IR 05000322/1985034
| ML20137R778 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 01/29/1986 |
| From: | Keller R, Kister H, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20137R749 | List: |
| References | |
| 50-322-85-34, NUDOCS 8602130407 | |
| Download: ML20137R778 (92) | |
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{{#Wiki_filter:, U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT N0. 85-34 (0L) FACILITY DOCKET NO. 50-322 FACILITY LICENSE N0. NPF-36 LICENSEE: Long Island Lighting Co.
P.O. Box 618 Wading River, N.Y.~11792 FACILITY: Shoreham Nuclear Power Station EXAMINATION DATES: September 16, 1985 to September 20, 1985 CHIEF EXAMINER: ~
&n48 / J [
Dave range, Le(#/ Reactor Sngineer Ddte '
(Examiner)
REVIEWED BY: ) / //27[N Robert M. Keller7 Chiefj Projects Section 1C Date APPROVED BY: 49' Ha'rry B. Kl(er, Chief, hate / Projects Branch No. 1 SUMMARY: Operator and Senior Operator Initial Cold License Exams were conducted at Shoreham Nuclear Power Station from September 16, to September 20, 1985. Four (4) Reactor Operator, four (4) Senior Reactor Operator, and one (1) Instructor Certification candidates were examined. All candidates passed the written and oral examinations.
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REPORT DETAILS-
~ ~ TYPE OF EXAMS: ' Initial b :
EXAM RESULTS:
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I R0 l SR0 . l Inst. Cert I ,.. l Pass / Fail l Pass / Fail 1 Pass / Fail l ' I I I l-
- 1. . . l .I I I l Written Examl 4/0' l .4/0 1 ^W / .I I I I I 'l l I l .
1 -l' < 1 Oral Exam 1. -4/0 1 4/0 l 1/0 I ik I I I I I I I l- 1 I ISimulator Exam l NA/' 1 NA/ 1 'NA/0' l-l~ l i I -I
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1. CHIEF EXAMINER AT-SITE: D. Lange, USNRC . '2 . .0THER EXAMINER: F. Crescenzo, USNRC !-
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1. No generic strengths or deficiencies were-noted on the written and oral exams administered.
2. Personnel present at Exit Interview: NRC Personnel D. Lange, Chief Examiner F. Crescenzo, Operator. Licensing Examiner
.J. Berry, Senior Resident Inspector (Shoreham)
Facility Personnel J. Scalice,-Operations Manager, LILCo L. Calone, Nuclear Training Manager, LILCo K. Rottkamp, Training Supervisor, LILCo J. L.' Smith, Manager, Nuclear Operations Support Dept., Shoreham 3. Summary of NRC Comments made at exit interview: None 4. Summary of facility comments and commitments made at exit interview: None 5. Changes made to written exam during examination review:
(See Attachments 3 and 4)
Attachments:
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1. Written. Examination and Answer Key (RO) 2. Written Examination and Answer Key (SRO)
-3. Facility Comments on Written Examinations 4. NRC Resolutionaof Facility Comments on Written Examinations
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U. S. NUCLEAR REGULATORY COHNISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SHOREHAh
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REACTOR TYPE - BWR-GE4 _________________________ DATE ADMINISTERED: 85/09/17 ___----__----------__-_.- ' EXAMINER: E;ANAVITCHel./LANGE,D APPLICANT: ___
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INSTRUCTIONE TO APPLICANT:
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Use separate paper for the answers. Write answe's on one side only.
~Stople question sheet on top of the answer sheets. Points for etch qr;estion are. indicated in parentheses after the question. The passing gr_ade requires at lestt 70% in each category and a final-grade of at _ least 80%. Examination papers sill be picked up six (6) hours after
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the enemination starts.
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%.0F CATEGORY % OF' AFPLICANT*3 'CATEG0kY UALUE TOTAL SCOPE VALUE CATEGORY
________ ______ ______-__-_ -_______ ________________-__-----__________-
- 1 __ _ 1 ___________ .._______
1. PRINCIPLES OF NUCLEAR F0WER PLANT OFERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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- _________-- ________ 2. PLANT DESIGN INCLUDING SAFETY AND ENERGENCY SYSTEnE.
- 1 __ _ 1 ___________ ________ 3 INSTRUMENTS AND CONTROLS 25.00 ^5.00 PR'CEDURES - NORMAL, ABNORMAL,
.______-- __'____ _________-_ ________ 4 O EMERGENCY AND RADIOLOGICAL CONTROL '100.00 100.00 TOTALE
________ ______ ___________ ________ FINAL GRADE _________________% All work done on this examination is my own. I have neither given nor received eid.
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____________________________________________ GUESTION 1.01- (2.50) a. Why-is conductivity measured on'a continuous basis.in the reactor coolant system, and why is it' maintained within specified limits? ( 1. 0 ')
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b. Will conductivity i~ncrease, decrease, or remain unchanged ift 1.-Chloride concentration increases.
2. Ionic impurities _ increase 3.:pH_ decreases from 7.0. (0.5 ea) GUESTIGN. 1.02 (1.50) A variable speed centrifugal pump.is running with its drive motor at 1800 rpm. The initial flow rate is 1000 spm, total head is 110 feet,
.and work input is 500 hp.
The flow rate is then changed to 1200-spm. Determine: a. The new drive motor speed (0.5).
b. The new total head (0.5) c. The new. work input (0.5) GUESTION 1.03 (2.50)
' Consider conducting a start UP at SNPS shortly after a scram from-full power, witn the moderator still near operating temper atur e and pressure. A reduced rod worth procedure is used, which~has rods in the center of the core withdrawn before edge r ods to ' shift' the neutron f i v >. to the outside area of the core.
Br a ef ly . enpl a a rs why this is a good idea. (2'.51
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GUESTION 1.04 (2.50) Water enters a centrifugal pump at 300 deg F and a pressure of 60 psis.
Based'on the available NFSH , oo you expect cavitation to occur? (I.5) 5HOW ALL WORK (xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)
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1. . PRINCIPLES OF-NUCLEAR POWER PLANT OPERATION, PAGE 3-
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____________________________________________ GUESTION- 1.05 (2.50) a. In terms of reactor power leve.1 control, explain why Shoreham's temperature coefficient is negative instead of positive. (1.0) b ~. How will'the Moderator Temperature-Coefficient change (ie., h0RE or LESS negative, . or no effect) for the following: 1.. SBLC is initiated 2. moderator. temperature increases 3. as the core ages. (0.5 ea) - GUESTION 1. 0 o - (2.00)~ With all other parameters held constante.If Recirculation Flow increases, how _ will the'followir!g parameters respond. initially?
(Increase, Decrease, or remain. constant.) (0.5'ea)
a. Thermal neutron population b. Thermal diffusion length c. Void Fraction d. Feactor Water Level 0UESTION' 1.07 - ( 1. 5 0 .- _uhy does core flow i r.i t i a l l y increase to approximately'110% of normal following a scram from full power? .(1.5) GUESTION '1.08 (2.50) a. Approximately what-percentage of neutrons produced from the fission
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of U-235 are born delayed? (0.5) b. How'does'this percentage change over core life? Why? (1.0) c. How'do the delayed neutrons affect the control of the reactor? (1.0)
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(*xxx*' CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx)
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1. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4
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-, -GUESTION 1.09' (1.50) :For each section marked on the graph below, state the reason for, the change in Shutdown. Margin (SDH) as core age increases. (1.5)
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SDH I I I I l___________________________________ Core Age GUESTION 1.10 (2.50) a. Fct a tubular heat enchanger', with all other parameters held constante.
would each of the'following changes (considered separately), INCREASE.
DECREASE, or have NO EFFECT-on the heat transfer rate? (0.5 ea) 1. Tube failure 2. An increase in1 flow velocity 3.. Fluid phase change with2n the heat exchanger b. State two reasons for a reduction in heat transfer rate when corrosion products build up on the tube s u r f a c e s .' (1.0) GUESTION 1.11 (2,50)
'The reactor.has been operating at 1007. Power for one month when a scram-occurs in which several contiol rods FAIL-10 FULLY ItJSERT. Enough rots DO insert to bring the reactor subtritical at the time of shutdown. If reactor. moderator teniper atur e is raaintained CONSTANT, and control rods are NOT. moved, about HOW LOUC will the cperater have to wait before he can be reasonably sure tnat the reactor will remain suberitical.?
EXPLAIN. (2.5) GUESTION 1.12 ~(1.001 Although steam is known to be a poorer heat conductor than water, NUCLEATE BOILING is a BETTER heat transfer mechanism than SINGLE-PHASE CONVECTION. Explain this apparent contradiction. (1.0) .
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 5 _______________________________________________________ t
._ .Th 3 '0UESTION- 2.01 (1.50)
Figure 2 is a' flow , diagram'of the HPCI system. .Upon auto initiation, state:-the final. position (OPEN'OR CLOSED) of the following valves.
a. MOV-047 Inboard Steam Isolation Bypass Valve (0.3 ea) b..MOV-042 Outboard Steam Isolation Valve c. MOV-039 Lube Oil Cooling Water Valve d. ADV-081fSteam Line Drain Valve o. LCV-095 Cond Pump Discharge to RW Level Control' GUESTION 2.02 (2.00) a. The RCIC system is in standby when the minimum flow valve (h0V-036) inadver tently opens. What would be the consequences of this situation if'it went'unnoticeo by the operators? (1.0) b. WhyLshould operation of tne RCIC turbine at speeds of less than 12:00 rpm. be avoided.? ( 1. C ) : GUESTION 2.03 (3.00)
/ '/ f a. How is feedwater.. flow controlled during i. ,-
1. Normal Operations (between 10 and 100*. power) ? (0.5) 2. Startop of the reactor (between 0 and 10% power) 7 (0.5) b. List thiee indications you have in the control room that can be used to verify proper feed pus.p operation. (1.0) c. .l. . When is-steam blanLe' ting of the hSR, (moisture separator reheater), necessary ' ~(0.50) 2. Why is this done? (0.50)
(***** CATECORY 02 CONTINUED ON NEXT PAGE *****)
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2. PLANT. DESIGN PAGE 6 _________________' INCLUDING SAFETY.AND EMERGENCY SYSTEMS ______________________________________ GUESTION 2.04 (3.00) Ccncerning the ADS System
.o.;What11s the: purpose of the SRV discharge line vacuum breakers to 1 g/C the torus?- (1.0)
b.1What is thefpurpose of the interlock that pr' events ADS initiation unless a-low pressure ECCS pump is running? _
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(1.0)
c.~If the automatic ADS. initiation-fails uhen required, it must be manually initiated.
1. What should you specifically verify or_do before depressing the four armed push button switches S15 A throvsh Do ( 0 . 5 ). 2.-After' depressing'the armed. push buttons, will the 105 second
~ timer-still have to timeJout before-the valves will open? . Explain (0.5)-
GUESTION. 2.05 (1.00) L/ Following an' accident in which the core geometry is-suspected to be _ damaged, state the best way to . de ter mine if co7e geomett/ Is still' intact.
(1.0) GUESTION- 2 06 (2.00)
' State the, specific design function provided by each'of the following-reactor safeovards, and briefly discuss.how it achieves.its function , ( 2. 0 )~
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a.. Main Steam Line Flow -Restrictors b. Containment Inerted bv Nitrogen
.00ESTION 2.07 '(3.00).
0.: What; signals will automatically cause the Emergency Diesel' Generators tc INITIATE ? (1.5) b. What signals will au atically cause the EDGs to TRIP if they.
are operating under NO Emergency ccnditions ? (1.5)
-(NOTE: for both a an be include set points where applicable.)
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2. PLANT DESIGN' INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 _______________________________________________________
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QUESTION 2.08 (1.50)
.0ther than.a LPCI initiation, what three (3) automatic actions will cccur by depressing the LPCI hanual initiation pushbuttons ?
(1.50)
QUESTION 2.09 (3.00) a. List four_'(4) Nuclear Safety Related system loads that are cooled
~by RBCLCW. (1.0)
Eb. What is an RECLCW ' system split', and when does it occur? (.75) c.. List five (5) station loads that are isolated from cooling water due to an RBCLCW . system split. (1.25) GUESTION 2.10 (1.00i State the normal operating value for the following parameters in the
. Control Rod Drive System. (Assume no rod motion) (1.0)
a..CRD Cooling Flow b. CRD-' Cooling Water dP c..-CRD Drive Water cP d. CRD' Drive Water Flow e..CRD System Flow-GUESTION 2.11- (2.00) With regard to the Remote Shutdcwn Panel (RSP), 1C61*PNL-001: a. HOW'would you, as a control room operatore know if a transfer switch.was taken to the essergency position on.the.RSP ( 2 inci-cations r equired)? (1.0) 6. List FIVE (5) , systems that h' ave selec'ted components which may be operated ficm the Remote Shutdown Panel (1.0)
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2.- PLANT DESIGN INCLUDING SAFETY:AND EMERGENCY SYSTEMS PAGE 8
-_______________________________________________________ 'GUESTION 2.12' (2.00)-
Fct eachoof_the following statements regarding the High Pressure-Coolant 1 Injection System (HPCI), indicate whether the statement is-TRUE'or FALSE, _and EXPLAIN your answer, ai In:the event Low-Pemp_Suetion Pressure is sensed during-HPCI' system operation,.the turbine will trip, and the signal must be manually reset before the turbine will restart, if initiation.
