IR 05000317/1987019
| ML20237E817 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/16/1987 |
| From: | Keller R, Norris B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20237E800 | List: |
| References | |
| 50-317-87-19OL, 50-318-87-21OL, NUDOCS 8712290204 | |
| Download: ML20237E817 (124) | |
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i U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO.
50-317/87-19(0L) and 50-318/87-21(0L)
FACILITY DOCKET NOS.
50-317 and 50-318 FACILITY LICENSE NOS.
Baltimore Gas and Electric Co.
P. O. Box 1475 Baltimore, Maryland 21203 FACILITY:
Calvert Cliffs Nuclear Power Plant, Units 1 and 2 EXAMINATION DATES:
October 27-29, 1987 CHIEF EXAMINER
12//g[67 arry S. Norris Date Senior Operations Engi er (Examiner / Inspector)
APPROVED BY:
/4 12//4/67
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Robert M. Keller, Chief Date PWR Section, Division of Reactor Safety SUMMARY: Written and operating examinations were administered to five Senior Reactor Operator (SRO) candidates and three Reactor Operator (RO) candidates.
Four SR0's and all R0's passed these examinations. One SRO candidate failed the written examination.
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I B712290204 871216 PDR ADOCK 05000317 V
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l REPORT DETAILS TYPE OF EXAMINATIONS:
Replacement EXAMINATION RESULTS:
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CHIEF EXAMINER AT SITE:
B. S. Norris (USNRC)
OTHER EXAMINERS:
L. S. Defferding (PNL)
P. T. Isaksen (EG&G)
F. S. Jaggar (EG&G)
R. M. Keller (USNRC)
1.
The following generic deficiency was noted on the operating examinations.
This information is being provided to aid the licensee in upgrading license and requalification training programs.
No licensee response is required.
Communication between the candidates during the simulator portion of the examinations was at two extremes. The operators were either very conscientious with correct orders and repeat backs, or very poor with vague orders and generic responses.
Examples of the poor communications are:
" bring it [the turbine generator] up fast", "How you doing?" or
"Okay" [in response to " bring it up fast"].
2.
The following is a summary of generic deficiencies noted from the grading of the written examinations. This information is being provided to aid the licensee in upgrading license and requalification training programs.
No licensee response is required.
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Reactor Operator Examination:
(by question number)
1.12 None of the R0 candidates could explain how the additional auxiliary feedwater would affect the accuracy of a secondary calorimetric.
2.05 All of the R0 candidates were weak with respect to the fail position of various valves after a loss of instrument air header pressure.
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Senior Reactor Operator Examination:
(by question number)
5.08 Three of the five SR0 candidates could not explain how thermal and fast neutron flux would change during core life.
5.09 Four of the SRO candidates were unable to determine the system flow and pressure for two centrifugal pumps operating in series or parallel.
6.01 All of the SRO candidates were unable to explain how and why a failure of a steam generator level detector would affect actual steam generator level.
6.05 All-of the SRO candidates were weak with respect to the fail position of various valves after a loss of instrument air header pressure.
6.10 None of the SRO candidates could correctly draw a diagram of the electrical power supplies to a vital 120vac bus.
7.10 None of the SRO candidates could describe how to retrieve a fuel transfer carriage if the cable were to become overloaded.
8.05 All of the SR0 candidates had difficulty in determining the correct Limiting Condition of Operation given a set of conditions and a copy of the Technical Specifications.
3.
Personnel Present a+, Exit Interview:
NRC Personnel
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i B. S. Norris, Chief Examiner i
P. H. Bissett, Senior Operations Engineer i
D. Wallace, Operations Engineer i
Facility Personnel l
i J. E. Gilbert, Supervisor Procedure Development l
J. R. Hill, Supervisor Operations Training K. Nietmann, General Supervisor Nuclear Training J. M. Yoe, Senior Operations Instructor I
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4.
Summary of comments made at the Exit Interview:
a.
The NRC summarized the examinations administered during the week and discussed the one generic weakness identified during the operating examinations.
See paragraph 1 above, b.
The NRC expressed concerns with respect to how the Emergency Operating Procedures (EOPs) were used durin'g the simulator
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examinations.
Th6 concerns are:
(1) The Control Room Supervisor (CRS) does not appear to adequately control the plant during the implementation of the E0Ps.
The CRS does not direct the completion of individual steps but instead directs that sections be completed.
Frequently, two or three sections at a time would be. assigned to one of the board operators.
During one of the scenarios, six sections were simultaneously assigned to the secondary operator, and there was no direction given by the CRS as to the priority of the sections. When asked by their examiner, the SRO candidates (while in the position of the CRS) were unable to show exactly where they were in the E0Ps.
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(2) When the board operators would report back that a section was completed, the CRS denoted completion of the section by putting an "X" in the procedure manual. This method of tracking could
, lead to confusion the next time that procedure was used.
(3) Because the board operators are directed to do a section of a procedure vice a specific step, they must carry the procedure manuals with them.
This process is very cumbersome for the operators and could lead to a manual being dropped on the panels, possibly causing the status of an emergency ccmponent to be changed.
5.
The written examinations were reviewed by the utility and discussed with the examiners after all candidates had completed the examination on October 27, 1987.
The facility's comments (Attachment 3) and the NRC resolution (Attachment 4) are enclosed. The following facility individuals reviewed the examinations:
R0:
J. Hill, D. Holm, R. Scott, and E. Chrzanowski SRO:
J. Hill, D. Holm, R. Scott, J. Yoe, and C. Andrews Attachments:
I 1.
R0 Written' Examination and Answer Key j
2.
SRO Written Examination and Answer Key l
3.
Facility Comments on Written Examinations
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4.
NRC Resolution to Facility Comments
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a U. S '. NUCLEAR REGULATORY COMMISSION
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REACTOR OPERATOR LICENSE EXAMINATION
i FACILITY:
CALVERT CLIFFS
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REACTOR TYPE:
PWR-CE DATE ADMINISTERED:
87/10/27
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EXAMINER:
ISAKSEN. P.
CANDIDATE 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
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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 100.0
%
Totals Final Grade All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature NEER COPY
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS i
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During the administration of this examination the following rules apply:
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1.
Cheating on the examination means an automatic denial of your applicatic!
and could result in more severe penalties, j
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.
I 5.
Fill in the date on the cover sheet of the examination (if necessary).
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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 only 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.
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.
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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 ofl the examiner only.
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17. You must sign the statement on the cover sheet that indicates that the j
I work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination hat been completed, i
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18. 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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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pace
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THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.01 (2.00)
The ratio of the PU239 and Pu240 atoms to U235 atoms increases over core _ life. Explain the effect this ratio change has on-
'the following and WHY:
a. Delayed neutron fraction (1.C b. Doppler Temperature Coefficient (1.C QUESTION 1.02 (2.00)
a.
How is Shutdown Margin (SDM) affected (Increase, Decrease, or No change) by a 50 ppm boron addition'while operating at 50% power?
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b.
List THREE factors, other than RCS boron concentration and rod position, which will affect SDM-and are used in-the SDM calculation.
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QUESTION 1.03 (1.50)
a.
One bases for the Technical Specification (TS) CEA Insertion Limit is to ensure sufficient SDM is available.
What are the othe2 TWO bases?
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The CEA Insertion Limit is a function of
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.(Fill in the blank, place your answers on your answer sheet.)
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QUESTION 1.04 (2.00)
l a. Explain HOW and WHY ASI is expected to change as power is increased
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from 20% to 70%, during a normal power increase at EOC.
b. Explain what TWO steps / methods are taken to maintain ASI within lim:
and WHY these actions are effective.
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
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THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.05 (2.00)
At BOC, power is reduced from 100% to 50% and stabilized, briefly explain HOW and WHY each of the following plant parameters will be affected over the next 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
Assume all systems are in automatic, rod control is in manual sequential, and no operator action is taken, a.
RCS temperature b. RCS pressure c.
S/G pressure d. Turbine Generator control valve position QUESTION 1.06 (3.00)
You have just completed a reactor startup and power level is at the point of adding heat. For the following situations, Indicate whether final power level will be (HIGHER, LOWER, OR THE SAME) in reference to initial power level and EXPLAIN your answer. (Assume the core is at mid-life, no operator action and treat each situation separately).
a.
Steam dump pressure setting is raised by 20 psig, b. A 1% steam leak develops outside of containment.
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c. An inadvertent 20 ppm boron addition is made.
QUESTION 1.07 (1.00)
Why is a main steam line break a more severe accident at EOC than BOC?
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
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THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
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l QUESTION 1.08 (1.50)
Compare the calculated Estimated Critical Condition (ECC) for a startup.
to be performed four hours after a trip from 100% power, to the actual control rod position, if the following events / conditions occurred.
Consider each independently.
Limit your answer to actual control rod i
position is (Higher than, Lower than, or Same) as the ECC.
a.
One reactor coolant pump is stopped two minutes prior to criticality.
b.
The steam dump pressure setpoint is increased to a value just below the steam generator 4PERY'setpoint.
S AF vW c.
All steam generator levels are being raised by 5% as criticality is being reached.
QUESTION 1.09 (3.00)
The plant is in a Natural Circulation Mode of core cooling.
As the fission product heat decays away, describe HOW and WHY you would expeci, the following RCS parameters to change. Assume that S/G pressure is being maintained constant at 900 psia.
a.
Teold b. Thot c. Core delta T d. Loop transit time
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
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THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW I
QUESTION 1.10 (2.00)
The discharge valve of a running, motor operated, centrifugal pump is-throttled (valve moved in the shut direction).
Indicate whether each of the following will INCREASE, DECREASE, or REMAIN UNCHANGED.
j a.
Pump motor amps.
b.
Pump discharge pressure.
c.
Pump discharge flow.
d.
Actual NPSH available at the pump (Assume fluid temperature remains constant).
QUESTION 1.11 (2.00)
a. List THREE of the RCS parameters on which the DNB Heat Flux (CHF)
is dependent?
b. At what location in the core, top, bottom, or middle, is the fuel the furthest from DNB? (i.e. Where is the DNB Ratio the largest?)
EXPLAIN your answer.
QUESTION'
1.12 (3.00)
After a secondary calorimetric and adjustment of the power range instruments, it is discovered that the Auxiliary Feedwater Pumps were operating.
State HOW and WHY the indicated power is more or less conservative than actual power.
THREE reasons required.
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page
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SYSTEMS
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QUESTION 2.01 (2.50)
a.
What is'the purpose of the CVCS Excess Flow Check Valve (EFCV)? ( 0.. !
b.
State the flow rate that will shut the EFCV.
(0.fj c.
Where is the EFCV located in the CVCS' system?
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How is the pressure equalized around the Excess Flow Check Valve so it may be reopened?
NOTE:
Include unit differences that may J
exist.
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QUESTION 2.02
.(1.00)
What automatic action occurs to the Containment Cooling Fans AND their associated support systems following a CSAS?
QUESTION 2.03.
(2.50)
a.
What TWO signals must be generated by the ESFAS in order to admit spray water into the containment?
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b.
How does the generation of a RAS signal affect the operation'of the containment spray system?
(1.t c.
How is the trisodium phosphate added to the containment spray system?
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QUESTION 2.04 (2.00)
Describe what automatically happens in each of the following systems upon receiving a SIAS signal.
a. Chemical and Volume Control system (Six actions required)
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b. Service Water system (Two actions required)
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page
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QUESTION 2.05 (3.00)
The following questions concern a loss of instrument air assume a normal at power lineup as the initial conditions.
a. How would a loss.of instrument air header pressure due to a ruptu just downstream of IA-144/146 (Air compressor isolation from air header) immediately affect the following components / systems.
Choose ONE of the following for each component / system:
A fail open or flow maximum B
fail closed or flow stopped C
fail as is or flow cannot change D
no immediate effect or system functions normally 1) Main Feedwater Regulating Valves er 2) Pressurizer spray valves 3) Letdown 4) Atmospheric Dump Valves i
5) AFW regulating valves 6) EDG service water supply valves 7) Auxillary Spray valve
[0.25 each]
8) Operating Turbine AFW pump b. Describe TWO means of interconnecting the IA system with backup sources of air pressure. Indicate automatic setpoints, if any.
Answer this question independently of part a, above.
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QUESTION 2.06 (2.50)
a.
State TWO design features which serve to prevent the loss of water inventory in the Spent Fuel Pool (SFP).
(1.t b.
List THREE sources of makeup water to the SFF.
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PLANT DESIGN INCLUDING S WETY AND EMERGENCY Page.1
SYSTEMS g
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QUESTION 2.07 (3.50)
i a.
.What condition (s) will generate a AFW pipe rupture signal?
Include >
setpoints.
(1.0 b.
Explain how the AFW system automatically responds to a faulted S/07:
Include setpoints and coincidence to generate the AFAS block signal ( 1. t i c.
The plant Fire Protection system (FPS) provides an emergency source of water to the AFW system.
How is this (FPS) emergency source of water lined up to the AFW system?
(1.C QUESTION 2.08 (3.50)
a.
What operator action is required to restart an Emergency Diesel Generator (EDG) that has experienced an engine trip after an auto-matic start and the automatic start signal is still present, the cause for the trip has been found and corrected.
(1.'
b.
What TWO EDG shutdowns are bypassed on a SIAS auto start?
(1.C c.
If a loss of power occurs without a LOCI signal (no SIAS), HOW doer the Shutdown Sequencer (SDS) respond?
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How are the CSAS and CIS subchannels affected, if a loss of power
. occurs when a SIAS signal is present?
(1.C QUESTION.
2.09 (2.50)
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a.
State the interlock (s) associated with Shutdown Cooling (SDC) hot leg suction isolation valves.
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There are TWO relief valves in the SDC return header, for EACH of these valves indicate the following:
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2. Where they relieve to.
3. State the portion of the SDC system protected AND the specific condition that the overpressure protection is based on.
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page'l
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QUESTION 2.10 (2.00)
i The following concern the failure of a CEA lift coil with the affected
CEA in a withdrawn configuration, assume the reactor remains at power.
a.
Explain why the affected rod will/will not drop.
b.
Explain why the affected rod will/will not move on a demand signal.
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INSTRUMENTS AND CONTROLS PagcL1
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-QUESTION 3.01 (2.50)
a.
. List the FOUR-alarm signals generated by the CEA Motion Inhibit Circuit.
Do not include the Out of Sequence signal.
(1.0 b.
What THREE conditions will cause a CWP interlock to activate?
(1.5 QUESTION 3.02 (1.50)
The. plant is operating at 85% power with pressurizer level transmitter LT-110x' selected and HS-100-3 in the X+Y position..
State the expected automatic system responses, in addition to receiving alarms, if LT.110x fails LOW.
QUESTION 3.03 (3.00)
a. How many Hot and Cold Leg Temperature instruments are.there in a single loop.
(0.E b. How many of.each type (T-hot and T-cold) are used for protec-
. tion AND what are they used for in the protection system?
(1.C c. What specific SYSTEM (s) are controlled from signals derived from the loop temperature instruments?
TWO required.
(1.E QUESTION 3.04 (3.00)
a.
What THREE inputs are used to generate the APD trip setpoint and WHERE is each input received from? Do not include RCP combination input.
(1.E b. What initiates an APD channel trip signal and HOW is the trip setpoint determined?
(1.C c. In addition to being used to provide a reactor trip, what other function does the APD signal provide?
Do not include indication.
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INSTRUMENTS AND CONTROLE Page 1
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QUESTION 3.05 (3.00)
What instruments are used to detect the following possible leakages:
a. Safety injection header check valve leakage.
b. Letdown heat exchanger tube leakage.
c. Pressurizer relief valve leakage.
