ML20148G483
| ML20148G483 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/19/1988 |
| From: | Moorman J, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20148G417 | List: |
| References | |
| 50-327-OL-87-01, 50-327-OL-87-1, NUDOCS 8801260519 | |
| Download: ML20148G483 (173) | |
Text
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ENCLOSURE 1 EXAMINATION REPORT 327/0L-87-01 Facility Licensee: Tennessee Valley Authority ATTN: 'Mr. S. A. White Manager of Nuclear Power 6N 38A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Facility Name: Sequoyah Nuclear Plant Facility Docket No.: S0-327, 328
- Written examinations and. operating tests were administered at Sequoyah Nuclear Plant near Soddy-Daisy, Tennessee.
' Chief Examiner: /'! b b /-/f-88
. 1 Moorman, III Date Signed Approved b : O _
/ 9/88
[ Tate Signed g . Munro, Section Chief Suninary:
Examinations on November 17-19, 1987.
Written and Operating tests were administered to five candidates; five of whom
-passed.
Based on the results described above, four of four R0's passed and one SR0 passed. ,
8801260519 880'120 PDR ADOCK 05000327 V PDR
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l REPORT DETAILS
- 1. Facility Employees Contacted:
- R. Joe Johnson, Director, Division of Nuclear Training
- C, H. Noe, Chief, Operations Training Branch
- J.-R. Walker, Assistant Operations Manager
' *:C. T. Benton, Supervisor, Simulator Training
- V. E. Keyser, Simulator Instructor
- R. C. King, Simulator Instructor
- Attended Exit Meeting 2._ Examiners:
- J. H. Moorman, III, RII S. D. Bitter, RII C. Y. Shiraki, NRC, Headquarters
- Chief Examiner
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided V. E. Keyser, with a copy of the written examination and answer key for review. The NRC Resolutions to facility comments are listed below.
- a. R0 Exam (1) Question 1.01b
.: Facility Comment:
Resolution: Comment acknowledged. Since interpretation of the question allows either increase or decrease to be accepted as a correct answer, the question has been deleted.
Question 1.21 Resolution: Comment acknowledged. Since there is no correct answer to the question as stated, the question has been deleted.
Question 1.27a Resolution: Comment accepted. The answer key has been annotated to allow +/ .05 F for interpolation.
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- Question 2.04c Resolution: Coment accepted. The answer key will be changed to accept "PRT via relief valve" as an additional correct response.
Queistion 2.06a Resolution: Coment accepted. The answer key has been changed to acce;t "Inability to reenergize the SRMs below P-6" as an additional correct response.
Question 2.09a Resolution: Comment partially accepted. The answer key will be changed to accept "Blackout with return of shutdown voltage, after a time delay" as an additional correct response.
"Control room handswitch not in the pull-to-lock position" is a restatement of the conditions given in the question and will not be accepted.
Question 2.10b Resolution: Comment accepted. The answer key has been changed to accept the answers proposed by the facility.
Question 2.12c Resolution: Coment accepted. The question has been deleted.
Question 3.01e Resolution: Comment accepted. The answer key has been changed.
Question 3.02 Resolution: Coment accepted. The answer key has been changed to accept the additional answers provided.
Question 3.05b, 3.05c Resolution: Coment accepted. The answer key has been changed to accept the more technically specific answers provided by the facility.
Question 3.06 Resolution: Coment accepted. The answer key has been changed to accept the additional answers provided by the facility.
3 Question 3.08d Resolution: Coment accepted. The answer key has been changed to accept OT delta T instead of high steam line flow SI as the the other most probable cause of the reactor trip.
Question 3.10d Resolution: Comment accepted. The answer key has been changed to accept BOTH as the correct answer.
Question 3.11a Resolution: Comment accepted. The answer key has been changed to accept the revised setpoint. The utility should update it's reference material to reflect this change.
Question 4.09a Resolution: Comment accepted. The answer key has been changed to reflect the most up-to-date information. . The utility should update it's reference material to reflect this change.
Question 4.14 Resolution: Coment acknowledged. The answer provided by the facility is an amplification of the existing answer key. The answer key will be annotated with the coment provided by the facility. ,
Question 4.22 Resolution: Comment accepted. The answer key has been changed to accept "a" as an additional correct answer. The facility should update it's reference material to reflect this change.
- b. SRO Exam Question 5.06 Resolution: Coment accepted. The answer key has been changed to require "d" as the correct answer.
Question 5.08d Resolution: Coment not accepted. TM word "initiates" clearly implies that the pressurizer begins spraying, i.e.,
prior to initiating, there was no spray flow (with the exception of spray bypass flow). Therefore, upon initiating spray, the j
pressurizer pressure drops and DNBR decreases. No change to the answer key.
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4 Question 5.12d
-Resolution: Coment not - accepted. Although most of the statements provided by the facility in the comment are true, they do no logically support the conclusion that the CCP NPSH remains the same. No change to the answer key.
Question 5.13d Resolution: Coment accepted. The correct answer should be DECREASE instead of INCREASE. Tha answer key has been revised to reflect this.
Question 5.17d Resolution: Comment not accepted. However, due to ambiquity in the definition of "80L to M0L" for this particular part of the question, the question will be deleted. The facility should update its training material concerning this item.
Question 6.05e Resolution: Coment accepted. The answer key has been changed.
Question 6.08a Resolution: Coment accepted. The answer key has been changed ,
to accept "False" as the correct answer. The facility should to correct the lesson plan from which the original response was obtained.
Question 6.14 Resolution: Coment not accepted. The lesson plan referenced in the answer key clearly states four purposes of the Ice Condenser system. Although it is true that the proposed additional responses are design features, considerations, and characteristics of the Ice Condenser system, they are not purposes. No change to the answer key.
Question 6.16 Resolution: Coment accepted. The answer key has been changed to accept the additional answers provided.
Question 7.11 Resolution: Coment accepted. The answer key has been changed to delete the 10 REM emergency exposure limit as a portion of the required response. The 1.5 total point value will be retained and redistributed over the remaining two portions of the required response. The facility should update its training material to reflect this information.
5 Question 7.12d Resolution: Coment partially accepted. The answer key has been changed to require the original answer and the additional answer provided by the facility since these are both correct as they apply to the two parts of step 13 in E-0. The point value has been redistributed.
Question 8.17 Resolution: Coment not accepted. Although the Inservice Inspection group is responsible for this particular surveillance, it is reasonable to expect a . Senior Operator to interpret all technical specifications. No change to grading.
Question 8.18 Resolution: Coment partially accepted. The answer key has been changed to require 6:00 p.m. as the correct answer instead -
of 11:15 p.m. based on the conservativeness of the facilities policy.
Question 8.20 Resolution: Coment accepted. The answer key will be changed to accept "Perfonn the actions of T.S. 3.0.3" as a correct response. The position taken by the facility in this situation, (i.e., that there are less restrictive requirements with two accumulator isolation valves out of cumnission than with only one) although not incorrect as the Technical Specifications ara currently written, is not consistent with the conservative nature of Technical Specification 3.5.1.1.
Due to the non-conservative interpretation postulated by the j facility, this item is identified as Inspector Follow-up Item 50-327/0L-71-01,
- c. Other Changes Question 5.13a, 5.13b, 5.13c Further technical review has revealed that the answer to these '
questions should be INCREASE instead of DECREASE. The answer
- key has been changed to require these as the correct responses.
L_____________________________________________._______________ ___ _____ ____
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- 4. Exit Meeting:
At the conclusion of .the site visit, the examiners met with members of your staff to discuss items pertinent to the operating exams and the examination process. The following items are discussed:
- 1. In preparation for the simulator examination, the examiners attempted to use Licensee Event Reports that had been generated as a result of incidents as the Sequoyah Nuclear Plant in the development of-simulator exam scenarios. Of ten LERs that could have been used to develop the ' scenarios, not one could be modeled on the Sequoyah simulator.
Additionally, it was noted that the Sequoyah simulator could not model a feed line break or a leak in the RWST.
- 2. The reference material provided to the examiners for preparation of the written exam was noted to be deficient and contradictory in some areas. Specific examples were provided to facility representatives.
The following generic weaknesses were noted among the reactor operator applicants as a result of the written examination and were not discussed at the exit meeting:
- 1. Question 2.05a Applicants did not know that only the Motor Driven Auxiliary Feedwater Pumps would start automatically.
- 2. Question 2.07a Applicants did not know the location of the 'k' phase regulating transformer.
- 3. Question 2.08 Applicants did not know the order of injection of the ECCS components in the event of a large break LOCA.
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There were no generic weaknesses noted during the oral and simulator examinations.
Four of 32 (12.5%) of the changes made to the written examination answer. key were due to inadequate of insufficient training material provided by your staff to the NRC for examination development.
The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was noted and appreciated.
The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.
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' s U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _SEQUOYAH_lh2____________ J REACTOR TYPE: _PWR-WEQ4________________
DATE ADMINISTERED: _@?!!1/1?________________
EXAMINER: _SH188K13_C1 _____________
CANDIDATE: bO bSTEIk INSIRUCI!9NS_IQ_C8L4D]D6IEi I Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__Y66UE_ _IDI6L ___SCOBE___ _y@(UE__ ______________Q@ LEG 08Y_____________
2e.s
- 29:x:__ _2Et99 ___________ ________ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 19.33
.I2:2i~__ _29399 ___________ ________ 2.
_ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
_}9199__ _2Et99 ___________ ________ 3. INSTRUMENTS AND CONTROLS
_}9199__ .29399 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 111.83 i x:::__ ___________ ________% Totals Final Orade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature I
8- 4 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the ade.inistration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and 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 l room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil gnly 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 gn gne side of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face 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 sathematical 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 tne examination has been completed.
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- 19. 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,
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions,
- d. Leave the examination area, as defaned by the examiner. If after 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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1 PRINCIPLES OF NUCLEAR POSER PLANT OPERATION 3 PAGE 2 ISE6dODyN6b]C@2_Og81_lB0N@[g8_@Np_ELUlp_[ LOW .
' s QUESTION 1.01 (1.50)
DIFFERENTIAL rod worth varies as a function of conditions in the core.
Considering each case below independently, does the DIFFERENTIAL rod worth INCREASE, DECREASE, or REMAIN THE SAME?
- a. Temperature decreases, with rod position and baron concentration g held constant.
}d',hT The rods are withdrawn from 150 steps to maintain temperature constant, with boron held constant.
- c. Boron concentration is diluted to maintain temperature constant, with rod position held constant.
QUESTION i.02 (1.00)
If the equilibrium count rate in a subtritical reactor increases by a factor of EIGHT from 100 counts per second to 800 counts per second due to a reactivity addition, by how much has the margin to criticality decreased?
QUESTION 1.03 (1.50)
Following the taking of critical oata, in the intermediate range, a stable startup rate of 0.15 DPM at 0.1% of full power is established. The reactor is at no-load Tave, and the Baron concentration is 800 ppa. The plant is below the point of adding heat.
- a. What will the reactor power be after 2 minutes?
- b. At what power level does the point of adding heat begin?
(No calculation required.)
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- 1. PRINCIPLES OF NUCLEAR PODER PLANT OPERATION 1 FAGE 3 IHE800DyggdlCS _HE@l_166NSEE8_68D_E(UID_E(OW 1
' i OUEST10N 1.04 (1.50)
A calculated Estimated Critical Position (ECP) is performed for a startup to be commenced 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a trip from 100% power.
Considering each case below independently, state whether the Estimated Critical Position (ECP) will be HIGHER, LONER, or the SAME AS the Actual Critical Position (ACP).
- a. The startup is delayed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the trip.
- b. The condenser steam dump pressure controller setpoint is i increased to just below the steam generator atmospheric steam dump controller setpoint.
- c. All steam generator levels are rapidly being raised by 5% while approaching criticality.
QUESTION 1.05 (1.00)
Which of the following statements best describes the change in moderator temperature coefficient (MTC) from BOL to EOL?
oncreuena
- a. The MTC becomes more negative due to d: ::: ; boron concentration, decreasing fission product inventory, and axial flux redistribution toward the edges of the core.
OGLeiksm b
- b. The MTC becomes more negative due to uc: :. ; baron concentration, increasing fission product inventory, and radial flux redistribution toward the edges of the core.
- c. The MTC becomes less negative due to increasing baron concentration, increasing fission product inventory, and axial flux redistribution toward the edges of the core,
- d. The MTC becomes less negative due to decreasing boron concentration, decreasing fission product inventory, and axial flux redistribution toward the edges of the core.
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' i QUESTION 1.06 (1.00)
Which of the following is the MOSI ACCURATE definition of the term reactivity?
- a. The rate of change of reactor power in neutrons per second
- b. The ratio between the populations of two successive neutron generations
- c. The difference between critical and all control rods withdrawn for a given core condition
- d. The measure of a reactor's departure from criticality 00EST10N 1.07 (1.00)
In a subtritical reactor, Xeff is increased from 0.880 to 0.965. Which one of the following is the amount of reactivity that was added to the core?
- a. 0.086 de'ta k/k
- b. 0.100 delta k/k
- c. 0.126 delta k/k
- d. 0.220 delta k/k QUESTION 1.08 (1.00)
During a reactor startup under xenon-free conditions, rod withdrawal is stopped at the -0.02% delta k/k position and the count rate is allowed to stabilize. In regard to the response of the count rate in the hour after stabilization, which one of the following statements is correct? (Assume N0 further operator actions are taken.)
- a. Count rate will remain essentially constant
- b. Count rate will rapidly decrease to its pre-startup level
- c. Count rate will slowly decrease because it is subtritical
- d. Count rate will slowly increase due to long-lived delayed neutrons l
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' s OUESTION 1.09 (1.00)
A. reactor is initially subtritical with a Keff of 0.95 and a source range count rate of 200 counts per second (CPS). Which one of the following is the final count rate if control rods are withdrawn to add 0.027 delta k/k reactivity?
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- a. 250 CPS l
- b. 300 CPS
- c. 350 CPS
- d. 400 CPS QUESTION 1.10 (1.00)
Which one of the following statements BEST explains changes in control rod worth as xenon concentration increases?
- a. The xenon increase has no effect on control rod worth because worth is a function of position
- b. The xenon increase will decrease control rod worth because the thermal neutrons are absorbed by the poison
- c. The xenon increase will increase the thermal utilization, and control rod worth increases
- d. The xenon increase does not affect control rod worth until the xenon peaks; after the peak, the control rod worth increases DUESTION 1.11 (1.00)
Which one of the following statements BEST explains why xenon concentration will increase after a power decrease?
- a. The migration length changes as moderator density increases
- b. Fewer thermal neutrons are available to burn out xenon
- c. The amount of xenon from fission does not decrease until the reactor is on a positive period
- d. The amount of Xe-135 produced from Sm-149 increases for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after power changes
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1 __Egl8CIE(Ep.gE_yyC(E@B_E9WE8_P(681_gfE861]Qy, PAGE 6 ISEBdQQyU8DICg3.dE81_IBBUSEEB_ bug _E(glg_E(gW _
0UESTION 1.12 (1.00)
A reactor trips from full power, equilibrium xenon conditions. Twenty f our hours later, the reactor is brought critical and power level is maintained at 5 x 10 -11 amps for several hours. Which one of the following statements is correct concerning control rod motion?
- a. Rods will have to be insertrd, since xenon will approximately follow its normal decay curve
- b. Rods will remain approximately as is, as the xenon establishes its equilibrium value for this power level
- c. Rods will have to be rapidly inserted, since the critical reactor will cause a high rate of xenon burnout
- d. Rods will have to be withdrawn, due to xenon build in QUESTION 1.13 (1.00)
Which of the following statements presents the two primary methods for Xe-135 production i'n the core?.
- a. Decay of fission products and activation of U-233
- b. Decay of Se-149 and activation of oxygen
- c. Decay of iodine and fission
Which one of the following will cause the axial flux difference to become more positive (less negati re)?
- a. Power increase with power defect compensated for by dilution only
- b. Power increase with power defect compensated for by rod withdrawal only
- c. Buildup of xenon in top portion of core
- d. Barnup of xenon in bottom portion of core
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s.- 1
l 1. PRINCIPLES OF NUCLEAR P0 DER PLANT' OPERATION 3 PAGE 7-ISE859pyd801CS3_UE81_IR8dSEE8_80p_E(Ulp_ELQy QUESTION 1.15 (1.00)
The reactor ic critical at 10,000 counts per second (cps) when a steam generator PORV fails open. Assuming BOL conditions, no rod motion, and no reactor trip, choose the answer below that best describes the values of
-Tayg and nuclear power for the resulting new steady state. (POAH) = Point Of Adding Hert)
- a. Final Tavg greater than initial Tavg. Final power.at POAH.
- b. Final Tavg greater than initial Tavg. Final power above PDAH.
- c. Final Tavg less than initial Tavg. Final power at POAH.
- d. Final Tavg less than initial Tavg. Final power above POAH.
QUESTION 1.16 (1.00)
With the plant operating at 95% power and all systems in a normal / auto configuration, the operator barates 100 pcm. Shutdown margin will
- a. Increase
- b. Increase until rods move
- c. Decrease
- d. Decrease until rods move OUESTION 1.17 (2.00)
Indicate whether the following will cause the poNer ranga instrument to be indicating HIGHER, LOWER, or the SAME AS actual power, if the instrument has been adjusted to 100% based on a calculated calorimetric,
- b. The reactor coolant pump heat input used in the calorimetric was ositted
- c. The steam flow used in the calorimetric was lower than actual i
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ISE8099yh85]9S 3_UE&l_IB8NSEEB_809_E(919_EL90 QUESTION 1.18- (1.00)
Which of the following statements correctly describes the power coefficient?
- a. At BOL',' the MTC is the largest contributor to the power coefficient due to the relatively high MTC caused by the low boron concentration
- b. The negative effect of the MTC outweighs the slightly positive effect of the FTC to cause the power coefficient to become more negative from BOL to EOL
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- c. At BOL, the FTC and MTC make roughly the same contribution to the power coefficient w4th the MTC becoming greatly more negative over core life
- d. The FTC is the largest contritator to the power coefficient at E0L due to the relatively small MTC caused by the low boron concentration QUESTION I 19 (1.00)
Select one statement from the following that correctly describes the reason density compensation of the main steam line flow measurement is necessary.
- a. Differer.tial pressure across the orifice is proportional to the volumetric flow rate. Volumetric flow rate is compensated with the fluid density to provide mass flow rate,
- b. Differential pressure across the orifice is inversely proportional to the volumetric flow rate. Volumetric flow rate is compensated with the fluid temperature to provide mass flow rate.
- c. The tem 9erature of the steam lines is proportional to the volumetric flow rate. Volumetric flow rate is compensated with I the fluid temperature to provide mass flow rate,
- d. The temperature of the steam lines is inversely proportional to the volumetric flow rate. Volumetric flow rate is compensated with the fluid density to provide mass flow rate.
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l L 1s_.EBIU91ELES_QE_dVQLE88 EggE8_E(8Bl_Q[58@llgN 3 PAGE 9 1: INE8Dgply80igS 3_SE8I 1880SEE8.80p_E(Ulp_E(pd l . ,
QUESTION 1.20 (1.00)
A recently calibrated pressure gauge sensing in the same area as a temperature instrument indicates 350 psig. Assuming saturated conditians, wh$c5 one of the following will the temperature instrument show?
- a. 420 degrees F
- b. 427 degrees F
- c. 431 degrees F
- d. 435 degrees F QUESTION 1.21 (1.00)
Which one of the following pa*ameters has NO effect on the , positive suction head (NPSH) of a centrifugal pump?
- a. The height of the coluan of water ab he eye of the tump
- b. TheamountofsubcqqgcI$ he column of water above the eye of the pump
- c. Pressura a the tank that is the water supply to the pump
, Number of pumps operating in parallel OUESTION 1.22 (1.00)
Operating a positive displacement pump with insufficient net positivr!
suction head will cause which one of the following to occur?
- a. Slip
- b. Decreased pump speed
- c. Viscosity loss
- d. Vapor binding
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- 1. ' PRINCIPLES OF NUCLEAR POSER PLANT OPERATION 3 PAGE 10
-ISEBdgpIN8dlCS 3_bE81_IB8HSEE8,8NQ_ELUlp_E(gy QUESTION 1.23 (1.00)
Which one of the f ollowing is the correct value f or the cooldown rate of a saturated steam system, if the initial pressure is 985 psig and the pressure 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> later is 385 psig?
- a. 80 degrees F/ hour
- b. 100 degrees F/ hour
- c. 120 degrees F/ hour
- d. 125 degrees F/ hour 00ESTION 1.24 (1.00)
Which one of the following is the correct order of heat transfer on the boiling curve, from the MOS? EFFICIENT method to the LEAST EFFICIENT method?
- a. Film boiling, transition boiling, nucleate boiling
- b. Nucleate boiling, film boiling, transition boiling
- c. Transitica boiling, nucleate boiling, film boiling
- d. Nucleate boiling, transition boiling, file boiling 00ESTION 1.25 (1.00)
Which of the following MOST ACCURATELY defines the tern NUCLEATE BOILING?
- a. Boiling that results in the bulk fluid temperature reaching saturated conditions
- b. Boiling that results in a thin layer of steam at the heated surface
- c. Boiling that occurs when small bubbles are forced at the heated surface and move off into the liquid
- d. Boiling that occurs when the critical heat flux is reached
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Iz__EB1Uc1P(ES_QE_ygg(898_899E8,P(801_gPE8811993 PAGE 11 IUE8099108djCS 2UE81_1888SBE8_80p_E(919_E(QB e
QUESTION 1.26 (1.00)
Which one of the following actions would help, rather than hinder, natural circulation?
- a. Lowering steam generator level
- b. Lowering RCS pressure
- c. Increasing RCS temperature
- d. Increasing pressurizer level QUESTION 1.27 (1.50)
Answer the following questions in reference to subcooling margin of the plant.
- a. What is the subcooling margin of the plant if the following conditions exist:
That = 587 F Tavg = 572 F Tcold = 557F Pp:r = 2235 psig Psg = 1033 psig
- b. If power is raised from 50% to 100%, why does the subcooling margin decrease?
(***** END OF CATEGORY 01 **+**)
d h 3A .Pg8Bl_pg!!@U jhgguplB9_g6 Eely _80p_EUgRGEUgy_Sy@Tgh$- PAGE 12
-QUESTION 2.01 (2.25)
For each-independent case below, explain the operation of the steam, dump system and how it affects RCS temperature.. Include in your answer the approximate final RCS Tavg. If a change in steam generator pressure results, speLify the new value. Assume all systems are normal except as stated and that no operator action is taken. .Specify any function or control that is blocked. Assume the pressure setpoints are for the steam dump system.
- a. The_ normal setpoint on the. steam dump system is reduced by 70 psi while in hot standby awaiting reactor startup.
- b. The steam dump control switch is taken to 0FF while stable at 5%
reactor power.
QUESTION 2.02 (1.00)
Complete the following statements:
- a. "The OT delta T calculated reactor trip setpoint is designed to protect the core from . . . *
- b. "The OP delta T calculated reactor trip setpoint is designed to protect the core from . . .
- QUESTION 2.03 (1.00)
- a. What are the two interlocks that must be satisfied to open the letdown isolation valves FCV-62-69 and FCV-62-707
- b. What.is the reason for these interlocks? A QUESTION 2.04 (3.00)
Answer the following questions in reference to excess letdown.
- 4. Froe which loop (number) and from which leg (hot or cold) does excess letdown come?
- b. What fluid system cools the excess letdown heat exchanger?
- c. After passing through the excess letdown heat exchanger, excess letdown flow can go to three possible destinations. What are these destinations?
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
2i__f(@UI,0g@][U_INC(UDjdG_@@[Ely_AND_gdERGENCy_!!!!Ed! PAGE 13 QUESTION 2.05 (2.00)
The unit is operating,at 90 percent power and one of the two running main feed.. pumps trips.
What four automatic control actions involving the auxiliary feed-pump (s), the main feed pump (s), the dain feed pump turbine condenser (s) and the turbine will prevent a unit trip from low-low steam generator level? (One automatic control action per componeat.)
QUESTION 2.06 (1.50)
One power range channel is taken out of service with the reactor at 50%
power while conducting a plant shutdown. When power drops below 10%,
a P-10 permissive interlock solid state bistable switch for one of the remaining power range channels does not reset.
- a. What is the effect on the solid state protection system (SSPS) of the failure of the P-10 bistable switch to reset?
- b. How would such a failure be detected in the control room?
QUESTION 2.07 (2.50)
The rod control system has main and auxiliary power supplies,
- a. Provide a one-line diagram of the main power supply to the rod control system, showing the following components:
MG sets, 480 V unit boards; reactor trip breakers, "A' phase regulating transformer, reactor trip bypass breakers
- b. Provide a one-line diagram of the auxiliary power supply to the rod control system, showing the following components:
i Primary fused disconnect, 400 V shutdown boards, 480V/120V transformer, vital transfer switch, 120 VAC instrument power distribution panel
(***** CATEGORY 02 CONTINUED ON NEXT PAGE **+++)
)
at__EL6BI RES[@8,[UC(UQ[N@,98EEIY,@ND_EdEBQEUCY_SY@l@M@ PAGE 14 QUESTION 2.08 (1.25)
In the event of a LARGE BREAK LOCA, what is the order in which the following ECCS systems will inject cooling water?
