ML17291B038

Insp Rept 50-397/95-12 on 950804-19.No Violations Noted. Major Areas Inspected:Eligibility Determination & Administration of Comprehensive Written Exams & Operating Tests

ML17291B038
Person / Time
Site:	Columbia
Issue date:	09/21/1995
From:	Pedlet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML17291B037	List: ML17291B036 ML17291B038
References
50-397-95-12, NUDOCS 9509270013
Download: ML17291B038 (19)

See also: IR 05000397/1995012

Text

ENCLOSURE

U.S.

NUCLEAR REGULATORY COHHISSION

REGION IV

Inspection Report:

50-397/95-12

License:

NPF-21

Licensee:

Washington Public Power

Supply System

3000 George Washington

Way

P.O.

Box 968,

HD 1023

Richland,

Washington

Facility Name:

Washington Nuclear Project-2

Inspection At:

Richland,

Washington

Inspection

Conducted:

August 4 - 19.

1995

Inspectors:

T. 0. HcKernon. Chief Examiner,

Operations

Branch

Division of Reactor Safety

S.

L. HcCrory. Examiner,

Operations

Branch

Division of Reactor Safety

R.

E. Lantz,

Examiner,

Operations

Branch

Division of Reactor Safety

Accompanying Personnel:

D. Draper,

Contractor

T. Bettendorf,

Contractor'.

Hitchell. Contractor

Approved

r

o

.

e

,

.

pera ions

rane

D

ision of

ea

or Safety

fd

ff

Ins ection

Summar

IRR

A~f

t d:

R tf

.

d f

P tf

f th

d lift tl

f

applicants

for operator licenses

at the Washington Nuclear Project-2 facility,

which included

an eligibility determination

and administration of

comprehensive written examinations

and operating tests.

The examination

team

also observed plant conditions incident to the conduct of the in-plant

applicant evaluations.

The examiners

used the guidance

provided in NUREG-1021,

"Operator

Licensing

Examiner Standards."

Revision 7, Supplement

1 including. Sections

201-303

'01-303

'01-403,

and 701,

issued

June

1994.

95092700i3

9509'2i

PDR

ADQCK 05000397

6

PDR

0

-2-

Results:

~Ocr ations

All eight of the applicants for senior reactor operator

licenses,

and

five of the six applicants for reactor

operator licenses satisfied the

requirements

of 10 CFR 33(a)(2)

(Sections

1.2 - 1.3).

The reference material provided by the training department for

examination

development

was adequate with some noted minor exceptions

(Section 1.1).

All applicants

except

one reactor operator applicant passed

the written

examinations.

with scores

ranging from a low of 77.3 percent to a high

of 95.9 percent,

with averages

of 84.2 percent for reactor

operator

applicants,

94.3 percent for senior reactor

operator

upgrade applicants,

86.9 percent for senior

reactor operator instant applicants.

and 86.5

percent overall (Section 1.2).

The examiners

observed overall good communication practices

by the crews

during performance of dynamic simulator evaluations;

however,

command

and control effectiveness

between

crews

was inconsistent

(Section 1.3.1).

A number of instances

of a lack of simulator fidelity issues

were

identified,

none of which impacted examination validity (Section 1.5).

Plant

Su

ort

The material condition of the plant observed during in-plant walk

through evaluations

was good (Section 1.4).

Summar

of Ins ection Findin s:

~

There were no findings identified during the course of this inspection

that were assigned

a tracking number.

Attachments:

Attachment

1 - Persons

Contacted

and Exit Heeting

Attachment

2 - Post-Examination Facility Comments/NRC

Response

Attachment

3 - Simulation Faci lity Report

Attachment

4 - Written Examination

and Answer

Keys

-3-

DETAILS

1

LICENSED OPERATOR APPLICANT INITIALQUALIFICATION EVALUATION (NUREG-1021)

During the inspection.

the examiners

evaluated

the qualifications of

14 license applicants:

6 for reactor

operator

(RO) and 8 for senior reactor

operator

(SRO), of which 2 were

ROs upgrading to SRO and

6 were unlicensed

SRO

applicants.

