ML20148M476
| ML20148M476 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/29/1988 |
| From: | Brockman K, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20148M455 | List: |
| References | |
| 50-321-OL-88-01, 50-321-OL-88-1, NUDOCS 8804050378 | |
| Download: ML20148M476 (229) | |
Text
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ ENCLOSURE 1 EXAMINATION REPORT 321/0L-88-01 Facility Licensee: Georgia Power Company P. O. Box 4545 Atlanta, GA 30302 Facility Name: Edwin I. Hatch Nuclear Plant Facility Docket No.: 50-321 and 50-366 Written examinations and operating tests were administered at Edwin 1. Hatch Nuclear Plant near Baxl y, Georgia. Chief Examiner: dNetMe# 48 N.1446 Date Signed
- p. E. BrjR:kmpsG Chief Operator Licensing Section 2 Approved by:
YM . F. Munro, Chief 29 m u. s s Date Signed Operator Licensing Section 1 Surrraary: Examinations were administered on February 8-11, 1988. Written and operating examinations were administered to eleven Senior Reactor Operator (SR0) and three Reactor Operator (RO) candidates. All SR0s and all R0s passed the written examination. All SR0s and all R0s passed the operating examination. Based on the results described above, eleven of eleven SR0s and three of three R0s passed the overall examination. l Of the nine technical corrections for the examinations, four (44%) were due l to inaccurate / incomplete materials provided to the Comission for examination f preparation. The facility is encouraged to ensure the accuracy and completeness l of facility reference material, especially in light of the numerous changes begin made within the plant procedures. Section 8 was administered in an "open book" format to 11 SR0 candidates. Overall, this section of the exam was well received. Eleven of eleven SR0s l passed the section. However, the time required to complete this section was much longer than anticipated from the time validation conducted by the Regionci staff. The cendidates were not adversely penalized as a result of this problem since additional time was allotted by the NRC proctor who closely l monitored the pace of the section. A qualitative evaluation will be made of l this testing format for lessons learned and the desirability for use in future examinations. l 8804050378 880329 ADOCK0500g1 PDR
I REPORT DETAILS
- 1. Facili1y Employees Contacted:
*E. Morris Howard, Manager, Nuclear Training EP & Security
*Harvey Nix, Plant Manager
- Lewis Sumner, Manager of Operations
*Curtis Coggins, Training & EP Manager
- Charles T. Moore, General Manager, Quality Assurance
*0. M. Fraser, QA Site Manager
- Steve Grantham, Operator Training Manager
*S. J. Bethay, Acting NSAC Manager
- Dan F. Moore, Nuclear Training Coordinator
*S. M. Crosby, Operations Training Coordinator
*C. L. Tully, Senior Nuclear Engineer
*T. H. Hunt, Senior Simulator Instructor
- Attended Exit Meeting
- 2. Examiners:
*K. E. Brockman, Region 11 D. C. Payne, Region II G. T. Hopper, Region II J. M. McGhee, EG&G J. F. Hanek, EG&G T. L. Morgan, EG&G M. A. King, EG&G (P. Holmes-Ray, Senior Resident Inspector, attended Exit Meeting)
- 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 NRC Resolutions to coments made by the facility reviewers are listed below.
- a. SR0 Exam (1) Question 6.06a l
Coment accepted. This portion of the question has been deleted from the exam. Section and total points have been adjusted accordingly. (2) Question 6.11b ! Coment accepted. Due to the recently installed modification to I the plant, this port (on of the question has been deleted from the exam. The utility is encouraged +.o ensure that training material is complete and accurate. Section and total points have been adjusted accordingly.
(3) Question 7.06b Comment accepted. Scanning the top of the core with the Fuel Grapple lowered to just above the fuel bundles will be added to the key as an acceptable answer. However, both answers are required for full credit. The utility is encouraged to ensure that material is complete and accurate. The point value remains unchanged. (4) Question 8.03c Comment accepted. The answer key has been revised to accept One Hour Report per 40AC-REG-002-0S due to the recent amendment to Technical Specifications deleting this requirement. The utility is encouraged to ensure that training material is complete and accurate. The point value remains unchanged. ! (5) Question 8.05 Coment not accepted. It is acknowledged that a failure of the TIP valve is an isolation valve failure. However, the LC0 for Primary Containment Integrity still requires all automatic valves to be operable with specific exceptions. These exceptions apply to both isolation valve failure and Primary Containment Integrity. Since the question stated that the TIP valve was only closed (but not deactivated), Primary Containment Integrity is NOT satisfied. For Unit 1, due to definition 1.0.T.3, the answer key has been expanded to allow deactivation of the valve as an acceptable answer. The point value remains unchanged. (6) Question 8.07 Comment partially accepted. The answer key has been modified'to reflect a response regarding operability for each part. For Part 1, no additional change was made since the question did not ask for the conservative response to the situation. For Part 3, continued operation is not allowed since, TS 3.0.3 applies. The point value remains unchanged. (7) Question 8.13 Comment accepted. The question has been deleted from the examination. Section and total points have been adjusted accordingly. l (8) Section 8 - General Coments acknowledged. A qualitative evaluation of the Section 8 open book format is being made. These comments will be duly assessed and incorporated, as appropriate, in lessons learned. l l l
a i
- b. R0 Exam (1) Question 3.04c l
Comment accepted. This part of the question has been deleted. Section and total points have been adjusted accordingly. (2) Question 3.09d Comment accepted. This part of the question has been deleted. Section and total points have been adjusted accordingly.
- 4. Exit Meeting At the conclusion of the site visit, the examiners met with representatives-of the plant staff to discuss the examination.
There were no generic weaknesses noted during the operating examination. Simulator weaknesses encountered during the operating examinations are noted in Enclosure 4. 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. n i
[*) SQ4 U. S. NUCLEAR' REGULATORY COMMISSION
- 8' v I REACTOR OPERATOR LICENSE EXAMINATION FACILITY: HATCH 1&2 REACTOR TYPE: BWR-GE4 DATE ADMINSTERED: 88/02/08 EXAMINER: dqGBEE. J.
CANDIDATE INSTRUCTIONS TO C6HDIMIEi. [ 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 VALUE- TOTAL SCORE V ALUE CATEGORY As.sr "
26.50 2 5. la 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 2 0 // 26.25 e4T68- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS A5*. 75 afbf 20.75- 45 38 3. INSTRUMENTS AND CONTROLS 24 77 e 26.00 24.04- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
/09'. 5
-105.5- % Totals j
Final Grade All work done on this examination is my own. I have neither given nor received aid. Uandidate's Signature l l l l WSTERCO?Y
r 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. l
- 4. Print your name in the blank srovided 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.
e-
'18. When you complete your examination,-you shall:
- a. Assembletyour 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.
I i l l l l 4 l l l l i I i
- 1. PRIEGIELES OF NUCLEAR POWER PLANT OPEBAllQEu Page 2 THERMODYNAMICS. HEAT TRANSFER-AND FLUI.D FLOW QUESTION 1.01 (1.50)
The reactor is taken to CRITICALITY from a cold condition and an 80 second POSITIVE period-is attained:
- a. From-control room nuclear instrumentation, HOW can the operator tell when the heating range has been reached?
(Rod position and recirculation flow are held constant.) (0.5)
- b. In which ONE of the following intervals'was the heating range entered? (1,0)
(1) Interval 1 - reactor power increased by a factor of 8 in 143.3 seconds. ! (2) Interval 2 - reactor power increased by a factor of 3 in 99.0 seconds ; (3) Interval 3 - reactor power increased by a factor of 5 in 128.8 seconds, i (***** CATEGORY 1 CONTINUED ON NEX'l PAGE ** *** )
QUESTION 1.02 (1.00) Which ONE (1) of the following thermal limits protects the fuel from clad rupture due to PLASTIC STRAIN (deformation)?
- a. APLHGR
- b. LHGR
- c. MCPR
- d. MAPRAT J
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
i QUESTION 1.03 (1.00) Choose the ONE (1) phrase below which BEST completes the following sentence. Without core orificing, the coolant flow through a high power bundle will be less than the flow through a icw power bundle because:
- a. the channel quality increases.
- b. the two phase flow friction multiplier decreases.
- c. the fuel rods expand due to thermal effects.
- d. the bypass flow increases.
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *4***)
r QUESTION 1.04 (1.50) Saturated steam with 100% quality enters the main condenser at 4.5 psia and with a flow rate of 6E+6 lbm/hr. Condensate exits as a saturated liquid. Circulating water enters the condenser at 82 deg F and exits at 77 deg F.
- a. Choose the circulating water flow rate (in Ibm /hr) from the list below: (1.0)
- 1. 4.01E+8 lbm/hr
- 2. 3.90E+8 lbm/hr
- 3. 3.65E+8 lbm/hr
- 4. 3.03E+8 lbm/hr
- b. STATE whether the condenser vacuam will would INCREASE, DECREASE, or remain the same if the circulating water flow rate were DECREASED. (0.5) l l
i l l l l l i l l L (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l I i l
QUESTION 1.05 (1.00) Choose the word (s) in parenti sis -nich best complete the following statement; j Shaping control rods nre (DEEP, INTERMEDIATE, SHALLOW) rods that are used to change the power profile because they (ARE, ARE NOT) affected by shadowing. (Choose one answer in each parenthesis.) l \ l l l l [ (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****> 1
E d > QUESTION 1.06 (1.00) i MULTIPLE CHOICE (Select the ONE correct answer.) , The Doppler Coefficient of Reactivity correlates the change in fuel temperature to a reactivity insertion. Which statement is TRUE concerning Doppler Coefficient? z
- a. The coefficient becomes less negative with fuel burnup, and more '
negative with control rod withdrawal.
- b. The coefficient becomes more negative with fuel temperature increase and less negative with void fraction increase,
- c. The coefficient becomes lees negative with control rod withdrawal, and more negative with fuel temperature increase,
- d. The coefficient becomes more negative with void fraction increane and less negative with fuel temperature increase.
(4444* CATEGORY 1 CONTINUED ON NEXT PAGE 4****)
QUESTION 1.07 (1.00) The reactor is critical at 106 cps. Which ONE (1) of the following BEST describes the behavior of neutron power following a prompt insertion of negative reactivity?
- a. Neutron power drops immediately to "Beta" (delayed neutron fraction) times the neutron power prior to the prompt insertion of negative reactivity. ,
1 l
- b. Neutron power decreases linearly with time after the initial prompt drop.
- c. After the initial prompt drop, neutron power decreases on a constant negative period; the magnitude of the period I determined by the amount of negative reactivity inserted. I
- d. Because only delayed neutrons are left immediately after a negative reactivity insertion, neutron power decreases on an 80-second period regardless of the size of the nega-tive reactivity insertion.
i l t ( I i l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l
QUESTION 1.08 (1.00) Reactivity is defined as which of the following?
- a. The ratio of the number of neutrons at some point in this generation to the number of neutrons at the same point in the previous generation.
- b. The fractional change in neutron population per generation.
- c. The factor by which neutron population changes per genera-tion.
- d. The rate of change of reactor power in neutrons per second.
1 l l l i i [ l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) I
QUESTION 1.09 (1.50) State HOW EACH of the below listed conditions will effect control rod worth. (Limit the answer to INCREASE, DECREASE, or REMAINS THE SAME.)
- a. Increasing moderator temperature
- b. Increasing the percent voids
- c. Increasing the fuel temperature j
?
i (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
, -- , , - s --.w - , - -
, ,7 7_-,--- - , , - -g -------- p- - -- - - --.,-
-g-- - - - -- , -p- y m -~-
QUESTION 1.10 (3.00)
- a. Define the term Critical Power (CP). (1.0)
- b. State how Critical Power would change for each of the following (2.0) events (i.e., INCREASE, DECREASE, or NO CHANGE).
Assume that the reactor is at full power. Consider each event separately.
- 1. Loss of a feedwater heater string
- 2. Reactor pressure increase from 950 psig to 1040 psig
- 3. Reciro Flow Control system fails to maximum demand
- 4. Feedwater Control system fails to maximum demand
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
QUESTION 1.11 (2.00) Figure 1 contains charts of several key reactor parameters following a Feedwater Controller Failure to Maximum Demand. For the areas marked, give the cause of each parameter change as stated below,
- a. State WHY reactor power rises at Point A then immediately (0.5) decreases.
- b. State WHY feedwater flow drops sharply at Point B. (0.5)
- c. State WHY core flow drops at Point C. (0,5)
- d. EXPLAIN the slight increase in reactor water level at Point D. (0,5) l
(***** CATEGORY 1 CONTINUED ON NEXT PAGE 44***) ; i j I
QUESTION 1.12 (1.00) The total amount of reactivity that must be added to bring a reactor to a critical condition is known as the:
- a. ' Reactivity Defect b ., Excess Reactivity
- c. Subcritical Factor
- d. Shutdown Margin l
l t { l l l l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE **4**) l l
4 QUESTION 1.13 (2.00) With regard to PCIOMR, IDENTIFY EACH of the following statements as TRUE or FALSE.
- a. PCI failures are dependent on absoluto power, increase in power, duration of power increase, previous power history and fuel .
exposure,
- b. Regardless of the stress state, strain rate, temperature, or state of irradiation, aircalloy fuel tubes will be ductile. 1
- c. If power level is reduced prior to completing the 12 hour soak, pre-conditioning is resumed at either the new power level or the highest power level which has soaked for 12 hours before the power decrease, whichever is higher.
i
- d. PCIOMR limitations are not required for barrier fuel because redesigned pellets cannot expand faster than the clad.
i-r (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
s QUESTION 1.14 (2.00) For each of the following, indicate whether the available NFSH at the suction of the recirculation pump would INCREASE / DECREASE / REMAIN THE SAME:
- a. The Feedwater Flow is INCREASED
- b. The Recirculation Flow is INCREASED
- c. The Vessel Pressure is INCREASED from 200 psig to 800 psig 1
i i a l' 1' (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
QUESTION 1.15 (1.00) In a boiling water reactor, non-condensible gases produced in the reactor are carried through the turbine and into the condenser. Non-condensible gases can also leak into the condenser. Which ONE (1) of the following correctly describes the effect of an increase in the amount of non-condensible gases in a steam turbine condenser?
- a. Condenser pressure decreases (vacuum increases)
- b. Circulating water outlet temperature decreases
- c. Steam cycle efficiency decreases
- d. Condensate depression increases
(***** CATEGORY 1 CONTINUED ON NEXT FAGE *****)
QUESTION 1.16 (2.00) Indicate whether each of the following statements are TRUE or FALSE:
- a. The delayed neutron fraction (beta) is defined as the ratio of thermal neutrons absorbed in the fuel to all thermal neutrons which are absorbed.
- b. The effective delayed neutron fraction of the core is greater at BOL than at EOL.
- c. When calculating reactor period, the delayed neutron term may be considered insignificant if reactivity added is less than beta,
- d. Delayed neutrons are fast neutrons, but are usually born at lower energies than prompt neutrons, r
1 l i l l j (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) 1
QUESTION 1.17 (2.00) For each of the indicated changes in core parameters (a. through d. below) INDICATE whether the Void Coefficient of Reactivity will:
- 1. Become more negative.
- 2. Become less negative.
- 3. Not Change.
- a. Increase in void fraction. (0.5)
- b. Buildup of fission product poisons. (0.5)
- c. Increase in fuel temperature. (0.5)
- d. Decrease in control rod density. (0,5) i
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
QUESTION 1.18 (1.00) Concerning control rod worth during a reactor startup with 100% peak Xenon versus a startup with Xenon free conditions, WHICH statement (ONE) below is CORRECT 7
- a. Peripheral control rod worth will be LOWER during the 100% peak Xenon startup than during the Xenon free startup.
- b. Central control rod worth will be HIGHER during the 100% peak Xenon startup than during the Xenon free startup,
- c. Peripheral control rod worth will be HIGHER during the 100% peak c- Xenon startup than during the Xenon free startup.
- d. Both central and peripheral control rod worths WILL BE THE SAME regardless of core Xenon concentration.
l (***** E6C CJ CATEGORY I *****J l ., . _ . .
2.__ELAHI_DESIGH_lHGLUDING SAEEII_AHD_EMEBGERG1 Pcgs 20 El&IEME QUESTION 2.01 (1.50) With regard to the Unit 1 RHR System, MATCH each of the items in COLUMN A with its ONE (1) associated setpoint OR interlock from COLUMN B. COLUMN A COLUMN B
- a. RHR HX bypass valve 1. 145 psi
- b. LPCI injection valves open 2. 2/3 core coverage permissive
- c. Recirculation discharge 3. 425 psi close valves 4. 105 psig
- d. Containment spray valves 5. Open for 3 minutes following LPCI
- e. ADS permissive initiation signal.
- f. Shutdown cooling valves close 6. 325 psig
- 7. 422 psig i
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
- n. x.
QUESTION 2.02 (3.00) Regarding the Standby Gas Treatment System:
- a. State FOUR conditions which will automatically initiate the SGTS trains, isolate both unit's refuel floor ventilation systems, and isolate Unit One's reactor building ventilation. INCLUDE ANY SETPOINTS. , (2.0)
- b. EXPLAIN WHY it may be necessary to provide continued cooldown flow after the train is no longer required (assume the train has beer operating for four hours). (1.0)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
QUESTION 2.03 (2.50)
- a. List FOUR (4) automatic trips /actuations which occur due to a High-High Main Steam Line Radiation signal ( 3 X Normal Background). (2.0)
- b. State the MSL Radiation Monitor High-High setpoint when Hydrogen injection is in progress (assuming it has been changed as required). (0.5) l l
l l l l r l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
QUESTION 2.04 (2.00) Determine if EACH of the following statements concerning the Standby Liquid Control System is TRUE or FALSE,
- a. In the event a remote (outside control room) reactor shutdown is required SBLC injection can be actuated by the local pump START switch,
- b. The pumps may be operated simultaneously if necessary to shutdown the reactor in an ATWS,
- c. When the control room handewitch is placed to "PUMP A RUN",
the "A" pump starts and all squib valves fire,
- d. Nitrogen-charged accumulators assure adequate suction pressure for the pumps.
l l 1 (***** CATEGORY 2 CONTINUED ON NEXT PAGE ***a4)
QUESTION 2.05 (1.50) LIST and EXPLAIN HOW the Control Rod Hydraulic system design features, components, and/or interlocks provide the following functions;
- a. Constant control rod speed / system flow during normal rod movement. (ONE COMPONENT / INTERLOCK REQUIRED) (0.5)
- b. Prevent pump runout while a scram signal is present.
(TWO COMPONENTS / INTERLOCKS REQUIRED) (1,0) l J ] (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
QUESTION 2.06 (2.00) RCIC is being used to control reactor level from the Remote Shutdown Panel ' when level is inadvertently allowed to increase to + 58".
- a. EXPLAIN HOW the RCIC system will respond to this level and WHY it responds as it does. (0.75)
- b. List FIVE (5) systems in addition to RCIC which can be operated (all or in part) from the Remote Shutdown Panel. (1.25)
I l l i l l l l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
QUESTION 2,07 (2.50) Regarding the Unit 2 SRVs and associated Low Low Set (LLS) Logic System; (SETPOINTS NOT REQUIRED). l , a. There are three lights associated with each SRV (RED, GREEN, l and AMBER). EXPLAIN what EACH of the three colored lights indicates when energized. (1.5)
- b. HOW and WHERE is the amber light reset after the conditions which caused it to be illuminated have cleared? (0.5)
I c. List the TWO (2) conditions (signals) needed to arm the LLS logic. (0,5) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
QUESTION 2.08 (2.00) State whether EACH of the following statements concerning the MSIV Leakage Control System is TRUE or FALSE.
- a. The system is electrically interlocked to prevent initiation until 10 minutes after a LOCA event has occurred,
- b. If the inboard system is actuated by the operator with the reactor at normal operating pressure and one MSIV failed open, the reactor will be vented to the Torus and will blowdown via the open steam line,
- c. Dilution flow is intended to lower the temperature of the leakage steam to prevent exceeding the operating limits vf the blowers.
- d. The MSIV Leakage Control System is interlocked to prevent initiation if all MSIV's are not full closed.
i i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) 1
QUESTION 2.09 (1.00) Which ONE (1) of the following statements BEST describes the purpose of the Exhaust Hood Spray?
- a. Provide additional scrubbing for steam to aid in removal of non-condensibles during startup or low load operations.
- b. Provide turbine blade cleaning during startup to remove accumulated silica deposits.
- c. Provide a spray path to aid the condenser in maintaining vacuum during high circulating water temperature conditions by condensing ,
steam as it leaves the last stage buckets. , l
- d. Provide cooling for the last stage buckets during low steam flow conditions. l i
j j e l 1 i (***** CATEGORY 2 CONTINUED ON NEXT FAGE *****) ,
\
QUESTION 2.10 (2.00)
- a. List THREE (3) of the four loads on Unit 2 RBCCW (Reactor Building Component Cooling Water) system which are not included in the Unit 1 i
system. (1.5)
- b. After a temporary loss of essential 600 VAC bus had been corrected, an operator attempted to restart the affected RBCCW pump by taking !
the pump breaker control switch to OFF and then to ON, but the pump j did not start. What additional action (s) must be performed in order to start the pump? (0 b) 1 l i 1 l i l I i (***** CATEGORY 2 CONTINUED 014 NEXT PAGE *****)
QUESTION 2.11 (2.75)
- a. List THREE (3) of the four parameters which cause isolation of the Containment Atmosphere Monitor Hydrogen /0xygen sample lines. (1.5)
- b. List FIVE (5) containment parameters monitored by the CAMS in addition to monitoring hydrogen and oxygen concentrations in the Drywell. (1.25)
(***** CATEGORY 2 CONTINUED ON NE'AT PAGE * * * * * )
l QUESTION 2.12 (2.00)
- a. If the Standby PSW pump is determined to be inoperable with the system in a normal configuration, STATE the action (s) required to be taken to ensure adequate cooling to components served.
- b. LIST the conditions / signals which will automatically close the PSW turbine building isolation valves.
(Setpoints not required.) (***** CATEGORY 2 CO!1TINUED ON NEXT PAGE ***4*)
QUESTION 2.13 .(i.50) While operating RHR Pump 2A (Loop A) in Shutdown Cooling Mode with pump 2C tagged out, a LPCI initiation signal is received. All actions / trips occur in accordance with system design. LIST all actions required to be performed by the operator to use the A Loop of RHR in the LPCI modo. (Valve numbers are not required, but valve descriptions should be as specific as possible.) { (***** END OF CATEGORY 2 *****) l _u
- 3. INSTRUMENTS AND CONTROLS Pcgo 33 QUESTION 3.01 (3.00)
- a. HOW is total core flow indication obtrined?
NOTE: BE SPECIFIC as to: ^
- 1. WHAT is measured
- 2. WHERE measured (sensing point (s)
- 3. HOW signal is converted and/or conditioned prior to control room indication (1.5)
- 6. Briefly EXPLAIN how the total core flow instrumentation compensates for a trip of one Recirculation pump (the other is still operating)? (1.0)
- c. Briefly EXPLAIN why a different method is used when one loop is operating and one loop is shutdown. (0.5) i I
l l l l l (***** CATEGORY 3 CONTINUED 08 NEXT PAGE *****) l , r l l
QUESTION 3.02 (3.00)
- a. Stato THREE conditions / signals which will cause an AFRM INOP trip. (1.5)
- b. List THREE actuations or signals which MAY result when an APRM It10P trip signal is generated. (1.5) d i
) 1 1 I J e i i 4 a t l
- (***** CATEGORY 3 CONTINUED Oli NEXT PAGE * * * * * )
l l l
d QUEST 1sN 3.03 (2.00) Concerning the Neutron Honitoring System (NMS), answer EACH of the following TRUE or FALSE.
- a. Removing the Unit 2 "shorting links" will place all NHS scram signals in a coincidence mode,
- b. All IRH trips are bypassed when the mode switch is in vun.
- c. The Unit 2 APRM flow biased scram is "clamped" at 118% regardless of recirculation flow,
- d. The APRM flow biase( scram is conditioned through a six second time constant. ,
i . I h I i l .) h L l i ; ! i i i , I l ? 4 , I l
*****)
i (***** CATEGORY 3 CONTINUED ON NEXT PAGE
/, 0 QUESTION 3.04 (4r60)
What effect would EACH of the following conditions have (INCREASE, DECREASE, or NO CHANGE) on indicated Reactor Vessel level indication?
- a. Seat leakage on a level transmitter equalizing valve,
- b. Increase in Drywell temperature.
eRecirculation loopw>peratrion en wide-rango inst 2rumentatri+1w c/effe i l l l l l l I I l I l l I i l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) ~
QUESTION 3.05 (1.00! Given: Unit 2 in control of D/G "B" D/G "B" Mode Switch in TEST (Surveillance being performed) Electrical distribution NORMAL (Full Power Lineup) l D/G "B" is at rated speed and voltage, but not synchronized, when all i off-site power to 4160 volt Bus 2F is lost. Which ONE (1) of the i following accurately describes the system operation? ;
- a. Bus 2F can be powered by D/G "B" when the operator takes the Output Breaker Switch to CLOSE and has the SYNC SCOPE activated.
