IR 05000443/1987019
| ML20236F816 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 10/19/1987 |
| From: | Dudley N, Keller R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236F791 | List: |
| References | |
| 50-443-87-19OL, NUDOCS 8711020306 | |
| Download: ML20236F816 (121) | |
Text
{{#Wiki_filter:-_ _ . _ _ ._ _ _ -. - _ o a i , . L U. S. NUCLEAR REGULATORY. COMMISSION' REGION'I , . OPERATOR LICENSING EXAMINATION-REPORT . . EXAMINATION. REPORT N0. -87-19-(0L) j.
' FACILITY DOCKET NO. 50-443 ' FACILITY LICENSE NO. NFP-56
LICENSEE: Public Service of New Hampshire P.
0.- Box 330 Manchester, New Hampshire 03105 ! , FACILITY: 'Seabrook Station o EXAMINATION DATES: August 10-14, 1987 and August 17-19, 1987- . CHIEF EXAMINER: W W' /M' M-
N. Dudley, Ledd Reac Engineer Date ) / [fN[ . APPROVED BY: . R. Keller, Chief Date l ' PWR Operations Section
, SUMMARY: Written and operating examinations were administered to thirteen senior reactor operator (SRO) and seven reactor operator (RO) candidates.
Thirteen SR0 and six R0 candidates passed these examinations.
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B711020306 871020 PDR ADOCK 05000443 V PDR ! ] i
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REPORT DETAILS TYPE OF EXAMS:. Replacement EXAMINATION RESULTS: l R0 l SRO l l Pass / Fail l Pass / Fail l l_ l I ' l l l l l Written i . 7/0 l 13/0 l l l l l l
1 I l Operating l 6/1 l 13/0 l
I I I I I I I l0verall l 6/1 l 13/0 l l l l
j 1.
CHIEF EXAMINER AT SITE: N. Dudley, Lead Reactor Engineer I 2.
OTHER EXAMINERS: F. Crescenzo, Resident Inspector ! E. Yachimiak, Reactor Engineer ! F. Jaggar, EG&G Idaho Inc.
l W. Hemming, EG&G Idaho Inc.
3.
The following is a summary of generic strengths or deficiencies noted from the grading of written examinations.
This information is being i provided to aid the licensee in upgrading license and requalification I training programs.
No licensee response is required.
STRENGTHS Individual and overall performance on the written examinations indicate a well trained group of candidates.
DEFICIENCIES a.
There are no learning objectives that define knowledge requirements I associated with the administrative aspects of the SRO position in , ' respect to the technical specifications and facility administrative l procedures.
l b.
Recent plant transients and clarifications to the administrative l procedures were not known by some candidates.
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Simulation Facility Fidelity Report.
During the conduct'of-the simulator portion of these operating tests, no performance and/or human factors discrepancies were observed.
5.
Personnel Present at Exit Interview: , NRC Personnel N. Dudley, Lead Reactor Engineer A. Cerne, Senior Resident Inspector Facility Personnel P. Richardson, Training Center Manager J. Grillo, Assistant Oper:tions Manager , R. Hanley, Operations Training Manager i L. Carlsen, Senior Simulator Instructor 6.
-Summary of NRC comments made at exit interview: The Chief Examiner reviewed the number and types of the examinations which had been administered, and stated, that due to the number of candidates, the results of.the examinations would not be available for five weeks. -The Chief Examiner informed the licensee that comments from i the facility's review of the written examination should be provided to j-the NRC within five working days of the date of the examination.
l The Chief Examiner noted that the Operations and Training Departments were professional and courteous during the examinations which contributed l to.the smooth conduct of the evaluations; that the simulator instructors l provided professional and responsive support during the simulator j . examinations; that there were no significant generic weaknesses noted during the operating examinations; and that some individual weaknesses had been discussed with the Training Department.
The licensee stated that simulator and plant walk through examinations I had been professionally conducted and that the facility appreciated the technique of " backing off" candidates during the simulator examinations.
The licensee stated that comments on the review of the written examination would be submitted to the NRC since there were some facility concerns about some questions on the Senior Reactor Operator examination.
7.
Evaluation made in accordance with TI 2515/79: During the simulator scenarios, the examiners evaluated the candidates' ability to satisfactorily implement the Emergency Operating Procedures (EOP's).
During emergency evolutions they were familiar with their i individual and team responsibilities, they were able to execute the E0P's with the minimum shift staff identified in the facility Technical ,
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Specifications, the candidates did not physically interfere with each other nor did they duplicate efforts (unless required), and they were able to transition from one E0P to another and to enter and exit as
required while assuring all necessary precautions were met and steps were completed.
8.
Licensee Action in Response to Examination Report No. 50-443-85-24 (0L): (Closed) Open Item (85-24-01) improvement to simulator malfunction 1tst.
During preparation for an October 14, 1985 licensing examination, NRC . examiners experienced difficulties in preparing simulator scenarios based .I on the documentation provided by the facility.
Since then, the facility has developed a Malfunction Cause and Effect Document.
The Document contains 169 malfunctions which are grouped according to system.
Each p malfunction has a title, description, severity options, initial alarms, reference procedures, expected plant response including strip chart recordings as appropriate, and expected operator response.
The Document includes a listing of 28 initial conditions.
Most components and valves can be overridden and a statement to this effect will be included in the Document This item is closed.
Attachments: 1.
Written Examination and Answer Key (RO) 2.
Written Examination and Answer Key (SRO) 3.
Facility Comments on Written Examinations after Facility Review 4.
NRC Response to Facility Comments i l ! )
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NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION ' FACILITY: _SEABRQQK_1______________ REACTOR TYPE: _PWR-WEC4________________ DATE ADMINISTERED: _g7fggfig________________ EXAMINER: YACHIMIAK E.
CANDIDATE: __ _Q _ h / _____ ,, V JNgIBUCIJgN@_19_C8NpJp@lEl Uso separate paper for the answers.
Write answers on one side only.
Steple question sheet on top of the answer sheets.
Points for each quostion are indicated in parentheses after the question.
The passing greda requires at least 70% in each category and a final grade of at locet 80%. Examination papers will be picked up six (6) hours after thn examination starts.
q % OF CATEGORY % OF CANDIDATE'S CATEGORY __YOLUE_ _lgl@L ___SCQBE___ _y@LUE__ ______________C@IEGQBl____________,, ~29t9@__ _29t@@ ________ 1.
PRINCIPLES OF NUCLEAR POWER , ___________ PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
! I _391@@__ _39 @@ ________ 2.
PLANT DESIGN INCLUDING SAFETY ___________ AND EMERGENCY SYSTEMS _29199__ _29199 ________ 3.
INSTRUMENTS AND CONTROLS i ___________ l _39199-_ _29199 ________ 4.
PROCEDURES - NORMAL, ABNORMAL, l ___________ EMERGENCY AND RADIOLOGICAL j CONTROL i
i ________x Totals j '199299__ ___________ Grade j Final
l All work done on this examination is my own.
I have neither given f nor received aid.
j j ___________________________________ Candidate's Signature
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NRC RULES'AND GUIDELINES FOR LICENSE EXAMINATIONS i , .g . i Duri'ng the administration of this examination the following rules apply: l 5.
Cheating on the examination means an automatic denial of your application ! cnd could result in more severe penalties.
R.
Rsotroom trips are to be limited and only one candidate at a time'may leave.
You must avoid all contacts with anyone outside the examination
room to avoid even the appearance or possibility of cheating.
2.' uss black ink or dark pencil gnly to facilitate legible reproductions.
S.
Print your name in the blank provided on the cover sheet of the examination.
B.
Fill in the date on the cover sheet of the examination (if necessary).
b.
Use only the paper provided for answers.
P.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
D.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a ngw page, write gnly gn gng sidg of the paper, and write "Last Page" on the last answer sheet.
2.- Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least thrgg lines between each answer.
Li. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
L2. Use abbreviations only if they are commonly used in facility literature.
13.'The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
L4. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
H5. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
hb. If parts of the examination are not clear as to intent, ask questions of the examiner only.
,7.
You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has
been completed.
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l10..Whcn you complete your examination, you shall ,. . cc.
Assemble your examination as follows: '( 1 ) - Exam questions on top.
j (2).
Exam aids - figures,: tables, etc.
' (3) Answer pages including figures which are part of the answer.
b.- ' Turn in ynur copy of.the examination and all pages used to answer.
! the examination questions.
t c.
Turn in all_ scrap paper and the balance of the paper that you did l not use for answering the questions.
j id.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still l in progress, your license may be denied or revoked.
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p- - 71.; Ffji;; PRINCIPLES OF NUCLEAR" POWER PLANT OPERATION .PAGE' '2.
LIdEbd99YNSU19EA_UEeI_IgeNgEEB_eND_E6UJp_ELgW l i - - ' p
E -QUESTION 1.01.
(.60)1 Choose the. answer which.most correctly completes the sentence: i'In the condensate system,1the operating point for two pumps operating in - '
- perallel will'be at --- (choose f rom below). --- as compared to the y
, opcrating point when one pump is operating and the other pump -is isolated" A.'the same. flow rate and the same head . HB, a higher flow rate and the same head C.Ja higher flow rate and a higher head D.
the same flow rate and a higher head
. QUESTION 1.02 (1.40) . l During plant operations, WHEN does the reactor vessel experience the createst-stresses and WHAT.are TWO parameters that can be controlled to . l i mi t these stresses? ! , i ! QUESTION 1.03 (2.75) ! e. List the -THREE-f actors that influence the ef f ective f uel temperature l les a' function of power AND state whether=their change over core life i INCREASES orLDECREASES the effective fuel temperature at 100% power.
(2.25) q WHAT is the change (INCREASE, DECREASE, NO CHANGE) in the DOPPLER POWER Tb.. COEFFICIENT over core life because of the overall change in effective ] fuel temperature 7 (0.50)
~ ' QUESTION 1.04 (2.40) ] l List the FOUR parameters that affect the Departure from Nucleate Boiling
Ratio (DNBR) and state whether an increase in that parameter' INCREASES or REDUCES the DNBR.
i l j ' QUESTION 1.05 (1.00) f- >; 'Upon a loss of offsite power, list THREE indications that natural . circulation has been established.
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QUESTION 1.06 (2.40) c. HOW and WHY does the, Moderator Temperature Coefficient (MTC) vary with temperature as the reactor is heated up from Mode 5 to Mode 17 (only consider the properties of the water) (1.20) b.
Briefly describe HOW and WHY MTC varies over core life.
(1,20) QUESTION 1.07 (2.20) a.
Would increases in reactor power be FASTER, SLOWER, or UNAFFECTED if the fraction of delayed neutrons created by the fission process INCREASED 7 Briefly explain WHY.
(1,10) b.
When a reactor trip from power occurs, WHAT steady state startup rate (SUR) can be expected AND WHERE do the neutrons that cause this SUR come from? (1.10) DUESTION 1.08 (2.20) A reactor trip from power has occured and the plant must be placed into Mode 5.
WHAT ane-TWO -p-i = v y p ar ameters wi4+ add positive reactivity to C' - the core, and WHAT action must be taken to account for these changes? QUESTION 1.09 (2.40) Compare the CALCULATED Estimated Critical Position (ECP) for a startup to bn performed 4 hours after a trip from steady state 100% power, to the ACTUAL control rod position if the f ollcwing events / conditions occur.
Consider each independently.
Limit your answer to RODS OUT MORE, RODS IN MORE, or RODS THE SAME than/as the CALCULATED ECP.
a.
RCS pressure is increased by 50 psi two minutes prior to criticality, b.
The startup is delayed until 8 hours after the trip.
c.
The steam dump pressure controller setpoint is increased to a value just below the steam generator ASDV setpoint.
d.
All steam generator levels are raised by 5% one minute prior to the ECP being reached.
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(--,,-- .- ' . - l . . ~PAGE . 4 ' l c.11 PRINCIPLES OF-NUCLEAR POWER PLANT-OPERATION1 2e.tBB55?BBBIDB :BE51:iBisi!iB:555:EEDID:EEDs (' ., , 's.
t o QUESTION 1.10 (1.50) ] c; Does INCREASING moderator temperature INCREASE, DECREASE, or NOT CHANGE the total worth'of control: rods? (0.50) b.
Briefly explain WHY bank overlap is used.
(1.00) { -QUESTION 1.11 (1.30).
) l The plant is in Mode 3 with pressure in one of the steam generators at 1000 i l poig.
Use steam tables to answer the following questions.
c.-WHAT is the steam temperature in the steam generator? (0.65) b.
If an' atmospheric steam dump valve (ASDV) is opened for one minute, l calculate the temperature of the discharged steam.. (assume discharge pressure is atmospheric and that the expansion process occurs at ! l constant enthalpy) (0.65) $ QUESTION 1.12 (2.25) A variable speed centrifugal pump is operating at 1/4 rated speed in a CLOSED system with the following parameters: 75 gpm Pump Flow = 50 KW Pump Power = , 300 psid
Pump Head = a.' Calculate the new values for these parameters when the pump speed is increased to 1/2 rated speed.
(1.80) b. HOW (INCREASES, DECREASES, NO CHANGE) does the available Net Positive Suction Head (NPSH) vary by an increase in system flow rate? (0.45) ! QUESTION 1.13 (1.80) { l A reactor is shutdown by a calculated amount of 5000 pcm.
HOW much boron j (in ppm) should be diluted to cause the countrate to double? Justify your
answer by calculation.
State any necessary assumptions and show your work.
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F ' J.# . .1 . . . . , . [lt._P6A l_QE@l@Njlyg(UQlNQiq@EE11,@NQ_EdER@ Edgy _gyg1Ed@ PAGE- '5- , 9? .. .. , n? > ] ^ , .... . LQUESTION 2.01 (.95) , WHAT;is-the. design basisifor' limiting the number of starts for large motors tduring a given-time period? List TWO.
< QUESTION 2.02 (2.80) a. L'ist'THREE Chemical and Volume Control (CVCS) valves that will automatically close when pressurizer level decreases to 17'/.. (0.90) b.
WHAT.is the function of the low-pressure letdown ~ pressure control valve.
' -(CS-PCV-131) during both normal.and solid plant operations? 10.70) ] ll. ba ) ' c.'If a' safety injection "S" isolation signal closes the centrifugal charging pumps minimum flow isolation valves (CS-V-196, CS-V-197), WHAT d' j are'the TWO conditions (including setpoints) that will automatically . cycle these valves lto the.open AND close positions.
(1. 993
( 0. 7 0) t QUESTION 2.03-(2.80) \\ -a. 'ListLTWO' safe'ty-related components which would cause degradation and/or i ' failure of their respective systems'upon a loss.of service water flow.
(0.80)- i ' b.
WHAT conditions.are required to initiate a train's "TA" signal? Provide appropriate setpoints.
(0.80) . c.
Describe-FOURLautomatic actions which take place upon the actuation of a l train's- "TA" signal.
(1.20) -QUESTION' 2.04 (3.00) a. Briefly explain HOW 4.16 kVAC Bus E5 supplies 125 VDC Bus 11A during normal' plant operations.
Include only major components.
(1.40) l b.
WHAT'are'FOUR emergency diesel electrical (EDE) system components that , would be'affected by a loss of 125 VDC Bus llA? (1.60) i (***** CATEGORY 02 CONTINUED ON NEXT PAGE
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! QUESTION 2.05 ( 3.~ 40 )
- c.-ListLFOUR paths'by.which seal injection flow can-leave the RCP shaft
.cea1Hassembly and> indicate WHERE each path.goes.
(2.40) b(,.Briefly explain'HOW the No. 1-RCP seal is cooled and lubricated if seal injection. flow is11ost.
-(1.00) ' QUESTION 2.06 (2.80)
- a.c Brief1'y explain WHY the containment spray system is needed during a Loss Of Coolant Accident or a Main Steam Line Break.
(1.30)
- b.J WHAT. are the minimum and maximum sump PH limits AND explain the basis.
- for each.
(1.50) QUESTION 2.07 (2.25)- For the following questions use Attachment 2A.
-
-a.LWHY is there an interlockJto prevent safety injection pump (SIP)
miniflow isolation valves (SI-V-89,90,93) and RH system to CCP/SI j ! isolation valves (RH-V-35 and 36) fromfall being open at the same time 7 (0.75) _ , ! b.'In WHAT position must SI-V-89,90,93-be aligned before RH-V-35 can be i opened? (0.75) c. WHY is component cooling water (CCW) used for the SIPS? (0.75)- l
) . QUESTION 2.08 (1.70) a.'WHAT is the normal maximum flow allowed by the steam generator EFW j supply ~ header flow control valves (FV-4214 thru FV-4244 A/B)? (0.40) J oC b.
Briefly describe HOW '(including setpLonts) a pipe rupture in the EFW c9 supply header, downstream of flow control valves FV-4214 A/B, results in f.
.the closure of both valves.
(0.80) ] l .c.
If both flow control valves fail to close on a pipe rupture, WHAT limits EFW flow through the break? (0.50) (***** CATEGORY 02 CONTINUED ON NEXT PAGE
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FAGE .L.__t%dLkuslGN_,[h,2 j_@da; s i,,,QbQ _gQgdQgu jl_,Q h 3.g _, , QyE.STIdN 2.09 (2.50) a. HOW long will an emergency diesel generatcr run at rated conditions if the fuel' oil l storage tank AND day tank are at their design basis levels? Consider each separately.
, ' (1.00) l b. WHAT are the THREE conditions which will automatically trip the emergency diesel gen-l erator af ter being emergency started? (D0 NOT include emergency stop) (1.50) ' ! l QUESTION 2.10 (2.80) a. Describe the flow path of containment air used for the control rod drive mechanism (CRDM) cooling system.
Include all major components.
(1.20) b. Describe the-TWO functions of the containment recirculating filter (CRF) (1.60) system and WHEN it is utilized.
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'_s . ! . l < l l QUESTION. 3.01 13.00) l l
'For each case below,.briefly explain HOW reactor coolant system (RCS) j temperature is. controlled by the steam dump system and/or: atmospheric steam i dump' valves (ASDVs), and. indicate the approximate final RCS Tavg.
Assume j _ i c11 systems are normal,'except as stated, and that no operator action is j teken.
Consider each case separately AND list the interlocks or control l cyetems1which.will actuate for each case.
c. The normal setpoint on the steam dump pressure controller (PK-507) is . reduced by 85 psi while in Mode 3 awaiting reactor startup.
! b.
The Train A steam dump interlock switch is taken to "OFF" with the plant-stable at 5% reactor power.
d l l c. With reactor power at 100%,.the pressure setpoint on the steam dump ! . controller (PK-507) is raised to 1200 psig and the steam dump control mode selector switch is placed in the steam pressure position.. At the / l came time, a reactor trip occurs.
! ' l l ' QUESTION 3.02 (3.00) j
a.
Driefly explain HOW neutrons produce detector current in a source range j (SR) excore nuclear instrumentation (NI)' detector.
b.
WHEN would an undercompensated intermediate range (IR) excore detector affect plant operations? Briefly describe HOW.
c.
WHEN would an overcompensated IR excore detector affect plant operations? Briefly describe HOW.
QUESTION 3.03 (2.80) For the following protective actions, state its blocking setpoint (if applicable), its trip setpoint, and its coincidence, a.
Source range high flux trip b.
Intermediate range high flux trip ' c. Power range high flux rate trip d.
Power range low power trip o.
Pressurizer high' level trip (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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' .. QUESTION-3.04 . (2.50) Procsurizer pressure control (PPC) channel 455 is selected for master-prcosure control and PPC channel 458 is selected for backup pressure control.
The plant is. in Mode 1 at 100% power.
//58 c. List THREE valves or. interlocks that thi-s PPC channel" controls in Cp automatic.
(1.50) v b.
WHAT are TWO PPC component responses that would occur immediately after PPC channel 455 f ailed HIGH7 (1.00) ' QUESTION 3.05 (2.40) A caf ety injection "S" signal can be initiated automatically by THREE nignals.
List these signals their setpoints, and coincidences.
, QUESTION 3.06 (3.40) For each of the following situations, indicate whether or not control rod motion (insert or withdrawal) is blocked AND justify your answer.
Assume rods are in automatic and NO operator action is taken, other than specified balow.
Consider each part separately.
e. Turbine load setpoint is increased from 15% to 25% reactor power.
b.
Reactor power is at 90% when turbine impulse pressure channel PT-505 fails low.
c.
Reactor power is at 75% when power range excore detector N-43 fails high.
d.
Reactor power is at 100% when reactor coolant system loop "A" narrow range Tc fail high.
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QUESTION 3.07 (2.90) c. List the TWO signals that generate a control room ventilation i sol ati on cignal.
(1.00) ' 6.' WHAT FOUR' automatic actions' occur in the control center HVAC (CBA) System as a result of a control. room ventilation Train A isolation si gnal ? Assume the system is aligned for automatic operation.
(1.40) c.
Once isolated, WHAT. local operator action must be performed to realign-the CBA system to draw air from the noncontaminated air intake 7 (0.50) i QUESTION 3.08 (2.50) ' ' ! a. List TWO initial automatic system actions that would be indicative of pressurizer level control (PLC) channel 459 failing HIGH.
Assume channel 459 is the controlling channel, the plant is at 100% power with all systems in automatic, and NO operator action i s taken.
(1.00) f 6.
List THREE initial automatic system actions that would occur if PLC channel 459 failed LOW.
Assume channel 459 is the controlling channel, the plant is at 100% power with all systems in automatic, and NO ' operator action is taken.
(1.50) QUESTION 3.09 (2.50) a. State the reason WHY the RC loop suction valves, RC-V-22,23,87,88 have open and close interlocks and list both interlock setpoints. (1.20) b.
Briefly explain WHY a failure of either RCS pressure detector PT-403 or PT-405 does not prevent both RHR trains from being isolated from the RCS.
(0.50) c.
In order to manually open RWST suction valve CBS-V-2, which TWO valves l must first be closed in order to meet the control logic interlocks? i (0.80) l l l (***** END OF CATEGORY 03
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. QUESTION' 4.01 (1.50) -c.
Procedure ON1233.01, Partial Loss of Vacuum, contains the following CAUTION statement: " Load-reduction for the purpose of restoring condenser vacuum should not go below 360 MWe if condenser vacuum remains greater than 5 in. Hg.
A.
(25 in. Hg. VAC)."
Briefly explain the basis for this cauti on.
(0.50) 6.
Attempts to restore condenser vacuum have not been successful.
Load has been reduced to 360 MWe and condenser vacuum is 41nne than 5 in. Hg. A.
( , ^ WHAT actions is the operator required to perform? - (1.00) l , i-QUEPTION 4.02 (2.00) In. Procedure FR-H.1, Response to a Loss of Secondary Heat Sink, steps 17 through 19 are a continuous loop until feed is reestablished to steam gnnarators (SGs).