' signals are still present. (1.0) b. If the HPCI turbine trips due to an overspeed condition, it-will restart when the speed coasts-downito between 3000 and.4000 RPM. (1.0)
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3.. INSTRUMENTS.AND CONTROLS PAGE 9 .____________________________ QUESTION 3.01 (2.00) What are-the purposes.of the following two precautions from Sp 23.609.01,
*LRod Sequence Control System' ?
c. *Do not bypass any control rod.in the RSCS having a failed ' FULL IN' or ' FULL OUT' limit switch unless the actual rod position.is known'(1.0) b. "Whil'e in the transition zone.(LPSP to LPAP), the operator must verify the alignment of.the rods within each group' (1.0) GUESTION 3.02 (1.50) Following a valid RCIC initiation, you receive a subsequent'RCIC turbine trip from high reactor watet level. Will the.RCIC autc start, witn no cperator action, if the water level'again decreases to the low level set point? E >:p l a i n (1 5) .0UESTION 3.03 (2.00) The-Reactor Water Cleanup System suction line is provided with auto-acting isolatior. .alves C h0V 33 & 34 ). Tnese valve.s auto close on leak detection. List the auto close signals and their setpoints. -(2.0) GUESTION 3.04 (2.00)
-When will'the following Reactor Protection System scram signals be bypassed ?-
c. MSIV closure b. TSV Fast Closure c. Mode Switch to Shutdown d. SDV High Level GUESTION 3.05 (3.00) With respect to the 120 VAC Uninterruptible Power Supply (UPS) n. What automatically happens within the system-in the event of an inverter failure? (1.0) b. If the vital bus inverter (1NV - 01) is lost, how will the hain
' Generator exciter switchboards be affected? (2.0) . (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx)
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3. INSTRUMENTS-AND PAGE 10 ____________________' CONTROLS ________ m JOUESTION 3.06- (1.50)' Upon a loss of Feedwater Control Signal from the M/A transfer' station to the Reactor Feed Pump Turbine control logic:
.o. WhatJwill be th.e status of the speed of the affected RFPT? (0.5)
b.-Will.the' operator be able to manually vary the speed of the .
:affected RFFT? If no, why not. If yes, how. (1.0) -GUESTION -3.07 (3.00)
Three'runbacks are provided in the Load Control Unit of the EHC System.
List,each of the three runbacks, and state to what level each runback
~will reduce 1 cad. (3.0) . QUESTION 3.08- -(3.00)
With regard to the Reactor Recirculation System.
a.-Why does the Recire Pump discharge valve have an auto close
. feature?_ (Provide approximate. actuating setpoint(sii. _
(1.5)
b. .RPT breakers' serve ~ three (3) functions. List the functions and setpoints which will actuate them. (1.5)
. QUESTION 3.09 (I.00)
Consider the Rod Block honitor ( R E.h )
' s. List the conditions which will cause a RBM to be bypassed. (1.0)
6.-A rod block will result f om a 'railure To Nul'l'. Explain what
' Failure to Null' means. (1.0) (xxxxx CATEGORY 03 CONTINUED ON NEXT.PAGE x***m)
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---------------------------- -QUESTION 3.10- (2.50) .The'following plant conditions' exist: . Reactor-Vessel Level is -140' (confirmed, both channels) and decreasing.
- The RHR System-is operating with a pump discharge pressure of
'>125 psis in system A and B.
- Reactor Vessel Pressure is 1000 psis and steady.
-- All ADS INIT INHIBIT ' switches in NORMAL position.
- ADS Timers have timed out (both A & B).
- On the Automatic Blowdown Relay. Panels ( 1H11xPNL-428 and 1H11*PNL-631), the white' indicating lights above each of the SRVs'show - ONE light DUT and DNE light LIT BRIGHTLY, for all valves.
a. What has h'appened 1 EXPLAIN. (1.5) b' . What'are your actions in this situation? (1.0) OUESTION 3.11 (1.00)
-What are the five (5) N0tJ-SAFETY RELATED LOADS that can be reenergized ,~
with a LOCA :23r=1 present and the r espective EDG Breaker closed for DIVISION I? . (1.0) GUESTION .3.12- (1.50) The reactor is operating at 100% power when a small area break occurs.
HPCI is out of ser.vice. RCIC fails to start automatically and mannot be started manually.. All signals are.present and valid for
'suto blowdown system actor. tion except the timer has NOT timed out.
c. RCIC is:now started and water level is raised tc above the low-low setooint. What effect does this have on auto blowdown initiation if the timer HAS-NOT timed out? If the timer HAS timed out? Briefly explain BOTH. (1.0)' b. If.the 125 UDC Electrical Distribution System is lost
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will auto blowdown accut- % EXFLAIN (0.5).
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. 4.- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12
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____________________ QUESTION 4.01 .(2.00) E:: plain ~ whether either of the f ollowing situations describes a case .where 10CFR20 Limits for individuals working in restricted areas have been violated.-(Both men have updated NRC Form 4s on file)
. c. A skin dose of- 7000 mrem is received by a 44 year old male over a full quarter. _ His lifetim'e dose is 55 rem. (1.0)
b. A-whole. body dose of 1500 mrem is_ received by a 44 year old male over a three (3) month period.. His lifetime dose is 22 rem. (1.0) QUESTION 4.02 (2.50) Concerning SP 29.012.01, ' Loss cf Condenser Vacuum': a. List'the four.(4~) Automatic Actions which occur on a loss of condenser vacuum. (Setpoints NOT. required) -(1.0) b. Step 4.9 of'this procedure instructs you to "haintain surveillance of Turbine Evilding airborne radiation levels'.-Why? (2 required) (1.0) c. Wh'y are the condenser vacuum breakers (IN11-MOV-013A E Bs opened if sealing steam is lost? (0.5) "GUESTION 4.03 (2.00) According to Procedure SF 20.004 01, 'Emer3ency Use of SLC'; s. WHO is responsible for determining if Standby Liquid Control initiation is necessary? (0.5) b. Name five (5) items which should be checked in the control room to verify that the SLC tank contents are being injected into the i vessel. (1.0) c. What is the purpose of the heat tracing that is provided on the SLC system piping? (0.5) OUESTION 4.04 (2.00) According to SP 23.116.01, ' Main and Auxiliary Steam', there are nine (9)
. indications in the control rcom that can be used to verify that an.
SRV is open. Name these nine indications. (2.0)
(***** CATEGORY 04 CONTINUED ON NEXT PACE *****)
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4. PROCEDURES - NORMAL,. ABNORMAL, EHERGENCY AND PAGE 13
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QUESTION _4.05 (3.00) List the Immediate. Operator Actions required for the Emergency Shutdown Procedure, SP 29.010.01. (3.0) GUESTION 4.06 (2.00) Cencerning Procedure SP 29.023.01, ' Level Control Emergency Procedure': 0.~ List-the entry conditions and setpoints for this procedure. _
(1.0) :b. List-the three items you should verify as occuring consistent with entry conditions, after you enter the procedure. (1.0)
OUESTION 4.07 (2.50) a. Fil'1 in the blants in the following:
"According to SP 29.023.02, 'Cooldown Emergency Procedure *, you .
are cautioned NOT to secure or place an ECC system in the manual mode unless misoperation in automatic is confirmed by ______(1) _________ or _________ (2> __________.' (1.Si Note: h0RE THAN ONE WORD I5 REQUIRED 4IN EACH BLANK
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b. In your own words, define ' t.;i s o p e r a t i o n ' as it is used above. (0,5)
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GUESTION 4.08 (2.00) According to Sp 23.604.01, 'APRn 5ysten,': a.-When is an AFRM ch:3nnel defined as inoperable? (1.0) b. During shutdown, when should the. operator switch the IRH/APRM recorder switches to the IRd position? (1.0)
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. 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PACE 14.
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____________________ GUESTION 4.09 -(3.00)
.o. List the five (5)-entry-conditions and setpoints for SF 27.023.03,; ' Containment Control Emergency Procedure *. (1.0)
b. According to ~ the procedure, what three ictions.should you take if drywell' temperature' approaches 296' degrees F? (1.0) c. 1. If Suppression Pool temperature is 100 riegreet .F, and RPV pressure is 600 psis, have you exceeded th= heat capacity limit of the ATTACHED figure 1? (0.5)
~2.'Under these ecnditions, would the procedure direct you to open all ADS valves? (YES or NO) (0.5)
GUESTION 4.10 (2.00) According to SP. 23.116.01, ' Main and Auxiliary Steam *: a. What should be your immediate actions if you h' ave an inadvertant MSIV closure and the Mode Switch is NOT in 'RUN'? (1.5) b. According to the procedure, waht is the limit on diff er ential pressure across the nSIVs prior to opening them? (0.5)
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GUESTION 4.11 ( 2.00 ). a. SP'22.001.01, 'Startup-Cold ShutdownEto 20% Power *, states-that both recirulation pumps must be running during startup. Why? (1.0) 6. Name the three locations at which reactor ccolant system temperature must.be maintained within 5 degrees F, once heatup power is schieved. (1.0)
(xx*** END OF CATEGORY 04 *****) (xxxxassxxsmsr END GF EXAMINATION ***************)
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l: I___ _ _ _ __ _ ____ _ _ ___ THERH0 DYNAMICS, HEFT- TRANSFER AND FLUID FLOW ____________________________________________ l l ANSWERS -- SHOREHAM -05/09/17-BANAVITCH,L./LAN'GE, i
~#
ANSWER 1.01- (2.50)
-( o ) Changes in conductivity indicate _ abnormal conditions. When conductivity is within limits, _the pH, chlorides, and other impurities must also be within their' limits. (0.5)
Exceeding any of these limits could cause higher corrosion rates.which would jeopardice reactor components. (0.5)
(b) 1. Increase (0.5 ea)
2.' Increase 3.-Increase REFERENCE (a): SNPS TS Bases 3/4 4.~4 Chemistry-(b) Chemistry _ Final Exam. question 4. SNPS v/ ANSWER 1.02 (1.50)
(a) Because flow rate varies directly with driving. speed speed' = speed (flow'/ flow) = 1800 (1200/1000) rpm = 2160 rpm (b) (N*N) is proportionel to the total head:
Head' = Head (speed'/ speed)**2
= 110 (2160/1800)**2 psi = 158.4 psi (c) .Nx*3 is proportional to the work input P' = P (speed'/ speed)**3 = 500 (2160/1500)**3 hp = 844 hp (0.5 ea)
REFERENCE SNPS Fluid Mechanics Module, page 3-74
F .
- . . .
1 1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 16
--- isEER55isisiCs- AEEi isissFEE As5 FE0i5 FE5-
____________________________________________ ANSWERS -- SHOREHAh -85/09/17-DANAVITCH,L./LANGE, L/ANSWER 1.03 (2.50) Prior to.the scram, most of the neutron flux is concentrated in the conter and bottom portion of the core. Xenon will be most predominant in these areas'because its concentration is a function.of power level. Xenon
' will therefore supress the flu:: in these areas, and the' top and edges' of the core will be,the most reactive areas.
This'is. opposite of the normal casei if edge rods were pulled first, excessively high incremental-rnd worths would be created in. areas which' usually have low worth. Pulling rods from the center first increases the flux there'and therefore lowers-the edge rod worths. (2.5) REFERENCE SNPS Reactor Phystes Mcdule, page 7-207
- ANSWER 1.04 (2.50)
.V
[-Saturaticn pressure from the-steam tables: 67.01 lbf/in2
= 9649.4 lbf/ft2 (0.5)
rho = 1/0.0175 = 57.14 lbf/ft3 (0.5)
- ; P1 = (60 + 14.7) = 74.7 psia = 10756.8 lbf/ft2 (0.5)
l" NFSH = (P1 - Ps)/ rho = (1075c.8 - c649.4) lbf/ft2 / 57.14 lbf/ft3 (0.5)-
= 19.38 ft Cavitation will NOT occur because the NP5h is a large. positive number.,
(0.5)
REFERENCE
$NPS, Fluid Mechanics hadule page 3-84 l
l
- .
( - . . . .
1. - PRINCIPLES 0F NUCLEAR: POWER PLANT OPERATION, PAGE 17
~~~~IUERR55isARICs? sEEi TEEssFEE As5 FL5i5 FtBE
____________________________________________
: ANSWERS-- SHOREHAM. -85/09/17-BANAVITCHsL./LANGE,
_ 'ANSWERf 1.05 -(2.50)
/,(c) If the. moderator' temperature increases, the negative temperature '(1 0)
coefficient'will cause~a negative reactivity insertion which will decrease the power ~ level. If the coefficient were positive,'an increase-in moderator: temper ature would cause a positive reactivity insertion which would increase the power and therefore temperature level. This series of increases would continue unless there.were outside intervent2on.
'(b) 1'. LESS negative 2. MORE nesstive 3. LESS negative. (0 5 ea) REFERENCE SNPS Reactor Physics Hodule, page 7 175 (a)_.
(b) Reactor-Physics-Final Exam, question 23'
.
1.
