ML20138R802
| 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: |
| References | |
| 50-454-85-42, NUDOCS 8511190053 | |
| Download: ML20138R802 (12) | |
See also: IR 05000454/1985042
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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
RFklarnlh
Approved By:
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 (H )
2
and Oxygen (0 ) in excess of that allowed by Technical Specifications;
2
(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
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 <br />0 hours <br />0 weeks <br />0 months <br />
March 7, 1985
13.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />
April 20, 1985
30.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />
April 23,1985
6.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />
May 30, 1985
30.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
May 31, 1985
5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
July 24, 1985
3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
July 24, 1985
1.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />
July 24, 1985
6.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />
3
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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
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 <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN
[ Mode 4] within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be
bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />) 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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
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 ) and Oxygen (0 ) concentrations present in
2
2
the Waste Gas system.
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On July 6, 1985, an Equipment Attendant recorded a H concentration of
2
5.5% and at 1140 a chemist recorded an 0 concentration of 3.9% on
2
Special Chemistry Data Sheet, BCP-400-T.60, Revision 0.
Technical
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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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />." 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 /H
2 2 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, "0 /H
2 2
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 /H2 concentrations
2
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 0 /H .
2 2
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 /H2 concentrations were determined to be less
2
than the limits of Technical Specification 3.11.2.5.a.
Failure of the licensee's management systems to identify the high
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concentrations of 0 /H on July 6, 1985, resulted in no action being
2 2
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 concentration in the OD
2 2
tank exceeded the limits of Technical Specification 3.11.2.5.b, subsequent
management decisions and management systems failed to reduce the 0 /H
2 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 /H2 present in
2
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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> . . . ."
With system operation continuing licensee personnel began taking and
analyzing grab samples every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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
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of the feedwater flow rate was not being accounted for.
In Byron's
Westinghouse Model "D-4" Steam Generators the feedwater flow is split
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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
" Power Range" NI were adjusted so that indicated power was lower than the
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actual reactor core power. Subsequently, licensee personnel reviewed the
plant computer records and determined that this error had caused the
licensed thermal power limit to be exceeded.
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The NRC's policy regarding exceeding licensed power levels is that the
average power level over any eight hour shift should not exceed the full
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steady-state licensed power level. While it is permissible to briefly
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exceed the full steady-state licensed power level by as much as 2% for as
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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
Condition 2.C(1) (454/85042-06(DRP)y Operating License NPF-37, License
is an apparent violation of Facilit
I
).
Additionally, the licensee failed
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
.
preparc surveillance procedures and licensee management failing to
l
adequately review surveillance procedures to ensure that all applicable
safety analysis conditions had been satisfied.
!
10
1
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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
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AMENDMEIET 44
DECDEER 1983
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BYRON /BRAIDWOOD STATIONS
FIN AL S AFETY AN ALYSIS REPORT
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FIGURE 6,3-2
!
DIAGRAM OF RESIDUAL HEAT IEMOVAL
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(SHEET 30F3)
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