Text

Exam Rept 50-324/OL-85-02 for Units 1 & 2 on 850923-26.Exam results:12 Candidates Passed & Three Failed.Requalification Program Evaluation Satisfactory
	ML20205F829
Person / Time
Site:	Brunswick
Issue date:	10/28/1985
From:	Munro J, Wilson B; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20205F819	List: BSEP-85-1758, Forwards Instructor Comments Re Questions & Answers for Reactor Operator & Senior Reactor Operator Requalification Exams Administered on 850623; ML20205F812; ML20205F829;
References
	50-324-OL-85-02, 50-324-OL-85-2, NUDOCS 8511130042
	Download: ML20205F829 (121)
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ENCLOSURE 1 EXAMINATION REPORT 324/0L-85-02 Facility Licensee: Carolina Power and light Company P. O. Box 1551 Raleigh, NC 27602 Facility Name: Brunswick Steam Electric Plant Facility Docket Nos. 50-324 and 50-325 Written and simulator examinations were administered at Brunswick Steam Electric Plant near Southport, North Carolina.

Chief Examiner:

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- to /2 4 / s s Date Signed

[ J./Nunro Approved by: k 8/pJ Bruce A. WilGr, Sectiorf Chief

/sDate2t!Jf

' Signed Summary:

Requalification Examinations on September 23-26, 1985.

Written and simulator examinations were administered to 15 licensed operators, 12 of whom passed.

The Requalification Program Evaluation for Brunswick Steam Electric Plant is satisfactory.

hI! DOCK N 4-0

REPORT DETAILS

1. Facility Employees Contacted:

J. Moyer, Training Manager G. Barnes, Senior Specialist - Operator Training

,C. Backus, Instructor (CP&L)

L. Dunlap, Instructor (RTS)

-*Provided Exit Briefing (Telephonically)

2. Senior Resident Inspector:

W. Ruland'

.3. Examiners:

J. Munro K. Brockman L. Wiens

Chief Examiner

4. Examination Review Meeting At the conclusion of the written examination, the examiners provided G. Barnes, with a copy of the written examination and answer key for review. The following comments were made by the facility' reviewers:

a. R0 Exam (1) Question #335 (1.09)

Facility Comment: No Tech Spec reference. for Part "C". 'Not an R0 question.

NRC Resolution: H2 and 02 ' limits are appropriatt for R0s as i delineated in the Mitigating Core Damage Lesson Plan. No change to. Exam..

.(2) Question #336 (1.10)

Facility Comment: Refer to 36 hrs - this is not in reference !

and may be a source for some confusion.

NRC Resolution: Time reference is covered by lesson plan, and is essential for identifying the dominant precursor. .

No change to Exam.

Enclosure 1 2 (3) Question #385 (2.07)

Facility Comment: Answer refers to full rated speed and flow as determined by the flow controller. The actual speed will be determined by reactor pressure, i.e., flow controls speed and varies with reactor pressure. No reactor conditions were stated.

NRC Resolution: Answer key assumes normal full power conditions. If other valid conditions are identified as assumptions, grading will reflect same. No change to Answer Key required.

(4) Question #391 (2.11 and 6.10)

Facility Comment: Only F004 closes on Hi outlet temp.

Reference is incorrect.

NRC Resolution: Answer Key modified to reflect only F004 closes; controlled reference DWG 9527-50056 verifies change.

(5) Question #283 (3.02 and 6.14)

Facility Comment: (a) 50% is optionai; actual is 44-49 percent.

Figure #283 refers to Hatch terminology -

answer for BSEp different than figure.

Alternate answer is "wherever the master controller ~is set." You have to assume master controller is at low limit to get answer key answer.

(b) The question and drawing are not clear.

The question-could be interpreted as one or the other failing. If..the tachometer failed low, the recirc pump will trip on exciter undervoltage due to the 70v/Hz programming voltage low relay K8.

Initially, pump attempts to speed up, then trips - should also be correct as an answer.

Contact Y1 opening and tachometer output fail low are two separate failures which would cause two different responses.

NRC Resolution: (a) Question asks for controlling component, not where it controls. No change to Answer Key required. ;

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, Enclosure 1 3 (b) The potential for confusion is acknowledged.

The alternate response is accurate for a total tachometer failure. Answer Key modified to accept either answer.

(6) Question #341 (4.03 and 7.03)

Facility Comment: Line-up given in procedure not an immediate action which would be reason for memorization.

NRC Resolution: Operators are required to be familiar with the methods and actions of "non-emergency" (no immediate action) procedures. This question investigates general familiarity only. No change to Exam.

(7) Question #345 (4.07)

Facility Comment: Per A0P-15, both loops of RHR can also be in Suppression Pool Cooling (preferred method I loop in SP cooling 1 injecting) can use Core i Spray.

. NRC Resolution: Answer Key modified to allow both RHR loops to be in Suppression Pool Cooling.

b. SRO Exam (1) Question #343 (7.06)

Facility Comment: These steps of the A0P would be conducted using the AOP, not from memory.

NRC Resolution: See comments for Question #341.

(2) Questions #352 (8.06), #356 (8.08), and #362 (8.11)

Facility Comment: You are requiring the SRO to memorize Tech Specs - this is not the case at BSEP. We are required to be able to use Tech Specs to determine requirements for various plant conditions. (Reference ES402, pg. 4)

NRC Resolution: Senior operators are required to memorize Limiting Conditions (LCOs), Definitions, and 1-Hour Action Statements. The actions requested by Question #352 were not 1-hour actions.

Questions #356 and #362 requested LCOs. Question

352 deleted. No change to Questions #356 nor

362.

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Enclosure 1 4

c. Generic Comments The facility also made numerous generic comments which did not recommend resolutions, nor address technical deficiencies. These comments were reviewed and found to be non-applicable to the exam and answer key.

No changes were made.

5. Exit Meeting After site visit, the examiners met telephonically with a representative of the plant staff to discuss the results of the examination. Those individuals who clearly passed the operating examination were identified.

There was one generic weakness (greater than 75 percent of candidates giving incorrect answers to one examination topic) noted during the operating examination -

normal procedure for Reactor Feedwater Pump Startup.

The cooperation given to the examiners was noted and appreciated.

Tne licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.

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, ENCLOSURE 3 l

l l U. S. NUCLEAR REGULATORY COMMISSION l REACTOR OPERATOR LICENSE EXAMINATION i

l FACILITY: BRUNSWICK 1R2 REACTOR TYPE: DWR-GE4 l

i DATE ADMINISTERED: 05/09/23 EXAMINER: K E BROCKhAN APPLICANT: _________________________

INSTRUCTIONS TO APPLICANT:

Use separate paper-for the answers. Write answers on one side only.

Stcple question sheet on top of the answer sheets. Points for each quostion are indicated in parentheses after the question. The passing Stade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4 ) hours after l tho examination starts.

% OF f CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 17 0 1.

___1.0.__ _I'4 29.I . ________... _____ .. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 17 00 24.29

..______ ... __ __________. ._____ _ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 18.00

___. ___ _I'_5.71 ___ ___________ ________ 3. INSTRUMENTS AND CONTROLS 10.00 25.71 4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 70.00 100.00 TOTALS

/INAL GRADE _.....___________%

All work done en this examination is my own. I have neither given not-received aid.

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LiCAAII ~SIGUEIVEE S

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2

~~~~iUEEE66YUd EC5~~HEdT ~TRd SEEE~5U6~ELuf6~EL6U GUESTION 1.01 (1.00)

A ' Periodic NSS Core Performance Log' (Figure 0 044) is attached for reference. Which statement is most accurately reflected by this printout?

a. Maximum LHCR(s) in the core is 12.00 Kw/ft.

b. Hanimum LHCR(s) in the core is 7.69 Kw/ft.

c. Maximum LHCR(s) in the core is 13.40 Kw/ft.

d. Maximum LHCR(s) in the core is 9.19 Kw/ft.

QUESTION 1.02 (1.00)

The fission process in a commercial reactor requires the neutrons that are ' born' by fission to be 'thermalized.' The interaction in the reactor core which is most efficient in thermali:ing neutrons for fission occurs with the ...(CHOOSE ONE)

a. ... OXYGEN atoms in the water molecules

b. ... CORON atoms in the control rods

c. ... ZIRCONIUM atoms in the fuel claddin3

d. ... HYDROGEN atoms in the water molecules (ass ** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3

~~~~iEERE035E E5C5~~5Ehi~TRIE5E5s~h 6~ELU56~IL65 GUESTION 1 03 (2 00)

Attached Figure 4213 shows a basic closed loop fluid system with its head vs. flow plot. The two pumps are identical, variable speed, radial, centrifugal pumps. Pumps 1 & 2 are initially operating at one-half speed to supply flow to ccmponents 1 A 7, as shown in BOLD LINES.

a. DEFINE (name) Point X (0.5)

b. With the system operating as shown in the diagram, if component 2 were throttled shut from its initial position, would the system flow INCREASE, DECREASE, or REMAIN THE SAME? (0.5)

c. Component 2 is valved out of service, thereby reducing the heat load on the system. How would power consumption be minimi ed? (0.5)

(1) Reduce BOTH pump's speed to one-fourth speed.

(2) Stop and Isolate ONE pump.

d. Which Pump Curve - A, B,C, or D - most accurately shows ONE PUMP operating at half-speed to supply flow to the INITIAL system lineup? (0.5)

GUESTION 1 04 (1.00)

Attached Figure 4 214 illustrates four potential temperaturo profiles (relationships) for a COUNTER FLOW heat enchanger. Which of these graphs most accurately shows the temperature profile (relationship)?

NOTE: THE TEMPERATURE PROFILES FOR EACH RESPONSE ARE PLOTTED ON THE SAME TEMPERATURE SCALE (i.e., The TOP line is the HOT flow and the BOTTON line is the COOLING WATER flow).

QUESTION 1.05 (1 00)

STATE for which condition the reactivity coefficient contribution would be NORE NEGATIVE. EXPLAIN your choice.

Doppler coefficient with a 25% Void Fraction in the core,

~0R-Doppler coefficient with a 75% Void Fraction in the core.

(sses* CATEGORY 01 CONTINUED ON NEXT PACE masas)

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4 i

.__.____ .;.t________________ ...--____ ________ I THERMDeVNAMICS, HEAT TRANSFER AND FLUID FLOW I

.> l OUESTION 1 06 (1 00)

. Attached Figure 4 219 shows a POWER HISTORY and four possible

,' XENON traces (Reactivity.vs Time). SELECT the most accurate curve for displayin3 the enpected XENON transient.

DUESTION 1.07 .(1 00)

' Attached Figure # 220 shows s POWER HI"(CRY arid f our possible SAMACIUM traces (Reactivity vs Time). SELECT the most accurate i curve for dier. laying the expvCted SAMARIUM transient.

, ( w i "00ESTION' 1.08 (1 00)

Attached Figure 4 285 is a simplified sketch of a' SJAE. For ;

each of the pressure relationships given belov. STATE whether the pregsu m listed first is GREATER THAN, CESS THAN, or EDUAL TO the ' pressure listed second.

NOTE: THE LOCATION OF THESE PRESSURES CORRESPOND TO THE POINTS INDICATED IN THE FIGURE.

a. P(1) as to P(3) (0.25)

b. P(1) as to P(5) *

(0.25)

c. P(2) as to.P(4) v (0.25) ,

d. P(2) as to,P(3)* *

(0.25)

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1 (smass CATEGORY 01 CONTINUED ON NEXT PAGE massa) 0 4 7 8

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5

-~~ istis55isisiCB- siAi iiissFER Es5 FEDi5 FEBs QUESTION 1.09 (1 50)

The Containment atmosphere is innerted with Nitrogen to limit the Oxygen content.

o. STATE the principal source of Oxygen ir. Lne containment following a LOCA. (0.5)

b. STATE the principal source of Hydrogen in the containment fo110 win 3 a LOCA. (0.5)

c. LIST the maximum permissible concentrations of Oxygen and Hydrogen (in volume percent) allowed in the containment following a LOCA. (0.5)

DUESTION 1.10 (1.00)

Your reactor has been shutdown for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The intrinsic neutron source which is the NAJOR CONTRIBUTOR to the background neutron population is ... (CHOOSE ONE)

a. ... Deuterium-Gamma reaction (Photo-neutron).

b. ... Spontaneous fission of Curium-242/244.

c. ... Alpha-Oxy 3en reaction.

d. ... Spontaneous fission of Uranium-238 OUESTION 1.11 (1.00)

There are two (2) criteria which determine the end-of-life for o control rod. FILL IN THE BLANKS as appropriate.

When the Control Rod has lost ___(a) ___% of its original nes-otive reactivity value due to the burn-up of the ___(b) .__.

When the gas pressure due to the ___(c) .__ and the ___(d) .__

formation reaches 6565 psig.

(*m*** CATEGORY 01 CONTINUED ON NEXT PAGE s****)

10 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

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. QUESTION 1.12 (1.00)

In a soberitical reactor, Keff'is increased from .880 to .965.

Which of the following is the amount of reactivity that was added to the core?

l a. .085 delta k/k

b. .100 delta k/k

c. .125 delta k / k I d. .220 delta k/k GUESTION 1.13 (1.00)

The tesctor. trips from full power, equilibrium XENON conditions. Twenty-four hours later the reactor is brought critical and power level is main-tained on range 5 of the IRMs for several houre. Which of the following statements is CORRECT concernin3 control rod motion?

a. Rods will have to be withdrawn due to XENON build-in.

b. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of XENON burnout.

c. Rods will have to be inserted since XENON will closely follows its normal decay rate.

d. Rods will approximately remain as is as the XENON estab-12shes its equilibrium value for this power level.

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(xxxrr CATEGORY 01 CONTINUED ON NEXT PAGE **xxx) s E

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 7 THERH0 DYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.14 (2.50)

Attached Figure 4 400 represents a transient that occur at a BWR.

