ML20207T444
| ML20207T444 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 02/20/1987 |
| From: | Casto C, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207T433 | List: |
| References | |
| 50-250-OL-86-05, 50-250-OL-86-5, NUDOCS 8703240021 | |
| Download: ML20207T444 (77) | |
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ENCLOSURE 1 l EXAMINATION REPORT 250/0L-86-05 Facility Licensee: Florida Power and Light Company Facility Name: Turkey Point Units 3 & 4
' Facility Docket No.: 50-250/50-251 Written and oral accele ated requalification examinations were administered at Turkey Point Nucle P1 t Florida Ci y, F o ida, i i / J# -[ 28 '
2 7 O ' Chief Examher: ChTrles A.'tasto '-
~Date Signed Approved by:
John F. Munro, Section Chief Ms rho f7 Date Signed i Summary: Accelerated requalification Examinations on December 15-17, 1986. Written requalification re-examinations were administered to two senior reactor operators (SR0s), one of whom passed. Written and oral requalification examina-tions were administered to six SR0s, two of whom passed the written exam and five of whom passed the oral exam. See Enclosure 4 for a discussion of these results on the NRC's evaluation of
- Turkey Point's accelerated requalification program.
Based on the results described above, three of eight SR0s passed, i [ N E E8883so PDR .
REPORT DETAILS
- 1. Facility Employees Contacted:
*C, R. Wethy, Onsite Vice President
*C. J. Baker, Plant Manager
*D Grandage, Operations Superintendent
*T. A. Finn, Assistant Operations Superintendent
*J. E. Crockford, Operations Training Superintendent
*G. A. Hollinger, Training Staff
*R. D. Hart, Licensing Engineer
*G. Salamon, R&C Engineering
*R. Brodnax, Training Staff
- Attended Exit Meeting
- 2. Examiners:
- Charles A. Casto William Hemming (EG&G)
J. Mcdonald, NRC, RI (Attended Exit)
- Chief Examiner
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided your training staff with a copy of the written examination and answer key for review. The comments made by the facility reviewers are included as Enclosure 3 to this report, and the NRC Resolutions to these comments are listed below.
SR0 Exam (1) Question 5.03 . Agree with facility comment. Question was modified from an existing question, but the answer key was not changed accordingly. (2) Question 5.05(c) Do not agree with facility comment. The question clearly states that the EOL Baron concentration was used vice the required BOL Boron concentration (which typically differ by several hundred ppm). The effect of this error in the ECC calculation is the focal point of the question. No change to answer key. ~
2 4 (3) Question 5.07(a) Agree with facility comment. . Even though theoretically, the count rate should be the same for identical reactivity additions, the given scenario may not allow sufficient time for count rates to stabilize as the reactor nears criticality. Unit B will also be accepted as a correct answer. , (4) Question 5.11(b) , Recomended response is a ecuivalent response to existing answer key. No change to answer key requ' red. (5) Question 6.06(a) Do not agree with facility comment. Referenced drawing shows reactor trip vice turbine trip sicnals. In accordance with Fig. 14 of System Description 127 " Main Turb ne Control" and Logic Drawing 5610-T-L1, the , original answer key is correct. No change to answer key. (6) Question 6.08(a) Agree with facility coment. Original answer key reflects the condi-tions at a different facility. Answer key will be modified as recom-mended with no penalty for an incorrect fourth answer. (7) Question 6.08(b) Agree with facility comment. The system description should be revised to correctly reflect information contained within Technical Specifica-tions. Answer key will be modified as recommended. (8) Question 7.05 Do not agree with facility comment. Plant supervisors should know all personnel performing actions under his direct supervision during a control room evacuation. Partial credit will be given for the recommended answer. No change to answer key required. (9) Question 7.11(c) Facility comment equivalent to existing answer key. No change required. (10) Question 8.06 Do not agree with facility comment. The question clearly states that inability to meet TS 3.4.4.b is causing the shutdown. TS 3.0.1 must be applied in choice "a". No change to answer key.
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sh 3 [ % / (11) Question 8,09
+
Agree vf th' facility comment. Based on possible misinterpretation of the type of in service testing that is being addressed in the question,
, i the additic al recommended answers will be accepted.
(32)i Question 8.10. Facility comment addresses information not asked for in the question.
. Exa.ninees will not be penalized for providing additional correct s information, but no change to the answer key is required.
~'
j , (13)' Question 8.16 r ~/ - Facility comment equivalent to existing answer key. No change required.
- 4. Exit Meeting At the conclusion of the sitt visit the examiners met with representatives of the plant staff to dicc m s the results of the examination.
There were no.ljeneric weaknesses noted during the oral examination.
, Twenty perceiit of the changes (orie,out of five) made to the answer key were the result of inadequote or incomplate reference material. This low number, as well as the minimc1 number of exam comments is a direct reflection of the ini,)rovement made in up acet racy and completeness of the facility's training naterial.
/ /
ihe cooperation ~given to :.he examiners and the effort to ensure an atmosphere in the control room conducive :to oral examinations was also noted and appreciated.
'helicenseedidnotidentNyasproprietaryanyofthematerialprovidedto e reviewed by .the exandners.
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p.-- A R C M A r/E A U. S. NUCLEAR REGULATORY-COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _ TURKEY _PQ1NI_gh4________ REACTOR TYPE: _PWB-WEg}________________ DATE ADMINISTERED _g6f12/1g____ _________ EXAMINER: _Qg8Nz_W M_____________ CANDIDATE: _____________________ _ INSIBUGI19NS_I9_GON9190IE1 Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination. Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff. 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 four (4) hours'after the examination starts. l l 7 % OF CATEGORY % OF CANDIDATE'S CATEGORY VebuE_ _IDIB6 ___SG9BE___ _Y0LUE__ ______________g61EgQ8Y ______ _==__ _19199__ _2Dz99. ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT
- OPERATION, FLUIDS, AND THERMODYNAMICS
_19199__ _2g199 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION . _10z99__ _20199 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL _19z99__ _25z99 ___________ ________ 8. ADMINISTR'A"TIVE PROCEDURES, CONDITIONS, AND LIMITATIONS _Z2199__ _ _ _ _ _ __ _ _ _ _ Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
~ ~ ~~~~~
Candidate s Signa [ure
~I~~
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 gely 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 gash section of the answer sheet.
IB . Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a Dgw page, write 90ly 90 90R Eidt 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 tht ge lines between each answer, 1 1.- Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Une abbreviations only if they are commonly used in facility litgtatutg.
- 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 examinet 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.
"c : 4 , .
- 10. 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 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.
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m Et__IME98Y_9E_NWGLEBB_E9 WEB _EL8NI_9EEBBIl9Nx_ELWID@s_8NQ PAGE 2 ISEBdgDYU6dlG@ QUESTION 5.01 (1.00) The reactor is critical at 10,000 cps when a S/G PORV fails open. Assuming.BOL conditions, no rod motion, and no reactor trip, choose the answer below that best describes the values of Tavg and nuclear power for the resulting new steady state. (POAH = point of adding heat). a.- Final Tavg greater than initial Tavg, Final power above POAH. i
- b. Final Tavg greater than initial Tavg, Final power at POAH.
- c. Final Tavg less than initial Tavg, Final power at POAH.
- d. Final Tavg less than initial Tavg, Final power above POAH.
QUESTION 5.02 (1.00) Attached Figure # 219 shows a power history and four possible Menon traces (reactivity vs time). Select (a, b,'c, or d) the curve that
-correctly displays the expected xenon transient for the given power history.
QUESTION 5.03 (1.50) Indicate whether the following will cause the power range instrument t'o be indicating HIGHER, LOWER or the SAME as actual power, if the instrument has been adjusted to 100% based on a calculated calorimetric.
- a. If the feedwater temperature used in the calorimetric was higher than actual feedwater temperature.
- b. If the reactor coolant pump heat input used in the calorimetric is omitted.
- c. If the steam flow used in the calorimetric was lower than actual.
QUESTION 5.04 (1.00) Which of the attached graphs correctly shows the relationship of the Isothermal Moderator Temperature Coefficient (MTC) to the Moderator Only MTC7 (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) I
Di__INE98Y_9E_NWGLEGB_E9 WEB _EL8HI_9EEB9II9Ni_ELUIDEi_8ND PAGE 3 THERMOQYG8 mig @ 4 . QUESTION 5.05- (1.50) An ECC is calculated for a startup following a reactor trip from 50% power equilibrium xenoni OL). Indicate if the actual critical rod position will be HIGHER, LOWER or the SAME from the calculated position for each of ' the f ollowing situations. Use attached curves as appropriate and treat each case individually. a) Xenon reactivity curve for a trip from 100% is used to calculate conditions to startup 20 hours-after the trip. b) Baron worth at EOL is used vice BOL worth. (Dilution is required .to reach desired Boron concentration in both cases) c) The EOL critical Boron Concentration is used instead of the BOL critical Boron concentration. QUESTION 5.06 (1.50) a) What is the primary factor at BOL that causes redistribution of the axial flux as power is increased? (0.5) b) Describe how the axial flux will shift as power is reduced from full to zero power at EOL, and the main cause of this behavior. (1.0) DUESTION 5.07 (1.00) - Unit A is at EOL while Unit B has just been started up af ter a ref ueling. Assuming a rod speed of 48 spm, both reactors are taken critical by pulling in 50 step increments and waiting 60 seconds before pulling again. Assuming all systems and parameters are identical at the commencement of the startup, and both units are initially shutdown by 2% (delta k/k): a) Which Unit will have the highest source range counts when criticality is reached? b) How will critical rod heights compare in the two Units? (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
Ut__IBEQ8Y_QE_NUGLE68_EQWEB_E68NI_QEgBOIlQNz _ELylQEz_@NQ PAGE 4 IEEBM90YN@ Migs QUESTION 5.08 (1.00) The reactor is operating at 100%' power with all rods out, near EOL with equilibrium Xenon conditions when power is to be reduced to 50%. The operator observes that AFD is within its band and decides to lower power by borating, leaving rods in the ARO position. Actual Tavg follows programmed Tavg. Describe the change that will occur in AFD and why it occurs, prior to changes in Xenon having a noticeable effect. QUESTION 5.09 (2.00)
- a. During natural circulation, EXPLAIN how it is possible to form a bubble in the reactor vessel head when indications show thet the RCS is subcooled?