- b. Bus 2F will be powered by D/G "B" automatically, after 12 seconds; appropriate loads will be picked up sequentially.
- c. Bus 2F can not be powered by D/G "B" while it is in the TEST mode, given these conditions.
- d. Bus 2F can be powered by D/G "B" when the operator resets the Lockout Relay, activates the SYNC SCOPE, and takes the Output Breaker to CLOSE.
l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *4***) { _ _
QUESTION 3.06 (2.00)
- a. In addition to having power available, what CRITERIA (interlocks) must be satisfied in order to start a Condensate Booster Pump using the control switch? (INCLUDE SETPOINTS) (1.5)
- b. List the SPECIFIC condition (s) which will cause a standby Condensate Booster Pump to automatically start if power is available and the control switch is in auto. (0,5)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
QUESTION 3.07 (2.50) List FIVE (5) of the seven Reactor Feed Pump Trip signals (INCLUDE ALL SETPOINTS for full credit). l t l l l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) l
QUESTION 3.08 (2.25) List THREE (3) control systems / components outside of the feedwater system which receive inputs from the Reactor Water Level Control system. For EACH signal identify BOTH the INPUTS and what FUNCTION that input is used for (IDENTIFY ANY INTERLOCKS, CALCULATIONS, OR TRIPS). l l l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) I -- ..
/o f QUESTION 3.09 (& OO)
LIST the pressure setpoint (including system) where each of the following UNIT 2 Plant Air System isolations or initiations occur,
- a. Nonessential Air Header ISOLATES.
- b. Service Air Header ISOLATES,
- c. Nitrogen backup valves OPEN to Noninterruptable Essential Instrument Air.
d, -Nitrogen-backup-valves-460LATEMeyweti-Pneumatrio-Systenw--e.- - cd/e/c I i l l l l
****^)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE 1
QUESTION 3.10 (1.50) STATE the THREE (3) conditions required for the Main Condenser Low Vacuum Isolation signal to be bypassed. [ Conditions may include operator action.] l l l l (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
QUESTION 3.11 (1.00) Choose ti parameter from the following list used by the Rod Sequence Control S. tem to determine the automatic bypass power level,
- a. APRM indicated power
- b. Process Computer calculated thermal power
- c. First Stage Turbine Pressure
- d. Steam Flow and Feed Flow l
l (***** CATEGORY 3 CONTINUED ON NEXT PAGE +4***) l l l
QUESTION 3.12 (1.00)
- a. Whose permission is required before manually bypassing the Rod Worth Minimizer? (0.5)
- b. List the automatic bypasses of the RWM. (0.5)
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
QUESTION 3.13 (1.50) Briefly DESCRIBE the consequences of closing the SUT supply breaker to a 4160V bus while the UAT breaker is closed (the control switch is released while both breakers are closed) and the "Cutout Switch" (panel 651) is not in thu Cutout position. (Include final position of both breakers and the status of the electrical bus.) i (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
\
QUESTION 3.14 (2.50) List FIVE (5) conditions which must be met for the Diesel Generator output breaker to close when the mode switch is in "AUTO". (***** END OF CATEGORY 3 *****) j
, . , ,. - ., . . . . ~ .-. . . _ . . ., . .
.. t;': s
- 4. PROCEDUPSS rNORLIAL, ABNORMAb, EMERGENCY Page<47
' AND RADIOLOG_ICAL_CONTRQL: -
e r
' QUESTION. 4.01 {2.00)
List the immediate operator actions required by 34AB-OPS-055-2S, "Control Room Evacuation-Unit Shut Down", to be performed prior to leaving the control room. (REACTOR IS IN OPERATIONAL CONDITION 3 AND NO SCRAM SIGNAL IS PRESENT) 1 l 1 l
- )
J q 2 d 4 J 1 i (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) ' i . _ , _ _ . . _ , .. _ ._ _ _ . .,~.._- ,___.. ....,,__ . _ ,_..~_-,-___-- ...- _ .-_.. . ...-. - , _ _ _ ____ .-- _ - _ _ . _ ..
.. .. - .- . , ~. .-. . . _ , .
QUESTION 4.02 (2.00) I STATE which Emergency Classification is appropriate for EACH of the ) following definitions. I
- a. Events are in progress or have occurred which involve actual or potential substantial degradation of the level of safety of-the plant.
- b. Events are in progress or have occurred which indicate a l potential degradation of the level of safety of the plant.
- c. Events are in progress or have occurred. which involve actual or imminent substantial core degradation'or melting '
with the potential for loss of containment integrity.
- d. Events are in progress or have occurred which involve ~
an actual or likely major failure of plant functions needed for protection of the public. 1 l 4 3 (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
QUESTION 4.03 (2.00) 34AB-OPS-038-2, "Control of Sustained Combustion in the Offgas Systems," lists several system parameters which would be expected to change significantly if ignition occurs. List FOUR (4) parameters indicated by system instrumentation which could be expected to change EXCLUDING system and component temperatures. (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) 1 l
QUESTION 4.04 (2,50) For each of the following sets of plant conditions, IDENTIFY the EOP flowpath (BY NUMBER) specifically designed to address those conditions.
- a. Reactor transients or failure of vital equipment while in hot standby or startup,
- b. High radiation, loss of vital power, failure of vital equipment, or stuck open relief valve.
- c. High radiation, loss of coolant, and loss of primary containment integrity,
- d. Reactor transients or failure of vital equipment while in the RUN mode.
- e. Failure of reactivity control systems.
1 i i (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
QUESTION 4.05 (1.50) The E0P contingency procedure for Alternate Pressure Control requires the operator to determine if adequate core cooling is assured. State THREE (3) criteria which may be used to verify adequate core cooling. (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
QUESTION 4.06 (1.00) 34SO-C11-005-2S, "Control Rod Drive Hydraulic System," contains a precaution stating the CRD Suction from Condensate Control Valve Bypass Valve, 2N1-F182, must remain LOCKED CLOSED while the Condensate system is in service. State the BASIS for this caution. l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
QUESTION 4.07 (1.00) Durind operation with the Condensate and Feedwater Cleanup Recire. Control valve (2NI-F165) open, limits are placed on how far the control valve may be opened with a Condensate pump or Condensate Booster pump EXPLAIN WHY it is necessary to limit the control valve position. I running. l l t i l l l l (***** CATEGORY 4 cot 1TINUED Oli NEXT PAGE *****) I
QUESTION 4.08 (1.50) RWCU system operating procedure, 32SO-G31-003-28, contains a caution statement that states: During operation of the.RWCU system CU & Demin Bypass V1v 2G31-F044 must not be fully closed for extended periods of time and left unattended.
- a. This valve 1is opened in anticipation of'WHAT plant or system transient? (0,5)
- b. State the function provided by this valve lineup. (1.0) i i
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) l
QUESTION 4.09 (1.00) State TWO (2) reasons Mechanical Vacuum Pump operation is not permitted above 5% Thermal Power, f I l l l l l (***** CATEGORY 4 ColiTIt1UED Oli t1 EXT PAGE *****) l l t
QUESTION 4.10 (1.00) Explain WHY it is not necessary to trip the Main Generator following a loss of both the Main and Emergency Seal Oil Pumps. J l l l (***** CATEGORY 4 CO!1TINUED ON NEXT PAGE *****) l i i
QUESTION 4.11 (1.50) List THREE (3) radiological conditions which require a Normal Operations Job Specific Radiation Work Permit be issued for work in that area. (SETPOINTS NOT REQUIRED) I l l l I l i I i (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) i l
QUESTION 4.12 (1.00) Determine if the following statements are TRUE or FALSE.
- a. After obtaining a new P1 printout, the operator can use the old P1 for scratch paper or discard it.
- b. Chart recorders are required to be checked within one hour of shift change and marked with date, time, and operator initials.
(***** CATEGORY 4 CONTINUED ON NEXT PAGS *****) I 1
QUESTION 4.13 (2.00) List TWO (2) conditions where removal of Danger Tags may be authorized from an Equipment Clearance Sheet (ECS) when all subclearances on that ECS have not been released. l l i l l l l l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
\
QUESTION 4.14 (3.50)
- a. You are verifying a valve line-up, STATE how you confirm position of EACH of the following:
- 1. Closed valve. (0.5)
- 2. Open valve. (0.5)
- 3. Motor-operated valve. (0.5)
- 4. Locked throttle valve. (0,5)
- b. List the actions required if a LOCKED VALVE is found in the wrong position while performing a valve lineup. (1.5)
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
QUESTION 4.16 (2.00) I 42FH-ENG-010-2, "Control Rod Movement", provides numerous STANDARD PRACTICES which apply when rods are being moved for the purpose of changing power level. Answer EACH of the following in accordance with 42FH-ENG-010-2.
- a. With LHGR > 8 Kw/ft, STATE the period of 1 you must wait between successive notch withdrawals of tl. ame rod. (0.5)
- b. With HIGH POWER and HIGH FLOW conditions, STATE which type of rod - SHALLOW, DEEP, CENTRAL, or EDGE - should NOT be withdrawn nor inserted. (1.0)
- c. With power < 30%, STATE the verification which is utilized when latching the first rod in any group. (0.5)
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
l QUESTION 4.16 (0.50) Determine if the following statement concerning a Cable Spreading Room fire and Procedure 34AB-FPX-053-2S if in TRUE or FALSE. For an exposure fire (in the cable spreading room) involving com-bustables WITHOUT electrical insulation involvement, the Control Room Operator in the affected unit will initiate a rapid load reduction, will trip the turbine generator, will manually SCRAM the reactor, and will place the Mode Switch in SHUTDOWN. i i l l (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION ****+4****)
i i , r1 b e b E
- S
"% O. .&
8 E
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--n.';.- ..
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l EQUATION SHEET f = ma v = s/t w = ms s = v,c + lsat 2 Cycle efficiency = Net g tJork (out) - E = aC - a = (v, - v )/t ; A=Aeg -At 2 KE = lysv v A = AN f=v o+ at PE = agh w = e/t A = In 2/tq = 0.693/tg i W ' v4P' eq (eff) = (t,,)(ts) AE = 931Am . ( 4 ) Q=[ncAT
, P ,
I = g ,4X o
', Q " UAAT I = I,e -UX , ,
Pur = W'g In I = I,lo-x/ M p=p 10 ER(t). Tyt = 1,3/g y=p o ,t/T HVL e 0.693/u
'SUR = 26.Co/t "
T = 1.44 DT SCR = S/(1 - K,gg)
/A ' o SUR = 26 CR g , x = S/(1 - K,gg,)
~
T = '(t*/o ) + [(i 'o)/A,ggo ] 1 off 1 * ~ eff 2
~~
y = t*/ (, . p M = 1/(1 - K,gg) = CR g/CR 0 I"I ~ 8)! eff' M = (1 - K,gg)0 II ~ Eeff)1
#*I eff II aff
- OEefflEeff SDM = (1 - K,gg)/K,g,
~
p= (L*/TKygg-] + [I/(1 + A,ggT )] t* = 1 x 10 seconds
? = I4V/(3 x 1010) A,ggA= 0.1 seconds I = No Idgg=1d22 WATER PARAMETERS Id g =Id22 ;
1 gal. = 8.345 lbm R/hr = (0.5 CE)/d 4,,t,,,) 2 , 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) . I 1 fc3 = / 48 gal. MISCEI.L\NEOUS CONVERSIONS , i Density = 62.4 lbm/ft 3 1 Curie = 3.7 x 1010dps i twnsity = 1 ge/c:n 3 1 kg = 2.21 1ha i Heat of valori:ation = 970 teu/lbm I hp = 2.54 x 103 BIV/hr ,' Hest of fusien = 144 Btu /lbr., 1 N = 3.41 x 100Btu /hr 1 Atm = 14,7 psi = 29.9 in. Ig. 1 Btu = 778 ft-lbf l 1 ft. H 2O = 0.433* Ibf/in 2 g inch = 2.54 c:a T = 9/5 C + 32
*C = 5/9 (Or - 32) '
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, Pcgo 63 THERMODYN AtjICS , HE_4T TRANSFER 6t{D__ FLUID FLOW MASTERC0?Y ANSWER 1.01 (1.50)
E4es d /fe fo/Lu;a- /gerpesses h eegnee / (c.r)
- a. Operatorcannoticetlatperiodhasbecomelonger>[n951_o_- og and that power change on IRMs/SRMs is leveling off (turning around due to power overshoot). (4r2%
- b. (2) (1.0) (From P = Poe(t/T) --> T = t/in (P/Po), in Interval 2 the period has lengthened from 80 seconds. The other intervals have 80 second periods)
REFERENCE EIH, GE Reactor Theory, LO 7.4.3, P. 7-10 CPS Introduction to Nuclear Reactor Operations, LO 4.1.1.2 INRO PP. 4-17 & 4-18 (3.8/3.6) 292008K112 292008K113 ..(KA's) ANSWER 1.02 (1.00)
- b. LHGR (1,0)
REFERENCE EIH, GE Heat Transfer & Fluid Flow, LO 9.3.3, P. 9-15 CPS Nuclear Power Plant Thermal Sciences, LO 12.1.1.1 NPPTS P. 12-2 i (3.0) l , I ' l I l 293009K108 ..(KA's) l (***** CATEGORY 1 cot 1TINUED ON NEXT PAGE * * * * * ) i MASTElCOPY
L___EBit[QlELESJE_HUCLEAR POWER _ELANT OPEB6IlQth. Pego 64 IHEBMQDINAMICS. HEAI_IBAHEEER_AND FLUID _ELQE ANSWER 1.03 (1.00)
- a. (1,0)
REFERENCE EIH, GE Heat Transfer & Fluid Flow, LO 8.9.2. P. 8-46 BSEP , HEAT TRANSFER, CH. 9 Page 9-51 Lesson Objective : Third from top of page 9-1A (no number assigned) (2.9) 293008K131 ..(KA's) ANSWER 1.04 (1.50)
- a. 1 (4.01E+8) [1.C)
- b. Decrease (absolute pressure would increase) [0.5)
REFERENCE EIH, GE Heat Transfer & Fluid Flow, LO 7.5.2, P. 7-40 (2.8/2.5) . 293004K113 293003K123 ..(KA's) ANSWER 1.05 (1.00) Shallow (0,5) ARE (0,5) i I (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) I
1.- EBl!LQIELES OF NUCLEAR POWER PLANT OEERATION t Pago 65 THERMODYNAMICS, HEAT TRANSEEB_&ED FLUID FLOW REFERENCE EIH, GE Reactor Theory. LO.5.3.3, P. 5-25 Monticello: Reactor Theory, Chapter 5 (2.8/2.4) 292005K111 292005K110 ..(KA's) s 4 ANSWER: 1.06 (1.00) j d [1.0] REFERENCE EIH, GE Reactor Theory, LO 4.6.3, PP-. 4-38 thru 4-42 WNP-2 REACTOR THEORY TEXT page IV-25 i (2.9) t 292004K105 ..(KA's) ANSWER 1.07 (1.00) , o [1.0] , REFERENCE EIH, GE Reactor Theory, LO 5.5.6 & 5.5.9, PP. 5-42, 43 NMP-2 Operations Technology, Module 1 Part . (3.3/3.7) 292003K106 292003K107 ..(KA's) l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
-11 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pcgo 66 IHEBBQDXHAMICS. HEAT TRANSFER AND FLUID FLOW
- ANSWER 1.08- . ( 1. 0 0 ) '
. b , l 4 ' REFERENCE EIH, GE Reactor Theory, LO 1.6.1, P. 1-38 DPC, Fundamentals of Nuclear Reactor Engineering, p. 96 GGNS: OP-NP-511 (3.2) . l ? 292002K111 ..(KA's) ANSWER 1.09 (1.50)
- a. Increase .
- b. Decreases ;
4 . c. Remains the Same l (3 @ 0.5 ea) i ;
- REFERENCE ,
I i .EIH, GE Reactor Theory, LO 5.2.5, PP. 5-12a, 5-13a, and 5-14a. , (2.5) 292005K109 ..(KA's) i i 'l
- I a I
?
(***** CATEGORY 1 CONTINUED ON NEXT PAGE 4****-) < i a I
- 1. PRINCIPLES OF NUCLEAR POWER ELANT OEEBAI1QEt Pcco 67 THERMODYNAMIC _Qu_ HEAT TRANSFER AND FLU _ID FLOW ANSWER 1.10 (3.00)
- a. Critical Power is the bundle power needed to produce the critical quality (or the bundle power needed to cause OTB to occur somewhere in the bundle. (1.0]
- b. 1. (inlet subcooling ^) CP increases [0.5]
- 2. (pressure ^) CP decreases (0.5]
- 3. (core flow ^) CP increases [0.5]
- 4. (inlet subcooling ^) CP increases (0.5]
REFERENCE i EIH, GE Reactor Theory, LO 9.5.1 & 9.5.7, PP. 9-26, 9-36 thru 9-39 SSES SCO23 G-3 Specific Objectives 3.3, 3.4, 3.6 ( 3. 3/2. 9 '2. 8/2. 7 ) 1 293009K124 293009K123 293009K122 293009K117 ..(KA's) l l ANSWER 1.11 (2.00)
- a. Rx power increases due to increased subcooling (+p) then (0.5) decreases sharply due to scram from turbine trip (hi level),
- b. FW flow decreases due to RFPTs trip on high RPV water level (0.5)
- c. Core flow decreases due to RPT breaker actuation. (0.5)
- d. Momentary increase from reformation of voids as pressure decreases and feed pumps coastdown. (0,5)
REFERENCE EIH Nuclear Training, 10.4-36 and Figure 10.4(33) i (3.8/3.5/3.8/3.7) 259002K302 259002K301 259002K104 259002K103 ..(KA's) l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) i
.I 1'. PRIEQiELES_OF NUCLEAR POWER PLANT OPERATION. -Pcas-68 ,
TliERMODYNAMICS. HEAT TRANSFER AND FLUID FLQH
' ANSWER 1.12- -(1.00) d [1.0)
REFERENCE EIH, GE Reactor Theory, LO 1.5.1, P. 1-36 EIH Nuclear Training, 10.103-22 (3.2) 292002K110 ..(KA's) i ANSWER 1.13 (2.00)
- a. True ,
- b. False '
- c. True
- d. False (4 0 0.5 each)
, REFERENCE {
- EIH Nuclear Training, Chap 10.2, PP. 18-21 GE BWR Academic Series. HT/FF LO 7.2, 7.5, and 8.2, PP. 9-49 through 9-52 (2.8/2.9) l 293009K132 293009K136 ..(KA's) l' l
l l l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l
a
- 1. PRINCIPLESLOF NUCLEAB POWER' PLANT OPERATION. Pago~-69
.THEBH9DXMAMICS, HEAT TRANSFEB_AND FLUID _ FLOW ' ANSWER 1.14 . (2.00)
- a. INCREASE'(More Subcooling at the-pump suction)
- b. DECREASE (Reduced pressure at the eye of the pump results in being closer to saturation pressure)
- c. DECREASE (Further to saturation temperature and increased density causing less static head) 0.66 (3 @ .OrCrnTach)
REFERENCE
~EIH, Heat Transfer & Fluid Flow, LO 6.10.10, P.6-81 (2.7/2.5) 293006K108 293006K110 ..(KA's)
ANSWER 1.15 (1.00) c (1.0) i REFERENCE l EIH, Heat Transfer & Fluid Flow, LO 5.5.2, P. 5-53 (2.7) 293007K107 ..(KA's) i i
- (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
- 1. PRINCIPLES OF NUCLEAR EQWER PLANT OPERATION.. Pcco 70 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLQR ANSWER 1.16 (2.00)
- a. FALSE [0.5]
- b. TRUE [0.5]
c.. FALSE -[0.5] d .- TRUE [0.5] REFERENCE GE_ Reactor Theory, LO 3.4.4, 3.4.5, 3.5.6, and 3.4.1, PP. 3-29 thru 36 (2.5/3.7/2.4/3.0) 292001K102 292003K103 292003K106 292003K104 ..(KA's) ANSWER 1.17 (2.00)
- a. 1 (more negative) [0.5]
- b. 1 (more negative) [0.5)
- c. 1 (more negative) [0,5]
- d. 2 (less negative) [0.5]
REFERENCE GE Reactor Theory, LO 4.4.3, pp. 4-19 through 4-23 (2.5/2.2) 292004K112 292004K111 ..(KA's) ANSWER 1.18 (1.00) o [1.0] (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) i
$ 1. - PRINCIPLES' OF HEGLEAR POWEB_EkMT . OPERATION, Peas 71 fliEBM0 DYNAMICS, HEAT TRANSEER_MD_ELULD..EkQE 5!
{;-REFERENCE-4 EIH, GE Reactor Theory, LO 5.2.4, P. 5-15 and 6-12 (3.1) 292006K114 ..(KA's) t i L f 4
.i , i
) : ! i 2 i i I !
- t 1
i l j : I 1 , i i l l ! l i , r t I t I i l t
)
l l r I (***** END OF CATEGORY 1 *****) l t i ! l I
- 2. PLANT DESIGN _IUCLUDING SAEEII_AND EMERGENCY Pcgm 72 EXEIElie ANSWER 2.01 (1.50)
- a. 5
- b. 7
- o. 6
- d. 2
- e. 4
- f. 1 (6 @ 0.25 each)
REFERENCE EIH LT-IH-00701-00, LO 11,12,&l4, TABLE 00701-2 PP. 48, 58-62 Nuclear Training, Main Steam, 5.1-5 Nuclear Training, Recirculation System, 4.1-38 (3.3/3.9/4.0/3.7) 205000K402 203000K411 203000K410 203000K402 ..(KA's) ANSWER 2.02 (3.00)
- a. Any 4 of the following @ 0.5 each;
- 1. Unit One Reactor building ventilation exhaust high radiation [0.25), 20 mR/hr [0.25)
- 2. Unit One Refueling Floor high radiation (0.25] 20 mR/hr [0.25)
- 3. Unit Two Refueling Floor high radiation (0.25) 20 mR/hr [0.25]
- 4. Unit One Drywell high pressure [0.25), 1.92 psig [0.25)
- 5. Reactor vessel low water level [0.25), -47 inches (0.25)
- b. To remove heat generated by radioactive particles contained in the idle train., [1.0)
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcgs 73 SYSTEMS REFERENCE EIH Nuclear Training, 3.3-6 and 3.3-9.
LP 13.2, LO 2, P. 4 LP 30.1, LO 3, P. 4 (3.7/2.6) 261000K402 261000K401 ..(KA's) ANSWER 2.03 (2.50)
- a. Any 4 of the following at 0.5 each;
- 1. Reactor Scram
- 2. Group I isolation
- 3. Swap Main Control Room Ventilation to pressurization mode.
- 4. Mechanical vacuam pump trip / isolates
- 5. Gland seal exhauster trip / isolates
- b. 6 X Normal Background [0.5)
REFERENCE EIH, LP 14.1, LO2,P.6 S0-OPS-01-1987 (3.3/3.2/3.2) 272000K501 272000K402 288000K105 ..(KA's) ANSWER 2.04 (2.00)
- a. FALSE [0.5)
- b. FALSE [0.5)
- c. TRUE [0,5) ,
- d. FALSE [0.5) l l
l l i t f (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l
'2; PLANT DESIGN IELUDING SAFETY AND EtGAGEE1 Pega 74 EXETEMS REFERENCE-EIH, LP LT-IH-01101-00, LO 5,8,9,& 11 '
4 Arnold System Description C-4, P. 15 (3.0/4.2/2.5/2.5) 211000K505 211000K409 ~ 211000K408 .211000K402 ..(KA's)
?
s ANSWER 2.05 (1.50)_ l . a. Stabilizing valves (0.25) maintain constant flow through the. - pressure control-valve (0.25) (thus maintaining RPV/ drive water i differential pressure constant) (0,5) , i b. 1) Restricting orifice (0.25) (in the charging water header)
- limits flow (0.?5) (to less than 200 spm) (0.5) i 2)- Flow element for Flow Control Valves (0.25) is located between ,
the pumps and the charging water header-(0.25) (so a high ' charging flow closes the FCVs) (0,5) , r I l i REFERENCE , EIH Nuclear Training. 4.2-7 I (2.5/2.6) ; ! 201001K402 201001K401 ..(KA's) f i I-i } f (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) i (
PLANT DESIGN INCLUDING SAFETY MD EMERGEHQX Pega 75 f J. SYSTEMS
-ANSWER 2.06 (2.00) i a. No response (operates normally) [0.25] because high water level trip is bypassed when operating from-the Remote Shutdown Panel ~.
