ML20236Q652
| ML20236Q652 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/04/1987 |
| From: | Arildsen J, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236Q640 | List: |
| References | |
| 50-413-OL-87-01, 50-413-OL-87-1, NUDOCS 8711200207 | |
| Download: ML20236Q652 (100) | |
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l I I i ENCLOSURE 1 EXAMINATION REPORT 413/0L-87-01 Facility Licensee: Duke Power Company 422. South' Church Street. Charlotte, NC 28242 Facility Name: Catawba Nuclear Station Facility Docket No.: 50-413 and 50-414 Written examinations and operating tests were administered' at Catawba Nuclear l Station near Clover, South Carolina. i Chief Examiner: h e A. Arildsen
-_ Nod V, - /937 Date Signed I
Approved by: M ~ Nos '/ ; 1947 p p in F. Munro, Section Chief Date Signed ] Summary: Examinations on September 14-17, 1987. Sperating examinations were administered to 14 candidates, 14 of whom passed. Written examinations were administered to 14 candidates,13 of whom passed. Based on the results described above, 4 of 5 R0s passed and 9- of 9 SR0s passed. l l l i 8711200207 871116 ! PDR ADOCK 05000413 V PDR ,
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l l REPORT DETAILS i
- 1. Facility Employees Contacted:
*H. B. Barron, Superintendent of Operations- !
*W. H. Barron, Operations Training Superintendent -
*C. T. Kiker, Operations Training Instructor - '
*0. Tower, Shift Operating Engineer
- Attended Exit Meeting ~j
- 2. Examiners:
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R. F. Aiello, RII NRC Examiner
*J. A. Arildsen, RII NRC Examiner R. S. Baldwin, RII NRC Examiner M. E. Ernstes, RII NRC Examiner
-B. C. Haagensen, Sonalysts I
- Chief Examiner 1
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided Mr. Wendell Barror, with a copy of the written examination and answer key for review. The NRC Resolutions to-facility comments are listed below.
j a. SR0 Exam 1 Question NRC Resolution 1.01 Agree with the facility comment. Answer "b" will be accepted as a correct answer for full credit. 1.10 Disagree with the facility comment. Facility reference material provided only supports "a" as the correct answer. No change will be made.to the answer key. 1.18 Agree with the facility comment. No change will be made to the answer. key. 1.25 Agree with the facility comment. Facility's recommended factor will be accepted as one of the required three factors.
o l 2 2.08 Facility comment acknowledged. Note that the referenced learning objective is quite explicit. In the training manual it states: " Identify.the (5) five major sections of the RN system (Intake, Pumps, Main Supply, Hx, Main Discharge)." However, the recommended additional answers are either equivalent answers or subsets of the existing ) answer key and will be graded accordingly. 2.16 Agree with the facility comment. Answer number 2 will be broken up into 3 separate answers. The candidate will be required to list 3 of the possible 6 answers. 3.05 Agree with the facility comment. The answer key will be changed to TRUE. ! 3.07 Agree with the facility comment. Automatic actuation logic and actuation relays will be deleted from the answer key. Additional changes to the answer key will be made to correct the inequality signs depicting the trip setpoints. 3.08 Facility comment acknowledged. Shell warming open bias is a form of " Bias Signal Voltage." The answer key will be changed to accept either answer for full credit. 3.10 Disagree with the facility, comment. For under-compensation, indicated power is greater than actual power. The magnitude of the indicated start-up rate (SUR) is lower than the magnitude of the actual SUR; however, it is conservative due to the fact that the sign is negative. Therefore, with
. respect to reactor safety, over-compensation is " worse" than under-compensation. Over-compensation can cause no electronic problems, but it is still worse than under-compensation due to the fact that the indicated power level is less than the actual . power level. No change will be made to the answer key.
3.12 Agree with the facility comment. The answer key will be changed to accept " Detects faults within slave cycler" as an additional correct answer. 3.14 Agree with the facility comment. However, in lieu of 82% and 78%, the answer will further be changed to accept 82.4% and 78.1% + or - p.5% respectively. 3.15 Agree with the facility comment. The answer key will be changed to read as follows: Interlocks on the LCC incoming breakers prevent closing both breakers together and prevent closing both feeders from the standby transformer at the.same time, o
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1 Facility comment acknowledged. The answer key will be 3.18 modified to give credit - for the following additional answers: a.1 Rods will only drive-in until the derivative circuit times out. ! a.2 Reactor will not trip, but Tave will remain low since l a R0D STOP (C-2) has occurred. 1 ! 3.20 Agree with the facility comment. The applicable portion of the answer key will be changed to read as follows:
- b. All four (4) breakers will trip (RT "A" & "B" and BY "A" & "B").
I c. RT "A" and BY "B" will trip. I 3.22 Agree with the facility comment. Tref will not drop 3 F for this load rejection. However, since the status of the Tavg - Tref differential prior to the instrument failure is not provided, it is possible that a dump actuation could I occur. The answer key will be changed to also accept " Arm l and Dump" as a correct answer. , u i 4.01 Facility comment acknowledged. The existing answer key is i sufficient. Additional correct information regarding l procedure usage will not result in credit deduction 1 l 4.n2 Disagree with the facility comment. The reference material l provided (i.e., the procedure) does not have entry l conditions. However, it is clearly stated in the training I material (0P-CN-H0-EP1, p.16) that at least one 4160 volt essential AC bus is also required to enter EP/1/A/5000/01. No change will be made to the answer key. 4.03 Agree with the facility comment. The answer key will be l l changed to reflect the terminology used in the procedure and will read as follows: Manually trip the reactor Verify reactor tripped Verify turbine tripped , Verify 4160 V essential power busses energized 1 Check if SI is actuated 4.06 Disagree with the facility comment. The facility's training material (OP-CN-H0-CSF, p.14), which provides justification for this procedural step, gives the full basis for tripping the RCP. No change will be made to the answer key. i
s ~ 4 4.10 Agree with the facility comment. This is a typographical error in the examination. The answer key will be changed as the facility recommended. ,
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4.14 Agree with the facility comment. This is a typographical ; error in the examination. The answer key will be changed ' to read "90% of the Admin quarterly limit." ! 4.22 Agree with the facility comment. The answer key will be changed to list 3 answers, and full credit will require 2 of the 3 answers. 4.23 Agree with the facility comment. The facility's recommended answer will be added to the answer key and full 1 credit will require 8 of the 9 answers. 4.24 Agree with the facility comment. The answer key will be changed to list 3 answers, and full credit will require 2 of the 3 answers.
- b. SR0 Exam Question NRC Resolution i
5.01 See 1.01 5.08 See 1.10 5.24 See 1.25 6.03 See 3.05 6.10 See 3.07 6.11 See 3.10 6.13 See 3.12 7.03 Agree with the facility comment. The question was changed from " minimum" to " maximum", and all candidates were informed during the_ examination. 1 i 7.17 See 4.01 8.06 Agree with the facility comment. The answer key will be changed to also accept that the violation would occur within the first two 16 hour days by adding "the A. O. can l not work more than 16 hours in a 24 hour period." l l
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5 8.07 Agree with the facility comment. The facility recommended j answer will be accepted as an additional correct answer. 8.09 Disagree with the facility comment. The question clearly- 1 states "all" power is lost, thus precluding .the situation , posed in the provided Technical Specification interpretation. ] No change will be made to the answer key. ! 8.11.a Agree with the facility comment. The facility recommended answer will be accepted as an additional correct answer. 8.14 Facility comment acknowledged. The facility's recommended answer is equivalent to the answer key. No change will be made to the answer key. l 8.18.a Agree with the facility comment. Part "a of the question will be deleted and thes (,uestion point value adjusted i accordingly. ; 1 1
- 4. Exit Neting l
At the conclusion of the site visit the examiners met with representatives I of the plant staff to discuss the results of the examination. ) There were generic weaknesses noted during the oral examination. ] These areas of below normal performance included:
- 1. Candidate's lack of basic familiarity with personnel dosimetry and radiation exposure limits.
- 2. Candidate's failure to consistently ensure escort's log entries in and out vital areas.
The cooperation given to the examiners and the effort to ensure an i atmosphere in the control room conducive to oral examinations were noted and appreciated. i l The licensee did not identify as proprietary any of the naterial provided i l to or reviewed by the examiners. 1 1 1
(_ t E 5
- U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CATAWBA L
REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 87/09/14 EXAMINER: BALDWIN R CANDIDATE INSTRUCTIONS TO CANDIDATE. i Use separate paper for the answers. Write answers on one side only. l Staple question sheet on top of the answer sheets. Points for each ] question are indicated in parentheses after the question. The passing i 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 l CATEGORY % OF CANDIDATE'S CATEGORY j VALUE TOTAL SCORE VALUE CATEGORY l 76 i 30.00 25.E6A' 5. THEORY OF NUCLEAR POWER PLANT l
OPERATION, FLUIDS,AND l THERMODYNAMICS Jo,oo a,G.9 s l
- 44.754. JLLWA4 Ae 6. PLANT SYSTEMS DESIGN, CONTROL, l AND INSTRUMENTATION Le 28.75 24.74## 7. PROCEDURES - NORMAL., ABNORMAL, EMERGENCY AND RADIOLOGICAL l CONTROL 434C fa
?& 7*3 3 23.79N8 .
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
\ lGS l 4&--7PM % Tota 1s l Final Grade l
All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature 1 i I k_.m. . ____m_. m
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i e' 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. j i
- 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. l
- 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 j examination.
l S. Fill in the date on the cover sheet of the examination (if necessary). l6 . Use only the paper provided for answers. l l7. Print your name in the upper right-hand corner of the first page of each i l section of the answer sheet. ) l l8. Consecutively number each answer sheet, write "End of Category __" as j 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. l9. Number each answer as to category and number, for example, 1.4, 6.3. i 1
- 10. Skip at least three lines between each answer. l l l l11. Separate answer sheets from pad and place finished answer sheets face j
down on your desk or table, j l l12. Use abbreviations only if they are commonly used in facility literature. { l l13. The point value for each question is indicated in parentheses after the i question and can be used as a guide for the depth of answer required. l l14. Show all calculations, methods, or assumptions used to obtain an answer , to mathematical problems whether indicated in the question or not, ) l15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE l GUESTION AND DO NOT LEAVE ANY ANSWER BLANK. ) l l16. If parts of the examination are not clear as to intent, ask questions of j ) the examiner only. ) l17. You must sign the statement on tne cover sheet that indicates that the l l 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. l l l l l l l l L -- - )
1,. 6. e'r f'
- 18. When you complete your examination, you shall:
li - a. Assemble your examination as follows:
)
(i) Exam questions on~ top. l
\
l j (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 f or 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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Ph l i I5 ., ~ THEORY OF NUCLEAR POWERP'LANT OPERATION, Page 4' 1) r: Ft>UIDS AND THERMODYNAMICS , l l
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f00ESTION 5.01 (1.00) i I 1 Which one of the following will NOT change over core life?: j l
- a. The minimum acceptable shutdown margin !)
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- b. The acceptable reactor power imbalance i
- c. Differential boron worth !
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- d. Doppler deficit
- e. Peak Sm worth after SD from full power 1
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l ' QUESTION 5.02 (1.00) i The Catawba Tech Spec LCO on minimum temperature'for criticality ensures I that the reactor will not be'made critical with the reactor coolant system average temperature less than 551 deg F. According to the bases for this LCO, this limitation is required in order to ensure-five (5) conditions. i Which one of the following is NOT one of--those conditions? I
- a. The moderator temperature coefficient is within its analyzed temperature range.
- b. The protective instrumentation is within its normal operating range.
- c. The P-12 interlock is above its setpoint. l d .. The pressurizer is capable of being in an operable status with a steam bubble.
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- e. The pressurizer pressure vessel is above its minimum RTNDT temperature i
l l l 1 l l '\ l l l 8 l l l 1 t (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
.5. THEqRY OF NUCLEAR POWER. PLANT OPERATION. Page 5
'. Fbu1DS AND THERMODYNAMICS
.QOESTION 5.03 (1.00)
The reactor trips from full power,' equilibrium xe'non conditions. . Six hours later'the reactor is brought. critical at 10E-8 amps on the inter-mediate range. If power level is main tained at 10E-8. amps which' one of the following statements concerning rod. motion requirements for the next two hours is correct?
- a. Rods will have to be withdrawn since xenon will closely follow its normal build-in rate following a trip..
- l. b. Rods'will have to be inserted since xenon.will closely follow its normal decay rate following a trip.
l c. Rods will have to be rapidly inserted since.the critical reactor will cause a high rate of burnout.
- d. Rods will have to be rapidly withdrawn since'the critical reactor will cause a higher than normal rate of build-in.
l QUESTION 5.04 (1.00) Under which one of the following conditions is the Moderator Temperature Coefficient Most negative?
- a. BOL, high temperature
- b. BOL, low temperature
- c. EOL, high temperature
- d. EOL, low temperature I
l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
(5;) THEORY OF NUCLEAR POWER PLANT OPERATION, Page 6 {-FLtflDS,AND THERMODYNAMICS
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, i, jl@ NOESTION 5.05 (1.00) Which one of the following statements best describes the relationship between integral and differential' rod worth?
- a. Integral rod worth (at any location) is the slope of the differential rod worth curve at that location.
- b. Integral rod worth (at any location) is'the total. area under the-differential rod worth curve from the end of the rod to that. location.
- c. Integral rod worth (at any location) is the square of-the differential rod worth at that location.
- d. There is no relationship between integral and differential rod worth.
QUESTION 5.06 (1.00) With the plant operating at 85% power and all systems in a normal / auto configuration, the operator borates 100 PCM. Shutdown Margin will: (choose one)
- a. Increase j
- b. Increase until rods move
- c. Decrease i
- d. Decrease until rods move '!
- e. Remain unchanged regardless of rod movement
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I, l i l (***** 3 CATEGORY 5 CONTINUED ON NEXT PAGE *****) ! l l I
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ji,$ THEORY OF NUCLEAR POWER PLANT OPERATION, Page 7 7 FLUIDS.AND THERMODYNAMICS b dlESTION - 5.07 (1.00)
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During a reactor.startup you have just verifiedEa constant.startup rate on.the SR Nuclear Instruments without any rod motion;or. boron dilution. The actual condition of'the core with this indication ist
- a. Prompt Critical
- b. Critical
- c. Supercritical
- d. Subcritical QUESTION 5.08 (1.00)
Which one of the following is the purpose of using soluble boronEto control the excess reactivity of the reactor?
- a. It does not significantly affect the flux shape
- b. It does not significantly affect the. rod worth
- c. It is more cost effective than adding more rods
- d. It increases reactor loading rates OUESTION 5.09 (1.00)
Which set of parameters below best describes centrifugal pump runout conditions?
- a. High discharge pressure,.high flow, high power demand
- b. High discharge pressure,, low flow, low power demand.
- c. Low discharge pressure, high flow, high power demand
- d. Low discharge pressure,- high flow, low power demand
- e. Low' discharge pressure, low flow, high power demand
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) : h __hm -m _ _ _ _ _ - . _ _ _ _ _ _ - _ _ _ _ _ __ m ._________..m - _ _ _ _ _ _ _ _.u_a___._________.--______-___.________.-___-mmm.--____._____a_m_ma___--___
m D. THEORY ~OF NUCLEAR POWER PLANT OPERATION, Page 8 i
'. FL'UIDS AND THERMODYNAMICS j
F t I 1 fOl)ESTION 5.10 (1.00) Which statement is'true if natural circulation is lost:
- a. Core delta T approaches zero, S/G level increases, S/G pressure decreases.
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- b. Core delta T is constant at approximately 80% full power value, l S/G level is constant, S/G pressure decreases. ;
- c. Core delta T exceeds 100% full power value, S/G 1evel increases, l S/G pressure decreases. ;
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- d. Core delta T will exceed 100%. full-power value, S/G level j decreases, S/G pressure increases.
l ' QUESTION 5.11 (1.00) 1 At normal operating temperature, a leak from the PZR water space to the -l containment atmosphere would consist of:
- a. Superheated steam.
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- b. High quality steam-
- c. Low quality steam. -
- d. Saturated water, l
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7 iS. THEORY OF NUCLEAR POWER PLANT OPERATION, Page t'.- FLUIDS.AND THERMODYNAMICS
- Il OCESTION- 5.12 (1.00) 1
-l I 'The Technical' Specifications allow. operations for a 2-hour time period )
l' with a Quadrant Power Tilt Ratio (GPTR) of greater than 1.02. Which I [ of.the following is the reason for allowing operations for these 2 hours? ) i 1
- a. To allow time for corrective action i e. the event of xenon j redistribution following power changes. l J
- b. To allow time'for correction of a dropped or misaligned l control rod.
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- c. To allow time for baron concentration changes to restore )
the OPTR to less than 1.02. 1
- d. To allow time for correction of RCS flow-imbalances.
QUESTION 5.13 (1.00) In the event of a rod ejection accident, which one of the following will be the first reactivity coefficient to insert negative reactivity? I
- a. Moderator temperature coefficient. l l
- b. Pressure coefficient,
- c. Void coefficient,
- d. Doppler coefficient.
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, Page 10 ;
. FL.UIDS.AND THERMODYNAMICS,
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00ESTION 5.14 (1.00) I When synchronizing the generator to the grid, OP/1/B/6300/01 " Turbine Generator Startup" directs the operator to regulate turbine speed to slowly rotate the synchroscope in the fast.(clockwise) direction. Which choice below CORRECTLY gives the two parameters that the synchroscope is' indicating 2 1
- a. Current and voltage differences d
- b. Voltage and frequency differences .1
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- c. Frequency-and phase differences )
- d. Phase and resistance differences
- e. Resistance and current differences l
l OUESTION 5.15 (1.00) Which one of the following statements is CORRECT concerning the paralleling l of electrical systems?
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- a. Although it is desirable to have speed and phase position matched, it is much more important to have voltages matched.
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- b. If voltages are not matched at the time the synchronizing switch is closed, there will be VAR flow from the lower voltage source to the ,
higher one. I
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- c. If the incoming machine is at synchronous speed but out of phase with the running bus when the breaker is closed, heavy currents will flow to either accelerate or retard the incoming machine.
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- d. If the incoming machine is in phase but slightly faster than j synchronous speed when paralleled, the system will tend to. speed up qq,4he system'to synchronous speed.
- e. If the resistances are not matched at the time the synchronizing switch is closed, heavy currents will flow to tend to speed'up the incoming machine to synchronous speed.
q i 4 (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) E__________._______. _ _ _ _ _ _ _ __ _ _ __m_ _ _.m_ .__ . _ _ . _ _ _ _ _ _ _.]
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- s FS[' THEORY OF NUCLEAR POWER PLANT OPERATIONe Page 11
- i. FL.UIDScAND THERMODYNAMICS r-
- 00ESTION 5.16 (2.50)
The plant is operating at 30% power, turbine in AUTO (IMP IN),'when~ loop -
#1 reactor coolant pump trips. Assuming no reactor trip, no operator action and rod control in MANUAL, indicate whether the following .
parameters will be HIGHER, LOWER or the SAME at.the end of the transient. -j compared to'their initial values. j
- 1) #2 S/G steam. pressure (0.5)
- 2) #3 RCS loop flow (0.5)
- 3) Tc in loop 41 (0.5)
- 4) Th in loop #2 (0.5)
- 5) Nuclear Power (0.5) i OUESTION 5.17 (2.00) l If steam goes through a throttling process, indicate whether-the j following parameters will INCREASE, DECREASE, or REMAIN THE SAME. l
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- a. Enthalpy -(0.5) !
- b. Pressure (0.5)
- c. Entropy (0.5)
- d. Temperature (0.5)
OUEST10N 5.18 (1.50) Indicate whether the following will cause the power range instrument to be indicating HIGHER, LOWER or the SAME as actual power, if the instrument has been adjusted to 100% based on a calculated calorimetric,
- a. If the feedwater temperature used in the calorimetric was higher than actual feedwater temperature.
- b. If the reactor coolant pump heat input used in the' calorimetric is omitted.
- c. If the steam flow used in the calorimetric was lower than actual.
4 (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) ______________a
i 5' . ' THEORY OF NUCLEAR POWER PLANT OPERATION, Page 12 t '. FbUIDS.AND' THERMODYNAMICS j l . l 1 ODESTION 5.19 (1.00) l A motor driven centrifugal pump is operating at rated flow. You then start closing down on the discharge valve.'How (INCREASE, DECREASE or: REMAIN THE SAME) will each of the following be affected7
- a. Flow
- b. Discharge Pressure
- c. Available NPSH I
- d. Motor Amps i
OUESTION 5.20 (2.00) I Indicate whether the following will cause the differential rod worth-to INCREASE, DECREASE or have NO EFFECT. 1
- a. An adjacent rod is inserted to the same height
- b. Moderator temperature is INCREASED
- c. Boron concentration is DECREASED
- d. An adjacent burnable poison rod depletes OUESTION 5.21 (2.00) l Indicate how each of the following will affect DNBR (INCREASE, l DECREASE, REMAIN THE SAME).