WHAT are the FOUR systems which an operator should use in en attempt to establish a secondary heat sink while in this loop? ! QUESTION 4.03 (3.00) Tho plant has experienced a large break LOCA snd procedure E-1, Loss of Racctor or Secondary Coolant, has just been entered.
a.
State the reactor coolant pump trip criteria.
(1.00) b.
WHAT are the FOUR criteria (include applicable setpoints) that need to be met in order to reduce ECCS flow? (2.00) QUESTION 4.04 (1.00) a..WHAT must be done to any batch type, li quid ef fluent waste bef ore sampling can be performed? l b.
WHO has the authorization to allow the release of liquid effluent waste I I to proceed? l
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RADIOLOGICAL CONTROL
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< .. . QUESTION 4.05 (3.00) c. State THREE symptoms or entry conditions for procedure 051202.04, Rapid Boration.
(1.50) b.
List TWO acceptable sources of borated water during rapid boration.
(0.50) c.
The plant is in Mode 5.
Briefly describe HOW rapid boration is , performed.
(1.00) QUESTION 4.06' (3.00) For the f ollowing 4aws questions, assume that the plant is at 40% power with c11 systems in their normal line up for this power level.
o.
If Primary compor nt cooling ' water (PCCW) loop A flow is lost, list.TWO [7 reactor coolant pump.(RCP) related alarms that the operator can expect to actuate (within approximately two minutes) due to this failure.
b.
WHAT is the operator response if PCCW flow can not be recovered? 4/st TWO.
c.
If RCP shaf t vibration alarm actuates and shaf t vibration is 16 mils and increasing, WHAT TWO actions is'the operator required to take? Q!)ESTION 4.07 (2.50) A reactor trip from 100% power has just occured.
c. WHAT is the response of the operator when a reactor trip can not be verified? (0.50) b.
The above action was not effective and procedure FR-S.1, Response to Nuclear Power Generation /ATWS, is entered.
List THREE indications which are used to verify that a reactor trip has NOT occured.
(1.50) c.. WHAT IMMEDI ATE ACTION is to be performed by the operator if the reactor can not be verified tripped (while in FR-S.1)? (0.50) l l l l (***** CATEGORY 04 CONTINUED ON NEXT PAGE
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l l
U
__ ' 4i_,PQQgEQUBE@_;_MO80@(t_Q@@OBd@(t_EUEQGENCX_6MQ.
PAGE
i RAD 40 LOGICAL CONTROL
- '
,., .' QUESTION 4.08 (1.00) c. WHAT is a reactor operator's 10 CFR 20 whole body exposure limit? b.
WHAT'is the annual administrative whole body limit,without extensions? QUESTION 4.09 (1.80)- Inntrument air pressure low alarm actuates with air pressure at 95 psig and idccreasing.
Per procedure ON1242.01, Loss of Instrument Air, c. Briefly explain WHY and HOW the reactor would automatically trip if no operator action is taken.
(1.20) b.
What effect (open, close, no effect) will the failure of instrument air have on MS-V-127 and MS-V-128, loop 1/2 steam supply valves to emergency feed pumps? (0.60) !
-QUESTION 4.10 (2.30) .The following questions deal with the Emergency Operating Procedures' usage rul sst. i ! a.
WHAT is the significance of bullets (o) when used to designate ACTION / EXPECTED RESPONSE subtasks? , b.
HOW is the operator made aware of tasks that must be f ully completed before proceeding to another instruction? c.
ARE CAUTION statements from E-0 still applicable after transitioning to E-17 d.
The Symptomatic Response / Unexpected Conditions page for ES-1. 2 i s applicable for which procedures? l l I QUESTION 4.11 (1.70) a. WHAT are TWO radiation alarms that would be indicative of a steam generator tube leak? (1.10) 6.
WHY is it necessary to re-route the condenser vacuuni pump discharge to ' the PAD if a tube leak exists? (0.60) (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) t - . _ _ _ _ _ _ _ _ _. _. _ _ _ _. _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _. _
. _ _ l' 4.
'P69CgpuRgp_ _N9RMBL _BgNpRMgL _gMgRggNCY_gNp PAGE
2
RADIOLOGICAL _ggNTRQL
+' .- l . ' ' ' . QUESTION 4.12 (2.20) Antwer the following questions concerning the precautions and limitations of procedure 0S1000.02, Plant Startup f r'om Hot Standby to Minimum Load.
c. HOW is the RCS differential boron concentration between the pressurizer cnd reactor coolant loops limited to less than 50 ppm? C b.
WHAT is the minemum reactor coolant system temperature allowed while in C4 Modes 1 and 2? U c.
WHAT is the maximum rate of heatup for tne reactor coolant system? d.
At WHAT power level do penalty minutes begin to accumulate? (***** END OF CATEGORY 04
- )
(************* END OF EXAMINATION
- )
. _ _ _ _. _ _ _. -. _ _. _ _ _ _ _ _ - - _ _ _ _. - _ _ _ -
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V,
- Cycle efficiency * (Netkorx-W h
'l-out)/(Energy in) ,
-
- s-= V t + 1/2 at w.
g.
. d E = me a ~ l
- KE = 1/2 mv a = (Vf - V )/t A = AN A=Ae o
g PE = ogn q V s'V,.+ at-w = a/t .A= in2/t1/2 = 0.693/t1/2 f t eff = [(tifp)(t)3 ] 1/2 h .. [(tl/2) + (t Il j b i = 931 mi J I'= 1 a * I o' . . .Q = mCp at q (j = UAat I*Iec Pwe = u sh i=L 10**I # f n j TYL = 1.3/u ! sur(t) HVL = -0.693/n P = P 10 P.= P e*/
'SUR = 26.06/T SCR = S/(1 - K,ff) CR = S/(1 - Xeffx) x l ' SUR = - 26a / t* + ( s - o )T CR;(1 - K,ffj) = CR (I ~ eff2)
T = ( t*/a ) - + [( a - o ) / o ] M = 1/(1 - X,ff) = CR /CR j g T = t/ ( o - a ) M = (1 - K,ffg)/(1 - Keffl) T ='(6 - o)/(lo) SDM = (1 - K,ff)/K,ff 10-5 seconds a = (X,ff-l)/Keff " #Keff/K t* = eff T = 0.1 seconds-I eff (1 + T)] o = [(t*/(T K,ff)] + [s / I d) =Id P = (r+V)/(3 x 1010 ) I d) 2,2 2 l gd j
2 I = oN R/hr = (0.5 CE)/d (meters) R/hr = 6 CE/d2 (feet) Water Parameters Miscellaneous Conversions I gal. = 8.345 lbm.
I curie = 3.7 x 1010dps l ga]. = 3.78 liters 1 kg = 2.21 lem = 7.48 gal.
I hp = 2.54 x 1 Stu/hr 1 ft4 , Oensity = 62.4 lbm/ft3 1 m = 3.41 x 1 Stu/hr l
lin = 2.54 cm i Density = 1 gm/cm Heat of vaporization = 970 Btu /lom
- F = 9/5*C * 32 l
Heat of fusion = 144 8tu/lbm
- C = 5/9 (*F-32)
{ l Atm = 14.7 psi = 29.9 in. Hg.
1 BT1J = 778 f t-lbf ' 1 ft. H 0 = 0.4335 lbf/in.2 p - _ _ - - - - - _. _ _ _ - _ - _ - -.. _. _. -. -. .. _ -. - -. . _ - - - - _ .J _
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- ISEddQQ1N851Q@i_dE81_IB@NSEEB_@NQ_E(ylQ_E(QW~
' , ? hdWh 9 E-A SEABROOK 1" -87/08/18-YACHIMIAK,;E..
- !3
. , -- -ANSWER; .1. 01.
('.'60 )
- C-C O. 603 >
l REFERENCE
- NRF;' Course Endbling Objective
- 36.l'-14.
~ LWastinghouse Thermal-Hydraulic' Principles'II pageL10-46' K/AL191004 K1.09.2.4 191005K109' ...(KA'S)' ANSWER 1.02' ( 1. 40), Lduring plant cooldown;CO.603 . temperature CO.403. 04 M M ' C o,Vo] ipressure CO.403, ' C7 . V-REFERENCE . . ' L ' NRF Course Ensbling Objectives'39.1-9.,29.1-10.
-
~ ' l Westinghouse' Thermal-Hydraulic Principles II page'13-67 l K/A-193010~K1.07 3.8 L 193010K107' ...(KA'S) - l 1'y ! ' ANSWER 1.03 (2.75) c.
. gap thermal 1 conductivity reduction from fission product gases CO.403 INCREASES CO.353 - fuel'densification'CO.403 INCREASES CO.353 . clad creep _CO.403 pf Co.Vo] q , _ ' DECREASES CO.353 1b. DECREASE'CO.503 i REFERENCE NRF Course Enabling Objectives 21.1-22.,21.1-23.
, Wzotinghouse Reactor Core Control for Large PWRs pages 2-45,2-46
K/AL192004 K1.07 2.9 192004K107 ...(KA*S) I j r - - - - - - - - - - - - _ _ _ _ ) .
_ ___ __ 'l __PglNCIPLEQ_QE_NQCLE@B_PQWEB_P(@N1_QPE8@l[QN PAGE
t THERMODYNAMIQQt_HE@T_ TRANSFER _ANQ_FLQIQ_ELQW ' AtjSWERS -- SEABROOK 1-87/08/18-YACHIMIAK, E.
ANSWER 1.04 (2.40) - roactor power E0.303 REDUCES CO.303 - RCS temperature E0.30] REDUCES E0.303 - RCS pressure E0.303 INCREASES [0.303 - RCS flow EO.30] INCREASES E0.303 REFERENCE L Simulator Course Enabling Objective 26.0-3.
W stinghouse Thermal-Hydraulic Principles II pages 13-23,13-24 ah[0 c r dhetr eaJlg K/A 193008 K1.05 3.4 193008K105 ...(KA'S) g e ()fe$ J d h TO f CO,G( - c s ep (i $dumM h CfC MgouM CO'0 d ANSWER O
- a constant or decreasing delta T across the reactor core less than the full load delta T E0.603 M Loop dT O ddecded2 C o.40] (g - core outlet temperatures are constant or decreasing E0.603 CA - a constant steam generator level with a constant feed rate eN' a constant d or decreasing steam pressure CO.603 04( a&m REFERENCE y NRF Course Enabling Objective 40.1-7.
Wantinghouse Thermal-Hydraulic Principles II page 14-27 K/A 193008 K1.22 4.2 193008K122 ...(KA*S) ANSWER 1.06 (2.40) c. MTC becomes more negative at higher temperature E0.603 because water's density change per degree of temperature change increases at higher temperatures E0.603 b.
MTC becomes more negative over core life C0.603 because of the { dscreasing l baron concentration (due to fuel burnup and the production of fission k product poisons) [0.603 REFERENCE NRF Course Enabling Objectives 22.0-7.,22.0-8,22.1-11.,24.0-10.
Westinghouse Reactor Core Control for Large PWRs pages 3-18,3-28 K/A 192004 K1.06 3.1 K/A 192007 K1.04 3.1
J
lit __Eh[NQlELES_QE_NQQ(EE8_EQWE8_EL@MI_QEE8@llQMt.
PAGE 17' ISE8dQQ1G@diqS _dE@l_18@NSEE8_@NQ_E(QLQ_E(QW' t kNNW S - SEABROOK 1L-87/08/18-YACHIMIAK, E.
~ ANSWER 1.07-(2.20) , , . bM - o.. SLOWER'[0.503 more delayed neutrons would result in a longer average fission" neutron ... lifetime resulting in a slowing down of the fission rate [0.603, d73 '3 b.
-1/3 DPM-CO.503 V-the longest-lived delayed neutron' precursors.CO.603 REFERENCE NRF' Course Enabling Objectives 18.0-7.,18.1-18.
, ' Westinghouse-Fundamentals of Nuclear Reactor Physics pages 7-30,7-68 K/A 192003 K1.07 3.0.
K/A'192003 K1.08 2.8 ANSWER.
1.08 (2.20) . --- x enon ' C O. 60 3 -!RCS temperature CO.603 Cp
- -the shutdown rods must be-fully' withdrawn C1.003 OR @MM DO v
REFERENCE-
NRF Cobrse Enabling Objective 28.0-1.
%MQ O, o6] ' W2stinghouse Reactor Core Control for Large PWRs page 7-14-K/A 192002 K1.14-3.8' 192005H;114 ...(KA'S) ANSWER 1.09 (2.40) . a.
RODS THE SAME b.
RODS OUT MORE [0.60 X 43 c.
RODS.OUT MORE d.
RODS IN MORE REFERENCE NRF Course Enabling Objectives 22.1-18.,23.0-1.
W2stinghouse Reactor Core Control for Large PWRs apages 3-15,3-38,7-28 thru 7-33 K/A 192004 K1.13 2.9 'K/A'192006~K1 10 3.1 , [ '
. . l - - _ _ _. _ _ _ _ _ _ - - _ - _ _ _ _
_ L -ic__P61NClP(ES_OE_NyC(E@g_POWEQ_P_(@N1_OEE6@llON PAGE
t THERMODYNAMICS _HEAI_IB@NSFE8_ANQ_FLUlQ_ FLOW t L ,< i l AT WERS -- SEABROOK 1-87/08/18-YACHIMIAK, E.
,
ANSWER 1.10 (1.50) ([MneGA) c.
INCREASE E0.50] ' b.
Bank overlap provides a more uniform DRW which ensures that rod motion produces a change in reactivity OR provides a more uniform axial flux distribution during control rod maneuvers [1.003 REFERENCE NRF Course Enabling Objectives 25.0-7.,25.0-8.
j W;ctinghouse Reactor Core Control for Large PWRs ) pcg3s 6-14 thru 6-16, 6-28 thru 6-29 K/A 192005 K1.07 2.5-K/A 192005 K1.09 2.8 ANSWER 1.11 (1.30) c. 546 F.
(+/- 2 F.)
E0.653 b.
300 F.
(+/- 2_EL ) [0.653 ' REFERENCE NRF Course Enabling Objectives 29.1-11,29.1-14.
Storm Tables K/A 193003 K1.14 2.4 K/A 193003 K1.25 3.3 ANSWER 1.12 (2.25) (N1/N2)^1 = Q1/02 [0.303 (0.25/0.50)^1 = 75.0/Q2 Q2 = 150 gpm [0.303 (N1/N2)^3 = P1/P2 [0.303 (0.25/0.50)^3 = 50/P2 P2 = 400 KW E0.303 (N1/N2)^2 = H1/H2 E0.303 (0.25/0.50)^2 = 300/H2 H2 = 1200 psid [0.303 ., b.
DECREASE [0.453 REFERENCE NRF Course Enabling Objectives 36.1-13.,36.1-15.
Wsetinghouse Thermal-Hydraulic Principles II pages 10-39,10-44 K/A 191004 K1.05 2.3 K/A 191004 K1.15 2.6 191004K105 191004K115 ...(KA'S) ' i _ _ _ _ - - _ _ _ _ _ _ - - _ _ - - _ _ _ _ _ _ _ _ - _ _ - _,
_ _ _ _ - _.
. - -. - - - - -- - -. 11.
- PRINCIPLES 3OF-' NUCLEAR ' POWER' PLANT OPERATIONg PAGE
'19' I 9; ;INE6099.LN8DJQgA_ BgSLIB8NgEgB_8Np_ ELUl p_ ELgW , u _ ;;.py. - . t. ANSWERS ~'-A SEABROOKL1-87/08/18-YACHIMIAK, E.
^
- 1 7
,- I s - ANSWER, I 1.13 -l (1.80) ' .; c,ccume>-10 pcm/ ppm-[0.303
y ! '
- .' C R 1 l'
K1 - 1 '.. - K 1 5000 pcm [0.503 --- =- '----- - = CR2 l' ~ K2 .CR1- ,2500 pcm!'[0.503 -- = 5000/1 - K2 'l --K2 - = 7.. - CR2< $(2500 pcm)/(10 pcm/ ppm)-= 250 ppm boron to be diluted [0.503 REFERENCE-NRF' Course. Enabling Objectives' 17. 0-7.',24. 0- 4., 26.1 - 12. - .Wantinghouse Reactor Core Contro1 for Large PWRs pages 5-3,9-10 . .K/A'192008'K1.'03 3.9 ' 192008K103- -...(KA'S) l' L l ^ , L.
- - _ .. . 2.
.____:-______-_--_________-___-_________--
_ -_ - _ rf2i_le(ANI_ DES {GN[[NC(QDlNG_q@ EELY _@ND_EUEBGENC1_Q1SIEdg.
PAGE
JANQWERSL--"SEABROOK 1-87/08/18-YACHIM1AK, E.- ,. . ANSWER.
2.01 (.95) . , . g
- provent motor winding damageL(overheating) CC.00] OR
du; to th: high motor starting current Ger454 CO.95) f-pg REFERENCE p ' NRF Course-Enabling Objective 59.0-5.
Dateiled Systems. Text page CD-20
- K/A 191005 K1.04 2.7 K/A 191005 K1.05 2.8 l
K/A 191005 K1.06 3.0 l
- 191005K104.
191005K105 191005K106- ...(KA'S) ANSWER' 2.02 (2.82) I a. CS-LCV-459, CS-LCV-460, CS-V-145 CO.30 X 33 Co.403 b.,-during normal operations maintains a crr.rt=n* pressure at the outlet of.
the~ letdown flow control valves Per60J and prevents the letdown flow ~from flashing to steam downstream of these valves [-G.3G3-(p.Vo3 during solid plant operations controls RCS pressure by varying the [ amount of letdown flow from the residual heat removal system Ger5B4 . CO. V0,) C, f ' c yc.
d le.open if CCP flow is less than 80 gpm 40.500 then reclose when flow o - f increases above 120 gpm CBrSO.1 (D.JS]'
REFERENCE Datailed Systems' Enabling Objectives 19.0-6.,19.0-10.
'Datailed' Systems Text pages CS-23,CS-81 thru CS-85 K/A.004000 K1.01 3.6 K/A 004010 KS.05 3.8 K/A 004010 K6.06 2.7 K/A 004020 K6.12 2.9
- 004000K101 004010K505 004010K606 004020K612
...(KA'S) -_ _-_ - - - - - - - - - - _ - - - - _ - - - - - - _ _ _ _
L __EL8NI_DEgJgN_JNC6!LDJNg_g3Egly_3ND_gd@SgENgy_ gyp]EUS PAGE
{ A,NSWERS -- 3EABROOK 1-87/08/18-YACHIMIAK, E.
{ i ANSWER 2.03 (2.80)
c._- component cooling water heat exchanger (CC HX E-17A/B) [0.403 ' L - diesel generator water jacket heat exchanger (DGWJ HX E-42A/B) [0.403
b.
cervice water discharge pressure < 22.5 psig CO.403 for atleast 3 raeconds [0.20] when atleast c,ne pump has been running 30 seconds [0.203 I c.
- service water pumps are tripped a-d '-d:d cut E0.303 - cooling tower pump and ion start [0.303 - supply end etu-" valves align for cooling tower E0.303 ' [ -- secondary heat loads are solated CO.303 l serare v.h em a ocM M cc>.sa3 Q/Ko.303 \\ - REFERE NCE '
Dstalled Systems Enabling Objectives 33.0-2.,33.0-7.,33.0-8.,33.0-9.
Dsteiled Systems Text pages SW-7,SW-11,SW-12 / . K/A 076000 K1.01 3.4
coolle twe/2. -feEI' dN hgry go[UtyC(oJeCo.fG j Wll./ fwra.
FWJ 31 orb Co 3 3 ' K/A 076000 K1.05 3.0 - K/A 076000 K3.07 3.7 r c to r 81 elae r o /,'y 71 A t' Co o // # ) ~[b uee 1 Co.FcQ - 076000K101 076000K105 076000K307 ...(KA'S) V fMo/y berf l000?? al' 'n Co. 5 Q
valVe Cfo3GJ Co.to3 J etyic e wu $ e G.
dif C ayc - ANSWER 2.04 (3.00) I a.
power from 4.16 kVAC Bus E5 is transformed down to 480 VAC CO.403 and powers battery charger (BC-1AJ CO.40] which converts AC to DC CO.203 which then feeds 125 VDC. Bus 11A CO.403 b.
- diesel engine control circuits
- static exciter / voltage regulator 00.40 X 43 I - diesel generator output breaker - emergency power sequencer REFERENCE Datailed Systems Enabling Objectives 7.0-1.,14.2-28.
Dstailed Systems Text pages EDC-8,EDE-2 j - K/A 062000 K1.03 3.5 i K/A 063000 K3.01 3.7 ) 062000K103 063000K301 ...(KA'S) { J
i 2i__P(AUT_Dgg]Qy_Jyghgg]Ug_g8ggly_8Bg_gyggggygy_gygIgyS 'PAGE
l
l l ' ( ANSWERS.-- SEABROOK 1-87/08/18-YACHIMIAK, E.. [ L ., - i i
ANSWER 2.05 (3.40) a.
- down the shaf t CO.30] to the impeller (casing) E0.30] l - No. 1 seal leakoff CO.30] CH charging pump suction [0.30] oR VCT C030)
- No. 2 seal ~ leakof f CO.30] reactor coolant drain tank (RCDT) [0.30] l - No. 3 seal leakoff CO.303 containment sump E0.303 I b.
flow reverses up along the shaft from the impeller (casing) to the No. 1 l geal CO.503.and-is cooled by the thermal barrier heat exchanger 00.503 REFERENCE Dsteiled Systems Enabling Objective 15.3-36.
Dateiled Systems Text pages RCP-20 thru RCP-25 K/A 003000 K1.03 3.3 K/A 003000 K4.04 2.8 003000K103 003000K404 ...(KA'S) ANSWER 2.06 (2.80) e.
- limits peak containment pressure C0.403 and temperature E0.403 - reduces the containment atmosphere iodine concentration below its design limit [0.503 b.
- minimum B.0 [0.25] keeps stripped iodine in solution E0.503 - maximum 10.5 E0.253 prevents caustic corrosion [0.503 REFERENCE Dsteiled Systems Enabling Objectives 31.0-1.,30.1-7.,30.1-8.
Dstailed Systemc Text pages CBS-3,CBS-7 K/A 026000 K4.02 3.1 K/A 026000 K4.04 3.7 026000K402 026000K404 ...(KA'S) ANSWER 2.07 (2.25) a.
to prevent a loss of water form the containment sumps to the RWST CO.753 b.
SI-V-89 and 90 closed wi+h "T-U-93 coen 4Mb or h6b covnhwd/on; regottech SI-V-93 closed wi u. E l - V-69 ana ve open [0.753 g pg/ cfpf'.y } c.
cool lubricating oil [0.753 REFERENCE Datailed Systems Enabling Objectives 30.1-14.,30.1-17.