,- -[0.5)
- b. Any 5 of the following @ 0.25 each:
- 1. Residual Heat Removal
- 2. RHR Service Water
- 3. Safety Relief
- 4. Plant Service Water
, 5. Control Rod Drive
- 6. Reactor Recirculation 4
REFERENCE i EIH, LT-IH-05201-00, PP. 8 & 14 (3.5/3.3) I 217000K402 217000K102 ..(KA's) , ANSWER 2 .'0 7 - (2.50)
- a. GREEN Light - power available for the solenoid control valve [0,5)
RED Light - solenoid control valve has energized [0.5) AMBER Light - high pressure in the relief valve tail pipe (tail pipe pressure swith actuated at greater j than or equal to 85 PSIG) [0.5) I
- b. Manually reset by key lock [0.25) at P602 [0.25) i
- c. 1. Any SRV has opened (tail pipe pressure switch activated) [0.25) i 2. Reactor pressure at high pressure setpoint t
(1044 or 1054 PSIG) [0.25) l l I l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) l l l
L__EL&HLDESIGN INCLUDING SAEEll_AHQ_EMERGEllGY Pcco 76 SYSTEMS REFERENCE EIH Nuclear Training, Chapter 5.1, Pages 7, 21 and 22 (3.7/3.0/3.4) 239002K402 239001K610 239001K125 ..(KA's) ANSWER 2.08 (2.00)
- a. False
- b. False
- c. True
- d. False (4 @ 0.5 each)
REFERENCE EIH Nuclear Training, Chapter 5.1, PP. 10-16 Lesson Plan 49.1, L.O. #7 and #4 P. 5 (3.1/3.2/3.0/2.4) 239003K407 239003K402 239003K401 239001K406 ..(KA's) ANSWER 2.09 (1.00) d (1.0) REFERENCE EIH Nuclear Training, Chapter 5.3, P. 8 (2.5/2.5/2.8/2.6/2.4/3,2) > 256000K301 256000K110 256000K101 245000K502 245000K306 245000K102 ..(KA's) (***** CATEGORY 2 CONTINUED ON NEXT PAGE *4***)
- 2. PLANT DESIGN INCLUDING SAFETY _AND_ EMERGENCY Pcgo 77 SYSTEMS ANSWER 2.10 (2.00)
- a. Any 3 of the following @ 0.5 each:
- 1. Recirculation Pump Motor Windings
- 2. Recirculation MG Set Motor Coolers
- 3. Recirculation MG Set Generator Coolers
- 4. Reactor Water Sample Coolers
- b. The 600V bus load shedding reset pushbutton must be reset. (0.5)
REFERENCE EIH LP 9.1, LO #3 and #4, PP. 3, 4, & 8 (3.3/3.5) 295018K101 295018K201 ..(KA's) ANSUER 2.11 (2.75)
- a. Any 3 of the following @ J.5 each:
- 1. High Drywell Pressure
- 2. Low Reactor Water Level
- 3. High Reactor Building Exhaust Radiation
- 4. High Refuel Floor Ventilation Exhaust Radiation
- b. Any 5 of the following G 0.25 each:
- 1. Particulate fission products in the drywell.
- 2. Gaseous fission products in the drywell.
- 3. Temperature in the drywell.
- 4. Pressure in the drywell.
- 5. Water temperature in the torus.
- 6. Water level in the torus.
- 7. Gaar mn 4,r446 /e<e/s ,b %. Q g
(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
- 2. PL6NT DESlGR_LHQL& DING SAFETY AHD_ EMERGENCY Pcgn 78 SYSTEME _
REFERENCE EIH LP 51.1, LO #4 and #9, PP. 3&5 (3.2/3.7) 223001K403 223001K103 ..(KA's) ANSWER 2.12 (2.00)
- a. Service water to 1B D/G must be aligned manually from Div. I or Div. II PSW. (1.0)
- b. 1. LOCA
- 2. LOSP
- 3. Condenser room flooding (3" above floor)
- 4. High division flow (20 paid increasing) i (4 @ 0.25 each)
REFERENCE EIH LP 33.1, LO 3, P. 5 & 9 (3.3/2.9) 295018K301 295018K201 ..(KA's) ANS'.1ER 2.13 (1.50)
- 1. Close RHR Shutdown Cooling Isolation or Pump Shutdown Cooling Suction Valve (F006A).
- 2. Open Pump Torus Suction Valve (F004A).
- 3. Manually start RHR pump.
(3 @ 0.5 each) i l I l l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
2a __ELAHI._DISIGtLlHQLUDING _ SAFETY _AHQ_EMEBQEHQ1 Pcca 79 SYSTEdd REFERENCE EIH LP-IH-00701-00, LO 21, Table 00701-5 (3.6/3.9) 205000K108 203000K114 ..(KA's)
- l
(***** END OF CATEGORY 2 *****)
32__INSIEUMENTS...AUD CONIBOLS Pego 80 ANSWER 3.01 (3.00)
- a. The 20 non-calibrated jet pump differential pressure signals (0.5) .from the jet pump throat pressure (0.25) and the below core plate tap (SLC injection line) (0.25), are converted to flow signals (by the square root extractors) (0.25).
The jet pump flow signals are summed to obtain two recire loop total flow signals and total core flow (0.25). [1.5 total)
- b. When one pump trips, the circuit selects the summer which calculates the algebraic difference between the two loop flow signals. (1.0)
- c. Flow being measured by the idle loop is backflow from the operating loop. (0.5)
REFERENCE EIH, LP44.1, P. 11 i (3.2/3.3) 216000K123 216000K110 ..(KA's) (3.00) ANSUER 3.02 4
- a. 1. Less than 11 LPRM inputs to the channel.
- 2. APRM mode switch not in "Operate",
i 3. Any internal module (APRM or flow channel) unplugged.
- (also accept the fellowing as one of the required three
Flow channel mode switch not in "Operate".) (3 0 0.5 each) l b. Any 3 of the following @ 0.5 each;
- 1. Reactor half-scram (full scram if 2 or more channels)
- 2. Rod Block i
- 3. Annunciator on P603 l
l l f (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) t
- 3. INSTRUMENTS AND CONTROLE Pcgo 81 REFERENCE i
EIH LP 12.3. LO 4 P. 13 , (4.0/3.6/3.3/3.7/4.1) 215005K402 215005K401 215005K116 215005K104 215005K101
..(KA's)
ANSWER 3.03 (2.00)
- a. True (0,5)
- b. False (0,5)
- c. False (0.5)
- d. True (0.5)
REFERENCE , EIH LP 12.1, LO 7, P.13 - LP 12.2, LO 5, P. 8 i LP 12.3, LO 3. PP, 11 & 12 ! (3.2/3.0/3.3/3.3/3.7/3.9) , t 212000K412 215005K407 212000K411 215005K116 215004A103 215004K406 ..(KA's)
/. o ANSWER 3.04 (4v609
- a. Increase
- b. Increase
-er---Increasetele f e 1
(4'@ 0.5 ea.) 6155t(/. c) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
Pego 82
- 3. ~INSTRUMEUIS_6HD_CQHIBQLS f
REFERENCE EIH, LP 44.1, LO 1, PP. 5-7 CPS: L. P. 74104 PP, 8 & 19. Enabling Objective 2.3. (3.6/2.9) 216000K509 216000K507 ..(KA's) ANSWER 3.05 (1.00) c (1.0) REFERENCE EIH: GPNT, Vol VI, PP. 7.2-20 (3.0/3.6/3.8) 264000K101 262001K406 262001K401 ..(KA's) ANSWER 3.06 (2.00)
- a. 1. Pump suction valve (0.25) must be full open. (0.25)
- 2. Suction pressure (0.25) must be greater than 34 psig. (0.25)
- 3. Pump oil pressure (0.25) must be greater than 5 psig. (0.25)
- b. Trip of a running pump. (0,5)
REFERENCE EIH Nuclear Training, Chapter 5. P. 15 LP 2.1, LO #8, P6 (3.4/2.8) 256000K403 256000K401 ..(KA's) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
J.._INSIBUMEHTS AND CONTROLS Pegn 83 . I ANSWER 3.07 (2.50) i Any 5 of the following at 0.5 each (0.25 for signal. 0.25 for setpoint):
- 1. Reactor vessel high water level - 58". '
- 2. Low Condenser Vacuum - 22.3".
- 3. Thrust Bearing Wear - 40 PSIG Lube Oil Pressure.
- 4. Turbine Bearing Low Lube Oil Pressure - 4 PSIG.
- 5. Low Pump Suction Pressure - 160 PSIG (4 second T.D.)
- 6. RFP Bearing Low Lube Oil Pressure - 4 PSIG.
- 7. Mechanical Overspeed - 110% (6325 RPM).
REFERENCE EIH LP 2.1, LO4,P.7
! (2.5) l 259001K405 ..(KA's)
ANSWER 3.08 (2.25) t ! Any 3 of the following; l 4 1. Reactor Recirculation System (0.5)
- a. Total feed flow to #1 speed limiter. (0.125) d b. Level and individual feedwater loop flow to #2 speed
- limiter. (0.125)
- 2. Process Computer (0.5) total steam flow and total feed flow
, for heat balance. (0.25)
- 3. Rod Worth Minimizer (0.5) total steam flow and total feed flow for power determination. (0.25) ,
- 4. Main Turbine (0.5) level for high water (58") trip. (0.25) r c
(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****) I i
- 3. Pcga 84 INSIBUMEBIS_AND_CONIBQLS REFERENCE EIH Nuclear Training, Chapter 9.5, P. 15 LP 2.2 (3.1/2.6/3.6/2.9/3.2/3.1) 202002K109 259002K115 259002K114 259002K105 259002K107 201006K103 ..(KA's)
/, $o ANSWER 3.09 (M)
- a. 50 PSIG Instrument Air Header Pressure,
- b. 70 PSIG Instrument Air Header Pressure.
- c. 90 PSIG Instrument Air Header Pressure. -c/ed/e, 4 -+1FP&IG-IG-Pneumatic-System-Header-Presures DL @ 0.5 EACH]
3 REFERENCE t EIH, LT-IH-03501-00, LO 11 & 12, Table 03501-9 (3.2/3.3/3.5/3.2) 295019K303 295019K302 295019K301 295019K214 ..(KA's) ANSWER 3.10 (1.50)
- 1. Keylock awi.tches (on P609 and P611) in bypass.
- 2. Turbine stop valves closed.
- 3. Mode switch not in Run.
(3 @ 0.5 ea : 1.5) (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
\
J
- 3. INSTRUMENTS AND CONTROLS Pcgo 85 REFERENCE EIH Nuclear Training, 5.7-5 DAEC - System Description. A-6, Main Steam, page 36.
(3.8/3.1) 239001K402 239001K401 ..(KA's) ANSWER 3.11 (1.00) C (1.0) REFERENCE EIH, LP LT-IH-05402-01, LO 8. P. 33 (3.3) 201004K404 ..(KA's) ANSWER 3.12 (1.00) i
- a. Operations supervisor (0.5)
- b. power above the LPSP (30% power) [0.5)
REFERENCE EIH, LP 54.2, LO 6 & 7, P. 10 (3.4/2.9) 201006K502 201006K404 ..(KA's) ANS!!ER 3.13 (1.50) The closed breaker will trip open (0.5) and the breaker being closed will trip open (0,5) resulting in a de-engergized bus. [0.53 (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
- 3. . ...IHEIRUMESTS AHD_QQHIBQLS Pcgo 86 REFERENCE EIH LP 27.1, LO 13, P. 13 (3.4/2.9/3.1) 262001K402 262001K103 ..(KA's) ;
262001K403 ANS!IER 3.14 (2.50) Any 5 of the following @ 0.5 each;
- 1. Emergency 4160V bus undervoltage
- 2. Proper output frequency
- 3. Proper output voltage
- 4. No electrical fault on bus.
- 5. Normal bus supply breaker open.
- 6. Alternate bus supply breaker open.
REFERENCE EIH LP 28.1, LO2,P.7 (3.8) 264000K101 ..(KA's) l i (***** END OF CATEGORY 3 *****)
- 4. PROCEDURES - NORMAL, ABNORMAL,__ EMERGENCY Pcgo 67 AER_B6Q1QLOGICAL CQtlIBQL ANSWER 4.01 (2.00) i
- 1. Close the following valves; (valve numbers not required for full credit) l 1
- All MSIVs (b21-F022A(B,C.D) and B21-F028A(B,C,D)) (0,5) 1
- Both Steam Drain Isolations (B21-F016 and B21-F019) [0.5) !
- Both Reactor Water Sample Isolations (B31-F019 and B31-F020) [0.5) !
- 2. Open the RFP Bypass (N21-F113) (0.5)
REFERENCE EIH, 34AB-OPS-055-2S i (3.8) 295016G010 ..(KA's) ANSWER 4.02 (2.00)
- a. Alert
- b. (Notification of) Unusual Event
- c. General Emergency
- d. Site Area Emergency (4 @ 0.5 each)
REFERENCE EIH: GET Handbook, pp 57, 58, 60, 61; 10AC-MGR-006-05 BFNP: BFN-IPD, IP-1, p 1; RQ 85/04/01 (4.3) 295030G011 ..(KA's) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
L _PROCEDUBES - NQBd6L, ABBQBdakt_EMEBG_Et{gl Pcgo 88 ARQ_BAQ1QLpfllCAL_CONTRQL _ ANSWER 4.03 (2.00) Any 4 of the following 9 0.5 each;
- 1. Profilter differential prorsure
- 2. Afterfiltar differential pr(ssure
- 3. System pressure
- 4. Post Accident radiation level
- 5. Offgas System flow rate
(., . 4/3 decyed etdeesk fid REFERENCE EIH, 34AB-OPS-038-2 (3.8) t 271000G015 ..(KA's) ANSWER 4.04 (2.50)
- a. 2
- b. 4
- c. o
- d. 3
- e. 1 (5 0 0.5 EACH)
REFERENCE EIH, LT-IH-20101-00, LO 7, PP. 18 & 19 (3.9/3.9/3.9/3.8) , 295006G012 295031G012 295025G012 2950240012 ..(KA's) (***** CATEGORY 4 CONTINUED ON NEXT FAGE *****) < J l
3 PR0gEDUBES - NGEMAL, ABNORMAL 2_ EMERGENCY Pcgo 89 AND_ BAD 19LMI. CAL _.09ETEOL ANSWER 4.05 (1.50)
- 1. At least one Core Spray pump is injecting, or
- 2. Reactor water invol is above the top of active fuel, or
- 3. Steam cooling is in progress.
(3 @ 0.5 each) REFERENCE i EIH, 31EO-EOP-001-2S 3.83, p. 3 ' (4.6/4.6) i 295031A204 295031K101 ..(KA's) ANSWER 4.06 (1.00) l I l Prevent overpressurization of the CRD suction piping. [1.0) REFERENCE EIH, 34SO-C11-005-2S (3.2) ' , 201001G010 ..(KA's) t i . ANSWER 4.07 (1.00) To avoid excessive vibration on the piping returning to the main condenser. [1.0) 4 { I - l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) { [
- n -
v - i Pego 90 4; PFdCEDUBES:- HQB ML AHHQBE L'. EMERGENCY ' JulD RAQlOLOGICAL CONTROL i REFERENCE EIH, 34SO2N21-007-2S, p. 15
, (3.1/2.4) 256000K506 256000G010 ..(KA's)
ANSWFr. 4.08 (1.50)
- a. -RWCU system isolation (in which both demineralizers transfer to hold) [0.5)
- b. Provide a pressure relief path for CRD water [1.0)
REFERENCE EIH, 34SO-G31-003-2S, P. 15 (2.7) 204000K402 ..(KA's) ANSWER 4.09 (1.00)
- 1. Significant amounts of H2 and 02 are present in the main condenser (which could result in a detonable mixture). [0.5)
-2. The mechanical vacuum pump bypasses the holdup volume. [0.5)
REFERENCE EIH, 34SO-N61-001-2S, p. 3 (3.1/2.7) 271000K507 271000G010 .(KA's) (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
- . .. - _ _ ~
P
- 4. PROCEDURES'- NORMAL. ABNORMAL. EMERGENCY Page_91
.AND RADIOLOGICAL CONTBQL ANSWER 4.10 (1.00)
Some machine gas pressure will'be maintained (25-30 psig) by the bearing header oi1~ supply to the seals [1.0] (so a load reduction.is all that will be required). REFERENCE EIH, 34SO-N42-001-2S, p. 6 (2.8/2.8) f 245000K610 245000K603 ..(KA's) l ANSWER 4.11 (1.50) Any 3 of the following @ 0.5 each;
- 1. High area radiation levels (greater than 100 hr).
- 2. High Airborne radioactivity concentration (grescer than 25% MPC) ,
i
- 3. Loose surface' contamination (levels greater than 50,000 dmp/sq.cm) l I 4. Breach of a contaminated system.
REFERENCE i ! EIH, 62RP-RAD-006-OS, P. 3 (3.3) t 294001K103 ..(KA's) t ANSWER 4.12 (1.00)
- a. FALSE [0.5]
- b. TRUE [0.5]
(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
. .-, . . . - . _ _ . - - _ . _ . ., ._-- ,- -, _ , ,.-_.- _. _ ,, ... _ --.--,~. .
- 4. PROCEDURES - EORMAL, ABNORMAL, EMERGENCY Pags 92 AND BADIOLQGLQAk CONIRQL REFERENCE EIH, 30AC-OPS-003-0S, PP. 16 & 17 l (3.4) j 294001A106 ..(KA's)
ANSWER 4.13 (2.00)
- 1. Tags included on a Temporary Release (for functional testing) [1.0)
- 2. Tags where all subclearances affecting them have been released and a Shift Supervisor (or Operations Supervisor) review has been performed. [1.0)
REFERENCE EIH, 30AC-OPS-001-OS, PP. 15 & 18 (3.9) 294001K102 ..(KA's) l l i l l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
~4. ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 93
- AND RADIOLOGICAL CONTROL-ANSWER 4.14~ ( 3'. 50 ) -
- a. 1. Turn valve in closed direction (1/4 turn max to seat) 2 '. Turn valve in open direction (1/4 turn max to' backseat)
- 3. Verify at remote (or local) position indication
- 4. Confirm locking device operability by attempting to move valve.
( 4 @ 0.5 each )
- b. l '. Reposition valve after getting SS concurrence
- 2. Install locking device and verify operability
- 3. Prepare a deviation report
( 3 0 0.5 each ) REFERENCE .; EIH, 34GO-SUV-001-0S, P. 2 L 10AC-MGR-004-0 (3.7) i 294001K101 ..(KA's) ANSWER. 4.15 (2.00)
- a. 2 Minutes (0.5]
3 b. Shallow [1.0]
- c. "Print Notch Error" function (of RWM) [0.5]
r I (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
,,-+-vm--, . - , , - - , , - ~ -y-.,. -,-w,v.-rvr....y,.nn,,,-,w, ~.-,..w.,-e,.. ,,,,,n,e_,.y,,,,--,n,,,,--,
, , - , , , , , , , , - - .,na ,--- . , , -,, , - , -
~ "' 4 . PBQCEDURES - NORMALi' ABNORMAL. EMEBGENCY Page 94.
AND RADIOLOGICAL CONTROL
REFERENCE:
EIH: 42FH-ENG-010-1/2 (2.8/3.2). 201004K405 201006K507 ...(KA's) ANSWER 4.16 (0.50) FALSE- [0.5)
- REFERENCE EIH: 34 B-FPX-053-2,Lpp 8 (3.5) 294001K116- ..(KA's) i l
l l l l y - l (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********) L
m-1 U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _UAlgd_1k2_______________ REACTOR TYPE: _@W8-@g4_________________ DATE ADMINISTERED: _ggdg240g________________ EXAMINER: _ bgPEER 1 _gz______________ CANDIDATE: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ INSIBUgIlgNS_IQ_CGNDIQ@151 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 eacn 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
__Y86Ug_ _IgI@L ___SGQBE___ _YG6UE__ ______________G81EGQBY_____________ _2Zsg0__ _ ggt 6g ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS _25t29__ _2}s22 ___________ ________
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_2@s@@__ 26 6g ___________ ________ 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 2Et0@__ _g}zZ$ ___________ ________ B. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 19512$__ ___________ ________% Totals Final Urade All work done on this examination is my own. I have neither given nor received aid. Candidate's Signature i
r_ _NRC, RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
~
During the_ administration of this examination the following rules apply:
-1. Cheating on the uxamination means an automatic denial of your application and could resul.t-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 goly 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.
G. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a ogw page. write gely 90 gag sidg of the paper, and write "Last Page" on t~1e last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least thtgg 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 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.
t l 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE l OUESTION AND DO NOT LEAVE ANY ANSWER BLANK. f
- 16. If-parts of the examination are not clear as to intent, ask questions of l
- the egamicet only.
l
- 17. You must sign the statement on the cover sheet that indicates that the l
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.
l l [
E 18.DWhen 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.
t l l l I l l-i
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _U@ICU_1.%2_______________ REACTOR TYPE: _ BUR-GE4_________________ DATE ADMINISTERED: _89202/08________________ EXAMINER: _UOPPER t _G.______________ CANDIDATE: __h_h_ I_h___________ INSIBUCIIgNS_IQ_CONDID81El Use separate paper for the answers. Write answers on one side only. Staple question sheet n +1 tcp 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
__MGLUE_ _IQIGL _ _ _SCQEE _ _ _, _Y@LUE__ ______________C01E@g8y_____________ 2Z390__ _g}.64 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS _251 35__ _}1 tS6 ___________ __,______ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION _29199__ _2St@2 ___________ ________ 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL _@@z29__ ____________ ________% Totals Final Grade All work done on this examination is mv own. I have neither given nor received aid. Candidate's Signature i 1 1 i l w
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the f ollowing 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 4
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 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 gach i section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate,' start each category on a ogw page, write gnly 90 gng sidg of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each an ser 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 litetatutg.
' 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. l 14. Show all calculations, methods, or assumptions used to obtain an answer l to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore. ANSWER ALL PARTS OF THE l QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
l 16. If parts of the examination are not clear as to intent, ask questions of l _the ggamingt only. 1
- 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 asststance in completing the examination. This must be done after the examination has been completed.
I I L
- 18. When you complete your examination, you shall: ;
1
- 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.
l l
- d. Leave the examination area, as defined by the examiner. If after l leaving, you are found in this area while the examination is still
]
in progress, your license may be denied or revoked. , 4 5
)
t d a t I a i i. I r l L
'. . . - - .,.._--,..._-.-,._,..,___..___._,___.....__.._,,..__.-...._._...-,___.m_.._____-,__,.,..-,_,, --f
.Di__IBE98Y_9E_NUg6568_EgWEB_E66NI_gEESSIlgdi_E691Dgt_Odp PAGE 2'
.IBEBd99YU8MIcs QUESTION 5.01 (2.00)
The reactor'is operating at 60 % power. ANSWER EACH ONE of the following questions as TRUE or FALSE.
- a. INCREASING recirculation flow will cause reactor power to INCREASE.
- b. DECREASING reactor pressure will cause a REDUCTION in reactor power.
- c. LOSS of a heater string will cause power to DECREASE due to less subcooling of the feedwater,
- d. INSERTING control rods will usually cause reactor pressure and generator output to DECREASE.
1 (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. ' THEORY' F NUCLEAR POWER PLANT OPERATIONx _ELy1QSt_8NQ PAGE 3 IHEBdQDYN8d1C@
l QUESTION 5.02 (2.00). Answer EACH ONE of the following statements TRUE or FALSE regarding reactivity coefficients.
- a. An increase in flow through the reactor core will. add negative ,
reactivity by decreasing the void fraction and thus increasing reactor power.
- b. As the burnable poison within a fuel bundle burns out,the VOID '
coefficient becomes more negative.
- c. LATE in core life, the large reduction in fuel molecules and the decrease in moderator density during a plant HEAT-UP can lead to a positive reactivity addition.
- d. As core age progressen, the DOPPLER coefficient becomes more negative due to plutonium-240 buildup.
i l l I (4**** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l I i l
.-. - . - _ ,___. -_..=.m__ . . . _ _ , , , _ , . . . _ . , _ _ _ , _ _ . . . _ . . . _ _ _ . . . . , _ _ _ . . , . . _ , , . _ _ _ _ . _ _ _ . _ . _ . _ . _ _ _ _ _ _ , ,
ht__ldE961_WE_UWGbEBB_bWWEb_Eb6dl_9thb811gNt_tLQ1QSg_93Q PAGE 4
'IHE6dODYNAdlCS QUESTION 5.03 (2.00)
Antwer EACH ONE of the following questions TRUE or FALSE concerning th3 graph of Contal Rod Worth During a Startup. control I rod I worth . .i 7 I 1 1
, l
.seG ___________J____________________
cold heat-up power operation shutdown complete %100 pwr 1% pwr -
- c. Control rod worth increases during heatup due to density decreases of the moderator which causes longer slowing down and thermal diffusion lengths, resulting in more thermal flux around a control blade.