- a. reduce reactor power l
- b. increase NC temperature
- c. increase NC flow I d. decrease NC pressure l
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) l I
g Si THEORY OF NUCLEAR POWER PLANT OPERATION. Page.13 t i- 'FL.UIDS.AND THERMODYNAMICS p 'OUESTION 5.22 (1.50) Answer EACH of the following as TRUE or FALSE'concerning a 60% load rejection. Assume no operator action..
- 1. Pressurizer pressure will spike high due to an increase in Tave and slowly decrease due to spray.
- 2. Generator load goes to zero immediately after the load rejection.
- 3. Steam Generator levels will initially decrease'due to " shrink" and increase from " swell" as the steam dumps actuate.
QUESTION 5.23 (1.00) During a reactor startup, equal increments of, reactivity are - added and the count rate is allowed to reach equilibrium each time. Choose' the bracketed ([]) words that describe what is observed on..the Source Range recorder and/or SUR meter.
- a. The change in equilibrium count rate is [ larger] [the same]
[ smaller] each time. (0.5)
- b. The time required to reach equilibrium is [ longer] [the same]
[ shorter] each time. (0.5) QUESTION 5.24 (1.50) List three factors which will affect the amount of core uncovering following a small break LOCA during power operations in the program pressure band. (***** END OF CATEGORY 5 *****)
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"Sf/ PL, ANT SYSTEMS DESIGN.-CONTROL. AND INSTRUMENTATION .Page 14 _j i
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QUESTION 6.01 (1.00) Which ONE.(1) of the following. statement's,concerning the CA' flow control ) valves is correct? I
- a. The controllers on Unit 1 are 0% (fullLclosed)-to 100 % (full open).
and on Unit 2 are 100% (full closed) to 0% (full open). i i
- b. The controllers'on both Units are 0% (full closed)'to 100%'(full l open).
- c. The controllers on Unit 1 are 100% (full closed) to-0% (full.open) l and on Unit 2 are 0% (full closed) to 100%'(full open), j
- d. The controllers on both Units are 100% (full closed) to 0% (full -i open). )
l ;, I QUESTION 6.02 (1.00) Indicate whether the following statements, regarding the Gamma Metrics l Flux Shutdown Monitor, are TRUE or FALSE. ) l
- a. Pulling of the fuses DOES NOT make the SDM inoperable. '
- b. The alarm setpoint tracks in the decreasing direction ONLY.
l QUESTION 6.03 (1.50) i Indicate whether the following statements concerning a resistance temperature detector (RTD) are TRUE or FALSE.
- a. An RTD is connected across one leg of a bridge circuit. As !
temperature that is sensed by the RTD changes, a proportional change in the output voltage (current) across the bridge occurs. ;
- b. When an RTD fails open, it will indicate a downscale (low) reading ;
on its meter.
- c. If an RTD is completely submerged, its ability to accurately' monitor i temperature is unaffected by flow rate. ,
1 l l i I (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) i I
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~. - . S .' PLANT' SYSTEMS DESIGN. CONTROL. AND INSTRUMENT &IlGH Page 15 1
QUESTION 6.04 (1.75) Match the following permissive interlocks.in column "A" with their i respective function in column "B". COLUMN _A COLUMN _B j
- a. P-11 1. Turbine Trip Feedwater Isolation < Low Tavg, arms condenser dumps. Allows block of Safety Injection signal after time delay block,
- b. P-9 2. Allows block of Source Range reactor trips.
. c. P-7 3. Unblocks "At Power" reactor trips. 1
- d. P-8 4. Unblocks the 1/4 Loop Loss.
- e. P-6 5. Allows block of the I.R. High Flux low Setpoint j reactor trip.
I l f. P-10 6. Allows reactor trip on turbine trip. I
- g. P-4 7. Blocks automatic and manual Rod Withdrawal. '
- 8. Actuates Turbine Runback.
I
- 9. Allows Manual Block of Low Pressurizer Pressure, !
I Safety Injection, and Low Steam Pressure Safety l l Injection. . l l l l QUESTION 6.05 (1.50) i List the IMMEDIATE trips that will shutdown the emergency diesel generator (EDG) during manual operation. (include tripe requiring operator action) QUESTION 6.06 (1.00) Reactor coolant (NC) system cold leg temperatures are less than or equal i to 285 deg F What " Secondary System" condition must be met prior to l starting a reactor coolant pump? l (4**** CATEGORY 6 CONTINUED ON NEXT PAGE *****) ws a____________--_-___-____
f,.- E1: ANT SYSTEMS DESIGN. CONTROL ANDLINSTRUMENTATION < -Page 16-b~
-QUESTION ~6.07 '(2.00)
. List FOUR reactor. coolant' system, leakage, limitations.
QUESTION 6,08- (1.50) What are the THREE.(3) load transients that-the pressurizer is designed'to satisfactorily operate under during normal operations? , l J QUESTION 3.09 (1.00) '; 1 List the FOUR (4) signals that actuate feed isolation. ) I QUESTION 6.10 (-Br00$ A*% taf' ! 1 List all the signals that will'eause a main steam line isolation and l their applicable trip setpoint. l 1 l QUESTION 1 6.11 (2.00) Compare the effects of an under-compensated-ion' chamber of the Intermediate Range nuclear instrument (IRNI) to a normal IRNI following l a reactor trip. Include in your answer the magnitude of the gamma signal of the inner volume verses the outer volume, under-compensation with i respect to reactor safety, and the rate of power decrease when passing through the intermediate range. (List FOUR) I QUESTION 6.12 (1.50) What THREE actions occur as a result of depressing the " TEST pushbutton on the chutdown monitoring panel (SDM)? l l l i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
,_m-m-_._-_aAu-mmu-- - -- - - '
=A
.6. PLANT SY$TEMS DESIGN. CONTROL. AND INSTRUMENTATION Page 17: i l QUESTION 6.13 (1,00) List two functions of'the slave cycler associated with the Rod Control Logic cabinet, i QUESTION 6.14 (1,50) List the total number of channels, channels to' trip and' minimum channels operable for the following reactor trip _ system instrumentation while'in mode ONE (1).
- a. Pressurizer Pressure - High
- b. Safety' Injection Input from ESF
- c. Power Range' Neutron Flux (P-8)
- d. Reactor Trip Breakers
- e. Automatic Trip and Interlock Logic.
QUESTION 6.15 (1.50) List,in order of preference, the sources of water to the auxiliary + feedwater (CA) system ? 4 QUESTION 6.16 (1.00) i l List the FIVE (5) Barriers that guard against the release of fission fragments to the environment. 1 1 l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) > 4
6 .- PL, ANT SYSTEMS DESIGN. CONTROL.-Allp-INSTRUMENTATION Page 18 QUESTION 6.17 (2.00) State the BASES of the following Reactor Trips.
- a. Power Range-HIGH Neutron Flux
- b. Power Range LOW Neutron Flux
- c. Power Range POSITIVE Rate Trip
- d. Power Range NEGATIVE Rate Trip
-QUESTION 6.18 (1.00)
State the reason (bases)'why TWO'(2) independent KC loops shall be. operable in modes 1,2,3,4 per T.S. 3.7.3. QUESTION 6.19 (1.00) l Why does T.S. (3.5.1.1) require 4 cold leg accumulators to be on line l when the contents of only 3 accumulators need be injected in accordance-I with safety analysis? QUESTION 6.20 (1.50) For the following instrument malfunctions, indicate whether the steam dump control system would ARM ONLY, ARM & DUMP or NOT BE AFFECTED.
- a. Tavg fails high with a 10% step load reduction in progress.
- b. Turbine impulse pressure (Channel I) fails high with a 25% step load reduction in progress.
- c. Turbine impulse pressure (Channel II) fails low with a 5% step-load reduction in progress.
1 l l l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)- I
( , 6 ~ - PL AILT SYSTEMS DESIGN . COtRROL . AND ILSJRUMENTATION~ Page119 1 l 1 QUESTION 6.21. (2.00)- ..f h : Draw the Rod Speed Program curve using the attached 'IK)D SPEED vs - COMBINED ERROR SIG GRAPH". Include the following:
- 1. Maximum RodLSpeed' band
- 2. Minimum Rod Speed band i 1
3, Proportional band
- 4. Lock up
- 5. Dead band
- 6. Insertion
- 7. Withdrawal
- 8. Steps / min VS degrees F I
l I
'I 1
1 1 I { l i (***** END OF CATEGORY 6 *****) ;
Page 20 fe. e (INTENTIONALLY BLANK)
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9
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 21 ANO RADIOLOGICAL CONTROL Q"UESTION 7.01 (1.00)
A caution listed in OP/1/A/6150/08, " Rod Control", Limits and ! Precautions, states
" Individual Control Bank positions on Bank Selector Switch should NOT be used to position rods manually."
What is the reason for this precautionary note? QUESTION 7.02 (2.00) l l In accordance with EP/1/A/5000/1A1, Natural Circulation Cooldown", ! what are the five (5) parameters (conditions) that support or indicate l natu"al circulation flow, during cooldown on natural circulation and l which way should they be trending? (or what should be the parameters' expected value, as applicable) l l l l l QUESTION 7.03 (1.25) i l Complete the following statement according to OP/1/A/6100/01,
" Controlling Procedure for Unit Startup". 1 1 1 i
I a). The baron concentration in the PZR should.be.within I +/- ppm of the NC baron concentration. A'4stwee y s/Mlty b). The . . ! - u ~- heatup rate of the NC shall be limited to 1__ degrees F. heatup rate limit for normal operation. Under abnormal or emergency conditions the NC heatup limit of 2 degrees F. in any one hour period hall not be i exceeded, c). The NC temperature should not exceed degrees F. until at least one NC pump is in service during solid I operation of the NC system. d). During heatup in Mode 3 the delta T between the PZR and the NC loops should be maintained approximately degrees F. to provide adequate subcooling while minimizing PZR and j spray flow delta T. (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) l l l w _-_ - ___ _ --___ _ -_ _ - _
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 22
- AND RADIOLOGICAL CONTROL l
. l GUESTION 7.04 (1.00) l In accordance with OP/1/A/6100/01," Controlling Procedure for Unit Startup", who must be contacted if criticality is attained-above the control rod insertion limits, but below the ECP band.
') i OUESTION 7.05 (1.00) ]
Referring to the attached excerpt from OP/1/A/6100/01,: Enclosure 4.1, 'k g for performing a Unit Startup, what is the significance of the "
" Bullets" preceding the substeps following 2.63?
OUESTION 7.06 (0.75) 1 a). Which two (2) individuals, by title, may authorize the q bypassing of Main Fuel Bridge interlocks? < b). Who by title, must be not2fied when an interlock is placed 1 in bypass or taken out of bypass? ] l OUESTION 7.07 (1.00) -l l While performing steps in EP/1/A/5000/03," Loss of All A/C power", an SI signal may be generated when power is restored. If an SI signal is generated, which one action below is required by the procedure?
- a. Reset the SI to permit the EDG's to energize the emergency buses.
]
- b. Place SI in test to prevent an overpressurization of the RCS. I
- c. No action is necessary es it has no effect.
- d. Reset the SI to permit manual loading of equipment on an A/C l emergency bus.
I (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) ) l l l l 1
'7. ~ PROCEDURES - NORMAL, ABNORMAL , - EMERGENCY- 'PageL23
- ANO RADIOLOGICAL CONTROL ll r
O'UESTION 7.08 (1.50)
.In accordance with EP/1/A/5000/01, " Reactor Trip or Safety Injection",
what are the three (3) automatic Safety. Injection signals?? (Include ( j in your answer setpoin ts ' associated with' these signals, coincidence is J not required)
}
.1 1
OUESTION 7.09 (1.25) What are the 5 Immediate Operator Actions for EP/1/A/5000/01
" Reactor Trip or Safety Injection" ? 1 1
?
l QUESTION 7.10 (1.00) l 1 In accordance with EP/1/A/5000/01," Reactor Tr'ip or Safety I n j ec t i on .". , ' NC pumps should be tripped if "At least one NV or NI pump; indicating flow AND NC subcooling is less than or equal to 0 degrees.F. q What is the basis for this step? OUESTION 7.11 (1.00) In accordance with EP/1/A/5000/1A, " Reactor Trip-Response",
"If one or more Rods NOT fully inserted, THEN, Emergency Borate 150 ppm for each control rod not fully inserted."
What is the basis for this step? QUESTION 7.12 (1.00) In accordance with EP/1/A/5000/1A1, " Natural Circulation Cooldown", when starting an NC pump, preference should be given to start the NC pump 1B first. What is the basis for this action? (*****: CATEGORY 7 CONTINUED ON NEXT PAGE *****) L_--______________________ - _ - _ - _ - . _ - _ - _ - _ - _ .- -
1 1 7. PROCEDURES - NORMAL,' ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
.Page 24 ]
I l O'UESTION 7.13 (1.00) What system-is of primary concern (concerning possible' leakage . location) when performing EP/1/A/5000/1C6 "LOCA Outside Containment"? I QUESTION' 7.34 (1.50) Concerning the EP's answer the following TRUE or FALSE: l i a). If all conditions are GREEN, when using the CSFST's, l monitoring frequency may be reduced to between 10 to 20 'l minutes. l b). If any ORANGE terminus is encountered, the Operator is not required to monitor the remaining trees. c). Once an FRP is entered due to a RED or ORANGE condition and ! no other higher priority path is encountered, that FRP is .) performed to completion, j l l 1 1 QUESTION 7.15 (1.00) l DEFINE: (DO NOT include symptoms or indications) l l l a). " FAULTED" S/G and l l b). "lSOLATED" S/G l \ as is used in the Westinghouse Owner's Group (WOG) ERG background l documents. l l l l i (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) l
7' . PROCEDURES - NORMAL. ABNORMAL. EMERGENCY' Page 25
- .- At4D RADIOLOGICAL CONTROL OllESTION 7.16 '(1.50)
In accordance with EP/1/A/5000/2B2, " Degraded Core Cooling", step 8 (see attached sheets) checks to verify if one NC pump should be stopped, a). Why is'this pump stopped? b). Which pump is stopped? lOUESTION 7.17 (1.00) If a contingency action (Response Not Obtained) in an EP can not be l performed or is not successful, and further' contingency instruction is NOT provided, what action should the operator take? QUESTION 7.18 (1.00) In accordance with Catawba Nuclear Stations Directives, at what , radiation level must an area in the Auxiliary Building be' designated l as a HIGH radiation area? l l l l QUESTION 7.19 (1.00) Which one of the following Critical Safety Functions lists is in descending order of priority? l a). Core Cooling, Heat Sink, Subtriticality, Containment, NC System Integrity, and NC Inventory, b). Core Cooling, Subcriticality, Containment, NC System Integrity, NC Inventory, Heat Sink. c). Subcriticality, Core Cooling, Heat Sink, NC System Integrity, Containment, NC Inventory, d). Subcriticality, Core Cooling, NC System Integrity, Heat Sink, Containment, NC Inventory. (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) c___-__-______-____________-_____.
J7. PROCEDURES -' NORMAL,- ABNbRMAL, EMERGENCY' -Pagel26
*t JAND' RADIOLOGICAL-CONTROL-QUESTION 7.20 (1.00)
- When the Reactor is shutdown, what must be done (with l respect to rod control), bef ore positi ve reactivity can be added by boron di l ut i on ?
QUESTION .7.21 (1.00) i Which one of the following statements is correct concerning'the.statias of the Nuclear Instrumentation Recorder. prior'to withdrawing control bank rods for a reactor startup? a). The highest reading source range channel and the. highest reading intermediate range: channel are selected.and the. NR-45 chart speed is set to "Hi" speed. b). The highest reading source range channel.and the highest I reading intermediate range channel are. selected and tlue NR-45 chart speed is set to "Lo" speed. c). The highest reading source range channel'and'the lowest reading intermediate range channel are selected and the NR-45. chart speed is set to "Hi" speed. d). The highest reading source range' channel and the lowest reading intermediate range channel are' selected and the NR-45 chart speed ~ is set to'"Lo" speed. 1 QUESTION 7.22 (1.00) AP/1/A/5500/04, " Loss of Reactor Coolant Pumps", contains a CAUTION } that states that all NC pumps should be.in operation above a certain. power level. Which one of the following is this power level? I a). 5% l b). 10% c). 15% d). 20% 1 l l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****). i k
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 27 AN9 RADIOLOGICAL CONTROL l
OUESTION 7.23 (2.00) In accordance with OP/1/A/6200/10, " Upper Head Injection",. limits and l precautions: l a). Why is it necessary for the UHI Surge Tank level to be maintained greater than 52%? b). UHI Surge Tank level should not be allowed to decrease below 0%. What is the basis for this? l OUESTION 7.24 (2.00) l In accordance with OP/1/A/6250/02, " Auxiliary Feedwater System" limits and precautions: a). What two actions must be performed if the CA pumps cavitate when suction is aligned to the hotwell?
\f $31-sV-(F b). What two automatic actions may result.1&am operation of the l CA pumps is attempted when supplied from the hotwell when i
the hotwell is under a vacuum? I . l l l b i \ l 4 l l l l l l . l l l l l (***** END OF CATEGORY 7 *****) ], 1 l 1 l \ l l E________________________________________________._____________ __ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _j
B. ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 28 e >
- AND LIMITATIONS l
t
~
l s OUESTION 8.01 (1.50) i State the basis for the Technical Specification limitation on ] Secondary Coolant Specific Activity. (Include.in your answer applicable conditions) OUESTION 8.02 (1.00) State the two (2) bases for the Technical Specification limitations on the Primary containment's internal pressure. (Include in your answer applicable conditions)
)
lOUESTION 8.03 (1.00) In accordance with Technical Specifications, the operability of the Main Steam Line Isolation valves is to ensure that no more than one Steam Generator will blow down in the event of a steam line rupture. ' What are the two (2) reasons for this restriction? ' l l I l l l QUESTION 8.04 (1.00) l l In accordance with Technical Specification LCO 3.9.3, Decay Time, the i reactor shall be subcritical for at least 72 hours. What is the bases for this time limit? l 'OUESTION 8.05 (1.50) State the basis for the Technical Specification limitation on Reactor-Vessel Water level minimum depth while in Mode 6. (Include in your answer applicable conditions) \ l l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) i l l 1 L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. Page 29 ANO LIMITATIONS
.I i
O'UESTION 8.06 (1.25)
'f M r .. R.D. Worker, an Auxiliary Operator, has been working 16 hours -1 (double shifts) per day for the last 5 days. He is going to be off ,
one day (24 hours) and then must return to 8 hour shifts. State j whether any Technical Specification has been or will be exceeded and ( cite the requirements if applicable. i GUESTION 8.07 (1.50) What four (4) conditions must be met in accordance with Technical Specifications, that allow the STA to assume the control room command function and serve as the SRO in the control room? (Including Time limitations) - QUESTION 8.08 (1.00) { If, while operating in Mode 2, the shutdown margin is less than the { , Technical Specification required value, the required action is: (State the action required by Technical Specifications) l 1 - QUESTION 8.09 (1.00) ! l A normal reactor heat up is in progress, (shutdown banks withdrawn) l with the unit in Mode 4, when all power to the DRPI display is lost. State what actions must be taken in accordance with Technical l Specifications. 1 i 'OUESTION 8.10 (1.00) In accordance with CNS Directive 3.1.19, " Action to take in case of Exceeding of Limits": If a shut down or power reduction is required per 3.0.3 a power reduction rate of (a) */. per hour or greater shall be used. If a plant cooldown is required per 3.0.3, a cooldown rate of (b) degrees F per hour or greater shall be used, s (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) l l k__________________.____.. ._ __ _ _ . . . - _ _ _ _ _ . _ _ - -
1 1
)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 30' ;
i .AMD LIMITATIONS-l OUESTION 8.11 (1.50) d. 1 In accordance with CNS Directive 3.1.1, " Safety and Delineation: Tags" ' i answer the followings
]
- a. What two actions must take place before a " Human" Red Tag _]
can leave the work station?
- 6. When can a Freeze Plug Operator transfer responsibility of ,
maintaining the Freeze Plug to another individual? OUESTION 8.12 (1.50) In accordance with Technical Specifications what are the six (6) ) factors considered when performing a Shutdown Margin calculation?
