ML20126L935

Exam Rept 50-259/OL-85-01 on 850409-12.Exam Results:Three of Six Reactor Operator Candidates & Three of Five Senior Reactor Operator Candidates Passed.Two Senior Reactor Operator Candidates Administered Reexams Failed

ML20126L935
Person / Time
Site:	Browns Ferry
Issue date:	05/10/1985
From:	Guenther S, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20126L868	List: ML20126L865 ML20126L935
References
50-259-OL-85-01, 50-259-OL-85-1, NUDOCS 8506200179
Download: ML20126L935 (189)

Text

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EhCLOSURE 1 EXAMINATION REPORT 259/0L-85-01 Facility Licensee: Tennessee Valley Authority 500A Chestnut Street Tower II Chattanooga, TN 37401 Facility Name: Browns Ferry Nuclear Plant Facility Docket Nos.: 50-259, 50-260, 50-296 Written and oral examinations were administered at Browns Ferry Nuclear Plant near Decatur, Alabama. Simulator examinations were administered at the Browns Ferry Simulator near-Soddy O isy, T nessee Sliulb i

Chief Examiner:

w Sie fried Guent r Date' Signed Ift o [S T Approved by:

AAAu_

Bruce /. Wilson,SectionChief Uate Signed Summary:

Examinations on April 9-12, 1985 Complete examinations (oral, simulator, and written) were administered to six RO candidates, three of whom passed. Complete examinations were also administered to five SRO candidates, three of whom passed. Two SRO candidates were adminis-tered written reexaminations; neither candidate passed.

t O

l REPORT DETAILS l

l 1.

Facility Employees Contacted i

l

• A. R. Champion, Instructor, Browns Ferry Hot License Training Unit

~

• W. D. Dawson, BWR Simulator Instructor

• E. S. Howard, Jr., Browns Ferry Licensed SRO l

• R. Hunkapillar, Operations Section Supervisor, BFNP (conference call) l

• J. D. Johnson, Browns Ferry, Requalification Unit Supervisor

• R. J. Johnson, Chief, Nuclear Training Branch (NTB)

• R. G. Jones, Browns Ferry Operations Training Section Supervisor

• G. Moody, Browns Ferry Engineering Training Supervisor

• C. H. Noe, Supervisor Operations Training (NTB)

~~

• L. Sain, Assistant Chief, Nuclear Training Branch

• J. Swindell, Assistant Plant Superintendent, Operations (conference call) 2.

Examiners

• S. Guenther, NRC, Chief Examiner

• J. Munro, NRC J. C. Kvamme, EG&G T. L. Morgan, EG&G

• Attended Exit Meeting 3.

Examination Review Meeting At the conclusion of the written examinations, the examiners met with R. G. Jones, J. D. Johnson, A. R. Champion, G. Moody, and E. S. Howard to review the written examinations and answer keys.

The following comments were made by the facility reviewers:

l a.

SRO Exam

{

1.

Question 5.2.a & e (1.25.a & e)

Facility Comment:

"The answer key response required two parts for full credit, but this was not stated in i

question (i.e., why core flow decreases and what caused this action to occur)."

NRC Resolution:

Contrary to the facility's comment, the question did state to " explain the cause(s) of" the designated recorder indications. The two part answers were derived directly from the facility's explanation for the plant response and are required for full credit.

The answer to part "a" has, however, been j

changed to allow full credit for stating that L

2 the recirculation pump would " runback" to minimum speed.

2.

Question 5.4.b (1.24.b)

Facility Comment:

"Concerning the fuel center line temperature with new fuel, your answer key looks for a decrease in fuel center line temperature.

This answer is correct if you consider some exposure time has been received to allow for compaction of the uranium oxide fuel.

Our MAPLHGR limits are adjusted to depress the new fuel Kw/ft limits to allow this compaction.

During this compaction the Helium gap increases, which decreases the heat transfer and will increase fuel center line tempera-ture.

After the compaction and the fuel cracks propogate (sic) and ratchet open, the helium gap decreases and increases heat transfer rate which will decrease the fuel center line temperature.

Reference page 9 Linear Heat Gen.

Rate &

Bases Lesson Plan thermo book.

D.5.a. new fuel exposed to operating temperature, densification occurs.

Reference transparency #2 Thermo-APLHGR Lesson plan tables of fuel types indicates depressee (sic) Kw/ft fuel limits from 0-200 mwd /t on exposures.

Reference page 171-172a-U-1 Tech Spec with fuel types from 0-200 mwd / ton exposure the heat transfer in KW/ft is depressed, due to poor heat transfer rates due to densification.

The question just addressed new fuel, it did not specify an exposure so based upon these references initially the fuel center line temperature of new fuel will increase due to poor heat transfer. After the cracks propo-gate (sic) and ratchet open fuel center line temperature will decrease. According to Tech Spec table this should have occurred by 200 mwd / ton exposure.

If a candidate responds with increase or - decrease, it should be accepted since no fuel exposure was specified."

l 3

NRC Resolution:

The reference quoted in the answer key states that the helium gap is essentially closed shortly after initial operation.

It was not considered necessary, therefore, to specify an exposure level in the question.

Based upon the references cited in the facility s comment, it is agreed that the most correct response of fuel centerline temperature to initial exposure is to increase due to fuel densification.

Any helium gap closure, and consequent fuel centerline temperature decrease, will occur after the fuel densifi-cation. The answer key has been changed to accept increase, rather that decrease, as the correct response.

3.

Question 5.4.d (1.24.d)

Facility Comment:

"Concerning question 1.24.d 'The effect on fuel center line temperature of operating with one MSIV closed.'

The question was stated ' operating with' this would indicate that the reactor had stabilized and survived the closure.

Initially the reactor pressure would increase and fuel center line temperature would increase during this transient. But as the EHC stabilizes the RpV pressure the center line temperature would effectively be the same while ' operating' with the one MSIV closed.

The candidates may respond with an increase, same as your answer key, if considering the fuel center line during the transient. Others will respond with ' remain the same' after the transient has stabilized."

NRC Resolution:

This comment is invalid in that it was not addressed during the allotted examination review period and should, therefore, have been made in writing to the Regional Branch Chief in accordance with Generic Letter 84-05.

It is also technically invalid in that it is not substantiated by the facility's reference material. The TVA BWR Simulator Instructor's Manual, malfunction #5, " Main Steam Line Isolation Valve (MSIV) Closure,"

indicates that the steady-state reactor pressure after an MSIV closure would be higher. Since a BWR operates under saturated conditions, this m

i 4

would result in a highar coolant temperature and, therefore, a higher fuel centerline temperature. No change to the answer key is warranted.

4.

Question 5.11 (1.19)

Facility Comment:

"This question's answers were all long and the time requirement to determine the most correct answer was too much for the assigned point value. Also the length of the answers make it too difficult to analyze each part."

NRC Resolution:

A candidate with a knowledge and understanding of the reactor heat balance equation should hav,e no difficulty in determining the one and only correct answer without having to perform l

a cifficult analysis of each part.

The i

question and answer key require no change.

l l

S.

Question 6.11 (3.13)

Facility Comment:

" Answer

'B' also could be correct if you consider that you are moving the last rod in a group.

This rod could then be moved two notches without violating any restriction."

l NRC Resolution:

The "b" distractor would be a correct answer to the question if the candidate assumed that the last rod in a group was being moved. The facility's reference material should be updated to reflect this anomaly. The answer I

key has been changed to include "b" as a correct response.

l 6.

Question 6.14 (2.4) i Facility Comment:

"Concerning question about what will trip a l

Diesel Generator during accident condition.

l Per 01 82 (Jan. 25, 1983) for UI & U2, i

differential and field failure both will trip l

D/G. The D/G lesson plan only list differen-tial, which is the correct answer. The OI 82 for U1 & 2 has been changed, but not in the material sent to NRC.

l Recommend that accept either answer as being correct."

I 5

(

i l

(

NRC Resolution:

This comment is invalid in that it was not addressed during the allotted examination review period (See resolution of Question 5.4.d).

The question and answer key accurately reflect the current equipment configuration and require no change.

7.

Question 6.17(2.13)

Facility Comment:

Per OI 84: Do not make-up nitrogen to the primary containment if venting is in progress, i

This would make answer 'c' incorrect."

NRC Resolution:

The question was intended to test the candi-dat.es' knowledge of the design of the Contain-ment Atmosphere Dilution System rather than its operating limitations.

The facility's I

comment is, however, valid and the question has been deleted from the examination.

8.

Question 6.26 Facility Comment:

"Concerning question 6.26 on the refueling interlocks with the moda switch in startup.

The copy of lesson plari 53 Figure 4 had a line inadvertently left off. Therefore the candi-dates response may be in more detail than the j

NRC answer key.

i 1.

Your answer key l

a.

refueling platform near or over core l

b.

service platform hoist loaded 2.

Response by candidates:

l a.

refueling platform near or over core with: and 1) monorail mounted hoist loaded l

or l

2) frame mounted hoist loaded or l

3) fuel grapple loaded or l

4) fuel grapple not full up l

l b.

service platform hoist loaded l

Reference attached drawing from BFNP i

FSAR" l

l l

6 NRC Resolution:

The Browns Ferry Nuclear Plant FSAR Figure 7.6.1 substantiates the facility's comment.

The answer key has been changed to include the additional information.

9.

Question 7.2 Facility Comment:

"A possible correct answer is 'a' per 0123 pg. 6, IV a.3..

Terminology of securing S. D.

cooling may be interpreted 'as removing that loop from service.'"

NRC Resolution:

The terminology, " secure RHR shutdown cooling," is quite explicit and also quite different from rerouting the RHR water through another heat exchanger as directed by OI-23.

Distractor "a" is not an acceptable answer to the question.

10.

Questien 7.4 (4.3)

Facility Comment:

"There doesn't seem to be any completely correct answer for this question if you considered that the RFPT was reset as given in direction to the class during the exam.

With the RFPT reset:

the AOP would be running, the suction valve open, the di:: charge valve closed, the minimum flow valve open, the blue lignt out, and in single element control."

NRC Resolution:

The candidates had been told to make an assumption based upon an incorrect interpreta-tion of 01-3, "Feedwater System," as verified on the Browns Ferry Simulator.

Since the stated assumption made none of the possible answers correct, this question has been M

deleted from the examination.

11.. Question 7.5 (4.4)

Facility Comment:

" Question concerning rubbing noises on Main Turbine at 100 rpm The correct answer (as shown on NRC answer key) contains statements from various sections of OI 47.

The correct answer contained an action statement requiring breaking vacuum.

This action statement was taken from Section III G.3.b which deals with actions to

t 7

be taken during emergency shutdown. The note is contained in the section which gives action to be taken at 1200-900 rpm range. Breaking vacuum at 100 rpm would have negligible effect on reducing the speed and could cause adverse effects on the unit due to pressure transient when the bypass valves go closed at 7"

vacuum."

NRC Resolution:

An immediate turbine

shutdown, per section III.D.I.e of 01-47, to protect plant equipment is considered to be an emergency shutdown (i.e. not normal) and thereby invokes

~ ~

the requirements of section III.G of that procedure. Although breaking condenser vacuum may not always be desirable, it is not incorrect. All the other possible answers to this question contain obviously incorrect actions per the OI; no change to the answer key is warranted.

12. Question 7.12 Facility Co=ent:

"See co=ents on R0 Section 4.11"

.NRC Resolution:

No co= ment on question 4.11 could be found in,

Facility Exam Co=ents.

The question and answer key were verified in accordance with the stated reference.

No change is considered necessary.

13. Question 7.22 (4.23)

Facility Comment:

" Question is requiring memorization of information within an appendix to a procedure.

The only immediate action per the procedure is to dispatch an operator to the backup control center."-

NRC Resolution:

The situation described in the question may be indicative of a simultaneous loss of' HPCI, ADS, and normal SRV protection.

The body of GOI-100-11 CAUTIONS the operator to perform the actions of the appendix concurrent with those of the procedure when such a situation occurs.

' Candidates able to conceptually describe the methods and objectives of the I

remedial / follow-up actions after dispatching an operator to the Backup Control' Center will receive the majority of the credit for the question.

l L'

E r

it. Oxstice 8.7 Fa:111:y' C:ement:

• 512.29 re=f res c:111: nice =f fem EF 125 f - detemici:q c;erattlity of e :f xeet vith a irc; D.fG

575 was r.c: rade avallatie :s l

• xasir.ne s.*

l MC Eescistics:

Fem EF 125 was act arafiable :: the e nnize-s is the refe e ce a.n.e-f a1. C: pies :f all :re a;;1icatie Te=:sf ral 5:erifica:1:cs, as ell i

as the *Lizitirg C ect:1:es f:r C:erati:e (LC3)* a-d *0;eratiitty" ceffritices re e l

';nrided *~ the entir+es. They a e es;+: e to, n:w := a:cly the c,. era:111:y :-iteria 1s varsi:q : Tar si: cati:cs. %: c~cqe := tne zesver Ley is ma-rarted.

15.n Cuesti:e 5.13 Fa:111:y C:mme:::

• 't'

.a -: - an ter :cssitie answer te.ich v:cid te scre fa.n111ar :: L e :pera;: s to:1C

me frent EDIs er :~:. G-I ;g 20

'Ec :: (sic) ser:~e er piare n i."'5 1: rie:.41 acce c less, by at leas:

.c 1 deArdett I

fedicati:cs (1) nis :eratic: != azt:matic scde is ::cfirned, er (2) a:e:2te :: e ::clie; 15 ass:-ec.

ht 'es:1c.: ter:

Sie e de etMicates seccid ect te everly

enalized f

r 1 ccesistercies 1

the ft:111:y's referecce waterizi, the uswer 1 Ley has bee: ch.azged := a11 v 70% credit f r t e EFG cr:ss-refererce ansver :.cted 1: the fa:111:y's

mmert.

Tre facility is ecco:.:-a;ed :: elinizate t*e 1,:orsistem:y ::

re

1:,ce pe=alizi:5 fc cre carcicazes.

15. Oastfee 5.14 (2.21)

Facility C:mme:::

"P.-eper reifef neus indiric a1 cces have ecostr.cted vie. Of 9-5 ard is 1: Z:ee 1.*

EC :.es:1ctice:

Tre fa:111:y's c:mme : is cc: side-ed valid.

Tne assuer Ley has t+e: c'eged :: 1=ci:.de

• ;e reliei by a.~.ather iice sed eqerat

  r*

as as acce;tatie aste.-. ate uswr.

t I

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9

17. Question 8.25 Facility Comment:

"This question requires memorization of an LCO. The Tech Specs are readily available in the control room for the SR0 to utilize in determining the LCO."

NRC Resolution:

The question is consistent with NUREG-1021,

" Operator Licensing Examiner Standards," in that it addresses an operating limitation as specified in the facility's Technical Specift-cations (Ref. ES-402). License candidates are expected to have a thorough knowledge of, and the ability to apply, significant Limiting Conditions for Operation (LCOs). No change to the question or answer key is warranted.

b.

R0 Exam Several facility comments apply to questions which appeared on both the R0 and SRO examinations. Those questions were addressed in section 3.a of this enclosure and will not be duplicated here.

1.

Question 3.14 Facility Comment:

"When 70% undervoltage is achieved on U-1 panel 9-9 the panel will automatically transfer to the U-2 unit preferred panel 11 breaker 1172. This transfer is for panel 9-9 only and is accomplished by an ASCO throw over switch if the. alternate supply has voltage.

Your answer was misleading in the fact that it addressed the U-1 unit preferred bus, which is in effect panel 11 in the battery board room, fed directly from the 1001, 1002 and 1003 breakers. Upon the loss of an MG set this panel will de-energize, but panel 9-9.in the control room will transfer to its alternate supply, U-2 breaker 1172, if potential is available. The result will be the U-1 unit preferred bus de-energized, which is panel 11 in battery board-room, and panel 9-9 energized from an alternate source.

ASCO - throws to alternate at 70% voltage with return to normal at 90% when normal potential is regained page 98 OI 57 U.R.2.b. addresses panel 9-9"

10 NRC Resolution:

The facility's comment is. substantiated by the additional reference material.

The question inferred that.the entire UPS bus transferred to an alternate supply when only the 9-9 panel does, in fact, transfer.

The question has been deleted from the examination.

2.

Question 3.28 Facility Comment:

"Concerning the question on the RWA latching into a group resulting in insert and withdraw errors.

The answer to part 'B' (after the rod in Group 1 has been withdrawn) would be depended (sic) upo,n whether the RWM was re-initialized or a red in Group 4 were selected or if all that was done was select a rod in Group 1 and withdraw it to 48.

Condition one:

Select rod in Group 1 and withdraw Results:

Group 3 remains latched group Only insert error is rod in group 2 Group 4 rod remains withdraw Condition two:

Select rod in group 1 and withdraw Select rod in group 4 Results:

Group 4 becomes latched group have two insert error (rod in Group 2 and 3)

No withdraw errors Condition three:

Select. rod in group 1 and withdraw re-initialize the RWM Results:

Same as condition two"

11 l

NRC Resolution:

The facility's comment is valid in that the candidates had to make an assumption to arrive at the answer specified in the answer key. If the candidate states his assumption and answers the question correctly based upon that assumption, then he will receive full credit for the question. If no assumption is stated, then the question will be graded per the answer key.

3.

Question 4.17 Facility Comment:

" Question does not call for response to first part of answer 'To produce sufficient pump head to create down thrust on the pump'"

NRC Resolution:

The brief explanation stated in the answer key was derived directly from 0I-27, " Condenser Circulating Water System." Both parts of the answer are required to fully explain the operating precaution. No change to the answer key is warranted.

4.

Exit Meeting At the conclusion of the site. visit, the examiners met with representatives of the plant and Nuclear Training Branch (NTB) staffs to discuss the results of the examination.

Those individuals who clearly passed the oral and simulator examinations were identified.

There were no generic weaknesses (greater than 75 percent of candidates giving incorrect answers to one examination topic) noted during the oral examinations.

Mr. R. J. Johnson, Chief, NTB, requested that copies of the written examinations and answer keys be left with the training staff. In accordance with established NRC policy, this request was denied. The meeting attendees were reminded, however, that copies of the master examinations and answer keys would be forwarded with the examination report.

The cooperation given to the examiners and the effort to ensure an atmo-sphere in the control room conducive to oral examinations was also noted and appreciated.

-MVpTEC. ~

y Snare:

UNITED STATES NUCLEAR REGULATORY COMMISSION o,,

2 o

REll N il 3

101 M ARIETTA STREET, N.W.

f ATLANTA, GEORGIA 30303 o

s,*****/

ENCLOSURE 3 U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION Facility:

Browns Ferry Reactor Type:

BWR-4 Date Administered: April 9,1985 Examiner:

S. Guenther Candidate:

INSTRUCTIONS TO CANDIDATE:

Use separate paper for answers. Write answers on one side only. Staple question sheet on top of the answers sheets.

Points for each question are indicated in parenthesis after the question.

The passing grade requires at least 70% in each cate o y and a final grade of at least 80%.

Examination papers will be picked up six 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination starts.

% of Category

% of Candidate's Category Value Total Score Value Category 30 25.b 1.

Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow 30_y 25 %

2.

Plant Design Including Safety and Emergency Systems 3023 25N g 3.

Instruments and Controls 30 v) 252+i 4.

Procedures - Normal, Abnormal, Emergency, and Radiological Control 12hg TOTALS Final Grade All work done on this examination is my own.

I have neither given nor received aid.

Candidate's Signature

o 1

1.

Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow.

1.1 A~significant amount of excess reactivity must be loaded into a core at BOL so that 100% power can be attained at the end of an 18-month fuel cycle.

From Column "B" select the correct reason for each positive reactivity requirement listed in Column "A".

(2.0)

Column "A" Column "B"

(+) AK/K Required Reason a.

4.77%

1.

Xenon Buildup b.

3.8%

2.

Samarium Buildup c.

1.0%

3.

Moderator Temperature Increase d.

3.0%

4.

Void Fraction Increase 1.2 MATCH the appropriate Thermal Limit (a-c),

(1.5) a.

Linear Heat Generation Rate (LHGR) b.

Average Planar Linear Heat Generation Rate (APLHGR) c.

Minimum Critical Power Ratio (MCPR) to each FAILURE MECHANISM AND to each LIMITING CONDITION given below:

FAILURE MECHANISM LIMITING CONDITION

' F1. Clad melting caused by L1.

Coolant transition decay heat & stored heat boiling following a LOCA F2. Clad cracking from the surface L2.

Clad plastic strain becoming vapor " blanketed"

<1%

F3. Clad cracking caused by high L3. Maximum clad tempera-stress from pellet expansion ture of 2200 deg. F 1.3 The " quality" of steam leaving the steam dryers refers to...

(1.0) a.

the ratio of the vapor volume to the sum of the liquid and vapor volumes.

b.

the ratio of the vapor mass to the liquid mass, c.

the ratio of the liquid mass to the sum of the liquid and vapor masses.

d.

the ratio of the vapor mass to the sum of the liquid and vapor masses.

i a

~

a 2

i 1.4 Which one of the following reactions is the most significant intrinsic neutron source in a highly exposed core?

(1.0) a.

-The high-energy gammas from the decay of fission fragments which cause increased spontaneous fission of Uranium and Plutonium.

b.

The significant increase in the Deuterium concentration which causes a concurrent increase in the Photo-Neutron source.

c.

The Alpha particles emitted from the decay of unstable fission fragments which cause a concurrent increase in the Alpha-oxygen reaction.

d.

The buildup of transuranic elements such as Curium-242 and Curium-244 which fission spontaneously.

1.5 Which one of the following is correct for constant speed centrifugal pumps?

(1.0) a.

The volume flow rate is directly proportional to the square of the pump speed.

b.

As the volume flow rate increases, the p' ump head increases.

c.

For two centrifugal pumps in parallel, the combined flow rate for a given head is equal to the sum of the individual capacities of the two pumps at that head.

d.

For two centrifugal pumps in series, the combined flow rate for a given head is equal to the sum of the individual capacities of the two pumps at that head.

1.6 Browns Ferry Unit 1 is operating at 100% rated thermal power. Which one of the following correctly describes the distribution of the fission energy which is being released?

(1.0) a.

Approximately 7% of the energy is being released as decay heat.

b.

No decay heat is being released; it begins at approximately 7% directly after a reactor trip.

c.

Approximately 80% of the energy is released instantaneously as prompt gamma radiation.

d.

Approximately 80% of the energy is released instantaneously as kt.stic energy of the-fission neutrons.

4 rr

w 3

1.7 Which of the following post accident containment hydrogen contributors is dependent on the radiation field intensity inside containment for the amount of hydrogen released?

(1.0) a.

Zr + 2H O + Zr02 + 2H2 2

b.

2Al + 3H O + Al 03 + 3H 2

2 2

c.

2H O + 2H

+0 2

2 2

d.

Fe + H.* Fe0 + H 2

2 1.8

.The change in reactivity associated with a change-in.Keff from 0.920 to 1.004 is approximately...

(1.0) a.

0.080 b.

0.084 c.

0.087 d.

0.091 1.9 Control Rod Worth...

(1.0) a.

increases as moderator temperature decreases.

b.

increases as K excess increases.

)

c.

increases as core void content increases.

~d.

increases as fuel temperature decreases, i

1.10 A BWR is operating at S85 psig with 217'F of feedwater subcooling. What is the feedwater temperature?

(1.0) a.

321.5*F I

b.

323.8'F c.

325.7'F d.

327.5'F 1.11 Which of the following is NOT a characteristic of i

subcritical multiplication?

(1.0) a.

The subcritical neutron level is directly proportional to the neutron source strength.

i b.

Doubling the indicated count rate by reactivity j

additions will reduce the margin to criticality by approximately one half.

l c.

For equal. reactivity additions, it takes longer for the new equilibrium SRM count rate to be reached as Keff approaches unity.

d.

A single notch of rod withdrawal will produce an j

equivalent equilibrium count rate 1.ncrease whether j

Keff is 0.88 or 0.92.

l 4

i j

i I

i

=

4 1.12 Which of the following is NOT one of the four basic neutron reactions of interest in reactor physics?

(1.0) a.

Elastic Scattering b.

Radiative Capture c.

Compton Scattering d.

Fission 1.13 Which of the following radioactive isotopes found in the reactor coolant would NOT indicate a leak through the fuel cladding?

(1.0) a.

Co-60 b.

Xe-133 c.

I-131 d.

Kr-87 1.14 Which of the following is NOT a correct statement concerning Xenon poisoning?

(1.0) a.

The concentration will bu'ildup and insert negative reactivity immediately following a reactor trip.

b.

Equilibrium Xenon reactivity worth at 50%

power is more than half of the equilibriuin Xenon reactivity worth at 100% power.

c.

Most Xenon is produced indirectly from the 1

decay of Tellurium and Iodine, d.

The time after a trip that Xenon peaks is independent of neutron flux before the trip.

1.15 Which of the following is NOT correct concerning delayed neutrons and their impact on reactor operations?

(1.0) a.

The delayed neutron fraction is the RATIO of the number of delayed neutrons produced to the number of prompt neutrons produced.

b.

When calculating reactor period, the delayed neutron term may be considered INSIGNIFICANT if the reactivity addition is GREATER than Beta.

c.

The magnitude of the average delayed neutron fraction is GREATER at BOL than at EOL.

d.

The presence of delayed neutrons causes the average neutron generation time to INCREASE.

5 1.16 Which of the following is a characteristic of the moderator temperature coefficient?

(1.0) a.

It tends to become less negative over core life.

b.

It tends to become less negative as control rod density increases.

c.

It tends to become less negative as moderator temperature increases, d.

It tends to become more negative as the void fraction is increased.

1.17 Which of the following describes the changes to the steam that occur between the inlet and the outlet of a REAL turbine?

(1.0) a.

Enthalpy DECREASES, entropy DECREASES, Quality DECREASES b.

Enthalpy INCREASES, entropy INCREASES, Quality INCREASES c.

Enthalpy CONSTANT, entropy DECREASES,-Quality DECREASES d.

Enthalpy DECREASES, entropy INCREASES, Quality DECREASES 1.18 Adding Latent Heat to liquid water at saturated conditions will...

(1.0) a.

increase the temperature of the water.

b.

change the water to steam at the same temperature.

c.

change the water to steam at a slightly higher temperature.

d.

decrease its subcooling by making it boil.

1.19 A reactor heat balance was performed (by hand) during the 00-08 shift due to the Process Computer being 00C.

Which of the following statements is correct concerning reactor power?

(1.0) a.

If the feedwater temperature used in the heat balance calculation was LOWER than the actual feedwater temperature, then the actual power is HIGHER than the currently calculated power.

1 b.

If the reactor recirculation pump heat used in the heat balance calculation has been OMITTED, then the ' actual power is LOWER than the currently calculated power.

c.

If the steam flow used'in'the heat balance calculation was LOWER than the actual steam 3

flow, then the actual power is LOWER than.

the currently calculated power.

d.

If the RWCU return temperature used in the heat balance calculation was HIGHER than the actual RWCU return temperature, then the actual power is LOWER than the currently calculated power.

m

6 1.20 Which one of the following statements is correct regarding counter-flow heat exchangers at Browns Ferry?

(1.0) a.

They are less coersonly used than parallel flow heat exchangers.

b.

Their heat transfer rate is lower than that of l

similar parallel flow heat exchangers with the same flow rates and temperatures.

c.

Their heat transfer rate is directly proportional to the temperature difference across the heat exchanger tubes.

d.

Their fluids flow in paths that are at an angle of 90' with respect to one another.

1.21 Which one of the following is an exa=ple of CONVECTIVE heat transfer in a BWR core?

(1.0) a.

Heat transfer across the helium gap in a fuel rod.

b.

Heat transfer across the laminar coolant boundary l

layer surrounding a fuel rod.

c.

-Heat transfer in the bulk coolant stream.

l d.

Heat transfer across the fuel pellet.

1.22 Attached figure 3-JJ, " Steady-state Natural Circulation l

Characteristics of a BWR,"

illustrates how " core flow" l

changes with respect to " reactor power" without forced circulation. EXPLAIN why incretental increases in power initially produce very rapid increases in core flow, but

~ eventually reach a point where further power increases l

produce no increase in core flow (1.5) 1.23 Using the enclosed Mollier Otagram, list the following property values for steam with an enthalpy of 1390 BTU /lbs and an entrepy of 1.568 BTU /lba

• F.

(1.5) a.

Pressure l

b.

Temperature c.

Superheat 1.24 Considering each of the following situations independ-ently (i.e., no other parameters change) STATE whether that condition would tend to INCREASE, DECREASE, or HAVE NO EFFECT on fuel centerline temperature.

(2.0) a.

A fuel bundle develops a zirc-oxide corrosion layer.

b.

A new fuel bundle is initially exposed..

s c.

The Pressure Setpoint on EHC is lowered by 10 psig.

d.

The plant is operated with 1 MSIV shut.

l l

m

7 I

1.25 The attached figures (SA & 58) represent a transient that could occur at a BWR.

Given:

(1) The "A" recirculation pun.p speed controller fails high at time T= 1.2 min 1

T.

(2) No operator actions are taken (3) Recorder speed of I division = 1 minute i

Explain the cause(s) of the following recorder indica-(2.5) tions:

a.

Core flow DECREASE (Point 3) b.

Reactor Pressure INCREASE (Point 6) c.

Reactor vessel level DECREASE (Point 13) d.

Reactor vessel level DECREASE (Point 14) e.

Total FW flow INCREASE (Point 17)

Write "End of Section 1" and start a new answer page 2

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8 2.0 Plant Design Including Safety and Emergency Systems 2.1 Which one of the following statements is correct regarding the closing operation of an outboard Main Steam Isolation Valve (MSIV)?

(1.0) a.

Air pressure is NOT used to close the MSIV.

b.

The drywell control air system supplies normal air pressure during the valve closing operation.

c.

Deenergizing either the AC or DC solenoid valve will admit air pressure to close the MSIV.

d.

During test operation the MSIV closes under spring pressure alone.

2.2 The Reactor Water Cleanup System (RWCU) is being operated in the " Hot Blowdown" mode to control reactor water level.

Which one of the following limitations is NOT a valid concern to the operator while controlling blowdown flowrate to the main condenser?

(1.0) a.

Exceeding the maximum allowable RBCCW system temperature exiting the NRHX.

b.

Exceeding the maximum allowable cleanup system temperature exiting the NRHX.

c.

Exceeding the FCV (#15 valve) downstream pressure setpoint.

d.

Exceeding the design cooling capacity of the regenerative heat exchanger.

2.3 Which one of the following is NOT correct regarding the "4" and "5" low pressure feedwater heaters?

(1.0) a.

Both the "4" and "5" heaters are horizontal U-tube heat exchangers with integral drain cooler sections.

b.

Both the "4" and "5" heaters are located inside the neck of the main condenser to minimize piping runs.

c.

Both the "4" and "5" heaters are continuously in service whenever the turbine is operating.

d.

Heater drains from the "3" heater cascade into the "4" heater but no heater drains cascade into the "5" he'ater.

9 2.4 Diesel generators A-D have started and are running in response to a high drywell pressure signal. Which one of the folloaing conditions will trip the diesel generator (s)?

(1.0) a.

Overcurrent b.

High differential current c.

Loss of field

,d.

Reverse Power 1

2.5 Backup scram valves provide a redundant means of venting air from scram pilot valves and scram dump valves. Which of the following correctly describes the design / operation of the Backup Scram valves?

(1.0) a.

They are normally energized and will deenergize upon a Reactor Protection System Scram signal.

b.

They are aligned such that two valves in series, one from each RPS trip system, must actuate to vent the scram air header, c.

They are designed such that both RPS channels must trip in order for any one of the valves to actuate.

d.

They are powered from the RPS buses A and B.

2.6 Which of the following correctly describes how the DRYWELL is prevented from exceeding its design i

external pressure?

(1.0) a.

" Suppression chamber - drywell" vacuum breakers vent air from the suppression chamber to the drywell at 0.5 psid.

b.

" Suppression chamber - drywell" vacuum breakers vent air from the drywell to the suppression chamber at 0.5 psid.

c.

"Drywell - reactor building" vacuum breakers vent air from the reactor building to the drywell at 0.5 psid.

d.

" Suppression chamber - reactor building" vacuum breakers vent air from the reactor building to the suppression chamber at 0.5 psid..

O

.~

~ - - -.

. ~_. - -

+

i I

10 2.7 The Standby Gas Treatment System (SBGTS) consists of three (3) redundant trains. Each SBGTS train...

(1.0)

I a.

discharges to a common header which is then piped to the i

plant stack for elevated release.

o j

b.

will auto start on 100 mres/hr in the reactor zone ventilation system and run until manually-secured.

c.'

has two HEPA filters designed to remove the Iodine fission products.

d.

has a decay heat damper which automatically opens on high charcoal bed temperature caused by fission product decay..

2 I

-2.8 Which one of the following statements correctly describes the design and operation of the Scram Inlet Valves?

(1.0)

J l

a.

They open slower than the Scram outlet. valves in order to prevent possible overpressurization of the CRD mechanism.

b.

They are normally held closed by air pressure and will open by spring pressure when either one of the two associated scram pilot air valves is deenergized.

i c.

They will not prevent their associated CR0 from' scramming if they fail to open, provided reactor pressure 'is greater than 200 psig, i

d.

They could cause a rod to drift in if they leak because the over piston area would be depressurized.

I 2.9 Which' one of the following statements is correct regarding j

the diesel generator motor-driven circulating oil pump?

(1.0) i l

a.

It, along with an immersion heater.in the oil sump, j

maintains the engine lubricated and warm while in standby readiness.

b.

It is secured once the shaft-driven oil pumps are operatin'g and supplying sufficient discharge head,

+

j c.

It takes its suction from the engine oil pan and supplies suction pressure to the piston and main oil pumps.

3

. d.

It provides the lube oil supply to the' engine's j

" turbocharger.

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11 2.10 Which one of the following statements is correct regarding the design and operation of major valves within the High Pressure Coolant Injection System (HPCI)?

(1.0) a.

All motor operated valves within the HPCI system are driven by 250 VDC motors.

The turbine stop and control valves will fail in the b.

closed position if the auxiliary oil. pump does not start upon system initiation.

c.

The HPCI steam supply valve (FCV-73-16) closes on a high reactor vessel water level signal (+54") and reopens on a low-low (-51.5") level signal.

d.

The HPCI suction valves will automatically shift from the suppression pool to the CST upon a hign level (7" above zero) in the suppression pool.

2.11 Which one of the following sequences of components correctly reflects the flowpath that a sample of water would follow as it traverses the Reactor Water Cleanup System?

(1.0) a.

"B" Recire. loop suction - RWCU pumps - NRHX - Regen.

HX - Filter Demins. - Regen. HX - Feedwater line b.

"A" Recirc. loop suction - RWCU pumps - Regen. HX -

NRHX - Filter Demins. - Regen. HX - Feedwater line i

c.

"A" Recirc. loop suction - Regen. HX - NRHX - RWCU pumps - Filter Demins - Regen. HX - Feedwater line d.

"B" Recire. loop suction - RWCU pumps - Regen. HX -

NRHX - Filter Demins. - Regen. HX - Feedwater line

12 2.12 The Off-Gas System consists of several major components connected in series to conduct noncondensible gases from the main condenser to the plant stack for release to the environment. This off gas flow...

