IR 05000454/1990003
| ML20012A716 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 02/23/1990 |
| From: | Beverly Clayton NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20012A711 | List: |
| References | |
| 50-454-90-03, 50-454-90-3, 50-455-90-02, 50-455-90-2, NUDOCS 9003120454 | |
| Download: ML20012A716 (17) | |
Text
~
i
..
..
..
j l
o
.
)
U.S. NUCLEAR REGULATORY COMMISSION i
'
REGION III
e Report No. 50-454/90003(DRP);50-455/90002(DRP)
Docket Nos. 50-454; 50-455 License Nos. NPF-37; NPF-66 l
Licensee: Commonwealth Edison Company
Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Station Units 1 and 2
.
Inspection At: Byron Site, Byron, Illinois Inspection Conducted: January I through February 16, 1990 Inspectors:
W. J. Kropp R. N. Sutphin J. D. Smith F. L. Brush D. R. Cal oun, Approved By:
rent Clay on Chief o7A.JJ, Reactor Projects Section IA Date Inspection Summary
. Ins)ection from January I through February 16.1990(ReportNo. 50-454/90003 (DR)); 50-455/90002(DRP))
Areas Inspected:
Routine, unannounced safety inspection by the. resident inspectors of actions on previous inspection findings; operational-safety; reactor startup; Jnit I refuel outage; onsite event follow-up; radiological controls; current material condition; safety assessment /
quality verification; maintenance activities; surveillance activities; TI 2515/101; engineering / technical support; and refueling / spent fuel pool activities.
Results: The licensee's performance in plant operations was overall considered good. The licensee's pre-outage planning was considered a strength. However, increased management attention. is needed in the area
!
of radiological controls to ensure an adverse trend is not established.
I Management has already initiated action to improve and independently assess
,
this area.
The licensee's site Quality Assurance (QA) organization has l
recently implemented a pilot surveillance program that required QA engineers i
to be in the field 50% to witness on-going work and material condition of the plant. This approach towards surveillance activities should produce performance based inspections that should be a viable management tool. The i
licensee's performance in the Maintenance / Surveillance area was overall
$[3bbK
'
.O E
.j
..
.
.
,
-
.
l considered good. The major surveillances witnessed by the inspectors were well coordinated with adequate shift briefings prior to performance. The licensee missed one surveillance during this inspection period which was the first missed surveillance in 427 days. The licensee's performance in Engineering / Technical Support was overall considered satisfactory. However, the interface between site nuclear engineering and corporate Nuclear Fuel Services was ineffective and resulted in an aborted reactor startup on November 11, 1989. Previous Inspection Report 50-454/89021; 50-455/89024 had also identified two other examples of inadequate interface between on-site and off-site engineering organizations. The inspectors will continue to nonitor this area. Of the thirteen areas inspected, no violations were identified.
.
m
...
,
,
,
.
.
.
DETAILS 1.
Persons Contacted Comonwealth Edison Company (Ceco)
- R.-Pleniewicz, Station Manager
- K. Schwartz, Production Superintendent R. Ward, Technical Superintendent
- J. Kudalis, Service Director D. Brindle, Operating Engineer, Administration T. Didier, Operating Engineer, Unit 0 T. Gierich, Operating Engineer, Unit 2
- T. Higgins, Assistant Superintendent, Operating J. Schrock, Operating Engineer, Unit 1
- M. Snow, Regulatory Assurance Supervisor
- D. St. Clair, Assistant Superintendent, Work Planning
'T. Tulon, Assistant Superintendent, Maintenance D. Winchester, Quality Assurance Superintendent
- D. Wozniak, ENC Project Manager E. Zittle, Regulatory Assurance Staff
- Denotes those attending the exit interview conducted on February 16, 1990, and at other times throughout the inspection period.
The inspectors'also had discussions with other licensee employees, including members of the technical and engineering staffs, reactor and auxiliary operators, shif t engineers and foremen, and electrical, mechanical and instrument maintenance personnel, and contract security personnel.
2.
Action on Previous Inspection Findings (92701 & 92702)
a.
(Closed) Violation 454/88019-01(DRP):
Failure of procedures to s Decify the use of an appropriate means of RCS level indications wille draining the refueling cavity to a level below the reactor vessel flange. The licensee initiated numerous immediate actions to correct and preclude recurrence. These actions were documented in a letter from the licensee to NRC Region III dated June 26, 1989. The inspectors selected several of the corrective actions for verification of implementation and no problems were noted.
The actions verified were:
Revision of pro W ure B0P-RH-9, Revision 4, " Pump Down of
!
Reactor Cavity to RWST", to require two functional methods of level indication to be used for any draining operations below the 403' elevation.
Revision of p'rocedure BOP RC-4a, Revision 2
SystemDrain.toincludespecificinstructionsfortygon i
installation.
!
!
!
. -.
.
..;
.
.
.
,
-
,
.
Revision of procedure BOP RC-4a to require use of the Chemical
and Volume Control System when draining below the reactor
'
vessel flange.
