Text

Exam Repts 50-456/86-03 & 50-457/86-03 on 861027,1111-14 & 18-22.Exam results:15 of 17 Senior Reactor Operators & 6 of 7 Reactor Operators Passed Exams
	ML20207M359
Person / Time
Site:	Braidwood
Issue date:	01/07/1987
From:	Burdick T, Reidinger T; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20207M304	List: ML20207M298; ML20207M359;
References
	50-456-86-03, 50-456-86-3, 50-457-86-03, 50-457-86-3, NUDOCS 8701130155
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U.S. NUCLEAR REGULATORY COM1ISSION

REGION III

Report No. 50-456/50-457/86-03

Docket Nos. 50-456/50-457

Licensee:

Commonwealth Edison Company

ATTN:

Mr. Cordell Reed

Vice President

Post Office Box 767

Chicago, IL 60690

Facility Name:

Braidwood Nuclear Station

Examination Administered At:

Braidwood Nuclear Station

Examination Conducted: October 27, November 11-14, 18-22, 1986

Examiners:

B. C. Haagensen, Sonalysts

T. P. Guilfoil, Sonalysts

G. D. Weale, Sonalysts

I. J. Kingsley, Sonalysts

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K. L. Parkinson, Sonalysts

F. W. Victor, Sonalysts

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Chief Examiner

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Approved By:

T. M. Burdick, Chie

Operator Licensing Section

Date

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Examination Summary

Examination administered on October 22, 1986, and during the period of

November 11-14 and November 18-22, 1986, (Report No. 50-456/0L 86-03

Examinations were administered to 17 senior reactor operators, and 7 reactor

operators.

Results:

All but 2 of the 17 senior reactor operations passed the examination.

All but 1 of the 7 reactor operators passed the examination.

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REPORT DETAILS

Examination Review Meeting

Licensee coments and their resolutions are attached to this report.

Exit Meeting

a.

On November 21, 1986, an exit meeting was held. The following personnel

attended:

Eugene E. Fitzpatrick, Ceco, Braidwood Station Manager

K. C. Kofron, CECO, Production Superintendent

Kevin Bartes, Ceco, Training Supervisor

C. W. Schroeder, CECO, Services Superintendent

Thomas M. Tongue, NRC, Senior Resident

Ronald L. Higgins, NRC, Licensing Examiner, Region III

Paul R. Sunderland, NRC, Region III

Brian C. Haagensen, NRC/Sonalysts, Lead Examiner

Francis W. Victor, NRC/Sonalysts, Licensing Examiner

Ivan J. Kingsley, NRC/Sonalysts, Licensing Examiner

b.

The Lead Examiner discussed the following comments with the licensee

representatives:

(1) The candidates exhibited a tendency to race through procedural action

and verification steps at the expense of accurately completing the

steps.

In several instances, operator candidates missed steps in

the emergency and abnormal procedures because of excessive haste

to complete the procedure.

(2) Many candidates did not demonstrate sufficient knowledge of basic

radiation and exposure monitoring practices. They did not understand

proper frisking techniques and the allowable count rates above

background levels to constitute personnel contamination, they did

not know how to read RWP survey maps and they did not have a solid

understanding of personnel dosimetry equipment and its uses.

(3) Many licensing candidates did not demonstrate knowledge of valve

lineup and verification requirements. They did not know how to

conduct valve lineup checks on locked manual valves, manual throttled

valves and automatically operated valves.

(4) Many candidates did not demonstrate effective control of axial flux

difference (AFD) during simulator exams. Confusion existed regarding

the use of the iconic display, the computer data table display and the

AFD meters on the main control boards. When the process computer was

not operating correctly, some candidates had difficulty tracking AFD

using control board indicators and could not keep track of AFD penalty

points per tech specs.

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(5) Although the overall trend shows improvement from previous exams, the

candidates' operation of the plant from the remote shutdown panel is

still weak. The candidates did not demonstrate familiarity with the

procedure for operating from the remote shutdown panel (PRI-5) and in

several cases, did not carefully verify that all local / remote switches

were selected to the local position.

(6) Many candidates did not have sufficient knowledge of GSEP procedures

and had difficulty classifying events and completing the NARs forms

correctly.

In many cases, tha candidates did not understand the

selection of protective action recommendations and were unaware that a

default set of PARS existed at the initial classification of a general

emergency event.

Additionally, many candidates would classify the

events by finding the first match between the EAL table events and

actual plant conditions, but would never check other EAL table events

to determine if a higher classification was appropriate under a

different event.

(7) Many candidates showed confusion over the use of the steam dump valves

in the steam pressure mode of operation.

(8) Many candidates lacked detailed knowledge of the rod control system

alarms, components and effects of malfunctions.

(9) The use of " orange dots" to identify tripped protective bistables was

not documented by an administrative procedure and each shift used the

" orange dots" differently.

(10) Many candidates had difficulty discussing the reactivity effects on

flux level, startup rate, shutdown margin and coefficient variation.

c)

Although it was not discussed at the exit meeting, the Braidwood Nuclear

Station Training staff is requested to provide a copy of their plant

specific modification of the Knowledges and Abilities Catalog for Nuclear

Power Plant Operators: Pressurized Water Reactors (NUREG-1122) to the

region.

This will facilitate the region with the development of site

specific examinations for the Braidwood Nuclear Plant.

It is requested

prior to the next scheduled examination.

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QUESTION 1.03 (2.00)

Considering their production and removal rates, explain why:

a.

Equilibrium concentration of Xenon increases as reactor power increases.

(1.0)

NRC ANSWER 1.03 (2.00)

a.

For equilibrium, Xenon removal by decay must increase as power increases

(0.5).

Xenon removal by decay is proportional to Xenon concentration (0.5).

(Therefore, Xenon concentration increases as power increases - optional.)

(1.0)

BRAIDWOOD CONTENTION:

1.03 a.

Westinghouse Reactor Theory Text, Chapter 4, Fission Product Poisoning

Effects describes Xenon characteristics.

These characteristics at

equilibrium conditions indicate that the rate of change of Xenon

concentration is equal to the production rate minus the loss rate.

As power is increased from an equilibrium condition, eventually Xenon

production exceeds Xenon removal due to the higher flux level

(production terms).

However, as Xenon concentration increases

the decay of Xenon and Xenon burnup become more significant and

establish a new equilibrium condition at a higher Xenon concentration,

since burnup is a function of Xenon concentration and flux, while

decay is a function of Xenon concentration only.

REFERENCE:

Westinghouse Text Fundamentals of Nuclear Reactor Physics

Pages 4-11 through 4-14.

1.03a

Resolution

The facility contention provides additional information concerning the

relationship between power level and xenon concentration.

Candidate answers

that provided similar amplifying information plus the cause-and effect

relationship stated in the NRC answer were not graded wrong.

QUESTION 1.05 (2.75)

a.

Select / match one equation from the right-hand column that applies to each

of the following heat transfer rates in a steam generator (SG).

(0.75)

(1) Rate of heat gain by feedwater/ steam

(a) h = [nCp At

(b) h=UAAH

(c) h=mCpAH

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(d) Q = wf AH (wf = Wp)

(e) h=UAAT

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b.

Define the following symbols / terms used in the above equations as they

apply to a steam generator.

(0.25)

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(7) Wf = Wp

NRC ANSWER 1.05 (2.75)

a.

(1) (d)

(0.25 each)

b.

(7) Wf feed flow rate or steam flow rate or secondary coolant flow rate

or mass flow rate.

(0.25)

BRAIDWOOD CONTENTION:

hfisnotastandardsymbolatBraidwoodforfeedflowrate.

The symbol used

for feed flow rate is mf or mfeed.

The us2 of Wf may cause some candidates to

be unable to recognize "(d)" as the correct match for "(1)" in Part a. and

unable to define Wf in "(7)" of Part b.

REFERENCE:

Thermal-Hydraulic Principles

1.05

Resolution

The equation Pwr = Wf Ah appears on the standard NRC license exam equation sheet

that was provided to the candidates with this exam.

Full credit was awarded to

answers that provided the Braidwood equation (Q = mf Ab) for " Rate of heat gain

by feedwater/ steam" and to definitions of m as feed flow rate or steam flow rate.

QUESTION 1.06 (.75)

In a simple " closed" cooling water system (similar to the CCW system, but with

just one centrifugal pump, a surge tank, and a heat exchanger loop), the

isolation valve in the line to the surge tank is inadvertently shut and the pump

discharge flow-control valve is fully opened.

Soon afterward the pump starts

operating noisily and vibrating excessively due to cavitation.

In this situation:

b.

What is one reason why opening the surge tank isolation valve halfway

open will stop the pump vibrations / noisy operations?

(0.25)

NRC ANSWER 1.06b.

b.

Restores the source of required NPSH.

(0.25).

BRAIDWOOD CONTENTION:

1.06b.

Answer should be " restores the source of NPSH available."

REFERENCE:

Thermal-hydraulic Principles, Volume II, Page 10-55.

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1.06

Resolution

Candidate answers that stated opening the surge tank isolation valve halfway

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open restores the source of required or available NPSH were awarded full credit.

QUESTION 3.07 (2.25)

For the following actions / occurrences, state the:

(1) Steam dump control system response,

(2) The plant response, and

(3) The resulting method of RCS temperature control.

Assume all systems operate normally except as stated and that no operator action

is taken.

Consider each case separately.

c.

The Train B Reactor Trip Breaker fails to open on a reactor trip signal

while at 78% power.

NOTE:

The Train A breaker opens properly.

(0.75)

NRC ANSWER:

c.

(1) Normal shift to the Turbine Trip controller will not occur.

(0.75)

(2) Plant will cooldown because Load Rejection controller will open

steam dumps.

(0.25)

(3) The Load Rejection controller will maintain RCS temperature near "No

Load" Tref (+4F deviation /deadband).

(0.25)

REFERENCE:

BW SYST TRNG MAN FIG 24-7, 9

BRAIDWOOD CONTENTION:

3.07 Part c.

C-8 is picked up and the turbine trip controller is activated

and will control the steam dumps to reduce Tave to no load value.

Since an actual Rx trip did occur the STM DUMP SYSTEM will

function normally.

REFERENCE:

BW SYST TRNG MAN.

ATT. A Page 24-32

Figure 24-9

3.07c

Resolution

The detailed circuit and logic diagrams needed to confirm the facility

contention were not included in the provided reference material package.

Since the provided reference material dose not contradict the facility

contention, full credit was awarded to candidate answers that stated the

following:

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(1) Steam dump control shifts to the Turbine Trip (Plant Trip)

controller.

(2) Plant temperature decreases to "no-load" Tavg.

(3) Steam dumping maintains RCS temperature near no-load Tavg.

QUESTION 3.15 (1.00)

Identify the control, protective, and permissive functions which use

individual loop Tave signals and NOT auctioneered Tavg.

ANSWER 3.15 (1.00)

1.

OT Delta-T calculator

(0.25)

2.

OP Delta-T calculator

(0.25)

3.

P-12 circuitry (Hi Stm. Flow SI permissive, Stm. dump block)

(0.25)

4.

Feedwater isolation circuitry

BRAIDWOOD CONTENTION 3.15:

List of setpoints vice names should be acceptable.

Part 3 Tave 550 = P-12

Part 4 Tave 564 = FW isolation setpoint

3.15 Resolution

The question requested " control, protective, and permissive functions," not

setpoints.

In Part 3 full credit was awarded to candidate answers that stated

P-12 or its equivalent functions (Steam Dump Block, Lo-Lo Tavg, or Hi Steam

Flow SI Permissive).

In Part 4 full credit was awarded to candidate answers

that stated feedwater isolation.

QUESTION 4.15 (1.50)

According to the BW Precautions, Limitations, and Setpoints book, all reactor

trip and safeguard actuation channels, except three, shall be placed in the

trip mode when the channel is out of service for any reason.

List the three

excepted circuits / trips that may be bypassed for maintenance.

(1.5)

ANSWER 4.15 (1.50)

1)

Source range hi flux trip

2)

Intermediate range hi flux trip

3)

Containment pressure hi-2 spray actuation (0.5 each)

REFERENCE:

BW PLS PG 7

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BRAIDWOOD CONTENTION:

Answer key Part 3

3.

Containment pressure hi-3 spray actuation.

REFERENCE:

BW PLS PG-10

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BW SYST TRNG MAN. 59-Page 21

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4.15 Rasolution

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Paragraph A.11.b of PRECAUTIONS, LIMITATIONS, AND.SETPOINTS FOR NUCLEAR STEAM

SUPPLY SYSTEMS (Revision 6, March 1986) for Braidwood Station Units 1 and 2

states, in part,

"The reactor trip and safeguards actuation circuits noted below may

be administrative 1y bypassed for maintenance on a single channel.

3)

Containment high-high pressure spray actuation."

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Full credit was awarded to candidate answers that stated

3)

Containment pressure Hi-3 spray actuation.

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NOTE:

The facility should designate the correction / resolution of

Section A.11.b.3) as an open item.

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Question 5.06 (1.50)

Compare the calculated Estimated Critical Rod Position (ECP) for a startup 15

hours after a trip to the Actual Critical Rod Position (ACP) if the following

events / conditions occurred.

Consider each independently.

Limit your answer

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ACP higher than ECP.

b.

ACP lower than ECP.

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c.

ACP is unchanged.

1.

Actual RCS temperature is 551*F at criticality.

(0,5)

NRC ANSWER 5.06

1.

b.

(ACP lower than ECP)

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BRAIDWOOD CONTENTION:

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The question does not state the temperature at which the ECP was calculated

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for.

Prerequisite number of one of BwGP 100-2 states that to perform a plant

startup, the plant must be in a Hot Standby (Mode 3) condition with all RCS

loop temperatures greater than 550'F.

Therefore, an ECO could be calculated

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with Tavg equal to 551*F and the reactivity effects would be corrected for the

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deviation form 557'F by the ECP calculation, BwGP 100-A8.

Therefore the ACP

could be equal to the ECP or even higher than the ECP if 550 F were used for

ECP calculation.

Reference:

BwGP 100-2, BwGP 100A8

5.06 Response

The Braidwood ECP (BwGP 100-A8) Item B.3 provides the operator with a step to

adjust the TAVE expected at startup if it is different from 557*F.

The

question explicitly did not provide this adjustment because part of the full

credit answer was to understand that the default TAVE for calculating the ECP

was 557'F. With no valve of "TAVE expected at Startup

provided,

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the candidate should use the default value of 557*F for comparison in arriving

at the correct answer.

If the candidate assumes a TAVE at startup that is different than 557 F, then

full credit will be given for the correct answer based on his assumed value.

Answer scores indicate that 13 of 14 candidates provided the correct response

and therefore, most candidates were not confused by the question.

QUESTION 5.11 (2.5)

b.

When natural circulation is established in the RCS, mass flow rate of the

coolant is proportional to:

(0,5)

1.

(delta-T)

2.

(delta-T)2

3.

(delta-T)3

4.

1/(delta-T)

c.

The reactor is producing 100% rated thermal power at a core delta-T of 60

degrees and a RCS mass flow rate of 100% when a station backout occurs.

Natural circulation is established and core delta-T goes to 28 F.

If

decay heat is 2%, what is the core mass flow rate (in %)?

(1.0)

NRC ANSWER 5.11

b.

2.

(0,5)

c.