Given (1) The Reactor is initially at 105% power

'(2) There is a COMPLETE LOSS of GENERATOR LOAD (3) There is 25% BYPASS capability (4) No operator actions occur EXPLAIN the cause(s) of the following recorder indications:

a. Variat' ion in NEUTRON FLUX ( GRAPH I) (Point A - Point B)

6. DECREASE in NEUTRON FLUX ( GRAPH I) (Point B)

c. Variation in VESSEL STEAM FLOW (GRAPH III) (Point C - Point D)

d. Variation in VESSEL STEAM FLOW (GRAPH III) (Point D - Point E)

e. Variation in VESSEL STEAM FLOW (GRAPH III) (Point E - Point F) i (xxxxx END OF CATEGORY 01 xxxxx) 1 a

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8

n-QUESTION 2.01 (1.00)

The plant is operating normally at power when the ' Rec' - ,, 4 Inner Seal _ Leakage and Seal Staging Flow' FS trips low -g +t LtCREASE in No. 1 Recire Pump seal pressure. Which of t ha. 4 s s i ,..>- f e .1 eres I would result in these indications?

a. Failure of No. 1 seal

b. Failure of No. 2 seal

c. Plugging of the No. 1 internal re cati ^J '*ifice

d. Plugging of the No. 2 internal r0 i -t:- , orifice 00ESTION 2.02 (1.00)

How would a loss of service air affect the operation of the Standby Liquid Control System (SBLC)?

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a. The SBLC tank level indication would be inoperable.

b. The SBLC tank air sparger would be inoperable.

c. The SBLC tank level indication and air sparger would be inoperable.

d. It would have NO impact since the instrument air system supplies all SBLC needs.

QUESTION 2.03 (1.00) l Which of the following axial location sequences correctly describe the axial locations of LPRMs in the core?

a. BAF - 'A' G +9' -

'B' G +27' -

'C' G +45' -

'D' G +63'

b. BAF - 'A' G +18' -

'B' G +54' -

'C' G +90' -

'D' G +126'

c. BAF - 'D' G +9' -

'C' G +27' -

"B' G +45' -

'A' G +63'

d. BAF - 'D' G +18' -

'C' G +54' -

B' G +90' -

'A' G +126'

(**xxx CATEGORY 02 CONTINUED ON NEXT SAGE *****)

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 QUESTION 2.04 (1.00)

Four 25% capacity centrifugal pumps per unit are used to DISCHARGE the circulating water effluent (vitimately) into the Atlantic Ocean.

These pumps ... (CHOOSE ONE)

a. ... will automatically trip if they lose their Ivbrication/

cooling. water supplied from the Service Water header.

b. ... are controlled by a supervisory system with the master sta' ion at Caswell Beach and the remote station in the Contr ol Room.

c. ...-are energized after the START command energizes the pump discharge open coil and an (50%) 'open" limit switch is activated.

d. ... are equipped with anti-reverse devices to limit high starting torques and currents during pump starts.

QUESTION 2.05 (2.00)

Answer the following with re3ard to the Vital Service Water System;

a. LIST three (3) components that receive their cooling water supply from the Vital Service Water header. (1.5)

b. EXPLAIN how corrosion and fouling are MINIMIZED in the Vital Service Water header. (0.5)

GUESTION 2.06 (2.00)

A LOCA and a Loss of Offsite Power (LOSP) have occurred simultaneously.

The Diesel. Generators have started and AUTOMATICALLY energized their respective emergency busses. The sequential loading relays will delay the automatic starting of emergency loads for 5, 10, 15, or 20 seconds to minimize starting surges on'the diesels. LIST the loads that will be energized at EACH of these time intervals.

NOTE: DO NOT ADDRESS EACH BUSi LIST'FOR A GENERIC BUS WITHOUT THE PARTICULAR DESIGNATOR FOR THE LOAD

(***** CATEGORY 02 CONTINUED ON NEXT PAGE axxxx)

< _ _ = _ - _ - -_.

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, 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 s

QUESTION 2.07 (1.00)

The HPCI system is in TEST operation with the Test Signal Generator Control Switch set to control HPCI speed at 2000 rpm. Select the statement that most accurately describes the actions (Manual -OR-Automatic) that would be necessary to place HPCI in ENERGENCY SERVICE follwing an AUTO initiation signal while in TEST.

NOTE: HPCI FLOW CONTROLLER IS IN AUTOMATIC

a. HPCI will AUTO align for injection at the preset speed of 2000 RPM.

b. HPCI will AUTO align for- injection at the FULL RATED speed aid flow as determined by the flow controller.

c. HPCI will AUTO align for injection but must be taken OUT OF TEST and placed in AUTO to actually inject into the vessel.

d. HPCI must be MANUALLY realigned for injection and placed in the AUTO mode to inject into the vessel.

QUESTION 2.08 (1.00)

Which of the following is the power supply for the HPCI Auxiliary Oil Pump?

a. 120 Vac (UPS)

b. 480 Vac

c. 125 Vdc

d. 250 Vdc QUESTION 2.09 (1.00)

There is a Check Valve and a Check Valve Bypass located in the discharge.line immediately downstream of each Core Spray Pump.

STATE the purpose of this Check Valve -AND- the Check Valve Bypass.

(xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 GUESTION 2.10 ( .50)

TRUE OR FALSE A train of the Standby Gas Treatment System (SBGTS) will AUTO START on receipt of a valid initiation signal, when the train's control switch is in the STANDBY position.

QUESTION 2.11 (1.50)

Each of the events listed below would cause the Reactor Water Cleanup Systes to ISOLATE. For each event, STATE whether the Inboard Isolation Valve (F001) AND/OR the Outboard Isolation Valve (F004) would automatically CLOSE.

c. A loss of cooling water to the NRHX, resulting in an NRHX RWCU outlet temperature exceeding 140 des F.

b. A MANUAL initiation of the Standby Liquid Contrcl Systee (SLC).

c. A loss of ALL AC power, concurrent with a LO-LO Reactor Water Level.

QUESTION 2.12 (2.00)

As the Instrument and Service Air header pressures DECREASE below normal, certain automatic actions and alarms occur. MATCH the pressures listed with the applicable automatic action or alarm.

, AIR HEADER PRESSURE ACTION / ALARM

a. 110 psig 1. ' INST AIR PRESS LOW' Alarm

b. 105 psis 2. Standby Rx Blds Air Compressors START

c. 100 psis 3. Interruptable Instr Air Hdt Isolation

d. 95 psis 4. ' SERVICE AIR PRESSURE LOW' Alarm

5. Scram Air Hdr Isolation

6. Service Air Hdr Isolation

7. A0G Air System transfers to Nitrogen

(***x* CATEGORY 02 CONTINUED ON NEXT PAGE *****)

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.- o y' l 2 6~ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 12 QUESTION 7. 13 (1.00)

The RED light on the RTGD that indicates an SRV is in the OPEN p@sition is activated by which of the following signals:

a. High SRV tailpipe temperature

b. High SRV tailpipe pressure

c. SRV Acoustic monitor

d. SRV Solenoid Valve energized QUESTION 2.14 (1.00)

Once ADS is AUTOMATICALLY initiated, the ADS valves will REMAIN OPEN until which of the following occur?

a. A Reactor Water Level HI Trip signal is received.

b. One of the ADS 120 second Timers are reset,

c. Reactor Pressure DECREASES to the LP ECCS injection permissive (<450 psig).

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d. Reactor Pressure DECREASES to approximately 50 psis above Primary Containment pressure.

(***** END OF CATEGORY 02 xxxxx) i f

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3. . INSTRUMENTS AND CONTROLS PAGE 13 OUESTION 3.01 (2.00)

Unit 2 is operating at 100% RTP when APRM 'A' fails upscale and results in a reactor half-scram. Utilizing the attached RPS trip logic diagrams (Figures $224 A thru F) DESCRIBE in c STEP-BY-STEP fashion (with regard to the opening / closing, ener-gizing/deenergizing of ALL applicable contacts and relays) how the APRM upscale. trip results in an actuation of the scram solenoid valves.

NOTE: IF-THE ATTACHED DIAGRAMS CAN NOT BE EASILY READ, ASSIGN THE CONTACTS / RELAYS, ETC NUMBERS AND REFER TO THEM IN YOUR ANSWER.

QUESTION 3.02 (2.00)

The plant is operating at 23% power and both Recirc Pump M/A Transfer Stations are in MANUAL and set for minimum speed.

The 'Recirc Flow B Limit' annunciator is CLEAR. For eat;. of the following instances, STATE how the speed of Recire Pump 'B' will change (i.e.r INCREASE, DECREASE, REMAIN THE SAME) and WHICH COMPONENT (S) of the control / positioning system is/are LIMITING.

NOTE: FIGURE 4 283 IS PROVIDED FOR REFERENCE

a. Recirc Pump 'B' M/A Transfer Station placed in ' AUTO' (1.0)

b. Tachometer output feedback signal fails low - Contact Y1 Opens (1.0) 1 (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE **xxx)

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3. INSTRUMENTS AND CONTROLS PAGE 14

- 0UESTION 3.03 (2.00)

The APRM scram function actually consis+= of two separate setpoints:

a. FILL IN THE BLANKS:

Flow Biased Scram -

__ (1) __ x w + _,, (2 ) __% (0.25)

Fixed Scram -

__ (3) __% (0.25)

b. LIST the specific location (s) of the s e r.s o r ( s ) which measure the variable 'w' . (0.5)

c. While operating at power, one MSIV fails shut resulting in a brief (~ 1 second) flux spike to 121% power. STATE which of the two scram setpoints mentioned above (one or both) should initiate a reactor scram. JUSTIFY your choice. (1.0)

(xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx) 7

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3. INSTRUMENTS AND CONTROLS PAGE 15 QUESTION 3.04 (1.00)

Unit 1 is operating at 100% RTP, with recite in Master Manual. An operator inadvertently DECREASES the ' Pressure Set

on EHC by 5 psis.

ASSUME: 1. No Further Operator Actions

2. All other EHC control settings are r.croal

3. Starting Parameters o TCV's - 100% Steam Flow Position o BPU's -

0% Steam Flow Position o Power - 100% Rated Thermal Power c Pressure - 1010 psig NOTE: FIGURE 4 374 IS ATTACHED FOR REFERENCE Which of the following most accurately describes both the INITIAL RESPONSE and FINAL STATUS of the different parameters and components.

a b c d 2

INITIAL RESPONSE o TCV's INO CHANGE ICLOSE (~83%) ICLOSE (~83%) i NO CHANGE o BPV's i NO CHANGE I OPEN (~17%) l NO CHANGE I OPEN (~10%)

o Power i DECREASE I NO CHANGE I INCREASE I DECREASE o Pressure i DECREASE i NO CHANGE I INCREASE I DECREASE I I I I FINAL STATUS I I I I I I l I

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o TCV's 10%(HSIV Shut)I ~ 83 % i ~100 % 1 100 %

o BPU's 10%(MSIV Shut)I ~ 17 % 1 0% 1 0%

o Power 10% (Rx Scram)! > 100 % i > 100 % 1 < 100 %

o Pressure IAs contro11edi >1010 psisi >1010 psigl <1010 psis Iby SRV's and i I I IHPCI/RCIC i i I (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx)

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3. INSTRUMENTS AND CONTROLS PAGE 16 QUESTION 3.05 (2.00)

For each of the following situations (i and ii) select the correct Feed-water Control System / plant response from the list (a through e) which follows. An answer may be used more than once, and NO operator actions are taken.

a. Reactor water level decreases and stabilizes at a lower level.

b. Reactor water level decreases and initiates a reactor scram.

c. Reactor water level increases and stabilizes at a higher level.

d. Reactor water level increases and initiates a turbine trip.

e. None of the above.

i. The plant is operating at 100% power, in 3-element control, when tha HFCI system inadvertently initiates and injects.

ii. The plant is operating at 100% power, in 3-element control, when one Feed Flow Detector FAILS DOWNSCALE.

QUESTION 3.06 (2.00)

Answer the following with regard to the RHR System

a. Reactor Pressure is 325 psis and DECREASING. STATE the permissive signal that is generated and STATE to where it is sent. (0.5)

b. A LPCI initiation signal is present and the operator takes a runnin3 RHR pump switch to the STOP position. STATE the indication that he will receive. (0.5)

c. LIST the operator actions which would be required to RESTART the pump secured in part (b). (1.0)

(xxxxx CATEGORY 03 CONTINUED-OH NEXT PAGE xxxxx) i I

3. INSTRUMENTS AND CONTROLS PAGE 17

~

QUESTION 3.07 (1.00)

The Mode Switch is in RUN and 11 of the 17 LPRM's assigned to APRM 'A' cre operable. Which of the following automatic actions will occur es a result of one of the operable 11 LPRM inputs FAILING DOWNSCALE.

ASSUME NO OPERATOR ACTIONS

a. A Rod Block

b. A Reactor Half-scram

c. A Rod Block and a Reactor Half-scram

d. None of the Above GUESTIDH 3.08 (2.00)

The reactor is critical at approximately 5 psis and the ' Press-uri:ation* phase of GP-02 is being performed. The Normal Control Range GEMAC LI's in the control room read the following ' approx-imate' values.

GEMAC A (N004 A) 187 GEMAC B (N004 B) 188 GEt,C C (N004 C)

187

a. The two Emergency System Range (Yarway) control room level indicators should read approximately ... (CHOOSE ONE) (1.0)

1. ... 150 *

2. ... 165 '

I

3. ... 188

4. ...>210

. b. The Shutdown Vessel Flooding Range control room level

. indicator should read approximately ... (CHOOSE'ONE) (1.0)

1. ... 150 '

2. ... 165 *

3. ... 188

4. ...>210 (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE mmmx*)

4

3. INSTRUMENTS AND CONTROLS PAGE 18 GUESTION 3.09 (1.00)

The Full-Core Display on the center panel has a BLUE scram light for each control rod. What is DIRECTLY indicated when this li3ht is illuminated?

a. That BOTH the Inlet and Outlet scram valves for that rod are OPEN.

b. That BOTH scram pilot air valves for that rod are deenergized.

c. That the beyond full-in overtravel reed switch for that rod is closed.

d. That the scram accumulator for that rod is depres-surized.

