ML20137F539
| ML20137F539 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 11/08/1985 |
| From: | Dimmack L, Dimmock L, Lang T, Mcmillin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20137F387 | List: |
| References | |
| 50-254-DL-85-2, 50-254-OL-85-02, 50-254-OL-85-2, NUDOCS 8512020049 | |
| Download: ML20137F539 (59) | |
Text
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U.S. NUCLEAR REGULATORY COMMISSION-REGION III Report No. 50-254/DL-85-02 Docket Nos. 50-254; 50-265 Licenses No. DPR 29; DPR-30 Licensee: Commonwealth Edison Company P.O. Box 797 Chicago, IL 60690 Facility Name: Quad Cities Nuclear Power Station Examination Administered At: Quad Cities Station Examination Conducted: August 27, 1985
/
Examiners: T. ang //[8!I6
\
h Dat'e /
J!Du f, L. Dimmock //br/tr Date Approved By: I 1 , C ief //[d 5 perator Licensing Section Dgte/j Examination Sumirary
~ Examination administered on August 27, 1985 (Report No. 50-254/DL-85-02)
On the week of August 27, 1985 examination were conducted for two Senior Operator Licensing Candidates and six Operator Licensing candidates.
Results: Two senior candidates and five operator candidates passed the examinations.
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REPORT DETAILS
- 1. Examiners T. Lang, Region III L. Dimmock, Region III
- 2. Examination Review Meeting Since the review policy has changed there was not a formal review. The utility had the full week to review the exam and submit comments.
- 3. Exit Meeting An exit meeting was held with the training staff and the plant Superintendent, in which all candidates who clearly passed the oral exam were divulgeo No generic problems were noted.
Resolution of comments on the Quad Cities SR0 exam of August 27, 1985. Also included are any other changes made to the answer key.
-Question No. 5.02 Comment: This answer is in error in describing the Latent Heat of Vaporization as a point.
This error exists in the lesson plan and students were made aware of this during their course, however the NRC copy of the lesson plan was not changed. The correct answer should be that the Latent Heat of Vaporization is the amount of heat required to change a liquid to a vapor at a constant pressure and temperature.
Resolution: Comment accepted answer key modified.
Question No. 5.~03 Comment: The Reactor Vessel and Internals lesson plan also lists decreased turbine efficiency and an increase in radioactivity carry over to the balance of the plant as undesirable affects of
. carry over. While all these should not be required, credit should not be taken off for listing these items.
Reference:
Reactor Vessel and Internals page 16.
Resolution: Comment accepted answer key modified.
Question No. 5.04c Comment: The term " critical quality" should not need to be mentioned in discussion of the effects of reactor pressure on critical power. This is not common terminology at Quad Cities and may be described better by explaining the enthalpy c+
saturated steam is less at high pressure as is described on the RO exam answer key.
2
4 Resolution: Comment accepted answer key modified.
Question No. 5.07.a Comment: The first sentence answers the question adequately. The second sentence is not asked for and should not be required for full credit.
Resolution: Comment accepted answer key modified.
Question No. 5.10.b.a Comment: The answer listed is T. The correct answer is F.
Resolution: Comment accepted answer key modified.
Question No. 6.01 Comment: Values should not be required.
Resolution: Values were noted in answer only in order to give credit if candidate stated value. No change to answer required.
Question No. 6.06 Comment: The operation of the IDI neater emergency drain valve should not be required in the answer to this question because the question asks for the inlet and outlet flow paths for the shell side of the ICI feedwater heater.
Resolution: Comment accepted answer key modified.
Question No. 6.07.d Comment: This answer is true, in that reference leg flashing will not occur. It seems however that the indicated level may be expected to increase due to the higher density water in the variable leg. Either answer should be acceptable if adequately explained.
Resolution: Comment accepted answer key modified.
Question No.7.05 Comment: This list is not complete. The list should be as follows:
- 1. Reactor Scram
- 2. MSIV's close
- 3. MSL drains (200-1 and 220-2) close
- 4. Recirc sample valve (202-44 and 45) close
- 5. Off gas suction valves (5401's and 5402's) close
- 6. Condenser mechanical vacuum pump trips
- 7. Off gas isolation valve (5406) closes
- 8. Off gas line drain valves (5408's) close
- 9. Off gas sample vial box suction valve closes valve numbers or valve description should be acceptable.
Resolution: Comment accepted answer key modified.
i 3
. . - - - - - - - _ - - - - _ _ - = -
1 Question No. 7.06 Coment: The question asks the meaning of the alarm tile TIP ISOLATION OFF LIMITS. The correct title of the alarm is TIP ISOLATION OFF NORMAL. Probably there will be no problem with this, however the i
question should be changed if it is in an exam j bank to prevent confusion in the future.
Resolution: Question changed to reflect coment however, i reference material should also be changed.
Question No.- 8.02.a Coment: After the procedures were sent to the NRC for i this exam this QAP was revised. This item in j
particular was involved in the revision. A copy of the revision is attached.
Resolution: Revised material will also be accepted.
Question No. 8.03.a and b i Coment: These answers are exactly correct in that they ask who must " concur" with the changes. The QAP however also states that procedures that do not change the intent must be reviewed by the Onsite Review and approval by the Station Superintendent within 14 days of implementation and also that those that do change the intent must be author-ized by the Station Superintendent prior to implementation. This information is beyond the scope of the question and should not be considered wrong if it is included.
Another concern with this question is that the exact table number and column number should not be required. If a student describes the table and/or column this should be acceptable.
Resolution: The review within 14 days is beyond the scope of i the exam and was not listed in the answer key.
Credit will be.given if stated correctly. Also, table numbers were not required as long as they were mentioned.
Resolution of coments on the Quad Cities R0 exam of August 27, 1985. Also included are any other changes made to the answer key.
, Question No.1.1 Coment: The Reactor Vessel and Internals lesson plan (page
- 16) lists two additional undesirable affects of carryover: (1) an increase in radioactivit carryover to the balance of the plant and (y) 2 a decrease in turbine efficiency. The HTFF also lists contamination of pure water by condenser breakdown (page 25). These answers should be acceptable if listed.
Resolution: Coment accepted. Answer key modified.
. 4
.- - - .. ~- - _ _ ~
, Question No. 1.8.b.2.
Comment: The answer listed is T. The correct answer is F. MFLPD being within limits ensures LHGR limits are met.
Resolution: Consent accepted. Answer key modified.
Question No. 1.12. Comment: In the explanation using the term " headless" vice "DP losses" should be considered acceptable.
Resolution: Comment accepted. Answer key modified.
Question No. 2.3.a Comment: While this may result in a potential airborne activity in the HPCI room, it may also result in high room temperatures resulting in a Group IV isolation. Either of these answers should be acceptable.
Resolution: Comment accepted. Answer key modified.
Question No. 2.3.d Comment: This answer is correct for normal plant
. pressure. If a student, however, explains that the HPCI will be at minimum speed, and if plant pressure is low enough, some injection may occur, this should be considered correct.
Resolution: Comment accepted. Answer key modified.
Question No. 2.11. Comment: The operation of the IDI heater emergency drain valve should not be required in the answer to this question because the question asks for the inlet and outlet flow paths for the shell side of the ICI feedwater heater.
Resolution: Comment not accepted. The question asked for the inlet flow path to the ICI heater. This flow path is diverted thru the 1D1 emergency drain valse when the 1D1 normal drain valve closes and as such is a required part of the answer.
Question No. 3.1.d. Comment: This answer is true, in that reference leg flashing will not occur. It seems however that the indicated level may be expected to increase due to the higher density water in the variable leg. Either answer should be acceptable if_
adequately explained.
Comment accepted. Answer key modified.
{- Resolution:
Question No. 3.3.c.and e.
Comment: In both these answers " low steam flow" should be
" low steam pressure".
5
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.- - - < . , - - m _ ,_,-, --
Resolution; Comment accepted. Answer key modified.
2 Question No. 3.5.b.
. Comment:
Resolut.on: No comment was made but a spelling correction was made to change "on" to "one". Required part of the answer.
Question No. 3.8.b.
Comment: The answer key says any "4" however, it appears 4 that LPCI (RHR at Quad Cities), CS, HPCI and RHR are considered as one answer. These should be
, considered separate answers since the question asks for " systems" that use reactor pressure.
Additionally, the process computer and the FWLC system (pressure compensation for GEMAC level indicators) should be added to the list of possible systems.
Resolution: The answers were always intended to be considered as separate. The two systems mentioned were added to answer key as well as the CRD system.
Question No. 3.9. Comment: Part "a" of this answer should not be required at all. To say that it bypasses all insert blocks except the RWM insert block is unnecessary because the RWM is the only system that applies insert blocks. To say that it bypasses any select block is totally irrelevant because if a select block is present, no rod may be selected so the " emergency in" would not even be used.
Credit for this question should be divided between b and c.
Resolution: Conment accepted. Answer key modified.
Points distribu?,ed to b and c.
Question No. 4.3 Consent: Instead of
... (C) incoming voltage should be slightly i (D) Higher than the _(E) running voltage.,
... (C) running voltage should be slightly (D) lower than the ]_EJ incoming voltage, should also be acceptable Resolution: Concent accepted. Answer key modified.
