N_ A M _T C S
- . g-
- i. r.y a.01
(, 0.. m, e-u __
Fcr the Fcilcuing transients, indicate which COEFCICIENT cf raattivity; alpha T, alpha D, or alpha Y tends to change reactor power FIRST and in what DIRECTICN.
A.
Fast cicsure of one MSIV.
(0.5)
B.
Isolation of a feedwater heater ctring.
(0.5)
C.
A contral rod drcp.
(0.5)
D.
Pelief valve lifting snd then receating.
(0.5)
QUESTION 5.02 (2.50)
E:: plain WHY reactor power INCREAEES as racirculaticn flow is INCREASED.
Ensure your e::l ai n at i on continues to a STABLE CONDITION after the flow change.
< i r E e., t r. n u e. w-(
.,0)
>. e Oricfly explain which time in ccre life GOL, Mid-of-life, or EOL) requir2s the least amcunt of pcsitive reactivity addition to achieve "pr:mp' critical".
l ni'E.'.'.oN.
- 'a4
.' l. " C '
What are t n r rc- (2) of the deci r er cperaticnal factors that insur: adequatu Not Positive Eu.c t i on Head (NPSH) fer the r-c ircul ation pumes?
(3 9 0.5 ma.)
. ju-.- - ms.
e.ge
,..,n-s. w f
......v1 o'
w
'ih i l ? p C r
- C r T:i r g r3 rCCCtGr "* t a r t 'J,p and N S itIJp CSr DCP 1 - 1,,
0it 2.
%:rmal Un't S t ar t.;.p. the MCOE 3WI'CH Mac cear, p12c 2d
,.,c._s.,,
- s. _ _,.,
..-._..,.4.
m.e--t.
. 7. u. = c =. < r -_
- m.. =.o..
~
. _s..
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.=
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.n l accur during this t--
- r s i a r t.
'.s s e -nc v; 27C# Lt0F 3 c" i O1'r 1.r d th 2 / 91 V? r ' ! 3 ~. # 35 i p C r C "~ 3 01.U.Jt2G.
I
~~ "CILd2
- 1. n / Cur d O '5 C r i p ". i c n th? Of0Si =. ; Or p '. 10 " pcwtr. p r T 3 3 *.. " 9.
29'501 idve1 i#
.~tn y trip 2CCUP7 ard t h ed fi Scl 3t02dy 5tato 3
- s.... A.;
.-.I
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( t t ':$ 3 CATCCCRY 05 CCNTINUED ON NEXT PAGE **trt)
1
,]
FLUIDu]. AND PAGE 3
THtIDRV O r~ h4TLEAR frv. DER P!. ANT OPERATICN 9.
x----
s----
TFERMCDYNAMICS s
GUEET!ON 5.06 (1.00)
The reacter is e::actl y cr i ti c al LOh' i n the intermediate range.
A control rcd is withdrawn one notch.
Describe what happens to indicated neutron level AND nhy?
(Ccntinue description until a steady state conditico ic reached.
Accume no futher cperator acticn other than ranging the IRM meters.)
(1.0)
CUERTION 5.07 (2.00)
Unit 2 presently uses the CCRE CONTROL CELL concept.
A.
What is a CGNTROL CELL 7 (1.0) 9.
Why 12 this operating strategy used?
(1.0)
GUESTICN 5.08 (3.20)
Assume the rzactcr in apare. ting at 100% pcwer and cne recirculation pump trips. Ir.dicate hcw each 1 sted indicated parameter would first char.ge (Increase or Decrease) and bric41y 2xp12in why the change occur.
P.. rec.ctcr power (1.0) 9.
reactcr ; ster level (1.0)
C.
fecW,ater 31cw (1.0) a-
.v-
..u. - r..v.
- c-n e o. -I
.3. E.f.,.- e a:ac-r.., Js., y y.
.. i, ;s:n.c,
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+
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4 u
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9__. _ TH_E__OF Y "_F__M_U_C_L_E_A_R__W-::E_C P'._A_NT_ _ O_ P E R_ A_ T_.I C_!8 _ F._L_ U_ _I D ? _ A N_ D_
PAGE 4
T H_ E_ 4_td O_ D_ v N_ A_.M _I C E_
-,.s e,i u..
u.O~
(1.m.0,)
., np e
u
(
control red movement).
l A reactor ctartup in in progrecc'(actual i
1 1
Hcu wculd each of the following conditions er events affect the ACTUAL i
critical red position' (mcre red withdrawal, lecc red withdrawal, cr no significant change)
A.
One r7circulatien pump is stcpped (with ne change in heat lestes)
(0.5)
S.
Xenon i s changing due to c:: tended pcwer cparaticn terminated 25 hcurs earlier.
(0.5)
C.
Reactor vascel water level is raised 12 inchen.
(0.5)
(,.. m Ca.
e.4.3
~,E.
i t u,J m
a e
Re prding MCPR (Minimum Critical Pcwer Ratic):
a.
4 hat PHENOMENCN could e:t i st in a fuel fundle if it were cper2ted at a MCPR LEES THAN ONE
(<
l.0) and WHAT would very 1:kel f be the CONSECUENCE cf the phenomencn?
(1.0) 5.
W H's must the Technical Specificaticn MCPR limi t be mcdfied when ccre ficw is LEEE THAN RATED? (Include in your answer
.shether MCPR 12 incre12ed ce decreased. )
(1.0) u,r.er n.y x..1.
(
.oC/-
i y C h Q P'*
YCr i: C V O f* 3 f,*, d '/ 'i. h f~.
'?
- C.~* c I 3 " h3Y bCen cpOrIIiny OI -' '.
C'SICLV Oduccc 0 3 C *". a t
- p Ou *? r to d-Z . 5 '/ r '? d u c i '* Q t h :2 3pOed j
CcGP2tCr i
9 the r eci cul.c t :. cn pumpc. During the c e::t 2-3 MINUTE 5 the ape ator j
nacic22 that the reactcr pcwor sicwly increanas ap prca ' mat el y 2 7.. EXP EIN the C2uce cf thi5 SJ'ECt.
/2.0) c.-.I,i.xn.v J r. s, wy. pse
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...e
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w 1-
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- t. 4 T..$.T. )
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(m) 5_?_. _. r peppv._O._F__N_U_.f_'LE_A_R__c h_WE_P__P_L_ ANT _OP_EP_A_T__I C_M_L_F_L__LI T.D_
L_A_N_D.-
PAGE 3
.,._s__.
