ML20210F790
| ML20210F790 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 02/06/1987 |
| From: | Bishop M, Burdick T, Dave Hills, Mary Spencer NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20210F772 | List: |
| References | |
| 50-331-OL-86-03, 50-331-OL-86-3, NUDOCS 8702110164 | |
| Download: ML20210F790 (55) | |
Text
i 8 1 U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-331/0L 86-03 Docket No. 50-331 License No. DPR-49 Licensee: Iowa Electric Light and Power Company IE Towers Post Office Box 351 Cedar Rapids, IA 52406 Facility Name: Duane Arnold Energy Center Examination Administered At: Duane Arnold Energy Center Examination Conducted: December 16-19, 1986 B.Nd L Examiner (s): D. Hills 2 7 Date M. Bishop 6 7
_Date M p 2 [ '
'Date Approved By:
h
- f. M. BUrdick, Chief bb_g
, Operator Licensing Section Date Examination Summan Examination administered en December 16-19_, 1986 (Report No.50_-331/0L-86-03)
Written and oral examinations were aaministered to two Reactor Operator (RO) i candidates.
Results: All candidates passed these examinations.
i 8702110164 870,'06 PDR ADOCK 05000331 V PDR
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s REPORT DETAILS
- 1. Examiners D. Hills', Chief Examiner M. Bishop M. Spencer
- 2. ' Examination Review Meeting Specific facility comments concerning written examination questions, followed by the NRC response, are, enumerated in the attachnent.
- 3. Exit Meeting At the conclusion of the examinations, an exit meeting was conducted.
The following personnel cttended this exit meeting:
Facility Representatives D. Mirieck, Plant Superintendent G. Van Middlesworth, Training Superintendent P. Roy, Supervisor, License Training R. Schlesinger, Training Instructor J. Morris, Training Instructor NRC Representatives
. . N. Chrissotimos, Deputy Director, Division of Reactor Safety 4 D. Hills, Operator Licensing Examiner
,J. Bjcrgen, Operator Licensing Examiner M. Spencer, Operator Licensing Examiner, INEL M. Bishop, Operator Licensing Examiner, INEL J. Wiebe, Duane Arnold Senior Resident Inspection As requalification examinations were administered concurrently with these replacement examinations, discussions, for the most part, concerned requalification program evaluation guidelines, criteria, and responses.
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ATTACHMENT Dt!ANE ARNOLD R0 REPLACEE NT AND REQUALIFICATION EXAMS DECEfEER 16, 1986 1.01a Facility Comment: (Replacement)
The answer requires " Flow Orificing", we request that you accept either " Flow Orificing" or " Core Orificing" since the two are used interchangeably. Reference General Electric - Heat Transfer and Fluid Flow p. 9-51.
Resolution:
Comment is valid. " Core Orificing" will be added to answer as an acceptable response since this term is used in the provided re-ference material. Point value remains the same.
1.07c Facility Comment: (Replacement and Requalification)
The question asks why significant radiation levels are not present in the offgas holdup piping like they are in the vicinity of the MSIVs.
Answer states that N-16 decays with a short half life so it is decayed by the time it leaves the hotwell.
We request that the answer reflect the "offgas holdup pipir.g".
Resolution.
Coment is valid. The answer will be changed to read as follows:
N-16 decays with a short half life (7 seconds) so it is essentially decayed off prior to reaching the offgas holdup piping. (0.5) 1 1.08c Facility Comment: Replacement (1.08b Requalification)
This question is an ex;remely complex question which we believe could be any of the three allowable responses. The answer depends on which of the parameters is believed to have the strongest effect. If you take the fact that as voids decrease the thermal flux increases, you will arrive at the conclusion that Rod worth will increase. But this assumes the total to ave flux ratio has increased which may or may not be true and it also ignores the thermal diffusion length change.
If there is only one correct answer though, it should be decrease. This is due to the thermal diffusion length decreasing as voids decrease and the local to ave flux ratio stays approximately constant, therefore Rod worth will decrease. See RXTH-IG Objective 6.3. Also see NED0-24810 B page 14-4 and H-17-17/H-18. In the region of interest it clearly shows that Rod worth decreases as power increases, but the change is relatively I
small, see " maximum Rod worth with single error" curve.
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The worth of a control rod depends on many factors and a proper discussior to support the candidates conclusions should be allowed.
Resolution:
Comment is valid. Since the process is complex and the candidate may interpolate from different points in the power increase, any reasonable response will be accepted.
1.09b Facility Comment: (Replacement)
Various means of increasing the subcooling may be generated from an operating point of view, vice a generalized theoretical viewpoint.
We request the following be considered as acceptable means of in-creasing the subcooling.
- 1. Starting one or more Cooling Tower Fans.
- 2. Placing one or more Cooling Tower Cells back in service.
- 3. Starting a second Circ. Water Pump.
- 4. Placing more air ejectors on or starting the Hogger. (Better vacuum will enable more energy to be extracted from the steam and, therefore more subcooling.)
- 5. Opening the blowdown valve to bring in more river water (cooler water). See System Description D-3, Figure 2, Valve M0-4253.
Resolution:
Comment is valid. Even though the present answer is not a generalized theoretical viewpoint but an operational reality, it is reasonable to expect the candidates to respond with specific methods of accomplishing the desired effect. Therefore any reasonable means of accomplishment will be accepted.
1.11 Facility Comment: (Replacement)
The question asks for TRUE or FALSE and EXPLAIN YOUR ANSWER. Both candidates never saw the " EXPLAIN YOUR ANSWER" and thus just answered the TRUE/ FALSE portion. The statement EXPLAIN YOUR ANSWER is hard to pick up even though it is capitalized. It gets lost being on the same line as the TRUE or FALSE. We believe that the " EXPLAIN YOUR ANSWER" would have been better off after the statement to which it was referring.
Resolution:
" EXPLAIN YOUR ANSWER" will be moved to the end of the statement in future use of this question.
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2.01a Facility Comment: (Replacement and Requalification)
Request you also accept "taking a suction on the CST when beth condensate pumps are secured". See attached SD A-1 page 56.
Resolution:
Comment is valid. This statement is in the referenced material and will be accepted for partial credit as an interface with the condensate system. The answer key will be changed to read as follows on Replacement Exam:
- a. "CRD pumps take suction from the condensate reject line to CST when at least one condensate pump is in operation (0.5). When neither condensate pump is operating, CRD suction is directly from the CST. (0.5)"
On Requalification Exam words will be the same but point values will be 0.25 each in lieu of 0.5 each.
2.01c Facility Comment: (Replacement and Requalification)
Request you also accept " Seal Leakoff from the Recirc Pumps" since it is supplied by CRD and drains to the Radwaste Sump system. See attached SD A-2 page 68.
Resolution:
Comment valid. "Recirc Pump Shaft Seal Staging Flow" will be added to the answer sheet. I assume this was the desired answer since it was the Highlighted response in the supplied reference. Point value remains the same.
2.04 Facility Comment: (Replacement and Requalification)
Low Steam Line Pressure is a direct Turbine Trip, but it is also a signal to isolate both the inboard and outboard steam supply valves.
It is not a true isolation in the sense that it will reset itself, but it does isolate HPCI from the Reactor with the same steam line See attached SD C-3 Figure isolation valves as a regular isolation.
9.
Resolution:
Comment is valid. Credit will not be deducted if Isolation is "ALS0" included for this answer. A notation to this effect will be added to the answer key.
