ML20107M895
| ML20107M895 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 08/30/1984 |
| From: | Fleishmann R PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Greenman E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20107M875 | List: |
| References | |
| NUDOCS 8411140327 | |
| Download: ML20107M895 (81) | |
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l August 30, 1984 Mr. Edward G. Greenman Chief, Projects Branch #1 U. S. Nuclear Regulatory Comnission, Region I King of Prussia, PA 19406
Dear Mr. Greenman:
Thank you for the opportunity to review the Reactor Operator and Senior Reactor Operator written examinations adninistered at the Peach Bottom Atomic Power Station on August 28, 1984. We believe that both examinations were straight-forward, emprehensive and operationally oriented,and were professionally and fairly adminis-tered.
The new review process was workable and constructive, resulting in resolving the majority of our concerns through open dialogue with the examiners. After review and discussion of both examina-tions, four cmments remain.hich require additional reference material for clarification of answers. Enclosed is a list of these connents by question nunber. For each connent, specific reference material supporting the clarification is provided.
l Veraf truly yours, Richard S. Fleischmann, II Superintendent Peach Bottan Atomic Power Station cc:
W. T. Ullrich R. W. Bulmer S. J. Mannix i
Enclosure i
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PHIIADELPHIA ELBCIRIC COMPANY CXMENTS ON NUCLEAR REGEAHRY CCMfISSICN REACIOR OPERAHR/ SENIOR REACER OPERAER LICENSE EXAMINATION PEAG BOP 10M AEMIC POWER STATICN August 28, 1984 3.04
'Ihis question asks the nochanisms of generating a reactor scram signal frm a Turbine Control Valve fast closure.
Cm ment: Answer key references EIS pressure frm ICT lesson plan. Lesson plan should state RETS (Relayed Emergency Trip Supply). PBAPS turbine uses Relayed EIS since drain lines are low capacity. Figure 1 to question 3.04 illustrates details of this arrangement. Answer key also discusses actuation of Fast Acting Solenoid in Control Pac. Fast Acting Solenoid is only actuated by Load Unbalance Relay (See Figure 2 to question 3.04) or last 10% of ' ICV test closure.
3.05b What are the effects of a condensate pump trip when Feed Flow is greater than 90%?
Coment: Reference lessca plan gives incorrect value to DiC load set runback triggered by the above conditions.
EHC load set runback only occurs (at 1%/sec) when total feed flow is greater than 95%. Figure 3 to question 3.05b shows the conceptual logic while Figure 4 to question 3.05b is an LER involving this circuit.
4.05c Question asks when actual Core Spray Injection to vessel would occur during a ILCA.
Coment: Answer key inplies injection would occur as soon as 450 psig injection valve pennissive is achieved.
Figure 5 to question 4.05c is provided to detail actual core spray purtp characteristics.
7.03b Question asks reason behind 50% speed limitation on running recirc punp when bringing on idle purrp.
Coment: Answer key discusses minimizing chance of APRM scram due to flow redistribution. Figure 6 to question 7.03b details consideration of vibration induced by flow reversal.
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P;ach Bottom Atomic Power Statirn Delta, Pennsylvania 17314 February 22, 1973 FROMi J. Winzenried 10:
BWR Trainees
SUBJECT:
LoadRejectScram(ControlValveFirstClosure/RETSOILLOPRESSURE) s The load reject condition (tripping of the 500KV generator output breakers 215 and 225) is detected by a " POWER. LOAD UNBALANCE" Relay which is in the turbine control package.
The Power. Load Unbalance Relay compares generator amps (all three phases) to turbine cross-around pressure (see P&ID M-303 area H.3 PT-2807). Cross-around~ pressure is proportional to turbine load.
When turbine load is 40% greater than generator amps the Power-Load Un-balance Relay trips and sends a trip signal to the control valve fast closure solenoid.
In the " Tripped" condition, the fast closure solenoid blocks the RETS oil supply and dumps the oil which holds the disk dump valve closed to drain.
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The pressure switch (PS) detects the decreasing oil precsure (setpoint 500 to 850 psig) before the disk dump valve opens and sends the scram signal.
Thus., this scram signal anticipates the pressure and resulting flux transient which will occur when the control valvec go shut in p.2 secondc.
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PHILADELPHIA ELECTRIC COMPANY J g f.
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PHILADELPHIA. PA.19101 March 7, 1984 Docket No. 50-278 i
Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
Licensee Event Report This LER deals with a reactor scram on APRM high flux caused by the failure of a relay in the turbine EMC runback logic.
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Reference:
Docket No. 50-278 g.
Report Number:
3-84-05 Event Date:
February 9, 1984 Report Date:
March 7, 1984 Facility:
?each Bottom Atomic Power Station RD #1, Box 209, Delta, PA 17314 This LER is submitted pursuant to the requirements of 10 CFR 50.73(a)(2)(IV).
l Very truly yours, l
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Superintendent Nuclear Generation Division ec:
Dr. Thomas E. Murley, Adn.Inistrator Region I, USNRC Mr. A. R. Blough t/( / A 7 / 7 /A I e-7
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lbach Bottom Atomic Power Station - Unit 3 e is ie la lo1217 8 1 o0 013 ist&8 443 Reactor Scram Due to APRM Bigh Flux avne, u,s isi un.v san asi asen, u,s evi l
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I A_b__s t.ra c t 3-84-O'i On February 9, 1984, at 6:0~1 p.m.,
a runback and subsequent relay failurg in the Main Turbine Electrohydraulic Control System (EHC) resulted in a Unit 3 reactor auto scram on high neutron flux.
With Unit 3 operating at about 92% power, the 3B reactor feed pump (RFP) turbine experienced high vibration.
Manually tripping the reactor feedpump turbine caused an auto.matic runback of the recirculation pumps and the turbine EHC system.
A faulty relay l
contact in the EHC system kept the closing circuit of the main.
turbine control valves energized which resulted in a reactor auto l scram on high neutron flux.
The f aulty EHC relay and a faulty. reactor feedpump turbine coupling which was causing the vibration were replaced.
A routine test will be instituted to functionally test this relay and the EHC runback logic each refueling outage.
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Description of the Event:
On February 9, 1984, Peach Bottom Atomic Power Station, Unit 3, was operating at about 921 power.
At approximately 6:00 p.m.,
the 3B reactor feed pump experienced high vibration.
The reactor operator, in an effort to decrease the vibration', reduced flow through the pump.
Vibration continued and the operator inanually tripped the pump.
The ensuing transient caused reactor water leve1 to decrease.
With feedwater flow greater than 95% and reactor water level below plus 17 inches, recirculation pump and EHC runback signals
.were correctly initiated.
The EHC runback is, designed to slowly pulse down the turbine load o
at approximately 14 per minute.
An Agastat relay (Model 2432 l
M PDC) failure caused the EHC system to runback at a rapid f
T-continuous rate.
The main turbine bypass valves opened to g
I compensate for the reduced turbine load.
When all bypass valves were fully open and the control valves continued to close, a high neutron flux auto scram occurred at 6:07 p.m.
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Consequences of the Event:
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The Reactor Protection System functioned properly and the reactor l'
successfully auto scrammed on high flux.
Since the Reactor Protection System functioned properly and no design. limits were exceeded, the* safety consequences of this event are considered minimal.
Cause of the Event:
The cause of the event was the failure of an Agastat relay (Model 2432 PDC) in the turbine EHC runback logic.
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The relay, if operating prope(
hould have decreased turbine load at approximately-1% per m.
until feedwater flow was less than 95%.
However, a contact within this relay remained closed after the EHC runback was initiated causTng a runback at a much fa=ts-rat.e and preventing the runback from stopping when it was no longer need,ed.
The cause of.the 3B reactor feedpump vibration problem was a faulty Zurn coupling.
Corrective Actions:
The Agastat relay was replaced.
A routine test has been written and will be instituted to functionally test this relay and the EHC runback logic e.ach refueling outage.
f The Zurn coupling on the 3B reactor feedpump turbine was replaced
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and the turbine was returned to satisfactory service.
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3.6.F & 4.6.F BASES M yymg h Mismatch Requiring the discharge valve of the lower speed loop to remain closed until the speed of faster pump is below 50% of its rated Epeed provides assurance when going from one to two pump cperation that excessive vibration of the jet pump risers will not occur.
Operation with one recirculation loop in service is permitted.
In such instances, the designated adjustments for APRM rod block and scram setpoints, RBM setpoint, HCPR fuel cladding integrity s2fety limit, MCPR operating limits, and MAPLHGR limits are required.
2 Figure 6 to question 7.03b t
-160-Amendt ent No.E.78/77
MASTERCOPY U.
S.
NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
_EEeCH_8DIIOU_213________
REACTOR TYPE:
_BUE=GEd_________________
DATE ADMINISTERED:_Boz0BZ28________________
EXAMINER:
_KUeBHE4_J.______________
APPLICANT:
INSIEUCIIONS_ID_eEELICeNIl Use separate Paper for the answers.
Write answeru on one side on1w.
Staple ouestion sheet on top of the answer sheets.
Points for each ouestion are indicated in Parentheses after the ouestion. The passing drade reouires at least 70% in each categorw and a final grade of at Isast 80%.
Examination Parers will be picked ur six (6) hours after the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY
__UALUE_ _IDIAL
___ SCORE ___
_UeLUE__ ______________CeIEGORY_____________
1.
PRINCIPLES OF NUCLEAR POWER
_25.00__ _25.00 PLANT OPERATION, T H ERMOliY N A M I C S,
HEAT TRANSFER AND FLUID FLOW
_25.00__ _25.00 2.
PLANT DESIGN INCLUDING SAFETY AND ENERGENCY SYSTEMS
._______ 3.
INSTRUMENTS AND CONTROLS
_25.00__ _25.00 4.
PROCEDURES - NORMAL, ABNORMAL,
_25.00__ _25.00 EMERGENCY AND RADIOLOGICAL CONTROL 100.00__ 100.00 TOTALS FINAL GRADE _________________%
All work done on this examination is mv own. I have neither diven nor ruewived aid.
APPLICANT'S SIGNATURE
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1.__ERINCIELES_DE_WUCLEeE_EDWER_EleWI_DEEEeIIONa PAGE 2
IWERHODXNAUICSa_WEeI_IEeNSEER_eWD_ELUID_ELOW QUESTION 1.01 (1.50)
Attached FIGURE 1 shows a basic closed loof fluid sustem with its head vs. flow Flot.
The two Pumps are identical, sindle speed, radial, Centrifugal Pumps.
InitiallUs assume Pump 1 is operatins to suPP1W flow to Component 1, as shoWn.
a.
WHAT is Poirit X on the Sustem Head vs. Flow Plot?
(0.5) b.
WHICH Pump curve, A or B, most securatelu shows BUTH PUMPS operating to suPP1W sUstem flow 7 (0.5) c.
WHICH WAY, to the LEFT or to the RIGHT, would the Sustem Curve shift if Component 2 was valved into the swstem, in addition to Component 17 (0.5)
GUESTION 1.02 (3.00)
A hot reactor is increased in Power DW WithdraWind Control rods.
The void fraction i rie r e a s e s 1.5% and fuel temperature increases 40 degrees F.
Assuming no change in moderator temperature, what was the reactivitw worth of the control rod?
Show all work and Otate anw assumptions wou make.
(3.0)
QUESTION 1.03 (3.00)
Classifw the heat exchander on the attached figure according to a.
operational mode (1.0) b.
tVPe of heat reJeetion (1.0) c.
flow Path (1.0) r
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s 1.__ERINCIELES_DE_WUCLEAE_EDWER ELeWI_DEEkeIIDWa PAGE 3
o IWEENODYWAHICSa_WEeI_IRANSEER_AWD_ELUID_ELOW QUESTION. 1.04 (3.25)
Concerning THERNAL LIMITS:
- a. Since MCPR is not'a direct 1w measurable parameter, WHAT cre THREE.
(3) measurable, core parameters needed bv the process computer to
. calculate MCPRT (1.0) b.
With regard to MAPRAT:
(2.25) 1.
WHAT'is the RELATIONSHIP between MAPRAT 1 MAPLHGR7 2.
The Process computer Prints out a MAPRAT of 1.05. Is this acceptable?
- 3. WHAT Phwsical conseouence could occur if the MAPRAT limit is exceeded?