- b. Control rod worth decreases as power exceeds 1% due to the effects of rod shadowing. Withdrawal of rods increases the thermal diffusion length thereby increasing the flux around a control blade.
- c. While heating-up, rod worth increase is due mainly to the effects of Bundle Coupling. Rod withdrawal couples fuel cells together making their effective size larger, resulting in increased leakage and a reduction in thermal flux.
- d. Since control rods are worth more when the moderator is hot, fewer control rods must be withdrawn to go critical when the reactor is hot than when cold.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) E 4
,~_
* ?
52_iTHEORY OF NUCLEAR POWER PLANT OPERATIONt ;E691pSg_8NQ PAGE 5 IUEBdgpyN@d1CS QUESTION 5.04 (2.00) Answer EACH ONE of the following statements TRUE or FALSE.
- a. Improper voiding and neutron thermalization caused by incorrect bypass' flow'could result in inaccurate indications of power.
- b. A portion of the care bypass flow is provided by coolant holes in the lower core plate.
.c. Increasing power by control rod withdrawal will increase core bypass flow.
- d. A change in core pressure drop has a greater effect on core bypass flow than on channel flow.
4 (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
Dz__IUE9BX_9E_ NUCLE 88_EQWEB_EL8NI_gEEB8IlgNt _ELUIDSz_8NQ PAGE 6 IUEBdggyN8dlCS QUESTION 5.05 (2.00) Match EACH ONE of the'following terms in COLUMN A with the best corresponding definition in COLUMN B. Choose ONE (1) definition for each term. COLUMN A COLUMN B
- a. Keff 1. An amount of fuel loaded into the core above that required for initial criticality to compensate for
- b. Shut Down Margin control rod motion.
- c. K-excess 2. A measure of the fractional change of the neutron multiplication factor,
- d. Reactivity
- 3. 1-keff/ Keff -1
- 4. A measure of how sub-critical the reactor is in terms of Keff.
- 5. The % of neutrons which are born delayed.
- 6. A measure of the availability of fuel loaded into the core above that is required for initial criticaN4g
- 7. A ratio = # neutrons produced from thermM f i ssi on
# neutrons proouced trom thermal fission preceeding generation l
i i l I l L (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) { r
Dz__IME96Y_9E_NUCLge8_EgWE8_ELGUI_ GEE 88IlgN1 _E6U1ggz_8NQ PAGE 7 IUESM99YUBMICg , OUESTION 5.06 (1.00) Match EACH ONE of the following statements in COLUMN A with the corresponding acronym (abbreviation) in COLUMN B. COLUMN A COLUMN B
- a. RPF*APF*LPF= 1) APLHGR
- 2) MCPR
- b. Thermal limit established to ensure 3) GEXL PEAK CLADDING TEMPERATURES do not 4) TPF exceed 2200 F in event of a design 5) LHGR basis LOCA. 6) PCIOMR
- 7) DNB
- c. Safety limit that is not analyzed 8) MAPRAT at <10% power or <e00 psia. 9) PCI
- d. A condition of maximum heat flux which occurs when steam bubbles combine to form a vapor film over a heated surface.
l 9 (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
'Dz__IUEg8y_QE_NyG6g68_EgWEB_E69UI_QEEBBIlgNt _ELUlpgx_6NQ PAGE 8 ISEBdggyNSdlGS
.QUESTION. 5.07 (1.00)
- FILL IN THE BLANKS.
(a) ____ is the primary means of heat transfer Following a LOCA, ___ fuel rod to a cooler fuel rod. from an uncovered (b) is'the process of transferring heat.between a fluid and a surface by the circulatian or mixing of a fluid. When heat is applied to a homogeneous material, the _____ (c ) _____ of the material's atoms is increased. A _____(d) ______ must exist in order for heat to be. transferred through a material. (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
Uz__IUE98Y_9E_NU9650B_E9 WEB _E60VI_9EEBOI1OUx_ELU1991 _00Q PAGE 9
~ISEBdgpyd8d1CS ,
i as
. OUESTION' 5.08 (2.00)
- a. LIST TWO (2) conditions that are used to determine WHEN the reactor is critical during a startup. (1.0) ;
4 i b.'WHICH ONE (1) of the following statements DEFINES REACTIVITY ANOMALY ? (1.0) , f
- 1. An unexpected change in critical rod configuration
; due to normal Xenon transients.
- 2. A change in the critical rod configuration due to the buildup of Samarium in a new core.
j 3. A reactivity equivalent of the difference between ; actual and expected rod configuration during steady ; state reactor power operation.
- 4. An unexpected change in reactor power induced via a reactivity change resulting from a plant transient.
t i
~
n o I . f \ F 4 t I I l I ' i t t t ! L 1 i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) I l-l f i ! i l [ l i P w.ra-r~ns w -sm,weww w .y--- g
5 t__IdE96Y_92_NUC6ETW_EQWEW_E(@NI_QEEB@llgN _E6UdQSz_@ND t PAGE 1N
'ISEBdgQyN@dlCS OUESTION 5.09 (1.50)
Utilizing the BWR Steam Cycle Diagram below, MATCH the line ecgments in COLUMN A with the associated plant component in COLUMN B.
! 3 TEMP 1 i
?
I I i
! i I
t i l 5
! i, 1 e l_____________________________________
ENTROPY COLUMN A COLUMN B 1-2 Low Pressure Turbino 2-3 Condenser 3-4 Pumps (condensate, booster.RFP's) 4-5 Heaters and Reactor 5-6 High Pressure Turbine 6-1 Motsture Separator (HP Turbine exhaust) l (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) 1 E l 'T '- I
~
l
- o ,
t
Sz__IUE96Y_gE_UgC6596_EQ2EB_E6@dI_QEEB@llggt_E6ylppt_@NQ PAGE 11
'ISEBd9QyN901CS QUESTION 5.10 (1.50)
EXPLAIN WHY the curve of Keff vs. Core Age changes as it does at;
- c. The DECREASE at point 1.
- b. The TURNING UP at point 2.
- c. The TURNING DOWN at point 3.
I 1 +-1 3 I / - Keff I g i 2 1 1 I l_____________________________ core age (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) E y .- 1 h $ e
51- IBE98Y 9E_NyGLE88_EQWEB_EL6MI_gEEB811QN t _E6919@t_9NQ PAGE 12
= ic , THERMODYNAMICS
' QUESTION S.11 (1.50)
A reactor scram occurs following several weel
__________y,g_________
- o. EXPLAIN what would happen to the system piping i f the valve were (.75) to suddenly (instantly) close.
- b. Can a similar effect be achieved by suddenly OPENING the valve? (.25) d J
i V e i I n o l e t 9 i I i
- (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) .
t I O i k '-- !
, f w
,p- g , -.e -- - - - - - - r--v-ye,,-w--s--, ,m --. _w----.y-.,m.----.w ~ e- wm--re,.g.-%,,-,-vww--we y,- y-~e,. ,,-,,a~, e------e v- ,r-w+- v-----
. Qt__IBEORY OF NUCLEAR POWER PLANT OPERATIONt _FLUlp@s_8NQ PAGE 14 IdgRMOQYN8d1CE ; ' OUESTION 5.13 (2.00)
- The attached FIGURE (1) represents parameter changes for a .
plant transient on UNIT TWO. Use this figure and t' , following [ iniarmation to answer the questions below. . (1) Initial Power Level = 100 % (2) The Main Turbine Generator Trips without the Bypass Valves opening f-(3) No operator actions are taken I EXPLAIN the specific cause(s) of-the following indications: '
. The INCREASE of Reactor Power (point 1) '
.. Core Flow DECREASE (point 3)
- c. Reactor Vessel Level DECREASE-(point 6)
, d. The INCREASE in Feedwater Flow (point 8) r p i 3 L J i t, 6 B P i 6 l , 1 1 k I I r 1 ! i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
Ut__IMEgBy_QE_NUGLE88_EgWEB_E68NI_9EEE8Ilgdi_ELUlpSt_8NQ PAGE 15 ISEBdODYN801gG _Q UESTION 5.14 (2.00) The attached FIGURE (2) represents parameter changes for a
~
plant transient on UNIT TWO. Use this figure and the following information to answer the questions below. (1) Initial Reactor Power = 80 % (2) The Master Retire Controller Fails to MAXIMUM (3) No operator actions are taken EXPLAIN the specific cause(s) of the following indications:
- a. The DECREASE in Reactor Power (point 2)
- 6. Reactor Water. Level DECREASES (point 3)
- c. Core Flow DECREASES (point 4)
- d. Reactor Steam Flow DECREASES (point 6) l I
i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. ' THEORY OF NUCLEAR POWER PLANT OPERATION x_FLUIQS t_Atjp PAGE 16 INE8dQQyd8dlCS.
QUESTION 5.15 (1.00) CALCULATE the equilibrium neutron count rate in a subcritical~ reactor after FOUR (4) generations given the following initial conditions: Source =.100 cps Keff = .2 Assume generation 0 consists of only source neutrons and equilibrium' is achieved af ter f our generations i t 4 t I l P i i l t
- i l >
1 l l ! l l ! i t i t f t' * * * *
- CATEGOHY 05 CONTINUED ON NEXT PAGE *****) l l
I i i h f
, . . . ~ . . - . _ - _ , _ _ , _ . - . _ - . _ _ _ . . . . , _ . . _ . _ - . _ - - ~ - _ . _ _ . . . - - . . -
5 t__IUE98Y_9E_ NUCLEAR POWE8_ELCUI_QEEB8110N1_ELUlpS1_8NQ PAGE 17
-IUEBOOQyN801CS QUESTION 5.16 (1.50)
During a reactor shutdown tne reactor vessel pressure was decreased from 800 psig to 350 psig in 45 minutes,
- a. CALCULATE the cooldown rate. (SHOW ALL WORK) (1.0)
, b. : STATE the technical specitization maximum allowable cooldown rate. (.5) d 4 1 i I i L 1 l 1 i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) I i l i l
[. ! . . i S-t THEORY' QF NUCLEAR POWER PL(3 N T OPEh3 TIONt _FLUIDSx_ANQ PAGE 10~ -l
. IHE8t]QQYN8t],[C@ :]
j' l 1 ! l i i
. OUESTION 5.17 (1.00) .
t
- a. DEFINE NPSH (Not Positive Suction Head).
4 >
- b. EXPLAIN why operation of the recirculation pumps is not allowed i l
i~ with insufficient NPSH. j e l.~ j i i 4 5 i 1 4 4 A L 1 l 1 4 .f i i i i i (***** END OF CATEGORY 05 *****)
6t Eb63I_SZSIEUS DESIONt_cDNIB96t_GND_IUSIBUd5NIBIlgN PAGE 19 i
. QUESTION 6.01 (1.00)
Which ONE of the following conditions will NOT result in a' shutdown of the SDGT System.
- a. Manual shutdown
- 6. High temperature-225 dog F cnarcoal bed
- c. High temperature 100 deg F heater inlet
- d. O*.<or l oad s in local. control panel e I
1 i ( I i i i (**'*** CATEGORY 06 CONTINUED ON NEXT PAGE *****) ! (
As- EL8NI_@y@l5NS_Qg@lGNi_CQNIBQL1_6UQ_1NSIBQdgNI@IlgN PAGE 20 QUESTION 6.02 (1.00) i The plant is operating normally at power when Pump A Controlled Leakage
-(FS "A") alarms LO (0.1 gpm) and you noto an INCREASE in No.2 Recirc 4 Pump seal pressure with NO CHANGE in No. 1 seal pressure. Which ONE of the following failures would cause those indications?
. a. Failure of No. 1 seal
- b. Failure of No. 2 naal
- c. Plugging of the No. 1 internal restricting / breakdown orifice d_ Plugoing of the No. 2 internal restricting / breakdown orifice NOTE: NO OTHER ALARMS ARE PRESENT FIGURE (3) IS ENCLOSED FOR REFERENCE i
i I l I l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
6 1__ELGUI_DYSIEUS_ DES 19Nt_QQNIBQL1_GUQ_INgIBQUENIGIlgN -PAGE 21 1 i
)
~
' QUESTION 6.03 (1.00)
The UNIT ONE Vital AC Power 120/240 V Distribution Cabinet 2A i is normally supplied from 600 V Dun 2D through a Dattery Charger [ and a Static Inverter. Which DNE of tree f ollowing most accurately ; describes the response to the static inverter failing. I i
- a. The power supply will automatically transfer to the alternate I 1 600 V Bun 2C / Vital AC Transformer 2A.
e
- b. The 125 VDC battery will maintain power to the Vital AC Cabinet for up to 5 hours. ;
- c. The power supply can be manually transf erred to the alternate 1 600 V Dus 2C / Alternate Static Inverter by depressing a a transfer pushbutton.
- l. d. The power supply can be manually transferred to the alternate i 600 V Bus 2C / Vital AC Transformer 2A by positioning the transfer switch to "Alternate". ,
i i i i l l l l l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) l
;6t__EL8HI_SYSIGde_ DESIGNt_QQUIBQLt_8NQ_IN@lByMENI8IlgN- PAGE 22 QUEST' ION 6.04 (1.50)
Answer EACH ONE of the falowing questions concerning the ADS, TRUE OR FALSE.
- a. On loss of power, the "B" logas circuit will automatically shift to the alternate 125 VDC ower supply.
i
- b. ADS is required to be operable when reactor pressure is greater than 150 psig.
- c. The ADS Timer Reset Pushbutton will stop ADS permanently until the ADS Logic is reset.
a T
)
l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
16t __E60NI_@ySIEMS_QESIGN1_CggIBg62_8NQ_1NSIBydENIBIlgN- PAGE 23 t i
. L QUESTION 6.05 (2.25) [
s ANSWER the following questions concerning PUMPS associated with the Condensato/Feedwater system.
- a. MA1CH the Pumps in column A with the Pump Trips in column B t NOTE: EACH PUMP MAY.HAVE MORE THAN ONE ANSWER (1.75) i COLUMN A COLUMN B Cond. Booster Pumps 1. Low suction Pressure of 34 puig Reactor Feed Pumps 2. Low Condenser vr.cuum - 22.~"
Cond. Pumps 3. Low suction pressure 18" Hg CRD Pumps 4. Low Oil Presst:re of 5 psig
- 5. Mechanical Overspeed - 110%
- 6. Low Hotwell Level < 39"
- 7. LOCA load shed 3 1
- b. Why is there a 50 sec T.D. on tne auto start of a Standby . (.5) ,
Condensate pump if any condensate pump is tripped ccincident with a LOCA signal 7 i ! k l a' i r f I i I I I i t l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) y [ 1 I'
i: -6t__ELONI_SYSIEd@_QES100t_QQNI6QLt_@NQ_INSIBydg@l@IlgN PAGE- 24 i 2-t- i OUESTION 6.06 ( .50) Answer the following questions concerning the Core Spray sparger I line break detection DP instrumentation: I l a. DELETED s-1
; b. Is the normal (sparger intact) indication at rated power (.25) j positive or negative with respect to zero ?
l c. Is it norma). for the A loop and B loop indications to have (.25)
.different values when operating at rated power ?
j l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
r; s, . E L. _ELeNI HYSIEDS_ DES 1QNx_GQUIBQ6_GUD,1NSI69dENI8IlON PAGE 25 QUESTION 6.07 (1.00) LIST the signals that will cause automatic isoletion of the Off Gas System. (1.0) t I i r k l 5 i < I i * ] ! 1 ? , 4 l 1 5 i k i l
! (***** CATLGORY 06 CONTINUED ON NEXT PAGE *****) t t
J f
. _ , . . . _ . , _ _ _ _ - _ . _ , .____.,____,_.._1
6 1__Et@UI_SXSIEdS_DESION1_CONIBOL1_80p_INSIBUDEUISIlQU PAGE 26 f I QUESTlON 6.08 (2.00) Concerning the Reactor Manual Control System: With the Refuel Platform over the core and the Reactor Mode Switch in Refuel, LIST FOUR (4) retueling interlocks which will result in a control rod withdrawal b l oc k: . (BE SPECIFIC) i (**444 C H T f_ til0 R i 06 L:ON1INUED UN f4Ex I F nt.E *es**) { l l 1
. .-.._.. - . .. . . . . . - . = _ - . - _ . - . . . _ _ . . _ - - - -. . _ - . . . . . . . - - -
65-_ELONI_SY@lgd@_QE@lGUt_GQUIS961_8Mp_INSI6WDEUIGI190 PAGE 27- t QUESTION 6.09 (2.00) ; LIST FOUR (4) CONDIT40NS that will AUTOMATICALLY TRIP a I Diesel Gonorator at ANYTIME. 4 4 f 4 , 4 i I r i 3 4 i b ] i ! I t 6 i t 1 1, ! i i , .i I 4 4 i ! i i ' i l A. ! I f 7 4 e
- t i
i. J t I I I f 1., e (***** CATEGORY 06 CONTit4UED ON NEXT PAGE *****) L e 1 I h ' i I i b h 1 i
?
j.-____.________..__......... _
- 6,.__EbeNI_SYSIEUS RES19 h_G9BIE96t_BND_INSIBUdENI8I19N PAGE 28 i I
1 1 , 4 ! OUESTION 6.10 (2.00) Answer the following questions concerning the CRD System
- a. LIST the two (2) causes of a CRD Accumulator Trouble alarm I (Setpoints NOT required) and EXPLAIN the action which must be I i taken to dotermine the cause. (1,0)
- b. Shortly after resetting a reactor SCRAM, it is reported :
that Cooling Water flow is LOWS however, the CRD flow indicator I 4 is reading FULL SCALE. EXPLAIN this apparent discrepancy. (1.0) r I , t
?
l I t i if .l f 1 i i i , l i l 1 i i ! I ! l I (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) , I I' ! t l i I I. .__ _ _ _ _ _ , _ . , _ _ _ _ _ _ _ , . , _ _ _ . . _ . _ . _ _ _ , , _ _ _ _ , . . . _ _ . _ _ . . . . . _ _ _ , _ . , . . . . . . _ . . . . . . . , . . _ . . . _
6t__ELONI_SYSIENS DESISut,QQNIBQLt_QUQ_10$1BydEN181190 PAGE: 29 QUESTION 6.11 ( l'. 00) Answer the following questions concerning Primary and Secondary Containment Systems for Unit is
- a. STATE the Bases for inerting the Drywell and the Torus. ('.0)
- b. DELETED
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
- Ar.__ELBMI_SYSIEd@_Qg@lG@t_CQNI6Q61_8NQ_IN@lBLJdEN18I19N PAGE 30 QUESTION 6.12 (2.00)
With regards to the Mann Steam Line Radiation Monitoring systems LIST-FOUR (4) automatic actions that occur when the trip setpoint of the MSL radi ation monitor is reached. (***** CATEGORY 06 CONTINUED ON NEXi PAGE *****)
- - i
< es__ELONI_SYSIEMS_ DES 198t_QQN18QLt_G@Q_ldQ16QMENIGIlQN PAGE- 31 i i i 4 i i ! .i OUESTION 6.13 (3.00) i
- a. LIST THREE (3) ways that the Rod Block Monitor (RBM) may be !
bypassed on UNIT 2. (1.50) : r i j b. Discuss how the LPRM'S input into the RBM CHANNELS (number j j and lovel A,B,C,D) when a control rod near the conter of : ! the core is selected. (1.0) i I c. Why cro the level "A" detectors not utilized by the RBM SYSTEM ? (.5) i l l
)
J l t f i i 1 l I i i i l. l l 1 4 i , a ? I t
?
(644** CATEGORY 06 CONTINUED ON NExi PAGE ****6) i I t
bz__PL@NI_Sy@l@ND_Q5 Sign 1_GQNISQLt_8NQ_idgI69dENIGIlgN PAGE 32 OUESTION 6.14 (3.00) With UNIT ONE operating at 100 % power, recire in Master Manual, an operator inadvertently DECREASES the " PRESSURE SET " by 5 pai. DESCRIBE the INITIAL response and FINAL status of tho
-following parameters due to thin action. Briefly EXPLAIN the reason for EACH of your answers.
NOTEt ASSUME NO OPERATOR ACTIONS FIGURE 9.4(12) IS ENCLOSED FOR REFERENCE
- a. TCV position
- b. BPV positi on-
- c. Power-
- d. Pressure
(***** CATEGORY 06 CONTINUED ON NEAT PAGE *****)
bz__E66MI_SYSIEUS_DE91901_QQNIB96t_80Q_19$J6905U181190 PAGE 33 QUESTION 6.15 (1.00) 1 The RCIC Barometric Condensor maintains a vacuum via steam condannation by no::zio spray and a condonner vacuum pump. If the condenser vacuum pump were to f all, could steam condensation maintain a vacuum indefinitely ? EXPLAIN your answer. i i 4 1 i I l l a i t l I l i (:**** CATEGORY 06 CONTINUED UN NEXT F' AGE ****s)
a 7
.hi__EL8NI SYSIgNS_pgsigN,_G9 NIB 96t_6ND_INSIB9dENIBI19N PAGE 34 1
1 l
- QUESTION t.'6 (1.00) 4 j
ANSWER the following questions concerning the MSIV Leakage Control Systemt l a. STATE the purposo of the MSIV-LCS. s 1
- D. EXPLAIN why Unit One does not have this system.
l e f h 1 I i 1 I i l I i i l (*4444 END OF CATEGQHY 06 *4***)
Zx__EBQGERUBE3_: EQB68bi_8089Bd86t_EdEBGE8GY_8NQ PAGE 35 , 680196991G86_GQUIB96 , i i t i QUESTION 7.01 ~(1.00) ;
.The following parameter changes / annunciators are abuerved by the Reactor Operator ;
RDCCW Temperaturo LOWER than normal. ! Low DP alarm (RBCCW to PSW) i RBCCW Surge Tonk level INCREASING , WHICH ONE (1) of the following malfunctions would cau !
- a. RBCCW 1eak in the Drywell ;
- b. Reactor Coolant leak into RBCCW via the NRHX l
- c. PSW 1eak in the RBCCW Heat Exchanger +
- d. RBCCW Fill Valva (F054) leakage into the RBCCW Surge Tank. ;
i F 0 1 23456789 : 1 i h I l < f
- i
. r i t a i I l ! (84*** CATEGORY 07 CONTINUER ON NEX T PAGE *****) (
- ~Zs__EB99E998gS_:_gggd86,_8pNQBd86t_gdgBQgNgy_8UQ PAGE. 36 ! 689196991986_G9BI696 : 1 {
; QUESTION 7.02 (1.50) f f
i ANSWER the following questions concerning the Condenser l Tube Leak abnormal operating procedure 34AD-OPS-028-2. ! (UNIT 2) ! I
- a. LIST TWO (2) ANNUNCIATOR ALARMS that would indicate a (.5)
, condenner tube leak. ! a
- b. If feedwater conductivity cannot be maintained less ,
than .2 umhos/cm, your immediate actions should be (1.0) [ 9 to (Select one answer) t i j
- 1. Reduce load as necessary until affected [
Hotwell Water Box is isolated. l 2. Perform a normal shutdown within O hours.
- 3. Scram the Reactor and trip the RFPTs, 4 Dooster and Condensate Pumps.
i
- 4. Recuce power below 30 % be in a Hot Standby
, Condition within 24 hours if the leak cannot be isolated. i l i t l l l l t (84*84 CATEGORY 07 CONTINUED ON NEXT PAGE *****)
4 j Zs__EBgGEQUBSD_ _Ug8dOb2_6QUQBd861_Edg8GENQY_6ND PAGE 37-4 BOD 196991GOL.GOUIB96 I i f. l 1 1 i QUESTION 7.03 (2.00) 3 ANSWER EACH ONE of the following questions TRUE or FALSE l concerning the Rod Soquence Control Systems
- a. The Group Notch Control Page self-test function is performod prior to reactor startup and before reducing i r err.t or power below 30% .
l b. RSCS rostricts rod movement to four (4) notches f rom all i other rods in selected group from 50 X rod density to a j preuct power level, 1 l c. When an operator preparcs to withdraw control rods i for reactor startup the Sequence Mode Selector Switch l; must be set to NORMAL.
- d. RSCS stops enforcing group notch constraints at 30%
powers all rods in the selected group remain backlighted i even when all rodu are at the same notch position. I t***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) _ _ . __ _ . .__- . .n
Zi_ EB9GEsWBE!L:_U9BDh 6DNQBUL,,gdgBQgNQLGNQ PAGE 38 I 809196991GOL_GQNIBQL { t t
?
QUESTION 7.04 (1.00) '. t I Determine if the following statements are TRUE or fat.SE. j
- a. Control Room computer printouts may be used for scratch paper >
once the operator has determined the printout data is of no i importance.