)
i 1 QUESTION 8.13 (1.25) State the two (2) bases for the Technical Specification limits on the Heat Flux Hot Channel Factor, Coolant flow Rate and Nuclear Enthalpy rise Hot Channel Factor. (Include in your answer applicable conditions) 1 1 1 QUESTION 8.14 (1.00) l In accordance with CNS Directive 3.1.2, " Access to Containment", what i two (2) conditions require the " Buddy System" to be used? l j 1 l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) l l 1
84 ADMINISTRATIVE PROCEDURES 0 CONDITIONS, Page 31 3
- AN9 LIMITATIONS I
i
~
l l 4 OUESTION 8.15 (1.00) l l Which one of the following CORRECTLY describes the required actions to retrieve a RED Tag Stub, if the Work Supervisor responsible for the job is NOT on site?
- a. Another Work Supervisor in that group may authorize retrieval, as long as he has phone approval from the. j responsible Work Supervisor. I I
- b. The Group Superintendent is the only individual who may sign )
authorizing removal, and he must inform the Work Supervisor I responsible for the work when he returns to the site. l
- c. The Group Superintendent must approve tag retreival, but he l may authorize this based on verbal approval and having.
another individual sign his name and initial authorizing tag ; removal.
- d. The Shift Supervisor is the only individual who can
, authorize tag retrieval in this situation. 1 OUESTION 8.16 (1.00) Which one of the following is NOT a valid method of performing an " independent verification?
- a. Two individuals independently observing the breaker for a component that is in the correct position.
- b. One individual observing that a valve is in the correct position LOCALLY, and another individual using a REMOTE indicator to verify valve position.
- c. Two individuals verifying valve position from a REMOTE r indicator only,
- d. One individual actually positioning the breaker, then the SAME individual verifying from a remote indicator that the breaker is in the correct position.
l l l l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) l L----_---__-------.-.____------------ - - - - - - _ -- - - - - - -- --_ _ _j
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, .Page 32 E. '-
l AND LIMITATIONS l ' 1 1 .L i dUESTION 8.17 (1.00)- In accordance with Operations Management 1-5-(Independent l Verification), at what radiation limit can independent verification'be waived and who may app' rove this waiver-(by ti t'l e ) ?
'l 1
1 i 1 'OUESTION 8.18 (1.50) ] 1 1 Answer the following regarding the use.of procedures: 4 1 l a. In addition to just prior to.useage, how oftenLis the.
- j accuracy of a working. copy be verified by comparision with' 'l the controlled copy of a procedure? ,
R
- b. What two people (by title) can reject a procedure change?
- c. If, for a Periodic Test,.the acceptance criteria-for.the test can not be met, where should this be documented? l i
l l l 1 QUESTION 8.19 (1.25) The plant was operating at 100% reactor power when a: cont'inuous rod withdrawal accident occur <ed. Reactor power. exceeded the trip setpoint and a trip did NOT occur. Upon, review of;the accident it was determined that Tave exceeded "625 deg. F. at 100 % power.and 2235 psig. Using given Technical Specificationcinformation: l
- a. State any limitation that were exceeded,
- b. What four administrative actions must be done in accordance with CNS Technical Specification (All NRC associated or l related actions are required for full credit) l 1
l l l l l l l t i (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
- - - - - - - _ _ _ - - - _ - _ _ - - - - - _ _ - - - - - - - - ---a- - - - - - - - - - - -----_a
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. Page 33
- AND LIMITATIONS e
EUESTION 8.20 (1.00) l Which one of the following has an associated Technical Specification Limiting Condition for Operation under 3.11 Radioactive Effluents? '
- a. Contaminated Oil Incineration j l
- b. Waste Gas Holdup Tank Oxygen concentration I
- c. Auxiliary Building Exhaust System gaseous effluent I
- d. Gaseous effluent air dose due to particulate QUESTION 8.21 (1.00)
In reference to CNS Directive 2.12.2," Fire Brigade Organization, Training, and Responsibilities", fill in the blanks.
- a. The (1) has final responsibility and authority in handling of a fire emergency,
- b. The (2) serves as the Chief of the Station's Fire Brigade.
} c. The (3) or (4) serves as the Fire Brigade Captain. (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) c__--_-___ - - . _ - - _ - _
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 34 AND LIMITATIONS O
.~
O'UEST I ON 8.22 (1.00) In reference to CNS Directive 3.8.5," Exposure Extensions and/or Exposure Limit Reduc tions", fill in the' blanks. i
- a. The (1) shall be ultimately responsible for approving the Req ues t for Exposure Extension" form if an individual is expected to exceed planned maximum exposure of ;
greater than (2) rem in a year. But not to exceed j the maximum yearly limit. i
- b. For planned Emergency Exposure, the whole body exposure limit may be increased to 5 rem, if it is necessary to remedy a situation immediately hazardous to life and property. Doses up to 25 rem should be authorized by the (1) or in his absence, by the (2) . j 1
QUESTION 8.23 (1.00) l In accordance with CNS Directive 3.8.8 " Radiological Work Practices", what are the requirements if an Extra High Radiation Area can not be i controlled by lock or guard? OUESTION 8.24 (1.50) ] For each of the following, state whether or not the condition (s) below l Will place CNS directly into an LCO. (State YES or NO) a). Loss of ONE (1) seismic monitoring instrument. b). Loss of ONE (1) meteorological monitoring instrument. c). Loss of ONE (1) PRNI channel a > 1 */. power. l l l (***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********)
]
5; ' THEORY OF NUCLEAR' POWER PLANT OPERATION.
) .
Page'35'
, , ' ' FLUIDS.AND THERMODYNAMICS ( I y
(,
' ANSWER 5.01 (1.00) gf
-( a ) .c (b)
REFERENCE 'ff /b v[ l
/. i 1
1 OC, OP 1103/15, enc 1 0.0, 5.1, 2.0. 1 OC,TS 3.1-23,.3.5-10. ' l OP-CN-RB-HO p. 7 >
.{
LPS,0 OP-CN-RB-HO #3 3.3/3.6 001010K535 ..(KA's) l i ANSWER 5.02 .(1.00) ! (e) 1 ll REFERENCE CNS T/S p. B 3/4.1-2 'i LPSO OP-CN-HO-IRX #13 > 1 3.3/3.7 193010K104 ..'(KA's) i ANSWER 5.03 (1.00) (a) y REFERENCE i HBR RXTH-HO-1 Session 39 LO#2 OP-CN-HO-RP p.12 LPSD OP-CN-HO-RP #8 ! 3.4/3.4 192006K107 ..(KA*s) l ANSWER 5.04 (1.00) i (c) i l (***** CATEGORY 5 CONTINUED'ON NEXT PAGE *****) l h____-_____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -__-__.-__m -
- 5. " THEORY OF' NUCLEAR POWER PLANT OPERATION, :Page 36 ..
.',a FLUI DS . AND THERMODYNAMICS ,c I
REFERENCE HBR'RXTH-HO-1, Session 26,.p 2, LO1-OP-CN-HO-RCO p'.11-LPSO OP-CN-RT-RCO #4A
,i 3.1/3.1 192004K106 ..(KA's)
L ! ANSWER 5.05 (1.00) ' I (b) , l i REFERENCE, Westinghouse Reactor Physics,-p. I-5.40' HBR, Reactor Theory,. Session 36, p. 2, LO2 j OP-CN-HO-RCO p.20 LPSO OP-CN-HO-RP #12 2.8/3.1 192005K105 ..(KA's) ANSWER 5.06 (1.00) i i 1 (a) q. REFERENCE HBR RXTH, session .50, 't
- p. 2, LO #2 OP-CN-RB-HO pp. 6,7 LPSD OP-CN-RB-HO #4B '
3.8/3.9 192002K114 ..(KA's) < ANSWER 5.07 (1.00) (c) ' I (***** CATEGORY 5 CONTINUED ON NEXT PAGE'*****) t . u _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .. _ ____..J
n-
.SS THEORY OF NUCLEAR POWER' PLANT OPERATIONS .Page.37 . ' , . ' FLUIDS.AND THERMODYNAMICS'
- REFERENCE
-Nuclear Energy Training, Reactor Operations, NUS Corp.;
-paragraph 12.4
'OP-CN-HO-RK p;12 LPSO'OP-CN-HO-RK #3e 2.8/3.1 001000K554 ...(KA's)'
ANSWER. 5.08 (1.00) (a)- REFERENCE HBR RXTH-HO-1 Session 33, p.2 LO2 OP-CN-HO-RP p.6 LPSO OP-CN-HO-RP #3 ' 3.0/3.2 192007K105 -
..(KA's)
! ANSWER 5.09 (1.00) (c) l l REFERENCE OP-CN-THF-FF p.10 1 LPSO OP-CN-THF-FF #9 2.5/2.7 191004K112 ..(KA's) 1 l ANSWER 5.10 (1.00) (c) i ' ! REFERENCE OP-CN-HO-HT p.7,8 LPSD OP-CN-THF-FF #4b 4.2/4.2 193008K122' ..(KA's) (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****) ( j E -_ _ _--_- - - - - - - - - - - - - - - - - - - - - - _ - - - J
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, Page 38
.', FLUIDS AND THERMODYNAMICS .~
ANSWER 5.11 (1.00) (c) REFERENCE Mollier diagram or steam tables LPSO OP-GA-THF-STM- #2e ; 3.1/3.4 193003K126 ..(KA's) ! ANSWER 5.12 (1.00) (b) REFERENCE CNS, TS, p. B 3/4 2-5 LPSO OP-CN-CTH-PD #5b - 3.6/4.2 001050A202 ..(KA's) l ANSWER 5.13 (1.00) (d) ! REFERENCE CN-OP-HO-RCO p.12 LPSD OP-CN-RT-RCO #3 3.4/3.8 ..(KA s) (ANSWER 5.14 (1.00) l l (C) (***** CATEGORY 5 CCNTINUED ON NEXT PAGE *****)
f54 THEORY OF NUCLEAR POWER PLANT OPERATION. Page 39, t
.? FLUIDS.AND THERMODYNAMICS REFERENCE l
OP-CN-DG-DG3 p.15 LPSO OP-CN-DG-DG3 #8 2.8/2.9 062000A403 ..(KA's) > a i l ANSWER 5.15 (1.00) i i , l (c) , ) REFERENCE OP-CN-DG-DG3 p.12 LPSO OP-CN-DG-DG3 #8 2.8/3.2 062000A215 ..(KA's). l 1 t j ANSWER 5.16 (2.50) l -
- 1) Lower (Higher Stm Flow >> P stm decreases)~
- 2) Higher (Less resistance to flow >> Other RCPs. speed up)~
- 3) Lower (Less total flow across core >>. delta T increases, Tc goes down with rods in manual)
- 4) Higher (as above, delta T increases, Th. increases)
- 5) Same (Primary power = secondary load) ~
REFERENCE General Physics HT&FF Part B, Chapter 1, pp.129-162- 'l OP-CN-HO-FF p. 15 LPSO OP-CN-TA-AT #8a ' 3.4/3.7 002000A105 ..(KA's) l ANSWER 5.17 (2.00) i
- a. RTS (0.5 each)
- b. DECREASE
- c. INCREASE
- d. DECREASE I
i (***** CATEGORY 5 CONTINUZD ON NEXT PAGE *****)- i i I L
L5. THEORY ~OF' NUCLEAR. POWER' PLANT OPERATION, 'PageL40l
- t. FLUIDS.AND THERMODYNAMICS ,
REFERENCE-
'OP-CN-HO-THF pp.18,19. 1 LPSO-OP-CN-THF-STM #6e l
2.6/3.0 010000K502' ..(KA's)- i
,; i ANSWER 5.18 ( l'. 50 )
a.. LOWER
. (0.5 each)'
- b. HIGHER ,
- c. LOWER lREFERENCE k
OP-CN-HO-SLC pp. 15-18 LPSO OP-CN-THF-SLC #6,7 3.5/3.8 015000A101 ..(KA's) 1 1 l ANSWER 5.19 (1.00)
- a. DECREASE (0.25 each)
- b. INCREASE 1
- c. INCREASE
- d. DECREASE REFERENCE
)
OP-CN-HO-FF p.3 l LPSO OP-CN-THF-FF #6 y t 2.7/2.9 191001K103 ..(KA's) ! ANSWER 5.20 (2.00) > a) Decrease (0.5 ea) b) Increase c) Increase d) Increase-l $ L i
'l 1
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)- 1 ; 1 l H l
-.____________________z_____________ _ _ _ _ _ _ _
j
5 5 .- - THEORY OF' NUCLEAR POWER' PLANT OPERATION. 1 s lPage?41;
.?,*'
,. . FLUI DS . AND : THERMODYNAM ICS ' <
iR$FERENCE , L SON /WBN License Requal. Training,." Core Poisons" CN-OP-HO-RCO'p. 191 LPRO OP-CN-RT-RCO #4 001000K502 (2.9/3.4) l 3.5/3.7' 001000K509 ..(KA's) I . ANSWER, 5.21- (2.00) ;
.. ..t .o
- a. INCREASE *
- b. DECREASE ]i c.' INCREASE
- d. DECREASE A 1
l REFERENCE l I l OP-CN-HO-HT p.11 LPSD OP-CN-THT-HT #6c , 3.4/3.6 193008K105 ..(kA's) L i l-l ANSWER 5.22 (1.50) l 1. FALSE (0.5 each) I :2. TRUE
- 3. TRUE l REFERENCE 'I OP-CN-HO-NT pp 7,8. s LPSO OP-CN-HO-NT #6a 010000A106 3.1/3.2 045000A304 3.4/3.6 i
4.2/4.2 035010A202 ..(KA's) I ANSWER 5.23 (1.00)
- a. larger (0.5) l' b. longer (0.5)
LI
'I
(***** CATEGORY 3 CONTINUED ON NEXT PAGE * **** ) f o I i i
- 5. THEORY-OF NUCLEAR POWER PLANT OPERATION, Pega 42
-( FLWIDS.AND THERMODYNAMICS ,
REFERENCE \ l-l OP-CN-HO-SCM p.18 - LPSD OP-CN-HO-SCM #9 d 3.9/4.0 192008K103 ..(KA's) l ANSWER 5.24 ! (1.50) )
-br.eak size (and geometry) (0.5 each) !
-location o,f break -
(3 , f 4 r e gu l<r./ )
-status of ECCS components ,
- HC f"~ps re ~a l, ef ecall.,3 or a,el
! REFERENCE ' OP-CN-HO-AT p.1 LPSO OP-CN-TA-AT #1 3.8/4.3 00009EA206 ..(KA's) , i ! l l
)
)
(***** END OF CATEGORY 5 *****)
S_, PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION Page 43
. O .~
ANSWER 6.01 (1.00) a REFERENCE CNS LP, (COND & FEED SYS), OP-CN-HO-CA, p 14, LPS0 LO 7. 059000K603 ..(KA's) ANSWER 6.02 (1.00)
- a. FALSE b.'TRUE (0.5 EA)
REFERENCE CNS LP, (I&C), OP-CN-HO-ENC, p 5-9, LPSO LO 2,3. 015020K402 .(KA's) ANSWER 6.03 (1.50)
- a. TRUE
- b. FALSE c.-rALCE 7Aud (0.5 EA)
REFERENCE CNS LP, (SERV SYS), OP-CN-HO-IG, p 5, LPSO LO 3. 002000K606 ..(KA's) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) l
. . x. .
'M J Ef : PLANT 2 SYSTEMS' DESIGN.~ CONTROL. AND INSTRUMENTATION <Page 44-t-.._. _ .j .
1 1 1 i LANSWER. 6.04- -(1.75) ].)
- a. 9 d. 4 g. 1
-b. 6 e. 2
- c. 3 f.5-
.q (0.25 EA) 1
- 1 REFERENCE I
CNS LP, (I&C), OP-CN-RO-IPX, p115,16,~LPSO LO 5.
.j
. 012000K610 ..'(KA's) -l l '
'i ANSWER- 6,05 -(1.50) 1
- 1. Engine overspeed .
- 2. Generator differential
- 3. Maintenance mode initiats
- 4. STOP/RUN button placed in stop position
- 5. Governor load limit set to zero
- 6. 51V voltage controlled over current (trips'D/G and, breaker)
(0.25 EA) REFERENCE CNS LP, (GENERATOR), OP-CN-HO-DG3, p 8, LPS0 LO 5,6. 064000K402 ..(KA's) 1 i j ANSWER 6.06 (1.00) i When the secondary water temperature of each S/G is lessLthan 50.deg F ; above each of the NC system cold leg _ temperatures. (1.0) -i i l l 5 i l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) -l i'
'Ef PLANT SYSTEMS DESIGN'. CONTROL. AND INSTRUMENTATION- . Page L 45L ' REFERENCE CNS LP, (PRI SYS), OP-CN-HO-NCP,.p 35',,LPS0 LO 6,7.
003000K614 ..(KA's) ANSWER 6.07 (2.00) (4 of 6 requi' red) a .- No pressure boundary leakage shall exist. i
. Maximum'1'GPM unidentified leakage.
b.
- c. 500 GPD total reactor-to-secondary. leakage through any one steam generator (S/G) and 1GPM total: reactor-to-secondary leakage through ?
all S/G.
- d. 40 GPM CONTROLLED Leakage'at a NC.. system. pressure of- 2235'+/- 20
~
psig.
- e. 10 GPM identified leakage in'the NCS.
- f. -1 GPM leakage at an NCS pressure of. 2235 +/- 20. psig from any;NCS pressure isolation valve specified in T.S.-table 3.4-1.
(0.5 each) REFERENCE CNS LP, (PRI SYS), OP-CN-HO-NCP, p 35, LPSO LO 8. 002000K405 ..(KA's) ANSWER; 6.08 (1.50)
.1) Loading or unloading 5% of full power per minute.
- 2) Step. load changes of +/- 10% of full' power with automatic reactor control.
'3) Step load reduction'of 95% of full power with automatic. reactor-control and steam dump.
(0.5 EA) (***** CATEGORY' 6 CONTINUED ON NEXT PAGE *****)
Ef PLANT SYSTEMS DESIGN, CONTROL. AND INSTRt1 MENTATION Page . '46 < REFERENCE CNS LP, (PRI SYS), OP-CN-IPE-HO, p 28, LPSO LO 8.a. 010000K602 ..(KA's) l l ANSWER 6.09 (1.00)
- 1. Safety injectio~n l 2. P-14 (Hi Hi S/G level)
- 3. Low Tave coincident with reactor trip
- 4. Manual (0.25 EA)
REFERENCE l CNS LP, (ECCS), OP-CN-HO-ISE, p 17, ISS LO 4.d. 059000A306 ..(KA's) l ANSWER 6.10 =(2.00) (1.75 ) SIGNAL TRIP _SETPOINT i l
- a. Manual Initiation (0.25) N.A. )
^
- b. ..u toma tic ^ctuation Lcgic and
" t. .
Actustier D elaye (9. 2 5 )-
- c. Containment Pressure-High-High (0.25) > 3 psig (0.25)
- d. Steam Line Pressure - Low (0.25) < or = to 725 psig (0.25)
I e. Steam Line Pressure y or : to 100 psi (0.25) l Negative Rate - High (0.25) l REFERENCE CNS LP, T.S. TABLE 3.3-4 ITEM 4, (STEAM SYS), OP-CN-STM-SM, LPS0 LO 8. l 039000K405 ..(KA's) l l l l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) I __-_-_-__-_A
E, PLANT SYSTEMS' DESIGN. CONTROL. AND INSTRUMENTATION. PaE@ 47
.~
ANSWER- 6.11 (2.00) ,, L
, ,j -
(4 of 5 required) l
- 1. The IRNI reading displays too high a reading. j l
- 2. P-6 may'neyer deenergize. ,l 3..The magnitude of the gamma signal from the outer volumefis. 3 greater than that from the inner. volume. (#) J
- 4. With respect to reactor safety, overcompensation is worse than-undercompensation. (#) 1 i
- 5. The rate of power decrease (DPM) is_ reduced lwhenLpassing:through' l the Intermediate Range. _(#)
(#) Required (0.5 each) j ' REFERENCE CNS LP, (I&C), OP-CN-HO-ENB, p 10, LPSO LO 8. 015000K502 ..(KA's) ANSWER 6.12 (1.50) 1 l l a. Performs a light test (should read all "8's"). l l b. Actuate SDM alarm.
- c. Causes the auto functions of that train to actuate'if " ENABLE".is' selected for that train.
l (0.5 each) l REFERENCE CNS LP, (I&C), OP-CN-HO-ENC, p 8, LPSO LO 4. 000008A121 ..(KA's) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****).
4 fr . PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Page .48 ANSWER 6.13 (1.00) (2 ef 3 << par el) 'l
- 1. The Slave Cycler generates current orders for the coils of the groups of rods to be moved.
