Text

Safety Insp Rept 50-454/85-42 on 850812-1018.Violation Noted:Improper Operation of Eccs,Failure to Follow Tech Spec Requirements & Reactor Core Licensed Thermal Power Rating Exceeded
	ML20138R802
Person / Time
Site:	Byron
Issue date:	11/13/1985
From:	Warnick R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20138R795	List: ML20138R794; ML20138R802;
References
	50-454-85-42, NUDOCS 8511190053
	Download: ML20138R802 (12)
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U.S. NUCLEAR REGULATORY COMISSION

REGION III

Report No. 50-454/85042(DRP)

Docket No. 50-454 License No. NPF-37

Licensee: Commonwealth Edison Company

Post Office Box 767

Chicago, IL 60690

Facility Name: Byron Station, Unit 1

Inspection At: Byron Station, Byron, IL

Inspection Conducted: August 12 through October 18, 1985

Enforcement Conference: Scheduled for November 22, 1985

Inspectors: W. L. Forney

P. G. Brochman

Approved By:

RFklarnlh

R. F. Warnick, Chief ////S/J'5'

Reactor Projects Branch 1 Da'te '

Inspection Summary

Inspection on August 12 through October 18, 1985 (Report No. 50-454/85042(DRP))

Areas Inspected: Special unannounced safety inspection by a regional inspector

and a resident inspector to review licensee performance in complying with the

Facility License and Technical Specification requirements. 'An Enforcement

Conference is scheduled for November 22, 1985. The inspection consisted of

69 inspector-hours onsite and at the Region III office by two NRC inspectors.

Results: _This report identified three apparent violations of NRC requirements:

(1) operation of the ECCS system designed to mitigate serious safety ;

events such that it could not have performed its intended safety function and

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failure to follow the applicable Technical Specification Action Requirements -

Paragraph 3; (2) failure of management controls necessary to assure compliance

with the Technical Specifications, 3 examples - Paragraphs 4, 5, and 6; and

(3) exceeding the reactor core licensed thermal power rating - Paragraph 7.

These violations are considered to be of safety significance with the potential

to effect the public's health and safety.

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DETAILS

1. Persons Contacted

Commonwealth Edison Company

R. Querio, Station Manager-

R. Pleniewicz, Production Superintendent

T. Tulon, Operating Engineer

D. Brindle, Operating Engineer

F. Hornbeak, Technical Staff Supervisor

C. Kilbride, Technical Staff

E. Wurtz, Technical Staff

2. General

This inspection was conducted as a result of Region III management's

continuing concern regarding Unit 1 unplanned reactor trips, missed

Technical Specification surveillances, failure to meet Technical

Specification Limiting Conditions for Operation Action Statement

requirements, and the large number of Licensee Event Reports (LER)

issued to date.

The inspection which began on August 12, 1985 and concluded on

October 18, 1985, included reviews of the LERs and the circumstances

surrounding: (1) operation of the unit in Mode 1 with both subsystems of

the Emergency Core Cooling System (ECCS) inoperable; (2) operation of the

unit in Mode 3 with Channel 8 of the Engineered Safety Features Actuation

System (ESFAS) inoperable for a period of time in excess of that allowed

by the Technical Specification Action Requirement; (3) operation of the

Radioactive Gaseous Effluent system with concentrations of Hydrogen (H2 )

and Oxygen (0 2 ) in excess of that allowed by Technical Specifications;

(4) failure to take grab samples when Radioactive Gaseous Effluent

Monitors for H2 and 02 were inoperable; and (5) operation of the unit at

re ntor core thermal power levels in excess of that allowed by the

Facility Operating License.

The inspector's evaluation of these 5 events consisted of a review of the

circumstances surrounding each LER and interviews with licensee personnel.