To determine flow in NC:

h=acpdelta-T

h = M1 cp1 (delta-T)

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m2 cp2 (delta-T)

Q2

b2 = h ml cp1 (delta-T)

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h

cp2 (delta-T)

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If the candidate assumes cp1 = cp2 then:

m2 = 2% x 100% x 60'F = 4.3% flow

100%

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28"F

If the candidate determines cp1 and cp2 from the figure provided, then:

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m2 = 2% x 100% x 145 x 60'F = 3.9% flow

100%

x

1.60 x 28"F

(accept - or - 0.2% from answer)

BRAIDWOOD CONTENTION:

5.11.b.

No correct answer is given.

Correct answer is (delta-T)b.

As

shown in the question reference (Thermal-Hydraulic Principles,

Volume II, Page 14-25),

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m a (delta-T)

Therefore m a (delta-T)b

Part b. of Question 5.11 should either be eliminated or any response

accepted.

REFERENCE:

Thermal-Hydraulic Principles, Volume II, Page 14-25.

5.11.c.

An alternative method for solving the problem should also be

accepted.

A commonly used thumb rule derived from natural

circulation testing is that decay heat equal to 6% full power

causes a stable natural circulation mass flow rate of approximately

6% of full flow (Page 14-23 of Thermal-Hydraulic Principles,

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Volume II).

The correct relationship is that Q o m , however since

the candidates were forced to select an incorrect relationship in

Part b, a calculation consistent with the selected Part b answer

should be accepted.

REFERENCE:

Thermal-Hydraulic Principles, Volume II, Page 14-23.

5.11 Response

1.

Part b.

A typographical error in the set of matching answers

provided to the candidate caused none of the correct

answers to appear.

Part "b" of the question is deleted.

2.

Part c.

The proposed alternative method provided by the supporting

documentation in the Braidwood contention is theoretically

valid.

However, this alternative method requires the

candidate to assume a set of initial conditions that were

not provided as part of the answer key.

The set of

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initial conditions that was provided to the candidates did

not allow the use of the alternative method because these

conditions were for forced circulation flow where the

Qm relationship does not apply.

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Based on a review of the response, the examiner concluded

that the candidates were taught the proposed alternative

method and a thumb rule of "6% decay heat yields 60*F

delta-T and 6% natural circulation flow" to provide initial

conditions for rationing the equation.

Based on the

validity of the method and the clear preference to reach

this alternative answer, full credit will be given if the

answer given in the Braidwood contention is used.

However, the licensee should note that the figure

(FN0-0PS-12) provided in the contention to support the

alternative method only supports the thumb rule and

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Q

m relationship for the theoretical curve.

The actual

4 loop test results appear as a linear relationship (Q m)

and 4.78% decay heat yields 6% flow (27% error between the

thumb rule and test results) instead of the 6% decay heat

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yielding 6% flow thumb rule.

Additionally, the material

provided does not make reference to this thumb rule as a

convention that is accepted (or validated) for Braidwood

and it is not clear that all candidates are held

responsible for knowing the thumb rule.

3.

The purpose of this question was to test the candidate's

understanding of the natural circulation thermal

hydraulics and, specifically, the sensible heat equation

Q = m cp 'T.

The initial conditions provided did not

allow solving this problem any other way.

The

relationship in Part "b" could not be used to solve

Part "c" because the initial conditions were given for

forced circulation flow and the m2 AT relationship does

not apply to forced circulation conditions.

Hence, the

candidate who attempted to use the relationship in

Part "b" with the initial conditions provided to solve

Part "c" would automatically arrive at an incorrect

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answer.

QUESTION 6.02 (2.00)

a.

Describe an IR instrument response if the circuitry in undercompensated

while performing a reactor startup, including any effects on SR

instrumentation.

Include any applicable setpoints.

(1.0)

NRC ANSWER 6.02 (2.00)

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Undercompensating results in a higher than actual reading, (0.50) and if

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IR instruments read > 1.0E-10 amps (P-6 set point) upon entering the

source range, then P-6 will prevent the SR detectors from energl2,ing and

providing high SR trip protection when high in the SR.

(0.5)

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BRAIDWOOD CONTENTION:

6.02

Answer in the Key is incorrect.

The questions asks for IR instrument

response during a reactor Startup, not a shutdown as answered in the

key.

With an IR instrument undercompensated during a reactor startup, the

a.

P-6 setpoint (10 10 amps) will be reached early as a result of

counting neutrons and gammas.

The operator may secure the source

range instruments when P-6 comes in which results in no indication

of neutron flux below the IR range.

6.02

Response

The proposed response in the Braidwood contention is correct.

This

change was inadvertently omitted from the copy of the answer key

provided to the utility but was made to answer key prior to

administering the exam and all candidates were graded on this

basis.

QUESTION 6.06 (2.50)

List FOUR signals (that are not generated from the same protective

a.

bistable) which will initiate a motor driven Auxiliary Feedwater Auto

Start (AFAS) signal.

NRC ANSWER 6.06 (2.50)

a.

1.

Manual

2.

S/G Iow-low level on 1 steam generator

3.

Safety Injection sequence signal

4.

Undervoltage on Bus 141 (sequenced on)

5.

Undervoltage on 2/4 RCP buses (loss of offsite power)

(any four at .25 each)

(0.50)

BRAIDWOOD CONTENTION:

6.06

a.

There are not five Auto start signals, only four.

1.

2/4 Low-Low Steam Generator

level in any SG.

2.

UV on 2/4 RCP Buses (Loss of offsite power).

3.

(IA only) UV on Bus 141 (Sequenced only).

4.

SI signal

REFERENCE:

BW STM VOL. 3 CH 26 (26-42)

6.06

Response

The documentation provided to support this contention underlines

the confusion.

" Manual initiation (1/2 coincidence)" appears to

be separated from the three other safety injection signals.

Upon

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closer review, the " " mark instead of a "0"

shows that " manual"

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is a cause of the SI signal, not the AFAS signal.

Logic diagrams

provided in the STM do not show which signal is the AFAS signal and

further confuse the answer.

Only one of 15 candidates listed

" manual" as a cause of an AFAS signal and therefore, the examiner

concluded that the training program provided sufficient

clarification on this answer.

Therefore, the answer " manual"

is deleted from the answer key.

The candidates must provide

4/4 signals that cause an AFAS signal for full credit.

QUESTION 6.08

b.

How is the minimum CST water volume required by Technical Specifications

ensured to be available at all times considering that several systems

take suction on the CST?

(0.5)

NRC ANSWER 6.08

b.

Other systems which use the CST have evaluated suction nozzles (which tap

in above the minimum required level).

(0.5)

BRAIDWOOD CONTENTION:

6.08 Part b.

By the T/S surveillance.

The CST does not have elevated nozzles supplying other systems.

REFERENCE: T/S 4.7.1.3.1.

General plan drawing of Equip Nos. ICD 01T.

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6.08

Response

Tech spec basis 4.7.1.3 for the condensate storage tank implied

that there were elevated discharge lines or other " physical

characteristics" that cause water to be not usable for supply to

some systems.

It is typical throughout the industry for CST suction

lines to be elevated to ensure sufficient water volume will be

available to supply AFW to meet tech spec requirements for

operability.

Indeed, some discharge lines do have elevated

nozzles in the CST as shown in general DrawTrig ICD 01T.

However, based on general Drawing 1 CD01T recently supplied by the

licensee to support the contention (a drawing not included in the

references provided to prepare the exam) the answer is modified to

state:

b.

"By Tech Spec surveillance."

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It should be noted that three of 15 candidates stated that the

CSF had elevated suction nozzles to ensure sufficient water was

available for AFW operability.

QUESTION 7.01 (2.25)

a.

18wEP.0 foldout lists three conditions (or sets of conditions) that

together require the operator to trip one or more RCPs.

What are these

conditions:

(1.25)

b.

What are the ERG bases for tripping the RCPs under these conditions (1.0)

(during a small break LOCA)?

ANSWER 7.01 (2.25)

a.

1.

Component cooling water to RCP lost (affected pump only)

(0.25)

2.

CNMT Phase B is actuated

(0.25)

3.

Both of the following conditions exist:

a.

RCS pressure is less than 1370 psig (1670 psig for adverse

containment)

(0.25)

and

b.

Coolant charging pump flow is greater than 200 gpm

(0.25)

or

Safety injection pumps have positive flow

(0.25)

BRAIDWOOD CONTENTION:

7.01

The question asks for " conditions that taken together require the

operator to trip RCPs".

This combined with the clarification of Part b. (that being the

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small break loca) could have misled the examinee to believe the

answer to Part a would only be.

1. - RCS Pressure < 1370 (167C adverse)

and

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2. - charging pump flow > 200 gpm

or

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3.

positive SI flow

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7.01

Response

A large majority of the candidates did not find this question to be

confusing.

12 out of 14 candidates achieved full credit on Part "a"

of this question.

Of the two remaining candidates, one had initially

listed a full credit response and then lined our Answers 1 and 2.

-

Full credit has been given for his answers because:

1.

It is clear that he knew Parts 1 and 2 to the question answer

key.

2.

The word "together" probably did not cause confusion and he

lined out the answers after reading Part "b" of the question.

Deleting the first two answers to Part "a" would only penalize the

candidates unnecessarily.

QUESTION 7.03 (4.00)

a.

List four Mode 1 situations or conditions which require the operator to

commence EMERGENCY B0 RATION of the RCS.

(2.0)

NRC ANSWER 7.03 (4.00)

a.

1.

Failure of more than one RCCA to fully insert following a

reactor trip.

2.

CRH below RIL

3.

Inadequate shutdown margin

4.

Unexplained or uncontrolled reactivity increase.

5.

Inability to borate normally.

(any four at .5 each)

BRAIDWOOD CONTENTION:

7.03

A sixth choice is an uncontrolled cooldown.

REFERENCE:

18w0A PRI-2 B.

SYMPTOMS OF ENTY CONDITIONS

7.03

Response

An " uncontrolled cooldown" is part of Answer 4, an " uncontrolled

reactivity increase."

Credit will be given of the operators provide specific conditions of

situations that are subsets of the general conditions in the answer

key provided these specific situations are not redundant.

The

16

1

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,

question did not require the candidate to list the six entry

conditions from the 18w0A Pri 2 B foldout and latitude will be

allowed for answers which are subsets of the general situation.

In addition, it should be noted that the Braidwood Systems

Training Manual Volume 2, Chapter 15b Appendix A does not include

uncontrolled cooldown as a separate situation or condition which

requires emergency boration.

QUESTION 7.07 (2.00)

a.

Complete the following statements regarding a reactor startup.

3.

Shutdown margin shall be greater than

delta k/k tech

spec limit) prior to commencing a startup.

(0.5)

NRC ANSWER 7.07 (2.00)

3.

1.3 delta k/k

(0.5)

REFERENCE:

BwGP 100-2, Pages 1, 2, 5, and 9

BRAIDWOOD CONTENTION:

7.07 a.3. Answer should be 1.3% ^k/k

REFERENCE:

BwGP 100-2, Page 1

Braidwood Tech. Spec. 3.1.1.1

7.07

Response

The "%" sign was omitted from the answer key by a typographic

error.

The answer key has been modified to accept the contention.

the correct answer is 1.3% ^K/K.

QUESTION 7.12 (2.00)

During a loss of all AC power, the operator must depressurize the RCS using

steam generator atmospheric relief valves at Step 16 of 18wCA-0.0.

How should these valves be operated without AC power and control air?

a.

(0.5)

NRC ANSWER 7.12 (2.00)

They have backup nitrogen accumulators and can be positioned from the

a.

control room.

(0.5)

17

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BRAIDWOOD CONTENTION:

7.12

a.

By using the hand pump loss of AC Procedure BwCA-0.0

Step 16-b.

locally dump steam using PORV per Bw0A Pri 5.

Bw0A Pri 5 Step 20-b directs you to Bw0P MS-6.

Bw0P MS-6 Step 5, operate the remote hand pump to move to

Atmospheric Steam Pump.

7.12

Response

During a loss of all AC power, the steam generator atmospheric

relief valves may be operated in the emergency mode to fast close

the valve with a nitrogen accumulator as stated in the answer key

referenced in the Systems Training Manual Volume 3 Page 23-33.

The

Braidwood contention provides an alternative method of handpumping

the valve closed or open from a local operating station.

The answer

key has been revised to accept the following answer based on the

information supplied:

Remote manual mode --- If the atmosphere relief valves were to lose

power they could be manually opened with a hand pump in the safety

valve rooms.

(0.5)

QUESTION 8.03 (2.00)

a.

How many members per shift are required on the fire brigade.

(0.5)

b.

Who may NOT be included as members of the fire brigade?

(1.0)

NRC ANSWER 8.03 (2.00)

a.

five

(0.5)

b.

The fire brigade shall NOT include the minimum shift crew required'for

safe shutdown of Unit 1 (0.5) and any personnel required for essential

functions during the fire.

(0.5)

REFERENCE: BwAP 1100-1

BW Technical Specifications, Section 6

BRAIDWOOD CONTENTION:

8.03

a.

Answer should be five, or cognizant Shift Foreman and remaining

four members composed of four of the following personnel:

Equipment Attend?nts, Equipment Operators.

Reference:

BwAP 1100-1, Rev. O, Page 7.

,

b.

Another answer that should be allowed is:

1-SE, 2-NSP's,

3-EA's, and 1 SCRE/STA.

REFERENCE:

BwAP 320-1, Page 2

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8.03

Response

The Braidwood contention is a subset of the more general answer

provided.

If the candidate provides the specific list by job

title, full credit will be give. However this information is

neither required nor asked for to receive full credit.

QUESTION 8.07 (3.00)

List SIX conditions or occurrences that require notification of the NRC within

ONE hour.

NRC ANSWER 8.07 (3.00)

1.

Declaration of any emergency classification.

2.

The initiation of any plant shutdown required by Technical

Specifications.

3.

Any authorized deviation (10 CFR 50) from Technical Specifications.

4.

Any event or condition during operation that results in the condition of

the plant or safety barriers being seriously degraded, or results in the

plant being:

a.

In an unanalyzed condition that significantly compromises plant

safety, or

b.

In a condition that is outside the design basis of the plant, or

c.

In a condition not covered by the plant's operating and emergency

procedures.

5.

Any natural phenomenon or external condition that poses an actual

threat to safety of plant or significantly hampers site personnel

while performing duties required for safe operation of the plant.

6.

Any event that results or should have resulted in ECCS actuation on a

valid signal.

7.

Any event that results in a major loss of emergency assessment

capability.

.

8.

Any event that poses an actual threat to the safety of the plant of

significantly hampers site personnel while performing duties necessary

for safe operation of the plant including fires, toxic gas releases, or

radioactive releases.

9.

Any violation of a safety limit.

(any six at 0.5 each)

!

(No. 4 counts as four separate conditions)

19

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BRAIDWOOD CONTENTION:

8.07 Addition correct answers should include:

10.

Exposure of the whole body of any individual to 25 rems or more of

radiation; exposure of the skin of the whole body of any individual of

150 rems or more or radiation; or exposure of the feet, ankles, hands or

forearms of any individual to 375 rems or more of radiation; or

11. The release of radioactive material in concentrations which, if averaged

over a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , would exceed 5,000 times the limits specified

for such materials in Appendix B, Table II of this part; or

12.

A loss of one working week or more of the operation of any facilities

affected; or

13.

Damage to property in excess of $200,000.

