ML19320D774
| ML19320D774 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/04/1980 |
| From: | Callahan L, Johnson W, Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML19320D759 | List: |
| References | |
| 50-313-80-07, 50-313-80-7, 50-368-80-07, 50-368-80-7, NUDOCS 8007220197 | |
| Download: ML19320D774 (18) | |
See also: IR 05000313/1980007
Text
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U. S. NUCLEAR REGULATORY COMMISSION
OF W E OF INSPECTION AND ENFORCEMENT
REGION IV
Report No. 50-313/80-07
License No. DPR-51
50-368/80-07
Licensee: Arkansas Power and Light Company
P. O. Box 551
Little Rock, Arkansas
72203
Facility Name: Arkansas Nuclear One (ANO), Units 1 and 2
Inspection at: ANO Site, Russellville, Arkansas
Inspection Conducted: April 22 - May 21, 1980
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Inspectors:
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W. D. Johnson, SenTor Resident Inspector
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Sk/fl0
/d L. J. Callan, Resident Inspector
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[h S. Dearf, Reactor Inspector
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Approved:
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T. F. Westerman, Chief, Reactor Projects Section
Date
Inspection Summary
Inspection conducted during period of April 22 - May 21, 1980
(Report No. 50-313/80-07)
Areas Inspected: Routine, announced inspection including Confirmatory
Order Follow-up, Follow-up on IE Bulletin 79-27, Follow-up on Open Item,
Operational Safety Verification, and Reactor Coolant Pump Seal Failure.
The
inspection involved 142 inspector-hours on-site by three NRC inspectors.
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8007220199
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Results: Within the five areas inspected, no items of noncompliance were
identified.
Inspection conducted during period of April 22 - May 21, 1980
(Report No. 50-368/80-07)
Areas Inspected: Routine, announced inspection including Follow-up on
IE Bulletin 79-27, Licensee Review of Power Ascession Test Results, Follow-up
on Licensee Event Reports, Inspector Review of Power Ascension Test Data,
Follow-up on Open Item, Isothermal Temperature and Power Coefficients of
Reactivity, Core Power Distribution Operational Safety Verification, and
Surveillance Observation. The inspection involved 126 inspector-hours
on-site by three NRC inspectors.
Results: Within the nine areas inspected, one item of noncompliance was
identified (infraction - adherence to procedure, paragraph 12).
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DETAILS SECTION
1.
Persons Contacted
J. P. O'Hanlon, ANO General Manager
G. H. Miller, Engineering & Technical Support Manager
B. A. Baker, Operations Superintendent
T. N. Cogburn, Plant Analysis Superintendent
E. C. Ewing, Plant Engineering Superintendent
P. Jones, Maintenance Superintendent
F. Foster, Operations and Maintenance Manager
J. McWilliams, Assistant Operations Superintendent
J. Albers, Planning and Scheduling Supervisor
D. D. Snellings, Technical Analysis Superintendent
L. Schempp, Manager of Nuclear Quality Control
M. Bishop, Acting Plant Administrative Manager
R. Tucker, Assistant Maintenance Superintendent
L. Bell, Assistant Operations Superintendent
G. Fiser, Radiochemistry Supervisor
V. Pettus, Assistant Maintenance Superintendent
I. Mosquito, Health Physics Planning & Scheduling Coordinator
D. Wagner, Assistant HP Sepervisor
G. Halverson, Assistant HP Supervisor
J. Lamb, Safety and Fire Protection Coordinator
H. Hollis, Security Coordinator
D. Lomax, Nuclear Engineer
C. Shively, Plant Performance Supervisor
T. Green, Training Coordinator
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D. Glenn, Health Physics Supervisor
G. Fiser, Radiochemistry Supervisor
P. Rogers, Nuclear Support Supervisor
R. Turner, Electrical Engineering Supervisor
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M. Stroud, Engineer
D. Trimble, Manager of Licensing
L. Humphrey, Manager of Quality Assurance
G. Charles, Secretary, Safety Review Committee
D. Sikes, Director, Generation Operations
The inspectors also contacted other plant personnel, including operators,
technicians and administrative personnel.
