ML19256D258
| ML19256D258 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/10/1977 |
| From: | Higgins J, Martin T, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19256D184 | List: |
| References | |
| 50-289-77-11, NUDOCS 7910170770 | |
| Download: ML19256D258 (12) | |
See also: IR 05000289/1977011
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U.S. NUCLEAR REGULATORY COMMISSION
0FFICE OF INSPECTION AND ENFORCEMENT
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Region I
Report No.
77-11
Docket No.
50-289
Category
C
License No. l'R-50
Priority
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Licensee:
Metropolitan Edison Comoany
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P. O. Box 542
Readino, Pennsylvania
19603
Facility Name:
Three Mile Island, Unit 1
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Inspection at: Middletown, Pennsylvania
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Inspection conducted: April 4-19, 1977
/o[77
Inspectors:
gins
date signed
/J.' C.p& k
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T. T. Martin
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' date signed
1-
date signed
Approved by:
P.6.1 M h
S*/ /ol7 7
E. C. McCabe, Jr., Chief, Nuclear
date signed
Support Section No. 1, RO&NS Branch
Inspection Sumary:
Inspection on Apri's 4-19, 1977
(Report No. 50-289/77-11)
Areas Inspected:
Routine unannounced inspection: refueling activities, including
fuel handling, core verification and operator training; diesel generator maintenance
and sampling system valve replacement; planned testing of systems disturbed during
refueling; snubber surveillunce; and, integrated leak rate testing.
The inspection
involved 119 inspector hours on site by two NRC inspectors.
Results:
Five areas were inspected, with no noncompliances found in four.
One
apparent item of noncompliance was found in one area (deficiency - failure to
implement integrated leak rate test procedure - Paragraph 8.g.(2)).
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DETAILS
1.
Persons Contacted
The below technical and superviscry level personnel were contacted.
a.
Metropolitan Edison
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Mr. R. Barley, Lead Mechanical Engineer
Mr. N. Brown, Administrator, Nuclear and Technical Training
Mr. J. Colitz, Unit 1 Superintendent
Mr. W. Cotter, Supervisor, Quality Control
Mr. R. Harper, I&C Maintenance Supervisor
Mr. C. Hartman, Engineer
Mr. N. Hernoisey, Maintenance Foreman
Mr. J. Heverling Security Specialist
Mr. J. Hillbish, Lead Nuclear Engineer
Mr. D. Huffman, Technical Support Engineer
Mr. G. Kunder, Supervisor of Operations
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Mr. E. Lawrence, I&C Foreman
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Mr. J. Moran, Licensing Engineer
Mr. J. O'Hanlon, Unit Superintendent of Technical Support
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Mr. J. Potter, QC Assistant
Mr. M. Shatto, Nuclear Engineer
Mr. B. Smith, Shift Superintendent
Mr. R. Summer, Nuclear Engineer
Mr. R. Van Stry, Administrator, Nuclear and Technical Training
b.
Other Licensee Personnel
Mr. R. Ely, Gilbert Associates Engineer
Mr. M. Nelson, General Public Utilities Service Corporation
Engineer
Mr. R. Shirk, Gilbert Associates Engineer
c.
NRC Accompanying Personnel
Dr. D. Lurie, Applied Statistics Group, Executive Director for
Operations, USNRC (April 14-16,1977)
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2.
Previous Inspection Item Uodate
(Closed) Unresolved Item 289/77-09-12:
Review of Surveillance
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Procedure 1303-6.1, Revision 3, dated April 9,1977, against
Appendix J to 10 CFR 50, American National Standard N45.4-1972 and
Technical Specification 4.4 showed that concerns noted in item 77-
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09-12 were adequately addressed. Additionally, several systems
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identified as not properly vented were corrected by the issuance of
a procedure Temporary Change Notice.
The inspector had no further
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questions in this area.
3.
Refueling Activities
Log and record review, personnel interviews, and video tape viewing
was employed to verify accomplishment of the following.
a.
Fuel handling crew training.
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b.
Pre-refueling equipment surveillance tests.
c.
Containment Integrity Establishment.
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d.
Core criticality monitoring during new fuel insertion.
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e.
Periodic Boron concentration determination.
f.