QUESTION 1.05 (3.00)
Three (3) minutes'following a reactor scram from high powere indicated
. reactor Power is 75 on range 4 and decreasing.
- a. WHAT will INDICATED Power be one (1) minute later?
(1.5)
(Show calculations)
L b.' Explain WMY Power decrecsed at this rate.
(1.5)
DUESTION 1.06.
(3.00) a.
List three(3) factors that cause excess-reactivity of the core to decrease over cwele life.
(1.5) b.
List three(3) factors that will ecuse Shutdown-Margin (SDM) to (1.5) increase.
b-GUESTION 1 07 (1.50) 8 a.
During Power operations, Samarium is not normalls considered an operationel concern as in Xenon.
State two(2) reasons whw Xenon is an operational concern while Samarium is not.
(1.0)-
b.
True or False:
At lower power levels, Xenon euvilibrium is reached FASTER than at higher power levels.
(0.5)
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1.__EEINCIELES_DE_WUCLEeE_EDWEE_EleWI_DEEEeIIONA PAGE 4
IHEEHDDYNAMICSa WEAI_IEANSEER_eUD_ELU1D_ELOW QUESTION 1.08 (3.00)
With Unit 3. et rated conditions the EHC Pressure setPoint (on the controlling Pressure regulator) is lowered to its minimum value with the DECREASE Pushbutton.
Assuming N0 further operator action, answer the following using attached FIGURE:
a.
WHY does APRM PoWCr graduC11U decrease in AREA 17 (0.5) b.
WHAT is causing total steam flow to be >100% rated flow at POINT 2?
(0.5) c.
WHY did total feed flow increase to full scale at POINT 3?
(0.5) d.
WHAT caused total feed flow to go to zero at POINT 47 (0.5) e.
WHAT is indicated be the useillations in the wide range reactor Pressure trace (AREA 5)?
(0.5) f.
.WHY do the Peaks in the Pressure oscillations occurring in AREA 5 become farther apart with time?
(0,5)
GUESTION 1.09
(.75)
The reactor is operating at a steem dome temperature of 536 degrees F when reactor power is increased so that steam dome tempercture increases to 544 degrees F.
WHICH of the following statements is cost correct?
a.
Steam Pressure increased, steam enthelPu increased.
b.
Steam Pressure remained Constant, steam enthalPU decreased.
c.
Steam Pressure increased, steam enthalPu decreased.
d.
Steam Pressure remained Constant, steam enthalPW increased.
4
4 1.__ERIWCIELES_DE_ NUCLEAR _EDWER_ELAWI_DEEEAIIDWA PAGE 5
IHERBODYNAHICSa_WEeI_IRAWSEER_AWD_ELUID_ELOW i
QUESTION 1.10 (3.00)
Followinst a normal reduction in Power from 90% to 70% with recirculation flow, HOW will the followinsJ chens:e (increase, decrease, or remain the same) AND WHY:
a.
The Pressure difference betWeen the recetor and the turbine steam chest.
(1 0) b.
Condensate depression at the exit of the condenser.
(1.0) c.
Final Feedwater temperature.
(1.0)
J t
--..--..m.,._..
-+,c.- - -,.,,.
2[__ELeWI DESIGW_IWCLUDING_SAEEIY_eND_EHERGENCY_SYSIEHS PAGE 6
QUESTION 2.01 (3.00)
Indicate whether the following statements cre TRUE or FALSE. If falsee EXPLAIN WHY.
1.
Scram valves are normal 19 held closed bu spring Pressure and opened bw air Pressure from a single air line which is controlled bw the scram Pilot valve.
2.
Back-up scram solenoids energize upon a scram.
3.
Cooling water flow enter the exhaust, Port and will leak Past seals into the reactor vessel.
QUESTION 2.02 (3.50)
Concerning the Standbw Liuuid Control Swstem:
A.
Whw is it necessarw for the swstem to be eaPable_of injecting the contents of the SLC tank in a MAXIMUM time of 125 (1.0) minutes?
B.
What are Three (3) uses of the SLC injection secreer OTHER (1.5)
THAN Poison injection?
C.
List Three (3) of four indications that an operator has availible to determine that Standbw Liuuid Control is indeed injecting into the Reactor Vessel.
(1.0)
QUESTION 2.03 (3.00)
Regarding the Standbw Gas Treatment Sustem:
A. What'are Three (3) of the four conditions which will auto initiate the swstem? SetPoints are reuuired.
(1.5)
B. Consider Unit 2 1 3 sePeretelw. Indicate the Primerv and back-up ' FAN' (AeBe or C) and the Primarv and back-up
' FILTER-TRAIN' (A or B) for an auto initiation signal.
(1.5)
GUESTION 2.04 (1.50)
^
What automatic tries are associated with the Fuel Pool Cooling water Pumps? SetPoints are NOT REQUIRED.
(1.5)
2.__ELANI_DESIGU_ INCLUDING _SAEEIY_AND_ EMERGENCY _SYSIEMS PACE 7
QUESTION 2.05 (3.75)
A.
List the automatic start signals for the emergeneu diesel generators. Give setpoints where applicable.
(2.25)
B. Brief1w EXPLAIN the recronse of the ESW/ECW sustoms with resPeet to the diesel generator on un auto start.
(1.5)
QUESTION 2.06 (3.50)
A.
What is the largest heat loed on the RBCCW sustem?
(.75)
B. What is the most imrottant load on the RBCCW sustem?
(.75)
C. Uron loss of the feeder buses to the drwwell chilled water chillers and eumes, what euuiement is ISOLATED AND SUPPLIED in the RBCCW Swstem?
(2.0)
QUESTION 2.07 (3.25)
A.
What are the two sources of steam for the Reactor Feed Pump Turbines? Be specific.
(1.0)
B.
For the following statements, indicate whether TRUE or FALSE.
If falsep EXPLAIN whv.
1.
The
'C' Reactor Feed Pume discharge bueass valve's purpose is to Provide the Pump with minimum flow for coolind during low flow operations.
(.75) 2.
The Motor Speed Changer (MSC) of the feed Pumr turbine can r e siul a te turbine speed on1w between 2200 rpm and full speed.
(.75) 3.
A reactor feed Pump turbine has 5 low pressure and 1 high Pressure Control vBlve.
(.75)
GUESTION 2.08 (1.00)
Identifw FOUR (4) Possible discharge flow Paths of the HPCI rumr.
(1.0)
'2[ ELAWI DESIGN INCLUDING SOEEIY oWD EHEECENCY_SYSIEMS PAGE 8
GUESTION 2.09 (2.50)
With regard to the Main Steam Safetw Relief Velve (SRVs):
'A.
EXPLAIN HOW/WHY an SRV discharge Fire (teil riPU) Could be damaged due to its vacuum breaker sticking shut during repeated actuation (lifting) of the SRV.
(2.0)
B. How (INCREASE, DECREASEe REMAINS the SAME) would Drwwell Pressure be e>:pected ' to respond to un SRV discharde line vacuum breaker STICKING OPEN during actuation of the SRVT
(.5) i I
r 1
A
.-,-..--..,w
.-c.3,m-,
,.,-,,ms,,w.m.
,w---.--..--c...
...-,,_-em-,.%
,m,,
'3$__IWSIEUBENIS_eWD COWIROLS PACE 9
QUESTION 3.01 (2.00)
What are the FOUR (4) conditions which will cause a PCIS Grour IIA (Reactor Watec Cleanup Swstem) isolation? Setroints NOT reuuired.
(2 0)
GUESTION 3.02 (3.00)
Refer to the attached Recirculation Flow Control Figure for the following:
A.
The plant is operating at 26% Power and both recire rump M/A transfer stations are in MANUAL and set for 28% speed. The recire flow
'A' limit annunciator is clear. Recire Pume
'A' M/A transfer station is then placed in AUTO. Indicate HOW the speed os Recire Pump
'A' would change (increase, decrease or remain the same) and WHICH component (s) of the control swstem is(are) limiting.
(1.5)
B. Following a ' runback' of the recire sustem from 100% power due to the trip of one feed Puer, WHAT action must be taken bu the control room operator Prior to recetting the " runback'T WHYT (Assume RFP is restarted prior to reset.)
(1.5)
GUESTION 3.03 (2.50)
What are the signals reouired to initiate the Automatic Depressurization Swstem (ADS) (setroints reuuired) and for oach identifw WHY each signal is used?
(2.5)
GUESTION 3.04 (3.00)
Brief1w describe the sensind method of a Turbine Control Valve fast closure and HOW does it interface with the Reactor Protection (3.0)
Swstem (RPS) Logic?
4 PAGE 10 3.__INSIEUMEWIS_eWD CONIEDLS QUESTION 3.05 (4.00)
Brief1w state what will occur from each of the below changes in the feedwater flow conditions. (NOTE: Consider on1w other cameonents cnd swstems and NOT integrated Plant response.)
A. Total Feedwater and Total Steam flow are > 25%. Total Feed-(1.0) water flow then drops below 25%.
B. Total Feedwater flow is > 90% and a condensete Pume tries.
(1.0)
C. Recite Pump speed is 40% cnd Total Feedwater Flow drops to (1.0) less than 20%.
D. Total Feedwater flow is >95%. Feedwater Flow
'A' then drops to < 20% and Reactor Water Level drops to < 17'.
(1.0)
GUESTION 3.06 (2.50)
Brief1w DESCRIBE each of the FIVE different ranges of Reactor Vessel Level Indication in terms of the following s.
The NAME of the indicating range.
b.
Its SPAN.
c.
Its ZERO REFERENCE.
- d. Its CALIBRATION TEMPPERATURE (Hot or Cold).
(2.5)
GUESTION 3.07 (3.00)
Consider the Neutron Monitoring Swstem:
A. Whw is it necessarw to gamma compensate the Source and (1.0)
Intermediate Range Monitor signals?
B. At what Percent Power should the APRM flow bicsed scram (1.0) occur with 50% recire loop flow?
C. What are FOUR conditions that will result in an 'APRM INOPERATIVE
- while in the Startue Mode 7 (1.0) i
' 32__IWSIEUt!ENIS_eWD_CONIROLS PAGE 11 QUESTION 3.08 (2.00)
Brief1w EXPLAIN the function of the 2 Second Auxilers Timer in the Reactor Manual Control Swstem ( R M C S ).' Include in wour oxPlenation what occurs when this timer times out.
(2.0)
GUESTION 3.09 (3.00)
Unit 2 is operating with the followirig conditions:
Reactor Power
=100%
Total Core Flow
=100%
Reactor Pressure 51010 Psig Turbine Load Set
=100%
Turbine Speed Select
=1800 rpm Pressure Set a920 Psig Max Combined Flow Set
=100 (Euuivalent to 125% steam flow)
Load Limit Set
=100%
BwPass CaPacitw
=25%
Recirculatiori Flow Control
= Master Manual Using the attached EHC Logic and Pressure Control Diagrams EXPLAIN the swstem response to each of the following below.
(NOTE: Show arew csiculations wou maw make and take the erablem to a stable Point Condition.)
A.
The bias sidrial to the
'A' pressure regulator is inadvertent 1w increased to + 5 Psi.
(1.5)
B.
Grid freauenew increases sudderilu ' o e value consistent with 1804.5 rpm turbine speed.
(1.5)
+t b
-,i.__EROCEDURES_=_WORHeLa_4kWORH4La_ EMERGENCY _e D PAGE 12 W
RADIOLOGICAL COWIROL QUESTION 4.01 (2.00)
One of the steps in the TRIP Procedure T-100, ' SCRAM', has the cPerator Place the Mode Switch in the ' SHUTDOWN' Position. State TWO soals accomplished bu this action.
(2.0)
GUESTION 4.02 (2.00)
You are the Reactor Operator dur ing high Power operations and oudden1w notice that Reactor Pressure is at 1050 esid. What are wour immediate actions? (NOTE: Assume a scram HAS NOT occur.)
(2.0)
GUESTION 4.03
(.50)
Choose the correct statement below:
(.5)
In accordance with OH-100, 'Failute of a Jet Pume', if a Jet Pump failure is confirmed:
A.
Secure the associated recire pump and isolate that recire loop.
B.~ Immediate1w stop both recite Pumps and scram the reactore C. Commence a controlled Plant shutdown in accordance with GP-3.
D. Immediate1w scram the Plant and CooldoWn to < 212 F.
QUESTION 4.04 (3.00)
WHEN MUST the Reactor-Operator initiate the Standbw Liould Control (SLC) Swstem "WITHOUT' the Shift Supervisor'c Permission?
