ML20133B987
| ML20133B987 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 07/18/1985 |
| From: | Constable G, Flippo T, Andrea Johnson, William Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20133B972 | List: |
| References | |
| 50-382-85-16, NUDOCS 8508060312 | |
| Download: ML20133B987 (8) | |
See also: IR 05000382/1985016
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APPENDIX B
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REGION IV
U. S. NUCLEAR REGULATORY COMbiISSION
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'NRC Inspection Report: 50-382/85-16
License: NPF-38
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Docket: 50-382
Licensee': Louisiana Power & Light Company (LP&L)
,'142 Delaronde Street
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New Orleans, Louisiana- 70174
Facility-Name: Waterford Steam Electric Station, Unit 3
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- Inspection At
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Inspection Conducted: April'l through May 31, 1985
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Inspectors:
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.G. L. Constable
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Senior Resident Inspector
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T. A. Flippo', Resident Inspector
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W. B. Jones',' Reactor Inspector
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A. R. Johnson, Reactor Inspector
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Approved:
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G. L. ConstableT Chief
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Reactor Project Section B
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Inspection Summary
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. Inspection. Conducted April 1 through May 31,-1985 (Report 50-382/85-16)
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. Areas Inspected: Routine, announced inspection'of:
(1) Control of
Design-Changes; (2) Radioactive Contamination of Secondary System; (3).Startup
Test Procedure Review; (4) Phase III Test Witnessing; ,(5) Test
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Results' Evaluation;.and (6) Survey of Licensee Responses to Selected Safety
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. Issues. ~The inspection-involved 638 inspector-hours onsite by 4 NRC
inspectors.
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Res'ults: Within the areas inspected, 3 violations were identified.
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DETAILS
1.
Persons Contacted
Principal Licensee Employees
- R. S. Leddick, Senior Vice President, Nuclear Operations
- R. P. Barkhurst, Plant Manager, Nuclear
- T. F. Gerrets, Corporate QA Manager
- L. F. Storz, Assistant Plant Manager, Operations and Maintenance
S. A. Alleman, Assistant Plant Manager, Plant Technical Staff
- 0. D. Hayes, Operations Superintendent
L. M. Meyers, Assistant Operations Superintendent
- J. R. McGaha, Maintenance Superintendent
- R. G. Pittman, Operations QA Audit Supervisor
- J. N. Woods, Plant Quality Manager
- K. L. Brewster, Onsite Licensing Engineer
G. E. Wuller, Onsite Licensing Coordinator
- T. C. Payne, I&C Supervisor
- G. F. Koehler, Operations QA Engineer
- Present at exit interviews.
In addition to the above personnel, the NRC inspectors held discussions
with various operations, engineering, technical support, and
administrative members of the licensee's staff.
2.
Plant Status
The Waterford 3 site is presently in the startup testing phase.
The 80%
testing plateau has been completed and the plant is in an approximate 3
week outage to perform preventive maintenance associated with a chemical
buildup in the main electrical generator.
3.
Control of Design Changes
On March 3, 1985, the NRC inspector observed that the licensee had
implemented the modifications described in Station Modification Package
(SMP) 760.
This SMP changed the automatic isolation of component
cooling water (CCW) to the reactor coolant pump (RCP) integral seal
cooler from high CCW discharge seal cooler pressure to high CCW
discharge seal cooler temperature.
In addition, SMP 760 added control
switches on CP-2 which allow the operator to reopen the CCW isolation
valves.
While reviewing LP&L's off-normal operating procedures on April 10,
1985, the NRC inspector noted that OP-901-022, Revision 1, "High
Activity in Component Cooling Water System," had not been revised to
reflect the above modifications.
The procedure indicated that CCW to
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the RCP integral seal cooler isolated on a 2/3 high pressure indication;
however, the station modification isolated the CCW line on a 1/1 high
temperature indication.
In addition, the procedure did not address the
installed switches on CP-2 for reopening an isolated CCW line and the
100 second interlock that will reclose the valve if the high temperature
indication is still present.
The NRC inspector reviewed Plant Engineering Procedure PE-2-006,
Revision 5, " Station Modifications," Attachment 6.14. " Document Update
Record," to determine if OP-901-022 had been identified as needing
revision during the SMP 760 review process.