For each LER the inspector developed a chronology; reviewed the functioning

of safety systems required by plant conditions; reviewed licensee actions

to verify consistency with the Facility Operating License, Technical

Specifications, and implementing procedures; reviewed the licensee

evaluation of the event; and reviewed previously identified problems of a

similar nature. Details of the events are provided in Paragraphs 3

through 7 below.

3. Operating With Both ECCS Subsystems Inoperable

(Closed) LER (454/85081-LL): This LER described events on March 6 through

July 24, 1985, while in Mode 1 (power operations greater than 5% power),

involving inoperability of both ECCS subsystems and the failure to follow

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Technical Specification Action Requirements. This event was discovered

by licensee personnel following identification of a similar problem at

the Callaway Nuclear Power Station by the NRC.

The low pressure injection portion of the ECCS consists of two Residual

Heat Removal (RHR) pumps, two RHR Heat Exchangers, and suction and

discharge flowpaths (see Attachment 1). Technical Specification 3.5.2

states, in part: "Two independent . . . ECCS subsystems shall be

OPERABLE. . . ," when in Modes 1, 2, or 3. The Safety Analysis, contained

in the Byron FSAR, for a Large Break - Loss of Coolant Accident (LB-LOCA)

assumes that each RHR pump is capable of injecting cold borated water

into all four Reactor Coolant System (RC) cold legs during the " Injection

Phase" of ECCS operation.

Both subsystems of the ECCS were rendered inoperable during the

performance of Byron Technical Staff Surveillances 1BVS 5.2.f.3-1,

"ASME Surveillance Requirements for Residual Heat Removal Pump 1RH01PA"

[A Subsystem] and 1BVS 5.2.f.3-2, "ASME Surveillance Requirements for

Residual Heat Removal Pump 1RH01PB" [B Subsystem] when valves 1RH8716A

and ISI8809A (see Attachment 1) were shut during the performance of the

RHR pump 1A surveillance and also when valves 1RH8716B and ISI8809B were

shut during the performance of the RHR pump 18 surveillance.

Byron FSAR, Figure 6.3-2 (see Attachment 1) and its notes define the

position of valves 1RH8716A, 1RH8716B, ISI8809A and ISI8809B as open

during the injection phase of the ECCS operation. With valves 1RH8716A

or ISI8809A shut and RHR pump 1A isolated, the B subsystem would have

only been capable of injecting water into a maximum of two RC cold legs

(1 and 2). Conversely, with valves 1RH8716B or 15I8809B shut and RHR pump

1B isolated, the A subsystem would have only been capable of injecting

water into a maximum of two RC cold legs (3 and 4). Consequently, with

this valve configuration both ECCS subsystems should have been considered

inoperable.

Both ECCS subsystems were inoperable on nine separate instances during

surveillance testing while in Mode 1. The dates of these events and

the approximate length of time the valves were shut (both subsystems

inoperable) is as follows:

Date Time Shut

March 6, 1985 13.0 hours0 days 0 hours 0 weeks 0 months

March 7, 1985 13.0 hours0 days 0 hours 0 weeks 0 months

April 20, 1985 30.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months

April 23,1985 6.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

May 30, 1985 30.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

May 31, 1985 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

July 24, 1985 3.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

July 24, 1985 1.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

July 24, 1985 6.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

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With both ECCS subsystems inoperable, Technical Specification 3.0.3

required that within one hour action should have been initiated to place

the unit in Hot Standby (Mode 3) within the next six hours and the unit

should have been placed in Hot Shutdown (Mode 4) within the following six

hours. Licensee personnel failed: (1) to initiate action within one hour

on the following dates: March 6, 7, April 20, 23, May 30, 31, and July

24; (2) to place the unit in Mode 3 within the next six hours on the

following dates: March, 6, 7, April 20, and May 30; (3) to place the

unit in Mode 4 within the following six hours on the following dates:

April 20 and May 30.