Revision of BGP 100-5, Revision 4, " Refueling Outage", to i
require training of operators shortly before entering a reduced inventory condition.
,
I b.
(Closed) Violation 454/89016-01(DRP):
Failure to perform a 50.59 safety evaluation and to place a " Caution Tag" at the Remote Shutdown Panel (RSP).
The inspectors reviewed the licensee's actions to correct and avoid further violations. The licensee's action to avoid further violations included declaring the centrifugal charging pump inoperable if the associated mini-flow l
s valve was closed in the future. Also, to. determine if a 50.59'
~
evaluation was required, the licensee committed to provide screening
-
criteria if any of the following conditions exists; 1) equipment
taken out-of-service 2) degraded equipment and 3) an abnormal line-up exists. Also, procedure BAP 330-6, " Caution Card Procedure" was revised to require a Caution Tag at the RSP if the same component / control exist at the Main Control Board.
'
c.
(Closed) Bulletin 88-02: Rapidly propagating fatigue cracks in i
steam generator tubes. The licensee response to this bulletin has been reviewed by NRR and found acceptable. The acceptance was documented in a letter dated January 17, 1990, from L. N. 01shan,
'
NRR Project Manager to T. J. Kovach, licensee Nuclear Licensing
Manager.
>
!
d.
NRC Region III management has reviewed the existing open items for
!
the Byron station and have determined that the following open item
will be closed administrative 1y due to the safety significance
>
relative to emerging priority issues. The licensee is reminded that commitments directly relating to the open item is the responsibility
,
of the licensee and should be met as committed.
NRC Region III will review licensee actions by periodically sampling administrative 1y l
closed items.
J l'
(Closed)OpenItem 455/89007-01.
3.
Plant Operations
Unit 1 operated at power levels up to 100% until January 5,1990, when the unit entered a scheduled 59 day refuel outage. See Section 3.d.
o l
of_ this report for further information about the outage.
Unit 2 operated at power levels up to 100% in a load following mode until January 18,1990, at 12:42 a.m., at which time a reactor trip /
safety injection occurred due to a spurious low steam pressure signal from channel 525 while channel 526 was in a tripped condition for calibration (2/3 coincidence for each S/G). For further details, see Section 3.c.1 of this report. The reactor was taken critical at 9:30
'
,
. - - - -
.
-.
-
.
...
,,
,
,
.
.
!
!
p.m. on January 18, 1990, and the turbine generator was synchronized to the grid. on January 19, 1990, at 3:30 a.m.
Since January 18, 1990, i
the unit had operated up to 100% power in the load following mode until the unit was shutdown _on February 9,1990, to repair cracks in the 2C accumulator fill line weld. The unit was returned to service on February 12, 1990, when the turbine generator was synchronized to the
.,
,
grid at 6:31 a.m.
l a.
Operational Safety (71707)
'
~
During the inspection period, the inspectors verified that the facility was being operated in conformance with the licenses and
,
regulatory requirements and the licensee's management responsibili-
-
ties were effectively carried out for safe operation.
Verification
,
was-based on routine direct observation of activities and equipment i
performance, tours of the facility, interviews and discussions with
licensee personnel, independent verification of safety system status
'
and limiting conditions for operation action requirements (LC0ARs),
j corrective action, and review of facility records.
On a sampling basis the inspectors daily verified proper control room staffing and access, operator behavior, and coordination of
!
,
plant activities with ongoing control room operations; verified
<
operator adherence with the latest revisions of procedures for ongoing activities; verified operation as required by Technical l
Specifications (TS); including (ESF) and ESF electrical alignment an compliance with LC0ARs, with emphasis
,
t on engineered safety features
'
valve positions; monitored instrumentation recorder traces and duplicate channels for abnormalities; verified status of various lit
,
i annunciators for operator understanding, off-normal conditions) and and compensatory actions; examined nuclear instrumentation (NI I
other protection channels for proper operability; reviewed radiation monitors and stack monitors for abnormal conditions; verified that
'
onsite and offsite power was available as required; observed the
'
!
frequency of plant / control room visits by the station manager, superintendents, assistant operations superintendent, and other
L managers; and observed the Safety parameter Display System (SPDS)
'
i for operability. No problems were noted.
b.
ReactorStartup(71707)
'
L On February 12, 1990, the licensee was successful in the restart of i-Unit 2 after a forced outage that was initiated on February 9,1990 L
due to a crack in a fill line weld for the 2C accumulator. For more details about the weld cracks see Section 3.c.2 of this report.
I The initial restart attempt on February 11,1990,at9:10p.m.(CST)
was aborted when critical rod position was predicted to be outside
,
the 500 pcm administrative limit based on the 8-fold count. Based
on the 8-fold count, criticality was predicted at 10 steps on bank
"D" instead of the estimated rod position of 85 steps on bank "D".
The difference in steps represented about a 750 pcm deviation. The licensee's Nuclear Fuel Services department determined that the l
control rod worth curves utilized to calculate the estimated l
L
_
..,
.
.
