ML20040G097
| ML20040G097 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/22/1982 |
| From: | Chung J, Fasano A, Haverkamp D, Young F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20040G093 | List: |
| References | |
| 50-289-81-32, IEB-80-24, IEB-81-01, IEB-81-02, IEB-81-1, IEB-81-2, NUDOCS 8202110254 | |
| Download: ML20040G097 (14) | |
See also: IR 05000289/1981032
Text
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U.S. NUCLEAR REGULATORY COMMISSION
50289-811008
50289-811009
Region I
50289-811125
50289-811116
Report No. 50-289/81-32
Docket No. 50-289
License No. DPR-50 -
Priority
__
Category
C
--
Licensee :
GPU Nuclear Corporation
P.O. Box 480
Middletown, Pennsylvania 17057
Facility Name: Three Mile Island Nuclear Station, Unit 1
Inspection at: Middletown, Pennsylvania
Inspection conducted: November 17, 1981 - January 8, 1982
Inspectors :
/
[
lfl9 [f t-
J. Chung, Reactor In$ hector
date signed
b/2 / wl---b
tfe4 f t,-
D. Haverkamp, Seniorflesident Inspector (TMI-1)
date signed
~
%
1)20lt2
F. You , , Resi
nt Inspector (TMI-1)
dat'e signed
Approved by: [
/oe
/ /f_z/lr2
A. Fasano, Chief, Three Mile Island Section
date signed
Projects Branch No. 2
Inspection Summary:
Inspection on November 17, 1981 - January 8,1982 (Report Number 50-289/81-32)
Areas Inspected:
Routine safety inspection by resident and regional based
inspectors (282 hours0.00326 days <br />0.0783 hours <br />4.662698e-4 weeks <br />1.07301e-4 months <br />) of. licensee action on previous inspection findings;
plant operations during long term shutdown including facility tours and log
and record reviews; steam generator tube degradation; small fire in TMI-l
Auxiliary Building; IE bul.letin followup; and licensee event reports - in-office
review.
Resul ts : No items of noncompliance were identified in six areas and one item
of noncompliance was identified in one area (failure to follow a welding and
cutting procedure, paragraph 5.c).
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8202110254 820125
{DRADOCK05000g
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Detail s
1.
Persons Contacted
General Public Utilities (GPU) Nuclear Corporation
B. Ballard, Manager TMI Quality Assurance (QA) Modifications /
Operations, Nuclear Assurance
R. Barth, Engineer-II -TMI-l
'
- M. Beers, Engineering Assistant Senior II, Nuclear Assurance
J. Colitz, Plant Engineering Director. TMI-l
- W. Heysek, Supervisor Site QA Audit, Nuclear Assurance
W. Miller, Nuclear Licensing Engineer, Technical Functions
- T. O' Conner, Lead- Fire Protection Engineer TMI-l
- C. Stephenson, Nuclear Licensing Engineer, Technical Functions
- D. Shovlin, Manager of Plant Maintenance TMI-l
. The inspector- also interviewed'several other licensee employees
during the inspection. They included control room operators,
maintenance personnel, engineering staff personnel and general
office personnel.
- denotes those present at the exit interview.
2.
Licensee Action on Previous Inspection Findings
(Closed) Unresolved Item 289/79-21-01 : Reactor Coolant System
(RCS) Leak Pate Surveillance Procedure (SP) 1303-1.1 incorrectly
accounted for operator induced changes to the Makeup Tank (MUT).
The inspector reviewed SP 1303-1.1, Revision 8, August 21, 1981,
Temporary Change Notices (TCNs) 1-81-0112 and 1-81-0122, and
1-81-0112 Attachment, and verified that the operator-induced
changes to MUT and RCS inventory were satisfactorily incorporated
into procedural and data sheet steps.
The inspector further
determined that temperature compensation and density corrections
were properly included in the procedure.
Based on these findings,
this item is closed.
(0 pen) Noncompliance 289/79-IR-23:
RCS inventory calculations for
unidentified leakage produced wrong values due to an error in the
calculational procedure. - Surveillance Procedure 1303-1,1 was
scheduled to be reviewed and changes implemented prior to restart
of TMI-1.
The inspector determined by review of SP 1303-1.1,
Revision 8, and independent calculations that the RCS leak rate as
determined by a manual calculation was adequate in the revised
procedure.