(1.0) a.

is composed mostly of hydrogen and oxygen in a 2:1 ratio which is catalytically burned in the Recombiner.

b.

receives its driving force from the third stage air ejector which receives its normal steam supply from the main turbine cross around header.

c.

is superheated to about 350' F by an electric preheater to maximize the efficiency of the catalytic recombiner.

d.

will automatically be directed through the charcoal adsorber beds (i.e., inlet valve opens and bypass i

closes) when either off gas pretreatment radiation monitor reaches the Hi trip setpoint.

l Ekid J 41'The Containment Atmosphere Dilbtion (CAD) system is used as a post-accident method of hydrogen control. Which of G; pest cn the following correctly describes its design / operation?

(1.0) a.

It is automatically initiated when drywell pressure reaches the "2 psig" ECCS initiation setpoint.

b.

Units 1 and 2 share two redundant tr'ains while Unit 3 has two independent trains.

c.

It uses the SBGTS to vent gases from containment while adding N2 to keep the 02 concentration less than 5%.

d.

It shares its nitrogen s0pply with the Containment Vent, Purge and inerting system.

4 f

13 2.14 Which one of the following lists of responses (a through d) correctly describes the main turbine's response to an overspeed condition?

(1.0)

(a)

(b)

(c)

(d) 105%

1, 3, 5 intercept 2,4,6 all intercept 1, 3, 5 valves begin to intercept valves begin intercept throttle valves begin to throttle valves to throttle begin to throttle 107%

All intercept 2,4,6 All intercept 1,3,5 valves full intercept valves full intercept closed valves full closed valves full closed closed 110%

Mechanical Electrical Mechanical Electrical O'spd trip 0'spd trip 0'spd trip 0'spd trip 110.5%

Electrical Mechanical Electrical Mechanical O'spd trip 0'spd trip 0'spd trip 0'spd trip 2.15 Which one of the following statements correctly describes the operation of the the Motor Gear Unit (MGU) in controlling Reactor Feed Pump (RFP) turbine speed?

(1.0) a.

The MGU will control the RFP turbine speed only if its speed demand signal is greater than that from the MSC.

b.

The MGU can be used to control feed flow rate over a turbine speed range of approximately 0-5600 rpm.

c.

The MGU is manually controlled from the control room at either a high or a low speed rate.

d.

The MGU will fail "as is" to prevent a ramp response if it loses its signal from the flow controller.

r l'

14 2.16 Which one of the following statements correctly describes l.

the design / operation of the main steam safety / relief valves (SRVs)?

(1.0) l a.

There are a total of 13 SiVs with a tctal capacity of approximately 84% cesign steam ficw.

[

b.

Each SRY is monitored for leakage by a pressure i

switch in the ta11 pipe.

.c.

They are divided into three sets, with the first set of four valves cpening at a;;roximately 1125 psig.

d.

Each SRY is equip;ed wita-an air accumulater to i

assure that the valve can be c;erated fc11cwieg I

a failu-e of its neraal air su;;1y.

i l

2.17 Whica cee of the folicwing statements correctly describes the cperation/ design of the condensate beester purps?

(1.0) i

.a.

Three pumps are acesally cperated at 100% power,

-[

hcwever, the pus;s are rated te enable full ;cwer

~

c;eration with caly two ; umps c;erating.

l b.

The becster punc auto start time delay is necessary i

to allow prelubrication of tr.e pump's mechanical seals.

c.

Each booster pump has its cwn lubricating oil systee with the cenoensate flow actieg as t.% ccoling sedium..

i l

l d.

Each pump has an auxiliary oli p.ap.51.:5 auto starts when the asscciated bocster pump is started.

t 2.18 The reacter recirculation pump seal cartrid;e asseeclies l

censist of two sets of sealing surfaces ar4 a breakcown l

l bushing assescly. Failure of the #2 seal asseely at

{

rated cer4ttiens would result in which of the felicwieg indicaticns?

(1.0)

I I

a.

-An IE REASE in #2 seal cavity pressure free a;;rext-sately 500 psig to a;;rcrizately 10C0 psig.

l b.

A CECREASE in #2 seal cavity pressure from approxtrately l

500 psig to approximately zero psig.

l l

c.

An IEREASE in #1 seal cavity pressure from a:;reximately 500 psig to approximately ICs00 psig.

d.

A CECREASE in #1 seal cavity pressure free a;;reximately 500 psig to a;;roximately zero psig.

I r

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15 2.19 Which one of the following statements correctly describes the operation of the Standby Liquid Control System (SBLC) during initiation / injection?

(1.0) a.

SBLC initiation will shut the RWCU inlet isolation valves but NOT the outlet isolation valves.

b.

As each squib valves fires, its continuity monitor light will illuminate, indicating that the valve has opened c.

Taking the keylock control switch to either position will start the associated pump and fire both squib valves.

d.

Indicated flow rate for two pump operation should be approximately 100 gpm.

2.20 Unit 1 4KV Shutdown Board B has a normal and three (3) alternate power sources.

Identify the normal source and the three alternate sources of power, in the order of preference.

(2.0) 2.21 With regard to the Unit 1 RHR System, MATCH each of the items in COLUMN A with all its associated setpoints/

interlock (s) in COLUMN E (3.0)

NOTE: There may be more than one match required for COLUMN A; also, items in COLUMN B may be used more than once.

COLUMN A COLUMN B a.

LPCI Outboard Injection Valves (OPEN) 1.

100 psig b.

Recirculation Pump Discharge 2.

185 psig Valves (CLOSE) c.

Containment Spray Outboard 3.

230 psig Valves d.

ADS Permissive 4.

450 psig e.

Head Spray Inboard Valve 5.

-39 inch post-(CLOSED) accident Yarway Permissive 6.

1.96 psig drywell pressure permissive I

16 2.22 Which one of the following RBCCW loads is considered an ESSENTIAL load?

(1.0) a.

Drywell floor drain sump heat exchanger b.

Non-regenerative heat exchanger c.

Drywell atmospheric coolers d.

Reactor water cleanup pump seal water coolers 2.23 The Circulating Water system is being operated in the HELPER mode. Which one of the following (a through d) lists the correct gate lineup for this operating mode?

(Figure 2.23 is enclosed for reference.)

(1.0)

(a)

(b)

(c)

(d)

Gate #1 Open Closed Open Open Gate #1A Open Closed Closed Closed Gate #2 Closed Open Open Closed Gate #3 Op4n Closed Open Open a

2.24 Answer the following with regard to the EECW system:

The system is normally in standby with the (which ones)

RHRSW pumps assigned to EECW.

In the event that any of these pumps are out of service, the (which ones) pumps can be valved in from (control room / local) and the (which ones) pumps can be valved in from (control room / local)

(2.0) 2.25 True or Falst:

a.

Control air compressors "A" and "0" are powered from 480V shutdown boards "1A" and "2A" respectively and will auto-restart when power is restored af ter a load shed.

(0,5) b.

If control air pressure decreases to 90 psig, the service air header will automatically cross connect to supply the control air header.

(0.5)

Continue Section 2 on next page m

17 2.26 Which one of the following statements is correct with regard to the RCIC turbine trip throttle valve?

(1.0) a.

The valve is normally closed during standby conditions, b.

The valve uses a DC motor operator to remotely reset the valve after a turbine trip.

c.

The valve must be locally reset after an electrical overspeed trip.

d.

If operating, the valve will fail shut upon loss of electrical power to the relay logic bus "A".

WRITE "END OF SECTION 2" AND START A NEW ANSWER PAGE i

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18 3.0 Instruments and Controls 3.1 Which of the following statements is correct concerning the Reactor Manual Control System (RMCS) CRD Notch Override (RONOR) Switch?

(1.0) a.

When used to override notch withdrawal, it bypasses the RSCS group notch logic, b.

When used to override notch withdrawal, it interrupts timer motor power at the beginning of the drive out cycle.

c.

When held in the " Emergency Rod In" position, it will bypass all rod insert blocks, except those imposed by RSCS.

d.

When held in the " Emergency Rod In" position, it interrupts timer motor power at the beginning of the drive out cycle.

3.2 Which of the following statements is correct regarding the Recirculation Flow Control System 75% speed limiters?

(1.0) a.

They are activated when total feedwater flow decreases below 20% and a reactor vessel low water level alarm is received.

b.

They are designed to prevent recirculation pump cavitation by ensuring adequate net positive suction head.

c.

They enable the FWCS to recover reactor vessel water level upon loss of a reactor feed pump.

d.

They automatically reset when the initiating signal clears to enable the operator to restore normal recirculation flow.

3.3 The core contains 172 LPRM detectors in 43 detector assemblies. These 43 detector assemblies...

(1.0) a.

consist of asymetrically arranged dry tubes.

b.

consist of asymetrically arranged wet tubes.

c.

consist of symetrically arranged wet tubes.

d.

consist of symetrically arranged dry tubes.

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19 3.4 Four SRM channels are used to monitor core neutron flux levels during shutdown conditions, refueling operations, and startup.

Each source range l

detector / channel...

(1.0) l l

a.

uses an operating voltage (350 VDC) which is too low to cause secondary ionization, so the j

detectors operate as ion chambers.

l b.

operates in the proportional region of the gas amplification (ionization) curve and is therefore very sensitive to changes in detector bias voltage.

c.

uses'a pulse height discriminator to eliminate the larger pulses produced by direct gamma ionization of the chamber's argon gas, l

d.

has a pulse height discriminator which is bypassed when shutdown since the fission gamma flux is so low that discrimination is no longer required.

3.5 Which one of the following conditions will cause the load rejection circuit to trip?

(1.0) a.

A 40% mismatch between generator stator amps and turbine first stage pressure, b.

A 40% mismatch between generator megawatts and HP turbine exhaust pressure, c.

A 40% mismatch between generator voltage and turbine first stage pressure, d.

A 40% mismatch between generator stator amps and HP turbine exhaust pressure.

3.6 Which one of the following statements correctly describes the normal functioning'of the Off-Gas system Pretreatment radiation monitor system?

(1.0) a.

The Off-Gas isolation valve (#28) will trip shut if either pretreatment radiation monitor channel goes up scale "Hi-Hi-Hi".

b'.

The system utilizes a gamma sensitive scintillation detector to aid in locating ruptured or failed fuel elements, c.

The mechanical vacuum pumps will auto trip if either pretreatment radiation monitor channel reaches the "Hi" trip setpoint.

d.

The system utilizes a gamma sensitive ion chamber to monitor a sample stream drawn from the inlet.to the six-hour holdup volume.

i

20 3.7 The main turbine is at 1800rpa in prepara 1cn for synchronizing the main generator to the grid (i.e.

the 500kv generator breakers are s:111 c;en). What will happen if the "All Valves Closed" pushoutten fs depressed?

(1.0) a.

Nothing vill happen since the synchronous speed select signal is sealed in, b.

The four turbine centrol valves and four nain stop valves will close, but the intercept valves will resain cpen.

c.

The ten centrol valves (TCVs and IVs) and four eain s:co valves will close.

d.

The ten control valves (TCVs and IVs) will close, but the four main step valves will remain c;en.

3.8 Rod Block Monitor (RSM) Channel "A" is functioning norsally and a two-string red is selected for neverent.

Which one of ne fo11cwing statements correctly describes the e;eration of RBM "A" in this configuration?

(1.0) a.

Four LPRM inputs, from the A and C level detecto-s in each LPRM string, are routed to the averaging circuit.

b.

Four LPRM inputs, from the 3 and 0 level detectors in each LPRM string, are routed to the averaging circuit.

c.

Since an edge rod is selected the "A" RSM channel is autcaatically bypassed.

d.

R3M channel "A" is automatically bypassed because the minimum number of LPRM inputs are not present.

3.9 Which of the fo11 ewing statesents correctly describes the c;eration 'of the SRMs during a reactor startup?

(1.0) a.

They are initially fully inserted to their upper electrical step which is 30" above the center of the active fuel.,

b.

An intericek prevents detector withdrawal unless the SRM count rate is above 100 cps.

c.

All SRM rod biceks are automatically bypassed when all IRM range switches are on 7 or above.

d.

The reactor period indication is still valid with the SRMs in the fully withdrawn position.

I 21 l

3.10 Unit 1 is in the process of a plant startup with reactor power at 3%. Which one of the following l

signals will result in a Group I isolation?

(1.0) a.

Reactor water level decreases to -40.5" below instrument zero b.

MSL tunnel temperature increases to 210*F c.

Reactor pressure decreases to 825 psig d.

Drywell pressure increases to 2.45 psig s

3.11 Which one of the following statements correctly describes the operation of the Rod Position Information System (RPIS)?

(1.0) l a.

When a CR0 is driven beyond the full-in position the S51 overtravel reed switch will be actuated and the "00" full-core digital display for that rod will be backlit green.

b.

Both the SS2 and 500 normal full-in reed switches must be closed in order to receive the green backlit "00" full-core digital display.

c.

The 500 and $48 normal full-in and full-out digital display reed switches also supply rod position input signals to the RSCS logic when at > 50% rod density.

d.

If an electronic malfunction occurs, an RPIS INOP maynotbecorrh;edtoindicatethatRPISdata signal is genertt t, and a rod withdrawal block l

is imposed, i

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22 3.12 The Main Steam Line Radiation Monitoring System is designed to detect and control the release of gross amounts of fission products from the fuel.

This system...

(1.0) a.

monitors the gross gamma radiation from the main steam lines at a location just upstream of the inboard MSIVs.

b.

will initiate a PCIS Group I isolation, but NOT a direct scram when the trip level of 3X normal full power background is reached.

c.

will activate a " Main Steam Line Downscale" annunciator to alert the operator of an equipment malfunction while at power.

d.

will trip the mechanical vacuum pump, if running, and close its suction valves when the trip level of 1.5x normal full power background is reached.

3.13 Which one of the following statements is NOT correct regarding operation of the Rod Sequence Control System (RSCS) while in the Group Notch Control (GNC) logic mode?

(1.0) a.

When the first rod in a group is selected, its rod select pushbutton will illuminate brightly and the campanion rods in that group will have their select PB dimly backlit, b.

The GNC logic will prevent any rod in the group from being moved more than one notch in the direction of movement of the first rod in the group.

c.

When the last rod in a group is moved (all rods in same position) all dim backlights for that group will extinguish.

d.

The GNC logic will allow selection and movement of ANY group as long as all the rods within that group remain within one notch.

1 I

o 23 h je, b W 1dhich one of the following statements is correct regarding the

() j,.

power feeds for the Unit 1 Unit Preferred Power Supply (UPS)?

(1.0) a.

It~is normally powered by a motor generator set but will auto-transfer to the Unit 2 UPS bus if bus voltage drops to <70%.

b.

It is normally powered by'a motor generator set which is normally driven by an AC motor powere'd from 480 VAC Shutdown Bd. IB.

c.

It is normally powered by a motor generator set which is normally driven by a DC motor powered from 250 VDC battery Bd. 1.

d.

It is normally powered by a motor generator set but will auto-transfer to the Unit-Preferred Transformer if bus voltage drops to <70%.

3.15 A LOCA signal is received with a concurrent loss of offsite power.

The diesel generators reach operating speed and voltage and their output breakers close at time t=o.

Which one of.the.following major components is matched with its correct load sequencing time?

(1.0) a.

Core Spray Pump-1A at time t=o sec..

b.

RHR pump 18 at time t=7 sec.

c.

RHRSW pump Al at time t=14 sec.

d.

Drywell Blower at time t=21 sec.

3.16 Which one of the following statements best describes an IRM's operation / performance during a reactor startup?

(1.0)

When the IRM'is reading-full cscale on range 10, the a.

f APRM should be reading approximately 20% power.

b.

Shifting from~ range 4, indicating 75, to range 5 will result-in a reading of 24 on range.5.

c.-

Reactivity feedback due.to the moderator. temperature co(#ficient.should begin at approximately range 5.

d.

. Failure to insert the'SRM shorting links after refueling will result in'non-coincident IRM scrams.

~

c, e

L.

,4 L.

2.\\

24 3.17 Main turbine first stage pressure switches provide permissive and/or control signals for several plant functions. Which one of the following is NOT one of those control functions?

(3.o) a.

Recirculation Pump RPT breaker trip permissive.

b.

RSCS bypass.

.c.

RWM LPAP.

d.

TSV, 10% closure scram bypass.

3.18 Which one of the following is NOT a difference between an IRM and an SRM detector designed to enhance the SRM's sensitivity?

(1.0) a.

The SRM's argon fill gas pressure is higher than the IRM's.

b.

The SRM detector is slightly larger than the IRM detector.

c.

The SRM's operating voltage is higher than the IRM's.

d.

The SRM detector has more active coating than the IRM detector.

3.19 Which one of the following components does NOT receive an input signal from the LPRM flux amplifiers?

(1,0) a.

The process computer.

b.

The Rod Block Monitor System.

c.

The LPRM (Channel A & B) averaging circuits.

d.

The upscale and downscale trip units.

4 i

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E' M

25 3.20 Which one of the following signals will cause the HPCI turbine to, trip but will NOT cause a HPCI system isolation?

(1.0) a.

High HPCI turbine exhaust pressure @ 150 psig.

b.

Low reactor pressure @ 100 psig.

c.

High HPCI area temperature @ 200*F.

d.

HPCI turbine exhaust inboard disc rupture @ 10 psig.

3.21 The reactor is critical at approximately 10 psig and the "Heatup and Pressurization" phase of G01-100-1 is being performed. The NR GEMAC LI's in the control room read the following " approximate" values:

(1.0)

NR GEMAC A 37" NR GEMAC B 38" NR GEMAC C 37" The two Emergency System / Accident (Yarway) control room indicators should read approximately...

a.

... O inches b.

15 inches c.

38 inches d.

60 + inches 3.22 Which one of the following statements correctly describes the design / operation of the reactor recirculation pump EOC-RPT trip?

(1.0) a.

The EOC-RPT trip protects the plant from overpressurization transients (e.g. MSIV Closure) by tripping the recirculation pumps when reactor pressure exceeds 1120 psig.

b.

A trip of the EOC-RPT breakers results in a concurrent trip of the MG Set Drive Motor Breakers.

c.

The EOC-RPT breakers interrupt power'to the MG Sets during over-pressurization transients.

d.

The EOC-RPT trip results in a recirc flow decrease rate equivalent to that caused by a trip of the MG Set Drive Motor Breaker.

26 For each of the following Feed Water Control System (FWCS) malfunctions (questions 3.23through 25) select the correct system / plant response from the list (a-d) which follows. An answer may be used more than once, and no operator actions are taken.

a.

Reactor water level decreases and stabilizes at a lower level.

b.

Reactor water level decreases and initiates a reactor scram.

c.

Reactor water level increases and stabilizes at a higher level.

d.

Reactor water level increases and initiates a turbine trip (w/ scram).

3.23 The plant is operating at 100% power in 3-element control when one steam flow input signal is lost.

(1.0) 3.24 The olant is operating at 100% power in 3-element contol when one SRV. inadvertently opens (and remains open.)

(1.0) 3.25 The plant is operating at 100% power in 3-element control when one feed flow input signal is lost.

(1.0) 3.26 Unit 1 is operating normally at 100% power when APRM "A" fails upscale and results in a reactor half-scram. Utilizing the attached RPS trip logic diagrams (Figures 2, 5, & 8) describe in a step-by-step fashion (with regard to the opening / closing, de/ energizing of all applicable contacts and relays) how the APRM upscale trip results in an actuation of the scram solenoid valves.

(2.5) 3.27 TRUE or FALSE: All GM detectors / survey meters at Browns Ferry (e.g., RM-14, Teletectors) will quickly go fu11 scale and then drop to a lower on-scale reading when exposed to a high radiation field.

(0.5)

27 3.28 Assume the following initial rod position distribution:

All rods in groups 1 through 3 are fully withdrawn, except for one rod in each group 55 in group 1, 46-55 in group 2, and 18-03 in group 3 - all fully inserted.

All rods in groups 4 through 30 are fully inserted to position 0 except for rod 34-27 in group 4 which if fully withdrawn.

Fill in the following table with the Rod / Rod Group number you would expect to see displayed in each RWM window for both situationsbelow(a&b).

If nothing will be displayed write " BLANK".

(2.0)

RWM-(a)

(b)

Window Initial Same as IC Condition but rod 22-55 (IC) withdrawn to 48 Rod Group Insert Error Insert Error Withdraw Error WRITE "END OF SECTION 3" AND START A NEW ANSWER PAGE W

I

~

o 28 4

Procedures - Normal, Abnormal, E=ergency, and Radiological Control 4.1 hhich of the following is a valid operating precaution /

requirement with regard to the Raw Cooling Water system as stated in BF-01-247 (1.0) a.

Do not run the 3C RCW pump in automatic to prevent the possibility of overloading the 3C D/G during a simultaneous LOCA with LOSP.

b.

Running more than one RCW pump per unit with no condenser circulating water pumps running could cause cavitation due to low suction head.

c.

If recirculation MG sets are operating and RCW header pressure remains above 55 psig then both RCW booster pumps may be run in standby.

d.

The two auxiliary RCW pu=ps should be secured if cold water channel te=perature increases to 85*F when operating in a closed cooling tower cycle.

4.2 Which of the following is a 10 CFR 20 exposure limit?

(1.0) a.

1000 mres/ quarter whole body.

b.

5 rem / year whole body.

c.

15 rem /qtr to extremities.

d.

1250 mrem /qtr of beta radiation exposure to the lens of the eye.

.hw M A reactor / plant startup is in progress and preparations

' bCb^'

are being made to start the first Reactor Feed Pump (RFP) in accordance with BF-01-3, "Feedwater System." Which of the following lists (a through d) correctly specifies the conditions which should be established, per OI-3, prior to rolling the RFP turbine?

(1.0)

(a)

(b)

(c)

(d)

FWCS Mode Switch single-eles.

single-elen single elen.

3-elen.

AC TGOP AUTO OFF AUTO AUTO RFP Suction Valve Open Open Open Open RFP Discharge Valve Closed Closed Open Closed RFP Minimus Flow Valve Open Open Open Open RFPT Trip Status On On Off Off (Blue Light)

~

29 4.4 A main turbine generator startup is in progress per BF-0I-47.

While conducting system checks at 100 rps, you receive a turbine High Vibration alarm and a report from the turbine building ADO of a squealing noise coming from the HP turbine. Which one of the following sets of actions is correct per 01-477 (1.0)

-a.

Check bearing oil flows, temperatures (oil and metal) and seal steam header pressure while maintaining turbine RPM to clear the rub.

b.

Trip the turbine if unable to_ verify / restore proper oil / seal steam flow and clear the rub / vibration within the. allowed 5 minute hold period.

c.

Immediately trip the turbine and verify lift pumps running but do not engage the turning gear.

d.

Immediately. trip the turbine, break condenser vacuum, verify lift pumps running, and engage the turning gear when the zero speed alarm sounds.

~

4.5 Which one of the folicwing actions in response to a stuck open relief valve is INCORRECT, per the Abnormal Operations section of BF-01-1, " Main Steam System"?

(1.0) a.

Operate the relief valve contro' switch two or three times in an attempt to close the valve.

b.

Initiate suppression pool cooling to maintain torus temperature below 95'F and to prevent local hot spots.

c.

Change the EHC pressure regulator setpoint in an effort to'close the valve.

d.

With the supervisor's approval, manually scram the reactor once torus temperature reaches 125'F.

4.6 A plant startup is in progress.and condenser vacuum is being established in accordance with BF-0I-66, "Off-Gas System." What is the proper sequence for component startups per 01-667 (1.0) a.

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

b' Seal Steam Regulator, Steam Packing Exhauster,

- Mechanical Vacuum Pump, Steam -Jet-Air Ejector.

c.

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

d.

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

o 30 4.7 Which one of the following is an acceptable method per BF-0I-3, "Feedwater System," for reducing thermal duty on feedwater nozzles during low power and/or hot standby operation?

(1.0) a.

Reduce reactor power to minimize the frequency of cold feedwater cycles.

b.

Reduce reactor pressure and temperature to reduce the differential temperature on the nozzles.

c.

Maximize RWCU blowdown flow in an effort to stabilize feedwater flow control.

d.

Secure RWCU flow to minimize reactor heat load and power level.

4.8 A plant startup is in progress per BF-G0I-100-1.

Under which set of conditions (a through d) would it be permissible to place 4

the mode switch it t.he STARTUP position and commence the reactor startup?

(1.0)

(a)

(b)

(c)

(d)

No. of SRMs > 3 cps 3

4 3

2 RHR system mode / status SDC/STBY SDC/STBY LPCI/STBY LPCI/STBY Recirc pumps running A

A&B A&B A&B Licensed Operators 2

2 3

3 present 4.9 Unit I has just been started and synchronized with the grid and power is being raised to rated in accordance with BF-G01-100-1.

Which one of the following operating restrictions / limitations MUST be observed per the G0I?

(1.0) a.

A recirculation pump speed differential of 6 - 8%

(in manual) must be maintained while passing through the recirc pump critical speed zones.

b.

Load cha'nges are prohibited during periods in which the process computer is inoperable.

1 c.

When above the PCIOMR threshold but below the envelope,. rod movement shal.1 be "notchwise" but recirc flow changes are unrestricted.

d.

Average power during an eight-hour shift shall not exceed 3293 MWt; however, brief power peaks to the design limit of 3440 MWt are authorized.

=

31 4.10 Which of the following describes a correct method for verifying the position of locked valves per OSIL-437 (1.0) a.

Remove the locking device if necessary; check closed valves in the close direction only; reinstall the locking device, b.

Remove the locking device if necessary; check open valves in the open direction only; reinstall the locking device.

c.

Attempt to move open valves in the close direction; if unable to move the operator due to the locking device, verify the lock's integrity and valve stem position.

d.

Remove the locking device if necessary; check closed valves in the close and open directions; reinstall the locking device.

4.11 Which of the following is NOT a symptom that you would expect to see as a result of a " Jet Pump Failure" per 0I-687 (1.0) a.

Decrease in failed jet pump flow b.

Decrease in core differential pressure c.

Decrease in reactor (APRM) power d.

Increase in indicated core flow 4.12 Per G01-100-12, " Normal Shutdown From Power," which one of the following is the preferred method of cooling down and depressurizing the reactor with the MSIVs open and the feedwater system available?

(1.0) a.

Reduce the pressure regulator setpoint.

b.

Open BPVs using the bypass valve opening jack, c.

Bleed steam through the MSL drains.

d.

Bleed steam through the HPCI and RCIC steam line drains.

o 32 4.13 Which of the following conditions establishes the reactor in HOT STAh0BY, as defined by BF-GOI-100-1, " Integrated Plant Operations"?

(1.0)

Parameter Condition (a)

(b)

(c)

(d)

Mode Switch S&HS S&HS S&HS S&HS Coolant Temp 210*F 300*F 553*F 536*F Rx Pressure O psig 52 psig 1060 psig 920 psig MSIVs closed open closed closed 4.14 Fuel loading is in progress when you notice an unexplained increase in SRM count rate and an indicated reactor period; you suspect that an inadvertent criticality event is taking place. Which one of the following actions is NOT CORRECT per BF-GOI-100-3, " Refueling Operations"?

(1.0) a.

Manually scram the Rx and verify all rods full in.

b.

If criticality is still evident after all rods in, then stop the CRD pump.

c.

Evacuate the refueling floor after withdrawing the fuel bundle which caused the criticality event.

d.

Upon the SE's or ASE's approval initiate SBLC.

4.15 HPCI flow testing is in progress.

List the four (4) entry con-ditions for BF-E0I-2, " Containment Control."

(2.0) 4.16 Fill in the Blanks:

SF-G01-100-1, " Integrated Plant Operations," limits the initial plant heatup rate to (a) *F/hr until the recirculating water reaches (b)

• F at which time the limit is changed to (c)

• F/hr.

The initial-heatup rate limit is established to (d)

(2.0) 4.17 BF-0I-27, " Condenser Circulating Water System," cautions the operator not to operate one CCW pump alone unless the condenser discharge valves are throttled. Briefly explain why this precaution is necessary?

(1.0)

33 4.18 The shift engineer has determined that the control room must be abandoned due to smoke and directs that the reactor be manually scrammed.

List seven (7) specific (e.g., start, stop, trip, check...) operator actions that should be performed per the pro-cedure, given sufficient time, prior to reporting to the unit backup control center?

(2.0) 4.19 Subprocedure No. 4, " Core Cooling Without Level Restoration,"

of E01-1, "RPV Control," identifies six alternate injection systems which should be used in an attempt to restore water level if both loops of CS are inoperable and normal injection systems or subsystems connot restore level to -150" on LI-3-46 A/B

(+18" on LI-3-52 and 62). List four (4) of those alternate injection sytems/ flow paths.

(2.0) 4.20 List the conditions under which Standby Liquid Control injection is MANDATORY per E0I-47, " Failure of Reactor to Scram when required...".

(2.0) 4.21 During power operation, each control room is manned by an assigned licensed reactor operator who must normally remain in " Zone Number 1."

Under what two (2) Conditions may he enter Zone Number 2" while at power?

(1.0) 4.22 LIST AM the immediate operator actions required by E0I-1, "RPV Control."

(2.5) 4.23 Unit I has just scrammed on a low reactor water level signal and the actions of G0I-100-11, II.A are being executed. Water level is <33" and decreasing, despite recovery efforts, when you receive a report of a possible fire from the roving cable tray fire watch. What operator actions are required per Appendix B to G0I-100-11, " Reactor Scram?"

(1.5)

WRITE "END OF EXAM" AND SIGN YOUR ANSWER SHEET

' ~

2 V,t + 1/2 at

, a mg sa E. me-A. A,e ' t

~

KE = 1/2 mv a = (Vf - 1 )/t A = a,4 3

PE = m9n vf = V, + a t

• = e/t a = zn2/t1/2 = 0.693/t1/2 I/2'If

• EIII / ) I I5))

2 I

w,,.p 33 A*

4

[(t1/2)

• Ito))

cE = 931 sn m = V,yAc

-gx Q = m,ah I = I,e Q = mCoat

= UAL T I = I,e"*

Fwr = W ah I=I 10~ * /D' f

TVL = 1.3/u P = P,10,r(t) su HVL = -0.693/u P = P e*/'

o SUR = 26.06/T SCR = 5/(1 - K,ff)

CR, = 5/(1 - K,ff,)

SUR = 26o/t= + (a - o)T CR (1 - K,ff)) = CR II ~ "eff2) j 2

T = (t*/o) + [(8 - oV Io]

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

j T = t/(p - s)

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

T = (a - o)/(Io)

SOM = (

- K,ff)/K,ff a = (K,ff-1)/K,ff = AK,ff/K,ff t=

10 seconas I = 0.1 seconds-I o = [(L*/(T K,ff)] + [a,ff (1 + IT)]

/

I d) 2,2 2

=Id j

P = (tov)/(3 x 1010)

I d) gd j

22 2

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

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

Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem.

I curie ='3.7 x 1010cps 1 gal. = 3.78 liters 1 kg = 2.21 lem 1 ft3 = 7.48 gal 1 np = 2.54 x 103 Btu /nr Density = 62.4 1

/ft3 1 mw = 3.41 x 100 Btu /hr Density = 1 gm/c lin = 2.54 cm Heat of vaporization = 970 Btu /lem

• F = 9/5'C + 32 Heat of fusion = 144 Btu /lbm

• C = 5/9 ('F-32) 1 Atm = 14.7 psi = 29.9 in. Hg.

1 BTU = 778 ft-lbf 1 ft. H O = 0.4335 lbf/in.

2 e = 2.718 q

N3-

Volume, ft'/lb f athelpy, St:t/lb Entropy. Sty /lb a F TyP wetic Evap 8 team Ccter Evop Ctscm Water Evap 8trem P

"[

e.,

e, a,

a,,

a, 32 0.08859 0.01602 3305 3305

-0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 85 0.09991 0.01602 2948 2948 3.00 1073.8 1076.8 0.0061 2.1706 2.1767 35 40 0.12163 0 01602 2446 2446 8 03 1071.0 1079.0 0.0162 2.1432 2.1594 40 45 0.14744 0.01602 2037.7 2037.8 13.04-1068.1 1081.2 0 0262 2.1164 2.1426 45 to 0.17795 0.01602 1704.8 1704.8 18.05 1065.3 1083.4 0.0361 2.0901 2.1262 50 60 0 2561 0.01603 1207.6 1207.6 28.06 1059.7 1067.7 0.0555 2.0391 2.0946 60 70 0.3629 0.01605 868.3 868 4 38.05 1054.0 1092.1 0.0745 1.9900 2.0645 70 80 0.5068 0.01607 633.3 633.3 48.04 1048.4 1006 4 0.0932 1.9426 2.0359 80 80 0.6981 0.01610 468.1 468.1 58.02 1042.7 1100.8 0.1115 1.8970 2.0086 90 100 0.9492 0.01613 350.4 350.4 68.00 1037.1 1105.1 0.1295 1.8530 1.9825 100 110 1.2750 0.01617 265.4 265.4 77.98 1031.4 1109.3 0.1472 1.8105 1.9577 110 120 1.6927 0.01620 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 130 2.2230 0.01625 157.32 157.33 97.96 1019.8 1117.8 0.1817 1.7295 1.9112 130 140 2.8892 0.01629 122.98 123.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8895 140 150 3.718 0.01634 97.05 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686 150 160 4.741 0.01640 77.27 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1.8487 160 170 5.993 0.01645 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170 180 7.511 0.01651 50.21 50.22 148.00 990.2 1138.2 0.2631 1.5480 1.8111 180 190 9.340 0.01657 40.94 40.96 158.04 984.1 1142.1 0.2787 1.514S 1.7934 190.'

200 11.526 0.01664 33.62 33.64 168.09 977.9 1146.0 0.2940 1.4824 1.7764 200 210 14.123 0.01671 27.80 27.82 178.15 971.6 1149.7 0.3091 1.4509 1.7600 210 212 14.696 0.01672 26.78 26.80 180.17 970.3 1150.5 0.3121 1.4447 1.7568 212 220 17.186 0.01678 23.13 23.15 188.23

% 5.2 1153.4 0.3241 1.4201 1.7442 220 230 20.779 0.01685 19.364 19.381 198.33 958.7 1157.1 0.3388 1.3902 1.7290 230 240 24.968 0.01693 16.304 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 250 29.825 0.01701 13.802 13.819 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 260 35.427 0.01709 11.74b 11.762 228.76 938.6 1167.4 0.3819 1.3043 1.6862 260 270 41.856 0.01718 10.042 10.060 238.95 931.7 1170.6 0.3960 1.2769 1.6729 270 280 49.200 0.01726 8.627 8.644 249.17 924.6 1173.8 0.4098 1.2501 1.6599 280 290 57.550 0.01736 7.443 7.460 259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 300 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300,,

310 77.67 0.01755 5.609 5.626 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 320 89.64 0.01766 4.896 4.914 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 340 117.99 0.01787 3.770 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 i

360 153.01 0.01811 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 380 195.73 0.01836 2.317 2.335 353.6 844.5 1198.0 0.5416 1.0057 1.5473 380 400 247.26 0.01864 1.8444 1.8630 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 420 30S.78 0.01894 1.4808 1.4997 396.9 806.2 1203.1 0.5915 0.9165 1.5060 420 440 381.54 0.01926 1.1976 1.2169 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 460 466.9 0.0196 0.9746 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 480 566.2 0.0200 0.7972 D.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4516 480 500 683.9 0.0204 0.6545 0.6749 487.9 714.3 1202.2 0.6890 0.7443 1.4333 500 520 812.5 0.0209 0.5386 0.5596 512.0.