The inspection

assessed

the eligibility and administrative

and

technical

competency of the applicants to be issued licenses to operate

and

direct the operation of the reactivity controls at

a ccmmercial nuclear

power

facility in accordance

with 10 CFR Part 55 and

NUREG-1021,

"Operator License

Examiner

Standards,"

Revision 7. Supplement

1. Sections

200 (series).

300 (series),

and 400 (series).

Furthermore,

the inspection included

evaluations of facility materials,

procedures.

and simulation capability used

in development

and administration of the examinations.

These areas

were

evaluated

using the guidance provided in the areas of NUREG-1021 cited above.

Finally, the examiners

observed

plant conditions during the conduct of

in-plant applicant evaluations.

After completion of the evaluations,

the examiners

recoranended

that five of

the six applicants

for

RO licenses

and all of the eight applicants

for

SRO

licenses satisfied the requirements of 10 CFR 55.33 (a)(2).

1. 1

Facility Haterials Submitted for Examination Development

The chief examiner

reviewed the licensee's

materials

provided f'r development

of the examination,

which included station administrative

and operating

procedures.

lesson

plans,

question

banks,

simulator scenarios,

simulator

malfunctions,

and job performance

measures

(JPHs).

The procedures

and lesson

plans were adequate.

Some changes to the examination were made during the

validation period as

a result of not having the most current procedure

revisions available during the development

phase.

The facility banks of written questions.

dynamic simulator

scenarios,

and JPHs

were adequate

in scope,

depth.

and variety.

The written examination question

bank was used only to a limited extent in developing the written examination.

Questions

used

from the bank were rewritten to change either the question

stem

or distractors.

Additionally. during examination development, it was observed

that

a number of the JPHs were not current with recently revised procedures.

The inconsistencies

applicable to the JPHs

used during the operating

examination were resolved during the validation period.

1.2

Written Examinations

The chief examiner provided the facility training staff with a copy of the "as

administered" written examinations

and the answer

keys immediately following

the completion of the examination

by the applicants.

The facility reviewed

the as administered

examination

and submitted postexamination

comments for NRC

consideration.

The licensee

was also provided

a copy of the preexamination

review comments for the postexamination

review.

'he

licensee

submitted postexamination

comments

on seven questions

on the

RO

and

SRO written examinations,

enclosed

as Attachment 2.

The'hief examiner

analyzed the comments.

using supporting information supplied by the facility

licensee

and other available material,

and determined that for five questions.

the facility comments

were technically correct

and the actions

requested

by

the facility were in accordance with NUREG-1021, Revision 7, Supplement l.

Therefore. five facility comments

were accepted

as requested:

three of the

questions

were deleted

(RO 31/SRO 30;

RO 15/SRO 16;

RO 27/SRO 26),

and two

questions

(RO 46 and

RO 59) were retained with two correct

answers'he

other two facility comments

were evaluated

as described

below.

Question

RO 57/SRO 52 was changed to accept only answer

"b" as the correct

answer rather than accepting

both answers

"a" and "b".

The facility comment

made clear that "a" would be correct only for a special

case.

Since no

clarifying information for the special

case

was provided in the question

stem,

the only correct

answer

was "b".

Question

SRO 98,

requested

by the licensee f'r deletion.

was retained.

While

applicants

are not expected to memorize all procedures.

they are expected to

have

a basic understanding of the emergency

plan and knowledge of emergency

action level definitions.

Application of this knowledge was sufficient to

correctly answer this question, without reading the exact procedure.

Overall, the applicants

performed adequately

on the written examinations.

Scores

ranged

from a low of 77.3 percent to a high of 88.7 percent with an

average of 84.2 percent for reactor operator

applicants;

and

a low of

81.4 percent

and

a high of 95.9 percent for senior reactor

applicants with an

average of 88. 1 percent,

and

a 86.5 percent

average overall.

All applicants

except

one

RO applicant

passed

the written examination portion of the license

examinations.

The chief examiner reviewed applicant performance

on individual questions

and

observed that the following questions

were missed

by 50 percent or more of the

applicants

attempting to answer the question.

The questions

are referenced

by

examination level

and question

number.

Refer to the master examinations.

in

Attachment 4, for the complete question.

SRO examination:

5,

17,

32,

38,

58,

78

RO examination:

5.

14.

16.