With one RHR pump isolated and the other RHR pump capable of only injecting

water into a maximum of two RC Cold Legs, both ECCS subsystems were

rendered inoperable and thus a system designed to mitigate serious safety

events [LB-LOCA] would not have been able to perform its intended safety

function. With both ECCS subsystems inoperable, the licensee failed to

initiate the required actions. These failures are an apparent violation

of Technical Specifications 3.5.2 and 3.0.3 (454/85042-01(DRP)).

If necessary, the licensed operators in the control room could have opened

the valves, upon receipt of a Safety Injection signal, with the valves

taking less than 10 seconds to open.

A previous violation of regulatory requirements in which both subsystems

of ECCS were inoperable is described in Inspection Report No. 454/85002(DRP).

In that report the violation concerned the isolation of both Safety

Injection pump flowpaths. The licensee's permanent corrective action in

response to violation (454/85002-02(DRP)) was submitted to the NRC in a

letter from D. L. Farrar to J. G. Keppler on July 10, 1985, and stated:

" Station personnel licer.:ed at the Senior Reactor Operator level conducted

a review of all operating prcradures involving ECCS systems, even as a

support system, to determine those prut.edures that could impact Technical

Specification LCO's. As a result of this review, affected operating

procedures were revised." The licensee's corrective action for this

violation does not appear to have been effective in that it failed to

identify that both ECCS systems would be inoperable during the performance

of BVS 5.2.f.3-1 and 5.2.f.3-2.

A previously identified violation (2 examples) of regulatory requirements

was described in Inspection Report No. 454/85016(DRP). Violation No.

454/85016-01(DRP) related to the failure to follow Technical Specification

Action Requirements within the specified time limits. This violation

concerned the failure to shut and de-energize the Pressurizer Power

Operated Relief Valve (PORV) block valves when the PORVs were inoperable

and the failure to place the Control Room Ventilation system in the makeup

mode with an inoperable radiation monitor.

The inspector identified a concern to the licensee that LERs 454/85017

and 454/85040 documented the failure to follow Technical Specification

Action requirements and LER 454/85011 documented the failure to maintain

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two operable ECCS' subsystems and questioned whether these LERs should have

been listed on LER 454/85081 as " previous similar events" as required by

10 CFR 50.73(b)(2)(ii)(J)(5). Additionally, the inspector questioned the

LER's lack of an assessment of the safety consequences and implications

of the event as required by 10 CFR 50.73(b)(2)(ii)(J)(3). These concerns

will be followed as an Unresolved Item (454/85042-02(DRP)).

The inspector identified to the licensee that the' valve identification

numbers and valve positions described in the notes attached to Byron

FSAR, Figure 6.3-2, Sheet 3 were not correct for the valves labeled as

numbers "22," "23," "24," "25," and "26." The licensee has committed to

issuing an amendment to the FSAR to correct this problem and accomplish-

ment of this action will be followed as an Open Item (454/85042-03(DRP)).

4 .~ Failure to Follow Technical Specifications With ESFAS Channel B Inoperable

(Closed) LER (454/85069-LL): This LER described an event on July 14-15,

1985, while in Mode 3, involving the failure to place the unit in the

applicable mode when required by Technical Specification 3.3.2,

Table 3.3-3, Action Statement 21.

At 1904 on July 14, 1985, an instrument mechanic shorted out the power

supply for Channel B of ESFAS causing a Reactor Trip. The channel was

declared inoperable and licensee personnel erroneously began following

the requirements of Table 3.3-3, Action Statement 14. Action Statement

14 required that the unit be placed in Cold Shutdown (Mode 5) within the

next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Licensee personnel failed to realize that Table 3.3-3,

Action Statement 21 was applicable and was more restrictive than Action

Statement 14.

