ML20137K610
| ML20137K610 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 01/21/1986 |
| From: | Dimmock L, Mcmillen J, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20137K585 | List: |
| References | |
| 50-331-OL-85-02, 50-331-OL-85-2, NUDOCS 8601240144 | |
| Download: ML20137K610 (30) | |
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U.S. NUCLEAR REGULATORY COMISSION REGION III Report No. 50-331/0L-85-02 Docket No. 50-331 License No. DPR-49 Licensee: Iowa Electric Light and Power Company Facility Name: Duane Arnold Energy Center Examination Administered At: Duane Arnold Energy Center Examination Conducted: November 5-7, 1985 bl Examiners: . Dimmock /!M//d Datd J. Mu ro /!I/ /d Jia Approved By:
Q M.)V kph I. McMillen, Chief / d/ I Operator Licensing Section hate Examination Summary Examination administered on November 5-7, 1985 (Report No. 50-331/0L-85-02)
Results: Eight SR0 candidates took the exam with-six satisfactory passing both the written and oral exam. One person failed the oral exam and an other failed the written exam.
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REPORT DETAILS
- 1. Examiners L. Dimmonk, Region III J. Munro, Region II
- 2. Exit Meeting Following the oral exams an. exit meeting was held at the training center with selected members of.the training and plant staff. At that time they were informed of the preliminary results of the plant orals i.e. , one candidate was not a clear pass. The training people also furnished us l with the exam comments at that time. The resolution of those comments is attached.
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Duane Arnold Energy Center November 5,1985 SR0 Examination Coments, Resolutions Coment: 5.1 "The primary purpose of a water rod is to increase neutron moderation near the center of a fuel assembly."
This is a quote from page 11 of System Description A-4(Fuel).
It goes on to say why we increase neutron moderation in the center of the bundle. Therefore, we feel that the first sentence of your answer should be all'which is needed for full credit.
Resolution: Comment not accepted. It is felt that flattening the flux should be part of the required answer.
Comment: 5.4 b. Reactor Pressure < 785 psig or Core flow < 10% of Rated; should be < in both cases.
Resolution: Coment accepted. Answer key modified.
Coment: 5.6 b. Plate neutrons should be Photo neutrons.
Resolution: Comment accepted. Answer key modified.
Coment: 5.11 b. The question only asked for a yes or no answer. It should have asked to explain your answer to elicit the desired response.
Resolution: Candidates are cautioned prior to exam that the value of the question is an indication of the expected depth of response. A 1.5 point response would require more than a yes or no. All candidates did answer with more than a yes or no. This question in the future should contain an
" explain."
Coment: 6.2 The purpose of the axial enrichment is to help control
, axial power. Without the use of shallow rods should not' be required for full credit.
Resolution: Coment accepted. Answer key modified.
Comment: 6.3 a. Narrow range GEMAC pressure compensated. This answer does not answer the question which asks for which level instruments are temperature compensated and how it is accomplished. Table 2 of System Description A-5 (Reactor Vessel Instrumentation) clearly shows the only compensation which is temperature is for the WR Yarways. If the questior, was worded: "Which level instruments are compensated for density and how is this accomplished?"
then your answer would be correct.
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Resolution: Comment accepted. Question was modified to ask for only one response and point value was dropped to 1.0.
Comment: 6.3 b. Drywell High temperature is also an abnormal condition which will render the Wide Range level instruments unreliable. See caution 5 of E0P-C and page 19 of System Description A-5 (Rx Vessel Instrumentation).
Resolution: Comment accepted. Answer key modified.
Comment: 6.4 It doesn't appear that the 20% number is .ieeded for full credit since the question is answered by using 30% as per the question.
Resolution: The 20% is not required but is included for information.
Comment: 6.5 Offgas stack Hi-Hi rad (1.5-E4 cps) is also a group 3 signal (recent modification). See pages 10 and 11 of SBGT System Description, (E-11).
Resolution: Comment accepted. Answer key modified.
Comment: 6.6 a. We request that you also accept as another answer
" unseating the ball check valve" in place of " cooling water pressure". Either would be a source of water to move the control rod.
Other answers such as creating a hazardous environment in RCIC room could also occur.
Resolution: Comment accepted. Answer key modified.
Comment: 6.7 a. Point values add up to 2.0 on the answer key, yet the question is worth 1.5. We request you weigh each part of "a." equally at .75 pts. each.
Resolution: Comment accepted. Answer key modified.
Comment: 6.8 b. Valve numbers are included in the system description for completeness, but are not required for memorization. We require that the operators know at what setpoints the high valve cycle between but nor the valve numbers. We request that valve numbers not be required for full credit.
Resolution: Comment accepted. Answer key modified.
Comment: 6.9 b. Add control of RCIC at IC208 which is one of the remote shutdown panels, as another means of controlling RCIC.
See page 42 of RCIC system description and page 18 of RCIC Instructor Guide. New system description for the Alternate Shutdown Control System (I-18) is not written at this time.
Resolution: Comment accepted. Answer key modified.
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Comment: 6.10 a. "The time requirement for inserting the boron solution was selected to override the rate of reactivity insertion caused by cooldown of the reactor following the xenon poison peak."
This is a quote from T.S. basis 3.4.1 page 3.4-4 Your answer includes " required shutdown margin" which is the basis for the boron concentration, not the time.
Resolution: Comment accepted. Answer key modified.
