ML20205N939

Exam Rept 50-259/OL-86-03 on 861117-21 & 861201-12.Exam Results:Seven of Eight Reactor Operators (RO) & Four of Five Senior Reactor Operators (SRO) Passed Replacement Exam & Four RO & Five of Six SRO Passed Requalification Exam

ML20205N939
Person / Time
Site:	Browns Ferry
Issue date:	03/09/1987
From:	Brockman K, Munro J NRC
To:
Shared Package
ML20205N848	List: ML20205N843 ML20205N939
References
50-259-OL-86-03, 50-259-OL-86-3, NUDOCS 8704030139
Download: ML20205N939 (492)

Text

r ENCLOSURE 1 EXAMINATION REPORT 259/0L-86-03 Facility Licensee: Tennessee Valley Authority 6N 38A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Facility Name: Browns Ferry Nuclear Plant Facility Docket No.: 50-259, 50-260 and 50-296 Written and oral examinations were administered at the Browns Ferry nuclear plant near Decatur,' Alabama. Simulator examinations were administered at the Power Training Operations Center near Soddy-Daisy, Tennessee.

Chief Examiner: M <sc[*w TWA U Date Signed Kef E. Broctmtfi /

Approved by: s John F. Munro, Settion Chief 3/Th7 Date Signed Summary:

Examinations on November 17-21, 1986 and December 1-12, 1986.

Oral replacement examinations were administered to 13 candidates,12 of 'whom passed; 13 candidates were administered simulator examinations, all of whom passed; 13 candidates were administered written examinations, 12 of whom passed.

Based upon these results 7 of 8 R0's and _4 ~ of 5 SR0's passed the replacement examinations.

Simulator requalification examinations were administered to 9 operators, all of whom passed; written examinations were administered to 10 operators, 9 of whom passed. Based upon these results, 4 of 4 R0's and -5 of 6 SR0's passed the requalification examinations.

Browns Ferry's progress in the accelerated requalification training program  !

continues to be satisfactory; this is an interim status report.

I 8704030139 PDR 870313 V

ADOCK 05000259 PDR l

i J

' im ..

/ ;,,

i.

,/'

, y REPORT DETAILS

1. Facility Employees Contacted: 1

• R. McKeon, Operations Superintendent

• R. J. Johnson, Director Nuclear Training '

• C. H.- Nee, Training Manager, BFNP

• R. G. Jones, Operations Training Supervisor, BFNP T. Mayfield, Simulator Instructor, BFNP

• Attended Exit Meeting 'i < ,

2. Examiners:

• K. E. Brockman, RII S. D. Stadler, RII D. J. Lange, RI L. Wiens, NRR M. Daniels, Sonalyst R. Miller, Sonalyst ,

• Chief Examiner

3. Examination Review At the conclusion of the written examinations, the examiners provided R. G. Jones with a copy of the written examination and answer key for review. The following resolutions -are provided to the facility comments:

(NOTE: Questions are referred to by Computer Cross Reference, due to multiplicity uong the four examinations administered.)-

(1) Question KEB 097 NRC Resolution: Comment Accepted. Part B deleted due to no correct response (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />) available. Section and tdtal points adjusted accordingly.

(2) Question KEB 869 l NRC Resolution: Comment Acknowledged. Answer choices would be 1 improved by setting time, t=0, on the vertical axis. However, )

,from t=0 to t=4, only response "d" starts at the appropriate value '

' and traces accurately. Answer "c" will not be credited. No change to answer key.

i i

S l

4 l

t . 61

Enclosure 1 2 (3) Question KEB 870 NRC Resolution: Comment Acknowledged. LIT 3-53 is equivalent to Level Indicator "A" and will be fully credited. Answer key annotated.

(4) Question KEB 874 NRC Resolution: ,

(a) Comment Acknowledged. 6-30 seconds (01-92) will receive half credit (The time delay can be less than six seconds.) 0 to 30-60 seconds will receive three quarters credit (The time delay could be greater than 5 time constants). A discussion of the circuitry operation (63.2% W power / time constant) .

would be a full credit response. No change to answer key required.

(b) Comment Acknowledged. Delta Power (P(f) - P(i)) is also an input to the time delay calculation and will be fully credited. Answer key modified.

(5) Question KEB 875 NRC Resolution: Comment Acknowledged. The utility resolution is accepted with the following exception " Required Vacuum not established" is not supported in the reference (0P171.030).

Answer key modified appropriately.

(6) Question KEB 876 NRC Resolution: Comment Acknowledged. Specific discussion of the component " Averaging Circuit" is not required; however, the justification must show why the power increase does not reach the Rod Block setpoint (46%). No change to answer key required.

(7) Question KEB 879 NRC Resolution: Comment Acknowledged. The platform being near or over the core is inclusive to the question stem and is not required to be specifically stated (Answer key annotated).

Stating the " Mode Switch in Startup" is not the same as " Mode Switch not in Refuel" and would receive half credit.

(8) Question KEB 880 NRC Resolution: Comment Acknowledged. The utility comment is )

equivalent to the answer key. No changes required.

l l

,. . . =

v ~i Enclosure 1 'x ,

3 (9) Question.KEB 881 l NRC Resolution: Coment' Acknowledged. The R0 'and SRO are

' respontible for-maintaining the reactor core operating within its licensed limits. The P-1 edit is the means~ by which this is verified. .They must be able to recognize whether their- thermal i limit data is' valid, or not. .This question only addresses a .

recognition of. validity - not an_ analysis of the process computer

problem. It is', therefore, appropriate R0 and SR0 knowledge. No changes required.

(10) Question KEB 885 NRC Resolution: Coment Accepted.. Partial credit weighting t

revised to show " equipment failure (0.3)" and " ADS blowdown....

(0.1)". Answer key modified.

(11) Question KEB 887 NRC Resolution:

(a) C'oment Acknowledged.- Answer. key expanded to address both i regions of the HCLL (14.6 ft. to 11.5 ft and less-than 11.5 ft). Each region will be weighted'at-0.5.

(b) Coment Acknowledged. No action required.

l \

(12) Question KEB 888 I

NRC Resolution: Coment Accepted. The time specificity- with respect to verifying operability is not' considered to be appro-priately elicited. Answer key modified.

(13) Question KEB 895 NRC Resolution: Coment Accepted. More detailed reference supplied by utility supports comment.- Answer key modified to

change part (b) response to "2".

(14) Question KEB 902 NRC Resolution: Comment Acknowledged. The utility recommenda-tions are alternate descriptions of the same location as in the.

answer key and would receive ful1~ credit. Answer key annotated with alternate responses.

(15) Question KEB 908

1 NRC Resolution:

j (a) Coment Acknowledged. Any accurate definition is acceptable..

!. No change to answer key required.

4 L._

_ a. . .

Enclosure 1 4 (b) Coment Not Accepted. The question asks for Available NPSH; considering ONLY recirculation flow, available NPSH would decrease.

(c) Coment Acknowledged. Available NPSH increases with increasing pressure; answer key changed to reflect correct response.

(16) Question KEB 911 NRC Resolution:

(a) Comment Accepted. The word " rapid" could be concentrated on as the specific point to be answered. Discussion of why only burnout is the key inventory depletor will be accepted for full credit. Answer key modified.

(b) Coment Accepted. Answer key only addresses the conditions at Point B. It will be expanded to include the changes from Point B to Point D.

(17) Question KEB 912 NRC Resolution:

(a) Comment Not Accepted. Inadequate justification.

(b) Comment Acknowledged. Since thermal stratification occurred with shutdown cooling in service, establishing natural circulation is the most correct response. G01 100-12 is not specific in directing an increase in SDC; however, if the candidate states to " increase SDC flow to reverse the stratification", partial credit (0.35) will be awarded.

Answer key adjusted.

(c) Coment Acknowledged. No change to answer key required.

(18) Question KEB 914 NRC Resolution: Coment Acknowledged. Typographical error corrected. Question weight changed to 2.0; category and exam totals modified.

(19) Question KEB 915 NRC Resolution:

(a) Coment Accepted. Instantaneous power is limited to 3358 MWt (G01 100-1), while average power is limited to 3293 MWt (Tech Specs). These two values will be accepted. Answer key modified.

Enclosure 1 5 (b) Connent Acknowledged. The question specifically asks for MWt, not % RTP; however, partial credit (0.25) will be allowed for "25% RTP". ' Answer key adjusted.

(20) Question KEB 919 NRC Resolution: Consnent Acknowledged. "Kf" is already shown as a full credit response. No change required.

(21) Question KEB 922 -

NRC Resolution: Comnent not Accepted. - While ~ " initial" wo'uld improve the question, only the 65 psig pressure limit is designed' to ensure complete / timely rod insertion. Question is adequate -

no changes made.

(22) Question KEB 923 NRC Resolution: Connent Acknowledged.- " Loss of water from the CST or the Torus" is an acceptable adverse consequence; Sump flooding is not. Answer key modified to reflect alternate answer.

(23) Question KEB 924 (a) Consnent Accepted. Either response is acceptable as one additional alternative. Answer key modified.

(b) Connent Accepted. Either consequence is acceptable for full credit. Answer key modified.

(c) Comment Acknowledged. The loss of vacuum in the barometric condenser is the precursor to the radioactivity release and

~

is not required. No change to answer key.

(24) Question KEB 925 NRC Resolution: Connent Not Accepted. The approximate equiva-1ency of the setpoints was not requested. The enclosed circuit diagram definitively supports the Bailey Drive as limiting;

~

OPL 171.008 specifies the Mechanical stop as- controlling. No change to answer key.

(25) Question KEB 926 NRC Resolution: Comment Not Accepted. The concern with, and corrective action for, the Minimum Flow Valve operation is specifically addressed in ' 01-74. Additionally, this was a-corrective action to an identified deficiency in BFNP operations.

Enclosure 1 6 There are no directed cautions concerning the Torus Isolation Valves (74-57 & 71) and addressing their closure is not appro-priate for this question. The inclusion of "and flushing" should not confuse a knowledgeable operator. No change to question or answer key.

(26) Question KEB 929 NRC Resolution: Comment Acknowledged. " Reset (HEA) Lockout Relay (86 Relay)" added to answer key since it must be reset to call EDG in standby readiness. Answer key modified to reflect new correct answer.

(27) Question KEB 930 NRC Resolution: Comment acknowledged. All alternatives describe the same parameter and will be fully credited. No change to answer key required.

(28) Question KEB 933 NRC Resolution: Comment not accepted. LPCI signal is a given and therefore, not an answer. The other three criteria are the only correct answers (2/3 core height is equivalent to -39"). No-change to answer key.

(29) Question KEB 935 NRC Resolution: Comment Acknowledged. Since there are four lights (Red / Yellow / Green / White) above the CS ' Pump Switch, the candidate must identify the one which will illuminate. Position description is an acceptable alternative to color. No change to answer key required.

(30) Question KEB 936 NRC Resolution: Comment Acknowledged. Since the potential for confusion does exist (single or multiple timer resets) either yes or no could be acceptable; therefore, question DELETED. Question, Section and Total points reduced accordingly.

(31) Question KEB 938 NRC Resolution: Comment Acknowledged.. " Local" added to answer key to increase clarity.

e .- - - - . - - - - . - - -

- Enclosure 1 7 (32) Question KEB 939 NRC Resolution: Coment Acknowledged. " Shutdown bus" referred

, to the EDG's appropriate 4160 v output, not the-BFNP 4160 v Bus.

, Answer key modified to increase clarity with "4160 volt Shutdown -

i Boa rd" .- .

(33) Question'KEB940-NRC Resolution: Comment Acknowledged.' Answer key allows - for additional appropriate responses. Alternatives;l & 2 would be fully credited; alternatives 3, 4, 'and 5 are-_not specifically-related to the relief valve failure and would not be credited.

Answer key modified.

' ~

(34) Question KEB 941

, NRC Resolution: Comment not Accepted. Assumptions made beyond j~ the given of the question cannot be verified. If the' candidate states his assumption concerning only partial LPCI response, he-will be graded accordingly; otherwise, TRUE is the only acceptable response. No change to answer key required.

(35) Question KEB 944 NRC Resolution: Coment Acknowledged. While the wording is explicit and consistent with that -in G01-100-1, it is possible for a candidate to erroneously assume a total loss of shutdown cooling.. If the assumption is definitively stated, full credit will be given for the utility recommended alternate answer. No -

change to answer key required.

(36) Question KEB 953 i

! NRC Resolution:

4

! (a) Comment Accepted. The DUAL limiter is the controlling component for "how high" the flow: increase will.go. Answer key modified to accept either response as a'. limiting component.

4 (b) Comment not Accepted. See resolution to KEB 925.

(37) Question KEB 955 I

NRC. Resolution: Comment Acknowledged. While the graph's clarity l is less than perfect, in the areas.of concern, it is adequate to answer part (a) within the allowed tolerance. Part (b) tolerance 1 is inc.reased to i40 MWe, per the uti_lity request.' This estab- '

! lishes two units' of measurement as the consistent level of_.

tolerance. Answer key modified.

I i

- ,,w - - . . - , . . , m,, , . ,., , ,,..w,m,... ..,..,,y. .m_.- , . , , , , , , - , , - _ . . - , - - , - v-,. ,-my,-,,-

Enclosure 1 8 (38) Question KEB 956 NRC Resolution: Coment Accepted. "480 volt Load Shed Logic initiation" is equivalent to a voltage loss with an accident signal present. Answer key modified.

(39) Question KEB 965 NRC Resolution: Coment Accepted. Typographical error corrected on answer key.

(40) Question KEB 966 NRC Resolution: Comment not Accepted. The procedure is explicit for two distinct activities - opening and tagging the switches.

Not defining the position is an inadequate response. No change to answer key.

(41) Question KEB 971 NRC Resolution: Coment Accepted. Answer key modified to reflect accurate response as supported by facility Technical Specifica-tions.

(42) Question KEB 972 NRC Resolution: Coment Accepted. Three criteria are, in fact, presented. Answer key modified to accept 2 of 3 at 0.5 each.

(43) Question KEB 974 NRC Resolution: Coment Acknowledged. There are no requirements for " buzz words". The concepts elicited will be credited, however they are described. No change to answer key required.

(44) Question KEB 976 NRC Resolution:

a. Coment Accepted. 5(N-18) added as a selection criteria.

Volunteers can, of course, have pre-consented and all rescue team members are, by definition, volunteers. " Older" is acceptable as an age reference answer key modified,

b. Coment Accepted. Answer key modified to reflect 200 Rem additional exposure, for 275 Rem total.

c. All recomendations are equivalents of the answer key. No changes required.

-~ -- --

+ . - , sm. e

t i '

1 l Enclosure 1

9 T

A j (45) Question KEB 975 NRC Resolution: : Coment Acknowledged. Question could have been worded more explicitly. ,

i The tested knowledge of a safety limit and the required actions

! if a safety limit is violated, are SR0 required knowledge. The

knowledge of the actions to be taken if a safety limit is violated i is a Learning Objective' '(TS #74). Standard Practice 12.8, 1 Section B. 7 specifically requires that if 1250 psig is exceeded, an evaluation.should be performed to determine if the 1375 psig safety limit was violated. The statement.in the comment that,

with a reactor pressure spike (indicated) of 1287 psig, "one

! could conclude that the pressure limit of 1375 psig had not been

. exceeded "is incorrect". . An SR0 should be aware that the pressure l at the bottom of the vessel would be greater than the measured

pressure in the steam dome d.ue to the varying head of water. .This-fact is addressed in the Technical Specification bases for the j Reactor Coolant System Integrity (1.2).

The safety limit of 1375 psig actually applies to any point in j the reactor vessel; however, because of the static water head, the

, highest pressure point will occur at the bottom of the vessel.'

i Because the pressure is not monitored at'this point, it cannot be j directly determined if this safety limit has been violated. .Also, j because of the potentially varying head level and flow pressure

,' drops, an equivalent pressure cannot be determined "a priori" for -

a pressure monitor higher in the vessel. . Therefore, following any j transient that is severe enough to cause concern that this safety i limit was violated, a - calculation will be performed using all

available information to determine if the safety limit was >

violated.

! The alternate responses proposed from SP 12.8 are administrative

! requirements for items which should be checked and documented

on the post trip review and not' really " unique". . While .these j responses are appropriate contributors for full credit, full

! credit could not be achieved without addressing evaluation of the

! safety limit. Answer key modified to reflect additional actions i (beyond NRC notification) of investigating potential safety limit

violation.

i (46)QuestionKEB977 i

NRC Resolution: Comment Acknowledged. This question is based upon

. a recent event at Browns Ferry - in fact, it is a direct quote of j the occurrence. The- operator should be knowledgeable of recent plant events and, therefore, the full credit answer remains

, unchanged. Discussion of only the electrical ramifications (TS -

3.9.c) will receive partial credit of 70% of the questions' value 1 (1.05). Answer key annotated.

f

- .m. . - , . - , . - , . , . . - , - -~ . , __.._-_,..,-.,y.,.,.-._,,,..._m_.-.._. ~ . , , , _ - . . , ....,,-,-.~~w-, .,---,w,

l Enclosure 1 10 1 (47) Question KEB 978 NRC Resolution:

l (a) Coment Acknowledged. The assumptions described by the utility are plausible. If these assumptions are stated the utility recomended answer would receive full credit. Answer key annotated to show alternative response.

(b) Comment Acknowledged. This is the exact event which recently occurred at Browns Ferry and, therefore, should be easily responded to. However, it is possible that "yes" or "no" would be answered, depending upon the actions of the review comittee. Since an explanation was not asked for, the question will be deleted. Section and total points adjusted appropriately.

(48) Question KEB 979 NRC Resolution: Comment Accepted. While not tied to grid, speed and frequency are equivalent. Answer key modified to annotate additional response.

(49) Question KEB 894 NRC Resolution: Coment Accepted. G0I 100-1 supports the requested revision. Answer key changed to TRUE.

4. Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination.

There were no generic performance weaknesses noted during the oral examina-tion.

There were two generic performance weaknesses noted during the simulator examinations,

a. The replacunent candidates were weak in identifying controller malfunctions (auto mode) and recognizing the need for manual control operations.

b. All candidates / operators were weak in keeping their team members notified when alarms were expected due to in-progress operations. This resulted in numerous instances of starting to respond to seeming malfunctions, when, in fact, the conditions were expected end controlled.

Enclosure 1 11 i

J

c. There were numerous instances of . r.imulator weaknesses - during . the examination. While the examination validity, in this case, was not felt to be challenged, the potential psychological impacts' on the

! candidates is unacceptable. Additionally, the simulation capabilities, j on the whole, become questionable when problems like those experienced

{ are seen. The correction of the software shortcomings should j receive the highest attention by BFNP training management. Examples ~of 4

these shortcomings were:

(1) The RCIC System can not pass its surveillance for Minimum Flow Valve Closure time.

(2) Main Steam Line High Radiation did not produce a half-scram.

(3) No pressure transient could be generated to produce a high neutron i flux condition.

(4) A & D diesel generator loadings were unrealistic.

t I (5) Off Gas model, at best, 1s poor.

(6) SRV's always are modeled to have steam flowing through them, even during core flooding conditions.

l The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted

and appreciated.

! The licensee did not identify as proprietary any of the material provided to

or reviewed by the examiners.

4 l

l i

i l

l l

i I

l I

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

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

U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _BRQWNS_ FERRY _1 1_2h]_____

REACTOR TYPE: _ SWR-GE4_________________

DATE ADMINISTERED: _Q6212291________________

EXAMINER: _BRgCiMAN f 3 _K.

CANDIDATE: _~____U _A__D 1 _E h _~__~____

INSIEUCIlgNS_Ig_CONDIDOIE1 Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination.

Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff. 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 four (4) hours after the examination starts.

% OF CATEGORY  % OF CANDIDATE'S CATEGORY

__Y8LUE_ _IgIG6 ___SCQBE___ _yGLUE__ ______________C@IEGQBy_____________

16.75 25.00 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

_161Z5__ _25199 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 1Z199__ _25:2Z ___________ ________ 3. INSTRUMENTS AND CONTROLS

_lbs59__ _23165 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_eZ199__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

A..

NRC RULES ANDyGUIDELINES_FOR LICENSE EXAMINATIONS During:the' administration of this examinatian the following rules' apply:-

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

'2. Restroom trips are to'be limited and only one candidate at a: time.may

~

leave. You mustLavoid all contacts with anyone outside the examination-room to avoid'even-the appearance or possibility of cheating. .

-3r Use black ink or dark pencil gely to facilitate' legible reproductions.

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

. examination.

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

6. Use only the paper provided for answers.

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

8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each' category on a gew page, write galy 90 gee side of the paper, and write "Last Page" on the last answer sheet.

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

10. Skip at least three lines between each answer.

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

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

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

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

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE DUESTION AND DO NOT LEAVE ANY ANSWER BLANK..

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

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

1

18. When-you. complete,your examination, you shall: 1

a. Assemble your; examination-as follows:

( 1 ) -- Exam questions on top.

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

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

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

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

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

e i

']

ti. PRINCIPLES OF NUCLEAR POWER PLANT' OPERATION 2 PAGE 2 IUEBUgDYN8dlCS2_HEGI_IB6NSEEB_8ND_E691Q_ELgy OUESTION 1.01 (1.50)

Concerning the Bypass Flow in the reactor core:

-a. Which one of the following most accurately indicates core bypass ,

flow at 100%_ thermal power and core flow? (1.0)

(1) 1%

.(2) 5%

(3) 11%

(4) 20%.

b. STATE the most significant consequence that would occur if bypass flow were significantly reduced at full power. (0.5)

DUESTION 1.02 (2.00)

Concerning the Net Positive Suction Head (NPSH):

a. DEFINE NPSH (0.5)

b. For each of the following, indicate whether the.available NPSH at the cuttien of the recirculation pump would INCREASE / DECREASE /

REMAIN THE SAME: (1.5)

(1) The Feedwater Flow is INCREASED (2) The Recirculation Flow is INCREASED (3) The Vessel Pressure is- INCREASED f rom 200 psig to 800 psig

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

c_ _ ___ u

l-

1. PRINCIPLES OF NUCLEAR POWER _PL8NT_QPER6TIQN i PAGE 3 ISEBdQQ1N80igS 1_UE@I_IB6NSEEB_6NQ_E(ylQ_E6QW QUESTION 1.03 (1.00)

Which one of the following most accurately.decribes when Control Rod Worth, during a REACTOR STARTUP, would normally be at its MAXIMUM 7

a. Cold Shutdown

b. Heatup in Progress (* 1% RTP)

c. Heatup Complete (* 1% RTP)

d. 50% RTP

e. 100% RTP QUESTION 1.04 ( .50) l STATE the relationship betweeen between Beta-effective and Reactivity which best describes PROMPT CRITICAL.

QUESTION 1.05 (1.50)

Attached Figure # 911 shows a representative Xenon concentration curve. The reactor is scrammed at Point A, resulting in a peak Xenon concentration approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> later, at Point B.

Line D - C represents the normal depletion of Xenon that would be expected.

If the reactor were RESTARTED at Point B (dotted line),. EXPLAIN:

a. The rapid decrease in Xenon concentration from Points B to D. (0.5)

b. The change in the distribution of the radial neutron. flux during the period from Points D to D (include reasons) . (0.5)

c. Any special precaution (s) that you would need to observe, as a reactor operator, between Points B and D (include reasons) . . (0.5)

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

'ig__EBINGIPLgg_gF NUCLEAR POWER PLANT OPERATION 2 PAGE .4 IUEBMggyUBMIgS2 _UE91_IB8NSEEB_80p_ELUIp_ELgW i

, " QUESTION 1.06 (2.00) l Concerning the effects of Control Rods on reactor powers

^

a. EXPLAIN how a control rod withdrawal could' result in a reactor power decrease (reverse power effect). (1.0)

b._ Which~one of the following would be the rod movement. sequence ~

l most likely'to cause the reverse power effect. described above? (1.O).

(1) Deep Rod - 10 notch movement I

j (2) Deep Rod - 1 or 2 notch movement j

(3) Shallow Rod - 10 notch movement l (4) Shallow Rod - 1 or 2 notch movement.

i QUESTION 1.07 (1.50) l Certain License / Technical Specification Limits are placed on the ,

) operation of the reactor to ensure that the design vessel, fuel power, thermal, and pressure limits are not exceeded during normal or transient operations.

STATE the exact minimum or maximum (as appropriate) value for each of the following limits.

NOTE: ASSUME FULL POWER OPERATION UNLESS OTHERWISE SPECIFIED l

j a. LHGR j b. MCPR i

c. MWt

d. Reactor Pressure i

j e. MWt (< GOO psia or < 10% Rated Core Flow) ,

f. Cooldown Rate for Shutdown (Technical Specifications) i i

i 1

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

+g . .

Ji t__PBINCIPLES OF NUCLEAR POWER PLANT OPERATION1 PAGE- 5_

IMEBdODYN8dlGS1 _UE81_IB8NSEEB_8ND_ELUID_E(QW t

QUESTION' 1.00 (1.75)

The reactor is'at the beginning of a fuel cycle (BOC).and exactly; critical. You' withdraw ~ control rcis to. add ~O.06% ^k/k.

NOTE: ASSUME BETA-EFF = 0.OO72.and LAMBDA-EFF.= 0.1.SEC E-1

a. CALCU' LATE-the resulting stable period. .SHOW

( ALL WORK)J (0.5)

'b. - CALC;; LATE how long - it would take. reactor. power to DOUBLE. (0.5)

c. If the reactor were at the End of Cycle (EOC) vice BOC, would the resultant stable period, for the same reactivity addition described in (a), be LONGER or SHORTER? JUSTIFY your response! (0.75)

QUESTION 1.09 (3.00)

Attached Figures t 5 A & B represent a transient that could occur at a BWR-  :

GIVEN: (1) One Recirc Pump Flow Controller. Fails High O Time t = 1.2Jmin (2) No operator actions occur (3) Doth Recirc Pumps were on line G Time t = 0.0 minutes EXPLAIN the cause(s) of the following recorder indications:

a. Core Flow DECREASE (Point 3) ( 0. 5 )~

b. Total Steam Flow DECREASE (Point 10) (0.5)

c. Reactor Pressure STABILIZATION (Point 12) (0.5)

(At a lower level)

d. Reactor Level DECREASE (Point 13) (0.5)

e. Reactor Power DECREASE (Point 16L) (0.5)

f. Since this transient occurred with power at less than rated - (70%) , '

STATE the factor (or program) which is utilized to ensure that ttue' MCPR limit is not exceeded. (0.5)

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

r. ,

~

4t__ERINCIPLES OF' NUCLEAR POWER PLANT ~ OPERATION 1 PAGE 6 IMEBdQQYNBdlCSx_SE61_IB6NSEEB_6ND_E6UID_E69W QUESTION 1.10 (1.00)

Which one of the following' correctly DEFINES " Void Fraction"?

a .- steam volume in mixture / total volume of.the'mixtu're

b. steam volume in mixture / liquid volume in mixture

c. steam mass in mixture / total mass of the mixture

d. steam mass in mixture / liquid mass in mixture QUESTION 1.11 (1.00)

Significant quantities of Hydrogen gas may be generated during, or subsequent to, a LOCA. This Hydrogen gas ... (CHOOSE ONE)

a. is generated primarily by the zircaloy-water reaction if fuel. clad temperature is allowed to exceed 2200 deg F.

b. is generated primarily tn/ the radiolytic decomposition of water.

c. could cause an explosive hazard in the'drywell if allowed to exceed the lower flammability limit of 1%. concentration.

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

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

\

1.

Ni__ELONI_DEg1GN_lNC(QDING_SQEEIY_QND_EUEBGENCY_SYSIEd@' PAGE 7 i'

i QUESTION 2.01- ( l '. 50 ) -

I

} Concerning the Control Rod Drive Hydraulic. ;(CRDH) System:

3

.For'EACH offthe following-(a - c)-SELECT the one response th'at most=

4

_. accurately. describes 1the proper system design.

- .a. The.. minimum Scram Discharge Volume which

would be necessary to-l -accomodate a FULL POWER reactor scram, at EOC, would be approximately'...

t (1) 50 gal i

(2) 200 gal

! (3) 600 gal j (4) 1100 gal ,

i i b. The Scram Discharge Volume is limited to a pressure of ________

i during a reactor scram to ensure complete rod. insertion.

.) -

l -(1) 65.psig l

1 (2) 125 psig i

I (3) 650 psig i

l (4) 1250 psig i

f

c. Following a reattor scram, the Scram-Discharge Volume.is protected j from overpressure by a relief valve which lifts at ________.

(1) 65 psig (2) 125 psig j (3) 650 psig

(: (4) 1250 psig l~

i i

t t

h i

4 t-i- (***** CATEGORY 02 CONTINUED ON NEXT PAGE.*****)-

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

J K

2 __ELONI_DggION_ INCLUDING _g8EEIy_8ND_ EMERGENCY _gy@IEMS .PAGE 8 OUESTION 2.02 (1.00)

Concerning'the RCIC System:

, a. STATE the adverse effect(s) that would occur if the RCIC pump Suction Relief Valve were to STICK OPEN following system initiation. (0.5) j b. .TRUE OR FALSE (0.5) l A RCIC. Low Steam Pressure Isolation will NOT automatically reset when '

! steam pressure is restored.

s 4

QUESTION 2.03 (2.25)

. For each HPCI System related component f ailura listed below, STATE:

(1) Whether, or not, HPCI will AUTO INJECT into the vessel 4

(2) If injection will not occur, WHY i

(3) If injection will occur, one adverse consequence of the system operating with this failed component. -)

i L

a. The Minimum Flow Valve fails to close after the HPCI start. (0.75) 4

b. The HPCI discharge flow element output' signal to the HPCI flow controller is failed to its maximum output. (0.75)

c. The SBGTS fails to operate. (0.75) i I

QUESTION 2.04 (1.00)

! The RHR pump suction valves from the Torus are interlocked to prevent

opening when the Shutdown Cooling RHR pump suction valves are open -

4 this prevents diverting reactor coolant to the Torus while in Shutdown i Cooling.

1 DESCRIBE another potential path for diverting reactor coolant to the j Torus during Shutdown Cooling (and flushing) operations, and i

-STATE the specific operator actions which are required by procedure to ensure that this path is unavailable, f

i l

1 I

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

4_, _ _ . -_ . _ . . _ _ . _ . _ . _ . _ . . . _ . . , _ _ _ _ . . _ , _ _ . _ . . _ _ . _ . - _ _ _ . _ _ . . _ , _ _ . . _ _ . - ,

J 2 i__E66NI_DESIGU_INCLUDIUQ_@8EEIy_8ND_EMEBGEUCy_gySIEMS PAGE- 9 ,

i i

I i

i i DUESTION 2.05 ( .50) ,

i EXPLAIN why'it is essential to ensure that the-Diesel Generator I governor is. positioned fully to the LOW SPEED STOP, prior.to

performing a local start.

i

! OUESTION 2.06 ( .50)

, While shutting down a Diesel Generator from the Control Room, j following a Surveillance Test, you inadvertently allow the. generator to trip on REVERSE POWER prior to opening the output breaker.

1 j DESCRIBE how you would return the Diesel Generator-to Standby

Readiness, given this condition.

1

)

I QUESTION 2.07 (1.50)

! Unit 2 is at f ull power when a Load Reject occurs due to a' trip of i the MTG Output Dreakers.

a. STATE the specific signals (and their sources) which activate the i Load Reject circuitry. (0.5)

! b. DESCRIBE the response of the Turbine / EHC System to this event. (1.0) 4 1

I QUESTION 2.08 (1.50) '

i STATE the response (Component and Reaction) of the Main Steam Isolation I j Valves to a LOSS of EACH of the following:

1 i

a. AC Power i i

i -b. Control Air I

c. Drywell Control Air i

4  :

j QUESTION 2.09 (1.50) l An automatic LPCI injection signal is received:

STATE the three specific conditions (setpoints, if applicable) i that acauld be required.to allow you to initiate Torus Spray.

(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

2 t__Eb8NI_DESIGU_ LUC 6UDING_S@EEIY_8UD_ENESGEUCY_SYSIEUS PAGE 10

'OUESTION 2.10 (1-00)

During a Loss of-Offsite Power (LOSP) with a concurrent LOCA condition on Unit 1 (Core Spray pumps in operation), another LOCA signal, occurs on Unit 2. Which one of the statements below most accurately reflects the response of the Core Spray Logic?

a. Unit.I Core Spray pumps 1A and 1C will. TRIP and Unit 2 Core Spray pumps 2A and 2C (only) will START.

b. Unit I Core Spray pumps 1B and 1D will TRIP and Unit 2 Core Spray pumps 28 and 2D (only) will START.

c. All Unit 2 Core Spray pumps will START, in addition to the four pumps already operating on Unit 1.

d. All Unit 1 Core Spray pumps will TRIP and Unit 2 Core Spray pumps must be MANUALLY sequenced ON, as the Diesel Generator capacity allows.

QUESTION 2.11 (1.50)

A Core Spray initiation signal is present on Unit 2. You MANUALLY secure the "C" Core Spray pump due to increasing reactor vessel level,

a. STATE the i ndi cati on (s) that would be provided on the control panel that tha pump was manually secured. (DO NOT address flow, pressure or other metered indications) (0.5)

b. LIST the conditions that would have to exist and/or the actions that you would have to take to permit the pump to automatically restart, if needed. (1.0)

OUESTION 2.12 (1.00)

STATE whether ADS would DEPRESSURIZE or NOT DEPRESSURIZE the reactor for each of the following conditions:  ;

CONSIDER THE TOTAL ADS RESPONSE AND ASSUME NO OTHER OPERATOR ACTIONS OCCUR

a. After initiation of the 120 second timer, but before it in timed out, Drywell Pressure DECREASES to 1.5 psig. (0.5)

b. After initiation of the 120 second timer, but before it is timed out, Reactor Water Level INCREASES to -90 inches. (0.5) ;

i

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

i 1

-- .. - .~ . . .

=2,__ELGUI_DESIGU_lHGLUQ1U9_E8 Eely _GUQ_EdEBgEUQY_SX@ led @ PAGE 11 i

i

. QUESTION 2.13 (2.00)

LIST the four (4) conditions that must exist for a Diesel Generator Output Breaker to Close following an AUTO start signal. ,

t 8

4 I

i i

i i

d i

j i

i l

1 s

i 1

4 i

t i

4 a

f i

s 1

i i

1 d

1 i 4 i

1 i

). (***** END OF CATEGORY O2 *****)

sz__IUSIBUUEUIS_80p_CQUIBgLS PAGE~ 12' l

DUESTION 3.01 (1.00)

Attached Figure # 869 shows the " Error Signal-Input" to the Reactor Recirculation System Flow Controller and shows four potential-Controller Output signals (a - d).

6-Which one of the output signals is the appropriate-response to'the input signal shown?

OUESTION 3.02 (1.00)

The reactor is at 100% Rated Thermal Power "A" Feedwater Level Control is Selected All Controls are in Normal / Automatic A failure in a Rx Vessel Level Narrow Range Instrument has occurred and resulted in the following related trips / indications:

"A" NR level indication reads at MINIMUM "D" & "C" NR level indications read at MAXIMUM ,

REACTOR WTR LEVEL A ADNORMAL Annunciator - ON t

REAC VESSEL WATER LEVEL LOW-LOW CHAN A Annunciator - OFF REAC VEGGEL WATER LEVEL LOW-LOW CHAN B Annunciator - OFF Two channels of Level 8 have TRIPPED One channel of Level 3 has TRIPPED Feedwater flow in at ZERO State which NR level transmitter has failed AND state in which direction it has failed (HIGH/ LOW).

l t

i

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

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

c

?z__IUSIBUMEUIg_AND_CONIBOLg! PAGE' 13 QUESTION 3.03 (1.00)

Concerning Level Instrumentation:

a. Attached Figure #15 shows the relationship'betweenLindicated'and

~

-actual water level on the' Scram Bartons. Each line represents the relationship at a different pressure - O, 200, 600, and 1000 psig.

NOTE: THE LINES ARE NOT MARKED BY PRESSURE INTENTIONALLY If actual level is 42 inches, with a reactor pressure of 600 psig, STATE the level which would be seen inLthe control room-.(indicated).. -(0.5)

b. Figure #4 shows a Yarway Column - STATE whether, or not,.this

-column is temperature compensated. JUSTIFY your response. (0.5)

QUESTION 3.04 (1.00)

Concerning the PCIS Logic:

a. State whether the normal PCIS sensor relays are ENERGIZED or DEENERGIZED when an UNSAFE condition exists. (0,5)

b. LIST the TWO systems whose isolation valve control logic is the opposite of the above referenced (part "a") norm. (0.5)

QUESTION 3.05 (1.00)

Concerning the Local Power Range Monitors (LPRM's):

a. TRUE or FALSE An LPRM's gamma sensitivity decreases with core life due to-the:

depletion of the ionized Argon gas as it is used up in estab-lishing a current flow. (0.5)

b. STATE the purpose of the AMPLIFIER RANGE SWITCH for each LPRM. (0.5)

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

k 33s- - INSTRUMENTg_AND_ggNTRgLQ -PAGE 14-

}~

DUESTION: 3.06_ _ (1.50)

}. Concerning the APRM I Circuitry:

i ~

7 a.- STATE the-(range of) time by which the flow-biased signal is time delayed. (0.5)

. 'l l b. -STATE what causes'the circuitry to arriveLat a particular time

' delay (within the range described above). (0.5) .

l c. The APRM flow-biased Scram trip is present when the: MODE SWITCH contacts are (OPEN/ CLOSED /BOTH OPEN AND CLOSED) - CHOOSE ONE!- (0.5)

{

E

! GUESTION 3.07 (1.00) i-

] The following plant conditions exist: i NOTE: FIGURES #-875 A & D ARE ENCLOSED FOR REFERENCE

]

j Inlet condensate valve to SJAE A intercondenser is FULLY OPEN Outlet condensate valve to SJAE A intercondenser is FULLY CLOSED Condensate pressure from SJAE A intercondenser is 45 psig

. Steam isolation valves to SJAE B are FULLY OPEN

j. Steam pressure at PS (1-151)' is 182 psig Steam pressure at PS (1-166) is 173 psig You take HS 66-14 (AOV FCV) to the OPEN position.

4 STATE whether FCV 66-14 will open. JUSTIFY your response.

J l

QUESTION 3.08 (1.00)

Concerning the operation of the Rod Block Monitor (RBM): ,

s -

You select a center rod for movement.

Recirculation Flow is 33%

I The Ref erence APRM reads 36%

The LPRM inputs to the Averaging Circuit read (an average of) 40%

l Withdrawal of the selected rod causes an average power increase of 6%.

i

With no operator actions (except the rod withdrawal), STATE whether l

or not a Rod Dlock would be received. JUSTIFY your response.

4 l

1 I

4

(*****

~

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

3:__INSIBUMENIS_AND_CONIBOLS- PAGE 15- q QUESTION 3.09 ( 1. 50)-

The~ plant'is shutdown and the reactor cavity is filled. . RHR-System 1 (Pump A) is operating in the Shutdown-Cooling mode, via Recirculation Loop B.

During maintenance, an IM causes a Group II isolation signal to be-erroneously generated.

a. STATE the specific-'RHR system-responses to the signal. (All' valves, pumps,etc. - identification numbers not required). (1.0)

b. LIST all, if any, operator actions required to rese't'the isol-ation (once the signal is verified as erroneous and cleared). (0.5)

DUESTION 3.10 (1.50)

Concerning the Fire Protection System:

a. MATCH the Detection System (COLUMN A) with its Useage.(COLUMN B) (0.75)

COLUMN A COLUMN B-

1. Rate of Rise Thermal Detector a. General area coverage throughout the plant

2. Continuous Strip Detector

b. Spreader Rooms and around any

3. Ionization Smoke Detector oil reservoirs

c. Cable Tray protection

b. Diven that a fire detection system provides an initiation function - STATE how the Fire Protection / Raw Service Water system responds / aligns. (0.75)

OUESTION 3.11 (1.00)

For each of the following indications on a P-1 edit, STATE whether it would indicate VALID or INVALID' data,

n. ITER = 2 (0.5)

b. IXYFLAG = 1 (0.5)

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

~-

yc

}. - '

J 4 it r _ <<

131__JNSIBUDENIg_6NQ_ggNIggLS ., PAGE. 16r t)

OUESTION- 3.12 (1.00) [

Answer the following with respect-to the Rod Sequence ~ Control System:

-During the performance of SI 4.3.D.3.a concerning RSCS Operability prior.to startup, a " TEST FAILED" light is received (Panal ~ 9-28) after depressing the "COMPARATOR CHECK A" Pushbutton. STATE whetner this-indicates a SATISFACTORY or UNSATISFACTORY' response. JUSTIFY your 1%

response! ,

QUESTION 3.13 (2.00)

The following plant conditions exiits l;

' Jet Pumps 1 -

10 Differential Pressure (Meter) = .15 psid-Jet Pumps 11 - 20 Differential Pressure (Meter) = 3 psid-RECIRC LOOP A ONLY OUT OF SERVICE (Annunciator) = ON

\

a. The TOTAL CORE FLOW Recorder would calculate core flow?by whfch 1 of the following methods? '

2 (1.0)

(1) Loop B Jet Pump Flow + Loop A Jet Pump Flow i.;

7 (2) Loop B Jet Pump Flow - Loop A Jet Pump-FlowI '

(3) Loop D Jet Pump Flow Only (4) Loop A Jet Pump, Flow Only

b. After 5 minutes, the operator opens the Discharge Valve of Recirc 3 Pump A to maintain the,A Loop'femperature. The TOTAL CORE FLOW l Recorder would calculate core flow by which one of the (611owing sl methods? ,

, (1.0)

(1) Loop B Jet Pump Flow + Loop A Jet Pump Flow l

(2) Loop B Jet Pump Flow - Loop A Jet' Pump Flow (3) Loop D Jet Pump Flow Only . l (4) Loop A Jet Pump Flow Only H

l 6

3 il

(***** CATEGORY'Op CONTINUED ON NEXT.PAGE *****) ,

i

sz__IUSIBWMEUIS_6dD_G9dIBQLF PAGE 17 OUESTION 3.14 (1.50)

For each of.the Recirculation Pump Flow Control Limiters-listed below:

,. LIST ALL input signals or-design features which will activate (enforce) it 2-

a. 20% Minimum Speed (0.5)

b. 20% Minimum Speed (0.5)

c. 75% Minimum Speed (0.5)

I J

i J

b 1

I 3,

h 1

i 1

d R

)

I [

! 7. w * * *

• END OF CATEGORY 03 *****)

~ .

.y r/ L, v ,

't ,

i 4 t__EBQGEQQBEQ_ _UQBd862_8BUQBd862_EdgBGEUQY_8BD ' PAGE 18 88 Dig 6QQlC66_QQt!,T.BQL E 4

i g (

] i I

s1 QUESTIDi['4h01 (1.00) <

I \s s ,

EDI-2, " Containment Control", states in Step PC/P-5:

l' IF drywell pressure exceed's 55 ' psig (70 pqia), T, HEN

... vent the primary containment . , . . topr' educe and maintain pressure below 55 psig as follows:

s.

's ..

1. . IF suppression pool water level-is below 20 feet, q THEN' vent the suppression chamber. ,

EXPLAIN why (if condittains allow) the suppression chamber is vented first (to reduce drywell pressure). Include the benefit acheived and how-this i s accompli shec'.

i>

i-OUESTION 4.02 (1.00) ,

Attached '

Figure "F" st}ows the Heat Capacity Temperature Limit (HCTL). ,

EXPLAIN the purpose for having the HCTL. (i . e. , what does the curve protect against) i 4s i

OUESTION 4.03 (1.00) ,

, I -

Special concern is required for rettte,ning MOV's, that have been '

MANUALLY CLOSED, to service. ,

g- ,

)

a. DECRIBE how to verif,y the operabi)4ty of an MOV in a safety  :

i . system h. hat has been manually closed. / .; (0.5)

? } \ '

\

b. DESCRIDE the malfunction that this operdbility check is-il, intendad to identify. - ( 0. 5 )

A

, \

J' OUESTION 4.04

's (1.50P

'( t,

.( ,

s i

- \

i, .

STATE the cri teria that are used to determir.ei i- ' an MR shoul d, be -

checked EMERGENCY, IniMEDI ATd ATT"pNTION, or' ROUTINE MAINTENANCE.

(i . e. , Define-each classificatioh) J ,

1 i

1 ,

i y + .

N .g

)

r

% :S ' .

t

( **** * ' CA Et'{0RY' -04 CONTINUED ON NEXT PAGE ***n)

I A. .- - - - . ., . , . . -...--..-;,..- , -

31__EBgCEDW8EE_2_U9BdG62_6BU9Bdeb2_EdgBGENCy_8ND PAGE 19

'BSD196991986_G9 BIB 96 QUESTION 4.05 (1.00)

Concerning a Plant Startup:

a. Step A.12 of GOI 100-1 states:

"RHR system in standby readiness condition.in accordance with 0I-74" TRUE OR FALSE

~

Step A.12 may be signed off with RHR pumps A $< C operating to provide Shutdown Cooling. ( 0. 5 )'

b. Step A.22 states:

" Reactor recirculation system ready for service, in accordance with OI-68" TRUE OR FALSE Step A.22 may be signed off with Recirculation Pump A Out of Ser-vice for Shutdown Cooling requirements (only). (0.5)

DUESTION 4.06 (2.50)

LIST the five (5) entry conditions to EDI-2, " Primary Containment

~

Control". (Include Setpoints, as appropriate). (2.5)

(***** CATEGORYE04 CONTINUED ON NEXT.PAGE.*****)

i. .

Y L Sz__EB9GEQUBES_;_NgBd862_8BNQBU662_EbEBGENCy_8Ng PAGE 20

-RADIOLOGIGAL CONTROL ,

d l QUESTION 4.07 (2.00)  :

E'! ATE which Emergency Classification is appropriate for the following

, definitions.

1

a. Events are in progress or have occurred which involve actual i or potential substantial degradation of the level of safety of the plant.

1

, b. Events are in progress or have occurred which could develop

! into, or be indicative of, more serious conditions which are not yet fully realiced.

c. Events are in progress or have occurred which involve I. actual or imminent substantial core failure with the poten-tial for loss of containment integrity.

, d. Events are in progress or have occurred which involve j an actual or likely major failure of plant functions needed 7, for protection of the public.

1 i

i

QUESTION 4.08 (1.00) 1 I Concerning the Routine Radiation Work Permit (RRWP)

j FILL IN THE BLANK FOR EACH OF THE FOLLOWING

a. No one is permitted to receive a dose greater than _________ per l day cm any RRWP or any combination of RRWP's. 'l q b. Personmnel with an exposure limit of ________ per quarter or less will not be allowed to enter en an RRWP.

c. No RRWP user will exceed ______ of their remaining allowable dose q (RAD) in one day as indicated on a current personnel listing, except i

by special permission of the Health Physics section supervisor or an assistant Health Physics section supervisor.

I d. Entries into dose rates of ________ or greater will require the.

f approval of-Health Physics supervision (see RCI No. 1) and the pre-

sence of a Health Physics technician. Additional HP surveys may be required prior to entry. These requirements will be noted on the RRWP.

i f

J l

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

-4 s_ ?8QGEQUBES_ _UQBd862_6pNQBd862_EMEBGENGY_6NQ .PAGE 21;

~1D1969 GIG 86_G9 NIB 96 t

- GUESTION 4.09 (1.50) 1 Concerning operations with Feedwater Heaters. removed from service, 3

as per OI-2, Condensate System:

. a. STATE whether kilowatt output will INCREASE / DECREASE / REMAIN THE.

SAME when the highest pressure feedwater heater is removed.from ser-vice. JUSTIFY you answer. (0.75)

b. STATE whether kilowatt output will INCREASE / DECREASE / REMAIN THE SAME when other than the highest pressure feedwater heater is removed from service. JUSTIFY your answer. (0.75)

QUESTION 4.10 (1.50) j STATE the three (3) conditions which must be present to allow the Unit i and 2 Diesel Generators to be pcralleled with the Unit 3 Diesel Generator. i QUESTION 4.11 (1.00)

Per 0I-24, " Raw Cooling Water":

a. STATE the header pressure at which the STBY (spare) RCW pump will automatically start. (0.5)

b. 01-24 cautions the operator NOT to run.the 3D RCW pump in automatic. STATE the reason for this caution. .(0.5)

JOUESTION 4.12 (1.50)

An Operability Test is conducted on a safety-related system, following the installation of an approved modification. LIST three-(3) sep-arate criteria which would procedurally require that.a TEST DEFICIENCY be. initiated (documented).

l

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

u-.

.(*****.+******* END OF EX AMINATION ***************)

1. PRINCIPLES OF~ NUCLEAR POWER PLANT OPER8TlgN2 PAGE 122 IHEBMgDYN8 MIGS2 _ HEAT TRANSEEB_8NQ_E6Mlp_E6gW ANSWERS -- BROWNS FERRY 1, 2&3' -86/12/01-BROCKMAN, K.

-ANSWER 1.01 (1.50)

a. c (1.0)

> - -b. (Excessive voiding'in the' bypass region resulting in) ~ unreliable LPRM readings (0.5) (1/2 credit.for LPRM overheating) (0.5) i REFERENCE General Electric Heat Transfer & Fluid Flow, Chapter.8

~

BFNP: HTFF, Chapter 8, L.O.'s 9.4 & 9.5 DWR K/A: 293008 K1.32 (2.5/2.6), K1.33 (2.4/2.6) 215005.K5.02 (2.7/2.8)

.i ANSWER 1.02 (2.00)

a. P -P - (0.5) act sat

b. (1) INCREASE (More Subcooling at the pump suction)

(2) DECREASE (Reduced pressure at.the eye of the pump results in being closer to saturatio pressure)

? (3) INCREASE (Further to saturation temperature and increased-density causing less static head) (3-@ 0.5 each)

REFERENCE 4

General. Electric Heat Transfer and Fluid Flow,' Chapter 6;.

. BFNP: HTFF, Chapter 6, L.O.'s 7.7, 8.1, & 8.2 I

i

BWR K/A

293006 K1.10 (2.7/2.8)

I h

t g- -y.. -.,q 3: g -g-- - - , - g - -

y i-&+,w y- ** f ^' F W~-M 4 7

• 1 4

111__EBINQ1ELES_QE_UyGLE8B_EgWEB_E66NI_QEEBBIlgN1 LPAGE 23 ISEBMgQYN6 MIGS 1;UEBI_IB6NSEEB_6NQ_Ebylp_E6gW

. ANSWERS'-- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN,'K.

i

, ANSWER 1.03 (1.00)'

C

. REFERENCE General Electric Reactor. Theory, Chapter 5

! BFNP: Fb: Theory, Chapter 5, L.O. 2.4 t

BWR K/A: 292005 K1.09 (2.5/2.6) 4 ANSWER 1.04 ( .50) 4 Reactivity GTE Beta-effective REFERENCE General Electric Reactor Theory, Chapter 3 l r BFNP: Rx Theory, Chapter 3, L.O. 5.6

, BWR K/A: 292003 K1.07 (3.3/3.3) f l

1

. I

I j

t 3

-: t i a

' i 1!

g- -

s _

L

-i

~ 1. PRINCIPLES OF NUCLEAR POWER PLANT CPERATION S PAGE . :24' ISEBMQQYN6 digs3 _SE61_IB6NSEEB_6NQ_ELUlD_E(QW,  ;

i .' ANSWERS -- BROWNS. FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER. 1.05- (1.50)

a. The decrease is due to the (decay of Xe-135.and the) increased burn-

., out (which is immediately seen by the.EXe]), being much greater than-the Xe production, (which is. time-delayed from the~1ower power.)

b. The flux will redistribute from outside (and2 higher) peaked and 4

center (and lower) depressedEto center peaked and outside' depressed.