Fuel accountability records maintenance,
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g.
Daily refueling equipment and cavity status surveillance
checks.
h.
Licensed operator assignments.
i.
Core assembly final verification.
The refueling was completed in five days with no significant pro-
blems encountered.
No unacceptable conditions were identified.
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4.
Refueling Maintenance
a.
Work Requests Selected for Revier
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(1) WR#16974 - Replacement of Diesel Control Circuit BFD
Relays.
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(2) WR#19361 - Inspection and Replace,.ent of Diesel Cam
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Shaft and Rollers.
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(3) WR#19524 - Replacement of Secondary Sampling System
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CA-V54 Valve Seats.
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b.
Acceptable Areas
No unacceptable conditions were found in the areas of work
approval, quality assurance requirements, redundant system
,
operability checks, post-maintenance test requirements,
jumper control, and maintenance personnel qualifications.
c.
Unresolved Area
}.
The licensee designated First Class Electrician and helper
performing the BFD relay replacement were observed by licens-
ee QA and the NRC inspector. A procedural fault was detected
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by the observers and a procedure change was effected in the
prescribed manner. When a second fault was detected, the
entire procedure was reviewed, resulting in identification
of a third fault. Completion of the procedure identified
two more faults, one of which had been caused by the pro-
cedure changes. Actual equipment performance was evaluated
to be proper. The inspector questioned the validity of the
already accomplished use of this procedure on the
"A" diesel,
especially the assurance of operability in all modes.
This
is Unresolved Item 77-11-04.
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5.
Startup Planning
Plans to verify, before startup, operability of systems disturbed
during the outage were reviewed, including post-maintenance testing,
selected precritical checklists, and the plant heatup and startup
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procedtres. No unacceptable conditions were identified by this
review,
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6.
Snubber Operability Surveillance
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Examination of the licensee's snubber test rig and observation of
a demonstration of its capabilities indicated that test capability
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was limited to the ability to check snubber stroke and verify lock-
up and bleed capability without defining the velocities involved.
.
The licensee later obtained and the inspector witnessed use of a
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functional test de/ ice which showed that some snubbers did not meet
test criteria. This area will be reinspected.
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7.
CILRT Chronology
1
The inspector witnessed the initial periodic CILRT at Three Mile
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Island, Unit 1.
A general chronology of the witnessed events
follows:
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4/13
Initial Containment Inspection - Reactor Building
interior not ready for CILRT.
4/14
Containment Inspection completed satisfactorily.
4/15 0500
Commenced pressurization.
1130
Pressurization stopped for internal inspection per
procedure at 12 psig.
No problems noted.
4/16 0600
Reactor Building p assure at approximately Pa + 0.2
psi. Commenced stabilization period.
1000
Stabilization period complete.
1230
Initial leak rate measurements indicate mass point
leakage of 0.2%/ Day.
Licensee began leak searches.
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2100
After 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> of data mass point technique leakage
is 0.1847%/ Day with a 95% confidence interval of
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0.0160%/ Day.
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2115
Licensee has corrected sore packing and fitting leaks
during the leak surveys.
(e.g. PP-V46 packing leak,
BS-V37C and BS-V37D reducer leaks)
2300
Leak rate appears to have decreased.
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4/17
0245
Pressure has dropped approximately 0.2 psi.
License
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repressurized Reactor Building to Pa + 0.2 psi.
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0730
After 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> leak rate is approximately Lt.
2130
17.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> data indicates a mass point leakage of
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0.1094%/ Day with a 95% confidence interval of 0.0083%/
Day.
2230
Commenced pressurizing steam generator "A" to 50 psig.
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2400
During the day, licensee had conducted extensive leak
searches using SNOOP leak detection fluid, ultrasonic
detectors, flowneters and pressure gages.
Valve line-
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ups have been shifted several times and some leaks have
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been repaired without performing local leak rate checks.
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These repairs included the atmosphcric flanges down-
stream of LR-V3 and seven leaks behind the leak rate
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test panel.
4/18
0100
There appears to be possible diurnal bias in the test
data with higher leak rates during the day (warm,
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clear and sunny) and lower leak rates at night (cold
and clear). The licensee is conducting various evolu-
tions that also could affect the leak rate, making it
difficult to sort out the various effects.