(3.0)
do__EBOCEDUEES_=_WDEMALa_eBNDEdeLa_ EMERGENCY _AND PAGE 13 EADIOLOGICAL_CONIEDL QUESTION 4.05 (2.00)
You are the Reactor Operator during a LOCA and wour attention is focused on the Core Sprew Swstem.
A.
How score should the Core Seraw Pumps start efter the receipt of all the initiation signals? (NOTE: Consider both with and without Normal Power uvailible.)
(1.0)
B.
What action should wou tche if the
'A' Core Seraw Pume did not auto start.
(.5)
C.
Core Spraw injection into the Reactor Vessel is blocked until the condition of low Reactor Pressure occurs. Pume discharge will be through the minimum flow recire valves until that time. At what Reactor PRESSURE will the Core Seraw Swstem inject into the Recetor Vessel?
(.5)
GUESTION 4.06 (3.00)
According to Procedure OT-114:
A.
List FOUR (4) observations an operator could mcke in order to verifw that a Safetw Relief Valve (SRV) is stuck oPen while at high Power.
(2.0)
B.
What immediate action (s) should the operator take for a stuck oPen SRV7 (1.0)
GUESTION 4.07 (2.00)
With the reactor shutdowre and cooled down to less than 212 F, secording to OP-12, ' Core Cooling Procedure':
w aH A.
If on1w one loop of Shutdown Cooling i s ON emi ND recire flow through the other loop, reactor 1cvel must be main-tained high enough to prevent WHAT from occurring?
(1.0)
B.
HOW would coolant temperature stratification be indicated bw reactor vessel metal surface temperatures?
(1.0)
A.__EEDCEDURES_=_WDEdeLa_eBWORMALA EHERDENCX_AND PAGE 14 EADIOLOGICAL_COWIEDL QUESTION 4.08 (2.00)
What are TWO (2) reasons for reuuiring bu Procedure that HPCI cnd RCIC NOT be operated at less than 2200 rem?
(2.0)
GUESTION 4.09 (3.00)
What are the FOUR (4) entrw conditions for the " Reactor Pressure Vessel (RPV) Control, ' TRIP Procedure, T-1017 (NOTE:
Seteoints are reouired where applicable.)
(3.0)
GUESTION 4.10 (2.50)
Regarding DN-113, ' Loss of RBCCW':
A. What THREE (3) automatic actions should be verified if the non-regen heat exchanger outlet temperature reaches 200 F?
(1.5)
B.
WHEN must the recire Pumps be triPeed following a loss of
(.5)
RBCCW7 C.
When shutting down the recirc Pumps, WHY are thew first runback to minimuni speed, then tripped 10 seconds apart?
(.5)
GUESTION 4.11 (3.00)
According to GP-2,
" Normal Plant StartuP*:
A.
During the approach to critical, WHICH rod (s) in a new Rod Group are likelv to exhibit high notch worth?
(1.0)
B. During heatue to rated temperature and Pressure after raising the EHC Pressure setPoint to 600 Psis, EXPLAIN WHY one turbine bweass valve is opened to 10 to 20% (with the bwPass valve Jack)?
(2.0)
FIGURE, for Question 5.01 XO v
Pug 1 cogonent cogonent 4
g y j
J L C
SYSTEM d
D j
3 l ~ System Curve D
N l
5
/
5 l
/
A%
~
E
/
[
X
+-B Pump 1 Curve
/
/
/
s Flow (Arbitrary Units) l SYSTEM HEAD VS. FLO11 PLOT l
l l
l
- - ~..
m L
- 2 k
FIGURE FOR QUESTION 1.03 Wat*f Steam
=
y M
O h
,1
=
l l
Pump 1
l I
r l
l l
l
..-.--e
..-n
.,_._.a.,
.--n,,
62__E(eNI_SYSIEdS_DESIQUt_CONIEQLt_eND_INSIEUUENICIIQU PAGE 18 ANSWERS -- BEP
-- 84/01/10
-- GRAVES, D.
I ANSWER 6.01 (1.50)
Reactor Pressure less than or eaual to 135 Psis Drwwell Pressure < 2 Psis Reactor level > 162 1/2 inches (1.5)
(0.5 each)
REFERENCE RHR Studw Guide Ps 9 DNG59 ANSWER 6.02 (3.00) a.
No effect(0.5).
The EHC swstem has a Permanent madnet senerDtor on the turinine which would continue to provide eower(0.5).
(1.0) b.
Reactor feed Pume controls lock uP(0.5) due to loss of RFP control signal (0.5).
(1.0) c.
Condenser vacuum would decrease (0.5) due to the air ejectors triPPins (0.23) and the loss of 2 CW intako eumps(0.25).
(1.0)
REFERENCE UPS Studw Guide Ps 6 DNG62 ANSWER 6.03 (3.00) a.
Input:
Ans of the LPRH inputs b.
Count:
The number of LPRH inputs Which are operable c.
Reference:
The reference APRM input d.
Block:
The trin level reference o.
Flow:
The flow input to the slope and bias circuit f.
Averase:- The RBM channel output (3.0)
(0.5 each)
REFERENCE DNG49 RBM Studw Guide Ps 19-20 1
9
,n
.- - -.., - _., _ _... ~
__,,.--f-.
y-
,_-,,y y.
y
, - _ +
FIGURE for Question 1.08 R
e n
i E
- I~
l!
a}
a
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is i
e i
l O
b I
y il
=
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w I
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z
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e a.
&W W
w e
e p
_=-
ww LJ 3
lE E'
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o f
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ss
- Es I
i'
=.
- l 1-l g
e
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a gs 5
z t
8 l'
8 II i
e i
r u
s
.s g
Time
- Time *
- Each time increment is one (1) minute
=
or--
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yo,
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,,.%.,v---wv,-_,_c
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STEAM TABLE PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)
Volume. f t /tb Enthalpy, 8tu/1b Eritropy, Stu/lb x F 3
water Evan Steam' Water Evap Steam Water Evap Steam
- ['
P ss.
[P h,
h sg syg s
Y hg q
g g
Yr Yrg e
32 0.08859 0.01602 3305 3305
-0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 35 0.09991 0.01602 2948 2948 3.00 1073.8 1076.8 0 0061 2.1706 2.1767 3;
40 0.12163 0.01602 2446 2446 8.03 1071.0 1079.0 0.0162 2.1432 2.1594 40 45 0.14744 0.01602 2037.7 2037.8 13.04 1068.1 1081.2 0.0262 2.1164 2.1426 45 50 0.17796 0.01602 1704.8 1704.8 18.05 1065.3 1083.4 0.0361 2.0901 2.1262 50 60 0.2561 0.01603 1207.6 1207.6 2S.06 1059.7 1087.7 0.0555 2.0391 2.0946 60 70 0.3629 0.01605 868.3 868.4 38.05 1054.0 1092.1 0.0745 1.9900 2.0645 70 80 0.5068 0.01607 633.3 633 3 48.04 1048.4 1096.4 0.0932 1.9426 2.0359 30 90 0.6981 0.01610 468.1 468.1 58.02 104.2.7 1100.8 0.1115 1.8970 2.0086 90 100 0.9492 0.01613 350.4 350.4 68.00 1037.1 1105.1 0.1295 1.8530 1.9825 100 110 1.2750 0.01617 265.4 265.4 77.98 1031.4 1109.3 0.1472 1.8105 1.9577 110 120 1.6927 0.01620 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 130 2.2230 0.01625 157.32 157.33 97.96 1019.8 1117.8 0.1817 1.7295 ' 1.9112 130 140 2.8892 0.01629 122.98 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 150 3.718 0.0163 A 97.05 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686 150 160 4.741 0.01640 77.27 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1.8487 160 170 5.993 0.01645 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170 180 7.511 0.01651 50.21 50.22 148.00 990.2 1138.2 0.2631 i.5480 1.8111 180 190 9.340 0.01657 40.94 40.96 158.04 984.1 1142.1 0.2787 1.5148 1.7934 190 200 11.526 0.01664 33.62 33.64 168.09 977.9 1146.0 0.2940 1.4824 1.7764 200 210 14.123 0.01671 27.80 27.82 178.15 971.6 1149.7 03091 1.4509 1.7600 210 212 14.696 0.01672 26.78 26.80 180.17 9703 1150.5 0.3121 1.4447 1.7568 212 220 17.186 0.01678 23.13 23.15 188.23 965.2 1153.4 0.3241 1.4201 1.7442 220 230 20.779 0.01685 19.364 19.381 198.33 958.7 1157.1 0.3388 1.3902 1.7290 230 240 24.968 0.01693 16.304 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 250 29.825 0.01701 13.802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 260 35.427 0.01709 11.745 11.762 228.76 938.6 1167.4 03819 13043 1.6862 260 270 41.856 0.01718 10.042 10.060 238.95 931.7 1170.6 0.3960 1.2769 1.6729 270 280 49.200 0.01726 8.627 8.644 249.17 924.6 1173.8 0.4098 1.2501 1.6599 280 290 57.550 0.01736 7.443 7.460 259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 300 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 310 77.67 0.01755 5.609 5.626 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 t
320 39.64 0.01766 4.896 4.914 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 340 117.99 0.01787 3.770 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 360 153.01 0 01811 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 380 195.73 0.01836 2.317 2335 353.6 844.5 1198.0 0.5416 1.0057 1.5473 380 400 247.26 0.01864 1.8444 1.8630 375.1 825.9 1201.0 0.5667 09607 1.5274 400 420 308.78 0.01894 1.4808 1.4997 396.9 806.2 1203.1 0.5915
'9165 1.5080 420 440 381.54 0.01926 1.1976 1.2169 419.0 785.4 1204.4 0.6161 J 8729 1.4890 440 460 466.9 0.0196 0.9746 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 480 566 2 0.0200 0.7972 0.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4518 480 500 680.9 0.0204 0.6545 0.6749 487.9 714.3 1202.2 0.6890 0.7443 1.4333 500 520 812.5 0.0209 0.5386 0.5596 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 540 962.8 0.0215 0.4437 0.4651 536.8 657.5 1194.3 0.7378 0.6577 13954 540 560 1133.4 0.0221 0.3651 0.3871 562.4 625.3 1187.7 0.7625 0.6132 1.3757 560 580 1326.2 0.0228 0.2994 0.3222 589.1 589.9 1179.0 0.7876 0.5673 1.3550 580 600 1543.2 0.0236 0.2438 0.2675 617.1 550.6 1167.7 0.8134 0.5196 1.3330 600 620 1786.9 0.0247 0.1962 0.2208 646.9 506.3 1153.2 0.8403 0.4689 13092 620 640 2059.9 0.0260 0.1543 0.1802 679.1 454.6 1133.7 0.8686 0.4134 1.2821 640 660 2365.7 0.0277 0.1166 0.1443 714.9 392.1 1107.0 0 8995 0.3502 1.2498 660 680 2705.6 0 0304 0.0808 0.1112 758.5 310.1 1068.5 09365 0.2720 1.2086 680 700 3094.3 0.0366 0.0386 0.0752 822.4 172.7 995.2 0.9901 0.1490 1.1390 700 705.5 3208.2 0.0508 0
0.0508 906.0 0
906.0 1.0612 0
1.0612 705.5
a-
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l I
FIGUREFOR]UESTION3.02 p ESC T'"I' Wester h From Tas Ges.
MlA 4 Control y
f h To j Froes y Fnes. Reg.
M1A
& Manual
( Always in M an.)
r-Closed Meo Gen.
/ held Bkr. Closed J
Qosed Men Total Feed Blow > 20%
~~
gm, 30%
/ Sig. Gen.
Speed J
Closed When Discharge Limiter
~~
Valve is Full Open Opens When Field Skr. Clones Clooed
==
==
Cloemd m me.
Rz. Wta.
". ~_2 B
> Individua! Feed Fussp
~~
O Flows > 20%
L,t.
> 17"
__- _2C N
~ ~ 2A 75%
Closed closed When Speed ma Total
~ ~_23
> Individual Cond.
Limiter Fumps Dnanke Feed Flow
< 90%
-j
_2C T
l Closed when Bypasa Function is
_[ Manually Reset Closed When Gen. Field skr. Closed If lf If
- e Opens When Field
~~
Speed Bkr. Closes ControBer If If Podtioner A
Gen.