,
J l
critical rod position did not reflect actual core burnup. After the station's nuclear engineers computed a new estimated critical position based on revised control rod worth curves, another startup j
was commenced on february 12, 1990, at 3:32 a.m.
The estimated critical.od position was 78 steps on bank "D".
At 4:07 a.m. the
,
reactor was made critical at 44 steps on bank "D".
A previous
!
startup on Unit 2 on November 21, 1989 also was aborted due to a large deviation (950 pcm) between calculated estimated rod position and critical rod position based on the 8-fold count. However, a Unit 2 startup on January 18, 1990, after a reactor trip, was i
normal with the actual critical rod position close to the estimated
rod position.
For further details on the nuclear engineering i
aspects of these events, see Section 8.a of this report.
-
,
c.
Onsite Event follow-up (93702)
(1) On January 18, 1990, at 12:42 a.m. (CST), a Unit 2 reactor trip i
and safety injection (SI) occurred with Unit 2 at 99% reactor
'
power.
The reactor trip /SI signal was initiated during a
!
calibration of PT-526, a steam pressure channel for the "B" steam gtnerator (50). With PT-526 channel in the tripped
condition for the calibration, a spurious low spike on another
'
"B" SG steam pressure transmitter, PT-525, resulted in the pick up of the bistable for a safety injection signal (2/3
-
coincidence safety injection signal on any SG). No other work i
was in progress at the time on PT-525. The licensee could
-
not duplicate the spike on PT-525 and decided to replace the pressure transmitter with a new transmitter prior to a reactor startup.
The licensee concluded that the most probable cause l
was a defective PT-525 transmitter. The licensee also placed
'
!
recorders on channel PT-525 to monitor the channel's
,
performance. After replacement of the pressure transmitter for channel PT-525, and other miscellaneous maintenance, a
'
l Unit 2 reactor startup was commenced. The reactor achieved
'
criticality on January 18, 1990, at 9:30 p.m.
At 3:33 a.m. on January 19, 1990, the turbine generator was synchronized to the grid.
-v
,
l (2) On February 6,1990, at 6:41 p.m. (CST) while equalizing the IC l
safety injection accumulator with the ID accumulator, water was observed leaking from the JC accumulator water fill Ifne at a weld. The IC accumulator was immediately vented. At the time the accumulators were being used as a source for pressurizing the discs of the loop stop valves. The licensee had previous problems with cracking of welds where the fill line enters the l
accumulator and had, (1) installed a U-bolt clamp on the fill
lines to prevent vertical movement and (2) initiated action to dye penetrant test (PT) the suspect welds just prior to and
<
immediately after an outage. On December 21, 1989, prior to the Unit I refuel outage, the weld on the IC accumulator had a PT examination with no indications identified. An inspection of the IC accumulator fill line, subsequent to the leak, on February 6,1990, identified that the U-bolt clamp was loose.
The licensee initiated action to PT the suspect welds on all
.
u.
.
.
.-
.
-
.
.
,
-
.
.
accumulator fill lines and to verify tightness of all U-bolt clamps. The inspection identified two other loose clamps on the fill lines for the ID and 2C accumulators. No indications during PT examinations of the suspect welds were identified except for the 2C accumulator fill line. The PT examination on the 2C accumulator identified four linear indications (3-1/4" and 1-1").
The licensee commenced a Unit 2 reactor shutdown on February 9,1990 and proceeded to Mode 3 with reactor pressure less than 1000 psi for repairs. Work activities included:
(1)
repair weld on 2C accumulator (2) torque nuts on all U-bolt clamps to 20 f t-lbs. and (3) tack weld the nuts on all U-bolt clamps. The work activities were completed, and the Unit returned to service on February 12, 1990, at 6:31 a.m.
The licensee has committed to PT the suspect welds and inspect the U-bolt clamps once a month until a re-evaluation of the-overall problem with cracked welds to ascertain if the U-bolt clamp was the appropriate engineering solution.
d.
Unit 1 Refuel Outage On January 5,1990, Unit I commenced a 59 day refuel outage. The return to service date was scheduled for March 5,1990. Major activities planned during the outage include; a complete off-load of the core, shot-peening of the cold legs for all steam generators (SG); eddy current testing of all SG tubes, replacement of two low pressure turbine rotors; 5 year inspection of "A" diesel generator and approximately 55 modifications. The inspectors reviewed the licensee's pre-outage planning activities and considered the activities above average. Some of the activities and/or methods utilized by the licensee in pre-outage planning included:
Several weeks prior to the outage, the status of priority
purchases that were needed to support the outage were discussed daily at the station's morning meetings. The status also was documented in the morning meeting minutes.
Development of a pocket size booklet that contained pertinent
outage information such as; milestone dates; outage telephone numbers, out-of-service information etc.
Outage kick-off meeting for key station and contractor
personnel that included; introduction of personnel, discussion of goals for the outage, identification of critical paths and a presentation by a consultant that pertained to motivation and team work.
At present, the outage has progressed generally on schedule with few problems being identified.
.
.
..
.