0300
Licensee has depressurized steam generator "A" and
rechecked system lineups.
Commencing new 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
CILRT.
0630
Mass values again dropping off with daylight.
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1320
Licensee has positioned equipment hatch shield in
an attempt to limit diurnal cycling of containment.
1800
Mass values leveling off with sunset.
2400
Mass vaks showing an increasing trend although
there is noticeably more data scatter than during
the day.
4/19
0300
After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of data, mass point leakage is
0.0553%/ Day with a 95% confidence interval of
0.0098%/ Day.
0730
Commenced superimposed leak test.
1200
Calculated leakage is below the lowest acceptance
criterion for the verification test after 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
This could be due to weather conditions negating
the diurnal effect previously measured.
Today is
cloudy and not as warm as were the previous 2 days.
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1530
Eight hour verification test completed. At approx-
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- imately 1330 the sky cleared and mass values began to
drop. This duplicated weather conditions during the
24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> CILRT.
Preliminary calculations indicate
that the superimposed leak rate falls within the
acceptance criteria.
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8.
CILRT Results
a.
Initial CILRT Failure
Test data taken on April 16, 1977, with the Reactor Building
at Pa (50.6 psig) indicated a leakage by mass point analysis
of 0.1847%/ Day.
Since packing and fitting leaks were repaired
without preliminary local leakage tests, the leakage from indi-
vidual paths is unknown. Also, valve lineups were shifted and
the steam generators were partially pressurized with instrument
air.
Both section III.A.1.(a) of Appendix J to 10 CFR 50 and pre-
cautions 3.5 and 3.6 to S.P. 1303-6.1, Three Mile Island Reactor
Building Integrated Leak Rate Test Procedure, prohibit leak
repair during a continuing CILRT.
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The initial CILRT was consequently evaluated by the inspector
as a failure due to:
approximately 2 days of data indicating a leak rate in
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excess of the acceptance criteria; and,
the many fitting and packing leaks corrected after the
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start of the test.
b.
Satisfactory CILRT
The Three Mile Island CILRT was conducted pursuant to Surveil-
lance Procedure 1303-6.1, Revision 3, dated April 9,1977.
The final CILRT was conducted from 0300, April 18,1977, to
0700, April 19,1977. A very noticeable diurnal effect in the
measured mass data was experienced.
Preliminary analysis of
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the test data from 0300, April 18,1977, to 0700, April 19,
1977, indicates that this CILRT met the acceptance criteria.
The inspector's independently calculated leak rate by the mass
point technique was 0.0553%/ Day with a 95% upper confidence
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level limit of 0.0651%/ Day. The data from 0500, April 18,1977,
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to 0500, April 19,1977, indicates a lower leak rate of 0.042%/
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c.
Verification Test
Verification test results appear to be satisfactory with a
measured , leak rate of 0.099%/ Day.
The verification test accept-
ance c.iteria are:
(Li + Lam
.25 La) < Lvm < (Li + Lam + .25 La) where:
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Li
= Superimposed Leak Rate = 207 SCFH = 0.056%/ Day
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Lam = Measured containment leak rate = 0.055%/ Day during
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the CILRT
La = Maximum allowable leak rate = 0.1%/ Day
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Lvm = Measured leak rate during the verification test
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= 0.099%/ Day
Thus:
0.086 4 0.099 < 0;136.
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d.
Acceptable CILRT Areas
Except as noted in other sections of this report, the following
areas were identified as acceptable.
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(1) General inspection of accessible interior and exterior
surfaces of the containment structures and components
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prior to the CILRT for indice2 ions of Structural dete-
rioration.
(2)
Inspection of reactor building to verify that no high
pressure sources were present and that instrumentation
was suitably located.
(3) Review of instrument calibration records.
(4) Witnessing of licensee data takers during the CILRT's.
(5)
Independent verification of raw data conversion to com-
puter inputs.
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(6) Review of various system lineups for proper system venting
and imposition of any artificial leakage barriers.
(7) Witnessing of quality control participation in CILRT.
(8) Verification that test prerequisites were satisfactorily
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met.
(9) Witnessing of performance of test and control room personnel
during the CILRT, including determination of proper manning
with qualified personnel, proper use of procedures and main-
tenance.of the Log of Events.