' ANSWER 1.06- (2.00)
'/
c. Increase (0.5 ea)
;b. Decrease-c. Decrease d. Decrease ' REFERENCE SNPS Fluid Mechanics anc Reactor Physics dadules ANSWER 1.07 (1.50) / Flow' increases beca:Jse there is less twc phase flow resistance'due to tne . power' reduction-after the scram which reduces the void fraction. (1.5)
REFERENCE , SNPS Fluid Mechanics Hocule, page 3-61
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
. . . .-
I- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 18
~ ~ ~~~~YUEEEU6 EE55C5I"UEET TREU5EER EU6~EEU56~EL6U
____________________________________________
' ANSWERS -- SHOREHAh' -85/07/17-DANAVITCH,L./LANGE.
- *
1, ANSWER (a) 0.641%
-1.08 (2.50)
,
#,[hs (0.5)
r.(b) DECREASES (0.5), due to Plutonium pr duction which has a lower
)j delayed neutron fraction than U-235. 1 b r'f) @ (l #) (c) Delayed neutrons increase the core everage generation time which allows-better control of the reactor. (1.0)
REFERENCE
.SNFS Reactor Physics Module, Lesson 15: Transient Reactor Responst . ANSWER 1.09 (1.50) ./ .a. Incr. eases initially due to fission product poisen buildup. (0.5ea)
b. Decreases due to poison burnout.rcte greater than fuel depletion rate.
c. Increases due +.o fuel depiction rate greater than poison burnout rate.
REFERENCE-SHPS Reactor Physics Module. page 7-222,223 t/ANShER 1.10 -(2.50)
(a) 1. decrease (0.5 ea)
2. increase 3. increase (b) 1. Water flow is restricted; mass flow rate drops.
2. The.lsver acts as an insulator to heat ttensfer. (0.5 ea;
::EFER ENCE SNPS Heat Transfer and Ther.taodynamics Module, Lesson 6: Heat Enchangers , ANSWER 1.11 (2.50) -70 hours (1.0)
It will take approx. 70 hours for the Xenon to peak and then decay'after the scram. If the-positive reactivity inserted by the decay of Xenon is less.than the shutdc reactivity due to rods, then the reactor will remain suberitical. (1 1 _
y , _ - . - _
. .
A
.
1.- PRINCIPLES OF NdCLEAR.POWERLPLANT OPERATION,. FACE 19
~~~~~YHEEbU6 EEI55I~5E3Y'IRIU5EEE~EU6~ELUi5 FL50
____________________________________________ ANSWERS -- SHOREHAM -85/09/17-BANAVITCH,L./ LANCE,
--
-;
'
' REFERENCE. .
. Reactor _Theorp - Xenon Transients j/ ANSWER 1.12' (1.00)
- The. bubbles _ caused-by nucleate boiling serve to agitat'e the stagnent
~ fluid film _next to the surface, thus improving thermal conductivity.CO.53
' Also, each bubble, as it leaves the surface, carries off more energy
-
than.is possible by natural convection.CO.53
. REFERENCE 5NFS Heat Transfer and$ Thermodynamics Module - Lessons 5, 7, and 8
-
' Student Objective.91, Lesson 5 Student' Objective ti, Lessen 7 Student Objective 41,-Lesson ~B ,
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- 2. PLANT DESIGN--INCLUDINGISAFETY AND' EMERGENCY. SYSTEMS PAGE' -26'
. -------------------------------------------------------
JANSWER3 ?-- 5HOREH Ari '- 85/09/17-SAN AVITCH e'L . /L ANCE , ANSWER 2.01- (1450)- i 1. G q.e se d L e s ' A '
-27 0 pen 3
*
-3.cOpen. ,
4.'. Closed .
* , '5. Closed ( 0. 3. S a )-
REFER'ENCE'
' 1SP 23.202.01-Rev 9 page 6 4
, w JANSWER 2.'02 t2.00)
(a) A drain path would be established from.the CST.to the Suppression-Pool... -(1.0)'
(br 1. Water hammer n,a y result in the exhsust line. (0.5)
.2. .At lower speeos. adequste cooling and lubriestion of the, turbine
.may net be ensured. (0.5)
51.[ick a( A 4 [ (d ( u e REFERENCE a Handcut 119 Se: tion 3.5.2.1 b SP_23.117.01;page 4 y c ,e! ,, 2 ^
- ANSWER : 2. 03 - -(3.00) ,
,.
. 'y ,, ;
(a) 1. By contrelling the speed of the feed * pumps' #.. -
(0.5)
.2."By two star tup - f eed water level c o r.t r o l valves (0.5)
'(b) Verify: ficw incteation, f,orbine speed, and discharge pressure ^ i (0.33.ea)
(c) 1 ~.- Elenketing-steam is. supplied to the~inside of the rehester. tubes
'
f r o c. the.' Aux E iler durIng shutdown. ) (0.75) 2. L Thi s is dorie . to' r educe corrosion of the heater tubes. '(0.75)'
-RE.ERENCE . . ;s'- SP23.109.03.Rev 5 page 2 , , ,
b- lesson plan R109, page 7 . -
,
*1'
.c- lesson plan 110, page 5 Lestning Objective C. \
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'2.' PLANTEDESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 21 .--------..----------------------------------------------.
JANSWERS --~SHOREHAh -85/09/17-BANAVITCHrL./LANGE,
'ANSWERI 2 04' (3.00)
.
(o) To prevent-) water.from beins drawn back into:.the discharge pipes'due-
~
-
'to steamJcondensing in-the~ pipes after an SRV actuation (1.0)-
..(b) .To' ensure a method.is available tc pump water into the vessel once the.
Iower: pressures are reached and before vessel' level is further decr eased by ADF actuation. (1.0)
-(c)"1.-Verify started or start at least two RHR pumps or one (1) CS pump.
- 2.-NO-'the timer'is bypassed by the armed pushbutton's. (0.5 each) REFERENCE.
.HL 201*:a .pa3c- 4,
-
b--page 13, c page 6. .q
' ANSWER- 2.05 , (1.00)
Core geometry.is still intact If the SRMs and the IRMs can inserted and withdrawn-successfully. (1.0) H
.REFERENCEE SNPS 'Mitigatir.g . Cor e Damsge Enam, Class 5 ANSWEFI 2.06' (2.00)
(a) 1.rTo Itait: steam flow to a maximum of 200% of rated steam line f1cu-f o111owin3- an MSL br eal. . . (0.5)~
.y2. By liraitin's the dP ,acecss the ' steam dryer and other vessel' internals as well;as tne mass. flow rcte from-the vessele (0.5)
(b) 1. To prevent reaching a flammable or explosive drywell atmosphere followingia LOCA.- .
.(0.5)-
2 By keepingcthe oxygen concentration in:the drywell below .the Tech Spec' limit of 4% by volume. (0.5) REFERENCE 6- HL;116 page 7 b- HL 654-page 7 TS Bases Section 3/4 6.6 l
.
I
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _________ _ _ _ _ , . . . . . . . ___ _ . - - .__ t L -
+ -
}. .
*
.. ! PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS,
-
2. PAGE 22 __'_____________________________________________'_______
-
r h ANSWERS -- SHOREHAM -85/09/17-BANAVITCH,L./LANGE, V ANSWER 2.07 (3.00) c/-
- (o) Undervoltage on respective 4160v' emergency. bus
-
(1.5)
High Drywell Pressure (+1.69 psig) Low Reactor Water Level (-132.5') [ (b) Overcurrent (neutral ground) Generator phase-differential-
~) (1.5) )
Overspeed (517 rpm increasina) ! Mode selector switch in LOCKOUT position { "*" L .hanual push button (Local or from Control Room' 1
-
Manual push button on DC skid 'STOP' depressed pf , Tc e eg - REFERENCE ,1 d ^
,,3 x .f j SP 23 307 01 Rev 10 page 13 D -
. !
. ANSWER 2.08 (1.50)
l' t (a)-1. EDG start i 2.' Emergency Pos Loading Progrem - 3r 3. Feed water check valve I s o l a t i c r. (0.99 ea) REFERENCE SP 23.204.01 Rev 2, page 2 ANEWER 2.09 (3.00) f-(a) RHR pump seal coolers (0.25 ea)-
; Spent fuel pool cooling water H:
Reactor Recirc Fump: Seal Cooler, Motcr Winding. Bearing Coolers (Four needed for full cr edit. )
(b) During:a LOCA signal cr lo lo head tank level. the system is auto
' separated into two independent loops. -
(0.75)
_
(c) During an accidents the ncn-nuclest safety related loacs are isolated by MOVs:
Reactor Recire Pump hG Set Oil Cooler RWCU Nonregenerative HX RWCU Pump Coclers Drywell Equipment Drain Cooler CRD Fump* Bearing Cooler and Gear 011 Cooler Drywell coolers (0.25 ea) (Five needed forfull credit)
. [ K 1/ak A~p4 pad avv L:A s q e~t>
_ _ - - - _ _ _
, _ . _ _ . . _ _ _ _ - _ _ _- -_ __
. . . . .
2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 _______________________________________________________ LANSWERS -- SHOREHAM -85/09/17-BANAVITCH,L./LANGE, REFERENCE orb- Lesson-Plan 118, page 5/'c- LP 118, page 6 and Learning Objective 7.3 ANSWER 2.10' (1.00)
1. 30-4X spm 2. 20 psid 3. 260 psid 4. O spm-5. 30-46 spm (0,2 ea)- REFERENCE-44L 106 Learn'ing Objective F, psge 2 of Student Handout Supplement j i ANSWER' 2.11 (2.00)
/a . -Alarm in the'contro1 roomE3.53-Loss of indicating-lights in the control room for the affected ccmponentCO.51 . (1.0) -ADS m. 52V- ' MA -RDCLCW d*S ' D'
b.
-Recirc * g* ' Compre-ssed Air System-RHR-
-RCIC- -Fuel Pool Cooling-Service Water. System C5 required at 0 2 each] (1 0; REFERENCE SHFS Procecure-5P'27.022.01, Rev 4 ANpWEh 2.12 (2.00)
3.-False CO.53 - Once the low suction pressure signal is clear, the turbine ~will -auto r estar t it.the initiaticn signals are-still present.CO.53-b. True CO.53 - The oil pressure will be restored when the turbine coasts down, thereby causing the stop valve to open.EO.53 REFERENCE HPCI SP-23.202.01
-. .. .. .
..
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-- , . . - .
e
.
PAGE 24 3.___ INSTRUMENTS AND CONTROLS
.
__ _______________________
' ANSWERS -- SHOREHAM -85/09/17-BANAVITCH,L./LANGE,.
ANSWER -3 01 (2.00)
c) c 'The. purpose of this precaution is to prevent crossing-tips and
: ensuring reactor engineering is aware of a possible-rod in an 'p' ' abnormal position for the selected sequence. (1.0) -dIue/k Yhh i Lb. This'must be accomplished before going below'the LPSP to avoid ' generating an insertion or withdrawal block. (1.0)
REFERENCE a. Lesson; Plan!607/609,-Learning Objective D.
b. SP 23.609.01, page o ANSWERI 3 02 (1.5C)
(0.5)-
1The operator must reset the trip and throttle ' valve (MOV - 44) by closing and reopening it.- (1.0)
: REFERENCE- * " A ' *- ~
J-SP 23.119.01 ANSWEP 3.'03- (2.00)
"' 7.5 : D-
_RWCU Area Hi Tempersture ( 4+4 degrees F)
-RWCU-High.Ficw'(44 s p e. ) -
Low Low Resctor' Water L'e vel (-3S') g7s f'5
' ~Matn Steam TunnelJPiping Area High Ten.perature ( Ett deg r e e s F ) (0.25 for p a r s a.e t e r , 0.25 for setpoint)
REFERENCE Student Handout 709, ptsc I L3.04 (2.00) fNSWER a. Mode switch in other than ' RUll '
.
b. Power < 25% c. 10-seconds after Mode Switch placed in S/D d. By the bypass switch or Mode Switch ii. 5/D or Refuel
. . ., .
b
. .
S
:
3. INSTRUMENTS AND CONTROLS PAGE 25 ____________________________.
. ANSWERS.-- SHOREHAM -85/09/17-BANAVITCHrL./LANGE,
. REFERENCE Student ~. Handout ~ 1 3'2/611, Learning Objective 7.5 ' ANSWER -
3.05 (3.00) o. The : static' transf er switch.will auto transfer the.AC Vital bus loads
.
to alternate AC source (1.0) ~ b. . Generator? amps, volts,-and temperature Jindications are lost ( 1. 0 ) . Generator' field ground detection trip and it's associated slarm are disabled. (1.0) Cg Q gQ g d . e
> REFERENCE Lesson Plan 313, pages 5 i 9, Learning Objective D ANEWER 3.06 (1.50)
a. Speed is lock.ed at the last requested speed (0.5) b..RFPT speed can be r educed f ron the RFPT'EHC panel, but 2t cannot
-
be increased.Q1.0) Q g & gM at fl5J
' REFERENCE 5tudent Hancout 656, page 9 15WER 3.07 (3.00)
1. : Less of Stater Cooling - 25*4 laad 2. SYNC speed not selected - rero 3. Power' Load Unbalanced - ero REFERENCE HL 657, EHC, oages.1-5
.. ' ' . . , . . :3. -INSTRUMENTS'AND CONTROLS PAGE 26 .____________________________ : ANSWERS - -SHOREHAM -55/09/17-BANAVITCH,L./LANGE, , ANSWER 3.08 ~(3.00)
a. The auto close feature is provided to close the discharge VLV in_the eventsof a LOCA and reactor pressure is <309 psig to
' ensure that.if the break is between the suction and discharge zVLV's, to allow LPCI to inject'to the reactor vessel and not flow back and out'the break. (1.5) .b. Functions * - . 20 1. Open on ATHS conditions of high reactor pressure of 1187 psig.
or low lou Rx level -30' (0.5)
'2. Open on-turbine trip when at greater than 25% power. (0 5)
3. Provide redundant fault protection for primary containment
' electrical penetrations. (0.5)
REFERENCE N'
'O " 76 o' . , . . . !,.
