ML20236B675

Forwards Util Comments Re Written Exams for Reactor Operator & Senior Reactor Operator Given at Plant on 881114

ML20236B675
Person / Time
Site:	Farley
Issue date:	11/18/1988
From:	Dennis Morey ALABAMA POWER CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20236B561	List: ML20236B558 ML20236B562 ML20236B675
References
FNP-88-0572-TRN, FNP-88-572-TRN, NUDOCS 8903210224
Download: ML20236B675 (47)

Text

_ _ _ _ . _ _ _ _ . _

Alabama Power Company J N ear Plant ENCLOSURE 3

.p F D la' er ,

$shford, Alabama 36312 #=

Telephone 205 899-5156 Y .

m Alabama Power the Southem eftXlnC SySlem FNP-88-0572-TRN November 18, 1988 i

The Regional Administrator, Region II Nuclear Regulatory Conmission 101 Marietta St., N.W.

Atlanta, GA 30323 ATTN: Mr. John F. Munro, Chief Operator Licensing Branch Enclosed are Alabama Power Company's comments concerning the written examinations for reactor operator and senior reactor operator given at Farley Nuclear Plant on November 14, 1988.

Questions appearing on both examinations have been addressed on the reactor operator exam comments only.

The courteous and professional manner which your staff displayed in preparing and administering this exam is appreciated.

For further clarification or discussion of these comments, please contact Mr. Lee Williams at (205) 899-5156, extension 6106.

D %u.f D.N.Morey,IIf General Manager - Nuclear Plant DNM,III/LSW/R3W:mjk Enclosures Lb <!$,$<jj  ;

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_ _ _ _ . . . . . . E

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REACTOR OPERATOR WRITTEN EXAMINATION i i

QUESTION 1.02 (2.00) l Choose / circle the appropriate word which completes the following j sentences: '

Given: maximum heat generation rate Peaking Factor =

average heat generation rate

a. The (axial / radial / local) peaking factor would be the one which accounts for the variance in power up and down the vertical axis of the core.

b. The (operating / design / accident) peaking factor would be the one which exists in a " worst case" condition which might be expected at any time in plant life.

c. The (axial / radial / local) peaking factor would account for variances in power produced at various distances out from the vertical axis of the core.

d. The (operating / design / local) peaking factor allows for the fact that power production varies in an assembly.

ANSWER 1.02  !

(2.00)

a. Axial (0.5)

b. Design (0.5)

c. Radial (0.5)

d. Local (0.5)

Reference:

NUS, Module 4, " Plant Performance", Chapter 10, "PWR Performance, Section 10.2, Farley LP OPS-52102J and Westinghouse; " Nuclear Power Distribution & Core Control" Material KAIR 2.9/3.1 2.9/3.3 192005L112 193009K107 ...(KA'S)

COMMENT The b) portion of the question addresses " operating / design / accident" peaking factors. This terminology is not consistent with that used at i FNP. The following reference, NUS, Module 4, " Plant Performance",

Chapter 10, "PWR Performance Section 10.2, is not taught at FNP as evidenced by the Thermodynamics, OPS-309, student proface which is attached. The second reference, Farley LP OPS-52102J, does not J

(

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': );.

reference peaking factors. The third reference, Westinghouse;." Nuclear

Power Distribution and Core Control" Material, is not taught at FNP.

~

~T he d) portion of the question also addresses " operating / design" peaking factors.- FNP license training addresses radial, axial, and local peaking factors. The terms " operating / design" are used only as distractors. Since a correct answer, " local," does exist, the use of-the terms in part d) is valid.

RECOMMENDATION-Delete part b) of the question and' reduce the total question point value from-(2.00)'to (1.50).

i 5

l

- - - - - - - _ _ _ _ _ _ __________o

, 6

. QUESTION 1.08 (1.50)

List.five (5) indications of natural circulation. '

l ANSWER 1.08 (1.50)_

a. RCS subcooling monitor indicates "becoming more subcooled" (0.3)

b. RCS T-hot leg temperatures stable OR trending down (0.3)

c. Core exit thermocouple stable OR trending down (0.3)

d. S/G pressure stable OR trending down (0.3)

e. RCS T-cold temperatures at saturation temperature for (0.3) i S/G pressures

Reference:

Farley LP OPS-30913, Obj 16 and Qual Req'mts ERP's/ ESP-0.2 KAIR 4.2/4.2 3.9/4.1 193008K122 193000K123 ...(KA'S)

COMMENT The key answer addresses only those indications which are delineated by the FNP Emergency Response Procedures. The FNP Abnormal Operating Procedures also address indications which the operator should observe during natural circulation. FNP-1-A0P-4.0, " Loss of Reactor Coolant Flow," step 6.1 and the preceding note address RCS AT < 65*F as an indication of natural circulation development. A copy of the applicable A0P-4.0 page and referenced Natural Circulation Log are attached.

RECOMMENDATION  !

Accept the following as one of the five (5) correct responses: .

RCS AT < 65'F (0.3) and change the answer key to read: (any five [5] of the following six

[6] indications - each worth 0.3 points) .

a. RCS subcooling monitor indicates "becoming more subcooled" (0.3)

b. RCS T-hot leg temperatures stable OR trending down (0.3)

c. Core exit thermocouple stable OR trending down (0.3)

d. S/G pressure stable OR trending down (0.3) i e. RCS T-cold temperatures at saturation temperature for (0.3)

S/G pressures

f. RCS AT <65'F (0.3) 1 l

7 QUESTION 1.11 (1.50) i SCENARIO: i During a recent outage, three (3) percent of the "A" S/G tubes were plugged.

"B" and "C" S/G's were not inspected nor plugged TRUE or FALSE Due to the reduction in the number of tubes available in the "A" S/G" ,

a. The total heat transfer rate of the "A" S/G will be less than that of the "B" or "C" S/G's.

b. Available (total) plant power output (MWe) is reduced because of the "A" S/G output pressure reduction and the subsequent effects to Psat versus Tsat.

c. The heat transfer coefficient ("U") will increase in the "A" S/G and decrease in the other two S/G's.

ANSWER 1.11 (1.50)

a. False (0.5)

b. False ',0. 5 )

c. False (0.5)

Reference:

Farley LP OPS-309078, Objective 1 and Farley LP OPS-30909, Objectives 3 and 9 and Farley LP OPS-30911, Objective 9 and LP OPS-521010, Objective 5 KAIR 4.1/4.2 002000K111 ...(KA'S)

COMMENT The answer to the a) portion of the question is incorrect. A reduction in the total number of S/G tubes in the "A" S/G reduces the heat transfer area (A) of that S/G. The heat transfer coefficient (U) of the remaining unplugged "A," "B," and "C" S/G tubes is unaffected by the plugging of S/G tubes in S/G "A." If it is assumed that,the AT is maintained constant, then the total heat transfer rate (Q) of the S/G's will decrease per the following formula:

i -, p, j

"A" S/G Q = UAAT

.,44 "B" and "C" S/G Q = UAAT

--_-_______________J

l> ,

The "A" S/G heat transfer rate will be'less than that of the "B" or "C"-

.S/G's.

J RECOMMENDATION Change'the answer key as-follows:

a)' True (0,5)'

I l

_______ _ _ _ _ . _ . . . - _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - - - - - _ - - - - - - - - - - - - - - - - - - - - --J

QUESTION 1.13 (1.00) .

Using the single " Pump Head (Hd) Versus Flow Rate (V)" graph provided below, sketch / draw:~

a. A " set" of two (2) Emergency Core Cooling. System (ECCS) pump curves-

- one (1) high head and one (1) low head - showing their relative position to one another. ,

b. A " set" of two (2) " system characteristic curves" depicting:

1. One.(1) " system curve" associated with initial reactor coolant  ;

system (RCS) pressure and; ,

l

2. One (1) other " system curve" which would be present during a '

major reactor coolant system (RCS) LOCA.

