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Exam Rept 50-443/85-04 on 850318-22.Exam results:16 Senior Reactor Operators & 3 Reactor Operators Passed All Portions of Exam.One Senior Reactor Operator Failed Simulator Portion of Exam
	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: IR 05000443/1985004; ML20129E193;
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: 8# ~

Reactor Engineer Examiner) Date

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REVIEWED BY: [ -) 1383 Chief, Projects Section 1C Date APPROVED BY:

Chfef, Profebts Branch N /

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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 averag ADOCK O 3

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6 2 REPORT DETAILS

. TYPE.0F EXAMS: Initial

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EXAM RESULTS:

I R0- l SR0 I

.I Pass / Fail- 1. Pass / Fail I

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.lWritten. Exa . 2/0 1 17/0 'l- 1 -l i I I l-10ral Exam I 3/0 - l 16/0 l I I I I I .1- l .

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~ISimulator Examl 3/0 1 15/1 1 I I I I l l .

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.10verall I 3/0 /1 I l- 1 I l 1 I I l

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1. ' CHIEF EXAMINER AT SITE: .D.-Ruscitto_(NRC)

2. 'OTHER EXAMINERS: R. _ Sailor -(EG&G)

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B. - Picker (EG&G) .

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 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 simulatio . INTERFACE WITH PLANT STAFF DURING EXAM PERIOD:

' The simulator instructors continue to perform their liaison duties in an outstanding manner, greatly enhancing the examination proces This allows last minute scenario modifications to examine new areas with little pause in the flow of the scenari 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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~4 g o PERSONNEL PRESENT AT EXIT INTERVIEW:

NRC Personnel D. Ruscitto, Reactor Engineer (Examiner)

~ R. Barkley, Reactor Engineer A. Cerne, Senior Resident Inspector R. Cochran,-Consultant

!' 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

' 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 wel 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 examinatio . 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 condition D. Schreiner~ S. Simonson J. Nichols R. Hanley J. Reagan R. Mayes

' Attachments: Written Examination (s) and Answer Key (s) (SR0/R0) 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 ke Reference Drawing was.provided at the exam revie _ Question 3.048 The FWP discharge valves do not close on High S/G water leve Reference _is Dwg. No. 9763-M-50358 Question 3.09B "Lo Lo S/G level" is appropriate als Referenc'e is FSAR, Table 15.0-6, Sheet Question 4.03A- " Rapid drop in nuclear power" is an- additional answer. IAW procedure for dropped ro Answer changed to. include above phras SRO EXAM COMMENTS AND RESOLUTION:

Question 6.038 _"0 TAT or Low Pressure Rx Trip" are acceptabl Reference is FSAR, Table 15.0-6, Sheet 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 ' Log 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 answer Question 8.08 Means of communicating may not match the answer given exactl 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 require 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 chang __

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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: .1EA3 GOOK 1_---- __

REACTOR TYPE: _

pWR-WEc4 _

. .DATE ADMINISTERED *_11493/is

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EXAMINER ISAKSEN. P.

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APPLICANT: ,

IB11gEI1QU_ID_&PPLICAE1 Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheet 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 80 Examination papers will be picked up six (6) hours after

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3 0F CATEGORY % OF APPLICANT'S CATEGORY vALUE_ TOTAL SCDgg___ VALUg__ ___CAlggggY_

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_25. 0D _ 25.00 _- _____ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

_25. 00 _ . .21aga ____

- PLANT DESIGN INCLUDING S AFETY AND EMERGENCY SYSTEMS l _22a90- _21agg __ ______ INSTRUMENTS AND CONTROLS l

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______ __ ___ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADI0 LOGICAL CONTROL

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__ TOTALS FINAL GRADE ______ ____%

All work done on this examination is my own. I have neither giv0n nor received al APPLICANTI S SIGNAiURE i MASTERCOPY

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- PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 2 IMER110D INAalC S4JfE A I_IRANSEER _ AND_ EL UID _ELOW QU5STIDM 1.01 (3.00)

Acstee . Rod control in manua . No operator actio . No protective function actuations.- 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

. ~Teve 'Tfuel

' Reactor power

-~ TRANSIENTS (consider each separately) Steam dump valve f alls open with the reactor critical below the point of adding hea . (1.45)

' Control rod drops at 502 power with turbine controls in aut (1.55)

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) Explain WHY the Moderator Temperature Coefficient becomes more negative over core life. (Two reasons required) (1.0)

! At 80L and as fuel temperature increasess WHY does the

! magnitude of the Doppler Temperature Coefficient decrease Per degree change in fuel temperature? (1.0)

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. lu PRI N CI PLELQE NUC LEARRML&HLD P E R AT I ON n PAGE 3 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months af ter a trip from 100* Powers to th3 ACTUAL critical control rod position if the following events / '

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 positio ,

' The FOURTH coolant pump is started two minutes prior to criticalit (0.5) The startup is delayed until 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , af ter the tri (0.5) The steam dump pressure setpoint is increased to a value Just below the Steam Generator PORV setpoin (0.5)

d.- Condenser vacuum is reduced by 4 inches of Mercur (0.5)

! All Steam Generator levels are rapidly being raised by 51 as criticality is reache (0.5)

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. eUESTIDM 1.05 (1.50)

l o.< Provide TWO reasons for Xenon contributing more negative i

reactivity than Samarium et full powe (1.0) 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 CHOIC (1.25)

s l- For two equivalent positive reactivity additions to a critical r reactors will the SUR be the Same, Largers or' Suslier at EOL as

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compared to 80L7 NO EXPL ANATION IS NECESSARY.- (0 5)

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.1 PRINCIPLES OF NUCLEAR POWER PLANT DPERATIO PAGE 4 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) Clad cree (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 independentl Reactor power.

i RCS flo '

< Tcol 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 indicate . Affected loop steam generator level (INITIAL change only). Affected loop steam generator ~ pressur . Affected loop cold les temperatur . Unaffected loop steam generator level (INITIAL change only). Unaffected loop steam generator pressur . Unaffected loop cold leg temperatur (2.4)

1 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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I .la_. PRINCIPLES OF_ NUCLE AR POWER PL ANT OPER ATIONn PAGE 5 THERMODYNAHICSn HEAT. TRANSFER AND FLUID FLOW

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, - 0UESTION 1.10 (2.75) 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) Enthalpy (h) . Pressure

! Entropy (s)

' Specific volume (v)