I A
U
..o.p y
=.
Loek ents Scoop Tube i
1 w
O O
O:
StAltI trP it A l(
4 i
s
.e :.s.
1
[
s't i
~_ 1 TUltHINE Tit lP 3.
gg g j M8 IV REGULAIION 6 og an l
lilltHINt _ _ - a gyn[G -
OPEN WitEN GEN OU TI'llT
% FLOW g
gy g4gg i
SPt i t)
SPE E D LOAD REJECT HHE AKEll IS CLOSED i
,,,,g y
,;g y E HitOft 0
!M 6-- swi s ces (OPilONAL)
CV REGUL ATION'N e s
I
/'
HitEM01 E INC/DEC SIGN AL 5T AH T IJP 8
.L ltAIE MASTER FLOW CON 1 ftOL LL R
/
_9g i
10 RE ClitC F L OW '
I-y ~
CONillOL y
--II--- StuNH ACK ON LO All REJECT
^
"b--- HilNit ACK SYNC SPEED NOT SEI ECTED w
e o,
7
- g ttjNUACK ON LOSS OF SI AI064 COOLING e
g y
g l TY lOPilON AL) w ST E AM TititOTTLE I
P ROM SPE ED PitESSURE SELECTOR
-v No asvor i
/
%, t' VALVE
,, 8 DEMAND I -
f I
F
~gg 5 PRESS il W A H MIN (,
TURHINE THIPPED CHtS8 1
[
Fruos v
v
^ 8 I
l PitESSullE SE 1
_ + 10 pse PHESS.
mmen Ni g.
j,.g, p g 3
,J g
E ott F
1
- 1. rIf,,'
,[STOP VALVES
]
U BIAS
, REG.
> t 0" S I E AM
+
r-o I CLOSED I
illitO T T L E
_cr' UYPASS
.I,J,,
f~y
,,v,,', c v,,vatyg e"
PH E SSUHE l
f l[ p
- r DFMAND 4
i LOW VAcutiM SM Al l ggypagg 1
j Cl OSE Jacg Hl AS 1
FIGURE 7,3-3 TURBINE CONTROL LOGIC (EllC) j I
G l
l
r'
?
N O
S 0679 m
O W C 2
EC WO g Jw
_Cp esc ma<
C: Bids e>:haust 16 mr/hr
- refuel floor e>:haust 16 ni r/ h t-(.5 ecch)(1.5)
B.
Unit 2: PRIMARY: fan A, trairi A B/U: fan B, trein A
(.75)
Unit 3: PRIMARY: fan C, trairi B B/U: Fan B, train B
(.75)
C)u '.,4.'4; M ; og bo-l k i sd d iursd op4 so bod k
-f( 4 s e A s w A REFERENCE LOT 0210, REV 0, Ps 3,5,7
2.__ELAWI_DESIGW_ INCLUDING _SAEEIY_eWD_ EMERGENCY _SYSIEMS PAGE 20 ANSWERS -- PEACH BOTTOM 213
-84/08/28-KVAMME, J.
ANSWER 2.04 (1.50)
(2 Psig)(.75) 1.
low suction Pressure 2.
Iow skimmer surge tank level - (213')(.75)
(1.5)
REFERENCE LOT-0750-REVO, Ps 6~
ANSWER 2.05 (3.75)
A.
1.
Triple low reactor water level
-130' 2.
High D/W Pressure 2 Psis
(.75 each)
(2.25) 3.
Loss of off-site Power 955 B.
On an auto start of the D/G, at BEF rpm the ESW discharge valve opens.(.5) ESW and ECW Pumps start after 22 seconds.(.5) 23 seconds later if the discharge Pressure is availibic from ESW, the ECW Pumps stoP.(.5)
(1.5)
REFERENCE LOT-0670, REV 0, Ps 3,19 ANSWER 2.06 (3.50) 1
(.75)
A.
NRHX B. Recirculation Pump seal and Anotor oil cooler
(.75)
C. RBCCW flow isolated to:
-Inst. N2 coolers (.33)
-NRHX(.33)
-RWCU Pump seal Coolers (.33)
(1.0)'
RBCCW flow sueelied to:
-recire PuRsP Rio to r coolers (.33)
-D/W air coolers (.33)
-D/W eauiPhent drain sump Coolers (.33)
(1.0)
REFERENCE LOT-0460, REV 0, Ps 7,5
4
-2 __EleWI_ DESIGN _ INCLUDING _SAEEIY_eWD_EHEEGENCY_SYSIEBS PAGE 21 ANSWERS -- PEACH BOTTON 213
-84/08/28-KVAMME, J.
ANSWER 2.07 (3.25)
A.
HP-
'A' MSL
(.5)
LP-cross around steasi(.5)
(1.0)
B.
1.
False (.25) This valve Provides for niore accurate vessel level control during low Power ops.(.5)
(.75) 2.
False (.25) MSC cars regulate between zero and full speed.(.5)
(.75) 3.
True(.75)
(.75)
REFERENCE LOT-0540, Rev 0, pg 3,6,8,11 ANSWER 2.08 (1.00) 1.
Normal discharge to "A'
feedwater line(.25) 2.
Mirsimum flow discharge to suppression ecol (.25) 3.
Test discharge to CST (.25) 4.
Test discharge bweass to suppression Pool (.25)
(1.0)
REFERENCE LOT-0340, REV 0, eg 41TP2 ANSWER 2.09 (2.50)
A.
Following the SRV's first actuatiors, the stecni in its discharge line(.5) would coridense causing a vacuum in the li rse (. 5 ) This would result in suppressiori ecol water being drawn up into the line(.5) which would cause overpressurizatiori of the line o re the next actuation (.5) (Also acceptable) Water hemmer end excessive Jet forces on the diffuser.
(2.0)
B.
Increases
(.5)
REFERENCE
~ LOT-0330, REV 0, es 4
3.__IWSIEUBENIS_eWD_CDWIROLS PAGE 22 ANSWERS -- PEACH BOTTOM 213
-84/08/28-KVAMME, J.
ANSWER 3.01 (2.00) 1.
Low water level 2.
High discharge temperature on non-regen HTX 3.
SLC initiation 4.
RWCU suction line high flow
(.5 each)
(2.0)
REFERENCE LOT-0110, REV 0, PM d ANSWER 3.02 (3.00)
A.
Increase (.5) Master l i sii t e r, low speed l i m i t t'1. 0 )
(1.5)
B.
The setpoint must be manuallw runback on each rumr (If M/A transferred to MANUAL) Prior to resetting the runbeck(.75)
Otherwise the recire e usie s will rame up to the previous setting causing a Possible sc ra Al(. 75 )
(1.5)
REFERENCE LOT-0040, Rev 0 ANSWER 3.03 (2.50) 1.
-130' vessel water level (.25) Indicates severe core degradstion(.25) 2.
+2 esig D/W Press (.25) Positive indicatiore of a rurture(.25) 3.
+6' vessel level Permissive (.25) Verifies triple low level from another instrument (.25) 4.
TD of 105 seconds timed out(.25) Allows HPCI time to recover (.25) 5.
One RHR G 50 Psis or two Core Srraw G 185 Psig(.25) Ensures low Pressure swstem are availible Prior to blowdown (.25)
(2.5)
REFERENCE LOT-0330, REV 0, es 5
34__INSIEUBENIS_AND_CONIEDLS PACE 23 ANSWERS -- PEACH BOTTOM 213
-84/08/28-KVAHME, J.
ANSWER 3.04 (3.00)
Turbine Control Valve fg#t glosure is determined bu Pressure ewitches which monitorAL[eikenew Trie SuPP1w 6ET S ) Pressure /lh te th; I;;t ^; t i ;5-C n' ~ "' d u
!!:e C en t ;,1 P.m J the T C '.'
'l 5_t There ase f o u t-Pressure switches, one for ecch RPS subchannel, A1, A2, B1, B2 (1.5)
As7 acTw6-socoso es cuc y /7c7047m 6 Y (3.0)
LOAD UAlssti A vu? RELA Y.(AroY R%vtRreo.)
REFERENCE LOT-0300, REV 0, PM 11 ANSWER 3.05 (4.00)
A.
R og u.t s ged o r
} gig AC4t#T 4C7gg@AMSWEX/fr en flow mismvrict*
s i n s.e react vg.t -d (1.0) g io resgo Nw-o B.
A recire eump runback to 60% is initiated and an EHC runback m 6 / )G[ecom uvres -/o-/d pg t/ow
- c. q gog (1.0) initiated.
C C.
A recirC PuhP runback to 30% is initiated.
(1.0)
D.
Recire Pump runback to 60% and EHC runback initiated until feedwater flow #4.
(1.0) 9 5~Yc, REFERENCE LOT-0550, Rev 0, Pg 11 l
l L.
.+:.
31__INSIRUMENIS_eWD_CONIEDLS PAGE 24 CNSWERS -- PEACH BOTTOM 213
-84/08/28-KVAHME, J.
ANSWER 3.06 (2.50) 1.
Feedwater Control Range
(.1)
O' to 60'
(.2) reference to instrument zero (538*)
(.1) calibrated hot
(.1)
(.5) 2.
Active Core Range
-325' to O' reference to instrument zero calibrated hot
(.5) 3.
Refuel Range:
-21' to +379' reference to instrument zero calibrated cold
(.5) 4.
Shutdown Range:
-178' to -78' reference to instrument zero calibrated cold
(.5) 5.
Yarwas Range:
-165' to +50' reference to instrument zero calibrated hot
(.5)
REFERENCE LOT-0050, REV 0, es 5-7 ANSWER 3.07 (3.00)
A.
At low Power levels, the signal Produced bw deceu or background samma overshadows the signal Produced be neutrons end fission samma.
(1.0)
B.
0.66W + 54 where W=% recire loop flow 8?%
(1,0) 33 + 54 0.66(50) + 54
=
=
C.
1.
High voltage low (.25) 2.
Hodule unplugged (.25)
- 3. Selector switch not in operate (.25) 4.<14 operable LPRH ineuts(.25)
(1.0)
REFERENCE LOT-0270, REV 0, 99 5 1 LOT-0240
3o__INSIRUMENIS_AND_CONIROLS PAGE 25 ANSWERS -- PEACH BOTTOM 2S3
-84/08/28-KVAMME, J.
ANSWER 3.08 (2.00)
The withdrawal Portion of a normal notch out seuuence takes about 1.5 seconds. The 2 second timer morsitors this interval (.5).
If the withdrawal signal is sent to directional control vlaves for > 2 seconds, the timer times out. It will generate a select block which deselects the rod (1.0) This Prevents a faultu master timer from causing an uncontrolled withdrawal signal (.5)
(2.0)
REFERENCE LOT-0080, REV 0, es 5,13 ANSWER 3.09 (3.00)
A.
'A' Pressure Regulator:
Steam throttle Pressure =950Psis Pressure set plus new bias =920+5=925 therefore 950-925=25aa signal out of
'A' pressure regulator (.5)
- B' Pressure Regulator:
Throttle Pressure =950 Pressure set plus bicci 920+10=930 therefore 950-930=20 ma signal out of
'B' regulator (.5)
- A' regulator signal Passes 25x3.3=82.5% signal. Control valves begin to close to 82.5% but as press increases reopen to 100%
with new throttle pressure. 30+925=955Psig and Rx Press =1020Psig(.5)
(1.5)
B.
In the speed circuits, both summers Will see a -4.5 rem speed error since the generator is in svne. and will follow the grid.
The LVG will pass one of these to the CV E IV regulation blocks.
The IV res. block will see a demand decrease of 12.5% but the oPen bias Will hold them open. The CV rec. block will develope a
-5% demand which sums with the load select to send a 95% signal to the Pressure load gate. CV's Will close to 95%. Since Rx demand has not changed, the BPV's will open to 5%. This condition will hold as long as the freuuenew holds at 1804.5 rpm.
(1.5)
REFERENCE LOT-0590 1 EHC Logic transients
"s e __EROCEDUEES_=_WDEMala_4BNDEMALa_EBERGENCY_AND PAGE 26 EADIOLOGICAL_CONIROL
. ANSWERS -- PEACH BOTT0H 213
-84/08/28-KVAMME, J.
CNSWER 4.01 (2.00) 1.
It seals in the scram signal bw irise r ting an additional scram signal f o r-ten seconds.
(1.0; 2.