,
,
4-
e, RadiologicalControls(71707)
'
The inspectors verified that personnel were following health physics procedures for dosimetry, protective clothing, frisking, posting, etc. and randomly examined radiation protection instrumentation for use, operability, and calibration. During this inspection period three of the Radiation Occurrence Reports (ROR) issued by the licensee were reviewed by the inspectors. The RORs reviewed (90-1, 90-2 and 90-5), pertained to:
(1)animproperUnit1 initial containment entry. (2) the spread of contamination when the 2A Residual Heat Removal system was vented and (3) a worker receiving a 18% body burden while transporting shot-peening equipment. The inspectors will review the completed RORs for adequate root cause and corrective action. However, the inspectors consider the licensee's performance in this area has reached a level where increased management attention is needed to ensure an adverse trend is not established in radiological controls. The
. licensee has already initiated some immediate actions that included:
(1) health physics personnel at shift briefings (2) nuclear station operator (NS0) monitoring activities that could affect radiological conditions of the plant (3) emphasizing the importance of good communications (interface) between operations and health physics and (4) requesting an assist' visit from INPO.
f.
CurrentMaterialCondition(71707)
The inspectors performed general plant as well as selected system and component walkdowns to assess the general and specific material condition of the plant, to verify that Nuclear Work Requests (NWRs)
had been initiated for identified equipment problems, and to evaluate housekeeping. Walkdowns included an assessment of the buildings, components, and systems for proper identification and tagging, accessibility, fire and security door integrity, scaffolding, radiological controls, and any unusual conditions. Unusual condi-tions included but were not limited to water, oil, or other liquids on the floor or equipment; indications of leakage through ceiling, walls or floors; loose insulation; corrosion; excessive noise; unusual temperatures; and abnormol ventilation and lighting. The
. inspectors concluded that the raaterial condition of the hardware for Unit 2 continues to be above average and housekeeping has improved since the last inspection period. Material condition of the hardware was not assessed for Unit I since the unit was in a refueling outage. Also, the housekeeping for Unit I areas was typical for an outage.
Assessment of Plant Operations The management involvement and control in plant operations overall was considered good. The licensee's pre-outage planning was considered a strength. The inspectors observed visits of licensee management to the control room and other areas of the plant.
The approach to the identification and resolution of technical issues by the operations department during this inspection period overall was good. However,
8
-,,... - _,
,
,-
.
.,
-
.
.
increased management attention is needed in the area of radiological controls to ensure an adverse trend is not established and to improve perfonnance in this area. Management has already initiated action to improve and assess this area of plant operations.
No violations or deviations were identified.
4.
Safety Assessment / Quality Verification (40500, 90712, 92700)
a.
The site Quality Assurance (QA) organization recently initiated a pilot program for the performance of QA surveillances. The program required the site QA enginee;s to be in the plant 50% of the time, to monitor ongoing work activities and to assess material condition of the plant. The surveillances performed to date under the pilot program have been performance based. For example surveillance 06-90-05, performed January 10 - 11, 1990, identified that the tygon tubing utilized for reactor vessel level (RVL) was not installed by maintenance in accordance with procedure IBOP RC-4a, " Reactor Coolant System D: ain", Revision 2.
The tygon tubing had been routed to a vent bottle rather to the pressurizer. The installation of the tygon tubing, even though completed, had not yet been walked down by the operations staff. QA determined that the work planner had utilized old Nuclear Work Requests instead of procedure IBOP RC-4a.
The tygon tubing installation was corrected and the activity of tygon tube installation for RVL will be controlled in the future by the licensee's computer program for surveillances to ensure the correct procedure, 180P RC-4a, is used for the installation.
b.
The inspectors reviewed the licensee's completed Deviation Reports (DVRs) generated during the inspection period. This was done in an effort to monitor the conditions related to plant or personnel performance, potential trends, etc. DVRs were also reviewed for proper issuance and disposition in a manner consistent with the applicable procedures and the QA manual. The following DVRs were reviewed:
06-01-89-155 - CD/CB pump motor fire.-
06-01-89-159 - Breakers accidentally tripped off.
06-02-89-004 - 2A RHR pump oil sample variation.
06-02-89-085 - Plant transient due to failure of excitation limiter card, c.
TheinspectorsreviewedtheMaintenanceProblemAnalysisData(PAD)
sheets that were implemented on a trial basis during January -
February 1989, via a maintenance memo and subsequently proceduralized in October 1989. The purpose of the PAD program was to provide a method for systematic analysis of maintenance problems. Procedure BAP 1600-8, " Maintenance Problem Analysis Program", Revision 0, delineates the process and methods utilized for the PADS.
The procedure was reviewed by the inspectors with the following results:
The procedure did not require that a scheduled completion
date be established for either determination of root cause
,
and corrective action and subsequently for completion of the corrective action.
- - - _
n
-
.
- .
.
,
-.
Methods were not established to routinely feedback the status
of the PADS to senior station management.
Methods were not established to escalate PAD resolution to
managenent, if necessary.
The inspectors reviewed the status of the PADS issued to date and
- determined that of the 28 PADS issued during January - February 1989, 10 had been closed.