' Temperature

' . State whether the steam will be subcooled, saturated or

superheated as it leaks ou (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) Choose the answer that mort correctly completes the sentence.

i "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." a higher head and higher flow rat . the same head and the same flow rate.-

> the same head and a higher flow rate..

a higher head and the same flow rat (0.5) 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 6 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 specifi . EFW supply header ventur (1.5)

, EFW supply header flow orific (1.0)

l l 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)

i What function is performed by the Air Relay Dump Val ve ( ARDV) on a j turbine trip AND why is this action necessary? (1.5)

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QUESTION 2 03 (2.50) The containment structure air is sampled by hydrogen analyzer Following a LOCA ("P" signal)s what automatic action ensures that a representative sample of containment air is sampled? (1.0) 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) What system serves as a backup for the hydrogen recombinerst (1.0)

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. PL A N I D ES IGil_IAICLUQ1BS_3AEEIX_ AND_EHE R G E N CX_ S Y S T E MS PAGE 7 0UESTION 2.04 (4.00) State the setpoints and coincidence which will automatically initiate Containment Building Spray (CBS). (1.0)

. What. is the purpose of inJactins sodium hydroxide (NaOH) into the spray water? (1.0) Descr ibe how NaOH enters the CBS Syste (1 0) What conditions must be met for containment recirculation sump Isolation valves (V-8, V-14) to automatically opent (0 5) True or False The CBS pumps will continue to operate during the switchover to the recirculation phase of operatio (0 5)

l GUESTION 2.05 (3.00) Assume the plant is shutdown at 500 F and 2200 psis. Will the

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Low Temperature Overpressure Protection System act to reduce plant pressure immediately if a loop Tcold instrument f alls LDWT BRIEFLY EXPLAIN your answers INCLUDING which train is affected, what action takes pl ace, and any applicable setpoint (2.0) 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 (1.5) What TWO relays are reset by the EDG Control Roset Switch on the NC (1.0)

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Z a__kkASI_ D E S I GM_ISCLilRIES_SAEEII_ AND_EHERraEtiCI_ SISIE53 PAGE 8 QUESTION 2 07 (2.50)

Tha following questions concern the Primary Component Coo, ling (PCCW)

Systems at the Seabrook Statio What TWO PCCW alarms could indicate a RCS to PCCW leakt (1.0) 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 Loo (Include any applicable setpoints.) (1 5)

QUESTION 2.08 (3 50)

An3ter the following questions utilizing the attached drawings Figure 2- Indicate the "f ailed" position for the valves labeled A through (2.1) 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) What are the TWO purposes of PCV-1317 (1 0)

QUESTION 2.09 (2 00) What THREE signals other than manual will cause Maln Steamline Isolation? (Include Setpoints) (1 0)

! What are TWO purposes of the Main Steam Flow Restrictor? (1.0)

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1a__'INSTRuggES AllD_CQ E RDL1 PAGE 9 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 separatel The normal steam pressure setpoint is reduced by 92 psi while in Hot Standby awaiting reactor startu (1.0) The train A steam dump selector switch is placed in 'of f' while at 55 reactor power awaiting turbine startu (1.0) Train 8 reactor trip breaker f alls to open upon a trip from 785 power. NOTE Train A breaker open (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 l

, Lower Quadrant Power Tilt Ratio (GPTR) Delta Flux (Axial Flux) Indication (Channel 4)

f S/G Feed Flow (Initial FRV response) (1.5)

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QUESTION 3 03 (2.50)

i l What associated operator action would be required during plant shutdown if an Intermediate Range Nuclear Instrumentation (IRNI)

channel was grossly undercompensatedt (1.0) List THREE Control / Protection functions provided by the IRNIf i (Include setpoints and logics.) (1.5)

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3.. ' INSTRUMENTS AND CONTRDLS PAGE 10 QUESTION 3.04 (3.50) 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) Provide ALL direct and immediate indirect additional automatic actions associated with P-14, other than feedwater isolation valve closur (1.8) What protection is provided (reason / basis) by the P-14 signalf (0 7)

GUESTION 3.05 (3.00) f What input signal is used to provide the programmed pressurizer level for pressurizer level controlf (0 5)

i The" CONTROLLING pressurizer level channel falls HIGH during 1005

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power operatio 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 result in reactor tri (2 5)

l-l QUESTION 3.06 (3.00)

Tha reactor is at 801 power with rod control in automati Assume a rapid lot load rejection occurs: Briefly EXPLAIN how a rod insertion signal is generated by the rod control systems - Power mismatch circult? (1.0) temperature mismatch circult? (1.0) 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 11 QUESTION 3.07 (3.50)

0.- List the FOUR plant parameter input signals to the Overtemper ature Delta-T (OTdT) protection circui (2.0) 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 provide (1.0)

GUESTION 3.08 (2 50) Describe the opening interlocks associated with the following valve Include setpoints where applicabl . Hot Leg Suction to RH (1.0) Recirculation Sump Suction to RH (0.5) State the TWO conditions that must be satisfied for Automatic Recirculation Sump valve openin (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 powe (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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, .4 . PROCEDURES - NORMALS ARNDRMALs EMERGENCY AND PAGE 12 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 ope (0.6)

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b. Whys when RCS pressure is <100 psis, must the RCP 81 seal

. leakoff valves be closedt (0.6) 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)

! 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

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which you may choose between to meet the Technical Specification requirements? Include time limitation (1.5) Assume that It is 0310 on 03/18/85 and the plant is presently

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at 452 power. Considering the AFD penalty history belows at what date and time may power be increased above 5057 EXPLAI (Show all work.) Assume no deviation outside the band af ter 0310 on 03/18/8 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 13 RADIDLDEICAL COMIROL QUES TION 4 03 (3.50)

Anscer the following according to 031210.5, Dropped Rod procedur List FIVE symptoms of a dropped RCC (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)

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QUESTION 4.04 (3.50)

Anster the following according to E-3 Steam Generator Tube Rupture (SGTR) procedure 05133 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? EXPLAI (0.5) 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 SGT (1.0)

de State the ECCS reinitiation criteria during a SGT (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 14 RADIDLOgig&L,ggglggt GUESTION 4.06 (2.50)

Anster the following according to 031200.02, Safe Shutdown and COstdown from the Remote Safe Shutdown Facilitie 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 boar (0.5)

GUESTION 4.07 (1.50)