Question No. 4.4.b. Comment: Credit should also be given if a student gives
- the formula of effective half life being equal i
i 6
,o to the sum over the product of the biological half life and the radiological half life because the question asks to define or explain.
Resolution: Comment accepted. Answer key modified.
Question No. 4.4.c Comment: Credit should also br given if a student gives a formula of REM is equal to RAD times the quality factor because the question asks to define or explain.
Resolution: Content accepted. Answer key modified Question No. 4.5.a. Comment: The question asks how a reactor s. ram is initiated from outside the control room on a control room evacuation. However, the answer key lists how an isolation is initiated. The answer should be that the scram is initiated from the auxiliary electric room by manually tripping the circuit breakers in the RPS distribution panel.
Resolution: Conment accepted answer key modified.
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/Y)A S TE R U.S. NUCLEAR REGULATORY COMMISSION
, , REACTOR OPERATOR LICENSE EXAMINATION FACILITY: Quad Cities REACTOR TYPE: BWR DATE ADMINISTERED: August 27, 1985 EXAMINER: L. Dimmock APPLICANT:
INSTRUCTIONS TO APPLICANT:
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 parei. theses after the question. The passing grade requires at least 70% ir, each category and a final grade of at least 801.
i of Categm'.- % of Applicant's Category
_Valut Total Score Value Category 2E 25 1. Principles of Nuclear Power Plant Operations, Thermo-dynmics, Heat Transfer and Fluid Flow 25 25 2. Plant Design Including Safety and Emergency Systems 25 25 3. Instruments and Controls 25 25 4. Procedures - Normal, Abnormal, Emergency and Radiological Control 100 IOC TOTALS Final Grade i
{ All work done on this exam is my own, I have neither given nor received aid.
j Applicant's Signature i
L
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SECTION 1 - PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONS, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 1.1 Explain the difference between carry over and carry under. ;
- Include in your answer any negative effects on plant i
~
equipment. (2.0) 152 For the following conditions, state if the FUEL TEMPERATURE COEFFICIENT becomes More or Less negative.
j
- a. Increase in fuel temperature (0.5)
- b. Increase in moderator temperature (0.5)
- c. Increase in void fraction (0.5) 1.3 Explain the effects of increasing the following core parameters on steady state critical power.
- a. Core flow (1.0)
- b. Inlet subcooling (1.0)
- c. Reactor pressure when above 800 psig '(1.0)
W 90 % w 1.4 You increase core power by pulling control rods around the center fuel bundle. Assuming that recirculation speed is kept constant would the flow through the center bundle increase, decrease, or stay the same? Explain your answer. (2.0) 1.5 If equilibrium xenon is obtained (the reactor has been operated at constant power for many hours), and the reactor power is doubled, will the new equilibrium xenon concentration be twice as great? Explain your answer. (2.0) 1.6 Prior to startup (all rods full in), the SRM count rate '
is 10 CPS and K effective is 0.96 _
- a. If control rods are pulled to give a AK of +0.035, what count rate on the SRMs should be expected when the period becomes infinite? (1.0)
- b. If additional control rods are pulled to give a AK of 0.003, would the time required to reach an infinite period be greater or less than the time in part "a"?
Give the reason for your answer. _ (1.0)-
1.7 Does the centerline temperature of the fuel pin located closest to a control rod change when the control rod is moved when operating near full power? Explain your answer using general concepts of heat transfer. (2.0)
- ', 1.8 You notice that MAPRAT is 1.02
,f Is the MAPRAT of 1.02 conservative? Explain
- a. :
(.75) your answer. ,
- b. In regards to MAPRAT which of the following
. statements are True and which are False.
4-
- 1. MAPRAT maintained within limits ensures that
' transition boiling will not occur in 99% of the fuel bundles. ( 25)
- 3. Maintaining MAPRAT limits ensures that peak clad temperature will not reach 2200'F during a LOCA. (.25) 1.9 a. List the three (3) reactivity coefficients in a BWR at 100% power and give a'pproximate values for each. (1.5)
- b. What effect (Increase, Decrease, or No Effect) do each of the three coefficients have on total core reactivity following a safety / relief valve failing open? Briefly explain why the dominant coefficient affects reactivity in the manner you indicate. (2.0) en6 1.10 Which of the following statements best describes the conditions known as " condensate depression"?' (1.0)
- a. Can lead to condensate pump civitation if condensate depression is too great.
- b. Decreases as hotwell level rises.
x
?c. Reduces Rankine cycle efficiency.
f .
.d. Increases as condensate temperature increases.
l l 1.11 As core exposure increases, plutonium-239 (Pu-239) concentration increases:
- a. Briefly explain the processes by which this buildup l occurs..(Note: a reaction-decay chain equation may
- (.75) be used.)
- b. Explain the effect on reactor behavior caused by (1.75)
~
the Pu-239 buildup.
SECTION 1 CONTINUED ON NEXT PAGE l 2 -
l
1.12 See attached figure 2 " Pressure - Steam Flow Relationship."
Why does the reactor pressure increase so much more from 0 to 100% steam flow than the turbine throttle pressure doesr? (1.0) 1 ,13 What are two (2) design features of your recirculation systepi (1.0) that assure an adequate NPSH for your recirc pumps?
END OF CATEGORY
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SECTION 2 - PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 2.1 What six (6) conditions must be met for the standby i Reactor Feed Pump to auto start? i (3.0) 222 If a complete loss of Instrument Air were to occur with e the plant operating at full power and with no operator action, what would be the effect on the following components or systems? (Note: Limit your answer to effects caused directly by instrument air only.) <
- a. CRD hydraulic flow control valve (0.5)
- b. CRD hydraulic instrument volume valves (0.5)
- c. Standby liquid control system (0.5)
- d. Main feedwater pump minimum flow valve (0.5)
- e. Reactor building to torus vacuum breaker (air operated) (0.5) 2.3 For each of the HPCI (High Pressure Coolant Injection) System component failures listed below, STATE WHETHER OR NOT HPCI WILL AUTO INJECT into the reactor vessel, IF IT WILL NOT INJECT WHY, AND IF IT WILL INJECT, provide ONE POTENTIAL ADVERSE EFFECT OR CONSEQUENCE of system operation with the failed component.
Assume NO OPERATION ACTION, and the component is in the failed condition at the time HPCI received the auto initiating signal.
- a. The GLAND SEAL EXHAUSTER fails to operate. (1.0)
The turbine AUXILIARY LUBE OIL PUMP fails to operate. (1,0) b.
- c. The MINIMUM FLOW VALVE fails to auto open (stays shut) when system conditions require it to be open. (1.0)
- d. The HPCI pump DISCHARGE FLOW ELEMENT output signal to the HPCI flow controller is failed at its maximum (1.0) output.
2.4 St' ate what problem would be associated with each of the following conditions in the CRD system:
Scram outlet valve fails to open on a scram. (1.0)
- a. _
- b. Failure of both CRD hydraulic pumps (two problems required). (2.0) 2.5 What are 5 conditions which will automatically trip the Reactor Building Ventilation supply and/or exhaust fans? (2.5)
~
2.6 Indicate if the following statements regarding the Fire Protection System are TRUE or FALSE.
~
- a. The wall-mounted hose-reel CARD 0X assemblies located throughout the plant each have a separate C02 storage ;
T, tank. i (0.5)
! b. The foam suppression system is used to protect the i recirc MG sets and the feedwater pump area. (0.5) 2.7 After LPCI loop selection logic has determined the intact recirc loop following a pipe break, what automatic valve actuations will occur to allow the LPCI system to perform its function? (1.5) 2.8 What are the two purposes of the third stage SJAE in the off gas system? (1.0) 2.9 What are eight (8) different indications available to you in the control room that could be used to determine that a reactor coolant leak was occurring in the drywell? (2.0) 2.10 In the hydrogen seal oil system the main seal oil pump normally supplies oil to the shaft seals. If the main seal oil pump were to fail, what are the two backups that would supply oil to the shaft seals? (1.0) 2.11 What will happen to the inlet and outlet flow paths for the shell side of the IC1 feedwater heater on an increasing level condition in IC1 heater? Assume level continues to (3.5) increase to the trip point. -
END OF CATEGORY 6
2
SECTION 3 - INSTRUMENTS AND CONTROLS 3.1 State whether the following conditions would (increase, :
decrease, not change) the indicated level of the Yarway ;
',' Instrument, and explain briefly. i. (2.0)
- a. Equalizing valve leaks .
- b. Subcooling in variable leg ( d e - W b
- c. Steam carry under
- d. Rapid decrease in reactor vessel pressure from 1000 psi to 600 psi 3.2 Answer the following questions in regard to LPCI loop select logic:
- a. How does the logic determine how many recirc pumps are running? (1.0)
- b. How does the logic determine which is the undamaged recirc loop? (1.5)
- c. If the logic determines that neither loop is damaged, which loop will select for LPCI injection? (0.5) 3.3 What 3 conditions would cause the off gas system to isolate either totally or partially? Include what parts of the system will isolate. (3.0) 3.4 LPRM output signals are sent to various systems or locations.
