ML20212N349

Exam Rept 50-456/OL-86-02 on 860716,0804-08 & 11-15.Exam results:11 of 16 Senior Reactor operators,7 of 8 Reactor Operators & 1 Retake Senior Reactor Operator Passed Exam

ML20212N349
Person / Time
Site:	Braidwood
Issue date:	08/18/1986
From:	Burdick T, Hemkming W, Hemming W, Higgins R, Riedinger T, Riedlinger T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20212N343	List: ML20212N340 ML20212N349
References
50-456-OL-86-02, 50-456-OL-86-2, NUDOCS 8608280193
Download: ML20212N349 (129)

Text

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c U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-456/0L-86-02 Docket No. 50-456 Licensee: Comonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690 Facility Name: Braidwood Nuclear Power Station Examination Administered At: Braidwood Nuclear Power Station <

Examination Conducted: July 16, 1986, and the weeks of August 4 and '

August 11, 1986 Examiners: W Date' k fgW% 1 R. L. Higgins Date mm g [

Dg/te Approved By: 6. M. Burdick, Chief Operator Licensing Section Date' Exemi_ nation Sumary Examination administered on July _16, 1936,- ~ ~ ~ -and the weeks of pgus_t_11-13, IV(6]Lep_ortT 50-43,6/0l -8'6-02T~~~~~~~~~~_ August 4-8 and Examinations were administered to 16 senior reactc.r operators, 8 reactor operators, and 1 retake senior reactor operator.

Results: All but 5 of 16 senior reactor operators passed the examination, all but 1 of 8 reactor operators passed the examination, and 1 retake senior reactor operator passed the examination.

8608280193 860826 PDR ADOCK 05000456 V PDR

REPORT 0ET_ AILS _

1. Examiners j W. Her:ning, INEL R. Higgins, Regici. III T. Reidinger, Region III

2. Examination Review Meeting Utility connents and their resolutions are attached to this report.

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3. E_xit Meeting J

a. On August 13, 1986, an exit meeting was held. The following personnel were present at this meeting:

i Eugene E. Fitzpatrick, Braidwood Station Manager James H. Harris, Production Training Manager - Simulator Kenneth Gerling, Production Training Simulator Operations Supervisor (PWR)

Kevin Bartes, Training Supervisor Charles Schroeder, Operation Superintendent Robert Ungeran, Unit 1 Operation Engineer Thomas M. Tongue, NRC Senio~r Resident Thomas E. Taylor, !!RC Senior Resident Timothy D. Reidinger, NRC Operator Licensing Examiner

b. The examiner discussed the following points with the utility:

(1) A major generic weakness noted was the consistent inability of the candidates to basically relate to any of the major electrical annunciators and alarms and associate their significance during normal power operations and abr.ormal evclutions.

(2) Tre candidates in general exhibited lack of operational feniliarity with the hot shutdown panel, i.e., not accustomed te reacting to malfunctions on hot shutdown panel.

(3) The reactor operators and senior reactor operators demonstrated a lack of awareness of Technical Specifications related to equipment out of service and related action statements associated with safety instrumentation out of service.

(4) The chief examiner expressed a concern relating to the senior reactor operators routinely operating switches while at the senior reactor operator position during normal simulator

/ operations. The senior reactor operator possibly could lose the awareness of the total integrated plant picture if he is concerned with operating switches. i 1 i 2

1 (5) The reactor operators and senior reactor operators exhibited a tendency to " outrun" the verification steps in the emergency procedures. The reactor operator once directed to perform a ,

valve isolation per emergency procedure would acknowledge to the senior reactor operator that the step had been completed, even though the step was not completed, and the senior. reactor operator did not verify that the step was completed. The senior reactor operator was reading the steps without the required verification of completion of those steps.

(6) The chief examiner expressed a concern _that the candidates when in a " red path" on " Response to Nuclear Power Generation /ATWS Procedure" (BWFR-S.1) appeared to digrcss from that red path procedure prior to being authorized to leave the procedure. The candidates would appear to use the Reactor Trip / Safety Injection procedure in parallel with the red path (BWFR-S.1) procedure.

There were no other major generic weaknesses noted during simulator / oral examinations. The candidates in general gave a strong performance during t

oral examinations in areas of reactor theory, and systems in general.

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,. BRAIDWOOD REACTOR OPERATOR FACILITY REVIEW COMMENTS

/ QUESTION 1.02 , (2.00)

Indicate whether the following will cause the differential rod worth of one control rod to INCREASE, DECREASE or have NO EFFECT.

c. Boron concentration is DECREASED 1 NRC ANSWER 1.02 (2.00)

c. Increase [0.50 each]

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BRAIDWOOD CONTENTION:

1.02 c. While the answer given is correct, partial credit should be given .

I for an answer which explains that as Boron concentration decreases, there is less spectrum hardening, so fewer neutrons are absorbed in the epithermal region, making differential rod worth decrease.

Reference:

Westinghouse Large PWR Care Control Pages 2-40 through 2-47 Resolution The contention statement is true, in that this is an effect; but the overall effect of boron concentration decrease ,1s an increase in rod worth. If a candidate states the spectrum hardening effect, he will receive 1/2 credit.

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• O QUESTION 1.04. _ (2.50)

b. How does each of the following affect the Fuel Temperature Coefficient (More negative, less negative, or no effect)? [1.5]

No explanation is oesired or required.

1. Accumulation of Xenon and Krypton gases in the fuel to clad gap.

2. Increase in the amount of fuel to clad contact._

3. Buildup of PU239 over core life. -

NRC ANSWER 1.04 (2.50)

b. 1. More negative

2. Less negative

3. More negative [0.5 each]

F BRAIDWOOD CONTENTION:

1.04 b.1 The answer should be less negative Explanation: ns the gap is polleted, the resistance to heat transfer will increase causing a resultant increase in fuel temperature. As fuel temperature increases, the Fuel Temperature Coefficient will become less negative. J

Reference:

Westinghouse Large PWR Core Control

1) Chapter 2, pages 2-40 to 2-49

2) Graph on page 2-46 b.2 The answer should be more negative.

Explanation: As the amount of fuel and clad contact (clad creep) increases, the resistance to heat transfer decreases causing a resultant decrease in fuel temperature. As fuel temperature decreases, the Fuel Temperature ~

Coefficient will bscome more negative.

Reference:

Westinghouse Large PWR Core Control

1) Chapter 2, pages 2-40 to 2-49

2) Graph page 2-46 ,

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QUESTION 1.04 _ (2.50)

BRAIDWOOD CONTENTION (cont):

b.3 Should be deleted from the examination.

Explanation: The question asks about the effect on Fuel Temperature Coefficient (FTC) of the buildup of Pu-239 over core life. The answer key states that FTC becomes more negativa. In actuality, it is the buildup of Pu-240 (not 239) which causes FTC to become more negative over core life. The buildup of Pu-239 is a result of the depletion of U-238, [

and the depletion of U-238 is the factor which will tend to cause FTC to become less negative over core life. Therefore,'either the answer key is incorrect or the question has a typographical error in asking for the isotope 239 vice 240. '

Reference:

Westinghouse Large PWR Core Control Pages 2-40 through 2-47 Resolution ,

Answers b.1 and b.2 were changed accordingly to the referenced document.

Answer b.3 was changed to accept either no effect or less negative for the reasons as stated in the contention.

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QUESTION 1.09 _ _ (1.00)

Part of the reactor thermal safety limit is based upon not allowing saturation conditions at the core hot leg. State the reasoning behind this.

NRC ANSWER 1.09 (1.00) -

(If saturation conditions were allowed to exist at the hot leg) further increases in core heat output would be undetected by the hot leg RTD [0.5] and protection teuld be degraded.

, BRAIDWOOD CONTENTION:

1.09 A discussion of the use of RCS Delta T as an indication of Reactor Power and the fact that if RCS Hot Leg temperature reaches saturation that Delta T is no longer indicative of actual i Reactor Power should be accepted as an alternative answer.

Resolution Agree with comment. The question was graded accordingly.

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QUESTION 1.10 (3.00)

What is the most significant type of heat transfer (conduction, convection, or radiation) taking place under each of the following conditions? Consider each condition separately.

b. Accident condition in whicn coolant is boiled and converted to steam in the reactor vessel.

NRC ANSWER 1.10 (3.00)

b. Radiation / convection (large Delta T)

[0.60 each]

BRAIDWOOD CONTENTION:

1.10 b. Answer key lists both Radiation and Convection as the Most significant form of heat transfer. Recommend accepting either answer as correct.

RESOLUTION:

Either answer is accepted for part b.

QUESTION 1.13.b.2 -

Answer should be decreases; the same question appears on the Byron R0 examination: per telecon.

RESOLUTION:

The answer is decreases and was changed on the Braidwood Answer Key.

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QUESTION 1.14 and 5.08 (1.50)

Use the steam tables and associated Mollier chart to answer the questions below, label quantities with proper units.

a. During cooldown and depressurization, you are required to remain 50 degrees F subcooled. As the pressure decreases through 2085 psig, what is the maximum Tavg allowed (nearest degree F)?

b. Steam is leaking from a pipe flange into a room. A thermocouple (TC) placed in the leakage stream reads 400 degrees F. How many degrees of superheat is this?

c. If the thermocouple in part b. had read 360 degrees F, and the steam pressure inside the pipe was 560 psia, what would you estimate the steam temperature to be at that pressure?

NRC ANSWER 1.14 (1.50)

a. 592 - 593 degrees F (dependirs on how round-off is done),

b. 198 degrees F of superheat per superheat tables.

c. 500 degrees F ,

BRAIDWOOD CONTENTION:

1.14 The answer key is very specific in' the values obtained from the .'

Mollier Diagram and the Steam Tables. During License Operator Training, a band of acceptable answers are normally orovided for any calculation-type questions. Further, license candidates are encouraged to show their work and, thereby, provide the mechanisms used to obtain the answer provided. Should the mechanism be presented correctly, credit is normally given.

Question 1.14 (and 5.08) should be graded based on the mechanisms used by the applicant and the answer key should include a range of permissible values. Recommend that a band of 450 to 520*F be acceptable. -

RESOLUTION:

A band of 185 to 190 degrees F accepted for part b, and 490-510 degrees F accepted for part c; cannot accept temperature below this because it cannot be assumed to be a saturated system.

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QUESTION 2.01 ,_

(1.00) 1 List the interlock requirements that must be satisfied in order to l close a Reactor Coolant Pump breaker. l NRC ANSWER 2.01 (1.00)

1. Oil Lift pump discharge pressure at least 600 psig .

2. Loop Isolation valve interlock [0.5 each]

BRAIDWOOD CONTENTION:

2.01 Or.e additional requirement to close a RCP breaker is tnst the oil lift pump be running. This should not be marked as an incorrect answer.

Reference:

BwCP RC-2, page 2 Resolution .

The facility supplied answer is rot an interlock and, therefore, is not required for credit. It will not be counted as a wrong answer, however, the original two answers are required for full credit. ,

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QUESTION 2.02 (.50)

TRUE or FALSE 7 i+N OPERATING RCP in en RCS loop will trip if the associated Icop Isolation Valve Interlock logic is not satisfied.

NRC ANS:JER 2.02 (.50)

TRUE BRAIDWOOD CONTENTION:

2.02 Answer could be FALSE.

Explanation: If the candidate stated the assumption that the 1000 bypass valve was open, then the RCP ,

will not trip Referer.ce: Drawing 20E-1-4030 RC 22 Resolution If the candidate states the above assumption with a falso answer, .

credit will be awarded.

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QUESTION 2.03 , _

(2.00)

TRUE or FALSE?

The following concern the Pressurizer Power Operated Relief Valves (PORV). ,

b. Pressurizer PORV's are required for overpressure protection during low temperature water solid operations.

ARC ANSWER 2.03 (2.00)

b. TRUE BRAIDWOOD CONTENTION:

2.03 b. While TRUE is the answer given, FALSE is also an acceptable answer based on Technical Specifications 3.4.9.3 (page 4-39),

which states that at least one of the following overpressure protection systems shall be operable:

a. Two residual heat removal (RHR) suction relief valves each with a Setpoint of 450 psig + 1%, or

b. Two power-operated relief valves (PORVs) with lift setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or

c. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2 square inches.

.' - l APPLICABILITY: Modes 4 and 5, and Mode 6 with the reactor vessel head on.

Resoluticn Because part b could be true or false depending upon conditions not provided in the question, the question is deleted from the exam. The point values for section 2 and the total are adjusted accordingly.

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f QUESTION 2.04 (3.00)

Refer to Figure 15 (attached), "CVCS Flow Diagram" For each number on the figure, provide the appropriate information on your answer page, for the following:

5. _ _ _ _ _ F (divert setpoint)

8. _ _ _ _ _ GPM per pump ur total

11. _ _ _ _ _ F NRC ANSWER 2.04 (3.00)

5. 138

8. 32 or 8 per RCP

11. 500 [0.25 each]

BRAIDWOOD CONTENTION:

2.04 5. Answer key is in error. The actual setpoint should read 133 F based on the PLS document (page 62) and the current revision of Braidwood System Descriptions (page 15a-21).

Answer key is correct per the text, however, we request that

8. .

a flow of 6-13 gpm be an acceptable answer based on Bw0P RC-2 (prerequisites).

11. Answer key is in error. It states that charging return temperature is 500*F. According to the system description chapter Iba (page 15a-35), this value should read approximately 518'F. We request that a band of 500-540*F be

'an acceptable answer based on simulator response.

RESOLUTION: -

5. Either 133 or 138 will be accepted for full credit.

8. 6 ,13 will be accepted.

11. 518 plus or minus 18 will be accepted because the diagram in system description states 500 and the verbiage states 518 degrees F.

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i QUESTION 2.06 &ND. 6.02 (2.00)

Concerning BTR$, state the maximum Dilution AND Boration rates (in ppm /hr) for both BOL AND EOL conditions.

NRC ANSWER 2.06 (2.00)

Boration: 80L - 40 ppm /hr EOL - 20 ppm /hr [0.5 each]

Dilution: BOL - 20 ppm /hr EOL - l'O ppm /hr [0.5 each]

BRAIDWOOD CONTENTIONS:

2.06 Object to this question due to the fact that operators use curves to determine system capacity when performing operations. Minimum and maximum design capabilities of the .

BTRS system should not be required as memorization numbers for the operator. A more appropriate question would have been "How and why do Boration/Cilution rates vary from BOL to EOL using BTRS?", since this better evaluates the operator's knowledge of the system. Recommend that credit be given for proper relationship and trends of Boration/Diluticn rates from BOL to EOL. ,

Resolution To receive full credit, candidate must provide exact numbers. Partial credit (1.0 points) will be given if candidate states that, both,

a. Bo. ration rate limits are twice as large as dilution AND

b. BOL rate limits are twice as large as E0L limits.

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i QUESTION 2.07 ,AND 6.03 (2.00)

The followirq concern valves in the Residual Heat Removal System.

a. State the FOUR conditions that must be satisfied in order to open valves 8701A and 8702A, RHR Suction Isolation Valves from RCS loops. [1.0] (Interlo~cks not administrative)

c. State the TWO requirements that must be present in order for valve 8811A, Suction Valve from the Containment Sump, to open automatically.

[0.5]

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. NRC ANSWER 2.07 (2.00)

a. 1. 8812 A closed

2. 8804 A closed

3. 8811 A closed

4. RCS Pressure J -360 psig [0.25 each]

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(0 pen signal from MCB)

c. (MCB switch in Auto)

1. "S" signal present ,

2. 2/4 Lo-Lo levels in RWST [0.25 each]

BRAIDWOOD CONTENTION:

2.07 a. The question asks conditions that must be satisfied to open valves 8701A and 8702A. These valves are in separate trains; 8701A is in train A of-RHR and 8702A is in train B of RHR. .-

The answer key assumes that they are in the same train.

Credit should not be denied if the candidate has listed the conditions required for both trains.

Credit should also be given for proper valve names and not just valve numbers. Credit should not be taken off if only valve names are used and not valve numbers.

c. ' Answer key has 2/4 logic on lo-lo RWST level. The question did not ask for coincidences. Therefore, full credit should be given without requiring coincidence. .

Resolution Parts a and c comments noted and graded accordingly.

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QUESTION 2.10 AND 6.04 (3.00)

a. Follewing a reactor trip, an "Overcrank" alarm is received on the 18 fuxiliary Feedpump. List the sequence of events that occurred to receive this alarm. ~[2.0]

NRC ANSWER 2.10 (3.00)

a. 1. Engine received on AUTO START SIGNAL

2. Starting motors engaged and CRANKED FOR 5 SEC

3. Engine DID NOT ACHIEVE 350 RPM IN 5 SEC

4. 10 SECOND TIME DELAY ACTIVATED

5. ANOTHER 5 SECOND CRANK ATTEMPTED

6. STARTING CYCLE ATTEMPTED 4 TIMES

[0.25 for each item: 0.5 for proper sequence]

BRAIC68000 CONTENTION:

2.10 a. Comments from the examinees indicate that there was much confusion during the examination in regards to this question. Since it is now impossible for us to know exactly what was requested in the question, recommend that this portion of the question be deleted from the examination. If '

not deleted, we request that credit. be received for any ,'

reasonable answer that describes the "Overcrank" feature of the Diesel-driven Auxiliary Feedwater Pump.

Resolution The question was graded as follows:

Credit was given for the following:

0.5 points for knowing maximum of 4 cranks (starts) .

0.33 points for knowing automatic strrt signal received 0.59 points for knowing that there is a 5 sec. crank and,10 second rest O.58 points for knowing there is an alternating crank-wa *t sequence The possible confusion was eliminated during the exam by eliminating the word six from the request for six events.

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QUESTION 2.11 ,AND 6.05 (3.50)

With RCS pressure starting at Normal Operating Pressure, describe each of the ECCS water injection system's response as pressure decreases to atmospheric, during a LOCA Safety Injection. Assume ALL components are operable and/or running. Include in you answer:

b. 1. The DESIGN flowrate (gpm) and associated pressure, AND the MAXINUM flowrate (gpm) and associated pressure.

OR

2. The MAXIMUM amount of water (gal.) INJECTED and associated pressure. -

NRC ANSWER 2.11 (3.500

3. Residual Heat Removal Pumps; b. 6000 gpm (5000 each) 0165 psig 10000 gpa (5000 each) 0 125 psig

[0.5 each]

4. Accumulators: b. 28,000 Gals. (approximately 7000 each ) 0 ,

approximately 625 psig

[0.5]

f BRAIDWOOD CONTENTION:

2.11 b.3. Answer key is in error. The design flowrate of the RHR pumps is 3,000 gpm each at 165 psig. Reference System Descriptions ,'

page 58-27.

It is also requested that full credit be given on this question if only the flowrate of one train is given vice the total flow of each system.

b.4. The answer key should allow for an acceptable bank for each accumulator based on Technical Specification 3/4.5.1 and System Descriptions Chapter 58, page 26.

602-647 psig -

6995-7217 gal each Resolution l

Comments noted and changes made to answer key and graded accordingly.  !

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QUESTION 3.01 , (3.00)

a. What is the meaning of the term "2/4" when indicated on a logic diagram? [1.00]

b. What is the purpose of the Shunt Trip in a Reactor Trip Breaker?

When is it energized? [1.5]

NRC ANSWER 3.01 (3.00)

a. (It meant that it) will require 2 of the 4 possible inputs to activate the particular function. [1.0] .

b. To insure the Reactor Trip Breaker opens if the UV coil fails to open it. [0.75] It is energized by use of the manual trip switch. [0.75]

BRAIDWOOD CONTENTION:

3.01 b. The second part of the answer is in error. A revision to the system has been made such that the shunt trip will be

. energized by use of the manual trip switch or by any automatic trip from SSPS. This issue was covered with the students and the System Description has been revised. .