Datailed Systems Text pages ECC-21,ECC-44 thru ECC-46 K/A 006000 K4.06 3.9 K/A 006000 K1.11 2.8 006000K111 006000K406 ...(KA'S)
1
_ _ __ - L2 ;_g69NI_Dgg1GN;JNC(yg}NQ_@@EgIy_9ND_ggggggNCy_Sy@ Igg @ PAGE. 23
, ANSWERS --LSEABROOK 1 '-87 / 08/18-YACHIM1HK, t:.. . l'of ,y ANS ER 2.08' (1.70) . . g a 'c.. 235 gpm-(+15 -10.gpm)LCO.'403., vec e b. oach. flow orifice transmitter (train A and B) sends a close signal to
its respective valve CO.503 at a flow of. greater than 450 gpm-[0.303 n c.
an in-line ventur.1 CO.503 " REFERENCE 061000A301 #2L061000A304 M/ ...(KA'S) f2 0 - f, f2,0 -/Q DeTIVU r,psWents Wgt ggy,7 ANSWER 2.09 (2.50) a.. storage tank'- 7 days [0.503 ~ day tank - 3 hours.CO.503 'b.
- overspeed - low lube oil pressure' [0.50 X 33 - generator differential lockout ' REFERENCE- .Datailed Systems Enabling Objectives 13.0-9.,13.1-15.,13.1-17.
. Detailed. Systems Text pages EDM-22,EDM-37 K/A 064000 K1.03 3.6 t K/A 064000 K4.02 3.9 064000K103 064000K402 ....(KA'S)
1 ANSWER 2.10 (2.80) a. containment air is drawn into the CRDM cooling shroud through vents
. CO.303 where it flows around the CRDMs up through connecting nozzles to a~ ring header [0.303 the air is drawn up by four vane. axial fans [0.303 end discharged through backdraft dampers into containment CO.303 b.
-- filters' the containment atmospheregprior to personnel entry E0.40] ) eft:r -~;...al plant ope-=tia-e E 9 A@F ' ' C o. Yo] l - reduces post-LOCA E0.403. hydrogen concentrations below design limits (4%) CO.403 l REFERENCE ( ~D2 tailed Systems Enabling Objectives 34.0-1.,34.0-2.
-Datailed Systems Text pages CHV-11,CHV-12 K/A 022000 K4.04 2.8 .K/A 022000 GO.07 3.3 022000G007 022000K404 ...(KA'S) ! ! L I
i i - e.
_ ._ __
1...lyp]SyDEyl@_8yp_gOyJROLp PAGE
' ANSWERS'-- SEABROOK.1; -87/08/18-YACHIMIAK, E.
p 3 eA s 3. o f (p1 ' ANSWER 3.01 (3.00)' c.! (the normal; steam pressure setpoint of 1092 psig maintains Tavg at "557 F.)
a decrease in the setpoint to 1007 psig would cause the dumps to open and cool Tavg to approximately.550 F..CO.50] where the P-12 interlock ' (low-low Tavg) would close all steam dumps [0.503 b.. steam. dump operation would be blocked 1[0.30], secondary pressure would rise to the setpoint of.the ASDVs which would maintain pressure at 1125 psig CO.403 causing primary temperature to steady out at 561'F. [0.303 c.
in pressure control mode, the output of the plant-trip and load-reduction controllers is blocked E0.303. ' Temperature would be-initi 11y be controlled at 561 F.
[0.303, corresponding to the ASDV f-setpoi nt, =til d o... dump valve op s, e t.. er, could P o r s= t en.n er = + 're ' - - is1a-E
- * * * * * ~
r ^ L- -S k k' Aenn f YIO ok bal k4D b* $ Yew YQ Cc & j)t2/)b(V12 REFERENCE @- Dateiled Systems Enabling Objectives 44.1-13.,44.1-14.,44.1-15.. Co. VQ Stosm Tables-Datailed' Systems Text pages SD-17 thru SD-19 K/A 041020 K3.02 3.8-041020K302 ...(KA'S)
Jefko CosQ ANSWER 3.02 (3.00) Oh c.:a neutron and a boron yield an io zed (+) lithium nucleus and an ' ionized (+) alpha particle CO.53 these ions create additional ion pairs CO.25] which migrate to the detector's charged electrodes E0.25] b.
during' shutdown-[0.503 source range (SR) NIs will not automatically ene.gize CO.503 , "c.-during startup [0.50] ) improper IR/SR overlap (may result in a reactor trip) CO.503 ) l REFERENCE
. Oa
. - Dstalled Systems Enabling Obj ective 26.0-9.
Dstalled Systems Text pages NI-7 thru NI-9,NI-12 - /g.
g K/A 015000'K5.01 2.9 K/A 015000 K5.02 2.7 Qg 015000K501 015000K502 ...(KA'S) g #fr wil 7d
- gd>e (wferoVff(;7h-CO.Sd)
~
__ __ _ ___ ___ _ __ - _ '9___JN@J690gNIg_OND_CgNISQLE-PAGE
AN5WERS -- SEABROOK 1-87/08/18-YACHIMIAK, E.. ,
- e
, I ANSWER 3.03 (2.80) a.
IR 10E-10' amps (P-6) [0.303, 1ES CPS E0.203, 1/2 CO.103 b.
10% reactor power (P-10) [0.303, 25% reactor power [0.203, 1/2 [0,103- {{'t c.
+/- 5% in.2 seconds CO.303, 2/4 [0.103 'C7 M TC C4tk N# '4'*i d.
10% reactor power (P-10) - [0. 30 3, 25% reactor power [0.203, 2/4 [0.103 o.
10% turbi ne power Jfb-1tM" [0. 30 3, 92% CO.203, 2/3 E0.103 c' p,13 ) or andet .(p.,) (p,9 s[ p.,o e REFERENCE D2teiled Systems Enabling Objectives 26.0-8.,26.0-9.
.Dateil ed Systems Text pages NI-47, NI-48, PPLC-36 thru PPLC-39 K/A 015000 K1.01 4.1 K/A 015000 A4.03 3.8 K/A 011000 K1.04 3.8-011000K104 015000A403 015000K101 ...(KA'S) ANSWER 3.04 (2.50) a.
pzr PORV PCV-4564 - pzr PORV PCV-4568 CO.50 X 33 - pzr block valve V-122 .b.
heaters turn off E0.503 mpray valves go full open CO.503 REFERENCE Datailed Systems Enabling Objectives 22.0-7.,22.0-9.
Detailed Systems Text pages PPLC-17 thru PPLC-19,PPLC-43 K/A 010000 K4.03 3.8 010000K403 ...(KA*S) ANSWER 3.05 (2.40) - high containment pressure (high-1) CO.503 2/3 [0.1634.3 psig [0.203 - low pressurizer (pzr) pressure E0.50] 2/4 C O.10 3 4fMK4 p si g [0.203 - low steam line pressure CO.503 E6& E0.103 585 psig [0.203 {y 2l3 Is75 (/ REFERENCE Datciled Systems Enabling Objective 54.0-10.
Dateiled Systems Text pages IS-22,IS-23,IS-33 K/A 013000 K1.01 4.2 013000K101 ...(KA'S) S _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ - - _ _ _ _ _ - _ _ _ - _ _ _ _ $7.- lUEl.6ydEN,1@_QN_D_QONIBOL@ PAGE
(\\NSWERS -- SEABROOK 1-87/08/18-YACHIMIAK, E.
7f c-s
- ckqMco.ss
' [o.S ANSWER 3.06 (3.40) O S , c. no CO.30] OS P-10 was blocked at 10% power CO.553 039 b.
yes CO.303 _.. power indicates below P-13 causing an auto rod withdrawal block CO.553 c.,per* C 0. 30 3 783 c,_Z, {o,SsQ L tu t'h $$ m, o,iy c Juicu lm = Ju out anput tc rad t, lock logic CO.553 d.
no 00.303 2 of 4 OT delta T channels high (within 3% of setpoint) are needed to cause a rod withdrawal block C0.553 ! REFERENCE Dateiled Systems Enabling Objective 27.1-13.
! ' Datailed Systems Text page CP-28 K/A'001000 K4.10 3.2 001000K410 ...(KA'S) ANSWER 3.07 (2.90) W14 'M VO. { - a.
- radiation detected at a remote air intake structure 00.503-X O'6 , - generation of a ESFAS "S" signal 00.503 l b.
- ventilation supply fan deenergizes (FN-27A) { - the non-associated discharge damper closes (DP-53B) CO.35 X 43- - the emergency cleanup fan starts (FN-16A) - M aOJ foA/ f//pf[[g-6) _the associated damper opens (DP-27A) c.
the manual isolation valve f or the contaminated intake must be closed. CO.503 REFERENCE Dateiled Systems Enabling Objectives 35.0-5,35.0-6.
D3 tailed Systems Text pages CAC-27 thru CAC-30,CAC-47 K/A 072000 K2.04 3.3 ] 072000K204 ...(KA'S) ! l
l l _ - -- - - - - -
. _ _ _ _ _ - - - _ _ _ _ _ _ - _ _ _ - - _ - _ - _ _ 1. l.NQI69Og,NI@ JN_D_ CONT _BOLQ PAGE 27.
- -
_... e: ,I ' ,oANSWERS;--SEABROOK;i-07/08/.18-YACHIMIAK,.E.
d
- ..
[ ANSWER - 3. 08 ' (2.50) l i c.?- charging flow to' minimum E0.503 ! .. backup heaters on [0.503 j . b.. - letdown 'isol ati on
- - charging' flow to maximum [0.50 X 33-l s heaters cutout ] '
REFERENCE
Dateiled Systems. Enabling-Objectives 22.0-9.,22.0-13.
- Datailed' Systems Text page PPLC-44
'K/A 011000 A2.10 3.4 K/A~011000 A2.11 3.' 4 d-j 011000A210 011000A211: ...(KA'S) j 'l ANSWER.
.'3. 0 9 ' (2.50)
c'.. isolate.the~1ow pressure RH. system from the RCS during normal plant operations.CO.63- ~ open - 365 psig [0.303 close - 660 psig CO.303 b.1the non-failed. detector sends a'close signal to a' RC loop suction - valve.
h inlboth loops.'(either-upstream-or. downstream) [0.503 c. RH-V35 (RH-CS/SI cross connect). CO.403 CBS-V8 (containment recirculation. sump suction) [0.403' REFERENCE 'Datailed. Systems Enabling Objective 29.0-3.
Datailed-Systems Text pages-RH-20,RH-22 K/A 005000 K4.07 3.2 005000K407 ...(KA'S) g' hoj77 6h hh oppin, so a)Ub!'C
- '
W~Y# $~W o wO ut go(afe a,1 c h*b af f# l [ i _:_=__-___--____-___. __ - _ _ _ -. _ _ _ _ _ _ _ -. _ ___ _ - -
. _ _ _ __._ _ __ _______ - _- - - - ~ d L _ _P 60 gE D yBE q _; _N O gd@( t _@@N O BU@(,, _EDE6 G E NC 1_@N D PAGE
-RADIOLOGIG@L_QQNT8QL . s ' i' j ANSWERS -- SEABROOK 1-87/08/18-YACHIMIAK, E.
l
.! ANSWER 4.01-(1.50) ! l l o.
Iow pressure turbine blade protection against overheating CO.503 ! ! b.
trip the reactor E0.53 OR DbIR CD's0] f trip the turbine EO.53 . REFERENCE (I _ Si mul at or Enabling Objective 63.1-7.
{ Dateiled Systems Text page TSS-29 { Procedure ON1233.01
K/A 000051 A2.02 3.9 C00051A202 ...(KA'S) ANSWER 4.02 (2.00) c. - emergency feedwater --Condensate CO.50 X 43 -- startup f eed pump - Fire pump REFERENCE Datciled. Systems Enabling Objective 39.0-3.
Procedure FR-H.1 , K/A 000054 EK3.04 4.4 000054K304 ...(KA*S) ANSWER 4.03 (3.00) fCSsubcoolingics(dion 2. y co, eo3 hace
< 40 degrees 46.303 c.
atleast one charging or SI pump running CU.DWJ l b.
- RCS subcooled'> 40 degrees based on core exit TCs j - total feed flow > 500 gpm OR one narrow range SG 1evel > 5% i - RCS pressure stable or increasing E0.50 X 43 - pzr level > 4HH'. 6 "fu (f' , REFERENCE Procedure E-1 K/A 000011 EA1.03 4.0 K/A 000011 EA2.11 3.9 E00011A103 000011A211 ...(KA'S) l l
_ $t__P8QQEDUgEg_ _ NOB 096t_@@yOBd@6t_EUE8QENQ1_ANQ PAGE
60 Dig 6QQLQQ6_gDNIBOL .e . ANSWERS -- SEABROOK 1-87/08/18-YACHIMIAK, E.
ANSWER 4.04 (1.00) c.
it must be recirculated CO.503 b.
USS or SS CO.50] REFERENCE Datciled Systems Enabling Objective 61.0-6.
! Administrative Procedure CP 4.1, Effluent Surveillance Program, pages 4,5 K/A 068000 A2.02 2.7 K/A 068000 GO.01 2.7 068000A202 068000G001 ...(KA'S) ANSWER 4.05 (3.00) c. - inadaquate shutdown margin (SDM) unexplained or uncontrolled reactivity addition (neutron flux - increasing, SDM monitor alarm, high flux at shutdown alarm) - failure of more than one shutdown or control rod to insert fully during a shutdown CO.50 X 33 b.
BAT CO.25] RWST CO.25] c.
BAT gravity feed valves (CS-V437,439,447,1207) to CCP suction open CO.3]
VCT outlet valves (LCV-112B/C) closed CO.30] charging path lined up (HCV-182,CS-V142,143,180,177 open) CO.303 / CCP started CO.103 REFERENCE Simulator Enabling Objectives 66.0-3.,66.0-4.
Procedure OS1202.04 K/A 000024 EK3.01 4.1 K/A 000024 EK3.02 4.2 000024K301 000024K302 ...(KA*S) yf harf hsrlC QCl /*0 71 (E fVM)0 D3*D open C s V - F'u G 'O CM 3 yep 6 borott'an {low C0J') qN r a_. s ecp rn.m s - -- - ! i i l
u___________.___ _ _ _ _ _ _ _ __
. _ _ --.. 3:_;Pggggggggg_;_NggggL _ggNpgg3L _ggggggNCy_9Np PAGE
I
2 RADIOLOGICAL CONTROL L 'B' l . 'AN9WERS -- SEABROOK 1-87/08/18-YACHIMIAK, E.
, l ANSWER "4.06 (3.00) c. high stator winding temperature [0.500 oA loSC o[ @bu b M P *k 8 high thrust bearing temperature E-tW5EKI ,t/ g g g c g(e gg b.
trip the reactor E0.503 trip the affected RCPs CO.503 [ Nf' l// btqTl6?1 Qgg c.
stop the RCP CO.503 .] chutdown the plant to MODE 3 [0.303 wi hin 1 hour [0.203 < et nucle fbc$ Med Aust'A. M 0 o 9- +df %, RX co,so3 ( L{ absumption REFERENCE
Simulator Enabling Objectives 68.0-4.,68.1-11.,68.1-15.
Procedure'OS1201.01, RCP Malfunction K/A 000007 EA1.04 3.6 h f 000007A104 ...(KA'S) ANSWER 4.07 (2.50) c. manually trip the reactor CO.503 all rod bottom lights not lit b.
- - reactor trip and bypass breakers not open CO.50 X 33 - neutron flux not decreasing ' c.
manually insert control rods [0.503 REFERENCE Procedure FR-S.1 K/A 000029 EA1.09 4.0 K/A 000029 EA1,14 4.2 000029A109 000029A114 ...(KA'S) i ANSWER-4.08 (1.00) l l-o.
3 Rem /Qtr [0.503 l b.,5000 mrem [0.503-REFERENCE NRF Enabling Objective 13.1-7.
Procedure RP 5.1 K/A 194001 K1.03 2.8 194001K103 ...(KA'S) ) ! .. _ _ _ _ _ _ _ _. _. _. _
__-_ _ _, ' Lii__E60QEDQ6E5_ _NQ80@(g_@BNOQU@Lg_EDEQGENC1_6NQ-
- PAGE. 31 y ' '
,.6@DIOLOGICAL CONTROL .<xe . JANSWERS -- SEAORDOK'1 '-87/08/18-YACHIMIAK,E.
.
.. [ I l.
JANSWER 4.09 - ( l '. 80 ) ' h 'c..Lf eed' ater ' control valves would f ail closed.[0;603 causing a reactor trip w f . on : steam generator. l o-l o 'l evel CO.603 b.. no of f ect. (valve normally open) [0.603 O R. - fdo . .O CO. 6 0] iREFERENCE gh.. Simulator Enabling Objective 77.0-2.
N
- d
.ck ' Procedure ON1242.01 /' K/A.000065 EA2.06.3.6 -
M!R E NC-f. M2' M 000065A206 ...(KA'S)
_! .
i ANSWER 4.10 (2.30) id ec. cequence of perf ormance is not: important'[0.603 ' ' Jb.'the step containing the taskoor an associated NOTE will explicitly state the requirement CO.603 c.,ysMr C 0. 50 3 ' /V O {' ~d.LAtt procedural, series [0.603 . REFERENCE
.Wantinghouse Owners Group Emergency Response Guidelines, l Executive Volume Users Guide K/A 194001 A1'.02 3.9
194001A102 ...(KA'S) l l ANSWER 4.11 (1.70) I _ _ b {o s u do vr) ('acl ink /ol' a fGf *1 g o,gg]. l a.
- condenser vacuum pump effluent alarm E-9-550 - ! - main steam line alarm E0.55J 'b.
minimize of f site release OR minimize plant' contamination CO.603
- -l REFERENCE ) Detailed Systems Enabling Objectives 73.1-12.,73.1-14.
Procedure'OS1223.02, " Steam Generator Tube Leak" K/A 073000 GO.07.3.4 K/A 073000 GO.10 4.2 073000G007-073000G010 ...(KA'S) .1 l l l l a-_---_-_______--___.._._:--- . _ _ _ _. _ . _ - _ _. _ ._.
_ - . - _. _ _
- _ _ _ _ _ -. _ _ _. _, R 4:__PggQEDURES'-' NORMAL _gBNgRMBL _gMgRggNgY_gNp PAGE 32-
2 609196991906_99NIBQL .t.
. . , ' . ANSWERS -- SEABROOK 1-87/08/18-YACHIMIAK, E.
! l l l ANSWER 4.12 (2.20) c. pressurizer sprays should be utilized [0.553
b., 551 degrees F.
CO.553 c'. 50 degrees F/Hr [0.553 l d.
15*/. power [0.553
REFERENCE Simulator Course Enabling Objectives 13.0-2.,13.1-11.
Procedur e 0S1000.02 . K/A 002000 GO.05 3.6 I K/A 002000 GO.10 3.4 CO2000G005 002000G010 ...(KA'S)
! ! ! ! ! _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _. _ __ __ _ .
_ _ _ _ _ _ . TEST CROSS REFERENCE PAGE-1 ,_,.* 4.
, l QUES.TICN' VALUE REFERENCE ____ at_ ______ __________ , l 01.01 .60 EAY0001180 01.02 1.40 EAY0001181.
'01.03 2.75 EAY0001167 l 01.04 .2.40 EAY0001175 , . 01.C5 1.80 EAY0001179 j II: 01.06 '2.40 EAY0001174 I L 01.07 2.20 EAY0001170 l 01.C8 2.20 EAY0001173 01.C9 2.40 EAY0001169 01.10 1.50 EAY0001168 01.11 1.30 EAY0001176 01.12 2.25 EAY0001177 01.13 1.80 EAY0001171 ______ -24.00- QgCO C C' 02.01- .95 EAY0001203 '02.02 2.80 EAY0001204 02.03 2.80 EAY0001205 02.04 3.00 EAY0001207 02.05 3.40 EAY0001208 02.06 2.80 EAY0001209 02.07 2.25 EAY0001210-02.08 1.70 EAY0001212 R02.09 2.50 EAY0001213 02.10-2.80 EAY0001214 l ______ 25.00 03.01 3.00 EAY0001193 03.02 3.00 EAY0001194 03.03 2.80 EAY0001195 03.04 2.50 EAY0001197 03.05 2.40 EAY0001198 03.06 3.40 EAY0001199 03.07 2.90 EAY0001200 03.00 2.50 EAY0001201 03.09 2.50 EAY0001202 ______ 25.00 04.01 1.50 EAY0001182 04.02 2.00 EAY0001183 .04.03 3.00 EAY0001185 04.04 1.00 EAY0001187 04.05 3.00 EAY0001188 04.06 3.00 EAY0001189 04.07 2.50 EAY0001190 04.08 1.00 EAY0001191 04.09 1.80 EAY0001192
l - _ _ _ _ _ _.
, _ _ - _ _ TEST CROSS REFERENCE PAGE
] ,g i QUES.T T VALUE REFERENCE i ---, --
04.10 2.30 EAY0001253 ~ j
- T4.11 1.70 EAY0001260 l
04.12 2.20 EAY0001261 j l
25.00
..---
100.00
, , .1 ! ! l l
i l ! ! i
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ j
. . _._.
. _. II T T S C H M E N T Q ' ~ ~ > U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SEABROOK 1 REACTOR TYPE: _EEB-WEC4 ' DATE ADMINISTERED: 87/08/18 EXAMINER: _ D.U_ D L E Y. N. CANDIDATE: MS E R > l INSTRUCTIONS.lQ_9aNDIDAIE_1 Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of.at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00..25.00 5.
THEORY OF NUCLEAR. POWER PLANT OPERATION, FLUIDS, AND { THERMODYNAMICS , 25.00 25.00 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00 25.00 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 25.00 25.00 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100.00 % Totals Final Grade All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature l _._ _-______________a
. - _ _ -_ _ _ - _ ____ _ _ _ _ _ _.
. ... l NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1.
Cheating on the examination means an automatic denial of your application I and could result in more severe penalties.
. 2.
Restroom trips are to be limited and only one candidate'at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil 9nlZ to facilitate legible reproductions.
! 4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
l l 7.
Print your name in the upper right-hand corner of the first page of 2ach section of the answer sheet.
8.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one sidg of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
q l 17. You must sign the statement on the cover sheet that indicates that the ) work is your own and you have not received or been given assistance in i i completing the examination.
This must be done after the examination has l-been completed.
i
i
- _ - _ - _ - .. .
- w
>18._When.you complete'.your examination, you shall: , a.
. Assemble your; examination as<follows: .(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) ' Answer pages? including figures which are part of the answer.
,.. Turn 'n.your copy of-the examination and all;pages used to answer 'b.
i - the examination questions.
c '. Turn in all scrap paper and.the balance ~offthe paper that.you~did .not use.for' answering the questions.
d '.' Leave the examination _ area,.-as defined by.the examiner.
If after leaving, you are'found in this' area-while~the' examination is still, in' progress, your license may be denied or revoked.
L- . , , _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ __ __ _m___ __ _ _ _ _ _ _. _ _ _ _ _ _ . _ _ _.. _ _. _ _ _
- .. .. - - _ _. - - - - - - -_ . . . 5, 1 THEORY OFNUC[ EAR POWER PLANT OPERATION.'FLUIDEl_6ED PAGE.