Comment: 7.7 We request that control rods which do not meet their scram times be an acceptable answer. We do not feel the numbers of 90% and 7.0 seconds should be required for full credit Should add as another acceptable answer "A control rod on which maintenance is being performed." See T.S. 4.9.A-2.
Resolution: If the question is answered as requested in comment above, this would be considered as combining answers 4, 5, and 6 in the answer key and this would be acceptable. Answer 7 was added.
Comment: 7.8 a. The IP0I states that pressure set is used then it can no longer maintain the CD rate, use the bypass jack. The question was confusing because it asked for the two methods to be used to cooldown from 920 - 50 psig. The two methods the question was attempting to elicit were in fact utilized together to achieve the goal. EHC pressure set will only go down to 150 psig (its lower end). See page 19 of 29 IPOI 4 and Figure 13 of EHC System Description.
Resolution: Comment accepted. Answer key modified.
Comment- 7.8 b. Depending on decay heat the following might be needed to augment the steam liae drains:
HPCI RCIC RHR Steam Condensing ADS See page 23 of 29 of IPOI 4 Resolution: Comment accepted. Answer key modified.
Comment: 7.9 c. Cooldown rate may also be controlled by running one or two RHRSW pumps or one or two RHR pumps. See OI49 pages 35 and 36.
Resolution: Comment accepted. Answer key modified. An additional answer was added to part b. as per the procedures.
Comment: 8.1 E0P-6 should be considered an appropriate answer since the SR0s are directed to evacuate the control room during certain conditions. See E0P-6 page 3.
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Resolution: Comment accepted. Answer key modified.
Comment: 8.4 b. Request you also accept-increasing RWCU flow or dump flow since that is. the manner in which bottom head drain flow is increased.
Resolution: Comment accepted. Answer key modified.
Comment: 8.11 b. IAW IPOI 4 should not be part of the required answer.
A discrepancy report (DR) must also be initiated and an immediate notification made. See ACPs 1402.2 and 1402.3.
! Resolution: Comment accepted. Answer key modified.
- Comment
- 8.12 a. There are more restrictions than in the Technical
- Specifications.
(1) Not allowed in changing the Safety Intent or function of controls or equipment as described in T.S. or UFSAR except'under emergency conditions as directed by the OSS.
- (2) Temporary revisions to STPs shall not allow two or more l , identical systems or components to be tested simultaneously.
(3) Temporary revisions are not allowed for 1400 level DAEC procedures. See ACP 1406.3 pages 2, 3 and 6.
Resolution: Comment accepted. Answer key modified.
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. ffa3 T-e r U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: Duane Arnold REACTOR TYPE BWR DATE ADMINISTERED: November 5, 1985 EXAMINER: L. Dimnock 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 2f'35G48P 25 5. Theory of Nuclear Power Plant Operations, Fluids and Thermodynamics 38:2Y 25 6. Plant Design, Control and Instrumentation 25 25 7. Procedures - Normal, Abnormal, Emergency and Radiological Control 25 25 8. Administrative Procedures, Conditions and Limitations 99 itTr 100 TOTALS Final Grade %
All work done on this exam is my own, I have neither given nor received aid, i
Applicant's Signature I
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Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics 5.1 What is the ,,rimary purpose of the water rod in a fuel bundle? (1.5) 5.2 Explain the difference between void fraction and quality. (1.0) 5.3 How would you expect critical power to change for each of the following conditions? Explain why. (Assume all other factors remain the same).
- a. An increase in core flow (1.0)
- b. Pressure decreases from 1000 psia to 800 psia. (1.0)
- c. Local peaking factor increases (1.0)
- d. The axial power distribution is changed from a pcak low in the core on a bundle to high in the core. (1.0) 5.4 What are the two(2) MCPR Safety Limits in the T.S.? (2.5) 5.5 APLHGR for any node can be found by multiplying Average LHGR by the RPF and APF for that node. Explain what RPF and APF are. (2.0) 5.6 What are the three types of intrinsic source reactions that can supply neutrons to your reactor? Include an example of each. (1.5) 5.7 Use the attached figure of " Effective Decay Constant Vs Reactivity." Explain why l changes. (2.0) 5.8 In the main condenser, circulating water flow rate is approx-mately 20 times that of the steam flow rate. Why are these flow ates different? (Primary heat transfer rate equals circulating water heat transfer rate). (Consider thermodynamic principles in your answer). (1.0) 5.9 HOW and WHY does the MAGNITUDE (reactivity added per change in degree F) of the FUEL TEMPERATURE COEFFICIENT (DOPPLER) change, given the following changes in core conditions:
- a. Core age (BOL to E0L) (1.0)
- b. A significant increase in fuel temperature. (1.0).
- c. A significant increase in core void function. (1.0)
SECTION 5 CONTINUED ON NEXT PAGE
. tuin 1 rag. 20.2 EFFECTIVE DECAY CONSTANT
.VS REACTIVITY l 0.6' 0.5
/
0.4 -
7 oa /
7 0.2
+P 0.1
_._ 0.08 /
k-1 O.06 0.05
\
Sec o,04 h 0.03 0.02 N
x l 0.01 0 .002 0.004 0.006 0.008 0.01 0 ,
Reactivity l 14
5.10 a. After making a rod notch withdrawal with the reactor critical, you notice a 100 second period. How much reactivity was added by the rod notch? (assume BOL) (1.0)
- b. Af ter a reactor scram from power the shortest stable period possible is -80 seconds. Explain this statement. (1.0)
- c. Is the initial period immediately following the scram shorter than -80 seconds? Explain your answer. (1.0) 5.11 APLHGR limits have been set to assure that peak cladding temperature of 2200 F are not exceeded following a postulated LOCA.