- b. How will pressurizer level respond, (INCREASE, DECREASE, or REMAIN THE SAME) if the backup heaters are energized with a bubble iri the reactor vessel head? Assume normal pressurizer level and briefly EXPLAIN your answer.
QUESTION 5.10 (1.00) The attached figurdbshows the change in pressure across a pressurizer PORV , and.its associated upstream and downstream piping for various valve positions. EXPLAIN why the major pressure drop occurs in piping segment P1-P2 when the valve is slightly open, whereas the major pressure drop occurs in piping segment P2-P4 by the time the valve is fully open. QUESTION 5.11 (1.50) Assume the plant is at a steady state power level of 75%, with rod control in MANUAL and the plant in BOL, when a 15% step increase in turbine load
-occurs. DESCRIBE and EXPLAIN the behavior of the following parameters during the first several minutes of the transient. Assume MTC is -3 pcm/F and NO RX TRIP occurs.
a) Tstm b) margin to DNB (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
-Di__IBEQBY_9E_NQQLg88_EQWg8_EL@NI_QEgB9IlgNz_ELylDEs_@ND PAGE 5 THgBMODYNAMICS
- 4 QUESTION 5.12 (1.50) a) Why is 2200 degrees F given as the 10CFR50 peak cladding limitation during an accident condition? (0.5) b) The Tech Spec limit on F(n) Delta h is 1.6201.0 + .3(1-P)] where P represents fraction of rated power. Why does this limit increase as power decreases for a given RCS flow rate? (1.0)
OUESTION 5.13 (1.50) Attached is a typical boiling, curve for water as it approaches, then exceeds, the DNB point.- What are the thermodynamic conditions that cause: a) The decrease in heat transfer rate in Region III7 l b) The increase in heat transfer rate in Region IV7 OUESTION 5.14 (1.00) Attached are the curves showing Baron Limits for Moderator Temperature Coef-licient Control (Unit 3, Cycle X). EXPLAIN why the curve can shift upward as Bank D rods are inserted. (***** END OF CATEGORY 05 *****)
c 6t__ELANI_EYEIEMS_DEEIENt_CQNIBQLg_ANQ_INEIBUNENI@IlgN PAGE 6 QUESTION 6.01 (1.00) MCC D automatically transfers to EDG.A (LC 4C) on a SI and loss of offsite power with EDG B in LOCAL CONTROL. Subsequently, EDG B is restored to normal control and supplying the 3B and 4B buses when EDG A fails. Which statement below correctly describes how power would be restored to MCC D7 a) It will automatically transfer back to LC 3D as soon as power is lost to LC 4C b) It will automatically transfer back to LC 3D even if power has NOT been restored to LC 3D, THEN both breakers will open due to no power source. c) It must be manually. transferred by physically operating the supply breakers on LC 3D and LC 4C. d) It will automatically transfer back to LC 3D after the lockout relay on LC 4C is reset. e) It must be manually transferred by physically operating the breaker on LC 3D after resetting the lockout relay on LC 4C. QUESTION 6.02 (1.00) Which of the following correctly describes the NORMAL status of the CCW valves associated with the Emergency Containment Cooling Units? INLET OUTLET BYPASS a) open open open - b) open open closed c) open closed open d) closed open closed e) closed closed open (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
Ai__P6&NI_gYSIEd@_QE519Nz_GQNI8962_8NQ_INSIBydgNI8IlgN PAGE 7 QUESTION 6.03 (1.00)
. Which statement below correctly describes what will happen if an IN SERVICE inverter for the 120 Vital AC system has a momentary failure, assuming the Inverter Auto Retransfer Switch is in the "OFF" position?
a) The system will transfer to its associated CVT and remain there. b) The system will transfer to the backup standby inverter and remain there. c) The system will transfer to its associated CVT and will transfer back to the inverter when it's restored to normal. d) The system will transfer to the backup standby inverter and will transfer back to the normal inverter when it's restored to normal. e) The system will transfer to either the associated CVT or standby inverter, depending on the position of the CVT ALTERNATE SOURCE TRANSFER SWITCH and remain there. QUESTION 6.04 (1.00) Use the attached. drawing of the AFW flow controllers to answer the following: a) Assuming an AFW initiating signal has occurred, what device DIRECTLY generates the OPEN signal for the solenoid valves which allow the HICS flow controller to position the flow control valves? . b) What is the backup to instrument air for Unit 4's Train 2 AFW flow control valves? OUESTION 6.05 (1.50) Answer the f ollowing regarding the Pressurizer Pressure Relief Systems a) What is the purpose of the loop seal associated with the Safety Valves? b) TRUE or FALSE: The safety valve inlets ALL tap off a common line which penetrates the upper head of the pressurizer. c) What two different signals will cause the "PORV/ RELIEF VALVE OPEN" annunciator to actuate? (Include any applicable setpoints) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
(- - 61__PL8NI_gySIgdg_QgglgN _CQNIBQLi_@NQ_lNEIBydgNI@IlQN z PAGE R QUESTION 6.06 (2.00)
.a) List the-6 devices / signals which will actuate the 20AST and 20ASB Auto Stop Trip Solenoids for the Main Turbine. (1.5) b) Describe how the device is used to prevent a turbine trip when testing the overspeed trip mechanism. (0.5)
QUESTION 6.07' (1.50) a)- Aside from operating the control switch or breaker, list the 4 direct causes of a RCP trip. Be specific in identifying any relays which cause the RCP to trip. (1.0) b) Why is there a 120 second time delay in starting a RCP once oil lift pressure has reached a sufficient pressure? (0.5) QUESTION 6.08 (1.50) a) What are the 4 sources of water available for the Spent Fuel Poo17(1.0) b) What design feature of the spent fuel racks ensures criticality does not occur in the Spent Fuel Pool? (0.5) QUESTION 6.09 (1.00) While transferring a new fuel assembly from the upender to the core, the Dillon Cell fails high. What actions must the operator take to disengage the gripper from the fuel when the assembly is properly placed in the core? (Include location of any controls / components operated) QUESTION 6.10 (1.00)
-Explain the basis for putting blocking devices on HHSI pump miniflow recirc valves 856 A/B.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
61__PL8HI_EYEIgd@_QEElGNi_CQNIBQLx_9NQ_INSIBQdgNI@IlgN PAGE 9 QUESTION 6.11 (1.00)
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Unit 3 is at 60% power with the turbine runback switch in the RPI/NIS position when Pzr Pressure protection Channel III is taken out of service for testing (all associated bistables are tripped). Subsequently, the upper detector for NI42 fails HIGH. How does this affect the operation of the. turbine? Explain your answer. QUESTION 6.12 (1.50) Provide the bases for the following precautions associated with the Emergency Diesel Generators: a) The soak back pump shall be in operation at all times when the engine is not' running. b) Avoid operating the engine unloaded for greater than 15 minutes at rated speed. , QUESTION 6.13 (1.50) A loss of voltage and bus stripping occurred 5 minutes ago. Both Units tripped and are being supplied with electrical power from the EDGs with ALL equipment operable. Unit 3 then undergoes a Safety Injection. What will happen to the Unit 4 equipment that is on line over the next minute? QUESTION 6.14 (1.50) a) Define a " Passive" failure as it applies to the Emergency Core Cooling Systems. (0.5) b) What combination of " Active" and " Passive" failures are allowed during BOTH the ' Injection
- phase and the ' Recirculation' phase of ECCS operation? (1.0)
(***** END OF CATEGORY 06 *****)
.s Zs__E80GEQWBEE_:_NQBdeLs_6ENQBdeks_EMEBEENGY_eNQ PAGE 10 809196gGIC66_CQNIBQL:
QUESTION 7.01 (1.00) Which one of the following situations requires emergency boration'per ONOP-046.1, " Emergency Boration"? a) One rod stuck in the-fully withdrawn position following a rx trip. b) ROD BANK D LOW LIMIT' alarm actuates as rods are driving in automatically. c) A steady, sustained increase in source range counts'with'keff=.95 with no rod motion or planned dilution in progress. d) RCS temperature f alls below 525 degrees F at a rate of 10 deg/hr. following a reactor trip. QUESTION '7.02 (1.00)
'Which one of the f ollowing is addressed in FR-Z.2, " Containment Flooding",
as a potential source of excessively high containment sump levels?. a) Condensed steam from a steam break b) RCS water from a LOCA c) RWST d) Accumulators e) CCW QUESTION 7.03 (1.50) Indicate whether each of the following statements regarding EOP usage is TRUE or FALSE: a) The REDIAGNOSIS procedure, ES-0.0, is ONLY used after departure from E-O and ONLY if an SI had been actuated. b) If an expected response of an EOP can NOT be verified and the Response Not Obtained action can NOT be performed, the operators may continue with the procedure. c) If a task is in progress when a transition to another procedure takes place, that task need NOT be completed. (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
