ML20128C610
| ML20128C610 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 05/20/1985 |
| From: | Reyes L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Scholand J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 8505280348 | |
| Download: ML20128C610 (43) | |
Text
{{#Wiki_filter:r .y MAY 2 01985 Westinghouse Nuclear Training Center ATTN: Mr. Jerry Scholand 505 Shiloh Blvd. Zion, IL 60099 Gentlemen:
SUBJECT:
EXAMINATION REPORT On May 2,1985, we mailed you a cover letter with the examination report attached. At that time, we neglected to send you a copy of the examinations and answer keys. Enclosed please find the above mentioned items. We regret any inconvenience this matter may have caused you. Sincerley, ORIGINAL' SIGNED BY 1.. A. REYES L. A. Reyes, Chief Operations Branch
Enclosure:
Examinations andAnswerKey(s)(SR0/R0) cc-w/ enc 1s: DMB Document Control (RIDS) T. West, Plant Training Mgr. B. Boger, Branch Chief, OLB 5 8505280348 850520 PDR ADOCK 050 2 gs g
V A5TER_ QoEsr m s i EMIEb3EC3!
.- [@Cff N bd .
- J b i-i (.k d i
_, { { p.Menf U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION
-h}' thal (.g E rs FACILITY: SNUFFS-1 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 85/04/15 EXAMINER: R.R.FERRELL APPLICANT: _________________________
INSTRUCTIONS TO APPLICANT: i U30 separate paper for the answers. Write answers on one side only. Stcple question sheet on top of the answer sheets. Points for each quaztion are indicated in parentheses after the question. The passing Stade requires at least 70% in each category and a final grade of at loost 80%. Examination papers will be Picked up six (6) hours after tha exaneination starts.
% OF CATE(JRY % OF APPLICANT'S' CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00
________ ______ ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS G.00 __"______ ______25.00 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 23.00 'S PROCEDURES - NORMAL, ABNORMAL, __-_____ __'_I_00 _ _-___-__-__ ________ 7. EMERGENCY AND RADIOLOGICAL CONTROL 2S 00 25 00 ADMINISTRATIVE PROCEDURES, ___I____ ___I__ ___-_______ ___-____ 8. CONDITIONS, AND LIMITATIONS 100.00 100.00 TOTALS FINAL GRADE _________________% All w0rk done on this examination is my own. I have neither sivCn nor received aid. APPLICANT'S SIGNATURE
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 QUESTION 5.01 ( .50)
Racetor power decreases from 1 watt to 0.1 watt in three minutes. The Startup Rate is____ DPM and the reactor period is____ seconds. Choose the correct answer. A.-3,-47 B.-1,-56
.C.-1/4,-96 D.-1/3,-78 i .
QUESTION 5.02 ( .50) If the Keff of a reactor is 1 05, then the reactivity is_____ Delta K/K. P.tcose the correct answer. A. 1378 B. 0005 ' ' C. 04762 D. 0007 GUESTION 5.03' (1.50) Explain how the moderator temperature coefficient changes with rods in the core verses out of the core.
, u QUESTION 5.04 ( .50)
Tho two factors that have the most si'gnificant effect on axial power dist-ribution following a transient are ___________ and ____________. A. Core Burnup, Boron Concentration B. Power Level, Boron Concentration C. Xenon, Power Defect D. Control Rod positione Xenon O
- 5. ' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 DUESTION 5.05 ( .75)
At the point where pump runout is reached, the volumetric flow rate goes to _______ and the system backpressure is ______ than the inlet pressure. It ocn be stated that pump runout is the volumetric flow rate at which the Cvailable NPSH is ______ than the required NPSH. Ch ose the answer which completes the above statement. A.cinimum, more, more 'B.canimum, less, less C.oinimum, less, more D canimum, more, less DUESTION 5.06 (1.50) What are the three bases behind the Control Rod Insertion Limits. QUESTION 5 07 (1.75) h%0ME'. 8Db5 ) hfD OPEG M D M Aisume one Reactor Coolant Pump trips at 30% power. Without a reactor prot-octive system actuation or a change in turbine load, indicate whether the followin3 Parameters will increase, decrease,or remain the same. A. Flow in the OPERATING reactor coolant loops. B.The ratio of the core flow compared to the total loop flow. (Core Flow / Total Loop Flow). C. Reactor Vessel Delta P. D. Core Delta T. E.An OPERATING LOOP Steam Generator P,ressure. C1.753 OU"STION 5.08 (1.50) Hcw does the Differential Boron Worth change with temperature and Doron Cencontration?, Explain and consider each seperately.
- 5. THEORY OF NUCLEAR POWER PLANT DPERATION, FLUIDS, AND PAGE 4
.--- g ; ggy g-------------------------------------- 9UESTION 5.09- (-.50) Th3 power distribution limits (Heat Flux Hot Channel Factors) are hei 3ht d3 pendent per the Technical Specifications. Ch2ose the response that explains why the above statement is true. 0.It is based on a small break LOCA and the limit is most restrictive at J ower levels less than 50%. B. Based on LOCA analysis and is most restrictive at 0-6 feet of fuel rod height. l C. Based on a LOCA where the top of the core is uncovered for a longer period between blowdown and reflood. D.More restrictive at power levels greater than 50% and must not exceed 1.9 kw/ft for the lower 6 feet of fuel rod height. OUESTION 5.10 (2.25)
.A. Define DNB and euplain what core parameters must be monitored to ensure that it does not occur.' C.753 B.How will the following affect the operating characteristics of centrifug-al pumps (IncreaseeDecrease, or Remain the same)
- 1. pump head as speed doubles
- 2. Total. flow as a second pump (in series) is started
- 3. Minimum required NPSH as flow rate increases C.50 es.3 OUESTION. 5.11 (2 00) .
Tho effects of Emergency Boration for 1 minute on power and Teve can be different depending on reactor power. Answer the followingt A.What happens to power and Teve if emergency boration is performed at 100% _ power? Assume Rod Control is in manual. C1.03 B.What happens to power and Teve if emergency boration is performed at 10
'to the -8 amps in the IR with Tave at 547' degrees F. C1.03 i
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 5. y 3 GUESTION 5 12 (1.50) Th? reactor is at 1% power with the Steam Dumps in the steam pressure mode. A rtd withdrawl accident causes bank D to move out at 40 SPM. Explain what h:ppens to the following. Assume no operator action. A.* ave B. Steam Pressure C.50 es.3 C.Ruactor power I OUESTION 5.13 (1.25) What is ' core stretchout' and how is it accomplished? In your answer also Oxplain the effects on the plant if stretchout is performed. QUESTION 5 14 (2.75) The reactor has been operating with the rods at i step above the rod inser-ticn limit for 12 hours wh'ile at 75% power. Answer the following.