Table 3.3-3, Action Statement 21 states: "With the number of OPERABLE

channels one less than the Minimum Channels OPERABLE. requirement, be in

at least HOT STANDBY [ Mode 3] within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN

[ Mode 4] within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be

bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification

4.3.2.1 provided the other channel is OPERABLE." Action Statement 21 was

invoked by Technical Specification 3.3.2, Table 3.3-3, Function Units

4.b, "Steamline Isolation, Automatic Actuation Logic and Actuation Relays"

and 6.b, " Auxiliary Feedwater, Isolation Automatic Actuation Logic and

Actuation Relays". Each of these functional units required a minimum of

two OPERABLE channels when in Mode 1, 2, and 3, or else follow Action

Statement 21.

At 1910 on July 14, 1985, Channel B was placed in the test position

[ bypassed condition]. At 0104 on July 15, 1985, the unit should have

been placed in Mode 4 due to the inoperability of the Steamline Isolation

and Auxiliary Feedwater functions. At 0320 on July 15 licensee personnel

discovered that Action Statement 21 was applicable and by 0512 had begun

a cooldown to place the unit in Mode 4.

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At 2307 on July 14, following replacement of the damaged power supply a

surveillance to verify Channel B operability was performed. The Main

Steam Isolation and Auxiliary Feedwater functions passed; however, several

other functions failed to pass the surveillance. At 0200 on July 15

licensee personnel voided the surveillance. A voided surveillance is not

an acceptable record to furnish evidence for activities affecting quality.

Licensee personnel failed to recognize that the voided surveillance could

not be used as evidence of the operability of the Main Steam Isolation or

Auxiliary Feedwater Functions.

At 0552 on July 15, ESFAS Channel B was placed in Normal (after having

been in Test for 10.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ) and the cooldown was terminated. At 0612 on

July 15 licensee personnel questioned the operability of the Auxiliary

Feedwater Function and resumed the cooldown. Mode 4 was entered at 1439

on July 15,19.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after Channel B was declared inoperable; Mode 5

was entered at 2048 on July 15. The failure to place the unit in Mode 4

within six hours and placing ESFAS Channel B in Test for greater than two

hours is an apparent violation of Technical Specification 3.3.2 and an

example of the failure of management controls necessary to assure

compliance with the Technical Specifications (454/85042-04a(DRP)).

ESFAS Channel A remained operable throughout the course of this event and

manual initiation of these ESF components could have been performed by

the licensed operators in the control room, if necessary.

This event is indicative of failure of corrective actions provided in

response to previous!y identified violations of regulatory requirements as

described in Inspection Report (454/85016(DRP)), to ensure that Technical

Specification Action Requirements are correctly identified and followed.

(See Report Section 3)

The inspector identified a concern to the licensee that LERs 454/85017

and 454/85040 documented the failure to follow Technical Specification

Action requirements and questioned whether these LERs should have been

listed on LER 454/85069 as " previous similar events" as required by

10 CFR 50.73(b)(2)(ii)(J)(5). This concern will be followed as an

Unresolved Item (454/85042-05(DRP)). An additional concern relating to

use of voided documents to provide an acceptable record to furnish

evidence of activities affecting quality will be followed as an Open Item

(454/85042-06(DRP)).

5. Explosive Gas Concentrations in the Radioactive Gaseous Effluent System

(Closed) LER (454/85067-LL): This LER described events on July 6-14,

1985, while in Mode 1, involving failure to follow Technical

Specifications Action Requirements for Radioactive Gaseous Effluents

relating to the Hydrogen (H 2) and Oxygen (02 ) concentrations present in

the Waste Gas system.

