ML20215N179

Forwards Util Comments on Reactor Operator & Senior Reactor Operator Written Exams Given on 860714.Questions Appearing on Both Exams Addressed on Reactor Operator Exam Comments Only

ML20215N179
Person / Time
Site:	Farley
Issue date:	07/18/1986
From:	Woodard J ALABAMA POWER CO.
To:	Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20215N166	List: ML20215N165 ML20215N174 ML20215N179
References
FNP-86-0348-TRN, FNP-86-348-TRN, NUDOCS 8611040404
Download: ML20215N179 (39)

Text

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~ ENCLOSURE 3

.I,

-Q Alabama Power Company J. M. Farley Nuclear Plant P. O. Drawer 470 Ashford, Alabama 36312 Telephone 205 899-5156 AlabamaPower me nem ecc us FNP-86-0348-TRN July 18, 1986 The Regional Administrator, Region II Nucler Regulatory Commission 101 Marietta St.,

N.W.

Atlanta, GA 30303 ATTN:

Mr. John Munro, Acting Chief Enclosed are Alabama Power Company's comments concerning the written examinations for reactor operator and senior reactor operator written examinations given at Farley Nuclear Plant on July 14, 1986.

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.

Sincerely,

[k

[(

^^1_

J. DL Woodard eral Manager - Nuclear Plant JDW/LSW/RBW:mgr Enclosures 8611040404 e61029 PDR ADOCK 05000348 V

PDR

]

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REACTOR OPERATOR WRITTEN EXAMINATION 1.05/5.06 QUESTION Assume one RCP trips at 30% power, without a reactor i

protection system actuation or a change in turbine load.

Indicate whether the following parameters will INCREASE, DECREASE, or REMAIN THE SAME.

a.

Flow in the operating reactor coolant loops.

b.

The ratio of core flow compared the total loop flow c.

Reactor vessel delta pressure d.

Core delta temperature e.

Operating loop steam generator temperatures ANSWER 4

a.

Increase b.

Decrease c.

Decrease d.

Increase e.

Decrease j

REFERENCE General Physics HT&FF, Part B, Chapter 1, pp 324-332.

i COMMENT The proctor define INCREASED, DECREASED, or REMAINED THE SAME as being "after the transient".

For Part 8, the proctor defined " total loop flow" as all three loops.

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..____..m.,

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1.09/5.10 QUESTION a.

If during a cooldown on natural circulation the PCS pressure was 1200 psig, what would be the maximum steam generator pressure to assure ade-quate subcooling, b.

During natural circulation cooldown, a steam bubble may form in the reactor vessel head area.

What is the primary indication of this bubble formation?

c.

What is the maximum core Delta Temp. which would be indicative of PROPER natural recirculation flow following a full power trip AND what is the approximate loop transient time?

ANSWER a.

Test for 1200 psig is 567F (from steam tables) 567 - 50 = 517F (subcoolina of 50F)

Psat for 517F is about 800 psig b.

Erratic pressurizer level indication c.

65F (+/- 3F) 10 minutes REFERENCE FNP E0P 7-1, pp. 12 & E0P 7-2 pp. 8 COMMENT For part A the proctor qualified " adequate sub-ccoling" as being "per procedure".

Per ESP-0.2, Nat-ural Circulation Cooldown To Prevent Reactor Vessel Head Voiding, there are several limits.

Step 7.3 ad-dresses a 78 F and 226 F limit depending on the availability of the CRDM fans.

Also figures two (2) and three (3) of the same procedure annotates 50 F and 200*F as being acceptable based on status of the CRDM fans.

This question deals with the candidates ability to use steam tables and an-assumed subcooling require-ment.

As you can see there are varied subcooling requirements within the procedure.

Therefore, the student should not be penalized for his selection of a value for subcooling, but tested on his ability to relate this number to steam generator pressure.

The reference for this question is E0P-7.1 and 7.2.

These procedures were replaced with the adoption of Emergency Response Guidelines in Mid 1985.

Refer-ences ESP-0.2 and ESP-0.4 should be used.

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'O 1.11/5.12 QUESTION During 100% power operation, it is decided to reduce power by 20% using control rods only for reactivity control.

a.

Explain HOW and WHY the axial flux shape will change for the first hour after the power reduction.

b.

Explain HOW and WHY the flux shape will change over the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Include the' effects of control rod movement to maintain power stable.

ANSWER a.

Flux will be depressed toward the bottom of the core [0.5] due to:

(1)

Lower control rod level [0.25] and (2)

Xenon buildup in top of core [0.25]

b.

