ML20214T746

Exam Rept 50-454/OL-86-01 on 860716 & During Wks of 860818 & 25.Exam Results:Five Senior Reactor Operator Candidates & Nine Reactor Operator Candidates Passed Exam.Review Comments,Exam & Answer Key Encl

ML20214T746
Person / Time
Site:	Byron
Issue date:	09/23/1986
From:	Jaggar F, Jenson N, Picker B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20214T742	List: ML20214T739 ML20214T746
References
50-454-OL-86-04, 50-454-OL-86-1, 50-454-OL-86-4, NUDOCS 8609300393
Download: ML20214T746 (116)

Text

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U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-454/0L-86-04 Docket No. 50-454 License No. NPF-37 Docket No. 50-455 Construction Permit No. CPPR-131 Licensee: Congnonwealth Edison Company.

Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Nuclear Power Station Examination Administered At: Byron Nuclear Power Station, Braceville, Illinois Examination Conducted: July 16, 1986, and the weeks of August 18 and August 25, 1986 N

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Examiners:

F. Jaggar Date 7[

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Approved By:

T. M. Burdick, Chief Operator Licensing Section Date_

_ Examination Summary Examination administered on July ~16 1986, and the weeks of AuSust 18 and August 25,191f67 Report Ifol Y0-45T/dT.~-Y6-04T

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~~~~

Examinations were administered to eVt senior reactor operator and twelve

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reactor operator candidates.

Results: Five senior reactor operator candidates and nine reactor operator candidates passed the examination.

8609300393 860924 PDR ADOCK 05000454 V

PDR

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REPORT DETAILS t

1.

Examiners F. Jaggar, INEL - Chief Examiner N. Jensen, INEL B. Picker, INEL 2.

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

3.

Exit Meeting a.

On August 29, 1986, an exit meeting was held. The following personnel were present at this meeting:

K. Gerlinri, PWR Operations Training-Supervisor, Production Training Center R. E. Querio, Ryron Station Manager T. K. Higgins, Byren Training. Supervisor R. Pleniewicz,- Production Superintendent T. Petelle, Simulator Instructor, Production Training Center S. Shankman, Operator Licensing. Branch, NRC Headquarters F. Jaggar, INEL - Chief Examiner B. Picker, INEL Examiner b.

The following generic weaknesses of the candidates were discussed by the Chief Examiner with the utility:

(1) Difficulty finding procedural requirements for sampling when blowdown or air ejector monitors are 00S.

(2) Need to supply logic diagrams for control room personnel.

(3). Noted uncertainty in locating power supplies for NI channels.

(4) When a candidate asks a question of the reactor operator on -

watch, (with examiner's approval) the reactor operator should give a complete and accurate answer.

2

ATTACHMENT BYRON SENIOR REACTOR FACILITY REVIEW COMMENTS QUESTION 5.15 Explain why a dropped control rod is worth approximately 200 pcm and a stuck cod 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.

(1019M/OlllM)

I QUESTION 6.01 ~ '

Refer to figure 15 "CVCS Flow Diagram" for each number on the figure, provide the appropriate information.on your answer page for the following:

NRC answer:

5.

138'F

~

11. 500*F CBCo answer:

5.

133'F

11. 518'F System Description, chapter 15a pages 35 and 21, Rev 3.

Reference:

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

5.

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

11.

QUESTION 6.03 Unit 2 has two additional installed solenoid Operated centrifugal Charging.

Pump mini-flow recire valves, 2CV8114 and 2CV8116.

a.Whatsignalandsetpointwillautomatic.sily 1.

Open r

)

2.

Close I

these valves? (1.0)

b. Why were the additional valves installed ? (0.5)

To prevent dead-heading the CCP's in a Low RWST level NRC answer:

b.

situation with high RCS pressure.

(0.5)

To provide full pump output to the RCS when RCS pressure CBCo answer:

b.

is Low and RWST level is above the Low-Low setpoint.

Byron Unit 1 and Unit 2 differences Page 12.

Reference:

Resolution:

As written in the referenced document there are'two reasons; one, as' stated.

The answer in the original answer and two, as stated above by the facility.

key was changed to require both for full credit.

QUESTION 6.05

b. Why is ths Narrow Range level span on Unit 2 move compressed then Unit 17 Describe this physical change.

NRC answer:

b.

Because of the higher recirculation flow in Unit 2 the S/G is less sensitive to level transients. The lower narrow range tap is higher.

CDCo answer:

b.

Because of the higher recirculation flow in U-2 F/G's, the level span was, compressed to prevent level indication fluctuations that might occur as the recire flow rate increased with power.

Reference:

Byron Unit l'and 2 Difference Book page 17.

Resolution:

Answer key changed to reflect facility clarification, and graded accordingly.

QUESTION 6.06 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 your answer:

a.

The NAME of the system, AND b.l.

The DESIGN flowrate (gpm) and associated pressure, and the maximum flowrate (gpm) and associated pressure OR 2.

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

NRC answer:

a.3.b.

6000 (5000 each) 8 165 psig CDCo answer:

a.3.b.

6000 gpm (3000 each) at 165 psig

Reference:

System Description, chapter 58, pgs 22-27 Resolution Comments noted and changes made to answer key and graded accordingly.

i

QUESTION 6.07

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

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

When is it energized?

c. True or False Both Reactor Trip Bypass Breakers can be racked in at the same time, but only one may be closed.

NRC answer:

a.

When control and protection are provided by the same parameter. With a~ channel failure 2/3 protection is still available.

CECO answer:

a.

The first sentence is correct. The second part doesn't answer WHEN a 2/4 logic is required and should be deleted.

NRC answer:

b.

To insure the Reactor Trip Breaker opens if the UV coil fails to open it.

It is energized by use of the manual' trip switch.

CECO answer:

b.

First sentence is correct.

Second part:

the shunt trip is energized on all RX trips.

References:

System Description, chapter 60A, pages 11 and 16 I

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.

i

QUESTION 7.01 What are the TWO conditions that must be monitored during a steam b.

generator tube rupture accident after a RCS cooldown is initiated, that require RCP's to be tripped?

NRC answer:

a.

1.

CC water to RCP lost. (affected pumps only) 2.

Phase B contmt, isolation.

CECO answer:

1.

CC water to RCP lost. (affected pumps only) 2.

Phase B cntmt. isolation.

3.

Pressurizer spray valve will not close (affected loops 4.

1. 1 RCP seal delta P of Less than 200 paid.

5.

1. 1 Saal leakoff flow less than 0.2. gpm.

References:

3.

1BEP3 Step 16c Response NOT Obtained.

4&5 1BEP ES-3.1 step 11 and BOP RCl page 4 6.

BOP RCl page 4 RESOLUTION:

The delta P The spray valve criteria is accepted as a correct answer.

and leak off criteria are also accepted because they appear in the post-cooldown procedure.

QUESTION 7.08 During performance of BOA ROD-4, " Dropped Rod Recovery" prior to a.

recovery of the dropped rod, state ALL methods that 'can be used to match Tref with Tave.

By reducing turbine load, diluting or moving rods NRC answer:

a.

CECO answer:

Manually adjust rods

-OR-Manually adjust turbine load

-OR-Manually adjust RCS boron concentration

Reference:

IBOA ROD-4, pg. 2 Resolution:

The CE Co. answer further clarifies the NRC answer and no change to the answer key is necessary.

t 5

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QUESTION 7.08 If a dropped rod cannot be recovered immediately, state the THREE b.

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 operator to continue.

NRC answer:

b.

Within 1 hour:

1. Restore rod to operable status, [0.3]

2. Rod is declared inoperable (0.3] and other rods in group aligned within 112 steps, (0.3]

3. Rod is declared inoperable and

a. Tech Spec SDM satisfied,

b. Power reduced to < = 75*4 The question is misleading by mentioning the one hour CECO answer:

In addition, it does not restrict the examinee requirement.

to Tech Spec.

1 BOA ROD-4 also supplies actions to be taken and should be included in the key.

1. Calculate the QPTR, if greater than 1.02, apply Tech Spec 3.2.4

2. Reduce power for dropped rod recovery.

3. Restore rod to operable status

4. Rod is declared inoperable and remainder of rods in the group are aligned within 12 steps of the inop.

rod while maintaining rod sequence and insertion limits.

5. Rod is declared inoperable and SDM is satisfied.

power operator may then continue provided that:

a) power reduced to < = 759.*within the next hour and within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Reactor Flux Trip Setpoint is reduced to less than or equal to 85% of RTD

Reference:

IBOA ROD-4, Rev. 51, pg. 3: Tech Spec 3.1.3.1 RESOLUTION:

Also accepted from BWOA R00-4 will be:

Calculate QPTR and Reduction of Power to 70%.

QUENTION 7.09

a. List the THREE actions, in the correct sequence, that are required, when using procedure 1 BOA RCP-1, Reactor Coolant Pump Seal Failure, if RCP bearing temperature reaches 225*F.

b. Accordf~ng to 1 BOA RCP-1 other than if RCP bearing temperature approaches the alarm level, when must the #1 seal bypass valve be opened?

c. What THREE conditions, all of which must be satisfied, before the

1. 1 seal bypass valve can be opened?

Trip the reactor (0.2).

NRC answer:

a.

Trip the affected pump (0.2).

Go to IBEP-0, Reactor Trip of Safety Injection (0.1)

(0.5)

When the #1 seal temperature approaches the alarm.

b.

1. Seal injection flow 8-13 gpm (0.1) c.

2. #1 seal leakoff < 1 gpm (0.2)

3) RCS pressure < 1000 psig (0.2)

Procedure has been revised.

CECO answer:

c.

1. Seal injection flow is between 8-13 gpm

2. No. 1 seal leakoff isolation valves are open

3. No. 1 seal leakoff flow is less than 1 gpm

4. RCS pressure is greater than 100 psig and less than 1000 psig

)

IBOA RCP-1, Rev. 51, pg. 3

Reference:

Resolution:

The additional correct response is added to the possible answers.

.-_-~._

QUESTION 7.10

a. State FOUR of the 8 symptoms that would indicate a need to enter 1 BOA PRI-1, Excessive Primary Plant Leakage.

(Setpoints not required.)

b. State the TWO specific conditions that would require the reactor to

^

be tripped and a transition from 1 BOA PRI-l to 1DEP-0, Reactor Trip or Safety Injection.

1. Containment Radiation Monitors high NRC answer:

a.

2. Increased charging flow

3. Increased VCT M/U frequency

4. Abnormal Containment pressure / temperature

5. Abnormal PRT conditions

6. Off-gas radiation monitors abnormal

7. Increase sum / cavity pump run times

8. Rx vessel flange leak off high temperature (0.25 each for any 4) b.

1. When pressurizer level cannot be maintained with all CCP's running (0.5).

2. SI 8801A and B open (0.5).

Additional answer includes Blowdown Radiation Monitor CECO answer:

a.

per 1 BOA PRI-l

1) Containment radiation monitors greater than alert alarn setpoint.

r 2)

Increased charging flow during normal operation.

3)

Increased VCT make-up frequency.

4)

Abnormal containment pressure or temperature.

f

5) Abnormal PRT conditions.

6) Off-gas radiation monitor greater than alert alarm setpoint.

7)

Increase sum / cavity pump run times.

Reactor vessel flange leak off temperature high.

8) 9)

Blowdown Radiation Monitor,

b. Only one condition requires this. The RNO on step 4. "If unable to maintain pzr level greater than 4*/....".

The answer key supplies two substeps out of seven that occur prior to step 4.

They are not

" conditions" and should be deleted.

If operator judgement was given as an answer, it should also be considered correct.

Reference:

IBOA PRI-1, Rev. 51

Resolution 7.10:

Part a.

" Blowdown radiation monitor greater than alert setpoint" is another acceptable answer, and is graded accordingly.

Part b.

The answer on the key was changed according to the facility comment after the reference was verified and the question was graded accordingly.

Operator judgement is not considered a specific (plant) condition to base a reactor trip requirement.

i f

a i

QUESTION 8.02 State the minimum number of gallons required per diesel to be in the a.

Diesel Oil storage tanks and the associated indicated level (in %) for each Unit.

b.

State the minimum number of gallons required to be in each unit's Diesel Generator Day Tank and the associated indicated level (in %).

NRC answerk a.

44,000 gallons Unit 1 - 93.8*4 Unit 2 - 90.0%

(1.0) b.

450 gallons Unit 1 - 35%

Unit 2 - 80%

(1.0)

CECO answer:

a.

44,000 gallons (1.0) b.

450 gallons (1.0)

Answer is as stated in Tech Specs. The percentages are not indicated on the Main Control Board. They appear only on the non-licensed operators round sheets. The round sheets have minimum levels stated on them.

Should the actual level' (in %) reach the minimum, the non-licensed operator circles the reading in~ red pen and directs this reading to the attention of the Shift Supervisor.

Reference:

Byron /Braidwood Tech Spec Section 3/4.8.

f Resolution:

Because the operator does not readily have this information presented to him, in percentage, for example on a main control board meter, the percentage values are not required for full credit.

t A

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QUESTION 8.05 a.

What is meant if an instrument number in Technical Specifications is preceeded by a zero?

b.

Refer to attached Figures 1-6a and 1-6b.

For Unit 1 to operate, what instruments must be operable. You may answer by section number if they all apply (i.e. all of #1).

NRC answer:

b.

All of #1-4.

Only Unit 2's instruments for #5.

ORE-PR009 and 2RE-PR009 for #6.

CECO answer:

b.

All of #1-4 is correct. Ilowever, it should be Unit l's instruments for #5 and ORE-PR009 and 1RE-PR009 for #6.

Reference:

Figures 1-6a and 1-6b of exam.

Resolution:

The correct numbers were inserted on the answer key and graded accordingly.

f I

I

q

• g, QUESTION 8.08 a.

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

[1.0]

b.

What must be done if the surveillance requirements for a piece of equipment is not performed within the specified time intervals?

[0.5]

c.

What is the interval for each of the designators below?

[1.0]

1. S

2. Z

3. SA ANSWER 8.08 a.

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

2. At least once every 92 days

3. At least once every 184 days

[1.0]

CECO answer:

C.2.

Z is not a frequency notation for Tech Specs.

A 92 day frequency is designated by a Q.

Delete C.2 from the exam.

RESOLUT0N:

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

> R1019M/0111MD

ATTACHMENT BYRON REACTOR OPERATOR FACILITY RE0IEW COMMENTS QUFSTION 1.04 why does the Doppler Defect Change as reactor power'is a.

How AND increased (1.0}.

How does each of the following affect the Fuel Temperature Coefficient b.

(MORE NEGATIVE, LESS NEGATIVE, or NO EFFECT)?

No explanation is desired or required.

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

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

3. Buildup of PU240 over core life.

NRC answer:

b.

1.

More negative 2.

Less negative Ceco answer:

b.

1.

Less negative 2.

More negative Westinghouse Reactor theory review text pages I-5.16 and

Reference:

I-5.21 Resolution Answers b.1 and b.2 were changed accordingly to the referenced document.

?

u i

i OrJERTION 1.10 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,

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.

e. Decay heat of fission products to clad surface.

NRC answer:

b.

Radiation / Convection (large Delta T)

CECO answer:

b.

Convection.

Answer Key is unclear as to which is the correct answer. The question asks for the most significant.

Reference:

HT & FF, chapter 3, page 100 Resolution:

Either answer is correct depending upon the point of reference.

t l

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l l

b QUESTION 2.02 TRUE or FALSE 7 The following concern the Pressurizer Power Operated Relief Valves (PORV).

a. One PORE is sufficient to prevent exceeding the fracture toughness limits of 10CFR50 Appendix G when water solid.

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

c. Sizing of the Pressurizer Code Safety Valves considers proper operation of one Pressurizer PORV.

d. The pressurizer can sustain a complete loss of load without relieving water, if at least one PORV operates properly.

NRC answer:

a.

True b.

True c.

False d.

False CECO answer:

b.

Can be either depending on the document referenced.

