ML20151X816
| ML20151X816 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 08/18/1988 |
| From: | Graves D, Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20151X814 | List: |
| References | |
| 50-458-OL-88-01, 50-458-OL-88-1, NUDOCS 8808260114 | |
| Download: ML20151X816 (58) | |
Text
{{#Wiki_filter:. i APPENDIX U.S. NUCLEAR REGULATORY COMMISSION REGION IV Operator Licensing Exam Report: 50-458/0L 88-01 Operating License: NPF-47 Docket No: 5C-458 Licensee: Gulf States Utilities P. O. Box 220 St. Francisville, Louisiana 70775 Facility Name: River Bend Station (RBS) Examination at: River Bend Station Chief Examiner: MAf _ D. N. Graves, Examiner P)dp/P/ Date Operator Licensing Section . Division of Reactor Safety Approved by: qy L h I Date 4 L~. Pellet, Section Chief Operator Licensing Section Division of Reactor Safety-1 l l Summary j NRC Administered Examinations Conducted During the Week of July 11, 1988 (Report 50-458/0L 88-01) NRC administered examinations to two reactor operator and one senior reactor operator candidates. Two candidates passed all portions of the examinations taken and have been issued the appropriate license. One candidate failed the , Integrated Plant Operations section of the operating examination. ! 88 8260114 080019 y ADOCK 05000450 PNV , l i
. i DETAILS
- 1. Persons Examined SR0 R_0, Total License Examinations: -Pass - 1 1 2 Fail - 0 1 1
- 2. Examiners D. N. Graves, Chief Examiner
- 3. Examination Report Performance 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 In general, editorial coments.or changes made during the examination, or subsequent grading reviews are not addressed by this resolution section. All coments proposed by the facility were accepted for incorporation into the examination and answer key. The full text of the comments is attached.
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- b. Site Visit Sumary (1) At the end of the written examination administration, the facility licensee was provided a copy of the examination and answer key for the purpose of comenting on the examination content validity. It was explained to the facility licensee that regional policy was to have examination results finalized ,
within 30 days. Thus, a timely response was desired to attain l this goal.- (2) At the conclusion of the site visit, the facility representative l and chief examiner discussed written examination review i coments. I
- c. Master Examination and Answer Key Master copies of the RBS license examinations and answer keys are attached. The facility licensee comments which have been accepted are incorporated into the answer key.
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s % Gulf States Utilities d. Facility Examination Review Coments The facility licensee comments regarding the written examination are attached. ,
- e. Simulation Facility Fidelity Report No simulator deficiencies or discrepancies were noted or observed during the simulator portion of the operating examinations.
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f U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _BIVEB_BEUQ_1____________ REACTOR TYPE: _QWB-GEh_________________ DATE ADMINISTERED: _HalQZtl2________________ EXAMINER: _GB6YElt_Qt______________ CANDIDATE: _________________________ INSIBUQIl0H1_IQ_Q68Q1Q61El Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__V6LVE_ _IQI6L ___1QQBE___ _Y6LUE__ ______________Q61EQQBI_____________ _21t00__ 100z00 ___________ ________ 1. PRINCIPLES OF NUCLEAR POWER ' PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW _21tQQ__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature i
- 14__EBINGIELES_QE_ NUCLE 68_EQWEB_EL8NI_QEE88IIQNt PAGE 2 IBEBdQQ1NedIGSt_UE6I_IB6NsEEB_6NQ_ELUID_ELQW 1
QUESTION 1.01 (2.00) ,
- s. Explain the effect that tends to make the Moderator Temperature .
i Coef ficient more positive when moderator temperature is increased.(1.0)
- b. Explain the effect on Fuel Temperature Coefficient due to increasing moderator temperature. (1.0)
ANSWER 1.01 (2.00)
- a. The positive effect occurs because moderator is expanded 'out of the core (0.5] and fewer neutrons are absorbed in the moderator at higher temperature. [0.5] (1.0) ,
- b. As temperature increases, slowing down time and length become longer. [0.5) Any change in resonance peaks will become more significant as neutrons spend more time in resonant energy spectrum. ,
[0.5] (1.0) REFERENCE River Bend Station, LOTM-TH-1.17-2 page 2 K102 Kill 2.5 2.5 NU OL 292004K102 292004K111 ...(KA's) i QUESTION 1.02 (1.00) l The reactor scrams from full power, equilibrium xenon conditions. Six (6)' E hours later the reactor is brought critical and power stabilized low in the intermediate range. If power level is maintained at this level, which one
- -o f the following statemants is CORRECT concerning control rod motion?
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- a. Rods will have to be withdrawn since Xenon will closely follow its normal buildup rate. ,
- b. Rods will have to be inserted due to Xenon decay.
- c. Rods will have to be rapidly inserted since the critical reactor i will cause a high rate of Xenon burnout.
- d. Rods will approximately remain as is as the Xenon establishes its e i equilibrium value for this power level.
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1&__EBINGIELES_QE_N99LE6B_E9 WEB _EL6NI_QEEB8IIQNt PAGE 3 IBEBdQDXUedIGS& _BE6I_IB6NSEEB_680_ELUID_ELQW ANSWER 1.02 (1.00) c (1.0) REFERENCE River Bend Station, LOTM-TH-1.22-2, HLO-147-2, LO.4b. K114 3.1/3.2 NU 87 292006K114 ...(KA'S) QUESTION 1.03 (2.00) List and briafly explain two (2) of the reasons that cause a high reactor pressure transient to be more severe at EOL than BOL. (2.0) ANSWER 1.03 (2.00)
- 1. At EOL, there is a slower scram reactivity insertion rate (0,5) because the control rods are further withdrawn than at BOL (0.5).
- 2. Beta, the delayed neutron fraction, is smaller at EOL CO.5). This results in a shorter reactor period at EOC than at BOL for the same positive reactivity addition (0,5).
REFERENCE l River Bend Tech Spec Bases 3/4.J.4 i River Bend Station, HLO-139-0, Lesson Objective I10 K106 3.7/3.7 K103 3.6/4.1 NU 87 292003K106 295001K103 ...(KA'S) I l I QUESTION 1.04 (3.00) l For each of the events below, state the order in which the three reactivity coefficients will respond (first, second, and third). ALSO indicate whether each coefficient will be adding + or - reactivity during the event. Use all three coefficients for each event. (3.0)
- c. Rod drop at power
- b. Turbine stop valve fails shut at power
- c. One recirc pump trips at power
. n l iz__EBINCIELES_QE_ NUCLE 68_EQWEB_EL8NI_QEEBoIIONt PAGE y4 ] INEBbQQ1 Hob 1 cst _UE6I_IBeNSEEB_6HD_ELVID_EL9W ANSWER 1.04 (3.00)
- e. Doppler (-), MTC (-), Void (-)
- b. Void C+), Doppler (-), MTC (_)
- c. Void (- ) , Doppler (+), MTC (+)
(0.25 for each + or - and 0.25 for the order) REFERENCE River Bend Station, LOTM-TH-1.17 thru 1.19 K101 3.2/3.2 K105 2.9/2.9 K110 3.2/3.2 NU OL-292004K101 292004K105 292004K110 ...(KA'S) QUESTION 1.05 (3.00) A reactor startup is in progress. You have been given an estimated critical roo position for the conditions at 0800. You start to pull control rods at 0800 for the approach to cr itical. HOW WOULD EACH OF THE FOLLOWING conditions or events AFFECT the ACTUAL CRITICAL ROD POSITION (more rod withdrawal, less rod withdrawal, or no significant effect)? (3.0)
- a. One reactor recirculation pump is stopped,
- b. Xenon is changing due to extended power operation, terminated 16 hours previously,
- c. Shutdown cooling is stopped (significant decay heat).
- d. Reactor head vent is inadvertently' closed. ]
- o. Moderator temperature is gradually decreasing.
- f. Reactor Water Cleanup System isolates (s ign if ic ant decay heat).
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It__EBINQIELES_QE_NUQLEeB_EQWEB_EleNI_QEEB6IIQNt PAGE 5 l l IBEBdQQYNedIQSt_UE6I_IBoNSEEB_oND_ELVIQ_ELQW i l ANSWER 1.05 (3.00)
- o. No significant effect
- b. Less rod withdrawal
- c. More rod withdrawal
- d. No significant effect
- e. Less rod withdrawal
- f. More rod withdrawal (0.5 EA.]