^
A. Explain what has occured between the top and bottom of the culo during this time and what happens over the next 12 hours after the rods are C1.503 withdrawn to 220 steps. B.What indication does the operator have of what is happenin3 in the core C.503 during this time? is this condition prevented when operating? C.753 C...sw QUESTION 5.15 (3 00) Hsu will the following paramater chanSes affect control rod worth? Explain. C1 003 A.Tave increases 5 de3rees F , C1 003 B. Boron Concentration increases 50 PPM C. Fuel Burnup over core life C1.003 w _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6 q
DUESTION 5.16 ( .50) Tha SHUTDOWN MARCIN REACTIVITY can best be described by which of the following statements. A. Sum of the Xenon, control rods,and temperature reactivity B. Difference between the Shutdown reactivity and Xenon reactivity C. Difference between the shutdown reactivity and reference reactivity data D. Sum of the Reference reactivity and Xenon plus temperature retetivity GUESTION 5.17 ( .50) Tho predicted critical rod hei3ht on a reactor startup is chosen to ensure which of the following? A. Ensure position is above rod insertion limits and selected to anticipate changing reactivity conditions. B. Ensure criticality is achieved within the liniits of the ECP and selected to anticipate changing reactivity conditions. C. Ensure position is above the rod insertion limits and baron chan3es at the POAH is niinimal. D. Ensures the boron chanSes at the POAH is niinimal and to ensure critical-ity is achieved within the liniits of the ECP. QUESTION 5 18 (2.25) A.What 4 parameters affect radial flux distribution in the core? C1.03 0.What 5 par ameters af f ect the astial power distribution in the core? C1.253 L p.;,Q C h*o p
, ,, -,. q ,
EDUAT.ONS s - REACTOR THEORT RADIATION Ft.UIDS/ THERM 0/ NEAT TRANSFER *
~
- t SUR*t -At .
P = P,e /t = P,10 N = N,e m = A191V1 = A2p2V2
= N
,h.,3-o ort =0-8 Q = A1V1 = A2V2
# A8 A8 I = I 0e * = I 10-x/TVT 0
E = E,g + AEstored k-1 k, - k i 0* k kak: 8 At 1Th = 0.693 E = KE + E + U + pV + Q + W R/hr 6 d feet = d -
*1 I CP.81 1 - k' k<1 point source sc cpst ,1-k2 Ind t* = 12d2 2 b " V, 3 Indt = I2d2 - line source g e -=1-k M R/hr x time = R reduced for - turbine. SG pump. nottle, p , Cps
- Rad x QF = Rea orifice condenser pipe. Rx M Cps flow a (dp a g T% = Rio x kad O
net
- AE8 doppler + " mod + " void Th3g, + TSRad headloss=fh or head loss a V*
c head loss a Ap
+8Ke + 8Se + 8Pu + p = h. + pambient =k D "
Boron + Drod
- 8 fuel +
AP2p hase . APaphase x K g 8 k = f(quality & Pressure) Poisons) Pump laws speed a flow k2 = ka + 3k (speed)2a pressure
&k=k-1 MATH (speed)3a power Q = kAAT = hAaT = UAa7 SUR = 26.06 a y =b -
4 = *
- pat log y b=a Q = mah I'Y e Q = cot y = 3.1 x 10te e los x = c log x AH = m e. AT
- P t . ya log x = log x - log y AU = m c,AT 4 = nv log xy'= log x + log y H = U + pV Defect = Coeff x A Parameter AS =
pV = nRT E111. = 2112 T T2 CV + C2V2 = Ci(V + V2)
~- 6 . PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7 QUESTION 4.01 _ ( .50)
Tha secondary source used in the core is made of ______ which must be re-g:nerated every _____ days. s , Chnose the correct answer. A.Sb-Be,150 B.Pu-Be,300 C.Cf-Be,200 D.Sb-Be,300 9 QUESTION 6.02 ( .50)
. Thimble Plus Assemblies used in the reactor can best be described in the following manner (Choose the correct answer):
- 0. Required to equalize the flow through the fuel assemblies. They project into the upper ends of the guide thimbles and consist of a round plate with long stainless steel rods suspended from the bottoni surface.
B. Required to equalize. flow through the assemblies and limit the bypass flow through empty thimble tubes. They are used in. guide thimble tubes which do not contain control rods. C. Required to bypass flow through the fuel assemblies and equalize flow throV3h empty thimble tubes. They are used in guide thimble tubes which do not contain neutron source rods. D. Required to limit the bypass flow through empty thimble tubes. They are cade of stainless steel and project into the lower ends of the guide thimble tubes. . DUESTION 6.03 (3.00) . List the functions of the following permissives. Include coincidence. A.P-9 B.P-10
'C.P-8 [1.0 ea.3 OUESTION 6.04 (1 00)
Give four inputs that will actuate the ' Computer Rod Deviation
- alarm.
J t
E
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 8 QUESTION 4.05 (1.00) 7 What four actions will initiate a Main Steam Line Isolation. Include set-points as applicable.
QUESTION 6.06 (1.50) Antwer the following TRUE or FALSE concerning the ECCS systems. 1.The ECCS cannot accept an active failure during the recirculation phase end still fulfill its design function. 2.The Charging pumps are desighed to deliver 150 GPM/ pump at 2500 psis and 550 GPM at 600 psis during an SI condition. 3.The SI pumps miniflow valves cannot be opened during hot les recireviat-ion unless the RHR discharge to charging and SI pump suction cross-ties (EJ-HV-8804A/B) are closed. 4.The Accumulators are maintained between 502 and 648 psig with a Boron concentration between 1900 and 2100 PPM. 3.lhe Containment Spray system no:rles are arranged such that one train of the CSS will spray at least 75% of the containment operating deck. 6.The hydrogen purge system would not be required unless there is a failure of both recombiners and would be manually initiated 9 days after a LOCA. QUESTION 6.07 (2 00) A.4 hat are two purposes of the No.2 RCP seal? [1.03 B.Why is the No.3 seal desisped as a double dam seal? Explain the operation [1.03 of the seal in your answer. QUESTION 6.08 (1 25) Explain how the Component Cooling Water system is designed to be protected from a leak / rupture in the Thermal Barrier Heat Exchanger.
. _ _ . . , . . _ _ _ . , . - . ,. ,, ._ ,___v_._. _ _ . _ _ _ . _ . _ _ _ . - _ , _ , - . _ ~ , , , - _ _ , , . , - - - , _ . . , -, _ - - --
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 9 QUESTION 6.09 (1.50)
For the following questions concerning the RCS Pressure Control System, fill in the blank with the correct pressure:
- 1. Normal RCS system pressure is______ psig.
- 2. Variable heaters are completely off at______ psig.
- 3. Pressurizer sprays start opening at _______ psis.
- 4. Pressurizer sprays are full open at _____ psis.
S.The Power operated Relief valves open at _______ psis. 6.A Reactor high pressure trip occurs at ______ psis. 7.The Pressurizer Safety Valves open at _______ psis. 8.The RCS design pressure is _______ psis. 9.RCS hydro test pressure is _____ psi 3 10.A low pressure safety Injection occurs at _____ psis. C".15 ea] QUESTION 6.10 (2.00) Tho reactor is operating at 75% reactor power with all systems in automatic control. For the following failures, explain the plants response and.indic-oto what reactor protection signal (if any) will cause the reactor to trip. Consider each independently and assume no operator action. A. Controlling Pressurizer level channel fails high. E1.0]' B.Contro11in3 Steam Generator level channel fails low. C1.03 QUESTION 6 11 ( .50) The automatic shift of the Auxiliary #Feedwater pump suction from the Condensate Storage Tank to the ESW system requires what conditions? Cncose the correct answer. A. Low level in the CST and AFAS present B. Low suction pressure on 2/3 sensors and AFAS present C. SIS present and low suction pressure on 2/3 sensors D. Low level in the CST and low suction pressure on 2/3 sensors t
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 10 QUESTION 6.12 (1 25)
A.What three actions are initiated on an undervoltage condition on a safe-guards electrical bus? [.753 B.What two conditions / events actuate the shutdown sequencer associated with the safeguards electrical distribution? E.53 QUESTION 6.13 (1.00)
' Whi t are 4 conditions that are required for the Emergency Diesel G e n e r a,t o r output breaker to automatically close?