QUESTION 3.06 (1.50)
What are the THREE positions of the main control board S/G Low Pressure Pressure Trip Bypass switch AND when are the positions used?
QUESTION 3.07 (1.00)
Explain the protection afforded by the CVCS Isolation signal AND how the protection is provided. (setpoints not required)
QUESTION 3,08 (2.50)
a.
What are the TWO signals that are auctioneered to provide a Turbint
bypass valve actuating signal?
(1.C b.
How are the atmospheric dump valves and turbine bypass valve
" Quick Opened"?
Include the condition necessary to cause a " Quick Opened" signal to be generated and the system / component response to perform the valve operation.
(1.f QUESTION 3.09 (3.00)
a.
What automatically happens when reactor power decreases to a point where EXTENDED RANGE MODE is activated for the wide range log channels?
Explain WHY.
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b.
At 10-4% increasing reactor power the level 2 bistable trips on.
What THREE automatic actions does this perform, other than energize the level 2 status lamp?
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INSTRUMENTS AND CONTROLS Page 1
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QUESTION 3.10 (2.50)
a.
What would be the effect on #11 S/G feedwater control AND level, if the #11 feedwater regulating valve differential pressure controller failed such that differential pressure went to a maximum?
(1.5 b.
Explain what will happen to the feedwater control system and S/G level as a result of a main turbine trip, while operating at 75% power.
(1.0 QUESTION 3.11 (1.50)
Indicate which of the following monitor channels have automatic actions'
associated them (other than indication and alarm)?
Briefly describe th automatic action.
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1. CCW radiation monitor 2. Liquid waste discharge monitor 3. Main vent APD's 4. Waste gas discharge monitor (***** END OF CATEGORY 3 *****)
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PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page :
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l QUESTION 4.01 (3.00)
a.
For each of the following operating conditions, state what instruments are used by the operator to determine RCS temperature, according to OP-1, Plant Startup From Cold Shutdown procedure.
1. RCP operating 2. Shutdown cooling in operation 3. No RCP running and shutdown cooling stopped b. How is core cooling provided if component cooling water is lost during a LOCA and cannot be regained as per AOP-7C Loss of Component Cooling?
QUESTION 4.02 (2.50)
What TWO (2) independent indications should be used to evaluate or corroborate each of the following plant parameters following a reactor trip?
a. All CEA's fully inserted b. Pressurizer level stabilized c. RCS subcooling greater than 30 F d. Proper operation of turbine bypass / atmospheric dump valves e. Main vent activity not increasing QUESTION 4.03 (2.50)
a. Under what TWO (2) general conditions should operators adopt
" manual" operation of automatically controlled systems?
According to E0P-0 Post Trip Immediate Actions procedure.
(1.t b. Under what THREE (3) conditions should the Functional Recovery Procedure, EOP-800, be implemented.
(1.'
QUESTION 4.04 (3.00)
List the SIX Safety Functions which are verified during the performanc-of the EOP-0 post trip immediate actions.
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4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 1
AND RADIOLOGICAL CONTROL
,
QUESTION 4.05 (1.50)
a. What are the Calvert Cliffs maximum administrative limits concerning weekly, quarterly, and yearly whole body radiation dose for individuals older than 18 years?
[0.9)
b. Whose approvals are necessary prior to exceeding the weekly, the quarterly, or the yearly whole body radiation dose?
[0.6]
QUESTION 4.06 (1.00)
The following questions concern AOP-9 ALTERNATE SAFE SHUTDOWN PROCEDURE / CONTROL ROOM EVACUATION procedure.
a. Where do the Unit 1 Control Room Operator and Reactor Operator go INITIALLY if the control room must be evacuated?
b. Who is designated to go to the Emergency Diesel Generator rooms?
QUESTION 4.07 (1.50)
During a natural circulation cooldown, RCS voiding is indicated and the AOP-3F NATURAL CIRCULATION COOLDOWN actions of stopping letdown, stopping the cooldown, and pressurizing the RCS to maintain 200 degree subcooling are NOT effective in eliminating the RCS voids.
What other TWO general methods could be used to reduce or eliminate the voided area?
QUESTION 4.08 (2.00)
Following an inadvertent reactor trip, TWO CEA's do not fully insert, what actions should be taken.
Include both the components to be opera:
and the point at which the steps are considered complete.
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 1!
'*-
AND RADIOLOGICAL CONTROL j
~
l-QUESTION 4.09 (2.50)
q
During a loss of off site. power, natural circulation must be establish ( !
a)
While establishing the SG'as a heat sink, increased loop transpo2 time causes a 5 to 10 minute delay in. temperature responses to a I plant change.
What TWO plant parameters provide better indicatic '
of RCS response during this period?
(1.C j b)
Give THREE plant conditions that can be observed'in order to l
verify natural circulation.
( 1. ! {
,
-QUESTION 4.10 (1.00)
State the TWO methods for restoring refueling pool level uponla cavity seal failure per AOP-6E recovery actions.
(Assuming'a fuel j
assembly may be: uncovered.)
QUESTION 4.11 (2.00)
According to GSO Standing Instruction #83-12, what actions are require in the event of a loss of power to 11 DC BUS?
WEY are these actions-i required 9
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CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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4.
' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 1
AND RADIOLOGICAL CONTROL
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QUESTION
.4.12 (2.50)
The following pertain to AOP-2A, Excessive Reactor Coolant' Leakage, a.
Define the following:
1.
Major Leak 2.
Minor Leak
'(1.C b.
During the performance of AOP-2A, the reactor must be manually tripped if certain minimum parameter limits are exceeded.
For the following parameters. state the limit which would require a' manual trip.
1.
Pressurizer Pressure
'
2.
Pressurizer Level 3.
Tavg (1.E (***** END OF CATEGORY 4 *****)
(********** END OF EXAMINATION **********)
l
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--. _ _ _
.a
,
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- /
EQUATION SMEET
.
,
-
f = ma v = s?t
,'.,,
"I '
,,yg+
g,g Cycle efficiency =
,
E = aC a = (vg - v,)/t
-
KE = bev v
A = AN g = v, + at A = A,e pg'= agh w = e/c 1 = in 2/tq = 0.693/tg
,
W = v&P-
. t (aff) = (t,.)(ts)
--
g AE = 931Am (
,
)
.
Il=[ scat I = I 48 P
,
,
o Q = UAAT I = I, ux
-
,
Pwr = U' 5" to / M-x g
I=I
,
g
').
M = 1.3/u P=P
-
y.y.t/T HVL = 0.693/u
.
.o
- SUR = 26.06/T
~
T = 1.44 DT SCR = S/(1 - K,gg)
/A oT
SUR = 26 CR = S/(1 - K,gg )
g x
1(
eff}1 (
eff)2 T = '(1*/o ) + [(i ' p)/1,g,p}
~
7. g*/ (, _ p M = 1/(1. - K,gg) = CR /CR
-
g
T = (I - 9)/ 1,ggo y, gy, gaff)0 Il ~ Esff)1 8 * CEsff'1)IEeff " AEsff/K,gg.
SDM = (1 - K,gg)/K,gg
[L*/TK,'gg.] + [I/(1 + 1,gg )]
1* = 1 x 10 ' seconds
~
T p=
,
P = I(v/(3 x 1010)
1,gg = 0.1 seconds A
I = Na
-
Idgy=Id22
.
WATER PARAMETERS Id =Id g
2 1 gal. = 8.345 lba R/hr = (0.5 CE)/d g,,,,,,)
I gal. = 3.78 liters R/hr = 6 CE/d (feet)
I ft3 = 7.48 gal.
MISCEI.L\\NEOUS CONVERSIONS
.
Density = 62.4 lbm/ft 1 Curia = 3.7 x 10 dps Density = 1 gm/cm i kg = 2.21 lba Heat of vaporization = 970 reu/lba 1 hp = 2.54 4 103 BTU /hr
Heat of fusica = 144 Bru/lkm 1 Hw = 3.41 x 10 Btu /hr 1 Atm = 14.7 psi = 29.9 in. I'g.
1 Stu = 778 ft-lbf
1 ft. H O = 0.4333 lbf/in g inch = 2.54 cm
F = 9/5*C + 32
"C = 5/9 (*F - 32)
__ _
>
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
Page 1 i THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW WfBCO?Y ANSWER 1.01 (2.00)
~
l a. Delayed neutron fraction decreases [0.5) because the beta is less for Pu239 as compared to U235. [0.5]
(1.0J b. Doppler Coefficient is more negative [0.5] because Pu240 has a higher resonance cross section than U235. [0.5]
(1.C REFERENCE CC LP #RO-302-2-1, Reactivity Factors, EO 3.2, 7.4 i
LP #RO-302-3-1, Rx Kinetics, EO 3.3
,
CE-Nuclear Physics, Reactor Theory and Core Opera *.ing Characteristics, i p 153 - 156.
192004K107 064050K606
..(KA's)
ANSWER 1.02 (2.00)
a.
SDM is increased. [0.5]
b.
[any 3, 0.5 each]
-RCS avg temp-Samarium-Fuel burnup-Power defect-Xenon concentration-Power level REFERENCE CC LP #RO-302-3-1,Rx Kinetics, EO 8.5.
192002K114 041020K603
..(KA's)
ANSWER 1.03 (1.50)
a.
1.
Core design peaking factors are not exceeded (acceptable power distribution limits).
2. The reactivity associated with a CEA ejection accident is accept able (within analysis).
[0.5, each]
b.
Core power.
[0.5]
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
Page 2~
,
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW REFERENCE
~
CC TS 3.1.3.6 and bases.
~
C-E Reactor Theory, pp 193-194.
192005K115 010000K607
..(KA's)
ANSWER 1.04 (2.00)
a.
Moderator density becomes less at the top of the core [0.F] causing the flux peak to move down in the core. [0.2J ([Since ASI is 1-u/1, 444GdiJgg.willbecomemorepositiveasthepowerisincreased. [0.f b.
1. Reduce power which creates less restrictive limits. [0.5]
(more neg. ASI due to density changes in moderator associated wit delta-T and Tc program)
2. Use rods to change flux shape which changes the value of ASI.[0.5]
REFERENCE CC LP #RO-302-4-0, ASI, EO 2.1, 2.2, 4.3 s
TS 3/4 3.1 192005K110 192005K114 011000K303 011000K302 011000K301
..(KA's)
ANSWER 1.05 (2.00)
a.
Decreases [0.25] due to buildup of Xe [0.25]
b.
Held constant [0.25] by PPCS spray and heaters [0.25]
c.
Decreases [0.25] due to the decrease of Tavg [0.25]
[ graded based upon answer in a, above]
d.. Remains the same [0.25] since load is lowered on valve position limiter.(and no of crator action is assumes.CoJ53
[2.C
~ % --
e+a m w vi=-ALo.u] w+,^
sfs p = - - _ ; t. u3
-
REFERENCE CC LP #RO-302-3-1, Rx Kinetics, EO 7.13; SD #5 p 25; SD #23-1 p 54 l
192006K106 011000K303 011000K302 011000K301
..(KA's)
l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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e 1.
FRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
Page 2 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
~
.
~
ANSWER 1.06 (3.00)
a. Lower [0.25]; the steam pressure increase causes RCS f
temperature to increase (MTC 44h4MF and FTC (Doppler)
both) adding negative reactivity to lower reactor power.[0.75]
b. Higher [0.25]; the increased steam flow results in a lower RCS temperature (MTC will) adding positive reactivity [0.75]
c. Lower [0.25]; the negative reactivity inserted by the boron will cause power to decrease [0.75].
REFERENCE CC LP #RO-302-3-1, Rx Kinetics, EO 7.11, 12, 15 CE Nuclear Physics, Reactor Theory and Core Operating Characteristics,
,
p 162-166, 178 192008K117 015000408 015000K407 015000K405 015000K402
..(KA's)
,
ANSWER 1.07 (1.00)
At EOC the MTC is more negative [0.5], the cooling of the RCS adds more positive reactivity to the core at EOC than at BOC [0.5].
REFERENCE CC LP #RO-302-2-1, Reactivity Factors, 20 6.7 192004K106 002000K603
..(KA's)
<
!
ANSWER 1.08 (1.50)
l i
a.
Same b.
Higher c.
Lower REFERENCE CC LP #RO-302-5-0 ECC 192008K108 013000K403
..(KA's)
l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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PRINCIPLES OF NUCLEAR POWER' PLANT OPERATION.
Page 2-THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
.
_
1:
~
ANSWER 1.09 (3.00)
a. Teold will remain constant [0.25] Since it follows S/G saturation temperature.
[0.5]
l b. Thot will decrease [0.25] since less fission product heat is being l
produced than is being removed by the steam generators.
[0.5]
c. Core delta T will decrease [0.25] since the amount of decay heat'is decreasing.
[0.5]
d. Loop transit time will increase [0.25] since the driving head for flow (core delta T) is decreasing.
[0.5]
REFERENCE
'CX: LP #RO-301-14-0, Natural Circulation, EO 14.2.3,-4 and 14.4.2 193008K122 045000K411
..(KA's)
ANSWER 1.10 (2.00)
a.
Decrease
'
b.
Increase c.
Decrease d.
Increase
[0.5 each]
REFERENCE CC LP #RO-301-11.1-0, Fluid Flow, EO 1.3 and 3.1 i
193006K113 014000K601
..(KA's)
i ANSWER 1.11 (2.00)
{
l a. Flow (
Temperature J
Pressure Power
[any 3, 0.33 each]
b. Bottom [0.5], because this is where the temperature (0.25]
is the lowest and pressure the highest [0.25].
(1.C i
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CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
Page 2 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
<
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REFERENCE
-
CC LP #RO-301-13-0, Rx heat Generation, EO 13.5,-6 and 13.6.3
~
CE Thermal Hydraulics, p 14 193008K105 035010K101
..(KA's)
ANSWER 1.12 (3.00)
The indicated power is LESS conservative [0.75]
1.
Actual feedwater temperature would be lower than that used in the calorimetric calculation.[0.75]
2.
The feedwater mass flow used in the calculation would be lower than actual. (Since AFW flow bypasses feedflow indication.)[0.75]
3.
Indicated power would be less than actual [0.75].
REFERENCE CC LP #RO-301-10-0, Plant Cycle Analysis, EO 10.2.2 193005K103 035010A204
..(KA's)
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(***** END OF CATEGORY 1 *****)
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2.
PLANT ~ DESIGN-INCLUDING SAFETY AND EMERGENCY Pagef2
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SYSTEMS
.
.
.
~
ANSWER 2'01 (2.50)
.
.a.
To minimize the' consequences of a CVCS letdown line rupture.
.[0.5 b.
210.(+/
20 gym.) [0.5]
c.
Downstream of the regenerative heat exchanger (and upstream of containment penetration)
[0.5]
l d.
(The letdown isolation valve is shut) Unit 1 has a bypass line l
[0.4] with an orifice [0.1] around the check valve.
Unit 2 has a hole'in the disc of the valve [0.5].
REFERENCE
~
l
.CC,.SD 6, PP. 12-13 and Fig. A-3.
004020A108 004010A402 035010K401
..(KA's)
'
ANSWER 2.02 (1.00)
All fans start or shift to low speed
[0.5]
and 8' inch service water valves on cooler outlets receive an open signc
[0.5]
REFERENCE S.D. 63 pg. 169 S.D.
