IR 05000289/1985031
| ML20153F324 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 01/28/1986 |
| From: | Dudley N, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20153F314 | List: |
| References | |
| 50-289-85-31, NUDOCS 8602250481 | |
| Download: ML20153F324 (41) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 50-289/85-31 (0L) FACILITY DOCKET NO. 50-
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FACILITY LICENSE NO. DPR-50 LICENSEE: GPU Nuclear Corporation P. 0. Box 480 Middletown, Pennsylvania 17057 FACILITY: Three Mile Island Unit 1 EXAMINATION DATES: December 13-18, 1985
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CHIEF EXAMINER: // / / /-2[-8h N. Dudley, Lea 6 Reactor Examiner Date REVIEWED BY: h [[ Robert M. Keller, Chief, Projects Section 1C Date
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APPROVED BY: - Harry B. $) ster, Ch14L
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Projects 1 ranch No. 1 SUMMARY: Five written, five oral, and four operating examinations were administered to four Senior Reactor Operator Candidates and one Instructor Certification Candidate. All candidates passed all portions of the examin-ations. No significant generic weaknesses were noted during the examinatio , i 8602250481 860130 PDR ADOCK 05000289 V PDR w _ _ _ _ _ _ - _ - - - - _ _ - . _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-- _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ - _ - - _ _ _ - _ _ _ _ _
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REPORT DETAILS TYPE OF EXAMS: Replacement EXAM RESULTS: I R0 l Inst. Cert l IPass/ Fail l Pass / Fail I I I I I I I I I Written Examl 4/0 l 1/0 I I I I I I I I I I Oral Exam I 4/0 1 1/0 I I I I I I I I I ISimulator Examl 4/0 1 -/- 1 I I I I I I I I I Overall l 4/0 1 1/0 I I I I I CHIEF EXAMINER AT SITE: N. Dudley, NRC OTHER EXAMINERS: W. Apley, PNL Summary of generic deficiencies noted from grading of written exams: Over the last two SR0 examinations, some of the licensed reactor operators who have been assigned to the Training Department have done poorly on Section 5, " Theory of Nuclear Power Plant Operations." Personnel'present at Exit Interview: NRC Personnel N. Dudley, Lead Reactor Engineer (Examiner) Facility Personnel ' M. Ross, Manager Plant Operations B. Leonard, Operator Training Manager Summary of NRC Comments made at exit interview: The NRC reviewed the number and types of examinations administered. A discussion was held to clarify some technical details relating to candi-
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dates' oral responses. The NRC then presented the-following training program weaknesses noted during the examinatio * The Training Center's copy of the-Technical Specifications and the copies of the Technical Specifications provided to the NRC for examination preparation were not up to dat * There is no basis for operating two Reactor Cooling Pumps (RCP) on the same bus vise separate buses during a control room evacuatio The Manager Plant Operations stated he would evaluate whether to-change the procedure to have two RCP operating from separate buse . Changes made to written exam during examination review: Question N Change Reason 7.2 Change "26-inches high Clarifies units of to "26-inches Hg." condenser vacuu .3 Change " Environmental Environmental Technical Specifications have been replaced by Appendix B" to " NPDES the NPDES Permi Permit".
Answer N Change Reason 6.6 Delete "STOP PUSH BUTTON Trip is not actuated (Engine and Control Room)". by ES signa .7 Add "It is also possible for Expands answer to allow BTU limit to clear as reactor additional correct power increases which causes respons Th to increase and therefore feedwater should start increasing".
6.8 Add " pump runout; starving Expands answer to allow other OTSG". additional correct respons Add "b". The responsibility of notifying offsite organizations has been assigned to Emergency Support Director by a recent Temporary Change Notice (TCN 1-85-0157).
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< Attachments: ,
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I- Written Examination and Answer Key (SRO) . . .
{- Facility Comments on Written Examinations made after Exam Review
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U. S. NUCLEAR REGULATORY COMMISSION
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SENIOR REAC10R OPERATOR LICENSE EXAMINATION
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Facility: TMI-l Reactor. Type: Babcock & Wilcox - PWR
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Date Administered: December 16, 1985
. Examiner: W. J. Apley /N. Dudley Candidate:
INSTRUCTIONS TO CANDIDATE.
< Use separate paper for the answer Write answers on one side onl Staple question sheet on top of the answer sheet Points for each question are indicated in :.orentheses af ter the questio The passing grade requires at least 701 in each category and a final gra.de.of at least 80%. Examination papers wi.ll be picked up six (6) hours after 'the examination start Ca tegory % of Candidate's % of Value To tal Score Cat. Value Category 25 25 5. Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics 25 25 6. Plant System Design, Control
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and instrumentation 25 25 7. Procedures - Normal, Abnormal, Emergency, and Radiological Control 25 25 ~ 8. Administrative Procedures,
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Conditions, and Limitations 100 TOTALS
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Final Grade % All work done on this examination is my own; I have neither given nor received ai CandidatF s signature
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1 THEORY OF NUCLEAR POWER PLANT OPERATIOh. FLUIDS. AND THERMODYNAMICS (25.0) - State whether each of the following statements is IRuE or > FALSE? EQ explanation required.