- a. Cold leg accumulators
- b. Safety inje.cica pumps
- c. Upper head injection accumulators
- d. Residual heat removal pumps
- e. Centrifugal charging pumps via the Bli QUESTION 2.09 (2.50)
Answer the following questions in reference to the essential raw cooling water system (ERCW).
- a. One condition required for automatic start of essential raw cooling water (ERCW) pump J-A is that it must be selected for automatic start, and pump Q-A must not be running. State the other.three conditions for pump J-A to automatic start,
- b. State the basis for the ERCW pump automatic start condition given above in part a.
- c. Which header (unit 1 or unit 2), is the primary cooling source for each of the diesel generator heat exchangers?
- d. Which two ERCW headers (IA, IB, 2A,.2B) are the normal caoling water supply to Component Cooling Water System heat exchangers A, B, and C?
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
i
2c__EL6BI_pgSigy_jNCLUp]U9.38EEIL,80p_EdE8@ENCY_@Y@lEb! PAGE- 19 QUESTION 2.10 (2.50)
Answer the' f ollowing questions in ref erence to the incore thermocouple system.
- a. -With the computer available, what is the range of indication for the incore thermocouple system?
- b. Positive indication that conditions of inadequate core cooling (ICC) exist when the incere thermocouples indicate greater than
_________ degrees F.
- c. Following a loss of natural circulation, actions must be taken to reestablish ECCS flow. Explain the immediate thermocouple response as ECCS flow is restored.
- d. Assuming all incore thermocouples are functioning properly, one thermocouple is reading abnormally high relative to the others. What fuel or flow channel condition does this indicate?
QUESTION 2.11 (2.00)
During a loss of coolant accident, all automatic safety injection systems functicn properly, pressurizer level stabilizes, and RCS pressure stabilizes at about 1400 psig. What is the approximate leak rate in gallons per minute? State all assumptions and justify your answer.
QUESTION 2.12 ::. b l 33 )
In the rod control system, auctioneered high Tavg is compared with Tref to develop a signal for rod speed and indication. For the systems listed below, briefly describe how auctioneered high Tavg is used to generate control signals. No setpoints are required in your answer,
- a. Steam dump control system
- b. Pressurizer level control system
- . c:ed=2! ' : ^t 21 :y:t; L OGL6TE D )
(***** CATEGORY 02 CONTINUED ON NEXT PAGE +++**)
1 2 3,_PL891_DE}]@h_lNC(UD]d@_@8 Eely _8ND_EdE6@ENCy_@y@ led $ PAGE 16 QUESTION 2.13 (2.50)
With the plant at 50% power and all systems in automatic, the selected turbine first stage pressure transmitter f ails low. Assuming no operator action, answer the f ollowing questions in ref erence to this event.
- a. State if the rods INSERT or WITHDRAW. Briefly explain why.
- b. Does pressurizer level INCREASE, DECREASE, or REMAIN THE SAME7
- c. Does pressurizer pressure INCREASE, DECREASE, or REMAIN THE SAME?
- d. If a reactor trip occurs, what will be the most likely cause?
DUESTION 2.14 (1.00)
Which one of the following is NOT a source of water to the pressurizer relief tank?
- a. RHR pump suction and discharge relief valves
- b. Pressurizer power operated relief valves
- c. Reactor vessel flange leakoff
- d. Reactor vessel head vent system QUESTION 2.15 (1.00)
Unit load is 95% when indication is received that the number 3 heater drain tank bypass valve to the condenser has left the fuity closed position. Which one of the following describes the plant response?
- a. A turbine trip occurs
< 8 0'h
- b. A turbine runback to s .c :D. power level is initiated
- c. The main feed pumps accelerate to their high speed stops.
- d. The turbine driven auxiliary feed pump auto starts.
(***** CATEGORY 02 CONTINUED ON NEXT PAGE ****+)
-as..P(@NI,QESlC8,1NC(UD18Q,S8FEl!,@@Q_EME8@EBCY,SYSIEd@ PAGE 17 OVESTION 2.16 (1.00)
Unit 1 is in mode 5 on RHR at about 150 degrees F. During the leak test of the unit 1 RVLIS sensing lines, they are pressurized to 3000 psig.
Which one of the following correctly describes the RHR system response to this pressurization?
- a. RHR pumps suction valve 74-3 will shut when pressure reaches 700 psig
- b. RHR inlet isolation valve 74-2 will shut when pressure reaches 700 psig
- c. The operating RHR pump will trip when pressure reaches 700 psig
- d. Heat exchanger flow control valve 74-16 will go fully open when pressure reaches 700 psig QUESTION 2.17 (1.00)
Unit 1 is in mode 5 with the residual heat removal system in operation using the "B" train pump (B-B RhR pump) and RCS temperature is about 100 degrees F. RCS indicated level is 695' 6". A swapover to the "A" train pump (A-A RHR pump) is desired. Unbeknownst to the operators, RCS indicated level is incorrect, and actual level is 695' 0". If the swapover is accomplished by starting the "A" train pump prior to securing the "B" train pump, which of the following statements best describes the possible system response?
- a. RHR indicated flow is zero due to air binding of the pump (s)
- c. Containment sump isolation valve 1-FCV-63-72 starts to open causing RHR pumps suction valve 1-FCV-74-3 to close
- d. The RHR pumpts) trip due to low aump discharge pressure t
(***** END OF CATEGORY 02 *****)
3:_.lN@I69MENIS_6ND.COUI80LS PAGE 18 QUESTION 3.01 (2.00)
For each'of'the following situations, indicate whether the OT delta T/0P l
. delta T-reactor trip setpoint will INCREASE, DECREASE, or REMAIN THE SAME.
- Consider each case separately.
- a. Tavg input to the OT delta T setpoint INCREASES.
l
- b. Pressure input to the 07 delta T setpoint INCREASES.
l l c. . Delta flux (function) input to the OT delta T setpoint INCREASES.
l l d. Tavg input to the OP delta T setpoint INCREASES.
i
- e. Delta flux (function) input to the OP delta i setpoint INCREASES.
l l QUESTION 3.02 (2.00)
One of the functions of the P-4 permissive signal is to actuate a turbine trip. List four other functions of the P-4 permissive.
QUESTION 3.03 (1.75)
How would the following components respond to a loss of Control Air?
- a. During normal power operation in Mode 1. Specify FAIL CLOSE, FAIL OPEN, TRIP, or ACCELERATE.
- 1) Normal letdown valves
- 2) RCP seal return valves
- 3) Charging flow control valve
- 4) Feed regulating valves
- 5) Main feed water pumps
- 1) MPAFW pump LCV's
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
J
}n..lySIByMEUIS_8hD QOU180(S PAGE 19 l
f -QUESTION 3.04 (1.50)
For the components listed below, state how each will respond to a HI-HI Containment Pressure signal received as a result-of a LARGE BREAK LOCA.
Specify TRIP, START, CLOSE, OPEN.
- 1) Containment Spray pumps
- 2) MSIV's
- 3) MS bypass valves
- 4) Containment ventilation
- 5) Main feed regulating valves
- 6) Main feed pumps DUESTION 3.05 (2.50)
Answer the f ollowing questions concerning the loss of 120V AC vital instrunent power board 1-!!. Assume a reactor trip has not taken place.
- a. Why has automatic control of the control rods been lost?
- b. Why has automatic control of steam generator water level on the affected steam generator (s) been lost?
- c. Why must feedwater pump speed be controlled manually?
- d. Why must a centrifugal charging pump be in service and not the positive displacement charging pump?
- e. Why must control of charging flow control valve (FCV-62-93) be transferred to the auxiliary control room?
DUESTION 3.06 (1.00)
One of the parameters monitored by the Post Accident Monitoring Instrumentation is steam generator level. List four of the remaining six parameters that are also monitored by the Post Accident Monitoring Instrueentation.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE ++++*)
31 lh!IBUMEU{$ 6BD.CghlBQL@ PAGE 20 L..' .,
QUESTION ~3.07 (1.50)
-Answer'the following questions in reference to the blocking of the high steamline flow coincident with either-low-low Tavg or low steamline pressure safety injection signal
- a. What criterion aust be met to block the safety injection signal?
(Give actual setpoint, not permissive number. State coincidence if applicable)
- b. -Does the block clear automatically, or is operator action required?
- c. What criter' ion aust be met for the block to clear (or be cleared)?
(Give actual setpoint,.not permissive number. State coincidence if applicable)
QUESTION 3.08 (2.50)
With the plant at 100% reactor power, the rod control system in automatic, and no operator action, a loop Tavg instrument fails high.
- a. What type of rod motion (if any) will ensue? (MOVE IN, MOVE OUT, or NO MOTION)
- b. Will pressurizer pressure INCREASE, DECREASE, or REMAIN THE SAME7
- c. Will pressurizer level INCREASE, DECREASE, or REMAIN THE SAME7
- d. What are the two most'likely causes of a reactor trip?
(***** CATEGORY 03 CONTINUED ON NEXT PAGE ****+) l
li. 1851&QUEyl@_68D_Q0 NIB 0($ PAGE 21 QUESTION 3.09 (1.25)
Hatch the following reactor ~ trip system interlocks with the appropriate function.
- a. P-6 1. Defeats auto block of reactor ': rip on turbine trip.
- b. P-7 2. Defeats auto block of reactor trip on low RCS coolant flow in a single loop.
- c. P-8
- 3. Enables manual block of the source range reactor trip.
- d. P-9
- 4. Defeats auto block of reactor trip on low flow in more
- e. P-10 than one primary coolant loop, reactor coolant pump undervoltage and underfrequency, turbine trip, pressurizer low pressure, and pressurizer high level.
- 5. Enables manual block of reactor trip on power range (low setpoint), intermediate range, as a backup block for source range, and intermediate range rod stops (i.e. , prevents premature block of the noted functions.)
- 6. Defeats manual block preventing automatic reactuation of safety injection.
QUESTION 3.10 (1.25)
For each of the following auxiliary feed pump automatic. start signals, specify whether it applies to the motor driven auxiliary feed pumps (MDAFWP), or the turbine driven auxiliary feed pump (TDAFWP), or both.
P
- a. Low-low level in any one steam generator (2/3 level transmitters)
- b. Loss of offsite power (immediate start)
- c. Loss of both main feed pumps
- d. Low-low level in 2/4 steam generators
- e. Loss of offsite power (25 second sequence)
(**+++ CATEGORY 03 CONTINUED ON NEXT PAGE +***+)
=32 . 18!IBudEUI!_899,ggBlBg(! - PAGE 22
-QUESTION- 3.11 (2.00)
Answer the following questions.in reference to the recirculation mode of the ECCS.
- a. What causes automatic initiation of the recirculation mode of the ECCS-components? (Include setpoints and/or coincidence if applicable.)
- b. What functions are served by the residual heat removal pumps during the recirculation mode?
QUESTION 3.12 ( 75)
Upon receipt of a safety injection signal, state whether the following valves OPEN, CLOSE, or REMAIN THE SAME.
- a. CVCS charging isolation valves 62-90 and 62-91
Which one of the f ollowing is initiated by a high level of 60% in a steam generator?
- a. The No. I heater divert valve opens, diverting feed flow to the condenser.
- b. A rapid closure of the NSIV for that steam generator occurs.
- c. A rapid closure of the feed regulating valve associated with that steam generator occurs.
- d. The main feed pump feeding that steam generator trips and the electric auxiliary feed pumps auto start.
(***** CATEGORY 03 CONTINUED ON NEXT Pp6E ***ce)
T7-- y wy , , ,-- wr v y
31._lUSIB9dEU?f,0ND,90 NIB 0(! PAGE 33 QUESTION. 3.i4 (1.00)
During normal operation, a pressurizer pressure control channel fails
-high. If it is the controlling pressure channel that falls, which one of the f ollowing statements correctly describes the immediate response of the listed components to this failure?
- a. Both spray valves go wide open, control and backup heaters come on, PORV 340 will not open due-to interlock with channel 322
- b. Both spray valves go wide open, control and backup heaters turn -r off, PORV 340 fails open
- c. One spray valve goes wide open, control and backup heaters turn off, PORV 340 will not open due to interlock with channel 322
- d. Both spray valves go wide open, control and backup heaters turn
( off, PORV 340 will not open due to interlock with channel 322 DVEST10N 3.15 (1.00)
During normal operation, a SECONDARY pressurizer lavel control channel fails low. Which one of the following statements correctly. describes the immediate response of the listed components to this f ailure?
- a. A letdown isolation valve will be closed, orifice isolation valves will be closed, pressuri:er heaters will ce turned off [
- b. A letdown isolation valve will be closed, orifice isolation valves will be closed, pressuri:er control heaters will turn off, backup heaters will turn on -
- c. Letdown isolation valves will remain open, orifice isolation valves will be closed, pressurizer heaters will be turned off
- d. A letdown isolation valve will be closed, orifice isolation valves will remain open, pressurizer heaters will be turned off t
l' te**** CATEGDRY 03 CONTINUED ON NEXT PAGE *****)
r 4
11..lU!IBUMEUI!,ANp_ggNIBg(! _PAGE 24 l .
QUESTION 3.16- .(1.00)
During normal operation, a staae flow transmitter fails low. Which-one of the following. statements correctly describes the response'to this
. failure?
- a. Alare received for to FW flow, feed regulating valve.goes fully open, steae generator level increases
- b. Alare received on Hi FW Flow, .f eed regulating valve goes f ully closed, steam. generator level decreases
- c. Alara received for Hi FW flow, feed regulating valve falls as is, steam generator level remains stable
- d. Alara received f or to FW flow, feed regulating valve goes fully open, feed water flow increases f QUEST 10N 3.17 (1.00)
I Which one of the following statements correctly describes the effect on rod action of an URGENT failure?
- a. Rods will not move in AUTO, rods will move in MANUAL, rods supplied by the unaffected power supplies will move in individual bank select
- b. Rods will not move in AUTO, rods will move in MANUAL, rods supplied by the unaf f ected power supplies will not move in individual bank select
- c. Rods will move in AUTO, rods will not move in MANUAL, rods supplied by the unaffected power supplies will move in individual bank select
- d. Rods will not move in AUTO, rods will not move in MANUAL, supplied by the unaffected power supplies will move in individual bank select
(***** CATEGORY 03 CONTINUED ON NEXT PAGE +++**)
m 3i2 18S1890!U19.88D,C0$1806) -PACE 29 QUESTION 3.18 (1.00)
Which'ofithe following statements is most correct concerning the upoer head injection system (accumulator) isolation valves?
- a. The isolation valves are solenoid operated, require operator action to close, have an alare in the main control room that Hill sound if RCS pressure is-above 1900 psig and valves in both paths are closed.
- b. The isolation valves are hydraulic cylinder operated, require operator action to close, have an alare in the main control room that will sound if RCS pressure is above 1700 psig and valves in both paths are closed.
- c. _The isolation valves are hydraulic cylinder operated, close upon low level in the UHI water accumulator,-have an alare in the main control room that will sound if RCS pressure is above 1900 psig and any valve is not fully ooen.
_d.
The isolation valves are solenoid operated, close upon low level in the UHI water accumulator, have an alare in the main control roca that will sound if RCS pressure is above 1700 psig and any valve is not fully open.
l
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
3r INSI@UdENIS,8$9,ggh]89L) PAGE 26
+
QUESTION 3.19 (1.00)
Which one of.the following statements is correct concerning containment isolation?
I .a. Low pressurizer pressure will cause a phase A isolation, high i differential preasure between any one main steam line and'one of the other three lines will cause a phase B isolation, one out of three high containment pressure signals will :ause a phase B isolation.
- b. Two out of three high containment pressure signals will cause a phase A containment isolation phase A always exists if containment isolation phase B exists, high differential pressure between any one main steam line and two of the other three lines will cause a phase A isolation.
- c. High differential. pressure between any one main steam line and one of the other three lines will cause a phase A isolation, low pressurizer pressure coincident with low pressurizer water level.
will cause_a phase B isolation, two out of three high containment pressure signals will cause a phase B isolation.
- d. Containment isolation phase A always exists if containment; isolation phase B exists, two out of'four high containment pressure signals will cause a phase A isolation, low pressurizer pressure coincident with low pressurizer water level causes a phase A isolation.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE **++*)
la._Jy!IBybEhl! 809_CONIB06)' PAGE_-27 QUESTION 3.20 (1.00)
Which one of the following statements correctly describes operation of the steam dump control system when;it is being operated.in the Tavg mode?
- a. Used greater than 15% load, load rejection-will cause a turbine
-trip without reactor tcip if less than 50% reactor power, . .
2 degrees F dead band provides steady conditions for rod control-to retain-control, turbine impulse pressure is an input.
- b. Used below 15% lead, load rejection will-cause a turbine trip without reactor trip if less than 75% reactor power, 2 degrees F dead band provides steady conditions for rod control to retain control, turbine impulse pressure is not an input.
- c. Used greater than 15% load, load rejection will cause a turbine trip without reactor trip if less than 50% reactor power, 5 degrees F dead band provides steady conditions for rod control to retain control, turbine impulse pressure is an input,
- d. 'used below 15% load, load rejection will cause a turbine trip without reactor trip if less than 50% reactor power, 5 degrees F dead band provides steady conditions for rod control to retain control, turbine impulse pressure is not an input.
QUESTION 3.21 (1.00)
Which one of the following correctly describes the initial response of feedwater flow if the steam pressure transmitter controllino the steam generator water level control system fails high while at 50% power?
- a. Feedwater flow would increase due to the maximum steam pressure input to the steam flow signal
- b. Feedwater flow-would increase due to the level mismatch error
. between actual and prograssed level caused by the pressure instrument failure
- c. Feedwater flow would decrease due to the eisaatch between I
steam and feedwater flow signals caused by the steam pressure instrument failure
- d. Feedwater flow would remain the same due to the dominance of the level error signal over the flow error signal
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
-- , , , -p. &
}z..ly!!890EUl@_6hp_C9 BIB 0(! PAGE 28 QUESTION 3.22 (1.00)
Which one of the following correctly describes the system response if the steam generator water level transmitter controlling the steam generator water level control system f ails low while at 50% power?
- a. Steam generator water level decreases because the f eedwater regulating valve modulates closed
- b. Steam generator water level increases because the feedwater regulating valve modulates open
- c. Steam generator water level decreases because the main feed pump speed decreases '
- d. Steam generator water level increases because the level error signal is not lagged
(***** END OF CATEGORY 03 ***++)
4:__88gCEpV8ES_;_N0808L3.8)UO8d86,_EdE8GESCy_AND PAGE 29 889196991086_cgNI896 QUESTION 4.01 (1.50)
For each of the statements below, fill in the blanks with one of the following facility / center:
- 2) Operations Support Center (OSC)
- 3) Local Recovery Center (LRC)
- 4) Site Decontamination Facility a) ________ is an area containing equipment and supplies that are required for radiological cleanup and/or patient evaluation and stabili:ation, b) The electrical, instrument and mechanical maintenance shops would be classified a/an ______ upon activation of the Radiological Emergency Plan.
c) The health physics and the radiochemical labs would be classified as a/an ______ upon activation of the Radiological Emergency Plan.
d) ______ may be used by the NRC during the event as an area near the site for assessment and assistance, e) ______ is located approximately 1.5 miles from the plant at Power Operations Training Center rooms 64 and 65.
f) ______ is the focal point of onsite activity and is the primary source of communication with offsite organi:ations during the event.
QUESTION 4.02 (1.50)
Assume that for each of the following failures, no operator action is taken
! and the reactor subsequently trips. For each failure, state the protection signal that causes t!.e trip (i.e. containment pressure, SI). Consider each case independently.
LEVEL
.) The pressuri:erAcontrolling channel fails high.
LEVE L b) The pressurizerAcontrolling channel fails low.
c) The pressurizer secondary level channel fails low.
l
(***** CATEGORY 04 CONTINUED ON NEXT FA0E ****+1
4 2..&BQCEQQBE@_ _NQBU@(s_@QNQBd@(3_EdEB@ENQY_ANQ PAGE 30 BAQlRLQQLQ8L_QQNIBQ(
o QUESTION 4.03 (1.00)
Administration Instruction A!-2, Authority and Responsiblities for Safe -
Operation and Shutdown, addresses the .uthority for the manipulation of controls. State WHEN and UNDER WHAT CIRCUMSTANCES an UNLICENSED individual is allowed to manipulate any control that directly affects reactor reactivity or power level.
QUESTION 4.04 (1.50)
Explain how a locked-closed valve is verified CLOSED, Assume that the locking device is locked and will not permit any movement.
QUESTION 4.05 (2.00)
In reference to the critical safety functions / trees, answer each of the following statements TRUE or FALSE.
a) An oran,1 path for SUBCRITICALITY has priority over a RED path for heat sink.
PTS b) The ':: '"T r!~ critical safety function has priority over the CONTAINMENT INTEGRITY critical safety function, c) If a YELLOW condition is diagnosed for one of the critical safety functions, the operator (s) may choose whether to continue with the optimal recovery in progress or to initiate function restoration to restore the critical safety function, j d) Once monitoring of the status trees has been initiated, they need to be conitored every 10 to 20 minutes during yellow or green conditions.
r l
l
(*++** CATEGORi 04 CONTINUED ON NEXT PAGE *****)
4z iB899E998E9 _U9806ks.0EU9800ks.[d[BggyCy,@Np PACE 31 809195991C0k.99dIB9k
- QU!STION 4.06- .(1.50)
Prior to_a reactor startup, with the RCS at normal operating pressure and temperature, the following RCS leakages exist. 'For'each~ leak listed below, state whether you could STARTUP or would.have to remain SHUTDOWN.
(Treat each leak Lalow as an indepedent event.)
- a. A leak from an unknown source ifo 1.5 GPM.
'b. 6.0 GPM from a manual valve packing gland,
- c. 0.4 GPM from one steam generator QUESTION 4.07 (1.00)
List f our means listed in E-3, Steam Generator Tube Rupture, which may be used to identify the ruptured steam generator.
QUESTION 4.08 (2.00)
Answer each of the following questions assuming the primary plant is solid
.in accordance with SON G01-1, Plant Startup free Cold Shutdown to Hot Standby.
- a. State the precautions taken with both safety injection pumps and the non operating centrifugal charging pump to prevent overpressurization of the RCS with an inadvertent safety injection signal.
3
- b. Name the component that provides overpressure relief protection for the RCS when primary pressure is less than 500 psig.
- c. Name the component that is used to control RCS pressure when the plant is solid with letdown from RHR.
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4t._E8QQEQUB[i.,: 6Q858(t.8%BQBUB(i_EUE8QEdgy_ANQ
' P AGE- 32 80Q19LQQ!GBl GQU18QL-
~
~
QUESTION 4.09 (1.50)
During a reactor startup per G01-2, Plant Startup from Hot Standby to Minimum Load, careful monitoring of the source range nuclear instruments is required.
- a. State the sinimum number of counts per second required on the highest reading source range instrument.
- b. State one reason _ f or having th'e einiaue number of counts specified in part-a,
- c. State how many doublings (approximately) of initial source range counts are expected by the time criticality occurs.
QUESTION 4.10 (1.50)
Fill in the following blanks in accordance with the Technical Specifications requirements:
The pressurizer heatup and cooldown rates shall not exceed
_________. degrees F/hr and ____ ....... degrees F/hr, respectively. The spray shall not be used if the tesperature difference between the pressurizer and the spray fluid is greater than .............. degrees F.
QUESTION 4.11 (2.50)
Answer each of the following questions assuming that it has been decided to abandon the control roce.
- a. State the lasediate Operator Action (s) per A01-27.
- b. State how each of the following plant parameters would be controlled from the auxiliary control room. (Naming the conponent(s) used in controlling the parameter will suffice.)
- 1) RCS pressure
- 2) Pressurizer level
- 3) Steam generator feed
- 4) Steam generator pressure
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J
'!__.EB99E99859 :_U98586i,0!NgBd8(i,[d[8@[UQy,8ND PAGE 33 68919k9919Bk.99dIB96 w- .-
QUESTION 4.12 (1.00)
In reference to the Critical Safety Function Status Trees, select the most correct operator action-if an extreme challenge (red) is diagnosed while an optimal recovery is in progress.
- a. Complete the optimal recovery and then initiate function restoration to restore the Critical Safety Function under extreme challenge.-
- b. Perform the optimal recovery in parallel with initiating function restoration to restore the Critical Safety Function under extreme challenge.
- c. Immediately stop optimal recovery and initiate f unction restoration to restore the Critical Safety Function under extreme challenge.
- d. Continue with the optimal recovery while continuously monitoring the Critical Safety Function under extreme challenge.