Flux continues to be depressed more and more towards the bottom as Xenon builds in top, then reverses as it decays off [0.75].

Control rod movement to compensate for Xenon changes reduces the flux shift [0.75]

COMMENT The proctor told the students to " assume control rods fully withdrawn as an initial condition" for question 1.11.and " additional rod motion only to remain at 80%, if required" for 5.12b.

a ----

i 1.12/5.01 QUESTION i

)

4 Briefly EXPLAIN how the addition ofo0.5% positive reactivity to a subcritical reactor would affect the following:

(No calculations are required) a.

THE.CH4NGE IN THE COUNT RATE:

(if the reactor was slightly subcritical [ shutdown margin = 1%]

as compared to greatly subcritical [ shutdown margin = 5%]

b.

THE TIME TO REACH A STABLE COUNT RATE:

(for the different shutdown margin conditions in (a) above)

ANSWER a.

The slightly (greatly) subcritical reactor will have a larger (smaller) increase in count rate.

b.

The slightly (greatly) subcritical reactor will take a longer (shorter) time to reach a stable count rate.

REFERENCE I

FNP Training Reactor Theory Manual, pp. H-4-20, H-4-21.

i COMMENT The proctor further defined." shutdown margin" as amount subcritical.

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1.14/5.14 QUESTION Explain how the starting of a Reactor Coolant Pump in a water-solid plant can cause a pressure transient.

ANSWER The idle RCP can develop temperatures in the seal area that are less than steam generator temperatures

[0.5].

When the cold slug goes through the steam generators it picks up heat and expands.

The thermal expansion in a solid plant causes a pressure increase

[0.5].

REFERENCE A0P-24 COMMENT Answer should also recognize that the core could also be referenced as the heat source.

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1 2.15/6.12 QUESTION under what plant conditions is the reactor. vessel.

level indication system be available for the operator's use.

ANSWER Available in normal and abnormal conditions in all modes except refueling.

REFERENCE FNP, License Retrng, ICCMS COMMENT This question was taken from the 1986 License Retraining Program materials.

The-Inadequate Core Cooling Monitoring System (ICCMS) was included in retraining anticipating the system being operable by j

May of this year on Unit #2.

The system was only partially installed and therefore is not being used.

i The candidates for this exam did receive some information on ICCMS but their training was greatly downscale once it was determined the system would not be operable for some time.

They will; however, have i

training as part of the 1987 Retraining Program prior to the ICCMS being made operable.

Considering these circumstances, this question about the ICCMS should be deleted.

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2.18 QUESTION a.

Which component (s) of the Control Rod Drive Mechanism act as the pressure boundary between the RCS and the Containment atmosphere?

b.

When paralleling the output of the two Rod Drive MG sets automatically, the " speed matcher" automatically changes the speed on which of the two MG sets?

(incoming / running) c.

Explain why the stationary coils for the CRDM are supplied with TWO DC voltages.

ANSWER a.

Latch ~ housing and rod tra' vel housing [0.25 ea]

b.

Incoming [0.5]

c.

The voltage on the coils is reduced to prevent overheating of the stationary coils which could cause damage to the insulation.

[0.1]

~

REFERENCE FNP, LP:

RCS; RDCNTRL i

COMMENT The two (2) components referenced in the key for part A of this question, as the pressure boundary associated with the Control Rod Drive Mechanism are commonly referred to as the lead screw or CRDM housing.

Since the question asks for component or components, either of these commonly used terms should also be acceptable.

l B part of this question refers to a " speed matcher"

~

associated with Rod Drive MG sets.

At FNP our motor generator sets utilize an induction motor and does not have capability of adjusting speed.

This question shculd be deleted.

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2.22 QUESTION Concerning HJTC,~ state two pur' poses of the heater j-power controller logic.

j ANSWER a.

Provides a sufficient delta T signal to give an indication of voiding when the heater junction is uncovered.

b.

Reduces heater power when the heated junction is i

uncovered to prevent heater burnout.

REFERENCE FNP, License Retrng i

l COMMENT l

Same as comment for Question 2.15/6.12 I

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l 3.07 QUESTION With the reactor at 100% power and the steam dump 1

l control systent in the Tavg mode, a.15% step loss of i

load occurs.

Assuming no reactor trip occurs the condenser is available, and the reactor operator manually OPERATES the control rods, which of the following would occur if Bank 1 steam dump valves failed to open?

l a.