REFERENCE ; River Bend Stacion, LOTM-TH-1.17-2; 1.22-2 K106 4.2 NU OL 292008K106 ...(KA'S) QUESTION 1.06 (2.50) While River Bend is operating at 90% power, extraction steam to the highest i pressure feedwater heater is removed. An engineer observed that the turbine load increased by 20 MW electric and concluded that this action has improved (increased) the plant's thermodynamic efficiency (not heat rate). In this conclusion correct? Explain your answer. (Include what caused ; electrical output to increase.) (2.5) ANSWER 1.06 (2.50) ! No (0.5). (Therme efficiency is a comparison of Energy In to Energy Out.) ! The increase in output results from no steam being diverted to the high I pressure feedwater heater (0,5) and increased Rx. power due to colder foedwater temperature (0.5]. Because the feedwater is now cooler, more ; onergy from the reactor is required to bring the water up to saturation I temperature (1.0) thus thermo ef f iciency is down. I l REFERENCE ) River Bend Station, HLO-314-0; LO-VI ! River Bend Station, Simulator Malfunction #93 i River Bend Station, HLO-114-2, LO.II.2 K105 2.7/2.8 NU 87 l 293005K105 ...(KA'S) l (***** CATEGORY 01 CONTINUE 0 ON NEXT PAGE *****) I
l 1 . l It__EBINCIELE3_9E_ NUCLE 68_E9 WEB _EL8NI_9EE86II9Nt PAGE 6 IBE80901NedigS _UE6I..IB8NSEEB_eNQ_EL910_EL9W t QUESTION 1.07 (2.50) The reector is operating at 100% power when one SRV fails open. Describe the response:of each of the parar.eters listed below. Include the initial change, the trend for that parameter over the next five minutes, and the reason (s) for the response. (2.5)
- a. Indicated Steam Flow (total)
- b. Turbine Steam Flow
- c. Reactor Power ANSWER 1.07 (2.50)
Indicated steam flow initially increases (.25) due to the SRV drawing more steam off the reactor ( 25). As reactor pressure decreases, the turbine control valves close down to control pressure (.25) resulting in reduced turbine steam flow C.25). Reactor power initially decreases (.25) due ia the depressurization (.25). Turbine steam flow decreasing causes a slight dscrease in FW temperature (.25) causing reactor power to begin increasing again (.25) which causes the pressure control system to increase turbine and total steam flow (.25). Final conditions will be slightly higher reactor power with the steam flow shared between the turbine and the SRV OR the reactor will scram on high power C.25). REFERENCE River Bend Station, A0P-0035 A107 2.9/3.0 A109 3.1/3.3 K122 3.5/3.6 NU OL 239002A107 239002A109 292008K122 ...(KA'S) 4 QUESTICN 1.08 (1.00) Describe the difference between cavitation and gas binding in a centrifugal pump. (1.0)
1 . s L . l Iz__EBINCIELE1_QE_ NUCLE 68_EQWEB_EL8NI_QEE86IIQNt PAGE 7 IBEBMQQXNedIQ1t_HE61_IB6NSEEB_6NQ_ELUID_ELQW ' ANSWER 1.08 (1.00) Cavitation occurs when water fladhes to steam in the pump eye. { Loss of NPSH) (0,5) Gas binding means that the pump is filled with a non-condensable gas. [0.5) REFERENCE River Bend Station, LOTM-TH-2.4-2 page 2 K112 2.9 TH OL 293006K112 ...(KA*S) QUESTION 1.09 (1.50) Indicate HOW each of the following changes will affect reactor CRITICAL
- '0WER (INCREASE or DECREASE). If CRITICAL POWER will not'be affected, state this. (1.5)
- a. Loss of extraction steam to a feedwater heater. i
- b. Mass flow rate through the core is increased.
- c. Reactor pressure is increased.
ANSWER 1.09 (1.501
- e. CRITICAL POWER increasen (0.5).
- b. CRITICAL POWER increases (0.5).
- c. CRITICAL POWER decreases (0.5).
I REFERENCE l LOTM-TH-3.7-0, Transition Boiling, pg 3 River end Station, LOTM-TH-4.8-2 ; K122 K123 K124 2.9 2.8 2.7 NU OL a 293009X122 293009K123 293009K124 ...(KA'S) ! l 0 (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****) _._ . . _ . _ _ _ _ _ _ . . - . _ __ _ . . _ . .__._. .. , .- _ _ . _ ~ . ~ . . _
11__EBIUCIELES_QE_UUCLE88_E9 WEB _EbeUI_9EEBoIIQUt PAGE 8 IBEBdQDXUedigS t_UgeI_IB6NSEEB_6ND_ELUID_ELQW QUESTION 1.10 (1.00) Using Figure 1, (next page), match the regions of the curve below (a - d) to the correct area (1 - 7) on the curve. (1.0) (a) Region I = 1. Nucleate Boiling (b) Region II = 2. Film Boiling (c) Region III = 3. Partial Film Boiling (d) Region IV = 4. Single Phase Convection
- 5. Forced Convection
- 6. Burn Out Flux
- 7. Onset of Transition Boiling ANSWER 1.10 (1.00) o = 4 b = 1 c = 3 j
d = 2 (0.25 each) I REFERENCE River Bend Station, LOTM-TH4.5-2 page 3 (Figure 3) l K108 2.9 TH NW l 293008K108 ...(KA'S) l I QUESTION 1.11 (1.00) l List two factors which determine the rate of heat transfer in a heat i oxchanger. (1.0) ANSWER 1.11 (1.00)
- 1. Surface area available. (0.5)
- 2. Temperature difference between the two riuids. (0,5)
REFERENCE River Bend Station, LOTM-TH3-10.2 page 13 K106 2.7 TH NW
L. - It__EBINCIELES_QE_NUGLE68_EQWEB_EL68I_QEEB6IIQNi PAGE 9 IHEBdQQ18601GSt_UE6I_IB6HSEEB_680_ELUID_EkQW 293007K106 ...(KA'S) QUESTION 1.12 (1.00) A temperature instrument with an out of date calibration sticker on it is roading 450 deg F. A recently calibrated pressure gage sensing in the same , eres indicates 390 psig. Is the temperature instrument reading accurately (within + or - 2 deg F)? If not, how close is it reading to the octual temperature? Assume the system is under saturated conditions. SHOW ALL WORK. (1.0) ANSWER 1.12 (1.00) 390 psig + 14.7 psig = 404.7 psia (0.25) Saturated temperature f or 404.7 psia: (456.28 deg F - 444.6 deg F) (4.7/50) + 444.6 deg F =445.7 deg F (0.5) The temperature instrument is reading 4.3 deg too high (0.25) ' REFERENCE Steam Tables 293003 K1.23 2.8/3.1 TH OL 293003K123 ...(KA'S) QUESTION 1.13 (2.00) ' Consider two RPV conditions: low power and low flow (<10%) OR high power l end high flow (>85%).
- e. During which condition is REQUIRE 0 NPSH for a racirculation pump greater? (0,5)
- b. During which condition is AVAILABLE NPSH f or a recirculation pump greater? (0.5)
- c. What provides the majority of the available Net Positivo Suctior. Head
- for the Recirculation System during: (1.0)
- 1. Low speed pump operation (low power)?
- 2. High speed pump operation (high power)?
It__EBINQ1ELES_QE NUQLE68_EQWEB_EL6NI_QEEB611QNi PAGE 10 IBEBdQQINed10St_UE6I_IB6NSEEB_6ND_ELV10_ELQW ANSWER 1.13 (2.00)
- e. High flow high power. (0,5)
- b. High flow high power (0.5)
- c. 1. Height of water in the reactor vessel (0,5)
- 2. Subcooling effect of the incoming feedwater (0.5)
REFERENCE River Bend Station, LOTM-TH-2.4-2 page 2 KA = 2.8 KAdO2 = 3.1,'3.2 TH OL 202001K402 293006K110 ...(KA'S) QUESTION 1.14 (1.50) Following an automatic initiation of LPCI at a reactor pressure of 400 i poig, reactor pressuae decreases to 100 psig. For each of the parametors listed below, determine any change (i.e. increase, decrease, or remain the ocme).
- o. LPCI injection flow '-
- b. LPCI pump discharge head (assume constant NPSH)
- c. LPCI pump power requirements l
ANSWER 1.14 (1.50)
- a. Increase (0.5) l
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- b. Decrease (0.5)
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- c. Increase (0.5)
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i REFERENCE I G.E. Heat Transfer and Fluid Flow, Ch. 6 pg. 6-95 & 6-96. SLO 6-10. l River Bend Station, LOTM-TH-2.4-2, p. 5 & p. 3; HLO-104-3, L.O.II.3. & i L.O.II.10 l KA 201004 K1.05 = 2.8/2.9 TH OL j 291004K105 ...(KA'S) ' l
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l l l t rtRC LICENSE EXN11 RATION llM000T EQUATIONS. CO:ISTMTS, NiD CONVERSI0HS 6'=r.%p*deltaT d=U*A*deltaT P = Po*10sur*(t) P = P *et /T SUR = 26/T T = 1*/p + (p-p)/I p T=1/(p-p) T = ($-p)/Xp p = (Xerr-1)/Ke rf = deltakerr/Ke rt p = 1*/TKeff + feff/(1+1T) . A = In2/tg = 0.693/tg K = 0.1 seconds-1 I = Io*e~"* CR = S/(1-Keff) 2 R/hr = 6*CE/d feet Water :'arameters 1 gallon = 8.345 lbm = J.87 liters 1 ft3 = 7.48 gallons , Oensity 9 STP = 62.4 lbm/ft3 = 1 gm/cm3 Heat of vaporization = 970 Btu /lbm Heat of fusion = 144 Btu /lbm 1 atmosphere = 14.7 psia = 29.9 inches Hg. Miscellaneous Conversions I curie = 3.7 x 101u disintegrdtions per second 1 kilogram = 2.21 lbm I horsepower = 2 54 x 103 Btu /hr 1 m = 3.41 x 105 Btu /hr i 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Stu = 778 ft-lbf i
4 U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _BlVE8_QENQ.l____________ REACTOR TYPE: _QWB-QEh_________________ DATE ADMINISTERED:_BQ20Zll2________________ EXAMINER: _QB6VE$t_Qt _____________ CANDIDATE: _________________________ IU51BVQIl005_IQ_Q6801Q61El Uce separate paper for s answers. Write answers on one side only. Staple question sheet on t p of the answer sheets. Points for each question are indicated ir trentheses after the question. The passing grade requires at l e as t '/ u, in each category and a final grade of at losst 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__Y6LVE_ _IQI6L ___5CQBE___ _Y6LUE__ ______________Q81EQQBl_____________ _25t00_- 25AQQ ___________ ________ S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS _25A00__ 25z00 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION _25s90__ _25200 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL _25tDQ__ _25t00 ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 10DiuQ__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. Candidate's Signatui
. -. j St__IHEQBI_QE_NUGLE68_EQWEB_2L6NI_QEE86IIQNt_ELUIDSt_6NQ PAGE 2 IBERdQQ186dlGS QUESTION 5.01 (1.50)
Compare withdrawing a center control rod at 90% rod density to withdrawing a center control rod at 40% rod density. In which situation is control rod worth greater for the withdrawn control rod? JUSTIFY YOUR ANSWER. (1.5) ANSWER 5.01 (1.50) Withdrawal of a center cor, trol rod at 90% density has greater worth (0.5). (The control rod worth is proportional to the local neutron flux / the core average neutron flux squared.) With 90% rod density the core average neutron flux is very small. (0.25) Withdrawing a central control rod, ancreases the local flux in the area of withdrawn rod substantially. Because the rod causes the value of the term f. l o c a l neutron flux / core average neutron flux) squared to ba large its worth for this condition is quite high. Higher than withdrawing the rod at 40% rod density, when core average flux will be higher. (0.75) REFERENCE River Bend Station, LOTM-TH-1.20-2; HLO-145-2, L.O.II.7 Kl.12 2.9 NU OL 292005Kil2 ...(KA'S) QUESTION 5.02 (2.00) f List and briefly explain two (2) of the reasons that cause a high reactor pressure transient to be more severe at EOL than at BOL. (2.0) l l
1 l, 5t__INE981_9E_UUCLE68_E9 WEB _ELeUI_QEE86I196t_ELU10St_6UD PAGE 3 IHEBUQD18601GS ANSWER 5.02 (2.00)
- 1. At EOL, there is a slower scram reactivity insertion rate (0,5) because the control rods are further withdrawn than at BOL (0,5).