QUESTION 4.13 (1.00)
In reference-to the Critical Safety Function Status Trees,' select'the most correct operator action if, during function restoration to address an extreme or severe challenge, a higher priority challenge is diagnosed.
4
- a. Complete the function restoration in progress, then initiate function restnration to address the higher priority Critical Safety Function Challenge.
- b. Continue conttoring the status trees. It is the operator's choice of whici challenge to address.
- c. Perform the function restoration for both the Critical Safety Functions under challenge in parallel.
- d. Terminate the ongoing response and i.)itiate function restoration to address the higher priority Critical Safety Function challenge.
QUESTION 4.24 (1.00)
Concerning FR-5.1, Response to Nuclear Power E2neration,/ATWS, explain why the reactor aust be tripped before the turbine is tripped.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
41__E89CEpyBES_:_U9Bd862.8!U98d8k3_gdEBGENCy_@Np PAGE 34.
66919699196k_cgyIgg6
- ~ .
QUESTION 4.15 (1.50)
Per FR-C.1, Response to inadequate Core Cooling, list the three methods, in order of priority, for regaining c ' cooling.
QUESTION 4.16 (1.00)
Explain why ECA 0.0, Loss of All AC Power, has priority over ALL FR6's.
0UESTION 4.17 (2.00)
Answer each of the following questions in reference to the performance of ECA 0.0, Loss of All AC Power.
- a. Explain the reason for this Caution in ECA 0.0:
If an SI signal exists or if an SI signal is actuated during this guideline, it should be reset to permit manual loading of equipment on shutdown board. ,
- c. The RCS is cooled down by using the SG PORV's. Explain how and from location the FORV's for SG's 1 and 4 are operated.
- d. What action is taken (stop depressurization o continue depressurization) if pressurizer sevel is lost and reactor vessel upper head voiding occur due to depressurization of the steam generators?
QUESTION 4.18 (1.00)
Answer the following questions in reference to the performance of the Emergency Recovery Guidelines (ERG's) l l
- a. If the contingency action (RNO coluan) cannot be performed or is not l successf ul, and f urther contingency instruction is not provided, what is l the operator's next action?
- b. Under what circumstances may the operator proceed to the next step prior to the completion of a task in progress? l l
l
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
l I
I
~$1_ 2699[pVBES_ _UQBU863,8!N9898(2 [d[BQ[NQy_@NQ PAGE 35 680196991986_G98169L l
QUESTION 4.19 (1.00) l From the following, select the one situation that has the highest priority
~
l for operator response.
- a. Pressurized Thermal Shock - Yellow
- b. Inventory - Red
- c. Core Cooling - Orange
- d. Heat Sink - Red
- e. Subcriticality - Orange
- f. Containment - Yellow OUESTION 4.20 (1.00)
From the following possible combinations of parameters, select the one combination that correctly describes characteristics of natural circulation in accordance with SONP ES 0.3, Natural Circulation Cooldown.
RCS SG That Cor e Exit Tcold Subcooling Pressure Thermocouples
- a. 10 F Decreasing Stable Decreasing Tset for 59 Press 7
- b. 20 F Stable Tsat for Decreasing Stable RCS Press
- c. 30 F Decreasing Decreasing Stable Stable ;
- d. 40 F Stable Decreasing Increasing Tsat for ,
SG Press
- e. 50 F Decreasing Stable Stable Decreasing I
I
(++*** CATEGORY 04 CONTINUED ON NEXT PAGE **+++1
$1__BBQCEDyBES_:_UQBb6(3_$@NgBd@(3 EbEBQENCY_AhD PACE 36 88DlQ(991C86_CQNIBQL
=
- 00ESTION 4.21 (1.00)
Choose the sequence of'the following four emergency baration steps that ]
places them in the correct order.
- 1. Verify ~ emergency boration flow.
- 2. If additional boration required, then align charging pump suction to RWST.
- 3. Open emergency borate valve.
- 4. Place boric acid transfer pumps to fast speed.
- a. 3,4,2,1
- b. 4,2,3,1
- c. 3,4,1,2
- d. 4,1,2,3 4 (
OUEST10N 4.22 (1.00)
From the following statements, select the one that correctly describes operation of the diesel generators per A01-3% in the event of a loss of offsite power.
- a. Each diesel generator is designed to carry up to 4400 kW during continuov5 operation
- b. If one diesel trips, its companion train diesel may lack sufficient cooling water a
- c. Equipment started b5- blackout will stop when offsite power is restored
- d. A diesel that has been loaded less than 40% for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> may be l shutdown without being purged 1
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
1
'It. EBIUQ[P(ES_QE_Nyq(E@8_PQQE8.P(@81_QEE8@l[0N 1 PAGE 37 18E880QYU68[CSt _8E81_18@NSEE8,@NQ,[(yl@_F(Og ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C.
O,00)
ANSWER 1.01 LL.ar)
- a. Decrease (0.50)
- b. E::ren '^.50; d-tN6efhe- 3 tLE 7 ED
- c. Increase (0.50)
REFERENCE SON, Operator Certification Train;ng, Reactor Theory, Reactivity Coefficients IV, Week 1-4, Training objective V.E 192005K107 ...(KA'S)
ANSWER 1.02 (1.00)
Margin to criticality decreases by 7/8 (1.00)
(Each time counts double, the reactor has gone halfway to criticality.
Therefore, 100 x 2 = 200, 200 x 2 = 400, 400 x 2 = B00. The counts have doubled three times, 1x 1/2 = 1/2, 1/2 x 1/2 = 1/4, 1/4 x 1/2 = 1/8. The reactor has gone 7/8 of i5e way to criticality.)
REFERENCE SON Prelicense training program lesson plans, week 1, Reactor Physics Revi ew, Training objective V.D 192000K104 ...(KA'S)
ANSWER 1.03 (1.50)
- a. P = Po10exp (SUR) (t) (0.5)
= (0,1 ) 10ex p ( 0.15) ( 2)
= 0.199526% (0.5)
Round off to 0.2 % is acceptable.
- b. About 1-2 x 10 -6 amps c.r 1% to 2% of full power. (0.5)
REFERENCE SON Prelicense Training progrta lesson plan, week 1, Reactor Physics Review, Training objective V.T 19200BK114 ...(KA'S)
(
(
It .EBINGitLES_0E.guq(E68 EQME8_E(@NI_QEE861[QNi PAGE 33 IBE80QQIy8dlCSt ,UE61,186dSE(8,@dQ E(UlQ,[(QQ
~
_ ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C.
AOSWER 1.04 (1.50)
- c. ECP higher than_ACP (0.50)
REFERENCE SDN prelicense Training program lesson plan, week 1, Reactor Physics Review, Training objectives V.K and V.E 192008K114 ...(KA'S)
ANSWER 1.05 (1.00) b REFERENCE SON Prel4 cense Training Program lesson plan, week 1, Reactor Physics Review, Training objective V.H 192004K106 ...(KA'S)
ANSWER 1.06 (1.00) i d
REFERENCE Bisic Nuclear Physics II, Training objective V.A 192002K111 ...(KA*S)
ANSWER 1.07 (1.00) b REFERENCE SON Prelicense Training Program lesson plan, week 1, Reactor Physics Review, Training objective V.D 192002K112 ...(KA'S)
11i,, PRINCIPLES OF NUCLEAR' POWER PLANT OPERATION i PAGE 39 L:
'IBE86QQ{N@dICSt ,yE@l 18@ESEE8,@NQ,E(Qig,E(OH
-ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C.
I i
ANSWER 't.08 (1.00) s a REFERENCE-SON Prelicense Training Program lesson plan, week 1, Reactor. Physics {
Review, Training objectives V.B and: V.C 192000K104 ...(KA'S) l l
-ANSWER 1.09. (1.001-d REFERENCE SON Prelicense Training Program lesson plan, week 1, Reactor Physics
~
Review, Para C 192008K103 ...(KA'S)
ANSWER 1.10 (1.00) b
, REFERENCE SDN Operator Certification Training Reacter Theory Reactivity Coefficients IV, week 1-4, training objective V.E 192005K107 ...(KA'S)
ANSWER 1.11 (1.00) b REFERENCE SON Operator Certification Training Reactor Theory Core Poisons V, week 1-5, training objective V.A 192006K106 ...(KA'S)
't. PRINCIPLES OFJNUCLEAR POWER PLANT _QPEB@llgN t PAGE 40
, THEgdggyN@dlCSx_HE81_{B6NSFE8_@NQ_E(UlQ E(QN
- ANSWERS -- SEQUOYAH 1&2 -87/11/17-SHIRAKI, C.
I l
ANSWER: 1.12' (1.00) a REFERENCE SON Operator Certification Training Reactor: Theory Core Poisons V, week 1-5, training objective V.D 192006K114 ...(KA'S)
ANSWER 1.13 (1.00)
-C-REFERENCE SON Prelicense. Training Program lesson plan, week 1, Reactor Physics Review,' Para K 192006K103 ...(KA'S)
AriSWER 1.14 (1.00) b REFERENCE SON Reactor Physics Review SON Reactor Theory 192005K114 ...(KA'S)-
ANSWER 1.15 (1.00) d REFERENCE SON Reactor Theory SON Reactor Physics Review 19200BK!!4 ...(KA*S)
li_LE81NC[P(ES,QE UUC(E@8 PQWE8 P(@h1_QEE861[QUt PAGE 41 IBE80QQY8@d!CSt _UE61_IB@yQEEB,@NQ_[(ylp,[(QW ANSWERS -- SEQUOYAH 1&2 -87/11/17-SHIRAKI, C.
ANSWER 1.16 (1.00) a REFERENCE SON Operator Certification Training Reactor Theory (
Operator application Training Objective V.B.4 192002K114 ...(KA'S)
ANSWER 1.17 (2.00)
I
- a. lower
- b. higher
- c. lower
- d. lower (0.50 each)
REFERENCE Licensed operator prelicense and certification training Thereocynamics, fluid flow, and heat transfer review Training objective V.B.14 193007K108 ...(KA'S)
ANSWER 1.18 (1.00) b REFERENCE SON Prelicense Training Program lesson plan, week 1, Reactor Physics Review, Training objective V.K t 192004K108 ...(KA'S)
ANSWER 1.19 (1.00) a REFERENCE SON Operator Certification Trainin;, ::ai., casm System, Para. X.E.
191002K102 ...(KA'S)
it._P81NCIP(ES_QE NQC(E@B,PQWE8_P(@NI_QP[8@llQys PASE 42 ISE60QQYN@dlCSt _HE@l,18@NSEE6,@NQ_[(QlQ,[(Q@
~'
' ANSWERS -- SEQUOYAH 1&2 -87/11/17-SHIRAKI, C.
ANSWER 1.20 (1.00) d REFERENCE Licensed Operator prelicense and certification training Technical stiff and managers training Thermodynamics, fluid flow, and heat transfer review Para. IX.F.
193003K125 ...(KA'S)
ANSWER 1.21 (1.00) d REFERENCE 9 Heat transfer, t ynamics, fluid flow fundamentals Section ' , rart B, Chapter 1
- K106 ...(KA*S)
ANSWER 1.22 (1.00) d REFERENCE Heat transfer, thermodynamics, fluid flow fundamentalsSection III, Part 8, Chapter 1 191004K106 ...(KA'S)
ANSWER 1.23 (1.00) b REFERENCE SON Prelicense Training Program lesson plan, week 1, Reactor Physics Review, Para X 193003K125 ...(KA'S) l l
l l
____________________J
It _E816Cl&(ES_QE_8QC(E88_E08E8_E(881.0&E8811081 PAGE 43 ISE88QQ1N88[CSi _8E81_188NSEE8,8NQ_[(Qlp,[(Q@
ANSWERS -- SEQUOYAH 1&2 -87/11/17-SHIRAKI, C.
ANSWER 1.24 (1.00) d REFERENCE Licensed Operator pre license and certification training, technical staff l and canagers, Thermodynamics, fluid flow, heat transfer review, Para 10 l 193008K103 ...(KA'S) l l
l l
ANSWER 1.25 (1.00) l c
REFERENCE Licensed Operator pre license and certification training, technical staff and managers, Thermodynamics, fluid flow, heat transfer review, Para 18 193008K103 ...(KA'S)
ANSWER 1.26 (1.00) d REFERENCE SON Natural circulation / inadequate core cooling 19300BK123 ...(KA'S)
ANSWER 1.27 (1.50)
- a. Tsat for 2250 psia (2235 psig)
= 652.67 F (0.50)
Subcooling margin = Tsat - Thot = 652.67 - 5B7 = 65.67 F (0.50) 1.05 F
- b. Subcooling margin decreases because Thot will increase as power increases (0.50)
REFERENCE STA, licensed operator training, Natural circulation and inadequate core cooling, Para X.B 001000K556 ...(KA'S) 1 l
l l
'2 _t P(@Ul_QEQ1QU,18Q(UQ10Q.S@EEIY_88Q_EUERQEUQi_@yQlEdQ PAGE G4 ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAXI, C.
ANSWER 2.01 (2.25)
- a. (The normal steam pressure setpoint of 1005 psig maintains Tavg at about 547 degrees F.) A decrease in the setpoint to 935 psig (950 psia) would cause the dumps to open (0.25) and cool Tavg to about 540 degrees F (0.34), where the P-12 interlock would close all steae dumps (0.34).
- b. Steam duep operation would be blocked (0.33), steam generator pressure would rise to the setpoint of the power operated relief valve (0.33) which would maintain steam generator pressure at 1040 psig (0.33), causing pr .;:t:- pre::_ : to stabilize at about 551 degrees F (0.33).
RCSn: TAv6 REFERENCE SON Operator Certification Training, Steam Dump Control System, Pa X.C.3 041020K105 ...(KA'S)
ANSWER 2.02 (1.00)
- a. ". . . low DNBR (due to adverse combinations of high temperature, low pressure, high flux differences, and power)." (0.50)
- 5. ". . . damage due to excessive reactor power output." (0.50)
(
REFERENCE EON, Operator Certification Training, LP: Reactor Coolant Temperature Instrumentation (week 4-6), Training Objective F 012000K402 ...(KA'S)
ANSWER 2.03 (1.00)
- a. 1. All orifice isolation valves must be closed. (0.25)
- 2. Pressurl:er level must be greater than 17%. (0.25)
- b. To ensure the regenerstive heat exchanger always has RCS system pressure in it to prevent flashing of high temperature water. (0.50)
REFERENCE
[
SON, Operator Certification Training, Chemical and Volume Control System, l Training objectiv_ F.1 004010K403 ...(KA'S)
7 h__Eb6Bl_D[Siq8,18C(yD!B@_SQEEIY.88D_EMERGEUCY_SYSIEH@ PAGE 45 ANSWERS -- SEQUOYAH 1&2 -87/11/17-SHIRAXI, C.
ANSWER 2.04 (3.00)
- a. Loop 3 (0.50)
Cold Leg (0.50)
- b. Component Cooling Water System (0.50)
- c. RCDT . %'
VCT 'O.50' Charging pump suction (VCT outlet) f9.T viA Muer- Vt>tve ' . 5 0 ' b h u1 3 of4 At o.5D c.Aul)
REFERENCE SON CVCS System Description, Learning objective V.B -
002000K106 004000K118 004000K405 . .(KA*S)
ANSWER 2.05 (2.00)
- a. Auto start of the electrical auxiliary feed pumps. (0.50)
- b. Acceleratinn of operating main feed pump to its high speed stop.
(0.bO)
- c. Isolation of affected MFPT :andenser. (0.50)
- d. Turbine runback to 75'/. unit load. (0.50)
REFERENCE SON SD Condensate and Feedwater. Learning objective V.C 059000K303 ...(KA'S)
ANSWER 2.06 (1.50)
- a. The intermediate flux trips will not be automatically reins (0.50) OR I ositiTV f o ugude. cite fne sRte's weeu Ogcow V-l tatedL BnMstf F-f o bismr>
in PeWEA sup?i.'i To TM C RM'i )
- b. By the status lights for the P-10 bistable switches (0.50) and for the P-10 permissive interlocks (0.50) (in both trains of the reactor protection system)
REFERENCE SON Operator Certification Training, Solid State Protection System, Para l X.F NRC IE notice 06-105 012000K401 ...(KA'S)
2 t__((6N1_DE@lGN_{NC(UD[NG_SAEEly_6ND_EUERg;UCy_SYSIENS PAGE 46 ANSOCRS -- SEQUOVAH 1&2 -87/11/17-SHIRAKI, C.
ANSWER 2.07 (2.50) {
l
- a. From left to right: 480 v unit boards, NG sets, reactor trip breakers in parallel with bypass reactor trip breakers, "A" phase regulating transformer in parallel with reactor trip breakers and reactor trip bypass ;
breakers (0.25 each)
- b. From left to right: 480 V shutdown boards, vital transfer switch, primary fused disconnect, 480V/120V transformer, 120 VAC instrument power distribution panel (0.25 each)
REFERENCE SGN Lesson Plan, full length rod control system, Figure 1 001000K202 ...(KA*S) l ANSWER 2.08 (i.25) c,a, e, b, d (0.25 each)
REFERENCE SON Operator Certification Training, ECCS Components and Operation (week 4-11), Objective B.2.1 013000A201 ...(KA'3) l ANSWER 2.09 (2.50)
- a. Safety injection signal from train IA or 2A 6tu.couf AFTgL h w if Wbeta TIMG MfuRH1 cF uuf boh)4 0 4LTM(#
No blackout signal Transfer switch gear in normal b M 1 E Pu.e s El'u t
- df M - M c ! ' ^Lh*WC-
-- ___t . ?e+teteeJ .
bhMV3 M O.336M.d)
- b. To prevent more than one rump at a time being fed from a single diesel generator (0.50)
- c. Unit 1 (0.50) l d. 2A and 28 (0.50) l REFERENCE SON System Description. Essential raw cooling water system 076000K101 076000K105 076000K201 076000K402 ...(KA*S)
Page 26 of 48 S.u E.f T/ou 7.07 SQN OP TRNG Revision 0 8/29/83
\
SEQUOYAH NUCLEAR PLANT LESSON PLAN Instructor's Notes (5) The "main" and "auxiliary" are set up as shown:
- Figure 1 .
m/s = Rm 3ra (
'ByA ByB a - u -- O & c -u U u e
m/6 *T y .
i.{h Lur Bd's. YI
- C 4 j,t FPAM 4BOv L4d ww Bd's I 150/'2cv mesuLATD c.4c
'A3 1
" ' \.'
qq/g, Y T^d *9** E A ip
_LsA e ,i fQ
, .;il i
3
._ _ .i 3 1 C
i'gv g l T" TN %.3 msr. 64. .7 i
MTRL ked blhT. PNL.
%ug[g D C*N k Swi
2 t,,P(@UI,QEsl@N_lNCLUQ[UQ_S@EEIY_QNQ E MRQ[NCY_SYSIEMQ PAGE 47 ANSWERS -- SEQUOYAH 1&2 -87/11/17-SHIRAKI, C.
ANSWER 2.10 (2.50)
- a. O to 1200 degrees F (0.50)
- b. 700 degrees F '^ 5^:
. ; C ou a t n TMO.M o c cutua l_f. ET'N > 1200 F
(_Gt S > 100 F w ifd <. 4o f Suaccotin 4. LA NY out f o t 0.60.) ;
- c. Slight increase (0.50) as steam from hottest core regions passes thercocouples as part of an advancing front of two phase flow (0.50),
- d. Hot spot or flow channel blockage (1 required for 0.50)
REFERENCE SON Natural circulation and inadequate core cooling, Para X.D.2 SON Incore Instrumentation, Para X.B.4 017020KiO2 017020K402 017020K403 017020K503 ...(KA'S)
ANSWER 2.11 (2.00)
Leakage would be the sum of the centrifugal charging pumps and the safety injection pumps (0.50).
Centrifagel charging pumps 2x ebout 400 gpm = 000 gpa (0.50) 2afety injection pumps 2x about 100 gpm = 200 gpm (0.50)
Leak rate = about 1000 gpa (0.50)
Note: Acceptable flow rate for CCP's 300-500 gpe Acceptable flow rate for SIP's 50 - 150 gpm REFERENCE SON Emergency core cooling, Para X.A.2. and flow capacity figures 006000K103 ...(KA*S) i 1
f l
2t__E($U1_QES16N_16Q(yQ1NQ_S6 Eely _6NQ_{UQRGENCY_SYSIEMS PAGE 48
, ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C.
ANSWER 2.12 '2.^^' I'
- a. ' Compared with Tref to generate deviation signal to control the modulation of the steam dump valves (0.67)
- b. Tavg used for level reference (0.66)
. Le- ': c; :i;r:' :0!n !de-t :t' :::te- t-!, ::u;;; 5 d:::t e-equi:ti ; ~c: t: :!=:: '^ .57' D e <_g T (D REFERENCE SON Condensate and feedwater SON Pressurizer and control systems SON Reactor coolant temperature instrumentation SON Rod control system SON Steam dump control systee 001000A101 011000A104 041020K411 059000A306 ...(KA'S)
ANSWER 2.13 (2.50)
-a. Control rcds insert (0.50) due to turbine / reactor power difference (0.25) and Tavg/iref mismatch (0.25)
- b. Decreases (0.50)
- c. Decreases (0.50)
J. Low pressurizer pressure (0.50)
REFERENCE SON Rod control system 011000K402 016000K111 016000K112 039000K104 ...(KA'S)
ANSWER 2.14 (1.00)
C REFERENCE SON Operator certification training, RCS 007000K103 ...(KA'S)
(2 t._Ek6dl QESl@N_18C(yplN@,SQEEIY QNQ.EdER@ENCY,@Y@ led! PASE 49
.,., ANSWERS -- SEQUOYAN 1&2 -87/11/17-SHIRAKI, C.
'NSWER-A 2.15 ( 1.' 0 0 )
b ,
i REFERENCE.
SON SD Condensate and Feedwater Para'E l 059000K402 ...(KA'S) l ANSWER 2.16 (1.00)
.b REFERENCE
. SON. Operator Certification Training, Residual Heat Removal System, SQNP LER'85-020-00 005000K407 ...(KA*S)
ANSWER 2.17 (1.00) a REFERENCE SON Operator Certification Training, Residual Heat Removal System, LONP LER.85-040-00 005000K407 ...(KA'S)
31._ldS18UdEBIS_ASD_C08180LS PAGE 30 ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C.
ANSWER 3.01 (2.00)
- a. Decrease (0.40)
- b. Increase (0.40)
- c. Decrease (0.40)
- d. Decrease (0.40)
R g AA A(u Tn6 5 AME
- e. ::: :: - (0.40)
REFERENCE SON, Operator Certification Training, LP: Reactor Coolant Temperature Instrumentation (week 4-6), Training objective F 012000K403 ...(KA*S)
ANSWER 3.02 (2.00)
Any fcur of the following at 0.50 points each:
- 2. Provides a signal to the SI block and reset logic
- 3. Locks in the circuit to prevent re-opening the MFW valves that were shut by either an SI or high steam generator level actuation signal
- 4. Provides a signal to the steam dumps so that the reactor trip controller controls the steam dumps vice the load rejection controller
- 5. Provides a signal to the process racks to reduce the HI steam flow program setpoint to its zero load setpoint REFERENCE SON, Operator Certification Training, LP: Solid State Protection System (week 6-6) 013000K115 013000K401 013000K412 ...(KA'S)
L, . l u ? u f' fo s PD5 Lia irisfes Co.iticnt 59perf Punct,ou .55ATuf Titce AufeMoric LAVD ATG
- 1. L u put To 7458 bMpuf6E blN ITIF1651M 560 0 6MM. 0 E E ueurt vsc a nca.)
e
li.,lh!IBydE81@_809_C0$1R0k! c PACE = 511 u
, ANSUERS.--iSE000YAH 1&2 ~ -87/11/17-SHIRAKI, C.
m
( ,
< 1 ANSWER. 3.03 m (1.75)
~.a - 1). Fall close (0.25) 2), Fall - cl ose- (0.'25) 3)r Fall,open' '(0.25)
- 4) ' Fall close (0.25):
- 5) Trip (0.25)
- b. 1) Fail.open (0.25)
-2) Fall-close (0.25)
' REFERENCE
SON ADI-10 Loss of Control Air 000065E208 ...(KA'S)
. ANSWER' 3.04- (1.50)
- 4. 1)-' Start' ( 0. 2 5 )'
- 2) Close '(0.25)
- 3) Close- (0.25)
- 4) Close (0.25)-
- 5) Close . (0.251
- 6) Trip (0.23;-
REFERENCE SON Energency core cooling, Para X.B.2 000011E301' 000011E302 000011E306 000011E308 ...(KA'S)
ANSWER 3.05 (2.50)
- a. Failure of a power range channel caused a High Flux Rod Stop.
- b. Loss of impulse pressure input and hence loss of program steam generator level o1L j Lh55 et unm itCW INMTIDMI b d I D $0 k T 6 0. C o M 1~In o k b4bTdd a
- c. L::: :' feed '!ce er ete:- '!c- c!;^21 *: t'r 'ced'2te' c c r- t ' 1
- , :' : . L ost o f Afgh m FLo w 314 t4 bL 'fo Fech Pu m p .svest, courRoc c.itcct ir.
