IR 05000259/1985032
| ML18029A749 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/03/1985 |
| From: | Brooks C, Cantrell F, Patterson C, Paul G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18029A748 | List: |
| References | |
| 50-259-85-32, 50-260-85-32, 50-296-85-32, NUDOCS 8508050024 | |
| Download: ML18029A749 (9) | |
Text
~pS REgy (4
Cy 0O IVl0 I
+
~O
+**~4 UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/85-32, 50-260/85-32, and 50-296/85-32 Licensee:
Tennessee Valley Authority 500A Chestnut Street Tower II Chattanooga, TN 37401 Docket Nos.:
50-259, 50-260, and 50-296 License Nos.:
DPR-33, DPR-52, and DPR-68 Facility Name:
Browns Ferry 1, 2, and
Inspection Conducted:
May 26 - June 20, 1985 Inspectors:
G.
L.
P k, Senior i
nt C.
A.
Pa erson, Resi e
a Signed 7 o7 f5 at Signed C.
R.
Br ks, Resid t Approved by:
F.
S. Cantrell, Sect n
ef Division of Reactor P
jects at Signed 7~sS Date Signed SUMMARY Scope:
This routine inspection involved 100 resident inspector-hours in the areas of operational safety, maintenance observation, reportable occurrences, licensee action on previous enforcement items, and review of licensee identified deficiencies.
Results:
No violations or deviations were identified.
8508050024 860703 PDR ADOCK 05000259
REPORT DETAILS Person Contacted Licensee Employees J.
A.
G.
T.
J.
E.
J.
R.
J.
H.
J.
D.
D.
C.
R.
Hu C.
G.
T.
D.
R.
E.
A.
M.
R.
E.
T.
L.
T.
F.
J.
R.
B.
C.
A.
L.
R.
R.
T.
M.
S.
R.
G, R.
M.
C.
A.
L.
R.
L.
Coffey, Site Director Jones, Plant Manager Swindell, Superintendent
- Operations/Engineering Pittman, Superintendent
- Maintenance Rinne, Modifications Manager Garison, guality Engineering Supervisor Mims, Engineering Group Supervisor nkapillar, Operations Group Supervisor Mages, Mechanical Maintenance Supervisor Cosby, Electrical Maintenance Supervisor Burns, Instrument Maintenance Supervisor Sorrell, Health Physics Supervisor Jackson, Chief, Public Safety Chinn, Senior Shift Manager Ziegler, Site Services Manager Clark, Chemical Unit Supervisor Morris, Plant Compliance'Supervisor Burnette, Assistant Operations Group Supervisor Smallwood, Assistant Operations Group Supervisor Jordan, Assistant Operations Group Supervisor Maehr, Planning/Scheduling Supervisor Hall, Design Services Manager Thomison, Engineering Section Supervisor Clement, Radwaste Group Controller Lewis, Senior Shift Manager Other licensee employees contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, publ.ic safety officers, guality Assurance, design and engineering personnel.
Exit Interview (30703)
The inspection scope and findings were summarized on June 21, 1985, with the Plant Manager and/or Assistant Plant Managers and other members of his staff.
The licensee acknowledged the findings and took no exceptions.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspectio.
Licensee Action on Previous Enforcement Matters (92702)
(Closed)
Open Item (259/83-09-04)
S. I. 4. 7. E.
has been revised to correctly identify the ventilation dampers.
This item is closed.
(Closed) Violation (260/83-33-02)
Two similar violations were given in this area for fai lure to follow procedure and take corrective action - fur ther tracking will be under the more recent violations (259/260/296/84-38-04 and-05).
This violation is closed.
(Closed)
Violation (259/260/296/83-60-04)
Instrument and Control drawing 47W600-133 has been revised to correct the drawing error.
This item is closed.
P'Closed)
Violation (260/83-27-07)
Plant procedure Modification/Addition Instruction (MAI)-13 was reviewed and an additional requirement to provide signoffs for housekeeping has been added.
This item is closed.
(Closed)
Open Item (260/83-23-01)
Valve 2-70-605A for the reactor building cooling water system heat exchanger
"A" sample line was inspected and found adequately supported.
This item is closed.
(Closed)
Open Item (259/83-52-04)
The control room abandonment procedure was reviewed and changes to correctly'dentify the panel'location of the RCIC control switches were made.
This item is closed.