687.0 1199.0 0.7133 0.7013 1.4146 520 540 962.8 0.0215 0.4437 0 4651 536.8 657.5 1194.3 0.7378 0.6577 1.3954 540 SCO 1133.4 0.0221 0.3651 0.3871 562.4 625.3 1187.7 0.7625 0.6132 1.3757 560 550 1326.2 0.0228 0.2994 0.3222 589.1 589.9 1179.0 0.7876 0.5673 1.3550 580 600 1543.2 0.0236 0.2438 0.2675 617.1 550.6 1167.7 0.8134 0.5196 1.3330 Goo 620 1786.9 0.0247 0.1962 0.2208 646.9 506.3 1153.2 0.8403 0.46S9 1.3002 620 640 2059 9 0.0260 0.1543 0.1802 679.1 454.6 1133.7 0.8656 0.4134 1.2821 640 660 2365.7 0.0277 0.1166 0.1443 714.9 392.1 1107.0 0.8995 0.3502 1.2498 660 l

640 2708.6 0.0304 0.0808 0.1112 758.5 310.1 1068.5 0.9365 0.2720 1.2086 680 700 3094.3 0 0366 0.0386 0.0752 822.4' 172.7' 995.2 0.9901 0.1490 1.1390 700 705.5 3208.2 0.0508 0

0.0508 906.0 0

906.0 1.0612 0

1.0612 705.5 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)

A.3

welume. lis/ib Er.thelpy. Dtv/lb E1tropy. Stu/eb e F Energy. giv/in P"'*-

Pm 1;'"P teter

Eve, Seeam W;ter Evep Steam Ceter Evep Seeom Ceser Storm pole I

pele

• r

'er

's h

h, e

o e,

e, e,

u, s

e.0486 32 Alt 0A1602 3302.4 3302.4 0.00 1075.5 1075 5 0

2.1872 2.8872 0

1021.3 e.0886 e.10 35 023 0.01602 2945.5 2945 5 3 03 10738 10768 0 0061 2.1705 2.1766 3A3 1022J e.10 0.15 45.453 0 01602 2004.7 20047 13.50 1067.9 1081.4 0 0271 2.1140 2.1411 13.50 1025.7 9.15 0.20 53.160 0 01603 1526 3 1526 3 21.22 1063.5 1084.7 0.0422 2.07?8 2.1160 2122 1028.3 320 i

9.30 64.444 0 01604 1039.7 1039.7 32.54 10b7.1 1089.7 0 0641 2.0168 2.0809 32.54 1032 0 0.30 0.40 72A69 0.01606 792.0 792.1 40.92 1052.4 1093J 0.0799 1.9762 2.0M2 40.92 1034.7 0.40 O.5 79.586 0.01607 641.5 641.5 47.62 1048 6 1096.3 0 0925 1.9446 2.0370 47A2 1036.9 9.5 0.6 85.718 0 01609 540.0 540.1 53.25 1045 5 1098.7 0.1028 1.9186 2.0215 53.24 1038.7 0.6 0.7 SO 09 0.01610 466.93 4 % 94 58 10 1042 7 1100 8 0.3 1.8966 2.0083 58.10 1040.3 0.7 OA 94.38 0.01611 411.67 411.69 62.39 1040 3 1102.6 0.1117 1A775 1.9970 6239 1041.7 0.8 9.9 98.24 0.01612 368.41 368.43 66.24 1038.1 1104.3 0.1264 1A606 1.9870 66.24 1042.9 0.9 1.0 131.74 0.01614 333.59 333 60 69.73 1036.1 11058 0.1326 13455 1.9781 49.73 1044.1 1.0 2.0 126.07 0.01623 173.74 173.76 94.03 1022.1 1116.2 0.1750 1.7450 1.9200 94A3 1051A 2.0 3.0 14147 0.01630 118.71 118.73 109.42 1013 2 1122 6 0.2009 14854 1.8864 109.41 10 %.7 8.0 4.0 152.96 0.01636 90 63 90 64 120.92 1006.4 1127.3 0.2199 1.6828 13626 120.90 1060.2 4.0 5.0 162.24 0.01641 73.515 73.53 130 20 1000.9 1131.1 0.2349 1.6094 13443 130 18 1063.1 5.0 6.0 170.05 0.01645 61.967 61.98 138 03 996.2 1134.2 0.2474 1.5820 12294 13BA1 1065.4' 6.0 7.0 176.84 0.01649 53 634 53.65 144.83 992.1 1136 9 0.2581 1.5587 13168 14421 1067.4 7.0 8.0 182.86 0.01653 47.328 47J5 150.87 988.5 1139.3 0 2676 1.5384 1A060 15034 1069.2 8.0 9.0 188.27 0 01656 42.385 42.40 1H.30 985.1 1141.4 0.2760 1.5204 1.7964 156.28 10703 94 10 193.21 0.01659 38.404 38 42 161.26 982.1 1143.3 0.2836 1.5043 1.7879 161.23 1072.3 10 -

14.696 212.00 0.01672 26.782 26 80 180.17 970.3 1150.5 0.3121 1.4447 1.7568 180.12 1077.6 14.896 15 213.03 0.01673 26.274 26.29 181.21 969.7 1150.9 0.3137 1.4415 1.7552 181.16 1077.9 15 20 227.96 0.01683 20.070 20.087 196.27 960.1 1156.3 0.3358 IJ962 1.7320 196.21 1082.0 to 30 250.34 0.01701 13.7266 13.744 218.9 945.2 1164.1 0 3682 1.3313 1.6995 218 5 1087.9 30 40 267.25 0.01715 10 4794 10 497 236.1 933 6 1169.8 0.3921 1.2844 1.6765 236.0 1092.1 40 50 261.02 0.01727 8.4967 8.514 250.2 923.9 1174.1 0.4112 1.2474 J.6586 250.1 1095.3 50 60 292.71 0.01738 7.1 M 2 7.174 262.2 915.4 1177.6 0.4273 1.2167 1A440 262A 1098.0 80 70 302.93 0.01748 6.1875 6205 272.7 907.8 1180.6 0 4411 1.1905 1A316 272.5 1100.2 70 80 312.04 0.01757 5 4536 5 471 232.1 900.9 1183.1 0.4534 1.1675 1A208 281.9 1102.1 80 90 320.28 0 01766 4.8777 4.895 290 7 894.6 1185.3 0.4643 1.1470 1.6113 290.4 1103.7 90 100 327.82 0.01774 4.4133 4.431 298.5 888.6 1187.2 0.4743 1.1284 1.6027 298.2 1105.2 100 120 341.27 0.01789 3 7097 3.728 312.6 877.8 1193 4 0 4919 1.0960 1.5879 312.2 1107.6 120 140 353 04 0 01803 3.2010 3 219 325 0 868.0 1193 0 0.5071 1.0681 1.5752 324.5 1109.6 140 160 363 55 0 0;815 2.E155 2.834 335 1 8590 1195.1 0.5205 1.0435 1.5641 335.5 1111.2 160 180 373 08 0 01827 2.5129 2.531 346 2 850.7 1196.9 05328 1.0715 1.5543 345.6 1112.5 180 200 351.80 0 01829 2.26S9 2.287 355.5 842.8 1198J 0 5438 1.0016 1.5454 3543 1113.7 200 250 40097 0.01665 1.8245 1.8432 3761 825 0 1201.1 0.5679 0 9585 1.5264 3753 1115.8 250 300 417 35 0 01859 1.5233 1.5427 394 0 803.9 1202.9 0.5SB? 09223 1.5105 392.9 1117.2 300

~

350 421.73 0 01913 1.3064 1.3255 409.8 7942 1201 0 0 6055 0 8939 1.4968 409 6 1118 1 350 400 444 60 00193 1.14162 1.1610 424.2 760 4 1204 6 0 6217 0 8630 1.4847 422.7 111E 7 400 450 456 28 0 0195 1.01224 1.0318 437.3 767.5 1204.8 06360 0.8378 1.4738 435.7 1118.9 450 i

500 467 01 0 0199 0 90787 0 9276 449 5 755.1 1204.7 06490 0.814S 1.4639 447.7 1118 8 500 S50 47694 00199 0 82183 0.8418 460.9 743.3 1204 3 06611 0.7936 1.4547 456.9 1118 6 550 600 485 10 0 0201 0.74962 0.769S 471.7 732.0 1203 7 0.6723 0 7738 1.4461 469.5 11IE.2 500 700

.503 08 0 0205 0.63505 0 6556 491.6 710.2 1201.8 0692R 07377 1.4304 488.9 1136.9 700 800 518 21 0 0209 0.54809 0.5690 509 8 689 6 1199 4 0 7111 0.7051 1.4163 506 7 1115.2 800 900 531 93 0 0212 0 4796S 05009 526 7 6E9 7 11 % 4 07279 06753 1.4032 523 2 1113.0 900 1000 5 *4.5B 00216 0 42435 0 4460 542.6 f 50 4 1192.9 07434 0 6476 1.3910 53";6 1110 4 1000 1100 55C 2<

0.0220 0.37Bf 3 04005 557.5 631.5 !!E91 07573 06216 1.3794 5531 1107.5 1100 1200 ' :67.19 0 0223 0 34013 0.3625 571.9 613.0 1164 8 07714 0 5969 1.36S3 556 9 1104 3 1200 1300 577.42 0 0227 0 30722 0.3299 585 6 544.6 1180 2 0.7843 0 5733 1.3577 580.1 1100 9 1300 l

l 1400 537 07 0 0231 0.278/1 03018 598 8 576 5 1175 3 0.7966 0 5507 1.3474 592.9 1097.1 1400 1500 5 % 20 0 0235 025372 0.2712 611.7 558 4 11701 0.8035 05233 1.3373 605 2 1093.1 1500 2000 635 80 0.0257 D16760 0.1883 672.1 465.2.1133 3 0 8625 04256 1.7881 662 6 10GS 6 2000 2500 ES$11 0 02c;f 010209 0 1307 731.7-361.6 1093 3' O9139 0 3206 1.2345 118.5 1032.9 2500 3000 695 33 0 0343 0 050/3 0.0650 801 8 218 4 10203 0 9728 01E91 1.1619 782A 973.1 3000 3208.2 701 47 0 050B 0

0 050d 906 0 0

906 0 1.0612 0

1.0612 875.9 875.9 3708.2 TABLE A.3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE)

A.4

Temposshwe. F 14bs pseen.

g to p) 300 200 300 400 500 000 700 000 900 1000 1100 3200 1300 4400 1500 v

0.0161 392 5 45:*.3 511.9 571.5 631.1 690 7 B

6 68 00 1150 2 1195.7 1241.8 1788 6 13M 1 1384 5 (101.74) s 0.1295 2.0609 2.1152 2.1722 2.2237 22708 2J144 e

0.0841 78 14 90.24 102.24 11421 126 15 13808 150 01 161.94 173 86 185 78 197.70 209 62 221.53 233 45 1

6 6

GS 01 lies 6 1144 8 1241.3 1788 2 3335.9 1384 3 1433 6 1483 7 1534.7 1586 7 1639 6 1993 3 1748.0 180).

1 (162.24) s 0.1295 13716.13369 1.9943 2.0460 2.0932 2.lM9 2.1776 2 2159 2 2521 2.28E 2.3194 2 3509 2.3811 IA101f 00161 38 84 44 93 51 03 57.04 63 03 69 00 74 98 80 94 86 91 9287 95 84 les to 110.76 116.72 30 6

68 02 1146 6 11937 1740 6 12873 1335.5 13840 14334 14835 1534 6 1586 6 1639 5 1693.3 1747.9 3803 4 (197.21) s 0.1295 1.7928 1.8593 1.9173 1.9692 2.0166 2.0603 2.1011 2.1334 2.1757 2.2101 2.2430 2 2744 2.3046 2 y

0 0161 0.0166 29 893 33 963 37.985 41.966 45.978 49 964 53 946 57.926 61.905 65 882 69858 73.833 77307 i

35 6

68 04 168 09

!!92 5 1239.9 1287.3 1335.2 1383.8 1433 2 1483 4 1534.5 1586.5 1639 4 18832 3747A 1800 4 (213.03) s 0.1295 0.2940 1.8134 18720 1.9242 1.9717 2A155 2.0563 2.0946 2.1309 2.1653 2.1982 2 2297 2.2599 2.2890 e

00161 0 01M 22.356 25428 28 457 31 466 34 465 37.458 40 447 43 435 46 420 49 405 52.388 55370 58 352 to 6

68 05 168 11 1191.4 1239.2 1286.9 1334.9 1383 5 1432 9 1483 2 1534.3 1586.3 1639.3 1693.1 17473 1803.3 (227.96) s 0.1295 0.2940 1.7805 1A397 13921 1.9397 1.9836 2.0244 2.0628 2A991 2.1336 2.1665 2.1979 2.2282 2.257 e

0.0161 0 0166 11 OM 12.624 14.165 15 685 17.195 18 499 20 199 21 697 23 194 24 589 26.183 27.676 29.1 40 6

48.10 168 15 1186 6 1236.4 1785.0 1333 6 1382.5 1432.1 1482.5 1533 7 15858 1638 8 1992 7 1747.5 1803 0 (267.25) s 0.1295 0 2940 1.6992 1.7608 14143 13624 1.9065 1.9476 1.9860 2.0224 2.0569 2.0899 2.1224 2.1516 2.180 e

0.0161 0.0166 7.257 8 354 9.400 10.425 11.438 12 446 13.450 14.452 15.452 16.450 17A48 18.445 19.441 to 6

68.15 168 20 list 6 1233.5 1283.2 1332.3 1381.5 1431.3 1481.8 1533.2 15853 1638.4 1882.4 1747.1 1802A (292.71) s 0.1295 0.2939 14492 1.7134 1.7681 1.8168 1A612 1.9024 1.9410 1.9774 2A120 2.0450 2A765 2.30E 0.01"61 0 0166 0.0175' 6 218 7.018 7.794 S.560 9 319 10.075 10 829 11 581 12.331 13.081 13 229 14.577 80 6

6821 168 24 269 74 1230.5 1281.3 1330.9 1380 5 1430.5 1481.1 1532 6 1584.9 1638 0 1692.0 1746A 1802.5 e

l (312.04) s 0.1295 0.2939 0.4371 1.6790 1.7349 1.7842 13289 13702 1 9089 1.9454 1.9800 2.0131 2.0446 2.0750 2.1041 e

0 0161 0.0166 0 0175 4 935 5.588 6.216 6.833 7.443 8050 8 455 9258 9360 10.460 11.000 11.659 500 4 48.26 168.29 269 77 1227.4 1279.3 1329.6 1379 5 1429.7 1480 4 1532.0 1584.4 1637.6 1891.6 1746.5 1802.2 (327A2) s 0.1295 0.2939 0.4371 1A516 1.7088 1.7586 13036 1A451 13839 1.9205 1.9552 1.9883 2A199 2.0502 2.0794 '

=

0 0161 0 0166 0 0175 4 0786 4.6341 5.1637 5.6831 6.1925 6.7056 7.2060 7.7096 S.2119 S.7130 9.2134 9.7130 i

120 4 6831 168.33 269 81 1224.1 1277.4 1328.1 1378 4 1428.5 1479.8 1531.4 1583.9 1637.1 1991 3 17462 1802A (341.27) s 0.1295 0.2939 0 4371 1.6286 1AS72 1.7376 1.7829 13246 13635 1.9001 1.9349 13680 13996 2.0300 2.0592 e

0 0161 0 0166 0 0175 34651 3 9526 4 4119 4 3585 5.2995 5.7364 6.1709 6.6036 7 2349 7.4652 7 2946 S.3233 140 4 68.37 168 38 26S 85 1220 8 1275.3 1326 8 1377.4 1428 0 1479 1 1530 8 1583 4 1636.7 1690.9 1745.9 1801.7 (353 04) s 0 1295 0 2939 0 4370 1.6085 1.6686 1.7196 1.7652 1.8071 13461 1.8828 1.9176 1.950S 1.9825 2.0129 2.0421 1

e 0 0161 0 0165 0 0175 3 0060 3 4413 3 8480 4.2420 4 6295 5 0132 5 3945 5.7741 6 1522 6 5293 '6 9055 7.2811 i

150 4 68 42 16642 269.89 1217.4 1273 3 1325 4 1376 4 1427.2 1478 4 1530.3 1582.9 1636.3 1690.5 1745.6 1801.4 4

i (363 55) s 0.1294 0 2938 0 4370 1.5906 1.6522 1.7039 1.7499 1.7919 1.8310 1.8674 1.9027 1.9359 13676 1.9980 2.0273 0 016! 0 0166 00!?4 2 6474 3 0433 3 4093 3.7621 4.1064 4.4505 4.7907 5.1289 5 4657 53014 6.1363 6.4704 150 6 68 47 165 47 269 9/

1213 8 1271.2 1324.0 1375.3 1426 3 1477.7 1529 7 1582.4 1635.9 1640 2 1745J 1801.2.

(373 C&1 s C 1294 0.2938 0 4370 1.5743 1.6376 1.6900 1.7362 1.7784 1A176 1.8345 1.8894 1.9227 1 9545 1.9849 2.0142 e

0 0161 0 0166 0 0174 2 3598 2.7247 3 0583 3 3783 3 6515 4 0008 4.3077 4.6128 4.9165 52191 5.5209 5.8219 200 6 68 52 16851 269 96 12101 1269 0 1322f 1374 3 1425 5 1477.0 15291 1581.9 1635 4 1689 8 1745.0 1800 9 (331.60) s 01294 0 293S 04359 1.5593 1.6242 1.6776 1.7239 1.7663 1.5057 1.8426 1.8776 1.9109 1.9427 1.9732 2.0025 e

OC161 0 0165 0 0174 0 0166 2.1504 24662 2 6872 2.9410 3 1909 3 4382 3 6837 3 9278 4.1709 4 4131 4.6546 250 6 68 66 168 63 270 05 3/5.10 1263.5 1319 0 1371.6 1423 4 1475 3 1527.6 1580.6 1634 4 1688.9 1744.2 1800.2 (400 97) :

01294 02937 04355 0 5567 1.5951 1.6502 1.6976 1.7405 1.7601 1.8173 1.8524 1.8d58 1.9177 1.9482 1.9776 00161 0 0165 0 0174 00186 1.7655 2 0044 2 2263 2 4407 2.6509 2 6555 3 0643 3.2688 3 4721 3.6746 3 8764 300 6 68 79 155 74 27u 14 375.15 1237 7 1315 2 1368 9 1421.3 1473 6 1526.2 1579 4 1633 3 1688 0 1743 4 1799.6 ej (417.35) :

0.1294 02937 0 4337 05%5 1.5703 1.6274 1.6758 1.7192 1.7591 1.7964 1.8317 1.8652 12972 1.9278 1.9572 e

0 0161 0 0105 0 0174 0 018G 1.4913 1 7028 13970 2 0332 2 2652 2 4445 2.6219 2.7980 2.9730 3.1471 3.32C5 250 6 68 92 161s B5 270 24 375 21 1251.5 1311.4 13662 1419 2 14718 1524 7 1578.2 1632.3 1687.1 1742.6 1798.9 (431.73; a 0 1293 0 29M 0 43G7 0.5664 1.5483 1.6077 1.6571 1.7009 1.7411.l.7787 13141 12477 12795 1.9105 1.9400 v

0 0161 0 0166 0 0174 0 0162 12841 1.4763 l'.6490 1.8151 1.9759 2.1339 2.2901 2.4450 2.5987 2.7515 2.9037 400 a 69 05 16897 270 33 375 27 12451 1307.4' 1363 4 1417.0 14701 1523 3 1576 9 1631.2 16S6 2 1741 9 17932 (444.60) :

0 1293 0 2935 0 4366 0 56C3 1.5782 1.5901 1.6406 1 6850 1.7255 1.7632 1.7988 1A325 1A64 I

0 0161 0 0166 0 0174 0 0186 0 9919 1.1584 13037 1.4397 1.5708 1 6932 1.8756 1.9507 2.0746 2.1977 2.3200 1796 500 6 69 32 109 19 270$1 34 38 1231.2 12991 1357.7 le!? 7 1466 6 1520 3 1574 4 16291 1684 4 1740 3 1.8998 (457.01) :

0 1292 0 2934 0 4364 C htto 149?! 1 5595 1 6123 1 65/8 1 6990 1.7371 1.7730 1.8069 13393 1 8702 TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE)

A.5

, se psees.

T:mperehwhF i

Clot n-pt. asap) 200 300 800 400 600 600 700 e00 000 1000 1100 3200 1300 It00 1500 v

0.0161 0 01 % 0 0174 0 0l86 0 7944 0 94 % 10726 1.1892 13008 14093 1.5160 16711 1.7252 1 8284 1.9309 g80 a 69 58 169 42 270 70 375 49 1215 9 1290 3 1351 8 3408 3 1463 0 1517.4 1571.9 3627.0 1682 6 1738 8 1795 6 (48620) s 0.1292 0.2933 04E2 0.M57 1 4590 1.5329 1.5844 1.63bt 16769 3.71 % 1.7517 1.7859 1A184 1.8494 12792 e

90161 0 0146 0 0174 0 0186 0.0204 0.7928 0 9072 1.0102 1.1078 1.2023 1.2948 1J858 1.4757 3.u47 1.6530 380 6 69 84 369.65 270 89 375 61 487.93 1281.0 1345 6 1403.7 1459 4 1514 4 1%94 1624.8 3680 7 3737.2 3794.3 (503.CS) s 0.1291 0.2932 0 4360 0 % 55 0.6889 1.5090 1.M73 1 6154 14580 3.6970 17335 1.7679 18005 38318 1.8617 i

e 0.0161 0 01 % 0 0174 0 0186 0 0704 0.6774 0 7825 0.8759 0 9631 1.0470 1.1289 1.2093 12825 1.3M9 1.4446 888 6 70.11 16988 271.07 37573 48738 1271.1 1339.2 13991 14558 15114 1%69 1622.7 1678 9 - 1735 0 1792.9 (5182.).

0.1290 0 2930 0 4358 0.% 52 0.6885 1.4869 15484 1.5980 1.6413 1 6807 1.7175 17522 1.7851 18164 1.8464 i

e 0.0161 0.0166 0 0174 0 0186 O C2' 4 0 5869 0 6858 0.7713 0 8504 0.9262 0 9998 1.0720 1.1430 1.?!31 1.2825 l

J 908 4 70.37 170 10 271.26 375A4 487.83 1260 6 1332.7 1394 4 1452.2 1508.5 IM44 1620 6 3677.1 1734.1 1791.6 (531.95) s 0 1290 0.2929 0.4357 0.M49 0.6881 1.4659 1.5311 1.5822 16263 1.6662 1.7033 3.7382 1.7713 1.6028 34329 d

e 0.0161 0 0166 0.0174 0.0186 0 0204 0 5137 0 6080 0 6875 0.7603 0 8295 0 8966 0 9622 1.0766 1.0901 1.1529 3000 6 70.63 170 33 271.44 375.96 487.79 1249 3 1325 9 1389.6 1448 5 1504.4 1 % 1.9 1618 4 1675.3 1732.5 1790.3 (544.58) s 0.1269 0.2928 0.43 % 0.5647 0.6876 1.4457 1.5149 1.5677 1.6126 1.6530 14905 1.72M 1.7589 1.7905 1.8207 i

e 0.0161 0 0lM 00174 0.0185 0.0203 0 4531 0 5440 0.6188 0 6865 0.7505 0 8121 0 8723 0 9313 0 9894 1.0468 1100 4 70.90 170.56 271 A3 376 08 447.75 1237.3 13188 1384.7 1444.7 1502 4 1559.4 1616 3 3673.5 1731.0 1789.0 (556J8) s 0.1269 0.2927 0.4353 0 5644 0.6872 1.4259 1.4996 1.H42 1.6000 1.6410 1.6787 1.7141 1.7475 1.7793 12097 e

0 0161 0.0lM 0.0174 0.0185 0.0203 0.4016 0 4905 0.5615 0.6250 0.6845 0 7418 0.7974 0 8519 0.9055 0.9584 3288 6 71.16 170.78 27122 37620 487.72 1224.2 1311.5 1379.7 1440.9 1449.4 1M69 1614.2 1671.6 1729.4 1787.E.

(567.19) s 0.1288 0.2926 0.4351 0.5642 0.6868 1.4061 1.4851 1.5415 1.5883 1.6298 1.6679 1.7035 1.7371 1.7691 1.7996 e

0 01(I O 0166 0 0174 0 0185 0 0203 0.3176 0 4059 0 4712 0 5282 0 5809 0.6311 0 6798 07272 0.7737 0 8195 1400 4 71.68 171.24 272.19 376 44 487.65 1194.1 1296.1 1369 3 1433 2 1493 2 1551.8 1609 9 16fA 0 1726.3 1785.0 j

(587.07) s 0.1287 02923 0.4348 0.5636 0.6859 13652 1.4575 1.5182 1.5670 1.6096 1.6484 1.6845 1.7185 1.7508 1.7815 e

0.0161 0.01M 00173 0.0185 0.0202 0.0236 0.3415 0.4032 0.4555 0.5031 0 5482 0.5915 0 6336 0.6748 0.7153 1600 &

72.21 171.69 272.57 376 69 487.60 616.77 1279.4 1358.5 1425.2 1486.9 1546.6 1605 6 1664.3 1723.2 17823 (604.87) s 0.1286 0.2921 0.4344 0.5631 0.6851 0.8129 1.4312 1.4965 1.5478 1.5936 1.6312 1.M78 1.7J22 1.7344 1.7657 e

0 0160 0.0165 0 0173 0.0185 0.0202 0.0235 0 2906 0 3500 0.3988 0.4426 0.4836 0.5229 0.5609 0.5980 0.6?43 3000 a 72.73 172.15 272.95 376.93 487.56 615.58 1261.1 1347.2 1417.1 1480.6 1541.1 1601.2 1660.7 1720.1 1779.7 (621.02) s 0.1284 0.2918 0.4341 0.5626 0.68*.3 0.8109 1.4054 1.4768 1.5302 1.5753 1.61H 1.6528 1.6876 1.7204 1.7516 e

0 0160 0.0165 0.0173 0.0184 0.0201 0.0233 0.2488 0.3072 0 3534 0.3942 0 4320 0.4680 0.5027 0.5365 0.5695 2000 6 7326 172 60 273.32 377.19 487.53 614.48 1240.9 1353 4 1408 7 1447.1 1536.2 1596.9 1657.0 1737.0 1777.1 (635 80) s 0.1263 0.2916 0 4337 05621 0 6834 0.8091 1.3794 1.4578 1.5138 1.5603 1.6014 1.6391 1.6743 1.7075 1.7369 e

0 01(0 0.0165 0.0173 0 0184 0.0200 0 0230 0 1681 0 2293 0 7712 0.3068 0.3390 0 3692 0.3980 D4259 0 452S 2500 6 74 57 173 74 27427 377.82 487.50 612.08 1176.7 1303 4 1386.7 1457.5 1522.9 1585.9 1647.8 1709.2 1770 4 (MB.11) s 0.1280 0.2910 0 4329 05609 0.6815 0 8048 1.3076 1.4129 1.47M 1.5269 1.5703 1.6094 1.6456 1.6796 1.7116 e

0 0160 0 0165 0 0172 0 0183 0 0200 0.0228 0 0982 0 1755 0.2161 0.2484 0.2770 0.3033 0.3282 0.3522 0.3753 3000 A 7563 17tEB 27522 378 47 467.52 610.06 10M 5 1267.0 13632 1440.2 15014 1574.8 1635 5 1701.4 17(1.8 (695.33) s 0.1277 0.29J4 0.4320 0.5597 0.6796 0 8009 1.1966 1.3692 1.4429 1.4976 1.5434 1.5641 1.621d 1.6561 1 6688 e

f10160 0 0165 0 0172 0.0183 0.0199 0.0227 0.0335 0.1588 0 1987 0.2301 0.2576 0.2827 0J06% 0.3291 0.3510 3200 h 76 4 175.3 27b 6 378 7 487.5 609 4 800 8 1250 9 1353.4 1433.1 1503.8 1570.3 1634A 1698.3 1761.2 (705 C4) s C1276 0 2902 0.4317 0.5592 0.6768 0.7994 0.9708 1.3515 1.4300 1.4866 1.5335 1.5/49 14126 1.6477 1.6806 l

e 0 0160 0 0164 0.0172 0.01E3 0 0199 0.0225 0.0307 0 1364 0 1764 0.2066 0 2326 0.2563 0.2784 0.2995 0.319P.

3500 6 77.2 176.0 276.2 3791 487.6 608 4 779.4 1224 6 1338 2 1422.2 1495 5 1563.3 1629.2 1693 6 17b7.2 s

0.1274 0.2899 0 4312 0 5585 0.6777 0.7973 0 9508 1.3242 1.4112 1.4709 1.5194 1.5618 1.6002 1.635S 1.6691 l

0 0159 0.0164 0.0172 0.0182 0.0198 0 0223 0 0237 0 1052 0.1463 0.1752 01994 02210 02411 0.2601 0.2783 i

1 4000 6 78.5 177.2 277.1 379.8 487.7 606 5 763 0 1174.3 1311.6 1403 6 1481.3 1552.2 1619.8 1655 7 1750 6 e

D1271 02m93 0.4304 0.5573 0.6760 0 7940 0.9343 1.2?54 1.3807 1.4461 1.4976 1.5437 1.5812 1.6177 1.6516 j

e 0 0159 0.0164 0 0171 0 0181 0.01 % 0 0219 0.0268 0.0591 0 1038 0.1312 0.1529 0.1718 0 1S*0 0 2050 0.2203 5000 4 El 1 179 5 279.1 381.2 488.1 604 6 746.0 1042.9 1252.9 13646 1852.1 1529.1 1600 9 1670.0 1737.4 s

0.1265 0.2861 0 4287 0.5550 0 6726 0.7880 0.9153 1.1593 1.3207 1.4001 1.4582 1.5061 1.5481 1.5663 1.6216 i

e 0 0159 0.0163 0 0170 0.0160 0 0195 0.0216 0.0256 0.0397 0.0757 0.1020 0.1221 0.1391 0.1544 0.!684 0.1817 6000 &

83.7 181.7 281.0 362 7 AP8 6 602 9 - 736 1 9451 !!S8.8 1323 6 1422.3 1505 9 15620 1654.2 17242 s

0.1258 0.2670 0 4271 0 5528 0.6693 0 7826,0.9026 1.0176 1.2615 1.3574 1.4229 1.4745 1.5191 1.5593 15 %.2 l

e 0.0158 0.0163 0 0170 0.0180 0 0193 0.0?!3 0.0248 0.0334 0 0573 0 031 A 0.1004 0.1160 0.129E 0.1424 0.1542 7000 4 86.2 1844 283 0 3ts4 7 489 3 601.7 729 3 901.8 1124.9 12813 1392 2 1482.6 15631 1639 6 1711.1 e

0 1252 0 2859 0 4256 0 5'.07 06563 0 7/77 0 8926 1.0350 12055 1.3 D 1 13904 1.44b6 14938 1.53'S 1.5735 i

i TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE) (CONTINUED) l A.6

is is 34 is n

in i

to si as v3 go 4%I NJ)M 8" 8"

y

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TA I

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, Entropy. Stu110. F FIGURE A.5 MOLLIER ENTHALPY-ENTROPY DIAGRAM l

A.7 i

l LESSON NOTES Page 3_79 Lrsson oUn,INE 11 0 10 0 h

90 1

8 0

80 c

w 70 T

i 60 j

q FORCED CIRCUL ATION L 50 cfw 3

oa 40 30 NATURAL CIRCULATION 20 10 0

O 10 20 30 40 50 60.70 80 90 100 110 120 CORE FLOW (7.)

Figure 3-JJ.

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4 8

12 16 i

REACTOR PRESSURE TURBINE STEAM FLOW IPSIGI (a108 tte/ht)

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l

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,,,,---~

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07/09/84 Less:n Plan 50

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FIGURE 2.'A3 COOLING' TOWER ARRANGEMENT

- ~

l

- Mcssh -~

Al 1.0 ANSWERS 1.1 a.

3 b.

4 c.

2 d.

1 (0.5 ea/2.0)

Ref: BFN Neutron Multiplication LP, P.1!

1.2 a.

F3, L2 b.

F1, L3 c.

F2, L1 (0.25 ea/1.5)

Ref:

BFN MCD, BWR Thermal Hyd. Review 1.3 (d)

Ref: BFN BWR Thermal Hydraulics LP, P.6 1.4 (d)

Ref: BFN Neutron Slowing Down and Diffusion LP, P.2-3 1.5 (c)

Ref: BFN Fluid Flow / Pump Char., Pump Heat, Pump Laws, P.4, 6 1.6 (a)

Ref: BFN Atomic and Nuclear Physics, P.16, Fig. 8 1.7 (c)

Ref: BFN Mitigating Rx Core Damage LP, P.6 6-9 1.8 (d)

Ref: BFN Neutron Multiplication LP. P.10 1.9 (b)

Ref:

BFN Control Rod Worth LP, P. 4-5 1.10 (d)

Ref:

BFN Phases of Matter LP, P.7 Steam Tables 1.11 (d)

Ref: BFN Subcritical Multiplication LP, P.5-7 s-.

A2 1.12(c)

Ref: BFN Atomic and Nuclear Physics LP, ?. 13, 20 1.13 (a)

Ref: BFN MCD, BWR, LP, P.17-18 1.14(d)

Ref: BFN Xenon and Samarium LP 1.15(a)

Ref: BFN Rx Power and Period LP, P.2 BFN Neutron Slowing Down and Diffusion LP, P. 4-8 1.16 (a)

Ref: BFN Reactivity Coefficients LP, P.2 1.17 (d)

Ref: BFN Entropy LP BFN Energy, Power and Enthalpy LP 1.18 (b)

Ref: BFN Phases of Matter LP 1.19 (b)

REF: BFN Reactor Heat Balance LP 1.20 (c)

Ref: BFN Applic. of the G.E. Eqn. LP, P. 9-10 BFN MCD BWR Thermal Hyd. Review LP. P.5 1.21 (c)

REF: BFN Mitigating Rx Core Damage LP, P.3-7, 8 1.22 Core flow increases due to increased voiding / buoyancy

[0.75] but then stabilizes as increased pressure drop cancels out the NC driving head [0.75]

(1.5)

REF: BFN Mitigating Rx Core Damage, P.3-77-80 L

)

.1 A3

'1.23 a.

1000 psia b.

800'F c.

255'F (0.5 ea/1.5)

Ref: BFN Mollier Diagram LP, P.6 i

1.24 a.

Increase D.

90070000 Increcoc c.

Decrease d.

Increase (0.5 each /2.0)

REF: BFN Fuel Element Temp. Profile LP, P.4-5 (m t m.~. --, s. -. a 1.25 a.

"B" Recirc Pump runback [0.25] at <20% TFW [0.25]

b.

Higher power increases pressure c.

Recire Pump (Suction) flow increases d.

Void collapse due to scram e.

FWC response to level decrease [0.25] and steam flow increase [0.25]

(0.5 each /2.5)

REF: Browns Ferry Transients, #5-e t

l Y

1

A4 2.0 ANSWERS 2.1 (d)

REF: BFN LP #9, p. 14 2.2 (d)

REF:

BFN LP #13, p. 8,12 2.3 (a)

REF:

BFN LP #11, p. 16, 41, 46-49 2.4 (b)

REF:

BFN LP #38, p. 11 2.5 (c)

REF: BFN LP #5, p. 35 2.6 (a)

REF: BFN LP #16, p. 14-15 2.7 (b)

REF: BFN LP #18, p. 5, 6, 11, 12, 14 2.8 (a)

REF: BFN LP #5, p. 10-13 2.9 -(d)

REF: BFN LP #38, p. 4 2.10 (b)

REF: BFN LP #42, p. 9-11, 17, 26 2.11(b) e' REF: BFN LP #13, p. 17 2.12 (a)

~ '

REF: BFN LP #30, p. 5, 6, 12, 17, Day,. -2.13 (c)

REF: BFN LP #16, p. 26, 32

A5 2.14(a)

REF: BFN LP #15, p. 13-14 BFN LP #14, p. 14-15 2.15 (d)

REF: BFN LP #12, p. 10-11, 25 2.16 (a)

REF: BFN LP #9, p. 7-8, 31 2.17-(d)

REF: BFN LP #11, p. 13-15 BF-0I-2, p. 15 2.18 (b)

REF: BFN LP #7, p. 8-9 2.19 (c)

REF: BFN LP #39, p. 7, 13, 15 2.20 Normal Shutdown Bus 1 Alternate 1 Shutdown Bus 2 Alternate 2 Diesel Generator B Alternate 3 Shutdown Bd. 3EB (0.5 ea/-2.0)

REF: BFN 0I-57, p. 105 2.21 a.