Bank 2 would open.

b.

Atmospheric dumps would open.

c.

S/G safeties would open.

d.

No other steam valves would open.

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-ANSWER 1

l d

REFERENCE i

FNP, LP:

Steam Dump Cont. Sys.

COMMENT The answer given assumes the operator' controls Tavg/ Tref within a 8 F mismatch.

At BOL, with a.

small moderator temperature coefficient, the 8 F assumption ~would probably be exceeded and answer A would be correct.'

The assumption of this question is not stated clearly and the question should be deleted.

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3.08/6.08 QUESTION Indicate what happens to the Rod' Control System (rods in, rods out, no change) and BRIEFLY explain why the change will or will not occur for the following conditions.

Rods are in auto unless otherwise specified.

i a.

Reactor power is 17% when the controlling turbine first stage impulse pressure transmitter fails high.

I b.

Reactor power is 100% and Loop 1 Thot fails i

high.

c.

N-44 is 50%, Rod Control is in manual.

Instrument testing is in progress on the turbine power input to rod control which has turbine power at 100%.

All indications have been stable for the last hour.

The Bank Selector switch is then placed in AUTO.

ANSWER a.

Rods out [0.25] Tref will be max so Tave/ Tref mismatch and NI/ Turbine power mismatch will both give a rods out signal.

[0.75]

b.

Rods in [0.25] Loop 1 Tave increases and auctioneered high Tave also increases.

1 Tave/ Tref mismatch gives a rods in signal.

[0.75]

l c.

Rods out [0.25] the power mismatch circuit of l

the reactor coolant unit responds only to rate

{

of change of deviation between turbine and nuclear power but rod motion will occur due to j

the Tave - Tref difference.

[0.75]

i REFERENCE J

FNP, LP:

Rod Cont. Sys.

I COMMENT Proctor changed "auctioneered high nuclear power" to "N-44".

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o 3.10/6.11 QUESTIDM For each of the following, give the set point and coincidence:

1.

P-12 2.

P-14 3.

High Main Steam Flow with' Low Low Tavg MSIV ANSWER 1.

543 deg F 2/3 < S.P.

i Lit < S.P. permission to Block SI

)

2.

75%

2/3 on 1/3 SG's > S.P.

3.

Variable:

1/2 > = S.P. in 2/3 Loops 40% for P<20%

Increases linearly 40 - 100% for P>20%

Tavg < P-12

[0.5] each st..pt.:

[0.5] each coinc.

1 REFERENCE 4

FNP, License Retrang, SG Protect.

COMMENT For answer No. 1 " Lit < S.P. permission to block SI" is not a setpoint or coincidence but a function of P-12.

The question did not ask for functions.

This

{

statement should be removed from the key.

Answer j

no. 3 states the steam flow setpoint " increases linearly 40 - 100% for.P > 20%".

It should read j

increase linearly 40 - 110% from 20% to 100% power.

Reference EEP-0 pg 12/62' Step 14.

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l 3.12 QUESTION List the following:

a.

Four protective functions provided by the pressurizer pressure PROTECTION channels (PT-455, -456, -457) other than MCB indications and alarms.

b.

Three control functions provided by pressurizer pressure CONTROL channel PT-444, other than MCB indications and alarms, c.

Two protection / control functions provided by the pressurizer level control and protection channels (LT-459, -460, -461) other than charging flow control and MCB indications and alarms.

ANSWER a.

1.

PZR Lo press, trip i

2.

PZR Lo press, SI 3.

PZR Hi press, trip 4.

P-11 or SI block perm. & PORV interlock (4 req'd [0.33] each) b.

1.

PZR heater group control 2.

PZR spray valve control 3.

PORV control (3 req'd [0.33] each) 2 c.

1.

Backup heater control 2.

CVCS and heater -interlock (2 req'd [0.33] each)

REFERENCE FNP, LP:

PZR Press. and Level Cont. Sys.

l COMMENT Part c of this question asks for two (2) protection /

4 control functions provided by level transmitter 459, 460, and 461.

The reactor trip at 92% level, should be added to the key.

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3.13/6.13 QUESTION a.

What 2 time delay actions occur in the condensate / main feedwater system when main feed pump suction pressure stays below 300 psig?

b.

What conditions cause automatic closure of the steam generator Main Feedwater Stop (Isolation)

Valves (3232A, B, C)?

4 c.

If the steam generator Feed Regulating valves l

are closed by a protection signal (SSPS), the signal must be cleared to reopen the valves.