What are four of these locations? (2.0) 3.5 What are two reasons for the interlocks, mechanical and key switch, associated with the reserve power supply to the RPS bus? (1.0) l 3.6 For each of the following, state whether a R0D BLOCK,
- HALF SCRAM, FULL SCRAM, or NO REACTOR PROTECTION SYSTEM ACTION is generated for that condition. (Note
- If two or more actions are generated, i.e., rod block and half-scram, state the most severe, i.e., half-scram.
l
- a. APRM B Downscale, Mode Switch in RUN (0.5)
- c. Flow Units I and B Upscale ~(>iEGREFlow),' Mode '
Switch in RUN
- (0.5) afu, A R A J ~ . g ) f,~ :2.wi otr
- d. Reactor water level 55", Reactor power 18%, Mode Switch in RUN (0.5) 3.7 The electromatic relief valves have three position control :
switches with the three positions being MANUAL, OFF, and ;
AUTO. How will the relief valves function in each of i (2.5)
{ these positions?
358 Answer the following questions in regards to Nuclear Boiler Instrumentation.
- a. What system uses the narrow range GE/MAC instrument input? (0.5)
- b. What are four (4) systems that use a reactor pressure input? (2.0)
- c. How does jet pump flow affect the wide range Yarway instruments reading actual verses indicated? (0.5) 3.9 What occurs when the " Emergency In" position is used in the Reactor Manual Control System? (1.5) 3.10 Match the recirculation flow control alarm with its setpoint. (2.5)
- a. Speed Signal Failure 1. Below 4 psid, 28 sec. after start
- b. Incomplete Sequence Trip 2. Feedwater flow below 20%
- c. Recirc Pump Low 3. Less than 1.0 ma output Differential Pressure from function generator
- d. Recirc Pump Locked 4. Below 4 psid Rotor Trip
- e. Recirc Loop Flow Limit 5. Below 4 psid, 30 sec.
after start 3.11 What are the four (4) conditions that can cause an APRM channel to be either physically inoperative or to be considered inoperative? (2.5)
END OF CATEGORY 6
2
SECTION 4 - PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
~
41 During shift change, the oncoming NSO shall perform ;
4 six items as a minimum per QAP 300-7, '! Shift Change i for Nuclear Station Operators." What are four (4) of these?
(4.0) e 4.2 On a loss of RBCCW cooling water the recirculation pumps are the most critical equipment to be monitored. What two recirc pump seal cooling water alanns would you expect (1,0) to receive on a loss of RBCCW? (Setpoints not required.)
4.3 When paralleling the main generator to the system, the synchroscope should be rotating A in the B direction and C voltage should be slightly 0 than the E voltage. (Fill in the blanks.) (2.5) 4.4 Define or explain the following:
High radiation area (1.0) a.
Effective half-life (1.0) b.
(1.0)
- c. REM 4.5 In a situation that the control room is evacuated and no scram was initiated prior to evacuating, where and how is each of the following performed?
Scram the reactor (1.0) a.
Trip the turbine (1.0) b.
Monitor reactor water level and pressure (1.0) c.
4.6 Assume an ATWS event has occurred:
What are the required immediate operator actions? (2.0) a.
- b. Under what conditions can the NSO inject SBLC without authorization from a supervisor? (2.5) e l
l
~^ ~ - -- -- . _ . _ . - . _-_ _ _ .
4.7 Match the appropriate conditions (i.e., power, press, or temp) (1 thru p with the following items a thru E
- a. Verify Rod Worth Minimizer Low Power 1. feed flow 40%
Setpoint (LPSP) window is lit. [
- b. Verify Rod Worth Minimizer Transition 2. feedhlow20%
[
g .;
window is lit.
- c. Remove the second feed pump from service. 3. 200 MWE Transfer the Reactor Mode Switch to 4. power between d.
startup. 5% and 10%
Open feedwater heater extraction drain 5. 30% power e.
valves. (2.5) 4.8 The reactor is operating at 90% power when condenser vacuum suddenly starts to decrease:
- a. WHAT are three (3) AUTOMATIC ACTIONS that occur on a LOSS of CONDENSER VACUUM 7 Include any applicable setpoints. (1.5)
E
- b. In accordance with 00A 3300-2, Loss of Condenser Vacuum, WHAT are six (6) Immediate Operator actions, OTHER THAN notification of the Shift Engineer and load dispatcher? (3.0)
END OF CATEGORY W
2
. EQUATION SHEET
~
. f = ma v = s/t Cycle efficiency = (Network
- out)/(Energy in) w = mg s = V ,t + 1/2 at 2 .
2 E = mc ~
KI=1/2mv a = (Vf - Vo )/t A = AN A= e PE = mgh V~ = V ,+ at w = e/t A = an2/t1/2 = 0.693/t1/2 1/2 eff = [(t1/2)(tb))
t NPSH = P in - Psat
[(t1/2) + (t b)3 m a p AV aE = 931 am I=Iec Q = mCpat I = I c
e~"*
Q = UAah Pwr = Wfah I=I g 10-x/TR TVL = 1.3/u P = Po l0 sur(t) HVL = -0.693/u t
P = Po e /T SUR = 26.06/T SCR = S/(1 - K,ff)
CR x = S/(1 - K,ffx)
SUR = 26p/t* + (s - p)T CRj (1 - Keff1) = CR2(1 kdf2)
T=(t*/p)+[(s-p)[Ap] M = 1/(1 - K,ff) = CR /CR j o T = 1/(p -8) M = (1 - Keffo)/(l - Keffj)
T = (s - p )/(Ap ) SDM = (1 - Keff)/Keff a2 (Keff-1)/K,ff = aKeff/K eff t* = 10 secon x = 0.1 seconds-p / + AT)]
= [(t*/(T K,f f)] + [s,f f (1 ldjj=1d22 2
P = (z,V)/(3 x 1010) Ij dj =Id p2 2
I = oN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2(feet).
NPhi=Statichead-ht-Psat
- '[
Water Parameters Miscellaneous Conversions I curie = 3.7 x 1010dps 1 gal. = 8.345 lbm.
1 gal. = 3.78 liters 1 kg = 2.21 lbm 1 ft3 = 7.48 gal. I hp = 2.54 x 103 Btu /hr Density = 62.4 lbgi/ft 3 1 mw = 3.41 x 106 Btu /hr Density = 1 gm/cv lin = 2.54 cm Heat of vaporization = 970 Btu /lbm *F = 9/5'C.+ 32 Heat of fusion = 144 Btu /lbm 'C = 5/9 (*F-32)
15' .
i os Table t Satufeled steam: Tempe7stufe Table 50ecalic ve4me Entna 07
$at Entrety $k Teme Aes Press Sal.
$at Sat Sal. Litme (see 90per fant
, Teme it Oer Lul.0 tes0 Water Lated Evat A fg vetor hg s, 8,, h I Fant 50 in. og h7
- p y, e gg 8 0000 2 1873 1 1s73 mi-I 00is022 uts 7 nos ? -acin 1975 5 ten S 0 0041 !!P62 2157 38 3
. 32 0
- 0 00059 0 016021 M619 3061 9 1 996 1974 4 1076 4 S W81 2 1651 2172 se MS 00N00 4 28 10732 10772 30 0 016020 MMO 2RM O 1972 1 19783 6 0122 2lWI 11883
. 30 0 10396 0 016019 NMI 2634 2 6 918
~,. 58 0 11249 Solu 2 1432 2 nes at M45 8 8 027 1977 0 IOne 0 0702 2 1325 1 8527 42 8 des 8 :2163 0 01:019 mas O 30 035 1598 1071 9 ass 00f60f9 2212 4 2272 4 1864 7 1000 7 8 0742 28217 f late at t 0 13143 0 016019 2112 8 2112 8 12 0s1 0 0292 2 till 2l3u 88 8 as t 01 192 1965 7 14 Os? 19676 INI6 0 0321 2l05 2 1327 48 8 .
0 086020 1965 7 1564 1523 48 8 48 8 0 15314 0 16514 8 016021 3430 0 1830 0 1651 0 0361 2 0001 11M2 30 8 b.k.
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- Critical temperstore
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ANSWERS - SECTION 1 1.1 Carry over is a condition which occurs when moisture :
droplets are swept into the steam exiting the core due i to a high water level resulting in a low exit quality. i.
These high velocity water droplets will cause excessive
- pipe erosion to occur. Damage by impingement to r
Carry un r is a condition which occurs o $ a->d er level is low and the water seal around the separator is broken and blowby to the annulus region occurs. This will result in steam bubbles being swept into the suction of the recirc pumps lowering NPSH and resulting in lower flow. (1.0)
Ref.: HTFF, page 25 1.2 a. Less
- b. More
- c. More Ref.: HTFF, pages 53, 54, and 58 1.3 a. As flow increases, critical power increases, at higher flow rate cooling improves and thus a greater power input is required to raise the coolant enthalpy to saturation conditions and change water to steam. (1.0)
- b. CP increases as inlet subcooling increases. Greater enthalpy rise is required to bring the coolant to saturation conditions thus higher bundle powers are required before boiling begins. (1.0)
- c. An increase in pressure will cause a decrease in CP.