QUESTION 3.10 (1.50)

The reactor is in STARTUPi RSCS is enforcing with respect to rod covement. The Reactor Manual Control System (RMCS) ROD OUT PERM-ISSIVE LIGHT is extinguished. Assuming a valid rod selection, STATE, for each of the following operator actions, whether rod covement WILL or WILL NOT occur.

a. The CRD Control Switch is placed in the WITHDRAW position. (0.5)

b. The CRD Control Switch is placed in the WITHDRAW position -AND-the CRD Notch Override Switch is placed in the OVERRIDE position. (0.5)

c. Place the CRD Control Switch in the INSERT position. (0.5)

GUESTION 3.11 (1.00)

The reactor is in the STARTUP model_the SRM's are FULLY INSERTED and show a count rate of 60 cps. The IRM's are set to Range 1.

STATE whether the SRh RETRACT PERMISSIVE would allow the SRM's to be WITHDRAWN. JUSTIFY your response.

QUESTION 3.12 ( .50)

Reactor pressure is 900 psis and LPCI is running in response to a valid initiation signal. STATE the-approximate expected FLOW INDICATION on the pump discharge flow meter on the Control Room Panel.

(****x END OF CATEGORY 03 ****x) l o

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4. PROCEDURES - NORMAL, ABNORMAL, - EMERGENCY AND PAGE 19

~~~~Rd656L5Ei5dL C6sTR5L QUESTION 4.01 (1.00)

A plant startup is in progress and condenser vacuum i s being established in accordance with GP-02, ' Approach to Criticality and Pressurization of the Reactor'. What is the proper sequence '

for component / subsystem startups.

a. Steam Packing Exhauster, Steam Seal Header, Mechanical Vacuum Pump, Steam Jet Air Ejector.

b. Steam Seal Header, Steam Packing Exhauster, Hechanical Vacuum Pump, Steam det Air Ejector.

c. Mechanical Vacuum Pomp, Steam Packing Exhauster, Steam Seal Header, Steam det Air Ejector.

d. Steam Packing Exhauster, Mechanical Vacuum Pump, Steam Seal Header, Steam Jet Air Ejector.

QUESTION 4.02 (1.50)

OP-27, " Generator and Exciter System Operating Procedure", cautions 2' the operator to minimize turbine generator operation below 100 MHe.

a. EXPLAIN the basis for t5is operating limitation. (1.0)

6. _Per OP-27, 'the power factor shall never be allowed to become less than _______.* (0.5)

OUESTION 4.03 (1.50)

LEP-03, ' Alternate Bor.on Injection *, identifies the RWCU pumps / system as one of six that may-be selected and used to inject boron if.the SLC system is NOT AVAILABLE when baron injection is required. DESCRIBE the method / flow path which is established to accomplish this injection.

NOTE: ASSUME LEVEL IS BEING MAINTAINED > 112'

(***** CATEGORY 04 CONTINUED ON NEXT PAGE x*xxx) l l

'& --*ei g y y P- y +w+ r -y r -

y

. ~. ._ - - _ = _ . - - - -

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 20

~ ~~~~~~~~~~~~~~~~~~~~~~~~

~~~~E 65ELUE5Ehl EUNTEdL QUESTION 4.04 ( .50)

Given the fr.llowing conditions

Mode Switch -

Shutdown Temperature -

180 des F Pressure -

O psig

Level -

330 inches (R605 Indication)

RHR -

SDC Mode Head bolts to the RPV are DETENSIONED STATE the above described Operational Condition.

i QUESTION 4.05 (1.50)

When increasing Recirculation Pump speeds with both controllers in MANUAL, their speeds should NORMALLY be maintained within

___ (a) ___% (of each other). The speed differential is LIMITED to ___(b)___% when below 75% core flow and ___(c) ___% when above 75% core flow.

i QUESTION 4.06 (1.00)

A LOCA has occurred and a high temperature steam environment exists in the drywell. EXPLAIN why the drywell sprays must NOT be initiated in the ' Unsafe

region of attached Figure # 344 "Drywell Spray Initiation Pressure Limit'.

(xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE *****)

l l

l' -. _ . . _ ..

i l

l

. 1

d. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 21 l

~~~~RI656LU555EL'E6NTR L'~~~~~~~~~~~~~~~~~~~~~~~

QUESTION 4.07 (2.50)

A reactor Cooldown is in progress and NEITHER loop of RHR can be placed in the Shutdown Cooling Mode. In accordance with AOP-15.0, Alternate Shutdown Cooling has been established to remove decay heat and continue the cooldown. STATE the condition / status of the following components / parameters when operating in this mode of shut-down cooling.

a. MSIV's (0.5)

b. SRV's (0,5)

c. RHR Loops A & B (0.5)

d. Reactor Pressure (Compared to Suppression Chamber Pressure) (0.5)

e. Reactor Level (Provide numerical value or component reference) (0.5)

DUESTION 4.08 (1.00)

Per the BSEP Radiological Emergency Plan

a. STATE the NORMAL OSC location. (0.5)

b. STATE the NORMAL EOF location. (0.5)

GUESTION 4.09 (1.00)

When the Rod Horth Minimi:er is inoperable or bypassed, commencement of Control Rod movement for a Reactor Start-up is ... (CHOOSE ONE)

a. ... PERMITTED, providing that a second qualified member L of the plant technical staff verifies the rod sequence.

b. ... PERMITTED, until the reactor thermal power reaches the Low Power Set Point (LPSP).

c. ... PERMITTED, if approved by the SOS and the Manager-Operations, and appropriate operating limitations are noted in the Pre-Startup Checklist.

d. ... NOT PERMITTED.

(xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx)

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

t

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 22

~~~~ ~~~~~~~~~~~~~~~~~~~~~~~~

RI65ELEEEE L'EENTEEL QUESTION 4.10 (1.00)

Unit 1 is operating at power when CRD Pressure DECREASES to 'O' psis.

STATE the conditions, per ADP 02.1, ' Inability to Move Control Rods',

which would require a MANUAL SCRAM.

QUESTION 4.11 (1.00)

There are two End Path Manuals at BSEP. STATE the respective OPER-ATIONAL OCCURANCES which will put an operator into LACH EPH.

QUESTION 4.12 (1.00)

Level Detectors NO36/NO37 are not included in the E0P-01/UG caution (CAUTION 46) concerning high temperatures near the reference les vertical runs. EXPLAIN WHY these instruments are EXCEPTED from this caution and WHEN, if ever, these i nstruments would develop excessive inaccuracies.

QUESTION 4.13 (1.00)

Per DI-13, ' Locked Valve Identification and Locking *, STATE the proper method for CONFIRMING valve position.

a. Turn the valve band wheel in the OPEN direction; confirm the position by observing that the stem travels in the OPEN direction.

b. Turn the valve hand wheel in the CLOSED directioni confirm the position by valve position indicator and/or firm tightness.

c. Turn the valve hand wheel in the DESIRED POSITION direction; confirm the position by valve position indicator and/or firm tightness.

d. Turn the valve hand wheel in the DESIRED POSITION directioni confirm the alve position by observing that the stem travels in the DESIRED direction.

(xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx) n -m , - , y - - , , ,, ,,- -

-w--- - - -

i t

- 4. PROCEDURES - NORMAL, ABNORMAL, EhiRGENCY AND PAGE 23

~~~~ ~ ~~~~~~~~~~~~~~~~~~~~~~~~

RI655LUE5EIL 56EiE6L QUESTION 4.14 (1.00)

_____________ clearance. (Definition) -

Issued to a foreman or lead man who is responsible and cognizant of two or more jobs being performed within the boundary of this clearance.

Fill.in the blank with one of the following.

a. Local

b. Individual

c. Multiple

d. Haster I

GUESTION 4.15 (1.50)

You enter an area posted with the following sign:

CAUTION HIGH RADIATION AREA LIST the three-(3) methods which can be used to monitor your exposure.

(****x END OF CATEGORY 04 *****)

(**xxx******** END OF EXAMINATION **xxxxxxxxxx*xx)

s

1. PRINCIPLES OF NUCLEAR POWER. PLANT OPERATION, PAGE 24

~~~~ ~ ~

T ERE66Y d EC5,~EEdT TRd EEER dU6 ELUf6 EL6E i . ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN J

ANSWER 1.01 (1.00) d REFERENCE General Electric, NEDE-24810 (Jun 81)

ANSWER 1.02 (1.00) d REFERENCE-EIH: L-RO-602, p9 ,

BSEP: 02-0G-A, pp 10 -11 ANSWER -1.03 (2.00)

, s. Operating Point

i. b. DECREASE

j. c. (1)

d. Curve C (0.5 each)

J g REFERENCE h- Pump Laws 4

ANSWER 1.04 (1.00) d

, REFERENCE ist and 2nd Laus of Thermodynamics BSEP: L/P 04-2/3-E, pp 7-15; HTFF, 8.30 - 8.31 EIH . L-RG-666, pp 15 - 18, Fig 6 J .,

i,

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

t

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 25

--- iAEERB5isERICsi sEEi iiissFEE EEB FEDi5 FE5s ANSWERS -- BRUNSWICH 182 -85/09/23-X E BROCHMAN ANSWER 1.05 (1.00) 75%. Void Fraction in the core.(0.5) This is because of the increased reasonance capture which would occur (due to the longer slowing down length).(0.5)

REFERENCE EIH: L-RG-604 GGNS: _ Reactor Physics L/Pr pp 1.7 - 9, 10, 13 BSEP: 02-0G-A, pp 39 -49 ANSWER 1.06 (1.00) c REFERENCE EIH: GPNTrVol VII, Chapter 10.1-83-86 BSEP: L/P 02-2/3-A, pp 172 - 176; 02-0G-A, pp 57 - 60 ANSWER 1.07 (1.00) d REFERENCE EIH* L-RG-606, pp 4, 5; Fig. 4 BSEP: 02-2/3-A, pp 177 - 180; 02-0G-Ar pp 60 - 61 ANSWER 1.08 (1.00)

a. GREATER THAN

b. GREATER THAN

c. GREATER THAN

d. LESS THAN REFERENCE Air Ejector Theory /Bernouilli's Equation EIH: L-RO-660 BSEP: HTFF, pp 5.63 - 5.68

.b

r S

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 26 THERMODYNAMICS, HEAT. TRANSFER AND FLUID FLOW ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN ANSWER 1.09 (1.50)

a. Radiolytic decomposition of water. (0.5)

b. Metal (Fuel cladding) to water reaction. (0.5)

c. Oxygen -

5 v/o (0.25 each)

Hydrogen - 4 v/o REFERENCE BSEP: MCD L/P, 5-2/3-C, pp 30 - 34 ANSWER 1.10 (1.00) a REFERENCE BSEP: 02-0G-A, p 26 ANSWER 1.11 (1.00)

s. 10% +- 1%

b. B-10

c. Li

d. He (c & d interchangeable) (0.25 each)

REFERENCE BSEP: 02-0G-A, p 53 ANSWER 1.12 (1.00) b REFERENCE NUS: Vol 3, pp 6.1-3 BSEP: 02-0G-A, pp 22 - 24 l-I

. s'

