ML20140C330

Insp Rept 50-285/97-11 on 970519-23.Violations Noted. Major Areas Inspected:Implementation & Maint of TS Program & Events Associated W/Control Rod Withdrawal to Bring Reactor to Critical Condition on 970512

ML20140C330
Person / Time
Site:	Fort Calhoun
Issue date:	06/04/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20140C322	List: ML20140C318 ML20140C324 ML20140C330
References
50-285-97-11, NUDOCS 9706090216
Download: ML20140C330 (9)

See also: IR 05000285/1997011

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ENCLOSURE 2

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket No: 50-285

License No: DPR-40

Report No: 50-285/97-11

Licensee: Omaha Public Power District l

Fort Calhoun Station FC-2-4 Adm.

P.O. Box 399, Hwy, 75 - North of Fort Calhoun

Fort Calhoun, Nebraska

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Facility: Fort Calhoun Station

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Location: Blair, Nebraska '

Dates: May 19-23,1997 i

Inspector: W. Walker, Senior Resident inspector

Approved: W. D. Johnson, Chief, Project Branch B  !

Attachment: Supplemental Information l

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9706090216 970604

PDR ADOCK 05000285

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EXECUTIVE SUMMARY t

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Fort Calhoun Station

NRC Inspection Report 50-285/97-11 ,

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' This special announced inspection included aspects of the implementation and maintenance  !

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! of your Technical Specification program. Specifically, the inspection focused on the events

- associated with the corcrol rod withdrawal to bring the reactor to a critical condition on .

May 12,1997.  !

Operation's

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-* The inspector concluded that there was a weakness in the operating crew briefing

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prior to the approach to criticality in that information concerning the potential to ,

have control rods fully withdrawn without being critical was not discussed

(Section 01.1.1). ,

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• - Operating Procedure OP-2A had a weakness in that it did not provide guidance on I

actions to take if the reactor was not critical with all control rods fully withdrawn -

(Section 01.1.2).

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• The reactor was maintained in a safe condition, but operators delayed driving in.  ;

Group 4 control rods while discussing the situation of having all rods fully l

l withdrawn without having reached criticality (Section 01.1.2). l

• A noncited, minor violation was identified for an inadequate plant startup procedure.

The procedure did not provide operator instruction for addressing a noncritical

reactor condition with all Group 4 rods fully. withdrawn (Section 01.1.2).

• The licensed operators failed to follow procedures when they did not change the

plant startup procedure to document the necessary acijustments made to the boron

concentration and control rods to achieve criticality. This was a violation of

Technical Specification 5.8.1 (Section 01.1.2).  ;

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Rep _ ort Details

Backaround

On April 21,1997, the reactor was manually tripped due to the rupture of a 12-inch

extraction steam line from the high pressure turbine to a low pressure feedwater heater.

On May 12 during the starmp from the forced outage, the licensee performed steps to

bring the reactor critical. On the morning of May 12, the inspector reviewed the control

room log book and noted that reactor criticality was not reacned with all rods fully

withdrawn. The estimated critical condition as predicted by the licensee was Group 4

control rods withdrawn 85 inches. The inspector questioned the licensee concerning the

sequence of events which had occurred during the approach to criticality.

I. Ooerations

01 Conduct of Operations (92901)

01.1 Reactor Criticality Controls

a. Insoection Scoce (92901) i

The inspector reviewed the circumstances and operator actions associated with the

May 12,1997, approach to criticality.

b. Observations and Findinas

On May 12,1997, at approximately 4 a.m. during performance of activities to bring

the reactor to a critical condition, a determination was made that tne reactor was 1

not critical when all rods were fully withdrawn. The estimated crit. cal position as i

calculated was that the reactor would be critical with Group 4 control rods ]

withdrawn 85 inches. l

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The following is a sequence of events for the approach to criticality on May 12,

1997.

Time Description

0249 Withdrew Group 1 control rods to 90 inches (1/M prediction

did not indicate criticality within the next withdrawal)

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0302 Withdrew Group 2 control rods to 90 inches (1/M prediction of

criticality between Group 3 control rods at 90 inches

withdrawn and Group 4 at 90 inches withdrawn)

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0317 Withdrew Group 3 control rods to 90 inches (1/M predicted

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criticality beyond all rods out)

. 0330 (approx) Withdrew Group 4 control rods to achieve criticality. The

l licensee discussed the potential to reach all rods out on

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Group 4 control rods and not be critical. A' decision was made

to continue based on knowledge that the rod worth curve was

not exact and, if under-predicted, could allow criticality prior to

all rods out. 3

0332 (approx) Group 4 control rods withdrawn to 110 inches, start up rate

and power were monitored. Group 4 control rods were

withdrawn in 2-inch increments while monitoring indications.