The enthalpy of saturated steam decreases as pressure increases, e.g., enthalpy at 1000 psi = 1192.9; enthalpy at 1100 psi = 1189.1. (1.0)
The drop in enthalpy means a given pound of coolant must acquire less energy in traveling through a bundle to reach transition boiling. Thus, CP decreases.
Ref.: HTFF, page 29 and General Thermodynamics Theory 1.-4 An fuel temperature is increased (due to Control Rod pull) [
more voiding is created. Therefore, more back pressure, which would mean less flow. However, because of core orificing back pressure is less significant and flow through the bundle is almost the same.
Ref.: QC Reactor Vessel Internals, page 24 -
1.5 No. The production rate is directly proportional to power -
level, but removal rate is proportional to xenon concentration and it contains a power dependent term, thermal neutron fluxi
'_ Since flux is directly proportional to power level, the i.
burnout term becomes more significant. This results in an equilibrium xenon value which is higher than the original 7, value, but not twice as high.
Ref.: QC Reactor Theory, page 68 and figure 58 1.6 a. CR N_1-Keffo 1-Keffi CR M _1.96 1 .995 CR CR2 = 80 cps M _ .04503
- b. The time to reach an infinite period would be greater, due to the fact that there are more generations, each representing a period of time, required to reach equilibrium.
Ref. : QC Theory, pages 34 and 35 1.7 Yes. As the rod is moved, neutron flux will change. Heat flux from fission is directly proportional to neutron flux. The water temperature next to the fuel rod will change only slightly if at all. (The only change would be heating up the subcooled water to saturation conditions in the subcooled water to saturation conditions in the lower part of the fuel.) Heat transfer is directly proportional to delta T. As water temperature stays approximately constant, centerline temperature must change as flux changes. (Will accept either up or down.)
Ref. : HTFF 1.8 a. The MAPRAT of 1.02 is not conservative. With a MAPRAT greater than one it means that the RAFLHGR has been exceeded because:
MAPLHGR (actual)
MAPRAT =
MAPLHGR (LCO)
- b. 1. F
- 2. -T- F' _
- 3. T Ref.: HTFF, page 31 2
1 i
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1.9 a. 1. Moderator temperature coefficient alpha T = -1x10 4 per degree change in temperature.
~
- 2. Moderator void coefficient alpha V = -1x10 8 per %: change in voids.
{
[ 3. Fuel temperature coefficient alpha 0 = -1x10 s per degree
, ; change in fuel temperature.
I b. Alpha T increases core delta K/K Alpha V decreases core delta K/K
- Dominant effect - relief valve opening results in decreasing reactor
, pressure which increased voids and decreases moderator density i'
resulting in more neutrons leaking out to the core and reducing l power.
Ref. : QC Theory, pages 48 through 58 General Theory 1.10 (c) l Ref.: Standard Thermodynamics and Fluid Flow Principles
- 1.11 a. Pu-239 is builtup by a sequence of neutron absorption by U-238 and two subsequent b- decays to Pu-239.
- OR -
l U-283 + ON1 ---->U-239---(B-)--->Np-239---(B-)--->Pu-239 (.75)
- b. Pu-239 is a fissile material and buildup to rather i
significant levels so that it accounts for approximately
- 35% of the fissions at EOL (0.5). The delayed neutron fraction for Pu-239 is 0.0021. This coupled with its i significant fission fraction causes the effective delayed neutron fraction for the core to decrease substantially l
- over core life (0.5). As the fraction of delayed neutrons drops the effective generation time tends toward the prompt neutron lifetime so that the period resulting from a given i
reactivity insertion is shorter near EOL. ( 75) (1.75) ,
4 1
Ref.: Reactor Theory, Chapter 19 p 1.12 As steam flow increases, the DP losses #due to friction increase and so a larger DP is needed at higher flows. - (1,0) i Ref. : Figure 2 and Standard Thermodynamic Principles 1.13 1. Can't increase pump speed if feed flow <2x10 4 lbs/hr l
[ 2. Pumps are in basement of drywell i
! 3 1
j 3. Feedwater mixes in annulus region subcooling recirc water
- 4. Increased feedwater flow at higher power increases :
percent of subcooling i
,~ Any 2 at (0.5) ea.
~3 Ref. : HTFF, pages 57 and 58 i
J l
I i
t i
4 l
4
.- . = - _ _ . . - . . - . - . -. - _ _
ANSWERS - SECTION 2 2.1 1. A supply breaker trips on the running RFP, other than via the control switch, and; j (0.5)
[ 2. A standby RFP is selected, and; (0.5) i 3. The suction pressure is above the trip setpoint (of 120 psig), and; (0.5)
- 4. At least one ventilation fan is operating, and; (0.5) a
- 5. Oil pressure is satisfactory (greater than 10 psig),
and; (0.5)
- 6. Reactor water level is less than the trip point (48 inches) (0.5)
Ref.: Feed Condensate Lesson Plan, page 38 2.2 a. Fail closed (valve is prevented from completely closing) (0.5)
- b. Vent and drain valves will close (0.5)
I
- c. Tank level indication would be lost (0.5)
- d. Fail open (0.5)
- e. Fails open (0.5)
Ref. : P&ID M-15, M-34, M-40, M-41 and generic system design 4
2.3 a. Will inject (.25).
in
+potentialairbornea Turbinesealleakageresulting)h ivity in the HPCI room (.75 .
n & Cp.wj+
(1.0)
M ~ e,.a)er .
- b. Will not inject (.25). urbine stop and control valves will not open (.75). (1.0) 1- c. Will inject (.25). Pump overheating and seal damage may result during low or no flow conditions (.75). (1.0)
- d. Will not inject (.25). Maximum signal from the flow element will cause the controller to keep turbine speed 3
at mi u .
g- -
(1.0)
Ref.: HPCI Lesson Plan 2.4 a. Either: Internal damage to mechanism or rod will scram slowly on seal leakage. (1.0)
- b. High temperatures in the CRD; inability to move the rod; discharge of scram accumulators (2 required). (2.0)
Ref.: CRD Hydraulic Lesson Plan i
2'5 a. Breaker trip i j b. High building pressure e
- c. High radiation in the exhaust duct
- d. High radiation on the refuel floor
- e. High radiation in drywell
- f. Low reactor water level
- g. High drywell pressure Any 5 at .5 ea.
Ref.: Ventilation Lesson Plan, pages 16 and 17 2.6 a. False (0,5)
- b. False (0.5)
Ref.: P&ID M-30 + 74 2.7 a. LPCI injection valves to broken recirc loop will close (0.5)
- b. Intact recirc loop pump discharge valve will close (0.5)
- c. When reactor press 1325 psig LPCI infection valves to intact loop open (0.5)
Ref.: RHR Lesson Plan 2.8 a. Adds dilution steam to dilute the H2 gas concentration to less than 4% by volume (0.5)
- b. Provides the driving force for the system (0.5)
Ref.: Off Gas Lesson Plan 2.9 There are numerous indications that could be used and any reasonable answer will be given credit. The list _
below may not be all inclusive and reasonable indications ~
or alarms may be acceptable.
Any 8 at .25 ea.
l 2
~
Drywell sump flow Primary Containment pressure Primary Containment atmosphere temperature -
Primary Containment radioactivity Drywell level i T. Torus level i
, Reactor level
- - Reactor pressure
'I Steam flow / feed mismatch Steam line pressure decrease Torus water temperature Ref. : Containment and Instrumentation Lesson Plans i
2.10 1. Emergency seal oil pump (0.5)
- 2. Bearing oil header (0.5)
Ref. : Generator Auxiliary System Lesson Plans 2.11 As level starts to increase in the 1C1 heater the drain valve will open further and the (emergency spill) to the condenser will start to open and open fully as level increases. As level increases further the high-high trip will be reached.
At this point the extraction steam valve will close and the extraction bypass valve to condenser will open, the drain from the 101 heater will close and the drain (emergency spill) from the 1D1 heater to the condenser will open. (3.5)
Ref. : Feedwater Lesson Plan 1
i -
3
ANSWERS - SECTION 3 3.1 a. Increase, measured AP will decrease, indicating :
higher level ,i (0.5)
~
- b. Increase, measured AP will decrease because of j denser water (0.5)
- c. Decrease variable leg density will decrease, so APwillincrease (0.5)
I
- d. No change. No flashing (0.5) wille<<,.x T/J 44 af <4, or aux 4AJchamber will refill Ref.: Rx Instrumentativn Lesson Plavl L/ w//~~ 4 J 3.02 a. By monitoring the differential pressure across each recirc pump. (1.0)
- b. By comparing the pressure in the riser pipes on one recirc loop with the pressure in the riser pipes of the other loop. The undamaged loop will have a higher pressure than the damaged loop. (1.5)
! c. Loop B (0.5) i
. Ref. : RHR System Lesson Plan 3.3 a. An off gas high-high radiation signal and the expiration of the 15 min. timer causes:
- 1. The pressurized drain tank valve to the pressurized drain pump to close;
- 2. The chimney isolation valve to close;
- 3. The off gas drain line drain valve to close.
l b. A high radiation (7x normal) from the main steam line will cause:
- 1. Chimney isolation valve to close; j 2. Air ejector suction valve to close;
- 3. Off gas line drain valve to close; ,
l
- 4. Off gas sample vacuum pump suction valve to ,
close (mechanical vacuum pump trips).
f,w ., -
. c. A low steam 4+ow to the third stage SJAE signal causes the main condenser off gas suction valves to close.