Reference:

Page 60A-16 Resolution

" Automatic trip signals" was added to the answer key and graded accordingly.

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QUESTION 3.04 . _ (2.50)

a. One of the selected Pressurizer Pressure Channel signals passes through Proportional Integral ("PI") controller. State FOUR pressurizer components that are operated by this signal (be specific). [1.0]

NRC ANSWER 3.04 (2.50)

a. 1. PORV 455A

2. All Ba.ck-up Heaters

3. Variable Heaters

4. #1 and #2 spray valves [0.25 each]

BRAIDWOOD CONTENTION:

3.04 a. Answer key is correct, however, an additional correct answer

• would be " alarm bistables", i.e., Pressurizer Pressure High.

Reference:

14-13a Resolution Alarm Bistables are not considered ~ pressurizer components.

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QUESTION 3.05 (2.00) >

The reactor is at 100% power with normal letdown and charging flow.

Charging flow is manually reduced to minimum and left in manual, no-other changes are made. List the sequence of SEVEN events that t will take place ending in a trip or SI. Be specific; no setpoints required.

NRC ANSWER 3.05 (2.00)

6. Variable Heaters re-energize [0.1]

BRAIDWOOD CONTENTION:

3.05 5. Answer key should read Backup Heaters vice Variable Heaters.

Reference:

Figure 14-15 (high deviation alarm)

Resolution Answer key changed to reflect correct nomencl.ature.

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QUESTION 3.06 (2.50)

One function of the Main Feed Pump D/P program ensures that sufficient feedwater pressure exists, despite the effects of increasing steam pressure the opening of the main feed regulating valve, during the design Iced rejection from 100% power,

a. State TWO other functions of the Feed Pump D/P Program. [0.5]

c. 1. List the THREE input signals that are used by the Main Feed Pump D/P Program control system. [0.75]

NRC ANSWER 3.06 (2.50)

a. Maintains FRV in linear flow region.

Minimizes wear on FRV. [2.50]

c. 1. Main Steam Pressure, PT 507 Main Feedwater Pressure, PT 508 Steam Flow (Total) [2.50]

1 BRAIDWOCD CONTENTION:

3.06 a. Another acceptable answer should be:

Reduce Main FW pump power requirements at reduced loads. ,'

Reference:

Braidwood System Descriptions Page 37b-5, Section B c-1 Credit should be given if the candidate did not specifically list the transmitter numbers. PT-507 can be referred to as Main Steam Line Header Pressure. PT-508 can be referred to as Main Feedwater Header Pressure. l Resolution Part a: Third possible answer will be accepted as or.e of the two required.

Part C-1: Credit will be allowed for either number or name of transmitters.

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QUESTION 3.08 __ (3.00)

Refer to Figure 33-1 attached, " Power Range Channel 41-44". On your answer sheet, state the label for each arrow point, on the figure, assigned a number (1-18). Include name, coincidence and setpoint (if applicable).

NRC ANSWER 3.08 (3.00)

5. Power Range High Flux Rate (Positive) 1/4 + /-5%/2 sec [0.15]

6. Power Range High Flux Rate (Negative) [0.15]

i (5 & 6 INTERCHANGEABLE)-

BRAIDWOOD CONTENTION:

3.08 5 & 6 The logic for the PR rate trips is 2 out of 4, not 1 out of four. An alarm is generated with the 1 out of 4 coincidence.

Reference:

System Description page 33-60.

Resolution Correction made to the answer key. ,-

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(

. QUESTION 3.09 ,

(3.50)

a. Briefly describe how the Reactor Vessel Level Indicating System detects a vessel level change. [1.5]

b. State FOUR inputs for the Subcooled Margin Monitor. Consider separate redundant transmitters of the same parameter as ONE input. [2.0]

NRC ANSWER 3.09 (3.50)

a. The basic principle of operation is the detection of a Delta-T' between adjacent heated and unheated thermoccuples [0.5]. The RVLIS sensor consists of a (Chromel-Alumel) TC near a heater and another (Chromel-Alumel) TC positioned away from the heater. In a fluid with relatively good heat. transfer properties, the Delta-T between adjacent TC's is small. [0.5] In a fluid with relatively poor heat transfer properties, the Delta-T betsmen the TC's is large. When a TC is uncovered, the Delta-T is large and the RVLIS indicates the level change. [0.5] ,

4. Representative CET Temperature (from CET processing) [0.5]

BRAIDWOOD CONTENTION: ,

3.09 a. Flexibility in answers should be considered. The question asks to "briefly describe". The answer key gives an in-depth explanation of RVLIS. Credit should be given for a description of the operation of RVLIS including how the Delta j T changes as the sensor is uncovered. )

b.4. Answer could also be stated as the average of the ten highest CET's.

Resolution -

Part a: Comment noted and answers graded in accordance with answer key and facility comment?. .

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Part b.4: ' Ten highest CET's is equivalent to representative. ,

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QUESTION 4.02 , _ ,

(2.00 State TWO RCS conditions that allow positive reactivity to be added without having the Shutdown Banks withdrawn.

NRC ANSWER 4.02 (2.00)

1. RCS borated to Hot, Xenon Free concentration and maintained at Hot Standby. [1.0]

2. RCS borated to Cold Shut. lown concentratior.. [1.0]

BRAIDWOOD CONTENTION:

4.02 The key implies thac there are only two acceptable answers to the question. A third acceptable answer should be:

9 If the shutdown banks cannot be withdrawn, che Reactor Coolant must be borated as conditions require and the baron concentration confirmed by sampling.

Referenca: BwGP 100-2 Revision 1, page 3.

Resolution: -

According to the new revision, the extra supplied answer is accepted as one of the two required.

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QUE* TION 4.05 _ _ (3.00)

The following concern BWAP 300-1, Conduct of Operations.

a. As Unit 1 NSO, what constitutes "at the controls". [1.0]

b. What must be done if the NSO must leave the "at the controls" l area during non-emergency conditions? Does this also include l going behind the Main Control Board for a valve manipulation or reading? [0.75]

1 s l NRC ANSWER 4.05 (3,00 ,

a. In line of sight of MCB front panels (so to be able to initiate prompt corrective actions when necessary). [1.0]

b. Obtain relief from a qualified operator. [0.50]

Yes. [0.25]

BRAIDWOOD CONTENTION:

4.05 3. Answer is correct. May also be answered by referring to the main control room layout drawing in BwAP 300-1 which includes the center desk and the fire protection status panel. ,

Reference:

BwAP 300-1A1

b. The question is very broad. Other acceptable answers should include any of the following:

1) Licensed operator must be specifically assigned the respor.sibility of monitoring the controls of the unattended unit. ,

2) This same operator must remain within line of sight of ~

the unit's front panels.

3) The licensed operator must (on a periodic basis) review the status of the unattended unit from within the "at-the-controls" area.

Reference:

Braid 4 cod's SRO examination Question 8.03b. <

B6.AP 300-1 page 10, Revision 51 Resolution:

The 4.05 q'testion deals with non-emargency conditions whereas the 8.03 question on the SR0 exam deals with an emergency at the other unit, therefore, the facility contention does not apply to this question.

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QUESTION 4.07- --

(2.50)

The following concern SWAP 380-1, Green Board Concept Control Panels.

a. What % power, above which, is the Main Control Board and Remote ,

Shutdown Panels Green Board configuration based on? [0.5]

NRC ANS'aER 4.07 (2.50)

a. 29% power [0.5]

BRAI:) WOOD CONTENTION:

4.07 a. A more correct answer is 30% power.

Reference:

BwAP 380-1 page 1 Resolution:

Correction made to the answer key.

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QUESTION 5.09 , _ (1.50)

The reactor is producing 100% rated thermal power at a core delta T of 42 degrees and a mass flow rate of 1000 when a blackout occurs.

Natural circulation is established and core delta T goes to 28 degrees. If decay heat is 2%, what is the core % mass flow rate?

NRC ANSWER 5.09 '(1.50)

Q = m Cp (delta T) 2% = m (28/42)

.02 = m (.67)

.02/.67 = .03 or 3%

)

BRAIDWOOD CuNTENTION:

5.09 Recommend deleting the question. )

Explanation: There is insufficient information given in the problem to elicit an answer from the applicant.

During forced circulation conditions, D = &cpAT/., During natural circulation conditions Q is proportional to 4 and toaT2 , .

the natural circulation thumbrules. Since the .

question provided paramsters from a forced circulation condition and compared these parameters to a natural circulation condition, the candidate could not determine the new mass flowrate by either set of proportionalities, neither forced circulation nor natural circulation. Because this question cannot logically be solved, it should be deleted.

Reference:

Thermal Hydraulic Principles Chapter 14, pages 20-26.

RESOLUTION:

Assuming Cp is constant and solving for Cp the equation becomes:

Cp=h/r'n(Th-Tc)

Equating the two conditions:

Q1/Ml(Th-Tc)1=h2/$2(Th-Tc)2 And solving for $2, a candidate can calculate the new flow rate to be 3%. ,

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QUESTION 5.11 _ _ (3.00)

During a startup the reactor is subcritical at 3000 CPS on the Source Range Instruments when a steam dump valve fails open,

b. How will the transient and final conditions differ if the transient in part "a" happened at EOL as compared to BOL7 l

'l Explain any differences.

NRC ANSWER 5.11 (3.00)

b. Power increase RATE is higher at EOL because of changes in Beta-Bar (MTC). [0.5] Final power is the same [0.5] but Tave will be higher [0.53 (closer to no-load temperature) because of the larger MTC.

BRAIOWOOD CONTENTION:

5.11 b. The question does not state "No reactor trip occurs". Thus, the answer key does not allow for other possible answers such as a reactor trip at 105 cps. Other reasonable answers should be accepted based on the candidates' assumptions.

RESOLUTION: _

The question asks the candicate to compare the BOL and EOL responses, i.e., how does core age causa a change in response, if the candidate correctly assumed the plant oower goes to the point of adding heat at BOL and assumes a trip occurs at E0L then he still needs to compare the two responses and credit will be awarded accordingly.

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QUESTION 5.15

, Explain why a dropped control rod is worth approximately 200 pcm and a I stuck rod is worth 1000 pcm even though the same rod could be considered in both cases. (Assume no trip.)

NRC answer: When a rod is stuck out with all other rods inserted, the flux profile is higher where the rod is out, therefore, that rod " sees" a much higher flux than average core flux.

(Because rod worth is a function of the relative flux '

difference between the rod and the core average flux, the rod is worth more (about 1000 pcm). (1.0)

If a rod is dropped just the opposite happens. The rod depresses the the flux in the area near the rod relative to the average core flux. (Worth about 200 pcm). (1.0)

Comment: The question states " Assume no trip". The answer assumes a trip. Replace key with:

Rod worth is dependent upon the relative flux the rod sees.

With the rod stuck out it sees a high flux in comparison to -

a dropped rod, which would depress the flux in its vicinity.

Reference:

WCAP 10315 Nuclear Case Design Characteristic's RESOLUTION:

The original answer key 'is,more detailed than the facility supplied answer because of the need to compare that answer with a number of ,

varied responses expected from many candidates. Because of this situation,it remains as is. Full credit would be awarded for an .

answer such as provided by the facility comment above.

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, QUESTION 6.06 (3.00)

a. When is a 2/4 trip logic required to be used in the Sclid State Protection System (SSPS)? [1.0]

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b. What is the purpose of the Shunt Trip in a Reactor Trip Breaker?

When is it energized? [1.5]

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NRC ANSWER 6.06 (3.00)

a. When control and protection are provided by the same parameter. '

[0.5] With a channel failure 2/3 protection is still available.

[0.5]

b. To insure the Reactor Trip Breakar opens if the UV coil fails to open it. [0.75] It is energized by use of the manual trip switch. [0.75]

BRAIDWOOD CONTENTION:

6.06 a. Answer key includes the reason. The question does not ask I for reason. Full credit should be given for answering when.

The applicant need not mention that a 2/3 logic will remain.

b. The second part of the answer is in error. A revision to the -

system has been made such that the shunt trip will be energized by use of the manual trip switch or by any automatic trip from SSPS. This issue was covered with the students and the System Description has been revised.

Reference:

Page 60A-16 RESOLUTION:

Part a. The 2/3 portion was removed from the required response. .

Part b. " Automatic trip signals" was added to the answer key and graded accordingly.

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QUESTION 6.07 and 3.02 (1.50)

The reactor has been shutdown without the reactor trip breakers opening and a manual SI has been initiated. If the SI is no longer required, would the SI signal reset? Explain your answer.

NRC ANSWER 6.07 (1.50)

Yes (0,5): Because the manual signal is only momentary, reset is possible without P-4. (The system, in fact, will return to full automatic operation.) (1.0)

BRAIDWOOD CONTENTION:

6.07 from the sequence of events presented in the question (first the reactor is shutdown with the reactor trip breakers closed and second a manual SI is initiated) it is not clear that the reactor trip breakers will not perform as designed and open with a manual SI. If indeed the breakers remain close4 then the answer key is correct. However, if the breakers open as should be the case from the sequence of events in the question then the system will not return to full automatic action. Credit should be given for both answers.

Reference:

Figure 61-15 (Rev. 0)

RESOLUTI0ft:

If the candidate states that he assumes the reactor trip breakers opened as a result of the manual safety injection and he correctly describes the P-4 logic then credit will be awarded.

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QUESTION 6.09 (3.00)

d. Which one of the belcw compensates the Reactor Control Unit for reactivity changes? [1.0]

1. Variable Gain Unit.

2. Non-Linear Gain Unit.

3. Lead-Lag Compensator.

4. Rod Speed Programmer.

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NRC ANSWER 6.09 (3.00)

d. a. [1.0]

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BRAIDWOOD CONTENTION:

6.09 d. Answer should be 1 (Variable Gain Unit)

Same as question 3.07 d.

RESOLUTION: ,

Typographical error corrected on 6.09.

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QUESTION 7.01 _

(2.00)

a. What are FOUR specific methods / symptoms that can be used, for icentifying the ruptured steam generator, during a steam generator tube rupture accident, in accordance with BwEP-37

b. What arc the TWO conditions that must be monitored during a steam generator tube rupture accident after a RCS cooldown is initiated, that require RCP's to be trioped?

NRC ANSWER 7.01 (2.00)

a. 1. Unexpected rise in any S/G narrow range level.

2. SG elowdown liquid radiation greater than alert alarm setpoint.

3. High activity from any one S/G sample.

4. Main Steamline radiation greater than alert alarm setpoint. '

[1.0] .

b. 1. CC water to RCP lost. (affected pumps only)

2. Phase B cntmt. isolation (1.0]

BRAIOWOOD CONTENTIONS:

7.01 a. Answer key is correct, however it is too limiting as possible responses. Good operators are trained to utilize all indications to diagnose adverse conditions. Per WOG background document E-3, page 4, the following should also be '

acceptable answers.

-Pressurizer level decreases at a rate which is dependent ~

upon size and number of failed tubes.

-RCS pressure decreases

-Pressurizer level continues to decrease

-Mismatch between steam flow and feed flow to affected S/G may be observed i

' -Increasing Temperature / Pressure in the ruptured Steam

• Generator RESOLUTION:

The question clearly stated for the candidate to respond in accordance with BWEP-3. The first tnree answers provided by the facility describe a symptom of a steam generator tube rupture but do not answer the question of identifying a stean generator. The mismatch answer will The in-0270g0pgggedifthecandidatealsostatesnolevelchange. creased Temp./ Pr the BWEP-3.

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QUESTION 7,0L _ (2.00) 1 BRAIDWOOD CONTENTION (cont):

b. The answer gives conditions which are taken directly from the foldout page for BwEP-3 (SGTR). It is our contention that other conditions are also monitored after RCS cooldown which would also required tripping RCP's during a SGTR. Alternate acceptable answers should include:

1. Following RCS depressurization using normal spray, if the spray valve will not close, the RCP supplying flow must be tripped.

2. CCP pump flow > 200 gpm or any SI pump discharge flow and RCS pressure < 1370 psig (1670 psig adverse containment).

Reference:

(1&2): BwEP-3 stop 16.C RNO column, BwEP-3 step 1

3. #1 RCP seal delta P of 200 psid

4. #1 Seal leakoff flow < 0.2 gre

5. VCT pressure < 5 psig

6. RCP bearing temperature > 225 F

Reference:

(3 thru 6): Sw0P RC-2 RESOLUTION:

The spray valve criteria is accepted as a correct answer. The delta P and leak off criteria are also accepted because they appear in the post-cooldown procedure. However, the CCP flow or SI pump flow criteria does not apply because'these are checked prior to the cooldown not af ter the i cooldown begins. The RCP bearing criteria is always in effect not in i

addition to the cooldown as a result of the tube rupture.

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QUESTION 7.02 _ _. (2.00)

What are FOUR plant conditions which place the plant on a RED PATH and requires the operator to utilize the status tree?

i NRC ANSWER 7.02 (2.00)

Nuclear power > 5% , ,

Core exit Tc > 1200 F All SGs <4% narrow range and total Feedwater flow < 500 gpm available.

Tc decrease > 100 F in 60 minutes and RCS cold leg < 246 F. l CTNT press > 50 psig

[4 0 0.5 ea.]

BRPIDWOOD CONTENTION:

7,02 Answer key is correct for Red path Summary Foldout page in BwEP-0. However, the Red Paths are stated differently in the BwST's. The answers may also be stated as: _

Nuclear power > 5%

Core exit TC's 1 1200*F All SG's 1 4% narrow range and total feedwater 1 500 gpm available Tc decrease > 103*F in 60 minutes and RCS cold leg temperature to the left of Limit A. g Containment pressure 1 50 psig

Reference:

BwST's RESOLUTION: ,

Reference verified, answer is acceptable and graded accordingly.

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QUESTION 7.03 , ,

. (2.00)

The following pertain to BwFR-S.1 " Response to Nuclear Power Generation /ATWS".

b. State the TWO entry symptoms or conditions for er,tering BwFP-S.I.

[1.5]

NRC ANSWERS 7.03 (2.00)

b. Entered from:

1. SwEP-0 Reactor Trip or Safety Injection. [0.25]

When Reactor trip is not verified [0.25] and manual trip not effective [0.25].

2. BwEP-1 Subcriticality. CSF on Red (Orange). [0.25]

BRAIDWOOD CONTENTIONS:

7.03 b. Full credit should also be given for an answer that states:

1) BwEP-0 step 1, response not obtained (RNO) and ,

2) BwST-1, Power range level > 5% or increasing, and/or intermediate range SUR positive RESOLUTION:

The additional two answers are added to possible responses and graded accordingly.

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6 QUESTION 7.07 (3.00)

c. If a dropped red cannot be recovered immediately, state the THREE conditions or actions, one of which, is required to be completed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, for power operation to continue. [1.5]

j NRC ANSWER 7.07 (3.00)

c. Within 1 hour:

, 1. Restore rod to oper'able status, [0.3]

2. Rod is declared inoperable [0.3] and other rods in group aligned within +/-12 steps, [0.3]

3. Rod is declared inoperable and

a. Tech. Spec. SDM satisfied [0,3]

b. Power reduced to </=75% [0.3]

BRAIDWX)D CONTENTION:

7.07 c. The answer key specifically requires the answer per technical .

specifications. Credit should also be given for action item steps in Bw0A R00-4, page 2 of 3 steps 1 through 7. ,

The answer key is also incomplete. The Tech Spec contains two additional items.