(0.25) This is because the Xe concentration shifts from.centerfpeaked to more peripherally peaked due to power production distribution.. (0.25)-

! c. Use caution when pulling peripheral rods (0.25) due to the L increased rod worths (increased thermal flux) (0.25). (0.5 each)-

REFERENCE j General Electric Reactor Theory, Chapter 6 j BFNP

Rx Theory, Chapter 6, L.O. 2 f

BWR K/A: 292006 K1.07 (3.2/3.2), K1.08.(2.8/3.0), K1.10 (2.9/2.9) i & K1.14 (3.1/3.3) f

' ~

i ANSWER '1.06 (2.00)

!' ~

i a. The negative reactivity added by the increased voids generated.by the rod withdrawal is greater than the positive reactivity added by-

the reduced rod'absorbtion. (1.0)

V

b. (4) (1.0) 1

-REFERENCE.

General Electric Reactor Theory, Chapter 5 BFNP: Rx Theory, Chapter 5, L.O. 3 i

i 1

~ . , . . , , . , . _ . _ , . . . . , - , . . _ . . ., . . , ..w ., ..,,-_,s,_, - , , - - _ -, __ _

t ig;_ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION 1 PAGE 25; ISEBUQDYN8dlCSz_SE8I_lB8NSEEB_8ND_EbulD_E6QW-ANSWERS -- BROWNS FERRY 1,~2&3 -

86/12/01-BROCKMAN, K.

BWR K/A: 292005 K1.04 (3.8/3.8)

ANSWER 1.07 (1.50)

a. LTE 13.4 Kw/ft b.: GTE 1.07

c. LTE 3293 MWt (8-hour) -OR- LTE 3358-MWt (Inst)

d. LTE 1375 psig

e. LTE 823 MWt

f. LTE 100 deg F/ hour (0.25,each);

REFERENCE General Electric Heat Transfer and Fluid Flow, Chapter 9 BFNP: Technical Specifications, L.O.'s 4 &.34; HTFF, Chapter-9, L.O.'s 2.2, 3.2, & 4.2; GOI 100-1 -

BWR K/A: 293007 K1.05 (3.3/3.5), K1.10 (2.7/7.9) 293009 K1.06 (3.4/3.8), K1.21 (3.1/3.6),-K3.8 ( 2. 7 / 3.'.1 )

~

W Y'- gr

• r -"Wgy w w'T 7 q h 9 f F- w 1 & -*MyM W W

- 1. PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION 2 PAGE 26 ISEBdg DYN8dICS2_ b E81_IB8 NS EE8_8N D_ E6 UID_ ELQLi ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN,_K.

ANSWER 1.' 08 . (1.75)

I a.- Period = (Beta - Rho) / (Lambda

• Rho)

= (0.0072.- 0.0006) / - ( Or .1 *-0.0006) = 110 seconds (0.5) b.- Period = 1.443

• DT -> DT = Period / 1.443 = 110 / 1.443 I

= 76.23 seconds (Error carries forward from (a)) (0.5)

c. Period will be SHORTER (0.25) This is due.to the fact that the Beta for EOC is shorter (~ 0.0058, due to the increased Pu contribution). (0.5) (0.75) f REFERENCE General Electric Reactor Theory, Chapter 3 BFNP: Rx Theory, Chapter 3,_L.O.'s 3.6, 4.6, &'5.9 BWR K/A: 292003 K1.08 (2.5/2.8), K1.09 (2.5/2.6) 4 i-d I

d

~

t f' s

=la__EBINCIE6ES_gE_NUC6E88_EgWEB_E68NI_gEEB8IIgN 2 PAGE 27 IUEBMggyN8MICS2 _Hg8I_IB8NSEEB_8ND_E691D_E6gW

.. ANSWERS - BROWNS. FERRY 1, 2&3 --86/12/01-BROCKMAN, K.

ANSWER. 1.09' (3.00)

a. Recirculation Pump Runback (From < 20% FW Flow) (0.5) b.- EHC' controlling reactor pressure (at 920 psig)' (0.5) t

c. EHC' controlling reactor pressure (at 920 psig) (0.5)-

d. Increase in the suction of the Recirc Pump (f rom _Downcomer) (0.5)

e. Reactor Scram (APRM Flow Biased G O.66*w + 54%) - (0. 5)

f. K-f is utilized to increase the steady-state MCPR limits at reduced flows (FLCPR Program). (0.5).

REFERENCE BFNP: Transient Analysis LP, Transient # 5,'L.D. ALB BWR K/A: 202001 A2.05.(4.0/4.0) 293009 K1.26 (3.3/3.3) K1.28 (3.5/3.5)

ANSWER 1.10 (1.00) a REFERENCE General Electric Heat Transfer and Fluid Flow, Chapter 8 1l l

BFNP: Thermal Hydraulics L/P, p 6; HTFF,~ Chapter 8, L.O. 6  !

I i

l

,. i

!- A__EBINGIEGES_QE_NyGLE88_EQWEB_E68MI_QEEB8IlgN 1 PAGE 28

.ISEBdQDYN8dlGS2_UE81_IB8NSEgB_8ND_E6UID_E60W , _

^

JANSWERS -- BROWNS FERRY.1, 2&3 -86/12/01-BROCKMAN,lK.

ANSWER 1.11 (1.00) a REFERENCE-BFNP: EGT106.015 (MCD, Ch 1), L. D. 's "B" & "D" BWR'K/A: 223001 A1.05 (3.1/3.3), SWG #7 (3.7/3.8)

PWG' K1.15 (3.1/3.3), A1.14 (2.9/3.4) l

.q u

2i PLANT [pgSIGNj_I_NGLUDIN9_E8FETY_ GNp_gMERggNgy_gygTgMS PAGE 29 ANSWERS -- BROWNS: FERRY 1, 2&3-' -86/12/01-BROCKMAN, K.

ANSWER 2.01 (1.50)

a. (3)' 600 gal

'b. (1) :65.psig Ec. (4) 1250 psig -(0.5 each)

REFERENCE BFNP: OPL171.OO5, L. O. 's "P", "O",- & "S" BWR K/A: 201001 K4.11 (3.6/3.6)

ANSWER 2.02 (1.00)

a. An open relief valve would decrease the available head f or pump suction (NPSH) (and might cause air binding of the - pump.') -- OR -

Loss of inventory (water) from CST or Torus. (0.5)

b. TRUE (0.5)

REFERENCE BFNP: OPL171.040, L. O. 's "C" & "D" BWR K/A: 217000 K1.01 (3.5/3.5), K4.05 (3.2/3.5), K5.01 (2.6/2.6), 1 A2.11 (3.1/3.2), A3.04 (3.6/3.5) 1

7 21__E66NI_ DESIGN _INgbyDING_S8EEIy_8ND_EMEBGENCy_gySIEMS PAGE- 30 ANSWERS -- BROWNS' FERRY 1,_2&3 -86/12/01-BROCKMAN, K.

ANSWER 2.03 (2.25)

~

a. WILL INJECT (0.25) HPIC injection will be at less than design-

-flow - (0.50) (which may not meet the_high pressure coolant makeup

, requirements of.certain accident conditions). - OR - Increase in

-Tcrus level (due to HPCI pumping thru MFV) - OR - Loss of CST ~~ Level i to the Torus.

b. WILL NOT INJECT (0.25) Minimum flow pr atection will .tue lost or the controller will keep the turbine speed at minimum (0.50).

c. WILL INJECT (0.25) There is a potential'for a (loss of vacuum in the HPCI Barometric Condenser and) release of airborne activity to the HPCI Room (0.50).

REFERENCE BFNP: OPL171.042, L.O. "A.B"; Accelerated Requal Exam 7.10 l

BWR K/A: 206000 K6.06 (3.1/3.2), A2.03 (3.5/3.5), A2.14 (3.3/3.4), i A2.15 (3.4/3.5) i I

ANSWER 2.04 (1.00)

During flushing and Shutdown Cooling operations, the RHR Minimum Flow Valve could open diverting reactor coolant to the Torus. (0.5)

(OI-74 has been revised to) require the the Minimum Flow Valve Handswitch be placed i.n the BYPASS position prior to flushing and Shutdown Cooling operations. (0.5)  !

REFERENCE BFNP: OPL171.044, L.O. "E.2"; OI-74; General Electric-Systems Review on RHR 1

BWR K/A: 205000 K1.08 (3.9/3.9), K3.02 (3.2/3.3), Ka.07 (2.7/2.0)

K5.02 (2.8/2.9), A2.10 (2. 9/2. 9) , A4. 05 .(3. 2/3. 2) .

a

~

n 1

4

2. PLANT DESIGN INCLUDING 7 SAFETY AND EMERGENCY SYSTEMS J PAGE - 131 :

ANSWERS -- BROWNS FERRY ~'1, 2&3 '-86/12/01-BROCKMAN, K.

4 a-e J

{ ~ ANSWER 2.05 ( .50)

Not having the governor at the' lowest setting can result in a mechanical.overspeed trip of the diesel.

REFERENCE BFNP

LER-86'-26-(7/28/C6 - EDG "D")

i

. i 5

t 1

BWR M/A: PWG (3.2/3.6)

ANSWER 2.06 ( .50)

] The Field Breaker would have to be (locally) reset, and i

the (HEA) Lockout Relay would have to be (locally) reset. (0.25 each)

REFFRFNCE BFNP: OPL171.038, L.O.'s "E" & "F"; OI-82, P 10

]

i BWR K/A: 264000 K4.01 (3.5/3.7), A1.09 (3.0/3.1), A2.02'(3.1/3.1)

I r

?

4 s

4 i

i

< . - - - - . - - - _ - _ - ,., m, , , . .+-w-.,am , ,,a,.-- +w - , - --ne + - n.

b.__EL8NIiQESIGU_IUCLUDING_S8 eel $_8MD_EdEBGENCY_@Y@IEUS 'PAGE :32 ANSWERS -- BROWNS FERRY-1, 23<3 -86/12/01-BROCKMAN,.K.

L ANSWER 2.07 (1.50)

-a. (Greater than 40% mismatch between the generator output and the l turbine power) as sensed by generator-stator amps and turbine crosse j -over pressure. (0. 5)'

7 b. (1)-TCV's Trip Closed (fast-acting solenoids)

. '(2) Bypass Valves Open

'(3) EHC Load Selector runs back (4)- Control Valves partially reopen G < 40% mismatch (5) CIV's throttle to control turbine speed 0 1800 rpm (0.20 each')_

i i REFERENCE EFNP: CPL 171.010, L.C.'s "J" 3< K" I,

i l

1 1

1-BWR K/A: 245000 K3.01 (3.4/3.7)

't i ANSWER 2.08 (1.50) l

):

a. No Response (Redundant DC solenoids keep valves open)

b. Outboard MSIV's Close

! c. Inboard MSIV's Close (0.5 each)

I REFERENCE BFNP: OPL171.OO9, L.O. "E" l

1 t

i BWR K/A 239001.K6.01 (3.1/3.3), K6.02~(3.2/3.2) f I

f h

., - ,z-... m -- .w.# - - , , , - -.,-e ,, __ - ,, ., ,.,,m n ,- ,- -e , , - < - , , . , , , ,

~2a__E68NI_QEgigN INGLUQINQ_SAEETY_ANQ_gdgRQgNgy_gy@Tgd@- 'PA'GE 33 --

ANSWERS._-- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER ~ 2.09 (1.50)

Reactor-Water Level (0.25) GTE - 39" (0.25)

Drywell Pressure (0.25) GTE + 1.96 psig (0.25)'

SELECT - RESET Switch (0.25) in SELECT (0.25)

REFERENCE BFNP:-OPL171.044, L.D. "E.9" BWR K/A: 226001 K4.03-(2.9/3.1) 203000 K4.10 (3.9/4.1)

ANSWER 2.10 (1.00) b REFERENCE BFNP: OPL171.045, L.O. "I" BWR K/A: 209001 K1.08 (3.2/3.4), K1.iO ( 3 . 7 / 3 . 8 )' , K2.03 (2.9/3.1)

K4.08 (3.8/4.0), K6.01 ( 3.' 4 / 3. 4 )

!2 __EL6NI_

t DESIGN _ INCLUDING _@@EEIX_6MD_EMEBGENCY_gy@IEMS. PAGE ~34;

' ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.-

3 -

l

' ANSWER 2.11 (1.50) l .

a. Yellow light above the Core Spray pump switch (on Panel 9-3).

b. Initiating conditions cleared (GT -114.5" RVL & LT 2.45 psig DWP).

Reset Core Spray Initiation signal. (0.5:each)

REFERENCE BFNP: OPL171.045, L.O.'s "F" & "H" s-BWR K/A: 209001 K2.03 (2.9/3.1), K4.OO (3.8/4.0), A3.02 (3.'8/3. 7 )

ANSWER 2.12 (1.00)

a. DEPRESSURIZE (DWP is a seal-in signal)

b. NOT DEPRESSURIZE (RVL i s not a seal-in signal) (0.5.each) 1 REFERENCE

, BFNP: OPL171.043, L.O. "E" t,

i BWR K/A: 218000 K:2. 01 (3.1/3.3), K4.01 (3.7/3.9), K4.03-(3.8/4.0)

! K5.01 (3.8/3.8) t I

i a

I I

3- < ,,.-,:,,,.--- n., ,. , - ,,w,lv,n , . . - . -~ n n,- - - . . , , , - - - - ,. . -n-e

y

'2c__E68NI_DE@lGN_ING6UQ1NG_S8EEIY_8NQ_EMEBGENGY_SYSIEMSl !PAGE: 35 ANSWERS -- BROWNS FERRY 1, 2843' -86/1'2/01-BROCKMAN, K.

ANSWER 2.13 (2.00)

(1) Diesel started and at' rated speed. -

'(2) All other supply breakers to the'4160 volt Shutdown Board Open.

- (3), No supply breaker lockout (EDG).

~

-(4) Undervoltage on the 4160 volt Shutdown Board. (0.5.each)

REFERENCE BFNP: OPL171.038; L.D. "E" BWR K/A: 269000 K1.01 (3.8/4.1), K5.05 (3.4/3.4).

O

_ ~ . . _ . . . . .-

31- lUSIBUMENIS_AND_QQUIBg($ -PAGE: 136-ANSWERS -- BROWNS FERRY:1', 2&3 -86/12/01-BROCKMAN, K.

l t

ANSWER- 3.01 (1.00) a-d

.1 REFERENCE.

BFNP: OPL171.OO8, L.O. "B.2" h

j BWR K/A: 202002 K4.02 (3.0/3.0) K5.03 (2.4/2.4)

J i-  ; A'NSWER 3.02 (1.00)

(1) A (LIT 3-53)

(2) Low .(0.5.each)

REFERENCE GGNS: OP-B21-501 BFNP: OPL171.OO3, L.O. "J"; OPL171.012, L.O. "E.3" ARP- 9 XA-55-5A-8, XA-55-5A-30, X A-55-5B-4 , : XA-55-5B-5

~

i BWR K/A: 259002 K5.01 (3.1/3.2) K6.5 (3.5/3.5) f a

4 f

4 w

+

.{ , * *

, ~ ., . . . , . .a, g--- ., ., - , - ,. . . . , ..

,. e .- ,,.u.e,, , - . .,-m-.

t-1 3k__INSIBudENIS_8ND_CONIBQLS P A G'E ' 37-

' ANSWERS - -BROWNS FERRY'1,'2L3 -96/12/01-BROCKMAN, K.

T ANSWER 3.03 (1.00)

a. 45.5 inches (+- 0.5 inches)' . (0. 5)

b. Yes-(0.1) It has a Heat TransferLClamp (0.4) 1 ( 0. 5')

REFERENCE BFNP: OPL171.OO3, L.O.'s "B" & "J" BWR K/A: 216000 K4.14 (3.3/3.4); K5.10 (3.1/3.3)

. ANSWER 3.04 (1.00) i 4

a. DEENERGIZED -(0.5) 1

b. HPCI and RCIC (0.25 each) i i

REFERENCE i BFNP: OPL171.017,Section X.C; L.O. C&D J

r i BWR K/A: 223002 K1.04 (3.5/3.5) K1.07 ~(3.4/3.6) K2.'01- (2.5/2.7)'

.i -

t 4

I 1

i i

_.- . . - . _ - _ . _ , _ _ _ , . _ . . - - . - . _ , . , _ _ . .. . _ . . _ ,--. :_ _, ,. . . 2. ._, .. , , , ,

e io

'}g__'INSIBUMEtjT@_ANQ_CQNTBQLS- PAGE 38 s.

. ANSWERS -- BROWNS. FERRY 1,'2&3 --86/12/01-BROCKMAN, K.

L t

i.

ANSWERL 3.05. (1.00): ~

I.

a. FALSE (0.5) i .

b. It allows for compensation for the U-5 depletion ( 0. 5') ..

(1/2 credit for " calibration"'

REFERENCE BFNP:-OPL171.021,Section X.B.1 & .4; L.O. C i

i BWR K/A: 215005 K4.06 (2.6/2.8)

J 4

ANSWER 3.06 (1.50) i O seconds (+3,-0) to 30 seconds

a. (+6,-6) -(0,5) i b. The rate of the power increase !ctermines the time. delay i (fast rise = short delay). -OR- The; absolute value of the l power increase (P(f) - P (i ) ) (0.5) 1.

l c. BOTH OPEN AND CLOSED (0.5) i i

i REFERENCE BFNP: OPL171.022,Section X.B.6 & .7; L.O. F&K '

1 i

i 1

BWR-K/A: 215005 K4.02 (3.7/3.7)

+

1 4

4

f. _

i.

., .-,c ,- ..% , e- ,9 9.., , 3 , , . . ._,.,g ,_,_#,,. , , , . u .c.. ~.s. ,r,-

Iz__IN@IBUMENIS_8ND_CQNIBOL@ .PAGE_ 39 9

ANSWERS -- BROWNS FERRY 1, 2&3- -86/12/01-BROCKMAN, K.

i ANSWER 3.07 (1.00)

NO (0,5) All requirements for.a start permissive are not met. ( 0. 5 ) ~-

(There is a 30 second TD on steam pressure in order for the PCV's.

to open and get pressure up to the setpoint y it.may or may not.cause

-the FCV to close. The Condensate outlet valve is closed, the.

condensate discharge pressure is inadaquate, and the SJAE "B"

$ steam MOV is open)

. REFERENCE BFNP: OPL171.030, Section B.S.g, L.O. "C" t

I i

4 i

i BWR K/A: 256000 K4.10 (2.7/2.7) K4.15 (3.1/3.0) l i ,

I

ANSWER 3.08 (1.00) t l A Rod Block would be received. (0.5) The Averaging Circuit is not

gained down to the reference APRM value; therefore, the 6% power j increase will be sensed as exceeding the rod block setpoint (0.66.* w 0 4 + 24%). (0.5) i REFERENCE BFNP

OPL171.035, Sections X.A.3 & .B.6; L.O.'s D & F t

j a

BWR K/A

215002 K1.04 (3.0/3.0), K4.01 (3.4/3.5),_K4.02 (2.9/3.0) 215005 K4.01 (3.7/3.7), K4.07'(3.7/3.7) i' l,

l 3

3.__IUSIBUMENIg_GUD_CgyIBgLS PAGE 40 ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 3.09 (1.50)

a. SDC Inbd Isol Valve Closes (FCV 74-53) (0.25)

SDC Suct Supply Valves Close (FCV 74-47 & -48) (0.50)

RHR Pump A Trips (0.25)

b. The Inbd Loop Valve Reset PB (XS74-126) must be depressed (0.50)

REFERENCE BFNP: OPL171.044; L.O.'s D.7, E, & F; OI-74,Section III.F & .J BWR K/A: 205000 K4.03 (3.8/3.0)

ANSWER 3.10 (1.50)

a. (1) b (2) c (3) a (0.25 each)

b. Fire pump (s) auto start Raw Water Storage Tanks Isolate Raw Service Water Pumps Shutdown (by interlock). (0.25 each)

REFERENCE BFNP: OPL171.049, Sections X . B. r . (2) & .(3) & .D.2; L.O. E BWR K/A: 296000 KS.06 (2.6/2.7), KS.07 (2.6/2.7), K4.01 (3.4/3.6)

K4.02 (3.3/3.5)

1

3. ' 1NSIBydENI@_6ND_CQNIBQL@~ PAGE 41' ANSWERS -- BROWNS. FERRY.1, 2&3 -86/12/01-BROCKMAN, K.

' ANSWER 3.11. (1.00)

a. VALID

b. . INVALID . (0.5 each)

REFERENCE General-Electric, NEDE 24810, Volume #2 BFNP: HTFF, Chapter 9, L.O. 5.g BWR K/A: 293009 K1.38 (2.7/3.1)

ANSWER 3.12 (1.00)

a. SATISFACTORY (0.5) - The Test Failed light ~ indicates-that the GNC test sequence has halted because of a failed comparison (which is the desired result) . (0.5)

REFERENCE BFNP: RSCS LP; RO 85/02/03; OPL171.025, Section B(10) .

BWR K/A: 201004 A3.05 (3.5/3.7)

~

a

?  %

[ '4.__INSIBQMEUIS_6MD_GOUIBOLS _.P GE -42

, ' ANSWERS -- BROWNS FERRY 1, 263. -86/12/01-BROCKMAN,'K.

t 4 $

ANSWER 3.13 (2.00).

a. (2) j b. ( 2 )' (1.0 each) '

i

$ REFERENCE.

'BFNP: ARP 9-4, XA-55-4A; OPL171.OO7 4 -

4 eI a

BWR K/A: 202001 K1.01 (3.6/3.7) 293008 K1.26 (2.9/3.1), K1.27 (3.6/3.7) i.

i j ANSWER 3.14 (1.50) .

l a. Limit Switch (0.25) and Mechanical Stop (0.25)

b. LTE 20% Feedwater Flow (0.25) or Pump' Discharge' Valve LT 90%'Open (0. 25),

c. Any RFP. tripped (0.25) and Level LTE +27" (0.25)

, REFERENCE BFNP: OPL171.OO8; L.O. "A"  ;

i I

l BWR K/A: 202002 K1.08 (3.1/3.2), K1.09.(3.1/3.2), K4. 06 - ( 3.1/3.~ 1')

A2.06 (3.3/3.3), A2.07 (3.3/3.3) l l

1 1

A 4

y ...-.r. . _ _ _ . . , . . ,,.. . .m,,...,_ , , . __ _ . . , , y , .,__.,.=_r ,. , --_.,e., . . . . . ...yc> 3

1.

4. ' PROCEDURES - NORMALt_8BNgBM8'L _EMEBGENCY_ANQ 2 PAGE 43 88DlQ60@lC86_CgNIBg6 ANSWERS - -BROWNS FERRY 1, 2843 -86/12/01L BROCKMAN,.id.-

ANSWER 4.01 (1.00)

Venting the suppression chamber first takes-adventage of the " pool scrubbing effect" (0.5) to help minimize the radiation released (0.5).

REFERENCE BFNP: EDI-2, PC/P-5,.L.O. "B.6" 4

BWR K/A: 259024 EK3.04 (3.7/4.1), EK3.07 (3.5/4.0).

[ EA2.03 (3.8/3.8), EA2.04 (3.9/3.9), EA2.08 (3.6/4.0) ,

i ANSWER 4.02 (1.00)

$ Maintaining pool temperature and reactor pressure < HCTL' assures .

j sufficient heat capacity in the Suppression Pool- (0.5) to prevent  !

t equipment failure (0.3) due to unstable steam condensation ~ (0.1) during an ADS blowdown of the reactor vessel (0.1).  :  ;

REFERENCE .i I

BFNP: SP/L, p 3; SP/T, p 1; OPL174.657; L.O.' B.7

.BWR K/A: 259024 EK2.08 (4.0/4.1),-EK3.04 (3.7/4.1)

EA2.04 (3.9/3.9), EA3.06 (3.4/3.9) i l I

i i

i I

j i

i

,-r , , - - - , , , - , - , , , , ,,,,r , ,-~ . . . ,. m ..- , , - ..--------r---, , . . , , ~ ., , #,,

1 1;c 4.__EBgCEQg8ES_;_NQBM662_6BNgBMSL2_EMEBgENCy_8ND PAGE. 44 B891gLggIC6L_CgNIBg6 ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

t ANSWER 4.03 (1.00) -

tl

a. It shall be cycled by the motor operator (0.6)(prior to being declared operational), (0.5)

b. Torque switch could become activated before the valve is off its seat (improper operation of the opening circuitry). (0.5)

REFERENCE BFNP: OSIL No. 47, p 2; L.O. "T" n'

N.

BWR K/A: 291001 K1.07 (3.4/3.4)

ANSWER 4.04 (1.50) I '( s EMERGENCY - Action is required to prevent imminent major equipment -

damage or to protect personnel from any imminent threat of bodily injury (0.35/O.15)

IMMEDIATE ATTENTION'- Work is 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 /> (, the next scheduled workday, or upon completion of necessary -

technical evaluations or material procurement) (0.5)

ROUTINE MAINTENANCE - Work is to be performed as manpower a$d circ-umstances permit. (0.5)

.{ >

REFERENCE ,

BFNP: SP # 7.6, Section 4.12 - .14; L.O. "A" BWR K/A: 294001 K1.02 (3.9/4.5) 1

._mh

4 t__EBQGEDUBEQ_ _NQBM862_8BNQBd862_EMEBGENCY'ANQ' PAGE: ZK5-

88 Dig 69GlG86_QQNIBQL ANSWERS -- BROWNS FERRY 1, 2&3- -86/12/01-BROCKMAN, K.

4 ANSWER- 4.05 ( 1~. 00 )

a. TRUE

.i

'b. TRUE - (0.5 each)-

i REFERENCE BFNP: . GOI 100-1, Sections A.12 & A.22 1

f BWR K/A: 203000 K1.14 (2.4/2.6), SWG #5 (3.5/4.4), SWG'#10-(3.7/3.9)

ANSWER 4.06 (2.50)

1) Suppression Pool Temperature (0.25) > 95 deg F (0.25) e

2) Drywell Temperature- (0.25) > 160 deg F (0. 25) I i

3) Drywell Pressure (0.25) > 2.45 # (0.25)

, 4) Suppression Pool Level (0.25) > -1 (0.25)

5) Suppression Pool Level (0.25) < -6.25 (0.25) (0.5 each)

REFERENCE BWROG EDI Users Guide BFNP: EDI-2, Section 2.1; EDI L/P, L.O. B-2 T

BWR K/A: 295024 SWG #11 (4.3/4.5) l

, , , , - - , g g-,y <++ ,--p ,- u,-,_-. , , - ~y,, - -

-w,- + ,y-, , ,-- ,y . -. , - . -e-g yf

3 - _ . -

3 g[I

.p - ^ - -

1 '6 )'

. ,r,9 t J T, -+ m.>; '

N

,;g'-j t. .. .n

.f9  :

._y w e - -y ?

N

-i

' N(>j,t' ,s - .

b I,'jid f  ;

w.

') .Np.,

F yN.__.P.RO_CE_bu_.RES - __Ndf.f.M. .A,_L.s~ ABNORM.A..L. u. EMERGENCY ND f ,- in -PAGE n46

'.. [: RADIOLOGICAL' CON' TROL "g 3 .]

{

)- s, , -,

m /-

ANSWERS BROWNS FERRY 1, 2843 iB6/i2/01-BROCKMAN),K..

s s ._

t i

j-k.s t -

r 4 sb f

f ',, .

J g A t .

..J'-

y , }f.

's G

, ANSWER .4.07 (2.00) . 3

~ ,

s -

Alert '

, . a. -, ,

(Notification ofi Unusual Ekent i i '

-'s - b. p

. c. General Emergency; 5

_.v ( ,

Site Area IEmergencyY

~d. .[ '(0.5'each) e ,

1, REFENENCE. .

/

p 1

. EIH: GET Handbook, pp.57, 58, 60, 61 ,.. x( s, HNP-x-4420, HNP-x-4520, HNP-x-4620', HNP-x-4 720 'c

. BFNP: BFN-IPD,,IP-1, p 1; RQ 85/04/01; REP L/P, L 0. C.2-  ! ,

n 't. , 4 L.

t, ' q g-

<t.,

+

r N, s t  :

s

( .

s s, , ,

DWR K/A: 294001' A1.16 (2.9/4.7) T

.s \ s -

vg

- 3 ANSWER 4.C8 si.OO) f f' I a. 150 mrem s

^,

, i ,

b.s 300 mrem , >

i f

I- c. , 50 X *

, s. ,

d. 1 Rem / hour , x (All +0%,-1C%) (0.25 each)

, s i k l

N.N$FERENCb ' '

BFNP: RCI-9, Secti on I r. E} d. p. "F",

• y ., .

4 . ., c N-*

I 1

-s t s 5

c, BWR K/,A: . 294001' K 1. 031,( 3.1/ 3. 8 ) -

I

,w ~

tt 1

(

, T 5 It '

p. &

', N .;

l I 's r _

\ '

1 iy

,g

,.:r ~['.

, ,-< ., \. '

jk ,, { t ,.',. [ ., . . . . , ,r, , , ,

. .a

c d 3 PBQQEDUBES_ _NOBd@61_8ENOBd662_EdEBGENQY_8NQ -PAGE 47 88D19600lG86_GQNIBg6

-ANSWERS -- BROWNS FERRY 1,.2&31 -86/12/01-BROCKMAN, K.

ANSWER 4.09 (1.50) 3

a. . Increase (0.25) Formerly extracted steam.now passes through

.ttue turbine (giving net work / watts .out) (0.5) .(0.75)

)

i

b. Decrease (0.25) Extraction steam to the next higher pressure-i heater will increase (0,5) (because the feedwater temperature rise across the. heater is greater - thus, less energy is available to q the turbine) (0.75 REFERENCE

.BFNP: OI-2,Section II.C.1 & .2, L.O.'s "A" & "B"; OPL171.011, Section E e

1 i

a

BWR K/A: 259001 A2.02 (3.1/3.3)(

1 ANSWER 4.10 (1.50)

a. Complete Loss of Offsite Power (LOSP)

b. Accident Signal present (LOCA on any unit) i ,

i c. The dedicated diesel generator (s) cannot supply sufficient c power for all of the required loads. (0.5 each) i 1 1 REFERENCE-

{ BFNP: DI-57, Secti on V. E. 4. j (5) ; L.O. "C"

?

i l b

i l BWR K/A: 264000 K5.05 (3.4/3.5), A2.01 (3.5/3.6) i a

t -

L - a

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

. s ,

14 2;_ PROCEDURES - NORMAL _8BNQRMOLi_EMERQENQY_8ND, 2 PAGE: 48 B8D196991C66_GQNIBQL i

' ANSWERS -- BROWNS. FERRY-1, 2&3 -86/12/01-BROCKMAN,'K.

ANSWER 4.11 (1.00)

a. 47 psig (+3, -2 psig)'

! b. . Possibility of overloading the 3C EDG (0.35) during a-simultaneous LOCA w/LOSP (0.15). (0.5 each) i .

REFERENCE

_BFNP: 'OI-24, Sections II.B & III.E, L.O. "A" ,

a 1

1 BWR K/A: 264000 K3.03 (4.1/4.2)

.86000 SWG #10 (3.2/3.5) i q ANSWER 4.12 (1.50) i (1) System fails-to operate.

l (2) System operates in a suspected adverse manner, t

I (3) System operates outside of the lisaits of the documented 4

acceptance criteria. (0.5 each)

REFERENCE l BFNP: Standard Practice 10.9, L.O. "B"; NOAM (Part 2), Section 4.9 J

i i

4 e

]

. TEST CROSS REFERENCE- PAGE '11 QUESTION! VALUE REFERENCE" 01.01 1.50' _KEBOOOO907 01.02 2.002 'KEBOOOO908-01.03 1.00 KEBOOOO909 "01.04, .50- KEBOOOO910 01.05 1.50 KEBOOOO911 01.06 2.00 ..K E B O O O O 9 1 4 0 1~. 0 7: 1.50 KEBOOOO915-01.08 1.75 KEBOOOO916 01.09 3.00 KEBOOOO919-

.01.10 1.00 KEBOOOO920, 01.11 1.00 KEBOOOO921 16.75 02.01 1.50 KEBOOOO922 02.02 1.00 KEBOOOO923 02.03 2.25 KEBOOOO924 02.04 1.00 KEBOOOO926 02.05 .50 KEBOOOO928 02.06 .50 KEBOOOO929 02.07 1.50 KEBOOOO930 02.08 1.50 KEBOOOO931 02.09 1.50 KEBOOOO933 02.10 1.00 KEBOOOO934 02.11 1.50 KEBOOOO935 02.12 1.00 KEBOOOO936 02.13 2.00. KEBOOOO939 16.75 03.01 1.00 KEBOOOO869 03.02 1.00 KEBOOOOB70 03.03 1.00 KEBOOOO871 03.04 1.00 KEBOOOO872 j 03.05 1.00 KEBOOOO973 03.06 1.50 KEBOOOOB74 03.07 1.00 KEBOOOO875 03.08 1.00 KEBOOOO876 03.09 1.50 KEBOOOOB77 03.10 1.50 KEBOOOOB78 03.11 1.00 KEBOOOO881 03.12 1.00 KEBOOOO884 03.13 2.00 KEBOOOO895 03.14 1.50 KEBOOOO925 17.00 04.01 1.00 KEBOOOO885 04.02 '1.00 KEBOOOO886 04.03 1.00 KEBOOOO888

Y s

TEST CROSS REFERENCE. PAGE: ~2 QUESTION VALUE REFERENCE

'04.04 ~ 1.50 KEDOOOOB901

.04'.05 1.00 KEBOOOOB94-04.06

~

2.50 KEBOOOO897 O4.07 2.00 KEBOOOOB98

'04.08 1.00 KEBOOOO903 04.09 1'.50 KEBOOOO904

- 04.10 1.50 'KEBOOOO905 04.11 1.00 KEBOOOO906

04.12' 1.50 KEBOOOO974 16.50

--g---

67.00 f.

1 i

i e

i i

a 1 2 1

I' 4

h i

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

U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _BBQWNS_FEBBY_11_2h}_____

REACTOR TYPE: _gWB-GE4_________________

DATE ADMINISTERED: _g6/12/91________________

EXAMINER: _gBggKMAN 3 K.

~~

CANDIDATE: ______ _h_ _1_h_ ~~ 2_

INSIBUCIIONS_IQ_CONDID6IE1 Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination.

Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff. Points fcr 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 four (4) hours after the examination starts.

% OF CATEGORY  % OF CANDIDATE'S CATEGORY

__YGLUE_ _IQI6L ___SGQBE___ _y6 lye __ ______________C8IEGQBY_____________

_1Z.20__ _20.sZ ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS

_16.Z5__ _2d163 ___________ _______._6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_1Z.59__ _25.Z3 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_16 59__ _23 26 ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS

_bSz99__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

• Candidate's Signature b

t

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration o'f this. examination the following rules apply:

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

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

3. Use black ink or dark pencil gely to facilitate legible. reproductions.

-4. Pri n t you.- name in the blank provided on the cover sheet of the examination.

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

6. Use only the paper provided for answers.

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

8. Consecutively number each answer sheet, write "End'of Category __"~'as appropriate, start each category on a gew page, write ggly gg gag sidg of the paper, and write "Last Page" on the last answer sheet.

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

10. Skip at least three lines between each answer.

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

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

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

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

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

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

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

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

a. Assemble your examination as follows:

(1) Exam questions on top.

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

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

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

.the' examination questions.

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

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

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

4

5. ' THEORY OF NUCLEAR POWER PLANT' OPERA' TION 2 _ELUIDS2iOND -PAGE .2 ISEBMODYN8MICS i:

I

QUESTION, 5.01' . ( 1. 50) '

Concerning the Bypass Flow in the reactor core:

a.--Which one of the following most accurately indicates co're bypass j .' flow.at 100%' thermal power and core flow? ' ( 1. 0)

(1) 1%

(2) 5%

(3) 11%

!~

f (4) 20%

b. STATE the most significant consequence that would occur if l

bypass flow were significantly reduced at full power. (0.5) i QUESTION 5.02 (2.00)

Concerning the Net Positive Suction Head (NPSH)

a. DEFINE NPSH . (0. 5) l b. For each of the following, indicate whether: the available NPSH I

at the suction of the recirculation pump would INCREASE / DECREASE /

j REMAIN THE SAME: (1.5) 1

! (1) The Feedwater Flow is INCREASED i

(2) The Recirculation Flow is INCREASED i

(3) The Vessel Pressure is INCREASED from 200 psig to 800 psig i

l i

i I

2 I

j I. (***** CATEGORY-05. CONTINUED ON.NEXT PAGE'*****)

$-( . k s

~

St__IBEQBY_QEJJUCLE88_EQWER ELANT_QPERAIIQN 2 _ELUlpS 2_ANQ. PAGE- 3 IBEBdQQYN8dlCS M

oc

, QUESTION 5.03 (1.50)-

Attached Figure # 911'shows a representative Xenon concentration curve. The reactor is. scrammed at Point A, resulting in-a peak Xenon concentration approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> later; at Point'B.

Line B - C represents the normal depletion of Xenon that would:

be expected.

'If the reactor were RESTARTED at Point B ; (dotted- li ne)', EXPLA'IN:

a. The rapid decrease in Xenon concentration from Points B to D. (0,5)

b. The change in the distribution of the radial neutron flux l during the period from Points D to D -(include reasons).- (0.5) i c. Any special precaution (s) that you would need to observe, as j a reactor operator, between Points B and D (include reasons) . (0.5) i

DUESTION 5.04 (1.50)

I j Unit 2 is in Cold Shutdown - The recirculation pumps are secured

for 4160 valt buswork - Shutdown Cooling is being throttled to

! maintain 150 - 160 deg F for a reactor restart. A review of ypur i control room instrumentation indicates the following. conditions:-

1 Bottom Head Temperature (RWCU) - 153 deg F l Reactor Pressure -

95 psig i Vessel Level - 33 inches ~

t a. DESCRIBE the indicated problem (s). (0.5) i

). b. STATE the corrective action (s) .which you should take. ' ( 0. 5) l c. If the pressure continues to rise - STATE what would be-your most immediate concern. (0.5) 4 i

l I l 1

~

i n

-1 4

'(*****

~

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

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

l

.5. ' THEORY OF NUCLEAR-POWER PLANT OPERATION 3 _FLUlQB1;AND- PAGE 4 j IHERMODYNAdlC@

d QUESTION 5.05. (2.00)

Attached. Figure'# 913 shows the variation in MAPLHGR'with respect to Exposure.

a. STATE the reasonLfor the INCREASE in the MAPLHGR limit.(Point A to Point B) early in core li f e.. (0.5)

b. STATE three (3) reasons for the significant DECREASE in the

~

MAPLHGR limit (Point C to Point D) at the end of core life. (1.5)

QUESTION 5.06 (2.00)

Concerning the effects of. Control Rods on reactor. power:

a. EXPLAIN how a control rod withdrawal could result.in a 'i reactor power decrease .

( reverse power effect). (1.0)

b. Which one of the following would be the~ rod movement. sequence most likely to cause the reverse power effect described above? (1.0)

(1) Deep Rod - 10 notch movement (2) Deep Rod - 1 or 2 notch movement (3) Shallow-Rod - 1C notch movement (4) Shallow Rod - 1 or 2 notch movement QUESTION 5.07 (1.75)

The reactor is at the beginning of a fuel cycle (BOC) and exactly critical. You withdraw control rods to add 0.06% ^k/k.

NOTE: ASSUME BETA-EFF = 0.0072 and LAMBDA-EFF = 0.1 SEC E-1

a. CALCULATE the resulting stable period. -(SHOW ALL WORK) (0.5)

b. CALCULATE how long it would take reactor power to DOUBLE.. (0.5)

c. If the reactor-were at the End of Cycle-(EOC) vice BOC, would the resultant stable period, for the same reactivity addition described in.(a), be LONGER or. SHORTER 7 JUSTIFY your response!- (0.75)

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

l Sz__IUEgBY OE_NUC6E6B_EgWEB_ELONI_gEEB6IlgN3 _E6UlpS 2_6ND PAGE 5. ,

1dhd0921000198 4

QUESTION 5.08 (2.00)

! The plant is at Full Power conditions, in 3-Element Control:

For an INADVERTENT HPCI INJECTION transient, STATE for each para-meter listed'below:

2

(1) . Whether the INITIAL value would INCREASE / DECREASE /

REMAIN THE SAME, and 1 (2) Whether the LONG TERM value would be LESS THAN/ GREATER THAN/

F THE SAME as the value before the transient.

]

. JUSTIFY each of your responses!

l a. Reactor Pressure 1

i b. Reactor Power

c. Total Feedwater Fl'ow

d. Vessel Level i

i

^

QUESTION 5.09 (3.00) 2 T Attached Figures # 5 A 8< B represent a transient that could occur at a BWR I GIVEN: (1) One Recirc Pump Flow Controller Fails High G Time t = 1.2 min

(2) No operator actions occur (3) Both Recirc Pumps were on line O Time t = 0.0 minutes

EXPLAIN the cause(s)'of the following recorder indications

a. Core Flow DECREASE (Point 3) (0.5)

b. Total Steam Flow DECREASE (Point 10) (0.5) i c. Reactor Pressure STABILIZATION (Point 12) (0.5) l l ( At a lower level)

~

d. Reactor Level DECREASE (Point 13) (0.5) j e. Reactor Power DECREASE (Point 16) ( 0. 5 ) -

j f. Since this transient occurred with power at less than rated (70%),  !

, STATE the factor (or program) which is utilized to ensure that the I l MCPR limit in not exceeded. (0.5) a l

i

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

Ag__ PLANT SYSTEMS _ DESIGN.1 CONTROL..AND INSTRUMFNTATTnN PAGE- 6-QUESTION- 6.01 (1.00)

The reactor-is at'100% Rated Thermal Power "A" Feedwater Level: Control is Selected All Controls are-in Normal / Automatic A failure in a Rx Vessel Level Narrow Range Instrument has occurred and resulted in the following related trips / indications:

"A" NR level indication reads at MINIMUM "B" 8< "C" NR level indications read.at MAXIMUM REACTOR WTR LEVEL A ABNORMAL Annunciator - ON PEAC VESSEL WATER LEVEL LOW-LOW CHAN A Annunciator.- OFF REAC VESSEL WATER LEVEL LOW-LOW CHAN B Annunciator - OFF Two channels of Level 8 have TRIPPED One channel of Level 3 has TRIPPED Feedwater flow is at ZERO State which NR level transmitter has failed AND state in which direction it has failed (HIGH/ LOW).

QUESTION 6.02 (1.50)

Concerning the APRM Circuitry:

a. STATE the (range of) time by which the flow-biased signal is time delayed. (0.5)

b. STATE what causes the circuitry to arrive at a particular time delay (within the range described above). (0.5)

c. The APRM flow-biased Scram trip is present when the MODE SWITCH contacts are (OPEN/ CLOSED /DOTH OPEN AND CLOSED) - CHOOSE ONE! (0.5)

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

\

'I

mosi_ "LONI_SYSIENS_DE219N _G9 2 NIB 9L2 0ND_INEISUMENIGIlgN- PAGE 7 i

QUESTIONL 6.03 (1.50)

Concerning the Fire Protection System:

1

a. . MATCH the Detection System (COLUMN A) with its Useage (COLUMN B) (0.75)

COLUMN A COLUMN B

1. . Rate of Rise Thermal Detector a. General area coverage throughout the plant

j. 2. Continuous Strip Detector .

b. Spreader Rooms and around any.

3 '. Ionization Smoke Detector oil reservoirs

c. Cable Tray protection f b. Given that a fire detection system provides an' initiation i function - STATE how the Fire Protection / Raw Service Water system

responds / aligns. (0.75) i l

QUESTION 6.04 (1.50)

]

Concerning Refueling Operations:

i 1

a. Attached Figure # B79 shows three 1/M plots for a hypothetical

reactor. SELECT the curve (#1 or #2) which is indicative of the l detector being TOO CLOSE to the source. (0.5)

! b. STATE the interlocks which will prevent movement of the Refueling Bridge over the core. (1.0) 4

)

1 4

! t i

i 4

i i

4 i

5 I

i

(***** CATEGORY 06 CONTINUED CR4 NEXT PAGE *****)

4 a

e

6. P66NI_SY@IEMS_QESIGN1_QQNIBQL1_6NQ_IN@IBUMENIGIlON PAGE: -G  ?

I i

a i

QUESTION 6.05 (1.50)

Concerning Refueling Equipment:

a. The ref ueling grapple is considered to be " NORMAL.UP" at a level of approximately 8 feet below the top of the refueling plat-

-form rails. STATE the reason-for this movement interlock. (0.5) l 1 b. STATE what provides the motive force for operating each of the following: (0.5)

]

(1) Channel Bolt Wrench (remove / install channel fastener assembly)

(2) Refueling Grapple (grab / release fuel elements for movement)

] c. STATE the indication (s) which you would use.to determine that an

. excessive pull was being exerted on the channel (during Fuel Assembly l Dechanneling). (0.5)

DO NOT ADDRESS PHYSICAL DEFORMATION, ETC., OF THE CHANNEL 1

(

l l

1 e

l t

1 i

I t

I i

J l

'rI l (***** CATEGORY 06 CONTINUED ON NEXT PAGE'*****) i l

_ _ . - _ _ . . . _ _ . _ . _ _ _ . - . . _ _ _ . . _ . . _ _ _ _ _ . _ . _ - - _ .. . _ . -_ ._ _.._.._.a

I 1.

6t__EL8MI_QYQIEOQ_DEylOU2_CONIBQL2_8MD_lNGIBUDENIBIICU PAOC 'f f

i QUESTION 6.06 (2.00)

The following plant conditions exist:

I .

Jet Pumps 1 - 10 Differential . Pressure (Meter) =- 15 psid Jet Pumps 11 --20 Differential Pressure (Meter) = 3.psid

RECIRC LOOP A ONLY OUT OF SERVICE (Annunciator) = ON

l. a. The TOTAL CORE FLOW Recorder _would calculate core ~ flow.by which

, of the following methods? (1.0) 1

)

1 (1) Loop B Jet Pump Flow + Loop A Jet Pump Flow l . .

(2) Loop B Jet Pump Flow - Loop A Jet Pump Flow

(3) Loop B-Jet Pump Flow Only i >

(4) Loop A Jet Pump Flow Only

]

4 t b. After 5 minutes, the operator opens the Discharge Valve of Recirc

} Pump A to maintain the A Loop temperature. The TOTAL CORE FLOW i Recorder would calculate core flow by which one of the following methods? (1.0) '

l (1) Loop D Jet Pump Flow + Loop A Jet Pump Flow 4

i (2) Loop B Jet Pump Flow - Loop A Jet Pump Flow i

(3) Loop B Jet Pump Flow Only i (4) Loop A Jet Pump Flow Only i

1 ,

I i

i OUESTION 6.07 (1.00) i 1 While venting Primary Containment through the Standby Gas Treatment ,

System (SBGTS), you receive a High Stack Gas Radiation Alarm, requiring l l you to discontinue venting. Which one of the following values is the j value at which you would expect this to occur 7 t

a. 3.8 E+1 cps i

i b. 3.8 E+2 cps r

l

c. 3.8 E+3 cps i

d. 3.8 E+4 cps i

i

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

a i I

lb Pt9EI_SiSIENS_DESION2_G9dIB9L 2 _8SD_INSIBudgNIcIIgg- PAee 10'

.i QUESTION

~

6.08' ( .50)-

EXPLAIN why it is essential to ensure that the~DieselnGenerator governor is-positioned fully to the LOW SPEED STOP, . prior.to-

. performing-a. local start. .

QUESTION 6.09 (1.50)

Unit 2 is at full power when a Load Reject occurs due to a trip of the MTG Output Dreakers.

a. STATE the specific signals.(and their sources) which activate the-Load Reject circuitry. '(0.5)

b. DESCRIBE the response of the Turbine / EHC System to'this event. (1.0)

DUESTION 6.10 (2.25)

An automatic LPCI injection signal is received:

a. STATE the three specific conditions (setpoints, 1f applicable) that would be required to allow you to= initiate Containment Spray. (1.50)

b. STATE whether each of these conditions can be bypassed. (0.75)

QUESTION 6.11 (1.50)

A Core Spray initiation signal is present on Unit 2. You MANUALLY secure the "C" Core Spray pump due to increasing reactor vessel level,

a. STATE the indication (s) that would be provided'on the control-panel that the pump was manually secured. (DO NOT address flow, pressure or other metered indications) (0.5)

b. LIST the conditions that would have to exist and/or the actions that you would have to take to permit the pump to automatically restart, if needed. ( 1. 0 )-

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

d t

i

ss_ n 00I aXeIEUa_DESIOU2_COUIB9L 2 _809_IngIBudgNIsIION- Pact :11. ,

1 1

i

' QUESTION 6.12 (1.00) i-STATE whether ADS would DEPRESSURIZE or NOT DEPRESSURIZE the reactor f 'for each'of the following-conditions:

CONSIDER THE TOTAL ADS RESPONSE AND ASSUME.NO OTHER OPERATOR ACTIONS; OCCUR' I

j a. After initiation of th'e 120 second timer, but.-before it is timed j- .out, Drywell Pressure DECREASES to 1.5 psig. (0.5)

b. After initiation of the 120 second timer, but before it is timed j out, Reactor Water Level INCREASES _to inches. (0.5) 4 i

4

}

i a

1 k

i I

ll i

'i.

I i.

a i

l l

1 i

1.

m

?

i i

1.

1 1

i

. (***** END OF CATEGORT 06 *****) )

l

, ., ,,,...,r,.-_,.,,. , . _ , , , _ . , . , , , , - _ . _ . . , , , , , _ - . . ._r--, - - _ , , , , . _ , _ . . , , , - , , , , _ - - , . - , . , , - , . - . + , , , _ _ . ~ . _ . -

Za PRQCEDtjRFS_ _tjQRMAL 2;AQNQRMAL _x EMERGENCY AND PAGE 12 88D1QLQGlC@L_CQNIEQL QUESTION 7.01 (1.00)

EDI-2, " Containment Control", states in Step PC/P-5:

IF drywell pressure exceeds 55 psig-(70_ psia), THEN

... vent the primary containment ... to reduce and maintain pressure'below 55 psig as follows:

~

1. 1F suppression pool water' level is below 20 feet, THEN vent the suppression chamber.