What are the THREE protection. signals that close the valves and HOW is each cleared?

ANSWER I

a.

Low FWP suction pressure (300 psig for 10 sec) starts the standby condensate pump [0.5].

Low FWP suction. pressure (300 psig for 30 sec) l causes FWP trip [0.5].

b.

Valve auto closes on FWP trip signal from both pumps with control switch in AUTO.

c.

1.

Hi Hi S/G level - cleared by closing reactor trip breakers 2.

SI - cleared by closing reactor trip breakers 3.

Lo Tave & P-4 -~ cleared by manual reset-i button (1/2) on the MCB to reset Lo Tave signal [0.5 each]

REFERENCE Farley Lesson Plans Volume 4, Tab 7, pp. 28, 29, 36 &

Volume 6, Tab 11, pp. 18, 19 I

COMMENT It is assumed that the information in parenthesis in answer A is not required for full-credit since the question did not ask for setpoints.

Part B answer should not require " control switch.in AUT0".

Thss is a spring return to AUTO switch from all p o s i t i o'n s.

Therefore, AUTO is not considered an

" interlock".

Reference SG Protection Lesson Plan, XQ, Pg 38 and Figure 16.

l Part C.1 of this question needs qualifying.

If reactor power is <35% (P-9), a reactor trip does not (CONTINUED ON NEXT PAGE)

L

f 3.13/6.13 COMMENT (CONT'D) a occur and the feed reg valves will reopen once the Hi Hi level clears.

The key answer C.1 is correct for a i

H1 Hi level condition when >35% power.

Reference S/G l

Protection Lesson Plan, XQ, pg 16 and Figure 8.

I For answer C.2 the SI signal must also be reset in order to reset the-reactor trip (P-4).

Reference S/G Protection Lesson Plan, XQ, Figure 8.

Answer.should read 2/2 verses 1/2 pushbuttons.

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Each button resets one train.

Both trains must be I

reset in order to get air to the valve operator.

j Reference S/G Protection Lesson Pl an, XQ, pg.10.

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3.14/6.14 QUESTION For each condition EXPLAIN which component (s) would be generating a rod movement signal and the response of Bank l

D rods to this signal.

Assume no other Rod Stop signals present, Reactor at power.

BANK SELECTOR SW.

IN-00T-HOLD LEVER PLANT PARAMETER "D"

POSITION i

a.

Manual In RIL LO-L0 ALARM 180 steps b.

"D" Hold Tavg-Tref +4 deg 180 steps i

c.

Manual Out

".. Urgent Failure" -200 steps d.

Auto Hold Tavg-Tref -4 deg.

222 steps

)

ANSWER a.

Manual signal calling for rod movement Rods move IN.

b.

Tave-Tref deviation calling for rod movement however with "D" selected rods DO NOT move.

]

i c.

Manual signal calling for rod movement however rods 1

00 NOT move 1

1 d.

Tave-Tref deviation calling for rod movement however 228 steps blocks movement, rods DO NOT move.

[caf]

REFERENCE i

FNP, LP:

Rod Cont. Sys.

j COMMENT Answer d should read 220 verses "228 steps".

Reference Rod Control System Lesson Plan XE page 8.

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3.16/6.15 QUESTION Describe the controls for the charging pumps.

Include control location interlocks, switch positions, and auto starts.

ANSWER Pumps A & C:

Controlled from MCB or HSP

[0.25]

MCB - STOP/AUT0/ START

[0.25]

- Operable only when remote selected at H3P [0.25]

Auto start:

1.

ESF sequencer if B pump bkr open

[0.25]

2.

LOSP sequencer

[0.25]

3.

B pump fault trip

[0.25]

--s g

Pump B Controlled from MCB or HSP

[0.25]

MCB - STOP/AUT0/ START

[0.25] one for each train

[0.25]

- Operable only when remote selected at HSP

[0.25]

Auto start:

1.

Selected train charging pump trips

[0.25]

2.

LOSP or ESF sequencer with selected train charging pump brk racked out

[0.25]

REFERENCE FNP, License Retrng, ECCS COMMENT We do not believe the wording of this question solicits the depth or scope addressed by the key.

The key answer is, for all intent and purposes, a direct copy of an objective used in the 1986 Retraining Program.

These candidates did not receive the retraining objectives.

It is reasonable to believe the students have this knowledge but it is doubtful that without having the objectives one would receive such a detailed response to this question.

We request the grader of this question be allowed liberal interpretation of the candidate's response rather than the strict, detailed response the key demands.

