IR 05000259/1986028
| ML18031A863 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/18/1986 |
| From: | Brooks C, Cantrell F, Patterson C, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18031A861 | List: |
| References | |
| 50-259-86-28, 50-260-86-28, 50-296-86-28, NUDOCS 8610100542 | |
| Download: ML18031A863 (21) | |
Text
~P,It RbCI
'Jp O~
ss It
.0 e cs 4~
~O
+**dr+
UNITEO STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/86-28, 50-260/86-28, and 50-296/86-28 Licensee:
Tennessee Valley Authority 6N 38A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Docket Nos.:
50-259, 50-260, and 50-296 License Nos.:
DPR-33, DPR-52, and DPR-68 Facility Name:
Browns Ferry Nuclear Plant Inspection Conducted:
August 1-31, 1986 Inspectors:
G.
L. au, Sensor ess nt A~4=
roo
, Res>dent Approved by:
antre
,
ect n
Division of Reactor Pr ects ate 1gne te sgne
~et r5 ate igne ate igne SUMMARY Scope:
This routine inspection was in the areas of operational safety, mainten-ance observation, surveillance testing observation, and reportable occurrences, whole body count procedures, and TMI action item status.
Results:
No violations or deviations.
8610100542 860929 PDR ADOCK 05000259
DETAILS Licensee Employees Contacted H.
P.
J.
G.
J.
P.
R.
L.
E.
A.
J.
E.
R.
M.
T.
D.
T.
F.
D.
C.
J.
G.
M. J.
R.
D.
A.
W.
R.
E.
Pomrehn, Site Director Walker, Deputy Site Director Stapleton, Project Engineer Lewis, Plant Manager Grimm, Assistant to the Plant Manager Swindell, Superintendent
- Unit Three McKeon, Superintendent
- Unit Two Cosby, Superintendent
- Unit One Ziegler, Superintendent
- Maintenance Mims, Technical Services Supervisor Turner, Manager - Site equality Assurance May, Manager - Site Licensing Schulz, Compliance Supervisor Sorrell, Health Physics Supervisor Jackson, Chief Public Safety Other licensee employees contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, public safety officers, quality assurance, design and engineering personnel.
Exit Interview (30703)
The inspection scope and findings were summarized on September 3,
1986, with the Plant Manager and/or Superintendehts 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 inspection.
Licensee Action on Previous Enforcement Matters (92702)
(Closed)
Violation (259/260/296/85-36-07)
The corrective action committed to in the licensee's response to this Violation was found to have been adequately implemented.
(Closed)
Unresolved Item (259/260/296/85-45-07)
Although the cause of the damaged cable in RPS panel 9-17 was attributed to heat generated from a loose termination at the breaker, the licensee further identified that the cable was undersized.
The No.
AWG cable, rated at 75,amps, was improperly applied in this instance since circuit protection is provided by a 100 amp (nominal) breaker.
Additionally, calculations show that at full power with all scheduled modifications completed, the cable would have been passing
amps.
The licensee performed a safety evaluation of this condition (cable overloaded with no protection)
and due to the fail safe and coincidence
design features of the RPS System it was concluded that safe shutdown of the reactor would not have been prevented.
The No.
AWG cable has been replaced with No.
AWG cable on all units.
(Open)
Inspector Followup Item (259/260/296/86-16-03)
Plant Procedure BF-12.24, Conduct of Operations, specifies that the shift engineer is responsible for insuring that proper records and logs are maintained in a
neat, legible manner for his assigned shift and that logs contain an accurate and complete account of events occurring on his shift.
This includes a review of all logs at least once per shift.
The shift engineer is to initial the logs indicating the review has been performed.
Although a non-licensed operator sometimes maintains the control log when the unit is defueled, a licensed operator is still responsible for the logs.
However, a
review of logs conducted on August 28, 1986, found that implementation of the review was only being done 50K of the time.
This item will remain open.
(Closed) Violation (259/260/296/86-06-04)
The licensee's training material concerning clearance procedure adherence was reviewed.
The licensee's corrective action was broad and recognized the problem as a lack of aware-ness by personnel concerning the importance of clearance precautions.
A site-wide familiarization course was undert'aken and large signs displaying the importance of hold orders and tagouts were placed throughout the plant.
This item is closed.
~
~
(Closed)
Unresolved Item (259/83-60-06)
A new procedure was written to control the issue of oil.
The procedure, Plant Manager's Instruction PMI-6. 1, Storage Control and Issue of Lubricants, requires that a quality assurance representative verify the issue of any oil for CSSC equipment.
This item is closed.
(Closed) Violation (259/260/296/84-26-04)
A general revision to Surveillance Instruction SI-4.5.C, Residual Heat Removal Service Mater System and Emergency Cooling Mater System, was made to clarify which valves require testing.
This item is closed.
(Closed) Violation (259/260/296/85-15-10)
The instrument checks and obser-vation Surveillance Instruction (SI-2)
was revised to ensure that compared reactor water level instruments are independent.
The calibration accuracy in inches is specified for each instrument.
This item is closed.
(Closed)
Unresolved Item (259/84-15-06)
Table 3. 11.A, Fire Protection System Hydraulic Requirements, has been corrected in the plant technical specifica-tions to include all fixed spray systems.
This item is closed.
(Closed) Inspector Followup Item (259/260/296/84-52-04)
Electrical Maintenance Instruction EMI-7, Maintenance of Medium and Low Voltage Switchgear, has been revised to include a test of the shunt field of direct current Limitorque operators on primary containment isolation valves.
This item is closed.
(Open) Unresolved Item (259/260/296/86-25-13)
This item will remain open as discussed in paragraph.
