IR 05000443/1985004
| ML20129E221 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/14/1985 |
| From: | Keller R, Kister H, Ruscitto D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20129E197 | List: |
| References | |
| 50-443-85-04, 50-443-85-4, NUDOCS 8506060388 | |
| Download: ML20129E221 (70) | |
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U.-S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING-EXAMINATION REPORT EXAMINATION REPORT NO. 50-443/85-04(0L)
FACILITY DOCKET NO. 50-443 FACILITY CONSTRUCTION PERMIT NO. CPPR-135 LICENSEE: Public Service'of New Hampshire
.P. O. Box 330 Manchester, New Hampshire 03105 FACILITY:
Seabrook 1
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EXAMINATION DATES: March 18-22, 1985 CHIEF EXAMINER:
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Reactor Engineer Examiner)
Date
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REVIEWED BY:
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1383 Chief, Projects Section 1C Date APPROVED BY:
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Chfef, Profebts Branch No.1 Datle (
SUMMARY:
Seventeen Senior Reactor Operators (SRO) and three Reactor Operators-(RO) candidates were examined. One SRO candidate failed the simulator portion of the exam. All other candidates passed all portions of the exam.
Five SR0 candidates were-identified as performing significantly above average.
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8506060388 850517 ADOCK O
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2 REPORT DETAILS
. TYPE.0F EXAMS:
Initial
" EXAM RESULTS:
I R0-l SR0 I
.I Pass / Fail-1.
Pass / Fail I
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l-10ral Exam I
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~ISimulator Examl 3/0
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1. ' CHIEF EXAMINER AT SITE:.D.-Ruscitto_(NRC)
2. 'OTHER EXAMINERS:
R. _ Sailor -(EG&G)
B. - Picker (EG&G).
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P. Isaksen (EG&G)
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R. Cochran.(NRC' Consultant-Observer)
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SUMMARY OF GENERIC DEFICIENCIES FROM ORAL AND SIMULATOR EXAMS:
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. Minor deficiencies with no safety significance were noted in the following areas:
Communications Classification of events using the Emergency Plan SS/USS liason during emergencies Verification of ERG parameters 2.
SUMMARY OF GENERIC DEFICIENCIES NOTED FROM WRITTEN EXAMS:
RO Exam (2 Candidates)
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-The following were areas of minor weakness:
Centrifugal Pump Operations Steam Dump Operation Dropped Rod Immediate Actions RCS Depressurization from RSSP SRO Exam (17 Candidates)
~The following were areas of minor weakness:
Worst Case Conditions for MSLB and Design Features which Limit Effects Indications and Automatic Action of EFW System T.S. Requirements / Bases for RHR System During Refueling Log Reviews by USS Dose Equivalent Iodine 3.
IMPROVEMENTS IN TRAINING PROGRAMS AS A RESULT OF PRIOR OPERATOR LICENSING EXAMINATIONS:
In response to comments from the December 1984 exam, additional emergency phones have been tied into the instructor's console providing more real-istic simulation.
4.
INTERFACE WITH PLANT STAFF DURING EXAM PERIOD:
' The simulator instructors continue to perform their liaison duties in an outstanding manner, greatly enhancing the examination process.
This allows last minute scenario modifications to examine new areas with little pause in the flow of the scenario.
The simulator malfunction list has a significant number of malfunctions with either no description or only a cursory description. These should be updatad. prior to the next exam in September.
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PERSONNEL PRESENT AT EXIT INTERVIEW:
NRC Personnel D. Ruscitto, Reactor Engineer (Examiner)
~ R. Barkley, Reactor Engineer A. Cerne, Senior Resident Inspector R. Cochran,-Consultant
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NRC Contractor Personnel R. Sailor, EG&G,-Idaho P. Isaksen, EG&G, Idaho B. Picker, EG&G, Idaho Facility Personnel P. Richardson, Training Manager
- R. Hanley, Training Supervisor L. Carlsen, Simulator Supervisor
.J. Grillo, Assistant Operations Manager
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SUMMARY OF NRC COMMENTS MADE AT EXIT INTERVIEW:
Generic deficiencies during oral / simulator exams were discussed. Training improvements and interface with plant staff were summarized as well.
Preliminary results of 15 candidates clearly passing and 4 candidates being marginal were presented.
5 candidates were identified as being clearly above average on their performance on the oral and simulator portions of the examination.
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EXAMINATION REVIEW:
At the conclusion of the written examinations, the examiners met with the -
following licensee personnel to review the exam and answer keys to identify any inappropriate questions relative to plant specific design and to ensure that the questions will elicit the answers in the key and that they reflect the most current plant conditions.
D. Schreiner~
S. Simonson J. Nichols R. Hanley J. Reagan R. Mayes
' Attachments:
1.
Written Examination (s) and Answer Key (s) (SR0/R0)
2.
Facility Comments on Written Examinations made after Exam Review
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ATTACHMENT 2
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-RO EXAM COMMENTS AND RESOLUTION:
_ Question 2.06B The start failure relay is also reset in addition to the _ answers provided 'on the key.
Reference Drawing was.provided at the exam review.
_ Question 3.048 The FWP discharge valves do not close on High S/G water level.
Reference _is Dwg. No. 9763-M-503582.
Question 3.09B
"Lo Lo S/G level" is appropriate also.
Referenc'e is FSAR, Table 15.0-6, Sheet 1.
Question 4.03A-
" Rapid drop in nuclear power" is an-additional answer. IAW procedure for dropped rod.
Answer changed to. include above phrase.
SRO EXAM COMMENTS AND RESOLUTION:
Question 6.038
_"0 TAT or Low Pressure Rx Trip" are acceptable.
Reference is FSAR, Table 15.0-6, Sheet 3.
Question 8.07 This question may _ be confusing in that Standing Order 83-004 requires the SS to. review the USS, A0, and two S/U ' Logs.
The turnover procedure (AQ10.003A) requires the SS to review the USS, Tagging, and Temporary ~ Modifications Log.
The reviews of AQ10.003A will be the accepted correct answers.
Question 8.08 Means of communicating may not match the answer given exactly.
In order that the candidates be evaluated on this area of the E-Plan that is not yet developed and. the appropriate equipment installed, discretion on the part of the examiner is required.
Parts B and C will be deleted as E
these procedures may change, however, parts A and D are still valid as these procedures are not expected to change.
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
.1EA3 GOOK 1_----
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REACTOR TYPE:
pWR-WEc4
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.DATE ADMINISTERED *_11493/is
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EXAMINER ISAKSEN.
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APPLICANT:
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IB11gEI1QU_ID_&PPLICAE1 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 guestion. The passing grade requires at least 70% in each category and a final grade of at losst 805.
Examination papers will be picked up six (6)
hours after
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the examination starts.
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3 0F CATEGORY
% OF APPLICANT'S CATEGORY vALUE_
TOTAL SCDgg___
VALUg__
___CAlggggY_
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_25. 0D _
25.00
_____ 1.
PRINCIPLES OF NUCLEAR POWER
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PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
_25. 00 _..21aga
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PLANT DESIGN INCLUDING S AFETY AND EMERGENCY SYSTEMS l
_22a90- _21agg
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INSTRUMENTS AND CONTROLS
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25.00
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PROCEDURES - NORMAL, ABNORMAL,
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EMERGENCY AND RADI0 LOGICAL CONTROL
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__ TOTALS
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FINAL GRADE
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All work done on this examination is my own. I have neither giv0n nor received ald.
APPLICANT S SIGNAiURE I
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE
IMER110D INAalC S4JfE A I_IRANSEER _ AND_ EL UID _ELOW QU5STIDM 1.01 (3.00)
Acstees.
1.
Rod control in manual.
2.
No operator action.
3.
No protective function actuations.-
4.
End of core life (EOL) conditions.-
Ccapare the FINAL values (HIGHER THANs LOWER THAN or the SAME AS)
. cith the INITIAL values of the parameters listed below for the following two transients.
EXPLAIN what CAUSES -these parameters
~to change from the inital values to the final values.
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PARAMETERS
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1.
~Teve 2.
'Tfuel
'3.
Reactor power
-~ TRANSIENTS (consider each separately)
c.
Steam dump valve f alls open with the reactor critical below the point of adding heat.
(1.45)
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b.
Control rod drops at 502 power with turbine controls in auto.
(1.55)
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i OUESTION 1.02 (1.50)
What are FIVE factors considered i n a MODE 3 or 4 Shutdown Margin Colculation?
(1.5)
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QUESTION '1.03 (2.00)
o.
Explain WHY the Moderator Temperature Coefficient becomes more negative over core life. (Two reasons required)
(1.0)
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b.
At 80L and as fuel temperature increasess WHY does the
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magnitude of the Doppler Temperature Coefficient decrease Per degree change in fuel temperature?
(1.0)
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ItiggggDINAgICsa,_tiEAT TRANSFER AlHLFLUID FLOW GUESTION 1 04 (2 50)
Cc pare the CALCULATED Estimated Critical Position (ECP) for a l
- otortup to be performed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> af ter a trip from 100* Powers to th3 ACTUAL critical control rod position if the following events /
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ornditions occurred.
Consider each independently.
Limit your answer to.ECP is HIGHER thans LOWER thans or the SAME as the ACTUAL critical centr ol rod position.
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The FOURTH coolant pump is started two minutes prior to criticality.
(0.5)
b.
The startup is delayed until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, af ter the trip.
(0.5)
c.
The steam dump pressure setpoint is increased to a value Just below the Steam Generator PORV setpoint.
(0.5)
d.-
Condenser vacuum is reduced by 4 inches of Mercury.
(0.5)
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All Steam Generator levels are rapidly being raised by 51 as o.
criticality is reached.
(0.5)
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. eUESTIDM 1.05 (1.50)
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Provide TWO reasons for Xenon contributing more negative i
reactivity than Samarium et full power.
(1.0)
b.
What would happen to the magnitude of the equilibrium Samarium Concentrations if reactor power was changed from $0% to 10011 (0.5)
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QUESTION 1.06 (1.75)
c..
Does Beta Effective Increase, Decreases or Remain the Same, fron BOL to EOL7 EXPLAIN YOUR CHOICE.
(1.25)
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For two equivalent positive reactivity additions to a critical r
reactors will the SUR be the Same, Largers or' Suslier at EOL as compared to 80L7 NO EXPL ANATION IS NECESSARY.-
(0 5)
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.1 PRINCIPLES OF NUCLEAR POWER PLANT DPERATION.
PAGE
THERMODYNAMICSm HEAT TRANSFER AND FLUID FLOW
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CUESTION 1.07 (2.001 Explain HOW and WHY the Doppler Power Coefficient' is * affected by the followinst-
" c.. Buildup of fission gasses in the fuel - to cl ad gap.,
(1.0)
b..
Clad creep.
(1.0)
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-GUESTION 1.08 (2.00)
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How is the margin to DN8 affected by a DECREASE In each of the fallowing.
Consider each independently.
o.
Reactor power.
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RCS flow.
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Tcold.
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d... Pressurizer pressure (2.0)
QUESTION 1.09 (3.00)
c.-
How are each of the following parame.ters af fected, if at alls (INCREASES DECREASE or NO CHANGE) If one main steam isolation valve (MSIV) closes with the plant at 50% load.
Assume all controls are in automatic and that NO trip occurs.- Compara to final steady state values UNLESS " initial. change only" is indicated.
1.
Affected loop steam generator level (INITIAL change only).
2.
Affected loop steam generator ~ pressure.
3.
Affected loop cold les temperature.
4.
Unaffected loop steam generator level (INITIAL change only).
5.
Unaffected loop steam generator pressure..
6.
Unaffected loop cold leg temperature.