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 27

~~~~ ~

T55Rk66fNdMfC5,~EEdT fkkNEEEE~5 E~fLUI6 fL6E ANSWERS -- BRUNSWICK 182 -85/09/23-K E BRUCKMAN ANSWER 1.13 (1.00) c REFERENCE BFNP XENON & SAMARIUM'LP, P.4,12 GGNS: LP OP-NP-514, p. 5-10 BSEP: 02-0G-A, pp 57 - 60 ANSWER 1.14 (2.50)

a. Increased Moderation (Pressure increase)

b. Reactor Scram (CV Fast Closure)

c. Turbine Trip (CV Fast Closure)

I d. BPU Cyclin 3

e. SRV's Cycling (0.5 each)

REFERENCE BSEP: HD 05-2/3-A, Section_2.2.2, Figure 3 i

t

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 28 ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN ANSWER 2.01 (1.00) c REFERENCE BFNP: LP47,P. 28 EIH: L-RO-714, Figure 714-6; HNP-2-2447 BSEP: SSM 10-2/3-A, Table 1, Figure 5, pp 64, 79 ANSWER 2.02 (1.00) b REFERENCE BFNP: LPt39,P.18 GGNS: OP-C41-501,P.5,20 BSEP SSM HO-14-2/3-H, Section 4.2.2, p 14 ANSWER 2.03 (1.00) b REFERENCE GGNS: LP'OP-C51-3-501

'. BSEP: SSM 25-2/3-C, p6 l ANSWER 2.04 (1.00) i c REFERENCE BSEP: SSM HD-22-2/3-A1, PP 17 - 25 i

- w - T e

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 29 ANSWERS -- BRUNSWICK 182 -85/09/23-K E BROCKMAN ANSWER 2.05 (2.00)

o. RHR Pump Room Cooler RHR Pump Seal HX's CS Pump Room Cooler (0.5 each)

6. Bypass Valves provide continuous flow thru the system (0.5)

REFERENCE BSEP: SD-43, p 3 ANSWER 2.06 (2.00) 5 seconds -

Nuclear Service Water Pump

10 seconds -

RHR Pump 15 seconds -

CS Pump 20 seconds -

Fire Pump Power Supply Breaker (0.5 each) i REFERENCE t

BSEP: SD 50.1, p 3; SD-17, pp 9, 33; HD 21-2/3-D, Section 3.3.6, pp 50 - 52 ANSWER 2.07 (1.00) b REFERENCE '

BSEP: HD 14-2/3-B, Section 3.2.4 (Obj. f) 4 ANSWER 2.08 (1.00) d REFERENCE BSEP: SD-51, Table 1.4.1; HD 14-2/3-B, Section 2 d

i 1

l l

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

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 30 ANSWERS -- BRUNSWICK 182 -85/09/23-K E BROCKMAN ANSWER 2.09 (1.00)

CHECK VALVE - (The check valve is located below the level of the Suppression Pool and) keeps the Core Spray line downsteam of it full of water (prevents draining to the Suppression Pool) (0.5)

CHECK VALVE BYPASS -

(This allows bypassing the discharse check valve) to drain the discharge line for maintenance. (0.5)

REFERENCE BSEP: HD 14-2/3-E, Section 2.2.2 (Obj. e)

ANSWER 2.10 ( .50)

FALSE - (Auto Start only in the SYSTEM PREF position) (0.5)

REFERENCE BSEP: OP-10, Section 5 ANSWER 2.11 (1.50)

a. F001 - Stays Open F004 - Closes

b. F001 - Stays Open F004 - Closes

c. F001 - Stays Open F004 - Closes (0.25 each) j REFERENCE j BSEP: 11-06-A, Section 3, pp 15, 18 I

ANSWER 2.12 (2.00)

a. 4

. b. 6 l c. 3 (0.25) & 1 (0.25)

d. 2 (0.5 each)

s i

i i

L

4

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 31 ANSWERS -- BRUNSWICK 182 -85/09/23-K E BROCKMAN REFERENCE BSEP: HO-21-2-Ai 15-2/3-6, Rev 18 AOP 20.0 CNSWER 2.13 (1.00) c REFERENCE BSEP: 11-06-A, Section 3, p 18 ANSWER 2 14 (1.00) d REFERENCE BSEP: D-2-VII,Section III.A.2.f s

~

, ,r ..

. .' .. ~v

3. INSTRUMENTS AND CONTROLS PAGE 32

^ -

ANSWERS 0- BRUNSWICK;182 -85/09/23-K E BROCKMAN ANSWER 3.01 (2.00~)

1) _APRM ,'A' fails upscale -> relay K12A deenersi:es (0.4)

2) -> NHS contacts K12A in RPS Trip Logic Al open (0.4)

3) -> Relays K14A"A E deenergi:e (0.4)

4) -> Contacts K14A & E open (0.4)

5) -> Occam pilot solenoid valves for RPS.A,#eenergize and vent (0.4)

. .< . n. , n-REFERENCE ,

BFNP: L/P J2d ,

EIH: L-RO-720, Fig 720-ta, -1b, -2a, -26, -3a, -3b CPNT, Vol. VI, Chapter ,9.3.1-2, 3, 4 BSEP: SSh-28-2,A ANSWER 3.02 (2.00)

e. EIH: INCREASE (45%) (0.5); Speed Demand Limiter (0 5)

BSEP: INCREASE (50%) (0.5); Dual / Master Limiter (0.5)

b. EIH/ INCREASE (0.5); Scoop-Tube Positioning Unit / Positioner (Full DSEP: Range , Mech /Elec S, tops) '(0.5) Partial Credit (0.25) for Error Signal Limiting Network.

REFERENCE EIH: L-RO-714, Fi3ure 4.1(8); GPNT, Vol V, Chapter 4.1 BSEP: SSM 10-2/3-A, Section 3.2.1.1, pp 25 - 30 ANSWER 3.03 (2.00)

a. .66

w + 54% , (0.25) 120% (0.25)

b. Recire Loop flow elements (pump discharge) (0.5)

c. Only the 120% fixed scram (0.5% This is because the flow biased ceram incorporates a time delay into its actuation (~ 6 seconds, repre-centative of the fuel temperature transient time) (0.5) (1.0)

REFERENCE BSEP: SSM 25-2/3-D

~

's er N.

f

'A

.s

, g s

?

e L

9 33

3. INSTRUMENTS AND CONTROLS PAGE ANSWERS -- BRUNSWICK 182 -85/09/23-K E BROCHMAN ANSWER 3.04 (1 00) d REFERENCE BSEP: RTN-033, 012i SD 26.2i SSN 19-2/3-B EIH: L-RQ-705, pp 18, 19; GPNT, Vol. VII, Chapter 9.4 ANSWER 3.05 (2.00)

1. e ii. d REFERENCE BFNP: LP412,P.24; TRANSIENT 820iOI-57,P.53 EIH: L-RO-726 BSEP: RTN 026; HD 17-2/3-B, Section 3.2 ANSWER 3.06 (2.00)

a. CLOSURE signal (0.25) for the Recirculation Pump Discharge Valve (0.25)

b. The white light associated with that pump will illuminate (0.5)

c. 1) Take the manual control switch to START (Spring return to AUT0)(0.5)

2) Reset the LPCI initiation signal (Depress the PS) (0.5)

REFERENCE BSEP: SD-17, pp 9, 33; SD-01, p 51; HD 14-2/3-0, pp 21, 47 ANSWER 3.07 (1.00) d REFERENCE BFNP: LP 422, p 16 BSEP: HD 25-2/3-C, Section 4.1.2, p 19; HD 25-2-D I

W.

3. INSTRUMENTS AND CONTROLS i PAGE 34 ANSWERS -- BRUNSWICK 182 -85/09/23-K E BROCKMAN ANSWER 3.08 (2.00)

a. 4 (1.0)

b. 3 (1.0)

" ~

REFERENCE BSEP: RTN 008; SSM 08-2/3-A

- ANSWER- 3.09 (1.00)

B REFERENCE BFNP LP429,P.6 EIH: GPNT, Vol VII, Chapter 9.2.1, Chapter 9.3 BSEP: RMCS Lesson Plan, p4 ANSWER 3.10 (1.50)

a. No Rod Movement (Rod Block)

q. b. No Rod Movement (Rod Block)

c. No Rod Movement (RSCS Blocks do not affect this light)

. d. Rod Insertion (Light has no effect on rod insertion) (0.5 each)

REFERENCE BSEP: RMCS Lesson Plan, PP 11, 12 ANSWER 3.11 (1.00)

The SRM's CAN be withdrawn. (0.5) The SRM RETRACT PERMISSIVE does not prevent SRM withdrawal (but it would generate a rod block with <100 cps). (0.5)

REFERENCE BSEP: 14-6-He' Section F.2.e.b f

3. INSTRUNENTS AND CONTROLS PAGE 35 ANSWERS -- BRUNSWICK 1A2 -85/09/23-K E BROCHNAN

' ANSWER 3.12 ( .50)

ero spa REFERENCE GGNS: OP-E21-501,'p 12, 17 BSEP HD 14-2-E

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36

~~~~~~~~~~~~~~~~~~~~~~~~

~~~~Rd656L665CdE 66NT66L ANSWERS -- BRUNSWICK 182 -85/09/23-K E BROCKMAN ANSWER 4.01 (1.00) a REFERENCE BFNP: BF-0I-66, pp 5-7 EIH: HNP-1001,.pp 19, 20 BSEP: GP-02, pp 16, 17, 21 ANSWER 4.02 (1.50)

a. To prevent cooling the LP Turbine's diaphragm (below 120 des F)

(0.5), which could cause severe turbine damage (0.5)

b. Less than UNITY (0 25), in the leading direction (negative or incoming VARS) (0.25).

REFERENCE BSEP: OP-27, pp.14, 15 ANSWER 4.03 (1.50) i

1) Isolate the Filter Demineralizers (Maximi =e F/D Bypass Flow) (0.5)

, 2) Feed the F/D Precoat Tank with the contents of the SLC Tank (using the locally stored submersible pump) (0.5)

3) Manually inject the contents of the Precoat Tank via RWCU (0 pen F/D Dome vent & Precoat Discharge vc1ve and place the Precoat Pump CS in ' HAND') (0.5)

4) (Run through the 'decoated' F/D for 10 minutes) (N/A)

- OR -

2a) Fill the F/D Precoat Tank with Borax (0.5)

REFERENCE BSEP: LEP-03r pp 4 -7

4. PROCEDURES - NORMAL,-ABNORMAL, EMERGENCY AND PAGE 37

~~~~Ed656E66fCdE~C6 TEEE~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN ANSWER 4.04 '( .50) 5 (Refuelins)

REFERENCE BSEP: GP-05, p 25; GP-06, pp 4, 6; TS, Section 1.0 ANSWER 4.05 (1.50)

a. 1%

b. 5%

! c. 10% (0.5 each)

REFERENCE BSEP: GP-04, p 4; GP-05, p6 ANSWER 4.06 (1.00)

Because sprayin3 the drywell may decrease containment pressure below atmospheric pressure at a rate beyond the capacity of the Reactor Building-to-Suppression Chamber Vacuum Breakers, resulting in negative containment pressures in excess of design.

REFERENCE BSEP: E0P-01/UG ,

l li J

4. ~ PROCEDURES - NORMAL,' ABNORMAL, EMERGENCY AND PAGE 38 ,

~~~~Rd6EdLdGEddL"C6 TR6L'~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- BRUNSWICK 182 -85/09/23-K E BROCKMAN ANSWER 4.07 (2.50)

c. Closed -(0.5)

b. 1 (or 2) Open (0.5)

c. 1 RHR loop injecting to the Reactor . o g. E \=gs oS RWE. in (0.25) 1 RHR loop in Suppression Pool Cooling E tt 'wmM Cw 3 (0.25)

d. 107 - 164 psig > Suppression Chamber Pressure (0.5)

o. Reactor level > +254 inches (Main Steam Line elevation) (0.5)

REFERENCE BSEP' AOP-15.0, pp 4, 5

, ANSWER 4.08 (1.00)

a. Service Building (0.5)

6. Training Building (0.5)

REFERENCE EIH: 63EP-EIP-061-0 BSEP: R.E.P, Sections 5.3, 5.4 t

ANSWER 4.09 (1.00) d REFERENCE BSEP: GP-02, p9 ANSWER 4.10 (1.00)

1) CRD Pressure cannot be restored (with either pump), AND (0.5)

2) Reactor Presure < 800 psis (0.5) t .

4

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 39

~~~~Rd656E66f6dE~C6 TREE ~~~~~~~~~~~~~~~~~~~~~~~~'

ANSWERS -- BRUNSWICK 182 -85/09/23-K E BROCKMAN REFERENCE BSEP: AOP 02.1, p3 ANSWER 4.11 (1.00)

EPH 4 1 - ATWS (0 5)

EPH # 2 - Successful SCEAM (has occurred) (0.5)

REFERENCE BSEP: EDP-01/UG, Para 7.0, p 23 l ANSWER 4.12 (1.00)

! These instruments have a cold reference les (0.4) and are 1 uncompensated-(0.1)

They are not affected until D/W temperature exceeds the RPV saturation temperature (0.5)

REFERENCE BSEP: E0P-01/UG, CAUTION 4 6, p 36 ANSWER 4.13 (1.00) b REFERENCE BSEP: OI-13, Sect 4.1.1, p2 i

ANSWER 4.14 (1.00) d REFERENCE BSEP: AI-58, Sect 3.1.6, p3 2

.a . .