Anster the following in accordance with Standing Operating Order No.84-001, Use of UNTESTED procedure 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

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to the UNTESTED procedure for the conditlen noted in as abovet (0.5)

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. PEDcEnggES - AIDRBALa_AAMDEBAla EBEREEHCI_ARQ PAGE 15 AAR1DLDEIGAL.GDMIEDL QUESTION 4.08 (2 00) 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 Superintenden . Personnel entering are given verbal Instructions and precaution . Continuous health physics coverag . Personnel entering must wear protective clothing over their personal clothin . The Job is documented on a RWP after completio (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 2 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 2 ,

KE = 1/2 av a=(Vf - V,)/t A = AN A = A g e'**

'

PE = agn Vf = V, + at w = e/t i = an2/t1/2 = 0.693/t1/2 g , ,3p

-

A=

,g 2 *1/2W = buiMtW

, [(g/2)*(*b}3

aE = 931 m *

- Y av A8

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Q = aCpat

, Q = UA A T I = I,e"#

Pwr = Wfah I = I,10"* M L -

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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,)

SUR=26e/s*+(s-e)T CRj (1 - K gfj) = CR2 (I - "e #2)

, T=(s*/s)+[(s-eVIs3 M*I/II~Kdf) = CAj A ,

T = 4/(, - s) M = (1 - Keffo)/(1-kdf1)

T=(s-e)/(Is) Kdf)/Edf o = (Kgf-1)/Kg f = 4Kgf/Kgf SOM"(h*secondj s' = 10 A = 0.1 seconds

-

e=((t*/(TKgf)]+[Tgf(1+IT)] /

Idl1*Id Id j 2 ,2 gd

P = (s4V)/(3 x 1010) 22 2 t = eN R/hr=(0.5CE)/d(,,g,,,)

l

' R/hr = 6 CE/d2 (f,,g) .

'

Water Parameters Miscellaneous Conversions

- 1 gal. = 8.345 le curia = 3.7 x 1010dps 1 ga;. = 3.78 liters I kg = 2.21 lba 1 ft* = 7.48 ga I hp = 2.54 x 10 3 8tu/hr 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. H STU = 778 ft-Ibf 1 ft. H 20 = 0.4335 lbf/i ,

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. la__ERINCIELE1_DE_HuCLEAE_EDER_EL AtiI_DEEEAIIDM PAGE 16 ItiE850AXHA51C14._ HEAT TRAN$ggg,AND_E(ylg_E(Q l ANSWERS -- SEABROOK 1 -85/03/18-ISAKSENs 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 Taves lower than initial (0 253 Tfuelt hi gher than initi al C0.253 Reactor powers higher than initial (0.253 (1.45) 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 Tavet lower than initial (0.25] Tfuels lower in fuel near dropped rod and higher in fuel remote f rom dropped rod (overall Tfuel is lower) [0.253 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 - 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

- ._ _ ._ _ __ _ - _ _ _ _ _ . . _ _ _ _ _ _ ~

- .,'

.

.la_.PRIMCIPLES OF MuCLEAE_fGMER_fLMI_QF.ERAI1QK PAGE 17 1H1150DIBABICla._tfEAI_IBAHIEEB ABD_ELilIR_ELQM ANSWERS -- SEABROOK 1 -85/03/18-ISAKSEN, 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 5 (1 0)

b FTC becomes less negative as fuel temperature lacreases because resonant band uldening is reduced at higher tempe r a tur e (1 0)

REFERENCE HD-RTR-76 to-78,-85 ANSWER 1.04 (2.50)

oo SANE

.

b ECP LOWER than ACP ECP LOWER than ACP de SAME ECP HIGHER than ACP [0.5 each) (2.5)

REFERENCE RN-1735 ANSWER 1.05 (1.50) . Higher fission ylei . Larger (thermal) absorbtlon cross section (1 0)

ho Renelns the same (0.5)

'

REFERENCE HD-RTR-111,-114

.

.. ...

1 2815CIPLES DE.BUCLEAE_EDMER_ELAtlI_DEERATIDMn PAGE 18 IHEREDDYMAMICSm_ HEAT TRANSFER AMD FLUID FLOW

' ANSWERS -- SEABROOK 1 -85/03/18-ISAKSEN, ANSWER 1.06 (1.75) Decreases [0.53 Pu 239 concentration increases (while U 235 concentration decreases) [0.75 (1 25)

bo Largey SU (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 [0.5 (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 5 (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 19 THERMODYNAMIC 1m HEAT TRANSEER AND FLUID.ELQM ANSWERS -- SEA 8 ROOK 1 -85/03/18-ISAKSENs i

,

~

ANSWER 1.09 (3.00)

ce DECREASE 2:. INCREASE INCREASE INCREASE DECREASE 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) . SANE DECREASE INCREASE INCREASE DECREASE C0.4 each3 (2.0)

b. Superheated (0.75)

REFERENCE THF-1 ANSWER 1.11 (3.00)

,

3 3 Power (2) = Power (1) * (NZ/N1) = 300 * (4) = 19.2 MW (0.5)

2 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) DECREASES (0.5)

de Pump efficiency and flowrate are reduced and mechanical pump damage (erosions pitting and vibration)., (0.5)

- . - . - . - . - _ - - _ - - - - - . - . - , . . . - - . - - . - . . . . - . _ _ . . .

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. 1 P R I NC I PLELQg_gELE Ag_gDMER_gL AgI_gggg&Ilgg3, - PAGE 20 THERMODYNAMICSa HEAT TRANSFER AND FLUTD__ FLOW ANSWERS - SEABROOK 1 -85/03/18-ISAKSEN, REFERENCE-THF-4

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.Z . PLAgI,gg31gg_1ggkj,lglgg_1&FETY AMD_EBEREENCL SYSTEMS

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PAGE 21-ANSWERS -- SEABROOK 1 -85/03/18-ISAKSENs ANSWER 2.01 (3.50) . 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) Provides input f or the hi gh flow isolation signa l of the

-

faulted header (closure of the FCV at 450 ppm) (1.0) One of the required (safety related) sources for makeup water to the Spent Fuel Poo (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 C0.75 (1.5)

REFERENCE HD-TGN-116, 117, 123 ANSWER 2 03 (2 50) The containment recirculating filter system fans star (1 0)

l (0.5) Containment Structure Purge Syste (1.0)

REFERENCE-

, NO-CGCS s20-22s 373 0S1023.40 L

i

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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22 ANSWERS -- SEA 8R00K 1 -85/03/18-ISAKSEN, ANSWER 2 04- (4.00) psig (HI-3) 2/ (1.0) To aid in lodine removal and retentio (1.0) From the S AT which gravity drains to the RWST mixing chamber (through two parallel lines) to the suction of the CBS pump (1.0)

do RWS T . LOW-L OW level in conjunction wi th SIS.; (0 5)

True (0 5)

REFERENCE HO-C BS- 18, 20s 23, 54 '

.