(2.4)
b.
If a reactor trip occurred as a result of the MSIV closing, which of the reactor protection system signals could be expected to cause the reactor to tript (If more than ones list the one that would reach the trip point first.)
(0.6)
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.la_. PRINCIPLES OF_ NUCLE AR POWER PL ANT OPER ATIONn PAGE
THERMODYNAHICSn HEAT. TRANSFER AND FLUID FLOW
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- 0UESTION 1.10 (2.75)
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c.
If steam goes through a throttling process,.specifically as in a leak from the main steam high pressure header to atmosphere, HOW will. the following parameters changet (2 0)
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1.
Enthalpy (h)
2.
Pressure
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Entropy (s)
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Specific volume (v)
5.
Temperature
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. State whether the steam will be subcooled, saturated or
superheated as it leaks out.
(0.75)
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0UESTION 1.11 (3.00)
01 A variable speed centrifugal pump i s operating at 1/4 rated speed in a CLOSED system with the following parameters:
Power = 300 KW Pump delta P = 50 psid Flow = 880 spa What are the new values for these parameters when the pump speed is increased to full rated speed?
(1.5)
b.
Choose the answer that mort correctly completes the sentence.
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"In soCLOSED system, two single stage centrifugal pumps operating i n parallel will have--(choose-f r on-be low)-, as
compared to the same system with one single -stage centrifugal pump operating with one pump isolated."
1.
a higher head and higher flow rate.
2.
the same head and the same flow rate.-
3.
the same head and a higher flow rate..
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a higher head and the same flow rate.
(0.5)
c.
How is the available NPSH affected by an increase in system flowratet (0.5)
de: Why is cavitation undesirablet (0.5)
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- ptANT DgilgN INCLUDlH&_3AFETY..ABa_EBEEEENCI_ H Hf53 PAGE
0UESTION 2 01 (3 50)
c.,
Other than EFW flow indication / recording state the function (s)/
purpose (s) for each of the following.
Please be specific.
1.
(1.5)
2.
EFW supply header flow orifice.
(1.0)
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b..
What is the purpose / design feature of the temporary hose connection on the suction of the turbine driven EFW pumpf (1.0)
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QUESTION 2 02 (1.50)
What function is performed by the Air Relay Dump Val ve ( ARDV) on a i
j turbine trip AND why is this action necessary?
(1.5)
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QUESTION 2 03 (2.50)
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c.
The containment structure air is sampled by hydrogen analyzers.
Following a LOCA ("P" signal)s what automatic action ensures that a representative sample of containment air is sampled?
(1.0)
b.
The containment air hydrogen concentration must be kept below
____ percent by volume to prevent the possibility of an explosion in containment. (Provide value)
(0.5)
c.
What system serves as a backup for the hydrogen recombinerst (1.0)
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.Z.z PL A N I D ES IGil_IAICLUQ1BS_3AEEIX_ AND_EHE R G E N CX_ S Y S T E MS PAGE
0UESTION 2.04 (4.00)
c.
State the setpoints and coincidence which will automatically initiate Containment Building Spray (CBS).
(1.0)
.b.
What. is the purpose of inJactins sodium hydroxide (NaOH) into the spray water?
(1.0)
c.
Descr ibe how NaOH enters the CBS System.
(1 0)
d.
What conditions must be met for containment recirculation sump Isolation valves (V-8, V-14) to automatically opent (0 5)
o.
True or False The CBS pumps will continue to operate during the switchover to the recirculation phase of operation..
(0 5)
l GUESTION 2.05 (3.00)
o.
Assume the plant is shutdown at 500 F and 2200 psis.
Will the Low Temperature Overpressure Protection System act to reduce plant pressure immediately if a loop Tcold instrument f alls LDWT
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BRIEFLY EXPLAIN your answers INCLUDING which train is affected, what action takes pl ace, and any applicable setpoints.
(2.0)
b.
What is the purpose of the PORV interlockt-(1.0)
l GUESTION 2.06 (2.50)
l o.-
List the THREE trip conditions which remain active when the i
EDG automatically starts as a result of a bus undervoltage con di t i oneoweew wm seinv.na4=m u.
(1.5)
b.
What TWO relays are reset by the EDG Control Roset Switch on the NCB.
(1.0)
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Z a__kkASI_ D E S I GM_ISCLilRIES_SAEEII_ AND_EHERraEtiCI_ SISIE53 PAGE
QUESTION 2 07 (2.50)
Tha following questions concern the Primary Component Coo, ling (PCCW)
Systems at the Seabrook Station.
c.
What TWO PCCW alarms could indicate a RCS to PCCW leakt (1.0)
b.
Describe how the RCP Thermal Barrier Loop (heat exchangers and other associ ated piping) i s protected from overpressure if either a leak OR rupture occurs into the PCCW Thermal Barrier Loop.
(Include any applicable setpoints.)
(1 5)
QUESTION 2.08 (3 50)
An3ter the following questions utilizing the attached drawings Figure 2-1.
c.
Indicate the "f ailed" position for the valves labeled A through G.
(2.1)
b.
If left in automatic control, what position should the Letdown Pressure Control Valve (PCV-131) be found in two minutes after a safety i nJection initiation?
(0.4)
c.
What are the TWO purposes of PCV-1317 (1 0)
QUESTION 2.09 (2 00)
a.
What THREE signals other than manual will cause Maln Steamline Isolation? (Include Setpoints)
(1 0)
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b.
What are TWO purposes of the Main Steam Flow Restrictor?
(1.0)
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GUESTION 3 01 (3 00)
Fcr each case below EXPLAIN the resulting method of reactor coolant cyates temperature control AND Indicate the approximate final RCS Tove.
Assume all systems normal except as stated, no operator C3tions AND consider each case separately.
o.
The normal steam pressure setpoint is reduced by 92 psi while in Hot Standby awaiting reactor startup.
(1.0)
b.
The train A steam dump selector switch is placed in 'of f' while at 55 reactor power awaiting turbine startup.
(1.0)
O.
Train 8 reactor trip breaker f alls to open upon a trip from 785 power.
NOTE Train A breaker opens.
(1.0)
f GUESTION 3.02 (1.50)
Th3 plant is operating at 505 power with all systems in automatic.
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H o does a HIGH failure of Power Range channel M-44 LOWER
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datector affect the following Indicationst
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Lower Quadrant Power Tilt Ratio (GPTR)
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b.
Delta Flux (Axial Flux) Indication (Channel 4)
f c.
S/G Feed Flow (Initial FRV response)
(1.5)
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QUESTION 3 03 (2.50)
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o.
What associated operator action would be required during plant shutdown if an Intermediate Range Nuclear Instrumentation (IRNI)
channel was grossly undercompensatedt (1.0)
b.
List THREE Control / Protection functions provided by the IRNIf i
(Include setpoints and logics.)
(1.5)
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' INSTRUMENTS AND CONTRDLS PAGE
QUESTION 3.04 (3.50)
c.
Provide TWO additional (different/ separate) AUTOMATIC signals other than High-Hi gh S/G-water level (P-14), which ulli c aus e the reactor protection and logic system to generate a feedwater isolation valve closure signal.
(Setpoints not required).
(1.0)
b.
Provide ALL direct and immediate indirect additional automatic actions associated with P-14, other than feedwater isolation valve closure.
(1.8)
6.
What protection is provided (reason / basis) by the P-14 signalf (0 7)
GUESTION 3.05 (3.00)
f c.
What input signal is used to provide the programmed pressurizer level for pressurizer level controlf (0 5)
b.
The" CONTROLLING pressurizer level channel falls HIGH during 1005 i
power operation.
Assuming NO operator action is taken, which
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reactor protection signal will cause the reactor to tript
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Include en explanation of the events, (causes and effects) which
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l-result in reactor trip.
(2 5)
l QUESTION 3.06 (3.00)
Tha reactor is at 801 power with rod control in automatic.
Assume a rapid lot load rejection occurs:
O.
Briefly EXPLAIN how a rod insertion signal is generated by the rod control systems
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1.
Power mismatch circult?
(1.0)
2.
temperature mismatch circult?
(1.0)
6.
What determines the rod insertion ratet (0.5)
se When will inward rod motion automatically ceasef (0.5)
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h _IH11&uSENIi_AND_CDHIRQ13 PAGE
QUESTION 3.07 (3.50)
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List the FOUR plant parameter input signals to the Overtemper ature Delta-T (OTdT) protection circuit..
(2.0)
b.
What core protection is provided by the OTdT protective ci r cu i tt (0.5)
co. State TWO additional functions (control / protection) other than reactor trip that the OTdT protection channel provides.
(1.0)
GUESTION 3.08 (2 50)
c.
Describe the opening interlocks associated with the following valves.
Include setpoints where applicable.
1.
Hot Leg Suction to RHR.
(1.0)
2.
Recirculation Sump Suction to RHR.
(0.5)
b.
State the TWO conditions that must be satisfied for Automatic Recirculation Sump valve opening.
(1.0)
QUESTION 3.09 (2.50)
YCar Reactor Protection System is designed so that a turbine trip will cause a Reactor Trip above P-9 (304 power).
o. Why is the system designed to do thist (1.0)
b. Provide a Reactor Protection signal that would act to give protection in the event that the Turbine Trip / Reactor Trip did not operate on a turbine trip from full power.
(0.5)
c. State the TWO ways that the Reactor Protection System senses that a turbine trip has occured?
(1 0)
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PROCEDURES - NORMALS ARNDRMALs EMERGENCY AND PAGE
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RADIDLOGICAL CONTROL
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j GUESTION 4.01 (3.00)
Tha foll.owing questions concern the precautions and Ilmitations of procedure OS 1001.05, RC P Operations.
I co Briefly explain whys when starting the first RCP while on RHRs both RHR suctions to the RCS must be open.
(0.6)
b. Whys when RCS pressure is <100 psis, must the RCP 81 seal
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leakoff valves be closedt (0.6)
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o. Ifs while operating at 50% powers PCCW is lost to the-RCP noteras what TWO options are you given as reactor operatorf (1 2)
de Why must RCP motor starting' limitations be adhered tof'
(0.6)
- GUESTION 4.02 (3 50)
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o.
Assume the plant is operating at full power and the Aslal Flux Difference (AFD) has been outside the target band for the last 5 minutes.
What are the TWO actions specified which you may choose between to meet the Technical
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Specification requirements?
Include time limitations.
(1.5)
b.
Assume that It is 0310 on 03/18/85 and the plant is presently at 452 power.
Considering the AFD penalty history belows at
'
what date and time may power be increased above 5057 EXPLAIN.
(Show all work.)
Assume no deviation outside the band af ter 0310 on 03/18/85.
TIME WENT GUT TIME BACK DATE OF BAND IN SAND POWER 03/17/85 0310 0318 852 03/17/85 1557 1637 655 03/18/85 0148 0310 452 (2 0)
'
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.is ERDC S11RES :.MDR5&La_&ANDEBALa EBEREENCX_AE PAGE
RADIDLDEICAL COMIROL QUES TION 4 03 (3.50)
Anscer the following according to 031210.5, Dropped Rod procedure.
c.
List FIVE symptoms of a dropped RCCA.
(2.0)
bb What specific operator action (s) is/are required if a reactor startup i s in progress when a RCCA drops?
(1.0)
ce The plant is at power when the RCCA dropse how is the operator directed to maintain programmed Tavet (0.5)
QUESTION 4.04 (3.50)
,
Anster the following according to E-3 Steam Generator Tube Rupture (SGTR) procedure 051330.
C.
The faulted S/G has been i dentified as
"B" S/G.
The "B" $/G ASDV controller setpoint is adjusted to 1125 psigs MSIV and bypass valves; blowdown isolation valves; main and emergency feed valves; and Main Steam drain valves are all closed.