w

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 40

~~~~Rd656E66fCdE~66 T6EC~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- BRUNSWICK 182 -85/09/23-K E BROCKHAN ANSWER 4.15 (1.50)

1) A radiation monitoring device which continuously indicates the radiation dose rate in the area. (" Pocket Dosimeter")

2) A radiation monitoring device which continuously integrates the

. dose rate in the area and alarms when a preset integrated dose is received. (' Chipper')

3) An individual qualified in radiation protection procedures who is eqttipped with a radiation dose rate monitoring device. (0.5 each)

REFERENCE BSEP: E & R C Hanuale Section 6.6, p 41 i

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TEST CROSS REFERENCE PAGE 1 1

QUESTION VALUE REFERENCE 01.01 1.00 KEB0000044 2

01.02 1.00 KEB0000159 01.03 2.00 KEB0000213 01,04 1.00- KEB0000214 01.05 1.00 KEB0000218 01 06 1.00 KEB0000219 4

01.07 1.00 KEB0000220 01.08 1.00 MEB0000285 01.09 1.50 KEB0000335 t 01.10 1.00 KEB0000336 101.11 1.00 KEB0000337 01.12 1.00 KEB0000338 01.13 1.00 KEB0000395 01.14 2.50 KEB0000400 17.00

, 02.01 '1.00 KEB0000282 02.02 1.00 KEB0000378 02.03 1.00 KED0000380 02.04 1.00 MEB0000381 02.05 2.00 KEB0000382 02.06 2.00 KEB0000383 02.07 1.00 KEB0000385

, 02.08 1.00 KEB0000387 02.09 1.00 KEB0000388 02.10 .50 'KEB0000390 02.11 1.50 KEB0000391

02 12 2.00 KEB0000392 02.13 1.00 KEB0000393

,I 02.14 1.00 KEB0000394' 17.00 03.01 2.00 KEB0000224 03.02 2.00 KEB0000283 -

03.03 2.00 KEB0000373 03 04 1.00 KEB0000374 03.05 2.00 KEB0000375 03.06 2.00 KEB0000376 03.07 1.00 KEB0000377 03.08 2.00 KEB0000379 03.09 1.00 KEB0000384 03.10 1.50 KEB0000386 03 11 1.00 KEB0000389 03.12 .50 KEB0000401 18.00 04.01 1.00 UEB0000228 i

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

y o....

TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE 04.02 1.50 KEB0000340 04.03 1 50 KEB0000341 04.04 .50 KEB0000342 04.05 1.50 ~KED0000343 04.06 1.00 KEB0000344 04.07 2.50 KEB0000345 J04.08 1.00 KEB0000346 04.09 1.00 MEB0000347 04.10 1.00 KEB0000357 04 11 1.00 KEB0000358 04.12 1.00 KEB0000359 04 13 1.00 KEB0000360 04.14 1.00 KEB0000366 04.15 1.50 KEB0000368 18.00 70.00 i

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, M A":.T E R I-4 ENCLOSURE 3 U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: BRUNSHICK 182 REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 85/09/23 EXAMINER: K E BROCKMAN APPLICANT: _________________ .. ____
{
INSTRUCTIONS TO APPLICANT:
Uso separate paper for the answers. Write answers on one side only.
Stcple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passinS Srede requires at least 70% in each category and a final grade of at locst 80%. Examination papers will be picked up luar 64) hours after tho examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 17 00 '4.64
________ '_____ _____..____ ...__.. 5. THEORY OF NUCLEAR POWER PLANT UPERATION, FLUIDS, AND THERMODYNANICS 18.00 26.09
.... ___ ______ ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 17.S0 25.36
_______. ._____ ____.. .... ...____ 7. PROCEDURES - NORMAL, A0NORMAle EMERGENCY AND RADIOLOGICAL CONTROL 16.50 '3.91
________ ..'.... ___...__... . ______ 0. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 69.00 100.00 TOTALS i FINAL GRADE _________________%
All eork done on this examination is my own. I have neither l given nor received aid.
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5. . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2
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-1 4
GUESTION 5.01 (1.00) 4hich of the following statements best describes the operating
bhsracteristics of an LPRh' detector? j
a. Depletion of the detUctor's Uranium coating causes both the neutron and the gamma sensitivity to DECREASE with detector age; the resulting neutron to gamma signal ratio remains relatively CONSTANT.
.. b. Since the detector functions as an_iontration chamber and the Argon gas pressure remains relatively CONSTANT, BOTH the neutron and the gamma sensitivity, as well as the neutron to gamma signal ratio, remain relatively CONSTANT as the detector ages.
c. Depletion of the detector's U,ranium coating causes neutron sensitivity to DECR?ASE,'tivt has an INSIGNIFICANT .
effect on gamma sensitivity; this~results in a neutron to samma signal ratio DECREASE as lJue detector ages. -
t .. 7
d. Depletion,of the detector /Sh Ur vitum coating has an INSIGNIFICANT effect on neutron sensitivity, but causes samma sensitiv,ity to DECREASEI this results in a neutron to gammma sigo.h ratio INCREASE as the detector ages.
OttESTION 5.02 '(1.00)
T:le fission' process,in a commercial reactor requires the neutrons that are ' born' by'ftssion,to be 'thermalized.' The interaction in the reactor core which is mast efficient in thermali:ing neutrons for fission occurs with the ...(CHOOSE ONE)
4. ... OXYGEN atoms in the water moleculos
b. ... BORON atoms in the contr ol rods 8
e. ... IIRCONIUM, atoms ih the fuel cladding
d. ... HYOROGEN' $ toms in the water molecules 4- ,
( =*<* CATEGORY 05 CONTINUED ON NEXT PAGE saava)
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3. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 QUESTION 5.03 (2.00)
Concerning General Electric's Preconditioning Interim Operating Management Recommendations (PCIOMR):
e. Starting with the fuel at a threshold of 11.0 kw/ft, a maximum ramp increase is begun at time 0000 and the final desired power of 13.0 kw/ft is achieved at 2000. The required soak is performed FOR 7 HOURS, at which time the load dispatcher directs an IMMEDIATE Power reduction that takes nodal power down to 11.8 kw/ft. SELECT the valid preconditioned value for this node.
ASSUME THE MAXIMUM RAMP RATE IS .10 Kw/ft/hr
1) 11.0 kw/ft
2) 11.8 kw/ft
3) 12.5 kw/ft
4) 13.0 kw/ft
b. After 12 nours the Load Dispatcher directs a return to full power.
SELECT the minimum time required to get back to 13.0 kw/ft, given the above ramp rate.
1) Immediate (Raise to 13.0 kw/ft, w/o restrictions)
2) 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 
3) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 
4) 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months 
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4
< ---- ggg 33 g3-}{g--------------------------------------
QUESTION 5.04 (2.00)
Attached Figure #213 shows a basic closed loop fluid system with its head vs. flow plot. The two pumps are identical, variable speed, radial, centrifugal pumps. Pumps 1 & 2 are initially operating at one-half speed to supply flow to components 1 & 2, as shown in BOLD LINES.
a. DEFINE (name) Point X (0.5)
b. With the system operating as shown in the diagram, if component 2 were throttled shut from its initial position, would the system flow INCREASE, DECREASE, or REMAIN THE SAME? (0.5)
c. Component 2 is valved out of service, thereby reducing the heat load on the system. How would power consumption be minioiced? (0.5)
(1) Reduce BOTH pump's speed to one-fourth speed.
(2) Stop snd Isolate ONE pump.
d. Which Pump Curve - A, B, C, or 0 - most accurately shows ONE PUMP operating at half-speed to supply flow to the INITIAL system lineup? (0.5)
GUESTION 5.05 (1.00)
Attached Figure 4 219 shows a POWER HISTORY and four possible XENON traces (Reactivity vs Time). SELECT the most accurate cerve for displaying the e::pected XENON transient.
GUESTION 3.06 ,
(1.00)
Attached Figure i 285 is a simplified sketch of a SJAE. For each of the pressure relationships given below, STATE whether the pressure listed first is GREATER THAN, LESS THAN, or EQUAL TO the pressore listed second.
NOTE' THE LOCATION OF THESE PRESSURES CORRESPOND TO THE POINTS INDICATED IN THE FIGUPE.
s. Pfi) as to P(3) (0.25)
b. P(1) as to P(5) (0.25)
c. P(2) as to P(4) (0.25)
d. P(2) as to P(3) (0.25) e
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
l 1
l
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5
_____ggg_gg_ggg7gg______________________________________
GUESTION 5.07 (1.00)
Your reactor has been shutdown for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The intrinsic neutron source which is the hAJOR CONTRIBUTOR to the background neutron population is ... (CHOOSE ONE)
a. ... Deuterium-Gamma reaction (Photo-neutron).
b. ... Spontaneous fission of Curium-242/244.
c. ... Alpha-Oxygen reaction.
d. ... 3pontaneous fission of Uranium-238 GUESTION 5.08 (1.00)
There are two (2) criteria which determine the end-of-life for a control rod. FILL IN THE BLANKS as appropriate.
When the Control Rod has lost ___(a) ___% of its original nes-stive reactivity value due to the burn-up of the ___ b) ( ___.
When the gas pressure due to the ___(c) ___ and the ___(d) ___
formation reaches 6565 psis.
GUESTION 5.O? (1.00)
The reactor trips from full power, equilibrium XENON conditions. Twenty-four hours later the reactor is brought critical and power level is main-tained on range 5 of the IRhs for several hours. Which of the following statements is CORRECT concerning control rod motion?
a. Rods will have to be withdrawn due to XENON build-in.
b. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of XENON burnout.
c. Rods will have to be inserted since XENON will closely follows its normal decay rate.
d. Rods will approximately remain as is as the XENON estab-lishes its equilibrium value for this power level.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE **x**)
1
. _ _ _ _ _ _ . _ _ _ - _ _ __. __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ m_ ____. _ _ _ _' _____-___.____.________.______.___i
l 1
5. THEORY OF NUCLEAR POWER FLUIDS, AND PAGE 6 l
_____ggg_gg_g_gggg__________' PLANT OPERATION, ____________________________
QUESTION 5.10 (2.00)
Attached Figure t 398-represents a transient that occur at a BWR.
i Givent (1) The Reactor is initially at 105% Power l t (2) The Reactor is in MANUAL Control (3) There is a Loss of 100 des F of Feedwater Heatins ;
(Heaters bypassed automatically)
(4) No operator actions occur
! EXPLAIN the cause(s) of the following recorder indications ,
j a. Variation.in NEUTRON FLUX ( GRAPH I) (Point A - Point B)
b. Variation in VESSEL LEVEL (GRAPH III) (Point C - Point D)
, c. Increase in REACTOR PRESSURE ( GRAPH V) (Point E - Point F) ,
d. Variation in DOPPLER REACTIVITY ( GRAPH VI) (Point G - Point H) ,
1 l
2 OUESTION 5.11 (3.00)
$. nttached~ Figure # 399 represents a transient that occur at a BWR.
, Given: (1) There is a-Feedwater Control System Failure (2) Both Feedwater Pumps TRIP (3) No operator actions _ occur I-EXPLAIN the cause(s)~of the fo11owin3 recorder indications:
s. Variation in NEUTRON FLUX ( GRAPH I) (Point A - Point B) j b. Variation in NEUTRON FLUX ( GRAPH I) (Point B -~ Point C)
c. Variation in-REACTOR PRESSURE. (GRAPH II) (Point D - Point ~E) i d. -Variation in CORE INLET FLOW ( GRAPH V) (Point F - Point G)
e. DECREASE in CORE INLET FLOW ( GRAPH V) (Point G)
! f. Variation in VESSEL STEAM FLOW (GRAPH VI)~ (Point H - Point I) h (**xxx-CATEGORY 05 CONTINUED ON NEXT-PAGE xxxxx) i-
-l 3
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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 7 QUESTION 5.12 (1.00)
In a suberitical reactor, Keff is increased from .880 to .965.
Which of the following is the amount of reactivity that was added to.the core?
a. .085 delta k / k
b. .100 delta k / k
c. .125 delta k / k
d. .220 delta k / k
(***** END OF CATEGORY 05 *****)
i
6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 8 OUESTION 6.01 (2.00)
Unit 2 is operating at 100% RTP when APRM 'A' fails upscale and results in a reactor half-scram. Utilizing the attached RPS trip logic diagrams (Figures $224 A thru F) DESCRIBE in a STEP-BY-STEP fashion (with regard ta the opening / closing, ener-S i zing /deenergiring of ALL applicable contacts and relays) how the APRM vpscale trip results in an actuation of the scram solenoid valves.
NOTE' IF THE ATTACHED DIAGRAMS CAN NOT BE EASILY READ, ASSIGN THE CONTACTS / RELAYS, ETC NUMBERS AND REFER TO THEM IN YOUR ANSWER.
QUESTION 6.02 (1.00)
The plant is operating normally at power when the 'Recire Pump A Inner ;
i Seal Leakage and Seal Staging Flow' FS trips low. You note a DECREASE in No. 1 Recire Pump seal pressure. Which of the following failures would result in these indications?
a. Failure of No. 1 seal t
b. Failure of No. 2 seal
c. Plugging of the No. 1 internal restricting orifice ;
d. Plugging of the-No. 2 internal restricting orifice GUESTION 6.03 (2.00)
The APRM scram f.>nction actually consists of two separate cetpoints:
a. FILL IN THE BLANKS:
Flow Biased Scram. - __( 1) -_
w + -_ (2) __% (0.25)
Fixed Scram -
__ (3) __% (0.25)
b. LIST the specific location (s) of the sensor (s) which measure the variable 'w' . (0.5)
}
c. While operating at power, one HSIV fails shut resulting in a brief (~ 1 second)fflux spike to 121% power. STATE which of thc- two scram setpoints mentioned above (one or both)-should initiate a reactor scram. JUSTIFY your choice. (1.0)-
