ML20137H246
| ML20137H246 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 08/23/1985 |
| From: | Lang T, Long T, Mcmillen J, Plettner E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20137H232 | List: |
| References | |
| 50-237-OL-85-02, 50-237-OL-85-2, NUDOCS 8508280233 | |
| Download: ML20137H246 (60) | |
Text
r U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-237/0L-85-02 Docket Nos. 50-237; 50-249 Licenses No. DPR-19; DPR-25 Licensee: Connonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Dresden Nuclear Power Station 2 and 3 Examination Administered At: Morris, IL Examination Conducted: June 10-14, 1985 Examiners: E. Plettner fy/23/#
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Examination Summary Examination administered on June 10-14, 1985 (Report No. 50-237/0L-85-02)
Results: Eleven candidates took the examination and seven passed.
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r- 1 REPORT DETAILS
- 1. Examiners E. Plettner, Region III, Chief Examiner T. Lang, Region III J. McMillen, Region III
- 2. Examination Review Meeting After completion of the written examination an exam review was conducted with Mr. S. Mattson, Mr. S. Stiles and Mr. 8. Zank. Facility comments were for the most part editorial and have been incorporated in the master examination and answer key for both R0 and SR0 exams.
- 3. Exit Neeting At the conclusion of the examinations, an exit meeting was held with the plant staff. They were informed of all candidates who clearly passed the oral and simulator exams.
r TiASTER COP 3 U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY Dresden 2/3 REACTOR TYPE: BWR GE 3 DATE ADMINISTERED: June 10, 198S EXAMINER: e. Plettner 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 25 25 1. Principles of Nuclear Power Plant Operations, Thermo-dynamics, 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 100 100 TOTALS Final Grade %
All work done on this exam is my own, I have neither given or received aid.
Applicant's Signature
Section 1 - Questions - Principles of Nuclear Power Plant Operations, Thermodynamics, Heat Transfer and Fluid Flow.
1.01 a. Excess reactivity is initially loaded into the reactor (1.0) to compensate for Keff decreasing from BOL to EOL. Give two (2) reasons why Keff will decrease at various times from BOL to EOL.
- b. For a period of time during core life, Keff will actually (1.0) increase (become more reactive). Give two (2) reasons for the core becoming more reactive.
1.02 EXPLAIN or DEFINE the following terms:
- a. Prompt Critical (0.5)
- b. Reactor Period (0.5)
- c. Suberitical Multiplication (0.5) 1.03 Which of the following radioactive isotopes found in the (1.0) reactor coolant would NOT indicate a leak through the fuel cladding?
- a. I-131
- b. Xe 133
- c. Co-60
- d. Kr-85 1.04 Prior to startup (all rods in) the SRM countrate is 20 CPS (1.0) and K effective in 0.96. If the control rods are pulled to give a delta K of +0.035, what count rate on the SRMs could be expected when the period becomes infinite?
- a. 40
- b. 160
- c. 80
- d. 120 1.05 Referring to the attached curve (Fig. 1), which of the (1.0) following regions on the curve is associated with the heat transfer mechanism known as " transition boiling"?
- a. A+B
- b. D
- c. E
- d. C+0
- e. B+C
r i I
1.06 A moderator is necessary to slow neutrons down to thermal (1.0) energies. Which of the following is the best reason for operating with thermal instead of fast neutrons,
- a. Increased neutron efficiency since thermal neutrons are less likely to leak out of the core than fast neutrons.
- b. Reactors operating primarily on fast neutrons are inherently unstable and have a higher risk of going prompt critical,
- c. The fission cross section of the fuel is much higher for thermal neutrons than for fast neutrons.
- d. Doppler and moderator temperature coefficients become positive as neutron energy increases.
1.07 Which of the following statements best describes the (1.0) condition known as " condensate depression"?
- a. Can lead to condensate pump cavitation if condensate depression is too great.
- b. Decreases as hotwell level rises.
- c. Reduces Rankine cycle efficiency.
- d. Increases as condensate temperature ince ases.
1.08 Which of the following statements correctly completes (1.0) the following sentence? Departure from nucleate boiling is the point where:
- a. Void fraction equals one.
- b. The heat transfer mechanism changes from nucleate boiling to single phase convection.
- c. Radiative heat transfer becomes insignificant.
- d. The heat transfer rate increases substantially when nucleate boiling reaches its maximum.
1.09 Which of the following is NOT correct concerning decay (1.0) heat?
- a. Is the heat produce'. Jy the energy released from the radioactive decay of fission products.
- b. Can be determined by the reading on the SRMs when reactor is shutdown.
- c. Is approximately 6% of the total energy released from fission.
- d. Is still a significant contributor to the energy in the reactor core for approximately two hours after the reactor has been shutdown.
2
n 1.10 Figure 1.21 is a representation of how the resonance (1.0) peaks of U-238 " flatten out" or Doppler broaden as fuel temperature increases. Which of the following are the correct labels for the X and Y axes?
- a. X is neutron flux; Y is interaction rate
- b. X is interaction rate; Y is neutron density
- c. X is atom density of U-238; Y is neutron flux
- d. X is neutron energy; Y is microscopic capture cross section 1.11 The ratio of Pu-239 and Pu-240 atoms to U-235 atoms (1.0) changes over core life. Which of the pairs of parameters listed below are most affected by this change?
- a. Doppler coefficient and beta
- b. Moderator .emperature coefficient and doppler coefficient
- c. Beta and moderator coefficient
- d. Moderator temperature coefficient and neutron generation time 1.12 What is " pump runout" and why is it an undesirable (1.0) condition?
1.13 Give three parameters which effect the NPSH of a (1.5) recirculation pump.
1.14 A reactor is exactly critical. Control rods are then withdrawn to insert .0005 Delta-k/k (assume Beta 0.007, Lambda =.1/sec)
Show all work.
- a. What is the resulting stable period? (0.5)
- b. How long will it take for power to increase by a (0.5) factor of 10?
- c. What would the period be for a further addition (0.5) of .0005 Delta-k/k?
1.15 a. Where do delayed neutrons come from? (0.5)
- c. Why does this fraction in part (b) change from BOC (1.0) ,
to E0C? i l
l 3 l 1
1.16 The reactor is started up after a refueling outage. Rods are pulled to the 100% line and power is then increased to 100%
with recirculation flow. After approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, reactor power has decreased to about 98% with no operator action.
- a. What is the primary cause for this reduction in power. (0.5)
- b. Briefly explain why control rod withdrawals are not (1.0) recommended at a high power levels.
1.17 Concerning the core thermal limits:
- a. For each condition (1-4) given below, INDICATE whether it will cause an INCREASE, a DECREASE, or have NO EFFECT on CRITICAL POWF.R Ratio.
- 1. Local peaking factor (LPF) INCREASES (0.5)
- 2. DECREASE in inlet subcooling (0,5)
- 3. INCREASE in reactor pressure (0,5)
- 4. Axial power peak shifts from BOTTOM to TOP of (0.5) channel
- b. With regard to MAPRAT:
- 2. IS a MAPRAT of 1.05 acceptable? (.25)
- 3. WHAT physical consequence could occur if the (.75)
MAPRAT limit is exceeded?
1.18 What are three (3) sources of neutrons other than (1.5) installed sources when the reactor is shutdown?
END OF SECTION 1 l
4 s
l i
l Section 2 - Questions - Plant Design Including Safety and Emergency Systems 2.01 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: (NOTE:
Limit your answer to effects caused in relation to instrument air only).
- a. CRD Hydraulic flow control valve (0.5)
- b. CR0 Hydraulic scram valves (0.5)
- c. CRD Hydraulic instrument volume (0.5)
- d. Main Feed pump minimum flow valve (0.5)
- e. Main Feed regulating valves (0.5) 2.02 What four (4) conditions must be met for the standby (3.0)
Reactor Feed Pump to auto start? (Include setpoints where applicable) 2.03 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 VACUUM PUMP fails to operate. (1.0)
- b. The turbine AUXILIARY LUBE OIL PUMP fails to operate. (1.0)
- c. The MINIMUM FLOW VALVE fails to auto open (stays shut) (1.0) when system conditions require it to be open.
- d. The HPCI pump DISCHARGE FLOW ELEMENT output signal to the (1.0)
HPCI flow controller is failed at its maximum output.
2.04 State what problem would be associated with each of the following conditions:
- b. Failure of both CRD Hydraulic pumps (two problems (2.0) required other than part c below).
- c. Isolating CRD Hydraulic water to the recirculation pumps. (1.0) 2.05 What are the four chain of events postulated to occur (2.0) in a control rod drop accident?
5
, e _ - , - - - - - -
a--
2.06 There are five (5) conditions which will automatically (2.0) trip the Reactor Building Ventilation supply and exhaust fans. WHAT are four (4) of these conditions?
2.07 Indicate if the following statements regarding the Fire Protection System are TRUE or FALSE.
- a. The ten wall-mounted hose-reel CAR 00X assemblies located (0.5) throughout the plant each have a separate CO2 storage tank.
- b. Although the Service Water Tie Line Valve (M02-3906) is (0.5) locked in the closed position, pressure in the Fire Protection Water System is normally maintained by the Service Water System.
- c. In the automatic mode of operation of the Halon (0.5)
Suppression System, an activation signal turns on the evacuation lights, sounds a siren and immediately commences a 3 minute injection of Halon.
- d. Once the Halon system begins to inject in the Aux. (0.5)
Electric Room, it is possible to secure the injection with an abort switch located adjacent to the door to the Aux. Electric Room.
2.08 a. List all the sources in order of preference, of makeup (1.5) water to the shell side of the isolation condenser.
- b. List three (3) of the four (4) conditions which will (1.5) cause valves (1301-17, -20) to auto close.
2.09 State whether the following conditions would (increase, (2.5) decrease, not change) the indicated level of the Yarway Instrument. (0.5 points for each answer.)
- a. Equalizing valve leaks
- b. Subcooling in variable leg
- c. Steam carry under
- d. Rapid decrease in reactor vessel pressure
- e. Flashing of condensate pot END OF SECTION 2 6
i
, I i
Section 3 - Questions - Instruments And Control 3.01 Answer the following questions in regard to LPCI loop select logic:
- a. How does the logic determine how many recirc pumps are (1.5) running? (Include Set Point)
- b. How does the logic determine which is the undamaged (1.5) recirc loop?