TtdE _ M O_ D_ Y_ N_ A M _T C. 3_
- e. 1,-
(,..,, n g a u._.T. f.f va,
.u a
~
Indic ate HOW each of the ccef fic-iento are effected CIncrease, Decreace or emain the came] by each of the three parameters listed?
Concider-each parameter separately.
2.
Rod Werth (delta K/K/ Bank) by:
1.
Moderatar temperature INCREASE 5 N C_'* R C 1 C C t <',, i. d..
t.
3.
Fuel temperature INCREASE 3
- 3 3 0.33 ca2 b.
Alph2 '/c t d s (dolta K/K/ . voids) by:
1.
Fuel temperature INCREASES 2.
Cer2 ac;c INCREASES 3.
Ccatrol Red Density INCREASES E3 Q 0.33 ca:
CL:ESTIGN 5.13 (C.00)
Actuming the rer.ctor has been operating at full power fcr an m:terdad pe-icd cf time when a scram cccurs.
Enssume time in core life is ECC2 Durin.; a restart eight hcurs after the scram:
.:.. Hcw will rod wcrth be af f ected? (overall cere)
(1.0) b.
Mcw. sill Radial and A:: i al #1ux distributicn be a4fected' (1.0)
W 'N..
'N.l
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?..f, C.
, Y.
Q " j. "N..~. ' \\
I vr us
-.r
I O
^
.. MEN W 1-------..: N PAGE 6
t
- CONT'iOt 4
-- s -------
1-A N D INSTRU
S..T. C MS.D.E G I G W d.
P t. AN'T SY u w,'**:*thi
.L. "J'.t.
/.*. n N 4 L"I : CO*
iw a s
w 4s.
In the performance Of a t'cct clecure an a MSIV, DESCRIEE what taken place whcn t".c test pushbutten is mcmontarily de;;resced AND when it is held in.
(Include in your discussion any differences betacen ncemal and test operaticn cf the MSIV's i.e.
speed, distance traveled etc.)
j l
m '_' E _ ~. ~i _ M.
a'. O '-
'm~. C C )
F j
I
- c.. What two devices or controls limit recirculation speed to l
20% as a mi n i murr.
CFluid ccupler stability 2 be cre the speed mismatch cir:uit e
b.
What two ccnditions must be wt will trip a Recirc Pump U - -. - n.v t. 0.
L.. n n )
4 o
u c.
INDICATE if the folicwing statements regarding the Ci;c Protecticn Eyst2m are TRUE cr ALSE.
A.
~5e tan.9all-mounted hose-reel CARDOX assemblien 1ccated trrcughcut the plant each naso a separate CO2 stcragc tank.
(G.5) 9.
A '. :' o u g h the Service Water Tie Line Valvo (MO2-2906) i s 1ccked in the c',ceed pacitien, pressure in the Fi.r Prcte:ticr. Water Sy2tu is r.crmall y mai ntai ned by the Service. Water 3 stem.
CO.5)
/
C. In the -utcmatic mcdo cf cp$rati:n of the Halon Supprassicr System, an activation c;gnal turns en the evacuaticr lightc, 3 cur-d 3 2 s.rin J r.G ' TiTiOd l a t 31 y C m.T.GM C 9s 2 3 iT *. ".u t d inJcCtion 2
-51:n.
(0.5) 3 C
c, cy 3 u21:n avstem 50: 12 t i.mject in the Au:. Electric 30C3
' t il pc2EitIe Ic CC: r 7 tbO in;Jcti2G
.41 t.' cD ibert
..sl t @ '. ". c 3 *. ? c 3 fj j i c '3" ~. tc C1G d.: c r-IC t h er O '.,, : '. El mctri c Rcoin.
0.5)
I i
l 1
%. - i O O b..n.,
b '.J l'
- m. '. !.., 3 k.- n,,
fs, b.
,,,,.,, C"' *;...I $, 3
%L 1 42
'G.
W1. t
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r a so.
4
- i l
i Q
l x-
1-AND INSTRI'MENTA, < N PAGE 7
DESIG,? CONTPGl.
x----.AMT SYSTEMS 6'
FL CCESTION 6.04 (2.00)
Given the fellcwing ccnditions and,the attached EHC logic diagram:
~
30" Generator cutput = 30".
Reacter power
=
Reacter prescure = 934 psig Throttle pressuro = 927 psig 920 psig Preccure regulatcr cetpoint
=
Qecirc pumps in MANUAL at 29% cpeed Lcad set
'O
=
195".
Load Li mi t 100".
Mx:imum combined ficw
=
1900 rpm Turbino speed
=
The generator cutput circui t breaker (OCB's) are =uddenly cpened manual 1 '/.
A.
Has a LCAD PEJECTICN cccurred? Juctify your answer.
(1.0)
D.
Which of the belcw subsystems of the EHC Pressure Contrcl and Lcgic System is the first to produce an error signal when the CCB's cpon? EXPLAIN.
(1.0) 1.
Pressure Centrol Unit 2.
Bypacs Centrol Unit 3.
Lead Ccnt ci Unit 4.
Speed and Acceleraticn Centrol Unit
-, Cw Lcn_rcl.Jni.
t.,51 */ 2 Q
eA m
r:~
Arswer the :0110 wing cuection in regard to LPCI Icop colect Icg;c:
A.
"~W doen the ic;i: detcrnine hcw many racirc pumps ai e unni,,q? P'OTc: Includc-setpcints wharc 2 p p '. t c a b l e '
( 1. Q
. uCN dcec tre Icgic dotcrai.2 wh :. c h ;c the 'JMC AMI'EED Pcci r c
- e.,,, e s
. ~. g
-a b
a 6
. m 6
$ =
-LS yt--
7 cc. rr~ 4 w.
- v. r.
.,. w'.,. n t
-. 6
-a 4 i
t.
A.....
.w a
- f. e. e..,
d
y 'I
. r, 7.7 I $ \\/.
- 3. t:
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u r-1<-
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_U._M.E N T r). _. O N PAGE 8
INSTR S'/97 CMS T)ES T G.v.-------------------- 1-AND CONTROL FL N ET
.d.
r; r c 7 7. t'.s1 A. 0.A_
( ".. e m )
w cr cr.ch of the IRM (Intermediate Range Monitoring) range ch anges C
listed below provide the ic11 cuing:
1.
The indicated level on the new range and l
C.
9.ny nutematic acticns ini tiated as a result of the indicated level en the new range.
t a.
Switching from range setting 5, reading 25. up to range setting 7 (1.0) b.
Switcning from range 6, reading CS, dcwn te range 5.