(***** CATEGORY 06 CONTINUED ON NEXT PAGE srxxx) i i
e 4
i
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6. PLANT SYSTEMS DESIGN, PONTROL, AND INSTRUMENTATION PAGE 9 GUESTION 6.04 (1.00)
Unit 1 is operating at 100% RTP, with recire in Master Manual. An operator inadvertently DECREASES the ' Pressure Set' on EHC by 5 psis.
ASSUME: 1. No Further Operator Actions
2. All other EHC control settings are normal
3. Starting Parameters o TCV's - 100% Steam Flow Position o BPV's -
0% Steam Flow Position o Pouer - 100% Rated Thermal Power o Pressure - 1010 psis NOTE: FIGURE 4 374 IS ATTACHED FOR REFERENCE Which of the following most accurately describes both the INITIAL RESPONSE and FINAL STATUS of the different parameters and components.
a b c d INITIAL RESPONSE o TCV's ING CHANGE ICLOSE (~93%) ICLOSE (~83%) i NO CHANGE o BPU's i NO CHANGE I OPEN ("17%) l NO CHANGE I OPEN (~10%)
o Power i DECREASE I NO CHANGE I INCREASE I DECREASE o Pressure i DECREASE I NO CHANGE I INCREASE I DECREASE
! I i i FINAL STATUS I I I I I I I I
~
o TCV's 10%(MSIV Shut)I 83 % i ~100 % 1
~
100 %
~
o BPV's 10%(MSIV Shut)I 17 % 1 0 % 1 0%
o Power 10% (Rx Scram)! > 100 % i - 100 % i 's 100 %
o Pressure IAs contro11edi <1010 psigt '>1010 psigt <1010 psis iby SRV's and i I I IHPCI/RCIC 1 1 I (xxxxx CATEGORY 06 c,0NTINUED ON NEXT PAGE xxx**)
l
6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 10
, GUESTION 6.05 (1.00) i TheLHode Switch is in RUN and 11 of the 17 LPRM's assigned to APRH 'A' i- are operable. Which of the following automatic actions will occur as a result of one of the operable 11 LPRM inputs FAILING DOWNSCALE.
l ASSUME NO OPERATOR ACTIONS 4
a. A Rod Block
b. A Reactor Half-scram
c. A Rod Block and a Reactor Half-scram
d. .None of the Above t
QUESTION 6.06 (2.00)
A LOCA and a Loss of Offsite Power (LOSP) have occurred simultaneously.
The Diesel Generators have started.and AUTOMATICALLY energized their respective emergency busses. The sequential loading relays will delay the automatic starting of emergency' loads for 5, 10, 15,
or 20 seconds.to minimize starting surges on the diesels. LIST the loads that will be energized at EACH of these time intervals.
NOTE: 00 NOT ADDRESS EACH BUS; LIST FOR'A GENERIC BUS WITHOUT THE PARTICULAR DESIGNATOR FOR THE LOAD QUESTION 6.07 (1.00)
The Full-Core Display on the center panel has a BLUE scram light for each control. rod. What is DIRECTLY indicated when this light is illuminated?
a. That BOTH the Inlet and Outlet scram valves for that rod are OPEN.
b. That BOTH scram pilot air valves for that rod are
! doenergized.
c. That the beyond full-in overtravel _ reed switch for that rod is closed.
d. That the' scram accumulator for that rod i s depres-surized.
4
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7-
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.- 6. PLANT SYETEMS DESIGNr CONTROL, AND TNSTRUMENTATION PAGE 11
- QUESTION 6.08 (1.00)
The HPCI. system is in-TEST operation with the Test Signal Generator Control Switch set to' control HPCI speed at 2000 rpm. Select the s.tatement that most accurately describes the actions (Hanual -OR-Automatic)~that would be necessary to. place HPCI in ENERGENCY SERVICE follwing an AUTO initiation signal.while in TEST.
NOTE
HPCI FLOW CONTROLLER IS IN AUTOMATIC
a. HPCI will AUTO align for injection at the preset speed of 2000 RPM.
b. HPCI will AUTO align.for injection at the FULL RATED speed and flow as determined by the flow controller.
c. HPCI will. AUTO align for injection but must be taken OUT OF TEST-and placed in AUTO to actually inject into the vessel.
d. HPCI most be MANUALLY realigned for injection and placed in the AUTO mode to inject into the vessel.
GUESTICN 6.09 ( .50)
TRUE OR FALSE A train of the Standby Gas Treatment System (SBGTS) will AUTO START on receipt of a valid initiation signal, when the train's control 4 switch is in the STANDBY. position.
- GUESTION 6.10 (1.50)
Each of the events listed below would cause the Peactor Hster Cleanup System to ISOLATE. For each event, STATE whether the Inboard Isolation Valve (F001) AND/OR the Outboard Isolation Valve (F004) would automatically CLOSE.
a. A loss of cooling water to the NRHX, resulting in an NRHX RWCU-outlet temperature e::ceeding 140 des F.
b.. A MANUAL initiation of the Standby Liquid Control System (SLC).
c. A loss of ALL-AC._powerr concurrent with a LO-LO Reactor ~ Water Level.
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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12 GUESTION 6.11 (1.004 Once ADS is AUTOMATICALLY initiated, the ADS valves will REMAIN OPEN until which of the following occur?
a. A Reactor Water Level HI Trip signal is received.
b. One of the ADS 120 second Timers are reset.
c. Reactor Presscre DECREASES to the LP ECCS injection permissive (<450 psis).
d. Reactor Pressure DECREASES to approximately 50 psis above Primary Containment pressure.
GUESTION 6.12 (2.00)
For each of the following situations (i and ii) select the correct Feed-water Control System / plant response from the list (a through e) which follows. An answer may be used more than once, and NO operator actions are taken.
s. Reactor water level decreases and stabilizes at a lower level.
b. Reactor water level decreases and initiates a reactor scram.
c. Reactor water level increases and stabilizes at a higher level.
d. Reactor water level increases and initiates a turbine trip.
e. None of the above.
i. The plant is operating at 70% power, in 3-element control, when one of the Steam Flow Detectors FAILS DOWNSCALE.
ii. The plant is operating at 90% power, in 3-element control, when HPCI inadvertently initiates and injects.
(****x CATEGORY 06 CONTINUED ON NEXT PAGE xxxxr) l l
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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 13 GUESTION 6.13 (1.00)
The reactor is critical at approximately 5 psis and the ' Press-uri:stion' phase of GP-02 is being performed. The Normal Control Range GEMAC LI's in the control room read the following ' approx-imate' values.
GEMAC A (N004 A) 187
GEMAC B (N004 B) 188 '
GEMAC C (N004 C) 187 '
The two Emergency System Range (Yarway) control room level indicators should read approximately ... (CHOOSE ONE)
a. ... 150
b. ... 165
c. ... 188
d. ...>210 QUESTION 6.14 (1.00)
The plant is operating at 23% power and both Recire Pump M/A Transfer Stations are in MANUAL and set for minimum speed.
The 'Recire Flow B Limit' annunciator is CLEAR. For the following instance, STATE how the speed of Recire Pump 'B' will change (i.e., INCREASER DECREASE, REMAIN THE SAME) and WHICH COMPONENT (S) of the control / positioning system is/are LIMITING.
NOTE: FIGURE t 283 IS PROVIDED FOR REFERENCE Tachometer output feedback signal fails low - Contact Y1 Opens (1.0)
(xxxx* END OF CATEGORY 06 *****)
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7. PROCEDURE'S . NORMAL, ABNORMAL,' EMERGENCY AND PAGE 14 i
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QUESTION 7.01 (1.00)
D A plant startup is in progress and condenser vacuum is being i established in accordance with GP-02, ' Approach to Criticality and Pressuri:stion of the Reactor *. What is the proper sequence for component / subsystem startups.
a. Steam Packing Exhauster, Steam Seal Header, Mechanical Vacuum Pump, Steam Jet Air Ejector.
b. Steam Seal Header, Steam Packing Exhauster, Mechanical Vacuum Pump, Steam Jet Air Ejector.
c. Mechanical Vacuum Pump, Steam Packing Exhauster, 1 Steam Seal Header, Steam Jet Air Ejector.
d. Steam Packing Exhanster, Hechanical Vacuum Pump, Steam' Seal Header, Steam Jet Air Ejector.
t QUESTION 7.02 (1.50)
OP-27, ' Generator and Exciter System Operating Procedure', cautions the operator to minimize turbine generator operation below 100 MWe.
a.. EXPLAIN the basis for this operating limitation. (1.0)
b. Per CP-27, 'the power factor shall never be allowed to become less than _______.* (0.5)
QUESTION 7.03 (1.50)
LEP-03, ' Alternate Bcron Injection *, identifies the RWCU i pumps / system as one of six that may be selected and used to inject boron if the SLC system is NOT AVAILABLE when baron injection is required. DESCRIBE the method / flow path which 1
is established-to accomplish this injection.
~
~ NOTE: ASSUME LEVEL IS BEING MAINTAINED > 112'
(***** CATECORY 07' CONTINUED ON NEXT PAGE ***xx) 4 i'
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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15
- ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~E 616 LUG 1Chl CUETEUL QUESTION 7.04 (1.50)
When increasing Recirculation Pump speeds with both controllers in MANUAL, their speeds should NORMALLY be maintained within
___ (a)___% (of each other). The speed differential is LIMITED to.___(b) ___% when below 75% core flow and ___(c) ___% when abcue 75% core flou.
GUESTION 7.05'- (1.00)
A LOCA has occurred and a high temperature steam environment
. exists-in the drywell. EXPLAIN why the drywell sprays must NOT be initiated in the ' Unsafe' region of. attached Figure # 344 "Drywell Spray Initiation Pressure Limit'.
GUESTION 7.06 (2.50)
A reactor Cooldown is in progress and NEITHER loop of RHR can be placed.in the Shutdown Cooling Mode. In accordance with A0P-15.0,
-Alternate Shutdown Cooling has been established to remove decay
~
heat and continue the cooldown. STATE the condition / status of the following components / parameters when operating in this made of shut-down cooling.
a. MSIV's (0.5)
b. SRV'.s (0.5)
c. RHR Loops A & B (0.5)
d. Reactor Pressure (Compared to Suppression Chamber Pressure) (0.5)
e. Reactor Level (Provide numerical value or component reference) (0.5)
QUESTION 7.07 (1.00)
Per the BSEP Radiological Emergency Plant
a. STATE the NORMAL OSC location. (0.5)
6. STATE the NORMAL EOF location. (0.5)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
'7. PROCEDURES - NORMAL, ABNORMAL, ENERGENCY AND PAGE 16 i RADIOLOGICAL CONTROL-
'GUESTION 7.08 (1.00) -
When the Rod Worth Miniminer is inoperable or bypassed, commencement of Control Rod movement for a Reactor Start-up l is ... (CHOOSE ONE)
'a. ... PERMITTED, providing that a second qualified member of the plant technical staff verifies the rod sequence.
b. ... PERNITTED, until the reactor thermal power reaches the Low Power Set Point (LPSP).
. c. ... PERMITTED, if approved by the SOS and the Manager-4 Operations, and appropriate operating limitations are noted in the Pre-Startup Checklist.
d. ... NOT PERMITTED.
GUESTION 7.09 (1.50)
Given that a required valve line-up HAS NOT been performed for l the present unit startup. Per GP-01, STATE the alternate require-3 cents which must be met to consider this valve line-up COMPLETE.
GUESTION 7.10 (1.00)
Unit 1 is operating at power when CRD Pressure DECREASES to "O' psis.
STATE the conditions, per AGP O2.1, ' Inability to Nove Control Rods',
which would require a MANUAL SCRAM.
GUESTION 7.11- (1.00) j There are two'End Path Manuals at BSEP. STATE the respectiv9 OPER-ATIONAL OCCURANCES which will put an operator.into EACH EPH.
QUESTION 7.12 (1.00)
Level Detectors NO36/NO37 are not included in the E0P-01/UG t caution (CAUTION-#6) concerning high temperatures near the reference les vertical runs. EXPLAIN WHY these instruments are EXCEPTED from this caution and WHEN, if ever, these instruments would develop excessive inaccuracies.
l
(***** CATEGORY 07 CONTINUED ON.NEXT PAGE *xxxx) i
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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17
~~~~E3356LUU5C3L'UUEIRUL~' ~ " "' - ~"
GUESTION 7.13 (1.00)
MATCH tne following Contamination Levels with the appropriate MINIMAL requirements for wearing PC Gloves
a. Alpha Contamiantion 1. 10 dpm / 100 cmacm
6. Beta-Camma Contamination 2. 100 dpm / 100 cmzem
3. 1,000 dpm / 100 cmacm
4. 10,000 dpm / 100 cmmem QUESTION 7.14 (1.00)
FILL IN THE BLANK with the correct response per FH-11, ' Fuel Handling Procedure.'
a. When moving the refueling platform with fuel loaded, ensure that the fuel grapple is at '_____(1) _____' befoae moving. (0.5)
b. Plant directions are azimuthally r ota ted appro::imately ___( 2 ) ___
(degrees) ___ (3) ___ (direction) from that of magnetic directions. (0.5)
(***** END OF CATEGORY 07 *****)
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 OUESTION 8.01 (1.00)
Unit 2 is operating at 75% rated thermal power. Channel Functional Tests are performed on all of the MSL Radiation Monitoring System channels. ' Channels A and D test UNSATi Channels B and C test SAT.
l Maintenance has no estimate of repair time and will not be able to commence troubleshooting and_ repair for at least 16 - 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .
Which of the following actions most correctly detail the allowances 4 and/or limitations imposed by the Technical Specifications in this ,
instance?
NOTE

