IR 05000387/1987005
| ML20213G559 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 05/07/1987 |
| From: | Collins S, Keller R, Kolonauski L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20213G509 | List: |
| References | |
| 50-387-87-05OL, 50-387-87-5OL, 50-388-87-05OL, 50-388-87-5OL, NUDOCS 8705180345 | |
| Download: ML20213G559 (98) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NOS. 87-05(0L) and 87-05(0L)
FACILITY DOCKET NOS.
50-387/388~
LICENSEE:
Pennsylvania Power & Light Co.
2 North Ninth Street Allentown, PA 18101
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FACILITY:
Susquehanna Steam Electric Station (Units 1 and 2)
EXAMINATION DATES: March 16-19, 1 7 CHIEF EXAMINER:
Yt/F)
Lynn Kolonauski actor Engineer (Examiner)
Date
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d([77 REVIEWED BY:
Robert M. Keller, Chief, Projects Section 1C Date APPROVED BY:
M(tMtfd 5 Y/87 S'amuel J. Collins, Deputy Director, DRP Date SUMMARY:
Operator licensing examinations were administered to five (5) Senior Reactor Operator candidates and four (4) Reactor Operator candidates during the week of March 16, 1987. Two candidates failed the R0 written examination. Two candidates failed the SRO operating examination. The remaining candidates passed their respective written and operating examinations.
Overall, three (3) SR0 and two (2) R0 licenses were issued.
8705180345 870500 ~
hDR ADOCK 05000387 PDR
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' REPORT DETAILS
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TYPE OF EXAMS:
Initial Replacement _X_
Requalification EXAM RESULTS:
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SR0 l
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Pass / Fail i
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4/0 l
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CHIEF EXAMINER AT SITE:
Lynn Kolonauski, NRC 2.
OTHER EXAMINERS: David Lange, NRC Allen Howe, NRC Brian Hajek, NRC Consultant 3.
Summary of generic strengths or deficiencies noted on oral exams:
While the examiners noted individual strengths and weaknesses, no generic strengths or weaknesses were identified.
4.
Summary of generic strengths or deficiencies noted from grading of written exams:
The SR0 candidates did very well in sections 5, 6, and 7. However, the SRO l
candidates indicated a weakness in answering the questions concerning the
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SSES Technical Specifications.
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The RO candidates were strongest in sections 2 and 3. However, no generic strengths or weaknesses were demonstrated by the R0 candidates as a group.
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5.
Comments on availability of, and candidate familiarization with plant reference material available in the control room:
All main control room reference material necessary to conduct the opera-ting examinations was readily available in the control room. While selec-ted SRO candidates were proficient in locating and using this material, the remaining SR0 candidates seemed somewhat unfamiliar with its location and use. These comments are documented in the individual operating exam-ination reports.
6.
Personnel Present at Exit Interview:
NRC Personnel Lynn Kolonauski, Reactor Engineer Examiner Tracy Lumb, Reactor Engineer Examiner (Trainee)
James Stair, Resident Inspector Facility Personnel Thomas R. Markowski, Day Shift Operations Supervisor Howard J. Palmer Jr., Supervisor of Operations Robert G. Byram, Plant Superintendent Kenneth Roush, Supervisor of Nuclear Instruction Arthur Fitch, Operations Training Supervisor William G. DiDomenico, Simulator Instructor William Lowthert, Plant Training Manager 7.
Summmary of NRC comments made at exit interview:
While the examiners noted individual strengths and weaknesses while administering the oral exams, no generic strengths or weaknesses were identified for the licensing class.
The personnel of the SSES Training and Operations departments were very cooperative throughout the examination period. The examiners acknowledged the cooperation and assistance of the simulator instructors.
In conducting the plant walkthrough examinations, the examiners experi-enced no access delays and noted that the plant appeared clean.
The training material provided for preparation of the examinations was well written and generally complete.
Despite the lack of malfunction flexibility provided by the SSES simula-tor, the machine was still acceptable for NRC operator examination pur-poses. The examiners recognized the age of the simulator and the facilty plans to improve the simulator's range of malfunctions within the next two years.
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Summary of facility comments and commitments made at exit interview:
The facility acknowledged the NRC comments and thanked the examiners for their cooperation and professionalism. The facility felt that the written exams were challenging and fair.
9.
CHANGES MADE TO WRITTEN EXAMS DURING EXAMINATION REVIEW:
All comments about the written examinations were resolved during the exam review conducted with the facility. The examiners left the site with no unresolved comments. The following list represents significant changes made to the written examinations.
R0 EXAM Answer No.
Change Justification 2.02 d.
Added throttling of These are additional methods HX outlet valve or as identified by the facility RHRSW valve.
2.05 b.
Accepted " false" with The Suppression Pool may also explanation, be used if the CST is unavailable.
2.06 a.
Added " rejecting This is an additional unfiltered water to condition identified in the Radwaste".
RWCU procedure.
2.07 a.
Also accepted valves ON-159 Attachments A and B listed in ON procedure.
list additional valve responses to the NS4 signal.
references).
2.10 Added requirement for Additional information candidates to include required for full credit.
LPCI injection logic.
3.04 a.
Added "When Cont In Additional correct answer PB is depressed" as identified during review.
acceptable answe _
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R0 Exam (Continued)
-Answer No.
Change Justification 3.05 Corrected runback to Corrected typo; took 30%.-Also noted that possible question SSES does not normally interpretation into increase recirc flow consideration.
until 60% power.
3.08 b.
Corrected for number Original answer incorrect.
of key switches.
3.08 c.
Changed "30%" to "24%"
Original answer incorrect of rated first stage due to lesson plan error, pressure.
4.02 a.
Corrected from "None" Original answer incorrect.
to "E0-100-101".
4.10 Accepted " increase".
Clarification of procedure provided by facility.
SRO EXAM Answer No.
Change Justification 5.09 b.3.
" Increases" Original answer incorrect.
6.02 c.
Removed "Rx pressure This interlock has been below 600 psig" from jumpered out at SSES.
answer.
6.04 a.
Accepted also-These are additional correct power level changes, answers not contained in the CRDH system parameter original question reference.
changes.
6.04 b.
Accepted also-This is an additional cause local SRI switch of a rod drift.
actuation.
6.05 4.
"e" added to key.
Each of these signals is a 5.
"a" added to key, possible RPS initiation signal for events 4 and 5.
6.06 a.1.
Deleted from exam; Question is inaccurate points redistributed.
because fuel zone range is calibrated without recirc flo,
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- SR0 Exam (Continued)
Answer No.
Change Justification 6.06 b.2.
" accurate" is also Upon ADS. actuation, acceptable.
pressure indication will be at the lower end of range.
6.07 b.
Rod block added to A rod block will occur answer.
due to the APRM flow unit.
6.09 a.
Added Circ Water Inadvertantly omitted pump trip.
.in original answer.
7.03 a.
Accepted additional The Discussion Section of alternate answers as the ON contains a broader marked on key, explanation for the basis of each step.
7.10 c.
Accepted also-Loss Alternate correct answer of NPSHA for LP ECCS.
identified in E0P Bases.
8.02 2.
Question changed It was necessary to be during examination.
more specific.
Answer changed to Answer changed to match Site Emergency.
modified question.
8.06 b.
Answer modified Original answer inaccurate.
as'shown on key.
With this specific TS problem, there are several possible and equally correct interpretations.
Attachments:
1.
Written Examination and Answer Key (SR0)
2.
Written Examination and Answer Key (RO)
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NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
S__----
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-_USQUEH A_N__N_A__1 &_2_--- _ __ _ _
REACTOR TYPE:
_gWR-gE4______________ __
DATE ADMINISTERED: 87/03/16 EXAMINER:
_KgLgNAugKI1_L.
________
CANDIDATE:
_
_
_ ______
IN@IgyCIJgNg_Ig_C@NQ1991El 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 trade requires at least 70% in each category and a final grade of at Icast 80%.
Examination papers will be picked up six (6)
hours after the examination starts.
% OF CATEGORY
% OF CANDIDATE'S CATEGORY
__YelyE_ _19106
___gCggE___
_ygLyE__ ______________C9IEgggY______
______
2St99-_ _23t99
________ S.
THEORY OF NUCLEAR POWER PLANT
___________
TH RMOD NAM C I
'
_24 99__ _24 99
___________
____
_ 6.
PLANT SYSTEMS DESIGN, CONTROL,
AND INSTRUMENTATION
_2@g99__ _2@ 99
________ 7.
PROCEDURES - NORMAL, ABNORMAL,
___________
EMERGENCY AND RADIOLOGICAL CONTROL
_22199-_ _22199
________ 8.
ADMINISTRATIVE PROCEDURES,
________-__
CONDITIONS, AND LIMITATIONS 199t99-_
x Totals
___-_______
_ _ _ _
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Final Grade All work done on this examination is my own.
I have neither given nor received aid.
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Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
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During the administration of this examination the following rules apply:
1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil only to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
B.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne side of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
I 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has been complete.
.
10. When you complete your examination, you shall:
a.
Assemble your examination as follows:
(1)
Exam questions on top.
(2)
Exam aids - figures, tables, etc.
(3)
Answer pages including figures which are part of the answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions.
c.
Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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QUESTIDN 5.01 (2.00)
o. According to the SSES Technical - Specification, " shutdown margin (1.5)
is the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods'are fully inserted encept for
"
...
List the three (3) additional conditions for this definition.
b.
Give the minimum acceptable value for shutdown margin as stated (0.5)
in SSES TS 3/4.1.1.
(Include units.)
QUESTION 5.02 (2.00)
The purpose of the end-of-cycle recirculation pump trip (EOC-RPT)
(2.0)
(as described in the SSES Tech Spec Bases) is to recover the loss of thermal margin that occurs at EDC.
List and briefly explain two (2) of the reasons that cause a high reactor pressure transient to be more severe at EOC.
QUESTION 5.03 (1.50)
SSES Unit 1 Procedure DP-164-OO1 (" Reactor Recirculation"), contains (1.5)
o prerequisite that limits the temperature differential between the Reactor Vessel Steam Dome and Bottom Head Drain.
If the Bottom Head Drain temperature recorder reads 395 degrees F, what is the maximum allowable reading on the narrow range RPV pressure instrument that would meet this requirement?
Show all work completed in arriving at your answer.
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QUESTION S.04 (3.00)
The HPCI system will automatically swap from the CST suction to the suppression pool suction on low CST level.
Failure of this cutomatic transfer could result in inadequate Net Positive Suction Head Available (NPSHA) for the HPCI pump.
The following information applies to the HPCI system:
HPCI pump suction elevation = 650.25 feet Transfer level elevation = 673.75 feet Suctica transfer point = 3.6 ft above bottom of CST Net Pasitive Suction Required = 15 feet Head loss = 7.1 feet CST diameter = 40 feet CST water temperature = 80 deg F c. Calculate the NPSHA for the HPCI pump if the level in the CST (2.0)
is at the suction transfer point.
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NPSHA = u (P-Psat)
b.
If the level in the CST is at the suction transfer point and the (1.0)
suction failed to transfer, how long could the pump operate before cavitation could be expected (i. e., before the CST would be emptied)7 Assume HPCI at rated flow conditions.
QUESTION S.05 (1.00)
SSES Tech Spec 3.1.4.1 states:
(1.0)
"The Rod Worth Minimizer (RWM) shall be operable in operational conditions 1 and 24 when thermal power is less than or equal to 20% of rated thermal power."
Why is the RWM not required to be operable if reactor power is greater than 20%7 QUESTION S.06 (1.00)
CSES procedure EO-100-113, " Level / Power Control", requires a (1.0)
reduction in RPV water level in order to reduce reactor power during an ATWS.
Explain how lowering reactor water level will reduce reactor power.
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QUCSTION 5.07 (2.00)
A startup is about to commence at SSES.
The Reactor Engineer has (2.0)
calculated the estimated critical position for this particular startup.
State the EFFECT on the ECP for each of the following situations.
That is, state whether the reactor will go critical with LESS rod withdrawal, MORE rod withdrawal, or that there will be NO CHANGE in the ECP.
Consider each case separately.
c. RWCU isolates (significant decay heat).
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Moderator temperature drops.
c.
Reactor head vent is inadvertantly closed.
d.
Shutdown Cooling isolates (significant decay heat).
QUESTION 5.00 (2.50)
c. Step 6.30 of SSES GO-100-02 (" Plant Startup and Heatup") directs (1.0)
the operator to continue with control rod withdrawal until a stable positive period of 100 seconds is achieved.
Assume that this period is maintained and that an operator wants to verify the accuracy of the period meter.
He decides to time l
the indicated power increase from 20 on IRM range 3 to 100 on IRM
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range 4.
How long should this take if the period meter is accurate?
l b.
Assume that the reactor scrams from full power.
Explain the (1.5)
response of indicated reactor power level as indicated by nuclear i
instrumentation versus actual core thermal output for approximately I
thirty (30) minutes after the scram.
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QUESTION 5.09 (3.00)
c. Define the term Critical Power (CP).
(1.0)
b.
State how Critical Power would change for each of the following (2.0)
events ( i. e., INCREASE, DECREASE, or NO CHANGE).
Assume that the reactor is at full power.
Consider each event separately.
1.
Loss of a feedwater heater string 2.
Main Turbine Trip (Consider for the time immediately prior to the reactor scram.)
3.
Recirc Flow Control system fails to maximum demand 4.
Feedwater Control system fails to maximum demand QUESTION 5.10 (1.50)
o. There is an orifice located downstream of each pump in the (0.75)
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Control Rod Hydraulic System (CRDH).
What condition are these orifices designed to prevent?
b.
What is the adverse effect to the pump if this condition occurs?
(0.75)
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QUESTION 5.11 (1.50)
i Tech Spec Table 3.4.4-1 specifies the Reactor Coolant chemistry (1.5)
limits.
Explain why the most restrictive chloride limits are l
given for startup conditions.
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DUESTION 5.12 (3.00)
Figure 1 contains charts of several key reactor parameters following a Feedwater Controller Failure to Maximum Demand.
For the areas marked, give the cause of each parameter change as ctated below.
A-State why reactor power rises at '20 seconds then immediately (0.75)
decreases.
B-State why feedwater flow drops sharply at '20 seconds.
(0.75)
C-State why core flow drops at '20 seconds.
(0.75)
D-Explain the variations in steam flow from '20 - 30 seconds.
(0.75)
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QUESTION 6.01 (3.00)
c.-SSES DN-024-OOl lists sixteen (16) Emergency Diesel Generator (1.0)
(EDG) trips for a non-emergency start. Of these, list those EDG trips which are NOT bypassed for an emergency start.
b.
What action (s) outside of the control room would an operator (1.0)
take in order to manually shutdown an EDG f ollowing an emergency start?
c.
Why is it undesirable to stop the EDG by using the Emergency (1.0)
Stop pushbutton (located on local control panel OCS21A) in any other case than an emergency?
QUESTION 6.02 (3.00)
Several RPG scram signals are listed below. For each signal, state (3.0)
whether or not the scram signal can be bypassed (either manually or cutomatically).
IF the signal can be bypassed MANUALLY, give the Reactor Mode Switch position (s) for which the bypass can be initiated and give the operator
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cction(s) taken to initiate the bypass.
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IF the signal can be bypassed AUTOMATICALLY, give the Reactor Mode Switch position and applicable plant conditions (if any) that initiate the automatic bypass.
a.
Scram Discharge Volume Hi Hi Level Scram b.
Drywell Hi Pressure Scram c. MSIV Closure Scram
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QUESTION 6.03 (3.00)
c. From the following list, choose the RCIC interlocks which (1.25)
are defeated when RCIC is controlled from the Remote Shutdown Panel.
1.
RCIC Turbine Overspeed Trip 2.
RCIC Turbine Trip on High Exhaust Pressure 3.
RCIC Turbine Trip on High RPV Water level 4.
Automatic Suction Transfer from CST to Sup Pool on Low CST level S.
RCIC auto initiation on RPV Level 2 (-30")
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b.
Using attached Figure 2,
"RCIC", trace the flowpath for Pressure (1.25)
Control using RCIC while cooling the plant down from the Remote Shutdown Panel, c.
Answer TRUE or FALSE:
If the CST level switch fails low, it is (0.5)
possible to switch the RCIC pump suction back to the CST by closing F031.