- b. Chart recordern are required to be checked within one hour of j shift change and marked with dato, time. and operator initials. j P
t i i l t I i I P 5 ( I l i i i f (4 *** CATEGORY 07 CONTINUED ON NEXT PAGE a $s) ; i I t l 1
~2A__BBggEggBES_r_NgBd863_8RUgBd861_gdgBggNgy_8bp EAGE 39 609196991G06_G9 NIB 96 OUESTION 7.05 (2.50) 4
^ ANSWER EACH ONE of the following with regards to the l Primary Containment on UNIT ONE: 1
- a. During a reactor. plant startup, WHEN must the Oxygen concentration be less than 4% 7 (1.0)
- b. Upon increasing temperature of the suppression pool, STATE
- the temperature at which a Technical Specification Limiting 1 Condition for Operation is first entered. (.5)
- c. The Reactor shall-be scrammed if Suppression Pool temperature reaches ___7____ . (.5)
- d. During Reactor isolation conditions, the reactor pressure vessel shall be depressurized to lean than 200 psig at normal cooldown rates if the suppression pool temperature i reaches ___?___ . (.5)
I i i 4 i 1 i (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
I Zo__EBQCEDUBEG_:_NQBdebt 0DNQBdGLt_EdEBGEugy_6NQ PAGE 49 809196901G06_G9 BIB 96 OUESTION 7.06 (2.50)
- a. LIS1 FOUR (4) separato visual indications used to confirm (2.0) correct orientation of a fuel bundio in the core.
- b. Driefly DESCRIBE how a bundic is checked for proper seating (.5) onto the fuel support picco.
(***** CATEGORY 07 CONTINUED ON NE)T PAGE *****)
Zt_ _ E B99 E 09 B E S_:_ NQ 8 d 863_8 D NQ 8 d 86, _ E d g 8 Q E N Cy._8 N D PAGE 41 BSD196901966_G9dI696 M M QUESTION 7.07 (1.00) ^ The EOP Flowcharts direct an operator to the appropri ate End Path Manual following any reactor scram. LIST TWO (2) canus (exceptions) where flowcharts direct the operator to a 4 Normal Operating-Proceduro. J d 4 I i i i 1 4 i 1 4 l (*8644 CATEGORY 07 CONTINUED ON NEXT PAGE *****)
. '2t__E89GEDVB69_ _U98d66t_GENQBM@(t_gdgBQgdgy_GNQ PAGE 42 B8919699190L_G9NI696 -l 1 QUESTION ?.08 (1.50) Excluding Reactor Power decreasu. LIST SIX (6) diffurent indications that should bo observed in the control room 2 following SLC initiation. I 4 s e i 8 1 0 i 1 (:**44 CATEGORY 07 CONTINUED ON NEXT PAGE 8**64)
i ! 2a__BS99EDUSES_:_N980061_0hu9BdeLi_EDEB9EBGY_609 PAGE 43 l 0 . BOD 196991GGL G9dIB96 t i i I
- i i !
s 3 i OUESTION 7.09 (2.00) In accordance with UNIT TWO procedure 34SO-D21-00102S,"Automatic . Deprousurization System and Low-Low Set Systum", the ADS may be ; initiated manually onl y if four conditions exist. LIST these FOUR (4) CONDITIONS. 4 i i ! 4 I 4 i l l (***** CATEGORY D7 CONTINUED ON NEXT PAGE *****)
3 yv . _ _ _ _ - _ _ -_____._____ _ _ _ _ _ _ _ Zs._ EB9GE99BES diUGBtf6Ld6pgirites(_geggggggy ggg pngg 44 689196991GOLG9dI69b' ; l
)
1 -QUESTION 7,10 ( 1, gkt; s STATE the TWO (2) alternative motheids of scramming the Reactor if a Manual Scram was not pouanble PRIOR to evacuation of the control room. 4 4 4 (t**e* CATEGORY 07 CONTINUED ON NEAT PAGE :$*48)
, . a i Z2__EB9CEDUBES - NORMALt_GDNQBtiSL1_g(jggggNpy_GNp PAGE 45 :
BGD196991GOL_G9N1896 l i d ! i ! l } QUESTION 7.11 (1.50) l 4 i
- I j LIST THREE (3) INDICATIONS which would be observable at the Rod [
Control Fanel upon failure of the RPIS while at 20% power. l l . 5 a ,
- i. . ,
I i I i i f 1 i
- 1 I
i 5 i . ( ! l 3 i I i i i I l i L I t (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) I I r I
-. .- . - . - -_ . - - - - - . _ - . _ .--.--.-s
Zt__E699EDUBES_:_U9BdOL ,_0DN9Bd86t_EdE69ENGLONQ PAGE 46 . BOR196991GOL_GONI696 i I J, QUESTION 7.12 (2.00) LIST the IMMEDIATE OPERATOR ACTIONS for a losu of Plant 1 Instrument and Service Air (34AD-OPS-020-2/15). If f differences exist between the proceduros of UNIT 1 and , - UNIT 2, be sure to includo actions for DOTH UNITS. {
; l t
i i
! I l .
t 1 ! i 4 1 i' i i ! i i I e I r
?
i f I f, I I t l i L , ! I I i i l ! I I ! i i I I l , f . r l tsas41 CA'IEGORY 0/ CONTINUED ON NEAT P4GE 44444) ! t i t i
Z,._ _EB99 E DUBE@,,r _ d Q Bd 8bi_8 D NQ B d861_ EM EB Q ENG y _89 Q PAGE 47
;B89196991G86_G9 BIB 96-I t vi-l 1
. \
' QUESTION 7.13 (1.00) l l
i EXPLAIN'what requirements must'be met to close-out an LCO. j 1 i l 4 4 e i 4 (***** CATEGOf ' 07 CONTINUED ON NEXT PAGE *****) i
. . - - . ~ - ~ ~ ~ . - . - . - . _ . . - - . . . . . . . . . . . . . .
'Zs__EBQG50UBES_:_NQBd86t_8BNQBd86t_Edq8QgNQY_8UQ- PAGE 48
- BODIQ6991G86_G9BI696 L
L QUESTION' 7.14- ( 1. 50) . ANSWER the fellowing questions concerning the Reactor Water Cleanep operating procedure (34S0-631-003-28).
- a. STATE WHEN'(under what conditions) the RWCU system can be 1
' operated without CRD Seal Purge flow. (.5)
- b. EXPLAIN WHY the !%mineralizer Bypass. Valve (2G31-F044'i j must remain throttled open during coeration of the (1.0)
RWCU system. (Assume a normal system lineup with the Reactor operating.at power.) J l, ) i l
~
l 5 l l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
'7. 'PRUCEDURES - NORMALx_ABNQRMAL 1_EME8QENCY_6NQ ~PAGE 49
-8801969QIC6L_CQNIBQL QUESTION 7.15 (1.00)
With the Reactor operating at 80 % power a single (one) Recirculation pump TRIPS. The affected pumps discharge valve isi required 1 to oe closed by the subsequent operator actions of'34AB-OPS-032-2S. This same valve is then required
~
to-be throttled open within 5 MINUTES of the time it went closed. ANSWER the following questions:
- a. EXPLAIN the reason for initially closing the pump Discharge Valve.
- b. EXPLAIN WHY the Dischar ge Valve is required to be Throttled open within 5 minutes.
W-1 i t i l l i (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) [ 1 l 1 ;
7.-' PROCEDURES -> NORMAL _@@NQBM861_gME6GENQY_8NQ 1 PAGE 50 B89196991986_G9 BIB 96 l l [ QUESTION 7.16 (2.50) For EACH ONE of the following sets of plant conditions, IDENTIFY the EOP' flowpath (BY NUMBER) specifically. designed to address those conditions.. ,_ a. Reactor transients or failure of vital equipment while in hot standby of startup.
- b. High. radiation, loss of vital power, failure of vital eqLipment, i or stuck open relief valve.
- c. High radiation, loss of coolant, and loss of primary containment integrity.
- d. Reactor transients or failure of vital equipment-while in the RUN mode.
- e. Failure of reactivity control systems.
I r i J t I i l l f l
- (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) {
i l
l Zt__EBRGEQUBES_:_NQ6d@65_8BNQBd@6t_Edg8QENQy_8Np PAGE 51- l 88Dlg6Q@lQ86_QQUIBQL 1 QUESTION 7.17 (2.50) , ,1 a STATE the exposure rate limits (f or a major portion of'the body) which characterize EACH ONE of the following: (1.5)
- 1. RADIATION (4REA
- 2. HIGH RADIATION AREA
- 3. LOCKED HIGH RADIATION AREA
- b. STATE the definition of extremities as it pertains to (1.0) radiation exposure of personnel.
L F i i
+
k ' p (***** END OF CATEGORY 07 * * * * * ) - (************* END OF EXAMINATION * * * * * * *********) l t r 1
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g, E 5 *: I 2 8,,3 A l 0 as l A o o l l PASTER RECIRC CONTROLLER FAILS TO PAXIMUM FIGURE 2 - 4 .- t
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, 1
.s EQUATION SHEET s
f = ma v = s/t v = ns s = v,c + ac 2 Cycle efficiency = Ne ut) - E = aC
- a = (vg - v,) / c KZ = mv 2 v g = v, + at A = AN A = A,e E
l PE = agh w = e/g A = In 2/tg = 0.693/tg , W = v&P-(e,,)(q) AE = 931 Am
- h (g+t) b 6=iC,at , t . 2 ,-Ix
~, k = UAAT g ._y -Vx
~ Pwr = Wf ' a I=I 10'*
- P=P 105UR(t). T7L = 1.3/u et/T P=P HVL
- 0.693/u
'SUR = 26.06/7 ~
T = 1.44 DT SCR = S/(1 - K,gg) o\ SUR = 26 f- A g {f CR x = S/(1 - K,gg,) T = '(**/o ) + [(i ~ o)/A,ggo ] 1
~
*ff 1
- 2 0 K eff T2
~~
T,= t*/ (, _ ;; M = 1/(1 - K,gg) = Ca t/CR n
~' eff M = (1 - x,gg) /(1 - x,gg)
P"I aff'I) eff * #eff eff /E Sp3 . (1 - K,gg)/X,gg a= (1*/TKyff -) + [I/(1 + A,gg7 )] 1* = 1 x 10"' seconds P = I$V/(3 x 1010) A,gf A= 0.1 seconds
-I E = No -
Idgg=Id22 WATER PARAMETERS Id =Id2 g 1 gal. = 8.345 lba R/hr = (0.5 CE)/d (meters) 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) - 1 ft = 7.48 gal. Mf SCELL\NEOUS CONVERSIONS , Density = 62.4 lbm/f t 3 1 Curia = 3.7 x 10 dps O Density = 1 gn/cm 1 kg = 2.21 lba Heat of var orizationi = 970 Itu/lbm I hp = 2.54 x 103 BTV/hr Heat of fusien = 144 stu/lbs 1 N = 3.41 x 106 stu/hr M' ' 1' fed = 14f psi = 29.'9 in. I g. 1 Blu = 778 f t-lbf
. ,.., .w -.
y gg,se g glg33 l'Ibf/in2* r ' 1 inch = 2.54 cm y .- T = 9/5'c + 32 . -
'C = 3/9 ('T - ,,32)
. - - - . - - . , , - , - , , . , - . - - - . - - , _ . . _ - . - _ - - , - - _ . , . ~ . , . - . - . . - _ . - . . - . - , . . - . - - . - , , - . . --
O H 0.9 pm A o
., H
;-*- - II A 0.1 p m A
MOTOR 0.6 on A - L F5
~
DWEUS .. y (CONTROLLED LE AEAGE . go SECOND SEAL (NO. 2) AT 0.7 6 pm) y
,, gg F1 i r
- DWEDS FIRST SEAL (NO.1)
T -- >> CONTROLLED PRESSURR SEAL PURGE mRuxDow rROu CRD sTsTEM, gg (D471RNAL AT SEAL) %RkAEDOWN BUSHING l Il iI l PUMP l . FAILURE OF SEAL NO 1 ONLY: NO. 2 SEAL PRESSURE WOULD APPROACH NO. . SEAL PRESSURE. LEAEAGE THRU NO. 2 ORIFICE WILL GO TO == 1.1 spa AND F3 "A" l WILL ALARM H3 A7 > 1.9 mm. FAILURE OF $E AL NO. 2 ONLY: NO. 2 SEAL PRE.55URT WOULD DROP DEFENDENT UPON WAGN17VDE OF FAILURE. LEAEAGE THRU FS "1" WOULD EXCEED 0.1 spa AND ALARM E1. FAILURE OF BOTH $EALS: TOTAL LEAZAGE OUT OF THE SEAL ASSEMBLY WOULD APPROACH
, 60 cpm AS LIWFED BY THE BREAIDOWN BUSHD4G 30TH FS "A" AND F3 "1" WOULD AL ARM NIGH. PRE.S$URE D4 BOTH SEALS WOULD DROP DEF ENDING UPON W AGN17UDE OF FAILVT.C. (NO.1 PRES $URE M1GHT NOT DROP SIGNIFICANTLY UNLE.55 FAILURE WAS LARGL.)
PLUGGING OF No.1 D4 TERN AL "10": NO. 2 PRESSURE WOULD GO TOWARD EERO AND FLOW THRU F5 "A" WOULD APPROACH ZERO AND ALARM LOW AT 0.3 sp m.
< PLUGGING OF NO. 2 TNTERN AL "RO": NO. 2 SEAL PRES $URE WOULD APPROACH NO.1 SEAL P2E.55URE.
CONTROLLED LEAEAGE WOULD APPROACH ZERO AND ALARM LOW AT 0.1 spa. l l ! = 1
- t""
I R %ore (3) _
.t '. 1 - ..n
8 8 8 8 8 i ' ' START UP RATE 4 - 4 d/ds 1 [ q ,
" e TURDI JE 'A 1 TURBINE TRIP SPEED
^
_E , ERR R INTERCEPT l' gE SP EO SELECT LVg ;
.1 g
g g g g ;- VALVE
% FLO5V CV REO TURBINE ~ i ,
E .IV REO SPEED 8 -- SPEED LOAD REJECT " *L OF ST ATOR COOLINO
- d/4 + I f -
ERROR o , !p g__ R 2% " ^tN Ml u. W CV REGULATIOtA / e
; REMDTE INC/DEC StGt1AL 2 + 1 e ST ART UP 7I O u N IN Ads)
^ MASTER FLOW
% " +Ih tuC CONTROLLER % f P
[(O .6 LOAD SELECTOR w MANUAL x- 4 AUT\ g& h TO RECIRC FLOW c CONTROL 1 ,
- HUNBACK ON LOAD REJECT (TO ZERO IN -45 SEC.)
GAIN 8.
- H I:NB ACK, SYNC SPEED NOT SELECIED T
STEAM y NO.2 STOP e { TilROTTLE ~ 4 E FROM SPEED Ny m VALVE
~ - SELECTOR OEMANO n PRESSURE -
A I g'fs TURBINE TRIPPED! CHEST g. 1 ' PRESS F-0 I WA"M'"G CONTROL
" uOT V e VO,
* ][gfag g
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+ gT PRESSURE SET DE lf p.
1 l*--*10 pal BIAS J % FLOW PRESS. RE G. LOAD l MAxtuuMI ggg,,ggg FLO" i
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STOP VALVES STEAM )+ OICLOSED I BYFASS TilROTTLE PRESSURE l ,, IT~ M
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V Al.VE DEMAND LOW VACUUM 4 SMALL BYPASS ~ CLOSE JACK BIAS g, O- *
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- I
- b D*4
- 5 t__ISEQBY_QE_NyGLE88_EQWEB_EL8MI_QEEB811QNt_ELy1QSt _88Q PAGE 52.
LIUEBd90XU001GS ANSWERS -- HATCH 1&2 -88/02/08-HOPPER, G.
. ANSWER 5.01 (2.00)
- a. True -(.5 each)
- b. True c.-False
- d. True REFERENCE G.E. REACTOR THEORY. Chpt 4, LO 1.5,3.6, Chpt 7, LO 5.6,8.3 3.1/3.2 , 3.3/3.4', 3.5/3.6 292008K119. 292008K120 292008K122 ...(;A'S)
K ANSWER 5.02 (2.00)
- a. Falte
- b. True
- c. True
- d. True (.5 each)
REFERENCE G.E. Reactor Theory, Chpt. 4, LO 1.5,3.6,4.3.6.3, Chpt 7, LO 5.6 2.5/2.6 , 2.1/2.2 , 2.5/2.6 , 1.9/2.12 292004K102 292004K109 292004K111 292004K113 ...(KA'S) ANSWER 5.03 (2.00)
- a. Ttue
- b. Falso c.. False j d. False (.5 each)
REFERENCE F.E. Reactor Theory, Chpt. 5, LO 7. 5 i 2.5/2.6 , 2.6/2.9 292005K109 292005K112 ...(KA'S) t
5 1__IBEgBy_gE_NgGLE86_EQWEB_ELONI_QEEB8IIQNz_ELgIQSt_6NQ PAGE 53 IbEBdgDYN8dlGS ANSWERS -- HATCH 1&2 -08/02/08-HOPPER, G. ANSWER 5.04 (2.00)
- a. True
- b. False
- c. True
- d. True (.5 each)
REFERENCE G.E. HTFF, Chpt 8, LO 9.1,9.2 9.4,9.5 2.4/2.6 293008K133 ...(KA*S) ANSWER 5.05 (2.00)
- a. 7
- b. 4
- c. 6
- d. 2 (.5 each)
REFERENCE G.E. Reartor Theory, Chpt 1, 1.0 4.1,5.1.6.1 2.7/2.6 , 3.2/3.5 , 2.4/2.6 , 5.2/3.3 292002K108 292002K109 292002K110 292002K111 ...(KA'S) ANSWER 5.06 (1.00) -
- a. TPF (4)
- 6. APLHGk (1)
- c. MCPR (2)
- d. DNB (7) (.25 each)
REFERENCE G.E. HTFF,Chpt 8. LO 2.7, Chpt 9, LO 1.1,4.1,5.1 l 2.2/2.6 , 3.3/3.7 , 2.8/3.6 , 3.0/3.2 l 293008K109 293009K104 293009K110 293009K119 ...(KA'S)
Dz__IUE96Y_gE_NygLg88_EQWEB_ELONI_QEEB9IlgN t_ELQ1QS x_8UQ PAGE 54 IUEBdQQyN8 digs ANSWERS -- HATCH 1&2 -88/02/08-HOPPER, G. ANSWER 5.07- (1.00)
- a. radiation
- b. convection
- c. kinetic energy
- d. temperature gradient (^ T) (.25 each)
REFERENCE G.E. HTFF, Chpt 7, LO 1.1.1.4,1.5 3.2/3.2 293007K101 ...(KA'S) ANSWER 5.08 (2.00) a.(Increasing power), stable positive period, no rod motion (.5 each)
- b. 3 (1.0)
REFERENCE G.E. Reactor. Theory, Chpt. 7, LO 2.2,9.1 EIH: 34GO-OPS-001-2 4.3/4.3 292008K105 ...(KA'S) ANSWER 5.09 (1.50) 1-2 Pumps (condennate, booster,feedwater) 2-3 Heaters and Reactor 3-4 High pressure turbine 4-5 Moisture Separator HP Turbine Exhaust 5-6 Low Pressure Turbine ( 6-1 Condenser (.25 each) ! REFERENCE l G.E. HTFF, Chpt 5, LO 6.1 l 2.0/2.2 293002K105 ...(KA'S) l l l l l l
Qu__IHEORY'OF NUCLEAR POWER-PLANT OPERATIONt _FLQIDQs_8ND' PAGE 55 ISE8dQQXN8d[CS ANSWERS - HATCH 1&2' -88/02/08-HOPPER, G. ANSWER 5.10 (1.50)
- 1. Fission product poison buildup.
- 2. Burnable poisons depletion is greater than fuel burnout.
(FP buildup is approx. complete)
- 3. Fuel depletion overcomes the effect of poison burnout. (.5 each)
REFERENCE G.E. Reactor Theory, Chpt 5, Chpt 6, LO 4.3 2.4/2.7 292007K103 ...(KA'S) ANSWER 5.11 (1.50)
- 1) Reactivity effects of Peak Xenon during startup. (.25)
Increased neutron flux during startup will cause an increase in the burnup of Xenon. Depletion of Xenon's poison effects will lead to a positive reactivity insertion. (.50)
- 2) Abnormal flux distribution. (High worth peripheral rods) (.25)
Xenon concentration will be highest in regions of the core where neutron flux was the highest during the. previous operational phase. Thermal neutron flux will be depressed in this region, and pushed toward prev'nusly Icw power regions (.25). (Peak thermal neutron flux . .. ally occurs in the central regions of the core-during startup.) High Xenon concentration causes the flux to be pushed to the periphery of the core, where previously low worth control rods now become high-worth control rods (.25).
- REFERENCE G.E. Reactor Theory, Chpt 6, LO 2.5,2.7 3.1/3.2 292006K114 ...(KA'S) i l
l I l l r
Uz__INE0BY_0E_NWGLE88_EQWE8_EL8NI_QEE68I1QNt _ELQlDSz_8dD PAGE 56 IHEBdQDYU8dlGS H ANSWERS'-- HATCH 1&2 -88/02/08-HOPPER, G. ANSWER 5.12 (1.00'
- a. (The piping system would undergoe a fluid shock caused by the rapid change in flow) - FLUID HAMMER (water hammer)(.5) - which could result'in extensive damage to the system.(.25)
- b. YES. (.25)
REFERENCE G.E. HTFF Chpt 6, LO 8.1 3.2/3.3 293006K105 ...(KA*S) ANSWER 5.13 (2.00)
- a. Increase due to pressure increase
- b. Decrease due to RPT trip caused by Turbine Trip above 30'/. power. (.5 each)
- c. Decrease due to void collapse on pressure increase.
- d. Increase due to level drop (on scram).
REFERENCE EIH: LP 200.1, VOL 8 CH 10.4 3.8/3.9 3.7/3.8 3.5/3.6 295005AA20 ...(KA'S) ANSWER 5.14 (2.00)
- a. Decrease due to Scram (.5)
- b. Decrease due to rush of Downcomer water into Recirc Suction (.5) c.. Decrease due to Speed Limiter # 1 on feedflow decrease to less than 20 */. (.5)
- d. Decrease due to scram (.15) and EHC controlling reactor pressure (.35)
- REFERENCE EIH: LP 200.1, VOL 8 CH 10.4 3.8/4.0 3.3/3.6 4.1/4.1 3.5/3.6 202001A205 202001K102 202001K122 202001K416 ...(KA'S) 1 i
5 l
---.e.+w 9-.rw------.+y
5 t__IHgQBy_QE_Nyg6g88_EQWEB_ELONI_QEEBBIlgN1_E6pIpgi_8ND -PAGE 57 IEEBdQDXU8 DIGS ANSWERS -- HATCH i t<2 -88/02/08-HOPPER, G. ANSWER 5.15 (1.00) S/1-Keff 100/(1 .2) = 125 or fission source total 0 0 100 100 1 20 100 120 2 24 100 124 3 25 100 125 4 25 100 125 REFERENCE-G.E. Reactor. Theory, Chpt 3, LO1.2,1.5 2.9/3.0 , 2.1/2.3 292003K101 292003K102 ...(KA*S) ANSWER 5.16 (1.50) a.. 800 psig = 814.7 psia = 520.3 deg F (.25) 350 psig = 364.7 psia = 435.5 deg F (.25) 520.3 - 435.5.= 84.8 deg F (.2b) c/d rate = C 84.8 deg/ 45 mi n 3460 min /hr = 113.. deg / hr (.25)
- b. 100 deg F / HR (.5)
REFERENCE Steam Tables 2.8/3.1 293003K123 ...(KA'S) I ANSWER 5.17 (1.00)
- a. NPSH is defined as the difference between total pressure at the eye of a pump (or inlet of a valve) and saturation pressure, or NPSH= Pi - Psat (.5)
- b. Cavitation would result in vibration (and noise) of the pump and pitting and corrosion of the pump parts,(especially the impeller ). (.5) i i
=5t__IBE98Y_9E_NUCLg68_EQWEB_E68NI_QEg6811QNt_ELUIDQ1_GNR PAGE 58 ISEBdQQyN8d1QS ANSWERS -- HATCH 1&2 -88/02/08-HOPPER, G.