It changes the plant mode such that if the MSL Pressure decreases to GP I isolation setpoint, MSIV's will NOT shut. This is desirable to keep the feedFume, SJAE, and off-gas therebw maintaining the condenser as a heat sink.
(1.0)
REFERENCE T-100, REV 0, PM 1,2 ANSWER 4.02 (2.00)
- 1. Run recite flow back to reduce Rx Pressure < 1040 Psic.
(1.0) 2.
Contiol RX Pressure <1040 esis with the buPass velvec using the Jack.
(1.0)
REFERENCE OT-102, REV 0, PM 1 ANSWER 4.03
(.50)
(.5)
'C'
,e REFERENCE ON-100, REV 0, PM 1 ANSWER 4.04 (3.00)
When the control rod swstem is incapable of shutting down the reactor erdi
(.75) 1 increasing as indicated bu NI's AND stm flow.
(.75)
IE. R x P oW e r i s 2.
If five or more adjacent CR's or thirts or-more CR'S can't be inserted past Position 06 AND level can't be meintained.
(.75) e1 3.
If five or more adjacent CR's or thirtu or more CR'c can't be insertd Past Position 06 AND suPPr'ession Poul temp reaches 110 F.
(.75)
.f, 4
A.__ERDCEDURES_=_NDEdeLz_eBNDEMAL4_EHERGENCY_eUD PACE 27 RADIOLOGICAL _CONIROL ANSWERS -- PEACH BOTTOM 213
-84/08/28-KVAMME, J.
REFERENCE S.B.6.B.,
REV 7, es 1 ANEVER 4.05 (2.00)
A.
1.
Normal Power availible-10 seconds
(.5) 2.
After diesel power availible - 5 seconds (.5)
(1.0)
B.
Manuallw start the eume.
(.5) f d icc*/o d C. < 450 psis W d ion
(.5) r A p(N V e*I O
)
f5e REFERENCE S.3.4.B.,
REV 5, Ps 1,2 ANSWER 4.06 (3.00)
,t A.
1.
Generator load reduction or bweass velve closure.
2.
' Blow Down Relief Valve HI Temp
- or 'Scretw Relief Valve Open' alarm.
3.
Relief Valve position lights.
4.
Inc reasinsi Torus temp.
Also Ac.ccoT S/2,t/ Ace s7;c
(.5 each) (2.0)
W)oWIT6 Le y y (4-m.
in setvice.
.5)
B.
1.
Place both loops of Torus cooling 2.
If Torus temp reaches 95 F enter T-102.(.5)
(1.0)
REFERENCE OT-114,ps 1,2 a
ANSWER 4 07 (2.00)
A.
All core flow being Jiverted through the idle Jet pump (burtssins) t h e c o r e ). W s cc . ~ AccE77 d *5c4SSIo d 07 AMWlX 00 5d fd fic
- 6d (1.0)
~
B.
Metal surface temperatures below water-level will be less than indicated coolant temperatuse.
(1.0)
REFERENCE GP-12, REV 5, Ps 9
'~
d.__ERDCEDURES_=_WDEbeLa_eBNDEMALa_EHEEGENCY_eHD PAGE 28 EADIOLOGICAL_CONIROL ANSWERS -- PEACH BOTTOM 213
-84/08/28-KVAMME, J.
CNSWER 4.08 (2.00) 1.
Ensures sufficient shaft driven lube oil Pump Precsure and flow.
(1.0) 2.
Ensures high enough steaming rate to avoid exhaust line check valve ' chatter' and Possib1w subseuuent damage.
(1.0)
REFERENCE LOT-1560, REV.
O, pg 14 ANSWER 4.09 (3.00)
- 1. RPV level below -48' or unknown.
2.
Drwwell Pressure is above 2 Psig.
3.
A Group I isolation occurs.
- 4. Scram condition with Power above 3% or unknown.
(.75 each)
(3.0)
REFERENCE LOT-1560, REV 0, Ps 7, 1 T-101 ANSWER 4.10 (2.50)
A.
1.
Group IIS isolation (OR MD-12-15, MD-12-18, and HO-12-68 close).
2.
RWCU Pumps trip.
- 3. RWCU domin hold PuniPs start.
(3 at
.5 each)
(1.5)
(.5)
B.
Within 10 minutes.
C.
To miniaize the reactor transient.
(.5)
REFERENCE ON-113e REV 09 PM 1 1 ON-113 BASESP Ps 1
A __EEDCEDURES_=_WOEHAL4_eBNDEMALa_EHEEGEWCX eWD PAGE 29 RADIOLOGICAL CONIROL f.NSWERS -- PEACH BOTTOM 213
-84/08/28-KVAMME, J.
ANSWER 4.11 (3.00)
A.
The first rod (s)
(1.0)
B.
To establish a constant flow iri and out of the reactor vessel (OR a constant feed flow desiand)(.66) to miniviize temperature fluxuations on the feedwater nozzles (.67)e thus minimizing the Probabilitw of thermal stess crackins(.67).
(2.0)
REFERENCE GP-2e REV 35, es 12,16
p
,t>
e TEST CROSS REFERENCE PAGE 1
00ESTION
-VALUE REFERENCE 01.01 1.50 JCK0000280 01.02 3.00 JCK0000281 01.03 3.00 JCK0000282 01.04 3.25 JCK0000284 01.05 3.00 JCK0000285 01.06 3.00 JCK0000286 01.07 1 50 JCK0000287 01.08 3.00 JCK0000288 01.09
.75 JCK0000289 01.10 3.00 JCK0000290 25.00 02.01 3.00 JCK0000291 02.02 3.50 JCK0000292 02.03 3.00 JCK0000293 02.04 1.50 JCK0000294 02.05 3.75 JCK0000295 02.06 3.50 JCK0000296 02.07 3.25 JCK0000297 02.08 1.00 JCK0000298 02.09 2.50 JCK0000299 25.00 03.01 2.00 JCK0000300 03.02 3.00 JCK0000301 03.03 2.50 JCK0000302 03.04 3.00 JCK0000303 03 05 4.00 JCK0000304 03.06 2.50 JCK0000305 03.07 3.00 JCK0000306 03.08 2.00 JCK0000307 03.09 3.00 JCK0000308 25.00 04.01 2.00 JCK0000045 L
04.02 2.00 JCK0000074 04.03
.50 JCK0000075 (U4. 04
.5 00 JCK00aDOTD 04.0b 2.00 JCK0000078 04.06 3.00 JCK0000309 04 07 2.00 JCK0000310 04.08 2.00 JCK0000311 04.09 3 00 JCK0000312 04 10 2.50 JCK0000313 04.11 3.00 JCK0000314 25.00 100.00
l MASTERCCPY
~
~
U.
S.
NUCLEAR REGULATORY COHHISSION s
SENIOR REACTOR OPERA 10R LICENSE EXAHINATION FACILITY:
_EEeCH_BUIIOU_213________
REACTOR TYPE:
_BUE=GEd_________________
DATE ADMINISTERED _84/08/28________________
EXAMINER:
_HILLa_D.________________
APPLICANT:
INSIEUCIIONS_ID_AEELICANII Uce separate paper for the answers.
Write answers on one side on1w.
Staple ouestion sheet on tur of the answer sheets.
Pointu for each cuestion are indicated in parentheses after the ouet, tion. The Passing crede reuuires at least 70% in each eategorw and a final grade of at losst 80%.
Examination eurers will be ricked up sin (6) hours after the exaniination starts.
% OF CATEGORY
% OF APPLICAN1'S CATEGORY
__UALUE_ _IDIAL
___ SCORE ___
_UALUE__ ______________CGIEGORY_____________
.25.00__ _25.00
________ 5.
THEOR) DF HUCLEAR PUWER PLANT OPERATION, FLUIDS, AND THERH0 DYNAMICS
.25.00__.25.00
________ 6.
PLANT SYSTEMS DESIGH, CONTROL, AND INSTRUMENTATION
_25.00__ _25.00
________ 7.
PROCEDURES - NORMALS ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
.25.00__.25.00
________ 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100400__ 100200
________ TOTALS FINAL GRADE _________________%
C11 work done on this examiination is me own. I have neither given not received aid.
APPLICANT'S SIGNATURE I
'5,o__IDEDEX DE_WUCLEGE_EDWER EL4WI_DEEEeIIDWo_ELUIDSA_0ND PAGE 2
.IWEEMDDYW4MICS s
QUESTION E.01 (1.50)
Regarding the core thermal limits!
4 a.
The Process computer output, CHFLPD, is used to monitor WHICH core thermal limit?
(.75) b.
WHICH core thermal limit ensures evek eiudding temperature will not exceed 2200 degrees F following a LOCAT
(.75)
GUESTION 5.02 (3.00)
Brief1w EXPLAIN WHAT happens to tiiv magnitude of tiie moderator temperature coefficient of reactivitw (MORE OR LESS NEGATIVE) and WHY considering the following changest a.
An increase its moderator temperature.
(1.5) b.
A decrease in control rod densitw.
(1.5)
GUESTION 5.03 (3.00)
Concerning heat transfer in the reactort a.
Brief1w EXPLAIN WHY nucleate boiling imrroves the heat transfer characteristics of the core over simelv forced convection?
(1.8) b.
Considering the heat transfer mechanism followind fuel drwout after c large break LOCA from a hidh reactor Power, WHY are the centesi fuel rods in a fuel bundle more likelv to exceed the 2200 degree F limite for Peuk cled temperature than the edge or corner rods which have higher local Peaking factors?
(1.2)
'5 __IWEDEY_DE_CUCLE4E_EDWER_ELCWI_DEEEAIIDWr_ELUIDSa_AWD PAGE 3
iIWEEMODYWeHICS QUESTION 5.04 (3.00)
Following an auto initiution of RCIC at a reactor PP&ssurU of 800 psis, reactor Pressure decreasec to 400 esis.
HOW ARE THE FOLLOWINO PARAMETERS AFFECTED (INCREASES, DECREASES, REMAINS CONSTANT) bu the change in reactor Pressure?
BRIEFLY EXPLAIN YOUR CHOICE.
ASSUME the RCIC sustem is opera-ting as designed.
a.
RCIC flow to the recetor.
(1.0) b.
RCIC Pure discharge head (cssuming NPSH remains constant).
(1.0) c.
RCIC turbine RPM (1.0) 00ESTION 5.05 (3.50)
Regarding thw xenon transient following a significant DECREASE in reactor power from high power operation:
a.
Brief1w, EXPLAIN WHY the xenon concentration will reak followins the manuever.
(1.0) b.
HOW will PWriPheral control rod Worth be affveted (INCREASE, DECREASE, REMAIN THE SAME) during the xenon pock? BRIEFLY EXPLAIN wour answer.
(1.5) c.
If the dvervase in recctor power wes from 100% to 50%,
would the new (50% Power) vuuilibrium xenon reactivitw be HORE THAN, LESS THAN or EQUAL TO one half the 100%
eauilibrium valuw?
BRIEFLY EXPLAIN YOUR ANSWER.
(1.0)
- 5._ IWEORY_DE_WUCLEAR_EDWER_ELAWI_DEEROIIDWa_ELUIDSa_eWD PAGE 4
.IWERMODXWeb1CS QUESTION 5.06 (3.00)
The reactor is operating at 70% Power with all sustoms functioning normallw when Recire Flow Controller
'A' fails HIGH. Using attached Figure 2, IDENTIFY the CAUSE of the recorder indiection chances et EACH of the NUMBERED POINTS described below.
NOTES: o Time intervals on the graph are in 1 minute increments o The transient begins 1 minuter 15 seconds from the beginning of each graph.
(1) The decrease in reactor water level.
(0.5)
(2) The increase in reactor power.
(0.5)
(3) The decrease in core flow.
(0.5)
(4) The increase in reactor Pressure.
(0.5)
(5) The iacrease in total feedwater flow.
(0.5)
(6) The slight increase in total steam flow.
(0.5)
QUESTION 5.07 (3.00)
Ottached Figure 3 shows selected Plant Parameter resPolises for a TURBINE TRIP transient initiated from rated conditions with NO OPERATOR ACTION.
NOTES:
(1) Time intervals on grarbs are 1 minute each.
(2) Use of graphs not direct 1w referred to in auestion MAY be reuuired to correct 1w answer all parts.
(3) Halfunction(s) other than the initicting one MAY be involved.
ANSWER the following:
a.
