ML20006E628

Forwards Comments on Written Reactor Operator & Senior Reactor Operator Exams Administered on 891023

ML20006E628
Person / Time
Site:	Farley
Issue date:	10/27/1989
From:	Dennis Morey ALABAMA POWER CO.
To:	Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20006E603	List: ML20006E602 ML20006E606 ML20006E628
References
FNP-89-0541-TRN, FNP-89-541-TRN, NUDOCS 9002260155
Download: ML20006E628 (21)

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); M. Farley Nuclear Plant t, ft+TP

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Ashford; Alabama 36312 I

Telephone 205 099-5156 y

1 ENCLOSURE 3 LAlabamaPower une souwn e:en swe, FNP-89-0541-TRN

^

October 27, 1989-p Th'e Regional Administrator, Region'II Nuclear Regulatory Consnission-

,101-Marietta St., N.W.

Atlanta,_GA130323 ATTN

Mr. John F.IMunro 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 t

iNuclear Plant on October 23,!1989.

All-of the questions addressed appear on both the SRO exam and the R0 exam;-

' ~ both numbers _ have been-placed at the-top of the question.

f The< pre-exam review conducted-in Atlanta appears to -have.b'een very

. beneficial in_ producing a high quality test.

s The courteous and professional manner which your staff displayed 11n

. 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 2%u/

a D.N.Morey,pl General ~ Manager - Nuclear Plant DNM,III/LSW/RJV:mjk Enclosures.

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-QUESTION R0.2.03/SR0_5.05

.(1.00)~

The refueling cavity seal ring has failed, causing-level in the l

refueling cavity to decrease rapidly. 'A fuel assembly is in the process of being withdrawn from the containment upender. Which of the following' is the appropriate' action to be taken?-

- a.

' Verify both containment sump pumps are operating and that the l

y

. liquid waste processing system is capable of receiving-discharge from the sump pumps.

n b.

~Run one RHR pump-in-cooldown alignment and align the other RHR

-pump,to the RWST.

'Immediately stop all RHR pumps.

c.

~

- d.'

PlaceLthe fuel assembly being moved into the core.

ANSWER 2.03/5.05-(1.00)

(b) i

Reference:

FNP Question bank no.-052521HT1001R 000036G010'

...(KA'S)

, i

'CONMENT L

l' This. question was originally written as an open reference-question.

It is based J on / FNP-1-A0P-30.0- (copy attached).

Choice b.

is based on subsequent action steps 5.7 and 5.7.1.

An. operator -with procedure in hand would use those steps ' of the subsequent actions to combat the casualty in progress making choice-b. a' correct answer.

Subsequent actions are not required to be memorized, so an operator without the.

procedure may not recognize this as a procedure requirement.

Choice d. is an-allowable action per-immediate action step 4.1.

It is not the most logical action to take with the-fuel being removed from the' upender, but. it - is the only response listed that is an intnediate operator action.

Unlike subsequent actions, immediate actions are required to be memorized, i

Based on the above information, there are two correct answers, choices i

b. and.d.-

b l:

RECOMMENDATION

- Based on two correct answers being available, the question should be deleted from the exam.

For future use, the question should be modified such that only one correct response is available.

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FNP-1-AOP-30.0 FARLEY' NUCLEAR PLANT-4 UNIT 1 ABNORMAL OPERATING PROCEDURE AOP-30.0 x

't

. ;t REFUELING ACCIDENT r,

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5, 1.0 Purpose l

2

.This procedure provides the symptoms, Automatic _

u Actions, Immediate Operator Actions and Subsequent jr Operator Actions for a refueling accident.-

.j 2.0

-symptoms 2.1 fDamaged fuel assembly and high radiation on

,p radiation' monitor (R-2, R-5, R-11, R-12~,

R-24A(B), R-25A(B)].

2.2 Rapid drop in refueling cavity level during refueling.

3.0 Automatic Actions None

'[

4.0 Immediate Operator Actions i

4.1 IF conditions (i.e.' radihtion levels or' rate'of cavity level decrease) permit as; determined.by the SRO in charge of fuel handling,'THEN attempt to. return any fuelibeing moved to-the reactor

-t vessel or to.the.upender-and' lower the upender to-its horizontal position.