Is the "8" S/G isolated? EXPLAIN.
(0.5)
b.
What operator actions are required to be perf ormed if the faulted S/G MSIV failed to shutt (1.0)
Go
' State the RCP trip criteria during a SGTR.
(1.0)
de State the ECCS reinitiation criteria during a SGTR.
(1 0)
GUESTION 4.05 (2.50)
Anster the following according to FR-51, Response to Nuclear PCacr Generation /ATWS procedure.
ao What THREE conditions must you observe to " verify" a Reactor Tript (1.5)
be The reactor has not tripped and attempts to manually trip the reactor at the MC8 have failed.
Assuming the turbine has tripped and EFW pumps are running, what are your required immediate actions?
(1.0)
,
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t a,__ERDCERURE1_=_5085Aka,_ AA5085&A,4,_E5EREESCI_ AMD PAGE
RADIDLOgig&L,ggglggt GUESTION 4.06 (2.50)
Anster the following according to 031200.02, Safe Shutdown and COstdown from the Remote Safe Shutdown Facilities.
c.
For a Unit 1 Control Room evacuation, what are the assignment responsibilities for the SCR0 and CR0f (Assume both RSS trains available)
(1.0)
ba How do you perform RCS depressurization during cooldown from the RSS panelf (0.5)
0.. What would be your indication of reactor vessel head steam voiding during RCS depressurization?
(0.5)
de
'True or Falsef Placing a selector switch at the RSS panel in LOCAL defeats most automatic controls and interlocks and divorces control and indicating lights at the main control board.
(0.5)
GUESTION 4.07 (1.50)
Anster the following in accordance with Standing Operating Order No.84-001, Use of UNTESTED procedures.
o.
What is your responsibility / action when it becomes evident that strict compilance with an UNTESTED procedure could result in equipment damaget (1 0)
be Who, by Job position / title, has the authority to issue a change to the UNTESTED procedure for the conditlen noted in as abovet (0.5)
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PEDcEnggES - AIDRBALa_AAMDEBAla EBEREEHCI_ARQ PAGE
AAR1DLDEIGAL.GDMIEDL QUESTION 4.08 (2 00)
c.
From the following, choose the provision (s) If any, required by procedure RP-5 0," Requesting and use of RWP's and SRWP's" for an entry into a high radiation area WITHOUT a RWP.
The entry is for an immediate/ critical action of short duration and has been authorized by the Shift Superintendent.
1.
Personnel entering are given verbal Instructions and precautions.
2.
Continuous health physics coverage.
3.
Personnel entering must wear protective clothing over their personal clothing.
4.
The Job is documented on a RWP after completion.
(1.0)
be You are required to periodically check your self reading Pocket dosimeter (SRPD) while in a RCA.
What SRPD " reading" would require you to leave the area immediately and notify HP personnel, according to RCA Access Requirements, procedura SS R P-8 01 (1.0)
GUESTION 4.09 (3 00)
D:ccment 10 CFR 20 provides regulations for radiation exposure at th] Seabrook Facility. Answer the following questions in accordance cith 10 CFR 20.
so What is your GUARTERLY Whole Body exposure limit?
(0.75)
b. What THREE criteria must be satisfled in order to exceed this limit under NON-EMERGENCY conditions?
(2 25)
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EQUATION SHEET
.
f a ma v = s/t Cycle efficiency = (Net work out)/(Energy in)
w = mg s = V,t + 1/2 at
g=m
,
KE = 1/2 av a=(Vf - V,)/t A = AN A = A e'**
g PE = agn
'
Vf = V, + at w = e/t i = an2/t1/2 = 0.693/t1/2
- 1/2W = buiMtW
-
g,,3p
,g A=
[(g/2)*(*b}3
,
aE = 931 m
- Y A8
""
'
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'
Q = aCpat Q = UA A T I = I,e"#
I = I,10"* M L
,
Pwr = W ah
-
f TVL = 1.3/u P = P 10 ""III HVL = -0.693/u
~
P = P,e*/
SUR = 26.06/T SCR = 5/(1 - Kgf)
CR,=5/(1-K,ff,)
gfj) = CR (I - "e #2)
SUR=26e/s*+(s-e)T CR (1 - K
j T=(s*/s)+[(s-eVIs3 M*I/II~Kdf) = CA A,
j
,
T = 4/(, - s)
M = (1 - Keffo)/(1-kdf1)
T=(s-e)/(Is)
SOM"(h*secondj Kdf)/Edf s' = 10 o = (Kgf-1)/Kg f = 4Kgf/Kgf A = 0.1 seconds gf(1+IT)]
e=((t*/(TKgf)]+[T
/
-
l1*Id Id 2,2 2 P = (s4V)/(3 x 1010)
Id gd j
2 l
t = eN R/hr=(0.5CE)/d(,,g,,,)
R/hr = 6 CE/d2 (f,,g)
'
.
'
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem.
1 curia = 3.7 x 1010dps
-
1 ga;. = 3.78 liters I kg = 2.21 lba 3 8tu/hr 1 ft* = 7.48 gal.
I hp = 2.54 x 10 Density = 62.4 lbg/ft3 1 or = 3.41 x 106 Stu/hr Density = 1 ga/cW lin = 2.54 cm Heat of vaporization = 970 Stu/les
- F = 9/5'C + 32 Heat of fusion = 144 Stu/les
- C = 5/9 (*F-32)
1 Ata = 14.7 psi = 29.9 in. Hg.
1 STU = 778 ft-Ibf 1 ft. H 0 = 0.4335 lbf/in.
-
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.
. la__ERINCIELE1_DE_HuCLEAE_EDER_EL AtiI_DEEEAIIDMa.
PAGE
ItiE850AXHA51C14._ HEAT TRAN$ggg,AND_E(ylg_E(Qg.
ANSWERS -- SEABROOK 1-85/03/18-ISAKSENs P.
MASTERCOPY ANSWER 1.01 (3*.00)
co. Steam flow increases, Tave decreases [0.13 Pos. reactivity (via MTC) [0 13 Reactor power increases (0.153 Tfuel increases (0 13 Neg. reactivity (vi a FTC) [0.13 Reactor power increase stops [0153 FINAL COMPARISDN 1..
Taves lower than initial (0 253 2.
Tfuelt hi gher than initi al C0.253 3.
Reactor powers higher than initial (0.253 (1.45)
b.
Rod drops, Neg. reactivity C0 13 Reactor power decreases CO.153 l
Tfuel decreases (013 Pos. reactivity ( v i a FTC ). IO.13 Tave decreases Co.13 Pos. reactivity (via NTC) C0.13 i
Reactor power increases (0 153 l
FINAL COMP ARISDN 1.
Tavet lower than initial (0.25]
2.
Tfuels lower in fuel near dropped rod and higher in fuel remote f rom dropped rod (overall Tfuel is lower) [0.253 3.
Reactor powers the same as initial Co.253 (1 55)
REFERENCE l
Scabrook Causes and Effects Manual 4-7, 35, 36.
'
Comprehensive Nuclear Training Operations--Reacter Theory and Tharmel Science, Ch 12 pp 31 - 5.
t ANSWER 1.02 (1 50)
Reactivities associated with -- 1. Control rod position 2. Samarium concentration
'
3. Xenon concentration 4. RCS Boron concentration l
S. Plant temperature (Tave)
6. Fuel burnup (core age)
(five required, 0 3 each)
(1 5)
REFERENCE
,
MASTERCOPY
~
- " "
-
--
- - - -
-
-._
_
._
_
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.
.la_.PRIMCIPLES OF MuCLEAE_fGMER_fLMI_QF.ERAI1QKa.
PAGE
1H1150DIBABICla._tfEAI_IBAHIEEB ABD_ELilIR_ELQM ANSWERS -- SEABROOK 1-85/03/18-ISAKSEN, P.
ANSWER 1.03 (2.00)
ce Reduction in boron concentration decreases the positive effects of boron removal dur ing temperature increases [0.53. The buildup of Pu-240 increases resonance absorbtion [0 53.
(1 0)
be.
FTC becomes less negative as fuel temperature lacreases because resonant band uldening is reduced at higher tempe r a tur es.
(1 0)
REFERENCE HD-RTR-76 to-78,-85 ANSWER 1.04 (2.50)
oo SANE
.
be.
[0.5 each)
(2.5)
REFERENCE RN-1735 ANSWER 1.05 (1.50)
c.
1.
Higher fission yleid.
2.
Larger (thermal) absorbtlon cross section (1 0)
ho Renelns the same (0.5)
'
REFERENCE HD-RTR-111,-114
.
..
...
2815CIPLES DE.BUCLEAE_EDMER_ELAtlI_DEERATIDMn PAGE
IHEREDDYMAMICSm_ HEAT TRANSFER AMD FLUID FLOW
' ANSWERS -- SEABROOK 1-85/03/18-ISAKSEN, P.
ANSWER 1.06 (1.75)
c.
Decreases [0.53 Pu 239 concentration increases (while U 235 concentration decreases) [0.753.
(1 25)
bo Largey SUR.
(0.5)
REFERENCE HO-RTR-23 to-27 ANSWER 1.07 (2.00)
ao Fission gasses pollute the Helium gas causing a reduction in sep thermal conductivity [0.53.
This results in increased fuel temp-erature change for a given power change, causing an increase in the magnitude of the coef fici ent [0.53.
(1.0)
be Clad creep effectively shrinks the clad into closer contact with the fuels increasing the gap thermal conductivity to.53.
This results in a fuel temperature decrease and a lower value for the coefficient [0 53.
(1.0)
REFERENCE
'
H D-R TR-87,-8 8
.
ANSWER 1.08 (2 00)
ao INCREASE be DECREASE se INCREASE do DECREASE
!
REFERENCE l
T HF-3
!
.
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paruciptEs nF MucLEAR POWER _RLAMI DEERAIIO4a~
PAGE
THERMODYNAMIC 1m HEAT TRANSEER AND FLUID.ELQM ANSWERS -- SEA 8 ROOK 1-85/03/18-ISAKSENs P.
,
~
ANSWER 1.09 (3.00)
ce 1.
DECREASE 2:.
INCREASE 3.
INCREASE 4.
INCREASE 5.
DECREASE 6.
DECREASE
[0.4 each3 (2 4)
b.-
Lo-Lo S/G level (0 6)
REFERENCE 300 brook Simulator Halfunction no. 40 ANSWER 1 10 (2.75)
c.
1.
SANE 2.
DECREASE 3.
INCREASE 4.
INCREASE 5.
DECREASE C0.4 each3 (2.0)
b. Superheated (0.75)
REFERENCE THF-1 ANSWER 1.11 (3.00)
,
3 0.
Power (2)
Power (1) * (NZ/N1)
= 300 * (4)
= 19.2 MW (0.5)
=
2 Dette P(2) = delta P(1) * (N2/N1)
= 50 * (4)
= 800 psid (0.5)
Flow (2) = Flow (1) * (N2/N1)
880 * 4 = 3520 spa (0 5)
=
-
be Answers fi (0 5)
O.
DECREASES (0.5)
de Pump efficiency and flowrate are reduced and mechanical pump damage (erosions pitting and vibration).,
(0.5)
-. -. -. -. -
- -
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-. - -. -.... -....
. '..
,
.
P R I NC I PLELQg_gELE Ag_gDMER_gL AgI_gggg&Ilgg3, -
PAGE
.
THERMODYNAMICSa HEAT TRANSFER AND FLUTD__ FLOW ANSWERS - SEABROOK 1-85/03/18-ISAKSEN, P.
REFERENCE-THF-4
,
4 -
.