- 2. Beta, the delayed neutron fraction, is smaller at E0L (0.5). This results in a shorter reactor period at E0C than at BOL for the same positive reactivity addition (0.5).
REFERENCE River Bend Tech Spec Bases 3/4.3.4 River Bend Station, HLO-139-0, Lesson Objective IID K106 3.7/3.7 K103 3.6/4.1 NU 87 292003K106 295001K103 ...(KA'S) . QUESTION 5.03 (3.00) For each of the events below, state the order in which the three reactivity coefficients will respond (first, second, and third). ALSO indicate whethe cach coefficient will be adding + or - reactivity during the event. Use all three coefficients for each event. (3.0)
- o. Rod drop at power
- b. Turbine stop valve fails shut at power
- c. One recirc pump trips at power ANSWER 5.03 (3.00)
- e. Doppler (-), MTC (-), Void (-)
- b. Void C + ), Doppler (-), MTC (-)
- c. Void (-), Doppler C + ), MTC (+)
(0.25 for each + or - and 0.25 for the order) REFERENCE River Bend Station, LOTM-TH-1.17 thru 1.19 K101 3.2/3.2 K105 2.9/2.9 K110 3.2/3.2 NU OL 292004K101 292004K105 292004K110 ...(KA'S) (1**** CATEGORY 05 CONTINUED ON NEXT PAGF *****)
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l Sa__IBEQBl_DE_NUQLELB_EQWEB_EL8NI_QEE86IIQNt_ELUIQHt_eNQ PAGE 4 IBEBdQQ1NedIDS QUESTION 5.04 (2.50) Answer the f ollowing with regard to the Controb Cell Core -(CCC) operating strategy: 1
- c. What constitutes a CONTROL CELL? (0.5)
- b. While operating at power, what is the position of the non-control cell control rods? (0.5)
- c. Why do control rod movements result in a lower kw/ft change per notch in a CCC than in a conventional cor ;? (1.0)
- d. TRUE or FAL3E. Using the CCC operating strategy eliminates the need for control rod sequence pattern exchanges. (0.FJ ANSWER 5.04 (2.50)
- e. Four fuel bundles and a control rod (0.5)
- b. Fully withdrawn (0.5)
- c. All cor, trol rod movements are associated witt low reactivity cells (1.0).
- d. True (0.5)
REFERENCE River Bend Station, LOTM-1-2 CVol. 1), Section M.2.p.17; HLO-002-2, L.O.II.8 K1.12 = 2.9 K1.23 = ?.1 NU OL 292005K112 292008K123 ...(KA'S) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
. i St__IBEQB1_QE_ NUCLE 68_EQWEB_EL88I_QEE86IIQNt_ELVIDSt_8NQ PAGE 5 IBEBUQDXNedICS QUESTION 5.05 (2.50) The reactor is operating at 100% power when one SRV fails open. Describe the response of each of the parameters listed below. Include the initial change, the trend for that parameter over the next five minutes, and the reason (s) for the response. (2.5)
- e. Indicated Steam Flow (total)
- b. Turbine Steam Flow
- c. Reactor Power ANSWER 5.05 (2.50)
Indicated steam flow initially increases (.25) due to the SRV drawing more steam off the reactor (.25). As reactor pressure decreases, the turbine control valves close down to control pressure (.25) resulting in reduced turbine steam flow (.25). Reactor power initially decreases (.25) due to the depressurization (.25). Turbine steam flow decreasing causes a slight decrease in FW temperature (.25) causing reactor power to begin increasing again C.25) which causes the pressure control system to increase turbine and total steam flow (.25). Final conditions will be slightly higher reactor power with the steam flow shared between the turbine and the SRV OR the recctor will scram on high power C.25). REFERENCE ' River Bend Station, A0P-0035 A107 2.9/2.0 A109 3.1/3.3 K122 3.5/3.6 NU OL 239002A107 239002A109 292008K122 ...(KA's) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) \ - - - - - - - - - - -
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i St__IVEQB1_QE_NUGLE68_EQWEB_ELeNI_QEEBeIIQNt_ELUIDSi_eNQ PAGE 6 1 IHEBdQQXNedICS l i QUESTION 5.06 (1.50) !
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A reactor water sample indicates the following isotohespresent: 7 Tritium.(H3) Kr_ 87 1 Mn 56 I 131 Ni 57 Xe 133 Fe 59 Cs 137 Co 60 Ba 140
- e. Which of the above are fission products and which are corrosion products? (1.0)
- b. Does the presence of fission products in the coolant mean a fuel element failure or defect exists? (0.5)
ANSWER 5.06 (1.50)
- c. Fission products: Tritium, I, Kr, Xe, Cs, 8a (0.1 each)
Corrosion products: Fe, Ni, Co, Mn (0.1 each)
- b. No (0.5)
REFERENCE River Bend Station, HLO-308-1, L.O.II.8.; LOTM-TH-5.4-2;- HLO-077-0, L.O.II.5 SG15 = 3.6 K3.01 = 3.6 K3.06 = 2.7 NU OL 204000G015 204000K301 204000K306 ...CKA'S) QUESTION 5.07 (1.00) ) Oriofly explain the bases for the LHOR limit. (1.0)
St__IBEQBl_QE_NUQLE88_EQWEB_EL88I_QEEB8110Nt_ELUIDSt_8NQ PAGE 7 IHEBdQQ1NedIQS ANSWER 5.07 (1.00) Due to the expansion of the fuel pellet, the gap between the pellet and the cladding can be eliminated (0.33). This friction between the two matesials will cause a concentration of cladding stress (0.33). This stress can oxceed the yield stress of the cladding (0.33). Accepted "to prevent exceeding 1% plastic strain on the cladding." REFERENCE River Bend Station, LOTM-TH-4.9.2 page 4 K107 3.6 TH NW 293009K107 ...(KA*S) QUESTION 5.08 (1.00) List two places in a BWR in which 1. eat is transferred by conduction, and the conducting material is not solid. (1.0) ANSWER 5.08 (1.00)
- a. The helium gap between the fuel pellet and the cladding. (The fuel pellet to the cladding material.)(0,5)
- b. The boundary or laminar layer between the cladding and the coolant (0.5).
REFERENCE River Bend Station LOTM-TH-4.2-2, Volume 6, Chapter 4, page 7 K1.01 = 3.2 TH NW 293007K101 ...(KA'S) QUESTION 5.09 (1.00) The APLHGR limit is to limit fuel cladding tumperature to a maximum of 2200 degrees F. Briefly describe why this value was chosen as opposed to a higher and lower value. (1.0)
l
-. 1 St__IHEQBl_QE_NUCLEeB_EQWEB_ELeNI_QEEBeIIONt_ELVIDS&_6NQ PAGE 8 IHEBdQQ1NedIGH
{ i ANSWER 5.09 (1.00) The value is high enough to allow reasonable fuel pellet temperatures for officient steam production (0.5). The selection was low enough to prevent fuel and cladding damage during a LOCA (0,5). (If ECCS water is injected, o thermal shock would occur causing the clad to brittle fracture. Below 2200 degrees F a zirc-water reaction is not self sustaining.) REFERENCE River Bend Station, LOTM-TH-4.10-2 page 2 I Kill 3.6 TH NW i 293009K111 ...(KA'S) 1 QUESTION 5.10 (2.50)
- c. What is the relationship between MAPRAT and MAPLHGR? (1.0)
- b. The process computer prints out a MAPRAT of 1.05. Is this acceptable (YES or NO)? (0.5)
, c. What physical consequence could occur if the MAPRAT limit is exceeded and under what conditions? (1.0)
ANSWER 5.10 (2.50)
- e. MAPRAT =
MAPLHGR (actual) / MAPLHGR (limit) (1.0)
- b. No (0,5)
- c. The clad temperature could exceed 2200 deg F (0.5) during a LOCA
-r core uncovery (0.5).