2.08c Facility Comment: (Replacement and Requalification)
The answer key has "CLOSE" as the answer, but it should be "N0 CHANGE".
The only heat exchanger throttle valve in the RHR system is the inboard 3
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valve. See attached 0.I. 149 page 16 of 143. This valve does not get any automatic signals to either open or close. The valves in question are MOV-1939 for Loop B and MOV-2029 for Loop A. See attached SD C-1 Figure 2. As shown on the attached SD C-1 Figures 11 & 13, neither of these valves get either an open or shut signal on a LPCI initiation signal or LPCI Loop Selection Logic Initiation.
Resolution:
Comment is valid. Answer key will be changed to "No Change" for part "C'.
2.09a Facility Comment: Replacement (2.09 on Requalification)
This question may elicit procedural interlocks in addition to the system interlocks. See attached 0.1. 183.3 page 4 of 37.
Resolution:
Comment is not valid. The question specifically ask for automatic interlocks therefore procedural requirements would not be acceptable.
The answer remains unchanged.
2.12 Facility Comment: (Replacement)
We request you also accept as another load the offgas jet compressors.
These are listed in the attached SD A-6 page 10 and this particular component is also referenced in the attache'd SD D-7 page 23.
Resolution:
Comment is valid. "Offgas jet compressor" will be added to the answer key as a possible response. Point value remains unchanged.
3.01a Facility Comment: (Replacement)
The pumps all start in the same manner. They will always have a delay bu'lt in whether or not there is a loss of offsite power. Therefore, pa-t a. should be the same answer as part b. See attached SD C-1 page
- 32. (Recent Revision)
Resolution:
! Comment is valid. The "A" part of Question 3.01 will be deleted and l
the point value reduced by 0.5.
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! 3.01b Facility Comment: (Replacement)
Yo;r answer states 10 seconds and 15 seconds after the standby diesel 9enerators come on line, but it is actually sensed "when essential bus voltage is greater than 65% of rated". See attached SD C-1 page 32.
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We request that either answer be acceptable.
Resolution:
Comment is valid. The answer will be changed to "after essential bus voltage is greater than 65% of rated" instead of "after standby diesel generators come on-line." Credit will not be given for either answer.
3.05 Facility Comment: (Replacement and Requalification)
There is no correct answer among the four answers provided.
Answer a. is wrong because the circuitry requires two condensate pumps running to start the second Reactor Feed Pump. See attached SD D-15 page 11.
Answer b. is wrong because tripping one condensate pump with both Reactor Feed Pumps running will cause its associated Reactor Feed Pump to trip, not both Feed Pumps. See attached SD D-15 page 11.
Answer c. is wrong because there is no time delay for lube oil cressure.
An Auxiliary Lube Oil Pump is provided to supply greater than 9 psig for the Reactor Feed Pump Start Logic. See attached SD D-15 page 11.
Answer d. is wrong because the low pressure suction trip has a 5 second time delay associated with it. Therefore, the feed pump will not trip immediately. See attached SD D-15 page 11.
l Due to no correct answer available, we request this question deleted.
Resolution:
Comment is valid. This question will be deleted and the point value adjusted accordingly. The reference material supplied was very confusing.
3.09b Facility Comment: (Replacement and Requalification)
Your answer is correct if you have the recirc system back to nornal.
If you assume that the "B" recirc pump is at 20% speed due to the problem in part "a", then you will only get the "A" recirc pum; to runback to 45%. We request that you accept either answer as correct.
Resolution:
This comment is addressed by the question itself (STATE ANY ASSUMPTIONS YOU MAKE). If the candidate states he assumes this condition then full credit will be given for actions based on those assumptions. For future use a statement will be added to the question to require consideration of each failure separately.
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3.12(3) Facility Causent: (Replacement)
Request you also accept < 9 LPRMs for APRMs A-D and < 13 LPRMs for APRMs E and F if stated in this manner. Your answer is correct, this is just the more specific case. See attached SD I-3 page 23.
Resolution:
Comment is valid. Answer key will be modified to also accept this answer for full credit for answer No. 3.
3.14 Facility Comment: (Replacement)
A. May get " Controlling Pressure Regulator" since the bias has been applied to the "B" regulator.
B. This may be called the " Pressure Average Manifold" vice " Steam throttle Pressure". Either should be correct since P & ID M-103 calls it the " Pressure Averaging Manifold" and the System Description D-11 Figure 4 shows it as the " Steam Throttle Pressure". Both references are attached.
D. May get "Non-Controlling" Pressure regulato: since this has a bias applied.
Resolution:
Comment is valid. These alternate answers will be added to the answer key.
3.15 Facility Comment: (Replacement)
We request that you also accept " Loss of 125 VDC" or " Loss of Control Power" as an acceptable answer. The Trip circuit and red indicating light are powered from Control Power which is normally 125 VDC and loss of control power will prevent a leitimate trip from occurring.
This would be a fault in the tripping circuit. See the attached typical 4160 volt breaker control, Bech E104-3.
Resolution:
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Comment is valid. These two alternate answers will be added to the answer key as acceptable for full credit.
4.06 Facility Comment: (Replacement and Requalification)
We request that you accept either your answer, what the Startup IP01 2 states or accept " Observe that at least two IRM channels in each trip system are increasing before the highest SRM count rate reaches 105 cps."
See IPOI 2 page 17 of 48 and 01 878.2 page 7 of 21. DAEC is in the process of changing to the statement in 01 878.2, but it is not complete.
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Resolution:
Comment is not valid. Even though it is unfortunate that a conflict exist between the procedures the question specifically ask for the action required by IP01-2. Therefore the answer must remain un-changed.
4.09 Facility Comment: (Replacement and Requalification)
The question asks "What will your actions be per IP01 7, to complete the electrical lineup, yet the answer key requires something that you do not do..." Leave the adjacent, "Init." column blank." We request that no points be deducted if this has not been stated. We do request that credit be given for:
- 1. Listing of all dis:repancies on the cover sheet.
- 2. Returns the lineup to the OSS for review.
These two actions are required per IP017, Rev. O, paragraph 5, step 5.1 (3), page 16.
Resolution:
Comment is partially valid. Leaving the "Init." column blank is a requirement of the procedure that reauries awareness by the operator and will remain in the answer. The additional two items will be added to the answer and will also be required for full credit. The point value remains unchanged.
Examiner Changes 208H (Replacement and Requalification)
This part of question will be deleted since it repeats part D.
Point value of the question remains 2.00 but each part now worth 0.285.
3.01b (Replacement)
Point value changed on answer key to be consistent with question value. This was a typo discovered during grading.
3.10 (Replacement and Requalification)
The following statement was added to the answer key to make this question consistent with the double jeopardy rule:
" Credit will be given for actions based on answer regarding the readings after the range change."
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- - MASTERCOPY U. S. NUCLEAR REGULATORY COMMISSION
- REACTOR OPERATOR LICENSE EXAMINATION FACILITY: duane arnold REACTOR TYPE: BWR-GE4 DATE ADMINSTERED: 86/12/16 EXAMINER: BISHOP. M.
CANDIDATE INSTRUCTIONS TO CANDIDATE:
Write answers on one side only.