Zi__EBQGEDWBEH_ NORMAL 5_A@NQBM862_EMg8GENgy_AND PAGE 11
.60DIOLgGIC86_GQNIBg6 ' QUESTION. 7.04 (2.00)
List, in their order of preference, the f our recovery techniques stated in FR-C.1, " Inadequate Core Cooling". QUESTION- 7.05 (2.25) What are all the Plant Supervisor immediate actions if the announcement is made " Fire in the control room, shift personnel report to assigned control room evacuation stations" ? QUESTION 7.06 (1.00) List the Four UNIT 3 Area Radiation Monitors critical to refueling that, if INOPERABLE, would halt refueling operations. QUESTION 7.07 (1.00) If a high temperature alarm on the CCW heat exchanger outlet actuates, what two verifications and associated actions on the Intake Cooling Water system must be made in accordance with ONOP 3408.1, "ICW Malfunction"? OUESTION 7.08 (1.00) . What actions are required per GNOP O208.14, " Deviation / Failure of Reactor Protection and Safety Related Hagan Instrumentation Channels", if a That protection channel spiked high MOMENTARILY and then returned to its original indicating level and behaved normally thereafter? QUESTION 7.09 (2.00) What are ALL the actions contained in the RESPONSE NOT OBTAINED column in FR-S.1, "ATWS", for the steps listed below? Ensure you discuss any contingency actions stated within the RNO step itself. a) Turbine is NOT verified as tripped. b) Pressurizer pressure exceeds 2335 psig. l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
wn--- PAGE 12 Zz__BBQGEQWBEE_:_NQBdekt_0ENQBdekt_EMEBEENGY_0NQ BGQ1969G1986_GQNIBQL QUESTION 7.10 (1.00) Aside.from termination of upper core boiling and boron precipitation, why is establishment of hot leg recirculation delayed 18 hours after a LOCA? QUESTION 7.11 (1.50) Provide the bases for the following precautions and limitations contained within OP O205.1, " Unit Shutdown": a) Minimize operation of the turbine at less than 40 MWe. b)- Allow sufficient time for steam dump to condenser valves to respond before calling for additional demand. c) Alternately place each charging pump in service with a 60 gpm orifice on line. QUESTION 7.12 ( .75) Describe the required operator actions if the following situation arises after the 3rd doubling during a 1/M approach to criticality. (Include any authorizations required to continue startup if no problems are discovered) a) Projected critical rod position is below the insertion limit. (.75) QUESTION 7.13 (1.00) Why is the S/G atmospheric PORV NOT isolated while performing E-3, "SGTR", but only has its setpoint increased? QUESTION 7.14 (1.00) EOP E-3, "SGTR", requires the operator to maintain AFW flow to the ruptured S/G until narrow range level is established. Provide two reasons for this procedural requirement. (***** END OF CATEGORY 07 *****)
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92__8Dd1N15IB8I1YE_EBQGEDUBEfi_ggNQ1IlgNSg_@NQ_ Lid 118IlgNE PAGE 13 OUESTION 8.01 (1.00)
-Which one of the following does NOT constitute an unreviewed safety question, involving 10CFR50.59, when evaluating proposed changes, tests and experiments to the facility or facility procedures?
a) If the margin for safety, as defined in the basis for any Tech Spec, is reduced. b) If a possibility for an accident or malfunction of a different type than initially evaluated for in the SAR is created. c) If the probability of occurrence of an accident or malfunction of equipment.important to safety previously evaluated in the SAR may be increased. d) If the suspension of a surveillance requirement for a Limiting Condition for Operation is instituted p'ursuant to section 4.0.2 Surveillance Requirements of Tech Specs. QUESTION 8.02 (1.00) Which of the f ollowing statements regarding plant clearances is correct?
- a. When executing a clearance order, each operation must be conducted in the order in which it was given.
- b. When returning systems to service, the electrical switching is done first.
- c. Written' permission from all clearance holders MUST be obtained before performing a temporary lift for test purposes.
- d. If conducting a clearance in a contaminated area, the ORIGINAL copy of the clearance is left at the entrance to the area and is initialed appropriately by the operator upon his exit.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
Es__eQd1NIEIB8I1YE_EBQGEQUBESt_G9NQ1119NEi_eND_LldlI8IlQUE PAGE 10 QUESTION 8.03 (1.00) Which statement below is correct regarding the administration of the temporary information tag system? a) White,_ green and red information tags are interchangeable in their use and application. b) If an information tag is found that is unauthorized, it may remain in effect if the PS-N then properly denotes his approval, c) The PS-N authorizes removal of a temporary information tag by signing the tag itself and routing it to the Operations Superintendent. d) Temporary Information tags may be used to serve the same purpose as a Caution tag, as long as this is noted on the Information tag itself. QUESTION 8.04 (1.00) Referring to the attached Tech Spec 3.20.1, which of the following is the basis for the requirement that there be >60,000. gallons in the DWST (assuming AFW is not available)? a) To allow for cooldown operations of DNE UNIT to Mode 5 in 12 hours. b) To allow for removal of decay heat from BOTH UNITS for up to 12 hours. , c) To allow time to establish RHR cooldown lineup. d) To allow time to establish makeup to the DWST. (***** CATEGORY OB CONTINUED ON NEXT PAGE *****) 1 l
Hz__8Dd1NIEIBOI1YE PBQGEQQB552_ggNQ1IlgNSa_8NQ_L151IBI1QNE PAGE 15 QUESTION 8.05 (1.00) When a fire suppression system and/or sprinkler system is INOPERABLE while the equipment it protects is required to be operable (eg. EDG Building Water Curtain) which one of the following correctly describes the required action? a) Commence a unit shutdown within 1 hour. b) Establish a Bi-hourly fire watch patrol for the affected area. c) Establish a continuous fire watch with backup fire suppression equipment around the affected area within 1 hour. d) Log ambient temperature readings for the affected area hourly. QUESTION 8.06 (1.00) During Unit 3 operation in. Mode 1, events occur that require application of Tech Spec 3.4.4.b (attached). Manipulations to bring Unit 3 to a Hot Shutdown condition are effected due to this Tech Spec ONLY. Which one of the following circumstances would require this course of action? a) A and C CCW pumps have been INOPERABLE for 18 hours when the B EDG is declared INOPERABLE. b)- A and B CCW pumps are operating when A CCW pump trips on over-current lockout. After 24 hours of being INOPERABLE, maintenance approximates another 18 hours to restore the A CCW pump. . c) CCW A pump and heat exchanger have been INOPERABLE for 10 hours. CCW C pump has been out of service and tagged for 48 hours. d) CCW pump A has been tagged for 48 hours when CCW B pump developes a seal leak which must be repaired. The pump is tagged and it is given a repair time estimate of 48 hours. QUESTION 8.07 ( .50) TRUE or FALSE: The Tech Spec definition of CORE ALTERATION includes a requirement that fuel must be in the vessel. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
Sz__8DMINIBIB8IIYE_EBQGED9BEU,_ggNQ1IlgNgz_@NQ_ Lid 118IlgNE PAGE 16 e QUESTION 8.08' (1.50) List the three meteorological conditions which would preclude conducting a routine Gaseous Waste release.
-QUESTION 8.09 ( .75) a) Where does the PS-N log requests for in-service testing.that must be delayed? (.25) b) What actions / notifications are required by the PS-N if a test.is "Unsat"? (0.5)
QUESTION B.10 ( .75) Fill in the blanks: Temporary Procedures (TP) to must be ______________. approved by the ___ _ _While active, for an amount of time not exceed a TP is retained in the _________ ___________ in the control room. QUESTION 8.11 (1.00) Using-the attached Technical Specifications, determine the Technical Specifications that apply (if any) for the following situation: Unit 4 is in-Mode 6, with B EDG out of service for the past 3
. hours for routine maintenance. 4A RHR pump, which has been in operation, was just secured to allow fuel movement along the periphery of the core. Reactor vessel level decreases due to undetermined reason, and falls below 23 feet 30 minutes after securing the 4A RHR pump.
QUESTION 8.12 (1.50) List 5 acceptable methods by which Independent Verification of electrical breaker alignments may be accomplished. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
Hz__8DMINISIB8Ilyg_P6QQEQQ6Eg2_QQNQ1IlgNgz_@ND_LidlI611QUE PAGE 17 QUESTION 8.13 (1.00) List.the three notifications that' the PS-N must make upon declaration of a significant event. QUESTION 8.14 (1.50) Technical Specification 3.9.2.c, "I-131, I-133, Tritium and Particulates",
' stipulates site boundary exposure. limits for members so the General Public.