( 1. 5)
CUESTION 6.07 (1.50)
Fcr each cf the folicwing, state whether a RCD ELGCK, HALF-ECRAM, FULL SCRAM, cr NC PRCTECTIVE ACTION is generated fer that condition.
NGTE:
IF two cr more actions are generated, i.e.
red biccit and a ha_;-scram, state the most severe,
.e.
half-scram.
Assume NO oper-ator ac*.i ans.
a.
APRM Ch 1 Downscale, Mode Switch in RUN (0.3) b.
?:t-- Fl cw Ccnv. Units Upscale (>125% ficw), Mcds Switch in RUN (0.5) c.
APRM Ch 1 and Ch 5 Upscale, Mcde Switch in STARTUF O.5) 1 c '_.r._* ?... _.,.
- .,.NC
, i
( 7.. )
The "Cafec1 M.rde One Red Per:u dsive" light i :s energi cd : hen what thr e (-)
ccndit;cn2 are 22t:sfted~
(1.0)
.w'
- , e.
c.
e_ c t r.
, w.
.,,, m,,
- 5 3
e w
a 8
e 125 VOi~ : On O '* : l pcWur i2.5}
'"e
- c,"' t u e ; g:- t on tM J > circ;. t 3ran er ' scal : nt-c! acrel
- .di:2t:2 W A ' t,..: ccnd:ticna 2:12:2 30.3)
{,,.K.y g. e ;.. t +.
.. _ t
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,pt p, -.. g,. g.,, 3
, - - r. r.,,.,,, -
C.
r 9
s 9.
P r._r N_T__S_V_7_T_EM_S__G_E_E.I._Gi,1 _C_O _N_T.P 0 L _ _A_N_D_ _I_ N_S.T_.R_UM E_N_T_A_~.. _u _N.
PAGE 9
__2 1
CUESTIGN 6.10 (1.50)
List three ins trucnen ts or indicatibnn that use the Standby Li quid Contra' injecticn sparger line as an i nstruinan t tap.
( 3 :D a. 5 ca. )
1 a m__=_ T. r r_ e,4 c_ 4..
(
.., o )
i Li:t four ecnditions-(cr signals) and SETPOINTS that cause an automatic cicauce cf the MSIV's. (Assume the MCDE switch is in "STAFTUP")
Q'.;EST I CN 6.12 (3.00) a.
The Rcd Wcrth Minimi:er System user what i nputs fcr reacter pcwor' (1.0) b.
What are the -equirements to automatica11/ bypass the RWM blocks? (1.0) c.
What is the significance of he TRANSITICM ZONE?
(1.0)
'tt2:5E END CF CATI 3C:V 06
- tt3*)
2_______________________________________m.
(';
e
- ----- --.E..R. G.E.- N C Y
.A..N..D PAGE 10
- NORMAL '
'1.-. c.O.N O.R M A L,
---.P. E 9
- 7.. PROr:EDtJ
--D. ICLOG f C AL. CON. TROL PA OUECTICN
~.01 (4.00)
Tb s' Unit 2 plant ic cperc, ting at ' 8'57. p c wer, for scme rassen the Mz.i n Generatcr fic1d broa!<ers trip cpen which recults in a turbine trip and a reactor scram.
WHAT are the 7 required cperatar immediate action uteps.
(3e >pecific)
'. A n action stop may ha.c mora than ene action item.)
i l
1 n L, e i ;. r t 1
,.m,
(,,. - e )
a c-v.
A.
Under.ahat conditicns can the NSO inject SELC without authoricatien from a superviscr?
(1.0) 9.
After the Star.dby Liquid Centrol Syntam is initiated, WPAT are the five (5) indications that the system is operating?
(1.25) 1 C.L'E. C._ "" ?. f* P.1
".(**
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ir cccardance with the Radiaticn Fratection Standards, WHAT are si:
'6) of the eight (3) ccnditicns/ situations that wculd require a..ceker to 1 eave a control 1ed area as quickly as pacsiblc, ccncictent with cafety' L.:
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?ROOEEURE9 - NOFMAL ' M NCRMA' EMERG"NCV AND PAGE 11 9 ----.---------1'------1---------
.P._A.D 7_O'_ C G..I _C..A..L C.O_.N._T__R _n_ t
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A. Befer-leaving the Centrol Rec,r[-during a Centrol Rocm-Evacuation, WHAT acticn2 are taken by the NEO?
(1.0)
B.
Where will the Shift Engineer establish his centrol center?
(1.0)
C.
How is the 12clation condensor cporaticn ver:4ied cutside the Ocnt ol Rocin AND where are thcze indications located?
(Tuc C2] are required cr full credi t. )
(1.0) c
.~.. r._e. T c y 7.30 L.. m )
t u
- a.. Operating Abncr:nal Procedure FECI':CULATION PUMF TRIP (DCA-202-1) reduces the speed cf the Operating -ecirc pumo to what value Walues for Unit 2 and unit 3 required.)
(1.0) b.
5.:pl ain why the cperating locp recirc flow must be reduced during single lecp cperaticn. (Your an:wer shall include ahv Unit 2 is (1.0)
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E;-i uf l y describ. the actirn r2qeired by precedure CGA-250-1.
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PAGE 12 l
!'.. FRCCEDURES - NCRMAL Art 1CRMAL EMERRENCY AND s
r-------- c--------------
--- ONTFOL C
7ADICI OG.ICat.
4 PLf C C T *. C A.1 7.13
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A 1csc of RECCW has occured. What 'ac t i o n must bc iM on if ficw cannot be restored in less than 2 minutes ?
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--.A OM *,N I C5 A7 i t r P.PO%dURES.1-CCNO I T I ON9 ANO t.IM< MTIGN5 PA.GE 13
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Wher a systen ic determined i ncperable solely because i ts cmergercy cr normal pcwer ecurce it inoperab'o, it may be considered operable 2cr the purpace of satisf yi ng the requirements of its Limiting Canditiens fcr Operations, provided WHAT two (C) conditiens are met?
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- n acccedance with the " Cut-of-Service and Paracnnel Pr ot ec t i on
pr:ccdurc OAP 3-5 t % requect for equipmer.t cutage, there are thr:e
(~1 cases which require additional approvals besides the 3hi t Euper,.scr's.
WHAT are these three (3) cases and WHO's
- pprov21 must be cbtained? (Titlec only on apprevals.)
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~ e Dres"en Technic.1 Specifications state a minimum and a ma: i mum w s.t cr '. a v a l and a ma::imum water temperaturn fcr the cuppression pool during reactcr operation.