- c. If the logic determines that neither loop is damaged, (0.5) which loop will select for LPCI injection?
fr g. e .-
3.02 Where does the signal orginate and under what conditions (2.0) will the OFF-GAS system isolate.
3.03 Describe how the EHC Pressure Control and Logic System (4.0) would respond if while operating at 100% power, the maximum combined flow fails to zero. Take your discussion to a final steady state condition. (i.e., Reactor scram or maximum combined flow at zero and state the final plant steady state condition. Assume no other operator action and all systems are in normal full power lineups.)
3.04 What happens when an " Edge Rod Selected" signal is (0.5) generated in the Rod Block Monitor System?
3.05 LPRM output signals are sent to various systems. (2.0)
List four (4) of the five (5) systems.
3.06 What are two reasons for the mechanical interlocks (1.0) associated with the reserve power supply to the RPS bus?
3.07 For each of the following, state whether a R00 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.
- a. APRM B Downscale, Mode Switch in RUN (0.5)
..,..t.-
- c. Flow U ets A and B Upscale (>108% flow), Mode Switch (0.5) in RUN
- d. Reactor water level 55", Reactor power 18%, Mode (0.5)
Switch in RUN 3.08 How is reduced sensitivity accomplished in the IRM (1.5) when compared to the SRM? ( 3 + a we. s e . 4: A 7
3.09 What are the automatic modes of operation for the (1.5) safety / relief valves? p ry+ hk 3.10 Answer the following questions in regards Nuclear Boiler Instrumentation.
- a. What system uses the narrow ranges GE/MAC instrument (0.5) input?
- b. What four (4) out of five (5) systems use rea tor (2.0) pressure input?
- c. How does jet pump flow affect the wide range Yarway (0.5) instruments reading actual verses indicated?
3.11 What occurs when the " Emergency In" position is used (1.5) in the Reactor Manual Control System?
3.12 Match the recirculation flow control alarm with its (2.5) setpoint.
- 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 Prewsure 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 END OF SECTION 3 l
8
Section 4 - Questions - Procedures - Normal, Abnormal, Emergency and Radiological Control 4.01 What five (5) actions shall the on-coming NSO perform (2.5) during the shift turnover?
4.02 What are an operators actions on the loss of cooling (2.0) by RBCCW?
4.03 When paralleling electrical sources the synchroscope (2.5) should be rotating A in the B direction and C voltage should be slightly D than the E voltage. (Fill in the blanks).
4.04 Define or explain the following:
- a. High radiation area (1.0)
- b. Effective half-life (1.0)
- c. Curie (1.0)
- d. REM (1.0) 4.05 Wnat is the whole body dose limit in an emergency condition that:
- a. Involves protecting equipment (0,5)
- b. Involves life saving action (0.5) 4.06 What are your actions upon orders from the Shift (2.5)
Supervisor to Evacuate the Control Room? (Assume 100% power normal operation) 4.07 Assume an ATWS event has occurred:
- a. What are the required immediate operator actions? (2.0)
(Identify the differences between Units 2 and 3 actions)
- b. Under what conditions can the NSO inject SBLC without (1.0) authorization from a supervisor?
- c. After the Standby Liquid Control System is initiated, (2.0)
WHAT are the five (5) indications that the system is operating?
4.08 What are an operators action upon receiving a Reactor (2.5)
High Pressure?
9 l _ ._
List the following items in the order of occurrence during (3.0) I 4.09 a normal unit shutdown (per procedure DGP 2-1) and match
, the approximate conditions (i.e. power, press, or temp) at which each is performed,
- a. Verify Rod Worth Minimizer Low Power 1. 40% power
, Setpoint (LPSP) window is lit. ;
- b. Verify Rod Worth Minimizer Transition 2. feed flow 20%
window is lit.
- c. Transfer level control to the Low Flow 3. 200 MWE Control Valve.
- d. Remove the second feed pump from service. 4. power between 5% and 10%
- e. Transfer the Reactor Mode Switch to 5. 35% power startup.
- f. Remove feedwater heaters from service. 6. feed flow less than or equal 10%
END OF SECTION 4 i
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in2/t 1/2 = 0.693/t 1/2 v2 1/2 eff = [(tp )(t ))
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P = Foe /T ,
SUR = 26.06/T , SCR = 5/(1 - K,7f) x = 5/(1 - K ef f X)
CR ..
SUR = 252/ t= + (5 - p)T CR)(1 - Kg g) = CRp (1 - k df 2)
T = (L*/o) + [(E -p)/lo) M = 1/(1 - Kgf) = CR)/CR l T = 1/(p - 8) M = (1 - Kg g)/(1 - KdH)
T = (E - p)/(10) SDM = (1 - Kgf)/Kdf p = (Kgf-1)/K ,ff = ar'eff/K eff t= = 10' seconds
-I T = 0.1 seconds o = [( t=/(T Keff)) + E Mf/(I + T))
I)d) =1d P = (I:v)/(3 x 1010) - 1)d) 2 ,2jd 2 22 2 I = o ff R/hr = (0.5 CE)/d (meters)
Water Parameters Miscellaneous Conversions l 1 gal. = 8.345 lbm.. 1 curie = 3.7 x 1010 cp3 1 gal. = 3.78 liters 1 kg = 2.21 lbm I ft3 = 7.48 gal. 1 hp = 2.54 x 10 3 Btu /nr Density = 62.4 lbm/f t 3 1 ms = 3.41 x 106 Btu /hr Density = 1 gm/cm 3 lin = 2.54 cm
'! Heat of vaporization = 970 Btu /lom 'F = 9/5'C + 32 Heat of fus. ion = 144 Btu /lbm *C = 5/9 (*F-32) 1 Atm = 14.7 psi =, 29.9 in. Hg.
Tcbla 2. Prgerties of Saturated Steam cud Satumttd Wat2r (Pressure)
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' e.7647 8.6091 1 373e *60.1 1806.e 555.63 0.02206 563 3 629 1 1387.9 e.7634 e.6115 1 37e9 558.7 3106 3 1320.0 0136641 0.388e7 561.9 425 9 18e7.8 1120.0 528.52 0.02203 e'.37041 0.29244 s.7620 e.6140 1 3760 557 3 3306 6 560.5 427.0 11ts.2 s.7e06 e.6365 1.3771 555.9 1806.9 1150.0 517.40 8 82199 0*.37%49 e.39648 559 0 629 6 1188.7 e.7592 0.6190 1 3783 554.5 1107 2 1100.3 556.2e e.02195 0137e63 0.40058 557.5 631 5 1889.1 8.7578 1090.0
- 555.14 S.02192 0*.35285 e.40476 e.6216 1.379. 353.1 1107 5 556 1 433 4 1189.5 e.7564 e.6241 1 2805 1020.0 **
- 5**.00 e.028te 0*.20714 0.40902 554.6 635 3 11eg.9 s.7550 551.7 Ills 7.0 550.2 1808 1 1070.0 552.06 0 02184 e'.39150 0 41335 553.1 e.6266 1.2317 637.1 11%C.3 e.7536 8.6292 1.222e Ste.e 1100.4 1060.0 551.70 8 021t1 C*.39594 0.41775 551.6 639 0 1150 7 e.7522 0.6318 1 28m0 !*7.4 1308*7 Sc50.0 520.53 0.02177 0*.40047 0.42224 550.1 640.9 1151.0 10 0.0 549.36 0.02174 e.7507 c.634e 1.2851 545.9 1109.0 0*.40507 c.42681 See.6 442.8 1151.4 e.7993 0.6373 1 2863 544.5 1109.3 1030.0 Set.le e.02170 0*.40976 0.63146 147.1 444 7 1851.8 1020.0 566.99 0.02866 0.41450 e.7478 0.6396 1 287e 5g3.0 1109.6 0.43620 545 6 646.6 1152.2 8.7e63 e.6423 1 22t6 !*1.5 1809.9 1010.0 545.79 e.02163 0*.41991 0.44103 544.1 648.5 1152.6 9.?t99 c.6449 1 2298 540 0 1810.1 2000.0 544.53 0 02159 0*.42436 0.49595 Se2.6 650.4 1192 9 8.7934
%g3.0 5s3.26 0.02155 0'.=2942 0.45097 c.6t?6 1.2910 538.6 1810.6 141.0 652 3 1153.3 a.7e19 0 6503 1 3922 537 1 1110.7 9to.0 562.1= 0.021*2 0*.g3457 C.45409 529.5 659 2 1863 7 570.0 5=0.50 e.0214 e s.7404 e.6530 1.293e 535.6 3115 0
- C*.439e2 c.46130 *37.9 656 1 115m.0 0.7349 SbO.c 524.t5 0.02165 0'.6451e e.6557 1 25e6 53s.0 1111.2 0 46662 536 3 658.0 11%e.4 e.7373 8.6584 3.2958 522.5 ,1111 5 450.0 529 29 0.02141 0*.6506% 0.47205 534.7 660.0 1154.7
%.C.0 527.13 0.02137 c.7J5e 8.6612 1 2970 521.0 1111.7 l 0'.m5623 0.=7759 123.2 661.9 1855.1 0.7302 0.66he 1.3922 *25.4 1112.0
%2*.0 525.et 0.0212* Okh6150 0.he22e 531.6 663 8 1155.4 ]
526.0 52. 56 e.02130 s.7327 0.6668 1 2995 527.9 1112 2 0146770 0.62501 520 0 665.0 1155.7 e.7311 0.1696 1.4007 526.3 1112 5 530.0 521.2o 0.02127 0 47362 0.49490 528.3 667.7 1856.1 S.7295 0 6724 1.4019 224.s 1112*7 500.0 528.95 0.02123 C*.4796e 0.50091 226.7 669.7 1196.4 t40.0 8.7279 e.6753 1.4032 523.2 1 13.0
- 0.43 0 02119 C.49526
~
0.50706 525 1 671.6 13%6.7 e.7263 0.67e2 1.4045 521.6 1113 2 teo.O 529.30 0.02816 C~.4521e 0.51333 223.4 673 6 1157.0 370.0 527.56 e.7297 0.6011 1.9057 520.0 1153 4 0.02112 0 49263 0 51575 *21.0 675.6 1857.3 S.7230 0.6240 1.4070 513.4 1113.7 etc.e 526.40 e.02109 0.50522 0.52631 520 1 677.6 1157.7 0 7214 0.6e69 1.4063 516.7 1813.9 F50.0 525.2% 0.02105 OkS1157 c.53202 518.4 679.5 1118.0 t.0.0 132.96 e.7397 c.6e99 1.4096 515 1 181*.1 e.02101 0*.51rP6 0 535te 586.7 681.5 1898.2 s.71ro 0.6529 1.4109 533** III"*3 420.0 522.66 0.02098 C*.52*S2 0.54tt9 *15.0 6e3 5 t20.0 521.06 1890 5 8.7161 0.6659 1.4122 533.8 133'*5 e.02054 0*.5331* 0.55600 113 3 685.5 11$e.8 8.71e 6 0.6590 1.4126 510.1 litt.8 e10.9 bl9.t* e.02091 e.54052 0.56143 518.6 687.6 1159.1 0.7129 a.7020 1.4149 50e** 181S*'
200.0 ** 518 21 .e.00007 n'.CC09 0.0156 0090 0 rh_nKnc n5w ^ mm n n - ea no.= a
l
' Table 2. Pnpertiss of Ss.turated Steam and Saturated hier(Pressure)
) )
4 Press. Temp. Volume, ft*/lbm Enthalpy, Bru/Em
. Entropy, Bru/lbm x R Energy, Btu /lb:n i psia F Water Evap. Steam Water . Evap. Steam Water Evap. Steam i v v v A Water Steam .