- a. FollowingaLOCA,whichrods(center, edge, corner) would be more likely to exceed this 2200 F limit.
Explain your answer. (2.0)
- b. Are these the same rods with the highest local peaking factors during normal operation? (1.5)
END OF SECTION 4
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Plant Systems Design, Control, and Instrumentation 6.1 On a slow failure of the No. 1 recirculation pump seal what 2 indications are available to the control room operator to alert him of the situation? (2.0) 6.2 In the Lead Test Assemblies that were loaded in Reload 7, one of the design features being tested was for extended exposure capability. The enrichment varies axially in the bundle.
Which portion (upper, central, or lower) of the bundle has the lower enrichment and what is the purpose of this test? (1.5) 6.3 a. Which level instruments b temperature compensated and /. O he. is this compensation accomplished? .( a f f
- b. The wide range level instrument is to be considered unreliable during certain abnormal conditions.
What are these conditions? (1.0) 6.4 Per the T.S., the Rod Worth Minimizer is required to be operable when below 30% rated power. Why is the Rod Worth Minimizer not needed when > 30% power? (1.0) 6.5 What are four signals (other than Iranual activation of the control switch) that will trip the Reactor Building supply and exhaust fans? (2.0) 6.6 a. What potential operational problem (s) could occur with a control rod if its scram outlet developed a seat leak? (1.5)
- b. How could control rod drive high temperatures following a scram be an indication of a scram discharge volume leak? (1.0) 6.7 a. From 50% rod density to the 30% power bypass, what two (2) restrictions are placed on rod movement by the Rod Sequence Control System (RSCS)? (1.5)
- b. From 50% rod density to 30% power, how does the RSCS determine rod position? ,
(1.0) 6.8 a. What conditions will activate the Safety Relief Valve Low-Low Set (LLS) logic and how are they sensed? (2.0)
- b. Describe the action produced by LLS logic actuation.
(Ee Specific and Include Setpoints) (2.0)
- c. Ho is LLS logic reset? (0.5)
SECTION 6 CONTINUED ON NEXT PAGE 3
6.9 a. After an auto initiation of RCIC, what is the process variable that controls pump output and where is it sensed? (1.0)
- b. When in manual control of RCIC, what two (2) other signals can control RCIC pump output? Include the switch positions necessary for these other signals to be in control. (2.0) 6.10 a. Why is it necessary for the Standby Liquid Control System to be capable of injecting, the contents of the SLC tank in a MAXIMUM time of 96 minutes? (1.0)
- b. What are four (4) other uses of the nozzle used for vessel penetration by the SLC sparger? (2.0)
END OF SECTION I
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l PROCEDURES-NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 7.1 During fuel movement operations, if fuel or core component moves are made within the reactor vessel or reactor vessel cavity, what people must be in the crew? (Note, there are two cases). (2.5) 7.2 During refueling outages, a functianal and subtritical check of every control rod shall be made. What does a functional and subcritical check consist of? (2.0) 7.3 What are the personnel restrictions on the refuel floor when a control rod or rods are moved and the head is off? (2.0) 7.4 The IPOI for startup (IP01-2) includes a list of systems required by Technical Specifications to be operable prior to commencing reactor startup. There is another list of 5 additional systems or modes of systems that must be operable prior to exceeding 212 F.
What are 4 of these systems or modes of systems that must be operable before exceeding 212 F? (2.0) 7.5 The Well Water /G5W system is one of the six Alternate Injection Subsystems per the E0P's. What are the other 5 Alternate Injection Subsystems? (2.5) 7.6 If during a startup the proper IRM/SRM overlap is not observed, what are the options available to the OSS? (3.0) 7.7 Control rods with inoperable accumulators are considered inoperable. What are three other conditions that will cause a control rod to be considered ir.c;;rable? (3.0) 7.8 a. During Plant Cooldown with MSIV's open, what are the two methods used to reduce reactor pressure from 920 lbs. down to 50 lbs. reactor pressure? (1.0)
- b. What is the normal method used if the MSIV's are closed?
(Assume the condenser is available and you do not want to open the MSIV's). (0.5) 7.9 a. Prior to placing RHR in Shutdown Cooling Mode, the RHR system is heated to no less than degrees F. How is this accomplished? (1.5)
- b. Ho. is the minimum flow valve prevented from opening when initiating Shutdown Cooling? (1.0)
- c. Once Shutdown Cooling is established, how is the cooldown rate controlled? (1.0)
SECTION 7 CONTINUED ON NEXT PAGE 5
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7.10 The Emergency Plan Implementing Procedure (EPIP 2.5) " Control Room Emergency Response Operation" lists the responsibilities of the Operations Shift Supervisor. What are three (3) of these? (3.0)
END OF SECTION ,
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 8.1 What are the two conditions when a senior licensed reactor operator is not required to be in the control room when there is fuel in the reactor? (1.0) 8.2 a. What is the difference in the use of the Hold card and the Warning Tag? (1.0)
- b. T/F
- 1. If two or more groups have clearances on the same piece of equipment, the same clearance number will be used for the two clearances. (0.5)
- 2. When a clearance is changed, the 055 is responsible for informing all the people working on the system under that clearance that it was changed. (0.5)
- 3. When the person whose name is on a Hold card is off-site and cannot be reached, the OSS can grant authorization to release Hold-0ffs if he deems necessary to do so for safe operation or shutdown of the plant after making sure there is no danger to personnel. (0.5) 8.3 a. Temporary cross connections between the demineralized water system and any other plant system must have the prior approval of one of two people. Who are these two people by title? (1.0)
- b. The temporary cross connect must have at least one specific device in it. What device is required? (0.5) 8.4 Following are two T.S. LC0's:
- 1. The pump in an idle recirculation loop shall not be started unless the temperatures of the coolant within the idle and operating recirculation loops are within 50*F of each other.