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QUESTION 6.04 (2.50)
c. While at power, the ROD DRIFT alarm at 1C651 annunciates. Name (1.0)
three (3) other indications or alarms in the main control room you would look for to verify the control rod drifts.
b.
List three (3) malfunctions within the Control Rod Drive (1.5)
Hydraulic System or in systems that interface with the CRDH system that could be causing the control rods to drift.
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QUESTION 6.05 (2.50)
Assume that a reactor scram occurs for each of the following events.
For each event, match the condition that sends the scram signal to RPS.
(Your answers should read 1-5 with a letter for each answer.
Conditions may be used more than once.)
-EVENTS-1.
Loss of Stator Cooling at 100% power (0.5)
2.
Main Steam Line Low Pressure in RUN Mode (0.5)
3.
Complete loss of Feedwater at 100% power (0.5)
4.
Closure of a single MSIV while at 100% power (0.5)
5.
Turbine trip (without bypass) at 20% power (0.5)
-CONDITIONS-a.
High Neutron Flux d.
RPV Low Water Level b. Low Reactor Pressure e.
High Reactor Pressure c.
Turbine Stop Valve Closure f.
MSIV Closure
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QUESTION 6.06 (2.50)
c. Assume that Drywell temperature increases to 200 deg F during an (1.5)
accident at SSES. For the following RPV water level ranges, state whether indicated level will be HIGH, LOW, or ACCURATE as compared to the actual water level during the accident.
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Wide Range
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Narrow Range 5.
Shutdown Range 3.
Extended Range L.
Upset Range b.
SSES Unit 2 is rapidly depressurized during an ADS blowdown. For the following RPV instrumentation, choose the phrase that correctly completes the sentence.
1.
The narrow range water LEVEL indication will indicate (0.5)
(A LOWER THAN ACTUAL, A HIGHER THAN ACTUAL, or AN ACCURATE)
RPV water level reading.
2.
The narrow range PRESSURE indication will indicate (0.5)
(A LOWER THAN ACTUAL, A HIGHER THAN ACTUAL, or AN ACCURATE)
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QUESTION 6.07 (2.50)
For the following events, choose the auto action or actions, if any, (2.5)
that will be initiated by the APRM Trip Units.
AUTO ACTIONS - ROD BLOCK / HALF SCRAM / SCRAM c. At 100% power, APRM C has thirteen (13) LPRM inputs b.
At 100% power Flow Unit A fails downscale l
c.
APRM B fails downscale at 75% power d.
Flow Unit A oypassed with joystick at 100% power o. APRM E fails upscale in Startup Mode
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QUESTION 6.00 (1.50)
a. List the automatic start signals for the ESW pumps A, B,
C, and D.
(1.0)
Include any applicable time delays.
b. Assume that a Loss of Offsite Power (LOOP) occurs and the ESW (0.5)
pumps are running. If a LOCA then occurs, will the ESW pumps continue to run?
DUESTION 6.09 (2.50)
a. Give the conditions for which the Recirculation Flow Control (1.0)
System will enforce the #2 Speed Limiter. (That is, limit demand signal to 45% of rated speed.) Include setpoints, if applicable.
b.
SSES Unit 1 is operating at 40% power. The Recirc pumps are at (1.0)
minimum speed in individual speed control. If the Manual / Auto Transfer Station is inadvertantly placed in AUTO, briefly explain what happens to the speed of Recirc pump A.
c.
Answer TRUE or FALSE: If a scoop tube lockout occurs at full (O.5)
power, the #1 and #2 Recirc speed limiters will still be able to runback Recire if their respective logics are satisfied.
QUESTION 6.10 (1.00)
SSES Unit 1 OP 149-005, "RHR Operation in the Suppression Pool (1.0)
Cooling Mode", recommends the use of RHR pumps 1P202C and D for use in Suppression Pool Cooling if no LPCI initiation signal is present.
Why are these pumps preferred?
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QUESTION 7.01 (2.50)
s. Assume that a reactor scram from full power is imminent. As the (1.5)
Control Room SRO, you could take certain actions to reduce the impact of the scram on the plant. Name three (3) of the actions that you would take.
b.
The bases of EO-100-01, " Scram", list two (2) purposes served (1.0)
by putting the made switch to SHUTDOWN after a scram occurs.
State the two reasons.
QUESTION 7.02 (1.75)
Assune that the control room must be evacuated due to dense smoke caused by a fire. If the control room is evacuated prior to scramming the plant manually, EO-100-OO9 instructs the operator to open the following breakers: CD2A on RPS distribution panel 1Y2OIA and CBBB on RPS distribution panel 1Y201B.
a.
State 2 purposes (i e., plant equipment response) accomplished (1.0)
by opening these breakers.
b.
Besides opening the breakers listed above, the operator must (0.75)
perform one other local immediate action. State the action.
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2 __P899gggBEg_;_Nggd@61_9pNQ858L _EDEggENCy_ONQ PAGE
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QUESTION 7.03 (3.00)
a. SSES Unit 1 is in Operating Condition 5 with the vessel head (1.5)
removed.
Both loops of the shutdown cooling mode of RHR are then lost. Section 3.3.1 of SSES ON-149-OO1 (Loss of Shutdown Cooling) suggests the following actions. Fill in each blank with the REASON for each suggested step.
3.3.1 a.
MAINTAIN water level 90 to 100 inches on shutdown indication to _
._ _ _ _ _
_ ____
_____________.
-OR-3.3.1 b.
OPERATE Reactor Water Cleanup System at its maximum flow rate in accordance with OP-161-OO1 to
.
3.3.1 c.
START Reactor Recirculation System with at least one pump in minumum speed in accordance with OP-164-OO1 to _______________________
_
_.
- NOTE:
DO NOT consider part "a" when answering parts
"b" and
"c".
b.
Section 3.3.2 a.
of DN-149-OO1, " Loss of Shutdown Cooling",
(0.75)
lists systems available to ADD water to the RPV. List three (3)
of these systems.
c.
Section 3.3.2 b.
of ON-149-OO1, " Loss of Shutdown Cooling",
(0.75)
lists systems available to REMOVE water from the RPV. List three (3) of these systems.
(***** CATESORY 07 CONTINUED ON NEXT PAGE *****)
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DUESTION 7.04 (2.00)
For each of the following plant conditions or events, (2.0)
1.. state the EDP or EOPs (Emergency Operating Procedures) entered (by title or procedure number), and 2.
give the specific EDP entry condition met. Include the setpoint, if applicable, a.
Suppression Pool Temperature = 105 deg F
,
b.
Drywell Pressure = 2 psig c.
Main Steam Line Rad Monitors read 7.5 times normal full power background.
d.
Reactor Pressure = 1100 psig e.
HVAC Zone III exhaust reads 3.0 mrem /hr.
-
QUESTION 7.05 (3.00)
The following questions concern GO-100-OO2, " Plant Startup and Heatup".
a. A CAUTION statement located in Section 6.2 of this procedure (1.0)
states: DPENING MSIVs WITH CONDENSER VACUUM ESTABLISHED AND
'
REACTOR _ HEAD VENT OPEN MAY RESULT IN ___(1) ___ OR ___(2)
.
Fill in each blank.
b.
According to this procedure, under what conditions is the reactor (1.0)
,
j considered critical?
l I
c.
Why is it important to monitor turbine first stage pressure (1.0)
during shell warming?
QUESTION 7.06 (2.25)
Tables SC-2 and SC-3 list Secondary Containment Maximum Operating values for area temperature and radiation levels. (See Figure 3.)
l a.
For Table SC-2, briefly explain the difference between (1.5)
L
" Maximum Normal" and " Maximum Safe" area temperatures.
l b.
For Table SC-3, define " Maximum Normal" area radiation level.
(0.75)
!
(***** CATEGORY 07 CONTINUED ON NEXT PAGE
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QUESTION 7.07 (1.50)
e. OP-143-OO1, "SJAE and Mechanical Vacuum Pump", states that the (1.0)
mechanical vacuum pump should not be operated when reactor power is greater than 5%. Give two (2) reasons for this precaution.
b.
Section 3.6 of OP-143-OO1, " Main Condenser Vacuum Breaker (0.5)
Dperation", states that if steam seals are lost and cannot be restored immediately, Main Condenser vacuum must be broken.
Briefly explain why.
QUESTION 7.08 (2.50)
e.
ON-135-OO1, " Loss of Fuel Pool Cooling", provides alternate (1.0)
'
methods of cooling spent fuel assemblies. Assume that the fuel pool
-
requires makeup because of a leak. List three (3) sources of makeup to the fuel pool.
.b.
Assume that a Loss of Fuel Pool Cooling has occurred as a result (0.5)
of PCV-11036 failing closed. All attempts to regain cooling (including the use of RHR in Fuel Pool Cooling Assist Mode) have failed. The only means of heat removal available is to allow the
,
fuel pool to boil.
Complete the following sentence with the limits specified in DN-135-OO1:
Boiling should not occur bef ore _(1) _ hours after loss of cooling and the fuel pool level must be maintained
_ 2) _ feet above the irradiated fuel bundles.
(
c.
Assume that you are following section 3.5.2 of DN-135-OO1, which (1.0)
contains the steps in providing fuel pool boiling. Briefly explain i
WHY you must ensure that Zone III exhaust ventilation must be aligned to exhaust the area directly over the fuel pool.
QUESTION 7.09 (1.50)
,
i Assume-that Startup Bus 20 was lost due to a fault from Startup (1.5)
Transformer T-20. In accordance with DN-OO3-OO2, " Loss of Startup
.
Bus:20", what are the three (3) conditions necessary for tie breaker OA10502 to close to f eed startup bus 2O?
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PROCEDURES - NORMAL _ABNgRMAL _gMERggNCy_ANg PAGE
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DUESTION-7.10 (3.00)
c. Using the attached graphs of the Heat Capacity Temperature Limit (1.5)
(HCTL) and the Heat Capacity Level Limit (HCLL) (Figures 4 and 5),
determine the suppression pool water level for which a controlled RPV depressurization would be required i n accordance with the Primary Containment Control Procedure (ED-100-103).
Given:
Reactor pressure = 500 psig Suppression Pool temperature = 170 deg F b.
Step SP/L-B of this procedure states that HPCI and RCIC should (1.0)
not be used if Suppression Pool water level decreases below 18.5 feet. Give two (2) reasons for this step.
c.
Why is rapid RPV depressurization required if Suppression Pool (0.5)
level reaches 12 feet?
QUESTION 7.11 (2.00)
OP-153-OO1, " Standby Liquid Control System", directs the operator to inject SBLC if required by EO-100-102 (RPV Control).
a. What is the limiting condition given by EO-100-102 at which (1.0)
I SBLC MUST be injected?
b.
Under what two (2) conditions (as stated in EO-100-102) may (1.0)
'
SBLC injection be terminated?
l l
(***** END OF CATEGORY 07
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Dz__0DglNigIB8IlVE_P89CEDUBEg1_CQNQlIlgNgi_8NQ_LIMlI@IlgNg PAGE
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QUESTION O.01 (3.00)
a.
State the four (4) Safety Limits as given in Section 2.1 of the (2.0)
SSES Technical Specifications. For each, include any applicable setpoints or conditions and the applicable plant operational condition (s).
b.
What IMMEDIATE action must be taken in accordance with the SSES (1.0)
Tech Specs if a Safety Limit is exceeded while the unit is at power?
QUESTION 8.02 (3.00)
Classify each of the following events as an UNUSUAL EVENT, an ALERT, a SITE AREA EMERGENCY, or a GENERAL EMERGENCY. A copy of the SSES Emergency Classification Guide (EP-IP-OO1) is attached as Figure 6.
i 1.5 1.
Building Vent Monitoring System indicates a total site 14TCI)
release rate for I-131 equal to 1.5 E3fcGi/ min.
break)mh(dt of Mm CmtainM t5
"C "
An RO reports 1,1<ti)
2. A h linea RPV water level at -180" on the Fuel Zone Range. He then reports that "SIV
"C" has failed to close.
ut nis/tr-chawjd V
OLA60dVd -
k y(
gyp for 5 minnb.
1209
QUESTION B.03 (2.00)
Answer the following TRUE / FALSE questions concerning the SSES Emergency Plan.
a.
If the Emergency Plan is initiated, the Shift Supervisor (SS)
(0.5)
!
assumes the role of the EMERGENCY DIRECTOR until relieved by the PLANT SUPERINTENDENT.
b.
The TSC is the central location for offsite emergency management.
(0.5)
c.
With regard to in plant radiation levels, The RESTORATION (0.5)
,
ORGANIZATION may be implemented as long as in-plant radiation
'
levels are decreasing.
I d.
It is the responsibility of the RECOVERY MANAGER to make the (0.5)
final determination regarding the establishment of the RESTORATION ORGANIZATION.
!
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QUESTION 8.04 (2.00)
a.~
List the minimum shift complement for Operations personnel as (1.5)
stated in DI-AD-015, " Minimum Shift Manning". Include position and type of NRC license held, if any.
b.
Fill in the blank: The Day Shift Supervisor may REDUCE the (0.5)
Minimum Shift Complement from that given in DI-AD-015, as long as
_ _
_ _
__________________
.
QUESTION 8.05 (1.50)-
MATCH the following tags with their designated meaning as given in (1.5)
SSES AD-DA-103, " Protective Permit and Tag System".
-TAG-
-MEANING-m.
Yellow 1.
The tagged device is not to be operated until properly cleared by the permit holder.
b.
Red 2.
The tagged device is in a controlled status-and is not to be operated unless the restrictions on the tag are met.
c.
Gtriped 3.
The tagged device is not to be operated except by request of the permit holder
,
AND on orders of the S.D.
representative.
]
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I
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QUESTION B.06 (3.00)
( cftcf sDY'
is bypassed.J SSES Unit 1 is operating at 90% power. APRM "B" APRM
"F" then fails upscale and is removed from service by the I & C Department.
c.
What actions are required by the SSES Tech Specs for the above (1.0)
situation?
b.
Assume now that in addition to the above failures, APRM "A" is (2.0)
bypassed. Also, the I&C Supervisor informs you that the Channel Functional Test on APRM
"C" i s due immediately.
What actions are required by Tech Specs in order to allow this
, surveillance to be completed?
j
NOTE: USE THE ATTACHED TECHNICAL SPECIFICATIONS TO ANSWER THE ADOVE QUESTION.
DE SURE TO REFERENCE ALL APPLICABLE TS THAT YOU USC IN DEVELOPING YOUR ANSWER.
QUESTION 8.07 (3.00)
SSES Unit 1 is at 100% power. The
"B" Main Steam Line Radiation Monitor fails downstale.
a.
Can reactor operations continue? What Tech Specs apply?
(1.0)
b.
What Tech Specs would be applicable if Main Steam Line Monit r (2.0)
"C" then failed downscale?
- ****
NOTE: USE THE ATTACHED TECHNICAL SPECIFICATIONS TO ANSWER HE ABOVE QUESTION. BE SURE TO REFERENCE ALL APPLICABLE TS THAT YOU USE IN DEVELOPING YOUR ANSWER.
- ********
LK1 (130 03/I'l87 Chomyd -h) ;
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\\
if MS wkod n KA' aFPM
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QUESTION 8.00 (3.00)
c. SSES Unit 1 is at 50% power. You learn from the previous Shift (1.0)
Supervisor that during individual control rod scram time testing, control rod 19-30 was f ound to be immovable in the fully withdrawn position. The testing is scheduled to continue during your shift.
What actions are you required to take in accordance with the Tech Specs if control rod 27-38 is also found to be immovable in the f ull y wi thdrawn posi tion?
(NOTE-A core map is attached as Figure 7.)
b.
If a single control rod was found to be immovable with normal (2.0)
Control Rod Drive pressure during Startup, could the startup continue?
Give any applicable Tech Spec requirements.
NOTE: USE THE ATTACHED TECHNICAL SPECIFICATIONS TO ANSWER THE ABOVE DUESTION. BE SURE TO REFERENCE ALL APPLICABLE TS THAT YOU USE IN DEVELOPING YOUR ANSWER.
3(iL{ V1 QUESTION 8.09 (2.50)
2, g l330 cbydM9W bas
&
While at power, 125 VDC battcry ch= g=c ID61/ and 250 V bettcry ID652 become inoperable.
DC kk#
a.