WHAT are the basec for these t:" ae ' -~ ) 1:mits en the supproccion pocl.
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'.c c c-d i n g to 10 CFF SC, the reporting raq irement of ONE HCUR for the im.
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Es,_6EULU1313CIIME_EE9CE JEEEt_ECO21112MEs_dME_'=15;; IlgUS PAGE 14 QUESTION 8.05 (3.02) a.
Any time irradiated fuel is in the reactor vessel and the reactor coolant temperatdro"is above 212 degrees F, che reacter ccclant leakago into the primary centainment fecm unidentified scurces chall not exceed 1
gpn and the total ccclant laakage shall not exceed ___2___ gpm.
5.
Wh-t 1 the baces behind the unidentified AND tctal coclant laaksge rates?
rL c_e, i I n_eJ e.me
(.3. e.n )
m m
What must be done in the event of a safety limit being c::cneded 9 nive c_. e. n_ N S.m, r, _. 0 0. )
u -
w State whether each of the folicwing statements ic TRUE or FALSE:
2.
A Licenced Scr. ice Reactar Cporatcr (SR) er a Foreman with an 9RO limited to fuel handling shall directly supervise the fuel handling operation frem the ro#ueling deck.
CO.5:
ucling floor, during refueling operations, is a b.
'5e ee rect :cted area.
C2.52 c..Du-ing c r? Elteration, CNE (1) SRM shall be Operable; cne in the care quadrant where fuel is being moved OR cne in an Id;2:cnt quadrant.CO.52 d.
"c rer2 than ene (1) of the four center IRM's may be byparsed cur m; cara al ter ati cns. CO. 5:
(2.0)
'm7: CATE3ORY 28 CONTINUE: GN MEXT PAGE Ft***)
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9_.
1 2
PAGE 15
_.A_D__M I _N_I S_T_R_A_T_I _V.E._ _P_ R_O _C r.9_U_R_ _E_9
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!n,2ccordanc2 with Operatin Crder #6-85
" Unit 2 and 3: Dr/wel' Equipment 3
Ficar Drain Pumping and Leakage 'Li6it ations. ", it states "When an equipment er floor drain sump high 1cvel al arm cccurs, determine radiolcgical ccnditions, using the criteria below prier to pumping dcwn the cump2."
MATCH the plant cceditions in part
'A' with the conditions that must c::i st prior to pumping, in p.2rt
'3",
i.e.
4,d r-A n.
e 1.
With the Fcactor Mcde Switch i n Run or Startup with steam flow to the turbinn ce condensor i.e.
generator on line or bypass valve open:
2.
When in Refuel or Ehutdown when reactor pressure is abcvo at:r.ospheric:
When in Startup cr Het Standby with nc steam 41cw to the condenser:
PART
- B' s,.
. :cacter water level above -59" when irradiated fuel is in the vessel cr check 0 ywell Cam for ncrmal activity, b.
Verify esctcr water level, Drywell pressure and main steam line Rad Mcnitarc indicata normal condition or ch ec !-
Drywell Cam for ncrmal actim ty.
c.
Ruertar water level abov2 -59",
Drywell pressure ncrmal, :..e.
less than T2i; cr check Drywell Cam for normal activity.
(3 9 T.5 ea.)
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4 A4ter c '. ; Feactor Ecramn in shich a Croup Cno i-sol a t i on si ;;n al han NOT I
cccurrea, the relay; fer th e.'tS I V ' 5,s AC Sol enci ds tr.ust be visually veri'ied, Whsn these colencids are found to be "decpped c u t ",, how are t.n n y raset*
(1.0) e T.r m s.
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m When cr the AUTOMATIC moda cf an ECCS System be defeat.nd" (1.5)
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PS*1 = u'S y p :M yi.t p t N g u-2 =-
ANSWERS -- ORE 3 DEN 2&3
-86/01/29-KING, M.
ANSWER 5.01 (2.00)
A.
Alpha V i ncreases power 9.
Alpha T i ncreaces power C.
Alpha D decreases powcr Olpha V decrecces pcwer then increasec power (4 9 0.5 ea)
D.
i Rce t,m..rc
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L-,.. 0 )
- ~-
a A.s ficw is i ncreased coolant ficw past the fuel elements tends to suecp away the voids more rapidly than thay are fermed.
This causes a peci tive reactivity addition by the void f raction (ccef.) of CO.52 reactiv ty.
Power level begins to rise immediately, which increases the fuel
.mplies mcre heat transfer to the clement tamparatures.
Th;s ccclart, thun i ncreacing the steam generation rate.
CD.52 With the ;:cwer increase, the void fraction incroaces.
CC.52 C.e r.c t c r pcwer i ncresces until the n2gative reactivity insertion ate due tc i n cr o ni.7 g /cid feacticn and Doppler cacfficiant cvercome2 tre ;;csi ti ve r eacti vi ty si:11 present in the cere.
C3.52 ha :cwor I a'.'el st2adias cut with nec r 2 2cti vi ty equal tc ce-o,
-r d
.'ciJ f ac ticn necr ts cr:gina! value.
CO.52 c _..c =_ = =.. w~- =._
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PAGE 18 AMD 7t.frR r.in n y N A M_.T r e u w_-
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They are phycically 1ccated as far below the normal water line as-pcssibic to prcvido the greatent static head.
2.
With feed ficw les: then 20*1 they are kapt on minimum speed.
j
- 2. At high power cperaticn cadoc;icato MPSH is cbtained from 2 ecdw ater subccclir.g.
A.
Law reacter Vessel water level trip 3 cc.vitaticn intericck.
j 5.
Sucti:n v al vo clcsad trip, cavitation intaricck.
c.
Pressure in the steam dome.
(~ O 0.5 ca)
- 4..ce. ~N. r e
- e. e OREECEN - Pccirc System Lescen Plan pg 16 & 10 GE Thermcdynamics, Heat transfer & Fluid Ficw, page 7-93 & 94 M5WER 5.05 (2.00) r W.h en the GRV c;: ens plan t cressure will accreaseEE.203. Power starts decreasingEO.20: due to increasud voidingEO.202, MSIV's close cn low h2cdcr pr etsur e [ 0. 20 resulting in a R:- S c r a m E O. 2 0 'i.
Vescal level N111 2 '_ u c tuat e.c 'p due tc increased voiding then ccwn after scram due to ~cid col l ap cc E O. 20 2.
Pl:nt centinuce a rapid c cidownEO.20] until 3
SP> clcuesEO. 202.