33. 34, 37. 39, 42. 51, 64. 78.

94

The chief examiner concluded that no generic

knowledge weakness

areas

existed.

Therefore. this observation

was provided to the facility training staff for

consideration.

1.3

Operating Tests

The examiners

developed

comprehensive

operating tests in accordance with the

guidelines of NUREG-1021, Revision 7, Supplement

1, Section

301.

The

operating tests

consisted of three parts:

an administrative portion.

a dynamic

simulator scenario portion,

and

a control room/plant walkthrough portion.

The

examination

team reviewed

and validated the various portions of the operating

tests in the Region

IV office during the week of July 24.

1995.

and on site

July 31 through August 4.

1995.

The licensee's

personnel

under

a security

agreement

assisted

in the onsite validation.

The examination

team

administered the operating tests

during the week of August 14.

1995.

1.3.1

0 namic Simulator Scenarios

The examiners

evaluated five crews; three crews with two SRO Instants

and

one

RO. one crew with one

SRO Instant.

one

SRO Upgrade.

and one

RO;

and

one crew with two ROs and one

SRO Upgrade.

The applicants

were rotated

through positions of primary operator,

balance of plant operator,

and control

room supervisor using two or three scenarios

depending

upon the crew makeup.

The examiners

evaluated

the applicants'ompetencies

by comparing actual

performance during the scenarios

against

expected

performance in accordance

with the requirements

in NUREG-1021, Revision 7, Supplement

1. Section 303.

The examiners

noted good overall comnunication practices

among the operating

crews evaluated in the simulator.

However, in a number of instances

on one

crew,

an applicant

was observed

leaving the at-the-controls

control board area

to perform actions

on back panels without informing other

members of the crew

and without informing other crew members of the status of the panels

for which

he was responsible.

Two of the crews evaluated

had difficultywith one of the simulator

scenarios.

In one scenario.

the crew was directed to reduce

power

from 100 percent to

80 percent using the reactor recirculation system flow control valves.

Through the misactions of the primary reactor operator,

a reactor

level

transient

was initiated when the operator

lowered core flow too quickly.

As

a

result.

a high level reactor pressure

vessel alert alarm annunciated.

While

the operator

acknowledged the alarm appropriately,

he made

no effort to stop

his actions

and wait for the perturbation to stabilize but rather

continued to

decrease

core flow, further per turbating level.

A preplanned

Feedwater

Master

Controller fai lure was actuated

at the same time the operator

resumed

core

flow reductions.

causing the plant to trip on high reactor vessel

level

(54.5").

The operator

was unable to prevent the trip in that his actions

initiated the transient

and further exacerbated

the condition by continuing to

reduce core flow.

In the second

instance.

a partial failure of reactor vessel

level

instrumentation

caused

some confusion

among the crew.

The control

room

supervisor directed the

RO to recover

vessel

level using the high pressure

core spray

pump while monitoring level indication on an already identified as

-6-

failed wide range level instrument.

The control

room supervisor relied on the

crew's

response to recognize this error and control level using other

indications.

While adequate

subsequent

plant control

by the operators

was

observed in both of these scenarios,

these

areas

were discussed

with the

licensee

as potential training issues.

Additionally, the examiners

noted inconsistent

command

and control

effectiveness.

Some

SROs demonstrated

strong crew leadership attributes while

others

needed

more support from the other crew members to determine action

plan alternatives

and governing technical specification requirements.

While some weaknesses

were observed during the dynamic simulator evaluation,

all applicants

passed this portion of the examination.

1.3.2

Walkthrou h Examinations

The examination

team evaluated

each of the

RO and

SRO applicants

using system

oriented

JPMs related to job tasks within the scope of the potential duties

as

appropriate in accordance

with NUREG-1021, Revision 7, Supplement

1.

This

included nonlicensed

operator tasks outside the control

room and performance

of some tasks

on the simulator in the dynamic mode.

Each of the applicants

were requi red to access

radiologically controlled areas to complete

one or

more tasks.

In addition to the tasks,

the examiners

asked prescripted

questions

related to the task system.

The facility administrative procedures

and practices

were also examined using

JPMs or questions.

Overall, the performance of the applicants

on the JPMs

was adequate.