39 pg. 20 022000A301 073000K101
..(KA's)
ANSWER 2.03 (2.50)
a)
Containment Spray Actuation Signal AND Safety Injection Actuation Signal [1.0]
.b)
Opens the containment sump isolation valves [h cnd :hutre-t-he*
ettnimum flow toc 14 alation line isolation valvo. [hE.Pt c)
Water on the containment floor dissolves it and carries it to the containment spray system [0.5]
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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~.2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page E,
'
SYSTEtiE.'
.
REFERENCE
~
~
026000K401-026020K403 026020K401 072000K401 072000K102
..(KA's)
ANSWER 2.04 (2.00)
CL,s.4-
, 0. 2 s '
'3 1. Boric @ acid pumps start -[0.25]%
a.
2. Charging pumps start [0. 25] e 3. Boric acid storage tank is lined up to inject boric acid
.
.
an oc u A-r-e t A w c cv co r, s e,) & -
!
-
' " ' ~ ~ ' ' N C + o s r,'i) u m y - A w%_-2 4. VCT makeup stop valve shuts E-0.251.
5. ' VCT outlet valve shuts [4rBR Ccue sts Letdown line loop isolation valve shuts -E-Gr35k., sis.)
(1.
6.
,
ALPA,t- + u vc. r C c a s oc, s a c.) &
'
1.
b.
1. Two service water pumps start [0.25]
2. The turbine building SRW isolation valve shuts [0.25]
004010A205 026000K101
..(KA's)
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CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page'E
"
SYSTEMS
.
~
ANSWER 2.05 (3.00)
a. 1)
C 2) Jy-D 3) #D
[0.25 each]
4)
D 5)
D 6)
A 7)
D 8)
A b.
1) Auto valve to plant air system X-ties FA to IA at 85 psig IA pressure.
2) Manual X-tie valve to Saltwater system air compressors.
3) X-connect Units 1 and 2 plant air systems.
[2 required, 0.5 each]
REFERENCE S.D.
32 pg 15 AOP-7D pp. 1,3-6
-
S.D. 41 Fig A-7 to A-9 S.D.139 pg 21 078000K302 008000K401
..(KA's)
ANSWER 2.06 (2.50)
a.
1. Stainless steel liner plate.
2.
No penetrations in pool wall below the normal water level (ex. fuel transfer tube)
3. Siphon breakers for penetrations above the normal water level.
q, W C r' gv,
[2 required, 0.5 each]
or A CnD b.
1.
SFP cooling and purification system 2. Refueling water tankst u-l 3.
Mrin
"nter syster-EM "'1 4. Fire hose
[3 required, 0.5 each]
033000K401 001000K402
..(KA's)
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'2.
PLANT' DESIGN INCLUDING SAFETY'AND EMERGENCY Page 2
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SYSTEMS
.
_
ANSWER 2.07 (3.50)
~
a.
Either turbine or. motor driven AFW line flow to S/G >100 gpm AND 175 psid or greater d/p between S/G and AFW line to S/G [1.0]
ci35S
'
b.
2/4 [0.25] d/p S/G >115,psid [0.25] will shut the 2 blocking valver in each AFW line to the faulted S/G [1.0].
c.
A fire hose is attached to the suction line connections of the mote'
driven AFW pumps (13,23) [1 0].
061000A304 012000K603 012000K103
..(KA's)
ANSWER 2.08 (3.50)
a.
Depress the ALARM RESET pushbutton (at the diesel gageboard) [1.0)
b.
-Jacket coolant low press.
'
-Jacket cooling high temp.
[0.5 each]
-
c.
The SDS will automatically energize selected equipment at (5 sec.)
programmed intervals (ie. service and saltwater pumps, instrument air compressors, CR and SWGR A/C units) [0.5]
d.
(SASB and LOCIS signals will work to) The CSAS, CIS subchannels will initially be blocked then unblocked at programmed intervals.
[1.0]
REFERENCE CC SD 48, p 152-167.
064000A307 064000K402 064000A401 045000SG11
..(KA's)
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' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 2-
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SYSTEMS
-
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ANSWER 2.09 (2.50)
a.
Valves will automatically shut if PZR press. exceeds 300 paia.[0.5]
b.
1.
2485 (+15/-100), 315 (+/- 15) psia. [0.25 each]
~
"C_ drain tank (LPOI vamy,ucLivu) [0.5] L 3. The piping between the SDC header isolation valves [0.25] from the pressure developed due to a sudden temperature increase in the containment [0.25]
The SDC flow path [0.2] from the overpressure transient due to
)
simultaneous operation of the charging pumps and SDC [0.15] with !
the PZR in a solid condition [0.15].
= A & = L L o *"~3
'
g.v % n v i & _ = n s -- + r. e-c. n -
REFrRENCE t-v %g u+_N w w a e~,=re. 9 e 3 '. xe mc m.way
,
La
.
CC SD 7, p 42, 43 and Figures 7-2 & 7-3.
005000K604 005000K407 000015SG07
..(KA's)
ANSWER 2.10 (2.00)
a.
The rod will NOT drop [0.5] due to the action of the lower
~
gripper [0.5].
b.
The rod will NOT move (up or down) [0.5] since the lift coil is uee to raise the upper gripper in either direction [0.5].
001000K402
..(KA's)
<
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3.
INSTRUMENTS AND CONTROLS Page 2:
,.
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l ANSWER 3.01.
(2.50)
_
a.
PDIL (Power dependant insertion limit)
Deviation (secondary group deviation)
MISH (Regulating bank withdrawal permissive)
_
)
'
MIRG (Shutdown bank insertion permissive)
[0.25 each]'
b.
1.
High power level pretrip 2.
High SUR pretrip 3.
TM/LP pretrip
[0.5 each]
001000K408 001000K407 003000K602
..(KA's)
ANSWER 3.02 (1.50)
1.
Starts all backup charging pumps 2.
Letdown signal to minimum 3.
Lo-Lo level heater cutout
[0.5 each]
REFERENCE
-
011000K513 011000A203 000025K303
..(KA's)
ANSWER 3.03 (3.00)
a. Five T-hot in each loop (located between Rx. Vessel and Steam
,
Generator). [0.4]
Three T-cold per loop (located between Coolant pump and Rx.
Vessel). [0.4]
(0.8 b. Four in each hot leg and two in each cold leg [0.5]. They provide temperature (and Delta-t) signals to develop the TM/LP trip setpoint[0.3] and high power level trip. [0.2]
-(1.C 2.= - v - 3 c. Pressurizer level, Steam dump and Turbine by-pass system [0.6'each]
m PT ( L.rDP).
(1.2 (*****
CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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3.
INSTRUMENTS AND CONTROLS Page S
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l
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REFERENCE
_
CC SD 62, p 3, tr-lo 4 D M-3 016000K501 001000SG11 035000SG11 016000SG11 010000SG11
..(KA's)
ANSWER 3.04 (3.00)
a.
1. Thermal power [0.25]-generated by the TM/LP calculator [0.25]
2. Nuclear power [0.25]-from the nuclear instruments [0.25]
3.
CEA Function [0.25]-fixed input from the safety analysis
[0.25]
b. A channel trip occurs if the axial shape index (YI) exceeds a positive or negative limiting value (Yp or Yn)
[0.5] determined by Qmax, the largest of either NI power or thermal power [0.5]
c. Also generates the axial offset used in the TM/LP calculator [0.5]
REFERENCE CC SD 59, p 21,23 and fig. A-8.
-
015000K405 041020A200 001000SG11
..(KA's)
'
ANSWER 3.05 (3.00)
a. Press Transmitters (4$dbetream of the check valves leak off indicators).
[0.5]
b. CCW head tank level + rad monitor. [0.5 each]
c. Hi temp alarm (temperature) [0.3]
QT level [0.3]
QT temp [0.3]
QT pressure [0.3]
Acoustic
[0.3]
002000K405 000001SG10 000037SG10 002000SG10
..(KA's)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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INSTRUMENTS AND CONTROLS Page 3
,
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_
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!
l ANSWER 3.06 (1.50)
,
BLOCK-Used during plant cooldown.
' RESET-Used to reset a trip signal after the condition clears.
NORMAL-Used during' normal operation (spring return to normal)
[0.5 each]'
012000K406 064000K102
..(KA's)
ANSWER 3.07 (1.00)
Designed to limit the consequences of a letdown line break outside containment [0.3]. Protection is provided by sensing the pressure in the West penetration and letdown HX rooms [0.3)
.the letdown isolation valves are shut on increasing pressure.[0.4]
004000K403 061000K505
..(KA's)
~
ANSWER 3.08 (2.50)
a.
1. Main steam pressure error signal 2. Programmed Tave error signal
[0.5 each]
4 r w s >sst'e b.
mmovc a specific rerer icvcita " quick opening" solenoid valve is energized when the turbine trips. [0.75] This solenoid provides air directly to the valve actuator opening the valves (bypassing the I/P converter which normally provides the modulating signal) [0.75).
REFERENCE CC SD 19, Figure A.S.
039000G007 076000K604
..(KA's)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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3.
INSTRUMENTS AND' CONTROLS Page 2-y
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ANSWER 3.09
.(3.00)
_
a.
The' dual'section proportional counter.is energized [0.5) to provide increased detection. sensitivity [0.5].-The indication changes from percent power to counts per second [0.5]
(1,f b.
1. Enables the high SUR trip.
2.EnablestheSURanqj$M/LPpretripinputstotheCWP circuit.
% Enables the metrascope PDIL circuit.
[,3%J0.5each]
012000K406 004000SG07
..(KA's)
ANSWER 3.10 (2.50)
a.
The feedwater pump would receive.a signal to cause the pump to slow down (to reduce the d/p) [0.75). This would result in a lower flowrate to the S/G and level would decrease [0.75]
(1.!
( em s=- FA Aw ev=-
= n. r : o y 4 n ~.= n )
b.
.The feedwater regulating valves shut and the bypass' valves go to.'
a' position equal to 5% of full feedflow [0.5].. Level will decrease rapidly due to shrink-[0.5]
(1.(
REFERENCE CC SD 32, p 16 & 21.
059000K418 059000A211 194001K103
..(KA's)
ANSWER 3.11 (1.50)
1. None[0.25]
2. Closes two discharge isolation valves to terminate the release [0. 5:
3. None[0.25]
4 '. Closes redundant waste discharge isolation valves [0.5]
(1.!
'
REFERENCE
- CC SD 15 p 20-48.
073000K401 194001A105
..(KA's)
(***** END OF CATEGORY 3 *****)
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a
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4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pcge 3
,
AND RADIOLOGICAL CONTROL
_
~
ANSWER 4.01 (3.00)
a.
1.
Cold leg temperature (Tc)
[0.5]
2. SDC temperature recorder (TR-351) [0.5]
3. Average of at least two core exit thermocouple
[0.5]
b. Line up containment spray pumps to provide core cooling. [1.5)
REFERENCE CC OP-1, p 1; AOP-7C.III.
000074E201 005000G008 008000G005 194001A103
..(KA's)
ANSWER 4.02 (2.50)
a. Rodbottom light (rod bottom reed switch)
Metroscope (reed switches)
b. Two independent level channels (RRS channels x and y)
c. Subcooling monitor RCS temp. and press. that do not feed the subcooling monitor d. S/G pressures stabilizes (between 850 and 920 paia.)
-
Tc stabilizes (between 525 and 535 F)
e. Main vent particulate monitorC M a *n-4 Main vent gaseous monitor
[any 2 for each part @ 0.25 each]
000007K301 000007A202
..(KA's)
ANSWER 4.03 (2.50)
a. Apparent malfunction [0.5]
Automatic action will not support the maintenance of a Safety Function [0.5]
b. Rx trip with no immediate apparent cause [0.5]
'
Conditions which threaten safety functions for which no procedural guidance can be immediately identified. [0,5)
Emergency procedure actions do not satisfy the safety function criteria. [0.5]
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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PROr'EnURES - NORMAL. ABNORMAL. EMERGENCY Page S
'-
.AND RADIOLOGICAL CONTROL
,
REFERENCE
-
~'
CC EOP-0, p 5; E0P-800, p 3,4.
l 012000G001 000029K312
..(KA's)
~
l..
l l
ANSWER 4.04 (3.00)
a) ' Reactivity control b)
RCS pressure and inventory control c)
Core and RCS heat removal d)
4 kv bus 11 or 14 energized (vit i c=We =)
l e)
Normal containment environment-f)
Normal radiation levels external to containment REFERENCE-l CC EOP-0, p 5-B.
l 012000G014 000007K301 194001A102
.,(KA's)
ANSWER 4.05 (1.50)
e a. Weekly 000 Lw.er-Quarterly * ^ "
- 9C N
Yearly 4.0 Rem
[0.3 each]
b. Individuals General supervisor AND General supervisor-radiation safe-
[0.f
,
REFERENCE P
CCI-800B, p 9-10; N w M F W " ~d'#** * *
000055K302
..(KA's)
-
t ANSWER 4.06 (1.00)
a. Unit 1 45 foot Switchgear room
[0.5]
b. Outside operator
[0.5]
(***** CATEGORY 4 CONTINUED ON NEXT-PAGE *****)
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4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pega 3.
- AND RADIOLOGICAL CONTROL
,
~
~ REFERENCE
~
00006BK318 103000A101
..(KA's)
.
ANSWER 4.07 (1.50)
1) Cycle RCS pressure (within the limits of EOP Attachment 1).
[0.75)
2) Operate Reactor Vessel head vent (per OI-1G)
[0.75)
REFERENCE
.
000074K311 103000A101
..(KA's)
.
ANS*2ER 4.08 (2.00)
eea3C*
Borate the RCS 4 W ppm.
[0.5)
by 1) opening charging' pump suction direct feed valve (CVC-514-MOV) - [ r.
,
2) starting a boric acid pump
[0,5)
3) starting all-available charging pumps
[0.5)
U-001000G014 192002K108
..(KA's)
l l
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
l l
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'4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 3
AND RADIOLOGICAL CONTROL
.
<
.
ANSWER 4.09 (2.50)
~
a)
Pressurizer level AND pressure [1.0]
b)
Any THREE required, 0.5 each.
--
1.
Thot - Teold between 10 degrees and 50 degrees F 2.
Teold' constant or decreasing 3.
Thot constant or decreasing 4.
CET temperature consistent with Thot 5.
Steaming rate affects primary temperature REFERENCE CC EOP-2, p 4,6 & 7.
000056K302 192005K107 192005K106
..(KA's)
ANSWER 4.10 (1.00)
~
1. Line up a spent fuel pool pump taking a suction on alternate RWT [0, 2. Line up a LPSI Pump recirculating spilled RCS fluid from the containment floor through the core and out the leak [0.5].
,
REFERENCE i
CC LOR-320-1-85 000036K303 192008K106 192008K103
..(KA's)
ANSWER 4.11 (2.00)
J Immediately station an operator at the Unit 2 turbine' front standard
[0.25] in direct communication with the control room [0.25] in order j
to manually trip the turbine if necessary [0.5].
Required because the remote and automatic electrical trip functions are lost [1.0].
]
(*****
CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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PROCEDI1RES - NORMAL ABNORMAL EMERGENCY Page 3
'
AND RAI M QGICAL CONTROL
,
REFERENCE
-
CC GSO Standing Instructions, 83-12.
~
-063000K302 045000K412 192003K102
..(KA's)
ANSWER 4.12 (2.50)
a.
1.
One charging pump is unable to maintain pressurizer level with minimum letdown flow.
[0.5]
2.
One charging pump is able.to maintain pressurizer level with minimum letdown flow but leakage is greater than Tech. Specs.
[0.5]
b.
1.
TM/LP pretrip value.
2.
101 inches 3.
537 F
[0.5 ea.]