Staacy st2te conditicn will bc, R:t $/C with precs & temp stead 1 / increcsing due to decay heatE2.22.
Fed t the reactcr frcm FFP's with level at normalEO 203.
(2.0) 1'
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s PAGE 19 h __l%E031 CE UMCLESE_EC'llEE._Eb6UI._GEEE611GU2._ELUlE!1Ld6U2 T_.H_.E__._ O_ D Y_t_.m M_ _I C S M
ANSWERS -- DREEDEN 253
--9 6 / 31/ 29-K I NG.,
M.
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A.
CCNTFOL CELL: cne centrol rod and fcur surrcunding fuel (1.0) bundlas.
E.
This strategy eliminates the need for r e-icdic control rod pattern "_ M.n r.g o s d u r i n g the operating cycle.
(1.0) c e r r_ a.L; p r4 I~.
4 CONTROL RCD ELADE AND DRIVE MECHANISM LP, pg 4
.43 Lev. v=o.
e 0e i n_.. G m, m
r u
A.
Decrease (.5) due to increased void centent in the core as flow decreaues(.5).
(1.0)
.oiding in the coret.25) and B.
Increase r.5' due te incrested racirc pump nc 1cnger taking a suction en the annulus (.25)
(1.0)
C.
Decreasa'.5) cue te level incerase (.5)
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.- _ - = _.= _ :- m ANEWERS - 2RESDEN 2L3
-96/01/29-KING, M.
\\
ANSWEP 5.10 (2.00) a.
T-2.ncition boi1ing may cccur which c an restal t in c1ad fni1ure.
(1.O) b.
To make the MCPR limit more conservativre to account fcr the pccsibility of a cudden ficw increase and a corresponding power increase. The MCPR is i ncreaced (cr mere conservative)
(1.0)
-'"Pecir:.
pump runaway" acceptable for "cudden ficw increase")
,x.c
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t.
NMP-1 Operations Techncicgy, Mod.X, p g. X 1, Tebh. Epccs3pg.70-7Da.
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The reacter i s now producing less stecm to go to the turbine. There wt11 be less e:: trac t i on steam and reheater drain steam gcing to the feedwater heat:r.(1.0) Therefore less feedwater heating will occur rewlting in ccider feedwater entering the vessel (.5) which will c auce reactor power to increate about 3% from the positive reactivity ad d ; t i c.m (alpha m). (. 5)
(2.0) c r_
c.e : r e_
"MP1, Eitelatcr Malfunctions Causes and Effects MS10 r_. p~e r. m e n.. - r _ m m_.... p t., n ac c . ' '.
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n II.El- ) U -. 2 '~ '2 0 " # 1/J r 2 CO';
'. CI-1C E!
P. 'f ~ ' ~ ~. 7 0 % ; '* 2.1
- g _ h *"
i e
5 e
e-i E___. _ _ T_ H,E_ O_ P_ Y O F__N_U_C_L_E_A_P__P_O_W E_R__P_L_A_N_T_ Q_P_E_R_A_T__I C_N__. _F_L_U__I D__E_ __ A_
FAGE 21 IMEEUQ21GQMlGS ANSWERS -- DREEDEN 2t<3
-96/01/28-KING, M.
ANSWER 5.1!
(2.00)
~
a.
I n cr er.c ed wceth of the peripheral rods and decreased wcrth of those (1.0) en the core center.
b.
Aniz1 flux distribution could be severly tcp peaked.
Radial flux dist_ribution will peak in the peripheral region and be substantially
'.cwer in the core cerater, t' 1. 0 )
REFERENCE Nine Mile Point Reacter Thecry Module I part 14 DREEDEN - Lescen P1an Back 4,
Ch 12 R:: Physics Review, pg 45 3
J
, lq ic_ E'JMI_SXEI505_EEIlhL.CQUIEG'u_0ND_lMSIE'R'EUIG1.d20,/
PAGE 22 ANSWERS - DREEDEN 2%3
-96/01/29fING, M.
1 I
.a~ tLa.wco, o. m.,
t. am.
r.
~
s
~
If the tect puchbutton is mcmentarily depressed, the AC L DC calenoid valves de-energien and the test calenoid valve is eno.gined to supply air to c1cse the valve. The MSIV then sicwly closes until the 90% open limit switch opens.
The sciencid vnivan then reposition and air 13 sipplied to open the MSIV.
(1.5)
I' the toct puchbutten is h21d in, thO sone'cccurs (abovo) encept the 7 0','
cpon limit switch is b yp a.s sed and the MSIV will Jull close in 45-6C cocandc. (Closing time not required fer credit.)
(1.5)
,: cc c: cw-.4.
DRE5 DEN - Leccen Plan Back 3, Chapter 3, page 13
.%,. Le i.,Lc..n c_. n --
(,.,Ln )
c.1 Limit switenesC3.53 2 Mc:henical stop CO.52,
( 1. 0) 5.l Spe:d m catch c::c c ads 10'. C 2. 3 :
2 The 1:wer speed racirc pump discharge valve in closedCO.53
( 1. m
- o. c.: = =.. c s kr~ =.
i.
w.
I DFEEDEN - Les:cn D1an Ecck 2,
Ch 2.
pg 3.'< 13, rev 14 l
l o
r.t, ~
- Lc -~
5'. m ~
.. _m. r.un.s te
-i 5 e C ( Il Id a
s *y 3..S C 3 9 I t.) s 3 $ CdOf *It 4
3 C f J.} C J
I 4
.l
((7 c:' o)
L~
s se e 4
~a L^
/q W)
- n,.3 Ls. -
- 3. 3I T C b 's' d 1Sn 31 (*jC dl C.3rt3 b
sCCCCd hfCd UCb.)**'}f) s
'"3Cr.
INhEn thG t J. T.Ef 10, t i,3Od Out The i '". ; t t 1 0 n '* t t r *. '? }
3
- u. 3 c' d.o
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o
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\\
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4 r.=
b.d>
w b - =hM
.=,m.
,,."p9, W ;f,
- i.. 3 de
-w c.
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--3 s
i
.m__.___________._.___._
)
G h.z.__ELGUI_S' SIEL5_DESIOLb_GGNI306_SUD_l!' SIS!JtEUIOl_:dB PAGE 23 I
J
}
}
AN3MERS -- DRESDEN OM
-96/01/29-RING, M.
ANGWER 6.04 (2.00)
A.
NCt.5) Load rejection in caused by a 40% micmstr.h between the first stage pressure and ctat or a.T.p s (. 5 ).
(1.0)
D.
