ML20195D735
| ML20195D735 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/26/1988 |
| From: | Graves D, Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20195D732 | List: |
| References | |
| 50-313-OL-88-02, 50-313-OL-88-2, NUDOCS 8811070126 | |
| Download: ML20195D735 (46) | |
Text
a APPENDIX U. S. NUCLEAR REGULATORY C0411SSION i REGION IV o *cator Licensing Exam Report: 50-313/0L 88-01 Operating License: DPR-51
- tet No.: 50 713 t" 4' Aryfts.+s Power & Light Company P . (l , x 551 L.ctie cock, Arkansas 72203 Fa 2 k b cansas Nuclear One Unit 1 (AN01)
Exa. ,ien at. ~ .A01 Chief Examiner: 8//4 eud ah r D. N. Graves, Examiner, Da t~e Operator L':ensing Section, Division of Reactor Safety Approved by: h*v ,Yd/). L .
nbVn J. L. Pellet, Chief, DLte Operator Licensing Section Division of Reactor Safety Summary A senior reactor operator written examination was administered to one candidate by the NRC. The candidate passed the examination and has been issued the at.propriate license.
G811070126 881027 PDR ADOCK 05000313 V PDC
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DETAIL 5' i
- 1. Persons Examined SRO i
License Examinations: Pass - 1 Fail - 0 l
- 2. Examiner D. N. Graves Chief Examiner r
- 3. Examination Report ,
t Perforn:ance results for individual examinees are not included in this ,
report as it will be placed in the NRC Public Document Room and these results are not subject to public disclosure.
- a. Examination Review Comment / Resolution l In general, editorial comments or changes made during the examination, ,
or subsequent grading reviews are not addressed in this section.
This section reflects resoiution of substantive coments made by AN01 staff personnel. Further.-this section addresses only those coments ;
that were not fully incorpoiated into the master examination key '
which is included in this report. Coments have been paraphrased for '
brevity in this section. The full text of the facility coments is l attached.
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6.05.d Request deleting part d from the question and redistributing the remaining point value. Not using the noun name for the device could lead to confusion and lead to right/ wrong 5 answer for the wrong reason. !
- Response
- Not accepted. Adequate inforNtion is available in that i question such that the candidate should understand what is occurring.
L Accept any four conditions that cause the unit to go into 6.11.a l Track or cause a Runback. I Response: Not accepted. The conditions listed in the coment do not directly cause the described condition in the question. If f the conditions in the comment occurred, and the unit did k not go into Track or runback, the effect described in the 6
- question would not occur. f i ;
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- b. Master Examination and Answer Ke2 A copy of the final AN01 written examination and answer key is attached.
The facility licensee coments accepted have been incorporated into the answer key.
- c. Facility Examination Review Comments The facility review consnents regarding the AN01 written examination are attached.
I 4'\ j ATTACHMENT l U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAhINATION FACILITY: 6886U183_UUQLE68_QUE:1___
REACTOR TYPE: EWB:B&W1ZZ_______________
DATE ADMINSTERED: HQtQ222h_________________
EXAMINER: QB6VE$t_Qa_______________
CANDIDATE _________________________
INSIBUCIIQUS_ID.CauDIQ8IE1 Uae 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 ir parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at loest 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY .
__VeLUE- _IDI6L ___SCQBE___ _VeLUE__ ______________C8IEQ9BI_____________
_2540D__ _25 ADD ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,AND THERMODYNAMICS
_25ADQ__ _25t0Q ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_25200__ 25100 ___________ ________ 7. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 25100__ _25xDD ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
_100AD__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received sid.
Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. -Cheating on-tha examination means an automatic den?.a1 of your application ond could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avol'd all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the
. examination.
- 5. Fill in the date on the cover. sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each cection 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 i 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 shcets from pad and place finished answer sheets face down on your desk or table.
j 12. Use abbreviations only if they are commonly used in facility literature.
! 13. The point value for each question is indicated in perentheses 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. Parti l credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
I 16. If parts of the examination are not clear as to intent, ask qi'estions of d
the examiner only.
j 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 l completing the examination. This must be done after the examination has
{ been completed.
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4 18.'When you complete your, examination, you shall:
- o. Assemble your examination as follows:
(1). Exam' questions on top.
(2) Exam'sida - 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 dented or revoked.
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Ei__IBEQBY QE_NUCLEeB_EQWEB_EL8HI_QEEBeIIQNA Page 4 ELUIQSteND_IBEBUQQYNed101 i
QUESTION 5.01 (1.00)
The reactor is critical at 10E-8 amps early in core life. These conditions were achieved about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to peak xenon occurring following a ;
reactor trip from 75% power. Assuming that feedwater and steam pressure are controlled automatically and no operator actions nor reactor trip.
occurs, which one of the following correctly describes the behavior of the reactor over the next several hours?
- a. The reactor stays critical due to the effect of the Moderator [
Temperature Coefficient.
- b. The reactor will go subcritical in about one hour, and will remain subcritical.
- c. The reactor will go subcritical in about one hour, will return to criticality in another hour and remain at e.pproximately 10-8 amps.
- d. The reactor will go suberitical in about one hour, will return to i criticality in another hour and then increase power to the.POAH.
i QUESTION 5.02 (1.00)
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Which one of'the following reactivity coefficients would be the first to turn power back down following a steam break accident? ;
a.- Moderator Temperature-Coefficient ;
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- b. Pressure Coefficient -
c .- Doppler coefficient j- r
- d. Void Coefficient i
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S&__IHEQBl_QE_NUGLE68_EQWEB_ELeNI_QEEBellQNi Page 5 ELUIQSioNQ_IHEBdQQ1NodICS QUESTION 5.03 (2.00)
Indicate whether each of the following will cause the differential rod worth to INCREASE, DECREASE or have NO EFFECT. (2.0) '
- a. An adj acent rod is withdrawn. l b.- Moderator temperature is DECREASED.
- c. Boron concentration is INCREASED.
- d. A Burnable Poison Rod depletes.
QUESTION 5.04 (2.50)
- a. What are the two (2) major components of the power defect from 0 to 15% power? (1.0)
- b. What is the major contributor to the power defect from 15 to 100%
power? (0.5)
- c. How does the power coefficient vary (more or less negative, more or <
less positive) from BOL to EOL? (0.5)
- d. How does the power coefficient very (more or less negative, more or less positive) from 50% power to 100% power? (0.5) i QUESTION 5.05 (0.50)
For equivalent positive reactivity additions to a critical reactor low in the intermediate range, will the SUR be the (SAME, LARGER, or SMALLER) at ;
BOL .is compared to EOL? (Choose one - no explanation required) (0.5) 1 QUE S T IO.4 5.06 (2.00)
Roactor power is reduced from 75% to 50% by boration. Control rod positions are not changed. How would power imbalance be expected to change (negative or positive)? JUSTIFY your answer.