Recirculation Spstem ' f ANSWER 3.09- (2.00) a. 1.' Manual operstion of bypass switch (0.33) 2. Reference APRH downscale (<30%. power) (0.33) 3. Edge rod selected. (0.33)E b. .A failure to null results i f. f l u:: is so severly depressed around the selected rod that even with the maximum gain enange-the REM cutput cannct be mace greater than or- aqual to the reference APRM. -(1.0) . REFERENCE 4 ,. rnh Te %eMa led fd* 5 O "1 f" - LP 606
<
I
~~
.. . . , :..
3.- INSTRUMENTS AND CONTROLS PAGE 27 ____________________________ ANSWERS - SHOREHAh -85/Oc/17-BANAVITChrL./LANGE, ANSWER 3.10 (2.50) a. All parameters are normal for an ADS intiation, but the ADS system has not initiated. The status of the white-indicating lights on the Automatic Blowdown Panels indicates that only one of the 2 channels in each of.the 2 logic systems energized. Since the
' logic systems are 2 out of 2 oneer the valves did not-open. (1.5).
b.-Attempt to mapually initiate the system by(verifying that at leas.t two RHR/ or . one-. CS pump is running, and try to open the ADS valves using the. individual control switches. -(1.0) REFERENCE 1.' Automatic Depressurication/ Safety Relief.-Valve System Student Handout 4 HL-201: Student Objectives =4 3& 10 2. SP 23.201.01
. ANSWER 3.11 (1.00-) -1. RFF 'A' Ivbe oli pump 2. RFF 'A' turning. gear 3. Bearing Lift Pumps A &C 4.-24V Batterv Charger 5. Emeroency Lighting Distr ibution Panr1 - (, . A d.czti PQ.. ( 5'at 0 2 pts each) . : REFERENCE Emergency Electrical Distribution Student ~ Handout Supplement Ob,iective No.-7.13 . /uf Zq . c t S, cl w ,II,fy 3 dyt 46 ANSWER 3.12 (1.50) ~
a. - Auto' blowdown will not initiate if the timer has not timed out E0.2] because the timer will reset and. not restar t until level drops below the setpoint CO.33.
--If the timer has timed out, auto blowdown will be initiated E0.253 a r.d ' will continue until completion (or reset) CO.253 b.1ha.-[0.253 Power to'the DC solenoids and logic initiation circuitry is powered from 125 VDC. CO.253 d.u, Go A oiv. s 5t L4 [,.Ad D L,._ 4 -
. .
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3. INSTRUMENTS AND CONTROLS PAGE 28 ____________________________ ANSWERS -- SHOREHAM -85/09/17-BANAVITCHrL./ LANCE, REFERENCE ADS Student Handout-Ovjectives 3 & 6
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-ANSWERS -- SHOREHAM -85/09/17-BANAVITCH,L./LANGE, ANSWER .4.01 (2.00)
f 44 = N- 5(N-18) = 130 rem.
Naither case has exceeded the lifetime dose under 10CFR21 (1.0) a. 7000 mrem to the skin does not exceed the'7500 mrem limit /qtr. (0.5) b. The 1250 mrem /qtr-whole body limit has not been exceeded because he has-an updated NRC Form 4 (0.5) REFERENC'E 10 CFR 20, Section 20.101 ANSWER 4.02 (2.50)
( e. MSIV and MSL-drains isclate.
Turbine BPUs close er r'emain c1csed Main Turbine trips Reactor Feed Pump turbines trip (0.25 each) b. 1. If condenser sit renoval pumps are used'to help n.aintair, vacuvan their exhsust is not treated by gaseous r a'd w a s t e so -it must.be-monitored.- (t ,
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2. Condenser outgassin3 covId increase the turbine building rad levels.f c. To prevent excessise cooling of.the turbine shaft and seals as air is-drawn in at these locations.
REFERENCE 3P 29.012.01 4.03 (2.00) pSNSWER a. The Watch Engineer (0.5) b. - Squib valve loss of continivtv a l a r n.
- SLC-pump discharge pressure >' reactor pressure Both squib valve lights out
-~ Selected pump running light en - SLC Tank level deeressing --Reactor power decreasing ( 5 of 6 G 0.2 pts each)
c. to prevent precipitation and crystallization of the solution in the pump suction lines (0.5) __ -
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____________________ ANSWERS -- SHOREHAM -85/09/17-BANAVITCH,L./LANGE, REFERENCE SP 29.004.01 and Student Handout " Standby Liquid Control' ANSWER 4.04 (2.00) 1. SRV leakingfannunciator 2. Hi' temperature on SRV discharge tailpipe 3. Hi pressure'on tailpipe pressure indicator 4. Relief Valve Open annunciator
.U. Suppression Pool Temperature increase 6.-- Suppr e ssion Pool water level increase 7. FW Flcw > Steam Flow B. Decrease in Turbine Generator Load-9. Temporary increase in Reactor water level prior to SF/FF mismatch .(0.22 each) g g ,c.2f.7u Ut-N1 So l "' ,S F 's 'M - -REFERENCE N[ . A b S-SP.23.11o.01, page 10 ANSWER 4.05 (3.00) '/ ' 1. Place the Moce Switch to Shutdowr, 2. Verify's rapid flux decrease 3. Verify all rods inser ted 4.-If not - refer to Transient with Failure to Scram Procedure 5. Monitor reactor vessel level 6. Initiate level control. procedure if necessary Six-components to answe fron. 4 immediate actions - Each component .5 each . REFERENCE .
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. ANSWERS"-- SHOREHAM- -85/09/17-BANAVITCH,L./LANGE,-
ANSWER l4.06 (2.00)
'a. RPV water level less than 12.5 inches Drywell pressure >l1.69~psis An isolation condition which requires cr initiates a scram ( 0.33 each) ..b. Grcup Sw-RA- -isolations Auto initiation of-ECCS' systems D/Gs start-(0.33 each) 3, q . REFERENCE--
SP 2 ?'. 0 2 3 . 0 1 , page 1
' ANSWER 4.07 (2.50)
1. 0
;o.l. at-least.two independent indications @r5-) ~o.2. adequate ccre cooling is assurred by at least two independent-indications. (g b. If-it.does not meet the Technical.Specificaticn definition of Operability. CAF FOR-0THER ACCEPTADLE-ANSWER 5. (0.5)
REFERENCE, 'I N u.As . % C J J -s q
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SP129.023.02, page 2
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ANSWER' 4.08 (2.00) s. When there are less thar 2 LPRM inputs per level or less than 11.LPRM. inputs total to the APRM (1.0) 6. When the first APRd Downscale alarm light illuminates (1.0) REFERENCE SP 23.604.01, pages 2 and 4
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-ANSWERS -- SHOREHAM -85/09/17-EANAVITCH,L./ LANCE, ANSWER .4.09 (3.00) . .jf lc. Suppression Pool Temperature > 90 degrees F .Drywell-Temperature > 145. degrees F-Drywell Pressure > 1.69 psis Suppression Pool Level >'+6 inches Suppression Pool Level < -6-inches. (0.2"each)
Lb.-Shutdown Recite Pumps-Shutdown Drywell Fans Initiate'Dryuell Spray . (0.33.each) c.1. NO (0.5) c.2.'NO (0.5i REFERENCE SP 29.023.03', pages 1 and 5
. ,f A. N S W E R . 4.10 (2.00)= .a. Monitcr and maintain reactor pressure using centrol-cods tc adjust power as r equir ed (0.75)
Start RCIC to assist in contrciling reactor pressure and level-(0.75) b. 200 psid (O'.5) REFERENCE-SP-23.116.01', pcse-10 m S- p q . (, . f. I . 2 , #1.
ANSWER 4.11 (2.00)
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a. Natural circulation startups imp'ose e:< c e s s i v e stress on 'the- control red housing.- (1.0)
'b. Reactor Vessel Gottom D r a i r, 'Recirt Loops A and B Reactor Vessel Bottom Head (0.33 each)
F:EFERENCE SP 22.001.01, pages 5.and 7
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4-g TEST CROSS REFERENCE PAGE 1 QUESTION- VALUE - REFERENCE ________ ______- __________
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01.01' 2.50. BAJ0000733
#01 02 1.50 BAJ0000734 01.03 2.50- BAJ0000735 01.04 2.50' 'BAJ0000736 01.05: -
2.50 BAJ0000737
~01 06 2.00~ BAJ0000738 01~07 . -1.50 BAJ0000739- '01.08 2.50 BAJ0000740 i
1 01.09- ' 1 ~. 5 0 'BAJ0000741.
~01'.10 2.50 BAJ0000742-01.11- 2.50 BAJ0000771 01 12: 1.00 BAJ0000772 ______
.25.00 02.01 1.50- BAJ0000743 02'.02 2.00 BAJ0000744 02 03: 3.00 BAJ0000745 02.04- 3.00 BAJ0000746 02.05 1.00. -DAJ0000747' -02.06 2.00 BAJC000748 02.'07' 3.00. BAJ00007.49 02.08. '1.50 BAJ0000750 02.09 13.00 DAJ0000751 02.10 la00 BAJ0000752 ~02.11 -
2.'00 BAJ0000773 02.12 2.00 BAJ0000774 ______ , 25.00 03.01: 2.00- BAJ0000764 03.02 :1.50- BAJ0000765
,03.03 2.00 BAJ0000766 '03.04 2.00 BAJ0000767 03.05 3.00 BAJ000076B 03.'06 1.50' BAJ0000769 .03.07 3.00 BAJ0000770 03.08 3.00 BAJ0000775-03.09 2.00 BAJ0000776-03.101 2.50 BAJ0000777 :03.11 1.00 BAJ0000779 ~03.12 1.50 BAJ0000778
______ 25.00 04.01- 2.00 BAJ0000753 04.021 2.50 BAJ0000754 04.03. 2.00 BAJOOOO755 04.04 2.00 BAJ0000756 04.05 3.00 DAJOOOO757
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.' g .. . ,-tt TESTLCROSS REFERENCE PACE. 2 .'0UESTION- ,VALUE- REFERENCE
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- --------- . - 04.06 2.00- BAJ0000758 . 04.07 2.50 BAJ0000759' ' 04.08 2.00 BAJ0000760 . 04.09 3.00 BAJ0000761- - 04.10 2.00 -BAJ0000762 - 04_.-11- 2.00- 'BAJ0000763 ------.
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L - FACILITY: SHOREHAM _---------___--_-_-_----_.
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-REACTOR TYPE: :BWR-GE4
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DATE ADMINISTERED: 85/09/17
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l' ! ** EXAMINER: -LANGE, D.
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APPLICANT:- _-----[f)_6_7__8__8_______ - j INSTRUCTIONEL TO APPLICANT
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p ! Usa _ separ a t.e. ' paper _f or' the' ar.s we r s . Write answers or. .one side only.
. 3taple' question sheet on top of the answet- sheets.' Pcints Yor each l- question are indicated ir, parentneses after the questtor.. The passing Stade' requites at least-70% in each category and a final grade of. at
;1 east'801.- Eaanination _ p aper s wi'll- be pickeci _up si: 36) hours after 1.he examinction< starts.
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l . CATEGOR) ' APPLICANT'S -CATEGDRY i: VALUE .% TOTALOF :5CCF E VALUE CATECCRY
. --_----_ -_-__- --___-_____ ____ -__ ____________-______________________,
25.00~ 25.00 ________ -_---- _________.__ -_______ 5. THEORY OF NUCLEAR F0WER-PLANT] ,' ~ OPERATION. FLUIDS. AND MDCC THERMODYNAMICS
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2 4 . 4+ ^-e$/; 25.~00 P"l-tO
-_----__ --~__--. _____--____-. ________ o. PLANT SYSTEh5 DESIGN. CONTROL, ANC INSTRuhENTATION- -25.00 25.00 -----___ _____. _______---_ -_-_-___
7. PROCE00RE5 - NGRHAL, AONORMAL.
EhERGENCY AND' RADIOLOGICAL-z c' CONTRCL 25.00 ! Tot
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. CONDITION 5, AND LIMITATION 5'
200.00 200.00 TOTALL
-__-_-_'- -_-__- -_-_-____-. ___-____ ' FINAL GRADE ________________,%
All ' wcrk1-done cr. this e:<smina t ion is my own. I have neither
' 9iven nor received sid.-
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~ 5.': THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2-
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______________ 00ESTION z5 01 (2.50)
: Control' rods which ate positioned close to the . intermediate / shallow rod boundary may. exhibit a * reverse power .tosponse".