REFERENCE:

10 CFR 20.403

8.07

Response

The 10 CFR 20.403 events require "immediate notification" of the

NRC not notification within one hour.

However, it is clear from

the candidate responses that the 10 CFR 20.403 conditions were

emphasized in the training program.

Based on the candidate response

and the documentation provided, the 10 CFR 20.403 events have been

included in the answer key for this question.

QUESTION 8.08 (1.00)

Complete the following statement with one of the provided terms.

Seal leakoff from the RCP No. 2 seal which is collected in the RCDT is

classified as

leakage.

a.

Controlled

b.

Pressure Boundary

,

c.

Identified

d.

Unidentified

i

NRC ANSWER 8.08 (1.00)

a.

(1,0)

BRAIDWOOD CONTENTION:

8.08

Tech Spec's define controlled leakage as that seal water flow

supplied to the RCP seals.

The No. 2 seal leakoff is collected in

the RCOT, which infers that it is Identified Leakage in recordance

with the Tech Spec. definition.

Therefore, the appropriate answer

is "c" - identified leakage.

20

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. -

. . - - - - _ . . - . - . -

- -

.

-

._-

-

.

--

-.

-

.

.. ,t,*.

REFERENCE:

Braidwood Tech Spec's Braidwood STM Chapter 13 Page 18.

>

8.08

Response

4

The term, " controlled leakage" is specifically reserved for the

seal water from reactor coolant pump seals because the leakage is

predetermined and controlled by ensuring that the gap between the

seal ring and runner is maintained constant.

The seal water

supplied to reactor coolant pump seals is also the same seal water

that lubricates and leaks by these seals. No distinction should be

4 -

drawn between the water going into the coming out of the seal

!

boundary (or else you will count this leakage twice).

The RCP No. 2 seal leakoff, although collected in the RCDT, cannot

.

{

be identified leakage because the definition of identified leakage

specifically excluded controlled leakage. The licensee has

'

interpreted the tech spec definition too literally.

The answer is

correct as provided in the answer key.

1

Reference:

BW Tech Spec Definition 1.8 and 1.15

j

BW STM Vol 2 Page 13-15 Paragraph 2

]

QUESTION 8.12 (1.00)

1

!

The Acting Station Director declares a general emergency event based on LOCA

l

into containment with a failure of ECCS to actuate.

No releases of

l

radioactive materials has occurred.

a.

What protective actions, if any, would the Acting Station Director

recommend to the offsite civil authorities?

(0.5)

1

NRC ANSWER 8.12 (1.00)

il'

a.

Shelter the two mile radius and five mile downwind sectors.

(0.5)

!

j

BRAIDWOOD CONTENTION:

)

>

At Braidwood we enforce that is it of the utmost importance that the SR0 be

aware of the fact that a plan exists for protective actions to be taken in the

i

event of a general emergency, and tnat he be able to locate and implement this

4

plan in a timely manner. We do not, however, believe or instruct that the

candidate should be required to commit to memory the chart in our BwZP's which

outlines the protective action guidelines for a general emergency.

Therefore,

we contend that this portion of the question is beyond the scope of required

material for SRO knowledge and the question should be deleted,

f

8.12

Response

l

The declaration of a general emergency classification at Braidwood

i

automatically requires that the SR0 assumes the role of Acting

Station Director, activate the GSEP and recommend protective

'

actions to the general public. The BWZP 380-4A2 Appendix B

Figure 6.3-1 flow chart provides detailed guidance, as mentioned in

!

4

21

E-_...,______._..____.___________

.

. .

.* 1 f

,

the contention, to determine protective action recommendations (PARS)

and certainly should not be committed to memory.

The correct answer

does not require the candidate to memorize this complicated flow

chart.

However, the intent of this question was to determine if the

candidate know that:

1.

Upon declaring a general emergency, there is only one set of

PARS that should be recommended immediately, and

2.

These PARS are to shelter the two mile radius and five mile

downwind sectors.

This basic level of knowledge is required for SR0s to demonstrate

adequate familiarity with the emergency plan.

The correct answer

did not require the candidate to specifically detail the NARs

paragraph numbers from the flow chart.

Instead, the candidate was

tested on his understanding of the basic concepts of protecting the

general public from exposure to accident release levels of radiation

by immediate sheltering of population at the general emergency

declaration.

NUREG-1122, Knowledges and Abilities Catalogue, assigned one of

to the candidates " ability

thehighestimportanceratings(4.7/5.0)tyEmergency

to take actions called for in the Facili

including . . . acting as the Emercency Plan Coordinator"

Acting Station Director) (Plant-Wice Generic K/A 36, Page 2-3)..e., The

candidate must have more than a cursory knowledge of PARS to meet

this ability and, therefore, this level of knowledge is appropriate

for SR0s.

QUESTION 8.14 (2.50)

When conducting valve lineups, the valve position must be determined by the

operator. How would the operator verify valve position on the following types

of valves?

e.

Manual valve in a throttled position.

(0.5)

NRC ANSWER 8.14 (2.50)

i

!

e.

Close valve counting the number of turns until the valve seats, restore

the valve to its proper position by pening it the same number of turns

that is was closed (check at facilit ).

(0.5)

BRAIDWOOD CONTENTION:

8.14

As requested, the reference for Answer "e" is BwAP 340-2 Rev. 0,

Page 3.

l

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8.14

Response

IBWAP 340-2 Page 3 Item 6 does not provide an explicit statement of

action to the operator conducting the value line up regarding how to

check the position of a throttled valve.

This was pointed out to

the licensee is both the answer key provided and in person to the

training department.

The examiner requested that the licensee

provide an explicit answer that paralleled the phraseology used

in BWAP 340-2 Page 3 Items 1, 2, 3, 4 and 5.

Instead, the training

department response was to provide a copy of the page of the

administrative procedure that the examiner had already referenced

in the answer key and had considered an inadequate answer to the

question.

Further discussions with the licensee have indicated that the

following answer is correct.

Answer:

1.

If the throttled valve position is listed on the valve line up

sheet by number of turns, then . . . (use answer key provided

to facility).

2.

If the throttled valve position is listed by stem position (%

travel on other indication) then the operator should compare

the indicated position with the valve line up sheet valve.

The operator should not operate the valve.

The key element required in 'this question was to see if operators

understood when to operate a throttle valve to check its position.

The examiner does not expect that the operators have memorized the

sentence in BWAR 340-2 Page 3 Item 6.

The successful answer should

have an action statement regarding how to check throttle valve

position that demonstrates their knowledge of the required actions

analogous to the action statements for manual valves.

If the candidate assumed that valve position was given by the number

of turns, then full credit was given for Answer 1.

If the candi gte

-

assumed that valve position was given by a stem position indicatig'n,

then full credit was given for Answer No. 2.

23

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B99 AN C. Hm6ENsaJ

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.

U. S. NUCLEAR REGULATOR

'

,

SENIOR REACTOR OPERATOR LICENSE EXAMINATION

,m,[ Q

FACILITY:

BRAIDWOOD 182

6

REACTOR TYPE:

PWR-WEC4

\\

= ped smocle, DATE ADMINISTERED: 86/10/22

-

b

EXAMINER:

HAAGENSEN, B./REIDIN

eErde. traole.

-

M w 4so. y. CANDIDATE:

INSTRUCTIONS TO CANDIDATE:

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

Staple question sheet on top of the answer sheets.

Points for eac'.

question are indicated in parentheses af ter the question. The passi :

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

CANDIDATE'S CATEGORY

VALUE

TOTAL

SCORE

VALUE

CATEGORY

25.00

1.

PRINCIPLES OF NUCLEAR POWER

PLANT OPERATION, THERMODYNAMICS,

HEAT TRANSFER AND FLUID FLOW

25.00

2.

PLANT DESIGN INCLUDING SAFETY

AND EMERGENCY SYSTEMS

25.00

3.

INSTRUMENTS AND CONTROLS

25.00

4.

PROCEDURES - NORMAL, ABNORMAL,

EMERGENCY AND RADIOLOGICAL

CONTROL

100.00

Totals

Final Grade

All work done on this exami .ation is my own.

I have neither given

nor received aid.

Candidate's Signature

_ _ _ _ _ _ _ .

_ _ .

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS

During the administration 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 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).

l

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.

Consecutively number each answer sheet, write "End of Category 'ne" side

8.

'

as

appropriate, start each category on a new page, write only on o

,

of the paper, and write "Last Page" on the last answer sheet?

'

9.

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

10. Skip at least three lines between each answer.

11. Separate answer sheets 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 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.

I

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE

QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

16. If parts of the examination are not clear as to intent, ask questions of

the examiner enly.

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.

,

I

4

,

-n

--.--.-.-,-.-n-.._.,--_,,,---,,,_.,,-.-.__,,._,-..,,,..-.,.,..n,.,

- . , .

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

n,, , - - - - ,. ,

. -

' -

.

18. When you complete your examination, you shall:

,

a.

Assemble your examination as follows:

(1) Exam questions on top.

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

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

b.

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

the examination questions.

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

c.

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.

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1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

PAGE

2

THERMODYNAMICS, HEAT 1RANSFER AND FLUID FLOW

., .

.

,

QUESTION 1.01

(2.00)

With the plant at 70% power and all systems in automatic, how AND why is

operating Shutdcwn Margin affected(INCREASE, DECREASE, NO CHANGE)

by the following occurrer.ces? EXPLAIN YOUR ANSWER.

a. RCS Boron concentration is increased by 10 ppm.

(1.00

b. Power is increased by 10% without dilution.

(1.00)

QUESTIOR 1.02

(3.00)

An Estimated Critical Position (ECP) is calculated for a reactor startup

to be performed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after a trip from a month at 100% power near EOL.,

If the following events or conditions occur, but the same ECP is used,

state whether the Actual Critical Position (ACP) is HIGHER THAN, LOWER THAN,

or the SAME as the ECP and explain WHY.

Consider each event or condition separately.

a. The startup is delayed for 8 more hours after the trip.

(0.75)

b. The startup is delayed for 8 days after the trip.

(0.75)

c.

The steam dump pressure setpoint is increased by 75 psig.

(0.75)

d.

All steam generator it:yels are being raised by 5% as criticality is

being approached.

(0.75)

QUESTION 1.03

(2.00)

Considering their production and removal rates, explain why:

a. Equilibrium concentration of Xenon increases as reactor power increases

(1.0)

b. Equilibrium concentration of Samarium remains constant regardless of

reactor power level.

(1.0)

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

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PRINCIPLES OF NUCLEAR. POWER PLANT OPERATION,

PAGE

3

THERMDDYNAMICS, HEAT TRANSFER AAD FLUID FLOW

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QUESTION 1.04

(1.50)

A motor-drivin, variable-speed centrifugal pump is operating at 1/4 speed

in a " closed" cooling water system with the following parameters:

(a) Current = 20 amps

(b) Flow = 50 gpm

(c) Pump delta-P = 8 psid

What are the new values for these parameters when the pump is shifted to

full rated speed?

(1.50)

QUESTION 1.05

(2.75)

a. Select / match one equation from the right-hand column that applies to

each of the following heat transfer rates in a steam generator (SG) (0.75)

.

a)

Q = mCp At

1) Rate of heat gain by feedwater/ steam

b)

h = UA AH

2) Rate of heat loss by reactor coolant

.

c)

Q = mCp AH

3) Rate of heat transfer thru SG tubewalls

.

d)

Q = WfAH

mh Idf = lJp

e)

h = UA AT

b. Define the following symbols / terms used in the above equations AS THEY

APPLY TO A STEAM GENERATOR.

(2.0)

.

1) m

2) Cp

3) At

4) U

5) A

6) A H

7)hf

sek hf3 $)p

8) AT

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

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

PAGE

4

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

,

.

QUESTION 1.06

( .75)

'

In a simple " closed" cooling water system (similar to the CCW system, but

with just one centrifugal pump, a surge tank, and a heat exchanger loop),

the isolation valve in the line to the surge tank is inadvertently shut and

the pump discharge flow-control valve is fully opened. Soon afterward the

pump starts operating noisily and vibrating excessively due to cavitation.

In this situation:

<

a. What are 2 reasons why shutting the pump discharge flow-control valve

halfway shut will stop the pump vibrations / noisy operations?

(0.5)

b. What is 1 reason why opening the surge tank isolation valve halfway

open will stop the pump vibrations / noisy operations?

(0.25)

QUESTION 1.07

(3.00)

With the reactor subcritical at a shutdown margin of 2.5% delta-k/k,

the stable count rate is 135 cps.

a. How much reactivity is required to be added to increase the stable

count rate to 405 cps? SHOW ALL WORK.

(2.25)

b. If the reactivity of part a is inserted in small, equal-reactivity

steps with the rods, will it take longer for the subcritical reactor

to reach a stable count rate near 150 cps or near 400 cps? WHY? (0.75)

QUESTION 1.08

(2.00)

With the plant operating at 100% power, a reactor trip and turbine trip

occur. Explain WHY steam generator levels change.

I

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

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

PAGE

5

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

.

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QUESTION 1.09

(3.00)

Concerning tegerature coefficients important to reactor control;

)< l Pea /*tc)

a. When Braidwood commences power operations (Cycle fl,BOL), the moderator

temperature coefficient (MTC) will be very small at the hot, zero power

n

condition. In view of this, explain WHY a large RCS dilution will be re-

quired for a power increas

o 100% power, even though Tavg will

increase during the power increase.

(1.50)

b. Explain why at end-of-life (EOL), the amount of RCS dilution required

for a power increaseyto 100% will be much more than at BOL.

(1.50)

0%

QUESTION 1.10

(2.00)

With the plant steady-state near 100% power at BOL on Cycle 1, Boron con-

centration at 1200 ppm and Tavg 587 F, a boration of 60 ppm is required to

compensate for a 600 pcm power defect due to a power decrease to 50% power.

Would the required boration be GREATER THAN, LESS THAN, or EQUAL TO 60 ppm

for the same 600 pcm gained by a power decrease if the following initial

condition change is made?(Only the listed condition changes; other initial

conditions stay the same.) EXPLAIN WHY FOR EACH ANSWER; CONSIDER EACH

CHANGE SEPARATELY.

a. Boron concentration is 900 ppm

(1.00)

b. Tavg is 572 F

(1.00)

QUESTION 1.11

(3.00)

With the plant stable at 80% power near BOL, which way will the Axial Flux

Difference initially change (MORE NEGATIVE or LESS NEGATIVE) for the

following occurrences? EXPLAIN YOUR ANSWER.

CONSIDER EACH OCCURRENCE SEPARATELY

a.

Xenon starts building in to the bottom of the core (1.00)

b.

OTdelta-T runback occurs with rods in automatic (1.00)

c.

A momentary large feed flow increase occurs with rods in manual (1.00)

(***** END OF CATEGORY 01 *****)

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2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

PAGE

6

.

,

.

[

f

N8

'

6

\\

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,

.

QUESTION 2.01

(1.00)

What are 2 purposes for the flow orifices in the Aux Feed Water supply

lines to the steam generators?

QUESTION 2.02

(2.00)

a. If the "0" Component Cooling Water (CCW) pump is initially in standby,

what CCW pump switch lineup condition /s are required for the "0" pump to

y

start automatically on an autostart signal? Q g gym) (0.5)

b. With the correct puI

eup, what 3 occurrences / signals will cause

the "0" Component Coofing Water pump to start automatically?