- 2. The Slave Cycler compares the count number to the "In" or "Out" signal from the Supervisory Memory Circuit.
femr5 W '/ NM St8vE c y e t t iq 3 ' (13 0.5 E reach) gers REFERENCE 1 CNS LP, (I&C), OP-CN-HO-IRE, p 11,12, LPSO LO 3. 001000K403
.(KA's)
ANSWER 6.14 (1.50)
- a. 4/2/3
- b. 2/1/2
- c. 4/2/3
- d. 2/1/2
- e. 2/1/2 (0.3 EA)
REFERENCE CNS T.S., TABLE 3.3-1, p 3/4 3-4., LPS0 LO 2,10 of OP-CN-HO-IPX, p4 of 20. 012000A306 .(KA's) ANSWER 6.15 (1.50) CA CST, UST, Hotwell, RN, BC (0.20 each, 0.05 for each in correct order) (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- 6. ~ PLANT SYSTEMS DESIGN. CO'NTROL.'AND INSTRUMENTATION' Page j 49 -
. REFERENCE'
'CNS-LP, (COND &. FEED SYS), OP-CN-HO-CA, pf9, ISSILO 2.
061000K401= ..(KA's') 1 l ANSWER' 6.16 ' ( 1. 00 ) '
'1. Pellet
^
- 2. Clad-
- 3. NC system
- 4. Containment vessel- "
- 5. VE system
'(0.20'EA). ,
I
.i l REFERENCE 1
'CNS LP, (CONTAINMENT), OP-CN-HO-CNT, p 4, LPS0 LO 2.
.103000K102 ..(KA's) ~
i l ANSWER 6.17 (2.00) i
- a. The High Setpoint trip provides protection during powei' operations to-mitigate the consequences of a reactivity excursion from all power levels.
- b. The Low Setpoint trip provides protection during.suberitical andLlow power operations to mitigate the consequences of a power ~ excursion beginning from low power. '
- c. The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection accidents. '
- d. The Power Range Negative Rate trip provides protection'for-control- '!
rod drop accidents. (0.5 each) ! l REFERENCE l CNS T.S., LSSS BASES, p B 2-4, OP-CN-HO-IPX.LPS0 LO 11. ' 012000G006 012000G004 ..(KA's) 1 (***** CATEGORY 6 CONTINUED ON'NEXT PAGE *****)
, .i L6: PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION. ' Page .50 8
,)
q ' ANSWER- 6.18 (1.00) } It! ensures' sufficient cooling-capacity.available for' safety'related. equipment (0.5) during normal and accident conditions.(0.5)-. ' j q REFERENCE l 1 CNS LP, (PRI'SUP SYS), OP-CN-HO-KC, p 11, LPS0 LO 8. ) 008010K006 ..(KA's)
.(
ANSWER 6.19 (1.00) 1 The fourth accumulator is assumed to dump out through the cold' leg break -{ (O'.5) .and bypass the core. (0,5)
.]
REFERENCE
.]
CNS LP, (ECCS), OP-CN-HO-CLA, p 8, LPSO LO 6. ) 006000K602 ..(KA's) j q ANSWER 6.20 (1.50) 1 1
- a. ARM & DUMP
- b. ARM ONLY i
- c. ARM ONLYu A## ,
f,Bcto,8) (0.5 EA) REFERENCE j CNS LP, (STM SYS), OP-CN-HO-IDE, p 8,9, LPSO LO 8. 041020K404 ..(KA's) i
'I i
ANSWER 6.21 (2.00) . 1 SEE ATTACHED SHEET f
-I
(***** CATEGORY 6 CONTINUED.ON NEXT PAGE *****) j
- 6. .
PLANT SYSTEMS PESIGN. 90NTROL'. AND INSTRUMENTATION Page 51
!1EFERENCE CNS LP, (I&C), OP-CN-HO-IBX, DWG CN-IC-IRX-6, LPSO LO.3.
001000K403 ..(KA's) l l l l l .- l (***** END OF CATEGORY 6 *****)
51a
- . I j
o* e
\
we OO 3
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ow w an
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5 N
- N g
gl L fB ! 1
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,, a 5
. Y g
.2 5-o <
4 4 I, i 151 s : o 2 .. ~ ! Z w% - e D E
- y e 4.
- a' t t. n b.
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~
1 o -=I E4 4N a y 4 3 oa
;; z
, 4 O#
T. - O t ' ~ 2
- O 1 ,
L .. . 3" ' 1 t O. E R.E.i 5.= l ag 9 o w- - _ . - _ _ - _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ . _ . _ _ . _ - - _ _
1
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Pcgm 52 AND RADIOLOGICAL CONTROL j 1
]
ANSWER 7.01 (1.00) l The automatic overlap function is disabled. (1.00) i REFERENCE OP/1/A/6150/08 ANSWER 7.02' (2.00) i
- 1. NC subcooling (0.2) > OoF (0.2)
- 2. S/G Pressure (0.2) Stable or Decreasing (0.2)
- 3. NC T-hot (0.2) Stable or Decreasing (0.2) 4., Core Exit T/C (0.2) Stable or Decreasing (0.2)
- 5. NC T-cold (0.2) Near Saturation Temp. for S/G press (0.2) I
( ! REFERENCE EP/1/A/5000/1A1 Natural Circulation Cooldown. p. 15 Enclosure 1. ! ANSWER 7.03 (1.25)
- a. +/- 50 (0.25 each)
- b. 1) 50 2) 100
- c. 160
- d. 100 .
l REFERENCE OP/1/A/6100/01 l l l ANSWER 7.04 (1.00) l Notify the Reactor Group Unit Engineer (1.00) (within 16 hours) 1 l (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES ~- NORMAL. ABNORMAL. EMERGENCY ~ ~Page'53
/. AND RADIOLOGICAL' CONTROL ,
} l lRdFERENCE l
-1 O P / 1 / A / 6'1 0 0 / 0 1 ,- p. 17' ]
'i l
' ANSWER 7.05 (1.00) l These substeps (marked'by " bullets")'canfbe~ performed in.any order' -1 \ (1.00) .
~
}
l REFERENCE , OMP 1-4, Operations Management Procedure
- l, i
'lj ANSWER, 7.06 (0.75)
- a. Shift Supervisor (0.25)
Fuel Handling Supervisor (0.25)
- b. Control Room Operator I (0.25) 1 REFERENCE j
1 CNS Directive (OP) 3.1.17 , p.2 d
'l ANSWER 7.07 (1.00) d (1.00 )
REFERENCE 1 EP/1/A/5000/03, Loss of All AC power, p.22 . l 1 l ANSWER 7.08 (1.50)
- a. S/G press (s) < 725 psig (0.5)
- 6. PZR press < 1845 psig (0.5)
- c. Containment press. > 1.2 psig (0.5)
(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) :
' '7 . - PROCEDURES -' NORMAL, ABNORMAL' EMERGENCY m
Page'54:
,' AND. RADIOLOGICAL CONTRO_L
~
REFERENCE:
EP/1/A/5000/01, Reactor Trip or. Safety 1 Injection p.3.1 ! ANSWER 7.09 (1.25) j .- .1. Ensure' Reactor trip (0.25)'- l 2. Verify Reactor Trip- (0.25) (Manually trip Reactor). 3.. Verify. turbine Trip .(0.25)
- 4. Verify,AC power'available (0.25)
- 5. Check SI actuated (0.25)
REFERENCE EP/1/A/5000/01, Reactor Trip or Safety' Injection OP-CN-HO-EP1 Training Objec tives LPSO, #6B-ANSWER 7.10 (1.00)- NC pumps are required to be tripped to prevent pumping NC system. inventory (0.5) out the break for a SBLOCA (0.5). REFERENCE ' OP-CN-HO-EP1, p.19 Training Objectives LPSO #8 p.4 ANSWER 7.11 (1.00) The relative reactivity should be made up by boration untill shutdown margin is equal to that required by Technical Specificaitons. (1.00) REFERENCE EP/1/A/5000/1A ERG Background Documentation p.4 (***** CATEGORY 7 CONTINUED ON NEXT PAGE * * * * * )L
7,- PROCEDURES'- NORMAL. ABNORMAL. EMERGENCY Page'55
'ANS RADIOLOGICAL CONTROL ' ANSWER? 7.12 (1.00)
(Single pump operation infthe-loopfthat) Provides'the best' spray ~is:
- pref erred' to obtain normal PZR spray- capability. (1~.00) . REFERENCE '
EP/1/A/5000/1A1, NaturalLCirculation Cooldown ANSWER 7.13 (1.00) ND (RHR) system (1.00) REFERENCE EP/1/A/5000/1C6 . Training Objectives LPSO #3, Discuss, equipment, switches'andi indications used to determine plant conditions. ANSWER 7.14 (1.50) a). TRUE b). FALSE c). TRUE l
' REFERENCE EP/1/A/5000/02, CSFST, p.3 ANSWER 7.15 (1.00) a). FAULTED refers to failure of the secondary pressure--boundary
' ( O .' 5 )
b). ISOLATED refers to closure of the steam and/or feed flow pa t h's
'( 0. 5 )
(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) t--- . _ _ _ _ _ _ _ . -
E7 . .PROCEDURESt - ' NORMAL'.' ABNORMAL; EMERGENCY" . P a g e f 5'6 :
-. ' ANS RADIOLOGICAL CONTROL , '
N FERENCE. User's' Guide for ERG /s and Background Documents p. 7 ! ANSWER 7.16 (1.50). a). The pump is stopped.to prevent l damaging it during possible. NC voiding-(0.5) AND to ensure is. availability'for future: use;(0,5)-('Gives best PZR flow lavailable) b). NC pump 1B -(0.5) ' REFERENCE EP/1/A/5000/2B2 , p.7 OP-f.N-HO-CSF p. 11 LPSO # 5, Discuss'the majo~r subsequent' actions specified irr the - EP's. . ANSWER 7.17 (1.00) The operator should' return to the next steplor substep in the-left-hand column. (1.00) REFERENCE User's Guide for ERG's and Background Documents p.5 OP-CN-HO-EP1 p.10 l ANSWER 7.18 (1.00) !
> 100 mrem /hr (1.00) i REFERENCE I CNS OP-CN-HO-HPM, p.20 ANSWER ~ 7.19 (1.00) c (1.00)
(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) 3
s.. . 70 PROCEDURES'- NORMAL.> ABNORMAL.' EMERGENCYI
-Pagej57:
. . AND RADIOLOGICAL CONTROL lREFERENCEL EP/1/A/.5000/02, p . 3, l ANSWER 7.20 (1.00)
A11 shutdown banks must be' fully withdrawn. (1.00)l
]
LREFERENCE l OP/1/A/6100/01, Contre 111ng Procedure for. Unit Start Uptp.n2 . 1
.OP-CN-CP-CP' LPSO Training Objective;#9 State the limits and-l precautions associated with each procedure.
1 ANSWER 7.21 (1.00) a (1.00)- REFERENCE OP/1/A/6100/01, Encl. 4.1 p. 15 step 2.78, p.16 step 2.84 OP-CN-CP-CP, Training Objective LPSO #8, State the major { procedure steps in proper sequence. IANSWER 7.22 (1.00) l b (1.00) REFERENCE AP/1/A/5500/04 p.4 l ANSWER 7.23 (2.00) a). To allow proper operations'of the UHI rupture disk leak detection system (1.00) b). N2 will come in contact with the water in the accumulator (0.5) thus resulting in out'of specification N2 concentration. (0.5) (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)'
'7 . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 58' ANB. RADIOLOGICAL CONTROL l
i REFERENCE OP/1/A/6200/10, p.1
- ANSWER 7.24 (2.00) a). Trip one- of the three pumps (0.5) and throttle flow as required. (0.5) b). Will result-in a pump trip (0.5) or switch over to RN (Nuclear Service Water) and RC (Condenser Circulating Water).(0.5) l REFERENCE l
OP/1/A/6250/02, p.4 1 l 1 1 1 1 (***** END OF CATEGORY 7 *****) l 1 w__. . . _ _ _ - - _ - ..
7 (
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. Page 59 l AND LIMITATIONS "
..f a . l ANSWER 8.01 (1.50) !
I The resultant offsite radiation dose (0.5) will be limited to a small ; fraction of 10CFR part 100 dose guideline values (0.5) in the event of { a steam line rupture.(0.5) REFERENCE I CNS Technical Specification 3.7.1.3, B 3/4.7.1.3 035'000G006 ,
..(KA's)
ANSWER 8.02 (1.00)
- 1. , Containment structure is prevented from esceeding its design negative pressure differential with respect to the outside atmosphere of 1.5 psig. (0.5)
- 2. The containment peak pressure does not exceed the design pressure l of 15 psig during LOCA conditions. (0.5) i9EFERENCE CNS Technical Specification 3.6.1.4, B 3/4.6.1.4 002000G006 ..(KA's)
, ANSWER 8.03 (1.00)
- 1. Minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blow down. (0.5)
- 2. Limit the pressure rise within containment in the event the steam line rupture occurs within containment. (0.5)
REFERENCE CNS Technical Specification 3.7.1.4, B 3/4.7.1.4 035006K106 ..(KA's) l l (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ __- ___- _-_-_--____-- a
Bi ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 60 l AND LIMITATIONS q 1 1 ANSWER 8.04 (1.00) I So sufficient time has elapsed to allow the radioactive decay of the j short lived fission products. (1.00) 1 1 REFERENCE I CNS Technical Specification 3.9.3, B3/4.9.3 192006K113 ..(KA's) ANSWER 8.05 (1.50) Sufficient water depth available to remove (99%) (0.5) of the assumed (10%) iodine gap activity (0.5) from the rupture of an irradiated fuel l assembly. (0.5) 1 l REFERENCE e CNS Technical Specification-3.9.9, B 3/4.9.9 092000G006 ..(KA's) ! ANSWER 8.06 (1.25) YES (0.25), there has been violation, the A.O. can not work more than j 72 hours in a 7 day period. (1.00) oc he A . o. can n,.f ut e k y,, c 1A , 16 hear, t,, a t. 3 n .m ,. gera./(e.co)) iREFERENCE CNS Technical Specifications p.6-2 194001A103 ..(KA's) i ANSWER 8.07 (1.50)
- 1. The shift supervisor is available to return to the control room within 10 minutes. (0.5)
- 2. The assumption of SRO duties by the STA be limited to periods not in excess of 15 minutes duration (0.25) total time not to exceed one hour during any 8' hour shift. ' O . 25 /(ce one heve in ,, ,4 3 12 hear .f (' g -
- 3. The STA has a Senior Operator license on the unit (0.5)
(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) l i
.3 8, -ADMINISTRATIVE' PROCEDURES. CONDITIONS. Page 61 1 ANS LIMITATIONS NCFERENCE-CNS' Technical Specification 6.2.2 f p. 6-5 l
194001A103 ..(KA's) l-l ANSWER 8.08 (1.00) Immediately initiate and. continue'boration ( 0.25) :at' 'greaterL than- or - g equ,a1 to 30 gpm (0.25)'of-a_ solution containing greater than'or equal ..; to 7000 ppen baron (0.25) or equivalent until' the required-S/DLmargin ; j is restored'(0.25) l I i) REFERENCE CNS Technical Specification _3.1.1.1, p. 3/4 1-1 001000G006 ..(KA's) I l ANSWER 8.09 (1.00) - Immediately open the reactor trip breakers (1.00) REFERENCE CNS Technical Specifications 3.1.3.3, p. 3/4 1-18 i
)
014000K301 ..(KA's)
]
ANSWER 8.10 (1.00) a) 1 0 '/. (0.5 ea) b) 10 degrees F I REFERENCE
.1 CNS Directive 3.1.19 p. 3 194001A102 ..(KA's) i
'f 4
(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) l
._-____._._____m_ ______ _ _ _ ___.__ _ _ _ _
i
$p ADMINISTRATIVE PROCEDURESc CONDITIONS, .Page 62
," AND LIMITATIONS ANSWER' 8.11 (1.50)
- a. 1. Work-is complete (0.5) ,
- 2. All normal work request functions are satisfied (0.5) I e , *'~ < t ,,,,c .; e < has n w <re I * -.I or?)
- b. hn e"'('tfe'"c~u r'renN, ree eh ug operator) notifies the Supervisor having Operational responsibility. (0.5) t l
REFERENCE l CNS Directive 3.1.1, Safety and Delineation Tags, p. 12, 26, 28 194001K102 ..(KA's) - i l l ANSWER 8.12 (1.50) l
- 1. RCS Boron concentration (0.25 each)
- 2. Control Rod position
- 3. RCS average temperature
- 4. Fuel burnup based on gross thermal energy generation
- 5. Xenon concentration R
- 6. Samarium concentration 1 .
REFERENCE CNS Technical Specifications, 4.1.1.1.1 , p. 3/4 1-1 , 192002K114 ..(KA's) I ANSWER 8.13 (1.25)
- 1. Design limits on Peak local power density (0.25) and minimum DNBR are NOT exceeded (0.25), and
- 2. In the event of a LOCA (0.25) the peak clad fuel temperature (0.25) will not exceed the 2200 deg. F acceptance criteria (0.25).
REFERENCE CNS Technical Specificaitons B 3/4.2.2 and 3/4.2.3 p. B 3/4 2-2a l l 001000G006 ..(KA's) 1 (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
ST' ->
.\\>
- 8.
'ADMI'NISTRATIVE PROCEDURES.'CONDITIONSi "
'Page!63-9 .. . AlD LIMITATIONS.
l
.~
e-IANSWER' O.14 (1.00)
- 1. 'If'the Reactor Building is to' be en tered during[ Modes. '1, 2,13,#(ihSo*h L{.
(0.5), or ' 1
- 2. - Anytime the ice condenser isfto be entered (0.5)'
l REFERENCE . 4 1 1 CNS, Directive 3.1.2 .p.4' 103000G001 ,
..(KA's) -1
' ANSWER 8.15 (1.00) l A
. c.
jREFERENCE l l CNS Directive, Saf ety Tags and Delineation Tags, p. 11 194001K102 ..(KA's)
']
ANSWER 8.16 (1.00) D. (1.00) l REFERENCE i ' Operations Management. Procedure 1-5, CNS Independent Verification p i '5 ; 194001A102 ..(KA's) l ^ lANSWER' O.17 (1.00). Can be waived if radiation greater than 100 mrem' dose,(0.5) ~and.may b'e: waived with the Unit Coordinator or Duty Engineer's approval.(0.5) l REFERENCE 1 Operations Management Procedure 1-5, CNS Independent Verificaiton, P.5 194001K103 ..(KA's) (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)' l l
- 8. ' ADMINISTRATIVE' PROCEDURES, CONDITIONS, Page'64 8 "ANE LIMITATIONS
. y l ANSWER. 8.18 ti rSet (l.co)
.. Wall be revicacd mor,thly to verify that thcy match the-Ur ent r.cntr:1 ccpy. (e.5e)
- b. Station Manager:or Group' Superintendent.'(0,5).
- c. Fill out a Procedure Discrepancies Progress Record (per Station directive 4.2.1) (0.5) l REFERENCE Operations Management-Procedure 1-4, p.4,.15, 13 194001A102 194001A101 ..(KA's)
ANSWER, 8.19 (1.25) I
- a. Safety Limit (thermal power) has been violated (0.25)
- b. 1. NRC operations center shall be notified by telephone as soon as possible and in all cases with in one hour. (0.25)
- 2. A safety limit violation report shall be prepared (0.25) l l
- 3. Safety limit violation shall be submitted.to the Commission l (with in 14 days of the violation). (0.25) l
- 4. Critical operation of the unit shall NOT'be resumed until authorized by the commission. (0.25) l REFERENCE Technical Specification 6.7.1 p. 6 -12 001000G005 ..(KA's) l4NSWER 8.20 (1.00) l b
l REFERENCE CNS Technical Specification 3.11.2.5 , p 3/4-11-16 194001A103 ..(KA's) i (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)
8 ADMINISTRATIVE 1 PROCEDURES. CONDITIONS. Page 6S', t SP - %NR L IMITAT IONS
/
i .'
.: j l ANSWER 8.21 (1.00)
I
- a. (1) Station Manager -(0.25.each) !
- 6. (2) Safety Supervisor
- c. (3) Assistant Shift Supervisor q (4)' Control Room Operator J i
3 REFERENCE - 1 i CNS, Directive 2.12.2, Fire Brigade Orgainzation,L Training, and Responsibilities- p.1- , 194001K116 * ..(KA's) l ! ANSWER 8.22 (1.00) j j
- a. (1) Station Manager . .i l
(2) 3 (0.25 each): )
- b. (1) Recovery Manager l (2) Emergency Coordinator ]
1 l l l REFERENCE l 'h CNS Directive 3.8.5, Exposure Extensions and/or Exposure. limit I Reductions. I 194001K104 ..(KA's) i l ANSWER 8.23 (1.00) i Area shall be conspicuously posted (0.S) and a flashing light shall be j activated to warn personnel to stay clear of the area (0.5). i l REFERENCE CNS Directive 3.8.8 , Radiological Work Practices p.3 194001K103 ..(KA's) (***** CATEGORY 8' CONTINUED ON NEXT PAGE'*****) t--_-__ _ - - . _ _ - _ - - _ - _ - _ _ - _ _ _ - _ - _ - _ _ - - _ - - _ _ .