REFERENCE-G.E. HTFF, Chpt 6, LO 10.0,10.9 2.7/2.8 293006K110 ...(KA'S) N e 4 i t t i h q 4 I o i r o i i I L
. . - . , , ,._... ._._-.-,__.-.___.,_-...,,._,.,.....m_._,_-..._ .._.,._.,,,.,,,_.,_._.,,,,,,,._.___.___.,___.,__,,..____._,_,__...,._.__!-
hs__ELONI_Sy@IEd@_QE@lgN t _CgNIBQ6t_@NQ_lNSIBUdENI8IlgN PAGE 59 U- ANSWERS -- HATCH 1&2 -88/02/08-HOPPER, G. ' ANSWER 6.01 (1.00) C REFERENCE EIH: VOL 5 CH~3.3, LP 30.1 LO 4 3.0/3.1 3.8/3.8 . 261000A205 288000A301 ...(KA'S) h ANSWER 6.02 (1.00) , d~ REFT?RENCE BFNP LP#7,P. 28 E7.H VOL 5 CH 4.1, LP 4.1 LO 24 i GGNS SD B33-1, pp 5, 6; OP-B33-1-501, p 5; ARI B33-FAL-L603A 3.3/3.3 202001A109 . . . ( JK A ' S ) ANSWER 6.03 (1.00) f d I- REFERENCE EIH: VOL 6 CH 7.4, LP 27.1 LO 20 3.6/3.9 3.4/3.7 262001A401 262001K406 ...(KA'S) ANSWER 6.04 (1.50) 4
- a. False l b. True
- c. False (.5 each) i REFERENCE EIH: LT-IH 03801-01 EO 4.14,6 3.4/3.6 4.2/4.3 3.7/3.9 218000G001 218000K401 210000K606 ...(KA'S) i
JA1__ELONI_SygIEd@_DEg1GN i _CgN18961_6ND_lNB189dgNIBIlgN ~PAGE- 60 ANSWERS --) HATCH 1&2 -88/02/08-HOPPER, G. i ANSWER -6.05 (2.25)
- a. Cond. Booster Pumps - 1,4
' Reactor. Feed Pumps - 2,5 Cond. Pumps - 6 (.25 each)
CRD Pumps .3,7
- b. Prevents overloading of the startup transformer (2D) when the ECCS pumps are auto starting. (.5)
REFERENCE EIH: VOL 6 CH 5.3, VOL 5 CH 4.2, LP 2.1 LO 4,7,8 LP 1.1 LO 7 4 2.8/2.8 3.3/3.3 3.4/3.4 3.2/3.3 256000K401 256000A201 256000K302 256000K403 ...(KA'S) ANSWER 6.06 ( .50)
- a. DELETED
- b. Negative. (.25)-
- c. Yes (.25)
REFERENCE BSdP SSM, Core Spray, P.8 Lesson Objective 9 EIH LT-IH-00801-00, 34AR-601-055-2, TO 200.64 3.0/3.2 2.4/2.6 2.8/3.0 209001K113 209001K404 209001V502 ...(KA'S) ANSWER 6.07 (1.00) Any combination of "Inop", "Downscale", or "Triple-HIgh" (.25 each) in both post treatment radiation monitor trip channels (.25). REFERENCE EIH: LP 31.1 LO 2, VOL 6 CH 6.8 , 3.1/3.3 i 271000K408 ...(KA'S) l
f65__ELONI_SYSIEdS_DESigNt_CQUI6QLt_60Q_INSI6QUENI8IlQN- PAGE 61 ANSWERS - . HATCH 1&2 -88/02/08-HOPPER, G. ANSWER 6.08 (2.00)
- a. Refuel Platform Grapple not full up.
' Fuel loaded on Refuel Platform Grapple Fuel loaded on Frame Hoist Fuel. loaded on Trolley Hoist Fuel Loaded on Service Platform Hoist Selection of a second rod for movement with any other rod r withdrawn L any 4 (.5 each) f REFERENCE
. EIH 'VOL 7.CH 9.2.1, VOL 6 CH 6.9, LP 1.2 LO 1, I 3.5/3.5 201002K402- ...(KA'S) ! I ANSWER 6.09 (2.00) Low lube oil ~ pressure I Engine.overspeed . Engine start failure Di f f erenti al Lrckout (.5 each) . REFERENCE EIH: VOL 6 CH 7.2, LP 26.1 LO 1 4.0/4.2- i 264000K402 ...(KA'S) , I i ANSWER 6.10 (2.00) [
- a. Low Nitrogen Pressure (of 965 pulg) (0.25) l '
l High Water Level in tne instrument block (of 60 ml) (0.25) t At f.%e local control panel, the back lit button must be depressed. , If te 2 light goes out, the cause i s water; if the light stays lit, , the cause is gas pressure. (0.5) l b. The CRD FCV is downstream of the flow element. (0.25) All of the indicated flow is going through the Charging line to recharge I the accumulators. (0.5) The sensed high flow is sending a signal t to close the FCV, (and thus Cooling Water flow is low). (0.25) l [ REFERENCE EIH: VOL 5 CH 4.2, LP 1.1 LO 10, 6 3.6/3.7 i I a j
- m. -
, . . __,,,.,m,-,-,_-..-_- . . . , _ _ _ _ _ . _ . , _ _ , . , _ _ _ , _ - _ _ . - _ _ , _ _ _ - . , _ _ , . , . ~ . . . . . _ . . _ . . - . - .
6,__eL8MI_@YSIEdg_QEgigNz_GQNIBQ6t_8NQ_lNSIBydENIGIlgN PAGE. 62 ANSWERS'-~ HATCH 162 -88/02/08-HOPPER, G. 201001G007 ...(KA'S) : ANSWER 6.11 (1.00)
-a. Reduce the concentration of oxygen to_ minimize the possibility of. hydrogen combustion following a LOCA. (1.0)
- 6. DELETED REFERENCE BFNP Lesson Plan 16, Objectives C, D, H. I, pp 1, 14, 26, & 32 E1H: LP 13.1 LO 4.20 3.7/3.8 223002G007 ...(KA'S)
ANSWER 6.12 (2.00) Reactor Scram Group I Isolation Main control room ventilation swaps to pressurization mode Mechanical vacuum pump trips and isolates ; Gland seal exhauster trips (and isolates) (4 @ .5 eac7) REFERENCE EIH: VOL 7 CH 9.7, LP 14.1 LO 2 3.6/3.9 3.7/4.1 3.8/3.9 272000A301 272000K402 272000K403 ...(KA'S) f-1
bi__E @@I_SYSIEdS_DE@lgN2_ggNIB9hx_6MD_INSIBudENI6IIgN' PAGE'~ 63-ANSWERS - HATCH 1&2 -88/02/08-HOPPER, G. ANSWER' 6.13 ' (3. 00)
- a. Power <27% UNIT 2 '
Selection of a peripheral rod Manual operation of - the RBM BYPASS joystick on P603 (.5 each)
- 6. One RBM channel will.use two level "B" and "D" LPRM detectors and all four level "C" detectors. (.5) The other channel ,
will use the remaining two "B" and "D" level detectors and the same four level "C" detectors. (.5)
- c. The level "A" detectors are located at the bottom of the core.
where transition boiling is not expected to occur.(.5) (Since the RBM protects against transition boiling, "A" level detectors are not necessary). REFERENCE EIH: VOL 7 CH 9.1.3, LP-12.3 LO 13 3.8/3.8 3.2/3.1 2.9/3.0 3.6/3.5 215002A304 215002G007 215002K102 215002K403 ...(KA'S) ANSWER 6.14 (3.00) l INITIAL RESPONSE: I a.'TCVs - Remain at 100% open
- b. BPVu - Open (16.5%)
- c. Power - Decreases i d. Pressure - Dec enses (.25 each)
! REASON: Above ( 3 .cc t') PCU calling for approx. 115% steam flow (.5).. i (950 - 91;, - J. 3 ) FINAL STATUS:
- a. TCVs - 100% position
- b. BPVs - Shut
- c. Power - Slightly lower
- d. Pressure - Slightly lower (.25 each)
! REASON: Above caused by the decrease in pressure and power causing i BPVs to shut -- PCU cycling to new equilibrium state. (.5) i ((945-915) x 3.3) i i REFERENCE j EIH: VOL 7 CH 9.4, LP 19.1 LO 2,5,10 3.4/3.4- 3.9/3.8 4.1/3.9 3.8/3.9 l 241000A101 2410004102 241000A114 241000K102 ...(KA'S) l l l l 5 l t
st__ELONI_@y@lEMS_QE@lGN z _CQUIBQ61_8ND_IUSIBydENI8IlON PAGE. 64
. ANSWERS -- HATCH 1&2 -88/02/08-HOPPER, G.
' ANSWER 6.15 (1.00) l NO. ('25) .
The condenser would lose vacuum due to accummulation of f noncondensible gases and-could possibly overpressurize. -(.75) REFERENCE : EIH: . VOL-5 CH 4.5, LP 39.1 LO 3 ; 2.9/3.0 3.2/3.5 ' 217000A200 217000K405 ...(KA'S) l ANSWER 6.16 (1.00) , s. The MSIV-LCS controls and minimizes the releawe of ' l fission products (which could leak through the closed MSIV'S and bypass SBGT) following a LOCA. (.5) r J
- b. Unit One does not have this system due to strict limits on MSIV leakage ( 11. 5 SCFH per^ valve) (which meet 10 CFR off-site dose limits) . (.5) i REFERENCE j EIH LP 49.1 LO 1,9 !
3.2/3.3 2.6/3.7 1 239003G004 239003G006 ...(KA'S) ) i i, ,
?
t i- ,
- i. '
l i i t a f 1 i !' h
-, o p.7 .,:r---w.'+3 -94,-,p.,.,e9%m,p.m . e % .-y,- ,e---yy. ,,,7w.e9__yy_,r_,_g-gm,. ,,-y., wpm.-m_ ,.,._-_,,r,epc,,r.-w.c ww, vwmc.---,e- ,w-- ,.
- _ . . . - _ ~
N Zz__eBQCEQQBES_ _NgBde61_epugad861_EDEBQENCy_8NQ PAGE 65 809196991 COL _CQUIBQL ANSWERS -- HATCH 1&2 -88/02/08-HOPPER, G. i ANSWER 7.01 (1.00) C REFERENCE , EIH: LP 9.1 LO 5 s
-ANSWER 7.02 (1.50)
- a. Powdex System Trouble Contaminated Feedwater Alarm Inlet High Conductivity j Polishing Demin High Conductivity (any 2 9 .25 each)
- b. 3 (1.0)
REFERENCE EIH: LP 25.1 LO 4, 34AB-OPS-028-2N 3.7/3.8 3.1/3.9 2.8/3.1 1 i' 256000A215 256000G011 256000G015 ...(KA*S) i ANSWER 7.03 (2.00) i
- a. True
- b. Falso
- c. False (.5 each)
- d. True REFERENCE
, EIH: LT-IH-05402-01 EO 2.h,6a,7,10.c
- 3.5/3.2 3.1/3.1 3.5/3.7 3.9/3.9 201004A305 201004A402 201004G007 201004K103 ...(KP'S)
) ANSWER 7.04 (1.00) 1 ' a. False
- b. True (.5 each) i REFERENCE EIH 30AC-OPS-003-OS l 3.2/3.4 l'
l t
7 Z,__eggggDURES - NORMAL _ABNgRUAL t _EMERGENgy t 8NQ PAGE <6 6 ~ B001g69 Gig 86_GQNIBQL
= ANSWERS:-- HATCH 1&2 -88/02/08-HOPPER, G.
294001A106 ...(KA'S) ANSWER 7.05 (2.50)
- a. Wtthin 24 hours subsequent to placing the reactor in the RUN MODE (f ollowing a shutdown) . (1.0)
- b. 95 deg F (.5)
- c. 110 deg F (.5)
- d. 120 dog F (.5)
REFERENCE EIH Technical Specifications 3.7.A.5 , 3.7 3.3/4.1 223001G005 ...(KA'S) ANSWER 7.06 (2.50)
- a. Channel Fantener in center of cell Identification Lug points to center of cell Spacer Buttons adjacent to control rod Serial Numbers readable from center of cell Cell to Cell symmetry Location of Gad Rod end plugs (any 4 @ .5 each)
- b. A TV camera is mounted on the fuel grapple and lowered for visual inspection (.25) (to a posi tion slightly above the f uel bail).
.The fuel grapple is lowered just above the bail handles and the core is scanned to ensure proper seating as indicated by the fuel grapple not striking any bail handles (.25)
REFERENCE ' EIH: RO 1.5 LO 11, 42FH-ENG-012-2 , 3.0/3.7 3.1/3.3 1 234000G13 234000K505 ...(KA'S) 1 I
Zs__ESQCgggBES_ _NQBd862_8BNQBd863_EMEBGENCY_8NQ PAGE- 67 *
-86DIQ69GIC86_CQNIBQL ANSWERS -- HATCH 1&2 -88/02/08-HOPPER, G.
ANSWER- 7.07 (1.00)
- 1. Scram signal was received while performing a scheduled test with the reactor shutdown. (.5)
- 2. Scram was initiated as part of a normal fast reactor shutdown procedure. (.5)
S REFERENCE EIH: LT-IH-20101-00 LO 11 4.1/4.0 4 295006G005 ...(KA*S) ANSWER 7.08 (1.50)
- 1. Amber Squib-continuity lamp of both explosive valves extinguish.
- 2. Loss of continuity alarm annunciates.
- 3. RWCU outboard isolation valve closed i
- 4. Selected SLC pump start light illuminates. I
- 5. SLC discharge pressure increases.
- 6. SLC storage-tank level decreases. (.25 each) ,
t REFERENCE l EIH: MOD 1.3 EO 1.3.1.1 - 3.8/3.0 4.1/4.2 4.0/4.1 4.2/4.2 2 211000A303 211000A305 211000A306 211000A308 ...(KA'S) , ANSWER 7.09 (2.00) ,
- 1. ADS cannot initiate automatically (.5) l
- 2. Reactor water level < -101 inches (.35) and is NOT INCREASIN3(.15). !
- 3. Any low pressure injection system (Core Spray or RHR or Condensate system) is running (.5)
- 4. Reactor Pressure is > 350 poig (.5)
REFERENCE f EIH: 34S0-821-001-25. LT-IH-03801-01 EO 9.6 ! 3.5/3.6 3.8/4.0 4.'1/4.2 [ 210000A204 218000G010 218000K402 ...(KA'S) ! 4 l t l l I
2t__BBQQEQQBEQ_ _NQBd861,6BNQBd86t_EdgBGENQi_6NQ PAGE 68 BOD 196gGIC66_QQNIBQL ANSWERS -- HATCH 1&2 -88/02/08-HOPPER, G. ANSWER 7.10 (1.00) l
- 1. Open RPS Bus' Breakers (CB3A,B) to Power Range Monitoring (APRM SCRAM)
- 2. Trip Mercoid Switches on Scram Discharge Instrument Volume Level Switches (HI-HI SCRAM) (Two swi tches on each side of Reactor Building cause full scram.)
- 3. Close scram air header isolation valve and vent the scram header.
- 4. Vent each CRD mechanism through its associated above piston vent valve.
(any 2 G .5 each) REFERENCE
.I EIH: LP 10.1 LO 9, 34AB-OPS-008-1(2) 3.8/3.9 4.1/4.2 295016AA10 295016AK30 ...(KA'S)
ANSWER 7.11 (1.50)
- 1. Loss of Four Rod Di spl ay
- 2. Rod Drift Indicator Light for all rods
- 3. Select rod Block
- 4. RWM Rod Block annunciator (if RWM in services (any 3 G .5 each)
REFERENCE EIH: 34AB-OPS-024-2 3.0/3.7 3.1/4.0 3.5/3.1 3.5/3.3 214000A303 214000G008 214000G011 214000G012 ...(KA'S)
. : Zx__ EB9CE DUB ES_ :_N9Bd 862_0 @NQBd@L t_gd gBQg NCy_6N D PAGE. 69. 88k196991ceL_C96IB96
. ANSWERS -- HATCH 1&2 G. '
-88/02/08-HOPPER, ANSWER 7.12 (2.00)
- 1. If.a Reactor Scram occurs, enter the EDP's. (.5)
- 2. Manually Scram the_ Reactor if any of the following ,
! or.c ur s , then enter the EOP's. (.25)
- a. Indication of 4 or more control rods .
. drifting into core. (.25) '
- b. Scram Valva Pilot Air Header high/ low pressure ,
coincident with CRD HYD High Temp. (.25) f
- 3. Maintain reactor water level between 15 and 45 inches (.25) i a.Utill:e LCV bypass as necessary on UNIT 2 (.25)
- b. Cycle the Startup Flow Control Isolation Valve and Feedwater Low Flow Control Bypass j on UNIT 1 (.25)
- .F REFERENCE EIH: LT-IH-035W1-00 EO 16, 34AB-OPS-020-2 < 3.7/3.4 3.3/3.9 3.4/3.3 , 295019AK20 295019G005 295019G010 ...(KA'S) 3 r i ANSWER 7.13 (1.00) ] I a ' (When a system or component is restored to an acceptable operable I
- condition), the LCO can be cleared after all functional tests associated with each MWO (in the LCO) are confirmed to have been performed (.5) and the results are satisfactory. (.5)
REFERENCE EIH: MOD 2.1 EO 2.1.3.2 4 3.9/4.5 "940001K10 ...(KA'8) i i ANSWER 7.14 (1.50) f , a. Operation i s permitted if the CRD system is Out of Service (.25) and Reactor water temperature is less i than 212 deg F (.25). I b. This va' ave provides a pressure relief path for CRD water (.5) in the event of a system isolation in which both domineralizers transfer to the HOLD condition (.5). REFERENCE [ EIH 34SO-631-003-25 LP 3.1 1.0 7.10
- 3.2/3.2 2.8/2.0 2.7/2.9 3.4/3.4 i
-;Zs__EB9CERUBES_:1N98d86t_GBNQBd86t_E iNgyldNQ PAGE 70 i 809196901G06_G9dIB96
. ' ANSWERS -- HATCH 1&2 -88/02/08-HOPPER, G. 204000A213 204000G010 204000K116 204000K402 ...(KA*S) ANSWER 7.15 (1.00)
- a. Prevents reverse rotation of the affected pump. (.5) l
- b. ihermal Binding of the discharge valve may occur (.5)
(if the valve is not reopened within 5 minutes) .i REFERENCE , EIHa'34AB-OPS-032-28 3.2/3.2 3.6/3.7 3.5/3.7 202001A203 202001A223 202001G010 ...(KA'S) i.
-ANSWER- 7.16 (2.50)
- a. 2 -
- b. 4
- c. 5
- d. 3
- e. 1
(.5 each) . i i REFERENCE
.EIH: ! .T-I H-20101 -00 LO 7 l 3.8/4.4 3.9/4.5 3.9/4.5 3.9/4.5 l 1
295006G012 295024G012 295025G012 295031G012 ...(KA'S) > k ANSWER 7.17 (2.50) - t ! a.
- 1. >5 mrem in any one tour or 100 mrem in 5 consecutive days i 2. > 100 mrem in any one hour }
- 3. > 1000 mrem /hr !
(3 & .5 each)
- b. Hands and forearns, including the elbows, feet, ankles, and lower legs, including the kneen. (equivalent answer accepted) (1.0)
REFERENCE EIH: 60Ar-HPX-004-CE, 10CFR20 3.3/3.8 , 294001K10J ...(MA'S) l 1 i i a t b f
. _ . - . - - . - - - ,,-m,... , m ,_ _ . ,m.,m..~,.,__,-..,-___,_-..,___,-~.-,,,,_... --
TEST CRGSS REFERENCE PAGE 1 QUESTION VALUE : REFERENCE 05.01' 2.00 GTH0000666
-05.02 2.00 .GTH0000658 05.03 2.00 GTH0000661 05.04 2.00 GTH0000669-03.05 2.00 GTH0000656 05.06 1.00 GTH0000659 05.07 1.00 GTH0000667 05.08 2.00 GTHOOOO662 05.09 1.50 GTH0000664
-05.10 1.50 GTH0000660 i 05.11 1.50 -GTH0000663 05.12 1.00 GTH0000668 05.13 2.00 GiH0000671 05.14 2.00- GTH0000672
- 05.15 1.00 GTH0000657 05.16 1.50 GTH0000670 05.17 1.00 GTH000066S i 27.00 ;
06.01 1.00 GTH0000679 06.02 1.00 GTH0000673 06.03 1.00 GTH0000685 06.04 1.50 GTH0000678 06.05 2.25 GTH00006G1 06.06 .50 GTH00C0676 06.07 1.00 GTH0000677 06.08 2.00 GTH0000684 06.09 2.00 GTH0000688 ' 06.10 2.00 G'iH0000674 06.11 1.00. GTH0000675 06.12- 2.00 GTH0000680 ; 06.13 3.00 GTM0000603 06.14 3.00 GTH000068o 06.15 1.00 GTH0000687 06.16 1.00 GTH0000682 [ 25.25 i i 07.01 1.00 GTH0000698 , 07.02 1.50 GlH0000703 07.03 2.00 GTH0000699 07.04 1.00 GTH0000705 . D7.05 2.50 GTH0000695 07.06 2.50 GTH0000690 07.07 1.0G GTH0000691 07.08 1.50. GTH0000693 07.09 2.00 GTH0000694 07.10 1.00 GTH0000696 .
- 6Dr.11 1.50 31HD000697 ;
=
TEST: CROSS REFERENCE PAGE 2 "OUESTION VALUE REFERENCE ., - - - - - - ~ ~ . ------ .---------- 07.'12 2.00 GTH0000702 07.13 1.00 GTM0000692
-07.14' 1.50 GTH0000700 ,
07.15 1.00 GTH0000701 07.16 2.50 GTH0000704 07.17 2.f0 GTH0000689 28.00 ; 00.25 DOCKET NO 321 A 1 b i
- I
= S l r i
- l. I t
I c i 4 1 4 m . - - - - ,-.n,- - .
,,,o ,- -- ,,,--e-., ,, ,,, _ n v
.----,-g-, , -, - - - . - . ., -, , ,- -.-~-
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _HOTQH_1&2_______________ REACTOR TYPE: _DWR-GE4_________________ DATE ADMINISTERED _88202/08________________ EXAMINEk: ,PAyNEt_Cz_,,_____________ CANDIDATE: _ M k%_I_(_k_____________ INSIbUGIIOUS_IO_CONDIDGIE: 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. Ti. 2 passing grade requires at least 70% in each category and a final grade of ut least 60%. Examination papers will be picked up sin (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__YGLUE_ _I0106 ___EGOBE___ _206UE__ ______________GGIEGOBY_____________ A T. c a _e6rMF__ 19?.tgg ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 2$~ca _3br@F_ _ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. Condidate's Signature i i
'as__89d]NISIBBI12g_E8QGEQQBES1;GQNQ1IlgNS,_6NQ_LidlIBIlgNS PAGE 2 ,
4 : I
; OUESTION 8.01 (1.00)
Which ONE (1) of the following describes the correct method for re-establishing the required SRM vninimum count rate during Unit 2 core reload. *
- a. Spiral reload the core until 3 cps is established.
- b. Load up to four now fuel assemblies next to each of the four SRMs to obtain the required count rate.
- c. Prior to fuel. reload, install _ neutron sources in:the the source tubes to establish 3 cps.
- d. Place offloaded' fuel bundles (up to four) into their previous positions around each SRM until 3 cps is established k
4
- I a
e 4 4 1 (***** CATEGORY 09 CONTINUED ON NEXT PAGE *****)
Oz__0901NISIB8IlyE_EBQGEQUBEQx_GQNQlligNQt_@dQ_(ld1I8IlgNQ PAGE 3 l 5
+
QUESTION 8.02 (1.00) '
. Unit 1 is in Operational Condition I with NO ob: standing deficiencies. ,
i The Surveillance Assignment Sheet identifies tomorrow as the "Latest . Date" for the RCIC quarterly flow test. Prior to performing the i scheduled surveillance, RCIC becomes INOPERABLE. Which ONE (1) of the statements below' accurately describes the surveillance requir ements f or this situation.- ! i >
- a. The surveillance must be performed immediately after roturning the system to an OPERABLE status.
- b. A MISSED SURVEILLANCE SHEET must be initiated since the surveillance will not be completed today as scheduled. [
t
- c. The surveillance shall NOT be documented as officially missed ,
until the "Latest Date" has elapsed. ! d. Since RCIC is already INOPERABLE. a MISSED SURVEILLANCE SHEET ! neod not be issued to track the missed surveillance. i l i l l I (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l
a2__end1NisigeIIVg_esgcggyBgg1_ggep1Ilgggi_eyp_ Lid 1I8I1gNg PAGE 4 . QUESTION 8.03 (1.50) STATE the reporting requirements for EACH of the following situations:
- a. Unit i reactor scrams from a Main Turbino Stop Valve Fast Clonare, following a turbine trip. The HPCI system receives an auto
-initiatson signal on low reactor water level but fails to inject to the reactor pressure vessel due to isolating on a Main Steam Lino high dP isolation signal.
- b. While making a tour of the Unit i Southeast Diagonal, a Plant Equipment Operator slips on the stairs at the 110 foot l evel resulting in a compound fracture of his left leg. The man is subsequently transported to Appling General Hospital for treatment.
- c. With Unit 2 at full power at 2424 MWt, the functional test for MSIV closure procedure is being performed. While testing the "B MSIV Not Full Open" logic, it is found that a Division I circuit de-energized a Division II relay _and the Division Il logic de-energized a Division I relay, thus violating the divisional circuit separation criteria as addressed in the FSAR.