. ta. What will: happen to the local flux and core average .(1.00) flux.when such a rod is withdrawn.
- b'. Why does the reactor respond this way?- -
(1.50) '0UESTION' -5.02 .3.00) (
Feedwater subcooling affects the reactor power respcnse.
(c) As resetor power incresses- what nappens to the. inlet.feec-water subcooling and WHY ? (1.00)
.(b) For a given gross power level, how does an increase in inlet-feedwater subecoling affect'the axial power distribution and WHY~? (1.00) (c) For a giser. elect-ical output, does the plant operate more eff-sciently with Inc. or dec. inlet feedwater subcooling end why ? (1.00)
GUESTION 5.03- (2.25) Hast Lbalance calibratior.s of the APFM's are nornally performed
: b y: the process computerr but n.ay need to be per f or med by hand.
(a) What are the five scurces of energy inputs required for e heat belance ca;culationi (1.25)
(b) What are the.three energy outputs or losses required for a heat b a l er.c e c a l c u l s t . c r:1 (;.00, GUESTION 5.C4 (3.00) .a. At EOL how much delta N/K must be added to place a critical reactor on a.60 second period. (State all or any assumptions.) (1.5)
b. Euplain why reactor power will decrease on a -6; <econd period shortly 1.o11owing a reactor scram. 41.51 (***** CATEGORY 05 CONTINdED ON NEXT PAGE ****ae
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* -5. LTHEORY: 0F NUCLEAR' POWER' PLANT OPERATION, FLUIDS,iAND PAGE 3
____ gg g ______________________________________ ______________ QUESTION 5.05 (1.50)
; Describe how' rod worth is affected by the followingi-(Indicate withi . increase,-decrease or stays-the same)
a. Moderator temperature increase (0.50) b. Void content increase (0.50's c. Neutron flux increase '(0.50) GUESTION: 5 0o. (2.25).
For each situation described below, determine if an SNPS Safety or Thermal Limit has been' violated. Name the' limit znd its value. (0.75 ea)
(a) Core flow = 2 5 '; . Steam come pressur e = ?50 ps13, MCPR = 1.10 (b) Lote flow = 60L, Ste am dor.e pressure = 780 psis, Thermal power = 70% -(c) CHFLFD = 1.001 GUESTION- 5.07 1I.00)
With all other parameters held const' ant, If Recirculation.Flcw increases, how will-the_fo11ouin3 pstsnatets respond initiallvi (Increase, Cecrease, or remain ccr.stant.) (0.66 ea) s.-Thermal neutt on pop ulatt or: b. Ther n.a l ._ d i f f us icn length c. Reactor' Water Level GUESTION 5.08 (2.50) The reactor has been operet:ng at 100* power for one nonth when a scram
. occurs-in which several control rods FAIL TO FULLY INSERT. Enough rods DO insert to brin 3Jthe reactor'svoeritical at the.tir.e of shutdown. If l reactor moder ator temperstore :,s stintained CONSTANT, end control rods .are NOT mcved, about HOW LONG w;11-the operator have to wait before he-can be reasonably sure that the reactor.will ren.ain suberitical?
EXPLAIN.. (2.50)- l (xrxxx CATEGORY 05 CONTINUED ON HEXT FAGE ***xx)
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-GUESTION 5.09 (3.00) -Following-a-reactor-scram, from 100 % powere,' explain what happens init- -
ially to the following parameters' (increase, decrease,.or remain the come) AND WHY ?
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y a. Flow through the. core. (0~.75) b.-Flowythrough'the-control rod drive pumps. (0.75) l -c. Control rod drive temperature. (0.75) d. Fressure drop' ire the steam lines. -(0.75i-QUESTION 5.10 _(3.00) , For each cf.the followir.s events, or~ changes in plant st'atus, state
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whether.the change will bririg tne system CLGSER TO, FARTHER FR0he.~or HAVE NO EFFECT ON the point at uhtch the Reactor Recirculation pun.ps
- 'will'eavitate. GIVE A BRIEF EXPLANATION flu TACH.
a. Increase in reactor water level- (1.0)-
- b.' Loss of a-feedwater hester. ( 1. 0 ) .
c.-Increase ii Recirculation Fump speer, (1.6)
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t i i 6. PLANT SYSTEMS' DESIGN, CGNTROL, AND INSTRUMENTATION PAGE 5
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QUESTION 6.01 (2.50) Th'e'uninterruptible AC power supply ,:(UPS) , provides continuous AC power.to non-safety relsted controls and instrumentstion.
(o) Under what conditions will power be supplied to the vital bus-,(UPS 41), directly-from the batt.6ty source ? 1(1.00')
(b) When would power be supplied to the vital bus from the alternate AC source ? (1.00) ;(c) What~is the purpose of the minuale. (Alternate Source)
by pass switch. (0.50)-
. 0UESTION 6.02 (2.50)
For each of the following Nucle &r Steam Supply Shutoff Syttem inclation signals, brieflv state, (1) the conoition for which
.protecticn is provided, (2) the concer n r equir ing tais protection, and (3) the setpoints'and reasons for salection of the setpoint values.
T(a) Hain' Steam Line High Ares Tempsrature (1.25)
(bi Reactor Vessel Low-Low L e v ed (1.25)~ -GUESTION 6.03 (2.50)
The APRM's receive and average signals fecn individual LFRM's.
(c) How .many LPRhs tre used in erch of the APRh channels?
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(0.50)
(b)'How would you determtne the flor. Level for a single
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LPRh G i v e ~t h r e e r.e t n o d s . (0.75)
(c/.What are the minimum LPRM inputs for each APEN channel accor ding to the ech Specs, and how . ar e they enforced? (1.00) (***** CATEGORY Oo CONTINUED ON NEXT PAGE *****)
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1a. ListLtwo conditions'which will automatically. bypass the ,
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Rod Block Monitor and for each condition briefly: explain why ? (2.00)~ ! _- b . How many LFRM inputs are required to prevent.an RBM instrument inoperative alarm-? -(0.50)
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GUESTION 6.05 (2.50) ; '
-The recirculation pumps are provided with tuo sets of mechanical /
cartridge' type seals'for i ts' se al- as semb.l y .
(a) Explain the purpose of both the~ Control Rod Drive and RSCLCW : water that is provided to the seals. After leaving tne seal = assembly where.does. water d.ischarje_to ? (0.75)
- (o)~At 300 % power -
l What is the normal. flow rate through the seal cavity ? (0.50s What.is the pressure in eath seal cavity ? (0.50)
(c) What woulc the pressures and flow rates be, and what clarms would be received, if only the No. 1 (intetnal' sesi were to fail ? (C.75)
GUESTION- 6.06 (2.00) l ' Core Spray anc RHP locp level system pumps snould be kept in " service at all times during' standby status to ai,sure that the RHRS discharge piping rema, ins full. WHERE and by WHAT INDICATION can an operator' verify that the discharge piping is inceed full.
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-GUESTION c. 0 7 (2.50)
Euplain the resocnse of the LPCI injection mode of the RHR system if a low # evel initiation signal occurs dur ing operation
of the RHRS in the shutdown. cooling mode. 'r o u r answer should include the response of the pumps, valves, and any operator action.
(2.50)
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GUESTION 6.08- (2.50) Concerning'the RBCLCW. system i a. What.three automatic conditions will cause a split of RBCLCW . loops.
(1.00)
-b._ List'six indications an operator has to-verify that a split of the-RBCLCWeloops'has occured. (1.50)
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'GUESTION' o.09 (1.50)
Concerning the Safetv Relief. Valves i
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fe1 TON AY# O o. How will-a safety relief ~ valve 5:11 ras failure affect the 1. f0 operation:of the valve. Consider all modes of operation. (+v60) t. - us i 4 ,u s r , + 4 n r. unni- .-o x om .e 4x: e <e, 7 . . e , p gg y ' _ (w) OUESTION :<6.10 (2.00) What sin (6; reactor protection system , ( RPS ) , trip functions are'never bypassed regarcless of MCDE SWITCH pcsition ? NOTE: Do not-include i osnual s c r a r. . (2.0
'0UESTION 6.11 (2.004 -
Refer to the attachcd' control panel diagram of the Cent ol Fod Drive system to answer the following.
-a. . State what has haFPenedf and li3t fi/E inO1 Cations which are off-normal w2 th an explair atiori of ee.ch.
(2.00)
(axax* END OF CATEGOR('06 muram)
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QUESTION - 7 ~. 01 ' (2.25)
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During a reactor shutdown from 20 L power procedure 22.005.01 cautions-
-you not.to use the Vacuum Breakers except for emergency conditions.
Explain-the-reason for.this caution and state when vacuum is broken
;por procedure. - ( 2 . 2 5 's .GUESTION 7L.02 (2.50)
la.-Listithe immediate operator- actions r equired upon reelept.- 0FF-' GAS.High Radiation alarm._ (1.00)- b.-What three-conttlcl room {indicatioJs'would you use to n.ahe a
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h A ' Tc M ' GUESTION. 7.03 (1.00)
(a) List the three entry condi-ions for the Level Control Emer3eney Procedu*e,(SP29.025.01). (2.00) ~ (b) Under what procedural :onditions would you enter the-Level Restoration Emergency Frocedure,(SP29.0:3.04)> (1.00)
GUESTION 7.04' (3 00) len.porary changet to appr ovec; statiori proc _edures sht11 be documented on the Temporary Procedure Change (TPCs Form. If during a startup on.a backsh2ft, a TPC is required to con.plete a CR0 coupling check ;
(a) What spprovsis are r equir ers pr icr to implementing the temporary change, And how ar e these approvals to be indic ted? (I'.00)
c(b)' What is the maximus. life o t' a TPC after approval arid are there any additional reviews required pilor to r eachirig ered of ~ lif e ? (1 00)
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____________________ GUESTION 7.05 (2.001 The : reactor is at 70 % power when you receive a trip of 'A' CRD.
pump due'to low suction pressure.-You direct'your operator to start the * B * CRD pump.'Two' attempts to start the.'B' pump fail. Should
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the. reactor be' scrammed at this time ? (briefly explain). -(2.00) QUESTION 7.06 (2.25) According to procedure SP-23.204.01 (LPCI), a caution exists that states _ to mair,tain Suppression Pool temperature in accctdance with the Tech. Spec. LCO. If no testing was'in prcgress , _ list-al!- , available indications an operator would have to determine if .the
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Supp. Pool temperature'had exceeced_tts Tech. Spec. LCO. (2.25) QUESTION 7.07 -(2.00) A step in the RPV Floocing Emergency Procedure (SP-29.023.07) directs you to the attacheo (Figure 1) maximum acceptable core uncovery Line ss time after-shutdown.
eS0 a. Which area of the gr aph is the unacceptable poriten--above or ,ef below-the curve? .(er&O) b. List'the entty' conditions to this procedure. . (4 ,5f)- IstT QUESTION. 7.00 (2.00) According to procedure, SF-21.004.01 , hain Control Room-Conduct of F ersonnel : What control reem censonnel can be designated to assume the control roon. coramand f unct2 cn during the absence of the Watch Engineer 0 Coes this requirement change for operationsi conditions'1 througn 5 (***** CATEGORY C7 CONTINUED ON NEXT PAGE ****m)
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0UESTION(7.09 (3.00) s. Drywell temperature has exceeded 296 degrees F, and SP 29.023.05 requires the operator to open all ADS valves. While attempting to comply with this requirement, he discovers.that only 2.5RVs-can be opened. List six~(6). alternate paths.to help cepressurire the reactor. (2.0)
-b. As Drywell temperature was increasing, Emergency Procedure 29.023.03,. Containment Control, directed that as. temperature approached 296 degrees the Reactor Recire. Pumps _and Drywell Fans be Shutoown. WHY? (1.0)-
w . GUE5TIDH- 7-.10 (3.001 s. List:tne five (5) entry cor.citions and.setpoints for.SF 29.023.03,
' Containment Control Emergency Proccdure'. (1.50) . b . - 1. If Suppression Fool temperature is 140 degrees F, and RFV pressure is 600 ps19,.have you exceedeo tne heat capacity limit of the ATTACHED figure 1? . 75)
2. Under these conditions, would the pr ocedure direct you to open
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i 8. ' ADMINISTRATIVE PROCEDURES, CONDITIONS,-AND LIMITATIONS. PAGE.'11-
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QUESTION .8.011 (2.00)
-The steady. state MCPR limit given in Tech. Spees. is multiplied by'a : flow biasing correction-factor Kf. Explain the bases f or this correction
, ' factoriincluding the events associated with it.
(2.00) DUESTION 8.02 (2.00) According to'the Tech Specs, define PRIMARY CONTAINhENT INTEGRITY.