.' The operator can increase the heat removal rate from the RCS by reducing steam pressure or increasing OTSG , leve .
(0.5) ^
4 A LOCA with no RCPs running can result in more inventor loss than a LOCA with RCPs runnin (0.5) ! A total and prolonged loss of OTSG feed can lead to a
loss of RCS liquid inventor (0.5) i The primary concern when fuel clad temperature reaches 1400 degrees F is the production of hydroge (0.5) i Give two (2) reasons why when a single feedwater pump ' trips ' at 92% power, there will be a small initial increase in ! generated Megawatt (2.0) i
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4 The condensate booster pumps discharge into parallel strings' of low pressure stage heaters (12,10, 8, 6). Why is the condensate temperature controlled below approximately 370 F at this point? (1.0) acd# p During power operations-at 45% mith three (3) reactor coolant I bcW ' pumps (RCP) in operation, the fourth RCP (loop A) is starte # Briefly discuss the following parameters during the tran- W'U y f C p.#- 7",,, g% 4 sient: 3. p .F) l Hot leg temp, Cold leg temp, and dTc (each loop) (1.5) I Feed Flow (each OTSG) (1.0) OTSG level (each OTSG) (1.0) Include the control actions of the ICS where applicabl .5 Why for the same reactivity change does the reactor respond
' more rapidly at End-of-Cycle (EOC) than at Beginning-of Cycle (BOC)?
(2.0)
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-Section 5 Continued on Next Page-
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. _ _ - - - < . . -, . . - 2 The reactor is subcritical with a K-eff of 0.9 Source channels are indicating 5 counts per second (cps). What is
K-eff when source channels show a count rate of 60 cps after rods are withdrawn? Show all work for full credi (2.0) Why is the worth of two (2) rods together not necessarily the-same as the sum of their individual worths? . (2.0) Control rod worth increases as moderator temperature in-creases. Describe how the amount of boron concentration affects the magnitude of that increas (2.0) ,
- Which of the following reactor conditions would least aggravate a Xenon oscillation? Select on Low Power, Positive MTC Low Power, Negative MTC High Power, Positive MTC High Power, Negative MTC -
MTC = Moderator Temperature Coefficient (0,5)
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5.10 One hour after shutdown, the heat generation in the fuel assemblies is approximately 1% of operating values, but in - the reflector and shield it is approximately 10% of operating values. Exolain the differenc (2.0)
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5.11 IRUE or FALSE: Departure from Nucleate Boiling (DNB) occurs when the surface temperature of the fuel assembly gxceeds
the saturation temperature of the liquid by 6 to 8 (0.5) d 5.12 If the pressure in a leaking subcooled water pipe'is reduced . by 50%, approximately how much will the leak rate decrease? Exolain any assumption (2.0) ' 5.13 Exolain what indication you should expect to see on the source range count rate monitors as the core is voided by a Loss of Coolant Ac.cident (LOCA). (1.0) 5.14 IRUE or FALSE: According to Operating Procedure 1104-9 (Circulating Water), Station Net Megawatt Generation is significantly improved by reducing to four (4) circulating water pumps during cold ambient condition (0.5)
-Section 5 Continued on Next Page-l
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.. . . . < -,. , . . ! - . 3 5.15 During a RCS cooldown and depressurization, the following condition occurs. A large rapid increase in pressurizer level occurs while reducing RCS' pressure, even though an adequate saturation margin is indicated between T-hot and RCS pressur According to the Plant Cooldown Operating Procedure (OP-1102-11), what is the cause of this condition? (1.5) Regardless of the cause, which action would you consider most appropriate: Select on (0.5) Increase RCS pressure Increase cooldown rate Increase letdown i Open the PORV-End of Section 5-
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. 4 PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATl0H (25.0). The Reactor Building purge rate is limited by ambient air temperatures. Idly? (1.0) The reactor is at 50% with three (3) Reactor Coolant Pump Load is reduced. Since one (1) loop's generator will go on level control before the other, how does the Integrated Control System (ICS) function to permit a further reduction of load after the one (1) generator is on level control? (2.0) Why is letdown entering the makeup tank sprayed in? (1.0) What are the power supplies to each of the nuclear services closed cooling pump motors (NS-P-1A,18, and IC). (2.0)
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pump turbines? Indicate which one(s) of the three are pre-ferre (2.0) Following an ESAS actuation: What must be done to transfer the AC Transfer Switcli -
(ABT) for IC ES Valves MCC from the IP to 1S 480 Volt ^'
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(1.5) List three' (3) trtp signals to the emergency diesel generator that are still in effect?