- 2. The reactor recirculation pumps shall not be started unless the coolant temperatures between the dome and the bottom head drain are within 145 F.
- a. What are the T.S. bases for 1 and 2 above? (2.0)
- b. If the ^T in 2 above is greater than 145 F, what are the 2 methods suggested for correcting this condition per IPOI-2? (2.0) 7
)
8.5 What 5 RPS trip functions must always remain operable when the reactor is subcritical and below 212*F? (2.5) i 8.6 During cold weather, what must be done to ensure that the following tanks don't freeze?
- a. Demineralized water Storage Tank (1.0)
- b. Condensate Storage Tanks (1.0) 8.7 The concurrence of what 2 people is needed to make a deviation from the Control Rod Withdrawal Sequence? (1.0) 8.8 The T.S's. use the term " Limiting Control Rod Patterns." When does a " Limiting Control Rod Pattern" exist? (2 conditions required). (2.0) 8.9 The T.S's. require that the APRM and LPRM neutron flux noise levels be checked as compared to baseline levels under certain conditions of core power and core flow. What LPRM's are required by T.S. to be monitored? (2.0) 8.10 Following a surveillance, the No. 3 TIP ball valve did not auto close. The TIP machine requires extra jogging to close this valve due to a malfunctioning " sticking closed" inshield limit switch. Is primary containment integrity satisfactory? If so why? If not, why not and how can it be made satisfactory? (2.0) 8.11 The reactor is operating at high power. A turbine trip occurs causing a scram on HIGH FLUX:
- a. Has a violation of the T.S. occurred? Why or Why not? (1.5) b What actions are you as the Operation Shift Supervisor required to take? (1.5) 8.12 You are the Operations Shift Supervisor on a weekend. Your crew is performing a procedure when it is determined that the procedure in incorrect. What are your responsibilities in regards to a temporary procedure change? (1.5)
END OF SECTION 8
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. EQUATION SHEET f = ma y = s/t Cycle efficiency = (Network out)/(Energy in) w = mg s = V,t + 1/2 at 2 2
E = mc KE,= 1/2 mv a = (Vf - Vo )/t A = AN A = Age' PE,1 mgh V'=V f o
+ at w = e/t x = an2/t1/2 = 0.693/t1/2 1
NPSH = Pin - P sat 1/2 eff = [(tl/2)(tb))
, [(t1/2) + (to )]
me oAV AE = 931 ara I=Ieg O Q = mCpat Q
- UAah I=le# o Pwr = Wfah I = 1, 10- */*
TVL = 1.3/u 3
P = Po l0 "#I*} HVL = -0.693/u t
P = Po e /T SUR = 26.06/T SCR = S/(1 - K,ff)
CR, = S/(1 - K,ffx)
SUR = 26o/t* + (8 - o)T CRj (1 - K,ff j) = CR2(1 kdf2)
T = (t*/o ) + [(a - p )/Ap ] M = 1/(1 - K,ff) = CR /CRj o M = (1 - K,ffo)/(1 - K,ffj)
T = 1/(o - s)
T = (a - 9 )/(10 ) SDM = (1 - K ,ff)/Keff o = (Kef f-l)/Keff = aKeff/K eff t* = 10-5 seconds x = 0.1 seconds o = [(t*/(T K,ffD + [s,f /f (1 + b]
Idjj=1d2p 2
P = (reV)/(3 x 1010) I)dj =Idg2 2
I = oN R/hr = (0.5 CE)/d (meters)
NP4H=Statichead-hg - P3,g R/hr = 6 CE/d2(feet) ,
Water Parameters Miscellaneous Converstons 1 gal. = 8.345 lbm. I curie = 3.7 x 1010dps 1 ga . = 3.78 liters I kg = 2.21 lbm I ft = 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/cm' lin = 2.54 cm Heat of vaporization = 970 Stu/lbm *F = 9/5'C + 32 Heat of fusion = 144 Stu/lbm 'C = 5/9 (*F-32)
STEAM TABl.E PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE) volume ft8sh Enthalpy. StuAb Entropy 8tunb a F
- ' Steam' water steem water Evec Steam w ater Evec Eveo 7
't '% 's hg h m h, sq s,r s, 32 0.08859 0.01602 3305 3305 -0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 35 0.09991 0.01602 2948 2948 3.00 1073.8 1076.8 0.0061 2.1706 2.1767 35 40 0.12163 0.01602 2446 2426 8.03 1071.0 1079.0 0.0162 2.1432 2.1594 40 45 0.14744 0.01602 2037.7 2037.8 13 04 1068.1 1081.2 0.0262 2.1164 2.1426 45 50 0 17796 0.01602 1704.8 1704 8 18 05 1065.3 1083.4 0.0361 2.0901 2.1262 50 60 0.2561 0.01603 1207.6 1207.6 28.06 1059.7 1087.7 0.0555 2.0391 2.0946 60 70 0.3629 0.01605 868.3 868.4 38.05 1054 0 1092.1 0.0745 1.9900 2.0645 70 80 0.5068 0.01607 633.3 633.3 48.04 1048.4 1096 4 0.0932 1.9426 2.0359 80 90 0.6981 0.01610 468.1 468.1 58 02 1042.7 1100.8 0.1115 1.8970 2.0086 90 100 0 9492 0.01613 350 4 350.4 68.00 1037.1 1105.1- 0.1295 1.8530 1.9825 100 110 1.2750 0.01617 265 4 255.4 77.98 1031.4 1109.3 0.1472 1.8105 1.9577 110 120 1.6927 0.01620 