2.
Follow-up on Previously Identified Item (Units 1 and 2)
(Closed) Open item 313/79-21-02; 368/79-20-01:
Overall review of the
QA Program effectiveness,
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The licensee has issued QA Administrative Procedure QAA-16 entitled
"QA Program Report." This procedure initiates a system for evaluating
the effectiveness of the QA Program.
3.
Inspector Review of Test Data (Unit 2)
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During this inspection, the inspector reviewed certain tests performed
at the 100% power level plateau.
Items considered in this review
included resolution of test deficiencies, licensee evaluation of
test results, acceptability of test data and administrative control
of testing.
Tests reviewed during this inspection included the following:
Test
Title
2.800.01 Appendix L
Process Variable
Intercomparison
2.800.01 Appendix M
Chemistry and Radiochemistry
Test
2.800.01 Appendix P
Core Performance Record
2.800.01 Appendix RR
2.800.09
Vibration and Loose Parts
Monitoring
2.800.01 Appendix U
Unit Load Transient Test
Appendix L was performed in January 1980, to compare process instru-
mentation readings obtained from the plaat computer, plant protective
system, the core protection calculators and console meters to verify proper
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agreement between systems. The inspector reviewed the results of this
test and identified no discrepancies which had not been identified and
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resolved by the licensee.
Appendix M was performed in February 1980, at 100% power.
This test
involves the recording of RCS chemistry and radiochemistry data and
comparing results with the process radiation monitor for correlation. The
results of this test did not meet the acceptance criteria.
The
resolution of this is an open item.
(368/80-07-02)
Appendix P was performed at 100% power in January 1980.
The test included
measurement of core radial power distribution and axial power distri-
bution.
Test' data was obtained in an INCA verification file which was
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transmitted to Combustion Engineering for use in the CECOR code to
obtain the power distributions. These measured distributions were then
compared to the predicated valves. The inspector reviewed the
test results and identified no deficiencies.
Appendix RR was performed on January 29, 1980, to demonstrate the
total system performance by the primary and secondary systems in
response to a full load turbine trip.
The inspector witnessed this
test and reported test results in Inspection Report 368/80-03.
The
inspector reviewed the results of this test and identified no
deficiencies which had not been identified and resolved by the
licensee.
Test Procedure 2.800.09 was performed in January 1980, to provide
baseline data ' r core vibration and loose parts monitoring.
The
inspector reviewed the results of this test and identified no
deficiencies.
Appendix U was performed in January-March,1980, to demonstrate
satisfactory operation of plant control systems in the automatic
mode to maintain plant parameters within acceptable limits
during steady state power operations and during transient
conditions.
The inspector revieved the results of this test and
identified no deficiencies.
4.
Licensee Review of Power Ascension Test Results (Unit 2)
The inspector attended a meeting of the Plant Safety Committee on
May 20, 1980. At this meeting, the status of the testing program was
reviewed.
Several tests have not yet been performed; review and analysis
of the data for several tests has not been ompleted; and several tests
need to be repeated.
The inspector will follow the licensee's progress
in completing these items.
5.
Isothermal Temp Coefficient of Reactivity Measurement and Power
Coefficient of Reactivity (Unit 2)
The purpose of this inspection effort was to verify that the measurements
of the Isothermal Temperature Coefficient of Reactivity Measurement
and Power Coefficient of Reactivity were technically correct and
performed in accordance with NRC and licensee requirements.
The
inspector reviewed the following test procedures and randomly sampled
data analysis for accuracy.
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Test Procedure
2.800.01 Appendix R
Variable T average
Revision 2
20% Poser Level 2/21/79 - 2/25/79
2.800.01 Appendix R
Variable T average
Revision 2
50% Power Level 7/29/79 - 8/8/79
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No items of noncompliance or deviations were identified.
6.
Surveillance of Core Power Distribution Limits (Unit 2)
The purpose of this inspection effort was to verify that the reactor
is being operated within the licensed power distribution limits. The
inspector performed a partial review of the following procedures.