4
---,,,nn.,--
, ,--,.n,. ~ - . -,-. . . . , . , _ . . . . _ , , , _ ___..,nn.,n.. ,.,,,,,,,,,.,--.,--m-,-- _m-_,w-.-.--- --,.-n-...-----
. s
- d. A high pressure /high temperature in the off gas holdup volume (explosion protection) signal causes the main condenser off gas suction valves to close. ,
i f ,, u., a I
- e. A low steam flow to the first and second stage SJAEs i
- i.
(100 psig) signal causes the main condenser off gas suction valves to close.
Credit will be given for e. even though it is always bypassed.
Any 3 at 1 point ea.
- Ref.
- Off-Gas Lesson Plan 3.4 a. RBMs
- c. The process computer
, d. The four rod display
- e. 90 x -37 back panel meters Any 4 at .5 ea.
Ref. : LPRM Lesson Plan 3.5 a. Prevents the MG set and the reserve power source from simultaneously supplying a RPS bus.
- b. Prevents selecting the reserve power supply to more than oneRPS bus.
- c. Prevent overloading the reserve instrument transformer Any 2 at .5 ea.
4 Ref.: RPS Lesson Plan 3.6 a. rod block (0.5)
- b. half-scram (0,5)
- c. rod block (0.5)
- d. no reactor protection system action 1 (0.5) l Ref.: RPS and APRM Lesson Plans 3.7 In the MANUAL position, the solenoid assembly is always energized to keep the valve open. (0.5) i 2
In the OFF position, the solenoid only energizes from an auto blowdown signal; it does not energize from a relief signal sent from the controller. -
(1.0)
In the AUTO position, the solenoid energizes from both ;
- an auto blowdown signal and a relief signal sent from i
{ the controller. (1.0)
~3 Ref.: ADS Lesson Plan 3.8 a. Feedwater level control (0.5)
- b. Safety / relief and electromatic relief valves ATWS RPS LPCI, core spray, HPCI, RCIC , C /z D (2.0) rwtc, 1"...... cc v r <.
Any 4 at .5 ea.
- c. Jet pump flow will cause level to indicate higher than actual (0.5)
Ref. : Reactor Vessel Instrumentation Lesson Plan 3.9 a. Bypas 1 interloc ' sert the rod except the rod worth m sert block and any select blo Oc /< 7 ( (tb4)
- b. Bypassing the timer to directly energize the directional control valves (0.75)
- c. No settle function, water forced past the seals (04)
Ref.: RMCS Lesson Plan 3.10 a. Speed Signal Failure 3. Less than 1.0 ma output (0.5) from function generator
- b. Incomplete Sequence 1. Below 4 psid, 28 sec.
Trip after start (0.5)
- c. Recirc Pump Low 4. Below 4 psid (0,5)
Differential Pressure
- d. Recirc Pump Locked 5. Below 4 psid, 30 sec.
Rotor Trip after start (0.5)
- e. Recirc Loop Flow 2. Feedwater Flow below Limit 20% or discharge valve not full open (0.5) 3
Ref.: Recirc Flow Control Lesson Plan 3.11 1. Module unplugged -
(0.5)
. , 2. Less than 50% of the assigned LPRMs in " operate." i:. (0.5)
. 3. Mode switch not in " operate." (0.5) i 4. Less than 2 inputs per LPRM level. (This is i procedural.) (1.0)
Ref. : APRM Lesson Plan i
t 4
. 4
ANSWERS - SECTION 4 I
~
4.1 a. Obtain the current operational status of the unit, -
including safety-related system status, equipment -
- line-up, and inoperable equipment. 3,
, O b. Obtain any information on general plant operations
'I that will be useful to future shift.
- c. Read and initial the log from the previous four shifts.
- d. Assure himself that he is aware of any abnormal operating conditions prior to accepting the position.
- e. Scan the panel's alarm boards and obtain information as to the reason why any alarms are up.
- f. Obtain information as to what major equipment has recently been taken out-of-service, and the reason why.
Any 4 at 1 point ea.
Ref.: QAP 300-7, Revision 5 4.2 Recirc pumps cooling water high temp (0.5)
Recirc pump seal cooling water low flow (0.5)
Ref.: Q0A 3700-1 4.3 a. Slow
- b. Clockwise (or fest)
- c. Incoming
- d. Higher
- e. Running i Ref.: QGP 1-3, Revision 20, page 11 4.4 a. An area where you could receive > 100 mrem in one hour (1.0) j b. A weighted half-life of a radioactive material which j - takesintoaccountthedecaycharacteristics(physicalT i half-life) of the material and retention of the material
- within the body (biological half-life). (1.0)
St'da ad! *~pN
~ -- - - - - - - _ - - - - - -- -- -
- c. Measure of dose of any ionizing radiation to body tissue in terms of its estimated biological effect ra (1.0) relative y:iB to we,Ja dose mm of 1 = roengten ca o x caof x dy.syy~ q Ref. : 10 CFR 20 .
, A sc. . . ~ (
~
. 4.-5 a. A itehtien is initiated in the auxiliary electrical
~
room yb op"erdng circuit breakers in the RPS distribution
-I= panel s! **A r *WI "'
t~ ;< l (1.0) fr a.t qcy.jf a.,/ j 7,
- b. This is done locally at the front standard with the manual trip. (1.0)
- c. Water level and pressure are monitored on the local indicators on panels 220X-5 and 6. (1.0) 4 Ref.: QOA 010.5, Revision 4 4.6 a. 1. Trip both recirc MG sets if they have not auto-tripped (0.5)
- 2. Scra'm the reactor manually by depressing both manual trip actuators. (0.5)
- 3. If a scram does not occur, initiate control rod scram by other means. (0.5)
- 4. If a scram cannot be made to occur, initiate SBLC. (0.5)
- b. If the control rod system is unable to maintain the reactor in a subcritical condition and either reactor water level cannot be maintained, or suppression pool water temperature cannot be maintained below the high temperature scram limit (110*F). (2.5) 4.7. a. -2
- b. -5
- c. -1
- d. -4
- e. -3 Ref.: QGP 2-1, Revision 25 4[8 a. 1. Reactor scram (21 in Hg vacuum) (0.5)
- 2. Main turbine trip (20 in Hg vacuum) (0.5)
- 3. Turbine bypass valve closure (7 in Hg vacuum) (0.5) 2
- b. 1. Reduce reactor recirculation flow as necessary to help maintain vacuum. Insert control rods as necessary to reduce power. .
- 2. Check and fill condenser loop seals. ,
- 3. If loss of vacuum is due to failure of the SJAE, O send an operator to the SJAE to correct problem.
~1
- 4. Verify that the condenser vacuum breaker is closed and has a water seal.
- 5. Verify the reactor scrams at 21 in Hg if mode switch is in RUN.
- 6. Verify the main turbine trips at 20 in Hg.
- 7. Verify that all turbine bypass valves close when vacuum is less than 7 in Hg.
- 8. Verify that turbine steam seal pressure is normal.
Any 6 at .5 ea.
Ref. : QOA 3300-2 3
o .
. . Maclec U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: Quad Cities REACTOR TYPE: BWR DATE ADMINISTERED: August 27, 1985 EXAMINER: T. Lang APPLICANT:
INSTRUCTIONS TO APPLICANT:
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%.
% of Category % of Applicant's Category Value Total Score Value Category 25 25 5. Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics 24.5 24.5 6. Plant Systems Design, Control, and Instrumentation -
- 25 25 7. Procedures - Normal, Abnormal, Emergency, and Radiological Control 25.E 25.5 8. Administrative Procedures, Conditions, and Limitations 100 100 TOTALS Final Grade %
All work done on this exam is my own, I have neither given nor received aid.
Applicant's Signature
e THEORY OF NUCLEAR POWER PLANT OPERATIONS, FLUIDS, AND THERMODYNAMICS 5.01 For the following conditions, state if the FUEL TEMPERATURE COEFFICIENT becomes more or less negative. _,
- a. Increase in fuel temperature (0.5)
- b. Increase in' moderator temperature (0.5)
- c. Increase in void fraction (0.5) 5.02 Define the following terms,
- a. Latent heat of vaporization (0.5) s._, b. Critical point (0.5)
- c. Saturated liquid (0.5) 4-5.03 Explain the difference between carry over and carry under. Include in your answer and negative effects on plant equipment. (2.0) 5.04 Explain the effects of increasing the following core parameters on steady state critical power,
- a. ~ Core flow (1.0)
- b. Inlet subcooling (1.0)
- c. - Reactor pressure ,
(1.0) 5.05 If equilibrium xenon is obtained (the reactor has been operated at constant power for many hours), and the reactor power is doubled, will the new equilibrium xenon concentration be twice as great? Explain your answer. (2.0) l 5.06 You increase core power by pulling control rods around the center fuel bundle. Assuming that l recirculation speed is kept constant would the flow through the center bundle increase, decrease, or stay the same? Explain your answer. (2.0) 5.07 Explain how and why Rod Worth changes for the following conditions.