(1.5) For the cases below, describe how ICS and the plant would respond to the conditions given. Assume each case in inde-pendent, all plant conditions are normal, and control systems are in automati The NSSS Demand from the SG - Reactor increases 10%
while a Btu limit from T(hot) exists on feedwate (1.5) The NSSS Demand from the SG - Reactor increases 10% while the CRD out relay is stuck (no outward rod motion available). (1.5)
.i.8 What are the two (2) design purposes of the cavitating ven-turis installed on the OTSG side of the EF-V-30s? (2.0)
-Section 6 Continued on Next Page-x , suamnwem; n wmmemnaru ,.wmmu,w.-mw : = 1 m ;,:m ; y am - -
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, IRUE or FALSE: The Group Out Limit cannot be bypassed fo any of the eight (8) rod group (0.5)
6.10 On a reactor trip, one {1{ intermediate range channel de-
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creases from 10 s to 10 1 in 7 minutes; the other channel-decreases initially, then levels out at 5 x 10-10 Describe what problems with the compensation settings could have caused these erroneous reading (2.0) 6.11 Under what conditions are the Reactor Coolant Inventory Tracking System (RCITS) hot leg level and Reactor Vessel level indications designed to be used? (1.0) 6.12 Dilute Permit 1 is terminated when Group 5 reaches 80%. Dilute Permit 2 is obtained when Group 6 reaches 95%'. Eng can dilution be enabled when Group 5 is greater than 80%, but Dilute Permit 2 has not yet been obtained? (1.0) 6.13 Where are the controls for the fuel transfer system (2 upenders and carriage travel) located? (1.0) > Describe the three (3) interlocks associated with the fuel transfer syste (1.5) 6.14 At 50% power, loss of the "A" DC Distribution System will . cause a reactor trip; loss of the "B" DC Distribution System will probably not. What causes the reactor trip on a loss of the "A" DC Distribution System? (1.0) 6.15 Why were the orifice rod assemblies (ORAs) removed from the core design prior to the current cycle 57 (1.0)
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6 PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL (25.0) What is the difference in the expected dose rate between a
"High Radiation Area" and a " Locked High Radiation Area"? (1.0) Select the condition from each of the following that requires the operator to manually trip the reacto . Loss of RCP (resulting in gnly 3 running)
< CRD stator temperature 165 F Pzr.levelof295igches Channel A Th of 602 F *
(0.5) . CRD stator temperature of 185 F Loss of 1 main feed pump Loss of Pzr. heaters Condenser vacuum of 26-inches W Hg (0.5) . Dropped control rod Btu limit (ICS) Two MSIVs close .25 gpm OTSG tube leak (0.5)
, What is the rated breathino time for a full air bottle in a " Scott Air Pack"? Select on (0.5) minutes minutes minutes 4 60 minutes
. Can the cylinder be changed safely in a contaminated atmosphere area? (0.5)
.1-Section 7 Continued on Next Page-
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7 Ellt in the following numbers from the list provided below: Select on During boration, boron concentration should be verified, every estimated 7 pp (0.5) Do not attempt to start a reactor coolant pump when power-is greater than ? %. (0.5) Do not exceed ? .% power unless both feed pumps ' and two (2) condensate booster pump pairs are in opera-
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tio (0.5) , Following a thermal power change exceeding ? % of i the rated thermal power within a 1-hour period, a reactor coolant sample shall be take (0.5) . 30 , .5 Why does Operating Procedure 1102-15 (Fill and Drain of Fuel Transfer Canal) prohibit filling the fuel transfer canal through the reactor vessel? (1.0)
: What are the four (4) criteria used to verify natural
, circulation in the RCS? (2.0) During a Natural Circu.lation Cooldown, why does the
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Operating Procedure (1102-16) state that it is better to use the motor- driven rather than the steam-driven emergency feedwater pump? (1.0) l . List the three (3) primary methods at power for determining which OTSG has the tube lea (2.0) ' Under which four (4) criteria can HPI be throttled? (3.0)
. List the immediate manual and automatic actions required in the event of a CRD malfunction where one (1) or more groups ,
of rods are driven out with no command for out motion pres-i ent. Assume the reactor is initially at steady-state, 50% powe (2.5)
-Section 7 Continued on Next Page-
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7.10 At What instrument air pressure should the reactor be tripped? Select on (0.5) pstg psig psig psig
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7.11 On a steam leak what determines whether the operator trips the reactor or shuts down at 10% per minute? (2.0) 7.12 Abnormal Procedure 1203-40 (Loose Parts Monitor System) lists twelve (12) unit parameters that should be checked for abnormalities if a loose parts monitor system alarm is received. Two (2) are RC flow and RC pump vibration. Name five (5) of the r,cwining ten (10) parameters that should be checke " '
(1.5) }
7.13 On an 075G tube leak (ATP-1210-5), if the condenser was not available and OTSG pressure cot not be maintained below 1000 psig in the contaminated generator, which action below should you order according to orocedure as shift supervisor? (0.5) '" Use the atmospheric dump valve to prevent MS safety valve actuatio Wait for automatic MS safety valveTctuation to preclude unnecessarily discharging additional amounts of con-