Test Procedure
Completed
Core Power Distribution, No. 2304.48, Revision 1
6/12/79
Core Power Distribution, No. 2304.48, Revision 1
6/30/79
Core Power Distribution, No. 2304.48, Revision 2
8/10/79
This inspection effort did not identify any items of noncompliance or
deviations in the areas reviewed. A future inspection will again
examine this area so as to complete the stated inspection purpose.
7.
Licensee Event Reports (LER's) (Unit 2)
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The inspector reviewed certain LER's to verify the following items:
Appropriate corrective action had been taken.
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The event did not involve operation of the facility in a
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manner which constituted ar unreviewed safety question as
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defined in 10 C7R 50.59(a)(2).
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The event did not involve continued operations in violation
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of regulatory requirements or license conditions.
Reporting requirements were met.
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The LER's included in this review are listed below:
79-21/03L-0
79-22/01T-0
79-23/03L-0
79-24/03L-0
79-27/03L-0
79-28/03L-0
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79-29/03L-0
79-30/03L-0
79-31/03L-0
79-32/03L-0
79-33/01T-0
79-37/03L-1
79-38/03L-0
79-39/03X-1
79-41/03L-0
79-42/03L-0
79-44/03L-0
79-45/03L-0
79-46/03L-0
79-47/04L-0
79-48/03L-0
79-49/03L-0
79-51/03L-0
79-52/03L-0
79-55/03L-0
79-56/03L-0
79-57/04L-0
79-58/03L-0
79-58/03X-1
79-59/03L-0
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79-60/03L-0
79-61/03L-0
79-62/03L-0
79-63/03L-0
79-64/03L-b
79-67/04L-0
79-68/03L-0
79-69/03L-0
79-70/03L-0
79-71/03L-0
79-72/03L-0
79-73/03L-0
79-74/03L-0
79-75/03L-0
79-75/03X-1
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79-76/03L-0
79-77/03L-0
79-78/03L-0
79-79/03L-0
79-80/03L-0
79-81/03L-0
79-82/03L-0
79-83/03L-0
79-84/03L-0
79-84/03X-1
79-85/03L-0
79-86/03L-0
79-87/03L-0
79-87/03X-1
79-88/03L-0
79-89/03L-0
79-90/03L-0
79-91/03L-0
79-92/03L-0
79-93/99X-0
79-94/03L-0
79-95/03L-0
79-96/03L-0
79-97/03L-0
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79-98/03L-0
79-99/03L-0
79-101/03L-0
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79-102/03L-0
79-102/03X-1
79-103/03L-0
79-104/03L-0
80-18/03L-0
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LER 368/80-18/03L-0 reported a loss of suction to the
Emergency Feedwater (EFW) pumps on April 7, 1980. The loss of
suction occurred about 15 minutes following a reactor trip from 98%
power caused by a loss of offsite power. Tb suction of the EFW
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pumps had been aligned to the Condensate Starage Tank (CST) and to
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the Startup and Blowdown Demineralizer effluent in parallel.
Prompt
operator action was taken to isolate the EFW pump suction from the Startup
and Blowdown Demineralizer and to vent the EFW pumps.
EFW flow was
reestablished within one minute.
Licensee investigation revealed that the
EFW pump suction loss was caused by flashing in the main feedwater train
forcing hot water through the Startup and Blowdown Demineralizers to the
EFW pump suction where it flashed to steam.
The steam caused
cavitation of the EFW pumps, and the pressure in the EFW suction
header prevented flow from the CST. Action to prevent recurrence
included revising the EFW system. operating procedure and the plant
startup prccedure to require shutting the EFW suction valve from
the startup and Blowdown Demineralizers during plant startup at about
5% full power after the Steam Generator feedwater source has been shifted
to a main feedwater pump.
In addition, the EFW suction valve from
the Startup and Blowdown Demineralizers will be verfied closed once
per shift during Mode 1 operation. Members of the NRC staff in Headquarters
are reviewing this event to determine whether it could occur at other
plants and to determine whether the licensee's action to prevent
recurrence is sufficient.
8.
Confirmatory Order Follow-up (Unit 1)
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a.