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On July 6, 1985, an Equipment Attendant recorded a 2H concentration of

5.5% and at 1140 a chemist recorded an 20 concentration of 3.9% on i

Special Chemistry Data Sheet, BCP-400-T.60, Revision 0. Technical '

Specification 3.11.2.5, " Radioactive Effluents Explosive Gas Mixture,"

states: "The concentration of oxygen in the WASTE GAS HOLDUP SYSTEM shall

be limited to less than or equal to 2% by volume whenever the hydrogen

concentration exceeds 4% by volume." Applicability of this specification

is "at all times." Technical Specification 3.11.2.5.a states: "With the

concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 2%

by volume but less than or equal to 4% by volume, reduce the oxygen

concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ." The OD Waste Gas

Decay tank was taken out of service on July 6, 1985, and records indicate

that the tank remained out of service, with concentration of 0 2/H2 greater

than that allowed by Technical Specification 3.11.2.5.a, until July 11,

1985. There is no record to indicate that the licensee initiated any

action to reduce the 02 concentration at any time prior to July 11, 1985,

in accordance with Byron Abnormal Operating Procedure OBOA PRI-8, "02/H2

Explosive Mixture Units 0, 1, 2."

Technical Specification 3.11.2.5.b states: "With the concentration of

oxygen in the WASTE GAS HOLDUP SYSTEM greater than 4% by volume,

immediately suspend all additions of waste gases to the system and reduce

the concentration of oxygen to less than or equal to 4% by volume then

take ACTION a above." At 2120 on July 11, 1985, the 0 2/H2 concentrations

of the OD tank were recorded as 10.8%/4.1% respectively and remained

greater than that allowed by Technical Specification 3.11.2.5.b until

1348 on July 14, 1985.

Review of the licensee records indicate that OB0A-PRI-8 was entered, for

tank OD, on July 11, 1985, to reduce the explosive mixture of 02/H2 .

Licensee personnel attempted to reduce the 02 concentration below the

limit of Technical Specification 3.11.2.5.b by releasing the tank;

however, the release was terminated when it was determined that Radiation

Monitor 0PR02J, which controls the Waste Gas discharge valve position, was

inoperable due to insufficient amount of vacuum above the low limit

alarm setpoint. Subsequently, a temporary alteration was installed on

July 12, 1985, which would allow the release to be accomplished. At 2100

on July 12, the release from the OD tank was recommenced; however, it was

observed that the pressure in tank OA was also showing a decrease and the

release was terminated once again. The reason the relear was terminated

was that Byron procedures do not allow for more than one Waste Gas Decay

tank to be released at the same time. A nuclear work request was

initiated to repair the 0A tank manual release valve, and after repairs

were completed tha OD tank release was recommenced and a nitrogen purge

was initiated. The talease and the purge were terminated at 1348 on

July 14, 1985, when the 0 2/H2 concentrations were determined to be less

than the limits of Technical Specification 3.11.2.5.a.

Failure of the licensee's management systems to identify the high

! concentrations of 02 /H 2 on July 6, 1985, resulted in no action being

taken by the licensee to reduce these concentrations below Technical

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Specifications limits from July 6 until July 11, 1985. After identifi-

cation by the licensee on July 11, that the 0 /H 2 2concentration in the OD

tank exceeded the limits of Technical Specification 3.11.2.5.b, subsequent

management decisions and management systems failed to reduce the 02/H 2

concentrations below Technical Specifications limits until July 14, 1985.

The inspector's review determined that the items listed below were

contributing factors to this event:

a. Incomplete / inaccurate Rad-Chem records,

b. Incomplete Limiting Condition for Operation Action Requirement

(LC0AR) data sheets.

c. Inadequate tracking of LC0AR conditions by management / supervision.

d. Inadequate review and assessment by management / supervision of

appropriate corrective actions to be accomplished.

e. Failure of management / supervision to ensure that corrective actions

identified were accomplished in a timely manner.

f. An apparent attitude of management / supervision to disregard Technical

Specification Action Requirements that do not provide specific

primary plant operational penalties.

The failure to reduce the explosive concentrations of 0 2/H2 present in

the Waste Gas system is an apparent violation of Technical Specification 3.11.2.5 and an example of the failure of management controls necessary

to assure compliance with the Technical Specifications (454/85042-04b(DRP)).