.1 s.

Hd .

V i

i

_ _ _ - _ _ _ _ _ _ - _ _ - _ - - - - _ - _ - - _ - - - _ = - _ _ - - - - - - _ - - - - - - - - - - - - - - - - - - - _ _ - - - - - - - - - - - - - - - - - - - - - - - - - -

i I

f j ANSWER 1.13 (1.00)

- USE ATTACHED FIGURE IN GRADING THE QUESTION --

Look for the following:

a. Correct placement / labeling.of Hi Head Pmp curve. (0.25)

b. Correct placement / labeling of Lo Head Pmp curve. (0.25)- y.

c. Correct placement / labeling of Init Sys curve. (0.25)

d. Correct placement / labeling of curve' depicting (0.25) system pressure reduction due to LOCA.

4

Reference:

Farley LP OPS-30910, Objectives 2, 4, 7 and 12; .LP'0PS-403020, Obj.10; LP OPS-52532H'and Qual Req'mt E0P's KAIR 2.8/3.1 3.5/3.9 006020K601 006000K506 ...(KA'S)

COMMENT The normal method of teaching system characteristic curves at FNP is with one pump curve and a system curve that originates at zero head and zero flow. The pump characteristics or system characteristics are then changed and the curves have to be redrawn for the. changes. This is indicated in the attached Figures 6.4-4 and 6.1-4.

Drawing the pump curves on the same graph should be no problem.

However, due to the scales involved, it will be very difficult to draw the system curves to coincide with the two pump curves, assuming the system curves start at zero.

RECOMMENDATION Grade the pump curves as shown on the key drawing. Grade the system ,

curves on an individual case basis. Full credit should be received if '

-it-is indicated that the RHR pumps will be at shutoff head at normal operations and increased flow as a LOCA develops, and that charging pumps initially have flow and flow increased as the LOCA developed.

i

a ,

i 7,7 . , ,

i I ' QUESTION 1.14 (1.50)-

TRUE or FALSE .

a. Since pressure stresses are higher on the RCS core vessel inner q wall during cooldown, restrictions on cooldown are more limiting than heatup restrictions. .

I

b. Due to the greater risk of brittle fracture, rather than cyclic i fatigue RCS. core vessel " tech spec" heatup/cooldown limits are i more restrictive than PZR pressure vessel limitations. '

c. The relief l capacity of a single PZR code safety (=345,000 lbm/hr)~ l

.is' adequate to relieve ANY overpressure condition (" potential PTS l problem") which could occur during shutdown (mode four (4) or mode 1 (5).

ANSWER l1.14' (1.50) I

a. True (0.5)

b. True (0.5) c.

True (0.5)

.]

Reference:

Farley LP OPS-30912, Objectives 4, 5, 6 & 7; Farley LP OPS-30913, Objectives 10 and 11; Farley Technical Specification 3/4.4.2 LC0 and Bases; Farley Technical Specification 3/4.10.1 LCO and Bases and Farley ..

Technical Specification 3/4.10.2 LCO and Bases '

KAIR 2.8/3.2 3.3/3.7 193010K101 193010K104 ...(KA'S) i COMMENT Part a. of this question is false'for two reasons: The Technical Specifications on heatup and cooldown are exactly the same 100* in any one hour period (Specification 3.4.10.1 attached). The heatup and .

cooldown curves show heatup as more restrictive above 180* and cooldown I more restrictive below 180', depending on rate used for cooldown.

In addition,.the pressure stress is always more tensile on the inner wall, re0ardless of whether the plant is undergoing a heatup or cooldown. The additional concern on cooldown is because the inner wall also receives an additional tensile stress. These stresses are shown on-pages 10.1-6 through 10.1-9 of NUS text attached.

RECOMMENDATION Change the correct answer on part a. to False for .5 points.

. . - 1 QUESTION 2.03 (2.00)

Unit One-(1) has experienced a LOCA resulting in a Safety Injection (SI)

. and all related equipment has functioned properly.

For the following Service Water System (SWS) loads, note if the designed !

SWS flow to.the load is " GREATER-THAN", "LESS THAN" or of the "SAME" magnitude as it was before the LOCA.

a. Component Cooling Water Heat Exchangers (0.2) l

b. Containment Coolers (0.2)

F

c. Control Room Air Coolers (0.2)

d. RHR (LHSI) Pump Room Coolers (0.2)

e. Charging (HHSI) Pump Room Coolers (0.2)

f. Auxiliary Feedwater Pump Room Coolers (0.2)

g. ' Battery Charging Room Coolers (0.2)

h. RCP Motor Air Cooler (0.2)

1. Turbine Building Heat Exchangers (0.2)

- j. Diesel. Generators (0.2) i ANSWER 2.03) (2.00)

a. Same (0.2) l

b. Greater than (0.2)  !

c. Less than (0.2)

d. Less than (0.2)

e. Less than (0.2)

f. Less than (0.2)

g. Same (0.2)

h. Less than (Note: " Nonexistent") (0.2)

1. Less than (Note: " Nonexistent") (0.2)

j. Less than (0.2) l

d 1

Reference:

i

, - Farley LP OPS-40101B, Objectives 5 'and '11 and Qual Reg'mt Elements #CRO-357 KAIR 2.9/3.4 3.7/3.7 076000K403 076000A302 ...(KA'S)

COMMENT Five (5) parts of the question key are incorrect for the question as stated. .It appears that the key assumes that a loss of a service water train has occurred due to the LOCA event. This~is the assumption that was made in the Service Water lesson plan (OPS-40101B), Table 3, which was derived from FSAR Tables 9.2-2, 9.2-3, and 9.2-5, which are

. attached.

t

! Answers parts c) through f) for the ESF pump room coolers and part j) for the diesel generators state that room coolers will have reduced

. flows following the LOCA. No' automatic actions occur on a LOCA event to

.cause the service water flow to these components to change from.their normal pre-LOCA flow rates.

RECOMMEN0ATION Correct the answer key as follows:

c) -SAME (0.2) <

d) SAME (0.2) e) SAME (0.2) f) SAME (0.2)

'j) SAME (0.2)

- t

. ,. j QUESTION 2.15 (2.00)

List the four (4) sources of normal / emergency makeup water to the I condensate storage tanks.

ANSWER 2.15 (2.00)

a. Water Treatment Plant (deoxygenated makeup) water (0.5) i

b. Hotwell Condensate (via the condensate pumps) (0.5)

c. Demineralized Water Storage Tank (0.5)

d. Service Water (0.5)

Reference:

Farley LP OPS-40201D, Objective 11 and FNP-1-SOP-21.0

" Condensate and Feedwater System" and Technical Specification LC0 and Bases 3/4.7.1.3 and Qual Reg'mt S0P's KAIR 3.9/4.2 061000K401 ...(KA'S)

COMMENT The reference "Farley LP OPS-402010, Objective 11" addresses the normal and emergency sources to the auxiliary feedwater system, which.does include service water as an emergency source. However, check valves in the system prevent the service water from reaching the condensate storage tank. Farley LP OPS-40201B and reference FNP-1-S0P-21.0 does state the water treatment plant, condenser hotwell, and demineralized water storage tank as sources of makeup to the condensate storage tank.

RECOMMENDATION Deletc from the key d) service water (0.5) and change the total point value to 1.5. Accept as a full credit answer a) water treatment plant (deoxygenated makeup) water (0.5); b) hotwell condensate (via the condensate pumps) (0.5); c) demineralized water storage tank (0.5).  !

These three sources have a design connection with the condensate storage tank. A fourth source could not be accomplished without bypassing many design systems. Do not deduct points from the fourth source given as an answer if it is a probable source (e.g., bypassing the water treatment plant,etc.). Attached is page 13 from Farley LP OPS-40201B and page 15 from FNP-1-SOP-21.0.