4.2 Evacuate Containment or Spent Fuel area depending-on area where accident has occurred (evacuate Containment and_ Spent Fuel ~ area if SG nozzle dam failure suspected)-and close all access doors.

5.0 Subsequent Operator Actions 5.1 consider actuating the plant emergency alarm and announce:over the public address system for all personnel to report to their designated assembly areas.

5.2 IF fuel damage has occurred in Containment, THEN TEop Containment Purge and Mini Purge supply and exhaust fans and close dampers, 1-CP-HV-3196, 3197, 3198 A,B,C,D, 2867 C&D and 2866.C&D.

Gen. Rev. 3 5

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4 FNP-1-AOP-3060

e 5.3' Ir, failure of a' Steam Generator primary nozzle 3am, Refueling. Cavity seal. ring f ailure OR othe r:

major Refueling Cavity leakage is suspecIed, THEN place control switches.(BOP) for Containment Sump E..

Pumps in " PULL TO LOCK".

5.4 Ir fuel damage has occurred in Containment, THEN

{

ig.

perform the following:

s 5. 4 ~.1 suspend all fuel handling' operations in the spent ruel Building.

,c b::

5.4.2 withdraw th'e fuel tra'nsfer' cart to.the spent ruel Building ifLthe damaged; fuel does not interfere with cart movement.

s 5.4.3 Close the fuel transferitube gate valve.

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5.4.4' Monitor R-2, 11 and 12 for indication of radiation' increase to aid in' deter-mination of the magnitude'of fuel damage.

5.5 IF-fuel damage has occurred in the spent ruel

' ~

Euilding, THEN perform the following:

5.5.1-Close all fuel handling hatches into the Spent Fuel area.

T 5.5.2

-Suspend all fuel handling operations in Containment.

i c

5.5.3 Monitor R-5 and R-25 (A and B) for indication of radiation increase to aid L

in determination of the magnitude of the-fuel' damage.

5.6 Initiate FNP-0-EIP-12, Alert, as necessary.

t 5.7 IF. Refueling Cavity. leakage is suspected due'to possible Refueling Cavity seal ring failure.OR to Steam Generator primary ~ nozzle dam' failure, THEN verify only one RHR' Pump running with its suction lined up to the RCS loops AND closely monitor running pump for evidence of cavitation.

CAUTION Monitor RWST level closely l

and provide makeup to it as required.

Monitor: running RHR Pumps for evidence of cavitation and secure RHR Pumps if necessary.

l 5.7.1 Line up one of the RHR Pumps to fill the Refueling Cavity from the RWST using the guidance of FNP-1-SOP-7.0.

Gen. Rev. 3 L

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-r FNP-1-AOP-30.0 5.7.2 IF: required,-THEN line.up a Charging Pump NE the Refueling Water Purification Pump EU supply a'3ditional makeup water.

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t 5.7.3 suspend all fuel' handling operations in Sl the spent Fuel Building.

l-5.7.4 IF donditions (i.e.: radiation levels'or:

rate of' canal level decrease)/ permit, as determined by the SRO in charge of fuel handling, ETF which may be in ths upender THEN' attempt to transfer any' Fuel-Assen to the spent Fuel Pool.

.1 S.7.5 withdraw the fuel ~ transfer cart to the~

~~-

Spent Fuel Building.

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5.7.6 Close the' fuel. transfer tube gate valve.

5.7.7 IF required, THEN supply makeup to the RWsT.

5.8 Initiate FNP-0-EIP-8, Notification Procedure.

5.9 Initiate FNP-0-EIP-14, Re-Entry. Procedures.

5.10 sample and analyze the4 Refueling Cavity water and 1[

spent Fuel Pool water.- IF required, THEN operate

.the'sFPC Purification SysEen per FNP-I;3DP-54.0..

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QUESTION R0 2'.12/SR0 5.15a 1(1.00)-

Using"the attached Table 03 from FRP-I.3, Response l.to Voids in Reactor -

m Vessel, determine theisize in,;% pressurizer level,- of the ; entire void -

if-initial-RCS pressure was 450 psig and.was raised to 550 psig. During-the pressure increase, PZR level dropped.from 45% to 310 in a.:

14%-

b.

65%

x c.

79%L 1

d.