O a
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e-
.
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.
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.
.
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PLAgI,gg31gg_1ggkj,lglgg_1&FETY AMD_EBEREENCL SYSTEMS PAGE
-
-ANSWERS -- SEABROOK 1-85/03/18-ISAKSENs P.
ANSWER 2.01 (3.50)
c.
1.
Provides backup- (passive) protection to the high flow isolation functions limiting flow in the f aulted headerE0 53 (750 spa) protecting the EFW pumps from runout [0 51 and prevents robbing all the EFW f low f rom other S/G's[0.53 (1.5)
2.
Provides input f or the hi gh flow isolation signa l of the
-
faulted header (closure of the FCV at 450 ppm)
(1.0)
b.
One of the required (safety related) sources for makeup water to the Spent Fuel Pool.
(1 0)
REFERENCE
,
HD-EFW 18s 23 I
l l ANSWER 2.02 (1 50)
'
Tho ARDV closess interrupting the instrument air to the extraction l
staan line non-return valves allowing them to close t0.753s which l
provents a reversal of steam flow that could cause a turbine
'
evorspeed condition C0.753.
(1.5)
REFERENCE HD-TGN-116, 117, 123 ANSWER 2 03 (2 50)
o.
The containment recirculating filter system fans start.
(1 0)
l b.
(0.5)
o.
Containment Structure Purge System.
(1.0)
REFERENCE-NO-CGCS p. 17s20-22s 373 0S1023.40
,
L i
,
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--
.
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2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
.
ANSWERS -- SEA 8R00K 1-85/03/18-ISAKSEN, P.
ANSWER 2 04-(4.00)
c.
18 psig (HI-3) 2/4.
(1.0)
6.
To aid in lodine removal and retention.
(1.0)
6.
From the S AT which gravity drains to the RWST mixing chamber (through two parallel lines) to the suction of the CBS pumps.
(1.0)
do RWS T. LOW-L OW level in conjunction wi th SIS.;
(0 5)
o.
True (0 5)
REFERENCE HO-C BS-18, 20s 23, 54
'
.
ANSWER 2.05 (3.00)
a.
- NO PRESSURE REDUCTION; The Train A PDRV only opens en auctioneered That that it provides.- [0.53
-- Auctioneered Tcold inputs to Train 8. [0.253.
l
- At about 350 F [0.253 the Train A PORY (456A) is armed and Its respective Isolatlen Valve (V-122) is opened. [1.03 (2.0)
l i
b.
It prevents depressurization of the RCS past the interlock
setpoint (2185 psis) If a pressure channel selected to a l
PORV falls high.
(1.0)
REFERENCE HD-PPLC-31, 33-35.
, ANSWER 2 06 (2.50)
l l
a.
1.
2.
Low lube oil pressure 3..
Generator differential lockout.
(1.5)
be 1.
Emergency start 2.
Engine trouble shutdown relays.
(1.0)
sm r=was %
i REFERENCE HD-EDm-61-64 j
l
.
-
.
- -.
.
..
..
.
. 2..
"PL ANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
ANSWERS -- SEA 8 ROOK 1-85/03/18-ISAKSEN, P.
ANSWER 2.07 (2.50)
Co-High head tank level-High head tank overflow rate-PCCW liquid radiation monitor
[0.5 each, too required)
(1.0)
b.
(leak) Overpressure protection via a 150s relief valve [0.53 (rupture High flow of 150 spa shuts the thersel barrier outlet isolation valve. A check valve will close on reverse flow to isolate'the thermal barrier. A 2500s relief valve is installed in this (high pressure) secti on of pipe.[1 0)
(1.5)
REFERENCE HD-PCC-18 to -24
ANSWER 2 08 (3.50)
c.
A (LCV 460s459) CLOSED 8 (TCV 3818) OPEN
.
C ( HC V 128 ) CLOSED 0 (PCU 131) OPEN i
l E (FCV-121) OPEN
!
F (HCU-182) OPEN G (V59) OPEN
[0 3 each)
(2.1)
l b.
Full closed (0.4)
ce
- Prevent flashing downstream of LD FCVs (Drag valves) (Normally)
- Control Plant pressure while solid (Shutdown)
(1.0)
REFERENCE HD-CVCS-32 to 34 and 171 ON 1242.01-Attachment A ANSWER 2.09 (2.00)
oo-Containment Pressure High-2 4.38-Steamline Low Press. (>P-11)
5858-Steamline High Press. Rate (<P-11)
100s/50 sec C0.33 each3 (1.0)
be-Limit steam flow on a steamline break accident-Provide d/p for steam flow measurement (0.5 each)
(1.0)
REFERENCE HD-MS-17s-18 and STCs8 52 (SetPoint Summary)
i
_ _,. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _,. _ _ _ _ _ _ _ _
. _ _ _
_. -. -. -
_
_. _.. _. _. _ _
_ _ _ _
_
_
..
..
.
.
.
-
3 d IMSIRUMEjQi.AHR.CDMIRQL1 PAGE
ANS WERS4-- SEASRDOK 1-85/03/1g-ISAKSEN,,P.
,
l
'
.
,
ANSWER 3.01 ( 3'. 0 0 )
c.
The normal ste a pressur3 setpoint of 1092 psig maintains Teve
.,at "557:F, a decr ease in the setpoint to 1000 psig would cause
'
'
the dumps to orien and cool Teve to ~550 F where the P-12 Interlock would close all steam dumps (Notes 0.25 oredit elven for correct converslen to Tsat "547 F)
(1.0)
^
be Secondary pressure would rise to the setpoint of the secondary i
. atmospheric relief valves to.53 unich would maintain pressure at 11252psis (0.253 and primary temperature 560 +/-1 F [0.253.
(1.0)
O.
A signal by the Load Rejectlen controller [0.53 would control primary temperature et,"Ho Load" Tref +2 F deviation (deed bands (559 F) (0.53.
(1.0)
_
.
REFERENCE
,
i Scabrook H0 SD p. 17, 27, 34, 41..
MS Dump Control Logic Print C-509050
.
.
<
,
- ANSWER 3.02 (1 50)
s.
INCREASE
,
i be NORE NEGATIVE se INCRE ASE (FRV OPENS)
[0.5 each3 (1.5)
,
REFERENCE Scabrook Instrument Failure Reference Manual, Sectlen K, p.
5,6,s,9
.
'
ANSWER 3 03 (2.50)
as. Operater will he required to uanually reinstate power to the Seures Range instrue.ents.
(1 0)
.
be
- I npu t to P-6 [ 0. 3 3 10-10 amps L O.13 1/2 [0.13
- High flux rod stop [0.13 leg 204 to.13 1/2 [0.13
- High flux Rx trip [0 33 Iag 254 C0 13 1/2 C0 13 (1.5)
,
REFERENCE 30abreak Instrument Failure Reference Manual, Section J, p.
2-4 i
NO-MIS-74
!
m.
.
.
V.3
.
.
"Ra-iMSTRugggli_Agg_gggIggk2 PAGE
ANSWERS -- SEABROOK 1-85/03/18-ISAKSEN, p.
ANSWER 3.04 (3.50)
Go 1.
SI 2.
Rx trip ooincident with Low Teve (1.0)
be 1.
Feedwater pumps trip 0,
- .e.s e ;7 PJ 0 ;;t.;;;e ;;!
- :
v..r
3.
Block start of start-up f eed pump
4.
Turbine trip 5.
Reactor Trip (>P9)
6.
FRV and bypass valves shut
[0.Sseach]
(1.8)
ce protect the Turbine from S/G moisture carryover.
(0.7)
.
REFERENCE Scabrook Logic Print C-509053, HO-RPS p. 79, HO-IS-52 ANSWER 3 05 (3.00)
so Auctioneered high Tave.
(0.5)
bo Charging flow decreases [0 53, pressurizer level decreases (0.51, letdown isol ates [0.53, and pressurizer level increases [0.53.
High Pressurizer level trip (924) [0.53.
(2.5)
REFERENCE Scobrook Instrument Failure Reference Manual, Section Ds p.
1-4
- ANSWER 3.06 (3.00)
c.
1.
Turbine impulse pressure dooreases [0.53 in respect to N-44, causing a rate of change slonal which drives rods in. [0.53 (1 0)
2.
Turbine impulse pressure decreases (Tref) (0.53 below Teve causing a temperature mismatch error whleh drives rod in. [0.53 (1.0)
.
be The magnitude of the power mismatch rate signal PLUS the temperature mismatch signal.
(0 5)
oo When Tevs and Tref are within 1 F.
(0.5)
REFERENCE Scabrook HD-FLRC p. 20-23
<:,
...
,
,:
la_IIMSIRUMENIS AND_ CONTROLS PAGE
ANSWERS ---SEABROOK 1-85/03/18-ISAKSENs P.
ANSWER 3.07 (3.50)
c.
Taves dts Pressures dI CO.5-each3 (2 0)
b.
Prevent exceeding DNB (0.5)
o.
Turbine runback Blocks automatic and manual rod withdrawal 0k$(0.5 each3 (1 0)
REFERENCE
,
Scobrook HD-RPS, p.54-59
,
ANSWER 3.08 (2 50)
j RCS pressure < 365 psig. (auto close at 660 psig).
c.
1.-
-
-- RHRS to CVCS/ SIS isolation valves (RH-V 35 and 36) closed.
(1.0)
2.
RCS suction valves to RHR closed.
(0 5)
b.
1.
RWST Low-Low level.
2.
"S" signal present.
(1.0)
REFERENCE Scobrook HD-RHRS p.
44-47-
!.
'
-AMSWER 3.09 (2.50)
c. Because the turbine serves as the heat alnk to the reactors a reactor trip follows a turbine trip to ainimize the RCS
,
temperature transiedt (and/or resultins safety valve operation).
(1.0)
b..- Hi gh pzr pr essure
- High pzr level
- 0T Delta T
[1 required]
(0.5)
. als w so uo m.
o. - All turbine stop valves shut
- Emergency oil (Emer. Trip Fluid System) pressure low (< 800 psig)
[0.5 each3 (1.0)
REFERENCE
'
HD-RPS-65 and 66 D g. 9783-C-589956
.
.
.
.
...
...
- .
e..
.
. ia__ PR OC E B U RE1_::_BDEB ALa_ ASHORBALa_EBEEEENCI_ ARQ PAGE
RADIDLDGICAL COBJ,RQL ANSWERS -- SEABROOK 1-85/03/18-ISAKSEN, P.
ANSWER 4.01 (3.00)
c.. Ensures both RHR reliefs are available to combat a pressure spike.(0.6)
b. Prevent backflow through the seals.
(0.6)
c. - Immed i ate ly res tore CCW.
- Trip the RCPs
[0.6 each]
(1.2)
d Prevent winding damage (from excessive starting current heat).
(0.6)
REFERENCE 05 1001.053-pages 2 to 5 ANSWER-4.02 (3.50)
c.
Within 15 (or next 10) minutes [0.53 either 1. Restore the indicated AFD to within the target band E0.53s or 2. Reduce the thermal power to <90% of rated thermal power.[0.53 (1.5)
-b.
Accumulated penalty over the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is 89 minutes.[1.01 The penalty will be reduced to 60 minutes at 1618 minutes on-03/18/85 and then power may be increased.tl.03
.
(2.0)
85%
0318-0310
=
[0.253 65%
1637-1557
=
CO.253 45%
0310-0148 82/2
[0.53
=
=
__
min totsi pen al ty 03/17/85s from 1557J 81 min left -60 = 21 min -> 1618 03/18/85
[1.03 REFERENCE TS3 3.2.1
.
~
.
.o
'
.4.