QUESTION 6 14 (1 50) A.What are the 4 permissives associated with the Steam Dump System? Explain the function of each in your answer? [1.03 D'.What is the function of the HIGH-1 and HIGH-2 bistables associated with the load rejection controller? [.53 QUESTION 6.15 (1.00) What 5 protective features will override the automatic SGWLCS signals? QUESTION 6.16 ( .75) Th* Gaseous Radwaste System collects radioactive fission gases which are pt .marily _________ (half-life of 9.09 hours) and __________ (half-life of 10.7 years) while the non-radioactive gas collected is mostly_________. Choose the correct answer. A.Xe-135, krypton-85, Helium B.Xe-135, krypton-85, Hydro 3en C.Cs-137, Krypton-85, Helium D.Xe-135, Radon-75, Hydrogen
, , ., --- . - . r-,..,v,- y a- y,- , - - . m,-- p-g-, w n e - ,-. -,,
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 11 QUESTION 6.17 (2.25)
A.What 3 conditions (interlocks) must be met to open the LETDOWN ISOLATION VALVES in the CVCS system from the main control board? E.753 B.What conditions must exist in order to manually close the sanie valves from the main control board? E.753 C.What is the basis for these interlocks? E.753 QUESTION 6.18 (1.00) Residual Heat Removal pump 'A',is used during the recirculation phase of a lots of coolant accident to supply water to the suction of the Centrifugal Charsins pumps and SI pumps. The supply of this recirculated water is thro-ush valve HV-8804A. Answer the followins based on the above statement (..What. conditions must be met to open HV-8804A? E.75$b B.What are the two bases of these interlocks? C'.'50Q sh QUESTION 6.19 (1.50) A.What two signals will senerate a Containment Spray Activation Signal (CSAS). Include setpoints and coincidence. E1.003 Dpf.WhattwowayscantheContainmentSpraypumpsbestoppedwhentheyare E.503 started on a CSAS signal?
d
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12
~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~ Rd656E6G5CAE~C6NTR6L r
DUESTION 7.01 (1.50) Cnower.the following TRUE or FALSE concerning Radiation Exposure. A.At a radiation exposare in the lethal range of 1000 REM, death is certain. B.Just above 200 REM, or the threshold for death, a small number of people will die within 30 days. C.The body burden of a particular radioactive species is simply the total amount of radioactivity accumulated within the body over a lifetime. D. Somatic effects ~of radiation can be passed on to offspring of the exposed person. E.A change in a cell's original genetic code is a mutation,and is passed on to subsequent generations. QUESTION 7.02 (2.00)
'A.What are three indications the SRO will use to identify which Steam Gen-
. orator has the ruptured' tube per E-3,' Steam Generator Tube Rupture" pro-cedure? [1.503 b.What is the RCP tripping criterion per the above procedure? E.503 f
QUESTION 7.03 (1.00)
. List three conditions when the shutdown banks must be fully withdrawn when ti' ? reactor is shutdown? [1.03 4
QUESTION 7.04 (1.50) List.5 conditions when emergency boration must be initiated. QUESTION 7.05 (2.25) i Por SF-0-01, ' Failure of Control Rod Bank (s) to Move', answer the following quastions. ! -A.What are three immediate operator actions if the rods fail to move in i cutomatic? C.753 l B.What conditions would have to be present to declare the rod inoperable? C.753 L C.What-are two ways Tave can be maintained when the rods fail to move?t.753 l I I l i
, ,. .., , . . . . , , , . -. , . . , . _ . , _ , _ _ .,, , , _ , , , , , . , . . - _ . ,.c.. . . . _ . . . . . . , , . , ,,,c,
- 7. PRCCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 13
~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~Rd656L6G5656~66NTR6L
. QUESTION 7.06 ( .75) Th9 ~10CFR20 limits for radiation exposure is as follows (fill in the blank) A.____ REM /Guarter,Whole Body B.____ REM / Quarter, Hands and Forearms C.____ REM / Quarter, Skin E.25 ea.] DbESTION 7.07 1.00) g Under what 3 conditions /c'ar -a-T'icensee permit an individual in a restricted cros to receive a to occupational dose to the whole body greater than the 10CFR20 limits? Assume normal operating conditions only. GL'ISTION 7.08 (1.50) Per procedure ES-0.2, ' Natural Circulation Cooldown*, answer the following quastions. , A.What cooldown rate limitation is imposed for the RCS cold less? E.53 B.What levels must be maintained in the Steam Generators? [.53 C.When depressurizing the RCS, what subcooling margin must be maintained? C.53 I.
. - , , v . . , . - - . - - - . - , , _ -
c- - . - .
- 7. ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 14
~~~~
T66L'~~~~~~~~~~~~~~~~~~~~~~~ RA6f6L66ICAE"_66 QUESTION 7.09 (2.00) Tha plant is in the following conditions:
- 1. Cold Shutdown, RCS on nitrogen pressure control (Pressurizer to PRT via the PORV's)
- 2. Letdown is via RHR to CVCS via crossconnect valve BG-HCV-128
- 3. Letdown flow is 75 GPM
- 4. Pressurizer level is 80%
- 5. Positive Displacement Charging pump running to maintain pressurizer level constant.