L

++ ,

' '~ QUESTION 3.01 (2.50)

a. What PROTECTIVE function can be manually. blocked by the operator-

' when 2/3 (TWO out of THREE) pressurizer pressure channels fall below 2000 psig (P-11)?

a. LIST the TWO (2) PROTECTIVE functions which are " auto" actuated when 2/3 (TWO out of THREE) T-AVG instruments fall below 543 degrees'F-AND high steam flow conditions exist (P-12). j

c. As RCS T-AVG. falls belowlthe "Lo-Lo T-AVG" setpoint, all open steam dump-valves shut and any further system operation:is " blocked".

How is this specific form of CONTROL restored if further plant.

cooldown is desired? )

{

ANSWER 3.01 (2.50) ]

a. Low Pressurizer Pressure Safety Injection (SI) (0.5)

b. 1) Main Steam Line Isolation (0.5)

2) Low Main Steam Line Pressure SI (0.5)

c. The Train A and Train B steam dump interlock switches are passed in the " BYPASS INTERLOCK" position. (1.0)

Reference:

Farley LP OPS-522011, Objective 14 and 22;.LP OPS-40301E; LP OPS-52201H,-

. Objective 2 and LP-52201G, Objectives 6 and 13 KAIR 3.0/3.3 3.7/3.9 041020K409 013000K412 ...(KA'S)

COMMENT The second part a. of the question, part b. on the answer key, asks for two protective functions of high steam flow and Lo-Lo Tave. There is only one protective function provided, and that is main steam line isolation. .The only five safety injection signals at Farley are listed on page 40 and Table 2 of LP OPS-52201I. A logic diagram, Figure-4 from L/P OPS-52201K, shows the only functions performed by high steam flow and Lo-Lo Tave to be main steam line isolation. In addition, the question states that P-12.is high steam flow, when, in fact, it is Lo-Lo Tave.

RECOMMENDATION Change answer key part b. to only require main steam line isolation for

.5 points.. Reduce. total points of question to 2.0 points; do not deduct credit if second answer to part b. is reasonable.

l'

? i QUESTION 3.05 (2.00)

Concerning the Full-length Rod Control System:

a . -- LIST F0VR-(4) of.FIVE (5) conditions which can activate a Power i Cabinet " ROD CONTROL URGENT FAILURE" alarm (annunciator). I d

b. LIST TWO-(2) of THREE (3) " automatic" actions which occur..upon ,

activation of the above URGENT FAILURE alarm. i

c. TRUE or FALSE Upon activation of the " ROD CONTROL URGENT FAIL'URE" alarm, rod motion is still available in the other "non-affected" power cabinets.

ANSWER 3.05 (2.00)

a. [Any FOUR (4) of the following FIVE (5) conditions - each worth 0.25]

1. Regulation. Failure

2. Phase Failure

3. Logic Error

4. Multiplex Error

5. Loose or Removed Printed Circuit Card

b. [Any TWO (2) of the following THREE (3) actions - eeach worth 0.25]

, 1. Stops ALL rod motion when in AUTO or MANUAL

2. Orders reduced current to associated stationary and movable gripper coils

3. Overrides any orders from Logic Cabinet for the associated Power Cabinet

c. TRUE

Reference:

Farley LP OPS-52201E, Objectives 12, 13, 14 and 15 KAIR 3.7/3.9 001050A201 ...(KA'S)

l- i L ..- ,

q 4

COMMENT When a " ROD CONTROL URGENT FAILURE" occurs, the following automatic actions occur:

Stops all rod motion, in or out, in manual or automatic Inhibits the pulser Overrides all current orders from the logic cabinet t Orders a current hold to stationary grippers and movable grippers in the l affected cabinet The important operational concern is for the operator to understand that a simultaneous current signal is sent to the stationary and movable grippers -- not the magnitude of the current order (i.e., reduced).

FNP-1-ARP-1.6 Page 1 of 2 is attached for reference.

RECOMMENDATION Change the answer key for the b) part to read:

[Any two (2) of the following four (4) actions - each worth 0.25]

1) Stops all rod motion when in auto or manual

2) Orders current to associated stationary and movable gripper )

coils

3) Overrides any orders from logic cabinet for the associated l power cabinet ~

4) Inhibits pulser 1

-l 1

._ 's QUESTION 3.10 (1.00)

Concerning the Emergency Diesel Generators (DGs):

What are the positions available on a swing diesel's " UNIT SELECTOR SWITCH" and what " protective condition" has the operator placed the swing diesel in for these positions?

ANSWER 3.10 (1.00)

Unit I OR Unit II position (0.25) - Only UV signals and SI signals from the selected unit will start the diesel. (0.5)

Unit I AND Unit II position (0.25) - Signals from either unit will start the diesel. (0.5)

Reference:

Farley LP OPS-40102C, Objective 8 Farley LP OPS-52102I, Objective 1 KAIR 4.1/4.4 013000K112 ...(KA'S)

COMMENT The question does not solicit the information required by the answer key. The term " protective condition" is not standard terminology and is not defined or explained in the question.

RECOMMENDATION Change the answer key to read as follows:

Unit I position - D/G starts on Unit I signals only (.33)

Unit II position - D/G starts on Unit II signals only (.33)

Unit I and Unit II position - D/G starts on signals from either Unit

(.33)

$' y-

- l a

QUESTION 3.12 (3.00)

Concerning the Steam Dump System Control Interlocks:

LIST, 1) all CONTROL ("C") INTERLOCKS, 2) associated LOGICS / CONDITIONS (if any) and (3) a BRIEF DESCRIPTION of what is accomplished by each interlock.

l ANSWER 3.12 .(3.00) i D '

[Each Interlock worth 0.33, Logic / Condition - 0.33; Description - 0.33)

a. C-7 (Loss of Load Interlock) - Impulse Pressure signal from PT-447 (NOTE: No "real" logic associated with this interlock) 10 percent step or 5 percent / min ramp - Arms steam dumps in T-AVE mode.after load rejection, IF C-9 is present.

b. C-8 (Turbine Trip Interlock) - Turbine Trip Indications; i.e. 4/4 Stop Valves closed OR 2/3 Auto Stop 011 signals - Arms steam dumps in T-AVE mode after turbine trip IF C-9 is present.

c. C-9 (Condenser Interlock) - Circ Pump BKR closed and >8 inches Hg condenser vacuum - Prevents overpressurization and subsequent damage to the main condensers.

Reference:

Farley LP OPS-52201G,_ Objectives 6, and 9 KAIR 2.5/2.8 2.4/2.7 2.8/3.0 041020K414 04102K407 041020G007 ...KA'S)

COMMENT The key answer is not structured in the format of the question and the requirements for full credit are not clear.

Answer part a) states incorrect logic / condition values for actuation of C-7. As' stated on attached pages 23-26 of the FNP Steam Dump lesson plan-(0PS-52201G), a 15% load reduction with a 120 second time constant will arm the dumps due to loss-of-load (C-7).

Answer parts a) and b) should not require the inclusion of C-9 as part of the answer since C-9 is covered as a separate control interlock in part c) of the answer for this question.

_,_______m _______m_.___-__ _._--------_------------

g RECOMMENDATION

. Change the answer key as follows:

1) 'a) C-7 (Loss of load interlock) (.33) l b) C-8 (Turbine trip interlock) (.33) j c) C-9 (Condenser interlock) (.33)

L Loaic/ Condition 1

2) a) 15%/120 seconds (.33) b) 4/4 stop valves closed or (.33) 2/3 low auto stop oil pressures c) Circulating water pump breakers closed and condenser vacuum < 8 in. Hg vacuum (.33) .