The inspector noted that procedural step 6.1 required
use of the process computer for the RCS leak rate determination
unless the process computer inputs and/or the program were inoperable.
A licensee representative stated that a computer software problem
was identifled during hot functional tests and that the RCS leak
rates were recalculated after correcting the programming problems.
The inspector requested a trial run during this inspection and an
additional . software problem was identified.
In addition, the
inspector conducted independent calculations.
The results are
compared in the following table:
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RCS LEAK RATE GPM
LICENSEE
INSPECTOR
PROCESS COMPUTER CALCULATIONS
DATE/ TIME
ORIGINAL
RECALCULATION
MANUAL
CALCULATIONS
<
8-31-81
G 2.24
0.169
0.18
--
01:59:49
U 2.01
- 0.045
-0.10
--
12:19:39
G 0.34
0.15
--
0.19
0 0.13
-0.06
--
-0.12
9-1-81
G 0.73
0.51
0.52
--
16:50:09
0 0.52
0.30
--
.0.22
9-2-81
G 0.39
0.04
0.34
--
00:48:14
U 0.17
-0.17
--
-0.05
9-3-81
G 0.71
1.05
1.05
1.13
16:01:27
U 0.50
0.78
0.79
0.57
9-4-81
G 0.76
0.17
0.50
--
04:36:36
0 0.54
0.23
--
0.41
15:39:39
G 0.77 -
0.41
0.62
--
0.42
U 0.54
0.17
--
9-5-81
G 0.93
0.49
0.70
--
07:58:54
U 0.71
0.27
0.58
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9-6-81
G 0.93
1.03
0.80
0.81
08:39:37
0 0.69-
0.82
0.59
.0.67
NOTE: G: Gross Leak Rate
U: Unidentified Leak Rate
Based on these results showing that the inspector's calculations
were consistent with the licensee's manual calculations, the
inspector determined that the manual calculations were adequate.
However, the process computer calculations were erratic.
This item
remains open pending an RCS Leak Rate demonstration of consistency
using the process computer during a subsequent NRC:RI inspection.
3.
Plant Operations During Long Term Shutdown
a.
Plant Status During Inspection Period
The plant has remained in the cold shutdown condition with Reactor
Coolant System (RCS) temperature less than 200 F per NRC
order of August 9,1979.
At the beginning of the inspection
period on November 17, 1981, the RCS was at 94 F and 0 psig
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with cooling to the reactor supplied by the
"B" Decay Heat '
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Removal (DHR) Loop.
The pressurizer was partially filled with
a nitrogen (N ) blanket.
Both Once Through Steam Generators
Z
(OTSG's) were in full wet layup.
During the course of the
inspection the following major plant changes or evolutions
occurred.
November 19, 1981:
Formed a steam bubble in the pressurizer
--
and raised RCS pressure _to 45 psig in order to test
Reactor Coolant Pump 1C. Shifted from "B" DHR Loop to
"A" DHR Loop in order to perform In-Service Inspection
(ISI) on Decay Heat Remov0 Pump 1B.
November 21, 1981 : OTS3 taoe leaks determined and RCS
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depressurized to 0 psig.
(Refer to paragraph 4-for more
information on OTSG tube degradation.)
November 22, 1981: Pressurizer par'.ially filled and N2
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blanket established in the pressurizer.
November 23, 1981: Diesel Generator 18 taken ~out of
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service due to high vibration readings on the' rotor.
--
November 25, 1981: Performed flushes on the secondary
side of the OTSG's and placed both OTSG's in wet layup.
November 26, 1981: Drained down RCS to perform N2
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bubble leak te.t and eddy current testing on both OTSG's.
December 1,1981 : Pressurizer partially filled and N2
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blanket established in the pressurizer.
December 12, 1981: RCS was partially drained for additional
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inspections of the OTSG tubes.
program redefined and recommenced using several different
'
type probes.
December 20, 1981: Drained secondary side of "B" 0TSG
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for tube removal.
Portions of four tubes were removed.
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December 23, 1981: Shifted from "A" DHR Loop to "B"
Loop due to a packing leak on Decay Heat Closed Cooling
Pump 1 A.
December 24, 1981 : Temporarily plugged holes from removed
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0TSG tubes and placed both OTSG's in wet lay up on secondary
side.