, taminant .14 What OTSG levels should be maintained after a reactor trip if: The 25 F subcooling margin is lost? (0.5) RCPs are on, adequate subcooling? (0.5) RCPs are off, adequate subcooling? (0.5)
7.15 IRUE or FALSE: If either the Diamond Rod Control or the Bailey Reactor Demand stations are in Hand and a feedwater cross limit occurs, the operator should use the " Raise-Lower" switch, and adjust the reactor power upward to be compatible with the total Feedwater Flo (0.5)
-Section 7 Continued on Next Page-w ._ - w_ , ._--, _ _ _ _ -m - -- -
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7.16 Exnlain why the following minimum power limits exist, a. The CR0 Control Panel should not be placed in AUTO unless reactor power is greater than 5% on the Nuclear Instru-mentation Channel feeding the IC (0.75) b. With the CR0 Control Panel in AUTO, the Reactor Demand Hand auto station should not be placed in AU10 until reactor power is greater than 15%. (0.75)
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8.0 ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS (25.0) 8.1 a. .Per Technical Specification 3.0, on general Limit'ing Conditions for Operation (LCO), what action must be initiated if an LCO is not met? (1.0) - , What is the minimum time pemitted for initiating this - action? (0.5) . ., 8.2 What two (2) special procedural requirements have been estab-
11shed especially for Special Temporary Procedures (STPs) which involve the Engineered Safeguards Actuation System .
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(ESAS) or the Emergency Feedwater System (EF)? (2.0) AJ 5 V i
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8.3 List two (2) of the parameters that are subject to Limiting D/ e(d #cppd 4 Conditions for Operation in the Environmental Technical d p @a-y Specifications, Appendix (2.0) g te ' g#y
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'. 8.4 Temporary Changgs to the valve line'-dps required by Procedure Valve Checklists da not require the use of a TCN (Temporary
- 1 Change Notification). However, what three (3) things must ~
] be done to allow those deviations? (1.5) 8.5 Anytime the Departmental Foreman is not on site, Wha i authorized to sign in the Departmental Foreman's space on the completed tagging application? (1.0) 8.6 A licensed Senior Reactor Operator, with no other concurrent responsibilities, must directly supervise all core altera-- tions. Where can he be located? Select on (0.5) i The Control Room Reactor Building Operating Floor Area Spent Fuel Pool Auxiliary Building-Section 8 Continued on Next Page-
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11 During heatup the plant is at 450UF when it is reported that the block valve for a PORV is shut and cannot be opened until repairs are completed in 2 days. What actions, if any, should be taken? Justify your answe (2.0) e8 The Emergency Director is initiaiiy responsibie fo:-: Classification of an emergency eYen Approving and directing official notifications to offsite agencie Approving and directing infonnation releases to the medi Approving and, if possible, personally conveying appro-s priate Protective Action Recommendations to the Bureau of Radiation Protectio Directing on-site evacuation at the Alert or hower level .~
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emergency classification based on potential hazard to non-essential personnel.
- Authorizing emergency workers to exceed 10 CFR 20 Radi-attor. Exposure Limit C -- Approving and directing deviation from established operat-ing procedures, emergen.cy operating procedures, normal equipment operating limits or technical specifications during attempts to control tne emergenc When the Emergency Support Director (ESD) arrives, which two (2) of the above responsibilities does the ESD assume? (2.0) Under what condition would a person injured in a radio-logically controlled area (RCA) be de-contaminated on-site? (1.0)
8.10 What is the minimum Emergency Classification that requires activation of the Technical Support Center? (1.0)
, Whera is the Technical Support Center located? (0.5) Wher.n is the backup location for the Technical Support Center if the primary location is uninhabitable? (0.5)
8.11 What are Exception and Deficiency (E&D) sheets used for? (1.0)
-Section 8 Continued on Next Page- .
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8.12 During STARTUP, HOT STANDBY, and POWER OPERATION, list three (3) of the six (6) allowable reasons for containment purging to reduce airborne activity to facilitate contain-ment entr (1.5) During STARTUP, HOT STANDBY, and POWER OPERATION, which of the containment parameters listed below may be con-trolled by conducting purge operations? (1.5)
. Temperature Pressure Humidity 8.13 For each of the following situations indicate what REQUIRE-MENT, if any, applies and what ACTION, if any should be taken. Consider each situation separatel s Diesel generator A's operability load test, which is required every 31 days, is scheduled for today. The last three tests were completed 36, 68, and 102 days ago, respectively. The plant is at 100L powe (1.5) ' The plant is at 295 F and heating'[ip at 1 F per minute, when a decay heat removal pump is found _ inoperabl (1.0) . , The piant is at 100% power when it is detennined that the discharge valves.for two emergency feedwater pumps are failed shu (1.0)
8.14 &ni often during refueling operations must startup checks be performed?