Ll oTA L17. ETb SMku PLold il( AL l's (156b AS A Loan kEFL f Rtuc& Pe t DELTA 7
- d. Positive displacement pump speed controller failed. 7006EA4A.)
e '. Flow transmitter FT-62-93 failed (opening FCV-62-93).
REFERENCE SQNP A01-25 000057E301 000057G010 ...(KA'S) e
3t__18SI8UDENIS_AND_CgNIB9L! PAGE 92
, ANSWERS -- SEQUOVAH 1&2 -87/11/17-SHIRAKI, C.
t 1
ANSWER 3.06 (1.00)
Containment pressure, RCS temperature, RCS pressure, pressurizer level, steam generator pressure, containment water level KW5T LCnf tL APk) ft_0W RME LCL Sukosuul M LUtu (0.25 each f or any f our),122%ulntL TOLv Thhfto!I , h AF(Tl VLbid ?ob tTiod, L REFERENCE SON Post Accident Monitoring Instrumentation, Para X.A 000009E101 ...(KA'S)
ANSWER 3.07 (1.50)
- a. Tavg less than or equal to 540 degrees F in 2/4 Tavg channels (0.50)
- b. Auto-clears (0.50)
- c. Tavg above 540 degrees on 3/4 Tavg channels (0.50)
REFERENCE SON Prelicense Training, Revi(x nf Solid Stste Protecticn System, Para X.F.2 013000K412 ...(KA'S)
ANSWER 3.08 (2.50)
- a. Nove in (0.50)
- b. Decrease (0.50)
- c. Decrease (0.50) c'I b6LTA T
- d. Low pressurizer pressure (0.50) or m ;' ;' e r ' ; 'I n E' (0.50)
REFERENCE SON 0;?rator Certification Training, LP: Review of Instrument Failures, Learning objective V.A 001010A101 ...(KA*S)
\
I
______________________________________________-__a
r -;
3a-_INSIByMENIS_BND_CQNISQLS PAGE 53 ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C. ,
1 ANSWER 3.09 (1.25)
- a. 3
- b. 4
- c. 2 1
- d. 1
- e. 5 (0.25 each)
REFERENCE SON Technical Specifications, Safety Limits 012000K406 012000K610 ...(KA'S)
ANSWER 3.10 (1.25)
- a. MDAFWP
- b. TOAFWP
-1 Both
- d. -T :- A.~ W P b 6T A
- e. MDAFWP (0.25 each)
REFERENCE SON Operator Certification Training, Auxiliary Feedwater System, ,
Learning objective V.A 061000K402 ...(KA'S)
ANSWER 3.11 (2.00)
Setpoints are RWST - less than 29%, and containment sump - greater t h a n 444 ( 0,50 )
11 . 1 5 C/o
- b. Net positive suction head to the centrifugal charging pumps and safety injection pumps as well as perforcing low head injection (1.00)
. I
hi__lyS18ydENTS_QUD_COUI80($ PAGE .54
-ANSBERS,-- SEQUDYAH 152 -87/11/17-SHIRAKI, C.
l t
. REFERENCE SON Operator Certification Training, ECCS Components and Operation,
-(week 4-11), Learning Objective B.3 005000K40S- ...(KA*S) ;
ANSWER 3.12 (. 75)
- a. Close (0.25)
- b. .Open (0.25)
- c. Close (0.25)
REFERENCE SON Operator training, ECCS components and operation, Para X.B.2 013000K111 ...(KA*S)
- ANSWER 3.1i (1.00)
C REFERENCE SON Operator Certification Training, LP: St eam Ger. erat or s. (week 5-3),
Training Objective V.F 057000K303 ...(KA*S)
' ANSWER 3.14 (1.00) d REFERENCE SON Operator Certification Training, Review of Instrument Failures, (week 6-15), Learning objective V.A.3 0100000010 ...(KA'S)
ANSWER 3.15 (1.00) a REFERENCE SON Operator Certification Training, Review of Instrument Failures, (week 6-15), Learning Objective V.A.4 011000A211 011000K302 ...(KA'S) 1
3z. lN@lBUdEUI!_86D.00N!BOL@ PAGE 55 ,
ANSUERS -- SE000YAN 1&2 -87/11/17-SHIRAKI, C.
ANSWER 3.16 (1.00) b REFERENCE SON Operator Certification Training, Review of Instrument Failures (week 6-15), Learning objective V.A.5 035010A204 035010K101 ...(KA'S)
ANSWER 3.17 (1.00) d REFERENCE
} SON Operator Certification Training, Full Length Rod Control, (Neek 6-2),
learning Objective V.A.5 001050A201 ...(KA'S)
ANSWER 3.1S (1.00)
C REFERENCE SON Upper Head Injection System, Para X.D 006000K602 ...(KA'S)
ANSWER 3.19 (1.00) b REFERENCE SQN Operator Certification Training, Containment Isolation, learning objective V.B ANSWER 3.20 (1.00)
(
a REFERENCE j SON Operator Certification Training, LP: Steam Dump Control System, learning objective V.C 1
It._lyglBUdEBlS,68Q,COU160L@ PAGE 56 l
ANSWERS - SEQUOYAH 1&2 -87/11/17-SHIRAKI, C.
041020A102 041020K603 ...(KA'S)
\
l ANSWER 3.21 (1.00) a i
REFERENCE l SON Operator Certification Training, LP: Review of Instrument Failures, (week 6-15), learning objective V.A.5 035010A203 ...(KA'S) l' l'
ANSWER 3.22 (1.00) b REFERENCE SON Prelicense Training, Steam Generator Water Level Control System, Para X.E 035010A203 ...(KA'S) l l
I 4i._P8QCEQUBE@_ ,NQBd@(t_@By0Bd@(t_E6E8QENCY_6ND PAGE 57 BBRIQL9GIC@(_CQN18QL ANSWERS -- SEQUOYAH'l&2 -87/11/17-SHIRAKI, C.
ANSWER 4.01 (1.50) a) Site Decontamination Facility (0.25)
.b) Operations Support Center (0.25) c) Operations Support Center (0.25) d) local Recovery Center- (0.25) e) Local Recovr ry Center e (0.25) f) Technical Support Center (0.25)
REFERENCE SON, Operator Certification Training, LP: Radiological Emergency Plan (week 8), Training objectives A, B.2-3 and B.10-13 194001A116 ...(KA'S)
ANSWER 4.02. (1.50) a) HI pressurizer level (0.50) b) HI pressurizer level (0.50) c) HI pressurizer level (0.50)
REFERENCE SON, Operator Certification Training, LP: Review of Instrument Failures (week 6-15), training objective A 011000A210 011000A211 035010A203 035010K101 ...(KA'S)
ANSWER 4.03 (1.00)
When it is part of the individual's training to qualif y f or an operator's license (0.50); and, only if, under the direct supervision of a licensed operator (0.50).
REFERENCE SON, Operator Certification Training, LP: AI-2 Authorities and Responsibilities for Safe Operation and Shutdown (week 10-1), Enabling objective B.5 001000G001 ...(KA'S)
I' 4 __PB0QEDyBES_ _UO8d6(1,8BN0858(i_EdE8GE8CY_AND PAGE_ 58 8@D[0(0GlC0(_C00160(
1 .
i ANSWERS -- SEQUOYAH 1&2 -87/11/17-SHIRAKI, C.
I l ANSWER 4.04 (1.50)
- 1) Remove the locking device (0.30)
- 2) Attempt to move valve operator in the CLOSED direction (0.30)
- 3) Feinstall locking device (0.30)
- 4) Verify it is locked (correctly reinstalled / locked) (0.30)
- 5) Second person must verify that locking device is (reinstalled and) locked) (0.30)
REFERENCE SON, G01-6, page 4 of 12; and, SON, Operator Certification Training, LPs GOI-6, Apparatus Operation tweek 6-14) 194001K101 ...(KA'S)
ANSWER 4.05 (2.00) a) False (0.50) b) True (0.50) c) True (0.50) d) Tree (0.50)
REFERENCE SON, Operator Certification Training, LP: Status Trees (Heek 7-7),
trainirig objectives B and D 194001A102 ...(KA'S)
ANSWER 4.06 (1.50)
- a. Shutdown (0.50)
- b. Startup (0.50)
- c. Shutdown (0.50)
REFERENCE SON TS 3.4.6.2 002020G011 ...(KA's) l
4 t._88QCEDyBES___$0806(i_6BUO80@(t_EdESCEUCY_AND PAGE 59 86 Ele (QGIC86_CQUIgg(
- 1 ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C. '
ANSWER 4.07 (1.00)
Any four of the following. 0.25 each
- 1. Unexpected rise in steam generator level
- 2. Steam generator blowdown monitors (RM-90-120, 121, or 124)
- 3. Rad Con survey of main steam lines
- 4. Rad Con survey of blowdown lines
- 5. Chem lab sample REFERENCE SONP E-3, Steam generator tube rupture, Step 3 000038A203 ...(KA'S)
/.NSWER 4.08 (2.00) i
- a. Power is tagged out to both SIP's (0.50) and the non operating centr.ivgal charging puep is in pull to lock (0.50).
- b. The RHR suction line relief valve (0.50).
- c. The low pressure letdown control valve (0.50).
REFERENCE SON G01-1, and SON, Operator Certification Training, LP: G01-1 (Heek 6-7),
objectives C and E 002000K410 005000K401 0060000007 0060000010 006000G014 006000K103 ...(KA'S) ;
4 t__P80CEDU8ES _80888(i_88N0888(3_g6{8@{UQy_AND PAGE 60 88010kOG.1C8(_CgN18Q(
ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C.
l l
l ANSWER 4.09 (1.50)
- a. 2 county per second (0.50)
- b. Two acceptable answers: (0.50 for either one)
Ensures that the source range channel is seeing neutrons.
Ensures that a change in counts can be witnessed.
- c. 5-7 times (0.50)
REFERENCE SON G01-2, Para V.N.
SON Operator Certification Training, LP: G01-2 (week 6-8), objectives C k E O!!000K505 ...(KA'S)
ANSWER 4.10 (1.50) ,
s 100 degrees F per hour (0.50) 200 degrees F per hour (0.50) 560 degrees F (0.50)
REFERENCE eCN Technical Specifications 3/4.4.9 010000G005 010000G010 010000G011 ...(KA'S)
4 t._P80CEDUBES_ _N0806(i_@BN0806(i_EdE86ENCY_AND PAGE 61 RADIOLOGICAL CONTROL
)
l ESWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C. l I
i ?
I ANSWER 4.11 (2.50)
! a. Trip the reactor (and refer to Emergency Instructions as required) i j (0.50).
i i
- b. 1) Pressurizer heaters and pressurizer spray (0.50) (normal or auxiliary spray)
, 2) Centrifugal charging pnap and charging flow control valve
( F C's - 6 2 -9 3 C ) (0.50)
- 3) MDAFW and TDAFW pumps (0.50) i 41 Steam relief valves (0.50) r REFERENCE r SQNP A01-27A, Control Room Inaccessibility at Power or Hot Standby, Para.
III.A, IV SON, Operator Certification Training, LP: A01-27 (week 14-9), objective 1 000068G010 000068K201 ...(KA'S) i ANSWER 4.12 (1.00)
C REFERENCE SON, Operator Certification Training, LP: Status Trees (week 7-7),
objectives B, C, D 194001A102 ...(KA'S)
ANSWER 4.13 (1.00) d REFERENCE SON, Operator Certification Training, LP: Status Trees (week 7-7),
objectives B, C, 0 194001A102 ...(KA'S) t i I
F St._P,8QCEQUBES__NQ858(t_p,BN08d8(3_EdsSGEhCY_AND PAGE 62 B8D1969BIC66_CQNIBQL ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C.
ANSWER 4.14 (1.00)
(If the reactor can not be tripped.) the turbine will be needed to remove the heat generated by the reactor (1.00) l.5feAu huur3 Ate L im irea rre 4e % op 3MAM kow Ann 7u20.4s Muif Keu Atu Ou (Jurr_ Reu.roa. PDWd 15 Wdd W rHa HE M MMDVAt. f.APAuh REFERENCE o G %f ME Am DUMPS)
SON FD-S.1, Response to Nuclear Power Generation /ATWS SON, Operator Certification Training, LP: Function Restoration Guidelines (week 7-8), objective B 000029K312 ...(KA'S) ,
ANSWER 4.15 (1.50)
(1) Reinitiation of high pressure safety injection (0.50)
(2) Rapid secondary depressurization (0.50)
(2) RCP restart and/or opening pressurizer PORV's (0.50)
REFERENCE SON, Operator Certification Training, LP: Function Restoration Guidelines (week 7-8), objective C 000074K103 000074K311 ... KA'S)
ANSWER 4.16 (1.00)
Because all FRG's are written assuming that one AC emergency bus is .
energized.
REFERENCE SQN, Operator Certification Training, LF: Emergency Contingency Actions (week 7-9), objective C 000055K302 ...(KA'S) l l
}s PROCEDUREE - NORMALS _AgNQRMA(y_Ef5RGENCf_AND PAGE - 63 8801QLQGIC6(,CQN18Q(
L ANSWERS'--'SEQUOYAH-1&2 -87/11/17-SHIRAKl. C.
s ANSWER 4.17 (2.00)
'a. To prevent SI_ equipment from automatically actuating upon AC power restoration.(0.50)
Eb. To prevent nitrogen injection lato the RCS from the cold leg accumulators (0.50)
- c. Using reach rods (0.25) in the 480-V SD BD room (0.25)
- d. Continue depressurization-(0.50) 7 REFERENCE SON, Operator Certification Training, LP: Emergency Contingency Actions (week 7-9), objective C 000055K302 ...(KA'S)
ANSWER 4.19 (1.00)
- a. Proceed to the next step (or substep) in the lef t hand column (0.50)
- b. Ensure the task in progress is proceeding satisfactorily (0.50)
REFERENCE SON, Operator Certification Training, LP: Introduction to Energency Recovery Guidelines (ERG's) (week 7-1), objectives E and F 000007G012 194001A102 ...(KA'S)
ANSWER 4.19 (1.00) d REFERENCE SON FR-0 Status Trees SON, Operator Certification Training, LP: Status Trees (week ?-7),
objectives B and D 194001A102 ...(KA*S)
I I
$t..P8gCEDy8ES_ _N08d@(i_@@8QBd@(i_8dE8@EUCy_6ND PAGE 64 8691g6ggicet_tggI8g(
ANSWERS -- SEQUDYAH 1&2 -87/11/17-SHIRAKI, C.
i ANSWER 4.20 (1.00)
{
a REFERENCE SONP ES-0.3 SON, Operator Certification Training, LP: Emergency Instructions, E-0 (week 7-2), objective H 000038K103 19300BK122 ...(KA'S) f ANSWER 4.21 (1.00)
)
C hEFERENCE SONP A01-34, Energency Boration, Para IV, V 000024G010 000024K302 004010K609 ...(KA'S)
ANSWER 4.22 (1.00) b on a REFERENCE SON ADI-35, Loss of Offsite Power 064000K102 ...(KA'S) h
y .
t g U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SEQUOYAH 1&2 REACTOR TYPE: PWR-WEC4 _ ___
DATE ADMINISTERED: 87 /11/14Ll7 EXAMINER: BITTER. S CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
Uee separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY uh/ 25.00 ._ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 30.00 25.00 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 30.00 25.00 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 30.00 25.00 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 120. __
% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature s
,l 4 NRC RULES AND GUIDELINES FOR LICENSE' EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and 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.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face 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.
I
e- y
'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 exam.: nation questions,
- c. Turn in all scrap paper and the balance of the paper that you did not use-for answering the questions,
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
l l
l l
s 4
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION._ FLUIDS. AND PAGE 2 THERMODYNAMICS
/
QUESTION 5.01 (1.00)
From the four (4) phrases listed below, sele:t the one that correctly completes the following statement: "During e. reactor startno, just prior to reaching criticality, the SUR meter indication will respond to a given amount of rod withdrawal by ..."
a) "... rising slowly, then slowly falling off to zero "
b) "... rising rapidly, then slowly falling off to zera."
c)
.. rising slowly, then rapidly falling off to ze o."
d) ".. rising rapidly, then rapidly falling of f to z rco. "
QUESTION 5.02 (1.00)
A system containing saturated vapor at 585 psig is leaking via a crack in a pipe wall. The pressure downstream of the crack is 285 psig. Assuming ideal throttling characteristics, select the one correct statement that represents the condition of the fluid downstream of the crack.
NOTE: The steam tables should be used in lieu of the Mollier Diagram, a) Wet vapor with a very high quality (quality is near 100 percent).
b) Wet .apor with a quality of approximately 60 percent.
c) Superheatad vapor.
d) Saturated liquid.
_ (***** CATEGORY 05. CONTINUED ON DEXT PAGE *****)
s e
3
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 3 THERMODYNAMICS QUESTION 5.03 (1.00)
The first reactivity addition to a suberitical reactor increases the steady state count rate from 150 cps to 300 cps. The second reactivity addition to the reactor increases the steady state count rate from 300 cps to 600 cps.
From the following statements, select the one that correctly describes the relative magnitude of the reactivity additions.
a) The first reactivity addition was twice as large as the second.
b) The second reactivity addition was twice as large as the first.
c) Both reactivity additions were equal, d) The first reactivity addition was larger than the second; however, there is insufficient information to tell by how much.
QUESTION 5.04 (1.00)
Sequoyah Nuclear Plant, Unit I, is operating with the following parameters:
Pressurizer pressure : 440 psig Thot = 350 degrees F Tcold : 340 degrees F Tavg = 345 degrees F From the following values, select the one that represents the correct subcooling margin.
a) 104 degrees F b) 107 degrees F c) 109 degrees F d) 112 degrees F
._(..***** CATEGQRY 05 CONTINUED ON NEXT PAGE.*****)..
I
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 4 THERMODYNAMICS
/
QUESTION 5.05 (1.00)
From the four statements below, select the one that correctly describes the xenon transient following a power increase, a) Xenon initially increases due to the increased xenon fission yield and then decreases due to increases in both xenon decay and xenon burnout.
b) Xenon initially decreases due to increased xenon burnout and then increases due to increases in both the xenon fission yield and iodine decay, c) Xenon initially increases due to an in increase in iodine decay end then decreases due to increases in both xenon decay and xenon burnout, d) Xenon initially decreases due to a decrease in xenon fission yield and then increases due to increased iodine decay.
QUESTION 5.06 (1.00)
N41 N42 N43 N44 Upper Detector - Actual Current Value 159.7 0* 139.5 147.1 Lower Detector - Actual Current Value 166.0 0* 145.1 150.3 Upper Detector - 100% Current Value 266.6 252.7 262.9 254.3 Lower Detector - 100% Current Value 278.6 236.7 270.0 252.3 Using the above data, select the correct QPTB from the choices listed below.
NOTE: Power Range Detector N42 has been properly taken out of service due to instrument malfunction, a) 1.025 b) 1.043 c) 1.052 d) 1.079
( (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) j
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' 5'. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 5 THERMODYNAMICS
/
QUESTION 5.07 (1.00)
From the four (4) statements listed below, select the one that correctly completes the following sentence: "As rods are withdrawn, the magnitude of the Moderator Temperature Coefficient (MTC) ..."
a) " remains constant because MTC is a function of boron concentration and temperature only."
b) " becomes more or less negative depending upon core age."
c) " becomes more negative because fewer neutrons are absorbed by the control rods."
d) " becomes less negative because MTC is less negative for an unrodded core."
QUESTION 5.08 (2.00)
For each of the following, indicate whether the Departure from Nucleate Boiling Ratio will INCREASE, DECREASE or REMAIN THE SAME. Consider each case separately, a) One reactor coolant pump trips resulting in three loop power operation.
b) Reactor power decreases.
c) One main steam isolation valve inadvertantly shuts. (Assume rod control is in manual and no reactor trip occurs.)
d) Automatic pressurizer spray initiates.
(***** CATEGORY 05 CONTINUSD ON NEXT PAGE *****)
s
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 6
( THERMODYNAMICS QUESTION 5.09 (2.00)
An estimated critical position has been calculated for a reactor startup that is to be performed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a trip from a 100 day full power run.
For each of the following events / conditions, indicate whether the ACTUAL critical rod position is HIGHER THAN, LOWER THAN or the SAME AS the ESTIMATED critical rod position. Consider each event / condition separately, a) Steam dump pressure setpoint is increased by 30 psig.
b) Startup is delayed for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
c) The present boron concentre. tion is 30 ppm higher than that used in the ECP calculation, d) Condenser vacuum increases by one (1) inch Hg.
QUESTION 5.10 (2.00)
For each of the following, indicate whether the differential rod worth of an individual control rod will INCREASE, DECREASE, or REMAIN THE SAME.
Consider each case separately, a) An adjacent rod is inserted to the same height, b) Moderator temperature is increased.
c) Boron concentration is decreased, d) Fuel adjacent to the control rod is depleted.
QUESTION 5.11 (1.50)
Two identical reactors are taken critical. Reactor A has a rod speed of 40 steps per minute. Reactor B has a rod speed of 30 steps per minute.
Assuming a continuous rod withdrawal in each case, answer: A. B, or THE SAME to each of the following questions.
a) Which reactor will achieve criticality first?
b) Which reactor will have the highest critical rod height?
c) Which reactor will have the highest source range count at criticality?
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
t
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 7 THERMODYNAMICS QUESTION 5.12 (2.00)
For each of the following conditions, indicate whether the available NPSH of the centrifugal charging pumps will INCREASE, DECREASE, or REMAIN THE SAME, Consider each case separately.
a) During normal CVCS operation, VCT level increases from 20 percent to 41.
percent.
b) During normal CVCS operation, hydrogen pressure in the VCT is increased from 17 to 25 psig.
c) During normal CVCS operation, the temperature of the tube-side of the letdown heat exchanger decreases from 127 degrees to 122 degrees.
d) During emergency boration, the filter downstream of the boric acid transfer pump becomes partially clogged from boric acid precipitation.
QUESTION 5.13 (2.00)
For each of the following events / conditions, state whether the value of the estimated critical boron concentration will INCREASE, DECREASE or REMAIN THE SAME. Consider each case separately, a) Estimated startup time increases from 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the trip to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after the trip.
b) Desired critical rod height changes from 130 steps / Bank D to 140 steps / Bank D.
c) The anticipated Tavg at startup is changed from 547 degrees F to 544 degrees F.
d) Reactor power history prior to the last shutdown is corrected from 404 to 50% power.
i
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 6 THERMODYNAMICS QUESTION 5.14 (2.00)
For each of the following situations, indicate whether differential boron worth will INCREASE, DECREASE, or REMAIN THE SAME. Consider each case separately.
a) Fission product inventory in'c re a s e s ,
b) Boron concentration increases, c) Burnable poison inventory decreases, d) Rods are withdrawn.
QUESTION 5.15 (2.00)
Sequoyah Nuclear Plant, Unit 1, is operating at 30 percent power with rod control in manual and the main turbine in auto ("imp-in"). For each of the following parameters, state whether tripping the LOOP 1 reactor coolant pump would cause the parameter to INCREASE, DECREASE, or REMAIN THE SAME, Consider each case separately.
NOTE: In each case, assume that no reactor trip occurs and no operator action is taken.
a) Reactor coolant system (RCS) loop 2 flow.
b) No. 3 steam generator pressure.
c) RCS loop 1 hot leg temperature.
d) RCS loop 2 cold leg temperature.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 9 THERMODYNAMICS QUESTION 5.16 (2.00)
Reactors "A" and "B" have been operating at 100 percent and 50 percent steady state power, respectively. Each reactor has a boron concentration of 1500 ppm. A simultaneous reactor trip occurs in both plants; all rods are inserted. Reactor "A" rods insert 100 pcm more negative reactivity than reactor "B" rods. Post-trip average temperatures are equal and remain the same throughout the shutdown. Assume that no operator action takes place.
For each of the following, identify the plant that has a larger shutdown margin (Denote as "REACTOR A" or "REACTOR B"]:
NOTE: 1) Figures 5.1, 5.2, 5.3 and 5.4 are enclosed for reference.
- 2) Figures 5.1, 5.2, 5.3 and 5.4 are applicable to BOTH REACTORS.
a) One (1) minute after the trip.
b) Nine (9) hours after the trip.
c) Fifty (50) hours after the trip.
d) 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> after the trip.