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January 5,1982 : Shifted to "A" DHR Loop from "B"
Loop due to Diesel Generator 18 still being out of service.
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At the close of this inspection per-iod the RCS was at 120 F
and 0 psig with cooling to the reactor core from "A"
DHR Loop.
The RCS was partially drained-for eddy current testing and
preparation for additional tube removal.
Both OTSG's were in
full wet layup.
b.
Plant Logs and Operat_ing Records
The inspector reviewed selected portions of the following
plant procedures to determine the licensee established require-
ments in this area in preparation for a review of selected
logs and records.
,
Administrative Procedure (AP) 1007, " Control of Records,"
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Revision 4
AP 1010. " Technical Specification Surveillance Program,"
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Revision 18
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AP 10l?, "ihift Relief and Log Entries," Revision 14
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AP 1033, " Operating Memos and Standing Orders," Revision 2
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AP 1037, " Control of Caution and [Do Not Operate] DN0
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Tags," Revision 2
AP 1044, "Etent Review and Reporting Requirements,"
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Revision 2
The inspector reviewed the following plant logs and operating
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records.
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Shift Foreman Log and Control Room Log Book
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Shift Turnover Checklist
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. Temporary. Change Notice (TCN) Log Book
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Active. Tagging Application Book
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)
Locked Valve Log
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Night Order Book
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Do Not Operate and Caution Tag Log
,
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_ Plant logs and operating records were reviewed to verify the
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following items.
Log keeping practices and log book reviews are conducted
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in accordance with established administrative controls.
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. Log ei.trie's inv.olving abnormal conditions provide sufficient-
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t
detail to communicate equipment status, lockout. status,
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correction and restoration.
Operating orders do not conflict with Technical Specifi-
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cations (TS) requirements.
Tagging operations are conducted in conformance with-
^
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established administrative controls.
No items of nonco..spliance were identified,
c.
Facility Tours
During the course of the inspection, the inspector conducted
- tours of the following plant areas.
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Control Room (daily)
. Auxiliary Building (November _18, 20 and December. 2, -19)
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Intermediate-Building (November 17 and December 16)
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Vital Switchgear Rooms
(November 19, 30 and December 14)
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Diesel GeneratokBuilding (November 18 and December 1, 30)
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Yard' Area (November 20 and'Iecember. 2,- 9)
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' Site" Perimeter ( November- 20:and December 2, 9)
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Reactor ' Building (November 24 and December 1, 10, 15, 23)
The following, observations / discussions / determinations were
made.
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Control room annunciators: Selected lighted annunciators
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(Radiation Alarm, B Diesel Trouble Alann, Uninterruptable -
' Power Supply _ Trouble ' Alarm) were: discussed with control-
- room operators to verify. that the reasons for them were.
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understood and corrective action, if required, was being
taken.
Control room manning: By frequent observation during the
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inspectioa, the' inspector verified that control room
manning .equirements of 10 CFR 50.54(k) and the Technical
Specifications were being met.
In addition, the inspector
observed shift turnovers to verify that continuity of
system status was-maintained.
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Technical Specifications: Through log review- and direct
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observation during tours, the inspector verified compliance
with Technical Specification LC0's associated with Decay
Heat Removal System and Fire. Protection Pump requirements
during shutdown plant operation.
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Plant housekeeping conditions : Unsatisfactory plant
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housekeeping conditions observed by the inspector had been
identified previously by station personnel and corrective
action had been initiated as necessary.
Monitoring instrumentation: The inspector verified that
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selected instruments (Source Range Instruments nil and
NI2, Auxiliary Stack Monitor RMA8, and RCS pressure) were
functional and indicated that parameters were within
Technical Specification limits.
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Valve positions : The inspector verified that selected
valves were in the position or condition required by
Technical Specifications for the applicable plant mode.
This verification included control board indication and
field observation of valve positions for the Decay Heat
Removal System and the Core Flood System.
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Fluid leaks: Fluid leaks observed by the inspector had
been identified previously by station personnel and
corrective action had been initiated as necessary.
Piping snubbers / restraints: Selected pipe hangers and
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seismic restraints were observed and no adverse conditions
were noted.