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o* 's Et__IBEQBl_QE_UVQLE88_EQWEB_EL6UI_QEEBoI1QUt Page 6 ELVIQSteUQ_IBEBdQQ188dIQS QUESTION 5.07 (1.00)
After an outage for refueling, the reactor is being heated with Teve at 100 deg F and the source range indicating 30 cps. In a few hours the reactor is at 532 des F Tave. There has been no change in rod position or boron concentration. How would you expect the Source Range indication to compare with the initial reading (higher, lower, or the same)? (1.0)
QUESTION 5.08 (1.00)
How many doublings of the Source Range indication should one expect prior s to reaching criticality?
- s. 5
- b. 7
- c. 9
- d. 10 QUESTION 5.09 (1.00)
The reactor is operating at 75% power steady state. Power is now increased to 100% and conditions are allowed to stabilize. What is the not reactivity in the core at the 100% power level and how does that compare to the net reactivity present at the 75% stable conditions? (1.0) ,
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O 15_ IBEQB1_QE_UUQLEeB_EQWEB_EL6HI_QEEBoIIQUt Page 7 ELVIQSteNQ_IBEBdQQ186dIQS QUESTION 5.10 (1.50)
For each of the following, indicate whether they will cauoe the power (1.5) csnge instrument to indicate HIGHER, LOWER or the SAME as actuel power, if ine instrument was adj usted to 100% based on a calculated heat balance:
- b. If the reactor coolant pump heat input used in the heat balance is OMITTED.
- c. If the feed flow used in the heat balance was HIGHER than actus1.
QUESTION 5.11 (1.00)
Which one of the following describes where the maximum fuel centerline temperature occurs in a core with 9 symmetric axial neutron flux about the core midplane?
- a. Top of the core
- b. Between the top and midplane of the core
- c. Core midplane
- d. Between midplane and bottom of the core
- e. Bottom of the core QUESTION 5.12 (1.00)
Which one of the following statements correctly describes the effect of adding Emergency Feedwater (EFW) during a Natural Circulation condition?
- n. It LOWERS the OTSG thermal center while INCREASING the strength of the heat sink.
- b. It LOWERS the OTSG thermal center while DECREA$1NG the strength of the heat sink.
- c. It RAISES the OTSG thermal center while INCREASING the strength of the heat sink.
- d. It RAISES the OTSG thermal center while DECREASING the strength of the heat sink.
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Et__IUEQBl_QE_UUCLE8B_EQWEB_EleUI_QEEB811QUt. Page 8 ELU10StauQ_IBEBdQQ1860101 QUESTION 5.13 (1.00)
Which one of the following correctly describes the behavior of RCS pres-sure, if a Small Break LOCA which was not large enough to actuate the ECCS, were to occur with no Feedwater available?
- a. Pressure initially decreases, then rapidly increases whan the OTSGs boil dry.
- b. Pressure decreases slowly until it levels off somewhere above ECCS actuation pressure.
- c. Pressure initially decreases slowly, then rapidly drops when the OTSGs are boiled dry.
- d. Pressure initially decreases, then when OTSGs boil dry, continues to
- decrease, but at a much slower rate.
QUESTION 5.14 (1.50)
With the Unit operating at 90% power with all control systems in automatic, a Turbine Bypass Valve fails full open. Indicate how the following parameters will change (INCREASE, DECREASE, REMAIN THE SAME) relative to their initial values when plant conditions stabilize:
- a. Tavg
- b. MWe
- c. Reactor power QUESTION 5.15 (1.50)
Indicate whether the following will INCREASE, DECREASE, or REMAIN THE SAME:
< b. Minimum required NPSH as volumetric flow rate increases,
- c. Available NPSH to condensate (hotwell) pumps as condenser subcooling increases.
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'It__IHEQBY_QE_NVQLE88_EQWEB EL6NI_QEEB811QNi Page 9 ,
ELUIQ1teNQ_IBEBdQQ1Ned1Q1' :
QUESTION 5.16 (2.00) {
Will the Departure from Nuclear Boiling ratio INCREASE, DECREASE, or [
REMAIN THE SAME if-the following plant parameters INCREASE during power !
operation? Consider each parameter independently. (2.0) l 1 !
- e. Reactor Coolant System (RCS) Pressure ;
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- b. RCS Temperature o
- c. RCS Flow 1 i
- d. Reactor Power [
I j QUESTION 5.17 (1.50) -
If the OTSG low level limit were decreased, indicate whether reactor power !
- would have to be HIGHER, LOWER or the SAME to maintain a TAVG of 579 ;
degrees. Justify your answer. (1.5) l
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I II QUESTION 5.18 (2.00) i i :
i While operating at 60% power, it is recognized that'a control rod has been (
j significantly miss11gned BELOW tts group average for several days. The rod [
is realigned, and a positive quadrant tilt develops in the quadrant of the 3 misaligned rod. (2.0) [
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- a. Why does a positive tilt exist, even though all rods are correctly ;
aligned? j i
- b. Assuming no operator action, how would this tilt change ovbr the next i ten hours? Explain your answer. !
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,s . t ht .EL6NI_111IEdl.QE11GNi_QQNIBQLt_600.IN11BudENI611QN Page 10 l
QUESTION 6.01 (2.50) l i
- a. If a control rod "in" signal is generated, and the rods move in the i "out" direction, what Diamond Control Panel lamp would illuminate AND j what automatic action would result in the rod control system from this l
- condition? (1.0) l l
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- b. What are three (3) conditions that will cause the Diamond Control Panel Auto Inhibit Lamp (amber color) to illuminate (Diamond Station i cannot be placed in automatic)? (1.5) T f
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- QUESTION 6.02 (1.00) !
Which one of the following is NOT a design difference between the APSRs and ,
the safety / regulating control rods and drives? (1.0) j
- s. APSR drives contain ball check bypass valves in the thermal barrier to (
I allow for in-flow of coolant during a trip. !
- b. APSR drives contain a small button on the lower portion of the segment ;
- arms which prevents the lead screw from being disengaged when power is {
f lost to the mechanism. ;
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- c. The female couplings of the APSRs are different from the female I j couplings of the safety / regulating rods to ensure one type of rod [
i cannot be attached to the other typw's leadscrew. i r
) d. Although a patch panel is available to reassign specific drives to !
control from a different power supply, the APSRs are permanently patched through the panel and cannot be reassigned.