Use separate paper for the answers. Points for each Staple question sheet on top of the answer sheets. The passing question are indicated in parentheses after theand question.
a final grade of at arade requires at least 70%
Examination papers in each will category be picked up six (6) hours after least 80%.
the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY SCORE VALUE CATEGORY VALUE TOTAL 23.39 1. PRINCIPLES OF NUCLEAR POWER 25.00 25.00 PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.JB 25.00 -E S . 00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 23so 232L 3. INSTRUMENTS AND CONTROLS 25.38 4. PROCEDURES - NORMAL, ABNORMAL, 25.00 -25.00 EMERGENCY AND RADIOLOGICAL CONTROL 98,50 400.0 Totals I have neither given All work done on this examination is my own.
nor received aid.
Candidate's Signature
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS i
During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application cnd could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
Number each answer as to category and number for example, 1.4, 6.3.
9.
- 10. Ship at least three lines between each answer.
- 11. Separate answer sheets f rom pad and place finished answer sheets f ace down on your desk or table. .
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
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- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- o. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions, Leave the examination area, as defined by the examiner. If after d.
leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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y Pcco 4 61 . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (2.50)
A design feature in the reactor vessel ensures proper flow distribution through the fuel bundles.
- a. What is this feature (or component)? (0.5)
- b. EXPLAIN why this feature is necessary in a BWR, including the consequences of a power increase if this feature were eliminated. (2.0)
QUESTION 1.02 (1.00)
The value of Beta (eff) changes over core life because: (choose one)
- a. As lambda (eff) decreases, beta must also decrease.
- b. There is an increased percentage of fissioning from Pu-239, which has a smaller delayed neutron fraction than U-238 and U-235.
- c. As U-235 is fissioned and effectively used up, there are fewer fast fissions which result in fewer delayed neutrons.
- d. The microscopic cross sections for the isotopes producing delayed neutrons progressively become smaller with core age.
QUESTION 1.03 (1.00)
A moderator is necessary to slow neutrons down to thermal energies.
Which of the following is the most correct reason for operating '
with thermal instead of fast neutrons? (1.0)
- 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 l thermal neutrons than for fast neutrons.
- d. Doppler and modertaor temperature coefficients become positive i
as neutron energy increases.
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(***** CATEGORY l CONTINUED ON NEXT PAGE *****)
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Pega 5 1: PRINCIPfTR OF NUCTFAR POWER PLANT OPERATION.
TFFRMODYNAMICS. HEAT TRANSFEx AND FLUID FLOW QUESTION 1.04 (2.00)
Assume a large centrifugal pump is operating normally with the discharge valve fully opened. What changes will occur in the below parameters if the discharge valve is throttled to the 50% open position? (Limit your answer to INCREASE, DECREASE, or NO AFFECT.)
- a. Pump motor current
- b. Pump discharge head
- c. Available Net Positive Suction Head (NPSH)
- d. Pump flow rate QUESTION 1.05 (2.00)
The attached figure 2 shows the " Normal Operating Map" of rated core flow versus reactor power.
- a. EXPLAIN WHY, on the natural circulation line, incremental increases in power initially produce very rapid increases in core flow, but eventually reach a point where further power increases produce no increase in core flow. (1.5)
- b. EXPLAIN WHY, on the pump constant speed line, core flow increases as reactor power decreases. (0.5)
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIOL Pcsa 6
. THERMODYNAMICS. HEAT TRANShER AND FLUID FLOW CUESTr% 1.06 (3.00)
For each of the three events listed below (a - c):
- 1. State which reactivity coefficient will respond first?
- 2. Why said reactivity coefficient responds?
- 3. Is the response POSITIVE or NEGATIVE?
- a. SRV opening at 100% power. (1.0)
- b. Rod drop at 100% power. (1.0)
- c. Isolation of a feed heater string at 100% power. (1.0)
QUESTION 1.07 (1.50)
The radiation level in the vicinity of the MSIV'S is normally very high due to the radioactivity caused by one particular isotope. .
a) What is this isotope? (0.50) b) How is it produced? (0.50) c) Why is it not present in any significant quantities in the Off- (0.50)
Gas holdup piping?
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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Pace 7 If PRINCIPLES OF NUCfJAR POWER PLANT OPERATION.
TERMODYNAMICS. RAT TRANSFu AND FLUID FLOW
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QUESTION 1.08 (3.00)
Indicate BOW (increase, decrease, unaffected) control rod worth changes for each of the situations listed below. EXPLAIN WHY the rod .
worth is affected, if applicable. ,
(1.0)
- m. The reactor is heated from 100 des F to 200 des F
- b. Reactor power is increased from 20% to 40% by control rod (1.0) withdrawal l
- c. Reactor power is increased from 70% to 90% by increasing (1.0) recirculation flow QUESTION 1.09 (2.00)
T-S diagrams of real plant cycles show a small amount of " condensate depression" (subcooling) in the condenser.
- a. How would cycle efficiency be affected if subcooling is (1.5)
Why?
,pbereased?
b.Howcouldtheoperator2bereasetheamountofsubcooling? ( 0. 5')
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QUESTION 1.10 (1.50)
In the event of a LOCA after extended operation at full power, which fuel rods (EDGE or CENTRAL) are more likely to exceed the 2200 deg F limit? Explain WHY.
(***** CATEGORY l CONTINUED ON NEXT PAGE *****)
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- 1.
- PRINCIPLES OF NUCWAR POWER PLANT OPERATION.
THERMODYNAMICS. HEAT TRANSFKR AND FLUID FLOW l
- ) } bis fu g,Sk
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QUESTION 1.11 (1.50)
I TRUE or FALSE EXPLAIN YOUR AN The effects of " Doppler Broadening" in U-238 result in a modified capture cross-section curve, but the area under both the original and the (higher temperature) broadened curve will theoretically be the ;
same. Overall neutron capture by U-238 will be about the same at any temperature.
I QUESTION 1.12 (2.00)
The reactor is operating at high power under steady state conditions.
The time required for power to A control rod is WITHDRAWN one notch. The stabilize is noted. Now the control rod is INSERTED one notch. Which transient should time for power to again stabilize is noted.
take longer for neutron power to stabilize and WHY?
QUESTION 1.13 (2.00)
In Operating Procedures it is stated that reactor period can be calculated by multiplying the time (in seconds) that it takes the power to double by 1.443. How is this factor (1.443) determined? SHOW ALL WORK.
(***** END OF CATEGORY 1 *****)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS QUESTION 2.01 (3.50)
How does the CRD system interface with the following systems?
- a. Condensate system (1.0)
- b. Reactor recirculation system (1.0)
- c. Liquid radwaste system (1.0)
- d. Reactor building cooling water system (0.5)
QUESTION 2.02 (1.00) r What two trip functions does the " Main Turbine Stop Valve Position Less Than 90% Open" signal initiate?
QUESTION 2.03 (3.00)
- a. State FIVE conditions which will cause the emergency diesel generator engine shutdown relay (SDR) to be energized. (2.5)
(Setpoints not required)
- b. State the EDG engine protective condition which IS NOT overridden when the emergency start relays ESA and ESB are actuated. (0.5)
QUESTION 2.04 (2.00)
Identify which of the following are direct HPCI turbine trips and which are HPCI system isolations:
- 1. Low pump suction pressure
- 2. High reactor vessel water level
- 3. Low steam supply pressure
- 4. Steamline high differential pressure
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) -
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- 1. - PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 63STION 2.05 (2.00)
When both recirculation loops are operating, the loop jet pump flow signals are added together to give a Reactor Jet Pump Total Flow Signal.
- a. How is jet pump total flow determined if one recirculation loop is operating and the other loop is shutdown?