PROCEDURES - NDRMALs ABNDR.MALs. EMERGENCY AND PAGE
RADIQLOGICAL_GQHIRQL i
ANSWERS -- SEABROOK 1-85/03/18-ISAKSENs P.
ANSWER 4.03 (3.50)-
c.
- Individual RCCA bottom light
- RCCA position deviation alarm
- PR NI Delta I alarm
- Rap i d dr op in T avg /musast we
- Detector or channel current comparator al arm
- Tave-Tref deviation alarm
- Control rods stepping out in AUTO C5 requireds 0.~4 each3 (2.0)
b.. Insert rods to shutdown the reactor [0.83'and investigate the.
problem.IO.23 (1 0)
c.
Adjustment of turbine load.
(0 5)
REFERENCE-Sochrook Dropped Rod Procedure 0S1210.05 ANSWER 4 04 (3.50)
!
l c.
NO -Steam supply to TDEFW pump (NS-V128)not closed.
(0.5)
b.
- Verify (turbine trip) Main Turbine stop valves closed (and main steam drains closed).
- Close remaining NSIV's and bypass valves.'
- (Use intact SG ASDV's for steem dump)
(1.0)
c.
- At least one high head pump running (CCP or SI)
- RCS subcooling-less than 30-F.
(1 0)
~d.
- PRZR level cannot be maintained >5% (30% adverse containment)
- RCS subcooling less than 30-F.
(1.0)
REFERENCE Scobrook E-3 SGTR procedure OS1330 p.
2-4 and Action Summary.
ANSWER 4.05 (2.50)
O.
Rod bottom lights 2.
Reactor trlP and bypass breakers open 3.
Neutron flux decreasing.
[0 5 eachT (1 5)
l l
b.-
1..
Manually insert control rods 2.
Emergency Borate.
[0.5 each)
(1.0)
l g
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- - - - - -
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'\\
44__IEROCEDURES. - NORMALa. ABNDRMAL. EMERGEHC1_Agg PAGE
RADIDLOGICAL CDMTROL ANSWERS - -SEA 8 ROOK 1-85/03/18-ISAKSEN, P.
REFERENCE
- Scobrook FR-5.1 p.
2,3
,
i ANSWER 4.06 (2.50)
0..
SCR0 - RSS panel (B) operator CR0 - RSS panel (A) operator (1.0)
b..
By opening.a PRZR PORV.
(0 5)
co-PRZR level increases or oscillations.
(0 5)
d.
True (0.5)
,
REFERENCE.
Scobrook 0S1200.02, p.
4s 5 and 19 ANSWER 4.07 (1.50)
l o.
Notify the SS or USS.
(1.0)
b.
SS or USS.
(0.5)
i f
REFERENCE l
Soobrook Standing Operating Order No.84-001
!
ANSWER 4.08 (2 00)
c.
Is 2 AND 4 (1.0)
b._
Unexpected OR excessive OR if it exceeds 3/4 scale (any one for full credit).
(1 0)
REFERENCE Scobrook Administrative Procedures RP-5.0 p. 5 and SSRP-8.0 p. 4 i
L
- T_ O a o ta__hgGCEDURES - _HDRBALg_ARBDR5&La_E5EggENCY ARQ PAGE
RADIDLDEICAL_CDSIRDL ANSWERS -- SEABROOK 1-85/03/18-ISAKSENs P.
ANSWER 4.09 (3.00)
c. 1 25 Rem /4uarter (0.75)
b. -3 Rem / Quarter is NOT exceeded. [0.753
,
i Total accumulated dose does not exceed 5(N-18). [0.753
-
l-Accumu ated exposure on recor d (NRC-4). [0.753 (2.25)
,
REFERENCE 10 CFR 20
,
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p a t.t b
U.
S.
NUCLEAR REGULA10RY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
SEABROOK 1
_________________________
REACTOR TYPE:
PWR-WEC4
_________________________
DATE ADMINISTERED: 85/03/18
_________________________
EXAMINER:
RUSCITT0rD.
APPLICANT:
_
_
__
____
____ ___
________ ____
'
Una ceparate paper for the answers.
Write answers on one side only.
Stcple question sheet on top of the answer sheets.
Points for each quantion are indicated in parentheses after the question. The passing groda requires at least 70% in each cate3ory and a final 3rade of at looct 80%.
Examination papers will be picked up six (6)
hours after tha examination starts.
.
g
% OF iATEGORY
% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY p______
______
___________
________ ___________________________________
.
25.00 25.00 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS s5.00 25.00 6.
PLANT SYSTEMS DESIGNr CONTROL, l
AND INSTRUMENTATION 5.00 25.00 7.
PROCEDURES - NORMAL, ABNORMALr
'
EMERGENCY AND RADIOLOGICAL CONTROL 5_I_00 25 0
___1__0
________ 8.
ADMINISTRATIVE PROCEDURESr
___
___________
CONDITIONS, AND LIMITATIONS 00.00 100.00 TOTALS FINAL GRADE _________________%
11-work done on this examination is my own. I have neither
'vcn nor received aid.
i IPPL5ChUT 5 555 ITURE I
~~~~~~~~~~~~~~
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDSr AND PAGE
____
______________________________________
______________
GUESTION 5.01 (2.00)
,
It is the responsibility of the reactor operator to ensure that the core pcwor distribution limits are maintained at all times.
Operation within
,
th9ce limits is reasonably assured when four conditions are met.
What are thace four conditions _that will prevent exceedin3 core power distribution lioits/ hot channel factors?
,
QUESTION 5.02
-
(1.50)
Explain why Reactor Coolant' System Delta T can be used as a measure of racetor power, bv.t secondary coolant Delta T cannot.
QUESTION 5.03 (.50)
Accu:e RCP's are tripped following a LOCA.
After the break has been icoloted which of the following situations would be MOST desirable?
l P PZR T PZR T HOT T COLD o.
1240 570 580 570 b.
1100 557 540 530
-
c.
750 520 520 520 d.
640 540 520 500 QUESTION 5.04 (1.50)
For oach of the following conditions, describe how each term of the heat transfer equation G = UA(DELTA T) chan3es.and why.
Assume power remains conotant.
l o.
Scale formation increases on steam senerator tubes. [0.753 b.
Additional steam Senerator tubes are pluS3ed. [0.753 fQUESTION 5.05 (2.50)
L Anowar the followins concerning the 1/M_plott o._What are two purposes of the 1/M plot during startup? C1.03 b. Why is the inverse count rate plotted rather than count rate? CO.753
.c.
Why is the time interval between reactivity changes important with-re3ard to constructing a 1/M plot? Co.753
.
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5.
THEORY OF' NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
___.
_--------------------------------
_____-____---_
QUESTION 5.06 (3.00)
Aftor operation at 100% power for several weeks near the end of cycler it
'
in decided to reduce power to 75% using rods only.
o.
After reaching 75% powere what rod motion would be required to
!
maintain the plant at 75% power over the next 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> assumin3 no change in baron concentration?
b.
Explain what causes these reactivity chan3es.
C1.5 each]
,
QUESTION 5.07 (3.00)
List the four parameters affecting Departure from Nucleate Boiling (DNB) and state whether the probability of approachin3 DNB is increased or decreased
,
as these parameter values increase.
QUESTION 5.08 (1.50)
Why is there a Technical Specification limit on Shutdown Margin?
.
QUESTION 5.09 (3.00)
l Answor the following concerning emergency boration f-j o.
Explain the response of reactor power and TAVE after 2 minutes of emergency boration at 100% power.
Assume rod control is in manual.
i b.
Explain the response of reactor power and TAVE after 2 minutes of emergency boration at 10 - 8 AMPS and no load TAVE.
[1.5 each3 L
.
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l 5. -- THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
____7;ggggg7;ggggg______________________________________
______________
~ QUESTION 5.10-(2.50)
- .
Ancwor~the following concerning fission product poisons:
o.
-State the production and removal mechanisms for Xe-135 and Sm-149 in i
the reactor core.[1 03 b.
State the reactivity value for peak xenon concentration following a trip from 100%. power and the time required to reach this peak condition.CO.753 c.
State the reactivity value for 100% equilibrium xenon concentration and the time to reach this condition from startup with a clean reactor core.CO.753
'
QUESTION 5.11 (1.50)
-
i A variable speed centrifugal pump is operating at 1/4 rated speed in a clocod system with the following parameters.
What are the new values of those parameters when speed is increased to rated speed?
o.
Power = 500 KW
.
b'.
Flow
= 900 GPM c.
Head
= 70 PSIG GUESTION 5.12 (2.50)
a.
Why does reactor coolant pump amperage increase during a plant cooldown? CO.53 l
b.
During primary system cooldownr why do steam dump valves have to be l
opened.further as temperature decreases to maintain the same cooldown rate? [1.03 r
c.
Why is vessel Delta T always less than core Delta T? C1.03 i
.
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-6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
QUESTION 6.01 (2.00)
'
Annwer the following concerning a high failure of RHR Train B' pressure transmitterLPT-405.
Assume the plant is at 300 DEG F coolins down usins both RHR trains.
o.
Explain the effects on the automatic actions / interlocks associated with the RHR isolation valves. C1. 03, b.
What other indications could be used to verify this failure?'C1.03 QUESTION 6.02 (3 00)
Answer the following concerning the Steam Dump Systent
-
c.
Why does the C-7 load rejection armins si nal lock in? C1.03
b.
Why must the Steam Dump Mode Selector Switch be reset upon completion of.a large load rejection? C1.03 c..
How do you know if the C-7 arming signal is actuated? C1.03
.
-QUESTION 6.03 (4.00)
Answer the following concerning the Pressurizer Pressure Control System.
Acoune that the plant is at 100% powers
o.
List the indications in the Main Control Room if the controlling
channel (455) of pressurizer pressure fails high.
Do not include alarms. C2.03
!
b..
What COULD cause a reactor trip on the above failure? C0.753 c.
What actions should be taken to restore the plant to a stable condition once the failed channel is identified? C1.253
.
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~ PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
.
QUESTION 6.04 (4 00)
Anower the following concerning a main steam line break o.
Explain how each of the following affect the severity of a main steam
'line break accident (1)
Mode (HSB, S/U, etc.) ti.03 (2)
Time in cycle (BOL, EOL, MOL, etc) C1.03 b.
What specific plant desi3n features were incorporated to limit the severity of a main steam line break? C0.53 c.
Following a main steam line break inside containment, which si nals
will cause safety injection actuation?
Include primary and backup signals, setpoints, and coincidence. C1.53 i QUESTION 6.05 (3.00)
Tha. plant is operating at_60% power when a T hot narrow ranse RTD fails hish.
Explain how this failure will affect the following.
Consider each itco independently.
Assume no operator action and that all control systems oro in automatic.
c.
Rod insertion limit.setpoint.
,
b.
Char 3 ng flow (initially).
i c.
Control rod bank position.
d.
Steam dump control system.
C0.75 each]
r
.
' QUESTION 6.06 (3.00)
'
Whore'are the source range, intermediate range and power range detectors i
lecoted vertically with respect to the core?
Why?
QUESTION 6.07 (1.00)
-
Hew could Vital Instrument Panel PP-1A be supplied if UPS 1A was not cvoilable?
r
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
< ______________________________________________________
'GUESTION 6.08 (3.00)
Ancwor the following concerning the Emergency Feedwater(EFW) System a.
What-automatic action occurs at 450 GPM in any EFW system feed line.
Why? 'Under what plant condition is this in effect? C1.03 b.
How are the'EFW pumps protected against runout? C0.753 How would an operator know if local control of either EFW pump has c.
been taken? E0.53
"
d.