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b 'e 6x _EL681.1111Edi_QE110Ut_CQUIBQLt_600_IUSIBudEUI6IIQU Page 11 QUESTION 6.03 (2.00)
- a. The auxiliary oil pump should be operated for what minimum length of time prior to starting or stopping a makeup pump (assume normal start ,
or stop)? (0.5)
- b. A makeup pump may be started, under extreme emergency veQuirements only, if the auxiliary oil pump is operated for what minimum length of i time? (0.5) i
- c. How long does the auxiliary oil pump need to be operated prior to the makeup pump starting due to an ES actuation? (0.5) i
- d. If a makeup pump fails to start after receiving an ES actuation signal, how much time should elapse before the operator is alerted, ;
via annunciator, that the pump failed to start? (0.5) i T
h QUESTION 6.04 (1.00)
Annunciator "RCP 1P32 SEAL WATER FLOW LO" alarms. What are two (2) methods of determining which RCP(s) has(have) the alarming condition? (1,0) 3 :
i QUESTION 6.05 (2.00) l For each condition below (a- d), select ALL of the numbered items (1 - 5) I that apply for that condition. (2.0) {
- a. ESF Actuation signal 1. Associated EDG starts i b. Undervoltage (UV) on 1A3 2. 1A3 bus is load stripped l c. ESF Actuation signal and (IV on 1A3 3. EDG breaker closes i d. 1A3 186 device actuates dse to a phase 4. Loads are automatically i l
overcurrent which in turn, causes an sequenced onto 1A3
- UV condition on 1A3 j 5. 1A3 reenergized, but (
loads must be manually !
restarted i
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O 6t__EL6UI_SYSIEd1_QESIGUt_CQUIBQLt_6UQ.IUSIBudEUIaIIQU Page 12 i QUESTION 6.06 (2.50)
- a. What provides the motive force for opening AND closing of the Main Steam Isolation Valves (MS!Vs)? (1.0)
- b. In which MSIV condition, open or closed, are all operating solenoids for the MS!V in the deenergized condition? (0.5)
- c. What is the function of the hydraulic system associated with each of the MSIVs? (1.0)
QUESTION 6.07 (2.00)
- a. What provides the normal and backup makeup sources of water for the Emergency Cooling Pond? (1.0)
- b. The Service Water System is operating with pumps A and B running.
Pump C is tagged out for repair. Indicate how the crosstie valves and loop supply / isolation valves will realign if an ES actuation signal is recefved. All crossties and loop supplies are open initially. (1.0)
QUESTION 6.08 (2.00)
- a. How many containment penetrations are utilized for supplying loads l with Intermediate Cooling Water (ICW) including what loads inside containment are supplied with ICW (3 major loads or groups of components required)? (1,0)
- b. How many containment penetrations are ICW outlet penetrations? !
IDENTIFY which lines combine or split. (1.0) ;
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O ht__EL6UI 111IEUS_QE1100t_CQUIBQLt_6UQ_1011BudEUI6IIQU Page 13 QUESTION 6.09 (2.00)
Match the plant area (a - h) with the type (s) of automatic fire protection (1 _ 5) available in that area. If more than one system is available in each area, list all that apply. (2.0)
- a. Control Room 1. deluge system
- b. EDG rooma 2. wet pipe sprinklers
- c. Reactor building 3. pre-action sprinklers
- d. Main turbine bearings 4. CO2
- e. Startup transformer 1 5. Halon
- h. Auxiliary control room QUESTION 6.10 (2.50)
- e. List the six (6) functions either imposed or bypassed when the Shutdown Bypass key switch is placed in BYPASS. INDICATE whether the function is inserted or bypassed. Setpoints NOT required. (1.5)
- b. What are two (2) conditions that require BOTH a reactor trip (from RPS) and actuation of the emergency feedwater system (from EFIC)?(1.0)
QUESTION 6.11 (2.00)
- a. What are four (4) conditions that will cause the unit master station to have both the manual and auto lights lit and the toggle switch not have any effect on the station? (1.0)
- b. The facility is operating at 90% power with the Reactor Demand Station in manual. One RCP trips simultaneously with one feedwater pump. !
What condition does this put the ICS into, AND what final power is the facility trying to obtain? (1.0) i
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6x._EleUI 11SIEdS_QES1 Qui-QQUIBQLt_eUQ 10SIBudENIeI1QU Page 14 QUESTION 6.12 (1.50)
- a. Once an engineered safeguards actuation signal has occurred, what must be done by the operator to take uanual control of a compnnent actuated by the ES signal? (0,5)
- b. Once manual control has been taken of a component, what are two (2) ways of removing the manual override? (1.0)
QUESTION 6.13 (2.00)
- b. Assuming a OTSG fill rate needs to be controlled, what parameter determines the fill rate for the OTSG? (0.5)
- c. What will be the status of EFW (available or not) to OTSG A and B if OTSG A is at 550 psig ar.d OTSG B is at 700 psig? (0,5)
- d. What will be 6.he status of EFW (available or not) to OTSG A and B if OTSG A is at 550 psig and OTSG B is at 500 psig? *0.5)
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b *e Zt__EBQQEQuBE1_ _UQBd8Lt_8BUQBdBLt_EdEBGEUCI Page 15 8UQ_B8010LQQ1Q8L_CQUIBQL QUESTION 7.01 (3.00)
Answer the following with regard to Procedure 1102.04, "Power Operations":
- a. What is the licensed power level for AN01? (0.5) I
- b. What is the maximum everage power allowed for en eight hour period and what is the maximum peak power allowed per 1102.04? (1.0) l
- c. The limit for quadrant tilt depends on which system is being used for monitoring. Which core monitoring system provides the most ,
restrictive limit and why is it more restrictive than the other? (1.0) <
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- d. What is the maximum time allowed for operation outside the power '
imbalance envelope? (0.5) i i
QUESTION 7.02 (3.00)
During a reactor trip recovery, per 1102.06, the Shift Supervisor maken !
the determination of the authorization required to restart by evaluating seven (7) criteria.
- a. If all criteria are answered "no", who (by j ob title) must authorize ,.
the reactor restart (2 titles required for full aredit)? (1.0)
- b. If one or more criteria are answered "yes", who (by job title) must authorize the reactor restart, IN A00! TION to those in part "a"? (0.5) l
- c. List three (3) of the seven (7) crite s mentioned above. (1.5) l t
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Zt__EBQQEQUBEl_:_NQBd6Lt_8ENQBd6L,._EdEBQENQ1 Page 16 8NQ_B8DIQLQQlG6L_QQUIBQL 1
QUESTION 7.03 (2.50)
- a. Per 1102.08, "Approach to Criticality", if a Count Doubling does NOT occur prior to 100% withdrawal of all safety rods (for this cese, the !
exceptions of Attachment A of 1102.08 are not applicable), what two i (2) actions should be taken prior to calling Nuclear Support? (1.0)
- b. If SR/IR overlap is determined to be < 1 decade during the power escalation, what action should be taken? (0.5) ;
- c. The reactor is critical during a startup with power at 5X 10E-11 amps when both SR instruments indicate downscale. What action should the operator take? (1.0)
QUESTION 7.04 (2.00)
- a. What conditions regarding control rod position must be met in order to perform a boron dilution on the RCS? (1.0)
- b. A boration has been calculated and initiated (batch method). What must be performed by the operator within 30 seconds after CV-124g is closed by the batch controller? (0.5) l
l QUESTION 7.05 (2.50)
- s. Following a reactor trip (no SGTR), what criteria (subcooling margin r and time) are used for stopping all reactor coolant pumps (2 sets of conditions required for full credit)? (1.0) L
- b. What negative consequence may result from leaving the RCPs running too long with a loss of subcooling margin, and then stopping all four .