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j 4.11/7.10 QUESTION l

List the immediate operator actions for a continuous j

control bank D withdrawal.

ANSWER

!~

1.

Place rods in manual 2.

Place turbine DEH in hold 3.

Insert control bank D to return Tavg to program j

[0.5 points each]

REFERENCE FNP-1-A0P-19.0 1

i' COMMENT l

This question has a total value of 1.00 point.

Yet the answer has three (3) parts 0 0.5 each.

Adjust j

point valve of answers.

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4.13/7.13 QUESTION l

a.

After a Residual Heat Removal (RHR) pump is i

started for-plant cooldown, but before placing l

the train in service, the pump is operated on miniflow recirculation for a minimum of 10 min.

WHY?

b.

How is a low boron concentration (in an RHR train to be placed in service) corrected?

i c.

Would starting an RHR pump, with the CVCS letdown pressure control valve (PCV-145) in manual result in a pressure INCREASE, DECREASE, or NO CHANGE in the Reactor Coolant System during solid plant operation?

i d.

When establishing a bubble in the pressurizer, WHY must both RHR trains be valved into their respective RCS hot legs?

i ANSWER 1

a.

To river water for sampling

[0.5]

l b.

Flowpath aligned from RWST to CVCS letdown (without exceeding 120 gpm thru LTDN HX) until boron concentration is' equal to or greater than RCS boron concentration.

[1.0) l c.

Increase

[0.5]

1 d.

RCS overpressure protection.provided by RHR inlet relief valves

[1.0]

i COMMENT Proctor further defined Part C of question to solicit effect on the system " initially".

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4.14/7.14 QUESTION While performing EEP-3, Steam Generator Tube Rupture procedure, how is lowering both the Pressurizor level AND ruptured Steam Generator level accomplishad?

ANSWER Reduce makeup flow REFEREhCE FNP-EEP-3 COMMENT The answer may be written as increase letdown or decrease charging.

4.16 QUESTION What reference is used to verify the settings on MCB Manual / Auto station potentiometers are correct when starting up the unit from cold shutdown?

Al'SWER Curve Book REFERENCE FNP-1-UDP-1.1 COMMENT The Curve Book is also commonly referred to as the Pot Setpoint Book and candidates may answer this question either way.

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4.17/7.15 QUESTION When would a transition be made from A0P-1.0, Excessive RCS Leakage, to the due U0P-3.1 to a loss of reactor coolant?

Assume no reactor trip or SI occurs.

ANSWER When charging pumps cannot maintain pressurizer level. (or leakge > T/S limits)

REFERENCE FNP-1-A0P-1.0 COMMENT The proctor changed "EEP Series" in the question to "UOP-3.1".

The answer key was also changed by the proctor to include "or leakage > T/S limits".

4.20/7.17 QUESTION Under what conditions are the adverse containment values for instrumentation used in the EEPs?

ANSWER

>4 psig in containment [0.5] OR > or = 10E5 in containment [0.5]

REFERENCE FNP-EEP-0, Fold Out Page COMMENT Add R/hr to "10E5 in containment"

e 4.21/7.18 QUESTION List three different allowable times an FRG may be exited if it was entered by direction of an ORANGE path.

ANSWER 1.

By the FRG directions 2.

Any red path occurs 3.

Any higher priority orange path occurs 0 0.5 points each REFERENCE FNP ERP Requal Training COMNENT The caution statement on page 2/28 of ECP-0.0, Loss Of All AC Power", also allows for exiting FRGs and should be considered as a correct response.

1 SENIOR REACTOR OPERATOR WRITTEN EXAMINATION 5.08 QUESTION Indicate whether the following will INCREASE, DECREASE, or HAVE NO EFFECT on the available (actual) l Net Positive Suction Head (NPSH).

a.

Increasing pump speed j

b.

Increasing pump suction temperature 4

c.

Increasing system pressure (consider a closed system) 1 4

ANSWER i

l a.

DECREASE i

l b.

DECREASE c.

INCREASE t

t REFERENCE

{

General Physics, HT&FF, p. 320 l

i COMMENT l

l The proctor added " consider a. closed system" to part j

c of the question.

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O 5.15 QUESTION a.

What is the definition of Shutdown Margin (SDM)?

b.

If a stuck rod exists while the reactor is at power, what adjustments, if any,'must be made to the SDM calculation?

ANSWER a.