Tech Spec 3.4.9.3 allows either 2 PORV's, or 2RH suction reliefs, or a 2 in2 vent, to provide overpressure protection. Request that part b. be omitted.

Reference:

BGP 100-5 step 3, Tech Spec 3.4.9.3 Resolution Because part b could be true OR false depending upon can'ditions notThe provided in the question, the question is deleted from the e.xam.

point values for section 2 and the total are adjusted accordingly.

QUESTION 2.03 Refer to figure 15 "CVCS Flow Diagram" for each number on the figure.

Provide the appropriate information on your answer page for the following:

NRC answer:

5.

138'F

11. 500'F CECO answer:

5.

133*F

11. 518'r

Reference:

System Description, chapter 15a pages 35 and 21, Rev,3.

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

5.

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

e

~

QUESTION 2.06 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, RHR Suction Isolation Valves from RCS loops.

(1.0] (Interlocks not administrative)

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)

NRC answer:

a.

1.

8812 A closed 2.

8804 A closed 3.

8811 A closed 4.

RCS Pressure </= 360 psig (open signal from MCB)

CECO answer:

8701A 8702A a.

1.

8812 A closed 1.

8812 B closed 2.

8804 A closed 2.

8804 B closed 3.

8811 A closed 3.

8811 B closed 4.

RCS pressure i 360 psig 4.

RCS pressure i 360 psig 8701 A is a suction isolation for A train whereas 8702 A is a suction isolation for B train, therefore, the interlocks are different.

Reference:

System Description, chapter 18, page 16 Resolution:

Comment noted and graded accordingly.

1

(

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QUESTION 2.08 Unit 2 has two additional installed Solenoid Operated Centrifugal Charging Pump Mini-flow Recirc Valves, 2CV8114 and 2CV8116.

a. What signal and setpoint will automatically 1.

Opan 2.

Close These valves?

(1.0)

b. Why were the additional valves installed?

(0.5)

To prevent dead-heading the CCP's in a Low RWST level NRC answer:

b.

situation with high RCS pressure.

[0.5]

To provide full pump output to the RCS when RCS pressure CECO answer:

b.

is Low and RWST level is above the Low-Low setpoint.

Reference:

Syron Unit 1 and Unit 2 differences page 12.

Resolution:

As written in the referenced document there are two reasons; one, as stated.

in the original answer and two, as stated above by the facility.

The answer key was changed to require both for full credit.

f

e QUESTION 2.11 With RCS pressure starting at Normal Operating Pressure, describe each of the BCCS wate,r injection system's response as pressure decreases to atmospheric, during a LOCA Safety Injection. Assume ALL components are operable and/or running.

Include in your answer:

a. The NAME of the system, AND

b. 1.

The DESIGN flowrate (gpm) and associated pressure, AND the MAXIMUM flowrate (gpm) and associated pressure.

OR 2.

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

NRC answer:

a.3.b 6000 (5000 each) 9165 psig CECO answer:

a.3.b 6000 gpm (3000 each) at 165 psig

Reference:

System Description, chapter 58, pgs 22-27 r

l Resolution Comments noted and changes made to answer key and graded accordingly.

1 i

QUESTION 3.01'

....____.n.

m. _. : _,,__2._

m._ t._

m.m... _..2 o m. __ m..

____.:__2

..__2

.m__

._m__m,_

,_m.__.

.... ~..,,.

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

When is it energized?

"'.:n ::ntr:1 :nd pr:t::ti

pr: Tided by th: ::::

. T :n;r:::

p::: :ter. '3ith : :M = 1 f:ilur: 2/2 pr:t:: tie w

till :::11:510, C D :::r:::

. The fir t ::ntene: i ::rr::t. Th: ::::nd p :t d:::n't-ncr::

2/4 logic i: r;;;ir:d nd ;h,uld 5:

.s_,.s.

2.

To insure the Reactor Trip Breaker opens if the UV coil NRC answer:

b.

fails to open it.

It is energized by use of the manual trip switch.

CECO answer:

b.

First sentence is correct.

Second part:

the shunt trip is energized on all RX trips.

References:

System Description, chapter 60A, pages -H-and 16 Resolution

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

QUESTION 3.05 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 with no operator action. Be specific, include all automatic func,tions, no setpoints required.

NRC answer:

1.

Charging < Letdown, Pzr level'will decrease 2.

Pressure decrease 3.

Variable Heaters full on, B/U Heaters on 4.

Letdown Isolates (all heaters off) 5.

Charging > Letdown, Pzr. level will increase 6.

Variable Heaters re-energize 7.

High level Reactor Trip Item #6 should be "Back-up heater re-energize" CECO answer:

System Description, Chapter 14, Figure 14-15, page 45

Reference:

Resolution:

Answer key changed to reflect correct nomenclature.

QUESTION 3.06 Why is the Narrow Range level span on Unit 2 more compressed than b.

l Unit 17 Describe this physical change.

(1.0)

Because of the higher recirculation flow in Unit 2, the S/G NRC answer:

is less sensitive to level transients. The lower narrow range ton is higher.

15ic)

Because of the higher recirculation flow in U-2 S/G's, the CBCo answer:

level span was compressed to prevent level indication fluctuations that might occur as the recirculation flow increased with power.

Reference:

Byron Unit 1 and 2 Differences Book, page 17 Resolution:

Answer key changed to reflect facility clarification, and graded accordingly.

QUESTION 3.10 b.

State RXJR inputs for the Subcooled Margin Monitor. Consider separate redundant transmitters of the same parameter as ONE input.

(2.0}

NRC answer:

b.

1. Pressurizer Pressure (P.T.'s 455, 456, 457 and 458)

2. Reactor Coolant Loop Pressure Wide Range, (Hot Legs A & C (P.T.s 403 and 405))

3. Maximum UHJTC Temperature (from RVLIS processing)

4. Representative CET Temperature (from CEI. processing)

The reference used by NRC was incorrect. SPDS revision CECO answer:

BY3-0, March 25, 1986 supplied by the CECO Computer Group lists:

1.

Containment pressure (PT-934, 935, 936, 937) 2.-

Incore T/C 3.

RC loop 1A and 1C WR pressure 4.

Rx trip bket and bypass breakers -

5.

Barometric pressure 6.

Pressurizer pressure (PT-455, 456, 457, 458) 7.

Containment hi range rad monitors 8.

Turbine impulse first stage pressure (PT-505, 506)

These are all the specific inputs that lead to determining subcooling.

Many are subroutines to develop the actual input to subcooling. The following,

1. Pressurizer preuure -(uo ^=M / *"eesed*

)

• [h therefore, should be the answer:

2. Containment presuure 7 - >s.5 6

3. Containment hi range rad monitors

4. Incore T/C'

Reference:

Safety Parameter Display System Byron Unit 1 l

Resolution:

Question asks for four inputs. Therefore, the question will be graded to accept four of the five inputs as stated above in the facility comment.

QUESTION 4.01 -

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

a. If during performance of BOA-RCP-1 "RCP Seal Failure", and the #1 Seal Bypass Valve needs to be opened, what THREE conditions must exist before opening #1 Seal Bypass Valve (1.0]

1. Seal Injection flow is 8-13 gym NRC answer:

a.

2. #1 Seal Leakoff flow is < 1 gym

3. RCS pressure is < 1000psig IBOA RCP-1 has been revised, FOUR conditions must exist CECO answer:

before the #1 Seal Bypass valve needs to be opened, therefore, any three of the four are acceptable.

1. Seal Injection flow is between 8-13 gym a.

2. No. 1 Seal Leakoff isolation valves on open 3, No. 1 Seal Leakoff flow is less than 1 gym

4. RCS pressure is gs. ster than 100 peig and less than 1000 psig

)

Reference:

IBOA RCP-1 Rev. 51, pg 3 Resolution:

The additional correct response is added to the possible answers.

i QUESTION 4.03 During performance of BOA ROD-4, " Dropped Rod Recovery" prior to recovery of the dropped rod, state ALL methods that can be used to a.

match Tref with Tave.

By reducing turbine load, diluting or moving rods NRC answer:

a.

CECO answer: Manually adjust rods

-OR-Manually adjust turbine load

-OR-Manually adjust RCS boron concentration

Reference:

IBOA ROD-4, pg. 2 Resolution:

The CE Co. answer further clarifies the NRC answer and no change to the answer key is necessary.

QUESTION 4.03 If a dropped rod cannot be recovered immediately, state the THREE conditions or actions, one of which is required to be completed within b.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for power operator to continue.

NRC answer:,

b.

Within 1 hour:

1. Restore rod to operable status, (0.3]

2. Rod is declared inoperable (0.3] and other rods in group aligned within 112 steps, (0.3]

3. Rod is declared inoperable and

a. Tech Spec SDM satisfied,

b. Power reduced to < = 75%

The question is misleading by mentioning the one hour CECO answer:

In addition, it does not restrict the examinee requirement.

to Tech Spec.

IBOA ROD-4 also supplies actions to be taken I

and should be included in the key.

1. Calculate the QPTR, if greater than 1.02, apply Tech Spec 3.2.4

2. Reduce power for dropped rod recovery.

3. Restore rod to operable status

4. Rod is declared inoperable and remainder of rods in the group are aligned within i 12 steps of the inop.

rod while maintaining rod sequence and insertion limits.

5. Rod is declared inoperab,le and SDM is satisfice.

power operator may then continue provided that:

a) Power reduced to < = 75% within the next hour and within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Reactor Flux Trip Setpoint is reduced to less than or equal to 85% of RTD l

b IBOA ROD-4, Rev. 51, pg. 3; Tech Spec 3.1.3.1

Reference:

Resolution:

The question states that there are three actions to be completed within one hour.

The Technical Specifications states these items are to be done within one hour and, therefore, is the basis for the answer.

The items the facility desires, have no time limit associated with them for per-formance (according to procedure 180A R00-4).

However, because the question did not state " Technical Specifications",

the first answer " calculate the QPTR" will be accepted as one action.

The second requested answer (power reduction) was in the original answer l

on the answer key.

l

QUESTICat 4.05 The following concern BAP 300-1, Conduct of Operators

a. As Unit 1 NSO, what constitutes "at the controls"?

NRC answer:

a.

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

a.. The NRC answer is correct, however, it could be answered CECO answer:

as "The At-The-Controls Area" is delineated in BAP 300-1A1. Or a sketch of BAP 300-1A1 may be given.

Reference:

BAP 300-1, Rev 51, pg.10; BAP 300-1A1 Resolution:

The sketch will be allowed as a correct answer.

QUESTION 4.07 Define the following, according to BAP 1450-2, Access to High Radiation

Areas,

b. Hot Spots.

NRC answer:

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.

i

-OR-Areas near equipment or pipes where the DOSE RATE AT 18 INCHES from the source would EXCEED 5 TIMES THE AMBIENT l

DOSE RATE for the GENERAL AREA.

(0.75) l Ceco answer:

b.

Hot Spots:

l Areas near piping or equipment where the dose rate at 18" from the source exceeds five (5) times the ambient does rate for the area.

-OR-Areas near piping or equipment where the dose rate at-I less than 18" from the source exceeds the applicable posted limited for the area.

Clarification - CECO answer to Question 4.07 b. Hot Spots does not include

" General" area just area and talks about a dose rate at "Less Than 18" not l

" Greater Than".

l

Reference:

BAP 1450-2, Rev. 2, pg. 1 Resolution:

Clarification made to answer key and graded accordingly.

e QUESTION 4.13_

According to BOh PRI-6, " Component Cooling Malfunction":

c. If Surge Tank level is INCREASING, STATE FOUR possible leakage sources into the Component Cooling System.

(2.0)

NRC answer:

c.

1.

RCP thermal Barriers.

2.

RH heat exchangers.

3.

Spent fuel pit heat exchangers.

4.

Letdown heat exchangers.

(2.0)

CECO answer:

1.

RCP thermal barriers 2.

RH heat exchangers 3.

Spent fuel pit heat exchangers 4.

Letdown heat exchanger 5.

Excess letdown heat exchanger

Reference:

Byron, IBOA PRI-6, Rev. 51, page 9 Resolution:

Additional accepted answer added to answer key.

t

c

~'

'.-,,~

U.

S.

NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:

_SyEQU_1_&_g_____________

REACTOR TYPE:

_PWR-Wgg3________________

DATE ADMINISTERED: _Sg4QZZ16________________

EXAMINER:

_J6R_ P p;

\\qlgJ o

APPLICANT:

k

,yg L_

IUSIBUGIIONS_IQ_6EELIG6 nil 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 sin (6) hours after the examination starts.

% OF CATEGOR'Y

% OF APPLICANT'S CATEGORY

__266UE_ _I0166

___EGQEE___

_YG6L'E__ ______________GGIEGOBY_____________

1.

PRINCIPLES OF NUCLEAR POWER

_2Dz99__ _2EzQQ PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 24.5

_25z99__

2EzQO

________ 2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

_2Dz99__ _20299

________ 3.

INSTRUMENTS AND CONTROLS

_SUz99__ _2Dz99

________ 4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 99.s TCTALS 122_2C__ Iggz99 FINAL GRADE _________________%

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

APPLICANT'S SIGINTURE

1

~

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the ad=inistration of this examination the folicwing rules apply:

1.

Cheating en the examination means an automatic denial 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 possibility of cheating.

' 3.

Use black ink or dark pencil only 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 categor o

one side of the paper, and write "Last Page{ on a new page, write Jn1 on the7 ast answer sheet.

9.

Nueber each answer as to category and number, for exaeple,1.4, 6.3.

10. Skip at least three lines between each answer.

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

12. Use abbreviations only if they are commonly used in facility 11tarature.

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

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

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

16. If parts of the examination are not clear as to intent, ask questions of 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 cosolating the examination. This must be done after the examination has 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, tantes, etc.

(3) Answer pages including figures which are a part of the answer.

b.

Turn in your copy of the examination and all pages used to answer the examination cuestions.

c.

Turn in all scrao paper and the balance of the paper that you did not use for answering the questions, d.

Leave tre examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

PAGE 2

12__EE10G1 ELE 5_0E_NQGLE@E_EQWEE_E(GUI_QEEE@IlQN 1 10ESdQQXNGdlGS1_SE91_IB6NSEEB_6NQ_ELQ1Q_E(QW QUESTION 1.01 (1.00)

During a Xenon-free reactor startup, critical data was inadvertently taken two decades below the required Intermediate Range (IR) level (1xE-10 amps).

The critical data was taken again at the proper IR level (1xE-B amps).

Assuming RCS temperatures and baron concentrations were the same for each set of data, which one of the f ollowing statements is correct?

a.

~The critical rod position taken at the proper IR level is LESS'THAN the critical rod position taken two decades below the proper IR level.

b.

The critical rod position taken at the proper IR level is THE SAME AS the critical rod position taken two decades below the proper IR level.

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 cri tical rod position taken two decades below the proper IR level.

QUESTION 1.02 (2.00) l Indicate whether the following will cause the differential rod worth ot one control rod to INCREASE, DECREASE or have NO EFFECT.

i 1

a.

An adjacent rod is inserted to the same height b.

Moderator temperature is INCREASED c.

Baron concentration is DECREASED o.

An adjacent burnable poison rod depletes i

I 1

OUESTION 1.03 (1.00)

TRUE OR FALSE?

As Boron concentration increases a.

Moderator Temperature Coefficient becomes less neoative due to increased neutron leakaae.

o.

Mcderator Temperature Coefficient tecomes mcre negative due to the increased resonance auscretion facter, s**++4 CATEGORY 01 CONTINUED ON NEAT PAGE +++++'

PAGE 3

Iz__EB1UGIELE5_9E_bWGLEBB_E9 WEB _ELGUI_9EEE6IIQN 2

IUE6dQDyd851CS3_Sg61_I68NSEg6_6NQ_E(ylp_E(QW 1

OUESTION 1.04 (2.50)

How AND why does the Doppler Defect change as reactor power is a.

increased?

C1.03 b.