Liste3 of the 4 pathways for these radionuclides to reach man that were used in the developement of these exposure limits. QUESTION 8.15 (1.00) If a General Emergency has been declared, due to offsite releases of undetermined levels, what is the minimum protective action that should be recommended? QUESTION 8.16 (1.00) As stated _in 10CFR50.54, under what conditions may actions be taken that depart from a license condition or a technical specification, and who, as a minimum, must approve such action? QUESTION 8.17- (1.50) What are the three borated water flow paths that can be considered to meet the requirements of Technical Specification 3.6.a (attached)7 l l 1 (***** END OF CATEGORY 08 *****) (************* END OF EXAMINATION ***************)
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10 LIMITING CONDITION 5 FOR OPERATION -' APPLICASE.ITY
.L 3.0.1 When a Limitin6 C 'ondition for Operation is not met,' except as provided in the associated AC lON requirements, within I hour action shall be initiated to place the unit in a MODE in whid the specification does not apply by placing it, as applicable, im a) At least HOT STANDBY wit %1n the next 6 hours,' .
b) At least HOT SHUTDOWN within the following 6 hours, and* c) At least COLO 5HUTDOWN within the subsequent 24 hours. Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the ; specified time limits as measured from the time of failure to meet the Limitirg Condition for Operation. Exceptions to these requirements are stated in the individual specifications. j l This specification is not applicable in MODE 5 or 6. I 3.0.2 Non-compliance with a soecification 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 l I Operation is restored prior to expiration of the specified time intervals, l completion of the ACTION requirements is not required. 3.0.3 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is requirbd during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the l Limiting Conditions for Operation, the associated ACTION requirements shall be met. 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This . ! provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications. For purposes of determining if a component, is operable for LCO J l 3.0.5 considerations, the . component need not be con'sidered inoperable due to inoperability of its normal or emergency power supply if all of its redundant
! components are operable with their normal or emergency power supplies operable.
' NOTE: Until full conversion to STS, when a LCO action statement requires a unit to be placed in HOT SHUTOOWN within 6 hours, refer to Table 1.1 and place
! the unit on the required status to meet the HOT STANDBY MODE. 1 1 3.01 Amendment Nos.114 and 108 i
s
,, ON(' emerjcricy tentainmeSt tooitng untt Say 03 out er scev1Co for a 'er'gd Jr la 9 Curs. de'or tJ *ntitattny natntcnance
' :ne otner TWO unt ts snel t se testeo to ammunstrate operactitty.
- 2. CNE contaimnent spray pump say ou out or service provices it is restoree to opereele status witnin 24 nours. The remaining containment spray pump snall ou tested to demonstrate operability terure initiating metatenance or tne' inoperacle pump.
- 3. Any salve in tne system may se inoperaole provideo repates are completec witnin 24 nours. Prior to initiating repetrs, all valves tnat provide tas dupiteate function snali ne tasted to demonstrate operanslity.
- 3. EMERGE' ICY CONTAINMENT FILTERING $YSTEM
- 4. The reactor snail not se mace critical, except for iow pcwor pnysics tests unless:
- 1. THREE wnergency containment rtitering units are ~
operacle.
- 2. All valves, interlocxs ano sising associated witn tne aoove* components anc required for ,;ost-acc1 cent operatton, are opersona. .
B. During power operation:
- 1. CNE unit say te inocertole for'4 ;erloc of 7 cays 1r t.e utner TWO are operaole.
- 2. An'y valve in tne system say se inoperaole provtceo repairs are completed witnin 7 cays. Prior to initiating saintenance, all valves tnat provide tne '
duplicate function snail ne tested to demonstrate opera 0111 ty .
- 3. If af ter 7 days tne unit is still inoperaole
' Specification 3.3.1 applies to 3.4.3.o.
- a. , COMPONENT COOLING SYSTEM
- a. The reactor snall not ne made critical, except for low power pnysics tests unless tan rollowing conoittuns are met:
- 1. THREE ccmponent cooling pumps are operaole.
- 2. ThAEE component cocitng neat 'excnangers are overaole.
- 3. All valves, interiocas anc piping assoc 14 tea witn tne aoove components are operacle. .
I 3.4 4 Amendments 83 & 77
- 3. Ouring :cwee scoratton, taa requirements or J.4.4.a may 33
-: SodtfiecLas states Scluw. [r CA3 system ts not rostaree to meet tne const tions or 3.4.4.a es tnin tne time sectoo speci rtec, tne reactor snall te pieces in the not snutoown t anoi t t un. Ir tne requirements or 3.4.4.a are not satt srisc witnin an acottional 48 nours. tne reactor snail na unaces in tne cole snutdown concitten. Toect rication J.0.1 appites . s to 3.4.4.0.
~ ~
- 1. ONE pump may be out of serv 1Cd for 7 odys.
- 2. ONC additional puso ane ONC tiest excnanger may se out ur 4
i service for perios et 24 nours. 1
- 5. IMTAKE C00LIMG WATER SYSTEM
.e
- 4. The reactor snall not te made cr) 1 cal unless tne rollowing conditions art met:
- 1. THREE intake cooling water pumps ano TWO neacers are l
operacle.
- 2. All valves, interlocks anc piptny associatec 1 n the operation of tnese pumos. anc required for post accident operetton, are oparacle.
- 5. During power operation, tan requtruments or 3.4.5.4.,
aoove may be modi rted to allow any one or tne rollowing camponents to De inocertole proviced tne remaining systems are in continuous cperation. It tne system 1s not restorec to meet tan requirements or 3.4.5.4. wt:nta tne time perloc specirtec, tne reactor stiali ce placec in the not snutcome concitton. [f tne requirements or 3.4.5.a are not sattsried witnin an eccit.onal 44 nours, i ' tne reactor shall na placec in tne coto snutouwn concitton. Speci r1 cation 3.0.1 appi t es to 3.4.5.s,.
- 1. One of tne two neaoers may me out of service rar a pertoe of 24 nours.
- 2. One intake cooling water pump say se out or service ror a pertoe of 24 nours.
l 3,45 Amendments 83 & 77 l a
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3.6 CH.DtICAL A.C VOL*'Mr. COSTROL SYS;2( Aeolicabiliev: App 3 tes to the operational status of the Chemical and Volume Control Systes. - . Obiec tive: To defina those canditions of the Chemical and Volume _ , Control Systes necessary to ensure safe reactor operation.
~
Specifiestion: a. When fuel is in the teactor there shall be at least one . b. flow path to tha. core'for baron injection.
- b. A reactor shall not be made critical unless the following Chemical and volume Cont:o1 systes conditions are sets
- 1. 740 associated charging pumps shall be opera' ole. .
- 2. 740 peric' acid transf er pumps shall be " operable.
- 3. The boric acid canks in servica shall contai= a total !
'of at least 3,080 gallons of a 20,000 to 22,500 pp=
l boren solution at a temperature of at least i i 145 T.
!. . System piping, interlocks and valves shall be operasia '
- to the extent of establishing one flow path from the boric acid tanks, and one llev path from 'the refueling water storage tank, to the 11eactor Coolant Systas. l l
l
- 5. 740 channata of heat tracing shall be operahla, for de j flow path from de horic acid tanks. >
,I
- l l
- 6. Th's pr1=ary votar st: rage ca:k conta'iss not less than .
1 30,000 gallons of vatar. ..
- c. he second reaccar shall not be =ade critical unless the .
following conditions are :st: 1 e i M endent Ms. 73 & C i 3.6-1 . J e =-
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3,10.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULAT10N 3.10.7.1 HIGH WATER LEVEL At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.' APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet. ACTION: . With no RHR loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosohere within 4 hours.
*The RHR loop may be removed from operation for up to I hour per S-hour period during the performance of CORE ALTERATIONS in
- the vicinity of the reactor vessel hot legs. -
! 3.10.7.2 LOW WATER LEVEL ,
Two independent residual heat re noval (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation. . 1 I APPLICABILITY: MODE 6, when the water level above the top of
! the reactor vessel flange is less than 23 feet.
- l. ACTION:
i a) Wit'h less than the required RHR loops OPERAB'LE, immediately initiate corrective action to return the required l RHR loops to OPERABLE status, or to establish greater than i or e , sal to 23 feet of water above the reactor vessel flange,
- as soon as possible.
j b) With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the 1 containment atmosphere to the outside atmosphere within 4 Mws 3.10.8 BORON CONCENTRATION The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be sufficient to ensure that the more restrictive of the following reactivity conditions is met; eithers l
- l .
3.10-3 Amendment Nos.11L and igg,
i i STANDBY PEEDWATER SYSTEM 3.20.1 Two standby feedwater pumps shall be available* and at least 60,000 gallons of water (available volume), shall be in the Demineralized Water Storage Tank.' APPLICABILITY: MODES 1, 2, 3 ACTION:
- 1. With one standby feedwater pump unavailable, restore the unavailable pump to available status within 30 days or submit a SPECIAL REPORT per 6.9.3.J.
- 2. With both standby feedwater pumps unavailable:
a) Within 24 hours, notify the NRC and provide cause for unavailability and plans to restore pump (s) to available status and, b) Submit a SPECIAL REPORT per 6.9.3.J.
- 3. With less than 60,000 gallons of water in the Demineralized Water Storage Tank restore the available volume to at least 60,000 gallons within 24 hours or submit a SPECIAL REPORT per 6.9.3.J.
- 4. The provisions of Specifications 3.0.1 and 3.0.4 are not applicable.
e
- These pumps are not safety related equipment and do not require plant safety related emergency power sources for availability.
** The Demineralized Water Storage Tank is non-safety grade.
3.20-1 Amendment Nos.118 and U2
jt__INEQBy_QE_Nyg6EeB_EgygB_EbeNI_QEE8911QNt_E(ylDS,_6ND)jk([ PAGE' 18 lbEBdggyN6dlGQ . k.