ANSWER 2.05 (3.00) 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 l

Its respective Isolatlen Valve (V-122) is opened. [1.03 (2.0)

i It prevents depressurization of the RCS past the interlock

setpoint (2185 psis) If a pressure channel selected to a l PORV falls hig (1.0)

REFERENCE HD-PPLC-31, 33-35.

, ANSWER 2 06 (2.50)

l l . Overspeed i Low lube oil pressure Generator differential lockou (1.5)

be Emergency start Engine trouble shutdown relay (1.0)

3 sm r=was %

i REFERENCE HD-EDm-61-64 j l

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. 2.. "PL ANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 ANSWERS -- SEA 8 ROOK 1 -85/03/18-ISAKSEN, 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) (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) A (LCV 460s459) CLOSED 8 (TCV 3818) OPEN .

C ( HC V 128 ) CLOSED i 0 (PCU 131) OPEN l E (FCV-121) OPEN

! F (HCU-182) OPEN G (V59) OPEN [0 3 each) (2.1)

l 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

_ _ , . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , . _ _ _ _ _ _ _ _

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3 d IMSIRUMEjQi.AHR.CDMIRQL1 PAGE 24 ANS WERS4-- SEASRDOK 1 -85/03/1g-ISAKSEN,, ; *

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ANSWER 3.01 ( 3'. 0 0 ) 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)

^

i be Secondary pressure would rise to the setpoint of the secondary

. atmospheric relief valves to.53 unich would maintain pressure at 11252psis (0.253 and primary temperature 560 +/-1 F [0.25 (1.0) A signal by the Load Rejectlen controller [0.53 would control primary temperature et,"Ho Load" Tref +2 F deviation (deed bands (559 F) (0.5 _

(1.0)

.

REFERENCE ,

i Scabrook H0 SD p. 17, 27, 34, 4 MS Dump Control Logic Print C-509050

.

<

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ANSWER 3.02 (1 50)

, INCREASE i

be NORE NEGATIVE se INCRE ASE (FRV OPENS) [0.5 each3 (1.5)

,

REFERENCE Scabrook Instrument Failure Reference Manual, Sectlen K, ,6,s,9

.

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ANSWER 3 03 (2.50)

as. Operater will he required to uanually reinstate power to the Seures Range instrue.ent (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, i NO-MIS-74 !

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"Ra- iMSTRugggli_Agg_gggIggk2 PAGE 25 ANSWERS -- SEABROOK 1 -85/03/18-ISAKSEN, ANSWER 3.04 (3.50)

Go SI Rx trip ooincident with Low Teve (1.0)

be Feedwater pumps trip 0, ;;.e.s e ;7 PJ v..r 0 ;;t.;;;e ;;!::: 1 Block start of start-up f eed pump

Turbine trip Reactor Trip (>P9) FRV and bypass valves shut [0.Sseach] (1.8)

ce protect the Turbine from S/G moisture carryove (0.7)

.

REFERENCE Scabrook Logic Print C-509053, HO-RPS p. 79, HO-IS-52 ANSWER 3 05 (3.00)

so Auctioneered high Tav (0.5)

bo Charging flow decreases [0 53, pressurizer level decreases (0.51, letdown isol ates [0.53, and pressurizer level increases [0.5 High Pressurizer level trip (924) [0.5 (2.5)

REFERENCE Scobrook Instrument Failure Reference Manual, Section Ds p. 1-4

ANSWER 3.06 (3.00) . 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) 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 signa (0 5)

oo When Tevs and Tref are within 1 (0.5)

REFERENCE Scabrook HD-FLRC p. 20-23

... <:, ,

,:

la_IIMSIRUMENIS AND_ CONTROLS PAGE 26 ANSWERS ---SEABROOK 1 -85/03/18-ISAKSENs ANSWER 3.07 (3.50) Taves dts Pressures dI CO.5-each3 (2 0) Prevent exceeding DNB (0.5) Turbine runback Blocks automatic and manual rod withdrawal 0k$(0.5 each3 (1 0)

,

REFERENCE Scobrook HD-RPS, p.54-59

,

j ANSWER 3.08 (2 50) .- -

RCS pressure < 365 psig. (auto close at 660 psig).

-- RHRS to CVCS/ SIS isolation valves (RH-V 35 and 36) close (1.0) RCS suction valves to RHR close (0 5) . RWST Low-Low leve . "S" signal presen (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

. _ _ . _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ . . . _ . . . _ _ _

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. ia__ PR OC E B U RE1_::_BDEB ALa_ ASHORBALa_EBEEEENCI_ ARQ PAGE 27 RADIDLDGICAL COBJ,RQL ANSWERS -- SEABROOK 1 -85/03/18-ISAKSEN, ANSWER 4.01 (3.00)

c.. Ensures both RHR reliefs are available to combat a pressure spike.(0.6)

b. Prevent backflow through the seal (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) 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)

- Accumulated penalty over the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 = 8 [0.253 65% 1637-1557 = 40 CO.253 45% 0310-0148 = 82/2 = 41 [0.53

__

89 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. .