3) a) Arms steam dumps in Tavg mode after (.33) load rejection b) Arms steam dumps in Tavg mode after turbine trip (.33)  ;

c) Prevents steam dump valve operation when they could cause overpressurization  !

and damage to the main condenser (.33) i

l

. , . - l

-QUESTION 4.05 (2.00)

LIST the FOUR (4) immediate operator actions of the Emergency Boration procedure FNP-1-A0P-27.0..

ANSWER 4.05 (2.00)

a. Start Boric-Acid Transfer Pump 1A or 1B (0.5)

b. Open Emergency Borate to Charging Pump Valve (05j

c. Verify ONE (1) Charging Pump running (0.5) 3

d. Verify Charging Pump. Suction Header Isolation Valves are open (0.5 1

Reference:

I Faricy LP OPS-52533A; Qual Reqm't Element "CRO-766; FRP-S.1,  !

'" Response to Nuclear Power Generation /ATWT" and l

FNP-1-A0P-27.0, " Emergency Boration" '

l KAIR 4.2/4.4 '

000024K302 ...(KA'S)

COMMENT We do not require the memorization of valve numbers, but will accept valve numbers as correct answers instead of the valve description for comonly used valves. Page 1 of FNP-1-A0P-27.0 is attached for reference.

RECOMMENDATION Change the answers to part b. and d. of answer key as follows:

b. Open emergency borate to charging pump valve ,

E open valve 8104 (.5) ,

.d) Verify charging pump suction header isolation 4 valves are open E

I Verify open 8131A, 8131B (.5) i

__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ J

1

'R0 QUESTION 4.09 (2.00)

SRO QUESTION 7.14 (2.00)

SCENARIO:

A fire exists in the Control Room (around the center desk section)

AND the Shift Supervisor decides that control of equipment from control is in jeopardy.

LIST FOUR (4) of FIVE (5) actions which should be performed - in an expeditious manner and in accordance with the " Fire in the Control Room" ,

procedure (A0P-28.2) - both PRIOR T0 and IMMEDIATELY AFTER the control room evacuation and BEFORE assemblage of personnel at the Hot Shutdown l' Panel.

ANSWER 4.09 (2.00)

[Any FOUR (4) of the following FIVE (4) actions - each worth 0.5 points]

a. Shutdown turbine as expeditiously as possible - IF the Shift  !

Supervisor deems it necessary, TRIP the turbine.  !

b. Shutdown reactor as expeditiously as possible - IF the Shift Supervisor deems it necessary, TRIP the reactor.

c. Verify reactor trip bkrs/ bypass bkrs are open at the reactor trip breaker cabinet, j

d. Verify main turbine tripped at the main turbine governor and j pedestal.

)

e. Announce (via PA system) " Fire in the Control Room; shifting control to the Hot Shutdown Panel." ,

Reference:

)

Farley Abnormal Op Procedure FNP-1-A0P-28.2, " Fire in the Control Room"; l Farley LP OPS-40204G, Objective 1 AND " Design Bases" Hot Shutdown Panel l AND A0P Qual Reg'mt (A0P-28.2) i 0 67K304 b0006 Ob 8 318 ...(KA'S) 1 COMMENT FNP requires the memorization of immediate operator actions for the  ;

Abnormal Operating Procedures. FNP does not want operators to memorize  !

subsequent actions, but to use the procedure for subsequent actions to {

ensure accuracy.

l This question clearly states the symptoms for A0P-28.2, and specifically  !

states to list actions from A0P-28.2. A0P-28.2 has no immediate i actions, and therefore this question should not have been used.  !

l l

l

• 1 A0P-28.0 does have immediate actions, but should not be used if control  ;

of equipment ~is in jeopardy from the fire as stated in the question. j RECOMMENDATION Delete question 4.09 from the R0 exam and question 7.14 from SR0' exam.

J i

, j QUESTION 4.10 (2.00)

Concerning Logging Requirements for Reactor Power Changes (FNP-84-0220) j

- Operations Memorandum 84-03:

Due to the reporting requirements of NERC-GADS and the " Performance Monitoring Report" (FNP-0-AWP-34), a means was established for reporting reactor power reductions.

a. What " magnitude" of power reduction (" change") must be reported to  !

FNP's Licensing Group?

b. What " units" are used when making this report? - i.e: i NI Indication (percent'P/R change)?

Percent of Gross MW Elec. Generated?

Percent Rated Thermal Power (RTP)?

Gross MW minus Reactive MW Generated?

c. LIST the FOUR (4) " pieces" of information that should be entered into the Reactor Operator's logbook.

ANSWER 4.10 (2.00) t

a. Power level changes of greater than or equal to (0.125)

FIVE (5) percent RTP to.... (0.125)

Power level changes of less than or equal to (0.125) ,

NINETY FIVE (95) percent RTP. (0.125) '

b. Percent Rated Thermal Power (RTP) (0.5)

c. 1) Power Level Reduced FROM (0.25)

2) Power Level Reduced T0 (0.25)

3) Time of the Reduction (0.25)

4) Reason for the Reduction (0.25)

Reference:

t Operations Memorandum 84-03 (3/3/84 AND Admin Proc. Qual Req'mts, Element CRO-399 AND AP-16, Section 6.1 l KAIR 3.4/3.4 3.6/3.5 194001A106 192008K118 ...(KA'S)

COMMENT Concerning part a), AP-16 (Conduct of Operations - Operations Group), .

step 6.1, states in part 5 that all power changes and power levels will l be logged by the Unit Operator. Operations Memorandum 84-03 states that the method for reporting power changes will be by the licensing group reviewing the Operator's log as the preferred method. Since the intent  !

of Operations Memorandum 84-03 is fully incorporated by the requirements l

of AP-16,'it is not_ reasonable to expect that a " magnitude" of power

-change requirement of 84-03 be committed to memory.

In addition, the Operations Standing Policy Book cover memorandum states-that personnel should "become familiar with the contents of.the policy book" and that the book will. remain in the Control Room for ready reference. The Operations Standing Policy. book, from which Operations Memorandum 84-03 came, was not provided to the examinees for reference.

Copies of the applicable portions of AP-16,and the Operations Standing Policy Book are attached for your reference.

RECOMMENDATION Change part a) of the answer key as follows:

a) Any magnitude of power reduction (0.5) i i

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-QUESTION 4.11 (1.50) l l

Concerning Surveillance Scheduling: I l

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NOTE: Refer to Figures 4-2 and 4-3 for the following questions.

l a.- What is the periodicity of surveillance listed on the "B-Schedule" (Figure 4-3)?

b. How are the "DUE DATES" - as listed on the "A-Schedule" (Figure  !

4-2) - determined? l i

c. How are the " GRACE PERIOD END DATES" - as listed on the l "A-Schedule" (Figure 4-2) - determined?

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ANSWER 4.11 (1.50) i

a. The "B-Schedule lists all STP's which must be performed every 31 days or more often - i.e.; any surveillance frequency greater than (0.5)

b. The "Due Date" will be the EARLIER of ... (0.1)

1) The date last performed PLUS the surveillance interval listed in the " Tech Specs" 0R.... (0.2)

2) The date previously scheduled PLUS the surveillance interval listed in the " Tech Specs". (0.2)

c. The " Grace Period End Date" is be the EARLIER of.... (0.1)

1) The date last performed PLUS the surveillance interval listed in the " Tech Specs" PLUS a " grace period (as allowed by Section 4.0.2 of the " Tech Specs") OR.... (0.2) t

2) The date performed TWO (2) cycles in the past PLUS d THREE (3) surveillance intervals as listed in the

" Tech Specs" PLUS a " grace period (as allowed by Section 4.0.2 of the " Tech Specs") OR.... (0.2)

NOTES: Reasonable Wording Accepted The statement "as allowed by Section 4.0.2 of the " Tech Specs" IS NOT required for full credit.