4(.5) When the CCB's cpen the turbine is virtually unicaded and bugins to cvarspeed. The speed and acceleration unit sences the overspeed ard produces a speed error si gnal (. 5)
(1.0) c c en.eu..r.r-
.~r-,-
c.
mg S. r.s" C.t.J.C f.N C T ". LO, F,e'/
7,,
_ ~.t r
ANSWEP 6.C5 (3.50)
A.
By r.cnitering the diffcrential pressure (.5) acrccs each recirc pump (. 5) for :
psid, indicating the pump is running (.5)
(1.5) 9.
By ccmparing the pressure in the riser pipes an ene recirc locp with the pressure in the riser pipes of the other 1ccp. The un d a m ag c-d loop will hava a higher pressure than the camaged Icep.
(1.5)
C.
Lce; B (0.5) i a c c :. c < :, k. c-CPE3 DEN - Lesson Plan ecck 3, Ch 14, pg 15 (r?. 5)
.w. L-- t.e_.c
- 5..,e n-f,.. u.e )
a.
2.5 cn
- nge
'7 No cutcmatic a:ticr, dowrrr',e at 2 *..
(1.0) t, CC cr r.2nge 5.
' 9d
- p red ticc: and I?" % ; h -+ i g h icle ccr uc.
(1.5)
= c =. c p =.~a.
=
m
.s ed M
1 d.
s. ' c. * ', --
- * - 's :a
.. :< c- -- -.: n
< - -. r C r.- a. -
t
,e u
..-~: -
s ;.
L.
O L..__E's SI_ 'fII 55_2Esis,._ccNIEL._eun_IUSIEuMESIere_d e
PAGE 24
('.NSWERS -- DRESDEN 2N
-06/01/28-KING, M.
.n,.La w.e.,.
6. n --
r,.. ccLs
.t m
a.
red bicck er half-screm if concu-rent hi IRM nssumed b.
rcd bicck (3 9 0.5 ca) c.
full scram R cr. e c,< r.,Lc c m.
v 51.ulatcr Syatem Manual, APRM, Ch 9d, pg 1 5 5 of June ic35 rev
' 4MP 1, m
between pages 4 and 10 DRE30EN - Leccon P1an Book 1,
Ch 12, APRM, pg 5, 7,
11, I
t a
l 4
ANSWER 6.08 (1.00) 4 1.
Mcde switch in REFUEL 2.
Mc centrol red selected for motion (select matri: power off) 3.
All control rods full in (3 3 0.33 ea = 1.0)
J
- e c r_ e,e iC -
+. -.
Dresder - Les-scn P1an, Ecok 1,
Ch ap t =:r 7,
page 11.(rev 11)
.:.1;e.
.c.
c:. on r<...m m )
m s
Lccal, at t"e breaker with a manus.1 pushbutton.
(0.3) 5.; " a br2_Wer is clased CO.252.
, 1._.., ; s s.. O.;,. ).
,, O.., v,s.
L.. u. r,_.
>w
- u..- :
a
.w n.,, t.a w. /
r.....
i..
crp _ m.. *er p
> - L
- 4. j
.c g a<
?. @w, pg 4 4 7,
f.'..
s T*
- % C S. _". O % j _ f.-
- e.... n.
~p.,p.!
g
()
- 3 b z E L G U.I _ 3 2 3 I 3 0 E_.E E E I @ b l E C 0 I 6 91
- 2 00E.1UEI8!=55U13.m00 PAGE 25 AM9WERS ---DRESCEN 25.3
-06/01/28-KING, M.
1 ANSWER 6.13 (1.50)
Care plato differ:ntial press. measurement Ccre cpray line break detection Jat pump diif. pros 1.
tap
.r,u3 ea.)
w..s O.
REFEFENCE DREEEDEN Lesson Plan Book 2, Ch 3, Rev li, pg 8 1
v.
1
- r.. 00 '/
L <
... e n. Le o, w
Lco-lon reactor water level CO.252 0 -59
('/- 2)inchen CO.25]
Main Staam Line high radiaticn C2.251 0 ; t (n=rmal 100% Iv1) C O. 25 3 ~
I Stran tunnel high temperature CO.25] 9 200
(+/-
10)F C3.253 Hi.;n Steam cl< a in any main steam line C2.253 @ 123
(+/- 5)% CO.252 1
fref2rence l
- 'RESDEN Lessen Plan Ecok 2, Ch 3, r e'.
14, Mc.in Steam, pg 15.
,.,.,0,,
., ~,. ~.
c
.-o w c.s 2.
t : t =.1 s e n, flow and t t:1 feed 'lew (1.0)
.... -,. m,., d 2,, *J
,qe".
- n.. -. 2
- ,M.
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~~
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q
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a a
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4 =. S.. }.'"'*"e'
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c Y%,,
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4.
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,,.g, 3
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- 4. m.
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en p.
4.
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.=
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a g,,i t.
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+"**r
- O s'- *
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, T.
- 3. g.*
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f p,. q..,
'== g n. g.,
f 4. w r.=., p. g.,
.' h m w-.-
.=g+.=.a.r.4.,6. J. = p
. y.e.
&. '=i ra m
c g
g. @ *,
/4
- f. O.,. U.
- cp..aef J
.p.*e Js=
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M.M.
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-.J.E yp
- w. em
- - *.=
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w, a
, '.; v.m e*t s ie.-.
w=
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/
- ... _.... 4 $.
- Y'".i 4
. y
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=.%@
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___N_G_P _M_AL.2_A__B_N_O.R._M A_ _L.__EM..EFG.E._N_C_Y__A_N_D PAGE 26 A
EBgigLggIg8L_ggNIBg6 ANSWERS -- DREEDEN 2&3
-86/Ol/CO-KING, M.
l l
ANS'. !ER 7.01 (4.C0) 1 (Generator tripe the Turbine, Turbine scrams the Reactor when j
grantar than 49% lead.)
1.
Press both scram buttent and place the mode switch to SHUTDOWN.
2.
A.
VERIFY eli rods are insertad to cr boycnd 04.
9.
Incert any rad not already at 00.
Reccrd al1 red net at 00 and nctify the Shift Superviscr.
3.
A.
Attempt to maintain level bet,veen +9 and % 40 inches,by multiple indication.
Control feedwater in AUTO unless control failure occurs.
9.
U necessary.
STOP the 3eed pumps tc pervert e::cessive reactor water level.
4.
IF all rods inserted to or beyckd positic 04, then verify that the turbino and generatcr have tripped.
5.
VER!FY that recirculation pumps A & B run back to the minimum pump cpeed.