- a. Rod Worth of a center rod compared to a peripheral rod. (1.0)
QUESTION CONTINUED NEXT PAGE 1
I
- b. Rod Worth when plant conditions change from
~
cold to hot at 1% power. (1.0)
- c. Rod Worth when plant conditions change from hot at 1% power to hot at 100% power. (1.0) l 5.08 Prior to startup (all rods full in), the SRM count rate is 10 CPS and k effective is 0.96
- a. if control rods are pulled to give a AK of of +0.035, what count rate on the SRMs should be expected when the period becomes infinite? (1.0)
- b. if additional control rods are pulled to give a AK of 0.003, would the time required to reach an infinite period be greater or less than the time in part "a"? Give the reason for your answer. (1.0) 5.09 Does the centerline temperature of the fuel pin located closest to a control rod change when the control rod is moved when operating near full power? Explain your answer using general concepts of heat transfer. (2.0) l 5.10 Your reactor operator informs you that MAPRAT is 1.02.
- a. Is the MAPRAT, as stated, conservative? Explain your answer. (.75)
- b. In regards to MAPRAT which of the following statements are True and which are False?
- 1. MAPRAT maintained within limits ensures that transition boiling will not occur in 99% of the fuel bundles. (.25)
- 3. Maintaining MAPRAT limits ensures that peak l clad temperature will not reach 2200*F during LOCA. (.25) 5.11 a. List the three (3) reactivity coefficients in a BWR
> at 100% power and give approximate values for each. (1.5)
- b. What effect (Increase, Decrease, or No Effect) do each of the three coefficients have on total core reactivity following a safety / relief value failing open? Briefly explain why the dominant coefficient effects reactivity in the manner you indicate. (2.0) 5.12 What effects does an increase in feedwater flow have on recirculation pump NPSH? Explain why. (1.0)
END OF CATEGORY 2
PLANT SYSTEMS DESIGN, CONTROL AND INSTRUMENTATION 6.01 What six (6) conditions must be met for the Standby Reactor Feed pump to auto start? (3.0) 6.02 If a complete loss of Instrument Air were to occur with the plant operating at full power and with no operator action, what would be the effect on the following components or systems? (NOTE: Limit your answer to effects caused directly by instrument air only.)
- a. CRD hydraulic flow control valve (0.5)
- b. CRD hydraulic instrument volume values (0.5)
- c. Standby Liquid Control System (0.5)
- d. Main Feedwater Pump minimum flow value (0.5)
- e. Reactor Building to torus vacuum breaker (air operated) (0.5) 6.03 State what problem would be associated with each of the following conditions:
- b. Failure of both CRD hydraulic pumps (two problems required). (2.0) 6.04 After LPCI loop selection logic has determined the intact recirc loop following a pipe break, what automatic actuations will occur to allow the LPCI system to perform its function? (1.5) 6.05 What are eight (8) different indications available to you in the control room that could be used to determine that a reactor coolant leak was occurring in the drywell. (2.0) 6.06 What will happen to the inlet and outlet flow paths for the shell side of the 1C1 feedwater heater on an increasing level condition in 1C1 heate.r? Assume level continues to increase to the trip point. (3.5) 6.07 State whether the following conditions would (Increase, Decrease, Not Change) the indicated level of the Yarway Instrument, and explain briefly,
- a. Equalizing value leaks (0.5)
- b. Subcooling in the variable leg (0.5)
QUESTION CONTINUED NEXT PAGE 3
- c. Steam carry under (0.5)
- d. Rapid decrease in reactor vessel pressure from 1000 psi to 600 psi (0.5) 6.08 Answer the following questions in regard to LPCI loop select logic:
- a. How does the logic determine how many recirc pumps are running? (1.0)
- b. How does the logic determine which is the undamaged recirc loop? (1.5)
- c. If the logic determines that neither loop is damaged, which loop will be selected for LPCI injection? (0.5) 6.09 According to the attached Figure 1 there are two interlocks (mechanical and key switch) associated with the feeds for the RPS buses. What is the reason or purpose of having each interlock? Be specific.
Treat each interlock individually. (2.0) 6.10 For each of the following, state whether a Rod Block, Half-Scram, or No Reactor Protection System Action is generated for that condition. NOTE: If two or more actions are generated, i.e., rod block and half-scram, state the most severe, i.e., half-scram.
- a. APRM B downscale, Mode Switch in Run (0.5)
- c. Flow units A and B upscale (>108% flow) Mode Switch in Run (0.5)
- d. Reactor water level 55", Reactor power 18% Mode Switch in Run. (0.5)
END OF CATEGORY i
4 l
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 7.01 What are three immediate operator actions required on a
" Failure of a Relief Valve to Close or to Reset Properly?" (1.5) 7.02 According to Q0A 202-1 " Jet Pump Failure" there are two indications or symptoms where simultaneous occurrence would indicate a failure of a jet pump. What are these two indications? Be specific. Include setpoints if needed. (2.0) 7.03 If the Standby Liquid Control System is initiated, what are five (5) indications or parameters that should be checked or verified to insure the system is operating properly? (2.5) 7.04 If plant conditions are such that the control room must be evacuated but an Immediate Evacuation is not necessary, what are three steps which should be performed prior to evacuating if possible? (1.5) 7.05 If a Main Steam line High Radiation condition were to occur, what are six automatic actions which would occur? Assume a High Radiation on both channels. (3.0) 7.06 If the following alarm were to annunciate, what event or condition would have caused its actuation?
xxxxxxxxxxxxXxxxxxxxxxxxxxxxx x x x TIP ISOLATION x x 0FF-LIMITS- x j x 00 % l x l
xxxxxxxxxxxxxxxxxxxxxxxxxxxxx (1.0) l 7.07 What are five automatic actions which will occur on a High-High Radiation in the main steamline of 7x normal? (2.5) 7.08 According to your ATWS procedure, what is the key to recognizing an ATWS event? (2.0) 7.09 LPCI loop selection will insure that the suction valves on both Recirculation pumps are kept open.
Explain why. (2.0) l 7.10 When is the use of Notch Override not permitted during a Normal Startup? (2.0) 5 l
1
7.11 During a unit startup and reactor vessel pressurization why is it necessary to maintain the pressure regulator setpoint 50 psig greater than reactor pressure? (2.0) 7.12 What are three things required to be verified prior to placing the Mode Switch in Run. (3.0)
END OF CATEGORY i
l r
t 6
ADPINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 8.01 a. Paragraph K of 10 CFR 50 states, "An operator or Senior Operator licensed pursuant to Part 55 shall be present at the controls at all times during the operation of the facility." What must an operator do to meet this requirement? (2.0)
- b. A control room supervisor may authorize a Unit Operator to leave his " stable and under control" reactor to help on a troubled unit if he ensures three (3) requirements are met. What are these three (3) requirements? (1.5) 8.02 In order to remove a safety-related system from service the outage must have an independent verification.
- a. Can anyone verify the system is removed from service? Explain your answer. (1.0)
- b. According to QAP 300-14 this independent verification should include verification of the status of two (2) types of plant equipment.
What are these two (2) types of equipment. (1.0) 8.03 a. Who must concur with Temporary Procedure changes which do not change the intent of the procedure? (1.0)
- b. Who must concur with Temporary Procedure changes which do change the intent of the procedure? (1.0) 8.04 Access to the Drywell is unlimited provided four conditions exist simultaneously. What are these four conditions? (2.0) 8.05 Other than on-shift personnel unlimited access is allowed for six (6) other persons. Who (by title)
- are five (5) of the six (6)? (2.5) l 8.06 Define the following terms per Technical Specifications r
- a. Hot Standby (1.0)
- b. Immediate (0.5)
- c. Instrument Checks (0.5) 8.07 According to Technical Specifications, what is the basis l for the following:
l l
- a. Condenser Low Vacuum Scram (1.0)
QUESTION CONTINUED NEXT PAGE 7
l l
- b. Main Steamline Isolation Value Closure isolation and scram (2.0)
- c. Turbine Control Valve Test Closure Scram (1.0) 8.08 a. Under what condition can a control rod drive, which cannot be moved by normal drive pressure be considered operable? (1.0)
- b. How is control rod drive coupling integrity verified? (2.0) 8.09 In regards to the RWM
- a. When is it required to be operable? (0.5)
- b. If it were to fail what action would be required? (2.0) 8.10 By Technical Specifications what are the DC electrical requirements which must be met prior to going critical? (2.0)
END OF EXAM l
l l
l
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l 8
i
4
- REACTOR PROTECTION SYE2M PO'at.R SL*PPI.Y r
(
16 ( 26 .) 19 (29 .)
' i r
,l 4, l)
I i
18-2 l (25-2) 19-2 ( 29-2 )
i i i t.