Whw does core flow decrecse [ Point 13 and whw doesn't it decrease to zero CPoint 237 (1.0) b.
Whw does reactor Pressure increase EPoint 23 and remain high CPoint 437 (1,0) c.
Whw does reactor level decrease initiallu CPoint 53 and what is causing the Peaks in level later CPoint 637 (1.0)
t
'54__IWEORY_DE_WUCLECR_E0WER_ELAWI_DEERAIIDWa_ELUIDSz_eWD PAGE 5
IHERHODYNAMICS QUESTION 5.08 (2.75)
Consider the attached process comeuter P-1 Printout, Figure B, part of an ACTUAL P-1.
a.
IS the output signal,from APRM 1 MORE or LESS conservative than the output signal from APRM 37
(.75) b.
If all the fuel has a design LHGR limit of 13.4 KW/ft.,
WHAT is the HIGHEST actual LHGR in the core?
(1.25) c.
IS the axial Power distribution bottom or top core reekedi
(.75)
QUESTION 5.09 (2.25)
The following statements are concerned with suberitical multirlica-t i o ri.
CHOOSE ONE of the capitalized Words to make each statement true.
a.
As K-effective approaches unitw, a (LARGER / SMALLER) change in neutron level occurs for a given chence in K-eff.
(.75) b.
As K-eff approaches unitw, e (SHORTER / LONGER) period of time is reauired to reach the euuilibrium neutron level for a diven change in K-eff.
(.75) c.
As k-eff approaches unitu, the count rate doublind technioue becomes (HORE/LESS) accurate.
(.75) l l
i I
f f
I
r 6.__EL'eWI_SYSIEOS_DESIGWa_CONIEDLa_eWD_IWSIEUMEWIDIIDW PAGE 6
QUESTION 6.01 (3.00)
When operating the Residual Heat Removal (RHR) sustem in SHUTDOWN COOLING MODE:
a.
WHY is it necessarw to have the rehire rumr in the return loop shutdown with its discharge 2n;qsJ..un..
u uuros, v..ves ve shut?
(1.0) b.
WHY must Power be secured to the minimum flow valvet (1.0) c.
HOW will the swstem be effected if reactor Pressure exceeds 75 esis?
(1.0)
QUESTION 6.02 (3.50)
With the Plant operating at 100% Power (Unit 2), recire in Master Manual, all operator inadvertent 1w DECREASES the ' Pressure Set
- bw 5 Psi.
WHAT will be the INITIAL receonse end FINAL status of the following Parameters due tu this action?
Brief1w EXPLAIN.
Assume NO operator action.
See attached Figure 0590-6, EHC Logie Diagram.
ANSWER on the attached handout page.
a.
TCV position b.
BPV position c.
Power d.
Pressure GUEST 10N 6.03 (2.50)
Concerning the Control Rod Drive (CRD) Hudraulie Sustem a.
Upon completion of a reactor scram with all CRDs fu11v inserted, WHAT are the TWO (2) sources of water contin-uing to fill the serasi discharge volusiv until the scram is reset?
(1.0) b.
Brief1w EXPLAIN WHY Pressure eOuali2cr Valves Were in-stalled between the cooling water line and the exhaust water header.
(1.5)
6.__EL'eWI_S%SIEMS_DESIGNo_CONIROLo_eWD_INSIEUMEWIAIIDW PAGE 7
QUESTION 6.04 (3.00)
With regard to the Reactor Recirculation Flow Control Sustem; a.
WHAT Plant conditions will cause a 60 % recire Pump runback?
(Seteoints reuuired)
(2.0) b.
With the Plant operating at 23% Power and minimum floWe an operator inadvertent 1w shifts the H/A transfer station for recite Pump
'A' from "Hanual' to ' Auto."
Assuming N0 further operator vetion, brief1w EXPLAIN WHAT will happen to the speed of
'A' recire Pump.
Continue wour discussion to the final steadw state speed.
(1.0)
DUESTION 6.05 (3.00)
WHAT are SIX (6) different automatic.act2ons which should occur if Main Steam Line Radiation Monitors
'C' and
'D' exceed their bish-high trip setpoint (3 x normal FPB) while operating at rated con-ditions?
Specifw individual component actions.
(3.0)
QUESTION 6.06 (3.00)
Concerning the Standbw Liould Control Swstem a.
WHY is it necessarw for the sustem to be capable of in-Jeeting the contents of the SLC tank in a HAXIHUH time of 125 seinutes?
(1.0) b.
WHY is the SLC Pump suction Piping heat tracedT (1.0) c.
WHAT are THREE (3) uses of the SLC injection sPurser, OTHER THAN Poison injection?
(1.0)
~ 6.__EL'eWI.SYSIEMS_DESIGWa_CONIROLa_eWD_INSIEUMEWIAIIDW PACE 8
1 QUESTION 6.07 (3.00)
With Unit 2 operating at 50% power in 3-element controle HOW will the Feedwater Control Swstem (FWCS) respond to each of the following instrument failures?
Your answer should include how the instrument eill fail (high or low), the FWCS recronse and effect on cetual water levele as well as anw automatic actions es a result of the change in actual vessel level.
a.
A steam flow detector's sensorw diarhragm ruptures.
(1.0) b.
The level channel selected for input to the FWCS loses instrument Power.
(2.0)
GUESTION 6.08 (4.00)
For each of the HPCI (High Pressure Coolant InJeetion) Sustem component failures listed below, STATE WHETHER OR NOT HPCI WILL AUTO INJECT into the reactor vessele IF IT WILL NOT INJECT WHYe AND IF IT WILL INJECT, Provide ONE POTENTIAL ADVERSE EFFECT OR CONSEQUENCE of sustem operation with the failed component.
Assume NO OPERATOR ACTION, and the component is in the failed condition ct the tinie HPCI receives the auto initiating signal.
a.
The GLAND SEAL EXHAUSTER fails to crerate.
(1.0) b.
The turbine AUXILIARY LUBE DIL PUHP fails to operate.
(1 0) c.
The HINIMUM FLOW VALVE fails to auto open (STAYS SHUT) when swstem conditions reuuire it to be open.
(1.0) d.
The HPCI PUHP DISCHARGE FLOW ELEMENT output signal to the HPCI flow controller is failed at its maximum outrut.
(1.0)
. 2 __EROCEDURES_=_WDEMALa_eBNDEHALc_ EMERGENCY _eWD PAGE 9
.EADIOLOGICAL_COWIROL 4
QUESTION 7.01 (3.00)
Cecording to GP-2, ' Normal Plant Startup*
a.
During the spesoach to critical, WHICH rod (s) in a new Rod Group is(are) likelw to exhibit high notch worth?
(1.0) b.
During heatuP to rated temPeratuiO and Pressure after raising the EHC Pressure setroint to 600 psig, EXPLAIN WHY one turbine bweass valve is oversed 10 to 20% (with the bwpass valve Jack)?
(2.01
~.
QUESTION 7.02 (2.00)
With the teactor shutdown end cooled down to less than 212 degrees Fe cecording to GP-12, ' Core Cooling Procedure':
a.
If on1w one loop of Shutdown Cooling is ON with NO recire flow thsough the other loop, reactor level must be main-tained high enough to prevent WHAT from occurring?
(1.0) b.
HOW would coolarit temperature stratification be indicated bw reactor vessel metal surfcce temperaturesT (1.0)
QUESTION 7.03 (3.00)
When starting a recire eume at Power (the other recire pume operating) in accordance with Start-Ur of a Recirculation Pume,'
Procedure S.2.3.1.A!
a.
WHAT ARE THE BASIS for the two (2) coolant temperature conditions that must be met erior to stceting a PumrT (2.0) b.
WHY must tiie speed of the operating loop be less than or eaual to 40%7 (1.0)
OUESTION 7.04 (2.00)
If an uriexpected or urienplained increase in drwwell Pressure occurs, WHAT TWO (2) ismediate operator actions should be rerformed per OT-101, 'High Drwwell Pressure Procedure," assuming Pressure did NOT reach the sciam setpoint?
2.__EEDCEDURES_=_WDEMALo_ABWDEMALo_EMERGEWCY_eUD PAGE 10
+ RADIOLOGICAL _CONIEDL QUESTION 7.05 (2.50)
Regarding DN-113, " Loss of RBCCW':
a.
WHAT THREE (3) automatic actions should be verified if the non-regen heat exchanger outlet temperature reaches 200 degrees F?
(1.5) b.
When shutting down the recire Pumps, WHY are theu first runback to minimum speede then tripped 10 seconds apart?
(1.0)
QUESTION 7.06 (3.00)
Brief1w discuss WHY wach of the overall Plant Preccutions below for the Transient Response Implementation Plan (TRIP) Procedures are necessarw.
a.
Do NOT dePressurize the reactor below 100 Psig unleCS motor driven pumps are readw to inject.
(1.5) b.
Do NOT initiate drwwell spraws unless torus water level is below 18.5 feet.
(1.5)
QUESTION 7.07 (3.00)
WHAT are the FOUR (4) entrv conditions for the 'Recetor Pressure Vessel (RPV) Control,' TRIP Procedurve T-101?
QUESTION 7.08 (2.00)
WHAT are TWO (2) reasons for reuuirine bw Procedure that HPCI and RCIC NOT be operated at less than 2200 RPM 7 OUESTION 7.09 (3.00)
WHAT are the entrw conditions for a ' Loss of Stator Coolinge' Per OT-1137
2.__EROCEDURES_=_t0E 4La_eBNDEHela_EHERGENCY_eWD PAGE 11
.R&DIOLOGICAL_CONIROL QUESTION 7.10 (1.50)
MATCH the following whole bodw Emergenew E>:Posure Limits with the CPProPriate emergenew function as provided in the Plant Emergenew Procedures.
EXPOSURE LIMIT ENERGENCY FUNCTION a.
5 REM 1.
Overation of euuiPuent to mitigate an emergenew.
b.
25 REM 2.
Life saving and reduction of inJurw.
c.
75 REM 3.
Protection of huolth and safetw of the rublic.
4
' 8.__ AD'f;IWISIR AIIVE_ERDCEDURESa_COWDIIICWSn_ AWD_LIMIIeIIDWS PAGE 12 QUESTION 8.01 (2.50)
Renarding the Techreical Srecificatiores for Cure Alterationst a.
HOW MANY SRM's must be urerable?
(0.5) b.
WHERE must the OPERABLE SRrl detectors be luectedT (1.0) c.
WHAT conditions must be satisfied before an SRM is con-sidered OPERABLE 7 LIST TWO (2).
(1.0)
DUESTION 8.02 (3.00)
Assume that Unit 2 is at aperuximatelv 85% Power and that all conditions are normal when wou relieve the shift at midnight.
Later in the shift WIDE RANDE DRYWELL PRESSURE RECORDER PR-4805 (0-70 rsig) feils low.
In cecordance with the Technical Specifications, WHAT ACTIONS HUST YOU TAKE IN THIS SITUATION 7 (3.0)
NOTE! USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS TO TO ANSWER THIS QUESTION.
FULLY REFERENCE ALL APPLICABLE SEC-TIONS OF THE T.S.
THAT YOU USE TO DEVELOP YOUR ANSWER.
QUESTION 8.03 (3.00)
In reference to the pruesdure for Control of Ignition Sourcest a.
WHEN is a Tech Spec fire watch reuuired?
(1.0) b.
HOW are the duties of a Tech Spec fire watch different from those of a Dedicated fire wetch?
(2.0)
QUESTION 8 04 (1.50)
WHAT is the significance of the WHITE, YELLOW, and GREEN color coding cn ALARN CARDS?
(1.5)
"Br.__eDHINISIEeIIWE_EROCEDUEESz_CONDIIIONSA_AND_LIMIIDIIDWS PAGE 13 QUESTION 8.05 (2.00) o.
WHAT are the THREE (3) parameters wou use to determine whether a Jet pump is OPERABLE 7 (1.0) b.
WHY must the Plant be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if one or more Jet Pumps is INOPERABLE 7 (1.0)
QUESTION 8.06 (1.50)
WHY is the reactor Pressure vessel, rather than the reeire Pirings the cost limiting Component of the reactor coolant swstem when considering the high Pressure safetw limit?
(1.5)
QUESTION 8.07 (3.50)
Unit 2 is at 65% Power and all conditions are normal With the folloWind oxception:
One Diesel Generator is INOP due to water being found in the fuel during the nionthlw fuel check.