EXPLAIN why (if conditioins allow) the suppression chamber is vented 7

first (to reduce drywell pressure). Include the benefit acheived-l and how this is accomplished.

i i OUESTION 7.02 (1.50) j Attached Figure "E" shows the Heat Capacity Level Limit (HCLL):

4 a. EXPLAIN the purpose for having an HCLL (i.e., what does the curve j protect against?) (1.0) l b. STATE the immediate action (s) required if plant conditions are out of the safe area. (0.5) l i

l

! OUESTION 7.03 (1.00) f'

Ccacerning recovering lost or unsecured articles introduced into

the primary system i t

j a. STATE when IMMEDIATE recovery is allowed. (0.5)

b. TRUE OR FALSE i

j_ If immediate recovery of an article has been performed, a Form

< DF-8 is not required; however, the Unit Superintendent must be notified. (0.5) 1 i

i f

I i

1 i

j (***** CATEGORY 07-CONTINUED ON NEXT PAGE *****)J

Zs__BBQQEQUBES_ _NQ8d861_8EdQBd862_EME8@ENQY_8NQ- PAGE 13 BG919690lG86_GQUIB96 OUESTION 7.04 (1.00)

Concerning plant operations:

a. STATE the. thermal power level which must not be exceeded under.

any conditions. (0.5)

b. STATE the maximum allowed average thermal pr.wer for an 8-hour.

shift. -(0.5)

OUESTION 7.05 (1.00)

Concerning a Plant Startup:

a. Step A.12 of GOI 100-1 states:

"RHR system in standby readiness condition in accordance with 0I-74" TRUE OR FALSE Step A.12 may be signed off with RHR pumps A & C operating to provide Shutdown Cooling. (0.5)

b. Step A.22 states:

" Reactor recirculation system ready for service, in accordance with OI-68" TRUE OR FALSE Step A.22 may be signed off with Recirculation Pump A Dut of Ser-vice for Shutdown Cooling requirements (only). (0.5)

OUESTION 7.06 (2.50)

LIST the five (5) entry conditions to EDI-1, "Rx' Control" (Include Setpoints, as appropriate). (1.0)

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

b g

-2 __E89EE908EE_I_d98dOb2_6EU9000bi_COGGQENGY_6UD 'FAGE 14

-BSDI96991GBb_GgNIBg6 QUESTION 7.07 (2.50) i

< Concerning_ Refueling Operations: .

j 1- a. Excepting notifications, LIST ALL of the actions to'be taken to shutdown-the reactor in the event of en Inadvertent? Criticality during Fuel Loading. (1.5)

b. Which one of the f ollowing is the lowest level person authorized to initiate Standby Liquid Control during an Inadvertent Criticality during Fuel Loading? (1.0) l (1) Any. Operator (License is not required) k (2) Unit Operator (RO License) i i (3) Assistant Shift Engineer ( RO License) i l (4) Assistant Shift Engineer (SRO License) l 1

OUESTION 7.08 (1.00) l l

,i j During Refueling operations, when the Moisture Separator-is trans-  ;

ferred fran the Equipment Storage Pool to the Reactor Well, the upper l j sections may temporarily break the surface of the water - a Refueling j Zone Isolation may occur.

i i Assuming that the (i sol at i on ) initiation signal clears, which one of j the following most accurately describes when the isolation may be

! cleared and the applicable systems returned to operation in accordance

! with OI-30?

i 1 a. Immediately i.

i b. Once HP verifies that the source of the initiation was the separator, via survey.

} c. When the separator is again submerged.

, i j d. Not until all transfer operations are complete and an HP j survey has been performed. l l

i 1

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

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

'F-

- 7. 7 PROCEDURES'- NORMAL 2_ADNORMAL _ 2 EMERGENCY _8ND' PAGE ~15 68DI969910BL_C9NIBg6

,. ' QUESTION. 7.09 (1.00) j.. Concerning the Emergency Plan:

a. 'Per IP-20, " Technical Support Center", STATE-the ALTERNATE location of the TSC. (0.5) f b. Per IP-15, " Emergency Exposure", STATE the allowable' emergency

! exposure'for each of.the following: (0.5)

(1) Life Saving Actions i.

l (2) Less Urgent Emergencies (f acility protection) 3 1

I QUESTION 7.10 (1.00) 1 5-j Concerning the Routine Radiation Work Permit (RRWP):

I i FILL IN THE BLANK FOR EACH OF THE FOLLOWING i

4 i a. No one is permitted to receive a dose greater than ___ _ per j day on any RRWP or any combinatica of RRWP's.

e

b. Personmnel with an exposure limit of ________ per quarter or j- less will not be allowed to enter on an RRWP.

{

c. No RRWP user will exceed ______ of their remaining allowable dose.

(RAD) in one day as indicated on a current personnel listing, except by special permission of the Health Physics section supervisor or an

assistant Health Physics section supervisor.

i i d. Entries into dose rates of ________ or greater will require the j approval of Health Physics supervision (see RCI No.'1).and the pre-

sence of'a Health Physics technician. Additional HP surveys may be l'

required prior to entry. These requirements will be noted on the RRWP.

1 ,

i l t

)

I J,

i l'

i l'

1

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

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

j 7. PROCEDURES - NORMAL x_6BNQBM662_EME8GENGY_6ND'. PAGE 16 88D196QGlG66_GQNI896

- QUESTION .7.11 (1.50)

~

i. Concerning operations with Feedwater Heaters removed from service, i i :as per DI-2, Condensate System:

a. STATE whether kilowatt output will INCREASE / DECREASE / REMAIN _THE l' SAMELwhen the highest pressure feedwater.~ heater is removed from ser- ,

) -vice. JUSTIFY you answer. (0.75) i .

i b. STATE whether kilowatt output will INCREASE / DECREASE / REMAIN _THE  !

! SAME when other than the highest prousure feedwater hecter is removed from service. JUSTIFY your answer. (0.75)

I l QUESTION 7.12 (1.50) i 1

STATE the three (3) conditions which must be present to allow the j Unit _1 and 2 Diesel Generators to be paralleled with'the Unit 3 Diesel Generator.

QUESTION 7.13 (1.00)

Per 01-24, " Raw Cooling Water"

1

a. STATE the header pressure at which the STDY (spare) RCW pump will automatically start. (0.5) j b. 01-24 cautions the operator NOT to run the 3D RCW pump in i automatic. STATE the reason for this caution. (0.5) i i

)

i \'

i l

i I

i i

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

I n Bu__8DdiU1EIB8IIVE_669CEDQBESx_QQNDlllQUS2_QUD_LldlI8IlQNS- PAGE .1'7 1

j ..

N ,

+

-QUESTION 8.01 (1.00)

Special concern is required for returning MOV's, that have'been MANUALLY CLOSED, to service.

1 a

a. DECRIBE how to verify the operability of an MOV in a safety system that has been-manually closed. (0.5)

'b. DESCRIBE the malfunction that this operability check 'is.

~

3 intended to identify. (0.5)

J i

j QUESTION 8.02 (1.50)

STATE the criteria that are used to determine if an MR should be checked EMERGENCY, IMMEDIATE ATTENTION, or ROUTINE MAINTENANCE.

(i.e., Define each classification)

)

i l QUESTION 8.03 (1.00' j With the reactor in STARTUP and at 10% RTP, the following Drywell to Suppression Pool Vacuum Dreaker monitoring light indications

) are noted for one of the Vacuum Breakers:

l Check Light -

ON 2

I Green Light -

ON i

l Red Light -

OFF STATE whether the Technical Specifications for Primary Containment

j. are sati s fi ed (Not exceeding LCO). JUSTIFY your decision!

i 5 i i

4 s

)

i l

1 i

j (***** CATEGORY 00 CONTINUED ON NEXT'PAGE *****)

9z__8DUINI@IBGIlyg_E8gCEDUBE@2_CQNQlIJQNS 3_GND_LidlIGIIQUg. PAGE 18 i

t OUESTION 8.04 (2.00)

Per Browns Ferry Standard Practice 12.8, following a reactor-scram and post trip review:

a. STATE two (2)-criteria that must be met before you, as the Shift j_ Engineer, could recommend a unit restart. (1.0) b.. STATE who (Job title) is responsible.for authorizing a restart (criticality). (0.5)

! c. LIST all additional actions, if any, which would be required j prior to restart if you, as the Shift Engineer,-and the STA disagreed

on the restart recommendation (i . e. , STA recommends startup delay).

(0.5) 1 f

! OUESTION 8.05 (1.00) ,

Units 1 & 3 are in Cold Shutdown; Unit 2 is in Startup, but STILL

, SUBCRITICAL. The ASE requests your presence in the reactor building to help resolve a 2ignificant problem with a CRD pump. As the Shift ,

j Engineer and the only SRO in the Unit 1/ Unit 2 Control Room, which one

{ of the following most accurately reflects your allowable response?

! a. You may respond to the request, providing there are j two licensed RO's in the Control Room and no additional

rods are pulled in your absence.

4 j b. You may respond to the request, providing there are f two licensed RO's in the control Room and ttue STA l replaces you.

1 i c. You may respond to the request, providing there are 3

two licensed RO's and an STA in the Control Room, and you remain in continuous radio contact and within 10 j minutes of the Control Room.

d. You may not respond unless relieved by another SRO.

}

! OUESTION G.06 (1.50) j An Operability Test is conducted on a safety-related system, following j the installation of an approved modification. LIST three (3) sep-

! arate criteria which would procedurally require that a TEST DEFICIENCY

be initiated (documented).

4 l

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

s L

j e. JeoMINISIB8IIVE_BBgCEgUBEg2_CgNDIllgNS _6ND_ 2 LIM 11BIlgNS~ PAGE :19 a

4 I

) QUESTION 8.07 (1.00) i t With the reactor at Full Power, an inadvertent IM error;results i in the closure of ALL FOUR MSIV's. Your past trip review' indicates

! that the Wide Range Vessel Pressure Recorder spiked at 1287 psig.

1

! STATE the unique actions, if any, that would be required prior.to i restart in this case.

l l

QUESTION 8.08 (2.50)

• \

l Following a power transient and subsequent reactor Scram, a mechanic j is reported to be severely injured and unconscious due to a ruptured-j RWCU line. HP technicians report that the readings.in the area are 4 as high as 225 Rem / hour.

l a. STATE three (3) criteria that you would utilize to select the Rescue Team personne). (0.75)

i. '

j b. CALCULATE the maximum Stay-Time that you.could permit to effect

! the rescue. (0.50) l c. STATE the dos e to the EXTREMITIES that you could permit to ef f ect i the rescue. (0. 25)-

l j d. Utilizing the enclosed classification table, Figure #976, STATE -

{ the applicable Emergency Classification and LIST the MOST RESTRICTIVE j NRC reporting requirements. (1.00) 1 1

i DUESTION 8.09 (1.50) i.

.I Units 1 and 2 are in refueling outages; Unit 3 is in an extended ,

I maintenance outage. While operating the Unit 1 and 2 standby ,

j diesel generator (EK) D, a mechanical overspeed trip occurs.

. Diesel generator A is also out of service at the time, due to i extended maintenance.

]

j STATE whether the Technical Specification limiting conditions f or

operations are met. If not, SPECIFY which TS's is/are not complied

with (TS Item number not required, but acceptable).' In either case, JUSTIFY your response!

4 i

l i l

i 1

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

y :e ._ ,(

,3 E .

.ll . ( ..

- : 82 _69dINI@IBBIlyE_BBggEQUBEg2_CQNDIllgNS 2 _8ND_(IMlIBIIgN@: PAGE 20 s

4 i r

QUESTION' 6.10 ( 1 ~. 50 ) -

l

-Unit 3 is in-an extended maintenance outage. RHR Pump 3D is running for the purpose of maintaining reactor coolant temperature above.-100

,. 'deg F. The following equipment, status is. reported:

3 RHR Pump 3B - Out 'of Service for. maintenance.

RHRSW Pump-D1 -'Out of~ Service for. maintenance

'RHRSW' PUMP D2 D/G - Preventive. maintenance in progress

, ~

Locp I of RHR is now declared INOPERABLE due to a damaged > pipe hanger on the-24".line-above the Torus. (S/D Cooling-and LPCI-Injection)-

l STATE whether the Technical Specification limiting conditions "

for,-. operation are met. If not, SPECIFY which TS's is/are'not'l 9 complied with. (TS-Item. number not required, but acceptable)- In ! ,.

either case, JUSTIFY your response! [ J(i.5)

)s i

i .

i I

i i .(

5 i

r

?

h, \

l i

.j a k N i

J i

1 "ji

}.

s R

1 s a ,

t-s

) /

. 1

! \

1 s

i i

s 4 f

, (***** CATEGORY.08 CONTINUED ON.NEXT PAGE *****)~ ,. -

1

az__0DUINISIBOIlyE_EBgCEDUBEg3_CgNQIIIONS 2 _8NQ_LIMII8IIgNg PAGE 21 QUESTION 8.11 (1.00)

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 statements most accurately describes the allowances / limitations as detailed in the Browns Ferry Technical Specifications and OSIL #35, " Secondary Containment Control."

a. Secondary Containment integrity is not required under the existing 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, sta-tioning a public safety officer allows both the inner and outer air lock doors to be open during equipment 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 doors at all times they are open.

NOTE: APPLICABLE TS'S ARE EF GSED FOR REFERENCE

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

\i es__8Dd191EIBGI12E_EBQCEDMBE@z_CQUD1IlQN@i_8ND_Lidll8119NS PAGE 22 OUESTION 8.12 (1.00)~

Units i and 3 are operating at 75% and 100% rated' thermal power, respec- ,

.tively. Unit 2 is in Cold 3hutdown. During performance of SI's (4.7.B) on SBGTS Trains A, B, and C, the following deficiencies are noted:

SDGTS Blower / Fan A is INGP f i

SBGTS Train C Decay Heat Damper (FCO 65-52)'is STUCK OPEN  !

Maintenance estimates 3 to 4 days to accomplish repairs.

Which one of the following statements most accurately describes the allowances and/or limitations imposed by the Drowns Ferry Technical Specifications.

NOTE: APPLICABLE TS'S ARE 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 Shutdcwn 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 Unita 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

i i

t o

-h

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

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

i Uz__IMEQBY_QE_NyCLE68_EQWE8_E66NI_QEEB611QN 2 _E6y1QSz_6NQ PAGE 23 IHEBdQDYN8dlCS ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/D1-BRDCKMAN, K.

]

-ANSWER 5.01 (1.50)

.- a . c (1.0)

b. (Excessive voiding in the bypass region resulting in) unreliable (0. 5) J

~

LPRM readings (0.5). (1/2 credit for LPRM overheating)

REFERENCE General Electric Heat Transfer & Fluid Flow, Chapter'8

BFNP

HTFF, Chapter 8, L.O.'s 9.4 & 9.5

)

e i

i I

BWR K/A: 293008 K1.32 (2.5/2.6), K1.33 (2.4/2.6) q 215005 K5.02 (2.7/2.8) i l ANSWER 5.02 (2.00) i a. P -P (0.5) j act sat a

b. (1) INCREASE (More Subcooling at the pump suction)

(2) DECREASE (Reduced pressure at the eye of the pump results in being closer to saturatio pressure)

(3) INCREASE (Further to saturation temperature and increased density causing less static head)- (3 8 0.5 each)

REFERENCE General Electric Heat Transfer and Fluid Flow, Chapter 6;'

l BFNP: HTFF, Chapter 6, L.O.'s 7.7, 8.1, & 8.2 7

4 BWR K/A: 293006 K1.10 (2.7/2.8) 4

- _- . . . - _. _ _ . . . - . - _;, . L _.-----,, , m. __.;-.---. m...,, , , . - . _ , . , - , , , - , , , ,

i

5 t__IBEQBY_QE_NQCLE68_EQWEB_EL6UI_QEEB811QN3 _ELylD$2_6NQ PAGE 24 ISEBdQQYU6dlGS

, . ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 5.03 ( 1. 50)

a. .The decrease is due to the (decay of Xe-135 and the) ' increased burn-out (which is'immediately seen by the EXe]), being much greater than the -

Xe production, (which is time-delayed .f rom the lower power. )

b. The flux will redistribute from outside (and higher) peaked and center (and lower) depressed to center peaked.and outside depressed.

(0.25) This is because the Xe concentration shifts from center peaked to more peripherally peaked due to power production distribution. (0.25)

c. Use caution when pulling peripheral rods (0.25) due to the ,

incrcased rod worths (increased thermal flux) (0.25). (0.5 each)

REFERENCE General Electric Reactor Theory, Chapter 6 BFNP: Rx Theory, Chapter 6, L.O. 2 BWR K/A: 292006 K1.07 (3.2/3.2), K1.08 (2.8/3.0), K1.10 (2.9/2.9)

L K1.14 (3.1/3.3) i

}. ANSWER 5.04 (1.50) ,

i'

a. Thermal Stratification (187 deg F Delta T)

b. Establish Natural Circulation (Raise level to 3rd stage of the separators *60")

c. Loss of Shutdown Cooling (due to pressure interlocks)

(- OR - Increase SDC to reverse stratification @ 0.25) (0.5 each) l l REFERENCE

General Electric Heat Transfer and Fluid Flow, Chapter 8 BFNP: HTFF, Chapter 8,-L.O. 10; GOI 100-12,Section IV l

l 4

, + --n-,- - , ~ .,e, ,-,--n-,,,-e ~-- -.e- + - - r + - - + - - - *n+-' s v ,

s 10t__IMEQBY_QE_UyCLE68_EQWEB_EL8NI_QEE86I1QUt_ELylDS 1_8NQ~ -PAGE 25

~IBEBdQDXU8dlGS LANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

1

~

'DWR K/A: 293008 K1.35 - 1.37 (3.1/3.3), K1.22 & 1.23 (2.9/2.8:0

) '

f .

ANSWER 5.05 (2.00)

.a._ Pellet to Clad contact (due to the fuel pellet expansion) (0.5)

b. (1) Fission Product gasses increase fuel rod gas' pressure and the stress on clad during dryout, j (2) Fission Product gasses-have a lower heat transfer coefficient

than Helium,.thus reducing the heat transfer rate.

(3) Crud buildup on the fuel clad surface with core age reduces

the heat transfer and increases the stored energy.

t j (4)- Fission product decay energy increases'with exposure and i

can increase the Zircaloy-water reaction rate near 2200 deg F.

l (3 0 9.5 each) i 4

REFERENCE

, General Electric Heat Transfer and Fluid Flow, Chapter 9 DFNP: HTFF, Chapter 9, L.O. 3.6 4

L 4

f BWR K/A: 293009 K1.14 (2.6/2.7)

E 1

i i

4 I

1 i

l l

l l

l

+ , . - , -

h a

4

5. THEORY OF NUCLEA8_BQWEB;ELONT_QEER@TigN2_ELUID@2_AND PAGE' 26

-IHE8dQDYN8dlC@

~

ANSWERS ---BROWNS FERRY 1', 2&3 -86/12/01-BROCKMAN, K.

' ANSWER 5.06- (2.00)

a. The-negative reactivity added'by the. increased voids generated by the rod withdrawal is greater ~than the positive reactivity added by the reduced rod absorbtion. (1.0)-

b. (4) (1.0)

REFERENCE General Electric Reactor Theory, Chapter 5 BFNP: Rx Theory, Chapter 5, L.O. 3 BWR K/A: 292005 K1.04 (3.8/3.8)

ANSWER 5.07 (1.75)

a. Period = (Beta - Rho) / (Lambda

• Rho)

= (0.0072 - 0.0006) / (0.1

• O.0006) = 110 seconds ~ (0.5)

b. Period = 1.443

• DT -> DT = Period / 1.443 = 110 / 1.443

= 76.23 seconds (Error carries forward from (a)) 1(0.5) ,

c. Period will be SHORTER (0.25) This is due to the fact that the Beta for EOC is shorter (* O.0058, due to the1 increased Pu _

contribution). (0.5) (0. 75).

. REFERENCE

. General Electric Reactor Theory, Chapter 3~

-BFNP: Rx Theory, Chapter 3, L.O.'s 3.6, 4.6, & 5.9 i

l .BWR-K/A: 292003 K1.08-(2.5/2.8), K1.09 ( 2. 5/2'. 6) -

i l

i l -

c _

1

Sz__IBEgBy_QE_ NUCLE 68_EQWg8_E66NI_QEgBSIJgN2 _E691gg2_6NQ .PAGE -27 ISEBUggyN6DIgg ANSWERS -- BROWNS FERRY 1, 2&3 -06/12/01-BROCKMAN, K.

' ANSWER 5.09 (2.00)

a. Reactor. Pressure decreases initially (0.125) 'due to HPCI steam-flow (0.125). Long term pressure is higher (0.125) due to the higher final reactor power (0.125) 4 3- b. Reactor Power decreases-initially (0.125) due to the additional voiding from the pressure decrease (0.125). Long term power.is-higher due to the colder water which enters the core (0.125).

c. _Feedwater. flow decreases initially ( 0. ,125 ) as FWLCS follows '

ttua decrease in steam flow (and increase in level) (0.125). Long term feedwater flow.is_less (0.125) to compensate for-the input of the HPCI injection (0.125).

1

] d. Vessel level increases initially (0.125)'due to the swell caused ,

by the pressure decrease (0.125). Long term level stabilizes at a I higher value to compensate for the ' steam flow / feed flow mismatch (in

. the FWLCS circuitry) .(0.125). (0.5 each)

-REFERENCF BFNP

Transient Analysis L/P, Transient # 20; OPL171.055, L.O. "A" BWR K/A: 206000- K2.07 (3.9/4.3) 295014 AA1.07 (4.0/4.1) s 1

b.

. _. 1.1,. 2 ~ . - _ - _

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

5. THEORY OF NUCLEAR POWER PLANT OPERATION 1 _ELylDS 2 _6HD .PAGE 28 IHEBUODYU8MICS ANSWERS --l DROWNS FERRY ~1, 2&3 -86/.12/01-BROCKMAN,,K.

't ANSWERL ,5.09 - (3.00)' ,

a. Recirculation. Pump Runoack (From < 20% FW Flow) ( O'. 5 )

E b.

~

EHC controlling r'eactor pressure (at 920 psig) (0.5)

c. EHC1contro11ing reactor pressure (at-920 psig) ( 0. 5).

d. ' Increase in'the suction of the Recirc Pump (from Downcomer) (0.5)

e. Reactor Scram (APRM Flow Biased'@ 0.66*w + 54%) (0.5)

f. K-f is utilized to increase the steady-state MCPR limits at reduced flows (FLCPR Program). E (0. 5) '

REFERENCE '

BFNP: Transient Analysis LP, Transient # 5, L.O. A&B BWR K/A: 20T001 A2.05 (4.0/4.0) 293009 K1.26 (3.3/3.3) K1.28 (3.5/3.5) i 4

J J~.

- 5___ELONI_gYgIgMg_ DESIGN3 _CQNIBQ62_8ND_INSIBQMENIGIION PAGE. 29

' ANSWERS ~-- BROWNS FERRY.1,-2&3

-86/12/01-BROCKMAN, _K.

ANSWER 6.01 (1.00)

(1) A (LIT 3-53).

(2)- Low -(O'.5 each)

REFERENCE GGNS: OP-D21-501 '

BFNP: OPL171.OO3,-L.O. "J";.OPL171.012, L.O. "E.3" ARP 9 XA-55-5A-8. XA-55-5A-30, XA-55-5B-4, XA-55-5B-5 BWR K/A: 259002 K5.01 (3.1/3.2) K6.5 (3.5/3.5)

ANSWER 6.02 (1.50)

a. O seconds (+3,-0) to 30 seconds (+6,-6) (0.5)

b. The rate of the power increase determines the time delay (fast rise = short delay). -OR- The absolute value of the power increase (P(f) - P (i ) ) (0,5)

c. BOTH OPEN AND CLCSED (0,5)

REFERENCE BFNP: OPL171.022,Section X.B.6 & .7; L.O. F&K

-BWR K/A: 215005 K4.02 (3.7/3.7)

-)

1

1 4

' bs__E GNI_SYSIgd@_QEQ1GN1_CQUIBQ61_8NQ_INSIBydENIATION PAGE 30

'A'NSWERS --' BROWNS-FERRY 1', 2&3 -86/12/01-BRCCKMAN, K.

i' i

6

. ANSWER 6.03 (1.50)

a. (1) b 1

(2) c.

(3) a (0.25 each)

.b. Fire pump (s) auto start Raw Water Storage Tanks Isolate i

Raw Service Water Pumps Shutdown (by interlock). .

'(0 25 :each) t I

REFERENCE i BFNP: OPL171.049, Sections X. B.r. (2) & .(3) & .D.2; L.O.-E l

1 A

l BWR K/A: 286000 K5.06 (2.6/2.7), K5.07 (2.6/2.7), K4. 01; (3. 4/3. 6);

K4.02 (3.3/3.5) 4 -

4 C

'l L

h h

J

+

I ,

s

. i 4

s

,- ~y,-r,'v-- --e e - -w -

3 -'-. . -r - + c- , ,-+ e, irww r * --w ,-.v- w -v< r ~w-

s I 6t__ELONILSYSIEMS_DESIGU2_GQNIBOL 1 _6ND_lNSIByMENIGIlgN PAGE 31!

ANSWERS -- BROWNS FERRY 1,-2&3 -86/12/01-BROCKMAN, K.

ANSWER 6.04 (1.50) l

a. Curve #1 ( 0. 5 ) '

b. (1). Any platform hoish (3) loaded or the grapple not fully up and all rods not full in-with the platform near or over the core. (0.5)

.(2) Platform near or over the core and not in the refueling mode ( 0. 25)'

(3) More than one rod withdrawn with the platform near.or over the core in the refueling mode. (0.25)

REFERENCE BFNP: OPL171.053,_ Sections X.B.1.m.(4) 8e .C; L.O.'s D & E ,

BWR K/A: 234000 K4.xx ANSWER 6.05 (1.50)

a. Preclude excessive radiation levels (ALARA) (0.25'each) i

b. (1) Manual (2) Air (0.25 each)

c. By the RED color area appearing on the tube of the Channel Handling Tool. (0.5)

REFERENCE Stearns/ Rogers Refueling Platform Technical Manual BFNP: GOI 100-3 -

i BWR K/A: 23400, K

6;__eLeNI_gygIENS_ DESIGN 1_QQNIBQL2_@ND_lNSIByMENIBIlQN PAGE 32 ANSWERS - BROWNS FERRY'1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 6.06 (2.00)

a. (2)

.b. (2)- (1.0 each)

REFERENCE-

-BFNP:'ARP 9-4, XA-55-4A; OPL171.OO7 BWR K/A: 202001 K1.01 (3.6/3.7) 293008 K1.26 (2.9/3.1), K1.27 (3.6/3.7)

ANSWER 6.07 (1.00)

C REFERENCE BFNP: OPL171.018; OI-65, p 9; Technical Specifications,' L.O. 62 BWR K/A: 261000 K3.02 (3.6/3.9), A4.01 (3.2/4.0) 272000 K3.02 (3.1/3.8) 1

'l 1

4

- 0z__ELGUI_SYSIEdS_QEQ1GN1 _CQNIBOL,_ANQ_1NSIBQMENI611QN - PAGE 33

[ ANSWERS --~ BROWNS FERRY-1., 2&3 -86/12/01-BROCKMAN, K.

i-

. i ANSWER 6.08 ( . 50)

Not having the governor at the lowest setting can result in a mechanical overspeed trip of the diesel.

REFERENCE l BFNP: LER 86-26 (7/28/86 - EDG "D")

A t

BWR K/A. PWG (3.2/3.6)

ANSWER 6.09 (1.50) .!

.i

a. (Greater than 40% mismatch between the generator output and the turbine power) as sensed by generator stator amps.and turbine. cross-

, over pressure. .0.5).

( ,

! b. (1) TCV's Trip Closed (f ast-acting solenoids) f (2) Bypass Valves Open (3) EHC Load Selector runs back (4) Control Valves partially reopen G < 40% mismatch' (5) CIV's throttle to control turbine speed G 1800 rpm (0.20 each)

REFERENCE i BFNP: OPL171.010, L.O.'s "J" & "K" t

i s

=BWR K/A: 245000 K3.01 (3. 4/3. 7) f'

'ei r w y-=-v3 - ~- w ,V -

-*-7vt M e s-- - n- -

4-e+' r --r - +---r-'m e -*e' --

• 4- - * - - * "

i 6,__E68MI_SYSIEd@_ DESIGN1 _QQNIBQ62_8ND_IN@IBQUENI8IlgN! PAGE E34 !

s' ANSWERS.-- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN,.K.

l 1

ANSWER 6.10 (2.25)

I- -a. Reactor Water Level (O.25) GTE - 39" (O.25) i Drywell Pressure (0.25) GTE + 1.96 psig (0.25) i SELECTT- RESET Switch (0.25) .in SELECT _(0.25)

b. Level can be bypassed

{ Drywell Pressure can not be bypassed 4

Select Switch can not be bypessed (0.25 each)

REFERENCE BFNP: OPL171.044, L. O. 's "E.10" & "E.11" i

i 3

BWR K/A: 226001 K4.03 (2.9/3.1)

. 203000 K4.10 (3.9/4.1)

ANSWER 6.11 (1.50) l.

l a. Yellow light above the Core Spray pump switch (on. Panel 9-3).

~

.b. Initiating conditions cloared (GT -114.5" RVL & LT 2.45 psig DWP).

l ~ Reset Core Spray Ini ti ati on '.si gnal . -(0.5 each)

J REFERENCE BFNP: OPL171.045, L.O.'s "F" & "H"-

i l

1 BWR K/A: 209001.K2.03 (2.9/3.1), K4.08 (3.8/4.0), A3.02 (3.8/3.7)

'I i

J-

-~c.

- ' - _ , , , - . _ , , , , - - ., , . . . . . , . . . , .- ,y . , .

'i l

I i

61__EL6NI_SYSIEUS_DEQ1QN z _QQNIBQ62_8UQ_lNSIBudENI8IlgN _PAGE. 35-ANSWERS -- BROWNS ~ FERRY 1, 2&3- ---86/12/01-BROCKMAN, K.

ANSWER- 6.12' (1.00)

a. DEPRESSURIZE (DWP is a seal-in signal)-

b. NOT DEPRESSURIZE (RVL is not a seal-in signal). (0.5 each)

REFERENCE-BFNP: OPL171.043, L.O. "E" I

BWR K/A: 218000 K2.01 (3.1/3.3), K4.01 (3.7/3.9), K4.03 (3.8/4.0)

K5.01 (3.8/3.8) r h

. > - , , - - . . , , , e ..~.,,e..

1 Z. EBQGEDQ8EQ_ _NQBd861_8BNQBdG61_EME8GENQY_@ND ..PAGE 36 86DlQLQQ1Q86_QQNIBQL

, ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

i l  ; ANSWER 7.01 (1.00)

Venting the suppression. chamber first takes. advantage of the " pool scrubbing.effect" (0.5) to_ help minimize-the radiation released (0.5).

REFERENCE BFNP: EDI-2,.PC/P-5, L.D. "B.6" BWR K/A: 259024 EK3.04 (3.7/4.1), EK3.07 (3.5/4.0)

EA2.03 (3.8/3.8), EA2.04 (3.9/3.9), EA2.08.(3.6/4.0)

ANSWER 7.02 (1.50) t

a. From 14.6 - 11.5 Feet Compensates for reduced heat capacity of the Suppression Pool due the reduced level (based upon HCTL assumption of level GTE~14.6 ft)- (0.5)

, Less than 11.5 Feet f.

Precludes incomplete steam condensation (0.3) due to downcomers uncovering ( 0.1 ) given a primary system break (0.1). (0.35 credit for stating only that the Containment overpressurizes) . (0,5)

b. Emergency Reactor-Depressuriztion is required.if you are'out of the safe area (Contingency 2). _ (0.5)

! i REFERENCE BFNP: SP/L, p 2; OPL174.657; L.D. B.7 ANSWER 7.03 (1.00) j a. Permissible f or floating items (to prevent sinking) . (0.5) i

! b. FALSE - (0.5)-

1

_.3

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

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

, ,.7,. .,,r..._., ,,, ,, .,m ,y .,_g-

I -

Zi__ PROCEDURES - NORMAL _ABNQBMOL 1 _EMESGENGY_GND 1 PAGE . 37 BOD 1960SIG86_GONIBg6 L ANSWERS -- BROWNS FERRYJ1, 2&3 - -86/12/01-BRO CKMAN, K.

I REFERENCE i BFNP: SP 7.7, Sections 4.1 & 4.2.1 (4) ; L.O. "B"-

ANSWER 7.04 (1.00)

\;

l a. 3358 MWt (+0,-17 MWt) 1 1

] b. 3293 MWt (+0,-16 MWt ) (0.5 each) i REFERENCE

_BFNP: GOI 100-1,Section IV.B.10; L.O.'s "F" & "G" BWR K/A: PWG #5 - Safety Limits (3.3/3.9) 1 i

j ANSWER 7.05 (1.00) 1 TRUE a.

i i b. TRUE (0.5 each)

J

! REFERENCE

! BFNP: GOI 100-1, Sections A.12 & A.22 4

BWR K/A: 203000 K1.14 (2.4/2.6), SWG #5 (3. 5/4.'4) , SWG #10 (3.7/3.9)'

1 1

t 7 , , - - 3 -w -,y 94y y - ,,.---._.,,mu.,y.-sr .w r . . - - .

_--,.y <.y. -..ya., - - , , p ,- --w - 4

C Zt__ PRO'EDURES - NORMAL 2_8BNQRMAL2_EMERGENGY_AND. PAGE- 38.

88Dlg6QGig86_GQNIBQ6 ANSWERS --1 BROWNS FERRY 1, 28c3 s-86/12/01-BROCKMAN, K.

ANSWER 7.06 (2.50)

1) Rx Water Level (0.25) > 11 "

(0.25) -(0. 5)

2) Rx Pressure (0.25) > 1043 # (0.25) (0.5)

3) D/W Pressure .(0.25) > 2.45 # (0.25) .(0. 5)

4) Rx Scram required and Power > 3%, or indeterminable (0. 5) :

5) MSIV Isolation Required -(0.5)

REFERENCE BWROG EDI Users Guide DFNP: EDI-2, Section 2.1; EDI L/P, L.O. A ANSWER 7.07 (2.50)

a. (1) Stop all Fuel Handling activities (and evacuate the Refueling Floor)

(2) Manually Scram the Reactor

.t (3) Stop the CRD Pump (s) (0.25) and Isolate RWCU (0.25) .(0.5 each)

b. (4) -(1.0)

REFERENCE BFNP: GOI 100-3,Section VI.D; L.O. "D" ANSWER 7.08 (1.00)

C REFERENCE.

BFNP: GOI 100-3, L.O. "S.5" l

L_ __

7. PROCEDURES'- NORMAL _ABNQRMAL 3 _1 EMERGENCY _ANQ. PAGE 39 B8RIOLREIC8L,_CQNIBQL ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

A t

ANSWER 7.09 (1.00)

a. Of fice Service Building' (OSD/PM Of c/Adm Bldg / Plt Of c Bldg)

b. -(1) 75 R (0.25); (2) 25 R (0.25) (0.5 each)-

REFERENCE BFNP: IP-15 L IP ,29; L.O. VI.E; RCI-1, L.O. "B" B t ANSWER 7.10 (1.00)-

a. 150 mrem

6. 300 mrem 7

c. 50 %

i

d. 1 Rem / hour (All +0%,-10%) -(0.25 each)-

1 i

REFERENCE j

BFNP: RCI-9,Section IX.E; L.O. "F" l

1 i

l l BWR K/A: 294001 K1.03 (3.1/3.8) l i

i f

s f

1

JZs__EBQCEDyBES_ _NQBd@65_GBUQBd862_EUEBGENQX_6ND PAGE '/K) 88DIOL9 GIG 86_GQNIBg6

. ANSWERS'-- BROWNS FERRYfi, 2&3 -86/.12/01-BROCKMAN, K.

ANSWER 7.11 (1.50)

a. Increat o -(0.25) Formerly extracted steam now passes through

-the turbine (giving net work / watts out) (0.5) (0.75)

b. Decrease (0.25) Extraction steam to the~next higher pressure heater will' increase (0.5) (because the feedwater tempercture rise across the heater is greater - thus, less energy is available to the turbine) (0.75 REFERENCE

~BFNP: OI-2,Section II.C.1 & .2, L.O.'s "A" & "B"; OPL171.011, Section E' BWR K/A: 259001 A2.02 (3.1/3. 3) (

ANSWER 7.12 (1.50)

a. Complete Loss of Offsite Power (LOSP)

b. Accident Signal present (LOCA on any unit)

c. The dedicated diesel generator (s) cannot supply sufficient power for all of the required loads. (0.5 each)

REFERENCE BFNP: 01-57, Secti on V. E. 4. j (5) ; L.O. "C" BWR K/A: 264000 K5.05 (3.4/3.5), A2.01 (3.5/3.6) m.

... n I g

Zz__EBQGEDQBES_ _UQBd862_8pNQBd862_EMEBGENQY_8ND JPAGE1 41

~ 88D1969 GIG 86_GOUIB96 ,

ANSWERS --- BROWNS ' FERRY -1, I2&3' ' -86/12/01-BROCKMAN,'K.

ANSWER' 7.13, (1.00) a.- 47 psig (+3,.-2 psig)

b.  : Possibility of overloading'the 3C EDG (0.35) during a.

simultaneous LOCA w/LOSP ( 0.15) . -  :(0.5 each).

REFERENCE BFNP: 01-24, Sections II.B & III.E, L.O. "A" BWR K/A: 264000 K3.03 (4.1/4.2) 286000 SWG #10 (3.2/3.5) i

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

8.__6pdINISIB8IIME_BBgCEgyBEgj CONDITIONS 2_AND_kIMIIAIIQN@ PAGE 42

i. .

j ANSWERS -- BROWNS FERRY'1,. .2&3..

. -86/12/01-BROCKMAN,:K.

I i

2

.; ANSWER 8.01 (1.00)

I.

a. It shall be cycled by the motor' operator-(0.4)' prior to being declared operational- (0.1) ( 0. 5 ) ~

j' b. Torque switch could become activated before the valve is off its seat (improper operation of the opening circuitry) . ( 0. 5) ,

i

}' REFERENCE BFNP: OSIL No. 47, p 2; L.O. "T" P

I BWR K/A: 291001 K1.07 (3.4/3.4) 1 I

ANSWER 8.02 ( 1. 50)-

EMERGENCY:- Action is required to prevent imminent major equipment i damage or to protect personnel from any imminent threat of bodily injury ( 0. 35/ O.15) .

I IMMEDIATE ATTENTION - Work is to be performed within the next 24'

! hours (, the next scheduled workday, or upon completion o'f ~ necessary i technical evaluations or material procurement) (0.5) f-ROl] TINE MAINTENANCE - Work is to be performed as manpower and circ-

! umstances permit. (0. 5)-

! REFERENCE BFNP: SP # 7.6, Section 4.12 - .14; L.O. "A"

,- i 4

4 DWR K/A: 294001 K1.02 (3.9/4.5) i 4

i.

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

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

8. ADMINISTRATIVE PROCEDURES2 _ CONDITIONS.~AND-LIMITATIONS- PAGE- zR5

. ANSWERS -- BROWNS FERRY'1, 2&3 -86/12/01-BROCKMAN,.K.

i J

C

~

+

ANSWER 8.03 (1.00)

A'r e ' ( 0. 3 ) (TS.(3.7.A.4.b) allows one Vacuum' Breaker to be Not '

i

_ Fully Closed, as long as it i s less -than 3 deg open. The red light being. 0FF indicates that the breaker -is less than 3. deg open;-the green light being ON-indicates that it is less than 80 i- deg open.-) These are the NORMAL CLOEED indications for the Vacuum )

' Breakers (0.7). (1.0)

REFERENCE BFNP: TS 3.7.A.4.b'.(and Bases); L.O.'s 51-& 55 BWR K/A: 223001 K5.01 ( - / 3.3), K6.09 ( - /3.6)

ANSWER 8.04 (2.00)

a. (1) The cause of the reactor scram has been identified and is j understead.

4 (2) There should be reasonable assurance that' the unit will not

l . scram again due to the same cause.

f

{' (3) The responses of safety related equipment are understood and .

4 acceptable for safe operation. (2 O O.5 each) F

} b. Plant Manager (0.5)

I s c. (1) Scram Committee will investigate the scram.

j l (2) PORC will make the recommendation on' unit restart .

l to the Plant Manager. (2 O O.25 each ,

I

REFERENCE L BFNP: Standard Practice 12.8, Section.6; L.D. C, D, &E i

t J

g .sa

.x y xn p.as. s . 9 u a .s 1u _ ..n.. a

.e 8:__0901NigIB8IlyE_BBggEgyBES2_CONDIllgdg2_80p_ lid 1IGIlgNS- PAGE. 44.

1

. ANSWERS --' BROWNS. FERRY 1, 2&3 -86/12/01-BROCKMAN,.K. ,

4=

~,

'BWR K/A: APE 296006' AA2.06 ( - /318), PWG 01 ( - /4.' 1 )

T-4

= ANSWER 8.05 (1.00)-

d

> REFERENCE

}7 BFNP: Standard Practice 12.24; 10CFR50.54 (iii); L '. O . B&C i

1

.i 1

BWR K/A: 294001 K1.03 ( - /3.7), A1.11 , - /4.3), Gen K1 ( - /2.9)

Gen K5 ( - /3.3) i l

]

ANSWER 8.06 (1.50) l (1) System fails to operate.

j (2) System operates in a suspected adverse manner.

5 (3) System operates outside of the limits of the documented ecceptance critaria. (0.5 each) f I REFERENCE BFNP: Standard Practice 10.9, L.O. "B"; NOAM (Part 2), Section 4.9 >

I 1

1 '

t i

i I: '

., . ..=-,_-_ _ .--- . ..-...c._. _ . , _ _ . ._ _ - - - , .

B.__0DUINISIB8IIVE_EBggEgg8ES 4 _CgNDIllgNS i_80p_LidII8IlgNS 'PAGE 45

' ANSWERS -- BROWNS FERRY 1, 263 -86/12/01-BROCKMAN, K.-

1 ANSWER 8.07 (1.00) ,

, (Shutdown and) calculate the maximum pressure at the bottom of.the vessel (considering the standing head of the water). (0.5)

If GT 1375 psig the NRC.must be notified (within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - NRC permission is required to restart). (0.5)

REFERENCE I

BFNP: Technical Specification 1.2 (and Bases) & 6.5, L.O. 74; Standard Practice 12.8, Section 6.'B.7, L.O. "C" j

I 1

., BWR K/A: 295025 EK 1.05 ( - /4.7), EA 2.01 ( - /4.3), Gen K5 ( - /4.4) a 295020 AA 2.04 ( - /3.9)

I ANSWER 8.08 (2.50)

a. Male or Non-reproducing Female l Lifetime Exposure (5(N - 18))

Volunteer Over 45  :(0.25 each) 4

b. 20 MIN (75/225 = 1/3 hr) (0.5)

! c. 200 Rem (0.25) i

^

l

d. ALERT (>1000Mr/hr)

J Immediately (10CFR20.403)  :(0.5~each)

REFERENCE 10 CFR 20.403 ,

l BFNP: RCI l I, L.O. V.-I . 3 ; Emergency Plan, L.O. B i

4 4

.- .- , .---,-,y _- -- -.4,e,-., - - - - - , - . . . . . . , ,7, , , ,c-... r s s y n e ,-+ ~ e g, ,- y- t .e - w r --

- .- - ~

83 __8QUINI@IB8IlyE_BBgCEpyBEg2_CONDIIIQNS,_8NQ_61dlI@IIQUS; PAGE 46 '

ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

BWR K/A: PWG K1.03 ( -'/3.8)

J

' ANSWER 8.09 (1.50)

Browns Ferry TS (3.7.B.1) requires two of three Standby Gas Treatment!

> (SBGT) (BH) train to be operable at all times when secondary con-

.tainment integrity is required. (0.5) The TS definition of-operable; requires both normal and emergency power to be available. . The l emergency power supplies :( A and D Diesels) for A and.B SBGT'~ fans were inoperable. (0.5) Thus, the=TS LCO could not be met. (0,5).

(Discussion of TS 3.9.c, Electrical Lineup, can be partially credited up to 1.05_ pts - Yes, it is operable, due to the meeting of the 1 requirement for two off-site power sources and two diesels being available) i REFERENCE BFNP: Technical Specification 3.7.B.1; L.O.'s 1.d, 54, & 55 LER 50-259-86026 i

j 1

i BWR K/A:

)

ANSWEP 8.10 (i.50) j a. Browns Ferry Technical Specification (3.5.B.9) requires at least:

j two operable RHR pumps for any unit containing irradiated fuel (0.5).

Technical Specification (3.5.C.7) requires at least one RHRSW pump. .

i for each RHR pump aligned for heat exchanger service (0.5). Thus, the j .TS LCO's are not met (0.5). (1.5) i j (Full-Credit can be given for evaluation under the criteria of TS 3.5.B.10 criteria)

REFERENCE

, BFNP: Technical Specification 3.5.B.9_and 3.5.C.7; L.O.'s 2, 23, & 25 LER 50-259-86009; LER 50-296-86007 & -85017 l

1 b

.-. ,. - , , , . , , - - - - . .: , , . . . . , , - - . - , - - , . --...-,,n,.,.: = a. . . . .

.f' 1 el__0DulNISIBOIIME_BBQQEQQBEQ2_CQNQ111QNQ2_8ND_61dlI6IlgNSt  ;-PAGEL 47

. ANSWERS -- BROWNS FERRY 1, 2&3'; -86/12/01-BROCKMAN,fK.'

BWR K/A:

i ANSWER. 8.11~ (1.00) d j ' REFERENCE BFNP: OSIL #35, L.O. "K"; Technical Specification 3.7.C-, L.O. #76 4

I i

j 4

1

BWR K/A

290001 SWG #10 (3.3/3.4), #11 (3.3/4.2) i PWG A1.02 (4.2/4.2) i ANSWER 8.12 (1.00)

C REFERENCE

!. BFNP: Technical Specification 3.7.D, L.O. #76 Technical Specification Definition-- LCO t

i 4

BWR K/A: 223001 SWG #6 (3.0/4.0), #11 (3.3/4.2)

PWG A1.02 (4.2/4.2) i f

1 1

I I

4 l-

[

\

~

TEST CROSS REFERENCE PAGE 15 ,

QUE'STION 'VALUE REFERENCE 05.01 05.02 1.50 2.00 KEBOOOO907-KEBOOOO908

[

05.03 1.50 KEBOOOO911 05;U4 1.50 KEBOOOO912 05.05 2.00 KEBOOOO913 4

G5.06 2.00 KEBOOOO914 05.07 '1.75 KEBOOOO916 05.OS 2.00 KEBOOOO917 j 05.09 3.00 KEBOOOO919 17.25 12 i 06.01 1.00 KEBOOCOB70

06.02 1.50 KEBOOOO974

06.03 1.50 KEBOOOO878 I

06.04 1.50 KEBOOOO879 -

06.05 1.50 KEBOOOO880 '

! 06.06 2.00 KEBOOOO895 I 06.07 1.00 KEBOOOO927 1 06.08 .50 KEBOOOO928 06.09 1.50 KEBOOOO930 l 06.10 2.25 KEBOOOO932 j 06.11 1.50 KEBOOOO935 06.12 1.00 KEBOOOO936 l- ______

l 16.75 1

07.01 1.00 KEBOOOO885

$ 07.02 1.50 KEBOOOOB87

! 07.03 1.00 KEBOOOOB91

07.04 1.00 KEBOOOO893 i 07.05 1.00 KEBOOOOB94

! 07.06 2.50 KEBOOOO896

! 07.07 2.50 KEBOOOO899 j 07.08 1.00 KEBOOOO900 j 07.09 1.00 KEBOOOO902 d

07.10 1.00 KEBOOOO903 j 07.11 1'.50 KEBOOOO904 i 177. 1 2 1.'50 KEBOOOO905 ,

07.13 1.00 KEBOOOO906 17.50 08.01 1.00 KEBOOOO898 08.02 1.50 KEBOOOO890

. 08.03 1.00 KEBOOOO971 TOB.04 2.00 KEBOOOO972 08.05 1.00 KEBOOOO973 08.06 1.50 KEBOOOO9741

! 08.07 1.00 REBOOOO975 i

,. . - , = - , - , . . , - , . . , . - , .- . . . . - . - - . .- z .- - '.-.---.:......;--.

I k' '

TEST CROSS REFERENCE PAGE 2 OUESTION VALUE REFERENCE 08.08 2.50 KEBOOOO976 08.09 1.50 KEBOOOO977 08.10 1.50 KEBOOOO978 08.11 1.00 KEBOOOO990 08.12 1.00 KEBOOOO991 ]

16.50

__M___

68.00

n. . ..

I U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _EBQWNS_FEBBY_11_2h3_____

REACTOR TYPE: _pWB-GEd_________________

DATE ADMINISTERED: _g6/12fg1________________

EXAMINER: _DBQgl' MAN 2 K.

CANDIDATE: ___~_$_d_b 1 _E__E Z_Z__

INSIBgCI1gNS_IQ_C@NDID8IE1 Use separate paper for the answers. Write answers on one side only.

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

% OF CATEGORY  % OF CANDIDATE'S CATEGORY

__Y8LUE_ _IQIOL ___SCOBE___ _Y@LUE__ ______________COIEGQBy__________,___

_25.Q9__ _22.@Z ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS

_2bz2M__ _25196 ___________ ________ 6. PLANT SYSTEt1S DESIGN, CONTROL, AND INE RUMENTATION

_2Z199__ _2_53 Z8 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

__26159__ _291}9 ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 1931Z5__ ______..____ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

___________________________________ l Candidate's Signature l

N

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

1. . Cheating on.the examination means an automatic denial of your application and_could result in more. severe penalties.

' 2. - Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside-the examination:

"rnom'to avoid even the appearance or possibility of cheating.

3. Use black in'k or dark' pencil ggly to facilitate legible reproductions.

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

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

6. Use only the paper provided for answers.

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

. 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gely 90 gag side of the paper, and write "Last Page" on the last answer sheet.

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

10. Skip at least three lines between each answer.

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

12. Use abbreviations only if they are commonly used i n facility litetatute.

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

1,.