Give all applicable Technical Specifications.
(1.0)
b.
Give all applicable Tech Specs, if, in addition to the above (1.5)
malfunctions, 480 V swing bus 19229 became inoperable.
NOTE: USE THE ATTACHED TECHNICAL SPECIFICATIONS TO ANSWER THE ABOVE QUESTION. BE SURE TO REFERENCE ALL APPLICABLE TS THAT YOU USE IN DEVELOPING YOUR ANSWER.
,
.
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QUESTION 8.10 (2.00)
SSES Unit 1 is in cold shutdown. A unit startup is scheduled to (2.0)
commence in three (3) days. The Maintenance Supervisor reports the failure of the outboard blower for the MSIV Leakage Control System.
He estimates that it will take ten (10) days to return the blower to service.
In view of the above malfunction, is it possible to maintain the present startup schedule without violating the SSES Tech Specs?
Explain your answer referencing all applicable Tech Spec requirements.
l
NOTE: USE THE ATTACHED TECHNICAL SPECIFICATIONS TO ANSWER THE ADOVE QUESTION.
DE SURE TO REFERENCE ALL APPLICABLE TS THAT YOU USE IN DEVELOPING YOUR ANSWER.
!
l QUESTION B.11 (2.00)
l For each radiation worker described below, determine if any SSES Administative or Federal (10 CFR 20) radiation exposure limits have been exceeded. Show all work and state which particular limits, if any, have been exceeded.
a.
45 year old male, with an NRC Form 4 (1.0)
lifetime exposure = 100 rem He has received 2000 mrem this current quarter.
l l
b.
35 year old male, with an NRC Form 4 (1.0)
lifetime exposure = 50 rem He is a new employee and has no documentation of his radiation exposure for the current quarter. He then receives 400 mrem during his first month at SSES, which is in the same calendar quarter.
(***** END OF CATEGORY 08
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THEORY OF NUCLEAR POWER PLANT OPERATION _FLUIpS _ANQ PAGE
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ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLUNAUSKI, L.
.
ANSWER 5.01 (2.00)
e.
1.
The single rod of greatest reactivity worth is withdrawn.
(0.5)
2. The reactor is cold; 68 deg F. y (0.5)
3.
The reactor is Xenon free.
-02 l
b.
O.38*/ delta k/k
.
(0.5)
(with the highest worth rod analytically determined, or O.28%
,
delta k/k with the highest worth rod determined by test.)
REFERENCE SSES SCO23 A-6, Reactor Control Specific Objectives 10, 11, 12 pages 7-10 Tech Spec 3/4.1.1 Bases
.
292002, Neutron Life Cycle (Group II Rx Theory)
K1.10 Define SDM 3.5 K1.14 Predict change in SDM with core param changes 2.9 i
ANSWER 5.~ O2 (2.00)
(2 required, 1.0 each)
1.
At EOC, there is a slower scram reactivity insertion rate (0.5)
because the' control rods are further 6ithdrawn than at BOC (0.5).
2.
The void coefficient is more negative at EOC (0.5), so more positive reactivity is added at EOC than BOC for the same change in core void content (0.5).
3.
Beta, the delayed neutron fraction, is smaller at EOC (0.5).
This results in a shorter reactor period at EOC than at BOC for the same positive reactivity addition (0.5).
REFERENCE SSES Transient Analysis; SCOO7C, Pressure Increase Transients Specific Objective 3.b pages 19-20 292003, Reactor Kinetics (Reactor Theory Group II)
K1.06 Explain effects of delayed neutron fraction on Rx power.
3.7 295001, Partial loss of forced core flow circulation (AE Group II)
AK1.03 704 of op implic of the thermal limit concept.
4.1
- -.
.
5:__IUgD8y_DE_NUg6E86_EgMEB_ELONI_DEEBATION _FLUIQS _ANQ PAGE
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ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSK1, L.
ANSWER 5.03 (1.50)
Maximum delta T = 145 deg F (0.25)
X - 395 deg F = 145 deg F (0.5)
X = 540 deg F Psat for 540 deg F -> 962.79 psia (0.5)
962.79 - 14.7 = 948.09 psig (0.25)
REFERENCE Saturated Steam Tables SSES Procedure OP-164-OO1 SSES SCO23 D-1 Rev O, Thermo Fundamental; Specific Objective 6 SSES SCO23 D-3.Rev O, Steam Tables; Specific Objective 1 293003, Steam (Group II Thermodynamics)
K1.23 Use saturated steam tables 3.1 ANSWER 5.04 (3.00)
e. NASHA = L) (P-Psat) +z h1 (2.0)
= [O.016ft3][(15 - 0.5)1bf3[144in2] + (673.75 - 650.25)ft - 7.1ft Ibm in2 ft2
= 33.4 + 23.5 -7.1 feet
= 49.8 feet b.
CST vol ume = Tr r2 h = TC (20 f t ) 2 (3.6 ft)
4524 ft3 (1.0)
=
HPCI flow = 5000 gpm Time = 4524 ft3 * 7.48 gal /ft3 * 1 min /5000 gal
= 6.77 min REFERENCE i
SSES Unit SCO23 E-4, Fluid Mechanics-Pumps; Objective 8 l
Saturated Steam Tables 293006, Fluid Statics (Group III Thermodynamics)
K1.10 Define NPSH 2.8
291004, Pumps (Components)
K1.06 Need for NPSH; effects of loss of suction 3.3
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i
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THEORY OF NUCLEAR POWER PLANT OPERATION _ FLUIDS _AND PAGE
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IBERMODYNAMICS
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ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 5.05 (1.00)
N The purpose of the RWM is to limit rod war (1.0)
.
When thermal power is greater than 20%,
here is no rod with high Gnough worth which, if dropped at the design rate of the velocity limiter, could result in cladding damage (a peak enthalpy of 280 cal / gram).
REFERENCE SSES SYO17 K-6 RWM, Objectives 1,
5, pages 2-3 Mitigation of Core Damage, SCOO6 L SSES TS Bases 29 314, Inadvertant Reactivity Addition (Group I EA Evol)
SG #3 KN of LCOs and safety limits 4.3 ANSWER 5.06 (1.00)
If RPV level is decreased, the natural circulation driving head (1.0)
>
is decreased.
This reduction in core flow will cause an increase in voiding and decreased reactor porter.
REFERENCE SSES SCOOO6 K SSES EUP Specific Objectives 2,
EA2.02 AB to relate RPV water level to ATWS 4.2 ANSWER 5.07 (2.00)
.
a. more rod withdrawal (O.5)
b.
less rod withdrawal (0.5)
c. no change in ECP (0.5)
d. more rod withdrawal (0.5)
REFERENCE SSES Licensed Operator Science, SCO23 A-7 Objective 3 292005 Control Rods (Group II Reactor Theory)
K1.09 Explain direction of change in CRW for a change in mod temp. etc..
5:__IUEg8y_gE_NUCLEOB_EgWE8_ELONI_9EEBOIlgN _ELUlygz_8NQ PAGE
z IUE8dggyN@dlC@
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ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 5.08 (2.50)
a. P = Po e**(t/T) --> t = T In (P/Po)
(1.0)
t = 100 sec * In(100/20)
100 sec * 1.61
=
= 161 sec = 2.7 min b.
The nuclear instrumentation will show a continual drop in the (0.75)
fission rate on a -80 second period.
The fission rate will continue to decrease until it drops into the source range (where subcritical multiplication effects will cause the fission rate to stabilize).
Core thermal output will drop to the decay heat level, which is (0.75)
about 67. of the original power level.
REFERENCE SSES SCO23 A-9 Licensed Operator Science pages 1-4
" Characteristics of an Operating Plant" Objectives 6, 7,
SSES Transient Analysis SCOO7 B, Section 2.6, Objective 3 292008 Reactor Operational Physics (Group I Reactor Theory)
K1.08 Describn power, period response once criticality reached 4.1 K1.27 Reactor power response to control rod insertion 3.5 K1.29 Define decay heat 3.6 ANSWER 5.09 (3.00)
a. Critical Power is the bundle power needed to produce the critical (1.0)
quality or the bundle power needed to cause OTB to occur in the bundle.
b.
1.
(inlet subcooling ^)
CP increases (0.5)
2.
(pressure ^)
CP decreases (0.5)
3.
(core flow ^)
CP derrrrr;; inenaeu (0.5)
4.
(inlet subcooling ^)
CP increases (0.5)
REFERENCE SSES SCO23 G-3 Specific Objectives 3.3, 3.4, 3.6 293009 Core Thermal Limits (Group 1 Thermo)
K1.17 Define CP 3.7 K1.22 KN of effect of subcooling on CP 3.3 K1.23 KN of effect of core flow of CP 3.2 K1.24 KN of effect of pressure of CP..
D __INEggy_gE_NgCLE@B_PQMEB_EL@NI_QPEB8IlgN _ELylg@t_@NQ PAGE
t IUEBDgQyN@gICS
.
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 5.10 (1.50)
c. Pump runout (excessive system flow demands when the scram (O.75)
accumulators are being charged).
b.
The increase in flow rate will cause the pump motor to draw (O.75)
a larger current amount which can result in overheating of the motor.
-REFERENCE SSES SCO23 E-4 Pumps page 22, Objective 5 291004 Pumps (Components)
Kl.12 Pump runout-corrective reasures, etc.
2.8 ANSWER 5.11 (1.50)
The limit on chloride concentration is to prevent stress corrosion (0.5)
cracking of the stainless steel.
During startup, there is not much deaeration taking place and the (0.5)
dissolved oxygen content in the coolant may be high.
With a higher oxygen concentration, the chloride limit must be (0.5)
lower to ensure that stress corrosion cracking is prevented.
t REFERENCE SSES System Lesson Plan, RWCU SYO17 L-1 page 18, Objective 6 204000 RWCU (Group II Plant Systems)
SG #6 KN of TS Bases for LCOs 3.4
,
l l
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I i
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i
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i
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_
_
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5-IMEQBy_QE_NyCLE98_EQWE8_ELONI_gEEg8IlgN _E(ylppg_@NQ PAGE
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' ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 5.12 (3.00)
A-Rx power increases due to increased subcooling (+p) then (0.75)
decreases sharply due to scram from turbine trip (hi level).
B-FW flow decreases due to FW pump trip caused by high RPV water (0.75)
level (+54").
C-Core flow decreases due to EOC-RPT initiation.
(0.75)
D-(Steam flow drops to zero upon turbine trip)then oscillates as (0.75)
the BPVs (and the SRVs) open to reduce reactor pressure.
REFERENCE SSES SCOO7 D pages 9-12, Objective 4 259002 Rx Water Level Control System (Group I Systems)
KN of effect of loss of FWCS on following parameters:
K3.02 FW system 3.7 / K3.06 Main Turbine 2.0 /
K3.07 RFC 2.9 Kl.lO KN of relation between RPS and-FWCS 3.9
.
.
6:__P(@NI_gYgIEgg_gEgigy3_CgglBg61_@Ng_ lng 16UDENI@IlgN PAGE
- ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 6.01 (3.00)
e.
1.
High Generator differential current (0.33)
2.
Low Engine lube oil pressure (0.33)
3.
Engine overspeed (0.33)
b.
By taking the mode selector switch to local and depressing the (1.0)
emergency stop pushbutton.
c.
Automatic cooldown is bypassed because the unit shutdown relay (1.0)
is opened, which prevents the cooldown circuit from energizing.
(Shid5 )
(>*d Mp )
REFERENCE SSES SYO17 G-1, Objective 5 DN-024-OO1 264000 Emergency Diesel Generators (Group I Systems)
K1.04 KN of E D/G relation to cooling water system 3.3 K6.03 KN of loss of lube oil effect on E D/G 3.7 K4.02 KN of E D/G trips 4.2 t
ANSWER 6.02 (3.00)
a. Manually bypassed (0.5) in the Shutdown or Refuel Modes (0.5) by placing the SDV high water level bypass switch in the BYPASS position.
(manual, keylocked, on benchboard IC651)
(0.5).
b.
Signal NOT automatically or manually bypassed (0.5).
c. Automatically bypassed (0.5) when mode switch not in RUN (0.5]
LR 3/(([f7 n_
r_
+._ _ ;. m L c i m. 600 gu y 40.5).
CA"~ ;
'
REFERENCE
- P"* " ~
SSES SYO17 L-5 Objective 6 i
l 212000 RPS (Group I Systems)
A4.04 AB to bypass SDV hi level scram signal 3.9 K4.12 KN of RPS interlocks which provide bypassing of 4.1
,
scram signals (auto or man)
!
l I
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i
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
.
ANSWER
.6.03 (3.00)
'c.
2-3-4-5 (1.25)
b.
See attached Figure.
(1.25)
c.
TRUE (0.5)
REFERENCE SSES SYO17 C-5, Objective 5d 217000 RCIC (Group I System)
K1.01 KN of RCIC relation with CST 3.5 295016 Control Room Abandonment (Group I EA Evol)
AA2.03 AB to operate to control Rx pressure 4.4 ANSWER 6.04 (2.50)
c.
SCRAM DISCHARGE VOLUME NOT DRAINED on panel IC651 (3 reqd, 0.33 each)
ROD OUT BLOCK on panel 1C651 SCRAM PILOT VALVE AIR HEADER LOW PRESS ALM on panel 1C601 Full Core Display Red Drift lights illuminated LK Wet-atto - C KW yeraW%r5 b.
1.
Scram valve leakage YoWtr Ltd CHAM *f8 (0.5)
2.
Cooling water pressure high (OrNow)
(0.5)
i 3.
Low Instrument Air Supply pressure (0.5)
acctyf atro Seltet Wd lnWf (Wl/IWN.M adlydtdl0 cal lj
REFERENCE SSES ON-155-OO6 Rev 2 SYO17 K-2 295019 Loss of Instrument Air (Group II AE Evol)
l AK2.01 KN of effect of air loss on CRD system 3.9 j
201001 CRDH (Group II System)
l A2.12 AB to predict impact of high cooling water flow on CRDH 2.9 l
A2.11 AB to predict impact of valve openings 2.7 I
!
!
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6 __PL@NI_gYSIEgg_pEQ1@N _CQNIBQL _@NQ_lNglByDENI@ll@N PAGE
t t
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 6.05 (2.50)
1.
c.
(0.5 each)
2.
(MSIV closure)
f.
3.
d.
4.
(hi steam flow-MSIV closure)
f. (or a.) (Or t.)
5.
e.
(or a.)
REFERENCE SYOl7 L-5 Objectives 5,6 SSES Transient Analysis, SCOO7C-Objective 3.2 295025 High Reactor Pressure (Group I EA Evol)
EK2.01 KN of relation between hi ex press and RPS 4.1 estutMcn w10 "<(" fl0W
" 'I' 8 I ANSWER 6.06 (2.50)
g g _. gag du b/ Wee-If k(cire, diu opediN>
o, c. 4.
la" 4.
accurate (O.
each)
2.
accurate 5.
high 3.
high 6.
high b.
1.
higher than actual g 3fg[37 (0.5)
RF ts0y'O
'
2.
lower than actual (0 5)
\\0wb~&o(-W (acewadt)
REFERENCE SYOl7 J-2 Objective 9 / SCOO6 G Objectivea 4,5 295028 High DW Temperature (Group II EA Evol)
EA2.03 AB to interpret RPV level 3.9 ANSWER 6.07 (2.50)
c. Half Scram, Rod Block (0.25 each)
b. Hali Scram (RPS A)3 Vodblock (0.5)
c. Rod B1ock
\\c h a. h f(c W W t (O.5)
d.
None d[ 3/IIIII (0.5)
O.
Rod Block, Half Scram (0.25 each)
REFERENCE SYO17 I-4 Objective 2 215005 APRM (Group I Systems)
KN of interrelations between APRM and:
- __
.__
.
6:__ PLAN 1_gYglEdg_QEglGN _CgNIQQL _8Np_lNgIQUDENI@llgN PAGE
t g
ANSWFRS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
'
K1.01 RPS (4.0)
K1.03 RUM (3.5)
ANSWER 6.08 (1.50)
c. Pump A and B start 40 seconds after respective DG start signal.
(0.5)
Pump C starts 44 seconds after C DG start signal.
(0.25)
Pump D starts 48 seconds after D DG start signal.
(0.25)
b.