4.
Unresolved Items (92701)
There were no unresolved items during this mport period.
5.. Operational Safety (71707, 71710)
The inspectors were kept informed on a daily basis of the overall plant status and any significant safety matters related to plant operations.
Daily discussions were held each morning with plant management and various members of the plant operating staff.
The inspectors made frequent visits to the control rooms such that each was visited at least daily when an inspector was on site.
Observations included instrument readings, setpoints and recordings; status of operating systems; status and alignments of emergency standby systems; onsite and offsite emergency power sources available for automatic operation; purpose of temporary tags on equipment controls and switches; annunciator alarm status; adherence to procedures; adherence to limiting conditions for operations; nuclear instruments operable; temporary alterations in effect; daily journals and logs; stack monitor recorder traces; and control room manning.
This inspection activity also included numerous informal discussions with operators and their supervisor General plant tours were conducted on at least a weekly basis.
Portions of the turbine building, each reactor building and outside areas were visited.
Observations included valve positions and system alignment; snubber and hanger conditions; containment isolation alignments; instrument readings; housekeeping; proper power supply and breaker; alignments; radiation area controls; tag controls on equipment; work activities in progress; radiation protection controls adequate; vital area controls; personnel search and escort; and vehicle search and 'escort.
Informal discussions were held with selected plant personnel in their functional areas during these tours.
Meekly verifications of system status which included major flow path valve alignment, instrument alignment, and switch position alignments were performed on the nuclear instrumentation systems.
0, A complete walkdown of the accessible portions of the primary containment instrumentation system was conducted to verify system operability.
Typical of the items checked during the walkdown were:
lineup procedures match plant drawings and the as-built configuration; hangars and supports operable; housekeeping adequate; electrical panel interior conditions; calibration dates appropriate, system instrumentation on-line; valve position alignment correct; valves locked as appropriate; and system indicators functioning properly.
All three units remained shutdown during this report period.
Maintenance Observation (62703)
Plant maintenance activities of selected safety-related systems and components were observed/reviewed to ascertain that they were conducted in accordance with requirements.
The following items were considered during this review:
the limiting conditions for operations were met; activities were accomplished using approved procedures; functional testing and/or calibrations were performed prior to returning components or system to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; proper tagout clearance procedures were adhered to; Technical Specification adherence; and radio-logical controls were implemented as required.
e.f.
Maintenance requests were reviewed to determine status of outstanding jobs and to assure that, priority was assigned to safety-Yelated equipment main-tenance which might affect plant safety.
The inspectors observed the below listed maintenance activities during this report period:
a.
Broken flange on fuel pool cooling pump lA b.
250 VOC design, deficiency analysis c.
Unit 1, 2, and 3 diesel air start system d.
Unit 3 LPCI MG set repairs Refuel floor blowout panel repairs Reactor building crane qualification
The licensee identified deficiencies related to the 250 VDC power system voltage limit specifications.
The deficiencies involve deviation from the FSAR design criteria and failure to document design calculations for the Staticon inverters used for the ECCS Analog trip system and the HPCI control system.
The inverters procured on contracts 81P6-826756 (ECN P0126),
83P6-831784 (ECN P0507)
and 66-90744 (original GE supplied HPCI equipment)
are rated for a
minimum voltage of 210 VDC.
Therefore, the equipment rating and lack of design calculations (to establish voltage limits resulting from cable voltage drops)
has rendered this a nonconforming condition.
a.
Contract 66-90744...Original GE supplied HPCI turbine speed controls and HPCI inverter (Topaz invertwr cat.
No.
b.
Contract 81P6-826756...Staticon inverter for the ECCS Analog Trip Unit panels (ECN P0126)
c.
Contract 83P6-831784...Staticon inverter for replacement of the existing HPCI Topaz inverter (ECN P0507)
There have been no equipment failures to date.
Design analysis has identi-fied possible deenergization of the HPCI and ECCS ATU inverters during transient conditions.
It was determined that there were no documented calculation to support the Staticon inverters (provided on ECN P0126) capability to adequately withstand all possible perturbation of the 250 VDC power system.
Further investigation identified the following:
a ~
b.
C.
d.
The FSAR Section 8. 6. 2 page 8. 6-1) states the following:
(1)
The 250 VDC battery may experience a final terminal voltage of 210 volts.