4

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

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

i-

16. If parts of the examination are not . clear as to intent, ask questions uf j- the agaminet only.

l 17. You must sign the statement on the cover sheet that indicates that the

! work i s your own and you have not received or-been given assistance in t completing the examination. This must be done after the examination has i been completed. ,

i I

t e m. . . r , , , , , , .v d. ~ .m -,ma9--

4

'18. When you complete your examination,gyou shall:

a. Assemble.your examination as follows:

(1) Exam questions on. top. -

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

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

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

the examination questions.

c. Turn in all- scrap paper and the balance of the paper that youLdid not use'for answering the quest .

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

L i

r I

T a

1 i

f 1 - -, < . . . , - - - . - , ,. . . - - . .. - - , . , - .1,_._,.-~..v, -.m ,, , , , , . , , ..vr,...,,+-

5. ~THEdRY OF' NUCLEAR POWER PLANT' OPERATION 2 _ FLUIDS 2 _8ND PAGE' 2 ISEBMppyN8MICS QUESTION 5.01 (1. 00)

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 7of 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 downfto 12.0 kw/ft. SELECT the valid preconditioned value for this node.-

ASSUME THE MAXIMUM RAMP-RATE IS .10 Kw/ft/hr

1) 11.0 kw/ft 2). 11.8 kw/ft

3) 12.5 kw/ft -

4) 13.0 kw/fc

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 /> DUESTION 5.02 (1.00)

Which one of the following describes the changes to.the steam that occurs between the inlet and the outlet of a-REAL turbine?

a. Enthalpy DECREASES, Entropy DECREASES, Quality DECREASES ,

b. Enthalpy INCREASES, Entropy INCREASES, Quality' INCREASES

c. Enthalpy CONSTANT , Entropy DECREASES, Quality DECREASES

d. Enthalpy DECREASES, Entropy INCREASES, Quality DECREASES

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

53 THgg8Y_OF__NUgl' GAR _PgWEB_ PLANT _gPgRATIgN 2 _ELUIgg2_AND [PAGE. 3-IHggdggyNAMICE QUESTION 5.03 (1.50)

Concerning the Bypass Flow in the reactor-core:

a. Which one of the following most-. accurately indicates core bypass flow at1100% thermal powerLand core flow?- (1.0).

'(1). 1%

(2) 5%

(3) 11%

-(4) 20%

b. STATE the most significant consequence that would occur if bypass flow were significantly reduced at' full power. (0.5)

QUESTION 5.04 (2.00)

Concerning the Net Positive Suction Head (NPSH):

a. DEFINE NPSH (0. 5)

b. For each of the following, indicate whether.-the ~available NPSH at the suction of the recirculation pump would INCREASE / DECREASE /

REMAIN THE SAME: -(1.5)

(1) The Feedwater Flow is INCREASED (2) The Recirculation Flow is INCREASED (3) The Vessel Pressure is INCREASED from 200 psig to 800 psig' i

a i

I I

i l

2 i

I j .

-(***** - CATEGORY 05jCONTINUED ON NEXT_PAGE *****) _

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

. ~ _.

M 5.-~THEDRY OF-NUCLEAR POWER PLANT OPERATION2 _ELUIDS3_AND PAGE .4'

. T H E R M O D Y N A M _I_C S QUESTION 5.05 (1.50)-

Attached Figure # 911 shows a representative Xenon concentration curve. The reactor is scramped at Point A, resulting in-a peak Xenon concentration appr oximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> later,7at Point B. -

Line B - C represents the normal depletion of Xenon.that would '

.be expected.

If the reactor were RESTARTED at Point B L (dotted ' line) , ~ EXPLAIN:

a. The rapid decrease in Xenon. concentration from' Points B to D. (0.5).

b. The change in the distribution of the radial neutron flux during the period from Points B to D (include reasons) . (0.5) c, Any special precauti on (s) that you would need to observe, as a reactor operator, between Points B and D (include reasons) . -( 0. 5 )

i '

,1 i

QUESTION 5.06 (1.50)

Unit 2 is in Cold Shutdown - The recirculation pumps are necured for 4160 volt buswork - Shutdown Cooling is being throttlhd to maintain 150 - 160 deg F for a reactor restart. A review of your control room instrumentation indicates the following conditions:

i j Bottom Head Temperature (RWCU) - 153 deg F Reactor Pressure -

95 psig

]; Vessel Level - 33 inches-

a. DFSCRIDE the indicated problem (s). (0.5) i b. STATE the corrective action (s) which you should take. (0.5)

c. If the pressure continues to rise - STATE what would tna your most immediate concern. -(0.5) 4 I

i i

l

[ (***r* CATEGORY 05' CONTINUED ON NEXT PAGEJ*****)

',_. .j

~

'k 5 __IBEQBY_QE_ NUCLE 88_EQWEB_ELONI_Q8EBBIlgNi_ELUIDS2 _8ND PAGE .5 IMEBMgDYN8MICS -

-QUESTION 5.07 (2.00) ,

Attached Figure # 913 shows the variation in MAPLHGR with respect to Exposure.

a. ' STATE the reason for the INCREASE in the MAPLHGR limit (Point A to Point B) early in core lif e. . (0. 5) .

b. STATE three (3) reasons for the significant DECREASE.in the MAPLHGR' limit (Point C to "oint D) at the end of core life. (1.5)

QUESTION 5.08 (1.50)

Concerning the effects of Control Rods on reactor power:

a. EXPLAIN how a control rod withdrawal could result in a reactor power decrease (reverse power effect). ~(1.0)

b. Which one of the following would be the rod movement sequence most likely to cause the reverse power effect described above? (1.0) ,

(1) Deep Rod - 10 notch movement (2) Deep Rod - 1 or 2 notch movement (3) Shallow Rod - 10 notch movement (4) Shallow Rod - 1 or 2 notch movement  ;

-QUESTION L.n9 (1.75)

I The reactor is at the beginning of a fuel cycle (BOC) and exactly critical. You withdraw control rods to add 0.06% ^k/k.

. NOTE: ASSUME BETA-EFF = 0.0072 and LAMBDA-EFF = 0.1 SEC E-1

a. CALCULATE the resulting stable period. (SHOW ALL WORK) (0.5)

b. CALCULATE how long it.would take reactor power to DOUBLE. (0.5) c ~. If the reactor were at the End of Cycle (EOC) vice BOC, would the resultant stable period, for the same reartivity addition described in (a), beLLONGER or SHORTER? . JUSTIFY your response! (0.75)

~(****r CATEGORY 05 CONTINUED ON-NEXT_PAGE'*****)

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

5:__IHEgBy_gE_ NUCLE 88_EgNEB_ELONI_gEEBOIIgN3 _ELUlgS1 _6ND PAGE 6 IHERMggyN8MICS a

QUESTION 5.10 (2.00)

The plant is at Full Power conditions, in-3-Element-Control:

For an INADVERTENT HPCI INJECTION transient, STATE for each para--

i- meter listed below:

l i (1) Whether the INITIAL value.would INCREASE / DECREASE /

REMAIN THE SAME, and (2) Whether the LONG TERM value would be LESS THAN/ GREATER THAN/'

j THE SAME as the value before the transient.

JUSTIFY.each of your responses!-

.a. Reactor Pressure b.. Reactor Power

c. Total Feedwater Flow

d. Vessel Level i.

}

j QUESTION 5.11 (3. 00 )

i Attached Figures # 5 A & B represent a transient that could occur at a BWR

GIVEN

(1) One Recirc Pump Flow rontroller Fails High G Time t = 1.2 min (2) No operator actions occur

]

(3) Both Recirc Pumps were on line G Time t = O.O minutes I

EXPLAIN the cause(s) of the following recorder in'dications:

a. Core Flow DECREASE (Point 3) (0.5)

I b. Total Steam Flow DECREASE (Point 10) (0.5) i

! c. Reactor Pressure STABILIZATION (Point 12) (0.5)

(At a lower level) ,

t. -

d. Reactor Level DECREASE (Point 13) (0.5) t e. Reactor Power DECREASE (Point 16) (0.5) l l f. Since this transient occurred with power at less than rated (70%),

STATE the factor (or program) which is utilized.to. ensure that the

MCPR limit .is not exceeded. (0.5) l' l

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

_ ._ .. _ - . - _ _ , . . . _ .;.. _ _ _ , . _ . _ _ . . . _ _ _ _ . . . . . , . _ . ~ . __ _ . _ . . . _ .

7 5s__IBEQBY_QE_UUQ6E88_EQWEB_E66MI_QEEB811QU2_ELUIDS1 _8NQ PAGE 7

.IHEBMQDYN8M1QS QUESTION 5.12 (1.00)

During a refueling outage, one Staandby Liquid' Control (SBLC) loop is-tested per the-Technical Specifications by firing a Squib Valve and pumping the contents of the Test Tank into the vessel. Due_to an oper-ational' error, the Test Tank contained a significant Baron concen-tration (as opposed to being purely deimineralized water). This resulted in a Boron concentration of approximately-100 ppm inside of the-vessel.

DESCRIBE the problem which could occur from the POAH (onward) if the reactor was started up in this condition.

QUESTION 5.13 (1.50)

For each of the following transients, STATE which of the reactivity coefficients (Moderator, Void, or Doppler) would respond first.

JUSTIFY your choice.

a. Rod Drop Accident from 10% thermal pcwer. (0.75)

b. MSIV Closure from 90% thermal power. (0.75) 1 QUESTION 5.14 (1.50)

A reactor heat balance was performed (by hand) during the 00-08 shift due to the Process Computer being OOC. The GAF's were computed, but the APRM GAIN ADJUSTMENTS HAVE NOT BEEN MADE.

FATE whether the Actual Power will'be HIGHER or LOWER than the alculated Power for each of the following:

a. The steam flow used in the heat balance calculation j was LOWER than the actual steam flow.  !

b. The reactor recirculation pump heat input used in the heat balance calculation was OMITTED.

c. The RWCU return. temperature used in the heat balance calcu-lation was LOWER-than the actual RWCU return temperature.

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

+ - . . . ~ . _- .- - - - -

> s J

, Sz__IBEQBy_QE_ NUCLE 68_EQWEB_E66NI_QEEBOIlgN 2 _ELUIDS1 _6NQ .PAGE 8 p .IBEBdQDYN6MICS i

1

-OUESTION 5.15 (1.00)

) During a Scram recovery, the reactor is cooling down_at the maxinium t administrative limit. 'The K-eff of the reactor is 0.99.

l CALCULATE when the reactor would theoretically achieve RECRITICALITY.- s IGNORE THE EFFECTS OF-XENON. SPECIFICALLY. LIST ALL ASSUMPTIONS!

QUESTION 5.16 (1.25)

SELECT the apropriate response for each of the folowing examples of variable Control Rod Worth:

i

! a. (MORE/LESS) control rods would need to be pulled to make the reactor critical at 545 deg F, as opposed to 140 deg F.

j b. An INCREASE in the Void Fraction will r esul' t in an (INCREASE / DECREASE)

, in individual control rod worth.

c. Control Rod Worth at End of Cycle would be (LESS/ GREATER) than at j the Beginning of Cycle.

d. Control Rod Worth will (INCREASE / DECREASE) with an INCREASE in

] moderator temperature.

e. Control Red Worth will (INCREASE / DECREASE) as the adjacent control rods are withdrawn.  !

l, i

i l

i j

1 i

1 1

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

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

5

6. PLANI _SygTEMS_DgglgN2 _CgNTRgL2 _AND_INSTRUMENIAIION' PAGE 9 1

i 1

QUESTION 6.01 (1.00) j l The reactor is at 100% Rated ~ Thermal Power q i "A" Feedwater Level Control is Selected All Controls are in Normal / Automatic A failure in a'Rx Vessel. Level Narrow Range Instrument has occurred and-

resulted in the following related trips / indications

} "A" NR level indication reads at MINIMUM "B" 84 "C" NR level indications read at MAXIMUM i REACTOR WTR LEVEL A ADNORMAL Annunciator - ON

! REAC VESSEL WATER LEVEL LOW-LOW CHAN A Annunciator - OFF 1

.I 1 REAC VESSEL WATER LEVEL LOW-LOW CHAN B Annunciator - OFF j Two channels of Level C have TRIPPED One channel of Level 3 has TRIPPED i

j Feedwater flow is at ZERO State which NR level transmitter has failed AND state in which direction J it has failed (HIGH/ LOW).

t c 1

1

}

l QUESTION 6.02 (1.50)

I Concerning the APRM Circuitry:

i

a. STATE the (ranga of) time by which the flow-biased signal is time delayed. (0.5)  :

i  !

b. STATE what causes the circuitry to arrive at a particular time I i delay (within the range described above). (0.5) I

! c. The APRM flow-biased Scram trip is present when the MODE SWITCH  ;

contacts are (OPEN/ CLOSED /BOTH OPEN AND CLOSED) - CHOOSE ONE! (0.b) -I l-i I

i t

i i

(***** .CATEGDRY.06 CONTINUED.ON NEXT-PAGE *****)' . ,

6s__PL8NI_QYSIEMS_ DESIGN2 _QQNIBQ63_8ND_lNSIBUMENI8IlgN; PAGE 10 QUESTION 6.03 (1.50)

Concerning the Fire Protection System:

a. MATCH the Detection System (COLUMN A) with its Useage (COLUMN B) - (0. 75)

COLUMN A COLUMN B

1. Rate of Rise Thermal Detector 'a. General area coverage throughout the plant

2. Continuous Strip Detector

b. Spreader Rooms and around~any-

3. Ionization Smoke Detector oil reservoirs

c. Cable Tray protection

b. Given that a fire detection system provides an initiation function - STATE how the Fire Protection / Raw Service Water system responds / aligns. (0.75)

DUESTION 6.04 (1.50)

Concerning Refueling Operations:

a. Attached Figure # 879 shows three 1/M plots fo'r a hypothetical reactor. SELECT the curve (#1 or #2) which is indicative of the detector being TOO CLOSE to the source. (0.5)

b. STATE the interlocks which will prevent movement of'the Refueling Bridge over the core. (1.0)

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

' l

__ J

6 ELONI_SYSIEUS_ DESIGN z _CQNIBg62_8ND_INSIBUdENI8IlgN .PAGE 11 OUESTION 6.05 (1.50)

Concerning Refueling Equipment:

a. The ref ueling grapple is considered to be " NORMAL UP" at a level of approximately 8 feet below the top of the refueling plat-form rails. STATE the reason for this movement interlock. (0.5)

b. STATE what provides the motive force for operating each of the

-following: '(0. 5)

(?) Channel Bolt Wrench (remove / install channel-fastener assembly)

(2) . Refueling Grapple (grab / release fuel elements for movement)

c. STATE the indication (s) which you would use to determine that an excessive pull was being exerted on the channel (during Fuel Assembly Dechanneling). (0.5)

DO NOT ADDRESS PHYSICAL DEFORMATION, ETC., OF THE CHANNEL

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

bz__EL8NI_SYSIEMQ_DESIGM1_CgNIBg62_8Ng_INSIBUMENI8IlgN PAGE 12 QUESTION .6.06 (2.00)

The following plant conditions exist:

~

Jet-Pumps 1 - 10 Differential Pressure (Meter) = 15 psid Jet Pumps 11 -'20 Differential Pressure (Meter) = 3 psid RECIRC LOOP A ONLY OUT OF SERVICE (Annunciator)' = ON

a. The TOTAL CORE FLOW Recorder would calculate core flow by which of the following methods? (1.0)

(1) Loop B Jet. Pump Flow + Loop A Jet Pump Flow (2) Loop B Jet Pump Flow - Loop A Jet Pump Flow (3) Loop B Jet Pump Flow Only (4) Loop A Jet Pump Flow Only

b. After 5 minutes, the operator opens the Discharge _ Valve of Recire Pump A to maintain the A Loop temperature. The TOTAL CORE FLOW Recorder would calculate core flow by which one of the following methods? (1.0)

(1) Loop B Jet Pump Flow + Loop A Jet Pump Flow (2) Loop B Jet Pump Flow - Loop A Jet Pump Flow (3) Loop B Jet Pump Flow Only (4) Loop A Jet Pump Flow Only QUESTION 6.07 (1.00)

While venting Primary Containment through the. Standby Gas Treatment.

System (SDGTS), you receive a High Stack Gas Radiation Alarm, requiring

.you to discontinue venting. Which one of the following values is the value at which you would expect this to occur?

a. 3.8 E+1. cps

b. 3.0 E+2 cps

c. 3.8 E+3 cps

d. 3.8 E+4 cps

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

_ a

6.__PL@NI_SYSIEUS_DEgigN3 _CgNIBg61_8ND_INSIBudENI8IlgN PAGE?=13 )

QUESTION 6.08 (-.50)

EXPLAIN why it is essential to ensure that the. Diesel Generator-governor is positioned fully to the LOW SPEED STOP, prior to performing.a local start.

DUESTION -6.09 (1.50)

Unit 2 is at full power when a Load Reject occurs due to a trip of the MTO Output Breakers.

a. STATE the specific signals (and their sources) which activate the Load Reject circuitry. (0.5)

b. DESCRIBE the response of the Turbine / EHC System to this event. (1.0)

DUESTION 6.10 (2.25)

An automatic LPCI injection signal is received:

a. STATE the three specific conditions (setpoints, if applicable) that would be required to allow you to initiate Containment Spray. (1.50)

b. STATE whether each of these conditions can be bypassed. (0.75)

GUESTION 6.11 (1.50)

A Core Spray initiation signal is present on Unit 2. You MANUALLY secure the "C" Core Spray pugp due to increasing reactor vessel level,

a. STATE the indication (s) that would be provided on the control

. panel that the pump was manually secured. (DO NOT address flow, pressure or other metered indications) (0.5)

b. LIST the conditions that would have to exist and/or the actions that you would have to take to permit the pump to automatically restart, if'needed. (1.0)

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

6.__P68NI_SYSIgMS_ DESIGN2 _CONIBg62_8ND_INSIBUMENIBIIQN PAGE 14

-QUESTIGN 6.12 (1.00)

STATE whether ADS would DEPRESSURIZE or NOT DEPRESSURIZE the reactor for each of the following conditions:

CONSIDER THE TOTAL ADS RESPONSE AND ASSUME NO OTHER OPERATOR ACTIONS; OCCUR

a. After. initiation of the 120 second timer, but before it is timed-out, Drywell Pressure DECREASES to 1.5 psig. (0,5)

b. A-fter-initiation of the 120 second timer, but before it is timed out, Reactor Water Level INCREASES to -90 inches. (0.5)

DUESTION 6.13 (1.50)

Concerning the Safety Relief Valve tailpipe Vacuum Breakers:

a. STATE the purpose of these Vacuum Dreakers. .(0.5)

b. Assuming accident conditions, LIST two (2) significant detri-

. mental effects that could occur if one (1), or more, of these Vacuum Dreakers STUCK OPEN. (1.0)

(***** cal'EGORY 06 CONTINUED ON NEXT PAGE *****).

l 1

t Ez__E68NI_SYSTEMg_ DESIGN _CQNTBQLi _8NQ_

1 INSTRUMENTATION LPAGE' 15 .

QUESTION 6.14 (1.00)

UnitT1.is operating at 100% RTP, with recirc'in Master Manual ~ The-"A" . l Pressure Regulator unit, which is governing,. FAILS LOW (i.e., the output (s) . of Regulator "A" FAILS DOWil) 4- ~ ASSUME: 1. No Operator Actions

2. All other EHC control settings are normal 3.. Starting Parameters o TCV's - 100% Steam Flow Position a BPV's -

OX Steam Flow Position o Power - 100% Rated Thermal' Power

) o Pressure - 1010 psig NOTE: FIGURE # 374 IS ATTACHED FOR REFERENCE Which of the following most accurately describes both the INITIAL RESPONSE and FINAL STATUS of the different parameters and components. j a b c d j INITIAL. RESPONSE o TCV's NO CHANGE l THROTTLE CLOSElTHROTTLE CLOSE! NO CHANGE o BPV's  ! THROTTLE OPENITHROTTLE OPEN 1 NO CHANGE I THROTTLE OPEN o Power  ! DECREASE  : NO CHANGE I INCREASE i DECREASE 1 o Pressure i DECREASE  : NO CHANGE I INCREASE  ! DECREASE

!  !  ! l 3 FINAL STATUS 1  ! l  ! I l 1 l 1 i

! o TCV's lOX(MSIV Shut): < 100 % i *100 %.  ! -NO CHANGE j o E'PV 's IO%(MSIV Shut): > 0%  ! O%  : 0%

r o Power 10% (Rx Scram): > 100 % > 100 % < 100 %

l o Pressure IAs controlled! >1010 psig i >1010 psig i <1010 psig i Iby SRV's and i l IHPCI/RCIC  :  :

1

)

i i

+

4 i

1 4

1 i

(***** '. CATEGORY 06 CONTINUED ON NEXT PAGE ****y)

1 I.

l

6. .eLQNI_SYSIEUS_ DESIGN 2 _CQNIBOL2_6MD_IN@IBudENI8IlgN PAGE 16 j

QUESTION 6.1S (2.00)

For each of the following situations (i' and ii) select 1 the correct Feed-water Control System / plant response from the list- (a through e) which

, follows. An answer may be used more than once, and NO operator actions are taken

! a. Heactor water-level' decreases and stabilizes at a lower level.

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

j c. Reactar water level ~ increases and stabilizes at a higher ' level.

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

e. None of the above.

i. The plant is operating at 70% power, in 3-element control, when One (1) MSIV Fails Shut.

ii. The plant is operating at 100% power, in 3-element control, when One (1) Feed Flow Detector FAILS DOWNSCALE.

s i

i l QUESTION 6.16 (2.00)

Answer the following with respect to the Rod Sequence Control System:

a. During the performance of SI 4.3.B.3.a concerning RSCS Operability prior to startup, a " TEST FAILED" light is received (Panel 9-28) after-depressing the "COMPARATOR CHECK A" Pushbutton. STATE whether this indicates a SATISFACTORY or UNSATISFACTORY' response. JUSTIFY your response! (1.0)

b. DESCRIBE how the lamp dimmer function differs when power is between the LPSP & LPAP, as compared to being below the LPSP. ( 1. 0 )

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

3 6t__E66dI_SYSIEdg_ DESIGN2 _CONIBg63_8NQ_INSIBUMENI8II9d PAGE 17-QUESTION 6.17 (1.00)

The RHR pump suction valves from the Torus are interlocked to prevent opening when the Shutdown Cooling RHR pump suction vlaves are open -

this prevents diverting reactor coolant tc the Torus while in Shutdown Cooling.

DESCRIBE another potential path for diverting reactor coolant to.the Torus during Shutdown Cooling (and flushing) operations,and STATE the specific operator actions which are required by procedure to ensure that this path is unavailable.

QUESTION 6.18 (2.00)

LIST the four (4) conditions that must exist for a Diesel Generator Output Breaker to Close following an AUTO start signal.

I 1

4 3

I.

4 k

i

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

I

Zz__EB9gEQUBES_ _bgBM862_8BNQBd862_EMEggENgy_8NQ PAGEL;18 B8DI9LQglg8k_ggNIBQ6-QUESTION 7.01 (1.00)

EDI-2, " Containment Control", states in Step PC/P-5:

IF drywell pressure exceeds 55 psig-(70 psia), THEN

... vent the primary containment ... to reduce and maintain pressure below 55 psig as follows:

1. IF suppression pool water level is below 20 feet,

,THEN vent the suppression chamber.

EXPLAIN why (if conditioins allow) the suppression' chamber is vented-first (to reduce drywell pressure). Include the. benefit acheived and how this is accomplished.

QUESTION 7.02 (1.50)

Attached Figure "E" shows the Heat Capacity Level Limit (HCLL):

a. EXPLAIN the purpose for having an HCLL (i.e., what does the curve protect against?) (1.0)

b. STATE the immediate action (s) required if plant conditions are out.

of the safe area. (0.5)

OUESTION 7.03 (1.00)

Concerning recovering lost or unsecured articles introduced _into the primary system:

a. STATE when IMMEDIATE recovery is allowed. (0.5)

b. TRUE OR FALSE If immediate recovery of an article has been performed, a Form' BF-8 is not required; however, the Unit Superintendent must be notified. (0.5)

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

t

7. PROCEDURES 1 NORMAL1 _ABNQBMAL2_EMEBGENQY_8NQ PAGE 19 88 Dig 60GIC66_CQNIBQL QUESTION E7.04 (1.00)

Concerning plant. operations:

a'. STATE the thermal power level which must not be' exceeded under

-any conditions. (0.5)

b. STATE the maximum allowed average thermal power'for an.8-hour shift. (0.5)

DUESTION 7.05 (1.00)

Concerning a Plant Startuo:

a. Step A.12 of GOI 100-1 states:

"RHR system in standby readiness condition in accordance with-01-74" TRUE OR FALSE Step A.12 may be signed off with RHR pumps A & C operating to provide Shutdown Cooling. (0.5)

b. Step A.22 states:

" Reactor recirculation system ready for: service, in accordance.

with 01-68" TRUE OR FALSE Step A.22 may be signed off with Recirculation Pump A Out.of Ser-vice for Shutdown Cooling requirements (only). (0.5)

QUESTION 7.06 (2.50)

LIST the five (5) entry conditions to EDI-1, "Rx Control" (Include Setpoints, as appropriate). .(1.0)

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

-=

s

. Zz__EBgCEQUBEg_ _NQBd862_8DNgBd862_EdESGENCY_8NQ PAGE" '20

.BSDIgLQQ1C86_CQNISQL-QUESTION 7.07 .( 2. 50 )

Concerning. Refueling Operations:

a.- Excepting notifications, LIST ALL of the actions to be taken to shutdown.the reactor in the event of an Inadvertent Criticality ~

during Fuel Loading. (1.5)

6. Which one of the following is the lowest level person authorized to initiate Standby Liquid Control during an' Inadvertent Criticality during Fuel Loading? (1.0)-

(1) Any Operator (License is not required)

(2) Unit Operator (RO License)

(3) Assistant Shift Engineer ( RO License)

(4) Assistant Shift Engineer (SRO Licease)

OUESTION 7.08 (1.00)

During Refueling operations, when the Moisture Separator is trans-Ferred from the Equipment Storage Pool to the Reactor Well, the upper sections may temporarily break the surface-of the water - a Refueling Zone Isolation may occur.

Assuming that the (i sol at i on ) initiation signal-clears, which one of the following most accurately describes when the isolation may be cleared and the applicable systems returned to operation in accordance with 01-307

a. Immediately

b. Once HP verifies that the source of the initiation was t!e ceparator, via survey.

c. . When the' separator is again submerged.

d. Not until all transfer operations are complete and an HP survey has been performed.

(***** CATEGORY _07 CONTINUED ON-NEXT PAGE *****)

L__-____-

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

i UNIT"LD STATES

1. jk# K8 og,# o NUCLEAR RESULATORY COMMIS813N

• • 'de l

O .f, o REGION il ,i I a 101 MARIETTA STREET. N.W.,5UITE 2900 .t

• 2 o, ATLANTA, GEORGIA 30323 s,.

} .

t t

t 1

i 1 F

'i j i t

i t

i f

1 I

) [

t i

e

{

I F

! I t

k I

i P

f

)

5 r

f e

t 9

% n .

e .N. e,.e-,

Z __EBgggpUBgg - NOBMAL _ASNQRMAL

_ 2 2_gMERggNgy_AND. 'PAGE 21.

BBD196991C86_GRNIBQL QUESTION 7.09 (1.00)

Per IP-24, " Earthquake Emergency Procedure", the magnetic playback system (SMP-1) starts on which one of the f ollowing signals? '

a. Manual Initiation PB

b. START OF STRONG MOTOR. ACCELEROMETER annunciator (0.01g)

c. SEISMIC TRIGGER annunciator (0.1g)

Activation of the Peak-Recording Accelerometers (PRA-103) (1.Og)

~

d.

OUESTION 7.10 (1.00)

Concerning the Emergency Plan:

i

a. Per IP-20, " Technical Support Center", STATE the AL TERNATE location l of the TSC. ' (0. 5) l.

b. Per IP-15, " Emergency Exposure", STATE the allowable emergency exposure for each of the following: (0.5)

(1) Life Saving Actions (2) Less Urgent Emergencies (facility protection)

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

a

l l

s

• 32303 AIGROEG,ATNALTA 0002 ETIUS ,.W.N TEERTS ATTEIRAM 101 8

/

3;

\ *

,3 5

li NOIGER . n NOISSIMMOC YROTALUGER RAELCUN t>

o s2 SETATS DETINU q % 8"*>gy

! LZ. _PBgCgguRgg___ NgRMAL2 _6gNgRMAL3_gMgRGENCy_ANg- PAGE f22

- B6D196gGIC66_CgNIBQL

.i 2

i l

1 QUESTION 7.11 (1.00)

Concerning the Routine Radiation Work Permit (RRWP):

FILL IN THE BLANK FOR EACH OF THE FOLLOWING

a. No one is permitted to receive a dose greater than _.__ per day on any RRWP or any combination'of RRWP's.

b. Personmnel with an exposure limit of. _ _ _ _ _ _ perfquarter or i less will not be allowed to enter on an RRWP.

, c. No'RRWP user will exceed ______ of their remainin'g allowable dose (RAD) in one_ day as indicated on a current personnel. listing, except

by special permission of the Health Physics section supervisor or an

assistant Health Physics section supervisor.

1 4

! d. Entries into dose rates of ________ or greater will require the approval of Health Physics supervision (see RCI No. 1) and the pre-

, sence of a Health Physics technician. Additional HP surveys may be required prior to entry. These requirements will be noted on the RRWP.

i i

OUESTION 7.12 (1.50)

Concerning operations with Feedwater Heaters removed from service, as per DI-2, Condensate System:

a. STATE whether kilowatt output will INCREASE / DECREASE / REMAIN THE i SAME when the highest pressure feedwater heater is removed-from ser-vice. JUSTIFY you answer. (0.75)

b. STATE whether kilowatt output will INCREASE / DECREASE / REMAIN THE i SAME when other than the highest pressure feedwater heater is' removed l from service. JUSTIFY your answer. (0.75)

I~ l

\

QUESTION 7.13 (1.50) l STATE the three (3) conditions which must'be present to allow the i j Unit 1 and 2 Diesel Generators to be paralleled with the. Unit 3-  !

! Diesel Generator. l s

4

(*****[ CATEGORY:07 CONTINUED ON NEXT~PAGE *****)

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

j Zz__EB9CEggBES_r_NgBd862_8BNgBM862_EMEBgENCy_8Ng- PAGE 23

'889196991C86_ggNIBg6 OUESTION 7.14 (1.00)

Per 0I-24, " Raw Cooling Water":

a. STATE the header pressure at which the STBY (spare) RCW pump willEautomatically start. (0.5) .

-b.

OI-24 cautions the operator NOT to run the 3D RCW pump in. .

automatic. STATE the reason for this caution. ' (0. 5) 1 QUESTION 7.15 (1.00)

With Unit 2 operating at 90% power, an Off Gas Hydrogen EH-2]

High alarm is received. Both Hydrogen analyzers indicate a Hydrogen' level of ~ 5.3%. Which one of the following responses ,

most accurately reflects the proper course of action you should take?

a. Change over to the alternate Recombiner.

b. Change over to the Alternate Off Gas train,

c. Start an additional SJAE to assist in Condenser H-2 removal,

d. Manually Scram the reactor.

QUESTION 7.16 (1.00)

The RPV Head is removed and a High Decay Heat condition exists. .One loop of Shutdown Cooling has been established to control temperature.

STATE the actions which should be- taken if Shutdown Cooling is lost and the vessel temperature increases to greater than 150 deg F.

QUESTION 7.17 (1.00)

Per the EDI's,. DEFINE " Adequate Core Cooling".

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

__ __ ____ _ _ _ ._ ~

Z. -PROggpURgg_ _NgRMAL 2_ADNgRMAL 2_gMgBGENgY_AND .PAGE 24 l 60DIOLOGIG66_G9NIBg6 QUESTION 7.18 (1.00)

You are executing EDI-1, " Reactor Control", in response to a reactor scram and subsequent low water level. Drywell cooling has been lost-due to an electrical malfunction to the fans. After 10 minutes, you receive High Drywell Pressure signal and note it'is 5 psig and still increasing.

Which of the following is the correct action .to take according to the generic guidance provided in using the EDI's.

a. Re-enter EDI-1 at the beginning (Only required action)

b. Continue in EDI-1 from where you are - it will direct you to the correct action.

c. Exit EDI-1 and Enter EDI-2, " Primary Containment Control" j

d. Re-enter EDI-1 at the beginning AND enter EDI-2.

QUESTION 7.19 (1.00)

EOI I-1, RC/L states:

If any control rod is NOT inserted to or beyond position (02), disable ADS auto blowdown function (reference Appendix 3) and enter C5 (LEVEL /

POWER CONTROL)

STATE the basis for disabling the ADS auto blowdown function in this plant condition. (1.0)

QUESTION 7.20 (1.00)

Per 0I-47, "Turbogenerator", the operator is required to perform certain actions immediately after a unit trip during Emergency Shutdown conditions. .

LIST the actions which are to be performed at/in the Breaker Control Cabinet.

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

y -l

~

l

. Zz__EBgCEQg8Ejk;_NgBd86.3_8pNgBU862_EtLEBQENGL8ND ' ~PAGE 25 )

88DI96991C8L CONTROL I

1 1

~

I QUESTION 7.21 (2.50)

LIST the five (5) entry conditions to EDI-2, " Primary Containment Control". (Include Setpoints, as appropriate). (2.5) l 1

1 ,

t i

t

,\

T t

I I

I i

i t

i

?

i h

I I

1 h

i i

i I (***** END OF CATEGORY 07 *****).

i

l e.__epMINISIBeIIVE_EEgCEDUBEg2_CONDIIlgNg2_6ND_LIMIISIlgNS PAGE 26 QUESTION 8.01. (1.00) ,

Special concern is required for returning MOV's, that have been MANUALLY CLOSED, to service.

a. DECRIBE how to verify the operability of'an MOV in a safety system that has been manually closed. (0.5)

b. DESCRIBE the malfunction that this operability check is

-intended to identify. (0.5) f QUESTION B.02 (1.50)

STATE the criteria that are used to determine if an MR should be-J checked EMERGENCY, IMMEDIATE ATTENTION, or ROUTINE MAINTENANCE.

4

( i . e. , Define each classification) 1  ;

a i OUESTION O.03 (2.00)

MATCH the Color of the Tags in CLOUMN A to the appropriate Usaege

! in COLUMN D. l; i

j COLUMN A COLUMN B ,

l j

a. Red Tag 1. Used to identify the' isolation boundaries of the clearance.

b. White Tag

2. Used as a master. tag for the

c. Blue Tag clearance - installed on the main j control point to isolate. equipment i d. Yellow Tag from all sources of energy.

i j 3. Used to call attention to the  ;

j existence of unusual circumstances  !

! - attached to equipment where ab-  !

normal operating conditions exist.  !

l

! 4. Used to identify equipment or -

I controls when the equipment is to q be operated by any person other j than its operator.

I l

i l

! 1 I

l i

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

I BA__69dINI@IBBIIVE_BBgCEDUBES 2 _CgNQIllgNg3_8ND_LidlI611gNS PAGE 27 QUESTION 8.04 (1.00)

With the reactor in STARTUP and at 10% RTP, the following Drywell to Suppression Pool Vacuum Breaker mor,itoring light. indications are noted for one of the Vacuum Breakers:

Check Light -

ON Green Light -

ON Red Light -

OFF STATE whether the Technical Specifications for Primary Containment are satisfied (Not exceeding LCO). JUSTIFY your decision!

OUESTION 8.05 (2.00)

Per Browns Ferry Standard Practice 12.8, following a reactor scram and post trip reviews

a. STATE two (2) criteria that must be met.before you, as the Shift Engineer, could recommend a unit restart. (1.0)

b. STATE who (Job title) is responsible for authorizir.g a restart (criticality). (0.5)

c. LIST all additional actions, if any, which would be required prior to restart if you, as the Shift Engineer, and the STA disagreed on the restart recommendation (i . e. , STA recommends startup delay). (0.5)

(***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)

i 4

1 8 t __8Qdidl@IB8IIME_BBQCEQUBES2_CQNQlll[j@s_8NQ_LidlI@IlgNS PAGE 28 i

L I

f j

i

, QUESTION B.06 (1.00) 1 i

j Units 1 & 3 are in Cold Shutdown; Unit 2 is in Startup, but STILL

SUDCRITICAL. The ASE requests your presence in the reactor building 4

to help resolve a significant problem with a CRD pump. As the Shift j'

Engineer and the only SRO in the Unit 1/ Unit 2 Control Room, which one of the following most accurately reflects your allowable response?

I j a. You may respond to the request, providing there are i two licensed RO's in the Control Room and no additional rods are pulled in your absence,

b. You may respond to the request, providing there are two licensed RO's in the control Room and the STA j replaces you, r

c. You may respond to the request, providing there are i two litenced RO's and an STA in the Control Room, and

} you remain in continuous radio contact and within 10

} minutes of the Control Room.  ;

1

! d. You may not respond unless relieved by another SRO.  ;

l i

1,

QUESTION 8.07 (1.50) i An Operability Test is conducted on a safety-related system, following I i the installation of an approved modification. LIST three (3) sep- I j arate criteria which would procedurally require that a TEST DEFICIENCY  !

j be initiated (documented). l I

l j OUESTION 8.08 (1.00)

J With the reactor at Full Power, an inadvertent IM error results ,

{ in the closure of ALL FOUR MSIV's. Your post trip r'eview indicates ,

that the Wide Range Vessul Pressure Recorder spiked at 1207 psig.

e t

i STATE the unique actions, if any, that would be required prior to j restart in this case. l l

4 l ,

1 '

1 i

a a

i

(****n

~

CATEGORY 00 CONTINUED ON NEXT PAGE *****)

l t

l l

82 __8QUINISIB8IIVE_EBgCEDUBgg3_CgNQIllgNg2_8Np_61dlI8IlgNS -PAGE .29 QUESTION 8.09 (2.50)

Following a power transient and subsequent reactor Scram, a mechanic is reported to be severely injured and unconscious due to a ruptured RWCU line. HP technicians report that the. readings in the-area are as high as 225 Rem / hour.

a. STATE three (3) criteria that you would utilize to select the Rescue Team personnel'. (0.75)

b. ' CALCULATE _the maximum Stay-Time that you could permit to effect the rescue. (0.50)

c. STATE the dose to the EXTREMITIES that you could permit to effect the rescue. (0.25)

d. Utilizing the enclosed classification table, Figure #976, STATE the applicable Emergency Classification and LIST the MOST RESTRICTIVE NRC reporting requirements. (1.00) -

QUESTION 8.10 (1.50) .

Units 1 and 2 are in refueling outages; Unit 3 in in an extended maintenance outage. While operating the Unit 1 and 2 standby dieuel generator (EK) D, a mechanical overspeed trip occurs.

Diesel generator A is also out of service at the time, due to  !

extended maintenance.

STATE whether the Technical Specification limiting conditions for operations are met. If not, SPECIFY which TS's is/are not complied  !

with (TS Item number not required, but acceptable). In either case, 1 JUSTIFY your response!

I l

l l

l

(***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)

y

'Di__0DdINigIB8112E_PBgCEDUBES2_CQNDIllgNS2_8ND_ Lid 1I8I1983 PAGE 30 QUESTION 8.11 (1.50)

Unit 3 is in an extended maintenance outage. RHR Pump 3D is running for thm purpose of maintaining reactor coolant temperature above 100 deg F. The following equipment status is reported:

RHR Pump 3D -

Out of Service for maintenance RHRSW Pump D1 - Out of Service f or maintenance.

RHRSW PUMP D2 D/G - Preventive maintenance in progress Loop I of RHR is now declared INOPERABLE due to a damaged pipe hanger on the 24" line above the Torus. (S/D Cooling and LPCI Injection)

STATE whether the Technical Specification limiting conditions for operation are met. If not, SPECIFY which TS's is/are not complied with. (TS Item number not required, but acceptable) In either case, JUSTIFY your response! (1.5)

DUESTION O.12 (1.50)

Unit 1 Technical Specifications specify for REACTIVITY CONTROL ...

"A sufficient number of control rods shall bu oper-able so that the core could be made subcritical in ..."

LIST the three conditions / assumptions which must be met to verify this " Reactivity Margin - Core Loadino." (i.e., the three annumptions for determining Shutdown Margin)

OUESTION O.13 (1.00)

A fuse blows for an UNKNOWN reason. It supplies power to a piece of equipment which is NOT REQUIRED for continued plant operation.

STATE the actions which you should direct to be taken, as per Standard Practice DF-12.24, " Conduct of Operations."

(***** CATEGORY OG CONTINUED ON NEX*r PAGE *****)

9:__8DdINISIB011ME_EBgCEDUBES 2 _CgNDlIlgNg3_8Np_LIUIISIlgNS PAGE 31 t

I QUESTION 8.14 (1.00)

,i

+

Concerning Standard Practice 14.25, " Clearance Procedure":

4

STATE the provisions for performing short-term emergency main-I tenance if the workers are not authorized to hold a clearance.

j Specifically address to whom the clearance is issued and what j " Remarks" are required on the clearance. .

i  :

i I QUESTION 8.15 (1.50) 1 Standard Practice 3.11, "Second-Person Verification", provides-guidance j for BYPASSING the second-person verification required by an instruction.

j ,

] NOTE: ASSUME EMERGENCY CONDITIONS EXIST  !

i j a. LIST who may do this. (0.5) i i b. LIST the reasons for which this may be done. -

(0.5)

J l

c. LIST who must be notified of this, and BY WHEN this notification i must be made. (0.5)

OUESTION O.16 (1.00) >

1 Which of the following scenarios requires application of the Power I Transient Fuel Cladding Safety Limit of the Unit i Technical Spec-3 tiications?

4

a. Reactor power is at 70% RTP; a steam leak to the Drywell occurs i and Drywell pressure rises; the roactor SCRAMS at

• 2 psig; Diesel

! Generator auto-initiation does not occur, but manual start is suc-

! cessful; the reactor is brought to a cold shutdown condition.

j b. Ruactor is in Start-Up, at 7% RTP; power is increased by rod I j pull; the reactor SCRAMS at-10.5% power, by APRM'u; level and preu-

! sure are maintained by normal systems for the plant status.

I j c. Reactor power is at 42% RTP: the main turbine trips due to

an EHC malf unction; the reactor SCRAMS based upon the Turbine Trip; j the DPV's control pressure thereafter, i

l d. The reactor in at 10% RTP; 1-1/2 DPV's are open in preparation

for turbine warmup; controller failuru reduces pressure to 875 puig; l MSIV's close; the operator then manually SCRAMS the reactorp level ,

j and pressure are maintained by normal systems for the plant status.

i 1

4

(***** CATEGORY 00 CONTINUED ON NEXT PAGE *****) '

I. _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ . _ . _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

t .

I e2__epblNISIB811ME_BBgCEgyBES4 _CQNDIIIQNS 2 _8Np_61dlIGIlgNg PAGE 32

{

l OUESTION G .17 - (1.00) e Which of the following correctly describes the TS definition .}

l of an " Instrument Functional Test."

a. The adjustment of an instrument signal' output so that i it corresponds, within acceptable range and accuracy, to a known value of the parameter which the instrument monitors,

b. The 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 i 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 the design intent.

QUESTION 9.10 (1.00)

The release rate of radioactive liquid effluents, excluding tritium and noble gases, shall not exceed _____ during any calender quarter.

E

a. 1E-7 Ci

b. 2E-4 C1

c. SE-1 Ci r

d. 2E+1 Ci i

)

1 l

(***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)

l.

l Dz__8DdINI@ISSIlyg_EBgCEDUBES2_CgNQ1IIONg3_6NQ_L1dII6IlgNS PAGE 33 i

OUESTION 8.19 (1.00)

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 statements most accurately describes the allowances / limitations as detailed in the Browns Ferry Technical Specifications and OSIL #35, " Secondary Containment Control."

a. Secondary Containment integrity is not required under the  ;

existing 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, sta-tioning a public safety officer allows both the inner and outer air lock doors to be open during equipment 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 doors at all times they are open.

NOTE: APPLICADLE TS'S ARE ENCLOSED FOR REFERENCE P

(***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)

t

e q ac__8DdlylSIBGIlyE_EBQGEDUSSSx_QgNQlligN@2_6ND_ Lid 118IlgNS PAGE 34 QUESTION 9.20 (1.00)

Units 1 and 3 are operating at 75% and-100% rated thermal power, respec- _

tively. Unit 2 is in Cold Shutdawn. During per. forma..ce of SI's (4.7.D) on SBGTS Trains A, B, and C, the following deficienctes are noted:

SDGTS Dlower / Fan A is INOP SBGTS Train C Decay Heat Damper (FCO 65-52) is STUCK OPEN Maintenance estimates 3 to 4 days to accomplish repairs.

Which one of the following statements most accurately describes the allowances and/or limitations imposed by the Drowns Ferry Technical Specifications.

I NOTE: APPLICABLE TS'S ARE 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 i and 3 in et least Hot Standby within sin 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 />. I

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 />.

l i

r

(**MM* END OF. CATEGORY 00 *****)

(************* END OF EX AMINATION * * * * ** * * * ** ** * * )

l . .

Qu__IBEQBy_QE_UQQLEQB_EQWEB_E(6UI_QEEB611QUx_E6Q1QS3_6NQ PAGE 35 ISEBd99100dlGS ANSWERS -- DROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

I i

ANSWER 5.01 (1.00)

a. 22222222222222222 E

REFERENCE General Electric NEDE 24010 i EIH: GPNT, STA Training Manual, Section 9; DFNP HTFF, Chapter 9, L.O. O.2 -

i DWR K/A 293009 K1.36 (2.8/3.4), K1.37 (2.6/3.3)

I ANSWER 5.02 (1.00)  ;

d REFERENCE Mollier Diagram (Steam Tables)

DFNP HTFF, Chaptor 4, L.O. 6.2 DWR K/A 293004 K1.09 (1.7/1.0)

L i

l

et__IMEQBY_gE_Nyg6E88_EQWEB_EL8MI_QEEB811gN z _ELUIDSi_80Q PAGE 36 ISE8090100dicg ANSWERS -- DROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 5.03 (1.50)

a. c (1.0)

b. (Excessive voiding in the bypass region resulting in) unreliable LPRM readings (0.5) (1/2 credit for LPRM overheating) (0.5) .

REFERENCE  !

General Electric Heat Transfer b Fluid Flow, Chapter 8 DFNP: HTFF, Chapter 8, L.O.'s 9.4 & 9.5 I

BWR K/A 293000 K1.32 (2.5/2.6), K1.33 (2.4/2.6) 215005 K5.02 (2.7/2.0)

ANSWER 5.04 (2.00) i

a. P -P (0.5) act sat ,

b. (1) INCREASE (More Subcooling at the pump suction) f (2) DECREASE (Reduced pronauro at the oyo of the pump ronults in being closer to caturatio prescure)

(3) INCREASE (Further to saturation temperaturo and increased l density causing less static head) (3 0 0.5 each)

REFERENCE General Electric Hoat Transfer and Fluid Flow, Chapter 63  ;

i DFNP HTFF, Cnaptor 6, L.O.'s 7.7, 8.1, & G.2 i t

r BWR K/A 293006 K1.10 (2.7/2.0)

L_.___.._____._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . _ . . . . _ _ _ _ . _ _ . _ _ _ . _ _ _ _ . _ _ . _ _ . _ _ . _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ . _

Uc__IUEQBY_QE_UUC6668_EQWEB_ELQUI_QEE88IlgNi_E(Q1932_6NQ PAGE 37 IUEBdggyd8 digs

{ . ANSWERS -- DROWNS FERRY 1, 263 -86/12/01-BROCKMAN, K.

i e

i i

1 j- ANSWER 5.05 (1.50)

a. The decrease is due to the (decay of Xe-135 and the) increased burn-  !

out (which is immediately seen by the [Xe]), being much greator than the j Xe production, (which is time-delayed from the lower power.)

{ b. The flux will redistribute from outside (and higher) peaked and  ;

center (and lower) depressed to center peaked and outside depressed.

{ (0.25) This in because the Xe concentration shifts from center peaked I to more peripherally peaked due to power production distribution. (0.25) l c. Una caution when pulling peripheral roda (0.25) due to tho 4

increased rod worths (increased thermal flux) (0.25). (0.5 each) l i  !

REFERCNCF General Electric Reactor Theory, Chapter 6 [

j DFNP: Rx Theory, Chapter 6, L.O. 2 A

l l

I DWR K/A: 292006 K1.07 (3.2/3.2), K1.00 (2.0/3.0), K1.10 (2.9/2.9) '

! & K1.14 (3.1/3.3) i l

i i

4 ANSWER 5.06 (1.50? l 1

{

a. Thormal Stratification (107 dog F Dolta T)

I j b. Establit.h Natural Circulation (Raiso level to 3rd stagu

] of tho separatorn ~60")

i

c. Loan of Shut down Cooling (due to prousure interlocks)

! (- OR - Incroano DDC to roverno stratification 0 0.25) (0.5 oach) l l

l HEFERENCE l General Electric Heat Transfer and Fluid Flow, Chapter O DFNP: HTFF, Chaptor G, L.O. log GOI 100-12, Soction IV

]

i i

I, l.

Ur._ ldEQBLQE_UUCLEGB_CQUEB__EL6UI_QEEB8Ilgu1_,ELylgG1_,00D PAGE 30 ISEEUQQYUQUlGS

^%

ANSWERS --- DROWNS FERRY 1, 2&3 -06/12/01-DROCKMAN, K.

DWR K/A: 293000 K1.35 - 1.37 (3.1/3.3), K1.22 & 1.23 (2.9/2.0:0 ANCWER 5.07 (2.00)

a. Pellet to Clad contact (due to the fuel pellet enpansion) (0.5)

b. (1) Finnion Product gauces increase fuel rod gas pressure and the strens on clad during dryout.

(2) F i ts s i o n Product gannen have a lower heat transfer coetficient than lielium, thuu reducing the heat transfer rate.

(3) Crud buildup on the fuel clad uuriace with core age reduces the heat tranufer and increases the atored energy.

( <l ) Fi t r.i on product decay energy increanen with unponure and can increane the Zircalay-water reaction rate near 2200 deq F.

(3 0 0.5 each)

REFERENCE liener al El c c t. r i c l le a t. Tr ann f er and F1uld F1ow, Chapte r 9 l'i NF': HTfF, Chapter '7 , l..U. 3.6 IMR l'/ A; ./ ,' 0 0 ' / f . l .14 ( '. ' . 6 / ', ' . / )

l l

l

m )

Dz__IUEQBY_QE_NUGLEQB_EQWEB_ELONI_QEE8611gN2_ELUlQdi_6NQ PAGE 39 ISEBt!ggyNQdlGS ANSWERS -- DROWNS FERRY 1, 2h3 -86/12/01-BROCKMAN, K.