On April 22, 1980, the NRC Office of Nuclear Reactor Regulation
issued a Confirmatory Order for Arkansas Nuclear One, Unit No. 1.
This order confirmed licensee commitments to take certain
actions as a result of experience gained from the Crystal River,
Unit No. 3 incident of February 26, 1980, in which a non-nuclear
instrumentation power loss resulted in a series of unexpected
events.
These actions were intended to reduce the probability of a
similar future power loss causing unexpected plant responses and
allow the plant operators to better cope with losses of instrumentation
and control functions.
b.
Part II of the above order specified required actions and referenced
licensee commitments. The inspector's review of these items is
described below:
(1) Required Action No. 1
Actions which will allow the operator to cope with various
combinations of loss of instruction and control functions.
This includes changes in (A) equipment and control systems
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to give clear indications of functions which are lost or
unreliable; (B) procedures and traiaing to assure positive and
safe manual response by the operator in the event that competent
instruments are unavailable.
(a) The licensee has performed design change DCP 80-1027 in
response to this item.
This change provides an alternate
120 volt power supply to the Non-Nuclear Instrumentation
(NNI) cabinets.
Automatic transfer switches were installed
which will transfer to the emergency power supply when the
normal source voltage fails and will transfer back to normal
source ter. minutes after normal source restoration.
The
inspector reviewed this DCP, which was completed under Job
Order 1-9079B-80-4 on April 28, 1980. The inspector checked
the modified wiring against drawing E-547, Sheet 3 of 3,
Revision 8-2, entitled " Connection Diagram-Plant Auxiliary
System Control Power, Y-C48."
Two jumpers in cabinet C-48
on terminal board P-TBY were not in accordance with the
drawing.
The licensee determined that the drawing was in
error and processed a field change to correct the drawing.
The licensee performed Work Plan 1407.01, " Checkout of NNI
Power System Modifications (DCP 80-1027)" on April 29, 1980.
The inspector witnessed the portion of this test which tested
the modifications for NNI "Y."
The test was properly conducted and the modified system
performed in accordance with design specifications.
(b) The licensee has issued Revision 6 to procedure 1203.12,
"Annuniciator Corrective Action." This revision adds
a section addressing the various NNI trouble alarms
annunciated on panel K14. The procedure lists the
causes and required action for the alarms, lists which
controls or indications are lost when various failures
occur, and includes variations in lost functions based
on Control Room selector switch positions. The
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Assistant Operations Superintendent has conducted
informal training sessions on this revised procedure
with operators, but at the end of this inspection, the
licensee had not yet conducted any formal training sessions
on this revised procedure.
(Open Item 313/80-07-01)
(2) Required Action No. 2
Determination of the effects of various combinations of loss of
instrumentation and control functions by design review
analysis and verification by test.
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The licensee's letter of April 18, 1980, to the Commission
provided the results of the design review analysis of the
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effects of the failure of various power supplies to the Non-
Nuclear Instrumentation (NNI) system. The inspector witnessed
the performance of Work Plan 1409.14 on May 4, 1980.
This work plan determined the available critical NNI under
conditions of loss of power to the NNI cabinets.
At the
end of this inspection period, the licensee had identified
one condition requiring correct action.
DCP 80-D-1048,
entitled " Modify A Steam Generator Startup Level Buffer Power
to NNI Y," was completed on May 20, 1980.
This design change
provides power for the A Steam Generator startup level buffer
to computer point P-003 from NNI Y in order to make A Steam
Generator startup level available on computer point P-003 if 24
volt DC Power is lost in NNI X.
The inspector will review the
licensee's written evaluation of the above test when it is
available (0 pen Item 313/80-07-02)
(3) Required Action No. 3
Correction of electrical deficiencies which may allow the
power operated relief valve and pressurizer spray valve to open
on non-nuclear instrumentation power failures, such as, the
even, which occurred at Crystal River, Unit 3 on February 26,
1980.
The licensee performed two design changes in response to this
item. DCP 80-1030, entitled " Modify NNI Control Circuitry for
Pressurizer Pilot Operated Relief Valve," added auxiliary
relays to detect the loss of the 1 24 volt power supply
and open contacts in the PORV control circuit to prevent
this valve from getting an "open" signal upon loss of NNI
power suppli s.