The review revealed that
OP-901-022 had not been identified as an affected document; however,
Off-Normal Operating Procedure OP-901-010, Revision 1, " Reactor Coolant
Pump Malfunction," and Operating Procedure OP-01-002, Revision 4,
" Reactor Coolant Pump Operation," were identified as affected documents
on Attachment 6.14 and were subsequently revised to reflect the changes
made in SMP 760.
This is a violation. (382/8516-01)
4.
Radioactive Contamination of Secondary System
On April 3, 1985, the licensee d tected small amounts of radioactive
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contamination, approximately 10
uCi/ml, at the discharge of the
condenser vacuum pumps.
A walkdown of the secondary system by the
licensee revealed that an auxiliary operator had failed to close Boron
Management Valve BM-435 A(B) as required by LP&L Operating Procedure
OP-07-001, Revision 4, " Boron Management System," after the last
operation of the boric acid concentrator (BAC).
Boron Management Valve
BM-435 A(B) normally isolates the boric acid concentrator steam chest
from the radioactive system relief header.
The relief header is fed by
various chemical volume control systems (CVCS) and BM relief valve,
including CVCS letdown relief valve CVC-115, and discharges to one of
the four holdup tanks.
During operation of the BMC, steam is fed to the BAC steam chest through
the auxiliary steam header.
BM valve BM-435 A(B) is then opened to
allow venting of noncondensables to the relief header. When the
auxiliary operator failed to close BM-435 A(B) after closing down the
BAC, a direct flow path from the radioactive relief header to the
auxiliary steam header and gland seal leak-off tank was provided.
Earlier, on April 3, 1985, the CVCS letdown line experienced a flow
transient which caused a pressure spike downstream of the letdown
orifice valves.
This resulted in CVCS letdown relief valve CVC-115
opening to reduce the pressure spike; however, the valve failed to
properly reseat after the pressure was reduced producing approximately
1 gpm RCS leakage to the relief header.
The relief header was properly
aligned to a holdup tank, but backpressure in the tank bypassed the RCS
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water to the BMC steam chest.
From the steam chest, the RCS water
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contaminating both.
In addition, the gland seal leak off tank.
discharged.to the condenser.
From the condenser the contaminated
condensate made its way into the polishers and steam generators via the
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' normal secondary system flow path.
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' Samples.of the condenser exhaust tank by the licensee revealed the
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following:
Xe - 133
6.8 x 10[ uCi/ml
Xe - 135
3.9 x 10
uCi/ml
Kr - 85u
1.1 x 10
uCi/ml
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In addition the licensee sampled the turbine building industrial waste
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sump.
The samples taken revealed the follow ng:
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Xe - 135
1.2 x 10[ uCf/mi
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Co - 58
7.'6 x 10,
uCi/ml
'Na - 24
7.5 x.10,
uCi/ml
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I - 133
1.4 x 10
uCi/ml
Zr - 97
8.5 x 10
uCi/ml
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Upon verifying the presence of contamination outside the radiation
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control areas, the licensee isolated the affected areas and established
controls in accordance with their approved procedures to prevent the
spread of contamination.
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.This is a violation..(382/8516-02)
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Startup Test Procedure Review
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The NRC inspectors reviewed the startup test procedures for power
ascension testing of the plant.
The procedures were reviewed for
technical content, compliance with Final Safety Analysis Report (FSAR),
and compliance with licensee's administrative procedures. ~The startup
test procedures reviewed are listed below.
SIT-TP-707
Atmospheric Steam Dump and Turbine Bypass Valve Capacity
Checks
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SIT-TP-726
Remote Reactor Trip with Subsequent Remote Plant Cooldown
No violations or deviations were identified.
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6.
Phase III Test Procedure Witnessing
The NRC inspectors observed the performance of portions of the following
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Phase III test procedures:
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SIT-TP-716
Core Performance Record
SIT-TP-718
Variable Tavg Test
SIT-TP-707
Atmospheric Steam Dump and Turbine Bypass Valve Capacity
SIT-TP-724
Temperature Decalibration Verification
SIT-TP-728
Loss of Offsite Power Trip Test
SIT-TP-726
Remote Reactor Trip with Subsequent Remote Plant Cooldown
SIT-TP-755
Natural Circulation Demonstration Testing
During the performance of the test, the NRC inspectors verified the
following:
a.
The personnel conducting the test were cognizant of the test
acceptance criteria, precautions, and prerequisites prior to
beginning the test.
b.