RCPs? (0.5) i
- c. List four (4) indications that may be used to determine a value for subcooled margin. (1.0)
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! QUESTION 7.06 (2.50) 1 What ere the E0P indications of a STATION BLACK 0UT and DEGRADE 0 POWER that would necessitate going to the appropriate tab in 1202.01? (2.5)
QUESTION 7.07 (2.50)
- a. State (YES or NO) if natural circulation and heat rej ection to the secondary system are indicated by each of the following l
conditions: (2.0)
- 1. RCS Hot to cold leg temperature differential stable at 40 F
- 2. Reactor coolant temperatures decreasing '
- 3. Turbine bypass valves open to control steam pressure ,
- b. If natural circulation flow is not established due to excessive voids, what should be attempted to try and start natural circulation flow? (0.5) I i
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QUESTION 7.08 (3.00) l l
- a. What action (s) should Unit 1 take if a fire in the Unit 2 control room I causes Unit 2 to perform an A". ternate Shutdown? (1.0) l
- b. When the decision is made to evacuate the Unit 1 control room, what I immediate actions should be taken by the operators? (2.0)
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o It_.BBQQEQVBEl_ _dQBd8Lt_6000Bd8Lt.EUEBic.NQX Page 18 880_B6010LQQ1C6L_CQUIBQL QUESTION 7.09 (2.00)
- a. What combination of conditions may cause vortexing at the DHR pump suction line when in the decay heat removal mode of operation? (1.0)
- b. How does the operator know vortexing is occurring? (0.5)
- c. What should the operator do if it is obvious that the pump is experiencing vortexing? (0.5) l QUESTION 7.10 (2.00) l i
Personnel may be permitted to enter the Reactor Building when the reactor !
l is critical without neutron monitoring badges under certain conditions.What i l are these four (4) conditions? "2.0) I l
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! As. aQUINISIB611V;_C20GEQUBESt_CQUD1I1001. Page 19 L 800 L10116I1061 t
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l QUESTION 8.01 (2.00)
- a. What members of the Fire Brigade Team are assigned by the Unit 1 Shift l Supervisor at the start of each shift? (1.0) t I b. What organization (s) assign the remainder of the Fire Brigade Team
! members? (1.0) l l
l .
l QUESTION 8.02 (3.00) ,
l l a. A fire occurs in an operating makeup pump with the plant at power.
l Personnel respond accordingly and the event is classified as an alert
! due to the potential loss of a single train of an Es system as a
~
l result of fire but the notifications have not yet been made. ihe fire is extinguished quickly and the SS decides that the damage is not nearly as severe as initially thought. TRUE or FALSE: Even though l the condition that resulted in the init.ial classification no longer exists, all notifications associated with an ALERT classification no-t st*11 be made. (0.o; i
- b. List five (5) criteria that should be evaluated prior to terminating l an emergency following a classification. (2.5) <
, i l \
QUESTION 8.03 (3.00)
- a. What are the minimum Protective Active Recommendations for any General l Emergency? ( 1. 0 's ,
- b. What minimum conditions must be met in order for an 0FFSITE evacuetion ;
to be recommended? (1.0) !
I
- c. "ader what conditions is a plant evacuation NOT required for a Site l l
Area Emergency and a General Emergency? (1.0) t i
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At__eQd181 SIB 6IIVE.EBQQEQUBES&_QQURIIl0 Ult Page 20 6UQ.LIdlI6Il001 QUESTION 8.04 (3.00)
- a. Except for physics testing, what is the minimum temperature for c*)tice11ty per the technical specifications? (0.5) ,
- b. The reactor shall be maintained subcritten1 by at least it delte K/k until what two (2) conditions are met in the pressurizer? (1.0)
- c. How much time is allowed to restore either of the conditions in part ,
"b" to within limits if it is exceeded? (0.5)
- d. What must be done if the condition cannot be restored within the
- allowable time of part "c" above? (1.0 QUESTION 8.05 (1.50)
Reactor building integrity shall be maintained wh(never what three (3) conditions exist (assume RCS is closed)? (1.5) l QUESTION 8.06 (1.50)
- e. During refueling op.retions, an unexpected increase in neutron count rate occurs and refueling is halted. What two epprovtis (by title) are required to resume refueling? (1.0) ,
- b. What is the minimum NRC operator license requirement for operation of i the fuel handling bridges? (0.5) .
l QUESTION 8.07 (2.00)
The ERV is leaking at a snown 18 gpm. Total combined leakaga has been calculated at 18.7 99m with the 0.7 gpm identified as OTSG tube leakage.
Have any Tech Spec leakage limits been exceeded? Justif y your answer.(2.0) ,
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I i i (esne* CATEGORY t CONTINUED ON NEXT PAGE *****) j
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fi__aQMINISIBel1VE_EBQGEQUBES&_CQNQ1IlQNit .Page 21 AND_LidlI6IIQN1' QUESTION 8.08 (1.50)
Givan the following types of Standing Orders (1 - S b6 low) answer questions a- c. (1.5)
- 1. Nonsafety-related Departmental Standing Order
- 2. Safety-related Departmental Stending Order _,
- 3. Nonsafety-related Station Standing Order
- 4. Safety-related Station Standing Order
- a. Which of the types of standing orders above must be approved by the Director, Site Nuclear Oprations?
b._ Which of the standing orders must be sent to the PSC for review?
- c. Which of the starding orders may be approved by the appropriate Department Manager? 5 QUESTION 8.09 (3.00)
Approximately twenty minutes prior to shift turnover, one of the R0s on the relieving-crew calls in sick and will not be reporting to work. Prior to the call, the relieving shift had two SR0s (one was the SS and one was the Shift SRO), two R0s (control board operators), one weste control operator (non-licensed), two auxiliary operators (non-licensed), one SAA (non-licensed) snd an GTA (non-licensed).