Amount by which core would be subcritical [+.25]

at hot shutdown conditions (540 deg F) [+.25] if all control rods were trip)ed assuming highest worth rod fully withdrawn :+.25] and no changes in xenon or boron concentration.

[+.25]

b.

No (already accounted for in SMD calculation) l

[+.5]

i REFERENCE FNP T/S, pp. 3.1.1 - 3.1.3 1

l COMMENT Both answers A and B are not FNP specific.

"A" should read as follows:

SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

Reference FNP Unit #1 Technical Specification Definition 1.28.

The Part B answer is; YES an adjustment is necessary.

Reference FNP Unit #1 Technical specifications page 3/41-1 item 4.1.1.1.1 a.

8.05 QUESTION Tech Specs defines Shutdown Margin as..." Shutdown Margin shall be the instantaneous amount of reactivity by which the Reactor is subcritical or would be subcritical from its present condition assuming...".

STATE the assumptions made for the l

plant conditions which complete the definition of j

Shutdown Margin.

i ANSWER j

All full length control element assemblies shutdown i

and reg. are fully inserted [0.5] except for the single assembly of highest reactivity worth which is

]

assumed to be fully withdrawn.

[0.5]

I REFERENCE l

St. Lucie Tech Spec def. 1.29 North Anna TS def.

FNP def.

i COMMENT Use FNP Technical Specification as key answer.

(Definition.1.28) i i

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1 8.07 QUESTION The following refer to " Emergency Plan Implementation Procedures", FNP-0-EIP's.

I a.

Who, by title, is responsible for the immediate and unilateral declaration of an emergency AND the initiation of emergency response during the initial phase of an emergency?

i b.

Who, by title, is the ONLY' individual authorized to downgrade an emergency level once an emergency has been declared?

,l c.

Which channel number on the public address system is designated for use during emergencies?

1, ANSWER a.

Shift Supervisor-b.

Emergency Director c.

Channel 5 1

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[3 0 0.5 ea.]

REFERENCE FNP-0-EIP-2, p. 2; EIP-3,.p. 1, EIP-8, p.

1 COMMENT During the simulator and oral examinations the week i

of 7-14, several people were quizzed regarding the l

downgrading of emergency level classification.

It 1

was pointed out during this time that per G0-EIP-115, DE-ESCALATION OF EMERGENCY CLASSIFICATION AND REC 0VERY INITIATION, the Recovery manager must also authorize the de-escalation from Site Area Emergency or General Emergency classifications.

Both Emergency Director or Recovery Manager should be considered a correct response for Part B.

1 Reference GO-EIP-115, page 1.

(COPY ATTACHED) 1 1

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l 8.10 QUESTION 1

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The Technical Specifications for reactor trip system l

instrumentation channels specifies if one channel of l

Power Range Nuclear Instrumentation is inoperable, a l

Quadrant Power Tilt Ratio must be done at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if power is at 100%.

a.

How is the Quadrant Power Tilt determined in this case?

b.

If the Quadrant Power Tilt Ratio is not-determined within the allowable time, what must be done?

ANSWER a.

The QPTR is determined by using the incore moveable detectors, b.

Reactor power must be reduced (to less than 75%)

[0.5] and the Power Range high neutron flux trip setpoint must be reduced (to < 85% within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />).

[0.5]

REFERENCE FNP Technical Specifications, pp. 3/4 2-13, 3-6 COMMENT We assume the information in parenthesis of answer B is not required for full credit.

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8.14 QUESTION During Mode 1 operation of unit 1 it is found that 2 of 3 channels for Pressurizer Pressure high reactor trip are inoperable due to a generic material deficiency (repair time 24 days).

Using Tech Spec LCO's provided, determine what actions must be taken as a result of this failure?

State specific LC0/ action steps which apply.

ANSWER LC0 3.0.3 applies i

REFERENCE North Anna LCO 3.0.3 FNP 3.0.3 COMMENT LC0 3.0.3 is commonly referred to as the " Motherhood Statement" and should be accepted.

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8.16 QUESTION Unit 1 is operating in Mode 1 with Diesel Generator Set A (DG 1-2A & 1-C) inoperable when it is identi-fied that the steam supplies.to the Turbine driven

'AFW pump are inoperable.

Specifically which Action /

LCO would be entered and carried out?

Assume re-maining DG's surveillances completed satisfactorily.

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ANSWER i

FNP-TS 3.0.5 REFERENCE I

FNP TS 3.0.5 COMMENT l

I This particular question is asking the candidate to interpret Technical Specifications provided him.