- b. WHY is a different method used when one recirculation loop is operating and one shutdown?
Q ESTION 2.06 (1.00)
What is the electrical power supply (bus and voltage) to each cf the RHR Pumps?
Q3STION 2.07 (2.00)
What four signals will cause the RWCU System to automatically isolate? Setpoints are required for full credit!
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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- 2. . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pc.se 11 l SYSTEMS l
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QUESTION 2.08 (2.00) i RHR Loop "B" is being used in the suppression pool cooling mode to cool the torus. RHR pump "B" and "B" and "D" RER service water pumps are in service.
List the changes that ocurr to the below equipment when a LPCI initiation and injection signal ocurres. (OPEN, CLOSES, NO CHANGE, STOPS, STARTS)
A. B RER Service water pump.
B. D RHR Service water pump.
C. Beat exchanger throttle valve. (M
- D. RER heat exchanger bypass valve. (MO-1940)
E. Suppression pool spray valve (MO-1932)
F. LPCI to suppression pool test valve (MO-1934)
G. RER "B" pump suction valve (suppression pool suction)
?_ S::t - e'.r; r 5- ::r - :Prc . L QUESTION 2.09 (1.50)
- a. What automatic interlocks must be satisfied prior to using the MSIV-LCS? Include setpoints if appropriate. (1.00)
I b. What is the purpose of the heater in the system? (0.50) i e
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(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
- 2. PLANT DESIGN INCLUDING EAFETY AND EMERGENCY Para 12
- SYSTEMS QUESTION 2.10 - (3.00)
- a. What are two design features that prevent inadvertent draining of the spent fuel pool? (2.0)
- b. Why must the temperature of the Fue3 Pool Cooling and Cleanup HX Outlet be maintained less than 130 deg. F.? (1.0)
QUESTION 2.11 ,
(2.00)
What four conditions are necessary for the ADS System to be AUTOMATICALLY initiated? Setpoints required for full credit!
QUESTION 2.12 (2.00)
In addition to the main turbine, the bypass, safety and safety relief valves, what are six (6) other stea= loads (systems) supplied by the main steam system? Do not use two.similar components (i.e.,
such as "A" and "B") as separate loads.
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(***** END OF CATEGORY 2 *****)
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Pcca 13 3 .- INSTRUMENTS AND CONTROLS l.co CUESTION 3.01 (t 60)
Assume a LPCI in.iection is initiated and all RHR pumps are available.
State the order of automatic pump starting for the following conditiong: (Include applicable time delays.)
, " 5 )'-
+ !!,.,sel ;;rrr ir ;c;il;bi: ':L o (1.0)
- g. A Loss of Offsite Power (LOOP) has occurred.
QUESTION 3.02 (2.00)
What are FOUR of the FIVE automatic actions which occur at the RPV Low-Low Water Level Trip, + 119.5 inches?
QUESTION 3.03 (1.50)
The main steam line flow restrictors are also used to determine the steam flow in the main steam lines. What are three uses of this main steam line flow signal?
QUESTION 3.04 (l.00)
Choose the best answer in regard to an auto initiation of SGTS.
- a. The train in auto will start on the first signal and the standby train will then be in readiness to start on any subsequent signal.
- b. Both trains will start and the standby train will trip on low flow.
- c. The train in auto will start and after the low flow alarm has cleared the standby train will start.
- d. Both trains will start immediately and run until some operator action is taken.
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
3s INSTRUMENTS AND CONTROLS Paco 14
~ ^
r 3.05 (1.00)
, QUESTION g
Choose the best answer concerning the reactor feed pump circuity; '
i a. A second feed pump can be started wit only one condensate {
pump run ga long as syltion press re remains greater than the
\ low pres nre tri point. I
/
feed pum s runn r g, a tri of o ly one condensate
- b. With both pump wilK cause both feed 'um to rip.
- c. The react r'Tuedphp o pre sure trip is automatic ally byp#sst or 10 seconds on a pump start to allow the shaf riven oil pump to build up pressure.
- d. If the a etion pressure of a running feed pump drops to less than the low pressure trip point, the feed pump will trip im=ediately.
QUESTION 3.06 (1.50)
During a failure of instrument and service air, what change (if any) will occur for each of the following parameters?
(Limit your answer to INCREASE, DECREASE, or NO CHANGE.)
(Assume no operator action is taken.)
- a. Steam seal regulator pressure.
- b. Final feedwater temperature,
- c. Standby liquid control tank level.
QUESTION 3.07 (2.00)
What are the NORMAL and ALTERNATE reference signal inputs to RBM Channels "A" & "B" ( Four items required)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
3 .. INSTRUMENTS AND CONTROLS Pcco 15 CUESTION 3.98 (0.50)
What is the logic arrangement for the Main Steam Line High Radition Level Scram?
QUESTION 3.09 (2.00)
The reactor is operating at 100% power.
h p\\ What will be the effect on/of the recirculation flow control
!, gaystemduetothefollowingconditions: (STATE ANY ASSUMPTIONS YOU MAKE )
%A n l Ca n ad e o.4 , acid J s.fdly )
.Y j, a. The full open indication on recire pump "B" discharge valve is
- lost.
.Q NOTE: The recire pump discharge bypass valve is open.
- b. The reactor feedwater pump "A" trips. (Assume reactor water level falls below 186 inches)
QUESTION 3.10 (2.50)
,For each of the IRM (Intermediate Range Monitoring) range changes (a. and b.) listed below, provide the following:
(Mode switch in STARTUP):
- 1. The indicated level on the NEW RANGE.
- 2. Any automatic actions initiated as a result of the indicated level on the new range.
- a. Switching from range setting 5, reading 25, up to range j setting 7. (1.0)
- b. Switching from range 6, reading 39, down to range 5. (1.5)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
~-r- -
3e INSTRUMENTS AND CONTROLS Pcan 16 QUESTION 3.11 (1.50)
What three conditions are required for the Main Condenser Low Vacuum Isolation signal to be bypast.ed?
Conditions may include operator action.
QUESTION 3.12 (2.N)
List four conditions that can cause an APRM channel to be either physically inoperative or to be considered inoperative?
QUESTION 3.13 (2.M)
Which RPV Water Level instruments are compensated and what parameter are they compenstated for?
QUESTION 3.14 (3. N )
Refer to Figure 1, Pressure Control Units, at the back of the test.
Identify the items specfied by the letters A through F.
QUESTION 3.15 ( 1. M )
Failure of the " red" indicating light to be "ON" when a breaker is Closed idicates what fault? (Bad buld is an unacceptable answer!)
f i
(***** END OF CATEGORY 3 *****)
4.- P?.OCEDURES - NORMAL. ABNORMAL. EMERGENCY Poco 17 l AND RADIOLOGICAL CONTROL QUESTION 4.01 (2.50)
- a. According to the Tagout procedure, what THREE actions are you ,
required to perform if while on a work assignment a Warning Tag or Hold Card is found to be missing? (1.5)
- b. Briefly explain how the use of the Warning Tag and Hold Card differs.
(1.0)
~
QUESTION 4.02 (1.00)
Which of the following would not necessarily be a symptom of unidentified leakage in the drywell of 5 GPM or higher? (Choose one)
- n. Excessively increasing drywell floor drain sump total flow at F.Q. 3707 on Panel 1C19.