Equipment tagging: The inspector selected plant components
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(Core Flood Tanks) for which valid tagging requests were
in effect and verified that the tags were in place and
the equipment in the condition specified.
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Security :
During the course of this inspection, obser-
vations relative to protected and vital area security
requirements were made, including access controls,
boundary integrity, search, escort, and badging.
No
adverse conditions were noted.
Licensee meetings: The inspector frequently attended the
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Plan-of-the-Day (P00) morning meetings and the daily.
status afternoon meetings held by licensee management and
supervisory personnel.
The inspector _ observed the meetings
to assess licensee evaluation of plant conditions, status
and problems and to review the licensee's plans for
conducting certain major plant operations and maintenance
activities.
The inspector also attended bi-weekly project
status meetings to assess licensee progress and difficulties
related to plant modifications required for restart.
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Acceptance criteria for the above items included inspector
judgement and requirements of 10 CFR 50.54(k), Regulatory
Guide 1.114, Technical Specifications, and the following
procedures.
AP 1002, " Rules for the Protection of Employees Working
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on Electrical .and Mechanical Apparatus," Revision 22
AP 1008, " Good Housekeeping," Revision 7
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AP 1037, " Control of Caution and DN0 Tags," Revision 2
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No items of noncompliance were identified.
4
St_eam Generator Tube Degradation
a.
Background
Late on November 20, 1981, routine chemistry sampling of the
Once Through Steam Generators (OTSG's) detected radioactivity
on the secondary side of the
"B"
0TSG.
The primary system was
at 45 psig pressure'in anticipation of raising pressure to
300 psig to run a reactor coolant pump.
On November 21, 1981,
a Reactor Coolant System (RCS) leak rate surveillance indicated
that the
"B" 0TSG had a primary to secondary leak between
0.3 gpm and 0.5 gpm.
The plant was immediately depressurized
from 45 psig to atmospheric pressure.
Nitrogen (N ) bubble
2
testing was performed on both OTSG's to evaluate the extent of
tube failure.
The licensee determined that there were 38 leaking
tubes in the "B" 0TSG and 86 leaking tubes in the "A" 0TSG. A
flush of both OTSG's upper tube sheet crevices was done to aid
in arresting the corrision attack.
After the _ flush, s second
N2 bubble test was conducted on both 0TSG's and an additional
six tube leaks were identified in the "B" 0TSG.
Portions of
four tubes have been. removed and shipped to B&W in Lynchburg,
Virginia (two tubes)'and Battelle in Columbus, Ohio (two
tubes) for metallurgical analysis.
Due to the severity of the problem, the Vice President and
Director of TMI-l established an OTSG Tube Leak Task Force.
The task force is composed of representatives from TMI-l
(engineering, operations and maintenance), Technical Functions
(corporate engineering and plant licensing), Maintenance and
Construction, Nuclear Assurance, Radiological Controls,
Communications and Babcock and Wilcox (NSSS engineering).
The
purpose of this task force is to coordinate and direct all
actions regarding the investigations and repairs to the steam
generator tube leaks.
At the close of this inspection the
task force was still compiling data to identify the failure
mode and establish the program to repair the OTSG's.
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b.
Review
Due to the significance and severity of the OTSG tube degradation,
the inspector has been closely monitoring the licensee's 0TSG
leakage evalaution program and initiated a review to verify
the below listed items.
Corrective action is appropriate to correct the cause of
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the *.ube degradation
Responsibility has been assigned to' ensure proper management
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attention in correcting the problem
The leakage / tube failures did not cause violation of
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Technical Specifications, license conditions or regulatory
requirements
,
Information related to the event submitted -to NRC is
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,
accurate
Stability of plant conditions, including provision for :
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Proper review and implementation of radiological' controls
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including ALARA
Selected sections of the following documents were. reviewed.
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Radiation Work Permits (RWP's) associated with' initial'
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entries into the upper portions of both OTSG's
ALARA review conducted for OTSG N2 bubble testing,
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installation of video equipment and eddy current testing
in OTSG's
Administrative Procedure (AP) 1103-11, " Draining and N2
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Blanketing RC System," Wevision 12
AP 1104-4, " Decay Heat Removal System," Revision 32
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AP 1106-16, "0TSG Secondary Fill Drain and Layup,"
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Revision 20
In addition to conducting interviews with personnel involved,
the inspector attended several licensee Steam Generator Tube
Leak Task Force meetings to assess the status of actions being
taken.