i I is e i A
le A a a s e, =
1 e i le a 8 1
200.0 3en.ep e.01839 2.2689 2 2e73 355 5 e42.s 11se.3 158.0 2eo.9m 0 01838 2.2912 2.3095 e.543e 1.ee16 1.5454 354.s 1113 7 '
354 6 a=3.6 119e.2 0.542e 1 0035 1.5463 353*9 1113*'
a 156.0 340.12 e.01836 2.3139 2 3J22 253.7 844.4 1158 1 14m.0 379.26 e.01035 2.3370 2.3554 0.5437 1.0054 1.5471 353.g 1113 5 352.s e45.1 1157.9 a.5406 1.0074 1 54ee 252 1 1813.4 192.0 378.40 e.01834 2.3604 2 3790 251 9 845 9 1157.e 0 5395 1 8094 1 5489 351 2 1113.2 190.0 377 53 e.01033 2.3847 2.483e 250 9 e46.7 1197.6 1e4.0 376.65 e.01832 2.4093 2.4276 0 52e4 1.85L3 1 54ge 350 3 1113 1 350.e 847.5 1157.5 0.5273 1.0133 3 5507 349.4 3113.s 4 125.0 375.77 e.01431 2.4344 2.4527 349 1 848 3 1157.3 1 les.O 374.8e 0 01330 2.4600 *2.47e3 0.5262 1.0153 1.5116 34e*4 111E*9 348.1 449.1 1157.2 0.5351 1.017e 3 5125 347.5 1812.0 182.e 373.58 0 01e28 2.4e62 2 5045 247 2 e49 9 1197.0 e.5339 1.o194 1 553, 346 5 1812 7 180.0 373.0a e.01827 2.5129 2 5312 246 2 850.7 1156.9 17c.0 372.16 c.01e26 2.5*02 2.5555 e.5328 1 0215 1 1543 245.6 1112 5 245 2 851.5 1156 7 8 5316 1 0226 1 5552 244.6 1112.4 176.0 371.2m 0 01e25 2.5681 2.5864 244 2 852.3 4
37 .o 370.31 0 01e24 2.5966 2.6149 1156.5 8.5305 1.0257 1.5542 343 6 1112 3 243 2 453 1 1156 4 e.5293 1.0279 1 5571 342.7 1112 2 172.0 369 37 0 01e23 2.6250 2 6440 242 2 e53.9 1156.2 e.5201 1.e200 1.5531 341 7 3112.0 170.0 368.42 0 01821 2.6556 2.6738 241 2 054 8 1196.8 367.47 c.01320 e.5269 1.8322 1 5591 348*7 111I*'
165.0 2.6961 2.1043 340 2 855.6 1155.s 3.5256 166.0 3t6.50 0 01819 2.7173 2 7355 229 2 85e.5 1.0244 1 5601 339.7 1111*8 164.0 365 53 3155.7 e.5244 1.e367 1 5611 338.6 3111 6 0 01818 2.7493 2.7674 '33a.2 857 3 1195.5 e.5232 162.0 34g.5m 0 01817 2.7820 2 3001 1.0389 1 5621 331.6 1111.5 227.1 e58 2 1155.3 0.5219 1.e412 1 5631 336.6 1111.3 160.0 363.55 e.01815 2.0155 2.0336 336.1 859.e 1195 1 150.0 362.55 0 01s14 2.e49e c.5204 1.e435' 3 5661 335 5 1131 2
' 2 9679 235 0 859.9 1154.9 e.5194 1 0458 1 56?2 324 5 1118.0 156.0 361.53 0 01e13 2.se*9 2.9031 233 9 e60.e 1154.7 154.0 3t0 51 c.01812 2.9210 8.51e1 1.04e2 1 5662 233 4 1110.9 2.9391 322.s 861 6 1154.5 0.5868 1.e506 352.0 359.4e 8 01010 2.9579 2 9760 1 1673 332 3 1180.7 231.s 862 5 1894.3 e.5154 1 0530 1 56et 333 2 1110.6 150.0 3Se.43 e.01809 2.9958 3 0139 230.6 863.4 3194.1 Igg.g 357.91 0 0130s 3.0151 0 5141 1.8554 1 5695 330.1 1130.4 3.0322 330 1 863.9 1154.0 e.5134 1 0566 1.5700 229.6 1110.3 r 148.0 357 3e s.01eos 3.03=7 3.0528 229.5 e64.3 147.0 316.eg 0.01007 3.05%5 3 0726 1153 9 e.5127 1 0579 3 57C6 225.0 1110 3 229.0 864 8 1153.s e.5120 1 0591 1 5712 324 5 1110 2 146.0 356.31 0 01806 *3.0746 3 0927 228.4 e65 2 1153.6 e.5114 1.0604 1.5717 327.9 1110.1 4
145.0 355.77 C.01e06 3.0950 3 1130 227.e 865.7 1193.5 144.0 355.23 0 01805 3.1156 3.1317 227 3 S.5807 1.0616 1.5723 327.4 1110*C 343.0 3*m.69 0.Cle05 3.1365 e66.2 1153.4 e.5 tee 1 8629 3.!?29 226 8 1109.9
'l 3 1566 226 7 sh.6 1153.3 0.5e93 1.0642 1.5734 226 2 1809.8 142.0 354.1* 9 01e34 3.1577 3.1757 326 1 e67.1 *1153.2 i
141.0 353.59 0 01803 3.1792 3 1972 e.50e6 1.0655 1.57 0 225 6 3109.'P 225 5 e67 5 1853.1 8.5079 1.0666 1 57*6 325.1 1109.7 l
140.0 353.0, t.01803 3.2010 3 2150 225.0 96e.0 1153.0 l
139.0 352.4e 0.01e02 3.2220 3.2=11 8.5071 1.04e1 1.5752 224.5 1199.6 351.52 0 01801 324.4 e6e.5 1152.s 8.5064 1.069* 3.571s 223 9 11NS 13e.0 3.2*54 3 2634 223.s e68.9 11$2.7 137.0 351.26 0 01001 3.2681 3.2861 e.5857 1.e707 1.5744 323*3 1389'"
223 2 e69.4 1152 6 0.5050 1.e720 1.577e 222 7 W 9.3 i 136.0 350.79 e.01000 3.2512 3 3051 222 6 e69.9 1152.5 ,
3.5e43 1.073J 1 5776 122 1 1109.2 135.0 350.23 a.01759 3.3145 3.3325 322 0 47e.4 1192.4 124.0 349.6% 0.01799 3.3142 3.3562 e.5e35 1.0747 1.5742 228 5 1139 1 e.0179e 321.4 370.e 1192 2 e.5e28 1 0760 1.5788 320.9 1109.0 133 0 349.00 3.3622 3.3802 320 8 871.3 1152 1 132 0 34a.50 0 01797 3.38 % 3 4046 e.Sete 1 8774 1 5794 320.3 Iles 9 320 2 371.s 1192.s 0 5e13 3.e7ee 339.7 Stee.o 131 0 3*7.52 8 01797 3.4113 3.4293 219.6 872 3 1.5ese 1191.9 c.5ee5 1.eeen 1 5007 319*1 1188*7 330.0 347.33 8 01796 3.4364 3.4544 219 0 472.s 1191.7 125.0 346.74 0 01795 3.4619 3.4799 e.490s 1 0915 1.5013 310 5 1138 6 e.01794 210 3 873 3 1191.6 e.4990 1.et29 3 5819 317*9 II'8*I 128.0 346.15 3.4870 3 5057 317.7 873.0 1191.5 127.0 3555 0.01794 3.5141 3.5320 4.49e2 1 0043 3 1e26 317*3 1888**
217 1 814 3 1191 3 e.4975 1.se50 1 5822 31&*7 1388*3 324.0 344.55 0 81793 3.5407 3 5586 316.4 474.s 1151 2 e.4967 1.se72 1 1839 31&*8 3888.1 325.0 344 35 e.01792 3.567e 3.5857 215.e 875.3 1191.1 i 343.14 e.01792 3.5953 S.4959 1.cee6 1 58*5 215 4 1106 1 324.e 3.6132 315.2 875.0 3 50.9 e.4953 1 0901 31m.e 1190.0 l 123.0 3*2.13 e.0!?91 3.6232 3.6411 214.5 876 3 1140.0 1.Tett 122.0 342.51 s.01790 3.6516 2.M95 e.4943 1 0515 1 545e 21"*1 1887*9 121.e 341.e5 313.9 sh.e 3150.7 e.4935 1.e930 1.5665 313 5 Sie?.e
.e.01750 3.6004 3.6583 213 2 477 3 1190.5 e.4927 1 8945 3 5372 312.8 1187.7 120.0 341 27 e.81789 3.7097 3.7275 312.6 877.0 1190.4 114.0 340.e4 e.017ee 3.7394 3.7573 0.4919 3.9960 1 5e79 212 2 11e7.6 e.01787 211.9 870 3 115e.2 e.4911 1.e975 3 5645 311.5 11o7.5
- 13e.0 240.01 3.7697 3.7e75 311 3 e78.0 117.0 319.37 S.017e7 3.3004 3.81e3 1890.1 c.4903 1.e990 1.5952 310.9 1107.4 116.e 310.6 879 3 1849.9 e.4e94 1.1005 1 5e59 310.2 18e7.3 328.73 8*01786 3.0386 3.84$5 209.9 479.9 11e9.e e.4806 1 1021 3 5906 209.5 1807 2 115.0 3:e .f,3 0 01785 3.e634 3. eel 3 209 3 000.4 11e9 6 i 114.0 327.s3 0.017e5 3.8957 3.9126 208 6 e.4077 1.1036 1 5913 200.4 1107.0 112.0 880 9 18e9.5 e.4e69 1 1052 3.5921 308.2 1806.9 3:6.7s 0 017th 3.9286 3.94e4 207 9 391.4 1149.3 0.4e60 1 1967 1.5928 2e7 5 1806.s 112.0 314 12 C.017e3 3.9620 3.9755 207.2 111.0 802.0 11e9.2 0.4052 1 1093 1 5915 306.e 1106 7 325.66 0 017e2 3.9560 4.013e 206 5 se2.5 1139 0 e.4e43 1 1059 3 55*2 306 1 180o.6
, 110.0 324.79 0 017e2 4.0306 4.84e= 205.e se3.1 1106.5 109.0 3!".11 S.011e1 e.4e34 3.1115 1.5950 205 4 4.065e 4.De37 '205 1 e83.6 11ee.