YES (0.5)
REFERENCE SSES SYO17 M-1 Objective 4 264000 Emergency Generators (Group 1 Systems)
AB to predict impact of following: A2.09 Loss of AC 4.1 A2.10 LOCA 4.2 ANSWER 6.09 (2.50)
(0.2cxh)
a. Condensate pump discharge pressure <100 psig OR (0.2 )
Individual FW pump flow is <20%
OR (O. 5)
1 or 2 FW Heater Hi H1 Level AND (O 25)
RPV level < 30" (Low level alarm point)
LK.' 3fgg[py (.25)
OR M Circ.Why pcketivL tYtp.
b.
The Recirc pump A speed will increase to the lower limit of (1.0)
the Master Controller (which is ed% speed).
c.
FALSE (0.5)
REFERENCE SYO17 L-9 Objectives 2,3 202002 Recirc Flow Control (Group 1 Systems)
K4.02 KN of RFC interlocks for Recirc pump speed control 3.0 K3.OS KN of effect of RFC malfunction on Recirc pump speed 3.3 l
.
$1__P(@@I_@Y@l@D@_Q@@l@Nt_Q9MI69(t_@@p_ INSTRUMENTATION PAGE
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 6.10 (1.00)
These pumps would trip if a LOCA signal was received for Unit 2, (1.0)
leaving RHR pumps 1P202A and B available f or LPCI initiation on Unit 1.
REFERENCE OP-149-OO5, Rev 3 219000 Suppression Pool Cooling Mode (Group II Systems)
K1.04 KN of relation between LPCI pumps and Sup Pool Cooling 3.9 i
!
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a Z.__PgggEggBEg_;_NggdOL _8gNQBM8(t_EdEggENGy_8NQ PAGE
t 809196991GOL_ggNIBgL
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ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 7.01 (2.50)
a.
Reduce recirc pump speeds to minimum.
(3 required, 0.5 each)
Shift Unit Auxiliary Busses to Startup Bus.
Start Main L-O Suction and Turning Gear Oil Pumps.
b.
1.
It seals in the scram signal by sending an additional scram (0.5)
signal to RPS.
2.
It changes the plant mode so that the MSIVs remain open if (0.5)
MSL pressure drops to 861 psig. [This maintains the steam supplies to equipment important for plant cooldown and shutdown (the FW pumps, the SJAEs, Offgas, and turbine seals) and keeps the Main Condenser available as a heat sink.]
REFERENCE EOP 100-101 Lesson Plan, page 2 295006 Scram (Group I EA Evol)
AA1.04 AB to operate / monitor Recirc related to scram 3.2 AA1.01 AB to monitor RPS-Scram 4.2 AK2.04 KN of turbine trip logic related to scram 3.7 ANSWER 7.02 (1.75)
a.
1.
The reactor is scrammed.
(0.5)
2.
The inboard and outboard MSIVs and MSL drains are isolated.
(0.5)
l b.
Manually close RFP Discharge Isolation Valves HV-10603A-C.
(0.75)
l l
REFERENCE EO-100-OO9 i
295016 Control Room Abandonment (Group I EA Evol)
AK2.02 KN of local control stations 4.1 l
l
-
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Z __PggCEDUggg_;_gggd@6t_@BNggd66t_EdEggENCy_8NQ PAGE
R_ AD_I O_ L_O_ G_ _I C_ A_ L_ _C_O_N__T R_ OL_
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ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
Wt b~ NIM N W"
ANSWER 7.03 (3.00)
c.
3.3.1. a - establish natural circulation. M pKttd hhtd $tyRh6CodY(16.5)
b - ensure maximum heat removal. 3 pd CtrC Glsc (0.5)
gg c
prevent thermal stratification.
cwt + m f ~ 4- ( 0. 5 )
- yiar(WA)r Wt Maved ) "' ' - ~
b.
RWCU return to vessel Condensate (3 required, 0.25 each)
CRD Core Spray Condensate Transfer LPCI c.
RWCU suction from vessel (3 required, 0.25 each)
RWCU letdown Skimmer Surge Tank Letdown Fuel Pool Cooling SRV Blowdown REFERENCE SSES ON 149-001 295021 Loss of SDC (RHR)
(Group II EA Evol)
KN of reasons for following responses: AK3.01 Raise Rx Water Level 3.4 AK3.04 Max RWCU flow 3.4 AK3.OB Alt Heat Removal 3.8 KN of oper implic of SDC concepts:
AKl.02 Thermal Stratification 3.8 AK1.03 Adeq Core Cooling 3.9 AK1.04 Natural Circulation 3.7 l
LE3AQfi gey,get a3 g, ye, e MWw, M - WO -W ANSWER 7.04 (2.00)
l c. Primary Containment Control - Sup Pool Temp > 90 deg F (0.4)
i b.
RPV Control and PC Control - DW pressure > 1.72 psig (0.4)
c.
RPV Control - 7 X NFPB-MSIV isolation signal (0.4)
d.
RPV Control - >1037 psig (0.4)
I a. Secondary Containment Control Zone III > 2.5 mrem /hr (0.4)
-
REFERENCE PPOO2 Specific Objective 1, page 3 l
223001 Primary Containment (Group I System)
SG #15 AB to recognize EDP entry conditions 4.2 290001 Secondary Containment (Group I System)
SG #15 AB to recognize EDP entry conditions 4.1
,
.
!
,
. _ _... _ _ _ _
_,.
. _ _ _
__r
- _.. -,.
_
. _ _ _ _ _ _, _. _ _ _., _.. _ _.
,
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BODI96991GOL_ggNIBg6
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ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 7.05 (3.00)
c.
1.
placing the drywell under vacuum, or (0.5)
2.
may cause long path recirculation to leak into vessel.
(0,5)
b.
The reactor is considered critical when a slightly positive stable ( 1. 0 )
period is achieved with increasing neutron flux and no rod motion.
c.
It is possible to inadvertantly "unbypass" the turbine stop and (1.0)
control valve closure scrams if first stage pressure is allowed to increase to Jd*/. of rated.
REFERENCE G0-100-002 292000 Reactor Operational Physics (Group I Rx Theory)
Kl.07 Define criticality as related to a Rx startup.
3.9 245000 Main Turbine (Group II Systems)
K1.04 KN of cause/effect relations between RPS and Mn Turb 3.7 ANSWER 7.06 (2.25)
g o.ty e.
" Max Normal" = the Tech Spec isolation setpoints of the Leakage (0.75)
Detection System.
" Max Safe"
= maximum environmental temperature at which (0.75)
continued operability of affected equipment is no longer assured.
b.
" Max Normal" = ten times the Area Rad Monitor alarm setpoint.
(0.75)
i
'
REFERENCE EO-LOO-104 Bases / PPOO2 Objective 3 290001 Secondary Containment (Group I System)
SG #10 AB to explain all system limits 3.4 i
<
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'Za__Pggggggggg_;_NgBd@61_9BNggd@61_EDEBgENgy_8Ng PAGE
R_ A D _I O_ L O_ G_ _I C_ A L _ C_ _O N _T R_ O L
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ANSWERS -
SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
f
, ANSWER 7.07 (1.50)
c.
1.
Hydrogen / oxygen explosion hazard (0.5)
2.
Unprocessed radiolytic gas discharges (0,5)
b.
To prevent drawing cold air across the turbine seals.
(0.5)
(to prevent bending of the shaft)
REFERENCE SSES OP-143-OO1, Rev 7 271000 Offgas (Group II Systems)
K1.06 KN of cause/effect relation between OG and Mn Steam 2.9 Plant Wide Generics K1.15 KN of safety procedures related to Hydrogen 3.0 ANSWER 7.08 (2.50)
c. Demineralized Water Storage
-
(3 required, 0.33 each)
Refueling Water Storage Tank RHR Service Water Emergency Service Water Fire Protection System b.
(1) - 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />, (2) - 22 feet (0.25 each)
c.
Because the airborne radioactivity will increase with boiling.
(1.0)
REFERENCE ON-135-OO1, SYO17 L-2 Objective 5 l
233000 Fuel Pool Cooling (Group III System)
K3.06 KN of effect on area rad levels if FPC lost 3.2 K4.01 KN of FPC design feature which maintain adequate level 3.2 SG #5 KN of limiting conditions and safety limits 3.4
.
b
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-22__PggCEDyBEg_;_NgBN@L _9pN9BDOL _EgggggNCy_@NQ PAGE
2
RA__D_I DL_OG I C AL___ CON __ TROL ___
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ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 7.09 (1.50)
~
1.
All tie bus lockouts reset (0.5)
2.
Control switches in normal after trip position (0.5)
3.
Unit Auxiliary Busses 12A and 12B not already being fed (0.5)
by SU Bus-20.
REFERENCE.
ON-OO3-OO2 262001 AC Elec Dist System (Group I System)
A3.03 AB to monitor automatic actions of AC Elec Dist-Load shedding 3.5 A1.05 AB to monitor changes in AC Dist-Breaker lineups 3.5
,
ANSWER 7.10 (3.00)
j I-a.-From the HCTLll urve, if Rx pressure = 500 psig, g4 y/$/37 (0.5)
l SP temp ->
deg Fg3 Delta T = 1-170 = pc deg F
[3 (0.5)
-
From the HCLL curve, if delta T = p6 deg F, G
(0.5)
,
!
ADS should be initiated prior to reaching JWi feet.
l-b.
1.
If suction is taken from the Suppression Pool, the HPCI and (0.5)
!
RCIC pumps will cavitate.
2.
Both system discharge spargers will uncover, and'will blow (0.5)
steam directly into the Suppression Chamber.
c.
Containment pressure suppression function is lost with suppression (0.5)
pool level below 12 feet. (Containment failure could result if a primary system break occurs.)
g gg.g 5tpct atse-toJs of-NWtg - Cf /LVCl Fuction,69 ECCS REFERENCE SSES EO-100-103 Bases g
,g Ly
,
295030 Low Suppression Pool Water Level (EA Evol Group I)
b"NA"I-EK1'.03 KN of Heat Capacity 4.1 EK3.07 KN of NPSH for ECCS pumps 3.8 EK2.08 KN of SRV discharge submergence 3.8
.
i
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$
l
!
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.
.
. - -.
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RA_D_IO_L_O_G__ICA_L_C_O_N__TR_OL_
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ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 7.11 (2.00)
a.
Before suppression pool temperature reaches 110 degrees F if not (1.0)
shutdown with rods.
b.
1.
If all control rods are inserted to Minimum Subcritical Banked (0.5)
Withdrawal Position (LE O2 with one control rod stuck out).
T~ ( dkk YCd,> in 2.
OR, SLC tank reaches 100 gallons.
(0.5)
REFERENCE OP-153-OO1 EO-100-102 211000 Standby Liquid Control (Group I System)
'A1.01 AB to monitor tank level changes 3.7 K5.03 KN of oper implic of SDM 3.5 C5 KN of LCOs and saf ety limits 4.4
!
>
l l
'
[
)
- r l
,
_.
_ _ _.
-
.
.
,
_
_. _ _ _
..
.
..
&
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~9 __89MINigIBOIlyE_P8QCEDUBE@t_CQNpillgN@t_8NQ_LIMlI@IlgN@
PAGE
- ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
&gh -
SL = 0.1 ANSWER B.01 (3.00)
Oycm = 0.2 c.
1.
THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with (0.5)
the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.
--
OP CON 1 and 2 2.
The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than (0.5)
1.06 with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.
OP CON 1,
3.
The reactor coolant system pressure, as measured in the reactor (0.5)
vessel steam dome, shall not exceed 1325 psig.
OP CON 1, 2,
3,
4.
The reactor vessel water level shall be above the top of the (0.5)
active irradiated fuel.
OP CON 3, 4,
, 0. L b.
Be in at least hot shutdown within two (2) hours (and comply with (1.0)
-
the requirements of TS 6.7.1.- Safety-Limit Violation).
REFERENCE SSES Tech Specs Su.Mem G.~1.1.0 295025 High Reactor Pressure (Group I EA Evol)
EK1.05 KN of exceeding safety limits 4.7 ANSWER B.02 (3.00)
t. S 1.
ALERT Il-rCT 2. % AL EMERGENCY f.1 rC1 sire r.5 (K M (6/ti REFERENCE SSES EP-IP-OO1 294001 Plant Wide Generic A1.16 AB to take actions in the facility E Plan 4.7
.,
-
.
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_ _.
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.__
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s___8DglNigIB@IlVE_PBQCEDUBE@t_CgNQlligN@z_@ND_LldlI@IlgNg PAGE
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ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER-8.03 (2.00)
a. TRUE.
(0.5)
b.
FALSE (EOF)
(0.5)
c.
TRUE (0.5)
d.
FALSE (VP Nuclear)
(0.5)
REFERENCE SSES E Plan 294001 PW Generic A1.16 AB to take actions as required by the E Plan 4.7 ANSWER 8.04 (2.00)
a.
Shift Supervisor SRO -1 Nuclear Plant Operator -5 (1.5)
Unit Supervisor SRO -2 Radwaste Operator -2 Assistant Unit Supervisor SRO (or RO)
-1 Plant Control Operator RO -4 Auxiliary Systems Operator -1 b.
the-Tech Specs limits (given in Table 6.2.2-1) for Minimum Shift (0.5)
Complement.are met.
REFERENCE SSES Tech Specs, DI-AD-015 294001 Plant Wide Generics A1.03 AB to use procedures related to shift staffing 3.7 ANSWER 8.05 (1.50)
c. 2 (0.5 each)
b.
C.
REFERENCE AD-QA-103 294001 Plant Wide Generics K1.02 Kn of tagging procedure 4.5
.,...
. - _, _. _.
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. _. - _ - -. _ - _,. _.
-,.
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9:___0951N1@l80IlyE_P89CEQUBE@z_CgNQlIlgNS _8Ng_L1811011gN@
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z
- L ANSWERS -- SUSQUEHANNA-1&2-87/03/16-KOLONAUSKI, L.
. ANSWER'
8.06 (3.00)
e. Per TS Table 3.3.1-1 (RPS Instrumentation), each RPS trip system (1.0)
requires two (2) operable APRMs.
This requirement-is not met.
Take acti on] TS 3.3.1. a3 > [ place the inop channel or trip system in the tripped condition within one (1) hour.3 ir ippeh con Itzon.)
b.
'~'C
"C" 1...
U u.:
This surveillance will requir removing APRM
"C" from service, (1.0)
so now RPS
"A" will have 1 s
than the required minimum number of operable channels per t p system because APRM
"A" is bypassed.
TS 3.3.1 b applies.
Action 4 of TS Table 3.
.1-must be taken-> Ebe in STARTUP withi six (6) hours.3
__
i BUT-Because APRM
"D" is opera e,
note "a" of TS Table 3.1.1-1 (1.0)
allows you to place RPS Chan
"B" in the inop condition without without tripping RPS Channe
"B" for up to two (2) hours in order to perform the requ' e surveillances.
!
i MCMMON d.tfCrildrA gg N SuW4iMLL CAhhei N 215005 APRM (Group I System)
d Oht, kCoA44 $I if N Y SG#11 AB to recognize TS entry conditions 4.1 ggg g 4 %4 y hgym A.o.rcum wih accar.
Lt 3/31/97 (2.o; k
Wr "c" ir A +nyp A cuhhce pu Tr 3. 3. t. &
i Wi* AfRH 'A bpmcl, a scuolley awsor a cw m Afrm "c" k ( M n-M Itrwm uAM occut.
, k(AAA,4 jfV f "G" clou ook hbt h bt planJ M M b7y<d CMClih(/n ir-
Ihear, (t is for% S c4 m surettmu cn "cums (W inr as A )
& surmiluu.
l C Aho, AWtn "C" cw w Actad iny wM prhmi$
was not pr 6mCl, A h4 W - 6dA r r ne uJodd nd occw.
H A surnibtt Achcms l %d 4 w(g,(du E 3 5 1 b kn yFlb - % W.
5. 3.1 - i (IAP A 6 heuro wim m rnu r<rMee uhtn 1
I
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9___0951NigIB811VE_PBgCEpuSESt_CgNQlIlgN@t_@NQ_LidlIGIlgNg PAGE
- ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 8.07 (3.00)
c.
YES.
From Table 3.3.1-1 (RPS Instrumentation),
(0.5)
- 6-2 MSL Rad Mon are required per trip system.