(1.5)

QUESTION 2.03

(2.00)

What VALVES in what PIPING LINES prevent main turbine overspeeds that may

occur when the generator has tripped and all main turbine throttle valves

and governor valves have shut?

QUESTION 2.04

(2.00)

Manual actuation of the Phase A Containment Isolation will close 4 sets of

valves associated with CVCS piping or components in the containment

building,

a. List the 4 sets of valves (noun name or function is sufficient)

(1.0)

b. Is long-term critical operation of the plant possible with Phase A Cont-

ainment Isolation manually actuated? EXPLAIN YOUR ANSWER.

(1.0)

QUESTION 2.05

(3.00)

List the 4 sources AND flow paths of boron solution used for Emergency

Boration. INCLUDE the approximate flow rate from each source.

1

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

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PLANT DESIGN INCLUDING SAFETY AND EERGENCY SYSTEMS

PAGE

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1

.

QUESTION 2.06

(1.50)

During a continuous RCS depressurization caused by a major LOCA at 1005

'

power, indicate the order (highest to lowest pressure) in which the

.

'

specific ECCS subsystems will inject into the RCS ANO show the approximate

pressure at which each subsystem will start to inject.

QUESTION 2.07

(2.00)

When a worker accidentally (cuts through the energized control power cable

to the 1A diesel generator DG) control panel, the resulting current surge

trips the 125vdc power supply breaker for Bus 111 and damages that breaker

such that it cannot be closed again. Neither the bus supply breaker nor

the 1A DG control power cable can be repaired or replaced for a week.

In the interim, until repairs can be completed, state the major steps

required to restore power to:

a. 125V Bus 111

(1.0)

b. IA DG Control Panel

(1.0)

QUESTION 2.08

(1.50)

a. Explain why operation ofthe main generator under reverse power cond-

itions is undesirable.

(0.75)

b. What are 3 reasons for the time delay on the main generator reverse

power trip?

(0.75)

QUESTION 2.09

(1.75)

a. What is the purpose of the #1 seal bypass ifnes on the RCPs?

(0.75)

b. What are the upper and lower pressures at which the bypass

valve (CV-8142) must be shut AND what is a reason for each pressure limit?

(1.0)

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

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2.

PLANT DESIGN INCLUDING SAFETY AND EERGENCY SYSTEMS

PAGE

8

,

.

QUESTION 2.10

(1.00)

a. List 2 CCW system indications in the Control Recm of a 1 gpm leak in

the in-use CVCS letdown heat exchanger.

(0.5)

b. Why is it likely that the leak is NOT in the non-operating ("not-in-use")

heat exchanger?

(0.25)

c. How can you confirm that the leak IS in the in-use heat exchanger?(0.25)

QUESTION 2.11

(1.50)

State 6 conditions required for the autoclosure of the diesel generator

feeder breaker ACB 1413 without synchronizing during an emergency.

(1.5)

QUESTION 2.12

(2.50)

a.1.ist 4 design features of the Spent Fuel Pool Cooling and Cleanup System

that prevent inadvertent draining of the Spent Fuel Pool below the tops

of the stored cells?

(2.0)

b. If all the NEW fuel cells for Unit 1 are temporarily stored in the

Spent Fuel Pool and the pool is inadvertently filled from the primary

water storage tanks, explain WHY an inadvertent criticality can/cannot

-

occur?

(0.5)

QUESTION 2.13

(1.25)

List 5 trips or automatic functions that occur when Breaker 41 (Main Gen-

erator Voltage Regulator Supply Breaker) trips.

l

r

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

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2.

TLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

PAGE

9

-

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,

'

QUESTION 2.14

(2.00)

Indic'te the water source (suction location) and SPECIFIC discharge

a

location (s) for the following 4 functions of the RHR system as part of the

Emergency Core Cooling System (ECCS).

a. Low head injection

(0.5)

b. Supply to other pumps

(0.5)

c. Cold leg recirculation

(0.5)

d. Hot leg recirculation

(0.5)

.

I

.l

l

(***** END OF CATEGORY 02 *****)

i

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

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3.

INSTRUMENTS AND CONTROLS

PAGE 10

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QUESTION 3.01

( .50)

Which one of the following conditions will actuate the Rod Control Urgent

Failure annunciator light?

a. A data error in one data cabinet

b. A logic error in one power cabinet

c. A failed oscillator in one power cabinet

d. A failed 25vdc power supply in one power cabinet

e. A failed 16.5vde power supply in the logic cabinet

QUESTION 3.02

(1.00)

During initial fuel load, the Shift Supervisor announces that an I&C tech

will be swapping the N42 and N43 power range Channel B drawers for trouble-

shooting. State 2 reasons why the SR0 should NOT permit the ISC tech to

do the power range swap.

QUESTION 3.03

(2.50)

Briefly .N. .R the interlock functions actuated whep the Alert Alarm

A

nc

cccurs on the following radiation monitors. (Nhaf

/

2.5)

a. ORE-AR055(FUEL BUILDING HANDLING INCIDENT)

b. ORE-PR009(CCW HEAT EXCHANGER "0" WATER OUTLET MONITOR)

c. ORE-PR031(CONTROL ROOM OUTSIDE AIR INTAKE A MONITOR)

d. 1RE-PR008(STEAM GENERATOR BLOWDOWN MONITOR)

o. 1RE-PR027(SJAE/ GLAND STEAM EXHAUST MONITOR)

l

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

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3.

INSTRUMENTS AND CONTROLS

PAGE 11

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QUESTION 3.04

(1.50)

a. List the 4 input signals to the S'Jbcooled Margin Monitor.

(1.0)

b. Under normal conditions at 100% power, what are the values of the

following? (Steam tables excerpts provided)

1) pressure margin to saturation

2)tenperature margin to saturation

(0.5)

.

QUESTION 3.05

(1.50)

On the steam dump control system, describe the difference between the

Loss of Load (or Load Rejection) controller and the Turbine Trip (or Plant

Trip) controller with regard to the following:

a. Arming requirements or signals

(0.5)

b. Input parameters or signals

(0.5)

c. Time delays or deadbands

(0.5)

QUESTION 3.06

(2.25)

During a plant power increase from 15% to 100% the steam pressure detector

for the selected steam flow channel on SG 1A sticks at the 15% position.

Explain the effect of the stuck steam pressure detector on the following:

a. IA SG inoicated steam flow versus actual steam flow

(0.75)

b. lA SG 1evel versus normal level at 100% power

(0.75)

c.1B main feed pump speed versus normal speed at 100% power

(0.75)

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

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3.

INSTRUMENTS AND CONTROLS

PAGE 12

-

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QUESTION 3.07

(2.25)

For the following actions / occurrences, state the:

1) Steam dump control system response,

2) The plant response, and

3) The resulting method of RCS temperature control

Assume all systems operate normally except as stated and that no operator

'

action is taken. CONSIDER EACH CASE SEPARATELY.

a.

With the plant in Hot Standby, steam pressure mode of control, waiting

for a reactor startup, the steam pressure setpoint is reduced by

200 psi.

(0.75)

b.

While in the steam pressure mode of control at 5% power, the train A

Steam Dump Interlock Bypass Selector Switch is placed in the

(0.75)

"off" position

~

c.

The train B reactor trip breaker fails to open on a reactor trip signal

while at 78% power. NOTE: The train A breaker opens properly. (0.75)

QUESTION 3.08

( .75)

Indicate whether the following statements concerning operation of the

reactor trip (RT) and bypass (BY) breakers are TRUE or FALSE.

a.

If RPS train A is placed in test while bypass breaker BYB is closed,

both reactor trip breakers and both bypass breakers will trip.

(0.25)

b.

If an attempt is made to close both bypass breakers at the same time,

both bypass breakers will trip, but the reactor trip breakers will

remain closed.

(0.25)

c.

A solid state protection system (SSPS) train A reactor trip signal will

'

trip all RT and BY breakers.

(0.25)

1

,

)

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

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3.

INSTRUENTS AND CONTROLS

PAGE 13

.

..

.

QUESTION 3.09

(1.25)

Match the following reactor protection and control signals in Column A

~

to their associated logic coincidence in Column B.

(1.25)

COLUMN A

COLUMN B

a.

RCS loop loss of flow trip (per loop)

1.

1/2

2.

2/2

b.

P-6(Source range turn-on on power decrease)

3.

1/3

4.

2/3

c.

P-10 (Nuclear at power)

5.

1/4

-

6.

2/4

d.

Pzr high pressure trip (pressure increase)

7.

3/4

e.

Power range high power rod stop (power increase)

QUESTION 3.10

(1.00)

What are 2 conditions (including setpoints if applicable) under which the

pressurizer level control system actuates the pressurizer heater controls?

Include WHY.

QUESTION 3.11

(1.50)

With the plant at 80% power and all systems in automatic control, state

the initial direction of rod motion for the following conditions or

occurrences. EXPLAIN your answer. Consider each condition separately.

a.

Loop 4 Tcold fails high.

(0.5)

b.

ure transmitter PT-505 fails high

(0.5)

c.

A spurious main turbine runback occurs

(0.5)

QUESTION 3.12

(1.00)

Why is a variable-gain circuit included in the rod control system power

cismatch circuitry?

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

_

^

3.

INSTRUMENTS AND CONTROLS

PAGE 14

. .

.

'

QUESTION 3.13

(1.00)

List the conditions or signals that will cause automatic closure of the

Main Feedwater Isolation Valves (1FW-009A,B,C D).

QUESTION 3.14

(2.00)

a. List all parameters or concitions WITH setpoints (as applicable) that

will prevent automatic AND manual rod withdrawal.

(1.4)

b. List all parameters or conditions WITH setpoints (as applicable) that

will prevent rod withdrawal in automatic ONLY.

(0.6)

QUESTION 3.15

(1.00)

Identify the control, protective, and permissive functions which use

individual loop Tavg signals and NOT auctioneered Tavg.

QUESTION 3.16

(1.50)

List ALL automatic start signals for the motor-driven aux feed pump 1A.

INCLUDE coincidence logic if applicable.

'

QUESTION 3.17

(1.75)

a. With the plant at 100% power and the Pressure Channel Selector Switch

set on Channel 1(IPT-455), list 3 insnediate pressurizer system responses

if pressure transmitter 1PT-455 fails HIGH.

(0.75)

b. If N0 operator action is taken for the IPT-455 failure, list 4 resulting

system actions.

(1.00)

QUESTION 3.18

( .75)

List the 3 separate functions of overspeed protection provided by the

everspeed protection controller in the Main Turbine Control and Protection

System.

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

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4.

PROCEDURES - NORMAL,. ABNORMAL, EMERGENCY AND

PAGE 15

.- , RADIOLOGICAL CONTROL

,,

.

QUESTION 4.01

(1.50)

Per the Technical Specifications, if the plant is at 2% reactor power,

and Tavg drops to 547F due to excessive warmup of the steam lines;

1) What action must be acconplished?

(0.75)

2) If action 1 cannot be accomplished, what other action must be taken?

(0.75)

(INCLUDE time limits in 1 and 2 if applicable)

1

QUESTION 4.02

(2.50)

A 24 year old male had a lifetime exposure of 23 REM through last quarter.

In addition to his lifetime exposure, he has also received I rem in

-

the present quarter.

a. In accordance with 10CFR20, what 3 provisions must be met to allow

this individual, in a non-emergency situation, to exceed the quarterly

regulatory limit?

(1.50)

b. Whose approval is needed to allow this individual to exceed the

quarterly regulatory limit?

(0.5)

c. How long may this individual work in a 200 mrem /hr radiation field

before he reaches the maximum quarterly limit allowed at Braidwood?

(0.5)

QUESTION 4.03

(1.50)

Per the Braidwood Tech Specs, if the plant is being maintained in hot

standby (Mode 3):

a.

What is the shutdown margin Ifmit?

(0.25)

l

b.

How often must the shutdown margin be checked?

(0.25)

c.

If the shutdown margin limit is not met, what required actions must be

taken? INCLUDE numbers (if applicable).

(1.0)

(***** CATEGORY 04 CONTINUED GN NEXT PAGE *****)

4.

PROCEDURES - NORMAL, ABNORMAL, EERGENCY AND

PAGE 16

,RADIDLOGICAL CONTROL

,

,,

.

QUESTION 4.04

(1.00)

List four (4) entry conditions from the Emergency Procedures which require

transitioning to IBwEP-3, Steam Generator Tube Rupture?

l

QUESTION 4.05

(1.50)

18wEP ES-1.3, Transfer to Cold Leg Recirculation, contains the following

Caution: SI PUMPS MUST BE STOPPED IF RCS PRESSURE IS GREATER THAN THEIR

SHUT 0FF HEAD PRESSURE. What is the reason for this caution and why is it

included only in the Transfer to Cold Leg Recirculation procedure?

QUESTION 4.06

(1.00)

List the four " Design Basis Accidents" which the ECCS systems are designed

to mitigate.

QUESTION 4.07

(1.50)

List all Critical Safety Functions from highest to lowest priority.

-

QUESTION 4.08

(1.50)

List the 3 methods provided to cool and depressurize a Steam Generator with

a tube rupture per 18wEP-3, Steam Generator Tube Rupture.

QUESTION 4.09

(1.75)

Bw0A PRI-9(Loss of Shutdown Cooling) instructs the operator to establish

alternate decay heat removal if BOTH RHR trains should fail and NOT be

capable of restoration to operability,

a.

List two (2) alternate decay heat removal methods provided in the

procedure for the following plant conditions: RCS temperature 300 F.

RCS pressure 300 psig, all systems other than RHR operable.

(1.00)

b.

List three (3) alternate decay heat removal methods provided for if the

vessel head is removed.

(0.75)

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

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4.

PROCEDURES - NORMAL, ABNORMAL, EERGENCY AND

PAGE 17

, RADIOLOGICAL CONTkOL

,

QUESTION 4.10

(2.25)

With the plant at 85% power and all systems in automatic, and N0 operator

actions in progress, the following symptoms are suddenly observed in

relatively quick succession:

,

a. Flow indicated on 1FY-0111, PRIMARY WATER CONTROL PREDET COUNTER,

,

b. Control rods inserting,

'

c. Actuation of annunciator 1-10-B6 ROD BANK LOW INSERTION LIMIT, and

d. Decreasing Boron Concentration Meter reading.

Shif ting the Makeup Control Switch to STOP does NOT change these symptoms.

What three (3) Main Control Board switch actions or operations should be

taken for these symptoms and WHY?

QUESTION 4.11

( .50)

For a dropped control rod with no reactor trip, the initial method of

removing / responding to the Tave/ Tref mismatch is which of the following?

a.

Controlling main turbine load.

b.

Taking manual control of individual control rod banks,

c.

Shifting to low load operation on the main turbine

d.

Boration and/or dilution of the reactor coolant system.

QUESTION 4.12

( .50)

Which of the following radiation exposures will inflict the greatest

biological damage?

a.

1 Rem of ALPHA

b.

1 Rad of NEUTRON

c.

1 Roentgen of BETA

d.

1 Rad of GAMMA

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

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PROCEDURES --NORMAL, ABNORMAL, EERGENCY AND

PAGE 18

,, , RADIOLOGICAL CONTROL

,,

.

QUESTION 4.13

(1.00)

Complete the following statements using info from the Precautions and

Limitations sections of the Plant Heatup procedure (BwGP 100-1).

a.