(2.00) 00ESTION S.03 -(2.50) According_to precedure SP-12.011.01 , Station Equiptent Clearance
. Permit, specific Limitations / Actions must be adher ed to. . Answer :the folicwing statemints either TROI or False. If false, eaplain the proper cl e ar ar.c e l i m i t a t i o r. . -
s. The on duty Watch Engineer can delegate, 60 the on duty.Wate'h Supervisor, approval or lifting of a SECPoss long at he is kept . fully'informec of the system status. . (0.50)
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b.- Cautiori tags may' be plac ed by ariyone who has been trained t r.
the use of the+e cards. ~0.50)
(
c. A Hold.Off. Ta3'and a Caution Card shall not be affixed to a specific component at the same time. (0.50) d. Equipment Information Carcs may be placed.by any LILCO employee with the Watch Engineers approval. .( 0.50) e. A Hold-Off.elearance end a Ceution clearante shall not be issued for a piece of equipment at the same time. (0.50) DUESTIDH 8.04 (3.00) Describe the three conditions that would necessitate you to
: direct, or be directed to initicte the Standby Liquid Control System. (3.00) (***** CATEGORY 08 CONTINUED ON NEXT PAGE m****)
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E8.: - ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 12-
.__________________________________________________________ "
GUESTION 8.05 -(3.00) Mid-way through'the 4PM to midnight shift, you are informed that, whilo-troubleshooting an: EHC ~ problem,L the Instrument Technicians discovered that the Bypass Valve Control Unit on the:EHC System is INOP, and the d: mand signal for.the Bypass-Valves is locked-in at ZERO demand. They.
octimate:that;it will take until tommorrow to repair the-circuitry.
-The plant is~ presently. operating at 80% power with; direction from the Plant' Manager to increase power to 100% at 10MW thermal / hour.- USING~THE ATTACHED TECHNICAL SPECIFICATIONS, deter mine what action (s) aust be taken in this situation. (3.0) QUESTION 8.06 (2.00) There-are numerous log's.that vou, as a Watch Enginecer a.ust review before assuming the waten. What ar e thesc log's. (2.001 QUESTION 8.07 (3.005
'Ac:ording to.the Radiation Work Permit Frocedure (SP 12.012.C1), there are six)(6) .cond2tiens when an Rut should be initiated. List 5.of these 4. conditions, . including raciation levels if' applicable. (3.00)
GUESTION 8.08 (2.50) LThroughout the 5,horeham Technical Specifications, in various sections, the_ f ollowing staterwent appear s;
" *
The provisions of of Specification 3.0.4 are not applicable.'
!What is Specification 0.0.4 , and what does this statement mean? (2 501-(***** CATECORY 08 CONTINUED ON NEXT PAGE *****)
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 13 , t 100ESTION 48;.09 (3.00) l l. -Uning the attached containment control figures'from SP.29 023.03, cnswer t.he following - j o. What'is the~ minimum suppression pool uater level given an RpV-pressure of 500=psig and suppression pool temperature of 160~F? =(1.0)
.b. If suppression pool level cannot be maintained above this levele the procedure. directs the. operator to open all ADS valves and depressurire. What is the reason for this procedural step? (1.0)
' c. During depressuri:stion, what must be, checked prior to-depressurizing.
below 110 psis. -(0.5) d. Can the cooldown limit of 100F/he be enceeded during this depressuri sticn (0.5) ! ! QUESTION E.10 ( 2 . 0(ei
,During operations at power itE ls discovered that'the 'A' RHR heat L exchanger: bypass valve is stuct.open and will not : lose. During
, subsequent. investigation, the outbo4rd 'B' LpCI injection valve (MOV-378) L is f ound .to tie inoper able due .o an electr ical malf unction. 'Using the attached Technical Spe 2fications, determine the' allowable
t2me the reactor may continue. operation. REFERENCE sections 2n the.
[ tech specs used'in~your.ar,swer. (2.0) i R
, (xxx** END OF CATEG3RY 08 *mann) (************* END OF EXAMINATION *****samax4ssmx)
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EQUATION SHEET , f = ma v = s/t i Cycle affichncy = (Net wrk cut)/(Energy in)
w=q s = V,t - 1/2 at
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E = x- * KE = 1/2 av a = (Vf - V,)/t A = IN A=Aeg PE = agn Vf = V, * at * = a/t t = sn2/tjjg = 0.693/t1/2
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y,,j - 2 1/2*" * U* m A= n04 [(g1 /2} * II b}3
d = 931 an -
-m a = V ,yAo , ,, s,te o . . t s
Q = mCoat * Q = UA A.T
[ = I e~q I = I,,'0** N j ,
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HVI. = -0.693/u p =p e/** t ~
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SUR = 25.06/T SG = 5/(1 - K,ff) Gx= S/(1 - K,ffx)
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SUR = 25e/t* + (s - o)7 G;(1 - K,ff3) = C22 U ~ "eff2) -
. T = ( &=/s ) + ((a a s'/Io] M = 1/(1 - K,ff) = CR;/G,
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T = 1/(s - 8) M = (1 - K ,ffe)/(1 - K ,ff;) T = (s - o)/('Is) SOM = (1 - K ,ff)/K ,ff
' 4 a = (X,ff-1)/X,ff * 4Keff/K,ff
-- t* = 10 secones I = 0.1 seconds"I e o = ((t*/(T K,ff)] * (s,ff /(1 + IT)]
Ij dj =2 I,2d2' Id j gd P = (14V)/(3 x 1010) 22 2 I = sN R/hr = (0.5 C2)/d (,,g,,,) R/hr = 6 CE/d2 (feet) , Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem. I curie = 3.7 x 1010dos 1 ga]. = 3.78 11 tars 1 kg = 2.21 lem 1 ff = 7.48 gal. 1 np = 2.54 x 10 3 8tu/nr Density = 62.4 lbe/ft3 1 m = 3.41 x 100 Stu/hr ,
^
Oensity = 1 gm/c9 lin = 2.54 cm \ Heat of vaporization = 970 Stu/lem *F = 9/5'C + 32 A Heat of fusion = 144 Stu/lem 'C = 5/9 ('F-32) ' 1 Atm = 14.7 Asi * 29.9 in. Hg.
I ft. H O = 0.4335 lbf/in.
1 BTU = 778 ft-1bf t/
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. /h9s7et? .0^5kJe2 A'ef 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 14
_ ______________________________________ ____7q g ______________ ANSWERS -- SHOREHAM -85/09/17-LANGE, D.
ANSWER 5.01 (2.50)
'
f c. Local power will increase Average reactor power will decrease. (1.00) b. This occurs because the negative reactivity added to the core due to increased boiling deeper in the core overcomes the positive reactivity added cue to rod withdrawal. (1.50) REFERENCE SNPS Reactor Phvsics Hodule pg. ~-201 ANSWEE 5.02 (3.00)
/ (c) It decreases.
As power increases, 1.o r e entraction steam is available for foeduatcr heating (1.00)
(b) Increased subcooling aill result in more power generation at the octtom of the core.
Eecase of temperature ccefficierat of reactivity. (1.20)
(c) An increase in tuocooling wil: require additional core heat to suppcrt a 31ven electrical outFut. Decrease succoolin] lead; to increased effacienev. (1.00)
FEFERENCE SUPS Heat Transfer / Thermodvnamics Module.
ANSWEP 5.03 (2.25)
(a) (1) Ceecuster (22 CRD return (3: Cleanup return (4)
C @ (5) Recirc pump hest (1.25)
(di. -(1) R eSteam ce (2) Cleanup system out (3) Losses to a m b i e r. t ( f t::ec heet losset) (1,00)
REF: LP Therms! A hydralute Design pg 24-25 Y ht - %4 ) Tc &~wd l- - , I
,
... , - -
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 15 ______________________________________ ____7 g g
.______________
ANSWERS -- 3HOREHAM --85/09/17-LANCE, D.- ANSWER '5'04
. (3.00) / a '. Using stable. period equasion T =~1ambda delta K/K 60 = (0.1)(delta K)
dolta K = 8-x :10 -4 delta K/K (1.5) NOTE: Req. to use EOL valve.of Beta.for full credit.
16. Following_the scram delayed neutrons will continue to be produced by the longer lived precur sor s. .The last remairiing courceJuill have a 55.6 second.h.sif life. Pericd is 1,44 times
,or 0.693 dividec" into the doubling tine or half life.
Upsilon ~= 55.6 seconds /-0.e93'= -50 seconds (Calculation not req.
' for full crediti (1.5) REFERENCE SNFS Reactor Physics Module.
ANSWER 5.05 (1.50) a.' Rod worth. increases as moderator temperature i r.c r e a s s = .
~
f/ (0.50) b. As void cor, tent increases r od worth. decr eases. (0.5) c. The worth of'a control rod.is a direct function of ' . chermal neutron flus to whicn it is expose so as fluu incresses, wor th increases. (0.5) REFERENCE SNPS' Reactor Physics, Ch. 10 & 13.
ANSWER 5.06 ( 2. 25 > . s (s) NO safety limits or thermal' limits viciated.
(b)~YES- the safety limit covering Thermal Power for Low Pressure has been violated. Steam dome pressure shculd 'r.ot be less than 785 psis.
.(c) YES- the thermel linit for LHGR has been violated. The LHGR LCO is violated sonewhere in the core if CHFLPD is greater than.1.0. . .
(0.75 ea)
REFERENCE SNPS Thermal Based Limitations Module, pages 10-36, 10-35, 10-44.
l l
.--
-.
-
,,
,
.
.
-
5. THEORY OF NUCLEAR' POWER PLANT OPERATION, FLUIDS,'AND PAGE 16 ____ ______________________________________ q ______________ ANSWERS -- SHOREHAM. -85/09/17-LANGE, D.
ANSWER 5.07 (2.00) c. Increase: b. Decrease c. Decrease (0.66 each) REFERENCE SNPS Fluid-Mechanics and Reactor Fhysics Modules -
. ANSWER 5.08 (2.50) / 70 hours (1.0)
It util take appron.~70' hours for the Xencn to peak and then decay after
'the scram. If the positive reactivity inserted by the cecay of Xenon is less than the shutdown reactivity due to reds,~then the reactor will remain suberitical. (1.53)
_ REFERENCE Reactor Thecry - Xenon Transients
!
ANSWEF 5. { g , (3. g , m. g c,,, Lea h ([.' Increase. ( 0. 25 . ) Cue to causing less two phase flow the void collapse when power decreases, E less flow resistance. ( 0.50 i 6.-Increase. ( O.25 ) .The. charging water head decreases due to-the re-charging of the scran, accumulators.-( 0.50 ) c. Increases. ( 0.25 ) When the drive'is in motion the cooling water flow is closed.off by the ball check valve. ( 0.50 ) d.. Decreases. ( 0.25 ' Steam velocity decreases due to the scrams-therefor-the fluid head ( pressure ) losses are lower. ( 0.50 )'- REFERENCE "" 4 Fluid Flow and-CRD System
.
L
-
: a
. -
.
.c
-
.5.- THEORY OF NUCLEAR POWER PLANT OPERATION,. FLUIDS,'AND, PAGE 17:
...----
ggggggggggggg--------------------------------------
,______________
l ANSWERS --cSHOREHAM -85/09/17-LANGE, D.
5.10 -13.00)
}/ENSWER
. - - . . . . .
,
~a.' Farther 1from cavitation (0.5). As thecreactor water 11 eve 1Lincreases, '
the static-headoof water component in:the NPSH determination is
.also increasing.which' adds NPSH. (0.5)
b, Farther ~from scavi tatiori . ( 0.5 ) . If a feedwater hester is lost, .then-
~the temperature of the water entering the reactor is lower, which brings the water. farther from;the saturation temperature. (0.5)
c. Closer.to cavitation (0.5). As pump spee.d increases. the pressure in-the eye of1thesimpeller/ decreases, which will.cause the pump
.
to cavitate _ earlier with:the same'NPSH. (0.5) REFERENCE Th e r modyn'a mi'e s
. .h
r .
's
!
.
.
-
.6, PLANT. SYSTEMS-DESIGN, CONTROL, AND. INSTRUMENTATION PAGE- 18
______________________________________________________ ANSWERS -- SHOREHAM -85/09/17-LANGE, D.
.
.AN'SWER 6.01 (2.50) - - (a) (1) Loss of. normal'480 V power supply "The operator would consider a loss of pwr to be a rectifier failure." (0.50) (2)nRectifier fails (0.50) (b) '(1) In event of inverter failure - lo inverter. output (2)
Manual push button depressed - ' Alt. Sourcesto Load" (1.00)
(c) lIt is in.the system to permit' maintenance work to be-performed on UPS. (0.50)
.REF: L.P.-4313, pg 5, S. 10, Fig. 1 ANSWEFf 6.02 (2.50)
'
(a)'(1)IProtects against a breach in hSL containment (2s Concern is release of radioactive material e::c es si ve loss of coolants (3) Setpoint is high eriough accve norn.al expected during operation to prevent spurious isolation, low enough to provide
.ocrly indication of a break. _
-(1.25)
185'.degr ees fahrenheit Shoreham also monitors main steam tunnel high delta T = 50 degr ees ~ f ahrenheit
, ( b )-- ( 1 ) Frotects against-potential. breach in nuclear primary pressure.bcondary (2) Concern is continued inventor / loss core overheating radioactive materials release (3) Setpoint is low (1.25)
enough to~ allow heat removal for a predetermined time following a scram, and high enough to pr ovide ECCS in event of a lerge leak.