(0.5)
8.15 While reviewing logs you note an obvious error. Exclain exactly how you would correct the entry (in accordance with AP-1001G: Procedure Utilization). (1.5)
-End of Section 8- -End of Exam- _ mammmeum e mm - _ m -
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EQUATION SHEET ___. ______..........____.............._....._. __...______._ ..._........
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Where mi = m2 (densityl i(velocity)1(area)1 = (density)2(velocity)2(area)2
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KE = mv2 PE = mgh PEi +KEi +P1Vi = PE +KE where V = specific Y 2 +P 2 Y22 volume P = Pressure
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Q = mcp (Tout-Tin) Q = UA (T ave -Tstm) Q = in(ht -h2 ) ___..__............. _______..____....__________________ .._____ .._...... P = Po1O(SUR)(t) P = Po et/T SUP, = 26.06 T = (B-p)t I p
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delta K =dKe f f-1) CR1 (1-Keffl) = CR 2 Il-Keff2) CR = S/(1-Keff)
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SDM = (1-K,ff)'x 100s Il'Eeff2) K eff _________________________2.ga____..__ ____________._______....._.. ____ decay constant = In (2) = 0.693 A1 = Ag e-(decay constant)x(t) t t 1/2 1/2 _____..__ _________. ______________________________ ...____ .________.. . Water Parameters Miscellaneous. Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons I hp = 2.54 x 103 Btu /hr
Density = 62.4 lbg/f t 1 MW = 3.41 x 106 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec2 ____ ._____...._____ .....__________________..._.._..____. _______________
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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION
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Facili ty: TMI-l Reactor Type: Babcock & Wilcox - PWR
. Date Administered: December 16, 1985 Examiner: W. J. Apley /N. Dudley Candidate: ANSWER KEY q INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. . Write answers on one side onl Staple question' sheet on top of the answer sheet Points for each question are indicated in parentheses af ter the questio The passing grade' requires at least 70% in each category and a final grade of at least 80%. Examination papers wi.11 be picked up six (6) hours after the examination start ~
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Category % of Candidate's % of Value- Total Score Cat. Value . Category- , l 25 _25 5. Theory of Nuclear Power Plant' Operation, Fluids an Thennodynamics 25 25 6. Plant System. Design, Control and Instrumentation , 25 25 7. Procedures - Nonnal, Abnormal, Emergency, and Radiological Control 25 25 8. Administrative Procedures, Conditions, and Limitations 100 TOTALS' i Final Grade % All work done on this examination is my own; I have neither.given nor received ai , Candidate's Signatur . e
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. 1 THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND THERMODYNAMICS (25 POINTSI TRUE FALSE TRUE
' TRUE Reference: Question Bank 013928 B&W ATOG Guidelines, Part II, Vol. .2 With a drop in feedwater flow, less steam is drawn through the aspir-ating port in the OTSG to heat the feedwater. That steam becomes available to the T . Steam used by the tripped turbine-driven feedwater pump is diverted to the T Higher pressure to turbine = increase in generated M Reference: Open Training Manual, Vol. VII, p. III- .3 If it was raised any higher you would not satisfy the minimum NPSH re-quirements for the main feedwater pump Reference: Operator Training Manual, Vol. VI, Condensate Lesson, p. .4 Th and dT will decrease, (+0.6) Tc will increase (+0.4). This is to maintain energy (heat) balance, Q = M dT (+0.5). Feed flow will be adjusted by ICS to maintain Tc's equal (+0.5). 'A'
OTSG will have increased feed flow and 'B' OTSG will have decreased feed flow (+0.5). (OTSG may reach low level limits.) OTSG level will follow feed flow to that OTSG (+0.5). 'A' OTSG
, level will increase and 'B' OTSG level will decrease, again to balance Tcs (+0.5) .,.
Reference: Question Bank 0014103 Oper. Training Manual, Vol. III, Heat Exchanger Heat Transfer, p. 15 ' 12226 ' ._ > ' _ .__ . _ . _ _-
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2 Pu builds up during the cycle until eventually at E0C it may be respon-sible for 40% of reactor power.. Pu has a beta of 0.0021, much less than that of U-235 (0.0065), thereby resulting in a much lower beta-(ffective later in cycle lif Reference: Oper. Training Manual, Vol. II, p. 3 .6 b , (1 - K2) formula sheet CR 2 (1 - K1) CR K2 = 1 - (1 - K1)
= 1 - 5/60(1 - 0.96) = 0.9967 Reference: Formula Sheet; Oper. Training Manual, Vol. II, p. 12 .7 The introduction of one rod distorts the flux distribution and thereby affects the worth of the second rod, since rod worth is approximately --
proportional to the square of the thermal neutron flux at any one poin Reference: Oper. Training Manual, Vol. II, p. 13 .8 As the moderator temperature increases, boron in solution spreads out as the moderation density decreases. A temperature increase with boron in t.ie moderator would cause boron density and moderator density to decrease. This results in an increase in the thermal diffusion length but also a decrease in the number of soluble boron poison atom The decrease in soluble poison in solution and the increase in thermal dif-fusion length, with no change in control rod density, causes more neutrons to be available to the control rods for capture; the higher the boron concentration, the greater the increas Reference: Oper. Training Manual, Vol. II, pp. 143-14 .9 low power, negative MTC Reference: Oper. Training Manual, Vol. II, p. 16 .10 The heat generation after shutdown is no longer primarily due to the kinetic energy of the fission fragments in the fuel; it is now due to the absorption of gamma radiation from fission-product deca Reference: Oper. Training Manual, Vol. II, p. 187.