84.5%

ANSWER 2.12/5.15 (1.00) l(c) a

Reference:

~

FNP question bank no. 052302F013.

l-FRP-I ~.-3 ~

000074A206

....(KA'S) 1 COMMENT

. In order to correctly answer this - question, the operator must know K.whether to use-initial RCS pressure 3(450 psig) or final. RCS pressure (550); to i enter Table 3 that -.was;provided in thestest < (attached).

The instructions for using final RCS pressure-to' enter the-table;comes from FNP-1-FRP-I.3 step 13.5.2- (attached).

This'. step was not:provided with the exam.

No -instructions. in. the stem of the ~ question were provided.

.This is not a procedure we would expect an operator to memorize.

j RECOMMENDATION i

Based on 'the ' lack of instructions for using the table, this -question p

should-be -deleted from the exam.

For future use, portions of step 13 of.FRP-I.3 should be provided.

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A'ction/ Expected Response Response NOT Obtained 4

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4 13 Determine miniman ' allowable pressurizer level during vent.

T CAttrION Pressurizer level must be maintained' greater than or equal to 20% (50%}.

IT pressurizer level begins to fall below IM- (50%} during.an evolution, DEN stop the evolution.

13.1 -verify charging flow control in manual and record one pressurizer level protection channel (LI-459A, 460 or 461) in Table 1 (below).

CHG..

FI G

_ Fr,-122 l

controller:in manual

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i FNP-1-PRP-I.3 Response to Voids in Reactor Ver.sel Revision 5

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step Action / Expected Response Response NOT Obtained i

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o NO:lY During the following RCS pressurization, pressurizer level change is used as an indication of void size, 'Iherefore, r

charging and letdown flows should NOT be adjusted during the evolution.

i l

13.2 Stop pressurizer spray and e

turn on pressurizer heaters l

to raise RCS pressure by 100 psi unless pressurizer level drops to 20% {50%).

_PI-402A (or 403A)

SPR VLV PVC-444C (D)

_PK-444C controller adjusted PK-444D controller adjusted

/].

.O PR2,R -

HTR BKUP (VARIABLE)

_1A on

_1B on

_1C on ID on i

_ 1E on 13.3 Record final RCS pressure l

(PI-402A or 403A), pressurizer i

level (using same channel l

recorded above) and complete the o pressurizer level in Table 1 (below).

TABLE 1 1

Pressurizer-Level L2 RCS Pressure Initial Final a Level Final (Step 13.1)

(Step 13.3)

(Step 13.3)

(Step 13.3) j

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c, FNP-1-FRP-I.3 Response to Voids in Reactor Vessel Revision'6

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step Action / Expected Response Response NOT Obtained l

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• • • • • • • • • • • • e******eneme****eme****ene************

j CAttrION IF RCS pressure exceeded 2000 psig,-

i

'!9EN reblocking of low pressurizer pressure EI should be performed when permissible.

• • • • • • • • • • • • • • • • • • • • • e****************************

13.4 Restore RCS pressure and pressurizer level to values prior to raising i'

RCS pressure.

t CHG

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Fi m rK-122 controller

~ manually adjusted as required SPR VLV RCS PVC-444C (D)

PRZR AUX SPR.

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~PK-444C CVC-HV-8145 used-p1 =

~

controller manually

~if required 1's ) <

adjusted PK-444D controller

~ manually adjusted PRZR HTR BKUP (VARIABLE)

GRP 1A, B, C, D or E

- turned on or off as required t

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FNP-1-FRP-I.3 Resp mse to Voids in Reactor Vessel Revision 5 g _I )

i Step Action / Expected Response Response NOT Obtained U

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13.5 Ir pressurizer level 13.5

'Ibe minim a allowable remained pressurizer level during 20% (50%) greater than

, ntal obtain the reactor head vent d, s mininum alloE5le pressurizer 20% (50%), proceed to level during reactor head step 14.

vent as follows.

13.5.1 Record existing pressurizer i

level in Table 2 (below).

i 13.5.2 Using the o Pressurizer j

level and final RCS pressure from Table 1, obtain the pressuriser level change required to remove entire i

void from Table 3 (attached).

13.5.3 subtract the required level change from the existing i

pressurizer level and record in Table 2 (below).