Subseauentlw, during the Core Spraw Operabilitw check, the
'A' Core Spraw Pump would not start.
In accordance with the Technical Specifications, WHAT ACTIONS MUST YOU TAKE IN THIS SITUATION?
(3.5)
- f**********************
Note: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS TO ANSWER THIS QUESTION.
FULLY REFERENCE ALL APPLICABLE SECTIONS OF THE T.S.
THAT YOU USE TO DEVELOP YOUR ANSWER.
0; "_ S T I O N 8.08 (2.50)
For the suppression chamber water temperatures listed below, WHAT ACTION (S) in(are) reuuired bv the Unit 2/3 Technical Srecifications with the unit MODE SWITCH in RUN.
(2.5) a.
97 degrees F.
b.
106 degrees F.
during HPCI testing c.
121 degrees F.
following a s e r a si with the MSIV'S SHUT
"B.__ADMINISIRAIIWE_EEDCEDUEESo_CONDIIIOUSa_AND_LIUIIeIIONS PAGE 14 QUESTION 8.09 (2.00)
With Unit 3 at 100% Power, an il)ConsistenCw in the temperature indica-tions for the siain steani line tunnel exhaust duct is observed bw the STA.
Investigation reveals that temperature elements, TE-5931A und TE-5931B are pulled out of the exhaust duct.
In accordance with the Tech-nical Specifications, WHAT ACTIONS MUST YOU TAKE IN THIS SITUATION 7 (2.0)
NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS TO ANSWER THIS QUESTION.
FULLY REFERENCE ALL APPLICABLE SECTIONS OF THE T.S.
THAT YOU USE TO DEVELOP YOUR ANSWER.
DUESTION 8.10 (3.50)
Preparations are underwaw for a reactor startue on Unit 2 when the 0125 VDC BATTERY CHARGER FAILURE' alarms.
Uron investigation, it is found that the 2BCA 125VDC Betterv Charser output is 90VDC.
In accordance with the Technical Spet4fications, WHAT ACTIONS MUST BE TAKEN DUE TO THIS INSTRUMENT FAILURE 7 (3.5)
Note: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS TO ANSWER THIS OUESTION.
FULLY REFERENCE ALL APPLICABLE SECTIONS OF THE T.S.
THAT YOU USE TO DEVELOP YOUR ANSWER.
- ct******
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ANSWER SHEET for Question 6.02_
INITIAL RESPONSE:
a.
TCV position (in % steam flow demand) b.
BPV position (in % steam flow demand) c.
Power (increase, decrease,or d.
Pressure remain the same)
Reason:
FINAL STATUS:
a.
TCV position (in % steam flow demand) b.
BPV position (in % steam flow demand) c.
Power (higher than, lower than, or d.
Pressure the same as the initial value)
Reason:
i' l
l l
l j
I-
EQUATION SHEET f = ma v = s/t Cycle efficiency = (Net work out)/(Energy in) 2 w = mg s = V t + 1/2 at o
2 E = mc KE = 1/2 mv a = (Vf - V )/t A = \\N A=Ae" o
g PE = mgn Vf = V, + at w = e/t x = tn2/t1/2 = 0.693/t1/2 1/2'N " U*U S
- 2 t
y,y 3p nD A=
[(t1/2)
- I*b)3 4
t.E = 931 am m = V,yAo
-Ex Q = mCpat d = UA A T I=Iec Pwr = W ah I=I 10-x/TVL f
n TIL = 1.3/u sur(t)
HVL = -0.693/u P = P,10 P=Peo SUR = 26.06/T SCR = S/(1 - K,ff)
CR = S/(1 - Keffx) x CR (1 - K,ffj) = CR (I ~ "eff2)
SUR = 26a/t* + (a - o)T j
2 T = ( t*/o ) + [(S - o V Io]
M = 1/(1 - K,ff) = CR /CR j
3 T = 1/(o - 8)
M = (1 - Keffo)/II ~ Keffl)
T = (a - o)/(Io)
SDM = ( -K,ff)/K,ff o = (K,ff-1)/K,ff = aK,ff/K,ff t* = 10 seconds I = 0.1 seconds-I o = [(t*/(T Keff)] + [S,ff (1 + IT)]
/
Idjj=Id P = (r*V)/(3 x 1010)
I d) 2,2 2 gd j
22 2
I = eN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (f,,g)
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem.
1 curie = 3.7 x 1010eps 1 gal. = 3.78 liters 1 kg = 2.21 lem 3
1 ft3 = 7.48 gal 1 np = 2.54 x 10 Stu/nr Density = 62.4 1 /ft3 1 mw = 3.41 x 106 Btu /hr Density = 1 gm/c, lin = 2.54 cm Heat of vaporization = 970 Stu/lom
- F = 9/5*C + 32 i
Heat of fusion = 144 Stu/lbm
'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg.
1 BTU = 778 ft-lbf I ft. H O = 0.4335 lbf/in.
2
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j 5 A__IBEDEY_DE_WUCLE AR_E0 JE s_ 'M DNg_ELUIDSz_eWD PAGE 15 IHERBODYWedICS ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 5.01 (1.50)
- c. LHGR.
(.75) b. APLHGR or HAPLHGR (.75) REFERENCE PBAPS L/P Vol. 6, Lot 1400, eg. 10
- 1410, pg. 2 EDH-378 ANSWER 5.02 (3.00) a.
More negative CO.53. Because each degree increase in moderator temperature results in a larger moderator densitu decrease E1.03. (1.5) b. Less nesative EO.53. Beceuse as control rod densitu decreases, neutrons leaking f rosi the volume near the fuel rods have less of a chance for non-fission absorption E1.03 (and a greater chance to cause fission). (1.5) REFERENCE PBAPS L/P Vol. 6, Lot 1450, eg. 4 EDH-379 ANSWER 5.03 (3.00) a. The formation of bubbles serves to agitate and break-up EO.63 the relativelu stagnant fluid boundarv film CO.63. As nucleate boiling progresses, relative 1w colder water rerlaces the gaps on the clad surface left bw the bubbles es thew detach and move into the coolant stream EO.63. (1.8) b. The edge and corner rods can dissipate heat bw radiation awaw from the bundle EO.63 while the central rods radiate much of their heat to other central rods [0.63. (1.2) REFERENCE PBAPS L/P Vol. 6, Lot 1340, eg. 1-6 EDH-380
~ l ' 5.__IWEORY_DE_WUCLEAE_EDWER_EleWI.0EERAIIDWa_ELUIDSa_AND PAGE 16 .IHERMODYNAMICS ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 5.04 (3.00) a. Remains constant E.253. Flow is controlled bu the RCIC flow con-troller which will attemet to maintain a constcnt output flow re-gardless of reactor Pressure E.753. (1.0) b. Decreases C.253. The flow controller functions to maintain e con-stant flow, thus Pump discharge Pressure is decreased along with the decreasing reactor Pressure to mairitain constant flow. OR Since .the flow controller maintains flow to the reactor, as reactor Pressure decreases, the eump discharge head must decrease to maintain a con-stant flow (constarit NPSH) 0.753. (1.0) c. Decreases E.253. Since Pump discharge head is decreasing to main-tain a constant flow, turbine RPM must also decrease E.753. (1.0) REFERENCE PBAPS L/P Vol. 2, Lot 0380 ANSWER 5.05 (3.50) a. The decrease in the burnout term EO.53 with the Production of Xenon from Iodine still at the higher Power rate dominutes E0.53 causirig the xenon concentration to increase. (1.0) b. Peripheral rod worth will increase EO.53 because the highest l xenon concentration will be in the center of the core l where the highest flux existed Previous 1w E0.53. This will suppress the flux in the center of the core and increase the flux its the area of the Peripheral rods, therebu, l 2ncreasing their worth E0.53. (1.5) l l c. More than one half the value at 100% Power. The removal of j xenon is direct 1w proportional to the xenon concentration but it also depends on flux (Power). Therefore the removal rate would become less significant. (1.0) l REFERENCE PBAPS L/P Vol. 6, Lot 1510, Ps. 6,7 EDH-384 l ~, m. _.
5 __IWEORY_DE_ NUCLE 6B_EDWER_ELANI_DEEEeIIONa_ELUIDSz_eND PAGE 17 .IHEEMODYNAMICS ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 5.06 (3.00) (1) Due to void collapse caused bw the high APRM scram er in-creased recire suction flow from the annulus. (0.5) (2) Due to the increase in recire flow. (0.5) (3) Due to the unaffected recire Pume runback to min. when feedflow decreases to <20%. (0.5) (4) Due to increasing reactor Power. (0.5) (5) Due to FWCS response to decreasing reactor water level. (0.5) (6) Due to TCVs opening to control reactor Pressure. (0.5) REFERENCE BWR-4 Transients BXY-4 EDH-385 ANSWER 5.07 (3.00) r). RPT on turbine trie [0.53. Natural circulation from decaw heat [0.5]. (1.0) b. Turbine BPV's fail closed CO.53. SRV's control Pressure at higher value CO.53. (1.0) c. Void collapse due to pressure increase and the scram E0.53. Level swell from SRV's lifting E0.5]. (1.0) (p. No itPT aecp). A.9t +<tr a.t <u.:n. p-9 s. REFERENCE BWR-4 Transients DXY-4 EDH-386 O ANSWER 5.08 (2.75) (.75) a. Less
- b. Maximum LHGR = 0.846 x 13.4 kw/ft = 11.34 KW/ft (1.25)
(.75) c. Top r REFERENCE l PBAPS L/P Vol. 6, Lot 1400, pg. 9-10 EDH-387
5.__IHEORY_DE_WUCLEAR_EDWER_ELAWI_DEEBAIIOWz_ELUIDSa_eWD PAGE 18 .IHERHODYNAMICS ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 5.09 (2.25) e s. L a rsie r (.75) b. L oriste r (.75) c. More (.75) REFERENCE PBAPS Reactor Theorwi Vol. 6, Lot 1430 EDH-388 ('
s .6a__ELeWI_SYSIEMS_ DESIGN 2_CONIEDLa_eWD_INSIEUMENIGIIDN PAGE 19 ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 6.01 (' 3. 0 0 ) c. To Preverst reciPC Pump bearing damage due to reverse flow. (1.0) b. To prevent draining the reactor vessel to the torus. (1.0) Inboard and outboard suction valves shut, RHR pumps trie. (1.0) c. REFERENCE PBAPS Lot Lesson Plan 0370, Ps. 16,17 EDH-356 ANSWER 6.02 (3.50) INITIAL RESPONSE: a. TCVs - Remain at 100% open (or open to 100%) C.353. b. BPVs - OPen 16.5% E.353. c. Power - Decreases [.353. d. Pressure - Decreases E.353. REASON: Above caused be PCU calling for "115% steam flow ((950-915) x 3.3) E.353. ( tmA t.d uk @ tog % - mc r se4 a tioto) FINAL STATUS: a. TCVs - At 100% Position (or initial) E.353. b. BPUs - Shut E.353. c. Power - Slight 19 lower C.353. d. Pressure - Slight 1w lower E.353. REASON: Above caused bw the decrease in pressure end Power causing BPVs to shut -- PCU cweling to new euuilibrium state ((945-915) x 3.3) E.353. REFERENCE PBAPS Lot Lesson Plan 0590, pg. 6-17 EDH-359 L
6.__ELAWI_SYSIEMS DESIGWa_CONIROLa_eWD_INSIEUMEWIoIIDW PAGE 20 ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 6.03 (2.50) s. Reactor water leakind Past the CRD seals E0.53 and chargins water from the CRDH Swstem E0.53. (1.0) b. To repressurize the exhaust header following & scramEO.53 and Prevent excessivelv high crd operating dPEO.53 during subseouent operation of a selected rodeo.53 (1.5) REFERENCE PBAPS Lot Lesson Plan 0070, Pg. 13,16,17 EDH-360 ANSWER 6.04 (3.00) a. 1. Reactor water level <17'EO.53 AND
- 2. An individual RFP flow <20%EO.53 OR 3.
Total feed flow >90%EO.53 AND 4. Anw condensate Pump not runninMEO.53 (2.0) b. Pump speed will increase to 50% at which time the Master Controller low speed limiter Will be limitinM. (1.0) REFERENCE PBAPS Lot Lesson Plan 0040, Pg. 9,10 EDH-361 ANSWER 6.05 (3.00) 1. Reactor Scram 2. MSIV closure
- 3. Recire sample line isolates (close) 4.