QUESTION 5.17 (2.00)
Answer each of the following statements TRUE or FALSE.
a) For a constant boron concentration, differential boron worth will decrease over core life because of the buildup of fission products.
b) For a constant power level, differential boron worth will increase over core life because of the decreasing boron concentration.
c) Power defect at BOL is smaller (less negative) than at EOL because MTC is much smaller (less negative) at BOL than at EOL.
d) Dopphr-coeff4cie.m %es--lese-negat-ive-frorrBOb-to-MOL10e7odue1-4entif4 cation of new-h' and become1CYo7nega__tiym-from= HOI,-to-ECI,-due nd rescrance cagure in-Pu-249-end-Pu-240. --
G cool )3r f %lf1 )
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 10 THERMODYNAMICS QUESTION 5.18 (1.50)
Assume that the power range channels have been adjusted based on a calculated calorimetric. Answer each of the following statements TRUE or FALSE.
a) If the blowdown flow had been ignored in calculating the calorimetric, then actual reactor power would be higher than indicated reactor power.
b) If the feedwater temperature used in calculating the calorimetric had been 10 degrees lower than actual feedwater temperature, then actual reactor power would be higher than indicated reactor power.
c) If a main steam atmospheric relief valve had been leaking by during the data-taking portion of the calorimetric, then actual reactor power would be higher than indicated reactor power.
QUESTION 5.19 (2.00)
One indication that single-phase natural circulation is occurring in the primary is that "the hot leg RTD should be indicating either a steady value or a slowly decreasing value." State four other indications of single-phase natural circulation (NC).
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- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 11 QUESTION 6.01 (1.00)
From the four (4) statements below, select the one that IS NOT a consequence of placing the two-position rotary "LEVEL TRIP" switch, on the source range drawer, in the BYPASS position. Refer to Figure 6.1 as an aid, a) Continuous power is fed to the RPS to maintain voltage to UV coils, b) The OPERATION SELECTOR switch is "enabled" (capable of being used),
c) The LEVEL TRIP BYPASS indicating lamp, on the drawer, is illuminated, d) The CHANNEL ON TEST indicating lamp, on the drawer, is illuminated.
QUESTION 6.02 (1.00)
Sequoyah Nuclear Plant, Unit 1. is operating at 30 percent steady state reactor power. Instrument maintenance (IM) personnel receive permission to perform a calibration on the power range (PR) channel N-41. The IM mechanic mistakenly pulls the instrument power fuses to PR channel N-42; then, realizing the error, he reinserts the N-42 fuses. The mechanic then pulls the fuses for channel N-41; the reactor trips. From the following statements, select the one that correctly describes the reason for the trip, a) PR neutron flux low setpoint trip - block removed when two consecutive fuses are removed - 2/4 logic met, b) PR Rate Trip - N-42 PR channel not properly reset - 2/4 logic met, c) Urgent Failure - rod control power mismatch problem - 2/3 logic met.
d) PR neutron flux high setpoint trip - H-42 PR channel not properly reset - 2/4 logic met.
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- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 12 QUESTION 6.03 (1.00)
From the four components listed below, select the one that is NOT isolated / tripped on a phase B containment isolation signal, a) Thermal barrier booster pump.
b) Excess letdown heat exchanger.
c) Reactor coolant pump upper oil cooler.
d) Thermal barrier.
QUESTION 6.04 (1.00)
From the following list of diesel engine / generator shutdowns, select the one that is enabled during an emergency start of the diesel, a) Low lube oil pressure.
b) Phase balance relay (46).
c) Voltage restraint overcurrent relay (51V).
d) Generator differential relay (87).
QUESTION 6.05 (2.00)
For each of the following situations, indicate whether the OT delta T / OP delta T reactor trip setpoint will INCREASE, DECREASE or REMAIN THE SAME.
Consider each case separately, a) Tavg input to the OT delte T setpoint INCREASES.
b) Pressure input to the OT delta T setpoint INCREASES.
c) Delta flux (function) input to the OT delta T setpoint INCREASES.
d) Tavg input to the OP delta T setpoint INCREASES.
e) Delta flux (function) input to the OP delta T setpoint INCREASES.
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- 8. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 13 QUESTION 6.06 (1.50)
For each of the following valves, state the position in which the valve fails upon loss of electrical power / air. Use the terms OPEN, SHUT or AS IS.
a) RHR inlet isolation valve 74-1.
b) RHR pump suction valve 74-3.
c) Letdown isolation valve FCV-62-69.
d) Letdown orifice isolation valve FCV-62-73.
e) RHR crosstie valve FCV-74-33.
QUESTION 6.07 (1.00)
For each of the following, state whether the rod height at which the rod insertion limit alarm setpoint is reached will be HIGHER, LOWER or REMAIN THE SAME. Consider each case separately, a) Loop 2 reactor coolant pump trips resulting in three-loop power operations.
b) No. 2 steam generator main steam isolation valve inadvertantly shuts.
QUESTION 6.08 (2.00)
Answer each of the following statements TRUE or FALSE.
a) Containment sump isolation valve 1-FCV-63-72, must be fully closed (as indicated by both the stem switch and the gear switch) to meet one of the conditions for opening residual heat removal (RHR) inlet isolation valves 74-1 and 74-2.
b) RHR inlet isolation valves 74-1 and 74-2 are in a series, rather than parallel configuration, to each another, c) During RHR cooling, RHR inlet isolation valves 74-1 and 74-2, if opened, will automatically close when RCS pressure reaches greater than 700 psig, d) RHR inlet isolation valves 74-1 and 74-2 are normally closed while in modes 1, 2, or 3.
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- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 14 QUESTION 6.09 (2.00)
In reference to the bank overlap unit (BOU), answer each of the following TRUE or FALSE.
a) The BOU tells the master cycler which slave cycler to signal, b) The BOU functions only with the BANK SELECTOR switch in the "automatic" or "manual" positions.
c) The BOU receives feedback signals from the slave cyclers.
d) The BOU provides a multiplexing signal to the power cabinets.
QUESTION 6.10 (1.50)
For each of the following circuits, indicate whether the circuit receives individual upper and lower power range nuclear instrument (PRNI) inputs, summed upper and lower PRMI inputs or no PRNI input. Answer INDIVIDUAL, SUMMED or NO INPUT. Consider each case separately, a) PR neutron flux low setpoint trip, b) OT delta T calculation.
c) Channel current comparator, d) PZR level control circuitry.
e) M-4 delta flux meter.
f) Rod control power mismatch, g) High power trip.
h) Turbine / nuclear power "imp-in".
- 1) Detector current comparator.
j) Power range rate trip.
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- 8. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION QUESTION 6.11 (1.00)
The BP-30 process monitor module has a normally illuminated green lamp as one of it's indications. This green lamp will go out if the instrument malfunctions. List two (2) OTHER conditions that will cause the lamp to be "out".
NOTE: "Burnt out light bulb" is'not an acceptable answer.
QUESTION 6.12 (1.50)
List the five (5) automatic start signals for the motor-driven auxiliary feed pumps.
QUESTION 6.13 (1.50)
STATE three (3) possible causes for receiving an Solid State Protection System (SSPS) General Warning alarm.
NOTE: ANSWERS HAVING REDUNDANT TRAINS COUNT AS ONE ANSWER.
QUESTION 6.14 (2.00)
LIST the four (4) purposes of the Ice Condenser System.
QUESTION 6.15 (2.00)
Assume that for each of the following failures, no operator action is taken and the reactor subsequently trips. For each failure, state the protection signal that causes the trip (i.e. containment pressure SI). Consider each case independently.
NOTE: Assume reactor is operating at 100 percent power; and all other systems are in automatic and operating normally, a) The pressurizer level controlling channel fails high.
b) The pressurizer level controlling channel fails low, c) The pressurizer secondary level channel fails low.
d) The pressurizer pressure controlling channel fails high.
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- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE lt QUESTION 6.16 (2.00)
One of the functions of the reactor trip interlock, P-4, is to actuate a turbine trip. List four (4) other functions of the P-4 permissive.
QUESTION 6.17 (1.00)
Complate each of the following statements:
a) "The OT Delta T calculated reactor trip setpoint is designed to protect the core from ..."
b) "The OP Delta T calculated reactor trip setpoint is designed to protect the core from ..."
QUESTION 6.18 (2.00)
The design basis for the Sequoyah Nuclear Plant steam generator water leve; program includes various safety considerations for having a minimum as well as a maximum value of water level. State the safety considerations for botl the minimum and maximum steam generator water levels.
QUESTION 6.19 (2.00)
In reference to the CVCS:
a) STATE the interlock condition that must be met in order to manually close the letdown isolation valves FCV-62-69 and 62-70.
b) EXPLAIN the reason for having the interlock stated in part a.
QUESTION 6.20 (1.00)
Figures 6.2 through 6.5 depict various fuel handling tools. Complete the statements below with the name of the tools, a) Figure 6.2 is a .
l l b) Figure 6.3 is a .
c) Figure 6.4 is a .
l d) Figure 6.5 is a .
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- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 17 RADIOLOGICAL CONTROL
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QUESTION 7.01 (2.00)
In reference to the critical safety functions / trees, answer each of the following statements TRUE or FALSE.
a) ORANGE path actions for SUBCRITICALITY take priority over RED path actions for HEAT SINK.
b) The INTEGRITY critical safety function has priority over the CONTAINMENT critical safety function.
c) If a YELLOW condition is diagnosed for one of the critical safety functions, the operator (s) may choose whether to continue with the optimal recovery in progress or to initiate action to restore the critical safety function.
d) Once monitoring of the status trees has been initiated, they need to be monitored every 10 to 20 minutes during YELLOW or GREEN conditions.
QUESTION 7.02 (1.50)
For each of the statements below, fill in the blanks with one of the following facilities / centers:
- 2) Operations Support Center (OSC)
- 3) Local Recovery Center (LRC)
- 4) Site Decontamination Facility a) is an area containing equipment and supplies that are required for radiological cleanup and/or patient evaluation and stabilization.
b) The electrical, instrument and mechanical maintenance shops would be classified a/an upon activation of the Radiological Emergency Plan.
c) The health physics and the radiochemical labs would be classified as a/an upon activation of the' Radiological Emergency Plan.
d) may be used by the NRC during the event as an area near the site for assessment and assistance.
e) is located approximately 1.5 miles from the plant at Power Operations Training Center roome 64 and 65.
f) is the focal point of onsite activity and is the primary source of communication with offsite organizations during the event.
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- 7. PROCEDURES - NORMAL. ABliORMAL. EMERGENCY AND PAGE 18 RADIOLOGICAL CONTROL QUESTION 7.03 (1.00)
GOI-SC, Unit Shutdown From Full Power to Minimum Level, specifies the power level at which certain equipment is removed from service. For each situation below, state the power. level at which it occurs.
a) Simultaneously stop both operating condensate-demineralizer booster pumps.
b) Remove from service one of the No. 3 heater drain pumps, c) Remove one of the two operating condensate booster pumps, d) Remove from service one of two operating MFP turbines.
QUESTION 7.04 (2.00)
State the four (4) conditions, as required by SOI 62.1, "Chemical Volume Control System", that must be satisfied prior to opening the reactor coolant pump No. 1 seal bypass valve.
QUESTION 7.05 (2.00)
Function Restoration Guide FR-S.1,"Response to Nuclear Power Generation /ATWS", contains steps which must be performed in event of an "anticipated transient without scram" situation. List, in order, the four (4) immediate actions of FR-S.I.
QUESTION 7.06 (2.00)
One type of area that requires a radiation work permit (BWP) prior to entry is a high radiation area. State four'(4) other types of areas that require an RWP prior to entry.
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- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 1E RADIOLOGICAL CONTROL QUESTION 7.07 (3.00)
For each of the following three (3) situations, list the IMMEDIATE operator actions according to AOI-16, Loss of Normal Feedwater.
a) Loss of normal feedwater control, b) Loss of main feedwater pump control.
c) Loss of one main feedwater pump above 80 percent turbine power.
QUESTION 7.08 (1.00)
State the two (2) IMMEDIATE operator actions per AOI-6, Small Reactor Coolant System Leak.
QUESTION 7.09 (1.50)
Explain how a locked-closed valve is verified CLOSED. Assume that the locking device is locked and will not permit any movement.
QUESTION 7.10 (1.00)
GOI-1, Plant Startup from Cold Shutdown to Hot Standby, contains a precaution that states, "Avoid changing the BCS temperature by greater than 50 degrees F whenever the reactor trip breakers are closed and any control rod bank or shutdown bank is not withdrawn at least 5 steps." State the reason for having this precaution.
QUESTION 7.11 (1.50)
List the Sequoyah Nuclear Plant emergency exposure limits and explain the conditions under which each applies.
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- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 20 RADIOLOGICAL CONTROL J
QUESTION 7.12 (2.00)
Emergency Instruction E-0, "Reactor Trip or Safety Injection", contains immediate actions to be performed in event of a reactor trip or safety injection. For each of the following steps, state the basis for the action.
a) Step 2: Verify turbine trip.
b) Step 7: Verify main feedwater isolation, c) Step 8: Verify auxiliary feedwater status, d) Step 13: Check Tave.
QUESTION 7.13 (2.00)
Emergency Instruction E-0, "Reactor Trip or Safety Injection", contains four (4) general criteria for safety injection termination. State these four (4) criteria.
QUESTION 7.14 (2.00)
Concerning General Operating Instruction GOI-2, "Plant Startup From Hot Standby to Minimum Load", state the reason / basis for each of the following precautions.
a) All shutdown rods must be fully withdrawn before the reactor is critical. <
b) Do not exceed a steady startup rate of +1.0 DPM.
c) Control rod banks should be withdrawn or inserted in the prescribed sequence. Overlap of the control banks may not exceed the prescribed setpoints for automatic overlap.
d) The reactor shall be made critical only when there is a steam bubble in the pressurizer.
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- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 21 RADIOLOGICAL CONTROL QUESTION 7.15 (1.00)
General Operating Instruction GOI-1, "Plant Startup from Cold Shutdown to Hot Standby", contains many prerequisites and precautions for the unit heatup. State the reason for each of the following prerequisites /precau-tions.
a) Main steam isolation valve (MSIV) FCV-1-4, 11, 22 and 29, and MSIV bypass valves FCV-1-147, 148, 149 and 150 are closed. (Prerequisite) b) Whenever the plant is solid, tag out the power to the cold leg accumulator isolation valves, and pull-to-lock one of the two centrifugal charging pumps. (Precaution)
QUESTION 7.16 (1.00)
In Emergency Instruction E-3, Steam Generator Tube Rupture, step 5 requiree-that a rapid cooldown of the reactor coolant system be accomplished. State the reason for this step.
QUESTION 7.17 (1.50)
Functional Restoration Guideline, FR-C.1, Response to Inadequate Core Cooling, contains the major actions to be performed in order to restore adequate core cooling. State the three (3) different methods used to restore adequate core cooling in the order of priority.
QUESTION 7.18 (1.00)
One of the precautions of SOI-68.1, Reactor Coolant System, states:
"Reactor coolant pump seal injection is required during all reactor coolant system filling operations and when water level in the reactor coolant system is above the reactor coolant pump seal elevation."
State the basis for this precaution.
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- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 22:
RADIOLOGICAL CONTROL QUESTION 7.19 (1.00)
Define each of the following terms.
a) Radiation Area, b) High Radiation Area.
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. . 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 23 QUESTION 8.01 (1.00)
From the following statements, select the one that most accurately defines
- a. CHANNEL CALIBRATION.
a) the qualitative assessment of channel behavior during operation by observation."
b) the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors."
c) (For Analog Channels) " the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including aleem and/or trip functions.
d) (For Bistable Channels) "
the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions."
QUESTION 8.02 (1.00)
Sequoyah Nuclear Plant, Unit 1, has been operating at 100 percent rated thermal power for an extended period of time. Twelve hours ago, reeldual heat removal (RHR) heat exchanger A was declared INOPERABLE. The maintenance supervisor now reports that the suction valve from the containment sump to RHR pump B is INOPERABLE; you concur. From the following statements, select the one that correctly describes the allowances and/or limitations imposed by the Technical Specifications that apply in this situation.
NOTE: APPLICABLE TECHNICAL SPECIFICATIONS ARE ENCLOSED.
a) Suspend all operations involving reductions in reactor coolant system (RCS) boron concentration and immediately initiate corrective action to return loop to operation.
b) Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, action shall be initiated to place the unit in at least HOT STANDBY within the next six hours and at least HOT SHUTDOWN within the following six hours.
c) Because of the inoperability of either the RHR heat exchanger or the RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain RCS average temperature at less than 350 degrees F by use of alternate heat removal methods.
d) Power operations may continue without restrictions.
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- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 24 i
QUESTION 8.03 (1.00)
In reference to Technical Specification 3/4.7.12, Fire Barrier Penetrations, select the one phrase below that correctly completes the following statement: " All fire barrier penetrations (including fire door) in fire zone boundaries protecting safety related areas shall be functional..."
a) "while in Modes 1 and 2.~
b) "at all times."
c) "whenever reactor coolant system (RCS) average temperature is greater than 350 degrees F."
d) 'whenever BCS average temperature is greater than 140 degrees F."
QUESTION 8.04 (1.00)
Sequoyah Nuclear Plant, Unit 1 Technical Specifications require that the overtemperature delta T Channel Functional Test be accomplished on a "monthly" basis. The last three dates on which this surveillance was performed are August 10, September 10, and October 8. From the dates listed below, select the latest date on which this surveillance can be accomplished without exceeding the periodicity required by technical specifications.
NOTE: August has 31 days; September has 30 days; and October has 31 days.
a) November 7.
b) November 8.
c) November 15.
d) November 18.
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. 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 25 l
QUESTION 8.05 (2.00)
In reference to the Radiological Emergency Plan and Implementing Procedures Documents (REP-IPDs), answer each of the following statements TRUE or FALSE.
a) The positions and responsibilities of the "Site Director" and the "Site Emergency Director" are NOT the saro.
b) The Shift Engineer is responsible for dec:aring an emergency and providing the initial activation of the REP.
c) The "crashing" of a helicopter on site would be grounds for activating the REP.
d) Unusual aircraft activit'; over the Sequoyah facility would be grounds for activating the REP.
QUESTION 8.06 (2.00)
AI-18, File Package 18 "Notification and Licensing Event Report (LER)",
contains reporting requirements to the huclear Regulatory Commission (NRC).
Ctate the four (4) criteria requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification of the NRC as delineated in AI-18.
QUESTION 8.07 (1.00)
In accordance with 10CFR50, actions that violate technical specifications or license conditions may be taken to protect the safety and health of the public during an emergency situation. State whose authorization is required, as a minimum, to take such actions.
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- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 26
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QUESTION 8.08 (1.50)
Sequoyah Nuclear Plant, Unit 1, Technical Specification 3.4.0.2 "Operational Leakage" provides limits for reactor coolant system (RCS) leakage. For each of the following categories of RCS leakage, state the allowable limit specified in the technical specification.
a) Pressure boundary leakage, b) Any one steam generator, c) Controlled leakage (RCS pressure is 2235 psig) d) Identified leakage, e) Unidentified leakage.
QUESTION 8.09 (1.00)
In accordance with the Sequoyah Radiological Emergency Plan, list the four (4) classes of emergencies that are utili.med by TVA.
QUESTION 8.10 (1.00)
AI-2, "Authorities and Responsibilities for Safe Operation and Shutdown",
specifies those individuals who can grant permission to take the reactor critical. List these three (3) individuals by title.
QUESTION 8.11 (2.00)
Technical Specifications 3.1.1.4 specifies that the minimum temperature for criticality is 541 degrees F. State the five (5) bases for this limit.
QUESTION 8.12 (1.00)
Administrative Instruction AI-2, Authority and Responsibilities for Safe Operation and Shutdown, addresses the authority for the manipulation of
! controls. State WHEN and UNDER WHAT CIRCUMSTANCES an UNLICENSED individual is allowed to manipulate any control that directly affects reactor j reactivity or power level.
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i t . 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 27
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QUESTION 8.13 (1.00)
As directed by Sequoyah Nuclear Plant Hight Order Book, Attachment 1 (Information for Temporary Instructions for the Operations Group) dated June 20, 1987, the swing in-PZR level is limited to less than 3 percent when there is a steam bubble maintained in the PZR. STATE the reasor. for this administrative limit.
QUESTION 8.14 (1.00)
As defined by Sequoyah Nuclear Plant, Unit i Technical Specifications, one of the conditions necescary to have CONTAINMENT INTEGRITY is:
"All penetrations required to be closed during accident conditions are either:
- 1) Capable of being closed by an OPERABL8 containment automatic isolation valve system, or
- 2) Closed by manual valves, blind flanges or deactivated automatic valves secured in their closed positions, except as provided with Table 3.6.2 of Specification 3.6.3."
STATE the other four (4) conditions necessary to have CONTAINMENT INTEGRITY.
QUESTION 8.15 (1.00)
Sequoyah Nuclear Plant, Unit 1, Technical Specifications 3.11.2.6 states:
"The quantity of radioactivity contained in each gas decay tank shall be limited to less than or equal to 50,000 curies of noble gas (considered as Xe-133). State the basis for this technical specification.
QUESTICN 8.16 (1.00) l Technical Specifications 3.2.2 and 3.2.3 place operating limits on heat flux hot channel factor, RCS flow rate and nuclear rise hot channel factor.
State the bases for these limits.
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- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 28 l
QUESTION 8.17 (1.00)
At Sequoyah Nuclear Plant, Unit 1, a steam generator tube inspection is to be performed on one steam generator with 3375 U-tubes. In reference to this inspection, answer the following questions.
NOTE: APPLICABLE TECHNICAL SPECIFICATIONS ARE ENCLOSED.
a) State the number of tubes that are required to be inspected, b) After completion of this inspection, it is determined that 3 percent of all the U-tubes inspected are degraded and one U-tube is defective.
State the CATEGORY that the result of this inspection falls into and state ANY REQUIRED ACTIONS.
QUESTION 8.18 (1.00)
At Sequoyah Nuclear Plant, Unit 1, the f( ' lowing events have occurred in the past seven hours:
- 1) At 5:00 AM, one PORV began leaking by while a reactor startup was in progress (reactor power at 10 exp -8 amps).
- 3) At 9:30 AM, Unit 1 was operating at 75% rated thermal power.
- 5) At 10:15 AM, an attempt to close the bicek valve for the second PORV failed and the block valve was declared inoperable.
- 6) At 10:30 AM, a plant shutdown commenced.
- 7) At 12:00 PM (present time), th3 plant was placed in hot standby.
In reference to the above events, answer each of the following questions.
NOTE: APPLICABLE TECHNICAL SPECIFICATIONS ARE ENCLOSED, a) Were any technical specifications violated? If the answer is YES, explain.
b) By what time tomorrow must the reactor be in COLD SHUTDOWN 7 QUESTION 8.19 (1.00)
Sequoyah Nuclear Plant Radiological Control Instruction 13 requires that positive access control be maintained for an area where the radiation dose is greater than or equal to 1000 mrem /hr (other than containment). One method by which positive access control can be maintained is by the direct and continuous surveillance of all doors to the area by an individual designated by the Shift Engineer. State the other method by which positive access control can be maintained.
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- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 29
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-QUESTION 8.20 (1.00)
While Sequoyah Nuclear Plant, Unit 1, is operating at 100 percent rated thermal power, the isolation valves for accumulator 1 and accumulator 2 are discovered shut. The reactor operator immediately attempts to open both isolation valves, but both isolation valves remain shut. State the required actions as directed by Technical Specifications.
NOTE: APPLICABLE TECHNICAL SPECIFICATIONS ARE ENCLOSED.
QUESTION 8.21 (1.00)
Sequoyah Nuclear Plant, Unit 1, Technical Specification 6.7 specifies the actions required to be taken if a SAFETY LIMIT is violated. State the two actions that are required to be accomplished within the FIRST hour.
QUESTION 8.22 (1.00)
Sequoyah Nuclear Plant, Unit 1, Technical Specification 4.7.12 states that fire barrier protection in fire zone barriers protecting safety related areas shall be verified functional at least once per 18 months by visual inspection. State the other time technical specifications require that visual inspection of fire barrier penetrations is accomplished.
QUESTION 8.23 (1.00)
Sequoyah Nuclear Plant, Unit 1, Technical Specification 3,7.12 "Fire Barrier Penetrations" requires that specific actions be taken within one hour if one or more fire barrier penetrations is nonfunctional. State these required actions as directed by this technical specification.
QUESTION 8.24 (1.50)
One of the four (4) bases (per Sequoyah Nuclear Plant Unit 1 Technical Specifications) for requiring the OPERABILITY of the protection and ESF instrumentation systems and interlocks is to ensure that "the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint."
State two (2) of the remaining three (3) bases.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 30 QUESTION 8.25 (2.00)
For each of the following situations, indicate whether temporary shift relief is REQUIRED or NOT REQUIRED, a) A Unit Operator (UO) who performed shift relief in accordance with AI-5 at the beginning of a shift on Unit 2 is requested to relieve a UO on Unit i during the assigned shift, b) A UO who performed shift relief in accordance with AI-5 at the beginning of the shift on Unit 1 as BOP is requested to assume lead operator position on Unit I during the assigned shift.
c) A 00 who performed shift relief in accordance with AI-5 at the beginning of the shift on Unit 1 as lead operator is requested to assume BOP position on Unit i during the assigned shift.
d) The Assistant Shift Engineer (ASE) (SRO) who performed shift relief in accordance with AI-5 at the becinning of the shift on Unit 2 is requested to assume BOP on Unit i during the assigned shift.