Which, if_any of the automatic initiation signals will start the EFW pumps under the conditions in part c above? E0.753
- QUESTION 6.09 (2.00)
Ancwor the following concerning the Combustible Gas Control System.
o.
Following a LOCA inside containment, what are three major sources of hydrogen? C1.03 b.
At what level does hydrogen in the containment become a problem?
Explain. C0.53 c.
How can hydrogen be removed from the' containment? CO.53 i
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~
RA65 E6G 6AE"66 TR6[~~~~~~~~~~~~~~~~~~~~~~~~
____________________
QUESTION 7.01 (3.00)
During 80% power operation, with rod control in automatier a Power Range NI channel fails low.
c.
'What immediate actions, if any must be taken?
Explain. C1.53 b.
What other actions must be taken in accordance with the Technical Specifications? C1.53 QUESTION 7.02 (1.50)
Follcwing a loss of Primary Component Cooling Water (PCCW) while at power, PCCW was isolated from the containment, the reactor was tripped and Reactor Ccolent Pumps (RCP) were tripped.
RCP A ti seal leakoff susequently was rcported to be 6.5 GPM.
What action, if any must be taken?
QUESTION 7.03'
(1.00)
Why cust the effluent of the letdown desassifier be directed to the CRIE for ten minutes prior to discharS ng the effluent to the VCT?
i
.
QUESTION 7.04 (3.00)
Following a fire in the Control Roon ventilation system while operating at powar, the followin3 actions are taken:
The reactor and turbine are ocnually trippede RCPs are stopped, all MSIVs and bypasses are shut.
You havo just manned Remote Safe Shutdown Panel A.
c.
How will the following parameters be controlled while cooling down to 450 DEG F and 1000 PSIG?
(1)
Steam pressure
.
(2)
Steam Senerator level
(3)
Pressurizer pressure (4)
Pressurizer level E0.6 each3 b.
How will the plant be depressurized to 1000 PSIG? E0.63 o
e J
i i
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~
RA6Y6[6G5CA[~66 TR6[^^~~~~~~~~~~~~~~~~~~~~~~
____________________
!
QUESTION 7.05 (3.50)
,
l Ancwer-the following concerning E-Or Reactor Trip or Safety Injection o.
.What are.the five conditions that require SI initiation? (Include
'
setpoints) C2.53
-
b.
The Main Turbine has not tripped and you attempt a manual trip as i
directed, with no response.
What additional action are you required to take in order to shutdown the turbine? C1.03 iQUESTION 7.06 (2.00)
'
Fill in the following precautions and limitations from various operating
}.procodores l
o.
The RHR System shall be isolated from the Reactor Coolant System (RCS)
'
before RCS exceeds ________ DEG F or ________ PSIG. [0.53 I
b.
The Reactor Coolant Pumps (RCP) shall not be operated when the number one seal differential pressure is less than ________ PSID or the Volume Control Tank pressure is less than ________ PSIG. [0.53 c.
RCP seal injection water temperature should not exceed _______ DEG F.
C0.25]
d.
At least one RCP should oe in operation with Tave greater than
________ DEG F.
Co.25]
o.
The boron concentration in the pressurizer should be maintained within ________ PPM of the RCS loop concentration. C0.253 f.
The pressurizer sprays sh'all not be used if the differential temperature between the in-service spray Teold and the pressurizer steam space exceeds ________ DEG F.
[0.25]
QUESTION. 7.07 (3.00)
Answer the following concerning intial core loading o.
What two specific conditions would require an immediate increase in boron concentration? C1.03 b.
What are four conditions that would require the suspension of core
' loading? (In addition to the above two) C2.03
.
.
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7.
PROCEDURES - NORMALr ABNORMAL, EMERGENCY AND PAGE
~~~~
RA6 6E6556AE 56RTR5E------------------------
____________________
GUESTION 7.08 (3.50)
Refueling Technical Specifications require that one RHR loop o.
be operating during refuelin3 What are two reasons for this requirement?.C1.53 b.
What is the reason for the requirement that two RHR loops be operable when water level is less than 23 feet above the vessel flange? C1.03 Under what conditions is it permissible to stop RHR flow during c.
refuelin3? [1.03 (
GUESTION 7.09 (3.00)
o.
In addition to alarms, last four indications of a dropped control.
rod? [1.03 b.
N' hat are the immediate operator actions required for a single dropped
-
rod? C1.03 c.
If the quadrant power tilt ratio is calculated to be 1 10, as a result of a dropped rode what is the time limit specified in the Technical Specifications for reducing power and how far must power be reduced? C1.03 iGUESTION 7.10 (1 50)
'
Anowor the following concerning radiation detection and effects a.
Consider two point gamma sources, each with 1 curie strength.
Source A gamma energy is 2 MEV and Source B gamma energy is 1 NEV.
If readings were taken at the same distance from each unshielded source with a Geiger Mueller(GM) type meter, how would the readings i
j compare?
Briefly explain. C1.03 L
b.
If a worker was exposed to a 1 RAD /HR neutron radiation field, would i
the biological dama3e be less thant 3reater than, or the same as if
-
the 1 RAD /HR field was due to samma radiation? CO.53
.
-
.
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.
~8.
-ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
'
.GUESTION 8.01 (3.00)
o During a normal reactor startup with the~ reactor at 3% powere a steam dump cyctes malfunction occurs resulting in a Tave decrease to 545 DEG F.
o.
What immediate actions must be taken in accordance with the Technical Specifications? [1.03
~
b.
What are the four bases for the Technical Specification Action requirements? [2.03 GUESTION 8.02 (2.50)
Tha plant is heating up from 310 DEG F at.35 DEG F/HR.
Maintenance reports that Residual Heat Removal (RHR) Pump 8B repairs will not be completed for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> but that RHR Pump 8A is operabler and the Technical Specifications Action Statement allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to repair an inoperable pump in Mode 3.
What action, if any, should be taken?
QUESTION. 8.03 (3.00)
During shift turnover, you notice that Axial Flux. Difference (AFD) has been cutoide of the target band for the last 10 minutes.
Power is 98% and you hovo accumulated 29 penalty minutes in the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
'
O.
What action, if any must be taken? C0.753 b.
What action, if any must be taken if power were 83%?
Why? E0.753 c.
What action, if any must be taken if you exceed 60 penalty minutes in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />?
How will this affect future operation? C1.53 GUESTION 8.04 (3.00)
At 0840, while operating at 99% powere the USS is informed that MSIV A is opcn and inoperable.
The estimated time of repair is four hours.
At 1230 tha Maintenance Supervisor states that repairs will take an additional hour
,
or co.
The USS orders load removed and power is stabilized at 98%.
The MSIV is returned to operable status at 1425 and power raised to 99%.
Did
.tho USS. violate Technical Specifications?
Explain your answer.
Refer to Figure 8 1.
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
.__________________________________________________________
QUESTION 8.05 (1.50)
'
Tho monthly surveillance test for the turbine driven emergency feedwater punp differential pressure was due to be completed on February 4.
A check
'
of the surveillance records showed that the test was not performed until Fcbruary 13.
The Operations Supervisor states _that it was still within
.
tho extension time allowed by the Technical Specifications.
Is he correct?
- - Explain your answer.
,
QUESTION 8:.04 (1.00)
Abnormal vibration in Primary Component Cooling Water Pump 11D causes the pu:p to be removed from service and ta33ed out.
Is that PCCW train cporable?
Explain y~our answer.
'
- QUESTION 8.07 (2.50)
Annwer_the following concerning los reviews by the Unit Shift Supervisor!
,
o.
What loss are reviewed? C2.03 b.
How often are they reviewed? C0.53 i
.
gGUESTION 8.08 (2.00)
-
,
'. What is the primary means of communication between the control room an'd occh of the following or3anizations during an emersency?
o.
NRC b.
Mass /NH State Police c.
Near-Tern Response Personnel j
d.
Station Security C0.5 each3 l
l QUESTION 8.09 (1.50)
'
What is the difference between the review process for intent and non-intent precoduce changes to SORC approved procedures?
QUESTION 8.10 (1.00)
What restrictions are placed on the composition of the Fire Brigade?
.
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
.
QUESTION 8.11 (2.50)
Ancwor the following concerning Radiation Work Permits (RWP):
Under what conditions may the Shift Superintendent authorize entry o.
into an area normally requiring a RWP without actually being provided a RWP prior to entry? C1.03 b.
What three actions must be taken in the above. situation to ensure that personnel safety and proper documentation are maintained? C1.53 QUESTION 8.12 (1.50)
Ancwer the following concerning Reactor Coolant System (RCS) limits:
o.
Technical Specifications provide steady state and transient RCS chemistry limits.
What is the period of time you are allowed to operate if the actual chemisty concentrations are between these two limits? C0.53 b.
RCS specific activity limits refer to the term Dose Equivalent I-131.
What is Dose Equivalent I-131? C1.03
.
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5. ~
____' THEORY OF NUCLEAR POWER PLANT OPERATIONr FLUIDS, AND PAGE
______________________________________
______________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITT0rD.
' ANSWER 5.01 (2.00)
All control rods in a single group are_ moved together with no single o.
rod in the group differing by more than 12 steps from group position.
b.
Control rod groups are sequenced with proper bank overlap.
c.
Control rod insertion limits are observed.
d.
Axial power distribution is maintained within specified limits.
'
to.5 each]
"
REFERENCE WNTC Rx Core Control, 8-32
.
ANSWER 5.02 (1.50)
Tho temperature rise in the coolant is directly proportional to the heat i
input as long as no phase change takes place. [0.753 As Vaporization
'tckoo place in the secondary, heat is added with no change in temperature.
,
.
C0.753
.
.
REFERENCE HO-THF-3
,
.
ANSWER 5.03 (.50)
-
b.
,
REFERENCE Stoco Tables HO-THF-2
[ ANSWER 5.04 (1.50)
o.
Scale reduces heat transfer ability, therefore U decreases and with constant powere Delta T must so up to compensate. CO.753 I
b.
Plugged tubes reduces heat transfer arear thereforer A decreases and with a constant powerr Delta T must so up. C0.753 i REFERENCE HO-THF-3 CAF for answer
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.
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.
.
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLU 1DS, AND PAGE
____
___________________________..__________
______________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITTO,D.
ANSWER 5.05 (2.50)
o.
11)
Criticality prediction based on count rate. C0.53 (2)
Ensures criticality does not occur below minimum Rod Insertion Limits and therefore ensures minimum Shutdown Margin. C0.53 (
b.
The approach to zero is easier to see than the approach to infinity.
C0 753 c.
As Keff approaches one, it takes lonSer for count rate to stabilize, because more generations are required to level count rate.
CO.753 REFERENCE
o.
HO-RTR-54 b.
HO-RTR-54
-
e.
CAF for answer ANSWER 5.06 (3.00)
o.
Rods will need to be withdrawn for about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> C0.753 and then inserted for th,e next 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />. CO.753 b.
After the power decrease, the production of xenon from fission Co.253 and from the, decay of iodine C0 253 is 3reater than the renoval by decay of xenon C0.253 and burnout by flux. C0.253 After five hours, the renoval rate is greater than the production C0.253 and positive reactivity is being added until equilibrium at about 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. C0.253 REFERENCE HO-RTR-109, 110 ANSWER 5.07 (3.00)
o.
Reactor Power Increases b.
Reactor Coolant Temperature Increases
.c.
Reactor Coolant Flow Decreases d.
Primary Pressure Decreases C0.75 each3
,
REFERENCE HO-THF-3 e
},'..... T
.
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e, i
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDSr AND PAGE
,
$~~~~TUERU66i AU5C5~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
______________
P ANSWERS -- SEABROOK 1-85/03i18-RUSCITT0rD.
,
.