I TS 3.3.1.at [With less than the minimum number of required operable channels per trip system (RPS B), place that trip system in the tripped condition within one hour.3 From Table 3.3.2-1 (Isolation Instrumentation),
(0.5)
[
TS 3.3.2.bt [With less than the minimum number of required operable channels per trip system, place that trip system in the tripped condition within one hour.]
b.(TR 3._3.1.b4 place RPS B in the tripped condition within one hour (0.5)
and take action given in TS Table 3.3.1-1.
- 6-[5d[ ion $3 [Be in Startup with the MSIVs closed in six hours, (6 33)
or at least in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.]
}TE 3.3.2.ch [ place at least one trip system in the tripped (0.5)
condition within one hour and take the-action given by TS Table 3.3.2-1.3 0.33
- 1e-l Action 20:)EDe in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and in cerS)
Cold Shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.]
0.33
- 3b-1 Action 21] EBe in at least Startup with the associated g>.95 )
isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least Hot Shutdoun within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.]
REFERENCE SSES TS 223002 PCIS (Group I Systems)
SG#11 AB to recognize TS entry conditions 4.1 212000 RPS (Group I System)
SG#11 AB to recognize TS entry conditions,
8___8QulNi@lB9IlVE_PggCEQUBE@t_CQNQ111QN@t_8NQ_LidlI@IlgNS'
PAGE-43
. ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER 8.08 (3.00)
. Declare the rod inoperable in accordance with]TS 3.1.3.1.a/
(0.33)
a.
Within one hour, verify:] action a.1 4)-- =hi 5 in NA, (0.33)
and disarm the associated DCVs.
Fol l ow acti on[ 3.1. 3.1. a. 2} -- f i x rod within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in (0.33)
Hot Shutdown within next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
UL (2.0)
b.
NO utr$), you may main in Op Con 2 as long as TS 3.1.3.1.a.1.b AND a.2 are met 03.
),
but you cannot go to the RUN mode (i e.,
3r34/j 7 change Op Con) because of TS 3.0.4 (QE).
g-g Or?.- $ 5 - Sfark h m CN h4A(,
( 2.c REFERENCE ( (WI' cadA M T ej to WW )
SSES TS M \\ CdM 8 T 3 ~5.l. 3.1. A. 1. f>
201001 CRDH (Group II Systems)
3, A SGW11 AB to recognize TS entry conditions 4.2 II 3 ' 3 *
2 yd$
ad nit.
ANSWER B.09 (2.50)
a. TS 3.8.3.1.b.1.a)
1) 1D612 lost DC Division I (1.0)
c) 1) 1D652 lost DE. Division I l TS 3.8.3.1 Action b} one DC dist system not energized reenergize within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or HOT SD within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and COLD SD within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.(TS3.8.3.1Actioneh480VACswingbusinop; (1.5)
declare LPCI loop inop-lTS3.5.1.b.j (TS 3.8.3.1 Action a-1 AC load group lost.
No fully applicable TS Action statement-->
TS 3. O. 3 )
REFERENCE SSES TS l
263000 Dnsite AC Power Distribution (Group II Systems)
SG#11 AB to recognize TS entry conditions 3.9
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- ANSWERS -- SUSQUEHANNA 1&2-87/03/16-KOLONAUSKI, L.
ANSWER B.10 (2.00)
NO (0.5). A reactor startup in three days would represent a violation of the SSES Tech Specs.
TSl3.6.1.4] requires that two MSIV LCS be operable in Op Con 1-2-3.
(0.75)
The action stateme t for TS 3.6.1.4 allows 30 days of continued operation, but Tech Spec l3.0./ does not allow entry into an Operational (0.75)
Condition while r61ying on an action statement.
REFERENCE SSES TS 223001 Primary Containment (Group I Systems)
SG #11 AB to recognize TS entry conditions 4.2 ANSWER B.11 (2.00)
e. NONE. [5 (N-18)
135 rem; not exceeded.]
(1.0)
=
[2000 mrem /qtr < 2500 mrem /qtr, < 3000 mrem /qtr3 b.
YES.
[5 (N-18) = 85 rem; not exceeded.3 (1.0)
[400 < 1250 mrem /qtr by 10CFR203 SSES Admin limit of LE 300 mrem /qtr if current quarter exposure not documented.
REFERENCE SSES AD-OO-735 External Dosimetry Program 10 CFR 20 294001 PW Generic K1.03 KN of 10 CFR 20 and related rad con requirements 3.8
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S.
NUCLEAR REGULATORY COMMISSION
.
REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
SUSQUEHANNA 1&2
,
REACTOR TYPE:
~BWR-GE4 DATE ADMINISTERED: 87/03/16 EXAMINER:
3.
K.' HAJEK CANDIDATE:
M4 STER INSTRUCTIONS TO CANDIDATE:
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 CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 1.
PRINCIPLES OF NUCLEAR-POWER PLANT OPERATION, THERMODYNAMICS,
'
HEAT' TRANSFER AND FLUID FLOW 25.00 25.00 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 3.
INSTRUMENTS AND CONTROLS 25.00 25.00 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00
%
Totals Final Grade All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
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'uring the administration of this examination the_following rules apply:
1.
Cheating on.the examination means an automatic denial of your application and could result in more severe penalties.
2.
Restroom. trips are to be Jimited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
I 3.
Use black ink or' dark pencil onIV to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
8.
Consecutively number each answer sheet, write "End of Category __" as-appropriate, start each category on a new page, write oniv on one side of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
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' 11. Separate answer sheets from pad and place finished answer sheets face
down on your-desk or table.
l 12. Use abbreviations only if they are commonly used in facility literature.
i 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
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14. Show all calculations, methods, or assumptions used to obtain an answer
'
to mathematical problems whether indicated in the question or not.
t l
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
l l
16. If parts of the examination are not clear as to intent, ask questions of the examirfr only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in I
completing the examination.
This must be done after the examination has been completed.
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18.'When you complete your examination, you shall:
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a.
Assemble your examination as follows:
(1)
Exam questions on top.
(2)
Exam aids - figures, tables, etc.
(3)
Answer pages including figures which are part of the answer.
b..
Turn in your copy of the examination and all pages used to answer the examination questions.
c.
' Turn in all scrap paper and the balance of the paper that you did not use for answering the questions, d.
Leave the examination area, as defined-by the examiner.
If after
- 1eaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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1.
PRINCIPLUS OF NUCLEAR POWER PLANT OPERATION, PAGE
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THERMODYNAMICS, HEAT TRANSF2R AND FLUID FLOW
.
.
QUESTION 1.01 (2.25)
With the reactor operating at full power, state which reactivity coefficient will cause the initial or greatest power change for each of the following malfunctions.
a.
A fault occurs in the EHC system that causes reactor pressure to oscillate.
(0.75)
b.
The recirculation system master flow controller fails such that recirc-flow demand decreases slowly.
(0.75)
c.
Extraction steam is blocked to a high pressure feedwater heater.
(0.75)
QUESTION 1.02 (2.00)
In the case of a loss of the shutdown cooling system, the Off Normal procedure, ON-149-001, Loss of RHR (Shutdown Cooling) System, recommends several operator actions that should be taken.
How will adhering to each of the following recommendations assure effective core cooling 7 a.
Restart at least one reactor recirculation pump OR operate Reactor Water Cleanup at its maximum rate. (1.0)
b.
Maintain vessel water level between 90 and 100 inches on shutdown indication instrumentation.
(1.0)
QUESTION 1.03 (2.25)
Explain why changes in the following three plant parameters or operating conditions cause reactivity effects that result in more severe pressure transients at the end of core life than at the beginning, thus requiring the EOC RPT protection function.
a.
Void coefficient of reactivity (0.75)
b.
Delayed neutron fraction (0.75)
c.
Control rod density (0.75)
.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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-1 PRINCIPLES OF NUCLEAR' POWER PLANT-OPERATION, PAGE
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.04 (2.00)
During an ATHS,. Emergency Operating Procedure EO-100-113 recommends lowering reactor level.
Briefly explain HOW and WHY this affects
,
J a.
Reactor power.-
( 1. 0 ) '
b..
SLC boron' concentration in the active fuel region. (1.0)
QUESTION. 1.05 (1.50)
'
a"p<.Fy)*'enthalpy of main steam at 1205 BTU /lbm, what tail pipe With the reactor operating at 100 percent power, and the
u S
' 's
,
p>dg temperature would you expect to see in order to verify that an SRV had cracked open and was continuously g #j**t'
leaking?. Assume a.ta11 pipe pressure of 75 psia.
ShowJs W41Me
.
all your work, and state all your assumptions.
(1.5)
I
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QUESTION 1.06 (2.00)
Concerning a potential rod drop accident,
{
a.
Why is a positive reactivity excursion due to a dropped rod more severe during a startup than when at power?
Give two reasons.
(1.0)
b.
Give two observations or methods that are required i
in the' process of withdrawing control rods to
'
minimize the chance and severity of a rod drop accident.
(1.0)
QUESTION 1.07 (2.00)
i Emergency operating Procedure, EO-100-112, Rapid l
Depressurization, provides for overriding ECCS pump initiations if more than one control rod is out further than notch 02.
What are the two reasons and the associated adverse consequences for this provision that
could result if ECCS injection was permitted?
(2.0)
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.08 (2.50)
For the changes in reactor operating conditions listed below, indicate whether NPSHA ( Available NPSH) for the reactor recirculation pumps INCREASES, DECREASES, or REMAINS THE SAME.
a.
Control rods are withdrawn from full in until reactor power reaches mid scale on Range 3 of the IRMs.
(0,5)
b.
Reactor power is maintained at mid scale on Range 6 of the IRMs, and the vessel heats up until the head vents need to be closed.
(0.5)
c.
Reactor power 10 maintained at mid scale on Range 6 of the IRMs, and reactor pressure increases from 100 psig to 920 psig.
(0.5)
d.
Reactor power is increased from 10 percent to 50 percent by withdrawing control rods.
Recirculation pumps are operated at minimum speed.
(0.5)
e.
Reactor power is increased from 70 percent to 100 percent by changing recirc pump speed.
Control rods are maintained at 100 percent rod pattern.
(0,5)
b#
QUESTION 1.09 (2.50)
O s.A+
,#' 8 IE*
A failure occurs in th Feedwater/ Level Control System
$M gy *""
which causes the level setpoint to instantaneously s, '
change from 37 inches to 31 inches while operating in
[,,,
three element control.
Explain the initial changes in s# eh'
,a j-j reactor power that result from this fault.
Include g
- ed considerations of changes in flow and the specific
.b g'
d
- reactivity coefficients involved.
(2.5)
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(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5,
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.10 (3.00)
State whether each of the following statements related to adequate core cooling is either TRUE or FALSE.
If a statement is FALSE, tell why it is false, a.
Adequate core cooling is assured if the core is covered with liquid.
(0.75)
b.
Adequate core cooling is assured if the core is covered with a two phase mixture, the larger percentage of which.is steam.
(0.75)
c.
If all level indication has been lost, operation of HPCI is sufficient to assure adequate core cooling.(0.75)
d.
If the reactor water level reaches the ADS initiation setpoint, and the RPV temperature is increasing, Blowdown Cooling must be initiated to assure adequate core cooling.
(0.75)
QUESTION 1.11 (3.00)
With regard to Xenon-135 in the core, a.
Explain why the xenon level initially increases after a power decreases, and initially decreases after a power increase.
(2.0)
b.
Why are the notch worths of control rods drastically altered by Xenon-135 for a startup ten hours after a reactor scram?
(1,0)
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(***** END OF CATEGORY 01 *****)
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
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e QUESTION 2.01 (3.00)
For each of the following ADS manual or automatic initiation conditions, indicate whether ADS will initiate, when, and why or.why not.
a.
Manual initiation attempted, one Core Spray pump running.
(0.6)
b.
ADS 102 see timer timed out.
No low pressure pumps running.
(0.6)
c.
ADS A/B Logic Control Switch in Inhibit position.
Then, all automatic initiation requirements become satisfied.
(0.6)
d.
ADS A/B Logic Control Switch in Inhibit position.
Manual initiation attempted.
(0.6)
e.
Level 1 and 3 signals received.
Hi D/W pressure signal not received.
Required low pressure pump (s)
running, and three minutes elapsed since receipt of low level signals.
(0.6)
QUESTION 2.02 (2.50)
When operating RHR in the Shutdown Cooling mode, a.
Why is it required that a flow rate of at least 4000 gpm be established within lo seconds after s arting the pump, and what adverse condition could occur if this was not accomplished?
(1.0)
b.
How is the rate of flow to the vessel controlled?
(0.5)
c.
Why must the flow rate through the heat exchanger be limited to less than 10,000 gpm?
(0.5)
d.
How is the cooldown rate controlled?
(0.5)
)
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PLANT DESIGN' INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
- c, QUESTION 2.03 (2.00)
For the Core Spray System, explain how a leak is detected in the piping between the inside of the RPV and the core shroud.
Include a description of how and why the monitored parameter will change if a break occurs in this piping.
(2.0)
QUESTION 2.04 (2.00)
J O' 714f What are the normal and alternate supplies for 4160 KV it SQ. 9 ESS Bus IA, AND in what order will power be supplied to 3'
( $'g, Bus 1A if the normal source is lost?
(2.0)
F.
QUESTION 2.05 (2.50)
State whether each of the following statements regarding the Reactor Core Isolation Cooling (RCIC) System is either TRUE or FALSE.
If the statement is FALSE, explain why.
a.
The turbine exhaust is directed to the barometric condenser where it is condensed and pumped to the Suppression Pool.
(0.5)
b.
Water from the Suppression Pool should only be injected into the vessel during automatic initiation or during an emergency condition.
(0.5)
c.
The RCIC System is capable of remote manual startup, operation, and shutdown from the Remote Shutdown Panel.
(0.5)
d.
If 110 pounds of pressure cannot be maintained in the RCIC discharge piping, it might be necessary to isolate the HPCI system discharge piping keeptill system.
(0.5)
e.
If the RPV water level reaches +54 inches and the F045 valve (Steam Admission valve) shuts, it cannot be reopened until the RPV water level reaches -30 inches.
(0.5)
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
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QUESTION 2.06 (3.00)
The Reactor Water Cleanup System (RWCU) is used for letdown during a reactor startup, a.
During'a startup, to what two locations can letdown be routed, AND under what conditions is each location used?
(1.0)
6.
How and why does the letdown rate limit the plant heatup rate?
(2.0)
QUESTION 2.07 (2.50)
For the Primary Containment Isolation System (PCIS),
O ta /S
_po a.
List five of the 1FystSES that will isolate from an 1_",7pyp N4S si al if a L A should occur when the reactor
"JO is ope ing at 40 percent power.
(1.5)
b.
RHR also receives isolation signals from N4S, but should not isolate under the conditions listed in l
Part a.
Why is this, AND under what conditions will the RHR isolation be active.
(1.0)
(*****
CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAG 3
2
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QUESTION 2.08 (1.00)
For each of the following conditions, match the correct statement for Recirculation System pump restart limitations, a.
Starting a recirc pump from AMBIENT conditions.
(0.5)
b.
Starting a recire pump from HOT conditions.
(0.5)
1.
A maximun of ONE start is allowed for the recirc pump.
After a 45 minute period has elapsed, another restart attempt can be made.
2.
A maximum of ONE start is allowed for the recire pump.
After a 25 minute period has elapsed, another restart attempt can be made.
3.
A maximum of TWO starta is allowed for the recirc pump.
After a 25 minute period has elapsed, another restart attempt can be made.
4.
A maximum of TWO starts is allowed for the recirc pump.
After a 45 minute period has elapsed, another restart attempt can be made.
QUESTION 2.09 (3.00)
What are the problems associated with operating HPCI under each of the following conditions 7 a.
Using the CST as suction if auto transfer to the suppression pool fails on low CST level.
(0.75)
b.
Suppression Pool level is above 23 ft. 9 in., and suction is being taken from the CST.
(0.75)
c.
Suppression Pool level is below 18.5 ft., and suction is being taken from the CST.
(0.75)
d.
HPCI turbine speed is below 2150 rpm.
(0.75)
b b*
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22. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS-PAGE 10 i
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LQUESTION 2.10
'(1.00)
au f.-w Manual start of AmHdt RHR pumps can be accomplished by depressing only one of the Manual Initiation pushbuttons.