'2- "2 TEEBMODYNAMICS: L ' i i QUESTION. 5.01-(2.00).
< iA reactor has the.following characteristics at.100%. power: Tave~is 573.5 F l and Tstm.is'513.8 F.. Assume that the plant!is shutdown.and that some j L steam generator tubes are plugged:so that only 95% of the original' heat- . transfer area is available.
If the-plant returns'to 100% power.and Tave is again 573.5:F,fdetermine!the pressure of-the steam leaving the steam . I ' generator.
Show all work: including assumptions made and equations used.
l . l QUESTION' 5.02 (2.00) i a. The plant is operating.at 85% power with rod controlfin manual"and all I other control systems-in. automatic.
.The operator inadvertently aligns
. charging pump suction to the RWST.
Explain how and why shutdown- ' margin (SDM) is affected by this action, b.-If'aLeontrol rod was stuck completely out of the-core on.a reactor- ' trip,:how would.SDM required by Technical-Specifications differ-from the normal case where all rods are. inserted: on a trip? d-l-h QUESTION 5.03 (2.40)' ' Listed below'are conditions which will affect a fuel loading 1/M Plot.
' For'each condition listed, indicate whether the number of fuel elements predicted.for criticality will be OVER or UNDER the number of fuel
elements required for actual criticality.
Provide a explanation for { -each answer.
- a. A fuel assembly is loaded near the detector.
l - b. The detector is too far from the newly installed fuel.
- c.
The detector is too close to the source.
-d'. The-' detector.is too far from the source.
I l ! l ! I i (***** CATEGORY 05 CONTINUED ON NEXT'PAGE *****) = _ - _ _ _ _ _ - _ _ _ _ - _ _ - - _ _ _ _ _ _ . ._ . - _ _ _ - _ _ -_.
- - _ - - _ --
, _. _. _ - _ _ _ - _. _ _ _ _ _ _ - __ - W ~ > d@1 - !THEQBY'OF NUCLEAR' POWER PLANT OPERATIONi-FLUIDS.-AND-PAGE-13.
' J5... ..,. IIIERMODYNAMICS ' , 11: . ! f- ,,e 1, ' l r ' (. QUES' TION (5'. 0'4L '(2.50)L a.: Describe HOW (Increase,.-Decrease,.or Remain'the Same) and WHY core . Delta T will respond tola loss of natural circulation flow,Ccaus'ed lbyfthe steam dump valves /failing closed,.following a'. reactor-trip'. Efrom-100% power.
(0.75): b.-Explain how Teold wide range and steam. generator pressure wi'll . respond.to'a loss of-natural circulation flow,-caused by;a' loss of~ .. i m feed flow, following.a reactorLtrip'from 100%xpower? (0.75) -c. How~would the operator change the'following parameters (Increase,
. Decrease, or:No Change) in order to' aid natural circulation.
- -i Consider each separately.
(1.0) 1. PressurizerLlevel at SM :. w , ' ' 2.RCScooldownrateat40\\F/(t%WR tr -
g 3. ' Steam generator lev.el ata S v t' c4. RCS pressure ath~;300 paQt ', a: , ' ,,- . {3 y) ' + c g n ,
V
i-QUESTION-.5.05 j(2 (Q). t
7g _ 'i, < > -3 How'(Increase, Decrrese,or Remain the Same) and why, would-each.of the following paramet9.r,phangen affect DNBR7-
,s, S, i ( ' ?- ' ' da. %pssurhaer #vmpe'rature Increases 5 degrees .} f- . , 3 b.
Mass floy ratd ir,"the core Increases-10%' f t 'c., AFD,incistpef tio +10% j
.! '~ N' ' t, 'l ' p (, \\ f '-p g- .a
- l VQUESTION,530 (
(2.00) ' ' . ( ' ~ a. -.Of.the_ coefficients that contribute to'the power defect, which<% i ! , 4s .,ppefficient contributes most to the change in power defect over, I core life and~what causes tbs-change? x \\ b bf 'the coefficient'that h'6 tribute to the power defect, which . coefficient sets first to' affect :teactivity: on a sudden power i A ~chana)andphydoesthecoefficient;reactbeforetheothers.
I o t ,4 g- .. s ) . 2-s
' ; t.
I . t i s v s - 1y - ' > - 'y . I .
u .s,A 't ) .i % j e %- ; i ./ [ ! >Q , Al \\ i t ! , , $ . g , m , /*- ,,,\\ V v ' ( , .A (***** CATEGORY 05, CONTINUED ON NEXT PAGE *****) l +- ( $ ., U _'.---_.._-__ _.-- _ b - ^ ..
l L L '5! 'THEQRY'OF NUCLEAB_POWEB_ PLANT OEEBATlON, FLUID L 6HD PAGE 4' EEBMQDH6MLQ1 -
QUESTION 5.07 (3.00) Consider the following plant conditions:
MODE 3, BOL.
Boron concentration is 900 ppm All~ shutdown banks withdrawn l-Actual reactivity present in the core is minus 4% delta-K/K ! Source range indication of.100 CPS ! Differential boron worth is minus 10 pcm/ ppm A boron dilution to 750 ppm increases the source range indication to 132 CPS.
During the dilution, Xenon concentration has changed.
How ' many PCM of reactivity did xenon contribute during the dilution? State all equations used and. assumptions made and show all work.
QUESTION 5.08 (2.00) .y Compare and explain the difference in the reactivity worth of a rod that , is dropped while at power to the reactivity worth of the-same rod' stuck l out while all the other rods are inserted.
! QUESTION 5.09 (1.50) ' '.. . - Explain how the flow element venturi-type flow restrictor will act to limit EFW flow if a line break occurs downstream of the venturi.
~ '4
- h i
QUESTION. 5.10 (1.60) l [', ANSWERLthe following TRUE or FALSE.
ei... Equilibrium xenon concentration at 50% power is one half of the i ~ equilibrium xenon concentration at 100% power.
! i tw b.. Xenon concentration initially increases after a reactor trip.
{ l . " c.
Final samarium concentration after a trip is a~ function of the l-previous power level, j v d. Final xenon concentration after a trip is a function i of the previous power level.
l l j (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l
x ,
_ _ _ _. k ' 5.
THEORY OF NUCLEAR POWER PLANT'OPERATIOH. FLUIDS. AND-PAGE
[ IHERMODYNAMLQS. ! . QUESTION 5.11 (2.00) A centrifugal charging pump is running with the discharge flow control . valve FCV.121 in mid position.
Indicate how each parameter will change ! (Increase, Decrease, or Remain the Same) if the discharge valve is fully opened.
a.
Discharge flow b.
Pump discharge pressure upstream of the discharge valve c.
Motor amps d.
Available NPSH to pump e.
Seal injection flow (Assume seal injection flow control valve is in manual) ) QUESTION 5.12 (2.00) How will the following parameters change (Increase, Decrease, or No Change) if one main steam isolation valve closes with the plant at 25% load.
Assume all controls are in automatic and that no trip occurs.
a. Affected loop steam generator level (initial change only) b. Affected loop cold leg temperature i c. Unaffected loop steam generator level (initial change only) '{ d. Unaffected loop steam generator pressure j e.
Unaffected loop cold leg temperature ] ! I
1
(***** END OF CATEGORY 05 *****) i
- - - _. ._ _ '6 PLANT SYSTEMS DESIGN. CONTROL'. AND INSTRUMENTATION PAGE
. QUESTION 6.01 (1.50) Assume a normal plant cooldown is in effect with the RCS.at 330 F and' 625 psig.
What is the effect, if any, of a low failure of a cold leg
RTD (T-413B) on RCS pressure? List the components that will be affected.
l - QUESTION 6.02 (2.50) The-plant is at 30% power, and all control systems are in automatic.
Explain the plant responses leading to a reactor trip if first stage . pressure transmitter.(PT-505) fails high at the beginning of life.
N i QUESTION 6.03 (2.80) a.
If the temperature controllers for the PCCW heat exchangers are ajusted to significantly increase the PCCW temperature, indicate how the following parameters will change (Increase, Decrease, or Remain the Same): (1.2) 1.
Thermal barrier-f]ow 2.
VCT temperature 3.
PCCW flow to the letdown heat exchanger , 4.
Surge tank level-j b. In the event of a thermal barrier leak, what two system design l features, both active and passive, will prevent the spread of ' reactor coolant throughout the cooling water system? (1.1) c. Circle the five valves on Figure 3.2 that will remain open if a low low' head tank level is reached.
(0.5) I (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) -__ - _ _ ___ _ _ _ _ - _
. _.
_ ___ -__-__ ______ _ _ - _.
, i FIGURE CC-3.2 e ' PRIMARY COMPONENT COOLING SYSTEM LOOP "B" Cp') ~ "g _=; f,7 g - @ 0. n g. -E so *a m to Ltvit ~ ,, aa't i-t I ' =. ..a . , h C6JM "N J, 4.G446 CC 8.tO ,y tr,s.3 @ CLOW one F lag 4A6 On meag tase. to seweg s =4 i _
. @ e6ou on a seas on sao,*== 6o to 6 ve6 l e..,
- c.
..... l
e. o......
,_, i... , FIGURE CC - 3.2 ,,,,,,,,,,, o PRIMARY COMPONENT ' ~ COOLING SYSTEM LOOP"B" ' sa'='"'c -'l ~ I l . (12/85) tr.
dr;m, .t -j - _,,, i
9 l , g - .- ,-, ,, .; - l Pr.fd f- -i r!== l
a a l m:': l-l =f., l,$, , ' " * ~
- I ** c-l
---{ f,i:l*Ja e,.l I = '" F l:C,,,~f,. a,,,; o
- , ., d E*E ~ ! ..C q Y '
g,,..,7,.,. i l = t l l
- g.. ~. ;
'" . --I:=. A l l --.! a ,2 i --{ ::= l
l =' ~c'- ! ! o I % l .. .. .., l . ! < r.pg CC-8 2/88 _ _ _ _ _ _ _
_ _ _ - _ _ _ _- . 6.
PLANT SYSTEMS DESIGN.' CONTROL. AND INSTRUMENTATION PAGE
~ . QUESTION 6.04' (2.60) ! c. Reactor Trip Breaker A is opened for RPS testing during Mode 1.
If Bypass Breaker B is closed prior to closing Reactor Trip Breaker A, . what actions,-if any, will occur? (1.0) b.'A spurious reactor trip signal occurs with the plant at 100% power, however Reactor Trip Breaker B fails to open.
What will be the effect, if any, on each of the following: (1.6) 1.
Turbine Generator 2.
Feedwater isolation valves 3.
Steam dumps 4.
SI reset [if SI occurs] QUESTION 6.05.
(3.00) a. How would RCP seal water return flow and RCP lower bearing water i temperature' change as a result of a failure of a RCP's number one seal? (1.2) i i b. What actions, if any, should be taken regarding RCS leakage and f reactor power, if a RCP seal leak:does occur? (1.8) QUESTION 6.06 (2.10) . What is the fail position of the following valves (Fail closed, Fail open, < Fail as is) as caused by a. loss of contol air.
l a. Condenser steam dump (MS-PV-3016) b. Feedwater regulating valve (FCV-510) l c. Charging Line to Regenative HX, Seal water flow controller (CS-HCV-182)
d. Charging-flow control valve (CS-FCV-121) l e. RHR Hx bypass flow control valve (FCV-618) f. PZR spray valve (RC-PVC-445A) g.
PCCW A Supply to Contm. (supply to RCP) (CC-V168) i QUESTION 6.07 (3.00) The plant is at 100% when a small break LOCA occurs.
What changes, if-any, occur in the main feedwater and the EFW systems as a result of the signals generated by the small break LOCA? Component numbers are not { required.
J (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) l
_ _ _ _ _ _ _ _ _ _ _ _ _ _ ._ J
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'
. . . 6.
PLANT SYSTEMS _ DESIGN. CONTROL. AND INSIRUMENTATIOH-PAGE
j l
I . QUESTION. 6.08 (2.50) { The plant is operating at 80% power when a bypass manifold Thot RTD ] (411A) fails high.
How, if at all, will this failure affect the following items.
Consider each item independently.
Assume no operator action and all control systems are in automatic.
a. Rod insertion limit:setpoint b. Charging flow (initially) l c.
Control rod bank position l d. Steam dump control system e. Overtemperature Delta-T trip setpoint '
QUESTION 6.09 (2.00) a. During a reactor startup prior to reaching 10 E-10 amps, a reactor trip will occur due to a loss of a single 120 VAC instrument bus.
What is the cause of the trip? l b. How, if at all, would a safety related 120 VAC bus alignment change l if a UPS AC output breaker trips open due to an electrical overload? l QUESTION 6.10 .(3.00) a.
If a Loss'of Power accident is in progress and the EDG output breaker is shut, what will be the sequence of stripping and energizing loads > if.a subsequent SI signal occurs? b. Under what conditions would the Remote Manual Bypass (RMO) switch associated with the EDG control circuitry be taken to the " bypass" l position? c. How is an emergency stop of the EDG activated in the Main Control Room? , d r-Wha-tr-THREE-oon d i t i on s /s i gn n 1 = n11nu tha Enn tn npnvnta thenngh al l,
.tlip-conditions r-except-overspee d, lov cil pressure.-and-generator r differential Icekout? s OELETED (***** END OF CATEGORY 06 *****) - - - _ _ _ _ _ _ _ _ _ - - _ _ _ _ - ,
o.
, . b-L '7.
PROCEDUEES - NORMAL, ABHQBMAL. EMERGENCY AND PAGE
' l' BADIOLo01 CAL CONTROL i QUESTION 7.01 (1.20) Match the class of fires listed below (A - D) with the materials involved (1 - 4).
CLASS OF FIRE MATERIALS INVOLVED A.
Alpha 1.
Flammable liquids, gases, or greases B.
Bravo 2.
Combustible metals ' C.
Charlie 3.
Ordinary combustibles (paper, wood, etc.)
D.
Delta 4.
Energized electrical equipment QUESTION 7.02 (2.00)' In order to maintain the plant at 100% power, work must be performed inside the containment in a radiation field of 400 mrem /hr gamma and 2.0 Rem /hr fast neutron.
The maintenance man selected is 24 years old and has a lifetime exposure through last quarter of 28 Rem on his NRC Form 4.
How long may.the man be permitted to work in this area per facility administrative guidelines without an exposure extension? l ! ! ! I ! l ! ! ! ! (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) _ _ _ _ - - _ _ _ _ _ _ _ _ _
_- ._ _ - - _ _ ._-_; Y PROCEQQRES --NORMAL'. ASHQBMAL. EMERGENCY AND PAGE
' - B6DIOLOGICAL CONTROL _ "
i QUESTION' 7.03'. (3.40) j i ' i l During'anLemergency condition.the STA reports the following: j 1.' Core Cooling - Orange Path I 2 '. Suboriticality'- Yellow Path 3. Integrity - Orange Path 4. Heat Sink - Red Path-l a. In what order should the above conditions be addressed? (1.0) b. Ten minutes later the-STA' reports that the.above conditions'still exist except'that Core Cooling is on a Red Path.
What actions should .lus taken and why? (1.2) - c. With the above conditions still present,'what optimal or functional: recovery procedure should the'_ operator follow,.if a loss of offsite occurs and both diesel generators fail to start?
- (10.6)-
d. After a Reactor Trip, what two circumstances initiate monitoring the.
Status Trees? (0.6)
QUESTION 7.04 (1.50) ' Answer the following questions regarding E-0, Reactor Trip or Safety Injection: a. What action should be.taken if the Turbine Generator does not . ' manually trip? (0.8) b. Why must the $lectrical output of the Turbine Generator be less than 100 MW before opening the generator breakers? (0.7) ! , (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
_ -- 7.
PROCEDURES - NORMAL. ABNORMAL ' EMERGENCY AND.
_PAGE
RADIOLOGICAL CONTBQL . . QUESTION' 7.05 (3.00) The' plant is in~ Mode 5.
RCS temperature is 100 F and the pressurizer is-vented to atmosphere via a small-vent (RC-V468).
While conducting a routine surveillance, the CRO inadvertently resets train A Low Steamline Pressure SI instead of-train A Feedwater Isolation.
Generally describe the operator actions <necessary to recover from the ensuing plant transient.
Aarume the.following equipment status.
PORV A is inoperable RHR pump B is in Full to Lock (PTL) RHR pump A is operating-CCP'A is in PTL CCP B is Off A11 other equipment is in normal Mode 5 lineup.
QUESTION 7.06 (3.50) Answer the following questions regarding OS1000.07, Reactor Startup: a.
Prior to commencing a reactor startup, what is the minimum nuclear instrumentation required to be operable? (0.9) b. Whose permission is required to take the reactor critical? (0.6) c.
List the three actions that should be taken if criticality occurs below the rod insertion limit.
(1.2) d. What action, if any, should be taken if an inadvertent steam dump actuation occurs just after criticality is achieved and Tave decreases to 540 F7 (0.8) QUESTION 7.07 (2.80) Answer the following questions regarding FR-H.1, Response to Loos of Secondary Heat Sink:
i a. What are the two symptoms or entry conditions for FR-H.17 (1.6) b. When, if at all, is it permissible to feed a faulted steam generator? j (0.6) ! c. Following bleed and feed operation, why does the procedure require a i transition out of FR-H.1 if RWST level decreases to less than 125,000 gallons? (0.6) l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) l l l u_____
p-s L V- . . _7.
PROCEDURES'- NORMAL, ABNORMAL. EMERGEHCY AND PAGE
. . RADIOLOGICAL CQNTROL i 'UESTION _7.08 (2.50) Q Anawer the following questions regarding E-3, Steam Generator Tube-Rupture (SGTR): a. List the three ways of identifying a ruptured steam generator?'(1.2) b. Why would it be necessary to bypass the low low Tavg interlock during a steam genetator tube rupture? (0.7) c. Why must ECCS be terminated when the termination criteria are satisfied? (0.6) l , i QUESTION 7.09 (2.10) ' ! .What four operator actions must be taken prior to evacuating the control ! room? QUESTION 7.10 (3.00) a.
In addition to alarms, list four indications of a dropped control rod if the rod control system is in manual.
b. What are the immediate operator actions required for a single dropped rod? l c.
If the quadrant power tilt ratio is calculated to be 1.10, as a result of a dropped rod while at 90% power, what is the maximum f allowable power level and wnat is the time requirement for reaching
this power? Technical Specification Section 3/4 2 is provided.
} ]
l I I l l (***** END OF CATEGORY 07 *****)
1 l _ _ _ _ _ _ _ _ _ _ _ _ _, - _ _ -_. . 8.
" ADMINISTRATIVE' PROCEDURES.~ CONDITIONS. AND ' LIMITMIQHe PAGE. 13 I . IQUESTION :8.01- '(1.60) a) An NRC inspector:from Washington'wants to go into containment while: -the plant is at 50% power.
In order for this' entry to-be made.who; 'if anyone', musteaccompany the inspector? b. According to Technical Specifications, what action'should be taken Lif containment 11ntegrity is' lost'while at power? n , QUESTION' 8.02 ( 2. 00 ) - a.LWhatLis the. bases for-the requirement that no more than two irradiated-fuel-assemblies be allowed in the cavity and' canal at.any one. time?: b.<During refueling, u' der what condition, if any,.can' residual heat n removal flow be suspended? . QUESTION 8.03-(1.30) -a.
What condition, if any, allows the Shift-Supervisor to grant an
exception to.the person who has' issued a tag order from conducting a: ' verification of every: tag? (0.6)- b. When is an extension control tag used? (0.7) LQUESTION 8.04 (3.00) ' i What-actions,.BOTH operational AND administrative, are required by the ' Technical Specification if Reactor Coolant' System pressure' reaches 2750 psig while in Mode 27 Include appropriate time' limits.
).t
- 1 I
i
s i (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) I l ' _ _ _ - _ - = _ _ _ _ _ _ _ _ _ _ _ _
-_ ____- - - - Sin
8...... ADMINISTRATIVE - PRQQEDHEES ; CONDITIONS; AND LIMITATIONS PAGET 14 ' K ' t y . ' w ' QUESTION-8.05' (2.00):
< , . Thelplant:is-operating ah 75%;powerLand.~the'latestileak-rate data shows: o '10;5 GPM 1-Total RCS' leakage ratel.
.. 1.5 GPH -' Leakage into.the Pressurizer Relief Tank . , - L 1.2:GPM ' " Leakage into the PrimarytDrains Transfer? Tank' ' c 11.5 GPMc --Leakage through'RH-V871 .[ 0.8'GPM Total primary to secondary 11eakage < 4.2 GPM - Leakagelpast RCP seals 1What~RCS' leakage limits',.ifiany,-have been exceeded? Justify your, answer-
- by1 referencing.the'appropriatefsections.fromithe Technical Specifications
> send by showing any calculations.- -
s jQUESTION -.8.06 .(2.10) , The;unitEis presently at 30%~ power and is increasing.
Due to'a.combina- .j tion 1of vacations, sick leave,1and inclement weather, the' oncoming? shift 1 ' 'will'be one member-short of the Technical Specifications' minimum crew-composition (a..reactorLoperator.will'be missing).
~ .ha.[WhaElshouldhappenat_shifttornoverandwhatotheradministrative ~
'requirementLshould be checked in.regards to meeting the' minimum-crew compositions?- . ( 1. 3 ). u b. If the1 minimum crew composition can not bei met, what Technical: _
Specification 1 action' statement is in effect? (0.8) { . ! l (QUESTION 8.07 (2.20) l The plant is in Mode 3 with a' routine startup in progress.
During surveillance testing on one of'the. Diesel Generators it is determined l-that its day tank level switch is out of service and will not control ' itsifuel. oil' transfer pumps.. The Maintenance" Supervisor assures you ' that repairs can be completed within 48 hours.
Explain why the startup 'SHOULD or SHOULD NOT proceed, I , (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) __z_=___ _ i (
.. , . < s 8.
' ADMINISTRATIVE' PROCEDURES. CONDITIONS.'AND LIMITATIONS.
PAGE 15.
s i T UESTION; ' 08 (3.00) ~ ~ Q . Forfeach'of the following situations,Lindicate whether a four hour ' notification'to the NRCDvia the Red' Phone;is required.
State-a basis for your decision.
a..While:in, mode 5,-a fee'dwater isolation (FWI)Loccurs inadvertently-while: filling steam generator reference' legs.
b. While in Mode 5, a -control: room ventilation isolation occurs due.
if . to a failed. west air intake radiation monitor, l ~ c. A reactor trip-occurs on OT delta.T while in Mode' 2 at 5 X-10 E-8 - amps...The cause of the trip was'due to two different instrument-failures.
1 QUESTION 8.09' (1.50) se. Under what conditions may work-be started without a work request?' - (1.0) -b.
Whose authorization is required to start work without a work request?: (0.5)- QUESTION 8.10 '(2.40) The1 concentration:of the boric acid: solution inLthe Refueling Water l Storage Tank (RWST)'shall be verified once per 7 days in accordance with Technical Specification 3.5.4.