4 b.

3 c.

5, 6 d.

1 e.

1 (0.5 ea/ 3.0)

REF: BFN LP #44, p. 17-24 BFN LP #43, p. 6 2.22l(c)

REF: BFN LP #47, p. 5

' 2.23 (d)

REF: BFN LP #50, p. 13-14

+

.u n

..,n,

...J

A6 2.24 A3,183, C3, 03 (0.5)

C1, D1 (0.5)

Control Room (0.25)

-A1, B1 (0.5)

Locally (0.25)

REF: - BFN LP #51, p.11 2.25 a.

FALSE (0.5) b.

TRUE (0.5)

REF: BFN LP #54, p. 5-6, 15 2.26 (b)

REF: BFN LP #40, p. 10, 20, 22 4

-BFN 0I-71, p. 9 I

k u

J

f A7 3.0 ' ANSWERS 3.1' (b)

REF: BFN LP #29, p. 22-23 3.2_-(c)

REF: -BFN LP #8, p. 7-8 3.3 (b)

REF: BFN LP #21, p. 4-6 3.4 (a)

REF: BFN LP #19, p. 23 3.5 (d)

REF: BFN LP #14, p. 10 3.6 (d)

REF: BFN LP #33, p.6-7 3.7 (c)

REF: BFN LP #14, p. 11, 19 3.8 (a)

REF: BFN LP #35, p.6-7 3.9 (d)

REF: BFN LP #19, p. 8, 17, 19, 23 3.10.(b)

REF: BFN LP #17, p.'9 3.11 (b)

REF: BFN LP #29, p. 16-18 3.12 (c)

REF: BFN LP #33, p. 4-5 3.13 (a) u 4)

REF: BFN LP #25, p. 25-26

~

A8 l

h w -3.14 (2)

REF: BFN AC Elec. Dist. Rev. 1, p. 10-11 3.15 (c)

REF: BFN LP #38, p. 19-20 3.16 (d)

REF: BFN LP #20, p. 12 GOI-100-1, p. 15 BFN LP #28, p. 13 3.17 (c)

REF: BFN LP #25, p. 17 BFN LP #7,_ p. 12 BFN LP #28, p. 19 BFN LP #24, p. 10 3.18 (b)

REF: BFN LP #20, p. 4 BFN LP #19, p. 7 3.19 (c)

REF: BFN LP #21, p. 10-11 3.20 (a)

REF: BFN LP #42, p. 25-26 3.21 (d)

REF: BFN LP #3 3.22 (b)

REF: BFN LP #8, p. 36 BFN LP #7, p. 12-13 3.23 (a) 3.24 (a) 3.25 (d)

REF: BFN LP #12, p. 21-24 BFN Transients #2

A9

.f 3.26 APRM "A" fails upscale relay SA-K12A deenergizes [0.5]

- letS centacts 5A-K12A in RPS Trip Logic Al c::en [0.5]

+ relays 5A-K14 A & E ceenergize [0.5]

+ centacts 5A-K14 A & E c;:en [0.5]

+ scras pilct sciencid valves for RPS A deenergize and vent [0.5]

(2.5)

REF: BFN LP #28 3.27 False (0.5)

REF: BFN E D, BWR LP, p..7-8 3.28 a.

03 b.

04 22-55 15-03 46-55 46-55 34-27 BLANK (0.25 eacn/2.0) 4

~

REF: 5FN LP #24, p. 24-25 l

1 r

e i

V

4 A10 4.

ANSWERS 4.1 (b)

REF:- BF-01-24, p. 2-3 4.2 (d)

REF: 10 CFR 20 tgyAx.

43 (e)

REF: BF-01-3, : p. 4 4.4 (d).

REF: BF-0I-47, p. 21, 38, 41, 44 4.5 (d)

REF: BF-01-1, p. 9 4.6 (a)

REF: BF-0I-66, p. 5-7 4.7 (b)

REF: BF-0I-3, p. 2-2A 4.8 (c)

REF: BF-G01-100-1, p. 9-11 4.9 (c)

REF: BF-GOI-100-1, p. 19-21 BF-TS, p. 4 4.10 (a)'

REF: BF-OSIL #43 - e. 4

.4

I All 4.11 (a)

REF: BF-0I-68, p. 28-29 4.12 (b)

REF: BF-GOI-100-12, p. 5-6 4.13 (d)'

REF: BF-G01-100-1, p. 22 4.14(c)

REF: BF-G0I-100-3, p. 84 4.15 - Torus temperature >105'F Torus water level <-6.25" or >-1" Drywell temperature >160*F Drywell pressure >2.45 psig (0.5 ea./2.0)

REF: BF-E01-2, p. 3 4.16 a.

50'F/hr.

b.

215*F c.

90*F/hr.

d.

reduce the On and hydrogen peroxide content of the coolant.

(0.5 ea/2.0)

REF: BF-GOI-100-1, p. 15 4.17 To produce sufficient pump head to create downthrust on the pump -[0.5] and prevent serious damage to the Kingsbury thrust bearing in the pump motor [0.5].

(1.0)

REF: BF-0I-27, p. 2

A12 4.18.1.

Check all rods fully in.

2.

Close MSIVs.

3.

Trip feedwater pumps.

4.

Trip main turbine.

5.

Trip condensate pumps.

I 6.

Trip condensate booster pumps.

7.

Start diesel generators.

(2.0)

REF: BF " Control Room Abandonment", p. 2 4.19 - CST charging water to RHR & CS SLC System - Suction from test tank RHR crossties to other units Standby coolant RHR drain pumps CS drain pumps (4 of 6 9 0.5/2.0)

REF: BF E01-1, p. 24 4.20 - Five or more adjacent rods not inserted below 06 position [0.5] or Thirty or more rods not inserted below 06 position [0.5] and either Rx water level cannot be maintained [0.5] or suppression pool water temperature limit of 110*F is reached [0.5]

(2.0)

REF: BF-E01-47, ' p. 3 4.21 - When another licensed operator is present [0.25] with an unobstructed view of panel 9-5 [0.25]\\ opmar Ch.i].

In a Emergency [0.25]g c.aAker-

• • nukg road them ce to attend to.or verify alarms and initiate various corrective actions-[0.25]

(1.0)

REF: BF Standard Practice 12.5, p. 4 g.

I A13 4.22 - If a Rx scram has not been initiated, initiate a Rx Scram.

Restore and maintain Rx water level between +11 and +54" if possible.

If Rx water level cannot be restored and maintained above

+11" Then maintain Rx water level above -114.5" (on LI -3

-46A & B)

If the ADS timer has initiated, then prevent depressurization by resetting the ADS timer.

To prevent MSRVs from cycling when RPV pressure approaches 1105 psig, manually open a sufficient number of MSRVs to reduce Rx pressure to ~950 psig.

(0.5 ea/2.5)

REF: BF-E0I-1, p. 6 4.23 -

Immediately dispatch an operator to the BCC and establish communications...

If reactor level decreases to -114.5" below inst. zero perform a rapid depressurization of the reactor using any six (6) availaole MSRVs...

Restore and maintain reactor water level with available makeup systems...

(0.5 ea./1.5)

REF: BF-GOI-100-11, p. 10 l

1 j

s,

- AAAsrErt -

caf '

ENCLOSURE 3 U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility:

Browns Ferry Reactor Type:

BWR-4 Date Administered: April 9, 1985 Examiners:

S. Guenther Applicant:

INSTRUCTIONS TO APPLICANT:

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

Staple question sheets on top of the answer sheets.

Points for each question are indicated in parentheses after the question.

The passing grade requires at least 70% in each category and a final grade of at least 80%.

Examination papers will be picked up six (6) hours after the examination starts.

Category % of Value Total Score Value Category 2v.t 30 24 fr 5.

Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics 3029 24.84 6.

Plant Systems:

Design, Control &

Instrumentation L

JFt3 25.6-7.

Procedures-Normal, Abnormal, Emergency Radiological Control

15. t 30 34 8.

Administrative Procedures, Conditions and Limitations 12 ins Totals Final Grade All work done on this exam is my own, I have neither given or received aid.

Applicant Signature 4

o(

.s*

1 f

5.0 Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics 5.1 The attached figures (21A & 218) represent a transient that could occur at a BWR.

Given:

(1) A break of the "B" Main Steam line down-stream of the flow restrictor at time

=.4 min.

(2) No operator actions are taken (3) Recorder speed of I division = 1 minute Explain the cause(s) of the following recorder indica-tions:

(3.0)

Core flow DECREASE (Point 1) a.

4 b.

Core Flow TURNS (Point 2)

Core Flow INCREASES (Point 3) c.

d.

Reactor vessel level INCREASE (Point 13)

Total FW flow INCREASE (Point 20) e.

f.

Turbine Steam Flow DECREASE (Point 8) 5.2 The attached figures (SA & 5B) represent a transient that could occur at a BWR.

Given:

(1) The "A" recirculation pump speed controller fails high at time T= 1.2 min (2) No operator actions are taken (3) Recorder speed of I division = 1 minute Explain the cause(s) of the following recorder indica-(2.5) tions:

Core flow DECREASE (Point 3) a.

b.

Reactor Pressure INCREASE (Point 6)

Reactor vessel level DECREASE (Point 13) c.

d.

Reactor vessel level DECREASE (Point 14)

Total FW flow INCREASE (Point 17) e.

l

ih 04 2

5.3 Answer the following with regard to the attached

" Periodic NSS Core. Perform ~ance Log" (P-1),

Which of the following most accurately reflects the a.

actual reading (in % power) on APRM "C"?

(1.0)

(1) 62.6%

(2) 63.9%

.(3) 65.2%

(4) 67.9%

b.

The maximum LHGR in the core is....

(1.0)

(1) 6.67 kw/ft (2) 7.66 kw/ft (3) 12.10 kw/ft (4) 13.40 kw/ft Draw a sketch of the relative axial neutron flux shape c.

based upon the attached P-1 printout.

(1.0) l 5.4 Considering each of the following situations independ-ently (i.e., no other parameters change) STATE whether that condition would tend to INCREASE, DECREASE, or HAVE NO EFFECT on fuel centerline temperature.

(2.0)

A fuel bundle develops a zirc-oxide corrosion layer.

a.

b.

A new fuel bundle is initially exposed.

The pressure setpoint on EHC is lowered by 10 psig.

c.

d.

The plant is operated with 1 MSIV shut.

a o

3 5.5 Water, the primary coolant used in a BWR, can exist in three phases. Which'one of the following is correct concernino the phases of water?

(1.0) a.

The Latent Heat of Vaporization is the amount of heat required to change one Ibm of water at 32*F to one 1bm of steam at 212 F.

b.

It is impossible to raise the temperature of liquid water at atmospheric pressure above **'*F.

c.

The Latent Heat of Vaporization increases as System pressure is raised from atmospheric to normal operating pressure.

d.

The condenser removes sensible heat from the LP turbine exhaust in changing the saturated steam to saturated liquid.

5.6 Significant quantities of hydrogen gas may be generated during, or subsequent to, a Loss of Coolant Accident.

This hydrogen gas...

(1.0) a.

... is generated primarily by the radiolytic decomposition of water, with much smaller quantities generated by the zircaloy water reaction.

b.

... is generated primarily by the zircaloy water reaction if fuel clad temperatures are allowed to exceed 2000*F.

c.

... could cause an explosive hazard in the drywell if allowed to exceed the lower flammability limit of 10% in air.

d.

... is of little concern since the containment is inerted and no oxygen should be present in the post-LOCA containment atmosphere.

,e l

4 5.7 Which one of the following is an example of CONVECTIVE heat transfer in a BWR core?

(1.0) a.

Heat transfer across the helium gap in a fuel rod.

b.

Heat transfer across the laminar coolant boundary layer surrounding a fuel rod.

c.

~ Heat transfer in the bulk coolant stream.

d.

Heat transfer across the fuel pellet.

5.8 Which of the following statements correctly describes the behavior of Xenon and Samarium?

(1.0) a.

After a reactor trip occurs, Xenon concentration initially INCREASES and Samarium initially DECREASES.

b.

After a reactor trip occurs BOTH the Xenon and Samarium concentrations will eventually DECAY TO ZERO.

c.

The Xenon and Samarium peak concentrations following a trip occur at a time INDEPENDENT of the previous power history, d.

Equilibrium Xenon concentration INCREASES with increasing power levels, but the equilibrium Samarium concentration remains the same.

5.9 Attached figure 3-JJ, " Steady-state Natural Circulation Characteristics of a BWR,"

illustrates how " core flow" changes with respect to " reactor power" without forced circulation.

EXPLAIN why incremental increases in power initially produce very rapid increases in core flow, but eventually reach a point where further power increases produce no increase in core flow?

(1.5) i

=.

.e l

5 5.10 The attached T-S diagram (fig. 5.10) approximates the " thermodynamic cycle" employed at the Browns Ferry Nuclear Plant.

Which one of the following expr'essions could be used to a.

calculate this thermodynamic cycle's efficiency?

(1.0)

(1) n = (area within) a - b - c - d - a (area within) a - d - f - e - a (2) n = (area within) e - a - b - c - d - f - e (area within) a - b - c - d - a (3) n=

(area within) a - d - f - e - a (area-within) e - a - b - c - d - f - e (4) n=

(area within) a - b - c - d - a (area within) e - a b - c - d - f - e b.

Which of the following statements is correct regarding Browns Ferry's thermodynamic cycle efficiency?

(1.0)

(1) The cycle is more efficient at 25% power than at 100% power.

(2) Increasing condensor vacuum from 25" to 29" will DECREASE cycle efficiency.

(3) Reduced feedwater heating will DECREASE cycle efficiency.

(4) Decreasing condensate depression will DECREASE cycle efficiency.

4 1

i

6 5.11 A reactor heat balance ~was' performed (by hand) during the 00-08 shift due t'o the Process Computer being OOC.

Which of the following statements is correct concerning reactor power?

(1.0) a.

If the feedwater temperature used in the heat balance calculation was LOWER than the actual feedwater temperature, then the actual power is HIGHER than the currently calculated power.

b.

If the reactor recirculation pump heat used in the heat balance calculation had been OMITTED, then the actual power is LOWER than the currently calculated power.

c.

If the steam flow used in the heat balance calculation was LOWER than the actual steam flow, then the actual power is LOWER than the currently calculated power.

d.

If the RWCU return temperature used in the heat balance calculation was HIGHER than the actual RWCU return temperature, then the actual power is LOWER than the currently calculated power.

5.12 Which one of the following statements best describes the operating characteristics of an LPRM detector?

(1.0) a.

Depletion of the detector's uranium coating causes neutron sensitivity and the neutron to gamma signal ratio to decrease significantly as the detector ages; this may impair the detector's ability to respond to changes in power.

b.

Since the detector functions as an ionization chamber and argon _ gas pressure remains constant, both neutron and gamma sensitivity, as well as the neutron to gamma signal ratio, remain relatively constant as the detector ages.

c.

Depletion of the detector's uranium coating causes both neutron and gamma sensitivity to decrease with detector age; however, the resulting neutron to gamma signal ratio remains relatively constant.

d.

Depletion of the detector's uranium coating causes a decrease in the neutron to gamma signal ratio as the detector ages, but this does not impair the detector's response since gamma discrimination is not required in the power range.

r_-

. =

7 5.13 The fission process in a commercial reactor requires that the neutrons " born" by fission be "thermalized." The interaction in the reactor core which is most efficient in thermalizing neutrons for fission occurs with which of the following?

(1.0) a.

OXYGEN atoms in the water molecules b.

BORON atoms in the control rods c.

HYDROGEN atoms in the water molecules d.

ZIRCONIUM atoms in the fuel cladding 5.14 A significant amount of excess reactivity must be loaded into a core at BOL so that 100% power can be attained at the end of an 18-month fuel cycle.

From Column "B" select the correct reason for each positive reactivity requirement listed in Column "A".

(2.0)

Column "A" Column "B"

(+) AK/K Reouired Reason a.

1.0%

1.

Moderator Temperature Increase b.

3.0%

2.

Void Fraction Increase c.

3.8%

3.

Samarium Buildup d.

4.77%

4.

Xenon Butidup 5.15 Which one of the following reactions is the most significant intrinsic neutron source in a highly exposed core?

(1.0) a.

The high-energy gammas from the decay of fission fragments which cause increased spontaneous fission of Uranium and Plutonium.

b.

The significant increase in the Deuterium concentration which causes a concurrent increase in the Photo-Neutron source.

c.

The Alpha particles emitted from the decay of unstable fission fragments which cause a concurrent increase in the Alpha-oxygen reaction.

d.

The buildup of transuranic elements such as Curium-242 and Curium-244 which fission spontaneously.

~

8 5.16 Which one of the following is correct for constant speed centrifugal pumps?

(1.0) a.

The volume flow rate is directly proportional to the square of the pump speed.

b.

As.the volume flow rate increases, the pump head increases.

c.

For two centrifugal pumps in parallel, the combined flow rate for a given head is equal to the sum of the individual capacities of the two pumps at that head.

d.

For two centrifugal pumps in series, the combined flow rate for a given head is equal to the sum of the individual capacities of the two pumps at that head.

5.17 Control Rod Worth...

(1.0) a.

increases as moderator temperature decreases.

b.

increases as K excess increases.

c.

increases as core void content increases.

d.

increases as fuel temperature decreases.

5.18 A BWR is operating at 985 psig with 217*F of.feedwater subcooling. What is the feedwater temperature?

(1.0) a.

321.5*F b.

323.8*F c.

-325.7*F i

d.

327.5*F l

{

9

~

5.19 Which of the following is NOT a characteristic of subcritical multiplication?

(1.0) a.

The subcritical neutron level is directly proportional to the neutron source strength, b.

Doubling the indicated count rate by reactivity additions will reduce the margin to criticality by approximately one half.

c.

For equal reactivity additions, it takes longer for the new equilibrium SRM count rate to be reached as Keff approaches unity.

d.

A single notch of rod withdrawal will produce an equivalent equilibrium count rate increase whether Keff is 0.88 or 0.92.

5.20 Which of the following in NOT correct concerning heat exchangers?

(1.0) a.

Heat transfer is by both the conductive and convective methods of heat transfer.

b.

The heat transfer rate is directly proportional to the heat transfer coefficient associated with the material from which the tubes are made, c.

The heat transfer rate for a parellel flow heat exchanger is higher than that of a counter flow heat exchanger given the same inlet temperatures.

d.

Most of the heat exchangers used at Browns Ferry are of the counter flow type.

~

10 5.21 Which one of the following statemen'ts is NOT correct regarding the Average Planar Linear Heat Generation Rate (APLHGR) thermal limit?

(1.0) a.

The limit grows continually more restrictive from BOL to E0L.

b.

Maintaining a MAPRAT less than 1.0 ensures that the peak cladding temperature will not exceed 2200 *F during a design basis LOCA.

c.

The average planar linear heat generation rate is limited to ensure proper radiative heat dissipation after a LOCA.

d.

The limiting value of MAPLHGR varies as a function of fuel type as well as fuel exposure.

WRITE "END OF SECTION 5" AND START A NEW ANSWER PAGE i

11 6.

Plant Systems Design, Control, and Instrumentation 6.1 Which of the following series of components most accurately reflects the flow path that the water pumped from the Drywell Equipment Drain Sump (DWEDS) would traverse on its way to the discharge canal?

(1.0) a.

DWEDS + Waste Collector Tank + filter + demineralizer + sample tank + discharge canal via Floor Drain Collector System rad.

monitor, b.

DWEDS + Floor Drain Collector Tank + filter + evaporator

+ sample tank + discharge canal via Waste Collector System rad, monitor.

c.

DWEDS + Waste Collector Tank + sample tank + discharge canal via Floor Drain Collector System rad. monitor.

d.

DWEDS + Radwaste Evaporator System + sample tank + discharge canal via Waste Collector System rad. monitor.

6.2 Which one of the following statements is correct regarding the Rod Block Monitor (RBM) system channel "A" nulling process?

(1.0) a.

A Trip Inhibit signal is generated that inhibits all RBM system trips during the nulling sequence.

b.

A " Nulling" INOP trip signal is generated which blocks rod movement but will not activate an annunciator alarm on the 9-5 panel.

c.

Bypassing the "A" APRM will reset the null sequence clock and initiate a new nulling sequence.

d.

Depressing the " Set Up" push button when the selected trip level is reached initiates a new nulling sequence.

12

)

6.3 Which of the following statements is correct regarding the Reactor Manual Control System "CRD Notch Override" switch?

(1.0) a.

It must be held'in the " Notch Override" position while the "CRD Control" switch is held in the " Rod In" p'sition if o

continuous rod insertion is desired.

b.

If held in the " Emergency Rod In" position, it will bypass all rod insert blocks except those imposed by the Rod Sequence Control System (RSCS).

c.

If held in the " Emergency Rod In" position, it will illuminate an amber light above the switch.

d.

When used for emergency rod insertion, it bypasses the automatic sequence timer and acts directly on.the directional control valves.

6.4 Which one of the following statements correctly describes the per-formance of a Control Rod Drive during a reactor scram?

(1.0) a.

Both the scram inlet and outlet valves for each rod /HCU will open simultaneously to initiate rod insertion.

b.

A control rod's scram insertion time can be adjusted by throttling the insert throttle valve for that HCU.

c.

At 300 psig reactor pressure a control rod will scram success-fully even if its accumulator is discharged or isolated.

I d.

There is no digital rod position readout displayed until after-the scram is reset or the SDV is full.

For each of the following Feed Water Control System (FWCS) malfunctions (questions 6.5 through 6.6) select the correct system / plant response from the list (a-d) which follows. An answer may be used more than once, and no operator actions are taken.

a.

Reactor water level decreases and-stabilizes at a lower level, b.

Reactor water level decreases and initiates a reactor scram.

c.

Reactor water level increases and stabilizes at a higher level.

d.

Reactor water level increases and initiates a turbine trip (w/ scram).

I 6.5 The plant is operating at 100% power in 3-element control when one staam flow input signal is lost.

(1.0)

)

6.6 Thu plant is operating at 100% power in 3-element contol when one SRV inadvertently opens (and remains open.)

(1.0) 1 l

13 6.7 The reactor is critical at approximately 10 psig and the "Heatup and Pressurization" phase of GOI-100-1 is being performed. The NR GEMAC LI's in the control room read the following " approximate"

~

values:

(1.0)

NR GEMAC A 37" NR GEMAC B 38" NR GEMAC C 37" The two Emergency System / Accident (Yarway) control room indicators should read approximately...

a.

.. 0 inches b.

15 inches c.

.. 38 inches d.

.. 60 + inches 6.8 A LOCA signal is received with a concurrent loss of offsite power.

The diesel generators reach operating speed and voltage and their output breakers close at time t=o.

Which one of the following major components is matched with its correct load sequencing time?

(1.0) a.

Core Spray Pump 1A at time t=o sec.

b.

RHR pump 18 at time t=7 sec.

c.

RHRSW pump Al at time t=14 sec.

d.

Drywell Blower at time t=21 sec.

6.9 Which one of the following signals will cause the HPCI turbine to trip but will NOT cause a HPCI system isolation?

(1.0) a.

High HPCI turbine exhaust pressure @ 150 psig.

b.

Low reactor pressure 9 100 psig.

c.

High HPCI area temperature @ 200*F.

d.

HPCI turbine exhaust inboard disc rupture 9 10 psig.

14 6.10 The main turbine is at 1800 rpm in preparation for synchronizing the main' generator to the grid (i.e.

the 500kv generator breakers are still open). What will happen if the "All Valves Closed" pushbutton is depressed?

(1.0) a.

Nothing will happen since the synchronous speed select signal is sealed in.

b.

The four turbine control valves and four main stop valves will close, but the intercept valves will remain open.

c.

The ten control valves (TCVs and IVs) and four main stop valves will close.

d.

The ten control valves (TCVs and IVs) will close, but the four main stop valves will remain open.

6.11 Which one of the following statements is NOT correct regarding operation of the Rod Sequence Control System (RSCS) while in the Group Notch Control (GNC) logic mode?

(1.0) a.

When the first rod in a group is selected, its rod select pushbutton will illuminate brightly and the campanion rods in that group will have their select PB dimly backlit, b.

The GNC logic will prevent any rod in the group from being moved more than one notch in the direction of movement of the first rod in the group.

c.

When the last rod in.a group is moved (all rods in same position) all dim backlights for that group will extinguish, d.

The GNC logic will allow selection and movement of ANY group as long as all the rods within that group remain within one notch.

6.12 Which of the following statements correctly describes the operation of the SRMs during a reactor startup?

(1.0) a.

They are initially fully inserted to their upper electrical stop which is 30" above the center of the active fuel.

b.

An intericek prevents detector withdrawal unless the SRM count rate is above 100 cps.-

c.

All SRM rod blocks are automatically bypassed when all IRM range switches are on 7 or above.

d.

The reactor period indication is still valid with the SRMs in the fully withdrawn position, f

15 6.13 The Reactor Water Cleanup System (RWCU) is being operated 4

in the " Hot Blowdown" mode to control reactor water level.

Which one of the following limitations is NOT a valid concern to the operator while controlling blowdown flowrate to the main condenser?

(1.0) a.

Exceeding the maximum allowable RBCCW system temperature exiting the NRHX.

b.

Exceeding the maximum allowable cleanup system temperature exiting the NRHX.

c.

Exceeding the FCV (#15 valve) downstream pressure setpoint.

d.

Exceeding the design cooling capacity of the regenerative heat exchanger.

6.14 Diesel generators A-D have started and are running in response to a high drywell pressure signal. Which one of the following conditions will trip the diesel generator (s)?

(1.0) a.

Overcurrent b.

High differential current c.

Loss of field d.

Reverse Pcwer 6.15 Backup scram valves provide a redundant means of venting air from scram pilot valves and scram' dump valves. Which of the following correctly describes the design / operation of the Backup Scram valves?

(1.0) a.

They are normally energized and will deenergize upon a Reactor Protection System Scram signal.

b.

They are aligned such that two valves in series, one from each RPS trip system, must actuate to vent the scram air header.

c.

They are designed such that both RPS channels must trip in order for any one of the valves to actuate.

d.

They are powered from the RPS buses A and B.

1 16 6.16 Which of the following correctly describes how the DRYWELL is prevented from exceeding its design external pressure?

(1.0) a.

" Suppression chamber - drywell" vacuum breakers vent air from the suppression chamber to the drywell at 0.5 psid.

b.

" Suppression chamber - drywell" vacuum breakers vent air from the drywell to the suppression chamber at 0.5 psid.

c.

"Drywell - reactor building" vacuum breakers vent air from the reactor building to the drywell at 0.5 psid.

d.

" Suppression chamber - reactor building" vacuum breakers vent air from the reactor building to the suppression chamber at 0.5 psid.

' NA

+-tf-The Containment Atmosphere Dilution (CAD) system is used q

as a post-accident method of hydrogen control. Which of the following correctly describes its design / operation?

(1.0) a.

It is automatically initiated when drywell pressure reaches the "2 psig" ECCS initiation setpoint.

b.

Units 1 and 2 share two redundant trains while Unit 3 has two independent trains.

c.

It uses the SBGTS to vent gases from containment while adding N2 to keep the 02 concentration less than 5%.

d.

It shares its nitrogen supply with the Containment Vent, 4

Purge and inerting system.

6.18 The reactor recirculation pump seal cartridge assemblies consist of two sets of sealing surfaces and a breakdown bushing assembly.

Failure of the #2 seal assembly at rated conditions would result in which of the following indications?

(1.0) a.

An INCREASE in #2 seal cavity pressure from approxi-mately 500 psig to approximately 1000 psig.

b.

A DECREASE in #2 seal cavity pressure from approximately 500 psig to approximately zero psig.

c.

An INCREASE in #1 seal cavity pressure from approximately 500 psig to approximately 1000 psig.

d.

A DECREASE in #1 seal cavity pressure from approximately 500 psig to approximately zero psig.

o 17 6.19 Which one of the following statements correctly describes the operation of the Standby Liquid Control System (SBLC) during initiation / injection?

(1.0) a.

SBLC initiation will shut the RWCU inlet isolation valves but NOT the outlet isolation valves.

b.

As each squib valves fires, its continuity monitor light will illuminate, indicating that the valve has opened c.

Taking the keylock control switch to either position will start the associated pump and fire both squib valves.

d.

Indicated flow rate for two pum;i operation should be approximately 100 gpm.

6.20 The Circulating Water system is being operated in the HELPER mode. Which one of the following (a throught d) lists the correct gate lineup for this operating mode?

(Figure 2.23 is enclosed for reference.)

(1.0)

(a)

(b)

(c)

(d)

Gate #1 Open Closed Open

',0ren Gate #1A Open Closed Closed Closed Gate #2 Closed Open Open Closed Gate #3 Open Closed Open Open 6.21 Which one of the following RBCCW loads is considered an ESSENTIAL load?

(1.0) a.

Drywell floor drain sump heat exchanger f

b.

Non-regenerative heat exchanger

[

c.

Drywell atmospheric coolers i

d.

Reactor water cleanup pump seal water coolers 6.22 Which one of the following statements is correct with regard to the RCIC turbine trip throttle valve?

(1.0) a.

The valve is normally closed during standby conditions.'

b.

The valve uses a DC motor operator to remotely. reset the valve after a turbine trip.

c.

The valve must be locally reset after an electrical overspeed trip.

d.

If operating, the valve will fail shut upon loss of electrical power to the relay logic bus "A".

i

-4 18 6.23 Unit 1 4KV Shutdown Board B has a normal and three (3) alternate power sources.

Identify the normal source and the three alternate sources of power, in the order of preference.

(2.0) 6.24 Answer the following with regard to the EECW system:

The system is normally in standby with the (which ones)

RHRSW pumps assigned to EECW.

In the event that any of these pumps are out of service, the (which ones) pumps can be valved in from (control room / local) and the (which-ones) pumps can be valved in from (control room / local)

(2.0) 6.25 Unit 1 is operating normally at 100% power when APRM "A" fails upscale and results in a reactor half-scram.

Utilizing the attached RPS trip logic diagrams (Figures 2, 5, & 8) describe in a step-by-step fashion (with regard to the opening / closing, de/ energizing of all applicable contacts and relays) how the APRM upscale trip results in an actuation of the scram solenoid valves.

(2.5) 6.26 What two (2) conditions will cause a refueling rod block if the Mode Switch is in STARTUP?

(1.0) 6.27 TRUE or FALSE: All GM detectors / survey meters at Browns Ferry (e.g., RM-14, Teletectors) will quickly go fullscale and then drop to a lower on-scale reading when exposed to a high radiation field.

(0.5)

WRITE "END OF SECTION 6" AND START A NEW ANSWER PAGE

/

I' I

h I,s i

19 I

7.

Procedures - Normal, Abnormal. Emergency, and Radiological Control 7.1 A Unit 1 plant shutdown and cooldown.are in progress per G01-100-12 and RHR shutdown cooling has been established per 0I-74. The Radiochemical Laboratory...

(1.0) a.

must be notified to begin taking daily samples since the RHRSW on-line process monitors lack the required sensitivity to detect small leaks which could be in excess of 10 CFR 20 release limits.

b.

must be notified to begin taking 4-hour samples since the RHRSW on-line process monitors are inoperable due to sample pump design problems.

c.

must be notified to begin taking 4-hour samples since-the RHRSW on-line process monitors lack the required sensitivity to detect small leaks which could be in excess of 10 CFR 20 limits.

d.

must be notified to begin taking daily samples since the RHRSW on-line process monitors are inoperable due to sample pump design problems.

7.2 Thirty minutes after taking a routine RHRSW sample, the Radio-l chemical Laboratory notifies the control room that-the sample contained radiation levels at 75% of the MPC/ Tech. Spec. release limit.

Per OSIL #19, " Liquid Effluent Monitoring for. RHRSW, RBCCW, and RCW," you must...

(1.0) a.

immediately implement the' Abnormal Operations section of 0I-23, "Hi Radiation RHRSW HX outlet," and secure RHR Shutdown cooling.

b.

increase the RHRSW sampling frequency and implement the Abnormal Operations section of 01-23, "Hi rad. on RHRSW HX outlet," when the 100% MPC/ Tech. Spec. release limit is reached, c.

increase the RHRSW sampling frequency and decrease reactor pressure / temperature at'the maximum permissible rate in an effort to reduce the RHRHX leak rate.

.d.-

immediately implement the Abnormal Operations section of 0I-23, "Hi Rad on RHRSW HX outlet" and implement IP-2,

" Notification of Unusual Event,"~if the. release. rate increases to the 100% MPC/ Tech. Spec. limit.

)

-k v

N 4

1 20 l

7.3 Which of the following is a 10 CFR 20 exposure limit?

(1.0) a.

1000 mrem / quarter whole body.

b.

5 rem / year whole body.

c.

15 rem /qtr to extremities.

d.

1250 mrem /qtr of beta radiation exposure to the lens of the eye.

hgg 4-A reactor / plant startup is in progress and preparations are being made to start the first Reactor Feed Pump (RFP) g "-

in accordance with BF-01-3, "Feedwater System."

Which of the following lists (a

through d) correctly specifies the conditions which should be established, per 0I-3, prior to rolling the RFP turbine?

(1.0)

(a)

(b)

(c)

(d)

FWCS Mode Switch single-elem.

single-elem single-elem.

3-elem.

l AC TGOP AUTO OFF AUTO AUTO RFP Suction Valve Open Open Open Open RFP Discharge Valve Closed Closed Open Closed RFP Minimum Flow Valve Open Open Open Open RFPT Trip Status On On Off Off (Blue Light) 7.5 A main turbine generator startup is in progress per BF-0I-47.

While conducting system checks at 100 rpm, you receive a turbine High Vibration alarm and a report from the turbine building AVO of a squealing noise coming from the HP turbine.

Which one of the following sets of actions is correct per 01-47?

(1.0)

Check bearing oil flows, temperatures (oil and metal) and seal a.

steam header pressure while maintaining turbine RPM to clear the rub.

b.

Trip the turbine if unable to verify / restore proper oil / seal steam flow and clear the rub / vibration within the allowed 5 minute hold period.

Immediately trip the turbine and verify lift pumps running c.

but do not engage the turning

• gear.

d.

Immediately trip the turbine, break condenser vacuum, verify lift pumps running, and engage the turning gear when the zero i

speed alarm sounds.

1

21 7.6 Which one of the following actions in response to a

stuck open relief valve is INCORRECT, per the Abnormal Operations section of BF-0I-1, " Main Steam System"?

(1.0) a.

Operate the relief valve control switch two or three times in an attempt to close the valve.

b.

Initiate suppression pool cooling to maintain torus temperature below 95 F and to prevent local hot spots.

c.

Change the EHC pressure regulator setpoint in an effort to close the valve.

d.

With the supervisor's approval, manually scram the reactor once torus temperature reaches 125 F.