This DCP was completed under Job Order
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1-7503-80-4 and tested in accordance with Work Plan 1408.07 on
April 22, 1980.
The inspector reviewed the completed DCP,
Job Order and Work Plan. DCP 80-1031, entitled " Add Contacts
to Control Circuit for Pressurizer Spray Valve (CV-1008)," was
performed to minimize the issuance of an "open" command to the
spray valve on loss of 1 24 volt power and issue a "close"
command on loss of 1 24 volt power. This design makes use
of the same auxiliary relays installed in DCP 80-1030.
This
DCP was completed under Job Order 1-7504-80-4 and tested in
accordance with Work Plan 1408.07 on April 22, 1980.
The
inspector reviewed the completed DCP, Job Order and Work
Plan.
9.
IE Bulletin 79-27 (Units 1 and 2)
Part III of the above order discussed IE Bulletin 79-27 and Part IV
required that the licensee provide a written response to IEB 79-27
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for Unit 1 by April 28, 1980. The licensee's response to IEB 79-27
for Unit I was dated April 23, 1980. The inspector has verified the
licensee's responses to items 1 and 2 of IEB 79-27 as described above.
The licensee's response to this bulletin for Unit 2 was dated
Februa ry 28, 1980.
The inspector rev ewed the licensee's
response and had no further questionr
10.
Reactor Coolant Pump Seal Failure (Unit 1)
On May 10, 1980, the seal on the "C" Reactor Coolant Pump (RCP) failed,
allowing excessive reactor coolant leakage to the Containment sump.
The unit was rapidly shut down and cooled down as required by Technical Specification 3.1.6.1.
The cooldown rate was within the limits
of Technical Specification 3.1.2.
An adequate margin to saturation
was maintained throughout the transient and no Technical Specification
Safety Limits for Operation were violated.
The NRC was notified of this
event as required by 10 CFR 50.72.
No measurable releases of radioactive
material to the environment were made as a result of this event on May 10,
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1980.
The inspector observed, in part, the plant cooldown and depressurization
and periodically monitored the plant stack radiation monitor to verify
that no non-routine releases were being made.
Licensee activities with
respect to this incident were observed to be well organized, with manage-
ment involvement, technical support and with a concern for the health and
safety of employees and the public.
About 59,000 gallons of reactor coolant water was spilled to the
Containment sump before the leak was stopped by depressurization and
partial draining of the reactor coolant system.
The inspector
observed, in part, the tranfer of this water to holding tanks in the
Auxiliary Building and its processing and clean-up for subsequent
use in the Borated Water Storage Tank and in the reactor coolant
system.
The inspector, together with other members of the NRC staff, reviewed
the licensee's analysis of the Containment atmosphere and the plans
for filtering and ventilating the Containment atmosphere. The NRC
staff performed independent calculations to verify that the licensee's
planned ventilation of the Containment Building would not result
in release of radioactive material to the environment in quantities or
rates in excess of those allowed by Technical Specification (Appendix B)
2.4.2.
The inspector observed the preparations for and the commencement
of the Containment Building ventilation on May 13, 1980.
The inspector
observed several segments of the release, conducted at hourly intervals
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until May 15, 1980, and confirmed that the pre-release calculations were
conservative and that the release rates were well within the Technical
Specification limits.
At the end of this inspection period, RCP seal replacement work was in
progress.
The inspector identified no items of noncompliance or
deviations related to this incident.
11.
Monthlv Surveillance
aspection (Unit 2)
The inspector witnes: ed the following core physics surveillance tests:
Procedure No.
Title
2103.15
Reactivity Balance Calculation
2302.01
Incore Detector Channel Check
2302.05
Core Power Distribution
2302.16
RCS Calorimetric Flow Rate
Calibration
Included in this inspection were the following items:
A review of each of the surveillance procedures for conformance
a.
to technical specification requirements and verification of
proper licensee review and approval.
b.
Observation of portions of each surveillance test.
A review of the test data for accuracy and completeness.
c.
d.