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. PROCEDURES - NDRMALs ABNDR.MALs. EMERGENCY AND PAGE 28 RADIQLOGICAL_GQHIRQL i ANSWERS -- SEABROOK 1 -85/03/18-ISAKSENs ANSWER 4.03 (3.50)- 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) . Insert rods to shutdown the reactor [0.83'and investigate th problem.IO.23 (1 0) Adjustment of turbine loa (0 5)

REFERENCE-Sochrook Dropped Rod Procedure 0S1210.05 ANSWER 4 04 (3.50)

!

l NO -Steam supply to TDEFW pump (NS-V128)not close (0.5) 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) At least one high head pump running (CCP or SI)

- RCS subcooling-less than 30- (1 0)

~ PRZR level cannot be maintained >5% (30% adverse containment)

- RCS subcooling less than 30- (1.0)

REFERENCE Scobrook E-3 SGTR procedure OS1330 p. 2-4 and Action Summar ANSWER 4.05 (2.50) Rod bottom lights Reactor trlP and bypass breakers open Neutron flux decreasin [0 5 eachT (1 5)

l l b.- Manually insert control rods Emergency Borat [0.5 each) (1.0)

l g

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44__IEROCEDURES. - NORMALa. ABNDRMAL. EMERGEHC1_Agg PAGE 29 RADIDLOGICAL CDMTROL ANSWERS - -SEA 8 ROOK 1 -85/03/18-ISAKSEN, REFERENCE

Scobrook FR-5.1 p. 2,3 ,

i ANSWER 4.06 (2.50) SCR0 - RSS panel (B) operator CR0 - RSS panel (A) operator (1.0) By opening.a PRZR POR (0 5)

co- PRZR level increases or oscillation (0 5)

, True (0.5)

REFERENC Scobrook 0S1200.02, p. 4s 5 and 19 ANSWER 4.07 (1.50)

l Notify the SS or US (1.0) SS or US (0.5)

i f REFERENCE l Soobrook Standing Operating Order No.84-001

!

ANSWER 4.08 (2 00) 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

a o

T_ O ta__hgGCEDURES - _HDRBALg_ARBDR5&La_E5EggENCY ARQ PAGE 30 RADIDLDEICAL_CDSIRDL ANSWERS -- SEABROOK 1 -85/03/18-ISAKSENs 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

,

-Accumu l ated exposure on recor d (NRC-4). [0.753 (2.25)

REFERENCE 10 CFR 20

,

l l

l

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nem&mem9 y ew,-tywm --w ~

F

. $ e - *

Y .h ': p a b U. S. NUCLEAR REGULA10RY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SEABROOK 1

_________________________

REACTOR TYPE: PWR-WEC4

_________________________

DATE ADMINISTERED: 85/03/18

_________________________

EXAMINER: RUSCITT0r APPLICANT: _ _ __ ____

____ ___ ________ ____ '

Una ceparate paper for the answers. Write answers on one side onl Stcple question sheet on top of the answer sheet 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 start .

g % OF iATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY p______ ______ ___________

________ ___________________________________ .

25.00 25.00 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS s5.00 25.00 l PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION 5.00 25.00 PROCEDURES - NORMAL, ABNORMALr

' EMERGENCY AND RADIOLOGICAL CONTROL 5 25 0

_I_00 ___ ___1__0 ___________ ________ ADMINISTRATIVE PROCEDURESr CONDITIONS, AND LIMITATIONS 00.00 100.00 TOTALS FINAL GRADE _________________%

11-work done on this examination is my own. I have neither

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, 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 canno 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 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 Scale formation increases on steam senerator tubes. [0.753 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

. Why is the time interval between reactivity changes important with-re3ard to constructing a 1/M plot? Co.753

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QUESTION 5.06 (3.00)

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Aftor operation at 100% power for several weeks near the end of cycler it in decided to reduce power to 75% using rods only.

! 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 0.0111 hours 6.613757e-5 weeks 1.522e-5 months assumin3 no change in baron concentration?

, Explain what causes these reactivity chan3e C1.5 each]

QUESTION 5.07 (3.00)

List the four parameters affecting Departure from Nucleate Boiling (DNB) and

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state whether the probability of approachin3 DNB is increased or decreased as these parameter values increas QUESTION 5.08 (1.50)

Why is there a Technical Specification limit on Shutdown Margin?

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QUESTION 5.09 (3.00)

l f-Answor the following concerning emergency boration j Explain the response of reactor power and TAVE after 2 minutes of i

emergency boration at 100% power. Assume rod control is in manua Explain the response of reactor power and TAVE after 2 minutes of emergency boration at 10 - 8 AMPS and no load TAV [1.5 each3 L

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~ QUESTION 5.10- (2.50)

. Ancwor~the following concerning fission product poisons: State the production and removal mechanisms for Xe-135 and Sm-149 in i the reactor core.[1 03 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 State the reactivity value for 100% equilibrium xenon concentration and the time to reach this condition from startup with a clean

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reactor core.CO.753

i QUESTION 5.11 (1.50) -

A variable speed centrifugal pump is operating at 1/4 rated speed in a clocod system with the following parameter What are the new values of those parameters when speed is increased to rated speed? Power = 500 KW

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b' . Flow = 900 GPM Head = 70 PSIG GUESTION 5.12 (2.50) Why does reactor coolant pump amperage increase during a plant cooldown? CO.53 l 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 Why is vessel Delta T always less than core Delta T? C1.03 i

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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 train Explain the effects on the automatic actions / interlocks associated with the RHR isolation valves. C1. 03, 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 - Why does the C-7 load rejection armins si 3nal lock in? C1.03 Why must the Steam Dump Mode Selector Switch be reset upon completion of.a large load rejection? C1.03 How do you know if the C-7 arming signal is actuated? C1.03

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-QUESTION 6.03 (4.00)

Answer the following concerning the Pressurizer Pressure Control System.

1 Acoune that the plant is at 100% powers List the indications in the Main Control Room if the controlling

channel (455) of pressurizer pressure fails high. Do not include alarms. C2.03

! What COULD cause a reactor trip on the above failure? C0.753 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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QUESTION 6.04 (4 00)

Anower the following concerning a main steam line break 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 What specific plant desi3n features were incorporated to limit the severity of a main steam line break? C0.53 Following a main steam line break inside containment, which si3 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 independentl Assume no operator action and that all control systems oro in automatic.

, Rod insertion limit.setpoin Char 3i ng flow (initially). Control rod bank position.

r Steam dump control syste C0.75 each]

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QUESTION 6.06 (3.00)

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i Whore'are the source range, intermediate range and power range detectors lecoted vertically with respect to the core? Why?

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QUESTION 6.07 (1.00)

Hew could Vital Instrument Panel PP-1A be supplied if UPS 1A was not cvoilable?

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'GUESTION 6.08 (3.00)

Ancwor the following concerning the Emergency Feedwater(EFW) System What-automatic action occurs at 450 GPM in any EFW system feed lin Why? 'Under what plant condition is this in effect? C1.03 How are the'EFW pumps protected against runout? C0.753 How would an operator know if local control of either EFW pump has been taken? E0.53

" 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 Syste Following a LOCA inside containment, what are three major sources of hydrogen? C1.03 At what level does hydrogen in the containment become a problem?