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 31 THERMODYNAMICS ANSWBRS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 5.01 (1.00) b) (1.0)
REFERENCE SQN, Operr. tor Certification Training, Lesson Plan: Basic Nuclear Physics II (Week 1 'Z), Training Objective B; and Lesson Plan: Reactor Power III (Week 1-3), Training Objective A.
015000K506 192008K103 192008K105 192008K194 ...(KA'S)
ANSWER 5.02 (1.00) c) (1.0)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Thermodynamics, Fluid Flow, Heat Trensfer Review (Week 2), Enabling Objectives 10e and 12.
193003K125 193004K115 .. (KA'S)
ANSWER 5.03 (1.00) d) (1.0)
REFERENC.,
! SQN, Operator Certification Training, Lesson Plan: Basic Nuclear Ph/ sics II
! (Week 1-2), Training Objective B.
l 192003K101 ...(KA'S) t ANSWER 5.04 (1.00) l b) (1.0)
REFERENCE NONE l 000074A201 000074K101 ...(KA'S) '
l l
l
- 5. THRORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 32 THERMODYNAMICS
/
ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 5.05 (1.00) b) (1.0)
REFERENCE SQN, Prelicense Lesson Plan, Lesson Plan: Reactor Physics Review (Section 1), Training Objective P.
192006K106 ...(KA'S)
ANSWER 5.06 (1.00)
~
.et" (1.0)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Technical Specifications 3/4.2 (Week 11-1), Enabling Objective B.1; and Technical Specifications, Section 3/4.2, Power Distribution Limits.
192005K113 ...(KA'S)
ANSWER 5.07 (1.00) d) (1.0)
REFERENCE SQN, Prelicense Lesson Plan, Lesson Plan: Reactor Physics Review (Section 1), Training Objective G.
001000K526 ...(KA'S)
ANSWER 5.08 (2.00) a) DECPSASE 'O.5) b) INChEASE (0,5) c) DECREASE (0.5)
- d) DFCREASE (0,5) 8BFERENCE SQN, Operator Certifi~ cation Training, Lesson Plan: Thermodynamics, Fluid l Flow and Heat Transwer Review (Week 2), Enabling Objectives B.22, B.23 and B.24.
193008K105 ...(KA'S) i J
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 33 THERMODYNAMICS ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 5.09 (2.00) a) HIGHER THAN (0,5) b) HIGHER THAN (0.5) c) HIGHER THAN (0,5) d) SAME AS (0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Reactor Theory, Operator Application VI (Week 1-6), Objective B.4.
001010A207 001010K526 192006K110 192006107 ...(KA'S)
ANSWER 5.10 (2.00) a) DECREASE (0.5) b) INCREASE (0.5) c) INCREASE (0.5) d) DECREASE (0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Reactor Theory, Reactivity Coefficients IV (Week 1-4), Objective E.
001000K502 001000K509 192005K107 ...(KA'S)
ANSWER 5.11 (1.50) a) A (0.5) l b) THE SAME (0.5) l c) B (0,5)
REFERENCE l SQN, Prelicense Lesson Plans, Lesson Plan: Reactor Physics Review (Section l 1), Training Objectives C and D.
l 192003K101 192008K103 192008K104 ...(KA'S) l l
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 34 THERMODYNAMICS ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 5.12 (2.00) a) INCREASE (0.5) b) INCREASE (0.5) c) INCREASE (0.5) d) DECREASE (0.5)
REFERENCE SQN, Licensed Operator, Prelicense and Certification Training, Technical Staff and Manager's (Advanced Phase) Training, Lesson Plan: Thermodynamics-Fluid Flow and Heat Transfer Review (Section 2), Enabling Objective B.19.
191004K106 ...(KA'S)
ANSWER 5.13- (2.00) a) SEGRBASE # 0 5) Gr>A' g/f/S b)-DECREAORf W 0.5) _/
c ) -SEGREASEf>>v#(0. 5 )
d)_IJ10DEASE (0.5)
DE CKGASE REFERENCE SQN, Operator Certification Training, Lesson Plan: Reactor Theory, Operator Application (Week 1-6), Enabling Objective B.4.
192002K114 192006K107 ...(KA'S) l l
ANSWER 5.14 (2.00) a) DECREASE (0.5)
! b) DECREASE (0.5) l c) INCREASE (0.5)
- d) INCREASE (0.5) l REFERENCE SQN, Prelicense Operator Training, Lesson Plan
- Reactor Physics Review (Section 1), Training Objective N.
i l
1 t
l
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 35 THERMODYNAMICS ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 5.15 (2.00) a) INCREASE (0,5) b) DECREASE (0.5) c) DECREASE (0.5) d) DECREASE (0,5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: A01-5 "Operation with One Reactor Coolant Loop Out of Service" (Week 12-8), Training Objective C.
003000K501 003000K504 003000K505 ...(KA'S)
ANSWER 5.16 (2.00) a) REACTOR "B" (0.5) b) REACTOR "A" (0.5) c) REACTOR "B" (0.5) d) REACTOR "B" (0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Ope:stor Application VI (Week 1-6), Enabling Objective B.4.
192002K114 192006K107 . (KA'S)
/
A ro ANSWER 5.17 ( 2-00-)
a) TRUE (0.5) b) TRUE (0.5) c) TRUE (0.5)
REFERENCE bd/ 5p/) /t/v/F7 SQH, Operator Certification Training, Lesson Plan: Reactor Theory, Operator Application VI (Week 1-6), Enabling Objectives B.1, B.2 and B.3.
192004K103 192004K108 192004K107 192007K104 ...(KA'S)
c' ,
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 36 THERMODYNAMICS i
ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 5.18 (1.50) a) TRUE (0.5) b) FALSE (0.5) c) FALSE (0.5)
REFERENCE Sequoyah Nuclear Plant, Surveillance Instruction SI-78, Power Range Neutron Flux Channel Calibration by Heat Balance Comparison (Daily).
015000A101 015000K504 193007K106 193007K108 ...(KA'S)
ANSWER 5.19 (2.00)
Any four (4) (0.5 each) of the following six (6):
a) RCS delta T should be approximately 25 percent to 80 percent of full power as indicated by WR RTDs.
b) Core exit thermocouples should be indicating either a steady value or a slowly decreasing value, c) Steam pressure should follow reactor coolant temperatures.
d) Cold leg RTDs should be indicating either a steady value or a slowly decreasing value.
e) RCS Cold leg temperatures should be slightly higher than S/G temperature.
f) RCS should be subcooled - hottest RCS temperature is less than saturation temperature.
REFERENCE SQN, Mitigating Core Damage For Operator License Certification and Prelicense Training, Lesson Plan: Natural Circulation / Inadequate Core i
Cooling (Section 4), Training Objective D.
l 193008K122 ...(KA'S) i i
i 1
l
)
c' .
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 37 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S x
ANSWER 6.01 (1.00) d) (1.0)
REFERENCE SQN, Opeator Certification Training, Lesson Plan: Excore Nuclear Instrumentation (Week 6-3).
015000A403 015000K405 015000K406 ...(KA'S)
ANSWER 6.02 (1.00) b) (1.0)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Excore Nuclear Instrumentation (Week 6-3), Training Objective B.4; LER (August 85)
Sequoyah 2 001.
012000K401 015000K603 ...(KA'S)
ANSWER 6.03 (1.00) b) (1.0) i l REFERENCE SQN, Operator Certification Training, Lesson Plan: Component Cooling Water System (Week 4-12), Training Objective E.
, 008000K601 008000K604 013000K108 ...(KA'S) l l ANSWER 6.04 (1.00)
( d. (1,0)
REFERENCE l
SQN, Operator Certification Training, Lesson Plan: Diesel Generators (Week l 9-11), Training Objective C.
l 064000K402 ...(KA'S) l t
l l
o .
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 38 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER,-S ANSWER 6.05 (2.00) a) DECREASE (0.4)
'b) INCREASE (0.4) c) DECREASE (0.4) d) DECREASE (0.4)
DEGREASE- (0.4)
REktnIN THE CAM 6 FERENCE SQN, Operator Certification Training, Lesson Plan: Reactor Coolant Temperature Instrumentation (Week 4-6), Training Objective F.
012000K403 012000K611 ...(KA'S)
ANSWER 6.06 (1.50) a) AS IS (0.30) b) AS IS (0.30) c) SHUT (0.30) d) SHUT (0.30) e) AS IS (0.30)
REFERENCE SQN, Operator Certification Program, Lesson Plan: Residual Heat Removal System (Week 4-9), Training Objective D.1 and 2; Lesson Plan: Chemical and Volume Control System (Week 4-8), Training Objective F.1 and 2.
004000G004 004000G007 004000K603 005000G004 005000G007
...(KA'S)
ANSWER 6.07 (1.00) a) HIGHER (0.5) b) HIGHER (0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Rod Control System (Week 6-2), Training Objective D.
001000K504 ...(KA'S)
O L PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 39 AN3WERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 6.08 _ ,
(2.00) a ) -T !!U " (
b) TRUE (0,5) c) TRUE (0.5) d) TRUE (0,5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Residual Heat Removal (Week 4-9), Training Objectives B, D and E.
005000K401 005000K402 ...(KA'S)
ANSWER 6.09 (2.00) a) TRUE (0.5) b) TRUE (0.5) c) FALSE (0.5) d) TRUE (0,5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Rod Control System (Week 6-2), Training Objectives Al and A2.
001000G007 001000K203 001000K402 001000K403 ...(KA"S) l ANSWER 6.10 (1.50)
- a) SUMMED (0.15)
! b) INDIVIDUAL (0.15) c) SUMMED (0.15) l d) NO INPUT (0.15)
I e) INDIVIDUAL (0.15) f) SUMMED (0.15) j g) SUMMED (0.15)
I h) NO INPUT (0.15) i) INDIVIDUAL (0.15) j) SUMMED (0.15) l REFERENCE SQN, Operator Certification Training, Lesson Plan: Excore Nuclear l Instrumentation (Week 6-3), Training Objective D.
l 015000K603 ...(KA'S) l
1 v' .
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 40 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 6.11 (1.00)
Any two (2) (0.5 each) of the following three (3):
a) Broken cable, b) Switch in "TRIP ADJUST" position.
c) Loss of "HI-VOLTAGE".
REFERENCE SQN, Operator Certification Training, Lesson Plan: Radiation Monitoring Systems (Week 3-1), Enabling Objective B.9.
073000G007 073000G008 073000G009 073000G011 ...(KA'S)
ANSWER 6.12 (1.50)
All five required at 0.3 point each:
- 1) LO-LO level in any steam generator.
- 2) SIS.
- 3) Loss of off-site power after 25 seconds.
- 4) Loss of either MFP above 80 percent load.
- 5) Loss of both MFPs.
REFERENCE SQN, Operator Certification Training, Lesson Plan: Auxiliary Feedwater System (Week 5-10), Training Objective A.
061000K402 ...(KA'S) l l
l I
I l
l l
i l J
s
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 41 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 6.13 (1.50)
Any three (3) of the following at 0.50 points each:
a) Loss of either 48 VDC power supply, b) Loss of either 15 VDC power supply.
c) Loose printed circuit card.
d) INPUT ERROR INHIBIT switch in INHIBIT position.
e) SLAVE RELAY TEST MODE SELECTOR switch in TEST position, f) MULTIPLEXING BLOCKING switch in INHIBIT position.
g) PERMISSIVE switch not in OFF position.
h) MEMORY switch not in OFF position.
- 1) LOGIC A switch not in OFF position, j) Reactor trip bypass breaker being closed.
k) Blown fuse in the input relay 48 VDC ground circuit.
REFERENCE SQN, Operator Certification Training, Lesson Plan: Solid State Protection System (Week 6-6), Enabling Objective B.4.
012000G004 012000G007 012000G008 012000G009 ...(KA'S)
ANSWER 6.14 (2.00)
All four (4) at 0.5 points each:
- 1) Absorbs (thermal) energy released during LOCA to control the peak pressure.
- 2) Hold pressure at a low value for an extended period of time.
- 4) Maintains proper pH (9.0-9.5) to convert iodine to a nonvolatile form.
REFERENCE SQN, Operator Certification Training, Lesson Plan: Containment Systems (Week 4-10), Training Objective D.
025000G004 ...(KA'S)
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 42 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 6.15 (2.00) a) HI pressurizer level (0.5) b) HI pressurizer level (0,5) c) HI pressurizer level (0.5) d) Low pressurizer pressure or OT delta T (Either answer accepted for full credit) (0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Review of Instrument Failures (Week 6-15), Training Objective A.
011000A210 011000A211 035010A203 035010K101 ...(KA'S)
ANSWER 6.16 (2.00)
Any four (4) of the following at 0.5 points each:
- 2) Provides a signal to the SI block and reset logic.
- 3) Locks in the circuit to prevent re-opening the MFW valves that were shut by either an SI or high steam generator level actuation signal.
- 4) Provides a signal to the steam dumps so that the reactor trip controller controls the steam dumps vice the load rejection controller.
- 5) Provides a signal to the process racks to reduce the HI steam flow p its 7.ero load setpoint.
9f? Ghogram setpoint fechb cm o ;t Srps 7)fyovu% an o W!Istjy' l p)p 7 REFERENCE al %$ P-29 wu[ h SQN, Operator Certification Training, Lesson Plan: Solid State Protection System (Week 6-6).
013000K115 013000K401 013000K412 ...(KA'S) 1 1
ANSWER 6.17 (1.00) l a) ".. low DNBR (due to adverse combinations of high temperature, low pressure,high flux difference, and power)." (0.5) b) "... damage due to excessive reactor power output." (0.5)
REFERENCE l SQN, Operator Certification Training, Lesson Plan: Reactor Coolant Temperature Instrumentation (Week 4-6), Training Objective F.
012000K402 ...(KA'S) l l
I
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 43 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 6.18 (2.00) a) Minimum level:
- 1. ensures a heat sink is avaliable (for the RCS). (0.5)
- 2. ensures that AFW pumps have time to start (0.25) (and pump water into the S/G) to control the level before tubes are uncovered (0.25).
b) Maximum level:
- 1. ensures that, in the event of a steam line break inside containment, the containment would not be overpressurized (due to the mass of water in the S/G). (0.5)
- 2. ensures that, in the event of a steam line break (upstream of the MSIV at higher power), the resulting positive reactivity insertion due to an uncontrolled cooldown would be limited (by the mass of water in the S/G.) (0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Steam Generator Water Level Control System (Week 10-7), Training Objective A.
035000G004 035000G007 035010K405 ...(KA'S)
ANSWER 6.19 (2.00) a) All orifice isolation valves must be shut. (1.0) b) Having the interlock ensures that the regenerative heat exchanger (0.25) always has BCS system pressure in it (0.25) to prevent flashing (0.50).
REFERENCE SQN, Operator Certification Training, Leseon Plan: Chemical and Volume Control System (Week 4-8), Training Objective F.'1.
004020K403 ...(KA'S)
ANSWER 6.20 (1.00) a) Thimble plug handling tool. ( 0 . 6') . 2f 9# n/1/87 b) Burnable poison rod assembly (BPRA) handling tool. ( 0. f) .2 f
- "" ' I c) Irradiation sample handling tool. (0.51,cf d) New fuel assembly handling tool. (0.S).If"lg;}{}{' }
REFERENCE c ye- 6n SQN, Fuel Handling Instruction (FHI) 6. t wraan 034000G007 0340000009 034000G013 034000K601 ...(KA' )
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 44 RADIOLOGICAL CONTROL ANSWERS -- SEQUOYAH 1&2 -87/11/16-BI TER, S ANSWER 7.01 (2.00) a) FALSE (0.5) b) TRUB (0.5) c) TRUE (0,5) d) TRUE (0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Status Trees (Week 7-7),
Training Objectives B and D.
194001A102 ...(KA'S)
ANSWER 7.02 (1.50) a) Site Decontamination Facility (0.25) b) Operations Support Center (0.25) c) Operations Support Center (0.25) d) Local Recovery Center (0.25) e) Local Recovery Center (0.25) f) Technical Support Center (0.25)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Radiological Emergency Plan (Week 8), Training Objectives A, B.2-3 and B.10-13.
194001A116 ..(KA'S)
ANSWER 7.03 (1.00) a) 60 percent (0.25) b) 60 percent (0.25) c) 30 percent (0.25) l d) 45 percent (0.25)
REFERENCE SQN, GOI-5C; and SQN, Operator Certification Training, Lesson Plan: 001-5C (Week 6-13), Enabling Objective B.2, 194001A110 194001A113 ...(KA'S)
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 45 RADIOLOGICAL CONTROL ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 7.04 (2.00) a) Beactor coolant system pressure greater than 100 psig and less than 1000 psig. (0.5) b) No. 1 seal leakoff valve is open. (0.5) c) No. 1 seal leakoff flowrate is less than one gpm. (0,5) d) Seal injection water flow rate to each pump is greater than 6 gpm.
(0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: SOI 62.1 "Chemical Volume Control System" (Week 9-3), Training Objective D.
003000G010 003000K602 004010A407 004020K607 ...(KA'S)
ANSWER 7.05 (2.00) 0.45 points for each correct response, and 0.05 points for each response in the correct order, a) Open reactor trip breakers.
b) Borate reactor coolant system, c) After reactor trip, verify turbine trip.
d) Ensure all AFW pumps running.
REFERENCE SQN, Operator Certification Training, Lesson Plan: Function Restoration l Guide (Week 7-8), Training Objective A.
- 000029G010 000029K312 ...(KA'S)
ANSWER 7.06 (2.00) l Any four of the following answers at O'.5 points each.
a) Airborn Redioactive Areas.
b) Contaminated Areas.
c) Areas posted "RWP Required for Entry".
d) Any area which Health Physics determines a need for a RWP.
o) Areas in which work activity may cause an increase in the radiological l health hazards (i.e. grinding on contaminated materials).
l f) Radiation area where the workers are expected to receive 50 mrem per l day. ,
REFERENCE SQN, Operator Certification Training, Health Physics Lesson Plan:
Principles of Radiation Protection (Section 10), Training Objective 1.
194001K103 ...(KA'S)
I
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 46 RADIOLOGICAL CONTROL ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 7.07 (3.00) a) Place MFW regulating or bypass valve in MANUAL (0.5) and return S/G 1evel to program (0.5).
b) Place MFW pump speed control in MANUAL (0.5) and return S/G level to program (0.5).
c) (If above 60 percent power), ensure that turbine runs back to 75 percent power. (1,0)
REFERENCE SQN, Operator Certification Training, Lesson Plan: AOI-16 "Loss of Normal Feedwater" (Week 11-6), Enabling Objectives B.1, B.2 and B.3.
000054G010 ...(KA'S)
ANSWER 7.08 (1.00) a) If pressurizer level is falling (0.25), (then) start centrifugal charging pumps as necessary to maintain level (0.25).
b) If loss of pressurizer level is imminent (0.2), (then) trip the reactor (0.15), initiate SI (0.15), (and go to E-0).
REFERENCE SQN, Operator Certification Training, Lesson Plan: A01-6 "Small Reactor Coolant System Leak" (Week 7-4), Enabling Objective B.3.
l 000009G010 ...(KA'S) l ANSWER 7.09 (1.50)
- 1) Remove the locking device. (0.3)
- 2) Attempt to move valve operator in the CLOSED position. (0.3)
- 3) Reinstall locking device. (0.3)
- 4) Verify it is locked (correctly reinstalled / locked). (0.3) l
- 5) Second person must verify that locking device is (reinstalled and) locked. (0.3) t l REFERENCE l SQN, GOI-6, page 4 of 12; and SQH, Operator Certification Training, Lesson Plan: GOI-6, Apparatus Operation (Week 6-14).
194001K101 ...(KA'S) l
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 47 RADIOLOGICAL CONTROL i
ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 7.10 (1.00)
(It limits the) possibility of thermal lockup of the rods. (1.0)
REFERENCE SQN, Operator Certification Training, Lesson Plan: GOI-1 (Week 6-7),
Terminal Objective C.
001000G010 ...(KA*S)
ANS 7.11 (1.50)
- bs-New 'O.25). for planned exposure during an eme rsency situat-ion-( 0. 25 ) .
25 Rem (0.261: to prevent serious damage to the plant or hazard to 2f personnel (0.G5-).,W 75 Rem ( 0. Gfr) : to say a_ life (0.25-).
35 WS / 2/'t
/ 27}
NO SQN, Operator Certification Training, Health Physics, Lesson Plan:
Radiation Standards and Guidelines (Section 5), Training Objective B.
194001A116 194001K103 ...(KA'S)
ANSWER 7.12 (2.00) a) Prevent an uncontrollable cooldown of the reactor coolant system (due to steam flow to the turbine). (0.50) b) Prevent overfilling of steam generators and an associated excessive RCS cooldown which could aggravate the transient (especially if it were a steam line break). (0.5) c) Provide feed to the RCS for adequate decay heat removal. (0.5) d) (RCS temperature stable at or trending to the no load value) indicates that the secondary steam dump system is operational as designed as a secondary heat sink. (44)
(o.t9 40 M &.8, wksu/ 2 al & As n REFERENCE &' mf lee (0.15-) @ / >/r/e?
SQN, Operator Certification Training, Lesson Plan: Emergehey natruction E-0 "Reactor Trip or Safety Injection" and Subsections ES-0.1, 8S-0.2 and ES-0.3 (Week 7-2), Training Objective F.
000007K301 ...(KA'S)
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 48 RADIOLOGICAL CONTROL v
ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 7.13 (2.00) a) Reactor coolant system (RCS) is subcooled (greater than 40 degrees F).
(0.5) b) RCS pressure is stable or increasing. (0.5) c) Adequate secondary heat sink exists (total AFW flow is greater than 440 gpm or narrow range level in at least one steam generator is greater than 10 percent). (0.5) d) Pressurizer level is indicating on level span (greater than 20 percent).
(0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Emergency Instruction E-0 "Reactor Trip or Safety Injection" and Subsection ES-0.1, ES-0.2 and ES-0.3 (Week 7-2), Training Objective G.
000008K305 000009G007 000009K325 000038K309 ...(KA'S)
ANSWER 7.14 (2.00) a) The shutdown rods must be fully withdrawn to ensure that the reactor will be shutdown in event of a reactor trip. (0.5) b) To ensure the operator has control of the reactor. (0.5) c) Proper sequence and overlap is required as part of the assumption in the accident analysis of reactivity addition rates for control rods. (0.5) d) A critical reactor has the capability of heating the moderator suddenly, therefore a bubble is required to prevent overpressure conditions in the RCS. (0,5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: General Operating Instruction GOI-2 "Plant Startup from Hot Standby to Minimum Load" (Week 6-8), Training Objective C.
001000G010 001010K501 001010K506 001010K536 ...(KA'S)
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 49 RADIOLOGICAL CONTROL ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 7.15 (1.00) a) (Isolation of the main steam l'ines) ensures that no accidental cooling of the reactor coolant system (RCS) can occur, (0.5) b) Minimize the possibility of overpressurization of the RCS from an safety injection signal. (0.5)
REFERENCE SQN, Operator Certification Training, Leeson Plan: General Operating Instruction GOI-1 "Plant Startup from Cold Shutdown to Hot Standby" (Week 6-7), Training Objective C.
002000G010 035000G010 ...(KA'S)
ANSWER 7.16 (1.00)
Ensure that there will be adequate subcooling margin (0,5) for a subsequent reactor coolant system pressure reduction (0.5). (Pressure reduction is required to reduce leakage rate.)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Emergency Instruction E-3 "Steam Generator Tube Rupture" (Week 7-5), Training Objective G.
000038K306 ...(KA'S)
ANSWER 7.17 (1.50) l l 0.1 point for each action in the correct position.
a) Reinitiation of high pressure safety injection. (0.4) i b) Rapid secondary depressurization. (0.4) c) Reactor coolant pump restart and/or opening pressurizer PORVs. (0.4) l REFERENCE I
SQN, Operator Certification Training, Lesson Plan: Functional Restoration l Guidelines (Week 7-8), Training Objective C.
000074K311 ...(KA'S) l l
l l
I l
o c
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 50 RADIOLOGICAL CONTROL ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 7.18 (1.00)
(The seal area offers a low point' in a portion of the reactor coolant system piping and continuous seal injection will) keep any suspended solids from settling out (0.50) and possibly blocking flow (0.25) or contaminating seal faces (0.25).
REFERENCE SQN, Operator Certification Training, Lesson Plan: System Operating Instruction SOI-68.1 "Reactor Coolant System" (Week 9-2), Training Objective A.
003000G010 003000K602 ...(KA'S)
ANSWER 7.19 (1.00) a) Any area (accessible to personnel, in which there exists radiation, originating in whole or in part within the license material) at such levels that a major portion of the body could receive (0.1) in any one hour a dose in excess of 5 mrem (0.2), or in any 5 consecutive days, a dose in excess of 100 mrem (0.2).
b) Any Area (accessible to personnel, in which there exist radiation, originating in whole or in part within the license material) at such levels that a major portion of the body could receive (0.1) in any one hour a dose in excess of 100 mrem (0.4).