6.
VER:FY thate the auu:liary pcwer has trsnsfered tc the reserve sun:liary transformer.
~ ~.
Insert SRM/IFM's.
Maintain the IPM's on scale to monitor shutdown.
(7 0 0. 44 ea) r et.. u.n!re
.. r ib w
n, sC.C ? P p hj _
L, A. F^
S -'
w w w _.
= +
r h '@
1
/N, m,
- 2.:__DCCEEUEE3_2_UGEU3bd3EUGEU36;_EtEEEENCZ_2ND PAGE 27 E0blG6021G06_GOUIEG6 ANSWERS -- DRESDEN 2L3
-86/01/28-KING, P1 A L,c s,. c.I,.
.r. o.,m Ln. o_ c, )
+.
A.
1.
All centrol rods are not icerted to or beyond pocition 04 E.05]
and tcrus temperature cannat be maintained below 110F.CC.53
- cr **
2.
It the rcacter cannot be chutdown befcre the tcrus temperaturo reaches 110 F.
All rods are at er beyond ;;ouition 04 c-ac Shutdcwn
=
3 determined by a qualified nuclear eng.
EITHTER ANSWER 1 GR ANSWER 2 WITH AT LEAST ONE DEFINITION OF SHUTDOWN IS RECUIRED FOR FULL CREDIT.
(1.0) 2.
1 Amcer pilot light of squib firing continuity circuit not lit.
- Ficw indication pilct light lit
~ Water clean-up system isolation.
4 Decreasing icvel of SELC stcrage tank.
5 EELC valve circuit fail announciater light lit.
6 Pume dincharge prescure increases.
(5 0 0.25ca)
- e. ~r
- w_. s.,e r.
CRE5 DEN -
DECP - 100-3, page 3, rev.
O.
DECP - 400-2 3 pcge 2,. rav.
O.
SPLC Lessco Plan pg 14 ra N S W Ei-
~.33 C. CO) 1.
'>!b en instructed ce signaled to do so by the Rad-Chem dcpartment.
2.
- :. i J.-e cr nicpect?d f3: bra cf perscnal p r o t,:c t i v e equipment.
3.
.!r e n ;: ac t 7 d d e t :- i or.at i c n s# raciclogical ccnditions.
In the e nnt t at the,;c - k e r c cur-ont accumulated dose equivalent I t M ;.5 a c C O.T.O c. r.C T ' 5 J i." fCF 3 C '/ '9AOcn cr dCCE C qu i V C.1 O n ". 1E
- C.1 '.
- c
'"O
.. ?:p o sLr O 3'J. th.c r i 2 0c f c '"
.h c JCb.
l 3,
" 2., - -.cg. ' 3, 9 7 ; g r. 7 3c.;nd7 f p c.g c g i ; g gr., - g.,31 -
l i:. 1:.c;: l a t i ;r c4 acrk acut;necrt.
~
+ _,.
n.
9 M
w a-
'we e
e e en unk'ichF,
- l. ; 3 2.5 72)
=.=.==.=..c.~r=.
u du a
p ou.
n-
-e E
mi n
.4 4
d i
l l
l a,,,,...
m
. _T y
zx-----------.AND
~
FAGE 20
.'
- PROCEDURE.S - N.ORMA! 1. ' ASNCP'd41 EMERGENCY 4
a PS D ICt _OG I C Al, CONTROL ANSWERS -- DREEDEN 2R<3
-86/01/28-KING, M.
ANEWER 7.04 (2.00)
A. Bl ace the Di scharge Volume High Water Eypasc/ Air Dump Bypass Swttch an Panel 902-5 in the BYPASS pcoition and VERIFY the ceram discharge volume vent and drain valves apen.
C.
INSERT rods manaally using the EMERGENCY IN position of the Rec-Out-Notch Overrida Switch.
D.
TR:P ;PE bus breaker at Centrol Rocm Panele 902(3)-15 and CO2 (3) -17.
E.
Individually SCRAM control rods from Control Room Eack Pcnel 9G2(3)-16.
Renat each rod after i t is fully innerted if individual 1y scrammed.
F.
TRIO FFS breakers in Au::ili ary Electrical Equipment Rocm which feed Central Rccm Panels 722(3)-15 and 902(C)-17.
G.
t'amentarily TPIP RPS M.G.
cet breakers at MCC's 28-2(38-2) and e_e. _. m < o,
- n..
- ~
H.
VRL VE Ct.'T and hieed of i nctrument air to Ecr am Solenoid Valves.
e I.
ancally CFEM ar BYPASC the 5: ram In strumen t Vclume Vent and Orcin '/el s es if possible.
s' a 2 3. =., a s- \\
t ~. m,.a' n C TT O Ch jf*"
ee +
"@e
%L g
n y
,F4 O
1.
g",4 a:.., <.. _ -
w.
y
.sg u
F l
l
[
-m,
- m
)
~L._ICOGZWEE3_ _UGEM0bl10202BdeL2_ECEEGEUGLSUD Pt'GE 29 P__A_D_I__O_!._C.O_I _C_AL__C_O.N_ TROL AMEWERS -- DRESDEN 2&3
-86/01/2G-KING M.
3 ANStiER 7.05 (2.00)
A.
1.
Manually scram the reactor 2.
Leave ecde switch in run 3.
Trip the CRD pumps 4.
Trip the turbine
'4 S 0. 25 esO (1.0) 9.
Ho sets up in the Tech. Suppcrt Cantar.
(1.0)
C.
Observing steam venting frcm the shell side of the Isclation Ocndonscr.
Scuth Side of R.E.
(1.0)
.w,.e : E.o c.N F C_
l j
~
Drceden EPIP 200-20 pg 1 2,
L 3 of 14 j
3 GNS!*iER 7.06 (2.00)
,, sJ. i 4,._./.
- c..
UNIT 3 63%
(2 S 0.5 EA.)
b.
The ficw must be reduced to prevent jet pump vibraticn.
Unit 2 has inadequate jet pump riser bracket design.
.e- _ c c, c_ %e._
w a.
m.
DRESDEN DGA-202, pg i and 4
.m{ 4 s2 wh L.e
- c. m_ e a. cv/
- euc w
s r.u/
Trip EGC (If cperating) cl ac2 r.u c t c r retir; in m eusl e.n d r:dnuca rectr; flow to minimum.
T"" an O f Or the Ru;' OcWer to kr Jr* 3 f Cr T10r 22 '22)
Ad; lot E;
.=J d t 2 r 19VG1 tC 'h? h2 ;h 19VT1 a l ra" 9 7211CW E; SCrdM O F C C 9 d' ir O 2nd
.S ai W a l '. y SCr ?.m thG re30tCP.