6. Failure to Take Grab Samples With Inoperable Radioactive Gaseous Effluent

Monitors

(Closed) LER 454/85082 described events on July 28 through August 4, 1985,

while in Modes 1 - 4, relating to the failure to obtain and analyze grab

samples from the Waste Gas system when two channels of Radioactive Effluent

Monitoring Instrumentation were inoperable.

At 2200 on July 16, 1985, Technical Specification 3.3.3.10, Table 3.3-13,

Instrument 3.a, 0AT-GW8000, " Hydrogen Analyzer" was taken out of service.

At 0720 on July 20,1985, Table 3.3-13, Instrument 3.b, 0AT-GW8003,

" Oxygen Analyzer" was taken out of service.

Technical Specification 3.3.3.10, Table 3.3-13, Instrument 3.a required a

minimum of one channel to be operable at all times, or else follow Action

Statement 38. Instrument 3.b required a minimum of two channels to be

operable at all times, or else follow Action Statement 38. Action

Statement 38 states, in part: "With the number of channels OPERABLE one

less than required by the Minimum Channels OPERABLE requirement, operation

of this system may continue provided grab samples are taken and analyzed

at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . . . ."

With system operation continuing licensee personnel began taking and

analyzing grab samples every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This requirement was listed on a

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status board in the Rad-Chem office. On July 27 this requirement was

inadvertently erased from the status board by licensee personnel. As a

consequence, the licensee failed to take and analyze grab samples on the

following dates:

a. While in Mode 3: July 28, 1985

b. While in Mode 4: July 29 - 31,1985

c. While in Mode 2: August 1, 1985

d. While in Mode 1: August 2 - 4, 1985.

The failure to obtain and analyze grab samples at least once every 24

hours is an apparent violation of Technical Specification 3.3.3.10 and an

example of the failure of management controls necessary to assure

compliance with the Technical Specifications (454/85042-04c(DRP)).

The failure to obtain samples required by Technical Specifications was

previously described in Inspection Report 454/85021(ORP). The licensee's

permanent corrective action in response to violation 454/85021-01b(DRP)

was submitted to the NRC in a letter from D. L. Farrar to J. G. Keppler

on July 19, 1985, and stated, in part: "A file organizer has been placed

in the Station counting room for initiated surveillance procedures.

Technicians are periodically instructed by the responsible foreman to

review the file for initiated surveillances. Initiated surveillances are

also tracked on the counting room shift turnover sheet . . . ." This

violation is indicative of the licensee addressing the specific violation

only, but not addressing the root cause of the problem. Consequently,

the action taken to avoid further violations was not effective.

These three examples (Paragraphs 4, 5, and 6) of apparent violations of

Technical Specifications are indicative of the failure of Management /

management systems and failure of corrective actions provided in response

to previously identified violations of regulatory requirements, as

described in Irspection Reports No. 454/85016(DRP) and No. 454/85021(DRP)

to ensure that Byron's operations are conducted in accordance with

regulatory requirements.

7. Exceeding the Reactor Core Thermal Power Limit

(Closed) LER (454/85080-LL): This LER described events on Auguit 6-7,

1985, while in Mode 1, involving exceeding the reactor core licensed

thermal power rating.

Licensee personnel monitor reactor core thermal p'ower with 4 channels

of Nuclear Instruments (NI). These channels of Power Range" NI are

calibrated by the performance of a secondary heat balan::e. Byron

Technical Specification Surveillance 180S 3.1.1-2, " Calorimetric

Calculation Surveillance", accomplished this heat balance. This

procedure compares the heat transferred into the steam generators with

the heat transferred out of the steam generators by calculating the

enthalpy of the water going in and out times its flow rate. Based on

this heat balance the core thermal power is determined and the " Power

Range" NI are adjusted so that 100% indicated power is equal to 3411

megawatts thermal (HWT).