REFERENCE CC AOP-2A, p 1, 3, 4.
002020SG10 192003K109
..(KA's)
=
(***** END OF CATEGORY 4 *****)
(********** END OF EXAMINATION **********)
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h T T6 clin 5 N T a.
U.
S.
NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
_CALVERT_CLIFFg__________
REACTOR TYPE:
_PWR-gE__________________
DATE ADMINISTERED: _@Zf1@f2Z________________
EXAMINER:
_NgRRI@2_@._@.___________
I
_ _ _ _I[~_ _ 4#__
CANDIDATE:
_
IN@l@pgQN@_Ig_ggNpiggIEi ggy , Usa sieparate paper for the answers.
Write answers on one side
. Staple question sheet on top of the answer sheets.
Points for each qusstion are indicated in parentheses after the question.
The passing grcde requires at least 70% in each category and a final grade of at loest 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY _ _ _ V_ A_ L_ U_ E _ _ T_ O_ T A_ L_ -__S_C_O_R_E___.. _ V_ A_ L_ U_ E_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ C_ A_ T_ E_ G_ O_ R_ Y_ _ _ _ _ _ _ _ _ _ _ ' _25 99-_ _251@@ ________ 5.
THEORY OF NUCLEAR POWER PLANT ___________ OPERATION, FLUIDS, AND THERMODYNAMICS _25:99__ _25 @@ ________ 6.
PLANT SYSTEMS DESIGN, CONTROL, ___________ AND INSTRUMENTATION _25199__ _291@@ ________ 7.
PROCEDURES - NORMAL, ABNORMAL, ___________ EMERGENCY AND RADIOLOGICAL CONTROL _25 @@__ _291@@ ________ 8.
ADMINISTRATIVE PROCEDURES, ___________ CONDITIONS, AND LIMITATIONS 199z99__ ________% Totals ___________ Final Grade All work done on this examination is my own.
I have neither given nor recei ved aid.
___________________________________ Candidate's Signature c_________
- l
) " NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1.
Cheating on the examination means an automatic denial of your application , l 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.
Consecuti vel y number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer 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.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbrevi ations onl y if they are commonly used in facility literature.
13. The point value f or 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 completed.
- _ _ _ _ _ _ _ _ _ _ _ - _. - _ _ _ _)
_ _ _ _ _ _ _ _ -. _ _ _ .
When-you complete your examination, you shall a.
Assemble your examination as f ollows: (1). Exam questions on top.
(2) Exam aids - figur es, 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.
... - _ _ - _ _ _ _ _ _ - -.
_ _ - _ _ _ _ 5.
THEORY OF NUCLEAR POWER PLANT OPERATION _FLUIpS _@Np PAGE
1
IHEBDggyN@DICS . l l QUESTION 5.01 (3.00) In accordance with the Calvert Cliffs Unit 2 Technical Specifications, there are four parameters that are monitored and can be controlled by the reactor operator to ensure that the reactor core safety limit is not GXCeeded.
a.
List THREE of these parameters and briefly explain HOW (i ncrease, decrease, or no change) and WHY a decrease in each of the parameters af f ects the Departure f rom Nucleate Boiling Ration (DNBR).
Assume the other parameters remain constant.
(2.25) b.
At what location in the core (top, bottom, or middle) is the DNBR the smallest? Justify your answer.
(0.75) DUESTION 5.02 (3.00) You have just completed a reactor startup and reactor power is at 1%. For the following situations, indicate WHERE the final power will be in relation to the initial power (higher, lower, no change).
Justify your cnswer.
Assume the core is at mid-life and no operator action.
Consider ecch situation separately.
c.
Steam dump pressure setting is raised by 20 psig.
b.
A steam leak equal to 1% reactor power dev>' ops outside containment.
c.
An 20 ppm boron dilution is made.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE
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THEORY OF NUCLEAR POWER PLANT OPERATION _FLUIpS _9Np PAGE
1
. THERDgQYN901CS QUESTION 5.03 (2.40) Given the below information and using the attached Figure 5.1, predict the critical boron concentration.
Show your work and state all assumptions.
REINITIALIZE at 1200 ppm.
BORON CONCENTRATION SOURCE RANGE COUNTS 1500 ppm 50 cps 1450
1400
1350
1300 100 1250 110 1200 125 1150 170 1100 215 1050 400 1000 1000 QUESTION 5.04 (2.00) During a natural circulation cooldown, the pressure in the steam generator is 835 psig.
What is the MINIMUM pressurizer pressure that would ensure that the RCS is 45 degrees F subcooled? Show your work and state all assumptions.
i QUESTION 5.05 (3.00) ' a.
Explain HOW (more positive or more negative) and WHY Axial Shape Index (ASI) is expected to change as power is increased from 20% to 70% during a normal power increase at EOC.
b.
Explain what TWO steps / methods are taken to maintain ASI within l i mi t s and WHY these actions are effective.
QUESTION 5.06 (1.60) Given the below conditions during a reactor startup, determine the final count rate: Initial count rate = 100 cps Initial reactivity = - 0.0526 Final reactivity = - 0.0204 (***** CATEGORY 05 CONTINUED ON NEXT PAGE
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ l - t 5:__IHggBy_g[_NgC6g@g_fgWgg_E69NJ_QEgg@]]QN _ELg]Qgz_9NQ PAGE
2 IHgBdggyN3d}Cg .
l l l QUESTION 5.07 (3.00) Using the Technical Data Book for Unit 2 (NEOG-11) provided, draw the curve for the xenon reacti vi ty associ ated with the power schedule shown on Figure 5.2.
Assume the reactor is in the middle of this cycle and that the reactor has been shutdown for 30 days.
QUESTION 5.08 (2.50) "or a reactor operat:ng at a constant power and temperature, HOW a.
- (increase, decrease, or no change) will the THERMAL neutron flux change from BOC to EOC? Justify your answer, b.
For a reactor operating at a constant power and temperature, HOW (increase, decrease, or no change) will the FAST neutron flux change from BOC to EOC? Justify your answer.
QUESTION 5.09 (3.00) Rsfer to Figure 5.3 attached: Determine the resultant SYSTEM flow and pressure for the below combinations of centrifugal pumps.
You may write on the figure and attach it to your answer sheets.
a.
Figure 5.3.a: a second pump identical to the first is added in PARALLEL with the first.
b.
Figure 5.3.b: Pump #1 is a booster pump in SERIES with Pump #2.
QUESTION 5.10 (1.50) What are the THREE reasons for loading excess f uel at the beginning of the cycle life (BOC)? l (***** END OF CATEGORY 05
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I -. _ _. _ _
- - . 6:__P(9NJ_gY@lgDS_pEglGN _CgNIBg61_9Np_IN@lBUdgN1911gN PAGE
1 . QUESTION 6.01 (3.00) The reactor has been operating at 25% power for 24 hours, all control cystems are in automatic with the exception of ggg control.
Turbine load is increased to 100% and the n rolling steam prcccure detector for #1 eteam generator sticks at the value.
Explain HOW (higher than, lower than, or same as normal l evel for 100% power) and WHY this will affect #1 SG level.
Assume no operator action.
State all assumptions.
NOTE: a numerical answer is not required.
QUESTION 6.02 (3.00) After a secondary calorimetric and adjustment of the power range instruments, it is discovered that the Auxiliary Feedwater Pumps were operating.
State HOW and WHY the indicated power is more or less conservative than actual reactor power.
THREE reasons required.
QUESTION 6.03 (1.50) a.
Explain what controls are used to maintain temperature while on Shutdown Cooling (SDC).
(0.50) b.
Other than r el i ef val ves, what overpressure protection is provided for the SDC system? Include setpoint.
(1.00) QUESTION 6.04 (3.00) Will the plant trip as a result of the following instrument failures? Assume no operator action.
Justify your answer.
a.
SUR channels A and B fail HIGH during a reactor startup when the reactor is critical at 10E-6% power, b.
SG#11 l evel channel A f ails LOW f ollowed by SG#12 level channel B failing HIGH while at 80% power.
c.
Loop #11 Tc channel A f ails HIGH f ollowed by l oop #12 Th channel B failing HIGH while at 100% power.
d.
The lower UIC detector for Safety channel B f ails HIGH f ollowed by Safety channel D upper detector failing HIGH at 50% power.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE
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2 l . l QUESTION 6.05 (3.00) l l Assuming a normal Mode 1 lineup for Unit 2, answer the f ollowing questions ] ! on instrument air.
Consider each part separately.
l c.
How would a loss of instrument air header pressure due to a rupture l just downstream of IA-144/146 (outlet isolation valves for the after filters) IMMEDIATELY affect the following components / systems? Choose ] ONE of the following for each component / system: (1.60) j A - f ails open/ maximum flow j B - fails closed /no flow C - f ails as is/no change in flow D - no immediate effect/ system operates normally 1.
Mai n feedwater regulating valves 2.
Pressurizer spray valves 3.
Letdown 4.
Atmospheric dump valves 5.
AFW regulating val ves 6.
EDG service water supply valves 7.
Auxiliary spray val ve 8.
Turbine AFW pump speed b.
Li st FOUR automatic actions that should occur to maintOin the pressure in ti,e IA header.
Include setpoints.
(1.40) QUESTION 6.06 (1.50) Assume that on each steam generator one level indicator which feeds the logic circuit for the Auxiliary Feedwater Actuation System (AFAS) were to fail as is while at power.
Subsequently, if l evel in both steam generators w:re to decrease to -170 inches, would an AFAS be generated? Justify your answer.
QUESTION 6.07 (1.50) e.
State TWO design features of the Spent Fuel Pool (SFP) which prevent substantial loss of water inventory.
(0.60) b.
List THREE sources of makeup water to the SFP.
(0.90) (***** CATEGORY 06 CONTINUED ON NEXT PAGE
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t . QUESTION 6.08 . ( 2. 40 ). Briefly describe how leakage is detected for each of the below conditions: a.
Safety injection header check valve leakage (ONE method) (0.40) .b.
Letdown heat exchanger tube leakage (TWO methods) (0.80) c.
' Pressurizer relief valve leakage (THREE methods) (1.20)- . QUESTION 6.09 (1.50) . l
What are the THREE postions of the main control board S/G Low Pressure Trip-Bypass switch and when is each position used? QUESTION-6.10 (3.10) Draw the electrical distribution for the #22 120V Vital AC Bus.
Start with the appropriate service transformer, include the diesel ganerator (s.). Show and label all buses and components; show all. breakers and transformers (numbers'for the breakers and transformers are not required).
Include the normal, alternate.and emergency power supplies to the #22 120V Vital AC Bus.
-QUESTION 6.11 (1.50) - .
- Explain WHY a saf ety injection would occur if the VCT level transmitter (LT-226) were to slowly f ail HIGH while at 50% power?
Assume no operator action.
Setpoints are not required.
(***** END OF CATEGORY 06 *****)
,_m.
_m._-__;u_ _
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7.- t t BBDigLQGIC@L_CQNIBQL . QUESTION -7.01 (3.00) . Answer'the following in accordance with OP-1 "Pl ant ' Startup f rom Col d ' . Shutdown ": a.. What instrument (s) are used to determine RCS temperature for the below conditions? 1.
One RCP is operating ~2.
Shutdown cooling (SDC) is in_ operation 3.- No RCPs running and SDC is secured b.
If the RCS-is -in a cold, solid water condition, what are THREE of the precautions taken to avoid an overpressure excursion? QUESTION 7.02 (1.50) ' Answer TRUE or FALSE, and justify your answer.
A reading of 1 0 0 */. o n the reactor vessel water level monitoring system '(RVLMS) is positive indication that there is NOT a bubble. in the vessel head.
QUESTION 7.03 (1.50) The maximum power increase rates for powers greater than 50% is described' below in'accordance with OP-3 " Normal Power Operation."
Why are the rates different? An increase to a level which has not been previously 3%/ hour sustained for a minimum of 3 hours within the last 60 days.
An increase to a level which has not been previously 20%/ hour sustained for a minimum of 3 hours within the last B days.
. An increase to a level which has been previously 30%/ hour sustained for 3 hours within the last B days.
l, (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) I a_r- -_- ___-_-_ -
_ _. _. - l l ' PROCEDURES - NORMAL _@BNQRM9L _gMgR@gNgY_9Np PAGE
7.
g 50919LQglC@L_CgNIggL ! -
i i QUESTION 7.04 (3.00) Srveral of the Calvert Clif f s AOPs require that the reactor be tripped if ccrtain limitations are reached.
For the below si tuati on s, state the l condition (value) requiring the reactor to be tripped.
c.
In accordance with AOP-1B "CEA Malfunctions" for MISPOSITIONED CEAs.
(1.50) b.
In accordance with ADP-2A " Excessive Reactor Coolant Leakage", in the event of a Steam Generator Tube Leak for PRESSURIZER LEVEL.
(0.75) c.
In accordance with AOP-2A " Excessive Reactor Coolant Leakage", in the event of a Steam Generator Tube Leak for AVERAGE TEMPERATURE.
(0.75) QUESTION 7.05 (3.25) Answer in accordance with AOP-9 " Alternate Safe Shutdow Procedure / Control Room Evacuation."
a.
Because of a fire in the Control Room, you, as the Shift Supervisor, have ordered an evacuation to the Unit 1 and Unit 2 Switchgear rooms on the 45' l evel. State WHERE you would send personnel to MONITOR and CONTROL the following plant parameters.
(2.00) 1.
Unit 1 steam generator l evel s 2.
Unit 1 steam generator pressure 3.
Unit 2 pressurizer level 4.
Unit 2 pressurizer pressure 5.
Unit 2 power level b.
If the Control Room evacuation were coincident with a loss of all offsite power, natural circulation would have to be established.
List the FIVE conditions that verify natural circulation cooling.
(1.25) QUESTION 7.06 (2.00) In accordance with EOP-0 " Post Trip Immediate Actions" what is the Reactor Cool an t Pump trip strategy for Unit 2? List FOUR separate requirements.
Include numerical parameter limits and pump trip requirements.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE
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689196991C86_C9N16g6 . QUESTION 7.07 (2.50) a.
In accordance with EOP-1 " Reactor Trip," under what TWO conditions should operators adopt " manual" operation of automatically controlled systems? (1.00) b.
What are THREE entry conditions for EDP-8 " Functional Recovery Procedure"? (1.50) QUESTION 7.08.
(.75) In accordance with EOP-8 " Functional Recovery Procedure", what THREE different combinations of spray trains and coolers is considered adequate to maintain the containment temperature and pressure within the limits of the Safety Functions Acceptance Criteria? QUESTION 7.09 (3.50) Answer the f ollowing in accordance with CCI-800B " Radiation Safety Manual ": Work needs to be performed inside the containment in a high radiation area.
You are going to accompany the maintenance man.
c.
What is the maximum dose you are allowed to receive without obtaining special permission? Assume you have received no exposure during the past year.
(0.75) b.
Due to unforseen circumstances, you are going to need permission to exceed the above limit.
Permission is required from what 4We i ndi vi duals? (1.00) c.
What TWO individuals, by posi ti on, must approve the Special Work Permit before work may begin? (1.00) d.
At what l evel of exposure would an IMMEDIATE notification of the NRC be required? Include whole body, skin and extremities.
(0.75) ! (***** CATEGORY 07 CONTINUED ON NEXT PAGE
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_ _ _ _ _ _, -
._ _ _.
-_.-. -- - PROCEDURES - NORMAL _ ABNORMAL _ EMERGENCY AND PAGE
) 7.
t t l RADIOLOGICAL CONTROL . j l QUESTION 7.10 (2.00) l
You are the SRO in charge of refueling at the Combination Control Console.