' ANSWERS ~- TURKEY POINT 3&4' -86/12/15-DEAN,.W M l
ANSWER 5.01 (1.00) , d L: REFERENCE Westinghouse Reactor Physics, Section I-5, MTC and Power Defect DPC, Fundamentals of Nuclear Reactor Engineering St Lucie Reactor Physics, Secti on 7. 6 t< 7. 7 039/000; A2.05(3.3/3.6) ANSWER 5.02 (1.00)
'b REFERENCE f EIH: GPNT,Vol VII, Chapter 10.1-83-86 BSEP: L/P O2-2/3-A, pp 172 - 176; O2-OG-A, pp 57 - 60 Westinghouse Nuclear Reactor Theory, pp. I-5.77 - 79
. Turkey Point, Reactor Core Control, pp. 4 28 OO1/OOO-K5.13 (3.7/4.0)
ANSWER 5.03 (1.50) keer
- a. .Hia;Asw. (+.5 ea)
- b. L-ower(6 9ker-
- c. :!!i ghr:he REFERENCE NUS Vol 4, pp 2.2-4 015/000 A1.01(3.5/3.8)
ANSWER 5.04 (1.00) C REFERENCE Westinghouse Reactor Core Control, pp 3-34/35 001/0003 K5.14(2.3/2.8)
h-. L.Et__THEDRY OF NUCLEA8_PQWEB_ PLAN 1_QPERATIQN t _ELylpSt _9NQ PAGE' 19 IHgBdQQ1N@dlGS-ANSWERS -- TURKEY POINT 3&4 -86/12/15-DEAN, WM ANSWER- 5.05 (1.50) a) ; Lower (+.5 ea) b) Lower c) Higher REFERENCE-ST Lucie OP 0030126 and Plant Curves CNTO Reactor Core Control, Section 7
.OO1/000;'A2.07(3.6/4.2)
ANSWER 5.06 (1.50) a) Density changes of the moderator with core height (+.5) Hb ) Fl u:< will' shift significantly towards the top (+.5), due to uneven fuel burnup (+.5) (higher density fuel at the top) REFERENCE Westinghouse Reactor Core Control, pp 3-51/53 001/000; K5.29(3.7/3.9) & K5.30(2.9/3.1) ANSWER 5.07 (1.00) a) Will be the same g(uy +.5J ea) b b) Unit B will be h gher REFERENCE Westinghouse RLactor Core Control, pp 6-23/26 Westinghouse Fundamentals of Nuclear Reactor Theory, pp 8-48/60 001/010; K5.08(2.9/3.2) & 001/000: K1.05(4.5/4.4)
.5 __IMEQBy_QE_NQCLEAR' 3 POWER PLANT OPER@TIQNz _ELylp53_@ND PAGE 20-INEBb9DXN951CS ANSWERS -- TURKEY POINT 3&4 -86/12/15-DEAN, WM ANSWER 5.08 (1.00)
Due to'the greater decrease in the temperature of the coolant exiting the core relative to the decrease of the inlet coolant, (+.5) more positive reactivity will be added in the upper core regions, resulting in a more positve (less negative) AFD (+.5) REFERENCE Westinghouse Nuclear Training Operations, Ch 8 Surry ND-86.2-LP-8, pp 8.14/15 001/000; K5.29 (3.7/3.9) ANSWER 5.09 (2.00)
- a. Subcooling'is based on core exit T/Cs or hot leg RTD readings. During natural circulation the mass of metal in the head can retain heat and keep local temperatures above saturation. The temperature indicators.
would not reflect this local saturated condition. (1.0)
- b. Pressurizer level decreases because the pressurizer pressure increase will compress the vossel void and force water out of the pressuricer.
(1.0) REFERENCE G.P. Heat Transfer and FF Pp 355-358 EPE-074; EA2.05(3.4/4.2) b EA2.07(4.1/4.7) ANSWER 5.10 (1.00) When slightly open, the pressure drop is across the valve itself as there is isenthalpic expansion (+.5). As the valve is fully opened, head losses in the piping become more significant (+.5) REFERENCE Westinghouse Thermal / Hydraulic Principles, pp 10-71/73 010/0003 K5.02(2.6/3.0)
=
.Qt__INE96Y_9E_N9GLE0B_E9 WEB _ELONI_QEEB811QN i_ELQ1QS _@NQ t .PAGE. 21 IUEBU99YN001GS-
-ANSWERS ---TURKEY. POINT 3&4 -86/12/15-DEAN, WM
' ANSWER 5.11 (1.50) a) Tstm will drop rapidly (+.25) as Tavg decreases in order to maintain the Delta T across the S/G tubes to provide the heat transfer to support 90% load demand (+.5) b) Even though-pressurizer pressure decreases and power increases, which decreases the margin to saturation, the large cooldown of the RCS (+.5) overcompensates for this and the margin to DNB increases (+.25)
REFERENCE Westinghouse Transient and Accident Analysis 035/010; K1.09(3.8/4.0) ANSWER 5.12 (1.50) a) prevent significant occurrence of =irc-water reaction (+.5) b) as power drops, the enthalpy rise across the core will be less (+.25) with smaller enthalpy increase, the coolant is further from DNB, allowing a higher F(n) Delta h (+.75) (This allows changes in the radial power shape f or all permissable rod insertion limits) REFERENCE TPT TS B3.2-5 Westinghouse Thermal / Hydraulic Principles II, pp 13-15/16 001/000; K5.46(2.3/3.6) ANSWER 5.13 (1.50) a) > DNB, have partial film boiling, where the fuel rod is alternately covered with steam and water (+.25) Steam has poor thermal conductiv-ity capabilities (+.25), so heat transfer rate drops and Delta T rises (+.25) b) As fuel surface temperatures rise, stable steam layer forms (+.25) causing a further increase in fuel rod temperatures (+.25). Eventually, significant radiative heat transfer occurs causing heat xfer rate to incarcase (+.25) REFERENCE Westinghouse Thermal / Hydraulic Principles II, pp 13-18/20
'.9z__1BE98Y_9E_ggCLg@g_ggWE8_EL@NI_QEgB8IlQN3 _ELylQS _@NQ i PAGE 22
. IHEBDQQyN@D1QS ANSWERS a TURKEY. POINT 3t<4 -86/12/15-DEAN, WM
.EPE-074;;EK1.02(4.6/4.8)
ANSWER 5.14 (1.00)
~ As rods are inserted, they provide an' additional leakage path (+.25) thus, MTC is more negative as rods are inserted.(+.25). Therefore, Baron concentration, which will make MTC less negative due to its absorbtion abilities can be increased while still maintaining MTC within limits (+.25)
LV*t REFERENCE Westinghouse Reactor Core Control, pp 3-16/22 001/000; K5.09(3.5/3.7)
r - 6E_ECONI_EYEIEUS_DESIGNg_CQNIBQLt_@NQ_lNSIBydgNI@llQN PAGE'- 23 ANSWER 3 -- TURKEY-POINT 3&O: -26/12/15-DEAN,.W M ANSWER 6.01 . ( 1. 00 ) d REFERENCE TPT LER-86-234 of 3 June 1986; TPT DWG 5610-T-L1, Sheet 4H2
-EFE-056; PWG-6(3.5/3.7)
ANSWER 6.02 (1.00) C REFERENCE TPT HD 1029-OL, APP A, pp 7 ! 022/000; K1.01(3.5/3.7) ANSWER 6.03 (1.00) l *
. REFERENCE TPT SD 144, pp 12-14 (TPT OBNK 20) 062/0003 K4.10(3.1/3.5)
ANSWER 6.04 (1.00) a) Limit switches on the steam supply MOVs (+.5 ea) b) N2 Station #4 REFERENCE TPT HD 1117-OL, APP A, pp 6/7 061/000; K6.01(2.5/2.8) & A2.07(3.4/3.5) [
- 6t__ELONI_EYSIEUS_ DESIGNt _GQNIBQLt_8NQ_INSIByMENIBIlQN PAGE 23 ANSWER] -- TURKEY POINT 3&4 -C6/12/15-DEAN, WM ANSWER 6.05 (1.50) a) Prevent damage from steam cutting or H2 gas acting on the seats'(+.5) b) False (+.5) c) > 25*/.~ flow signal on ANY safety valve acoustic monitor (+.25)
PORV leaving the closed position (+.25) REFERENCE TPT SD 9, pp 27-29 010/000; PWG-4(3.6/3.7), K1.05(3.4/3.6), K4.03(3.8/4.1) ANSWER 6.06 (2.00) a) (1) Rx trip bkr "A" and Bypass "A" open (+.25 ea) (2) "B" "B" (3) Generator Lockout Relay Tripped (4) Turbine Trip Pushbutton (5) $ kii S/G 1evel (6) SI Signal b) Test handle is put into the " Test" position (+.25) and held there(+.25) during the test. REFERENCE TPT LER-86-431 of 21 oct 1986; TPT SD 127, pp 24, Fig 14 045/050; K1.01(3.4/3.6) & K4.07(2.5/2.8) ANSWER 6.07 (1.50) a) overcurrent (+.25) 4KV Bus stripping relay underfrequency 4KV Bus lockout relay b) Ensure adequate oil flow and distribution prior to pump start (+.5) REFERENCE TPT SD 8, pp 24, Fig 17 003/000; PWG-7(3.5/3.9) 062/000; K4.02(2.5/2.7)