How does each of the following affect the Fuel Temperature Coefficient (MORE NEGATIVE, LESS NEGATIVE, or NO EFFECT)?

C1.53 No explanation is desired 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 PU240 over core life.

I OL'EST ION 1.05 (1.50)

Compare the calculated Estimated Critical Position (ECP) for a startup 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after a trip to the actual Critical Rod Position (ACP) if the following events / conditions occurred.

Consicer each independently.

Limit your answer to:

a.

ACP higher than ECP.

4

b. ACP lower than ECP.

ACP would not be significantly different than ECP.

c.

1.

One Reactor Coolant Pump is stopped one minute prior to criticality.

2.

The steam dump pressure setpoint is increased to a value just below the code saftiss setpoints.

3.

The startup is delayed 2 more hours.

i

(**++* CATEGCRY 01 CONTIfluED ON NEX T F AGE + + + + <l

PAGE 4

1s__EBluGIELEE_QE_UUGLE68_EQ' DES ELBUI_QEESSIl002 ISEBdQQXN@d1CQi_dE81_188NEEg8_6NQ_E(ylQ_E(QW OUESTION 1.06 (1.00) answer from the choices Complete the sentence by choosing the correct below.

Delayed neutrons play a major role in the operation of the core because they a.

are born at (thermal) slow energy levels (less than 1 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 the core.

c.

are born so much later than the prompt neutrons and provide controlability during steady 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 reutrons.

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?

L1.03 b.

Will it take the same amount of time to raise power from 40%

to 60% i f the same startup is maintained?

EXPLAIN.

[1.53 rde.

t'..+++ CATEGGP) 01 CONTINUED ON t4E(T PAGE *++++)

PAGE 5

Iz__EBING1 ELE 5_QE_NUQLEGB_EQWEB_EL6MI_gEg6911QN 2 IbEEDQDYN001G52_UE01_IB8N5EEB_6ND_ELVID_ELQW d

OUESTION 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.

The doppler effect adding positive reactivity due to the d.

temperature decrease following a trip.

QUESTION 1.09-(1.00)

Part of the reactor thermal safety limit is based.upon not allowing i

saturation conditions at the core hot leg.

State.the reasoning behind this.

i, OUESTION 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.

a.

Nucleate bolling.

b.

Accident condition in which cociant is boiled and converted i

to steam in the reactor vessel.

c.

Heat from fission thru the fuel rod.

4 d.

Decay heat removal by natural circulation of coolant, e.

Decay heat of fission products to clad surf ace.

4 i

(+++++ CATEGORY 01 CONTINUEL ON NEXT FAGE *++++>

r PAGE 6

it__ESluCIELEE_QE_UUCLEGS_EQWEB_ELGUI_QEEB611QN i ISEBdQQXU951GEx_bEBI_ISBNEEEB_GNQ_ELUIQ_ELOW OUESTION 1.11 (1.00) from the choices Complete the sentence by choosing the correct answer below.

The 2200 degrees F maximum peak cladding temperature limit is used because a.

it is 500 degrees F below the f uel cladding melting point.

b.

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

a =ircalloy-water reaction is accelerated at temperatures above c.

2200 F.

d.

the thermal conductivity of rircalloy decreases at temperatures above 2200 F causing an unacceptably sharp rise in the fuel centerline temperature.

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?

a.

Enthalpy decreases, entropy decreases, quality decreases.

b.

Enthalpy increases, entropy increases, quality increases.

c.

Enthalpy constant, entropy decreases, quality decreases.

d.

Enthalpy decreases, entropy increases, quality decreases.

+++++ CATEGOhi 01 CONTIhUED ON NEXT PAGE *++++)

T PAGE 7

Iz__E 510GIE L E 5_9E_ UUC L E 86_ E 9 W E B_EL GUI_9E EB6I1902 IUEEd90XN@dIQg2_dg@l_IE983EEE_9ND_ELUID_ELOW OUESTION 1.13 (3.00)

Since DNB cannot be measured directly, what FOUR parameters are a.

monitored to assure that DNB is not exceeded?

C2.03 b.

Assuming the reactor is operating at 85% power indicate how the following changes in the plant condition would affect DNBR (INCREASES, DECREASES, REMAINS THE SAME),

Consider each case separately.

E1.03 1.

.The operator withdraws control rods without changing turbine load.

2.

Steam Generator PORV fails open.

3.

Reactor Coolant pressure increases.

QUESTION 1.14 (1.50)

Use the stet.- tables and associated Mollier chart to answer the questions below, label quantites with proper units.

During cooldown and depressurization, you are required to remain 50 a.

degrees F subcooled.

As the pressure decreases through 2005 psig, what is the maximum Tavg' allowed (nearest degree F)?

A thermocouple (TC) b.

Steam is leaking f rom a pipe flange into a room.

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 inside the pipe was 560 psia, what would you estimate the pressure steam temperature to be at that pressure?

i+++++ CATEGOFY 01 CONTINUED ON NEXT PAGE +++++

8 PAGE B

12__EB1NCIELEE_QE_UWGLEBB_EQWEB_ELONI_QEEBelighi ISEEdQDXW6d1QEi_dEGI_IE6NSEE8_66Q_ELQ1D_ELQW OUESTION 1.15 (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.

Following a reactor trip from equilibrium conditions, Xenon peaks b.

because delayed neutron precursors continue 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.

QUEETION 1.16 (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.

CAn answer may be used more than once.]

It takes about ____ hours to reach the maximum Xenon concentration a.

after a reactor trip.

b.

The decay half-life of Xenon 155 is approximately ____ hours, c.

It takes about ____ hours to reach equilibrium Xenon concentration after a step increase from 0 to 50% power.

d.

The decay half-life of Promethium 149 to Samartum 149 is acproximately

____ hours.

Available Answers:

0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />s:

5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />s:

10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />s:

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 />; 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />.

i

(++++* END GF CATEGOPY 01

• ++++s

PAGE 9

St__ELGUI_DE5106_INGLUDING_56EEIX_GUQ_EdEEGEUGY_SX$1Ed5 OUESTION 2.01 (1.50)

A seal water heat exchanger outlet high temperature condition exists.

a.

Other than low CCW flow, list TWO other causes of this condition.

can the Unit 2 operator verify that low CCW flow is not a possible b.

How cause?

c.

How can the Unit 1 operator verify that low CCW flow is not a possible cause?

s. Ko OUESTION 2.02 Creec)

TRUE OR FALSE 7 The following concern the Pressurizer Power Operated Relief Valves (PORV).

a.

One PORV is suf ficient to prevent exceeding the f racture toughness limits of 10CFR50 Appendix G when water solid.

b_

c-eeeu-ize-coo"'c 2rr cquired for overper Ourc pretcetic-du-ing

-I rr t perature c:2t:r Oclid Optreti;r.:.

the Pressurizer Code Safety Valves considers proper operation ha.

Siring of of one Pressurizer PORV.

load without relieving The pressuri:er can sustain a complete loss of c.1W.

water, if at least one PORV operates properly.

<+*+++ CATEGGRY 00 CONTINUED ON NEXT PAGE

+++++t

=

PAGE 10 2___EL6UI_DESIEU_INCLUDINQ_$$E[IIY_@Up_Edg85ENGY_SYSIEd5 s

OUESTION 2.03 (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 i

4.

________ PSIG 5.

________ F (divert setpoint) 6.

________ GPM (maximum allowable for each kind) 7.

________ GPM B.

_ _ _ _ _ _ _ _ G P M p a c1r e, 7 u l

9.

________ GPM

10. ________ GPM

11. ________ F

12. ________ GPM p see o. T M 1

U OUESTION 2.04 (3.00)

The following concern the Reactor Makeup Control System.

a.

State the maximum flow rate (in gallons per minute) allowed by the Boric Acid Flow Controller.

CO.53 s

J l

b.

State the flow rate (in gallons per minute) out of the blender if the makeup system is in automatic.

CO.53 I

c.

At what level is automaLic makeup to the VCT started and stopped?

C1.OJ

,f j

d.

State all conditions that will generate a " flow deviation" alarm.

C1.OJ OUESTION 2.05 (2.00)

Cencerning BTRS. state the mattimum Dilution AND Boration rates (in ppm /hr) for botn BOL AND ECL conditions.

~

CATEGORY O2 CONTINUED ON NEAT FuGE +++++)

(+++++

i i

PAGE 11 O __ELGUI_DESIGU_INGLUDIUD_SGEEIZ_GUD_EdEEGEUGY_SYSIEd5 OUESTION 2.06 (2.00)

The following concern valves in the Residual Heat Removal System.

State the FOUR conditions that must be satisfied in order to a.

RHR Suction Isolation Valves f rom RCS open valves 8701A and 8702A,

[ 1. 0 3 ( I.wk ).ek s, co* Aa mwWh WW oad loops.

b.

State the TWO signals that will close these same valves.

CO.53 State the TWO requirements that must be present in order for valve c.

8811A, Suction Valve from the Containment Sump, to open automatically.

CO.53 OUESTION 2.07 (1.50)

I and Unit 2 State the pressure source used to pressuri:e the Unit than pressurizer FORV accumulators.

Why is the source for Unit 2 different that of Unit 17 OUESTION 2.08 (1.50)

Unit 2 has two additional installed solenoid operated centrifugal Charging Fump mini-flow recirc valves, 2CV8114 and 2CV811e.

a.

What signal and setpoint will automatically 1.

Close 2.

Open these valves?

C1.03 b.

Why were the additional valves installed?

CO.53 OCESTION 2.09 (2.00)

State, for each of the below, if they are ACTIVE or FASSIVE failures, a.

Failure of a cump to start.

t.

Loss of packing in a valve.

c.

An electrical relay does not respond.

d.

A valve stays open when called on to close.

(**+++ CATEGORY O2 CONTINUED ON NEXT FAGE + +++)

PAGE 12 8t__ELeUI_DEE1Gu_1GGLUDIUQ_SeEEIY_eUD_EdEE9EUGY_5XSIEd5 OUESTION 2.10 (3.00)-

a.

Following a reactor trip, an "Overcrank" alarm is received on the 1B Auxiliary Feedpump.

List the sequence of JHMC events that occurred to receive this alarm.

[2.03 b.

Other than "Overcrank", list FOUR other conditions that will trip and lockout the 1B Au::iliary Feedpump.

(Setpoints not required.)

C1.03 OUESTION 2.11 (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 your answer:

a.

The NAME of the system, AND b.

1.

The DESIGN flowrate (gpm) and associated pressure, AND The MAXIMUM flowrate (gpm) and asscciated pressure.

OR 2.

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

(+++++ END OF CATE6ORY O2 +++++)

~

z PAGE 13 7t__JNSIggdgNI5_@BQ_QQUIBGL3 QUESTION 3.01 (3.00) a.

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

[1.03 b.

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

When is it energized?

[1.53 c.

TRUE or FALSE?

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 recuired, would the SI signal reset?

Explain your answer.

QUESTION 3.03 (2.00)

The following concern the Remote Shutdcwn Panels.

TRUE or FALSE?

a.

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

Reactor Coolant Pumps cannot be started f rom the Remote Shutdown b.

Panels.

c.

If local control cf 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.

(+++++ CATEGORY O! CONTINUED ON NEXT PALE ++++++

PAGE 14 t__JNgIBUDEUI5_ANp_ggNTBQLS OUESTION 3.04 (2.50) a.

One of the selected Pressuri:er Pressure Channel signals passes through Proportional Integral

("PI") controller.

State 2de FOUR pressuri:er components that are operated by this signal (be specific).

C1.03 b.

What are the TWO specific input control signals for each Pressurizer PORV 455A AND 456, when selected for Cold Overpressure Protection?

C1.0J If the pressure sources to the Pressurizer FORV's are lost, c.

appro>:imately 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 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 cnanges are made.

List the sequence of SEVEN events that will take place ending in a trip or SI with no operator action.

Be specific. include all automatic functions, no setpoints required.

QUESTION 3.06 (2.50) a.

State the S/G Narrow Range level setpoints (in percent) for the following:

[1.53 UNIT 1 UNIT 2

-=

High High Level Trip -

Normal Operating Level at 100% Power -

Lo-Lo Level Trip -

l b.

Why is the Narrow Range level span on Unit 2 more compressed than l

Unit I?

Describe this physical change.

C1.03 l

t I

i l

I l

i f

l

[

(+++++ CATEGORY 03 CONTINUED ON NEAT PAGE ++++->

l l

i PAGE 15 It__1U51BWDEUIS_eUD_G9dIB9L5 OUESTION 3.07 (3.00)

State the inputs that are used to generate the Power Mismatch a.

signal in the Reactor Control Unit.

[0.53 b.

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

[1.OJ c.

The Summing Unit can only function using temperature signals.

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

[0.53 d.

Which one of the below compensates the Reactor Control Unit for reactivity Changes?

[1.03 1.

Variable Gain Unit.

2.

Non-Linear Gain Unit.

3.

Lead-Lag Compensator.

4.

Rod Speed Programmer.

QUESTION

3. 08 /3.09 (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 l

applicable).

b I

PAGE 16 st__IUSIEUMEUIS_ Gyp _GQUIRgt S l

I OUESTION 3.10 (3.50) 1 t

Briefly describe how the Reactor Vessel Level Indicating System a.

detects a vessel level change.

[1.53 b.

State FOUR inputs for the Subcooled Margin Monitor.

Consider separate redundant transmitters of the same parameter as ONE input.

[2.0 QUESTION 3.11 (2.00)

For each type of 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.

i b.

Gaseous, c.

Particulate (Gas streams).

d.

Iodine (Gas streams).

l i

i i

i i

I I

(*++** END OF CATEGORY 03 +++++)

PAGE 17 dz__E50GEDUEES_:_UDED662_6800EdeL2_EUEBGENCY_SNQ 60D196991G96 G901696 OUESTION 4.01 (1.50)

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

If during performance of BOA RCP-1, "RCP Seal Failure", and the

1. 1 Seal Bypass Valve needs to be opened, what THREE conditions must e::ist bef ore opening #1 Seal Bypass Valve?

[1.03 b.

If during performance of BOA RCP-2, " Loss of Seal Injection", what TWO conditions must exist such that the #1 Seal Bypass Valve must be closed?

[0.53 OUESTION 4.02 (1.00)

Which ONE of the following statements, concerning Technical Specifications actions required for Nuclear Instrument malfunctions, is correct?

a.

If a source range channel fails while a startup is in progress end reactor power is belcw P-6, insert all control banks to

ero steps.

intermediate range channel fails while a startup is in b.

If an the progress and reactor power is above P-6 but below P-10, increase may continue using'the operable intermediate power range channel, Failure of one power range channel during shutdown precludes c.

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

c.

Failure of both source range channels while shutdown requires shutdown margin requirements to be verifled within i hour.

I OUESTION 4.03 (2.00)

Durino the performance of 50A ROD-4, " Dropped Rod Recovery".

a.

prior to reccvery of the dropped rod, state ALL methods tnat can be used to match Tref with Tave.

i b..

If a dropped rod cannot be recovered immediatelv. state the THREE conditions or actions, one of which. is required to ce completed within 1 h ot.t r, for power operation to continue.

l

(+++++ CATEGORY 04 CONTINUED CN NE/T FAGE +++++,

PAGE 18

$t__ESQCEQUBES_ _NQEMA62_ARNQEM9L,_gMESGENQY_AND 69D196001G96_GQNIBQL QUESTION 4.04 (2.00)

The following concern information found in BOA PRI-2, Emergency Beration.-

a.

State the TWO conditions which if either are encountered, while in mode six, would require Emergency Boration.

CO.63 b.

If Emergency Boration flow of 30 GPM is required, state the THREE flowpaths that are available.

QUESTION 4.05 (3.00)

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

a.

As Unit 1 NSO, what constitutes "at the controls"?

C1.03 b.

What must be done if the NSO must leave the "at the controls" area

~

during non-emergency conditions?

Does this also include going behind the Main Control Board for a valve manipulation or reading?

[0.753 c.

1.