The personnel in the f uel storage building have informed you that a fuel essembly is ready to be transf erred to the containment.
l c.
What interlock would prevent the transfer carriage f rom moving? (0.75) b.
If the Cable Overload alarm were to be received while the carriage were in the transfer tube, HOW and from WHERE could the carriage be retrieved? List TWO methods.
(1.25) QUESTION 7.11 (2.00) Answer the following in accordance with GSO Standing Instruction 83-12 " Loss of 11 DC Bus": a.
Why must an operator be stationed IMMEDIATELY at the Unit 2 turbine? h.
What are the operator's TWO responsibilities? . (***** END OF CATEGORY 07
- )
_____-_r
_ - - - - --- - --_- _ _ __-. _ -_ - _ - _ _ _ . - - -_ - _._ - , . az__89dINigIBSIJVE_P89pEpuSE@2_CgNpillgN@2_8Np_L]d]]@llgNg PAGE
. QUESTION B.01 (2.50) Define Containment Integrity in accordance with the Calvert Cliffs Technical Specifications.
List FIVE requirements.
-QUESTION 8.02-(2.00) Unit 1 is.in Mode 4 (Hot Shutdown).
The CRO informs you that the shutdown mnrgin has been calculated as 3.0% Delta K/K.
In accordance with the Unit 1 Technical Specifications, what actions should you order as the Control Room Supervisor? QUESTION 8.03 (3.00) Unit 1-is in_the process of a plant startup, reactor power is about 2%. You note that the loop temperatures are all 500 F.
In accordance with the Unit 1 Technical Specifications: a.
What actions must be taken? (1.00) b.
What are the FOUR bases for the Minimum Temperature for Criticality? l (2.00) l QUESTION B.04 (2.00) Using Section 3/4 of the Unit 2 Technical Specifications provided, assuming that the plant is at 90% power, and given the below information: BAST #21 120 7 wt% boric acid 95 F " BAST #22
B wt% boric acid 100 F " RWT 450,000 gal 2400 ppm boric acid 90 F a.
Why are the above conditions NOT acceptable? (1.50) l l b.
What actions must be taken? Reference to the paragraph number in the l TS is sufficient.
(0.50) l (***** CATEGORY 08 CONTINUED ON NEXT PAGE
- )
_ _ _ _ - _ _ - _ _ - -
. 9 __B9 DIN 1@lBSIlyE_PBQCEpuBES _CgNpillgNg2_gNp_Lidl]@llgN@ PAGE
1 . QUESTION 8.05 (3.75) Rafer to Section 3/4 of the Unit 2 Technical Specifications.
The plant is operating at 100% power, all control systems are in automatic with the exception of rod control which is in manual sequential.
Except as noted, all equipment is operable.
For each si tuati on bel ow, state if the plant is in an action statement and if so what LCO is violated; if the eituation is not a violation of an LCO, state why not.
Reference to the page(s) and SPECIFIC paragraph number (s) is sufficient for justification,. Consider each part separately, c.
- 22 AFW Pump failed its last surveillance.
b.
- 22 HPSI Pump circuit breaker is racked out for replacement c.
- 21 Charging Pump motor is being replaced d.
- 12 Diesel Generator fuel transfer pump has seized bearings e.
Assume that all four of the above happen consecutively.
QUESTION B.06 (1.50) How can an i ndi vi dual determinegif a locked valve should be open or shut? .f,,,,,, (mko'y o$ $ ve lvt QUESTION 8.07 (1.50) Both Units are at 100% power.
Fifteen minutes before shift turnover, the on-coming Plant Watch Supervisor calls in sick.
The on-shift Shift Supervisor (SS) calls in a replacement who said he will arrive in about two hours.
Because the overtime situation on his crew has been high due to illness on other shifts, the SS decides to allow his crew to go home.
Are the actions of the SS appropriate per the Calvert Cliffs Technical Specifications and Administrative Policies and Procedures? If YES, why; if NO, what actions should have been taken? l l (***** CATEGORY 08 CONTINUED ON NEXT PAGE
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I ! _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . 9:__9901Nig]3911yg_PBgCgpuggg1_C999111gNg1_gNp_ LID 11gIlgNg PAGE
. QUESTION 8.08 (2.25) TRUE or FALSE; if false, state WHY Answer the f ollowing in accordance with CCI-112C " Safety Tagging": a.
Any individual licensed as an RO on BOTH units at Calvert Cliffs is qualified as a Senior Safety Tagger.
b.
When a system is restored to normal lineup, any individual qualified to operate the system is allowed to verify the lineup.
c.
Yellow tags may be used f or personnel protection onl y if the work is on a low energy system.
QUESTION 8.09 (1.50) Answer the f ollowing in accordance with CCI-133H " Fire Protection Plan": a.
WHO, by position, is designated as the Fire Brigade Leader? (0.50) b.
WHO, by posi ti on, is designated as the Operational Technical Advisor? Include alternates.
(0.50) c.
WHO, by posi ti on, are restricted f rom f unctioning as part of the Fire Brigade? (0.50) (***** CATEGORY 08 CONTINUED ON NEXT PAGE
- )
- _ _ _ _ _ _ _ _
- - - - . B __99dINISIBBIlyE_BBQCEDUBESz_CgNQlligNgi_BNp_(JDlIBIlgNS PAGE
. QUESTION B.10 (3.00) You are the Shift Supervisor, it is 2:00 a.m.
Unit 2 has been operating at 100% for the last month when the following occurs: Pressurizer l evel and pressure start decreasing rapidly and cause a reactor trip / safety injection.
Pressurizer level suddenly indicates maximum with all HPSI pumps injecting.
Steam Generator #22 pressure is 1000 psig, and an Outside Operator reports that a SG saf ety valve is jammed open.
Containment pressure is normal.
c.
Using the EAL Criteria attached, classify the above event.
Justify your answer.
State all assumptions.
(2.00) b.
Based on the classification made in part a., will the Technical Support Center and the Operational Support Center be manned? (0.50) c.
TRUE or FALSE As the Shift Supervisor, you can upgrade the event as conditions warrant, but to downgrade the event you must have the concurrence of the NRC.
(0.50) QUESTION 8.11 (2.00) State whether each of the f oll owing events requires a ONE HOUR noti f icati on por 10CFR50.
Consider each separately.
c.
The plant is in a condition NOT covered by operating and emergency procedures.
b.
The loss of the offsite notification system.
c.
A valid automatic initiation of the Reactor Protection System.
d.
A shutdown was commenced because the plant was in vi ol ati on of the Technical Specifications.
(***** END OF CATEGORY 08
- )
(************* END OF EXAMINATION ***************) - _ ~ -
_ _ _ _ _ _ _ _ _ _ _ _ .
, f . . " ' Ecuations Q = Mh p= + tK,7f I + lt 6=mcaT 26 p p SUR = t* + (S p) t .. Q = UAat C (1-K ) = C2 (I-K )
3
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. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ ._ _ _ _ _ _ _ _ _ _. . ERPIP M . l Rev.12 Ch.7
.. EMERGENCY ACTION LEVEL EAL) CRITERIA-52!1- - . IF EAL Criteria are met or escreded THEN 0 to page 17
C ATEGOR Y GENERAL EMERCENCY SITE EMERCENCY 1.
R ArHOACTIVITY Actual or potential offalte ')35E Actual or potential projected DOSE RELEA5Es 71000 mrem whole body /PSJ00 mewm at Protected Area fence 7500 mrem thyroid under actual meteorological whole body /72500 thyroid under conditions.
actual meteorological conditions.
Whole body site boundary DO5E RATE Whole body sate boundary DOSE RATE P 1000 mrom/h under actual meteor-7 230 mesm/h ameer actual meteor-ological condation:. siegical conditions.
2.
F155 ION PRODt?CT Any TWO of the following three, AND LOCA with 51 Tank discharge.
MAR 8tlER OEGR A-ootential of occurance for the - OATION: THIR D.
Total loss of main & auxiliary
- RC5 activity > 300 uC1/cc i-131 feedwater for longer than 10 min.
Done Eouivalent.
- EOP 5 (Loos of Coolant Acc dent),
or EOp 6 6 team Generator Tute CTMT pressure 725 seig.
FUEL R upture) implemented.
RCS CTMT ave. temp 7130*F.
CTMT
- CTMT degradation (anv af the 5elow)
- Ecuo. hatch not closed / sealed - Either aarlock inoperable.
- CTMT pressure >25 pois.
- All penetrations not closed or capable of being closed remotely by automatic signal or manual initiation.
1.
SECURITY: Security threat resultinst in toes of Secas ity threat resultang in imminent ability to achieve and maintain safe loss of ability to achieve and maintain shutdown of either reactor.
safe shutdown of eitner reactor.
- . . 4.
FIRE: Maior fire wheen defeats both safety trains or functions.
- . . 5.
GENER AL 5 AFETY: Plant conditions exist which could Plant conditions exast which could i I result in imininent core degradation.
result an gross plant contamination.
6.
STEAM LINE 9RE AK:, l SCl3 and botn MSIV's fast to close.l
AIRCRAFT /MI55tLE: $EVERE damage to any of the belows-Auxiliary Sidg.
-Containment-intake 5tructure-500kv Sw. Yaro 412 Cond. Str. Tk.
13 kv Sw. Yard -#21 FO Str. Tk.
-R WT t.
WEATHER: Earthquake >0.15g hors or 0.10g vett.
Flood? 6F above mean sea level.
Wind?90 moh. predicted 7150 mph.
j 9.
ELECTRICAL: Loss of all vital AC/DC fortl5 min.
Annunciators are not functioning for 15 man. AND transient affecting RC5 or Waste Processang occurs.
10. OTHER H A* ARDS: Shutdown control not regasned withan EXPf.05 TONS 15 min. af ter 40P-9 (Alternate Safe CASES Shutdown / Control Room Evacuation) LlOUIDS implemented.
Hazardous substance rendering safety related eouipment inoperable in any of the following-Corttrol Room-Cable Spiet.d Rm-Centamment-Diesel Gen Rm 5hutdown Panels Switchgear km l6 r_-_-_-_____ _ _ - - _ _ .. . -, -....
, _7 . . _ _ _ _ _ _ _ _ _ _ _., _.. _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ ERPIP 3.9 . Rec.13 Ch.8 EAERGENCY ACDOPE LEVEL GAL) CRITERIA ' .ggg.
- IF EAL Criteria are met er easseded THEN go to page 17, CATEGORY ALERT WU5UAL EVENT 1.
RADIOACTIVITY Y Any vahd RM5 readings > lasted for Any wahd RM5 readsngolasted ior RELEASE: longer than 15 mlnutes & espected to lesger then I lueur.
eentlense for lesger than I hour.
NAME NUMBER (uC1/s) NAME NURSER (uCi/s) U-l/2 WRNGM I/2-RIC-5415 9J ES U-1/2WRNGM 1/2 RIC-MIS 9.2 E4 NAME NURSER (CPM) NAME NUMBER (CPM) U-l Main Vent 3 8E-3413. -SAE4 411 atele, mit 3 RE-MIS 8.9E3 U 2 Main Vent 3.RE-3415 1.ES U.2 tech. kent 3 4 E-3415 1.0E6 U l Waste Proc 3 RE-3410 LSES U-2 Wasse Proc 3-RE M10 LSE4 U 2 Waste Proc SAE-9410 LSES U.2 Waste Proc 34E-3410 LSE4 Access Centrol 0 RE-Se25 4AES U l ECCS PP Rm 1 RE Sa06 2.0ES ' i 04 BCC5 PP Em 34E-3406 LOES fB sel Mendsg 8 RE M20 1.4ES e Assoas Centrol 94E-3423 EAE4 other plant emnetlens with actual / Lagdd weses desdierge monitor >180 nwom Whole 948 3301)hipialarm trip falls />S00 mrom thyreld (Easieseng es shut hath testation valves, eenesealed areal.
2.
rIansON PRODUCT EUP 3 ELees of Caetant Aleident), or Esta actistlen durang power operat.
. SARRF.R DEGRA.
ECP 6 tateam Generator ' ime eLuotwe), jan impelvifg initiatieg signal j DATION: er AOP 6D yesel Mensana kicadent) residting in ECC5 water being injected '-; t.R lese RC5.
FUEL RC5 RC5 activity >S00 WCl/cc I-ISI DE.
ECC5 Bow verified by LP51/HP51 CTMT indention en 1/2 C08 & 09.
^ Ctmt presswe >4 peig.
Cent temperstwe 6ews)>1SO*F.
Failure of RP5 to bring reacter sub-critical after a salid trip condition.
~ C_ ' 3.
SECURIT Ys Forced entry of ietsuthorised Forced entry of tessuthorised pereennel pereennel ante the vital area.
Ante the protected area.
Sabotage of vitalares :7'--at Sahetage of protected area espalpment in proaress.
In areareas.
4.
FIRE: Batety Related Egulpment Ilre Esfety Relatetf Equipment area peqpsiring efinite assistance.
Are met extinguished within 10 minutes after Gre Sahtina eiferts besin.
5.
EiENERAL 5ATETY Cn'^.; ens enhet which have on-oneer amonge required by Tech. 5pec.
stantially degraded plant safety.
3.0.3 for not snoeting LCO.
Canettens warrant increasing efielte esency eweroness.
L.
BTEAM LINE BREAKS Main 5 team Line break with 3Gu.
Resisena ti anstematic reactor trip.
J.
AIRCRAFT /Mu3II.Es Crash Anside of, _ 4 ares, er Akerart erseh an site, but sortelde ente any persnenent plant structure.
preescond area and not hapacting enr sient servetwes.
L PEATHER: E.arthquehe) 0.555 Mora. er Ibs333 vert.
. Fleed>4F above sneen een level Wind >90 mph preq5ceed CISO saph.
Terviede strildma M%. s.
ELE 4 ssubAL: Lees of SII 123 VDC to eHher emit.
Lene of effeite power and lose of all snelse AC power to eteher sent.
Unplanned lose of aM/most anmencisters for > I hour dwing power a seration.
10. UTrER MAZARD5: ADP-T 5ection IV, W'lant " _: _.. Espleelen er release et hasarcous EXPLO510N5 Fellowina Non Fire InsuovCentrol edetence potentially affecting a CASES Room Evacuation) denpet nonted.
IJQLEDS ' vital area er pereennel safety.
Egassion er renesse of hasardous edetence rendering safety related eenisment ineserable.
IS - _ _ _ _ _ - _ _ _ _ -.
.' - ~ ~ - ' t.q, w , CCI-501D '.,, - . - , , '-
ATTACHMENT (1) . - , l T NUCLEAR FUEL PROCEDURES $ TECHNICAL DATA BOOK (U-2) i NEOG-11
__ ; - , Prepared By Date Reviewed By Date POSRC MTG# Approved By - Date -.., - ORIG W. J.
!2/3/ S. W. Long 12/3/ 76-145 R. M.
12/5/ Lippold
76 Douglass
_ REVI R. L. Bigelow12/21/ W. J. Lippold 12/22/ 78-169 L. B.
12/29/ -
78 Russell 7R REY 2 J. P.
12/20/ J.
A.' 12/20/ 79-197 L. B.
12/10/ . ' Steelman
Mihaleik
Russell
, REV 3 . B.
J. P.
3/31/ J. B. Couch 3/31/ 81-52 . Q37 . Russell erm 3 _ g, p Rgy 4 M. Trigger 1/4/83 J. P.
1/4/83 83-02 L. B.
1/5/ Steelman Russell
O l 5;REV5 Richard Moore $/26/ J. P.