Specification 3.7.1.2.b requires shutdown to hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for 2 inoperable AFW pumps.

Specification 3.0.5 could also be applied to this situation.

3.0.5 allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to fulfill the two conditions of the specification or be in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Therefore 3.0.5 is less restrictive than 3.7.1.2.b for the situation addressed in the j

question.

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Considering the candidates were not provided specification 3.0.5 in its entirety and the fact that j

specification 3.7.1.2 is more restrictive, either specification should be accepted as a correct answer.

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GO-EIP-115 D

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L DE-ESCALATION OF EMERGENCY CLASSIFICATION AND' RECOVERY INITIATION 1.0 Purpose The purpose of this procedure is to delineate authorities for de-escalation of FNP emergency classification and for initiation of recovery actions, establish criteria f or such actions and identify notifications to be made concurrent with such action.

2.0 Scope This procedure applies to de-escalation of emergency classi-fication or initiation of the recovery phase following entry into an Alert, Site Area or General Emergency classification s

at Farley Nuclear Plant.

3.0

.teferences 3.1 FNP Emergency Plan 3.2 FNP-0-EIP-9 i

3.3 GO-EIP-101 4.0 Authority 4.1 De-escalation f rom Site Area Emergency or General Emer-gency cl assi ficati ons mu st be authorized by the Recov-ery Manager.

4.2 Termination of emergency classification and e n t ry into the Recovery Phase must be authorized by the Recovery Manager.

5.0 De-escalation of Emergency Classification 5.1 Meteorological and Plant Pa rameter Review Rev. O

GO-EIP-il5 s

"O 5.1.1 The decision to de-escalate emergency classifi-cation will be based on a comprehensive review of plant system parameters, radfological inven-tory / release potential and cu rrent and projected meteorological conditions.

5.1.2 The Emergency Director is responsible f or initi-ating recommendation of emergency classification de-escalation.

Such recommendation shall be based on a review of parameters that include, but are not necessarily limited to, the followirig as applicable to the existing emergency condition:

1)

Stability of the reactor system (mode, shut-O down margin, subcooling margin, pressure, etc.);

2)

Quantity and integrity of intact barriers preventing or mitigating radioactive releases (e.g., cladding, RCS vessel, piping & valves, containment, HEPA filters and charcoal fil-ters) including any potential tnreats to barrier integrity (e.g. cu rrent and projected containment H 2 concentration);

3)

Availability and operability of a heat sink; 4)

Operability, accuracy and integrity of plant instrumentation, including effluent monitors and radiation monitoring equipment; 5)

Availability and reliability of offsite and emergency power sources;

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G0-EIP-115

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6)

Status of natural phenomena involved in accident initiation or affecting accident mitigation (e.g. seismic events, flood, tornado, acc.)

7)

Radiological and meteorological parameters listed in 5.1.3.

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5.1.3 The Dose Assessment Director is responsible for advising the Emergency Director regarding de-escalation advisability based on a review of

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radiological and meteorological parameters that

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include but are not necessarily limited to the following:

1)

Current and anticipated effluent release rates (both monitored effluent paths and best estimate of unmonitored paths);

2)

Radioactive material inventory constituting potential release source; l

3)

Current knowledge regarding isotopic makeup i

of effluents and radioactive material inven-tories; 4)

Current offsite dose rates (calculated and measured);

5)

Meteorological forecasts and resultant pre-dictions of changes in atmospheric stability class, deposition rate, population affected, etc.

5.2 De-escalation Criteria

GO-EIP-115 A OO The Recovery Manager will analyze input from his advis-ors in the areas listed above to decide on emergency classification de-escalation.

The following criteria shall be considered appropriate f or initiating a reduc-tion in emergency classification.

5.2.1 Reduction from General Emergency to Site Area Emergency 1)

Potential exposu re to non-evacuated offsite areas (based on exposure to current time, cu rre nt release rate and dose rates, projec-tad release rate and dose rates and project-ed release duration) is less than 1 Rem whole body and less than 2.5 Rem Thyroid.

Qualitative judgement of projections should include assessment of dose assessment model accuracy utilizing Radiation Monitoring Team measurements.

2)

The plant is stable and subcritical with either no substantial core degradation existing or expected OR If core damage has occurred, no fu rtner - de-gradation is expected and either no realis-tic potential exists for loss of containment 7-integrity or the ground l e v e l.,ra l. e a s e o f

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radioactive materials availaole for release from containment would not cause criteria 1) C 1

GO-EIP-115 T_.