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5.11 False - actually that is the definition of nucleate boiling Reference: Oper. Training Manual, Vol. II, p. 19 .12 Leak rate is proportional to the area of leak times velocity of fluid leaving pipe. From Bernoulli's equation, pressure is proportional to the square of velocity, so halving pressure will only reduce velocity by square root of two (or drop of about 29.3%). Reference: Oper. Training Manual,-Vol. III, p. 9 .13 Count rate would significantly increase (by a factor of _102 to 104, number not important), and would be erratic as voiding oscillations took place ... both due to increased leakag Reference: Oper. Training Manual, Vol. VII, TMI-2 Accident Description,
. p. 1 _ _ _
5.14 False Reference: OP-1104-9, p. .15 Voids are occurring in the' reactor vessel head due to head water temperature being higher than RCS temperatur Definitely #1 - Increase RCS pressure, other alternatives would aggre-vate the proble .- Reference: OP-1102-11, p. p v:c a c: m a m : --vu.+ - wm.+. s_:.a - -
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4 PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION (25 POINTS) - The containment isolation valves have a relatively high NDTT (60 F).
Number not require Reference: Oper. Training Manual, Vol. VII, p. 23, RB Ventilation System .2 ATc control is blocked and the total flow control circuit is release This total flow controller will modify the loop demand (for genera'.or not on level control) to ensure that the change in feedwater flow is sufficient to satisfy the unit deman Reference: .0per. Training Manual, Vol . VI, p. IV- .3 To absorb hi h gen for subsequent scavenging of oxygen in the primary -- - system ( W c<d;+ ook g for allotM 43 d.c% Reference: Oper. Training Manual, Vol. VI, p. 20, Makeup and Purification System _ . NS-P-1A receives power from the IP 480 volt engineered safeguards bus and motor NS--P-1C receives power from the1480 volt engineered safe-guards bus. Motor NS-P-10's power source can be selected from either
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the IP or IS 480 volt engfiieered safeguards bus. During normal operation, motor NS-P-1B receives power from the IP 480 volt engineered safeguards bu Reference: Oper. Training Manual, Vol. VI, p.10, Nuclear Services Closed Cooling . LP steam from aux, boiler LP from extraction steam (4th and 6th moisture separator) ' HP from main steam LP steam is preferred Reference: Oper. Training Manual, Volume VI, Feedwater System Lesson, p. . p pm - ., , w*e * v Y
* * . . . . . '
5 . Reset ESAS Push ABT Reset button on panel PCR Select ABT to 1S Any three (3) of:
* 2 out of 3 low lube oil pressure running * 2 out of 3 high crankcase pressure running * engine overspeed - -stop-pushbutton-(engine-and-control-room)- Ne 6 +<'P * 86/G Reference: Oper. Training Manual, Vol. VI, Electrical Distribution System, p. 33/ Eme<pc3 'D iot t LoM pk a ((p.1l+L@ a Rx power will increase (+0.3) with FW staying the same (+0.3) until FW-demand > FW-flow + 5% (+0.3) which then causes a cross limit (+0.3)
which will reduce Rx power.to keep power withinJ% of FW flow (+0.3). FW flow will 4dcrease (+0.3) 'and a rod out demand will occur (+0.3) until FW-flow > FW-demand /Rx power +5 % (+0.3) which causes a cross limit (+0.3) to reduce FW flow to within 5% of Rx power (+0.3).
Reference: STM-1-64, p. 17, 1 Question Bank 0013947 To limit EFW flow to an OTSG on a steam line rupture accident: limit containment pressure / Fm: rummt /Uw"^; *" ^W prevent restart if OTSG is overfed Reference: Oper. Training Manuasl, Vol. VI, AFW, p. 3 .9 False - Group 7 (group not required) Reference: Oper. Training Manual, Vol. V, CRDM System, p. CRD-2 .10 The rapid decrease represents the over-compensated instrument; the level-ing off is an indication of no compensation at al Reference: Oper. Training Manual, Vol. II, p. 26 * I4- a also @ p BTa UmLE to ch e & W4 M m T, 1 :. N & M L& * nw - - - ~ = . - - -
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6.11' After a reactor trip: With MQ reactor coolant pumps operating (full credit) RCITS will not be reliable if associated loop vent valve is open (hot leg level) or vessel head vent valve is open (vessel level)
(+0.25 pts credit if only answer #2 is provided) (This response not required for full credit.)