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 130 2.2230 0.01625 15732 157.33 97.96 1019.8 1117.8 0.1817 1.7295 1.9112 130 140 2.8892 0.01629 122.98 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 150 3.718 0 01634 97.05 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686 150 160 4.741 0.01640 77.27 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1.8487 160 170 5.993 0.01645 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170 180 7.511 0.01651 50.21 50.22 148 00 990.2 1138.2 0.2631 1.5480 1.8111 180 190 9.340 0.01657 40.94 40.96 158.04 984.1 1142.1 0.2787 1.5148 1.7934 190 200 11.526 0.01664 33.62 33.64 168 09 977.9 1146.0 0.2940 1.4824 1.7764 200 210 14.123 0.01671 27.80 27.82 '178.15 971.6 1149.7 0.3091 1.4509 1.7600 210
, 212 14 696 0.01672 26.78 26.80 100.17 970.3 1150.5 0.3121 1.4447 1.7568 212 120 17.186 0.01678 23.13 23.15 188.23 965.2 1153.4 03241 1.4201 1.7442 220 230 20.779 0.01685 19.364 19.381 198.33 958.7 1157.1 03388 1.3902 1.7290 230
- 240 24 968 0.01693 16304 16321 208 45 952.1 1160.6 0.3533 1.3609 1.7142 240 250 29.825 0.01701 13.802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 260 35 427 0.01709 11.745 11.762 228.76 938.6 1167.4 0.3819 1.3043 1.6862 260 270 41.856 0 01718 10.042 10 060 238.95 931.7 117t. 6 0.3960 1.2769 1.6729 270 280 49 200 0.01726 8.627 8.644 249.17 924 6 11731 0.4098 1.2501 1.6599 280 290 57 550 0.01736 7.443 7.460 259.4 917.4 1176.8 0 4236 1.2238 1.6473 290 300 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 310 77.67 0.01755 5.609 5.626 280 0 902.5 1182.5 0.4506 1.1726 1.6232 310 320 89.64 0.01766 4.896 4.914 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 340 117.99 0.01787 3.770 3.788 311.3 878.8 1190.1 0 4902 1.0990 1.5892 340 360 153.01 0 01811 2.939 2.957 332.3 862.1 1194 4 0.5161 1.0517 1.5678 360 380 195.73 0.01836 2.317 2.335 353.6 844.5 1198 0 0.5416 1.0057 1.5473 380 400 247.26 0.01864 1.8444 1.8630 375.1 325.9 1201.0 0.5667 0.9607 1.5274 400 420 30838 0.01894 1.4808 1.4997 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 440 381.54 0.01926 1.1976 1.2169 419.0 785 4 1204 4 0.6161 0.8729 1.4890 440 460 466 9 0.0196 0.9746 0 9942 441.5 763.2 1204 8 0.6405 0.8299 1.4704 460 480 566 2 0.0200 0.7972 0.8172 464.5 739.6 1204.1 0 6648 0.7871 1.4518 480 500 680 9 0 0204 0.6545 06749 487.9 714.3 1202.2 0 6890 0.7443 1.4333 500
$20 -. 812.5 0 0209 05396 0.5596 512.0 607.0 1199 0 0.7133 0.7013 1.4146 520 540 $ 962 8 0.0215 04437 0.4651 536 8 657.5 1194.3 0 7378 0.6577 1.3954 540 560 1133 4 0 0221 0.3651 0.3871 562 4 625.3 1187.7 0.7625 0.6132 1.3757 5G0 580 0 1326.2 0.0228 0.2994 0.3222 589.1 589.9 1179.0 0.7876 a.5673 1.3550 580 600 1543.2 0.0236 0.2438 0.2675 617.1 550 6 1167.7 0.8134 0.5196 1.3330 600
, 620 1786.9 0.0247 0.1962 0.2208 646.9 506.3 1153.2 0 8403 C.4689 1.3092 620 540 2059 9 0 0260 0.1543 0.1802 679.1 454.6 1133.7 0 8686 0.4134 1.2821 640 >
640 2365.7 0 0277 0.1166 0.1443 714.9 392.1 1107.0 0 8995 0.3502 1.2498 660 680 0.0304 0.0808 0.1112 758.5 310.1 1068.5 0 9365 0.2720 1.2086 680 270E.6 700 3094.3 0.0366 0.0386 0.0752 822.4 172.7 995.2 0.9901 0.1490 1.1390 700 705.5 3208 2 0.0508 0 0 0508 904.0 0 906.0 1.0612 0 1.0612 705.5
+
fas"t-er-ANSWERS - Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics 5.1 It increases neutron moderation near the center of the fuel assembly. This tends to flatten the flux distribution across the entire assembly thereby reducing the local peaking factor. (1.5)
Ref.: Fuel Lesson Plan 5.2 Quality is the mass fraction of steam in a mixture and void fraction is the volume fraction of steam in a mixture. (1.0)
Ref.: HTFF Lesson Plan, page 10-4 5.3 a. Critical power increases. A greater power input is needed to raise the coolant enthalpy to saturation and change water to steam. (1.0)
- b. Critical power increases. The enthalpy rise required to boil water is higher at the lower pressure. (1.0)
- c. Critical power decreases. This highest quality would be expected to occur around the hottest pin. As local peaking factor increases it is expected that the hottest pin is closer to OTB. (1.0)
- d. Critical power decreases. In a bundle with a bottom peak,