-
ANSWER 5.08 (1.50)
A SDM ensures that the. reactor can be made suberitical from all operating ccnditions C0.53 and that the reactivity transients associated with i
pootulated accident conditions are controllable. C0.53 It ensures that the rocctor will be maintained sufficiently suberitical to preclude inadvertant
'eriticality in the shutdown condition. C0.53
,
REFERENCE RX 1707, p.2
- ANSWER 5.09 (3.00)
c.
Power decreases initially due to the boron addition.CO.53 The
primary to secondary mismatch causes TAVE to decrease.CO.53 The decrease in TAVE inserts positive reactivity and restores power to level sli htly lower than or the same as initial power.CO.53 a
(Possible low pressure tripe LO-LO TAVE implies low pressure)
b.
Tave is determined by the amount of pump heat C0.53 and the steam dump setting thus it does not change.CO.53 After the initial transient, power decreases (at a -1/3 DPM rate) to the multiplied source level.CO.53 REFERENCE HO-RTR-95/97 ANSWER 5.10 (2.50)
o.
(1)
Xe-135 production directly from fission and the decay of iodine C0.253 and removal by decay and burnout. C0 253
.
(2)
Sa-149 production from the decay of promethium C0.253 and removal l
by burnout. C0.253 b.
5300 pcm at 7-10 hours C0.753 t
c.
2800 pcm at 40-60 hours C0.753
'
REFERENCE HO-RTR-104/116
+
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' 5.
THEORY OF NUCLEAR POWER PLANT OPERATIONr FLUIDSr AND PAGE - 17
. ----
---_-----_-_--
-
ANSWERS --.SEABROOK 1-85/03/18-RUSCITT0rD.
ANSWER 5.11 (1.50)
3 o.
P2 P1(N2/N1)
= 500 X 4
= 32'MW
=
i b.
Q2
=
G1(N2/N1)
= 900 X 4
= 3600 GPM
2 c.
H2
= H1(N2/N1)
= 70 X 4
= 1120 PSIG REFERENCE-HO-THF-4
ANSWER 5.12 (2.50)
o..
As temperature decreasest. the density of the fluid increases.
More
. power is required to pump the heavier fluid.C0.53 b.
As temperature decreases, both steam density and steam velocity decrease.CO.53 The steam dump valves must be opened.further to maintain a constant mass flow rate.CO.53
.
,
c.
About 5% of the total RCS-flow bypasses the core and picks up very little heat.C0.53 In the upper plenuar these-two flows mix, reducing vessel outlet temperature.C0.53-REFERENCE HO-THF-2/4
'
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6.
PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITT0rD.
ANSWER 6.01 (2.00)
.o.
Train B provides interlock and automatic isolation for the upstream valves of both trains.
Thereforer RHR Train A (RC-V23) and RHR Train B (RC-V88) upstream isolation valves will shut and be prevented from opening. C1.253
.
t b'.
Pressure meter and' recorder on-the Main Control Board. [0.753
CCredit given for indications of RHR system isolation and pump
'
cavitation as well as indications of PORV actuatuation (If assumption is made that LTOP is armed)3-REFERENCE HO-RHRS-25/26 005000K1.09 RO.
SRO 3.9
,
005000A2.01 RO 2.7 SRO 2.9 ANSWER 6.02 (3.00)
Prevents Steam Dump Valves from shutting once impulse pressure stops o.
decreasins. C1.03 b.
The dump will remain armed and a failure low of the other impulse channel (PT-505) or a hi h failure of any temperature channel input
'
to Tave would cause dump actuation. E1.03 TURB IMPULSE CHAMBER PRESS PERMISSIVE status light on the Main c.
Control Board. C1.03
,
REFERENCE o.
HO-SD-39 b.
HO-SD-39,40 c.
HO-SD-40
.
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6.
PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITT0rD.
.
ANSWER 6.03 (4.00)
o.
(1)
Spray valves open. [0.673 (2)
PORV 456A opens until pressure channel indicates 2185 PSIG, C0.673 then PORV 456A closes. CO.673 b.
The rapid pressure transient could cause Low Pressure Trip j2/4 at
_
ggd gM//gry/gg/ pfyg4 1945 PSIG) or OT Delta T (2/4). [0.753 c.
(1)
Select alternate channel (457) on Channel Selector. [0.53 OR (2)
Turn heaters off, manually close spray valver check PORVs closed or close block valves until transient stops. [0.753 CCrodit given for an explanation of using heaters to restore pressure.]
REFERENCE Inctrument Failure Manual, pp. C1-C6 HO-PPLC-24/33 ANSWER 6.04 (4.00)
a.
(1)
Hot Zero Power C0.503 because of the greatest mass in the SG results in the largest RCS cooldown. [0.503
,
(2)
EOL CO.503 because HTC is at its maximum negative value. [0.503 b.
Flow restrictors in the S/G outlet nor:les (and BIT) [0.53 c.
(1)
Steam line low pressure 585 PSIG 2/3 any steam header (2)
Pressurizer low pressure 1850 PSIG 2/4 (3)
High containment pressure 4.3 PSIG 2/3 C0 5 each3 REFERENCE HO-ECCS-44 HO-RPS-70/72 Satpoint Summary f o s>l & n s t e wY
!
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
' ANSWERS'-- SEABROOK 1-85/03/18-RUSCITT0rD.
.
k ANSWER 6.05 (3.00)
gg,
Increases due to auctioneeredgDelta T input. (Tave input set to zero)
o.
b.
Increases to raise pressurizer level to 100% program, because of the higher temperature input.
,
c.
Rods move in, because of the Auctioneered Tave/ Tref mismatch.
!
d.
No effeetr the demand signal is present (Tave/ Tref) but
'
-there is not an armin3 SiSnal present.
[0.75 each3
,
REFERENCE Instrument Failure Reference Manual pp. Al-A5
.
ANSWER 6.06 (3.00)
o.
During startup, the flux is mainly in the bottom of-the core. [0.53
,
therefore, the source range detector is located at the botton.C0.53
,
-b.
At low power ~
'
levels with control rods inserted and during much of reactor startup operation'the flux remains low in the core. [0.53 thereforer the intermediate range detector is located at the aid l
plane. C0.53 c...The power range detector covers the full core height to.53 which enables them to determine relative axial power production. [0.53 REFERENCE HO-NIS-20/22 l ANSWER 6.07-(1.00)
Frco NCC-E531 via a 480/120 VAC transformer (and the mechanically intorlocked supply breaker)
REFERENCE HO-E'AC-24 l
.
..
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'. 4
.
4.
PLANT SYSTEMS DESIGNr C NTROL, AND INSTRUMENTATION PAGE
______________________________________________________
ANSWERS'-- SEABROOK 1-85/03/18-RUSCITT0rD.
ANSWER 6.08 (3.00)
o.
At 450 GPM, when the motor-operated flow control valves are in remote (MCB) contro1rCO.33 they will shut and the remaining three sets will be blocked open C0.43 to ensure isolation to a failed steam generator (SC) and provide feedflow to the unaffected SGs. [0.33 b.
The flow venturi in each line limits flow to 750 GPM. C0.753 c.
Alare on the MCB. C0.53 d.
Noner all automatic initiations are blocked in local operation.CO.753 REFERENCE HO-EFW-16/25,38 ANSWER 6.09 (2.00)
o.
(1)
Radiolysis of water (2)
Zirconium / Water reactions (3)
Coolant H2 inventory (4). NaOH/Zine/ Aluminum reactions CO.33 eache 3 req.]
b.
Hydrogen is explosive in concentrations above 4%. [0.53 c.
(1)
H2 recombiners (2)
Compressed air puse and vent C0.25 each3 REFERENCE ya/ acup[//demadk,or
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f 7.
LPROCEDURES-NORMAL, ABNORMAL,EMNRGENCY-AND PAGE
--- RA5iBE55iEAE 55ETEBE------------------------
____________________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITTO,D.
,
.
~r ANSWER-7.01 (3.00)
'
No immediate actions are required since all control funetions except o.-
-
rod stop are 2/4 coincide 6ce.'C1.53 b.
The' inoperable channel cust bi tripped within one hour. C1.53 REFERENCE Instrument Failure Reference Manual p. K-11 TS p. 3/4 3-6 ANSWER 7.02 (1.50)
Cloco' seal leakoff valveC1.03 w'ithin 5 minutes. C0.53
. REFERENCE 081200.01 p.3 ANSWER 7.03 (1.00)
LTo prevent reactor power variations due to changes in boron concentration.
REFERENCE 0S1002404, page 2
-
ANSWER 7.04'
(3.00)-
<
o.
(1)
Steam pressure will.be maintained (less than 1000 PSIG) with the
>
atmospheric dump valves.
(2)
Stesa generator level will be maintained (45 - 95%) with the emergency feed system.
(3)
Pressurizer pressure will be. maintained using heaters.
(4)
Pressuizer level will be. controlled by adjusting the cooldown rate.
C0.4 each]
b.'
The plant is depressurized using the pressurizer PORV. [0.63
,
' REFERENCE 081200.02, pp.2-19
.
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~ A5iBE55 5 E 55aTR5E------------------------
R
____________________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITT0rD.
ANSWER 7.05 (3.50)
o.
(1)
Pressurizer pressure less than 1850 PSIG and decreasinS.
(2)
Containment pressure greater than 4.0 PSIG.
(3)
Steanline pressure less than 585 PSIG and decreasing.
'
(4)
Reactor Coolant System subcoolin3 less than 30 DEG F.
(5)
Pressurizer level less than 5% and decreasing.
[0.5 each]
b.
-( 1 )
Close the MSIVs (2)
Open the Generator Breaker CO.5 each3 REFERENCE F-0, pp.2,3 ANSWER 7.06 (2.00)
O.
350 DEG F/425.PSIG E0.53 b.
220 PSID/15 PSIG C0.53 c.
220 DEG F E0.25]
.
d.
350 DEG P C0.253 o.
50 PPM
[0.253
'
f.
320 DEG F CO.253 REFERENCE OS1000.01 pp. 3,4
-
051001 05 P.
l
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.
-
-.
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.
>
.
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~
RA6E L6656AL 66 TR6L'~~~~~~~~~~~~~~~~~~~~~~~
~
____________________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITT0rD.
.
ANSWER 7.07 (3.00)
o.
(1)
Reactor Coolant System boron concentration less than 2000 PPM or Keff greater than 0.95. [0.53 (2)
The 1/M plots indicate that criticality is immin.ent. E0.53 b.
(1)
An unexplained decrease in the boron concentration greater than i
20 PPM from nominal.
(2)
An unanticipated. increase in the count rate by a factor of five occurs on any channel during any step after loading of the eight initial fuel assemblies.
(3)
Audible count rate in the containment or control room is lost.
(4)
Loss of communications between the control room and the refueling station.
(5)
The containment evacuation alarm is sounded.
(6)
Mechanical damage to a fuel assembly occurs.
f (7)
Containment integrity is lost.
l (8)
Less than 1/2 CPS attributable to core neutrons exist on three of the four channels following installation of the initial eight assemblies.
-
l (9)
An unanticipated increase in count rates by a factor of two L
occurs on all channels during any step after loading of the l
eight initial assemblies.
[0 5 eache 4 required 3 REFERENCE 1-ST-4 pp.. 7-9 l
lANSWER 7.08 (3.50)
o.
(1)
Sufficient capacity to remove decay heat and maintain temperature less than 140 DEG F.
[0.753 (2)
Sufficient circulation to minimize the effects of a boron dilution accident and prevent stratification. [0.753 b.
Ensure that a single failure will not result in a loss of heat removal capability (with a diminished heat sink). E1.03 c.