Answer the following questions concerning a plant heatup: A.Name 2 indications the operator will use to determine that a bubble
-exists in the pressurizer. E.753 B.Name 2 conditions that must be satisfied prior to starting a RCP durin3 heatup. [.753 C.How is RCS temperature maintained while on Nitrogen pressure control?E.53 I
DUESTION 7.10 (2.25) Por E-0, Reactor Trip or Safety Injection,how are the following verified by the operator to have actuated properly as part of the immediate actions? A. Reactor Trip E33 C.753 B. Turbine Trip E23 E.50] C.eeedwater Isolation E4] [1.03 ENOTE Number in [3 is required number,of responses] QUESTION 7.11 (1.25) Whon enterin3 ES-0.2, NATURAL CIRCULATION C00LDOWNethe first thing attempted is to restart a RCP after establishing the proper conditions. List 5 of the conditions that have to be established prior to starting the RCP. QUESTION 7.12 (1.00) List 4 symptoms the operator would have on a loss of a Main Feedwater Pump with power at 75%. l l i n w ------tw-1g ,e --g-e-y ,9-- w- pc---e m- *.*w-- -p,-- - -- .v. -r,w gw--. m o - - - - , ew
- 7. PROCEDURES -' NORMAL, ABNORMAL, EMERGENCY AND PAGE 15
~ ~~~~R I656L6656kL 66UTR6L'~~~~~~~~~~~~~~~~~~~~~~~
GUESTION 7.13 (2.50) A.In addition to possible annunciator alarms, list 4 indications of a drop-ped rod. [1.03 B.What are the immediate operator actions specified in SF-0-5, Dropped Rod Procedure? [.753 C.If the QUADRANT POWER TILT RATIO is calculated to be greater than 1.10 due to a dropped rod at a reactor power of 65%,what is the time limit specified in Tech Specs for reducing power and how far must power be red-Oced in order to satisfy the' action statement.? E.753 00ESTION 7.14 (1.25) A.During recovery from an inadvertent SI, list the 3 conditions which require the operator to manually reinstate SI [.753 b.Why can't the operator rely on automatic reinitiation? [.503 QUESTION 7.15 (2.25) During a normal startup, A.Why is the reactor brought up to 6-8% power before the generator is sync-hronized on to the power grid? E.753 B.Wher. are-the Steam Dumps normally switched to the Tave mode? E.753 C.What limits the loading rate of the turbine /senerator? Elist 33 E.753 2 QUESTION 7.16 (1.00) WSct are 4 conditions that have to be established in the plant before GEN-N
-08, ' HEAT BALANCE CALCULATION', can be performed? [1.03
1
. l
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 16 QUESTION 8.01 ( .50)
Tha Technical Specification Limit for the Guadrant Power Tilt Ratio is less than or equal to_____. A.1.07 B.1.05 C.1 02 D.1.04 QUEbTION 8.02 (1.75) A.What are the four categories of 4Lte demergencies? [.503 B.When is the on-site emer3ency organization called out? [.253 C.When is the corporate en.ergency organization activated? C.253 D,.What are 3 responsibilities the emergency coordinator can not dele 3 ate? E.753 QUESTION 8.03 ( 1. 0 0-) Concerning EIP-ZZ-00101,'Callaway Plant Emergency Plan Implementing Pro-Ledure*, answer the following TRUE or FALSE. A.The above procedure is initiated when alarms, abnormal instrument readings or reports of conditions that indicate an emergency situation (either real or potential) has occored. B.9ecisions of who will receive doses in excess of occupational limits in
' .ife-saving situations will be made only by the Emergency coordinator.
C.Upon classification of an emergency,the Shift Supervisor will assume overall emergency management resportsibilities as the emergency coordinat-or. D.All non-essential personnel have to be evacuated from the site within 30 cinutes of a declared SITE EMERGENCY. C.253 DUESTION 8 04 (1.50) List 6 items'the Operating Supervisor and Unit Supervisor (SRO) have to do in order to relieve the off-going shift per CAP-01,' Shift and Relief Turn-ovGr'.
l l l B. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 17 QUESTION B.05 (1.00) ,. Precedure CAP-02,* Routine (Shift and Daily) Surveillance Requirements *, sets tha surveillance frequency for the following items which have limits per Technical Specifications. What are the limits established in Tech Specs? A. Minimum Temperature for Criticality. + B. Shutdown Rods fully withdrawn. C.RCS pressure /Tave D. Pressurizer Water level. QUESTION 8.06 (1.00) A.What is the required action if the Tech Spec LCO for the Centrifugal Charging pumps can not be met when in Modes 1,2,3, and 4? C.503 B.What are the surveillance requirements for the Centrifugal Charging pumps when in modes 1,2,3, and 4? C.50] GUESTION 8 07 (1.00') 3 List ,A' Technical Specification limits on the Accumulators when in Modes 1,2 ond 3 (Greater than 1000 psig). co0WW "" QUESTION 8.08 (1.50) What are the Technical Specification requirements for A.C. Sources when in Mndes 1,2,3,and 4? Be specific and list all requirements for operability. QUESTION 8.09 (2.00) Par Section 6.0, Administrative Controls, answer the following TRUE or FALSE. _A.The Fire Brigade can include more than 3 members of the minimum shift crew necessary for safe shutdown of the unit. B.The composition of the Fire Brigade may be less than the minimum require-ments to account for unexpected absences of on-duty personnel. C.The minimum shift crew composition when in Modes 1,2,3,or 4 shall be 1 STA,2 RO's,1 SRO,1 SS and 2 E0's. D.The shift crew composition may be 1 less than the minimum requirements for a period not to exceed 2 hours.
- E.The Operating Supervisor may leave the MCR when the controls are being canned by a licensed Reactor Operator to confer with the Shift Supervisor for a period not to exceed 5 minutes.
I . l l
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 00ESTION . lB .10 (2.00)
Answer the following TRUE or FALSE concerning Tech Specs. A. Noncompliance with a Specification shall exist when the requirements of the LCO and associated ACTION requirements are not met within the specif-ied time intervals. B. Entry into an OPERATIONAL MODE or other specified condition shall not be onde unless the conditions for the LCO are met without. reliance on pro-visions contained in the ACTION requirements. C. Failure to perform a Surveillance Requirement within the specified time interval does not constitute,a failure to meet the OPERABILITY require-cents for a LCO. D.It'all Pressurizer power-operated relief valves ( PORV's) and their assoc-isted block valves are not operable in Modes 1,2,and 3,the reactor shall be placed in hot standby within the next 6 hours. E.With any RCS leakage greater than the limits specified in Tech Specs,exc-luding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours. DULSTION 8.11 (1.50) Khot are 3 requirements that have to be met to make changes to the Emergen-cy or Security Plan implementations? [1.503 QUESTION 8.12 (2.50) AC4 hat are the two safety limits associated with Technical Specifications? Explain the basis of each in your answer. [2.03 B.If these' safety limits are exceeded)the Tech Spec requirement is to be in HOT STANDBY within ______ . C.53 Choose the correct answer. A.15 minutes B.30 minutes C.1 hout D.2 hours
E . -
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19 QUESTION 8.13 ( .50)
HOT SHUTDOWN is defined as bein3 a Heff.of______or less, a % rated thermal power of 0% and an average coolant temperature between_____ degrees F. Choose the CORRECT answer to complete the sentence. A. .95, 350 and 200 B. .95, 200 and 140 C. .99, 500 and 350 D. .99, 350 and 200 l QUESTION B.14 (1.75) A.At a power level of 80% it is noted that AFD has exceeded the target band-by 2% for 78 minutes on two channels. What corrective action is required? [.753 F.If the condition in part A above had occured at 36% power,what would have been required? Explain the reason for your answer. [.503 C. Explain why the AFD target band changes from plus/minus 5% at BOL to plus a. 3 minus 12% at EOL. [.503
' QUESTION 8.15 (1.00)
Accume a weekly surveillance is performed according to the followin3 sch-odule at noon : l Monday,B/1 Tuesday,8/9 Tuesday,8/16 ' j Wednesday,8/24 His the Tech-Spec surveillance interval been properly complied with? Explain your answer. QUESTION 8.16 (2.50) A.How many hours is a licensed operator permitted to work in a 48 hour per-iod? [.53 B.How many consecutive days is a licensed operator permitted to work? [.53 C.How many fire teams must be on duty at all times? C.53 D.Does the Fire Brigade Leader (Fire Chief) have any authority to require the plant to be shutdown? '[.53
'E.Who assumes the duty of Fire Brigrade Leader on the back shifts when the
- need arises? [.53
-.-,--w..,w.--.--,...,-,...-,,~.,,,._e ,, ,_-,.,,..m.y , ,_,_...,.__v., _ . . , , , _ _ _.. , , _ , . _ , , _ _ _ _ _ , _ , . .
f, ' . B. ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS PAGE 20 QUESTION 8.17 (2.00) A.For each of the following leak locations, state the maximum allowable rate of leakage of reactor coolant specified in the Technical Specifications.