LIMIT YOUR RESPONSES TO LESS THAN 15 DAY REPORTING REQUIREMENTS. 4***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
e
- St _0DdlL41 SIB 8I1YE_EBQGEDUBEQu_QQNQlliQNSt_@dQ_Lidll@llQNS PAGE 5 l
f OUESTION 8.04 (1.50) ! i Concerning Temporary Procedure Chtnges: . [
- a. You are the Unit 2 Shift Supervisor on duty and an inscrument technician brings you the procedure for "Barksdale Pressure Switch y Cal i b r a t i on " . He points out a part of the procedure that he wants
- 4 you to approve as a temporary change. He i.. forms you that the work was just completed in accordance with the "changed", but not approved, procedure.
(1) STATE whether you can approve this TCN and the work it covers (YES or NO). (0.25) , (2) JUSTIFY your response. (0.50) j j +
- b. On August 19, the "PRIMARY CONTAINMENT ATMOSPHERE CONTROL SYSTEM"
- procedure was temporarily changed by the Shift Supervisor. On l August 24, a procedure revision request to the procedure, [
i consisting of the same items as the temporary changes, was reviewed by the PRB and approved and signed by the Plant Manager d on September 4. ) (1) STATE whether any problem exists (YES or NO) with the administrative processing of the changes to the "PRIMARY CONTAINMENT ATMOSPHERE CONTROL LYSTEM" procedure. (0.25) (2) JUSTIFY yoar response. (0.50) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
LQa 00dIU1HIBOI1YE_eBQGERLJBES _GQUQlIl0NSt_0NQ_LidlI6IlgNS. PAGE ~6 I i QUESTION O.05 (2.00) ; I \ 1- Following a TIP trace on Unit 2. the ball valve for the "C" drive > mechanium fails to auto close. Tho TIP machine requires extra jogging to close this valvo due to.a sticking limit switch. The ball valve is , subsequently closed. a- a. STATE whether Primary Containment Integrity is satisfied. JUSTIFY your response and STATE any ACTION required. (1.5) ;
- b. EXPLAIN any differences in your response if this situation l had occurred on Unit 1. (0.5)
E f r l a i . f' i t i i , i I i i l 1 'l r , 1 i ? t l' t i I l t I ; ! (***** CATEGORY 08 CONTINUED ON NEXT PAGE $4***) , l i 1 , r- i I i I _ _ , _ . . . , _ _ _ _ _ , , _ , . . . . _ , . _ . . _ . . . _ _ . . - . . , . _ _ . . _ . _ . . _ _ _ .,___.__,_.,.___...._.__..._______-_.i
es__6Dd1BISIBBI1YE_EBQGEQUBg5x_GQUpillON@t_8NQ_ Lid 11GI199@ PAGE 7 QUESTION 8.06 (1.00) You are the Shift Supervisor- while performing a normal reactor startup of Unit 2. At 10% power with the mode selector switch in STARTUP/ HOT STANDBY, the Rod Worth Minimizer is declared INOPERABLE. A second licensed op;rator is stationed at the reactor control console to verify compliance with the reouired rost sequence check off list'and the startup is continued to 100% power. EXPLAIN WHY Technical Specifications HAVE or HAVE NOT buen violated. w (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
_ _ _ _ , _ _ _ m._..
'OtJk6DbidlE16611ME_669CEQU669t_GQNDLIlgN$t_6NQ_LidlTATIONS PAriE ~8
- I I
i- !
- i I
.- 1 QUESTION 8.07 (3.00) !
l For'EACH of the following situationut I
- a. STATE whether the TS allown continued operation in Operational Condition 1.
.b. JUSTIFY the basis, per TS, for.your decision. !
NOTE: ASSUME ALL OTHER COMPONENTS ARE FULLY OPERABLE l' 1. A fuel oil transfer is dismantled for maintenance on Diesel l Generator 2A. ; i , L f'
- 2. It is reported that 18 Standby Li qui d Control Pump will not meet the minimum flow requirements per Technical Specifications.
1- - i 3. TWO (2) Suppression Pool-Drywell Vacuum Breakers in Unit 2 cannot .! be CLOSED.
- 4. The leads are lifted from the motor controller for the "A" Fuel Oil Transfer Pump on Diesel Generator-18.
t ( L t I (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) f
. . t
92__89dINISIB8I1YE_EB99EDUBEst_G9NDII19BSt_6ND_ Lid 11eligNS PAGE 9 QUESTION 8.08 (3.00) Unit 1 is operating at 48% power and 35% flow following a , Recirculation Pump trip. LIST by section and paragraph ALL ' requirements per TS that must be considered to permit continued operation.on a single loop. INCLUDE a summary of any 4CTION you muut initiate in order to comply with these requirements. 4 i e i i I
?
1 l l l (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l
al__GQd1NISIBBIl$E_E8QQEQUBg5g_GQNQlIlgNgt_8NQ_ Lid 1I611gNS PAGE 10 i
'l I
4 OUESTION 8.09 (2.00)- i Unit 1 is nearing the end of a major refueling and maintenance outage j .' that'has already overrun by two months. Significant pre-startup ! 1 _' testing of plant systems is planned for the coming week. Management intends to cover them through use of entonsive overtimo by the wtaff.
- As Shift Supervisor you receive the f ollowing work schedule f or one of your Plant Operators.
MONDAY 0700-1900 TUESDAY 0700-1700 WEDNESDAY 0700-2000 THURSDAY 0700-1700 FRIDAY 0000-0900 1700-2300 SATURDAY 0600-2300 SUNDAY - OFF - STATE the overtime restrictions, if any, that would be violated if the operator worked as scheduled and LIST the timeframe during which each violation would occur. 1 J 4 ) n 1 l 1 l 4 1 s (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
95__0Dd1NIglBOIIVg_PBQQgpyBg@g_QQNQlllQN@t_6NQ_61d1I611QN5 PAGE 11-QUESTION O.10 12.00) [ i Wnile operating at 100% power, a Group I isolation and reactor SCRAM occur on Unit 1. Data collected from the plant Process Computer and [ the plant operators indicate the f ollowing occurred (1) The Group I isolation was caused by technician error. , i
'(2) The reactor SCRAM was caused by high reactor pressure.
(3) Both Reactor Feed Pumps tripped.
?
(4) Reactor water level decreased causing both HPCI and RCIC to i auto start and inject to the reactor vessel. 4 (5) An operator secured HPCI, took manual control of RCIC and ; maintained reactor vessel level. l
~
(6) A feud pump was restarted. After level control was transferred to the feed pump, RCIC was secured. f (7) Pl ant was placed in a normal HOT SHUTDOWN condition. i Which DAE (1) of the following. based on the information given above. is the correct statement conc 7rning cubsequent reactor operation. JUSTIFY your response ; i I
- a. Power % oration cannot resume because HPCI auto initiated and I injected to the reactor vessel. ,
I
- b. Power operation cannot resume because the f eed pumps should [
not have tripped. } l c. Power operation cannot resume because a safety limit may have i been violated.
- d. Power operation cannot resume until the MSIVs have been -
- inspected due to the Group I isolation signal.
- .
1 i L ! I i i h t i I (***4* CATEGORY 08 CONTINUED ON NEAT PAGC *****) . I ! l i l
Br__e90181SIBGI1YE_EB96EpyBggi_GQU91IlgNgx_GND_LidlIBIlgNg- ~PAGE 12 f i F f i QUESTION 8.11 (2.00) i ). ! j- You have just assumed the 0000-0800 (2/0/88) chift as the Unit i Shift ' Supervisor. The plant is operating at 100% power with 98% core flow and-the following equipment out of service.(OOS): l Date 00b t
'SLC Tank remoto level indication 01/25/88 ;
i- RHR Service Water Pump A 01/31/88 RBCCW Pu np C 02/01/88 , CRD Pump B 02/01/88 HPCI Equipment Area Coolers 02/06/88 6 Core Monitor 02/05/88 ! Turning Gear Motor (Main Turbine) 01/15/88 Condensate Pump A 02/05/88 4 Annwar EACH of the following questions L. sed on the above informations
- a. LIST the out of service equipment which should havo Limiting j
] Ccnditions for Operation (LCOs) in effect and STATE the ; i i surveillance requirements that must be performed for uach j per Technical Specifications in order to allow continued [ j operation in these conditions. (1.0) t i
- b. STATE how long the reactor may remain in operation if NO
. repairs are completed. (0.5) , l c. While perf orming the RHR quarterly f ull flow test. RHR STATE the TS requirements ; Pump C is decl ared inoperabic. i concerning plant operability. (DE SPECIFIC and REFERENCE ! THE TS LEING APPLIEDt) (0,5) ! I b i i i I . i s , i t i ] f l l l t i t t (***** CATEGORY 08 CONTINiiED ON NEXT PAGE *****) l l t i
- Y
e . Qu__6Dd1NISIBOIIME_EB9CERUBESt_G9N91I19651.8ND_ bid 1I6119NS PAGE 13 , i i i s t QUESTION 8.12 (1.00) l t You are.the Unit 1 Shift Eupervisor with the reactor operating et 80% ! power. CRD Pump A is out of service due to a faulty motor controller. l Maintenance personnel inform you that they wish to commente work and ! will be disconnecting the electrical leads on the controller. You have the Maintenance Work-Order-(MWO) in front of you and notice that there is rua Temporary Modification Sheet associated with the MWO. , t LIST TWO (2) conottions that must be met in order for the work to be i performed without a Temporary Modification Sheet. [ t L I L 5 A i l l i I i i f l t [ t t (4**** CATEGORY 08 CONT 4>UED ON NEXT PAGE *****) i t t
,-oo, e ww-+wne, - - -a m--,,w-
- 1. ,
Di__G901NISIBOI1YE_EB99EDUSESt_G96D1119BSx_8NQ_6101I611gNS PAGE- 14 OUESTION 8.13 (1.00) [ t You are the Unit 2 Shift Supurvisor with the plant in a Cold Shutdow f Condition. Both loops of LPCI and "B" loop of CS have been l ac t sequentially aligned to take suction f rom the torus so that I system functional tests can be performed. The following - ipment is ; out of service: A t< B Main Condensate Pumps ( HPCL / x- I f,g\1 2A Standby Diesel Generator All Suppression Pool level igsk mentation RWCU
-{)
An I&C technician comes to you and requests you align "A" CS loop [ to the torus so they n complete the logic system functional _ tests. i 4 a. STATE who er you would align "A" loop of CS to the torus , in th situation (YES or NO). (0.25) . i
- b. USTIFY your response. (0.75) i
~ f f i f i P l t
?
L i l (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) ! I i
..-m,, ,. _ - - - . . - - , - , -
-Qi_200dlulSIBellyg_eBQGEQUBESt _GQNQlllQN$t_6NQ_LidlIGIlGNS PAGE. 15 i
[ t e i : QUESTION 8.24 (2.00) !
- t i
- a. Unit 2 is operating at 85% power and Plant Service Water (PSW)
Pump "2A" has been out of service for two days. Maintenance . continues, and the pump is expected to be returned to service not I earlier than four days from today. A streable pipe break occurs in the "2B" and "2D" PSW pump room, removing both pumps from service. ; Repairs on these pumps are expected to take two weeks. , STATE the ACTION (S) you will take and REFERENCE all TS you use to I develop your. answer.
- b. STATE what your ACTION (S) would be for the above scenario if f 2nstead of Operctional Condition 1. the plant was in the REFUEL ;
~
Condition. ASSUME heat losses are sufficient to maintain Operational Condition 5. REFERENCE all TS you use to develop your answer. i h a h I 1 l t n I (***** CATEGORY 00 CONTINUED ON NEXT FAGE *****)
! -9t__0Dd1NISIB011YE_EB9GEQUBggi_ggNQlIlQN3t_8NQ_b10118IlgND- PAGE 16 l
- I i
i l f h' l -QUESTION 8.15 (2.00) 1 j For EACH of the situations below, STATE whether the plant is in an LCO j and if so. STATE what ACTION must be taken: l a. You are the Unit 2 Shift Supervisor during normal full power , j opurations when you receive the following information l i ,
- SLC Tank Temperature 82 deg F i i SLC Tank Concentration
- 11.3 weight percent (1.0) ;
$ b. On a different occasion, you notice the SLC Tank level as indicated ' !' on Panel 2H11-P603 i s 70%. (1.0) i SHOW ALL WORK AND REFERENCE ALL APPLICABLE SECTIONS OF THE TECHNICAL ! I l SPECIFICATIONS. CONSIDER EACH CASE SEPARATELY. t I !
- i 1
i-i i i f I s i t f i f (***** END OF CATEGORY 08 *****) (************* END OF EXAMINATION ***************) f i i e
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s I f
Bi__8DdlulHIBellYE_EBQGEQUBEgt_CQUQlIlgNSt_8NQ LidlISI1QNS PAGE 17
-ANSWERS -- HATCH 1&2 -88/02/08-PAYNE, C.
I ANSWER 8.01 (1.00) 4 d 3- , REFERENCE. EIH: U2 TS 4.9.2 LER #2-82-036 3.2/3.9 3.2/3.9 215004G005 215004G011 ...(KA'S) ANSWER 8.02 (1.00) a a -or- c REFERENCE EIH: 40AC-REG-001-05 3.1/3.7 -3.4/3.6 . . 217000G602 294001A106 ...(KA*S)
- i i
j ANSWER 8.03 (1.50) . a. Within one hour (via ENS for valid signal for HPCI injection) ! (4 hr report for RPS ectuation) (4 hr report for failure of HPCI)
- b. None (if assumo man NOT contaminated) 4 hr report (tf assume man IS contaminated)
- c. 1 hr report (f or situation beyond design basis of FSAR or unanaly:ed condition that significantly compromises plant safety)
(0.50 each) i l REFERENCE 10 CFR 50.72 EIH: 40AC-REG-002-OS LP 300.4, LO MB Module 3.1. EO #3.1.3.1 LER #1-82-069,R1 LER #2-82-098 I 3.4/4.5 2.9/4.3 212000G003 261000G003 ...(KA*S) l l l l
at__0DU181BIBOI125 EB9GE99BE5t_G9BD1I19N5t_6NQ_LidlI@Ilg@$ PAGE 18 . ANSWERS -- HATCH 1&2 -88/02/08-PAYNE, C. l ANSWER 8.04 (1.50) g
- a. (1) No (0.25) l (2) Procedure changes must be approved prior to performing the work. (0.50)
- b. (1) Yes (0.25) ,
'(2) Approval of the changes took too long (greater than 14 1 days from the original date of change). -(0.50)
REFERENCE ! fi!H e 10AC-MGR-003-OS [ U2 TS 6.8.3.c i LER M2-82-107 LER #1-82-077 LP 300.4, LO #A.7 2.9/3.4 294001A101 ...(KA'8) i i i
et__GDd1NIEIS8IlYE_EB9GERUBEst_G00RIIl00St_0NQ LidlI8IlQUS PAGE 19
-ANSWERS -- HATCH 1&2 -88/02/08-PAYNE, C.
ANSWER 8.05 (2.00)
- a. (Primary Containment Integrity is) NOT satisfied. (0,5)
(Per TS 4.6.1.1) Primary Containment Integrity is violated if all penetrations to the containment are not capable of being closed by OPERABLE containment automatic isolation val ves. (0,5) (To meet containment integrity and allow continued operation per TS 4.6.1.1 and 3.6.3), the ball valve must be deactivated in its isolated position. (0.5) (1.0)
- 6. (Per TS 1.0.T.3, Primary Containment Integrity is satisfied if all inoperabic automatic isolation valves are deactivated in the isolated position.) No change from Unit 2 requirements! (0.5)
NOTE: Closing the shear volve accomplishes same effect and will be accepted in lieu of deactivating the ball valve. . ALSO ACCEPT: If neither deactivating the ball valve nor closing the shear valvo can be accomplished, an orderly shutdown shall be initiated and the reactor placed in the Cold Shutdown Condition within 24 hours. REFERENCE EIH: U2 TS 3.6.1.1 TS 4.6.1.1 (note 1) U1 TS 3/4.7.D and Table 3.7-1 LP 300.1, LO #2 2.4/3.3 2.7/3.4 215001G005 215001G011 ...(KA'S) ANSWER 8.06 (1.00) I TS havu NOT been violated. TS 3.1.4.1 ACTION statement excepts TS l 3.0.4 from being applicable. (0.5) Thus startup may continue and l Condition 1 entered after a second operator is stationed. (0.5) l l REFERENCE EIH U2 TS 3.1.4.1 I TS 3.0.4 i LP 300.1. LO 42 3.2/4.0 3.5/4.2 201006G005 201006G011 ...(KA'S) l
t
@t__OpblNIDIBGIlyg_EBOCE DyBE@ s _GQUplI198@t _8N Q_l,1 M I T ATigNW ' PAGE 20 l
- ANSWERS -- HATCH 1&2 -88/02/08-PAYNE. C. !
)
L l l ANSWER 8.07 (3.00)
- r 4
I
- 1. Continued ops allowed (0.25). Only one of the two fuel oil transfer [
pumps is required por TS 3.6.1.1 (0.5).
'i l .
- j. 2. Continued ops allowed (0.25). Por TS 3.4.D, plant is in a 7 day LCO provided the redundant component is operablo (0.5). :
- 3. Continued ops NOT al3 owed (0.25). This situation addresses circum- !
stanCOs in excess of those given in TS 3.e.4.1. Therefore, TS 3.0.3 " applies and facility shall be placed in Het Ghutdown within 6 hours ! } and in Cold Shutdown within the next 30 hours (0,5). !
- 4. Continued ops allowed (0.25). (TS 4.9.A.2 requires both fuel oil r trensfer pumps to be operable for the D/G to be considered operable.) l I
{ DG 1B is INOP and TS 3.9.B.2 applies limiting operation to 7 days i if two 230 kV offsite transmission lines are available, both !
- remaining DG's and associated buses are operable and increased SV requirements implemented per TS 4.9.B.2 (0.5).
I REFERENCE l EIH: U2 TS 3.8.1.1 l j U1 TS 3.4.B & BASES , l
- U2 TS 3.6.4.1 i U1 TS 3/4 9.A.2 i l LP 300.1, LO #2 i 4
3.4/4.1 3.4/4.1 ! 1 264000G005 264000G011 . ..(KA*S) [ ' i r t i 1 l i i ' I I t i i f
?
l f I r I L zweme,ww,v-~---.,-~,-~.-w,- -- - _ ,,-,,-- m _ ,----,- --a,~~---se,.,-,-,--- - - ,
Dt._0Dd1NISIBOIIVE_EBgCggyBgg,_Cggp111gggi_0ND_LidlIGI19BS PAGE 21 l ANSWERS -- HATCH 1&2 -88/02/08-PAYNE, C. I i
. - r ANSWER 8,08 (3.00) 1
- 1. 3.6.J.3 - requirements of Sections 1.1.A, 2.1.A. 3.1.A, 3.2.G, .
3.11.A. and 3.11.C as applicable to single loop ops shall be mot f (or unit pl aced in HOT SHUTDOWN within 12 hours). ; t
- 2. 3.6.J.4 - initiato action within 15 minutes to place power / flow (
below limits of Fig. 3.6-5.
- 3. (1.1.A. 3.11.C) --increaso MCPR limits by 0.01 (to 1.08). ;
i
- 4. (2.1.A.1.C.(1). 3.1.A table 3.1-1) - adjust APRM setpoints. j r
- 5. (3.2.G table 3.2.7) - adjust APRM rod block setpoints.
- 6. 3.6-J.2 - verify operation below limits of Fig. 3.6-5 at least !
once per 24 hours and whenever thermal power has been changed by ' at least 5% of RTP and SS conditions have been reached. (0.5 each) f
- 7. 3.11.A - none (unless power is increased above 52% rated thermal i power). (ANSWER NOT REQUIRED) i NOTE: Listing TS 3.6.J.1, 3.6.J.5, 3.6.D, & 3.6.E will not result in ;
+ or - credit !
REFERENCE f EIH: U1 TS 3/4. 6.J. 2 i LP 300.1, LO #2 ! 3.4/3.4 3.3/4.0 3.4/4.2 3.6/3.7 3.4/4.2 3.4/4.2 202001A203 202002A201 2020026005 202002G011 ...(KA'S) : i i i [ l i 5 f t i i r [
}
. - _ , . - _ _ _ . - . - . - - . _ . , . , . . _ _ _ . . - _ , . , - , . , . - . ~ _ , _ . . _ . , _ - . _ . _ . _ , _ . , . . - . . . _ . _ , , _ _ . . - . . - . . - - - - . - - , , - _ - ,
at _0Rd1NISIB8I1Y.5_EBQQEQUBE@t_GQNQlllONQi_6NQ_bidlI@Il(US PACE' 22 ANSWERS -- HATCH 1&2 -88/02/08-PAYNE, C. i i ANSWER 8.09 (2.00) I
- 1. 0700 Thurs - 0700 Fri (0.25): work more than 16 hrs in 24 hr I period (0.25).
i l
- 2. 1700 Thurs - 0000 Fri (0.25): less than 8 hr break between' work periods (0.25). l
- 3. 0600 - 2300 Sat (0.25): work more than 16 hrs straight (0.25). !
- 4. 0700 Mon - 2300 Sat (0.25): Work more than 72 hrs in seven day ueriod (0.25). (Occurred at 2200 Friday)
NOTE: There are other violations that may have occurred in addition to those listed above. Any four = Full Crodit. REFERENCE EIH: U1 TS 6.2.2.g 2.7/3.7 294001A103 ...(KA'S) t
' 4WER 8.10 (2.00)
! c (1.0) i On a Group I isolation, the SCRAM signal should come from MSIV closure and NOT high reactor pressure. (0.5) [ ! TS (1.1.c) states that a Safety Limit shall be assumed to be exceeded when a SCRAM is accomplished by a means other than 3 the expected scram signal. (0.5) { r i i (Also, TS (1.0.DD) indicates that exceeding a Safety Limit t requires Unit shutdown and review by the NRC before resumption of Unit operation.) i , ! REFERENCE EIH: U1 TS 1.0.DD, 1.1.C ' i, 42EN-ENG-011-03 t 4.0/4.1 4.0/4.2 4.2/4.3 3.8/4.0 3.1/4.3 3.2/4.1 3.3/4.1 239001A203 237001A212 239001G001 239001K127 ...(KA'S) i l I i I _ , _ _ _ , ____._,_r_____.__,.. , . _ . , _ , _ , , _
i Br_.8Dd1NISIBOIlyE_EBQQEQUBE@s_GQNQlIlONSt_@UD_61d11011gNS PAGEL :23 ANSWERS -- HATCH 1&2 -88/02/08-PAYNE, C. ANSWER- 8.11 (2.00)
- d. RHR Service Water Pump A (TS 4.5.C.2) (0.1)
- remaining active components of both RHR subsystems shall be demonstrated to be operable immediately. (0.2)
- an operable service water pump shall be demonstrated to be operable daily. (0.2)
HPCI Equipment Area Cooler (HPCI INOP per TS 4.5.D.2 via 3.5.K.2) (0.1)
- - the ADS actuation logic, the RCIC system, the RHR system i LPCI modo, and the CS system shall be demonstrated to be
- operable immedi ately. (0.2)
- the RCIC system and ADS logic shall be desnonstrated to be operable daily. (0.2)
- b. 14 days past 2/6/88 (when HPCI pump was declared INOP)
(or 2/20/88). (0.5)
- c. Per TS 3.5.C.4, the reactor shall be placed in the Cold l Shutdown Condition within 24 hours. (0.5)
TERMINAL TIMEOUT IN 30 SECONDS REFERENCE 3 EIH: U1 TS 3.5 LP 300.1, LO #2
'3.6/4.5 3.7/4.4 3.'/4.3 4
! 203000G011 206000G011 217000G011 ...(KA'S) e i 4 i h i 4 4 1 l - - . _ , - - . - . , , _ . _ . . _ . . . . _ _ _ - . . . _ . - . _ _ . . - , , , .
._.m_ , , - , , . - - . - _ . , , _ _ . - . . , . , - . - , . _ _ _ - . , - , , - - _ . , ,
at__8DdINISI68IlVg_d6QQEQUBES t _QQNQlligN$t_8NQ_LidlIGllgNS PAGE 12 3 ANSWERS -- HATCH 1&2 -88/02/08-PAYNE, C. ANSWER O.12 -(1.00)
- 1. A functional test must be defined on the MWO that will test proper operation of the motor.
- 2. Work must be controlled by a clearance..
- 3. If functional test of a NWO requires that all wires and links be red-lined, or if it verifies the designed functions-and/or setpoints of the subcomponent.
, 4. If the activities described (in 30AC-OPS-005-OS, Sects.on 2.1) are adequately covered in another approved plant procedure.