May be stopped for one hour per eight hour period for core alterations in the vicinity of the hot legs. ti.03 REFERENCE TS p. 3/4 9-8, 83/4 9-2 1'
!
.
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. - -
.
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- -.
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.z 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~EI656E66565E~665TR E~~~~~~~~~~~~~~~~~~~~~~~~
____________________
' ANSWERS -- SEABROOK 1-85/03/18-RUSCITT0rD.
ANSWER 7.09 (3.00)
o.
'(1)
Change in Delta I indication
'
(2)-
Control rod bank rapidly stepping out(Auto only)
(3)
Decreasing Tave (4)' Decreasing pressurizer level (5)
Decreasing pressurizer pressure (6)
Decreasing flux on one or more channels (7)
Rod position indication / Rod bottom light CO.25 each, 4 required)
(8)
Tave deviates from Tref (9)
Decreasing steam pressure
b.
Place rod control in manual CO.53, and reduce turbine load to maintain Tave/ Tref. C0.5] 46 3 c.
Power to be reduced by 30% of rated thermal power CO.53 within 30 minutes. C0.53 A;//! 8CCf M
/
O REFERENCE'
S1 10 5 p
'guyb".
. ANSWER 7.10 (1.50)
o.
The readings would be approximately the same CO.53.
GM meter readin3s are not dependent on the ener3y level of the source since each interaction results in complete ionization of the gas in the detector, giving a pulse. CO.53 (If a compensatin3 filter is assumed in GM tuber full credit will be given for Source A reading approximately twice Source B reading)
b.
Greater-than. CO.53 REFERENCE CAF
,,
.
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.
-
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.
,
,
..,
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.
,
8.
. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE
4.__________________________________________________________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITT0rD.
.
' ANSWER 8.01 (3.00)
- o.
Restore Tave to greater than 551 DEG F within 15 minutes Co.53'or be in hot standby within the next 15 minutes C0.53 b.
This limitation is required to ensure that:
'
(1)
MTC is within its analyzed temperature range.
(2)' Protective instrumention is within its analyzed temperature range.
(3)
The pressurizer is capable of being operated with a steam
- -
bubble.
(4)-.The reactor pressure vessel is above minimum RTNDT. [0.5 each3
.
i REFERENCE TS pp. 3/4 1-6, B 3/4 1-2 ANSWER 8.02 (2.50)
.Tha Technical Spcifications require that all LCOs be satisfied prior to
]
cntry into an operational mode. C0.753 Since you are about to enter Mode 3 CO.753 the heatup must be disc.ontinued CO.5-3 and temperature' held at less then 350 DEG F until RHR Pump 8B is proven operable. C0.53 REFERENCE
,
TS p. 3/4 0-1,2
- ANSWER 8.03 (3.00)
l o.
Reduce power to less than 90%. C0.753 b.'
None required because accumulated penalty is within the limits of the Technical Specifications. (60 MIN) CO.753
'c.
Reduce power to less than 50% within 30 minutes. [0.753 Operation at greater than 50% may not resume until penalty minutes fall below 60 in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. CO.753 REFERENCE j.RN 1740, p.4
' TSr p.3/4 2-1 i
.. -.. ~.. -,, _, _ _,, - _.., _. - _ _ _.. _ _ _.
,.-_,_...--_...___,__.-,_.______,___,.u._-..-_,,,m.,
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.
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.
.
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.
3.
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.
~8.
ADMINISTRATIVE PROCEDURESr C
!':_________________________________ONDITIONSr AND LIMITATIONS PAGE
_________________________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITT0rD.
ANSWER 8.04 (3.00)
Tho~ Technical Specification was violated because it requires the operator to take action to have power below 5% two hours after the four hour repair tico has expired.
Lowering power one percent commenced load reduction but by 1425, load could not be reduced fast enough to achieve 5% within the n0xt 15 minutes and he was therefore in violation.
REFERENCE TS p.3/4 7-9 ANSWER 8.05 (1.50)
Nor because the maximum extension is 25% of the specified time intervale
.
which in-this case has been exceeded.
+
REFERENCE TS p.3/4 0-2 ANSWER 8.06 (1.00)
,
Yoor E0.5] Standing Order 84-D04 sives operational guidelines on the Tcchnical Specification interpretation. [0.53 REFERENCE Standing Operating Order 84-004 ANSWER 8.07 (2.50)
o.
(1)
Tagging Order Log (2)
-
(3)
Station Loss (4)
USS Journal CO.5 each]
!
b.
Once per shift E0.53 REFERENCE AG 10.003'
,
,
,
+
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-
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..
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.-. - -.... - _. -, -. - - -. ~.
,. i s.
.a
.. *....
.
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
__________________________________________________________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITT0rD.
ANSWER 8.08 (2.00)
g/[NNh SN1M c.
ENS Hotline
[b.
Nuclear Alert System
c.
Radio Pasing System fN
/
'
i.
Telephone /PA System C)wC each3'
REFERENCE M
Ecorsency Plan Figure 3.1 ANSWER 8.09 (1.50)
'
.Prict to implimentation, intent changes must be reviewed by SORC and signed by the Station Manager CO.753, whereas non-intent chanSes are reviewed by SORC within 14 days but approval only requires concurrance of both a Station Staff Supervisor knowledsable in the area affected and either the Unit Shift Supervisor or Shift Superintendant. C0.753 REFERENCE AQ1.002, p.15, 16
,
ANSWER 8.10 (1.00)
Tho Fire Brigade shall not include the Shift Superintendent and the three othcr members of the minimum shift crew necessary for the safe shutdown of tho unit (and any personnel required for other essential functions during a firo emersency).
REFERENCE T.S. 6.2.2
- ANSWER 8.11 (2.50)
a.
When critical and immediate action is required C0.53 for work of a short duration. C0.53 b.
(1)
The workers are given verbal instructions and precautions.
(2)
Continuous health physics coverage is provided.
(3)
The job is documented on a RWP after completion.
CO.5 each]
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8.
. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE
__________________________________________________________
ANSWERS -- SEABROOK 1-85/03/18-RUSCITT0rD.
.
REFERENCE RP-1 0, p. 5'
ANSWER-8.12 (1.50)
c.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C0.53 b.
Dose Equivalent I-131 is that concentration of I-131 which alone would produce the same thyroid-dose as the quantity and isotopic mixture of I-131, I-132r I-133r I-134 and I-135 actually present.
[1.03-REFERENCE
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TS 3/4 4-22 b.
TS 1-2
.
l I
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TFST CRDSS REFERENCE'
PAGE
IUESTIDN VALUE-REFERENCE
._______
______
-_________
'05.01-2.00 DGR0000816 05.02 1.50 DGR0000817
~
- 05,03
.50 DGR0000818 05.04 1.50-DGR0000819 05.05 2.50-DGR0000820 05.04 3.00 DGR0000821 05.07 3.00 DGR0000822
'05.08 1.50 DGR0000823 05 09-3.00 DGR0000825 05.10 2.50--
DGR0000826 05.11 1.50
. DGR0000827 05i12 2.50 DGR0000828
______
25.00 06.01 2.00 DGR0000832
. 06.'02 3.00 DGR0000833
' 06.03 4.00 DGR0000834 04.04 4.00 DGR0000835 06.05 3.00 DGR0000836 06.04 3.00 DGR0000839
.06.07 1 00-DGR0000840 06.08 3.00 DGR0000841
- .
'06.09 2.00 DGR0000842
______
,
25.00 07.01 3.00 DGR0000838 07.02 1.50 DGR0000846 07.03 1.00 DGR0000858 07.04 3.00 DGR0000859 07.05 3.50 DGR0000860
-
07.06 2.00 DGR0000864 07.07 3.00 DGR0000865
,
i07.08
'3.50 DGR0000866 L07.09 13.00 DGR0000867 07.10 1.50 DGR0000868
______
25.00 08.01 3.00 DGR0000824
'
08.02-2.50 DGR0000837
- 08.03 3.00 DGR0000843 08.04 3.00 DGR0000844 08.05 1.50 DGR0000845
- 08.06 1.00 DGR0000850
'
08.07 2.50 DGR0000851 08.08 2 00-DGR0000852 08.09 1.50 DGR0000854 08.10 1.00 DGR0000855
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TEST CROSS REFERENCE PAGE
'UE VALUE REFERENCE L_STION _______
__________
_____
08.11 2.50 DGR0000856 08.12 1 50 DGR0000857
______'
25.00
______
______
100.00 t
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EQUATION SH E T
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f = ma y = s/t Cycle efficiency = (.1stwork out)/(Energy in)
.
w = ag s = V,t + 1/2 at
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E = mc
,
,
A = A,e^
KE = 1/2 av a = (Vf - V,)/t A = AN PE = agn Vf = V, + at w = e/t 1 = an2/tifg = 0.593/t1/2
t eff * C(tm)-(~%)]
w=,g rD 1/Z
-
A=
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[(c /2) * (E )J l
b
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d = 931 as
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Q = scaat g = UA47 I=I,e*
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w. up i = t,10-#!"-
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TYL = 1.3/s
.
sur(t)
MVL = -0.693/s
P = P 10 P = P,e*/
SUR = 28.06/7 SG = 5/(1 - Kgf)
p G * 3/(I ~ "Wfx)
x
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gf3) = G (I ~ " #f2)
SUR = 25e/s* + (s - e)T G (1 - K
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T = ( c/s) + [(s - s yI.]
M = 1/(1 - Kgf) = CR /G,
j
,
T = a/(s - s)
M = (1 - Kgf,)/(1 - Kgfi)
T = (s - s)/(T.)
SDM=(}-Kgf)/Kgf s= = 10 seconds a = (Kgf-1)/Kg f = 4Kgp/Kgf I = 0.1 seconds-gf(1+IT)]
e=[(s*/(TKgf)] + [T
/
Idli*Id 2 =2 2
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P = (r4V)/(3 x 1010)
Id I4 j
2 l
t = eN R/hr = (0.5 G )/d (,,g,,)
l R/hr = 6 2/d2 (f,,g)
-
,
W4 tar Parameters Miscellaneous Conversions 1 gal. = 8.345 lem.
I curie = 3.7 x 1010dps
1 gal. = 3.78 liters 1 kg = 2.21 lho 3 Stu/ar 1 ft" = 7.48 gal.
1 np = 2.54 x 10 Density = 62.4 ling /ft3 1 mv = 3.41 x 106 Stu/hr Density = 1 ga/c:e lin = 2.54 cm Heat of vaporization = 970 Stu/les Y = 9/5'C + 32 '
Heat of fusion = 144 Stu/les
- C = 5/9 ( T-32)
1 Ato = 14.7 psi = 29.9 in. Hg.
1 STU = 778 ft-lbf 1 ft. H O = 0.4335 lbf/in.
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MAIN STEAM LINE ISOLATION VALVES
.
.
-
LIMITING CONDITION POR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.
f APPLICABILITY:
MODES 1, 2 and 3.
.
ACTION:
.
%
-
.
MODE 1 - With one' main steam line isolation valve inoperable but open. POWER
'
o j
OPERATION may continue provided the inoperable valve is restored to
'
OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise reduce power to less than or equal to 5 percent of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
,
MODE 2 - With one main steam line isolation valve inoperable, subsequent and 3'
-operation in MODES 2 or 3 may proceed provided:
a.
The isolation valve is maintained closed.
'
i b.
The provisions of Specification 3.0.4 are not applicable.
-
-
i
.
Otherwise, be in MOT STANDdY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MOT
'
SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS
.
4.7.1.5 hach main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 5.0 seconds when tested pursuant to Specification 4.0.5.
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.
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-.
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FIG.8.I
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Seabrook - Units 1 & 2
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