The ins)ectors reviewed the ten closed PADS and determined eight of ten t1at were closed had not met the criteria for a PAD. Therefore, of the 20 PADS issued during January -
February,1989, that met the PAD criteria, only two had been closed.
Since the issuance of 8AP 1600-8 on October, 1989, 28 PADS had been issued and were still open. Also, the inspectors noted several PADS had been logged several months earlier, but copies were not routed to the maintenance staff personnel administratively responsible for the PAD program...The inspectors discussed results of-the review with the licensee. The licensee implemented the following actions to resolve the inspectors' concerns with the PAD procedure:
(1) Issued diaintenance memo 400-11, " Problem Analysis Data Sheet ProcessingGuidelines"thatspecifies(a)prioritizationof PADsand(b)estimateddatesforscheduledcompletionof PADS based on prioritization of the PAD.
(2)
Issued a PAD status report to management that identified the status of each PAD by department, d.
Assessment of Safety / Quality Verification In conclusion, the inspectors considered the licensee's overall performance in this area as improving.
The QA pilot surveillance program presently implemented at the station is considered peformance based and should be a viable management tool.
However, management involvement in the PAD program apparently was not sufficient to ensure the program was adequately established and implemented.- The licensee implemented immediate action to correct any weaknesses in the PAD program. The root cause analysis and corrective actions identified in DVRs continue to be above average.
No violations or deviations were identified.
5.
Maintenance / Surveillance (62703&61726)
a.
Maintenancefetivities(62703)
Station maintenance activities that affected safety-related and associated systems and components were observed or reviewed to ascertain compliance with approved procedures, regulatory guides and industry codes or standards, and in conformance with Technical Specifications.
The following items were considered during this review:
the limiting conditions for operation were met while components or systems were removed from and restored to service; approvals were
W
,
,
.
r",
I
-
,
.-
,
l
.
.
.
'
s tained prior to initiating the work; activities were accomplished-using approved procedures and were inspected as applicable;
'
functional testing and/or calibrations were performed prior to a
'
returning components or systems to service; quality control a
records were maintained; activities were accomplished by qualified
personnel; parts and materials used were properly certified;
.
j
!
radiological controls were implemented; and fire prevention controls
'
'
' were-implemented. Work requests were reviewed to determine the
status of cutstanding jobs and to assure that priority was assigned
,
.m to safety-related equipment maintenance which may affect system performance.
,
'
Portions.of the following maintenance activities were observed.and reviewed:
,
NWR B 68497 - 5 year. tear-down/ inspection of IA Diesel Generator.
s NWR B 68498'- 18 unth inspection and test of IB Diesel Generator..
.t
. NWR:B 69802 - Trevitest, per BMP 3114-15, of IMS017D', Main Steam
-
Safety Valve.
NWR B 69824 - Trevitest, per BMP 3114-15, of INS 013C, Main Steam Safety Valve.
,
NWR B 69828 - Trevitest, per BMP 3114-15, of IMS014C, Main Steam t
Safety Valve, hWR B -72965 - Rework to remove excessive Boron buildup, and test,
.
Letdown Orifire Isolation Valve ICV 8149A.-
t NWR B 73382 - Maintenance on 2PT-FW0525, replace transmitter as.
required.
NWR-B 73588 - Replace 3L Jerk Pump on Diesel Generator e'id test.
r The insp^ectors periodically monitored the licensae's' work in progress and verified performance was in accot A nce with proper j
procedures, and approved work packages, that 10 CFR 50.59 and other applicable drawing updates were made-and/or planned, and
'
that operator training was conducted in a reasonable period of time.
-
a
.
During the maintenance and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance runs of the'1A
.
diesel generator (DG),.several fuel leaks-on fuel injector pumas-
(jerk pumps) were encountered. - Following is a chronology of tie-fuel leaks during.the 1990 calendar year:
' January 25 - fuel leak on jerk ) ump for the 4R cylinder
.
.
,
January 26 - 9:10 a.m. - a crac t was identified on jerk pump
,
L connaction for cylinder 10L. Pump replaced.
11:53 a.m.
- fuel leak on jerk pump for the 10L cylinder.
January 28 - fuel' leak on jerk pump for the 3L cylinder.
January 30 - fuel leaks on the jerk pumps:for the 3L and 7L cylinders.
L Investigation by the licensee identified the preliminary cause as a
-
deficier,t installation procedure. The procedure utilized to connect the Mgh pressure fuel line hose at the jerk pump required a torque of 15.ft-lbs and then another 1/4 turn on the connector.
the
licensee concluded that the torque applied was of sufficient force to cause a crack in the connector on the jerk pump. The procedure
!
,
.
.
.
.
i
_
- _ _ _ _ _ _ _ _ _ _ _ _ __
r.
.;
.
,
'
-
,.