11a .79 e.4426 1.1332 3.!S57 20**7 13 " *3
, ICt.0 323.6 e.017e0 4.1017 4 1195 204.4 107 0 322.7) etw.1 Iges.5 e.4eg7 1.11ge 1 55m5 20m.0 1306 2 0.01779 4.1382 4.1560 203.7 etw.7 11ee.4 s.4600 3.1165 3 5972 303 3 1106.1 106.8 322.C6 0.01779 4.1753 4 19J1 2e3 0 805.2 11ee.2 e.4799 1 1103 1 5See 302 6 1106.0 103.0 331 37 0.01178 4.2132 4.2309 202 2 885.8 Ste.e 330.67 0 01777 Iles.O e.4790 1.115e 1 5948 201 9 1185.e 4.2517 4.2695 201 5 886.4 11e?.9 e.47e9 1 13g5 1 5955 201.2 1805.7 102.0 229.57 e.01776 4.2980 4 3087 200 8 ese.9 102.0 229.26 13e7.7 e.4771 1.1232 1.e013 200.4 1185 6 S.01776 4.3310 9.3487 200.0 407.5 1187.5 0.4762 1 12m9 1 6011 299*7 1385**
AW1 0 320.!* 9 01775 4.3717 4.3e55 399*3 See.1 1187 3 e.*752 3 1267 3*6019 89'd II'$*
i m 3,7.= ... m t 4.4133 .4m 15. 5 .. .m72 e 743 3.u.4 1.m 7 2a.2 3 :*=
A_ - - - - - -
MA STEk COia Section 1 Answers - Principles of Nuclear Power Plant Operations, Thermodynamics, Heat Transfer and Fluid Flow.
1.01 a. Keff decreases due to fission product buildup (0.5) and fuel burnup (0.5).
- b. Burnable poison is burning out (0.5) and Pu-239 is building in (0.5).
Reference:
Standard Nuclear Principles 1.02 (a) Reactor critical on prompt neutrons alone. (0.5)
(b) Time in seconds required for power to change (0.5) by factor of 'e'.
(c) The multiplication of neutrons by the fuel in a (0.5) subcritical reactor. . 3 - . . + , . .
4 u - ,
< -.., n m. % . , , , , ,
Reference:
Standard Nuclear Principles 1.03 (c) (1.0)
Reference:
Standard Nuclear Principles 1.04 (b) (1.0)
Reference:
Standard Nuclear Principles 1.05 (b) (1.0)
Reference:
Standard Thermo-Hydraulic Principles 1.06 (c) (1.0)
Reference:
Standard Nuclear Principles 1.07 (c) (1.0)
Reference:
Standard Thermodynamic and Fluid Flow Principles 1.08 (d) (1.0)
Reference:
Standard Thermal Hydraulic Principles and Applications 1.09 (b) (1.0)
Reference:
Standard Nuclear Principles 1.10 (d) (1.0)
Reference:
Standard Doppler Coefficient Principles 11
l l
1.11 (a) (1.0)
Reference:
Standard Nuclear Principles 1.12 Increase in pump flow due to loss of backpressure. (0.5)
The increased flow causes the motor to draw more (0.5) current and possibly damage the motor winding.
Reference:
Dresden Heat Transfer and Fluid Flow 1.13 a. Pressure due to water column (0.5)
- b. Pressure as measured in the steam dome (0.5)
- c. Pressure loss due to irreversible flow losses (friction) (0.5)
- d. Temperature will also be accepted.
Reference:
GE Manual, Section 7, pages 94-96 1.14 a. tau === 330 sec. Lambda-ef f Delta-k/k .1(.0005) (0.5)
- b. P = P o e t / tau implies 10 = e t/130 implies in (0.5) 10 = t/130 implies t = 130 in 10 = 299 sec.
- c. tau === 60 sec. 1(.0001) .0001 (0.5)
Reference:
GE Reactor Theory, page 11 and 12 1.15 a. Neutrons produced indirectly from the fission process. (0.5)
They are born from the decay of the fission programs or delayed neutron precursors.
- b. 80C .0072 (.010) EOC .0055 (.010) (0.5)
- c. The . Pu-239 has a Beta of .0002 and Pu-239 accounts (1.0) for 30% of all fissions at the E0C. The EOC value is a weighted average of P239 and U235 fractions and 1 smaller than BOC because Beta Pu-239 is smaller than Beta U238.
Reference:
Q&A Bank Reactor Theory Section 1.16 a. As the reactor operates at power, Xenon builds into (0.5) equilibrium, adding negative reactivity, causing power to decrease.
- b. A rod withdrawal from a high power region will cause (1.0) a power increase in the adjacent fuel rods because of being closer to thermal limit and therefore cause damage.
Reference:
Dresden General Physics BWR RX Theory 12
1.17 a. a Decreases (0.5) b Decreases (0.5) c Decreases (0.5) d Decreases (0.5)
- b. 1. MAPRAT is the ratio of APLGHR TO Limit APLHGR OR (0.5) the ratio of APLHGR (act) to MAPLHGR (LCO)
- 2. NO (.25)
Reference:
HT&FF, page 16 & 17, GE Thermodynamics, HT&FF, pages 9-85, to 9-89 1.18 Photo-neutron source (Naturally occuring deuterium and fission (0.5) or decay gammas react to form hydrogen and a neutron).
Spontaneous fission (Uranium, plutonium and curium undergo (0.5) spontaneous fission. Curium is the most significant producer of neutrons).
Alpha-neutron reactions (Oxygen-18 is uranium oxide fuel (0.5) reacts with an alpha particle to produce a neutron).
Reference:
Theory Review, page 28 & 30 END OF SECIION 1
! 13 w _ _ - _ _ __-
Section 2 Answers 2.01 a. Flow control valve would close (0.5) j b. Scram valves would open under spring pressure and (0.5) control rods would be inserted.
l
! c. Instrument volume vent and drain valves would close. (0.5)
- d. Main feed pump minimum flow valve would open. (0.5)
- e. Both the main feed regulation valves would fail as is. (0.5)
Reference:
Instrument Air LP, Rev. 3, page 14.15 2.02 a. Standby FRP must be selected
- b. Suction press must be greater than 120 (+/-2) psig
- c. Vent fan must be on
- d. Oil press must be greater than 20 (+/-2) psig (4 @ 0.5 ea + 2 s.p. @ 0.5 ea) a
Reference:
Dresden Cond & Feed Lesson Plan Table 2 2.03 a. Will inject (0.25). Turbine seal leakage resulting (1.0) in potential airborne activity in the HPCI room (0.75).
- b. Will not inject (0.25). Turbine stop and control valves (1.0) will not open (0.75).
- c. Will inject (0.25). Pump overheating and seal damage (1.0) may result during low or no flow conditions (0.75).
4
- d. Will not inject (0.25). Maximum signal from the flow (1.0) element will cause the controller to keep turbine speed at minimum (0.75).
Reference:
Dresden HPCI Lesson Plan 4
2.04 a. Either: Internal damage to mechanism or rod will scram (1.0) i slowly on seal leakage.
I l b. High temperatures in the CRD; inability to move the rod; (2.0) l discharge of scram accumulators (2 required).
I
- c. Either crud buildup in the recire. pump seals or reduced (1.0) seal lifetime.
Reference:
CRD Hydraulic Lesson Plan i
I 14 l
l j
l 2.05 a. The CROM separates from the control rod blade. (0.5)
! b. The drive mechanism is withdrawn. (0.5)
- c. The blade sticks in the core and does not follow (0.5) the drive mechanism,
- d. The blade then breaks free and drops to the position (0.5)
! of the drive mechanism, thus rapidly inserting reactivity and generating heat in the fuel.
Reference:
Dresden RWMS Lesson Plan, page 2 2.06 a. Low flow through an operating fan (will accept) any one butterfly damper is closed,
- b. A group II isolation.
- c. R.B. exhaust or refueling floor high radiation,
- d. Low voltage on a bus.
- e. Abnormal reactor building air pressure.
- f. R. B. exhaust high radiation.
(Any 4 @ 0.5 ea)
Reference:
Dresden RB Vent Sys Lesson Plan, page 7 2.07 a. False (0.5)
- b. True (0.5)
- c. False (0.5)
- d. True (0.5)
Reference:
Fire Protection L.P., Rev. 7, pages 3, 7, 12, 16, 17 2.08 a. Clean, domineralized water system (0.5)
Condensate transfer system (0.5)
Fire protection / service water system (0,5)
- b. Valve (1301-3) is open.
/? !.'
Reactor pressure is 10% psig after 15 second time delay:
1 A group I isolation signal is present.
A group V isolation signal is present .
(Any 3 0 0.5 ea) i
Reference:
Isolation Condenser L.P., page 3 & 4 4
15 e.p ---.---------,-.--,w-r--,- - - - , -A -
wmew ,- .--_-.--w-y--+ > - . ,--.,- , - aw -.we t ,, .->-g-----w.------ - - - -
I.
I i
2.09 a. Increase (0,5)
- b. Increase (0.5) -
] c. Decrease (0.5)
- d. Increase (0.5) l l e. Increase (0.5)
)
Reference:
BWR Training Center, Morris Book 1, Lesson Plan #4, pages 6-13 e
i
! END OF SECTION 2
!I a
I i
1 0 \
1
)
i i
i l
4 4
4 i
i 1
4 T
i I
i r j i i i L
J 4
4 t
?