Behavior was indicative of being close to critical (power at

1.OE-05 percent and very slowly increasing, but startup rate

was not a constant positive value).

0347 Withdrew Group 4 control rods to all rods out position.

Behavior was indicative of being close to critical (power at

1.OE-05 percent and very slowly increasing, but startup rate

was not a constant positive value).

0400 (approx) The shift supervisor, reactor engineer, and licensed senior

operator discussed reactor indications and confirmed that l

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criticality was not achieved as indicated by:

• No constant positive startup rate (indications still read

negative occasionally).

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• Power was not steadily increersing'without control rod  !

motion. _

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• Power was less than 1.0E-04 percent.  !

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At this point, discussions on how to proceed involved:  ;

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• How much to dilute: this should be enough to bring  ;

Group 4 control rods to 85 inches withdrawn, which l

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was .26 percent delta-rho or 30 ppm. This was derived

from the Technical Data Book Figure ll.B.2.b. This was

the target estimated critical position.

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• How f ar to insert Group 4 control rods: this should be

l approximately twice the added reactivity of the dilution i

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'or 40 inches withdrawn. This was also derived from j

j Technical Data Book Figure ll.B.2.b. I

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• Calculating the amount to dilute (850 gallons water). l

The licensee used Operating Instruction Ol-CH-4,

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" Chemical and Volume Control System Makeup I

Operations." '

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• The need for a new estimated critical condition: This  :

-was not necessary since the calculation would result in l

the same value as before of 1395 ppm and Group 4 j

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control rods at 85 inches withdrawn.

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0405 Inserted Group 4 control rods to 40 inches withdrawn.  !

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0410(approx) _ Began dilution addition of 850 gallons of water. l

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0450(approx) Chemistry sample determined new boron concentration. i

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0500(approx) Restarted approach to criticality. New base count for 1/M plot  :

was taken for Group 4 control rods at 40 inches withdrawn. l

Withdrew Group 4 control rods to 65 inches withdrawn. j

0515(approx) 1/M predicted criticality between 85 and 100 inches

withdrawn. Withdrew Group 4 control rods to achieve i

criticality,

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0524 Criticality was achieved with Group 4 control rods withdrawn

94.5 inches. Comparison of estimated critical condition and

actual critical condition is as follows

• Total deviation between predicted and actual critical I

condition = 0.35 percent delta-rho I

• This was less than administrative review limit

(0.5 percent delta-rho) and less than the Technical l

Specification limit (1.0 percent delta-rho)

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01.1.1 Ooerator Performance Issues Asso_giated with the Acoroach to Criticality

a. Insoection Scope (92901)

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The inspector reviewed the estimated critical condition calculation and the approach 1

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to criticality briefing. In addition, interviews were conducted with the shift

. supervisor, the licensed senior operator, and the reactor engineer who provided

direct oversight of bringing the reactor to a critical condition,

b. Observations and Findinos

The inspector performed a review of the estimated critical condition calculation and

verified that the calculation as performed indicated that the reactor would reach

criticality with Group 4 control rods withdrawn 85 inches. During the review of the

estimated critical condition worksheet, the inspector noted that the potential

existed to reach all rods out without the reactor being in a critical condition. The

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inspector oubMioned the shift supervisor and the reactor engineer concerning

'r whether daring the approach to criticality briefings this potential for reaching all

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rods out whhout being critical was discussed. The reactor engineer and the shilt

!- supervisor stated that this was not discussed; however, the reactor engineer stated

he was aware of the possibility. The inspector considered it a weakness that the

operating crew and especially the licensed senior operator were not reminded that

the possibility existed to be in an all rods out condition without being critical.

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The inspector discussed with the reactor engineer the count rate predictions

obtained when Group 3 control rods were at 90 inches which indicated that the

reactor would go critical beyond all rods out on Group 4. The reactor engineer

stated that discussions were held with the operating crew regarding this and a

decision was made to continue. This decision was based on the knowledge that

the rod worth curve was not exact and criticality could be achieved prior to

reaching all rods out on Group 4.

The inspector reviewed the estimated critical condition worksheets from the

previous three approaches to criticality and no anomalies were noted,

c. Conclusions

The inspector concluded that there was a weakness in the operating crew briefing  ;

prior to the approach _to criticality in that information concerning the potential to

reach all rods out on Group 4 control rods without being critical was not discussed.