- b. DRYWELL HIGH TEMPERATURE alarm on Panel 1C04-C
QUESTION 4.03 (2.50)
What are the five entry conditions for EOP-1, RPV CONTROL 7 QUESTION 4.04 (2.50)
What are the five entry conditions for EOP-2. PRIMARY CONTAINMENT CONTROL?
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
4 ,
4'- PROfTDURES - NORMAL. ABNORMAL. EMERGENCY PO4m 18 AND RADIOLOGICAL CONTROL j
QUESTION 4.05 (1.00)
AN LCO listed as "3/30" on part three of the NSOE and ANSOE Shift i Turnover Form designates what? l
\
QUESTION 4.96 (1.00)
Briefly describe how proper IRM - SRM " overlap" is observed per IPOI 2.
, QUESTION 4.07 (1.50)
List three of the four requirements of IPOI 2 prior to placing the MODE Switch to "RUN".
QUESTION 4.08 (2.50)
Briefly explain the REASON (S) for each of the following cautions in j the Emergency Operating Procedures:
- c. Manually trip Standby Liquid Control pumps at 95 in the SBLC tank. (0.5)
QUESTION 4.09 (2.00)
You are performing a normal system electrical lineup and discover a A Hold Card / Warning Tag is breaker is NOT in the required position.
attached to the breaker.
What will your actions be per IPOI 7 for completing this lineup?
~( " - ; p. % ,.c M
i I
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
i
Pega 19
.4 " PROCEDURES'- NORMAL. ABNORMAL. EMERGEN~Y AND RADIOLOGICAL CONTROL QUESTION 4.10 (2.50)
- a. How is a control rod drive electrically disarmed? (1.0)
What are TWO instanccs requiring a control rod drive to be b.
electrically disarmed? (1.0)
- c. What is the maximum number of inop. control rods during reactor power operation?
(0.5)
QUESTION 4.11 (2.00)
- a. STATE the exposure rate limits which characterize the following: (1.5)
- 1. Radiation Area
- b. Assuming you have an approved NRC Form 4 (exposure history) on file, WHAT is the maximum whole body exposure you are allowed in ONE DAY without additional authorizations. (Also assume it is the i
first day of a new calendar quarter. ) (0.5)
CUESTION 4.12 (2.50)
The reactor is operating at 90% power when condenser vacuum suddenly starts to decrease:
In accordance with IPOI 6.4.0, Loss of Condenser Vacuum, WHAT are FIVE of the SIX Immediate Operator actions, OTHER THAN acknowledgement of annunicators and announcement of the condition?
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
- 4. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY Pcca 20
. AND RADIOLOGICAL CONTROL QUESTION 4.13 (1.50)
Briefly explain the REASON for each of the following cautions in Integrated Plant Operating Instruction No. 4 (Shutdown):
- a. Insert the SRM detectors one at a time,
- b. The reactor should not remain in a hot shutdown condition longer than necessary.
- c. Maintain RPV level below +211 inches.
l l
l l
(***** END OF CATEGORY 4 *****)
(********** END OF EXAMINATION **********)
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/ Constant i
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Master Manual 04 Conuel Lies Power 60 - }
(percent) Typics: Low Po'e' Flow Contret Lime ~
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- Minimum Power Line 30 _
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Normal W M*P
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EQUATION SHEET r
f = as v = s/t 2 Cycle efficiency = Ne ok e) -
w = ms e = v,t + hac K = uc* -
a = (v, - v ) / t ,
- at sg = guy1 ,, , ,
, ,'c A . xy A = 4, yg = agh ,= efg a = la 2/cg = 0.693/ g ,
W = v&P-t%(eff) = (t,)(q) . . . .
AE = 931&m .
(t,g+e) b k
- k AT p - ,
I = I,e ..
" IAAT I = I,e*88 .
~Pur = Ug E -
I=I o10 "
m(t), tyL = 1.3/u y = y* to
,= , .t/T -
ivi. 0.9u.
o . ,
SUR = 26.06/T ,
T = 1.44 DT SCR = S/(1 - Egg) f1'grh o CR, = S/(1 - Edh)
SUR = 26 g,,
I ~
eff)1 " 2 CI ~~Inff)'2 T = W /o ) + [(i 's)/1gg] e I
t = t*/ (, . y; M = 1/(1 - Egg) - C1./CEo ,
T = G - p M 1,ggo y ,(g ,gdf)0lII ~ Inff)1 8*E sff-1)/K,gg = R ,gg4eff SDM = (1 - Edf)IId*
p= [1*/TE;gg 3 + [I/(1 + AggT )] g* - 1 x 10 -3 seconds
~I
,=,I(V/(3 x 10 0) Ag g A? 0.1 seconds E = Me -
. Idgg=122 WATER PARAMETERS Idg =Id 22 2
1 gal. = 8.345 1ha R/hr = (0.5 CE)/d g,,t,,,)
1 sal. = 3.78 liters R/hr = 6 CE/d (feet) -
1 gg = 7.48 gal. MTSCE11ANEOUS CONVERSIONS .
3 Density = 62.4 lbs/ft 1 Curie = 3.7 x 1010,,,
3 Bensity = 1 ge/cm 1 kg = 2.21 1ha Best of vaiorization = 970 ttu/lba 1 by = 2.54 x 103STC/hr gaat of fusica = 144 Stu/lba 1 W = 3.41 x 106 Bes/hr 1 Ata = 14.7 Psi = 19.9 in. I's. 1 Stu = 778 f t-lbf 2 g inch = 2.54 cm 1 ft. N 20 = 0.4333 lbf/in
'T = 9/5'c + 32
- C = 3/3 (87 32)
. - - . . . _ , - , . , - . - - - . . . , - - ~ .-.,,--..,_,.,_....,.,,._,---..._._,-_-,.,.,__,,_---.,.-n.. . . . _ , _ , . _ . _ , . . . _ _ _ , ,
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- r svorvALvEs 8 CLOsEO
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i i Pressure Control Units i
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1
1*. PRINCIPfER OF NUCfJAR POWER PLANT OPERATION. Pcco 21 THERMODYNAMICS. MAT TRANSra.x AND FLUID FLOW ANSWER 1.01 (2.50)
- a. Flow orificing er G, . , , ,'[. ., h; (0.5)
- b. Higher bundle power causes increased voiding and therefore increased resistance to coolant flow [1.0], if no orificing, high power (central) bundles would be starved of cooling, while more coolant flow would be diverted to lower power (peripheral) bundles [1.0]. (2.0)
REFERENCE DAEC - System Description A-3, p. 14 and 15 General Electric - Heat Transfer and Fluid Flow. PP. 8-45 ANSWEF. 1.02 (1.00) b REFERENCE E! HATCH- General Electric, Reactor Theory, Chapter 3, page 3-73 GE DAEC - Lesson Plan, RT-19b, Prompt and Delayed Neutrons, II, pp.19h-8 and 19b-9.
l ANSWEF. 1.03 (1.00) c REFERENCE DAEC - Reactor Theory, Page 8-7 l
l l
t
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
1: PRINCIPLES OF NUCLRAR POWER PLANT OPERATION. Pcca 22 THERMODYNAMICS. HEAT TRANS m AND FLUID FLOW ANSWER 1.04 (2.00)
- a. Decreases
- b. Increases
- c. Increases (due to lower friction losses)
- d. Decreases (4 8 0.5 ea.)