- s. Will the shift meet the minimum required staffing per 1015.01, Conduct of Operations? (0.S)
- b. How Jong may a shift crew remain below the minimum required staffir.g I and what action must be taken if this time is exceeded? (1.5)
- c. What should be done to ensure adequ.ste staffing if the situation -
detailed above would not have met the minimum requirement? (1,0)
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flt__eQUIN11186IIVE_EBQQEQWBEft_00601110N1.
Page 22
+ ANQ_LidII8IIQN1 QUESTION 8.10 (2.00)
What-two (2) conditions DO NOT require that a "Category E Valve Log Sheet" form be filled out when manipulating Category E velves? (2.0) l l
l l
QUESTION 8.11 (2.50)
- a. What should be performed prior to returning a safety-related motor operated valve to service following manuel operation? (0.5)
- b. List three (3) items that may be checked to ascertain the ability of control valves to operate when performing. valve lineups? (1.5)
- c. What is the minimum level of qualification or position requirea to review a completed valve lineup sheet?. (0.5)
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(********** END OF EXAMINATION **********)
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"; ., 10 Lii__IHEQBY QE_NWQLEeB_EQWE8_EL88I_QEEB8Il0Ni 'Page 23 1 ELVIQSteNQ_IBEBdQQ1Ned101 ANSWER. 5.01 (1.00)
,d ( 1. 0 )'
REFERENCE
'ANO Plant Specific Reactor Theory, pg 200-205 192006K107 3.4/3.4 192006K107 ..(KA's)
ANSWER 5.02' (1.00) ,
c (1.0) .
REFERENCE ,
AN01 Plant Specific Reactor Theory, Chapter 17 192008K121 3.6/3.8 -
192008K121 ..(KA's)
ANSWER 5.03 (2.00)
- a. Increase (0.5 ea) ,
- b. Decrease
- c. Decrease
- d. Increase r
' REFERENCE AN01 PSRT, pg 166 192005K107 2.5/2.8 192005K107 ..(KA's) (
ANSWER 5.04 (2.50)
, s. MTC and FTC (doppler) 0.5 each
- b. FTC (doppler) (0,5)
- c. More negative (0.5)
- d. Less negative (0.5)
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-REFERENCE 'l AN01.PSRT, pg 18'l-185 _
192004K108 3.1/3.1 192004K113 2.9/2.9- e
_ 192004K113 192004K108 ..(KA's)
ANSWER '5.05 -(0.50)
Smaller (0.5) t REFERENCE AN01 PSRT, pg 216 192003K107 3.0/3.0 192003K106 3.2/3.3 '
192003K107 192003K106 ..(KA's)
ANSWER 5.06 '(2.00)
.Will change in a positive direction (0.5). As reactor-power is reduced, T Hot decreases in the top of the core and adds positive reactivity (0.5).
'T Cold increases'in the bottom of the core adding negative reactivity (0.5)
Power in'the top of the core increases relative to the power in the bottom of the core (imbalance is power in the top minus power in the bottom) (0.5) thus imbalance changes in the positive direction.
REFERENCE AN01 PSRT, pg 114 192008K120 3.8/3.9 192007K105 3.0/3.2 192008K120 192007K105 ..(KA's)
ANSWER 5.07 (1.00)
Higher than 30 cps (1.0)
REFERENCE AN01 PSRT, pg 181 AN01 EQB AA-61008-000-02L-1 !
015000A108 3.3*/3.4 015000A108 ..(KA's)
P P
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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y JEf__IHEQBl_QE_NuGLE6B_EQWEB_EL6NI_QEEB6IIQNt. Page'25
' ELUIQ1t&NQ_IHEBbQQ1Ned1GS s
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ANSWER 5.08 (1.00) b'or 7 (1.0)
. REFERENCE AN01 PSRT, pg 148 AN01.EQB AA'61008-000-03P-1 192008K103 3.9/4.0 192008K104 3.8/3.8 '
192008K104 192008K103 ..(KA's)
ANSWER 5.09 (1.00)
Net reactivity at 100% = 0 (0.5) which is the same at 75% (0.5)
REFERENCE ANO PSRT 192002K112 2.4/2.5 192002K112 ..(KA's)
ANSWER 5.10 (1.50)
- a. Lower (0.5 ea)
- b. Higher
- c. Higher 1
REFERENCE l ANO Heat Transfer, Thermodynamics, and Fluids Handbook (HTTF), pg 112 015000A101 3.5/3.8 015000A101 ..(KA's) l l \
l \
l ANSWER 5.11 (1.00) b-(1.0) !,
1 l
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. S t.._ IB E Q Bl_ QE _U u Q L E 6 B _ E QW E B _ E L 6 U I_ Q E E B o II Q U t Page 26 ELul0StoUQ_IBEBdQQXUed1QS REFERENCE ANO HTTF, pg 125 193008K127 2.2/2.4 193008K127 ..(KA's)
ANSWER 5.12 (1.00) c (1.0)
REFERENCE ANO HTTF, pg 144 193008K123 3.9/4.1 193008K123 ..(KA's)
ANSWER 5.13 (1.00) a (1.0)
REFERENCE AN01 ATOG Technical Document, App F, pg F-23 000074A207 4.1/4.7 000074A207 ..(KA's)
ANSWER 5.14 (1.50)
- a. Remain the same (0.5)
- b. Remain the same (0.5)
- c. Increase (0.5)
REFERENCE STM 1-64, Integrated Control System Simulator respotise to the event 041020A202 3.6/3.9 041020A202 ..(KA's)
+
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. it__INEQBY-QE_NUQLE68_EQWEB_ELANI_QEEBal1QNt. .Page127 l ELUIQ1teNQ_lHEBdQQ1NedIQ1
.I ANSWER 5.15 (1.50)
- a. Decreases -(0.5 ea)
- b. Increases
[0.5 each)
- s. INCREASE- < c l b. DECREASE ;-
. c. INCREASE l
- d. DECREASE-i >
REFERENCE ~
{
b' ANO HTTF, pg 124 ,
193008K105 3.4/3.6
+
193000K105 ..(KA .?' i-
'l ANSWER 5.17 (1.50)
Lower-(0.5) due to the heet transfer area decreasing, the Delta T across the OTSG must increase to achieve the same power. (1.0) (QsuA DeltaT) i REFERENCE i 9 ANO Heat Transfer, Thermodynamics, and Fluids Handbook, pg 142, 143 035010K109 3.8/4.0 <
035010K109 ..(KA's) i i
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'It-_IHEQB1_QE_HUGLE68 20 WEB _EL68I_QERB81IQNA Page 28 ELVIQEtoNQ_IBEBdQQ188dICS:
4
- ANSWER 5218 (2.00)
, e. .
The lower level of xenon in that quadrant results -in' a h'igher power.
production (0.5).- Also accept the exposur~e of new fuel.