(2)
The Engineered Safeguards Systems that are supplied from the 250 VDC power system shall be designed to operate at a minimum voltage of 200 volts.
The inver ters (Staticon)
provided on contract 81P6-826756 ECN P0125)
and contract 83P6-831784 (ECN P0507)
are rated for 210-280 volts.
The original GE supplied inverters (Topaz) which are being replaced by ECN P0507 are rated for 210-280 volts (ref.
instrument data sheet:
234A9300 Sh.
264).
Informal coordination has identified the original GE supplied HPCI turbine speed controls as being rated for 230 t 20 volt D An analysis by the licensee of the 250 VDC power system supports the following conclusions:
b.
The FSAR statement that the battery terminal voltage may experience a final value of 210 volts appears conservative for steady state conditions.
Current calculations reflect a worst case terminal voltage (excluding transients)
of approximately 217-218 volts for approximately 1/2 hour, 228 volts from 1/2 hour to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 235 volts for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to four hours.
During transient conditions the battery voltage may drop to approximately 207 volts, but this is for short transients only.
The Staticon inverters provided on contract 81P6-826756 (ECN P0126)
can satisfactorily operate for the current
"as-designed" unit
configuration and for the current unit 3 restart configuration.
This will require resetting their undervoltage detector circuitry to operate at 200 +
VOC.
These determinations were documented in design calcu-lations B43 850515 908 and B43 850502 935.
Additional calculations are necessary to establish satisfactory operation for the unit 1 and
"as-designed" configuration, although these calculations are fully expected to document the acceptability of the inverters.
C.
The Staticon inverters for the HPCI system provided on contract 83P6-831784 and the original GE provided inverters (Topaz)
on the NSSS contract are capable of satisfactory operation with the current Unit.3-restart and the current unit 2 configuration.
This will require resetting the Topaz inverters low-voltage sensor to operate at 185 +
VOC and the Staticon inverters undervoltage detector circuitry to operate at 200 L 1 VDC.
The Staticon inverter undervoltage detector circuitry has a time delay of approximately one second while the Topaz inverter low voltage circuitry has no inten-tional time delay.
It is anticipated that these inverters will be adequate for Unit 1 and Unit 3 eventually.
The determination that these inverters are capable of satisfactory operation is based upon informal coordination with the vendors concerning their equipment's ability to withstand the voltage trans-ients and not the ability to experience continuous operation below the equipment's voltage rating of 210 volts.
Lowering the undervoltage detectors is providing assurance that transient voltage conditions will not result in operation of the detectors.
d.
The HPCI turbine speed control circuitry is rated for 230 + 20 YDC.
Informal coordination with Woodward Controls has identified that the speed control circuitry will function properly for 200 VDC.
Operation at 200 VDC should only be for infrequent event Voltage drop calculations reflect that a possibility exists for the existing HPCI inverter to trip and automatically re-energize (approxi-mately 3 seconds later) during transients.
The particular time frame associated with this possibility is prior to the HPCI turbine operation and will not prevent the turbine from proper operation.
The transient occurs in the HPCI autostar t initiation time frame.
The actual HPCI turbine operation which requires proper inverter operation will be approximately 5 seconds after the inverter disturbance.
Actions to be taken by the licensee include:
Revise the FSAR to accurately describe the plant design.
Change the undervoltage setpoint for the Staticon and Topaz inverters to prevent possible tripping during system transients.
Document system design basis in a design document.
The item will remain open for additional inspector followup.
(259/85-32-01)
On May 8, 1985, while reassembling the 1A spent fuel pool cooling system pump, the discharge flange was broken across 2 bolt holes.
Overtorquing was suspected.
Evaluation revealed that the pump casing has a raised flange, while the adjoining pipe has a flat flange.
The maintenance personnel torque bolts based on the bolt type, since it is common industry practice not to supply specific torquing values for each component.
The licensee had assumed that the pump casing would be strong enough that the bolts would be the limiting element in the connection's strength, since it was designed to ANSI B 31-1.
The licensee's evaluation indicates the flange doesn't meet the ANSI standard.
However, their analysis indicates that a seismic event would not have produced a guillotine break, but might produce a high leak rate, less than the full line flow rate.
The same deficiency was found in the spare pump in storage.
The pump is a Peerless 83AD10 Horizontal Split Casing Model.
The design error was brought about by common industry practice of not providing flange joint bolting torque requirements to construction and maintenance personnel.