6/26/ 84-90 L. B.
6/27/ "la ' J. F. Williams 84 Steelman
. Russell
s. I - REV6 J. B. Couch 12/18/ Penney File 12/18/ 85-155 L. B.
12/18/ l~ B5
Russell
.; .- [ .'..*? fMy h OI W~
! Y~
- ,.g5
' ' ) - ,
- 1
- ' JNE0RMATION OY Y - , t- , W,g....NOT TO BE USED 1, FOR TEST /OPEMTim - . ,a:
- "
f , , , M 1sa 'e ,
________ I
- _ - _ _ - _ _ - _ _ ' ' NEOG-11, R;v. 7 Pag 2i NEOG.11 - 'IECHNICAL DATA BOOK (U-2)
- _ 1.0 PURPOSE .. The purpose of this operator guide is to collect technical data In a sirgle source for use In plant operatiors. It is Intended that this guide form a Technical Data Book to which appropriate information may be added over"the Ilie of the plant as required.
2.0 REFERENCES . 3.0 INITIAL CONDITIONS 4.0 PRECAUTIONS AND LIMITATIONS The purpose of this guide is to provide the most representative techr.lcal Information ' available at the time. It is intended that all information contained in this guide shall.
be continually updated to refle - .gnificant changes on an as-required basis. Any section of this procedure may be superceded by a Post Startup Test Procedure.
g 5.0 PROCEDURE This procedure consists of the following Table of Contents and associated figures - ,and tables.
. . ,l-m * . - . . O
, . _,, . -. -me== .,....... ....., . - - --
_ _ _ - - _ _ _ ._. _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~ 0; _ . . . l- - NEOG-11, Rcv. 7
- Page 2 , - u -
- TABLE OF CONTENTS
i . , , . !. ,L FUEL ASSEMBLY, CEA AND OTHER CORE COMPONENT LOCATIONS ,. ': 5 F- - Figure Number !. -1 :TR A.
Core Maps i1, - . .. ~ - . "^y s.e ~ ..,, _ . 1.
Fuel Assemblies by Serial Numbers 2-I.A.1 '
2.
CEDM's By CEDS Number 2-I.A.2 - . - 3.. CEA's by Serial Number 2-1.A.3 , 4.
CEA Groups 2-I.A.4 . . 5.
Incore Detector Locations - 2-I.A.5 . (Computer arv' JNCA) ~~~
, 6.
Incore Det&ctor Octant Locations j d' -I.A.6 , ... .. , w j..- Computer) ... '
. ,
- . ' E.
REACTIVITY PARAMETERS A.
Boron
. _- 1.
Soluble Boron Concentration vs. Burnup, HFP,
y 7,7. p w/ Equilibrium Xenon and Samarium, ARO 2-H.A.1 k "i%%,j .f 2.
Inverse Soluble Boron Worth vs. Burnup 2-U.A.2
j W3-3.
Shutdown Boron Concentration vs. Burnup, ARI 2-H.A.3 J3 b J. 71 . 4.
Shutdown Boron Concentration vs. Burnup, ARI 2-H.A.4 ~. except most Reactive CEA Stuck Out s < " - --' < 5.
Shutdown Boron Concentration, more. 2-II.A.5 !* than ONE CEA Stuck Out k,
- 6. ~ Refueling Boron Concentration 2-H.A.6 g..
.; . > [: c' b-.: - - i B.
Control Rods - . pm n 1.
Integral CEA Group Worth, with Overlap 2-U.B.1 i
\\ > a n. 4..,
- O 4'7
C.
Reactivity Coefficient - e' / . '. , ,. . p .M.
, , <,.~c' 1.. Power Defect Curve - 2-H.C.1 v
y.
' ',;e D.
Xenon and Samarium - .
...,..? . . , ^ 1.
Equilibrium Xe Worth vs. Power Level 2-U.D.1 '
, ~h.'Of 2.
Xe Worth After Tripping from Various Power Levels vs. Time After Trip 2-H.D.2 ,
i 3.
Sm Worth 2-U.D.3 - .
.
0 _ _. _ _ . _ _ _ _. _ _.. -. _ _
r------- . . . . ! NEOG-il, Rev. 7 '
. .Pagn 3 -
III. LOG SHEETS {. 1.
Incore Alarm Setpoints, Log Sheet No.1.
(Attachment NEP-4-6) ) 2.
Log of Operable Quadrant CETs (Attachment NEP-4-18)
IV. MISCELLANEOUS FIGURES ' '~' ~ ~ ' ,y , A.
Power Monitoring . - .
- q ' \\..,_
' 1.
Excore LHR Power Monitoring Vs. Peripheral 2-IV.A.1 Axla! $hape Index
2.
Excore LHR Power MonitaJ,ng vp. Peripheral, 2-IV.A.2 Axlal Shapp Lndex e, - 3.
Excete DNBK Power Monitoring vs?[eripheral ..c4-IV.A.3 G' AidEl $hape 16dex, ..g.*
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_ _ _.. __ -__-- .. . RESPONSES TO NRC LICENSE EXAM ' l l SECTION 1 _ 1.04 a Should not require "ASI=L-U/L+U" since the question did not ask for the definition of ASI, only for its response to power distribution changes.
1.05 d Should also accept: Increase (opens) since LOAD set will automatically open valves to maintain set load.
(LOAD SET is an alternate turbine control mode.
Question stated controls in auto.)
1 / 'l i _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _. _. _ _ _.. - - -
- . SECTION 2 2.03 b Shutting of mini flow recirc line isolation valve requires operator action.
' OM 74 sh 1 of 3 - 1-E-76 sh 31 2.05 a 2) D all 3) D supplied 4) D by 7) D Cont. Accumulator 2.06 a Add seismic structure (sys. description) Same as referenced on key.
I 2.06 b sources of make-up to SFP
1.
DI water via purification system Ref. OI-24 Attached Can use RWTs 2.09 b 2. RV'467 Relieves to south containment trench i l 2. RV 468 Relieves to waste processing equipment and area (miscellaneous waste system).
OM-74 sh 2 of 3
i o__________ _
.. ________- _ _ ___ _ _ _ _ _ _ L
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'SECTION 3 3.05 a See OM-74 sheet 2 with attached system description
- 7 page 26.
Pressure transmitters are upstream of RCS. loop check valves and downstream of SIT outlet MOV.
- L 3.08 b Not a specific power level but a Tave valve > 557 F as referenced by sys. Des. #59 sheets 15 & 16 and~setpoint manual sheet 2-56-4 for reactor.
regulating identifying signal and valve for " quick-open" feature.
3.10 a .If the candidate assumed the controller was not j selected to maintain SGFP speed the answer would be no effect as the FRV D/P controller output signals are fed to an auctioneer circuit that selects higher of the two and uses-it to control SGFP. speed.
The controller not selected (manual output signal- .at.20-30%) failing high would cause its. controller output signal to go to. minimum if-it were in automatic operation.
Since selected to manual this change in output signal is never seen by feedpump speed control.
'If assumed controller is in automatic the answer given is correct operation of. system.
Ref. OI-12A, In Lesson Plan 32-1-1 and figure 7C ! '] ! -
'___________________i_.____.__._._____________ _ _ _ _ _
.
. I SECTION 4
4.08 a Recent CCOM change #87-293 to EOP-0 changed boron value to 2300 ppm if more than one CEA fails to fully insert (Sheet attached ) - l )
_ _ _____ ___ ! ( SECTION 5 5.01 b Should also accept power peak location in lieu of ) pressure for reason for minimum DNBR.
j l - _ 5.05 a Should not require "ASI=L-U/L+U" since the question did not ask for the definition of ASI, only for its response to power distribution changes.
! , i i
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- _ - - _ _ _ - _ _ _ l SECTION 6 6.01 Flow transmitters use a derivative signal meaning that they only input the system during a change in 'the measured parameter.
Therefore, an initial - blip in the recorder trace would be seen as the flow error changes.
When the signal goes to a , constant output (no change) the level error becomes dominant returning level to the programmed value.
The error signal is processed in the e' lead / log circuit.
System Description 32 page 18 6.05 a 2) D all 3) D supplied 4) D by 7) D Cont. Accumulator 6.07 a Add seismic structure (system description) Same as referenced on key.
6.07 b Sources of make-up to SFP 1.
DI water via purification system 2.
RWTs 6.11 Failure of LT-226 high will result in VCT level alarm and no automatic make-up.
(Although we operate in manual) Letdown will not divert as diversion occurs via
LC-227A.
' , 1.
However VCT level will decrease due to normal system leakage.
2.
At 0" VCT level LC-227B will actuate to shift charging junction from VCT to RWT 3&4. Adding Boron to core will cause power to decrease and a shrinking of the RCS so pressurizer level and pressure will decrease.
This will occur when charging shifts to RWT (LC-227B) U 5.
Will get TM/LP trip and stabilize at 532 F _ _ _ _ _ _ _ _ _ _ _ _ _ _
- _ _ - - _ - - _ - - - l SECTION 7 7.3 Answer key : Check with facility These requirement are a part of the fuel - preconditioning guidelines issued by Combustion Engineering.
OR The purpose of these guidelines are to prevent fuel damage.
We ask that you accept either answer.
Ref. OP-3 page 7 045000K405 KSA 7.5 a Answer Key: Check at facility 1.
U-1 Steam Generator Levels 1C43 45' Switchgear Room U-l item #19 lE 101 sheet 43C 2.
U-1 Steam Generator Pressure 1C43 45' Switchgear Room U-1 item #20 lE 101 sheet 43C 3.
U-2 Pressurizer Level 2C43 45' Switchgear Room U-2 item #11 1E 101 sheet 43C 4.
U-2 Pressurizer Pressure 2C43 45' Switchgear Room U-2 item #12 1E 101 sheet 43C 5.
U-2 Power Level 2C43 45' Switchgear Room U-2 items #35 & 36 1E 101 sheet 43C Reference 1E-101 Sh. 43C Attached Wiring Diagram Aux. Shutdown Local Control Panel 7.9 a The maximum dose you can receive by CCI-800 without "special" permission is 300 MREM /WK.
b In order to exceed 300 MREM in a given week the individual needs permission from his/her immediate Supervisor (CCI-800B page 10 par. 2).
This question however asked for "TWO" individuals to exceed the limit in "a" above.
In this case, the answer to "a" should be 2 REM /QTR to match the required answer presented in the key.
OR If someone put 300 MREM as an answer to "a" they should have listed only one individual for part "b" in order to be correct.
L_________
_ _ _ _ _ _ ___ ______ _________________ _ . '. - PAGE 1 INCONSISTENCIES WITH NRC SR0 EXAM SECTION 8, 10/27/87: 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS ~ QUESTION 8.01 Define Containment. Integrity in accordance with the Calvert Cliffs Technical. Specifications. List FIVE requirements.
ANSWER 8.01 1.
All penetrations required to be closed are.either: a.
Capable of being closed automatically, or b.
Closed by manual valves, etc.
2.
All equipment hatches are closed and sealed.
3.
Each airlock is operable.
4.
The containment leakage rates are within limits.
5.
The sealing mechanism for each penetration is operable.
REFERENCE CC TS Unit 2, pg 1-2 000069A201 ...(KA'S) COMMENT 8.01 Question is worded such that it implies 5 requirements beyond the i definition.
' REFERENCE: CC TS Unit 2, p 1-2 i l _ _ _ _ - - _ \\
. _ __ ___ _ ___ _-- PAGE 2 , QUESTION 8.02 Unit 1 is in Mode 4 (Hot Shutdown). The CR0 informs you that the shutdown margin has been calculated as 3.0% Delta K/K.
In accordance with the Unit 1 Technical Specifications, what actions should you order as the Control Room Supervisor? ,_ ANSWER 8.02 1.
Verify the shutdown margin.
2.
Initiate immediate boration at >/= 40 gpm of 2300 ppm boric acid solution (orequivalent).
3.
Until the SDM has been restored.
REFERENCE CC TS U-1 para 3.1.1.1 000024K301 ...(KA'S) COMMENT l Answer 1.0 isn't correct, IAW T.S. borate.
REFERENCE CC TS U-1 para 3.1.1.1 ! QUESTION 8.04 Using Section 3/4 of the Unit 2 Technical Specifications provided, assuming that the plant is at 90% power. and given the below infonnation: .
BAST #21 120 " 7 wt% boric acid 95F BAST #22 90 " 8 wt% boric acid 100F RWT 450,000 gal 2400 ppm boric acid 90F a.
Why are the above conditions NOT acceptable? b.
What actions must be taken? Reference to'the paragraph number in the l TS is sufficient? l l _ _ _ _ _ - - - _ _._
_ _ _ _ - _ _ _ _ _ _. ._ PAGE 3 ' d ] ANSWER 8.04 a.
BAST #22 is below minimum temperature.
TS LCO 3.1.2.8 requires either both BASTS or #22 BAST & RWT.
b.
TS Action Statement 3.1.2.8.b _ REFERENCE CC U-2 TS pg 3/4 1-16 024000G008 ...(KA'S) COMMENT 3.1.2.8.b doesn't apply, should be 3.1.2.8.a REFERENCE CC U2 TS p 3/4 1-16 QUESTION 8.05 Refer to Section 3/4 of the Unit 2 Technical Specifications.
The plant is operating at 100% power, all control systems are in automatic with the exception of rod control which is in manual sequential.
Except as noted, all equipment is operable. For each situation below, state if the plant is in an action statement and if so what LC0 is violated; if the situation is not a violation of an LCO, state why not. Reference to the page(s) and SPECIFIC paragraph number (s) is sufficient for justification.
Consider each part separately.
a.
- 22 AFW Pump failed its last surveillance.
b.
- 22 HPSI Pump circuit breaker is racked out for replacement.
c.
- 21 Charging Pump motor is being replaced.
d.
- 12 Diesel Generator fuel transfer pump has seized bearings.
e.
Assume that all four of the above happen consecutively.
j l l
J I - - - -
- -_ PAGE 4 , ANSWER 8.05 a.
IN an action statement.
Pg 3/4 7-5, para 3.7.1.2.a.2 b.
'IN an action statement.
_ Pg 3/4 5-3, para 3.5.2.a c.
NOT in an action statement.
Pg 3/4 1-9, para 3.1.2.2.a Pg 3/4 1-11, para 3.1.2.4- - d.
IN an action statement.
Pg 3/4 8-1, para 3.8.1.1.b.3 e.
IN an action statement Pg 3/4 0-1, para 3.0.5 REFERENCE CC U-2 TS as noted above 194001A102...(KA'S) COMMENT b is incorrect, 22 HPSI not required.
(21,23) REFERENCE CC U2 TS pg 3/4 5-3.
para 3.5.2.a , QUESTION 8.07 Both Units are at 100% power.
Fifteen minutes before shift turnover, the on-coming Plant Watch Supervisor calls in sick. The on-line Shift Supervisor (SS) call in a replacement who said he will arrive in about two hours. Because the overtime situation on his crew has been high due to illness on other shifts, the SS decides to allow his crew to go home.
Are the actions of the SS appropriate per the Calvert Cliffs Technical , l Specifications and Administrative Policies and Procedures? If YES, why? NO, what actions should have been taken? l l . - - - _ _ _ _ _
- - - - - - - - - - - - _ _ - _ _ _ _ _ _ - PAGE 5
, . ANSWER 8.07 No The SS should have had one of his SR0s stay until the relief arrived to take the shift.
REFERENCE - C01-1400, pgs 1-2 l Admin Policy 84-4 CC U-2 TS, pgs 6-4 to 6-5 194001A103 COMMENT ! Answer is wrong / for half credit.
l 1.0 T.S. allows 2 hours.