- b. Tilt will increase (0.5)due to rapid burnout of the' existing xenon in- '
that quadrant (0.5), and then decrease es. xenon concentration increases (0.5)
REFERENCE AN01 Plant Specific Reactor Theory (PSRT), pg 114 192005K111 2.8/3.2 192006K1083.3/3.4 192006K108 192005K111 ..(KA's) 4 i
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L61__ EL 6NI_ SI S I Ed i _ Q E S IGN c _C Q NI B DA t _8N Q _ IN S I B u d EN I 6Il 0 N 'Page129 ANSWER. 6.01 (2.50) ,
- a. The Motor Fault (0.25). lamp illuminates and the Diamond station shifts to' manual (0.75).
~ '
- b. - Safety groups not.at'the out limit (0,5)
- Neutron. error signal > 14 (0.5) .
- ICS power not available (0.5)'
' REFERENCE STM-1-02,. Control Rod Drive System, pg 12-13 001000K402 3.8/3'.8 001000K402 ..('A's)K
' ANSWER 6'.02 (1.00) a (1.0)
LREFERENCE STM-1-02, Control Rod Drive System, pg 3, 5, 7 3 001000K103 3.4/3.6 001000K103 ..(KA's)
ANSWER 6.03 (2.00)
- a. one minute (0.5)
- b. one second (0.5)
- o. O seconds , not required to stort at all (0.5)
- d. 30 seconds (0.5)
REFERENCE
, STM-1-04, Primary Makeup and Purification, pg 9 004000K604 2.8/3.1 013000A401 4.5/4.8 013000A401 004000K604 ..(KA's)
ANSWER 6.04 (1.00)
- Check the 4 red indicating lights near the RCP control switches (0.5)
- Check flow indication for each RCP (0,5) on panel C04
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hiIiEL8NI_SYSIEMS_QE11GNt_QQNIBQLt_6NQ INSIBudENIeIIQN .Page 30 REFERENCE
'. S T M 0 4, Primary Makeup.and' Purification, pg 13 1203.126, Annunciator K08 Corrective Action,.Rev 19, pg 7 003000A109 2.8/2.8 003000A109 ..(KA's)-
ANSWER 6.05 (2.00)
- a. 1
- b. 1, 2, 3, 5 ,
.c. 1, 2, 3, 4
- d. 1, 2 (11 at 0.182 each)
REFERENCE STM-1-31, EDG, p g .1 STM-1-32, Electrical Distribution, pg 26 062000K102 4.1/4.4 062000A211.3.7/4.1 4- 062000K102 062000A211 ..(KA's)
ANSWER 6.06 (2.50)
' a. Opening - pneumatics (0.5) 100 psig instrument air Closing - spring (0.5)
- b. Open (0.5)
- c. Controls the MSIV closure time (1.0) to prevent valve damage REFERENCE STM-1-15, Main Steam, pg 10, 20 039000K405 3.7/3.7 039000K405 ..CKA's)
ANSWER 6.07 (2.00)
! a. normal - natural surface drainage (0,5) backup - Russellville water supply line to the site (0.5) b' . The crosstie valves between the A and 8 pumps (CV3646 and CV3644) will shut (1.0).
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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.hi _ELoNI_111IEd1_QE11GNt_CQUIBQLt_8NQ_INSIBudENI6110N Page~31 REFERENCE STM-1-42, ServiceLWater, ,pg 5, 8, 9 076000K119l3.6*/3.7*
'076000K119 ..(KA's)
ANSWER 6'.08 (2.00)
- a. 3 penetrations (0.25). Also-accepted 5 due to wording of question.
RCP motor air and lube oil cooler
~
CRD cooling water Letdown coolers RCP seal coolers
'(3 loads. required at 0.25 each)
- b. 2 penetrations (0.25)
The CR0 and RCP cooling lines combine to form a. single line (0.375)
The letdown cooling and RCP seal cooling lines combine (0.375)
REFERENCE STM-1-43, Intermediate. Cooling Water, pg 13 PAID M234
-008000K102 3.3/3.4 103000K102 3.9/4.1*
103000K102 008000K102' ..(KA's)
ANSWER 6.09 (2.00)
- n. 5 Halon
- b. 3 pre-action
.c. 3 pre-action
- d. 4 002 e, 1 deluge
- f. 1 deluge
- g. 4 C02
- h. 5.Halon (eight at 0.25 each)
REFERENCE i STM-1-60, Fire Protection, pg 4-9 086000G009 2.9/3.3 0860000009 ..(KA's) i
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hr- EL6HI 1XIIEd1LQE11GNL-QQNIBQLt eND IU17.LQUENI8IlQN' s-ANSWER _6.10 ' .(2.'50)
- a. :- low pressure trip. bypassed
- power imbalance flow trip bypassed
- power to pump trip bypassed
- variable'- pressure trip bypassed
- new high pressure trip incerted
- new high flux trip inserted
- (6 at 0.25 eachl
REFERENCE STM-1-63, Reactor' Protection System, pg'19, 21 012000K105 3.8*/3.9 012000K604.3.3/3.6 012000K604 012000K105 ...(KA's)
' ANSWER 6.11 (2.00)
.a. - Unit in truck
- Runback in progress
- High load limit exceeded
- Low load limit exceeded (0.25 each)
- b. A runback should be initiated (0,5) to run the facility back to 40%
load (0.5) at 50%/ minute.
I REFERENCE l STM-1-64, Integrated Control System, og 1, 3 l 045000K412 3.3/3.6 l, 045000K412 ..(KA's) t-ANSWER 6.12 (1.50) l
- a. Depressing the MANUAL pushbutton (0.5)
-s ' . ,
- b. Depressing the AUTOMATIC pushbutton (0.5) or resetting of the trip signal (0.5).
I
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ht__EL6UI_111IEdi_QE11QUt_QQUIBQLt_480_INSIBudENI8I19U Page 33 REFERENCE STM-1-65, Engineered Safeguards Actuation System, pg 7, 8 013000A401 4.5/4.8 013000A402 4.3/4.4 013000A402 013000A401 ..(KA's)
ANSWER 6.13 (2.00)
- a. Whether at least 1 RCP is running or not (0.5)
Also accept: position of the NC/ Reflux PB on 009.
- b. OTSG pressure (0,5)
- c. Available to OTSO B (0.25)
Not Available to OTSG A (0.25)
- d. Available to both OTSGs (0.5)
REFERENCE STM-1-66, Emergency Feedwater Initiation and Control, pg 2, 3 1105.05, EFIC, pg 12, rev 13 061000K411 2.7*/2.9* 061000A102 3.3*/3.6* 061000A303 3.9/3.9 C61000K411 061000A303 061000A102 ..(KA's) l i
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ANSWER 7.01i (3.00) ,
- s. 2568 Mw't (0,5)
- b. The average power over any eight hour period (or normal work' shift) shall not exseed the licensed power level (0.5). In no case should.