M M 1A-1 1B-1 EPA EPA LA-2 IB-2 EPA EPA i ,
b-- q r- -- -e (l MECE. RPS A I I RPS B MECH.
I INTL.
INTL . I j - t '
- - - - -- -- -c ei _ _ _m mEawCt______.)_-_ i LAB-2 A .
LAB-1 EPA 1-AB-3 REG l
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( - f
.,- W .1.) 12C WT 400 (V '
15-2 i) c25-2) l I FIGURE 1 1
EQUATION SX ET f = ma . y = s/t Cycle efficiency = (Network out)/(Energy in) w = ag s = Vot + 5 at2 E = mc2 KE = 1 mv2 a = (Vf - Vo )/t A = AN A = Ace-At PE = agh Vf=Vo + at w = e/t x = in2/tg = 0.693/tg W= 6 tgeff = [(tg) (t b)3
[(tg) + (t b)3 AE = 931 am y ,yo,-Ix Q' = m' Cpat 6=UAat I = Ic e -"X Pwr = W f s I = Io 10- N TVL = 1.3/p P = Po10 M HVL = -0.693/p P = Po et/T SUR = 26.06/T SCR = S/(1 - Keff)
CRx = S/(1 - Keffx)
SUR = 26p/r* + ( a- p )T CRt (1 - Keff1) = CR 2 (1 - keff2)
T = (t*/p) + [(a - p)/Ip] M = 1/(1.- Keff) = CR /CRo 1 T = f/(p 's) M = (1 - Keffo)/(1 - Keffj)
. T = (8 - p)/(Ap) , SDM = (1 - Keff)/Keff p = (Keff-1)/Keff = AKeff/Keff t* = 10-5 seconds I = 0.1 seconds-1 p = [(1/(T Keff)] + [s eff;/(1 + IT)]
'I d) t = 1 d22 2
P = (IeV)/(3 x 1010) 11d1 2=1d22 I = aN R/hr = (0.5 CE)/d2(meters)
R/hr = 6 CE/d2 (feet)
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm. I curie = 3.7 x 1010dps l
! I gal. = 3.78 liters 1 kg = 2.21 lbm
! I ft3 = 7.48 galt 1 hp = 2.54 x 103 Btu /hr l Density = 62.4 lbm/ft 1 mw = 3.41 x 106 Btu /hr l
Density = 1 gm/cm3 1 in = 2.54 cm Heat of vaporization = 970 Btu /1bm *F = 9/5'C + 32 Heat of fusion = 144 Btu /1bm 'C = 5/9 ('F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-1bf I ft H 2O = 0.433 lbf/in2 l
' l
. , Na s e r ANSWERS Q.C. EXAM 5.01 a. Less
- b. More
- c. More Ref.: HTFF, pages 53, 54, and 58 5.02 a. The point where further heat addition or removal to a substance will cause it to boil or condense without a change in tem rature
+ +4 ese.>~f*#
at,.a, a .cpytapt fressure.
a., . i 44.,Asu
.. a ,~ ., ~ n ./ A,- [* f /* C *~b 1
- b. The critical pressure and corresponding critical temperature at which the density of the vapor formed is exactly that of the liquid.
- c. 100% liquid at the saturation temperature and pressure.
Ref.: HTFF Definitions 5.03 Carry over is a condition which occurs where moisture droplets are swept into the steam exiting the core due to a high water level resulting in a low exit quality. These high velocity water droplets will cause excessive pipe erosion to occur. Damage by impingement to turbine and co amination of pure wa er b cond nser breakdown. 44. ....pd-deu ms J L ~< e S e-~ s rv ...n s aa CarryunderisaconditionwhicEoccursintheseparatora're2w..,m. hereas core water level is so low the water seal around the separator is broken and steam blowby to the annulus region occurs. This would result in steam bubbles being swept into the suction of the recirculation pump lowering the pumps NPSH and resulting in a loss of flow.
Ref.: HTFF, page 25 5.04 a. Increasing core flow will cause critical power to increase due to the increased heat removal capability.
- b. Increasing subcooling causes an increase in critical power. The inlet er.thalpy will be reduced and the heat which can be removed will increase.
- c. Increasing reactor pressure reduces the energy to be at critical e quality therefore power will be lower. #c. duwd .rd, a / Jebd/
- 34. - ,1 /sas a/ A y y e.tr w <. .
Ref.: HTFF, page 29, and Genera Thermodynamic Theory 5.05 No. The production rate is directly proportional to power level, but removal rate is proportional to xenon concentration and it contains a power dependent term, thermal neutron flux. Since flux is directly
proportional to power level, the burnout term becomes more significant.
This results in an equilibrium xenon value which is higher than the original value, but not twice as high.
Ref.: QC Reactor Theory, page 68, and figure 58 5.06 As fuel temperature is increased (due to Control Rod pull) more voiding is created. Therefore, more back pressure, which would mean less flow.
However, because of core orificing back pressure is less significant and flow through the bundle is almost the same.
Ref.: QC Reactor Vessel Internals, page 24 5.07 a. Control rods in the center of.the core are exposed to a higher thermal flux than those at the core periphery and, therefore, have a greater worth. Red ucrth would be increased :np here in th: cer-e whs e e radial th;r=1 neutren #!ux peak existed ever if it were an edge rud.
- b. As moderator temperature increases, its density decreases resulting in longer thermal diffusion lengths. This allows thermal neutrons to travel further and be absorbed by the rod; thus, rod worth increases with an increase in moderator temperature.
- c. As voids increase, again less moderations takes place. Again, thermal neutrons travel further; however, voids tend to depress thermal flux because of the very poor moderation. Thus, as voids increase, rod worth decreases.
Ref. : QC Reactor Theory, pages 59 and 60 ,
5.08 a. CR k_1-Keffo 1-Keff i CR M _1.96 1 .995
.04 CR2 = 80 cps aCR 10 -_ M l b. The time to reach an infinite period would be greater, due to the l
fact that there are more generations, each representing a period
! of time, required to reach equilibrium.
Ref.: QC Theory, pages 34 and 35 r 5.09 Yes. As the rod is moved, neutron flux will change. Heat flux from i fission is directly proportional to neutron flux. The water temperature next to the fuel rod will change only slightly if at all. (The only l change would be heating up the subcooled water to saturation conditions l
i 2
in the subcooled water to saturation conditions in the lower part of the fuel.) Heat transfer is directly proportional to delta T. As water temperature stays approximately constant, centerline temperature must change as flux changes. (Will accept either up or down.)
Ref.: HTFF 5.10 a. The MAPRAT of 1.02 is not conservative. With a MAPRAT greater than one it means that the HAPLHGR has been exceeded because:
MAPRAT =
MAPLHGR (actual)
MAPLHGR (LCO)
- b. 1. F
- 2. -T- F
- 3. T Ref.: HTFF, page 31 5.11 a. 1. Moderator temperature coefficient alpha T = -1x10 4 per degree change in temperature.
- 2. Moderator void coefficient alpha V = -1x10 3 per % change in voids.
- 3. Fuel temperature coefficient alpha D = -1x10 5 per degree change in fuel temperature.
- b. Alpha T increases core delta K/K Alpha V decreases core delta K/K Dominant effect - relief valve opening results in decreasing reactor pressure which increased voids and decreases moderator density resulting in more neutrons leaking out to the core and reducing power.
Ref.: QC Theory, pages 48 through 58 General Theory 5.12 NPSH will increase because of increased subcooling.
Ref.: QC HTFF, page 57 3
ANSWERS - Q.C. EXAM 6.01 1. A supply breaker trips on the running RFP other than via the control switch.
- 2. A standby RFP is selected.
- 3. The suction pressure is above the trip setpoint (120 psig) and
- 4. At least one ventilation fan is operating, and
- 5. Oil pressure is satisfactory (greater than 10 psig) and
- 6. Reactor water level is less than the trip point (48 inches)
Ref.: Feed Condensate Lesson Plan, page 38 6.02 a. Failed closed (Valve is prevented from completely closing)
- b. Vent and drain valves will close
- c. Tank level indication would be lost
- d. Fails open
- e. Fails open Ref.: P&ID M-15, M-34, M-40, M-41 and generic system design 6.03 a. Either: Internal damage to mechanism or rod will scram slowly on seal leakage.
- b. High temperature in the CRD; inability to move the rod; discharge of scram accumulators (2 required)
Ref.: CRD Hydraulic Lesson Plan 6.04 a. LPCI injection valve to broken recirc loop will close.
- b. Intact recirc loop pump discharge valve will close.
- c. When reactor pressure < 325 psig LPCI injection valves to intact loop open.
Ref.: RHR Lesson Plan 6.05 There are numerous indications that could be used and any reasonable answer will be accepted. The list below is just a sample.
- a. Drywell sump flow
- b. Primary Containment pressure
- c. Primary Containment atmosphere temp.