Verification that test results met technical specification
requirements.
Verification that the test was done by qualified personnel.
e.
f.
Verification that the surveillance schedule for the test
was met.
No items of noncompliance were identified.
12.
Operational Safety Verification (Units 1 and 2)
.The inspectors performed certain activities to ascertain that the
facility is Leing operated safely and in conformance with regulatory
requirements and that the licensee's management control system is
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effectively discharging its responsibilities for continued safe
operation.
The inspectors activities and findings in this regard are
described in the following paragraphs.
A.
Certain inspection activities were performed frequently (several
times per week).
(1) Control room observations were made which normally included
the following items:
Verification of licensee adherence to selected Limiting
a.
Conditions for Operation (LCO).
b.
Observation of instrumentation and recorder traces for
abnormalities.
Verification of proper control room and shift manning.
c.
d.
Verification of operator adherence to approved operating
procedures.
(2) 3 elected logs and operating records were reviewed to obtain
information on plant operations, detect trends, determine com-
pliance with regulatory requirements and assess the effectiveness
of communications provided by the logs and records.
B.
Certain inspection activities were performed on a weekly basis.
(1) The operability of selected emergency safeguards features
systems was verified by noting valve positions, breaker positions,
instrumentation availability and general condition of major
system components.
Systems selected for review during this
inspection were both trains of Unit 2 Diesel air start system,
Unit 1 Train A of Emergency Feedwater, and both trains of
Unit 2 High Pressure Safety Injection (HPSI) system.
During the inspection of Unit 2 HPSI system on April 29, 1980,
the inspector noted that 2 SI-5091-3, the bypass isolation
valve around the combined HPSI discharge isolation valve
(2CV-5091), was not mentioned 'in the procedure for going into
shutdown cooling mode.
2SI-5091-3 is a normally locked open
valve and must be unlocked and shut if shutdown cooling
is to be effective.
The inspector determined from discussions
with several operators that, although not required by existing
procedures, the valve is, in fact, unlocked and shut when
the HPSI system is lined up for shutdown cooling mode (0 pen
Item 368/80-07-33).
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Additionally, the in -Letor noted that the electrical junction
box cover (2TB903) was missing for HPSI valve 2CV-5076,
the normally shut valve that opens on an ESF signal to allow
HPSI flow to 'D'
reactor coolant loop. Also, the junction
box door on the supply end was open.
This valve had undergone
recent corrective electrical maintenance under job order
2-2012-80 that was documentei complete on April 7, 1980.
The above findings are contrary to licensee procedure 1004.14,
step 4.20, which requires that ". . . any maintenance work
completed requires checkout prior to returning to service .
. .
the cognizant supervisor makes the determination of checkout
requirements .
. the signature of the assigned person of
.
Item 23 (a line item on the job order form) indicates that
the required checkout is complete and the equipment is ready
to be returned to service." This is an apparent item of
noncompliance (368/80-07-01).
(2) The licensee's equipment control was reviewed for proper
implementation of performance of the following inspection
activities:
Review of tag out records to determine that the licensee
a.
has complied with LCO's with respect to removal of
equipment from service.
b.
Independently verifying the proper return to service of
selected safety-related components or systems.
c.
Independent verification of proper conduct of
selected safety-related tagouts currently in effect.
(3) The inspectors conducted tours of accessible areas of the
facility to assess equipment conditions, plant conditions,
radiological controls, security, safety, and adherence to
regulatory requirements.
During these tours, the inspectors
made observations in the following categories:
General plant / equipment conditions including operability
a.
of standby eqaipment.
b.
Maintenance requests had been ir.itiated for equipment
in need of maintenance, and the appropriate priority
had been assigned.
c.
Fire hazards.
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d.
Control of ignition sources and flammable materials.
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e.
Conduct of activities in progress in accordance with the
licensee's addinistrative controls and approved procedures.
f.
Condition of the futerior of selected electrical and
control cabinete.
g.
Physical Security.
The inspector verified that the security plan is being
implemented by observing:
(1) The security organization is properly manned and
that security personnel are capable of performing
their assigned functions.