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to be exceeded under worst case meteorological conditions.

5.2.2 Reduction f rom Site Area Emergency to. Alert 1)

Cu rrent and projected of f site dose ra tes due to cloud passage are less than 1 mR/hr.

Tnis value excludes background due to shine from p ri or deposi ti on.

Qualitative judgement should be applied based on dose assessment model accuracy utilizing Radiation Monitoring Team measu rements.

2)

Current and projected ef fluent releases do

.not exceed the limits established by 10CFR20.

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3)

If the emergency condition involved the p ri-mary plant, the reactor is subcritical and proceeding to a cold shutdown (or equivalent) condition.

4)

The plant is stable.

5)

If the emergency event resulted in core dama ge or the release of. radioactive material to containment, either containment' integrity is established and.unthreatened or the dose rate 4

that would result from loss of containment integrity under worst case meteorological conditions would be less than 1 mR/hr at the

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site boundary.

NOTE THAT. MEETING THESE CRITERI A 00ES NOT AUTOMATICALLY INDICATE THAT OFFSITE PROTECTI'IE ACTI0f4 MEASURES SHOULO BE RELAXED.

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i GO-EIP-115 AO CRITERI A MEASURE THE POTENTI AL FOR ADDITIONAL RELEASES AND TH E POTENTIAL IMPACT OF ADDITIONAL RELEASES ON THE EPZ, NOT THE IMPACT PRIOR RELEASES HAVE HAD.

OEPOSITIUN OF EFFLUENTS IN THE EPZ MAY PREVENT RELAXATION OF 0FFSITE PROTECTIVE ACTION UNTIL FURTHER EVALUAT20N IS COMPLETED.

DECISIONS TO RELAX OFFSITE PROTECTIVE ACTIONS WILL BE MADE SY OFFSITE AUTHORITIES AND SHOULD INCLUDE CONSIDERATIONS IN ADDITION TO THOSE DISCUSSED ABOVE.

5.3 Notificati,on of De-escalation The following agencies should be notified of the decision to de-escal ate emergency cl as s i fi cati on.

5.3.1 Nucl ea r Re gul a t o ry Commi s s i on Laj t l a i s / D a t e 5.3.2 Alabama Radiological Health initials /Date 5.3.3 Georgia Environmental P rotection Division initials /Date 5.3.4 Florida Department of Natural Resources initials /Date 6.0 Activation of Recovery Organization and Recovery Phase 6.1 Plant Parameter Revi ew The decision to move from an Aler condition to tne Re-covery Phase will be based.on a comprenensi ve review of plant parameters and radiological inventory / release po-tential.

This review shall include but not necessarily be limited to the following:

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1)

Stability of tne reactor snutdown condition (if involved in tne event) i.e., successful movement toward a cold shutdown condition.

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GO-EIP-115 s

2)

Integrity of intact barriers preventing or mitigating radioactive releases.

3)

Operability of radioactive waste systems and decon-tami nati on f aciliti es.

4)

The availability and operability of a neat sink.

5)

The integrity of power supplies and electrical equip-ment.

6)

The operability and integrity of instrumentation including radiation monitoring equipment.

In the latter instance this shall include portable equipment assigned to the emergency.

7)

Availability of trained personnel and support s/

services.

6.2 Criteria for Entering Recovery Phase The Recovery Manager will analyze the input from his advisors in the areas listed above to determine if plant restoration efforts can begin.

The following criteria shall be considered appropriate f or the initiation of recovery measures:

1)

Plant parameters of operation no longer indicate a

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potential or actual emergency exists.

2)

The release of radioactivity from the plant is con-

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trollable and no longer exceeds permissible levels a

and no danger to the public from tnis source is credible.

3)

The plant is capable of sustaining itself in a long term snut-down condition.

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G0-EIP-115 4)

Plant ent ry and cl ean-up is possible without workers j

receiving in excess of their permissible exposures.

6.3 Notification 6.3.1 The Recovery Manager shall n o t i fy the Plant Manager and company management that a decision has been reached to initiate a recovery operation.

He shall then notify offsite agencies' represen-tatives ensuring the NRC,. and state and local authorities are provideo with the same informa-tion.

He shall also inform tnese agencies if any change in the structure of the recovery organiza-tion is to occu r.

5.3.2 The Administrative Support Director or nis designate shall noti fy all APCo departments initially notified at the onset of the accicent, informing-tnem of the current status of the affected unit, changes in the structure of the reco ve ry orga ni zat i on, a nd any information pertinent to that department.

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