Why then is it necessary to depress both pushbuttons when manually initiating RER?
(1.0)
' QUESTION.R2.11 (2.50)
For an EMERGENCY START of the Emergency Diesel Generators, which of the following will cause an engine trip?
(2.5)
a.
Low turbocharger oil pressure b.
Low' lube oil pressure c.
Main bearing high temperature d.
High jacket water temperature e.
Overspeed f.
Generator overvoltage g.
Generator overexcitation
,
h.
High Generator Differential Carrent 1.
Reverse power J.
Incomplete start sequence i
.
.
(***** END OF CATEGORY 02 *****)
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INSTRUMENTS AND CONTROLS PAGE
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o QUESTION 3.01 (2.00)
What methods, are available to insert control rods if they fail to insert on a scram signal, and if normal insertion actions fall?
Consider the actions recommended in E0P-100-102, RPV Control.
Give four.
(2.0)
QUESTION 3.02 (3.00)
For each of the following malfunctions or changes in plant parameters, state whether the Narrow Range Reactor Level instrumentation will indicate ACTUAL, HIGHER THAN ACTUAL, OR LOWER THAN ACTUAL LEVEL.
a.
A leak in the reference leg occurs that allows a continuous flow of water through the leg.
(0.5)
b.
Temperature in the Drywell during a LOCA increases and remains at 100 degrees F above normal.
(0.5)
c.
A steam leak occurs outside the Drywell, and one Main Steam line fails to isolate.
The reactor water level is maintained by HPCI.
(0.5)
d.
A leak occurs across the instrument equalizing valve.
(0.5)
e.
A rapid vessel depressurization occurs that causes an elevated Drywell temperature.
(0.5)
f.
An instrument technician has mistakenly calibrated the Narrow Range Level instrumentation instead of the Wide Range Level instrumentation while the reactor is operating at 75 percent power.
(0.5)
QUESTION 3.03 (1.00)
<
Calibration of the APRMs is checked after an LPRM is bypassed.
Since count and averaging circuits automatically adjust the output for the bypassed LPRM, why is it necessary to check the calibration after the LPRM is bypassed?
(1.0)
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o 3.
INSTRUMENTS AND CONTROLS PAGE
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O QUESTION 3.04 (3.00)
a.
Under what two conditions will the Rod Position and Information System detect a control rod drift?
(1.0)
b.
How would you determine which control rod is drifting?
(1.0)
c.
Give two conditions that could cause a rod to drift.
(1.0)
j, yl$'p { 5
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,.2 ewy QUESTION 3,05 (3.00)
pdt b
W. l
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p*
The plant is operating at 60 percent power, and the-bdak
.
feedwater flow signal to the recirculation flow control system fails to zero.
Consider each of the following questions independently.
Se sure to consider the
effects that will occur in BOTH UNITS.
a.
How will the speed of BOTH recire pumps be affected?
Why?
(1.0)
b.
What will happen to the speed of SOTH recirc pumps if the signal failure clears in about 30 seconds before any operator action is taken?
Why?
(1.0)
c.
How will the speed of BOTH recirc pumps be affected if the speed controller output on A MG set fails just prior to the loss of the feedwater signal?
Why?
(1.0)
QUESTION 3.06 (2.00)
Concerning the IRM system, a.
What four signals will cause a reactor scram?
(1.0)
o dm1*e, b.
Under what conditions is this scram function4 bypassed?
(1.0)
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(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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o 3.
INSTRUMENTS AND CONTROLS PAGE
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o QUESTION 3.07 (1.00)
Reactor Feedpump turbine speed is controlled with two controllers, the motor speed changer (MSC) and the electric automatic positioner (EAP) or Auto Speed Controller, a.
Which controller will be in control?
(0.5)
b.
Where must the MSC be positioned for the EAP to have control over the full range of turbine speed? (0.5)
QUESTION 3.08 (3.00)
List WHEN and HOW (automatically or manually) each of the following Reactor Cerams are bypassed, a.
MSIV Closure (0.75)
b.
Scram Discharge Volumn:Hi Hi (0.75)
c.
Turbine Control Valve Fast Closure (0.75)
d.
Mode Switch in shutdown (0.75)
QUESTION 3.09 (3.00)
Using the attached diagram of the EHC Pressure Control Unit (Figure 5), for reference, n.
Explain the purpose of the zero and three psi biases.
(1.0)
b.
Explain what will happen to control valve position 9ah g,g should a failure occur that causes the output of 1 Y
~~
g t h e f g oliiir n'i n g i e gli1~a t'6 r t o F A I L L O W.
(1.0)
s_
_.
-
g.D, p 3,J
c.
Explain what will happen to control valve position
., v should a failure occur that causes the output of-QA sdD-
yh t k
-
.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
3.
INSTRUMENTS AND CONTROLS PAGE
=, -
s QUESTION 3.10 (3.00)
With the plant operating at 75 percent power, consider each of the following channel / component operating situations for Rod Block Monitor Channel A.
a.
APRM A has been bypassed due to maintenance on the amplifier that has caused the unit to be taken out of Operate.
APRM C has just failed upscale and caused a half scram.
Relative to operation of the RSM, and to assure the safest completion of any necessary rod movement, would it be preferable to have APRM A or C bypassed, and to take the trip with the other APRM7 EXPLAIN.
(1.0)
b.
An LPRM input to the selection matrix has failed high.
The LPRM is also an input to APRM D.
How will bypassing or not bypassing of this LPRM in the APRM cabinet affect operation of the Rod Block Monitor?
EXPLAIN.
(2.0)
QUESTION 3.11 (1.00)
~
What are four AUTOMATIC ACTIONS (not alarms) that occur on receipt of a Refuel Floor High Exhaust Hi Hi Radiation signal?
(1.0)
.
.
(***** END OF CATEGORY 03 *****)
o 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
RADIOLOGICAL CONTROL
.
QUESTION 4.01 (2.00)
State whether each of the following statements is either TRUE or FALSE according to AD-QA-300, Conduct of Operations.
If the statement is FALSE, state why it is FALSE.
a.
Only licensed operators are permitted to manipulate controls that directly affect reactivity or power level of the reactor.
(0,5)
b.
In the Off-Normal Procedures, the Operator Actions are written in a logical sequence to provide guidance, and are to be performed as necessary in the sequence as directed in the procedure.
(0.5)
c.
ECCS actuation may be inhibited or overridden if specifically directed to do so in a Symptom-oriented Emergency Operating Procedure.
(0,5)
d.
When a nonradioactive oyotem to or becomeo contaminated, further use of the system may proceed no soon no shift Supervision provides for continuous radiation monitoring of the system.
(0,5)
QUESTION 4.02 (3.00)
For each of the following conditions, indicate whether or not EMERGENCY CPERATING PROCEDURE entry is required.
If entry is required, state which procedure (s) to enter.
If entry la not required, state "None."
Consider each sub part no a separate item.
Aonume no additional abnormal conditions are present for each individual item.
a.
RPV level la 10 inches.
(0.25)
b.
Reactor power la 12 percent, Startup mode.
(0.25)
c.
Reactor power 10 93 percent seven minutes after a load reject.
ggpv (0.25)
d.
Power operationo, Orcup I-loolation occuro.
(0.25)
e.
Supprecolon Pool level is 23.3 feet.
(0.25)
f.
Drywell preocure la 2.5 poig.
(0.25)
g.
CRD-HCU North Area Monitor indicating 130 MR/HR.
(0.25)
h.
Supprecolon Pool temperature is 95 degrees F.
(0.25)
1.
Reactor shutdown, RPV pressure is 1090 psig.
(0.25)
j.
Drywell temperature 10 160 degreco F.
(0.25)
k.
Zone III HVAC Exhaust Radiation Level is 4.5 MR/HR.(0.25)
1.
Suppression Pool level is 24.7 feet.
(0.25)
.
'
(***** CATEGORY C4 CONTINUED ON NEXT PAGE *****)
-
-
-
4.
PROCEDURES - NORMAL, ABNORMAL, EME9CENCY AND PAGI
.
,
,
RADIOLOGICAL CONTROL
QUESTION 4.03 (3.00)
While taking hourly instrument readings, you note that the condenser is slowly losing vacuum.
f a.
According to ON-143-002, Loss of Main Condenser Vacuum, which of the following possible symptoms j
should you check to either confirm that vacuum is decreasing, or to identify the cause of the vacuum loss?
(2.0)
'
1.
Off gas isolation valve closure
,
2.
Generator output decreasing 3.
Circ water system pump trip and alarm 4.
Reactor pressure decreasing 5.
Recombiner isolated 6.
Low RBCCH discharge pressure 7.
SJAE inlet stea?. pressure decreasing 8.
Cooling Tower basin level low 9.
Steam seal regulator malfunction 20.
Reactor water level high b.
Explain why reducing reactor po.ar can slow the decrease in the rate of loss of condenser vacuum.
(1.0)
i QUESTION 4.04 (3.00)
According to AD-QA-300, Conduct of Operations,
.
Attachment E, Immediate Operator Action List, what are the immediate operator actions to be taken if the Main Steam Line radiation levels are increasing, but the trip point has not yet been reached?
Include all alternative actions until the reactor is in a safe isolated condition.
(3.0)
t i
,
i
'
.
)
i i
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
,
'
.
.-.
..
-
- -.
-.
. - - - -. - - -
- -.
-... -
. -
- -. -. -.
_ _ _ _ _ - _ _
,
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
.
RADIOLOGICAL CONTROL
QUESTION 4.05 (2.50)
According to the Precautions in OP-100-002, Remote Shutdown - Normal Plant Operating Lineup, a.
What adverse operating and monitoring conditions may result if a transfer switch at the Remote Shutdown panel is placed in the EMERG position while the plant is operating?
(2.0)
b.
What may occur if a handswitch position at the Remote Shutdown Panel is changed while its associated transfer switch is in the NORM position?
(0.5)
QUESTION 4.06 (3.00)
According to the Precautions and Cautions in GO-100-002, Plant Startup and Heatup, a.
Where should the " Pressure Set" and " Bypass Jack" be positioned before condenser vacuum reaches 7" HgV (22.2" HgA)?
(1.0)
b.
What two adverse effects could occur if the MSIVs were opened with condenser vacuum established and the head vent open?
(1.0)
c.
How is SRM/IRM overlap established?
(0.5)
d.
HPCI and RCIC must be operable prior to exceeding what RPV pressure?
(0.5)
)
_. -
,
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
,
- ...,-
4.
PROCEDURES -' NORMAL. ABNORMAL. EMERGENCY AND PAGE
RADIOLOGICAL CONTROL
<
,
,
QUESTION 4.07 (3.00)
According to ON-118-001, Loss of Instrument Air, how do each of the following components respond (FAIL OPEN or FAIL CLOSED) to low Instrument Air header pressure?
a.
Outboard MSIVs (0.5)
i b.
Feedwater heater normal drain valves (0.5)
'
c.
Service Water outlet throttle valves from RSCCW and TBCCW heat exchangers (0.5)
d.
Emergency Service Water supply and return valves for RBCCW and TBCCW heat exchangers (0,5)
e.
Drain and vent valves for the Scram Discharge Volume (0.5)
f.
Condensor reject and makeup control valves (0.5)
QUESTION 4.08 (1.50)
According to AD-00-735, External Dosimetry Program, a.
What is the station administrative quarterly whole body exposure limit for a Radiation Worker with an up to date NRC Form 4 on file?
(0.5)
b.
If you are working in a radiation field of 200 mR/hr, and your Self Reading Dosimeter has a full scale range of 500 mR, when do you need to return to Health Physics to have it rezeroed?
(1.0)
QUESTION 4.09 (2.00)
'
According to ON-155-001, Stuck Control Rod, a.
What are two indications of a stuck rod that could be observed while attempting to move rods?
(1.0)
b.
Which of the following conditions could be causing the rod to stick?
(1.0)
1.
Drive water pressure at 220 paid 2.
Scram Discharge Volume Not Drained Alarm 3.
Drive water flow at 2 gpm while attempting to insert the rod one notch 4.
HCU valve alignments
>
-+
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
,
g
4.
PROCEDURES - NORMAL,' ABNORMAL, EMERGENCY AND PAGE
~
RADIOLOGICAL CONTROL
'
.
' QUESTION 4.20
.(2.00)
Describe how each of the following parameters will
'
change if a Safety Relief valve inadvertently opens and remains open, according to ON-183-001, Stuck Open Safety / Relief Valve.
a.
Indicated feedwa.ter flow relative to indicated steam flow-(0.5)
b.
Generator Load (0.5)
c.
Reactor power (0,5)
d.
Suppression' Pool level (0.5)
l l
l l
l
!
i I-
.
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
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1.
PRINCIPLES OF NUCLEAR POWER PLAp?T OPERATION, PAGE
,
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW e
ANSWERS -- SUSQUEEANNA 1&2-87/03/16-3. K.
HAJEK ANSWER 1.01 (2.25)
a.
void b.
Void c.
Moderator temperature REFERENCE SCO23 A-7, ppg. 12 - 13.
Specific Objectives 3, 13, 14, 15.
K&A 292008, K1.08 (4.1/4.1), KI.20 (3.3/3.4), K1.22 (3.5/3.6)
ANSWER 1.02 (2.00)
thermal stratification. [RWcu bgi4ul( Alco amer ++ h N a.
Prevents
% % m-e.Ly Hx-w h b.
Promotes natural circulation.
-h, gg 4 g
REFERENCE SCOO6 K (Mitigation of Core Damage - Natural Circulation), pg.
8.
Specific Unit Objective 3.
K&As 293008, K1.37 (3.2/3.4)
ANSWER 1.03 (2.25)
a.
The void coef. is greater at EOC than at BOL, and will add more reactivity should the voids collapse.
b.
Beta is smaller at EOL, and the reactor will respond faster to positive reactivity insertions.
c.
Control rods are further out of the core and must initially travel through a low flux region which will reduce the initial negative reactivity insertion rate.
REFERENCE SCO23 G-3 (Reactor Core Thermo:Th. Limits), ppg. 18 -
19.
Specific Objective 3.7.
K&A 293009 K1.24 (2.7/3.2), 202001 K1.28 (3.9/4.1),
K4.13 (3.7/4.0), K5.05 (3.5/3.6).
..
,
.-.3-
- - -
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
,-
,
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW.
- ANSWERS ---SUSQUEHANNA 1&2-87/03/16-B. K. HAJEK
' ANSWER 1.04 (2.00)
a.
Power is reduced [0.5) due to a decrease in natural circulation which causes more voiding in the core. [0,5)
b.
SLC mixing is retarded [0.5) due to less natural circulation in the core and increased voiding which reduces the boron concentration. [0.5)
REFERENCE SC006 K (Mitigation of Core Damage - Natural Circulation), pg.
9.
Specific Unit Objective 4.
K&As 293008, K1.34 (2.9/3.1), K1.37 (3.2/3.4), 292008, K1.20 (3.3/3.4).
ANSWER 1.05 (1.50)
This is a throttling process, which is adiabatic.
Therefore h1 = h2.
[0.5)
The Superheat Stean Tables [0.5) can be used to find a temperature of 350 F, [0.5]
or The Moller Diagram (0.5) can be used to find the same temperature. [0.5)
No interpolation is necessary for either method.
REFERENCE SCO23 D-8, ppg. 2 - 7.
CE Steam Tables Specific Objective 4.
K&As 218000, A1.01 (3.4/3.6), A3.02 (3.6/3.7), A4.06 (3.5/3.6), SG 15 (4.2,4.4), 239002, K4.06 (3.5/3.7),
K5.04 (3.3/3.5), etc.
.'
.
,-
-
-
,
--
-, - -
,,-,
,
j.;
2..
' PRINCIPLES OF NUCLEAR POWER' PLANT' OPERATION, PAGE
THERMODYNAMICS,-HEAT TRANSFER'AND FLUID FLOW I*l
~ANSWEFP~-
SUSQUEHANNA.2&2-87/03/16-B.
K. HAJEK
.
ANSWER'
.1.06 (2.00)
-a, 1.-
Rods'are worth more at startu'p 2.
There is no void feedback at startup
~
-b.
1.
Assure proper rod sequencing ~by use of. pull sheets, RWM, or.RSCS.
2.
Monitor core response during rod pull.
3.