7.7 A plant startup is in progress and condenser vacuum is being established in accordance with BF-0I-66, "Off-Gas System." What is the proper sequence for component startups per OI-66?

(1.0) a.

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

b.

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

c.

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

d.

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

7.8 Which one of the following is an acceptable method per BF-0I-3, "Feedwater System," for reducing thermal duty on feedwater nozzles during low power and/or hot standby operation?

(1.0) a.

Reduce reactor power to minimize the frequency of cold feedwater cycles.

b.

Reduce reactor pressure and temperature to reduce the differential temperature on the nozzles.

c.

Maximize RWCU blowdown flow in an effort to stabilize feedwater flow control.

d.

Secure RWCU flow to minimize reactor heat load 'and power level.

22 7.9 A plant startup is in progress per BF-GOI-100-1. Under which set of conditions (a through d) would it be permissible to place the mcde switch in the STARTUP position and commence the reactor startup?

(1.0)

(a)

(b)

(c)

(d)

No. of SRMs > 3 cps 3

4 3

2 RHR system mode / status SDC/STBY SDC/STBY LPCI/STBY LPCI/STBY Recirc pumps running A

A&B A&B A&B Licensed Operators 2

2 3

3 present 7.10 Unit I has just been started and synchronized with the grid and power is being raised to rated in accordance with BF-G01-100-1.

Which one of the following operating restrictions / limitations MUST be observed per the GOI?

(1.0) a.

A recirculation pump speed differential of 6 - 8%

(in manual) must be maintained while passing through the recirc pump critical speed zones.

b.

Load changes are prohibited during periods in'which the process computer is inoperable.

c.

When above the PCIOMR threshold but below the envelope, rod movement shall be "notchwise" but recirc flow changes are unrestricted.

d.

Average power during an eight-hour shift shall not exceed 3293 MWt; however, brief power peaks to the design limit of 3440 MWt are authorized.

7.11 Which of the following if NOT a valid valve operation guideline as stated in OSIL #477 (1.0) a.

Gate valves are not recommended for throttling service and should normally be used in either their full-open or full-closed positions.

b.

Always backseat valves (except throttling valves) as tightly as possible, without using a " cheater," to isolate the packing from line pressure.

c.

Wrenches or " cheaters" are not to be used on motor-operated valve handwheels or on any valve with a-gear drive.

d.

Make all efforts not to operate MSIVs when there is not steam flowing-through the valves.

23 7.12 Which of the following is NOT a symptom that you would expect to see as a result of a " Jet Pump Failure" per 0I-68?

(1.0) a.

Decrease in failed jet pump flow b.

Decrease in core differential pressure c.

Decrease in reactor (APRM) power d.

Increase.in indicated core flow 7.13 Per GOI-100-12,

" Normal Shutdown From Power,"

which one of the following is the preferred method of cooling down and depressurizing the reactor with the MSIVs open and the feedwater system available?

(1.0) a.

Reduce the pressure regulator setpoint.

b.

Open BPVs using the bypass valve opening jack.

c.

Bleed steam through the MSL drains.

d.

Bleed steam through the HPCI and RCIC steam line drains.

7.14 Which of the following conditions establishes the reactor in HOT

STANDBY, as defined by BF-GOI-100-1,

" Integrated Plant Operations"?

(1.0)

Parameter Condition (a)

(b)

(c)

(d)

Mode Switch S&HS S&HS S&HS S&HS Coolant Temp 210*F 300 F 553*F 536 F Rx Pressure 0 psig 52 psig 1060 psig 920 psig MSIVs closed open closed closed 7.15 Fuel loading is in progress when you notice an unexplained increase in SRM count rate and an indicated reactor period; you suspect that an inadvertent criticality event is taking place. Which one of the following actions is NOT CORRECT per BF-GOI-100-3, " Refueling Operations"?

(1.0) a.

Manually scram the Rx and verify all rods full in.

b.

If c.riticality is still evident after all rods in, then stop the CRD pump.

c.

Evacuate the refueling floor after withdrawing the fuel bundle which caused the criticality event.

d.

Upon the SE's or ASE's approval initia'te SBLC.

24 7.16 HPCI flow testing is in progress.

List the four (4) entry con-ditions for BF-E01-2, " Containment Control."

(2.0) 7.17 Fill in the Blanks:

BF-GOI-100-1, " Integrated Plant Operations," limits the initial plant heatup rate to (a) *F/hr until the recirculating water reaches (b)

'F at which time the limit is changed to (c) 'F/hr.

.The initial heatup rate limit is established to (d)

(2.0) 7.18 a.

The shift engineer has determined that the control roem must be abandoned due to smoke and directs that the reactor be manually scrammed.

List seven (7) specific (e.g., start, stop, trip, check...) operator actions that should be performed per the procedure, given sufficient time, prior to reporting to the unit backup control center?

(2.0) b.

Had you been unable to scram the reactor prior to abandoning the control room, how should you initiate a reactor scram per the " Control Room Abandonment" procedure?

(0.5) 7.19 Subprocedure No. 4,

" Core Cooling Without Level Restoration,"

of E01-1, "RPV Control,"

identifies six alternate injection systems which should be used _in an attempt to restore water level if both loops of CS are inoperable and normal injection systems or subsystems connot restore level to

-150" on LI-3-46 A/B

(+18" on LI-3-52 and 62).

List four (4) of those alternate injection sytems/ flow paths.

(2.0) 7.20 List the conditions under which Standby Liquid Control injection is MANDATORY per E0I-47,

" Failure of Reactor to Scram when required...".

(2.0) 7.21 LIST ALL the immediate operator actions required by E0I-1, "RPV Control."

(2.5) 7.22 Unit I has just scrammed on a low reactor water level signal and the actions of GOI-100-11, II.A are being executed. Water level is <33" and decreasing, despite recovery efforts, when you receive a report of a possible fire from the roving cable tray fire watch. What operator actions are required per Appendix B to GOI-100-11, " Reactor Scram?"

(1.5) 7.23 a.

Define " Adequate Core Cooling" per the Browns Ferry Nuclear Plant E0I Cross-Reference to the Emergency Procedure Guidelines.

(0.5) b.

List, in order of preference, tiie three-(3)' viable mechanisms of adequate core cooling.

(1.0)

WRITE "END OF SECTION 7" AND START A NEW ANSWER PAGE

25 4

L 8.

Administrative Procedures, Conditions, and Limitations 8.1 Unit 1 is in a Hot Shutdown condition following a reactor trip. As Shift Engineer, you receive a request to move some equipment through the Unit 1 565' elevation access hatch (air lock). Which one of.the following actions complies with the BF Technical Specifications and OSIL #35,

" Secondary Containment Integrity"?

(1.0) a.

Secondary containment integrity is -not required under the existing plant condition so both.the inner doors and outer doors can be opened to facilitate movement.

b.

Although secondary containment integrity is not required under the existing plant condition, OSIL #35 still requires the outer doors to be closed on a SE hold order while the inner doors are open.

c.

Although Secondary containment integrity is required,

-stationing a public safety officer allows both the inner and outer air lock doors to be open during equip-ment movement for up to one hour.

d.

Secondary containment integrity is required and the outer air lock doors shall not be opened unless the inner doors are shut, the SE outer door hold order is cleared, and a public safety officer is stationed at the airlock outside door at all times they are open.

8.2 Browns Ferry OSIL #16 governs the control of plant keys.

Every high radiation area'...

(1.0)

~

a.

...with a dose in excess of 100 mrem /hr will be locked, and the keys to these areas will be maintained under administrative control of the SE on duty.

b.

...with a dose rate in excess of 1000 mrem /hr will be locked, and the keys to these areas will be maintained under administrative control of the Health Physics representative on duty.

c.

... key will be surveyed at th'e beginning of each shift by the SE clerk and the.SE will' log'in his daily journal that all keys are in their ass,igned storage locations.

-d.

... key in'the SE key cabinet of Units 1, 2, &.3 CRs~may E

be individually signed out on a." Key Control Record" sheet by operations or health _ physics personnel.

i

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26 8.3 Who is designated as tha Browns Ferry medical assistance team leader per OSIL #12, " Fire and Medical Emergency"?

(1.0) a.

The Unit 1 ASE b.

The Unit 2 ASE c.

Public Safety Officer d.

The station nurse 8.4 Which of the following correctly states the. emergency radiation exposure guidelines per BFN-IP-15, " Emergency Exposure"?

(1.0) a.

Planned doses to the whole body for search and rescue of injured personnel shall not exceed 100 rem.

b.

Planned doses to the whole body for the control of fires and protection of facilities shall not exceed 75 rem.

c.

Planned emergency exposures shall be strictly voluntary and within the 5(N-18) limitations.

d.

The amount of emergency exposure permitted during a mission shall be determined by the Emergency Director.

8.5 Which of the following statements correctly reflects the delegation of authority at Browns Ferry per OSIL # 97 (1.0) a.

The Unit 1 ASE is responsible for the Unit 1 and 2 diesel generators.

b.

The Unit 2 ASE is responsible for the switchyard and will normally perform the Sunday 0700-1500 shift switchyard inspection.

c.

The BOP SE will supervise the refuel floor when the refuel floor SE is not at the plant site.

d.

The Unit 3 ASE is responsible for the intake building, the stack and the SBGTS.

27 8.6 Which of the following correctly describes the guidelines for use of the " System Abnormal Status" sheet in maintaining systems in standby readiness per OSIL #43?

(1.0) a.

The ASE is responsible for evaluating all " System Abnormal Status" sheets to determine if operability is maintained and that appropriate action has been taken per the Technical Specifications.

b.

" System Abnormal Status" sheets are only used to maintain control of ECCS System alignments.

c.

" System Abnormal Status" sheets need not be completed for routine operations that do not extend beyond the control of the operator changing the status, provided the operations do not extend beyond the working shift.

d.

The " System Abnormal Status" sheets should be removed from the files and discarded when the abnormal status condition has cleared.

8.7 Unit 1 is operating normally at rated conditions with one outstanding LCO on its A Diesel Generator (6 days ' remaining).

During your shift as SE the IB standby liquid control loop fails its monthly functional test and must be declared inoperable. Which of the following actions is correct per the Technical Specifications?

(1.0)

Note: Applicable TS enclosed for reference.

a.

No new TS operational restrictions and/or actions are initiated by this condition.

b.

Place the reactor in a shutdown condition with all operable control rods fully inserted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

Place the reactor in a Hot Standby condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, d.

Operation at rated conditions may continue for six days, as limited by the DG LC0; then an orderly shutdown shall be initiated and the reactor shall be shutdown and in the cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

i

28 8.8 " Controlled Zones" are established at Browns Ferry in areas requiring protective measures from actual or potential radiation or radioactive materials hazards. Which of the following state-ments is correct regarding these " Controlled Zones"?

(1.0) a.

" Regulated areas" are buffer zones surrounding radiological hazard areas; sating and drinking are generally permitted, except in posted areas.

b.

" Radiation areas" are established where dose rates are between 5 mrem /hr and 100 mrem /hr; an RWP is always. required any time a worker may receive greater than 100 mrem in five consecutive days.

c.

" Contamination areas" use step-off pads (SOPS) and RWPs to control access; the SOPS are part of the contaminated area.

d.

" Airborne radioactivity areas" use an RWP to control access; the Health Physics Representative completing the RWP may prescribe the wearing of respiratory protection devices.

8.9 A Site Area Emergency exists and the TSC organization is fully manned. TSC evacuation / relocation becomes necessary. What is the first choice for the Site Emergency Director's new location?

(1.0) a.

The Plant Manager's Office b.

The affected unit's control room c.

The Plant Maintenance Superintendent's Office d.

The Hypochlorite Building

29 8.10 Which of the following statements is correct with respect to the issuance of equipment clearances per BF Standard Practice 14.25?

(1.0) a.

An AVO is only authorized to act as a first or second party verifier in special circumstances and then, only when under direct UO supervision and communication.

b.

Clearances shall normally only be issued to employees named on the official clearance list.

For short-term emergency mainten-ance the SE may hold a clearance for an unauthorized person performing the work.

c.

The ASE and the cognizant department supervisor may jointly release a clearance during an emergency if the person holding the clearance cannot be contacted.

d.

A single white hold notice tag is used on any given clearance while clearance boundaries are identified by multiple red hold order tags.

8.11 Which of the following statements c'orrectly reflects the procedures and responsibilities established by OSIL #21 for the handling of " Maintenance Requests" at the Browns Ferry Nuclear Plant?

(1.0) a.

The SE or the ASE will initial the back of the MR form to acknowledge completion of the necessary repairs.

b.

A square yellow sticker, containing an MR number, date and operator's initials is used to identify the controls of equipment experiencing maintenance trouble.

c.

Form BF-40, " Urgent Corrective Maintenance List", will be completed by each unit ASE prior to 0500 each morning.

d.

Before any corrective maintenance begins, the foreman or craftsman will obtain the SE's, ASE's, or UO's signature on the back of the MR.

i I

i

30 8.12 Which of the following shift turnover practices is NOT correct per the Standard Practices for " Administrative Control of Plant Opera-tions" (12.17) and " Shift Turnover" (12.7)?

(1.0) a.

Individual System panel checklist for core cooling systems shall be reviewed by both the U0 and the ASE of each shift and documented on the shift turnover cover sheet.

b.

Prior to relieving the off going shift both the VO and the ASE will acquaint themselves with the status of ongoing activities.

Panel walkdowns are required, but may be performed following shift relief.

1 c.

Prior to relieving the off going shift, all operators will review journal entries back to the person's last shift or back five calendar days, whichever is less, d.

Temporary reliefs during a shift must be logged in the operator's journal including the time, the relief's name, and his initials upon assuming responsibility.

8.13 BF Standard Practice 12.17, " Administrative Control for Plant Operation," establishes plant policy for the control of contain-ment isolation and safety systems during an emergency.

a.

What evaluation shall be made prior to resetting a primary containment isolation?

(1 0) b.

What are the two conditions which allow operators to override automatic operations of engineered safety features?

(2.0) 8.14 During power operation each control room is manned by an assigned licensed reactor operator who must normally remain in " Zone Number 1."

Under what two (2) Conditions may he/she enter " Zone Number 2" while at power?

(1.0)

I w

31 8.15 Which of the following correctly describes the Procedure (per RCI-9) for an immediate and emergency entry into an airborne radioactivity area to prevent damage to plant equipment?

(1.0) a.

The SE can direct entry without an RWP. Notification of entry is required to be made to HP Shift Supervisor.

b.

The SE can direct entry with a Routine RWP.

c.

The HP shift supervisor can permit entry only with an approved RWP.

d.

The HP Shift Supervisor can permit entry without an approved RWP if continuous HP coverage is made available.

8.16 Units 1 and 3 are operating at 75% and 100% rated thermal power, respectively. Unit 2 is in Cold Shutdown.

During the performance of sis (4.7.8) on SBGTS Trains A, B, and C, the following deficiencies are noted:

SBGTS Blower / Fan A is INOP SBGTS Train C decay heat damper (FC0 52) is stuck open Maintenance estimates 3 to 4 days to accomplish repairs.

What, if any, is the TS requirement in this situation?

(1.0)

Note: Applicable TS enclosed for reference.

a.

No TS restrictions and/or actions are required by these deficiencies.

b.

Reactor operation and fuel handling is permissible only during the succeeding 7 days.

c.

Place Units 1 and 3 in at least Hot Standby within six hours and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

d.

Place Units 1 and 3 in at least Hot Standby within six hours and in Hot Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, i

i 4

32

'8.17 Which of the following correctly describes the TS definition of an " Instrument Check"?

(1.0) a.

The adjustment of an instrument signal output so that

~

it corresponds, within acceptable range and accuracy, to a known value of the parameter which the instrument monitors.

b.

Tne injection of a simulated signal into the instrument's primary sensor to verify the proper instrument channel response, alarm, and/or initiating action.

c.

The qualitative determination of acceptable operability by observation of instrument behavior during operation, including, where possible, comparison of the instrument with other independent instruments measuring the same variable.

d.

A test of all relays and contacts of a logic circuit to insure all components and instruments are operable per design.

8.18 Unit 1 is operating at 75% rated thermal power. Operability sis are performed on all of the MSL Radiation Monitoring System Channels. MSL Radiation Monitoring System Channel A tests UNSAT per the SI. Maintenance has no estimate of repair time and will not be able to commence troubleshooting and repair for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to the lack of a crucial part. What is the TS requirement in this situation?

(1.0)

Note: Applicable TS enclosed for reference, a.

Trip the RPS and PCIS trip systems associated with MSL Rad Monitor Channel A; power operation may continue.

b.

Place MSL Rad Monitor Channel A in an INOP status but do not trip the associated RPS and PCIS trip systems for four hours; power operation may continue.

c.

Trip the RPS and PCIS trip systems associated with MSL Rad Monitor Channel A; initiate an orderly load reduction and have Main Steam lines isolated within eight hours.

d.

Trip the RPS and PCIS trip systems associated with MSL Rad Monitor Channel A; and initiate insertion of operable rods and complete insertion of all operable rods within four hours.

33 8.19 Which of the following situations requires application of the Power Transient Fuel Cladding Integrity Safety Limit of the Unit 1 Technical Specifications?

(1.0) a.

Reactor power is at 42% RTP; the main turbine trips due to an EHC malfunction; the reactor SCRAMS on HIGH PRESSURE; the BPVs control pressure thereafter.

b.

Reactor power is at 70% RTP; a steam leak to the Drywell occurs and Drywell pressure rises; the reactor SCRAMS at 2.05 psig; HPCI auto-actuation does not occur, but manual start is successful; the reactor is brought to a cold shutdown condition.

c.

Reactor is in Start-Up, at 12% RTP; power is increased by rod pull; the reactor SCRAMS at 12.5% power, by APRMs; level and pressure are maintained by normal systems for the plant status, d.

The reactor is at 18% RTP; 3-1/2 BPVs are open in preparation for turbine warmup; controller failure reduces pressure to 875 psig; reactor SCRAMS on closure of MSIVs; level and pressure are maintained by normal systems for the plant status.

8.20 Which of the following correctly describes the utilization of an Operating Permit (TVA 6271)?

(1.0) a.

The person holding an operating permit is authorized to perform maintenance on the equipment.

b.

The operating permit authorizes operation of equipment from a control board only by its normal operators.

I c.

More than one person can hold the same operating permit.

d.

When an operating permit is used in conjunction with a hold order, a person other than the one holding the operating permit can be issued the hold order.

8.21 Who are the two individuals charged to determine if a one-hour or four-hour red phone report is required per BF SP 15.2, 4

Licensee Event Report (on Form BF-19)?

(1.0)

ShiftEngineerabdOperationsSupervisor a.

b.

STA and Compliance Supervisor c.

Shift Engineer and STA d.

Operations Supervisor and Compliance. Supervisor

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s.

p-

34 8.22 Whose approval is required prior to removing from service any component which renders a fire protection system incapable of performing its intended function in its intended manner for any reason other than testing?

(1.0)

NOTE:

Assume condition of protected equipment is such that Fire Protection is required.

a.

Operations Supervisor / Superintendent b.

Plant Manager c.

80P SE d.

Public Safety Services Supervisor 8.23 Per OSIL # 21, Maintenance Requests, Emergency Maintenance...

(1.0) a.

...is maintenance requiring work to be performed within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or on the next scheduled workday.

b.

...is only allowable for safety related equipment or CSSC components.

c.

...on safety related equipment must be reviewed by plant supervision for adequacy of procedures and hold points.

d.

...on safety related equipment may begin prior to issuance / completion of appropriate clearances if clearances will be completed within one hour after the maintenance begins.

8.24 Concerning General Electric's Preconditioning Interim Operating Management Recommendations (PCIOMR):

a.

Starting with the fuel at a threshold of 11.0 kw/ft, a maximum ramp increase is begun at time 0000 and the final desired power of 13.0 kw/ft is achieved at 2000.

At this time, the required soak is performed for 10 MINUTES, at which time the load dispatcher directs a power reduction that takes nodal power down to 12-9 it.V kw/ft. SELECT the valid preconditioned value for this node.

(1.0)

ASSUME THE MAXIMUM RAMP RATE IS.10 kw/ft/hr 1.

11.0 kw/ft 2.

11.8 kw/ft 3.

12.5 kw/ft 4.

13.0 kw/ft

o 35 b.

SELECT the minimum time which would be required to raise power back to 13.0 kw/ft, given the above maximum ramp rate.

(1.0) 1.

Immediate (Raise to 13.0 kw/ft, w/o restrictions) 2.

5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 4.

20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 8.25 The following data was taken during two days of operation at approximately 100% rated thermal power on Unit 1.

The unit had been at that power for two weeks.

Identified Unidentified Leakage Leakage Day 1, 0800 9907 gal.

1488 gal.

1600 10,560 gal.

1008 gal 2400 10,440 gal.

1824 gal.

Day 2, 0800 9398 gal.

2198 gal.

1600 9600 gal.

2198 gal.

2400 9278 gal.

2496. gal.

STATE the TS Coolant Leakage LCO limit (s) applicable in this plant condition and IDENTIFY any that were exceeded during these two days.

(3.0)

WRITE "END OF EXAM" ANO SIGN YOUR ANSWER SHEET

ATTACHED ARE THE FOLLOWING TECHNICAL SPECIFICATIONS EXCERPTS FOR USE i.

IN ANSWERING CERTAIN SECTION 8 QUESTIONS.

4-I

-. Selected Definitions Section 3.1 -Reactor Protection System i

Section 3.2 Protective Instrumentation j

Section 3.4 Standby Liquid Control System

- Section 3.7 Containment Systems Section 3.9 Auxiliary Electrical System l

}

l NOTE: SELECTED PAGES AND SETPOINTS HAVE BEEN DELETED FROM THE HAND 0UT.

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Raviccd 2-6-81

.f 1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications any be achieved.

A.

Saf ety Limit - The safety limits are limits below which the reason-able maintenance of the cladding and primary systems are sasured.

Exceeding such a limit requires unit shutdown and review by the Atomic Energy Commission before resumption of unit operation.

Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.

B.

Limitina Safety System Settina (LSSS) - The limiting safety system setting are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit. sad these settings represent margin with normal operation lying below these settings. The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.

C.

T4=4 tina Conditions for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup sad operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.

g 1.

In the event a Limiting Condition for Operation and/or associated requirements cannot be satisifed because of circumstances in excess of those addressed in the specif1-cation, the unit shall be placed in at least Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are completed that permit operation under the permissible discovery or until the reactor is placed in an operational condition in which the specification is not applicable. Exceptions to these requirements shall be stated in the individual specification's. This provides actions to be taken for j

circumstancas not directly provided for in the specifications and where occurrence would violate the intent of the 1

specification. For example, if a specification calls for two systems (or subsystems) to be operable and provides l

for explicit requirements if one system (or subsystem) is i

inoperable, then if both systems (or subsystems) are inoperable the unit is to be in at least Hot Standby in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> if the inoperable condition is not corrected.

l

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4 e

a 1.______

• I_*"

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Reviced 2-6-81 1.f 1.0 DEYINITIotts (continued) 2.

When a system, subsystem, train, component or device As determined to be inoperable solely because its onsite power soutes is inoperable, or solelybecause its offsite power source is inoperable, it any be considered operable for the purpose of satsadyeng the requirements of its applicable Limiting condition For Operation, provided:

(1) its corresponding offsite or diesel power source is operable; and (2) all of its redundant systea(s), subsystem (s), train (s),

I component (s) and device (s) are operable, or likewise eatisfy these requirements. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least Bot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This is not applicable if the unit is already in Cold Shutdown or Refueling. This provision describes what additional conditions ~ mast be satisfied to permit operation to continue consistent with the specifications for power sources, when an offsite or onsite power source is not operable. It specifically prohibits operation when one division is inoperable because its offsite or diesel power source is inoperable and a systaa, subsystem, train, component or device in another division is inoperable for another reason. This provision permits the requirements associated with individual systems, subsystems, trains, components or devices to be consistent with the requirements of the associated electrical power source. It allows operation

/

to be governed by the time limit of the requirements associated with the Limiting Condition For Operation for the offsite or

' diesel power source, not the individual requirements for each system, subsystem, train, component or device that is determined to be inoptrable solely because of the inoperability of its offsite or diesel power source.

D.

DC ETED 2a

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.,g' 1.0 DETINITIONS (cent'd)

FEB 061981 A systen, subsystem, train, component, E.

Operable - Deerability or device shall be Operable or have operability when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the systen, subsystem, train, component or device to perform its fune: ion (s) are also capable of perfor=1ng their related support function (s).

F.-

Operatine - Operating means that a system or component is performing its intended functions in its required manner.

G.

Inmediate - Immediate means that the required action vill be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.

H.

Reactor Pever Operation - Reactor power operation is t.ny operation with the mode switch in the "Startup" or "Run" position with the reactor critical and above 12 rated power.

I.

Het Standbv Conditien -

/

J.

Cold Condition - Reactor coolant te=perature equal to or less than K.

Hot Shutdown - The reactor is in the shutdown mode an'd the reactor coolant te=perature greater than L

L.

Cold Shutdown - The reactor is in the shutdown mode and the reactor coolant te=perature equal to or less than M.

Mode of Oeeration - A reactor mode switch selects the proper interlocks for the operational status of the unit. The following

.are the modes and interlocks provided:

1.

Startue/ Hot Standbv Mode - In this mode the reactor protection scras trips initiated by condenser low vacuum and main steam line isolation valve colsure, are bypassed when reactor pressure is less than the reactor protection system is energized with IRM neutron nenitoring system trip, the APFM 15I high flux trip, and control red withdrawal interlocks in service. This is often referred to as just Startup Mode. This is intended to imply the startup/ Hot Standby position of the mode switch.'

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AUG 131984 1IllITI!!C COMDTTIONS FOR OPERATION SimVEII.I.ANCE }(EOUIREME?TTS

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3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR P!!OTECTION SYSTEM i

Applicability Applicability Applies to the instrumentation Applies to the surveillance of and associated devices which the instrumentation and asso-initiate a reactor scram, ciated devices which initiate reactor scram.

Objective Objective To assure the operability of the l

To specify the type and frequancy reactor protection system.

of surveillance to be applied to the protection instru=entation.

Specification Specification A.

When there is fuel in the vessel, A.

Instrumentation systems shall the setpoints, minicum number of be functionally tested and trip systems, and minimum number calibrated as indicated in of instrument channels that must Tables 4.1.A and 4.1.B respee-be operable for each position of tively.

the reactor mode switch shall be as given in Table 3.1.A.

T B.

Two RPS power monitoring channels for each inservice' RPS MC sets or alternate sourte shall be operable.

When it is determined that l

1. With one RPS clectric a channel is failed in' the unsafe condition, the power monitoring channel other RPS channels that-for inservice RPS MG set monitor the same variable or alternate power supply shall be functionally inoperable, restore the tested immediately before inoperable channel to the trip system containing operable status within the failure is tripped.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> er remove the The trip system containing the unsafe failure may be associated RPS MG set or alternate power supply untripped for short from service.

periods of time to allow functional testing of the other trip system. The trip system may be in the untripped position for no more than eight hours per functional test period for this testing.

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DEC 121983 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM B.2 With both RPS electric power B.

The RPS power monitoring monitoring channels for an sy st em instrumentation inservice RPS MG set or alter-shall be determined operable:

na te power supply inoperable, restore at least one to 1.

At least once per 6 months operable status within 30 by perf ormance of channel minutes or remove the f unctional te sts, associated RPS MG set or alternate power supply f rom service.

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DEC1 1983 TAllt.E 3 1. A REACTOR PROTECTION SYSTEll (Scil Atti IllSTDt!!1ENTATION REOlliltr.Ilf.NT

(

Hin. No.

or l

Operable Iloden in Which Function Inst.

ilust ne Operahic Channels Startup/Ilot y

shnt-Actlen(l)_

dnun 11eruel(7)_

Standb L Bun I

Per Trip System (1)(23) Trip Function Trip p _ vel Setting f'

X X

X X

1.A j

1 tiode Switch in Shutdown X

X X

X 1.A 1

Hanual Scram 3

Illgh Flux 6120/125 Indicated X(22)

X(22)

X (5) 1.A IllH (ife) l on scale 1

X X

(5) 1.A 3

Inoperable APIIH (16) (24)(25)

X 1.A or 1.D u

High Plux (Flow Blased) See Spec. 2.l.A.1 X

1.A or 1.R 2

X(21)

X(17)

(15) 1.A or 1.B 2

High Flux (Fixed Trip) 6120 %

2 High Flux

$151 rated power (13)

X(21)

X(17)

X 1.A or 1.8 2

Inoperative (11)

(11)

X(12) f.A or f.D t

2 Downscale

?) Indicated on Scale

{

X(10)

X X

1.A I:

2 High Reactor Pressure l!

2 High Dryvell X(R)

X(8)

X 1.A

[

Pressure (14)

L Reactor Low Water X

X X

l.A d

2 Level (188) 1538" above vessel zero Ucat feram Discharge 6 50 Callons X

X(2)

X X

1.A 2

.High Water Level in j;

Tank I

(l.S-85-45 A-D) 2 111gh Water Level in East $50 callons X

X(2)

X X

l.A h

Scram Discharge Tank (l.S-85-4 5 E-II) j v

k i

' S i

l TABLE 3.1.A REACTOR PROTECTION SYSTEM (SCRAMI INSTRUMarrATION REQUIRDIENT k

Min. tio.

ct Opercble Modes in which nanction I

Isst.

Most se operable l

• Ch:nnels Shut-Start up/tsot Trip i.evel Sett ing M Refuel (71_

Standtv Run Actionf1)

Per Trip rystsa (1L(23) Trio Function Main Steam Line Isola-X (3) (6)

X (3) (6)

X (6) 1.A or 1.c 4

tion valve closure 5 105 Valve closure 2

Barbine Cont. Valve X (4) 1.A or 1.D or 2 550 isla Fast Closure Turbine Trip 4

Turbine Stop Valve X(4)

1. A or 1.D 5 105 Valve Closure Closure W

2 Turbine First Stage X(18)

X(18)

X (18)

(19) not 2154 psis Pressure Permissive 2 23 In. Ug, Vacuum 7(3)

X(3)

X

'1.A or 1.C 2

~ Turbine Condenser low Vacuum 2

Main Steam Line High 3X Hormal Full Powtr X(9)

X (9)

X(9) 1.A or 1.C l

Radiation (14)

Background (2())

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MAY 191382 i

NCT!S FOR TABLE 3.1-A 1.

here shall be two operable or tripped trip systems for each function.

If the minimum number of operable instrument channels per trip system cannot be met for both trip syetems, the appropriate actions listed below shall be taken.

A.

Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

In refueling mode, suspend all cperations involving cera alterations and fully insert all operable control rod 1 within one hour.

B.

Reduce power level to IRM range and place mode switch in the Startup/:fot Standby position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C.

Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

D.

Reduce power t'o less than 30% of rated.

2.

Scram discharge volume high bypass may be used in shutdown or refuel to bypass scram discharge volu=e scram with centrol rod block for reactor prcsection system reset.

3.

Bypassed if reactor pressure psig and mode switch not in run.

4 Bypassed when turbine first stage pressure is less than 154 prig.

f 5.

IRM's are bypassed when APRM's are onscale and the reactor mode switch is in the run position.

6.

The design permits closure of any two lines without a scram being ini tiated.

7 When the reactor is suberitical and the reactor water temperature is less than only the following trip functions need to be operable:

A.

Mode switch in shutdown B.

Manual scram C.

High flux IRM D.

scram discharge volume high level E.

APRM 15% scram 8.

Not required to be operable when primary containrent integrity is not required.

9.

Not required if all main steamlines are isolated.

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JUL 22882

10. Not rey'ared to be opera'le whc3 the re,ictor pressure vessel u

'f heaad is not l'r I t e* l to tJt e vessel.

i 11 The. A.'RM down rale trap function 1 ". only active w fien the reactor mode swa ten is in run.

12. The APHM downntrale t. rip is automatically "bypa:. nerd when the IRM instrument.ition in operable and not high.

13. Less th.2ri 14 mmrable LPfe a will cause a trip syctem trip.

1 83 Channel shared by Reactor Protection System and Primary Containment and Reactor Vessel Isol t-ion Control System.

A channel failure may be a channel failure in each system.

15. The ApBM 15% scram is bypassed in the Run Mode.

16. Channel shared by Reactor Protection System and Reactor Manual Conerni Sy.st r m (Ro.1 filoch Port ion). A chauncl failure nay be a channel f ailure in each nynten.

If a channel in alloweil to be U:0P!;ltAr!.C per Table 3.1. A, the corre mendinv. funct ion in th.it tace channel nay be innperable in the Reactor Manual Cont rol System (Rod lilock).

17. riot required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not, to exceed 5 M,3(t).

13. This function must inhibit the automatic bypassing of turbine co..trni valve fast closure or turbine trip scram and, turbine stop valve closure

[j scram whenever turbine first stage pressure is greater th.n or equal to 154 psis, r

19. Action 1.A or 1.D shall be takeri,only if the permissive f ails in such a manner to prevent the affected RPS logic from perforning its intended function.

Otherwice, no action is required.

o

20. Tne nor. sinal setpoints for alarn and reactor trip (1.5 and 3.0 times bachgroand, respectively) are establishci based on the normal background at full power. The allowable sa:tpoints for alarm anel reactor trip are 1.2-1.8 and 2.4-3.6 tinen background, respectively.

21. The APRM High Flux and Inoperative Trips do not have to be operable in the Refuel Mode if the Source Range Monitors are conr.ected to give a non-coincidence, Hi;;h Flux scram, at S x 105 eps, The SRM's shall be operabic per Specification 3.10.B.I.

The removal of eight (S) shorting links is required to provide non-coincidence high-flux scrai protection from thu Source Range rionitors.

22. The three rerpet red IlOl's per trip channel is not required in the Shutdown cr Refuct Modes if at least four IRM's (onc in cach core quadrant) are connected to give a non-coincidence, Ilir,h Flux scram.

The removal of four (4) shorting links is required to pt.uide non-coincidence hi;;h-flux scram protection from the IRM's.

23. A channel may be placed in an inoperable status 'for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for required surveillance without; placing the trip system in the tripped

[

condition provided at least one OPLRA:;LF. channci in the same trip j

systen is monitoring that parar eter, f

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Dr e' 1 1983

,f The Average Power Range Monitor scram function is varied (ref.

24 Figure 2.1-1) as a function of recirculation loop flow (W).

The trip setting of this function must be maintained in accordance with 2.1.A.

25. The APRM flow biased neutron flux signal is fed through a time constant circuit of approximately 6 seconds. This time constant may be lowered or ecuivalently removed (ne time delay) without affecting the operatil1* y of the flow biased neutron flux trip channels. The APRM fixed high neutron flux signal does not incorporate the time constant but responds directly to instantaneous neutron flux.

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(tjtTIMr:_CONtitTIONS FOR OPERATION SURVEli.I.ANCt REQUIRD4ENTS

3. )

PROTECTIVE INSTRUMENTATION 4.2 PROTECTIVE INSTRUMElffATION Applicability Applicability i

Applies to the plant.instrumen-Applies to the surveillance re-tation which initates ar.d con-quirement of the instrumentation I

trole a protective function.

that initiates and controle pro-tactive function.

l Objective Objective To assure the operability of To specify the type and frequency protective instrumentation, of surveillance to be applied to protective instrumentation.

Spec if ic.s cion Specification A.

primary Containment and Reactor.

A.

Primary Containment and Reactor Build in e, tsotation Functions Building Isolation Functions l

When primary containment inte-Instrumentation shall be func-j arity is required, the limitinte.

tionally tested and calibrated conditions of operation for the as indicated in Table 4.2.A.

instrumentation that initiates primary containment isolation ~

Syaces logic shelf be function-are given in Table 3.2.A.