Explain. C0.53 How can hydrogen be removed from the' containment? CO.53 i

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QUESTION 7.01 (3.00)

During 80% power operation, with rod control in automatier a Power Range NI channel fails lo 'What immediate actions, if any must be taken? Explain. C1.53 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 trippe 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 i ng the effluent to the VCT?

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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 How will the following parameters be controlled while cooling down to 450 DEG F and 1000 PSIG?

(1) Steam pressure

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(2) Steam Senerator level

(3) Pressurizer pressure (4) Pressurizer level E0.6 each3 How will the plant be depressurized to 1000 PSIG? E0.63 o

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QUESTION 7.05 (3.50)

l Ancwer-the following concerning E-Or Reactor Trip or Safety Injection

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o. .What are.the five conditions that require SI initiation? (Include setpoints) C2.53 - The Main Turbine has not tripped and you attempt a manual trip as i

directed, with no respons What additional action are you required to take in order to shutdown the turbine? C1.03 iQUESTION 7.06 (2.00)

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Fill in the following precautions and limitations from various operating

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' The RHR System shall be isolated from the Reactor Coolant System (RCS)

before RCS exceeds ________ DEG F or ________ PSIG. [0.53 I 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 RCP seal injection water temperature should not exceed _______ DEG C0.25] At least one RCP should oe in operation with Tave greater than

________ DEG F. Co.25] The boron concentration in the pressurizer should be maintained within ________ PPM of the RCS loop concentration. C0.253 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 [0.25]

QUESTION. 7.07 (3.00)

Answer the following concerning intial core loading What two specific conditions would require an immediate increase in boron concentration? C1.03 What are four conditions that would require the suspension of core

' loading? (In addition to the above two) C2.03

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GUESTION 7.08 (3.50) Refueling Technical Specifications require that one RHR loop be operating during refuelin3 What are two reasons for this requirement?.C1.53 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 refuelin3? [1.03 (

GUESTION 7.09 (3.00) In addition to alarms, last four indications of a dropped contro rod? [1.03 N' hat are the immediate operator actions required for a single dropped

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rod? C1.03 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)

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Anowor the following concerning radiation detection and effects Consider two point gamma sources, each with 1 curie strengt Source A gamma energy is 2 MEV and Source B gamma energy is 1 NE If readings were taken at the same distance from each unshielded i source with a Geiger Mueller(GM) type meter, how would the readings j compare? Briefly explain. C1.03 L

i If a worker was exposed to a 1 RAD /HR neutron radiation field, would 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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.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 What immediate actions must be taken in accordance with the Technical Specifications? [1.03

~ 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/H Maintenance reports that Residual Heat Removal (RHR) Pump 8B repairs will not be completed for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months but that RHR Pump 8A is operabler and the Technical Specifications Action Statement allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to repair an inoperable pump in Mode 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 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ' What action, if any must be taken? C0.753 What action, if any must be taken if power were 83%? Why? E0.753 What action, if any must be taken if you exceed 60 penalty minutes in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ? 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 hour At 1230 tha Maintenance Supervisor states that repairs will take an additional hour ,

or c 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 i

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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

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. Fcbruary 1 The Operations Supervisor states _that it was still within

tho extension time allowed by the Technical Specifications. Is he correct?

- Explain your answe ,

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 ou Is that PCCW train

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cporable? Explain y~our answer.

QUESTION 8.07 (2.50)

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Annwer_the following concerning los reviews by the Unit Shift Supervisor! What loss are reviewed? C2.03 How often are they reviewed? C0.53 i

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gGUESTION 8.08 (2.00) ,

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'. What is the primary means of communication between the control room an'd occh of the following or3anizations during an emersency? NRC Mass /NH State Police Near-Tern Response Personnel j Station Security C0.5 each3 l

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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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QUESTION 8.11 (2.50)

Ancwor the following concerning Radiation Work Permits (RWP):

o. Under what conditions may the Shift Superintendent authorize entry into an area normally requiring a RWP without actually being provided a RWP prior to entry? C1.03 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 limit What is the period of time you are allowed to operate if the actual chemisty concentrations are between these two limits? C0.53 RCS specific activity limits refer to the term Dose Equivalent I-13 ; What is Dose Equivalent I-131? C1.03

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ANSWERS -- SEABROOK 1 -85/03/18-RUSCITT0r ;

' ANSWER 5.01 (2.00) All control rods in a single group are_ moved together with no single rod in the group differing by more than 12 steps from group positio Control rod groups are sequenced with proper bank overla Control rod insertion limits are observe Axial power distribution is maintained within specified limits.

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REFERENCE WNTC Rx Core Control, 8-32

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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

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'tckoo place in the secondary, heat is added with no change in temperature.

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REFERENCE HO-THF-3

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ANSWER 5.03 ( .50) -

, REFERENCE Stoco Tables HO-THF-2

[ ANSWER 5.04 (1.50) Scale reduces heat transfer ability, therefore U decreases and with constant powere Delta T must so up to compensate. CO.753 I 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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ANSWERS -- SEABROOK 1 -85/03/18-RUSCITTO, ANSWER 5.05 (2.50) ) 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 ( The approach to zero is easier to see than the approach to infinit C0 753 As Keff approaches one, it takes lonSer for count rate to stabilize, because more generations are required to level count rat CO.753 REFERENCE * HO-RTR-54 HO-RTR-54 - CAF for answer ANSWER 5.06 (3.00) Rods will need to be withdrawn for about 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months C0.753 and then inserted for th,e next 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months . CO.753 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 0.0111 hours 6.613757e-5 weeks 1.522e-5 months . C0.253 REFERENCE HO-RTR-109, 110 ANSWER 5.07 (3.00) Reactor Power Increases Reactor Coolant Temperature Increases

. Reactor Coolant Flow Decreases Primary Pressure ,

Decreases C0.75 each3 REFERENCE HO-THF-3 e

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P ANSWERS -- SEABROOK 1 -85/03i18-RUSCITT0r .