Who has the authority to allow the NSO to leave his "at the controls" area to assist the other unit?

[0.53 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 NSO.

[0.753 OUESTION 4.06

(.50)

TRUE or FALSE?

A Safety Injection Pump with its control switen in " pull-to-lock",

is still considered operable if a dedicated operator is stationec at its control switch.

(++++* CATEGGb'Y 04 COf 4TINUED ON NEXT F AGE

++++-J

=

I PAGE 1@

4___EEOGEDUEE5_:_NQEU@L2,$@NQEUg64_gDEE@gNgY_9ND 60DIQLQQ1G66_G9NISQL QUESTION 4.07 (1.50)

Def ine the f ol' lowing, acccrding to BAP 1450-2, Access to High Radiation Areas.

a.

High Radi ation Area.

b.

Hot Spots.

QUESTION 4.08 (2.00)

State the DAILY whole body dose limit for any individual at a.

. Byron 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?

E1.O]

c.

If the limit in b., above, needs to be exceeded, who must approve this additional increase?

EO.53

-QUESTION 4.09 (1.00)

Assume the plant has experienced a small LOCA, SI h'as been initiated, Procedure 1BEP-1, reset, and only the charging pumps remain running.

Loss of Reactor or Secondary Coolant, is in effect.

What action would be required if pressuriner level began to decrease and could not be maintained above 4%.

QUESTION 4.10 (2.50)

State the THREE conditions, if one of which existed, that would require the NEO to trip the RCP's when procedure 1BEP-0, Reactor Trip or Safety Injection, is in affect.

l 1^

(+++++ CATEGCRy 04 CONTINUED ON NE.t 7 F AGE +++ ++ )

l

l PAGE' 20 Sz__tB9CEDUSE5_:_U9600L2_GENQBU662_EMEBGEUGy_6ND 860196001GG6_GQNIBQL QUESTION 4.11 (0.00)

State the SIX Critical Safety Function Status Trees (CSFST) a.

in order of priority.

Include the single letter designator assigned to each tree.

C2.43 b.

True or False?

An ORANGE PATH in the CORE COOLING CSFST takes priority over a RED PATH in the CONTAINMENT CSFST.

CO.63 OUESTION 4.12 (1.50)

Describe the effect that a loss of DC Bus 111 will have on the f ollowing.

Feed Water Regulating Valves.

a.

b.

Feed Water Regulating Valve bypasses.

c.

Pressurl:er PORV accumulator Nitrogen supply.

QUESTION 4.13 (3.50)

According to BOA PRI-6, " Component Cooling Malfunction":

a.

If Surge Tank level is DECREASING, at what LEVEL must operator action be taken?

EO.53 b.

State TWO of the 4 actions that must be taken in

'a' above.

[1.03 c.

If Surge Tank level is INCREASING. STATE FOUR possible leakage sources into the Component Cooling System.

E2.03

(+++++ ENE CF CATEGCFY OA ++++++

END OF EXAnINATICN +++++++*+++++++)

(+++++++++++++

EQUATION SHEET Cycle efficiency = (Net w G v = s/t f = ma cut)/(Energy in) z s = Y,t + 1/2 at

,=y 2

.x A = A e' A = \\M c

E = 1/2 mv a = (Vf - 1 )/t 3

PE = agn t = us2/t1/2 = 0.693/t1/2

+ at

• = */t Yf=Y C

eff = ((tt f.,)( ts) 3 2

1/Z

((c/2)*(*b)3 3$

N*"#

A=

1

-Ex

'I

• 93 I "

m = Y,yAo g,g Q = nCaat I = I e~"*

c Q = UAa.T I=I 10-x/TVL a

pwr = W sh f

TVt. = 1.3/u sur(t)

HYL = -0.!;.iin p. p to p,p,t/T SG = S/(1 - K,g)

SUR = 25.06/T G.= S/(1 - K,gx) x G (1 - K,ff3)

• G II ~ "eff2) j 2

SUR = 25s/t* + (s - o)T T = (t*/a) + [(a - sy Io]

M = 1/(1 - K,g) = G)/G, M = (1 - K,g,)/(1 - K,ff))

T = 1/(s - a)

SUM = ( - K,g)/K,g T = (a - o)/(Is) t' = 10 seconcs a = (X,g-1)/K,g =.:X,g/K,g I = 0.1 seconds-I o = ((t*/(T K,y)3 + CI,ff (1 + II)3

/

Id1i*Id P = (:4V)/(3 x 1010)

I)d; 2 =2 2 Id 22 A/hr = (0.5 CZ)/d'(meters)

R/hr = 6 CE/c2 (feet) g = :n Miseslianeous 0:nve.sions Water P ar*. meters 10 I curie = 3.7 x 10 cas 1 gal. = 8.345 tem.

1 %g = 2.21 Itm 3 Stu/hr 1 gal. = 3.78 11:ars 1 no = 2.54 x 10 1 fd = 7.48 gal.

1 mw = 3.41 x 100 5tu/hr 3

Oensity = 62.4 lett/ft lin = 2.54 ::n Gensity = 1 g:n/c9

• F = 9/5'C + 32 Heat of vacorization = 970 Stu/lem

'C = 5/9 (**-32)

Heat of fusion = 144 Stu/lem 1 STU = 778 ft-lbf 1 At:s = 14.7 csi = 29.9 in. Hg.

1 ft. H O = 0.4335 ltf/in.

2

s e__. !

o-x Q- 'C w

~ m f

f V

N i

1 f

}

E q *)

2

~~ '

b f

.b i

5 s

"1 3

.=

_.N -

\\'

0,-

02 :

@e

~i

-e 0-X E

a

.u M

N S

Ge 4

oqtra 0-qD=

=

4

Qa O_ _.

A.._

,=

1 i

W(E 9

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

.{

~

ulC UPPER UIC LOWER METER RANGE l

METER R ANGE A

OETECTOR e

DETECTOg l

SELECTOR SHuMTS h TEST SIGNAL S

.I.5 1 See

.I 5 1 See

)

4

.S I '

AMMETER HIGH VOLTAGE AMMETEli l**

b POWER SUPPLY g

4 >

300-1500,4s v

4,

,y 2

s.

J

'au O'

I I

~E~

ISOL b

~

AMP i MP GAINAOJUST '

A

--e @

~

+-

-e@

e-

-e. @

2 /4; H;

g E

OTHER THREE 1/4 g PR CHANNELS *

~

2*/4 10 % PWR.

j-j j ISOL j,, AUCTIONEER h

r 2/4 J

g CIRCuti POWER MISMATCH

""~'~

DEFE AT SWITCH l

2/4 ; 100% PWR.

r ISOL

'g h

I/4 J

AMP I

l L_, g A

~

l a

i 1/4 2/4; 30% PWR.

g b ;iO3 % PWR.

I j

t

,i4 g% 5 I

@D ROD STOP BYPASS SW.

ADJUSTAGLE j

+ S% IN 2 SEC.

I /4

[

]

I i

% FULL POWER ON NIS PANEL

~

l COMPMI

-5 % IN 2 SEC.

.ATOR (0-120 %)

COMPARES PWR. LEVEL h

I 3/4

[

NOW WITH WHAT IT WAS s

2 SEC.*S AGO.

i Figure 33-1

• n,

~

PAGE 21 Iz__EBINGIELES_9E_UUCLE86_E9 WEB _EL9NI_9EEBBI1QN 4 ISEBU99XU651GS2_UE9I_IE9NSEg8_gNQ_E(yIQ_E(QW ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

cw ' '

pr k

, CT $ Wi C

Y#W ANSWER 1.01 (1.00) b REFERENCE NUS, Nuclear Energy Training, Module 3, Unit 6 Westinghouse Reactor Physics, Sect.

3, Neutron Kinetics and Sect.

5, Core Physics HBR, Reactor Theory, Sessions 20 and 24 - 31 Zion, NUS book 3, sections 6.5, 6.6, 6.7, 12.4, 12.5.

BYN, Westinghouse Large PWR Core Control, Ch.

7, Pp. 24-34.

ANSWER 1.02 (2.00) a.

Decrease b.

Increase c.

Increase d.

Increase CO.50 each]

REFERENCE SON /WBN License Requal Training, " Core Poisons" BYN, 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 PEFEFENCE IP-3 EC1 Ex Theory: Chapter 7. Pages 21, 22, and 27 DCC R:: Theory Review Text, pp. 1-5.42

.50 EHNP, RT-LP-1.13. Pp. 11-18.

ROWE Reactor Operator Training Manual, p.

3-237 BYN, Westinghouse Large PWR Core Control, Ch.

3.

4 l

l l

~

PAGE 22 It__EBINGIELES_DE_UUGLEGB_EDWEB_ELGUI_DEEEBIIQN 2

IdEEUQQyN951Q$2_dg8I_IB6USEEB_9dp_ELylp_E6QW

-86/07/16-JAGGAR, F.

ANSWERS -- BYRON 1 ANSWER 1.04 (2.50)

(aasng.. cm Abe.. mss.% W rwe aus As power increases, fuel temperature increases, the Doppler Juer.%we

• ".jg Q 3**

a.

Defect becomes more negative C1.OJ g

Les s b.

1.

Meee negative.

2.

5@ Sis negative.

CO.5 each]

3.

More negative.

REFERENCE CPLL Reactor Theory Chap. 14 pp. 14-7 thru 14-11 ROWE Reactor Operator Training Manual, Sec. 3,.pp 189-202

~

Ch.

2, Pp. 23-48.

BYN, Westinghouse Large PWR Core Control, ANSWER 1.05 (1.50) 1.

c (same) 2.

a (ACP higher) 3.

b (ACP lower)

CO.5 ea.]

REFERENCE SONP, Review of Core Poisons, pp. 4 - 7 KAOO1/OOO,K5.18,4.2.

Cook Theory, Pp.

I-36-45.

Zion, NUS book 3, section 12.5.

BYN, Westinghouse Large PWR Core Control, Ch.

7, Pp. 24-34.

ANSWER 1.06 (1.00) c.

REFERENCE KAOO1/OOG.K5.49,2.9.

SONP. Review of Neutron Kinetics, p.

5 Cook Theory, Pp.

I-3.3-10.

Zion. NUS book 3, section 5.5.

BYN. uJestinghouse Large PWR Core Control, Ch.

7.

Pp. 23-30.

PAGE 23 1&__EEldCIE6E5_QE_NQQ(E68_EQWEB_E(@N1_QEEB@llgG1 ISESdQDYded1GSz_SEGI_ISBNSEEB_@UQ_ELylp_ELgW ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 1.07 (2.50)

a. 36 seconds.

(+/-

2)

E1.03

b. No.C.53 Power escalation is a log function and therefore increases at an increasing rate.

E1.03 REFERENCE KAOO1/010,K5.37,3.2.

Cook Theory, Pp. 13.15-16.

Zion, NUS book 3, section 6.4 BYN, Westinghouse Large PWR Core Control, Ch.

7, Pp. 12-22.

ANEWER 1.08 (1.00) a REFERENCE VEGP, Training Text, Vol.

9, p.

21-47 I-3.17 & 19 Westinghouse Reactor Physics, pp.

DPC, Fundamentals of Nuclear Reactor Engineering, p.

106 BYN, Westinghouse Large PWR Core Control, Ch.

7, Pp. 23-30.

OO1/OOO-K5.49 (2.9/3.4)

ANSWER 1.09 (1.00) at the hot leg) further (If saturation conditions were allowed to exist increases in core heat output would be undetected by the hot leg RTD E0.53 and protection would be degraded CO.53.

(=% hJic h ot'e w d REFERENCE fer PWR., Ch.13, Fp.13-53.

Westinghouse Thermal-Hydraulic Principles ANSWER 1.10 (3.00) a.

Convection l

b.

Radiation / convection (l arge Del ta T) c.

Conduction d.

Convection (natural) e.

Conduction (o, r 4 WMow 4p c\\md e wd ce*Ed* EO. 60 each 3 hewgkc\\eJ) l l

l l

I

~

}

PAGE 24

~1c__EB10GIELE5_9E_UUGLEGB_EDWEB_EbeUI_9EEEGI1982 IbEBdQDyd$dIQQ2_dgeI_IgeN@EEB_@NQ_ELylp_E(QW

-86/07/16-JAGGAR, F.

ANSWERS -- BYRON 1 REFERENCE Ch.

3.

BYN, Westinghouse Thermal-Hydraulic Principles for PWR, ROWE Reactor Operator Training Manual, Sec.

2, pp 64-69 ANSWER 1.11 (1.00)

C REFERENCE MNS Thermo-Core Performance, p.2.

Ch. 13.

BYN, Westinghouse Thermal-Hydraulic Principles for PWR, ANSWER 1.12 (1.00) d REFERENCE MNS OP-SS-HT-2, p.12.

Ch.

7.

Westinghouse Thermal-Hydraulic Principles for PWR, B(N, ANSWER 1.13 (3.00)

a. Nuclear Power, RCS temperature (Tave), RCS Loop Flow,

[0.50 each]

RCS Pressure.

b.

1. DNBR decreases 2.DNBR decreases 3.DNBR increases

[1.OJ REFERENCE McGuire Question Bank Ch.

13.

Westinghouse Thermal-Hydraulic Principles for PWR,

BYN, ANSWER 1.14 (1.50) a.

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

b.

424 degrees F of superheat per superheat tables.

its no E0.5 each]

c.

500 degrees F 4io -Sio FEFEPENCE Steam Tables and Mollier chart

i PAGE 25-Iz__ESIUCIELEE_QE_UWGLE86_ERWES_ELGUI_QEEE811QN 2

ISEEdQDYNGd1GSi_dEGI_IBONSEEB_@NQ_E(y1Q_E(QW

-86/07/16-JAGGAR, F.

ANSWERS -- BYRON 1 ANSWER 1.15 (1.00) d.

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

Reactor Theory, Sessions 38 and 39.

.HBR,Section VI.

DPC, Fundamentals of Nuclear Reactor Engineering, 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 />).

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 />).

[0.25 each]

d.

50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.

REFERENCE Westingouse NTO Nuclear Physics, pp.

I-5.70-79.

PAGE 26 82._ _ EL G NI_ pg SJ QN_IUC L UQJ NQ_ S6E EIX_6UD_ E dg B GENQX_ SX SIEDS ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 2.01 (1.50)

1. n; w(oe 33)Labb. N

1. n eew T=,

a. s. Hi temperature CV pump recirculation flow.

(Tu,

..g'.! ]

4.E:: cess #1 seal leakoff flow /Te= &W e-CO.25 each]

b.

By noting the correct CCW flow on the MCB meter.

C O. 5,1 By checking the correct CCW flow locally.

CO.53 c.

REFERENCE Byn, Byron Differences Book, PP. 14, 27 1.50 ANSWER 2.02

( 2. :.,0 ;

a.

TRUE t.

TRUE

b. e.

' FALSE CO.5 each]

c. e.

FALSE REFERENCE

BYN, S.D.

Ch.

14, Pgs. 12-14 ANSWER 2.03 (3.00) 1.

75 5.

138/s*3 9.

12 2.

2235 6.

120 ' Mi::ed, 75 Catton 10.

55 3.

557-55 7.

87 11.

500-534 4.

370t5 8.

32 or 8 per RCP (L-0) 12.

3 per RCP or 12 CO.25 each]

REFERENCE

BYN, S.D.

CH.

15a., Figure 15a-19 i

i I

PAGE 27 S&__EbeUI_DE@l@y_1Ng(ygly@_g6Egly_68Q_Edg$@gNQX_S131E05 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 2.04 (3.00)

E 0. 5.3 a.

40 GPM E0.53 b.

120 GPM c.

Start -37% VCT level

[0.5 each3 Stop

-55% VCT level d

BA flow deviation:

+/

.8 gpm of setpoint PW flow deviation:

+/- 8 gpm of setpoint E0.5 each3 REFERENCE BYN, S.D.

Ch.