4t__ELONI_EYSIEd@_QE@l@N g _GQNI6QLg_6NQ_lNSIBUDENIsligN PAGE 25
~ ANSWER 3 -- TURNEY' POINT 3&4 -26/12/15-DEAN, WM ANSWER 6.08 (1.50) a) PG Syst'em .25 ea) Wwrsl tic} OJM (f.3 3 to ) 1Ng n CVCS B1 er gp Boro ecovery Tanks b9 S# q%
M F e Main SY#5 N#I_.S b) -The center-to-center distance ensures < .95 Keff for spent fuel, (even-if unborated water is used) + . 5.) REFERENCE
'M # " b N MN TPT SD41 " Spent Fuel Pool Cooling, Purification and Ventilation", pp 4-6 Surry ND-92.5-LP-6, pp 6.5/8 033/000; K4.05(3.1/3.3)
ANSWER 6.09 (1.00)
-Operate the solenoid bypass switch to he " Bypass" position in the cabinet under the control consol e (2.5-::) (AM(J g
-Use the gripper control knob on the console to operate the gripper [4.UJ REFERENCE TPT SD 44, pp 33 t< Fi g 13 034/000; PWG-6(2.7/3.3)
ANSWER 6.10 (1.00) If there were a slow RCS depressurization in conjunction with a loss of instrument air (+.25) the miniflow recire valvos would close (+.25) and the SI pumps could then overheat and incur damage due to operating at shutoff head too long (+.5) REFERENCE TPT LER-86-269 006/020; K4.01(2.7/3.0) (Plant specific emphasized question) l L
iz__eL9NI_gy@Igdg_QEgigNg_CQNIBQ63_8NQ_INgI890gNI811gN PAGE 2S [ - ANSWERS -- TURKEY POINT 3&4- -86/12/15-DEAN, WM ANSWER 6.11 (1.00) Turbine Trips (+.5) due to 2/3 OTDelta T logic being made up (+.5) REFERENCE TPT DWG 5610-T-D-14 012/000; PWG-10(4.4/4.7) ANSWER 6.12 (1.50) a) Providg turbocharger prestart lubrication lndpostshutdownbearing cooling) (+.75 ea) b) Cause oil to collect in turbocharger and cause a fire REFERENCE TPT HD 1137-OL, APP A, pp3 064/050; PWG-7(3.6/4.0) ANSWER 6.13 (1.50) CCW and ICW pumps trip (+.5) The Sequencer timer resets and starts after
.a 36 second time delay (+.5) then normal sequencing starts the ICW and CCW pumps (+.5)
REFERENCE TPT SD 170, pp 15-17 (TPT ODNK design 25) 064/000; K4.10/4.11(3.5/4.0) l ANSWER 6.14 (1.50) a) Structural failure of component resulting in leakage (+.4) >50gpm (+.1) b) Injection: 1 Active only (+.5 ea) Recirculation: 1 Active OR 1 Passive l REFERENCE l TPT HD 1021-OL, APP A, pp 2 10CFR50, APP A ! 006/050 PWG-4(4.2/4.3)
.______a
.Zs.__E80GEPQBE5_ _NQB5ebi_0ENQ8d@ba_EDE8QENQy_8NQ PAGE 27 --
60Q10b091GOL_QQN18Qb ANSWERS - TURKEY POINT 38e4 -86/12/15-DEAN, WM ANSWER 7.01 (1.00) C REFERENCE TPT ONOP 046.1, pp 4 EPE-024g EK3.01(4.1/4.4) ANSWER- 7.02 (1.00) e
. REFERENCE FR-Z.2 EPE-069; EK3.01(3.8/4.2) l ANSWER 7.03 (1.50) a) True (+.5 ea) b) True c) False REFERENCE Westinghouse User's Guide, pp 5, 17, 18 PWG-22(4.3/4.3)
ANSWER 7.04 (2.00) y^ 1) Increase Injection Flow (+.3 for technique, +. 2 f or posi tion)
- 2) Depressurice S/Gs
' 3) Start RCPs
- 4) Open RCS vent paths (PORVs)
REFERENCE FR-C.13 (TPT OBNK H-5) l EFE-074; EK1.03(4.5/4.9)
'.Zt__E60GEDQBEE_:_NQ656Lt_@@NQ6d@Lt_EDEBGENQy_8NQ PAGE 28 689196901G06_GQNIBQL ANSWERS -- TURKEY POINT 3&4 -86/12/15-DEAN, WM ANSWER 7.05 (2.25)
- 1) Ensure f ollowing personnel are dispatched with appropriate attachments:
RCO 1 & 23 N.O. 1; N.O. 2; 3rd licensed RCO; Nuclear Turbine Operator (
- 2) Take these items from control room and distributes radios; d4 ICCS Valve keys; Electrical Penetration Room Keys; plant logs _for both units; Plant Supervisor Logs 1 pair of wire cutters (* ! . 0)
- 3) Proceed to TSC (+.25) ,&F9)P)(FF0
\
ANSWER 7.06 (1.00) j R-2 (Containment Operating Floor- 58' level) (+.25 ea) R-7 (SF Bldg Xfer Canal Area- 58' level) R-19(SF Bldg Exhaust Duct) R-21 (SF Bldg North Wall ) REFERENCE TPT OP 16002.6, pp 8/9 072/000; PWG-7(2.9/3.3) ANSWER 7.07 (1.00)
- 1) Check basket strainer for high dp and backflush (+.5 ea)
- 2) Check temperature control valve and put in manual or bypass REFERENCE TPT ONOP 3408.1, pp 2 008/000 PWG-11(3.9/4.1)
ANSWER 7.08 (1.00) The bistables associated with the instrument should be placed in the test position (+.5) within 30 minutes (+.25) for further investigation by I&C personnel (+.25) REFERENCE TPT ONOP O208.14, pp 2 012/000 PWG-7(3.7/4.2)
o Z t _ _EB9G E DWB EE _:,NQB d@( t_@@NQ65@( g _E DE 6@gNQ y _@N Q PAGE. 29 B60106901G86_GQNIBDL ANSWERS -- TURKEY POINT 344 -86/12/15-DEAN, WM I ANSWER 7.09 (2.00) a) -Manually trip turbine (+.33 ea)
-If can't, then manually runback turbine
-If can't, close main steam line isolation and bypass valves b) (1) Verify PORV and block valves open (+.25 ea) if not, open them (2) Verify containment vent isolation i if dampers not closed, manually close them REFERENCE FR-S.1 EPE-029 PWG-11(4.5/4.7)
ANSWER 7.10 (1.00) i To ensure decay heat is low enough (<12 MWT)to prevent ( +3 5 ) steam / water binding in the S/Gs A$$$$P if the break is in the cold leg (+.25) REFERENCE TPT HO 1021-OL, pp 9 EPE-011; EK3.13(3.8/4.2) ANSWER 7.11 (1.50) a) Prevent overheating the turbine (+.5 ea) b) Prevent causing an inadvertant saf ety injection c) Ensure all charging pumps can be used to control pressuri:er level REFERENCE TPT OP O205.1, pp 2, 3, 7 PWG-7;(3.5/4.0) l l _ . _ _ _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _
Zt__880CEDQ8EE_:_NQBd6L t_0ENQ6d@bt_EDESQENGL6NQ - PAGE 30 6801DLQQ1GOL_QQNI6QL ANSWERS -- TURKEY POINT 3&4 -86/12/15-DEAN, WM ANSWER 7.12 ( .75) a) Reinsert the control banks and borate as needed, PS-N authorize startup (+.75) REFERENCE TPT OP O202.2, pp 3 001/0103 A2.07(3.6/4.2) ANSWER 7.13 (1.00) This allows S/G Pressure protection without having to depend on the code safeties (+.75) which could lif t and not reseat causing an unisolable steam leak (+.25) REFERENCE Westinghouse Background Documents (TPT OBNK E-16) EPE-03Og EK3.06(4.2/4.5) ANSWER 7.14 (1.00)
- 1) (Promote thermal stratification) so ruptured S/G doesn't depressurire during cooldown (+.5 ea)
- 2) Ensures S/G available as a heat sink if required REFERENCE Westinghouse ERG Bbckground Documents (TPT OBNK 10)
EPE-030; EK3.06(4.2/4.5)
p y
~ '9t__eDNIN1EIB011W EBQGEDUBEEt_GQND1110 net _8ND_ Lit!1IGIl0NE PAGE 31
- ANSWERS -- TURKEY POINT 38,4 -86/12/15-DEAN, WM -
4 ANSWER O.01 (1.00) ! d ' REFERENCE 10CFR50.59, TS 4.0.2 , PWG-23(2.8/3.5) i ANSWER 8.02 (1.00) a REFERENCE TPT ADM 0103.4, pp 13-15 PWG-14: Tagging / Clearances (3.6/4.0) ANSWER 8.03 (1.00) b REFERENCE TPT AP 0103.36, pp 2-5 PWG-14(3.6/4.O) ANSWER 8.04 (1.00) d . REFERENCE TPT TS B.20.1 061/000 PWG-5(3.3/4.1) L
p< x,
. y xq. v y 7
- e (r
_ _ ,7 _ p
'Es..l_eQUIN1818eI1YE_EBQGEDWBEE _GQUD1I19NE _OND_ Lit!1IeI1QUE PAGE 32 wi 4 . , ,
-26/12/15-DEAN, WM
., k,,"'yz .1+y'JSW.Fi! pg~~ TURKEY POINT 36(4
, j f '( '< s !., .;
Y. '? . .,
A i
.f' j p/,U_,
Y ANSWT.R ,' 8.05 .c c_(1.00)
! IU ,y n e.