, the most rapid enthalpy rise occurs in areas of low quality, thus providing a large margin to OTB. In a top peaked bundle, the highest enthalpy rise occurs in a region which is closer to OTB. (1.0)
~
Ref.: HTFF Lessor. Plan, pages 11-2 and 3 5.4 a. Reactor Pressure > 785 psig and Core Flow > 10% of Rated The existence of a minimum critical power ratio (MCPR) less than 1.07 for two recirculation loop operation (1.10 for SINGLE LOOP OPERATION (SLO)) shall constitute violation of the fuel cladding integrity safety limit. (1.5)
- b. Core Thermal Power Limit (Reactor Pressure s 785 psig or Core Flow 5 10% of Rated When the reactor pressure is.5 785 psig or core flow is less than or equal to 10% of rated, the core thermal power shall not exceed 25 percent of rated thermal power. (1.0)
Ref.: HIFF Lesson plans, pages 13-1-4 and T.S. 1.1-1 9
5.5 RPF is Radial Peaking Factor. The radial peaking factor is found by making a comparison of the bundle power of interest to the core average bundle power. By virtue of its radial position, the bundle power will be different from the average power.
Or RPF = bundle power of interest core average bundle power (1.0)
APF is Axial Peaking Factor Certain axial positions of the fuel will have a higher heat flux or power than the average for the fuel rod or assembly.
Axial peaking factor defines the ratio between the power at a point of interest (node) and the average nodal power of the assembly.
Or APF = nodal power average nodal power (1.0)
Ref.: HTFF Lesson Plans, pages 15-4-5 5.6 a. Spontaneous fission - Curium or Uranium (0.5) fhc 7c
- b. 14huhe neutrons y induced dissociation of deuteruim (2H) (0.5)
- c. Alpha neutrons - (a, n) reaction of 0-18 (0.5)
Ref.: Rx Theory, RXTH-SH-16 5.7 When power is changing, the relative concentrations of the precursors change because short-lived (large A .) precursor concentrationschangefasterwhileapproachingtheirfinal, steady-state values, than do the long-lived precursor. (2.0)
Ref.: RXTH-Sh-20, page 8 5.8 Circulating water is maintained subcooled while the steam undergoes a change in phase. The heat removal required to condense the steam (i.e., latent heat of condensation) accounts for the large difference in flow rates. (1.0)
Ref.: Standard Thermodynamics 10
l l
. . I 5.9 a. INCREASES or becomes more negative (0.25) due to the buildup of resonance absorption materials, such as Pu-240 and fission products not present at BOL (0.75). (1.0)
- b. DECREASES or becomes less negative (0.25) due to smaller fractional change in the neutrons being ,
resonantly captured (0.75). (1.0)
- c. INCREASES or becomes more negative (0.25) due to an increase in neutron slowing down length, causing neutrons to spend more time in the resonance energy spectrum (0.75). (1.0)
Ref.: Standard Nuclear Theory 5.10 a. T=E p/Ap so p=B/AT + 1 A = Lambda assume B = .0072 (BOL) and A = .1 4 p = .0072/(100)(0.1) +1 = 6.545 x 10E-4 delta k/k (1.0) 1
- b. After the initial prompt drop, power cannot decrease l
faster than the longest lived delayed neutron appears. (1.0)
- c. Yes. The initial drop in power will only be due to the prompt neutrons. (1.0) 5.11 a. The central rods are more likely to exceed the 2200*F limit. (0.5)
In the event of a LOCA, the fuel would dry out rather quickly and the primary heat transfer mechanism prior to rewetting would be thermal radiation. The edge rods can radiate heat away from the bundle, while the central rods radiate much of their heat to other central rods. (1.5)
- b. No. The edge rods, and the corner rods in particular, have higher local peaking factors. This is due to the water gaps. (1.5)
Ref.: Heat Transfer and Fluid Flow, page 15-2 and Standard Nuclear Theory END OF SECTION t
l 11 l
i
ANSWERS - Plant Systems Designs, Control, and Instrumentation 6.1 Number 2 seal pressure would increase (1.0)
The seal high flow alarm would annunciate (1.0)
Ref.: Recirculation System Lesson Plan, figure 3 6.2 The lower portion of the bundle has the lower enrichment. (0.5)
The purpose is to help control axial power without the use of 2:!! ieds. (1.0)
Ref.: Fuel Lesson Plan, page 23 6.3 a.tc.RYarways - heat clamps and lagging. (1.0)
N: r; = g: C "AC prenure cemgem oied.- MT 1 /"'- 4L4)-
- b. Rapid RPV depressurization below 500 psig. (1.0) e ,. oc a s os is < 7 t - u'.