1) A power distribution map is obtained

2) Perform an accident analyses reevaluation Credit should not be taken off if these items are included in

' the candidates answer.

Reference:

Braidwood Tech Spec 3.1.3.1 ,-

RESOLUTION:

Also accepted from BWOA R00-4 will be: Calculate QPTR and Reduction of Power to 70%.

l The two responses listed in the facility contentions are not required to be done within the one-hour cuideline and, therefore, are not expected responses nor will they be allowed for credit unless time statement is .

included. No credit will be deducted if they are provided in additior to the correct answers.

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QUESTION 7.09 (2.00)

a. State all the conditions (in all Modes) that required Emergency Boration accordino to Sw0A PRI-2.

NRC ANSWER (2.00)

a. 1. Inadequate Shutdown Margin

a. ( 1.3% kd/K Modes 1-4 [0.1]

b. < 1% dk/K Mode 5 [0,1]

c. < 0.95 Keff or < 2000 ppe boron Mode 6 [0.1]

2. Control rods below bank insertion limit. [0.3]

3. Failure of more than 1 control rod to fully '

insert following a reactor trip. [0.3]

4. Unexplaf.ned or uncentrolled reactivity increase [0.3]

5. Uncontrolled cooldown. [0.3]

6. Inability to borate normally. [0.3]

b. 3w0A PRI-11, Uncontrolled dilution. [0.23 BRAIDWOOD CONTENTION:

7.09 b. Question deleted per examiners direction.

RESOLUTION:

No resolucion is necessary for this contention because the examiner deleted this question during the administration of the examination due to the large number of potential answers for or.ly 0.2 points.

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QUESTION 7.11 (3.00)

Unit 1 is in its initial ascension to full power. Power is increases l

from 0-40% at a constant rato over 8 nours. Barik 0 Control Rods move l from 75 steps to 110 steps also at a constant rate. After remaining at 40% for 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />, power is increased to 80% at a cor.stant rate over 8 hoJFs and Bank 0 rods move from 110 steps to full out again also at a constant rate. Pouer remains at 80% for 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> at which time the reactor trips

b. State any rod withdrawal liNits that were violated and where they were violated.

NRC ANSWER 7.11 (3.00)

b. Limit - 3 steps per hour after 50% power when the 3%/ hour rate is applied [0.5]. Violated from 50-LO% power [0.5].

I BRHIDWX)0 CONTENTION:

7.11 b. This question is not applicable to Braidwood Station

Reference:

BwGP 100-3 page 2 RESOLUTION:

Because this limit does not exist at Braidwood the question part b is deleted from the exam and the poin',s adjusted accordingly.

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QUESTION 7.12 . (2.50)

b. According to SwGP 100-2, What are the requirements if the Estimated Critical Control Bank Height is below the Lo-Lo insertion limit?

NRC ANSWER 7.12 (2.50)

b. 1. Insert all rods

2. Recalculate ECC [.5 ea]

BRAIDWOOD CONTEN1 ION:

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7.12 b. Answer key is correct for eight-fold ECC method. Reference SwGP 100-2, page 8. This question could be interpreted as ,

initial ECC. For this case an acceptable answer could be to '

change boron concentration thus adjusting rod height to above lo-lo insertion limit and continue to perform the reactor-startup.

RESOLUTION:

The referenced material,BwGP 100-2, does not support the facility contention of changing boron concentration nor adjusting rod height. The question qcn-cerns the estimated not the actual critical condition. The material supplied with the facility colaments does not address the initial ECC.

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QUESTION 7.13 , _ (2.00) _

The following pertain to Precautions and Limitations found in BwGP 100-1, " Plant Heatup".

b. Would starting an RH pump while using RH letdown with the RCS solid cause an inadvertent RCS pressure INCREASE or DECREASE if CV131 is in AUTO? [0.5]

NRC ANSWER 7.13 (2.00)

b. Decrease [0.5]

BRAIDWOOD CONTENTION:

7.13 b. Another possible answer based on interpretation of the o question as stated could be "Yes". Normally CV131 is placed in manual when starting or stopping a RHR pump.

RESOLUTION:

The qLestion asks;would starting the punp cause an increase or decrease

• in RCS pressure? The more correct way of asking this question is to .

add " choose either increase or decrease".

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QUESTION 8.01 a RESOLUTION:

Because the question did not specify the type of event, i.e. GSEP criteria met or not, reportable or non-reportable, immediately re-portable or not immediately reportable, potentially significant or not potentially significant, normal working hours or backshift; t,he question did not have a specific answer and, therefore, could not be consistently graded. Therefore, the question was deleted from the exam which caused Section 8 to be worth 24 points.

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QUESTION 8.05 -

(2.50)

a. (Accordire to Tech Specs). state the minimum number of personal required for each position below with Unit 1 in Modes 1-4 and Unit 2 Modes 5 - 6 or defueled.

(Place your answers on your answer sheet.) [0.75]

Shift Engineer Shift Foreman Reactor Operator Auxiliary Operator STA or SCRE

b. What is the maximum allowable period for the manning level in part "a" to be below minimum?

What is the maximum number of persons that is allowed to be absent during this period?

State the EXCEPTION to the minimum manning allowance. [0.75] .

NRC ANSWER 8.05 (2.50)

a. 1 1

3 3

1 [0.15 each]

b. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1

During shif t turnover when a crew member is late or absent.

[0.25 each]

1 BPAIDWOOD CONTENTION:

8.05 a. This question had a verbal clarification made during the exam. Not everyone was made aware of "According to technical .-

specifications" for this question. Therefore an acceptable answer could be based on BwAP 320-1, which is more limiting than tech specs. Additionally, the answer key has the chart on 6-5, another acceptable answer should be the chart on 6-Sa since Unit 2 has no fuel / license.

Reference:

Technical Specifications tables 6-1, 6-la

b. The third question to this part should be deleted due to confusion. There is no exception. Only that you can have one less for up to two hours as the previous questions in this part already asked for.

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8.05 a RESOLUTION: '

Either the Technical Specifications or BwAP numbers will be allowed for full credit.

8.05 b RESOLUTION:

Both Technical Specifications and BwAP provide the answer when the minimum manning allowance is not allowed (exception).

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QUESTION 8.08 ' -- (2.50)

c. What is the interval for each of the designators below? [1.03

1. S

2. Z

3. SA NRC ANSWER 8.08 I. 2 . LO)

c. 1. At least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

2. At least once every 92 days

3. At least once every 184 days BRAIDWOOD CONTENTION:

8.08 c. There is no interval "Z".

Reference:

Technical Specifications, Table 1.1, page 1-8.

RESOLUT0N:

Part C.2 is eliminated from grading because of the specified reason provided by the facility.

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QUESTION 8.09 (2.00)

According to BWAP 100-1, Station Director Implementing Procedure, the Station Director has certain responsibilities that CANNOT be delegated . List them.

NRC ANSWER 8.09 (2.00)

1. Declaration of an Unusual Event, Alert, Site Emergency or General Emergency cordition.

2. The decision to notify and recommend protective actions or offsite authorities in the case where a Site Emergency or General Emergency condition exists and the Recovery Manager and/or Corporate Command Center are not prepared to do 50.

[1.0 each]

BRAIDW000 CONTENTION:

8.09 2. The decision to notify and recommend protective actions to offsite authorities should be sufficient for full credit.

Also, on 6/13/86 at an NRC Emergency Preparedness exit ..

meeting here at Braidwood, Mr. Tom Ploski of Region III requested that Braidwood consider expanding the responsibilities that cannot be delegated to four. These would be the two given in the answer key plus:

. 3. Authorization to use Emergency Dose Limits  ;

4. Decision to request assistance from Federal i Emergency Management Agencies This information was provided to our License class as likely additions to our procedures. If the candidate included these in his answer, we request that he receive full credit.

RESOLUTION:

The required answers are acceptable for full credit.

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. o QUESTION 8.11 , (1.50)

a. According to Sw2P 400-1, Technical Support center Implementing Procedure (TSC), when is the activation of the TSC required.

NRC ANSWER 8.11 (1.50)

a. When: directed by the Station Director an alert is declared a Site Emergency is declared a General Emergency is declared [0.25 each]

BRAIDWOOD CONTENTION:

8.11 a. An acceptable answer could also be " Alert or above" in lieu of Alert, Site Emergency, General Emergency.

RESOLUTION:

The facility contention is simply a clarification of the correct answer.

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A i U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: BRAIDWOOD 1 REACTOR TYPE: PWR-WEC4 --

DATE ADMINISTERED: 86/07/16 AP I AN : , . _ __ _ u_ _ _ _

INSTRUCTIONS TO APFLICANT:

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

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the qaestion. The passing drade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY  % OF APPLICANT'S CATEGORY VALUE _IQTAL SCORE VALUE CATEGORY __

b 25.00 25.00 5. THEORY OF NUCLEAR POWER PLANT OPl: RATION, FLUIDS, AND THERMODYNAMICS 25.00 25.00 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

24. 30 PROCEDURES - NORMAL, ABNORMAL,

25. T 25.00 7. .

EMERGENCY AND RADIOLOGICAL CONTROL l

2 L4. 00 25.T 25.00 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS qs.so 15.= 100.00 TOTALS FINAL GRADE  %

All work done on this examination is my own. I have neither civen nor received aid.

APPLICANT'S SIGNATURE i

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the edninistrction of this examinatico the following ruics apply:

1. Cheating on the examination means an attomatic denial of your application

- and could result in more severe penaltias.

2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3. Use black _ ink or dark pencil-o g to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the examination.

5. Fill in the date on the cover sheet of the examination (if necessary).

6. Use only the paper provided for answers.

7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write jon) ge sTde of ,

the paper, and write "Last Page" on the Tast answer sheet.

9. Number each answer as to category and number, for example,1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets free pad and place finished answer sheets face down on your desk or table.

12. Use abbreviations only if they are commonly used in facility literature.

13. The point value for each questica is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain an answer ,

to mathematical problems whether indicated in the question or not.  !

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION ANO 00 NOT LEAVE ANY ANSWER BLANK.

16. If parts of th.e examination are,not clear as to intent, ask questions of l the examineI only.

17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has i been completed.

18. When you complete your examination, you shall:  ;

a. Assemble your examination as follows:

(1) Exam questions on top.

(2) Exam aids - figures, tables, etc. l (3) Answer pages including figures which are a part of the answer.

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b. Turn in your ctpy of the examination and all pages used to answer l the examination questions.

c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.

d. Leave the examination area, as defined by the examiner. If after i leaving, you are found in this area while the examination is still >

in progress, your license may be denied or revoked. .

. 5. THEORY OF NUC WAR POWER PLANT OPERATION. FLUIDS. AND PAGE 2

. THERMODYNAMICS QUESTION 5.01 (1.00)

Which one of the following statements concerning Xenon-135 production and removal is correct?

a. At full power, equilibrium conditions, about half of the Xenon is produced by Iodine decay and the other half is produced as direct fission product.

b. Following a reactor trip from equilibrium conditions, Xenon peaks becausc delayed neutron precursors con'tinue to decay to Xenon while neutron absorption (burnout) has ceased. ,

c. Xenon production and removal increases linearly as power level increases; i.e., the value of 100% equilibrium Xenon is twice that of 50% equilibrium Xenon.

d. At low power levels, Xenon decay is the major removal method. At high power levels, burnout is the major removal method.

QUESTION 5.02 (1.00)

Which one of the following statements concerning SAMARIUM reactivity effects is correct?

a. The equilibrium (at power) value of Samarium is dependent upon power level. The peak value of Samarium following a shutdown is dependent upon power level prior to shutdown.

b. The equilibrium (at power) value of Samarium is dependent upon power level. The peak value of Samarium following a shutdown is f Independent of power level prior to shutdown.  :

c. The equilibrium (at power) value of Samarium is Independent of power level. The peak value of Samarium following a shutdown is dependent I upon power level prior to shutdown.

d. The equilibrium (at power) value of Samarium is Independent of power level. The peak value of Samarium following a shutdown is Independent of power level prior to shutdown.

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

1 .

. 5. THRORY OF NUCf. WAR POWER PLANT OPERATION. FLUIDS. AND PAGE 3 THERMODYNAMICS L QUESTION 5.03 (1.00)

The following statements concern fission product poisons. Complete the ,

statements with the available answers provided below. Place the answers on your answer sheet. [An answer may be used more than once.]

a. It takes about hours to reach the maximum Xenon concentration after a reactor trip.

b. The decay half-life of Xenon 135 is approximately hours.

c. It takes about hours to reach equilibrium Xenon concentration l I

after a step increase from 0 to 50% power.

i d. The decay half-life of. Promethium 149 to Samarium 149 is ap' proximately i hours.

Available Answers':  ;

O hours; 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />; le hours; 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />; 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />; 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. j I

QUESTION 5.04 (1.00) i I

Which one of the following describes the changes to the steam that occur l between the inlet and outlet of a real (not ideal) turbine? j

a. Enthalpy decreases, entropy decreases, quality decreases. l t

b. Enthalpy increases, entropy increases, quality increases. i i

c. Enthalpy constant, entropy decreases, quality decreases. I i

d. Enthalpy decreases, entropy increases, quality decreases. i i

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(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) i i

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. 5. THEORY OF NUC MAR POWER PLANT OPERATION. FLUIDS. AND PAGE 4

. IBFRHQDYNAMICS QUESTION 5.05 (1.00)

Why would the Reactor Protection System become unreliable for DNB protection from the OT delta T trip if voids were allowed to form in the Reactor Coolant System? Choose one of the following.

a. The heat transfer coefficient of the cladding is reduced significantly,

b. The specific heat capacity of the reactor coolant inventory changes when voiding occurs and is not measurable by the RTDs.

c. The critical point of water is reached and is not measurable by the RTDs.

d. Entropy becomes more limiting than enthalpy, which is not within the design considerations of the Reactor Protection System.

QUESTION 5.06 -

(1.00)

The 2200 degrees F maximum peak cladding temperature limit is used because ..... (Choose one of the following.)

a. it is 500 degrees F below the fuel cladding melting point. ,

b. any clad temperature higher than this correlates to a fuel center line temperature at the fuel's uelting point.

c. a zircalloy-water reaction is accelerated at temperatures above 2200 F.

d. the thermal conductivity of zircalloy decreases at temperatures above 2200 F causing high centerline' temperatures.

QUESTION 5.07 (1.50) l A variable speed centrifugal pump is operating at 1/4 rated speed in a closed system with the following parameters:

l Power = 300 Kw Pump delts P = 50 psid Flow = 880 rpm What are the new values for these parameters when the pump speed is increased to full rated speed?

l l

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l

5. THEORY OE_FUCLEAR POWER PLANT OPERATION FLUIDS. AND PAGE 5 1 THERMODYNAMICS l QUESTION 5.08 (1.50)

Use the steam tables and associated Mollier chart to answer the questions below, label quantities with proper units.  ;

a. During cooldown and depressurization, you are required to remain 50 degrees F subcooled. As the pressure decreases through 2085 psig, what is the maximum Tavs allowed (nearest degree F)?

b. Steam is leaking from a pipe flange into a room. A thermocouple (TC) placed in the leakage stream reads 400 degrees F. How many degrees of superheat is this?

c. If the thermocouple in part b had read 360 degrees F, and the steam j pressure inside the pipe was 560 psia, what would you estimate the steam temperature to be at that pressure?

l QUESTION 5.09 (1.50)

The reactor is producing 100% rated thermal power at a core delta T of 42 degrees and a mass flow rate of 100% when a blackout occurs.

Natural circulation is established and core delta T goes to 28 degrees. If decay heat is 2%, what is the core % mass flow rate?

QUESTION 5.10 (3.00)

How are esch of the following parameters affected (INCREASE, DECREASE or NO CHANGE) if one main steam isolation valve c1cses with the plant at 50% load. Assume all controls are in automatic that no .

trip occurs.

1. Affected loop stean generator level (INITIAL change only) ,

2. Affected loop steam generator pressure

3. Affected loop cold leg temperature

4. Unaffected loops steam generator pressure

5. Unaffected loops cold les temperature

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

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. 5. THEORY OF NUCTRAR POWER PLANT OPERATION. FLUIDS. AND PAGE 6 THERMODYNAMICS QUESTION 5.11 (3.00)

During a startup the reactor is subcritical at 3000 CPS on the Source Range Instruments when a steam dump valve fails open.

a. EXPLAIN what happens to reactor power and Tave. Continue your explanation until stable conditions'are reached with no operator action. (Assume the reactor is undermoderated, at BOL and no reactor trip occurs.)

b. How will the transient and final conditions differ if the transient in part "a" happened at EOL as compared to BOL7 Explain any differences.

QUESTION 5.12 (2.50)

a. Of the coefficients that contribute to the power defect, which contributes most to the change of power defect ever core life? EXPLAIN. D.o]

b. Explain why power defect is desireable for reactor operation at power.[bol

c. Of the coefficients that contribute to power defect, which coefficient reacts first to a sudden power change due to rod movement? f o 53 QUESTION 5.13 (1.00)

Explain why, as moderator temperature increases, the magnitude of MTC increases.

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(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

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5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 7

. THERMODYNAMICS QUESTION 5.14 (3.00)

Compare the CALCULATED Estimated Critical Position (ECP) for a startup to be performed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a trip from 100% power, to the ACTUAL critical control rod position if the following events /

conditions occurred. Consider each independently. Limit your answer to ECP is HIGHER THAN, LOWER THAN, or the SAME AS the ACTUAL critical control rod position.

a. The FOURTH coolant pump is started two minutes prior to criticality.

I

b. The startup is delayed until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the trip.

c. The steam dump pressure setpoint is increased to a value just below the Steam Generator PORV setpoint.

d. Condenser vacuum is reduced by 4 inches of Mercury.

e. All Steam Generator levels are rapidly being raised by 5% as criticality is reached.

QUESTION 5.15 (2.00)

Explain why a dropped red is worth approximately 200 pcm and a stuck rod is worth 1000 pcm even though the same rod could be considered in both ,

cases.

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(***** END OF CATEGORY 05 *****)

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 8 ,

QUESTION 6.01 (3.00)

Refer to Figure 15 (attached), "CVCS Flow Diagram".

For each number on the figure, provide the appropriate information on your answer page, for the following:

1. GPM (Normal operating)

2. PSIG

3. F

4. PSIG '

5. F (divert setpoint)

6. GPM (maximum allowable for each kind) l J

7. GPM

8. GPM p* 9-e o To4 \

9. GPM

10. GPM

, 11. F

12. GPM v.e 9-e o W QUESTION 6.02 (2.00)

Concerning BTRS, state the maximum Dilution AND Boration rates (in ppm /hr) for both BOL AND EOL conditions.

QUESTION 6.03 (2.00)

The following concern valves in the Residual Heat Removal System.

a. State the four conditions that must be satisfied in order to open valves 8701A and 8702A, RER Suction Isolation Valves from RCS loops. [1.0) L 4e som e s,n.A A basbu ve- CowaWaas

b. State the TWO signals that will close these same valves. [0.5)

c. State the TWO requirements that must be present in order for valve 8811A, Suction Valve from the Containment Sump, to open automatically.