REFERENCE SQN, Operator Certification Training, Lesson Plan: 10CFR20 "Standarda for Protection Against Radiation" (Week 15-2), Training Objective E.
194001K103 ...(KA'S)
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 51 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 8.01 (1.00) b) (1.0)
REFERENCE Sequoyah Nuclear Plant, Unit 1, Technical Specifications, Section 1.0, Definitions.
012000A301 012000G001 012000K101 012000K602 012000K605
...(KA'S)
ANSWER 8.02 (1.00) b) (1.0)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Technical Specification 3/4.5 "Emergency Core Cooling System" (Week 11-4), Training Objective B.3.
006000G005 006000G011 ...(KA'S)
ANSWER 8.03 (1.00) b) (1.0)
REFERENCE SQN, Prelicense Training, Lesson Plan: Technical Specifications 3/4.7.12 "Plant Syatems" (Week 6-7), Enabling Objective B.1.
000067A215 000067G003 000067G007 000067G008 ...(KA'S)
ANSWER 8.04 (1.00) c) November 15 (1.0)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Surveillance Program SI-1 (Week 9-12), Enabling Objective B.4 and Docket 50-327, LER 86-034.
012000G011 ...(KA'S)
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 52 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 8.05 (2.00) a) TRUE (0.5) b) TRUE (0.5) c) TRUE (0,5) d) TRUE (0,5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Radiological Emergency Plan (Week 8), Training Objectives A, B.6 and B.7; and Implementing Procedures Document IP-1.
194001A116 ...(KA'S)
ANSkER 8.06 (2.00)
All four answers at 0.5 points each, a) Exposure of the whole body of any individual to 5 rems or more of radiation; exposure of the skin of the whole body or any individual to 30 rems or more of radiation; or exposure of the feet, ankles, nands or forearms to 75 rems or more of radiation, b) The release of radioactive material in concentrations which, if averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, would exceed 500 times the limits specified for such materials (in 10 CFR 20, App. B, Table II).
c) A loss of one day or more of operation of any facilities affected, d) Damage to property in excess of $2,000.
REFERENCE SQN, Operator Certification Training, Lesson Plan: AI-18, File Package 18 "Notification and License Event Report (LER)" (Week 15-1), Enabling Objectives B.3.
000060G002 ...(KA'S)
ANSWER 8.07 (1.00)
(1.0)
Licensed Senior Operator.
REFERENCE SQN, Operator Certification Training, Lesson Plan: 10CFR50, (Week 15-3)
Training Objective H.
001000G001 ...(KA'S)
- 8. ADMIiilSTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 53 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 8.08 (1.50) a) zero leakage. (0.30) b) 500 gpd. (0.30) c) 40 gpm (0.30) d) 10 gpm. (0.30) e) 1 gpm. (0.30)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Yechnical Specification 3/4.4 "Reactor Coolant System" (Week 10-5), Enabling Objective B.1; and Lesson Plan: AOI-6 "Small RCS Leak /LOCA Discussion" (Week 7-4), Enabling Objective B.1.
002000G005 ...(KA'S)
ANSWER 8.09 (1.00) a) Alert. (0.25) b) Notification of Unusual Event. (0.25) c) Site Area Emergency. (0.25) d) General Emergency. (0.25)
REFERENCE SQN, Operator Certification Training, Leccon Plan: Sequoyah Radiological Emergency Plan (Week 8), Enabling Objective B.S.
000036G002 C00060G041 000067G002 000076G002 ...(KA'S)
ANSWER 8.10 (1.00) a) Plant Manager. (0.34) b) Plant Superintendant. (0.33) c) Operations Supervisor. (0.33)
REFERENCE SQN, Operator Certification Training, Lesson Plan: AI-2 "Authorities and Responsibilities for Safe Operation and Shutdown" (Week 10-1), Enabling Objective B.4.
00100CG001 ...(KA'S)
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 54 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 8'11 (2.00)
(This limit ensures that:)
a) The moderator temperature coefficient is within its analyzed temperature range. (0.4) b) The protective instrumentation is within its normal operating range.
(0.4) c) The P-12 interlock is abovc its setpoint. (0.4) d) The pressurizer is capable of being in an OPERABLE status with a steam bubble. (0.4) e) The reactor pressure vessel is above its minimum BT-NDT temperature.
(0.4)
REFERENCE SQN, Prelicense Training, Lasson Plan: Technical Specification 3/4.1 "Reactivity Control System" (Week 6-1), Enabling Objective B.2.
001000G006 .. (KA'S)
ANSWER 8.12 (1.00)
When it is part of the individual's training to qualdfy for an operator's license (0.5); and only if under the direct supervision of a licensed operator (0,5).
REFERENCE SQN, Operator Certification Training, Lesson Plan: AI-2 "Authorities and Responsibilities for Safe Operation and Shutdown" (Week 10-1), Enabling Objective B.5.
001000G001 ...(KA'S)
ANSWER 8.13 (1.00)
"A concern (has been expressed) that the pressurizer surge line may exceed the temperature limits covered in the (reference) tech spec." (1.0)
REFERENCE SQN, Night Order Book, Attachment 1 "Information for Temporary Instructions for the Operatons Group" dated June 20, 1987.
000028G003 ...(KA'S)
^
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 55 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 8.14 (1.00) a) All equipment hatches are closed and sealed. (0.25) b) Each air lock is OPERABLE (~ pursuant to Specification 3.6.1.3).
(0.25) c) The containment leakage rates are within the limits of Technical Specifications (Specification 3.6.1.2). (0.25) d) The sealing mechanism associated with each penetration is OPERABLE.
(0.25)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Technical Specification Overview (Week 9-1), Training Objective C.7.
000069A201 ...(KA'S)
ANSWER 8.15 (1.00)
(Restricting the q'lantity of radioactivity contained ir. each gas decay tank prevides assurance that) in event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. (1.0)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Technical Specification 3/4.11 "Radioactive Effluents" (Week 14-7), Enabling Objective B.2.
071000G006 ...(KA'S)
ANSWER 8.16 (1.00)
Ensure that:
a) design limits on peak local power density and minimum DNBR are not exceeded; (0,5) b) in event of a LOCA, the peak fuel clad temperature will not exceed the 2200 degrees F ECCS acceptance criteria. (0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Technical Specification 3/4.2 "Power Distribution Limits" (Week 11-1), Enabling Objective B.2.
002000G006 ...(KA'S)
t
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 56 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 8.17 (1.00) a) 405 U-tubes must be inspected. (0,5) b) Category C-2. Plug the defective tube and inspect an additional 810 U-tubes in this steam generator. (0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Technical Specification 3/4.4 "Reactor Coolant System" (Week 10-5), Enabling Objective B.1.
002000G011 .(KA'S)
ANSWER 8.18 (1.00) a) No technical specifications were violated. (0.5) b) By 1MS--FM _tommorrow , the plant must be in cold shutdown. (0,5)
REFERENb SQN, Operator Certification Training, Lesson Plan: Technical Specification 3/4.4 "Reactor Coolant System" (Week 10-5), Enabling Objective B.3.
002000G005 010000G005 ...(KA'S)
ANSWER 8.19 (1.00)
Doors leading to the area are locked (0.5) (with Rad Tumbler or Bad Security Lock) and keys to the locks are maintained by the Shift Engineer and Health Physicist (0,5).
REFERENCE SQN, Operator Certification Training, Lesson Plan: 10CFR20 "Standards for Protection Against Radiation" (Week 15-2), Training Objectives E, F and G.
194001K103 ...(KA'S)
ANSWER 8.20 (1.00)
Be in hot standby within one hour and be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. (1.0)
REFERENCE SQN, Operator Certification Trainin6, Lessan Plan: Technical Specification 3/4.5 "Emergency Core Cooling System" (Week 11-4), Enabling Objective B.3.
006050G005 ...(KA'S) t
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4 ~
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 57 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, S ANSWER 8.21 (1.00) a) The unit shall be placed in at least hot standby. (0,5) b) The NRC Operations Center shall be notified (by telephone as soon as possible and in all cases within one hour). (0.5)
REFERENCE SQN, Operator Certification Training, Lesson Plan: Technical Specification 2.0 "Safety Limits and Limiting Safety Settings" (Week 10-8), Terminal Objective A; and SNP Technical Specification 6.7.
002000G003 002000G005 ..(KA'S)
ANSMER 8.22 (1.00)
(Prior to returning a fire barrier penetration to functional status) following repairs or maintenance. (1.0)
REFERENCE SQN, Prelicense Training Program, Lesson Plan: Technical Specification 3/4.7 "Plant Systems" (Week 6-7), Terminal Objective A.
086000G005 194001K116 ...(KA'S)
ANSWER 8.23 (1.00) a) Establish a continuous fire watch (0.25) on at least one side of the affected penetration (0.25)
(or) b) Verify the operability of fire detectors ( 20) on at least one side of the nonfunctional fire barrier (0.10) and establish a hourly fire watch patrol (0.20).
REFERENCE SQN, Prelicense Training Program, Lesson Plan: Technical Specification 3/4.7 "Plant Systems" (Week 6-7), Terminal Objective A.
086000G005 194001K116 ...(KA'S)
e., .
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 58 ANSWERS -- SEQUOYAH 1&2 -87/11/16-BITTER, C ANSWER 8.24 (1.50)
Any two of the following three for 0.75 points each.
a) (It ensures that the specified) coincidence logic is maintained, b) (It ensures that sufficient) redundancy is maintained to permit a channel to be out of service for testing or maintenance.
c) (It ensures that sufficient) system functional capability is available for protective and ESF purposes from diverse parameters.
REFERENCE SQN, Prelicense Training, Lesson Plan: Technical Specification 3/4.3
' Instrumentation" (Week 6-3), Enabling Objective B.2.
012000G006 013000G006 ...(KA'S)
ANSWER 8.25 (2.00) a) REQUIRED (0.5) b) NOT REQUIRED (0.5) c) NOT REQUIRED (0.5) d) REQUIRED (0.5)
REFERENCE SQN, Prelicense Training, Lesson Plan: AI-5 "Shift and Relief Turnover" (Week 3-2), Enabling Objective B.2.
002000G001 194001A110 194001A111 ...(KA'S)
TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE .,
05.01 1.00 DCLOOO4999 05.02 1.00 0CL0005000 05.03 1.00 DCLOOO5001 05.0< 1.00 DCL0005003 05.05 1.00 DCLOOO5005 05.06 1.00 D('LOOO5007 05.07 1.00 DCL2005008 05.08 2.00 DCLOOO4994 05.09 2.00 DCL0004995 05.10 2.00 DCL0004996 05.11 1.50 DCLOOO4998 05.12 2.00 DCLOOO5002 05.13 2.00 DCL0005004 05.14 2.00 DCLOOO5096 05.15 2.00 DCL0005009 05.16 2.00 DCL0005011 05.17 /.sc-E-0@ DCLOOO4997 05.18 1.50 DCLOOO5012 05.19 2.00 DCL0005010 h n/h>
06.01 1.00 DCLOOO5022 06.02 1.00 DCLOOO5014 06.03 1.00 DCL0005018 06.04 1.00 DCL0005031 06.05 2.00 DCLD005015 06.06 1.50 DCLOOO5028 06.07 1.00 DCLOOO5032 06.08 2.00 DCLOOO5017 06.09 2.00 DCL0005023 06.10 1.50 DCLOOO5013 06.11 1.00 DCLOOO5020 06.12 1.50 DCL0005024 06.13 1.50 DCLOOO5026 06.14 2.00 DCLOOO5027 06.15 2.00 DCLOOO5029 06.16 2.00 DCLOOO5030 06.17 1.00 DCLOOO5016 06.18 2.00 DCLOOO5019 06.19 2.00 DCL0005025 06.20 1.00 DCLD005021 30.00 07.01 2.00 DCLOOO5034 07.02 1.50 DCLOOO5038 07.03 1.00 DCLOOO5036 07.04 2.00 DCLOOO5041 07.05 2.00 DCLOOO5042
A., .
TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE >
07.06 2.00 DCLOOO5045 07.07 3.00 DCL0005050 07.08 1.00 DCLOOO5051 07.09 1.50 DCLOOO5033 07.10 1.00 DCL0005035 07.11 1.50 DCLOOO5037 07.12 2.00 DCLOOO5040 07.13 2.00 DCLOOO5043 07.14 2.00 DCLOOO5044 07.15 1.00 DCLOOO5046 07.16 1.00 DCLOOO5047 07.17 1.50 DCLOOO5048 07.18 1.00 DC's0005049 07.19 1.00 DJL0005039 30.00 08.01 1.00 DCLOOO5053 08.02 1.00 DCLOO95054 08.03 1.00 DCLOOO5055 08.04 1.00 DCL0005068 08.05 2.00 DCLOOO5058 08.06 2.00 DCL0005059 08.07 1.90 DCLOOO5064 08.08 1.50 DCL0005065 08.09 1.00 DCL0005069 08.10 1.00 DCLOOO5070 08.11 2.P0 DCLOOO5077 08.12 1.00 DCLOOO5052 08.13 1.24 DCLOOO5056 08.14 1.00 DCL0005057 08.15 1.00 DCL0005060 08.16 1.00 DCLicG05061 08.17 1.00 DCL0005062 08.18 1.00 DCLOOO5063 08.19 1.00 DCLOOO5066 08.20 1.00 DCL0005067 08.21 1.00 DCL0005071 08.22 1.00 DCLOOO5072 08.23 1.00 DCLOOO507J 08.2'. 1.50 DCLOOO5074 08.25 2.00 DCL0005075 30.00 l ______
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, . ENCLOSURE 3-Written Examination Review Comments GUESTION Jf01 (1.50) Part b. only DIFFERENTIAL rod worth varies as a function of cor.itions in the core.
Considenng each case belowindependently, does the DIFFERENTIAL rod worth INCREASE, DECREASE, or ret 4AIN THE SA!4E?
- b. The rods are withdrawn from 150 steps to maintain temperature constant, with boron held Constant.
ANSWER
- b. Decrease (0.50)
REFERENCE SON, Operator Cernfication Trainmg, Reactor Theory , Reactivity Coefficients IV, Week 1-4, Trainmg objective V.E 192005K107 . (KA'S)
RESPONSE
INCREASE should also be accepted as an appropnate response because 6fferentialrod worth curves for an EOL core indicate an increas e in ifferential rod worth in the area above 150 steps. Refer to the atached copy of a iff erential rod worth curve for an EOL core for supporting documentation.
i
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. -OUESTION 1 21 (1.00) .
,Which of the following parameters has NO effect on the net positive suction head (NPSH) of acentrifugalpump.
- a. ' The height of a column of water above the eye of the pump.
- b. The amount of subcooling of the column of water above the eye of the pump.
- c. Pressure in the tank thatis the water supply to the pump.
- d. Number of pumps operating in parallel. '
' ANSWER ,
- d. -
-REFERENCE Heat transfer, thermodynamics, fluid flow fundamentals Section 111, Part B, Chapter 1 191004K106 (KA'SJ
RESPONSE
4 Reference indicated does not deal with multi-pump operation .
' Classroom tralmng involves the use of a fluid flow test bench to illustrate multi-pump operation
, (both parallel and series). With total system flowrate held constant or allowed to inerease some with the same system resistance, parallel pump operation results in increased NPSH to each pump due to the reduced flowrate through each pump.Fbw through each individual pump would be less than 1/2 ofits originalvalue.Therefore addmg or removing pumps change the NPSH to the other runrdng ;
pumps. Refer to the attached head / flow curves used to teach pump operation.
f
s g OUESTION 121 (1 50) Part a. only Answer the followmg questions in reference to the subcoolmg margin of the plant.
- a. Whatis the subcooling of the plantif the following conditions exist Thot = 587F Tavg = 572 F Tcold = 557 F Ppzr = 2235 prig Psg = 1033 psig ANSWER
- a. Tsatfor2250psis (2235 psig)
= 652.67 F (0.50)
Subcooling margin = Tsat-Thot = 652.67 - 587 = 65.67 F (0.50)
RESPONSE
Unlizing the ASME Steam Tables ,Imeat interpolation errors amount to an error band of approximately + or .05 F.
Using Table 2 and interpolatmg betwe en 2240 and 2260 psia results in a Tsat of 652.695 F and a subcoohng margm of 65.695 F.
A range of + or .05 P from the answer given on the answer key should be accepted as an appropriate response.
Attachedis a copy of the appropriate page from Table 2 of the ASME Steam Tables.
l
QUESTION 22 (3.00) Part c. only Answer the followmg querions in leference to excess letdown.
- c. After passing through the excess letdown heat exchanger, excess letdown flow can go to three possible destinations. What are :hese destinations?
ANSWER
- c. RCDT (0.50)
VCT (0.50)
Charging pump suction (VCT outlet) (0.50)
RESPONSE
Answer of PRT via relief valve (from seal return header) should also be accepted as an appropriate response for one of the flow paths.
-the relief valve will relieve to the PRT when the 5eal return header isclates on a phase A isolation.
Refer to the copies of the attached TVA drawings:
47W5091 47W813-1 47W611-62-1 l
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QUESTl011 25 (1.50) Part a. only One power range channelis taken out of service with the reactor at 50% power while conducting a plant shutdown. When power drops below 10%, a P-10 permissive interlock solid state bistable switch for one of the remaining power range channels does not reset.
. a. Whatis the effect on the solid state protection system (SSPS) of the failure of the P 10 bistable switch to reset?
A11SWER
- a. The intermediate flux trips will not be automatically reinstated (0.50)
RESPOllSE:
Inability to re energize SRM's when below P-6 because P-10 disable 5 the power supply to the SRMs, should not be counted off for if stated as an answer.
Refer to the attached excerpt from IE Information Notice No. 86-106:
Potential for loss of Reactor Trip capability at intermediate Power Levels.
\ \
G
QUESTION q LOi (2.50) Part a. only Answer the following questions in reference to the essential raw cooling water system (ERCW).
- a. One condition required for automatic start of the essential raw cooling water (ERCW) pump J-A is that it must be 5electedIor automatic start, and pump Q A must not be running. State the other three conditions for pump J-A to autontatic start.
ANSWER
- a. Safety inje ction signal from train 1 A or 2A No blackout signal Transfer switch gearin norrral (0.33 each)
REFERENCE SQN System Desenption, Essentialraw toolmg water system 076000K101 076000K105 076000K201 076000K402 . (KA'S)
RESPONSE
The following responses are also correct and should be accepted.
- 1) Blackout with return of shutdown voltage, af ter a tune delay.
- 2) Control room handswitch not in the pull-to-lock position.
Refer to the attached excerpt from TVA logic pnnt 47W611474 for pump J-A as & reference.
QUESTION 21Q (2.50) Part b. only Answer the following questions in ref erence to the incere thermocouple system.
- b. Positive indication that con 6tions ofinadequate core cooling (ICC) exist when the incere thermocouples indicate greater than degrees F.
ANSWER
- b. 700 degrees F (0.50)
REFERENCE SON Natural circulation and inadequate core coo!:ng , Para X.D.2 SON Incoreinstrumentation, Para X.BA 017020K102 017020K402 017020K403 017020K503 _.(KA'S)
RESPONSE-Either of the following two answers should be accepted as correct
- 1) Core exitthermocouples > 700 F w:th < 40 F subcoolira.
- 2) Core exit thermocouples > 1200 F Per the critical safety function status tress and FR-C.1 either of these two parameters are indications of ICC and will transition you to the ICC instruction.
Refer to the attached copy of FR-C.1 and the status tree on core cooling.
i i
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QUESTION 1 12 L2.0gl Part c. only
,in the rod control system, auctioneered high Tavg is compared with Tref to develop a signal for rod speed and indication. For these systems listed below, briefly describe how the auctioneered high Tavg is used to generate control signals . No setpoints are required in your answer.
- c. Feedwater control system.
ANSWER
- c. Low Tavg signal coincident with reactor tnp causes feedwater regulating valves to close (0.67)
REFERENCE SON condensate and feedwater 001000A101 011000A104 041020K411 059000A306 ._(FAS)
RESPONSE-Part c. should be deleted based on the follow.ng.
The question asked to briefly desenbe how the auenoneered high Tavg is used in the stated control circuits . Feedwater control circuit do es not utilize the auctioneered high Tavg signal.
The suggested response is not totally valid because a failure of the auctieneered value will not inhibit the to Tavg signalinto t* . feedwaterisolation cucuit.
Refer to the attached lesson plan on condensate and f e edwater and excerpt from TVA logic pnnt 47W6113-1.
4 OUESTION lql (2.00) Part e. only For each cf the following situations, indicate whether the OT delta T/0P delta T reactor tnp setpointwillINCREASE, DECREASE, or REMAIN THE SAME. Consider each case separately.
- e. Delta flux (function) input to the OP delta T setpoint !!! CREASES.
ANSWER
- e. Decrease (0.40)
REFERENCE SQN, Operator Cerufication Training, LP : Reactor Coolant Temperature Instrumentation (week 44), Trairung objective F 012000K403 ...(KA'S)
RESPONSE
Answer to part e. is REMAlti THE SAME. Delta 1is an input to the OP delta T setpoint calculator butits gain or multiplieris 0.
Refer to the attached copy of SQN Technical Specification Section 2.0, Safety Limits and Reactor Tnp System Instrumentaton Setpoints. The equation for Or deltaT 5etpoint calcu!stiort
O QUESTION 1 92 (2.00)
One of the functions of the P-4 permissive signalis to actuate a turbine trip. List four other functions of the P-4 permissive.
ANSWER Any four of the folowing at 0.50 points each-
- 2. Provides a signalto the Si block and resetlogic.
- 3. Locks in the circuit to prevent re-opening the MPW valves that were shut by either an SI or high steam generator level actuation signal.
- 4. Provides a signal to the steam dumps 50 that the reactor trip controller controls the steam dumps via the load tejeetion tnp.
- 5. Provides a signalto the process racks to reduce the HI steam flow program setpoint to its zeroload setpoint.
- REFERENCE SON, Operator Certification Tralmng, LP
- Solid State Protection System (week 6-6) 013000K115 013000K401 013000K412 .-.(KA'S)
RESPONSE
The following answers should also be acceptei
- 1) Input to SPDS (initiates Critical Safety 5' unction Status Tree automatic update)
- 2) Input to P-250 computer (Initiates the sequence of events recorder)
Refer to the attached copy of the SQN Operator Certification Lessen Plan on the Tecbr.ical Support Center Computer / Safety Parameter Display System (SPDS).
l t
l
QUESTION LQi (2.50) Part b. and c. only Answer the following questions concerrdng the loss of 120V AC vital instrument power board 1-!!. Assume a reactor inp has not taken place.
- b. Whyhss the automatic control of steam generator water level on the affetted steam generator (s) beenlost?
- c. Why must feedwater pump speed be controlled manually?
At!SWER
- b. Loss of impulse pressure input and hence loss of pro;; ram steam generator level.
- c. Loss of feed flow cr stearn flow signal to the feedwater control system.
RFERENCE SONP A01-25 000057E301 000057G010 ..(KA'S)
RESPONSE-Answer for part b. should also acceptloss of steam flowinput to the feedwater control system Refer to AO!-25.2 section B.3.
Answer for part c. should be changed to : Steam flow signal to fe ed pump speed control circuit.
(Tota!ced steam flow signalis used as a lead reierence for the delta P progrant)
Refer to the attached copy of AO!-25.2 and IVAlogic print 47W611-3 2 for supporting informatiort l
i
OUESTION LCM (1.00)
One of the parameters monitored bythe Post Accident Monitonng Instrumentationis steam generator level. List four of the remairung six parameters that are also monitored by the Post Accident Monitonng Instrumentation.
ANSWER Containment pressure, RCS temperature, RCS pressure , pressunzer level, steam generator pressure, contamment waterlevel.
(0.25 each for any four)
REFERENCE SON Post Acc: dent Momtonng Instrumentation, Para X.A 000009E101 (KA'S)
RESPONSE
The followmg answers should also be accepted as correct responses:
RWSTlevel, Auxiliary Feedwater flow rate, RCS subcooling margin, PZR PORV position, Safety Valve Position, RVLIS.
Refer to the attached copy of Tech Spec LCO 3.3.3.7, Post Accident Monitonng Instrumentation, for support documentation.
OUESTION MQ 12.50) Part d. only With the plant at 100% reactor power, the rod control system m automatic, and no operator action, a loop Tavg instrument fails high.
- d. What are the two most hkely causes of a reactor trip?
ANSWER
- d. Low pressunzer pressure (0.50) or high steamline flow SI (0.50)
REFERENCE SQN Operator Certfication Traming , LP: Review of Instrument Failures, Leaming Objecove V.A 001010A101 _ (KA'S)
RESPONSE
Answer to part d. is Low Pressurizer pressure or OT delta T as the most probable cause of a reactor tnp.
Hi steimline flow SI would not be more probable because Trei would remain relatively constant (due to impulse pressure remaming at a high value) and the Hi flow setpoint would rermin at approximately 110%. Aditionally, as Tref and the flow setpomt decreased so would the steamline flow due to decreasing steam header pressure.