Perform coupling check at position 48.
[0.5] for.any two or other resonable responses.
REFERENCE SCO23 A-6-(Rx Th-Rx Control), pg. 12.
Specific Objectives.24, 25, 26.
SC006 L (MCD).
Specific Unit objective 2, 3.
K&A.201003 K3.01 (3.2/3.4), K4.02 (3.8/3.9), A2.02 (3.7/3.8).
!
i
'
l ANSWER 1.07 (2.00)
.
1.-
To prevent a. cold water. injection [0.5) that could
'
-result in a pccitive reactivity insertion causing a power increase. [0.5]
\\
2.
To prevent boron dilu. tion (0.5] and preclude a
~
power increase or fluctuation.- [0.5]
!
REFERENCE EO-100-112, Bases, pg.3.
PP002 Specific Unit Objective 14, 16, 19.
,
i
'SC006 L (MCD).
Specific Unit Objective 2, 3.
!-
K&A 295037 EK1.03 (4.2/4.4), EK1.06 (4.0/4.2), EK2.04 (4.4/4.5).
l
'
i is
!
-
.
l l
.
-
. --
..
- _ _
.-.
.. -...~
_
...
- - - - -.,. -. -. -. _. -... -.. - _ - -
,c.
2-PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B.
K.
HAJEK
-ANSWER 1.08 (2.50)
a.
Remains the same, b.
Decreases [due to decreasing water density)
c.
Decreases [due to decreasing water density]
d.
Increases [due to increased FW subcooling)
e.
Decreases [due to flow losses exceeding the increase in FW subcooling)
REFERENCE SCO23 E-4, ppg. 35 - 43.
Specific Objectives 8, 9.
K&As 202001, K4.02 (3.1/3.2), K6.07 (3.3/3.3), A2.11 (3.7/3.9), A2.12 (3.6/3.8), 291004 K1.06 (3.3/3.3)
ANSWER 1.09 (2.50)
t*t When the level setpoint is lowered, feed flow will decrease. [0.5]
This will reduce subcooling, and cause warmer water to enter the core. [0.5]
Power will decrease [0.5) due to the moderator temperature coefficient [0.5], and due to increased voiding (void coef). [0.5)
REFERENCE SC007B (Plant Reactivity Evolutions), Specific Objective 2.
K&A 259002 A1.03 (3.8/3.8)
ANSWER 1.10 (3.00)
a.
True b.
True c.
False.
Low pressure ECCS is required to assure sufficient volume.
d.
False.
Wait until reaching TAF because as long as water covers the core, adequate cooling is available.
REFERENCE SC006 L (Review of Systems and Procedures used for Mitigating Core Damage), ppg.
2, 3,
19.
Specific Objective 1.
.
. _ _ _ _
1.,-PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE. 24
.,.
.TFERM0 DYNAMICS, HEAT TRANSFER'AND FLUID FLOW L
6,
' ANSWERS -- SUSQUEHANNA'I&2-87/03/16-B.
K. HAJEK K&A 295031 EK1.01 (4.6/4.7), EA2.04 (4.6/4.8)
ANSWER 1.11.
(3.00)
a.
Xenon is produced primarily from the decay of Iodine-135. [0.5)
It is removed primarily from burnup. [0.5].
On a power decrease, it is initially still.being produced at the higher power production rate, while the burnup rate has decreased. [0.5)
The opposite is true for a power
. increase. [0.5)-
b.
Xenon will have built up or peaked in regions of the core that previously had high flux, and will be.
.
much lower in regions that previously had low flux.
[ 0.' 5 ]
This will result in a shift in flux during the startup condition to the previously low flux regions, and a corresponding increase in the reactivity of the control rods in those locations.
(0.5]
REFERENCE SCO23 A-8 ~(Rx Th/Xe and Sm), ppg. 5-9.
Specific objectives 1, 2,
4, 5,
9.
K&As 292006 K1.02 (3.1/3.1), K1.03 (2.9/2.9), K1.04 (2.9/2.9), K1.06 (2.7/2.7), K1.07 (3.2/3.2), K1.08 (2.8/3.2).
'
.
_ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _. _ _... _
l 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 l
,
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B.
K. HAJEK e
ANSWER 2.01 (3.00)
a.
ADS will not start. [0.3)
Both Core Spray pumps must be running (or at least one LPCI pump]. [0.3)
b.
Nothing happens [0.3] until pump requirement is met. [0.3)
c.
ADS does not initiate [0.3) because Inhibit prevents satisfying the [drywell and level)
permissives. [0.3]
d.
ADS initiates provided the pump permissive is satisfied. [0.3]
Manual start overrides the other permissives. [0.3)
e.
ADS initiates [0.3] after seven minutes to allow initiation for a leak outside the D/W. [0.3)
REFERENCE SYO17 C-4 (ADS), pg. 7 and Table 4.
Specific Unit Objectives 1, 3,
6, 7.
K&As 218000 K4.02 (3.8/4.0), K4.03 (3.8/4.0), K5.01 (3.8/3.8), K6.01 (3.9/4.1), K6.02 (4.1/4.1).
ANSWER 2.02 (2.50)
a.
To keep the min flow valve (FQj7) from opening
[0.5) and draining the vessel to the suppression poo3. [0.5]
b.
By throttling the F017 valve [ located upstream of 10.5 )
the F015 injection valve).
c.
To preclude vibrations in the heat exchanger.
(0/5)
d.
By throttling the heat exchanger bypass valve (OS9 (F048), odl4k Mn, rs, et,HetSW. Q4 % f "3 gar
,
Valve numbers OR descriptions are acceptable.
Both are not required, buts deduction will be made if wrong numbers are given or incorrect descriptions are given.
REFERENCE YO17 C-1, Attachment D and Figures 3 and 6.
pecific Unit Objectives 3, 4,
5, 8,
9, 12.
' K&As 205000 K1.02 (3.6/3.6), K3.02 (3.2/3.3), K4.05
~ 08-i#9 001 dg 17
!
'
~
.
-, -
,
w
bLANT DESIGN INCLUDING SAFETY AND EMERGE!!CY SYSTEMS PAGE
,.
a ANSWERS -- SUSQUEEANNA 1&2-87/03/16-B.
K.
HAJEK
.
(3.6/3.7).
ANSWER 2.03 (2.00)
A differential pressure switch measures the dP between the bottom of the core plate (SLC above core plate tap)
[0.5] and the inside of the Core Spray sparger pipe just inside the RPV shroud. [0.5)
-If a break occurs, the dP
-
will increase-[0.5] because the pressure drop across the steam separator will be included. [0.5] - -. _.
[ Normal dP is -3.5 psid.
Alarm at. Ave psidJ2 p;f,,,m y g g g o,$-
'
.
,
"d,~( _
REFERENCE SYO17 C-2 (Core Spray), ppg. 15 - 16.
Specific Unit Objective 8.
K&As 209001 K4.04 (3.0/3.2)
ANSWER 2.04 (2.00)
Normal source is ESS101 Transformer OX201 from S/U Bus 10 OA103. [0.5)
First alternate (0.25] is ESS 201 Transformer OX203 from S/U Bus 20 OA104. [0.5)
Second alternate (0.25] is D/G A. [0.5)
REFERENCE SYO17 G-5 (AC Dist), pg.7 and Figures 4 and 10.
Specific Objectives 2, 6,
8.
K&As 262001 K1.01 (3.8/4.3), K1.03 (3.4/3.8), K4.06 (3.6/3.9), K6.02 (3.6/3.9).
i e
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
,
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B. K. HAJEK
.
ANSWER 2.05 (2.50)
a.
False. [0.2]
The turbine exhaust goes directly to the SP.
OR Gland seal exhaust, valve steam leakoff, and exhaust line drain pot water is
...
[0.3)
True. bbC. ok $ ( di ll pd t v h'~4ApWh N u rM Sc e. '
b.
CST elen d h [1
.
c.
True.
-
M * M. M M.3%f. klIOh d.
True, e.
False. [0.2]
It won't open automatically until level reaches -30 inches, but it can be opened manually any time after the high level condition clears. [0.3)
REFERENCE SYO17 C-5 (RCIC) ppg. 1 &
6, 24, 2,
8, 10.
Specific Unit Objectives 3, 5,
6, 7,
10.
K&As 217000 K1.03 3.6/3.6), K1.08 (3.3/3.4), KK4.01 (2.8/2.8), K4.02 (3.3/3.3), K4.06 (3.5/3.5), K4.07 (3.6/3.6).
ANSWER 2.06 (3.00)
a.
To the condenser [0.25] if vacuum is established Nedds ad dh
[0,25], and to Radwaste [0.25] if vacuum has not
- mNO" been established. [0.25]
hM M b E' % ^ M m m.sp Q h. y d '?thssf n o.T p s.
"YG'E *
d k
b.
Since some water is not being returned to the O t^ -
vessel, it is not passing through the regenerative
' #~
I heat exchanger. [0.5]
This limits the capacity of the regen Hx and the Non Regen HX must maintain system below 130 degrees F on its own. [0.5]
Otherwise, the RWCU will isolate. [0.5]
A slower heatup rate will produce less excess water, allowing more to be returned to the vessel. [0.5]
"
~#
^
?O S 0 7 L-1 (RWCU), pg. 2g
'
Specific Unit Objectives 2,~3, 4.
K&As 204000 K1.01 (3.1/3.3), K1.06 (2.8/2.8), K1.07 (2.6/2.7), K1.09 (3.2/3.3), K3.02 (3.1/3.1), K4.03 (2.9/2.9), K4.04 (3.5/3.6), K4.07 (2.9/2.9), A2.12 j
(2.7/2.8), SG7 (3.4/3.3), SG10 (3.2/3.2).
.
.
2.
PLANT DESIGN INCLUDING' SAFETY AND EMERGENCY SYSTEMS PAGE
,
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B.
K.
EAJEK
.
LANSWER 2.07 (2.50)
a.
MSIVs A)So c~uetb RWCU System VAlvti IU4#
TIP System in OWDM Containment Atmosphere Control System 7%g g.
Drywell Floor Drain / Equipment Drain System Reactor Building Chilled. Water to Rx Recirc pumps HVAC Isolation /[ Actuation) Mat.
5+ypALssim foot % ~1 sad *m
.ar6 each for any five
b.
RHR will not isolate because it is needed for the LPCI injection mode. [0.5)
If it is operating in Shutdown Cooling (different operating mode than in Part a), it will isolate. [0.5)
MA*O REFERENCE SYO17 E-3 (PCIS),. ppg. 3-10. T4 57
_ oM-1 W O*** pol 1,id5 -
Specific objective 5/
- - -
K&As 223002 K1.01 (3.8/3.9), K1.02 (3.3/3.5), KI.03 (3.0/3.2), K1.08 (3.4/3.5), K4.04 (3.2/3.6), SG4 (3.6/3.7)
tl W5)
z.0 8 a
b 2.
(o.5)
N GYol7 - L-8 (Rsd 6pefc Ob'fcSA T jdfAo
.
,
....
-.
.c.,
,, -
e 2-PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS-PAGE
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B.
K. HAJEK
.
ANSWER 2.09 (3.00)
a.
Continued operation of HPCI will quickly empty the (o,pgj CST.
(Low CST level alarm at 45" and swap to SP at 3'
7.5"]
b.
The combination of HPCI steam discharge and high (gy gg pool level could result in exceeding the design loads of the Suppression Pool.-
c.
The HPCI steam exhaust spargers become uncovered.
( 0. 76)
d.
Below 2150 rpm, oil pressure will be too low to maintain the turbine stop valve in its open g,35)
position, [and to prevent governor valves from operating too close to their seats.]
REFERENCE SYO17-C6 (HPCI), ppg. 17 ff.
Specific Unit Objectives 5,
6, 8,
9, 10.
SC006 L.
Specific Objectives 2, 3.
K&A 206000 K4.09 (3.8/3.9), K4.10 (3.7/3.8), K4.19 (3.7/3.8), K5.05 (3.3/3.3), K6.04 (3.5/3.7), K6.05 (3.5/3.7).
^
ANSWER 2.10 (1.00)
(Q BecauseAonly one ESW Division will start on depressing only one pushbutton, [ leading to RHR pump motor damage ] [o.73 (~0 6 iil n d b d. A LICII^'cc.Wm % b% biAdm mios 6N,*4 g g, 4.,-
G a
&n REFERENCE G -* * t N
N
"'
D
~ cog. 32 - 33.
SYO17 C-1 (RHR),
Specific Unit Obje'ctives 5, 8,
14.
W Fol7 A/S AJd bM K&As 203000 K1.16 (3.1/3.2), K4.04 (2.6/2.7), A4.05 g&) 'a g Q
(4.3/4.1), SG7 (4.2/4.3)
'
l de ley..c
- gS g
,
M pc.M d D 8 ~'
'$
-
!
i
-
l
.
b, e, h l
l L-
,
-
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
,
,
' ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B. K. HAJEK
>
ANSWER 2.11 (2.50)
b, e, and h (will cause trips during both emergency and non emergency starts.
All others will cause trips only during non emergency starts.]
-
'
0.25 for each response - not listing a response implies it is not an emergency trip.
REFERENCE SYO17 G-1 (DGs), ppg. 29 - 30.
Specific Unit Objactive 6.
K&As 264000 K4.01 (3.5/3.7), K4.02 (4.0/4.2)
,-
'
.
I o
. *,
- -,, - - -
. -,, -
. -,
-
,
.,, - -, - -
-
-.,
-,,. -,,
,
c 2-s--..-
-- ---,--
INSTRUMENTS AND CONTROLS PAGE
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B.
K.
HAJEK s
ANSWER 3.01 (2.00)
TrytodrivearodwithincreasedCRDflow.(esivgdadd4Iad5k 1.
L' k ' # D dJ 2.
Remove air from the top of the scram valves to t
allow them to fail to'the scram position.
(This is done by closing the Inst Air filter valves and opening the Scram Air Header Drain valve.]i.e.,IulskertA t e M 3.
De-energize the scram solenoid valves.
.
[This is done by opening breakers CB2A in RPS Trip] 14 drE4 da 9'~'
Sys A and CB8B in RPS B.]
J n+d*++ H5 8 4 x Lb22-4.
Alternately scram rods and empty the SDV.
N 5.
Pull fuses to de-energ4% the scram pilot solenoids.
This is recommended in SSES-EPG, but considered to 'Gd, k n am Yodo be less reliable than Item 3 above due to hot Jif%H5/V4%
shorts.
CAF to determine if credit should be g
given.
Scram rods individually. h SgI a Sc % fend sw'd[
Cl 'D 6.
"
2 14C.h REFERENCE SYO17 K-2 Specific Objectives 5, 8,
9.
SC006 L.
Specific Objectives 2, 3.
PP002 Specific Unit Objectives 14, 16.
EO-100-102 Flow Chart and Bases K&A 295037 EK3.07 (4.2/4.3)
-
!
.
,
,
-3..
INSTRUMENTS AND CONTROLS PAGE
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B. K. HAJEK
$
ANSWER.
3.02 (3.00)
a.
HIGHER THAN ACTUAL [because reference leg temperature will not cool to ambient, density will be less
.]
..
b.
HIGHER THAN ACTUAL (due to lower density
]
...
c.
HIGHER THAN ACTUAL [due to colder than normal water in the vessel and a higher variable leg pressure)
d.
HIGHER THAN ACTUAL (due to d/p decreasing across the valve to zero)
e.
HIGHER THAN ACTUAL [due to reference leg flashing]
L w'.Ll M4,Lph oscill fcAto sq $w f.
ACTUAL (since both are calibrated at normal
temperature and pressure]
o REFERENCE SYO17 J-2 (RPV Inst), ppg.
8-10.
Specific Objective 9.
--
SC006 G (MCD-RPV Level Inst), ppg. 8 - 14.
Specific Unit Objective 4.
K&As 259002 K2.09 (2.9/3.0),
295012 AK1.02 (3.1/3.2),
216000 K4.05 (3.9/4.1), KS.01 (3.1/3.2), K5.06 (3.4/3.6), K5.07 (3.6/3.8), K5.10 (3.1/3.3)
ANSWER 3.03 (1.00)
Bypassing a low reading LPRM could cause the APRM to read high.
Likewise, bypassing a high reading LPRM
could cause the APRM to read Iow.