REFERENCE DAEC - Lesson Plan, Fluid Flow, Section 4, Pg. 4-11 ANSWER 1.05 (2.00)
- a. Core flow increases due to natural convection driving head [0.5]
but then stabilizes due to increased pressure drop [0.5] from voiding and friction loss (0.5]. (1.5)
- b. Reduction of two phase (or voiding) losses. (0.5)
REFERENCE DAEC - Beat Transfer and Fluid Flow. P. 3-3 DAEC - System Description A-2, P. 33 1
.1
.i
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
I t
-s - . - . . .__ -_ _ ,,., _.- . . . . , . . . . - - , . , , . - . _ , . . _ , _ , , _ _ _ . . _ _ _ _ _ _ _ - - - .y._-.__,
,1 : PRINCIPLES OF NUCLFAR POWER PLANT OPERATION. Pccs 23 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWER 1.06 (3.00)
- a. 1. Void Coefficient is first.
- 2. The opening of SRV causes a system pressure decrease and an increase in voids.
- 3. Negative reactivity.
- b. 1. Fuel temperature coefficient is first.
- 2. The rapid addition of positive reactivity due to the rod removal caused power to increase which increases fuel temperature.
- 3. Negative reactivity
- c. 1. Moderator temperature coefficient is first.
- 2. Removal of feed heating causes a decrease in feed water temperature.
- 3. Positive reactivity.
(9 0 0.33 = 3.0)
REFERENCE Reactor theory, chapters 23, 24, and 26 DAEC - Lesson Plan, Reactor Theory, R.T. 23, 24, and 26.
l l
ANSWER 1.07 (1.50) a) Nitrogen 16 (0.5).
b) produced by the activation of oxygen in the reactor water (0.5).
c) N-16 decays with a short half life (7 second) so it is esed#p decayedvtr th; tie- it 1:1.es th; hehil (0.;)::
d5 (k. de r chi J -il. o //jo., JJbr gig. (0. 5')
e
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
1 .- PRINCIPfTR OF NUCr.rAR POWER PLANT OPERATION. Pace 24 THERMODYNAMICS. MAT TRANStr_x AND FLUID FLOW REFERENCE Vermont Yankee Nuclear Power Corporation, SCRO-02-010, Reactor Theory DAEC - CHECK AT FACILITY ANSWER 1.08 (3.00)
- a. Incrcsse (0.25). As the moderator density decreases, the neutrons travel a greater distance and are more likely to interact with a control rod (0.75).
- b. Decrease (0.25). The voids depress the thermal neutron flux which in turn decreases the rod worth (0.75).
- c. Increase (0.25). Rod worth decreases as void content increases.
A slight decrease in void content means the thermal flux is not quite as depressed and control rod worth subsequently increases (0.75). (in o,. een py ($ ><g) by a.4) gg,)
REFERENCE GE Reactor Theory Review, pg 37 DAEC - Lesson Plan RT-27, Control Rod Worth, pp 27-11, 27-10, 27-13.
f ANSWER 1.09 (2.00)
- a. Cycle efficiency would be increased by a decrease in subcooling.
As less heat is rejected to the condenser, the returning condensate requires less reactor heat to produce steam.
Therefore cycle efficiency will increase. (1.5)
- b. By controlling the temperature and flow of the cooling water to the condenser, the operator can directly affect the overall cycle efficiency. (0.5)
(Mll ee<re q r" *V~Yw sp tw of &c uaI skesllq )
REFERENCE Standard thermodynamic principles DAEC - Lesson Plan, Thermodynamics, Section 16, Plant Effeciency, pp 16-5
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
Poca 25 j 1e PRINCIPT.RM OF NUCf.KAR POWER PLANT OPERATION. ;
THERMODINAMICS, HEAT TRANSFER AND FLUID FLOW ANSWER 1.10 (1.50)
Central rods [0.5] because the edge rods would radiate heat away from the fuel bundles while the central rods radiate much of their heat to other central rods [1.0]. (1.5)
REFERENCE DAEC - Heat Transfer, Pg. 15-2 ANSWER 1.11 (1.50)
FALSE (0.5)
The self-shielding effect [0.5] causes more U-238 atoms to be available for resonance absorption of neutrons at higher temperatures
[0.5], therefore, more overall neutron capture will occur at higher temperatures.
REFERENCE DAEC - Reactor Theory, Doppler Effect, RT - 25, pp. 7.
ANSWER 1.12 (2.00)
The downpower transient should take longer to stabilize (0.5). On a downpower transient, the rate of power change is limited to the rate of decay of the longest lived delayed neutron precursors (1.5).
REFERENCE Reactor Theory Review, pg 21 DAEC - Lesson Plan, RT -21, Use of Period Equation, pp. 21-11.
(***** CATEGORY 1 CON"INUED ON NEIT PAGE *****)
- 1. PRINCIPLES OF NUCTFAR POWER PLAST OPERATION.
Pocs 26
- THERMODYNAMICS HEAT TRANSFu AND FLUID FLOW AN5UER 1.13 (2.00)
Solving P = (Po)eE(t/T) [0.5 pts), where P = 2Po [0.5 pts) 2Po = (Po)eE(t/T) 2 = eE(t/T)
In 2 = t/T [0.5)
.693 = t/T T = t/.693 = t (1/.693) = 1.443: [0.5)
REFERENCE Reactor Theory Review, pg 20 DAEC - Lesson Plan, RT - 20. Period Equation, pp. 20-6.
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(***** END OF CATEGORY 1 *****)
2.- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pra 27 SYSTEMS ANSWER 2.01 (3.50)
- b. d5tD *1.5system (C. TsD 0 ~,Yl seal suppfiels
~
ls'Y pur$.?)
- ~"*'"~ '#ge water to recirc pumps .g" " '" #
- c. Scram discharge and instrument volumes drain to radwaste.@- 1so vents from the drive water filters go to liquid radwastept)/ua /q 5% #4 S.-/
Siss sp M ] a* r. h t.j-.J eu%4~ (4.0)
- d. CRD GEAR box, pump bearings and seals are cooled by RBCCW. (It is sufficient to say pump is cooled.) (0.5)
REFERENCE DAEC - Lesson plan A-1, pages 56 thr 58.
ANSWER 2.02 (1.00)
- 1. Reactor scram signal (0.5)
- 2. Recirculation pump trip signal (0.5)
REFERENCE DAEC - Lesson plan D-9, page 31.
ANSWER 2.03 (3.00)
- a. 1. Low lube oil pressure
- 2. Jacket coolant pressure low
- 3. High crankcase pressure
- 4. Jacket coolant temperature high
- 5. Diesel engine start failure
- 6. Diesel engine over-speed
[5 required 0 0.5 each] (2.5)
- b. Diesel engine overspeed (0.5)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
~, ,
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Poca 28 SYSTE"J5 REFERENCE DAEC - System Description, G-2, p. 33 and 34.
ANSWER 2.04 (2.00)
- 1. Turbine trip
- 2. Turbine trip
- 3. Turbine trip ( A #ct ecsd oris[ JiL/E h., /hle/)
- 4. Isolation [4 9 0.5 each) (2.0)
REFERENCE DAEC - System Description C-3, pp. 26 and 28.