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Reference:

Farley Technical Specifications Section 4.0.1 through 4.0.5, FNP-0-AP-5,

" Surveillance Program Administrative Control", FNP-0-AP-52, " Equipment Status Control and Maintenance Authorization AND Farley qual Req'mt, Element CRO-378 KAIR 2.5/3.4 3.6/3.1 1 194001A103 194001A108 ...(KA'S) l 1

_ - _ _ - _ -- --- - ---------- -- -- -1

COMMENT Scheduling STPS from the A schedule and B schedule is a job performed by an SR0 in the Shift Foreman position. (See attached SR0 Qual Requirements pages SR0 11 and 7, items 7, 51, 52, 53, and 54.) The R0s never get involved in the scheduling. The reference cited of Farley

-Qual Requirement CRO-378 deals with diagnosing abnormal events and has nothing to do with STPs.

RECOMMENDATION Delete the question from the exam. '

n QUESTION 4.13 (2.00)

Use Figures 4.4 thru.4.9 (Emergency Procedure CSF Drawings FNP-1-CSF-0.1 thru CSF-0.6) for the following question:

Unit ONE has just experienced an ATWT event, however, all rods are not inserted, the turbine is tripped, and NO adverse containment conditions are experienced.

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SPDS is unavailable but fromn other indications you note that the following Critical Safety Function conditions exist-PRESSURIZER LEVEL IS =63 PERCENT AND " STEADY" WITH NO NOTICEABLE FLUCTUATION IN LEVEL CONTAINMENT SUMP LEVEL IS =THREE (3) FT. AND RAD MONITORS  !

INDICATE ~120 MREM j l " INTEGRITY" CSF DETERMINED TO BE " SAT" i NR LVL IN ALL S/G's LESS THAN FIVE PERCENT - TOTAL FDWTR FLOW TO ALL S/G'S =400 GPM - AVG PRESS =1077 PSIG NC SYSTEM SUBC00 LING IS =ZERO (0) DEGREES F AND FIFTH HOTTEST  ;

CORE EXIT TC AVG TEMP ~715 DEGREES F POWER RANGE INSTRUMENTS INDICATE THREE (3) PERCENT -

INTERMEDIATE RANGE STARTUP RATE SLIGHTLY POSITIVE

a. Which of the above conditions, require at least " PROMPT" ("0 RANGE")

actions / responses?

b. If the Power Range instruments suddenly indicated greater than FIVE (5) percent reactor power, the operator would:

1) Continue with functional restoration of the " loss of heat sink" condition THEN comence functional restoration of the

" loss of suberiticality" condition.

2) Proceed with functional restoration of the " loss of suberiticality condition" THEN continue with restoration of i the " loss of heat sink" condition.

3) Continue with functional restoration of the " loss of core  ;

cooling" condition THEN commence functional restoration of the {

" loss of subcriticality" condition.  !

4) Proceed with functional restoration of the " loss of suberiticality condition" THEN continue with restoration of the " loss of core cooling" condition.

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

J , E. ; q gy. o i

ANSWER 4.13: (2.00)  !

i L

l a) 1) POWER RANGE INSTRUMENTS INDICATE THREE (3) PERCENT l

~

INTERMEDIATE RANGE STARTUP RATE SLIGHTLY POSITIVE [ Highest Priority;'" Prompt (Orange) Action"] .(0.5)

2) NC SYSTEM SUBC00 LING IS =ZERO (0) DEGREES F - FIFTH HOTTEST  :!

CORE EXIT TC AVG TEMP ~715 DEGREES F [2nd Priority] (0.5)= ,

NOTE:. Priority of actions NOT necessary for. full credit, HOWEVER,'IF  !

i< ' core cooling is expressed as a higher priority " prompt action" deduct 0.25 points.

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" Heat Sink" is a " Yellow" CSF condition b)' 2) Proceed with functional restoration of the " loss of-subcriticality condition" THEN continue with restoration of the " loss of core cooling" condition. (1.0) i

Reference:

Farley LP OPS-52534,-Objectives 1, 2, 3 and 4; FNP-1-CSF-0, " Critical

. Safety Function Status Trees" KAIR 4.4/4.7 4.0/4.4 4.4/4.6 000029K312 000074K311 000054K304 ...(KA'S)

COMMENT l

The distractor used in the question is not plausible, and, therefore, the inconsistent critical safety function conditions are confusing. The diagnosis of critical safety function conditions is more effective when the conditions relate to the initiating event. Also the answer key to the (b) part is not correct. Initially, our priority is a suberiticality orange path and a core cooling orange path, and the only 4 possible path for heat sink is a yellow path. An event is given which escalates subcriticality to a red path. . The most correct answer would-be choice #4. This choice addressed subcriticality.first and then core cooling. The choice should be worded to " continue with functional restoration of the " loss of subcriticality condition", then commence functional restoration of the " loss of core cooling" condition.

RECOMMENDATION Change the answer key to the (b) part to read: i

4) Proceed with functional restoration of the " loss of suberiticality condition", then continue with restoration of the " loss of core cooling" condition (1.0) l 1

J

QUESTION 4.15 (1.50)

LIST SIX (6) " parts /" sections" which makeup Farley's Technical

-Specifications.

ANSWER 4.15 (1.50)

[ANY SIX (5) of the following - worth 0.25 pts each]

l 1)- Definitions 6) Design Features i

2) Limiting SEfety System 7) Admin' Controls Settings (Safety Limits) l

3) Limiting Conditions for 8) Appendix B - Environmental Operation (LCOs) Protection Plan i 4). Surveillance Requirements- 9) Tech Spec Interpretations

5) Bases - Sections 3.0/4.0 10) Bases - Section 2.0 ,

1 I

Refererence: l Farley Technical Specification LCO & BAsdes 2.1.1 and 2.1.2 AND Administrative Controls 6.7.1 KAIR 2.6/3.1 2.9/4.0 2.9/4.,1 194001A108. 006000G006 012000G006 ...(KA'S)

COMMENT The answer does not reflect the vague nature of the question. FNP Technical Specifications has numerous " parts"/" sections". The index of FNP Unit I Technical Specifications is included for reference.

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RECOMMENDATION Change the answer to the following: (Any six [6] of the following -

worth 0.25 points each)

1) Definitions

2) Safety Limits

3) Limiting Safety System Settings

4) Safety Limits (Bases)

5) Limiting Safety System Settings (Bases)

6) Limiting. Condition for Operation l

7) Surveillance Requirements l

8) Reactivity Control Systems

9) Power Distribution Limits

10) Instrumentation

11) Reactor Coolant Systems

12) Emergency Core Cooling Systems

13) Containment Systems

14) Plant Systems

15) Electrical Power Systems

16) Refueling Operations

17) Special Test Exceptions

18) Radioactive Effluents

19) Radiological Environmental Monitoring

20) Bases - LCOs and Surveillance Requirements

21) Design Features

22) Administrative Control

23) Index

SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION 5.03 (2.00)

At 0001 on November 1, 1988 a reactor shut down from 100 percent power was initiated. Over the next 3 days the reactor was shutdown and cooled down to 140 degrees F. During the cooldown, boron concentration was increased by 100 ppm. Given the following absolute values of reactivity change over the 3 days:

Xenon = [ ] 2575 pcm Rods = [ ] 6938 pcm Temperature = [ ] 550 pcm Boron = [ ] 1140 pcm Power Defect = [ ] 1525 pcm A. What was the shutdown margin at 0001 on November 1, 1988? [ SELECT i THE CORRECT VALUE] )

a. - 5413 pcm

b. - 5963 pcm

c. - 6938 pcm

d. - 7913 pcm B. What is the shutdown margin after all the reactivities were added?