15b., Pgs. 36-38 ANSWER 2.05 (2.00)

Boration:

BOL - 40 ppm /hr

'h M. b b EOL - 20 ppm /hr M - 1x t'L u L.,C0.5 each3 4

suo Dilution:

BOL - 20 ppm /hr h

.+, 2 A Eob udh E0.5'each3 EOL - 10 ppm /hr REFERENCE BYN, SD. CH. 16, Pg. 25 ANSWER 2.06 (2.00) a.

1.

8812 Ahclosed 2.

G804 Asclosed 3.

8811 Asclosed E0.25 each]

4 RCS Pressure </= 3o0 psig (Coen signal from MCB) b.

1.

Close signal from MCB 2.

RCS pressure at 662 psig (MCB switch in Auto)

[0.25 each3 c.

(MCB switch in Auto) 1.

"S" signal present E0.25 each]

2.

9&+ Lo-Lo l e' vel s i n RWST REFEFENCE BYN.

5.D.

CH. 18. Pgs. 17-18

PAGE 28 Et__ELGUI_DEE106_INGLUDIUQ_EGEEIY_00D_EUEEGEUGY_EYEIEd5 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 2.07 (1.50)

Unit 1 - Nitrogen CO.5 each]

Unit 2 - Instrument Air Instrument Air is less e>; pensive #(constant losses).

[0.5]

REFERENCE Byn, Byn Differences Book, P.

11 ANSWER 2.08 (1.50)

,Close - RCS pressure 1448 (+/-10) with SI signal a.

Open - RCS pressure 1643 (+/-10) with S1 signal

[0.5 each]

b. s,To prevent dead-heading the CCP's in a low RWST level situation with high RCS pressure.foaO Eth-5-]

2.T..N ue 4w E to - wm %w (2.cs pes sm. T o.zs3 REFEFENCE Byn, Byn Differences Book, P.

12 ANSWER 2.09 (2.00) a.

Active b.

Passive c.

Active CO.5 eacn]

d.

Active REFERENCE

BYN, S.D.

CH. 26. Pg. 7

FAGE 2C Et__EbeUI_DE51GU_lNGLyQlN@_E6Egly_6NQ_EDEEGENCy_S15IEUS ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

2 ANSWER 2.10 (3.00) a.

1.

Engine received an AUTO START SIGNAL [o.10 l' GI 2.

Starting motors engaged and CRANKED FOR 5 SEC

% ce. M b= M o S*c b 5 2.

Engi ne DID k!OT Aru1Eug -.mn npM tu m crrm 4.

10 SECOND TIME DELAY ACTIVATED.

g ag4 l

e,w,. g.

5.

ANOTHER 5 SECOND CRANK ATTEMPTED.,

6.

STARTING CYCLE ATTEMPTED 4 TIMES.lo.s3 EO.25 for each item; O.5 for proper sequence]

High Water Temperature (2os*D b.

1.

2.

Low Oil Pressure 60psy Over speedJioo.p-J 3.

(Low-Low) Pump - Suction Pressurebi 54C %g v..)

CO.25 eachJ 4.

REFERENCE

BYN, S.D.

CH. 26, Pgs. 16-17

[t io % v.L acr@N ANSWER 2.11 (3.50) a.

1.

Centrifugal Charging Pumps:

b.

300 gpm (150 each) @ 2500 psig 1100 gpm (550 each) @ 600 psig

[0.5 each]

2.

Safety Injection Pumps:

b.

800 gpm (400 each) @ 1200'psig 1300 gpm (650 each) @ 800 psig CO.5 each]

3.

Residual Heat Removal Pumps:

b.

6000 gpm ( 000 each) @ 165 psig 10000 gpm (5000 each) @ 125 psig CO.5 each]

(>c t-7 tiv 4.

Accumulators:

b.

28,000 Gals.(approximately h each) trI4 ps CO.53 appr ox i mat el y (cou-w)i g gg_63oA h,g REFEFENCE EW N. S.D.

CH. 56. Pgs. 22-27

~6W " Tech. Spe >

=

Ri__1USIEUdEUIE_GUD_GQUISOLS PAGE 30 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 3.01 (3.00) a.

(It means that it) will require 2 of the 4 possible inputs to activate the particular function.

[1.03 b.

To insure the Reactor Trip Breaker opens if the UV coil fails to open it.

CO.753 It is energized by m -e4-4he4-u1 tri ps. Eoa 53 rai tc 5 -40.751

.u c.

True

[0.53 REFERENCE BYN,S.D. CH. 60A, Pgs. 11, 16 ANSWER 3.02 (1.50)

Yes E0.53:

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

[1.03 N., ace @ Trow' M h W WaM 6 <uh d u enbaa 4%. p4.,

cA-A T 84 P hun 4*--

BYN, S.D.

CH. 61, Figure 61-17 ANSWER 3.03 (2.00) a.

TRUE b.

TRUE c.

TRUE d.

TRUE E.5 ea]

REFERENCE

BYN, S.D.

CH. 62, Pgs. 14-17

PAGE 32 2t__16518UdEUIE_GUQ_GQUIBOLE ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 3.04 (2.50) a.

1.

PORV 455A 2.

All Back-up Heaters 3.

Variable Heaters

[4,egJ3 4.

1. 1 and #2 spray valves E0.25 each]

4 5.

b.

455A - 1.

W.R.

Pressure 2.

W.R.

Low Avg Tcold It % p*> v.ooru.A e3 - w to.G 1.

W.R.

Pressure 6 T. 551 456 2.

W.R.

Low Avg That CO.25 each]

A u. + =.....i c.

7 (+e wu Ate b

s0

/- 3), Open

[0.25 each]

. = 1. I. 7.M e Soc Ait REFERENCE BYN, S.D. CH. 14, Figure 13a, 13c, Pg. 22

%W.

t.m.s %,S w..t L.,

I? h su ptcic.

ANSWER 3.05 (2.00) 1.

Charging s Letdown, P:r. level will decrease EO.53 CO.13 2.

Pressure decrease 3.

Variable Heaters full on, B/U Heaters on CO.13 4.

Letdown Isolates (all heaters off)

CO.63 5.

Charging > Letdown, P:r. level will increase.

[0.13 CO.13 6.8/#a -i 91' Heaters re-energi=e EO.53 7.

High level Reactor Trip REFERENCE BYN, S.D.

CH. 14, Figure 14-2, 14-4 ANSWER 3.Oo (2.50) a.

UNIT 1 UNIT 2 81.4%

78%

66%

50%

40.8%

17%

(all values +/-1%)

[0.25 each]

recirculation flow in Unit 2.[the S/G is b.s.Because et the higher transients.)neeebe. & WaN g g h g J b

• O 9e*

T less sens2tive to level I*

2. The lower narrow range tap is higher.

Lu.o ee PEFERENCE BYN. Byn D1 44erences Sock. Pgl7,18 ano Figure 9-1.

PAGE 32 Iz__IUSISWDEUIg_edp_CgyIBObg ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 3.07 (3.00)

Auctioneered High Nuclear Power a.

Turbine load (Pimpulse)'

EO.25 each]

b.

Summing Unit (adds three temperature error signals together to) generate a total temperatur.e, error for the Rod Speed and Direction g

Programmer.lo51 fe.as) to.ac 6..va c.

Non-linear Gain Unit CO.53

1. (VwMi* Ga.w C'M E1.OJ d.

REFERENCE BYN,S.D. CH.

28,' Pgs. 26-29 ANSWER 3.08 (1.50) 1 6.

1.

Open - RCS pressure 1643 (+/-10) with SI signal each]

/

2.

Close - RCS pressure 1448 (+/-10) with SI sign b.

To prevent dead-heading the CCP's in a low RWS evel CO.5]

situation with high RCS pressure.

REFERENCE BYN, Byn Differences Book, P.

12 ANSWER 3.09 (1.50) a.

1.

Hi temperature CV pump recirculation flow.

2.

xcess #1 seal leakoff flow' EO.25 ea.

[0.53 b.

noting the correct CCW flow on the MCB meter.

By checking the correct CCW flow locally.

CO.5]

REFERENCE B'r N. Byron Differences Boch, PP.

14, 27 See Aw c La 9.

e

PAGE 00 0;__INg169dENIS_@$p_CQNISQL9 ANSWERS -- SPSIDWOOD 1

-86/07/16-JAGGAR, F.

3y < ten lh ANSWER 3.08 $09 (3.00) 1.

Flux Level High Rx Trip (Low Range) 2/" 25%

C3.152 2.

C-2 Rod Stop 1/4 103%

l0.152 (1 & 2 INTERCHANGEABLE) 3.

P-10 Permissive 2/4 10%

te.152

)

4.

Flux Level High Rx Trip (High Range) -2 / A i m c */

rm_1s, (3 & 4 INTERCHANGEABLE) 5.

Power Range High Flux Rate (Positive) 4/4 */ *M/0 err Ee.15]

6.

Power Range High Flux Rate (Negative)

CO.152 (5 & 6 INTERCHANGEABLE) 7.

P-8 Permissive (0 loop flow) 2/ t 30%

49.152 8.

Power Range Channel Current Comparator 2 detecterr/27. ET.152 O 1 e==b 9.

Over Power Recorder

0.15; 4.c ib-4 (8 L 9 INTERCHANCEABLE)

1. l + io 10.

Rod Control CO.153 11.

NIS Power Range Loss of Detector Voltage

( 2'.1[p o,qg,,,g 12.

Summing and Level Amp

( 2 qp0 4,, g g 15.

Delta Flux to OP and OT Delta T

[ 2 2$3 a g _,,,

t e 2l5]

14.

NR-45 15.

Currrent Recorder

( 2 2!51

[ 2 25]

16.

Computer 17.

Delta flux meter

( 2 253 18.

Detector Current Comp. 2 Detectors /2% of Avg.

[2.,1 3 (14-18 INTERCHANGLEABLE)

REFERENCE BWD, S.D.

30, Figure 33-1

f PAGE 3!

2___IU5IEUDEUI@_8Up_GQNIB9LS ANSWEPS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 3.10 (3.50)

The basic principle of operation is the detection of a Delta-T a.

The between adjacent heated and unheated thermocouples CO.53.

RVLIS sensor consists of a (Chromel-Alumel) TC near a heater and another (Chromel-Alumel) TC positioned away from the heater.

In the Delta-T a fluid with relatively good heat transfer properties, between adjacent TC's is small. CO.53 In a fluid with relatively poor heat transfer properties, the Delta-T between the TC 's i s l arge.

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 Pressure Wide Range, (Hot Legs A & C.(P.T.s 403 and 405))

( f. wu i;VLIC g.

cmmi..y; 3.

Me,. i.T.sa. U; ;J TC Te.T.p c. e tm-c

[4 vg*M

2. N.

Representative CET Temperature (f rom CET processing)

[0.5 each]

9.

cwwA.h w 5

CM. % <ahr %. WW REFERENCE BYN. S.D.

CH. 34B, Pgs. 10-11, 21 BW. S?DS, me vc w 15, 6 st.

ANSWER 3.11 (2.00) a.

G-M, Gamma b.

Scintillation, Beta c.

Scintillation. Beta CO.5 each]

d.

Scintillation, Gamma REFERENCE BYN. S.D.

CH. 49, Pg. 17, 62

J PAGE 34 Sz__BBQGEDUBES_:_UQBdeb1_GEUQBdeL2_EdE6@EUCY_@ND bed 19L001GOL_COUIBQL

-86/07/16-JAGGAR, F.

ANSWERS -- BYRON 1 ANSWER 4.01 (1.50) a.

1.

Seal injection flow is 8-13 gpm.

bM4 v%w\\f d) 2.

1. 1 Seal Leakoff flow is < 1 gpm.

3.

RCS Pressure $ M 1000 psig.

C1.03

'J.

Si Sen.) k.L.4% w.k%.w wau %.

b.

1.

RCS Pressure is < 100 psig and 2.

Seal injection water is not supplied.

CO.53 REFERENCE BYN, BOA RCP-1, Pg.

2, BOA RCP-2, Pg. 2 ANSWER 4.02 (1.00) d.

REFERENCE BYN, Technicel Specification Table 3.3-1 ANSWER 4.03 (2.00)

By reducing turbine load, diluting, or moving rods.

CO.53 a.

hour s) We.e_ cegwired a3r EO S e-aeQ b.

(Within 1 1.

Restore rod to operable status, IC.31 2.

Rod is declared inoperable O:.33 and other rods in group aligned within +/- 12 steps, E

-+3 3.

Rod is declared inoperable and Spec. SDM satisfied.

EM3 Tech.

a.

Power reduced to < /= 75%bo*/o)

C ::. 3:

a4.b.

5.

c \\che QMn 4"

• Y)"*

^ Y

• PI' i

REFERENCE

• M\\*b i

• "4' M d 'd BYN. BOA ROD-4 and TS 3/4.1.3.

c <.4n.)

r l

PAGE 05-di__EBQGEQUEEE_:_UQEd862_6EUQSd862_EDEEEEUGl_6ND BGD1Q60 GIG 06_GQUIBQL ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 4.04 (2.00) a.

1.

Keff of.95 or greater OR

2. Boron concentration of less than 2000 PPM.

C.4 each]

b.

1. Valve CV 8104.

2. Valves CV 110A and 110B.

C.4 each]

3.

RWST high head path.

REFERENCE BYN, BOA PRI-2, Pg i and 2.

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).

C1.OJ (s W kk ace.se W) b.

Obtain relief from a qualified operator.

CO.503 CO.253 Yes.

c.

1.

The SCRE/ Control Room Supervisor.

CO.50]

Injury to Company personnel or Public.

2.

a.

b.

Release offsite in excess of T.S.

c.

Damage to equipment that could affect public 3

CO.25 each]

heal th / saf ety.T.Q C.ts3 REFERENCE BYN. BAP 300-1, Fg. 8 l

ANSWER 4.06

(.50)

False REFERENCE BYN, BAP 300-1, Pg. 15 l

l i

PAGE 36 6t__EEOCEDUBE5_:_UQEMGLi_6EUREdGL4_EdESGENGY_GUQ SGDIOLOGIGGL_CQNIERL

-86/07/16-JAGGAR, F.

ANSWERS -- BYRON 1 ANSWER 4.07 (1.50)

ANY AREA ACCESSIBLE TO PERSONNEL in which there exists RADIATION a.

at such LEVELS that a major portion of the body could receive in ANY ONE HOUR a dose IN EXCESS OF 100 MREM.

[0.753 Areas near equipment or piping where the DOSE RATE AT GD 18 INCHES b.

from the source EXCEEDS THE applicable posted limits for the CCNCRAL AREA.

OR Areas near equipment or pipes where the DOSE RATE AT 18 INCHES the from the source would EXCEED 5 TIMES THE AMBIENT DOSE RATE for

[0.753 OCNCR^L AREA.

-REFERENCE BYN, BAP 1450-1, Pg. 1-2 B4 W, s*P P150 - 2 pt i 4.

2 ANSWER 4.08 (2.00)

E0.53 a.

50 mrem b.

Supervisory approval to 100 mrem.

E1.03 Radi ati on - Chemistry Supervisor.(%\\4L% s s.5 Sesa.3 E0. 53 3

c.

REFERENCE BYN, Radiation Protection Standards, Pg. 24 ANSWER 4.09 (1.00) to restore level.

Manually operate ECCS pumps as necessary REFERENCE BYN. 1BEP-F.1

PAGE 37 t__E6QGEDUBES_ _UQBd6Lx_@@UQEd@L4_EdE8EEUCX_@NQ 4

89D1960 GIG 96_CQNIBQL

' ANSWERS -- BiRON 1

-S6/07/16-JAGGAR, F.

ANSWER 4.10 (2.50) 1.

CC water to RCP lost (affected pumps only).

[0.53

[0.53 2.

Cntmt Phase B actuation.

Controlled RCS C/D NOT in progress [0.53 with 3.

a.

b.

RCS Pressure < 1370 psig c.

CCP's > 200 GPM (SIP positive flow exists)

[1.53 REFERENCE BYN, 1BEP-F.O ANSWER 4.11 (3.00) a.