01.1.2 Procedure Usaae Durina Rod Withdrawal to Criticality

a. Insoection Scoce (92901_1

The inspector reviewed Operating Procedure OP-2A, " Plant Startup," and Standing

Order S0-0-1, " Conduct of Operations." The review was conducted to evaluate l

operator performance and adherence to procedures during the rod withdrawal to

criticality.

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b. Observations and Findinas

! The inspector performed a review of the Operating Procedure OP-2A, " Plant

Startup." Attachment 2, "CEA Withdraw to Criticality Mode 2," provides i

! instructions on performing the rod withdrawal to criticality. Step 10e states,

L " withdraw Group 3 control rods to 90 inches, and verify Group 2 all rods out and

l Group 4 at approximately 14 inches." Step f states, " wait 5 minutes while

l monitoring count rate." Step g states, " withdraw Group 4 as required to achieve

l- criticality." The next step is number 11 which states, "when reactor power is

greater than 1.0E-4 on all wide range nuclear instrumentation channels then place

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each zero power mode bypass switch on the reactor protection system cabinets to

off." The inspector found that the procedure did not address the reactor condition

encountered, that is, not critical with all Group 4 rods fully withdrawn. This was a

minor violation per Section IV of the NRC Enforcement Policy and is described in

- this report because of its impact on the cited violation below. This violation

constitutes a violation of minor significance and is being treated as a noncited

violation consistent with Section IV of the NRC Enforcement Policy.

The inspector also performed a review of Standing Order SO-O-1," Conduct of

Operations." Step 7.3.1 states "it is the responsibility of the on-duty shift

supervisor to direct all scheduled and planned reactor power changes in accordance

with approved procedures." Step 7.3.2C states "all scheduled or planned power

changes will be made in accordance with the applicable operating procedure and

operating instructions." Additional guidance is provided in the Standing Order under

the Procedure Adherence section (12.1.2). Section 12.1.2B states in part that, if

while performing a procedure it is discovered that the anticipated response was not

received, the following actions should be carried out:

• Place the system / component in a safe condition.

• Contact the shift supervisor and inform him of the situation and status of the

component / system.

Evaluate the situation to determine the cause of the unexpected response

and initiate a temporary or permanent procedure change in accordance with

G-30, " Procedure Change And Generation," to allow usage of the procedure

for the current situation.

The inspector determined that, when the licensee failed to reach criticality with all

Group 4 rods out, the anticipated response (i.e., criticality) was not received. The

failure to immediately place the reactor in a safe, stable condition and then to

initiate a procedure change before continuing to perform reactivity adjustments was

a violation of Technical Specification 5.8.1 (50-285/97011-01).

c. Conclusions j

Operating Procedure OP-2A had a weakness in that it did not provide guidance on

actions to take if the reactor was not critical with all control rods fully withdrawn.

The inspector determined that the licensed operator actions to drive Group 4 control

rods back into the core when all rods were out and criticality had not been reached

l were technically sound and maintained the reactor in a safe condition. However,

the operators delayed driving in Group 4 rods from 3:47 a.m. until 4:05 a.m. while

a discussing the situation. The operators violated procedures when they failed to

stop and change the procedure to document the necessary adjustments made to the  ;

boron concentration and control rods to achieve criticality,

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II.Manaaement Meetinas

X1 Exit Meeting Summary

The inspector presented the inspection results to members of licensee management

at the conclusion of the inspection on May 23,1997.

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ATTACHMENT

SUPPLEMENTAL INFORMATION

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PARTIAL LIST OF PERSONS CONTACTED

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' Licensee -l

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J. Bishop, Assistant Plant Manager

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C. Br'u nnert, Manager, Quality Assurance and Quality Control

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D. Dryden, Station Licensing Engineer '

l- T2 Dukarski, Supervisor, System Chemistry

S. Gambhir, Division Manager, Production Engineering

J. Gasper, Manager, Nuclear Projects l

l- W. Gates, Vice President, Nuclear '{

S.'Gebers, Manager, Radiation Protection

B. Hansher, Supervisor, Station Licensing

R. Jaworski, Manager, Design Engineering, Nuclear

.R. Phelps, Manager, Station Engineering

R. Ridenoure, Supervisor, Operations

H. Sefick Manager, Security Services '

L C. Stafford, Principal Reactor Engineer

J. Tills, Manager, Nuclear Licensing -

D. Trausch, Manager, Nuclear Safety Review Group

NRC

-W. Walker, Senior Resident inspector

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INSPECTION PROCEDURES USED I

IP 92901: Followup - Operations

ITEMS OPENED AND CLOSED

Ooened

50-285/97011-01 vio failure to follow procedures during approach to criticality i

(Section 01.1.2)-  ;

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