ANSWER 5.08 (1.50)

a. The negative reactivity added by the increased voids generated by the rod withdrawal is greater than the positivo reactivity added by the reduced rod absorbtion. (1.0)

b. (4) (1.0)

REFERENCE General Electric Ruactor Theory, Chapter 5 DFNP: Rn rhoory, Chaptor 5, L.O. 3 DWR K/A 292000 K1.04 (3.0/3.0)

ANGWER 5.09 (1.75) l

a. Period a (Deta - Rho) / (Lambda

• Rho) ,

= (0.0072 - 0.0006) / (0.1

• O.0006) = 110 seconds (0.5) ,

b. Period = 1.443

• DT -> DT = Perfod / 1.443 = 110 / 1.443

= 76.23 oucands (Error carrios f orward f rom (a)) (0.5)

c. Period will bu GHORTER (0.25) This in duo to the fact that the Data for EOC is shortur (* O.0050, due to the increased Pu contribution). (0.5) ( 0. '75 )

l REFERENCE General Electric Reactor Thoory, Chaptor 3 DFNP Rn Theory, Chaptor 3, L.O.'o 3.6, 4.6, & 5.9 DWR K/A 292003 M1.00 (2.5/2.0), K1.09 (2.5/2.6)

I l i

l l

, I l

! Dx__IUE98Y_9E_U9GLEGB_E9 WEB _E60UI_9EEBBIIQN2 _ELylpS4_6NQ PAGE 40 l IdEBd9DYN6dlGS ANSWERS -- BROWNS FERRY 1, 263 -86/12/01-BROCKMAN, K.

ANSWER 5.10 (2.00)

a. Reactor Pressure decreases initially (0.125) due te HPCI steam flow (0.125). Long term prousure in higher (0.125) due to the higher final reactor power (0.125)

b. Reactor Powar decreases initially (0.125) due to the additional voiding from the pressure decreasu (0.125). Long term power is highur due to the colder water which ontors the coro (0.125).

c. Feedwater flow decreases initially (0.125) au FWLCS follows t the decroase in steam flow (and increano in level) (0.125). Long  ;

term feedwater flow is less (0.125) to componnate for the input t of tho HPCI injoction (0.125).  !

d. Vennel level increaseu initially (0.125) due to the swell caunod ,

by the prensurn decrease (0.125). Long term lovel utabilizes at a higher value to compensato for the ntoam flow /fced flow miumatch (i n ,

the FWLCS circuitry) (0.125). (0.5 nach)

REFERENCE 1+ NP Transient Analysin L/P, Transient # 20; OPL171.055, L.O. "A" ,

DWR U/A 206000 K2.07 (3.9/4.3) 295014 AA1.07 (4.0/4.1) l i

l

{

i

[

t i

\

Et__IBEQ8Y_9E_UUCLE88_E9WE8_E68BI_9EE88I1Gui_ELUIDS2 _800; PAGE 41 ISE800QXUQUlgS ANSWERS -- DROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 5.11 (3.00)

a. Recirculation Pump Runback (From < 20% FW Flow) (0.5)

b. EHC controlling reactor pressure (at 920 poig) (0.5)

c. EHC controlling reactor pressurn (at 920 psig) (0.5)

d. Increase in the auction of the Rocirc Pump (from Downcomor) (0.5) l

o. Reactor Scram (APRM Flow Bicued 0 0.66Mw + 54%) (0.5)

f. K-f is utilized to locroano the steady-stato MCPR limits at reduced flows (FLCPR Program). (0.5)

REFERENCE DFNP Tranutent Analyniu LP, Transient # 5, L.O. A&D DWR h/A: 202001 A2.05 (4.0/4.0) 293009 K1.26 (3.3/3.3) K1.20 (3.5/3.5)

ANSWER 5.12 (1.00)

This could renuit in a POGITIVE MODERATON TEMPERATURE COEFFICIENT (0.5).

The ponitivo roactivity addition by a dacruano in the density of thri moderator and the Doron (on a huat up) can bn grnator (0.25) than the negative roactivity addition from the incruanud leakago, reranonanco capturn, and control rod worth.(0.25)

REFERENCE liFNP OPL171.039; L.O. "G" ,

DWR k/A 2110d0 K5.01 (2.7/2.9)

a 4

Ut__IUE0BY_0E_NyGLE88_EQWEB_E66UI_QEgB811QN1 _ELU1QS1_8NQ. PAGE 42 ISEBdgQYU8d1GS ANSWERS ~~ DROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 5.13 (1.50)

a. Doppler Coefficient (0.25) The rod insertion will be first "seen" by the adjacent fuel rods. (The immediato flux changen will cause fuel temperature to change before the effects are seen in moderator temperature or void fraction.) (0.5)

b. Void Coefficient (0.25) The MSIV Closure will immediately cause a pressure increase which will result in a change in the void fraction.

(This change will be "seen" before any changes in moderator temperature or fuel temperature.) (0.5)

REFERENCE General Electric Reactor Theory, Chapter 4 DFNP Rx Theory, Chapter 4, L.O.'s 4.3 & 6.3 DWR K/A 292004 K1.00 (2.2/2.4), K1.11 (2.5/2.6)

ANSWER 5.14 (1.50)

a. Higher I

b. Lower i l

c. Lower (0.5 each) 1 1

1 REFERENCE General Electric Heat Transfer and Fluid Flow, Chaptor 7 l l

EIH: L-RO-667, p 10 '

DFNP R): Hunt Dalance LP; RO 05/03/05 <

HTFF, Chaptor 7, L.O.*s 5 & 6.4; HTFF, Chapter 8, L.O. 5.1 l I

1 UWR U/A 293007 K1.11 (2.6/3.1), K1.13 (2.3/2.9) l l

l 1 )

t l I .

l

5. THEORY OF NUCLEAR POWER PLANT OPERATION 3 _ELVIQS2 _6ND PAGE. 43 IbEBdgpyd8d1CQ ANSWERS -- DROWNS FERRY.1,'2&3 -86/12/01-BROCKMAN, K.

ANSWER 5.15 (1.00)

Cooldown Limit = 90 deg F/ hour (0.25)

Moderator Coefficient = - 1 E-4 Rho /dag F (0.20)

Rho = (1 - K-eff)/K-eff *= K-eff (0.20)

Time = E(Rho / Mod Coeff)/Cooldown Limit] (0.25)

= 1.11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> (67.3 min) _(0.10)

REFERENCE General Electric Reactor Theory, Chapter 4 DFNP Rx Theory, Chapter 4, L.O. 2.4 DWR K/A: 292000 K1.01 (3.8/3.9) 292004 K1.04 (3.3/3.4), K1.11 (3.7/3.8) 3 ANSWER 5.16 (1.25)

a. Moro r

b. Decrease

c. Loss ,

d. Increase

o. Incruasu (0.25 each)

REFERENCE Gunural Electric Reactor Theory, Chapter 5 DFNP Rx Thoory, Chapter 5, L.O. 2.4

r.

! 5- THE98Y_9E_UUGLE68_E9 WEB _EL8NI_QEEB811gN 3 _ELylDS2_60Q PAGE 44 IBEBU9DYN8dlGS l

(_ ANSWERS -- DROWNS FERRY 1, 2t<3 -G6/12/01-BROCKMAN, K.

i

~

l DWR K/Ai 292005 K1.09 (2.5/2.6), kl.10 (2.8/3.3) l I

l

\

l i

l

( . .

y 4

6. ELONI_SY@IEUS_DEglGN 2 _CQNIBQL 1 _8ND_lNSIBQUENI8IlQN PAGE 45 ANSWERS -- BROWNS FERRY 1, 2L3 -86/12/01-BROCKMAN,1(.

ANSWER 6.01 (1.00)

(1) A (LIT 3-53)

-(2) Low (0.5 each)

REFERENCE GGNS 'OP-B21-501 BFNP: OPL171.OO3, L.O. "J"; OPL171.012, L.O. "E.3" ARP 9-5 --XA-55-5A-8, XA-55-5A-30, XA-55-5D-4, XA-55-58-5 .,

BWR K/A: 259002 K5.01 (3.1/3.2) K6.5 (3.5/3.5)

ANSWER 6.02 (1.50)

a. O seconds (+3,-0) to 30 seconds (+6,-6) (0.5)

b. The rate of the power increase determines the time delay (f etst ri se = short del ay) . -UR- Ihe absolute value of the power increase (P(f) - P (i ) ) '(0.5)

c. DOTH OPEN AND CLOSEE (0.5)

REFERENCE DFNP: OPL171.022,Section X.B.6 & .7; L.D. F&K l

DWR K/A 215005 K4.02 (3.7/3.7) l l

u

=

n

./

6t__E66NI_SY@lEMS_DEgl@N 2_QONIBQL1_@ND_iN@lBQUENI@IlON- PAGE- 46' ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BRDCKMAN, K.

ANSWER 6.03' -(1.50)

a. (1) b ---

(2) c (3) a (0.25 each)

b. Fire pump (s) auto start _

Raw Water Storage Tanks Isolate Raw Service Water Pumps Shutdown (by interlock). (0.25 each)

REFERENCE DFNP OPL171.049, Sections X.B.r.(2) & .(3).& .D.2; L.D. E DWR K/A: 286000 K5.06 (2.6/2.7), K5.07 (2.6/2.7), K4.01 (3.4/3.6)

K4.02 (3.3/3.5) i l

t I

l 6t__EL6NI_gY@IEd@_DEgigN2_GQNIBQL2 _8NQ_INQIBQUENIBIlQN PAGE 47-ANSWERS--- BROWNS FERRY 1,.2&3 -86/12/01-BROCKMAN, K.

ANSWER ,

6.04 (1.50)

a. Curve #1 -

.(0.5)

b. (1) Any platform hoist (3) loaded or the' grapple not fully.

up and-all rods not full in with the platform near or1over the core. (0.5)

(2) Platform near or over the core and not in the refueling mode (0.25)'

(3) More than one rod withdrawn with the platform near or over the core in the refueling mode. (0.25)

REFERENCE DFNP OPL171.053, Sections X.B.1.m.(4) & . C; L.O.'s D & E i

a DWR K/A: 234000 K4.xx ANSWER 6.05 (1.50)

a. f>reclude excessive radiation levels (ALARA) (0.25 each)

b. (1) Manual (2) Air (0.25 each)

c. By the RED color area appearing on the tube of the Channel Handling Tool. (0.5)

REFERENCE Stearns/ Rogers Refueling Platform Technical Manual.

DFNP: GOI 100-3 BWR K/A: 23400, K

1 .

6s__ELONI_g1SIEU@_QE@l@Nz _CQNIBQL2_6NQ_lNSIBQdENI611QU_ PAGE~ 48 l

ANSWERS -- DROWNS' FERRY 1, 2843 -86/12/01-BROCKMAN, K.

1 ANSWER 6.06 (2.00) 4

a. (2)  :

b. (2) (1.0.each) l

+

REFERENCE

, BFNP: ARP 9-4, XA-55-4A; OPL171.007 L

1 BWR K/A: 202001 K1.01 (3.6/3.7) 293008 K1.26 (2.9/3.1), K1.27 (3.6/3.7) ,

I i i 1 ANSWER 6.07 (1.00) l C

REFERENCE BFNP: OPL171.018; OI-65, p 9; Technical Specifications,.L.O. 62 I

i j

i BWR K/A: 261000 K3.02 (3.6/3.9), A4.01 (3.2/4.0).

272000 K3.02 (3.1/3.8)-

j i

1 i

h A

y + - ,- - . - - - ,e s ,n- . - - - -- -, . - , - . , - - = p m

i,

6. E6801_SYSIEd@_QEg1GNx_CQNIBQL1_8NQ_INSIBQUENI8Ilgd PAGE 49 ANSWERS'-- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 6.08 ( .50)

Not having the governor at the lowest setting can result in a mechanical overspeed trip of the diesel.

REFERENCE BFNP: LER 86-26 (7/28/86 - EDG "D")

DWR K/A: PWG (3.2/3.6)

ANSWER 6.09 (1.50)

a. (Greater than 40% mistaatch between the generator output and the-turbine power) as sensed by generator stator amps and turbine cross-over pressure. (0.5)

b. (1) TCV's Trip Closed (f ast-acting solenoids)

(2) Bypass Valves Open (3) EHC Load Selector runs back (4) Control Valves partially reopen @ < 40% mismatch.

(5) CIV's throttle to control turbine speed-@ 1800 rpm (0.20 each)

REFERENCE BFNP: OPL171.010, L.O.'s "J" & "K"'

BWR K/A: 245000 K3.01 (3.4/3.7)

I

q 6t_ PLANT _SYSIEM@_DESIGNz_GQNTRQL 2 _AND_INSIRtJMENIATlQN PAGE 50' ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 6.10 (2.25)

a. Reactor Water Level (0.25) GTE - 39" (0.25)

Drywell Pressure (0.25) GTE + 1.96 psig (0.25)

SELECT - RESET Switch (0.25) in SELECT.(0.25)

b. Level can be bypassed Drywell Pressure can not be bypassed Select Switch can not be bypassed (0.25 each)

REFERENCE DFNP: OPL171.044, L. O. 's "E.10" & "E.11" BWR K/A: 226001 K4.03 (2.9/3.1) 203000 K4.10 (3.9/4.1)

ANSWER 6.11 (1.50)

a. Yellow light above the Core Spray pump switch (on Panel 9-3 ) '.

b. Initiating conditions cleared (GT--114.5" RVL & LT 2.45 psig DWP).

Reset. Core Spray Initiation signal.- (0.~5 each)

REFERENCE BFNP: OPL171.045, L. O. 's "F" & "H" BWR-K/A: 209001 K2.03 ( 2. 9 /3.1 ) ', K4.08 (3. 8/.4. O) , - A3. 02 -(3. 8/3. 7) we

4t__EL8MI_SX@IENS_ DESIGN1 _CONIBOL1 _6ND_INSIBUDENI6TlQN PAGE .51 ANSWERS -- BROWNS' FERRY'1, 2&3

-86/12/01-BROCKMAN, K.

T

. ANSWER 6.12 (1.00)

a. DEPRESSURIZE (DWP is a seal-in signal)

b. NOT DEPRESSURIZE (RVL is not'a seal-in signal) (0.5 each)-

REFERENCE BFNP: - OPL171.043, L.O. "E" l

i' i

BWR K/A: 218000 K2.01 (3.1/3.3), K4.01 (3.7/3.9), K4.03 ( 3. 8 /4.' O )

K5.01 (3.8/3.8) -

i i

ANSWER 6.13 (1.50)

a. To prevent creation of a vacuum in the tailpipe of the SRV (f rom -

i condensation) following an actuation and reseat. Torus water would i be drawn into the tailpipe and a subsequent actuation could result l,

in excessive forces placed on the downcomer. (0.5)

b. (1) Could result in uncondensed steam release to the

{ primary containment.

(2) Could result in high radioactivity releases due to j the loss of the Suppression Pool " scrubbing". (0.5 each)  ;

Other responses, as appropriate.

(e.g., Increase in Drywell Pressure / Temperature)  ;

REFERENCE ,

BFNP: OPL171.OO9, L.O. "B" '

i

~ BWR K/A: 239002 ' f04.03 (3.4/3.6), A2.01 (3.0/3.3) i I

,i i

,~ ..--r. -

.m ..umr~.,,w,,,,,w,w .u ..s.,%i_%,,

, , - .,-,..7 % ,,,r., w J.g,.,. -

p se, ..

61__E60NI_SYSIEUS_DEglGN1 _CQNIBQL2_6NQ_INSIBydENI6IlON LPAGE -ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

L ANSWER 6.14 (1.00)

C

REFERENCE BSEP

'RTN-033, 012; SD 26.2; SSM 19-2/3-B EIH: L-RQ-705, pp 18, 19; GPNT, Vol. .VII, Chapter 9.4  :

BFNP: Simulator Malfunctions 8, 108; Operational Transient LP, #15;

OPL171.014, L.O. "E.1" a

BWR K/A': 259002 K6.05 (3.5/3.7), A2.01 (3.4/3.9) 1 ANSWER 6.15 (2.00)

1. e I ii. d REFERENCE 7 BFNP: Operational Transient LP, #20; OPL171.012, L.O. "E" EIH L-RQ-726 BSEP: RTN 026; HD 17-2/3-B, Section 3.2 f

~

! BWR K/A: 259002 K6.05 (3.5/3.5) 1 1

f. ANSWER' 6.16 (2.00)

a. SATISFACTORY (0.5) - The Test Failed light indicates that.:the GNC test sequence has halted because of a failed comparison (which is-the j desired result).(0.5)

.i b .- Illuminates backlighting for companion rods (0.5), regardless of_their sensed position. (0.5)

REFERENCE BFNP: .OPL171.025, L.D. "D" 1__ . _ _ _ _ . ___-__._______1____

6,__eL8NI_SYSIEMS_ DESIGN2 _CQNIBQL1 _QND_INSIBUMENI@IlgN PAGE 53 ANSWERS -- BROWNS FERRY 1,'2&3 -86/12/01-BROCKMAN, K.

ANSWER 6.17 (1.00)

During flushing and Shutdown Cooling operations, the RHR Minimum Flow Valve could open diverting reactor coolant to the Torus. (0.5)

(OI-74 has been revised to) require the the Minimum Flwa Valve Handswitch be placed in the BYPASS position prior to flushing and Shutdown Cooling operations. (0.5)

REFERENCE BFNP: OPL171.044; OI-74: General Electric Systems Review on RHR BWR K/A: 205000 K1.08 (3.9/3.9), K3.02 (3.2/3.3), K4.07 (2.7/2.8)

K5.02 (2.8/2.9), A2.10 (2.9/2.9), A4.05 (3.2/3.2)

ANSWER 6.18 (2.00)

(1) Diesel started and at rated speed.

(2) All other supply breakers to the 4160 volt bus Open.

(3) No supply breaker lockout (EDG).

(4) Undervoltage on the 4160 valt bus. (0.5 each)

REFERENCE DFNP: OPL171.038; L.O. "E" 1

BWR K/A: 269000 K1.01 (3.8/4.1), K5.05 (3.4/3.4) l l

l 1

I

\

Zt__EBOGEDUBES_;_NQBd861_8BNQBd861_EME8GENQy_8ND: PAGE 54 88D106991G86_GQNIBQL ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 7.01 (1.00)

Venting the suppressior. chamber first takes advantage of the " pool scrubbing effect" (0.5) 'to help minimize the radiation released (0.5).

REFERENCE BFNP: EDI-2, PC/P-5, L.O. "B.6" BWR K/A: 259024 EK3.04 (3.7/4.1), EK3.07 (3.5/4.0)

EA2.03 (3.8/3.8), EA2.04 (3.9/3.9), EA2.08 (3.6/4.0)

ANSWER 7.02 (1.50)

a. From 14.6 - 11.5 Feet Compensates for reduced heat capacity of the Suppressicn Pool due the reduced level (based upon HCTL assumption of level GTE 14.6 ft) (0.5)

Less than 11.5 Feet Pr(pcludes incomplete steam condensation (0.3) due to downcomers un4overing (0.1) given a primary system break (0.1). (0.35 credit fod stating only that the Containment overpressurires) (0.5)

b. Emergency Reactor Depressuri: tion is required if you are out of the safe area (Contingency 2). (0.5)

REFERENCE BFNP SP/L, p 2; OPL174.657; L.O. B.7 ANSWEF. 7.03 (1.00)

a. Permiss'.ble for-floating items (to prevent sinking) (0,5)-

b. FALSE (0.5) s_. a

c- .q

-Z1__ESQQEDQBES_ _NQBM862_6BNQBd@b1_EME8@ENQY_6ND PAGE 55 88 Dig 6QQ1G66_GQNI696

. ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

REFERENCE BFNP: SP 7.7, Sections 4.1 & 4.2.1(4); L.O. "B" ANSWER 7.04 (1.00) ao 3358 MWt (+0,-17 MWt)

b. 3293 MWt (+0,-16 MWt) (0.5 each)

REFERENCE BFNP:-GOI 100-1,Section IV.B.10; L.O.'s "F" & "G" BWR K/A: PWG #5 - Safety Limits (3.3/3.9)

ANSWER 7.05 (1.00)

a. TRUE

b. TRUE (0.5 each)

REFERENCE BFNP: GOI 100-1, Sections A.12 & A.22 BWR K/A: 203000 K1.14 (2.4/2.6),' SWG #5 (3.5/4.4), SWG #10 ( 3.~7 /3. 9 )

I 1

7

-Zs__EBQGEDQBES_ _UQBd862_8ENQBM862_EMEBGENGY_8NQ. PAGE. 56'

B8D196901G86_G901896-ANSWERS -- BROWNS FERRY 1,.2&3 -86/12/01-BROCKMAN, K.

ANSWER- 7.06 (2.50) 1)- Rx Water Level (0.25).> 11 (0.25) (0.5)

2) Rx Pressure ' (0. 25) > 1043 # (0.25) (0.5)

3) D/W Pressure (0.25) > 2.45 # (0.25) (0.5)

4) Rx Scram required and Power > 3%, or indeterminable ( 0. 5 ) ~

5) MSIV Isolation Required (0. 5) -

REFERENCE BWROG EDI Users Guide BFNP: EDI-2, Section 2.1; EDI L/P, L.O. A ANSWER 7.07 (2.50)

a. (1) Stop all Fuel Handling activities (and evacuate ~the Refueling Floor)

(2) Manually Scram the Reactor (3) Stop the CRD Pump (s) (0.25) and Isolate RWCU (0.25) (0.5 each)

b. (4) (1.0)

REFERENCE BFNP: GOI 100-3,Section VI.B; L.O. "D"

~ ANSWER 7.09 -(1.00)

C REFERENCE BFNP: GOI 100-3, L.O. "S.5"

? i)

A

Z___EBQGEDUBES_

_N9Bd861.8BNQBd8L _EUEBGENQY_8NQ_ 2 -PAGE 57 B8DlQLQGIGO6~QQUIBQL

. ANSWERS ~~- BR' OWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K..

h ANSWER 7.09 (1.00) b REFERENCE BFNP: IP-24, Section 3.3.1 ANSWER 7.10 (1.00)

a. Office Service Building (OSB/PM Ofc/Adm Bldg / Plt Ofc Bldg)

~

b. (1) 75 R (10. 25) ; (2) 25 R (0.25) (O'.5 each)

REFERENCE BFNP: IP-15 & IP-29; L.O. VI.E; RCI-1, L.O. "

B" ANSWER 7.11 (1.00)

a. 150 mrem

b. 300 mrem

.c. 50 % ,

d._ 1 Rem / hour (All +0%,-10%) -(0.25 each)

-REFERENCE BFNP: RCI-9,Section IX.E; L.O. "F"

-BWR K/A: 294001 K1.03 (3.1/3.8) u.um.- __---m_ m_---_.m_.____--__.___mm.___-__.m.__.__

Zz__E80GEQQBEQ_ _NQBd@(x_8BNQBd861_EdgBQENQY_8NQ. PAGE 58 88 Dig 6QQ1Q66_QQNIBQL ANSWERS -- BROWNS FERRY 1, 263 -86/12/01-BROCKMAN, K.

ANSWER '7.12 (1.50)

a. Increase (0.25) Formerly extracted steam now passes through the turbine (giving net work / watts out) (0.5) (0.75)

b. Decrease (0.25) Extraction steam to the next' higher pressure heater will increase (0.5) (because the feedwater temperature rise across the heater is greater - thus, less energy is available to the turbine) (0.75 ,

'i REFERENCE BFNP: DI-2,Section II.C.* & .2, L.O.'s "A" & "B"; OPL171.011,' Section E BWR K/A: 259001 A2.02 (3.1/3. 3) (

ANSWER 7.13 (1.50)

a. Complete Loss of Offsite Power (LOSP)

b. Accident Signal present (LOCA on any unit)

c. The dedicated diesel generator (s) cannot supply sufficient .

l power for all of the required loads. (0.5 each)

REFERENCE BFNP: OI-57, Scction V.E.4.J(5); L.O. "C" l

)

BWR K/A: 264000 K5.05 (3.4/3.5), A2.01 (3.5/3.6)

{

l l

"7. PROCEDURES - NORMALi_ABNgEM6L 2_EMERGENGy_ANQ PAGE 59

E8D196991G86_G9NIEQL i -

ANSWERS -- BROWNS FERRY 1, 2t<3 -86/12/01-BROCKMAN, K.

ANSWER 7.14- (1,00) i

a. 47 psig (+3, -2 psig)'

b. Possibility of overloading the 3C EDG (0.35) during a

simultaneous LOCA w/LOSP-(0.15). (0.5 nach) i ~ REFERENCE

BFNP

0I-24, Secti ons ' II. B 8< III.E, L.O. "A" 1

l '

l i-l J BWR K/A: 264000 K3.03 (4.1/4.2) 286000 SWG #10 (3.2/3.5) 4 f

I ANSWER 7.15 (1.00) i-d REFERENCE i BFNP: 01-66, pp 36 - 37, L.O. "B" & "K" r

l 4

BWR K/A: 271000 K4.04 (3.3/3.6), K5.04 (2.9/3.1), PWG.10 (3.1/3.2)

I A1.13 (3.2/3.7), A2.06 (3.5/3.9), A4.09 (3.3/3.2) i r

i 4

t 4

- . - - . _ _ - . . _ _ _ _ _ -- - - - _ - _.---- - ___-.__ .--E_-_---___--_-______--_ __ -----e

m

~-

~}

r

~Zz__EBOCEDURES'- NORMAL2_'ADNgBM8L1 _EMgBgENCY AND' PAGE: :60~-

BBD196901CeL_ggNIBQ6-

. ANSWERS.- BROWNS FERRY _1, 2h3 -86/12/01-BROCkMAN, K.

ANSWER 7.16 (1.00)

1) Evacuate the Refuel Zone

-2) Place the other loop of Shutdown Cooling in service immediately, (without flushing). (0.5'each)-

REFERENCE BFNP: GOI-100-12,'Section'III.E; GOI-100-12 L/P, L.O. D BWR K/A: 205000 K3.03 (3.8/3.9), K3.05 (2.6/2.7)-

294001 K1.03 (3.7/3.8)

ANSWER 7.17 (1.00)

Heat removal from the reactor sufficient-to restore and maintain the peak fuel clad temperature (PCT) LTE 2200 deg F. - (Heat up is  ;

allowed as long as PCT will nowhere exceed 2200 deg F). (1.0) a REFERENCE BWROG EDI Users Guide BFNP: OPL174.657 (EDI L/P), L.O. D i

l l

1

-BWR K/A: 295031' Ekl.01 (4.6/4.7)

ANSWER 7.10 (1.00) d ,

.t I

t e er ' - , --en <,,y--r, - s- -

w, e-m., mer

7. PROCEDURES NORMALx_8QtjgBM8Lx_EMEBQENQY ANQ PAGE 61-

.B891960G1G66_GQNIBQL.

. ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

REFERENCE '

BWROG EDI Users Guide,/Section.2.1

-DFNP: EDI-1, Sections 2.1 & 3.3; EDI-2,-Section 2.1; EOI.L/P, L.O. I. >

ANSWER 7.19 (1.00).

( ADS is disabled as long as a Reactor Shutdown is not assured, or may be contingent upon SLC) to-prevent the injection of ECCS (large volumes of relatively. cold water) which may result in a reactor power excursion leading to substantial core damage.. (1.0)

REFERENCE BWROG Users Guide, EDI RC/L, p3 BFNP: OP' .174. 6',7 (EDI L/P), C-5, L.O. "B.'6" BWR K/A: 29503' EK1.03 (4.2/4.4), EK1.06 (4.0/4.2), SWG #7_(3.7/3.9)

ANSWER 7.20 (1.00)

1) Open knife blade switches (CS-1, 2, & 3)

2) Tag each with a SE Hold Tag (0.5 each)

' REFERENCE DFNP: OI-44,Section III.G.1.f

C q

7. __P8QQEQUEES_ _NQBd8L1 _8@NQBMAL1 _EdgBGENQY_8NQ PAGE 62 88DIQLQGlG86_QQNIBQL ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 7.21 (2.50)

1) . Suppression Pool Temperature (0.25) > 95 deg F (0.25)

2) Drywell Temperature (0.25) > 160 deg F (0.25)

3) D'rywell Pressure (0.25) >~2.45 # (0.25)

4) Suppression Pool Level (0.25) > -1 (0.25) I

5) Suppression Pool Level (0.25) < ~6.25 " (0.25) (0.5 each)

REFERENCE BWROG EDI Users Guide DFNP: EDI-2, Section 2.1; EDI L/P, L.O. B

Ot__0DMINISIBOIIME_BBQCEQUBES 2 _CQNQlIlQNS 1 _QNQ_LIMIIGIlgNQ PAGE 63 ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

i ANSWER 8.01 (1.00)

a. It shall be cycled by the motor operator (O.$)(prior to being declared operational). (0.5)

b. Torque switch could become activated before the valve is off its seat (improper operation of the opening circuitry). (0.5)

REFERENCE BFNP: OSIL No. 47, p 2; L.D. "T" BWR K/A: 291001 K1.07 (3.4/3.4)

ANSWER 8.02 (1.50)

EMERGENCY - Action is required to prevent imminent major equipment damage or to protect personnel from any imminent threat of bodily injury (0.35/O.15)

IMMEDIATE ATTENTION - Work is 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 /> (, the next schedu'.ed workday, or upon completion of necessary technical evaluations or material procurement) (0.5)

ROUTINE MAINTENANCE - Work is to be performed as manpower and circ-umstances permit. (0.5)

REFERENCE BFNP: SP tt 7. 6, Section 4.12 - .14; L.O. "A" l

l l

BWR K/A: 294001 K1.02 (3.9/4.b) '

l l

1

g - - - -

y B. ADdlNISIB8IlyE_EBQQEQUBES 2 _QgNplIlgN33_8ND_61dlI611gNS PAGE '64

' ANSWERS -- BROWNS FERRY 1,.2&3 -86/12/01-BROCKMAN, K.

' ANSWER 8.03 (2.00)

a. 2

b. 1

c. 4

d. 3 (0.5 each)

REFERENCE BFNP: SP 14.25, Sections 4.4 - 4.7 DWR K/A: 294001 K1.02 (3.9/4.5) t ANSWER 8.04 (1.00)

Are (0.3) (TS (3.7.A.4.b) allows one Vacuum Breaker to be Not ,

Fully Closed, as long as it is less than~3 deg open. The red light being OFF indicates that the breaker is less than 3 deg open; the green light being ON indicates that it is less than 80 deg open.) These are the NORMAL. CLOSED indications for the Vacuum )

Dreakers (0.7). (1.0)

REFERENCE BFNP TS 3.7.A.4.b (and Bases); L.O.'s 51 L 55 i

DWR K/A: 223001 K5.01 ( - / 3.3), K6.09. ( - /3.6)

t et__0D01NISIB8IlyE_ESQggDUBE@z_GQUDlIlQN@1_8ND_LIMlI8IlgN@. PAGE 65 i ' ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 8.05 (2.00) i a. (1) The cause of the reactor scram has been identified and is understood.

(2) There should be reasonable assurance that the unit will not scram again due to the same cause.

' ~

(3) The responses of safety related equipment are understoo'd and acceptable for safe operation. (2 O O.5 each)

6. Plant Manager (0.5) o i c. (1) Scram Committee will investigate the scram.

f

(2) PORC will make the recommendation on unit restart j to the Plant Manager. (2 0 0.25 each REFERENCE j BFNP

Standard Practice 12.8, Section 6; L.O. C, D, &E I

i 1

I t

BWR K/A: APE 296006 AA2.06 ( - /3.8), PWG 01-( - /4.1) i ANSWER 8.06 (1.00)

' d i

l REFERENCE j BFNP: Standard Practice 12.24; 10CFR50. 54 . (i i i ) ; L.O.-B & C I

f 1

'BWR K/A: 294001 K1.03 ( -./3.7), A1.11 ( -

/4.3), Gen-Ki_( - /2.9)

-i Gen K5-( - /3.3) i

et__6Dd1NIEIB6I1YE_EBQGEDUBEQ2_CQUDlI1QN@2_6MD_LidlIOI1QU@ .PAGE 66 ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 8.07 (1.50)

(1) System fails to operate.

(2) System operates in a suspected adverso manner.

(3) System operates outside of the limits of the documented acceptance criteria. (0.5 each)

REFERENCE BFNP: Standard Practice 10.9, L.O. "B"; NOAM (Part 2), Section 4.9 ANSWER 8.08 (1.00)

(Shutdown and) calculate the maximum pressure at the bottom of the vessel (considering the standing head of the water). (0.5)

^

If GT 1375 psig the NRC must be notified (within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - NRC permission is required to restart). (0.5)

REFERENCE BFNP: Technical Specification 1.2 (and Bases) & 6.5, L.O. 74; Standard Practice 12.8, Section 6.B.7, L.O. "C" BWR K/A: 295025 EK 1.05 ( - /4.7), EA 2.01 ( - /4.3),. Gen.K5 ( - /4.4) 295020 AA 2.04 ( - /3.9)

Ot__6DMINISIB6IlyE_EBQCEQQBE@2_CQNQlIlgNQ3_8NQ_(1M1I@l1QNg PAGE. 67 ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.-

ANSWER O.09 (2.50)

a. Male or Non-reproducing Female Lifetime Exposure (5(N - 10))

Volunteer Over 45 (0.25 each)

b. 20 MIN (75/225 = 1/3 hr) (0.5)

c. 200 Rem .(0.25)

d. ALERT (>1000Mr/hr)

Immediately (10CFR20.403) (0.5 each)

REFERENCE 10 CFR 20.403 DFNP: RCI I, L.D. V.I.3; Emergency Plan, L.O. D DWR K/A: PWG K1.03 ( - /3.0)

Ex__8Dd161EIB011ME_BBQGEQQBES 1 _GQNQlIlQNS 2 _GUQ_LidlI611QNQ- PAGE 68 ANSWERS -- BROWNS. FERRY 1, 2&3 ~B6/12/01-BROCKMAN, K.

ANSWER O.10 (1.50)

Browns Ferry TS (3.7.B.1) requires two of three Standby Gas Treatment (SDGT) . (BH) train to be operable at all times when secondary con-tainment integrity is required. (0,5) The TS definition of operable requires both normal and emergency power to be available. The I emergency power supplies (A and D Diesels) for A and B SBGT fans were inoperable. (0.5) Thus, the TS LCO could not be met. (0.5)

(Discussion of TS 3.9.c, Electrical Lineup, can be partially credited up to 1.05 pts - Yes, it is operable, due to the meeting of the requirement for two off-site power sources and two diesels being available)

REFERENCE DFNP: Technical Specification 3.7.B.1; L.O.'s 1.d, 54, L 55 LER 50-259-06026 DWR K/A:

ANSWER 8.11 (1.50)

a. Browns Ferry Technical Specification (3.5.B.9) requires at least two operable RHR pumps for any unit containing irradiated fuel (0,5).

Technical Specification (3.5.C.7) requires at least one RHRSW pump for each RHR pump aligned f or heat e>tchanger service (0.5). Thus, the TS LCO's are not met (0.5). (1.5)

(Full Credit can be given for evaluation under the criteria of TS 3.5.B.10 criteria)

REFEF .NCE BFNP: Technical Specification 3.5.B.9 and 3.5.C.7; L.O.'s 2, 23, & 25 LER 50-259-06009; LER 50-296-86007 L -85017 l

l

.i BWR K/A:

es__800lNISIB8IlyE_ESQgEQQBES2_GQUQ1IIQUS3_8NQ_LINIIGIlQU@ 'PAGE 69 ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 8.12 (1.50)

(1) Highest worth rod (0.25) fully withdrawn (0.25)

(2) Xenon free '

(3) Cold (68 deg F) (0.5 each)

- OR -

(1) Most reactive condition during operating cycle (2) Strongest control rod fully withdrawn (3) All other operable control rods fully inserted ,(0.5 each)

REFERENCE General Electric Reactor Theory, Chapter 1 EIH: U2 TS, 1.0 "SDM" l DFNP Technical Specification 3.3/4.3.A, L.O. 9 Rx Theory, Chapter 1, L O. 5.1 ANSWEP 8.13 (1.00)

1) Check the associated equipment, panel, MOV Daard, etc., for abnor-mali ties to an extent practicd1 under the existing plant conditions.  !

2) ..nitiate an MR to Electrical Maintenance to investigate the over-load. (DO NOT REPLACE THE FUSE (0.3)) (0.5 each)

REFERENCE DFNP SP-12.24, Section 5.9.11; SP-12.24 L/P, L.O. O ANSWER 0.14 (1.00)

The clearance for the equipment shall be held by the SE for the persons performing the work. (0.5) He shall so note under " Remarks" '

on the clearance sheet the person's name, purpose, time, and date.(0.5)

REFERENCE ,

BFNP: SP-14.24, Section 6.1.7; SP-14.25 L/P, L.D. C l l

- Sz__0DUJNISIB8IIVE_EBgCEDUBES2_GONp1IlgNg2_8ND_LldII8IlgNS: -PAGE 70

ANSWERS -- BROWNS' FERRY'1, 2&3 -86/12/01-BROCKMAN, K.

i l'

ANSWER 8.15 (1.50) i

a. Shift Engineer

b. Plant Emergency Conditions (0.25), or for Personnel Safety (0.25)

I

c. . Brought to the attention of the PORC, no later than the next work day. (0.5 each)

J j REFERENCE l BFNP: SP-3.11, p 5; SP-3.11 L/P, L.D. C i

ANSWER 8.16 (1.00) d 4 REFERENCE l EIH: U1TS, 1.1.c BFNP: Technical Specification 1.1.B, L.D. 3 J

ANSWER 8.17 (1.00) b i

! REFERENCE

} DFNP: U1TS, 1.0.V.3 1

I ANSWER 8.18 (1.00) d 4

REFERENCE BFNP: Technical Specification 3.8.A, L.O. 60 t

l l

4 l

1 4

- - - =_ - -- -.

.= . - , , . ,.

Da__8DMINISIB8IIVE_BBQgEQQBES2 _ggNQlIlgNS2 _8ND_LIMIISIlgN@ PAGE 71 ANSWERS -- DROWNS FERRY 1,.2&3' -86/12/01-BROCKMAN, K.

ANSWER 8.19 (1.00) d REFERENCE BFNP: OSIL #35, L.O. "K"; Technical Specification 3.7.C, _L.D. #76.

DWR K/A: 290001 SWG #10 (3.3/3.4), #11 (3.3/4.2)

PWG A1.02 (4.2/4.2)

ANSWER 8.20 (1.00)

C REFERENCE BFNP: Technical Specification 3.7.B, L.O. #76 Technical Specification Definition - LCO BWR K/A: 223001 SWG #6 (3.0/4.0), #11 (3.3/4.2)-

PWG A1.02 (4.2/4.2) i

a 1

TEST CROSS REFERENCE .PAGE 1-

-QUESTIOb~ VALUE- -REFERENCE

.05.'01 1.00 .KEBOOOOO97 05.02 1.00 KEB0000154 05.03 1.50 KEB0000907 1

05.04' 2.'00 KEB0000908 05.05 1.50 KEB0000911 05.06 1.50 ~

KEBOOOO912 05.07 2.00 KEB0000913 05.08 1.50 KEB0000914

'05.09 1.75 KEBOOOO916 05.10 2.00 KEB0000917-05.11 3.00 KEB0000919

05.12 ' 1.

00 KEBOOOO937 j 05.13 1.50 KEBOOOO962

05.14 1.50 KEB0000963 4

05.'15 .1.00 KEBOOOO968 j 05.16 1.25 -KEB0000969 25.00 06.01 1.00 KEBOOOO870 3

106.02 1.50 KEB0000874 l 06.03 1.50 KEB0000878 j 06.04 1.50 KEB0000879 06.05 1.50 KEB0000880 <

06.06 2.00 KEB0000895 06.07 1.00 KEBOOOO927 i 06.08 .50 KEBOOOO928 i 06.09 1.50 KE80000930 06.10 2.25 KEB0000932 06.11 1.50 KEB0000935 06.12 1.00 KEBOOOO936

! 06.13 1.50 KEB0000940 j 06.14 1.00 KEB0000949 06.15 2.00 KEB0000951

] 06.16 2.00 KEBOOOO967-

06.17 1.00 KEB0000987 06.18 2.00 KEBOOOO988 i

26.25 .

d 07.01 1.00 KEB0000885 07.02 1.50 KEB0000887 07.03 1.00 KEB0000891

-07.04 1.00 KEB0000893 07.05' 1.00 KEB0000894,.

l- 07.06 2.50 KEBOOOO896 07.07 2.50 KEB0000899 l

-j 07.08 _ 1 1. 0' 0 KEBOOOO900 07.09 1.00' kEB0000901 07.10 1.00 KEB0000902

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

I l

TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE 07.11 1.00 KEBOOOO903 07.12 1.50 KEBOOOO904 07.13 1.50 KEBOOOO905 07.14 1.00 7B0000906 07.15 1.00 ._B0000942 07.16 1.00 KEBOOOO944 07.17 1.00 KEBOOOO945 07.18 1.00 KEBOOOO946 07.19 1.00 KEDOOOO947 07.20 1.00 KEBOOOO948 07.21 2.50 KEBOOOO966 27.00 08.01 1.00 KEBOOOOB88 ,

08.02 1.50 KEDOOOO890 '

08.03 2.00 KEBOOOO892 08.04 1.00 KEDOOOO971 08.05 2.00 KEBOOOO972 08.06 1.00 KEBOOOO973 08.07 1.50 KEBOOOO974 08.08 1.00 KEDOOOO975 08.09 2.50 KEBOOOO976 08.10 1.50 KEDOOOO977 08.11 1.50 KEBOOOO978 08.12 1.50 KEDOOOO979 08.13 1.00 KEDOOOO980 08.14 1.00 KEBOOOO981 08.15 1.50 KEBOOOO982 08.16 1.00 KEBOOOO983 08.17 1.00 KEDOOOO984 08.18 1.00 KEDOOOO986 08.l'1 1.00 KEBOOOO990 08.20 1.00 KEBOOOO991 26.50 104.75 l

l l

\

U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _BRQWNS_EEBBy_11_2k]_____

REACTOR TYPE: _pWR-GE4_________________

DATE ADMINISTERED: _@6/12Z91________________

EXAMINER: _pRggKM8N 2 K.

CANDIDATE:

~~

b

___h__h__b__~__~__

IUSIBycIIgNS_Ig_C8UDID81El Use separate paper for the answers. Write answers on one side only.

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

% OF CATEGORY  % OF CANDIDATE'S CATEGORY

__Y8LUE_ _IQIOL ___SCQBE___ _yGLUE__ ______________CGIEGQBy_____________

_2bt99__ _231Z9 ___________ ________

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

_201Z5__ _2d1SZ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

_2Z1gg__ _26 3 1; ___________ ________ 3. INSTRUMENTS AND CONTROLS

_2bs99__ _23129 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 19512D__ _________,_ ________% Totals Final Grade -

All work done on this examination is my own. I have neither given '

i nor received aid. ,

_________________________ A____'____

Candidate's Signature f

P a

L

L NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS' During the administration of this examination the following rules apply:-

t

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

4

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

l 1 3. Use black ink or dark pencil ggly to facilitate legible reproductions.

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

1 1

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

j 6. Use only the paper provided for answers.

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

i

8. Consecutively number each answer sheet, write "End of Category __" as j appropriate, start each category on a ogw page, write gely_gg goe sidg

.i of the paper, and write "Last Page" on the'last answer sheet.

I i

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

l j 10. Skip at least threg lines between each answer.

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

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

. 13. The point value for each question is indicated in parentheses after the

} question and can be used as a guide for the depth ofl answer required. l

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

l 4 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE

{ QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

3

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

the examinet_only. j

j. 17.-You must sign the statement on the cover sheet that indicates that.the.

i work is your own and you have not received or been given assistance.in completing the examination. This must be done after the examination has been completed.

f,.

~ - -

7, : - ~,yy g .p < +

t ' *y 'e s< g ~

(, ," '

% t. -

.w  %.

^ i- , (

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

s

a. Assembly your examination as Tallaws:-

-(1) Exam questionI on top. '

s

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

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

~

b.- T' urn in your copy of the examination-and all.pages.used to answer  !

the_ examination. questions.

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

- not use'for answering the questions.

5

) '\

d. Leave the examinat. ion area, as defined by the exininer. If after leaving, you are found in this area while the examination i s still in progress, your license may be denied or revoked.

i en I

.l

~i l

i 4

I L

4 l

l 3

}

)

I

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

, ,n,, - - , , - - , , , - -,,,-,.-v-p- v g,,,-- e..w.,v. . - - - - -

n e w ,., , -,,n - - p .r+,-+y y

.11__EBINGIELES_gE_NUCLEGB_EgWEB_ELGNI_QEEB811gN 2 - PAGE 2 ISEBOOQYN801CS1 _SEOI_IB6NSEEB_6ND_E691D_ELOW .

t DUESTION 1.01 (1.00)

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

a. Enthalpy DECREASES, Entropy DECREASES, Quality DECREASES

b. Enthalpy INCREASES, Entropy INCREASES, Quality INCREASES

c. Enthalpy CONSTANT , Entropy DECREASES, Quality DECREASES

d. Enthalpy DECREASES, Entropy INCREASES, Quality DECREASES QUESTION 1.02 (1.50)

Concerning the Dypass Flow in the reactor core:

a. Which one of the following most accurately indicates core bypass flow at 100% thermal power and core flow? (1.0)

(1) 1%

(2) 5%

(3) 11%

(4) 20%

b. STATE the most significant consequence that would occur if bypass flow were significantly reduced at full power. (0.5)

)

l l

I i

l l

1

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION 2 PAGE 3' IBEBdQDYN8MICS2 _UE81_IB8NSEEB_8ND_ELUID_ELQW.

s QUESTION 1.03 (2.00) i . .

Concerning the Net' Positive Suction Head (NPSH):

~

a. DEFINE NPSH (0.5) j b. -For each.ofithe'following, indicate whether the available NPSH at the suction of the rec'irculation pump would INCREASE / DECREASE /.

j . REMAIN THE SAME: ( 1. 5) -

(1) The Feedwater Flow is INCREASED l

, (2) The Recirculation Flow is INCREASED (3) The Vessel Pressure i s INCREASED from 200 psig to 800 psig J

QUESTION 1.04 (1.00)

Which one of the following most accurately decribes when Control Rod Worth, during a REACTOR STARTUP, would normally be at i ts MAXIMUM 7-

a. Cold Shutdown

b. Heatup in Progress (* 1% RTP)

c. Heatup Complete (* 1% RTP)

I

d. 50% RTP l

)

e. 100% RTP QUESTION 1.05 ( .50)

STATE the relationship betweeen between Beta-effective and Reactivity which best describes PROMPT CRITICAL.

l' i

1 4

4 i

i i

l

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

1 c... - . . ,g.r .-,,w. w . ~ - - - . - . . . . _ , - + , . . . , ,,,-,e-2 . _y.. m -.,._ ,u.,- yy. , . , - ,,,,p., , , . . . , , -m, . . , . , ,- . - , . , . ,-,r,,,-

'1.__EB18CIELES_9E_NUC6EGB_EgyEB_ELGNI_gEEBBIIgN2 PAGE 4.

.IBEBdgDYN8MIC@2_ME8I_IB8BSEEB_8ND_E691D_E6gW

+

- QUESTION 1.06 (1.50)

Attached Figure # 911 shows a representative Xenon concentration curve. The reactor is. scrammed at Point A, resulting in a peak Xenon concentration approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> later, at Point B.

Line B - C represents the normal. depletion of Xenon-that would beLexpected.

t If the reactor were RESTARTED at Point B (dotted line), EXPLAIN:

a. The rapid decrease in Xenon concentration from Points B to D. -(0.5)

b. The change in the distribution of the radial neutron flux during the period from Points B to D (include reasons) . (0.5) f' c. Any special precaution (s)'that you would need to observe,.as a

a reactor operator, between Points B and D (include reasons) . (0.5) 1 QUESTION 1.07 (2.00)

Concerning the effects of Control Rods on reactor power:

a. EXPLAIN how a control rod withdrawal could result in a reactor power decrease (reverse power effect). (1.0)

b. Which one of the following would be the rod movement sequence most likely to cause the reverse power effect described above? (1.0)  ;

(1) Deep Rod - 10 notch movement (2) Deep Rod - 1 or 2 notch movement 2 (3) Shallow Rod - 10 notch movement l (4) Shallow Rod - 1 or 2 notch movement I

1

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

lt__EBINCIELES_OF_NUCLEAB_PgWEB_ PLANI _gPgBAllgN 2 PAGE 5 ,

IHEBMggyN8MICS2_HE81_IB8NSEEB_8NQ_E6UIQ_E69W H

.i 4-OUESTION 1.08 (1.50)

I j Certain License / Technical Specification Limits are'placed on the operation of the reactor to ensure that the design vessel, fuel power, thermal, and pressure limits are not exceeded during normal 4

or transient operations.

STATEfthe exact minimum or maximum (as appropriate) value'for each

! of the following limits.

NOTE: ASSUME FULL POWER OPERATION UNLESS OTHERWISE SPECIFIED

a. LHGR j b. MCPR i

i c. MWt I. d. Reactor Pressure

e. MWt (< 800 psia or < 10% Rated Core Flow)

f. Cooldown Rate for Shutdown (Technical Specifications) l 4

4 OUESTION 1.09 (1.75) t i

{ The reactor is at the beginning of a fuel cycle (BOC) and exactly critical. You withdraw control rods to add 0.06% ^k/k.

7 NOTE: ASSUME BETA-EFF = 0.0072 and LAMBDA-EFF = 0.1 SEC E-1

a. CALCULATE the resulting stable period. (SHOW ALL WORK) (0.5)  ;

l

b. CALCULATE how long i t would take reactor power to DOUBLE. (0.5) l~ c. If the reactor were at the End of Cycle (EOC) vice DOC, would the resultant stable period,-for the same reactivity addition described in (a), be LONGER or SHORTER? JUSTIFY your responce! (0.75) ,

l, 1

i I

i i

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

i 4

- , . - . , - ~ . . . _ , .,_-.,,-,_e--.-.-,.,,m. -.,., y,_--,-yv,,.- , , - , , , . - . . - , , . _---,,or,. ,,_rn.,- , - , , . , - , , , ...e, .,..-,w -.m,_. -e -

11 PRINCIPLES OF NUCLEAR POWER PLANT' OPERATION3 - PAGE 6 ISEBdggyN851CS 3_ BEAT TBONSEE8_8ND_E6UID_ELgW i

i l OUESTION 1.10 (3.00)

Attached Figures # 5 A & B represent a transient that could occur at a BWR-GIVEN: (1) One Retire Pump Flow Controller FaiJs High 0 Time t = 1.2 minL

,f (2) No operator. actions occur l (3). Both Recirc Pumps were on line 0 Time t = 0.0 minutes

EXPLAIN'the cause(s) of the following recorder indications

i

a. C re Flow DECREASE (Point 3) (0.5)

b. Total Steam Flow DECREASE (Point 10) (0.5)

I'

c. Reactor Pressure STABILIZATION (Point 12) (0.5)

( At a lower level)

d. Reactor Level DECREASE (Point 13) (0.5) i e. Reactor Power DECREASE (Point 16) (0.5) 1

f. Since this transient occurred with power at less than rated (70%),

i STATE the factor (or program) which is utilized to ensure that the

! MCPR limit is not exceeded. (0,5) 1 l ,

QUESTION 1.11 (1.00) 1

! Which one of the following correctly DEFINES " Void Fraction"?

l r a. steam volume in mixture / total volume of the mixture

! b. steam volume in mixture / liquid volume in mixture

]

c. steam mass in mixture / total mass of the mixture

d. steam mass in mixture / liquid mass in mixture i

l i  !

l l i

1 1

i I l t

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

4 w____ . _ . . . . , , , , . , _ _ , _ , - . . - . - , , , , ~ .

l 12__eBINCIPLES OF' NUCLEAR POWER PLANT OPERATION 2 PAGE 7' IUEBdODYN8dlCS2_bE81_IB8NSEEB_8ND_ELUID_E69W l

-QUESTION' 1.12 (1.00)

Significant. quant'i ties of Hydrogen gas may be generated during, or subsequent to, a LOCA. This Hydrogen gas ... (CHOOSE ONE)

.a. is generated primarily by the circaloy-water reaction if fuel clad temperature is allowed to exceed'2200 deg F.

b. is generated primarily by the radiolytic decomposition of water.'