(SELECT THE CORRECT VALUE]

a. - 3428 pcm

b. - 4863 pcm l

c. - 6478 pcm

d. - 9053 pcm ANSWER 5.03 (2.00)

A. a. [- 5413 pcm]

B. a. [- 3428 pcm]

[1.0 each]

Reference:

NUS, Reactor Operation, Unit 11, Reactor Core Characteristics, Para 11.1. Pages 11.1-1 thru 11.1-6 3.8/3.9 192002K114 ...(KA'S) h I

.1 .

- t 8 .,

COMMENT l

Shutdown margin, as defined in FNP Technical Specifications (definition attached), includes an allowance for most reactive rod. Using a -l ballpark value of 1000 pcm for the most reactive rod, the correct answer 'l to part a. would be -4413 pcm and part b. would be -2428 pcm.

The answers provided were shutdown reactivity without'an allowance for the most reactive rod. Technically, there was no. correct answer.

Answer a. is the answer closest to the correct answer for both part a.

and b. j RECOMMENDATION Accept answer a. for part A. and B., but also consider other answers if calculation is shown.

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_ - _ _ _ - _ _ _ _ - _ _ ._--__-_--_-____-_________b

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f QUESTION 5.07 (1.00)

A reactor startup is in' progress. The source range count rate indication was 11 cps when the startup conynenced and now indicates 2386 cps. What'is-the expected count rate response resulting-from a brief rod. withdrawal during the approach to criticality? [ SELECT THE CORRECT' COUNT. RATE RESPONSE].

A. - An intnediate rapid rise continuing to criticality

-B. An immediate rapid rise followed by a gradual increase to a higher steady state value j C. A gradual' increase followed by a rapid decrease when rod withdrawal is stopped D. A gradual increase continuing to criticality ANSWER 5.07 -(1.00)

8. [An immediate rapid rise followed by a gradual increase to a higher  !

steady state value] l

Reference:

NUS, Reactor Operation, Unit 12, Fuel Loading and Startup, Para 12.4.

Pages 12.4-1 thru 12.4-4 3.9/4.0 129008K103 ...(KA'S)

COMMENT The question states that source range count rate indication increased i from 11 cps to 2386 cps. Criticality will normally be achieved after 5 to 7 doublings of count rate. (See attached from FNP Startup Certification Training' lesson plan, OPS-31307.) If the examinee determines that the reactor is already critical or critical after the brief rod withdrawal stated in the question, then no correct answer exists for selection. This is a possibility'since count rate after 7 3 doublings would be 1408 cps and the reactor power level is approaching the eighth doubling at 2816 cps.

RECOMMENDATION j Request grader discretion based on the examinee's assumption. If the assumption is made that the reactor is not yet critical and does not reach criticality during the brief rod withdrawal, then the key answer B. is correct. If the assumption is made that the reactor is critical or reaches critical, then full credit should be awarded for recognizing that no correct. answer exists.

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_ - _ - _ _ _ _ - - - - - _ _ _ _ J

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' QUESTION 5.20 (2.00)

Unit 1 is operating at 30 percent power with rod control in manual. For each of the following parameters, state whether an inadvertent closure of S/G A main steam isolation valve would cause the parameter to INCREASE, DECREASE or REMAIN THE SAME. Consider each case separately.

Assume power is maintained at 30 percent, j A. Departure from Nucleate Boiling Ratio (DNBR).

I B. Reactor Coolant System Subcooling Margin (RCS SCM).

C. RCS Loop B Hot Leg Temperature.

D. Quadrant Power Tilt Ratio (QPTR).

ANSWER 5.20 (2.00)

A. DECREASE.

B. DECREASE.

C. INCREASE.

D. INCREASE,.

[0.5 each]

Reference:

TPNP, REV 1. ILP 0057-OL, App I, First E0; TPNP, Thermo. .

NUS, Plant Performance, Para 8.2, Pages 8.2-1 thru 8.2-4 '

3.4/3.6 3.6/3.8 193008K105 193008K115 ...(KA'S)

COMMENT For part D. of this question, the answer key assumes uneven mixing of the water returning to the core, causing power to shift from one quadrant to another. This could cause QPTR to increase. If 100'/. mixing is assumed, then it would have no effect on QPTR at all. QPTR is not measured until power is at least 50% (T.S. enclosed). At 30% power, any QPTR measured would be of questionable quality.

RECOMMENDATION Delete part D. from exam.

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.._--_ __ _ _____- _ _________ ___-___ _ __ _ ________ _ _ - A

c. . ., !j

.. w b :,

I //I . QUESTION,6.10 (1.80). j Refer to the attached drawing of the Reactor Coolant System (RCS) to j

, identify the following RCS penetrations: [ IDENTIFY /NAME;0R STATE THE-D PURPOSE OF PENETRATION AND INDICATE T0/FROM AS APPROPRIATE]

A. - Penetration number 3?

B.- Penetration number 47 i C. Penetration' number 107

.D. . Penetration number 11?

E. Penetration number 12? i F. Penetration number 14?

ANSWER 6.10 (1.80) a

.A. From safety injection system accumulator tank?

B. CVCS alternate charging line?

C. From high head safety injection, residual heat removal pumps i D. From baron injection. tank high head safety injection and residual '

heat removal pumps E. To residual heat removal pump F. From high head safety injection pump

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[0.3 EACH]

Reference:

Reactor Coolant, OPS-52101A (OLT), OPERATOR LICENSE TRAINING OBJECTIVE 3, Figure 12 3.7/4.0 4.5/4.6 002000K106 002000K108 ...(KA'S) -l J

l COMMENT '

The RCS drawing provided was never meant to be used as an identification j tool. .With this drawing completely unlabeled, it requires the  ;

memorization of penetration sizes and orientation in the loops, which is  ;

beyond the scope of the objective references. '

1 The drawing provided also had an error which could have led to some  !

confusion. Number 14 on the B cold leg should be labeled number 11. i The B cold leg penetration, mislabeled #14, is a penetration for both  ;

high head safety injection and residual heat removal. This drawing i i

(. ' . , ' i o

l1 ' error, therefore, led to confusion for parts D. and F. since:they addressed penetration numbers 14 and 11.

l RECOMMENDATION l l

Delete parts D. and F. from the question and change the point value for the remaining parts to 0.45 each.to allow the total value of the question to remain the'same.

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QUESTION 6.11 (2.00) {

Refer to the attached Pressurizer Pressure Protection And Control b.

diagram to answer the following if the master pressure controller fails .

l7- at 43% [10.32 ma] while selected to automatic: [ ANSWER TRUE OR FALSC]  !

[ ASSUME NO OPERATOR ACIT0N] 4 i

A. Variable heater group C . voltage will decrease to cause pressure to l decrease.  !

B. Spray Valve PCV 444C wil open to maintain pressure within the  !

operating band. j t

C. PORV 444B will open when pressure increases to setpoint.

D. PORY 445A will open when pressure increases to setpoint. 1 f

ANSWER 6.11 (2.00) l i

A. False '

B. False i l

C. False D. True i

[0.5 each]

Reference:

Pressurizer Pressure And Level Control, OPS-52201H (0LT), OPERTOR 1 LICENSE TRAINING OBJECTIVE 9., Pages 6 thru 12, l 3.6/3.7 3.9/4.1 l 010000K103 010000K101 ...(KA'S)  !

COMMENT  !

Memorization of the master pressure controller output setpoint is not ,

required at FNP. Since no curve (such as the attached Figure 4) was i provided for reference, the examinee was required to make an assumption  ;

concerning the failed output of 43% as stated in the question. If 43% i was assumed to be a controller failure low, then the following answers ,

would be correct: I A. False B. False ,

1 C. False j D. True 1

1

If 43% was assumed.to be a controller failure to the reference pressure setpoint, then.the following answer would be correct:-  ;

o -A. False '

B. False i

C. False D. True If 43% was assumed to b5 a controller failure high, then the following answers would be correct:

A.. _True-B. True (if failure high assumed to,be greater than 2260 psig '

C. True (if failure high assumed to be greater than 2335 psig) 1 D. True Note that the answer to question D. is always true based on the fact that PORV-445A is not affected by a failure of the master pressure controller. If pressure increases to the PORV setpoint of 2335 psig,-

then PORV-445A will open.

i RECOMMENDATION-

'l Request grader discretion based on the examinee's assumption of the failure-to 43% controller output.