For normal heatup of the pressurizer, a rate of

F/ hour

i

will not be exceeded.

(0.25)

i

b.

Spray flow into the pressurizer will NOT be initiated if the

temperature difference between the Pzr and Spray fluid exceeds

i

F.

(0.25)

l

c.

Administratively, RCS oxygen must be in specification prior to

exceeding

F.

(0.25)

'

d.

Heatup must be terminated or spray initiated if pressurizer

boron concentration approaches

ppm less than RCS loop (0.25)

concentration.

QUESTION 4.14

(2.00)

According to function restoration procedure IBwFR-S.1 " Response to Nuclear

4

Power Generation /ATWS" Step 2, the operator is to verify turbine trip. If

the turbine has not tripped, he should manually trip it. List the sequence

cf actions he should take if the turbine does not respond to his actions.

QUESTION 4.15

(1.50)

l

According to the BW Precautions, Limitations, and Setpoints book, all

reactor trip and safeguard actuation channels, except 3, shall be placed

in the trip mode when the channel is out of service for any reason. List

the 3 excepted circuits / trips that may be bypassed for maintenance. (1.5)

l

i

l

QUESTION 4.16

(3.50)

a. List 2 criteria (WITH limits if applicable) that require emergency

boration during refueling operations.

(1.0)

!

b. List 5 criteria (WITH limits if applicable) that require emergency

boration during operations at 98% power.

(2.5)

,

l

4

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

l

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

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.

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.o

.

EQUATION SHEET

f = ma

v = s/t

Cycle efficiency = (:tet work

'

out)/(Energy in)

2

w = ag

s = V,t + 1/2 at

2

E = mc

2

-at

KE = 1/2 av

a = (Vf - V,)/t

A = AN

A = A,e

PE = agn

Vf = V, + at

w = e/t

i = an2/t1/2 = 0.693/t1/2

1/2*ff = C(tui)(t)3

2

d

w

, ,p.

,n

A=

[(t1/2)*III)

4

b

E " 93 I "

m = V,yAo

-b

y,

.

Q = mCoat

Q = UA A T

I " I ',

o

/TVL

Pwr = W ah

I*I

10

f

o

TV1. = 1.3/u

sur(t)

HVI. = -0.693/u

P = P 10

P = P e*I

o

SUR = 26.06/T

SCR = S/(1 - K,ff)

CR, = S/(1 - K,ffx)

CR (1 - X,ffj) = G ( -Eeff2)

SUR = 26a/t= + (a - p)T

j

2

T = (t*/o) + ((s - oV Io]

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

'

j

T = 1/(o - s)

M = (1 - K ,ff,)/(1 - K,ffj)

T = (a - o)/(Io)

SDM = ( - K ,ff)/K,ff

a = (K,ff-1) A,ff = K,f/K,ff

t* = 10

seconds

I = 0.1 seconds-I

o = [(t*/(T K,ff)] + [i ff (1 + IT)]

/

Idjj=Id

2 =2 2

P = (I4V)/(3 x 1010)

Id

l

Id

jj

22

2

I = oN

R/hr = (0.5 CE)/d (meters)

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

Water Parameters

Miscellaneous Conversions

I gal. = 8.345 lbm.

1 curie = 3.7 x 1010dps

1 ga]. = 3.78 liters

} kg = 2.21 lbm

I ftJ = 7.48 gal.

I hp = 2.54 x 10 Stu/hr

Oensity = 62.4 lbm/ft3

1 mw = 3.41 x 10 Stu/hr

,

3

lin = 2.54 cm

Density = 1 gm/cm

Heat of vaccrization = 370 Stu/lem

F = 9/5'C + 32

. test of fusion = 144 Stu/ltm

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

1 Atm = 14.7 psi = 29.9 in. Hg.

1 BTU = 778 ft-lbf

I ft. H O = 0.4335 lbf/in.

2

_ _

_ -

..

_-

.

. . - - -

. - - - .

- . - . -

- _

-

.-___ --

_ __, _

_

_ _

.._

_

.

<

.

-

,

i

i

-

l

Table 1.

Saturated Steam: Temperature Table

i

Abs Press.

Specific Volume

Enthalpy

Entropy

Temp

Lb per

Sat.

Sal.

Temp

j

fahr

SqIn.

Liquid

Evap

vapor

Liquid

Evap

Vapor

Liquid

Evap

Vapor

Fahr

,

i

t

p

vi

veg

vg

he

h fe

h

se

sig

s

t

g

1

32 0

0 08859

0.016022

33043

0.0179

1075 5

0 0000 2.1873 2 1873

32 8

I

34 0

0 09600

0 016021

3061.9

1.996

1074.4

1076.4

0 0041

2.1762 2.1802

34 8

l

36 8

0 10395

0 016020

2839.0

2839 0

4.008

1073.2

1077.2

0 0081

2.1651

2.1732

35 0

i

38 0

0.I1249

0.016019

26341

2634.2

6.018

1072.1

10781

0 0122

2.1541 2.1663

3e s

1

'

i

et t

4.12163

0 016019

2445.8

8.027

1071.0

1079 0

0 0162

2.1432 2 1594

48.9

i

42 8

0.13143

0 016019

2272.4

10 035

1069 8

1079.9

0 0202

2.1325 2.1527

42 0

44 8

0 14192

0 016019

2112 8

2112.8

12.041

1068 7

10803

0 0242 2 1217 2.1459

44 e

4

j

46 I

O15314

0 016020

1%57

1%5.7

14 047

1067.6

1081 6

0 0282

2.1111

2.1393

46 8

40 0

0.16514

0.016021

1830 0

16 051

1066 4

1082.5

0 0321

2.1006 21327

48 0

>

.

1

!

50 8

0.17796

0.016023

1704 8

18 054

1065 3

1083.4

0.0361

2.0901

2.1262

30.0

52 0

0.19165

0 016024

1589.2

1589 2

20 057

1064 2 .1084 2

0 0400 2 0798 2.1197

52 8

54 0

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215 220

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2.1126

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392.s

225 516

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270 600

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415 e

423 3

308.780

0 01894

1.4808

1.4997

396 9

806 2

1203.1

0 5915 0 9165

1.5080

428 0

424.3

322 391

0 01900

1.4184

1.4374

4013

802 2

1203 5

0 5964

0 9077

I5042

424I

'

l

428.9

336 463

0 01906

1.3591

1.3782

405 7

7980

1203.7

0 6014

0 8990 15004

428 s

l

432 e

351 00

0 01913 1.30266

1.32179

410 1

793 9

1204 0

0 6063

0 8903

1.4966

432 9

435 8

366 03

001919 1.24887

1.26806

414 6

789 7

1204 2

0 6112

0 8816 1.4928

43E 8

440 e

381.54

0 01926 119761

1.21687

419 0

785 4

1204 4

0 6161

0 8729

I4890

443 e

444 0

397 56

0 01933 114874

116806

423 5

781 1

1204 6

0 6210 0 8643

14853

444s

44f.a

414 09

0 01940 110212

1.12152

4 29.0

776 7

12041

0 6259

0 8557

14215

4470

44#

43]I4

0 01947 1 05764

1 07711

pt5

772 3

1204 8

0 6308

0 8471

1 47/8

4g.O

456 s

44813

001954 10151R

103472

437 0

767 8

1204 8

06356

0 8385

14741

45 6

i

.

Abs Press

Specific Volume

Enthalpy

Entropy

Temp

tb per

Sal.

Sal

Temp

Fahr

SqIn.

Liquid

Evap

Vapor

liquid

Evap

Vapor

Liquid

Evap

Vapor

fahr

vr

hg

h ig

h

s,

59,

5,

t

p

vi

ver

g

4ES I

466.87

0.01 % I

0.97463

0 99424

441.5

763.2

1204 8

0 6405 0 8299

1.4704

468.0

464 8

485 56

001%9 0 93588

0 95557

446.1

758 6

12043

0 6454

0 8213

1.4667

464.0

468 e

504 83

0 01976 0.89885

0 91862

4503

754 0

1204 6

0 6502 0 8127 1.4629

468.0

472 8

524 67

0 01984 0 86345

0 88329

455 2

749.3

1204 5

0.6551

0 8042

1.4592

472.I

47E g

545.11

0.01992

0 82958

084950

459.9

744.5

1204.3

0.6599 03956 1.4555

475.8

48s O

566 15

0 02000 0 79716

0 81717

464.5

739.6

1204.1

0.6648 03871

14518

esee

484 8

587 81

0 02009

036613

0 78622

4691

7343

1203.8

0 66 % 01785 1.4481

4s4.s

4sg a

610.10

0.02017

033641

035658

473 8

729 7

1203.5

0.6745 0.7700 1.4444

488.0

437 3

633.03

0 02026 030794

032820

4785

724 6

1203.1

0 6793

03614

1.4407

492.8

496 I

656 61

0 02034 0 68065

0 70100

483 2

719 5

12023

0 6842 03528 1.4370

495 e

,

'

!

500 5

680 86

0 02043 0 65448

0 67492

487.9

714.3

1202.2

0 6890 0 7443

1.4333

500.0

504 8

705 78

0 02053 0.62938

064991

492.7

709 0

1201.7

06939

03357

1.42 %

504.8

508 0

731 40

0 02062 0 60530

0 62592

497.5

7033

1201.1

0 6987

03271

1.4258

500.8

I

512 8

757 72

0 02072 0 58218

060289

502 3

698 2

12005

0 7036

03185 1.4221

517 8

SIE e

78436

0 02081

0.55997

0.58079

507.1

6923

Il99L8

0 7085 03099 1.4183

516 8

570 8

812 53

0 02091

0 53864

0 55956

512 0

687.0

1199 0

0.7133

0 7013

1.4146

529.8

5745

84104

0 02102 0 51814

0 53916

516 9

681.3

1198.2

03182

0 6926 1.4108

524.8

578 0

870 31

0 02112 0 49843

051955

521 8

675 5

1197.3

0 7231

0.6839 1.4070

528 3

537 8

900 34

0 02123 0 47947

0 50070

526 8

669 6

11 % 4

03280

0.6752 1.4032

532.0

535 0

931.17

0 02134

0 46123

048257

531 7

663.6

1195 4

0 7329

0.6665 1.3993

536.8

548 8

962 79

0 02146 0 44367

0 46513

5368

657.5

1194.3

0 7378

0 6577 1.3954

See e

544 8

995 22

0 02157 0 42677

0 44834

541 8

651.3

1193 1

0 7427

0 6489 1.3915

5448

5480

1028 49

0 02169 0 41048

0 43217

546 9

645.0

1191.9

03476

0 6400 1.3876

548.0

5578

1062 59

0 02182 039479

0 41660

552 0

638.5

1190 6

0 7525

0 6311

1.3837

5520

555 8

1097.55

0 02194 0 37966

0 40160

5573

632.0

1189.2

03575

0 6222 1.3797

556.8

568 8

1133 38

0 02207 0.36507

0 38714

562.4

625 3

11873

0.7625

0 6132 1.3757

560.8

54 0

1170 10

0 02221

0.35099

0 37320

567.6

618.5

11861

0 7674

0 6041

1.3716

564e

568 0

1207 72

0 02235 0 33741

0 35975

572 9

611.5

1184 5

0 7725

05950 1.3675

568 e

577 6

1246 26

0 02249 0 32429

0 34678

578 3

604 5

1182 7

03775

0 5859 13634

5720

515 0

1285 74

0 02264 0 31162

033426

5837

597.2

1180.9

03825

0.5766 1.3592

575.s

a

See 8

1326 17

0 02279 0 29937

0 32216

589I

589.9

1179 0

0 7876

0 5673 1.3550

500.0

584 0

13677

0 02295 0 28753

0 31048

594 6

582 4

1176 9

0 7927

0.5580 1.3507

544.5

5st O

1410 0

0 02311 0 27608

0 29919

600 1

5743

1174 8

07978

0 5485 1.3464

588 3

5970

1453 3

0 02328 0 26499

0 28827

605 7

566 8

1172 6

0 8030

0.5390 1.3420

597I

596 0

1407 8

0 02345 0 25425

0 27770

611 4

558 8

1110 2

0 8082

0 5293 1.3375

59E 8

-

.

!

Abs Press.

Specific Volume

Enthalpy

Entropy

Temp

tb per

Sal.

Sat.

Sal.

Sat.

Sal

ten

Fahr

SqIn.

Liquid

Evap

Vapor

Liquid

Evap

Vapor

liquid

Evap

Vapor

Fa

I

p

vg

vtg

vg

hg

h gg

h

s,

5 ,,

s

I

g

t

i

sees

1543 2

0 02364 0.24384

036747

617.1

550.6

1167.7

0.8134

0.5196 1.3330

000.0

see e

1589 7

0 02382 0.23374

0 25757

622.9

542.2

1165.1

0.8187

0.5097 1.3284

804.0

seg 3

16313

0 02402 0 22394

0 247 %

628 8

533 6

1162.4

0.8240

0.4997 13238

530.0

-

s12 8

16861

0 02422 0.21442

0 23865

634 8

5243

1159 5

0.8294

0.48 % 1.3190

512.0

l

316.5

1735 9

0.02444 0.20516

0.22960

640.8

515.6

1156.4

0.8348

0.4794 1.3141

615.8

t

$28 8

1786.9

0.02466 0.1%I5

022081

646.9

506.3

1153 2

0.8403

0.4689 1.3092

820.0

524 8

1839 0

0 02489 0.18737

021226

653.1

406.6

1149 8

OE'5!

0.4583 1.3041

524.0

$23 8

1892 4

0 02514 0.17880

0 20394

659 5

4863

1146.1

0.8514

0.4474 1.2988

820 0

l

332.s

19470

0 02539 0.17044

0 19583

665.9

476.4

1142.2

0.8571

0.4364 1.2934

532.0

l

636.3

2002 8

0 02566 0.16226

0.18792

672.4

4653

1138.1

0.8628

0.4251

1.2879

636.0

548 8

2059 9

0 02595 0.15427

0 18021

679I

454.6 11333

0.8686

04134 1.2821

sese

544 8

2118 3

0 02625 0.14644

0.17269

685 9

443.1

1129 0

0.8746

0.4015 1.2761

544.8

648 e

21781

0.02657 0.13876

0.16534

692.9

431.1

1124.0

0.880E

03893 1.2699

648.0

552 3

2239 2

0.02691

0.13124

0.15816

700 0

4183

11183

0.8868

03767 1.2634

652 e

$56 I

23013

0.02728 0.12387

0.15115

707.4

4053

1113.1

0.8931

03637 1.2567

456.0

568 I

23653

0.02768 0.11663

0 14431

714 9

392.1

1107.0

0.8995

03502 1.2498

808.8

,

'

564 8

2431.1

0 02811

0.10947

0.13757

722 9

377.7

1100.6

0.9064

03361 1.2425

554.s

568 I

24981

0.02858 0.10229

0.13087

731.5

362.1

1093.5

0.9137

03210 1.2347

ses.g

572 0

256E 6

0.02911 0.09514

0 12424

740 2

3453

1085.9

0.9212

03054 1.2266

572.8

376 8

26368

0.02970 0 08799

0.11769

749.2

328.5

1077.6-

0.9287

0.2892 1.2179

876.0

$80.0

2708 6

0.03037 0.00080

0.11117

758 5

310.1

1068.5

0 9365

02720 1.2086

8e0.0

See B

2782.1

0.03114 0.07349

0.10463

768 2

290.2

1058.4

0.9447

0.2537 1.1984

604.0

588.I

2857.4

0 03204 0.065 %

0.09799

778.8

268 2

1047.0

0.% 35

0.2337 1.1872

les.8

592 8

2934.5

0 03313 0 05797

0 09110

790 5

243.1

1033.6

0.9634

0.2110 1.1744

892.0

896.I

3013.4

0.03455 0.04916

0.08371

804.4

212.8

1017.2

0.9749

0.1841 1.1591

895.8

700 0

30943

0.03662 0.03857

0.07519

822.4

1723

995.2

0.9901

0.1490 1.1390

700.0

782 s

3135.5

0.03824 0 03173

0 06997

835 0

1443

9793

1.0006

0.1246 1.1252

782 0

.