102% (0.5) be exceeded.
- c. The excor& monitoring system has the most restrictive limit (0.5) because it contains larger uncertainties (0,5). >
- d. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (0.5) ,
REFERENCE 1102.04, Power Operations, Rev 18, pg 4 001050A202 3.6/4.2 001000G005 3.7/4.1*
001050A202 001000G005 ..(KA's) -
ANSWER 7.02 (3.00)
- u. Shift Supervisor, Operations Superintendent (0.5 each)
- b. Operations Manager (0.5)
- c. 1. Did automatic ESAS occur?
i 2. Did any train of EFIC actuate?
- 3. Did any major equipment damage occur?
4 Is the cause of the trip unknown? !
- 5. Were any RAC's written as a result of "off-normal" plant or !
equipment performance?
- 6. Were any sections of the Emergency Plan initiated?
- 7. Did plant parameters go beyond the bounds of normal transient
, limit envelope of the SPDS (1600 psig > RCS press < 2450 psig, i RCS temperature < 510 F, subcooling margin < 50 F)? i (3 required at 0.5 each) ;
I REFERENCE 1102.06, Reactor Trip Recovery, Rev 7, pg 4 (
000007G001 3.4/3.8 194001A103 2.5/3.4 '
194001A103 000007G001 ..(KA's) r j (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
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Page 35
.6NQ_B8010LQElG8L_QQNIBQL
+ c..
,, ' ANSWER- 7.03. (2.50)
- a. - insert Group 4 safety rods (0.5)
- re-evaluate boron concentration and/or ECP (0.5)
- b. Reduce power to " 10E5 CPS on'the SR (0.5)
- c. Trip the reactor (1.0)
REFERENCE 1102.08,, Approach to Criticality, Rev 7, pg 3-1203.21, Loss of Neutron Flux. Indication, Rev.1, pg 3, 4 015000A303 3.9/3.9 0000320010 2.9/3.1 015000A303 000032G010 ..(KA's) t ANSWER 7.04 (2.00) !
- a. At least one safety rod group shall be at its-upper limit (0.5) except L when critical or during approach to critical when-all four safety ,
- groups must be withdrawn (0.5).
Also accepted: Op 1-4 100% Gp 5< 80%
Gp 1-4 100% Gp 6 > 95%
Gp 1-4 100% Gp 6 < 75%
- b. Boric acid pump (s) must be stopped (0.5).
d
- c. when they are within 30 ppm boron (0.5).
REFERENCE ,
i i
1103.04, Soluble Poison Concer,tration Control, Rev 9, pg 5, 7 i 1105.09, Section 3.9 -
0040200014 3.9*/3.8 004020A401 3.8/3.3 t 004020G014 004020A401 ..(KA's) l
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4 8NQ 86010LQQ196L_QQNIBQL ANSWER 7.05 '(2.50) -
1
- a. - Subcooling margin in < 30 F (0.25) and < 2 minutes have elapsed (0.25) OR '
- Subcooled margin.has not recovered to ) .or = 50 F (0.25) within 2 minutes (0.25)
- b. core may uncover (0.5)
- c. - Subcooled margin recorder (on 004)
- SPDS CRT display ICC display
- RC pressure and temperature indications ,
- C486 digital readout
' ( 4 at 0.25 ersh) ;
i REFERENCE 1202.01, Emergency Operating Procedure, pg 4, Rev 11 1105.12, RCS Saturation Margin Monitor Operation, pg 1, Rev 2 ,
000007A104 3.6/3.7 000007G011 4.1*/4.3 0070000010 4.2*/4.1* '
000074K101 4.3/4.7 I 000074K101 000007G011 000007G010 000007A104 . . (KA's) i ANSWER 7.06 (2.50) t Degraded Power: - loss of normal control room lighting (0.5) F
- EDG(s) auto start and supplying ES bus (es) (0.5)
- No voltage indicated on 6.9Kv and non-vital 4160V buses (0.5) ,
t Station Blackout: - Loss of all control room lighting except emergency DC [
lights (0.5) r
- No voltage indicated on any 6.9Kv or 4160V bus (0.5) l REFERENCE f 1202.01, Emergency Operating Procedure, Rev 11, pg 8 j 000055G011 4.1*/4.1* 000056G011 3.5/3.8* j 000056G011 0000550011 ..(KA's) b
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88Q_B8010LQQ1G8L_GQUIBQL ANSWER 7.07 (2.50)
- a. 1. no
- 2. yes
- 3. yes
- 4. yes (0.5 each)
- b. Bump an RCP (0.5) to try and establish natural circulation flow.
REFERENCE 1202.01, Emergency Operating Procedure, Rev 11, pg 47, 48 000017A211 3.4*/3.8 000015K101 4.4/4.6 000017A211 000015K101 ..(KA's)
ANSWER 7.08 (3.00)
- a. Begin an immediate plant shutdown at the maximum safe rate (1.0).
- b. - Manually trip the reactor (0.3)
- Close CV-2691 and CV-2692 (MSIVs) (0.3)
Close CV-2630 and CV-2680 (Main Feedwater Isolation Valves) (0.3)
- Manually initiate EFW on the EFIC matrices (0.3)
Close CV-1221 ( Letdown Isolation Valve) (0.3)
- Trip the following breakers and place in Pull-to-lock:
- All H-1 and H-2 (6.9 KV) bus feeders) (0.1)
- A-309 (Bus A-1 to A-3 tie breaker) (0.1)
- A-301 (B-5 bus feeder breaker) (0.1)
A-409 (Bus A-2 to A-4 tie breaker) (0.1)
- A-401 (8-6 bus feeder breaker) (0.1)
REFERENCE 1203.02, Alternate Shutdown, Rev 5, pg 4, 5 000068G010 4.1*/4.2 0000680010 ..(KA's)
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ZA__EBQQEQUBE1_ _NQBdeL&_etHQBdeL&_EdEBGENQ1 Page 38 eNQ_BeQ1QLQGlGal QQNIBQL ANSWER 7.09 .(2.00)
- s. Low RCS level (0.5) and high OHR flowrate (0.5)
- b. Loss of DHR flow or erratic flow (0.5)
- c. Stop the pump (0.5). Also accept reduce pump flow.
REFERENCE 1203.28, Loss of Decay Heat Removal, Rev 3, pg 4 000025A207 3.4/3.7 000025G010 3.9/3.9 000025G010 000025A207 ..(KA's)
ANSWER 7.10 (2.00)
- 1. The SS and RO are aware that personnel are in'the Reactor Building and the exemption is in effect (0.5).
l 2. Reactor power is not allowed to exceed 10E-7 emps (0.5).
- 3. Control rods are in manual (0.5).