[0.5]

- 1

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

. 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 9 l

QUESTION 6.04 (3.00)

a. Following a reactor trip, an "Overcrank" alarm is received on the 1B Auxiliary Feedpump. List the sequence of DGE; events that occurred to receive this alarm. [2.0]

\

b. Other than "Overcrank", list FOUR other conditions that will trip and lockout the 1B Auxiliary Feedpump. (Setpointe not required.) [1.0)

QUESTION 6.05 (3.50)

I With RCS pressure starting at Normal Operating Pressure, describe each of the ECCS water injection system's response as pressure decreases l to atmospheric, during a LOCA Safety Injection. Assume ALL components are  ;

operable and/or running. Include in your answer: l f

a. The NAME of the system, AND i

b. 1. The DESIGN flowrate [gpm] and associated pressure, AND The MAXIMUM flowrate [ gym].and associated pressure.  :

OR  ;

2. The MAXIMUM amount [ gal.] of water INJECTED and associated {

pressure.

i i QUESTION 6.06 (3.00) {

a.. When is a 2/4 trip logic required to be used in the Solid State Protection System (SSPS)? [1.0]  !

b. What is the purpose of the Shunt Trip in a Reactor Trip Breaker?

• When is it energized? [1.5]  :

c. TRUE or FALSE 7 Both Reactor Trip Bypass Breakers can be racked in at the same time,  !

but only one mey be closed. [0.5]

QUESTION 6.07 (1.50) g The reactor has been shutdown without the reactor trip breakers opening and i a manual SI has been initiated. If the SI is no longer required, would i the SI signal reset? Explain your answer. l

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) f i

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE le QUESTION 6.08 (2.00)

The following concern the Remote Shutdown Panels.

TRUE or FALSE?

a. The MCB vill-to-lock feature is overridden when operation is from the Remote Shutdown Panels,

b. Reactor Coolant Pumps cannot be started from the Remote Shutdown Panels.

c. If local control of the MSIV is taken at the Remote Shutdown Panels, no Control Room alarm will sound.

d. Reactor Containment Fan Coolers (RCFC's) can be controlled from the Remote Shutdown Panels in high speed only. .

QUESTION 6.09 (3.00)

a. State the inputs that are used to generate the Power Mismatch signal in the Reactor Control Unit. [0.5]

b. State the purpose of the Summing Unit in the Reactor Control Unit.

[1.0] ,

I

c. The Summing Unit can only function using temperature signals.

In what system component is the Power Mismatch signal converted to a temperature signal? [0.5]

d. Which one of the below compensates the Reactor Control Unit for reactivity Changes? [1.0]  ;

l

1. Variable Gain Unit.

2. Non-Linear Gain Unit.

3. Lead-Lag Compensator. l

4. Rod Speed Programmer.  ;

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(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

. ._, D

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 11 QUESTION 6.10 (2.00)

For each type of radiation monitor, list the MAJOR type of detector used (G-M Tube, ion 6hamber, scintillation etc...) and MAJOR radiation type detected (alpha, beta, gamma etc...).

a. Area Monitors,

b. Gaseous.

c. Particulate (Gas streams).

d. Iodine (Gas streams).

4

(***** END Or CATEGORY 06 *****)

7. PROCEDUDM - NORMAL. ABNORMAL. EMERGENCY AND PAGE 12 RADIOLOGICAL CONTROL QUESTION 7.01 '

(2.00) op- ed

a. Whe.tareFOURspecificjmethods/symptomsthatcanbeused, for identifying the 6e=66ed steam generator, during a steam generator tube rupture accident, in accordance with BwEP-37

b. What are the TWO conditions that must be monitored during a steam generator tube rupture accident after a RCS cooldown is initiated, that require RCP's to be tripped?

QUESTION 7.02 (2.00)

What are FOUR plant conditions which place the plant'on a RED PATH and requires the operator to utilize the status tree?

QUESTION 7.03 (2.00)

The following pertain to BwFR-S.1 " Response to Nuclear Power Generation /ATWS".

l a. Why is manual SI actuation not advisable during performance of EwFR-S.17 [1.0]

b. State the TWO entry symptoms'or conditions for entering BwFR-S.1.

M Do3 QUESTION 7.04 (1.50) ,

The following pertain to issuance and use of Type 1 and Type 2 RWPs.  !

a. State the Shift Engineer's responsibility for Type 2 RWPs PRIOR l to any work signing in on the RWP. [0.2]  ;

b. State the FOUR reasons that a Shift Engineer may use to terminate  !

an RWP. [0.8)

c. For how long is a Type 2 RWP valid? [0.25]

d. State the whole body equivalent dose, greater than which, a Type 2 RWP is required. [0.25]

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(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

4

-. . - - - . - . , , . . - . , - . . , . . - . - . - , y _ - - . , _ , . . , . . --,,,,,.,.._,.m -

7. PROCEDURES - NORMAL. ABNOEliAL. EMERGENCY AND PAGE 13 RADIOLOGICAL CONTROL l

QUESTION 7.05 (1.00) '

l l

Which one of the following statements, concerning Toch Spec actions i required for Nuclear Instrument malfunctions, is correct? l

a. If a source range channel fails while a startup is in progress I and reactor power is below P-6, insert all control banks to l zero steps.

b. If an intermediate range channel fails while a startup is in progress and reactor power is above P-6 but below P-10, the power increase may continue using the operable intermediate range channel,

c. Failure of one power range channel during shutdown precludes

. reactor startup until the failed channel is returned to operable status,

d. Failure of both source range channeI>s while shutdown requires shutdown margin requirements to be verified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

QUESTION 7.06 (2.00)

The following pertain to use of the Emergency Procedure (BwEP, BwST, BwFR, BwFS) Network. Assume an emergency situation exists.

a. When is the initial scan of the Critical Safety Function Status Trees performed? [0.5]

b. A BwFR is being performed after an ORANGE condition was identified.

If a higher sequence priority GRANGE condition is identified during the evolution, what actions should be taken by the operator? [0.5]

c. Which of the following procedures may be entered directly?

(Without being entered from another procedure.) Note: More than one procedure may be correct. [1.0]

1. BwEP-0

2. BwEP-3

3. BwCA-0.0

4. BwCA-1.1 l

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

. 7. PROCEDU m - NORMAL. ABNORMAL. EMERGENCY AND PAGE 14 RADIOLOGICAL CONTROL QUESTION 7.07 (3.00) ,

i

a. During use of BwOA ROD-3 " Stuck or Misaligned Rod", under l what conditions must Attachment B, " Alignment of Bank to l Rod", be used? [1.0] j

b. During the performance of BwOA ROD-4, " Dropped Rod Recovery",

prior to recovery of the dropped rod, how is Tref balanced with Tave? [0.5]

c. If a dropped rod cannot be recovered immediately, state the THREE conditions or actions, one of which, is required to be completed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, for power operation to continue. [1.5]

QUESTION 7.08 (1.50)

The following concern operation of #1 Seal Bypass Valve (CV8142) on the Reactor Coolant Pump.

a. If during performance of BwOA RCP-1, "RCP Seal Failure", and the

1. 1 Seal Bypass Valve needs to be opened, what THREE conditions

> must exist before opening #1 Seal Bypass Valve? [1.0)

b. If during performance of BwOA RCP-2, " Loss cf Seal Inject 1on", what TWO conditions must exist such that the #1 Seal Bypass Valve must be

-closed? [0.5]

(i . s o)

QUESTION 7.09 (2.00)

)(. State All the conditions (in all Modes) that require Emergency Boration according to BwOA PRI-2. , ,

b. Frs= uh:t precedure is M A PRI-2 enter:d7 QUESTION 7.10 ( .50)

True or False?

According to BwGP 100-5, " Plant Shutdown and Cooldown", cooldown of the RCS '

CANNOT b'e initiated while boration to the Cold Shutdown Xenon-free Boron Concentration is in progress.

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

7. PROCEDU m - NORMAL. ABNORMAL. EMERGENCY AND PAGE 15 RADIOLOGICAL CONTROL QUESTION 7.11 {3.00) i Unit 2 is in its initial ascension to full power. Power is increased from 0-40% at a constant rate over 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />'s. Bank D Control Rods move from 75 steps to 110 steps also at a constant rate. After remaining at 40% for 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />, power is increased to 80% at a constant rate over 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and Bank D rods mov: from 110 steps to full out e.sain also at a constant rate.

. Power remains at 80% for 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> at which time the reactor trips,

a. What fuel conditioning power increase limit was violated and at what point in the above scenario was it violated? C1.5]

n.c.h

b. State any rod withdrawal rates 3that were violated and where they were violated. to.s3

c. After the trip, to what power level may the reactor return without any fuel conditioning limits applying? How is this new level determined, i.e. what is the basis for the new level? El.o1 QUESTION 7.12 (2.50)

a. What are the THREE requirements according to BwGP 100-2, " Plant Startup", if criticality is achieved with the control bank below the Lo-Lo insertion limit? Ei.53

b. According to BwGP 100-2, What are the requirements if the Estimated Critical Control Bank Height is below the Lo-Lo insertion limit? D o3 e

QUESTION 7.13 (2.00)

The following pertain to Precautions and Limitations found in BwGP 100-1,

" Plant Heatup".

a. WHAT is the maximum pressure and temperature that should be maintained in the RCS when the RH System is in service? [0.5]

b. Would starting an RH pump while using RH letdown with the RCS solid cause an inadvertent RCS pressure INCREASE or DECREASE if CV131 is in AUTO 7 [0.5]

c. All Reactor Coolant Pumps and RH pumps may be deenergir,ed during Mode 5 operation providing two conditions are met and maintained. State these TWO conditions. [1.0]

(***** END OF CATEGORY 07 *****)

. 8. ADMINISTRATIVE PROCEDURRR. CONDITIONS. AND LIMITATIONS PAGE 16 8.01 fs2.53,

s. s o)

QUESTION ,

Referring to BwAP 1250-5, Potentially Significant Event.

c. St:t: the duties of the Shift Euninwor.L.e!

a. b . State the duties of the Station on-call Duty Person (Duty Operating Engineer). Be specific if discussing notifications including person notified and reason for notification.tio1 b .tr. If agreement on the evaluation of the event is not obtainable, who must be contacted for final event classification? to.s3 i

QUESTION 8.02 (2.00)

BwAP 300-1, Conduct of Operations, contains four actions required to be taken by the Shift Engineer on a reactor trip. State 'the four actions.

QUESTION 8.03 (2.00) .

A situation has arisen that calls for one unit's NSO to leave the "at the controls" area of his " stable and under control" reactor to help with an emergency on the other unit.

a. Who must approve this action? [0.5] l

b. If the decision is made to allow one NSO to assist the other, what THREE compensatory actions must be taken on the " stable -

and under control" unit? [1.5] ,

i QUESTION 8.04 (2.50) l

a. Who can approve operating outside of the Technical Specifications, l

^

procedures, or operating orders?

b. The decision to operate outside of established guidelines in part "a" should be done only to solve an immediate problem in order to prevent what three things?

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(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

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8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 17 t

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+

QUESTION 8.05 (2.50) l s neemauj

• 7ea-we sp e Ae.4, .: l 3r. State the minimum number of personnel required for each position below with Unit 1 in Modes 1-4 and Unit 2 in Modes 5 - 6 or defueled.

(Place your answers on your answer sheet.) [0.75]  ;

Shift Engineer  !

. Shift Foreman Reactor Operator  !

1 Auxiliary Operator  ;

STA or SCRE  ;

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b. What is the maximum allowable period for the mannsna level in part "a"  !

to be below minimum?

What is the maximum number of persons that is allowed to be absent [

during this period?  ;

State the EXCEPTION to the minimum manning allowance. [0.75] l

?

c. During Modes 1-4, if the Shift Engineer is to be absent from  !

the Control Room, what must be done to ensure continuity of control? [

t If he/she is to be absent for Modes 6 or 6, what must be done to ensure continuity of control? [1.0] {

f QUESTION 8.06 (3.00) ,

l According to Technical Specifications:  ;

r

a. what are the exemptions from the RWP issuance requirements during the l performance of their duties in a High Radiation Area? ,

i

b. what must be done for areas accessible to personnel with radiation levels greater than 1000 mr/hr? l

c. what must be done for individual high radiation areas accessible  !

to personnel with radiation levels greater than 1000 mr/hr that  !

are located within large open areas, where the entire area is  !

not a high radiation area?  ;

?

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(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) f

8. ADMINISTRATIVE PROCEDURER. CONDITIONS. AND LIMITATIONS PAGE 18 QUESTION 8.07 (2.50) i

a. What action (s) must be taken if a failure places the unit  !

outside of any applicsble LCO's? Be specific as to any l applicable time limits.f3.o3 l I

b. If, 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the discovery in part "a".-the unit is in  ;

such a condition that operation is allowed under a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action Statement, how much time is available in the newly t entered action requirements? Do.53 i

QUESTION 8.08 (2.50) I

a. Surveillance requirements must be performed within specified time intervals with specified maximums. State all the maximums allowed. [1.0]

What must be done if the surveillance requirements for a piece  :

b.

of equipment is not performed within the specified time intervals? l

[0.5]

c. What is the interval for each of the designators below? [1.0]

l

1. S r

2. 2  :

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2. S . SA f QUESTION 8.09 (2.00)  ;

According to BwZP 100-1, Station Director Implementing Procedure, the Station Director has certain responsibilities that CANNOT be delegated.

List them. ,

i QUESTION 8.10 (2.00)

The following concern BwZP 300-1, Initial Notification and GSEP Response. f

a. What is the time limit for notification of off-site authorities? [0.5]

v I

b. When does the clock start for notification of off site authorities?  ;

[0.5]  ;

i

c. If the off-site authority wants verification of authenticity of the notification, what action is to be taken? What information is not given? [1.0]

f

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) ,

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8. PAGE 19 ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS QUESTION 8.11 (1.50)

a. According to Bw2P 400-1, Technical Support Center Implementing Procedure (TSC), when is the activation of the TSC required.

I

b. Per BwZP 400-1, what must happen before the TSC can be activated.

(***** END OF CATEGORY 08 *****)

(************* END OF EXAMINATION ***************)

4 EQUATION SHEET m

Cycle efficiency = (Net work f = ae v = s/t

_ cut)/(Energy in) 2 w = m; ,

s = V,t

• 1/2 at .

~

5 = x- 2 A = 13 A = A3 e '*

.<E = 1/2 mv .a = (Vf - 13 )/t FE = mgn

• = e/t x = &n2/t1/2 = 0.693/t1/2 Vf = V,

• at 2 :1/28 "

• U *T

((c /2) * (; )J

.g , , j 20 A= 4 1

i = 931 m

-Dc m=V gAo ,

Q = mCsat I = Ioe'#

d = UAa r I*I o 10.xf77L p e = a ,4h -

in.= 1.3/u s HVL = 0.693/n l P = o,. 7 10,ur(t) 7 = 7 e' '

SUR = 25.06/T IG

• II#I ~ #eff}

G, = 5/(1 - K,ffx)

CRz (1 - K ,ff3) = G 2 (I ~ eff2}

SUR = 25s/t' * ($ - o)T M = 1/(1 - K,ff) = CRj /G 3 r = ( t*/o ) + ((a - o '/ Io]

M = (1 - K ,ff,)/(1 - K ,ff3) 7 v(o - a)

SCM = ( - K ,ff)/K ,ff T = (a - s)/(Is) t' = 10 seconos a = (K,f,-1)/K,ff = M,ff/K,ff I = 0.1 seconds

-I o = ((1*/(T K,ff }] * [T,ff (1 + IT)]

/

Idli*IdId I;d) 2 =232?,

P = (uV)/(3 x 1010) 2

?/hr e (0.5 CE)/c (meters)

• *N ~

R/hr = 6 CE/d2 (feet) ,

Miscellaneous C:nversiens

' dater 9ar wetam I curie = 3.7 x 10 10 dos 1 gal. = 8.345 lem. I kg = 2.21 lem 1ga}.=3.7811 tars Ino=2.54x103Stu/nr 1 f. = 7.48 gal. 1 = = 3.41 x 100 5tu/hr Oensity = 62.4 les/ft 3 lin = 2.54 cm Gensity = 1 gm/c9. *F = 9/5'C + 32 )

Heat of vacorization = 970 Stu/lem 'C = 5/9 ('F-32)

Heat of fusion = 144 Scu/ lcm 1 STU = 778 ft-lbf 1 Atm = 14.7 pst = 29.9 in. Hg. .

l 1 ft. H 2O = 0.4335 luf/in.

3

~~

o{

s-, C- mn _

N U}

E

]

g_. -e .

e-= =-

L -

s d 5

"a W

e

_ _ ii . 1 t

. e, (H :

D

- =

@e f .-e 0-X s a

w s.,

M b[ N S Ge

- n .,

Q 0--@D V lt a N

. y

~ 14 -

c -

"" a O_..

. b. -

l (E)1 -

u v

@('j @

FIGURE 15 a-19 CHEMICAL AND VOLUME CONTROL SYSTEM FLOW BALANCE

5. THEORY OF NUCMAR POWER PLANT OPERATION. FLUIDS. AND PAGE 23

.. TEERMODYNAMICS 9

ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

rD Y 7"'

ANSWER 5.01 (1.00) Hi. : Li k Vdl I i

D.

REFERENCE Westinghouse Reactor Physics, pp. I-5.63-76.

BBR, Reactor Theory, Sessions 38 and 39.

DPC, Fundamentals of Nuclear Reactor Engineering,Section VI.

001/000-K5.39 (3.5/4.1)

ANSWER 5.02 (1.00)

C. ,

REFERENCE Westinghouse NTO Core Physics, pp. I-5.77-79.

DPC, Fundamentals of Nuclear Reactor Engineeria.2, p. 170 001/000-K5.13 (3.7/4.0)

ANSWER 5.03 (1.00)

a. 5 (or 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />).

b. 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />,

c. 50 for 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />).

d. 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. [0.25 each]

REFERENCE Westingouse NTO Nuclear Physics, pp. I-5.70-79. i i

ANSWER 5.04 (1.00) t d

REFERENCE MNS OP-SS-HT-2, p.12. ,

BWD, Westinghouse Thermal-Hydraulic Principles for PWR, Ch.7.

l . - _. . _ .

PAGE 21

5. THEORY OF NUCr. WAR POWER' PLANT OPERATION. FLUIDS. AND THERMODYEAMICS -

ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 5.05 (1 54) b REFERENCE MNS Thermo, para. 2.6.

BWD, Westinghouse Thermal-Hydraulic Principles for PWR, Ch. 2 & 13. .

?

ANSWER 5.06 (1.00)  ;

e REFERENCE

\

MNS Thermo-Core Performance, p.2.

• BWD, Westinghouse Thermal-Hydraulic Principles for PWR, Ch. 13. i ANSWER 5.07 (1.50) i 3

Power (2) = Power (1)(N2/N1) cubed = 300x(4) = 19.2 Hw l' 2 2 Delta P(2) = Delta P(1) (N2/N1) = 50x(4) = 800 psid Flow (2) = Flow (1) (N2/N1) = 800x(4) = 3520 gym (0.5 each]  ;

REFERENCE GPNT Vol. III, Ch. 2, Sect. H, p. 2-234.  !