'~'

13.5.4 Ir calculated level is Iess than 20% (50%), THDI r

mininum allowable level is 20% (50%), otherwise use calculated value.

i TABLE 2 l

Pressurizer Level Existing 6 Level Required Min Level Calculated I

(Step 13.5.1)

(Step 13.5.2)

(Step 13.5.3) l-l l

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no-1-emp-2.3 mesponse to veids in meester vessel movision 6 n

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sEs3 CNNE El Em 155. M355 'E3 3505 BEDE M (a m 1 ICR N ML15 5. M 1% - M )

i m CNME El EmIBM. (%)

1 2

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7 8

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100 1.30 2.36 3.4E 4.3s 5.75 6.ast

8. m 9.116 10.323 11.C R - 30 2.1C 4.34 6.4E

8. 5 10.7 5 12. 5 2 15.0 3 17.116 19.323 21.C 1

C 20 3.14 4.34 9.4 4 12. M 15.7 5 3.m

22. 9 3.116 3.323, 3L.C

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s 80 4.14 8.34 12.44 16.M E.75 24.ast 3.029 33.116 N.323 a.C

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20 5.147 10 2 4 15.4E 3. 5 3.75 Nm M.m E.116 4.323 SL.C i

P 800 6.1 4 12.3 4 2.4 4 24.5 5 30.75 36.aB2 C.029 4.176 5.323 E.C R 7tB 7.147 14.34 2L.44 3.M B.75 4.M M.9

57. m 64.323 71.C

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57. m 8.176 73.323 81'.47 lO

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E 1300 13.1C 3.294 5.44 52.M S.75 78. 5 2 32.029 105. m 118.323 131.C l

1400 14.1C 3.294 4.44 56.55 10.7 5 84.882 99.0 3 113.176 127.323 14.47 i

l P 1500 15.147 30.294 6.44 m.5BB 75.7 5. 90.882 106.029 12L.176 136.323 151.47 s 1800 M.14 31.294 4.44 64.!M 00.7 5 96.882 113.029 129.176 1 6.323 18L.C I 1100 17.1 4 34.3 4 EL.4 E S. M 8.75 102.M 13.9 1 N.116 154.323 11C G 15 5 2.147 N.34 M.4E 72.5 90.7 5 108.5 2 127.03 1 4.176 183.323 181.47 l

1918 19.147 5.3 4 57.4 4 76. 5 95.75 114.M 134.029 153.176 172.323 151.C 2000 2.147' W.294 60.44 80.588 100.75 120.882 14.029 181.176 181.323 201.C O

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36. 4 7 E.154 W.21 72. 5 77.25. E.52 M.89 E.646 M.793 lot.M

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t 1300 144.E7 157.164 170.911 24.5 1M.25 210.352 223.69 236.646 28.793 262.94 1400 15.E7 19.M4 183.911 138.5 212.25 236.352 240.89 254.646 268.793 282.M P 1500 186.E7 18L.M4 196.911 212.5 227.25 2E.52 257.499 32.646 287.33 302.94 3 1800 177.E7 133.184 25.21 23.5 242.25 25.5 D4.89 290.646' 306.793 322.M I 17tB 15.E7 25.M4 221.9L1 2a.S 257.25 274.E EL.89 25.646 325.793 342.M o 1soO is.E7 217.7s4 2s.21 254.3 272.2as as0.351 3as.e9 326.646 344.793 362.94 1910 210.E7 229.164 2a.51 28.5 297.25 306.E 325.89 344.646 363.793 M2.M 2l00 22L.E7 2E.164 26L.SL1 31.5 301.25 322.362 342.89 362.646 M2.793 402.94 O

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QUESTION R0 2.15/SRO 5.18 (1.00)

Intennediate Range channel N-35 is severly UNDERcompensated.

After a i

reactor trip from 100% power, the source range channels:

1 a.

Will automatically reenergize at P-6, normally 15 to 20 minutes after the trip.

b.

mast be reenergized by the operator.

They nonna11y would have 1

auto energized 15 to 20 minutes after the trip.

1 c.

will automatically reenergize at P-6, normally 20 to 25 minutes after the trip.

d.

must be reenergized by the operator. They normally would have 1

auto energized 20 to 25 minutes after the trip.