The test was conducted in accordance with an approved procedure and
the test procedure was used and signed off by personnel conducting
the test.
c.
Data was collected and recorded as required by the test procedure
instructions.
No violations or deviations were identified.
7.
Test Results Evaluation
The NRC inspectors reviewed Phase III test results to verify that: (1) all
changes, including deletions to the test program, had been reviewed for
conformance to the requirements established in the FSAR and Regulatory
Guide 1.68; (2) deficiencies had been adequately addressed and corrective
action completed; (3) the licensee correctly analyzed the test data and
verified that it met the established acceptance criteria; and (4) the
startup organization as well as the plant operating review committee
(PORC) had reviewed and accepted the test results.
The following test
packages were reviewed:
SIT-TP-650
Low Power Physics Testing
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SIT-1P-716
Core Performance Record
SIT-TP-717
CPC/COLSS Verification
The NRC inspectors detennined that each of the above test packages was
properly reviewed by the licensee and met the applicable acceptance
criteria.
No violations or deviations were identified.
8.
Survey of Licensee Responses to Selected Safety Issues
During this inspection period, the NRC inspectors reviewed the actions
taken by the licensee of safety issues identified in I. E. Information
Notices 83-75 and 84-06. The selected review included possible steam
binding of emergency feed water pumps and power operations with
mispositioned control rods.
The NRC inspectors reviewed each of the above safety issues to determine
whether the licensee had implemented procedures and provided training
programs to prevent, detect, and recover from a steam bound EFW pump and
mispositioned ccntrol rods.
In addition, the NRC inspectors compared the
Waterford 3 EFW system design with the designs at H. B. Robinson 2, Farley
1, and Surry 2 which have all experienced steam binding of the auxiliary
feedwater pumps.
The review revealed that the EFW system design at Waterford 3 incorporates
additional isolation valves between the EFW pump and main feedwater line
that are not used in the other auxiliary feedwater designs. Specifically,
the discharge of the EFW pump is isolated from the main feedwater line by
two check valves and two isolation valves, all in series.
The isolation
valves do not receive a signal to open until after the EFW pumps are
running. The licensee performs an operational test of the EFW pumps every
31 days in accordance with Surveillance Procedure OP-903-019, Revision 2
" Emergency Feedwater Flow Verification," to verify the operability of the
In addition, the licensee performs Surveillance Procedure
OP-903-045, Revision 1. " Emergency Feedwater Flow Path Verification" every
31 days to ensure the isolation valves are properly aligned and free from
external leakage.
The NRC inspectors determined that the licensee's procedures for recovery
of a mispositioned control rod and verification of rod position, when one
form of indication is lost, have been adequately addressed in Off-Normal
Operating Procedure OP-901-009 Revision 3, "CEA or CEDMS Malfunction."
The licensee has also provided the operators with training in the proper
movement of control rods, and the consequences of improper movement and
operating with a mispositioned control rod.
No violations or deviations were identified.
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9.
Site Tour
At various times during the course of this inspection period the NRC
inspectors conducted general tours of the reactor auxiliary building,
turbine building, and reactor building to observe ongoing maintenance and
testing.
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On May 15, 1985, while on tour in the reactor auxiliary building, the NRC
inspector found Fire Door 0-221 to the Boric Acid Concentrator Room B had
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been removed from service. After further investigation, it was found that
LP&L Fire Protection Procedure FP-1-015, Revision 1, " Fire Protection
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System Impairments," paragraph 4.2, required in part that the shift
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supervisor / control room supervisor (SS/CRS) evaluate the fire impairment
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impact of Technical Specifications and insure that applicable ACTION
statement are complied with.
In addition, the SS/CRS is required to
complete a fire appliance impairment form and note the same in index log.
Contrary to the above statement, the NRC inspector could find no evidence
of a completed fire appliance impairnent form or that operability of the
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fire detectors on at least one side of the inoperable assembly had been
verified. Thisisaviolation(382/8516-03).
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10. Cpen Items
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Three violations have been identified in this report.
8516-01
Control of design changes
Paragraph 3
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8516-02
Failure to close boron
Paragraph 4
managenent valve during BAC
operation
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8516-03
Failure to follow procedures
Paragraph 9
for evaluation of removal of
fire doors
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11. Exit Interviews
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The NRC inspectors met with the licensee representatives at various times
during the course of the inspection.
The scope and findings of the
inspection were discussed.
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