This practice is compounded by the wording in ANSI B31.1 para.
108.3 and 108.5.2 that leads the user to believe a joint assembled in accordance with the code descriptions would be bolt strength limiting.
Calculations performed during this investigation prove this not to be the case.
Calculations that a torque of 78.6 ft. lbs., could cause a flange failure.
Since it is not known what torque was applied to this joint nor what torque has been applied to the other joints of 'this configuration, and assuming the worst condition of having the joint torqued to near its failure point, then it is assumed that the joint could not withstand the added load of a seismic event.
This item will remain open for additional inspector followup.
(259/85-32-02)
The licensee identified a deficiency with the reactor building crane during this report period.
The 125-ton reactor building crane was designed for earthquake loads which were calculated on the assumption that the crane and supporting structure were rigid in the vertical and longitudinal (N-S)
directions of motion.
Also, the reactor building main steel framing, runway beams, and crane rails were designed for the seismic loads from the crane analysis.
During an evaluation in July 1982 to determine if the crane could support a 105-ton spent fuel cask, it was found the assumptions that the crane and support structure are rigid were not correct.
The horizontal inertial loads from an OBE or SSE level earthquake would be transmitted by the crane through the end trucks to the crane rails.
The clamps and bolts securing the rails in position on the runway beam could be overstressed and the cradle rails could overturn and/or be displaced resulting in the crane dropping up to 6 inches on the runway beam.
The failure is expected to be in the load carrying structural system for excitation in the longitudinal direction of motion.
The affected parts of the crane and support system are the crane wheel housing (the end truck),
the crane rail, rail clamps, and bolts.
These vary in the amount of over-load they will experience depending on the position of the crane during a seismic event.
The crane is a 125-ton Ederer Corporation crane, serial No.
E-5589, TVA contract No.
This deficiency resulted from a failure to consider in the original seismic'nalysis the dynamic interaction between the reactor building crane and the supporting structure.
In the crane.analysis, the supporting beams were assumed to be rigid.
The building designer designed the supporting beams to meet allowable stresses and deflections without considering stiffness requirements and did not understand how the crane loads were determined.
Also, crane designers did not understand how the response spectra for the crane was generated and they did not followup to determine that the infor-mation provided to the building designers was used properly.
This item will be left open for further followup by the inspector.
(259/
85-32-03).
Surveillance Testing Observation (61726)
The inspectors observed and/or reviewed the below listed surveillance procedures.
The inspection consisted of a review of the procedures for technical adequacy, conformance to technical specifications, verification of test instrument calibration, observation on the conduct of the test, removal from service and return to service of the system, a review of test data, limiting condition for operation
'met, testing accomplished by qualified personnel, and that the surveillance was completed at the required frequency.
a.
S.I. 4.8.A.3 Strontium, Fe-55, P-32, Tritium, and Gross Alpha Analysis - Liquid Effluent b.
TI-38 Proportional Composite Sample Preparation
c.
S.I. 4.1.B-2 Reactor Protection System APRM Output Signal Adjustment d.
S. I. 4. 1.8-15 Flow Bias Adjustment for Changes in the "R" Factor 8.
Reportable Occurrences (90712, 92700)
The below listed licensee events reports (LERs) were reviewed to determine if the information provided met NRC requirements.
The determination included:
adequacy of event description, verification of compliance with technical specifications and regulatory requirements, corrective action taken, existence of potential generic problems, reporting requirements satisfied, and the relative safety significance of each event.
Additional in-plant reviews and discussion with plant personnel, as appropria'te, were conducted for those reports indicated by an asterisk.
The following licensee event reports are closed:
LER NO.
"296/84"11
"296/84-13 Date 10-02-84 11-22-84 Event RHR Testable Check Valve Leakage Limitorque Motor Pinion Gear Inspection
"296/85-03
- 296/85-04
"296/85-12 1-11-85 1-22-85 4-25-85 HPCI Inoperable Late Surveillance Test Loss of Secondary Contain-ment
- 296/85-13 296/85-14 296/85-15 4-30-85 4-29-85 5-15-85 Excessive Closure Time on Purge Valves Containment Isolation Turbine First Stage Pressure Setpoint Drift
"296/85-16 5-17-85 Diesel Generator Fuel Line Support Weld Crack No violations or deviations were identified in the above area.