- 4. A portable neutron survey meter is in use by HP in the Reactor Building (0.5).
i REFERENCE l
l 1622.09,' Requirements for Reactor Building Power Entries, Rev 5, pg 4
! 194001K103 2.8/3.4 I 194001K103 ..(KA's) t l
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.At__8QdlNISIB8IIVE_EBQQEQUBESt_QQNDIIl0NSt Page 39 8NQ_LIdlI8Il0NS a I
ANSWER 8.01 (2.00)
- a. - Fire Brigade Team Leader (0.5)
- Fire Brigade Team Member (0.5) '
- b. - Unit 2 Fire Brigade Team member assigned by Unit 2 SS (0.5)
- Security (0.5) assigns two team members i REFERENCE 1903.02, Assignment of Plant Emergency Team / Group Members, Rev 8, pg 4 (
194001K116 3.5/4.28 194001A103 2.5/3.4 f 194001K116 104001A103 ..(KA's) l ANSWER 8.02 (3.00) {
6 4
- a. True (0.5) j b. 1. Plant conditions no longer meet the emergency action level l
? criteria AND it appears unlikely that current conditions'will l degrade further requiring reinstitution of an emergency
- classification.
- 2. Non-routine releases of radioactive material to the environment are terminated or under control.
- 3. Any fire, flood, earthquake, or similar emergency condition is controlled or has ceased.
i !
l 4. All specified corrective actions have occurred OR the plant has
! been placed in the appropriate operational mode. ,
t !
! 5. All required notifications have been made. ,
- 6. Agreement with NRC and state officials that downgrading is ;
appropriate if their emergency response organizations have been j activsted as a result of this event.
(5 required at 0.5 each) i i
REFERENCE i 1903.10, Emergency Action Level Classification, Rev 22, pg 5, 7 194001A116 3.1/4.4* 00006/0002 3.2*/4.1 l l 194001A116 000067G002 ..(KA's) !
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-At__8Dd1N1118811VE_EBQQEQUBElt_CQHQII1QN1m Page <40 8NQ_LidlI8Il0NA. J ANSWER 8.03 (3.00)
- e. Shelter 2 mile radius (0.5) [
Shelter 5 miles downwind (0.5) (downwind sector + two adj acent ,
sectors) t
- b. General Emergency (0.25) and ,
> 10% fuel.cled failure (0.25) actual or projected t 1 R whole body (0.25), 5 R thyroid (0.25) past the exclusion eres i
- c. SAE: Ongoing security threat within plant buildings (0.25) but not within the control room or vital eres (0.25)
GE: Ongoing security threat (0.25) within the control room or vital i area (0.25) i Accepted Security threat for full credit f REFERENCE 1903.011, Emergency Response / Notifications, Rev 1, pg 35' 1903.030, Evacuation, Rev 15, pg 4, 5 ,
194001A116 3.1/4.4*
194001A116 ..(KA's) l t
r ANSWER 8.04 (3.00)
- a. 525 F'(0.5) !
- b. -
steam bubble is formed (0.5) [
indicated water level is established between 45 and 305 inches (0.5) [
- c. 15 minutes (0.5)
- d. Be in at least Hot Shutdown (1.0) within the next 15 minutes. I f
REFERENCE i r
AN01 Technical Specification 3.1.3 ;
0010100005 3.7/4.18 001010G014 4.1*/3.7 .
0010100014 0010100005 ..(KA's) l f
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R&__6 QUIN 11188IIVE_E80GEQuBES&_GQ8Q1110811 Page 41 eNQ_ Lid 1I8Il0N1 ANSWER 8.05 (1.50) l
' i
- Reactor coolint pressure > or = 300 pulg (0.5).
- Reactor coolant temperature > or = 200 F (0.5)
- Nuclear fuel is in the core (0.5) l l
REFERENCE ]
1 AN01 Technical Specification 3.6.1 H l
103000G005 3.3/4.1 1030000005 ..(KA's) i l
ANSWER 8.06 (1.50)
- a. - SRO in charge of fuel handling (0.5)
- Reactor Engineering-(0.5)
- b. N) license is required to operated the bridge, but must be directly I supervised by at least en RO license (0.5).
REFERENCE 1502.04, Refueling Shuffle, Rev 15, pg 4 0340000001 2.3/2.9 1 0340000001 ..(KA's) l ANSWER 8.07 (2.00)
No (0.5). ERV leakage is considered returnable leakage (0.5) and up to 30 gpm is allowable (0.5). The limit on total RCS leakage of 1 gpm is not exceeded (0.5).
REFERENCE AN01 Technical Spec ;fication 3.1.6 002000G011 3.3/4.0 0020000011 ..(KA's)
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'It__aQd1N11IBAIIVE_EBQGEQUBEft_GQNQ111QNit .Page 42
- 'AND_ lid 11eIIQN1 ANSWERL '8.08 (1.50)
- a. 2, 3, 4 (0.5)
- b. 2,~4-(0.5)-
- c. 1 (0.5)
(6 at'O.25 each)
REFERENCE 1000.26, Stand'ing Orders, .Rev 2, pg 5 194001A103 2.5/3.4
'194001A103 ..(KA's)
ANSWER 8.09 (3.00)
- a. no (0.5)
- b. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (0.5). Place the plant in a mode where the minimum crew composition is met (1.0).
- c. An RO from the off-going shift should work over en additional four hours and an RO from the next shift should be called in four hours early (1.0).
l REFERENCE l 1015.01, Conduct of Operations, Rev 33, pg 29 - 30, Attachment i
! 194001A103 2.5/3.4 194001A103 ..(KA's)
\
l ANSWER 8.10 (2.00) l 1. When a plant heatup is begun from cold shutdown, and all Category E
! valve positions are documented. In other words, the sheets are not required to be filled out when in refueling or cold shutdown (1.0) since they will all be verified prior to or during the heatup.
I 2. When the valves are manipulated per approved procedures with specific steps requiring signoff and independent verification (1.0)
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At__aQd1811IB811VE_EBQQEQUBEit_QQUQIIlQNit ,Page 43 8NQ_L1dIIeI1QN1 REFERENCE i
1015.01, Conduct of Operations, Rev 33, pg 40 - 41 194001K101 3.6/3.7 194001K101- ..(KA's) l J
i ANSWER 8.11 (2.50) i ,
- s. The valve should be stroked using the motor operator (1.0)
- b. - Pins are removed from valve operators
- Air.is available to pneumatically operated valves l
- Manual / auto bypass levers are in auto
- All manual overrides are removed 1 -
Power is available to control valve solenoids or motors (3 required at 0.5 each)
- c. SRO (0.5) .
i REFERENCE i
4 1015.01, Conduct of Operations, Rev 33, pg 37, 38 194001K101 3.6/3.7 2
194001K*.01 ..(KA's)
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