The chemist sampled the RWST onithe following schedule.
(All samples taken at 1200 hours.)
-April 1 --- April 8 --- April 16 --- April 24 --- April 31 a. EXPLAIN why the surveillance time interval requirements were or were not exceeded on April 16.
b. EXPLAIN why the surveillance time interval requirements were or ! were not exceeded on April 24.
!
I r.~ (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) _ - _ _ - _ _ _ -
- - _ _ ,8.i" ADMINISTRATIVE.PROCEDJJBES. CONDITIONS.~AND LIMITATIONS ~- PAGE K16' m ,
- 4 QUESTION T8.11.:
(2.00)
- a. Under whaticircumstances is it
- permissible'to violate a: Technical Specification-Limiting Condition'of Operation'(LCO)?.
(1.2)
b..Whoishould be notified.if an'LCO is. intentionally' violated? (0.8) QUESTION _-8.12-(1.90) Answer the'following questionsfregarding the Control.of Locks and- , Keys.; a. What-action should be takent if a key is los't to a' lock that is in an' inaccessible area?
- (1.3)
b. Who is responsible for authorizing the issuance of security keys during an emergency condition? (0.6) ,
1 f ! l' '. ! ! .. ! (***** END OF CATEGORY 08 *****) l (************* END OF EXAMINATION ***************) { l i _ _ _ _ _ _ _ _ _ _ _ - - _ __ _ _ - - - _ - - - a
p 15.
THEORY'OE_NUCLEAk POWER PLANT OPERATION. FLUIDS'.'AND .PAGE' 17 E ' THERMODYNAMICS' '
- ANSWERS -
SEABROOKJ1 ~-87/08/18-DUDLEY,.N.
'
, .. l /, ANSWER 1 5.01-(2.00) .Q'= U1 A1 (Tavg - Tatml) = U2 A2-(Tavg - Tstm2) [0 ' 7]. . L . U1 -= U2: l -- e
- A2 =.95 Al L-
.(Tavg - Tatml) =. 95 (Tavg Tstm2) . '573.5 - 59.7/.95 = Tstm = 510.6'F [0.7] , l .(573.5 - 513.8)'=.95-(573.5 - Tstm2) 1From: steam tables Patm = 748'.5 psia- [0.6] REFERENCE Thermal-Hydraulic-Principles and applications to the PWR Vol.
1, Chap 5, D p.
57-Obj:'NRF p 32.0, EO'9 , Importance: 2.8 i l 041020K502 ...(KA'S) ANSWER-5.02 (2.00) [ ;. bf,; ~ - : : = : ' ^:' ^ = '! '~ " !^ ^" 1 a. SDM :::.:1== the :u.~ The decreene'in the reactivity by the boren "ill be equal to the #
- inoreese in reastivity frcr. MTC [075-] * "'"
necrases, uvs obr ~ - n- ~M"jf#"$,,,Ro,_ f g,77, "I' " , c . Mon ; an-un E0.3]psv~wn powet -- - a.,,, r,,,;., .a <b. required SDM increases (0.5] by the worth of the. stuck rod [0.5]d" n
REFERENCE
Reactor Core Control For Large PWR, Chap 7, p 7-13 ! T.S., p 1-5', 3/4 1-1-Obj: NRF p 26.1, EO 6,11,14 i 3.1 001 010 K 5.35 3.3/3.6 .) 004 000 K 5.19 3.5/3.9
001010K535 004000K519 ...(KA'S)
' SDtt ; TWc d' cases [Q,3] .Yyc Ceckciasc to rec Arscrsvz rr' ny rer 61'ix' wrc s ne ec>aa ro r n* e i n cesse re rire nee rrrrry Mm ruc rc,veur.us wj, pc.weg recogse, Cy fff[ TIVJA L ES3 ACSrTJVf $CffC 7TVTr?' 43Lt 6C AdCCd 8t' rWs' /h'wr4 i ourr w/a nr c [a7] s
_ _ _ 5.
THEQRY OF NUCLEAR POEER PLANT OPERATION.' FLUIDS. AND PAGE
'THERMODYEAdigg ANSWERS -- SEABROOK 1-87/08/18-DUDLEY, N.
, i ' ANSWER 5.03 (2.40) l c.
(Loading a-fuel assembly into the core close to a neutron detector { increases the fraction of neutrons in-the core reaching the detector).
The detector count rate increases more than the core neutron population-l increases [0.3]. UNDER [0.3]. b._The detector will not see neutrons until there are a great number.
[0.3] OVER [0.3] - -j c.
The initial count rate is too high and the detector is insensitive to f core changes. [0.3] OVER [0.3]
d. The detector will not see neutrons until there are a greater number - [0.3] OVER [0.3] { ' REFERENCE Fundamentals of Nuclear Reactor Physics, Chap 8, p 8-28 to 8-31 jl Obj: NRF p 19, EO 10 3.1 001 010 K 5.16~ 2.9/3.5 001010K516 ...(KA'S) !, , ANSWER 5.04 (2.50) a.
Delta T increases ~[0.25] as Thot increases (as boiling occurs in core) and Teold remains relatively constant [0.5] l b.
Peteam will decrease as boiloff occurs in the SGs [0.4] while Tcold
remains relatively constant due to stagnant RCS conditions [0.35]. j i c.
1, Increase l 2. D;;rcccc o 1sttensc ! 3.
Increase j 4. No Change [0.25 pts each] i l REFERENCE ! Thermal-Hydraulic Principles and Applications to PRR Vol 1, Chap 14, p 14-26 to 14-29 Obj: NRF p 40.0, EO 7,
l l 3.4 000 015 EK 1.01 4.4/4.6 l 000015K101 ...(KA'S)
, L _ _ __ ._
__,- .-_. ? 1 (- % . t , 7 g -. W.' L THEORYlOF NUCLEAR POWER-PLANT' OPERATION'.~ FLUIDS. ANDJ PAGE ::19 0 ,- THERMODYNAMICS 'ANSWERSi--~SEABROOK 1l ' 187/08/18-DUDLEY, N'. ~ -
' t '- .u + iANSWERL' ~5;05: 1(2.00)J
a.
11ncrease's1[0. 2]. -As PZR temperature' rises, so:does; pressure, hence the margin-to' saturation increases.[0.46]. t b.x Increases-[0.2]. The Delta.T across the' core:will-be lower;to~ pro - duce the same power,.so Th will decrease (and'the margin to saturation ' increas es) [0.-46),
c.
. Decreases - '[0. 2].. Since more power-is being produced-in thett'op half: _ of the coreiand since' head loss increases with core. height [the marginL ito saturation. decreases) [0.46]. ' ' - ' REFERENCE . . ._ . to the PWR Vol.
2,. Chap.
.. 'ThermaliHydraulic' Principles and Applications .13, - p.13-2 3 -
- Obj:.4NRF p.39.0, EO 3
.'3 4 003. 000.K : 5. 01 c3.3/3.9, . ~003000K501 ...'(KA'S)- ' ANSWER.
5.06: (2.00) La.
Moderator Temperature Coefficient (MTC) [0.4] due:to an. increase '(more negative) in MTC.as. boron concentration is reduced over core lif e..[0. 6]. ' ? -b.. Doppler :(FTC) [0.4]. Fuel temperature changes before the other.
- parameters-change [0.6].
. ' cA t1TCL : se n e c aW M & Y JWDUCED _ TRh WSJ ggf 3S g Sg umg g, = REFERENCE-Reactor Core Control for Large PWR, Chap 3, p 3-5, 41 .Obj: NRF p 22.1, EO 18, 20 p 21.1, EO 19 G 3.1 001:000 k 5.49 3.4/3.7 .001000K549- ...-(KA'S) i -
- -,. - .,-, ~.- - - - -- - _ _.. -. - 5.
THEQRY OF~NUELEAR POWEB_ ell @T OPERATION. FLUIDS. AND PAGE
4 IHEBdQDXHAMICS ' ANSWERS -- SEABROOK 1-87/08/18-DUDLEY, N.
- ANSWER 5.07 (3.00)
Koff1 = 1/((1-(-0.04)) = 0.9615 . [0.3] R 100(1 .9615) 132(1 - Keff2) Keff2 = 0.9709 [0.7] = , rho 2 = (0.9709 -1)/0.9709 = -0.03 [0.3] delta rho = rho 2 - rhol = -0.03 -(-0.04) = 0.01 = 1000 pcm [0.6] Boron delta rho = -150 ppm x -10 pcm/ ppm = 1500 pcm [0.5] l Xenon delta rho = 1000pcm - 1500 pcm = -500 pcm [0.6]
REFERENCE Fundamentals of Nuclear Reactor Physics, Chap 5, p 5-21 to 5-23 Obj: NRF p 21.0, EO 2 f p 24.0, EO-4 3.1 001 000 K 5.28 3.5/3.8 001000K528 ...(KA'S) ANSWER 5,08 (2.00).
The stuck rod'would be worth more [0.8]. Reactivity worth is proportional to the relative flux squared [0.4]. For a dropped rod, the flux is depressed adjacent to it [0.4] whereas if the same rod was stuck out, while the others were inserted, it would be exposed to a much higher flux than the flux in the rest of the core [0.4]. REFERENCE Reactor Core Control for Large PWR, Chap 6, p 6-27 'Obj: NRF p.
25.0, EO 7,.9 .f 3.1 000 003 EK 1.03 3.5/3.8 005 EK 1.05 3.3/4.1-f 000003K103 000005K105 ...(KA'S) j j
l - i , I _ - _ _ _ _
._ .- -- _.
-. . j; < < L >5.
- THEQBYOF: NUCLEAR-POWER PLAET OPERATION. FLUIDS. AND.
PAGE-2 11
- !
. > H 'IHEBMQDYNAMICS- .
.
' JANSWERS:--LSEABROOK 11-07/08/18-DUDLEY, N.
.j! i [ i ANSWER-. 1 5.09- - ( 1.' 50 ) ' Fluid;velocityfthrough'the'venturl. increases (0.3];which causes: pressure-to' decrease to the saturation: pressure ~and cavitation occurs [0.6] which . lower the'dp.across'the venturi![.0.3] and thus:' reduces flow since flow L i s. p r o p o r t i o n a l'. t o s ths-dp [0.3]. i
- REFERENCE..
. .. . . d '
- Thermal-Hydraulic Principles.and Applications to-the PWR Vol.
2,' Chap 10, /p 10-72.
-l Obj:'NRF kp.35 0, EO 5.- ' < '193006? Fluid statics and dynamics K 1.11 3.1/3.2- -K-1.15 3.1/3.2- '3.5 061 000.-K'5.05 2.7/3.2 . . . - 061000K505' .193006K111 193006K115 ...(KA'S) , (ANSWER; '5.10
- ( 1. 60 ) -
_ a.: False
- .
Al b.?True c..True .. j d.1 False -[0.4 pts each).
REFERENCE . -i ' Reactor. Core Control-for Large PWR, Chap 4, p 4-19,21,22,23~ ] Obj:.NRF p 23.1, EO;10,12,17 c 3.11001 000 .K 5.33 3.~2/3.5 3.1--001 000 K 5.34-2,1/2.2 i 3.:1 001 000 K 5.35 2.1/2.3
001000K533 001000K534 001000K535 ...(KA'S) ! l I i i
1
.q l
_ - ____- -__._-_- _.--
_ _ - _ _ - - - _ - - - - _ , .] c5.' TREORY-OF~ NUCLEAR POWER PLANT OPERATION. FLQIDS. AND PAGE: 22' tIHEBMQDYNAMICS q . -ANSWERS.--iSEABROOK:1-87/08/18-DUDLEY,LN.
' < ANSWER' 5 ','11-( 2. 00 ) - a.
' Increase b.
Decrease
c.
Increase d.
Decrease e.- Increase .[0.4. pts each] , . REFERENCE ~ i iThermal-Hydraulic Principles and Applications to PWR Vol 2,. Chap 10, ! p 10-42, 44 i ' ' CS, p CS-6,.CS-87 Obj:.NRF p 36.0, EO 8 Components:191004 Pumps 11.05 2.3/2.4 1.06 3.2/3.3 l _ . . 1.07-2.9/2.9
- 191004K105 191004K106 191004K107
...(KA'S) .1 ! ' ANSWER 5.12 (2.00)- . ' a.
Decrease b.- Increase c.
Increase-
- d.
Decrease o.
' Decrease [0.4 pts each] REFERENCE Thermal-Hydraulic Principles and Applications to the PWR Vol 2, Chap 12, P -12-41, -- 5 2,
Obj: NRF p 31.1, EO 10 3.2 002 000 K 5.01 3.1/3.4 5.09 3.7/4.2 5.11 4.0/4.2 002000K501 002000K500 002000K511 ...(KA*S) l < l i l i i ! C___--_--_---------_---------
.- - -. ._ Q.. L6. 4 PLANT:SEIEMS~ DESIGN. CQHTROL AND INSTRUMEHIATION ' ' PAGE / 23 ~ : ANSWERS --.SEABROOK 11-87/08/18-DUDLEY, N.
'
. h-N ' ANSWER 16.01'~ (1.50) When;TL413B) fails low the Train'B PORV-(456) will.open [1.0]. ~ < . , RCS pressure.will decrease [0.5]. i ' REFERENCE.
.. . , , System Description'PPLC,..p PPLC-20
Obj: Lesson 1108, EO 19 '
33.3~010 000 K 4.03 3.8/4.1 '"
- K 1.03. 3.6/3.7
- 010000K103 010000K404
...(KA'S)' i l i i / ANSWER E6.02.- (2.50)
Control rods Ewithdraw-in" response to '[0.03 i (0 7]
the. temperature error [0,3 sand power mismatch [0.3]r(0,f7.
- which adds. positive reactivity.[0.1] fLO.f.7
-;] 'dith.; small MYC, reactor percr will reiro rapidly [ 0. S ] _* nd a pcuer oveeshoct recultc in a hi-eh neutren flux trip. [0. S]J d L 0 kl MTC e A uSES ' Tus' to w ar*.SC CQ G ~ Av0 Rt: Wus rkte og ung fu iem, .) REFERENCE- .. . . 'GN A$'* % re System Description CP, p CP-9, 10'(4/86) o r A 7' g .
, Obj: Lesson.#1113, EO 5 3.9 016.000 K.3.'01 3,4/3.6 K.4.03 2'.8/2.9 'A 2.01 3.0/3.1.
q 016000A201-016000K301 016000K403 ...(KA*S)
1 I q u q l g
l
l l l l
'.. ' ,
. .-- - - _. -. ._ _ c -. a.
g; 'kI . '6.
PLANT SYSTEMS? DESIGN'. CONTROL.' AND' INSTRUMENTATION PAGE-242 .) ' c . . ANSWERS-~.SEABROOK;1: -87/08/18-DUDLEY',- N.
, - - . . .
u.. ) - .
- I
, .) -i
- ANSWER
~6.03-
- (2.80)
, , . M.1 :RemainIthe.Same ' ! T ~ -2L Increase ' 3:. Increase l <4: Ecrri.: the h = 3- '[0.3 pts.each]' 'l ' gNCRE A 6 E: -l b'?close's isolation valves [0.6]-(due.to high f1'w),. _. [ 0 ~. 5 ] - j o as pressure increases-a check valve on theLflow inlet seats . - ] ' c. See' attached figure.[0.5]~ REFERENCE- ! System Description: p:CC-11, 12,'41 - .Obj: Lesson.#1118,-EO 3,'8,,10 . .1 3.10.008 000.K 1.02 -3.3/3.4: j 1008000K102 008 OK3bi ...(KA'S) - '
- t ANSWER-6.04 (2.60)
. . i 'a.': Bypass breaker.A will open resulting in a' reactor trip [1.0] i ~.b. l'. TG trips , '2... Valves'close 3. Steam dumps:will operate-in the load rejection mode .. . 4. Mc c'#ect o' [0.4 ptsJeach]~ uun ou s to geSer rggry y-a REFERENCE
System' Description: RP, p RP-12 I lObj: Lesson #1138, EO 6,
3.'1 001 000 K 1.05 4.5/4.4 K 2.02 3.6/3.7 . . K 6.03 3.7/4.2 '001000K105 001000K202 001000K603 ...(KA'S) _ _ _ _ . _. . _..
,... _ ) ' ! I
FIGURE CC-3.2 - l PRIMARY COMPONENT COOLING SYSTEM LOOP "B" .. C".~'~) "q p=. g "& ~ . e u.
..., -..... b h ' JS(. q ' vG4 6 gg ,,Q
- .to e ca,.,,
s.o-. a., ~,0 .... e-,- e c,oo o.,.- on.... so...,,, e...,,_ l c.
....a.. l e e.. -.. , _w., , ' " " ~ FIGURE CC - 3.2 'O ' ,,,,,,,. i PRIMARY COMPONENT ' ' lW=c 'l c-/ \\ COOLING SYSTEM LOOP"B" (12/85) tr.., e'..., J g,. c-.. ; a-m ,- -
i COOL... t a$ l w.00.
etcee { d *?.'?"o:M'*; I <
o IW
l "a' .O l-
I lac ewe c emo I we co= osseum
g.ao uotoa on c,ay ,
I ,0 se a.,o .oes L' ' .em, ; ,..,, I wo= ca I la.c,o*we e ano I i = motoa os eta g y g
_ - _. a a oo,... w 2 ' I l. ec e ss I ___ - i-.. .w i g,,,....,; l
- 0'.".'"
l cy,.g j r-H=='. l l = ~ - l _,.a , i..._ c._ i i cowea. ssoas i i 1 M8=8 I . t O , ! l , , .a a _.
. - W IN - .i , j l , !
l I I ! !, ! s ! .) f.
, ., i + CC-6 2/86
i
- _.
, --- 13. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 25-ANSWERS ---SEABROOK 1-87/08/18-DUDLEY, N.
ANSWER 6.05 (3.00) a.
Return flow increases (above 5 GPM) [0.6] Lower behring water. temperature increases (approaching 230 F): [0.6] b.
Close the #1 RCP Seal Leakoff Isolation Valve ( AV-8141A) 60 83 Reduce power to < 37.5%s(P-8)f0,5J [0.0 pts cach] e l STOP R C P [o.5] REFERENCE OS1201.01, p 3 System Description: p RCP-21, 24 3.4 003 000 K 1.03 3.3/3.6 K 6.02 2.7/3.1
000 017 EK 2.07 2.9/2.9 ! EK 3.03 3.7/4.0 ! EK 1.05 2.8/3.0 000017E105 000017E207 000017E303 003000K103 003000K602 ... - ( KA ' S ) ! ! ANSWER 6.06 (2.10) i c. Fall closed b. Fail closed c. Fail open d. Fail open ' o. Fail closed -
f. Fail-c1csed
g.
Fail closed [0.3 pts each] j ' REFERENCE I ON 1242.02, p 5 ON 1242.01, p 5, 6, 11, 12 PID-RH-B20662 ! Obj: Lesson #1104, EO 28 - 3.8 078 000 K 3.02 3.4/3.6 078000K302 ...(KA'S) i
! ! ! ! f
2 , L_ - _- - - - - - - - - - - - - - - - -.
m - ; , .. 3 - ,
y - - -- ----,;- -
. i .. f'h -' l'- S , ,_ f6. : PLANT SYSTEMS: DESIGN; CONTROL.' AND~ INSTRUMENTATION
- PAGE j26-
ANSWERS:-;SEABROOK'1j-87/08/18-DUDLEY,:N.. i i - d s.
d - LANSWER ! 6 '.~ 0 7;. '(3.00) '
i MFW pumps-trip._ j Feedwater' regulating valves'close: ) -
- MFW bypass-valves
- close
' L . Main feedwater isolation. valves.close j > L MDAFW: pumps; start 1: . ., TDAFW^ pump steam; supply. valves'open' [0.5 pts:each] o .. .-i . REFERENCE 'I -
- FW, p.FW 12,'57i f
d EFW, p EFW-2,11,15,16-L , Obj:, Lesson?#1139,-EO:12 H '3.5 059.000.K 4.'16 3.1/3.2 K 4.19.3.2/3.4 '061 000'K 4.02 4.5/4.6 .~059000K416- .059000K419 061000K402 ...(KA'S) ANSWER.
6.08i (2.50) y a'... Raises.the limit'[0;5] (because'high dT indicates'a. higher-power) l 'b',' Increases.[0.5]'(to raise pressurizer level to 100% program,:because . of the higher-Tave) ic; Rods. move in [0.5].(because of.the Auct. Tave/ Tref mismatch) d. NoJeffect'[0.'5] (the demand' signal is present (Tave/ Tref)'but there is no' arming signal)-
- e ~ Decreases setpoint [0.5]-(teccuse hie:h dT indicatc; reduction in DNDR)*"
k r u n r w edeAss vince Lowess s er nwvr s e cm c uc,,, ryny' . , ') -REFERENCE System Descriptions: p CP-9, 20 ! " p PPLC-35
p SD-14 p RP-32 .Obj: Lesson #1107, EO 6 - 3.9:016 000 K 3.02 3.4/3.5
.
K 3.03 3.0/3.1 K 3.01 3.4/3.6 016000K301 016000K302 016000K303 ...(KA'S) ! l ! i j m
__ _-- 6.
PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION LPAGE. 27 ANSWERS --'SEABROOK 1- -87/08/18-DUDLEY, N.
i- ' ANSWER 6.09 (2.00) ' e. Source range or intermediate range high flux. trip [1.0) " b. No change-[1.0) (vital UPS does not have a static transfer switch)- p i i REFERENCE ! RPS, p RP-30-EAC, p EAC-28 Obj: Lesson #1098, EO 5,
3.7 062 000 K 2.01 3.3/3.4 K 4.10 3.1/3.5 3.9-012 000 K 2.01 3.3/3.7 00K410 062062000K 2012010620 ...(KA'S) ANSWER 6.10 (3.00) e. The EPS will sequence on equipment required by the new condition without stripping the running loads. [0.45]
RA relay will-trip its affected loads [0.3] b. to allow UAT and RAT _ incoming breakers to be shut if the RMO bistable is energized. [0. 7 5]- c. Simultaneously depress both Emergency stop pushbuttons on the MCB.
l ostern [0.75] j dr-Emergeney-Start [0. 25] / i =SI Signal [0. 2 5 ]-j G4VLOP cignal [0. 25] s.
J REFERENCE-EDG, p EDE-21,35 HO-EDM-64, 12/83 Obj: Lesson 1100, E0 11, 24 i 3.7 064 000 K 4.10 3.5/4,0 , K 4.11 3.5/4.0
K 3.01 3.8/4.1 I 064000K301 064000K410 064000K411 ...(KA'S) ' l-l <
, ! l ,
= - - -
- -
-7=--------=-------- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - k.Y' ;; iipWje , e - j , i 7{ 7i PlQCEDURES? : NORMAL'.-ABNORMAL. EMERGENCY AND'- ..PAGE: 28l ' ' ' '
, RADIOLOGICAL' CONTROL -' N@W ~, jANSWERS:-..SEABROOK-1.