ANSWER 2.15/5.18 (1.00) i (b) i

Reference:

Lesson Plan OPS-52201D 20-2 1

000032A204

...(KA's)

QUESTION RO 3.01/SRO 6.01 (1.00)

Which one of the following is an indication of an Intermediate Range channel being undercompensated?

a.

-Automatic energizing of the source range when the first intermediate range channel drops below the P-6 setpoint.

.o b.

Actual neutron level falls into the source range without indication.

c.

Indication remains above 10E5 cps in the source range on both channels.

+

d.

Automatic energizing of the source range when the P-10 interlock was cleared on both channels.

l ANSWER 3.01/6.01 (1.00)

(b)

Reference:

FNP Exam Questions from the 1987 Licensed Operator Retraining Program FNP 1987 License Retraining Objectives Cycle 1, OPS-522010, item 3 (3.9/3.9) 015000A303 015000K401

..(KA's) e.,--a

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COMENT.

,e These two questions rely on. the same knowledge in order to be answered correctly.L This would penalize the candidate twice if he or she did not" j

3 know how an intennediate range instrument would respond if under-D-

compensated.

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RECOMENDATION

.)

1 The use of multiple questions testing the same' _ knowledge should be 1

avoided in the future.

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. QUESTION R0 2.20/SR0 5.22 (1.00)

With the pressurizer level control selector switch in position "I/II", a failure causes the following plant events:

(assume no operator actior.5) j 1.

Charging flow reduced to minimum l

2.

Letdown secured and heaters off 3.

Level increase until high presurizer level trip Using the attached drawing, which of the following failures occurred?

l a.

LT-459 failed low b.

LT-460 failed low c.

LT-461 failed low d.

LT-459 failed high e.

LT-460 failed high f.

LT-461 failed high ANSWER 2.20/5.22 (1.00)

(d) i l

Reference:

FNP question bank no. 052201H09015 000028A202

...(KA'S)

COMMENT Using the figure provided on the test, pressuri:er level protection and control Figure 7 (copy attached), it is clear that LT-459 failing high would cause the events listed to occur in the sequence provided.

Choice

d. is a correct answer.

L If LT-460 were to fail icw, it would cause letdown to isolate and turn i

heaters off (event 2).

With letdown isolated, actual pressurizer level would increase.

LT-459, sensing the increase, would reduce charging y

flow to minimum (event 1).

Even with minimum flow, pressurizer level

/-

would continue' to rise until actual level reached the high level L

setpoint, causing a reactor trip (event 3).

The question does not l

state the events occurred in the sequence listed.

With all events l

occurring for LT-460 failing low, choice b. is also a correct answer.

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REC 0WlENDATION 1

Based on two correct answers, the question should be deleted from the

,. e exam.

The stem of the question should be changed to state that the t

events occurred in sequence for future use.

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QUESTION R0 2.22/SR0 5.04 (1.00)

Your Unit has experienced a Small Break LOCA accident.

Briefly explain why RCS pressure might stabilize at 1300 psig for a long period of time, even though break flow rate is greater than SI flow rate.

(Assume the i

pressurizer has emptied)

ANSWER 2.22/5.04 (1.00)

More energy is being put into the RCS from decay heat than is being removed by energy removal means (1.0) 1

Reference:

lesson plan 052702A, Obj. 5-004 FNP exam question bank 000009K306

...(KA'S)

COMMENT The question does not necessarily solicit the answer key response.

WOG Background information for E1 (pages 16 & 25 attached) on page 16 for 1-2 inch breaks discuss the energy balance and energy removal causing pressure to remain elevated.

Or page 25 for a 3 inch break (last paragraph), it discusses the RCS going into saturation to stabilize and maintain RCS pressure elevated.

RECOMMENDATION Accept for full credit the answer key answer, including any reasonable i

discussion of energy balance.

Also accept for full credit any reasonable discussion that discusses the RCS going into saturation without an energy balance discussion.

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Breaks ~ 1" < diameter <- 13-1/2" (IFT )

For break sizes of one to two-inch in equivalent diameter, the RCS will rapidly depressurize early in the transient, and an automatic reactor trip and safety injection signal will be generated based on low pressurizer pressure.

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o During the early stages of the depressurization, when the system is still full of two phase liquid, the break flow, which also will be mostly liquid, is not capable of removing all the decay heat.