Subcriticality (S)

Core Cooling (C)

Heat Sink (H)

RCS Integrity (P)

Containment (Z)

Inventory (I)

CO.35 each name, 0.05 each letter]

b.

False C.63 REFERENCE BYN, BAP 340-1, Pg.

8, 11 ANSWER 4.12 (1.50) a.

Valves c1cse.

[0.53 b.

Valves close.

CO.53 c.

Isolates (PORVs will have limited Nitrogen supply).

[O.53 REFERENCE BYN, BOA ELEC-1 ap. 2 and 6

, ~.

PAGE 35 Os__ESQGEQUEE5_:_UQSd6Li_6ENQBd6La_EDEEEEUGY_6UD E60106RGIGOL_GQNIEQL ANSWERS' - BYRON 1.

-86/07/16-JAGGAR, F.

ANSWER 4.13 (3.50)

L(O.53L..t.~

L...d C

a.

13%.

b.

1.

Trip the reactor.

2.

Trip the RCPs.

Ensure CC pumps are tripped'.

3.

4.

Go to BEP-0.

C2 @ O.5 ea.]

{h'q,w.43 g y o.y,%q c.

p.

1.

NCP thermal Barriers.

2.

RH heat exchangers.

4 3.

Spent fuel pit heat exchangers.

4.

Letdown ~ heat exchangers.

C 2. 0 2 ^ -

G.

Leess law w%.

REFERENCE PYN. BOA PRI-6 pp. 8 and 9 1

i l

,.._n

[.

~

8., >

U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:

BYRON 1 17 REACTOR TYPE:

PWR-WEC4 DATE ADMINISTERED: 86/07/16 EXAMINER:

JAGGAR__F__

APPLICANT:

J L

INSTRUCTIONS TO APPLICANT:

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 VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 25.00 6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00 25.00 7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 25.00 25.00 8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100.00 100.00 TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.

APPLICANT'S SIGNATURE

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the attainistration of this examination the following rules apply:

1.

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

2.,

Restroom trips are to be limited and only one candidate at a ties any 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 on,,1g 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.

1 8.

Consecutively number each answer sheet, write "End of Category

" as appropriate, start each categorg on a new page, write Joni gne side of the paper, and write "Last Page on th 7 east answer sheet.

9.

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

10. Skip at least three lines between each answer.

l

11. Separate answer sheets free pad and place finished answer sheets face l

down on your desk or table.

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

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

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

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

16. If parts of the examination are not clear as to intent, ask questions of 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 conolete your examination, you shall:

a.

Assemble your examination as follows:

(1) Exam questions on top.

(2) Exam aids - figures, tables, etc.

(3) Answer pages including figures which are a part of the answer.

i b.

Turn in your copy of the examination and all pages used to answer the examination questions.

j l

c.

Turn in all scrap paper and the balance of the paper that you did 4

I not use for answering the questions, d.

Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

~

-.--.,_m,

---

er

-'w,--=-"s--ww---'---

wre--------

-<,ww+--

r

~=~

5.

THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 2

THERMODYNAMICS QUESTION 5.01 (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?

Enthalpy decreases, entropy decreases, quality decreases.

a.

b.

Enthalpy increases, entropy increases, quality increases.

c.

Enthalpy constant, entropy decreases, quality decreases.

d.

Enthalpy decreases, entropy increases, quality decreases.

QUESTION 5.02 (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 the correct answer.)

The heat transfer coefficient of the cladding is reduced significantly.

a.

The specific heat capacity of the reactor coolant inventory changes b.

when voiding occurs and is not measurable by the RTDs.

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

RTDs.

d.

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

I i

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

5.

THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 3

THERMODYNAMICS

. QUESTION 5.03 (1.00)

Complete the sentence by choosing the correct answer from the choices below.

The 2200 degrees F maximum peak. cladding temperature limit is used because it is 500 degrees F below the fuel cladding melting point.

a.

b.

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

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

2200 F.

d.

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

QUESTION 5.04 (1.50)

A variable speed centrifugal pump is operating at 1/4 rated speed in a closed system with the following parameters:

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

(Show all work.)

f

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

5.

THEORY OF NUCLFAR POWER PLANT OPERATION. FLUIDS. AND PAGE 4

THERMODYNAMICS QUESTION 5.05 (1.50)

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

During cooldown and depressurization, you are required to remain 50 a.

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?

If the thermocouple in part b had read 360 degrees F, and the st'eam pressure inside the pipe was 560 psia, what would you estimate the c.

steam temperature to be at that pressure?

QUESTION 5.08 (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 If decay heat is 2%, what is the % core mass flow rate ?

degrees.

QUESTION 5.07 (2.50)

How do each of the following parameters change (INCREASE, DECREASE or NO CHANGE) if one main steam isolation valve closes with the plant at 50% load. Assume all controls are in automatic that no trip occurs.

Affected loops steam generator level (INITIAL change only)-

a.

i b.

Affected loops steam generator pressure Affected loop cold leg temperature c.

l d.

Unaffected loops steam generator pressure Unaffected loops cold leg temperature e.

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

l

PAGE 5

5.

THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND THERMODYNAMICS QUESTION 5.08 (2.00)

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

An adjacent ro'd is inserted to the same height a.

b.

Moderator temperature is INCREASED Boron concentration is DECREASED c.

d.

An adjacent burnable poison rod depletes QUESTION 5.09 (1.00)

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

At full power, equilibrium conditions, about half of the Xenon is a.

produced by Iodine decay and the other half is produced as direct fission product.

Following a reactor trip from equilibrium conditions, Xenon peaks b.

because delayed neutron precursors continue to decay to Xenon while neutron absorption (burnout) has ceased.

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

increases; of 50% equilibrium Xenon.

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

d.

high power levels, burnout is the major removal method.

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

G.

THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 6

THERMODYNAMICS QUESTION 5.10 (1.00)

The following statements concern fission product poisons. Complete the Place the answers on statements with the available answers provided below.

your answer sheet.

[An answer may be used more than once.]

hours to reach the maximum Xenon concentration a.

It takes about after a reactor trip.

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

b.

c.

It takes about hours to reach equilibrium Xenon concentration after a step increase from 0 to 50% power.

The decay half-life of Promethium 149 to Samarium 149 is approximately d.

hours.

Available Answers:

0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />; 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 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 />.

QUESTION 5.11 (3.00)

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

EXPLAIN what happens to reactor power and Tave.

Continue your a.

explanation until stable conditions are reached with no operator action.

(Assume the reactor is undermoderated, at BOL and no reactor trip occurs.)

"a" b.

How would final conditions differ if the transient in part happened at EOL as compared to BOL7 Explain any differences.

QUESTION 5.12 (2.50) which Of the coefficients that contribute to the power defect, a.

contributes most to the change of power defect over core life?

EXPLAIN.

[1.0]

Of the coefficients that contribute to power defect, which b.

coefficient reacts first to a sudden power change due to rod movement?

[0.5]

Explain why power defect is desireable for reactor operation at c.

l power.

[1.0) l l

l

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

l l

l

5.

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

THERMODYNAMICS QUESTION 5.13 (1.00)

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

QUESTION 5.14 (2.50)

Compare the CALCULATED Estimated Critical Position (ECP) for ato 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, 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.

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

criticality.

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.

The steam dump pressure setpoint is increased to a value just c.

below the Steam Generator PORV setpoint.

d.

Condenser vacuum is reduced by 4 inches of Mercury.

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

criticality is reached.

QUESTION 5.15 (2.00)

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

(Assume on trip.)

no

(***** 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)

GPM (maximum allowable for each kind) 6.

7.

GPM 8.

GPM Pa-Re9 ** %'

9.

GPM 10.

GPM 11.

F 12.

GPM y., eae ce 7M QUESTION 6.02 (1.40)

State the pressure source used to pressurize the Unit 1 and Unit 2 pressurizer PORV accumulators.

Why is the source for Unit 2 different than that of Unit I?

QUESTION 6.03 (1.50)

Unit 2 has two additional installed solenoid operated centrifugal Charging Pump mini-flow recirc valves, 2CV8114 and 2CV8116.

What signal and setpoint will automatically a.

1.

Open 2.

Close these valves?

[1.0]

b.

Why were the additional valves insta11ec7

[0.5) i

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

6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE O

QUESTION 6.04 (1.50)

A seal water heat exchanger outlet high temperature condition exists.

Other than low CCW flow, list TWO other causes of this condition.

a.

How can the Unit 2 operator verify that low CCW flow is not a possible b.

cause?

How can the Unit 1 operator verify that low CCW flow is not \\ possible c.

cause?

QUESTION 6.05 (2.50)

State the S/G Narrow Range level setpoints (in percent) for the a.

following:

[1.5]

UNIT 1 UNIT 2 High High Level Trip -

Normal Operating Level at 100% power -

Lo-Lo Level Trip -

b.

Why is the Narrow Range level span on Unit 2 more compressed than Unit 17 Describe this physical change.

(1.0]

QUESTION 6.06 (3.60)

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 your answer:

The NAME of the system, AND a.

b.

1.

The DESIGN flowrate (gpm) and ascociated pressure, and The MAXIMUM flowrate (gpm) and associated pressure.

OR 2.

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

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

6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 10 QUESTION 6.07 (3.00)

When is a 2/4 trip logic required to be used in the Solid State a.

Protection System (SSPS)?

[1.0]

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

b.

When is it energized?

[1.5)

TRUE or FALSE?

c.

Both Reactor Trip Bypass Breakers can be racked in at the same time, i

but only one may be closed.

[0,5)

QUESTION 8.08 (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 i

SI signal reset?

Explain your answer.

l QUESTION 6.09 (2.00)

The following concern the Remote Shutdown Panels.

TRUE or FALSE i

The MCB pull-to-lock feature is overridden when operation is from the

.i a.

Remote Shutdown Panels.

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

Panels.

If local control of the MSIV is taken at the Remote Shutdown Panels, l

c.

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.

l 4

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

-L I

f 6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 11 QUESTION 6.10 (3.00)

State the inputs that are used to generate the Power Mismatch a.

signal in the Reactor Control Unit.

[0.5]

b.

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

[1.0)

The Summing Unit can only function using temperature signals.

.c.

In what system component is the Power Hismatch signal converted to a temperature signal?

[0.5]

d.

Which of the below 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.

QUESTION 6.11 (2.00)

For each type of 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.

b.

. Gaseous.

c.

Particulate (Gas streams).

d.

Iodine (Gas streams).

(***** END OF CATEGORY 06 *****)

j

=

PAGE 12 7.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL QUESTION 7.01 (2.00) v wpk-.J WhatareFOURspecific[ method / symptoms,thatcanbeused,for a.

identifying the fault.. steam generator, during a steam generator tube rupture accident, in accordance with BEP-37 What are the TWO conditions that must be monitored during b.

a steam generator tube rupture accident after a RCS cooldown is initiated, that require RCP's to be tripped?

QUESTION 7.02 (2.00)

The following concern information found in BOA PRI-2, Emergency Boration.

State the TWO conditions, which if either are encountered while in a.

mode six, would required Emergency Boration.

[0.8]

If Emergency Boration flow of 30 GPM is required, state the THREE b.

flowpaths that are available.

[1.2]

QUESTION 7.03 (2.00)

Refer to attached Figures 22-1 and 22-2, (RCS Subcooling Margin).

Explain the reason why the " Adverse CNMT" curve is less restrictive on a.

Unit 2 when compared to Unit l's curve.

b.

What TWO factors were accounted for, in establishing the Unit 2

" Adverse" curve on the RCS subcooling margin curve, Figure 22-2.

QUESTION 7.04 (2.50)

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

Why is manual SI actuation not advisable during performance of a.

BFR-S.1?

[1.0]

State the TWO entry symptoms or conditions for entering BFR-S.I.

[1.5]

b.

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

7.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 13 RADIOLOGICAL CONTROL QUESTION 7.05 (1.50)

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

State the Shift Engineer's responsibility for Type 2 RWPs PRIOR a.

to any work signing in on the RWP.

[0.2]

b.

State the FOUR reasons that may be used to terminate an RWP.

[0.8]

For how long is a Type 2 RWP valid?

[0.25]

c.

d.

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

[0.25]

QUESTION 7.06

(.50)

The lower Steam Generator Narrow Range tap is at different levels for Unit 1 and Unit 2.

What is the reason for requiring at least 4% Narrow Range S/G 1evel when verifying a heat sink is available in both Unit 1 and Unit 2 procedures.

QUESTION 7.07 (2.00)

The following pertain to use of the Emergency Procedure (BEP, BST, BFR, BFS) Network.

Assume an emergency situation exists.

When is the initial scan of the Critical Safety Function Status a.

Trees performed?

[0.5]

b.

A BFR is being performed after an ORANGE condition was identified.

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

[0.5]

Which of the following procedures may be entered directly?

c.

(Without being entered from another procedure.)

Note:

More than one procedure may be correct.

[1.0]

1.

BEP-0 2.

BEP-3 3.

BCA-0.0 4.

BCA-1.1 I

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

n PAGE 14 7.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL QUESTION 7.08 (2.00)

During the performance of BOA ROD-4, " Dropped Rod Recovery",

a.

prior to recovery of the dropped rod, state ALL methods that can be used to match Tref with Tave.

[0.5]>

If a dropped rod cannot be recovered immediately, state the b.

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 operation to continue.

[1.5]

QUESTION 7.09 (1.50)

List the THREE actions, in the correct sequence, that are required, a.

when using procedure IBOA RCP-1, Reactor Coolant Pump Seal Failure, if RCP bearing temperature reaches 225 F.

According to IBOA RCP-1 other than if RCP bearing temperature b.

approaches the alarm level, when must the #1 seal bypass valve be opened?

c.

What THREE conditions, all of which must satisfied, before the #1 seal bypass valve can be opened?

QUESTION 7.10 (2.00)

State FOUR of the 8 symptoms that would indicate a need to enter IBOA a.

PRI-1, Excessive Primary Plant Leakage.

(Setpoints not required.)

State the TWO specific conditions that would require the reactor b.

to be tripped and a transition from IBOA PRI-1 to IBEP-0, Reactor Trip or Safety Injection.

i I

(*****

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

PAGE 15 7.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL QUESTION 7.11 (3.00)

Assume the following:

~

Unit 2 is in its initial ascension to full power, power is increased Bank D Control Rods move from from 0-40% 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 />.

75 steps to 110 steps 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 move from 110 steps to full out 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, What fuel conditioning power increase limit was violated a.

and at what point'in the above scenario was it violated?

um;4s b.

State any rod withdrawal rates,that were violated and where they were violated.

c.

After the trip, to what new 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?

QUESTION 7.12 (1.00)

WHAT After the initial power ascension requirements have been met, is the basis for the new preconditioned power level.

QUESTION 7.13 (3.00)

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

" Plant Heatup".

WHAT is the maximum pressure and temperature that should be maintained a.

in the RCS when the RH System is in service?

[1.0]

Would starting an RH pump while using RH letdown with the RCS solid b.

cause an inadvertent RCS pressure INCREASE or DECREASE if CV131 l

is in AUTO?

[0.5) i All Reactor Coolant Pumps and RH pumps may be deenergized i

c.

i during Mode 5 operation providing two conditions are met and i

maintained.

State these TWO conditions.

[1.5) i i

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

8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 16 QUESTION 8.01 (2.50)

The following pertain to information found in BAP 1250-2 Deviation Reporting, and BAP 1250-6, Reportable /Potentially Significant Event Screening and Notification.

What is the basis for determining if a plant event or condition is a.

a Potentially Signigicant Event?

[1.0]

b.

If an event is determined to be NOT reportable, who, by job title, is required to screen the event for significance?

[0.5]

If an event is determined to be a reportable event, but not c.

a GSEP event, state the TWO notifications that must be made by the Shift Engineer.

[1.0]

QUESTION 8.02 (2.00) p., w.. s.) G.- Ja-State the minimum number of gallons required,"to be in the Diesel a.

Oil storage tanks and the associated indicated level (in %)

for each Unit.