.
for the installation of the high' pressure fuel line hose had been revised in April 1989, to include information obtained from an owner's group meeting. Previous installation of the hose to the jerk pump was considered within the skill of the craft and consisted
of tightening the connection finger tight and then turning the connector another 1/4 turn. The utilization of the skill of craft method had not resulted in any prevalent leaks as experienced with tha torquing method. After the leaks between January 25 - 30,_1990, the licensee performed a dye penetrant on all jerk pump connections
'
and connected the high pressure fuel line hose to the jerk pump with the previous skill of craft method. DeviationReport_(DVR) 6-1-90-021 m
was issued to document the fuel leaks _that occurred between January 25-
- 30,'1990. The completed DVR wil'1 be evaluated by the resident inspectors for appropriate root cause and corrective action.
The material control of 1A DG parts during the 5 year inspection was noted by the inspectors as above average. Maintenance personnel had 1A DG parts segregated, labeled and protected to prevent damage.-
Use of bins for storage and segregation were utilized to the maximum extent possible, b.-
SurveillanceActivities(61726)
The. inspectors observed or reviewed surveillance tests required by Technical Specifications during theJinspection period and verified that tests were performed in accordance with adequate procedures, test instrumentation was calibrated, limiting conditions.for operation were met, removal and restoration of the affected components were accomplished, results conformed to Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and any deficiencies identified during the tests were properly reviewed and resolved by appropriate management personnel.
The inspectors also witnessed portions of the following activities:.
IBVS 8.1.1.2.e-2,
"1B Diesel Generator 18 Month Preventative Maintenance Check", Revision 4.
"
IBVS 8.1.1.2.f-13, " Unit 1-1A Diesel Generator 24. Hour Load Run and Sequence Test - 18 Month", Revision 6.
IBVS 8.1.1.2.f-14, " Unit 1-1B Diesel Generator 24 Hour Load Run
and Sequence Test - 18 Month", Revision 6.
- IBVS 8.1.1.2.f-15, " Unit 1-1A Diesel Generator Safe Shutdown Sequence and Single Load Rejection Test -
18 Month", Revision 7.
The above surveillances were we11' coordinated with shift briefings conducted prlor to surveillances IBVS 8.1.1.2.f-13 and.BVS 8.1.1.2.f-15.
Adequate coverage by the technical staff existed with l
any anomalies discussed with operating shift personnel. During the
_
- -,
- - -.....
Y
.- -
.
,
a
~\\
review of activities for IBVS 8.1.1.2.e-2, the inspectors noted. -
the use of a Turnover Log by personnel that performed the 18 month inspection of the IB diesel generator (DG). The Turnover Log was completedattheendofeachshift_anddocumented;(1)workcompleted on the shift (2) suggested work activities for the next shift and (3) any additional data the oncoming shift should know. The use of the Turnover Log ensures good communications between shift and could
- be utilized for work planning on future DG work.-
During this inspection period, the licensee failed to perform a tsurveillance in the required time. Technical Specifications 4.9.9 required the Containment Purge Isolation system to be demonstrated operable within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to start of core alterations. On January 18.-1990, the Technical Staff was notified at approximately 7:10-a.m. that the outage activity of unlatching control rods (core alteration)wasinprogress. The last time the containment purge
,
isolation was' demonstrated operable was on January.12,1990, at 9:47 a.m.,'approximately 142 hours0.00164 days <br />0.0394 hours <br />2.347884e-4 weeks <br />5.4031e-5 months <br /> prior to the commencement of core-alteration 'on January 18; 1990. The inspectors will review the applicable DVR/LER for adequate root cause and corrective action.
c.
Assessment of Maintenance / Surveillance The licensee.'s overall performance in this-area continues to be above average. The surveillance program continues to be well managed. -The missed surveillance in this inspection period was the first. surveillance missed in 427 days. The major surveillances
- witnessed by the inspectors were well coordinated with adequate shift briefings-prior to performance.
No violations or deviations were identified.
6.
Engineering &TechnicalSupport(37700)
a.
. As stated in Section 3.b of this report, a reactor startup was aborted on February 11, 1990, due to the difference between the predicted critical position and the estimated: critical position:.
exceeding the 500 pcm administrative ~ limit. On January 18, 1990, a successful reactor startup was performed after a reactor trip, with the estimated critical position close to the actual critical'
position. The inspectors determined the estimated critical position for the January 18,1990, _ reactor startup was calculated with revised xenon and rod worth curves furnished by the licensee's Nuclear Fuel Services organization that was based on actual core burnup. The decision to revise the xenon and rod worth curves was based on the fact that the. core was in transition between'" middle of life" to
"end of life".
However, the estimated critical position for the aborted reactor startup on February 11, 1990, was calculated with a revised xenon curve but not with a revised rod worth curve. The failure of the licensee's nuclear engineers to use a revised rod worth curve based on actual core burnup did not appear appropriate since the successful startup on January 18, 1990, which utilized a revised rod worth curvc preceded the February 11, 1990, aborted
.o
..
.
-
.
_ _::
0,
-l x
.
.
.