Main Steam sample line isolates (Close) 5. RHR sample line isolates (close)
- 6. Main Steam line drains isolate (close)
(6 reuuired 0 0.5 each) (3 0) 7-OkJu Au.\\ vxs-pomp +<.y REFERENCE PBAPS Lot Lesson Plan 0180, TP 43 EDH-362
64._EL'eWI_SYSIEMS_DESIGWA_CDWIEDLA_eWD_IWSIEUMEWIeIIDW PAGE 21 ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 6.06 (3.00) o. Poison injection must be fast enough to overcome reactivitw due to cooldown. (1.0) b. To ensure that the PPison solution does not solidifw its the lines and make the swstem inoperable. (1.0) c. o Core Plate dP. o Core Spraw Swstem line break detection. o Jet Pump dP. (3 at 0.33 each) (1,0) REFERENCE PBAPS Lot Lesson Plan 0310e pg. 6,7 and TP 42 EDH-363 ANSWER 6.07 (3.00) a. Fails lowE.253 causing SF/FF mismatchE.253 decreasing feed pump speedE.253. Actual level will be maintained a few inches lower E.253. (1.0) b. Fails lowE.253 causes level error [.253 increasing feed pump speed [.253 to HSSE.253. Actual level increases [.253 resulting in Main Turbine tripe.253, Feed pump triest.253, and Reactor scramE.253. (2.0) REFERENCE PBAPS Lot Lesson Plan 0550p Pg. 19,20 EDH-364
t 4 6.__ELAWI_SYSIEMS_DESIGWa_CDWIEDLa_AND_INSIEUMEWIDIIDW PAGE 22 ANSWERS -- PEACH BOTTOM 283 -84/08/28-HILL, D. ANSWER 6.08 (4.00) c. Will inject [0.253. Turbine seal 3-akage r e s u l t i n t.! in potential air-borne activitw in the HPCI room E0.753. (1.0) b.- Will not inject E0.253. Turbine stop and control valves will not open [0.753. (1.0) c. Will inject E0.253. Pume overheeting and seal damuse meu result during low or no flow conditions [0.753. (1.0) d. Will not inject [0.253. Maximuni sisinal from the flow element will cause the controller to keep turbine speed at m i ni sium EO.753. (1.0) REFERENCE PBAPS L/P Vol. 2, Lot 0340 4 - - ~. - - -.,,-...,,-..,-.,.,,,y-.... -,.----,y- ..er--, .,-,-------~~e--
2 __EkDCEDURES_=_WORMALa_4BWORMela_EHERGENCY_AND PAGE 23 .RADIOLODICAL_CONIROL ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 7.01 (3.00) s.- The first rod (s). (1.0) b. To establish a constant flow in and out of the reactor vessel (OR a constant feed flow demand) [0.663 to minimize temrerature fluxuations on the feedwater nozzles [0.673, thus siinimizing the Probabilitw of thermal stress cracking [0.673. (2.0) REFERENCE GP-2, Rev. 35, PP. 12 8 16. ANSWER 7.02 (2.00) e. All core flow being diverted through the idle Jet Pumps (beesssing the core). (1.0) b. Metal surface temperatures beloW Water level Will be less than indicated coolant temperature. (1.0) REFERENCE GP-12, Rev. 5, Pe 9e ANSWER 7.03 (3.00) s. Assures that the changes in coolant temperature at the reector vessel nozzles and bottom hecd region are cceeptable. (1.0) Assures that ASME Codes are not exceeded when recire Pumps are started and the colder Water in the bottom of the vessel is forced uP to the upper regions of the vessel. (1.0) b. To avoid re ce i vi nsi a APRM Hi flux scrani from the PoWor spike following the Pump blart. (1.0) <cacA\\ ' Abc, ug"vkdh =% 4k' Ad P m?' *M d b 419 REFERENCE S.2.3.1.Ae Rev. 4, p. 2, and LOT-0030, P. 26.
2.__ERDCEDURES_=_WORMALa_eBWORHOLa_EHERDENCY_eWD PAGE: 24 .RADIOLDDICAL CONIROL -CNSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. CNSWER 7.04 (2.00) o Maximize drwwell cooling (1.0) o Terminate drwwell inerting (1.0) REFERENCE OT-101, Rev. O, P. 1. ANSWER 7.05 (2.50) a. o OrouP IIs isolation (OR HD-12-15, MD-12-18, and MD-12-68 close). o RWCU Pumps trip. o RWCU demin hold Pumps start. (3 at 0.5 each) (1.5) b. To minimize the reactor t rcosierit. (1.0) REFERENCE ON-113, Rev. O, p. 1, and DN-113 Bases, P. 1. ANSWER 7.06 (3.00) e. Ensures depressurization below the HPCI and RCIC isolation setpoints will not occur unless motor driven inJuction Pumps available. (1.5) b. The suppression chamber to drwwell vacuum breakers would be covered with water E0.53 making it possible to exceed the design DW negative internal Pressure CO.53 following the drastic redJetion in DW internal Pressure caused bU DW SPrgw initiation CO.53. (1.5) ' REFERENCE LOT-1560, Rev. O, PP. 8 1 9. m-
2a__ERDCEDURES_=_WORdeLa_eBNDRHSLa_ EMERGENCY _eWD PAGE 25 . RADIOLOGICAL _CONIRDL ANSWERS -- PEACH BOTTON 213 -84/08/28-HILL, D. i ANSWER 7.07 (3.00) 1. RPV level below -48' or unknown. 2. Drwwell pressure is above 2 esis. 3. A Group I isolation occurs. 4. Scram condition with power above 3% or unknown. (4 reouired 9.75 each) (3.0) REFERENCE LOT-1560, Rev. O, p. 7, and T-101. ANSWER 7.08 (2.00) 1. Ensures sufficient shaft driven lube oil Puinn pressure and flow. (1.0) 2. Ensure high enough steaming rate to avoid exhaust line check valve " chatter" and possiblU subseGuent damage. (1.0) REFERENCE LOT-1560, Rev. O, p. 14. ' ANSWER 7.09 (3.00) A stator coolant swstem troubic s l a r si in conjunction with (.75) o an automatic turbine generator runback OR (.75) o the trie of both recire Pumps OR (.75) o e ' Generator Stator Slots Hi Teme.' clarm (.75) REFERENCE OT-113.
2.__ERDCEDURES_=_NORMALa_eBWORMALa_EHERGENCY_AND PAGE 26 .ReDIOLOGICAL_CONIROL ANSWERS -- PEACH BOTTuM 213 -84/08/28-HILL, D. i ANSWER 7.10 (1.50) s.' 3 b. 1 l c. 2 (3 at 0.5 each) REFERENCE EP-103, Rev. 9, P. 9. v s ,e,,-.e,,, -,,,,--,,,,._..,,,.,,,,,.,y,m,,-, ,..n., ,,,y,,n_
B.__eDMINISIRAIIVE_EEDCEDURESz_CONDIIIONSz_AND_LIMIIoIIONS PAGE 27 . ' ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 8.01 (2.50)
- s. Two (0.5) b.
One in the core cuadrant where fuel or control rods are being moved and one in the adjacent uuadrant. (1.0) c. 1. The SRM shall be inserted to the normc1 operating level. (0.5) 2. The SRM shall have a minimum of 3 ces with all rods fullw inserted in the core. (0.5) REFERENCE Technical Specifications 3.10.B.1, Ps. 227,228 EDH-347 ANSWER 8.02 (3.00) Table 3.2.F. SURVEILLANCE INSTRUMENTS E1.03 reuuires a minimum of 2 operable instruments E0.53 however NOTE 41 to the table L1.03 states that if the Parameter is reduced to one indication that continued oPer-otion is permissible on1w during the succeeding 30 daus unless such inst-rumentation is sooner made operable [0.53. REFERENCE Technical Specifications Table 3.2.F., pg. 77-78 EDH-348 ANSWER 8.03 (3.00) a. When an ignition source is to be used in an area, governed bw T.S. 3.14, which has, or will have as a result of the work scope inoperable fire Protection eGuiPment. ET.S. number not reuuired] (1.0) b. The Dedicated fire watch is not reuuired during Periods when the ignition source is not in use and which are at least 30 minutes after the most recent use of the ignition source. The Tech Spec firewatch is continuous regardless of whether or not the ignition source is in use, until SS/SSV advises that the inor fire swstem is back in service. (2.0) REFERENCE Admin Procedure A-12, pg. 3,4 EDH-350
~~ l Bi__eDMINISIR AIIUE_ERDCEDURESz_CDWDIIIDWSa_eUD_LIMII AIIDWS PAGE 28 ' ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 8.04 (1.50) White-common to both units (0.5) Yellow-unit 2 on1w (0.5) Green-unit 3 on1w (0.5) REFERENCE Admin Procedure A-11, eg.1 EDH-351 ANSWER 8.05 (2.00) o. 1. Recirculation Pume Flow 2. Total Core Flow 3. Diffuser-to-Lower Plenum dP o.- 3=t 7 m A7 (3 reauired for full credit) (1.0) b. In the case of a DBA, the blowdown area is inereesed and the caPabilitw for reflooding the core is reduced. (1.0) REFERENCE Technical Specifications 4.6.E, eg. 148i Bases, pg. 159 EDH-352 ANSWER 8.06 (1.50) Even-though the recirculation suction rivins design pressure is lower (1148), the ANSI code allows 120% (1377.6) of design for PiPins. ASME codes allow 110% of design pressure for maximum allowable Pressure transients. (1250 x 110% = 1375) (1.5) REFERENCE PBAPS L/P Book VIII, Lot 1820, eg. 15 EDH-353
_ 8.__eDMINISIRGIIVE_EBOCEDUEESz_COWDIIIONSz_AND_LIMIIeIIONS PAGE 29 ' ANSWERS -- PE ACH BOTTON 213 -84/08/28-HILL, D. ANSWER 8.07 (3.50) T.S. 3.9.B.3 E1.03 states continued reactor operation-is Persiissible in accordance with 3.5.F. if 3.9.A.1 is satisfied EO.53. Both off-site cources and both the startue and emergenew transformers are available so 3.9.A.1 is satisfied E.253. 3.5.F.1 [1.03 states that with one D/G inoPF all low Pressure Core coolitsM must be operable EO.53. Therefore on orderlw shutdown of the reactor must be initiated and the reactor in Cold Shutdown in 24 hours [.254 cust be (Rduence 4o T.S 3,4 4 r9l red 4.e MI wedib REFERENCE Technical Specifications 3.9.B.3, 3.5.F.1p Pde 132F221 ANSWER 8.08 (2.50) a. Restore the temperature below 95 decrees F. ie. initiate suppression Pool coolinM. (0.5) b. Stop HPCI testing and reduce temperature below 95 decrees F. le. initiate suPPressiott Pool cooling. (1.0)
- c. DeFressurize to less than 200 Psis at normal couldown rates.
(1.0) REFERENCE Technical Specifications 3.7.A.Cp Pg. 165a EDH-355 ANSWER 8.09 (2.00) T.S. Table 3.2.A E0.53 reouires two operable instrument channels Per trip swstem and one of.two instrument channels is defeated in each trie swstem E0.53. Note 2B for Table 3.2.A [0.53 states to initiate an orderlu load reduction and have Main Steam Lines isolated within eight hours E0.53 (2.0) REFERENCE Technical Specifications, Notes for Table 3.2.A, and Table 3.2.A
3 8.__&'lNISIRAIIVE_ERDCEDURES4_CONDIIIDUS2_eUD_LIMIIel10NS PAGE 30 j " ANSWERS -- PEACH BOTTOM 213 -84/08/28-HILL, D. ANSWER 8.10 (3.50) T.S. 3.9.A.4 E1.03 reauires 4 unit batteriee and their chargers to be operable if the reactor is to startur CO.51 T.S. 3.9.B.5 E1.OJ allows continued reactor operation during the suceeedini: 3 dawu within electri-cal safetw considerations, provided repair work is initiated EO.53. T.S. 3.0.D allows operation to be soverned bw the LCOs for the power source, not the individual action statements for each swstem determined to be in-CPeratiVe solelv because of the Power source E0.5J LF;ra4 :,e k e u. Q b\\* %- Co erek V) REFERENCE Technical Specifications, 3.9.A.4, 3.9.B.5, 3.0.D}}