TM s

31772

0 04108 0 02192

0 06300

854 2

102.0

956 2

1.0169

0 0376 1.1046

704.0

~

705 0

31983

0 04427 0.01304

0.05730

873 0

61.4

934.4

1.0329

0 0527 1.0856

785 8

785.47*

32082

0.05078 0 00000

0.05078

906 0

00

906.0

1.0612

0 0000 1.0612

7e5.47*

I

-

. - _ _ _ _ _ _ _ _ _ - _ .

--

_

- . . .

..

-_

. _ - - - -

-

.

-

..

O

!

.

l

.

Table 2: Saturated Steam: Pressure Table

.

,

}

Specific Volume

Enthalpy

Entropy

Abs Press.

Temp

Sat.

Abs Press.

i

Lb/Sq In.

Fahr

liquid

Evap

Vapor

Liquid

Evap

Vapor

Liquid

Evap

Vapor

LblSq In.

p

t

vg

v

hg

hgg

h

sg

s ig

s

p

eg

g

i

,

l

g.g3065

32.018

0.016022

3302 4

3302.4

0 0003

1075.5

1075 5

0 0000

2 1872

e39355

i

0 25

59 323

0 016032

1235 5

27.382

1060.1

1087.4

0 0542

2.0425

2.0967

5 25

i

e 5e

79 586

0 016071

64I 5

641.5

47.623

1048 6

1096.3

0 0925

1.9446

2.0370

e 53

l

1e

10134

0016136

333 59

333 60

6913

10361

1105 8

01326

1.8455

I9781

Ie

1

50

162.24

0.016407

73.515

13 532

130 20

1000 9

1131.1

0 2349

1.6094

I8443

5g

l

18 I

193 21

0 016592

38 404

38.420

161.26

9821

1143 3

0 2836

1.5043

17879

le O

'

14 898

212.00

0.016719

26382

26399

18017

970 3

1150.5

0 3121

1.4447

17568

14 596

15.0

213.03

0 016726

26274

26.290

181.21

%97

1150.9

0.3137

1.4415

13552

15 8

2ee

227.96

0016834

20 070

20 087

196.27

9601

1I56.3

0 3358

1.3%2

13320

23 0

30.0

250 34

0.017009

13 7266

133436

218 9

945 2

1164.1

0.3682

1.3313

16995

30 e

i

oga

267.25

0017151

10.4794

10.4965

236.1

933 6

1169 8

0 3921

1.2844

16765

4e e

i

50 g

281 02

0 017274

84%7

8.5140

250 2

923 9

1174.1

0 4112

12474

16586

See

!

Ise

29231

0.017383

7.1562

7.1736

262.2

915 4

1177.6

0 4273

1.2167

1.6440

Is e

I

is e

302.93

0 017482

6.1875

6.2050

2723

907 8

1180 6

0 4411

1.1905

16316

7e e

i

33.8

312 04

0 017573

5.4536

5.4711

282.1

900.9

1183.1

0 4534

1.1675

16208

30 g

33.0

320 28

0 017659

4.8779

4.8953

2903

894.6

1185.3

0.4643

1.1470

1.6113

Se e

lesg

327.82

0 017740

4 4133

4.4310

298.5

888 6

1187.2

04743

1.1284

1.6027

lese

!

lig 8

334.79

001782

4 0306

4.0484

305 8

883.1

1188.9

0 4834

1.1115

1.5950

110 e

'

12g a

341.27

0.01789

33097

33275

312.6

877.8

1190.4

0 4919

1.0960

1.5879

120 g

l

130.0

347.33

0.017 %

3.4364

3.4544

319 0

872.8

1191.7

0.4998

1.0815

1.5813

13eI

14g.g

353 04

0.01803

3 2010

3.2190

325 0

8680

1193.0

0 5071

10681

1.5752

les O

'

150.0

358 43

0.01809

2.9958

3.0139

330 6

863.4

1194.1

0 5141

10554

1.% 95

15ee

its e

363 55

0 01815

2.8155

2 8336

336.1

859.0

1l95 1

0 5206

10435

1.5641

Ils e

17e g

368.42

0 01821

2 6556

2.6738

341 2

854 8

II96.0

0 5269

I0322

1.5591

17e s

133.0

373 08

0 01827

2.5129

2.5312

346.2

8503

11 %.9

0 5328

1.0215

I.5543

les e

130.0

377.53

0.01833

2.3847

2.4030

350.9

8463

1197.6

0 5384

1.0113

1.5498

19eI

!

2gg g

381.80

0 01839

2.2689

2.2873

355 5

042.8

1198.3

0.5438

1.0016

I.5454

230 e

l

210.0

385.91

0.01844

2.16373

2.18217

359.9

839.1

1199 0

0 5490

0 9923

1.5413

21g g

i

220.0

389.88

001850

2 06779

2.08629

364 2

835 4

1I99.6

0 5540

0 9834

1.5374

225 0

'

233 0

39330

0 01855

I97991

1.99846

368 3

831 8

12001

0 5588

0 9748

15336

238 g

l

240 e

397.39

0 01860

1.89909

I91769

372 3

828 4

1200 6

05634

0 9665

I5299

248 8

a

250.0

400 97

001865

1.82452

1.84317

376.1

825 0

1201.1

0 5679

0 9585

1.5264

250 0

!

260 0

404.44

C01870

1 75548

137418

379.9

821 6

1201.5

0 5722

0 9508

1.5230

260 0

l

270.g

407.80

0 01875

I69137

131013

383 6

818.3

1201.9

05764

0 9433

15197

275 0

4

238 I

411.07

0 01880

163169

1.65049

387.1

815.1

1202.3

05805

0.9361

15166

280 e

'

298.9

414.25

0 01885

1 57597

1.59482

390 6

812.0

1202.6

0.5844

0 9291

1.5135

290 0

333 e

417.35

0 01889

I 52384

1.54274

394 0

808 9

1202.9

05882

0 9223

I5105

300 e

350 0

43173

0 01912

130642

1.32554

409 8

794 2

1204 0

0 6059

0 8909

I4%8

358 8

488 0

444 60

0 01934

1.14162

1.16095

424 2

780 4

1204 6

0 6217

0 8630

14847

400 0

,

i

t

.

- - .

.

-

_ - -

. - .

. -

_ . - - - - - - . -

!

-

i

!

%

!

Specific Volume

Enthalpy

Entropy

,

o

l

Abs Press.

Temp

Sat.

Abs Press.

i

Lb/Sg in.

Fahr

liquid

Evap

Vapor

Liquid

Evap

Vapor

Liquid

Evap

Vapor

tblSq In. .

i

g

p

l

p

t

vi

vig

v

h l

h

g

ig

g

sg

s ,g

s

.

'

458.8

456 28

0 01954

101224

1.03179

437 3

767.5

1204.8

06360

0.8378

14738

458 8

580.0

46701

001975

0 90787

0.92762

449 5

755.1

12043

06490

0 8148

1.4639

500 0

l

t

550 0

476 94

0 01994

0 82183

0 84177

460.9

743.3

1204.3

0 6611

03936

1.4547

550 0

l

ses8

486 20

0 02013

074%2

0 76975

4713

732 0

12033

0 6723

03738

14461

See 0

,

j

650 8

494 89

0 02032

0 68811

0.70843

481.9

720.9

1202.8

06828

07552

14381

658 I

i

700.0

503 08

0 02050

0 63505

0.65556

491.6

710.2

1201.8

0 6928

03377

1.4304

fee t

750 0

510 84

0 02069

0 58880

0 60949

500 9

699 8

1200 7

0 7022

0 7210

14232

758 8

j

BIO O

518 21

0 02087

054809

0.568 %

509 8

689 6

1199.4

03111

0 7051

I4163

000 0

850.0

525 24

0.02105

0 51197

0 53302

518 4

679.5

1198 0

03197

0 6899

1.4096

350 0

'

900 8

531.95

0 02123

0 47968

0 50091

5267

669 7

1196 4

0 7279

06753

14032

93g e

958 8

538.39

0 02141

0 45064

0.47205

534 7

660 0

1194 7

0 7358

0 6612

1.3970

95s g

1000 0

544.58

0 02159

0 42436

0.44596

542 6

650.4

1192 9

0 7434

0 6476

1.3910

lege s

1950 0

550.53

0.02177

0.40047

0 42224

550.1

640.9

1191.0

01507

06344

1.3851

lessI

'

l

110s 8

556 28

0 02195

0 37863

0 40058

557.5

631.5

1189.1

03578

06216

13794

Ilse e

i

1150.8

561.82

0 02214

0.35859

0.38073

564 8

622 2

1187.0

03647

06091

13738

115eI

1200.0

567.19

0 02232

0.34013

0.36245

571.9

613 0

1184.8

0.7714

05%9

1.3683

12000

'

j

1250 9

572.38

0 02250

0 32306

0.34556

578 8

603 8

1182.6

01780

0 5850

1.3630

1258 8

j

1388 I

577.42

0.02269

0 30722

0 32991

585 6

594 6

1180 2

0 7843

0 5133

1.3577

13ss e

13500

582.32

0 02288

0.29250

0 31537

592.3

585 4

1177.8

03906

0.5620

1.3525

135e e

i

lose e

587.07

0 02307

0 27871

0.30178

598 8

576 5

1175.3

03966

0 5507

1.3474

lega 5

1458 8

591.70

0.02327

0 26584

0 28911

605 3

567 4

1172.8

0 8026

0 5397

1 3423

1450 0

150s 0

596 20

0 02346

0 25372

0 27719

611 3

558 4

1170.1

0 8085

0 5288

1.3373

15eg g

15580

600 59

0 02366

0 24235

0 26601

618 0

549 4

1167.4

0.8142

0.5182

1.3324

155II

ISOS S

604.87

0 02387

0 23159

0 25545

624.2

540.3

1164.5

0.8199

05076

1.3274

163g g

l

1658 8

609 05

0 02407

0 22143

0.24551

630.4

531.3

1161 6

08254

0 4971

1.3225

1650 0

1

1700 0

613.13

0.02428

021178

0.23607

636.5

522.2

1158 6

0.8309

0.4867

I.3176

Ilse g

l

1750.8

617.12

0.02450

0 20263

0.22713

642.5

513.1

1155.6

0 8363

0.4765

1.3128

17500

4

Igge 3

621 02

0 02472

0 19390

011861

648 5

503 8

1152.3

0 8417

0.4662

1.3079

Iges e

i

185g e

624.83

0 02495

018555

0.21052

654 5

494 6

1149 0

0.8470

0 4561

1.3030

is5e e

i

1900e

628.56

0 02517

0.17761

0.20278

660 4

485 2

1145 6

0 8522

0 4459

1.2981

19ee e

'

19500

632.22

0 02541

0.16999

0.19540

666 3

475.8

1142.0

0.8574

0.4358

13931

19500

!

2000.0

635 80

0.02565

0 16266

0 18831

612.1

4663

1138.3

0.8625

0 4256

12881

2000 5

1

2100.3

64236

0.02615

0.14885

0 17501

683.8

4463

14 30.5

0 8727

04053

1.2780

21ee e

.

2200.3

649.45

0 02669

0.13603

0.16272

695 5

4263

1122.2

0.8828

0.3848

1.2676

2200.8

l

2300.0

655.89

0.02727

0.12406

015133

707.2

406 0

1113.2

0.8929

0.3640

12569

2388 5

!

2400.0

662.11

0.02790

0.11287

0.14076

719 0

384.8

1103.7

0.9031

0.3430

1.2460

2480 8

l

i

2500 e

668.11

0.02859

0 10209

0.13068

731 7

361 6

1093 3

0 9139

0 3206

1.2345

7500 e

l

260s e

673 91

0 02938

0.09172

0.12110

744 5

337 6

1082.0

0 9247

0 2917

12225

2500 0

2700 0

679 53

0 03029

0 08165

0.11194

757 3

312 3

1069 7

0 93 %

02741

12097

2100 0

2000 e

684 96

0 03134

0 07171

0.10305

770 7

2851

1055 8

0 9468

02431

1.1958

20050

2900 8

690 22

0.03262

0 06158

0 09420

7851

2543

1039 8

0 9588

02215

1.1803

2988 8

30000

695 33

0 03428

0 05073.

0 08500

801 8

218 4

1020 3

0 9728

01891

1.1619

3seg g

31st8

700 28

0 03681

0 03771

0 07452

824 0

169.3

993 3

0 9914

0 1460

1.1373

3100 I

i

370sI

705 08

0 04472

0 01191

0 05663

875 5

56 1

931.6

1.0351

0 0482

1.0832

320s e

3200.2*

70547

0 05078

0 00000

0 05078

9E4

00

906 0

1 0612

0 0000

1 0612

1788 7-

Critical pressure

!

-

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

PAGE 19

fHERM0 DYNAMICS,HEATTRANSFERANDFLUIDFLOW

B.

ANSWERS -- BRAIOWOOD 132

-86/10/22-WEALE, G./REIDINGER

MASTEri COPY

'

ANSWER

1.01

(2.00)

a. Increase (0.25) - increase in shutdown margin due to higher rod position

while available power defect reactivity remains constant (0.75)

(1.00)

b. No change (0.25) - decrease in shutdown margin due to more available

power defect matched by increase in shutdown margin due higher rod pos-

ition.(0.75)

(1.00)

REFERENCE

REACTOR THEORY REVIEW I-5.31/38, BW TECH SPECS I-5

ANSWER

1.02

(3.00)

a. (ACP) HIGHER)THAN $CP)(0.25)- Because Xenon will have built up (to nea

-c

peak value adding negative reactivity (0.5)

b. lACP) LOWE,R THAN (ECP)(0.25)- Because Xenon will have decayed off(to a low -%ddaf

value) adding positive reactivity (0.5)

(0.75)

c. I6CP) HIGHER THAN (ECP)(0.25)- Because with steam dumps maintaining higher

steam generator pressure / temperature, RCS temperature increases (due

decay heat and input from RCS pumps-optional) to add negative

reactivity (0.5)

(0.75)

d. lACP) LOWER THAN @C4(0.25)- Decause RCS temperature decreases (due to heat

removal in SGs-optional) and adds positive reactivity (0.5)

(0.75)

REFERENCE

REACTOR THEORY REVIEW FIG I-5.13/52

_

__

- _ _

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

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

PAGE 20

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

,

.>

s

ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE, G./REIDINGER

,

ANSWER

1.03

f2.00)

a. For equilibrium, llenon removal by decay must increase as power

-

increases (0.5). Xe.aon removal by decay is proptfitional to Xenon

concentration (0,5). (Therefore, Xenon concentration increases as

power increases - oottonal.)