.'
i-startup by approximately 21 days (472.2 effective full power hours),
l The licensee issued Deviation Report (DVR) 6-2-90-006 for the aborted startup on February 11, 1990. The inspectors will review t
the' closed DVR for proper root cause and corrective action. The licensee has stated that any future reactor startups would use-
,
the appropriate revised curves in the calculations for estimated
. critical positions. This item will be tracked by Open Item
-
,
-455/89024-01(DRP)thatwas-issuedfortheNovember 21,'1989,-
.,
aborted startup.
-(
i b.
Installation and Testing of Modifications (37828)
}
The inspectors reviewed several of the licensee's modifications packages and performed visual walkdown inspections to' assess-the
,
i implementation'of the licensee's modification program.
e
L The inspectors verified that modification work psckages were properly controlled and documented, that prior management approval
!
was received before initiation of work, work activities were l
performed by qualified personnel and that adequate 50.59 reviews-
L were performed.
'
The inspectors also ascertained that approved procedures were used-for modification installations, the correct equipment model and
,
materials were used, and post-modification testing was performed
'
and acceptance criteria were met.. Visual field inspections of the
,
following modifications were performed to verify that installations
'
conformed to the as-built drawings: M6-1-88, " Installation of
Flushing. Lines With Manual Isolation Valve", and M6-2-87-102,
<
" Auto Make-up to Component Cooling Surge Tank." No discrepancies
'
L were identified.
c.
Assessment of Engineering / Technical Support
The licensee's performance in this area was acceptable. However, j
,
the failure to utilize a revised rod worth curve that resulted in an aborted reactor startup co February -11,1990, is another-example t
,
of the importance of adequate interface between off site engineering
-
organizations and the station's technical staff.
Inspection Report
50-454/89021; 50-455/89024 also identified a concern with adequate
interface with off site engineering organizations.
No-violations or deviations were identified.
'
7.
Refueling and Spent Fuel Pool Activities (60710, 86700)
s
The_ inspectors observed or reviewed the Unit i refueling and associated
Spent Fuel Pool activities to verify the licensee had implemented
controls for the conduct of refueling operations and for maintaining
control of plant. conditions, in accordance with the requirements of
TechnicalSpecifications(TS)and10CFR50,AppendixA.
.
1
o
,.....
.
p
e
m
...
a
>
. *
-
.
P
'l
t
-
.
<l4
'
.
.
The inspectors interviewed key licensee and contractor personnel
regarding responsibilities, understanding of administrative and
surveillance requirements and responses, prerequisites for refueling,
equipment checkout, fuel receipt and inspection, and overall management
direction and involvement. Observations of the activity were completed
in the control room, fuel building and the containment.
During this Unit 1 outage, all of the fuel was unloaded from the. reactor.
.
nioved to the spent fuel pool, ultrasonically tested for indications of
fuel: leaks, stored in the spent fuel pool, and then reloaded into~ the
,
reactor as required for the next fuel cycle.
The refueling activity was initiated on schedule, completed essentially
-
on schedule and proceeded in accordance with the plan and requirements
except for the two following anomalies which the licensee identified:.
a.
On January 22, 1990, the licensee discovered that spent fuel rack
,
location C-005 was already occupied by another fuel assembly. Fuel
moves were stopped and a. licensee investigation was conducted.
.l
The review determined that no fuel assemblies should have been at
'
location C-D05. Tag boards and spent fuel maps were compared to the
spent fuel rack. The comparison determined that fuel assembly, DZ1,
had been pla d foi ' *:ation C-DOS instead of the assigned location
of B-005.
The liceva
revised the records to reflect the new
location for fuel asseitly 021. The licensee plans to conduct an
i
" error. board", whose me1bers include upper management of the station
.l
and the individuals invtivec in movement of fuel assembly DZ1, to
ascertain root cause and di', cuss possible corrective actions. The
inspectors. reviewed the UpUated Final Safety Analysis Report and
determined that the wrong location for fuel assembly DZ1 was bounded
by the analysis for misplaced fuel assemblies in the spent fuel
storage rock.
i
b.
During the reconstitution of fuel assembly C40, in the spent fuel
pool,- the licensee attempted to remove fuel rod A12 from assembly
-C40. ' Fuel rod, A12, was-removed because of a previous identified
.'
abnormal: condition:of the rod.
During the rod removal process the
rod sovered at a point of 104 inches from the bottom.1This. portion
of the rod was stored in the failed rod storage basket.
The site of
the defect was observed to be at the bottom' grid location and was
apparently due to grid rod fretting. The defect allowed additional
water to enter the rod and cause secondary hydriding of the rod.
The secondary hydriding contributed to the development of a brittle
condition of the rod which severed during rod removal.
Visual-
examination by the licensee of the rod segments showed no evidence
that fuel pellets had fallen from the rod segments. The inspectors
reviewed Onsite Review (OSR) 90-03 and determined the content as
acceptable. The OSR included a Westinghouse letter that addresses
the industry. experience with failed fuel rods and proposed storage
requirements for the failed fuel rod.
c.
Assessment of Refueling / Spent Fuel Pool Activities
The licensee overall performance in this area was satisfactory.
.
9l,
.
- - -. - -
.
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ^ ^ - - - - - - - - ^ ^ ^ ' - " " ^ ~ ^
-
m
1..