REFERENCF, River Bend LOTM-TH-4.10-2: HLP-125-2, 0.L. II.1 29300a. X1.11 2.8/3.6, K1.13 3.1/3.6 TH OL 292009K111 293009K113 ...(KA'S) i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) _____.____________m-____--__-- _ - - _ - - - - - - - -
11__IHEQBl_QE_NUQLE68_EQWEB_EL6NI_QEE86IIQNt_ELUIDSt_6NQ PAGE 9 I8E80001N60191 l l QUESTION 5.11 (2.50) While River Bend is operating at 90% power, extraction steam to the highest pressure feedwater hester is removed. An engineer observed that the turbine load increased by 20 MW electric and concluded that this action has improved (increased) the plant's thermodynamic efficiency (not heat rate). Is this conclusion correct? Explain your answer. (Include what caused electrical output to increase.) (2.5) ANSWER 5.11 (2.50) No (0.5]. (Thermo efficiency is a comparison of Energy In to Energy Out.) The increase in output results from no steam being diverted to the high pressure feedwater heater (0.5) and increased Rx. power due to colder feedwater temperature (0.5]. Because the feedwater is now cooler, more energy from the reactor is required to bring the water up to saturation temperature (1.0) thus thermo efficiency is down. REFERENCE River Bend Station, HLO-314-0; LO-VI River Bend Station, Simulator Malfunction #93 River Bend Station, HLO-114-2, LO.II.2 K105 2.7/2.8 NU 87 293005K105 ...(KA'S) QUESTION 5.12 (1.00) A temperature instrument with an out of date calibration sticker on it is reading 450 deg F. A recently calibrated pressure gage sensing in the same area indicates 390 psig. Is the temperature instrument reading accurately (within + or - 2 deg F)? If not, how close is it reading to the actual temperature? Assume the system is under saturated conditions. SHOW ALL WORK. . (1.0) ,
m Sz__IHEQBl_QE_NUQLE88_EQWEB_EL6HI_QEE86Il0NA_ELVIDS&_8NQ PAGE 10-IBEBdQQ1N601GS ANSWER 5.12 (1.00) 390 psig + 14.7 psig = 404.7 psia (0.25) Saturated temperature for 404.7 psia: (456.28 deg F - 444.6 deg F) (4.7/50) + 444.6 deg F =445.7 deg F (0,5) The temperature instrument is reading 4.3 deg too high (0.25) REFERENCE Steam rables 233003 Kl.23 2.8/3.1 TH OL 293003K123 ...(KA'S) QUESTION 5.13 (1.00) A reactor heat balance was performed (by hand) during the midnight shift due to the Process Computer being 000. The GAF's wero computed, but the , APRM GAIN ADJUSTMENTS HAVE NOT BEEN MADE. Which of the following statements is TRUE concerning reactor power? (1.0) SELECT ONLY ONE ANSWER COnly one is truel)
- e. If the feedwater temperature used in the host balance calculation was LOWER than the actual feedwater temperature, then the actual power is HIGHER than the currently calculsted power.
- b. If the reactor recirculation pump heat input used in the heat balance calculation was OMITTED, then the actual power is LOWER than the currently calculated power. '
- c. If the steam flow used in the heat balance calculation was LOWER than the actual steam flow, then the actual power is LOWER than the currently calculated power.
- d. If the RWCU return temperature used in the heat balatace cal lculation was HIGHER than the actual RWCU return temperature, then the actual power is LOWER than the currently calculated power.
I
l 51__INEQBl_QE_NVQLE68_EQWEB_EL6NI_QEEB6IIQNt_ELVIQSt_eNQ PAGE 11 l I8E8000106d103 ) L ANSWER 5.13 (1.00)
- b. (1.0)
REFERENCE l River Bend Station, HLO-115-2, L.O.II.1&2, V. #1; LOTM-TH-3.5-2; l LOTM-TH-3.13-2. l K1.11 = 3.1 NU -0L l l 293007K111 ...(KA'S) 1 QUESTION 5.14 (2.00) Consider two RPV conditions: low power and low flow (<10%) Ok high power end high flow (>85%).
- e. During which condition is REQUIRED NPSH for a recirculation punip greater? (0.5)
- b. During which condition is AVAILABLE NPSH for a recirculation pump greater? C0.5)
- c. What provides the majority of the available Net Positive Suction Head for the Recirculation System during: (1.0)
- 1. Low speed (low power) pump operation?
- 2. High speed (high power) pump operation?
ANSWER 5.14 (2.00)
- c. High flow high power (0,5)
- b. High flow high power (0.5)
- c. 1. Height of water in the reactor vessel (0.5)
- 2. Subcooling effect of the incoming feedwater (0.5)
REFERENCE River Bend Station, LOTM-TH-2.4-2 page 2 K1.10= 2.8 KA4.02 = 3.1/3.2 TH OL 202001K402 293006K110 ...(KA'S) (***** END OF CATEGORY 05 *****)
6t__EL6HI_311IEda_ DES 10Nt_CQNIBQLt_6ND_INSIBubENI6I1QN PAGE 12 QUESTION 6.01 (1.50) While operating RHR Pump A in the Shutdown Cooling Mode, a LPCI initiation
~
signal is received. All actions / trips occur in accordance with system design. LIST the actions required to be performed by the operator to use the A Loop of RHR in the LPCI mode. Component numbers are not required, but component descriptions should be specific enough to accurately identify the components. (1.5) ANSWER 6.01 (1.50)
- 1. Close Pump A Shutdown Cooling Suction Valve (F006A).
- 2. Open Pump A Suppression Pool Suction Valve (F004A).
- 3. Take the pump control switch to 0FF/0 PEN and then to START /0N.
(3 0 0.5 each) REFERENCE RBS, HLO-021-1 LO 4 & 7, LOTM-19-2, PP. 45 - 48 (3.6/3.7, 3.9/3.9, 4.2*/4.2) 203000K114 203000K401 205000K108 ...(KA'S) QUESTION 6.02 (2.00) The HPCS Diesel Generator Mode Select Switch was inadvertently left in the "MAINTENANCE" position (all other controls normal). A valid LOCA signal is received. The Division III bus, E22-S004, becomes deenergized and the operators take the following mitigating actions in the indicated order:
- 1. Place the HPCS Pump Control Switch on 1H13-P601 to TRIP.
- 2. Place the HPCS D/G Mode Select Switch to Automatic.
EXPLAIN the responses of BOTH the HPCS D/G AND the HPCS Pump to the operator actions. Limit your discussion to diesel and pump responses to the electrical signals / control manipulations AND the reasons for those responses. (2.0)
61__EL6NI_SXSIEUS_QESIGNt_QQNIBQLt_6NQ_INSIBudENI6IIQN PAGE 13 ANSWER 6.02 (2.00) , The HPCS 0/G would start and tie to the Div 3 bus (0.5) due to-LOCA (0.25) and undervoltage signals -(0.25). The HPCS. pump would not auto start (0.5) , taking the control' switch to the trip position (breaks the auto start signal) puts the pump in manual override. (0.5) REFERENCE RBS, HLO-019-1 LO 2 & 4, LOTM-18-2 P. 12 HLO-075-1 LO 5, LOTM-59-2 P. 10 (3.8/3.8, 2.4/2.4) 209002K104 209002K107 ...(KA'S) QUESTION 6.03 (1.00) i The LPCS system piping inside the reactor vessel is monitored for integrity by the measurement of the pressure differential between the LPCS injection piping inside the drywell and: (1.0).
- a. the A RHR system inj ection piping inside the drywell.
- b. the above core plate pressure tap inside the drywell.
- c. the HPCS injection piping inside the drywell.
~
- d. the SLC inj ect ion line below core plate pressure tap inside the drywell.
i ANSWER 6.03 (1.00) f a (1.0) REFERENCE RBS, HLO-020-1 LO 1, LOTM-17-2 P. 4 GGNS, OP-LO-SYS-LP-E21, LO 3.b, P. 8 (3.0/3.2) 209001K404 ...(KA'S) QUL3 TION 6.04 (1.00) List FOUR conditions which allow retraction of the SRM detector without actuating a rod withdrawal block. (INCLUDE ANY APPLICABLE SETPOINTS) (1.0) (
Et__EL6NI_SISIEUS_DESIGNt_QQNIBQLt_8HQ_INSIBubENI6IIQN PAGE 14 ANSWER 6.04 (1.00)
- 1. The SRM is indicating greater than 100 counts.
- 2. .IRMs are on range 3 or above.
- 3. The reactor mode switch is in run.
- 4. The SRM is manually bypassed.