This ally cented se indicated in includes instrumentation that Tabla 4.2.A.

initiates' isolation of.the reac-g{

tor vessel, reactor building.

(

main steam lines, and initiates 9...' standby gas treatment system.

R.

Core and Containment Coolina 5.

Core and Containment Coolint Systems - Initiation & Control Systems - Initiation. & Control I

1 The limitica cenditions for Instrursentation shall be fune-operation fer the instrumenta-tionally tested, calibraced and tion that initiates or controls checked as indicated in Table the core and containment coo' ling 4.2.3.

systems are given in Table 3.2.3.

This instrumentation must be Systes logic shall be function-operable when the system (s) it ally tested as indicated in initiates or controls are re J Table 4.2.3.

quired to be operable sa speci-fied in Section 3.3.

.[

Whenever a system or loop is rade inoperable beesuse of a required test or calibration, the other syntess or loops that

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I-4 TAtiLE 3.2.A PRIMARY CONTAINMENT AND REACTOR BUILDING I!41ATION INSTRUMENTATION Minimum No.

Instrument Channels Operable

.l Per Trip Sys(1)(11) runction Trin uvel settina Ace ton _ (11 Remarks

• 2 Instrument channel -

2 538" alm >ve vessel zero A or 1.

Below trip set. ting does the i

(B snd E) following:

Reactor Low Water Level (6) a.

Initiat re Mact >r Duildang Isolation t

b.

In st iat es I. im.sr y Containcient Isolation ( Croups 2,3,and6) c.

Initiat es SGTS D

1.

Above trip setting isolates the t

1 Instrument Channel -

100 + 15 psig shutdown cooling suet an valves Reactor High Pressure of the RilR s ys tam.

2 Instrument Chann*1 -

2 3 78" atm)v e= ve*s,el zero A

1.

Below trip e,etting snatiates min Steam Line* I sola tiori Peactor Low Water Level (LIS-3-56A-D, SW 01) h A or 1.

Above trip set t t r.q W. the 1

2 Instrument Ch.nnn.1 -

l (H and C) following:

High Dr ywell Pt e.esure* (6) a.

Inst a at es React 'it Hie ild an 3 (PS-64-56A-0)

Isointion 4

b.

Init i.it +%

Primst'/ Con t.sinme n*

I Isolattoh f

c.

Initiat s SGTc j

2 Instrinwne Ch annol -

3 times normal rated D

1.

Above trip setting irtttates Maan High Radtat ion M. sin Steam fi11 power tackgrour>s (13)

Steam Line Isolation Line Tunnel (6) 2

' Instrument Chennel -

2 825 psig (4) 8 1.

Below trip settinq snitiates Main Steam Line Isolatiot.

Iow Pressure N. tin Steam Line g

2 (3)

Instrument Ch.'nnel -

5 1405 of rated steam flow B

1.

Above trip setting initiates fr.ai n Steam Line Isolation Iligh Plow M.iin St eam Lane (D

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TABLE 3.2.A l

PRIMARY COtGAINMENT id4D REACTOR BUILDING ISO 1ATION INSTRUMENTATION l

Minimum No.

Instrument Channels Operable' runct ion Trio I4 vel setting _

Action ( 1 L_.

Rwru per Trip Sys(1)(11) o a

1.

Above t r i p r.et t i r.<i i n i t ia t t s I

2 (12)

Instrument Channel -

< 200er Hain Steain Line I +> L *

• t en.

Main Steam Line Tunnel tilgh Temperature 16G - 180*r C

1 Move trap setting anttnates I

i 2 (14)

Instrument Channel frolu tar-

>f 9e.netor M * >a r Reactor Water Cleisaup cuanup

.r.

f r om F ? rt og

.m !

4 System Ploor Draan

• csctor

.*cr feturn I.ine.

High Temperature 2

Instrument Cnannel -

1b0 - 180*r C

t.

.;ne.2::

4 ave Reactor Entar Cleanup System Space fliqts Temperature 1

Instrument Channel -

5 100 mr/nz or downscale G

1.

1 upscale or 2 downnc il t will e.

In s t a a *. <a CGTS l

1ation High R..diat ion -

b.

Isol ta reactor one and Reactor Buililing Venti-retualing floor, control sytt m. g Deactor Zone c.

Clou-

• .'osphoto O

100 'ar/h or downocale r

1.

1 upsc. ale a:r 2 downscale will d

I i

y 1

Instrument channel

.n. Initiate SGTS Reactor Duilding venti-h.

Irole. refuelani floor.

i c.

Clo ;<= ata.osphe r" control system O) lation liiqh R.vli.at ion -

Setu.*1ing Ion.

aD 2 (7) (8) ~

Instrument channel

.harcosl lleaters 5 2000 cfm H and 1.

Selow 2 M O cim, trip r,etting ct..ir-G)

A SGTS Flow - Train A

4. ?t. If e,.t or s T 2000 cfm (A or F) coal hoat. to will turn on.

2.

Below 20 Y. Cf.7, t r ip :.et ting P. il.

Heaters heaters will shut off.

2 (7) (8)

Instrument Channel Charcoal ileaters 5 2000 cfm H and 1.

Selow 2000 cim, t r i p s et t ing cha r -

l-SGTS Flow - Train B R.H. Heaters 5 2000 cfm (A or F) coal heat.ers will turn on.

2.

!elow 2C30 cfm, tr i p e,et t inq a.!!.

Heaters heater *. will r. hut off.

):

2 (7) (8)

Inntrunent Channi1 Charcoal lie.iters 52000 cf m H anel 1.

f*elcw.000 cfn. trap sat t an<i en e r -

i.

SGTS Flow - T:ain C R. it. treaters 5 2000 cf m (A or F) coal ha.it r1 wil l t.ir n. n.

2.

Solow 2000 cim, trip nc* t a ng %.H.

r' Heaters

i. eater:a wt 11 cnut

t. t t.

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T AI LE 3.2.A PRIM ARY CO:4T AltsMr.".* 4t*D R EACTOR BUILDING ISOIATIOrd Iti.TRUMEfC ATIJta

!!irnimu:n h*o.

g Instrument y

e Channels Operable per Trip Sys Q)]) Punct ton _ Trio revel Setties _getton (11 aenirks 1 Reactor Buil.1ing Isolataan J$t s 2 seca. H or i 1. La.los trap setting prevents rpinous t rips ont sy9tas pertur-Timer'(refuelin3 floor) L e vina trssa anse iat isw asolataan i H or i 1 inviteil in unit 1 only P* t i in v. f or t.t at ic l'a cssur e 1 In st.rumen t Channel - Static Pressure Control co.. rol (CGTS A. 8, or. C on). Periaissave (a ef uelin. Chtanel

• h.or ed t y ge re 1stve on floor) rsactor zone st.it ic g r... sur e cont.

H or I 1. l oca t +1 in unit 1 only 1/ 2" !! o 2. sentrols ut st le prest.ur-of 4 1 Static Press are Control Pressure Regislat or (Re-re*uelin. floor durtn1 r e.seto r fuelang Floor) t uildan s s tol tt son with SGTS T ur st a n.t. f 1 Reactor butl.lierg I sol a t a..in ') $t 5 2 accs. G or A 1. $ tow trip setting prevents or H ' hpurious trips an] sy.tes pertur-Timer (re actor zona) c.ition5 t rem initiat i 1 ksoittion. I 1. Par:=tssive for statte Fressure 1 (9) Instrument channe1 - Te/ A control (CGTS A, B or C on). St at ic. P r <.s r.ure. Cont rol channel sn.tred by perntssive on permissave (re wtor r.ruliv; floor stati.- pressura i zone) ecntrol. I 1. Controls static press *:re of 119) Static Pressure Contaol 5 1/ J" ll o re setor zone diaring r. setor Pressure P. palator (t e.vt os tuildinI a mlat a cn w t' n G4T.i } zone) running. A 1. heter to Table 3.7. A f ' r list of 2 (10) Group 1' (Initiat anql 1,& g a c f4/ A valves. 19 1 Refer to Tnble 3.7.A Lor list tot c -L Group 1 ( Ac tu.s t aon) I.o g i c N/A C 1 . valves. t~ CO .I N. ~ 1 4 3 3 f

p i .h e Ti.htC 3.2.A PRIMAf Y CONTAll. MENT Af4D REACTOH BUILDING ISOIATION I:CITHHM EN TATIOtt Minirt ri !;o. g g. Inst remen t j Channels Operable j ptr Trip Sys(1)(1} } runct ion Trip t. eve 1 cattino Action til Panarks ~ j 2 Group 2 (Initiatir. p Logic N/A A or 1. Refer to TJtle 3.7.A tot list of (8 and El valves. y i Group 2 (RHR 1s0144 ion-N/A D Actuation) togac r' I Group 8 (Tip-ActEation) 23/A. J togte 1 Gr oup 2 (Drywell rump N/A ft l -- Dr a ins-Act u.et. ion) t.. ma s c 1 . Group 2 ( Re.sc t est amaldi53 N/A F and G 1. l'a r t of Ca n'op 6 Imit. & pef ta+=1 &ng Floor, and Dry-well Vent ased t'un 9. - Actuation) log ie: 2 Group 3 (Init l e* ar: tl M3i c N/A C 1. Imfer to Tat:la J. 7. A t ot 11.t o! valves. 1 Group.3 ( Act uat t im) 8.o l t e FI/A C w 1 Group 6 Loqic 13/A F and G 1. Peter to Tit.l* 3.7.A for list i J of valves. 1 Group 8 (In i t iat ! n lo11c. 1 . Reactor Buildin.) I olation N/A H or r 1. 1.ogic tus p *r1miscive to taf ueling frefueling floor) 1.> >q i c floor st.atic pressura r "<su l.s t o r. 1 Reactor Buildire Isolation 13/A H or G 1. Logic has paraissive to reactor p (reactor zone) Logic or A zone statae pressure re 91.stor. g C) I (JI L p l l i

y TARLE 3.2.A PRIMARY CONTAll#tElfr At:D dEACTOR BUILDING ISOLATIOtt IISSTRUstBIrfATION flinim h j Ins trur.,en t , Channels Operable per Trlin Sys(1X11) runction Trw Levet settina action til Pe m t L!.._ o 1(1) (e) SGTS Train A logic N/A L or (A and F) L or

• (7) (6) sGTS Train B Logic N/;.

(A and F) 1(7) (e) scTs Train C togic N/A L or (A and F) 5 or F 1. Located in i. nit 1 only. 1 Static Pressure control N/A (refuelino floor) Logic 1(9) Static Pressure Control N/l. I (reactor sone) Logic-Ref e'r to Table 3. 2.B for DCIC and HPCI f unction s in:1uding Groups 4, 5, and 7 valves. ^$ I cCr~ s tQ I-g

1 l IeOTES FOR TABLE 3.2.A gg 1. Whenever the respective functLons are required to be operable, there shall be two operable or tripped trip systems for each function. if the first column cannot be met for one of the trip systems, that trip system or logic for that function shall be tripped (or the 1 appropriate action listed below shall be taken). If the column cannot j bc met for all trip systems, the appropriate action listed below shall be taken. A. Initiate an orderly shutdown and have the reactors in Cold Shutdown Cond.ition in 2f+ hours. S. Initiate an orderly load reduction and have Main Steam Lines tooLated within eight hours. C. Isolate Reactor Water Cleanup Systre. D. Isolate Shutdown Cooling E. Initiate primary containment isolation within 24 hours. F. The handling of opent fuel will be prohibited and all operations over spent f uels and open reactor wells shall be prohibited. C. Isolate the reactor building and start the standby gas treatment system. / H. Immediately perform a logic system functional test on the logic in th. ether trip getamesad daily thereaf ter not to exceed 7 days. 1. No action required. Reactor sone welle and ceiling designed above suction pressure of the SGT3. .l. Withdraw TIF. K. Manually isolate the affected lines. Refer to section 4.2.E for the

t.. N

'Dk*t [do*pe"raoke'I'anE*sctions H er a:tien A and T. If two SCTS in is t trains are ineperable tage actiens A and F 2. When it is determined that a channel la failed in the unsafe condition, the other channels that monitor the asse variable shall be functionally tested innediately before the trip system or logic for that function is tripped. The trip system of the logic for that function may remain untripped f or short periods of time to allow functional testing of the ether trip system or logic f or that function. 3. There are four sensors per steam line uf wreich two must be operac;e. 4 Only required in nun Mode (interlocked with Mode Switch). Not required in Run Mode (bypassed by mode ywitCh). l 60 /

• es

.n. A ..L7:.*. m.

OCT 161984 6. Channel shared by Rps and Primary Containment & Reactor vessel Isolation control system. A channel failure may be a channel failure in each systen. 7. A train is considered a trip system. 8. Two out of three SGTS trains required. A failure of more than one will require action A and F. 9. There is only one trip system with auto transfer to two power sources. 10. Refer to Table 3.7.A and its notes for a listing of Isolation Valve Groups and their initiating signals. gg* A channel may be placed in an inoperable status for up to four haurs for required surveillance without placing the trip system in the tripped condition provided at least one OPER,ABLF. channel in the same trip system is monitoring that parameter.

12. A channel contains four sensors, all of which must be operable for the channel to be operable.

?over entra:icus pe=1:ted fer up =c 30 days vi:h 15 of :he 16 ta=seratura 's.d ::hes operahlt. In the event t' hat normal ventilation is unavailable in the main steam line tunnel, the high te=perature channels may be bypassed for a i period of not to exceed four hours. During periods when nor=al ventilation is not available, such as during the performance of secondary contai. ment leak rate tests, the control room indicators of

he affected space te=peratures snall be monitored for indications of imall steam leaks. In the e. vent of rapid increases in ter:perature (indicative of steam line break), the operator shall.promptly close the main steam line isolation valves.

13. The nc.inal serpoin:s for alarm and reactor :1p (1.5 and 3.0 ::mes backgrcund, respe:::vely) are es:ablished based on :he ne=al :' ck-

cu

d at full ;over.

he c.llevaole setpcin:s fer ala = and reae::: trip are 1.2-1.8 and 2.I.-3.6 times background, resee:::vely. 14. Requires two independent channels from each physical location, there are two locations. 61 " " " ^ _, _ e.

• LIMITINC CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

[ 1.4 STAND 3Y LIQUID CONTROL SYSTEM 4.4 STANDBY LIQUID CONTROL SYSTEM Applicability Applicability V Applies to the operating etstus Applies to the surveillance require-of the Standby Liquid Control mente of the Standby Liquid Control System. System. Objective Objective To assure the availability of a To verify the operability of the Standby erstem with the capability to Liquid Control System. ohut down the reactor and main-tain the shutdown condition with-4 out the use of contro rode. spee ficatien specification A. Normal Svetem Availability A. Normal System Availability 1. The standby itquid con-The operability of the Standby trol system enall be opera-Liquid Control System shall be veri-ble at all times when there fied by the performance of the to fuel in the reactor ves-following teste: sel and the reactor is not in a shutdown condition 1. At leest once per month each with all operable control pump loop shall be function-t rode fully inserted except ally tested. s as specified in 3.4.3.1. i \\*= ' 2. At least once during each operating cyclat

s. Check that the setting of the system relief valves to 1425 + 75 pois.

b. Manually initiate the eye-tem, except explosive valves.

] Pump boron solution through the recirculation path and ) back to the Standby Liquid l Control Solution Tank. Mini-I num pump flow rete of 39 spe i l L 1 l 135 /' ) NW . ~. -.. _.... 3 ..a

l LIT 171kC C0Nu1710N5 FC'. OPERATION SURVEILLANCE REQUIREMENTS 3.4 STAwngY L1 QUID CONTROL SYSTIM 1.4 STANDBY LIOUID CONTROL SYSTDi against a system head of 1275 psiB shall be w. verified. After pumping baron solution, the sys-tem shall be flushed with de-ineralized wa ter. c. Manually initiate one of the Standby Liquid Con-trol System loops and pump demineralized water into the teactor vessel. This test check explosion of the charge associated with the tested loop, proper operation of the valves, and pump opera-bility. Replacement charges shall be selected such that the age of charge in service shall not exceed five years from the manufacturers assembly date. d. Both systems. includine both explosive valves. [ shall be tested in the course of two operatine, cycles. g B. Operaty n with Imoperabit B. Surveillance with Inonerable Cnaponente: Comoonente: h From and after the date 1. When a component is round to that a redundant compo-be inoperable. Its redundant nent is made or found to component shall be demon-be inoperable, Specifica-strated to be operable tion 3.4.A.1 shall be con-innediately and daily there-sidered fulfilled and con-after until the inoperabic tinued operation permi ted "*P'"*"" i' P'if'd* 2 provided that the component is returned tu an operable condition within seven Sys. 136 u w_ - E A r b e.d -

EstLTINf*. cnem! TION % FOR Or, ERA r10N, SURVEILLANCE REQUIRD4ENTS 3.4. staunsY LIQUID ODNTROL SYSTEM 4.4 STAND 8Y LIQUID C00fTROL SYSTEM / \\ / C. Sodium Pentaborate Solution C. Sodium Pentaborate solution At all times when tne Standby The following teste shall be Liquid Control System is re-performed to wrify the avail-quired to be operable the foi-ability of the Liquid Control i loving conditions shall be met: Solution: 1. The net volume - concentra-1. Volume: Check at least tion of the Lig sid Control asce per day. solution in the liquid een-trol tank shall be mata-2. Temperature: Check at tained as required in least once per day. Flaure 3.4.1. 2. The temperature of the 3. Concentration: Check at liquid control solution least once per month. shall be maintained above Also check concentration the curve shown in Figure any time water or, boron 3.4.2. This includes the is added to the solution piping between the standby ,i or solution temperature liquid control tank and to below the temperature the euction inlat to the required in Figure 3.4.2. pumpe. D. If specificatter 3.4.A throush C cannot be met, the reactor [; shall be placed in a Shutdown Condition with all operable control rods fully inserted V within 24 hours. l 137 ( V y .. _.. m...a_ ~. w- -a

• • w A-

1 SEP 191984 LIMITING CONDITIC33 FC3 CP 3ATION GURVEII1X::CZ Itl!Qt7TREMEMas i 3.7 COhM INME?PF SYST2MS 4.7 Com*AliOtENT SY3TZ48 B. Standbv Gao Treatment 8vetes B. Standby Gao Treetsent Svatem 1. Except as specified in Specification 1. At least once per 3.7.B.3 belcw, all year, the f ollofing three trains of the conditions shall be stancly gas treatment demonstrated. system a. Pressure drop across the shall be combined MEPA operable at all times filters and when secondary charcoal containment integrity adsorter banks is required. is less than 6 inches of water at a ficw of 9000 cfm (+ 105). b. The inlet boaters cm each circuit are teetet in ac::orcance with AUGI H510-1975 and a're capable of an ou*get of at 4 least 40 kw. c. Air dietribution is unifom within 2C% across d2PA filteru and charcoal adsorbers. 216 -=

REVISED: 5-4-79 (ITING CONDITION 8 FOR OPERATION SURVEILI.ANCE REQUIREMENTS 4.7 CONTA;HMENT FYPT!ws 3.7 CO NT A I NM EN" SYST EMS 2. a. The tests and 2. a. The results of sample analysis the in-place of Specification cold DOP and 3.7.B.2 shall be halogenated performed at hydrocarbon least once per tests at 2 105 operating cycle design flow on or once every 18 REPA filters and months whichever charcoal occurs first for adsorber banks standby service shall show 2995 or af ter every DOP removal and 720 hours of 2991 halogenated system operation hydrocarbon and following removal when significant tested in painting, fire accordance with or chemical ARSI H510-1975. release in any (r' ventilation tone (, b. The results of communicating laboratory with the system. carbon sarple analysis shall b. Cold DOP testing show 2901 shall be radioactive Performed af ter i methyl iodide each complete or removal when partial tested in replacement of accordance with the BEPA filter ANSI N510-1975 bank or after (130*C, 951 any structural R.E.). c. Systen shall be the system shown to operate housing. within +10s design flow. c. Ralogenated hydrocarbon l testing shall ba performed af ter J each complete or partial replacement of the charcoal adsorter bank or after any s tructural maintenance on 237 the system housing. Amendment tio. 50

SEP 281983

,IMITING CONDITIONS FOR OPERATION SURVETI1ANCE REQUIREMENTS l

3.7 coMAIMMENT SYSTJJ1E 4.7 COMAI14 TENT SYSTIP8 d. Each train shall be operated a total of at least 10 hours 1 every month. e. ' rest sealing of gaskets for housing doors shall be performed utilizing chemical smoke generators during each test perf ormed for compliance with specification 4.7.B.2.a and Specification 3.7.B.2.a. 1 ( I 3. From and af ter the 3. a. Once per date that one train Operating cycle i of the standby gas automatic i treatment sycten is initiation of made or found to be each branch of inoperable for any the standby gas reason, reactor treatment system operation and fuel shall be i handling is demonstrated permissible only from each unit's during the succeeding controls. 7 days unless such l circuit is sooner b. At least once i made operable, per year manual I provided that during operability of i such 7 days all the bypass valve 1 active components of for filter the other two standby cooling shall be i gas treatment trains demonstrated. shall be operable. J 1 238 ww 1 ass

N6 LIMITING CONDITIONS 7CR OPERATION SURVEILI.ANCE 'REQUIRIMENTS O s J J. 7 CotrrAMMENT SYSTD95 4.7 CONTA INM EN1 S YST EMS c. When one *:ain of the standby gas tr ea tmant system becomes inoperable the other two trains shall be demonstrated to be operable within 2 hours and daily thereafter. 4 If thcee conditione cannot be met, the reactor shall be placed in a condition f or which the standby aas trea tment sys tem is not required. A ( l 1 \\ 239 O AMENDMENT No. 50 ._._..-,..:........---.---~""'~'^^^^^^"~~ ^^ ?

NOV 161981 a,V-LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3 9 AUXILIARY ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTEM Applicability Applicability

plies to all the auxiliary Applies to the periodic testing electrical power system.

requirements of the auxiliary electrical systems. Objective Objective To assure an adequate supply of Verify the operability of the electrical power for operation of auxiliary electrical arstem. those systems required for safety. Specification Specification A. Auxiliary Electrical Equipment A. Auxiliary Electrical Equipment A reactor shall not be started

1. Diesel Generators up (made critical) from the cold condition unless four units 1

a. Each diesel generator shall and 2 diesel generators are be manually started and operable, the requirements,of loaded once each month to s

3 9.A.4 through 3 9.A.7 are met demonstrate operational f and two of the following off-readiness. The test shall site power sources are available continue for at least a one as stated with no credit taken hour period at 75% of rated for the two 500-kV Trinity load or greater. lines. During the monthly - Eoih 161-kV lines, both common sta-generator test, the. diesel tion service trar.sformers, and both generator starting air start buses provided the second compressor shall be checked source is from the 500kV system. for operation and its ability to recharge air - The unit 1 unit station service receivers. The operation transformer TUSS 1B is available. of the diesel fuel oil If the unit 2 station service transfer pumps shall be transformers is the second demonstrated, and the source, a minimum of tuc 500-kV diesc1 starting ti=e to lines must be available. reach rated voltage and speed shall be legged. - The unit 2 unit station service transformer TUSS 2B is

b. Once per operating cycle, a available. If the unit 1 test will be conducted station service transformers is simulating a loss of the second source, a minicum of.

offsite power and similar two 500kV lines must be conditions that would exist /,, available. With the presence of an actual safety-injection - Both 161-kV lines, both cooling signal to demonstrate the 292 . h.--- - - - - --

K. NOV 18 E81 r LIMITI G CC:!DITIINS FOR OPERATICN SURVEILLANCE REQUIREMEN*S 4 9 AUXILIARY ELECTRICAL SYSTEM 3 9 AUXILIARY ELECTRICAL SYSTEM tower transformers, and the bus following: tie board energized and capable

1. Oeenergi:ati:n of the of supply power to the units 1 and 2 shutdown boards, provided emergency buses and 1:a:

shedding from the that the second source is from emergency buses. the 500-kV system.

2. The diesel starts frc:

- Both 41-kV lins, ont CC : n ambient condition on tne station service transformer supply-auto-start signal, ing a start bus, and one cooling e i es t e em g tower transformer (through the bus g tie board) provided the cooling connected loads, tower transformer is not parallel tc energi:es the auto-the common station service trans-connected emergency former and provided that the second loads through the lea: source is frc= the 500kV system. sequencer, and operates ~' for greater than or A reactor shall not be started equal to five minutes while its generator is up (made critical) from the Hot loaded with the Standby Condition unless all of the following conditions are emergency loads. g. satisfied:

3. On diesel generator breaker trip, the ich..

1. One of the off-site power are shed from the sources listed above is available and capable of emergency buses and the diesel restarts on the supplying auxiliary power to the shutdown boards.

auto-start signal, the emergency buses are

2. Three units,1 and 2 diesel energized with generators shall be permanently connecte:

loads, the auto-operable. connected emergency loads are energized

3. An additional source of power consisting of one of the through the load sequencer, and the following:

diesel operates for

a. A second off-site power greater than or equa' to source available and five minutes while 1 s capable of supplying power generator is loaded with to the shutdown boards.

the emergency loads.

b. A fourth operable units 1

c. Once a month the qu~sntity and 2 diesel generator.

of diesel fuel available shall be logged.

4. Buses and Boards Available

d. Each diesel generator sb .

be given an annual (deleted) inspection in accordar a. N 293

E LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS g#" 3 9 AUXILIARY ELECTRICAL SYSTEM 4.9 AUXILI ARY ELECTRICAL SYSTEM with instructions based on

b. The units 1 and 2 4-kV the manufacturer's shutdown boards are recommendations.

energized.

e. Once a month a sample of

c. The 480-V shutdown boards diesel fuel shall be associated with the unit checked for quality. Tne are energized.

Quality shall be within acceptable limits specified

d. The units 1 and 2 diesel in Table 1 of the latest revision to ASTM D975 and auxiliary boards are energized.

logged.

e. Losa of voltage and degraded

2. D. C. Power System - Unit voltage relays operable on Batteries (250-Volt) Diesel 4-kV shutdown boards A, B, C, Generator Batteries (125-Volt) and Shutdown Board Batteries and D.

(250-volt)

f. Shutdown busses 1 and 2

a. Every week the specific energized.

gravity and the voltage of

g. The 4807 Rx. HOV Boards the pilot cell, and D&E are energized with H-G temperature of an adjacent i

cell and overall battery i Sets IDN, IDA, IEN, and TCA in service. voltage shall be measured and logged. /'

5. The 250-volt unit and shutdown board batterien and

b. Every three months the measurement shall be made a battery charger for each battery hoards are operable.

of voltage of each cell to nearest 0.1 volt, specific gravity of each cell, and

6. Logic Systems temperature of every fifth cell. These measurements

a. Commnn accident signal shall be logged.

lo61c system is operable.

c. A battery rated discharge (capacity) test shall be

b. 480.V load shedding logic performed and the voltage, system is operable.

tine, and output current measurements shall be

7. There shall be a minimum of logged at intervals not to 103,300 gallons of diesel exceed 24 months.

fuel in the standby diesel generator fuel tanks. 4 293a e

• 1. +M*

-N .4 h w e M - - -

V' 1,I_MITING CONDITIONS' FOR OPERATION SURVEILI.ANCE REQUIREMENTS REP 0 3 C 8', 3 9 AUXILI ARY ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTEM je

3. Logic Systems

a. Both divisions of the common accident stF,n11 logic system Shall be teste* ciery 6 e r.'_hn t.

demonstrate that it will function on actuation of the core spray system of each reactor to provide an automatic start signal to all 4 units 1 and 2 dienel generators.

b. Once every 6 months, the conditinn under which the 480-Volt load shedding log.c system is required shall be simulated using pend.1nt test switches and/or pushbutton test switches to demonstrate that the load shedding logic system would initiate jr load shedding signals on the diesel auxiliary boards, reactor MOV boards, and the 480-Voit shutdown boa rit s.

4. Undervoltage Relays

a. (deleted) e

b. Coce every 6 monthn, the conditions under wh',ch tha loss of voltage and degraded voltage relays are required shall be simulated with an undervoltage on each shutdown board to 294 e

w

• - S a he as ake. gg 9,

,44 ,g.g gg ,g.y, m,

I 9 03 C81 SURVEILLANCE REQUIRF.MENTS LIMITING CCtIDITIONS FOR OPERATION 4.9 AUXILI APY ELECTRICAL SYSTEM i ,.9 AUYILI ARY ELECTRICAL SYSTEM demonstrate that the associated diesel generator will start.

c. The loss of voltage and degraded the veltage relays whicr. start diesel generators fro = the 4-kV shutdown boards shall be cali-brated annually for trip and reset and the measurements logged. These relays shall be calibrated as specified in Table 4.9. A.4.c.

d. 4-kV shutdown board voltages shall be recorded once evcry 12 hours.

5. 480-V RMOV boards D and E

a. Once per operating cycle the automatic transfer feature for 480-V RHOV boards D and E shall be

.J functionally tested to verify auto-transfer capability. 9 294a I l ++ ~ - - -. ~~'- - v----

~" LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.9 AUXILIARY ELECTRICAL SYSTEM NOV 1 3.9 AUXILIARY ELECTRICAL SYSTEM r B. Operation with Inoperable ,3. Operation with Inoperable Equipment Equipeent Whenever a reactor is in Startup

1. When only one offsite power mode or Run mode and not in a source is operable, all cold condition, the availability units 1 and 2 diesel of electric power shall be as generators and associated boards must be demonstrated to specified in 3 9. A except as be operable immediately and specified herein.

daily thereaf ter.

1. From and after the date that

2. N.A.

only one off-site power source is available, reactor operation is permissible for seven days.

2. From and after the date that

3. When one or the units 1 and 2 the 4-kV bus tie board diesel generator is found to becomes inoperable, reactor be inoperable, all of the CS, RHR (LPCI and Containnent operatien is permissible indefinitely provided one of Cooling) Systems and the the required off-site power remaining diesel generators source is not supplied from and associated boards shall be the 161-kV system through the demonstrated to be operable bus tie board.

Lmnediately and daily thereafter. ,r'

3. When one of the units 1 and 2 diesel generator is

4. When one 4-kV shutdown board inoperable, continued reactor is found to be inoperable, all operation is per=issible remaining 4-kV shutdown boards during the succeeding 7 days, and associated diesel provided that two offsite generators, CS, and RHR (LPC1 power sources are available, and Containment Cooling) and all of the CS, RHR (LPCI Systems supplied by the and Containment Cooling) remaining 4-kV shutdown boards Systems, and the renaining shall be demonstrated to be three units 1 and 2 diesel operable, immediately and generators are operable. If daily thereafter.

this requirement cannot be met, an orderly shutdown

5. When one shutdown bus is found shall be initiated and the to be inoperable all 1 and 2 reactor shall be shut down diesel generators shall be and in the cold cendition proven operable ic=ediately within 24 hours.

and daily thereafter.

4. When one units 1 and 2 4-kV

6. When one units 1 and 2 Diesel shutdown board is inoperable, Aux. board is found to be continued reactor operation inoperable, the remaining is permissible for a period diesel Aux. board and each not to exceed 5 days, unit 1 and 2 diesel generator

/ ? 295 v s e . 1. D

~ OM O3105tl LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3 9 AUXILIARY ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTEM l provided that two off-site shall be proven operable l power sources are available, immediately and daily an the rer. ining 4-kV theresrter. d shutdown boards and associated diesel generators, CS, RHR (LPCI and Containment l Cooling) Systems, and all 480-V emergency power boards are operable.- If this. I requirement cannot be met, an orderly shutdown shall be initiated and the reactor i. shall be shut down and in the cold condition by the end of the fifth day.

5. When one of the shutdown busses is inoperable, reactor operation is permissible for a period of 7 days.

i

6. When one of the 480-V diesel

</ Aux. boards becomes inoperable, reactor operation l is permissible for a period of 5 days.

7. From and after the date that one of the three 250-Volt unit batteries and/or its associated battery board is found to be inoperable f*or

~ any reason, continued reactor operation is permissible during the succeeding seven days. Except for routine surveillance testing, the NRC shall be notified within 24 hours of the situation, the precautions to be taken during this period, and the plans to return the failed component to an operable state.

8. From and after the date that t

one of the 250-volt shutdown board batteries and/or its I / associated battery board is found to be inoperable for l 296 ~.a.--.-

LIMITINC CONDITIC*iS FOR C"ERATICM SURVEILLANCE eECUTPEMNTS 3.9 AUI LIAST ELEC PICAL SYSnM 4.9 AUTILIA?Y ELECTRICAL SYSU M SEP 03 ;gg; any reason, continued rea:tcr operation is permissible during the succeedir4 five days in accordance with 3 9.E.7.

9. When one division of the Logic System is incpe-able, continued reactor operation is persissible under this condition for seven days, provided the CSCS requirements listed in specification 3 9.3 3 are satisfied. The NP.C shall te l

notified within 24 hours of I the situation, the l precautions to be taken during this peric4, and tne plans to return the failed component to an cperatie state.

10. (deleted) f i

1 1

11. The folicwing limiting conditions for operation exists fer the undervoltage relays which start the diesel generators on the 4-kV shutdown boards.

a. The loss of voltage relay channel which starts the diesel generator for a coeplete loss of voltage en a 4-kV shutdevn beard r.sy be inoperable for 10 days pro-vided the degraded voltage relay channel en that shut-devn board is operable (with< in the surveillance schedule of 4.9. A.4.b.) 1 t g 297 O 3 --. E ' .m..- .e....

SURVFIt.t.f.NCE REQUIRFtEt:TS I,1MpINC CON!)fT10N3 FOR OPr.MATION 1l 0 33tLJA M MCTFml. SMW 4.') pUYU.IARY F.t.ECT y t. S N EM f

b. Tie degradud voltage relay channel which starte the diesel generator for degrad-ed voltage on a 4-kV shutdown beard cry be inoperable f r 10 days provided the loss of voltage relay channel on that shutdown board is operable (within the surveillance schedule of 4.9.A.4.b).

c. One of the three phase-to-Phase degraded voltage relays provided to detect a degraded voltage on a 4-kV shutdown board may be inoperable for 15 days provided both of the fo11 ewing conditions are satisfied.

1. The other two phase-to-phase degraded voltage

/ relays on that 4-kV shut-down board are operable (within the surveillnnee schedule of 4.9.A.4.b).

2. The loss of voltage relay channel on that shutdown board is operable (within the surveillance schedule of 4.9.A.4.b).

d. The degraded voltage relay channel and the lose of voltage relay channel on a 4-kV shutdown board may be inoperable for 5 days provid-ed the other shutdown boards and undervoltage relmye are operable. (Within the 297a G

.+ ..we=== . LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3 9 AUTILIARY ELECTRICAL SYSTEM 4.9 AUTILIARY ELECTRICAL SYSTEM , /" surveillance schedule of 4.9.A.4.b). ggEP 03 t9BI i

12. When one 480-volt shutdown board is found to be inoperable, the reseter will be placed in hot standby within 12 hours and cold shutdown within 24 hours.

13. If one 480-V RMOV board M-G

' set is inoperable, the reactor may remain in operation for a period not to exceed seven days, provided the remaining 480-V RMOV board n-g sets and their associatec loads rer.ain operable.

14. If any two 480-V RMOV beard M-G sets become inoperable, the reactor shall be placed in the cold shutdown condition within 24 hours.