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1 QUESTION 7.10 (3.00) i EIPs-12, 17, 18, & 19 ARE ATTACHED: i i

A. Refueling operations are in progress. A fuel assembly has been fully withdrawn from the core. When the operator starts to move I

the manipulator crane towards the upender,.the manipulator crane fails. The operators can neither lower the fuel assembly back into the core nor move it to the refueling canal. Immediately after the manipulator crane failure the refueling cavity water level rapidly decreases to the top of the core.

What is the emergency classification [ NOTIFICATION OF UNUSUAL EVENT, ALERT, SITE AREA EMERGENCY OR GENERAL EMERGENCY] for the given plant conditions?

8. At 0925 a private aircraft crashed onsite. Approximately 10 4 seconds after the crash a loss of off site power occurs. ECCS actuation and high secondary collant activity occurs.

What is the emergency classification [ NOTIFICATION OF UNUSUAL EVENT,-ALERT, SITE AREA EMERGENCY OR GENERAL EMERGENCY] for the given plant conditions?

C. At 0924 power was removed from all main control board annunciators to isolate a ground. At 0925 a private aircraft crashed onsite.

Approximately 10 seconds after the crash a loss of off site power occurs. The operators properly execute the Emergency Event Procedures and Emergency Contingency Procedures. At 0935 the aircraft pilot has been removed _from the aircraft. At 0945 main  ;

control board annunciators are restored. At 0946 the injured pilot '

is transported to the hospital by ambulance.

What is the emergency classification [ NOTIFICATION OF UNUSUAL EVENT, ALERT, SITE AREA EMERGENCY OR GENERAL EMERGENCY] for the given plant conditions?

ANSWER 7.10 (3.00)

A. General Emergency [GE]

B. Site Area Emergency [SAE]

C. Site Area Emergency [SAE]

[1.0 each]

Reference:

Emergency Plan Implementing Procedures, OPS-53002, Objective 28, FNP-0-EIP-19, GENERAL EMERGENCY, Para 3.2.4, Page 3, FNP-0-EIP-18, SITE AREA EMERGENCY, Para 3.2.3 & 3.2.10, Pages 2 & 3 3.1/4.4 194001A116 ...(KA'S)

_ _ _ _ - - _ - - - - - - - _ . _ - - - _ - _ - - 1

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COMMENT-Part A. of this question is not directly referenced by any of the EIPs. l The key answer assumes the examinee will go to EIP-19, page 3, item '

3.2.4. If the assumption is made that the upender can be moved to the  ;

spent fuel pool and the gate valve shut, then containment integrity will  !

be satisfied and no gross release of radioactivity will occur. The j hanging fuel assembly has not yet overheated, so at the present time, j there are still two boundaries. .A General Emergency should not be J declared unless there is a real need for it due to the effect on the

-surrounding population.

A0P-30 (copy attached) addresses fuel handling accidents and it l reconinends declaring an Alert. If the examinee assumes a high off-site {

dose will occur but not a core melt, he could use EIP-18, Item 3.2.8 to declare'a Site Area Emergency.

Part B. of this question does not provide sufficient information to accurately make a declaration. If the examinee assumes R-15 or R-19 go off-scale high, then paragraph 3.2.3 of EIP-18 would require a S.A.E.

If crash is assumed to affect vital structures by fire or impact, then 3.2.13 of EIP-18 requires S.A'E. If these assumptions are not made, 3

then item 3.2.19 of EIP-12 would require an Alert declaration.

l RECOMMENDATION l

For part A., accept General Emergency, Site Area Emergency, or Alert i based on assumptions.

For part B., accept Site Area Emergency or Alert based on assumptions.

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1 QUESTION 7.11 (3.00)

A reactor _ trip has occurred and the operators are performing i FNP-1-EEP-0, Reactor Trip or Safety Injection. The operators have l progressed- to intnediate action Step 4. Check-if SI actuated. '

[SEE ATTACHED PAGE FROM FNP-1-EEP-0]

A. - How do the operators perform Action / Expected Response Step 4.1 "Any ~

SI actuated indication"? [ LIST TWO DIFFERENT INDICATIONS] [0.9]

B. How do the operators perform Response NOT obtained Step 4.1.1,

" Trip status light box bistables meet coincidence for SI", and Step i 4.1.2, " Parameter indicators have reached an SI setpoint"? [ LIST . 1 FIVE (5) INCLUDING SETPOINTS AND COINCIDENCES AS APPROPRIATE] [2.1].

i ANSWER 7.11 (3.00)

A. SAFETY INJECTION ACTUATED window [0.3] on 1

BYP & PERMISSIVE panel lit [0.3]

MLBI 9-1 or 19-1 lit [0.3]  ;

B. Pressurizer pressure low [0.15] 1850 psig [0.15] 2/3 [0.15]

Steam Line Differential pressure [0.15] 100 psid [0.15] 1 steam line 100 psig less than other two on 2/3 protection sets [0.15]

Low Steam Line pressure [0.15] 585 psig [0.15] 2/3 [0.15] '

Containment Pressure High [0.15] 4 psig [0.15] 2/3 [0.15]

i Manual [0.15] 1/2 [0.15] l f

Reference:

FNP-1-EEP-0, Reactor Trip or Safety Injection, Pages 4/68 and 6/68  !

4.2/4.1- 4.1/4.3 000007G010 '000007G011 ...(KA'S)

COMMENT A part of the answer given for the "B" part is manual, 1/2. This may or may not be a response since the Response Not Obtained Steps 4.1.1 and 4,,1.2 address the trip status light box bistables and the parameter indicators. There are no bistables or indicators requiring manual.

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RECOMMENDATION Delete manual 1/2 as an answer to the B. part since there are no bistables or indicators associated.

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00ESTION 7.12 (3.00).

A... As the Reactor. Operator,'what action would you take in accordance

.with~AOP-19.0, Malfunction of Rod Control System, if.an unexplained outward rod motion resulting in a significant positive reactivity

, . addition' occurred? [ LIST =TW0' ACTIONS] [1.0]

B.- .Given the following indications:

One [1] Rod Bottom Light.

Rod at bottom alarm Sudden drop in Tavg

1. .What are the automatic actions that-should occur? [ LIST TWO]'

[ ASSUME AUTO R0D CONTROL] [1.0]

2. What imediate operator actions should be taken in accordance with FNP-1-A0P-19.0, MALFUNCITON OF R0D CONTROL. SYSTEM? [ LIST TWO ACTIONS] [ ASSUME AUTO ROD CONTROL] [1.0]

~ ANSWER 7.12'. (3.00)

A. -1. Attempt to stop rod motion [0.25] by placing the Bank Selector switch in AUTO or MANUAL [0.25]

2. If rod motion does not stop [0.25]~, THEN trip the reactor

[0.25]

B. 1. a. Power range high negative flux rate trip [0.25]-

may occur [0.25]

b. Rods will. step out in AUTO [0.25] due to Tavg - Tref mismatch [0.25]

2. a. Place' rod control [0.25] in MANUAL [0.25]

b. Place turbine [0.25] on HOLD [0.25]'

Reference:

1 FNP-1-A0P-19.0,. MALFUNCTION OF R0D CONTROL SYSTEM, Para 2.3, Page 1, Para 5.2 & 5.3, Page 3 4.4/4.6 3.6/3.8 3.9/3.8 3.9/4.0 000001A205 000003G011 000003G010 000001G010~ ...(KA'S)

COMMENT l

The answer B.1.a. may not be received because the question states what automatic actions should occur for one (1) dropped rod. A power range high negative flux rate trip is not expected based on the attached j references from FNP-1-A0P-19.0, Section 5.0, FNP-FSAR-7 page 7.2-4, and  ;

FNP-FSAR-15, pages 15.2-11 through 15.2-12. These references address a '

power range high negative flux rate trip from multiple dropped rods.  ;

_ _ _ - _ _ __-_________A

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• ; , _ * =' .._:.., ; ,

y .-c-.