(4 0 0.25 each) REFERENCE RBS, HLO-051-1 LO 12, LOTM-9-2, P. 20 GGNS, OP-LO-SYS-LP-051-1-03, LO 4a, PP. 13-15 S0-051-1/REV 2, P. 11 (2.8/2.9, 3.7/3.7, 3.2/3.2) 21G004K401 215004K404 215004K406 ...(KA'S) QUESTION 6.05 (2.50) , List FIVE controls available to the control room operator which would stop Recirculation Flow control Valve movement if the valve starts to ramp open while operating in Loop Manual Control. Refer to Figure 1 for a drawing of the H13-P614 Controls and Indications. (2.5) l ANSWER 6.05 (2.50)
- 1. Shutdown HPU using manual pushbutton at H13-P680 (0.5) ,
- 2. Shutdown HPU using the 3HUTDOWN pushbutton at at H13-P614 (0.5) f 3. Shutdown HPU at H13-P614 by placing both subloops in Maintenance (0.5) l I
- 4. Shut Orywell isolation valves at H13-P808. (0.5) l l S. Depress the Drywell Pressure Test pushbutton at H13-P614. (0.5) k 1mnopq rs t uvw x y6. Depress the CRVICS BOP pushbuttons on t l ho P680 panel (0.5) l (5 required at 0.5 each)
REFERENCE RBS, HLO-058-2 LO3 & 4, LOTM-8-2 PP 5-8 & Figure 4 (3.8/3.5, 2.6*/2.6*, 2.8/2.8) 202002G009 ...(KA'S) (***** CATEGORY 06 CONTINUE 0 ON NEXT PAGE *****)
e _, 62__EL6NI_SISIEUS_DESIGNt_CQNIBQLt_6ND_INSIBudENI6IIQN PAGE 15 QUESTION 6.06 (2.00) Given the following data for APRM channel C: LPRM level: A B C D LPRMs assigned: 4 4. 4 4 LPRMs bypassed: 1 3 1 0 c.. If APRM Channel Selector Switch on the local instrument is placed to the COUNT-position, what would be the expected meter reading? Describe HOW you arrived at your answer. (1.0)
- b. Based on the above information, is APRM C operable? Answer YES or NO i and EXPLAIN WHY. (1.0)
ANSWER 6.06 (2.00)
- c. (11 operable channels)(5% per operable channel) = 55% !
J OR (55/125)(10v) = 4.4V Either percentage or voltage is acceptable (1.0)
- b. No (Inoperable) (0.5) due to < 2 LPRMs per level (0.5)
REFERENCE RBS Technical Specit ications 3.3.1, pg 3/4 3-5 LOTM-12-2, pg 4 (3.6/3.6, 3.2/3.4, 3.4/4.1 215005A208 215005G011 215005K104 ...(KA'S) 1 (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) l
I 6t__ELeNI_HYSIEdf..QESIGNt_QQNIBQLt_eNQ_INEIBudENI8IIQN PAGE 16 QUESTION 6.07 (3.00) With the unit operating at 75% power, with recirc control in Flux Manual, en electrical fault causes the maximum combined flow setpoint to drop to its minimum output. How will the following parameters respond after the fault and why? Refer to Figure 2 and limit your discussion to response within one minute of the fault. Assume no operator action,
- c. Turbine control valve ,sosition.
- b. Bypass valve position.
- c. Rx power and pressure.
ANSWER 6.07 (3.00)
- a. TCV's will close to minimum (30%) steam flow (0.5). The TCV LVG passes a MCF signal of 30% rather than the higher signal from the pressure controller (0.5). Value not required.
- b. The BPV's will remain closed through the t r ans ie nt . (0.5) The MCF summer will send a zero signal to the BPV LVG. (0,5)
- c. (Rx power and pressure will rapidly increase following the fault.) The reactor will scram on high flux /or high pressure due to TCV closure (0.5]. Reactor pressure will be controlled by the TBV's. [0.5)
REFERENCE RBS, HLO-059-1 LO6, LOTM-27-2 pp. 70-71 and Figure 5 (4.1/4.1, 4.2*4.3*, 4.1*/4.1, 3.7/3.7) 241000K301 241000K302 241000K306 241000K308 ...(KA'S) f QUESTION 6.08 (1.50)
- a. What are FOUR parameters monitored by the instruments mounted on the meteorological tower? (1.0)
- b. What is the emergency power supply for the meteorological tower instrumentation? (C.5) l
62__EL8NI_SYSIEdS_ DESIGN &_GQNIBQLt_6NQ_INSIBudENI6IIQN PAGE 17 I
.)
ANSWER 6.08 (1.50) i l o. 1. . Temperature l 2. Wind Speed i
- 3. Wind Direction
- 4. Dewpoint L
- 5. Precipitation (4 4 0.25) l
- b. Emergency generator at the Met tower (0.5).
l l REFERENCE ! RBS, HLO-070-1 LO3, LOTM-74-2 p. 1 l (2.5/3.8, 3.8/4.3*) j 262001K101 295017A205 ...(KA'S) l l QUESTION 6.09 (2.50) List TEN (10) Balance of Plant (80P) systems which penetrate primary l containment and receive isolation signals from CRVICS on low reactor water lovel at -45" or high drywell pressure st 1.69 ps ig. (2.5) l i ANSWER 6.09 (2.50) Any 10 of the following: 1 l 1. Instrument Air l 2. Service Air l 3. Fire Protection i 4. Service Water i
- 5. Floor Drains
! 6. Equipment Drains
- 7. RPCCW l 8. Hydrogen Mixing
! 9. Fuel Pool Cooling and Cleanup I 10. Chilled Water l 11. Condensate Makeup l 12. Containment and Drywell Purge l ~13. Reactor Plant Sampling
- 14. Annulus Pressure Control (10 required at 0.25 each) l REFERT.NCE l RBS, HLO-062-1 LO 7 & 8, LOTM-51-2 pp. 51 & 52 (3.1/3.2, 2.8/3.1, 2.7/2.9, 3.6/3.7) i 2230020004 223002K110 223002K114 223002K119 ...(KA'S) l
(***** CATEGORY 06 CONTINUED ON NEXT PAGE ** **)
64._EL8HI_SISIEdS_ DES 10Nt_CQNIBQL&_880_INSIBudENI8IIQN PAGE 18 QUESTION 6.10 (2.00) Answer the following questions concerning the Reactor Recirculation System:
- e. While MANUALLY transferring Recirculation pumps from fast speed to slow speed, CB-5A opens but CB-5B does not. Explain the response of the control system including the final expected condition of the two pumps. (1.0)
- b. List FOUR signals, including setpoints, which will cause the Recirculation Pumps to AUTOMATICALLY transfer from fast speed to slow speed. (1.0)
ANSWER 6.10 (2.00)
- a. Pump A will transfer to LFMG) (0.25). The INCOMPLETE SEQUENCE RELAY will operate (0.25] (after the 40 see time delay) and trip CB-58
[0.25). Pump B will then coast to a stop (0.25).
- b. 1. FW flow less than 30% for 15 seconds. (0.25]
- 2. RPV Level III. [0.25]
- 3. Steam dome to recirc suction at less than 8 degrees F for 15 seconds. [0.25]
- 4. EOC-RPT turbine stop valve or control valve closure with power greater than 40%. (0.25]
REFERENCE RBS, HLO-005-2, LO 4 & 15, LOTM-7-2, pp. 18 & 19 (3.3/3.6, 3.1/3.2) 202001K402 202001K416 ...(KA'S) QUESTION 6.11 (3.00) List the THREE vessel thermal shock interlocks and explain the purpose of each including vessel and system components each is designed to protect.
~
6t__EL88I_SYSIEd1_DE210Nt_CONIBQLt_880_INSIBudENI8I1ON PAGE 19 ANSWER 6.11 (3.00)
- 1. RPV bottom head drain / steam dome delta T must be less than or equal to 100 degrees F (0.5) to prevent thermal shock to the bottom head (due to stratification at low flow conditions) (0.5).
- 2. Steam dome /recirc pun.p suction delta T must be less than or equal to 50 degrees F CO.5) to prevent thermal stress on the pump, piping, and vessel riser nozzles (0.5].
- 3. Loop suction differential temperature between A and B loops must be less than or equal to 50 degrees F C0.5) to protect the pump from thermal stress (excessive heatup rate) and prevent damage to the bearings and seals. (0.5)
REFERENCE RBS, HLO-005-2, LO 10, LOTM-7-2, pp. 13 & 14 (3.1/3.3, 3.2/3.5) 202001K117 202001K412 ...(KA'S) QUESTION 6.12 (2.00) List FOUR expected automatic actions tlist occur on a loss of the flow reference signal to APRMs A & E with the reactor at rated power. (2.0) ANSWER 6.12 (2.00) Any 4 of the following at 6.5 each:
- 1. APRM A or E upscale trip or INOP alarm / console light.
- 2. RPS trip on affected channel.
- 3. Rod withdrawal block.
- 4. APRM Recirc flow upscale alarm.
- 5. Flow reference upscale console light.
- 6. Neutron Monitoring System alarm.