State the minimum number of gallons required to be in each unit's b.

Diesel Generator Day Tank and the associated indicated level (in %).

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]

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]

QUESTION 8.04

(.50)

On the Technical Specifications Table 3.3-1.1.b, Unit 2 Fire Detection Instruments, the words "(Unit 1)" appears several times.

What is the significance of this?

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

i 8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 17 QUESTION 8.05 (2.00)

What is meant if an instrument number in Technical Specifications a.

(0.5) is preceeded by a zero?

b.

Refer to attached Figures ~1-6a and 1-6b.

For Unit i to operate, what instruments must be operable.

You may answer by section number if they all apply (i.e. all of #1).

(1.5)

QUESTION 8.06 (2.50) b 74Eebwhe.\\ S$ set'b' h

" S *-

Ateced%

a'.

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 Auxiliary Operator STA or SCRE i

b.

What is 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)

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

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

If he/she is to be absent for Modes 5 or 6, what must be done to ensure continuity of control?

[1.0) l l

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

8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 18 QUESTION 8.07 (3.00)

According to Technical Specifications:

what are the exemptions from the RWP issuance requirements during the a.

performance of thei'r duties in a High Radiation Area?

b.

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

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

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?

QUESTION 8.08 (2.50)

Surveillance requirements must be performed within specified a.

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?

[0.5]

What is the interval for each of the designators below?

[1.0]

c.

1.

S 2.

2 25.

SA L

i QUESTION 8.09 (2.00)

According to Technical Specifications, how will an operator know if an a.

LCO applies only to one unit?

How will different operating parameters, setpoints or equipment b.

for each unit be identified in Technical Specifications?

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

8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 19 QUESTION 8.10 (2.00)

The following concern BZP 300-1. Initial Notification and GSEP Response.

What is the time limit for notification of off-site authorities?

a.

[0.5]

b.

When does the clock start for notification of off-site authorities?

[0.5]

If the off-site authority wants verification of authenticity c.

of the notification, what action is to be taken?

What information is not given?

[1.0]

QUESTION 8.11 (2.00)

According to BAP 380-2, Handling of Long-Term Annunciator Alarms, a.

what constitutes a "long term alarm"?

b.

List THREE actions that must be taken for an alarm that is valid, and the condition causing the alarm is a desired means of operation.

QUESTION 8.12 (2.00) a.

During off-hour shifts, weekends, and holidays when the Station Security Director is not on site, who assumes the responsibility for Station security?

[1.0) b.

How many visitors may a single authorized individual escort in protected area?

How many in the vital areas?

[0.5]

c.

TRUE OR FALSE 7

[0.5)

Badged personnel with a status level LOWER than is required i

to access into a particular area CANNOT be escorted into that area by a person who has the proper status level.

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

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

EQUATICN 5HEIT Cycle afficiency = (!Iet W A v = s/t f = ma cut)/(Energy in) 2 s = V,5

• 1/2 at

, = wg 7

{ =,gc-A = A e' A = tN g

.<E = 1/2 my a = (Vf - V )/t 3

PE = mgn A = &n2/tjjg = 0.693/t1/2

• = e/t v = V,

• at 1/2*" " b*U' 3

f 2

I * (I I3 20

((t1/2 y,,j A=

b 4

-I:x ti = 931 ms a = V,yAo Q = mCost I = I,e " #

6 = UA A T I = I,10"* 6#'

Pwr = W 2 y

TVL = 1.3/u sur(t) gyg,,g,gg37,

~

P = P 10 O

7 = P e / *'

t SG = S/(1 - K,ff) o SUR = 25.06/T G, = S/(1 - K,m)

I SUR = 25e/ t + (s - o)T G;(1 - K,ffj) = G II ~ "eff2 2

T = ( L/s ) + ((s - o '/ Io]

M = 1/(1 - K,ff) = G /G 3

3 M = (1 - K,ff,)/(1 - K,ff;)

7 = s/(o - a)

SCM = { - K,ff)/K,ff T = (s - o)/(Is) t' = 10 secancs a=(K,ff-1)/K,f,*.K,fgK,ff I = 0.1 seconds' o = C(t-/(r K,ff)] + (7,ff (1 + It)]

/

l1*Ik Id 2 =2Id P = (:4V)/(3 x 1010)

Idj 22 2

R/hr = (0.5 CI)/c (,,g,73y R/hr = 6 CE/d2 (feet)

=N Miset11aneous Ocnve sions Watar Aar meters I curie = 3.7 x 1010cos 1 gal. = 8.345 lem.

I kg = 2.21 lem 1gja.=3.7811 tars 1noa2.34x103Stu/hr 1 f.

= 7.48 gal.

1 m = 3.41 x 100 5tu/hr 3

Oensity = 62.4 lbrp/ft lin = 2.54 ::n 3ensity = 1 g:n/c::r8

'F = 9/5'O - 32 Heat of vacerization = 970 3tu/ tem

'C = 5/9 ('F-32)

Heat of fusion = 144 Stu/lem 1 STU = 778 ft-lbf 1 4tm = 14.7 asi = 29.9 in. Hg.

1 ft. H O = 0.4335 luf/in.

2

H

.=.

e-G,,.

w

1. h [

M N

i g

>,), g i

(

=

g

. s -S i

N 0--

1 C-E L

i E

w 5

W W

_,E -

W e

g 6

.g y

--l

-O 0-X E

__3 (H: 0-K X

gi

'O T.

4 N

P--f9-D 7

(

O

( =,i I

u v

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

-r--

~_.m-.,-,--.__.-,,,._.._w..,,,,y...

_.,-.m__.,_,

y

. i,..

G u+iew ~.es 7

l REV. lA REACTOR TRIP OR SAFETY INJECTION 1BEP-0 WOG-1 UNIT 1 260 2500 2400 2500 i

2200 2l00 I

2 1900 1800 1700 1400 CEPT m t f 1500 i

lg i400 i

la00 1200 i

1100 1000 i

900 i

a gg f

N 800 l

700 600 500 l

400

\\

'00 l

200 l

10 0 600 7t 0

[

i 100 -

200 500.

400 500 TEWERATURE (*F) i llO FIGURE 1BEP 0-1 i

RCS SUBC00 LING MARGIN l

't Figure 22-1 I

Page 25 of 26 l

Ad, 7.o s 4

2600 2500 2400 2300 2200

/l 2l00 2000

~

.I ~

1900 i

! ~

l800 i

1700 l

l 1600 i

.i l n

1500

[ /[

b f400 l j^ ;^

g 1300

/

/ ;'

(

Q l200 l

ll 1100 ACCEPTABLE i

/.

t l000 l

l l' e

900 ADVERSE

/

-'/

E 800 CNMT

.' / j' 600 CNMT Nf-

',/,

NOT ACCEPTABLE E 700 NORMAL

_e'

,- /

l

_m _- -.-

500

^

^

400 300 Sg'TURATION 200 10 0 10 0 200 300 400 500 600 700 TEMPERATURE (*F)

RCS SUSC00 LING MARGIN Figure 2 2-2

i....'

TABLE 3.3-13

] ]

RAOl0 ACTIVE GASEGUS EFFLUENT HONIIORIIIG INSTRUNENTATI0fl

!! :P MINIMUM EHANNELS

'l3 INSTRUNENT OPERABLE APPLICA81LITY ACil000 cg xo 30 1.

Plant Vent Montt ring Systee - Unit 1 0

a.

Noble Gas Activity Monitor-6

[

Providing Alare 33 1) liigh Range (IRE-PR0280) 1 39 l

2)

Low Range (IRE-PR0288) 1 40 b.

Iodine Sampler (IRE-PR028C) 1 40 c.

Particulate Sampler (IRE-PR028A) 1 w

d.

Effluent System Flow Rate a

36 1

Measuring Device (LOOP-VA019)

I w

36 Sampler Flow Rate Measuring Device 1

h e.

(IFT-PR165) 2.

Plant Vent Monitoring System - Unit 2 a.

Noble Gas Activity Moniter-I Providing Alarm 39 '

1) liigh Range (2RE-PR0280) 1

,8 39 2) low Range (2RE-PR0288) 1 i

40

{

b.

Iodine Sampler (2RE-PR028C) 1

(.

)

40

4. r-c.

Particulate Sampler (2RE-PR028A) 1 r~ ~.,

.c M'

d.

Effluent System Flow Rate 36 b

.Tl HeasurIng Device (LOOP-VA020) 1 ei i

C a

36 e.

Sampler flow Rate Measuring Device 1

g (2fT-PR165) 1 1

d TA8LE 3.3-13 (Continued) e en RADIDACTIVE GASEGUS EFFLUENT NONITORING INSTRUMENTATION 4

s e(

MINIMuH CilANNELS gg INSTRIMENI OPERA 8LE APPLICA81LITY ACTION 3.

Gaseous Waste Management System i

34 Hydrogen Analyzer (OAT-GW8000) 1 e-a.

i!

FJ b.

Oxygen Analyzer (OAIT-GWOO4 and 38 DAT-GW8003) 2 4.

Gas Decay Tank System i

a.

Noble Gas Activity Monitor - Providing Alara and Automatic Termination of a

35 Release (ORE-PR002A and 28) 2

,s

[,

5.

Containment Purge System i

U a.

Noble Gas Activilty Honitor - Providing 8

37 Alarm (RE-PR0018) 1 b.

Iodine Sampler a

40 (RE-PR0010) 1 c.

Particulate' Sampler (RE-PR001A) 1 40 a

=-.

g.

Radioactivity Honitors Providing Alara and Automatic Closure of Surge Tank Vent-Component 41 c 3 j, Cooling Water Line (ORE-PR009 and RE-PR009) 2

,=. w 7.

g 4.*1,'-

..q.h,,

b

.. - 7, p

1g

8.o s TA8LE 3.3-13 (Continued)

TABLE NOTATIONS

• At all times.

i

""During WASTE GAS HOLDUP SYSTEM operat orr.

1. All instruments required for Unit 1 or Unit 2 operation.

ACTTON STATEW1. TT N

ACTION 35 - With the number of channels OPERA 8LE less than required by the Minimum channels OPERA 8LE requirement, the contents of the tank (s) may be released to the environment for up to 14 days provided that prior to. initiating the release:

a.

At least two independent samples of the tan,k's contents are analyzed, and b.

At least two Jechnically qualified members of the facility staff independently verify the release rata calculations and discharge valve lineup.

Otherwise, s'uspend release of radioactive effluents via this pathway.

ACTION 36 - With the number of channels OPERA 8LE less than required by the Minimus Channels OPERA 8LE requirement, effluent releases via this pathway say continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

ACTION 37 - With the number of channels OPERABLE less than required by the Minious Channels OPERe8LE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

ACTION 38 - With the number of channels OPERA 8LE one less than required by the Minimum Channels OPERA 8LE requirement, operation of this system may continue provided grab samples are taken and analyzed at least once

~

I per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With both channels inoperable, operation say continue provided grab samples are taken and analyzed at least once per

4. hours during degassing operations and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during other operations.

ACTION 39 - With the numeer of channels OPERABLE less than required by the Minimus Channels OPERA 8LE requirement, effluent releases via this pathway say continue for up to 30 days provided gran samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• O QM' w 'i, q y

BRALDWoco eveeM-- UNITS 1 & 2 3/4 3-73

+

wW v--

e.

---

e se---

-,---w--

r ---

r------

- - - - +

.s p s

.y i

TA8LE 3.3-13 (Continued)

ACTION STATE @f75 (Continued) i ACTION 40 - With the number of channels OPERA 82 less than required by the Minimus Channels CPERA8LE requirement, effluent releases via the affected pathway may' continue for up to 30 days provided samples i

are continuously cs11ected with auxiliary sampling equipment as required in Table 4.31-L j

l ACTION 41 - With the number of channels GPERA8LE f ess than required by the i

Minimum Channels OPERA 8LE requirement, effluent releases via this pathway may continue for up to 30 days provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for radioactivity at a lower limit of detection of no more than 10 7 microcurfe/rl.

i 4

s I

i i

J. Kien

-,% v, 2n us.6<

N I

~

SRMDwo0D 9440M - UNITS 1 & 2 3/4 3-74

o

.'t s'

5.

THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 20 THERMODYNAMICS ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

///,CTn en, i

1%r [ gf e-y jl ANSWER 5.01 (1.00) d REFERENCE MNS OP-SS-HT-2, p.12.

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

BYN, HT&FF Review, Pg 137-143.

ANSWER 5.02 (1.00) b REFERENCE MNS Thermo, para. 2.6.

2 & 13.

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

BYN, HT&FF Review, Part C.

ANSWER 5.03 (1.00) c REFERENCE MNS Thermo-Core Performance, p.2.

13.

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

BYN, HT&FF Review, Part C.

ANSWER 5.04 (1.50) 3

=

19.2 Mw Power (2) = Power (1)(N2/N1) cubed = 300x(4) 2 2

800 psid Delta P(2) = Delta P(1) (N2/N1)

= 50x(4)

=

Flow (2) = Flow (1) (N2/N1)

800x(4)

3520 gpm

[0.5 each]

REFERENCE GPNT Vol. III, Ch.

2, Sect. H, p.

2-234.

ROWE Reactor Operator Training Manual, Sec.

2, pp 49-~50 10.

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

BYN, HT&FF Review, Pgs. 322-324.

5.

THEORY OF NUCLEAR POWER PLANT OPERATION FLUIDS. AND PAGE 21 THERMODYNAMICS ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 5.05 (1.50) a.

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

b.

t&& degrees F of superheat per superheat tables.

i ss-n o c.

600 degrees F.

[0.50 each]

4 to - sio REFERENCE Steam Tables and Mollier chart ANSWER 5.06 (1.50)

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

.02 = m (.67)

.02/.67 =.03 or 3%

[1.5)

REFERENCE General Physics, HT'& FF, Section 3.2 ROWE Reactor Operator Trainin's Manual,' Sec. 2, pp 54-63 BWD, Westinghouse Thermal-Hydraulic Principles for PWR, Ch. 14 BYN, HT&FF Review, Pgs. 351-355.

ANSWER 5.07 (2.50) a.

Decrease b.

Increase c.

Increase.

d. Decrease e.

Decrease

[0.50 each]

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

BYN, HT&FF Review, Part B, Sections 2 & 3.

.= ___

5.

THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 22 IHERMODYNAMICS ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 5.08 (2.00) a.

Decrease b.

Increase c.

Increase d.

Increase

[0.50 each]

REFERENCE SQN/WBN License Requal Training, " Core Poisons" BYN, Rx Theory Review Text, Pgs. 5.36-5.52.

ANSWER 5.09 (1.00)

D.

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

HBR, Reactor Theory, Sessions 38 and 39.

Section VI.

DPC, Fundamentals of Nuclear Reactor Engineering, ANSWER 5.10 (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 (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 />.

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

5.

THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 23 THERMODYNAMICS ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 5.11 (3.00)

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

inserts positive reactivity [0.25], and power increases. [0.25]

(At the POAH), increased power will increase temperature which inserts negative reactivity via FTC. [0.25] Power will stabilize higher than the POAH [0.25] and Tave will lower than the no-load value (minus the number of degrees needed to overcome FTC).

[0.25]

b.

Power increase RATE is higher at EOL because of changes in Beta-Bar (MTC). [o S]

Final power is the same;Tave will be higher [0.5] (closer to no-load temperature) because of the different Beta-Bar (MTC) [0.5].

REFERENCE Millstone Reactor Theory, RT-18.

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

BYN, Rx Theory Review Text, Pgs 5.2-5.26 ANSWER 5.12' (2.50)

Moderator Temperature Coefficient (MTC).[0.5]

Because boron a.

concentration is reduced [0.5].

b.

Doppler (FTC) [0.5].

Power defect has a stabilizing influence on reactor operation c.

because it resists power changes.

(As power increases, power defect adds negative reactivity and as power decreases, power defect adds positive reactivity).

[1.0]

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

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

BYN, Rx Theory Review Yext, Pg 5.2-5.26.