' 87/08/18-DUDLEY, N.
, - , . W ' < ! ? ANSWER- ,7.01E - ( 1, 20 ) : ' T..<3-.. Ayin A TB..
.C; :4, P < ,'D.
' [ 0 '.'3 : p t s e a c h ]! V U ' REFERENCE ! ' s ' Plant Wide Generic' 194001-KL1.16L.3.5/4.2- ' '
194001K116: _...(KA'S)L I ' - q j . .
4 iANSWER.
7.02: ~(2.00) - 5(24-18) =-30 Rem;. Lifetime < limit-= 3 - 28 =:2 Rem- . i With' Form;4 on' file he'is-permitted ~1 Rem /Qtr [0.7] " -0.4 Rem /hr'+ 0.2 Rad /hr2x110' Rem / Rad:= 2.4 Rem /hr
- [0.7]
O .1;0. Rem /2'4: Rem /hrj= 0'417; hrs.=,;25-minutes [0.6] .j . . l-' tREFERENCE- ' LRP-5 4,4p 3,-Rev.J2 ' ' ,, 194001'- PlantLWide'. Generic K 1.03.-2.8/3.4' K 1. 04' 3.3/3.5 ?194001K103-194001K104 ....(KA'S) EANSWER 7.03 (3.40) a.- 4,1,3,2 .[1.0] b'. -The'.. Core Cooling Red Path should be immediately addressed [0.6] - becaused it'is of higher priority than,the Heat' Sink Red Path [0.6] - RestorepowerfirstIAWECA.O.f(andthe do4,1,3,2)[0.6] c -. , .d.
When directed to in E-0 [0.3] or when transferr,ing out of E-0 [0.3] REFERENCE . "(Usage Rules) i ' .ECA-0.0.pg.3
Obj: Lesson #1343, EO 1, 3 194001 Plant-Wide Generic A 1.02 4.1/3.9 1, L_.. 2. - .
g_ f?%R' l' f&QQ:le ' - , .. I j, a , -p PROCEDURES: ! s w w ? Y .
.. . .. ,, 7 :7; 2 P'MJ L,'ABM R E._. EMERGENCY'AHQ::/ PAGE. 29 N.
RADIOLOGICAL 3sTdEQL
as w ' " < hAbih$lERS j-- EABROOKJ.O I -87/08/18-DUDLEh,'N.
4 %: - 7s1 j13 w
- c
. s\\.
>; %,dO1 M 02t _. f, (M.',S).
'. x w a y @ 104 . ... <-
, ' ' 4 %v , 5 u.
.
.- - s , , .'%,
p y -; ' (' b* a Q :. v V;/ o ' y4 !
. V .. ..l: ,. y ') g: .\\% s
' y > @ANNNER' !7104 . pf(.1.50 P 4 l/ 7 ', .. ." F . . a .: av .i y >;. 1+7.acShut thefMSIV's:[0.8]f c. h;9- %,l g ,.. ,
. jg , . u , b.' Prevents overspending the turbini. LDj] l' , ' i n A r-a.. 7 RENTRENCE /%, E b'- // ' ,
"
- -t g
'C L.,: D
m LE-0L p;2 v. ' - - , ,s . i i f ' I.
' 103): ~ j1 '" " " s d . e
di5'd45L00d?K~4.02-2.5/2f/h/. I .' 3' . 'K 4'.13L 2.6/2.8 / ' , pf ,- . ,1 f' '- 3q " 4;3/413i.NE000K413 ff...i(KA'S) j' s. / '
- 3.1.000 007 EA 1.07
// ., ~./Q045000K402:
- 000007E107 q.
< < a llg Qq ', , 3la s , s V.
, i y l
O je 't
- {
. !J Zf - / < a.
.. Ltr " it j , , LANSWER; ~ '7;05 L j Ft.iG0f f~s. } l.
.. r .* 4- ^ ' RE'S ET ' Sf' CC 5].. . . Q}. ?? W l. j!.ij.
f.o . x 'LBlock low MSLIS,[ '. Je .c .. L' . _ Restore sRHR lineup by shut, ting suction valve @ rom.:tWST [0,-7e]" , ~ Shutdown EDG-[0.5 N !O .q ; 4 - / (' a Lg3-c - Restore FN lineup for recircing'the SG [0.5 F
3l0SM Stop'CR:reciro gan A:[0.5]v ~ ,, ES#f /Wbp$( m ca'p,
- A'
T sN t , wode-c
FERENCE V tcf . 12-14-60,A , LER 50-443/88-002-00, OBJ: ' Inadvertent SI, l , . .. , 001050G10.. 003000G10-010000G10 ~01SD00K107 ..(KA'S) . - .Dernad} % r k Tur Pms, y-2 j g4 - - /:] '
a e - %. - ' ,n 'f i A.
! ,. }
. -,, s t .z) I I [t i ,\\ ,(' ~ ' ,. . . ,4j , ,,e \\m , d, i ($ ,J . .. , , (- ^ (t
.
, .) d .s, i, '
a /
t ! / ' , h ii-! , \\.) C b ~ t $ j a y + /, , . ii s1 .. d , '
,
v t , a ' e ' , I % L l ,\\ , , >3 I . f.
' ' s > < s ' s , t
' y ( .h IN ' .f , s
t r.
. , % g- > ,1: __ _ .___.__________i ' li '
p 3# 4f ~~ - - ~ ~ ,1 ,. l 7.
PROCED$BE.d'._EQBMAL, ABNORMAL. EMERGENCY AND PAGE-30 l .BADIOLQGICAu:CONIBQL ANSWERS -- SEABROOK 1 l' -87/08/18-DUDLEY, N.
x ANSWER 7,06 (3.50) a.. Two.SRs,-Two irs, Three prs [0.3 pts each] b.
.USS [0.6] a c.
Insert control banks ,
Verify SDM ^ > < Obtain RCS boron sample Recalculate ECP [any 3 @ 0.4 each] Wrtry trarce EWHsitt d.
Return Tave within limits within 15 minutes [0.4] } or.put plant in HSB in 15. minutes [0.4] l r REFERENCE
TS, p 3/4 1-6 i OS1000.07 Rev 1, p5 I .Obj: Lesson #1170, EO 1,.4 3.1>001 050 G 2 3.2/3.9 3.9.015 020 G 11 3.1/3.8 , 3.4~003 000 G 5 3.4/3.8 s: 3.1 000 024 EK 3.01 '4.1/4.4 ' 000024E301 001050G2 003000G5 015020C11 ...(KA'S) .a.
, ANSWER 7.07 (2.80) j -, y a.
EFW flow. not verified in E-O LO 93 01-toD n aar l ostetu TreenforF-0.3JFeedwaterflow<500GPM"andthree'SG<65%)'LOS] ' [0.8. pts ccch]f b.
If a non-faulted SG is not available [0.6] l c.
To' cool the core using cold leg recirculation [0.6] REFERENCE FR-H.1,.pgs 2, 8 Obj: Lesson.# 1420', EO 4,
- 3049, EO 4, 7 3.5 000054'EK=3404 4.4/4.6 G 11 3.4/3.3 G 12 3.2/3.2 000054E304 000054G11 000054G12
...(KA'S) , _.-.__---.__--A _-
. .-. _ _- .. - _ _.. - - - _ _ - _ _ - -. I .' g ' 7i PROCEDURES C NORMAL'. ABNORMALi EMERGENCY AND-- r PAGE. '31' - , - BADl%QQ1 GAL _GQHIBQL iANSWERSi-'- SEABROOK 1-87/08/18-DUDLEY, N.
. L ' ANSWER 17.08 . 2 ; 5 0 ).- ( t ia.o-Unexpected: increase inLany SG NR-level.
, if Hig' radiation from:any SG^ sample !L
- High. radiation:from any steam'line
' .[0.4' pts =each]
- b.-
To' maintain steam dump' operation during:-RCS cooldown' [0'.7] c; To prevent' overfilling the ruptured SG [0.6] REFERENCE E-3, pga.4;10,15 'ObJ: ELesson:#:1414,'.E04,_6 3'.3 000 037.EK.3.07 4.2/4.4 000037E307 .G.12; 3.5/3.8 . 000037G12.
...(KA'S)- ' ANSWER 7.09 (2.10) ' trip-the reactor [0.6] shut the-MSIVs [0.6]. . trip.the.RCPs'[0.6] plecc ASDV 'centrcl ::: itch te 'clerc [0.3] 9Mct Born rewy, ssp 5 z u ygg r REFERENCE SD-95-10 OS1200.02-(87-res-0191) ~0bj: Plant-Wide Generics 194001 A 101 3.3/3.4 '194001A101 ...(KA'S)
l l l i l ' ' QL __-._ - -_ _--. s: . '
'j S* h s7.
' PROCEDURE.- NORMAL. ABNORMAL. EMERGENCY 'A@ PAGE '32 L, B@IQLOGICAL CONTROL h ~ ANSWERS:-- SEABROOKal-87/08/18-DUDLEY, N.
, ' m < , / ANSWER-7.10' (3.00)- a.1 Change'in"DeltacILindibation.
- Decreasing'Tave Decreasing ~ pressurizer level'
- Decreasing pressurizer pressure.
Decreasing flux--on one or moreschannels- ' Rod position indication / Rod bottom-light-Tave deviationofrem' Tref Decreasing-steam pressure- [any-4 @ 0.25'each] c M NNEG cuptrar cottp4 Rated Abd&) b.= Place red contr:1 in annual [0.5] se !! . Reduceturbine' load'to.maintainTave/ Tref'{0.5]af/.0} 7D7, 1c. Power to &O4-e [0. 7]: H-within;2 hours'[0.3] REFERENCE -
- TS p.3/4 2-12 OS1210.5.p 1-
. '000003G004 000003G005 000003G010 ...-(KA'S).
H k ._E_'___.-__.__
m
- - - - _
. _ _ - - _ .- .. . - _ - - - - -- - -. -- fbi ' ADMINISTRATIVE PROCEDURES. CONDITIONS.I AND:LIMITATIOHg- ~PAGE :33- -
. ,. LANSWERSL- >SEABROOK.15-87/08/18-DUDLEY) N.
- I
. d as
, 1, JANSWER-8.01 (1.60) Ja.LAnLHP; tech [0.4] Land.another permanent staff; person (0.4]. ?C -b.-: Restore 1CNMT. integrity,(within 1-hr)[0.4] ~or beLin HSB;(within:6. hours) [0.4]- . + - REFERENCE z TS;'p13/4.6-9-r 3,6 103 000.A 2.05r 2.9/39; G 15' 3.8/4.1- - L G - 1-3.5/3.8< G5 3 '. 3 /4.1 103000A205 1103000G15 '103000G1 103000G5 (KA'S)' < ... ANSWER .8.02: (2.00) ' a.:To: avoid having'more than two irradiatedcfuel' assembly out of.the core . fin:the event of' loss of' cavity seal event,[1.0]. b. Flow gan be-suspended'(for. up to;1 hr per.8 hr period) [0. 5]- during performance of core alterations in the vicinity of the reactor ~ pressure vessel hot' legs [0.5]. REFERENCE-
- OS1013.03, p 3
.OS1000.09-(Rev 1),.-p 5 . 3.11 034000 K 1~.01-2.5/3.2 K 1.02 2.5/3.2 G'10
- 2.7/2;9
, . . G~13 2.7/2.9 . . . . 034000G10 034000G13 034000K101 034000K102 (KA'S) ... . i b p o ,
. ._ . _. _ _ _ _ _ _____.______.___m_ . _ _ _ --
_- - ,
'a.
~ .
- 8; PADMINISIRATIVEiPROCEDURES, CONDITIONS. AND LIMITATIONS PAGE r34 s, 3.
/ 'IANSWERSV-- SEABROOK :- l'.. ---87/08/18-DUDLEY, N.
~ , p , ANSWER.
8.03.
(1.30) si , , a.JUnderfno. conditions [0.6] ' c. On remote operators to indicate the associated equipment is-L tagged i[0 '; 7] n iREFERENCE MAJ4.3:Rev 6,1p 7 c MA':- 4. 2 Rev 3, -; p 4,.~ 111-Obj:. .. Plant-Wide Generc. 194001 K 1.02 3:7/4.1-i 194001K102- ...-(KA'S)- I ' l ANSWER-8.04 ~(3.00.)' ~ ' .Beiin hot _ standby: [0.7]L. . j with' pressure within limits /in one hour [0.8]
' Notify the NRC,in 1 hour [0.5]
Notify the Vice President-Nuclear Production and NSARC (in 24' hrs)[0.5]
Safety Report prepared and submitted (in 14. days)[0,5]. ] ^ REFERENCE TS pgs. 2.1,.6-15 ,
- 3.3 000 027~G 4 2.4/3.4
.! EK 3.03 3.7/4.1 ~ 000027E303-000027G4- ...(KA'S) I . ANSWER: 8.05 (2.00).
b i l UNIDENTIFIED Leakage limits exceeded 3.4.6.2.49 [1.0] ' .10'51- (1.5+1.2+1.5+0.8+4.2) = 1.3 gpm which is > 1.0 gpm [1.0] REFERENCE R TS'3.4.6, p 3/4 4-20 to 4-24 'Obj: Lesson.# 3029, EO 2 ! 3.2 002 020 Sys gen 5 3.6/4.1 ' 002020G5 ...(KA'S) i i L ___ ___=_ _ _ )
,----. 18, ADMINISTRATIVE PROCEDUBES. CONDITIONS....AHD LIMITATIONS PAGE
ANSWERS -.SEABROOK 1-87/08/18-DUDLEY, N.
ANSWER 8.06-(2.10) a.
An RO from the previous shift will remain on duty [0.65] but work hour limits must'be considered [0.65] b.
Immediate action [0.3] must be taken to restore the minimum shift crew composition within two hours [0,5]. REFERENCE TS pg 6-5 Plant-Wide Generic 1P4001 A 1.03 2.5/3.4 194001A103 ...(KA'S) ANSWER 8.07 (2.20) With diesel fue1~ oil transfer train is out of service-[0.4], the diesel generator will also be out of service [0.4] and thus the plant is in an action statement [0.4]. Entry into an operating mode is not allowed un-less all LCO's are met for that operating mode without reliance on an action statement [0.7], therefore the startup cannot procede [0.3]. REFERENCE , TS 3.8.1.1 and 3.0.4.
3.7 064 050 G 5 3.4/3.9 06400005 ...(KA*S) i ANSWER 8.08 (3.00) No [0.4] FWI is not required to be operable in Mode 5.[0.6] a.
nun Hb Gccv HHHEd Fh1 S CHIC C b. Yes [0.4] inadvertent ESF actuation when CRVI required to be in service.[0.6] ' c. Yes [0.4] all reactor trips not due to testing [0.6] REFERENCE 'tH4UC p 4.43 i 10CFR50.72 l NUREG 1022 Sup 1, pp 7,8 194001K102 ...(KA'S) t> v : s rn sn ro kun, he 30, I987 l , - L- __ _ - _ -
r _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ' f8; ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE
Ls . ANSWERS -- SEABROOK 1-87/08/18-DUDLEY, N.
ANSWER 8.09-(1.50) a. When work is of an emergency nature (priority'E work request)-[1.0] 'b.
Shift Superintendent [0.5] REFERENCE SSMA Procedure MA 3.1 Rev. 7 194001A116 ...(KA'S)- ' ANSWER 8.10 (2.40) a.
Interval requirement not exceeded [0.6]. Eight days does not exceed 1.25 times the specified interval [0.6]. b.
Interval requirement exceeded [0.6]. The last 3 consecutive intervals exceed 3.25 times the.specified interval [0.6]. REFERENCE TS 4.02, pgs 3/4 0-1, 02 TS 3.5.4, pg 3/4 5-9-Obj: 3.2 006 050 G 5 3.5/4.2 006050G5 ...(KA'S) ANSWER 8.11 (2.00) a. During an emergency to protect public health and safety [0.8] when no other action consistent with TS can provide adequate or or equivalent protection is immediately apparent. [0.4] b. The NRC should be notified [0.8] REFERENCE 10 CFR 50.54 (x) I 10 CFR 50.73 TS pg 6-15 Obj: Plant-Wide Generics 194001 A 1.16 3.1/4.4 194001A116 ...(KA'S) , _ -. - - _ _ - - - - - - - _ _ -
- -. . _. _ _ 8.
ADMINISTRATIVE PROCEDURES. CQNDITIONS. AND LIMITATIONS PAGE
l ANSWERS -.SEABROOK 1-87/08/18-DUDLEY, N.
I ! ! ANSWER-8.12 (1.90) e. Lock 1the acess point to inacessable area [0.8) replace lock (within 5 days)CWof regaining access to lock [0.5] b.' Unit Shift: Supervisor [0.6] REFERENCE OPMM Rev 2, p 7-1.3
Plant-Wide Generics 194001 K 1.05 3.1/3.4 ) 194001K105 ...(KA'S) i ! l l ! i i i i ! ,
! l l
n a L:__ __.
_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _. _ - -QUESTION ~7.10 3/4.2 POWER DISTRIBUTION LIMITS 3/4. 2.' l AXIAL FLUX OIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following-target bands (flux difference units) about the target-flux difference: a.
t.5% for' core average accumulated burnup of less than or equal to i 3000 MWD /MTU; ! . b.
+ 3%, -12%Lfor core average accumulated burnup of greater than 3000-MWD /MTU; and c.
+ 3%, -12% for each subsequent cycle a
The indicated AFD may deviate outside the above required' target band at greater than or equal to 50% but less 'than 90% of RATED THERMAL POWER provided the indi-
! .cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative penalty' deviation. time does not exceed 1 hour during the previous 24 hours'. The indicated AFD may deviate outside the above required target band at greater ' than 15% but-less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed 1 hour during the previous 24 hours.
- APPLICABILITY
- MODE 1, above 15% of RATED THERMAL POWER.*
.; , ACTION: J With the indicatea AFD outside of the above required target band and with.
a.
THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either: 1.
Restore the indicated AFD to within the target band limits, or 2.
Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.
.I b.
With the indicated AFD outside of the above required target band for more than 1 hour of cumulative penalty deviation time during the previous a 24 hours or outside the Acceptable Operation Limits of Figure 3.2-1 and' with THERMAL POWER less than 90% but equal to or greater than 30% of . I RATED THERMAL POWER, reduce: - 1.
THERMAL POWER to less than 50% of RATED. THERMAL POWER within 30 minutes, and 2.
The Power Range Neutron Flux * ** - High Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
"See Special Test Exceptions Specification 3.10.2.
- Surveillance testing of the Power Range Neutron Flux Channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits of Figure 3.2-1.
A total of 16 hours' f SEABROOK - UNIT 1 3/4 2-1
- _ _ - -
POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CON 0! TION FOR OPERATION 3.2.1 ACTION: (Continued) With the indicated AFD outside of the above required target band for c.
more than 1 hour of cumulative penalty deviation time during the previous 24 hours and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by: Monitoring the indicated AFD for each OPERABLE excore channel at a.
least once per 7 days when the AFD Monitor Alarm is OPERABLE, and b.
Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereaf ter, when the AFD Monitor Alarm is
inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of: . One-minute penalty deviation for each 1 minute of POWER OPERATION a.
outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and b.
One-half-minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.
4, 2.1. 3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full-Power Days.
The provisions of Specification 4.0.4 are not applicable.
4.2.1.4 The target flux dif ference shall be updated at least once per 31 Effective Full-Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the predicted value at the end of the cycle life.
The provisions of Specification 4.0.4 are not applicable.
- (Continued) operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.
SEABROOK - UNIT 1 3/4 2-2 l t
i j l
1
! p 120 ' l
I ell I i l lijs I ! i l I $5__ l I lx i l Il
l U,NACCEPYABi.E OPEAATION l UNACCEPTABLE $ OPERdTIO'N j ! l i-ti.9e> l al.sei! l \\ l ! Ei i
I / '
,
i l l / T ' i k ' a 80 i l-1/ \\ - ! O ' a ! ! / ! - i \\ l ' [ ! l l /l ACCseTAete. oe:esATioN \\ l l l $ ' ]
- 60 '
I /i l I l\\ l a I e s tu i t I $ I-31.5 0) (31.5 0) l
i ! I I , ^ ! ! I I I I I I I I i l I l " a i l i l l l l x
I II I i l l I I I II
. ! I l I I I j ' i l l l l
i i
I l l l l l l l
-50-40-30-20-10
10
30
50 FLUX DIFFERENCE (aIW.
I [. FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF ) RATED THERMAL POWER SEABROOK - UNIT 1 3/4 2-3
Q r _:- : -
m- >
<6' q - . POWER DISTRIBUTION LIMITS-n 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION
3.2.2 F (Z) shall be limited by the following relationships:
F'(Z) $ 2.32 K(Z) for P > 0.5
P , F (Z) 5 (4.64) K(Z) for P 10.5
q . THERMAL POWER , and Where: .P = RATED THERMAL POWER t-K(Z) = the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY: MODE 1.
,y ACTION: j With F (Z) exceeding its limit:
a.
Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the q limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower aT Trip Set-
points have been reduced at least 1% for each 1% F (Z) exceeds'
_ j the limit, and l b.
Identify and correct the cause of the out-of-limit condition ! prior to increasing THERMAL-POWER above the reduced limit re- ' i quired by ACTION a,, above; THERMAL POWER may then be increased, provided F (Z) is demonstrated through incore mapping to be
9 within its limit.
-l ! ! ! - .. ' SEABROOK - UNIT 1 3/4 2-4
. ~________i___________.________.__
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2
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CORE HEIGHT (FT) ! I l FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT .; q SEABROOK - UNIT 1 3/4 2-5 l k ____ _ _ _. - _ _ _ _ - _. _.
.3
_ i r . ( POWER DISTRIBUTION LIMITS { -) HEAT FLUX HOT CHANNEL FACTOR - F (Z) SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.
4.2.2.2 F shall be evaluated to determine if F (Z) is within its limit by: q Using the movable incore detectors to obtain a power distribution a.
map at any THERMAL POWER greater than 5% of RATED THERMAL POWER, b.
Increa;ing the measured F component of the power distribution map xy by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, c.
Comparing the F computed (F ) obtained in Specification 4.2.2.2b., above, to: 1) The F limits for RATED THERMAL POWER (FRT ) for the appropriate measured core planes given in Specification 4.2.2.2e. and f., below, and 2) The relationship: F =FRTP [1+0.2(1-P)], ' Where F ' is the limit for fractional THERMAL POWER operation
- Y RTP'
expressed as a function of F and P is the fraction of RATED xy THERMAL P.0WER at which F was measured, x d.