J'

15. If t5e requirements for operating in the conditions specified by 3 9.B.1 throu6h l

3.9.B.14 cannot be met, an orderly shutdown shall be initiated and the reactor shall be shutdown and in the cold condition within 24 hours. / 297 b b ~.e a.. =

Si!P O31981 1.IMITING CONDITIONS TQ OPERATION SURVEILLANCE REQUIREMENTS 3 9 AUIILIARY ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTEM C. Operation in Lold Shutdown Whereve-th? -'?eter is in cold snuccoan condition with 4 irradiated fue' in the reactor, the availability of electric power shall be as specified in Section 3.9.A except as specified herein.

1. At least two units 1 and 2 diesel generators and their associated 4-kV shutdown boards shall be operable.

2. An additional source of power i

consisting of at least one of the following:

a. The unit 1 or 2 unit station service g.

transformers energized.

b. One 16?-kV transmission line and its associated 4

3 j common station service transformer energized.

c. Either 161-kV line, one cooling tower transformer 4

and the bus tie board energized and capable of supplying power to the units 1 and 2 shutdown boards energized.

d. A third operable diesel I

generator.

3. At least one 480-V shutdown board for each unit must be operable.

4. One 480-V RMOV board motor-generator (M-G) set is required for each RMOV board (D or E) required to support operation of the RHR system

/' in accordance with 3 5.B.9 O 298 = - - ~_+

2 , o mg 5 o V t

• 1/2 at o

[ = mC* (E = 1/2 my a = (Vf - 1 )/t A = AN A=Ae** 3 3 PE = ogn Vf = V, + at. . * = e/t 1= an2/t1/2 = 0.693/t1/2 2 W = v :P nD 1/2 l 1 A* [(t1/2)

• II I) 4 D

t.E = 931 un m = V,yAo -Ix Q.=ph I"Ie n Q = mCpat h = UA A T I=Ien I = I,10-*/D L Pwr = W ah f TVL = 1.3/u sur(t) P = P,10 HVL = -0.693/u P = P e /T t o SUR = 26.06/T SCR = S/(1 - K,ff) CR, o S/(1 - K,ffx) SUR = 26o/t* + (s - o)T CR (1 - K,ffj) = CR II ~ "eff2) j 2 T = (t*/s) + [(8 - oVIo] M = 1/(1 - K,ff) = CR /CR, j T = 1/(o - s) M = (1 - K,ffo)/(1 - K,ffj) T = (s - o)/(Io) SOM ='( - K,ff)/K,ff a = (K,ff-1)/K,ff = g,ffA,ff t= 10 seconcs I = 0.1 seconds-I o = [(t=/(T K,ff)] + [i,ff (1 + IT)] / d Ijj=Id P = (reV)/(3 x 1010) I d) 2,2 2 gd j 22 2 I = cN R/hr = (0.5 CE)/d (m,t,73) R/hr = 6 CE/d2 (f,,g) Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem. 1 curie = 3.7 x 1010aps 1 ga:. = 3.78 liters I kg = 2.21 lbm 1 ft* = 7.48 gal. I np = 2.54 x 10 6tu/hr Density = 62.4 1M/ft3 1 mw = 3.41 x 10 Stu/hr Density = 1 gm/cW lin = 2.54 cm Heat of vaporization = 970 Stu/lem 'F = 9/5'C + 32 Heat of fusion = 144 Btu /lbm

• C = S/9 (*F-32; 1 Atm = 14.7 psi = 29.9 in. Hg.

1 BTU = 778 ft-lbf I ft. H O = 0.4335 lbf/in. s 2 e = 2.718 \\ )

EVENT STEAM LNGE FREAK NCIDE C/W POWgg_ 100% i l/ l i l J r 3 k ' * = i = / \\ \\ t w / A e .l ( [ h / / 2 1 \\ \\ / ,f "n-- j 6

• =

, p 4 O 20 40 EO 40 100 120 0 o S 12 16 l CORE FLOW TOTAL STE AM FLOW l (X 1 MILLION Ibs/hr) (X 1 MILLION lbsthr) i i l 4 l I I / t I 960 970 990 1010 1030 1050 0 4 8 12 16 RE ACTOR PRESSURE TURBINE STE AM FLOW (P810) ( X 1 MILLION Ibs/hr)

_ ~ _ a EVENT STEAM LINE GREAK IN DE D10 00WER 100% I en l m1 A t'""""" ^ m l 1 \\ 2 4 l l 1{ 16 "4 ~ ~ , ~ ~. r 12 4 . ~ ~ ~ ' J 11 I i ~ g 1 s L ) um-. '~~---...,,,,,,""""=- 3 g a ~ g - 15 I [, ,,,-.-----~~~ y y L 9 14 a ~ 1% n ----~~~ \\ l \\ m l l l l j O 300 600 e00 1000 1500 0 10 20 30- 40 50 60 A RE ACTOR PRESSURE RE ACTOR VESSEL LEVEL (PSIG) (IN C H E S) q 7 23 l j.- me 2 1 ...---w r.... 2, .---...-3 22 ~ T t = a l f I -i f / j m to m to g 17 ] ar l

• >~

55 ~' 18 3 0 25 50 75 100 125 0 4 a 12 16 l{ APRM TOT AL FW FLOW (% POWER) (X 1 MILLlON lbs/hr)

• e i

m m

g 1 RECtR. FLOW FAILED Hi?H pgwen 70% 1 gyszy r 4 I 3 q - ~~,,,,, p 2 5 I, / p s '~ ,,-.a 1 0 20 40 so so 100 120 0 4 a 12 18 CORE FLOW TOTAL STEAM FLOW (a to*1bs/ht) (a 10* tte/hr) 'l 1 6 'el ei' t' 9 10 ? l r I _i__ 1 l l 7 / 9 S r y i, g .pr =-- gp 4 e e ^"". i 6 -~~~ 8 i 950 970 990 1010 1030 1050 0' 4 8 12 18 r ~: REACTOR PRESSURE TURBINE STEAM FLOW { (PSIGI (a 10e Ibe/hrt i 5 g L i

5B syset RECIR. PLOW FAILED MtGH powen 70% r i y46* yan* se l L f %* % g e % g 6 -w a" 2.L..,_ 13 b s* p** 11 k } I uoo s oo o 50 no REACTOR PHES$URE REACTOR VEW WE (PSlG) (seecHEs) < I.7,%, I .t ~- ^ 54 i i ' k 1 s, % g N gg f s 'pr' h i 4, 11 p I hii \\ t! l i i i i k la is e dP ~ ~ 'E '" ** =, 16 17 o 25 so 75 too 125 o 4 s 12 M TOTAL FW FLOW 3 is ic' Ite/ht)

e e 'I 115 12,,MOST LIMITING BUNDLES FOR NFLCFR FOR MFLPD FOR MAPRAT PLCFR LOC MCPRl CPRLIM MFLPD LOC MRPD RI'DLIH MAPRAT LOC ' MAPLHCR LINLHGR >569 17-36 2.195 1.250 0.572 21-48-19 7.66 13.40 0.572 19-48-19 5.79 10.12 ~, ,569 43-26 2.195 1.250 0.572 21-14-19 7.66 13.40 0.572 41-14-19 5.79 10.11 ,569' 43-36 2.197 1.250 0.571. 39-14-19 7.66 13.40 0.572 19-14-19 5.79 10.13 ,568 17-26 2.200, 1.250 0.571 39-48-19 7.65 13.40 0.172 41-48-19 5.78 10.11 .565 15-38 2.213 1.250 0.569 19-50-19 7.63 11.40 0.553 9-36-20 6.64 12.00 565 - 45-24 2.213 1.250 0.569 19-12-19 7.62 13.40 0553 51-26-20 6.65 12.00 ,564 45-38 2.215' 1.250 0.569 41-12-19 7.62 13.40 0.552 51-36-20 6.63 12.00 ,565 15-24 2.218' 1.250 0.568 41-50-19 7.61 13.40 0.552 9-26-20 6.62 12.00 ,562 21-48 2.224l 1.250 0.552 9-36-20 7.40 13.4,0 0.552 21-48-19 6.67 12.10 ,562 39-14 2.226 1.250 0.552 51-26-20 7.40 13.40 0.551 19-50-19 6.67 12.10 56I 21-18 2.227 1.250 0.552 51-36-20'7.39 13.40 0.551 21-14-19 6.67 12.10 561 39-44 2.227 1.250 0.551 9-26-20 7.39 13.40 0.551 39-14-19 6.67 12.10 IE NUMBER OF BUNDLES WITH MFLCPR GREATER THAN 1.0 = 0 15 NUMBER OF BUNDLES WITH MFLPD GRE ATER Til AN 1.0 = 0 lE, CUMBER OF BUNDLES WITH MAPRAT GREATER TIIAN 1.0 = 0 'I l i TRANSPARENCY #4a '.' t 'I h -ll;' ~ .s t-y.. j l .. g.... by,

p..

l-

ito 'S-24-84 TIME 0800 BROWNS FERRY - 2 SEQ. No. 5

• • • PERIODIC NSS CORE PERFOKHANCE LOG ***

Oc cico' 1 2~ 3 4 -5 6 7 8 9 10 11 12 CHWT 2103. IAL REL PWR 0.42 0.75 0.81 0.88 0.96 1.06 1.13 1.21 1.27 1.34 1.25 0.90 PCT PWR 63.9 310] REL PWR 0.88 1.07 0.88 1.09 1.20 1.09 0.89 l'. u / 0.88 CMWE 662.5 BC CEL PWR 1.08 1.24 1.24 1.22 1.26 1.23 1.03 0.40 CMFCP 0.56% RM CAF 0.98 0.98 0.98 0.98 0.98 0.98 CHFLPD 0.572 CHAPR 0.57. GICJ 1 2 3 4 5 6 7 8 9, CMPF 2.18i LCF3 O.561 0.548 0.562 0.569 0.555 0.569 0.562 0.548 0.561 CAEQ 0.082 C 21-18 35-10 39-14 17-36 25-30 43,-26 21-48 35-52 39-44' CAQA 0.09; SU a 0.1209 0.1215 0.1212 0.1204 0.1211 0.1204 0.1212 0.1215 0.1209 CAVF 0.249 F 1.44 1.40' 1.42 1.46 1.42 1.46 1.42 1.40 1.44 CAPD 31.301 LP3 0.572 0.539 0.571 0.552 0.511 0.552 0.572 0.539 0.571 CRD 0.00c C. 21-14-19 35-10-19 39-14-19 9-36-20 25-30-16 51-20-20 21-48-19 35-52-19 39-48-19 CRSYN 2. FL 2.18 2.06 2.18 2 11 1.95 2.11 2.18 2.06 2.18 PR 1011.77 PEAT 0.572 0.540 0.572 0.553' O.502 0.553 0.572 0.540 0.572 DPC-M 17.61 3 19-14-19 25-10-19 41-14-19 9-36-20 25-30-16 51-20-20 19-43-19 35-52-19 41-48-19'DPC-C 21.26 ra - 1.60 1.87 1.60 1.92 1,77 1.92 1.60 1.87 1.60 RWL 32.98 DHS 17.32 ] WFW 8.10 [LSD SENSORS 2 WD 34.68 7 WTSUB 102.63 ~ FAILED LPRM LIST BASE Cd1T CODE WTHE -1.00 WT 102.15 f,.. 1609,A,b 2409,5,1 2409,D,2 4008,C,2 PCTWTR 99.7 4809,A.1 241.7,D,5 0825,C,2 2425,C,1 WTFLAG 2.0 '4825,C.2 4825,D,2 0833,C,2 1633,C,2 ITER 1.0 3233.A,1 4033,C.2 4833,D,2 2441.C,2 IREC 0.0 f. 2441.Dl2 3241,8,1 3241.C,1 4041 D,2 IEQL 0.0 5h 1649,A.1 2449,A,1 2449,E.1 2449,C,2 IXYFLG 0.0 f' 4849,C,1 14 a l C.. TRANSPARENCY v4 d '. :. lU: l f..-j 0:1-p... s%:E p c p A ,k.-

e e [J jggg f e g Fiyee, r.io I i -.___e,,-

LESSON NOTES p.g. 3,79 LESSON CUTLINE 11 0 l 10 0 r i 90 / G ~ 80 h c 70 Y 60 j q FORCED CIRCU L ATION L 50 Cw 3Oa 40 30 NATURAL CIRCULATION 20 10 s O O 10 20 30 40 50 60 ~ 70 80 90 100110 120 CORE FLOW (%) Pigure 3-JJ. Steady-State Natural Circulation Characteristics of a BWR

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Page 22 of 31 08/01}E Q Lesses Plan 23 r A T

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W 'O.':'l !l 4 L 4 446 (B_, s c D<D<D<t=. m..... g3 ; FIGURE,2.7.3 C00$INE TOWER ARRANGEMENT ~

tuhmes.9t*M E8ehstry. Oss4 Emerapy. Santan a F 7 weeer Evop Steem Water E**p 9 teem Wette Ever Bewn T A A A s, s, s, 't 't 's t t s i 32 SAB856 E01602 3305 3305 -0A2 1075.5 1075.5 CADOC 2.1873 2.1873 32'- N 4.N091 Salm 3988 Stes 3A0 18F33 1876A SN61 3.1786 2.1M7 M de 4.12143 98840t 3446 Me6 SAS ISF1A 1879A tale 2 3.148f 2.1994 as as R14744 8A1402 30373 20372 13D4 1068.1 1081 2 SA062 2.1164 2.1426 as te 0.17796 SA1602 1704A 17042 18A5 1965.3 1083.4 0A361 1A801 2.1M2 se 80 02541 0A1603 1207A 1207.6 2S46 10591 1087J 04555 2A391 2.0946 80 70 0.3629 041605 258 3 968 4 38 45 10540 1092.1 0 0745 1.9900 2 0645 70 00 0.5068 041607 6333 6333 48.04 1048A 1006 4 0 0932 1.9426 2 0359 30 90 0.6061 0A1610 468.1 468.1 54.02 1042J 11002 0.1115 13970 2.0C46 50 300 03492 0A1613 350.4 350.4 68 00 1037.1 1105.1 0.1295 13530 1.3825 1g0 118 1.2750 021617 265A 265A 77.98 1031A 1109 3 0.1472 12105 1.9577 110 130 1A927 CA1620 203.25 20326 57.97 1025.6 1113.6 0.1646 1.7693 1.9339 120 130 2.2230 OS1625 15732 157.33 97.96 1019 2 11172 0 1817 1.7295 12112 130 140 22892 0A1629 122.98 123.00 107.95 1014 0 11224 01985 14910 1.8895 140 150 3J18 041634 97.05 97.07 117.95 1008 2 1126.1 02150 14536 13666 150 160 4341 0.01640 77.27 77.29 127.96 2002.2 1130.2 01313 14174 12487 160 179 5393 OA1645 62A4 62.06 137.97 996.2 1134.2 0.2473 1.5822 13295 170 300 7411 841661 90.21 50.22 14SAO 930.2 113L2 0.3631 14400 13111 180 ISO 9.340 041657 40.94 40.M 158.04 984.1 1142.1 0.2787 1.5143 1J934 130 - 200 11.526 0.01664 33.62 3344 168 09 977.9 1146.0 0.2940 14824 IJ764 200 210 14.123 0 01671 27.80 2722 178.15 971.6 !!49 1 03091 1A509 1J600 210 212 14.696 0A1672 2578 26.30 180.17 9703 1150 5 03121 1A447 1.7568 212 220 17.186 0.01678 23.13 23.15 188 23 M 5.2 1153.4 03241 1A201 IJ442 220 230 20J79 0.01685 19364 19381 19833 958.7 1157.1 03388 1.3902 17290 230 240 24.968 041693 16304 16321 20E45 952.1 1160.6 0.3533 L3609 1J142 240 250 29225 041701 13302 13A19 218.59 945A 11644 0J677 1.3323 1J000 250 1 260 35 427 0.01709 11.745 11J62 22SJ6 938 6 1167A 0J819 13043 1.6862 260 270 41.356 0 01718 10.042 10.060 23L95 931.7 1170.6 03960 12769 1A729 270 280 49.200 041726 E627 E644 249.17 924.6 1173A 0.4098 1.2501 14599 290 290 57.550 041736 7.443 7A60 259A 917A 1176.8 CA236 1.2238 14473 290 300 67.005 0 01745 5448 6.466 269.7 9104 11793 CA372 1.1979 1.6331 300 320 77E7 001755 5609 5 E26 223 0 902.5 11E2.5 04506 1.1726 1.6232 310 223 89.6% CC1766 4 695 4 914 2K4 E h4 E. 11552 0 4tG 11477 1.6116 323 343 117.99 0 01737 3 773 3385 311 3 E75 S 1190.1 049 2 14993 1.5552 343 360 153 01 0.01811 2.939 2.957 3323 562.1 1194 4 0.5161 14517 1.5675 363 340 195J3 0A1836 2317 2335 353.6 844.5 1196.0 0.5416 14057 1.5473 383 400 247 26 041564 1 A444 12633 375.1 825.9 1231 0 0 5567 0.9607 1.5274 400 ( 473 30578 0 01E94 1.4S'.a 1A997 396.9 806.2 1203.1 0.5915 02165 1.5063 420 ~ 443 331.54 041926 1.1976 1.2169 419.0 7&5A 1204.4 06161 03729 1A593 443 460 466 9 00196 0.9745 0 9942 441.5 7632 1234 8 06405 08299 1.47GL 460 4E3 565 2 00200 0.7972 0 3172 464.5 739.6 1234.1 0 6648 0.7871 1.4515 483 SGC 650 9 C.C2:4 06545 C E749 457.9 714 3 1202.2 06290 C 7443 1 4333 503 5:3 812.5 0C239 05356 05595 5120 657.0 1199 0 0 7133 07313 1.4146 52;: 543 952 8 0C215 0 4437 04E51 536 8 657.5 1194.3 0.737E O6577 13954 540 SEO 1133 4 0C221 03E51 03571 562.4 625 3 1157.7 0.7625 06132 1.3757 550 553 1326.2 0C223 0.2994 03222 5S9.1 589.9 1179 0 0.7576 0.5673 13553 55s 400 1543.2 04236 0.24?S 02675 617.1 550 6 1167J 02134 0.5196 13333 '4 0 620 1796.9 0 0247 0.1962 02208 646.9 5063 1153.2 03433 OA6S9 13092 623 640 2C59 9 0.0260 0.1543 0.1802 679.1 454.6 1133.7 03666 OA134 1.2821 640 560 23653 0 0277 0.1166 0.1443 714.9 392.1 1107.0 02995 03502 1.2458 640 640 2708 6 0 0304 n nana 0.1112 75E 5 310.1 1066.5 0.9365 0.2720 1.2006 Geo 7c0 30943 0 0366 043e6 0 0752 s22A' 1723 995 2 02901 0.1490 1.1390 700 705.5 320s2 04508 0 0 0508 906.0 0 936 0 1.0612 0 14612 705.5 TABLE A.2 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE) A.3 l

tuhans. 98/lb 8eshelpy.ShWlb Sakepy. Deuhb a F Emergy.Seu/lb Pwas. Temp g,g, g,,, geoem toter Ever Steam We'ar twop tesem Weere Steem Pwes. Psb F pele A A A a o s, og er, 't 't 's t 4s s f g j 0.0686 32.018 0.01602 3302.4 3302.4 0.00 1075.5 1075.5 0 2.1872 2.1872 0 1021.3 0.OsSs &le 3L0f3 SA1002 3945.5 9945.5 3A3 1973.8 3076 3 0.0061 2.1706 2.1766 has left.3 tse als 46.453 SAIGot 2004.7 3004.7 13.50 1967.9 3081A OA271 2.1140 3.1411 1830 lat.7 g33 ) mac 63.100 944003 1526.3 1526.3 3122 8063A 2004.7 93422 SA738 3.1960 31A2 tema gas. i sJe 64.484 0A8604 30N.7 1039.7 32.54 3057.1 3009.7 0.0641 2AIGS SAe09 m as 3032.0 ase See 72369 OA1606 792A 792.1 40.92 1052A 1093J 03799 1.9762 2A M2 4022 1034.7 cAO e.5 79.586 0.01607 441.5 641.5 47.62 1048.6 1996.3 00925 1.9446 2.0370 47A2 1036.9 a5 8.6 85.218 0 03609 540 0 540.1 53.25 1045.5 1098.7 0.1028 1.9186 2A215 53.24 1038.7 0.6 0.7 90 09 0.01610 466.93 466 94 58 10 1042 7 1100.8 0.3 1.8966 2.0083 58.10 1040.3 0.7 i SA 94J8 0.01611 413.67 411.69 62.39 1040.3 1102.6 0.1117 12775 1.9970 6229 1041.7 0.8 9.9 98.24 OA1612 368.41 368.43 66.24 1038.1 11043 0.1264 13605 1.9870 86.24 1042.9 E9 1.0 101.74 0.01614 333.59 333.60 49.73 1036.1 1105.8 0.1326 13455 1.9781 09.73 1044.1 1A 2.0 126.07 0.01623 173.74 173.76 94.03 1022.1 1116.2 0.1750 1.7450 1.9200 94A3 10512 2A 3.0 141.47 0.01630 118.71 118.73 109.42 1013.2 1122.6 0.2009 14854 1.8864 109.41 1056.7 3.0 4.0 152.96 0.01636 9063 90 64 120.92 1006.4 1127.3 0.2199 1.6428 1.8626 120.90 1060.2 4.0 5.0 162.24 0.01641 73.515 73.53 130.20 1000 9 1131.1 0.2349 1.6094 13443 130 18 1063.1 5A 6.0 170.05 0.01645 61.967 61.98 138.03 996.2 1134.2 0.2474 1.5820 12294 13821 1065.4' 6.0 7A 176.84 0.01645 53 634 53.65 144.53 992.1 1136 9 0.2581 1.5587 12168 14421 1067.4 7A ES 182.86 0.01653 47.328 47.35 15027 988.5 1139.3 0.2676 1.5384 12060 15034 1009.2 a.0 SA 189.27 041656 42.385 42A0 156J0 985.1 1141.4 0.2760 1.5204 1.7964 156JB 10702 Et, 10 193.21 0.01659 38.404 38 42 161.26 982.1 1143.3 0.2836 1.5043 1.7879 161.21 10723 10 - 14.696 212.00 0.01672 26.782 26 80 180.17 9703 11bO.5 0.3121 1.4447 1.7568 180.12 1077.6 14.696 15 213.03 0.01673 26.274 26.29 181.21 969.7 1150.9 0.3137 1.4415 1.7552 181.16 1077.9 15 30 227.96 0.01683 20.070 20.087 196.27 960.1 1156.3 03358 1.3962 1.7320 196.21 1082A to 30 250.34 0.01701 13.7266 13.744 218.9 945.2 1164.1 0.3682 1.3313 1.9995 218A 1087.9 30 40 267.25 0 01715 10.4794 10.497 236.1 933.6 1169A 0.3921 1.2844 1A765 236 0 1092.1 40 50 281.02 0.01727 8.4967 8.514 250.2 923.9 1174.1 0.4112 1.2474 JA586 250.1 10953 50 80 292.71 0.01738 7.1562 7.174 262.2 915.4 11774 0.4273 1.2167 1A440 262A 1038.0 00 70 302.93 0A1748 6.1875 6.205 272.7 907A 1180.6 0.4411 1.1905 1A316 272.5 1100.2 70 to 312.04 0.01757 5 4536 5.471 232.1 900.9 1183.1 0.4534 1.1675 14208 281.9 1102.1 30 90 320.28 0.01766 42777 4.895 290.7 894.6 1185.3 0.4643 1.1470 1.6113 290.4 1103.7 90 100 327A2 0.01774 4.4133 4.431 298.5 888.6 1187.2 0.4743 1.1284 1A027 298.2 1105.2 300 120 34;.27 0 01789 3 7097 3.728 312.6 877.8 1190 4 0 4919 1.0960 1.5879 312.2 1107.6 120 140 353 04 0 01603 3 2010 3 219 3250 BSE O 1193 0 0 5071 1.0681 15752 324 5 1109.6 140 160 363 % 0 0;615 2.E155 2 634 33E 1 6550 1195 1 0t2LE 1.0435

1. % 41 335.5 1111.2 160 380 373 03 0.01827 2.5129 2.531 346.2 850 7 1196.9 0 5328 1 0215 1.5543 345.6 1112.5 180 3'l.80 0 01829 2.26S9 2.287 355.5 842.8 1198.3 0 5438 1.0016 1.5454 3542 1113.7 200 200 s

250 400 97 001865 1.8245 1.8432 3761 825 0 1201.1 0.5679 0 9585 1.5264 3753 1115.8 250 300 417 35 001859 1.5233 1.5427 394 0 SOS 9 1202 9 05S82 09223 1.5105 392.9 1117.2 300 350 431.73 0 01913 1.3064 1.3255 409.8 794 2 12440 0 6055 08909 1.4968 406.6 1118.1 350 400 444 60 0.0193 1.14162 1.1610 424.2 760 4 1204.6 0 6217 0.8630 1.4847 422.7 111F 7 400 450 4t6.28 0 0195 1.01224 1.0318 437.3 767.5 1204.8 0 6360 0.8378 1.4738 435.7 1118.9 450 500 467 01 0 0199 0 90787 09276 449.5 755.1 1204.7 06490 0.814S 1.4639 447.7 1118 8 500 55? 47194 00199 0521E3 0 Sale 400 9 743 3 1204 3 0 6511 07936 1.4547 4 55.9 Ills E 550 ECO 4E

• 10 0 C20; O74962 0.7E95 471.7 732 0 1203 7 06723 0 7738 1.4461 469 5 11IE.2 600 700

.503 08 0 0205 0.63505 06536 4916 710 2 1201.8 06928 07377 1.4304 4S8.9 1116 9 700 i 833 518 21 0 0209 0 54809 05690 509 8 689 6 1199 4 0 7111 0.7051 1.4163 506 7 II15.2 800 900 Ei! 95 0 0212 0 47968 05009 526 7 669 7 1196 4 07279 06753 1.4032 5232 1113.0 900 1000 5 ~4.5B 0.0216 042435 0 4460 542.6 650 4 1192.9 07434 0.6476 1.3910 53:;4 1110.4 1000 1100 5EE 2d 0.0220 0 37803 0.4006 557.5 631.5 1169.1 07573 0.6216 1.3794 553.1 1107.5 1100 1230 i M7.19 0 0223 0 34013 0.3625 571.9 613.0 1184 8 07714 0.5969 1.3683 5669 1104.3 1200 1500 577A2 0 0227 0.30722 0.3299 585.6 594.6 1180.2 0.7843 0.5733 IJ577 580.1 1100 9 1300 1400 $27.07 0 0231 0.27871 0 3018 5988 5765 1175.3 0.7966 0 5507 1.3474 b92.9 1097.1 3400 3500 596 20 0 0235 025372 0.2772 611.7 550 4 1170.1 0.8035 0 5233 1.3373 605 2 1093.1 1500 2000 635 80 0.02L 7 01626f., 0.1883 672.1 466.2 1138.3 0 8625 0 4256 1.7881 6624 10GS 6 2000 2500 65d ll 0 0286 0 10209 0.1307 731.7-361.6 1093.3 0.9139 0 3206 1.2345 718.5 1032.9 2500 3000 69513 0 0343 0 05073 0.0850 801 8 218.4 1020.3 0 9728 0.1891 1.1619 782 2 973.1 3000 3208.2 70147 0 050B 0 0 050a 906 0 0 906 0 1.0612 0 1.0612 875.9 875.9 3208.2 TABLE A 3 PROPERTIES OF SATURATED STEAM AND SATURATED WATER (PRESSURE) I A.4 1

WJ he psene. 200 200 300 a00 900 e00 700 000 000 1000 1100 1200 1800 1400 as00 MA e 0 0161 392 5 452.3 511.9 571.5 631.1 690 7 f 3 6 68 00 3150 2 I!95.7 1241.8 1288 6 1336.1 1964 5 0 01.74) s 0.3295 2.0609 2.1162 2.1722 2 2237 2 2798 2J144 e2:41 Ps.14 esse 188.N IMJI 886.16 ISB.aB 390A1 968.94 173AS ISS.M 897.39 333A2 323A3 ggBAS s e esel Slasa_1994A 3241J 13082 lag 6A 13e43 Mass 1483.7 ISMJ 39067 38394 333.3 SpeSA agasa D6224) s 0.1296 14716 1AMG SAMS 2A400 2AB32 2.3369 2.1776 22160 92621 SJ006 SJ1M 3J000 2381 I e 00161 30 84 44 98 51.03 57A4 6323 SD 00 M 98 80.M SE 91 92 87 98A4 les 30 110.76 116.7 30 6 68 02 1144 6 1193 7 1240 6 12873 1335.5 1954 0 1433 4 1483.5 15M 6 1506 6 1639 5 1993 3 1 (192.21) s 0.1295 1J928 1A593 1.9173 1.9692 2.0166 2.0603 2.1011 2.1394 2.1757 2.2101 2.2430 22744 2) e 0A161 0.0166 29399 33.M3 37.905 41386 45378 49M4 S3.946 57.926 61D05 65A82 68358 73A3 35 GSAs MSOD 1192.5 32393 1387J 3335.2 1383A 3433.2 3403.4 ISM.5 1946.5 1639.4 less.2 174 l pl3AS) s 0.1295 02940 1A134 14720 13242 13717 2A155 2At63 2AB46 2.1309 2.1653 2.1982 2J397 22599 SJego e 0.0161 0.0166 22.3M 25.428 28457 31A66 M.465 37A58 40447 43.435 46.420 49405 52 JOB 953 30 t GS.05 168 11 1191.4 1239,2 12863 13M.9 1383 5 3432.9 3483.2 1534.3 1546.3 1639.3 1603.1 17472 18033 327.M) s 0.1295 0.2940 1.7005 14397 12921 13397 1.9836 2.0244 2.0628 2AD91 2.1336 2.1665 2.1979 2 v 0.0161 0 0166 11 035 12.624 Id.165 15.685 17.195 18 699 20.199 21.697 23.194 40 6 64.10 168.15 1186 6 1236.4 1285.0 1333.6 1382.5 1432.1 1482.5 1533.7 15853 16388 1992.7 1 (267.2S) s 0.1295 0.2940 1.0992 1.7608 13143 13624 13065 1.9476 1.9060 2A224 2.0569 2AB99 2.1224 2 e OA161 SA166 7257 8.354 9 ADO 30A25 llA3B 32A46 13A50 14AG2 REAS2 16 ABO 17A48 14A46 to 4 48.15 148 20 1181 6 1233.5 1283.2 13323 1381.5 1431 3 3481A 1933 2 1905 3 IS3BA 352.4 1747 0 92.71) s 0.1295 0.2939 1.6492 1.7134 1.7481 1A168 1A612 13024 13410 1.9774 2A190 2.0450 28765 2.1 e 0.0161 0.0166 0.0175 6.218 7.018 7.794 S.560 9.319 10.075 10.829 11.581 12.331 13.081 13.829 14.577 to 68.21 168.24 269.74 1230.5 1281.3 1330.9 1380.5 1430.5 3481.1 1532.6 15843 1638.0 1692.0 1746A 1802 (B12A4) s 0.1295 0.2939 0.4371 1.6790 1.7349 1.7542 1A289 13702 13089 13454 1J000 2.0131 2A446 2.0750 2.1041 o 0.0161 0.0166 0.0175 4 935 5.588 6.216 6233 7.443 8 050 S.655 9258 9A60 10A60 11A00 11A59 300 m 68.26 168.29 269 77 1227.4 1279.3 1329.6 1379.5 1429.7 1480.4 1532A 1984.4 1637.6 1891A 1746.5 190 927A2)s 0.1295 0.2939 0.4371 1 4516 1.7008 1.7506 13036 13451 13839 13205 1.9552 13003 2A199 2A502 2A794 v 00161 0.0166 0 0175 4.0786 4.6341 S.1637 5.6831 6.1929 6.7006 7.2050 7.7096 S.2119 S.7130 9.21M 9.7130 120 6 68.31 168.33 269 81 1224.1 1277.4 1328.1 13784 left 8 14792 1531.4 1983.9 1637.1 18D13 17462 1802A ] (341.27) a 0.1295 0.2939 0.4371 1A286 1A872 1.7376 1.7829 12246 13635 13001 13349 13600 13996 2A500 2.0592 0.0161 0 0166 0.0175 3 4651 3 4526 4.4119 4.8585 5.2995 5.7364 6.1709 6.6036 72349 7A552 7AD46 S.3233 l 140 & 68.37 168 38 269 85 1220 8 1275 3 1326 8 1377.4 1428 0 1479.1 1530 8 1583 4 1636.7 1990.9 17453 j (353 04) e 01295 02939 04370 1.EOS5 1.668E 1.7196 1.7652 1.8071 1A461 1.8828 13176 13503 1.9825 2.0129 2.0421 e 0 0161 0 0166 0 0175 3 0060 3 4413 3 8480 4 2420 4 6295 5 0132 5.3945 5.7741 6.1522 6 5293 ~6.9055 7.281 l 150 6 68 42 168 42 269.89 1217.4 1273 3 1325 4 1376 4 1427.2 1478 4 1530.3 1582.9 1636.3 1690.5 1745.6 1801.4 06355)s 0.1294 0.2938 0 4370 1.5906 1.6522 1.7039 1.7499 1.7919 1.8310 1A678 1.9027 13359 1.9676 1.9980 2.0273, e 00161 0 0166 0 0174 2 6474 3 0433 3 4093 3.7621 4.1064 4.4505 4.7907 5.1289 5.4657 52014 6.1363 6.4704 too 6 68 47 168 47 269 9/ 1213 8 1271.2 1324.0 1375.3 1426.3 1477.7 1529 7 1582.4 1635.9 1640.2 1725.3 1801.2 g (373.C81 s C.1294 0.2938 04370 1 5743 1.6376 1.6900 1.7362 1.7784 1A176 1.8545 1.8894 1.9227 1 9545 1.9049 2.0142, e 00161 0 0166 00174 2 3598 2.7247 3.0583 3.3783 3 6915 4.0008 4.3077 4.6128 4.9165 5.2191 5.5209 5 2219 200 > 68 52 IE8 51 269 96 12101 1269.0 1322.f-1374.3 1425 5 1477.0 15291 15813 1635.4 1689 8 1745.0 1800.9 (3sL60; s 0 1254 02535 0 4359 15593 1.6242 IE77G 1.7239 1.7653 1.8057 1 8426 1.8776 1.9109 1.9427 1.9732 2.0025 CC161 0CICE O0174 0 016E 21504 2 4662 2 6872. 2.9410 3 1909 3 4382 3 6837 3 9278 4 1709 4 4131 4 6546 250 h 68 66 166 63 270 05 3/5.10 1263 5 1319 0 1371.6 1423 4 1475 3 1527.6 1580 6 1634.4 1688.9 1744 2 1800.2 (400 97) s 01294 0.2937 0 4355 0 5567 1.5951 1.6502 1 6976 1.7405 1.7E01 1 8173 1.8524 1.8858 1.9177 1.9482 1.9776 e 0 0161 0 0165 00174 0 0186 1.7665 2.0044 2.2263 2.4407 2.6509 2 6555 3 0643 3.2688 3 4721 3.6746 3 8764 300 4 68 79 156 74 27u 14 375.15 1257.7 1315 2 1368 9 1421.3 1473E 1526.2 1579.4 1633.3 1688 0 1743 4 1799.6 (417.35) s 0.1294 0.2937 0 4307 C5%5 1.5703 14274 1.6758 1.7192 1.7591 3.7964 13317 13652 13972 1.9278 1.9572 - e 0.0161 0 0166 0 0174 0.018G IA913 1.7028 12970 2.0332 2.2652 2 4445 2A219 2.7980 23730 3.1471 3.3205 - 350 4 68 92 1ES 85 270.24 375.21 1251.5 1311:4 1366.2 1419.2 1471A 1524.7 1578.2 1632.3 1887.1 1742.6 1798.9 (431.73) e 0 1293 02935 0.43G7 0.5664 1.5463 1.6077 1A571 1.7009 1.7411 1.7787 13141 13477 12798 1.9105 1.9400. e 0 0161 0 01E6 0 0174 0 01C2 1.2841 1.4763 1.6499 1 A151.1.9759 2.1339 2.2901 2.4450 23987 2.7515 2.90 400 a 69 05 168 97 270 33 375.27 1245 1 1307.4 1%34 I417.C 1470.1 1523 3 1576 9 16)).2 1686.2 1741.9 1798 (444.60) s $ 1293 0 2935 0 4365 0 %C3 1.5282 1.590( l.6406 1.6850 1.7255 1.7632 1.7988 13325 1A647 1.8955 1325 e 0 0161 0 0106 0 0174 0 0186 0 9919 1.1584 1.3037 1.4397 1.5708 1 6992 1A256 1.9507 2.0746 2.1977 2.3200 S00 h 69 32 1E919 270 51 3438 1231.2 1299.1 1357.7 1412 7 1466 6 1520 3 1574 4 16291 1684 4 1740 3 17 %.9 (457.01) s 0 1292 02934 0 4354 05E60 149?! l.5595 1.6123 1 65/8 1 6990 1.7371 1.7730 1.8069 13393 1.8702 1A998 l TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED I WATER (TEMPERATURE AND PRESSURE) A.5 l

Mepuses. Sesupeestem F SWeeIn-gae.essip) 300 300 300 400 800 400 700 000 000 2000 3100 3300 3300 1400 1300 o 0 0161 0 0166 0 0174 0 0186 0 7944 0 94 % 107M 1.1992 1.3008 14093 1.5160 14711 1.??52 1 8784 1.9309 880 6 69.58 169 42 270 70 3Fh eir 121b t 1290 3 1351 8 14083 1463 0 3517.4 1571.9 1627.0 1682 6 1738 8 1795 6 pg,2g s 0.1292 02933 04362 OM57 1.4590 1.5329 1.5444 143b1 16?H 1.73 M 1.7517 3.78h9 3A184 I A4M 13792 e eAiol asles SAIM actes assos avese ae072 later 3.197s IJea3 IJess Saalt RAMF IM47 BA500 338 4 GBAS 388m 27038 SM61 48738 ISSta latte 840B.7 M80.4 1584 4 3000 4 38943 2008.7 1737A 37343 l gang o E1891 32932- 0A300 ESHS 0.6889 1.5000 IA673 14154 IA800 3.M70 3.7335 3.7579 33806 Sall8 33617 i e Osl61 00lu SA174 0.0lM 0.0PO4 04774 0 7823 0A759 0 9631 1.0470 1.1999 12003 IJW85 1.3M9 1A446 l 388 4 70.11 169 88 271.07 375 73 487A8 1273.1 1339.2 1999.1 leM S 1511 4 IM69 1622.7 1678 9 1735.0 1792.9 I (5182.)