- -3 8 ' ' : wide range REF: LP 4650 Appendix I ANSWER 6.03 (2.50)
/ (a) 17 in A, C, &E (RPS A) 14'in B, D, &F (RPS B) .(0.50) (b) (1) Select a coritr ol rod & obser ve 4 cod display (2) Select at back panel (3) Obtain a computer run (4) Run a T2F- trace (5)
Quadraul Symmetry ( 3 at .25 each)
(c) 11. min / channel. If < 11, get INOP 2 hin/ level. (0.50)
Administrative 1y controlled (0.50)
. REFERENCE SNPS. LP 4603 & 604.
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6.. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 19 ______________________________________________________
~ ANSWERS -- SHOREHAM -85/09/17-LANGE, D.
' ANSWER .6.04 (2.50)
(a) (1)' Reference APRM indicating < 30%. (0.50) (2) Edge rod selected (0.50) (1)'Iffaveragefcore power is < 30%, local power conditions which could leak to fuel damage from single rod withdrawal cannot be achieved.
(2) At the edge of the core,. local fuel thermal limits cannot be exceeded by rod' withdrawal.
(b) 50% of expected inputs
' REFERENCE-SNPS LP-4.606 pg 1-16.
ANSWER 6.05 (2.50) aOk :} , j
(a) :(1) RBCLCW cools the seals, then coes to the RDCLCW Hi' 83 # (.375) -(2) CRD pr ovides seal purge and then leaks past the breakdown pbushir.g into the P :: , or goes to the DWEDS (.375) '(b) 3J:sp n. of CRD water /pu_mp , . < M yk -0.75 spn. is se al (s t a g ir7D #g#7 3 (0.50)
nu J . Seal #1 cavity pressure = 1000 psis Rx pressure (jyjgg : Seal 42 cavitv pressure = 500 psig r 1/2 of 41 (0.50)
(c) (1) Pressures in both cavities would be at Rx pressure (.375) -(2) Flow rate would increase over destgr; flow rate, and the ' Recite Pump Seal Etagir.g Flow Hi/Lo' wouldannunciate(at0.9h( .375 L spm.
REFERENCE SNPS LP-120 (Recirc.)
LP'-113 ;RBCLCWe
.SNPS 120.01.1288, alarm resconse procedure.
pHSWER 6.06 (2.00) Loop fill presure of appro::. (40 psis) can be seer. on 1E21-PI-001A, and E-for the CSS on panel -601. (1.00), and by observing that. annunciators, Line Fil-1 Pump A/E discharge Low Pressure (1116) and Line Fill Pump A/B Flow High-(1117) are not illuminated.- (1.00) REFERENCE
~SP- 23.121.01 rev. 10 pg #3, precaution 4.2.
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6.~ ' PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE' 20 ______________________________________________________ ANSWERS -- SHOREHAM. -85/09/17-LANGE, D.
NSWER' 6.07 (2.50) ~ The RHRS will automatically allign itself to the LPCI injection mode oxcept'the pumps will not start due to no suction path open from the suppression pool. .. (' O.50). The Operator must close-( MOV's 032 A-D )
-chutdown cooling suction, (0.75) , open HOV's 031 A-D upon which the pumps.
will automatically start, (0.75) and reset the-SDC isolation logic for NOV-037 A/B'by depressing the reset switch . (0.50) REFERENCE S F -23.121. 01, - p s .~ 4 10 all. Operation of RHRS in SDC.
ANSWER ~
~
6~.08 (2.50) v
* pS .
a.-RSCLCW splits on a LOLA signal (0.33), and on LO-LO Head Tank level (.33) and RX. Low level - 132.5. (0.33) b. 1. RBCLCW pumps' continue to run.
57~ . 2. Non-safety related loops are isolated.
i/[*[4[- 3. ~
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HX, that is cut-cf service is put into. service 4. Both- RBCLCW outlet' valves and ser vice water valves auto open.
5. NG' set cooler circ.. pumps are isolated and trip on Ic-suction.
6. PCU is isolated.
7. AUV check valves close.
S. HX outlet valves will open. (6 reeuired et 0.25 each) REFERENCE '"'" "Y ~ *
%I '
RBCLCW 1esson plan. # 118.
eo ANSWER- 6.0c ( 1.EC ) "
.
k . s te ~ a. If the be ll-su s - has failed the self actuation (safety _ mode) is lost but relief mode is still operable. (1.0) M MM. Se11ous failure-i+-tndie+ted 4Mw-s+stm M C AF ) 10. 5 ) REFERENCE SNPS. LP-116'. A%S- 2ol, NO ,
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SYSTEMS DESIGN,' CONTROL'_, AND INSTRUMENTATION PAGE 21
- :6.__ PLANT
__ _____________________________ ____________________
' -ANSWERS -- SHOREHAh -85/09/17-LANGE, D.
' A'NSWER', 6.10 '2.00)-
(
Scrans uneffected by-mooe switch.
1. Low-Rx water level (.33) 2. High Rx pressure (.33) 3... Hi gh D.W. pressure (.33) 4. Main. steam line.high radiation (.33) 5.
APRM flow biased HI: (.33) 6. APRh INOP (.33) REFERENCE
=SNPS- LP 4-~611.. ' ANSWER ~ 6.11 (2.00)
' a. Reactor scram prior to reset. (0.25)
-
Off-normal'2nd2 cations, with euplaination i 1. Pump an.ps high due to high flow to.SDV.
2. Pilot air header icw al ar ra due to scram F ilot valves doenergized.
-3.'CRD sys; flow is high due to high flow to.SDU 4. Flow; control valve closed due to high sys flow - 15. CRDidrive water D/F low due-to flow control valve closed 6.: CRD cooling water D/P. low due to flow control valve closed 17.-CRD1ccoling' water flow-lcu due to Icw colling flcu.
(4 reqd 0H0.5 es) REFERENCE
- S H P S -. L PL 1 0 6 ,
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L7, : PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 22
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~~~~Rd656L6656dL I C6UTR6L
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . ANSWERS -- SHOREHAH- -85/09/17-LANCE, 3.
ANSWER 7.01 '~(2.25)
(1) Openins.the-vacuum breakers imposes excessive loads on the turbine last stase buckets. (1.00) . (2) Vacuum shall~not be broken until the unit shaft rotation.has decreased to 1200 RPM. (It is in the' procedure t'o be after reaching cold S/D (8.1.3.6).J (1.25)
REFERENCE SNPS SP-22.005.01. 8.1.9.7 a 8.1.3.8.
-ANSWER- 7.02 12.50)
(a) 1. Verify the alarm on~the 3f'-Gas radiation monitor-r ecor der .- (0.50>
2.. ' Reduce, reactor power te clear the. alarm. (0.50e (b) . Whenever .the HI-HI setpoint is icached in any of the fo11 cuing.)
1. -Main Steam Line Radiation Monitor Recorder 2.- Off-Gas Log.
Radiation Monitor Recorder (SJAE Outlet Rad Mon) 3. Off-Gas Vent Pipe-Radiation nonitor Recor der (Cnarcoal Ben Outlet or Off-Gas Outlet Rad honitor) (0.50 each) gt,63l , gj REFERENCE "'"~' SNPS.: SP- 29.002.01
. ~ ANSWER' 7.03 (3.00)
e) The three conditions ar e any of the following 1. RPV water' level less than 12.5' (0.66)
~2. D/W pressure > 1.69 psig. (0.66) '3. An isolation condition exists which requires or initistes R> . scram. (0.66) -REF' Proc.429.023.01 jg/ b > Enter from 29.023.01 or 29.023.02 when-(level control) -(cooldown)
1. Level cannot be maintained above TAF = -153 on fuel-:one. (0.50) .
.'2. . Level cannot be determined. (0.50)
M c7
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RADIOLOGICAL CONTROL
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. ANSWERS -- SHOREHAM -85/09/17-LANCE, D.
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' ANSWER- 7.04 (3.00)
La)1 Approval ini . . _
.
.1.-{ via member-ofPlantManagementStaff (PMS) 2.T licensed SRO
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signatures on'the form.
2H gi . Verbal ' approval of PHS is OK if directly given to.the SRO signing /+(g the .. b).31TPC (and such is indicated on the~TCP-forn @ g . days - unless approved for permanent change.'
(2.00)
(0.50)
ROCJreview. required within l'4-days. (0.50)
~
REFERENCE
.SF-22.006.01r pg4 9 f 10.
- ANSWER- 7.05 ( 2. 0 0-) - NO.'(0.50).
- The reactor must be scrammed after the second accumulator light is lit.
., (0.75).
'At greater.;tha :600.psig ieactor pressure is suficcent to' insert rods-
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without accumu atorfpressur'e.
( 0.~p7 5 ( G %(g sc( . a .-, - f.3 gu-cs REFERENCE-
:SP-23 106.01i rev.8, pgf.15-18.
ANSWER 7.06 (2.25) '
:[1..-(04670(8) Supp.'Poo1 DIV. 1 (2) Instm. Temp. High. (0.75)
2~ .Supp.' Pool temp.' indicator. PNL-602.
. (0.75) 3. Temperature recorder.s on back panet. (0.75)~
. REFERENCE SP..23.204.01--(LPCI), .Contrel Room Instrumentation, Alarm's.
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- 74- PROCEDURES - NORMAL,-ABNORMAL, EMERGENCY AND PAGE 24 ~ ~~~~R I656E655EIL E6sTRUE~~~~~~~~~~~~~~~~~~~~~~~~
____________________ ANSWERS'-- SHOREHAM -85/09/17-LANCE, D.
ANSWER 7.07 (2.00)
. ., v . ,
J' a.'Above the curve. (Or7S) 4 b.(1) Temperature near_the cold reference les instrument vertical runs-exceeds the RPV saturation limit and indicated RPV level is'less than - 38. (O'.+ S) fo- ru
'(2) RPV water level'cannot be determined. (0.95) (3) Suppression; chamber pres'sure exceeding-pressure suppression .
limit. ( 0. F5 )
. ' .ge REFERENCE : S P.-2 9. 0 2 3. 0 9 , ' ANSWER 7.08 (2.00)
7 4.During operational' condition 1-2-3, an individual, other than the STA, with a curr ent SRO license. -(1.00) b. Dur.ing operational condition 4-5, an individual with.an RO or SRO current license. (1.00) If candidate answers according to control room position. acceptable.
REFERENCE-SNPS-'SP-21.004.01.
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7. ~ PROCEDURES - NORMAL, ABNORMAL, ENERGENCYlAND PAGE 25
~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~RI656L655Ch[~56NTR6L
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: ANSWERS:-- SHOREHAh -E5/09/17-LANCE, D.
ANSWER 7.09 (3.00)- 1. RCIC 2.'HPCI
:3. Main Turbine' Bypass valves 4.- Steam Jet Air Ejectors ~5. RFPTs 6.' Steam-Seal Evapor ator 7. Main Condenser Deaerating System S. RPV Head Vent 9. Main-Steam Line Drains 10. RWCU.Elowdown ho'de .
11. RHR Steam Condensing Mode
'( ANY d G .33 pts each)
b..Th'is is in anticipetion of the starting cf the drywell sprays due to high temperature.- If. spray occurs, you want to secure e ectrical-equipment in containment' to prevent damage to shor. ting out '( 1. 0 ) ~
: REFERENCE-23.023.05 -- Rapid RPV Depressurizatior: Emergency Frocedure
]
. ANSWER 7.10 .(3.00)
a. Suppression Fool Temperature e 90 degrees F Drywell Temperature ? 145 degrees F Drywell P.~ essure > 1.~49 psis Suppression Pool Level > +6 inches Suppression Pool Level <-6 i r.c h e s (0.3 each)-
col. N0l (0.75) c . 2 '. NO -(0.73) REFERENCE SPc29.023.03,-pages 1 and 5
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8. ADMINISTRATIVE PROCEDURES, $ CONDITIONS, AND LIMITATIONS PAGE 26 __________________________________________________________ ANSWERS -- SHOREHAM -85/09/17-LANGE, D.
ANSWER 8.01 (2.00) This flow adjustment-factor increases the'MCPR limit at core flows less than rated. Events such as loss of FW heating and tur bine trip without
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bypass become less severe when initiated from power levels less than the design value. This is due to decreased steam flow. .But events such as inadvertant start up cf an idle recire. pump, recire. flow controller failure (increased flow ) and FW flow controller failure ( m a:: ) can be-come more severe than transient 3 which are limiting at -design condit-ions.- dJs % d d i ft W - ( 2.00 ) REFERENCE ' df* ~ *'N' SNPS - Thermal Limits, Student Module pg.10-22 . ANSWER B.02 (2.00) PRIMARY CONTAINMENT INTECRITY shall exist when:
'a . All primary containment during accident conditions ! '1. by an OPERA 5LE primary containment isolation system, 2. by at11 east one manusi valve, blind flange, or. deactivated automatic valve secured in its closed position, except as provided in T able (3. 6.3-1\ of Speci f ication, 3.6.3.
n b. All primary containment are cicsed and sealed.
c. Each primary containment is.in compliance with the requirements of to Specification 3.6.1.3. - both closed except when using - then one closed. Leakage rates' within spec.
d. The primary containment are within the limits of Specification 3.6.1.2. - i.e. within spec.
e..The is in compliance with the requirements of Specification 3.6.2.1. - level, leakage, H2 seel f.'The ; e.g., welds. bellows or 0-rings, is . REF: T.S. pg 1-5 7, y, p g ,bg <%.d'[ '/.IE "8* ANSWER 8.03 (2.50) a, TRUE b.-TRUE c. TRUE d.' TRUE (0.50 esco) G.