The

applicants

were aware of plant activities and the location of plant

components.

All applicants

passed

the walkthrough portion of the examination.

1.4

Observations

During the walkthrough portion of the examination,

the examiners

observed

good

material condition of the plant.

The examiners

observed that the licensee

had

taken additional

measures

to improve the overall housekeeping

in the plant and

has recently carpeted

the main control

room to help reduce the background

noise level.

1.5

Simulation Facilit

During the preparation

and conduct of the operating examinations,

the

examination

team observed

some discrepancies

in simulator fidelity, as

documented

in Attachment 3.

The observed

discrepancies

did not impact

examination validity.

ATTACHHENT 1

1

PERSONS

CONTACTED

1.1

Licensee

Personnel

"J. Baker, Director of Training

• 0 ~ Hueller. Initial License Class Coordinator

%. Schaeffer,

Operations Training Manager

• C. Schwarz.

Operations

Manager

• D. Swank, Licensing Manager

1.2

~IIRC

2

• T. HcKernon, Chief Examiner.

Operations

Branch. Division of Reactor Safety

• Denotes personnel

that attended

the exit meeting.

In addition to the personnel

listed above.

the inspectors

contacted other

personnel

during this inspection period.

2

EXIT HEETING

An exit meeting

was conducted

on August 19.

1995.

During this meeting. the

inspector

reviewed the scope

and findings of the report.

The licensee did not

express

a position on the findings documented in this report.

The licensee

did not identify as proprietary any information provided to,

or reviewed by,

the inspector.

ATTACHMENT 2

FACILITY POSTEXAMINATION COMMENTS/NRC RESPONSE

WASHINGTON PUBLIC POWER SUPPLY SYSTEM

8/19/1995

Question

C.

The affected annunciator should be deenergized

and placed

under a danger tag clearance order.

The affected annunciator should be deenergized

and placed

under a caution tag order.

A Problem Sticker (PS) should be placed adjacent to the

affected annunciator window.

A Component Status Change Order (CSCO) should be

initiated on the annunciator.

Reference:

PPM 1.3.1, Department Policies, Programs and

Practices,

Rev. 19, Section 4.16.2h., Page 45 of 67.

Question: RO f59

A faulty pressure switch has resulted in an erroneous instrument air

header low pressure alarm.

WHICH ONE (1) of the following

describes

a required action for this deficiency2

WNP-2 Comments/References/NRC

Response

Actions specified in both answers 'b'nd 'c're procedurally driven:

WNP-2 References:

ANSWER

'b'.

PPM 1.3.1, Rev 20, Section 4.16.4.f, p.47, allows spurious alarms to be

deactivated.

2.

PPM 1.3.8A, Rev 4, section 3.2.4, p.5, states that a caution tag should be

used to flag any annunciator taken out of service.

3.

PPM 1.3.9, Rev 16, Section 4.4.7.b, p.10, requires a caution tag for a

deactivated alarm unless a TMR is generated.

ANSWER

'c'.

PPM 1.3.1, Section 4.16.4.h, Rev 20, p. 47, states that a Problem Sticker

should be placed adjacent to alarm- caused by faulty equipment.

NRC Response:

Accept 'b'r 'c's a correct answer.

Recommendation

Accept 'b'nd 'c'

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0

Questton

a.

b.

C.

d.

Unusual Event

Alert

Site Area Emergency

General Emergency

References:

PPM 4.12.1.1 Control Room Evacuation and Remote Cooldown,

Rev. 22, Section 4.1, Step 8, Page 6 of 53.

QUESTION: SRO if98

A Halon System initiation has forced the Shift Manager to evacuate

the Control Room and perform a remote shutdown and cooldown.

WHICH ONE (I) of the following describes the emergency

classification associated

with this event:

B

WNP-2 Comments/References/NRC

Response

Operators arc expected to perform only Immediate Actions from memory.

Performance of Subsequent Actions or emergency classification is accomplished with

procedures available.

The classification of an Alert is directed in Step 10 of the

Subsequent

Operator Actions of the referenced procedure.

WNP-2 Reference:

1.

PPM 4.12.1.1, Control Room Evacuation And Remote Cooldown, Rev 23,

Section 4.1, Step 10), p.6,

Classify the emergency

as an ALERT and

initiate notifications."