(Acceptability of other plant procedure is based on 4 specific guidelines given in-30AC-OPS-005-OS.) (any 2 G 0.5 each) REFERENCE EIH: 30AC-OPS-005-OS LP300.4 LO #10 3.9/4.5 294001K102 ...(KA*S) l
; ANSW'dR 8.13 (1.00)
J
- a. No '.25)
- b. TS 3/4.5.3 requires an OPERABLE flow path (in Condi n4 or 5) from either the SP or the CST. (0.25)
< Since SP level instrumentation is )I4 ', f-'\
. en all systems
! lined up to the SP are consider dO . (0.25) 4 Therefore, he must maintaink "A" loop of CS aligned to the CST (until "B" loop CS is re-aligned to the CST. (0.25) 4 REFERENCE EIH: U2 TS 3.5.3 LER #2- 048 l 2.8/o . 3.3/3.4 3.3/4.2 3.4/4.2 I 20~-01G005 209001G011 209001K410 209001K603 ...(KA'S) i l l l l
'ex__8Dd161EIBOI12E_EBQGEDUBESt_GQND1IlRNSt_8NQ_ bid 110119NS. ~PAGE 25 ANSWERS -- HATCH 1&2 -88/02/08-PAYNE,-C. ' ANSWER 8.14 (2.00)
- a. NONE of the ACTION statements listed in TS 3.7.1.2.a apply (0.5)
TS 3.0.3 must therefore be applied and the unit must be placed in at least HOT SHUTDOWN within 6 hours and in COLD SHUTDOWN within the following 30 hours. (0.5)
- b. Per TS 3.7.1.2.6 since they cannot' restore both PSW loops with at least one pump in each loop to an OPERABLE status within 7 days, must declare the CS system, the LPCI system and the associated diesel generators inoperable and tahu the ACTION required by TS 3.5.3.1, 3.5.3.2, and 3.8.1.2. (1.0)
REFERENCE EIH U2 TS 3.7.1.2 4.2/4.2 294001A101 ...(KA'S) ANSWER 8.15 (2.00)
- a. From Fig. 3.1.5-1 in TS, the temperature vs. concentration is SATISFACTORY. NOT in an LCO. (1.0)
- b. Converting percent level to gallons:
100% = 5150 gal => 70% = 3605 gal. (0.25) Per TS 3/4.1.5, tank level must be greater than or equal to 3802 gal. for the system to be OPERABLE. (0.25) Therefore, ARE in LCO 3.1.5.a & ACTION a.2 applies (restore system to OPERABLE status wi thin 8 hrs. or be in at least HOT SHUTDOWN within the next 12 hrs.) (0.5) REFERENCE EIH: U2 TS 3.1.5 LT-1H-01101-00 EO #17 3.6/4.4 3.4/4.1 211000G005 211000G011 ...(KA*S)
. TEST CROSS REFERENCE PAGE 1 ;
1 l QUESTION- VALUE REFERENCE- ! l - i i 08.01 1.00 DCP0001435 08.02 1.00 DCP0001436 : o08.03 1.50 DCP0001431 l
'98.04- 1.50 DCP0001433 !
'08.05 2.00 DCP0001425 98.06 1.00 DCP0001426 08.07 3.00 DCP0001428 i 08.08 3.00 DCP0001430 ;
08.09 2.00 DCP0001438 ! 08.10 2.00 DCP0001440 l 08.11 2.00 DCP0001444 7 08.12 1.00 DCP0001445 ' 08.13 1.00 DCP0001446 08.14 2.00 DCP0001447 08.15 2.00 DCP0001429 t 26.00 26.00 DOCKET NO 321 i i l k r 5 f t i t i I ! 1 [ l l i i i (
L
'~
GeNghibwer Company
., , - E: tam 1. Hatch Nuclear Ptant ENCLOSNRE 3 Fbst Off<e Box 439
- Bailey Georg.a 3t513 Te!ephcyg 912 367 7781 912 537 9444
- J. T. Beckham, Jr. GeorgitiNwer vee nescent ,, ,
February 15, 1968 PLANT E. 1. HATCH Plant Hatch Operator Licensing Examination Rtype: A04.25 Log: LR-VPH-014-0288 U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W. Atlanta, Georgia 30323 ATTNf_Caudle A;"Julian,*
' Chief, Operations Branch '
Division of Reactor Safety Gentlemen:- In accordance with NUREG 1021, I have enclosed the formal coment submittal for the Plant Hatch Operator Licensing examination administered on February 8, 1988. Please call C. L. Coggin at 912-367-7851 if further information or clarification is desired. [, s // J. T. Beckham, Jr. 7 Vice President Plant Hatch JTB/RSG/gkb xc: C. L. Coggin E. M. Howard (w/o enclosure)
__-_w_ __ __ -- _ _ ___ ____ Plant E.I. llatch Utility Comments Reactor Operator and Senior Operator Written License Examination Comments February 11, 1988 t 3.04c Utility Comment: r i The question requires the effect of Recire Loop operation on 3 wide range instrumentation. The stem does not delineate any ' r specific change in the Recire operation. This causes the candidate to make assumutions that may not be consistent with the intent of the question. It is also noted that at flatch l' the Wide Range Instrumentation" is called Emergency Range i Instrumentation. This may have caused some confuulon. Recommendation: i It is requested that 3.04c be deleted.
Reference:
E.I. Ilatch,. LP 44.1 t i 3.09d Utility Comment _L The question requires the pt esaure setpoint to i solate Nitrogen backup valves to Drywell Pneumatics System. The Drywell Pneumatic Air Compressors are not used at Plant ' lia tch . Therefore nitrogen is manually aligned to the Drywell Pneumatic System and any isolation is inconsequential to the , operation of the system. ! Recommendation: ; } It i s requested that 3.09d be deleted t
Reference:
E.I. ita tch LT-Ill-03501-00 l i 6.06a Utility Comment: 1 The question requires the high and low side gressure sensing points for Core Spray Line Break Instrumentation Detection System. The ques tior, does not state plant conditions, which will vary high and low side sensing. The answer key is given for cold conditions. At power, the answer is reversed due to , Pitot Tube Effect. i Recommendation: [ It i s requested that 6.06a be deleted. i
Reference:
_ t E.I. Ita tch LT-Ill-00801-00 ) i'
-n ,-p,,,7 - --
w,,,,----,v---ggmg- m----gryn-,,,,,(-, yryyy,,y,-wwwms, m,w-wgy-- ,wy y,4www-w-m-x-,ew,-+ m4 m w wm ry, o, wye m m-- yavmwmw g -m ry ec
6.11b Utility Comrent: The question requires the reason for maintaining a Drywell to i Torus Differential Pressure. Due to recent Hatch ( modifications, this is no longer required by the normal plant startup procedure. Recommendation: It is requested that 6.11b be deleted.
Reference:
34GO-OPS-001-1S , 7.06b Utility Comment: , The question requires a description of the verification process for proper seating of fuel onto fuel support piece. Procedure allows for the use of the fuel grapple, in a scanning mode, to verify seating. Recommendation: It is requested that the key be revised to include the use of the Fuel Grapple for the verification. ,
Reference:
42FH-ERP-014-OS, 7.5.15 through 7.5.20 i I 8.03c Utility Comment: The question requires the necessary reporting requirements if t RPS Divisional Circuit Separation criteria has been violated, j Technical Specifications Amendment #86 deleted prompt and follow-up notifications from Technical Specifications. All necessary raports would be made per 40AC-REG-002-OS. i ) Recommendation: L Revise the answer key to accept One Hour Report per l 40AC-REG-002-OS. , l Re fe rence : Unit Two Technical Specifications sections 6.9.1.12 and i 6.9.1.13 Amendment 86. ' 40AC-REG-002-0S I l I l.,..... _ _ . . . . _ _ _ _ . , _ . _ _ _ _ _ . _ _ _ _ _ . _ , _ _ , _ _ _ . . _ _ _ _ _ _ , _ , _ _ _ . _ . . _ _ _ _ _ . _ _ . , _ _ ._
8.05 Utility Corment: The question requires the Technical Specifications actions for the failure of a TIP ball valve. The failure of a TIP ball valve does not constitute a Primary Containment failure. It is the failure of a Primary Containment isolation valve. . l Answers should address requirements for inoperable Containment isolations. Recommendation: It is requested that the answer key be revised to accept the < following. A. Yes, Primary Containment Integrity is maintained due to Ball Valve being closed. Only actions required would be those per Technical Specifications Section 3/4 6.3. B. Unit One Technical Specifications would not require the deactivation of the Primary Containment Isolation Valve. It would only require the adherence to Section 3.7.D.2
References:
Unit 1 Technical Specifications; 3.7.D.2. Table 3.7-1 Unit 2 Technical Specifications; 3/4 6.3 4 8.07 Utility Comments: The question requires an interpretation of Technical Specifications to determine if continued operation is allowed. The answer key states whether the equipment is operable / inoperable, and does not address if continued operation is allowed. Recommendation: , It is requested that the key be revised to incorporate the Utility comments as follows: , Part 1; Yes - Operation may continue due to Technical Specifications requiring only one Fuel Oil Transfer Pump for Diesel Generator Operability.
-or-Yes - Student may respond with a conservative interpretation of Technical Specifications Operability as defined in the definition section. This will result in the plant being in a 72 hr LCO due to an inoperable Diesel Generator.
Part 2; Yes - IB SBLC Pump being inoperable. puts the plant
- into a 7 day LCO per 3.4.B.
Part 3; Yes (this would place the plant in a shutdown LCO) - The plant will be in a 6 hours to Hot Shutdown and 30 hours to Cold Shutdown LCO per Technical Specifications section 3.0.3. Part 4; Yes - One Diesel Generator being inoperable places t
- the plant into a 7 day LCO per Unit 1 Technical Specifications ,
i section 3.5.G or should accept a 72 hour LCO if candidate addressed unit operation on Unit 2 with "B" Diesel Generator i inoperable per Technical Specifications section 3.8.1.1. l 1
References:
Jnit 1& 2 Technical Specifications. i
Y 8.13 Utility Comments: The question requires an interpretation of Technical Specifications to allow alignment of "A" loop of Core Spray to the Tores. Since the plant is in condition 5, 3.0.5 does not apply. With the 2A Diesel inoperable, Core Spray "A" is inoperable, resulting in the initial conditions not being allowed by Technical Specifications. This situation may lead the candidate to make assumptions that may not be consistent with the intent of the question. The conditions specified in the referenced LER were without the "A" Diesel Generator being inoperable. Recommendationt It is requested that 8.13 be deleted.
References:
Unit 2 Technical Specifications Section 8 Utility Comment: Two primary factors are believed to have contributed to the greater than anticipated time required on section 8. First, the arrangement of the questions required frequent use of alternate unit Technical Specifications. This aspect could be , improved by grouping questions dealing with each unit's Technical Specifications . Second, the importance of bounding questions becomes greater for open book examinations. Some questions may lead the candidate to go into greater depth than expected, by not specifying the scope or depth closely enough. Examples include: Question 8.03, regarding required reports, and question 8.09, regarding overtime restrictions. l l l l i I J l 1 1
O' .- v . 3 h
--e g TABLE 3.7-1 (Cont'd) .
- PRIMARY CONTAlHMENT ISOLATIOff VALVES W98tCil RECEIVE A PRIMARY C001TAllW4ENT ISOLATIOst SdCNAL a f m ~
ActTon on isolation Group llumber of Power Oserated Velves '\ \ OpeMaximum ra ti ng 16o rma l Position Initiating (b) Velve Identification fdl . Inside Outside N Yime fsoci fel Slanel 1 2 Suppression chamber exhaust volve 2 5 C SC bypass to standby gas treatment ' (T48-F339, T48-F338) i 2 Suppression cheeber nitrogen 1 5 C SC make-up I Ine (no rme 1 opera t ion) (T48-F1188) 2 Drywe l l and suppression cheeber' 1 5 C SC nitrogen supply line (Inerting) (T48-F103) 2 Dpil and suppression chenber nitrogen 1 5 C SC mehe-up IIno (normel operation) (T48-F104) W 2 O rywe l l equ i pmen t d ra i n sump d i scha rge
- 2~ 15 0 GC f y (E11-F019, C11-F020) 6 g 2 15 0 CC
[ 2 Drywell floo r d re in sump d i scha rge (cil-Foos, cli-Foo4) l 2 TIP Guide Tube 1 each NA C SC (C51-J004) ~ --- 1Ine . (c) Drywe i I pneusetIc system 2 _5 0 cc ( P70-F002, P70-F003) y 5/~ 5004 ;- <, p gc p Sf<ED Mowr -EQortwr-a t9
- h. .
3
- 1 2; -
r3
- t- TABLE 3.7-1 (Cont'G) e
_ PRIMARY CONTAINMENT ISOLATION VALVES WHIC11 RECEIVE A PRIMARY CONTAINMENT ISOLATION SICNAL
$--e -s isolation Group Number of Power Maximum Normal Action on (b) Velve Idgniffication idl 9ppr8ted Valyg1 Operating Position I n i t !s t in.g inside Outside Time (sect tal Slanal fel 6 RifR reactor shutdown cooling suction ( supply) 1 1 24 C SC (E11-F008, E11-F009) 6 RHR reactor head spray (E11-F022, [11-F023) 1 1 20/12 C SC f
3 H PCI - turbine steam 1 (E41-F002, E41-F003) 1 50 0 GC 4 MCIC - turbine steam 1 (E51-F007, E51-F008) 1 20 0 GC 5 Reactor water cleanup from 1 reci rculation loop 1 30 0 GC (C21-F001, C31-F004) I# 2 Post-accident sampling system supply 2
]J (B21-F111, B21-F112) 5 C SC b 2 Post-accident samp f log system return 2 (E41-F122, [41-F121) 3 C SC 2 Core spray test line to suppression pool 1 nach 50 C SC
([21-F015A,B) line E 3 . O. o
. . --- s -
LIMITING CONDITIONS FOR OPERATION SURVElttANCE REQUIREMENTS 4.7.0.1. Surveillance of Ooerable Valves (Continued)
- b. At least once per operating Cycle the reacto'r coolant system instrument line excess flow check valves shall be tested for proper operation.
- c. At least once per quarter:
(1) All normally open power-operated isolation valves n (except for the main steam line power-operated iso-lation valves) shall be fully closed and reopened. (2) With the reactor power less than 75% of rated,
*the main steam line isolation valves shall be tripped (one at a time) and closure time verified,
- d. At least once per week the main steam line power-oper-ated isolation valves shall be exercised ont at a time by partial closure and sub- -
sequent reopening. 3.7.D.2. Operation with inoDerable Valves 2. Surveillance of lines with an Inoperable valve in the event any isolation valve Whenever an isolation valve specified in Table 3.1-1 becomes listed in Table 3.7-1 is inoperable, reactor power operation inoperable the position of at may continue provided at least one least one other isolation valve isolation valve in each line having in each line having an inoperable an inoperable valve is in the mode isolation valve shall be verified corresponding to the isolated con- to be in its isolated position dition, daily. ;
- 3. Shutdown Recuirements if Specificatich 3.7,0.1. and 3.7,0.2.
cannot be inet, an orderly shutdown shall be initiated and the reactor shall be placed in the Cold Shut-down Condition w: in 24 hours. O HATCH - UNii 1 3.7-14 i
,' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.C.I. Surveillance while integri_t,y Maintained (Cont'd)
- b. Secondary containment capability to maintain a minimum 1/4-inch of water vacuum under calm wind ( 5 mph) conditions with each filter train flow rate not more than 4000 cfm shall be demonstrated at each refueling outage, prior to refueling.
, 3.1.C.2. Violation of Secondary 2. Surveillance After intearity Violated ) Containment Integrity
- a. Without Hatch-Unit I second- After a secondary containment viola-ary containm nt integrity, tien is determined the standby gas restore Hatch - Unit I sec- treatment system will be operated
' ondary containment integ
- inenediately af ter the affected zones rity within 4 hours, o'r per- are isolated from the remainder of form the following (as appli-
' the secondary containment. The cable): ability to maintain the remainder (1) Suspend irradiated fuel of the secondary containment at and/or fuel cask handling 1/4-inch of water vacuum pressure in the Hatch, Unit I sec- under calm ( 5 mph) wind conditions ondary containment, shall be confirmed.
- g (2) Se in at least Hot Shutdown .
within the next 12 hours and tpeet the Conditions of - 3.'.C.1.a. within the next 24 , hours.
- b. Without Hatch-Unit I secondary ,
containment, refer to the follow- ' ing Hatch-Unit 2 Technical Specification, for LC0's to be followed for Hatch-U.11t 2: , (1) Section 3.6.5.1. - (2) Section 3.9.5.1. D. Primrv Containment isolation Valves D. Pr1Nry containtwnt Isolation Valves
- 1. Valves Recuired to be Deerable 1. Surveillance of Operable Valves During reactor power operation, Surveillance of the primary con-all primary containment isolation tainment isolation valves shall be valves and all reactor coolant \ performed as follows:
system instrument line excess flow check valves shall be operable except a. At least once per opersting as stated in Specification 3.7,0.2. cycle the operable isolation valves that are power operated and automtically initiated shall be tested for simulated automatic initiation and the [Nv) closure times. l HATCH - UNIT 1 ,, 3.7-13 Amendment No. /0, 58, JJ, Jeg,125
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HATCH - UNIT 2 3/4 6-24 Amendment No. 59
CONTAINMENT SYSTEMS 3/4.6.3 PRIMARY CONTAlhMENT IS01.ATION VALVES LIMITING CONDITION FOR OPERATION
'3.6.3 The primary containment isolation valves and the reactor instru-mentation line excess flow check valves specified in Table 3.6.3-1 shall be OPERABLE with isolation times as shown in Table 3.6.3-1.
APPLICABILITY: CONDITIONS 1, 2 and 3. ACTION:
- a. With one or more of the primary containment isolation valves specified in Table 3.6.3-1 inoperable, operation may continue and the provisions of Specification 3.0.4 are not applicable, provided that at least one isolation valve is maintained OPERABLE in each af fected penetration that is open, and either:
- 1. The inoperable valve (s) is restored to OPERABLE status within 4 hours, or
- 2. Each affected penetration is isolated within 4 hours by use of at least one deactivated automatic valve secured A i in the isolation position, or
~
- 3. Each affected penetration is isolat.ed within 4 hours by
- use of at least one closed manual valve or. blind flange.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. With one or more of the reactor instrumentation line excess flow check valves specified in Table 3.6.3-1 inoperable, opera-tion may continue and the provisions of Specifications 3.0.3 and 3.0.4 are not appitcable provided that within 4 hours;
- 1. The inoperable valve is returned to OPERABLE status, or
- 2. The instrument line is isolated and the associated instru-ment is declared inoperable.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLO SHUTOCWN within the following 24 hours.
' 5HeML mc s rr w cn ,ag, 1
O HATCH - UNIT 2 3/4 6-15
CONTAltNENT SYSTEMS SURVEILLANCE REQUIREMENTS __ 4.6.3.1 Each primary containment isolation valve specified in Table 3.6.3-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of 'ull travel and verification of specified isolation time. 4.6.3.2 Each primary contain'aent automatic isolation valve specified in Table 3.6.3-1 shall be demonstrated OPERABLE during COLD SHUTDOWN or REFUELING at least once per 18 months by verifying that on a containment isolation test signal each automatic isolation valve actuates to its isolation positicn. 4.6.3.3 The isolation time of each power operated or automatic valve specified in Table 3.6.3-1 shall be determined to be within its limit when tested pursuant' to Specification 4.0.5. 4.6.3.4 Each reactor instrumentation line excess flow check valve shall be demonstrated OPERABLE at least once per 18 months by verifying that the . valve stops excess flow. O O HATCH - UN'T ; 3/4 C-16
e GEORGIA POWER COMPANY l
~
l l PLANT E.I. HATCH I l PAGE 17 0F 41 DOCUMENT TITLE: l DOCUMENT NUM8ER: l REVISION NO: FUEL MOVEMENT OPERATION ~ l 42FH-ERP-014-OS I O O 7.5.3 Set up the video equipment and lighting. 7.5.4 Attach the underwater television camera to the Refueling Platform Fuel Grapple. 7.5.5 POSITION the Refueling Platfona over the fuel bundles to be observed. ' 7.5.6 OPERATE the Grapple LOWER Control to position the camera close enough to the fuel bundle to clearly discern the serial numbers. 7.5.7 Start the video recorder and indicate the coordinate of the first bundle in ! the row being observed and the direction of scan (i.e., North, South, East, ' j or West) with an audible statement on the video tape. 7.5.8 Scan the core slowly, one row of fuel at a time, and record the serial ! ! number of each fuel bundle in the corresponding coordinate blank of the core , ! map. 1 7.5.9 Record any discrepancies found on Attachment 9. 7.5.10 OPERATE the Grapple RAISE Control until a four bundle cell 'can be clearly discerned. ] ( 7.5.11 Indicate the coordinate of the fuel cell in the row being observed and the . . j direction of scan with an audible statement on the video tape. - c 7.5.12 Scan the core slowly, one row of cells at a time, and confirm the following: ! 7.5.12.1 Each fuel bundle in the cell is oriented properly (the channel fastener ] clip is oriented toward the center of the control rod). ! 7.5.12.2 The twelve uncontrolled bundles on the. core periphery have their channel l fastener clips oriented radially outward from the core edge. 7.5.13 Record any discrepancies found on Attachment 9. . 7.5.14 0PERATE the Grapple RAISE Control and raise the camera out of the water. ' 7.5.15 Remove the camera f rom the Fuel Grapple. 7.5.16 OPERATE the Grapple LOWER Control until the Fuel Grapple is located just above the fuel bale handles. 4 s 7.5.17 Scan the core slowly, one row of fuel at a time, and confirm that all fuel l bundles are seated correctly as indicated by the Fuel Grapple not striking . l any fuel bale handles. J L j 7.5.18 Record any discrepancies found on Attachment 9. l l 7.5.19 OPERATE the Grapple RAISE Control and position the Refueling Platform as j s necessary. 1
- 7.5.20 Complete Attachment 9 and attach the core map used. .
] MGR-0001 Rev. 1
- _ - - - - _ _ . -. _ - . - - . - - - _ ~ . - _ - _ - - . . . .. - _. . . . - - . . . _ - ~ _
. . - . c Jh-s - . -
ThWD4UL'.RR LT-IH-00801-00 uut/ . ACTIVlw OUILDE & IEIl0CrKN IE'Il0Cl(R 10I15 >
- a. During cold conditions:
I - Pressure at point 2 is due to the weight of the water colsses within the core shroud (applied to the Ili side of the DP ., H I inst rianent) . j - Pressure on the Im side of the DP Instrsanent is due to the weight of the l water colunn in the piping fran the Core Spray lleader. 1 DP is low or zero. l
- b. As rated tenperature and pressure is reached the density of the water in the vessel decreases, therefore less applied pressure to the 111 side of the DP instrianent resulting in a negative DP reading.
- c. As core iIow is increased, the pressure at point 3 increases, therefore nore applied 9 pressure to the lo side of the DP instrssnent resulting in a greater negative DP reading.
(naninal DP at rated is approximately -3.5 psid).
- d. If a Core Spray line breaks internal to the vessel, pressure sensed will be between points 2 and 5: DP becanes less negative.
- c. Alann at I psid above nonnal DP (1) loca1 indication (2) Annameiator 4
Page 30 of 63
s 0 fiTEAM DRYER DRYWELL WATER LEVEL d
- y F037 F005 F004
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F007 G [ SD Nl9W O5 F006 E S A ___ _______ W 4r PAH
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! it- PLATE dpA H ge PRESSURE
- l PRESSURE IristDE THE l
k*- UPPER PORTICri 0F die . 1 LOW (ph HI CORE SHROUD. l = STANCBY LICUlO CONTROL I I i TR ANSPARENCY O O 8 01 - 9 l CORE SPRAY SYSTEM PlPE BREAK DETECTION INSTRUMENTATION i 1 I FOR TA' M ~'
.,t.
c C (This page is intentionally left blank.) I i i l l C lC HATCH - UNIT 2 6-15 Amendment No. /B, 86 t
l .. 1 O C C C (This page is intentionally left blank.) O C u. C ( O HATCH - UNIT 2 6-16 /cendrent No. 37.18.48,86
ENCLOSURE 4 SIMULATION FACILITY FIDEllTY REPORT Facility Licensee: Georgia Power Company Facility Licensee Docket No.: 50-321 and 50-366 Facility Licensee No.: DPR-57 and NPF-5 Operating Tests administered at: Edwin 1. Hatch Nuclear Plant Operating Tests Given On: February 9-11, 1988 During the conduct of the simulator portion of ti,e operating tests identified above, the following apparent perferrance and/or human factors discrepancies were observed:
- 1. *B" Recire Controller has a dead spot that prevents shifting to AUTO without help from the Simulator Instructor.
- 2. The Annunciator Response Procedure for RBCCW Expansion Tank Low Level alarm was not available in the simulator control room.
- 3. The system lineup for the RCIC coolers was not in accordance with the procedure for all the IC's used on the simulator.
i
- 4. If RCIC trips (e.g. during SV testing) and is then reset, a subsequent restart of the system will result in a spurious injection into the reactor vessel.
j}}