ANSWER 2.05 (2.00)
- a. The idle loop jet pump flow signal is subtracted from the operating loop jet pump flow signal. (1.0)
- b. Flow being measured by the idle loop jet pump is backflow from the operating loop. (1.0)
I REFERENCE DAE0 - System Description A-2, P. 64 l
ANSWER 2.06 (1.00)
Puss.s A & C - 4160 ESSENTIAL BUS 1A3 Pumps B & D - 4160 ESSENTIAL BUS 1A4 (2 at 0.5 ea = 1.0)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
- - - - - - - - - - - . - - w-- , ,-----.--.-,---%. , - - - - - - - - - - , - , - - - - - - - - - . _ - - - - - - - , . - . - , , , - - - ~ _ - - - - , c we, , .- ~--,-------.--,-r-
~. .
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paco 29 i
SYSTEMS i
\
REFERENCE DAEC - System Description C-1, RHR, pp 48 ;
ANSWER 2.07 (2.00)
- 2. RWCU High Diff. Flow e 40 spm after 15 see time delay.
- 3. RWCU Area High Temp. 9 130 deg. F.
- 4. RWCU Area Diff. High Temp. at 14 deg. F. dT above 100% power operation ambient temp.
- 5. Non Regen. HX Outlet Temp. 9 140 deg. F.
- 6. Initiation of SBLC. (4 0 0.5 ea = 2.0)
REFERENCE DAEC System Description B-4, RWCU, pp 17 & 18 ANSWER 2.08 (2.00)
A. STOP B. STOP C. -414:00E p.Ch*-
D. OPEN E. CLOSE F. CLOSE .
G. NO CHANGE 1 0.2 85 C. CC "- (t 9.0.%"ea = 2.0)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
. , _ - , _,,.,___-,.__y, _ . , , _ , , _ . _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ . , , , , - . _ _ _ . . , _ _ , , . . . . . _ . , . - . _ , - , _ . , . --.._m_.m., . _ . _ _ - m,- -__ --
Pega 30 ;
2/ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS I
REFERENCE DAEC - Lesson plan C-1 and F-6 ANSWER 2.03 (1.50)
- a. Pressure in reactor vessel <35#.(0.5) The inboard MSIV in the associated line is closed.(0.5)
- b. Boil cff any condensate that collects in the line. (0.5)
REFERENCE DAEC - Lesson plan A-6, page 43 ANSWER 2.10 (3.00)
- n. 1. No connections low in the pool
- 2. Return lines to pool have vac. breakers.
- 3. Liner drains are provided for seam weld monitoring (Any 2 4 1.0 ea = 2.0)
- b. Preven,s damage to filter demin resin. (1.0)
- REFEREN.'E i
DAEC Syste:n Description , B-1. Fuel Fool Cooling and Cleanup, pp 10
- i i
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
i i
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcsa 31 SYSTEMS ANSWER 2.11 (2.00)
- 1. RPV W/L Low-Low-Low G 46.5"
- 2. RPV W/L Low 9 170" .
- 4. 120 sec. time delay timed out.
(0.25 for parameter)
(0.25 for setpoint)
(4 0 0.5 ea = 2.0)
REFERENCE 4
DAEC System Description A-7 ADS, pp 13 ANSWER 2.12 (2.00)
- 1. RCIC pump turbine
- 2. HPC.I pump turbine
- 3. SJAE's
- 4. Moisture separator / reheater
- 5. Gland seals
- 6. Turbine building sampling system
- 7. Off gas preheater ,
( any 6 0 0.333 ea = 2.0 )
( 8. Ofl y*> j et ca., pean n REFERENCE DAEC - Lesson plan A-6, page 10.
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(***** END OF CATEGORY 2 *****)
,3 . INSTRUMENTS AND CONTROLS Pace 32 IO ANSWER 3.01
_ ( ? . *r E y -
- . .'.11 ,_ ,. .t..t 12..;di +-1y e -
- b. A and B pumps start 10 seconds after i.udo, diesels
+ -o..tcr: :::: e n l i n e .:. 0.075
[Ew.S2]=
C and D pumps start 15 seconds after ndt die els vu lin .- [0.0753' h 5] g (s.0)
- -_ :_tv . w.
" udw % 452 'b '" W E ' m ag ... ,
-...a. i s
.n. % Jaikuveth, is "Lr
~..
~
~ --- ' ~~ '
REFERENCE 32.
DAEC - System Description C-1, P.4 0 .'."D ??.- .
ANSWER 3.02 (2.00)
- 1. Initiate HPCI
- 2. Initiate RCIC
- 3. Recirculation Pumps Trip (ATWS)
- 4. PCIS Group 8 Isolation
- 5. LPCI loop selection (any 4 0 0.5 ea = 2.0)
REFERENCE DAEC - System Iiescription A-5, P. 27 ,
I ANSWER 3.03 (1.50)
- 2. Input to feedwater control system
- 3. Control room indication.
- 4. RWM input (any 3 0 0.5 ea = 1.5)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
A_ _ _ _ _ _ _ . _ . _ _ _ - _ . _ . . _ _ _ _ _ _ _ . _ . _ . . _ _ _ _ _ _ . . . _ _ . _ . _ _
Pcco 33 3.- INSTRUMENTS AND CONTROLS REFERW CE DAEC - Lesson plan A-6, page 48 & 49. System Description I-10 pg 8 ANSWER 3.04 (1.00) d REFERENCE DAEC - Lesson plan E-11, page 3.
~
l ANSWER 3.05 (1.00) 1 b
?
REFERENC
- IAEC - Lesson plan D-15, page 11.
_=
ANSWER 3.06 (1.50)
- a. INCREASE
- b. DECREASE
[3 0 0.5 each) (1.5)
- 3. INSTRUMENTS AND CONTROLS Pc2 34 ANSWER 3.07 (2.00)
RBM Channel A -
Normal = APRM "A" Alt. = APRM "E" RBH Channel B -
Normal = APRM "D" Alt. = APRM "F" (4 0 0.5 ea = 2.0)
REFERENCE DAEC System Descrip[ tion I-5, Rod Block Monitoring System, pp 34 ANSWER 3.08 (0.50)
One Out of Two Taken Twice REFERENCE DAEC System Description I-12 , PRM. pp. 13 ANSWER 3.09 (2.00)
- a. (Loss of fully open indication of the discharge valve removes the bypass of the 20% speed limiter.) Runback to 205. The "B" (1.0) recire pump speed will decrease to 20%
- b. Both recirc pumps will decrease speed to 45%. (1.0)
REFERENCE DAEC - System Description A-2, Figure 19.
I
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3.- INSTRUMENTS AND CONTROLS Peas 35 I
. l 1
, ANSWER 3.10 (2.50) I
- a. New reading on Range 7 is 2.5 [0.53 ,
No auto actions [0.5] (1.0) j I
- b. New reading on Range 5 is 39 [0.5]
IRM high rod block and HI-BI half-scram will be in [1.03 (1.5)
,.4,% L..J..n.u
( c.* . . .n. ,,,1I L e , , , ., . A, ,. ,,);.7 Jh-REFERENCE r* A y ,h., W rnf., 4"7< )
DAEC - System Description I-2, PP. 13, 24-25 ANSWER 3.11 (1.50) 4 1. Ke71ock switch in bypass.
- 2. Turbine stop valves closed.
- 3. Mode switch not in Run. (3 0 0.5 ea = 1.5)
REFERENCE DAEC - System Description, A-6, Main Steam, page 36.
ANSWER 3.12 (2.00)
- 1. APRM channel mode switch not in operate.
- 2. APRM module (circuit card) removed 3.LPRMinputcountcircuitindicatestoofewLPRMsarebeingaveraged[,r
(< L11M; Ar /M m Ef f nJ<9sf40 F- MVSAYte)
- 4. Less than two inputs are available from any LPRM level in the core.