ROWE Reactor Operator Training Manual, Sec. 2, pp 49-50  :

BWD, Westinghouse Thermal-Hysraulic Principles for PWR, Ch. 10. l ANSWER 5.08 (1.50)

a. 592 - 593 degrees F (depending on how round-off is done).  ;

b. 186 degrees F of superheat per superheat tables. j s15-AO

c. fHiPJ degrees F. [0.50 each]

q qo- sio REFERENCE ROWE, Steam Tables and Mollier chart 1 - - - - - --- ._. ____ _ ._

5. THEORY OF NUC MAR POWER PLANT OPERATION. FLUIDS. AND PAGE 22 ,

. THERMODYNAMICS ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

i ANSWER 5.09 -

(1.50) .t Q: m Cp (delta T) 2% = m (28/42) [

.02 = m (.67) >

.02/.67 = .03 or 3% l REFERENCE General Physics, BT & FF, Section 3.2 ROWE Reactor Operator Training Manual, Sec. 2, pp 54-63 '

4 BWD, Westinghouse Thermal-Hydraulic Principles for PWR, Ch. 14  :

J ANSWER 5.10 3.00) 3

[

1. Decrease j

2. Increase  ;

3. Increase i

4. Decrease l

5. Decrease (0.60 each]

REFERENCE 4

BWD, Westinghouse Large PWR Core Control, Ch. 12.  ;

ANSWER 5.11 -

(3.00) j

a. (The excess steam flow causes) Tave to decrease [0.25] which inserts ,

positive reactivity (0.25], and power increases. [0.25] (At the POAH) .

increased power will increase temperature which inserts negative  !

reactivity via the FTC. [0.5] Power will stabilize higher than POAH I

[0.25] and Tave will be lower than the no-load value (minus the number of degrees needed to overcome FTC). [0.25]

l

b. Power increase RATE is higher at EOL because of changes in Beta-Bar (MTC). [0.5] Final power is the same [0.5] but Tave will be i higher [0.5] (closer to no-load temperature) because of the  !

larger MTC. i I

I REFERENCE Millstone Reactor Theory, RT-18.

BWD, Westinghouse Large PWR Core Control, Ch. 2 & 3.

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

PAGE 23

5. THEORY OF NUCr. TAR POWER PLANT OPERATION. FLUIDS. AND

.T_RERMODYNAMICS ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 5.12 (2.50) a.. Moderator Temperature Coefficient (MTC).[0.5] Because boron i boron concentration is reduced. [0.5]  ;

i

b. Power defect has a stabilining influence on reactor operation

~

because it resists power changes. (As power increases, power def ect adds negative reactivity and as power decrea-ses, power .

defect adds positive reactivity). [1.0)

c. Doppler (FTC) [0.5]. ,

REFERENCE Millstone Reactor Theory, RT-13, Pp 6-7 and RT-12. '

BWD, Westinghouse Large PWR Core Control, Ch. 3.

r ANSWER 5.13 (1.00)

The change in water density per degree F increases as as temperature increases.

REFERENCE BWD, Westinghouse Large PWR Core Control, Chapter 2'p. 2-3 to 41, Chaptar 3 ,

p. 3-20 to 23.

ANSWER 5.14 (3.00)

a. SAME AS

b. (ECP) LOWER THAN (ACP)

c. (ECP) LOWER THAN-(ACP)

d. SAME AS ,

e. (ECP) HIGHER THAN (ACP) [0.6 each]

REFERENCE BWD, Westinghouse Large PWR Core Control, Chapter 7 p. 7-24 to 28. I i

PAGE 24

5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND

. TRRRMODYNAMICS

-86/07/16-JAGGAR, F.

ANSWERS -- BRAIDWOOD 1 ANSWER 5.15 (2.00)

When a rod is stuck out with all other rods inserted, the flux profile is higher where the rod is out and therefore that rod " sees" a much higher flux than the average core flux. (Because rod worth is a function of the relative flux difference between the rod and the core average flux, the rod is worth more (about 1000 pcm)). [1.0] ,

If a rod is dropped just the opposite happens. The rod depresses the the flux in the area it is in relative to the rest of the core and so is only worth about 200 pcm. [1.0]

REFERENCE BWD, Westinghouse Large PWR Core Control, Ch. 6.

IP2 Reactor Theory pg 7-69,60.

t i

PAGE 25

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 6.01 (3.00) i

1. 75 5. 1386n) 9. 12

2. 2235 6. 120 Hixed, 75 Cation 10. 55

3. 557-sst 7. 87 11. 500-5%

4. 370!5 8. 32 or 8 per RCP (6 'O 12. 3 per RCP or 12

[0.25 each]

REFERENCE BWD, S.D. CH. 15a., Figure 15a-19, CVCS drawing.

ANSWER 6.02 (2.00)

Boration: BOL - 40 pps/hr 'l2 eve 4 H For EOL - 20 pps/hr i . %.wo. L i ~*

• 2 = 'd b ka- [0.5 each]

ums

'A

• 5 a ' 'i ~;*

Dilution: BOL - 20 pps/hr 2. 3os u-a & =

EOL - 10 ppm /hr [0.5 each]

REFERENCE BWD, SD. CH. 16, Pg. 25 ANSWER 6.03 (2.00)

a. 8812 A/sclosed 8804 Alec losed 8811 Agclosed RCS Pressure </= 360 psig [0.25 each]

(Open signal from MCB) l

b. Close signal from MCB RCS pressure at 662 psig (MCB switch in Auto) [0.25 each]

l

c. (MCB switch in Auto)

"S" signal present (51) e-/+ Lo-Lo levels in RWST [0.25 each]

REFERENCE BWD, S.D. CH. 18, Pgs. 17-18

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 26 6.

ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 6.04 (3.00)

a. 1. Engine received an AUTO START SIGNALEo 3G

2. Starting motors engaged ar.d CRANKED FOR 11'73 a' WM

3. (Engine DID NOT ACHIEVE 350 RPM IN 5 SEC)5  % e <emSECV-< Sha 5'c h s

4. 10 SECOND TIME DELAY ACTIVATED. %c.auwa%=%

cv= *u res4-

""^~"" " ~ ^OND - n na t- im

6. STARTING CYCLE ATTEMPTED 4 TIMES.T.o.s1

[0.25 for ::ch ite=; 0.5 for pre;cr : qacace]

b. 1. High Water Temperature bos*O

2. Low Oil Pressure bo pr,d

3. Overspeed 69oo e4

4. (Low-Low) Pump Suction Pressure (4 4S" S ved [0.25 each]

REFERENCE BWD, S.D. CH. 26, Pgs. 16-17 ANSWER 6.05 (3.50) {tio.j,go, $ p_ .3 psq u =twn , = cce eW 3

a. 1. Centrifugal Charging Pumps: b. 300 gpm (150 each) 9 2500 psig 1100 gpm (550 each) 9 600 psig

[0.5 each]

2. Safety Injection Pumps: b. 800 gpm (400 each) @ 1200 psig 1300 gym (650 each) @ 800 psig

~

[0.5 each]

3

3. Residual Heat Removal Pumps: b. 6000 gym (5000 each) @ 165 psig 10000 gym (5000 each) 0 125 psig

[0.5 each]

69 s s -nn

4. Accumulators: b. 28,000 Gals.(approximately J000 each) @

(T.S. L',4h u. O*/. Wed*approximately S2S psig [0. 5]

Go ul REFERENCE BWD, S.D. CH. 58, Pgs. 22-27 l

l

PAGE 27

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 6.06 , (3.00)

a. When control and protection are provided by the same paramet. erd 4 [C. 5] 'Jith e c.heLLei f ailurc 2/3 protection is still

.. 4,.w,.

......... r ,m . a 1

b. To insure the Reactor Trip Breaker opens if the UV coil fails to open it. [0.75] It is energized by use of3 the manual trip switch. [0.75] .gg oug%., sp g

c. True [0.5]

REFERENCE BWD,S.D. CH. 60A,' Pgs. 11, 16 ANSWER 6.07 (1.50)

Yes [0.5]: Because the manual signal is only momentary, reset is possible without P-4. (The system, in fact, will return to full automatic operation.) [1.0] It % c u ase sbhes %.b ke a ss-asc.~de r a m . m a s 1. . ,a %

<.qwb<- M9 b=

J =e4.s so.bars op.,.a u ,%.Jui. o,_

9-w W c., A s a b.a~.d.J h c"4W REFERENCE BWD, S.D. CH. 61, Figure 61-17 AriSWER 6.08 (2.00) i

a. TRUE

b. TRUE

c. TRUE

d. TRUE [.5 ea]

REFERENCE BWD, S.D. CH. 62, Pgs. 14-17  ;

i i

f i

i I

PAGE 28

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION

-86/07/16-JAGGAR, F.

ANSWERS -- BRAIDWOOD 1 .

ANSWER 6.09 (3.00)

a. Auctioneered High Nuclear Power Turbine load (Pirnpulse) [0.25 each]

b. Summing Unit (adds three temperature error signals together to) generates a total temperature error for the Rod Speed and Direction Programmer. [1.0]

c. Non-linear Gain Unit [0.5]

d. x. i. [1.0]

REFERENCE BWD,S.D. CH. 28, Pgs. 26-29 ANSWER 6.10 (2.00) raisi

a. G-M, Gammaro.as3 r 24

b. Scintillation, Betalo.2s1 to.sn

c. Scintillation, Beta La153 ro.sc

d. Scintillation, GammaCo451 [0.5 each]

REFERENCE BWD, S.D. CB. 49, Pg. 17, 62 k

- - - - - - - .~. - - - - . . - - - -

PAGE 29

7. FROCEDUMA - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 7.01 (2.00) g gg

a. 1. Unexpected rise in any S/G narrow range level.

2. SG Blowdown liquid radiation greater than alert alarm setpoint.

3. High activity from any one S/G sample.

4. Main Steamline radiation greater than alert alarm setpoint. [1.0]

s. s w m._ > s...av % wak % s .s c.w% e.

CC water to RCP lost. (affected pumps only) b ** *

b. 1.

4, RC P " I seaA A P t 2oo ps td

2. Phase B entmt. isolation [1.0] 5. A* P

• SM V 4 h P5

3. N. spmy v. We mcw op.m C=trc cw a e4)

REFERENCE [ o . s ,mq BWD BwEP-3 p. 3 & 4; Fold Out Page ANSWER 7.02 (2.00)

Nuclear power >,5% ,

Core exit Tc >,1200 F All SGs (4% narrow range and total Fedwater flow <500 gym available.

Tc decrease > 100 F in 60 minutes and RCS cold leg < 246 F.( +. ten oC ti~t+4)

CTMT press >f 50 psig

[4 @ 0.5 ea.]

REFERENCE Fold Out Pages.

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 30

- RADIOLOGICAL CONTROL ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 7.03 (2.00)

a. A possible loss of heat sink due to main feedwater isolation while the reactor is at power. [1.0]

b. ' Entered from: (Two Rei'd l

1. BwEP-0 IReactor Trip or Safety Injection. [0.251 When Reactor trip is not verified f0-963 and C o . 7 53 tmanual trip not eff ective -[0.25] .fo.an .i on s +.,1 - a wo

2. BwST-1 Suberiticality. CSF on Red (Orange). [0.2E]

09 [ Pown 7 5 % W/ r, ,q , $ we > o }

REFERENCE BWD BwFR-S.1 pp. 1 & 2.

ANSWER 7.04 (1.50)

a. Read, understand, and initial his approval for that date and shift. [0.2]

b. Job Cancellation Job Completion Expiration Changed conditions [0.8]

c. For the length of the job. [0.25]

d. >50 mrem / day. [0.25]

REFERENCE BYN, RP Standards, Pp. 12 - 17 BWD, RP 5tandards, Pp. 14 - 17b ANSWER 7.05 (1.00) d.

REFERENCE BWD, Technical Specification Table 3.3-1

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 31 RADIOLOGICAL CONTROL ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 7.06 (2.00)

a. After departing BwEP-0 unless directed by BwEP-0. [0.5]

b. Suspend the lower priority BwFR and address the higher priority BwFR. [0.5] ,

c. 1 and 3 [1.0]

REFERENCE EwAP 340-1, Pg. 8, 9 WOG, Users Guide, Pg. 14; BwEP-0, Pg. 1; BwCA-0.0, Pg. 1 ANSWER 7.07 (3.00)

a. 1. Misaligned rod is in a controlling bank, AND

2. Misaligned rod is above core bottom but below the rest of the bank. [1.0]

b. By reducing turbine load or diluting. [0.5]

c. (Within 1 hour:) At se=s& s.1 o< 3 = ~1 La o W <s % h <* = '* % #*

1. Restore rod to operable status, f0-3-]

2. Rod is declared inoperable [".3] and other rods in group aligned within +/- 12 ,

steps, feral

3. Rod is declared inoperable and

a. Tech. Spec. SDM satisfied. EO-&]

8. t . Power reduced to </= 75%.(M 43r33 S. C m \c M QYCL

• • P0-3 m Jmpa aP O* A > "

REFERENCE BWD, BwOA ROD-3, Pg. 3 ee s d o.9 , w pacme\y & W h e d *d Bw0A ROD-4, Pg. 2 g , g g .3 Tech. Spec. 3/4.1.3

7. PROCEDU m - NORMAL. ABNORMAL. EMERMNCY AND PAGE 32 o RADIOLOGICAL CONTROL ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

(1.50)

ANSWER 7.08

a. 1. Seal injection flow is 8-13 gpm.E.G

2. #1 Seal Leakoff flow is < 1 gpm.r.11 .

3. RCS Pressure is > 100 psig but < 1000 psig.f,.ul ti-91

b. 1. RCS Pressure is < 100 psig and

2. Seal injection water is not supplied. [0.5]

REFERENCE BWD, BwOA RCP-1, Pg. 2, BwOA RCP-2, Pg. 2 (i. so)

ANSWER 7.09 (2,00)

K. 1. Inadequate Shutdown Margin

a. < 1.3% dk/K Modes 1-4 [0.1]

b. < 1% dk/K Mode 5 [0.1] -

c. f0.95Keffor<2000preboronMode6 '[0.1]

2. Control rods below bank insertion limit. [0.3]

3. Failure of more than 1 control rod to fully insert following a resctor trip. [0.3]

4. Unexplained or uncontrolled reactivity increase. [0.3]

5. Uncontrolled cooldown. [0.3]

6. Inability to borate normally. [0.3]

b. OwOA PRI-11, Uncentrolled dilution. [a 9]

REFERENCE BWD, BwOA PRI-2, Pg. 1 ANSWER 7.10 ( .50)

False REFERENCE BWD, BwGP 100-5, Pg. 2

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 33 RADIOLOGICAL CONTROL ANSWERS -- BRAIDWOOD r -86/07/16-JAGGAR, F.

~

ANSWER 7.11 (3.00)

a. Limit - 3% of full power per hour after 20% power [0.5]. Violated from 20-40%5'5khen 40-80% [0.5].

b. 'Liioit 3 steps per haur after 50% pcu;r when the 3%/heur rete 1: ;;11;d [0.5]. Violated-froa 50-80% pcu;r [0. 5] . u. um vMM Eo.51 c . ao 461% [0. 5] . New level is determined by highest power level achieved for any 72 consecutive hours during any 7 day operating period ~"[0.5]

'O A.

b l h o. L<bhum  ?= ^ b, W P=A=>"a""*"

REFERENCE c. 1 p m., 4., c . 5 g s . % q , to f. .A s %wW BWD, BwGP 100-3, P. 2 &

ANSWER 7.12 (2.50) ro.s:2

a. 1. Emergency Borate a100 ppm.Eo.G i

2. Reinsert LCOAR

3. . Perform all controlotankstos3 to, for Insertion Limit Surveillance. 1 g.5 ee]

Won % s. O

b. 1. Insert all rods

2. Recalculate ECC [.5 ea]

REFERENCE -

BWD, BwGP 100-2 Pg. 8 ANSWER 7.13 (2.00)

a. 350 F, 400 psig [0.25 each]

b. Decrease (0.5]

c. 1. No (operations are permitted that could cause) dilution of the Reactor Coolant System (boron concentration) AND l

2. Core outlet temperature is maintained at d' least 10 F below saturation temperature. [0.5 each]

REFERENCE BWD, BwGP 100-1, Pg. 3, 5; Tech. Spec. 3.4.1.4.2

PAGE 34

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS 5

ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

1. 5 0 ANSWER 8.01 , '2.50)

p. Er-"re timely netificatien cre made per S=.^.P 1250-5T1, Si.- .ificant E tent Nettf-icatic Fle Chart. [1.0] ,

a. V. Screen the eventOS i-4 AA .e**,w)[o.23 [g,33 Notify - Nuclear Division on-call Duty Person to get concurrence on the evaluation of the event. E o.91 [0.25]-

- Station Superintendent so that he may assess  :

the significance of the event.Eo d E4-251 b..af. Division Vice President and General Manager [0.50]

REFERENCE i

BWD BwAP 1250-5, Pg. 2 -

ANSWER 8.02 (2.00) [ k <r. ot & he h@ ced

1. Ensuretheplantisplacedinasafecondition(byobserving that the necessary operations are performed in accordance with approved procedures.)

2. Notify the SCRE (if not present in the control room.) f

3. Determine subsequent actions to be taken W otify OAS duty person on call)and (other appropriate personnel and (or agencies.)  ;

C. A . EnsureReactorTripdocumentationiscomplete(inaccordance  !

with BwGP 100-A13," h.cvwe h4 hs. J -W ** ) {

i

[0.5 each] f i

REFERENCE _t BWD,BwAP 300-1, Pg. 4-5 ,

I l

i I

l l

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 35 ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 8.03 ,

(2.00)

a. SCRE/ Control Room Supervisor. [0.5]

b. 1. Licensed operator must be specifically assigned the responsibility of monitoring the controls of the unattended unit.

2. This sams operator must remain within line of sight of the unit's front panels.

3. The licensed operator must (on a periodic basis) review the status of the unattended unit from within the "at the controls" area.

[0.5 each]

REFERENCE -

BWD, BwAP 300-1, Pg. 8-9 i

ANSWER 8.04 (2.50)

a. StationShiftEngineer,koriftime oes not permit, the licensed SRO immediately available. [1.0)

b. 1. Injury to the Public or Company personnel.

2. Releases off-site above T.S. limits.

3. Damage to equipment if damage may have adverse . t effect public health and safety.

[0.5 each]

REFERENCE BWD, BwAP 300-1, Pg. 13

[

t

-~ - --- _

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 36 ANSKIRS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 8.05 (2.50)

[E, h o.cc.pw b A.

• • 1 I I

1. uae  %%.  %. b.% ,,

1 t u-t 3 ,, l OR 1 "" S* **' d l 3 z 1 , [ (0.15 each]

b. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1

During shift turnover when a crew member is late or absent.

[0.25 each]

c. Designate an individual with a valid SRO license to assume control. [0.5]

Designate an individual with a valid Operators License to assume control. [0.5]

REFERENCE '

BWD, Tech. Spec. Section 6, Pg. 6-5 oo, 3 A7 ANSWER 8.06 (3.00)

( weaq o mWU

a. Individuals qualified 4in radiation protection proce ures (or personnel continuously escorted by such individuals [1.0]

b. Locked doors with controlled keys.h aG (1.0]

r80 Area must be barricaded (by more than rope),

c.

conspicuously posted, and a flashing med light ,

shall be active. (1.0]

REFERENCE BWD, Tech. Spec., Section 6, Pg. 6-24 ANSWER 8.07 (2.50) a.i.Within one hour initiate the following:

2.At least hot standby in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> f 1.At least hot shutdown in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

9. At least cold shutdown in the next 24 hrs. [0.5 each] I

b. 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> (0.5]

l I i 1

8. ADMINISTRATIVE PROCEDURM. CONDITIONS. AND LIMITATIONS PAGE 37 ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

REFERENCE BWD, Tech. Specs., Section 3, Pg. 3/4 0-1 ANSWER 8.08 (2.50)

a. A maximum allowable extension not to exceed 25% of the surveillance interval [0.5], but the combined time interval for any 3 consecutive, e

intervals shall not exceed 3.25 times the specified interval [0.5].

b. The equipment must be declared inoperable. [0.5]

c. 1. At least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (sG4v) 2 At 1:::t ::::  ::--; 92 d:;:

J. 3. At least once every 184 days (h**~ 0 (1.0)

REFERENCE BWD, Tech. Specs., Section 3, Pg. 3/4 0-2 i

ANSWER 8.09 (2.00) {h a..pe } .

