ML20246E882

Forwards Comments on Reactor Operator & Senior Reactor Operator Written Exams Administered on 890626

ML20246E882
Person / Time
Site:	McGuire, Mcguire
Issue date:	06/29/1989
From:	Travis R DUKE POWER CO.
To:	Curtis Rapp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20246E857	List: ML20246E854 ML20246E877 ML20246E882
References
NUDOCS 8908300053
Download: ML20246E882 (6)

Text

. _ _ _ _ _ _ _

s -

ll l DukeIbwer Company (704] 875-4000 -

R McGuire Nuclear Station P0. Bar 488

- Cornelius, NC 280310488 ENCLOSURE 3

, DUKEPOWER June 29, 1989 Mr. Curt Rapp Operator Licensing Section U. S. Nuclear Regulatory Commission, Region II 101 Marietta Street N.'W.

AtlanLS Georgia

Subject:

McGuire Nuclear Station Review of Written License Examinations

Dear Sir:

Attached are our specific comments on the Reactor Operator and Senior Reactor Operator written examinations administered June 26, 1989 at McGuire Nuclear Station.

In addition, McGuire would'like to commend the NRC on performing a pre-exam review with McGuire Operations and Operations Training personnel. McGuire

. feels this helps both the utility and the commission to administer better quality exams. McGuire encourages the commission to continue this worthwhile practice.

Very truly yours,

~

% f( r2Wu R. B. Travis Superintendent of Operations McGuire Nuclear Station RBT:tkb cc: T L. McConnell L. E. Weaver D. A. Baxter D. M. McGinnis P.F.: 9.2 0908300003 890823  ?!

PDR ADOCK 05000369 y PNV y t __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

l n:r

' i_ a [ .

.l

'b E ,

14 u,' NRC EXAM QUESTION POST REVIEW SRO/RO EKAM 1989 1.. 2.l'1a/5.10a QUESTION 2.11 (2.00)

Assume McGuire Unit l.has-sufferedLa large SG tube!

rupture that.results in a reactor trip and SI L actuation. In. addition to high activity levels'in the secondary:

p a. WHAT'are>TWO (2) SG indications that can be _ .

used to. determine which SG has a. tube rupture? ~ (1.0) lI

b. STATE.TWO.(2) ressor.s .hy ir.olation.'of the SG is necessary'. (1.0).

ANSWER 2.11 (2.00)

a. 1. rapidly increasing water level in the affected m SG [+0.5]

2. steam flow /feedwater flow mismatch in the affected SG [+0.5]

b. 1. limit the spread of contamination [+0.5]

2. allow maintaining ruptured SG pressure above the pressure of the intact SG's, as a prerequisite for' stopping break flow [+0.5]

Answer key should also include Boron Sample of S/G water. (FF/SF mismatch is valid prior to Rx trip or Feedwater Isolation, and question implies Rx trip and Ss has already occurred.)

REFERENCE:

OP-MC-EP-EP4 Pages 18 of 55 23 of 55 1

___m. __ ._ _..___._m.m________ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ __ _

.?,

c - s. s u

r'

2. 3.08/5.13 Y

QUESTION 3iOB (1.00)

'WHICH;ONE (1) of.the-'following cause the CLA discharge-

' isolation valves to receive an open1 signal? (1.0) i~

~

(a.) asafety.knjectionsignal

-(b.) anLautomatic diesel' generator start.

(c.)' a' blackout signal (dj) NC pressure increasing above 1955 psig-ANSWER 3'.08 (1.00)

. (a.) [+1.0]

" Accept A or D as correct REASON:- Cold leg accumulator valves will auto open when NC system pressure is > P-11 (2/3 NCS >

1955'psig)

~

REFERENCE:

OP-MC-ECC-CLA.

Pages 9 of 15

~ 10 of 15

l c

p Enclosure 4 NRC Resolution of Facility Comments R0/SR0 Examinations Question 2.11a/5.10a:

NRC Resolution: Facility comment noted. . Since the stem of the question states a reactor trip has already occurred, the additional response will be added to the answer key. However, Feed Flow / Steam Flow mismatch will be accepted only if the candidate states prior to the reactor trip.

Question 3.08/*

t.

NRC Resolution: ' Facility comment accepted. Since both choices (a) and (d)

( are correct, the answer key will be modified to accept either for full credit.