Remeasuring F according to the following schedule: RTP 1) When F is greater than the F limit for the appropriate measured core plane but less than the F relationship, additional power distribution maps shall be taken dF compared to F and F either: ) a) Within 24 hours after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which F was last
- Y determined, or b)
At least once per 31 Effective Full-Power Days (EFPD), whichever occurs first, a SEABROOK - UNIT 1 3/4 2-6 i _ __
- - - __ -__ ___ _ -___ ) l POWER DISTRIBUTION LIMITS i ' HEAT FLUX HOT CHANNEL FACTOR - F (Z) SURVEILLANCE REQUIREMENTS l 4.2.2.2d. (Continued) P 2) W'ien the F is less than or equal to the F limit for the appropriate measured core plane, additional power distribution P maps shall be taken and F compared to F and F at least once per 31 EFPD.
e.
The F limits for RATED THERMAL POWER (F P) shall be provided for all core planes containing Bank "0" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specifica-tion 6.8.1.6; , f.
The F limits of Specification 4.2.2.2e., above, are not applicable xy in the following core planes regions as measured in percent of core height from the bottom of the fuel: 1) Lower core region from 0 to 15%, inclusive, ! ). 2) Upper core region from 85 to 100%, inclusive, 3) Grid plane regions at 17.8 2%, 32.1 2%, 46.4 2%, 60.6 2 2%, i and 74.9 1 2%, inclusive, and 4) Core plane regions within 2% of core height (t 2.88 inches) about the bank demand position of the Bank "D" control rods, g.
With F exceeding F ', the effects of F on F (Z) shall be evaluated xy q to determine if F (Z) is within its limits.
q 4.2.2.3 When F (Z) is measured for other than F determinations, an overall
measured F (Z) shall be obtained from a power distribution map and increased
by 3% to account for manufacturing tolerances and further increased by 5% to { account for measurement uncertainty.
' l l % SEABROOK - UNIT 1 3/4 2-7 l , _
, - - _ - - _ _ _ - _ - _ _ _ _ _ _ I / l l POWER DISTRIBUTION LIMITS L 3/4.2.3 NUCLEAR ENTHALPY RISE' HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION Fhshallbelessthan 1.49 [1.0 + 0.2 (1-P)]. '3.2.3 Where: P=' THERMAL POWER RATED THERMAL POWER APPLICABILITY: MODE 1.
ACTION: q WithFhexceedingitslimit: 3.
Within 2 hours reduce the THERMAL POWER to the level where the LIMITING CONDITION FOR OPERATION is satisfied.
. b.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the limit required by ACTION a.l, above;THERMALPOWERmaythenbeincreased,providedF$g is demonstrated.through incore mappin'g to be within its limit.
m .L SURVEILLANCE REQUIREMENTS, 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.
4.2.3.2Fhshallbedemonstratedtobewithinitslimitpriortooperation j above 75% RATED THERMAL POWER after each fuel loading and at least once per j 31 EFPD thereafter by: ' a.
Using the movable-incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% RATED THERMAL POWER.
UsingthemeasuredvalueofFhwhichdoesnotincludeanallowance b.
for measurement uncertainty.
,
! - L i i k ' SEABROOK - UNIT 1 3/4 2-8 Q---.--^-L-_-_-- .- _
. - _ _ - - - _ _. _-- _-_ - _ _ - ._.
_ _.. _ - _. - _ _ - _ _ _ _ _ -___ _ - - - _ _ _ _ - _ _ - - _ _
POWER DISTRIBUTION LIMITS , . ! 3/4.2.4'~ QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION i 3.2.4 The QUADRANT POWER -TILT RATIO shall not exceed 1.02.
APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER *. ACTION: With the QUADRANT POWER TILT RATIO determined to exceed 1.02: a.
Within 2 hours reduce THERMAL POWER at least 3% from RATED THERMAL ~ POWER for.each 1% of indicated QUADRANT POWER TILT RATIO in excess of I and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours.
, b.
Within 24 hours and every 7 days thereafter, verify that F (Z) (by
F,y evaluation) and F are within-their limits by performing Surveil-H lance Requirements 4.2.2.2 and 4.2.3.2.
THERMAL POWER and setpoint ' reductions shall then be in accordance with the ACTION statements of.
Specifications 3.2.2 and 3.2.3.
SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by: a.
Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and b.
Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable.
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm indicated QUADRANT POWER TILT RATIO at least once per 12 hours by either: a.
Using the four pairs of symmetric thimble locations or b.
Using the movable incore detection system to monitor the QUADRANT POWER TILT RATIO subject to the requirements of Specification 3.3.3.2.
.
- See Special Test Exceptions Specification 3.10.2.
SEABROOK - UNIT 1 3/4 2-9 , ~~.--_-.---:..____.___.,__
_ _ _ _ _ t . POWER DISTRIBUTION LIMITS 3/4.2.5 ON8 PARAMETERS
LIMITING CONDITION FOR 0PERATION
3.2.5 The following DNB-related parameters shall b.e maintained within the the following limits: Reactor Coolant System T,yg, 5, 594.3 F a.
a b.
Pressurizer Pressure, > 2205 psig c.
Reactor Coolant System Flow, > 391,000 gpm** ' APPLICABILITY: MODE 1.
ACTION: - Wit'h any of the above parameters exceeding its limit, restore the parameter to' within its limit.within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours.
d SURVEILLANCE REQUIREMENTS i-4.2.5.1 Each of the parameters shown above.shall be verified to be within its limits at least once-per 12 hours.
Additionally, Reactor Coolant System flow shall be demonstrated to be within its litnit prior to operation above 75% of ) RATED THERMAL POWER after each fuel loading.
The provisions of Specifica- ' tion 4.0.4-are not applicable for the verification that Reactor Coolant System flow is within its limit.
4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL CALIBRATION j at least once per 18 months.
4.2.5.3 The RCS total flow rate shall be determined by precision heat l balance measurements at least once per 18 months.
- Limit not applicable during either a THERMAL POWER ramp in excess of 5% of l.
RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10%
of RATED THERMAL POWER.
l j
L
- Includes a 2.1% flow measurement uncertainty.
SEABROOK - UNIT 1 3/4 2-10 i _ _ - _ _ _ _ - _ - _ _ _. k
- - - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ _ _ _ _ _ _ _ __ __ _ _ QUESTION 8,05 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE i LEAKAGE CETECTION SYSTEMS LIMITING CONDITION FOR OPERATION l 3.4.6.1 The following Reactor Coolant System Leakag6 Detection Systems shall ' be OPERABLE: The Containment Atmosphere Particulate Radioactivity Monitoring a.
System, b.
The Containment Drainage Sump Level Monitoring System, and c.
Containment Radioactive Gas Monitor APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION: With only two of the above requirec Leakage Detection Systems OPERABLE, operation may continue for up to 30 days provided grab samples of the contain-ment atmosphere are obtained and analyzed at least once per 24 hours when the required Gaseous or Particulate Radioactive Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following c' ours.
SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by: Containment Atmosphere Gaseous and Particulate Monitoring Systems - a.
performance of CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and b.
Containment Drainage Sump Level Monitoring System - performance of CHANNEL CALIBRATION at least once per 18 months.
, SEABROOK - UNIT 1 3/4 4-20 _-
_ k I ' -REACTdRCOOLANTSYSTEM l L REACTOR'C00LANT SYSTEM' LEAKAGE ,- , ! OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION a 3.4.6.2 Reactor Coolant System leakage shall be limited to:- a.
No PRESSURE BOUNDARY LEAKAGE, 'b.
1 gpm UNIDENTIFIED LEAKAGE,
1 gpm total reactor-to-secondary leakage through all steam generators c.
and 500 gallons per day through any one steam generator, d.
10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, e.
40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of , 2235 nsig-20 psig, and I f.
0.5 gpm leakage per nominal-inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2235 ! 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.
' APPLICABILITY: H0 DES 1, 2, 3, and 4.
? ACTION: a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours, b.
With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE ~and leakage from . Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
, $ With any Reactor Coolant System Pressure Isolation Valve leakage.
c.
- greater than the above limit, isolate the high pressure portion of
the affected system from the low pressure portion within 4 hours by ' use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, i l } ' SEABROOK - UNIT 1 3/4 4-21 - ____:__:_________ __ - _
__ - _ __- - _ _ _ _ l I l . , REACTOR COOLANT SYSTEM 4' REACTOR COOLANT SYSTEM LEAKAGE OPERATIONAL LEAKAGE
, SURVEILLANCE ~ REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by: , t a.
Monitoring the containment atmosphere particulate radioactivity f monitor at least once per 12 hours; l b.
Monitoring the containment drainage sump inventory and discharge at l least once per 12 hours; - c.
Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump J seals when the Reactor Coolant System pressure is 2235 1 20 psig at ' least once per 31 days with the modulating valve fully open. The
provisions of Specification 4.0.4 are not applicable for entry into i MODE 3 or 4; i d.
Performance of a Reactor Coolant System water inventory balance l within 12 hours after achieving steady-state operation * and at least: ! once per 72 hours thereaf ter during steady-state operation, except that not more than 96 hours shall elapse between any two successive i ' inventory balances;,and Monitoring the Reactor Head Flange Leakoff System at least once per e.
24 hours.
. ! i l
- T being changed by less than 5 F/ hour.
avg SEABROOK - UNIT 1 3/4 4-22 !~
_.. _ _ _ _..... _ _ _ _ _ _ _ _ _
_ - _ _ - _ - _ _ _ _ _ - _ __ . _ _ _ _ - _ - _ _ - _ _ _ - _ - _, _ _ _ _ - - - ____ I ' J I , l ! { REACTOR COOLANT SYSTEM i REACTOR COOLANT SYSTEM LEAKAGE , OPERATIONAL LEAKAGE i
l SURVEILLANCE REQUIREMENTS l 4.4.6.2.2 Each Reactor Coolant System Pressure Isol'ation Valve specified in Table 3.4-1 shall-be demonstrated OPERABLE by verifying leakage to be within its limit: a.
At least once per 18 months, r b.
Prior to entering'H0DE 2 whenever the plant has been in COLD ' SHUTDOWN for 72 hours or more and if leakage testing has not been performed in the previous 9 months, , Prior to returning the valve to service following maintenance, c.
repair, or replacement work on the valve, and Within 24 hours following valve actuation due to automatic'or manual . I d, action or flow through the valve.
As outlined in the ASME Code, Section XI, paragraph IWV-3427(b).
e.
The provisions of Specification 4.0.4 are not applicable for entry into. MODE 3 or 4.
4 i SEABROOK - UNIT 1 3/4 4-23 _.
_ _ _ - _ - _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ - -
_ _ - - _ __ _ _ - _ _ - _ _ _ _ _ .. TABLE 3.4-1-REACTOR COOLANT SYSTEM PRESSURE ISOLATION V'ALVES.
.. MAX, ALLOWABLE . VALVE VALVE NUMBER SIZE FUNCTION LEAKAGE (GPM1 > SI-Vik4' 1-1/2" SI'to RCS Loop 1 Cold-Leg Injection 0.75 .SI-V148 1-1/2" SI to RCS Loop 2 Cold-Leg In'jection 0.75 l SI-V152 1-1/2" SI to RCS Loop 3 Cold-Leg Injection-0.75 il 1 < - ' SI-VI56 1-1/2" SI to RCS Loop 4 Cold-Leg Injection 0.75 . SI-V81-2 SI to RCS Loop 3 Hot-Le'g Injection 1.0 A SI-V86 2" SI to RCS Loop 2 Hot-Leg Injection 1.0 SI-V106 2" .SI to RCS' Loop'4 Hot-Leg _ Injection-1.0 SI-V110 2" SI to RCS Loop 1 Hot-Leg Injection' 1. 0 l SI-V118 2" SI to RCS Loop 1 Cold-Leg Injection 1.0 SI-V122 2" SI to RCS Loop 2 Cold-Leg Injection 1.0 - SI-V126 2" SI to RCS' Loop 3 Cold-Leg Injection-1.0 SI-V130 2"- 51 to RCS Loop 4 Cold-Leg Injection 11. 0 SI-V140.. .3" --51 to RCS Cold-Leg Injection 1.5 . SI-V82 6" SI to RCS Loop'3 Hot-Leg Injection 3.0 SI-V87-6" SI to RCS ~ Loop 2 Hot-Leg Injection 3.0.
RH-V15-6" RHR to SI Loop 1 Cold-Leg Injection 3.0-' RH-V29 6" RHR to SI Loop 3 Cold-Leg Injection' 3.0 RH-V30 6" RHR to SI Loop 4 Cold-Leg Injection 3.0 RH-V31 6" RHR to SI Loop 2 Cold-Leg Injection 3.0 RH-V52 6" SI to RCS Loop 1 Hot-Leg-Injection 3.0 RH-V53 6" SI to RCS Loop 4 Hot-Leg Injection 3.0 RH-V50 8" RHR to RCS Loco 4 Hot-Leg Injection 4.0 RH-VS1 8" RHR to RCS Loop 1 Hot-Leg Injection 4.0-SI-V5
SI to RCS Loop 1 Cold-Leg Injection-5.0 i d SI-V6.
10" SI Tank 9A Discharge Isolation 5.0 l SI-V20 10" SI to RCS. Loop 2 Cold-Leg Injection 5.0 i SI-V21 10" SI Tank 98 ?ischarge Isolation 5.0 31-VS5 10" SI to RCS Loop 3 Cold-Leg Injection 5.0 l SI-V36 10" SI Tank 9C Discharge Isolation 5.0 l .' SI-V50 10" SI to RCS Loop 4 Cold-Leg Injection 5.0 SI-VS1' 10" SI Tank 90 Discharge Isolation 5.0 RC-V22 12" RHR Pump 8A Suction Isolation 5.0 RC-V23 12" RHR Pump 8A Suction Isolation 5. 0 , RC-V87 ,12" RHR Pump 88 Suction Isolation 5. 0
RC-V88 12" RHR Pump 88 Suction Isolation 5.0- < . I o iz O l SEABROOK - UNIT 1 3/4 4-24 h . ' ' _ _ _. _ _. _ _ _. _ _, _ _ _ _ _ _
g- -. .u ' f = ma' v = s/t-Cycle efficiency i (Network l-IJ..,, out)/(Energy in) s...mg s = V t + l/2 at - o
E = mc
L KE =.1/2.'mv a = (Vf - V )/t A = AN A=Ae g g . PE =lm9n ! , g , . Vf = V, +. a t w = e/t A = tn2hl/2 = 0.693/t1/2 ' 1/2*ff " b(tin)(t)3 ] ' ' t h - ((t1/2) + (t )) ' b , AE = 931 am ~
. I =-I e O n . - { Q = mCp a t i = UA a T I = I e~"* g , Pwr' = W an I = I, 10-x/TVL ' f TVL = 1.3/u P = P 10 "#(*) HVL = -0.693/u
g El ' P-= P e o ' SUR = 26.06/T SCR = S/(1 - Keff) CR = S/(1 - Keffx) x SUR = 269/t* + (8 - o)T CR)(1 - Keffl) = CR (1 - keff2)
T = (t*/o) + [(8 - p)/ o] M = 1/(1 - K,ff) = CR)/CR g ~T = 1/(o - 8) M = (1 - Keffo)/{I ~ Keffl) T = (a - o)/(lo) SOM = ( -Keff)/Keff " o = (K,ff-1)/K,ff = d,ff/K,ff t' = 10 seconds T = 0.1 seconds-I o. =-[(t*/(T K,ff)] + [T,ff (1 + $T)] j / P = '( I+ V )/ ( 3. x'.'1010 ),. I)d) 2,,2 2 =1d . I ) d )' , cg d ' ' -' '
2 I = oN R/hr = (0.5 CE)/d (meters) i Water Parameters Miscellaneous Conversions ' 1 gal. = 8.345 lbm.
1 cur'ie = 3.7 x 1010eps , 1gaj.=3.78 liters 1 kg = 2.21 lem, Btu /hr ' = 7.48 gal.
I hp = 2.54 x 10' 1 ft Density = 62.4 lbm/ft3 1 mw = 3.41 x 106 Btu /hr - Density = 1 gm/c.d lin = 2.54 cm Heat of vaporization = 970 Btu /lom 'F = 9/5'C + 32 Heat of fusion = 144 Btu /lbm 'C = 5/9 (*F-32) l Atm = '4.7 psi = 29.9 in. Hg.
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R :e l[ ' t ' ( ' TEST CROSS REFERENCE PAGE
\\ ! QUESTION VALUE REFERENCE 05.01 2.00 DUD 0001786
105.02 2.00 ' DUD 0001788 05.03 2.40 DUD 0001789 05.04 2.50 DUD 0001790 05.05-2.00 DUD 0001791 05.06 2.00 DUD 0001794 l 05.07 3.00 DUD 0001795 i 05.08 2.00 DUD 0001796 l ~05.09 1.50 DUD 0001797 I 05.10 1.60 DUD 0001787 05.11 2.00 DUD 0001792 ' 05.12 2.00-DUD 0001793 ______ 25.00 06.01 1.50 DUD 0001802 06.02-2.50 DUD 0001803 06.03 2.80.
DUD 0001806-06.04 2.60 DUD 0001808 06.05 3.00 DUD 0001814 06.06 '2.10 DUD 0001798 06.07 3.00 DUD 0001801 06.08 2.50 DUD 0001804 06.09 2.00 DUD 0001799 06.10 3.00 DUD 0001800 ______ 25.00 07.01 1.20 DUD 0001809 i 07.02 2.00 DUD 0001810 ' 07.03 3.40 DUD 0001811-07.04 1.50 DUD 0001815 07.05 3.00 DUD 0001816 07.06-3.50-DUD 0001817 07.07 2.80 DUD 0001818 07.08 2.50 DUD 0001819 07.09 2.10 DUD 0001829 07.10 3.00 DUD 0001831 s ______ 25.00 08.01 1.60 DUD 0001812 08.02 2.00 DUD 0001813
08.03 1.30 DUD 0001820 { 08.04 3.00 DUD 0001821 j 08.05 2.00 DUD 0001822 - 08.06 2.10 DUD 0001823 ! 08.07 2.20 DUD 0001824
08.08 3.00 DUD 0001825 l 08.09 1.50 DUD 0001826 ! I i k i - - - - - a
_ . -. - __ __ ~ j TEST CROSS REFERENCE . ;.__ r ' PAGE
! , ! ' QUESTION-VALUE REFERENCE , ! - ! ________ ______ ___.______ 08.10 2.40 DUD 0001827 08.11 2.00-DUD 0001920 ' 08.12 .1.90 - DUD 0001850- ) ______ 24.00 ______ _ _ WW WIBEWe died 100.00 i ' .,
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, , _ , -_ _________-___ ___ - l >* ^ ATTAClit1EWT 3; L *" .i " g g Vice President Nuclear Production p-r , Pihte Service of New Hampshire N9w Hampshire Yankee Division August 25, 1987 Mr. Robert M. Keller, Chief . Projects Branch No. 1 ' Division of Reactor Projects U.S. Nuclear Regulocory Commission gf.
(631 Park Avenue
T' -Ring of Prussia, PA 19406 s .<- ' D bject: Review of NRC Written Exam d
. -n e N$$hMr.Keller: l' . n <.s \\ f <do11owing 'the administration of the Performance and Written exams ' administered between August 11-and August 19 at Seabrook Station, the Training H 'Centen Staff was provided'an opportunity to review the written exam.
'As a result of this review, several comments were generated. Since the Written exam was administered after completion of the Performance exams', the examiners left the_ site with very little time available for resolution of the comments, as has been past practice. For this reason, we are forwarding all-of the comments generated durinb the exam review for your~ consideration during the grading process as an enclosure to this letter. References to support comments are provided where appropriate.
Should you havesany questions concerning this information, please contact i Mr. Peter Richardson l Seabrook Training Manager, at (603) 474-9521, extension ' 2605.
Very truly yours, J -
eo S. Thomas Enclosure r
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P.O. Box 300. Seabrook, NH 03874,. Telephone (603) 474 9574 t -_ ___ __ ___ ____________ _ ________ _ _._____
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, 1987 NRC COLD LICENSE EXAM COMMENTS i i ]s RO '4 > , s , > < s
q @ction 1' 1.0 ~ ] Should aise accept cooldowry rate in lieu of temperature.
'l 9 ? ) J + DISPOSitCN: Comment will be,incorporeteo.
, t hi <l
t - s i I s =\\ [ t. O ' 1.03 in additida to the thr #. factors listed, f ue.[ pollet swell or fracture couldi be listed.
Since :ldensification occurs ' relatively early in core ,./ ' life. ) dnd swell occurs later In core life time with the opposite effect { ' ', on fuel temp (FT + as swell occurs) either answer should be accepted / , - as correJt.
s<' / ./ . ,, ' DISPOdlTION: Cleti creep will cover this.
- i
' y' . ' ) ' . 1.05 Additionally the five items on the attachment ES-0.1 ' OlSPOSITION: Comment will be iTorporated.
l ) s c i , l 1-
,, e ( y < 1.07 Additt30all;/s.)rocess similar to the following should be acceptable: ' r,
S ( ,T '- , ,. if delayed n%trosi faction increases => Beff increases, \\( ' g ( and P h : As B Increases => T increases => response will 6e' slower.
' , A-p gy J a DWPDtth0N: Comment wil! be incorpor6:eds 34 - - < )
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1) < i )', tf l }
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f ., .( 1.08 The correalve ac.tlon (accounting) for the positivq gactivity addition J Should be to borum to cold shutdown boron concentrabn. The withdrawal
, of the shutdown banks is an early warning system which will provide rapid ! . , capability to insert -p in the event that counts increase dt. iring cooldown, , thus rod withdrawal should not be required.
, b j ,. , DISPCOTION: Comment will be incorporated.
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.. .j j f ( l I ' { l ,a l 1.10b To p;ovita more linear reactivity addition durir.g contrei rod withdrawal.
I ' ' +
, , , DISPOSITION: Comment t rill be incorporated.
(
> . , l , j ' ' ( , ' l 1.11b Since use of th h rr,ollier diagram is an acceptable methcc 'of determining { ' t?o restibtive.
Recommend t or - 5 F.
i 12 F \\lfs temp, a band of ' j^ wo-w oi ' l-1-J ' , , m ed _ _ _ _ _ _ _ _ _ _
m ., ,
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M%p: 4- . 1987 NRC COLD LICENSE EXAM COMMENTS' Dj0 k . .RO- ' yw J' c Sectiotk 2-
- 4
,, 2.01 '- . Question asks. for two, design bases, ~ answer gives. only_ one basis broken-4 into two parts.
<
, If$
- DISPOSITION:-. Comment will'be incorporated.
(4
- 2.02a Should accept equivalent noun names for valves in lieu of numbers.
DISPOSITION: Comment will be ' incorporated.
'! j ,,
2.02b Word ?" constant" should'. not be required-for full credit, valves function.
pjhu
to maintain back pressure.
DISPOSITION: Comment'will be incorporated.