• 0.1290 0 2930 0.4358 0.5652 0 4805 1.4869 1.5484 1.5980 1.6413 16807 1.7175 1.7522 1.7851 1 8164 3.8464 i

e 0.0161 OAIM 00174 0 0186 0 0234 O M69 0 6854 0.7783 0 8504 0.9262 0 9998 1.0720 1.1430 1.2131 1.2825 388 e 70.37 170.10 271.26 37534 48723 1260 6 1332.7 1984.4 1452.2 35CS.5 1564 4 1620 6 1677.1 17M.1 17914 $3135) s 0.1290 R2929 OA357 OM49 E6881 1A6H 14311 13822 34263 1.4M2 1.7033 1.7382 3.7713 13024 13329 e 00161 00166 OAl?4 0 0186 0 0204 0 5137 0 4080 0 4475 0.7603 0 8295 0 3966 0.9672 1.0966 1.0901 1.1529 38eo & 70.63 170.33 271A4 375.96 487.79 1249.3 1325.9 13894 1448.5 1504.4 I M t.9 3618 4 1675.3 1732.5 1790.3 (544.58) s 0.1289 0.2928 0.4355 0.5647 0 4476 1A457 1.5149 1.5677 1.6126 1.6530 1.6905 1.72 M 3.7509 1.7905 1.8207 e 0.0161 0 01 % 0 0174 0.0185 0.0203 0 4531 0 5440 0 6188 0 6465 0 7505 0 3121 0 8723 E9313 0 9894 1.0468 1180 4 70.90 170.M 27143 376 08 487.75 1237.3 13188 1384.7 1444.7 1502 4 I M 9.4 1616 3 3673.5 1731.0 1789.0 55628) s 0.1269 0.2927 0.4353 0.5H4 0.6872 1.4259 1A996 3.M42 1.6000 1.6410 14787 1.7141 1.7475 1.7793 12097 e SA161 R0lM OAIM OA185 0.0008 E4016 04905 OMIS 04260 0.6445 0.7418 0.7974 03819 &9065 R9884 138 4 71.16 170.78 27132 3MJO 487.72 1224.2 1311.5 1379.7 1440.9 3449 4 1556.9 3614.2 36714 4739A 1737.6 9 67.19) s 0.1288 0.2926 0.4351 0.5642 OA468 1A061 IAS51 1.5415 1.5883 1A298 S M 79 1.7035 1.7371 1.7691 1.7996 e 0 01(.1 0 0166 0 0174 0 0185 0 0203 0.3176 0 4059 0.4712 0 5282 0 5809 0.6311 0 6798 0 7272 0.7737 0 8195 N00 & 71.68 171.24 272.19 376 44 487.65 1194.1 1296.1 1369.3 1433 2 1493 2 15512 1609.9 1668 0 1726.3 1785.0 OS7A7) s 0.1287 0 2923 0.4348 0.5636 0.6859 1.3652 1A575 1.5182 1.5670 1.6C36 14484 1AS45 1.7185 1.7508 1.7815 e 0.0161 0.0166 0.0173 0.0185 0.0202 0.0236 0.3415 0.4032 0 4555 0.5031 0.5482 0.5915 0.8336 0.6748 0.7153 3488 4 72.21 171.69 272.57 376 69 48740 616.77 1279A 1358 5 14252 1486.9 1546.6 1605.6 1664.3 1723.2 1782J 9 04.87) s 0.1286 02921 0.4344 OM31 0.6851 03129 1.4312 1.4965 1.5478 3.5916 1.6312 1A678 1.7022 1.7344 1.7657 e 0 0160 0.0165 0 0173 0.0185 0.0202 0 0235 0.2906 0.3500 0.3988 0 4426 0.4836 0.5229 0.5009 0 5990 0 6743 3800 a 72.73 172.15 272.95 376.93 487.56 615.58 1261.1 1347.2 1417.1 1480.6 1541.1 1601.2 1660.7 1720.1 1779.7 5 21.32) s 0.1284 02918 0.4341 0.5626 E68*.3 0A109 1A054 1AMS 1.5302 1.5753 1.61 % IA628 1.6876 1.7204 1.7516 e 0.0160 0.0165 0.0173 0.0184 0.0201 CA233 0.2488 0.3072 0.3534 0.3942 0.4320 0.4600 0.5027 0.5365 0.5695 3000 6 73 26 172 60 273.32 377.19 487.53 614 48 1240.9 1353 4 1408 7 1447.1 1536.2 1596.9 1657.0 1717.0 1777.1 (635 80) s 0.1253 0.2916 0 4337 05621 06834 0.8091 1.3794 14578 15138 1.5603 1.6014 1.6391 1.6743 1.7075 1.7359 e 0.01E0 0.0165 0.0173 0 0164 0.0200 0 0230 0 1681 02293 02712 0.306B 0 3390 0.3692 0.3980 0 4259 0 4525 2500 & 74.57 173 74 274.27 377.82 487.50 612.08 1176.7 1303 4 1386.7 1457.5 1522.9 1585.9 1647A 1709.2 1770 4 (468.11) s 0.1280 0.2910 0 4329 0.5609 0 6815 0 8048 1.3076 1.4129 1.47M 1.5269 1.5703 1.6094 14456 1.6796 1.7116 e 0 0160 0 0165 0 0172 0 0183 0 0200 0.0228 0 0982 0 1759 0.2161 0.2484 0.2770 0 3033 0.3282 0.3522 0.3753 3000 A 75 53 17tEB 275.22 37847 487.52 610.06 1060.5 1267.0 13632 1440.2 1503.4 1574.8 1635 5 1701.4 17f 1.8 (EM.33) s 0.1277 0.2904 0.4320 0.5597 0.6796 0 8009 1.1966 1.3692 1.4429 1.4976 1.5434 1.5641 1.621A 1.0061 1 6888 e 0.0160 0 0165 0.0172 0 0183 0.0199 0.0227 0.0335 0.1588 01987 0.2301 0 2576 0.2827 0.3065 0.3291 0.3510 3200 h 76 4 175.3 275 6 378.7 487.5 609 4 800.8 1250 9 1353 4 1433.1 1503.8 1570.3 1634A 1698.3 17612 (705 08) s 0 1276 0 2902 0 4317 05592 0 6768 07994 0 9708 1.3515 14300 1.4866 1.5335 1.5749 14126 1.6477 1.6806 e 0 O1M C 0164 0 0172 0 01E3 0 0199 0 0225 0 0327 01364 0 1764 020!6 02326 02563 0.2784 02995 0 319*. 3500 6 77.2 176.0 276.2 3791 487.6 608 4 779 4 1224 6 1338 2 1422.2 1495 5 1563.3 1629.2 1693 6 17b7.2 s 0.1278 02899 0 4312 0 5555 0 6777 0 7973 0 9508 1.3242 1.4112 14709 1.5194 1.5618 1.600? 1.635S 1.6691 e 0 0159 0.0164 0.0172 0 0182 0.0198 0 0223 0 0287 0.1052 0.1463 0.1752 01994 0.2210 0.2411 0.2601 0.2783 4000 4 7s.5 177.2 277.1 379.8 487.7 606 5 763 0 1174.3 1311.6 1403 0 1481.3 1552.2 1619.8 1685.7 1750.6 s D.1271 0.2a93 0.4304 EM73 0 6760 0 7940 0.9343 1.2754 1.3807 1.4461 1A976 1.5417 1.5812 14177 1.6516 e 0 0159 0.0164 0.0171 0 0181 0A196 00219 0.0268 0.0591 0.1038 0.1312 E1529 0.1718 0.18mo 0.2050 0.2203 8000 a 31.3 179.5 2791 381.2 488.1 604.6 746.0 1042.9 1252.9 1364 6 1452.1 1529.1 16039 1670.0 1737.4 s 0.1265 0.2861 0.4287 0.5550 0 6726 0.7800 0.9153 1.1593 1.3207 IA001 IA582 1.5061 1.5481 1.5863 IA216 e 00159 0.0163 0 0170 0.0160 0 01 % 0.0216 0.0256 OIO397 0.0757 0.1020 0.1221 0.1391 0.1544 0.!a84 0.1817 6800 & 83.7 181.7 281.0 342.7 488 6 602 9 7361 945.1 1188.8 1323 6 1422.3 1505.9 1M20 1654.2 17242 s 0.1258 0.2670 0 4271 0.M28 O M93 0.7826,0.9026 1Al?6 12615 1.3574 1A229 1A745 1.5194 1.5593 1596.2 e 0.0158 0.0163 0 0170 0 0180 0 0193 0.0213 0.0248 0.0334 0 0573 0 031 A 0.1004 0.1160 0.1298 0 1424 a1542 7000 & 86.2 184 4 283 0 384.2 489.3 601 7 729 3 901.8 1124.9 1281.7 1392 2 1482.6 1M31 1639 6 1711.1 s 01252 0.2859 04256 0 5'.07 0 6563 0.7/77 0 8926 1.0350 12055 1.31)I 1.3904 1.44u6 14938 1.53'5 1.5735 TABLE A.4 PROPERTIES OF SUPERHEATED STEAM AND COMPRESSED WATER (TEMPERATURE AND PRESSURE) (CONTINUED) A.6

/V\\ aster -- t A1 5.0 ANSWERS 5.1 a. RPT breaker operation. b. LPCI starts injecting c. Recirc pump discharge valves close (and reactor pressure decreases). d. Level swell due to decreased pressure. e. Condensate booster pumps begin injecting. f. Turbine trips on high reactor level. (0.5 each/ 3.0) REF: Browns Ferry Transients, #21 c< u ~~+ n., :-r "A" Recirc Pump runback [0.25] at <20% TFW [0.25] 5.2 a. b. 1.1 sher power increases pressure c. Rocirc P6mp (Suction) flow increases d. Void collapse due to scram e. FWCS response to level decrease [0.25] and steam flow increase [0.25] (0.5 each /2.5) REF: Browns Ferry Transients, #5 5.3 a. (3) b. (2) c. (1.0) er CL' e,r. (s tan,vg. sows REF: P-1 printout BFN APRM LP (#22), P. 23 BFN Process Computer LP 5.4 a. Increase b. 02 Te?Se T.urvu c. Decrease d. Increase (0.5 each /2.0) REF: BFN Fuel Element Temp. Profile LP, P.4-5

A2 5.5 (b) REF: BFN Phases of Matter LP Steam Tables 5.6 (b) REF: BFN Mitigating Rx Core Damage LP, P.6-6 through 6-10, 6-32 5.7 (c) REF: BFN Mitigating Rx Core Damage LP, P.3-7, 8 5.8 (d) REF: BFN Xenon and Samarium LP, Rev. 1 5.9 Core flow increases due to increased voiding / buoyancy [0.75] but then stabilizes as increased pressure drop cancels out the NC driving head [0.75) (1.5) REF: BFN Mitigating Rx Core Damage, P.3-77-80 5.10 a. (4)

b. (3)

REF: BFN Rankine Cycle LP, p. 7-8, 5 5.11 (b) REF: BFN Reactor Heat Balance LP 5.12 (a) REF: BFN LPRM LP (#21), P.4 5.13 (c) REF: BFN Neutron Slowing Down and Diffusion LP, pp 10, 12 5.14 a. 3 b. 4 c. 2 d. 1 (0.5 ea/2.0) REF: BFN Neutron Multiplication LP, P.14 5.15 (d) REF: BFN Neutron Slowing Down and Diffusion LP, P.2-3 s-i

A3 5.16(c) REF: BFN Fluid Flow / Pump Char., Pump Head, Pump Laws LP, P.4, 6 5.17(b) REF: BFN Control Rod Worth LP, P. 4, 5 l 5.18 (d) REF: BFN Phases of Matter LP, P. 7 5.19(d) REF: BFN Subcritical Multiplication LP, P. 5-7 5.20(c) REF: BFN Application of the G.E. Eqn. LP, P. 10 BFN MCD LP, P. 5 5.21 (a) REF: BFN MCD Supplement LP, P. 4-6 BFN MCD LP, P. 32-33 1 4 4 i l

A4 6. Answers 6.1 (a) REF: BFN Liquid Radwaste LP, p. 5-10 6.2 (b) REF: BFN LP #35, p. 17-20, 30 6.3 (d) REF: BFN LP #29, P. 9, 23 6.4 (d) REF: BFN LP #29, p. 7 BFN LP #5, p. 7, 12 BFN LP #6, p. 27 6.5 (a) 6.6 (a) 6.7 (d) REF: BFN LP #3 t 6.8 (c) REF: BFN LP #38, p. 19-20 6.9 (a) REF: BFN LP #42, p. 25-26 6.10 (c) REF: BFN LP #14, p. 11, 19 6.11 (a) w (b) REF: BFN LP #25, p. 25-26 6.12 (d) REF: BFN LP #19, p. 8, 17, 19, 23 6.13 (d) REF: BFN LP #13, p. 8, 12 i

A5 6.14(b) REF: BFN LP #38, p. 11 6.15(c) REF: BFN LP #5, p. 35 6.16 (a) REF: BFN LP #16, p. 14-15 c m ?.17 (c) REF: BFN LP #16, p. 26, 32 6.18 (b) REF: BFN LP #7, p. 8-9 6.19 (c) REF: BFN LP #39, p. 7, 13, 15 6.20 (d) REF: BFN LP #50, p. 13-14 6.21(c) REF: BFN LP #47, p. 5 6.22(b) REF: BFN LP #40, p. 10, 20, 22 BFN 0I-71, p. 9 6.23 Normal Shutdown Bus 1 Alternate 1 Shutdown Bus 2 Alternate 2 Diesel Generator B Alternate 3 Shutdown Bd. 3EB (0.5 ea/ 2.0) REF: BFN 0I-57, p. 105 6.24 A3, B3, C3, D3 (0.5) C1, D1 (0,5) Control Room (0.25) A1, B1 (0.5) Locally (0.25) ) REF: BFN LP #51, p. 11

A6 6.25 APRM "A" fails upscale + relay 5A-K12A deenergizes [0.5] + NMS contacts 5A-K12A in RPS Trip Logic Al open [0.5] + relays SA-K14 A & E deenergize [0.5] + contacts 5A-K14 A & E open [0.5] + scram pilot solenoid valves for RPS A deenergize and vent [0.5] (2.5) REF: BFN LP #28 6.26 - Refueling Platform near or over core Service Platform Hoist Loaded (0.5 ea./1.0) REF: BFN LP #53, p. 21 i 6.27 False (0.5) REF: BFN MCD, BWR LP, p. 7-8 s L. A i) ebor.s.g e%M bt La\\,1 t) Er.. % di o..a k\\ bsk L\\ed

5) La C%peu LxwA d) L\\ Gerqv 4W Oh _ u r u

O

A7 7. Answers 7.1 (c) REF: BF-OSIL #19 7.2 (d) REF: BF-OSIL #19 BF-0I-23, p. 6 BF-TS-3.8 BF-IP-1, p. 6 7.3 (d) REF: 10 CFR 20 DMc 7.4 (e)- REF: BF-01-3, p. 4 7.5 (d) REF: BF-0I-47, pp. 21, 38, 41, 44 7.6 (d) REF: BF-01-1, p. 9 7.7 (a) REF: BF-01-66, p. 5-7 7.8 (b) REF: BF-01-3, p. 2-2A 7.9 (c) REF: BF-GOI-100-1, p. 9-11 7.10 (c) REF: BF-G01-100-1, p. 19-21 BF-TS, p. 4 7.11(b) REF: BF-OSIL #47

A8 7.12(a) REF: BF-DI-68, p. 28-29 7.13 (b) REF: BF-GOI-100-12, p. 5-6 7.14 (d) 'REF: BF-G01-100-1, p. 22 7.15 (c) REF: BF-GOI-100-3, p. 84 7.16 - Torus temperature >105'F Torus water level <-6.25" or >-1" Drywell temperature >160*F Drywell pressure >2.45 psig (0.5 ea./2.0) REF: BF-E01-2, p. 3 7.17 a. 50* F/hr. b. 215*F c. 90*F/hr. d. reduce the 0 and hydrogen peroxide content of the coolant. 2 REF: BF-GOI-100-1, p. 15 (0.5 ea/2.0) 7.18 a. 1. Check all rods fully in. 2. Close MSIVs. 3. Trip feedwater pumps. 4. -Trip main turbine 5. Trip condensate pumps. 6. Trip condensate booster pumps. l. 7. Start diesel generators. (2.0) b. Close the MSIVs from the BCC. (0.5) REF. 6F " Ce4ro\\ Ro.m Abn.A enmed ", P. 2.

A9 7.19 - CST charging water to RHR & CS SLC System - Suction from test tank RHR crossties to other units Standby coolant RHR drain pumps CS drain pumps (4 of 6 @ 0.5/2.0) REF: BF E0I-1, p. 24 7.20 - Five or more adjacent rods not inserted below 06 position [0.5] or 4 Thirty or more rods not inserted below 06 position [0.5] and either Rx water level cannot be maintained [0.5] or suppression pool water temperature limit of 110*F is reached [0.5] (2.0) REF: BF-E0I-47, p. 3 7.21 - If a Rx scram has not been initiated, initiate a Rx Scram. Restore and maintain Rx water level between +11 and +54" if possible. If Rx water level cannot be restored and maintained above +11" Then maintain Rx water level above -114.5" (on LI -3 -46A & B) If the ADS timer has initiated, then prevent depressurization by resetting the ADS timer. To prevent MSRVs from cycling when RPV pressure approaches 1105 psig,' manually open a sufficient number of MSRVs to reduce Rx pressure to ~950 psig. (0.5 ea/2.5) REF: BF-E0I-1, p. 6 4 )

A10 7.22 Immediately dispatch an operator to the BCC and establish communications... If reactor level decreases to -114.5" below inst. zero perform a rapid depressurization of the reactor using any six (6) available MSRVs... Restora and maintain reactor water level with available makeup systems... (0.5ea./1.5) REF: BF-GOI-100-11, p. 10 7.23 a. Sufficient heat removal to restore and maintain PCT at or below 2200'F. (0.5) Core Submergence b. Spray Cooling Steam Cooling (1.0) REF: BFNP E0I Cross-Reference, p. 2 l

All 8.0 ANSWERS 8.1 d. (1.0) Ref: TS 3.7.C BF OSIL #35 (1.0) 8.2 c. Ref: BF OSIL #16 (1.0) 8.3 a. Ref: BF OSIL #12 BFN, IP-10 8.4 d. (1.0) Ref: BFN, IP-15 8.5 b. (1.0) I Ref: BF OSIL #9 8.6 c. (1.0) Ref: BF OSIL #43 8.7 b. (1.0)- Ref: TS 3.4.0, BF OSIL #20 BF SP 12.20 8.8 d. (1.0) Ref: BF RCI-1 8.9 a. (1.0) Ref: BF IP-20 8.10 b. (1.0) Ref: BF SP 14.25 8.11 a. (1.0) Ref: BF OSIL #20 ~

A12 8.12 b. (1.0) Ref: BF SP 12.17 BF SP 12.7 8.13 a) Ensure inadvertent transfer of significant amounts of (1.0) contaminated fluids will not occur. k) 1. Continued operation of the engineered safety (1.0) features will result in an unsafe plant condition ~ with regard to either personnel or operability of safety features. QR 2. The plant is in a stable condition in which technical (1.0) specifications clearly indicate that operability of the engineered safety features is no longer required. cM. b) Fm., ue - 8. D d Ref: BF SP 12.17 7c,. 8.14 1. When another licensed operator is present [.25] with (1.0) an unobstructed view of panel 9-5.i ic.,[d.25] ,c.xa e. L (1 an.. w w Prem <a.wa % <,%we 2. In an emergency [.25] to attend or verify alarms and initiate various corrective actions. [.25] Ref: BF SP 12.5 8.15 d. (1.0) Ref: BF RCI-9 w a

j A13 8.16 c. (1.0) Ref: TS 3.7.8 i TS Def. LC0 8.17 c. (1.0) Ref: TS Def. 8.18 a. (1.0)

Ref

TS 3.1 and 3.2 8.19 a.

(1.0) Ref:-TS Def. 8.20 b. (1.0) Ref: BF SP 14.25 8.21 c. (1.0) i Ref: BF SP 15.2 8.22 b. (1.0) Ref: BF SP 14.15 8.23 c. (1.0) Ref: BF OSIL #21 8.24 a. -2 (1.0) b. -3 (Note: Answer will be graded in accordance with (1.0) answer to a.) i ^ Ref: BF TI 13 8.25 1. Any time irradiated fuel is in the reactor vessel (0.5) and reactor coolant is above 212F, reactor coolant i leakage into the primary containment from unidentified sources shall not exceed 5 gpm. 2. In addition, total reactor coolant leakage into the (0.5) primary containment shall not exceed 25 gpm. 3. In RUN mode, reactor coolant leakage into the primary (0.5) containment from unidentified sources'shall not increase by more than 2 gpm average over any 24 hour period. Limit I above exceeded day 2, 2400 - 2496 gal = 5.2 gpm (0.5) L

~ A14 Limit 2 above exceeded day 1, 2400 - 10,440 gal + 1,824 gal (0.5) = 12,264 gal = 25.55 gpa Limit 3 above exceeded day 2, 0800 and 1600 with respect (0.5) to day 1, 1600 - 2198 gal = 4.58 gpm is greater than 2 gpm increase compared to 1008 gal = 2.1 gpm (CAF) Ref: BF TS 3.6.C I I en e e 9 9 1 k

I ENC 1.05URE 4 1 INTRODUCTION STATERENT The TTA exas evaluators considered the scope of the exam was within

the knowledge level expected of the candidates. However, the method used to ensure that the correct answer wasn't obvious created the following problem areas:

1., Required indepth reading of answer which proved to be too time consuming for the SRO candidates. One word or letter designation made the answer right or wrong. 2. In more than one incident the question required detailed memorization of :satorial that would normally be readily available to the operator, and was not stated in the procedure as an immediate operator action. 3. In many cases, questions and/or answers did not contain sufficient information and required the candidates to make undocumented assumptions that affected his determination of a valid answer.

NRC Instructions Written on Markerboard to Class l 1 1. 1.23 Change OR to 0F 2. T/F will say so - if does not, pick one answer 4 3. 2.14 Pick one of A, B, C, D. Only one column is fully correct for each speed 4. 3.26/6.25 If in doubt about relay contact number, then assign a number to it and refer to that number 5. 4.3/7.4 Assume RFPT has been reset 6. 8.25 Add: All integrators (FQ 77-6 & 16) are inop and coolant leakage determinations are being done manuall 7. 8.24a Change power reduction to 11.8 Kw/FT from 12.0 Kw/FT r e 4

Qu:stian 1.19 f l This question's answers were all long and the time requirement to determine the most correct answer was too much for the assigned point value. Also the length of the answers make it too difficult to analyze i each part. l 1 Question 1.24/5.4 t Concerning the fuel center line temperature with new fuel, your answer key looks for a decrease in fuel center line temperature. This answer is ~ i correct if you consider some exposure time has been received to allow for compaction of the uranium oxide fuel. Our MAPLHGR limits are adjusted to depress the new fuel Kw/ft limits to allow this compaction. During this compaction the Helium gap increases, which decreases the heat transfer and 1 will increase fuel center line. temperature. After the compaction and the fuel cracks propogate and ratchet open, the helica gap decreases and increases heat transfer rate which will decrease the fuel center line temperature. Reference page 9 Linear Heat Gen. Rate & Bases Lesson Plan thermo book. D.S.a. new fuel exposed to operating temperature, densification occurs. Reference transparency #2 Thermo-APLHGR Lesson plan tables of fuel types indicates depressee Kw/ft fuel limits from 0-200 mwd /t on exposures. j Reference page 171-172a-U-1 Tech Spec with fuel types from 0-200 mwd / ton l exposure the heat transfer in Kw/ft is depressed, due to poor heat transfer rates due to densification. The question just addressed new l fuel, it did not specify an exposure so based upon these references initially the fuel center line temperature of new fuel will increase due to poor heat transfer. After the cracks propogate and ratchet open fuel center line temperature will decrease..According to Tech Spec table this should have occurred by 200 NWd/ ton exposure. If a candidate responds with increase or decrease, it should be accepted'since no fuel exposure was specified. i 4 Concerning question 1.24.d "The effect on fuel center line temperature of operating with one'NSIV closed." The question was stated " operating with" this would indicate that the reactor had stabilized and survived the closure. Initially the reactor pressure would increase and fuel center line temperature would~ increase during this' transient. But as the EHC stabilizes the RPV pressure the center line temperature would effectively be the same while " operating" with the one NSIV closed. a yv tw y 5' erp*i e m e-T+- 3-m? + - * * - - f W '5%'=" M

Qu :tica 1.24/5.4 (ccatinued) The candidates may respond with an increase, same as your answer key, if considering the fuel center line during the transient. Others will respond with " remain the same" after the transient has stabilized. Question 1.25 (a) and (e) The answer key response required two parts for full credit, but this was not stated in question (i.e., why core flow decreases and what caused this action to occur). Question 2.13/6.17 Per OI 84: Do not make-up nitrogen to the primary containment if venting is in progress. This would make answer 'c' incorrect ~

Qu:stien 3.13/6.11 Answer 'B' also could be correct if you consider that you are moving the last' rod in a group. This rod could then be moved two notches without violating any restriction. Question 3.14 When 70% undervoltage is achieved on U-1 panel 9-9 the panel will automatically transfer to the U-2 unit preferred panel 11 breaker 1172. This transfer is for panel 9-9 only and is accomplished by an ASCO throw over switch if the alternate supply has voltage. Your answer was misleading in the fact that it addressed the U-1 unit preferred bus, which is in effect panel 11 in the battery board room, fed directly from the 1001, 1002 and 1003 breakers. Upon the loss of an MG set this panel will de-energize, but panel 9-9 in the control room will transfer to its alternate supply, U-2 breaker.1172, if potential is available. The result will be the U-1 unit preferred bus de-energized, which is panel 11 in battery board room, and panel 9-9 energized from an alternate source. ASCO - throws to alternate at 70% voltage with return to normal at 90% when normal potential is regained page 98 OI 57 U.R.2.b. addresses panel 9-9 l 5 i t -.. ~.,, ,y,. v. ,,-.7 -y,

r Qu:stica 4.3/7.4 There doesn't seem to be any completely correct answer for this question if you considered that the RFPT was reset as given in direction to the class during the exam. With the RFPT reset: the AOP would be running, the suction valve open, the discharge valve closed, the minimum flow valve open, the blue light out, and in single element control Section 4 & 7 Question concerning rubbing noises on Main Turbine at 100 rpm The correct answee (as shown on NRC answer key) contains. statements from various' sections of OI 47. The correct answer contained an action statement requiring breaking vacuum. This action statement was taken from Section III G.3.b which deals with actions to be taken during emergency shutdown. The note is contained in the section which gives action to be taken at 1200-900 rpm range. Breaking vacuum at 100 rpes would have 4 j negligible effect on reducing the speed and could cause adverse effects on the unit due to pressure' transient'when the bypass valves go closed at 7" vacuum. 1 1 l 1 ,.r, ,,.-w.-~-< v y p.,-, -,.y,-<-

S:ctica 4 Question 4.17 1 Question does not call for response to first part of answer "To produce j sufficient pump head to create down thrust on the pump" 1 l l 5 a Question 4.23 Question is requiring memorization of information within an appendix to a procedure. The only immediate action per the procedure is to dispatch an operator to the backup control center. e i n en.. n. y--. y

Qu stien-6.26 Concerning question 6.26 on the refueling interlocks with the mode switch in startup. The copy of lesson plan 53 Figure 4 had a line inadvertently left off. Therefore the candidates response may be in more detail than the NRC answer key. 1. Your answer key a. refuelins platform near or over core b. service platform hoist loaded 2. Response by candidates: a. refueling platform near or over core ~ with: and

1) monorail mounted' hoist loaded or

2) frame-mounted hoist loaded or

3) fuel grapple loaded or

4) fuel grapple not full up b.

service platform hoist loaded Reference attached drawing from BPNP FSAR l 4 i

r- _ ~ _.. WONOR AIL MOUNTED H AWE 430%TED ' FUEL GRAPPLE FUEL GRAPPLE MONORAll M0bNTED FRAIE MOUNTED'. FUEL GRAPPLE FUEL GRAPPLE ~ ' HOIST LOADEO H0isT LOACED: ~ LOADED h0T FULLY UP HOIST LOADED HOIST LOADED - LOADED NOT FULLY U? I-1 1 I .I ' ONE FC0 ' WiTrtCFAs% - + ig REFUELING - REFUELING REFUELING ORE DAN ONE 5ERv1CE PLATFORM . SELECTION OF. PLATFORM NEAR PLATFORM NEAR - SERVICE PLATFORM PLATFORM NEAR NOT ALL RODSIN RCD WITHDRAWN H0 TIT LOACED SECOND ROD OR OVER CORE ' OR OVER COPE HOIST LOADED OR OVER CORE I I I I I

MODE SwtTCH MODE SWITCH PL AR IN " REFUEL" IN **STARTUP" OR OVER CORE-1 I

REFUELING MODESWITCH MODE SulTCH PLATFORM NEAR ROD BLOCK IN "STARTUP" OR OVER CORE "IN REFUEL" 1 I STOP REFUELING FRAME MOUNTED ' MONORAll MOUNTED PLAT R EL HOIST LOADED H0t3T LOADED I-I , NOT ALL RODS IN NOT ALL RODS IN - NOT ALL R005 IN NOT ALL RODSIN NOTE: ~ l l 'l l

• ' LOADED" 15 DEFINED AS SLIGHTLY LESS TH AN THE PRE-MEASURED AVERAGE 7ElGHT OF ONE FUEL ASSEMBLY REFUELING REFUELING REFUELING SERytCE PLATFORM PLATFORM NEAR PLATFORM NEAR '

PLATFORM NEAR. HOIST LOADED OR OVER CORE. OR OVER CORE OR OVER CORE BROWh5 FERRY MUCLEAR PLANT . PIMAL SAFETY ANALYSIS REPORT STOP FUEL ' STOP SERVICE- . STOP FRAME ' STOP MON 0RAll Refueling Interlocks, GEAPPLE-PL ATFORM HOIST ' MOUNTEC HOIST MOUNTED H0tST p,,,,;,n., gioeg p;,,,,, OPERATION OPERATION RA!!E OR LIFT rat 3E OR LIFT FIGURE 7.6-1 e 4 0

-Section 7 Question 7.2 A possible correct assuer is *a* per OI 23 pg. 6 IT a.3.. Terminology of securing S.D. cooling any be laterpreted *as removing that loop from service." Question 7.12 See comment os 80 sectica 4.11 Section 8 Question 8.7 BF 12.20 requires utilizative of form BF 126 for determinias operability-of equipment with am laop D/G. RF 126 taas act made available to examisees.

Qu:stic2 8.13- "b" part - another possible answer which would be more familiar to the operators would come from E0Is or EPG-X pg 20 "Do no secure or place an ECCS in manual mode unless, by at least two independent indications (1) misoperation in automatic mode is confirmed, or (2) adequate core cooling is assured. Question 8.14 Proper relief means individual does have unobstructed view of 9-5 and is in Zone 1. i Question 8.25 This question requires memorization of an LCO. The Tech Specs are readily available in the control room for the SRO to'ut111:e in determining the LCO. i l i 4 -s-m- .~u -m ,y, ,~- ..w q -- -,y.. y 3,,-.-w, +, - ,m, -,-- y-~- e -

r a C::ccenirs tha quastice en th) RWR Ictching into a secup resulting in insert i and withdraw errors. =The answer to part 'B' (after the rod in Group 1 has been withdrawn) would be depended upon whether the RWR was re-initialized or a rod in Group 4 were selected or if all that was done was select a rod in Group 1 and withdraw it to 48. Condition one: Select rod in Group 1 and withdraw Results: Group 3 remains latched group Only insert error is rod in group 2 Group 4 rod remains withdraw Condition two: Select rod in group 1 and withdraw Select. rod in group 4 Results: Group 4 becomes latched group have two insert error (rod in Group 2 and 3) No withdraw errors Condition three: Select rod in group 1 and withdraw re-initialize the RWR Results: Same as condition two -Question # Concerning question about what will trip a Diesel Generator during accident condition. Per OI 82 (Jan. 25, 1983) for U1 & U2, differential and field failure both will trip D/G. The D/G 1esson plan only list differential which is the correct answer. The OI 82 for U1 & 2 has been changed, but not in the material sent to NRC. Recommend that accept either answer as being correct. l

c-s a SlHDIARY STATEMENT Some'of the above mentioned problem areas would have been avoided had the exam review process proceeded the exam. .}}