RECOMMENDATION -

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I, . Delete the answer B.1.a. because thisLis not an expected' action from a L single dropped rod.

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.. . 1 QUESTION 8.03 (1.00)

Which of the following statements is correct regarding manual operation of MOVs? [ SELECT THE CORRECT STATEMENT]

A. DO NOT hold the declutch lever in the depressed position while the motor is running. This prevents motor overspeed damage.

B. DO NOT at any time depress the declutch lever. This prevents damage to the clutch _ internals.

C. DO NOT use the manual operator to force the valve any further against its seat than the motor operator will drive it. The -

motor may not be able to drive the valve off the seat without

' damaging the operator.

D. DO NOT use the manual operator to force the valve any further  !

against its seat than the motor operator will drive it. The' l seat may be damaged by the additional mechanical advantage of j the mechanical operator. '

ANSWER 8.03 (1.00)

C. [D0 NOT use the manual operatoro t' force the valve any further against its seat than the motor operator will drive it. The motor may not be able to drive the valve off the seat without damaging the operator]

Reference:

FNP-0-SOP-0, GENERAL INSTRUCTIONS TO OPERATIONS PERSONNEL, Para 9.0, Pages 8 & 9- 3.3/3.7 191001K106 ...(KA'S)

COMMENT The answer given in the key only addresses potential damage to the operator of an MOV from using the manual operator. FNP-0-SOP-0, General Instructions to Operations Personnel, Section 8.0 addresses the possibility of damaging valve internals (i.e., valve seat) from the use of added mechanical advantage to assist in operating a valve.

RECOMMENDATION  ;

1 Accept answer C. or D. as .a correct answer since operator damage and valve seat damage can potentially occur by manually forcing an MOV l against its seat.

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QUESTION 8.04 (1.00)

How would you determine the use/ purpose of a key reported as found in the turbine building and numbered DE-243. [ STATE THE PROCEDURE YOU WOULD CONSULTJ]

ANSWER 8.04 (1.00)

FNP-0-SOP-0 [ GENERAL INSTRUCTIONS TO OPERATIONS PERSONNEL]

Reference:

FNP-0-SOP-0, GENERAL INSTRUCTIONS TO OPERATIONS PERSONNEL, Table 2 3.1/3.4 194001K105 ...(KA'S)

COMMENT The answer states that FNP-0-SOP-0, Table 2, should be referenced to determine the use/ purpose of a key reported as found in the turbine building. Multiple documents exist for reference in determination of the use/ purpose of the key. Additional references as follows are attached:

Locked Valve Key Checkout Sheets STP-64.0, Safeguard System Locked Valve Verification STP-64.1, Non-Safeguard System Locked Valve Verification RECOMMENDATION Request that you add the following references to the present answer and accept any of the four for full credit:

!ocxed Valve Key Checkout Sheets STP-64.0, Safeguard System Locked Valve Verification STP-64.1, Non-Safeguard System Locked Valve Verification

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QUESTION 8.14

(2.50) 1 1

FARLEY - UNIT 1 Technical-Specification 3.2.1, Axial Flux Difference is attached.

A. . -If the AXIAL FLUX DIFFERENCE Monitor. Alarm-is inoperable, how often is AXIAL FLUX DIFFERENCE monitored and logged?

B. How long may operation continue at 45% thermal power with AXIAL FLUX DIFFERENCE indicating +5% before 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of penalty deviation accumulates? [ ASSUME INITIAL PENALTY DEVIATION IS 30 MINUTES]

C. How.long may operation continue at 55% thermal power with AXIAL FLUX DIFFERENCE indicating -5%? [ ASSUME INITIAL PENALTY DEVIATICN IS 30 MINUTES]

D. If AXIAL FLUX DIFFERENCE IS -20% at 75% thermal power, what action should be taken to continue power operation? [ ASSUME INITIAL PENALTY DEVIATION IS 30 MINUTES] [1.0]

ANSWER 8.14 (2.50)

A. Once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> [0.25] and at least once per 30 minutes thereafter [0.25]

B. One (1) hour [60 mninutes] [0.5]  !

i C. Indefinitely [0.5]

D. . Reduce thermal power to < 50% [0.2] within 30 minutes [0.2] and .

reduce the Power Range Neutron Flux-High Trip Setpoints [0.2] to '

less than or equal'to 55% of rated thermal power [0.2] within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> [0.2]

[ Technical Specification 3.2.1 action a.2.a)2) accepted for full credit) 1

Reference:

.I Technical Specification 3.2.1, AXIAL FLUX OIFFERENCE 3.7/4.1 1 001000G005 ...(KA'S) -l L

COMMENT l

The 8. part has an answer of one (1) hour of power operation allowed.

The applicability of axial flux difference, Tech Spec 3.2.1, is for mode 1 above 50%' rated thennal power; therefore, power operation may continue ,

indefinitely in thia condition (regardless of initial penalty  !

i deviation). '

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i C. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____ ______ _ __ ._

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5 In the D. part, there are two possible answers which address the actions that could be taken to continue power operation. AFD can be returned to within the acceptable operation region of Figure 3.2-13 and then operation can continue for 30 more minutes (based on 4.2.1.2.a and a.2.a.1 and an initial penalty deviation of 30 minutes). If this action cannot be taken, then power must be reduced to less than 50% within 30 minutes (based on a.2.a.1).

RECOMMENDATION j Change the answer key on the B. part to read indefinitely (0.5) since the AFD limit is not applicable less than 50% power.

l Add the following to the answer key D. part as an acceptable action:

Return AFD to within Figure 3.2-1 limits immediately.

2 QUESTION 8.16 (3.00) c . UNIT 1 IS AT 93'/. LOAD The following equipment is out of service:

Charging Pump A Component Cooling Water Pump A Residual Heat Removal Pump A Start up auxiliary transformer 1A Diesel Generator 1-2A l Component Cooling Water Pump B key interlock [ pump aligned to Bus j 1F] ,

As the Shift Supervisor, what action would you take?

f j ANSWER 8.16 (3.00) t within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> [0.5] initiate action to be in [0.25]

at least HOT STANDBY [0.5] within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> [0.25] l at least HOT SHUTDOWN [0.5] within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> [0.25],

and at least COLD SHUTDOWN [0.5] within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> [0.25]

Reference:

FARLEY - UNIT 1 Technical Specification 3.0.3 3.5/4.2 006000G005 ...(KA'S)

COMMENT The conditions of the question result in invocation of the Applicability Action Statement 3.0.3, which is attached for reference. Since this Technical Specification was not provided for examinee reference, grader discretion should be allowed.

RECOMMENDATION Request that full credit be given for the following 3.0.3 response:

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, initiate action to be in at least HOT STANDRY within the next S hours.

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ENCLOSURE 4 D

g .. SIMULATION FACILITY FIDELITY REPORT i

Facility Licensee: Alabama Power Company Facility Licensee Docket No.: 50-348 and 50-364 Facility Licensee No.: NPF-2 and NPF-8 Operating Tests administered at: J. M. Nuclear Plant Operating Tests Given On: November 15-17, 1988 During the conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies were observed:

-The DEH system model crashed 4 times.  !

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