Reference: OP-1103-1, p. .12 A keylock switch labeled DILUTE SIGNAL is provided to enable dilutio This keylock switch is located in System Logic Cabinet 2. (Location not required for full credit.)
- Reference: OP-1105-9, p. 3 .13 One, located i.n the fuel storage building, controls the carriage travel and the upender in the spent fuel pool. The other, in the Reactor Building, controls the upender in the fuel transfer cana Interlocks are provided to prevent moving the carriage when the ' upenders are in an upright position, to prevent moving the upenders when the carriage is not stopped in the proper position, and to prevent lowering the fuel hoist grapple tubes over the upender loca-tions if the upenders are not vertica Reference: RP-1507-7, p. .14 Loss of DC power to EHC system; reactor trips on turbine tri , Reference: EP-1202-9A, p. , 6.15 There were mechanical oroblems with the ORA latching mech . ,. mainers have needed to be installed to ensure positive retention of the source cluster Reference: Cycle 5 Reload Report, p. 6-1.
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7 PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL (25 POINTS) Locked - 1000 mrem /hr High Radn - 100 mrem /hr
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Reference: Oper. Training Manual, Vol. VII, Module 8, Radn. Protection Program, p. 9 .2 I Reference:- Question Bank 0010310 CR Ques. Cat. 4 and 7, No. 39 (0P-204, AP-521) , #2 - 30 minutes No - m.. --ust be changed in a clean area Reference: Oper. Training Manual, Vol. VII, Module GET-103, p. 4 .4 ppm (37 - ,. %
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Reference: OP-1102-10, p. ' OP-1102-2, p. OP-1102-4, p. OP-1102-4, p. 1 .5 It causes a CRUD burst which causes high refueling-' radiation levels.
Reference: OP-1102-15, p. .
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g . RCS AT increases to approximately 30 F to 50 F (degendent on Decay Heat) and stabilizes and Th is less than 600 . Incore thermocouple temperatures stabili;:e and are tracking TH - Cold leg temperatures approach saturation temperature for secon-dary side pressure (normally within 5 minutes). Verify heat removal from OTSG Steam flow indication Feed flow indication Reference: ATP-1210-10, pp. 10 - 1 It eliminates the RCS uncontrolled heat loss (steam to the EFW tur-bine).
Reference: OP-1102-16, p. _ . Sample OTSG (Beta-gamma, H-3, Na-24, I-133, CS-137; Isotopes not d required). (+0.75) ~ Surveying steam lines (+0.5) ~~ Observing OTSG levels and feed rates (+0.75) Reference: ATP-1210-5, p. .8 HPI must be throttled to prevent pump runout (550 gpm/ pump). HPI must be throttled to prevent violation of the applicable brittle fracture / thermal shock curve limitation . HPI may be throttled if LPI flow is greater than 1000 gpm in each line and stable for 20 minute . HPI may be throttled if the required 25 F subcooling margin ' exists and pressurizer level is established greater than 0 i Reference: ATP-1210-10, p. 4.
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9 *a. Verify that the CRD operators console is in manua b. Select jog spee *c. Verify that the GROUP and SINGLE select switches are OF d. Select sequence overrid ~ e. If out motion continues, select the affected group'with the GR0ljP select switch and place the Manual Command Switch in the Insert positio ~
* Expected Automatic Action Reference: EP-1202-08, p. 1 .10 .'. 60 psig Reference: EP-1202-36, p. .11 If SLRDS actuates on either SG or . Continued operation presents a hazard to [ersonnel or equipment required for safe shutdown Then manually trio the reacto If Continued operation is not posing a hazard to personnel or equipment required for safe shutdown but.is severe enough to require shutdown or - RB pressure exceeds 2 psig Then reduce load at 10% oer minut ~
Reference: AP-1203-24, p. . a@
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7.12 R.C. Pressure: R.C. Pump Seal Flow , R.C. Pump and Motor Bearing Temperatures Power Range Power Level b CRDM Drive Temperatures
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' RM-L1 Letdown Filter d/p Letdown Sample Obtain Secondary System Samples for activity 1 Incore thermoccuple indication for possible blocked channel Any five (5): da.3 pts each Reference: AP-1203-40, p. ' 7.13 Reference: ATP-1210-5, p. .14 to 95% Operating Range in. Startup Range % Operating Range
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Reference: ATP-1210-10, p. 5.
' 7.15 False - Reactor power should never be raised to clear FW x limit Reference: OP-1105-4, p. .16 Going into AUTO at a very low power level could result in a continuous rod withdraw signal.