The inspector also conducted' field observation of N2
bubble testing in the
"B" 0TSG, Eddy Current testing on both
OTSG's on several occassions and fiber optics video recording
on the "A" 0TSG to independently evaluate the situation.
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c.
Findings '
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(1) On December 11,1981, at .1 :25 PM, the inspector entered
the Control Room on a daily tour. While reviewing the-
status of the Steam Generator tube leak problem, the
inspector noticed that the pressurizer heaters were not
tagged out as required by Operating Procedure (0P) 1103-11,
step 2.1.3.
At the time the licensee was in the process-
of draining both OTSG's in accordance with OP 1103-11.
The inspector informed the shift supervisor who immediately
ordered a tagout to be issued for the pressurizer heaters.
Discussion with control room operators revealed that the
tagout of the heaters were to back up a low level pressurizer
heater cutout switch.
The inspector concluded that there
had not been a degradation in plant safety and considered
the item to be an isolated case.
On December 14, 1981,
the inspector audited the tagout for. pressurizer heaters.
and found it to be correctly performed in accordance with
the applicable station procedure.
The inspector had no
further comments in this area..
(2) On January 5,1982, a meeting was held on site between
NRC representatives and key members of the Steam Generator
Tube Leak Task Force to review the scope and status of
the licensee's program.
At the conclusion of the meeting
it was determined that substantial work remained and that
another meeting will be required after more data has
been collected.
At the close of the inspection, the tube failure and
repair program was still in the developmental stage.
The
licensee's continuing efforts to identify the cause of tube
degradation and to complete adequate corrective measures
,
will be reviewed during subsequent NRC inspections (289/81-32-01).
5.
Small Fire in Unit 1 Auxiliary Building
a.
Background
_
At 1:35 PM on December 16, 1981, during welding on the 305' elevation
of the auxiliary building, weld slag material fell through an
open penetration to the 281' elevation and ignited a cloth
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(Chem-wipe) laying on a cable tray below.
The fire watch
immediately put the fire out using a dry chemical extinguisher.
The licensee responded to the fire which included dispatching
the onsite fire team and sounding the onsite fire alarn.
No
offsite assistance was needed.
Dust and smoke resulted in the
area necessitating local evacuation.
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The welding was on a support for a shield with respect to the
additional pipe shielding restart modification.
The licensee
reported no significant damage to the cables in the cable
tray. Also, there were no indications of explosive gas or a
release of radioactivity.
An investigation has been conducted
by the TMI Safety and Health Department staff. Several
recommendations were made in the investigation which have not
been implemented fully at this time,
b.
Review
Licensee activities in response to the fire were reviewed to
assess items listed below.
Safety significance of the event, and compliance with TS
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or other licensee requirements
Reportability of the event and licensee plans regarding
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a press release
Amount of radioactivity released, if applicable
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Event internal management review and followup including
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event description, cause and systems or plant components
affect.ed and overall sequence of events
Immediate and subsequent corrective action
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c.
Findings
After the fire alarm was activated, the inspector proceeded to
the Unit 1 Control Room and then observed the licensee's
actions to extinguish a small fire on the 281' elevation of
the auxiliary building.
The inspector noted that licensee
response to putting out the fire was timely and adequate.
All
members of the fire brigade were present at the scene of the
fire with designated equipment as defined in the licensee's
procedures.
Followup fire site observation by the inspector revealed that
a floor opening located within 40 feet of the welding location
had not been covered.
This is a violation of item D listed on
the Welding and Cutting pennit which had been completed and
signed by proper site personnel .
The failure to cover the
floor penetration is also a violation of section 6.3.1.c of
TMI-l Maintenance Procedure 1410-Y-26, Revision 5, " Control of
Welding, Flame Cutting, Grinding, Brazing and Soldering.
Althougl the fire was not severe, the failure to follow these
fire prevention procedures is a serious matter and is considered
an item of noncompliance with respect to 10 CFR 50, Appendix B,
Criterion V (289/81-32-02).
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Subsequent to the fire, an investigation was conducted by the
TMI Safety and Health Department staff.
The following findings
were identified by the licensee in the investigation report.