He most hkely cause of a reactor tnp after lo'n pressurizer pressure would be OT delta T due to the decreasing RCS pressure andincreasing delta I, caused by the inserton of concol reds and power being skewed to the bottom cf the core. Both of these factors would cause the setpoint for OT delta T to reduced while the actual delta T would increase or remain the same cue to decreasing Tcold.
QUESTION ]JO (1.25) Part d only For each of the following auxdiaryieed pump automatic start signals, specify whether it applies to the motor dnven auxillary feed pumps (MDAFWP), or the turbine driven auxdiary Ieed pump (TDAFWP), or both.
d Low-low levelin 2/4 5 team generators.
ANSWER i TDAFWP (0.25 each)
REFERENCE SON Operator CerticationTraming, Auxbary Feedwater System, Learmng Objective V.A 061000K402 ..(KA'S)
RESPONSE
BOTH should be accepted as an appropriate response to part i because 10-10 level on 2/4 steam generators implies 10-10 level on any one steam generator which would start the motor dnven auxiliary feedwater pumps and 2/4 starts the turbine driven pump.
Refer to the attached TVAlogic drawing 47W611-3-2 for supporting documentation.
QUESTION JJ.1 (2.00) Part a. only Answer the following questions in reference to the rec:rculation mode of the ECCS.
t ' What causes automaticinitiation of the recirculation mode of the ECCS components? (Include setpoints and/or concidence if applicable.)
ANSWER
& Lowlevelin the RWST coincident with a high containment rump level (0.50)
Setpoints are RWST-less than 29%, and containment sump - greater than 10% (0.50)
REFERENCE SQN Operator Certification Traimng , ECCS components and operation,(week 411),
Learning Objective B.3 005000K408 . (KA'S)
RESPONSE
- a. Containment sump level setpoint for swapover is now 11.25%
Refer to the attached copy of ES-1.2 step 3 revised Apn129,1987
r OUESTION 1 02 (1.50) Part a. oni/
During a reactor startup per G01-2, Plant Startup from Hot Standby to Minimum Load, careful monitoring if the source range nuclear instruments are required.
- a. State the mmimum numberif counts per second required on the highest reading source range instrument.
ANSWER
- a. 2 counts per second (0.50)
REFERENCE SON G01-2, Para V.N.
SON Operator Certfication Traimng , LP: GO!-2 (week 6-8), objecoves C&E 015000K505 . (KA'S)
RESPONSE
- a. Correct response should be 1/2 counts per second.
Referto the attached copy of the recentrevision to G012. RevisionM August 5,1987.
QUESTION Ms (1.00)
Concerning FR-S.1, Response to Nuclear Power Generation /ATWS, explain why the reactor must be tripped before the turbine is tripped.
ANSWER (if the reactor cannot be t'ipped), the turbine will be needed to remove the heat generated by the reactor (1.00)
REFERENCE SON FR-S.1, Response to Nuclear Power Generation /ATWS SON, Operator certification Trammg,LP. Function Restoration Guidelines (week 7-8), objective B 000029K312 -.(KA'S)
RESPONSE
Should also accept an answer pertaining to lirnited capacity of the steam dumps of 40% and turbine must remam on untilthe reactor poweris within the heat removal capacity of the steam dumps.
Reier to the attached copy of an excerpt from the lesson plan for Operator Certification on the steam dump system.
QUESTION 4_12 d.QQ1 From each of the followmg statements , select the one that correct]y desenbes operation of the dieselgenerators per A0135 in the event of a loss of offsite power.
- a. Each dieselgeneratoris designed to carry up to 4400 kW during continuous operation
- b. If one dieseltrips,its companion train dieselmaylack sufficient cooling water.
- c. Equipment started by blackout will stop when offsite power is restored
- d. A diesel that has been loaded less than 40% for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> may be shutdovm without being purged ANSWER b.
REFERENCE SON A01-35, Loss of Offsite Power 064000K102 (KA'S)
RESPONSE-Answer a or b should be accepted due to recent changes to the load ratngs for the 6esel generators. The dieselgenerators have vpgraded to a continuous rating of 4400 kW and a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ratng of 4800 kW.
P.efer to the attached copy of the change to S! 26.la WN h dictates the change and also in6 cates that a Tech Spec revision has been subnutted i
QUESTION 515 0 110.01 LM LM N43 N44 Upper Detector - Actual Current Value 159.7 0* 139.5 147.1 Lower Dete ctor - Actual Current Value 166.0 0* 145.1 150.3 Upper Detector 100% CurrentValue 266.6 252.7 262.9 2543 Lower Detector-100% Current Value 278.6 236.7 270.0 25? 3 Using the above data, select the correct QPTR from the choices lif ted below.
NOTE. Power Range Detector N42 has been properly taken out cf service due to instrument malfunction.
a.1.025 b.1.043 c.1.052 1 1.079 ANSWEB 515 IJpQQ}
d) (1.0)
REFERENCE SON, Prelicense Lesson Plan, Lesson Plarr Reactor Physics Review (Section 1), Traming Objecuve G.
00100K526 (KA'S)
SES Po MGE '.
5.% Cerrect answeris i 1.079 N41 N42 N43 N44 Lower Detector- Actu11 Current 166.0 0 145.1 150.3 Avg = 153.8 166/153.8 1.079.32 l
t
QUESTIOl] ig L221 Part D Only For each of the following, indicate whether the Departure from Nucleate Boilmg Ratio will DJ.C_RJASE. DECREASE or REMAIN THE SAME Consider each case separately.
- a. One reactor coolant pump trips resulting in three loop power operation.
- b. Reactor rower decreases.
- c. One main steam isolation valve inadvertantly shuts. ( Assume rod controlis in manual and no reactor tnp occurs.)
- d. Automatic pressuri:er sprayinitiates.
ANSWER fB (2.00)
- d. DECREASE (0.5)
BEEERENCE SQN, Operator C ertdicaton Training, Lesson Plarc Thermodynamics, Fluid Flow and Heat Tran!!er Review (Week 2), Enabhng Objeetives B.22, B.23 and B. 24.
193009K105 (KA'S)
RESPONSE
- d. Should also accept the answer remain the same based on the assumption that the pressun:er spray valves are modulatmg to maintain a constant RCS pressure. By keeping pressure at a constant vaWthe DNBR would remain the same.
s, o OUSSTION 5.12 (2.00) PART D ONLY For each of the following conditions, indicate whether the available NPSH of the centnfugal charging pumps willINCREASE, DEOREASE, or REM /.IN THE SAME. Consider each use separately.
- d. Dunng emergency baration, the fdter downstream of the bonc acid transfer pump becomes partially clogged from bodc acid precvitation..
ANSWER EJ2 (2.00)
- d. DECREASE (0.5)
REFERENCE SQN, Licensed Operator, Prelicense and Certification Trauung, Technical Staff and Manager's fAdvanced Phase) Training, Lesson Plart thermodynamics Fluid Flow and Heat Transfer Review (Section 2), Enabling Objective B.19.
191004K106 (KA'S)
RESPONSE; REMAIN THE SAME should be the accepted enstner.
Reference General Physics - Thermodynamics, Heat Transfer and Fluid Flow Manual; Section 111, Part A. chapter 2, pag e5 296 - 308.
For flow to occur in the Emerg ency Boration (EEi line into the CCP suction hne, the EB hne pressure must be higher than the suction line. The suction hne acts as a receiving "tank" of water from the EB Ene and contains a static pressure baseo on VCTlevel and pressure in the EB hne, the EB hne would contain the same pressure as the sucticn line up to the first closed line or major restriction.
If the BA f1!ter becomes clogged, the increased flow resistance will be reflected upstream the fi!ter and not downstream the filter. The BA transfer pump riischarge pressure willincrease to map"1p for the.i creased flowresistance of the clogged filter.
The ove a result of the clogged filter is that EB flowwould decrease,8 AT pump discharge pressure willincrease and there willbe no change in NPSH to the CCP'S. Sucti;n line pressure will not decrease. Refer to the attached copy c,f 41W809 2 & 5 for flow path ahgnment.
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.QUESTI N SJJ L2R Part D Only Ir '
For each of the following events / conditions, state whether the value of the estimated citicalboron concentrationwillINCREASE, DECREASE or REMAIN THE SAME. Consider each case
' separately.
1
- d. - Reactor power history prior to the last shutdownis corrected fromA0% to 50% power.- :l
,1 1
fl1
~ ANSWER (2.00)
- d. INCREASE .(0.5).
1
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REFERENCE b v SON, Operator Certification Trauung, Lesson Plarc Reactor Theory, Operator Application (Week 1. l 6), Enabling Objective BA. I 192002K114 192006K107 (K.'.'S) l RESPONSR Decease should be accepted as a correct response. If Reactor is assumed Xenon Free with ir. ceased power detect the critical baron concentration would decease refer to attached except from TI 21.
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- QUESTION . :fjl (2.001 - Part D Only
= Answer each of ti.e following statements TRUE or FALSE.
. d. Doppler coefficient becomes less negative from BOL to MOL due to fuel densification of
. new fuel, and becomes more negative from MOL to EOL due to cooler fuel and resonance -
capture in Pu-239 and PU 240.
ANSWER 5 17 L2&Q1
- d. TRUE (0.5)
REFERENCE; SON, Operator Certification Tranung, Lesson Plarc Reactor Theory, Operator Application VI
. (Week 14), Enabhng Objectives B.1,-B.2 and B.3.
192004K103 192004K106 192004K107 192007K104 '(KA'S) e
' RESPONSE:
Correct answer to part d. Shauld be FALSE.
4 Reascnis Doppler coefficient becomes more negative from BOL to MOL due to fuel pellet o expansion. Per recent information from Westinghouse fuel densification actually occurs over approximately the first 1000 MWD /MTU exposure on new fuel
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QUESTION 1.QE G001 PART E ONLY
- e. Delta flux (function) input to the OP delta T 5 etpoint INCREASES.
ANSWER &&5 (2.00)
- e. DECREASE (0.4)
REFERENCE SQN, Operator Certificatien Training, Lesson P. art Reactor Coolant Temperature instrumentation (week 44), Training Objective F.
012000K403 012000K611 (KA'S)
RESPONSE;
- e. ANSWER, REMAIN THE SAME,is the correct response. Delta Iis aninput to the OP delta T setpoint calculator but its gain or multiplie1 is 0.
Refer to the attached Table 2.2.1 from the SON Tech Specs for validation.
t r
l 00ESTION 633 110010 PART A ONLY -
Answer each of the following statements TRUE or FALSE.
- j. a. Containment sump isolation valve 1-FCV-63 72,must be fully closed (as' indicated by both I - the stem switch and the gear switch) to meet one of the conditions for opening residual heat removal (RH R) inlet isolation valve s 74-1 and 74-2 1).
l-ANSWER 5Og 1221 I~
1 A TRUE (0.5)
REFERENCE; i SQN, Operator Certification Traming, Lesson Plan Residual Heat Removal (We ek 4-9), Traimng Objectives B, D and E.
l 005000K401.005000K402 -(KA'S)-
RESPONSE;
- a. Correct answeris FALSE.
1 PCV43-72 is an interlock to FCV-74-1 only, FCV43-73 is an interlock to FCV-74-2 only Refer to the attached excerpt from TVA logic print 47W61174-1 l
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q - QUESTION : :62 ILQQ1
? LIST the iour (4) purposes of the Ice Concenser System. -
T
[ ANSWER ' j.B ILOQ1 AllIour(4) at 0.5 points each-
', - 1.; Absorbs (thermal) energy released during LOCA to control the peak pressure.
- 2. Hold pressure at a low value for an extended period of time
. 3. ' Uses sodium tetraborate to remove elementallodine from the containment atmosphere -
- 4. Maintains proper pH (9.0-9.5) to convert iodine to a nonvolatile f orm
' REFERENCE SQN, Operator Certification Trammg, Lesson Plan' Containment Systems (Week 4-10), Traming Objective D.
,. 025000 0004 (KA'S)
= RESPONSE; The following responses should also be acceptei.
' 1) Control PH in the containment sump to reduce corrof.on. By keepmg pH between 9-10 corrosionis limited and thus H2 productionis reduced.
'- Refer to attached Lesson Plan on Mitigating Core Damage for ref erence information.
- - Refer to attached excerpt from SON Tech Specs on Ice Condenser basee .
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QUESTION 6jk (2.001 One of the functions of the reactor trip interlock, P-4,Is f., actuate a turbine trip. List four (4) other functions of the P 4 permissive.
ANSWER 6.16 (2.00)
Any four (4) of the following at 0.5 points each-
- 2) Prcvides a signal to the SI block and reset logic
- 3) Locks in the circuit to prevent re-opening the MFW valves that were shut by either an Sl or high 5 team generator level actuation signal
- 4) "rtvides a signalto the steam dumps so that the reactor trip controller controls the steam du:rps vice the ic'd rejection controller.
- 5) Provices a signal to the process tacks to reduce the HI steam flow program setpoint to its zero load setpcint.
REFERENCE; SQN, Operator Certification Training, Lessen Plan Sohd State Protection System (Week 6-6).
013000K115 013000K401013000K412 (KA'S)
RESPONSE; The following responses should also be acc pted as correct answers:
- 1) Input to SPDS; initiates Critical Safety function Status Tree automatic updat c
- 2) Input to P-250 computer;intiates the sequence of events printout.
Refer to the attached Operator Certification Lesson Plan on Techdcal Support Center Computer / Safety Parameter Display system (SPDb) l l
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I
'00ESTION M 1501
' ."- J List the Sequoyah Nuclear Plant emergency exposur' ilmitiand explain the conditions under which '
eachapplies.
. -ANSWER M (1.50)
' 10 Rem (0.25): for plar.ned exposure during an emergency situation (0.25).
25 Rem (0.25): to prevent serious damage to the p' ant or hazard to personnel (0.25).
75 Rem (0.25): to save alife (0.25).
. REFERENCE; SON, Operator Certification Training, Health Physles, Lesson P! art Radiation Standards ano
- Guidelines (5eetion 5), Trauung Objactive 3.
- 194001A116 194001K103 (KA'S)
RESPONSE;
- Correct answeris:
25 RE!4 - To protect facilities, eliminate furthur escape of effluents, or to control fires, 75 RE14 - For lifesaving for individuals or to prevent senuus injuries to alarge ntunber of persons.
Refer to attached excerpt from IP-15, Emergency Exporere Guidelines Rev. 6, July 10,- 1987.
-_-n - - . _ _ __ _ _ - - - - _ . _ . . _ _ - _ _- _ _ - ___________ _ _ ____
r . . .
QUESTION L}2 (2.00) FART r gjg Emergency Instruction E 0, "Reactor Tnp or Safety Injection", contains anmediate actions to be performedin event of a reactor trip or safetyinjection. for each of the following steps, state the basis for the action injection. For each of the following steps, state the basis for the action a) Step 2. Verify turbine trip b) Step 7: Verify main feedwater isolation c) Step 8: Verify auxiliary feedwater status d) Step 13: Check Tave ANSWER L){ (2.00)
- d. (RCS temperature stable at er 4ending to the no load value) indicates that the secondary steam dump systemis operational as designed a: a secondary heat sink. (0.5)
REFERENCE; SQll, Operator Certification Training, Lesson Plan Emergency Instruction E-0 "Reactor Trip or Safety Injectiorc" and Subsections ES4.1, ES4.2 and ES-0.3 (Week 7-2), Trauung Objective F.
000007K301 (KA'S)
RESPONSE;
- d. Following answer should be accepted-The basis for the check T-avg step is to venfy Tavg is approximately 547 F. and is stable *.nd controlled. If Tavg is decreasing in an uncontrolled manner it is indicative of a secondary side break.
Refer to copy of step 13 of E-0 the RNO column for - vn.
gm
-_h) f OUESTION ' M 11 00_1 y, *'
At Sequoyah Nuclear Plant, Unit 1, a steam generator tube inspectionis to be performed on one steam generator with 3375 U-tubes. . In reference to this inspection, answer the iollowing questions.
NOTE: APPLICABLETECHNICAL SPECIFICATIONS ARE ENCLOSED.
- a. State die number of tutas that are required to be inspected
- b. - After completion of this inspection, itis determined that 3 percent of allthe U-tubes .
inspected are degradeiand one U tube is dcfective. State the CATEGORY that the result of this
. inspection fallsinto and state ANY REQUIRED ACTIONS.
ANSWER M 0 IL001
- a. 405 U tubes rrustbeinspeeted. (0.5)
- b. C .4 gary C-2. Plug the de'ective tube and inspect an additional 810 U-tubes in this steam L ,
generator. (0.5) ,
REFERENCE; SON, Operator Certification Training, Lesson Plan; Technical Specificatten 3/4A "Reacter Coolant System"(Week 10-5), Enabling Objectiva B.1.
002000G011 (KA'S)
RESPONSE; Question should be deletei Surveillar.ce responsibility for this L.C.O. belo'igs to the ISI (Inservice
. Inspection) group. Determination of surveillance results is not directly perfomled by the SRO as part of his job responsibility.
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= _ < . -m-- , ,. ,- - , - . , , ,. ~ . m. - ,
/
,,5 QUESTION M Igl At Sequoyah Nuclear Plant, Unit 1, the following events have occurred in the past seven hours:
- 1. At 5:00 a.m, one PORV beganleaking by while a reactor startup was in progress (reactor power at 10 exp'-8 amps).
~ 2. At 5:45 a.m the leaking PORV's block valve was shut and power removed.
- 3. . At 9:30 a.m., Unit I was operating at 75% rated thermal power.
- 4. At 9:45 a.m, the 5econd PORV valve beganleaking by.
- 5. At 10:15 a. m, an attempt to close the block valve for the second PORV failed and the block valve was declaredinoperable.
- 6. At 10:30 a.m, a plant shutdown commenced.
- 7. At 12:00 p.m., (pre s ent time), the plant was placed in hot standby.
In reference to the above events, answer each of the followAg questions.
NOTE; APPLICABLETECHNICAL SPECIFICATIONS ARE ENCLOSED.
a.- Were any technical specifications violated? If the answeris YES, explain.
- b. By what time tomorrow must t!.e reactor be COLD SHUTDOWNR ANSWER M (1.00)
- a. No technical specifications were violated. (0.5) l
- b. By 11:15 p.m tommorrow, the plant must be in cold shutdown. (0.5)
REFER 9NCE-
' SQN, Operator Certification Training, Lesson Plan- Technical Specification 3/4.4 "Reactor Coolant Systent (Week 10-5), Enabling Objective B.3.
002000G005 010000G005 (KA'S) 4 RFC ONSE
- b. Question should be deletei Answer given is inconect. Correct answer is 6:00 p.m. the ,
l following day. SQN policyis that Tech Spec Action Times are n:t additive. From the time you enter Hot Standby you must be in Cold Shutdown in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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lOUESTION ' QR 1.00)
While Sequoyah Nuclear Plant, Unit 1,is operating at 100 percent rated thermal power,' the isolatifon ;
valves for accumulator 1 and accumulator 2 are discovered shut. The reactor operator immediately -
attempts to open both isolation valves, but bothisolation valves remain shut. State t6he required actions as di!ected by Technical Specifications.
NOTE; APPLICABLE TECHNICAL SPECIFICATIONS ARE ENCLOSED.
ANSWER 8J figl Be in hot standby within one hour and be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. (1.0)
REFERENCE; SQN, Operator Certification Training, Lesson Plarr Technical specification 3/4.5 "Emergency Core Cooling systent (Week 11-4), Enabling Objective B.3.
006050G005 (KA'S)
' RESPONSE; Action requirements for L. C. O. 3.0.3 should be accepted as a correct respons e. The action requirement for 3.5.1.1 stipulates with one accumulator inoperable while the condition you are in .
involves two accumulators. With more than one accumulator inoperable you don't comply with the e.ction of L. C. O. 3.5.1.1 and you are in 3.0.3.
3.0.3 requirements
I hour to correet 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to Hot Standby 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to Hot Shutdown Refer to attached Tech Spec LCO's 3.0.3 and 3.5.1.1
Ye us w Gwe.sko, ca . a o .
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3/4.5 EMERGENCY CORE COOLING
- SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS COLO LEG INJECTION ACCUMULATORS LIMITING CONDITION FOR OPERATION .
3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with:
- a. The isolation valve open,
- b. A contained borated' water volume of between 7857 and 8071 gallons of borated water,
- c. Between 1900 and 2100 ppe of coron, and
- d. A nitrogen cover pressure of between 385 and 447 psig.
APPLICABILITY: MODES 1, 2 and 3.*
ACTION:
- a. With one cold leg injection accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to
- OPERABLE status within one hour or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- b. With one cold leg injection accumulator inoperable due to the isola-tion valve being closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUT 00WN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each cold leg injection accumulator shall be demonstrated OPERABLE:
- a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:
- 1. Verifying, by the absence of alarms or by measurement of levels R16 and pressures, the contained borated water volume and nitrogen cover-pressure in the tanks, and
- 2. Verifying that each cold leg injection accumulator isolation valve is open.
^ Pressurizer pressure above 1000 psig. .
MAR 20 w82 SEQUOYAH - UNIT 1 3/4 5-1 Amendment No. 12 s
De
. l g 9. c 6 l
\ 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS l
3/4.0 APPLICABILITY i LIMITING CONDITI0f t FOR OPERATION l
3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting conditions for Operation, the associated ACTION requirements shall be met.
3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.
3.0.3 When a Limiting Condition for Operation is not met,.except as provided in the. associated ACTION requirements, within one hour action shall be initiated
' to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in:
- 1. At least HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- 2. At least HOT SHUTDOWN within 'tha following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- 3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Where corrective measures are completed that permit operation under the ACTION I requirements, the ACTION may be taken in accordance with the specified time limits p*
as measured from the time of failure to meet the Limiting Condition for Operation.
Exceptions to these requirements are stated in the individual Specifications.
3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operatien are met without reliance on provisions contained in the ACTION requirements. This
-provision shall not prevent passage through OPERATIONAL MODES as required to \
comply with ACTION requirements. Exqeptions to these requirements are stated in the individual Specifications.
3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency i
' power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s),
train (s), ccmponent(s) and device (s) are OPERARLE, or likewise satisfy the requirements of this Specification. Unless both conditions (1) and (2) are satisified, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> action shall be init.iated to place the unit in a
[ MODE in which the applicable Limiting Condition for Operation does not apply by placing it as applicable in:
- 1. At least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- 2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- 3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This Specification is not applicable in MODES 5 or 6.
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ENCLOSURE 4 o
SIMULATION FACILITY FIDELITY REPORT Facility Licensee Docket No.: 50-327, 50-328 Facility License _No.: DPR-77, DPR-79 Operating Test Administered at: Sequoyah Nuclear Plant Operating Test Given On: November 18-19, 1987 During the conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies were observed.
Simulator Modeling Weaknesses During one exam-scenario, candidates were borating for a normal plant shutdown-when the simulator began to over-borate the RCS. None of the simulator instructors involved with the exam could explain why the simulator responded in this way.
Simulator' Malfunction Weaknesses Malfunction 74 Volume Control Tank Level Control fails low. This malfunction did not perform as stated in the malfunction description. With level transmitter LT-62-130A failed low, make-up should have started and continued until the operator took corrective action. Instead, the make-up stopped at 40% level as if LT-62-129 were controlling make-up.
During the development of the simulator examinations an attempc was made to duplicate LER events which have occurred on Unit 1 or Unit 2. The following is a list of those events which could not be duplicated by the simulation facility due to a lack of modeling capability:
LER 84-014 Valve 2-FCV-3-47 (feedwater) failed to close when signaled.
The simulator cannot model this mode of valve failure.
LER 84-015 Double lit valve (2-FCV-3-47). The capability to double light valves does not exist.
LER 85-002 LCV-6-106A Heater Drain Pump Discharge valve failure. Three similar events have occurred related to this type of valve failure.
Train "A"Reactor Trip Breaker. Although the simulator can prevent both breakers from opening, it cannot prevent a signie breaker from opening.
Enclosure.4 2 LER 85-009 All RPIs lower approx. 20 steps due to degraded power supply.
The simulator cannot model this event.
Stator Coolant Pump Trip.
LER 85-035 Failure of control power for an EDG, LER 84-006 One bank of backup heaters fail to operate.
LER 84-054 Failure of relay rack 1-R-15 power supply; VCT divert valve failure; failure of PI-68-66.
LER 85-007 Failure of EGTS Room Cooler to start LER 85-016 Degraded controller PC-1-72 Impulse Pressure. (The simulator is capable of failing the controller catastrophically); leaking feedwater isolation valves.
LER 85-029 Tripping of IB 480v MOV unit bd.; failure of valves on MFPT condensor to close.
Additionally, several other failures could not be simulated such as:
RWST leak and Feedline breaks
-The inability to model these events prevented the NRC examiners from evaluating the facility operators on actual reportable events which have occurred at Sequoyah.
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