• This. question did nct appear on the administered SRO examination.

l f

(

L i

g ,

W" '

y p, 1

ENCLOSURE-5

'SIl10LATION' FACILITY REPORT-

Facility Licensee

' Duke Power' Company Facility: Docket Nos.: :50-369'and 50-370 Operating T.ests Administered.On: June 27 - 29, 1989 Thisiform is to be used.only to report observations. These observations do not

? constitute audit or inspection findings and are not, without further-verifica-tion: and review,11ndicative of. _ noncompliance ' with 10 ' CFR 55.45(b). These observations do '.not -affectL NRC certification 'or. approval ~of the simulation facility ;other . than to ' provide' information which may- be used inL future

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

~ During the conduct of the simulator examinations, the following items -were observed.

Pressurizer pressure control is- more sensitive than in the reference plant.

Low. pressure alarms are received on minor RCS temperature changes which.could indicate unrealistic changes in Tave. This distracted the candidates and impacted; examination administration.

Dropping sny control rod or spider assembly, other than the center control rod

-(H-8) results in a reactor trip regardless of reactor power 11evel.. This impacted examination development.

Malfunction CF6, Feedwater Line Break Outside Containment Downstream of rheck Valve, '. states a Reactor Trip will occur on Lo-Lo Level and a Feedwater Isolation will occur on Hi-Hi doghouse level. 'When _this malfunction was inserted, the affected steam generator blew dry and dog house level increased.

The . facility P& ids show the steam generator should not have blown dry unless the . check valve inside containment also failed' open. The malfunction description does not state this effect. Additionally, the status description states thisr is a break inside containment. No increase in containment pressure,- sump levels, or humidity was observed. This impacted examination development.

When a primary leak is' introduced, radiation monitors either did not respond or responded very slowly. Wb',n a Letdown Heat Exchanger tube leak into KC was

inserted, the simulator operator .had to manually increase the associated radiation monitor level until it alarmed. This impacted examination administra-tion.. Also, the radiation monitors for steam generator blowdown, EMF-14 HI

- and EMF-34 LO are reversed causing the HI range to alarm before the LG range indicates any increan in activity. This impacted examination development.

I Enclosure 5 2 During a Loss of All AC, all four PCBs opened when only the two PCBs intering the plant should open. This resulted in the candidates not being able to use approved plant procedures to restore power and impacted examination administration.

During a loss of All AC following reset of DG1B, DG1B closed on the dead emergency bus and began to sequence loads. DG1B then tripped ft.r no apparent reason. DG1B continued start attempts even after the sequencer wcs reset and the emergency bus energized from offsite power. This impacted exenination development.

During recovery from a Loss of All AC, power was restored to one of the J incoming off-site buslines. Power was restored to the DRPI display and the feedwater control section before the PCBs were closed and the in-plent .

distribution busses energized. This impacted examination administration.

Additionally, when the PCBs were closed and power restored to plant equipment, pressurizer level instantly increased to greater than 100 percent and NC system pressure increased to greater than 3000 psig. This forced early termination of the examination and impacted examination administration.

The VCT level controller is simulated as a direct acting controller but is an inverse acting controller in the reference plant. This confused the candidates when manual control was required and impacted examination administration.

The Rod Bank Lo Limit alarm was received approximately 50 steps above the RIL.

According to the facility PLS, the setpoint for this alarm is ten steps above the RIL. The candidates recognized this was incorrect and became confused by the alarm. The required action for this alarm is to borate until the alarm clears. This resulted in power decreases occurring almost completely by boration and impacted examination administration.

When NC pump #1 seal was failed and the leakoff isolation valve shut, indicated seal dP rapidly oscillated between 250 psid and 150 psid. This impacted examination development. As a result, the candidates tripped the NC pump in accordance with reference plant procedure which requires tripping of an NC pump whenever seal dP is less than 200 psid. This impacted examination administration.

During the NV suction header leak malfunction, charging flow would slowly increase to approximately 100 gpm then rapidly decrease to about 50 gpm on a cyclic basis. Seal injection flow was affected, but pressurizer level remained constant. This rendered the malfunction unusable and impacted examination development.

Miscellaneous HVAC alarms (VC/YC) would alarm, then clear, on a ten-minute cycle.

The H2-KG Panel Trouble alarm would also alarm intermittently. These alarms distracted the candidates and impacted examination administration.

_ _ - - _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ -