ML19257A841
| ML19257A841 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/17/1979 |
| From: | Kellogg P, Ridd M, Webster E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19257A838 | List: |
| References | |
| 50-338-79-39, NUDOCS 8001090200 | |
| Download: ML19257A841 (12) | |
See also: IR 05000338/1979039
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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REGION 11
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101 MARIETTA ST., N.W., SUITE 3100
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ATLANTA. G EORGIA 30303
Report No. 50-338/79-39 (with attached Investigation Report)
Licensee: Virginia Electric and Power Company
P. O. Box 26666
Richmond, Virginia 23261
Docket No. 50-338
Facility Name: North Anna Unit 1
License No. NPF-4
Inspection at North
na. Sit.e near Mineral, Virginia
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Inspectors:
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M. S. Kidd, Resident Inspector
Date Signed
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E. H. Webster, R,eactor Inspector
Date Signed
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Is R. W. 2ava oski/, Radiation Specialist
Date Signed
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k' xS. C. Ewa'1d, Radiation Specialist
Date Signed
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Accompanying Personnel on Settember 25-26, 1979:
N. C. Moseley, Director, aivision of Reactor Operations Inspection, IE
D. F. Ross, Deputy Director, Division of Project Management, NRR
A. Schwencer, Chief, Operating Reactors Branch No. 1, Division of Operating
Reactors, NRR
J. D. Neighbors, Project Manager, Division of Operating Reactors, NRR
B. W. Sheron, Reactor Engineer, NRR
G. Kwalina, Environmental Branch, Division of Operating Reactors, h3R
G. R. Mazetis, Div,ision of System Safety, NRR
K. R. Mahan, Ope
or L'ce s' g Branch, NRR
Approved by:
. C'
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P. J/ Kefl<fgg ' dert i#rhi af RONS_ Branch
~ yate $igried
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Dates of Inspection. September 25 - October 5,1979
SUMMARY
Areas Inspected
This special, announced inspection involved 78 inspector-hours onsite in the area
of a reactor trip with safety injection and release of radioactive gases to the
environment on September 25, 1979.
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d 0 010 90-EE7
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e
Summary
-2-
Results
Within the area inspected, one apparent item of noncompliance with two examples
was identified (infraction - failure to conduct safety evaluations on changes to
components as described in the safety analysis report
paragraph 5).
>
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DETAILS
1.
Persons Contacted
Licensee Employees
- E. A. Baum, Executive Manager, Licensing and Quality Assurance
- W. R. Cartwright, Station Manager
- J. D. Kellams, Superintendent of Station Operations
- S. L. Harvey, Operating Supervisor
- E. W. Harrell, Superintendent of Maintenance
- D. W. Hopper, Health Physics Supervisor
- E. R. Smith, Superintendent of Technical Services
- B. R. Sylvia, Director of Nuclear Operations
- J. M. Stallings, Vice President-Power Supply and Production
Operations
Three shif t Supervisors
Other licensee employees contacted included five operators.
- Attended one or more exit interviews.
2.
Exit Interview
Meetings were held with licensee management on September 26, 27, 28, and
October 4, 1979, for those persons indicated in Paragraph I above.
Open
and unresolved items denoted in there details were discussed on September 28,
as was the apparent noncompliance involving blocking of the switch for
LCV-1115A. This modification plus the one involving liquid waste line
LW-81-152 and orifice R0-LW-104 were discussed in detail on October 4,
1979. The inspector's comments were acknowledged by station management.
3.
Licensee Action on Previous Inspection Findings
Not inspected.
4.
Unresolved Items
Unresolved items are matters about which more information is required to
determine whether they are acceptable or may involve noncoapliance or
deviations.
A new unresolved item identified during this inspection is
discussed in Paragraph 5.
.
5.
Event Description
On September 25, 1979, Unit 1 experienced a turbine trip / reactor trip with
a subsequent automatic initiation of safety injection (SI), due to excessive
cooldown of the reactor coolant system (RCS). The following chronology of
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significant occurrences during the event was developed by the inspector
using recorder printouts, traces, log reading and operator interviews.
Times given in the left hand column are approximate.
Time (am)
Event
0609
Turbine trip received due to high level in
low pressure feedwater heater 5B; reactor
trip from turbine trip.
0613
Automatic SI initiated on low pressurizer leve
and pressure due to excessive cooldown of the RCS
caused by a condenser steam dump valve sticking
open.
Reactor coolant pumps were manually tripped in
accordance with IE Bulletin 79-06C.
0619
One charging (HHSI) pump was manually tripped
because RCS pressure was returning to normal.
0620
Pressurizer power-operated relief valve (PORV)
actuated several times due to high RCS
pressure. One HHSI pump still operating.
0627
Letdown orifice valves opened in process of
re-establishing normal makeup and letdown flows,
and termination of SI.
0639
Estimated time of last PORV closure.
0648
Volume control tank (VCT) relief valve lifted,
gases / water were vented to high level waste drain
tank.
0652
Ventilation system radiation monitor reading
increased. Auxiliary building evacuated, air
sampling was initiated.
0717
VCT pressure decreased below relief valve setpoint.
0900
Ventilation system monitor readings returned to
approximately backgi ound values.
Response of plant control systems appeared to be as designed following the
turbine / reactor trip except that condenser steam dump valve, TCV-1408G,
remained open after RCS temperature had returned to the normal hot standby
value (547*F), causing excessive cooldown of the RCS. The cooldown resulted
in automatic SI initiation due to low pressurizer level and pressure. This
will be the subject of a fourteen day written report per Technical Specifi-
cations (0 pen Item 338/79-39-01 identified for followup purposes). During
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the event, other problems were identified. These are discussed briefly
below and will be reviewed in more detail at a later date.
The RCS hot leg cooldown rate of 110 F in approximately thirty minutes
a.
exceeded the 100 F per hour limit of Technical Specification 3.4.9.1.b
(0 pen Item: 338/79-39-02).
b.
The control room bottled air systems did not actuate. is designed upon
receipt of SI signal (Unresolved Item: 338/79-39-03).
c.
A turbine reheat stop valve failed to close when the turbine tripped,
contrary to design (0 pen Item: 338/79-39-04).
d.
A containment Hi-Hi air particulate radiation alarm was received.
This was apparently due to increased leakage across the seals of
reactor coolant pump "C" (0 pen Item: 338/79-39-05).
Following receipt of the SI, all reactor coolant pumps were tripped as
required by IE Bulletin 79-06C. SI was allowed to function for twenty
minutes as required by IE Bulletin 79-06A. In the process of re-estab-
lishing normal RCS makeup and letdown and terminating SI, the volume
control tank (VCT) overfilled, forcing some of contents through the VCT
relief valve, RV-1257
to the high level waste drain tanks (FSAR Figure
11.2.2-1).
The gas spaces of the drain tanks are piped to the process vent
system by design, but plant personnel found that the one-inch line
LW-81-152 had been opened at restricting orifice R0-LW-104 and an elbow
installed such that the tanks were open to auxiliary building atmosphere.
This abnormal configuration was the subject of an investigation condu^ted
October 2-4 and is documented in the Investigation Report appended to this
report.
Station management was informed on October 4 that the abnormal
configuration appeared to be in noncompliance with 10 CFR 50.59(b)
requirements and paragraph 14.5.2 of the Nuclear Power Station Quality
Assurance Manual (NPSQAM) in that it constituted a change to equipment
discussed in the FSAR and had not been evaluated and documented (Infraction
339/79-39-06).
The orifice was reinstalled upon discovery by station
personnel on September 25 and observed that same date by Region II inspec-
tors.
Review of the system by station personnel and Region II inspectors revealed
another potential release path to the auxiliary building which is discussed
in detail in paragraph 10.
.
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Regard.ag overfilling of the VCT, station management informed the inspector
on Septamber 26 that the apparent cause was blocking of the VCT level
control valve, LCV-1115A, to the VCT position by operators using a paper
clip and pencil to keep the control from returning to the auto position.
It was stated that this action had been practiced in the past because
LCV-1115A allowed leakage through to the boron recovery system when in the
auto position. Management further stated that the clip would be removed
from the switch for LCV-1115A and operators instructed not to block the
switch in the future. The necessity for controlling and reviewing such
" modifications" to component functions was discussed by the inspectors. On
October 1, during a tour of the control room, the inspector observed the
switch for LCV-1115A again blocked to the VCT position. Discussions with
members of the operating staff on duty revealed that they did not fully
understand the requirements of 10 CFR 50.59 regarding evaluation required
before changing a system or procedure described in the safety analysis
report. Station management was informed that blocking of LCV-1115A to one
position negated the control function as described on pages 9. 3. 4-7, 9. 3. 4-8,
and 9.3.4-24 of the FSAR and that failure to perform a safety evaluation on
this blocking action appeared to be another example of noncompliance with
10 CFR 50.59(b) and paragraph 14.5.2 of the NPSQAM (Infraction 338/79-39-06).
The inspector stated that a similar citation had been addressed in IE
Report 50-338/79-08.
On October 2 station management informed the inspectors that it had been
determined that LCV-1115A could not have been blocked to V:; position at
the time of the event because a continuous dilution of the RCS requiring
that letdown flow be diverted to the boron recovery system, had been in
progress during the entire shif t except for a brief time. This assertion
was addressed as part of an investigation and is covered in the Investiga-
tion Report appended to this report, which conciaded that the VCT overfilled
due to actuation of the letdown relief valve (RV-1209), which is piped to
the VCT. This relief actuated upon re-establishing letdown when the boron
recovery system gas stripper was isolated by trip valve, (TV-BR-Illa)
leaving no flow path for letdown except via the relief to the VCT.
6.
Radiological Sequence of Events
NRC radiological specialists reviewed plant event recorder printouts,
radiological sample results, radiation monitor printouts, and interviewed
plant staff to determine the sequence of events during the incident.
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Date
Time
Event
Results
_2
9/23
0730
Reactor Coolant Sample
I-131 @ 3.86 x 10
pCi/cc
9/25
0000-0400
Routine Auxiliary Building
No significant activity
Air samples
revealed.
0609
Reac; ,r Trip
0652
Auxiliary Building Ventila-
Monitors response increased
tion Montiors Alarm
by factor of 1000 over back-
ground and returned to normal
by 0900.
0700
Evacuation of Auxiliary
Building
0705-0710
Auxiliary Building Air
259' elevation:
155 times MPC*
Samples
NOTE: MPC for noble gases
274' elevation:
101 times MPC*
are based on submersion
doses
_5
0720-0803
Ventilation Stack Samples VVA: Xe-133 E 2.0 x 10_s pCi/cc
Xe-135 * 9.6 x 10,3 pCi/cc
VVB: Xe-133
1.0 x 10
pCi/cc
_6
Proc Vent: Xe-133 9.5 x 10_s pCi/cc
Xe-135 3.6 x 10_s pCi/cc
Kr-85m 4.5 x 10_7 pCi/cc
Rb-88
1.1 x 10
pCi/cc
0730
Posted Auxiliary Building as
" Airborne Radioactive Area"
0830-1433
Auxiliary Building Air
Activity reduced by factors
Samples
of 100 and 30 compared to
0750 samples for 274' and 259'
elevations respectively.
_2
0940
Reactor Coolant Sample
I-131 @ 8 x 10
pCi/cc
1400
Unit Containment Entry and
Exposures a 10 mrem
Air Sample
Air activity a 240 times MPC
(This is typical of an
operating facility)
_1
1600
Reactor Coolant Sample
I-131 @ 1.4 x 10
pCi/cc
Approx. 1700
South Fence TLDs removed
No exposures above background.
- MPC is maximum permissible concentration.
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7.
Estimate of Activity Released
The inspectors reviewed results of the gaseous and particulate grab samples
taken from the auxiliary building vents (A&B) and the process vent. Below
is a tabulation of the results obtaine'., including flow rates.
Released Concentrations (pCi/ml)
Aux. Bldg. Vent A
Aux Bldg. Vent B
Process Vent
Isotope
(50,000 CFM)
(27,000 CFM)
(270 CFM)
2.02 x 10 5
1.01 x 10 5
9.47 x 10 5
Xe-135
9.60 x 10 6
-
3.59 x 10 5
Rb-88
-
-
1.10 x 10 7
Kr-85m
-
-
4.54 x 10 6
From a review of the strip chart recordings on RM-VG-112, the inspectors
ascertained that the duration of the release was no longer than a three
hour period commencing at approximately 7:00 a.m. on September 25, 1979.
Within one hour of the incident, licensee's representatives estimated that
no more than 7.5 curies of Xenon-133 equivalent could have been released
over the three-hour period. The inspectors reviewed the licensee's estimate
and determined that the estimate represented an upperbound because the
release was assumed to occur at a relatively high concentration for the
entire three hour period. The inspectors also determined that: (1) no radio-
active iodine w:. detected in the grab samples from the vents, and (2) no
technical specification release limits were exceeded.
The release rates
were less chan ten percent of the allowable limits as defined by the tech-
nical specifications.
An offsite dose assessment of the released activity was made by licensee's
representatives immediately after the release. Assuming a release of 7.5
curies of Xenon-133 for three hours and actual site meteorology, (average
wind speed of approximately 7.5 miles per hour from the north with two
hours in Pasquill's Stability Class D and one hour in Class E), licensee's
representatives estimated that the maximum site boundary dose to the south
was 3.5 x 10 3 millirad gamma and 1.1 x 10 3 millirad beta and the maximum
nearest residence dose to the south was 1.7 x 10 4 millirem whole body and
,
5.7 x 10 4 millirem skin.
Supporting data was obtained by licensee's
representatives who read the TLDs which were hung on the south protected
area fence. No noticeable increase above background was discernable for
the TLDs. The inspectors had no further questions or observations in this
area.
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8.
In Plant Exposures
Personnel exposures during the event were determined by routine use
a.
of TLDs aud self-reading pocket dosimeters. Pocket dosimeter readings
for the workers initially in the auxiliary building and those who
re-entered for air sampling, etc. , ranged from 5 to 10 millirem. The
workers TLDs were also read, with indicated doses ranging from 3 to
157 mrem. The doses, however, represent exposures for the previous
month, quarter, or for the individual with 3 mrem, the previous two
weeks. An inspector calculated skin (beta) exposures for these workers
using TLD system vendor equations indicating doses from 4 to 13 mrem.
b.
After evacuating the two workers from the auxiliary building at approxi-
mately 7:00 a.m. , several entries were made to draw air samples (partic-
ulate, gas, charcoal) from the auxiliary building general areas, and
reactor containment.
An inspector reviewed the results of these air
samp'es, summarized below and had no additional questions.
Approximate
Time
Location
MPC Fraction
Isotopes
0150
274' Aux. Bldg.
0.007
Rb-88
0405
259' Aux. Bldg.
0.008
Rb-88
0705
274' Aux. Bldg.
101
Kr-88, Xe-133,
& Xe-135
0715
259' Aux. Bldg.
155
Kr-87, Kr-88,
Xe-133, Xe-135
0830
274' Auz. Bldg.
1.1
Xe-133, Xe-135,
& Rb-88
0840
259' Aux. Bldg.
6.0
Xe-133, Xe-135,
& Rb-88
0955
259' Aux. Bldg.
0.7
Xe-133, Xe-135,
& Rb-88
1200
259' Aux. Bldg.
0.003
Rb-88
1400
291' #1 Contain-
ment Building
240
Xe-133, Xe-135
The Rb-88 activity measured before the transient was the result of an
identified small leak from the Gas Stripper. It should also be noted
that the MPC for all the isotopes above are based on submersion doses
and any personnel exposures resulting from the air activity would be
accounted for by TLD dosimetry.
9.
Effluent Monitoring
a.
The release points for airborne activity from the facility, the vent
monitor system, and monitor responses were reviewed.
The release
points and type of monitoring provided are summarized below.
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(1) Ventilation Vent A (VVA) (System flow is about 50,000 cfm). This
system ventilates the auxiliary building general areas, the
control area (Volume Control System) of the auxiliary building,
laboratory vents, sample sink, laundry and health physics areas.
The monitor system includes an isokinetic probe in the vent, a
gas monitor, particulate monitor, and a charcoal filter for
analysis of iodine releases. The monitor response is recorded,
and alarms in the control room.
(2) Ventilation Vent B (VVB) (System flow is about 27,000 cfm). This
system ventilates the Safeguards Building, Fuel Building, and
Decon Building. The Containment Purges are also routed through this
vent.
The monitor system includes an isokinetic probe,
a gas
monitor, particulate monitor and charcoal filter. The monitor
reads out, alarms and is recorded in the control room.
(3) Process Vent (PV) (System flow is about 270 cfm). This system
provides a release path for various tank vents and other sources
that might have airborne activity. The air is processed through
HEPA and charcoal filters prior to discharge. The monitor includes
an isokinetic probe and particulate and gas monitors with readout
recorder and alarms in the control room.
A separate sample
system incorporating a charcoal cartridge is also used.
b.
The inspectors reviewed strip chart recordings of the responses of the
three monitors discussed above. The strip chart (RR-150) recording of
VVA and PV had been removed for examination and were not reviewed by
the radiation specialists. They were reviewed at a later date by the
Resident Inspector.
A review of the computer printout indicating
various alarms revealed VVA (gas) alarmed at 6:52 a.m.
and PV (gas)
alarmed at 6:53 a.m.
The strip charts for VVB and VV' indicated monitor
response increased from 50 cpm to 50,000 cpm (particulate) and from 40
cpm to 10,000 cpm (gaseous) at approximately 6:55 a.m.
Using cali-
bration data of June 8,
1979, an inspector correlated the monitor
response for VVB with isotcpic data, discussed in paragraph 7, and had
no questions.
10.
Verification of Release Pathway
Accompanied by the licensee's representatives, inspectors traced the volume
control tank relief valve discharge pathway. The relief. valve discharges
into line 4"-DH-245-153A-Q3 which flows to the high level ~ waste drain tank
(1-LW-TK-2B) thru line 4"-LW-217-152.
The two high level waste drain tanks
have a six inch crass connect line (6"-LW-63-152) from which a one inch tap
(approximately 10 feet long and designated 1"-LW-81-152) goes through an
orifice (RO-LW-104). At the time of the transient the orifice was not in
place and the open vent line vented to the high level waste drain tank
cubicle on the 259'-6" elevation. The inspectors observed the overflow
cross connection between the high level waste drain tanks and the low level
waste drain tanks (1-LW-TK-3A and 3B).
The four inch cross-connect (line
numbers 4"-LW-91-52, 4"-LW-92-152, 4"-LW-249-159, 4"-LW-94-152 and
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4"-LW-95-152) contains approximately 50 feet of piping, with a locked open
diaphram valve (1-LW-787), between the high and low level waste tanks. The
low level waste drain tanks each have a four inch overflow which branches
off the cross-connect line. During normal operations the low level waste
drain tanks are vented through the process ventilation system. However,
during the discharge of the volume control tank to the high level waste
drain tanks, the high level waste drain tanks as well as the low level
waste drain tanks became pressurized.
The pressurization of the tanks
allowed radioactive gases to escape via the one inch vent line on the high
level tanks (line 1"-LW-81-15) and through the overflow lines (line Nos.
4"-LW-94-152 and 4"-LW-95-152) on the low level waste tanks. Smears taken
by the inspectors on the inside of the overflow lines on the low lev 21
waste tanks indicated the presence of activity.
Readings of several
hundred counts above background were obtained.
Based on pressure drop
considerations, the majority (>90%) of flow of contaminated gases should
have come from the low level waste drain tank overflow lines.
The
inspectors discussed the potential leak path of the low level drain tank
with licensee's representatives.
Licensee's representatives acknowledged
the leak path potential and stated that they were actively pursuing an
engineering evaluation of the problem. The inspectors will followup on
licensee evaluation in future inspections (338/79-39-07).
With the
exception of the orifice, which was identified above, the inspectors
verified the system was installed in accordance with the FSAR.
11.
Fuel Integrity
The inspectors reviewed the reactor coolant sampling program. Below is a
table listing the reactor coolant concentrations of significant isotopes
prior to and after the transient. From the data presented, it is obvious
that no significant fuel degredation resulted from the transient.
The
small increases in the iodine levels are well within the projected increases
for the transient. Based on the review of the data, the inspectors deter-
mined that no technical specifications for reactor coolant activity were
exceeded. The inspectors had no further questions or observations in this
area.
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REACTOR COOLANT ACTIVITY DETERMINATION (pCi/ml)
Isotope
Date: Time
Date: Time
Date: Time
9/23/79:0730
9/25/79:0940
9/25/79:1600
3.86 x 10 2
8.13 x 10 2
1.42 x 10 1
I-132
2.03 x 10 2
5.97 x 10 2
4.29 x 10 2
I-133
4.77 x 10 2
1.24 x 10 1
1.59 x 10 1
1-134
1.77 x 10 3
1.07 x 10 2
N/D
I-135
1.27 x 10 2
5.46 x 10 2
4.51 x 10 2
3.68 x 10 1
2.19 x 10 1
2.16 x 10 1
Xe-135
1.81 x 10 1
1.50 x 10 1
1.23 x 10 1
N/D
N/D
1.63 x 10 2
Cs-138
1.30 x 10-3
N/D
N/D
N/D = Not determined
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