• 1

' could cause an explosive hazard in the drywell if allowed to c.

i exceed the ?ower flammability limit of 1% concentration.

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

QUESTION 1.13 (1.00)

During a refueling outage, one Staandby Liquid Control (SBLC) loop is tested per the Technical Specifications by firing a Squib Valve and pumping the contents of the Test Tank into the vessel. Due to an oper-ational error, the Test Tank contained a significant Baron concen-tration (as opposed to being purely deimineralized water). This l resulted in a Doron concentration of approximately 100 ppm inside of the vessel.

l DESCRIBE the problem which could occur from the PDAH (onward) if the j reactor was started up in this condition.

i t

4 OUESTION 1.14 (1.00) j The THRESHOLD power below which PCI failures do not occur is.known to l DECREASE with fuel burnup. STATE two (2) reasons for this decrease in the

} PCI threshold.

I i

l 1

l I I

i 4

i i

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

1 1

li__EBINClebES_QE_NgCLE68_EgWEB_E68NI_QEEgeIIgN2 PAGE' 8-

-ISEBdQQyNGdICS3_ DEST TB8NSEEB_ONQ_E6UID_E698 QUESTION 1.15 (1.00)

With--the reactor initially at a K-eff of 0.99,1a certain reactivity.

change causes the count rate to double. If this same amount of reactivity is again added to the reactor, which of the following will be the status of the reactor 7

a. Subcritical

b. Critical

c. Supercritical

d. Prompt Critical OUESTION 1.16 (1.50)

For each of the following transients, STATE which of the reactivity coefficients (Moderator, Void, or Doppler) would respond first.

JUSTIFY your choice.

a. Rod Drop Accident from 10% thermal power. (0.75)

b. MSIV Closure from 90% thermal power. (0.75)

OUESTION 1.17 (1.50)

A reactor heat balance was performed (by hand) during the 00-08 shift due to the Process Computer being OOC. The GAF's were computed, but the APRM GAIN ADJUSTMENTS HAVE NOT BEEN MADE.

STATE whether the Actual Power will be HIGHER or LOWER than the Calculated Power for each of the following:

a. The steam flow used in the heat balance calculation was LOWER than the actual steam flow.

b. The reactor recirculation pump heat input used in the heat balance calculation was OMITTED.

c. The RWCU return temperature used in the heat balance calcu- )

lation was LOWER than the actual RWCU return temperature, j

~1 l

1 i

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

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

l 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION 1 PAGE- 9 THERMODYNAMICSg_h2AT TRANSFER AND FLUID FLOW j i

l 1

i

! OUESTION 1.18 (1.00) i f- During a Scram recovery, the reactor is cooling down at the maximum

administrative limit. The K-eff'of the reactor is 0.99.

1 i i CALCULATE when the reactor would theoretically achieve RECRITICALITY.

1 IGNORE THE EFFECTS OF XENON. SPECIFICALLY LIST ALL ASSUMPTIONS!

}

?

DUESTION 1.19 (1.25) iL j SELECT the apropriate response for each of the folowing examples of

} variable Control Rod Worth:-

I

a. (MORE/LESS) control rods would need to be pulled to make the reactor critical at 545 dog F, as opposed to 140 deg F.

i

b. An INCREASE in the Void Fraction will result in an (INCREASE / DECREASE)

! in individual control rod worth. ,

J

c. Control Rod Worth at End of Cycle would be (LESS/ GREATER) than at the Beginning of Cycle.

d. Control Rod Worth will (INCREASE / DECREASE) with an INCREASE in y

moderator temperature.

e. Control Rod Worth will (INCREASE / DECREASE) as the adjacent' control l rods are withdrawn.

i-i i l l l l

i i

i  :

2 1

i i

1 I

i

! 1

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

i l'

2r__E66NI_pggigN_ INCLUDING _g8EEIY_6ND_ EMERGENCY _gYSIEde .PAGE 10 i OUESTION 2.01- (1.50)

Concerning the Control Rod Drive Hydraulic (CRDH) System:

'For EACH of the following (a - c) SELECT-the one response that most accurately describes the proper system design.

a. The minimum Scram Discharge Volume which would be necessary to accomodate a FULL POWER reactorLscram, at EOC, would be approximately ...

_ (1) 50 gal

_(2) 200 gal (3) 600 gal (4) 1100 gal

b. The Scram Discharge Volume is limited to a pressure of ________

during a reactor scram to ensure complete rod insertion.

(1) 65 psig (2) 125 psig (3) 650 psig (4) 1250 psig

c. Following a reactor scram, the Scram Discharge Volume is protected from overpressure by a relief valve which lifts at ________.

(1) 65 psig (2) 125 psig (3) 650 psig (4) 1250 psig

(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

I m___________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _

- 2z__EL9NI_Dgg1GN_18CLyQ1NG_S6EEIY_6ND_EdEBGENCy_gySIEd3 PAGE> 11

.i QUESTION 2.02 (1.00)

Concerning the RCIC System:

i a. STATE the adverse effect(s) that would occur if the RCIC pump Suction Relief Valve were to STICK OPEN following system initiation. '(0.5)

b. TRUE OR FALSE (0.5)

I A RCIC Low Steam Pressure Isolation will NOT automatically reset when steam pressure is restored.

I QUESTION 2.03 (2.251 i

j For each HPCI System related component failure listed below, STATE:

(1) W h e t'h e r , or not, HPCI will AUTO INJECT into the vessel -

i

. (2) If injection will not occur, WHY

) (3) If injection will occur, one adverse consequence of the

system operating with this failed component.

a. The Minimum Flow Valve f ails to close af ter the HPCI start. (0.75) i j b. The HPCI discharge flow element output signal to the HPCI

{ flow controller is failed to its maximum output. (0.75) i j c. The SDGTS fails to operate. (0.75) i l

! OUFSTION 2.04 (1.00) i The RHR pump suction valves from the Torus are interlocked to prevent i i opening when the Shutdown Cooling RHR pump suction valves are open -

this prevents diverting reactor coolant to the Torus while in Shutdown 1 Cooling.

1 I

DESCRIDE another potential path for diverting reactor coolant to the Torus during Shutdown Cooling (and flushing) operations, and STATE the specific operator actions which are required by procedure

. to ensure that this path is unavailable.

t I

l

(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) I 1

1 i

r l

L- 2__E60NI_DghlgN_

1 INCLUDING _g8EEIY_8ND_gdEBGENCY_gYSIEdg PAGE- 12:

i l

l QUESTION 2.05 ( .50) i EXPLAIN why it is essential to ensure that the Diesel Generator governor is positioned f ully to the LO'J SPEED STOP, prior to performing a local start.

QUESTION 2.06 ( .50)

While shutting down a Diesel Generator from the Control Room, following a Surveillance Test, you inadvertently allow the generator

, to trip on REVERSE POWER prior to opening the output, breaker.

I DESCRIDE how you would return the Diesel Generator to Standby Readiness, given this condition.

QUESTION 2.07 (1.50)

Unit 2 is at full power when a Load Reject occurs due to a trip of the MTG Output Breakers.

a. STATE the specific signals (and their sources) which activate the Load Reject circuitry. (0.5)

b. DESCRIBE the response of the Turbine / EHC System to this event. (1.0)

DUESTION 2.08 (1.50)

STATE the response (Component and Reaction) of the Main Steam Isolation Valves to a LOSS of EACH of the following:

a. AC Power

b. Control Air

c. Drywell Control Air QUESTION 2.09 (1.50)

An automatic LPCI injection signal is received:

STATE the three specific conditions (setpoints, if applicable)  ;

that would be required to allow you to initiate Torus Spray. <

(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) I l

I 2i__EL8NI_DE@lGN_1NCLUDid@_S8EEIY_8ND_EMEBGENCy_@ySIEMS PAGE: 13 V

i QUESTION 2.10 (1.00)

! .During a Loss of Offsite Power (LOSP) with a concurrent LOCA. condition I on Unit 1 (Core Spray-pumps in operation), another LOCA signal occurs

on Unit 2. Which one'of the statements below most accurately reflects

-the response of the Core Spray Logic?

! a. Unit I Core Spray pumps 1A and 1C will TRIP'and Unit 2 j Core Spray pumps 2A and 2C (only) will START.

a x j b. Unit I Core Spray pumps 1B and 1D will TRIP and Unit 2 Core Spray pumps 2D and 2D (only) will START.

4

c. All Unit 2 Core Spray pumps will START,-in addition to l the four pumps already operating on Unit 1.

l

d. All Unit i Core Spray pumps will TRIP and Unit :2 Core

< Spray pumps must be MANUALLY sequenced ON, as the Diesel j Generator capacity allows.

1 l

! QUESTION 2.11 (1.50) k

A Core Spray initiation signal is present on Unit 2. You MANUALLY I secure the "C" Core Spray pump due to increasing reactor vessel level, i

a. STATE the indication (s) that would be provided on the control panel that the pump was manually secured. (DO NOT address flow,  ;

i pressure or other metered indications) (0.5)

[ .b. LIST the conditions that would have to exist and/or the actions ,

I that you would have to take to permit the pump to automatically

restart, if needed. (1.0) i l

ni!FSTION 2.12 (1.00) i j STATE whether ADS would DEPRESSURIZE or NOT DEPRESSURIZE the reactor j for_each of the following conditions:

CONSIDER THE TOTAL ADS RESPONSE AND ASSUME NO OTHER OPERATOR ACTIONS ~ OCCUR i

i

a. After initiation of the 120 second timer, but before it is timed out, Drywell Pressuro DECREASES to 1.5 psig. (0.5) o 1

i b. After initiation of the 120 second timer, but before it is timed out, Reactor Water Level INCREASES to -90 inches. (0.5)

?

i (***** CATEGORY O2 CONTINUED GN NEXT PAGE *****) l

! i i

i

. i

c ag__PL8NI_DESIGU_lNGLUDING_SOESIy_8ND_EMER@gNGy_gygIgMS PAGE 14 l QUESTION 2.13 (1.00)

DESCRIBE the difference (s) in resetting a Mechanical Overspeed Trip-on the HPCI turbine as compared to the RCIC turbine.

QUESTION 2.14 (2.00)

LIST the four (4) conditions that must exist for a Diesel Generator Output Breaker to Close following an AUTO start-signal.

I QUESTION 2.15 (1.50) 2 I

Concerning the Safety Relief Valve tailpipe Vacuum Breakers:

i a. STATE the purpose of these Vacuum Breakers. (0.5) l

) b. Assuming accident conditions, LIST two (2) significant detri-mental effects that could occur if one (1), or more, of these Vacuum Breakers STUCK OPEN. (1.0)

) QUESTION 2.16 (2.00)

Answer EACH cf the following, concerning RHR, as TRUE or FALSE:

I a. RHR-LPCI will AUTO initiate on a loop which is in Shutdown

. Cooling if Level is =< -114.5" or Drywell Pressure is >= 2.45 psig. (0.5) i

b. AUTO initiation of RHR-LPCI will not occur on a High Drywell i

Pressure signal (>= 2.45 psig) UNLESS the Core Spray Drywell Pressure

, Test Switch is in AUTO. (0.5) i f c. When utilizing RHR Drain Pump "A" for supplemental Fuel Pool j Cooling, only Loop II of the RHR System is available for LPCI. (0.5)

! d. On a Loss of Offsite Power (LOSP) with a LPCI Initiation Signal, the RHR pumps are sequenced on at 7 second intervals to prevent overloading the Diesel Generators. (0.5) i

(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

f

2t__ELOUI_ DESIGN _INGLQDING_E8EEIY_8NQ_gdgBGENGY_SYSIEdg PAGE 15 4

QUESTION 2.17 (1.00)

. The main generator is on line at 800 megawatts when a hydrogen leak j in the generator reduces hydrogen pressure to 45 psig. Using attached Figure # 451 (Estimated Capability Curve):

j

a. STATE the maximum leading REACTIVE load allowed on the generator.

if a power factor of 0.00 is to be maintained. (0.5)

b. STATE the maximum REAL load allowed on the generator if a power.

j facter of UNITY-(1.0) is to be maintained. (0.5) i 1

> I i

] UUESTION 2.18 (1.00)

I l STATE ALL of the automatic signals which will CLOSE the RDCCW l Ractionalizing Valve (FCV-70-48). .

l

! OUESTION 2.19 (1.00) i l Reactor Feed Pump (RFP) turbine speed is controlled by either j a Motor Speed Changer (MSC) or an Motor Gear Unit (MGU). The '

, MSC ...(CHOOSE ONE) d a. ... will control the RFP turbine's speed only if its speed  !

j signal is greater than that from the MGU.

, b. ...is only able to control feed flow rate over an approx-

) imate turbine speed of 2600 - 5650 rpm.

1

c. ..., like the MGU, does afford the capability of manual speed control by use of a local handt. heel,

d. ...will lock in place to prevent a ramp response to a false *

] signal, if the signal from the flow controller in lost. ,

t i l i

i 1

j i

.I

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

I f r

I Bi__E68dI_DESIGU_ING6UDIUQ_SBEEIY_889_EdEBQENGY_SYSIEUS PAGE 16 i

QUESTION 2.20 (1.50)

For each of'the following RHR-systems, STATE the RHR system (s),

(unit and system), IF ANY, with which it can be DIRECTLY cross-connected. Consider only Unit-to-Unit cross-tie capabilities.

l a. Unit 1,' System 1

b. Unit 2, System 1 c .~ Unit 3, System 1 i

i i

r l

l l

\

t I

(***** END OF CATEGORY O2 *****)

l'

l- 'Nz__10SIBUMggIS_8ND_CQNIBQL@ PAGE 17

) QUESTION 3.01 (1.00) i Attached Figure. # 869 shows the " Error Signal Input" to the Reactor Recirculation System Flow Controller and shows four potential Controller _

Output signals (a - d).

i i Which one of the output signals is the appropriate response to the i input signal shown?

4 i i, QUESTION 3.02 (1.00) i  !

! The reactor is at 100% Rated Thermal Power

{ "A" Feedwater Level Control is Selected I All Controls are in Normal / Automatic j i

i

, A failure in a Rx Vessel Level Narrow Range Instrument has occurred and

resulted in the following related trips / indications

i .

"A" NR level indication reads at MINIMUM j "D" & "C" NR level indications read at MAXIMUM i I REACTOR WTR LEVEL A ADNORMAL Annunciator - ON l

t j REAC VESSEL WATER LEVEL LOW-LOW CHAN A Annunciator - OFF a

i REAC VESSEL WATER LEVEL LOW-LOW CHAN D Annunciator -'OFF j Two channels of Level 8 have TRIPPED One channel of Level 3 has TRIPPED 1

Feedwater flow is at ZERO State which NR level transmitter has failed AND state in which direction it has failed (HIGH/ LOW). l I

i l

i l

i i i

I l

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

l i

~-

r- .

-3 t;_INSIByMENIS_6ND_CONIBOLg. PAGE- 18 l

l OUESTION 3.03 (1.00)

Concerning Level Instrumentation:

a. Attached Figure.#15 shows the relationship between indicated'and

actual water level on the Scram Bartons. Each line represents the j relationship at a different pressure - O, 200, 600, and 1000 psig.1 NOTE

THE LINES ARE NOT MARKED BY PRESSURE INTENTIONALLY If actual level is 42 inches,'with a reactor pressure of 600 psig, STATE the level which'would be seen.in the-control room (indicated). (0.5)

~b. Figure #4 shows a Yarway Column - STATE whether, orLnot, this column is temperature compensated. JUSTIFY your. response. . (0. 5) -

QUESTION 3.04 (1.00)

Concerning the PCIS Logica

a. State whether the normal PCIS sensor relays are ENERGIZED or.

DEENERGIZED when an UNSAFE condition exists. (0.5)

b. LIST the TWO systems whose isolation valve control logic l 1s the opposite of'the above referenced (part "a") norm. (0.5) l OUESTION 3.05 (1.00)

Concerning the Local Power Range Monitors (LPRM's):

a. TRUE or FALSE l An LPRM's gamma sensitivity decr. eases with core life due-to the.

l depletion of the ionized Argon gas as it is used up in estab-l lishing a current flow. (0.5)  ;

l

b. STATE the purpose of the AMPLIFIER RANGE SWITCH for,each LPRM. (0.5) l l

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

l 1

c 3 1__INSIBUMENIg_GUD_CONIBgL@ PAGE 19.

}

a 1

, ' QUESTION 3.06 (1.50) 4 Concerning the APRM Circuitry:

)

a. STATE the (range 'of ) time by which the flow-biased signal is time delayed. (0.5) i i b. STATE what causes the circuitry to arrive at a'particular time i ~ delay (within the range described above). (0.5) j

$ c. The APRM flow-biased Scram trip is present when the MODE SWITCH

, contacts are (OPEN/ CLOSED /BOTH OPEN AND CLOSED) - CHOOSE ONE! (0.5).

1 i

t

] QUESTION 3.07 (1.00) i i The following plant conditions exist:

NOTE: FIGURES # 875 A & B ARE ENCLOSED FOR REFERENCE

- Inlet condensate valve to SJAE A intercondenser is FULLY OPEN Outlet condensate valve to SJAE A intercondenser is FULLY CLOSED 4

Condensate pressure from SJAE A intercondenser is 45 psig j Steam isolation valves to SJAE B are FULLY OPEN ,

1 Steam pressure at PS (1-151) in 182 psig j Steam pressure at PS (1-166) is 173 psig i

l .

You take HS 66-14 (AOV FCV) to the OPEN position.

i l STATE whether FCV 66-14 will open. JUSTIFY your response.

1 j- OUESTION 3.08 (1.00) l Concerning the operation of the Rod Block Monitor (RBM):

i' You select a center rod for movement.

Recirculation Flow is 33%

j The Reference APRM reads 36%

The LPRM inputs to the Averaging Circuit read (an average of) 40%

j Withdrawal of the selected rod causes an average power increase of 6%.

! With no operator actions (except the rod withdrawal), STATE whether l or not a Rod Block would be received. JUSTIFY your response.

I

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

I

f 3 t__INgIBUMENIS_AND_CONIBOLS PAGE 20 t

j i

. QUESTION 3.09 (1.50)

} The plant is shutdown and the reactor cavity 'is _ filled. RHR-System 1

-(Pump A) is operating in the Shutdown Cooling mode, via Recirculation l Loop,B.

During. maintenance, an IM causes a Group II isolation signal to be i erroneously generated.

t a.- STATE the specific RHR' system responses to the signal. (All valves, pumps,etc. - identification numbers not required) (1.0) i'

b. LIST all, if.any, operator actions required to reset the isol-1 ation (once the signal is verified as erroneous and cleared). (0.5)

OUESTION 3.10 (1.50) l Concerning the Fire Protection System:

a. MATCH the Detection System (COLUMN A) with its Useage (COLUMN D) (0.75)

]

COLUMN A COLUMN D >

] 1. Rate of Rise Thermal Detector a. General area coverage throughout the plant

2. Continuous Strip Detector j b. Spreader Rooms and around any 1 3. Ionization Smoke Detector oil reservoirs

c. Cable Tray _ protection

'l i b. Given that a fire detection system provides an initiation ,

function - STATE how the Fire Protection / Raw Service Water system

responds / aligns. (0.75) i OUESTION 3.11 (1.00) i For each of the following indications on a P-1 edit, GTATE whether j it would indicate VALID or INVALID data.

1 j a. ITER = 2 -(0. 5)

]

b. IXYFLAG = 1 (0.5) l

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

3 i

__ _ _ . _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ . _ . _ _ _ . . _ _ . _ _ . _ _ _ _ . . _ . , _ . m __ _ . _ ____ . . _ . _ _ _ _ __

3 1__INSIBudEUIg_6ND_QQUISQL@ PAGE 21 QUESTION 3.12 ( 1. 00)

Answer the following with respect to the Rod Sequence Control' Systems During the performance of SI 4.3.B.3.a concerning RSCS Operability prior to startup, a " TEST FAILED" light is received (Panel 9-28) after depressing the "COMPARATOR CHECK A" Pushbutton. STATE whether this indicates a SATISFACTORY or UNSATISFACTORY response. JUSTIFY your response!

OUESTION 3.13 (2.00)

The following plant conditions exist Jet Pumps 1 - 10 Differential Pressure (Meter) = 15 psid Jet Pumps 11 - 20 Differential Pressure (Meter) = 3 psid RECIRC LOOP A ONLY OUT OF SERVICE (Annunciator) = ON

a. The TOTAL CORE FLOW Recorder would calculate core flow by which of the following methods? (1.0)

(1) Loop D Jet Pump Flow + Loop A Jet Pump Flow (2) Loop B Jet Pump Flow - Loop A Jet Pump Flow (3) Loop D Jet Pump Flow Only (4) Loop A Jet Pump Flow Only

b. After 5 minutes, the operator opens the Discharge Valve of Recirc Pump A to meintain the A Loop temperature. The TOTAL CORE FLOW Recorder would calculate core flow by which one of the following methods? (1.0)

(1) Loop D Jet Pump Flow + Loop A Jet Pump Flow (2) Loop B Jet Pump Flow - Loop A Jet Pump Flow (3) Loop D Jet Pump Flow Only (4) Loop A Jet Pump Flow Only

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

E

l Jz__INDIBUDENIa_8BD_G9 BIB 963 PAGE 22  ;

l i

i l

t OUESTION 3.14 (1.50)

For occh of the Recirculation Pump Flow Control Limitern listed below .

LIST ALL input uignals or donign features which will activate (enforco) it

a. 20% Minimum Speed (0.5) ,

b. 20% Minimum Spoed (0.5)

c. 75% Minimum Speed (0.5) ,

OUESTION 3.15 (1.00)

Unit i is operating at 100% RTP, with rocire in Mantor Manual. The "A" [

Prossure Regulator unit, which in governing, FAILG LOW (i.e., the  ;

output (s) of Regulator "A" FAILS DOWN) ,

ASSUME: 1. No Operator Actions

2. All other EHC control sottings are normal  !

3. Starting Paramotors o TCV'n -

100% Steam Flow Position o DPV's -

0% Steam Flow Position o Power -

100% Rated Thormal Power i o Prousuro - 1010 psig ,

NOTE: FIGURE M 374 IG ATTACHED FOR REFERENCE  !

Which of the following mont accuratoly describen both the INITIAL i RESPONSE and FINAL STATUS of the difforont paramoters and compononts. 1 a b c d INITIAL RESPONSE  ;

o TCV's i NO CHANGE ITHR0TTLE CLOSEITHR0TTLE CLOSEI NO CHANGE 4 o DPV's ITHROTTLE OPENITHROTTLE OPEN I NO CHANGE I THROTTLE OPEN o Power i DECREASE I NO CHANGE I INCREAGE I DECREASE o Preunuro 1 DECREAGE I NO CHANGE I INCREASE I DECREASE i i l I i FINAL STATUS I I I l I I I I I o TCV's fo%(MGIV Shut)I < 100 % 1 *100 % i NO CHANGE l o DPV's 10%(MSIV Shut)l > 0% 1 0% 1 0% ,

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

l o Pronauro IAs controlludl >1010 pulg I >1010 poig i <1010 psig  ;

lby GRV's and I l  !

l IHPC1/RCIC I l l l

P

(***** CATEGORY 03 CONT INUED ON NEXT PAGE * * * **) i

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

It__1NSIBudEUIS_0ND_GQUIBOLS PAGE 23-QUESTION 3.16 (2.00)

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

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

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

c. Roactor watur level increases and stabilizou at a higher level.

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

e. None of the above.

[ 1. The plant is operating at 70% power, in 3-element control, when One (1) (1SIV Fails Shut.

11. The plant is operating at 1007. power, in 3-element control, when One (1) Feud Flow Detector FAILS DOWNGCALE.

l.

UUESTION 3.17 (2.50)

Listed below aru five (5) parameters which can indicate a Failed Jet Pump.

GTATE whether thoue parameters will INCREASE, DECREASE, or REMAIN THE' SAME for a Jul Pump Failuru.

NOTE: THIS IS NOT A RISER DREAK

a. Reactor Power, as indiented by APRM's. (0.5)

b. Core Flow, as calculated from Core Plato dP. (0.5)

c. Failed Jet Pump (* low. (0.5)

d. Companion Jet Pump Flow (Othur Jet Pump on the common riser). (0.5) i e. Companion Jet Pump Loop RECIRCULATION FLOW (Other JP Loop). (0.5) i i

i r

(***u* CATEGURY 03 CONTINUED ON NEXT PAGE *** n) t

(

3 t__INSIEUMENIS_0NQ_GQUIBQLS PAGE 24 ,

QUESTION 3.10 (2.00)

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

The "Recirc Flow D Limit" annunciator is CLEAR. For each of the f ollowing instar.ces, STATE how the speed of Recirc Pump "B" will change (i . e. , INCREASE, DECREASE, REMAIN THE SAME) and.

WHICH COMPONENT (S) of the control / positioning system is/are LIMITING.

NOTE: FIGURE 8 474 IS PROVIDED FOR REFERENCE

a. Recire Pump "D" M/A Transfer Station placed in " BALANCE". (1.0)

b. Recirc Pump "D" M/A Transfer Station manual potentiometer is turned fully in the counter-clockwise direction. ( 1. 0) .

QUESTION 3.19 (1.00)

Answer the following with respect to the Rod Sequence Control System DESCRIDE how the lamp dimmer function differs when power is between the LPSP & LPAP, as compared to being below the LPSP. (1.0)

QUESTION 3.20 (2.00)

When paralleling a Diesel Generator with the 4Kv Bus:

a. STATE the purpose of the governor control -

(1) Defore the EDO Output Dreaker is closed. (0.5)

(2) After the EDG Output Breaker is closed. (0.5)

b. STATE the purpose of the voltage regulator -

(1) Defore the EDO Output Dreaker in closed. (0.5) i (2) After the EDG Output Dreaker is cloud. (0.5)

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

Sz__EBQGEDUBES - NORMAL _2 ABNORMAL _

2 EMERGENCY AND PAGE 25  ;

=B8Dlg6QGIG86_GQNIBQL I i

J. .

1 1 i

QUESTION 4.01 (1.00)

EOI-2, " Containment Control", states in Step;PC/P-5:

. IF drywell pressure exceeds 55 psig (70 psia) , THEN

... vent the primary containment ... to reduce and i maintain pressure below 55 psig as follows

1. IF suppression pool water level is below 20 feet, j THEN vent the suppression chamber.

J' EXPLAIN why (if conditioins allow) the suppression chamber is vented

first (to reduce drywell pressure). Include the benefit acheived -

3 and how this is accomplished.

1

.i

! 'OUESTION 4.02 (1.00)

]

j Attached Figure "F" shows the Heat Capacity Temperature Limit (HCTL).

1 EXPLAIN the purpose for having the HCTL. (i.e., what does the curve protect against)

I j -

t e

l QUESTION 4.03 (1.00) 4 Special concern is required for returning MOV's, that have been j MANUALLY CLOSED, to service.

I a. DECRIDE how to verity the operability of an MOV in a safety system that has been manually closed. (0.5)

{ b. DESCRIBE the malfunction that this operability check is

intended to identify. (0.5) i 1

i OUESTION 4.04 (1.50) i STATE the criteria that are used to determine if an MR should be j checked EMERGENCY, IMMEDIATE ATTENTION, or ROUTINE MAINTENANCE.

(i.e., Define each classification)

]

l l

2 i (***** CATEGORY 04 CONTINUED ON NEXT PAGE.*****)

> d2__EBgCgDyBES_ _NOBM862_8BNgBM862_EMEBGENCy_8ND PAGE 26 RADIOLOGICAL CONTROL

, QUESTION 4.05 (2.00)

MATCH the Color of the Tags in CLOUMN A to the appropriate Usaege in COLUMN B.

COLUMN A COLUMN B

a. Red Tag 1. Used to identify the isolation boundaries of the clearance.

, b. White Tag

2. Used as a master tag for the

c. Blue Tag clearance - installed on the main j control point to isolate equipment i d. Yellow Tag from~all sourcec of energy.

1 1 3. Used to call attention to the existence of unusual circumstances

- attached to equipment where ab-

! normal operating conditions exist.

l

4. Used to identify equipment or i controls when the equipment is to be operated by any person other than its operator.

i QUESTION 4.06 (1.00)

Concerning a Plant Startup:

a. Step A.12 of GOI 100-1 states:

, "RHR system in standby readiness condition in accordance with 01-74" TRUE OR FALSE 1

! Step A.12 may be signed off with RHR pumps A & C operating to provide Shutdown Cooling. (0.5)

!_ b. Step A.22 states:

f " Reactor recirculation system ready for service, in accordance with 01-68" i

i TRUE OR FALSE I

Step A.22 may be signed off with Recirculation Pump A Out of Ser- -

vice for Shutdown Cooling requirements (only). (0.5)

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

i

.=..,-.r-,-,. -,

--,--% ,- - , _ . - ,y. -- -, .yy-y , .-.-, , r--,. . , , ,,.,-,.% .-4

4 2__EBgCEQUBES_ _UgBU862_6009800(2_EUEBQENCY_,30Q , PAGE 27 BODI96991Ce6_C9dIB96 I  %

QUESTION 4.07 (2.50) .

LIST the five (5) entry conditions to EDI-2, " Primary Containmant Control". (Include Setpoints, as appropriate). -

(2.5)

, _s QUESTION 4.08 (2.00)

STATE which Emergency Classification is appropriate f or the-f.ollowing definitions.

a. Events are in progress or have. occurred which anvolve actual or potential substantial degradation of the level'of safety of _

the plant. '

b. Events are in progress or haye occurred wnich could develop into, or be indicative of, more sericas condi 31o'is which, are not yet fully realized.

c. Events are in progress or havG' occurred which invo;ve actual or imminent substantial core failure with the pcton-tial for loss of containment integrity.

d. Events are in progress or have occurred which invol;u ,

an actual or likely major failure of plant functions needed for protection of the public.

s s

+1

.\

(*rw** CATdGDRY 04 CutlTINUED ON NEXT PeibE * * *

• a- )

, _ ~ _ $_ . .* .-_ _ _ _ q

4 2__ PROCEDURES-- NORMAL 2_6DNgRMAL 3_EMERGENgY_AND -PAGE 28 B09196991G06_ggNIBg6 j'

i f'

QUESTION 4.09 (1.00)

I J Concerning the Routine Radiation Work Permit (RRWP):

/ FILL IN THE BLANK FOR EACH OF THE FOLLOWING 1-

a. No one is permitted to receive a dose greater thar. _________ per day on any RRWP or any combination of RRWP's.

b. Personmnel with an exposure limit of ________ per quarter or

( less will not be allowed to enter on an RRWP.

g;. No RRWP user will exceed ______ of their remaining allowable dose j (RAD) in one day as indicated on a current personnel listing, except by upecial permission of the Health Physics section supervisor or an assistant Health Physics section supervisor.

d.. Entries into dose rates of ________ or greater will require the approval of Health Physics supervision (see RCI No. 1) and the pre-i sence of a Health Physics technician. Additional HP surveys may be required prior to entry. These requirements will be noted on the RRWP.

QUESTION 4.10 (1.50)

Contarning operations with Feedwater Heaters removed from service,

.aa per 0I-2, Condensate System G. STATE whether kilowatt oittput will INCREASE / DECREASE / REMAIN THE SAME when the highest pressure feedwater heater is removed from ser-vice. JUSTIFY you answer. (0.75)

b. STATE whether kilowatt output will INCREASE / DECREASE / REMAIN THE SAMr when other than the highest pressure feedwater heater is removed f rotn service. JUSTIFY your answer. (0.75)

QUESTION 4.11 (1.50)

STATE the thren (3) conditions which must be present to allow the Unit 1 and 2 Dx.auel Generators to be paralleled with the Unit 3 Diesel Generater.

A '

/ f l' (***-**

i f CATEGORY 04 CONTINUED ON NEXT PAGE *****)

i  : U

\ ll i g

1 '

s 4'.-

,) -

/

/. g n - '

3 t__ EB99E puBEg_;_ NgBd862_8 p NgBd862_Ed EBG E NCY_8Np : PAGE '29 E8DI969GI986_99NIBg6 QUESTION 4.12 (1.00)

Per 0I-24, " Raw Cooling Water":

a. STATE the header pressure at which the STBY (spare) RCW pump will automatically start. (0.5)

6. 'OI-24 cautions the operator NOT to run the 3D RCW pump in automatic. STATE the reason for this caution. (0.5) s GUESTION 4.13 (1.00)

With Unit 2 operating at 90% power, an Off Gas Hydrogen EH-2]

High alarm.is received. Both Hydrogen analy:ers indicate a Hydrogen level of

• 5.3%. Which one of the following responses most accurately reflects the proper course of action you should take?

a. Change over to the alternate Recombiner.

b. Change over to the Alternate Off Gas train.

c. Start an 'sd,di ti onal SJAE to assist in Condenser H-2 removal.

d. Manually Scram the reactor.

1 OUESTION 4.14 (1.00)

STATE why the Field Breaker should be left in the TEST position when tagging a Recirculation Systehi MG Set for maintenance.

QUESTION 4.15 (1.00)

The RPV Head is removed and a High Decay Heat condition exists. -One loop of Shutdown Cooling has been established to control temperaturo.-

STATE the actions which should be taken.if-Shutdown Cooli-ng is last-and the vessel"t'emperature increases ~to. greater than 150 deg F.

4-QUESTION 4.16 *

(1.00) ,

g.

Per the EDI's, DEFINE " Adequate Core Cooling".

~ l' ,

~,: s ._ _

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

di__BBgCgDuBgS_ _N9Bd862 0DNgBd862_gdggggNCY_8ND PAGE_ 30 BSDIg69@lC86_CgNIBQL

OUESTION 4.17 (1.00)

EDI I-1, RC/L states:

_If any control rod is NOT inserted to or beyond position (02), disable ADS. auto blowdown function (reference Appendix 3) and enter.C5 (LEVEL /

POWER CONTROL)

STATE the basis for disabling the ADS auto blowdown function in-this plant condition. (1.0)

QUESTION 4.18 (2.50)

LIST the five (5) entry conditions to EDI-1, "Rx Control" (Include Setpoints, as appropriate). .(1.0)

QUESTION 4.19 (1.50)

An Operability Test is conducted on a safety-related system, following the installation of an approved modification. LIST three (3) sep-arate criteria which would procedurally require that a TEST DEFICIENCY be initiated (documented).

l l

t l

i 1

f j

i i

(***** END OF CATEGORY!O4 *****) 1

(************* END OF EXAMINATICN ***************)  ;

t

. . ~

1. PRINCIPLES OF NUCLEAR POWER' PLANT OPERATION2 PAGE : 3'1

~IUEBMQDYNGMICB2_SE61_IB8NSEEB_8ND_ELUID_EbgW

~

ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN,' K.

ANSWER 1.01 (1.00) d

, REFERENCE i Mollier Diagram (Steam Tables)

BFNP: HTFF, Chapter 4, L.O. 6.2 e

A BWR K/A: 293004 K1.09 (1.7/1.8) i ANSWER 1.02 (1.50)

)

a. c  ;( 1. 0)

b. (Excessive voiding in the bypass region resulting in) unreliable LPRM readings (0.5) (1/2 credit for LPRM overheating) (0.5) i,

.j REFERENCE General Electric Heat Transfer & Fluid Flow, Chapter 8 i

l BFNP: HTFF, Chapter 8, L.O.'s 9.4 & 9.5 1

1 f

BWR K/A

293008 K1.32 (2.5/2.6), K1.33 (2.4/2.6) 215005 K5.02 (2.7/2.8) i 1

r 1

l

E ili__EBINQlELES_QE_ NUCLE 88_EQWE8_E68NI_QEEB8IlON2- -- PAGE '32-ISEBMODYN8MIGSx_UE6I_IB8NSEEB_8ND_ELylD_E6QW ANSWERS - . BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 1.03 (2.00)

a. P -P (0.5) act sat

b. (1) INCREASE (More Subcooling at the pump suct' ion)

(2) ' DECREASE (Reduced pressure at the eye of.the pump results in being closer to saturatio pressure)

(3) INCREASE (Further to saturation. temperature and increased density causing less static head) (3 @ O.5 each)

REFERENCE General Electric Heat Transfer and Fluid Flow, Chapter 6; BFNP: HTFF, Chapter 6, L.O.'s 7.7, 8.1, & 8.2 t

BWR K/A: 293006 K1.10 (2.7/2.8)

ANSWER 1.04 (1.00) i c

I REFERENCE j General-Electric Reactor Theory, Chapter 5

DFNP: Rx Theory, Chapter 5, L.O. 2. 4 ~

t i

l BWR K/A: 292005 K1.09 (2.5/2.6) c

. . . - - , ,,- - - ,__,--m.--._-,-, __.-.,_.e _ _ _ - - - . , , - , . , ,

~

. 1,_ LPRINQlELES_QE_NUQ6 EAR _PQWEB_P68NI_QPEBAllgN2 PAGE' 33 ISEBdQQYN8diQS2 _SE8I_IB8NSEEB_8Mp_E6Ulp_E6QW ANSWERS --_ BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.'

ANSWER 1.05 ( .50)

Reactivity GTE-Beta-effective REFERENCE General Electric ~ Reactor Theory, Chapter 3 DFNP:~ Rx Theory, Chapter 3, L.O. 5.6 BWR K/A: 292003 K1.07 (3.3/3.3)

ANSWER 1.06 (1.50)

a. The decrease is due to the (decay of Xe-135 and the) increased burn-out (which is immediately seen by the EXe]), being much greater than the Xe production, (which i s time-delayed f rom the lower power. )

b. The flux will redistribute from outside (and higher) peaked and center (and lower) depressed to center peaked and outside depressed.

(0.25) This is because the Xe concentration shif ts f rom center peaked to more peripherally peaked due to power production distribution. _ ( 0.' 25 )

c. Use caution when pulling peripheralfrods (0.25) due to the

increased rod worths (increased thermal flux) (0.25). (0.5 each)

REFERENCE General Electric Reactor Theory, Chapter 6 BFNP: Rx Theory, Chapter 6, L.O. 2 BWR K/A: 292006 K1.07 (3.2/3.2), K1.08.(2.8/3.0), K1.10 (2.9/2.9)

& K1.14 (3.1/3.3) 4 t -

+ -- -

-- a , . - - - , , 3 --

e

11__EBING1ELEG_QEiNUQ6E8BiEQWEB_E68dI_QEg88IlgN2 PA'GE i34

? :IdE8dQDyd8dlGS2_Bg81_IB8NSEE8 AND FLUID FLOW -1 l

2 ANSWERS'-- BROWNS-FERRY.1, 2&3 -86/12/01-BROCKMAN, K.

l 7 ANSWER' 1.07 (2.00)

a. The negative reactivity added by the increased' voids-generatdd by the rod' withdrawal is-greater than the positive reactivity added by the reduced rod absorbtion. ( 1.' O )

-b. (4) (1.0) i  ;

REFERENCE General Electric Reactor Theory, Chapter 5 i

l BFNP: Rx Theory, Chapter 5, L.O. 3 1

1 i

I i

J BWR K/A: 292005 K1.04 (3.8/3.8)

ANSWER 1.08 (1.50) j a. LTE 13.4 Kw/ft

-f b. GTE 1.07 i

c. LTE 3293 MWt (8-hour) -OR- LTE 3358 MWt (Inst) i

d. LTE 1375 psig

e. LTE 823 MWt

f. LTE 100 deg F/ hour (0.25 each)

REFERENCE

General Electric Heat Transfer and Fluid-Flow, Chapter 9 BFNP

Technical Specifications, L.O.'s 4 & 34;~HTFF,; Chapter 9,-

L.O.'s 2.2,-3.2, & 4.2; GOI 100-1 BWR K/A: 293007 K1.05 (3.3/3.5), K1.10 (2.7/7.9) i 2

--L__L__-_.___.-- __ . _-..___..____ _- _. .._--__._ _____.- -._.--_ _... _ - -_~-- . - __- - -._____ _. - _ L _ .

ig__ PRINCIPLES:OF NUCLEAR POWER' PLANT OPERATION 1- PAGE 35:

-IBEBdQDYN6dlCS2_BEGI_IBedSEEB_8dD_E6MID_E6QM ANSWERS -- BROWNS FERRY 1, 2&3- -86/12/01-BROCKMAN, K.

~

293009 K1.06 (3.4/3.8), K1.21 (3.1/3.6), K3.8 ( 2.' 7 / 3.1 )

i ANSWER 1.09' (1.75)

~ ~

a. Period = (Beta - Rho) / (Lambda

• Rho)'

= (0.0072 - 0.0006) /-(0.1

• 0.0006) = 110 seconds (0.5)'

b. Period.=-1.443

• DT -> DT = Period / 1.443 = 110 / 1.443

= 76.23 seconds (Error carries forward from-(a)) (0.5)

c. Period will be SHORTER (0.25) This is'due to the fact that the Beta for EOC is shorter (* O.0058, due to the increased Pu j contribution). (0.5) (0.75) i REFERENCE i

General Electric Reactor Theory, Chapter 3

)

BFNP: Rx Theory, Chapter 3, L.O.'s 3.6, 4.6, & 5'9 .

r BWR K/A: 292003 K1.08 (2.5/2.8),_K1.09 (2.5/2.6)

L S

b t-r

11__EBINCIE6ES_gE_ NUCLE 88_EgWEB_E68NI_QEEB8IIQN 3 PAGE. 36 IHEBdgpyN8dlCS3 _HE8I_IB8NSEEB_8ND_ELUID_E69W ANSWERS -- BROWNS-FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 1.10- (3.00)

a. . Recirculation Pump Runback (From < 20% FW Flow) (0.5)

b. EHC controlling reactor pressure (at 920 psig) (0.5)

c. EHC controlling reactor pressure (at 920 psig) (0.5)

d. In fease in the suction of the Recirc Pump (f rom Downcomer) (0.5)

e. Reactor Scram ( APRi1 Flow Biased @ 0.66*w + 54%) (0.5)

f. K-f is utilized to increase the steady-state MCPR limits at

. reduced flows (FLCPR Program). (0.5)

REFERENCE BFNP: Transient Analysis LP, Transient # 5, L.O. A0 B BWR K/A: 202001 A2.05 (4.0/4.r 293009 K1.26 (3.3/3.3) K1.28 (3.5/3.5)

ANSWER 1.11 (1.00) a REFERENCE General Electric Heat Transfer and Fluid Flow, Chapter 8 BFNP: Thermal Hydraulics L/P, p 6; HTFF, Chapter 8, L.O. 6

~l l

i l

l l

. = . -

-le - PRINCI'PLES OF-NUCLEAR POWER IPLANT OPERATIONS .PAGE .37:

IBE8dODYN8dlCS2_SE81_IB8NSEE8_8dD_E(y1D_E60W JANSWERS - BROWNS-FERRY 1,m2&3. -86/12/01-BROCKMAN, K.

m l

L ANSWER 1.12 (1.00) a REFERENCE

.BFNPi EGT106.~015-(MCD, Ch 1), L.O.'s '"B" & "D" ' -l 1

1 ii BWR K/A: 223001 A1.05 (3.1/3.3), SWG #7 (3.7/3.8)

PWG K1.15 (3.1/3.3), A1.14 (2.9/3.4)

ANSWER 1.13 (1.00)

^

This could result in a POSITIVE MODERATOR TEMPERATURE COEFFICIENT 4b).5).

The positive reactivity addition by a decrease in the density-of the moderator and the Baron (on a heat up) can be greater (0.25) than the negative reactivity addition from the increased

. leakage, reasonance capture, and control rod worth.(-0. 25) .

REFERENCE BFNP: OPL171.039; L.O. "G" BWR K/A: 211000 K5.01 (2.7/2.9) i 4

s 4

'ic__EBINCIELES_QE_Nyg6E88_EgMEB_EL8NI_QEEB8IlgN2 PAGE 38 nIMEBdQDYN8digSg_Ug81_IB8NEEEB_8ND_ELUID_ELQM I

-ANSWERS---l BROWNS FERRY I', 2&3- ~-86/12/01-BROCKMAN, K.

' ANSWER 1.14' ( 1. 00 ) '

. l'. Neutron embrittlement of the cladding.

2.' Thermally induced pellet growth.

I , 3. Inward motion of the cladding walls (creepdown).

L

4. Fission Product (Chemical) embrittlement of the cladding 1(f rom Cd/I) .

~5.' Fission Product Gas Buildup (Pressure) (2 @ O.'5 each)-

j REFERENCE General Electric Heat Transfer ~and Fluid Flow, Chapter 9 GGNS: MCD, PCIOMR,' p 7 EIH: L-RO-673 BFNP: HTFF, Chapter 9, L.O.'s 7 & 8 i

BWR K/A: 293009 K1.36 (2.8/3.4), K1.37 (2.6/3.3)

ANSWER 1.15 (1.00)

C REFERENCE General Electric Reactor Theory, Chapter 3 i

HBR: Reactor Theory, Session 42, pp. 3&4 DPC: Fundamentals of Nuclear Reactor Engineering I

BFNP

Rx Theory,1 Chapter 3, L.O.'s 3.4, 3.5, & 3.6 PWR K/A: OO4/OOO-K5.08 (2.6/3.2)

BWR K/A: 292002- K 1.- 11 _ (3. 2/3. 3 ) , K1.141(2.6/2.9) 292003 K1.01 (2.9/3.0),~K1.08 (2.7/2.8) 1 h

y -

~

• g .~,y -e

1 .- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION 2- PAGE 39 ISEBMgDYN8MICg2_UE8I_IB8NSEEB_8NQ_E6UID_E69W.

~ ANSWERS.-- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

,- ANSWER 1.16 (1.50)

a. Doppler Coef ficient. (0.25) The rod insertion will be first "seen" by the adjacent fuel rods. (The immediate flux changes will cause fuel temperature to change before the effects are'seen in moderator

temperature or void fraction.) (0.5)

b. Void Coefficient (0.25) The MSIV Closure will immediately cause a pressure increase which will result in a change in the void' fraction.

g (This change will be "seen" before any changes-in moderator temperature or fuel temperature.) (0.5)

REFERENCE General Electric Reactor Theory, Chapter 4 i BFNP: R:: Theory, Chapter.4, L.O.'s 4.3 & 6.3 1

1 1

BWR K/A: 292004' K1.08 (2.2/2.4), K1.11 (2.5/2.6)

ANSWER 1.17 (1.50)

a. Higher l

b. Lower ,

c. Lower. (0.5 each)

. REFERENCE ~

General Electric Heat Transfer and Fluid Flow, Chapter 7

{ EIH: L-RO-667, p 10 BFNP: Rx Heat Balance LP; RQ 85/03/05 l HTFF, Chapter 7, L.O.'s 5 & 6.4; HTFF, Chapter 8, L.O. 5.1 BWR K/A: 293007 K1.11 (2.6/3.1), K1.13 (2.3/2.9) l t-~.,, , - - , , -,, ,.- - .+ w . - - .

, , _ ~ ~ ~ , 3 ew_ % .,2,,

Iz__ERINCIELES_gE_NgCLE68_EgWEB_E(6NI_QEgB611gN2 'PAGE 4 0 --

,IME6MggyN6MICS2_SE6I_IBONSEEB_6ND_E691g_E6gW t

ANSWERS -- BROWNS FERRY.1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER l '.18 - (1.00)

Cooldown-Limit = 90=deg F/ hour (0.25)

-Moderator Coe'fficient

= - 1 E-4 Rho /deg F _( 0. 20)J Rho =-(1 - K-eff)/K-eff *= K-eff - (0. 20) .

l . ..

Time = E(Rho / Mod Coeff)/Cooldown Limit] (0.25) i

= 1.11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> (67.3 min) -(0.10) 4 REFERENCE-General Electric Reactor Theory, Chapter 4 ,

i i BFNP: Rx Theory, Chapter 4, L.O. 2.4 7

d BWR K/A: 292008 K1.01 (3.8/3.9)

292004 K1.04 '(3.3/3.4), K1.11 (3.7/3.8)

ANSWER 1.19 (1.25) a a. More

b. Decrease i

c. Less

d. Increase

e. Increase (0.25;each)

REFERENCE General Electric Reactor Theory, Chapter-5 BFNP: Rx Theory, Chapter 5, L.O. :2 . 4

' -I l

l l

5

'r-.

-, , n ,, --

% __EBINGIE6ES_QE_NUC6E88_EQWEB_E68MI_QEg88IlgN 2 PAGE 41' THERMODYNAMICS2 _ HEAT _IBANHEEB_8NQ_ELUID_ELQW

'.'ANSWERS -- BROWNS FERRY 1', 2&3 -86/12/01-BROCKMAN, K.

-BWR K/A: 292005 K1. O9 (2.5/2.6) , K1.10 (2.8/3.3) i i

i q

l 4

2 J

k

.,a

_____.-----_,.__--.._-_-__-___.-____.--.--__..l_-_

2i_;ELONI_QESIGN_ INCLUDING _S6EgIY_6ND_EdgBGENCY_SYSIEMS . PAGE' 4:2 LANSWERS - -BROWNS FERRY;1, 28<3 -86/12/01-BROCKMAN. K.