Ref.: Reactor Vessel Instrumentation Lesson Plan and E0P Cautionc, page 6 6.4 Regardless of the rod pattern, it is impossible to reach 280 cal / gram in the event of a control rod drop occurring at > 20% power. (1.0)
Ref.: T.S. 3.3-13 6.5 Drywell high pressure (2 lbs)
Reactor water level low (170")
Reactor Building exhaust high radiation (11 mr/hr)
Refuel Floor exhaust high radiation (9 mr/hr)
Tripping cfr Gn the n ~Lock ^ HOut relay
'* M i ' (s
" (2) on the Standby Gas Treatment Panel
!FNcn)
Any 4 at 0.5 each (2.0)
Ref.: Secondary Containment lesson Plan, page 5 6.6 a. C~ f , d er pressure on the under piston area can cause the device to drift in as the over pressure are3 depressurizes. (1.5)
,< . . L, . fw-~
' L ifD nt'< .:/ 4Z m-4 _jw y4m%
- b. Withascramdichargevolumeleak,r'e[ctorwat7ra.cr.[.J{ t hi temperature continuously leaks past CRD seals to the SDV. (1.0)
Ref.: CRD Lesson Plan .
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6.7 a. 1. Any rod can be moved only a single notch at a time. .(D-57 e 75
- 2. All rods assigned to a notch group have to be within one notch of one another. % .7I
- b. The RSCS determines rod position by sensing the direction in which the Rod Movement Control Switch is moved and sensing the settle function of the RMC system. (1.0)
Ref.: RSCS Lesson Plan 6.8 :. . Reactor high pressure scram signal present as sensed by RPS and (1.0)
An SRV opened as sensed by two of the three tailpipe pressures switches (1.0)
- b. Activating LLS logic affects S/R valves PSV 4401 and 4407.
S/R valve Open Close 4401 1020 900 (1.0)
M 4407 1025 905 (1.0) si c. nica.c -
- c. Logic is reset using switches on IC03. (0.5)
Ref.: LLS Lesson Plan 6.9 a. RCIC pump discharge rate from a flow element in the pump discharge line. (1.0)
- b. 1. Condensate in "A" RHR Ht exchange - Mode selector switch in HxA position.
- 2. Condensate in "B" RHR Ht exchange - Mode selector switch in HxB position.
- 3. Test speed adjust potentiometer with RCIC turbine speed test switch in test.
- 4. With flow controller in manual, can directly control&e6 f ;f.in flow. h b. e7 &~.
Any 2 at 1.0 each Ref.: RCIC Lesson Plan, pages 29-32.
I 13 i
B l
6.19 a. To overcome the maximum positive reactivity resulting from cooldown and xenon decay after a complete shutdown afbee-pre ide u . reyu;.e4 et, tde .. ,! . (1.0)
- b. 1. Total jet pump developed head
- 2. Core plate dp
- 3. CRD drive water dp
- 4. CRD cooling water dp
- 5. Jet pump dp j 6. Core spray system line break detection l
Any 4 at 0.5 each END OF SECTION t.
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ANSWERS - Procedures-Normal, Abnormal, Emergency, and Radiological Control 7.1 a. An SR0 An assistant A Reactor Engineer (1.0)
- b. If RE not available An SR0 An R0 or SRO An assistant (1.0)
This is provided the SR0 is fully cognizant of the requirements of fuel moving accountability and can assume all the tasks performed by the RE. (0.5)
Ref.: FRCHP No. 5, page 8 7.2 Notching the control rod to its fully withdrawn position (0.5) verifying subcriticality (0.5) verifying coupling of the CR0 by attempting to drive to the overtravel position (0.5) and then measuring the time required for the rod to drive in. (0.5)
Ref.: FRCHP No. 5, page 9 7.3 Whenever a control rod is not fully inserted or is in motion in a cell containing fuel all personnel will be at least out of line of sight of loaded fuel. (1.0)
Anytime more than one control rod is withdrawn in control cells that contain fuel, all personnel will be off the 855' level. (1.0)
Ref.: FRCHP No. 5, page 9 7.4 River Water Supply RHR (Containment Spray Mode)
RHR Service Water Emergency Service Water Leak Detection System (2.0)
Any 4 at 0.5 each Ref.: IPOI-2, page 6 15
l
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l 7.5 RHR Service Water Crosstie.
Fire System SBLC ESW Condensate Service Water (2.5) 5 at 0.5 each Ref.: E0P1, page 14 7.6 a. Bypass the IRM channels not exhibiting proper overlap. (1.0)
- b. If more than one IRM channel in either RPS trip channel must be bypassed, insert a manual half-scram in the re-spective trip channel. (1.0)
- c. Red.se reactor power to 10 4cps as read on SRM's. (1.0) 7.7 1. , Control rods which cannot be moved with control rod drive pres su're.
l 2. Control rods whose position ~cannot be positively determined.
- 3. Uncoupled control rods.
- 4. Control rods whose 90% scram insertion time exceeds 7.0 seconds.
- 5. Control rods that cause the average scram insertion time for all operable control rods to exceed the limits.