- 1. Unknown location 2.RHR valve packing leak with leakoff line 3.Through a wall crack in the line between the pressurizer code safety valve and the pressurizer 4.A Steam Generator tube E.375 ea.]
B.What is the basis for the Technical Specification limit on maximum RCS octivity? C.503
+ .
f y .w-- --, e - ey rm.,- - - - - - . -r.,- -- -- r-*- - - -, ,_ , -- , , , -. ,- v e ---
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 21
~~~~ TUERE66ENAU C5~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 5.01 ( .50) D. REF. Fundamentals of Nuclear Reactor Physics, p.7-20 ANSWER 5.02 ( .50) C. I REF. Fundamentals of Nuclear Reactor Physics, p.5-21 ANSWER 5.03 (1.50) T72 moderator temperature coefficient is more negative when the control rods are in the core. Increasing the temperature decreases the waters mod-orating ability and thus increases the mi3 ration length of the neutrons. In
'ths rods out condition, the increased migration length increases leakage only at the cores periphery. At elevated temperatures the migration length increases the rods ' sphere of influence". With the rods'in the core, increasing the temperature will increase the probability of neutron leakage into the rod and loss of fission chain reaction.Therefore, for a Siven tem-parature change, more negative reactivity is inserted when control rods are in the core.
Wlth rods out, the PWR has little Buckling because of its size. Therefore o.i increase in slowing down length and. diffusion lenSth has little affect on the moderator temperature coeffici.ent. Placing rods in the core forces flux to the perephery causing Buckling to increase. The change in Ls and L+h due to a temperature change becomes more signficiant causing the
'anderator temperature coefficient to become more negative.
Caither answer acceptable] l REF. Reactor Core Control for Large PWR,P.3-22,23 l
' ANSWER 5.04 ( .50)
D. REF.Rx Core Control For Large PWR, p.8-19 i l 4
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2?
CNSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 5.05 ( .75) B. REF. Thermal-Hydraulic Principles and Applications to the PWR II,p.10-43,44 ANSWER 5.06 (1.50)
- 1. Sufficient negative reactivity must be available to achieve the required SDM at all times,especially in the event of a steam rupture accident
- 2. Minimize the positive reactivity which could be inserted should a rod ojection accident occur
- 3. Provide an acceptable radial flux distribution to minimize peaking factors E.50 ea.]
REF.SNUPPS Control and Instrument Systems,p.3-4 AN WER 5.07 (1.75) A. Increase B. Decrease C. Decrease D. Increase E. Decrease E.4 ea.3 REF. Reactor Core Control for Large PWRep. ANSWER 5.08 (1.50)
-With little boron concentration,there would be little competition for nGutrons and a high probability that each boron atom would absorb a neut-von. -As the concentration increases baron ato.ms are in greater competition for n2utrons which shields other boron atoms -Therefore,as boron conctntration increasec,the Differential boron concen-tration decreases [0.53 -As temperature decreases,there is a greater mass of boron in the core for the same PPM and more reactivity per PPM -A3 temperature decreases, Differential Baron Worth increases -Ac temperature increases, Differential Boron Worth decreases REF.SNUPPS Plant Operations,p.3-38,39
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 23 g
ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 5.09 ( .50) C. REF.SNUPPS Plant Operations,p.4-9 ANSWER 5.10 (2.25) A.The point of maximum heat tr'ansfer sustainable with nucleate boilinst.253 cperator monitors coolant temperatures, primary system pressure, core flow rates,and reactor power level E.503 B.1. Increases Ep.10-373
- 2. Remains the same Ep.10-473
- 3. Increases [p.10-563 E.5 ea.]
REF. Thermal-Hydraulic Principles app.to the PWRep.13-23,24 ANSWER 5.11 (2.00) A. Power decreases at first due to boron Primary to secondary mismatch causes Tave to decrease Decrease in Tave adds positive reactivity Power is restored to original value with a lower Tave B.Tave is determined only by RCP heat and Steam Dumps power decreases at a -1/3 DPM rate to the source ranse REF. Reactor Core Control for large PWR's 4 ANSWER 5.12 (1.50) A. Increases with power B. Stays constant as Steam dumps actuate,to m,ain,tain .
'-[. ;',g (n. ,
wa pin ,t> g , . t A t. d ; r< .S. erc t-t. sni6 o i,6o N C. Power increases due to rod withdrawl until P-10 which causes a trip REF.
F-
.5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 24 ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 5.13 (1.25)
A.EOL when it is possible to continue operations at less than full power to oxtend core life [.203
'A110 win 3 Tave to decrease to the minimum allowable level E.203
+ reactivity from decreased Tave will maintain power for limited period When Tave is at lower limit, power is decreased to maintain Tave in allow-able ranSe C.453 B. Increases possibility of shorter life in subsequent fuel cycles [.203 oay create undesireable neutron flux distribution in next cycle C.203 REF. Reactor Core Control for large PWRep.2-16 ANSWER 5.14 (2.75)
A.1. Flux is redistributed and forced toward the bottom (with rods in)
- 2. Production rate in top decreases and Xenon concentration increases due to decrease in burnoutiburnout in bottom increases and formation of I-135 increases to decrease Xenon for next 6.7 hours.
- 3. As forn ation of Xenon in bottom increases, flux is pushed to top and Xenon oscillations will occur between top and bottom for a period of approximately one day (after rods are withdrawn) [1.503 B. Operator can observe Axial Flux Difference (Delta-I meters) CO.53 C. Operate with rods out esure Delta-I is maintained in band C.753 REF. Reactor Core Control for Large PWRep.4-29
)"A.Ineweases'"'Eam*s.
c, P d?k.s ' * [.253:mederator-becomes-less a s M dense, -a Aneutrons
& f' can travel further,
" hisheo-pr-obabili-ty of-reaching a control rod (increased sphere of influ-ence) C.753 B .op ressys E.253:iperease egncentration shifts flux spectrum more to the I .;ft keYdil e n e r g y Ea'r'ig e'.' 'd o'ni r 8 I r o d s a r e g o o d a b s o r b e r s i n t h e e p i t h e
~
cal range E.753 C. Increases E.253 Explanation same as above E.753 REF. Reactor Core Control for Large PWRep.6-22 thru 28 {'A 1:%-E baJt50
50 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 25 g g g-------------------------------------- ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 5.16 ( .50) Co REF.Rx Core Control for Large PWR'sep.7-13 ANSWER 5.17 ( .50) Ao / REF. Reactor Core Control for Large PWRep.7-24 ANSWER 5.18 (2.25) A.1. Fuel assembly loading pattern f(u( D 8j TiW , ,)
- 2. Burnable poision loading pattern [,.tj :, M:. IV. :its ^It?. i. hbi { " . ~
- 3. Control rod pattern , p tait.,pp.(p,pelfy G.DlM 4 C E.25 ea3
- 4. Neutron leakage Boi. Rod Height
- 2. Moderator density variations
- 3. Power level
- 4. Fission product poision concentration
- 5. Fuel depletion variations along the length of the rod E.25 ea3 REF.Rx Core Control for Large PWR's, p.29-32 i
l i
. , . - - - . _-,e , _ - - - - . _ _ - , .,, - _ -, - , _ . .