REFERENCE RBS, HLO-054-1, LO 6, LOTM-12-2, pp. 6, 7, a 13 ARP-680-06, pp. 3, 4, & 14 (3.4/3.5, 3.3/3.4) 202001K123 215005K116 ...(KA'S) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
6t__EL8NI_SYSIEUS_DESIGNt_QQdIBQLt_860_INSIBudENI6IION PAGE 20 QUESTION 6.13 (1.00) State the difference between control rod insertions using the IN-TIMER SKIP cnd the INSERT pushbuttons. (1.0) ANSWER 6.13 (1.00) The IN-TIMER SKIP pushbutton bypasses the timer circuit thereby oliminating the settle function (1.0). REFERENCE RBS, HLO-057-1 LO 12, LOTM-6-2 P. 28 (3.2/3.2) 201002K408 ...(KA'S) l l l l (***** END OF CATEGORY 06 *****) l
ZA__EBQQEQUBE1_:_NQBdeLi_8HNQBdekt_EdEBQEUQ1_88Q PAGE 21 86DIQLQQ108L_QQUIBQL QUESTION 7.01 (1.00) According to FHP-0003, Refueling Platform Operation, what action (s) should the platform operator take if pool level begins to drop with a fuel bundle cuspended from the grapple? Be specific. (1.0) ANSWER 7.01 (1.00) Immediately place the bundle in a fuel rack or in the reactor vessel (whichever is closer) (1.0) REFERENCE RBS, HLO-227-1 LO 2, FHP-0003 REV 2 P. 2 (3.1*/3.5*) 234000G014 ...(KA'S) < QUESTION 7.02 (1.50) The Plant Startup procedure, 00P-0001, requires all turbine bypass valves to be fully closed when withdrawing control rods with reactor power above the Low Power Setpoint. What is the basis for this requirement? (1.5) r ( ANSWER 7.02 (1.50) The Rx power input for rod control is determined from first stage pressure (0.5). With the bypass valves open, the RCS senses Rx power as less than octual and the potential exist for a non-conservative rod withdrawal.(1.0) l REFERENCE RBS, HLO-500-0 LO 4, GOP-0001 P. 3 (3.8/3.9, 3.7*/3.8*) l 241000K101 24100DK405 ...(KA'S) l 1 l QUESTION 7.03 (2.00) Duting Cold Shutdown conditions with the RPV head in place, what are TWO CONDITIONS that shall be met to minimize temperature stratification which coy cause coolant temperatures to exceed 212 deg F and inadvertent RPV pressurization? (2 0)
Zi_..EBOCEQUBES_: UDBdeLi_6BU9806Lt_EUEBQEUCX_80Q PAGE 22 , 86919L90106L CQUIB0L ! ANSWER 7.03 (2.00)
- 1. At least one shutdown cooling loop should be in operation.
(Forced circulation maintained.)
- 2. RPV water level shall be maintained greater than 75" (using Shutdown Level Instrumentation when RR pumps are secured).
- 3. In the event forced circulation is unavailable (RR pumps off) and water level is at or below the Minimum Natural Circulation Level for any reason, then periodic monitoring of vessel metal temperatures above and below the intended water level should be initiated.
(2 required at 1.0 each) REFERENCE RBS, HLO-021-1, SOP-0031 Rev S P. 4 (2.8/3.1, 3.1/3.2, 3.G/3.6) 205000K503 205000K603 205000K604 ...(KA*S) QUESTION 7.04 (1.00) During initial startup of an RHR pump in the Shutdown Cooling mode, the operator is cautioned to establish greater than 750 gpm within 8 seconds of pump start. Why is this important? (1.0) ANSWER 7.04 (1.00) (If greater than 750 gpm is not established within 8 seconds), the pump minimum flow valve to Suppression Pool (1E12-F064A/B) will open (0,5) dumping reactor water to the suppression pool. [0.5) REFERENCE RBS, HLO-021-1 LO 10, SOP-0031 p. 16 (2.7/2.8, 3.6/3.2) 305000G013 205000K407 ...(KA'S) ( (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) ,
Zt__EBQQEQUBES_:_UQBd6Lt_6DNQBdekt EdEBQENQ1.800 PAGE 23 86019L00196L_Q9N180L QUESTION 7.05 (3.00) Answer the following questions in accordance with AOP-0009, Loss of Normal Service Water.
- c. When is an immediate manual reactor scram NOT required? (1.0)
- b. If header pressure drops to 76 psig, non-essential service water loads will be isolated. List EIGHT non-essenti 51 loads isolated. (2.0)
ANSWER 7.05 (3.00)
- c. If turbine 1 not on the line (0.5] and bypass valves are not being used to control reactor pressure (0.5].
- b. Any 8 of the following e 0.25 each:
- 1. Turbine lube oil coolers.
- 2. Generator hydrogen coolers.
- 3. Generator alternator cooler.
- 4. CCS heat exchangers.
, S. Turbine building chillers.
- 6. Radweste building chillers.
- 7. Steam bypass HPU coolers.
- 8. Turbine EHC coolers.
- 9. Turbine sample room chiller.
- 10. SJAE intercoolers.
REFERENCE RBS, HLO-510-0 LO3 & 4, A0P-0009 pp. 2&4 (3.3/3.4, 3.4/3.6, 3.3/3.4) 295018K201 295018K202 295018K302 ...(KA'S) QUESTION 7.06 (2.00) Excluding control room annunciators and condenser vacuum, what are FOUR (4) symptoms of a loss of main condenser vacuum? (2.0)
Zi__EBQQEQUBES_:_NQBdelt_8BNQBdekt_EdEBQENQX.6NQ PAGE 24 BeQ10LQQIQeL_QQUIBQL ANSWER 7.06 (2.00) Any 4 of the following at 0.5 each:
- 1. Circulating Water Pumps or fans Tripped.
- 3. Loss of Oland Seal Steam pressure.
- 3. Increasing Off Gas System Flow (increased air inleakage).
- 4. Blockage in Off Gas System.
- 5. Circulating Water rupture in Turbine Bldg.
- 6. Generator load decrease
- 7. Turbine trip
- 8. MSIV or TBV closure REFERENCE RBS, HLO-510-0 LO2, A0P-0005 p. 2 (3.7/3.8) 395002G011 ...(KA'S)
QUESTION 7.07 (1.00) Why does A0P-0004, Loss of Offsite Power, caution the operators to refill l the f uel oil storage tanks on a staggered basis for a power outage expected I to exceed 24 hours? (1.0) l l ANSWER 7.07 (1.00) , i Prevent loss of more than one diesel if contaminated fuel oil is used (to , refill the tanks) (1.0). , REFERENCE . RBS, HLO-510-1 LOS, A0P-0004 p. ' 8 (3.2/3.6) 3950030007 ...(KA'S) QUESTION 7.08 (1.00) , SOP-0007, Condensate System procedure, cautions the operators that the I canual Startup Bypass valves (1CNM-V142 and V143) used to preheat the LP , heater string must be closed after the LP Heater Inlet Isolation Valves ! (1CNM-MOV33A and 338) are opened during startup. What are the potential , consequences of failing to close V142 and V1437 (1.0) l l
l o . Zi__EBQCEDUBES_:_NQBU8La_8HNQBd8Lt_EdEBQENG1.8HD PAGE 25 B8QIQLQQ198L_GQUIBQL ANSWER 7.08 (1.00) Heater string will not be isolated on heater high level (0.5) and water: induction into the turbine may occur (0.5). REFERENCE RBS, HLO-046 1 LO9, SOP-0007 p. 2 4 (3.2/3.2, 2.6/2.6, 3.1/2.0) 3560000010 256000K301 .. (KA'S) QUESTION 7.09 (2.00) List TWO Secondary Containment conditions which require emergency depressurization of the Reactor Pressure Vessel per E0P-1. (2.0) ANSWER 7.09 (2.00)
- 1. If a primary system is dischar9ing into secondary containment, any ares temperature, radiation level, or water level is exceeding its I maximum esfe operatin9 value (from Table 6 of E0P3) (1.0).
- 2. Offsite radioactivity release rates requiring a etneral emergency with a primary system discharging outside primary and secondary containment (1.0)
REFERENCE RBS, E0P-1 (3.6/3.9, 3.6/4.5*, 3.3/3.5, 3.5/3.8, 2.6/2.8) 295032K301 295033K301 295036K301 29503(K101 295038K304
...(KA'S)
(***** ' CATEGORY 07 CONI!NUED ON NEXT PAGE *****)
Zi__EBQGF9UBES_:cbQBdelt_6RNQBb8Lt_EdEBQENC1_6NQ PAGE 26 BeQ10LQQ198L_GQNIBQL _ J l QUESf!0N 7.10 (1.50) Answer the fol?owing questions concerning ALARA and radiological controls at River Send TRUE or FALSE: (1.5)
- e. The ALARA ceJncept has been applied if, instead of reaching dose limits during the first week of a quarter, a worker reaches his dose limit on the last dcy of the quarter.
- b. Personnel who enter the RCA on an Access RWP are tesued a color coded sticker by RP that indicates they have the proper training and dosimetry requirements for entry.
- c. Shif t Supervisor's approval is not required for RWP revisions.
ANSWER 7.10 (1.50)
- 1. False (0,5)
- b. TruG [0.5)
- c. Tr9e (0,5)
REFLRErJCE RBS, General Employee Training, pp. 7-43 R3P-0200 pp. 18 and 22 (3,3/3 8, 3.3/3.6) 2^v)001K103 294001K104 ...(KA'S) QUESTION 7.13 (1.00) Determine if th4 following otatements concerning SOP-0092, Offges System, ate TRUE or FALSE. (1.0)
- o. TL6 charcoal beds are bypsused during initial system startup to p r e.e.it radioactive isotope releases due to high system air flows.
L. G*ycol syntems are potentially contaminated and any leakage should be routed 30 the .iearest floor drain for collection in Radweste.