I f
l i
i i
l
+ i l
i 16 i
i l
Section 3 Answers 3.01 a. By monitoring the differential pressure (0.5) across each recire pump (0.5) for a 2 psig or greater dp, indicating the pump is running (0.5).
. b. By comparing the pressure in the riser pipes on one (1.5) l recirc loop with the pressure in the riser pipes of j the other loop. The undamaged loop will have a higher pressure than the damaged loop.
- c. Loop B (0.5)
Reference:
LPCI LP, Rev. 5, page 15 3.02 . An off gas high-high radiation in the 15 minute holdup volume.+ (0.5) Both detectors > high-high (0.5) or one >
high-high and the other downscale (0.5) starts a 15 '
'M A. a l
minute time delay (0.5).... we" -su ,t Y sd a +.. A > '
V.',T y t N '4
- m o, ( v-sm .; & (e.n <9 S
C "(
Reference:
Dresden Off-Gas Sys Lesson Plan, page 13 .y g.g. ,j, g i .,,.,,,...ar w,. . ,.
3.03 As the maximum combined flow decreases below 100%, the control valves will start to close and the bypass valves will remain closed (1.0). As the control valves continue to close, reactor pressure increases (0.5). The reactor will scram on high flux -
high pressure (0.5). Reactor pressure will continue to increase until the safety / relief valves lift and then reseat (0.5). The isolation condenser will initiate (0.5). The final steady state conditions will be the reactor shutdown and pressure maintained with the isolation condenser (1.0).
Reference:
Dresden EHC Lesson Plan.
3.04 Automatically bypasses both the RBM channels. (0.5)
Reference:
Dresden RBM Lesson Plan, page 6 3.05 a. RBMs
- c. The process computer
. d. Full core display
- e. 902-37 back panel meters.
(Any 4 @ .05 ea)
Reference:
Dresden Lesson Flan LPRM, page 18 4
1 17
_ _ _ ~. _ _ _ _ _ __ ..
a e
6 3,06 a. Prevents the MG set and the reserve power source (0.5) from simultaneously supplying a RPS bus.
- b. Prevents selecting the reserve power supply to more (0.5) than one RPS bus.
Reference:
Dresden Lesson Plan RPS 3.07 a. rod block , (0.5) co o a - '
- b. Aalf-scram ne re*3 w ; "' o n (0.5)
- c. rod block (0.5)
- d. no reactor protection system action (0.5) i
Reference:
Dresden Lesson Plan RPS, APRM Systems 3.08 a. A lower argon pressure (0.5)
- b. Less total Uzas in the U30s coating (0.5)
- c. Reduced detector operating voltage (0,5)
Reference:
Dresden Lesson Plan IRM i 3.09 a. Self-actuation on high pressure (0.5)
- b. Pilot actuation on ADS signal (0.5)
- c. Pressure relief function (0.5)
Reference:
Dresden Lesson Plan ADS 3.10 a. Feedwater level control system (0.5)
- b. ECCS control circuitry Reactor Protection System j Isolation condenser control circuit Safety / relief and electromatic relief valve controls (Any 4 @ 0.5 ea)
- c. Jet pump flow will cause level to indicate higher than (0.5) actual
Reference:
Dresden Nuclear Boiler Instrumentation lesson Plan, page 12 & 23 3.11 a. Bypasses all interlocks to insert the rod except (0.5) the rod worth minimizer insert block and any select block.
- b. Bypassing the timer. (0.5)
- c. No settle function. (0.5) 1 j
Reference:
Dresden RMCS Lesson Plan, page 7 l 4
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, _ _ . _ _ _ . . _ . ~ _.. . _ _ _ . _ __ . _ _ _ _ _ . _ _ . _
f 3.12 a. Speed Signal Failure 3. Less than 1.0 ma output (0.5)
, from function generator
- b. Incomplete Sequence Trip 1. Below 4 psid, 28 sec. (0,5) after start
- c. Recire Pump Low 4. Below 4 psid (0.5) 4 Differential Pressure i
- d. Recirc Pump Locked 5. Below 4 psid, 30 sec. (0.5)
Rotor Trip after start
- e. Recirc Loop Flow 2. Feedwater Flow below (0,5)
Limit 20% or discharge valve
- not full open
Reference:
Dresden Ricirc flow control lesson plan - page 15 i
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END OF SECTION 3 4
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Section 4 Answers A
4.01 a. Complete Emergency Systems checklist (Appendix A) (0.5)
I
- b. Review the previous shifts Daily Surveillance sheets. Verify (0.5) results as logged are within Technical Specification limits.
- c. Review the NSO's log from the last date on shift or for the (0.5) preceding four (4) days whichever is less.
- d. Review Night Orders and Degraded Equipment Log. (0.5)
- e. Comp lete the NSO's turnover checklist for the on-coming NSO (0.5)
(Appendix A)
Reference:
Dresden Operating Orders DAP 7-2, pages 2 & 3 4.02 a. Attempt to restore RBCCW flow. (0.5)
- b. If R8CCW flow is lost and cannot be restored within one (0.5) ,
minute, TRIP Recirculation Pumps A and B. '
- c. Verify that a reactor scram occurs when drywell pressure Ger5) reaches +2 psig (0.5) or MANUALLY SCRAM the reactor if l equipment damage appears imminent (0.5).
)
Reference:
Dresden DOA 3700-1, Page 1, RBCCW 4.03 a. Slow I
- b. Fast (clockwise)
- c. Incoming
- d. Higher
- e. Running j (5 @ 0.5 ea) 1
Reference:
Dresden DGP 1-1, page 19 of 26 steps 5.1 and 5.m 4.04 a. An area where you could receive > 100 mrem in one hour (1.0)
- b. A weighted half-life of a radioactive material which (1.0) takes into account the decay characteristic (physical half-life) of the material and retention of the material within the body (biological half-life). ;
I
- c. 3.7 x 10 to the 10 power disintegrations /sec. (1.0) 4 20
}'
- d. Measure of dose of any ionizing radiation to body (1.0) tissue in terms of its estimated biological effect relative to a dose of I roengten of x-rays.
Reference:
10 CFR 20 4.05 a. 25 rems (0.5) l b. 75 rems (0.5)
J
Reference:
Dresden Radiation Protection Standards, page 26 4.06 a. Manually scram the reactor (0.5)
- b. Verify all rods in (0.5)
.i
- c. Leave the Mode Switch in Run (0.5)
- d. Trip the Control Rod Drive pump (0.5)
- e. Trip the turbine (0.5)
Reference:
Dresden EPIP 200-20, page 1 4.07 a. 1. Scram the reactor manually by depressing both l manual trip buttons.
l 2. Trip both A and B Reactor Recirculation MG Sets.
j 3. Place the Mode Switch to Shutdown.
4 Monitor nuclear instrumentation for decreasing neutron flux and local areas of high reactivity.
l 5. Trip hydrogen addition (Unit 2 only) .
(5 @ 0.40 ea)
- b. If either two (2) or more adjacent rods are not inserted (1.0) past the 06 position OR thirty (30) or more rods are not
! inserted past the 06 position AND if reactor water level j cannot be maintained OR suppression pool water temperature cannot be maintained below 110'F.
j c. 1. Amber pilot of Squib firing continuity circuit not lit
- 2. Flow Indication Pilot Light lit
! 3. Water clean-up system isolation
- 4. Decreasing level of Standby Liquid Storage Tank i
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- 5. Star.dby Liquid Squib Valve Circuit Fail Annunciator Ligh: lit
- 6. Punp discharge pressure increases (5 9 0.40 ea)
Reference:
Dresden DGA 18, page 7, 00P 1100-2, page 1 of 3 and 3 of 3
! SBLC Lesson Plan, page 14 1
4.08.a. Verify that the Main Steam Bypass Valves are controlling (0.5) reactor pressure. MONITOR reactor pressure by multiple indications.
- b. VERIFY that the Reactor SCRAMS when reactor pressure (0,5) reaches 1060 psig. Follow the Unit 2/3 Reactor Scram
. procedure, DGP 2-3.
i
- c. VERIFY that Relief Valves OPEN when relief valve (0.5) setpoints are reached.
- d. VERIFY isolation condenser initiation if reactor pressure (0.5) ,
t remains greater than 1070 psig for longer than 15 seconds. '
- e. VERIFY that A and B Recirculation M.G. Set Field Breakers (0.5)
. TRIP ff reactor pressure reaches 1240 psig.
Reference:
Dresden DGA-6, page 1 & 2, Reactor High Pressure 4.09 d. 40% power (0.5)
- b. 35% power (0,5)
- f. Convenient time below 200 MWE (0,5)
I a. 2 x 10 E+6 lbm/hr feed flow or 20% (0.5) +
- c. Feedwater total flow is less than or equal to 10% of rated (0.5)
- e. Reactor power is between 5% and 10% (0.5)
(Items f and a may be reversed)
Reference:
Dresden DGP 2-1 page 4 of 13 thru 10 of 13 END OF SECTION 4 22
, -- .. . . - - . . - - . - - - - . _ . - . - - - . . . _ _ - _ - . . _ . - _ ~ -
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i j U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION k f .
FACILITY: Dresden 2-3 l
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p REACTOR TYPE: BWR-3
) [
j V DATE ADMINISTERED: June 10, 1985
! EXAMINER: McMillen/Lang APPLICANT:
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2NSTRUCTIONS 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 i parentheses after the question. The passing grade requires at least 70% in each j category and a final grade of at least 80*.
4 i
% Of -
Category
% Of Applicant's Category Value Total Score Value Category
_24 24.1 5. Theory of Nuclear Power Plant Operation, Fluids, and Themodynamics 25 25.3 6. Plant Systems Design, Control, j !
i and Instrumentation b
25 25.3 7.
Procedures - Normal, Abnormal, Emergency, and Radiological Control j f 25 25.3 8.
i Administrative Procedures, Conditions, and Limitations I
l 100 100
- j. TOTALS Final Grade %
i n All work done on this exam is my own, I have neither given nor received aid. I l
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- 5. Theory of hucicar Power Plant Operations, Fluids and Thermodynamics 5.1 The change in reactivity associated with a change in Keff (1.0) frem 0.920 to 1.004 is approximately: -
- a. 0.091
- b. 0.084
- c. 0.087
- c. 0.080 5.2 Which of the following is NOT a characteristic of subcritical (1.6) ruitiplication,
- a. For equal reactivity additions, it takes longer for the equilibrium subcritical multiplication level to be reached as Keff approaches unity,
- b. If the reactor is shutdown long enough, the source rarse instruments will lose their ability to detennine the subtritical multiplication level even though the core may still be at the middle of life,
- c. If 2 notches of rod withdrawal increases the sub-critical multiplication level by 10 cps, then 4 r.ctches or rod withdrawal will increase the subcritical multiplication level by approximately 20 cps.
