ML20246C903

Forwards Addl Comments Re Senior Reactor Operator Exams. Letdown Backpressure Control Continues to Have Large Fluctuations Producing Alarms Which Distract Operators While Trying to Operate or Control Plant

ML20246C903
Person / Time
Site:	Arkansas Nuclear
Issue date:	08/02/1989
From:	Force E ARKANSAS POWER & LIGHT CO.
To:	Mccrory S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
Shared Package
ML20245L217	List: ML20245L211 ML20245L222 ML20246C903
References
RER-89-02173, RER-89-2173, NUDOCS 8908250174
Download: ML20246C903 (7)

Text

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ARKANSAS POWER & LIGHT COMPANY Reeves E. Ritchie Nuclear Training Center Arkansas Nuclear One Rt. 3, Box 137G Russellville, Arkansas 72801 August 2, 1989 l

4 RER-89-02173 l

Mr. Steve McCrory U.S. Nuclear Regulatory Commission Region IV j

611 Ryan Plaza Drive l

Suite 1000 j

Arlington, TX 76011 i

SUBJECT:

Arkansas Nuclear One Docket No. 50-368 License No. NPF-6 Comments on NRC SR0 Exam of August 1, 1989

Dear Mr. McCrory:

In a.ncordance with NUREG-1021, I have attached additional comments on.

the SRO examinations for your review and consideration.

If you have any questions, please contact me.

Sincerely, Ed Yorce Manager Training EAF: PDC:lah Attachment ec: ANO-DCC

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8908250174 890814 PDR ADOCK 05000368 V

PDR l

L MEMBEA MiOOLE SOutw utsutiFs system

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4-Attachment-to:

RER-89-02173 Page 2 In addition to the pre-examination review, the following examination answers require correction:

QUESTION 6.32 (3.00)

Using the extract from EPIP 1903.10 provided, classify from column 2, as applicable, each of the conditions or events.in column 1.

If you feel more than one classification is possible for any. condition / event, select.a single classification and justify your answer.

EVENT / CONDITION CLASS a.

Both shutdown cooling trains become 1.

NOUE inoperable but at least one can be 2.

ALERT restored within 45 minutes.

3.

SAE 4.

GE 5.

None-b.

Control room evacuated and plant shutdown in progress locally.

c.

Area radiation monitors in Containment reading 2500 mR/hr 90 minutes after plant trip following full power opreations for 90 days.

d.

Site experiencing sustained winds of 80 mph.

e.

Steam line break outside containment (nonisolable) with a 12 gpm primary to secondary tube leak.

f.

' Dose equivalent I-131 in the RCS is 350 pCi/gm, and there have been no power or thermal transients in the plant.

g.

All annunciator power is lost concurrent with a generator trip.

h.

RCS leak. age of 75 gpm.

ANSWER 6.32 (3.00) a.

5. none b.

2. ALERT (6.8) c.

5.

None d.

3. SAE (8.3) e.

2. ALERT (3.3) f.

1. NOUE (1.2) 3

5. none h.

2. ALERT (2.2)

(0.375 EA)

7 Attachment to:

l

' RER-89-02173 Page.3 REFERENCE ANO EPIP 1903.10, REV 25, pgs15-125 194001A116

..(KA's)

Revision 26 of procedure 1903.10 was supplied to the candidates. The question was written from Rev. 25.

The revision changed the correct answer for item "C" of the exam. The correct answer should be "5.

None" since the event / condition did not specify a loss of reactor coolant. The revised page is attached to this package.

QUESTION 6.25 (1.00)

Fill in the blanks in the following statement concerning the containment Combustible Gas Control System.

A containment dome sample may be taken by placing the hydrogen purge manifold selector switch to position 6 and _(a)_ with handswitches on Panel 2C33.

The hydrogen analyzers are placed in service after a(n) _ b)_ event to

(

monitor any change in hydrogen gas concentration inside the Containment Building.

Each hydrogen recombination unit is capable of maintaining containment

_ c)_ volume percent following any

(

hydrogen concentration below postulated _(d).

ANSWER 6.25 (1.00) a.

opening dome sample valves (2CV-8351 & 2SV-8341) b.

LOCA c.

3.5 or 3.8 d.

LOCA (0.25 each)

REFERENCE AN02 STM 2-06, pgs 3-6 028000K601 028000G015 028000A403

..(KA's)

STM 2-06 page I states "The system is designed to limit the hydrogen concentrations to 3.8% during any postulated Loss of Coolant Accident (LOCA)".

Page 5 of the same STM lists 3.5 volume percent.

.7

)

Attachment to:

RER-89-02173 Page 4 4

Recommend you accept either value due to their close proximity and inconsistency of reference material. This was overlooked in the i

initial review.

QUESTION 5.23 (1.00)

Which one of the radiation monitors below is NOT required by Technical Specifications to be operable during fuel movement in the spent fuel pool?

a.