(1.0)

b. Samarium does not decay (0.5). Samarium production and burnup rates

are proportional to power (0.5). (Therefore, Samarium concentration

is constant at power - optional.)

(1.0)

REFERENCE

REACTOR THEORY REVIEW I-5.68/77/78

huk62 b (1.50)

of

1.04

ANSWER

(a)(Current (2) = j nfak

r

Current (1 X (N2/N1)**3_= 20 X (4)**3 = 1280 amps (0.50)

9

(b)

low (2) = Flow (1) X (N2/N1) = 50 X 4 = 200 gpm

(0.50)

-

,

(c) Delta-P(2) = Delta-P(1) X (N2/N1)**2)= 8 X (4)**2 = 128 psid

(0.50)

REFERENCE

THERMAL HYDRAULIC PRIN 10-36

,

, - . , , -

. - , . - , , , _ ,

.- - . . - - , .

. . , - . - -

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

, . - . - - . . ,

, . . - , - . - - . - , - .

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

PAGE 21

THERMDDYNAMICS, HEAT TRANSFER AND FLUID FLOW

,

,,

ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE, G./RE!DINGER

.

ANSWER

1.05

(2.75)

'

a. 1) d)

2) a)

3) e)

(0.25 each)

k. wbhdeact)rcoolanh(mass)flowrate

(0.25)

.

b. 1) m

2) Cp

specific heat capacity of reactor coolant

(0.25)

3) At reactor coolant delta-T or (Thot-Tcold)

(0.25)

4) U

overall heat transfer coefficient for SG

(0.25)

5) A

total area of SG tubes

(0.25)

6) AH feedwater/ steam enthalpy change or (Hstm-Hfeed)

(0.25)

7)hf

feed flow rate or steam flow rate

(0.25)

8) AT delta-T across tubewalls or (Tavg-Tstm) or (Tavg-Tsat) (0.25)

REFERENCE

THERMAL-HYDRAULIC PRIN 12-12/13/14

-

QM A / ( .75)

ANSWER

1.06

..

a. (1){educi'1g pump flow reduces the minimum / required NPSH

(0.25)

(2)hedue

tem flow' increases the available NPSH

(0.25)

J

wd!4.

b. Restores the source of require NPSH

(0.25)

REFERENCE

THERMAL-HYDRAULIC PRIN FIG FND-FF-85, 10-60/61

I

i

)

_ . , - _ - -

-

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

-

_ . , . _ . . _ -

_ . _ . . _ , -

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

PAGE 22

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

.

,,

..

ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE,G./REIDINGER

'

\\

ANSWER

-1.07

(3.00)

a. Rhol = -2.5% delta-k/k

Keffl = 1/(1-rho) = 1/(1-( .025)) = .9756 t,0o2.

(0.75)

CR1/CR2 = (1-Keff2)/(1-Keff1)

135/405 = 1/3 = (1-Keff2)/(1.9756)

Keff2 = .992 t .002-

(0.75)

Rho 2 = 1-1/Keff2 = -0.81% delta-k/k

Reactivity added = -0.81% - (2.5%) = 1.69% delta-k/k i.,1*/o

(0.75)

b. Near 400 cps (0.25), because as Keff approaches 1, more neutron

generations are required to stabilize the neutron level (0.25).

(0.75)

1(I I I Iti NCl.

RLACIOR THLORY REVIEW 1-4.27/28

ANSWER

1.08

(2.00)

~

1) When the steam flow out of each SG decreases

SG temperature increases

,

Qo match RCS temperature](0.5), causing SG pressure to increase (0.5),

compressing / collapsing the steam bubbles in the riser (0.5). As a

result the SG levels drop or " shrink".(0.5)

(2.0)

OR (accept either answer)

(2) The loss of steam flow and recirculation flow through the SG moisture

separators removes the recirculation flow delta-P(0.5), removing the

need for recirculation flow driving head (0.5), allowing the downcomer

level to drop (0.5),and equalize with the lower level in the

riser. (0.5)

(2.0)

REFERENCE

THERMAL-HYDRAULIC PRIN 12-53

i

-

. -

-._.

.

_.

.

__ _ __ -

--

1.

PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

PAGE 23

,. ,'

- JHERMDDYNAMICS, HEAT TRANSFER AND FLUID FLOW

86/10/22-WEALE, G./REIDINGER

ANSWERS -- BRAIDWOOD 182

-

.

ANSWER

1.09

(3.00)

a. The fuel temperature coefficient (Doppler-only power coefficient)

overrides the small NTC(0.50), making the total power coefficient large

negative (0.50), requiring large positive reactivity (dilution) to

increase power.(0.50)

(1.50)

b. The large negative change in MTC over core life (0.50) makes the total

power coefficient much more negative at EOL than at BOL(0.5), requiring

a much larger positive reactivity addition (more dilution) to increase

power (0.50)

(1.50)

REFERENCE

REACTOR THEORY REVIEW FIG I-5.9/13/20/22/23

ANSWER

1.10

(2.00)

a. LESS THAN (0.25) - Differential Boron worth is greater because Boron

concentration is less, so less ppm required. (0.75)

(1.00)

,

b. LESS THAN (0.25) - Differential Boron worth is greater because Tavg is

less, so less ppm required. (0.75)

(1.00)

REFERENCE

REACTOR THEORY REVIEW FIG I-5.24/25/27

ANSWER

1.11

(3.00)

a.

Less negative (0.25) because Xe inserts negative reactivity in the

bottom of the core and flux moves to the top of the core (0.75).

b.

More negative (0.25) because .~ods are inserted and push the flux to the

bottom of the core (0.75)

c.

More negative (0.25) because more moderation will occur in the bottom

of the core due to sudden influx of colder Tcold (0.75)

'

REFERENCE

REACTOR THEORY REVIEW FIG I-5.34/35/36

%

-,,

- --+~,

a

- - - - - - , - . - ,

n-e,--n--.-

-

-- -

a

-

- , - - - - - - - - , - - , , - - - - - - - - . , , - - , - - , , - , - - - - - - - - - - - , , , , ,

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

PAGE 24

ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE, G./RE!DINGER

.

ANSWER

2.01

(1.00)

1) To maintain adequate flow to the 3 remaining SGs if one SG/AFW line is

ruptured OR keeps all AFW flow from going to ruptured SG/AFW line break.

(0.5)

2) Used for AFW flow detectors [p2 $gsh

(0.5)

,

REFERENCE

BW SYST TRNG MAN 26-23

ANSWER

2.02

(2.00)

a. Start /stop switch for other CCW purp on same 4.16KV ESF electrical bus

must be in " pull-to-lock" position.

(0.5)

b. Low CCW discharge header pressure

(0,5)

Safety injection signal

(0.5)

Power return after station blackout

(0.5)

REFERENCE

BW SYST TRNG MAN 19-12

ANSWER

2.03

(2.00)

(1) Interceptor control valves / reheat stop valves (0.5 for either)

in hot reheat steam lines to LP turbines (0.5)

(1.0)

(2) Non-return check valves / extraction steam isolation valves (0.5 for

either) in extraction steam lines to feedwater heaters (0.5)

(1.0)

REFERENCE

PW SYST TRNG MAN 35-16,36-16

!

.

_ _ - - _ - .

_ - _ . _ .

_ _ . . __

-

. . .

- -

. - - . _ - _

- . .

- -

. - - - - - - -

-

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

PAGE 25

ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE,G./REIDINGER

-

,

ANSWER

2.04

(2.00)

a.1) CVCS letdown entmt isol valves (CV-8160,8152)

(0.25)

2) CVCS letdown orifice isol valves (CV-8149A,B,C)

(0,25)

3) RCP seal water return entmt isol valves (CV-8100,8112)

(0.25)

4) Cntmt isol valves for CCW to excess letdown HX(CC-9437A,B)

(0.25)

b. No(0.25); without letdown, pressurizer level will increase (0.25) due to

RCP seal flow into RCS(0.25) until plant must be tripped due to high

przr level (0.25)

(1.0)

REFERENCE

BW SYST TRNG MAN 15A-15/16/38/41, TECH SPECS 3/4 6-18

ANSWER

2.05

(3.00)

1) BAST, boric acid xfer pump /CV-8 4

gn

50-60gpm

2) BAST, boric acid xfer pump and RMCS manual or borate (mod

0-40gpm

3) RWST, thru CV-1120/E to centrif chg pump suction; 105gpm

4) BAST, boric acid xfer pump thru CV-8439(throgalg) to charg pump

suction; less than 10gpm

(0.4 for each path, 0.2 per source, and 0.15 per rate)

(3.0)

(10gpm tolerance on flow rates)

REFERENCE

BW SYST TRNG MAN 158-27

ANSWER

2.06

(1.50)

.j u e

04.SEdnSalien

1) Centrif Charg Pump injection

2235 psig3((0.25)(10% tolerance on

2) Safety Inject Pump injection

1500 psig 0.25)

all pressures)

3) Accumulator injection

640 psig (0.25)

4) RHR Pump injection

200 psig (0.25)

(0.5 for order)

i

REFERENCE

'

BW SYST TRNG MAN Fig 61-16,58-63

.

_

___

_ _ _ _ _ . _ - _ . . -

- - . . _ - _ -

-

'

-

~2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

PAGE 26

~

ANSWERS -- BRAIDWOOD 182

-'

86/10/22-WEALE,G./RE!DINGER

-

.

ANSWER

2.07

(2.00)

1) Power to Bus 111 can be supplied from Unit 2 Bus 211(0.5) by closing

the 111/211 bus tie breakers (0.5)

(Optional - Check both bus grounds < 70vdc

Check bus voltages within 20vde

Close bus tie breakers 111 and 112

)

Isolate affected battery and battery charger)

2) Power to the 1A DG cont panel can be restored by shifting the removable

fuse block /no-blow link (0.5 for either) and closing the reserve feeder

,

breaker (0.5)

'

(Optional - Verify control power from reserve feeder

Open main and reserve supply breakers

Shift removable fuse block /no-blow link from main supply

receptacle to reserve supply receptacle

Close reserve supply breaker )

REFERENCE

BW SYST TRNG MAN 8A-13,21

ANSWER

2.08

(1.50)

a. After the turbine has tripped (0.25), turbine windmilling without steam

cooling (0.25)could cause overheating of the turbine blading(0.25).

(0.75)

b.1) To keep RCPs energized longer to maintain RCS flod if the turbine

trip was caused by a reactor trip.

(0.25)

2) To prevent RCP overspeed/ damage during steamflash on major LOCA(0.25)

3) To prevent overspeeding the turbine if some steam still present after

turbine trip.

(0.25)

REFERENCE

BW SYST TRNG MAN 5-29, 5-40

J

<

!

i

l

i

.-

- - - - , _ -

,,,,m_,_,._,_.-.,mm

~,.,__-..._,-._,,-.-_____,_.,__._._..__._.__,,__,.______=_-_,...-.,--_.,.__..m

. -

- - -

-

I

2.

PLANT DESIGN INCLUDING SAFETY AND EERGENCY SYSTEMS

PAGE 27

ANS14ERS -- BRAIDWOOD 182

-86/10/22-WEALE, G./REIDINGER

'

.

ANSWER

2.09

(1.75)

a. Provides sufficient leakoff flow to cool lower pump radial bearing (0.5)

at low RCS pressures (0.25).

(0.75)

b. Upper-1000psig(0.25);1eakoff flow should be sufficient above 1000psig;

if not, something wrong (0.25)

(0.5)

Lower-100psig(0.25);to prevent backflow of potentially dirty water

from VCT/CVCS system OR contaminants from the seal water return

filter,(0.25 for ett er)&gnUtadu </00&

(0.5)

DC MdA kn

REFERENCE

BW SYST TRNG MAN 13-37

ANSWER

2.10

(1.00)

a. Increasing CCW surge tank level

Increasing rad reading on CCW HX outlet Process Rad Monitor (PR009)

Surge tank vent closure on PR009 ALERT

(any 2, 0.25 each)

b. CVCS inlet valve to nonoperating HX is normally shut and check valve on

outlet should prevent pressure in tubes from exceeding shell side

pressure.

(0.25)

(Optional-the non-operating HX is normally isolated at power OR the CVCS

supply valve to the HX is shut and assume no leakage o

valve)

c. Shif t HXs (also acceptable-increase on local thermometer reading on CCW

outlet of operating HX)

(0.25)

REFERENCE

BW SIM MALF CCW-1; SYST DIAG M-64 SH 5

ANSWER

2.11

(1.50)

1) No locko(uts on/d)

ai. *o DG feeder brkr(1413) or SAT feeder brkr(1412)

(0.25)

3

2) DG at rated frequency and voltage

(0.25)

3) ESF cross-tie breaker from Unit 2 (ACB 1414) open

(0.25)

4) SAT feeder breaker (ACB 1412) open

(0.25)

5) ESF-to-non-ESF bas cross-tie breaker (ACB 1411) open

(0.25)

6) Control switch for DG feeder brkr ACB 1413 in After Trip

(0.25)

--.

. - -

, - ___

. - - - -

.

._.

__

.-

-

. , .

- . -

- _ - _ -

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

PAGE 28

'

dNSWERS--BRAIDWOOD182

-86/10/22-WEALE, G./REIDINGER

'

,

'

REFERENCC

BW SYST TRNG MAN 4-101

ANSWER

2.12

(2.50)

a.1) Cooling pump suction lines are located 4 feet below normal level (0.5)

2) The downward range of the skimers is limited to about 4 inches

below normal level (0.5)

3) Cooling pump discharge line's terminate about 6 feet above fuel

assemblies (0.5)

4) Cooling pump discharge lines have anti-siphon holes (0.5)

b. An inadverfnt criticality cannot occur (0.25) because the design of the

storage racks provides sufficient center-to-center spaging to prevent

inadvertent criticality @ven with no

gnthewatej(0.25)

REFERENCE

BW SYST TRNG MAN 51-13, 51-25

ANSWER

2.13

(1.25)

b

"#

~ b Mb

h#"kk

77;?p[uhr %@ d543 ry h

1) WTA voltage regulator trips

2) Base adjuster goes to no-load position

T41 unr

4 e -ESF 4/4o 6m

3) Main transformer cooling equipment trips

4) Auto 24-hour ground detector trips

n'

oca4 I-r,7-f

5) Bus duct cooling equipment trips

ggQgeg

g

%TO

RFFERENCE

BW SYST TRNG MAN 6-31/40

ANSWER

2.14

(2.00)

a. From RWST(0.25), to all RCS cold legs (0.25)

b. From containment sump (0.25), to suctions of SI and centrifugal charging

pumps (0.25)

c. From containment sump (0.25), to all RCS cold legs (0.25)

d. From containment sump (0.25), to RCS hot legs A and C(0.25)

-

.

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

PAGE 29

.

-

ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE,G./REIDINGER

.

,

-

REFERENCE

8W SYST TRNG MAN 58-51,52,53

.

l

!

-

- -

-

.

- . -

- -

- . - -

. .

,

3.

INSTRUMENTS AND CONTROLS

PAGE 30

.* ANSWERS.-- BRAIDWOOD 182

-86/10/22-WEALE, G./REIDINGER

'

1

-

ANSWER

3.01

( .50)

--b--

REFERENCE

BW SYST TRNG MAN 28-64,65,66;29-19

ANSWER

3.02

(1.00)

1) With 2 prs deenergized simultaneously, the P-10 signal will be actuated

to deenergize both Source Ranges (0.5).