9 4
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 26 ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 6.01 ( .50)
D. REF.Rx Core Control for Large PWR e p .1 -i3 7 ANSWER 6.02 ( .50) B. REF.Rx Core Control for Large f'WRep.1-39,40 ANSWER 6.03 (3.00) A.P-9: Enabled below 50%,2/4, auto blocks reactor trip following turbine trip B.P-10:1. allows block of PR low reactor trip at 25%,IR high reactor trip at 20% power,20% rod stop
- 2. auto block of SR high level trip at 10e5 CPS
- 3. Input to P-7 12/4 PRNI]
C.P-8: auto blocks single loop loss of flow reactor trip-enabled below 48% on 2/4 channels REF.SNUPPS Control and Protection Instrument Systems,p.3-11 ANSWER 6.04 (1.00) 1 Rods operating out of sequence
- 2. Deviation of + or -12 steps between any rod and its bank demand
- 3. ' '
two rods in the same bank 4.Any shutdown rod below 220 steps REF.SNUPPS Control and INSTRUMENT sYSTEMSep.3-11 ANSWER 6.05 (1.00)
- 1. Low steam line pressure at 585 psis 2.High steam press rate-110 psi in 50 see (with stm press SI blocked)
- 3. Containment pressure,hi 2-17 psis
- 4. Manual REF.SNUPPS Engineered Safeguards Systems,p.1-21
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 27 ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL
~
ANSWER 6.06 (1.50)
- 1. FALSE [p.2-43 2.TRUE [p.2-123 3.TRUE [p.2-163
- 4. FALSE [p.2-243
- 5. FALSE Cp.3-83 6.ThdE [p.4-13,143 REF.SNUPPS Engineered Safeguards Systems 1
ANSWER 6.07 (2.00) A. Backpressure to force water from No.1 seal into return line Backup for No.1 seal E.5 ea3 E. Permits injection of Reactor Makeup Water at a slightly elevated press-ure between the dams to provide the No.3 seal with clean water for lub-rication E.333 and to prevent dissolved radioactive gases in the dischar-se fluid of the No.2 seal from entering the containment atmosphere C.333. A portion of the flow goes into the cavity between the No.2 and 3 seal and then out the No.2 seal leakoff. The remaining flow is discharged through the No.3 seal leakoff into the normal containment sump E.333 REF.SNUPPS Reactor Systems and Components,p.4-16 ANRWER 6.08 (1.25) 1.Lheck valve in the CCW supply line to prevent backflow of RCS
- 2. Pipe from check valve to flange connection at RCP is designed for 2485 PsiS 3.FT-17,18,19,20 will shut MOV (BB-HV-13,14,15,16) downstream of the thermal barrier on a high sensed CCW flow in excess of 50 GPM
- 4. Common return line from TBHx MOV (EG-HV-62) will automatically close on o combined CCW flow of greater than 200 GPM as sensed by FT-62
- 5. Relief valve located between check valve upstream of each thermal barrier heat exchanger and the motor operated high flow isolation valve (BB-HV-13,14,15,16) downstream of each heat exchanger. Relief valve will relieve excess pressure caused by thermal expansion if high flow isolation valves cre closed. C.25 ea.3 REF.SNUPPS Reactor Systems and Components,p.4-19 l
- 4. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
.____' PAGE 28 ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER ~ 6.09 (1.50) 1.2235
-2.2250 3 2260 4.2310 5.2335 6.2385 7.2485 8.2485 ,
9.3107 10.1849 E.150 ea3 REF.SNUPPS Systems and Components,p.6-3 ANSWER 6.10 (2.00) A. Charging flow decreases
.Dressurizer level decreases (Letdown isolation / heater cutoff at 17%
Pressurizer level increases Reactor trip on high level B. Level increases until S/G high level (78.1%) Feedwater isolation MFP and Main Turbine Trip Turbine Trip (P-{)> 50% power causes Reactor Trip REF.SNUPPS Control and Protection Instrument Systems (NPS) 227 Chapter 5,NPS 215 and 223, Chapter, 6 ANSWER 6.11 ( .50) B. REF.SNUPPS Steam Cycle Systems,p.5-15
1 T
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 29 ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 6.12 (1.25)
A.1. Diesel Generator start signal generated
- 2. Load shed signals remove all loads from bus except 480 load centers and Centrifugal Charging pumps
- 3. Blocks LOCA sequencer from starting directly on SIS or CSAS [.25 ea.3
! B. Diesel Generator output breaker closed Preferred,lsource breakers open E.25 ea.3 Tua w.u REF.Snupps Electrical Systemsetp.3-14 ANSWER 6.13 (1.00) 1.Both feeder breakers to associated ESF bus open a 2.DG Master Transfer Switch in auto
- 3. Generator up to voltage E4.16 KV3
- 4. Generator up to speed [>471 RPM 3 5.No lockout relays energ.ized E.2 ea.]
REF.SNUPPS Electrical Systems,p.4-21 ANSWER 6.14 (1.50) A.P-4: shifts Tavs mode from load rejection to plant trip function E.253 C-9:(Condenser available interlock) protects the condenser from overpres-surization by blocking steam dump actuation during periods of insuff-icient vaccum C.253 C-7:(Loss of Load interlock)-arms the steam dumps for operation followins a load rejection of greater than 10% in 120 seconds E.253 e P-12:(low-low Tavs interlock)-closes al dumps when Tave reaches 550 degrees F, preventing an uncontrolled cooldown from occuring B.HIGH-18 trips at deviation signal equal to 50% steam dump demand; trip solenoid valves for groups 1 and 2 dump valves energize to open valves HIGH-2: Trips at deviation signal equal to 100% demandigroups 3 and 4 dump valves full open Eboth bistables reset when condition clears 3 REF.SNUPPS Steam Cycle Systems,p.4-6
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 30 ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 6.15 (1.00) 1.canual 2.P-14 3.SI 4,Rx trip P-4 coivncident w/a low low Tave signal (564 degrees F) 5.S/G lo-lo level on 2/4 channels on 1/4 S/G's [.2 es.]