. _ _ _ , _ , . - . _ _ . _ . - , . _ _ . . _ _ _ - . _ _ ~ . -
I Zt__EB9CE99BES_:_UDBdeLt_6BUQBd6Lt_EdEBQENQX_68Q PAGE 27 l 86019699106L_G9 NIB 9L ANSWER 7.11 (1.00)
- 3. False (0.5)
- b. False (0,5)
REFERENCE RBS, HLO-047-1 LO 7. SOP-0092 Rev. 4 P. 3 (2.6/2.8, 3.1/3.2, 2.3/2.5) 268000K115 2710000010 271000K511 ...(KA'S) 0,UESTION 7.12 (2.uo) SOP-0052, HPCS Diesel Generator, cautions the operator to depress the
S TOP" pushb utton cn the engine panel or P601 after the diesel has tripped cutomatically. What function does depressing the stop button perform and why is this action necessary? (2.0)
ANSWER 7.12 (2.00) Start relay (K1) seals in on diesel start and depressing "STOP" deenergizes this relay. [1.0) Failure to do eo results in an automatic r0 start of the engine when the trip condition clears. ri,o] REFERENCE RBS, HLO ^6-1 LO 10, SOP-0052 Rev. 4 p. 3 (3.8/3.7, 3.2/3.6) ' 264000G010 264000K408 ...(KA'S) QUESTION 7.13 (2.00) E0P-0001, RPV Control Emergency Procedure, conta!ns an enclosure which provides for alternate injection paths for sodium pentaborate should Standby Llauid fail to inj ect . Briefly describa how boron is inj ected via this method incleding plant systems which can be utilized. (2.0)
- _ _ - - _ _ _ _ _ _ _ _ - - - _ - - _ .- --- --- - . - - - - - - - .--------_J
Z2_ _ P B Q Q E Q U B E 3_ :_N Q B d 6 L t _6 H N Q B d 6 L t _ E d E 8 Q E N Q X,,6 N Q PAGE 28 86DIQL2QIQ66-QQNIBQL
<NSWE' 7.13 (2.00) f , odium pentaborate is mixed in a drum (0.5) and pumped into an injecting ECCS system vent (0,5) using a portable hydro pump (0.5). Systems that can ,
be used are RHR A, RHR B, RHR C, LPCS, OR HPCS [5 e 0.1 EACH). [ l } REFERENCE j- RBS, E0P-0001 REV 7, PP. 65-66 (3.4/4.1) f 295037K213 ...CKA'S) l l QUESTION 7.14 (2.50) List the entry conditions (including setpoints) for E0P-0002, Primary Containment Control Emergency Procedure. (2.5) ANSWER 7.14 (2.50)
- 1. Drywell/ Containment differential pressure above 1.68 psid.
- 2. Containment pressure above 15.7 psia.
- 3. Suppression Poc1 level not between -0.25 ft. and +0.25 ft.
- 4. Primary Containment hydrogen concentration above 0.5%.
- 5. Drywell hydrogen concentration above 0.5%.
[5 9 0.5 each) REFERENCE RBS, E0P-0002 REV 6, P. 2 (4.1*/4.3*, 4.2*/4.5*, 4.2*/4.5*, 4.3*/4.5*) 295007G011 295010G011 295029G011 295030G011 ...(KA'S) QUESTION 7.15 (1.50)
- a. Explain the basi;. for the Heat Capacity Temperature Limit graph which is part of the CDP flow charts. (1.0)
- b. What suppression pool level is assumed as the basis for this (0,5) graph?
- ~ _ - _ - _ _ - - _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
t Zz__.PB9CEDUBgS_:_N9Bd6Lt_6BNQBd6Lt_EdEBQENQX_6NQ PAGE 29 B6019L99106L_QQNIBQL ANSWER 7.15 (1.50) )
- a. The line on'the graph represents the maximum suppression pool temperature at which actuation of ADS will not rescit in eitt.er unstable steam condensation at or near the SRV discharge devices (0.5]
or e suppression pool temperature in excess of the containment air space design temperature while the RPV is above the shutdown cooling interlocks. [0.5]
- b. At the low level LCO. [0.5)
REFERENCE E0P Bases Documents (4.3*/4.6*) 295027G012 ...(KA'S) (***** END OF CATEGORY 07 *****) _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _. I
8t __6031NISIB6IIVE_EBQQEDMBESt_LQNRIIIQUSt_8NQ_LIMIIoIIQNg PAGE 'O QUESTION 8.01 (3.C3) Given the below listed facility conditions, Reactor power = 85% Reactor Flow = 90% 12th day of operation at these conditions. One ADS valve (1821*RVF051G) has failed. The SS has declared it inoperable because he/she does not believe it will function properly.
- c. List the Technical Specification L00(s) exceeded and the appropriate i Action statement (s) allowing continued plant oparation, (2.0)
- b. During this period of operation, the Maintenance Supervisor brings'to you.- attention the load test surveillance'for diesel 10 was not perf orroed and the grace period has expired. List the Technical i Specification LC0(s) exceeded and the appropriate Action statement (s) allowing centf.nued plant operation. (1.0)
ANSWER 8.01 (3.00) -
- e. 3.5.1.a.3 (0.25) 3.5.1.b.2 (0.25)
Action e.1 (0.5) 3.4.2.2 (0,5) Action a. (0.5)
- b. 3.8.1.1.b (0.5)
Action d (0.5) REFERENCE River Bend Station, Technical Specifications River Bend Station, HLO-406-0, Lesson objective #2 K3.01 = 4.4; SG = 4.1: ADS-SG = 4.3 (***** CATr. GORY 08 CONTINUED ON NEXT PAGE *****)
e
-az__6Db1812T.861;yg_PBQQggyBE$t_QQUQIIIQN$t_6NQ_LIMII6IIQU$ PAGE 31 QUESTION 0.02 (1.50)
Answer the following with regard to the River Bend Emergency Plan Implementing Procedures dose calculation requirements: l 0 Who is responsible for the initial and follow-up offsite dose ! calculations (at least until relieved by personnel in the TSC)? (0,5) '
- b. A determination has been made that offsite dose calculations are necessary.
- 1. When should the initial assessment be completed? (0.5)
- 2. When should follow-up calculations be completed? (0.5) t AMSWER 8.02 (1.50)
- a. Shift Foreman (0.5)
- b. 1. Initial assessment should be completed within approximately 15 minutes of the declaration (0.5).
- 2. Follow- calculations should be performed approximately every 30 minute. .ollowing initial projections (0.25), or sooner if release rates change significantly (0.25).
REFERENCE RBS, EIP-2-025 Rev 2 PP. 2&3 (2.9*/4.7*, 3.83/4.3*) 294001A116 295017K102 ...(KA'S) QUESTION 8.03 (2.50) Answer the following questions pertaining to the River Bend Emergency Implementing procedures Protective Action Recommendation Guidelines.
- a. If a General Emergency is declared, what offsite protective actions shall be recommended immediately, unless other, more extensive, protactive actions have already been recommended? (2.0)
- b. When should protective attions less stringent than those recommended by the Protectiv. Action Recommendation Guideline Procedure, EIP-2-007, be recommended by the Emergency Director? C0.5)
- n. - ,, . -,v,. . , - - - .,,,.,,,,..,.n._ ._.._,,,.-n- ,.n.-.. m..
i j 8t__8001NISIB6IIVE_EB9CEQUBESt_GQNDIIIONSt_6NQ_ lib 1I6IIONS PAGE 32 ANSWER 8.03 (2.50)
- a. 1. Shelter to the 2 mile radius around the station (1.0)
- 2. Shelter to the 5 mile radius in the downwind and adj acent sectors (1.0)
- b. Only if constraints (time available, traffic congestion, weather hazards, etc) make the required actions a greater hatard to public health (0.5).
REFERENCE RSS, EIP-2-005, Rev 4, pg 6 EIP-2-007, Rev 5, pg 2 (2.9*/4.7*, 4.2*/4.4*) 294001A116 295038K102 ...(KA'S) QUESTION 8.04 (2s00) Answer the following questions in accordance with Technical Specifications requirements.
- a. What is the operator's responsibility when the Limiting Condition for Operation (LCO) cannot be met? (0,5)
- b. When is the operator allowed to enter an operating condition where an LCO cannot be met without relying on provisions of an action statement (two required)? (1.0)
- c. What 9 Non-Compliance with a Technical Specification with regard to Section 3 of the Technical Specifications? (0.5)
ANSWER 8.04 (2.00)
- e. To comply with the ACTION STATEMENT (0.5)
- b. When it is required to change modes to ,neet an action statement (0,5),
or TS 3.04 is specified as N/A (0.5),
- c. LCO's not met (0.25) and action statements not complied with within the specified time interval (0.25).
REFERENCE RBS, HLO-400-0 LO 3 & 4, T.S. Sect. 3 P. 3/4 0-1 (3.7/3.7) 2010010001 ...(KA'S) (***** CATEGORY 08 CONTINUED ON ffEXT PAGE *****)
{ 8t __6DdINISIB6IIVE_EBQCEQUBESt_CQUDIIIgust_6Hg_LIdII6II"US PAGE 33 QUESTION 8'.05 (2.00) Answer the following with regard to Temporary Change Notices (TCNs):
- a. What.is the maximum number of TCNs that can be active against a single
-procedure? (0.5)
- b. How would the next TCN be written against a procedure that already had the maximum number of TCNs allowed written against it? -(1.0)
- c. How long are "One Time Only" TCNs effective (maximum)? CO.5)
ANSWER 8.05 (2.00)
- a. 3 (0.5) '
- b. The next TCN would have to supersede one of the previous TCNs in its entirety or it would not be allowed (1.0).
1
- c. 14 days (0.5)
REFERENCE RBS, HLO-202-2 LO 5 & 6, A0M-0003 Rev 14 P.11 (2.9/3.4) 294001A101 ...(KA'S) QUESTION 8.06 (2.50) Technical Specification defines the composition of the site fire brigade in Section 6.