i k,'q d/= -,) ). L - i,, t ' 4 --
s l's
'NkFERENC
'iPT*TO3'.l'M.3 dj,Q -I g 3 006/000; Fwg-5(2.,8/3.7)
N - ,< g . V fh -
/si .
jANSWER r\ " O . 06. ' ' (1.00) 1{2 / d
- i
'p y
'9 #
'ItEFEFt3NCE h
g (TPTdTJ t.4.4.6 - 9 i, j' 008 '0303 PWG-t(2.9/3.7) r e(, - ! . ANT,WER O.07 ( .50)
' Uue (+.5) 6 ,'/ i
, ;- REFERENCC l 7e/.T&T,TS Definition 1.10
- t ;
PWG-5(2.9/3.9) y ANSWER 8.08 (1.50) 1
- 1) wind speed (+.4) < 10 mph (+.1)
- 2) Hind direction (+.3) from the south or northeast (+.2)-towards scout
- 3) precipi,tation is falling (+.5) camps REFERENCE TPT GD50 " Gaseous Waste Disposal System", pp 25p 071/000g PWG-7(3.2/3.7) 0 I q
-)
+
s i 4 O
St__0051NIEIBBI1YE_E60GEDVBEEi_GQNQlllQNSg_8NQ_(ldlI@llQNE PAGE 33. ANSWERS'-- TURKEV POINT 3&4 -86/12/15-DEAN, WM F
.f. . t ANSWER 8.09 (. . 75 )'
f M'l6 a) Equipment out of service log "(+.'25 ea) b) -Describe malfunction in Remarks section of the test sheet
-Notify Maintenance W
a) prwch wk I" NMcLS Ccd<!~ DE fa2yal4thvJ REFERENCE . b) Halt dMTesMwako fe,r &,s4tNap grek-1 Ot03 prf 7-/ T(Tosi('uc,tggo
- SYM S In Con $"W l09I PWG-1(3.5/3.9)
ANSWER- 8.10 ( .75) Pl ant ' Manager; one year; Temporary Procedure Log (+.25 ea) REFERENCE TPT AP 0109.6, pp 2-4 f PWG-23(2.8/3.5) I ANSWER 8.11 (1.00) TS 3.10.7.2 (a) and (b) apply (+.5 ea) REFERENCE TPT TS 3.10.7.2 EPE-025; PWG-5(3.0/3.9) ANSWER 8.12 (1.50)
- 1) Visual inspection of breaker position (+.3 ea)
- 2) Bieaker light indication
- 3) Functional Test (eg. voltmeter)
'4) Local (or Remote) Instrumentation
- 5) Annunciators
- 6) Switch positicn REFERENCE TPT AP 031, pp 8 PWG-13: Conduct / Verify Valve Lineups (3.7/4.0)
3 m 7_ - -
- et__0DUINigI69Ilyg_889Qgpg85@g_QQNQlI1QN@g_9NQ_ Lid 1I9IlgNS PAGE 34 ANSWER 3 - . TURKEY. POINT 3&C -C6/12/15-DEAN, WM ANSWER 8.13 (1.00)
- 1) Duty Call Supervisor (+.33)
- 2) VP, Nuclear Operations (NEDO during off hours) 3)- NRCOC Duty Officer REFERENCE TPT AP 1 0 3 . 1 2 ,~ pp 2 PWG-2(2.7/3.8)
ANSWER 8.14 (1.50)
- 1) Individual inhalation- (+.5 ea for any 3)
- 2) Deposition on leafy vegetables with subsequent ingestion by man
- 3) Deposition on grassy areas, eaten by milk / meat animals then by man
- 4) Deposition on the ground with subsequent exposure to man REFERENCE TPT TS B 3.9.4 EPE-060; PWG-5(2.9/3.9)
ANSWER 8.15 (1.00) Shelter all people (+.5) within a 2 mile radius and (+.25) 5 miles in the downwind sectors (+.25) REFERENCE TPT EP 20101, pp 2 PWG-36(2.9/4.7) ANSWER 8.16 (1.00) In an emergency when' the action is needed to protect the health and safety of the public (+.75), approved by at least a licensed SRO (+.25). REFERENCE
-10CFR50.54 PWG-36(2.9/4.7)
. *8t__8Dd1N1518811YE_EBQQEDWBEQi_QQNQlIlgN@g_@NQ_LidlI@I1QNE PAGE' 35 ANSWERS -- TURKEY POINT 3&4 -@6/12/15-DEAN, WM
' ANSWER 8.17 (1.50)
- 1) Boric Acid Tanks to-Boric Acid Xfer Pumps to charging pumps (+.5 ea)
.2) Alternate suction of charging pumps to the RWST
- 3) SI pumps taking a suction from RWST and inject via BIT line (BIT no longer contains borated water)
REFERENCE TPT TS B3.6 004/020; PWG-5(2.9/4.1) i i
~
12/15/86
, ENCLOSURE 3 Pcge1 NRC Exam Section 5 QUESTION SRO 5.03 (1.50)
Indicate whether the following will cause the power range 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.
- a. If the feedwater temperature used in the calorimetric was higher than actual feedwater temperature.
- b. If the reactor coolant pump heat input used in the calorimetric is omitted.
- c. If the steam flow used in the calorimetric was lower than actual.
ANSWER:
- a. Higher (+.5 ea)
- b. Lower
- c. Higher
RESPONSE
i i- Request that the Answer Key be changed to reflect the following: I
- a. Lower
- b. Higher
- c. Lower i '
l
REFERENCES:
l i l See Following Attachment.
- RI-5/cas l
~~
12/15/86 QUESTION SRO 5.03 (Attachment) This question says the instrument was adjusted to 100% based on a calculated calormetric. This means that the calormetric came out to 100%. The only thing that is wrong is the answer. Correct answers should be:
- a. Lower
- b. Higher-
- c. Lower The assumption in this question is "Ifinstrument has been adjusted to 100% based on a calculated calormetric." Implicit in this assumption is that calculated calormetric is 100% for each case but that an error was rnade each time.
a. 1Q gg = 3ig.,.,, (hg73, - h, ,L) + 31gg @gg - hp ,L ) Analyze how actual conditions affect each term, actual feedwater temp. is lower, thus actual hFw is lower. Actual power is higher than indicated or indicated is lower than actual power. b. L Q ,,,,,,= Qsc ~ G ecP i=U" QRCP = 0 for calculated, actualis 8MW. So actual power is lower than indicated or indicated is higher than actual power. c. l t Qsa
- I AIsrat(hsrar -h g j + Sig(hgp - hfg) l Actual steam flow is higher than calculated RX
- ASG ~ RCP
! Actual power > indicated power indicated is lower than actual power.
*R1-5/cas
- i
12/15/86
. Pags2 NRC Exam Section 5
. QUESTION SRO 5.05. (1.5)
An ECC is calculated for a startup following a reactor trip from 50% power equilibrium xenon (BOL). Indicate if the actual critical rod position will be HIGHER, LOWER or the SAME from the calculated position for each of the following situations. Use attached curves as appropriate and treat each case individually.
- a. Xenon reactivity curve for a trip from 100% is used to calculate conditions to startup 20 hours after the trip.
- b. Boron worth at EOL is used vice BOL worth. (Dilution is required to reach desired Boron concentration in both cases)
- c. The EOL critical Baron Concentration is used instead of the BOL critical Baron concentration.
ANSWER:
~
- a. Lower ( +.5 ea)
- b. Lower
- c. Higher RESPONSE: 5.05c j We request that the Answer Key be expanded to accept Lower also. If the Boron Letdown Curve for End of Life was used, this would require a RCS Dilution to a lower than required Boron concentration prior to withdrawals. Due to the lower than required Boron concentration the critical rod position would be much lower than the predicted rod position. -
REFERENCES:
l l
*R1-5/cas
~
12/15/86-
. Page 3 NRC Exam Section 5 QUESTION SRO 5.07 (1.0)-
Unit A is at EOL'while Unit B hasjust been started up after a refueling. Assuming a rod speed of 48 spm, both reactors are taken critical by pulling in 50 step increments and waiting 60 seconds before pulling again. Assuming all systems and parameters are identical at the commencement of the startup, and both units are initially shutdown by 2%(delta k/k);
- a. Which Unit will have the. highest source range counts when criticality is reached?
- b. How will critical rod heights compare in the two Units?
ANSWER:
- a. Will be the same (+.5 ea)
- b. Unit B will be higher
~
RESPONSE:5.07a l Recommend SAMEhnit B HIGHER be accepted. Since unit B has a higher rod height,it will take longer to reach criticality. Since 60 seconds may not be sufficient l to establish completely stable countrate, the longer startup time may result in a l higher countrate when the reactor attains criticality, i
REFERENCES:
l *RI-5'cas
~
12/15/86
- Page4 NRC Exam Section QUESTION SRO 5.11 (1.50)
Assume the plant is at a steady power level of 75%, with rod control in MANUAL and the plant in BOL, when a 15% step increase in turbine load occurs. DESCRIBE and EXPLAIN the behavior of the following parameters during the first several minutes of the transient. Assume MTC is -3 pcm/F and NO RX TRIP occurs.
- a. Tstm
- b. margin to DNB ANSWER:
- a. Tstm will drop rapidly ( + .25) as Tavg decreases in order to maintain the Delta T across the S/G tubes to provide the heat transfer to support 90%
load demand ( +.5)
- b. Even though pressurizer pressure decreases and power increases, which decreases the margin to saturation, the large cooldown of the RCS ( +.5)
~
biercompensate for this and the margin to DNB increases ( +.25) RESPONSE: 5.11b Recommend accepting margin increases due to temperature decrease if student explains using the temperature change.