NRC RESPONSE:

Deny.

While it is agreed that Step 10 of PPM 4.12.1.1

"Control Room Evacuation and Remote Cooldown

is a subsequent

action step, the

question tests the candidate's

understanding of the basis of the emergency

classification and understanding of the

Alert EALdefinition. An Alert

is

defined, in part,

... as an event(s) which involves substantial actuaVpotential

degradation of level of safety..."

Because control room evacuation is needed,

additional support, monitoring and direction through the TSC and/or Operations

Support Center would be needed.

The candidate's knowledge of the EAL definitions

alone permit answering the question.

An Unusual Event are those events involving a

potential degradation, while an Alert involves an actual/potential degradation of level

of safety of the plant.

Choices 'c'nd 'd'an be eliminated since no significant

component failures or core degradation

is indicated in the stem of the question.

Recommendation

Delete

References:

PPM 13.1.1A Revision 0, Attach 4.1, pg 94 of 153,

EAL Basis

DOE/RI 9442, Revision 0, Tbl 4-1,

Summary of Emergency Classes

Emergency Plan, Revision 15, Tbl 4-1, Category 7.2, "Emergency Classification

Intttatmg Conditions"

PPM 13.1.1, Revi ion 22, pg 4 of 33,

Emergency Plan Implementing Procedures

Question

CRD pumps

Drywell pressure

RPV level

Reactor pressure

Reactor Power

Both tripped

6 psig

45 inches

250 psig and stable using SRVs

5%

With Full Core Display indicating all HCU accumulators have

depressurized,

WHICH ONE (1) of the following methods can be

used to add negative reactivity?

Question RO //31, SRO NO

During full power operations,

a LOCA occurs accompanied by an

ATWS. The plant status is a follows:

B

WNP-2 Comments/References/NRC

Response

There are no correct selections for this question.

The RPV pressures

specified are

not suflicient to cause rod insertion.

WNP-2 References:

1.

PPM 5.5.11 (Alternate Control Rod Insertion) Tab 6, Page 17, requires a

RPV pressure of 400 psig or greater to perform overpiston vent.

2.

Training text 82-RSY4300-Tl (CRDM), states that ifreactor pressure

is

less than 375 psig, the rods do not go in at all without CRD accumulator

assistance.

NRC Response:

Accept; delete question.

Recommendatloa

Delete

Clear any hydraulic lock and open the scram inlet and

outlet valves for each rod that has failed to insert.

Vent the control rod drive mechanism overpiston area for

those rods that failed to insert.

Reset the scram and insert rods using the individual scram

test switches.

Close SRVs as necessary

and increase RPV pressure to

370 psig.

Reference:

l.

82-RSY4306-TI, Nov. 92, p. 11

2.

LO: 82-RSY4300-Ll CRDM, 6981

Qnestfon

Question: RO //15, SRO P16

WHICH ONE (I) of the following describes

the indicating light

status at panels P601 and F631 for a PARTIALLYCLOSED SRV

that had opened on high RPV pressure?

Panel 601

Panel P631

a.

Red light ON

Red light ON

Green light OFF

Green light ON

b.

Red light OFF

Red light ON

Green light OFF

Green light OFF

WNP-2 Comments/References/NRC

Response

D

There are no correct selections for this question.

The indication on P601 is as stated in answer 'd':

Red light ON, Green light ON.

The indication on F631 is determined by the status of the ADS initiation signal, i.e.,

whether or not the associated ADS solenoid is energized.

For the conditions given,

the P631 indications willbe Red light OFF and Green light ON.

Reference:

Print EWD-IE433, Rev. 14

NRC Response:

Accept; delete question.

Recommendatlon

Delete

C.

Red light OFF

Green light ON

Red light OFF

Green light ON

d.

Red light ON

Green light ON

Red Light OFF

Green light OFF

Reference:

1.

82-RSY4)100-T4, March 95, p. 12 &, 13

2.

LO: 82-RSY4100-L4 NBI, 5528, p.53

Question: RO //57, SRO P52

A spurious half scram has occurred.

It is just past midnight and the

shift engineer can not determine the cause of the half scram.