1. Declaration of an Unusual Event, Alert, Site Emergency i or General Emergency condition.

l

2. The decision to notify and recommend protective actions to offsite authorities (in the case where a Site Emergency or  ;

General Emergency condition exists and the Recovery Manager and/or Corporate Command Center are not prepared to do so.)  ;

3. A%n t. L.,T y 4 -p,sm L.ak C 2. 4 3 i

[1.0 each] t

9. Rp4  % - .1 % -; c v.wb-c e-  ;

REFERENCE BWD, Bw2P 100-1, Pg. 1 .

I h d. AcO A hs p,.m c ,

l ANSWER 8.10 (2.00) j

i

a. 15 minutes (0.5]  !

b. When the event is calssified by the Station Director. [0.5] ,

i

c. Have them call back to an outside phone line [0.5]  !

(do not provide) outside phone number information (0.5]. j S

REFERENCE BWD, BwZP 300-1, Pg. 2 ,

t i

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

ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE 38 8.

ANSWERS -- BRAIDWOOD 1 -86/37/16-JAGGAR, F.

ANSWER 8.11 , (1.50)

a. When: directed by the Station Director i

an alert is declared a Site Emergency is declared l a General Emergency is declared [0.25 each) i

b. A formal turnover between the Station Director and Shift Engineer, , (0.53 REFERENCE BWD, BwZP 400-1, Pg. 1

e 4

U. S. NUCLEAR REGULATORY COMMIGSION REACTOR OPERATOR LICENSE EXAMINATION 4

FACILITY: _BRAIDWQQD 1 __ _________

- REACTOR TYPE: _PWR-WEC4_ _ _ __________

DATE ADMINISTERED: _@bf@Ifib________________

EXAMINER: _JAGGAR2 _E _____________ _

APPLICANT: 4

} 04 f_1__r; 4 PTT"n; p 6 ..r.'

' ~

l'i. % $ U i ., v .' . 3 INSIEUCIlgN@_Ig_@EELICSNIl Use separate paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY  % OF APPLICANT'S CATEGORY $

__YeLUg_ _IgI@L ___SCOBE___ _y@LUE__ ______________g61EGOBY_____________

_EE @@__ _2E 9@ ___________ ________

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 29 5 .

_2 EASE __ _2E1@@ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

_2@z@@__ _2Ez@@ ___________ ________ 3. INSTRUMENTS AND CONTRCLS

________ 4. PROCEDURES - NORMAL, ABNORMAL,

_2EzEE__ _2Et@@ ___________

EMERGENCY AND RADICLOGICAL CONTROL 99.5 122_0E__ 199 99 ___________ ________

TCTALS FINAL GRADE _________________%

All work done on this examination is my own. I have neither given nor received aid.

3PFCIC3UT 5~55G55TU55-~~~~~~~~~~~~~

o NRC' RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the acair.htratica cf'this examination the following rules apply:

1. Cheating on the examination means an automatic dental of your application and could result in more severe penalties.

2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or pussibility of cheating.

3. Use' black ink or dark pencil o,nly l to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of tres examination.

5. Fill in the date on the cover sheet of the examination (if necessary).

6. Use only the paper provided for answers.

7. Print your name in the upper right-hand corner of the first page of g section of the answer sheet.

8.

" as Consecutively number each answer sheet, write "End of Category Tde of appropriate, start each category on a new page, writa goni ong s the paper, and write "Last Pags" on th7 east answer stuet.

9. Number each answer as to category and number, for example,1.4, 6.3.

10. Skip at least three Ifnes between each anever.

]

11. Separate answer sheets free pad and place finished answer sheets face down on your desk or table.

12. Use abbreviations only if they are commonly used in facility 11teraturg.

13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain an answer to esthematical problems whether indicated in the question or not.

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QtMSTION AN0 00 NOT LEAVE ANY ANSWER BLANK.

16. If parts of th.e examination are not clear as to intent, ask questions of i the examiner only.

17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

18. When you complete your examination, you shall:

a. Assencia your examination as follows:

(1) Exam questions on top.

1 (2) Exam aids - figures, tantes, etc. I

\

(3) Answer pages including figures which are a part of the answer.

D. Turn in your cooy of the examination and all pages used to answer tne examination questions.

c. Turn in all scrap Daoer and the Dalance of the paper that you did '

not use for answering the questions.

d. Leave the examination area, as dafined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

l

. 1___ESINClELES_gE_NUG(E@B_EgMEB_E6@NI_gEEE@IlgN 2 PAGE 2 IME6dgQYN@ digs1_SE@I_IE9NSEE6_@NQ_E6Ulp_E6gM QUESTION ,

1.01 (1.00) s During a Xenon-free reactor startup, critical data was inadvartently taken two decades below the required Intermediate Range (IR) level (1xE-10 amps) .

l

~

The critical data was taken again at the proper IR level (1xE-8 amps) .

. Assuming RCS temperatures and baron concentrations were the same for each set of data, which one of the following statements is correct?

a. The critical red position taken at the proper IR level is LESS THAN the critical rod pcsition taken two cecades below the proper IR level.

b. The critical rod position taken at the prcper IR level is THE SAME As the critical red position taken two decades below the proper IR 1evel.

c. The critical rod position taken at the proper IR level is GREATER THAN the critical rod position taken two decades below the proper IR level.

d. The critical rod position taken at the proper IR level CANNOT BE COMPARED to the critical rod position taken two decades below the proper IR level.

CUESTION 1.02 (2.00)

Indicate whether the following will cause the differential rod worth of one control rod to INCREASE, DECREASE or have NO EFFECT.

a. An adjacent rod is incerted to the same height

a. Mcderator temperature is INCREASED B

c. Baron concentration is DECREASED

d. An adjacent burnable poison rod depletes j

4 QUESTION 1.00 (1.00)

TRUE CR FALSE 7 i

As Baron concentration increases:

a. Moderator Temperature Coefficient becomes less negative due to '

increased neutron leakage.

b. Mcderator Temperature Coefficient becomes more negative due to the increased resonance absorption factor.

(***** CATEGORY 01 CONTINUED CN NEXT PAGE *****)

i PRINCIPLES OF NUCLEAR POWER PLANT OPERATRON2 PAGE 3 1.

IUEBD9DXN@ DICE,t_dg8T_IB@NSEES_@NQ_E(U1g_E(Qg f i

QUESTION .1.04 (2.50)

. a. How AND why coes the Doppler Defect change as reactor power is increased 7 ti.03 i

i

b. How does each of the following affect the Fuel Temperature Coefficient (More negative, less negative, or no effect)? C1.53 l

No explanation is desired or required.

. 1. Accumulation of Xenon and Xrypton gases in the fuel to t-ciao gap.

l 2. Increase in the amount of fuel to clad contact.

i i 3. Buildup of PU209 over core life.

4 l

QUESTION 1.05 (1.50) l i

Compare the calculated Estimated Critical Position (ECP) for a i

startup 15 hcurs after a trip to the actual Critical Rod Position I (ACP) if the following events / conditions occurred. Consider each independently. Limit your answer tot 4

I a. ACP higher than ECP.

b. ACP lower than ECP.

f

c. ACP would not be significantly different than ECP.

i

1. One Reactor Coolant Pump is stopped one minute prior to criticality.

2. The steam dump pressure setpoint is increr. sed to a value just -

i below the code safties setpoints.

3 C. The startup is delayed 2 more hcurs.

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

- 1 __ESINCIE(Eg_gE_MyC6E@B_EggE6_E6@NI_ GEE 8@IlgN 2 PAGE O IEEEdggyN@dlCS 3_SE@I_IE@NgEE8_@NQ_E(glQ_E(QM OUESTION 1.06 (1.00)

Complete the sentence by choosing the correct answer from the choices below.

Delayed neutrons play a major role in the coeration of the core because they ...

a. are born at (thermal) slow energy levels (less than i ev) and therefore are more apt to cause a fission as compared to being absorbed by a poison.

b. are considered as epithermal neutrons and therefore they will not travel far enough to leak out of tne core.

c. are born so much later than the prompt neutrons and provide controlability during stesdy state operations and power transients.

d. provide 70% of the fission neutron inventory and have higher importance factors associated with them as compared to prompt neutrons.

QUESTION 1.07 (2.50) ,

a. If the reactor is operating in the power range, how long will it take to raise power from 20% to 40% with a +0.5 DPM Start-up rate? C1.03

b. Will it take the same amount of time to raise power from 40%

to 60% if the same startup is maintained? EXPLAIN. C1.53

(****+ CATEGORY 01 CONTINUED ON NEXT PAGE *****)

[ 1___ESINCIE6Eg_gE_NUC6E88_EgWEB_E68NI_gEg88IlgN2 PAGE 5 IHERMODYNAMICg2_ HEAT _TRANgFER_ANg_FLUIQ_E6gW QUESTION 1.08 (1.00)

The -1/3 DPM SUR following a reactor trip is caused by which one of the following?

a. The decay constant of the longest-lived group of delayed neutrons.

b. The ability of U-235 to fission with source neutrons.

c. The amount of negative reactivity added on a trip being greater than the Shutdown Margin.

d. The doppler effect adding positive reactivity due to the temperature decrease following a trip.

CUESTION 1.09 (1.00) f Part of the reactor thermal safety limit is based upon not allowing saturation conditions at the core hot leg. State the reasoning behind this.

QUESTION 1.10 (3.00)

What is the most significant type of heat transfer (conduction, convection, or radiation) taking place under each of the following conditions? Consider each condition separately.

I

a. Nucleate boiling.

b. Accident condition in which coolant is boiled and converted to steam in the reactor vessel.

c. Heat from fission thru the fuel rod.

d. Decay heat removal by natural circulation of coolant.

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e. Decay heat of fission products to clad surface.

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(

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(***++ CATEGORY 01 CONTINUED ON NEXT PAGE ***++)

, i f

. 1. - P6 ingle 6ES_QE_Nyg(E@B_EgWEB_E6@NI_gEEB@IlgN 2 PAGE 6 IEEBt99XN@ digs 1_bE@I_IE@NSEEE_@NQ_E(ylQ_E6gg QUESTION 1.11 (1.00)

Complete the sentence by choosing the correct answer from the choices below.

The 2000 degrees F maximum peak cladding temperature limit is used because ....

a. it is 500 degrees F below the fuel cladding melting point.

b. any clad temperature higher than this correlates to a fuel center line temperature at the fuel's melting point.

c. a =ircalloy-water reaction is accelerated at temperatures above 2200 F.

d. the thernal concuctivity of zircalloy decreases at temperatures above 2200 F causing an unacceptably sharp rise in the fuel centerline temperature.

1 QUESTION 1.12 (1.00)

Which one of the following describes the changes to the steam that occur between the inlet and outlet of a real (not ideal) turbine 7

a. Enthalpy decreases, entropy decreases, quality decreases.

b. Enthalpy increases, entropy increases, quality increases.

c. Enthalpy constant, entropy decreases, quality decreases.

d. Enthalpy decreases, entrcpy increases, quality decreases.

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

PAGE 7 I - 1. LESINg1E6ES_gE_Ngg65@$_EQWE8_E6@NI_gEEE@IlgN 2 i

IHESQQDYNAMIgg2_ HEAT _TRANgFER_ANp_FLUIp_FLgW f

1 I

j' CUESTION 1.13 ,

(0.00)

I,

a. Since DNB cannot be measured directly, what FOUR parameters are monitored to assure that DNB is not exceeded? C2.03 i

i b. Assuming the reactor is operating at 85% power indicate how the '

j following changes in the plant condition would affect DNBR i (INCREASES, DECREASES, REMAINS THE SAME). Consider each case ,

i separately. C1.03 .

1. The operator withdraws control rods without changing turbine ,

i load.

t l 2. Steam Generator'PORV fails open.

j 3. Reactor Coolant pressure increases. ,

f CUESTION 1.14 (1.50)

, 1 Use the steam tables and associated Mollier' chart to answer the questions '

t below, label quantites with proper units.

. a. During cooldown and depressurization, you are required to remain 50 degrees F subccoled. As the pressure decreases through 2085 poig,

[ what is the maximum Tavg allowed (nearest degree F)? ,

) i

b. Steam is leaking from a pipe flange into a room. A thermocouple (TC) l j'

placed in the leakage stream reads 400 degrees F. How many degrees ,

of superheat is this? j

c. If the thermoccuples in part b. had read 360 degrees F, and the I steam pressure inside the pipe was 560 psia, what would you estimate l

' the steam temperature to be at that pressure? l i

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I  !

• 1:== PRINCIPLES _QE_NUGLEAR_PgWER_P6 ANT _QPERATigN A PAGE 8 l .

IHgRMQDYNAMI_QSz_ HEAT _ TRANSFER _AND_ELUID_FLQW 1

1 d

QUESTION 1.15 ,

(1.00) 3 Which one of the following statements ccncerning Xenon-105 production _and removal is correct?

I j a. At full power, equilibrium conditions, about half of the Xenon is

+ produced by Iodine decay and the other half is produced as direct

fission product.

i l b. Following a reactor trip from equilibrium conditions, Xenon peaks

because delayed neutron precursors continue to decay to Xenon while neutron absorption (burnout) has ceased.

f i

j c. Xenon production and removal increases linearly as power level i increases; i.e., the value of 100% equilibrium Xenon is twice that It of 50% equilibrium Xenon. l

! d. At low power levels, Xenon decay is the major removal method. At i high power levels, burnout is the major removal method.

i i

I QUESTION 1.16 (1.00)

)

l The following statements concern fission product poisons. Complete the

statements with the available answers provided below. Place the answers on

! your answer sheet. CAn answer may be used more than once.3 l

I a. It takes about ____ hours to reach the maximum Xenon concentration  :

after a reactor trip.  ;

i i

b. The decay half-life of Xenon 105 is approximately ____ hours.

l I

1 c. It takes abcut ____ hours to reach equilibrium Xenon concentration f i

after a step increase from 0 to 50% power.

i d. The decay half-life,cf Promethium 149 to Samarium 149 is approximately l ____ hours. i i

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Available Answers

0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />s: 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />; 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />; 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />; 50 hour5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />s: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. l i i

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t PAGE 9

• h__% ANT _ DES I GN __ I NCLUD __h'G I _S AFET,Y_AND _EMER@ENCY_gY@T,[ME QUESTION 2.01 (1.00)

List the interlock requirements that must be satisfied in order to close a Reactor Cociant Pump breaker.

! OUESTION 2.02 ( .50)

! TRUE or FALSE 7 ,

i AN OPERATING RCP in an RCS loop will trip if the associated Loop Isciation Valve Interlock Logic is not satisfied.

1

}

.so i QUESTION 2.03 Hihree-)

TRUE OR FALSE 7 r

The following concern the Pressurizer Power Operated Relief Valves (PORV).

J

a. One PORV is sufficient to prever.t exceeding the fracture toughness j

limits of 10CFR50 Appendix G when water solid.

I b p-e--u-i w eon"*e are re uired fer :;:rpr : u = pr:tectier during I c- tz ;rr:tur; x;te,- ssi.d wr.,.i.iwns.

b. g . SL:ing of the Pressurizer Code Safety Valves considers proper operation

of one Pressurizer PORV.

c. d . The pressurizer can sustain a complete loss of load'without relieving water, if at least one PORV operates pcoperly.

) .

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(***** CATEGCRY 02 CONTINUED ON NEXT PAGE *****) l l

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- 2___E6eNI_DEgIGN_INC6UDING_S@EgIy_6ND_EMgRGENCy_gygIEMg PAGE 10 QUESTION 2.04 (3.00)

Refer to Figure'15 (attached), "CVCS Flow Diagram".

For each number on the figure, provide the appropriate information on your answer page, for the following:

1. ________ GPM (Normal operating)

2. ________ PSIG

3. ________ F

4. ________ PSIG

5. __ ____=_ F (divert setpoint)

6. ________ GPM (maximum allowable for each kind)

7. ________ GPM

8. ________ GPM pw p-p o To%.\

9. ________ GSM

10. ________ GPM

11. ________ F

12. _______ GPM p., m eo Toh\

CUESTION 2.05 (0.00)

The following concern the Reactor Makeup Control System.

a. State the maximum flow rcte (in gallons per minute) allowed by the Boric Acid Flow Controller. CO.53

b. State the flow rate (in gallons per minute) out of the blender if the makeup system is in aut or.iat i c . CO.53

c. At what level is automatic makeup to the VCT started and stopped?

[1.03

d. State all conditions that will generate a " flow deviation" alarm.

C1.03 CUESTION 2.06 (0.00)

Concerning BTRS, state the maximum Dilution AND Boration rates (in ppm /hr) for both BCL AND ECL conditions.

l

'***** CATEGCRY 02 CONTINUED ON NEXT PAGE *****)

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4 PAGE 11

-- 2t__E60NI_pgSigN_JNC69DlNG_ggEEIY_gNp_Edg6G[NCY_EYSIgdg-n QUESTION 2.07 (2.00) q l

i The following c6ncern valves in the Residual Heat Removal System.

a. State the FOUR conditions that must be satisfied-in order to open valves 9701A and 9702A, RHR Suction Isolation Valves from RCS ti.03 Lw,\ ock s no& Adm. ash we.co-d m.As

) loops.

b. State the TWO signals'that will close these same valves. t2.53

c. State the TWO requirements that must be present in order for valve 8811A, Suction Valve from the Containment Sump, to open automatically.

! t0.53

) . .

)

< q QUESTION 2.09 (3.02)

! State how the Component Cooling Water System (CCW) pumps will respond to 2

t each situation below.

a. Reactor Trip - No Safety Injection (SI) -

No Station Blackout (SBO)

b. Reactor Trip - No SI - With SBO t

c. Reactor Trip - With SI - No 553

! d. Reactor Trip - With SI - With 590 1

QUESTION 2.09 (2.00)

! State, for each of the below, if they are ACTIVE or PASSIVE eatlures.

t, I a. Failure of a pump to start.

4'

b. Loss of packing in a valve.

c. An electrical relay dees not respond.

d. A valve stays open when called on to close.

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

o 2___E6eNI_DESl@N_1NC69 DIN @_S@EgIY_@ND_gd[E@gNCy_SYSI[d@. PAGE 12 1 OUESTION 2.10 (3.00)

a. Following a' reactor trip, an "Overcrank" alarm is received on the 1B Auxiliary Feedpump. List the sequence of JN4Levents that occurred to receive this alarm. C2.03

b. Other than "Overtrank", list FOUR other conditions ths.t will trip and lockout the 1B Auxiliary Feedpump. (Setpoints not required.) C1.03 i

OUESTION 2.11 (3.50)

With RCS pressure starting at Normal Operating Pressure, describe each of the ECCS water injection systum's responsa as pressure decreases to atmospheric, during a LOCA Safety Injection. Assume ALL components are operable and/or running. Include in your answers

a. The *

Af'.c c .- the system, AND

b. 1. The DESIGN flowrate (gpm) and associated pressure, AND The MAXIMUM flowrate (gpm) and associated pressure.

OR .

7 l 2. The MAXIMUM amount of water (gal.) INJECTED'and associated pressure.