- d. Primary Containment radioactivity
- e. Drywell level
- f. Torus level
- g. Reactor level
- h. Reactor pressure
- i. Steam / Feed mismatch
- j. Steamline pressure decrease
- k. Torus water temperature 6.06 As level starts to increase in the 101 heater the drain valve will open further and the emergency spill to the condenser will start to open and open fully as level increases. As level increases further the high-high trip will be reached. At this point the extraction steam valve will close and the extraction bypass valve to the condenser will open, the drain from the 1D1 heater will close ead-the drain (amara-acy spill) f =
the 1n1 hastar te tha ennriangor will nnaa-I Ref.: Feedwater Lesson Plan 6.07 a. Increase, measured delta P will decrease indicating higher level
- b. Increase, measured delta P will decrease because of greater water i density ,
- c. Decrease, variable leg density will decrease, so delta P will increase
/ 4cuf b lo m <se
- d. wNow chan age,j-eno flashing+ woy~y a-or aux chy<nel will refill. - 4 s*
swarsa .
i l Ref.: Rx Instrumentation Lesson Plan l
l 6.08 a. By monitoring the differential pressure across each recirc pump
! b. By comparing the pressure in the riser pipes on one recirc loop with the pressure in the riser pipes of the other loop. The undamaged
, loop will have a higher pressure than the damaged loop.
l
- c. Loop B Ref. : RHR System Lesson Plan l
l 2
6.09 Mechanical Interlock prevents the MG set and the Reserve power source from simultaneously supplying an RPS bus, ney Switch Interlock prevents selecting the Reserve power supply to more than one RPS bus. Will also accept: prevents overloading the Reserve Instrument Transformer.
Ref. : RPS Lesson Plan and figure 1 6.10 a. Rod Block
- b. Half-Scram
- c. Rod Block
- d. No Reactor Protection System Action Ref. : RPS and APRM Lesson Plans I
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ANSWERS Q.C. EXAM 7.01 1. Attempt to properly close or seat the valve by opening and reclosing it using the appropriate key-lock switch.
- 2. If the valve does not properly close, scram the reactor.
- 3. If the suppression pool temperature exceeds 95 F, place the suppression pool cooling mode of RHRS into service i.Tenediately and closely monitor suppression pool temperature.
Ref. : Q0A 201-2 7.02 1. The recirculation pump flow differs by more than 10% from established speed-flow characteristics.
- 2. The indicated total core flow is more than 10% greater than the core flow valve derived from established power-core flow relationships.
Ref.: Q0A 202-1 7.03 Monitor the following to insure injection:
- 1. Amber pilot light of squib firing continuity circuit not lit.
- 2. Flow indicating pilot light lit.
- 3. Reactor Water Cleanup system isolation.
- 4. Decreasing level of SBLC storage tank.
- 5. Standby liquid squib valve circuit fail annunciator lit.
Will also accept: Power decreasing or any other reasonable answer.
Ref.: QOP 1100-2 7.04 If possible before control room evacuation:
l 1. Manually scram the reactor
- 2. Observe proper operation of the feedwater level co.ntrol system during the post-scram transient. It is desirable to leave the system in automatic on the low flow bypass valve if possible when control room is evacuated.
- 3. Notify the Chicago L.D. that the reactor has been scrammed and the control room evacuated.
Ref.: Q0A-010-5
7.05 Automatic actions if both channels High Radiation.
- 1. Reactor Scram
- 2. MSIVs close b
- 4 y s, J .~ L .
- N %u and Los s .o
- 3. Off gas isolation valves close c{ . m ,
- 4. Air ejector suction valves close
- 5. Mechanical vacuum pump trips, if running
- 6. MSL drains close
- 7. Primary sample valve close (0.5 each) only six (6) required for full credit Ref.: Q0A 1700-5 7.06 One or more TIP isolation valves open with a GRP II containment isolation signal present.
Ref.: Q0A-900-3-A alarm A-16 (79) 7.07 1. Reactor Scram
- 2. Group I isolation
- 3. Stack isolation valve isolates
- 4. Air ejector suction valve isolates
- 5. Mechanical vacuum pump trips Ref.: QGA-16 7.08 Reactor pressure and/or neutron flux indication increases abruptly, and may go off scale on recorder.
Ref. : QGA-17, page 1 7.09 Reactor recirc pump suction valves will remain in the open position to prevent inadvertent closing on the affected loop to assure reactor vessel blowdown.
Ref.: QGA-1 7.10 Use of Notch Override switch is not allowed between positions 04 and 12 on control rod arrays 3 and 4 and between position 00 and 24 from half control rod density until the reactor pressure is 920 psig with at least one bypass valve partially open.
2
Ref.: QGP 1-1 7.11 If the pressure regulator setpoint should become less than reactor pressure with condenser vacuum less than or equal to 7 inches, upon attaining 7 inches of vacuum, a reactor scram could occur from the sudden opening of the turbine bypass valves.
Ref.: QGP 1-1, page 6 7.12 1. Verify that all APRMs are indicating between 4% and 12%.
- 2. Verify all APRM downscale lights are not illuminated.
- 3. Verify that the main condenser back pressure is less than 7 inches Hg. (greater than 23 inches Hg. vacuum).
- 4. Verify that the channel A/B low vacuum alarm is cleared.
Any three for full credit.
Ref.: QGP 1-1 3
ANSWERS Q.C. EXAM 8.01 a. In order to comply with this paragraph, an operator will be considered to be at the reactor controls if he (she) is physically within the operating area in front of the unit panels.
- b. 1. A licensed operator has specifically been assigned the responsibility of monitoring the controls of the unit and responding to all unit alarms.
- 2. This same Licensed Operator remains within line of sight of the unit's front panels, and
- 3. The Licensed Operator, on a periodic basis, (approximately 5-10 minutes) reviews the status of that unit from within the area designated as being in close proximity to the main control panels of the unit.
Ref.: QAP 300-2, pages 11 and 12 8.02 a. e44 Mustbeverifidbha[%0wledgt:ble:nagementpersca.
=e
%o efase.
l 1..,.. eg e *b*
- b. First hand check of valve positions and electrical feeds.
Ref.: QAP 300-14, page 3 8.03 a. Temporary Procedure changes which do not change the intent require concurrence of one SR0 and one of the individuals identified in QAP 1100-T1, column 3.
- b. Temporary Procedure changes which do change the intent require concurrence of the Technical Staff ~" Supervisor and one of the individuals identified in QAP 1100-T1, column 2.
Ref.: QAP 1100-7, page 1 8.04 1. Reactor is subcritical
- 2. Mode switch in " shutdown" or " refuel"
- 3. Reactor vented
- 4. Tip system is out-of-service Ref.: QAP 1150-2, page 1 8.05 Any five for full credit
- 1. Station Superintendent
- 2. Assistant Superintendent
2
- 3. Operating Engineer
- 4. Technical Staff Supervisor
- 5. Perman3ntly assigned onsite NRC inspector
- 6. Station Quality Assurance personnel Ref.: QAP 1900-3, page 6 8.06 a. Hot Standby means operation with the reactor critical, system pressure less than 1060 psig, the main steam isolation valves closed, and thermal power not exceeding 15%.
- b. Immediate means that the required action will be initiated as soon as practicable, considering the safe operation of the unit
, and the importance of the required action.
- c. An instrument check is qualitative determination of acceptable operability by observation of instrument behavior during operation.
This determination shall include, where possible, comparison of
, the instrument with other independent instruments measuring the same variable.
, Ref.: TS Definitions 4
8.07 a. Loss of condenser vacuum occurs when the condenser can no longer handle the heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in surface heat flux. To prevent the cladding safety limit from being exceeded if this occurs a reactor scram occurs on turbine stop valve closure
- in the Run mode.
- b. The low pressure isolation of the main stem at 825 psig was provided
( to give protection against rapid reactor depressurization and the rapid cooldown of the vessel. Advantage was taken of the scram feature in the Run mode which occurs when the main steamline isolation valves are closed to provide for reactor shutdown so that high power operations at low reactor pressure does not occur, thus, providing protection for the fuel cladding inte! ity safety limit.
- c. The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass.
l l 8.08 a. Control rod drives which are fully inserted and electrically
- disarmed shall not be considered to be inoperable.
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._. - . , __ _ - . , _ . _ _ . . . , . _ _ _ _ , _ . . - _ . _ _ . . . - _ _ - _ _ - ~ _ . - --
- .,s,
- b. The coupling integrity shall be verified for each withdrawn control rods as follows:
- 1. When the rod is withdrawn the first time subsequent to each ,
refueling outage or after maintenance, observe discernible l response of the nuclear instrumentation; however, for initial 1 rods when response is not discernible, subsequent exercising of these rods after the reactor is critical shall be performed to verify instrumentation response.
- 2. When the rod is fully withdrawn the first time subsequent to each refueling outage or after maintenance, observe that the drive does not go to the overtravel position. 1 l
Ref.: TS page 3.3/4/3-la and 3.3/4.3-2 l 8.09 a. The RWM must be operable below 20% power.
- b. A second operator or qualified technical person may be used as a substitute for an inoperable RWM which fails after withdrawal of at least 12 control rods to the fully withdrawn positions.
Ref.: TS, page 3.3/4.3-3 8.10 a. Unit 24/48 volt batteries
- b. two station 125 volt batteries
- c. two station 250 volt batteries
- d. Battery chargers for each required battery Ref.: TS, page 3.9/4.9-2 l
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