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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, ; ANSWER 5.09 (3.00) 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 a level sli3 htly lower than or the same as initial power.CO.53 (Possible low pressure tripe LO-LO TAVE implies low pressure) 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) (1) Xe-135 production directly from fission and the decay of iodine C0.253 and removal by decay and burnout. C0 253

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(2) Sa-149 production from the decay of promethium C0.253 and removal l by burnout. C0.253 pcm at 7-10 hours C0.753 t pcm at 40-60 hours C0.753

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REFERENCE 1 HO-RTR-104/116

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ANSWERS --.SEABROOK 1 -85/03/18-RUSCITT0r ANSWER 5.11 (1.50)

3 3 P2 = P1(N2/N1) = 500 X 4 = 32'MW i Q2 = G1(N2/N1) = 900 X 4 = 3600 GPM 2 2 H2 = H1(N2/N1) = 70 X 4 = 1120 PSIG REFERENCE 3 -HO-THF-4 ANSWER 5.12 (2.50) As temperature decreasest. the density of the fluid increase More

. power is required to pump the heavier fluid.C0.53 As temperature decreases, both steam density and steam velocity decrease.CO.53 The steam dump valves must be opened.further to

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maintain a constant mass flow rate.CO.53 .

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

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ANSWERS -- SEABROOK 1 -85/03/18-RUSCITT0r ANSWER 6.01 (2.00)

. Train B provides interlock and automatic isolation for the upstream valves of both train 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

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is made that LTOP is armed)3-REFERENCE HO-RHRS-25/26 005000K1.09 RO .3 6 SRO K4.01 RO SRO ,

005000A2.01 RO SRO ANSWER 6.02 (3.00) Prevents Steam Dump Valves from shutting once impulse pressure stops decreasins. C1.03 The dump will remain armed and a failure low of the other impulse

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channel (PT-505) or a hi3 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

, Control Board. C1.03 REFERENCE HO-SD-39 HO-SD-39,40 HO-SD-40

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ANSWERS -- SEABROOK 1 -85/03/18-RUSCITT0r .

ANSWER 6.03 (4.00) *

4 (1) Spray valves open. [0.673 (2) PORV 456A opens until pressure channel indicates 2185 PSIG, C0.673 then PORV 456A closes. CO.673 The rapid pressure transient could cause Low Pressure Trip j2/4 at 1945 PSIG) or OT Delta T (2/4). [0.753

_

ggd gM//gry/gg/ pfyg4 (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) (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 Flow restrictors in the S/G outlet nor:les (and BIT) [0.53 (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 be ba) Absson U J/0w NaA3/6014$

be ora

,

I l

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_

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.

9 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 20

______________________________________________________

' ANSWERS'-- SEABROOK 1 -85/03/18-RUSCITT0r .

ANSWER 6.05 (3.00) gg k, Increases due to auctioneeredgDelta T input. (Tave input set to zero) Increases to raise pressurizer level to 100% program, because of the

, higher temperature inpu Rods move in, because of the Auctioneered Tave/ Tref mismatch.

!

' No effeetr the demand signal is present (Tave/ Tref) but-there is not an armin3 SiSnal presen [0.75 each3

,

REFERENCE Instrument Failure Reference Manual pp. Al-A5

.

ANSWER 6.06 (3.00)

, 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

'

- 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

.

. .

..-;.*,.

. '. 4 PLANT SYSTEMS DESIGNr C NTROL, AND INSTRUMENTATION PAGE 21

______________________________________________________

ANSWERS'-- SEABROOK 1 -85/03/18-RUSCITT0r ANSWER 6.08 (3.00) 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 The flow venturi in each line limits flow to 750 GPM. C0.753 Alare on the MCB. C0.53 Noner all automatic initiations are blocked in local operation.CO.753 REFERENCE HO-EFW-16/25,38 ANSWER 6.09 (2.00) (1) Radiolysis of water (2) Zirconium / Water reactions (3) Coolant H2 inventory (4) . NaOH/Zine/ Aluminum reactions CO.33 eache 3 req.] Hydrogen is explosive in concentrations above 4%. [0.53 (1) H2 recombiners (2) Compressed air puse and vent C0.25 each3 REFERENCE

"*~****

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.

f LPROCEDURES-NORMAL, ABNORMAL,EMNRGENCY-AND PAGE 22

--- RA5iBE55iEAE

____________________

55ETEBE------------------------

ANSWERS -- SEABROOK 1 , -85/03/18-RUSCITTO, .

~r ANSWER- 7.01 (3.00)

-

o.- No immediate actions are required since all control funetions except '

rod stop are 2/4 coincide 6ce.'C1.53 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 ANSWER 7.03 (1.00)

LTo prevent reactor power variations due to changes in boron concentratio REFERENCE 0S1002404, page 2 -

ANSWER 7.04' (3.00)- < (1) Steam pressure will.be maintained (less than 1000 PSIG) with the >

atmospheric dump valve (2) Stesa generator level will be maintained (45 - 95%) with the emergency feed syste (3) Pressurizer pressure will be. maintained using heater (4) Pressuizer level will be. controlled by adjusting the cooldown rat C0.4 each]

b.' The plant is depressurized using the pressurizer PORV. [0.63 ,

' REFERENCE 081200.02, pp.2-19

.

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.

, *. . . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 23

~~~~R A5iBE55 5 E

____________________

55aTR5E------------------------

ANSWERS -- SEABROOK 1 -85/03/18-RUSCITT0r ANSWER 7.05 (3.50) (1) Pressurizer pressure less than 1850 PSIG and decreasin (2) Containment pressure greater than 4.0 PSIG.

'

(3) Steanline pressure less than 585 PSIG and decreasin (4) Reactor Coolant System subcoolin3 less than 30 DEG (5) Pressurizer level less than 5% and decreasin [0.5 each] ( 1 ) Close the MSIVs (2) Open the Generator Breaker CO.5 each3 REFERENCE F-0, pp.2,3 ANSWER 7.06 (2.00) DEG F/425.PSIG E0.53 PSID/15 PSIG C0.53 DEG F E0.25] . DEG P C0.253

' PPM [0.253 DEG F CO.253 REFERENCE OS1000.01 pp. 3,4 -

051001 05 l 1 i

- ___-. ___

_ . ." i-

.

> *:

. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 24

~~~~ ~

RA6E L6656AL 66 TR6L'~~~~~~~~~~~~~~~~~~~~~~~

____________________

ANSWERS -- SEABROOK 1 -85/03/18-RUSCITT0r .