(2) Protected area barriers are not degraded.
(3) Isolation zones are clear.
(4) Persons and packages are checked prior to entry
into the protected area.
(5) Vehicles are properly authorized, searched, and
escorted or controlled within the protected area.
(6) Persons within the protected area display photo
identification badges.
Persons requiring escort
are properly escorted.
h.
Radiation Protection Controls.
i.
Plant housekeeping - the inspector noted that many
of the locked rooms in the Unit 1 and Unit 2 auxiliary
buildings, such as the Upper and Lower South Piping
Penetration Rooms in the Unit 2 auxiliary building,
exhibits poor cleanliness, especially with respect
to debris remaining from completed maintenance.
J.
Radioactive waste system.
(4) The inspectors reviewed the licensee's trouble tickets
to verify the operability of this problem identification
system.
(5) The inspectors conducted discussions with operators and
other plant personnel and observed several shift turnovers.
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C.
Certain inspection activities were performed once during this
reporting period.
(1) ESF System Operability Verification.
(2) The inspector verified that a selected portion of containment
isolation lineup was correct.
Containment penetrations inspected
were:
2P-5, 6, 7, 8, 9, 10, 11, 14, 15, 17, 18, 19, 20, 21, 23, 24,
25, 31, 32, 33, 34 (all penetrations listed are in the Upper
South Piping Penetration Room of Unit 2 auxiliary building).
The inspector checked that motor operated valves were not
mechanically blocked and power was available.
The piping
between containment and the isolation valves were visibly
inspected for leakage or leakage paths.
(3) The inspector verified that plant conditions, equipment status
and operating parameter, fulfill the following LCO's.
Unit 1
3.1.2.3 Reactor Heatup and Cooldown Limitations
3.1.2.5 Pressurizer Heatup and Cooldown Limitations
3.1.9.2 Rod Control Operation
3.2.1.2 Boron Injection Flow paths
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Unit 2
3.1.2.2 Boron Injection Flow paths
3.1.2.4 Charging Pumps - Operating
3.1.3.1 CEA Position
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3.1.3.2 CEA Position Indicator Channels
(4) The inspector reviewed the licensee's Jumper and Bypass
logs and no conflicts with Technical Specifications were
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identified.
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(5) Radioactive Waste Handling
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The inspector witnessed a radioactive liquid release and
a.
verified the following items:
(1) The release was conducted in accordance with
approved procedures.
(2) The required release approvals were obtained.
(3) The required samples were taken and analyzed.
(4) The effluent release control instrument was
operable and in use during the release.
The inspector also reviewed a sample of the liquid and
gaseous effluent records for March 1980.
b.
The inspector witnessed portions of a compactable trash
processing operation and verified the following items:
(1) Waste products processed complied with shipping
regulations specified in 49 CFR Parts 100-199 as
required by the Department of Transportation.
(2) Tne licensee ensured all waste products compiled
with disposal site acceptance criteria as
established by Chem-Nuclear Systems, Inc.'s,
Barnwell Site Disposal Criteria, effective
December 1, 1979.
The inspector noted that existing licensee approved
written procedures were not sufficiently detailed to
adequately cover solid waste processing and preparation
for shipment.
Licensee represen atives indicated
that new procedures covering solid waste processing are
being written. This item will remain open pending
inspector review of the new procedures (open item
313/80-07-03; 368/80-07-04).
(6) The inspector verified the implementation of the licensee's
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radiation protection controls by:
a.
Observing portions of an area survey performed by health
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physics personnel.
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b.
Examining randomly selected radiation protection
instruments that are in use and ve.ifying operability
and adherence to calibration frequency.
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Verifying by observation and review that the requirements
c.
of one current RWP were being followed.
d.
Verifying compliance with requirements of 10 CFR 20
regarding posting.
Observing that licensee's procedures are Seing followed.
e.
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13.
Exit Interviews
The inspectors met with Mr. J. P. O'Hanlon (Plant General Manager)
and other members of the AP&L staff at the end of various segments
of this inspection. At these meetings, the inspectors summarized the
scope of the inspection and the findings.
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