< There is a 15% low limit in the Reactor Demand signa Reference: OP-1105-9, p. . e
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8.0 ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS (25 POINTS) 8.1 Place the plant in a condition where the TS LCO does not appl Within 1 hou Reference: Tech. Spec. 3.0.1, p. 3- Question Bank 001403 .2 The STP must contain a step which requests the Shift Supervisor / . Shift Foreman verify that the redundant system (s), component (s), ' channel (s), etc. is operable (satisfied T.S. requirements) prior to removing the system, component, channel, etc. affected by the STP, from servic The STP must contain a step which requires an independent valve and/or switch position verification check be performed within the -
. boundary of the system p(s) component (s), chann(1(s), etc. affected -
by the STP prior to returning the system (s), component (s), channel (s), . etc. to service. This independent check will provide positive assur-ance that the system is returned to a fully operational statu Reference: AP-1001A, p. 2 .3 d "' Thermal ('c n m ui m * A M W W Y" Chemical Full credit if any two of' three chemical parameters (chlorine, pH, sus-pended or dissolved solids) are liste Reference: Enviror.eent01 Technical Specificctions/ MPDE5 Pe<ma Lc!waPh4 The deviations must be r.eted in indelible marker on the applicable check off sheet and must be reviewed and app.wed by two Licensed Reactor Operators, one of whom must be a Shift Supervisor / Shift Foreman with a SRO License. The deviations mu t be noted in the Shift Foreman's Lo Reference: AP-1001A, p. 2 .5 Shift Supervisor Reference: AP-1102, p. 1 : l
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12 RB Operating' Floor Area Reference: RP-1501-1, p. .7 Verify PORV shut (+1.0) Remove power from PORV (+1.0) Reference: Tech. Spec., pp. 3-18c, 3.1.1 .8 c. and (abe @ ' Reference: EPIP-1004.2, p. 2. (TCM I -?6 - 015'd Tces n e cttterIsl 4 - br ~A Mg4 ip % CUNT The injury was minor enough that it could be treated on-site; if they have to be moved off-site for treatment, any decontamination would be incidental to ensuring the condition of the perso ' Reference: EPIP-1004.16, p. .10 Alert Remote S/D Panel Room (R2 O e ex '- ;c d]) I and C Shop Reference: EPIP-1004.28, pp. 1, .11 To record problems encountered during surveillance testin Reference: AP-1001J, p. 14.
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8.12 During STARTUP, HOT STANDBY and POWER OPERATION: Any three (3) of
, Non-routine safety-related corrective maintenance Non-routine safety-related surveillance Performancc of Technical Specification required surveillances Radiation surveys Engineering support of safety-related modifications for pre-outage planning Purging prior to shutdown to prevent delaying of outage commence-ment (24 hours prior to shutdown) Temperature - No Pressure - Yes Humidity - No Reference: Tech. Spec., p. 3-4/a, .13 Each test is within 25% of required time (+0.35) and each three (3)
consecutive tests are not within 3.25 of required time (+0.4).
Declare DG A inoperable (+0.25). Prove operability of DG B within I hour (+0.3). Conduct load test on DG A (+0.2). Decay heat removal pumps are not required until Rx is critical. No action besides repair of pump is required. (+1.0) c. Within 1 hour initiate shutdow (+1.0) Reference: Tech. Spec., pp. 1-8, 3-2 Question Bank 00791 .14 At least once per shif Reference: RP-1507-3, p. .
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8.15. Asterisk it . Put correct entry by it Sign your name by the explanation Time and date the explanation Have supervisor (1 level higher) review and ensure person making _ error is informe Reference: AP-1001G,_p. 9.
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Where mi = m2 ( den si ty)1( vel oc i ty )1 ( a rea )1 = ( den si ty )2 ( vel oci ty ) 2 ( a rea )2 __ _______________-___________-___________-____-_-______-______--_________ KE = mv2 PE = mgh PEi +KEi +P1V1 = PE +KE where V = specific 2 +P 2 V22
'i volume P = Pressure
_-_______-_______ __________--__-________________-_--_-____-______--____-- Q = mcp (Tout-Tin) Q = UA (T ave -Tstm) Q = m(ht -h2 I ______________-____-_-______________________________________-_____________ P = Pg10(SUR)(t) p = p9et/T SUR = 26.06 T = (8-p)t
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__-_____________________-______________-__ ---_____-________-___________ delta K = (Ke f f-1) CR1 (1-Keff1) = CR2 (1-Keff2) CR = S/(1-Keff)
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M = (1-Keff1) SDM = (1-Keff) x 100% Il-Keff2) K eff x _____________________--______ -_____________________-_-________________-. decay constant = In (2) = 0.693 A 1 = Age-(decay constant)x(t) t t 1/2 1/2 _______________________-____ -_______-_______.__________ _____________-___ Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons I hp = 2.54 x 103 Btu /hr
Density = 62.4 lbg/f t " 1 MW = 3.41 x 106 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/i >f-sec2 __________________-__________-_-__-___ ____-____________ .________________ M . --- .
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