Welding permit had been filled out correctly and signed
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by the proper site personnel
Floor openings located directly beneath the welding
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location were not tightly covered which was in direct
violation of the welding permit and Site Procedure 1410-Y-26.
Plywood had been used to cover two of the tnree floor
penetrations and one penetration was not covered at'all
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Contrary to the welding permit, combustible materials
were located within 40 feet of the welding operation with
no protection provided by asbestos curtains, metal guards
or flame proof covers
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Flameproof cloth covers were not available for use in
sufficient quantities from the licensee's maintenance
tool room
Fire extinguisher issued to the fire watch failed to
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function properly.
It was subsequently determined that
the extinguisher had a faulty C02 charge cartridge
Source of the fire was apparently a muslin cloth laying
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on top of a cable tray located underneath the penetration.
It was surmised that the fire was produced by welding
sparks dropping through the floor penetrations and
igniting the muslin cloth
The following recommendations were made by the Safety Department
to plant management.
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Contractor should be required to describe in detail what
actions will be taken to prevent a recurrence of this
event
Fire Protection Engineering should be tasked with reviewing
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contractor's action for adequacy
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An adequate supply of welding cloths should be established
and maintained in both the licensee's and the contractor's
tool room
At the close of the inspection, management's final evaluation
and correcti'.e actions were still under review.
Immediate
steps taken by the contractor had been to terminate the
employment of the two craftsmen responsible for the fire.
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6.
IE Bulletin Followup
The inspector reviewed the licensee's followup action regarding the
IE Bulletins (IEB) listed below.
IEB 80-24, " Prevention of Damage due to Water Leakage Inside
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Containment," dated November 21, 1980
IEB 81-01, " Surveillance of Mechanical Snubbers," dated January 27, 1981
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IEB 81-02, " Failure of Gate Type Valves to Close Against
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Differential Pressure," dated August 18, 1981
With respect to the above bulletins, the inspector verified that
licensee management forwarded copies of the bulletin response to
appropriate onsite management representatives, that information and
corrective action discussed in the reply was accurate and implemented
as described, and that the reply was submitted within the time
period described in the bulletins.
.
Acceptance criteria for the above review included inspector judgement
and requirements of applicable Technical Specifications 'and facility
procedures.
Licensee followup to the above bulletins was acceptable.
7.
Licensee Event Reports (LER's) - In-Office Review
The inspector reviewed the LER's listed below, which were submitted
to the NRC Region I office, to verify that the details of the event
were clearly reported, including the accuracy of the description of
cause and the adequacy of corrective action.
The inspector determined
whether further information was required from the licensee, whether
the event should be classified as an Abnormal Occurrence, whether
the information involved with the event should be submitted to
Licensing Boards, whether-generic implications were indicated, and
whether the event warranted onsite followup.
The following LER's were reviewed:
LER 81-010/03L-0, dated November 10,1981 (Quality control
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audit determined that a Reactor Building prestressing tendon
had an unacceptable lift-off force.
Technical Specifications
require that two adjacement tendons be inspected, which was
not performed)
LER 81-Oll/0lT-0, dated November 4,1981 (Review of a catwalk
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structure for application of additional loads revealed that
seismic. design calculations were not performed by the original
architect engineer)
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LER 81-012/04T-0, dated November 30,'1981 (River Water
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Chlorinator malfunctioned' causing Environmental Technical
Specification limiting condition for total chlorine concen-
tration to be exceeded)
LER 81-013/01 T-0, dated December 9,1981 (Primary to secondary
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tube leakage occurred in both Once Through Steam Generators)
Several blocks and codes on LER 81-010/03L-0 and LER 81 -011/01 L-0
were noted to have minor errors. .The discrepancies were identified
to the licensee and the licensee is in the process of resubmitting
corrections to both LER's.
The inspector has reviewed the draft
resubmittals and found them to be correct.
The above LER's are
closed based on this in-office review 'and correction of reporting
errors.
8.
Ex._i_t Interview
Meetings were held with senior facility management periodically
during the course of the inspection to discuss the inspection scope
and findings.
The inspectors met with the licensee representatives
(denoted in paragraph 1) at the conclusion of the inspection on
'
January 8,1982, and summarized the purpose and scope of the
inspection and the findings.
The licensee representatives acknowledged
the findings.
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