2) Per Tech Specs both SRs must be operating for fuel load operations.(0.5)

REFERENCE

BW SYST TRNG MAN FIG 33-11, TECH SPEC 3/4.9.2

ANSWER

3.03

(2.50)

a. One charcoal booster fan starts and the filter bypass damper shuts (0.5)

b. Unit 1 and 2 CCW surge tank vent valves shut (0,5)

'

c. Outside air intake A dampers close, fan starts, and air intake A dampers

from turbine building open(0.5)

d. Steam Gen blowdown sample valves (1PS-179A-D - optional) close(0.5)

o. Energizes off-gas vent filter system (0,5)

REFERENCE

BW SYST TRNG MAN 49-64-68

.

-.y

-

. , , - . .

. - - - - , - . . - ,,

-_

,s,,

,.r-

-

w,--- - , - _. , . - .. . , .,y-.-.

-

,,,m-

-.

._,,, ,-

-m_.

e

y

.

3.

INSTRUMENTS AND CONTROLS

PAGE 31

ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE, G./REIDINGER

.

ANSWER

3.04

(1.50)

a.1) top 3 sensors of RVLIS channels A and B

2)10 hottest core exit thermocouples

3) average pressurizer pressure (NR)

4) average RCS loop pressure (WR)

(0.25 each)

(1.0)

b. 1)490 psig

2)35 F

(20f, tolerance, 0.25 each)

(0.5)

REFERENCE

BW SYST TRNG MAN 348-35

ANSWER

3.05

(1.50)

a. L of L is armed by C9 and C7(0.25); TT is armed by C9 and C8(0.25)

b. L of L conpares Tavg to Tref (0.25); TT compares Tavg to Tno-load (0.25)

c. L of L has 4F deadband(0.25); TT has no deadband(0.25)

REFERENCE

BW SYST TRNG MAN 24-31

ANSWER

3.06

(2.25)

a. Indicated steam flow will be higher than actual flow (0.25) because

steam pressure compensation is using too high pressure multiplier.(0.5)

b. Erroneous high steam flow will try to bring feed flow up(0.25), but

level-dominant SGWLC system will reduce feed flow as necessary to main-

tain constant programmed level (0,5)

c. Erroneous high steam flow will cause programmed feed pump delta-P to

go higb(0.25) causing feed pump to speed up to suply more delta-P than

normal at 1007, power (0.5)

(Give full credit on b and c if effect traced correctly even though

initial effect on steam flow reasoned incorrectly.)

REFERENCE

BW SYST TRNG MAN FIG 27-4,27-8

_ _ - _ _ _ - .-

.

._ _

_

3.

INSTRUMENTS AND CONTROLS

PAGE 32

.* A'NSWERS -- BRAIDWOOD 182

-86/10/22-WEALE,G./REIDINGER

'

-

1

ANSWER

3.07

(2.25)

l

a.

1) The steam dumps open(.25)

J

2) Cooling Tavg to about 550F(0.25)

3) The P-12 interlock will control RCS temperature by cycling the steam

dump valves around 550 F (0.25)

b.

1) All steam dump valves will shut (0.25)

2) Steam pressure will rise to the setpoint of the main steam

atmospheric relie* valves (0.25)

3) MS atmospheric reiief valves will cycle to maintain steam pressure

and RCS temperature.(0.25)

c.

1) Normal shift to t'

Turbine Trip controller will not occur (0.25)

'

2) Plant will cooldow because load Rejection controller will open

steam dumps (0.25)

3) The Load Rejection c troller will maintain RCS temperature near

'

"No Load" Tref (+4F d fation/deadband)[0.25]

O%d

contwlQ h TuQ(fLJ74)cos, ties (.es)

2M M

Mm M So-MW 62O

W SYS

RNG MAN FIG 24-7,9

3) %

& tLs V 4 m u n m e. food Tav' 6 rs)

ANSWER

3.08

( .75)

a.

TRUE

b.

FALSE

c.

FALSE

[0.25 each]

REFERENCE

BW SYST TRNG MAN FIG 608-10

ANSWER

3.09

(1.25)

'

a.

4

(0.25)

b.

2

(0.25)

c.

6

(0.25)

d.

5

(0.25)

o. ' S

(0.25),

,

'

REFERENCE

'

BW SYST TRNG MAN 658-35

.

$

. . -

- . _ _ _ . _

. . - -

. -

.

3.

INSTRUMENTS AND CONTROLS

PAGE 33

. . -.- ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE,G./REIDINGER

.

ANSWER

3.10

(1.00)

1) Turns all heaters off at 17% level (0.25) to prevent heater burnout (.25)

2.

Turns on backup heaters if actual level varies above program by 5%(.25)

in anticipation of an outsurge after an insurge of relatively cold

water [0.25].

REFERENCE

BW SYST TRNG MAN FIG 14-2

ANSWER

3.11

(1.50)

a.

IN(0.25), because Tavg is higher than Tref (0.25)

b.

OUT(0.25), because Tref becomes more than Tavg(0.25)

c.

IN(0.25), because the power mismatch circuit sees turbine power

decreasing below Rx power (also, Tref decreasing below Tavg)(0.25)

'

REFERENCE

j

BW SYST TRNG MAN FIG 28-14

!

ANSWER

3.12

(1.00)

Because the change in 1 power is greater at higher power levels for a given

change in reactivity (0,5), it reduces signal multiplier at high power (0.5)

(optional - prevents overshoot (0.5) which is more likely to occur at high

power levels (0,5))

REFERENCE

BW SYST TRNG MAN 28-27

,

ANSWER

3.13

(1.00)

,

1) Low feed flow ( < 1700gpm - optional)

(0.25)

f

2) Feed reg valve or bypass (1FW510,1FW510A) not closed, with(0.25):

Either SG low press or low level or low feed flow (0.25)

3) Any fee &ater isolation signal

(0.25)

i

M3 AM"

SC Hi-lh Lud % a 4 scs

Le M (%+) W P-4 sl4

'

,

..-.,,..,.m,.

- .. _--.-,- ,

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

_,._,v

_.....,-,.,,__,.,_.m.,-,n.n._.,,_...,,.n_

a..

,.. _ . ,,-,

-

3.

INSTRUMENTS AND CONTROLS

PAGE 34

'.

' ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE, G./REIDINGER

l

.

REFERENCE

BW SYST TRNG MAN 25-68

l

ANSWER

3.14

(2.00)

a.

Power range high flux - 103%

[0.3]

Intermediate range overpower - current equiv. to 20%

[0.3]

OP Delta-T - 3% below setpoint

[0.3]

,

OT Delta-T - 3% below setpoint

[0.3]

Urgent Failure Alarm (no setpoint required)

[0.2]

b.

Turbine power < 15%

[0.3]

Control bank D withdrawal stop - 223 steps

[0.3]

REFERENCE

BW SYST TRNG MAN 28-70,71

i

ANSWER

3.15

(1.00)

1.

OT Delta-T calculator

(0.25)

2.

OP Delta-T calculator

(0.25)

3.

P-12 circuitry (Hi Stm. Flow SI permissive,gStm. dump block)

(0.25)

4.

Feedwater isolation circuitry

(0.25)

REFERENCE

BW SYST TRNG MAN FIG 12-14

ANSWER

3.16

(1.50)

1) Safety Injection Signal (3.25)

2) Undervoltage on RCP busses (0.25)

(2/4)

(0.25)

3) One SG low-low level (0.25)

2/4 channels (0.25)

4) Undervoltage on bus 141 (0.25)

REFERENCE

BW SYST 1RNG MAN 26-42

- _ _ _ _ _ _ _ _ _ _

_

__

3.

INSTRUMENTS AND CONTROLS

PAGE 35

ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE, G./REI0INGER

.

,

9

ANSWER

3.17

(1.75)

a. Spray valves open fully

(0.25)

PORV-455A is opened

(0.25)

All pressurizer heaters interlocked off

(0.25)

bo Resuau

KV-45cA od 41ay yk

F. Backup pr% du.4s eius h(IPT-458I'will shut'PORV-455A

essure transditter

Pressure will continue to drop due to open sprays

Reactor will trip on pressurizer low pressure

Safety injection will initiate on pressurizer low pressure

Injection flow will fill pressurizer and stop spray effect

Pressure will cycle about PORV-455A backup pressure setpoint

(any 4, 0.25 each)

REFERENCE

BW SYST TRNG MAN 14-54

MSWER

3.18

( .75)

1) Overspeed (e4 cetAul dg*Ay$b

2) Interceptor valve fast timin{(e M(/MM)tfby

dh

y

3) Load drop anticipatorfat Nd hh25 a h)

-

REFERENCE

BW SYST TRNG MAN 37A-74

i

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

__ ____

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

. . _

__

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

4.

PROCEDURES - NORMAL, A8 NORMAL, EMERGENCY AN0

PAGE 36

RADIOLOGICAL CONTROL

.. . , . .

ANSWERS -- BRAIOW000142

-86/10/22-WEALE, G./REIDINGER

.

ANSWER

4.01

(1.50)

1. Restore Tavg > 550F(0,5) within 15 minutes (0.25), or

(0.75)

2. Be in hot standby (0.5) within next 15 minutes (0.25)

(0.75)

REFERENCE

TS, pgs. 3/4 1-6 8 B3/4 1-2

ANSWER

4.02

(2.50)

a.

Dose to whole body must not exceed 3 rem per quarter

(0.5)

The 5(N - 18) limit must not be exceeded

(0.5)

Individual's exposure history must be documented on NRC Form 4

(0.5)

b.

Administrative and Support Services Assistant Superintendent

(0.5)

c.

3000 mrem /Q - 1000 mrem = 2000 mrem at 200 mrem /hr = 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months

(0.5)

REFERENCE

CW RAD PROT STDS 24,25

ANSWER

4.03

(1.50)

a.

1.3% delta-k/k

(0.25)

b.

Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (0.25)

c.

Immediately initiate and continue boration(0.25) at > 30gpm(0.25) of

solution at > 7000 ppm (0.25) until shutdown margin ilmit is met.(0.25)

REFERENCE

TS 3/4 1-1

..

.

_ _ - . _ _ _ _ .

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

_-

_ - _ _ _ _ . - _ . _ _ _ _ _

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND

PAGE 37

RADIOLOGICAL CONTROL

.

ANSWERS -- BRAIDWOOD 142

-86/10/22-WEALE,G./REIDINGER

,

ANSWER

4.04

(1.00)

1) Uncontrolled increase in steam generator level

2) SJAE radiation abnormal

3) SG blowdown radiation abnormal

4) SG radioactivity abnormal

5) Main steamline radiation

(any 4 at 0.25 each)

REFERENCE

18wEP-3

ANSWER

4.05

(1.50)

The miniflow isolation valves for the SI pumps are shut during this

procedure (0.5). If the RCS pressure is greater than the SI pump discharge

pressure, there will be no flow thru the pumps (0.5) to remove pump heat

and resulting overheating could damage the pumps (0,5)

(1.5)

REFERENCE

18wEP ES-1.3; BW SYST TRNG MAN 58-37

,

ANSWER

4.06

(1.00)

1) Rod Ejection

(0.25)

2) Loss of Steam / Secondary Coolant Accident

(0.25)

3) Steam Generator Tube Rupture

(0.25)

4) Loss of Coolant Accident

(0.25)

REFERENCE

8W SYST TRNG MAN 58-6

,

. . . - - - . -

-

. - . _ _

.

-_.

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

. _ -

4. - PROCEDURES - NORMAL, ABNORMAL, EERGENCY AND

PAGE 38

RADIOLOGICAL CONTROL

j

=<

ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE,G./REIDINGER

t

-

ANSWER

4.07

(1.50)

.

1) Subcriticality

2) Core Cooling

3) Heat Sink

4) RCS Integrity

5) Containment Integrity

6) RCS Inventory

(0.2 for each LS7, 0.3 for order)

REFERENCE

SwST BOOK

ANSWER

4.08

(1.50)

a.

Blow it down to the blowdown system. [0.5]

b.

Blow it down (backfill) to the RCS.

[0.5]

c.

Cooldown with steam dumps.

[0.5]

REFERENCE

IBwEP-3

ANSWER

4.09

(1.75)

a.

Use AFW and steaa to steam dumps (0.50).

Use AFW and steam to steam generator PORVs(0.50).

(1.0)

b.

Draining and charging

Refueling pool cooling / recirculation

SI pump injection

Inject accumulators

(any 3, 0.25 for each)

REFERENCE

18w0A PRI-9,18w0A REFUEL-4

.

4.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND

PAGE 39

RADIOLOGICAL CONTROL

,y

s i

ANSWERS -- BRAIDWOOD 182

-86/10/22-WEALE,G./REIDINGER

,

ANSWER

4.10

(2.25)

1) Place a Centrifugal Charging pump in operation if not operating (0.25)

to have high flow rate for borated water (0.5)

(0.75)

2) Open IMOV-CV112D/E(RWST TO CHG PUMPS SUCT VLV)(0.25) to have source of

borated water (0.5)

(0.75)

3) Close IMOV-CV112 B/C (VCT TO CHG PUMPS SUCT YLV)(0.25) to try to stop

path of dilution. (0.50)

(0.75)

REFERENCE

IBw0A PRI-11

,

.

ANSWER

4.11

( .50)

--a.--

REFERENCE

IBw0A R00-3

ANSWER

4.12

( .50)

--b.--

REFERENCE

10CFR20

ANSWER

4.13

(1.00)

a.

50

b.

320

c.

180

,

d.

50

(0.25 each)

'

REFERENCE

BwGP 100-1

1

i

l

.

-

,

4

PROCEDURES - NORMAL, ABNORMAL, EERGENCY AND

PAGE 40

RADIOLOGICAL CONTROL

,

,3

%

-

ANSWERS -- BRAIDWOOO 182

-86/10/22-WEALE,G./REIDINGER

- ,

,

ANSWER

4.14

(2.00)

1) Manually run back the turbine at max rate by:

a) Pressing TURB MAN (0.33)

b) Pressing FAST ACTION (0.33) and GOV LWR (0.33) simultaneously

(1.00)

2) If turbine cannot be run back, Pull-to-Lock EH pumps (0,5)

3) If turbine cannot be run back, initiate Steamline Isolation (0.5)

REFERENCE

18wFR-S.1

ANSWER

4.15

(1.50)

1) Source range hi flux trip

2) Intermediate range hi flux trip

3) Containment pressure (hi-hpray actuation

(0.5 each)

REFERENCE

BW PLS PG 7

ANSWER

4.16

(3.50)

a. (1) Inadequate shutdown margin (Keff)(0.25), limit 0.95 (0.25)

(2) Inadequate shutdown margin (Boron)(0.25), ifmit 2000 ppm (0.25)

b. (1) Inadequate shutdown margin (delta k/k) (0.25), limit 1.3% (0.25)

(2) Rods low (0.25)below bank insertion ilmit (0.25)

(3) Failure of rods to fully insert on trip (0.25), limit one (0.25)

(4) Reactivity increase (0.25), unexplained / uncontrolled (0.25)

(5) Uncontrolled cooldown (0.5)

(6) Inability to borate normally (0.5)

(any 5, 0.5 each)

REFERENCE

18w0A PRI-2

,

- , , , . - . - - - .

,

, , . _ - . . -

- - - - . .

,

-

..-

- - - - , - ,- -

, .

ML20207M359

Text

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