REFERENCES:
i f' l { l
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12/15/86
. Page 5 NRC Exam Section 6 QUESTION SRO 6.06 (2.00)
- a. List the 6 devices / signals whi:h will actuate the 20AST and 20ASB Auto Stop Trip Solenoids for the Main Turbine (1.5)
- b. Describe how the test handle cavice is used to prevent a turbine trip when testing the overspeed trip mech inism. (0.5)
ANSWER:
- a. (1) Rx trip bkr "A" and Bypass "A" open ( + .25 ea)
(2) Rx trip bkr "B" and Bypass "B"open (3) Generator Lockout Relay Tripped (4) Turbine Trip Pushbutton (5) Hi Hi S/G Level (6) SISignal
- b. Test handle is put into the " Test" position (+.25) and held there (+.25) during the test.
RESPONSE: 6.06a
- 1) Should read Rx Trip BKR "A" and Bypass "B" open
- 2) Should read Rx Trip BKR "B" and Bypass "A" open
REFERENCES:
Drawing: Reactor Trip Signals,5610-T-L1, Sheet 2
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. -12/15/86 Page 6 NRC Exam Section 6
_ QUESTION SRO 6.08 (1.50)
- a. What are the 4 sources of water available for the Spent Fuel Pool? (1.0)
- b. What design feature of the spent fuel racks ensures criticality does not occur in the Spent Fuel Pool?
ANSWER:
- a. PG System (0.25)
CVCS Blender Boron Recovery Tanks Fire Main
- b. The center-to center distance ensure < .95 Keff for spent fuel,(even if unborated water is used)(+.5)
RESPONSE
6.08a Recommend changing answer to: (1) Demineralized Water (2) Borated Water from RWST (3) Borated Water from CVCS HUTS i 6.08b Recommend expanding answer key to include sufficient center-to-center j spacing or a combination of spacing and poison. l
REFERENCES:
l 608A, SD 041, pg 18-19 608B, Tech Spec Bases 3.17 l l
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. Page7 NRC Exam Section 7 QUESTION SRO 7.05 (2.25)
What are all the Plant Supervisor immediate actions if the announcement is made
" Fire in the Control Room, shift personnel report to assigned control room evacuation stations"?
ANSWER:
- 1) Ensure following personnel are dispatched with appropriate attachments: RCO 1 & 2; N. O.1; N. O. 2; 3rd licensed RCO; Nuclear Turbine Operator ( + 1.0)
- 2) Take these items from control room and distribute; radios; ICCS Valve keys; Electrical Penetration Room keys; plant a logs for both units; Plant Supervisor Logs; 1 pair of wire cutter ( + 1.0)
- 3) Proceed to TSC ( +.25)
RESPONSE
Change Enswer Key to not require listing each Operator. Accept such answers as dispatch operators to NTO Shelter to perform their preassigned attachments.
REFERENCES:
0-ONOP-103, page 5 - 6 l
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.. Pags8 NRC Exam Section 7.
QUESTION SRO 7.11 1.50 Provide the bases for the following precautions and limitations contained within OP-0205.1," Unit Shutdown":
- a. Minimize operation of the turbine at less than 40 MWe.
- b. Allow sufficient time for steam dump to condenser valves to respond before calling for additionaldemand.
- c. Alternately place each charging pump in service with a 60 gpm orifice on line.
Answer question as the basis for the " step" ANSWER:
- a. Prevent overheating the turbine ( +.5 ea)
- b. Prevent causing an inadvertent safety injection
- c. Ensure all charging pumps can be used to control pressurizer level RESPONSE: 7.11c Recommend changing answer to "To verify charging pump operability during plant shutdown so that malfunctioning pumps can be identified and repaired."
REFERENCES:
General Operating Procedure 103, page 5, step 5.24. i. I i I
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12/15/86
. Paga9 NRC Exam Section 8 QUESTION SRO 8.06 (1.00)
During Unit 3 operation in Mode 1, events occur that require application of Tech Spec 3.4.4.b (attached). Manipulations to bring Unit 3 to a Hot Shutdown condition are effected due to this Tech Spec ONLY. Which one of the following circumstances would require this course of action?
- a. A and C CCW pumps have been INOPERABLE for 18 hours when the B EDG is declared INOPERABLE estimated repair 2 hours.
- b. A and B CCW pumps are operating when A CCW pump trips on overcurrent lockout. After 24 hours of being INOPERABLE, maintenance approximates another 18 hours to restore the A CCW pump.
- c. CCW A pump and heat exchanger have been INOPERABLE for 10 hours.
CCW A pump has been out of service and tagged for 48 hours estimated repair 2 hours.
- d. CCW pump A has been tagged for 48 hours when CCW B pump develops a seal leak which must be repaired. The pump is tagged and it is given a repair time estimate of 48 hours.
ANSWER: - d
RESPONSE
l Recommend expanding answer key to include A or D because A would immediately _.. place the plant into Tech Spec 3.0.1 requiring the unit to be shutdown.
~
l
REFERENCES:
__ Tech Spec 3.0.1, page 3.0-1 l
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~" . 12/15/86 Paga10 NRC Exam Section 8 QUESTION SRO 8.09 .75 a) Where does the PS-N log requests for in-service testing that must be delayed?
(0.25) b) What actions / notifications are required by the PS-N if a test is "Unsat"? (0.5) ANSWER a) Equipment out ofservice log (+.75) b) - Describe malfunction in Remarks Section of the test sheet
- Notify Maintenance
RESPONSE
Based on the question, two answers may be appropriate depending on the interpretation ofinservice testing. The answer above is correct for post maintenance testing to call a piece of equipment inservice after maintenance. The second answer could include or be periodic inservie-testing for routine operability requirements.
- 1) OSP 200.l answera) provides a note in Remarks section of the daily attachment to allow tracking b) make determination for conintued operation and document
- 2) AP 0190.28 a&b) answer key correct
- 3) AP 0103.4 a) insures clearance order and test sheetr. are in E006 log, no documentation required at this time b) b in answer key correct -
REFERENCES:
OSP 200.1 page 10 (new procedure) AP 0190.28 page 2 (test answer) AP 103.4 page 21
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. 12/15/86 , _- Page11 NRC Exam Section 8
_ QUESTION SRO 8.10 .75 Fillin the blanks: -: Temporary Procedures (TP) must be approved by the for an amount of time not to exceed . While active, a TP is retained in the in the control room. ANSWER: Plant Manager one year; Temporary Procedure Log ( +.25 ea)
RESPONSE
Expand Answer Key to accept such answers as "to include PNSC". The Plant Manager will not approve TP until PNSC review.
REFERENCES:
AP 0109.6, Step 5.2.1. l s I
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-~ Page12 NRC Exam Section 8 QUESTION SRO 8.16 1.00 As stated in 10CFR50.54, under what conditions may actions be taken that depart from a license condition or a technical specification, and who, as a minimum, must approve such action?
ANSWER: In an emergency when the action-is needed to protect the health and safety of the public ( + .75), approved by at least a licensed SRO ( +.25).
RESPONSE
Expand Answer Key to accept PSN as a licensed SRO.
REFERENCES:
P i
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ENCLOSURE 4 ACCELERATED REQUALIFICATION PROGRAM EVALUATION REPORT-Facility: Turkey Point Units 3 & 4 Nuclear Plant Examiners: W. M. Dean C. A. Casto W. Hemming W. A. Douglas T. Rogers L. L. Lawyer
- Dates of Evaluation: April, September, December 1986 Areas Evaluated: X Written X Oral Simulator Examination Results:*
R0 SR0 Total Evaluation Pass / Fail Pass / Fail Pass / Fail (S, M or U) Written Examiation 10/2 14/12 24/14 M Operating Examination
' Oral 5/0 18/2 23/2 S Simulator N/A N/A N/A N/A Evaluation of f;cility written examination grading N/A *The numbers reflect the number of examinations passed / failed during the accelerated requalification program evaluations, not the number of individuals examined.
Overall Program Evaluation With an overall pass rate of 63.2% (24/38) it would appear that the accelerated requalification program has not been totally successful in correcting the deficiencies noted as a result of the requalification examinations conducted in February, 1986. However, upon closer examination of the individuals who passed / failed,- an important result becomes obvious. Only one out of seven training department examinations was passed. If these results are extracted from the overall results, then a 74.2% (23/31) pass rate for operators and technical staff (Reactor Engineers, STAS) exists. This indicated the accelerated requal-ification program has been effective in restoring the level of knowledge and
l J Enclosure 4- 2 operational ability of those directly responsible for the daily operation of the facility, especially in light of the significant deficiencies that existed initially. Several noticeable improvements observed during this past year were in the areas of knowledge /use of emergency and abnormal operating procedures and in the quality and accuracy of the facility training material, particularly system descriptions / lesson plans. The poor performance of training staff personnel on these examinations is a major concern. FP&L needs to examine the reasons for this problem, as well as any other meaningful trends or concerns that may arise based on an in-depth analysis of the overall results. Corrective actions either planned or already taken need to be identified. A response documenting any problem identified by FP&L and associated corrective actions shall be submitted to the NRC. The review of this report will be carried as an open item, IFI OL-65-01. c hubm ttsd: Fo ded: Appr ved: Q,, / ' - 1an
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