-v - - -
j ,I. L , 8;' ADMINISTRATIVE' PROCEDURES,' CONDITIONS. Pagaf66'
'.(AND LIMITATIONS J
W= i'sY ANSWER 8.24 (1.50) . a). Yes (0.50 each) 1 b). Yes l( c). Yes '
-{
REFERENCE j CNS. Technical Specifications Table 3.3.1, LCO 3.3.3.2', 3.3.3.3,;3.3.3,4 OP ,CN-HO-IPX, LPSO'#10 OP-CN-HO-IEE, LPSO #5-l OP-CN-HO-END, LPSO #10 OP-CN-HO-ENA, LPSO #9 015000G005 ..(KA's) u
- i j
I l
>j j
.q l
I f l (***** END OF CATEGORY 8 *****) (********** END OF EXAMINATION **********) I L- u _ - - - _ - - - - _ - _ - - - - - - - - . .
"< .h$>,
U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENS.E EXAMINATION FACILITY: CATAWBA ; l REACTOR TYPE: PWR-WEC4 1 DATE ADMINISTERED: 87/09/14 EXAMINER: AIELLO. RF CANDIDATE j l' 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 i least 80%. Examination papers will be picked up six (6) hours after the examination starts. l 1 l % OF l CATEGORY % OF CANDIDATE'S CATEGORY j VALUE TOTAL SCORE VALUE CATEGORY !
- i 30.00 24. 9d 4' 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ;
28.25 23.4 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
$Z.co v(
32w26I# 26.36#5 3. INSTRUMENTS AND CONTROLS 30.00 24.9d 6 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL j lf 120 W M % Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature l i l i l
- - - - - - - - - -_ _ J
f $$Yb{l p,,, ' NRC RULES AND' GUIDELINES FOR LICENSE EXAMINATIONS . L a ! During,the administration of . this: examination the1fo11owing rules apply: l : 1.- Cheating.on the examination'means an automatic denial.of your application end could result in more' severe penalties.
- 2. _Restroom trips.are to be limited;and only one candidate at a time,'may 1 cave. You must avoid all contacts with anyone outside the examination a room to avoid even the appearance or possibility of cheating.-
- 3. Use black ink or' dark pencil 1only to facilitate legible' reproductions.
- 4. Print your name in the blank provided onfthe 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.
I 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, j
- 13. The point value for each question is indicated in, parentheses after~the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask. questions of the examiner only.
- 17. You must. sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given' assistance in completing the examination. This must be done after the examination-has been completed.
l I
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- 18. When you complete your examination, you shall:
- a. Assemble your exam'ination 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.
l l l l l l \. ..x _. . _ _ _ _ _ _ . . _ _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.1 ' . . .RRINCIPLES-OF NUCLEAR POWER PLANT OPERATION.- Pcga 4 L THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW I
QUESTION l'.01 (1.00) ; Which one of the following will NOT change over core life?
- a. The minimum acceptable shutdown margin
- b. The acceptable reactor power imbalance
- c. Differential boron worth
- d. Doppler deficit
- e. Peak Sm worth after SD from full power i
l l QUESTION 1.02 (1.00) ,
)
Which one of the following is NOT true concerning rod worth? l a. Over core life rod worth will decrease where flux increases and j increase where flux decreases.
- b. Rod worth will increase as moderator temperature increases. l
- c. High concentrations of Xenon will cause a noticable change in ;
integral rod worth. I
- d. Differential rod worth will decrease when inserting adjacent rods j to the same height. '
QUESTION 1.03 (1 00) l l l The reactor is critical at 10,000 cps when a S/G PORV fails open. l Assuming BOL conditions, no rod motion, and no reactor trip, choose ~ the answer below that best describes the values of Tavg and nuclear power for the resulting new steady state. , (PDAH = point of adding , l heat),
- a. Final Tavg greater than initial Tavg, Final power above PDAH. ,
l
- b. Final Tavg greater than initial Tavg, Final power at POAH. ;
- c. Final Tavg less than initial Tavg, Final power at POAH. l l l l d. Final Tavg less than initial Tavg, Final power above PDAH.
l j 1 1 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l _ _ _ - _ _ _ _ _ _ _ _ _ . - _ _ _ . - _ - _ _ _ _ - _ _ _ _ - - _ _ _ _ - _ - . _ _ _ _ _ . _ _ _ _ _ - _ . -_ . _ _ _ _ - _ . - ________-_______.___a
a el. ~RRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Page S b THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.04 (1.00) The reactor trips from full power, equilibrium xenon conditions. Six hours later the reactor is brought critical at 10E-O amps on the inter-mediate range. If power level is maintained at 10E-B amps which one of the following statements concerning rod motion requirements for the next two hours is correct?
- a. Rods will have to be withdrawn since xenon will closely follow its normal build-in rate following a trip.
- b. Rods will have to be inserted since xenon will closely f follow its normal decay rate following a trip.
I
- c. Rods will have to be rapidly inserted since the critical reactor will cause a high. rate of burnout,
- d. Rods will have to be rapidly withdrawn since the criticai l
reactor will cause a higher than normal rate of build-in. l l QUESTION 1.03 (1.00)
)
The major constituent of the power coefficient at BOL is the: ) l a. Doppler coefficient , 1 i l b. Moderator temperature coefficient
- c. Boron coefficient '
l
- d. Void coefficient l
l l QUESTION 1.06 (1.00) l Under which one of the following conditions is the Moderator Temperature Coefficient most negative?
- a. BOL, high temperature
- b. BOL, low temperature 1
l l c. EOL, high temperature 1 l d. EOL, low temperature l 1 l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l
)
al.* RRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pago 6 l THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW j 1 1 [ QUESTION 1.07 (1.00) With the plant operating at 85% power and all systems in a normal / auto 1 configuration, the operator barates J OO PCM. Shutdown Margin will: I (choose one) l 1 i
- a. Increase
- b. Increase until rods move j
- c. Decrease j k
- d. Decrease until rods move j 1
\
l e. Remain unchanged regardless of rod movement l i I QUESTION 1.08 (1.00) I I Choose the correct order of the boiling phases listed below as they would occur in a coolant channel with high heat flux?
- 1. Critical Heat Flux l 2. bulk boiling l
- 3. film boiling l 4 sub-cooled nucleate boiling '
- a. 2, 4, 3, 1 I
- b. 2, 4, 1, 3
- c. 4, 2, 3, 1
- d. 4, 2, 1, 3 l
l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) I l 1 I __ . _ _ . . . _ _ _ ______________u
J el .~ -RRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Page 7 1 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW 1 l QUESTION 1.09 (1.00) k l; During a reactor startup you have Just verified a constant startup rate on the SR Nuclear Instruments without any rod motion or boron dilution. l The actual condition of the core with this indication is l a. Prompt Critical i
- b. Critical i l
- c. Supercritical l l d. Subtritical l !
1 l QUESTION 1.10 (1.00) Which one of the following is the purpose of using soluble boron to control the excess reactivity of the reactor?
- a. It does not significantly af fect the flux shape
- b. It does not significantly affect the rod worth j i
- c. It is more cost effective than adding more rods
- d. It increases reactor loading rates QUESTION 1.11 (1.00)
Which set of parameters below best describes centrifugal pump runout conditions?
- a. High discharge pressure, high flow, high power demand
- b. High discharge pressure, low flow,.Iow power demand
- c. Low discharge pressure, high flow, high power demand
- d. Low discharge pressure, high flow, low power demand
- e. Low discharge pressure, low flow, high power demand
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
'1 f5 1 ARRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pags 8 THERMODYNAMICS, HEAT TRANSFER AND FLUID' FLOW l
QUESTION 1.12 (1.00)' l Which statement is true if natural circulation is losts
- a. Core delta T approaches zero, S/G 1evel increases, S/G pressure decreases.
- b. Core delta T de constant at approximately 60% full power value, S/G level is constant, S/G pressure decreases. t
- c. Core delta T exceeds'100% full power.value, S/G. level increases, S/G pressure decreases,
- d. Core delta T will exceed 100% full power value, S/G 1evel decreases, S/G pressure increases.
QUESTION 1.13 (1.00) l At normal operating temperature, a leak from the PZR water space to the containment atmosphere would consist of: (choose one) l
- a. Superheated steam.
i
- b. Saturated steam,
- c. Low quality steam. l
- d. Saturated water, i
QUESTION 1.14 (1.00) In the event of a rod ejection accident, which one.of the following will be the first reactivity coefficient to insert negative i reactivity?
- a. Moderator temperature coefficient.
- b. Pressure coefficient.
- c. Void coefficient.
- d. Doppler coefficient. l
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
E M'1,"" PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pago 9 4 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW GUESTION 1.15 (1.00) When synchronizing the generator to the grid, OP/1/B/63OO/01 " Turbine-generator Startup" directs the operator _to regulate turbine speed to slowly rotate the synchroscope in the fast (clockwise) direction. Which choice below CORRECTLY gives the two parameters that the synchroscope.is indicating?
- a. Current and voltage differences
- b. Voltage and frequency differences
- c. Frequency and phase' differences
- d. Phase and resistance differences
- e. Resistance and current differences QUESTION 1.16 (1.00)
Which one of the following statements is CORRECT concerning the paralleling of electrical systems?
- a. Although it is desirable to have speed and phase position matched, it is much more important to have voltages matched.. ;
1
- b. If voltages are not matched at the time the synchronizing switch is closed, there will be VAR flow from the lower voltage source to the higher one.
- c. If the incoming machine is at synchronous speed but out of phase with the running bus when the breaker is closed, heavy currents will flow to either accelerate or retard the incoming machine,
- d. If the incoming machine is in phase but slightly faster than synchronous speed when paralleled, the system will tend to speed up
-th : ;yctcn to synchronous speed.
h%
- e. If the resistances are not matched at the time the synchronizing switch is closed, heavy currents will flow to tend to speed up the incoming machine to synchronous speed.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) e _ _._ _ _ _ ____ A ___ _ ______._ _ __._._______ _ . _ _ _ _ _ _ _ _ _ _ _ _ __ .______.__________.________________J
4."-PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Paga 10 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW I
. i
.I QUESTION 1.17 (1.00) j Which one of the following is true concerning Xenon during a power 1 increase? ,
1
- a. Xenon concentration will dip reaching a-low in 4 to 8 hours.
- b. Xenon concentration will increase reaching a high in 4 to 8 hours.
- c. Xenon concentration will dip reaching a low in 12 to 18 hours.
- d. Xenon concentration will increase reaching a high in 12 to 18 ;
j hours. l l i QUESTION 1.18 (2.50) , l ! l The plant is operating at 30% power, turbine in AUTO (IMP IN), when loop
#1 reactor coolant pump trips. Assuming no reactor trip, no operator action and rod control in MANUAL, indicate whether the following parameters will be HIGHER, LOWER or the SAME at the end of the transient j compared to their initial values.
i l 1. #2 S/G steam pressure l 1 l 2. #3 RCS loop flow ; l
- 3. Tc in loop #1
- 4. Th in loop #2
- 5. Nuclear Power QUESTION 1.19 (1.00)
Assuming all other DNB parameters remain constant, how will the following changes affect DNBR? (INCREASE, DECREASE, NO CHANGE)
- 1. Reactor thermal power increases 1
- 2. Average NC temperature increases
- 3. NC pressure increases
- 4. NC flow increases 1
1 (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) i l
1 l 1,' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pcge 11 i THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW J l i i I QUESTION 1.20 (1.50) I Indicate whether the following will cause the power range instrument ! to be indicating HIGHER, LOWER or the SAME as actual power, if the , instrument has been adjusted to 100% based on a calculated calorimetric.
- 1. If the feedwater temperature used in the calorimetric was higher than actual feedwater temperature.
- 2. If the reactor coolant pump heat input used in the ca.orimetric was omitted.
l
- 3. If the steam flow used in the calorimetric was lower than actual. l l i l \
l l l QUESTION 1.21 (1.50) 1 Answer EACH of the following as TRUE or FALSE concerning a 60% load rejection. Assume no operator action.
- 1. Pressurizer pressure will spike high due to an increase in Tave and slowly decrease due to spray.
- 2. Generator load goes to zero immediately after the load rejection.
- 3. Steam Generator levels will initially decrease due to " shrink" and increase from " swell" as the steam dumps actuate.
QUESTION 1.22 (1.50) Answer each of the following as TRUE or FALSE concerning natural circulation.
- 1. Feeding steam generators too fast can reduce NC system flow.
- 2. Two phase flow (NC system at saturation) can be sufficient to ensure heat removal.
- 3. An indication of natural circulation will be NC T-hot near saturation temperature for steam generator pressures.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
l 1., PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Page 12 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l 1 - i OUESTION' 1.23 (1.50) j Match the heat transfer equation in Column 1 with type of heat transfer that it represents in Column 2. (answers in Column 2 may be used more than once.) i Column 1 Column . :2
- 1. b = UA delta T a. radiation
- 2. b = MC delta T b. conduction 1 1
- 3. b = MC delta h c. convection l
GUESTION 1.24 (1.00) i During a reactor startup, equal increments of reactivity are added J i and the count rate is allowed to reach equilibrium each time. Choose the bracketed ([]) words that describe what is observed on the Source Range recorder and/or SUR meter. 1
- a. The change in equilibrium count rate is [ larger] [the same] i
[ smaller] each time,
- b. The time required to reach equilibrium is [ longer] [the same]
[ shorter] each time, i ! QUESTION 1.25 (1.50) List three factors which will affect the amount of core uncovering following a small break LOCA during power operations within the program pressure band. QUESTION 1.26 (1.00)
- a. If the reactor is operating in the power range, how long Will it take to raise power from 20% to 40% with a +0.5 DPM Startup-rate?
i
- b. How long will it take to raise power from 40% to 60% with the same
+0.5 DPM Startup rate?
i (***** END OF CATEGORY 1 *****) I 1 l \ - _ - _ - - _ _ _ - _ -
< tp . .
i:2W PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Fags 13 1 Y'f ' SYSTEMS 1
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QUESTION 2.01 (1.00) Which ONE (1) of the following statements, regarding the interlock between the GTA(B) breaker and the FTA(B) incoming feeder from ETA (B), is correct?
- a. If GTA(B) is opened, FTA(B) cannot be closed.
- b. If FTA(B) is opened and the GTA(B) is then closed, FTA(B) will trip.
.I
- c. If GTA(B) is closed, FTA(B) cannot be closed. H
]
- d. If FTA(B) is closed or'the GTA(B) is closed, FTA(B) will trip.
QUESTION 2.02 (1.00) l l When conducting a cooldown on the ND system, the operator will manually adjust which of the following valves in order to control the'cooldown rate j] per OP/1/A/6100/027 ' i
- a. KC outlet isolation valves from ND heat exchanger (VC-57A & B28).
- b. ND heat exchanger outlet valves (ND-26 & 60).
- c. ND heat exchanger bypass valves (ND-27 & 61),
- d. ND heat exchanger outlet valves AND bypass valves.
I 1 1 l l 1 i 1 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) i i
i 2L PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 14 i SYSTEMS' i 1 L) i
'i
.4 QUESTION 2.03 (1.50) l
,l With regard to the chemical and volume control-system, state to what q position (OPEN, CLOSE, AS IS, or. COMPONENT they divert to) the following j valves fail on a loss of control signal. ]
1
- a. Letdown isolation valves-(NV-1A & NV-2A) f q
- b. ND control valve (NV-135)
- c. Charging line isolation valves (NV-312A/314B) i
- d. Letdown temperature control divert valve (NV-153A) l
- e. Excess letdown isolation valves (NV-1228) q l
- 1. Excess letdown divert valve (NV-125B) ]
1
)
i QUESTION 2.04 (1.00) I l b l State whether the following statements regarding control room l l ventilation are TRUE or FALSE.
- a. The System is required to be' operable by Tech Specs in ALL modes of j operation. (T.S. 3/4 7.6) I 3
- b. An interlock is provided such that only ONE (1) train will operate at a time.
1 l 1 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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2.' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paga 15 SYSTEMS QUESTION 2.05 (1.00) State whether the following statements regarding NCP operation are-TRUE-or FALSE.
- a. ALL NC pumps and ND pumps may be deenergized for up to 1 hour provided: No operations are permitted that would cause dilution of the NC system baron concentration, OR core outlet temperature is maintained at least 10 deg F below saturation temperature.
- b. A NC pump shall NOT be started with one or more of the NC system cold leg temperatures less than or equal to 285 deg F unless the secondary water temperature of each S/G is less than 50 deg F above each of the NC system cold leg temperatures.
QUESTION 2.06 (1.50) l Match the chemistry control in Column "A" to the chemical used for that control in Column "B". l l COLUMN A COLUMN B
- a. Control pH during startup 1. Hydrazine l
- b. Scavenge Oxygen during 2. Hydrogen startup from cold conditions i
- c. Control Oxygen during full 3. Lithium Hydroxide
, power operations l 4. Ammonium Hydroxide QUESTION 2.07 (1.50) List the IMMEDIATE trips that will shutdown the emergency diesel generator (EDG)during manual operation. (include trips requiring j operator action) 1 (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) 1 i
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_~__ -- - - [!2 ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcgo 16 SYSTEMS
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l 1 I QUESTION 2.08 _(1.00) { 0 The Nuclear Service Water System (RN) consist of FIVE (5) functional ! sections. List them. I QUESTION 2.09 (2.00) I State FOUR conditions required for the existence of containment integrity. I GUESTION 2.10 (1.00) List the FOUR (4) core bypass flows and the percentage of each. ) l j 4 QUESTION 2.11 (2.00) J l List FOUR reactor coolant system leakage limitations. l j l QUESTION 2.12 (1.50) What are the THREE (3) load transients for which the pressurizer is ! designed to satisfactorily operate during normal operations? l 1 I I OUESTION 2.13 (1.00) List the FOUR (4) signals that actuate feed isolation. i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
h
- 2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pega 17 SYSTEMS QUESTION 2.14 (1.75)
List the FOUR (4) main steam isolation ac tuation signals. Include in , your answer applicable coincidences / number of channels to trip.
' QUESTION 2.15 (1.50)
List,in order of preference, the sources of water to the auxiliary feedwater (CA) system 7 i OUESTION 2.16 (1.50) l List THREE of the four resets that occur as a result of depressing the Rod Position Startup Push Button. I I 1 QUESTION 2.17 (1.50) State the TWO alarm conditions which require opening the NC pump #1 seal bypass valve, and the FOUR conditions that must be met to open this bypass valve. QUESTION 2.18 (1.00) State the minimum time the operator must wait prior to attempting another EDG start following S/D from an EDG operability test? QUESTION 2.19 (1.00) State the reason (basis) why TWO (2) independent KC loops shall be operable in modes 1,2,3,4 per T.S. 3.7.3. (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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((#' PLANT' DESIGN INCLUDING 5AFETY AND EMERGENCY Pcgs 18 SYSTEMS p 1 1 QUESTION 2.20 (1.00) Why should Caution be used when equalizing and opening MSIV's with the { reactor critical and S/G pressure > 725 psig? ! I i j j QUESTION 2.21 (1.00) ) Why are pressurizer heaters placed in MANUAL and energized during j boration or dilution actions? 1
'1
)
QUESTION 2.22 (1.00) Why does T.S. (3.5.1.1) require 4 cold leg accumulators to be on line when the contents of only 3 accumulators need be injected in accordance I with safety analysis? l 1 (***** END OF CATEGORY 2 *****)
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13[ INSTRUMENTS AND CONTROLS Page 19 p ' n R. LQUESTION. 3.01 (1.00) Which ONE (1) of the following statements regarding the nuclear ~ instrument system is correct?
- a. Tests may be performed on only ONE-(1) of the four nuclear instrumentation. system protection channels at a time. The redundant protection channels not'under test.must be~ capable of performing.the trip. logic,
- b. An interlock between the. level Trip and Operation Selector Switches on the Intermediate and Power Range drawers requires that the level trip switch be in the Bypass position before testing can be performed on the associated channel.
- c. The neutron detectors and associated cables must not be subjected to more'than 135 deg F continuously. A short term excursion for up tcr ten (10) hours to a maximum of 175 deg F-is permissible.'
- d. Do not reset the Source Range' Reset-Block switches when below 4 x-10 exp-10 amperes indication on either Intermediate' Range channel.
QUESTION 3.02 (1.00) l i Which ONE (1) of the following statements concerning the CA flow control
. valves is correct?
- a. The controllers on Unit i are 0% (full _ closed) to 100 % (full open) and on Unit 2 are 100% (full closed) to 0% (full open).
- b. The controllers on both Units are 0%'(full closed) to 100% (full open).
- c. The controllers on Unit 1 are 100% (full closed) to 0% (full open) and on Unit 2 are 0% (full closed) to 100% (full open).