R E F E R E N C E T 'l N SYO17, ppg. 25.
Specific Unit Objective 6.
K&As 215005 K3.05 (3.8/3.8), K6.03 (3.1/3.3)
.
+
- A
INSTRUMENTS AND' CONTROLS
'PAGE.33
.,-
~ ANSWERS -- SUSQUEHANNA'I&2-87/03/16-B. K. HAJEK
.w ANSWER 3.04 (3.00)
a.
1.
When any rod other than the selected rod is not
-
at an even numbered notch 3 n. On %,JJfuelr.d4d p*cq up.
C195,$
2.
When the selected rod is not at an even notch
[n, A+g and no rod motion is requested, or if it doesn't get to the next notch before the timer
'E times out.
3. 10b Orvdinat/R4 In fMWb 55 kp*M..
b.
Depress the DISPLAY ROD DRIFT pushbutton [0,5] and check for the red light on the Full Core Display.
[0.5)
c.
1.
Excessive CRD seal leakage 2.
Excessive cooling water pressure 3.
Air in the system 4.
Scram valve leakage 5.
Directional control valve malfunction 6.
Collet mechanism failure to latch 7.
Depressing Cont Insert Rod pushbutton 0.5 each for any two REFERENCE SYO17 K-7 (RMCS), ppg.
3, 6,
Figure 2.
Specific Objectives 3, 4.
ON-155-006 Rod Drift AR-104-001 pg. 50 Rod Drift K&A 201002 K4.03 (3.6/3.6)
.
'
.
m
- _ -. - _ - - -. -, ~. -, -. _
_ _ - -
_
_ - -.
- _.
3.
INSTRUMENTS AND CONTROLS PAGE
-,
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B. K. HAJEK e
hl0 nrm wCMus Atc.vt Jyr ANSWER 3.05 (3.00)
M > 6 0 'lo
Both pumps will run back to 456 percent (0.33]
a.
because the feedwater flow signal to the recirc system is less than 20 percent (or - the flow limiter-is not bypassed at less than 20 percent feedwater flow). [0.67]
N($4bbypuNd'
b.
On Unit #
1, the condition will clear automatically, and the pumps will return to their l
> gd).
previous speed. [0.5]
(
' nafe c m.hss-On Unit # 2, nothing will happen because the limiter must be reset by operator action after the 4Y' f3
,
condition clears. [0.5]
)
c.
Pump B will run back as in Part a. [0.25]
Pump A speed will remain the same (0.25] because of a scoop tube lockup caused by the loss of speed signal. [0.5]
REFERENCE SYO17 L-9 (Rx Recirc Control), ppg. 3 - 8.
,
Specific Unit Objectives 3, 4,
6, 10.
K&As 202002 K1.08 (3.1/3.2), K3.05 (3.2/3.3), K4.02 (3.0/3.0), K6.04 (3.5/3.5)
ANSWER 3.06 (2.00)
a.
1.
120/125 of scale [0.25]
2.
HV low [0.25]
3.
Module unplugged [0.25]
4.
Function switch not in operate [0.25]
b.
Mode switch in RUN [0.4]
AND [0.2]
Companion APRM on scale. [0.4]
REFERENC SYO17-I2, ppg. 17 - 19.
Specific Objective 3.
K&A KI.01 (3.9/3.9), K4.02 (4.0/4.0), K6.06 (3.2/3.4)
i
-
.
-
3.
INSTZUMENTS AND CONTROLS PAGE
,
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B. K.
HAJEK e
.
ANSWER 3,07 (1.00)
a.
The controller with the lower of the two electrical output signals.
b.
At the high speed stop.
REFERENCE SYO17 D-3 (Rx FN-Sys), ppg. 11 - 13.
System-Unit Objective 9.
K&As 259001 K5.03 (2.8/2.8), A1.04 (2.8/2.7)
ANSWER 3.08 (3.00)
a.
Automatically (0.25] bypassed when the mode switch is not in RUN. [0.5)
M b.
Manually bypassed [0.25] with tese keylocked switched when mode switch is in Shutdown or Refuel. [0.5)
Tku-
-wq c.
Automatically bypassed [0.25] when turbine first stage pressure is less than X percent of its rated valuel. [0.5)
- hos Ibs n 2.Y pned pvv44;q d.
Automatically bypassed [0.25] after a 10 seconds delay. [0.5)
Dofe: PoirdSnoY M "p REFERENCE
.
hMcT,4FCd n mA*W h
SYO17 L-5 (RPS), ppg. 31 - 36.
Specific Objective 6.
y K&A 212000 K4.12 (3.9/4.1)
ggf,*m
,
, p t40 f
-
.
_
3.
INSTRUMENTS AND CONTROLS
.PAGE
. ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B. K.
HAJEK e
ANSWER 3.09 (3.00)
a.
Since the regulators are identical and redundant, the biases assure that only one regulator will normally be controlling.
$t ed, '
b.
When the A output falls at least 3 psi, the 3,
backup, regulator will take control and limit the closure of the control valves.
g h ' 'k
c.
There is no backup for a failure that causes an increase in output signal, and-the control valves will open.
REFERENCE SYO17 A-8 (EHC Pressure Control & Logic), ppg.
7-8, and Figure 5.
Specific Objective 1, 2,
3.
K&As 241000 K1.01 (3.8/3.9), K1.08 (3.6/3.7), K3.01 (4.1/4.1), K3.08 (3.7/3.7), K4.01 (3.8/3.8)
ANSWER 3.10 (3.00)
a.
It would be better to bypass APRM C [0.5] because with C bypassed, APRM E would automatically become the reference APRM. [0.5]
b.
'If the LPRM is not bypassed, the high reading will be averaged with the other LPRM signals. [0.75]
However, if the LPRM is bypassed, its output would drop to zero, and would trip the dow scale trip unit. [0.75]
This would remove it from the count circuit, and result in a RBM input based on the readings of the properly operating LPRMs. [0,5)
REFERENCE SYO17-K5, ppg. 5 - 7.
Specific Objectives 4, 6.
KSAs 215002 K1.01 (2.9/3.0), K1.02 (3.2/3.1), K4.01 (3.4/3.5), A2.03 (3.1/3.3).
.
e
-
-
3r
~ INSTRUMENTS AND CONTROLS PAGE
.
.. ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B.
K. HAJEK ANSWER.
3.11 (1.00)
1.
RB Zone III Recirc Fan A/B starts.
2.
SEGT Train A/B starts.
3.
Control Room Emergency Outnide Air Supply Fan A/B starts.
4.
Isolation of RB Zone III dampers.
5.
RB Zone III Recirc Dampers open.
0.25 each for any four.
REFERENCE Sy017 B2 (Process. Rad Mon Sys), pg.
8.
Specific Objective 4.
K&As 288000 K1.03 (3.7/3.7), K1.02 (3.4/3.4),.K1.05
~(3.3/3.6), K4.01 (3.7/3.9), K4.02 (3.7/3.8)
l
,
l
.
t
.
I e
g
'4.
PROCEDURES ~-' NORMAL, ABNORMAL, EMERGENCY AWD PAGE'.38
.
RADIOLOGICAL CONTROL
'
ANSWERS -- SUSQUEHANNA 2&2-87/03/16-B.
K. HAJEK ANSWER 4.01 (2.00)
a.
False. [ 0.' 2 5 ] - An exception is when an individual is receiving license training. [0.25]
b.
True.
'c.
True.
d.
False.
[0.25]
Operation shall be restricted until the cause of the contamination is identified and corrected and the system has been decontaminated or'if the system'Is required, a safety evaluation
.of the system as a radioactive system must be performed.-[0.25]
REFERENCE AD-QA-300, Conduct of Operations, ppg. 28, 23, 30, 33.
PWG A1.02'(4.2/4.2),' EOP SGs 12 (Varies 3.5 and above),
Abnormal Procedures SGs 12 (Varles 3.4 and above)
ANSWER 4.02.
(3.00)
a.
h 101 59w b.
None
- g g g g-l)
'
c.
102 - R P V M 'r$
d.
102 - 55 ES d,*44 e cry,3 dagWs e.
None f.
102, 103 (0.125 each]
g.
10 4 - 5 ei %d% ced.CMv4 h.
103 - edm M. cud re-t 1.
102 j.
103 k.
104 1.
103 REFERENCE PPOO2 (EOPs)
Specific Unit objective 1,
17.
EOPs 570, 571, 572, 580.
KAAs 295024-38, System Generic K/As II (4.3/4.5).
.
=
.
i 4..
PROCEDURES - NORMAL, AB4ORMAL, EMERGENCY AND'
PAGE
i
,
RADIOLOGICAL CONTROL o
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B. K. HAJEK
,
l i
ANSWER 4.03 (3.00)
a.
YES:
1, 2,
3, 5,
7, 8,
9.
i NO:
4, 6,
10.
j 0.2 for each correct answer.
If candidate only lists the YES answers,'NO will be assumed for the others, b.
Reducing power assists in maintaining vacuum by reducing the input of the non-condensible gasses (0.5] and lowering the condenser heat load. [0.5)
REFERENCE ON-143-001, Loss of Main Condenser Vacuum K&As 295002-AA2.02 (3.2), G5 (3.2).
ANSWER 4.04 (3.00)
1.
Reduce Recirc pump speed to minimum
[0.5]
,
2.
Insert CRAM Array rods (0.4]
[0.2)
Insert in-sequence rods.
[0.4)
3.
If power decrease did not stop the radiation increase,
[0.5)
Manually scram the reactor
[0.4)
AND
[0.2)
Isolate the MSIVs.
[0.4]
REFERENCE AD-QA-300, Conduct of Operations, Attachment E.
K&As 272000 A2.01 (3.7/4.1), SGI (3.6/3.9), SG8 (3.5/3.5), SG14 (3.4/3.4), SG15 (3.7/4.2)'.
l l
[
!
l
,
o
- - _ -
,
-
.4..
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
o
.
RADIOLOGICAL CON ~ROL
ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B.
K.
HAJEK ANSWER 4.05 (2.50)
a.
Changing the transfer switch to EMERG may 1) change the position of associated valves, (0.5)
or 2) change the operation of associated equipment, (0.5]
or 3) cause a loss of Control Room indication for the components controlled from the Remote Shutdown panel. [0.5]
4) Also, it may result in operation of equipment that cannot be monitored at the Remote Shutdown panel. [0.5)
b.
Nothing.
RdFERENCE OP-100-001, Remote Shutdown - Normal Plant Operating Lineup K&As 295016 AK2.01 (4.4/4.5), AK2.02 (4.0/4.1), AK 3.03 (3.5/3.7).
ANSWER 4.06 (3.00)
a.
Pressure set should be above reactor pressure (0.5]
Bypass Jack at zero [0.5)
b.
A vacuum could be pulled on the Drywell. [0.5]
Long path recire could leak into the vessel. [0.5)
b g'T~. S,
C c.
With 0.5 decade of overlapp 1""i" 202; LG;.-:J.::~)
(0.f)
d.
150 psig.
REFERENCE GO-100-002, Plant Startup and Heatup, ppg.
7, 8,
17, 20.
K&As 241000 K1.06 (3.8/3.9), 239001 K1.17 (3.1/3.2),
215004 K4.06 (3.2f3.2), A2.05 (3.3/3.5), 206000 SG1 j
(4.3/4.4), SG10 td.7/4.4).
Mok N Memel M
%C,b ' O 3 3.es A h g w c_. L
, % $ady 4 eadAQscahML4Q 4% A 4 7.Jt 3m 2. J.Re. b kaf sum m suJe eM 4 6
--
tod<Asex4.7 % + h m 444.f4
$ MM'"
aw k 4 m % - <se.k sera n.u w o a nw-p d A 14*b y h ao N "
k to m 2 q q.
}#
W
-
..
..
.
-
_ - _ _ - _ - _ - _ _ -.. _ _ - _ - - _ _ _ _ _ _ _ _
__ __ - _-
lj 4.-PROCEDURES - NORMAL,' ABNORMAL, EMERGENCY AND PAGE.41 L
RADIOLOGICAL CONTROL
-o ANSWERS -- SUSQUEHANNA 1&2-87/03/16-B.
K. HAJEK
- -
L
/
ANSWER 4.07
'(3.00)
a.
Fall closed b.
Fall closed c.
Fall open d.
Fall closed e.
Fall closed f.
Fall closed REFERENCE ON-118-001, Loss of Instrument Air, ppg.
5-7, Paragraphs D, B,
L, M,
C.
K&As 295019-AK2.01 (3.8/3.9), AK2.02 (2.9/3.0), AK2.03 (3.2/3.3),.AK2.05 (3.4/3.4), AK2.07 (3.2/3.2)
=
ANSWER 4.08 (1.50)
[n.te. Espom syIdelids ud erkJ-)
a.
2500 mrem b.
When it reaches 75 percent of scale (1.0)
OR when it reaches 375 mR [1.0)
OR in about I hr 52.5 min after starting the job. (1.0)
REFERENCE AD-00-735, External Dosimetry Program, ppg.
8, 12.
K&A 294001 K1.03 (3.3/3.8)
ANSWER 4.09 (2.00)
b'
a.
1.
Rod position indication does not change (0.5]
2.
No indicated change in NI response (0.5){p 6 4;) q w 3 M for each correct evaluation. p) is poe.]
C b.
1, 3, and 4.
(0.25 (l.0 )
REFERENCE ON-155-001, Stuck Control Rod, pg.
2.
K&As-201003 A2.01 (3.4/3.6),
-
.
_ _ _.. _....
. _ _ _ _ _ _ _. _. _
4 PROCEDURFS - NORMAL, ABNORMAL, EMERGENCY AND PAGE
2
,
RADIOLOGICAL CONTROL O
ANSWERS --'SUSQUEHANNA 1&2-87/03/16-B.
K. HAJEK ANSWER 4.10 (2.00)
,
Feed flow greater than steam flow a.
b.
Decreases Reactor power remains the same(Lg +> @d is ** bcM.Ae. '% prwC4.
c.
Oscillates k i: 6 obtv M d not M gJ,gly,",
' #
'
F d.
}
g $g4g gg]
"
REFERENCE ON-183-001I, Stuck Open Safety / Relief Valve, pg.
2.
K&As 2390d2 K4.06 (3.5/3.7), A1.06 (3.7/3.8), A2.03 (4.1/4.2).
i Y t-
& rytAk 6%Yts 'Mj 4 SD & Y 44.44 u c. AM rise.
i
.
,-,- --- - - -..
- -,, -.. - - - -,. _ _
,
TEST CROSS REFERENCE PAGE
9
.
,
QUESTION VALUE REFERENCE QUESTION VALUE REFERENCE o
________
______
__________
________
______
__________
01.01 2.25 BRH0000320 03.01 2.00 BRH0000342 01.02 2.00 BRH0000321 03.02 3.00 BRH0000343 01.03 2.25 BRH0000322 03.03 1.00 BRH0000344 01.04 2.00 BRH0000323 03.04 3.00 BRH0000345 01.05 1.50 BRH0000324 03.05 3.00 BRH0000346 01.06 2.00 BRH0000325 03.06 2.00 BRH0000347 01.07 2.00 BRH0000326 03.07 1.00 BRH0000348 01.08 2.50 BRH0000327 03.08 3.00 BRH0000349 01.09 2.50 BRH0000328 03.09 3.00 BRH0000350 01.10 3.00 BRH0000329 03.10 3.00 BRH0000351 01.11 3.00 BRH0000330 03.11 1.00 BRH0000352
______
______
25.00 25.00 02.01 3.00 BRH0000331/
04.01 2.00 BRH0000353 02.02 2.50 BRH0000332 04.02 3.00 BRH0000354 02.03 2.00 BRH0000333 04.03 3.00 BRH0000355 02.04 2.00 BRH0000334 04.04 3.00 BRH0000356 02.05 2.50 BRH0000335 04.05 2.50 BRH0000357 02.06 3.00 BRH0000336 04.06 3.00 BRH0000358 02.07 2.50 BRH0000337-04.07 3.00 BRH0000359 02.08 1.00 BRH0000338 04.08 1.50 BRH0000360 02.09 3.00 BRH0000339 04.09 2.00 BRH0000361 02.10 1.00 BRH0000340 04.10 2.00 BRH0000362 02.11 2.50 BRH0000341
25.00
25.00
______
100.00 l
.
(
^
s