REFERENCE , CC U-2 TS, pgs 6-4 to 6-5 I QUESTION 8.10 You are the Shift Supervisor, it is 2:00 a.m.
Unit 2 has been operating at 100% for the last month when the following occurs: Pressurizer level and pressure start decreasing rapidly and cause a reactor trip / safety injection.
Pressurizer level suddenly indicates maximum with all HPSI pumps injecting. Steam Generator #22 pressure is 1000 psig, and an Outside Operator reports that a SG sagety valve is jammed open. Containment pressure is normal.
a.
Using the EAL Criteria attached, classify the above event. Justify your answer. State all assumptions, b.
Based on the classification made in part a., will the Technical Support Center and tiie Operational Support Center be manned? c.
TRUE or FALSE As the Shift Supervisor, you can upgrade the event as conditions warrant, but to downgrade the event you must have the concurrence of the NRC.
- _ _ _ _ _. . _ _ _ - .
PAGE 6 -
, ANSWER 8.10 a.
General Emergency Category 5, General Safety: A bubble has'f1ormed in the reactor vessel l which could result in core degradation.
b.
'Yes (bases on a classification in part a of alert or higher).
~ c.
False (CAF) REFERENCE CC E-PLAN, Chpt 3.0, pgs 14 & 15; Chpt 4.1.3, pg 1: Chpt 4.1.4, pg 1 194001A116 . COMMENT a.
alert: ESD, SGIS or general safety.
REFERENCE CC E-PLAN Chpt 3.0, pgs 14 & 15 . i ! I l
....----...u.-.-- -..- - - h
_ - _ _ _ _ - _ _ _. ? ATTACHMENT 4 NRC RESOLUTION TO FACILITY COMMENTS ON WRITTEN EXAMINATIONS Question Resolution 1.04.a Comment accepted. Other correct explanations of how and why ASI changes will be acceptable for full credit. Answer key corrected.
1.05.d Comment accepted. Answer key modified.
2.03.b Comment accepted. Original material provided for the examination preparation was in error. Answer key changed.
2.05.a Comment accepted. Answers 4 and 7 were already D, answers 2 and 3 were changed to d.
Original material provided for the examination-preparation was in error.
2.06 Comment accepted. Answer key modified.
2.09.b Comment accepted. Answer key modified.
3.05.a Comment noted.
Location was not required for full credit. Answer , key corrected.
3.08.b Comment accepted. Answer key modified.
3.10.a Comment'noted. Question will be graded based on candidate's assumptions.
Partial credit will be given if the candidate assumes plant conditions other than controlling in automatic.
4.08.a Comment accepted. Answer key corrected.
5.01.b Comment accepted. Answer key modified to any two of three.
5.05.a Comment accepted. Other correct explanations of how and why ASI changes will be acceptable for full credit. Answer key corrected.
6.01 Comment accepted. Answer key corrected.
6.05.a Comment accepted. Answers 4 and 7 were already D, answers 2 and 3 were changed to d.
Original material provided for the examination preparation was in error.
6.07 Comment accepted. Answer key modified.
6.11 Comment accepted. Answer key corrected.
7.03 Comment accepted. Answer added to key.
-- _.
. _ - _ - _. _.
_ . _ _ _. ____-__________________-_____-_a
, _ _ _ _ _ _ _. .. . ATTACHMENT 4 . Question Resolution 7.05 Comment accepted. Answer added to key.
7.09.a Comment noted.
New procedure issued after material sent to examiners. Answer key modified to accept either a weekly or quarterly limit.
7.09.b Comment noted.
Considered in grading.
8.01 Comment noted.
8.02 Comment accepted. Answer key modified.
8.04 Comment accepted. Answer key changed.
8.05 Comment accepted. Answer key changed.
8.07 Comment not accepted. Technical specifications and procedure require three SRO licensed personnel on shift. The technical specifications two-hour requirement is applicable only to unexpected absence of on-duty shift crew members.
8.10 Comment noted. Will be considered in grading.
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.. . -____ _ - _ - _ _ _ _ _. ., 5.
-THEORY OF NUCLEAR POWER PLANT OPERATIONt_ FLUIDS _AND-PAGE
t IHE8dODyNAdlCS . ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
ANSWER 5.01 (3.00) l a.
any THREE of the;below at a maximum of 2.25 i ' 1.
Thermal power-[0.253 Decreasi ng ' power results in decreased heat flux [0.253 DNBR increases CO.253 2.
Pressurizer pressure [0.253 Decreased pressure decreases the subcooling [0.253 DNBR decreases E0.253 3.
Highest operating loop Tc [0.253 Decreased temperature will increase the subcooling [0.253 DNBR increases E0.25] 4.
Coolant flow [0.253-Decreased flow increases the blanketing of the steam bubbles E0.253 DNBR decreases [0.253 b.
Top.of the core [0.253 (EBecause the pre ure is the lowest :0.253 an the temperature is the
y f4c fw g[ac/(y A G 9.a $~ ca ch.7 highest -:0.253 Arab pfem REFERENCE CC TS-U2, pgs 2-1, B2-3 to'92-3 i CC' Thermodynamics & HTFF, pgra 13-20 to 13-21 193008K105 ...(KA'S) ANSWER 5.02 (3.00) a.
Lower CO.253 Steam pressure increases => steam temperature increases => RCS temperature increases => adds negative reactivity [0.753 b.
Higher [0.253 Steam flow increases => RCS temperature decreases => adds positive reactivity E0.753 c.
Higher [0.253 Boron concentration decreases => adds positive reactivity CO.753 REFERENCE 'CC.LPMRO-302-3-1 " Reactor Kihetics" EOs 7.11, 12, 15 CE " Nuclear Physics, Reactor Theory, & Core Operating Characteristics" pgs 162-166, 178 004000K520 004020K507 039000K508 19200BK117 ...(KA'S) / - -- - -- /- . o
- _ - - - _ _. _ - -. . _ -. ' i
THEORY OF NUCLEAR _PgWgR_PL@NT_gPgRSTJQN _FLUJp32_9N9 PAGE
5
,THERdggyN9digg . ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
, ) ' ANSWER 5.03 (2.40)
S e Figure 5.1 (Key) attached REFERENCE CC LPMRO-302-3-1 " Reactor Kinetics" EO 7.8 192008K106 192008K120 ...(KA'S) ANSWER 5.04 (2.00) 835 psig = B50 psia E0.503
850 psia => 525.24 F [0.503 . 525.24 + 45 2 570.24 F for'Th [0.25] 568 F ='1207.71 psia > 570.24 F = x > [0.253 572 F= 1246.26 psi a > x =.1229.3-psia (1214.3 psig) [0.503 REFERENCE CC LP#RO-301-9-0 TO e.2 193003K117 193003K125 ...(KA'S) J ANSWER.
5.05 (3.00) a.
As the coolant temperature increases towards the top of the core, the moderator density becomes less _EO.403 causing the flux peak to move down in the core E0.20] Since ASI is (1-u)/(1+u) [0.403 it will become more positive as the power is increased. [0.503 L o b e.re do'in h /4 4.*,*a knl/ b.
1.
Reduce power E0.25] g ' g' " " 9# . "' Creates less restrictive limits CO.503 - 2.
Rods E0.253 Change flux shape to change value of ASI [0.503 REFERENCE CC LP#RO-302-4-0 "ASI" EOs 2.1, 2.2, 4.3 & TS 3/4 3.1 ! 192005K110 192005K114 ...(KA'S) , l a i J
- - - _ _ _ - - - _ - - - _ - - - - - _ _ _ _ - - - - - - - - - -.. - - - - - - - - - . _ - - - - . - - _ - - - - _ - - - - -.. - - . . t_ FLUIDS _AND PAGE 18. 5.
- THEDRY OF NUCLEAR POWER PL ANT DPERATION t ISEBdDDyN@dlCS . ANSWERS -- CALVERT CLIFFS-87/10/27-NDRRIS, B.
S.
ANSWER' 15.06 (1.60) Kaff'= 1/(1-p)'[0.503 Keffi 1/(1-{-0.0526)) Keff2 = 1/(1-{-0.0204)) = = 0.95 CO.203 = 0.98 [0.203 'CR2 = CR1({1-Keff1}/(1-Keff2)) [0.503 '100((1 - 0.95}/(1 - 0.98)) = = 250 cps 00.203 REFERENCE CC LP#RD-302-3-1 " Reactor Kinetics" ED 7.4 pg.56 192008K104 ...(KA'S) -ANSWER 5.07 (3.00) Rsfer-to Figure 5.2,(Key). attached REFERENCE CC LP#RD-302-2-1 " Reactivity Factors" ED 12.6,pgs 47-49 192006K109 ...(KA'S) ANSWER-3.08 (2.50) n.- Increase [0.503-Due to.the decrease amount of fuel in the reactor, the thermal flux must increase,to maintain the same reaction rate. CO.753 b.
No change CO.503 Fast neutron ~ flux is proportional to power E0.753 REFERENCE CC LPWRD-302-0-0 " Basic Nuclear Concepts" ED VIII.A to VIII.D, pgs'63-66 192001K108- ...(KA'S) ! E __--- _ __ _ _ - - - _
- hi__IBEgBy_QE_NQCLE@B_EQWEB_EL@N1_QEEB@llgN _E(glgS _@NQ PAGE' 19 t t ISE5d99Yb'ed1GS ' ANSWERS'--'CALVERT CLIFFS' -87/10/27-NDRRIS, B.
S.
i ANSWER 5.09 (3.00)- c.
Fl ow - = 3600'gpm (+/-150) Pressure = 160 psig (+/-10) b.
Flow' = 16,500 gpm (+/-750) ) Pressure = 1025 psig (+/-50) [0.75 each] i REFERENCE CC LP#RD-301-11.1-0 " Fluid Flow" ED 2.1, pgs 17-19 , 191104K109 191104K110 191104K114 ...(KA'S) , ANSWER 5.10 (1.50) 1.
Override temperature effects [0.50 each] 2.
Fuel-depletion 3.. Fission product poison buildup REFERENCE CC LP#RO-302-2-1 ED 3.1, pgs 10-11 192007K104 ...(KA'S) , ' ,
- '"'L--J---.-._.
- - -. - - - - - - - - - - -
__ -.._ , . 6 __PL@gI_@y@IEdg_QEglgN _CQU169L _@NQ_ lng 169dENI@llgN PAGE
t t . ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
I I . i i ANSWER 6.01 (3.00) dee page J70a.- l n- - -.. _ _ 4_--_--m-n_,o-o ________ em me,
,$ g_ 1., n.LL',,~1 al, n, le ~ '~~ ~~~ he correch uf We r.
l ~ '~ i en ~e m 3 l J M.
Il uW retc - Canatant C t i o., p iLaw- ? C yw w. 6 FuOt OI Culte I
,o - vroe+msrenmt+, osgm_57 _ m-m +sm -+m ,_-----..--~+,m --- +,m+ ra+ -w~.ta am---,- s
3m _s-i [ A [- ' UAC e A AA A E1 Em [15 bb ihbb tbEb thb the bb5ta ibEhbSrbc o d ,-+..,1 -+m,_ 1 1 ~., rm vn, y Anr99 & $4) / $ y The flow error will send signal to 4MHu% the feed control valve !?_^?2[a.ju(/ c /dc C2."224c(pi[[ The level error wil signal to c1 :: the valve ofer? Eventually, th errors will cancel out and the control valve will be positio d such that steam flow equals feed flow at a 'i;hr-than norm level for 100'/. [0.75] lo wer REFERENCE CC SD#32, pgs 17-22 035010A102 035010K401 ...(KA'S) ANSWER 6.02 (3.00) The indicated power is less conservative.
[0.75] i.
Actual feedwater temperature would be lower than that used in the cal cul at i on.
[0.75] 2.
Actual feedwater flow would be greater than the value used in the calculation.
[0.75] 3.
Indicated power would be less than the actual power.
[0.75] REFERENCE CC LPMRO-301-10-0 " Plant Cycl e Anal ysi s" ED 10.2.2 015000K504 193005K103 ...(KA*S) - _ - _ _ - _
_, _ _ _ _.,.. - _ _ _ _ _,. _ _. _. _ m e C.?L f 3. od-,..-a ,... - = . .,4,., ..m_ .. ..m,_ ,...,.. _,.,. _ -,., _.., _ ,.,,w,. ,,,, s-4.a.c,, . -.,,._w .. ... _ _ _. _ ,m,- ....,., a-484'*' e-e a % _.-f M _ A fe w__e.,,,,..sy a / r.ya.
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< ' 6:__P(9NI_@y@IEd@_DE@l@Nt_C@UIBQLt_@ND_lN@lBydENIGIlON PAGE 21- .. ANSWERS - CALVERT CLIFFS ~B7/10/27-NORRIS, B.
S.
. ANSWER' 6.03 (1.50) t i a.
By: adjusting'the bypass flow around the SDHx.- [0.503 b.
Containment i sol ati on valves (SI-651/652) shut CO.503 300 psig.EO.503 REFERENCE SD #7, pgs 5, 41-47 005000K401 005000K402 ...(KA*S) ANSWER 6.04 (3.00) a.
-No trip E0.253 The trip will not-occur until power reaches 10E-4*/. power' 0 0. 50 3 b.
No trip [0.253 Only the auctioneered LOW signal is selected, theref ore onl y channel A will trip [0.503 - c -. TRIP [0.25] Due to Delta.T decreasing and causing setpoint to change [0.503 , d.. TRIP E0.253 Both APD channels will trip E0.503 REFERENCE SD.No 59, pgs 10-31 - 012000K403.
012000K603 ...(KA'S) f.
l ) _ _ _ _ _ - - - _ _ - - - - - - - _ - - - - - ) _
_ - _ - _ _ _ _ _ _ _ _. . - .
- 4:__P6@NI_@y@lEDS_ DESIGN _CQNI696t_@ND_lN@l69dENI@IlgN
.PAGE
t , ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
I
J l ANSWER 6.05 (3.00)- . a.
1.
C FA 3-5.
D #8#"- [0.20 each] 2.
4>) NsWI 6.
A Ps 3. 4@WO-7.
D V0WE 4.
D No6 B.
A po b.
Any 4 required at 0.35 each NOTE: tolerance on pressure is +/-'3 psig 1.
Standby IA compressor starts-[0.253 at 90 psig [0.103 2.
Pl~ ant airfto instrument air-cross-connect valve opens [0.253.at 85 psig [0.103 3.
PA header isolation val ve closes [0.253 at 85 psig [0.10] 4.
Unit'1'PA compressorfstarts to supply air to' Unit 2 PA header.
[0.253 at 90 psig.[0.103 5.
IA to containment nonessential loads isolates [0.25] at 75 psig [0.10] REFERENCE
- CC SDs #32, pg 15; H39, pg 21; #41, Fig A-7 to A-9.
J CC AOP-7D, pg 1 078000K302 078000K402 ...(KA'S) < ANSWER 6.06 (1.50) Yes [0.753 l 2/3 redundancy would still. be available to produce the signal [0.75] REFERENCE-SD no 34,.pgs 45-47, 63-64, 86 061000K402 ...(KA'S) L 1.
_ - _ - - - _ - - _ _ _ _ _
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. ' 6 __PL@NI_SISIEDS_pEgl@N _ggNI@gL _@NQ_lNgl@QUENI@Ilgy PAGE
t t . ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
ANSWER 6.07 (1.50) a.
1.
Stainless steel liner plate 2.
No penetrations in pool wall below normal water level 3.