[ctincidence/setpoints not necessary for full credit] REF.SNUPPS Steam Cycle Systems,,p.6-22 ANSWER 6.16 ( .75) A. REF.SNUPPS Reactor Support Systems Part II,p.3-5 ANSWER 6 17 (2.25) ..e , A.1.All LETDOWN ORIFICE ISOLATION VALVES must be W. open 2.Pressuri:er level greater than 17%
- 3. Proper air pressure and control evoltage must be available B. LETDOWN ORIFICE ISOLATION VALVES must be closed C. Ensure Regenerative Hx always at RCS pressure which will prevent steam '
flashing and dama3e the tubes RT .SNUPPS Rx Support Systems Part Iep.1-7 ANSWER 6.18 (1.00) l A.1. Valves HV-8701A or PV-8702A, suction valves from RCS hot les loop 1 must be closed AND 2.Either both of the SI pump recirculation line isolation valves (HV-8814A/B) must be closed OR 3.The SI recirc line header isolation valve (HV-8813) must be closed B. Prevents overpressurization of Charging /SI pump suctions whenever one of the RHR pump svetions are aligned to its respective hot legs and prevents radioactive recirc water from being pumped to the RWST via the SI mini-flow lines REF.SNUPPS Rx Support Systems Part I,p.4-18
60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 31 ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 6.19 (1 50) A.1.2/4 Containment pressure high at 27 psig 2.Either of 2 sets of 2 switches on MCB are taken to activate position E.375 ea] Ba1. Pull to lock
- 2. Reset CSAS and stop E.375 ea3 REF.SNUPPS Engineered Safeguards Systems,p.3-11 l
4
i
~
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32
~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~Rd6I6L6656AL'66NTR6L ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 7.01 (1.50)
A.TRUE C.3-233 B.TRUE Cp.3-233 C. FALSE Ep.3-283 D. FALSE [p.3-303 E. fat 4E [p.3-303 E.3 ea.3
.T E S ,
REF. Radiation, Chemistry,Corros' ion Manual ANSWER 7.02 (2.00) A.1. Unexpected rise in iny SG narrow range level 2.High radiation from any SG sample 3.High radiation from any SG steaniline 4.High radiation from any SG blowdown line E3/4 at .5 ea.] B.less than 1320 psis v - a- c s e'. a ?.* v ' 'i i w *^ E.503 REF.E-3,SG Tube Rupture ANSWER 7.03 (1.00) Whonever positive reactivity is being added by*1. boron cone. changes
- 2. Xenon cone. changes 3.RCS temp. changes
- 4. control bank rod movement
[2/4 at .5 ea.] REF. GEN-N-01,p.2 2 ANSWER 7.04 (1.50) A. Control rod bank height below th ROD BANK LO-LO LIMIT alarm with the reactor critical B. Failure of 2 or siore rods to insert on a trip or shutdown indicated by IRPI C. Uncontrolled cooldown of RCS following a trip D. Uncontrolled or unexplained reactivity increase E. Failure of RMCS to entent that it has to be bypassed to borate the RCS [.3 es.] REF.BG-0-01, Emergency Boration t f
9 i 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33
~~~~Rd656E665CIE~66OTR6E~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- SNUPPS-1 - 85/04/15-R.R.FERRELL ANSWER 7.05 (2.25) A.1. Terminate turbine load and/or boron concentration changes
- 2. Transfer rod control in attempt to maintain Tavs/ Tref within +/- 2 3.If contro11ons bank will not move, adjust turbine load / boron concentrat-ion to restore / maintain Tave/ Tref C.753 B. full length rod that will not move in manual mode C.753 C.odjust turbine load ,
edjust baron concentration C.75] REF.SF-0-01, Failure of Control Rod Bank (s) to Move
. CNSWER 7.06 ( .75)
A.1.25 B.18.75 C.7 5 REF.10CFR20.101' l , AN3WER 7.07 (1.00) 1.The whole body limit shall not exceed 3 REM
- 2. Dose shall not exceed 5(N-18) REM where N= individuals age 3,NRC Form 4 has been completed. C.33 es.3 REF.10CFR20.101 e
s
-,, .. , . _ . , , . _ _ _ . ~ . -
. - , ._.,,__._...4_ _ _ _ . , , _ ._______c_____.___ __ . _ , ., . _ _ _ _ . _ _ . , _ , , _ , - _
- 7. ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34
~ ~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~R d6f6L66f6AL C6NTR6L ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 7 08 (1.50) .
A.less than 50 degrees F/ hour 75Cff0k B.between 45 and 55% C.50degreesF[:6. CI E.50 ea.3 REF.ES-0.2, Natural Circulation Cooldown
/
CNSWER 7.09 (2.00) A. Pressurizer vapor space temperature increases PRT temperature increases Close PORV and use aux spray to see if a pressure dewcrease is observed E.753
-B. Pressurizer pressure greater than/ equal to 320 psis Pressurizer level less than 70% ,. ,j i , .h' /
t'.t 'l? I y- / E.753
/ ce ., pg.pg// f 'r' ' -H-l r
ueal S,u Injection /, return
..u.c. 3 z . . ,- t, , , c. lined
.. c . ~' F ~ cup/ t > ;c. c .'u a --
C.RHR system u[ sins flow throV3h the RHR hl eat exchangers and bypass flow control valveiCCW is cooling medium E.53 REF.SNUPPS Plant Operationsep.5 thru 9 c ANSWER 7 10 (2.25) A.1. Rod bottom lights lit
- 2. Reactor trip and bypass breakers open
- 3. Neutron flux decreasing i, C.25 ea3 B.1. Turbine stop valves closed
- 2. Turbine Control valves and CIV's closed E.25 ea3 C.1. Flow control valves closed
- 2. Flow control bypass valves closed 3.S/G blowdown isolation valves closed 4.S/G sample isolation valves closed C.25 ea3 REF.E-0,Rx Trip and SI
- 7. . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~RIDf6L6EECdL C6NTR6L ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL
~
ANSWER 7.11 (1.25) 1.CCW supplied to RCP motor and thermal barrier in,o-qc/" 2.41 seal leak off greater than .2 spm a'{5 A 9 ' b '#*d 3.41 seal differential pressure greater than 200 psid. 4.RCPO lift oil pump running for at least 2 min.
- 5. Reactor makeup to contyainment isolation valve (BL-HIS-8047) open
- 6. Power to AC Service bus 01
\.' t."\ m s. g...; l +... k W ;
(PA si k 'or .' PA-02) E5/6 at .25 ea.3 REF.ES-0.2iNATURAL CIRCULATION C00LDOWN ANSWER 7.12 (1.00) " 2.MFPA/Btripalarmon[SIyy,g*ec.~:3:af6'W.'. 2.S/G A,B,C,D flow mismatch alarm , _ , 3.S/G A,B,C,D level deviation alarm As u.n' 4.MFP header pressure decreasin3
- 5. Decreasing S/G 1evels n #. v' n ,,, ,'
4 6.Feedwater flow decreasing W N,td*i\* 7.Lo,ad [4/7 at .25 ea3 n . :. .rej,ection.>to
~ > c. w 60% ,- .,. t. , ,
. r 1 'J , .
REF.AE-0-01, Loss of a Feedwater Pump ANSWER 7.13 (2.50) A.1. Rods stepping out rapidly in auto 2.Tavs decreasing , 3.P:t level decreasing 4.Pzt pres decreasing 5 1 or more PR neutron flu:: decrease
- 6. Rod position indication / Rod bottom LED E4/6 at .25 ea.]
B.If Control rods reach C-11, reduce turbine load to maintain Tavg= Tref E.53 C.30 min E.253; power to be reduced by 30% of rated thermal power E.253, cpplies only above 50% of rated thermal power E.253 E.753 REF.SF-0-5, Dropped Rod and Technical Specificaqtions p.3/4 2-12 P
F
- 7. . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36
~~~~ ~ ~~~~~~~~~~~~~~~~~~~~~~~~
RA656L6G5CdL C6NTR6L ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 7.14 (1.25) A.1.Pzr drops below 1849 psis
.2.pzt level drops below 10%
3.RCS subcoolins drops below 20 degrees F C.25 ea.3 B.FI will not automatically reinstate until the reactor trip breakers are reset E.503 REF.ES-03 ' , ANSWER 7.15 (2.25) A.When the senerator is synchronized to the grid it picks up about 5% power Jwhich causes no change in nuclear power since the steam dumps close most of the way. This will minimize the transient on the Rx. C.753 B.When load is picked up and the steam dumps close fully. [.753
-C.Most limiting of -
- 1. Load Dispatcher request
- 2. Turbine loading curve
- 3. Fuel conditioning requirements C.753 te u .c. nne , ox . r ' . . .