, f ' 2.03c j 2nd part of answer gives two components,'should;only require one for full credit. (Question asks for four total.). eu DISPOSITION: Comment will be incorporated.. . o, ..:., ' % - 4, 2.04 Battery charger number should not be required for full credit, i DISPOSITION:. Comment will be incorporated.
- j i
i i i 2.08a Question improperly worded - Normal Design flow ls 235 gpm; normal . maximum flow is limited by hi-flow isolation logic which is 450 gpm.
'i DISPOSITION:' Comment will be incorporated.
2 2.10 b ' Fan.does not remove H, recombiners remove H. DISPOSITION: Comment will be incorporated.
, (A10-141) ' -2-i 1T
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I ! 1987 NRC COLD LICENSE EXAM COMMENTS RO Section 3 3.01a - Question is worded to elicit a calculation, not a discussion of P-12.
DISPOSITION: Comment will be incorporated.
3.01b 12%oF of 561oF should be accepted based on steam table interpretation and calculation of AT (Tave-Tstm).
DISPOSITION: Comment will be incorporated.
3.01c Same as "b" above and Steam Dumps will not restore Tave to 5570F until operator action is taken to restore the normal lineup.
j DISPOSITION: Comment will be incorporated.
. 3.02a " Lithium" should not be required for full credit. - (Just a description of positive and negative ions).
DISPOSITION: Comment will be incorporated.
l l 3.02b Also may effect startup based on overlap.
DISPOSITION: Comment will be incorporated.
i . 3.03c 15% with 2 second time constant should be accepted.
{ DISPOSITION: Comment will be incorporated.
3.03e - 10% turbine or 10% Rx allows un block of P-7.
DISPOSITION: Comment will be incorporated.
(A10-14 2) -3- _ - _ _ _ _ _
r.. ,... e I 1987 NRC COLD LICENSE EXAM COMMENTS RO [., a , Section 3 continued
- 3.05 Low pressurizer pressure Si setpoint is 1875 psig.
! DISPOSITION: Comment will be incorporated.
'3.06 Operator has - no idea.what 'has been performed prior to ~ statement,- especially part A.
DISPOSITION: Comment will be incorporated.
3.06a Operator has no way.of knowing If rod blocks have been properly defea'ted.
DISPOSITION: Comment will be incorporated.
3.06b - P-13 is incorrect,, C-5 is the proper designation.
DISPOSITION: Pending NRC review, . 3.06c - Wrong answer - Westinghouse functionals C2 Rod Block 103% power,l/4 power , J range detectors.
DISPOSITION: Comment will be incorporated.
! I i l 3.07.
Control room ventilation isolation does not occur as a result of an 'S' signal.
High radiation of the intake is the only auto isolation signal.
DISPOSITION: Pending NRC review.
l , h j t i i (At0-14 3) -4-o___________ __ i
_ -_ - - ' g -, . . . , . , E 1987.NRC COLD LICENSE EXAM COMMENTS: RO +
J J
- Section 4 '
. 4.01 a ' ' A' discussion on vibration problems:should be acceptable.
DISPOSITION: Comment will be incorporated.- 4.03.
SI' termination - PZR level > 5% per E-1, ES-1.1 DISPOSITION: - Pen' ding NRC review.
'
- lf MS-V127.or,MS-V128 Is closed, then they will open upon loss of air.
c4.09 DISPOSITION: Comment will be incorporated.
'4.10c No is the correct answer.
Reference: _ Executive Volume Users guide, page 2 and page 5 DISPOSITION: Comment will be incorporated.
4.10d The. operator actions summary eage applies to a given guideline series, - i.e. E-1, ES-1.1, ES-1.2, ES-1.3 : and : ES-1.4, but not to a different guideline series. Additionally, " Symptomatic Response / Unexpected Conditions Page" is - not plant specific terminology." Operator Action Summary" is the ' correct , plant specific terminology.'
DISPOSITION: Comment will be incorporated.
i i , (A10-14 4) -5- ! _ _ _ _ _ _ _ _ _ _ _ _ _ __._ _ _ _ _ _. _
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- , - , 1987 NRC COLD LICENSE EXAM COMMENTS - < SRO- ' q , ' Section 5 ~ . ~ 5.01 . Tolerance band for answer not stated on answer key. (some deviation from j answer should. be. acceptable.
. Provided. ' solution method -is correct).- ~l . l ! ! DISPOSITION: Comment will be incorporated.
! ! 5.02a SDM, as defined by Tech Specs, should Increase as RCS boron. concen-tration is increased, even though net reactivity remains constant (i.e., reactor remains critical =>Keff = 1 => pnet = 0) This is illustrated' by the following calculation:' initial conditions:. 85% power, manual rod control ARO, 'Tavg on program (i.e. Tavg (85%) = 583.7), DBW = - 10 pcm/ ppm.
[ NOTE: all Rx physics data taken from Seabrook Nuclear Design Report'- NDR).
' SDM before Boration: SDM = -prods'+ Pdoppler + Amtc . " -Prods + Ppower' defect (Tavg on program) SDM = - 7700 pcm..+ 1170 pcm (NDR,LP. 6-9). (NDR, Fig. 5.15) SDM = 1.6530 pcm = - 6.53%. Ak/K ~ ! SDM after Boration: Assume operator inadvertently increases RCS CB by 10 ppm I 10 ppm X - 10pcm/ ppm = 1100 pcm To remain critical, Tavg decreases (and power drops somewhat due to lower available Pstm), thus adding enough + p to counteract - 100 pcm.
Assume MTC = -8 pcm/oF. at these conditions (NDR, Fig. 5.10) = -10 pcm/% at these conditions (NDR, Fig. 5.12) Assume BD ) A reasonable combination of these effects is to lower Tavg by approximately 100F (+ 80 pcm), and to ' reduce reactor power by approximately 2% (919 psig available Pstm instead of 1000 psig) -> +20 pcm.
Conditions after boration : CB = 1000 ppm Tavg = 573.7 Power = 83% {A10-14 5) ' -6-I __ _-_
-. _ = N a , . , i < ' ' -1987 NRC' COLD LICENSE EXAM COMMENTS .SRO Section 5 continued , SDM = -7700 pcm + MTC feedback + Doppler feedback' (trip) . (trip)
- SDM.=
7700 pcm + 25 pcm +.1000 pcm (NDR, Fig 5.5)(NDR, Fig 5.13)- . SDM = - 6675 pcm = - 6.675% Ak/k Thus, boration has caused T/S SDM to increase, due.to a smaller available instantaneous trip. reactivity from -defects (lower. Tavg and Power).for the same available rod worth, DISPOSITION; : Pending NRC review.
. 5.04b ' The question states that natural circulation has been lost, yet the answer . key states, that Pstm will decrease as bolloff occurs" in the ? SGs.
If - In fact the SGs f are still steaming, then there has to be heat transfer from the RCS Into the SG, which would serve as RCS natural ~ circulation thermal driving head.
Loss of; feed flow by itself will not result'iri a'short term loss of natural circulation.
Thus, the student = should not be penal'ized' for making a general statement that Tc and ~.Tstm will' no longer be approximately equal, without further - Identifying individual trends.
DISPOSITION: Pending NRC review.
5.04c This question is difficult to interpret in that it does not define the.
expression "ald natural ciruclation", (i.e., increase actual RCS flow, prevent any degradation of exising flow, or merely allow the operator to feel more confident without actually affecting RCS flow?). In addition, even if the license condidate fully' understands natural theory,. the answers identified on the answer key may not be elicited for the following reasons.
(A10-14 61-7- _ = _
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I 4: 1987. NRC COLD LICENSE EXAM COMMENTS l ' SRO- , ^;; -. Section 5 continued LITEM 2L "RCS cooldown rate 400F/Hr."
According to the E-0 and ES-0.2 background documents, RCS cooldown rates lessq then 500F/hr.are satisfactory for maintenance : of natural ' circ- - ulation (N.C.) conditions (along with various other parameters). The answer key' states.that decreasing the cooldown rate aids N.C.
In fact the thermal ~ driving head will begin to decrease as the cooldown rate is decreased, resulting - in less RCS flow. Thus, a better answer is.that increasing the cooldown , rate (up to 500F/Hr) results in a higher rate for RCS flow.
' El ITEM 4' "RCS pressure at 1900 psig" The answer key. states that N.C. cannot be aided by changing RCS pressure.
Depending upon the parameter of concern, this may not be true. While RCS pressure.itself does not appear on attachment A to ES-0.1, subcooling does, and so it 'may '.be desireable under certain conditions (l.e. no CRDM fans ' available) to increase RCS pressure to increase subcooling.
Furthermore, even.though Step 10 of ES-0.2 tells the. operator to maintain.
_ 1900 psig, the purpose of this. step is to stabilize plant conditions.following-the depressurization'in Step 8 to allow low Pstm and low Pzr pressure 'SI . . blocking.
Once. this has occurred, 1900 psig is maintained only long enough to verify cooldown rate, PZR level, and correct pressure - temperature ' (i.e., ' brittle fracture) relationships.
Once all of ~the N.C. Indicators have-been reverified (ES-0.2, Step 11),- further RCS depressurization is directed in Step 12, subject to subcooling and brittle fracture concerns.
_ In any case, since neither E-0, attachment A, nor ES-0.2 is called out in
~ the question, it is unreasonable to expect : the students :to be able to "second guess" which concepts are being examined.
DISPOSITION: Pending NRC review.
5.05b ." Margin to saturation" should not be part of the required answer since the definition of DNBR does not include this term.
DISPOSITION: Comment will be incorporated.
l t i 5.05c Same as above.
DISPOSITION: Comment will be incorporated.
. as ' ( A10- I 4.7) ' -8- _ _ _ _ _ _ _ _ - -- --
.. . _ - _ -, - __ -- - - _. . _ _ _ _ _ - - _ _ _ _ _ _ - - - .. . 1987.NRC C'OLD' LICENSE EXAM COMMENTS c SRO.
!. - ,
LSection'6 ,, - 6.02' The.. training staff feels that asking.'a license 1 candidate to explainiwhy ' 'a reactor trip occurs as a result of this i trenslent. Is misleading : the candidate. The NRC examiner concurred:and committed,to evaluating the ~ candidate based on expected riant response to the transient.
, DISPOSITION: Comment will be incorporated.
6.03a Based on operating experience.at Seabrook, rals'ing PCCW temperature results in an increase in surge tank level; ' DISPOSITION: : Comment will be incorporated.- - ' 6.04b With the "B" reactor trip breaker falling to open, the. "B*' reactor tripf SI-7- signal could not be reset.
DISPOSITION:. Comment will be incorporated.
6.05b Since there was no' specific RCP mentioned, simply stating th'at the associate'd ' ] leak-off isolation should be closed.
-I . i Also the P-8 setpoint at Seabrook is 50%. j DISPOSITION:. Comment will be incorporated.
6.08c OTAT setpoint decreased due to increased Tave vs dt.
DISPOSITION: Comment will be incorporated.
I . 6.09a The answer key is looking for repsonses that could occur only as a result of loosing PP-1A or PP-1B, The question stated that a trip could occur as a result of loosing any vital instrument bus. As a result the NRC examiner committed to considering any valid ' trip caused by' loss of any vital
instrument bus.
I j.
' DISPOSITION: ' Comment will be incorporated.
' ! l ! l , (A10-14 8) -9-l !. ! L_-_x-_-__ -
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( i 1987 NRC COLD LICENSE EXAM COMMENTS. SRO Section '6 continued.
l , 6.10 b '- The words "or" should be used in place of "and" between "UAT" and * RAT".
- DISPOSITION: Comment will be incorporated.
t 6.10d This question is. misleading in that there is only one condition that by-passes all trips except the.three mentioned, namely "Sl".
' .The EPS "looks" for an Si signal.
'If there. is one then all. the trips except the. three mentioned will be blocked.
if 'there is no. Si signal, regardless of other plant conditions, all trips will be functional for. the - EDG s.
Reference 1-NHY-310857, sheet E93/8e Ref rence 1-NHY-310102, sheet AS4d and AS4e DISPOSITION: Pending NRC review.
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_ _ _. _ _. - _ _ - _ - - < . s-g 1987 NRC COLD LICENSE EXAM COMMENTS ~ SRO' - Section 7 - 'l - 'j I ' ' '7.02i . Answer: -Question -does not' ask for NRC-4 limits, top two lines of ' key ~ are.
not asked for.
DISPOSITION: Comment will be incorporated.
I
7.03c - Typo in question. ' Assume it asks for " loss of offsite power".
Answer: Do not agree. Transition should be to go to ECA-0.0 to restore: AC power and monitor CSFTS.
l DISPOSITION: Comment will be incorporated.
l
k . . . t 7.05 Question - Asks for " general" actions, not specific.
Response.will vary greatly.
! . . a Answer should also accept a summary of ES-1.2, "Si termination" steps.- 'l Do not agree with all actions of key. What is reference for key?
- j DISPOSITION
- Pending NRC review.
l 'l 7.06a Question asks for NIS required by MPE. Answer gives NIS required by T.S.
j See prerequisites in MPE OS1000.07 (#5.2 and 5.5) j DISPOSITION: Comment will be incorporated.
.! 7.06c Answer should also accept - Notify Rx Engineering; or When error found ! and corrected, go to Step 7.1 . DISPOSITION: Comment will be incorporated.
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- . _ _ - - - - - _ - _ _ _. . , a '1987 NRC COLD LICENSE EXAM COMMENTS SRO
Section 7 continued
'
- 7.07a Based on wording, the candidate could assume the question asked for Heat Sink Red Path, also accept Red Path of Low Levelin S/G and < 500 gpm EFW flow to S/G.
! DISPOSITION: Comment will'be incorporated.
7.09 Question does not specify' Remote Safe Shutdown action, but assumes answer should be Step 1 of OS1200.02.
.q Answer should be changed to Steps a,b,c,d, see attached.
I DISPOSITION: Comment will be incorporated, j . 7,10b No immediate actions in Abnormal Procedures, question asks 'for immediate
action of dropped rod.
' Answer is correct for first two actions of Abnormal.
' DISPOSITION: Pending NRC review.
1 l I '! I l j,. . m-u i n - 12 - i a---_-_-. -
.. - - _. . . i )'.. l, 1987 NRC COLD LICENSE EXAM COMMENTS < . SRO ' , Section 8 - 8.01 a. This is a special circumstance which would be properly researched prior to - performance. - This is the. type of infrequent process that would require pro-- l
- cedural asistance and as such is not the sort of thing we have the operators
^
memorize.
DISPOSITION: Pending NRC review.
! . 8.02b This is a note to an LCO in the refueling section of T.S.
It'is not an 1 hour " ACTION" statement and therefore should not be required memory know-ledge. An operator would be able to look this up if implementation was ,.necessary. Also the questions-does not ask.for times and time is required in the answer.
I DISPOSITION: Pending NRC review.
j l a ' l 8.04 Everthing beyond the one hour time frame is information which we do not' l require the operator to memorize, the 24' hour and 14-day time frames both.
provide ample time to research an allow operations management the time to properly prepare the required reports.
DISPOSITION: Pending NRC review.
l l .j 8.05b Some of the tank names are not Seabrook specific and may add an element of I confusion. Also, RH-V87 does not exist, it should be RC-V-87.
I .. DISPOSITION: Comment will be incorporated.
I l ) 8.06b This does not involve an " ACTION" statement as implied in the question.
l This is a notation in the crew composition table and the operator would have this information available.
The question should state "What action is necessary...". [- DISPOSITION: Pending NRC review.
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-. _ _ _ _ . '4 1987 NRC COLD LICENSE EXAM COMMENTS SRO 'l Section 8 continued j
Il 8.08a These items are' not "immediate notification" (1 hour or less) and are not required to be known from memory. The operator would have adequate time to research and make the proper determination within the four hour time frame.
Answer is YES. Required operability for the mode does not matter if the system is still capable of performing its intended function.
This is ' a reportable 'occurance.- See NUREG-1022, Supplement No.1, Question 6.9.
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DISPOSITION: Pending NRC review, 8.12a ' Again' the requirement to replace the lock within 5-days of re-establishing access is' not.the type of activity that the training staff emphasizes. The procedure would be used for this type of activity.
DISPOSITION: Pending NRC' review.
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l l i . (A10-1413) - 14 - _ _ _ - _ _ _ _ - - _ _ _ _ _.
_ _ _ = L Attachment 4 L.- NRC Response to Facility Comments Question (RO) 'NRC Resolution
' 1.02 Incorporated facility comment.
1.03 Accepted either clad creep or fuel swell.
1.05 Incorporated facility comment.
1.07 Incorporated facility comment.
.j 1.08 Incorporated facility comment.
1.10 b Incorporated facility comment.
1.11 b Incorporated facility comment.
2.01 Incorporated facility comment.
2.02 a+b Incorporated facility comments.
2.03 c Incorporated facility comment.
In addition, answer key was expanded to include responses in accordance with System Description SW Pages 7, 11 and 12.
i 2.04 Incorporated facility comment.
2.08 a The question was extracted verbatim from facility training
material. An answer of 450 gpm will be acceptable with appropriate justification.
2.10 b Incorporated facility comment.
3.01 a No change to the answer key.
The question states "... explain HOW reactor coolant system temperature is controlled..." and does not require a calculation.
3.01 a,b,c Allowable deviation will be 12 F.
3.02 a+b Incorporated facility comments.
3.03 c+e Incorporated facility comments.
3.05 Incorporated facility comment.
Changed coincidence of low steam line pressure from "2/4 to "2/3" based on proper system configuration.
3.06 a,b,c Incorporated facility comments.
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Attach' ment.4?
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- t.O 3.07c Inco.rporated' facility' comment..However, the facility training material 'contains information the ' licensee stated was incorrect.
~ ' , 3.09.b The answer. key was: expanded to include the'following. answer;. ' "one train of RHR valves'are deenergized.open, so 'a failure.of PT-403.~or PT-405 will nottisolate'one train'of. RHR.",The~ change to:the answer key was based,on Technical Specification , Chapter 4.4.9.3.2.
) .) 4;01 a- . Incorporated facility. comment.
'j . .4.03'a: - . ,i The answer key was expanded to include Phase B isolation based- < on actual plant configuration.
4'
- .03'b-
' Incorporated facility comment.
- 4.' 05 ; c -
- The: answer key was expanded,to accept " Start boric acid:
ntransfer pumpi-open'CS-V-426, verify boration flow," ba' sed on .procedur'e PS-1202.04.
a 4.06 a The:. answer key was. expanded to include "or loss'of-flow'to RCP motor. oil and/or' air coolers high vibration" based on the system
' description.
'1 4.06 c_ 'The answer _ key was expanded to include "or trip the Rx'(1f i assumption is made that plant power > 50%)," as an alternate - acceptable: answer, ,4.09' Incorporated facility, comment.
-_4.10 c+d.
.' Incorporated facility comments.
4.11.a The answer. key was expanded to include blowdown radiation alarm.
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' i f- -3; Attachment'4 ' Question (SRO).
NRC Response to Facility' Comments 5.01'
- No~ change was made t'o the answer key,.howeve'r the comment was
, considered during grading of the. examination.
. 5.02 a Change answer key to read "SDM increases (0.3).
The decrease 'in the reactivity by'the boron.will be equal to the increase in the reactivity from the temperature and/or power < decrease. On the trip'less positive reactivity will be added.by the' power defect and/or MTC. (0.7)."
Facility' assumes that power is zero-and Taug is 557 instantaneously on a reactor trip.
Therefore, .the effects of MTC and doppler must be considered in calculating ! ' shutdown margin,
. 5.04 b No change was made to the answer key, however, the comment was'- ' considered during grading of the examination.
Partial credit was awarded based on the assumptions stated by each individual candidate.
- 5.04 c1 .The answer key was changed from decreased to increased based on
facility comment.
-5.04 c4 No change was made to the answer key.
Subcooling is a concern due to-DNB considerations and not directly associated with natural circulation.
5.05 b+c Incorporated facility comment.
. 5.06-b The answer key was expanded to include "MTC.if a secondary induced transient is assumed," since the coefficient which acts a e ' first to effect reactivity on a rapid power change is dependent on how the power change is initiated.
i [ 6.02-The answer key was changed to award the majority of credit for , ' the plant response to the transient and accepted any one of ( l three reactor trips.
i 6.03 Incorporated facility comment.
6.04 b Incorporated facility comment.
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6.05 b Changed answer key to reflect the proper set point for P-8 and added that the RCP must be stopped within 30 minutes in accor-dance with OS 1201.01.
- 6.08 e Incorporated facility comment.
6.09 a No change was made to the answer key, however, the comment was considered during grading of the examination.
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, . , { l4 ' Attachment 4 , f6._10 b-Incorporatedffacility comment.
,6.10'd; ' Question was deleted based on the facility comment.
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- 7.02-
- Answer key, as written doe's not require NRC Form 4 limits.
, - 7.03 c.
Incorporated facility comment.
7.05 Thecanswer key was rewritten to' read: ' Black: low MSLIS {0.5) . Reset SI (0.5)- - Restore RHR lineup by shutting suction valve from RWST {1.0) [any 3 @ 0.33.each)
- Shutdown'EDG Restore FW lineup for recircing the SG Stop'CR recirc fan A Restore COP 'or CAP:
.-Reset Phase "A" ' Answer is ba' sed on LER 50-443/86-002-00' and Detailed System , Text Pages IS-21 thru IS-31.
7.06 a No change made to the answer key.
MPE OS1000.07 prerequisite 5.2-requires. that " Technical specification ' mode change requirements for Mode 2 operation are complete."; Therefore, the NIS required by T.S. are the minimum number required to be' operable.
7.06 c Incorporated facility. comment.
7.07 a- . Incorporated facility comment.
' 7.09 Incorporated facil.ity comment.
27.10 b The answer key was changed to require " Reduce turbine load to maintain Tave/T ref."
Technical Specifications require the rod .j to be recovered within 1 hour and the candidate is expected to
'know how reactivity is controlled in this abnormal situation.
8.01 a No change was made to the answer key.
Candidates are expected -) to know escort responsibilities during plant operations.
' '8.02'b Answer key was. changed to remove the time requirement.
m 8.04 No change was made to the answer key.
Candidates are responsible to know generally the administrative requirements in Section 6 of the Technical Specifications.
-8.05 b-Facility comments incorporated.
The answer key was changed from "3.4.6.2.a" to "3.4.6.2.b" to correct a typographical error.
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Attachment 4
8. 06 b No change was made to the answer key.
Candidates are j responsible to know generally the administrative requirements ! in section 6 of the Technical Specifications.
8.08 No change was made to'the answer key.
Candidates are responsible to know the contents of internal memorandum which provide direction on how to perform their duties.
Reference memorandum Stasek to Walsh January 30, 1987. Additionally, candidates are responsible for understanding the causes and corrective actions taken for plant events which resulted in LER's, i , 8.12 a The answer key was changed to remove the time requirement.
Candidates are responsible to understand the administrative , requirements associated with key control.
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