- c. What is the minimum number of fire brigade members required on Site at all times? (0,5) i l
- b. List FOUR shift personnel who are specifically excluded from being on j the fire brigade. (2.0) I
_ - . _. . . _ . . _ _ - _ . , - . _ _ _ , _ . ,- ._ . - . _ _ . ~ , ._ . _ _ _ _ . . . . _ _ .
n . . at__eQdINISIB6IIVE_EB9CEQUBESt_CQNDIIIONSt_8NQ_LIdII6IIGN$ PAGE 34 ANSWER 8.06 (2.50)
- e. 5 [0.5)
- b. Any 4 8 0.5 each;
- 1. Shift Supervisor
'2. Shift Technical Advisor
- 3. Control Operating Foreman
- 4. At-the-controls Operator
- 5. 1 Nuclear Equipment Operator NOTE: Two members of the minimum shift manning required for safe shutdown of the plant, i. e. NCO, NEO (answers 4 and 5 above)
REFERENCE RBS, HLO-415-0 LO6, T.S. 6.2.2.e P. 6-1 AOM-022 Rev. 10, P. 11 (3.5/3.8) 294001K116 ...(KA'S) QUESTION 8.07 (2.00) For each of the f ollowing s ituat ions, determine if a Condition Report is required to be submitted by ADM-0019. Answer Yes or No for each situation.
- a. Removal of plant equipment from service which constitutes a voluntary entry into an LCO action statement.
- b. Fire i n a. clean trash container.
- c. An individual with significant internal contamination.
4
- d. Failure of a component which has failed three times within the last month.
ANSWER 8.07 (2.00)
- a. No (0.5)
- b. Yes (0.5)
- c. Yes (0.5) d, Yes (0.5) l L -_ - - - - - - _ _ _ - - - - - - - - - - - -- - - - - - -
at__eQblNISIBoIIVE_PBQQEQQBE$t_CQUQIllQU$t_8UQ_LidlI611QU$ PAGE 35 REFERENCE RBS, HLO-205-0 LO 3, ADM-0019 Rev. 6 PP. 12-14 (2.7/3.7) 294001A103 ...(KA'S) QUESTION 8.08 (3.00) Answer the following questions pertaining to Surveillance Test Procedures.
- c. What are THREE methods available to correct a Surveillance Test i Procedure (STP) which is incorrect as written and cannot be (1.5) !
performed? I
- b. What actions should be performed if a typographical error is detected during the performance of an STP if the step is otherwise correct and acceptance criteria is not affected? (1.5)
ANSWER 8.08 (3.00)
- a. 1. Temporary Change Notice (0.5]
- 2. Preliminary Change Notice (0.5)
- 3. Procedure Revision (0.5]
- b. 1. Obtain concurrence from on shift SS/C0F (0.5)
- 2. Note the typo in the comments section of the procedure (0,5)
- 3. Initiate a comment control form and place it in the history folder (0.5)
REFERENCE RBS, HLO-221-1 LOS, HLO-202-2 LO6, Rev. 11, P. 10 & 12 (2.7/3.7) 294001K103 ...(KA'S) i QUESTION 8,09 (1.00) In order to depressurize the vessel under some plant conditions, it will be necessary to defeat isolation interlocks in accordance with approved E0Ps or AOPs.
- a. Where are the j umpers for bypassing these interlocks maintained? (0.5) b, Who controls the issuance of these jumpers? (0.5)
8t__6DdINISIB6IIVE_EBQGEQUBESt_GQNDIIIONSt_8NQ_LIdIIBIIQNg PAGE 36 i l ANSWER 8.09 (1.00)
- a. Locked storage in the Main Control Room-(0.5)
- b. Shift Supervisor (0.5)
REFERENCE RBS, ADH-0022 Rev. 10 P. 18 (2.7/3.7) 294001A103 ...(KA'S) QUESTION 8.10 (1.50) Fill in the requirements (a) - (c) in each of the following statements describing Technical Specification requirements if a safety limit is violated.
- a. The NRC operations center shall be notified by telephone as soon as possible and in all cases within (a). (0.5)
- b. The Senior Vice-President RBNG and the NR8 chairman shall be notified within (b). (0.5)
- c. A Safety Limit Violation ' Report shall be submitted to the. Commission, the NRB, and the Senior Vice-President - RBNG within (c) of the violation. (0.5)
ANSWER 8.10 (1.50)
- a. I hour (0,5)
- b. 24 hours (0.5)
- c. 14 days (0.5)
REFERENCE RBS, HLO-415-0 LO 11, T.S. 6.7.1 P. 6-13 (3.4*/4.5*) ; 212000G003 ...(KA'S) J (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
l l 81__60dINISIB6IIVE_EB9CEDUBESt_GQNDIIIQNSz_eND_LIdII6IIQN3 PAGE 37 1 1 QUEFTION 8.11 (1.00) Define each of the following per Technical Specifications:
- a. Limiting Uontrol Rod Pattern (0.5)
- b. Identified Leakage (0.5)
ANSWER 8.11 (1.00.)
- a. A pattern which results in the core being on a' thermal hydraulic limit (0.5).
- b. 1. Leakage into collection systems (such as pump seal or valve packing leaks) that is captured and conducted to a sump or collecting tank (0.25), or
- 2. Leakage into the drywell atmosphere from sources that are specif,1cally located and known not to. interfere with the operation of the leakage detection systems (0.125) or to be Pressure Boundary Leakage (0.125).
REFERENCE RBS, HLO-400-0 LO id a le, T.S. Section 2.0 (2.9/3.5, 3.3/4.1) 201001G002 2230010005 ...(KA'S) QUESTION 8.12 (2.00) What are the limits on recirculation loop flow mismatch and what is the basis for the limits? (2.0)
.at__6DMINISIBoIIVE_E899EQUBESt_CQNDIIIQNSt_6ND_LIMII6IIQN1 PAGE 38 ANSWER 8.12 (2.00)
Mismatch shall be within:
- a. 5% with core flow greater than or equal to 70% of rated (0.5) 1
- b. 10% with core flow less than 70% of rated (0.5)
(Maintaining mismatch within. limits will) ensure an adequate core flow coastdown from either recirc loop following a LOCA (1.0) a REFERENCE RBS, HLO-405-0 LO 6 a 1, T.S. 3.4.1.3 (2.6*/3.7, 3.3/4.0) 202002G005 202002G006 ...(KA'S) QUESTION 8.13 (1.00) Until the new computer is installed for Emergency Notification, how does the Shift Supervisor actuute (or cause to be actuated) the emergency sirens? (1.0) ANSWER 8.13 (1.00) Shift clerk calls Security Shift Supervisor who dispatches personnel to the EOF (Training Center) to enable the siren system (1.0). REFERENCE RBS, Standing Order #60 dated 4/26/88 (2.9*/4.7*) 294001A116 ...(KA'S) \ . . - . _ - _ - _ - _ _ _ _ - - _ - _ - _ _ - - - _ _ _ _ _ - - _ - - _ _ - - _ - - _ _- . - - -
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O C> NRC LICENSE EXN4IfLATI0tt HANDOUT EQUATIONS, 00:lSTANTS. NiD C0HVERSIONS 6'm*C*deltaT p 6 = U*A*deltaT P = Po*10 sur*(t) P = P *et /T SUR = 26/T T = 1*/p + (p-p)/X p T=1/(p-p) T = (p-p)/X p p = (Kerr-1)/Keff = deltaKe rg/Ke rr p = 1*/TXerf + jerr/(1+1T) . A = In2/tg = 0.693/tg K = 0.1 seconds-1 I = Io*e "* CR = S/(1-Keff) 2 R/hr = 6*CE/d feet Water Parameters 1 gallon = 8.345 lbm = 3.87 liters 1 t 1 ft3 = 7.48 gallons l Density @ STP = 62.4 lbm/ft3 = 1 gm/cm3 l Heat of vaporization = 970 Btu /lbm - Heat of fusion = 144 Btu /lbm 1 atmosphere = 14.7 psia = 29.9 inches Hg. Miscellaneous Conversions 1 curie = 3.7 x 101v disintegrations per second l 1 kilogram = 2.21 lbm I horsepower = 2 54 x 103 Btu /hr 1 mw = 3.41 x 105 Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Stu = 778 ft-lbf , i i
l . ~ . $, I I RIVER BEND STATION FACILITY COMMENTS ON NRC EXAM GIVEN JULY 12, 1988 l 5.07 Request additional correct answer to be: tio t to exceed 1% plastic strain of the cladding. Ref.: River Bend Station, LOTM-TH-4.9.2 pg. 5 of 5 1 6.05 Request additional correct answer to be: Depress CRVICS BOP isolation pushbuttons on the P680 panel. Ref.: River Bend Station LOTM-51-2, Table 2, pg. 29 & 31 6.09 Request additional correct answer to be:
- 1. Containment and drywell purge systens
- 2. Rx plant sampling systems
- 3. Annulus pressure control system Ref.: River Bend Station LOTM-51-2, Table 2 ;
i 6.10 Part a. Pump A will downshift to the LFMG. On depressing the Transfer switches, CB LA opens immediately and CB-1A closes. CB-2A will then close when its permissives are met. Ref.: River Bend Station LOTM-7-2, pg. 18 19 7.06 Question does not specify reference to any procedure. Answer addresses causes rather than symptoms of a loss of main condenser vacuum. Request question be deleted or acceptance of the following answer:
- 1. Generator load decrease
- 2. Turbine trip
- 3. Bypass valve closure 4, Main Steam isolation valve closure Ref.: River Bend Station LOTM 29-2, Table 2 and pg. 1
. =) . .
8.06 Part b Question addresses T.S. specifically, answer to part b references ADM-022 personnel requirements. Answer should include 2 members of the minimum shift manning who are' required for safe shutdown of the plant, i.e. i NCO, NEO.
.Ref.: River Bend T.S. Sect. 6.2.2.E 8.13 Request training center.be included in answer as location of siren actuation system. EOF is located in Training Center.
Ref.: Standing Order f60 dated 4/26/88 f.'
.-,7 .-. - ._ .- _ , - , - _ - --}}