REFERENCE
'SF-12.001.01
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< * / 8. . ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS' FAGE 27
__________________________________________________________ ANSWERS --~SHOREHAn -85/09/17-LANGE, D.
~ LANSWER 8.04 (3.00)
-SBLC would be-initiated under fo11owing.(3) . conditions .n. EitherLof the.following two answers '
11.'The.Rxcis' critical.and/or power is. increasing as indicated by . ncutron count rate or' steam flow-and the operator is~ unable to l
'chutdown with control rods.
2. From ATWS' procedure: If Rx' power is:above 6% or RPV level
:cannot be maintained abov'e 12.5' or suppression ~ pool temperature reaches 110 degrees. fahrenheit. (1.0) ..b. ~ Cr iticalityf i s predicted to occur within one hour. based on cooldoun'and/or-xenon decay. (1.0)
c. Hazard exits to plant personnel and the environment,.and plant- - abandonment is required. (1.0) PEF: SNPS. Lesson Plan #123 & SP'29.004.01
: ANSWER- - B'. 0 5 (3.00)
!. '/ T.S. 3/4.7.10 requires that the' turbine bypass' system be operat'icnsi
.when. thermal power is greater than or equal to 25%.of rated.
LWith'the-system INOP, and unrestcrable within one hour, take the action
.
required by T.S. 3.2.2 T.S. 3.2.~3' requires MCPR to be determined to be gree,er than or equal to the-hCPR limit as a f unction of ' aver age scramDtime as shown in-Fig.
3-2.3'1 times'the Kf'shown in Fis'3.2.3-2 If these conditions are raet, T.S. 3.0.4 is not-applicable, and operation can continue.
F.EFERENCE SNPS Technical Specifications 3/4.7.10 and 3.2.3 r
ANSWER 8.06 (2.00)
/ ..
1. Outstandin3 LCO's 2. Temp. proed. los
'3. Lifted lead log-L4 HWR : l o's . ~5. Outstanding RWP file (0.25 each )
6. SECF los 7.-Night Orders-B.' Watch Engineers los ,
; REFERENCE lSF-21.002.01,'rev. 10,-pg.46.
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND' LIMITATIONS PAGE 28 _______________________________'___________________________ ANSWERS -- SHOREHAh -85/09/17-LANGE, D.
p WER 8.07 (3.00)- 1. Maintenance in an area where radiation exposures are in excess of 5 arem/hr 2. Entry into an area where radiation exposure rates are in' excess of 100 mrem /hr 3. Entry into an airborne area 4. Entry into a. contaminated area ( 500dpm/100 cms squared) 5. Work in-an. area ~with neutron dose rates greater than 2 mrem /hr 6. Where radiological conditions are unknown C 5 of 6,at 0.5 pts'each] REFERENCE SP 12.012.01 ' Radiation Work Permit' ANSWER -8.08 (2.50) Entry into an Operational Condition or other specified condition-shall not.be made unless the conditions for the LCO are met without reliance on' provisions' contained in the Action requirements. This provisi~cn shall not prevent passage through or to Operational Conditions as required to comply with Action requirements. . Exceptions to these requirements are stated in the individus1 specifications.
The statement means that you can enter any operational condition-with' that LCO in an action statement ststus.
. (2.50) REFERENCE Shoreham Technical Specifications m s ANSWER 8.09 (3.00) [s 7.g.y.f,i' _- m. From fig 1 Thc=10 F,-min level for Thc=10 from fig. 5 is -4 ft. (1.0) 6. To ensure the suppression chamber could absorb the energy released from the reactor and not exceed containment design values. (1.0)~ c. Ensure-motor driven pumps sufficient to maintain RPV water level are running and available for injection. (0.5) d. Yes- (0.5) REFERENCE SP 29.023.03. Containment Control l'
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'* ' , . . . > .ll8.~ADMINISTRATIVEPROCEDURES,-CONDITIONS, AND-LIMITATIONS. PAGE- 29 --------------------------------------------------------- ,
ANSWERS -- SHOREHAM -85/09/17-LANGE, D. - NSWER 8.10 ' ( 2. 0 0 ) -
~ . 'Must : restore _the inop' bypass' valve within 72 hrs or be shutdown in 12Lhrs..Ref. TSL 3. 6.2.2 or- 3. 6.2. 3. (2.0) ' -REFERENCE TS 3.6.2.2 or-3'- . ' -
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. -L^' ^s E , TEST. CROSS: REFERENCE' PAGE, 1 :
QUESTION; lyALUEJ ' REFERENCE
- . ________ ______ __________ -
05s01 'DJL0000404-
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2.50' 05 02 ~ 3. 0 0. DJL0000405
'
05.03 :2.25- DJL0000407
.05.04- 3.00: .DJL0000424 ' ,05.05' 1.50= -DJL0000425'
05.06= .2.25- DJLOOOO438
. 05 07.. '
2.00 DJLOOOO439
. 105.08 ~2.50 DJL0000442'
05.09' ~ 3.00 DJLOOOO463
'05.10- 1 3.00- DJL0000464
______ 25.'00 0s.01- . 2 .' 5 0 DJLOOOO403 06.02 2.'50 DJL0000409 06.03 2.50 DJL0000410 06.04. 2.50 .DJL0000411
. J06.05 . 2.50 DJLOOOO413 :06.06 2.00 DJL0000427- <06.07 -2.50- -DJL0000428 06.08- 2.50 DJL0000431 06.09 1.50' DJLOOOO459-06.10~ :2.00 DJLOOOO460-06.11 12.00 DJL0000473
______ 25.00'
;07.01 '
2.25 DJL0000414
:07.02 -2.50 DJL0000415 07.03 3.00- DJLOOOO420 07.04' 3.00 DJL0000423' :07.05 2 .' 0 0 ' DJLOOOO429 107.06 2.25 DJLOOOO430 07.07; 2.00 DJL0000453 '07.08- H2.00 DJL0000465 07.09 3.00 DJL0000468 07.10 3.00 'DJL0000470 .______
25.00 108.01- 2.00 DJL0000003 08.02 2.00 DJL0000421 08.03- ~2.50 -DJL0000455 08.04 3.00 DJL0000457 08.05- 3.00' DJL0000461
.08.06> 2.00 DJL0000466 05.07- .3.00- DJL0000467 .08.08 '2.50 DJL0o00469'
108.09 3.00. DJL0000471-
-08.10 2.00 DJL0000472--
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'25.00
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-_____.
._ 100.'00.
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ANaeAme ,4 A i September 18, 1985 NTS-85-0239
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Mr. David Lang USNRC Region 1 , 631 Park Avenue - j:
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King of Prussia, PA 19406 Subject: NP.C Reactor Operator. Exam Comments (9/17/85)
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Dear Mr. Lang:
Attached are our comments on the 9/17/85 NRC Reactor Operator i
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Examination. Where necessary, we have included copies of 4 applicabic reference material to support our challenges on specific answers.
If you have any questions or comments, please contact me at (516) 929-6700.
Ver trul ours,
. : . !.O M^
Ken Rottk mp Station Training Supervisor KR/bv Enc.
cc: NTS File NOSF SR2
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lh September 18, 1985 NTS-85-0239 Mr. David Lang
.USNRC Region 1 631 Park Avenue King of Prussia, PA 19406 Subject: NRC Senior Reactor Operator Exam Comments (9/17/85)
Dear Mr. Lang:
Attached are our comments on the 9/17/85 NRC M ior Reactor Operator Examination. Where necessary, we have included copies of applicable reference material to support our
,, challenges on specific answers. ,
**
If you have any questions or comments, please contact me at (516) 929-6700.
Very truly yours, H Ken Rot kamp Station Training Supervisor KR/bv Enc.
cc: NTS File NOSF
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SR2
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m y @ 7F?gC, _ 7 p 1.4 ' Change existing answer key: steam tables supplied to candidates: work may consist of finding SAT pressure and comparing with available pressure with no
, calculation required..
2.2 For part b) give full credit' for either answer listed since exam question did not specifically ask for 2 answers.
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'2.6- a) Accept either answer, " limit p on internals" or " limit mass flow rate", ;as question did not ask for multiple reasons.
2.8 Rather than listing bus program, the candidates may list actual actions caused by
. the- bus: loading program.
Ref - Student Handout #309 3.5- b) ~ Accept " loss of generator indications".
4.3 Accept. precipitation and/or crystalization.
4.4 Add to answer key the following: _ 1. If solenoid actuated ~the SRV, red indicating. light will be on.
-2. Turbine control valves will close down.
3.- Rx power will decrease and then return to its original valve.
4. . A steady state level'off-set from feedwater controller setpoint.
,
.Ref ADS Student Handout HL-201 Rev 3.Page 17 4.8 a). Candidate may also give:
1. Low high volts.
2 .' Switch out of operate.
3. Module unplugged -
'and should not lose credit fo'r these,
. -
b) Also accept prio p going to startup with mode switch.
. 4.11 a) Add the following alternative answers: 1. Tech. Spec. recire pump LCO requires both loops in operation.
2. As per Tech. Spec. bases for recire LCO, there is no ECCS analysis for single recirc pump operation.
{
, b) Question'is confusing to candidate as there is no requirement to maintain any temperatures within five degrees of each other, during S/U as per SP 22.001.01 Rev 11. SP 22.000.01 Rev 11 does require temperatures to be ' logged on SPF 22.001.01-3 during startup. Therefore, any three temperatures listed.on the SPF should be acceptable fo'r full credit.
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5.3 Delete answer " Reactor" from answer key. Heat balance is done to solve for reactor power and therefore is not generally considered an energy input.
5.7 For Part B - either remain the same or decrease should be acceptable for full . credit due to the fact that at the bottom of the core diffusion length will decrease however at the e.op of the core little if any effect will be seen.
6.10 Add to answer key the scram discharge volume high level scram, Main Turbine Stop ' Valve closure, Main Turbine Control Valve fast closure, APRM fixed flux and IRM hi hi because these are not bypassed by the mode switch position alone.
7.3 Delete part 'B' completely. This procedure is entered from either Level Control or Cooldown Emergency Procedures and should not be entered directly. Entering - Level Restoration directly could result in bypassing important steps in the previously mentioned Emergency Procedures.
8.4 At SNPS six (6) conditions require the use of SBLC, any three (3) of 'these six should be acceptable. Three are contained as symptoms of SP 29.004.01-2 and the remaining three are contained in SP 29.024.01-4 Step 3.6.
8.8 When grading question, please take into account the fact that the question can be interpreted as either: 1) What does 3.0.4 mean? or ' 2) What does "3.0.4 not applicable" mean? Depending upon the interpretation the candidate took, the answer to 1) above should be, "The statement means that you can not enter any operational conditi<n with that LC0 in an action statements status", the answer to 2) above should be, ,
"The statement mFans that you can enter any operational condition with that LCO in an action statement status." In addition the definition of 3.0.4 should not be required since the question did net ask for it, tC \
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a , t ATTACHMENT 4 . The following represents the NRC resolution to those comments made by the facility as a result of the current exam review policy.
Only those comments resulting in significant changes to the master answer key, or were "not accepted" by the NRC, are listed and explained below. Comments made that were insignificant in nature and resolved to the satisfaction of . both the examiner and the licensee during the post exam review are not listed.
i.e.: typo errors, relative acceptable terms, minor set point changes.
/ 1.4 Not Accepted. If candidate arrives at correct answer, he/she will
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have to validate. All calculations will be graded.
2.2 Not Accepted Two correct. answers required for full credit.
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-2.6 (a) Not Accepted Question specifically asked for 1) the design t
function and 2) how it achieves its function. Both parts needed for full credit.
. 2.8L Accepted Only if candidate answers correct actual actions caused by W bus prog.
3.5(b) Not-Accepted Alternate answers considered during grading only in i addition to correct answers.
l t. 4.3 Accepted And/or crystallization graded.
4.4 Not Accepted Alternate answers considered for partial credit during
. grading.
"- 4.8 Accepted Candidate will not lose credit for these answers, only in s addition to correct answer.
4.11(a) Not Accepted Will consider Technical Specification requirement for partial credit.
(b) Not Accepted Question was not confusingly worded. Question specifically asked for three locations where temperature must be
,
maintained. Answer Key remains as is.
5.3 Not Accepted Question asked for inputs and outputs. Heat' generated; from the reactor must be considered.
' 5.7 (b) Not Accepted Decrease is t.he correct answer.
! 6.10 Not Accepted Question asked, "regardless of mode switch position", not regardless of mode switch position alone".
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'2 7.3 Not Accepted The NRC realizes the concern of bypassing procedural-l< steps in previous procedures. Question asked for procedural
}.' ' conditions, and answer is correct as stated.
L 8.4 Accepted Candidate will be required to-state the symptoms and the l' procedu.re they are addressing.
, 8.8 .Not Accepted Question explicitly asked,' "What is the meaning of. ! l specification 3.04, and What does NOT applicable mean in various '
- LCO's." The definitions (word for word) was not asked for, only the i meaning of.'it.
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