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(10) !?tness of leak searches of the pressurized containment
volume with leak detection fluid, ultrasonic detector,
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pressure gages and flowmeters.
(11) Icposition of the superimposed leak rate during the verifi-
cation test.
(12) Independent calculation of test results from raw. data.
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e.
Type C Testing
The licensee currently does not Type C test CIV's in systems
not vented during the Type A test and normally filled with
water and open to the containment under post accident con-
ditions as required by Section III.A.I.(d) of Appendix J.
He has, however, requested an exemption from NRR. The in-
spewtor had no further questions in this area.
f.
Reactor Building Inspection
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During the initial Reactor Building inspection on April 13,
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1977, the inspector identified several areas of concern:
bottles of pressurized gas in the Reactor Building, some " lay
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flat" type plastic was impeding venting of containment isola-
tion valves; some instrumentation was not placed per the pro-
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cedure; puddles of water were on the reactor building floor;
and, weld channels used during.preoperational testing of the
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liner were not vented.
During the second inspection on
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April 14,1977, the inspector confirmed that the above areas
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were corrected except for the weld channel venting. The
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licensee had vented readily accessible weld channels (34 of
an estimated 75 total).
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Since the licensee was unable to show that the ueld channels
would survive a Design Bas {s Accident, they constituted a
potential artificial leakage barrier during the CILRT. This
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item (77-11-02) will be referred to NRC management for further
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review.
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g.
Valve Lineups
Inspector sampling check of valve position conformance to pro-
cedure valve lineup 3.heets and of venting of systems external
to the test vohme to prevent artificial leakage barriers
identified no inadequacies, except as noted in the following.
(1) During the leak search, pressure gages were placed on
the vents of two zones of the PP system to aid in leak
detection.
These constituted artificial leakage barriers
and were removed for the final CILRT.
During a Reactor
Building tour the inspector noted a pressure buildup in
part of the PP zone to the equipment hatch door seal's.
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That zone of the PP system was vented but a shut pressure
control valve isolated the portion between the door seals
from the vent path.
The licensee vented this area and
also the corresponding area of the PP system associated
with the personnel air lock. The inspector had no further
questions in this area.
(2) Valve LR-V7
At approximately 5:30 a.m. on April 17, 1977, valve LR-V7
was found shut when required to be open by step 6.4.e.(4)
of S.P. 1303-6.1.
This valve provides the vent path for
the containment isolation valve LR-V1.
The licensee sub-
sequertly opened valve LR-V7.
This is an item of non-
compl'ance of the Deficiency level (77-11-01).
h.
Type C Correction to Test Results
The licensee operated the Reactor Building Industrial Cooling
System during the Type A test.
The containment isolation
valves of this system were therefore not subjected to the post
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accident differential pressure, necessitating correction of
the Type A test results based on Type C tests of containment
isolation valves RB-V2A and RB-V7.
The recently completed
Type C tests of the valves totaled 12,380 SCCM or 0.0071%/ Day.
This leak rate was added to the preliminary Type A results to
obtain a corrected mass point leak rate of 0.0624%/ Day, which
is less than the acceptance criteria of 0.075%/ Day.
The
corrected leak rate at the 95% upper confidence level is
0.0722%/ Day, which is also less than the acceptance criteria
of 0.1%/ Day. The inspector had no further questions in this
area.
9.
Future CILRT Schedule
Section III.A.6.{a) of Appendix J to 10 CFR 50 requires that the
Commission review and approve the licensee's schedule applicable to
subsequent Type A tests, if any periodic Type A test fails to meet
the acceptance criteria. As stated in detail 7.a. the licensee
failed his initial CILRT attempt; the schedule for future Type A
tests is designated unresolved item 77-11-03.
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10. Unresolved Items
Items about which more information is required to determine accept-
ability are considered unresolved.
Paragraphs 4.c, 8.f and 9 of
this report contain unresolved items.
11. Exit Interview
A management meeting was held at Three Mile Island Nuclear Station
on April 19, 1977, with Mr. J. Colitz to describe the findings of
the inspection, including the Item of Noncompliance, the unresolved
items and the CILRT failure.
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