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(***** END OF CATEGORY 0 *****)

t

,__y -m- ..,-.r----- w,, - .,  %-- - - _ _ _ - - . - - .,_y -

-mr,, ..r,. , - - . - - - - -, -----.e -

i %y e .. --c- - - , - .y

PAGE 10 3.__IN@IBUDENIS_9ND_CQNIBQ6g QUESTION 3.01 (3.00) s

a. Wnat is the,meanirg of the term "2/4" when indicated on a logic diag am? C1.03

6. What is the purpose of the Shunt Trip in a Reactor Trip Breaker?

When is it energized? C1.53

c. TRUE or FALSE 7 Both Reactor Trip Bypass Breakers can be racked in at the same time, but only one may be closed. CO.53 QUESTION 3.02 (1.50)

The reactor ~has been shutdown without the reactor trip breakers opening and a manual SI has been initiated. If the SI is no longer required, would the 4

SI signal reset? Explain your answer.

QUESTION 3.00 (2.00)

The following concern the Remote Shutdown Panels.

TRUE or FA,LSE7

a. The MCB pull-to-lock feature is overridden when operation is from the Remote Shutdown Panels.

! b. Reactor Ccolant Pumps cannot be started from the Remote Shutdown Panels.

2 c. If local centrol of the MSIV is taken at the Remote Shutdown Panels, no Centrol Rcom alarm will sound.

d. Reactor Centainment Fan Coolers (RCFC's) can be controlled from the Remote Shutdown Panels in high speed only.

1

(***** CATEGCRY 00 CCNTINUED ON NEXT PAGE *****)

., -#..-,.4 - . - - - - - - - + - -% ,. + - - .y.. , y

PAGE 14

- 2 __INgIBUMgyIg_9ND_CgNIRQLg QUESTION 3.04 (2.50)

a. One of the selected Pressuri:er Pressure Channel signals passes through Proportional Integral ("PI") controllar. State t>e.FOUR pressuri:er components that are operated by this signal (be specific) . C1.01

b. What are the T'40 specific input control signals for each Pressuri:er PORV 455A and 456, when selected for Cold Overpressure Protection?

C1.03

c. If the pressure sources to the Pressuri:er PORV's are lost, approximately how many times will the accumulator allow each PORV to cycle? Which direction (OPEN or CLOSE) does.the nitrogen cause the valve to operate? CO.53 OUESTION 3.05 (2.00)

The reactor is at 100% power with normal letdown and charging flow.

Charging flow is manually reduced to minimum and left in manual,.no other changes are made. List the sequence of SEVEN events that will take place ending in a trip or SI. Be spacific, no setpoints required.

QUESTION C.06 (0.50)

One funtion of the Main Feoo Pump D/P program ensures that sufficient feedwater pressure exists, despite the effects of increasing steam pressure the opening of the main _ feed regulating valve, during the design load rejection from 100% power.

a. State TWO other functions of the Feed Pump D/P Program. CO.53 1

6. Explain why, as stated above, the feed regulating valve will open during the load rejection. CO.753

c. 1. List the THREE input signals that are used by the Main Feed Pump D/P Program control system. C3.753

2. What does the D/P Program system control on each of the Main Feed Pumos to vary the feodwater pressure. CO.53 i

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

3___INSIEUMENIS_gND_CgNISQL@ PAGE 15 QUESTION 3.07 (3.00)

a. State the inputs that are used to generate the Power Mismatch signal in ths Reactor Control Unit. CO.53

b. State the purpose of the Summing Unit in the Reactor Control Unit.

C1.03

c. The Summing Unit can only function using temperature signels.

In what system compcnent is the Power Mismatch signal converted to a temperature signal? CO.53

d. Which one of the below compensates the Reactor Contr'o1 Unit for 1 reactivity Changes? C1.03

1. Variable Gain Unit.

2. Non-Linear Gain Unit.

3. Lead-Lag Compensator.

4. Rod Speed Prcgrammer.

QUESTION 3.08 (3.00)

Refer to Figure 30-1 attached, " Power Range Channel 41-44". On your answer sheet, state the label for each arrow point, on the figure, assigned a number (1-18). Include name, coincidence and setpoint (if applicable).

QUESTION 3.09 (3.50)

a. Briefly describe how the Reactor Vessel Level Indicating System dutectc a vessel level change. C1.53

b. State FOUR inputs for the Subcooled Margin Monitor. Consider separate redundant transmitters of the same parameter as ONE input. C2.03

\ .

t

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

3.__IN!IEUMENIg_QND_GQUIB96S PAGE 16 -.

A QUESTION 3.10 (2.00) 1

, For each type os radiation monitor, list the MAJOR type of detector used (G-M Tube, ion chamber, scintillation etc...) and MAJOR radiation type detected (alpha, beta, gamma etc...).

a. Area Monitors.

4

b. Gaseous.

4

c. Particulate (Gas streams).

t

d. Iodine (Gas streams).

j l

l L

1 -

I i.

1 i

i i

(***** END CF CATEGORY 0; *****)

?

4 z__EBggEgyBES_;_NQBd@62_@BNgEd@62_gdgBGENgY_6d9 PAGE 17 o

5Ac;OLOGI996_G9NI$gh J

QUESTION 4.01 (0.00)

Answer the following according to BwGP 100-1, Plant Heatup.  ;

a. In the precautions of GP 100-1, there are 3 specific points (or times) in the procedure when ADV-CV131 (Letdown Pressure Control Valve) must be placed in manual. List these THREE points. C1.53 When using the RHR system, the RCS must be maintained below b.

__________ psig and _________ F. [0,53

c. All RCP's and RH Pumps may be stopped for up to one hour provided TWO conditions are met. State these TWO conditions. C1.03 i

OUESTION 4.02 (2.00)

State TWO RCS conditions that allow positive reactivity to be added without having the Shutdown Banks withdrawn.

i OUESTION 4.0~ (1.50)

As the NSO, what are the THREE requi'. ements according to BwGP 100-2, Plant i

Startup, if the reactor achieves criticality below the Low Low insertion limit? r P

CUESTION 4.04 (1.50)  :

Per BwGP 100-6, Refueling Outages ,

a. Emergency Beration shall be initiated if boron concentration ,

decreases to ___________ ppm.

b. Shutdown Margin shall be greater than or equal to ____________ ,

(include units) for Mode 5.

c. Baron Concentration of RCS and Re3ueling Canal shall be such that the more restrictive of Keff is less than or equal to ___ _ ___ _ or boron concentration is greater than or equal to __________ ppm.

i

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

• da__EBgggpu$gg_;_NQ$d@62_@@NQBM@61_[dgB@gNQY_@ND PAGE 18 B89196901G86_QQNIBQ6 QUESTION 4.05 (3.00)

The following concern BwAP 000-1, Conduct of Operations.

C1.03

a. As Unit 1 NSO, what constitutes "at the controls".

b. What must be done if the NSO must leave the "at the controls" area Does this also include going during non-emergency conditions?

behind the Main Control Board for a valve manipulation or reading? CO.753

,c . 1. Who has the authority to allow the NSO to leave his "at the controls" area to assist the other unit? CO.53

2. State THREE basic guidelines that are used to determine if a unit's emergency is serious enough to warrant assistance from the other unit's NSC. [0.753 QUESTION 4.06 ( .50)

TRUE or FALSE?

A Safety Injection Pump with its control switch in " pull-to-lock",

is still considered operable if a dedicated operator is staticned at its control switch.

1

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

PAGE 19

- d___PBgCEQUEES_;_Ng80862_@BNgSd@62_gdEEGENCY_@ND

. 68D10'0G1986_990I696 -

QUESTION 4.07 (2.50)

The following concern BwAP 080-1, Green Board Concept-Control Panels.

a. What % power, above which, is the Main Control Board and i Remote Shutdown Panels Green Board configuration based on? CO.53

b. 1. A valve, normally open when the system is in its normal configuration, would indicate ________ (color) when open and

______ _ (color) when shut. CO.53

2. For a normal line-up, what is the positier. of a valve that 4

has a green bar and a green circle as its MCB sndication? CO.53

3. What color is used on the MCB to indie.mte a tripped condition? [0.253

4. Turbine :ero speed being reached is indicated by a (color) light. CO.253

5. A motor that has a standby or backup motor would indicate

___________ ( color) for stop and __________

( color) for running.

CO.53 ,

QUESTION 4.08 (2.00)

Define the following, according to BwAP 1450-2, Access to High Radiation i

! Areas.

1 .

a. High Radiation Area. ,

b. Hot Spots.

o QUESTICN 4.09 (2.00) i I

a. State the DAILY whole body dose limit for any individual at Braidwood without further approvals, to increase this limit. CO.53

b. Who can approve exceeding the daily limit AND what is the new limit with this approval? C1.03

, c. If the limit-in b., above, needs to be exceeded, who must approve i this additional increase? CO.53 i .

> 1 l

(***** CATEGCRY 04 CONTINUED ON NEXT PAGE *****)

A I

PAGE 20 4.__EBgCgDuggg_:_NgBd@L _@gNgBd@L 2 2_gdgE@gygy_6NQ

. 68D196991Ce6_CgNIBg6 OUESTION 4.10 ,

(1.00)

Assume plant has experienced a small LOCA, ProcedureSI has been initiated, reset, 19wEP-1, Loss of and only the charging pumps remain running.

Reactor or Secondary Coolant, is in effect.

What action would be required if pressuri:er level began to decrease and could not be maintained above 4%7 OUESTION 4.11 (0.00)

State the THREE. conditions, if one of which existed, that would require the NSO to trip the RCP's when procedure 18wEP-0, Reactor Trip or Safety Injection, is in affect.

i OUESTION 4.12 (3.00)

a. State the SIX Critical Safety Function Status Trees (CSFST) in order of priority. Include the single letter designator assigned to each tree. [2.43

b. True or False?

An ORANGE PATH in the CORE COOLING CSFGT takes priority over a RED PATH in the CONTAINMENT CSFST. CO.63 i

C+++** END OF CATEGORY 04 *****)

(******+++++++ END OF EXAMINATION ***********+++*)

. EQUAT!CN SHEIT C/cle efficiency = (.1 etwrt f = .na v = s/

, cut)/(Energy in)

,=q -

s = V3 :

• 1/2 at .

7 g = x- z ,17; 4 , 4,,-A:

<g = 1/2 av 2 3 , (yf ,7,)j, ,

i PE = agn  ;

• = e/t A = m2/t1/2 = 0.693/t1/2 vf = V, - at C 8#

• U *U  !

2 1/2 4,,j A=

2D ((c ,) + p5)J 4 .

ei = 931 :m

-Ex l m=V avAo g I,e Q = mCast I = I,e h = 347 I = I,10** ##l Pw = W fah T/t. = 1.3/u  ;

sur(:) gyg, , ,g,gg37, 1 7 = 7 O* 10 7 = 7,e'/ ' , SG = S/(1 - Xgf) l l

5:.:R = 25.06/T G, a S/(1 - <gfx)

G j (1 - 4,ffy = G2(I ~'#eff2)

SG = 25s/L' * (s - a)T M = 1/(1 - K,g) = CRt /U s

~ = ( I=/s ) + C(a - s '/ Ia ] '

T = A/(s - 3) M = (1 - Xdfo)/II ~ #eMI) '

SrA = (1 - < df)/Ke#

7 = (3 - o)/(Ta) * = 10 sec:nas l s = (Kg,-I)/Kgf = JXgf /Xgf I = 0.1 secones-! ' ,

s=[(t*/(TKdf)3 * [Idf /(I

• II)3 Idl1*Ek 2 Id j 2 ,2g,3,, -- ,

P = (:47)/(3 x 1010) R/hr = (0.5 CE)/c"(metars) 7 I

=d R/hr = 6 CE/c' (feet) ,

Miset11aneous 0:nversiens f Watar 2mmeters l 1 curia = 3.7 x 1010,33 1 gal . = 3.245 tom. 1 4g = 2.21 lem I 1 gaj. == 7.483.7811:ars gal.

1nc=2.34x103Stu/hr

• 1 f. ,

1 m = 3.11 x 100 5tu/Mr Density = 52.4 Terp/ftJ lin = 2.34 :::n i.

Jensity = 1 ;m/c9

• F = 9/5=0 - 32 i Heat of vaccr1:stien = 970 3:u/lem *C = 5/9 (*F-32)

. eat of fuston = 144 3:u/lem M I ST1J = 778 ft-lbf 1 Atm = 14.7 :si = 29.9 in. Hg.

,' 1 ft. H 23 = 0.4335 1:f/in. t i

l

9 0-X m _

e -.:C m N $ 0

@(

3 g _

O  ::A:

E 3

ed

• ) '

~6 g

0-X WC-

.l 9

I

~~Y s u 3 5

__j. EM -'

A -s _, 0-! :

- s ,

N -

I g

0-X a

w w

I m -

e, 4 4

,) V , a s

. y

=

W(! 9 '

FIGURE 15 a-19 CHEMICAL AND VOLUME CONTROL SYSTEM FLOW BALANCE

o * .

u WETER RANGE e CTOR DE ton SELECTOR SHUNTS METER RANGE SELECTOR SHUNTS e- - TEST SIGN AL S ism g ,3 TEST SIGNAL Am A .S

, AMMETER HIGH v'OLTAGE AMMETER ,

l

<; POWER SUPPLY < 4 i'

,, i <

300.iS00w. 1

<l ,,

dt d '

. . . 5 =

_g_ eT _pe rc .

. .m _ _ . ,

,.A,g,f,,

M GAINADJUST l

~_

p ~@

-e 2/4; H; g OTHER THREE

• 1/4 g- -

. PR CHAl#IELS 2'/4 ; 10 *4 PWR- j- j j g 2/4 YA y *h

~

POWER MlSMATCH DEFE AT SWITCH 2/4 " IOS*4 PWR

& =

~

W g ~- Q A

La I/4 2/4; S0*4 PWR.

g w_h;1/4 10S */n PWR- @D

, ,, y % J. I ROD STOP RYPASS SW. ADJUSTASLE

-t S */. IN 2 S E C.

6) . =

i I /4 r

,J 7 O ~

COMRug *4 FULL POWER ON NIS PANEL ATOR ( 0 -12 0 *4 )

- S *4 I N 2 S- E C. .

h '

, s /4 [ d COMPARES PWA. LEVEL NOW' WgTH WHAT IT WAS e _ 2 SEc. 8 Aso.

Figure 33-1 .. * . . ..

n,

1 __EBINCIE6ES_QE_ NUCLE @B_EgWEE_E68NI_QEEB@ TION 2 PAGE 21 r-IHgEggDYNgdlC@2_ HEAT TRANSFER _AND_ FLUID _ FLOW ANSWERS -- BRAIDWOOD 1 -96/07/16-JAGGAR, F.

.c m 4T "' S ANSWER 1.01 (1.00) g i b

REFERENCE NUS, Nuclear Energy Training, Module 3, Unit 6 Westinghouse Reactor Physics, Sect. 3, Neutron Kinetics and Sect. 5, Ccre Physics HBR, Reactor Theory, Sessions 20 and 24 - 31 Zion, NUS book 3, sections 6.5, 6.6, 6.7, 12.4, 12.5.

BWD, Westinghouse Large PWR Core Control, Ch. 7, Pp. 24-34.

ANSWER 1.02 (2.00)

a. Ducrease

b. Increase

c. Increase

d. Increase [0.50 each3 REFERENCE SQN/WBN License Requal Training, " Core Poisons" BWD, Westinghouse Large PWR Core Control, Ch. 6.

001/000; K5.09 (3.5/3.7) & K5.02 (2.9/3.4) & K5.10 (3.9/4.1)

ANSWER 1.03 (1.00)

a. FALSE

b. FALSE REFERENCE IP-3 ECI Rx Theory; Chapter 7, Pages 21, 22, and 27 DCC R:: Theery Review Text, pp. I-5.42 - .50 SHNP, RT-LP-1.13, Pp. 11-15.

RCWE Reactor Operator Training _ Manual, p. 3-233 BWD, Westinghouse Large PWR Core Centrol, Ch. 3.

I l

. 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATTON 2 - PAGE 22 ISE3MODYN6MIC32_dE@I_IB@NSEEB_@ND_E6UID_E60W ANSWERS -- BRAIDWOOD 1 -96/07/16-JAGGAR, F.

q 4

6 ANSWER 1.04 (2.50)

a. As power increases, fuel temperature increases, the Doppler Defect becomes more negative C1.03 Less

b. 1. Mer-e n eg at i v e.

2. M negative.

3. g negati(,

ve Becau

. - e, mo etL=A- 4ead oC- CO.5 o i- Pu-1M %

3 W -tu each3 o)

REFERENCE CP&L Reactor Theory Chap. 14 pp. 14-7 thru 14-11 ROWE Reactor Operator Training Manual, Sec. 3, pp 189-202

• BWD, Westinghouse Large PWR Core Control, Ch. O, Pp. 23-48.

ANSWER 1.05 (1.50)

1. c (same)

2. a (ACP higher)

3. b (ACP lower > CO.5 ea.3 REFERENCE SONP, Review of Core Poisons, pp. 4- 7 KA001/000,K5.19,4.2.

Cook Theory, Pp. I-36-45.

Zion, NUS book 3, section 12.5.

BWD, Westinghouse Large PW4 Core Control, Ch. 7, Pp. 24-34. ,

ANSWER 1.06 (1.00)  ;

c.

REFERENCE SONP, Review of Neutron Kinetics, p. 5 KA001/000,k5.49,2.9.

Cook Theory, Pp. I-3.3-10.

Zion, NUS book 3, section 5.5.

BWD, Westinghouse Large PWR Core Control, Ch. 7. Pp. 23-00.

f e

o

- 1.__ PRINCIPLE @_QF_ NUCLEAR _PQWER_ PLANT _QPERATigN2 PAGE 23 ISEEdQDYN@DICg1_dE6I_IE@NSEEB_@MD_ELQ1Q_ELQW ANSWERS -- BRAIDWOOD 1 ,-86/07/16-JAGGAR, F.

ANSWER 1.07 (2.50)

a. 36 seconds. (+/- 2) C1.03

b. No.C.53 Power escalation is a log function and therefore increases at an increasing rata. [1.03 REFERENCE Cook Theory, Pp. !3.15-16. KA001/010,K5.37,3.2.

Zion, NUS book 3, section 6.4 BWD, Westinghouse Large PWR Core Control, Ch. 7, Pp. 12-22.

ANSWER 1.09 (1.00) a REFERENCE VEGP, Training Text, Vol. 9, p. 21-47 Westinghouse Reactor Physics, pp. I-3.17 & 19 DPC, Fundamertals of Nuclear Reactor Engineering, p. 106 BWD, Westinghouse Large PWR Core Control, Ch. 7, Pp. 20-30.

001/000-K5.49 (2.9/3.4) f ANSWER 1.09 (1.00)

(If saturation conditions were allowed to exist at the hot leg) further increases in core heat output would be undetected by the het leg RTD CO.53 and protection 4would be degraded. fm 5]

(ip,co<a powee Mg% c 4., ud h W V N 2<)

REFERENCE Westinghouse Thermal-Hydraulic Principles for PWR., Ch.13, Pp.13-53.

ANSWER 1.10 (0.00)

a. Convection

b. Radiation / convection (large Delta T)

c. Conduction

d. Convection (natural)

e. Conducti on (wr. e ua ) /on R=41=M e b \=4I CO.60 each3

1___EBINg1E6gg_gE_ggg6g@g_Egyg8_E69NI_gEg6@IJgN2 PAGE 24 IEEBM9DyN8Migg2_bgA7_IB@NSEEB_@Ng_E(YID _E(gW ANSWERS -- BRAIDWOOD ! -86/07/16-JAGGAR, F.

REFERENCE ,

BWD, Westinghouse Thermal-Hydraulic Principles for PWR, Ch. C.

ROWE Reactor Operator Training Manual, Sec. 2, pp 64-69 ANSWER 1.11 (1.00)

C REFERENCE MNS Thermo-Core Performance, p.2.

BWD, Westinghouse Thermal-Hydraulic Principles for PWR, Ch. 13.

ANSWER 1.12 (1.00) d REFERENCE MNS OP-SS-HT-2, p.12.

BWD, Wastinghouse Thermal-Hydraulic Principles for PWR, Ch. 7.

ANSWER 1.10 (3.00)

a. Nuclear Power, RCS temperature (Tave), RCS Loop Flow, RCS Pressure. CO.50 each3

b. 1, DNBR decreases 2.DNBR wwsc-eases 3.DNBR increases Deauaes C1.03 REFERENCE McGuiro Question Bank BWD, Wastinghouse Thermal-Hydraulic Principles for PWR, Ch. 13.

ANSWER 1.14 (1.50)

a. 592 - 393 degrees F (depending on how round-off is done).

b. FEF degrees F of superheat per superheat tables.

t SS- Wo

c. 500 degrees F ygo-sio REFERENCE m Steam Tablas and Xollier chart E

1___ESINCIE6ES_QE_NUC6E@B_EgWEB_E6@MI_OPEBATigN1 PAGE 25 -

IMERMODYNAMIC@1_ HEAT _TRAN@FER_ANg_ FLUID _FLgW ANSWERS -- BRAIDWOCD 1 -96/07/16-JAGGAR, F. .

ANSWER 1.15 (1.CO)

D.

REFERENCE Westinghouse Reactor Phy si cs , pp. I-5.60-76..

H9R, Reactor Theory, Ses:sions 08 and 39.

DPC, Fundamentals of Nuclear Reactor Engineering,Section VI.

ANSWER 1.16 (1.00)

a. 5 (or 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />). ,

l

b. 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

c. 50 (or 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />).

d. 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.

t' REFERENCE Westingouse NTO Nuclear Physics, pp. I-5.70-79. 1 i

?

f r

1 l

l

t

2,__P6@NT DEg]QN_JNGLUDINg_@@FgTy_AND_gdggggNgY_gYgIgdS 'PAGE 26 3

ANSWERS -- BRAIDWOOD 1 -96/07/16-JAGGAR, F.

d ANSWER 2.01 (1.00) i caux

1. Oil Lift pump discharge pressure at least,600 psiglo.83

2. s Loop Isolation valve interlock saMsked ,

[0.5 eachJ REFERENCE BWD, S.D. CH. 13, Pg. 30 ANSWER 2.02 ( .50)

TRUE REFERENCE BWD, S.D. CH. 13, Pg. 30 +53, i.so ANSWER 2.03 Gr404 i

a. TRUE

b. 9 02 b W. FALSE

0. W. FASLE CO.5 each3 s REFERENCE ,

BWD, S.D. Ch. 14, Pgs. 12-14 ANSWER 2.04 (3.00)

1. 75 5.. 138 />33 9. 12

2. 22 5 6. 120 Mixed, 75 Cation 10. 55

3. 557-75s 7. 87 11. 500-534 4 370 t 5 8. 32 or 9 per RCP(6 G W"*0 12. O per RCP or 12

[0.25 each] -

REFERENCE '

SWD, S.D. CH. 15a., Figure 15a-19 s

O e

2. __ PLANI _ DESIGN _ INCLUD,I,N@_@AFEYY _AND_EbER@ENCy_@Z@ TEM @ PAGE 27 ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

e

~

ANSWER 2.05 (3.00)

a. 40 GPM CO.53  ;

b. 120 GPM CO.53

c. Start -37% VCT level Stop -55% VCT level CO.5 each]

i

d. BA flow deviation: +/ .8 gpm of setpoint PW flow deviation: +/- 9 gpm of setpoint CO.5 each3 i

REFERENCE BWD, S.D. Ch. 156., Pgs. 36-38  ;

ANSWER 2.06 (2.00)

Beration: BOL - 40 ppm /hr 'la c < *d 'i k Roc EOL - 20 ppm /hr L %.M c -. Liwa 2 x bil be , CO.5 each] r uma Dilution: BOL - 20 ppm /hr 7g y ,, 7 g ugu EOL - 10 ppm /hr CO.5 each]

iTEFERENCE -

BWD, S.D. CH. 16, Pg. 25 ANSWER 2.07 (2.00) ,

a. 1. 8812 A/sclosed

2. 8804 Akclosed  ;

3. 8811 A4 closed

4. RCS Pressure </= 360 psig CO.25 each] l (Open signal from MCB)

b. 1. Cicse signal 4 rom MCB

2. RCS pressure at 662 psig (MCB switch in Auto) C0.25 each]

c. (MCB switch in Auto)

1. "S" signal presentbI)

2. 96+ Lo-Lo levels in RWST CO.25 each3 REFERENCE BWD, S.D. CH. 1G, Pgs. 17-18

2__ E68NI_DE@lGN_INC6UDING_S@EEIy_@ND_EdESgENCY_@XSIEd@ PAGE 28 ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F. ,

ANSWER 2.08 (3.00)

a. Ccntinue operation as before trip. (No effect on system.)

6. All CCW pum trip and restarted automatically upon regaining ESF power. [ps ho-will93 t/

c. All non-running CCW pumps will start ( e:t cep t "O" pump) by SI actuation requence.

d. All CCW pumps will trip and then restart from the SI actuation sequence when power is restored tc 6he ESF bus.

[0.75 each3 REFERENCE BWD, S.D. CH. 19, Pgs. 42-43 ANGWER 2.09 (2.00)

a. Active 1.

i

b. Passive

c. Active

d. Active CO.5 each3 REFERENCE .

BWD, S.D. CH. 26, Pg. 7 l

l F

• 2 __ PLANI _ DESIGN _INCLUDINQ_@AFEIY_AND_EMER@ENCY_@YgTEM@ PAGE 29 ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 2.10 (3.00)

a. 1. Engine received an AUTO START SIGNALlo D3 U"I

2. Starting motors engaged and CRANKED 5:OR 5 GCC ' * " #' '**'**

3. (Engine DID NOT ACHIEVE 350 RPM IN 5 SEC.3 "[. S

4. 10 SECOND TIME DELAY ACTIVATED. h e r ea n E. e a%=WS

=.

- n u. n v. u.e m. -- r_r _r__ r .m.i n., r , m ,. . -

m,.,~.m,-.

m, ,m , .. g, *

6. STARTING CYCLE ATTEMPTED 4 TIMCS.Eo 53 C0_25 fer e ach--i-t em ; -O v5 -for- pecp a- ccquc cc3

b. 1. High Water Temperature 65o*O

2. Low Oil Pressure bo p=Q l

3. Overspeed 6toon>T

4. (, Low-Lovd Pump Suction Pressure (4.4 5" 43 ac) CO.15 each3 -

REFERENCE BWD, S.D. CH. 26, Pgs. 16-17 ANSWER 2.11 (3.50) [ t l o*/. L 91. 4 paw w=h*s3

a. 1. Centrifugal Charging Pumps: b. 300 gpm (150 each) e 2500 psig 1100 gpm (550 each) G 600 psig CO.5 each]

2. Safety Injection Pumps: b. 800 gpm (400 each) 0 1200 psig 1300 gpm (650 each) 8 800 psig CO.5 each3 3

3. Residual Heat Removal Pumps: b. 6000 gpm (TOOO each) 9 165 psig 10000 gpm (5000 each) @ 125 psig CO.5 each3 L98 5- ran 4 Accumulators: 6. 28,200 Gals.(approximately 2000 each) @

hy T.s. 31-63o/. Last)' approximately a psig CO.5]

(p o% . L O

~

REFERENCE BWD, S.D. CH. 58, Pgs. 22-27 i

- s __1NgIBUdENIS_AND_CONIBOb@ PAGE 30 ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER C.01 (3.00)

a. (It means that it) will require 2 of the 4 possible inputs to activate the particular function. C1.03 t

b. To instre the Reactor Trip Breaker opens if the UV coil fails to open it. CO.753 It is energi::ed by use of3 the manual trip switch. [0.753 a )) a.J. ~.W s tyd s e 4

c. True CO.53

+

REFERENCE BWD,S.D. CH. 60A, Pgs. 11, 16

ANSWER O.02 (1.50)

Yes [0.53: Becauce the manual signal is only momentary, .eset is possible without P-4. (The system, in fact, will return to full automatic ,

operation.) C1.03It ** M hh S M e **b

• 455"~*5 44 e vencAecT*ie D'S eM ducetw % P-4 a.s a e u m o ot- A. ww \ sr ua w ee,v, u E" '"E REFERENCE , N C. A k " W b * -

BWD, S.D. CH. 61, Figure 61-17 ANSWER 3.03 (2.00) d

a. TRUE

b. TRUE j c. TRUE

d. TRUE t.5 ea]

REFERENCE BWD, S.D. CH. 62, Pgs. 14-17 i t

i

PAGE 31

., 3 __INSIRUMENTS AND_ CONTROLS ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 3.04 (2.50)

a. 1. PORV 455A

2. All Back-up Heaters

3. Variable Heaters

4. #1 and #2 spray valves CO.25 mach 3

b. 455A - 1. W.R. Pressure

2. W.R. Low Avg Tcold 456 - 1. W.R. Pressure

2. W.R. Low Avg That CO.25.each]

c. 50 (+/- 0), Open CO.25 each3 1+Forus a w (G -T % Nd REFERENCE BWD, S.D. CH. 14, Figure 13a, 13c, Pg. 22 ho, t.<sso h , E p . Em Gabe ANSWER 3.05 (2.00)

1. Charging < Letdown, P:r. level will decrease CO.53

2. Pressure decrease CO.13

3. Variable Heaters full on, B/U Heaters on CO.13 A. Letdown Isolates, (all heaters off) [0.63

5. Charging > Letdown, P:r. level will increase. [0.13 i 6.4LVerieble Heaters re-energi:e CO.13

' 7. High level Reactor Trip CO.53 i REFERENCE BWD, S.D. CH. 14, Figure 14-2, 14-4 ANSWER 3.06 (2.50)

[% %dl

e. Maintains FRV in linear flow region. { 2 .4 T a.

t. Minimizes wear on FRV. CO.25 each3 L %6 % W yop,3 powar v ege- +- aM==. Ws

b. A load rejection will create shrink in the S/G which causes a level error (calling for the valve to open). CO.753

c. 1. Main Steam Pressure,(PT 507)

Main Feedwater Pressues,(PT 508) -

Steam Flow (Total) CO.25 each]

2. Turbine Driven: Speed of the turbine (or Governce Valves).

i Motor Driven: D;--harge Valvo. CO.25 mach 3 i

m w

3. _INETRUMENIg_AND_CgNTR06@ - PAGE 32 ANSWERS -- BRAIDWOOD 1 -96/07/16-JAGGAR, F.

REFERENCE BWD, S.D. 27, Pgs. 15, Figure 27-8 and Appendix D ANSWE9 3.07 (3.00)

a. Auctioneered High Nuclear Power Turbine lead (Pimpulse) CO.25 each3

b. Summing Unit (addts three temperature error signals together to) generates a total temperature error for the Rod Speed and Direction Programmer. C1.03

c. Non-linear Gain Unit CO.53

d. 1. (Variable Gain Unit) C1.03 REFERENCE BWD,S.D. CH. 29, Pgs. 26-29 I

i i

9 l

p 6

PAGE 33 3.__INSI89dENI@_@ND_CQNI89L@

ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

~

ANSWER 3.08 (3.00)

Flux Level High Rx Trip (Low Range) L'4- 25" C 4 ,4 5 3--

1.

2. C-2 Rod Stop E/4-403% C O . i SG-(1 & 2 INTERCHANGEABLE)

3. P-10 Permissive -274-10% CO.133

4. Flux Level High Rx Trip (High Range) 2/3 109% C3 -1&l-(0 & 4 INTERCHANGEABLE)

5. Power Range High Flux Rate (Positive) 'i / 4 / SE/2-sec CO.153 Power Range High Flux Rate (Negative) 44,-154 6.

(5 & 6 INTERCHANGEABLE)

7. P-9 Pernissive (0 loop flow) C/* 30% CO.153

8. Power Range Channel Current Comparator C detecte--/2!'. CO.153 -

9. Over Power Recorder te_153 (G & 9 INTERCHANGEABLE)

CO.153 l'M S' [* 2 ***k3

10. Rod Control

11. NIS Power Range Loss of Detector Voltage C .

Summing and Level Amp C . 3 12.

3

10. Delta Flux to OP and OT Delta T C

14. NR-45 CO $3

15. Currrent Recorder C 3

16. Computer C . 53

17. Delta flux meter C .0 3

18. Detector Current Comp. 2 Detectors /2% of Avg. 0 0.1}3 "L E '"'*"k-H ~ 'S (14-19 INTERCHANGLEABLE)

REFERENCE BWD, S.D. 33, Figure 33-1 l

4 4

L

• PAGE 34 i . 3 = INSTRUMENTS _AND_ CONTROLS ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

ANSWER 3.09 (5.50)

a. The basic principle of operation is the detection of a Delta-T between adjacent heated and unheated thermoccupies CO.53. The RVLIS sensor consists of a (Chromel-Alumel) TC near a heater and another (Chromel-Alumel) TC positioned away from the heater. In a fluid with relatively good heat transfer properties, the Delta-T between adjacent TC's is small. CO.53 In a fluid with relatively poor heat transfer properties, the Delta-T between the TC's is large.

When a TC is uncovered, the Delta-T is large and the RVLIS indicates the level change. CO.53

b. 1. Pressurizer Pressure (P.T.s 455, 456, 457, and 458)

2. Reactor Coolant Loop Pressur s Wide Range, (Hot Legs A & C (P.T.s 403 and 405))

3. Maximum UHJTC Temperature (f r om RVLIS processing)

4. Representative CET Temperature (f rom CET processi.7g) CO.5 each3

( Av3 .C h k; pud REFERENCE BWD, S.D. CH. 348, Pgs. 10-11, 21 l

ANSWER 3.10 (2.00)

a. G-M, Gamma

b. Scintillation, Beta

c. Scintillation, Beta

d. Scintillation, Gamma CO.5 each3 REFERENCE BWD, S.D. CH. 49, Pg. 17, 60 i

i

PAGE 35

4z__PggCEgygES_;_Nggd@(2_gBNggj@(1_EggBGENCY_gNQ 8891969GIC86_C9NIgg6 ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F. .

ANSWER 4.01 (3.00)

a. 1. Starting a RCP.

2. Starting RH Pump.

3. Stopping a RH Pump. [0.5 each]

b. 400, 350 C2.25 each]

c. 1. No operations that could cause RCS dilution AND

2. Core ou.tlet temperature is maintained 10 F below saturation.

CO.5 each3 t

REFERENCE BWD, BwGP 100-1, Pgs. 2-3, 5 ANSWER 4.02 (2.00) 4

1. RCS borated to Hot, Xenon Free concentration and maintained at Hot Standby. C1.03

2. RCS borated to Cold Shutdown concentration. C1.33 (on A s c.4h2 Rgww e. Co.s} mod co.4. .,,> b, s.-p m Co.s1 )

REFERENCE BWD, BwGP 100-2, Pg. 3 ANSWER 4.03 (1.50)

Co.53 Emergency Beraten100 PPMfe.il Re-insert all CONTRO@IBANKS[oS1 Notif y the Shi4t Engineer b tce*O f o.G te,5 :: h:

REFERENCE BWD, SwGP 100-2, Pg. 9 4

4_=_PRgCEQg8E@_;_NOBd@(2_@gNO$d@(1_Ed[B@gNQY_@ND PAGE C6 o

B691969GIC66_CQNIBQ6 f ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGTAR. F.

ANSWER 4.04 (1.50)

a. 2000 ppm CO.53

b. 1*/. Del ta K/K CO.53

c. .95, 2000 CO.25 each]

REFERENCE

• BWD, BwGP 100-6.

ANSWER 4.05 (3.00)

a. In line of sight of MCB front panels (so to be able to initiate ,

prompt corrective actions when necessary) . (A c ned C1.03_

S'es e M dd\ b o_cc ept ,3 )

b. Obtain relief from a qualified operator. CO.503 Yes. CO.253

c. 1. 'The SCRE/ Control Room Supervisor. CO.503

2. a. Injury to Company personnel or Public.

b. Release offsite in excecs of T.S.

c. Damage to equipment that could affect public health / safety. CO.25 each3 REFERENCE BWD, BwAP 300-1, Pg. 8 ANSWER 4.06 ( .50)

False REFERENCE BWD, BwAP 300-1, Pg. 15 5

l l

, Q.__:_PRQCgQQBgS - NQBd@(2_@BNQBd@(2_EdgBgENCY ANQ PAEE 37 RAplgLQ@lCAL CgNTRgL ,

ANEXERS -- ERAICWOOD 1 -86/07/16-JAGGAR, F.

l ANSWER 4.07 (2.50)

"ho %

a. 297. power CO.53

b. 1. Green - (Open)

Red (shut) CO.25 each3

2. a Open or closed (%p. Jug go. pb,4- cowc)'.%s) CO.53

3. Amber CO.253

4. White CO.253

5. Blue (stop)

Green (run) CO.25 each]

REFERENCE BWD, BwAP 080-1, Pg. 1-3 ANSWER 4.08 (2.00)

a. ANY AREA ACCESSIBLE TO PERSONNEL in which there exi sts RADI ATION at such LEVELS that a major portion of the body could receive in ANY ONE HOUR a dose IN EXCESS OF 100 MREM. C1.03

b. Areas near equipment or piping where the DOSE RATE AT > 18 INCHES from the source EXCEEDS THE applicable posted limits for the GENERAL AREA.

OR Areas near equipment or pipes where the DOSE RATE AT 18 INCHES from the source would EXCEED 5 TIMES THE AMBIENT DOS 2 RATE for the GENERAL AREA. C1.03 REFERENCE BWD, BwAP 1450-1, Pg. 1-2 ANSWER 4.09 (2.00)

a. 50 mrem CO.53

b. Supervisory approval to 100 mrem. C1.03

c. Radiation - Chemistry Supervisor. CO.53

'.s t 0.__EBgCEDyBES_;_NgBd@62_@BNg5d@62_EDEEGENCy_@ND PAGE 30 6ADIOLggig@6_ggNIBg6 ANSWERS -- BRAIDWOOD 1 -86/07/16-JAGGAR, F.

REFERENCE BWD, Radiation Protection Standards, Pg. 24 ANSb!ER 4.10 (1.00)

Manually operate ECCS pumps as necessary to restore level.

REFERENCE BWD, 19wEP-F.1 i

ANSWER 4.11 (3.00)

1. CC water to RCP lost (affected pumps only). CO.753

2. Cntmt Phase B actuation. CO.753

3. a. Both RCS Pressure < 1370 psig

b. CCP's > 200 GPM (SIP positive flow exists) C1.53 REFERENCE BWD, 1BwEP-F.0 ANSWER 4.12 (3.00)

a. 1. Suberiticality (S)

2. Core Cooling (C)

3. Heat Sink (H)

'4 . RCS Integrity (P)

5. Containment (Z)

6. Inventory (I) CO.05 each name, 0.05 each letter 3

b. False C.63 REFERENCE BWD, BwAP 340-1, Pg. 9, 11 i

- - - - - -