+ REF. Lesson Plan Pl, ant Transient Operation . ANSWER 7.16 (1.00) 1.nx has been at a constant power level and unit has been at a steady load for 5 minutes 2.S/G 1evels and steam pressure has been constant for 5 minutes 3.No rod motion for 5 minutes 4.Feedwater flow and temperature has been constant for 5 minutes 5.S/G Blowdown secured for duration E4/5 at .25 ea3 ( 4
c b
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 37
) ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 8.01 ( .50) C. REF.Rx Core Control for Large PWRep.8-29 ANSWER 8.02 (1.75) A.1. Unusual Event
- 2. Alert
- 3. Site Emergency
- 4. General Emergency E.125 ea3 B.ilert or higher E.253 C. Site or General Emersency E.253 D.1. Directing notifications to off-site agencies
- 2. Making protective action recommendations to off-site authorities
- 3. Requesting off-site assistance including federal, state, and local E.25 ea3 REF.EIP-ZZ-00101,Callaway' Plant Emergency Plan, p.5-7 ANSWER 8.03 (1.00)
A.TRUE [p.1 of 393 - ' B. FALSE Ep.6-293 OC T ^* C.TRUE Ep.5-83 D. FALSE Cp.6-21] REF.Callaway Plant Emergency Plan ANSWER 8.04 (1.50)
- 1. Read Unit Supervisor's los back throvsh last previous day on shift
- 2. Read night orders
- 3. Discuss plant status, planned evolutions or maintenance w/ Duty Supervisor
- 4. Check the Duty Supervisor has completed / signed off plant surveillance
- 5. Review forms of any releases of radioactive wastes
- 6. Review jumper and lifted lead los
- 7. Initial outstanding RWP's and terminate RWP's that are-at time limit
- 8. Check minimum manning requirements are met
- 9. Read Standin3 Order Boob through last date on shift
- 10. Walk MCB for review of plant status
[6/10 at .25 ea.3
+
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 38 ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 8.05 (1.00)
A. greater than 551 degrees F B.228 steps C. greater than or equal to 2205 psig/less than or equal to 595 degrees F D.less than/ equal to 92% 0.25 ea3 i REF. CAP-02, Routine Surveillance Requirements ANSWER 8.06 (1.00) ,' A.with only 1 inoperable, restore both to operable within the next 72 hours a or be in at least HOT STANDBY and borated to a SDM equivalent to at eas 1% delta K/K ( at a RCS temperature of 200 degrees F) within the next 6 hours E.253; restore both to operable within the next 7 days or be in COLD SHUTDOWN within the next 30 hours [.253 E.503 B.Both demonstrated operable by verifying,that on recire flowethe pump dev-olops a discharge pressure of greater than or equal to 2390 psis. [.503 REF. Technical Specifications,3/4 1-11 ANSWER 8.07 (1.00)
- 1. Isolation valve open
- 2. Volume between 6500 and 7000 gallons 3.9oron concentration between 1900 and 2100 ppm
- 4. Nitrogen cover pressure between 602 and 648 psis E3/4 at.33 ea.]
REF.TS,3/4 5-1 ; ANSWER 6.08 (1.50) A.Two physically independent circuits between the off-site transmission network and the on-site Class 1E distribution. [.53 B.Two seperate and independent diesel generators each with: [.553
'1.Seperate day tank with minimum volume of 390 gal of' fuel C.153 2.Seperate fuel oil storage system with minimum volume of 85,300 gale.153 3.Seperate fuel transfer pump E.153 REF.TS 3/4 8-1 1
- l l
,- c , - . . , - , - , . - . - , , , - - , - . . , , , , . ,....,,...,,_,,m .n, ---+ .,n-,-- - . n ,., . , , - -
=
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 39 ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL
~
ANSWER 8.09 (2 00) A. FALSE B.TRUE C.TRUE D.TRUE E. FALSE E.4 ea.] REF.TS,Section 6, Administrative Controls,p.1-5 s ANSWER 8.10 (2.00) A.TRUE [3/4 0-13 B.TRUE [3/4 0-13 C. FALSE [3/4 0-13 D. FALSE E3/4 4-103 E.TRUE [3/4 4-183 E.4 ea] REF. Tech Specs . ANSWER 8.11 (1.50)
- 1. Intent of the original procedure is not changed
- 2. Approved by 2 members of the plant management staffeat least 1 of whom is the Operating Supervisor, holding an SRO license on the affected unit 3 Change is documented, reviewed by the ORC, and approved by the Plant Superintendent within 14 days of implementation E.5 ea.]
REF.TS, p.6-17 * . d
. .,-. 7.,_, - _ _ , . . - . . _ _ , _ - . . _ . .., . , , - _ _ _ . . _ . _
g
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 40 ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL ANSWER 8 12 (2 50)
A.1. Combination of thermal powerepressurizer pressuretand the highest oper-ating loop coolant Tave shall not exceed the limits of the curve. BASIS prevents overheating fuel and clad damage by preventing DNB E1.03 2.RCS shall not exceed 2735 psis BASIS protects integrity of the RCS from overpressurization E1.03 REF.T/S B 2-3 i B.C REF.TS,2-1 ANSWER 8.13 ( .50) D R .T/S p.1-8, Table 1.1 Ai'SWER 8.14 (1.75) A.If AFD has been outside the +/- 5% target band for more than 1 hour E.253 reduce thermal power to less than 50% of rated thermal power within 30 ainutes E.253 and reduce the Power range neutron flux high trip setpoints to less than or equal to 55% of rated thermal powe6'E.253.'Il'~/ ' ' E.753 B.Nothing-return to band since only earn i for 2 penalty minutes E.53 C.AFD-Flux at top / flux at bottomiThus over core life flus shifts toward the top of the core due to fuel depletion. [.53 REF.T/S-3/4 2.1 ANSWER 8.15 (1.00) NO. The normal surveillence interval is 7 days,which can be extended by 25% to 8 and 3/4 days,provuded 3 consecutive intervals do not exceed 325% ( 22 cnd 3/4 days) of the normal surveillance interval. In this case,no single interval exceeds 8 daysebut the total of the 3 intervals is 23 dayseexceed-ing the allowable time limit. REF.T/S
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 41 f _B. ANSWERS -- SNUPPS-1 -85/04/15-R.R.FERRELL f - ANSWER 8.16 (2 50) A.24 hours B.14 days without having 2 consecutive days off C.2 cc i D.yes c 4 ,' ',. ' ' ' ,p,' E. Operating Supervisor . tD i4 E.5 ea] REF.T/S,Section 6.1-2 ANSWER 8.17 (2 00) A.1.1 spa 2 10 spm 3.0 spa 4 500 spd through any 1 SG/1 spm total E.375 ea.] B.The. potential release of activity to the atmosphere is below limits to protect the public [.50] in the event of a SG tube rupture E.253 E.503 REF.T/S 3.1.1.4 and bases' 4 e 9
, - - - , , _ . , - . . . , . . -}}