Containment Purge & Exhaust Isolation Monitor b.

Spent Fuel Pool Area Monitor c.

Control Room Ventilation Intake Duct Monitor d.

RCS Gaseous Activity Leakage Detection.

ANSWER 5.23 (1.00) d REFERENCE AN02 TS Table 3.3-6 000036K202

..(KA's)

This question assumes but does not state that the plant is in mode 6.

During the initial review the mode 6 assumption was made, however, fuel may be moved in the spent fuel pool during mode 1 and Technical Specifications require RCS Gaseous Activity Leakage Detection.

Since there is no right answer, recommend elimination of question.

QUESTION 6.30 (1.00)

Choose the statement below which best represents the definitions, conditions, or requirements found in the ANO ALARA Manual, OPAP 1000.33.

a.

Judgement as to what is ALARA is based upon the benefits from the improvements and regulatory requirements.

b.

The goal of the ALARA program is mainly to maintain the individual exposure of ANO and contractor personnel "as low as reasonably achievable".

l

y

',_ U$l t L g) 7'V ' S -,-

Attechment to:

m (RER-89-02173' Page 5-r g..y 4

c.

Jobs are classified into one of four categories for ALARA consideration on the basis of radiation levels at the work site.

d.

Individual ALARA reviews are not required for jobs of a

-repetitive nature.and similar radiological conditions.if a Standing ALARA: Review exists.

ANSWER-6.30.(1.00) d.

REFERENCE

'ANO OPAP 1000.33, Rev. 5, pgs 3, 8, 10, 13 194001K104

..(KA's)

.Although.this question is. technically correct, it is asking the-examinee to recognize that "mainly" was added and " annual exposure" and "both collective" were deleted from the sentence. Recommend' adding.

"b" as~a correct answer.

.After careful review of the examination by the training staff, the following-.two questions are provided as" examples of questions that we

' feel go beyond the scope of SRO required recall. We are not asking

.for'any. change in this exam but request:that this type of question does'not. appear'on future exams.

5.11 (1.00)

QUESTION 7 Which one of the following red light alarms on panels 2C343-1 thru 4 is positive indication that fire suppression media is being admitted to the affected space or equipment assuming proper system response or

. actuation?

a.

1-1-4(B): Cable spreading room b.

1-2-4(T): EDG room #2 c.

3-3-4(T): TG bearing and lube oil piping d.

3-3-4(B): Generator exciter ANSWER 5.11 (1.00)

D.

l O

Attachment to:

RER-89-02173 Page 6 l

1 l

t REFERENCE AN02'AOP-2203.09 REV 8, STM 2-60 000067A106

..(KA's) i This question asks for memory knowledge of discrete sections of f

a sixty-seven page procedure which has no immediate actions.

j i

QUESTION 6.33 (1.50) 1 Mat.ch the badge background colors in the right column to the type of individual desiring controlled access entry in the left column. Base your answer on the type of access generally granted or required by the individual. The numbered items in the right column may be ased more than once or not at all.

a.

General visitor 1.

White b.

AP&L corporate personnel 2.

Blue c.

NRC Commissioner 3.

Yellow d.

INPO personnel 4.

Green e.

Plant administrative personnel 5.

Red f.

Contract personnel supporting an outage ANSWER 6.33 (1.50,1 a.

5 d.

3 b.

3 e.

I c.

5 f.

3 (0.25 ea)

Although badge color is important from the security classification standpoint, the SRO's primary concern is whether the badge allows escorted or unescorted access.

o-

,d SIMULATION FACILITY FIDELITY REPORT facility Licensee: Arkansas Power'& Light Company.

Facility Licensee Docket No.:

50-368 s

Facility License No.:. NPF-6 Operating Tests Administered at: ' Arkansas Nuclear One Unit 2 Operating. Tests given on: August 1-3, 1989 During the conduct of the simulator portion of the operating tests identified' above, the following apparent performance and/or human factors discrepancies-were observed:

1.

Letdown backpressure control continues to'have large fluctuations (not L

representative of actual plant response) producing alarms which distract operators while trying to operate or control the plant.

2.

.The. plant experiences'high temperature alarms in certain switchgear rooms regularly. This condition is inconsistently modeled in the simulator.

Starting a simulation-from a high power initial condition resulted in these alarms within. fifteen minutes. However, starting from a lcw power initial condition did not result in these alarms actuating during a'l-hour simulation.- According to the simulator instructors, the alarms would not actuate for several hours'when a simulation is begun from'certain initial conditions, but that the occurrence in the plant is not power level dependent.,

'3.

The simulator " locked up" toward the end of a simulation. whi-segan at about 63~ percent power, end-of-life conditions and progressed through a steam generator tube rupture with loss of condenser vacuum event. The cause 'of the " lock up" was not readily evident.. However,;it occurred far enough into the simulation to allow an adequate evaluation of the applicant with a few followup questions without needing to reset the simulator.

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