- d. Doubling the indicated count rate by reactivity accitiers will reduce the nargin to critical by approximately one half.
5.3. Which reactivity coefficient is the most dominant when (1.0) '
pulling rcds in startup and 150*F.
- a. Pressure coefficient
- b. Mcderator coefficient
- c. Doppler coefficient
- d. Void coefficient 5.4 The highest internal stresses placed on a pressurized (1.0) systen boundary such as reactor vessel is:
- a. on the thickest components during a heatup
- b. on the thinnest ccmponents during a heatup
- c. On the,thTin'est) n components during a cooldown 4M
~
d r 'on the th't m 651 ccmponents during a cooldown ,
5.5 The need to change the RTNDT of the reactor vessel (1,0) cver the life of the plant is the result of:
- a. therral cycles (heatup and cooldown transients)
- b. pressure cycles (charges in pressure)
- c. gancia radiation
- d. neutron radiatien 2
. - -. - - _ . -. . --- - . - - - . - - _ _ - _ - . - _ - . _ . --_ - =-.- . .
1
- 5.6 During a design basis accident of a loss of coolant, (1.0)
I the therral limit that protects the fuel is i
- a. Lh6L ,
4 eb. QPLGHR) HG '
{ -c. Mbi W s0 C.
i d. MCPR i
l 5.7 The quality of steam to the turbine refers to: (1.0) 1
- a. the ratio of the vapor mass to the sum of the - -
l liquid and vapor masses.
- b. the ratio of the liquid mass to the sum of the liquid and vapor masses.
- c. the ratio of the liquid mass to the vapor mass.
j d. the ratio of the vapor mass to the liquid mass.
5.8 hhich of the following statements describes the effect (1.0) the r:agnitude of the initial level of source range counts would have on critical rod position and the power level (count rate) at criticality:
l a. The critical rod position would not be affected ,
- by the source range count rate and the power level l (ccunt rate) when criticality is reached would be higher.
~
- b. The critical rod position would be lower since the 3 scurce range count rate is higher and the power s
level (ccunt rate) when criticality is reached
- would be higher,
- l. c. The critical rod position would be lower since the j source range count rate is higher and the power '
{ level (count rate) when criticality is reached i would be lower.
)'
- d. The critical rod position would not be affected by the source range count rate and the power level (cour.t rate) when criticality is reached would !
be icv.er.
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~5.9 Which of the following statements describes the reason (1.0) for the change in critical power as the reactor pressure i ir. creases from 800 to 1100 psi.
- a. a decrease in local quality occurs which causes t the margin between actual and critical qualities to increase and thus power increases.
- b. Stean bubbles increase in quantity and collect more j readily at the !. eat transfer surface, thus making the transition between nucleate and film boiling - !
j casier, i 1
) c. Local quality decreases along the boiling length
! ard thus a greater critical power is necessary to
! drive quality up to the critical limit.
i l d. Eoiling lenoth decreases so the voids are formed i Icwer in the core and thus the critical quality
! is greater Icwer in the core and transition boilings !
j occurs at a point below the midpoint of the core.
j j 5.10 EXFLAIN or DEFINE the following terms: ,
! a. Prcept Critical 0.5
- b. Reactor Period 0.5 j c. Suberitical Multiplication 0.5 5.11 (Assun.e 100 power) Then reactor power is reduced by (1.0)
- driving rods. The recirculation pump speed remains 4 constant. Ccre flow changes because of the actions j taken. Cheese the proper reason for the core flow change.
- a. Flcw will decrease because of aii increase in two .
j phase flow conditions. i 4
- b. Flow will increase due to the increased natural
- circulation ,
- c. Flow will increase because of a reduction in two
- phuse ficw conditions.
- o. Flow will increase because of an increase in '
two-phase flow conditions.
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i 5.12 The MAPLhGR curve increases early core life and as (1.0) .
expcsure increases the limit begins to increase at a '
- decreastro rate and then decreases. Choose the concition'that is NOT responsible for this: .
s l a. Burnable poison depletion j b. Fission gas build-up
- c. Local peaking factors :
) d. Reduccd heat transfer rate !
) 5.13 Which of the following statements is correct concerning -
(1.0) l control roc worth? -
l i a. It is proportional to reactor power l It is Icwer in regions of higher relative
- b.
j neutron flux i ,
j c. It is proportional to rod speed l d. It is higher in regions of higher relative ;
- neutron flux l i !
- 5.14RegardingMCPR(MinimumCriticalPowerRatio)
- [
j i i a. What FHEf.0MEh01 could exist if a fuel bundle were (1.0) l operateo at a MCPR LESS THAN OhE and WHAT would I very likely be the CONSEQUENCE of the phenomenon? {
. i.
l b. WHY cust the Technical Specification MCPR limit (1.0) ;
include a 'K' factor when core flow is LESS THAN -
! RATED? :
?
! c. HOW is the margin to MCPR changed (INCREASES, (0.5) '
! DECREASES, or REMAINS CONSTANT) when inlet sub-1 cooling cecreases?
i <
5.15 Which of the following radioactive isotopic found in (1.0) !
1 the reactor coolant would not indicate a leak through j i the fuct cladding? ;
i l a. 1-131 f j '
- b. Xe 133 i
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- s. t
- t 1
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e ,__ . .
f 5.16 A moderator is necessary to slow neutrons down to thermal (1.0)
I energies. Which of the following is the most correct reason for operating with thermal instead of fast neutrons.
i a. Increased neutron efficiency since thermal neutrons j are less likely to leak out of the core than fast
- neutrons.
$ b. Reactors operating primarily on fast neutrons are
- inherently unstable and have a higher risk of going j prompt critical. .
I
- c. The fission cross section of the fuel is much higher for thermal neutrons than for fast neutrons.
- i i d. Doppler and moderator temperature coefficients become ;
l positise as neutron energy increases.
. 5.17 hhtch of the following statements best describes the (1.0) l conditten known as " condensate depression?" ,
! a. Can lead to condensate pump cavitation of condensate l depression is too great.
i b. Decreases as hotwell level rises. ,
- c. Reduces Rankins cycle efficiency,
- d. Increases as condensate temperature increases.
5.18 Which of the following statements most correctly completes ("i .0) ,
the f olicwing sentence? Departure from nucleate boiling !
I is the point where:
)
i a. Void fraction equals one l
- b. The heat transfer mechanism changes from nucleate boiling to single phase convection.
l c. Raatatise heat transfer becomes insignificant, i
4
- d. The heat transfer rate sustainable with nucleate boiling reaches its maximum, 4
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. - . . - , _ _ _ - - - - - _ _ . _ . _ _ - . - - - _ - - . - - - . - - - - _ _ . _ - - ~ -- . . . _ . - - . - .
S E 19 Which of the following is NOT correct concerning decay (1,0) i heat?
i 1 a. !s the heat produced by the energy released from the .
j radioactive decay of fission products.
! b. Can be determined by the reading on the SRM's when i the reactor is shutdown. '
l I c. Is approximately 6% of the total energy released i from fission. ' '
- d. Is still a significant contributor to the energy in the reactor core for approximately two hcurs af ter the reactor has been shutdown.
l 5.20 Figure 1.21 is a representation of how the resonance (1.0) i peaks of U-238 " flatten out" or Doppler broaden as fuel j terperature increases. Which of the following are the ccrrect labels for the X and Y axes?
- a. X is neutron flux; X is interaction rate
- b. X is neutron energy; Y is microscopic capture i cross section
- c. X is atom density of U-238; Y is neutron flux X is interaction rate; Y is neutron density.
- d. -
) 5.21 The ratio of Pu-239 and Pu-240 atoms to U-235 atoms (1.0) f changes over core life. Which of the pairs of parameters I listed belcw are most affected by this change? '
j a. Mcderator temperature coefficient and Doppler coefficient
! b. Doppler coefficient and beta i
j c. Beta and moderator coefficient i
l d. Moderator temperature coefficient and neutron j ger.eration time.
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- j. 5.22 Prior to startup (all rods in) the SRM countrate is (1.0)
- 20 CPS and K effective in 0.96. If'the control rods
- are pulled to give a delta K of 0.035, what count rate on the SRM's could be expe ted when the period ,
becomes infinite? .
- a. 40 i b. 160 j c. 60
- d. 120 i a5 5.J# Referring to-the attached curve (Fig. 1), which of the -'
folicwirg regions on the curve is associated with the t q(
i heat transfer rechanisa known as " transition boiling?" "_
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1 END OF CATEGORY 5 l
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- 6. Plant Systems Design, Control and Instrumentation 6.1 a. An operator is driving in a control rod. Explain (1.5) khY during motion of the drive mechanism there is
- no cooling water reaching the drive mechanism.
- b. Indicate whether the following statements are TRUE or FALSE. O in-rt0- p-)
There are two pairs of Alternate Rod Insertion #' )V 1.
(ARI) Valves located on the Scram dump valve air da g/.
supply and upstream of the backup scram valves. (0.75)
- 2. Depressing both of the pushbuttons on one side of the rod select matrix will energize either the "A" group or the "B" group ARI valves. (0.75) 6.2 hith regard to the ADS system:
- a. What is the initiation logic? (setpoints required) (1.0)
- b. What is the source of power to the relief valve solenoids? (0.5)
- c. Why ccn't the ADS valves operate at less than 150 psig? (0.5) 6.3 Fcr each of the HPCI system component failures listed belcw, STATE WHETHER OR NOT HPCI WILL AUTO INJECT into the reactor vessel; ~IF IT WILL NOT INJECT, WHY, ard, IF IT WILL INJECT, provide ONE POTENTIAL ADVEliSE EFFECT OR CONSEOUENCE of system operation with the fai'eo component. Assume NO OPERATOR ACTION and the component is in the failed condition at the time HPCI receives the auto initiating signal.
- a. The Gland Seal Exhauster Vacuum Pump fails to operate (1.0)
- b. The Turbire Auxiliary Lube Oil Pump fails to operate (1.0)
- c. The Minimum Flow Valve fails to auto open (stays shut)
- wher systems 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 output. (1.0) 6.4. Answer the following questions regarding the LPCI loop select logic.
- a. How does the logic-determine how many recirculation pumps are running? (Include setpoints where applicable). (1.5)
- b. How does the logic determine which recirculation loop is UNDAMAGED? (1.5)
- c. If the logic determines that neither loop is damaged, which loop will it select for LPCI injection? (0.5) 9
, - . _, ,- , , - . ,.v. , - . - . . . - , - ,
6.5 Describe the control action of the recirculation (2.5) ficw control system, if both recirculation pumps were operating at 50'. when an operator closes the discharge ,
valve on the "6" recirculation pump. (Assume the ,
discharge bypass valve is open).
[..t,- a. khat are three sources of makeup water to the shell side of the Isolation Condenser?
~
(1.5)
- b. What is the relationalship of the Isolation -
Condenser to the 480VAC Bus 28 and 29? (0.5)
- c. Hcw icr.g can the isolation condenser operate withcut shellside makeup if the minimum level requirerent was met? (0.5) 6.7 If a cceplete loss of instrument air where to occur with the plant operating at full power and no operator action, what would be the affect on the folicuing cceponents (Limit your answers to affects caused in relation to instrument air only).
- a. CRD Hydraulic flow control valve (0.5)
- b. CR0 hycraulic scram valves (0.5)
- c. CRD Hydraulic instrument volume (0.5)
- c. Main Feed pump minimum flow valve .(0.5)
- c. Main feed Regulating valve (0,5) 6.8 State whether each of the following statements about the AC electrical distribution system are True or False.
- a. The Unit 2 auxiliary transformer (TR-21) will lockout -
following a main generator trip. (0.5)
- b. Curing normal operation with the main generator synchrcnized to the grid, Buses (22 and 24) are normally powered from reserve auxiliary transformer TR-22. (0.5)
- c. There is no automatic transfer feature on RPS A. (0,5)
- c. Reactor Feed Pump 2A is powered by the 4.16 KV Bus 21. (0.5) 10
. .. . _ . _ . _ . _ . . _ _ _ _ _ ._ __ _ . ____= _-- ~ _ _ _ ._. _ _ .
t 6.9 The electrical portion of the Feedwater Level Control System is a GE/MAC. Discuss how the vessel level ;
instrument implements the logic of a GE/MAC controller 4 by answering the following questions: .
l a. A full range deflection of the vessel level input !
signal produces an output from 10-50 ma. Why i
is a 10 ma output maintained? (0,5)
} b. khat is the range of vessel level input that the
! 10-50 ma output represents? (0.5)
- c. If a 3 inch change in vessel level input causes a 2 r.a change in the controller output, what is the l
output signal in millamps with an input signal i at the normal reactor vessel water level. (Show any calculations you may use). (1.0) 6.10 The Stancby Liquid Control System has a minimum and (1.0) 1 maximum injection time designed into it. What is the.
bases for these times?
END OF CATEGOP,Y 6 l
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- 7. Procedures-hormal, Abnormal, Emergency, and Radiological Control 7.1 State whether each of the following statements concerning 10 CFR 2U is TRUE or FALSE. -
- a. Exposure of the whole body of any individual to 5 rems or more of radiation requires immediate notitication to the Director of the appropriate hRC Office. (0.5)
^
- b. The maximum permissible level of radiation in an unrestricted area is two millirems per hour. (0.5)
Y
- c. 0.1radduetoDeutronsisequivalenttoonerem. (0.5)
- c. A high radiation area is any accessible area where a major portion of the body could receive a dose in excess of 100 mrems in five consecutive days. (0.5) 7.2 List the following items in order that they would occur (3.5) during a norral ccid plant startup, (in accordance with CGP l-1 and give the approximate ( 10 psi) reactor pressure at which each is done.
- a. Line up HPCI for auto initiation
- b. Place first feed pump in service
- c. Place steam seals in service A. Start steam jet air ejectors .
-e. Shift to RWCU recirculation pump Ec', '
-f. Fully withdraw the' SRM's ' '
7.3 C0A 300-5 Inoperable or Failed Control Rod Drive States:
"If reactor power is less than or equal to 20% and
, a control rco is determined to be uncoupled, fully insert the rod to position 00."
Gise three indications of an uncoupled or stuck
- control rod. (1.5) 7.4 The Unit 2 is operating at 85% power and the Main Generator (4.0)
- field breakers trip open which results in a turbine trip and a reactor scram. What are the required irrediate action steps. Be specific, an action step may have more than one action item and if so it will have more value.
7.5 What are two (2) general instructions concerning rod- (2.0) rovements to minimize the risk of inadvertent short periods?
12
_ .= _ -. - ._ _ . _ - . - - -. . ..
I 7.6 Assume an ATUS event has occurred:
- a. Under what conditions can the NSO inject SBLC q withcut authorization from a supervisor.? (1.0) b.
What is the difference acticns required between for Unit 1 compared the imediate [.
_to Unit (0.5) c.
A h3 After the SBLC system has been [~ initialed, give four indicaticns that show the system is operating. -(1.0,)
7.7 State the reasons for each of the following precautions from 00F 200-1, Recirculation Startup:
- a. After a pump trip, shut the pump discharge valve il ar.c pump discharge bypass valve. (1.0) l
- b. Af ter approximately five minutes, the discharge valve / ,
are discharge bypass valve should be opened. Jna ad . (1.0)
- c. Establish flow from the seal purge system to l recirculation pump seals only after the pump i discharge and suction valves are verified open. (1.0) 7.8 hcw will the following parameters vary as a result of i a failed jet pump. (i.e., increase, decrease, remain
- ccostant; vary more or less than normal readings)?
j a. Core thermal power -
.(0.5)
- b. Recirculation pump flow for a given speed (0.5)_
- c. Individual jet pump flow indication (0.5)
- c. Electrical output (0.5).
- c. Core Flow (0.5) ;
7.9 In accorcance with the refueling procedure DFP-800-1, (1.5) how many SRM's are required and where are they to be located during fuel movement?
7.10 A. Before leaving the control room during a control room evacuation, what four actions are to be taken
, by the ftS0? .
(2.0)
B. Where will the shift foreman establish his control center?' (0.5)
END CATEGORY 7 1
J 13
. _ _ - ~-- _ __ _
- 8. Acc.inistrative Procedures, Conditions, and Limitations 8.1 The technical specifications for Dresden contain two (2.0) leak rate lients. List these two limits.and state J
.w1 u(M~~
the basis for each. "[d W e4 b *'
8.2 What is the function of the low pressure closure of the (3.0)
P.SIV's while in RUti and WHY is there a reactor scram interlecked with this closure?
8.3 According to 10 CFR 50, the one hour reporting requirement -
for each of the following conditions is applicable. (True or False).
- a. The plant is in a condition not covered by operating and crergency procedures. (0.5)
- b. The loss of the off-site notification system. (0,5)
- c. A valid automatic initiation of the Reactor Protection System. (0.5)
- d. A shutccwn was commenced because the plant was in violaticn of the Technical Specifications. (0.5) 8.4 List four ccnditions that require the post-accident (4.0) lineup of the high radiation sampling system and use of tne post-accident sampling procedure.
~
8.5 A. What is a Limiting Control Rod Pattern? (1.0)
B. If a Limiting Control Rod Pattern exists and a RBM Channel becomes inoperable, what action must be taken? . (1.0) 8.6 According to the "Out-of-Service and Personnel Protection" (2.0) procedure DAP 3-5, the request for equipment outage, there are cases which require additional approvals besides the Shift Supervisor. List two of these cases and who must approve. (Titles only needed) 8.7 There are several provisions that must be met in (4.0) orcer to operate with one recirculation loop out of service. What are four of these provisions?
Set points not required.
8.8 What are the personnel requirements for the fire brigade? (2.0) 8.9 List four of the irnediate operative actions for (4.0)
DGA-16 Coolant High Activity Fuel Element Failure.
END CATEGORY 8 14
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KE = 1/2 mv a = (Vf - Vo )/t A = IN A= Ace' PE = mgh Vf = V, + a t w = e/t x = an2/t1/2 = 0.693/t)fp NPSH = Pin - P 33t t l/2 eff = [(tl/2)(tb )] '
[(t1/2) + (t b)3 m o S AV AE = 931 aa I"I'o Q = UAah I = Ig e'"*
Pwr = Wfah I=I 10-x/ m g
TVL = 1.3/u P=P o l05 "" ) HVL = -0.693/u P = Poe /T .
SUR = 26.06/T SCR = S/(1 - K,ff)
CR x = S/(1 - K,ffx)
SUR = 260/t* + (e - o)T CR)(1 - Keffl) = CR2(1 keff2)
T = (t */p ) + [(e - c )hp ]
M = 1/(1 - geff) = CR)/CR g T = t/(o -e) M = (1 - K,7fo)/(1 - K,ffj)
T = (e - o)/(lo) SDM = (1 - K,ff)/Keff
" " IKeff-II/Eeff * #Keff/Neff t*= 10-5 seconds 1 = 0.1 seconds ~
o = [(t*/(i Keff))*b"eff/II +
- 3 I)d) = 1d 22 P = (IsV)/(3 x 1010) I)d) 2 =1d22 I = e ri 2 R/hr = (0.5 CE)/d (meters)
NPNi = Static head - hg - P3 ,g R/hr = 6 CE/d2(feet).
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm. I curie = 3.7 x 1010dps I gal. = 3.78 liters 1 kg = 2.21 lbm i f t = 7.48 gal. I hp = 2.54 x 10 Btu /hr Density = 62.4 lbg/ft 3 1 mw = 3.41 x 10 Btu /hr Density = 1 gm/cr 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) 1 Atm = 14 7 n c i = '20 C in Wa
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f
] CATEGORY 5 ANSWERS i 5.1 (a)
General fiuclear Principles
- 5.2 (c) j General fluclear Principles y 5.3 (b)
] General .tiuclear Principles >
5.4 (t)' Cl General Thermodynamics and Stress Analysis 5.5 (d)
General fauclear Principles 1
5.6 (c)
General fiuclear Principles i
, 5.7 (a)
General Thenral Principles l 5.8 (a) i General fiuclear Principles
- +
5.9 (b)
General fiuclear Principles
~
l 5.10 (a) Reactor critical on prompt neutrons alone. (0.5)
$ (b) Tire in seconds required for power to change (0,5)
'e'.
by a factor of (c) The multiplication of neutrons by the fuel in a (0.5) subcritical reactor.
1 Standard tiuclear Principles 5.11 (c), '
4, Standard fiuclear Principles t ,
5.12 (a)
Standard fluclear Principles 5.13 (d)
Standard fluclear Principles l
I i
1 15 1
l 4 s 5.14 (a) Transition Boiling may occur which could result in (1.0) clad failure.
(b) To make the MCPR limit more conservative to acccunt (1.0) for the possibility of a sudden flow increase Lnd a corresponding power increase. i (c) Decreases. (,
j Star. card Thermo-Hydraulic Principles ;
5.15 (c) -
Standard f uclear Principles 5.'16 (c)
Standard f.uclear Principles ^
! /7
,,' 5. X (c) 4 , Standard Thermodynamic and Fluid Flow Principles 1 5 5.J9 (d)
. Standard Therral Hydraulic Principles and Applications s1
< 5.20 (b)
Standard fiuclear Principles ,
j e
- c 5.11 (b)
Standard Doppler coefficient principles 4 21 5.22 (b), .
, ,. Standard fiuclear Principles .
'5 (b)
StandardfuclearPrinciples4) j ._. -b -6 5.2I4b)9 l Standard Thermo-Hydraulic Principles -
-4 M 4
e i
i 16 i
e CATEGORY 6 ANSWERS
! 6.1 (a) Ccoling water normally foll s from the cooling water header through a check valve to the drive f insert line at about 20 psi above reactor pressure. (0.5) hhen a drive is in motion the pressure in the insert lir.e is at 280 psi greater than reactor pressure. (0.5)
The dp between drive water and cooling water will keep ccoling water check valve closed and prevent cooling water flow. (0.5)
(b) (1) False i (b) (2) True Ref: CRD Hydraulics LP pg.6, 7, 27.
6.2 (a) (1) Hi drywell greater than 2 psig 5 (2) Low level less than 59 in. /' g (3) 120 Sec timer expired (4) Low pressure. ECCS pump running ,,
(b) 125 VDC c.
(c) Rx pressure inadequate to overcome spring forces. c. '
! 6.3 (a) Will inject . (0.25) 1 Turbine seal leakage resulting in potential airborne activity in the HPCI room. (0.75)
(b) Will not inject (0.25) t Turbine stop and control valves will not open (0.75)
(c) Will inject (0.25)
Pump overheating and seal damage may result during icw or no flow conditions (0.75)
(d) Will not inject (0.25) 4 Maximum signal from the flow element will cause the controller to keep turbine speed
. at minimum (0.75) t Ref: HPCI Lesson Plan.
I I
17
e 4
6.4 (a) By monitoring the DP (0.5)
- Across each recirculation pump (0.5) fcr a 2 psid or greater dp, indicating the purp is running (0.5) j (b) By comparing the pressure in the riser pipes -
on ene recirculation loop with the pressure in /-
the riser pipes of the other loop. The undamaged ;
lecp will have a higher pressure than the s damaged loop. ,
(c) Loop B. .:
Ref: LPCI Lesson Plan. Rev. 5, pg. 15.
6.5 (1) The limiter would try to reduce pump "B" speed to 28%.
(2) The mismatched circuit would stop the scoop tube insertion
- at 40
- , (10'; mismatch). -
(3) The "B" pump would be running at 40% with the discharge c) . 3 valve shut (pump damage could result). [
(4) The mismatch circuity trips the low speed pump when the discharge valve goes shut. y Ref: Dresden Recirculation Flow Control, Lessen Plan, pg.13.
6.6 (a) Clean demineralized H2O system, Condensate Transfer system, /3 service water / fire protection. -
(b) Pcwers rotor operated inboard isolation valves, o0 < '
(c) F i ve +inutes. a.c & . p 3< e- C-~F l" - q g-Qo-j 6.7 (a) Valve wculd close (b) Scram valves would open -
(c) Instrurent volume vent and drain valves would close 'A 7
! (d) Minimun flow valve would open
?
(e) Regulating valves would stay as is j Ret: Instrument air, Lesson Plan pg. 14-15 6.8 (a) False 1'
(b) True '
(c) True (d) True Ref: Lesson Plan, Bk 2, April 13, pa0es 7-10
, 6.9 (a) Maintaining a minimum output allows the detection . .'
of an electrical fault in the device.
(b)'0-60inchvessellevel s
, (c) fiormal level 30 inches midway between 0-60 inches, s 7
i therefore this would be half the difference in the
- range of the output or 20 ma (0.33)
- 10 ma 20 ma = 30 ma (0.33) l Ref: Feedwater Control, Rev. 8, Page 6-7 18
6.10 The minirun tire minimizes imperfect mixing and reactivity digging (0.5)
The maximum tire is fast enough to overcome any positive reactivity effects due to cooldown and xenon decay from peak. (0.5) 19
5 CATEGORY 7 ANSWERS 7.1 (a) False 4 (b) True -
(c) True -cO&
(d) False k Ret: 10 CFR 20 q.h % @
7.2 C,F,A,0,B.E 25, 50, 90, 200, 300, 920 (~V }V - LE) AD#of 3co+
OF ,Q 5) 60,,7#a g
Acceptable answer for C is also, when sufficient steam available
, Ref: DGP-1-1 pg 5-12 7.3 (1) tio nuclearalarm instrumentation response during rod withd -
(E) Rcd overtravel (3) t;c position indication past position 48 on a withdrawal sicnal after drive ficw decreases to stall flow.
(4) tio' position indication response to control rod movement j Ref: DGP 300-5, pg. 3-4
)
7.4 (1) Place rade switch in shutdown (or refuel) (0.4)
(2) Check all rods fully inserted (0.4)
(3) Verify APRM's decreassing or downscale (0.4)-
(4) Maintain feed water in auto unless controller failure
- occurs. Control level 20 and 140 inches by observing
- r. ore than cne available indication
~
(0.8)
(5) Verify turbine and generator have tripped and speed i is decreasing and OCB;s are open (0.8)
(6) Verify bypass valves are controlling Reactor Pressure (0.4)
(7) Start Er:ergency bearing oil pump (0.4)
(E) Verify auxiliary power transfers to transformer 22 (0.4) 7.5 (1) All reds in BPLS groups 3 and 4 must be notch withdrawn between positions 04 and 12 (1.0)
(2) f;otch cverride shall not be used after a black and white pattern is reached until at least one Bypass valve is partially open or unit is on'da line (1.0) .
Ref: DGP-3-4, Rev. 5, pgs. 2-6 l
i 20
e 7.6 (a) If either two or more adjacent rods are not inserted
'['
past 66 or 30 or more rods are not inserted past 06 J; '
ana if reictor water level cannot be maintained or '
suppression pool water temperature cannot be matiitained below 110 F. '
(b) Unit 2 rust trip Hydrogen addition
(c) (1) Ccrtinuity circuit pilot light not lit '
(2) Flow pilot light lit
(, ! < {'
(3) Clean-up system isolated ,
(4) Fump discharge pressure increasing ) ,
(5) Decreasing level in SBLC tank /
Ref: CGA 18, p.7; DOP 1100-2, pgs. 1-3.
7.7 (a) Present reverse rotation of the pump (b) Paintain temperature in idle loop (c) Prevent overpressurization of the recirculation loop Ref: LGP 202-1, pg. 2-3; DOP 202-4 pg. 2 7.8 (a) Decrease (b) Increase (c) Vary less (c) Cecrease (e) Increase Ref: Dresden C0A 201-1, symptoms pg. 1 7.9 2 SEM's one in quadrant where fuel or rods are being rovec ar.d cne in adjacent quadrant.
Tech Spec. 3.10 7.10 (a) Manually scram reactor, leave mode switch in run, trip CR0 pumps, trip turbine (b) EB second floor.
21
r i
CATEGORY 8 ANSWERS 8.1 (a) 5 gpm unioentified (b) 25 gptr. total -
(a) crack propagation (b) surrp pump capacity Ref: T.S. 3/4.6 6.2 (a) The icw pressure closure protects against a rapid i cool dcwn due to a failure of a pressure regulator (1.6)
(b) The scran is in anticipation of the pressure and flux trarsient which would occur following the MSIV closure (1.5) 8.3 (a) True (b) True (c) False (o) True Ref: 10 CFR 50.72 i 8.4 (1) Grcup 1 isolation on 3x normal MSL Rad level (2) Group II isolation on Drywell High Rod level (3) Reactor water level decreased below or is below the l TAF as a result of an operational transient or accident condition.
(4) Reactor water level cannot be detennined or was inocterminate at some time during an event.
(5) Failure of the RPS to initiate or complete a scram cnce a scran signal has been initiated or once a LSSS has been exceeded.
(6) Other conditions as deemed appropriate by the Rad / Chem Director or Station Director.
l Ref: EPIP 300-Ti 8.5 (a) A pattern which results in the core being on a thermal hydraulic limit (operating on a limiting value for APLHGR, LHGRcrMCFR).
(b) Place inoperable RBM channel in tripped condition within one hour.
Pef: DOA 700-3; T.S. 3.2.C 8.6 (1) Outage of ECCS and primary components. Ops Eng.
(2) Outage of buses, transformer, DG's and bus til breakers (Until support Ops Eng.)
(3) Reliability related Equipment (0ps Eng)
(4) Outages of Fire Protection systems or equipment (Station Fire Marshal)
Ref: Dresden DAP 3-5, pg. 1-2 22
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8.7. (1) Steacy state thermal limit less than 50% i
- (2) MPCR linit increased by 0.03 (3) MPCR LCO is increased by 0.03 l (4) MAPLHGR limit is reduced to 70% of current value (5) APPM Scram anc Rod Block setpoints and RBM setpoints j reauced by 3.5%. ;
(6) Suction valve in idle loop closed (7) APRM flux noise is less than 5% peak to peak over i a half hcur average i
(8) Core plate delta p noise is less than lpsi peak to peak -
Ref: T. 5. Amend. 75, pg. 6 i 8.8. (a) At least five persons that do not include shift crew i recessary for safe shutdown of plant and any personnel required for essential functions during the fire emergency. .
8.9 (1) Stop any pcwer changes in progress i (2) Ccepare readings. Main Stream Line radiation versus off gas, i ar.d flux tilt monitors and recorders.
j (3) If tain steam or off-gas bi rad alarms annunciate, adjust J pcwer down to keep activity below trip set point. (Reduce l recirculation flow and/or insert rods) 1 (4) If off-gas Hi-Hi rad monitor annunciates, rapidly reduce
] pcwer below the monitors trip point and reset off-gas i monitor, j (5) If main steal line Hi-Hi rad alarms annunciate follow scran procedure and verify the automatic actions procedures, l (hote 4: Reset the monitors as soon as activity is below the j set points. Off gas line will isolate if monitors are not
, reset within 15 minutes.
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