Therefore, the early depressurization L

is limited by energy removal considerations, and the RCS pressure will i

I temporarily hang up above the steam generator safety valve set pressure, assuming no steam cump is available.

The RCS pressure stays at this level in i

order to provide a temperature difference from primary to secondary so that core heat may be removed by the steam generator.

At this energy-balance controlled pressure, however, pumped safety injection flow is less than the break flow, and there is a net loss of mass in the RCS.

Voiding throughout the primary side occurs and eventually the RCS begins to drain, starting from the, top of the steam generator tubes.

The rate of RCS drain is determined by the net loss of liquid inventory, a function of both SI flow and break size.

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Prior to the occurrence of draining, heat is removed from the steam generator through continuous two phase natural circulation, with two phase mixture flowing over the top of the steam generator tubes.

As the draining continues, the natural circulation mode of heat removal as just defined ceases, and core heat is removed through condensation of steam in the steam generator.

This method of heat removal is called reflux and is discussed in Reference 2.

The condensation mode of heat removal is almost as efficient as continuous two phase natural circulation in removing heat.

However, condensation heat transfer coefficients may be lower than continuous two phase natural circulation heat transfer coefficients.

Thus, as the steam generator tubes k

drain, a slight increase in primary system pressure occurs to give a greater AT from primary to secondary in order to remove all the decay heat.

The steam generator secondary side pressurizes to the safety valve set pressure l

.E-1 16 HP-Rev. 1 6998B:1b L

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i This analysis was terminated at 83.3 minutes and does not illustrate the maintenance of long term equilibrium pressure where pumped SI equals break flow, the concept that was described previously.

The analysis presented here, however, does represent a stable condition since the core will remain covered and RCS voids are decreasing at the end of the transient.

Pressure is stable 1

and the steam generators are not required for heat removal.

This range of break size is large enough that even if maximum safeguards were injected, RCS repressurization would not occur.

The characteristic behavior mode would be either as just described, where RCS pressure stabilizes below the steam generator pressure, or as described previously in Category 2, where the RCS pressure stabilize's above the steam generator pressure.

The SI flow delivery characteristics and break size will determine the transient behavior mode.

Breaks in the two-inch to thirteen and one-half-inch (13-1/2") diameter range result in significant draining of the RCS and could result in some core uncovery depending on the status of safeguards equipment, break size, and*

break location.

Analyses of cold leg breaks in this category are presented in each plant's SAR to bound the limiting small break for peak clad temperature considerations.

For a three-inch equivalent diameter cold leg break with one train of safeguards equipment operating (one train of pumped safety injection and auxiliary feedwater) the reactor coolrat system initially depressurizes rapidly and reactor trip and safety injection are initiated on their respective pressurizer set pressures.

The reactor coolant system continues to depressurize relatively rapidly to the pressure where the cold legs become soturated (e.g., P,,g at 567'F is ~ 1200 psig). At this point essentially the entire reactor coolant system is at saturation so that c6te exit voiding

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and liquid flashing to steam tend to stabilize and maintain pressure.

This response is similar to breaks of smaller size where the RCS stabilizes above a

the steam generator pressure.

During this stabilization period liquid is discharged through the break since the RCS has not drained down to the point where steam can be vented through the crossover legs.

Since the break is E-1 25 HP-Rev. 1 6998B:1b

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ENCLOSURE 4

. SIMULATION FACILITY FIDELITY REPORT

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i Facility Licensee: Alabama Power Company I

Facility Docket Nos.:

50-348 and 50-364 Operating Test Administered On: October 24 - 25, 1989 This fom is to be used only to report observations.

These observations do.

1 not constitute. audit or ; inspection findings _ and are - not, without further J

verification and review. indicative of noncompliance with '10 CFR 55.45(b).

These observations do 'not affect NRC certification or. approval of: the simulation facility other than to provide 'information which may be used in j

future evaluations.

No licensee action'is required in response to these observations.

i During the conduct of the simulator portion of the operating test, the i

following items were observed-a Step 7.6 of FRP-H.1. Response to Loss of Secondary Heat, directs operators to o

defeat the feedwater isolation signal -to main feedwater bypass valves usino

- Attachment-I to the procedure.

This cannot be simulated on the current model.

This has been previously identified by the training department, and there~are plans to include this in-the updated model.

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