- t. r-k.
m q e s
we J
,n.6 e /
C_ k I"'u'.*
! T..*- ".~"" 7
.u
._w_.e m
L., -.;,, a O t.,
.s j
W
l l
m l
)
L'
't.'
=ROCEDUPES - NPRMrLT_GENCEMBLx_EMEE@ Edgy _eND PAGE 30
~'
E02106021C66_G2MIBG6 ANSWERS -- DRESEEN 2L3
-86/01/29-KING, M.
I l
ANSWER 7.03 (1.00)
~.
When level increases above the range of the narrow range instruments.
REFERENCE DREEDEN - DOA-600-1 ANSWER 7.09 (2.00) 1.
If the shell side level inst. approaches full scale (1.0) 2.
If the radiation level at the vent o::ceeds 20 mr.
(1.0)
REFERENCE ERESEEN - DCA-12OO-1, pg 1 ANSWER 7.10 (1.00)
If RECC W :n Icst fer > 1 minute the recirc pumps must be tripped.
FEFEFENCE DRESDEN - DOA-3700-1, PG 1 I
l 1
A m
i h _'109U1MlEIESIlME_SECCEs'JEEEs._GQUElIl9EEs._6UQ_wlM r M I002PAGE 21 ANSWERS - DRE5 DEN 2C
-86/01/29-KING, M.
ANSWER 2.01 (2.50) 1.
It2 cerrccpending normal cr emergency pcwcr scurce i: cperable.
(1.25) 2.
All of ;ts redundant systam(s) in the othar division are operable, or likewise satisfy the requirements of the sspeci ficati on.
(1.25) 4 c.e c e_ r c r c e wi L-4 Dresden Tech. Spect. T.3 9 pg Cla ANSWER 3.CC C.CC)
- 1. Outages of ECCS and primary centainment ccmponents C3.5:
must be appreved by the Operating Engineer for the offected un4+
r_,u.a.,
( 1. 3. )
s.
2.
Out:qcs of buses, transfcrmers, diocel generatcrs and bus tis b r o r.h er n CO.53 muct be appreved by the Unit Supprurt Oper ating Engineer CO.52.
(1.0) 3.
Cut ager of Fire Prctection Systems or equipment CO.52 must be appecved by the Staticn Fire Marshal CC.53.
(1.0) pre. rn. e_y. re_
1 Cresecn DAF T-5 pg 1 '< C cf 17 f
J
.a N.e y. c e m-4.._. m C. 5
>~
m e
r 4
a I 4
e 4.
i, 4
4 tha sie 2m itu -i.9g 3 EDA.
(l.C)
M;.: t muc-1292) i: to mininin the amount of precaure bu 1 dup during a
$ 5 A, Ma'nhTin prCO'3ur7 1 9 3'5 t h 3.F cCntJir. Cont O Cs i F.1
',1. C }
'C : i mL* m P. O mp F it '. r 2 1 ; 9 i '.
72Cr~ IC Pp1 Ot 3Ds;rtt Cn Cf tha e ri M g y ulc22nd d' r t. g. C5A : -d tn2 tcrr:Oraturc.- i s c ' a i 1 1 rot c::caed the
)
C %? f' ' 0 4 7 CCrd702itiCr I?"p dr 3.t ur a (cf 170 F) im
.h e tupprocatan
.c.,
f.
<J.
I ** T T. *.*. C. * ! ". C t
4,-~1f
&p %.
a
<*4wF y,-.-
s p
- e
+s.
W
..a e
7 l
A l
+
t Es__.6,EE11MlSIE6Il2E_E8EChdl!EEss. EGUDlllCUE:._602.1 1E.li_ IlgUS PF.GE 32 ANSWERS -- DRESDEM 2&3
~86/01/2S-MING, M.
I AN3NER 9.04 (2.00)
A.
true C.
false 3.
true D.
true
(.5 each)
(2.0) o.rt.-
e.e.m.re
.m 3.,. ~.
< m.,
e c c..
e.
.1 AN?WER 3.C5
(~.C3) 2.1 5 gpmCO.5]
.2 25 gpmCC.53 (1.0) 5.
5 gpm based cm crack propagaticnE1.03 25 gpm based on capacity of sump pumps (100 gpm) to removeC1.02 (2.0) leakage.
_,e3Jc -
nerrm_.uc c..
. e, *4.6 e.D 1 D. g m gJ n au x.
i.
.m,.. m_, s = =..
=. G. t
(.<. c. 0 )
~
If a nfety limit is 2:t c ceded, the reactor chall be shutdown
. meditaely and reacter operations shall nct be r:au: nod until authori:cd b/ the NRC.
(ThO condition of the shutdcwn shall be prompt'.y repertad t; the divicico.T:an ag er r.uc l e ar sta!;iana or his dec1;natcd alternate.)
=. =.= =, I = w.. ~ c w.
C. f..
h.
.] f}
b b.
== i. h,
de w,
f
-. +,
y s ' C t, : r'* *"*
2.m.
I m.
~9. m ',
-~
~
.c u t r '1 C
'. " " *X4 A.
5 $1
-2 or,
n.
...s l
2 v.
1 ~_.T '"_" ? 7. *. ; f".* *.**
J.
- ~.....
ss 4
9 y _s s
-9
_m__._-
7 m
- h. -1$51UlEIEOll'2E_ ECGIEWEEEx COU21112MEsfe1D LIMi aTIONS PAGE 33 f
i ANEWERS -- DRESDEN 2h3
-96/01/28-KING, M.
4
.n,.m_,.1n_e. s, c.no
- e. <. c. n )
n, w-1 b
a (C 9 C.5 Ca.)
C C
RE ERENCE DFESDEN Operating Order #6-55 pg i of 3 July 1,
1985 AN3WER 9.29 (2.20) e.
The addition of hydrogen causes increased radiaticn levels in the plant due to generation of N-16. This increased radiation ic "short" lived and reduced to ncemal level at chutdcwn or cocuring of hydrogen injecticn.
(1.0) b.
1.
The MSIV Hi-radiation scram setpoint must be lowered befcre decreasing pcwer below 20%.
- cr #18 2.
Check ARFM gain adjustment factor cn the computer.
If any AGAF greater than or equal to 1.22, adjust por DIS 700-17 (How the AGAF is checked or DIE 11 are not required for full credit)
(1.0) ee.rr ~ cure
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