APPLICABLE TS's ARE ENCLOSED FOR REFERENCE
6. Be in at least HOT SHUTDOWN within 6 hours
b. Be in at least-HOT SHUTDOWN within 6 hours and in i COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
1'
c. Place MSL Rad: Hon Channel 'A' in the tripped condition within one hour -AND- be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
4 d.- Place MSL Rad Mon Channel 'A' in the tripped condition j within one hour -AND- be in at least HOT SHUTDOWN within i 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
1 OUESTION E,02 ( .50)
Given the following conditions:
Mode Switch -
Shutdown i Temperature -
180 des F Pressure -
O psis Level -
330 inches (R605 Indication)
RHR -
SDC Mode Head bolts to the RPV are DETENSIONED STATE'the above described Operational Condition.
L
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,,. , _ . , . _ , . . . . _ . , , , , _ , . _ _ _ , _ , - . - . , . v. . . . _
- _ _ _. _ . ~ . _ _ _ m.. . _ _ _ _ _ . . . . _ . _ . _ . _ _ . . _ _ _ __ _ _ .- _ . ..- _ .
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. 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19 i 00ESTION- 8.03 (1.00)
OI-13 provides the operator with the following CAUTION:
'When performing' valve checks or line-ups on systems that are normally operated at high temperatures, valves should NOT.be positioned on their backseats.'
, EXPLAIN the basis for.this CAUTION.
OUESTION 8.04 (1.00) i Procedure OI-01, '
Operating Principles and Philosophy', states that during the performance of normal evolutions by two persons I AT DIFFERENT LOCATIONS, both persons should have a copy of the
procedure. In which of the following situations would i t be allowable, per OI-01, for only ONE INDIVIDUAL to have a copy of a the procedure?
s i a. During an evolution'that requires only a limited number
! of manipulations by an individual under the direction of the ;
! controlling individual, only the individual CONTROLLING the l'
! evolution need_have a copy of the procedure.
j b. During an evolution that requires a limited number of
manipulations by an individual under thg direction of the j controlling individual, only the individual PERFORHING the
manipulations need have a copy of the procedure.
l c. During an evolution in a CONTAMINATION AREA that will ;
t be completed within one hour by the individual performing
the manipulations, irregardless of the number (of manipo-i lations) involved, only the individual CONTROLLING the evolution n- d have a copy of the. procedure. ;
d. During an. evolution in a CONTAMINATION AREA that will l be completed within one hour by the individual performing j the manipulations, irregardless'of the number (of manipu-j lations) insolsed, only the individual PERFORHING the j aanipulations need have a copyu of the procedure.
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 20 GUESTION 8.05 (1.00)
The Control Room Operator is unable to perform one of the PT's on the DSR.
a. EXPLAIN how he must document the non-performance of the PT. (0,5)
b. EXPLAIN how he must reschedule the PT, prior to routing the DSR to the Shift Foreman for review. (0.5)
QUESTION 8.06 (1.00)
Concerning the Unit 2 EOC-RPT in the Reactor Recirculation System
a. EXPLAIN the phy sical phenomenon which necessitates the EOC-RPT. (0.5)
b. STATE the operational limitation (s) that the plant accepts to be able to operate at power conditions with the EOC-RPT bypassed. (0.5) i QUESTION S.07 (1.00)
Per the Unit 2 Technical 5pecifications, LIST two of the conditions which must be satisfied prior to BYPASSING the refueling interlock to permit removal of any Control Rods.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 21 GUESTION 8.08 (1.00)
Per DI-13, ' Locked Valve Identification and Locking *, STATE the proper method for CONFIRMING valve position.
a. Turn the valve hand wheel in the OPEN direction; confirm the position by observing that the stem travels in the OPEN direction.
b. Turn the valve hand wheel in the CLOSED direction; confirm the position by valve position indicator and/or firm tightness.
c. Turn the valve hand wheel in the DESIRED POSITION direction; confirm the position by valve position indicator and/or firm tightness.
d. Turn the valve hand wheel in the DESIRED POSITION direction; confirm the valve position by observing that the stem travels in I
the DESIRED direction.
OUESTION 8.09 (1.00)
Admin Procedure, Section 11.7, lists the systems which require that two operators perform any TAGGING orerations due to these systems being safety-related and requiring Independent Verific-ation.
LIST the two OTHER INSTANCES when two operators are required for TAGGIMO cpersti r.:, par AI 50.
GUESTION G.10 (1.00)
The concentration of radioactive material r e l e a ser; in liquid effluents to UNRESTRICTED AREAS after dilution in the discharge canal shall be limited to the concentrations specified in 10 CFR, Part 20... For dissolved or entrained noble gases, the concentra-tion shall be limited to ... (CHOOSE ONE)
a. ... 2E-4 ve/ml
b. ... 1E+2 ut/mi
c. ... 1E+1 Ci/l
d. ... 1E+6 Ci/l
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 22 GUESTION 8.11 (1.00)
WEEKLY and DAILY PT's are tracked by which of the following?
a. RCI-02.4 Only
b. 0I-03 (DSR) Only
c. RCI-02,4 -AND- DI-03
d. RCI-02.4 -OR- 0I-03 GUESTION 8.12 (1.00)
You are conducting a Drywell Closecut Inspection and notice RED PAINT on the Drywell Cooler Dampers. STATE the purpose of this red paint.
GUESTION 8.13 (1.00)
You, as Shift Forsman, must prepare a NEW OWP to remove a system from service for maintenance. WHO, of the following, is required to approve this OWP for TS interpretation, per 0I-10, ' Operations Work Procedures.'
a. A second licensed operator
b. The SOS
c. Engineer - Operations
d. PNSC
(*xx** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
9. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 23 QUESTION 8.14 (1.00)
_____________ clearance.' (Definition) -
Issued to a foreman or lead man who is responsible and cognizant of two or more jobs being performed within the boundary of this clearance.
Fill in the blank with one of the following.
a. Local
b. Individual
c. Multiple
d. Haster GUESTION 8.15 (1.00)
Unit 2 is operating at 92% Rated Thermal Power, with one outstandins~LCO:
The HPCI Aux Gil Pump motor has been removed for repairs (It has been out for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months )
Three hours into your shift, a routine surveillance by the instrument shop determines that Channel C-2 of the RCIC Reactor Vessel Water Level - HIGH has a setpoint of 213 inches. It can not be readjusted by the i.n s t r u men t technician.
Which of the following actions most correctly detail the allowances and/or limitations imposed by the Technical Specifications in this instance.
a. No new limitations or TS Operational Condition restrictions are initiated.
b. Place the INOPERABLE channel in the tripped condition within one hour - OR - declare RCIC INOPERABLE. Power operations may continue.
c. Be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
d. Be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 24 GUESTION 8.16 (1.00)
Unit 2 Technical Specifications require that the coupling intes-rity of a Control Rod be demonstrated by withdrawing the rod to the Fully Withdrawn position and verifying that the rod does not so to the overtravel position. LIST TWO (2) of the three circum-stances under which the Technical Specifications REQUIRE that this coupling check be performed.
QUESTION 8.17 (1.00)
Core SHUTDOWN MARGIN must be determined by measurement within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior tot or during, the first STARTUP after completing CORE ALTERATIONS.
a. SDM must show the core to be Subcritical byr at leastr
'R + ____ % delta k / k' (0.5)
6. Define 'R' (0.5)
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 25 ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN ANSWER 5.01 (1.00)
C REFERENCE Nuclear Power Reactor Instrumentation Systems Handbook,Harrer &
Beckerly, p 44.
BSEP: L/P 25-2/3-C, pp 3,4 EIH: GPNT, Vol. II, Chapter 3.E ANSWER 5.02 (1.00) d REFERENCE EIH: L-RO-602, p9 BSEP: 02-0G-A, pp 10 -11 ANSWER 5.03 (2.00)
a. 3
b. 2 (Or as appropriate per the part (a) response)
REFERENCE General Electric NEDE 21493 (Rev 5)
EIH: GPNT, STA Training hanual, Section 9; L-RO-673, pp 6, 7 BSEP: L/P 06-2/3-B, p 1 - 16; HTFF, Chapter 9 ANSWER ~5.04 (2.00)
a. Operating Point
b. DECREASE
c. (1)
d. Curve C (0.5 each)
REFERENCE Pump Laws h.
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 26 ANSWERS -- BRUNSWICK 1&2 -85/09.'23-K E BROCKMAN ANSWER 5.05 (1.00)
C REFERENCE EIH GPNT,Uol VII, Chapter 10.1-83-86 BSEP: L/P 02-2/3-A, pp 172 - 176; 02-0G-A, pp 57 - 60 ANSWER 5.06 (1.00)
a. GREATER THAN
b. GREATER THAN
c. GREATER THAN
d. LESS THAN REFERENCE Air Ejector Theory /Bernouilli's Equation EIH: L-RG-660 BSEP: HTFFr pp 5.63 - 5.68 ANSWER 5.07 (1.00) a REFERENCE BSEP: 02-0G-A, p 26 ANSWER 5.08 (1.00)
3. 10% +- 1%
b. B-10
c. Li
d. He (c & d interchangeable) (0.25 each)
REFERENCE BSEP: 02-0G-A, p 53
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 27 ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKHAN ANSWER 5.09 (1.00) e REFERENCE BFNP: XENON & SAMARIUM LP, P.4,12 GGNS: LP OP-NP-514, p. 5-10 BSEP: 02-0G-A, pp 57 - 60 ANSWER 5.10 (2.00)
a. Positive reactivity a Jdition caused by colder feedwater (Increased moderation).
b. Increased voiding (follewing scram).
c. Increased reacter power.
d. Increased power (causing increase in fuel temperature). (0.5 each)
REFERENCE BSEP: HO 05-2/3-A, Section 3.1, Figure 12 ANSWER 5.11 (3.00)
a. Increased core voiding.
b. Reactor Scram on Low Water Level (165')
c. MSIV Isolation on Low Water Level (112.5')
d. Recire Pump Runback (45 %)
e. Recirc Pump Trip on Low Water Level (112.5')
f. SRV's cycling to control pressure. (0.5 each)
REFERENCE BSEP: H0 05-2/3-A, Section 5.3.2, Figure 18, pp 61, 62
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 28 ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKHAN ANSWER 5.12 (1.00) b REFERENCE NUS: Vol 3, pp 6.1-3 BSEP 02-0G-A, pp 22 - 24
6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 29 ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN ANSWER 6.01 (2.00)
1) APRM 'A' fails upscale -> relay K12A deenergines (0.4)
2) -> NMS contacts K12A in RPS Trip Logic At open (0.4)
3) -> Relays K14A & E deenergire (0.4)
4) -> Contacts K14A & E open (0.4)
5) -> Scram pilot solenoid valves for RPS A deenergi:e and vent (0.4)
REFERENCE .
BFNP L/P #28 EIH: L-RO-720, Fig 720-1ar -1br -2a, -2br -3ar -36 GPNT, '!o l . VI, Chapter 9.3.1-2, 3, 4 BSEP: SSN 28-2-A ANSWER 6.02 (1.00) e REFERENCE BFNP: LPt7,P. 28 EIH: L-RO-714, Figure 714-68 HNP-2-2447 BSEP: SSn 10-2/3-Ar Table ir Figure 5r pp 64, 79 ANSWER 6.03 (2.00)
a. .66 *w + 54% (0.25) 120% (0.25)
b. Recire Loop flow elements (pump discharge) (0.5)
c. Only the 120% fixed scram (0.5) This is because the flow biased scram incorporates a time delay into its actuation (~ 6 seconds, repre-sentative of the fuel temperature transient time) (0.5) (1.0)
REFERENCE BSEP: SSM 25-2/3-D l
._ . _. . . . - _ _ _ . . . _ - _ _ _ _ _ . _ _ _ _ - . , ~. .
p+
f
. 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 30
. ANSWERS -- BRUNSWICK 1&2 -95/09/23-K E BROCKMAN 1
ANSWER 6.04 , (1.00) , c d
r i i REFERENCE
, BSEP: RTN-0)33, 012; SD 26'.2i SSM 19-2/3-B
.c EIH: L-RG-y05, pp 18, 19; GPNT, Vol. VII, Chapter 9.4 6,
ANSWER 6.05 (1.00)
J?'
d ,
w ,
I kEFERENdE
,/f BFNP: 4LP#22,#jp 16 fi BSEP: HD 25-223-C, Section 4.1.2, p 19; HO 25-2-D e ,
/
I' ' ANSWER.' 6.06 (2.00) * '
.i
5 seconds -
Nuclear Service Water Pump i
10 seconds- 7 RHR' Pump i, ,'
" 15 seconds . -I' i ,
CS Pump
./
20 sbec;nds -
Fire Pump Power Supply Breaker t (0.5 each)
REFERENCE * <-
BSEP:'SD 50,,1a p 3i SD-17, pp 9, 33; HD 21-2/3-D, Section 3 3~.'6, pp 50 - 52 1.r L .
[,(
t ;- ;
ic ANSWER;C 6.07 (1.00)
.v . v ,
-;? B,- h c rREFEPENCE i BFNP