.
- .
[
e
.
..x
e
No violations or deviations were identified.
'
i8.
Temporary Instruction (TI) (2515/101)
-
(Closed) TI 2515/104:
Inspection of licensee actions to prevent and, if
necessary, respond to loss of decay heat removal during operations with
the reactor conlant system (RCS) partially drained. The inspectors
,
'
verified the following:
,
a.
Training had been conducted _by the licensee to ensure personnel were
aware of the' risks associated with a partially drained RCS during
shutdown activities.
The inspectors randomly selected six
individuals from the' operations staff and-reviewed the training
records, included in the review was the= lesson plan. The lesson
plan was comprehensive and included the pertinent aspects of Generic-
Letter 88-17, industry experience (events) associated with a
,
partially drained RCS and various operation configurations affecting-
or potentially affecting a partially drained RCS.
i
L
b.
Controls were established to ensure that containment closure would
_be achieved before uncovering the core. Procedure 1/2 BOP RC-4a,
L.
'! Reactor Coolant System Drain", Revision 2, requires that
containment closure capability be established prior to a reduced
>
inventory level in RCS. The licensee utilized a " Containment
L
- Closure capability Checklist", that required all cables and hoses
through the air lock doors to have quick disconnects. Also, the
l
checklist required that at least one barrier (isolation valve, blank
flanges, etc.) was av'ilable and capable of isolating any path of~
a
radioactive material release from containment atmosphere to areas
outside containment. The procedure also stated that the Shift
control Room Engineer should update the checklist as required to-
maintain a current status of all containment penetrations while the-
RCS was in a reduced inventory condition.
Inspectors also verified
~
that abnormal procedure, 1/2 BOA-PRI-10, " Loss of RH Cooling Unit
i
1(2)", Revision 54, required necessary containment closure.
'
' Procedure 1(2)BGP100-6,"RefuelingOutage", Revision 4.-requires:
c.
two independent core exit thermocouples in service whenever the RCS
,
was intentionally drained-and the reactor vessel head was in place.
'
Also,1(2)BGP100-6,"RefuelingOutage", Revision 4.requiresthat
the time be minimized between when the temperature indicators were
disconnected and the reactor vessel head was removed. The inspector
verified that the temperature was required to be recorded every
shift.
'
d.
Two independent RCS water level indications were required to be
,
available when the RCS was in a reduced inventory condition.
J
-
Specifically,' procedure 1(2) B0P-RC-4a requires two functional
independent methods of level indications prior to any draining of
the RCS below 405'(397' and below considered reduced inventory
condition). Discrepancies between the level indications were
required to be resolved prior to further RCS draining. With level
between 399' and 399'6", an hourly check of the tygon level was
required, and with level below 397' (reduced inventory condition),
,
I
,
. -
. -
o
A
y
,
- ,;
.
q.i
f
y
s
,
w
-
.
,
'9
,
the tygon level has to be continuously monitored. Another
independent level transmitter to monitor level in the mid-loop
' region'(392' --402' 6") was planned for installation en Unit 1
.-
during the current outage and on Unit 2 during the refueling outage-
scheduled for September 1990. Also, the licensee designed a
graphic. display that depicts real time data such as the four
nP
refueling level instruments, pressurizer level,-RHR pump suction
-
and discharge pressure, discharge temperature and motor bearing
-temperatures.. The data was displayed on a computer screen located
'on the Main Control Board next to the RHR system controls. The
,
inspectors also walked down the tygon used for level indication
'
with no problems noted.
e.
Procedure B0P-RC-4a restricts any' parallel outage activities which
could affect the stability of the RCS support systems or containment
-
closure capability. Any such. activities must have approval from the
SCRE.
,
.
.
.;
f.
Procedure 80P-RC-4a requires that two means of _ adding inventory to
the RCS-must be _ established prior to entry into a reduced inventory
condition.
-t
.
p
g.
Procedure 80P-RC-4a requires that appropriate vent paths ba
L
established.if loop nozzle drains were utilized. Prior to entering
p~
'
a reduced inventory condition and every four hours thereafter,
acceptable RCS configuration is verified with a "RCS Configuration
Check Flowchart",
i
No violations or deviations were identified.
[4
L-
9.
Meetings and Other Activities
,
i+
n
- ,
a.
ManagementMeetinis(30702)
L
On January 11, 1990, B. Clayton, Chief, Reactor Projects. Section IA,
p
and the NRC _ resident inspectors toured the Byron plant and met with
.
H
L11censee management to discuss plant perfonnance and plant material
L,
condition
k
b.
ExitInterview(30703)
- The inspectm4 met with the licensee representatives denoted in
-
paragraph 1 during the inspection period and at the conclusion of
n
the inspection _ on February.16,1989. The inspectors summarized
the scope and results of the inspection and discussed the likely
_ content of this inspection report. The licensee acknowledged the
.!
L,,
information and did not indicate that any of the information
1: '
disclosed-during the inspection could be considered proprietary in
nature.
!
s
b
ls
.
,
'
,
,
j
[