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Byron Station Facility Operating License NPF-37, License Condition 2.C(1)

states, in part: "The licensee is authorized to operate the facility at

reactor core power levels not in excess of 3411 megawatts thermal (100%

power) . . . . "

A licensed operator perfonning the procedure identified that a portion

! of the feedwater flow rate was not being accounted for. In Byron's

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Westinghouse Model "D-4" Steam Generators the feedwater flow is split

into two paths. One of these paths, the tempering line feedwater flow

rate was not accounted for in the surveillance procedure; consequently,

a nonconservative error was introduced into the surveillance and the

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" Power Range" NI were adjusted so that indicated power was lower than the

actual reactor core power. Subsequently, licensee personnel reviewed the

plant computer records and determined that this error had caused the

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licensed thermal power limit to be exceeded.

The NRC's policy regarding exceeding licensed power levels is that the

l average power level over any eight hour shift should not exceed the full

steady-state licensed power level. While it is permissible to briefly

l exceed the full steady-state licensed power level by as much as 2% for as

!

long as 15 minutes, in no case is it permissible for 102% power to be

exceeded. Power excursions to less than 102% for periods longer than 15

minutes are permissible (i.e.,101% for 30 minutes,100.5% for one hour,

! etc.) provided that the power level, averaged over an eight hour shift,

does not exceed 100%.

Reactor core power (MWT) was greater than 100% power, averaged over an

eight hour shift, on the following three instances:

l a. 1500 - 2259 on July 26, 1985 - 100.06%/3413 MWT

b. 2300 - 0659 on July 27, 1985 - 100.18%/3417 MWT

i c. 0700 - 1459 on July 27, 1985 - 100.30%/3421 MWT.

! Additionally, the average reactor core power equaled or exceeded 100.5%

for greater than one hour on two instances:

a. 2221 - 0013 on July 26, 1985 - 100.50%/3428 MWT

b. 0924 - 1105 on July 27, 1985 - 100.68%/3434 MWT.

The failure to maintain reactor core power less than or equal to 3411 MWT

I is an apparent violation of Facilit

Condition 2.C(1) (454/85042-06(DRP)y Operating

). Additionally, the License

licensee NPF-37,

failed License

to submit a written report of this event within the 30 day time limit

requirement of Facility Operating License NPF-37, License Condition 2.F.

Both this apparent violation and the apparent violation described in

Paragraph 3 are examples of licensee personnel failing to correctly

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preparc surveillance procedures and licensee management failing to

l adequately review surveillance procedures to ensure that all applicable

safety analysis conditions had been satisfied.

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8. Enforcement Conference Scheduled For November 22, 1985

An enforcement conference is scheduled for November 22,1985, to be held

at the Region III office. . The inspectors met with licensee representatives

on October 28, 1985 and summarized the purpose and scope of the inspection

and the apparent findings. The inspectors discussed the likely informa-

tional content of the inspection report with regard to documents or

pr?: esses reviewed by the inspectors during the inspection. The licensee

did not identify any such documents / processes as proprietary.

9. Unresolved Items

Unresolved items are matters about which more information is required in

order to ascertain whether they are acceptable items, violations, or

,

deviations. Unresolved items disclosed during the inspection are

discussed in Paragraphs 3 and 4.

10. 'Open Items

Open items are matters which have been discussed with the licensee, which

will be reviewed further by the inspectors, and which involve some action

on the part of the NRC or licensee or both. Open items disclosed during

the inspection are discussed in Paragraphs 3 and 4.

11

O

,

.

i

l

3 gg i AMENDMEIET 44

Hot I.egs DECDEER 1983

Free SI pumps

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24 ISI8809A

26 15188095 RER Subsystem 'B' Discharge Flowpath - - -- - - - - - BYRON /BRAIDWOOD STATIONS

FIN AL S AFETY AN ALYSIS REPORT

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FIGURE 6,3-2

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DIAGRAM OF RESIDUAL HEAT IEMOVAL

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(SHEET 30F3)

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ML20138R802

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