ANSWER 7.07 (3.00) (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 (1) An unexplained decrease in the boron concentration greater than i

20 PPM from nomina (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 assemblie (3) Audible count rate in the containment or control room is los (4) Loss of communications between the control room and the refueling statio (5) The containment evacuation alarm is sounded.

f (6) Mechanical damage to a fuel assembly occur (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 assemblie l (9) An unanticipated increase in count rates by a factor of two L

l occurs on all channels during any step after loading of the eight initial assemblie [0 5 eache 4 required 3 REFERENCE 1-ST-4 pp.. 7-9 l

lANSWER 7.08 (3.50) (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 Ensure that a single failure will not result in a loss of heat removal capability (with a diminished heat sink). E1.03 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'

!

. .. . . - - .- - -.

'

-e t ..

t

, .z PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 25

~~~~EI656E66565E~665TR E~~~~~~~~~~~~~~~~~~~~~~~~

____________________

' ANSWERS -- SEABROOK 1 -85/03/18-RUSCITT0r ANSWER 7.09 (3.00) '(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 Place rod control in manual CO.53, and reduce turbine load to maintain Tave/ Tref. C0.5] 46 3 Power to be reduced by 30% of rated thermal power CO.53 within 30 minutes. C0.53 REFERENCE' A;//! 8CCf M / 8 O S1 10 5 p

'guyb".

. ANSWER 7.10 (1.50) 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) Greater-than. CO.53 REFERENCE CAF

,,

. _- . -

, . . . , - .

,

,. . . ,

.

,

8. . ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 26

.__________________________________________________________

ANSWERS -- SEABROOK 1 -85/03/18-RUSCITT0rD.

.

4 ' ANSWER 8.01 (3.00)

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 This limitation is required to ensure that: '

(1) MTC is within its analyzed temperature rang (2)' Protective instrumention is within its analyzed temperature rang (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 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

' 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . CO.753 REFERENCE j .RN 1740, ' TSr p.3/4 2-1 i

. . - . . ~ . . - , , _ , _ _ , , - _ . . , _ . - _ _ _ . . _ _ _ . ,.-_,_...--_...___,__.-,_.______,___,.u._-..-_,,,m., . . . _ . , - . , _ ~ _ . , - , ,

. . _ . . _ -. .~

_

.

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.

~ ADMINISTRATIVE PROCEDURESr C AND LIMITATIONS PAGE 27

!':_________________________________ONDITIONSr _________________________

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 violatio 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 exceede 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) (1) Tagging Order Log (2) Temporary Modifications Los -

(3) Station Loss (4) USS Journal CO.5 each]

! Once per shift E0.53 REFERENCE

,

AG 10.003'

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. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 28

__________________________________________________________

ANSWERS -- SEABROOK 1 -85/03/18-RUSCITT0r ANSWER 8.08 (2.00) ENS Hotline g/[NNh SN1M

[ Nuclear Alert System 6

' Radio Pasing System Telephone /PA System fN / * C)wC each3'

REFERENCE M Ecorsency Plan Figure 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) When critical and immediate action is required C0.53 for work of a short duration. C0.53 (1) The workers are given verbal instructions and precaution (2) Continuous health physics coverage is provide (3) The job is documented on a RWP after completion. CO.5 each]

.

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. . ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 29

__________________________________________________________

ANSWERS -- SEABROOK 1 -85/03/18-RUSCITT0r .

REFERENCE RP-1 0, p. 5' *

ANSWER- 8.12 (1.50) hours C0.53 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 presen [1.03-REFERENCE * TS 3/4 4-22 TS 1-2

.

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. . . ..

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.

TFST CRDSS REFERENCE' PAGE 1 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 2

'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

.

f = ma y = s/t Cycle efficiency = (.1stwork out)/(Energy in)

. w = ag s = V,t + 1/2 at '

2 ,

E = mc ,

KE = 1/2 av a = (Vf - V,)/t A = AN A = A,e^

PE = agn Vf = V, + at w = e/t 1 = an2/tifg = 0.593/t1/2 1/Zeff * C(tm)-(~%)]

- 2 t w=,g A= rD,

,

[(c l/2) * (Eb)J d = 931 as *

,y

,

a I = I,e ,

'

Q = scaat

-

g = UA47 I=I,e* -

'

w . up i = t,10-#!"- -

.

TYL = 1.3/s P = P 10sur(t) MVL = -0.693/s P = P ,e*/

SUR = 28.06/7 SG = 5/(1 - Kgf)

p Gx* 3/(I ~ "Wfx)

'

G j(1 - K gf3) = G 2(I ~ " #f2)

'

, SUR = 25e/s* + (s - e)T T = ( c/s) + [(s - s yI.] M = 1/(1 - Kgf) = CRj /G ,

,

T = a/(s - s) M = (1 - Kgf,)/(1 - Kgfi)

T = (s - s)/(T.) SDM=(}-Kgf)/Kgf a = (Kgf-1)/Kg f = 4Kgp/Kgf s= = 10 seconds I = 0.1 seconds-e=[(s*/(TKgf)] + [Tgf(1+IT)] /

Idli*Id

[ P = (r4V)/(3 x 1010) Id j 2 =2I4 2

l t = eN R/hr = (0.5 G )/d2 ( ,,g ,,)

l R/hr = 6 2/d2 (f,,g) ,

-

W4 tar Parameters Miscellaneous Conversions

1 gal. = 8.345 le I curie = 3.7 x 1010dps 1 gal. = 3.78 liters 1 kg = 2.21 lho 1 ft" = 7.48 ga np = 2.54 x 10 3 Stu/ar 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. H STU = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in.

!

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~ -

'

-

I .'.A :: SYSih %

( r. f

. '

.

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 .

.

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise reduce power to less than or equal to 5 percent of RATED THERMAL POWER within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months !

,

MODE 2 - With one main steam line isolation valve inoperable, subsequent and 3' -operation in MODES 2 or 3 may proceed provided:

!

The isolation valve is maintained close l

'

i The provisions of Specification 3.0.4 are not applicabl i

.

'

Otherwise, be in MOT STANDdY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MOT SHUTDOWN within the following 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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. .

.

-

. .

..

FIG. Seabrook - Units 1 & 2

.

.__ ._..----- -

v t e Inspection Report - Seabrook - 1985004
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1985 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 8(0) 13(0) 14(0) 15(0) 16(0) 17(0) 18(0) 19(0) 20(0) 21(0) 22(0) 23(0) 24(0) 25(0) 26(0) 27(0) 28(0) 29(0) 30(0) 31(0) 32(0) 33(0) 35(0) 99(0) Security Operator Exams Emergency Planning Other

IR 05000443/1985004

Text

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