. ANSWER 2.01 (1.50)

~

^

a.- (3)' 600 gal b.- (1) _65 psig

c. (4) 1250 psig (0.5~each)

REFERENCE-BFNP: OPL171.005,.L.O.'s "P", "Q" , 8< " S "

BWR K/A: 201001 K4.11 (3.6/3.6)

ANSWER 2.02 (1.00)

a. An open relief valve would decrease the:available head for pump.

cuction (NPSH) (and might cause air binding of the pump.) - OR'-

Loss of inventory (water) from CST or Torus. (0.5)

b. TRUE (0.5)

REFERENCE BFNP: OPL171.040, L.O.'s "C" 8< " D "

l

-1

l

'l

~

l BWR K/A: 217000 K1.01-(3.5/3.5), K4.05 (4.2/3.5),- K5.01 (2.6/2.6), i A2.11-(3.1/3.2),. A3.04'(3.6/3.5) i 1

i J

1

2x__E66NI_DEgl@N_IUC69 DING _SOEgIy_8ND_EMEBQENCy_gySIEMS. PAGE. 43 i' -

, . ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

i l

I ANSWER 2.03 -(2'25)

a. WILL INJECT (0.25)- HPIC injection will be at less than design flow (0.50) (which may not meat the high pressure-coolant. makeup

requirements of certain accident conditions). - OR - Increase in -

Torus-level (due to HPCI pumping thru MFV) - OR - Loss of CST Level

to the Torus, j

j b. WILL NOT INJECT (0.25) Minimum flow protection will be lost-or the controller will keep the turbine speed at minimum (0.50).

1 I c. WILL INJECT (0.25) There is a potential for a (loss of vacuum in

the HPCI Darametric Condenser and) release of airborne activity to the HPCI Room (0.50).

l REFERENCE BFNP: OPL171.042, L.D. "A.8"; Accelerated Requal Exam 7.10 1

T i

i i

I f BWR K/A: 206000 K6.06 (3.1/3.2), A2.03 (3.5/3.5), A2.14 (3.3/3.4),

l A2.15 (3.4/3.5) l

!t l

l ANSWER 2.04 (1.00) i During flushing and Shutdown Cooling operations, the RHR Minimum Flow Valve could open diverting reactor coolant to the Torus. (0.5)

(OI-74 has been revised to) require the the Minimum Flow Valve  ;

Handswitch be placed in the BYPASS position prior to flushing and Shutdown Cooling operations. (0.5) f

REFERENCE j BFNP: OPL171.044, L.O. "E.2"; 0I-74; General Electric Systems Review on RHR I,

j t

BWR K/A: 205000 K1.08 (3.9/3.9), K3.02 (3.2/3.3), K4.07 (2.7/2.8)

K5.02 (2.8/2.9), A2.10 (2.9/2.9),.A4.05 (3.2/3.2) i i

4

2.. PLANT DES'IGN INCLUDING' SAFETY'AND EMERGENCY SYSTEMS PAGE- 44

' ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN,-K. ,

ANSWER 2.05 ( .50)

Not having the governor.at the lowest setting can result in a mechanical overspeed trip of the diesel.

REFERENCE BFNP: LER 86-26 (7/28/86 - EDG "D")

BWR K/A: PWG (3.2/3.6)

AN5WER 2.06 ( .50)

The Field Breaker would have to be (locally) reset, and the (HEA) Lockout Relay would have to be (locally) reset. (0.25 each)

REFERENCE BFNP: OPL171.038, L.O.'s "E" & "F"; OI-82, P 10 BWR K/A: 264000 K4.01 (3.5/3.7), A1.09 (3.0/3.1), A2.02-(3.1/3.1)

12t__ELOUI_ DESIGN _INC6gDlbG_S8EEIY_8BD_EdEBGENCY_Sy@IEUS. PAGE 45.

ANSWERS -- BROWNS FERRY 1,.2&3 -86/12/01-BROCKMAN,'K.

ANSWER 2.07' (1.50) ai (Greater than 40% mismatch between the generator output and the turbine power) as sensed by generator stator amps.and turbine cross-over pressure. . (0. 5)

6. (1) TCV's Trip Closed (fast-acting solenoids)

(2) _ Bypass Valves Open (3) EHC Load Selector runs back

. (4) Control Valves partially reopen.@ < 40% mismatch (5) CIV's throttle to control turbine speed @ 1800 rpm (0.20 each) 4

'j-REFERENCE BFNP: OPL171.010, L. O. 's "J" & "K" 1

BWR K/A: 245000 K3.01 (3.4/3.7) r l.

4 ANSWER 2.08 (1.50)

a. No Response (Redundant DC solenoids keep valves open) 5 I

b. Outboard MSIV's Close

c. Inboard MSIV's Close (0.5 each) l REFERENCE BFNP: OPL171.OO9, L.O. "E"

}

i-

, BWR K/A 239001'K6.01 (3.1/3.3), K6.02 (3.2/3.2) 1 i

1 1

J2.~LE66MI_QESIGN_ INCLUDING _SOEEIY_8NQigMgBGENCY_SY@IEMS JPAGE' 46 ANSWERS -- BROWNS FERRY;1,'2&3- -86/12/01-BROCKMAN, K.

1 ANSWER 2.09 (1.50)

Reactor Water Level (0.25) GTE - 39" (0.2L)

-Drywell Pressure (0.25) GTE + 1.96 psig (0.25)

SELECT - RESET Switch (0.25) in SELECT (0.25) i REFERENCE BFNP: OPL171.044, L.O. "E.9" 4

4 i

BWR K/A: 226001 K4.03 (2.9/3.1) 203000 K4.10 (3.9/4.1)

ANSWER 2.10 (1.00) b REFERENCE BFNP: OPL171.045, L.O. "I"

~

BWR K/A: 209001 K1.08 (3.2/3.4), K1.10 (3.7/3.8), K2.03: (2.9/3.1)

K4.08 (3.8/4.0), K6.01 (3.4/3.4) 3 t

ie f

z

12c__E68MI_DEglGN_INC6QDING_S6EgIY_@ND_ENEBGENCY_SYSIENg PAGE - 47;

' ANSWERS -- BROWNS FERRY 1,'2&3 -84/12/01-BROCKMAN, K.

1

. ANSWER- 2.11 (1.50)-

1 1

~ a. Yellow light above_the Core Spray pump switch (on Panel 9-3).

- b .; Initiating conditions cleared (GT -114.5" RVL & LJT 2.45 psig DWP).

Reset Core Spray Initiation signal. (0.5 each)

' REFERENCE BFNP: OPL171.045, L.O.'s "F" & "H"'

BWR K/A: 209001 K2.03 (2.9/3.1), K4.08 (3.8/4.0), A3.02 (3.8/3.7) l l ANSWER 2.12 (1.00)

a. DEPRESSURIZE (DWP is a seal-in signal)

b. NOT DEPRESSURIZE (RVL is not a seal-in signal) '(0.5 each) i i

REFERENCE BFNP: OPL171.043, L.O. "E" 1

l 4

BWR K/A: 218000 K2.01 (3.1/3.3), K4.01 (3. 7/3. 9 ). , K4.03 (3.8/4.0) l K5.01 (3.8/3.8)

I

+

,7 - ,, - - . - . - - , . < -

r , ,,,-e..~

2___E68NI_DE@l@N_INCLUDIN@_E6EETY AND_EdEBGENCY_SYSIEUS PAGE 48 ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER- 2.13 (1.00)

The HPCI mechanical-overspeed automatical'ly resets-(@ LT 5000 rpm) (0.5)

The RCIC mechanical overspeed. trip requires local resetting of the Trip. Valve _(71-9) with subsequent reopening (local or-control room).. (0.5) i j

REFERENCE

-BFNP: OPL171.042, L.O. "D.2"; OPL-171.040, L.O. "B.1"

.i 4

BWR K/A: 206000 K4.03 (4.2/4.1) 217000 A2.02 (3.8/3.7) 1 l

l ANSWER 2.14 (2.00) l i

(1) Diesel started and at rated speed.

i j (2) All other supply breakers to the 4160 volt Shutdown Board Open.

(3) No supply breaker lockout (EDG).

(4) Undervoltage on the 4160 volt Shutdown-Board.- (0.5 each). .

l 1

} REFERENCE l BFNP: OPL171.038; L.O. "E"

{

l l

l

i BWR K/A: 269000 K1.01 (3.8/4.1), K5.05 (3.4/3.4)

P 4

i_.m__

12ai_E68MI_QESIGNildCLUDIN@_@8EEIY_8NQ_EMEBGENCY_SYSIEd@ . PAGE 14 9 ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

T i

l- . ANSWER' . 2.15 (1.50)

a. To. prevent creation of a vacuum in the tailpipe of the SRV (from condensation) following an actuation and reseat. - Torus water.would be' drawn into the tailpipe and a subsequent actuation could result in excessive forces placed on the downcomer. (0.5),

~

b. (1) Could result in uncondensed steam release to the primary containment.

(2). Could result in high radioactivity releases due to

. the loss of the Suppression Pool " scrubbing". (0.5 each) j Other responses, as appropriate.

J (e.g., Increase in Drywell Pressure / Temperature)

REFERENCE BFNP: OPL171.OO9, L.O. "B"

?

i f

BWR K/A: 239002 K4.03 (3.4/3.6), A2.01 (3.0/3.3)

ANSWFR 2.16 (2.00)

a. False

b. True 4

c. False

d. False (0.5 each)

REFERENCE BFNP

OPL171.044, L.O.'s "D" & "E"; OI-74 i

i

)

i I v

4 4

=w ,y r v r --

rr-- ,c,, .w-3 = 4 , --,.H- , r- -v-- -

' tm - 1r r

-- 2x__E66NI_DE@lGN_INCLUplN@_@8EEIYiGND_EMEBGENCy_SygIEMS PAGE? 50

~

ANSWERS ,-- BROWNS FERRY - 1, 2&3 -86/12/01-BROCKMAN, K.

A BWR K/A: 2030900 K1.14 (3.6/3.7), K4.01 (4.2/4.2)-

1 ANSWER 2.17 (1.00) t

a. 510 MVAR +- 10 MVar (0.5)

! b. 1060 MWe +- 40 MW (0.5)

'f O

1 REFERENCE l EIH: L-RQ-729

• i' BFNP GOI-100-1 1

I i

l BWR K/A: 245000 K4.07 (2.5/2.6), K5.02 (2.8/3.1)

ANSWER 2.18 (1.00) 4 Loss of Normal AC Power (0.25) la conjunction with

.; an Accident Signal (0.25) -OR- Initiation of 480 v Load

Shed Logic (0.5).

3 --

Low pump discharge header pressure (0.50)-

i REFERENCE j -BFNP: OPL171.047, L.O. "D"; OI-70. L.O. "B" i

I DWR K/A: APE 295018 AK1.01 (3.5/3.6), AK2.01 (3.3/3.4) i

)

i l

},

i, i

' l 1

i

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

.at__EL8MI_ DESIGN _INGLUQ1NG_g8 Eely _8ND_EUEEQENgy_@ySIgdg PAGE 51_ ,

ANSWERS -- BROWNS FERRY 1, 2&3 '-06/12/01-BROCKMAN,.K.-

) .

' ANSWER 2.19 (1.00)

I

't c

i' REFERENCE

USNRC BWR-4 Systems Manual, pp 3.3 3.3-10

EIH

HNP-x -1001 ; HNP-x-1286 BFNP: OPL171.011, L.D. "B" 4

i j BWR K/A: 259001 K4.04 (.2.5/2.6), A4.02 (3.9/3.7) 1 . ANSWER 2.20 (1.50) _

) a. None l b. Unit 1, System 2 l c. Unit 2, System 2 (0.5~each) 1

] REFERENCE I

BFNP: OPL171.044; Technical Specifications, L.O. 24 l

i i

1

.! BWR K/A: 203000 K4.06 (3.5/3.5) 1 f

i l

4

_.__________.___________i_____________.Z_________

& __ INSTRUMENTS'AND CONTROLS: PAGE~ :52 L

. ANSWERS --- : DROWNS - FERRY 1, 2&3 -86/12/01-BROCKMAN,'K. -

ANSWER -3.01 ( l'. 00 )

d 1

REFERENCE e BFNP: OPL171.008, L.O. "B.2" 1

i

l t- BWR K/A: 202002 K4.02 (3.0/3.0) K5.03 (2.4/2.4)

I i

I ANSWER 3.02 (1.00) i j (1) A (LIT 3-53) a 1

(2) Low '.(0.5 each)- ,

l REFERENCE j GGNS: OP-B21-501 BFNP: OPL171.003, L.O. "J"; OPL171.012, L.D. "E.3"

ARP 9 XA-55-5A-8, XA-55-5A-30, XA-55-58-4, XA-55-5B-5 t

i l BWR K/A: 259002 K5.01 (3.1/3.2) K6.5 (3.5/3.5) a i

4 l

4

-4

~

t

.---_..___--____.-_-.__----_.--.__--~____..-__.____-._.__..-_...

^. - _ _ . _ . _ - -.-.._-..._.---_.-..--L-_..

M__INSIBy!1ENIS_8ND_CQNIBgLS PAGE 53 ' -

ANSWERS -- BROWNS FERRY.1,-2&3 ' -86/12/01-BROCKMAN, K._.

ANSWER 3.03 (1.00)

a. .45.5 inches (+~ O.5 inches) (0.5)

6. .Yes (0.1) ' It has a Heat Transfer Clamp (0. 4) - (0.5)

REFERENCE BFNP: OPL171.OO3, L. O. 's "B" & "J" i

h BWR K/A: 216000 K4.14 (3.3/3.4); K5.10 (3.1/3.3).

ANSWER 3.04 (1.00) f I a. DEENERGIZED (0.5) i i HPCI and RCIC

b. (0.25'each)-

1 4

REFERENCE BFNP: OPL171.017,Section X.C; L.O..C & D I

i i

4 BWR K/A: 223002 K1.04 (3.5/3.5) K1.07 (3.4/3.6)- K2.01 (2.5/2.7)

J l

i

+

4 w _._w.__ u- --..-__.-_m _____.____-w . _ . _ _ . _ - _

P

. 3[__INSIBydENIS_8ND_CQNIBQ6S PAGE- 54 IANSWERS'-- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 3.05 (1.00)-

a. FALSE -(0.5)

b. It allows for compensation for the U-5 depletion- (0.5)'

(1/2 credit for " calibration")

REFERENCE BFNP: OPL171.021,Section X.B.1 & .4; L.D. C BWR K/A: 215005 K4.06 (2.6/2.8) 1 ANSWER 3.06 (1.50) 4

a. O seconds (+3,-0) to 30 seconds (+6,-6) ' (0. 5)

b. The rate of the power increase determines the time delay (f ast rise = short delay). -OR- The. absolute value of the power increase (P (f ) - P(i)) (0.5)

c. BOTH OPEN AND CLOSED (0.5) l REFERENCE BFNP: OPL171.022,Section X.D.6 & .7; L.O. F&K P

BWR K/A: 215005 K4.02 (3.7/3.7)

.a >

l

', \'

3. INSTRUMENTS AND l'pNTROLF - - -

Pn c' j'i5 ANSWERS -- BROLJNS FERRY 1, 28<3 -86/12/01-YROCKMAN, K. -

. . ~

9 ANSWER 3.07 ^<.PO) -

s .x NO All requirements for a start perm ssive U-e not gy t . ( 0. P (0.5)

(There is a 30 second TD on steam pressure .tn' order for the PCV't% w to open and get pressure up to the setpoint - it may or -mayc.,tf cause 3 the FCV tc close. The Condensate outlot valve is closed,. ths, ..

condensatt siischarge pressure is i n ad aq . tat e , and.the SJAE

~

"D" \

steam MOV i s -epen ) ' '

4

- _' l REFE'r.'EN CE -

BFNP: OPL17L.030, ~~Section

. B.'5.g, L.D. 'C" s s.

e '

BWR K/A: '?"6000 K4.10 (2.7/2.7)-K4.15 , 3.1/3.0) ,. .

r ,1 .

ANSWER 3.08 (1.OOs A Rod Block woulci be received. ( 0.;,5 ) il e Averaging Circutt Jm not giained down to, th e ref erence APRM value; therefore, the 6% pos v increase will t; c u.rsed as exceeding ?he rod block setpo. int '

.t .6B A w

+ 24%). (0.5)  %

,, 1 f

, s REFERENCE DFNP: O Pl. 1 7 1 . 0 3 5 ; Ecct!cns X . A . 3 !< . B., 6 i L . O . ' s b u 'F

_ t m

D'!R K/A: 215002 Ki 04 (3.0/3.0), K4 . 0.1 . ( 3 . 4, -' . 5 ) , k 4.M ( 2. 9:> 7. 0) ,

I 215005 4T2 (3.7/3.7), K4.07 ( 3 . 7 /., . 7 : '

-T I

4 -

i h O N

\

/

~

4 I

.~.

0-w y y s vDF -- ^

- ~ -

, e I

s

+ A l_INEil69[gNISiy1D_GONIBOL S PAGE 56

,.4b541ERS -- BOO'WS FhRRY 1, ( 2%.3 y ,

. ,)

--86 /12/01 -BROCKMAN , K.

, i s,r - \

I t

(

i-  !,'

.i '.

AN5'JCR '5.09 l (1.50b ,

!. / ,

a.

SDJ'Inbd Isol Valve Closes (FCV 74-53) (0.25)

SLC Suct Supplyila'7ees Close (FCV 74-47 & -48) (0.50)

RHR Pump A T(ips (0.25)

The Inbd Loop Val ver .6 3 sit PD (XS74-126) must be depressed

/s b.

., (0.50) 4EFERENCE F DFNP: OPL171.044; L.D.'s D.7,' E, .'l F ; OI-74,Section III.F & .J 4

1

, A Db!9 , ' + - 305000 K t. 03 (3.8/3.8) '

i/

't ANSWER 3.10 '

1.53) 1

-- "A.

A1) b

'2) c (3) a (0.25 each)

5. I' . r: oump(s) auto start raw'dTter Storage Tanks Is ol att Daw Su-r o ce later Puracu Shutdown ( b -. interlock). (0.25 each)

REFERENCC BFNP: GPL171.049, Sections X.B.r.(2) & .(3) & .D.2; L.O. E a.

/

A

(

B;,R K/4: 286000 s K5.06 (2.6/2.7), K5.07 (2.6/2.7), K4.01 (3.4/3.6)

K4.02 (3.3/3.5)

' i j ,

e i

\

s I

f -

t k

i k.

~~

,,' {g D E * -

2 4k__10SIByMEUI'd_8MD_QQUIBQ6S -PAGE 57

- ArdSWERSt--

-BROWNS FERRY-1,52&3. -86/12/01-BROCKMAN, K.

ANSWER 3.11 (1.00)

a. VALID-

-b. INVALID

.(0.5 each)

REFERENCE General Electric, NEDE.24810, Volume it2 BFNP: HTFF,. Chapter 9, L.O. 5.g BUR K/A: 293009 K1.38 (2.7/3.1) e ANSWER 3.12 '(1.00) a.

SATISFACTORY (0.5) - The Test Failed light indicates that the GNC test sequence has halted because of-a~ failed comparison-(which is'the  ;

desired result) . (0.5)

REFERENCE BFNP: RSCS LP; RQ 85/02/03; OPL171.025, Section B(10) l DWR K/A: 201004 A3.05 (3.5/3.7) l i

s .

Y f3. INSIBUdENIQ_6ND_CQNIBQLS~ '.PAGE- 58

[' ANSWERS - BROWNS FERRY-1,'2&3 ~-86/12/01-BROCKMAN, K.

i -ANSWER '3.13- (2.00) a.. (2)

b. (2)

-(1.O'each)

REFERENCE-1BFNP: ARP 9-4, XA-55-4A;s OPL171.OO7

.BWR K/A: 202001 K1.01 (3.6/3.7) 293008 K1.26 (2.9/3.1), K1.27 (3.6/3.7)

ANSWER 3.14 (1.50)

a. Limit Switch (0.25) and Mechanical Stop (Os 25)

b.

LTE 20%_Feedwater Flow (0.25) or Pump Di scharge Valve LT 90% Open -'(0. 25)

I c. Any RFP tripped (0.25) and Level LTE +27" (0.25)

T REFERENCE BFNP: OPL171.OO8; L.O. "A" 4

4 BWR'K/A: 202002 K1.08 (3.1/3.2), K1.09 (3.1/3.2), K4.06'(3.1/3.1) j A2.06 (3.3/3.3), A2.07 (3.3/3.3)

I

e 1[. . .INSIByMENIS_6ND_GQNIBOLS '39'

~

.PAGE

ANSWERS -- BROWNS FERRY 1, 28<3 .-86/12/01-BROCKMAN, K.

4 1

. ANSWER 3.15 (1.00)

C REFERENCE BSEP: RTN-033,-012; SD 26.2; SSM 19-2/3-B EIH: L-RQ-705, pp 18,.19; GPNT,.Vol. VII, Chapter 9.4 BFNP: Simulator. Malfunctions 8,.108; Operational Transient LP, #15;-

OPL171.014, L .' O . "E.1" BWR K/A: 259002 K6.05 (3.5/3.7), A2.01 (3.4/3.9) 1 ,

ANSWER 3.16 (2.00)

i. e ii. d 4

4 REFERENCE

BFNP

Operational Transient LP, #20; OPL171.012, L.O. "E"-

EIH: L-RQ-726 BSEP: RTN 026; HD 17-2/3-B, Section 3.2

} .

! BWR K/A: 259002 K6.05 (3.5/3.5) i i

4 4

4 F

,. . .r . . . . . , - , . _ . . _ , . , , . , - - - - . . ...-~1-. . . _ . . , _ , ,. , . , . , . . - - .

3 t_;1NSTRUMENI@_ANQ_QQNIRQL@ PAGEf 60^

- - ANSWERS .-- - BROWNS ' FERRY11 ', 2&3 -86/12/01'-BROCKMAN , K. ;

- ANSWER :3.17 (2.50) a.- Decrease _

b. Decrease '

c. ' Increase

d. Decrease

e. Increase (0.5 each)

REFERENCE BFNP: OI-68; OPL171.OO7 BWR K/A: 202001 K6.01 (3.5/3.7), A2.01 (3.4/3.9)

ANSWER 3.18 (2.00)

a. Increase (45%) (0.5); Speed Demand Limiter (0.5)

b. Decrease (20%) (0.5); Limit Switches on Bailey Positioner (Low Speed Mechanical Stop) (0.5)

REFERENCE EIH: L-RO-714, Figure 4.1 (8) ; GPNT, Vol V, Chapter 4.1 BSEP: SSM 10-2/3-A, Section 3.2.1.1, pp 25 - 30 BFNP: OPL171.OO8

.BWR K/A: 202002 K4.02 (3.0/3.0), A1.01 (3.2/3.2) k u_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _______m_m' _ _ _ . _ _ . _ _ . _ _ _ _ _ - _ _ _ _ _ _ _ . _ _ _ _ _ _ - _ __.__.__..__________s.___ _ _ _ _ _ _ . . _ _ . _ _ , _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ . _ . . . _ _ _ _ _ _ . . ___S.'__.

4 8

3._; INSTRUMENTS _AND_CgNTBQLg' PAGE 61-

. ANSWERS?- . BROWNS FERRY'~1, 2t<3. -86/12/01-BROCKMAN, K.

i t

1 -

-ANSWER- 3.19 (1.00)'

-Illuminates-backlighting for comnpanion. rods:(0.5),:regardless of their' sensed position.-(0.5)

{ -REFERENCE BFNP: OPL171.025, L.O. D" i  ;

3 s

i l BWR K/A: 201004 K4.06 (3.3/3.4), A3.02 (3.~1/3.1 A3. 05 (3.5/3.7)

ANSWER 3.20 (2.00)

a. (1) Speed (f requency) Control

{ (2) Load Control t

]

b. (1) Voltage Control (2) VAR Control (0.5-each)

REFERENCE BFNP: OPL171.OB3 1

4 4

4 4

BWR K/A: 264000 K5.05 (3.4/3.4), A4.01 (3.3/3.43), A4.04 (3.7/3.7) i

)

1 I

2

!di__EBgCEQUBES_;_N98M862_6BN9BM8L2 _gMEB9ENgy_6N9 PAGE= 62 86DJ96991C66_CgNIB96

ANSWERS -- BROWNS FERRY 1, 28<3 -86/12/01-BROCKMAN, K..

ANSWER 4.01 (1.00)

Venting the suppression chamber first takes advantage of'the'" pool scrubbing effect" (0.5) to help minimize the-radiation released (0.5).

REFERENCE BFNP: EDI-2, PC/P-5, L.O. "B.6" t

BWR K/A: 259024 EK3.04 (3.7/4.1), EK3.07 (3.5/4.0)

EA2.03 (3.8/3.8), EA2.04 (3.9/3.9), EA2.08 (3.6/4.0)-

ANSWER 4.02 (1.00)

Maintaining pool temperature and reactor pressure < HCTL assures sufficient heat capacity in the Suppression-Pool (0.5) to prevent '

equipment failure (0.3) due to unstable steam condensation (0.1) during an ADS blowdown of the reactor vessel (0.1).

REFERENCE BFNP: SP/L, p 3; SP/T, p 1; OPL174.657; L.O. B.7 DWR K/A: 259024 EK2.08 (4.0/4.1), EK3.04 (3.7/4.1)

EA2.04'(3.9/3.9), EA3.06 (3.4/3.9)

4 1__EBggEgUBES_;_NgBM862_8BNg8M862_EME89ENCy_8ND PAGE 63 B6 Dig 69g1C66_CgNIBg6 ANSWERS -- BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

ANSWER 4.03 (1.00)

a. It shall be cycled by the motor operator (0.5 ) (pri or to being declared operationall (0.5)

b. Torque switch could become activated before the valve is off its seat (improper operation of the opening circuitry). (0.5)

REFERENCE BFNP: OSIL No. 47, p 2; L.D. "T" BWR K/A: 291001 K1.07 (3.4/3.4)

ANSWER 4.04 (1.50)

EMERGENCY - Action is required to prevent imminent major equipment damage or to protect personnel from any imminent threat of bodily injury (0.35/O.15)

IMMEDIATE ATTENTION - Work is 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 /> (, the next scheduled workday, or upon completion of necessary technical evaluations or material procurement) (0.5)

ROUTINE MAINTENANCE - Work is to be performed as manpower and circ-umstances permit. (0.5)

REFERENCE BFNP: SP # 7.6, Section 4.12 - .14; L.O. "A" DWR K/A: 294001 K1.02 (3.9/4.5) 2

1

<4,  ::PBQGEQljBE@_ _NQBMAL2_8@NgBdAL 2_EMEB@gNCY_8ND- PAGE- 64

'88DIRLQelG86_GQNIBQL

. ANSWERS - . BROWNS FERRY 1,;.2&3 -86/12/01-BROCKMAN, K.

-ANSWER .4.05 -(2.00)

a. 2'

b. -1

c. 4

d. 3 -(0.5 each)

REFERENCE BFNP: SP 14.25, Sections 4.4 - 4.7 BWR K/A: 294001 K1.02 (3.9/4.5)

ANSWER 4.06 (1.00)

a. TRUE

b. TRUE (0.5 each)

REFERENCE BFNP: GOI 100-1, Sections A.12 & A.22 BWR K/A: 203000 K1.14 (2.4/2.6), SWG #5 (3.5/4.4), SWG #10 :(3. 7/3. 9) .

')

l

c CSA__EBOGEDUBES_;_NQBU861_8BNQBd862_EMEBGENQY_8ND PAGE: 6"L-188D106QGlQ86_QQUIBQL ANSWERS -- BROWNS FERRY 1', 2&3 -86/12/01-BROCKMAN, K.

L ANSWER 4'.07 ( 2. 50')

I 1) . Suppression Pool Temperature (0.25) > 95'deg F (0.25)

2) Drywell Temperature- (0.25) > 160 deg F ( 0. 25)'

3) Drywell Pressure (0.25) > 2.45'# (0.25) l Suppression Pool Level "

4) (0.25) > -1 (0.25)

5) Suppression Pool Level (0.25) < -6.25 "' (0.25) - (0. 5' each) .

REFERENCE BWROG EDI Users Guide BFNP:~EOI-2, Section 2.1; EDI L/P, L.O. B-2 4

i

BWR K/A: 295024 SWG #11 (4.3/4.5)

~

I 1-ANSWER 4.08 (2.00)

a. Alert

b. (Notification of) Unusual Event

c. General Emergency

d. Site Area Emergency (0.5'each)

REFERENCE EIH: GET Handbook, pp 57, 58, 60, 61 HNP-x-4420,- HNP-x-4520, HNP-x-4620, HNP-x-4720 BFNP: BFN-IPD, IP-1, p 1; RQ 85/04/01; REP L/P, L.D. C.2.

i DWR K/A: 294001 A1.16 (2.9/4.7) a

di__EBQGEDUBES -;NQBd862_ABUQBd@61_EMEBGENQY_86DL :PAGE..66~

8091969 GIG 06_G90IB96 ANSWERS -- BROWNS FERRY.1, 2&3 -86/12/01-BROCKMAN,fK.

ANSWER- 4.09. (1.00)

a. 150 mrem

b. 3OO' mrem

. c . .. _.50 %

d. 1 Rem /hourj (All +0%,-10%) (0.25 each)

' REFERENCE.

BFNP: RCI-9,Section IX.E; L.O. "F" BWR K/A: 294001 K1.03 (3.1/3.8)

ANSWER 4.10 (1.50)

a. Increase (0.25) Formerly extracted steam'now passes through the turbine (giving net work / watts out) (0.5) (0.75)

b. Decrease (0.25) Extraction steam to the next higher pressure heater will increase (0.5) (because the feedwater temperature rise across the heater.is greater - thus, less-energy is available to the turbine) (0.75 REFERENCE BFNP: 01-2,Section II.C.1 & .2, L.O.'s "A" & "B"; OPL171.011, Section E BWR K/A: 259001 A2.02 (3.1/3. 3) (

' 4. EBgGEDyBES_ _NQBd862_6BNgBd662_EdEBGENgy_8ND. PAGE '67

-86D1960 GIG 86_QQNIBg6

, ANSWERS ^- . BROWNS FERRY 1, 2&3' ~ - 8 6 / 1'2 / 0 1 - B R O C K M'A N , K.. H

'l

'l ANSWER 4.11 '(1.50)

a. Complete Loss of Offsite Power-(LOSP) b.. 'Accihent' Signal present (LOCA on any-unit)

c. The dedicated diesel generator (s) cannot supply sufficient power for all of the required loads. (0.5 each)

REFERENCE BFNP: 01-57,Section V.E.4.j(5); L.O. "C"

'BWR K/A: 264000 K5.05 (3.4/3.5), A2.01 (3.5/3.6)

ANSWER 4.12 (1.00)

a. 47 psig (+3, -2 psig)

b. Possibility of overloading the 3C EDG (0.351 during a simultaneous LOCA w/LOSP (0.15). (0.5 each)

REFERENCE BFNP: OI-24, Sections II.B & III.E, L.O. "A" l

l

'l l

BWR K/A: 264000 K3.03 (4.1/4.2) 286000 SWG #10 (3.2/3.5) i l

1 m___.________________.__ _ _ _ _ _ _ _ _ _ _ _i__.__.____.____.._________________m.____.E_.__.i___ ____.___.___.______._._________m____.__________ _ _ _ _ _ _ _ . - -

.44 ' PROCEDURES ~- NORMAL2 _ABNgRMAL2 _EMERggNgy_8ND: PAGEL 68 B891969E'C86_CgNIBg6

, ANSWERS -- BROWNS FERRY 1,_2&3 -86/12/01-BROCKMAN, K.

ANSWER 4.13 (1.00) d REFERENCE BFNP: 0I-66_, pp 36 - 37, L.O. "B" & "K" BWR K/A: 271000 K4.04 (3.3/3.6), K5.04 (2.9/3.1), PWG.10 (3.1/3.2)

A1.13 (3.2/3.7), A2.06 (3.5/3.9), A4.09 (3.3/3.2)

ANSWER 4.14 (1.00)

The Recirculation Pump Motor Heaters will not operate unless the Field Breaker is in the (INSERTED or) TEST position. (1.0).

REFERENCE BFNP: OI-68,Section III.A.7; 01-68 L/P, L.O. C BWR K/A: 202001 SWG #10 (3.5/3.7)

, 294001 K1.02 (3.9/4.5) l

4t__ PROCEDURES - NORMALx_ABNQRM8L 2_EMEBGENQY_ANQ .PAGE' 69 88D1069GIC86_CQNIRQL ANSWERS -- BROWNS FERRY 1, 2&3. '-86/12/01-BROCKMAN, K.

ANSWER' 4.15 -(1.00)

1) Evacuate the Refuel Zone

2) Place the other loop of Shutdown Cooling in' service immediately, (without flushing). (0.5 each):

REFERENCE BFNP: GOI-100-12,Section III.E; GOI-100-12 L/P, L.O. D DWR K/A: 205000 K3.03 (3.8/3.9), K3.05 (2.6/2.7) 294001 K1.03 (3.7/3.8)

ANSWER 4.16 (1.00)

Heat removal from the reactor sufficient to restore and maintain the peak fuel clad temperature (PCT) LTE 2200 deg F. -(Heat up is allowed as long as PCT will nowhere ettceed 2200-deg F). (1.0)

REFERENCE BWROG EDI Users Guide BFNP: OPL174.657 (EDI L/P), L.O. D l 1

i i

l BWR K/A: 295031 Ekl.01 (4.6/4.7) 1

% _ . _ _ .x . .- N

4. JP8QGEDUBEg_ _NQBMA61_ABNQRMAL1 _gMERggNGy_AND. PAGE '70 -

88D196901GBb_G9 BIB 96 ANSWERS '-'- BROWNS FERRY 1 , '2343 -86/12/01-BROCKMAN,_K. .

ANSWER 4.17 .(1.00)

(ADS is disabled as long as a Reactor- Shutdown is not assured, or may be contingent upon -SLC) to prevent the injection of ECCS (large 4 volumes'of-relatively cold water) which may result in a reactor power {

excursion leading.to substantial core damage. (1.0) i 1

REFERENCE

-BWROG Users Guide, EDI RC/L, p3 .

i

'BFNP: OPL174.657 (EDI L/P), C-5, L.O. "B.6" l BWR K/A: 295037 EK1.03 (4.2/4.4), EK1.06 (4.0/4.2), SWG #7 (3.7/3.9)

ANSWER 4.18 (2.50)

1) Rx Water Level (0.25) < 11 "

(0.25) (0.5)

2) Rx Pressure (0.25) > 1043 # (0.25) (0.5)

3) D/W Pressure (0.25) > 2.45 # (0.25) (0.5)

4) Rx Scram required and Power > 3%, or indeterminable (0.5)

5) MSIV Isolation Required (0.5)

REFERENCE BWROG EDI Users Guide l BFNP: EDI-2, Section 2.1; EDI L/P, L.O. A l BWR K/A: PWG #11 (4.0/4.0) l

S.__E89GEQQBES _-' NQBt]862_6BNQBt!@b2_gt!EBGENQY_GNQ PAGE - 71 889196991G86_GQNIBQL ANSWERS --~ BROWNS FERRY 1, 2&3 -86/12/01-BROCKMAN, K.

4 ANSWER 4.19 (1.50) ,

(1) System f ails to operate.

(2) System operates in a suspected adverse manner.

(3) System operates outside of the limits of the documented acceptance criteria. (0.5 each)

REFERENCE BFNP: Standard Practice 10.9, L.O. "D"; NOAM (Part 2), Section 4.9 s

~.

l N -___._.s__-__._.___J

-TEST. CROSS REFERENCE PAGE. 1 QUESTION

'VALUE. REFERENCE' l

' 01.01' 1.00 -KED0000154 01.'02 1.50 KEB0000907

-01.03 2.00 KED0000908

-01.04- 1.00 KEBOOOO909 01.05 .50- KED0000910 -

'. ~01.06 1.50 KEB0000911

! _01.07 2.00 KEB0000914 l 01.08 1.50 KEB0000915 01.09 1.75 KED0000916 01.10 3.00 KEB0000919 I 01.11 1.00 KED0000920 01.12 1.00 KED0000921 01.13 1.00 KED0000937 01.14 1.00 KEB0000959 l~ 01.15 1.00 KED0000960

'01.16 1.50 KEB0000962 01.17 1.50 KED0000963 01.18 1.00 KEB0000968 01.19 1.25 KED0000969 26.00 02.01 1.50 KEB0000922 02.02 1.00 KED0000923 02.03 2.25 KED0000924 02.04 1.00 KED0000926 02.05 .50 KED0000928 02.06 .50 KED0000929 02.07 1.50 KEB0000930 02.00 1.50 KED0000931 02.09 1.50 KED0000933 02.10 1.00 KED0000934 02.11 1.50 KEB0000935 02.12 1.00 KED0000936 02.13 1.00 KED0000930 i 02.14 2.00 KED0000939 I

O2.15 1.50 KED0000940 02.16 2.00 KED0000941 02.17 1.00 ' KED0000955 02.10 1.00 KED0000956 02.19 1.00 KED0000957 02.20 1.50 KED0000958 l

25.75 03.01 1.00 KEB0000069 03.02 1.00 KED0000070 03.03 1.00 KED0000071 03.04 1.00 KED0000072 03.05 1.00 KEB0000073 l

l

r .. r- , - ;.

Yi TEST' CROSS REFERENCE ~ PAGE 2 QUESTION VALUE- REFERENCE 03.06 1.50 KEB0000874' 03.07 '1.00 KEB0000875 03.08 1.00 KEB0000876 03.09 -1.50 'KEB0000877

'03.10 1.50 KEB0000878 03.11' 1.00 KEB0000881-03.12 1.00 ~KEB0000884 03.13 2.00 KEB0000895 l 03.14- .1.50 'KEB0000925' l ~03.15 1.00 KEB0000949 l 03.16 2.00 KEB0000951

03.17 2.50 KEB0000952 l 03.18 2.00 KEB0000953 l

03.19' 1.00 KEB0000954 03.20 2.00 KEB0000989 27.50 04.01 1.00 KEB0000885

• 04.02 1.00 KEB0000886 04.03 1.00 'KEB0000888 04.04 1.50 KEB0000890 l 04.05 2.00 KEB0000892 l '04.06 1.00 KEB0000894 04.07 2.50 KEB0000897 04.08 2.00 KEB0000898 04.09 1.00 KEB0000903 04.10 1.50 REB 0000904 04.11 1.50 KEB0000905 04.12 1.00 KEB0000906 04.13 1.00 KEB0000942 04.14 1.00 KEB0000943 04.15 1.00 KEB0000944 04 16 1.00 KEB0000945 04.17 1.00 KEBOOOO947 04.18 2.50 KEB0000965 04.19 1.50 KEB0000974 26.00 105.25 l

l L

l l

ENCLOSURE #3 RO-MLT 3

SRO-HLT 5.01 RO-RQ REh $b97 SEO-RQ t TVA CONNENT b) Power was dropped to 12.0 KW/ft. Not 11.8 I

TVA RESOLUTION b) Based on starting a ramp at the power dropped to (12 KW/ft to 13 KW/ft) the minimum time to get to 13 would be 10 hrs. None of the choices are l

correct. Please correct prior to using again. Credit should be given i for selection of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> since this was the only logical selection.

I

RO-HLT 3.01 SRO-HLT RO-RQ 3.01 -

KEb BGS l

SRO-RQ s t 4

TVA CONMENT The attached NRC figure 869 gave the " error signal input" at 50% initially.

It would be zero initially and then start increasing. The way the figure is drawn (error signal), the controller output would be increasing initially. The drawing could have lead to confusion.

TVA RESOLUTION Need to correct drawing to initially show error signal at zero. This would then make answer 'D' totally correct. Answer 'D' is the only logical selection for the information given, however answer 'C' does show a response to the error signal increase at t=0. It might be an acceptable answer.

Ref. OPL171.008, page 10 and 38.

l 1

l

RO-HLT 3.02 ,

SRO-HLT 6.01 RO-RQ 3.02 SRO-RQ 6.01 J t TVA ColetENT Part a - accept LIT 3-53 as correct answer. Ref. Lesson Plan OPL171.003 LIT 3-53 is level indicator 'A'.

I l

TVA RESOLUTION Accept LIT 3-53 for full credit.

I 1

l l

l

\

1

r RO-HLT 3.06 SRO-HLT 6.02 RO-RQ 3.06

+

i REE M 4 l

SRO-RQ 6.02 s t TVA CONNENT Part 'a' - See Attachment 1 TVA RESOLUTION l Part 'b' - See Attachment 2 ,

l i

1 e

r Attachment 1 COIDIENT:

a) The range of time in the answer which was taken from our Lesson Plan is not an orgat time. The network in the APRM circuitry is used to model the fuel thermal time constant as discussed in Chapter 8 cf NT & FF.

(See Attached). The response of the power input to the Thermal Scram trip unit would be the same as that shown in Figure 9-19. (Tc).

The range of the " delay" will be determined by the circuitry time constant (i.e., 6 sec's). This " delay" could be 0 sec. or after 5 to 7 time constants (30 to 42 secs), which would be the time for the output to reach a new equilibrium value. It is important to keep in mind that the time constant network effects the core power signal from the APRN averaging amp and not the flow biased trip setpoint. (.66W+54).

Some students may respond to this question with a time delay of 6-30 sec.

based on 0I-92. This time is not 100% accurate and a procedure change request will be submitted to delete the reference to a " time delay" and reference a 6 sec. " time constant".

b)

The amount of time delay will depend upon the below three items, not just the one referenced in the answer key,

1. Rate of power increase.

2.

The amount of Power increase (ie. Pfinal - Pinitial)

3. How close you were to the trip setpoint when the power increase started.

Some students may have based their answer to part "a" and "b" on a discussion of how the circuitry actually works. The correct response should describe that the output of the time constant network will reach 63.2% of its final outcome in 6 second (ie it will increase 63.2% of the APower in 6 seconds) The same response will be seen every 6 seconds

l

[

Attachment 2 RESOLUTION:

a) Accept the following for full credit:

1. 6-30 secs based on 01-92.

2. O secs. to something between 30 and 60 secs.

3. Give credit if the student discusses how the circuitry works.

b) Correct answer should be: i

a. rate of power increase (NRC Key)

b. P final - P initial (AP)

c. How close you are to the trip point when the power increase started.

j

.; e 4 0

THERMAL TIME CONSTANT

(. ,

A rapid change in fuel temperature is measured in terms of fuel thermal time constant (Figure 9-19). In plots (a) and (c), the fuel centerline temperature has made a ' step increase from T to T Plots (e) and (f) fi f2 show the time response of the cladding surface temperature to the step i increase in fuel centerline temperature. ' Tc lags behind the increase in T ,

f rising exponentially to a new steady-state temperature (Tc2). The change in Tcwith time is given by T

c" cl + (Tc2 - Tc!) gj _ , t/T) (Equation 9-10)

$ tat where:

Tc ' c m h & OfMs .

)

Tc = cladding surface temperature as a function of time (Analh *

, Tcl = initial cladding surface temperature (PQ .

l Tc2 = final cladding surface temperature (P+)

t = time T = thermal time constant

• t The thermal time constant (T) is time required for cladding temperature to reach 63.2% of the increase (or decrease) resulting from an instantaneous change in fuel temperature. A short time constant will result in a rapid response to power changes. A long time constant will result in a slow response.

e e

O b y

e-D 9-42

.....- -- x N= - . . . . . . . . . ~  : .. . - -~-- " *N~

I Tgg a enettet Fue centeessne Temperature Te a laitial cieddine surface Temperature

( Tgf a Fines Fuel centerline Temperature .t ,

, Tg s Final Cledding Surrece Temperature e Tg = Fuel centerline Temperature e Function of Time

$ To a caeddins surrece Temperature se a Function of Time gL t s Time g .

f a Thermal Time Cenetent 2

W .

Tc; .

T g

, {

(a) (b) TIME Tg 2 Tg2 * "

• e E ,

.. t,' . . _ . . . . . .

w Tfg %

• Tg g T

Teg ct ,

. (c) , .(d) TIME w

8 A T' Tf2 R Tf2

• C2 *2 g sss e g

\ \\ ' I

'\\ 63% of (Tc2 -Tc g)  !

I

\ \N[ Tct Tf3 f TIME CONSTANT (f)

Tct , g (e) teo T (g) TIME

- 1 Figure 9-19. Thermal Time Constant For 8 x 8 fuel, the thermal time constant is about 5 to 6 seconds (i.e., if Equation 9-10 is used to calculate the cla'dding surface temperature after ,

one time constant (t = T), the cladding surface temperature will reach 63.21, . -

of its final temperature increase after 5 to 6 seconds). Thus, although fission rate response to a power change is very rapid , the. rate of ..

\ temperature change is governed by the fuel thermal time constant. Thl

• concept is important in analyzing core response to transients and accidents.

t 9-883 .

h g '

_ agapmM .h h _ % h4 -- -- ^

^

R0-HLT 3.07 3

-i SRO-HLT R0-RQ 3.07 N SRO-RQ ,j t TVA Colel5NT i

Answer key is not correct. To get the outlet valve open requires:

1. < 160*

2. < 10 psig

3. HS90-155 to reset position i 4. Vacuum 1 6" Ng

5. sufficient steam pressure i

j

' The requirement for steam pressure is: must increase to 1 190 pelg 4

(switch makes at 1 190 increasing). The switch will open at s 150 psig decreasing. Thus valve would not come open since pressure was given at i 190 pais, a

In order to get the PCV open, and thus steam pressure up, the following are required:

3 i 1. HS1-150 in auto or open 1 2. steam pressure 1 190 psig i 3. SJAE inlet and outlet condensate valves open 3 4. condensate pressure 1 60 psig

5. Other SJAE steam NOVs closed.

i i

There is a 30 second time delay on pressure in order for the PCVs to open and l get pressure up to setpoint. Since condensate pressure is < 45 psig and condensate outlet valve is closed and SJAE 'B' steam NOVs are open, the PCVs will not open.

Thus pressure will not increase and the valve will not open.

I j TVA RESOLUTION I

Change answer key to : NO Reason: Agt of the following should receive full credit:

j 1. condensate pressure < 60 psig i 2. Other steam NOVs open

3.

I Condensate outlet valve closed

4. Steam pressure < 190 psig

5. Required vacuum not established 1

Reference:

Lesson Plan OPL171.030 pages 7,9,31 l 1

1 i

f

}

, , , _ _ , u a e ws *.

i mt J A

,f R0-HLT 3.08 SRO-HLT RO-RQ _3.08 - -

SRO-RQ __ -

TVA COMMEll7 .

Requiring "the Av9 raging Circuit is fot gained down (( it) reference APRM" should not be required for full credic since'the quest yn doesn't addests the gain change circuit. ,

m TVA RESOLUTION Acespt as justifIcatici: '

the rod block'netpoint. the power increase will be sensed as reaching '

q r

f I

e

/

i 0

t pA g a ~* ""

,5 . .

(-

3, s

. . 9,4 .? : /. ,

3 7

s .-

3 3 .RO-NLT i

i 1-

'3RO-HLT t ._04 i

• I ,

< , \

5' t

i RO-RQ -R -

NO N3 j - JRO-RQ p 6.04 J -

I, x, ,

4 s '*

. i. , . t i l ,[ P ' TVA COPPEhT f

.s*

I Part B.r2 answer: Should accept as correct: Refueling platform near or over l ,

,c core and mode switch in startup.

Based upon the wording of the question (i.e. prevent r..N'- m)vement of the Refueling Bridge over the core), the answer i l

should not require; for full credit: " Platform near or i" over the core." i

! i

' i

/ ,

Reference:

OPL171.053, page 8 and 21.

a .-

i (

} i l i,

, 1 a

1 i

e .\ -

\'

4 14 SOLUTION

i Accept either the NRC answer key response or the above comment as fully correct i

1/

l i

') ,

is

\

1 1 I

I a ,

l

, s

"i t

\

I;. ,

q . ,

f t

i l- ,a Vi ,t : '

!. l .

jy2 ~s

- r,

<, (t ,

RO-HLT 9

SRO-HLT 6.05 R0-RQ REE SBO SRO-RQ _ 6.05 t

TVA C0f0 TENT Part A:

Answer key needs to include: The 8' upper limit assures adequate water shielding of the fuel on grapple. (REF: OPL171.053, p 7)

RESOLUTION Part A:

Any answer that adresses preventing excessive radiation levels or to maintain the required level about a fuel bundle should receive full credit. (Ref: OPL171.053, p 7)

=

9 1

RO-HLT 3.11 ,

SRO-HLT RO-RQ 3.11 SRO-RQ ~ t TVA C0f0 LENT a) This question is taken from material that R0s and SR0s are not taught.

A person who is a nuclear engineer does get the training.

b) The IXYFLAG printout on a P-1 will either have a "0.0" printed or a four digit number which corresponds to the coordinates of a bundle at which the CPR-calculation failed to converge. It is not clear what is meant by " invalid data". Is it the P1 printout or the data used to calculate thermal limits.

(Ref: NEDE-24810, page 4-15)

TVA RESOLUTION As a minimum part"b" should be deleted.

1 I

l l

R0-HLT 4.02 SRO-HLT i

RO-RQ 4.02 NE3 885 SRO-RQ L

TVA C010fENT Weighting of sections of answer appears inappropriate.

The curve protects against " equipment failure (i.e. containment)" due to unstable steam condensation during an ADS blowdown.

REF: OPL171.057, Appendix A, p19 (E0I Lesson Plans)

RESOLUTION .

Reverse point values or make all parts of question equal point value.

The above should be given full credit. .

l g es s

.,y. , g - .

c._- , . . _ - . . - e _ _ . . _ . - - . . , ,,., ..

1 i

R0-HLT m

SRO-HLT 7.02

RO-RQ SRO-RQ 7.02 -

TVA CONNENT Part A. The purpose of the HCLL curve is really two-fold.

1. From 14.6 ft, down to 11.5 ft. the HCLL compensates for reduced heat-capacity of the pool due to a reduced level.

The HCTL is calculated based on a level 114.6 feet.

2. Below 11.5 ft. the downcomers from the drywell to pool are uncovered and incomplete steam condensation would occur on a primary system break. (i.e. overpressurization of the containment could occur.

Part "a" really is two distinct questions in that the purpose of the HCLL and what the curve protects against is two separate discussions.

The portion of the NRC answer which applies to a level <11.5ft, is not worth a valid point value. (i.e. each of the two arguments should be worth equal value).

B.

The use of the phrase "immediate action (s)" is not appropriate when dealing with the E0I's.

The E0I's have no immediate actions and none of the steps are " time dependent." To do so would defeat the total concept of " symptomatic" procedures. The original E0I's at BFN contained "immediate" actions and they were intentionally taken out because the classification of steps in this manner defeated the whole precept upon which the E0I's are based. During training on the EOI's it is emphasized to the operators to not " memorize"the steps, because the sequence of execution can change from one set of conditions to another. See attached excerpts from Appendix B of the EPG's.

RESOLUTION Part A. 1. Revise point value of current answer as stated earlier.

2. Accept for full credit present answer or the answer presented in 1 and 2 above.

Part B.

Discontinue the referencing use of actions the the from word "immediate" on future exams when E0I's.

t

, - - - - - ..---_,3--- , , - - - - , , -- . . . - , _, __.,m_... _ _ _ . . . _ _ . _ - . _ - . _ .--_,m,,m_.__e., -% -. , _ , . . , - , _ _ ,

. . A; l CI' "' * ~~ ' ~;'- '

s

&.x%$ft?Yh${A.