LP429,P.~6
/ EIH GPNT, Vol VII,. Chapter 9.2.1, Chapter 9.3 BSEP: RMCS Lesson Plar., p4
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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 31 ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN ANSWER 6.08 (1.00) b REFERENCE BSEP: HD 14-2/3-B, Section 3.2.4 (Obj. f)
ANSWER 6.09 ( .50)
FALSE - (Auto Start only in the SYSTEM PREF position) (0.5)
REFERENCE BSEP: OP-10, Section 5 ANSWER 6.10 (1.50)
a. F001 - Stays Open F004 - Closes
b. F001 - Stays Open F004 - Closes
c. F001 - Stays Open F004 - Closes (0.25 each)
REFERENCE BSEP: 11-06-A, Section 3, pp 15, 18 ANSWER 6.11 (1.00) d REFERENCE BSEP: D-2-VII,Section III.A.2.f 1
)
O 60 PLAhi SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 32 ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN ANSWER 6.12 (2.00)
i. a ii. c REFERENCE BFNP: LPt12,P.24; TRANSIENT t20;0I-57,P.53 EIH: L-RG-726 BSEP: RTN 026 ANSWER 6.13 (1.00) d (1.0)
REFERENCE BSEP: RTH 008; SSM 08-2/3-A ANSWER 6.14 (1.00)
EIH/ INCREASE (0.5); Scoop Tube Positioning Unit / Positioner (Full BSEP: Range , Mech /Elec Stops) (0,5) - OR - Volt / H: differential causing the pump to trip. (0.5) Partial Credit (0.25) for Erro. - *nal Limiting Network.
REFERENCE EIH: L-RG-714, Figure 4.1(S); GPNT, Vol V, Chapter 4.1 BSEP: SSM 10-2/3-A. Section 3.2.1.1, pp 25 - 30 l
I I
7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33 RADIOLOGICAL CONTROL ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN ANSWER 7.01 (1.00) a REFERENCE BFNP: BF-0I-66, pp 5-7 EIH: HNP-1001, pp 19, 20 BSEP: GP-02, pp 16, 17, 21 ANSWER 7.02 (1.50)
a. To prevent cooling the LP Turbine's diaphragm (below 120 des F)
(0.5), which could cause severe turbine damage (0.5)
b. Less than UNITY (0.25), in the leading direction (negative or incoming VARS) (0.25).
REFERENCE BSEP: OP-27, pp 14, 15 ANSWER 7.03 (1.50)
1) Isolate the Filter Demineralizers (Maximize F/D Bypass Flow) (0.5)
2) Feed the F/D Precoat Tank with the contents of the SLC Tank (using the locally stored submersible pump) (0.5)
3) Manually inject the contents of the Precoat Tank via RWCU (0 pen F/D Dome vent & Precoat Discharge valve and place the Precoat Pump CS in ' HAND') (0.5)
4) (Run through the 'decoated' F/D for 10 minutes) (N/A)
- OR -
Za) Fill the F/D Precoat Tank with Borax (0.5)
REFERENCE BSEP: LEP-03, pp 4 -7 l
1
, o
7. PR9CEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R d6f6E66fddL'Ud TkdL ANSWERS -- BRUNSWICK 182 -85/09/23-K E BROCKMAN ANSWER 7.04 (1.50)
a. 1%
b. 5%
.c.-10% (0.5 each)
REFERENCE-BSEP -GP-04, p 4; GP-05, p6 ANSWER 7.05 (1.00)
Because spraying the drywell may decrease containment pressure below atmospheric pressure at a rate beyond the capacity of the Reactor Building-to-Suppression Chamber Vacuum Breakers, resulting in negative containment pressures in excess of desi 3 n.
REFERENCE
, BSEP: E0P-01/UG ANSWER 7.06 (2.50)
a. Closed (0.5)
b. 1 (or 2) Open (0.5)
I
c. 1 RHR loop injecting to the Reactor (0.25) 1 RHR_ loop in Suppression Pool Cooling (0.25)
~
I -
d. 107 - 164 psig > Suppression Chamber Pressure (0.5)
e. Reactor level > +254 inches (Main Steam Line elevation) (0.5)
! REFERENCE
) BSEP: 'AOP-15.0, pp 4, 5
____.__.___m ___.________m_____ _ _ _ _ _ _ _ _ _ _ _ .______.___m___._.________.____.___-___._______.______.___________.____.____m_ _______.,__.______._._____.___________________-__.___.___.._mm.__
7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35
- ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~Rd656L5siEdL E5NTR6L ANSWERS -- BRUNSWICK 1&2 -85/09/23-K-E BROCKMAN ANSWER 7.07 (1.00)
. :a.: Service Building (0.5)
~
b. Training Building (0.5)
REFERENCE
$ EIH: 63EP-EIP-061-0 BSEP: R.E.Pr. Sections 5.3, 5.4 i
ANSWER- 7 08 (1.00) d i REFERENCE BSEP: .GP-02, p9
ANSWER 7.09 (1.50)
~
, The system has not undergone any significant maintenance or l aodification (0.5) and there.is not sufficient reason-to believe
the status'has changed significantly since the last valve line-up performance (0.5) and the last line-up has been completed within the last refuel cycle (0.25). All control room controls and I indications.(for-this system) are checked to ensure proper line-up (0.25).
I REFERENCE BSEP: GP-01, 3.2.1.1 & 3.2.1.2 ANSWER 7.10 (1.00) l 1) CRD Pressure cannot be restored (with either pump), AND (0.5)
2) Reactor Presure < 800 psig (0.5)
REFERENCE BSEP*-AOP.02.1,-p'3 1
i s
. ,_ ,, .-...,, , -. . , - - . , , , , . . . - . . . , . - . . . . , . . .,. ..- , n.-., , , ,
l
7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36
~
~~~~R dEf6E6Efd5L Ed TREE"~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN ANSWER 7.11 (1.00)
EPH # 1 - ATWS (0.5)
EPM i 2 - Successful SCRAM (has occurred) (0.5)
REFERENCE BSEP: E0P-01/UGr Para 7.Or p 23 ANSWER 7.12 (1.00)
These instruments have a cold reference leg (0.4) and are uncompensated (0.1)
They are not affected until D/W temperature e:<c e e d s the RPV saturation temperature (0.5)
REFERENCE BSEP: E0P-01/UG, CAUTION 4 6, p 36 ANSWER 7.13 (1.00)
a. 1
b. 3 (0.5 each)
REFERENCE BSEP: E E R C Manual, Section 6.5.3, p 40 ANSWER 7.14 (1.00)
a. Normal Up (Full up) (0.5)
b. 45 des (+- 15 deg) (0.25)
Clockwise (East) (0.25)
REFERENCE BSEP: FH-ilt Section 4.0, p3
8. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 37 ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKHAN ANSWER 8.01 (1.00) d REFERENCE EIH: U2 TSr 3.3.1, 3.3.2 BSEP: U2 TS, 3.3.1, 3.3.2 ANSWER 8.02 ( .50) 5 (Refueling)
REFERENCE BSEP: GP-05, p 25; GP-06, pp 4, 6: TS, Section 1.0 ANSWER 8.03 (1.00)
Thermal expansion (of valve internals upon heatup) may cause valve binding and/or damage.
REFERENCE BSEP: DI-13, Section 4.0.7, p i ANSWER 8.04 (1.00) a REFERENCE BSEP: 0I-01 ANEWER 8.05 (1.00)
a. He must enter the reason for the non-performance on the Completion / Exception Form. (0.5)
b. He must, tentativelyr reschedule the PT, in RED, on the Rescheduling PT Sheet. (0.5) i l
- . . . . - - . . . .=
8. ADMINISTRATIVE PROCEDURES,-CONDITIONS, AND LIMITATIONS PAGE 38 ANSWERS -- BRUNSWICK 1&2- -85/09/23-K E BROCKMAN
' REFERENCE BSEP: 0I-13, Section 1.1, p 1 ANSWER' 8.06- (1.00)
, a. The physical phenomena is that void reactivity feedback due to a pressurization transient can add positive reactivity at a faster rate than'the Control Rods can add negative reactivity. (0.5)
b. A OLMCPR restriction. (0.5) 4
. REFERENCE BSEP: TS 3.0.4,-3.2.3.1, & 3.3.6.2 ANSWER 8.07 (1.00)-
~
1. Th'e Mode Switch must be locked in the REFUEL position
2. All the fuel assemblies in the 2 x 2 cell containing the Control Rod to be withdrawn must be removed.
3. 2 SRM's must be OPERABLE 4.' SDM requirements must-be satisfied 5.- All other Control Rods-must be inserted (2 0 0.5 each) i REFERENCE BSEP: TS, 3/4.9.10.2 ANSWER '8. 0 8 ' (1.00) b
+
. REFERENCE BSEP: OI-13, Sect 4.1.1, p2 e
I 1
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8. ~ ADMINISTRATIVE' PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE- 39 ANSWERS - BRdNSWICK 112 -85/09/23-K E BROCKMAN
' ANSWER- 8.09 (1.00)-
1) Processin3 a-Ststion Clearance (0.5)
2) Rackins In/Out 4 4160 Vac Breaker (0.5)
REFERENCE BSEP:'AI-58, Sect 4.1.3, p 7-AN3WER 8.10- (1.00) a REFERENCE-BSEP: TS 3/4 11.1.1 ANSWER 8.11 (1.00) b-
" REFERENCE BSEP: OI-03, Sect 1.1, p 1 ANSWER- lB.12 (1.00)-
The dampers are marked for proper position with the red peint.
REFERENCE BSEP: DI-08, Item i 22, p3 ANSWER 8.13' (1.00) t d
REFERENCE BSEP:-DI-10,. Sect D.1, p-4 j
8.. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 40 ANSWERS -- BRUNSWICK 1&2 -85/09/23-K E BROCKMAN 1 ANSWER 8.14 (1.00) d i
~
REFERENCE BSEP: AI-58, Sect 3.1.6, p3 4
ANSWER 8.15 (1.00) d REFERENCE BSEP: TS 3.0.3, 3.3.7, & 3.7.4 ANSWER 8.16 (1.00)
i. 1) Prior to reactor criticality after completing CORE ALTERATIONS-that could have affected the control rod drive coupling integrity.
i 2) Anytime the control rod-is withdrawn to the ' FULL DU1' position i in subsequent operation.
I 3) Following maintenance on or modification to the control rod or control rod drive system which could have affected the control rod drive couplins. integrity. ( 2 0.0.5 each)
REFERENCE
BSEP: TS 3.1.3.6 ANSWER 8.17 (1.00) .
a. 0.38 (%) (0,5) *
b. The difference between the calculated value of the maximum core reactivity during the operating cycle'(0.25) and calculated 1
BOC reactivity. (0.25) (0.5)'
REFERENCE BSEP: TS, Section B-3/4-1-1 i
f l
l 4
i i
. . _ . _ . , _ . . . . _ _ . _ . _ . _ . , , _ . , _ . . . . . _ . . ~ . _ . . , _ - . _ _ _ _ _ _ _ _ ,_
TEST CROSS F.EFERENCE PAGE 1 -
.OUESTION ' VALUE REFERENCE 05.01 1.00' KEB0000068 05.02- 1.00 KEB0000159.
05.03 2.00 - KEB0000211 05.04 2.00 KEB0000213 05.05' 1.00 - KEB0000219 05.06 1.400 KEB0000285 05.07 1.00 KE90000336 05.08 1.00. KEB0000337 05.09 1.00 . KEB0000395 05.10 2.00. . KE80000398-05.11 3.00 - KEB0000399 105.12 1.00. KEB0000338.
17.00 06.01-- 2.00 MEB0000224 06.02 1.00 KEB0000282
' 06.03 2.00 KEB0000373 06.04- 1.00 KEB0000374 06.05 1.00 KEB0000377 06.06- 2 00 KEB0000383 06.07 1.00 - KEB0000384 06.08 1.00 KEB0000385 06.09 .50 KEB0000390.
06.10 1.50 KEB0000391 06.11- 1.00 KEB0000394 06.12 2.00 - KEB0000397 06.13 1.00 ' KEB0000402
'06.14 1.' 0 0 - - KEB0000403-18.00 07.01 1.00 KEB0000228 07.02' 1.50' KEB0000340 07.03 1.50 KE80000341 07.04' 1.50 KEB0000343 07.05 ~1.00 KEB0000344 07.06 2.50 KEB0000345
- 07.07 .1.00 KEB0000346-07.08- 1.00 KEB0000347 07.09 1.50 MEB0000348 07.10 1.00 KEB0000357 07.11 1.00 KEB0000358 i 07.12 1.00 KED0000359
-07.13 1.00f KEB0000369 07.14 1.00 MEB0000370 17.50-08.01 1.00 KEB0000293 L
t
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7-TEST CROSS REFERENCE PAGE 2 GUESTION VALUE REFERENCE 08.02 .50 KEB0000342 08.03 1.00 KEB0000349 08.04 1.00 KEB0000350 08.05 1.00 KE80000351 08.06 1.00 KEB0000353 08.07 1.00 KEB0000356 08.08 1.00 KEB0000360 08.09 1.00 KEB0000361 08.10 1.00 KEB0000362 08.11 1.00 KEB0000363 08.12 1.00 KEB0000364 08.13 1.00 KEB0000365 08.14 1.00 KEB0000366 08.15 1.00 KEB0000371 08.16 1.00 KEB0000372 08.17 1.00 KEB0000396 16.50 69.00 t
see. is0. 3 TIE 1505 1mIIT 1 SATE 1-11-79 .
energBIODIC MSS CO M PERFORMANCE LOG ***
to 11 12 CMwr 1779.
6 7 8 9 600.6 LOCATICII 1 2 3 4 5 1.16 1.15 0.91 0.52 cwe 1.01 1.12 1.11 1.11 1.11 1.08 1.13 CtFCF 0.7; ARIAL REL PWR 0.60 1.09 0.92 1.02 0.42 0.92 1.02 0.92 1.09 1.01 CHFLFO 0.M 3805011 REL fWe 1.15 0.61 1.09 0.91 1.13 1.14 1.27 CHPF 2.F alus est Pts 0.95 0.96 0.98 atees caF . 0.99 0.95 0.97 CP.E0 0.2; 8 9 Caso 0.1.
l 4 5 6 7 ass 10s 2 3 0.715 CA04 0.1
1 0.723 0.602 0.724 0.715 0.635 serLcta 0.715 0.634 0 .7 15 9-36 19-42 43-36 Cavr 0.3 43-18 11-34 33-34 41-34 37.3' LOC 9-18 19-12 0.0960 0.0966 0.0960 CAPO l
0.0960 0.0947 0.0984 0.0947 0.0
FLCIf 0.0960 0.0966 1.45 1.51 1.29 1.51 CnD 1.29 1.51 1.45 1.23 0.688 Cas1H 2.
PEF 1.51 0.525 0.662 0.688 0.613 0.649 0.613 0.684 0.661 43-36-20 rn 981
satre 13-32-17 33-34-19 39-22-17 9-33-20 25 5 Loc 9-18-20 .25 5 43-18-20 2.00; 2.25 Drc-M 12.0 2.16 1.75 2.16 2.25 12.0 PKF1. 2.25 2.00 2.25 0.226 0.187 0.226 Drc-c 0.226 0.219 0.172 0.219 sea E4 0.226 0.187 .
9-36 25-44 43-36 nut 36.e 43-14 11 34 33-34 41-34 19.
Loc 9-ia 25-10 0.0947 0.0944 0.0947 0.0960 0.0983 0.0960' Dits 7.
Ftow 0.0960 0.0943 0.0960
1.51 1.32 1.51 WFW 1.51 1.45 1.23 1.45 25.
Petr 1.58 . 1.32
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. s* se esseed TRIP CHANNEL 'A2' FOR NEUTRON MONITOR SYSTEM
! FIGURE # 224 B i
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____________________7_____.___________-__.
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ERROR 4 sesso i 3 .g _
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-8 7 % F LOW IV REGULATION 4 es T U n e188E _ < i CVREG SPEED 's _ IV RE3 g SPEED LOAD REJECT ' '
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e v r WALVE ofesasso PRMSURE SET , % Flow d
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lCLOSED -
i gypagg PRESSURE VALVE 7;,
DERAAND LOW VACUUtd SasALL gypagg CLOSE JACK a AS FIGURE # 374 ELECTRO-HYDRAULIC CONTROL LOGIC
. . . ......ia .
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!'l. pleure e z33 Recirculation Mow Contrel
j NOTE.
l SUPPRES$10N CHAMBER AIR SPACE TEMPERATURE IS DETERMINED BY:
UNIT ONE ONLY UNIT TWO ONLY THE AVERAGE OF POINTS 17,18,19 AND THE AVERAGE OF POINTS 2 AND 3 ON CAC-(
20 ON CAC-TR-1258 E THE AVERAGE TR-4426-1 AND POINTS 2 AND 3 ON CAC-OF COMPUTER POINTS WlO6, WlO7, T R-44 26-2 M THE AVERAGE OF COMPUTER D115 AND Wil6. POINTS WlO6,WlO7,Wil5 AND Wil6.
300 1
280 -
\
>t 260 .
E 240 1 i
g 220 UNSAFE D
b-
<[
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c.
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('
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h WE Z 140 o
! Z o 120
$ 100 -
fr -
l 5 80 60 -
40 I O 10 20 30 40 50 60 70 80 90 SUPPR E SSION CHAMBER PRESSURE (PSIG)
DRYW ELL SPRAY (
INITI ATION PRESSURE LIMIT BSEP/VOL. VI/EOP-01/UG FIGURE # 344 Rev. 1 l 9
y b
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i 1
ENCLOSURE 4 1
( REQUALIFICATION PROGRAM EVALUATION REPORT Facility: B'eunswick Steam Electric Plant
~ i I
Examiners: K. Broca 7a'n , J. Munro, and L. Wiens v
Dates of Evaluation: ' September 23-26, 1985
\
Areas Evaluated: X hritten Oral X S imu' a to r Written Examination
1. Evaluation of examination: N/A _
2. Evaluation of facility examination administration: N/A
3. If NRC examination was substituted for facility examination (or sections thereof), attach examination summary sheet to this form.
Summary sheet attached X :
4. Evaluation of examination grading: N/A OrakExamination r
1. Overall etaluation N/A
2. Number observed N/A 1 Number conducted N/A Simulator Examination
> t 4
1. Overall evaluation N/A '
2. Number observed N/A $ umber conduc ed 15
,.> s Overall Program Evaluation \ i Satisfactory X Marginal Unsatisfactory o (
Written Exam: SRO-9/11 = 82% R0-4/4 = 100% TOTAL-13/15 r 87%
Simulator: SR0-9/11 = 82% R0-4/4 = 100% TOTAL-13/15 = 87%
Overall:
SRO-8/11 = 73% R0-4/4 = 100% TOTAL-12/15 = 80%
~,.' ,
Submitted: '3 Forwarded:
Approved:
[hfj j '
WEx ami tier , , _ Sectirr Chief Branch Chief 4 > Lt ' l' i
l
[ t
f. ' i.,..

ML20205F829

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