WHICH ONE (I) of the following is the LOWEST level of

authority, by title, that can authorize the half scram to bc reset

under these conditions?

Prior to resetting any half scram signal, the

Control Room Staff should investigate

and correct the cause.

In a case where the cause is indeterminable, the Shift

Manager would normally be a part of this process.

The procedure statement to reset at the direction of the

CRS/SM

allows for the

reset to still occur in the event that the Shift Manager is not available.

The CRS

would not give the direction to reset a I/2 scram of indeterminable cause ifthe Shift

Manager were present in the Control Room.

Accept 'a'nd

'b'.

b.

C.

d.

Control Room Supervisor

Shift Manager

Operations Manager

Plant Manager

Unless it is assumed that the Shift Manager is not available, selecting

CRS'n

answer to this question would be counter to Control Room practice and the intent of

the referenced procedure.

Reference:

1.

PPM 1.3.1, Department Policies, Programs and Practices, Rev.

19, Section 4.6.2.a, Page

19 or 67.

WNP-2 Reference:

PPM 1.3.1, Section 4.7.6, Rev. 20

NRC Response:

Accept only 'b's the correct answer.

Insufficient information in

stem of question to allow 'a's the correct answer.

Question

Question: RO ¹27, SRO ¹26

The control room supervisor directs the reactor operator to insert a

rod from P603 using the insert pushbutton.

WHICH ONE (I) of

the following describes the correct Directional Control Valve (DCV)

position while the rod is in motion?

WNP-2 Comments/References/NRC

Response

The question indicates that a rod willbe inserted using the Insert Pushbutton,

then

asks for thecorrect

DCV position while the rod is inmotion.

[underline added]

As long as the Insert Pushbutton is depressed,

DCV 121 and 123 willbe open, DCV

120 and 122 willbe closed.

This was not an available selection.

Recommendation

Delete

Under Piston

DCV 120

Over Piston

DCV 121

Withdraw

DCV 122

Reference:

I. 82-RSY4507-TI, Jan. 82, p.12

2.

LO: 82-RSY4300-LI CRDM, 5214

a. OPEN

CLOSED

CLOSED

b. OPEN

CLOSED

OPEN

c. CLOSED

OPEN

OPEN

d. CLOSED

CLOSED

OPEN

Insert

DCV 123

CLOSED

CLOSED

OPEN

CLOSED

In asking for the DCV lineup the question is, in effect, asking for the lineup during

the insert stroke of rod movement rather than the ancillary lineup for the senle

function. To be technically correct, the question should be worded to ask for "a

correct

DCV lineup during the rod movement.

A prewxam comment to specify rod

settle" motion to clarify the question was not

incorporated.

WNP-2 Reference:

1. Prewxam Review Comments, RO Question ¹26[SRO ¹27] mark-up.

NRC Response:

Accept; delete question.

ATTACHHENT 3

SIHULATION FACILITY REPORT

Facility Licensee:

Washington Public, Power Supply System

Facility Docket No.:

50-397

Operating Tests Administered on:

August 14-19,

1995

These observations

do not constitute audit or inspection findings and are not,

without further 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 future evaluations.

No licensee action is required in response

to these observations.

While conducting the simulator portion of the

operating tests,

the following discrepancies

were observed:

ITEH

DESCRIPTION

Reactor Trip

During an

ATWS scenario.

with no apparent

(Reactor Protection

System)

actions

from the operators.

the reactor

scrammed

from 20 percent

power .

Reactor Building Radiation

Honitoring Instrumentation

Emergency Diesel Generators

Reactor Recirculation

Pumps

Reactor

Building Radiation monitoring

instrumentation did not accurately

mimic

simulator

computer inputs.

RISC-4.6.8 meters

did not accurately indicate 10'r /hr when

simulator computer inputs were activated at that

radiation level.

When taking the EDG's

mode switches to Control

Room position, the loss of 125V dc and 250

V dc

power annunciators

alarm.

(This item was

observed during the validation week.)

During a scenario in which it was desired to

trip the "B" RRC pump, the simulator tripped

both "A" and "B" RRC pumps

when the "B" RRC pump

trip was actuated.

ATTACHHENT 4

WRITTEN EXAHINATIONS AND ANSWER KEYS

t'