- d. The controllers on both Units are 100% (full closed) to 0% (full open). j l
1 2
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(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) I
[hhINSTRUMENTS'ANDCONTROLS- r
- Page 20 ,
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QUESTION 3.03 (1.00) iWhich ONE (1).of:the'following malfunctions-would causera p'ressuri'zer level indication of 0% ?' ]_ 1 j
- a. DP cell diaphragm rupture,
- b. Reference leg rupture
- c. Impulse line rupture
- d. Equalizing valve leakage.
1
'1 i
i OUESTION 3.04 (1.00) Indicate whether the following statements, regarding the Gamma Metrics Flux Shutdown Monitor, are TRUE or FALSE.
- a. Pulling of the SDM fuses DOES NOT make the'SDM inoperable. ]
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- b. The alarm setpoint tracks in the decreasing direction ~ONLY.
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l l 1 OUESTION 3.05 (1.50) l Indicate whether the following statements concerning a resistance : temperature detector (RTD) are TRUE or. FALSE.
- a. An RTD is connected.across one leg of a. bridge circuit. As temperature that is sensed by~the RTD changes, a proportional-change in the output voltage (current)~across the bridge occurs.
- b. When an RTD fails open, it will indicate a downstale (low) reading-on its meter.
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- c. If an RTD is completely submerged, 1.ts ability to accurately monitor temperature is unaffected by flow rate.
I (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) 1 1 l l
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D3. . INSTRUMENTS AND CONTROLS Page.21 : I i OUESTION 3.06 (1.75) ) Match the following permissive interlocks in column "A". with their respective function in column "B". l COLUMN _A COLUMN _B
- a. P-11 1. Turbine Trip Feedwater Isolation < Low Tavg, arms !
condenser dumps. Allows block of Safety Injection l signal after time delay block.
- b. P-9 2. Allows block of Source Range reactor trips,
- c. P-7 3. Unblocks "At Power" reactor trips.
- d. P-8 4. Unblocks the 1/4 Loss of Loop Flow.
l e. P-6 5. Allows block of the I.R. High Flux low Setpoint. reactor trip. l
- f. P-10 6. Allows reactor trip on turbine trip.
- g. P-4 7. Blocks automatic and manual Rod Withdrawal.
- 8. Actuates Turbine Runback.
- 9. Allows Manual Block.of Low Pressurizer Pressure, Safety Injection, and Low Steam Pressure Safety Injection.
GUESTION 3.07 4-;.mos (l.77) l List all the signals that will cause a main steam line isolation and l their trip setpoint (if applicable). l I I QUESTION 3.08 (1.50) What 3 signals are summed to produce a CV tlow reference signal for the I flow control unit to the generator electro - hydraulic control system? (Computing circuits (CN-SYS-EHC 9 through 11)) l l b l 1 i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l
- 3. INSTRUMENTS 2AND' CONTROLS- Page 22
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-j QUESTION 3.09 '(i.50) .]1 Listuthe. Pressurizer pressure control ~(IPE)-or-protective action -
l associated with.each of the following.IPE and protection setpoint,- l
- a. 2385.psig' 1
'b. 2335 psig j
.c. 2315 psig
- d. 2210.psig
- e. 1955 psig
- 1. 1845 psig 1 ,
OUESTION 3.10 (2.00) 1 Compare the effects of an under-compensated ion chamber of.the i Intermediate' Range nuclear instrument (IRNI) to a normal IRNI following. a reactor trip. Include in your answer the magnitude of the. gamma signal-of the inner volume verses the outer volume, under-compensation With l respect to reactor safety, and the rate of power decrease when passing-through the intermediate range. (List FOUR) l QUESTION 3.11 (1.50) What THREE actions occur as a result of depressing the " TEST-pushbutton on the shutdown monitoring panel (SDM)? OUESTION 3.12 (1.00) ' List two functions of the slave cycler associated with the Rod Control Logic cabinet. (*****' CATEGORY 3 CONTINUED ON NEXT PAGE *****) l f' O _ _ _ _ _ _ _ _ . - - _ . . _ _ _ _ _ _ _ _ . _ - _ _ . _ _ _ . _ _ _ _ _ _ _ - - _ _ _ _ _
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- 3. ', INSTRUMENTS AND CONTROLS Page 23 QUESTION 3.13 (1.50)
List the total number of channels, the number of channels needed to trip and the minimum channels required to be operable for the following reactor trip system instrumentation while in mode ONE (1).
- a. Pressurizer Pressure - High i
- b. Safety Injection Input from ESF
- c. Power Range Neutron Flux (P-8) j
- d. Reactor Trip Breakers i
- e. Automatic Trip and Interlock Logic. j l 1 I
l 4 l OUESTION 3.14 (1.50) l; I
- a. What are the High High S/G level Feedwater Isolation Setpoints for l Unit 1 and Unit 2 respectively? !
l \ > l b. With a High High S/G level signal present, can Feedwater Isolation tne reset? ] 1 l { GUESTION 3.15 (1.00) , g /g/ l What is the function of the Kirk - Key interlocks on the 600Vpd Essential l Load Center? l lOUESTION 3.16 (1.00) Excluding manual initiation, list the signal (s) that cause each of the l following:
- a. A containment Phase "A" isolation.
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- b. A containment phase "B" isolation
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l i L--______----___--_-----_-__--_-----_-----_ . - _ _ - - - - - _ . - - - - _ -- - - - . --_-_ - . _ - .--__.____-_---J
- 3. -INSTRUMENTS AND CONTROLS Page 24
+ . ' QUESTION 3.17 (2.00) l ListRthe. motor driven CA pump auto' start signals. (Include coincidence- ]
logic when' applicable.) GUESTION 3.18 (1.50). State the effects for the following. Power' Range nuclear instrument channel failures on the Rod Control and Reactor Protection systemst. l (Assume 75% power, EOL, the failing channel is the controlling channel, Rod Control is in' automatic and no-operator' action)
- a. Failed High )
i I
- b. Failed Low l'
l a l l U OUESTION 3.19 (1.00) d With the Pressurizer Level-Control. Selector Switch in position 1-2, a failure causes the following plant events: (Assume no operator actions taken.) l
- 1. Charging flow reduced to minimum
- 2. Pressurizer level decreases
- 3. Letdown secured and heaters off
- 4. Level increases until high level trip WHAT level channel failed and HOW did it fail? (i.e. high, ' low, or as is) l l
l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l l l _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ __w
3 L3 . INSTRUMENTS AND. CONTROLS Page' 25
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l ; LOUESTION 3.20 (1".50) j 1 LState.'the' disposition of the reactor tripi(RT.) and-bypass (BY) breakers -l
.given the following THREE; sets ofLinitial-conditions:' i
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- a. If one train is placed-in test while the other train's BYl breaker;is- ;
closed.- 1 f j
- b. An attemtt'is made'to close both BY breakers 1at.the same time.
- c. A solid state protection system (SSPS) train "A' -reactor trip i signal is generated.
I i QUESTION 3.21 (1.00) l l From the selection of LOGIC system failures below,'classif'.whether y they. l are URGENT or NONURGENT failures.
- a. Failed' Pulser (oscillator failure). !
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- b. Receipt of a "GO" pulse to the Slave Cycler before completion of the ;
previous cycle-(Slave Cycler' failure). I
- c. Loose or removed circuit card.
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- d. Loss of any 1 of 6 DC power supplies-to the control circuits-(power supply failure)
.1 4-
, OUESTION 3.22 (1.50) l 1 1 For the following instrument malfunctions, indicate whether the steam ! dump control system would ARM ONLY, ARM & DUMP or NOT BE AFFECTED. ' l
- a. Tavg fails high with a 10% step load reduction in progress.
- b. Turbine impulse pressure (Channel I) fails high with a 25% step load reduction in progress.
- c. Turbine impulse pressure (Channel II) fails low with.a 5% step load. 1 reduction in progress. l t
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) H
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- 3. INSTRUMENTS'AND CONTROLS Pagel 26 >
TodESTION'. 3.23 (2.00) Draw the Rod Speed Program curve using'the attached:" ROD; SPEED-vs COMBINED ERROR SIG GRAPH". Label the followingi'-
- 1. Maximum Rod Speed = band 1
- 2. Minimum Rod Speed band
- 3. Proportional band ; i
- 4. Lock up j
- 5. Dead band ..
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- 6. Insertion
- 7. Withdrawal
{
- 8. Steps / min- VS. degrees F l 1
l s
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u 1 1 i i (***** END OF CATEGORY 3 *****) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._2_______________________ _______________1_ . _ _ _ _ _ _ _ _ . ______1._____ ___._)
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-4.*. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY Page 23 AND RADIOLOGICAL CONTROL l
i QUESTION 4.01' (1.00) j l If a contingency action (Response Not Obtained) in an EP can not be ; performed or is not successful, and further contingency instruction is ; not provided, what actio.1 should the operator take? I QUESTION 4.02 (1.00) In the event of a Reactor Trip.or Safety Injection what additional l j condition (if any) is required for entry into EP/1/A/5000/01, " Reactor Trip or Safety Injection"? i I QUESTION 4.03 (1.25) What are the 5 Immediate Operator Actions for EP/1/A/5000/01, " Reactor Trip or Safety Injection"? i QUESTION 4.04 (1.00) R While ac ting on a RED path for Heat Sink, an ORANGE path for Criticality comes in, a). Which one will take priority? b). What action (if any) should be taken by the operator? QUESTION 4.05 (1.25) List in order of severity the four (4) levels of Emergency Classifications? (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
4." PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 29 AND RADIOLOGICAL-CONTROL QUESTION 4.06 (1.00) 1
-In accordance with EP/1/A/5000/2C1. " Loss of Secondary Heat Sink",
there is a CAUTION that requires, "...any NC pump' running in a loop 1 with a DRY S/G .... should be stopped." ) d What is the basis for stopping that NC pump? ] i l QUESTION 4.07 (1.00) I l In accordance with EP/1/A/5000/2F3, " Void in Reactor Vessel": a). What are the three (3) methods used to collapse a condensible void in the Reactor Vessel? b). What is the method used to collapse a non-condensible void in the Reactor Vessel? QUESTION 4.08 (1.00) l Fill in the blanks for the following statements: In accordance with EP/1/A/5000/03," Loss of All AC Power": l l a). After starting a Diesel Generator, RN flow must be l established within minutes to its related KD heat exchanger. l b). NC pump seal cooling must be restored within minutes l to prevent damage to NC pump seals. l l l l l 1 l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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- 4. ^ PROCEDURES'- NORMAL. ABNORMAL. EMERGENCY Paga 30 AND RADIOLOGICAL CONTROL 4
QUESTION 4.09 (1.00) Fill in the blanks for the following statements: a). If Reactor control is in manual, Tave should be maintained j within +/- degrees F. of Tref to prevent receiving J
" Tref /Tauct. Hi/Lo" alarm.
b). The difference between the Baron concentration of the PZR and the NCS should be maintained within ppm. c). The NC system temperature and pressure should not exceed (1) degrees F and (2) psig respectively, when the ND system is in operation. 1
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QUESTION 4.10 (1.00) l On a power increase from 15% to 75%, the following actions are performed. a). Transfer steam seal supply from Auxiliary steam to Main ; steam. ] b). Verify SSRH high load valves automatically open, c). Start a second CF pump. 1 d). Ensure that OPTR is < or = to 1.02 1 , l What is the CORRECT order of performance of the above actions? l j l ! l l l l 1 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
- 4. * ~ PROCEDURES - NORMAL . ABNORMAL, EMERGENCY Pcge 31 AND RADIOLOGICAL CONTROL j
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. .l GUESTION- 4.11 (1.00) i During normal. plant operations, which one of the following is the
. maximum S/G primary (NC) to Secondary (SM) differential pressure? f I
a). 1000 psid l 4 b). 1200 psid j j c). 1400 psid d). 1600 psid I
)
i 'l l QUESTION 4.12 (1.00) 1 l I Referring to the attached excerpt from OP/1/A/6100/01, Enclosure 4.1, l for performing a Unit Startup, what is the significance of the
" Bullets" preceding the substeps following step 2.63?
l l l l QUESTION 4.13 (1.00) l What must be done when leaving a contaminated area'(RCA) if the background ra'diation level on the closest frisker is greater than 240 I cpm on the X-1 scale? l 1 QUESTION 4.14 (1.50) What are the Catawba Maximum Administrative Guarterly Exposure Limits beyond which exposure limit extensions are required, for: a). whole body ; b). Skin of the whole body C). Extremities l l l l l j (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) l I L__---_--______. . _ _ _ _
L 4.. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pcgs 32 . l AND-RADIOLOGICAL CONTROL ' 1 i . l i QUESTION 4.1S (0.50) ) What is the maximum neutron exposure per quarter for a radiation 1 worker 7 i I l l l I l l QUESTION 4.16 (1.00) j Whenever access to an Extra High Radiation Area cannot be controlled j by a lock or guard, what must be'done? j l ) J 1 QUESTION 4.17 (1.00) In accordance with OP/0/A/6400/06C, Nuclear Service Water System, Enclosure 4.1, what is used to supply the RN pump seals when'no RN ] pumps are running? ! l J i' QUESTION 4.18 (1.00) If a fire should occur in the Auxiliary Building Exhaust Filter train, what must be done by the operator? i 1 QUESTION 4.19 (1.00) In accordance with OP/0/A/6450/05, " Instrument Air System", which one of the following represents the starting limits for the air compressor motors. a). 1 start in a 1 hour period ! b). 2 starts in a 1 hour period c). 3 starts in a 1 hour period d). 4 starts in a 1 hour period 4 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ -. ___ __ . _ _ _ _ . _ . _ _ _ _ _a
I . Il 4 ,i PROCEDURES - NORMAL'. ABNORMAL. EMERGENCY Pcgo 33 i AND RADIOLOGICAL CONTROL GUESTION 4.20 (1.00) l A reactor coolant pump should not be started unless the # 1 Seal delta i P is greater than , 1 a). 100 psid b). 200 psid-1 c). 300 psid d). 400 psid 1 a GUESTION 4.21 (1.00) 1 In accordance with OP/1/A/6100/01, " Controlling Procedure for Reactor ! Startup", during heatup in Mode 3, the delta T between the PZR and the 1 NC loops should be maintained approximately 100 degrees F. What is the basis for this limitation? ) i l QUESTION 4.22 (1.50) j l In accordance with OP/1/A/6100/01, " Controlling Procedure for Startup, i what two (2) conditions are verified prior to transferring to the main I feed nozzles?
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j GUEST 10N 4.23 (2.00) 1 In the event of a loss of control room emergency requiring immediate i evacuation, what eight (8) actions are taken before evacuation of the control room? 1 2 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) i t I
4., PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 34 AND RADIOLOGICAL CONTROL QUESTION 4.24 (1.50) In accordance with EP/1/A/5000/01, " Reactor Trip or Safety Injection", what 2 Criteria are used following a safety injection to trip the NC pumps? QUESTION 4.25 (1.00) In accordance with OP/1/A/5000/01, " Reactor Trip or Safety Injection", as an immediate action, there is a step that verifies the Turbine has tripped. How is it verified (by the procedure) that the turbine has tripped? QUESTION 4.26 (1.00) What Radiation Monitor alarming would indicate a LOCA outside containment? QUESTION 4.27 (1.50) In accordance with Technical Specifications what are the six (6) factors considered when performing a Shutdown Margin calculation? (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********)
G I DS E NR IO BR MR OE C e, a , L A , i N . G I '. S R O R R E D DE D OE E RP N S I B M O C 1 s .. v D ' E n. E P - S i. D O R i.
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i. i.
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!i!llilll[lliii!Illi;I, L ;ij!! l((tL
p - rsge'13 of'25 i R:typs # 13 CONTROLLING PROCEDURE FOR UNIT STARTUP ft OP/1/A/6100/01-UNIT STARTUP , ENCLOSURE 4.1 1
.; Date 1 Time / Initials i 2.61 With NC pressure between 1700 and 1800 psig,: verify that remote PZR Pressure Indication on the Aux
)
R Shutdown Panels is operable per T.S. 3.3.3.5'(Remote !
' Shutdown System) and document in PT/1/A/4600/16 '
.(Surveillance Requirements For Unit Startup).
CAUTION: Do not exceed 1955 psig NC pressure prior to'S/G j pressure reaching > 725 psig. This will cause a- ; safety injection on Lo Steam Pressure. 2.62 At 1955 psig (P-11) NC. pressure, verify.the following l signals: : 4 "PZR LO PRESS S/I TRAIN A -(B) BLOCKED" status light goes out. , j "STM LINE S/I TRAIN A (B) BLOCKED" status light goes out. l'
+
i 2.63 When NC pressure reaches 2235 psig, ensure that the following controllers are in "AUT0": , l PZR Mtr Ctrl 1A (B) (D) ' PZR Pressure Master PZR Spray Controllers 1 INC-27 (PZR Spray Ctrl Frm Loop A) INC-29 (PZR Spray Ctrl Frm Loop B) PORV's 1NC-32B (PZR PORV? 1NC-34A (PZR P0h.) INC 36B (PZR PORV) 1 Pressurizer Level Master PD Pump Speed Ctrl or INV-294 (NV Pmps A & B Disch Flow Ctrl) 2.64 If the NC has been opened for refueling or maintenance (ex. , seal replacement, PORV repair, pressure boundary repair, etc.) perform PT/1/A/4150/01A (Reactor Coolant System Leak Test). s
a p b M Y a s/t Cycle Efficiency o (Net Work out)/(Energy in) s a Y,t + 1/2 (et2 ) W = Mg/ge .
- E = Mc2 ' a = (Yr - Y,)/t Scarnot = T4 - Tg _7 g-KE = MY2 /2 Vf = Y, + At T 4
PE = Ngh v = e/t P = ggh/ge A = (np2 )f4 -i Wk = PAY M = Y,yA 9 turbine = hj-h2r'= vkt-2r
' AE = 931( AM) 9 = 1/v h-h wk 3 21 1 - 21 4 = A&h 2
C = ACpAT 921+Yt + P jv; + u j + q 1 - 2 " 922+Y2+PY2 2 + v2 + vk't 6 = UA AT F @ J Q -@ J~ J f# Cy = Au/AT
, u t + q1-2 = U2 + vk 1- 2 I"I'Po a
Cp = Ah/AT Wk = hy -h U; + Q1-2 = U2 + Wk t- 2 '" -x/TYL TYL = 1.3/5 l J
. 1 Wk = F1(hg -h2 ) h = hr + % hrg HYL = -0.693/N Wkt -2 = P dv h = u + Pv/J l1 g J 2 I gd,2 al d 22 Y )
$13=PY22 R/hr = (0.55CiE)/d (meters) 2 '
R/hr = 6CiE/d2 (feet) A=AN P = P,10 ~ A = Ag e-At P = P, et /T A = 0.693/t SUR = 26.06/T SUR = 26( Ak/k)/i'+(3,77-( Ak/k))T 1/2 t t/2 eff = ((t , 1/2)-(tb ))/((t1/2) + (tb )) E=6N T = (L*/AK/K) + ((),rr - ( AK/K))/((kyr-( AK/K) + (d( AK/K)/dt))
, , ( AK/K) = (Kerr - 1)/ Kerr A = c/T ( AK/K) - (1*/ T) + (5,fr/(1 +X erg T))
N = 6.02 x 1023,p i, , o. 4 A
**# " EE RRD = EI = N 6 l
'.If="0.05 seconds-1 (down pover .e n.-g.; , '
P = (Elv)/(3 x 1010) = 0.0125 seconds-I (screm) ) -
= 0.08 seconds ~ 1 (critical)
N = $ -( 1 - K,rt )/ ( 1 -K,fr) N = No-(K,77)" M = 1/(1-K,yf) = CR3/CR, CRx = S/(1-Keryx) M = (1-K,fr,)/(1-K,frj ) CR;(1-K,rr,) = CR2(I~ Keff2) SDM = (1-K,fr )/Kerr D
r_ <
, Water Peremeters Miscel'1eneous Conversions _ rj i
1 gel. = 8.345 lb m 10 I watt = 3.12 x 10 fissions /second 1 gel. = 3.78 liters 1 curie '= 3.7 x t o J o dps 3 1 f1 = 7.48 gel. 1 kg = 2.21 lb m- i Density = 62.4 lbm/ft I hp =.2.54 x 103 Stu/hr. 6 Density = 1 gram /cm3. It1W = 3.413 x 10 Stu/hr Heat of Vaporization = 970 Btu /lb 1 inch = 2.54 cm m Heat of Fusion = 144 Btu /lb m *F = 9/5'C + 32 ; 1 ft. H2O = 0.4335 lb / inch 'C = 5/9(*F - 32) f Cp = 1 Stu/lb m *F 'R = 'F + 460
.P_hysics, Thermo, end Heet Transfer 1 Stu = 778 f t-Ib g I
~0 d = 0.172 x 10 Btu /hr-f t *R e = 2.718
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mp= 1.67264 x IO gram I ev = 1.60219 x 10 Coulomb-Volt - mn= 1.67495 x 10
~
gram 1 Coulomb-Volt = 1.60219 x l'O
~I9 Joule
-28 1 m, = 9.10956 x 10 gram 10 c = 2.99 x 10 cm/sec Op = 1.6 x 10"I9 Coulombs .
-34 h = 6.626 x 10 Joutes-sec
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Q, =(-) 1.6 x 10 Coulombs . N-m2/kg
~II '
G = 6.67 x IO K = 9.0 x 10 N-m2 /C , u
, l Therto. Forcula Shoot. y _;s p . 1,,,_,. g _.y,a p, gc 21 + V1,82ge p =+ A ge Z + 2ge +p
& _Vaz pg,1 jg y,a p,,,
scJ + 2gcJ + J + ua + q1_ = gcJ + 2gcJ + J + un + wkt.: p= 1 PE = MZg v ge hf g g KE = 1/2 MV8 6 - b2R bt
- WK[* y, _ hg f h=u+ J P = phi ga_: = J,2.Tds se wkt_: = f,, Pdv
,Qs _: = aCp (AT)
. . ,47 Pu = RT NPSH = (Pdy,- F,) gc.
PV = MRT psg l-1-" - 7,, _ 8in - sur _ wha Q = a(ah) h/N h/M
~
4 = UsaT) m T*'" - T~r_
"Dw Vater Parameters Conversions Cp = 1 BTU lba 0F _ 1 hp = 2.54 x 108 BTU he p = 62.4 lba 1 MW = 3.41 x 105 BTU ft' hr 1 gal. = 8.3 lba 1 inch = 2.54 ca.
1 gal. = 3.78 1 1 BTU = 778 ft -lbf 1 ft* = 7.48 gal of = 9/50C + 32 - ot = F + 460 ot = oC + 273 0273o 4
e' * *
, 6 t.
O Y k 1 4 f 1 Y i Eyk st/ G W' .f I 1 d s 1 i 1
)
i l 1 I I j i i,n, 1 l
)
1
1, ' PRINCIPLES OF NUCLEAR POWER PLANT CPERATION, Pcgn 35
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW L s; _
$0 {& A0lNb NSf ANSWER 1.01 (1.00) i (a) or (b)
REFERENCE OP-CN-RB-HO p. 7 LPRO OP-CN-RB-HO #3 1 3.3/3.6 OO1010K535 ..(KA's) i ANSWER 1.02 (1.00) (a) REFERENCE l OP-CN-HO-RP p.17 j LPRO OP-CN-HO-RP #12 l 3.5/3.7 OO1000K509 ..(KA's) 3 f ANSWER 1.03 (1.00) l l (d) REFERENCE ( Westinghouse Reactor Physics, Section I-5, MTC and Power Defect ; LPRO OP-CN-HO-RP #8 i 3.1/3.1 192OO8K114 ..(KA's) ANSWER 1.04 (1.00) (a) I (***** CATEGORY 1 CONTINUED ON NEXT PAGE ****A) I
I5,* PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Page 36 i C~ THERMODYNAMICS, HEAT TRANSFER AND FLUID' FLOW i , REFERENCE 1 HBR RXTH-HO-1 Session 39 LOH2 i OP-CN-HO-RP p.12 l LPRO OP-CN-HO-RP #8 -J 3.4/3.4 192OO6K107 ..(KA's)
)
ANSWER 1.05 (1.00) ; (a) i REFERENCE HBR RXTH-HO-1 Session 32 LO2 Fig. CN-RT-RCO-7; OP-CN-HO-RCO p.12 LPRO OP-CN-RT-RCO #4d 3.1/3.1 192OO4K108 ..(KA's) l ANSWER 1.06 (1.00) (c) REFERENCE HBR RXTH-HO-1, Session 26, p 2, LO1 OP-CN-HO-RCO pp. 10,11 LPRO OP-CN-RT-RCO #4A 3.1/3.1 192OO4K106 ..(KA's) ANSWER 1.07 (1.00) (a) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
7 LI. PRINCIPLES OF NUCLEAR POWER PLANT ~ OPERATION. -Paga 37 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW REFERENCE F l HBR RXTH, session 50, p. 2, LO #2 ! OP-CN-RB-HO pp. 6,7 LPRO OP-CN-RB-HO #4B 3.8/3.9 192OO2K114 ..(KA's) l; ANSWER 1.08 (1.00) (d) REFERENCE l OP-CN-CTH-CTH P. 12 l LPRO OP-CN-CTH-CTH #7 2.8/3.1 193OO8K103 ..(KA's) l l ANSWER 1.09 (1.00) (c) i REFERENCE Nuclear Energy Training, Reactor Operations, NUS Corp. paragraph 12.4 1 OP-CN-HO-RK p.12 j LPRO OP-CN-HO-RK #3e 2.8/3.1 OO1000K554 ..(KA's) ] l ANSWER 1.10 (1.00) (a) 1 J REFERENCE l HBR RXTH-HO-1 Session 33, p.2 LO2 ; OP-CN-HO-RP p.6
- LPRO OP-CN-HO-RP #3 !
l l l 3.0/3.2 192OO7K105 ..(KA's) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) I l i l 4
'1.' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pzgs 30 ;!
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW I i l l I
]
l- ANSWER 1.11 (1.00) j (C) J REFERENCE OP-CN-THF-FF p.10 LPRO OP-CN-THF-FF #9 2.5/2.7 191004K112 ..(KA's) i ANSWER 1.12 (1.00). ) 1 (c) ' REFERENCE I OP-CN-HO-HT p.7,8 ! LPRO OP-CN-THF-FF #4b ] 4.2/4.2 193OO8K122 ..(KA's) j ANSWER 1.13 (1.00) (c) REFERENCE i Mollier diagram or steam tables ; LPRO OP-GA-THF-STM #2e 1 3.3/3.4 193OO3K126 ..(KA's) , l 1 ANSWER 1.14 (1.00) l (d) I l i i (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l
- - - _ _ _ _ _ _ _ - _ _ _ 1
R-Y$y,- - . - 41', ' PRINCIPLES OF NUCLEAR' POWER PLANT OPERATION. .Pego ' 39 ' THERMODYNAMICS. HEAT TRANSFER AND-FLUID FLOW ,a,
.9 REFERENCE CN-OP-HO-RCO p.12 LPRO OP-CN-RT-RCO #3 3.4/3.8 ..(KA's)
ANSWER 1.15 (1.00) (c) REFERENCE OP-CN-DG-DG3 p.15 LPRO OP-CN-DG-DG3 #7 2.8/2.9 062OOOA403 ..(KA's) ANSWER 1.16 (1.00) (c) REFERENCE OP-CN-DG-DG3 p.12 LPRO OP-CN-DG-DG3 #8 2.8/3.2 062OOOA215 ..(KA's) ANSWER 1.17 (1.00) (a) REFERENCE OP-CN-HO-RP p.11 LPRO OP-CN-HO-RP #8 3.1/3.1 192OO6K111 ..(KA's) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
w - 6 .- i NI.*~ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pcgo 40 L7 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWER 1.18 (2.50)
- 1) Lower
- 2) Higher
- 3) Lower
- 4) Higher
- 5) Same REFERENCE General Physics HT&FF Part B, Chapter 1, pp.129-162 '
OP-CN-HO-FF p. 15 ; LPRO OP-CN-TA-AT #8a 3.4/3.7 OO2OOOA105 ..(KA's) ANSWER 1.19 (1.00)
- 1. Decrease (o.2r each )
- 2. Decrease
- 3. Increase
- 4. Increase REFERENCE General Physics, Rx PWR Limits p.243 General Physics, Boiling Heat Transfer p.122 OP-CN-HO-HT p.11 LPRO OP-CN-THF-HT #6e 3.4/3.6 193OO8K105 ..(KA's) l l
1 ANSWER 1.20 (1.50) j
- 1. LOWER (0.5 each) i
- 2. HIGHER ,
- 3. LOWER l
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
p 15 ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. ' Pegs 41 E THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ie REFERENCE OP-CN-HO-SLC pp. 15-18 LPRO OP-CN-THF-SLC #6,7 3.5/3.8 015000A101 ..(KA's) ANSWER 1.21 (1.50)
- 1. FALSE
- 2. TRUE
- 3. TRUE REFERENCE al OP-CN-HO-NT pp.7,8 LPRO OP-CN-HO-NT #6a 010000A106 3.1/3.2 045000A304 3.4/3.6 4.2/4.2 035010A202 ..(KA's)
ANSWER 1.22 (1.50)
- 1. TRUE
- 2. TRUE ,
- 3. FALSE '
REFERENCE OP-CN-HO-HT LPRO OP-CN-THF-HT #4d 3.9/4.1 193OO8K123 ..(KA's) ANSWER 1.23 (1.50)
- 1. b '
- 2. c
- 3. c
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
l1.* PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Page 42 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
' REFERENCE OP-CN-HO-HT p.3,4 LPRO OP-CN-THF-HT #1A 2.5/2.5 193OO7K101 ..(KA's)
ANSWER 1.24 (1.00)
- a. Iarger (0.5)
- b. longer (0.5)
REFERENCE OP-CN-HO-SCM p.18 LPRO OP-CN-HO-SCM #9 3.9/4.0 192OO8K103 ..(KA's) ANSWER 1.25 (1.50)
-break size (and geometry) (0.5 each)
-Iacation of break ( 3 ef 4 rey / red)
-status of ECCS components
-NC p.. mp re u; n opw Ls ce no+
REFERENCE OP-CN-HO-AT p.1 LPRO OP-CN-TA-AT #1 3.8/4.3 OOOO9EA206 ..(KA's) ANSWER 1.26 (1.00)
- a. 36 sec. (+/- 1sec) (0.5)
- b. 21 sec. (+/- 1sec) (0.5)
REFERENCE OP-CN-HO-RK p.8 LPRO OP-CN-HO-RK #1d 2.5/2.5 192OO7K101 ..(KA's) (***** END OF CATEGORY 1 *****)
N '- 1 f;f.' 2, ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paga 43 C' SYSTEMS S { l J ANSWER 2.01 (1.00) c 1 i REFERENCE I l CNS LP, (ELECT), OP-CN-HO-EP,.p'25, LPRO LO 21. I J 062OOOA401 ..(KA's) 4 ll i' ANSWER 2.02 (1.00) b REFERENCE CNS LP, (PRI SYS), OP-CN-HO-ND, p 24,25, LPRO LO 6.B. j i OO5000A101 ..(KA's) i j l
' ANSWER 2.03 (1.50)
- a. Close
- b. Close
- c. As is
- d. To VCT
- e. Close
- f. To VCT (0.25 EA)
REFERENCE CNS LP, (PRI SYS), OP-CN-HO-NV, p 7-35, LPRO LO 2. OO4000K405 ..(KA's) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
.g ---
ij
- PLANT-DESIGN INCLUDING SAFETY AND EMERGENCY Page 44' iW . SYSTEMS si fi.f Y
ANSWER 2.04 (1.00)
- a. TRUE
- b. TRUE
, (0.5'EA) ;
REFERENCE CNS LP, (PRI SUP SYS), OP-CN-HO-VC,.p 7, ISS LO 3,4. ANSWER 2.05 (1.00)
- a. FALSE I
- b. TRUE REFERENCE CNS LP, (PRI SYS), OP-CN-HO-NCP, p 35, LPRO LO 6,7.
OO3OOOK614 ..(KA's) ANSWER 2.06 (1.50) a.3 b.1 c.2 (0.5 each) REFERENCE CNS LP, OP-CN-HO-PC, p 13, LPRO LO 14; OP-CN-HO-SC, p 4,.LPRP LO 3. j OO4000K402 OO4000K401 ..(KA'r) 1 1 I e l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) , I I
_ .~. ._
- 2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 45 SYSTEMS ANSWER 2.07 (1.50)
- 1. Engine overspeed
- 2. Generator differential
- 3. Maintenance mode initiate
- 4. STOP/RUN b'utton placed in'stop position
- 5. Governor load limit set to zero
- 6. 51V voltage controlled over current (trips D/G and breaker)
(0.25 EA) i REFERENCE l ( CNS LP, (GENERATOR), OP-CN-HO-DG3, p 8, LPRO LO 4,5. I f 064000K402 ..(KA's) l l 1 ANSWER 2.08 (1.00)
- 1. The source and intake Section l 2. RN pumphouse section j 3. The main supply section
- 4. The heat exchanger section 1 5. The main return section 1
! (0.2 EA) l REFERENCE j i CNS LP, (PRI SUP SYS), OP-CN-HO-RN, p 9, LPRO LO 3. i l 076000 GOO 4 ..(KA's) l l l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) m--._._._-__.__m -
2.. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pege 46 SYSTEMS i I ANSWER 2.09 (2.00) I (4 of 5 required)
- a. All penetrations required to be closed during accident conditions are either: )
l
- 1) Capable of being closed by an' operable Containment automatic 1 isolation valve system, or
- 2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided by T.S. I
- b. All equipment latches are closed and are sealed.
- c. The sealing mechanism associated withreach penetration is operable.
- d. Each Air Lock is in compliance with T.S. 3.6.1.3
- e. Containment Leakage rates are within limits of T.S. 3.6.1.2.
(0.5 each) i REFERENCE CNS LP, (CONTAINMENT) OP-CN-HO-CNT, p 15. 103OOOK102 ..(KA's) i i ANSWER 2.10 (1.00) l
- 1. Head cooling bypass flow (0.20), 4 "+/- 0.5"% (0.05) l i
- 2. Nozzle bypass flow (0.20), 1 "+/- 0.5"% (0.05) i
- 3. Control rod & instrument thimble bypass flow (0.20), 2 "+/- 0.5"% I (0.05)
- 4. Baffle bypass flow (0.20), 0.5 "+ 0.5 - 0"% (0.05)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l l
2.' PLANT DESIGN INCLUDING ~ SAFETY AND EMERGENCY Pags 47 SYSTEMS !
' REFERENCE F
L .CNS LP,,(PRI SYS), OP-CN-HO-RVI, p 15, LPRO LO 3.A. L . I OO2OOOA105 ..(KA's) l l
]
i 1 ANSWER 2.11 (2.00) (4 of 6 required) j j
- a. No pressure boundary leakage shall exist. ]
- b. Maximum i GPM unidentified leakage. j l
- c. 500 GPD total reactor-to-secondary leakage through any one steam l
generator (S/G) and 1GPM total reactor-to-secondary : leakage through all S/G. 1
- d. 40 GPM CONTROLLED Leakage at a NC system pressure of 2235 +/-'20 Psig- 'I
- e. 10 GPM identified leakage in the NCS.
]
- f. 1 GPM leakage at an NCS pressure of. 2235 */- 20 psig from any NCS j pressure isolation valve specified in T.S. table 3.4-1.
(0.5 each) REFERENCE CNS LP, (PRI SYS), OP-CN-HO-NCP, p 35, LPRO LO 8. OO2OOOK405 ..(KA's) ANSWER 2.12 (1.50)
- 1) Loading or unloading S% of full power per minute. _l
- 2) Step load changes of +/- 10% of full power with automatic reactor-control.
- 3) Step load reduction of 95% of full power with automatic reactor control and steam dump.
(0.5 EA) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) i j
-- _-- _o
2.' PLANT DESIGN INCLUDING' SAFETY AND EMERGENCY Paga 48 i SYSTEMS REFERENCE [) CNS LP, (PRI SYS), OP-CN-IPE-HO, p 26, LPRO LO 8.a. { 010000K602 ..(KA's) l ANSWER 2.13- (1.00) ; I l'. Safety injection .
- 2. P-14 (Hi Hi S/G 1evel) l
- 3. Low Tave coincident with' reactor trip l
- 4. Manual j (0.25 EA) , l J
REFERENCE i CNS LP, (ECCS), OP-CN-HO-ISE, p 17, ISS LO 4.d. 059000A306 ..(KA's) j 1 ANSWER 2.14 (1.75)
- 1. Hi Hi containment pressure (0.25), 2/4 channels (0.25).
- 2. Lo steam line pressure (0.25) from 2/3 channels (0.25) on any one S/G.
- 3. Hi steam pressure rate of decrease (0.25) from 2/3 steam pressure channels (0.25) on any one steam line.
- 4. Manual (0.25) I i
l REFERENCE CNS LP, (ECCS), OP-CN-HO-ISE, p 18, ISS LO 4C. 039000K405 ..(KA's) ANSWER 2.15 (1.50) CA CST, UST, Hotwell, RN, RC (0.20 each, 0.05 for each in correct order) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
2'.' PL'AN'T DESIGN INCLUDI'GN SAFETY AND EMERGENCY Paga 49 ( SYSTEMS -l J
~
{ REFERENCE f CNS LP, (COND & FEED SYS), OP-CN-HO-CA, p 9, ISS LO 2. 1 061000K401 ..(KA's) j ANSWER 2.16 (1.50) f6 ) (3 of 6 equired) Resets demand position counters. # A (1) ] (2) Resets master cycler (3) slave cyc1'er andhbank overlap unit to "zero". 1 (5) Resets ALL alarms associated with rod control except the Urgent l Failure Alarm. l (6) Resets P/A converter. (0.5 each) REFERENCE CNS LP, (I&C), OP-CN-HO-IRE, p 18. OO1010K604 OO1010K402 ..(KA's) ANSWER 2.17 (1.50) The bypass is only opened if either pump bearing temperature (0.25) or seal leakoff temperature approaches alarm setpoint (170 & 200 deg F respectively) (0.25) and all the following are met: l 1) NC Press > 100 psi but < 1000 Psi (0.25) l l 2) #1 seal leakoff valve is opened (0.25) l l 3) #1 seal leakoff flowrate < 1 gpm (0.25) . l ) 4) Seal injection flow to each pump > 6 gpm (0.25) l l l l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
- 2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pege 50-SYSTEMS i i
REFERENCE 2 CNS-LP, (PRI SYS), OP-CN-HO-NCP, p 19. ISS LO 2, 2.A. OO3OOOK103 ..(KA's) ! l ANSWER 2.18 (1.00) 2 min REFERENCE CNS LP, (GENERATOR), OP-CN-NO-DG3, p 13, CNS memo dated 04/23/85. , 064000G010 ..(KA's) l ANSWER 2.19 (1.00) d It ensures suf ficient cooling capacity available for saf ety related k equipment (0.75) during normal and accident conditions (0.25). l REFERENCE i CNS LP, (PRI SUP SYS), OP-CN-HO-KC, p 11, LPRO LO 8. J i OO8010KOO6 ..(KA's) ! l l ANSWER 2.20 (1.00) .; l The low steam line pressure safety injection bistable is a rate 1 sensitive bistable (0.5) and can cause safety injection at pressures well above P-11 if steam line pressure is dropped rapidly. (0.5) l REFERENCE l CNS LP, (STM SYS), OP-CN-HO-SM, MAIN STM HO p 5, LPRO LO 9. 035010K601 ..(KA's) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l ) w - - - - - - - - a
I j 2." PLANT DESIGN INCLUDING SAFETY AND EMERGENCY' Page 51 ) SYSTEMS l I
~
l l ANSWER 2.21 (1.00) ) This is done to allow better mixing of NC and pressurizer water (0.5) to keep the boron concentration within T.S. limits (of 50 ppm). (0.5). l l i REFERENCE CNS LP, (PRI SYS), OP-CN-IPE-HO, p 7, LPRO LO 9.a. 010000K603 ..(KA's) l
'I ANSWER 2.22 (1.00) i l i
The fourth accumulator is assumed to dump out through the cold leg-break-i and bypass the core. (1,0) 4 l l l REFERENCE I l CNS LP, (ECCS), OP-CN-HO-CLA, p 8, LPRO LO 6. 1 OO6000K602 ..(KA's) l ! (***** END OF CATEGORY 2 *****)
. _ - --_-_A
.3. INSTRUMENTS AND CONTROLS Page s- 52 -ANSWER ~3.01 (1.00)
? a. REFERENCE CNS LP, (I&C), OP-CN-HO-ENB, p 16,'LPRO LO.6. 015000K604 ..(KA's) d ANSWER 3.02 (1.00)
^
' REFERENCE
- CNS LP, (COND & FEED SYS),'OP-CN-HO-CA, p 14, LPRO.LO 7..
059000K603 ..(KA's)
]
1 ANSWER 3.03 l (1.00) 3
'J l
]
REFERENCE 1 l , CNS LP, (SERVICE SYS), OP-CN-SS-IG, p 9, LPRO LO'6c. I ( l 011000A211 ..(KA's). 4
- ANSWER 3.04 (1.00) l i
- a. FALSE -i l b. TRUE (0.5 EA)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) I 1 l __a-_--.---._.__._-..--. - -
- 3. ' INSTRUMENTS AND CONTROLS 'Page 53 REFERENCE-CNS LP, (I&C), OP-CN-HO-ENC, p 5-9, LPRO LO 2,3.
015020K402 ..(KA's) ANSWER 3.05 (1.50)
- a. TRUE
- b. FALSE
- c. fab 6E raut (0.5 EA)
- REFERENCE CNS LP, (SERV SYS), OP A CN-HO-IG, p 5, LPRO LO 3.
l 002000K606 ..(KA's) l ANSWER 3.06 (1.75)
- a. 9 d. 4 g. 1
- b. 6 e. 2 i
- c. 3 f.'S (0.25 EA)
REFERENCE CNS LP, (I&C), OP-CN-HO-IPX, p 15,16, LPRO LO S. 1 L 012000K610 ..(KA's) l l l l l 1 , (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l l'
- 3. ,
INSTRUMENTS-AND CONTROLS- Page. 54 ANSWER 3.07 (IM[)
-2.00)
SIGNAL TRIP ;SETPOINT
- a. Manual Initiation (0.25) N.A.
'b. ^ a t a= = t'i ^-tuatic~ L.cgic and ".^ ' t hbbi
^r'u2 tier o- 2 ry c 'm 25'
- c. Containment Pressure-High-High (0.25) > 3 psig (0.25)
J
- d. Steam Line Pressure - Low (0.25)' < O 725.psig (x29)
(0.25) -
- e. Steam Line Pressure y '100 psi (0.25)
Negative Rate - High (0.25) REFERENCE CNS LP, T.S. TABLE 3.3-4 ITEM 4, (STEAM SYS), OP-CN-STM-SM, LPRO LO 8. 039000K405 ..(KA's) ANSWER 3.08 (1.50)
- 1. Speed error (0.5)
- 2. Load reference (0.5)
- 3. Bias signal voltage (0.5) or (S ign u. wn Am i n,6 oMn f4 s n 5)
REFERENCE CNS LP, (GENERATOR), OP-CN-HO-EHC, p 20, LPRO LO 7. 045000G004 ..(KA's.) l
- (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
1 b________-_________._--__._.-__-- . . _ - - - _ . _ _ . _ . - - - - - - - _ -- - - _ . - - . - - - - - - - - - -
< 3 .; INSTRUMENTS AND CONTROLS' Page: 55 N
' ANSWER '3.09 (1.50)
- a. Reactor Trip
- b. PORV's open
- c. PORV's closed
- d. Backup Heaters on
- e. P-11 Permissive for Safety Injection B' lock
- f. Safety Injection (0.25 EA)
REFERENCE CNS LP, (PRI SYS), OP-CN-IPE-HO, p 24, LPRO LO 4. i , i
' 011000K103 ..(KA's)
]
i l ANSWER 3.10 (2.00) f l 1 (4 of 5 required) ' 4
- 1. The IRNI reading displays too high a reading.'
- 2. P-6 may never deenergize.
- 3. The magnitude of the gamma' signal from the outer volume is greater than that from the inner volume. (#) l
- 4. With respect to reactor safety, overcompensation is worse than i undercompensation. (#)
- 5. The rate of power decrease (DPM) is reduced when passing through 3 the Intermediate Range. (#) ,
(#) Required (0.5 each) i REFERENCE CNS LP, (I&C), OP-CN-HO-ENB, p 10, LPRO LO B. 015000K502 ..(KA's) i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
- 3. INSTRUMENTS AND CONTROLS Page 56 ANSWER 3.11 (1.50)
- a. Performs a light test (should read all "8's"),
- b. Actuate SDM alarm,
- c. Causes the auto functions of that train to actuate if " ENABLE" is selected for that train.
(0.5 each) REFERENCE CNS LP, (I&C), OP-CN-HO-ENC, p 8, LPRO LD 4. 000008A121 ..(KA's) ANSWER 3.12 (1.00) (2.<G3 /7LQ)
- 1. The Slave Cycler generates current orders for the coils of the groups of rods to be moved.
- 2. The Slave Cycler compares the count number to the "In" or "Out" signal from the Supervisory Memory Circuit.
S '- A S A d rif f .f lutrAiw 5 ( 4 tif (p g ,4 REFERENCE C CNS LP, (I&C), OP-CN-HO-lRE, p 11,12, LPRO LO 3. 001000K403 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) I _ _ _ _ __ _ _ ----_-_ _- - - - - - _ - - - _ - --- - - - - - - - - - _ - - - - _ - - - - -_ - - - - - - - 2
3 ., , INSTRUMENTS AND CONTROLS Page 57 ANSWER 3.13 (1.50)
- a. 4/2/3
- b. 2/1/2
- c. 4/2/3
- d. 2/1/2
- e. 2/1/2 (0.3 EA)
REFERENCE CNS T.S., TABLE 3.3-1, p 3.'4 3-4., LPRO LO 2,10 of OP-CN-HO-IPX, p4 of 20. 012000A306 ..(KA's) ANSWER 3.14 (1.50)
- a. Unit 1 > or = to 8'.4% 2 4 0 5 (0,5)
Unit 2 > or = to 78.1%
- 0 f (0,5)
- b. NO (0.5)
REFERENCE CNS LP, (ECCS), OP-CN-HO-ISE, p :.7, LPRO LO 4. 013000K115 ..(KA's) ANSWER 3.15 (1.00) vj a -- v g .7 ,,; t e 3 3ct- g m. p7ay;ggg tg g, ,- _,.,_; ggty ; g _ ,7, . g
r. 'sc 8 @- c cc, t , O I c m , L er m m e mu t u l uud at the a, a n e ti=c.
b u _. . ) ( 1. 0 ) /3 e / # M#S /# M #' .^ 9%' ' ,.
- n f: - c' ;m. .'. '-
dmc. f# 0 isv rMu r.k S ea t n 6. t c c. ,Ncem,nt ti. 9 R C /* /2 6 y /:J/~ C cerME gKWH /~4 EMS [ N, (Mm r oad S r4:/ ^fmq A r r ar 5,,9 m E ?>mG, ! REFERENCE CNS LP, i (ELECT)', OP-CN-HO-EP, p d4, LPRO LO 2h Id* U isiturieu L a ra e rtz l i 062000K403 ..(KA's) j (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) ) a
- 3. ' INSTRUMENTS AND CONTROLS- Page' 58
- m. ,
;A'NSWER-. (l'.00')
3.16
- a. Safety Injection' signal ..( 0 , 5 )
- b. Containment High High Pressure signal (0 '5) ,
I REFERENCE
.CNS LP, (ECCS), OP-CN-HO-ISE, ENG SAFEGUARDS SIGNALS, CN-ECCS ISE-04, LPRO LO 4.
013000K101 ..(KA's) . ANSWER 3.17 (2.00) a.'2/4 low / low levels (0.25) in any-one S/G. (0.25)
- b. (Train related blackout) .2/3 undervoltage relays.(0.25).on the Essential bus, (0.25)
I
\
l
- c. Safety Injection signal (load group 7) (0.5) 'l i
, d. Loss of both main feedwater pumps .(0,5) 1 i REFERENCE 1 CNS LP, (COND & FEED SYS), OP-CN-HO-CA, p 10,11, .LPRO LO.4. 1
'I 061000K402 ..(KA's)
ANSWER 3.18 (1.50)
.1
- a. 1. Rods drive in enV Dar+g >&
l 2. Pressurizer Low( 0. Pressure 5 )(Aj,eg yv;t# t brtrip _(0.5) L* ^ r* L /AC nat?/J/lrta e k?~ j react '
/? + te t t. over ra in 4ar T vit t /?).n s r e Gew S in ce sq A ce[ S Y dA [O i
- b. No effect because highest readYng channel is used for' control. ( 0. 5 ) f 4 f wm**4 1
i REFERENCE CNS LP, (I&C), OP-CN-HO-IRX, DRAWING CN-IC-IRX-8, LPRO LO 7.
'015000A202 ..(KA's)
<l 1
(***** CATEGORY 3 CONTINUED ON NEXT-'PAGE *****) 1 l 1 1 o____ _i__ _ - . _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ . - - _ _ __ .-__ __ -. 9
3. INSTRUMENTS AND CONTROLS Page 59
. i ANSWER 3.19 (1.00)
Level channel I (0.5) failed high (0.5) 1 I REFERENCE l CNS LP, (PRI SYS), CN-CP-ILE-HO, PS ILE-4. [ 011000A210 ..(KA's) { l J ANSWER 3.20 (1.50) l 1
- a. BOTH RT (0.25) and BOTH BY breakers will trip (0.25).
1 1
+ho -1 l b. OCTll CY brra!:r : .1 1 ' trip (m 991 hvi+ nr weaker; ;il: c a ffi l
a l ac.cd (C.25). /ft c 4} SkA< W L C-- FAs0 & r *R " (*/2 " &Q
. *A's ned'Td")
- c. The RT "A" (0.25) and BY "V 8 (0.25) breakers will trip.
l l l REFERENCE CNS LP, (I&C), OP-CN-HO-IPX, p 9, L. PRO LO 7. 012000A307 ..(KA's) l l ANSWER 3.21 (1.00) i
- a. URGENT
- b. URGENT
- c. URGENT
- d. NONtfRGENT l
(0.25 EA) i l i a i I (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) i
3 INSTRUMENTS AND, CONTROLS Pa9". 60 8EFERENCE I CNS LP, (I&C), OP-CN-HO-IRE, p 12,.13, ARP-OP/1/B/6100/1DC, LPRO LO 11. ! 001000K403 ..(KA's)
.i i
i ANSWER 3.22 (1.50)
- a. ARM & DUMP
- b. ARM ONLY ,
- c. ARM ONLYor( $A m , ( Dc461,4). .
(0.5 EA)
.1 l~ REFERENCE 1
CNS-LP, (STM SYS), OP-CN-HO-IDE, p 8,9, LPRO LO 8. l: 041020K404 ..(KA's) - l { l ANSWER 3.23 (2.00) SEE ATTACHED SHEET (0.25 each) l REFERENCE CNS LP, (I&C), OP-CN-HO-IRX, DWG CN-IC-IRX-6,LPRO LO 3. 001000K403 ..(KA's) i 1 I 1 I l 1 i (***** END OF CATEGORY 3.*****) i l \ . l i L---__---------------_---z------ - - - - - - - - - - - - - - - - - -- - - ---- - - - ---- ----- =----------------- l
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- 8
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)
de e npe N DMI i s DE/ b OES RPP 2 1 . mnd oa SE 9 c T D n S N P f ( A U K oi to 8 D C ec A O d er L E 1 mtui ai d 4 D
/ 5 r
ggo oor r nd l D I O N Pme din D T I deor nm ME 2 R e e UED - pt yt MP I S N S Si rd e A IN DB N dl a l MO I ooi l R R.P.w I 2 4 3 u s D- - u t A
=
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)
- 4. ' PROCEDURES - NORMAL. ABNORMAL.' EMERGENCY Pcga 61 !
AND RADIOLOGICAL CONTROL I 1
)
i i
. ANSWER 4.01 (1.00) 1 1
The operator should return to the next step or substep in the left ) hand column. (1.00) I i REFERENCE i I Users Guide for ERG's and Background Documents p.5 OP-CN-HO-EP1, LPRO Training Objectives #3 , 194001A102 ..(KA's) ) l l ANSWER 4.02 (1.00) I At least one 4160 volt essential AC bus energized (1.00) REFERENCE , 1 l EP/1/A/5000/01, Reactor Trip or Safety injection ! l OP-CN-HO-EP1 p. 16 l l LPRO Training Objective #3 l l 194001A102 ..(KA's) a 1 i 1 ANSWER 4.03 (1.25) i l l 1. Ensur; Ocactor 'i p (0.25)tw aall3 trir n e. eenc /c r-
- 2. Verify Reactor Trippu(0.25) (Manus 11, trip neactcr) !
! 3. Verify Turbine Tripp&(0.25) l 4. VE, lii nC pae r availabic ( 0. 25 ) Ve r'Is 4llo o t ewen}Ia l race f $>> sses enerjenl l S. Check SI actuated (0.25) l REFERENCE EP/1/A/5000?O1, Reactor Trip or Safety Injection l OP-CN-HO-EP1, Training Objective #6B 194001A102 ..(KA's) l ANSWER 4.04 (1.00) l a). Any RED path takes priority over any other color or position (0.5) b). Continue in the Heat Sink RED Path (0.5) l l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) h > ---_ __ - -
1
- 4. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pega 62 l.* AND RADIOLOGICAL CONTROL l
. 1
. REFERENCE OP-CN-HO-CSF p. 5 LPRO Training Objective #2 194001A102 ..(KA's)
]
ANSWER 4.05 (1.25) 1
- 1. Notification of Unusual Event (0.25 each classification
- 2. Alert 0.25 for correct order) I
- 3. Site Area Emergency I
- 4. General Emergency l 1
REFERENCE l OP-CN-HD-SEP p.2 LPRO Training Objective #1 A 194001A116 ..(KA's) i ANSWER 4.06 (1.00) Allows time for the operator in establishing NCS feed and. bleed (0.5) by removing a heat source from the system (0.5) REFERENCE i EP/1/A/5000/2C1 p.2 i OP-CN-HO-CSF p.14 LPRO Training Objective #5 003OOOGOO1 ..(KA's) l ANSWER 4.07 (1.00) a). 1. Collapse by increasing NCS pressure (0.25 each)
- 2. Collapse by starting one NC pump
- 3. Vent to containment b). Vent to the containment i
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) L . _ _ _ _ _ _ _ _ _ _ _
t
'4; PROCEDURES - NORMAL. ABNORMAL' EMERGENCY . Pagn 63 ,
AND' RADIOLOGICAL CONTROL (a l 2 l REFERENCE'
]
EP/1/A/5000/2F3 p. 4,5,7
.OP-CN-HO-CSF. p. 26 LPRO Training Objectives #5 OOOOO9G012 ..(KA's) l ANSWER 4.08 (1.00) q a). 2 (0.5 each) b). 10 REFERENCE i d
EP/1/A/5000/03, Loss of All AC Power OP-CN-HO-EPS LPRO Training Objective #5 ) OOOOS6E302 ..(KA's) ) i i 1 ANSWER 4.09 (1.00) l a). 2 (0.25 each) b). 50 c). (1) 350 (2) 385 REFERENCE Controlling Procedure for Unit Operations OP/1/A/6100/03, p.1 Controlling Procedure for Unit Shutdown OP/1/A/6100/02, p.2 OO2OOOKO10 ..(KA's) i ANSWER 4.10 (1.00) d b a, c, b,,+ (1.0 for correct sequence, 0.75 for 3 correct, 0.5 for 2 correct) REFERENCE OP/1/A/6100/03, Controlling Procedure for Unit Operations 194001A102 ..(KA's) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) w _- _ -____- -_ -
=<
4.. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pagn 64' j AND RADIOLOGICAL CONTROL j , i L. a j l l ANSWER l 4.11 (1.00) i i d (1.00) l j REFERENCE i
)
OP/1/A/6100/02, p.1 'l 035000A010 ..(KA's) ANSWER 4.12 (1.00) These substeps (marked by " bullets") can be performed in any order (1.00) : 1 REFERENCE -O
)
OMP 1-4, Operations Management Procedure 1 194001A102 ..(KA's) { i ANSWER 4.13 (1.00) I Proceed to another location within the RCA to frisk. (1.00) REFERENCE J I CNS Directive 3.8.3, Decontamination Responsibilities, Enclosure 1. l 194001K103 ..(KA's) ANSWER 4.14 (1.50)
- 3 cc gop l)fe Ad~l<.. y~< bria f*H a). 2500 - c r^ ' q . r (0.5 each) b). 6 rem /qtr c). 15 rem /qtr REFERENCE CNS Directive 3.8.5, Exposure Extensions and/or Exposure limit reductions p.2 194001K103 ..(KA's)
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
- l. PhDCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 65 AND RADIOLOGICAL CONTROL e i i
1 ANSWER 4.15 (0.50) ] i Not to exceed 300 mrem /qtr I
. REFERENCE !
I CNS Directive 3.8.6, Radiation Exposure Control- p.3 194001K103 ..(KA's) J 1 l l ANSWER 4.16 (1.00) Area shall be conspicuously posted (0.5) and a flashing light shall be ) activated to warn personnel to stay clear of the area. (0.5) REFERENCE CNS Directive 3.8.8, Radiological Work Practices p.5 19400K103 ..(KA's) i ANSWER 4.17 (1.00) l l \ Portable unwatering pump (1.00) i j REFERENCE I i OP/0/A/6400/06C, Nuclear Service Water System p.1 j 076000 GOO 1 ..(KA's) I 1 J
\
i ANSWER 4.10 (1.00) J Manually, (0.5) open the deluge valve (located at the filter train) (0.5) l REFERENCE l OP/1/A/6450/03, Auxiliary Building Vent. System OOOO67 GOO 7 ..(KA's) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) e___---__
i 4. PMOCEDURES - NORMAL. ABNORMAL. EMERGENCY Pagp 66 q AND RADIOLOGICAL' CONTROL 1
)
ANSWER 4.19 (1.00) d (1.00) REFERENCE { OP/0/A/6450/05, Instrument Air System. 078000G010 ..(KA's) l ANSWER 4.20 (1.00) j i b (1.00) REFERENCE OP/1/A/6100/01, Enclosure 4.1, 2.17 ) OO3OOOK103 ..(KA's) l 1
)
ANSWER 4.21 (1.00) i i i Provides adequate subcooling (0.5) while minimizing PZR and spray j fluid delta T (0.5) ) REFERENCE OP/1/A/6100/01, Controlling Procedure for Reactor Startup 011000G010 ..(KA's) ANSWER 4.22 (1.50) ( 2 o 4" 3 recp/re./ ) a). Final feed temperature for each S/G > 250cF (0.75) b). S/G reverse purge flow has been in service for greater than or equal to 30 minutes. (0.75) c). CF Flow > 17 % (o,'75) l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
PIROCIDURES - NORMAL. ABNORMAL. EMERGENCY Pegn 67
*4.' AND RADIOLOGICAL CONTROL 1
I n i o l l REFERENCE OP/1/A/6100/01, Controlling Procedure for Startup, p.23 0-CP-CP-43 059000KOO7 ..(KA's) ANSWER 4.23 (2.00) (8 d 9 rep ec i /) a). Trip the reactor (0.25 each) b). Verify the turbine tripped ! c). Trip BOTH CFPT's d). Trip ALL NC pumps e). Check 1 ETA and 1ETB f). Ensure INV-10 Closed (Letdn Orif IB Otlt Cont. Isol) g). Ensure all CRDM fans running h). Announce over PA system i), Un1+ Sape r v:n c +., Le, eve, oATC a >>d b ispa k bec NCO 1. Asp + pccG<m Enc ic w<e ,i , REFERENCE AP/1/A/5500/17, Loss of Control Room, p.2 OOOO68K318 ..(KA's) ANSWER 4.24 (1.50) (2 ( 3 e,guire) } a). Subcooling less than or equal to OoF (0.75) b). At least one NV or NI pump indicating flow (0.75) c), VC Shep l 3 Ilde Flw h NCP 6cp Lu (ons) REFERENCE EP/1/A/5000/01, Reactor Trip or Safety Injection, p.9 OP-CN-HO-EP1 LPRO Training Objective #8 CNS Exam Bank-Emergency Procedures i 19400A102 ..(KA's) ANSWER 4.25 (1.00) l ALL main stop valves are closed. (1.00) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) 1 1 k_ _ _ __ _ _ _ _ _ . _ _ _ _ _ _ _
a i
.% e
- 4. PROCIDURES - NORMAL. ABNORMAL. EMERGENCY Pcge 68 AND RADIOLOGICAL CONTROL i
i REFERENCE OP/1/A/5000/01, Reactor Trip or Safety Injection p.3 OP-CN-HO-EP1, LPRO Training Objective #4 ; 194001A102 ..(KA's) i ANSWER 4.26 (1.00) i j High radiation on EMF 41 (Auxiliary Building Monitor) (1.00) (either j answer acceptable) l REFERENCE j i OP/1/A/5000/01, Reactor Trip or Safety Injection p.17 j 0-EP-EP-08, CNS Exam Bank j 073OOOGOO7 ..(KA's) ) ANSWER 4.27 (1.50)
- 1. RCS Baron concentration (0.25 each) ;
- 2. Control Rod position l
- 3. RCS average temperature ;
- 4. Fuel burnup based on gross thermal energy generation ]
- 5. Xenon concentration ;
- 6. Samarium concentration REFERENCE i
CNS Technical Specifications, 4.1.1.1.1 , p. 3/4 1-1 192OO2K114 ..(KA's) l l (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********) _ - _ _ _ _ _ _ _ _}}