Siphon breakers on penetrations above normal water l evel E2 reauired at 0.30 eachl 4.
J e ir = le. s te e s 4.s e < 87[Mc/Ig we de M A Mg 7_ --d pi-iei "+imn .b.
1.
E " cer!! 3 e"e+- 2.
Refueling water tankt &4n/f j - 3.
Demin water cyctr~ v ie-fa rih'r e ft h .$yJ M 4.
Fire hose [3 required at 0.30 each3 REFERENCE 'CC SD #10, pg 31 033000K401 ...(KA'S) ANSWER 6.08-(2.40) c.
Pressure transmitters downstream of the check valves [0.403 b.
CCW tank level CO.403 CCW Radiation monitor [0.403 i c.
High temp alarm (on discharge piping) [0.403 j QT (l evel / pressure / temperature) [0.403 ' Acoustic-[0.403 REFERENCE CC SDs #3, pg 7; #7, pg 26; #40; #62, pgs 5-6-002000K405 ...(KA'S) l ANSWER 6.09 (1.50) 1.
BLOCK [0.253 - used during plant cooldown [0.253 2-. RESET [0.253 - used to reset a trip signal after the trip condition clears [0.25] 3.
NORMAL CO.253 - normal operation (spring return to normal) [0.253 REFERENCE CC SD #60 pgs 25-29 012000K406 ...(KA'S)
. _ - _. -
- _-_ - _ _ _____- _ .. 6 __PL@NI_SYSIEd@_gESl@N _CQNIBgL _@Ng_lNS169dE31@IlgN PAGE '24 t t ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
l ANSWER 6.10 (3.10) Sne attached answer key for grading.
'- Normal: Battery chargers to the DC bus Alternate: The inverter backup bus Emergency: Battery REFERENCE CC SD #54, pgs 3-5, Figs 54-1,4,8
- 53, Fig 53-6
- 52, Figs 52-1,2
- 51, Fig A-1 062000K102 062000K407 062000K410
...(KA'S) . ANSWER 6.11 (1.50) II. Letdown would divert to the CO.303 2.
Charging pumps would lose uction E0.303 3.
Pressurizer level wou decrease CO.303 14.
Pressurizer press would decrease [0.303 (5.
ESFAS on low pr surizer pressure CO.30] REFERENCE CC SD#6, pgs 22-24
'SDh63, pgs 45-46 004000K101 ...(KA'S) l s ' l', t/ t.
ite.r level wi# deci< ara a'.cl '& s ysr %,, feaky < Ces, e r.]
- 1. f7Y (3") lm lekf <%
ear gia,, pm d'au.-fe),s Y/ f //6 ' [o.[252 a wi fAs V*/ /* -/dt A?W 7 l 3. 'Boem.MMn wr Y/ caus e p,w de ara e,e <. 9 eaa.r2s' fe.,ye rn /ux 4 das. m a.
f* *
- /*
,, ,, - c.. w => p gu,,,,,,,,. /*erI wiW efecieesa [o.tr_.7 s'. /M.uu<A.<< paJrw a wi Y/ ele c.rea.1 e.
\\ =>'T*/7',lAA /~t* l*f Y' [** *b f.no press u,,, u pm; a << [o. z[] s FJenJ m l t ) - i -. - - - _ - _ _ - - - _ - - _ _ _ _ _.
_____- _ __ _ - . 7.
PROCEDURES - NORMAL _ ABNORMAL _ EMERGENCY _AND PAGE
t t l R_ A_ D_I O_ L_ O_ G_ _I C_ A L_ _C_ O NT R_ O_ L , _ _ ___ _
. ' ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, D.
S.
ANSWER 7.01 (3.00) c.
1. - Tc of.THAT l oop [0.503 2.
Read on the SDC temperature recorder (TR0351) CO.503 3.
Average of at least two CETs CO.503 b.
1.
Two charging pumps are tagged out Cany 3 at 0.50 each] 2.
The four RCPs are tagged out 3.
Two HPSI pumps are tagged out 4.
Pzr heaters are tagged out 5.
Establish a computer alarm below the maximum allowable pressure 6.
Two PORVs are operable in MPT ENABLE REFERENCE OP-1, pgs 1-2 002000K410 005000K401 ...(KA'S) ANSWER 7.02 (1.50) FALSE CO.503 p. co) 3 100% in 'che RVLMS i s less than the top of the reactor vessel head .C.50 Cicp12 r r-t e the w te-2 b e t'r the 100% : r cul
- w4 t r 2r. : r : ce*+
r . the perse*izv-I rvrb 'er ~^0 1 chrt? C-C _-143 --- REFERENCE CC 01-1I, pg 3 000074A206 ...(KA'S) , i ANSWER 7.03 (1.50) Ch ;', It f :i'..ty . ,..-_1 _ _,_m _1 ___._m __, - - --... - - - -<---.__________________________________________ f_1 e d .e._.
- r E e /_,o_ m__f u l _ %a j., _f u /T L___4_______________________
e=J4a ______________________________________ _____________________________________________________________________ REFERENCE OP-3, pg 7 rue E45555Ri55-~~~~~~~ IRE 5) .. - _ _ _ _ _ - _ _ _ _ _ - -
__ _ _ _ _._ _ _ ._ _ _ _ _. ~~Z___PBQgEQQ6ES_;_UQBd86t_@BNgBd@6t_EdEB@ENgy_@NQ PAGE
889196901G86_G9 NIB 96 ' ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
ANSWER '7.04 (3.00) a.
Two or more CEAs [0.753 (misaligned from their respective groups) by
15" or more-[0.753 ' b.
Pzr level < 101" [0.753 c.- Tavg < 537 F CO.753- , l REFERENCE CC ADP-1B, pg 5 AOP-2A, pg4 000005A203 000037K307 001050204 ...(KA'S) ANSWER 7.05 (3.25) a.
Chr !- 2t ";;ility K-L__[dl_$qt_Se_[lC 4Cb 1.
[0.40 each3 H -L_.Ksf l e a& __ C.C V3) 2.
F .}( Q _g>__ &_,c h,__ is_a vs> 3.
LW-@_ g fit _C V3) 4.
K-3_fE l_f n._fdsWJ) A D ' b.
1.
Core delta T between 10 F and 50 F and stable CO.253 2.
Tc constant or decreasing E0.253
Tn' constant or decreasing [0.253 4.
Minimum of 30 F subcooling f0.253 5.
Steaming rate af f ects GCS temperature [0.253 I d R$FERENCE CC AOP-9, pgs 29, 35 000017K101 00006BK201 ...(KA'S) l i _ - _ - - - - _ _ _ _ . -
_ ___ - - _ _ - - - , .
P :__PBgCEQUBES_;_NgBd@6t_@@NgBd@6t_EdEB@ENCy_@ND PAGE
Z 609196991C86_C9 NIB 96 ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
. ANSWER 7.06 (2.00) 1.
If RCS pressure. decreases to 1725' psia EO.303 then trip one.RCP in each loop-[0.203 2.
If LOCA indications exist E0.153 and RCS pressure decreases to'1300 psia.CO.153 thenftrip all RCPs [0.20] 3.
If-LRCS temperature and pressure are less than the minimum limits.EO.303 then trip all RCPs~EO.20] 4.
-If CIS has' initiated [0.303 then-trip all RCPs CO.203 REFERENCE CC EDP-0, U-2, pgs.3, 6 CC LP#RO-201-1-0, ED 1.3 000007K301 003000A202 ...(KA'S) ' ANSWER 7.07 (2.50) =a.
Apparent malfunction [0.503 Automatic operation wi)1 not support a Bafety Function [0.503 b.
EDP-3 has been completed but-the event cannot be diagnosed E0.503 An event diagnosis was rnade but multiple safety functicns are.not meeting their acceptance criteria.. [0.503 An event diagnor,is was made but all parameters for a single safety-function are not. meeting ~their acceptance criteria.
[0.503 ' REFERENCE 'CC EOP-1, pg S EDP-8, pg.4 000007A202 000007G012 ...(KA*S) ANSWER 7.08 (.75) 1., two spray trains CO.253 2.
three coolers [0.253 3.
one spray train and two coolers [0.253 REFERENCE ! CC EDP-8, pg138 222000A102 026000K404 ...(KA'S) .. . . _ - _ - - _ -
_ _ _ _ - _ _ - _ _ _ _ _ _ - _ - _ _ - "Z___EBggEggBEg_;_NgBD961_@@NgBD961_EDE69ENgy_9Ng PAGE
889196991986_99NI696 ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, 8.
S.
l i-ANSWER 7.09 (3.50) . 4r 900ter**/yk f0 7d s.
300 mre 0.753 b.
Individual's General Supervisor ! ?. 52 3 C t. p*7 j Or r 21 S up r-';i c e-rdi2 tie-92rrty re_593
o c.
Originating. work supervisor [0.503 Rad-Con operationsCO.503 d.
Whole body > 25 rem [0.253 Skin > 150 rem [0.253 Extremities > 375 rem CO.253 REFERENCE CC CCI-800B, pgs 9, 10, 26, 43 194001K103 194001K105 ...(KA'S) cl. RS P 3-10*i PJ 6 ANSWER 7.10 (2.00) c.
The. containment upender is not in the horizontal positon [0.753 b.
1.
At the.SFP Console [0.253 the overload setpoint could be increased CO.253 and the winch motor speed could be set to high CO.25] l 2.
Manual E0.253 by installing the handwheel on the gear shaft ] extension EO.253 l REFERENCE j CC OI-25E, pg 4 ( CC SD#13, Fi g A-16 034000K402 ...(KA'S) ANSWER 7.11 (2.00) a.
All trip functions from the control room are lost [1.003 b.- The operator must stay in direct contact with the CR CO.503 to manually trip the turbine. [0.503 REFERENCE CC GSO 83-12 E00058K302 ...(KA'S) __
_ _ _ _ - _ _ _
9t__09dINigIggIIVE_PgggEQUBE@t_ggNpillgNgt_@NQ_LidlIQIlgNg PAGE
!
ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
. i ANSWER 8.01 (2.50) 1,. All penetrations required to be closed are either: or [0.253 j a.
Capable of being closed automatically, b.
Closed by manual valves, etc.
[0.253 2.
All equipment hatches are closed and sealed [0.503 j 3.
Each airlock is operable [0.503 ^
The containment leakage rates are within limits [0.503 5.
The sealing mechanism f or each penetration is operable [0.503 REFERENCE CC TS' Unit 2, pg 1-2' 000069A201- ...(KA'S) ANSWER 8.02 (2.00) !_ t'r-i ng the chutdr -
- ;i-
!?.52 f..lih Initiate immediate boration at >/= 40 gpm 44,4iWN [6,7f,,/ of 2300 ppm boric acid' solution 48 -EWH {e,Jd (or equivalent) J, 6.
Until the SDM has been. restored [0.503 REFERENCE CC TS U-1 para 3.1.1.1 000024K301 ...(KA*S) ANSWER 8,03 (3.00) e.
- ..
Restore Tavg to > 515F [0.253 within 15 minutes [0.253 2.
or be in Hot Standby [0.253 within the next 15 minutes [0.253 b.
1.
Ensure MTC is within analyzed range [0.503 2.
Protective instrumentation is within normal operating range [0.503 3.
Pressurizer is capable of being operable with steam bubble [0.503 4.
Reactor vessel is above minimum RT-NDT temperatrue CO.503 REFERENCE CC TS U-1 pgs 3/4 1-7, B3/4 1-2 .CO2000G005 002000G011 ...(KA'S) .__ -_ _- . _ _ _ _ _ _ - _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - ~
_ - _ _ _ _ _ _ -. _ l '8n._BDdlNigIBBIlyE_PBQCEQUBE@t_CQNQlliQNgg_@NQ_LidlIQIlgN@ PAGE. 30 , ANSWERS - CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
l ANSWER ~ ~B.04 (2.00) g . ten \\. i.
l 'a.
BAST #22 is below minimum' temperature [0.753 .TS LCO 3.1.2.8 requires either both BASTS or #22 BAST & RWT CO.753 & b.
ETS. Action Statement 3.1.2.8.y CO.503 REFERENCE CC U-2.TS pg 3/4 1-16 024000G008 ...(KA'S) ANSWER 8.05 (3.75) . ,. a._ .IN'ansaction statement [0.253 Pg 3/4 7-5, para 3.7.1.2.a.2 [0.503 wet ' s. = - w-b.. +M.an action statement. [0.253 Pg 3/4 5-3.
para 3.5.2.a E0.503 .c.
NOT in an action' statement [0.253 Pg-3/4 1-9, para 3.1;2.2.a CO.253 Pg 3/4 1-1!, para 3.1.2.4 E0.25] d.
IN an action statement (0.253 'Pg'3/4 8-1, para 3.8.1.1.b.3 E0.50] e.
IN an action statement [0.253 o Pg 3/4 0-1, para 3.0.5 E0.503-REFERENCE CC U-2 TS as noted above 194001A102 ...(KA'S) ~ ANSWER-8.06 (1.50) Locked open - red tag E0.753 Locked shut - green tag [0.753 REFERENCE CCI-309A, pg 3 '194001K101 ...(KA'S) _ - _ _ _ _ -
-_ _ _ _ _ _ _ I*Bi__0901NISIB@llVE_BBgCEQWBE@t_CgNQlligNS _@NQ_(ldll@llgNE PAGE
t , ANSWERS -- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
. ANSWER 8.07 (1.50) No [0.503 The SS should have had one of his SROs stay CO.503 until the relief arrived to take the shift [0.503 ' REFERENCE CCI-140D, pgs 1-2 Admin Policy 84-4 CC-U-2 TS, pgs 6-4 to 6-5 194001A103 ...(KA'S) ANSWER-8.08 (2.25) a.
False CO.253 Must also meet the requirements of the Safety Tagger Qualification Record [0.503 b.- False E0.253 Only licensed individuals may verify the restoration of locked valves E0.503 c.
False E0.253 Yellow tags may NEVER be used for personnel protection -[0.502 REFERENCE CCI-112C, pgs'3 & 4' 194001K102 ...(KA'S) ANSWER 8.09 (1.50) a.. Fire and Safety Technician (FAST) CO.503 b.
Plant Watch Supervisor [0.253 Alternate: Reactor Operator or above [0.25) , c.
Shift Supervisor [0.253 ' Members of the minimum safe shutdown crew [0.253-REFERENCE CCI-133H, pg 4 194001K116 ...(KA'S) -- .. . ..
_ _ _ _ _ _ _ _
- B __8901NISIBBIlyE_B69gEQyBE@t_ggNQlligN@t_@UQ_LidlI@IlgNg PAGE
e ANSWERS'-- CALVERT CLIFFS-87/10/27-NORRIS, B.
S.
. ANSWER 8.10 (3.00) c.
General Emergency [1.003 Category 5, General Safety: A bubble has formed in the reactor vessel .00.503 which could result in core degradation E0.50] of (basefoW PRODUCT ~ BitRRIsRS l oST (SSTR + s6. S u $7bc Ye(3 % Fis EN b.
on a classification in part a of alert or higher) [0.
c.
False T.^ " ' E0.503 REFERENCE CC E-Plan, Chpt 3.0, pgs 14 & 15; Chpt 4.1.3, pg 1; Chpt 4.1.4, pg 1 194001A116.
...(KA*S) y ;p })SSve<BD Lo% of S v 6 Gootu/6. l, S J TE SMER&5Ncy ANSWER 8.11 (2.00) a.
Yes E0.50 each3 b.
Yes c.
No d.
Yes REFERENCE 10CFR50.72 001000G003 194001A116 ...(KA'S) ( i ! l
ll _: _.
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