1,f. .s,4 5 : , . . . . .

• Il g-5.&? $!h$,h .

5 ' b$3Ndh: f&. ~

_ 'l 2 -

. g.3..c., Document 8300-2 Seetion 2, EPG Structure ."-/,, .,, .

3,, ,';h,"

.y S

j.W. ..

t %.,m,. .

,.y a ;.

,<v

. :~~ q

a. :1. :y. . .~urt , <

l r , G5 ' .(;. .%e 4*. s ,gN t .4:.-Ty . : . . W, . .. w . . . . .*

4 ... .. y .i/.$.. . :

P, D 6,, ,

,,9 ,. e,

%.;. t.y[.gI p. I

.h,. .. $ .

m-f v?lN h.. ;h'n u>.. 4a(19 0f w m, f s.j?{ w, ?

-b G.../%{*E.

. ~

\

parameter is exited and. the appropriate Contingency is entered. ~9.'d)e$.,. k;.

' ;A,h p.'

. , . .~ . . v . . c m . . ;-

.d.. ,N...; e' This lg'1s' accomplished by use of a statement at the ?- beginning t

('

fi smawww of M' 4 n.section w.mm.1n an example .of which is illustrated below: 4 ww mww -

1. I n

Q,,QWgypp,y:w \ ~ wn w w r. h W W % g .Wi' ^ ' ' g'b~rL.

uJ '

w u n a n ------------- yas n e a -------- -----------------------

,-- g h.Je M e,.. If while executing the following steps Emer- I ' M" W@ '

pe

'. ichM .).:t .M. gency RPV Depressurization is required, enter %p. g

! [ procedure developed from Contingency #23. #

I di

~ , . I . q G.'I,. -

^%.gti, gt ;' _________________________________________________ w

..c; g. , l .

t , , j,,

0'?,.fl,?!' Q{r&.;,. .. f&. &.; Qlf., . Q f

Q'

.-Q. g. g . .,

' g.. This ..

structure effects the transfer of RPV pressure control when. ,%q .

p . , .

. }. {g.).v appropriate . m ;. . and precludes concurrently ef f ective but conflicting . ,$j.png., 4 ,g'&p{ ,

Efl

, -)';, ' '.  ; . , d"'

.h{ '@.Q 3 nstructions.

i

3. o m .

' /+ Q,, Q;i'%g@ .

..3 ysc cwMW.i " 'WW "' **.**k .'" " '" 4Wh*'.W3pW/dM;m ' $3,6.dY;w',0 .j., ....!

. ;; . .,5 4 . 3.

M,%pg&6<[,",f

, . s .a' The EPGs based upon the symptomatic approach to emergency' '

are (. . # ' .,)

.T

..': . .q;ge b,Y 'i. if r response as discussed in Section 2.2. To achieve the ' ' ~

s' ..

event-

.......~> t b W6 independence required by this approach, certain operator actions J

..y f e, , -

9;a,.,.,

.%.6 , ,

for monitoring and controlling plant parameters must be executed <

M Jf e.

~. ~

_-;a.fi,Q .: .

Q ;j,9 g.},

in parallel. This results from the e fact-q%.wJs.V,: that,,the_  :. W.?.ioptimum .1 u*zi 5 j*W;i -

,.~ ,, .

sequence 'of all actions appropriate to an emergency cannot be .h,Yhr,3 yr ' 4 'i. . . a + "

.+ ,.,,,, ,

. .:jQ fy.*y E fy determined without knowledge of the sequence of events ,hich w

["

. p . .y . . . :w .. .t ,

i,.

9 _., 4 . ,

. p. g. f@

precipitated the emergency. Therefore, since these actions cannot 'I W .Tcr N _

i

• i g . ,

.a .r.< . . , ,

f&Q ll be sequenced or prioriti:ed in advance, the values and trends of W'n

.i,.,.~,.~.e+

,, *r .

parameters and the status of plant systems and equipment t.-

' ' u,; ,

re- ,

y. ,p.f,'[,(

. sul t i ng from any transient that requires the use of emergency ,3 :,4Q}d proccedures developed from the EPGs dictate the relative *=l % W'j,U i mpor- - vf ,. j

. + ... w. . .;q tance of ind2vidual Guadeline steps and the order in which they 1.c .m .a W. J Pjk

- should be accomplished. .v

~. .

.$e.d.::2 $:

.. d aW .M *as9 M (r-- ,

2 a .g ,Q%!! '-17j

..<.-..,y,...y.,_g.,

, ,. e4 :-o. -.j f ,y,;.pt ,e  :-

7 3 h, .'<' e : %.:t. y6 e -,

qdi .e * >;?. . Q .,; y.;.= mn.~ . :. ,,*.s. y;:~. .+.W.Q2<

{*".,;,

r.y$qgg,

~

,. .p O ,

~ U'

. b

l' q; "g,. - sC$' 2-5 h'. P.

,- March 1984 g.p;..g& Wg.T. - ". W ?A:.~Dra'ft, Revision 3A J.; '. .1 3-.- r , c . .. %gg y ,

. . g :.J >. . - ,

e a < * - - 3,p e p.:.vc 3g.m;twy w d

• q[

., .W-

.. ' 4 . . , . - .,

pay.i' '}y n ?T

.J ~ i.s ,.' ~

~"

. , . g ;khh '

. '.r ..ge .': .././o ':.~.. .. -

. .:%k;*$kW:

.o . ., .

r. 4.. .#. ~

. av;W ' '

l,};,'Q : ' .

, 9, .,::

g.1 Document B308-2 Section 2, EPG Structure .y . .Q . . ;T s

, $ p;n.:. g.'. J ", ,

,' , 9,k W 3 s) dhD M y.D 'Y$

s%w- waby . u.M- 41M;g%

[;/ph fj 14i.i n m:

Jf: ,

.. ,p .e . r.~. m ,. ay,g.t . . .>9 , y.. ;

. J..c 0 ',. 2. 3 Format and Oroaniration -~,

,.~ . , i.g;

.,,'....4" , c,,. . . . .m,. . rj (.y

.,, . ; - c '..

. u;4 ,- . e . ..4.. .. .

A.s .M

. w r 3. {QQw.w. . wg.Q..

p a ':r. -

n::q. '

Each., of the four Guidelines is organi:ed under the fallowing :g  ;,

if(jg'g 1 ,: w Y.b M: .q :tf.y W. .

f .

• f 4.,.,. J

%. _g'.gg,(section headings: . 4{c 4.jpej.Wyg'

~,. g . g'$ ( ,

g'G}MSUp. ,p . : ;_ }' , ', ,, id) g , ,

-.o. ...

• , a J g , s .-

r, M. . '

' . i ' [,. . . . . < ,N '

• • ' ' a .,

, ' . , 7 g .; . g ..

i

~ ..3 .; .i f. '

59 4 1 ..?I

.. ,. y 74*; , . g . 7,.

, .u.. , ~. ' . ,

. e.. . .

PURPOSE '. .;,[Y[s r-a gt.N .s .oSr F.', 'aM 4Q;>.p g ,. ,. , l

. @.h *- ,P (%,'$Q.

• • . -  ; L: .:

.g'j.' p :.n... ENTRY CONDITIONS

' ,' ' .' g g' ,g.)M',. , g; g:.< k..;; f. C ,;

'ie

@..fy 'N..,, ,T

... .. . M ,d ,...~..,.....St' .&

.- . .,..bcg. l ,

. - - m. . . p .

c.6,#,d . xw. , . .' .o. ., ,.3

y. . ..., . %,, . . .

-1 . . ...x g. A. . . .-

%..:w .

, v. 's;  ; , ,- g. .;

/.T a .

t.

c. .. . .

"q . , . ..

T .g\IN

i. i ..,"~:N.' OPERATOR ACTIONS ~il,I *'NF .Y. .T. b. , D,, .b

.. \; g.ty ,.:<:r~. I4 .. A~ T';t!"' .. ~

. .. v: . . . ... :,; . V

s .c . n ,.., :

y ,, . i }

y

(. , , ,, 9 7 3 ,,

.. .--;WMrtrM.XM 4e W 3'P.' W @ % 4*. 4 * '-

No differentiation between "immediate" and " subsequent" Operator

~; , av W

J.$};W;g

. 2- .J.R . b +$

Actions is made because such a distinction, '

if 1t can be made at . Pr ~ ~ ~ .

all, requires consideration of a specific event and

.$.f this would be t 4m:. .

4

. ,j. .7, s> inconsistent with the event-indeoendent basis upon wh[ch the EPGs

. w. .

y.f'.'

m ..

• are developed.

s - *2 . , & c..:<l k.,

.i ' . ,t. .w i , ,  ;

.. . ;a.c - .

NLA- -

No section which lists AUTOMATIC ACTIONS is included)

.*..c. where . 'c'.*% .. Mf: c.*

.y . .s .p..s >.- . t .W*.

r.

'.u .; p . g'n.x ,. . b

.'.,f confirmation of automatic actions is appropriate, .these instruc-

, g- y, .. .. ,

tions are included directly in the steps under OPERATOR ACTIONS. Afi- M'

~ a ?w.i .W n ., ....

v?q.dfh.,.- M.f.<

yun:

, . wa e.;-p ,-T..m.  :,&. . .@w..d.a nn. . y.M..j.?g;G;Q4-  : v y .:. - :L: . * :.:. .m

.~w-...4 -

-3 y'

. .w . s,,

EPG steps are identified by a letter group signifying the v

. f. cg., . ..

u c

physical space and parameter controlled and by a number

.[9.

d. e.:.6 : ';

t.

p indicating the sequence of the step within the set of steps M'.'i , '

.V ' gn..

controlling the specific parameter. "For e:: amp l e , DW/T-O is the 77 5

_ . . .. c.d.3 ,.4 drywell" - ~ p]

second step in the set of steps for control of x , o. v..: .m; temperature. -m

. p:.c W. -

. , . . - Q. ,. wn a' -:w :* z' j. & .%,

,,..,, s ,' . . M

..y  ; ,

.. . . g,,g -

~) / - 1,-

.r.

.;, ,. y f .g. .r y. ,p o e;. ~ g.:4,p,, :3.

. . A. . . , .

y ..,. mn.ggy , . . qq. . .y.. .,,.s u -

'Marb1984 k W .W;k'.y se d.,< :-10 Y ' UN Dhaft, Revisi . A

.. . - . . ~. .c. ,,w. n . w ,.

• f G. 'Nlll;3-fp+.#iyfn. ~. a.,d.. dll:.s;'ifiq,~%.).}^ fin-f.Q'p!,.a 2+ , G.-nlQ,qli,.w .

v?V? .%f:*.:. 9, .p s q . _r s. . , e. . . y

.s

,.s- . .i'.$i:

.. . - .. - .r, . s..

~ ; * . %? ^

3 I' '

.l<f .* . _ . , W. *

.* s ,

ht.

RO-HLT 4.03  %

SRO-HLT 8.01 R0-RQ 4.03 f RED bSB SRO-RQ 8.01 .J L

TVA COMMENT Part A: Question ask to " Describe how to verify operability of MOV."

l l

RESOLUTION

]

Part A: The phrase " prior to being declared operational" should not be required for full credit since the question ask how to verify operability.

l 9

i RO-HLT 3.13 SRO-HLT 6.06 R0-RQ 3.13 KEB e 5 SRO-RQ 6.06 t TVA COMMENT The correct answer for Part 'a' and 'b' is "2". As long as either the field breaker is open or the discharge valve is closed with the other pump field

1. breaker closed and its discharge valve open, the flow from the idle loop will be substracted from the core flow signal.

Ref. GE 730E483 and 723E841 (attached)

TVA RESOLUTION Change answer key for part 'b' to '2' l

i l

l l

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

(s .

rS O I I i 1

~ CLOSES ON CLOSED 10-101 A Z[ [ [101C ~ K9A CLOSES

~ KS A _ _ K58 ~ LO RX __ (( ON _ ,

~ ~K35A ((K7A [_~ K78 ~ ~ SEAL M LEVEL FROM LORX . RX PRESS RX LEVEL ON HI DV CS LOGIC j PRESS BUS A ~ ~K6 A ~ ~K6B PRESS

.; 250V DC

• T T " ' " " ~~

CS LOGIC

~

DV k'o^ ~ ~k'^ ::x*8 RX LEVEL

- / PRESS Y~DV TEST

K5A SVITCH '

6A K7A 8A, 35A WITIATION FROM CS 14 A-S10 A h. K36A K73A 9A SIGNAL K10A (Y)

LOGIC " SEALED IN I I '

HiDV PRESS DV PRESS RELAY LO RX LEVEL 3 LO RX PRESS. INITIATION SIGNAL DEVELOPMENT I

~

j[ ~

j[ [_" [_~ K9A K110 A , K111 A,, K112 A K113A 1 ,

OPENS VHEN CLOSED VHEN -~ -- -- --

i D/G PVR NML AUX PVR BUS A AVAILABLE AVAllABLE K238 K104A

~

K24B K105A j 250VDC ~

y ^2 ~ NVAj~[DWA((NVA

- #K74A y U2 K25A K18A

• B B A A K104 A K23A K104C K105A K24A K105C K110A KillA K113A K112A BPUMP A PUMP

1

, TDPU TDPU TDPU

i TDPU

.1 SEC 7 SEC .1 SEC

.1SEC

' POWER SUPPLYINTERLOCKS I

PUrP START RELAYS

}

RESIDUAL HEAT REMOVAL SYSTEN INITI ATION LOGIC GE730E920 171.044-TP7 g 7/22/86

i t

I  ?  !

CLOSES ON CLOSED K98

. [~ (( .~ [ K5 A _" [ K58 ~ [ LO RX __ (( ON [

~

~ "K358 - "K7 A ~ [ K7B ~ ~ SEAL M i

f '

CLOSES LEVEL FROM LORX RX PRESS RX LEVEL i 1- ON Hi DV CS LOGIC i'

PRESS -- - RESET PRESS BUS B i

? 250V DC

~ FROM [ ["K108,-[K8A ((K88

, CS LOGIC DV RX LEVEL

,$ - # PRESS I # ~DV TEST SVITCH i INITIATION B K7B 8B K358 K36B K73 B K9B FROM CS 14 A-S10 B h SIGNAL  ;

TO U2 K100 LOGIC r SEALED N

I I I HiDV PRESS DV PRESS REL AY LO RX LEVEL LO RX PRESS. INITIATION SIGNAL DEVELOPMENT I

4 I y#

y - - -- -- - K9B K110B K119B K111B K1129 OPENS WHEN CLOSED VHEN --

6 D/G PYR NML AUX PVR BUS B AVAILABLE AVAILABLE K23A K1048 K24A K105B

{ -

250 V DC y DGVA(( NVAj-[DGVA((NVA

- B y K258 KI8B

' O-K104B K23B K104D K1058 K248 K105D A A B B h h

.r s Y Y K110B TDPU K119B TDPU K112B TDPU K111B TDPU I

A PUMP B PUMP j .1 SEC .1 SEC .1 SEC 7 SEC j} POWER SUPPLYINTERLOCKS PUMP START RELAYS li RESIDUAL HEAT REMOVAL SYSTEM INITI ATION LOGIC GE730E920 171.044-TP8 l 7/24/86

f f i VALVE it 1

. . VALVE l CLOSED l VALVE l 174-25

, 74 EXCEPT_ (4-48 _K15 _K16B_1 B_ _ ,

~* _ K72 B VHEN CLOSED g

NE -

EXCEPT

-' L VHEN YALVE I HS f- .OPEN FULL OPEN h'~ BUS B - Kf f 2 8

[Kf f 18 _K25 A - STOP - 4 I j , 250V DC # ~ VALVE 74-24 K72B K15 B K168 ,

'? '

(Y) h h h I h'

DIS AGREEMENT LOGIC I

• PUMP STARTING PERMISSIVES B PUMP B PUMP

.f

.i i

4 AUTO RESTART INHIBli s i

K73B

--U2 - -- 0 -

- K74 B -- - HS _ HS ~[K114B K19B ACCIDENT EMERO. TD0 START STOP j BUSB SIGNAL 250VDC 4 f.

f}- gy K18 B1 K25A

. AVAIL. -

p[K105 D h' LS CLOSED i5 K1 4 B VHEN

k.: B TDPU FCV W W D D i [" K116 B K1158 K74 B K1138 CLOSED TDPU TDPU hTDPU h60 SEC.(Y)U-2 SEALED IN 52C 2

l ;. 60SEC. .1 SEC. .

.1 SEC.

l I

UNIT 2 INimTION PUMP BREAKER CONTROL ( 1 B )

u '

• RESIDUAL HEAT REMOVAL SYSTEN INITI ATION LOGIC GE730E920 171.044 TP9

.: 45N765-4 7/24/86

.t l

i (N rS .

I

[ YALVE VALVE l CLOSED 1 VALVE llEXCEPT _A ,,.74-48_

74-47 , , K15 A _ ,,,,K16 _174-2 l --

VHEN CLOSED l- VALVE EXCEPT j- FULL VHEN VALVE f' ~

-- [K112 A _~ C K113 A [ .~, K25B ~ ~ TOP BUS A 250 V DC

/

g [

l F.

~

V RVE

- 74-1 K72A K15A K16A .

K19A (Y) l5 l

[

'  % DIS AGREEMENT LOGIC I PUMP STARTING PERMISSIVES i j A PUMP A PUMP i

.A l

K73A I A A

U2  ; ;g74A f -

ACCIDENT 22 - HS [ ~ HS ~

j[K115A_~[K19A i

BUS A EMERG. STOP TD0 SENAL START

,. 250V DC kJ _ NV 9 K25B AVAE-. ""

y 6A l UI

. j[K105C DC LS CLOSED K1 5A t-' VHEN A TDPU

[- C C U-2 FCV WW t

i K117A K116A K74A K114A CLOSED TDPU TDPU TDPU h60 5EC.(Y)IN SEALED 52C 60SEC. .1 SEC. .1 SEC.

f UNIT 2 INITIATION I f

j g, PUMP BREAKER CONTROL ( 1 A )

r. .

'0 -

RESIDUAL HEAT REMOVAL SYSTEM INITI ATION LOGIC

[t ,y' '{.... .> GE730E920 171.044 TP10

,3 . ,

I . 7/24/86 it.

1

i I

t-8 I VALVE i' I K9A K118 A K119A VALVE_EXCEPT_

74-47_ 1 CLOSED,(4-48 l VALVE 1

_ K15A.1 2 4-13

. _K16 A _ _

-- -- -- _ _ K76A VHEN

') K74 A U-2 C

VALVE CLOSED EXCEPT

( BUS A 4i FULL VHEN VALVE

[ 250V DC IK -- - HS OPEN

~

~ STOP FULL OPEN K105C

- p [ ~K118 A _~ _~ K119 A _ g g.

j y vnw 4

l p[.' K104C 74-12

(. C C K21 A K76A K16 A K15A h h h h K22A t

K118Ah TDPU hK119ATDPU h fA W

14 SEC. .1 SEC.

I

• PUMP START REL AY DIS AGREEMENT LOGIC PUMP STARTING PERMISSIVES j* C PUMP C PUMP i

s 6

~

C

~

C

1. [K76A

'i' i E~ E CHS [ _~ HS pdK117A Z Z K22A EMERG. STOP TD0 START

).

]p* BUS A

- K21A ' f LS CLOSED i m'.* 250VDC

~~

! I.!

f VHEN - K116 A

~~

FCV 74-104 TDC CLOSED 4W "

'A .

, 52C i

l

4 PUMP BREAKER CONTROL ( 1 C )

RESIDUAL HEAT REMOVAL SYSTEM INITI ATION LOGIC l(I l,. GE730E920 171.044 TP11

,, l 45N765-4 7/24/86 6

i,

{ f T

D PUMP I ' I VALVE K98 U-2 .

K1178 K1188

, K76 B VALVE l CLOSED 74 47_ _EXCEPT_1 _74-48 _ K15 1 174-36 VALVE B_ _K16 B,1 _ ,

VHEN

,' ACCIDENT VALVE -

CLOSED D EXCEPT

] k.' 41 FULL VHEN YALVE

~

K4 [ZK118 9 K1178 TOP _ g g

~

!:J4, --

y Y ~ VALVE

, , ', s K105D K104D 74-35 lt 6B 6B 5B  ; K228 K1178 h hK118Bh21B TDPU 21 SEC.

TDPU

.1 SEC.

h s

8 PUMP START REL AY 8 PUMP STARTING PERMISSIVES DlS AGREEM(NT LOGIC D PUMP D PUMP DPUMP l

{

.l 1

1: .

- 4

~ K76 B ~~ D D E CHS [ ~ HS [ [ K115B ZK228

, EMERG. START STOP TDC

.j BUS B

~ K21 B ' / LS CLOSED J ""

250 V DC fVHEN

- / K1168 5

~

FCV 74-106 / _ TD0 CLOSED i

2 52C

! g PUMP BREAKER CONTROL ( 1 D )

} RESIDUAL HEAT REMOVAL SYSTEM INITI ATION LOGIC i- GE730E920 171.044-TP12

j.  ;., j 45N765-4 8/6/86 4 5

~

R0-HLT ,

SRO-HLT 7.10 RO-RQ F REB 9o1 SRO-RQ 7.09 .s L

TVA CCMMENT The alternate location of the TSC, technical support center, is the plant managers office wing of the plant office bids.

REF: BFN IPD, IP-20, pages 1 and 3 of Attachment I.

This area is the old administration building office wing.

RESOLUTION Accept Plant Managers Office, Plant managers office wing Upstairs old administration building Plant office building Second floor office building

' 4,

h e 2 r.:.

Page 1 of 3 ~$

ci . 31 SFN - IPD

e.

.

7: :i 3FN, IP-20 47

ns c. .3:: .i t Attachment I '" l l -

1000 Operatiori With TSC and OSC Relocated REY 0001 466d i

".:.5 To Office Building 'J!U EOS 'I

. .:d, t * (See Sketch) '

g p = ,2 -

,, 4 sto - r e ri ;n! :::3seR ',");

ca t Tw att, -T C :nen.nymA

1. TSC personnel will relocate to plant office building as stiown on drawing.

eoar\n.at , .r . - ; sStt a:penzaci c.se, :n n avaca ensis- s- 7lT.{ ---

2. TSC Comanunicator remain in Control Room or Backup Control Room.

,7... u g c

=  :~v.~ v?

3. Public Safety Supervisor to set up access control for relocated TSC.'l

. 3;I;sTC22 - *:C L"30'N M .

, Et; .tc--

t,s: C TiSni303 i;3I:20912 -

Phone Numbers RE p. TOC:

~

?f *;C 4 *32*M IC0530 3M

• Position DIM PAX 2,; t g . e c ;; -Sic m 31ay:s ~;. M

,.,,.t.....,. pn: i e 2.w-: .: :.' 3::

i

.. Site Emergency Director 3703/3704 .]

2212 ,

L aM j

Operations Manager - *: 3707/3708

. : t r-22'21 ,.

)9 z :1, Maintenance Manager

.y 3797

- 2235/2207 s Technical Assessment Managed

. v 3786 2405/2406 i

. cit . . . 20L -

CECC Connunicator 3701 2202

.- I NRC Coordinator 3701 2202  !

Secretaries 3701 2202 -

TSC Communicator 3628 (U-1) 2191/2192 (U-1) 3629 (U-2) 2291/2292 (U-2) ' 'n- -

3630 (U-3) 2391/2392 (U-3)

Chemical Engineer 3788 2405/2406 i

, Operations Specialist 3788 -

2405/2406

^* )

j ~

~

• Page 19 1

,  ? .. . 1 r! q I

y 0105p

. .w.. . -

.. t

_m ~ _. _ = ,._&.l. . _. .

'

• L.$'$, @$Y n . SS ' A c- +e.-- - - - - - - - o e ..-----r - - - - - -

Page 3 of~3 3FN - IPD SFN, IP-20

~

Attachment I RELOCATION AREA FOR TSC REY 0001 -

- . ; ,* 't 1

i - -

i .

.svrs

h.  ; - '

i russeggej & Off/A?'M N3 contarsez .orretrawa- swposrr c w yss MQA { $ ####### A**" ^

.rx:rtswsw r pro risers /me s -

(l, fritfrsAf p5::?n ,t A4**7" M**'7M Q

(: J, g ,, .- ,

connoaan y'

. lj.

l $,rl

?

~ l. _. .f. k I Mj

[' '

~ '

QL "S

/ I her uurr $"

C h &( &Y i /I {Q j""#" ].

3 -

Ni?$r

((h[t g

't 1 h

G

- ',t r 1 e .e ~~ C' )d(

s nu noau

,,,h b rosur tilho k

$ } ej i(q

}M . !

8 b M

(,_.,- J u .

-h - Ik3 , k' )9 lh

. _4_ -. n t2 s g uswn N. 3

- W s.r.rk. ~"

lc=woon .

t l.L . Hr n

u, p ,' j f g i . * 'M \

lu (l

~ ' , d- prxrx es seys- , yace: Il - .

S l,,';l -

G g

,4 ..

A7;rks

-- s

.ysse.r ?do b'"l

-e. .

4[ g g <:

?

n. , .

. .i,,..... _

l ,  ; ,,,j 3 . ,x t_

~" -

-; --~

4g Z s - . -

--w-Q__ y

$ l0 .

(1) Site Area and General Emergencies Only Maint. Engr. Desk Assignment Technical Assessment "

Staff Desk Assig'nment

2. PSO Engineer

5. Electrical Engineer 1. Rehetor Engineer Plant

9. Mechanical Engineer Assessment Team I,eader Engineer 2. Systems and Test Engineer

4. Operations Specialist

• 13. Instrument and Controls 5. Chemical Engineer

6. Plant Assessment Team Engineer h

0105p Page 21 ..

RO-HLT 1.03 9

SRO-HLT 5.04 RO-RQ l.02 N SRO-RQ 5.02 t TVA CONNENT ALLOW ALTERNATE FORMS OF NPSH EQUATION AS SHOWN BELOW PART A NPSH = PACT - PSAT OR NPSH={Psys + P

• h - head losses) - PSAT OR NPSH = Paystem + Pwater column - PSAT - Plosse, OR NPSH = (Psystem + Pwater column - PSAT - Plosses )_Ke_

P88 OR NPSH = (Pauction - PSAT) _ge 98 TVA RESOLUTION Part b.1.

Either increase or decrease would be correct depending upon what caused the FW flow increase / power increase (See Attached).

Part c.

The clarification by examiner about the pressure increase, when  !

asked 'Real or Theoretical', was ' Theoretical', and should have been

'Real'. Theoretically with a system not at TSAT, NPSH will increase if all that changes is system pressure..

r ,

i l

R0-HLT 1.06 SRO-HLT 5.05

'% i J

RO-RQ 1.05 bblb )\!

SRO-RQ 5.03 t TVA CONNENT a) Question asked for the " rapid" decrease from B to D. " Rapid" decrease occurs because you have a large Ng , and then go to 100% $th (which trace shows). The Burnout being so large, rapidly decreases Xe to 100% equilibrium. In 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> AxeNxe would do very little.

b) Question asked for change in flux distribution during period from B-D.

Reference:

Reactor Theory, Chapter 6, Page 6-9 TVA RESOLUTION a) Full credit should be given for just addressing burnout, b) Change in flux would be a redistribution from outside peaked and depressed in the middle (due to Ze) to middle peaked, and depressed on outside. Answer describes beginning flux pattern, not changes in flux distribution.

l l

l

i I

R0-HLT 3

SRO-HLT 5.06 RO-RQ l

SRO-RQ 5.04 t TVA COMMENT i

Part c. '

"immediate concern" is relative, l i

Part b.

Raise level is the correc't answer if I have no Recire. and no RHR SDC flow. If I have RHR flow which I did, the correct response would be to increase RHR flow /SDC flow. )

i l

l l

TVA RESOLUTION  !

Part a.

i No longer in ' cold shutdown' is also an indicated problem Part b.

Should also accept increase SDC system flow for full credit.

(Ref. GOI 100 page 9)

Part c.

delete or accept anything within reason.

I

RO-HLT N

SRO-HLT 5.08 RO-RQ reb 9)L}

SRO-RQ 5.06 s L

TVA C0lOfENT Point value of Part 'B' should be .5 instead of 1.0. Examination indicated question was worth 1.5 and not 2.0.

i RESOLUTION Change value of Part B to .5.

4 k

-n -,- . --...c., - - - - - , - , - - , - . - - . . . . , , -

RO-HLT 1.08 SRO-HLT

) (

R0-RQ 1.07 REE MS SRO-RQ t l

TVA CONNENT l l

C. Maximum could be 3358 MWT since no specific times are given.

E. Could have been given in %  !

i TVA RESOLUTION C. Should also accept for full credit 3358 NWT (Ref. GOI 100-1)

. E. Should also accept 25%.

Reference:

Tech. Specs.

1 l

l l

l I

l s

-,e - , - , , - - - . = - , . - - - , . , , - . -, --- -- ---,,r. -,- -, -

l l

i RO-HLT 1.10 D

SRO-HLT 5.11 1 ,

RO-RQ 1.09 SRO-RQ 5.09 .) t TVA ColOIENT Part f.

Asked to " State the factor" not to " state and explain use of the factor".

l 4

TVA RESOLUTION

Part c.

Should except for full credit "Eg".

i i l

I l

RO-HLT 2.01 SRO-HLT

)

RO-EQ 2.01 KEb 92.L SRO-RQ L

TVA COMMENT 4

Part B: -

65 psig is the maximum SDV backpressure allowed " initially" during scram to insure against affecting drive scram time.

I RESOLUTION Part B - Should be deleted since the question did not state " initial" volume pressure on a scram signal. The only thing that would limit pressure after the scram signal is present would be the relief valve. Thus 1250 (answer 4) should be accepted as correct. (Ref. OPL171.005, p. 29) l I

l e

_ . _ - _ _ _ , _ _ _ _ _ , , _ . _ . . . , .,_ .m- - . _ _ _ _ _ _ , ., yy.

i RO-HLT 2.02 m

SRO-HLT l RO-RQ 2.02 bD D SRO-RQ s t

l I

TVA CONNRNT Part A: I doubt that a 1" relief valve failing open would have any i significant effect on the NPSH available to RCIC. Until the suction supply is lost, flow would be out of system and air would not enter. RCIC is protected by a low suction pressure trip also.

i RESOLUTION Part A: Other acceptable answers should be:

1. Loss of water from the CST or Torus (depending upon suction path)

2. Flooding of the sump to which the Relief valve vents l

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

RO-HLT 2.03 3

SRO-HLT RO-RQ 2.03 b KCD 92 4 1

SRO-RQ -

L 4

i TVA CONNRNT '

Part A: Since the flow element is downstream of the minimum flow (OPL171.042, p. 32), the control system will increase speed of turbine to get rated flow. Since the minimum flow line has an orifice in it, the system may or may not be able to reach rated flow. With the minimum flow valve open, the torus will start to fill up.

Part B: The question asked for only one adverse affect '

Part C: The question asked for only one adverse affect 1

l t

l 1

RESOLUTION l Part A: Accept as another correct response:

(1) increase in Torus level due to HPCI pumping water thru minimum flow valve to Torus.

(2) Draining of the CST to Torus 1

j Part B: Accept for full credit either:

(1) Ninimum flow protection lost or, 4

(2) Turbine speed will be at minimum which could violate the j

minimum speed requirements for HPCI Part C: Full credit should be given for " release of airborne activity to the HPCI room or Rx Bldg. Should not required a statement concerning " loss of vacuum" for full credit.

I j

1

, . - - - , - = - , . - - - - ,. -- , - - , . , , - - , . , - -- - , - - - - - - - , - -. . . .e e--v.---

v-t

. y<- ,: ,r <jo ,

q_-

q . . . s,

, g t

n ,. ', j '

d RO-HLT 3.14 m d

i i(se ,

SRO-HLT "

r. I\

y,, _ y cs I RO-RQ 3.14 Y ~ 'd O D M i

- f<,'c .

, i i

x SRO-RQ 1

s s

s, y-

\ '

,g TVA CONNENT

\,\ i s - i Could get as answer: Lower electrical or mechanical stops os the.#!eiley '

drive. They are set at approximately the same position. , C ,)

i .

Ref. OPL171.008, page 11. '

i

[<4 .' N t

' s. ,

~

. , y,; y\

L

( N. 's . \ .  !

5:,

s s s.

s . , . .,

i

~- 1 4

2

\'

\

r .'s,t

\

i

% v s r.,, ,

• +

' , s i s ,e s.,

TVA RESOLUTION c' -

[ ' t }', ' l s i l

Accept for full credit either: Lower electrical or mechanical stop. t-

{t s j ss l .,

\

s( t's s.

s,

, I .e

, ' ' , , t,

'<!,+ ,

t( 4 t t

• y ',;

\ ' 1zlv r

1, u 'yl

.=

\ \

s.-

s, ( ' '

4 '; s s

't b [

'(

T( yts W .

l x

t vic . s ,x y},:. l a f 3

g g I

a N 4 #

/i

g. o g' ,=

=s -

,. 3

'k ;L'

/

\' \

, s , ., q

+ '

kj 3 r 4 s, s ,\.

.g-[

-h ,/,,

t *)r a I

g'\ '.-

(  %

RO-HLT 2.04 m

SRO-HLT 6.17 RO-RQ i

2.04 f J

l' .'

SRO-RQ J

t

(

i 1 TVA CONNENT I

(. ,

g

,Tiand to expand answer key to include another potential flow path (i.e. Test i

' ve.tvos to the Torus) . The test valves (74-71 & 73 on Sy 2 or 74-57 & 59 on Sy l

1) could be opened during shutdown cooling operation with adverse affects.

, 1 I

I

.;I t

• ! , (I

')

, i RESOLUTION 3

N e.s

' Accept as correct answer the " Test valves opened" The procedure (OI-74) ei

, doesn't address any specific action, except closing of the test valves after i " .g' flushing complete. This should be acceptable since the question addressed shut'down cooling with "and flushing" in parenthesis. The parenthesis could

, lh have lead to confusion as to what was being asked.

l l

• e e s i

1 j

b 0

i f,

/' s t

,k 7 ',o y , , , - - . , _ . . -., , .- ,-- _,_.e.- , , ~ , -..-, . , --

-s RO-HLT 2.06 SRO-HLT

)

RO-RQ 2.06 r REE 9 M SRO-RQ s t

TVA COMMENT

" Reverse Power" would require HEA LOCKOUT RELAY locally to be reset alonE with Field Breaker to call D/G in standby readiness.

Examinee may answer: " Reset HEA LOCKOUT RELAY RESOLUTION Accept for full credit: Roset lockout relay. Either the NRC answer key answer or the above should receive full credit. By procedure (OI-82) the field breaker must be reset. Also the lockout relay must be reset.

(Ref:45N767-4)

.9. ,,

O

R0-HLT 2.07

)

SRO-HLT 6.09 RO-RQ 2.07 f KEE 9%

SRO-RQ 6.09 J L

TVA CONNENT Part A: - Turbine cross-over pressure is physically same as LP turbine inlet pressure. Also could be cross-under pressure since this is the physical layout.

RESOLUTION Part A: Accept (1) crossover or (2) crossunder or (3) LP turbine inlet or (4) HP exhaust as correct (OPL171.014, p 10 and OPL171.010, p 35)

RO-HLT 2.09 _

SRO-HLT 6.10A

(

R0-RQ 2.09 f neb 933 SRO-RQ 6.10A .)

L TVA CONNgNT Need to expand answer key to include LPCI initiation signal present. In order to spray the Torus, the in-line valve interlock must also'be bypassed. To bypass the in-line valve interlock requires that an LPCI initiation signal be present in addition to the items given in the answer key. At BFN, we require that they give this to receive full credit if a number of responses is not requested. This is because, we might want to continue to spray the Torus even after the LPCI initiation signal has cleared and been reset (i.e. alert the oportor as to what will occur if it is reset while spraying). Also 2/3 core height is equal to -39 inches.

RESOLUTION Give full credit for any three of the four following conditions (even though LPCI signal was given in question, it is still a requirement to spray unless bypass switch is used):

i 1. LPCI initiation signal

2. 2/3 core height (-39 inches)

3. D. W. pressure >1.96 psig

4. Roset switch to reset i

Also accept 2/3 core height or -39 inches for full credit.

REF: OPLl71.044 p 19 & 20 OPLl71.003 p 16 i

I

RO-HLT 2.11 SRO-HLT 6.11 R0-RQ 2.11 h M.3 33 D~

SRO-RQ 6.11 y L

TVA COMMENT The color of the light should not be required for full credit.

RESOLUTION Give full credit if the answer addresses a light indication above the control switch.

l I

l i

9

.RO-HLT 2.12 SRO-HLT 6.12 RO-RQ 2.12 EB 9%

SRO-RQ 6.12 s t

TVA CONNENT .

Part C: Is unclear as to if we only reset once or each time the timer is initiated.

The question did not require a discussion as to why a person answers the way they did so partial credit would be difficult to get.

RESOLUTION Give credit on part "c" for either response since the question was unclear and the student should not be punished if he/she assumed the wrong thing.

If the student does write a reason for his/her response then credit (full)

! should be given if reason is valid.

l Another action would be to delete part "c" and increase valve of part A & B.

t

]

,_ . . _ , _ . , , _ , - -- --w- e ~ ' ~ ' ' - ' ' ~ ' ' ' ' ' ' ' ' ' ' ^ ^ ~~" '

1 RO-HLT 2.13 3

SRO-HLT RO-RQ y REE 93% 1 l

SRO-RQ s l

4 i

TVA C0f0fENT l Clarify answer key for RCIC mechanical overspeed trip j l

1 l

I l

l RESOLUTION The mechanical overspeed on RCIC must be reset locally first and then run the 71-9 back open.

Reference:

OPL171.040. p 21 l

,- . . , . - - , - . ..,, ,,,,-_.e.n.--, .- - , ,. - ,

RO-HLT 2.14 SRO-HLT 6.18

)

R0-RQ 2.13 KEE %9 SRO-RQ s L

TVA COMNRNT The requirement is loss of voltage on the associated S/D Bd-Not Bus.

RESOLUTION l

i Accept loss of voltage or undervoltage on the board as correct answer.

Reference; OPL171.038, p 11.

L l

3

RO-HLT 2.15 SRO-HLT 6.13 R0-RQ SRO-RQ s TVA COMMENT I could not find a reference for the answer on the NRC key for part 'B' of the question. There could be many different responses, as indicated on the key.

RESOLUTION Some possible correct answers for Part 'b':

1. Increase in DW pressure

2. Increase in DW temperature

3. Damage to electrical components in Drywell

4. Increased radioactivity in the Drywell S.

Increased leakage from containment as indicated by sump pump run times.

There could be others, evaluate each response as to its merit.

l l

1 j

l l

I

RO-MLT 2.16 7

SRO-HLT RO-RQ SRO-RQ d L

TVA CONNENT Part B: This is a true statement if certain assumptions are made. To be true for all four RHR pumps, both test switches must be out of auto.

With only one switch out of auto, three RHR pumps will still start.

The preferred pumps (i.e. A & B RHR on Unit 1) get a start signal from both logic systems and one other pump would start from the non-affected logic system.

i RESOLUTION 4

Part B:

The answer could be TRUE or FALSE depending upon what the student assumed. One RHR pump will not start for LPCI-TRUE. The other three RHR pumps will start-FALSE. See attached drawings.

l l

4 1

i

_ , - __, _ . ,m. _ _ ,- __m., . . _ _ , _ , . . , . _ , . .-._c ____.,,.r... _ _ . _ _ . . _ , _ . _ _ . _ , , _ , _ - - - _ _ _ _ . , . , _ , , _ , , . , _ _ . _

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

RO-HLT 4.15 SRO-HLT 7.16 R0-RQ R63 3Y SRO-RQ s L

TVA COMMENT Due to the way the question was worded, someone could have thought that all shutdown cooling was lost. This would have changed their answer.

i RESOLUTION Accept for full credit the following

1. Raise Rx level to +60 inches

2. Increase RWCU flow for mixing REF: GOI 100-12, 3F l

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

RO-HLT 3.18 7

SRO-HLT RO-RQ IND Db SRO-RQ ) t TVA CONNRNT Part a:

The answer key states " Speed Demand Limiter" as the correct response to

" which component". The Dual Limiter will be the controlling component for this event. (Ref. OPLl71.008, page 37 and NRC exam figure 474)

Part b:

I The decrease in speed will be limited by the lower elect. stop (limit switch) or the low speed mechanical stop. These are set at approximately the same.

' l l..

TVA RRSOLUTION Part a:

Accept Dual Limiter as a totally correct response also.

Part b:

j Accept either lower electrical (limit switch) or lower mechanical stop as correct answer. (Ref. OPL171.008, page 11) i I

,.w.- .- - . _ - - . . _ . s___.,_, . . ..- . . - . _ _ , . _ .m,

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

RO-HLT 2.17 m

SRO-HLT R0-RQ ' N E.1 9bb SRO-RQ b L

TVA CONNENT The figure given out was of poor quality and very difficult to read. The answer key needs to be changed to allow a greater range of response. The object of the question was to verify their ability to use the chart.

RESOLUTION Change answer key to Part a: 510 + 20 Part b: 1060~1 40 l

9

RO-HLT 2.18 SRO-HLT RO-RQ reb 35G SRO-RQ t

TVA C0lOIENT Accept another answer for full credit RESOLUTION Should also accept for full credit " initiation of 480 volt load shed logic."

OPL171.047 pg. 13 l

RO-HLT 4.18 SRO-HLT 7.06 RO-RQ l ND 9bb '

SRO-RQ 7.06 L

i TVA CONNENT j

Antwer key wrong for Rx level.

Should be "Rx water level <11" vice >11" I

\

i RESOLUTION

, Change answer key to <11" 1

REF: E0I-1 i

i 1

I

R0-HLT SRO-HLT 7.20 RO-RQ NE3 9 6(,

SRO-RQ J L

TVA CONNENT A clearance is the means used to ensure that lines or equipment remains de-energized, depressurized, and isolated for work to be safely performed.

This is done by racking out or opening all sources of energy, circuit breakers, or switches, pulling fuses, or lifting wires, etc. that are necessary to complete the clearance. (REF: BF 14.25, p. 6). Therefore, if a student responsed with tag the knife blade switches, full credit should be given. The switches would not be tagged until they were opened.

RESOLUTION l Accept for full credit, tag the knife blade switches.

't l

I i

RO-HLT SRO-HLT 8.04 RO-RQ -

K E 3 c3 7 /

SRO-RQ 8.03 TVA COIDIENT The question as stated indicates a valve that is less than 30 open.

(RRF: Tech. Spec., P. 271)

Check light-on Green light-on Red light-off i

Tech. Spec. is met, check light is on indicating valve fully closed and Red light off verifies that valve is less than 30 open.

4 l

l RESOLUTION l Answer should read as follows: yes. Light indications given indicate the

! vacuum breaker is fully closed (in a normal lineup).

a l

1 .

R0-HLT SRO-HLT 8.05 RO-RQ q K Gh M 1 l

SRO-RQ 8.04 s TVA CONNENT Part A. Question ask for 2 criteria that must be met. The answer key list two criteria in the same statement which are really two different items.

1 RESOLUTION A. Any 2 of the following should receive full credit.

1. Cause of scram identified and understood..

2. Reasonable assurance the unit will not scram again due to the 1

same cause.

3.

Response of safety related systems are understood and acceptable.

RO-HLT 4.19 SRO-HLT 8.07 RO-RQ 4.12 OD EN SRO-RQ 8.06 s L

.6 TVA COMMENT NRC should evaluate each response for technical merit and not verbatim response from procedure, i

l l

RESOLUTION l

Possible answers in addition to answer key could be:

1. Responded abnormally as determined by operator or skilled craftsman

2. Time limits exceeded

3. Component falls to operate

RO-HLT SRO-HLT 8.09 R0-RQ -

KE% 9 W-SRO-RQ 8.08 ,s i

t TVA CONMENT Part A: Add the criteria of an individual's lifetime exposure compared to the allowable exposure of 5(N-18)

Ref: RCI 1 pg 15 footnote (1)

After " Volunteer" answer key should also accept "or members of an emergency team and have previously consented to receive this exposure." Ref: RCI 1 pg. 15 footnote (2) and IP-15 page 1.

1 Should also accept " older volunteer" or >45 for full credit since we might not have anyone available that is older than 45.

! Part C: Based on IP-15, page 1, the allowable dose to the extremities permitted is 275 rem (75 rem whole body +200 rem additional to extremities)

Part D: Expand answer key as indicated below j

f I

RESOLUTION Accept c.

addition to answer key for part a and correction of answer key for part Part D: Should also except for full credit "within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />" (via red phone). Per IP-3 procedure and explanation of "immediate" contained in 10CFR50.72.

l Also IP-3 should be accepted as well as ' Alert'.

4

-~---,.y- - - ,,~_, 7 -w-mr , - - ,-,,-,-,.,----e ---w -

r---e ,w- ..-m_,---._ --

7 ,,. vww.v-_ -- .---- . =,..,-w.-wn-. -,,-,s-g.-,,,-w-,-e-- - - ,

RO-HLT SRO-HLT 8.08 i

RO-RQ 5 KE% T75 SRO-RQ 8.07 >

L TVA CONNRNT Question is very ambiguous, and the answer could involve many things. Part of the answer requires the operator to conclu'de a safety limit had been violated, which cannot absolutely be concluded based on the information provided. He can conclude that a check definitely needs to be made to insure the Power Transient Safety Limit was not exceeded.

The answer only contains the " unique actions" required regarding the Reactor Vessel Pressure Safety Limits and not all the unique actions required. On a timed test with a question that is as "open ended" as this, a lot of valuable time and effort can be wasted in an attempt to cover all angles. If you are specifically looking for " unique actions" pertaining to a specific limit and/or system, the question should be worded to ensure it solicits that response.

The first part of the answer is taken from a sentence in SP 12.8. The Shift Engineers responsibility for Post Scram Report does not require him to review the charts / graphs but to provide them to the STA. The write-up and evaluation of plant response is the responsibility of three individuals as stated on page 47 of BFSP 12.8.

Since RPV press, spiked to 1287 psig (an exact pressure) one could conclude that the pressure limit of 1375 psig had not been exceeded and no unique actions are required.

RESOLUTION Delete the question is one option or, accept any of the following as an answer based on actions required per SP 12.8 because of the pressure of 1787psig.

1. Verify received a high pressure scara signal (11055 psig).

i

2. Verify recirc. pumps tripped (1120)

3. Verify all MSRV's opened.

4. Evaluation made to ensure safety limit of 1375 psig not exceeded.

5. Verify / ensure MSIV closure initiated / caused scrammed.

high pressure, appears it did not. Based on the l

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

RO-HLT SRO-HLT 8.10 RO-RQ SED 977 SRO-RQ 8.09 )

t TVA CONMENT The question asked to state whether the Tech. Spec. LCO's are met. The answer key strictly addresses only one specific LCO. The way the question is written, one would think that we are concerned with electrical power requirements. The question should have been more limiting by specifying the LCO's to be addressed. To be able to answer the question per the key would require detailed knowledge of the feeds to each piece of equipment. If this occurred, the operator would use Attachment 20 of BF 12.24 or the electrical print out do determine what was lost or what failed to meet electrical supply requirements. Since the question addressed loss of two Diesels with all three

' units in cold shutdown, the electrical requirements as listed in Tech. Spec.

3.9c. would be of concern.

l i

RESOLUTION

' If the power supply requirements are addressed (i.e. two Diesel operable with one of the offsite power sources available), full credit should be given.

, R0-MLT SRO-MLT 8.11 1

I 20-RQ XEh U 8 a

. SRO-RQ 8.10 J t

I 1

TVA CONNENT See 8.09 l

A. Question should state which LCO's need to be addressed and give additional information as to the status of the unit.

1

. B. This plant condition is obviously a safety concern. LER's are written when violations occur with regards to safe operation of the plant. If i

outside a Tech. Spec. LCO, how can it not be a safety concern. In addition, if the head is torqued and RNR is keeping the temperature above 100*F, then there is a concern about keeping temperature up.

4 (REF: Tech. Spec. 3.6A.5) j l

l 1

l 4

i l

RESOLUTION i

part A:

i Another possible answer for full credit would be. If the plant is in the following conditions LPCI and containment cooling are i

not required (REF: Tech. Spec. 3.5.B.10).

l I

a. Head is removed (not sure with conditions given)

b. No work being done with potential to drain vessel

! c. Fuel pool gates open and level in pool above low alarm j d.

j At least one RHRSW pump operable for standby coolant (still 1

have B1 and 82 RNRSW)

Part B

Answer should be "yes".

i See TVA comment Part B. As an SRO, it 1 would be a safety concern, until the plant staff makes a determination it isn't.

4 f

i j

i i

ew,en---v-o.w,- ,.w<,- ---+-,-----e-,vwc--

ev--------m--v-te-=, e-e -m 3,--.----em,-. ----w,-.w my-,e-----,m,,--- -,---*.-et---v--r

4 RO-HLT 3.20 SRO-HLT RO-RQ REB 07g 3 SRO-RQ t TVA ColOIENT Part a.1. - The operator would be watching the frequency meter while

! adjusting the speed of the D/G.

i TVA RESOLUTION Part a.1. - Accept frequency or speed control as totally correct.

i I

1 l

l l

l l

RO-HLT 4.06 l

SRO-MLT 7.05 RO-RQ 4.05 SRO-RQ 7.05 U L

TVA COIDIENT Part A: Should be true. Note on page 2 of GOI-100-1 states no more than 4

1 loop of RHR can be inoperable due to closed torus suction valves. RHR A and C make up one loop.

RESOLUTION i

Change answer in Part 'A' to TRUE.

REF: GOI 100-1, p2 l

1 J

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

l ENCLOSURE 4 REQUALIFICATION PROGRAM EVALUATION REPORT Facility: nrowns verry Nuclear Plant 1 Examiner: K. E. nrockman l Date(s) of Evaluation: November 17 - December I, 1986  !

Areas Evaluated: x Written Oral x Simulator 1 Examination Results:

R0 SRO Total Evaluation Pass / Fail Pass / Fail Pass / Fail (S. M or U)

Written Examination 4/o 5/1 9/1 S*

Operating Examination Oral  %

Simulator 4/o 5/o 9/o S*

Evaluation of facility written examination grading N/A Overall Program Evaluation Satisfactory

• Marginal Unsatisfactory (List major defi-ciency areas with brief descriptive This is an iterim evaluation of the Browns Ferry Accelerated Requalification Program. Progress is satisfactory at this time. A final program evaluation will be made af ter the March 1988 examinations.

Submitted: Forwarded: Approved:

/bYbYe Ex ner Frv Q (kph& W '

0

/5ect16FChJef Branch Chief K. E. Brockman qJ. P.-Munro - C. A. Ju1L u