(4 4 0.5 ea = 2.0) l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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- 3. INSTRUMII'NTS AND CONTROLS Pcca 36 REFERENCE DAEC - Lesson plan, page 23.
ANSWER 3.13 (2.00)
NR GEMAC (0.5) Pressure (0.5)
WR Yarwar (0.5) Temperature (0.5)
. REFERENCE t
DAEC - System Description, A-5. RPV Instrumentation, page 33, Table 2.
ANSWER 3.14 (3.03)
A = pressure regulator A cr- (b /~/e3 [ ,.sr* ['y*/ de
/
B = Steam Throttle Pressure or [co w + gaar,-p+ /7/*"'O/2 C = Pressure Set D = Pressure Regulator B er A<,-d M d7 [ * *r* l '"'
E = Bigh Value Gate F = Low Value Gate (6 0 0.5 ea = 3.0)
REFERENCE DAEC System Description D-11 EHC, Fig 4 ANSWER 3.15 (1.00)
Indicates a fault in the tripping circuitf'oy /ess o[ /JS Vd'. g /.5s o8 C<~iral ps sw.
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- 3. INSTRUMENTE AND CONTROLS Pace 37
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REFERENCE i l
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4.* PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pace 38 AND RADIOLOGICAL CONTROL AN5WER 4.01 (2.50)
- a. 1. Stop affected work if a personnel or safety hazard exir .
- 2. Make a duplicate tag.
- 3. Note " duplicate tag issued" next to the pertinent entry on the Equipment Tagging Form. (3 9 E.5 ea = 1.5)
- b. Hold Card is used to safeguard human life [0.5], while a Warning Tag is used for operatinal reasons where human life is not endangered [0.5]. (1.0)
REFERENCE DAEC - Tagout Procedure, PP. I and 15.
ANSWER 4.02 (1.00) b REFERENCE DAEC - IPOI volume C-2.0, page II.B.2. (EP-13) l l
l ANSWER 4.93 (2.50)
- 1. RPV W/L below +170"
- 2. RPV Pressure above 1955 psig
- 3. Drywell Pressure above 2.0 psig
- 4. A Condition which requires MSIV isolation
- 5. A Condition which requires a reactor scram and reactor power remains above 54 or cannot be determined.
(5 e f.5 ea = 2.5)
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Pcsa 39 d? PRO <xnuxE5 - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL REFERENCE DAEC, EOP-1, RPV CONTROL, pp 1 of 80.
ANSWER 4.04 (2.50)
- 1. Torus Water Temp. above 95 deg. F.
- 2. Drywell temperature above 150 deg. F.
- 3. Drywell pressure above 2.0 psig.
- 4. Torus water level above 60%.
- 5. Torur., water level below the minimum level allowed for the existing drywell to torus dP.(Graph 12 Page 2)
(5 0 0.5 ea = 2.5).
REFERENCE DAEC EOP-2 , PRIMARY CONTAINMENT CONTROL, pp. 1 of 28 ANSWER 4.d5 (1.00)
The third day on a thirty day LCO. (1.0)
REFERENCE DAEC - Shift Organization Operation and Turnover, rev 1. Page 9 procedure number 1410.1 ANSWER 4.06 (1.00)
Observe each IRH channel increases by a factor of three from the (1.0) initial reading prior to the SRM count rate exceeding 10E5 cys.
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- . . - - - - - - - - - - , - . - - , , . , . , , , . .,__,c,-,--_-.-- , . - - - . . , . . . - - . - . . . - - _ , . , . . . . . . . - - . _ - . - - - , - - - - - - - , - - - - - , . . . - - . - - - .
e 4.. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pasa 40 AND RADIOLOGICAL CONTROL REFERENCE DAEC - IPOI 2, rev. O, page 17 ANSWER 4.97 (1.50)
- 1. Verify all operating APRM recorders are indicating between 5 and 12 % reactor power.
- 2. Verify no APRM DNSC lights are ON.
- 3. Verify the CHANNEL A[B] MAIN STEAM LINE LO PRESSURE annunciators are RESET.
(any 3 0 0.5 ea = 1.5)
REFERENCE DAEC - IPOI 2, rev. O, page 27.
ANSWEF. 4.08 (2.50) .
- a. Prompt removal of heat from the primary containment (0.5) and minimizes suppression pool heatup(0.5) (1.0)
- b. Avoid cycling of turbine exhaust check valve (0.5), assure i adequate turbine lube oil pressure (0.50). (1.0)
- c. Assures SBLC pump availability by protecting pumps should it be needed again. (0.5) l REFERENCE I
l General Electric - Emergency Operating Procedure Fundamentals DAEC - NEDC - 30796, Caution 4, pp. 4-6: Caution 11, pp. 5-5:
Caution 17, pp. 5-13.
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? Pcca 41 4.. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL ANSWER 4.09 (2.EO) lo. 5)
Note the clearance and leave number "Init."
the adjacent. of the column tag in theblank. ~ 44,8/A "Discrep" column, t+dt)(acei j,u .// .
S.,. m (*. s)
L*' N
$6 I,Jy k e n
. , _.. _ 1o REFERENCE ,.
-7 ~g DAEC IPOI 7, rev. O, paragraph 5, step 5.1 (2), page 16.
ANSWER 4.10 (2.50)
- a. Disconnect the amphenol connections for the j directional control valves. (1.0)
- If a control rod is uncoupled.
- b. 1.
4
- 2. If a " full-in" rod cannot be moved with drive water pressure.
- 3. Position indication for " full in" and " full out" is unavailable.
- 4. Control rod is INOP. ,
- 5. Control rod has INOP accumulator.
- 6. Control rod position cannot be positively determined.
- 7. Control rod on which maintenance is being performed.
(any 2 0 0.5 ea = 1.0)
- c. Eight (0.5)
REFERENCE l DAEC - AOP 255.1, P. 11 DAEC - Tech. Specs. P. 3.3-3
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! Paga 42 p PROMtnURES - NORMAL ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL ANSWER 4.11 (2.00)
- a. 1. > 2.5 mrem / hour or 100 mrem in 5 consecutive days
- 2. > 100 mrem /hr but < 1000 mrem /hr
- 3. > 1000 mrem / hour [3 4 8.5 ea.]
- b. 150 mrem [0.53 (2.2)
REFERENCE DAEC - Bealth Physics Procedure 3106.1, P.2 DAEC - Health Physics Procedure 3102.1, P. 2 ANSWER 4.12 (2.50)
Verify automatic actions have occurred (0.5). If turbine trip has occurred, carryout turbine trip actions (0.5). If react"or scram occurred, carryout scram actions (0.5). If reactor has not scrammed, rapidly reduce reactor power using recire pumps (0.5).
Place standby SJAE unit in servive (0.5). Place Mechanical Vacuum Pump in service (0.5)
( any 5 0 0.5 ea = 2.5)
REFERENCE DAEC - Check at the facility - - IPOI C-40, Section 0, Page 0-2-3.
! ANSWER 4.13 (1.50)
I a. Simultaneous insertion may cause a scram due to electronic i noise on the IRH's.
- b. To minimise thermal stresses at the feedwater nozzles.
To prevent tripping the operating RFP, HPCI, and RCIC.
c.
[ 3 0 0.5 )
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Pcga 43 d.' PROum.uUfui.S - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL REFERENCE DAEC - IPOI-4, PP. 14-17 I
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