ANSWER 5.13 (1.00)

The change in water density per degree F increases as

[1.00]

as temperature increases.

REFERENCE BWD, Westinghouse Large PWR Core Control, Chapter 2 p.

2-3 to 41, Chapter 3 p.

3-20 to 23.

5.

THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 24 j

THERMODYNAMICS

)

ANSWERS -- BYRON 1

-86/07/16-JAGGAE, F.

i BYN, Rx Theory Review Text, Pgs 5.2-5.26.

ANSWER 5.14 (2.50) a.

SAME AS b.

(ECP) LOWER THAN (ACP) c.

(ECP) LOWER THAN (ACP) d.

SAME AS (ECP) HIGHER THAN (ACP)

[0.5 each]

e.

REFERENCE 7-24 to 28.

BWD, Westinghouse Large PWR Core Control, Chapter 7 p.

BYN, Rx Theory Review Text, Ch.

5.

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, therefore, that rod "se's" e

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]

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

IP2 Reactor Theory pg 7-59,60.

BYN, Rx-Theory Review Text, Pg 5.36.

6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 25 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 6.01 (3.00) 1.

75 5.

138/#33 9.

12 2.

2235 6.

120 Mixed, 75 Cation 10.

55 3.

557-sst 7.

87 11.

500 -536 4.

37025 8.

32 or 8 per RCP(b-id 12.

3 per RCP or 12

[0.25 each]

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

ANSWER 6.02 (1.40)

Unit 1 - Nitrogen Unit 2 - Instrument Air

[0.5 each]

Instrument Air is less expensive #(constant losses).

[0.4]

REFERENCE BYN, Byn Differences Book, P. 11 ANSWER 6.03 (1.50) a.

1.

Open - RCS pressure 1643 (+/-10) with SI signal

[0.5 each]

2.

Close - RCS pressure 1448 (+/-10) with SI signal b.i.To prevent dead-heading the CCP's in a low RWST level situation with high RCS pressure.Eo.2s1

-[0,5]-

emmawawT.oasi

2. r %.,e Au si-L5 6 9.....

REFERENCE BYN, Byn Differences Book, P. 12 ANSWER (mga t weess 1.uo-w( 1. 50 )

6.04 3.

N

< % cewr.m,mwe a.

1.

High temperature CV pump recirculation flow.

{n.,,,, gal 2.

Excess #1 seal leakoff flow / w e== b s

[0.25 ea.]

b.

By noting the correct CCW flow on the MCB meter.

[0.5]

By checking the correct CCW flow locally.

[0.5]

c.

REFERENCE BYN, Byron Differences Book, PP. 14, 27

~,

\\

6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 26 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

j ANSWER 6.05 (2.50) a.

UNIT 1 UNIT 2 81.4%

78%

66%

50%

40.8%

17%

(all values +/-1%)

[0.25 each]

b. i. Because of the higher recirculation flow in Unit 2 khe S/G is less sensitive to level transients)g'$*gj'O75 each]** M aa9"554 b "

• f Y

a.The lower narrow range tap is higher.

REFERENCE BYN, Byn Differences Book, Pg 17,18 and Figure 9-1.

ANSWER 6.06 (3.60)

,,3,,

a.

1.

Centrifugal Charging Pumps:

b.

300 gpm (150 each) # 2500 psig 1100 gpm (550 each) @ 600 psig

[0.51 each]

2.

Safety Injection Pumps:

b.

800 gpm (400 each) @ 1200 psig 1300 rpm (650 each) @ 800 psig

[0.51 each]

6000gpm(dOOOeach)@ 165 psig 3.

Residual Heat Removal Pumps:

b.

10000 gpm (5000 each) @ 125 psig

[0.51 each]

G19 5-nr1 Accumulators:

b 28,000 Gals.(approximately P&OO each) 4.

(3s-6 3% teutQ -*#. 6 approximately 426 psig.

[0.54]

Lo1 643 REFERENCE BYN, S.D. CH. 58, Pgs. 22-27 W, T.A. spoo

(

/

6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 27 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 6.07 (3.00)

When control and protection are provided by the same a.

parameter.D=3 [G. 5]

' JILL e channel f ailurc 2/3 pretcetier. is still ava14able.

[0.5]

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 3

switch.

[0.75]

au.J.muc 4dp(sign =$0==4 c.

True

[0.5]

REFERENCE BYN,S.D. CH. 60A, Pgs. 11, 16 ANSWER C.08 (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] uo a.ed.J tW wa.h concey d=enbes Me. o pan =A %

o4 A% N perasno.

REFERENCE BYN, S.D. CH. 61, Figure 61-17 ANSWER 6.09 (2.00) a.

TRUE b.

TRUE c.

TRUE d.

TRUE

[.5 ea]

REFERENCE BYN, S.D. CH. 62, Pgs. 14-17

6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 28 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 6.10 (3.00)

Auctioneered High Nuclear Power a.

Turbine load (Pimpulse)

[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.Ios3 rur3 t.c{1.01 c.

Non-linear Gain Unit

[0.5]

J W.

1. (Variable Gain Unit)

[1.0]

REFERENCE BYN,S.D. CH. 28, Pgs. 26-29 ANSWER 6.11 (2.00) a.

G-M, Gamma b.

Scintillation, Beta c.

Scintillation, Beta d.

Scintillation, Gamma

[0.5 each]

REFERENCE BYN, S.D. CH. 49, Pg. 17, 62

7.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY A@

PAGE 29 RADIOLOGICAL CONTROL ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 7.01 (2.00) r g v,g,s; % 5:. 3 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.

9

-St.w>

w c*o w_m.w. w.sewme,

b.

1.

CC water to RCP lost. (affected pumps only)

@ayT** * * *d 2.

Phase B entmt. isolation

[0.5 en.]

y.pepa q At troops 94RSm u.we mw.p(=hw P+*4) s.eep*i w vo 4 c.q,3 2.

REFERENCE BYN BEP-3 p.

3 & 4; Fold Out Page ANSWER 7.02 (2.00) a.

^1.

Keff of.95 or greater OR

2. Boron concentration of less than 2000 PPM.

[.4 each)

1. Valve CV 8104.(E%=<t%4to- *We i

-b.

2. Valves CV 110A and 110B.(W e 4 h M

3. RWST high head path.(int e+E)

[.4 each]

REFERENCE BYN, BOA PRI-2, Pg 1 and 2.

ANSWER 7.03 (2.00)

Unit 2's wide range pressure transmitters are located outside a.

containment [0.5] thereby eliminating the need to account for adverse containment conditions in the pressure input to the saturation lines.

[0.5]

b.

1.

Adverse conditions of the Core Exit Thermocouples 2.

The " normal" inaccuracy (instrument error) of the wide range pressure instrument (+/-90 psig).

[1.0]

REFERENCE BYN, Syn Differences Book, PP. 26-27

\\

7.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 30 RADIOLOGICAL CONTROL ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 7.04 (2.50)

A possible loss of heat sink due to main feedwater isolation a.

while the reactor is at power.

[1.0]

b.

Entered from:

1...BEP-0 (Reactor Trip or Saf ety Injection.) [0.5]

When Rx trip is not verified and manual trip not effectivero.xs3

{s..]

3 W33-res3 (od) na

a E 9-o 54-*p \\ ao it wo cow 2...BST-1 Suberiticality.

CSF on Red (Orange).

[0.5]

c,e L. %.c >< s e.

. J /..

nte. Wiwa su n REFERENCE BYN BFR-S.1 pp. 1 & 2.

ANSWER 7.05 (1.50) a.

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

and shift.

b.

Job Cancellation Job Completion Expiration

[0.8]

Changed conditions

[0.25]

For the length of the job.

c.

[0.25]

d.

50 mrem / day.

l REFERENCE BYN, RP Standards, Pp. 12 - 17 l

BwD, RP Standards, Pp. 14 - 17b I

ANSWER 7.06

(.50)

Co.w3 l

(Considering instrument inaccuracies) S/G 1evel is ensured to be in the i

narrow range.0oas]

REFERENCE BYN, Byn Differences Book, P. 28

7.

PROCEDURES.- NORMAL. ABNORMAL. EMERGENCY AND PAGE 31 RADIOLOGICAL CONTROL ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 7.07 (2.00) a.

After departing BEP-0 unless directed by BEP-0. [0.5]

b.

Suspend the. lower priority BFR and address the higher priority BFR.

[0.5]

c.

1 and 3

[1.0]

REFERENCE BAP 340-1, Pg.

8, 9 WOG, Users Guide, Pg. 14; BEP-0, Pg. 1; BCA-0.0, Pg. 1 ANSWER 7.08 (2.00)

By reducing turbine load, diluting, or moving rods.

[0.5]

a.

b.

(Within i hour)

[ h, o p.4 b [o.5 e 4 1.

Restore rod to operable status, fe-St 2.

Rod is declared inoperable-[0.0} and other rods in group aligned within +/- 12 steps, EO-33 3.

Rod is declared inoperable and a.

Tech. Spec. SDM satisfied.

fG-3-3

'i n.

Power reduced to </= 75%,Do%)

[0.3]

hVM. 3 d*95 E

C'\\ *M' OM

( rii. wTwT or. 4yosa w mM.S be Mi N

REFERENCE g

'T g BYN, BOA ROD-4, and TS 3/4.1.3.

y.,5 ANSWER 7.09 (1.50) a.

Trip the reactor [0.2]

Trip the affected pump [0.2]

Go to IBEP-0, Reactor Trip or Safety Injection [0.1]

b.

When the #1 seal temperature approaches the alarm.

[0.5]

c.

1.

Seal injection flow 8-13 gpm [0.1] ]

2.

1. 1 seal leakoff < 1 gpm EG-2-}

M 4 0 M ech3 3.

RCS pressure < 1000 psig {0 2-]-

[3 %

V.

• 1 5.J L. L.t r yu u o j

REFERENCE BYN, 1 BOA RCP-1, Pg 2 and 6.

X PAGE 32 7.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 7.10 (2.00) a.

1.

Containment Radiation Monitors high 2.

Increased charging flow 3.

Increased VCT M/U frequency 4.

Abnormal Containment pressure / temperature S.

Abnormal PRT conditions 6.

Off-gas radiation monitors abnormal 7.

Increase sump / cavity pump run times 8.

Rx vessel flange leak off high temperature AJ e.a.

.ww > a t.+

s. 44=h4

[0.25 each for any 4]

T.

When pressurizer level cannot be maintained =ith all CCP's > tf %D.ol b.

1.

rn==ing [0.5].

2.

CI 0001A ond D vpun [C.5].

REFERENCE BYN, 1 BOA PRI-1, Pg 1,3 ANSWER 7.11 (3.00) 3% of full power per hour (after 20% power) [0.5].

Violated a.

from 20 - (40%), and 40-80% [0.5].

3 steps per hour (after 50% power when the 3%/ hour rate b.

[0.5].

Violated from 50 - (80%) power [0.5].

is applied) New level is determined by highest power level achieved for c.ao@6% [0,5].

any 72 consecutive hours during any 7 day operating period [0.5].

(A..< u,

..x..ua b 6,.s a.3 3 wo+ 7)

REFERENCE BYN, BGP 100-3, P. 2 BGP 100-3, Power BYN, PWR Initial Licensing Training Lesson Plan, Ascension, PP. 8-9 ANSWER 7.12 (1.00)

It is the highest power achieved for a cumulative 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period during the preceeding 30 days of operation.

REFERENCE BYN, BGP 100-3, Pg. 2

{

i 7.

PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 33 RADIOLOGICAL CONTROL ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 7.13 (3.00) a.

350 F, 400 psig

[1.0]

[0.5]

b.

Decrease c.

1.

No operations are permitted that could cause dilution of the Reactor Coolant System boron concentration, AND 2.

Core outlet temperature is maintained at least 10 F below saturation temperature.

[.75 each]

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

8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONE PAGE 34 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 8.01 (2.50)

(Reportable or non-reportable which) warrants immediate response, a.

intensive investigation, and aggressive, timely corrective

[1.0) action.

b.

SCRE/[Also p me 3; st / sz>o / opry /,po+)

[0.5]

NRC and SDO/opty(by hv sv.M

[1.0) c.

REFERENCE BYN,BAP 1250-2, Pg. 3, BAP 1250-G p.3, and BAP 1250-T3.

ANSWER 8.02 (2.00) a.

44,000 gallons Unit 1 03.0%

" nit 2 00.0%

[ 1. 0 ]-

b.

450 gallons Unit 1 35%-

Unit-2--80%---

[1.0]

REFERENCE BYN, Byn Differences Book, Pgs 9-10.

ANSWER 8.03 (2.00)

[0.5]

a.

SCRE/ Control Room Supervisor.

Licensed (operator (must be specifically) b.

1.

assigned the responsibility of monitoring the controls of the unattended unit.)

2.

This same operator must remain within line of-sight of the unit's front panels.

The licensed operator must (on a periodic basis) 3.

review'the status of the unattended unit (from within the "at the controls" area) 4.

kW b o.6 m s 0.5 each]

REFERENCE BYN, BAP 300-1, Pg. 8-9

N 8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 35 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 8.04

(.50)

It shows that the particular instrument is physically located on the Unit 1

~ ide (and must be applied to Unit 2 Tech. Specs.)

[0.5]

s REFERENCE BYN, Byn~ Differences Book, P. 7 ANSWER 8.05 (2.00) a.

It is a common instrument between units.

[0.5]

~

b.

All of #1-4, only Unit I's instrument for #5, ORE-PR009 and XRE-PR009 (in #6) i

[1.5]

1 REFERENCE BYN, Tch. Specs. P. 3/4 3-73 Byn Difference Book, PP. 7-8 ANSWER 8.06 (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 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 BYN, Tech. Spec. Section 6, Pg. 6-5 i

,e

N

.i 8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 36 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

1 ANSWER 8.07 (3.00) p,weuJ.s 964.,,<.Aen)

Individuals qualified'in radiation protection procedures (or a.

personnel continuously escorted by such individuals)

[1.0]

b.

Locked doors [0?S] with controlled keys.

[0F5].

Area must be barricaded (by more than rope) [0.4],

c.

conspicuously posted [0.4], and a flashing auwk light shall be active [0.2].

REFERENCE BYN, Tech. Spec., Section 6, Pg. 6-24 ANSWER 8.08 (2.50)

A maximum allowable extension not to exceed 25% of the surveillance a.

interval [0,5], but the combined time interval for any 3 consecutive 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 /> 2.

At least once every 02 days 43.

At least once every 184 days

[1.0]

REFERENCE BYN, Tech. Specs., Section 3, Pg. 3/4 0-2 l

l l

ANSWER 8.09 (2.00)

It will be stated in the applicability section of the specification a.

4 a

[1.0]

i re% D 1 w.o.Au 4 mquirementsarewritteninthetextand) Unit 2 b.

(Unit are in parenthesis.

[1.0]

l REFERENCE BYN, Byn Differences Book, P. 7

,as.

8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 37 ANSWERS -- BYRON 1

-86/07/16-JAGGAR, F.

ANSWER 8.10 (2.00) a.

15 minutes

[0.5]

b.

When the event is classified by the Station Director.

[0.5]

c.

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

(do not provide) outside phone number [0.5].

REFERENCE BYN, BZP 300-1, Pg. 2 ANSWER 8.11 (2.00) a.

(An alarm that is in an) alarm condition for longer than 1 shift whenb.e3 greater than the P-8 setpoint.IoS1 si es b.

1. Submit an operator aid.(130-i)

2. Place a green plastic on the annunciator window.

3. Verify the annunciator illuminates with a green hue.

[1.0]

REFERENCE BYN, BAP 380-2, Pg. 1 ANSWER 8.12 (2.00) a.

On duty Shift Engineer.

[1.0]

b.

10.

5.

[.5]

c.

TRUE

[.5]

REFERENCE BYN, BAP 900-1, pg 1 and BAP 900-5, pg 1.

.,,.-..____m.-

- -. - - -