- 6. Control rods that cause the average scram insertion times for the thren fastest control rods of all groups of four control rods in a 2x2 array
- 7. 7 e-~ / h e d to exceed the limits.
t' Pnf<d 3 at 1.0 acn .
Ref.: T.S., pages 3.3.1-6 7.8 a. EHC pressure setpoint is reduced as necessary. (0.5)
The. Bypass Valve Opening Jack. jf' (0.5) 0sb. i w ,f nit.u / -~ Z,* we.t ,/ a n W s 7 -'"g ,,,,_
O en and throttle t'he steamline drain valves. (0.5)
Ref.: IPOI 4, page 19 l
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16
7.9 a. 150*F (0.5)
Hot reactor water is allowed to drain through the RHR piping to RW. (1.0)
- b. The LPCI outboard flow control valve 3should be throttled open within 10 seconds to oravanLofIening the minimum flow valve. g,; g g gm ve v, (1.0)
- c. Control the cooldown rate by adjusting the Hx bypass flow control valve and the Hx inlet flow control valve. (1.0) c y GL d ,( 1 2d e er L M Jr'p u y 7.10 1. Take action, as specified in Emergency Operating Procedures or as otherwise may be required, to restore the plant to a safe condition within the limitations pres 6ribed in the DAEC Operating License and Technical Specifications.
- 2. Evaluate plant conditions, initially classify the event and recommend reclassification of the event based upon plant status and/or projected trends.
- 3. Develop initial protective action recommendations, as appropriate, and initiate notification of Iowa Electric and off-site agency personnel.
- 4. Initiate evacuation of the plant, as required, and insure accountability for all operating personnel on shift.
- 5. Delegate any (or all) responsibilities described in this procedure, as necessary, to properly respond to the emergency situation.
Will also accept
- 6. Shall function as the Emergency Coordinato and site rad protection coordinator until relieved.
Any 3 at 1.0 each Ref.: EPIP 2.5 END OF SECTION 17
ANSWERS - Administrative Procedures, Conditions, and Limitations 8.1 Cold shutdown mode (0.5)
Cold refuel mode N' ^ 6 -
(0.5) w.iti Ju u<.:,.1 Ref.: 1410.0, page 2 8.2 a. The Hold card is used to safeguard human life and the Warning Tag is used to safeguard equipment. (1.0)
- b. 1. F (0.5)
- 2. F (0.5)
- 3. T (0.5) 8.3 a. Radiation Protection supervisor or the Operations supervisor. (1.0)
- b. Check valve or other flow reversing device. (0.5)
Ref. : 1410.5, page 8 8.4 a. 1. This assures that the changes in coolant temperature at the reactor vessel nozzles and bottom head region are acceptable. (1.0)
- 2. This assures that the vessel stresses will not exceed allowable per the ASME code. (1.0)
- b. Decreasing CRD cooling water flow. . -
Increasing bot. tom head drain flow. h "w~7 ^'"Y"[k (1.0) (1.0)
Z.~
Ref.: ~ chT.S.~rl3-6.17-18 and IP01-2, page 12 8.5 Mode switch in shutdown Manual Scram High flux IRM Scram discharge volume high level APRM 15% flux 5 at 0.5 each Ref.: T.S. 3.1-6 and IP01 4, page 6 18
,. c.- .- , a ., y.-_....., .., .., ,. , .~ , . . , . .-.,,
l 8.6 a. Either demineralized water transfer pump A or B (1P-13A or B) should be kept running to circulate water through heater (IE-14) which must be in service. (1.0)
- b. Jockey Pump (1P-11) or its backup, Cond Service Pump A or B (1P-12 A or B) should be kept running to circulate water through the heater (IE-15) which must be in service. (1.0)
Ref.: IPOP-6, page 4 8.7 The Reactor Engineer and the OSS. (1.0)
Ref.: IP01 3, page 5 8.8 a. A Limiting Control Rod Pattersexifsts when core thermal power is greater than or equal to 30% of rated and less than 90% of rated (30% < P < 90%) and the MCPR is less than 1.70. or (1.0)
- b. Core thermal power is greater than or equal to 90% of rated (P>90%) and the MCPR is less than 1.40. (1.0)
Ref.: T.S. 3.3-17 I
8.9 Detector levels A and C of one LPRM string per core octant plus detector levels A and C of one LPRM string in the center of the core shall be monitored. (2.0)
Ref.: T.S. 3.3.7 8.10 Primary Containment is not satisfactory. (0.5)
The TIP ball valve is an isolation valve (PCIS GP 2) and it is malfunctioning. (0.5)
Either of the following: Restore the valve to operable status or deactivate the TIP ball valve in the isolated position. (1.0)
Ref.: T.S. Definitions 8.11 a. YES, a T.S. violation has occurred. (0.5)
A Safety Limit is assumed to have been exceeded if the scram is accomplished by a means other than the primary source signal. (1.0)
- b. The reactor shall be shutdown -HM-!*UF4 (0.5)
An immediate report shall be made to the Director-Nuclear
, Generation and to the Safety Committee. (1.0)
Ref.: T.S. 1.1.C and 6.7 /
l}[]tz n.J A 2~f D ' " L - Q J 2tu. 4 l
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.. .o 8.12 a. Insure that the intent of the procedure is not altered. (SPM . l 1
- b. Ensure that two members of the plant management staff, at least one of whom holds a valid SR0 license, concur with'the
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