Unresolved. Items" (92701)
An unresolved item is identified in paragraph 5.b.
and 5.d relating to the scram air header blowdown special test conducted on July 7, 1986, and circuit breaker control power, respectively.
Operational Safety (71707, 71710)
The inspectors were kept informed of the overall plant status and any significant safety matters related to plant operations.
Daily discussions were held with plant management and various members of the plant operating staff.
The inspectors.
made routine visits to the control rooms 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 supervisors.
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'ower supply and breaker; alignments; radiation area controls; tag controls on equipment; work "activities in progress; and radiation protection controls.
Informal discussions were held with selected plant personnel in their functional areas during these tours.
Weekly verifications of system status which included major flow path valve alignment, instrument alignment, and switch position alignments were performed on the Residual Heat Removal Service Mater and Emergency Equipment Cooling Mater Systems.
A complete walkdown of the accessible portions of the Control Rod Drive Scram Air Header 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.
t An Unresolved Item ss a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviatio In the course of the monthly activities, the inspectors included a review of the licensee's physical security program.
The performance of various shifts of the security force was observed in the conduct of daily activities to include:
protected and vital areas access controls, searching of personnel, packages and vehicles, badge issuance and retrieval, escorting of visitors, patrols and compensatory posts.
In addition, the inspectors observed protected area lighting, protected and vital areas barrier integrity.
A confidential security concern was brought to the attention of the TVA Personnel Security Chief on August 29, 1986.
During a routine tour on August 13, 1986, the inspector found a diesel generator fuel oil transfer valve, 3-18-611, unlocked and closed.
This valve connects the fuel oil storage tanks to the Unit 3 diesel generator building and associated seven day fuel oil tanks.
The valve is located outside the plant in the yard underneath grating.
A chain and padlock were laying next to the valve.
Plant operations personnel stated the valve was in the required position by plant Operating Instruction OI-18.
No require-ment was known for the valve to be locked.
The normal valve position is open on plant drawing 47W840-1.
The licensee initiated steps to correct the drawing discrepancy.
This will remain an Inspector Followup Item pending correction of the drawing (296/86-28-01).
Technical Specification Interpretation The licensee's Technical Specification Interpretation Committee (TSIC)
issued an interpretation of T. S. l. 0. AA regarding surveillance intervals on July 31, 1986.
The interpretation stated that the maximum allowable extension of surveillance intervals (25K)
as specified in T.S.
1.0.AA could be applied to surveillance requirements such as 4.7.E. 1, 4.7.E.3, 4.7.F. 1, 4.8.D. l.a and 4. 11.A.5.
These surveillance requirements specify intervals on a "not to exceed" or an "intervals no greater than" basis.
Following discussions with regional management and the NRR Licensing Project Manager, the inspectors informed licensee representatives that no extensions were allowed for intervals so explicitly stated.
b.
Special Test 86-10 The licensee performed Special Test 86-10, Backup Scram Valve Blowdown Test, on July 7, 1986.
The purpose of this test was to determine scram air header blowdown time in order to complete the design input for alter nate rod injection (ARI) system ATWS design alternatives.
Normally,. following actuation of the backup scram valves, the scram outlet valve opens in about 4 seconds followed by the scram inlet valve in another 3 seconds.
During the special test which was performed on Unit 1, several scram outlet valves delayed opening for about
seconds and another
seconds elapsed until the scram inlet valves opened.
This data would indicate the presence of some restriction in the scram air header or a problem in the scram solenoid pilot valves.
The licensee intends to determine the cause of the anomalous behavior and also evaluate the potential for Units 2 and 3 concerns.
This will be tracked as as Unresolved Item pending resolution by the licensee
(259/260/296/86-28-02).
These problems may be related to scram solenoid pilot valve discrepancies reported in General Electric's Service Information Letter No. 441, Control Rod Drive Scram Anomaly.
While reviewing this event, the inspector noted several deficiencies in the scram air headers at Browns Ferry.
These are identified as follows:
(1)
An unidentified pressure gauge installed above hydraulic control unit (HCU) 30-43 on Unit 1 was reading 150 psig.
This is impossible since the air pressure input to the header is controlled at about 70 psig and several other pressure gauges were indicating about
psl g.
(2)
Electrical conduit was being supported by bailing wire wrapped around the scram air header above HCU 26-55 on Unit 3.
(3)
Many pipe clamps are missing on the Unit 3 air headers above the HCUs.
These deficiencies were reported to licensee representatives during a routine daily meeting on August 20, 1986.
Thermal Overload Program The licensee identified in December 1985, that design documents do not reflect the overload element ratings for Motor Control Center (MCC)
Starters.
It is unclear whether the overloads have been properly specified by designers in the absence of such documentation.
The problem is detailed in Significant Condition Report (SCR)
BFNEEB8536.
Corrective action is to consist of the following for each overload element:
(1) field verify the rating, (2) design review for adequacy of application, (3) replacement of any improperly applied elements, and (4) revision to reflect element ratings on drawings.
As of this month, the program has progressed to the field verification phase with the majority of Unit 2 and common electrical boards being completed.
This program will be tracked as an Inspector Followup Item (259/260/296/86-28-03) to record the progress and follow any adverse findings.
Confusion Over Switch Labeling During a routine tour on August 19, 1986, the inspector found a circuit breaker identi fication 1 abel on the floor of both the 3A and 3B 480 volt shutdown board rooms.
This was discussed with plant management on August 20 in the daily noon meeting.
Also, discussed was a problem with the adhesive'olding the labels on the circuit breaker cabinet as several labels were becoming loose.
On August 28, 1986, the inspector toured the 3A room and found the identification label still on the floor.
The label was for the 3A standby liquid control pum At
A more detailed inspection conducted by the inspector of the 480 volt shutdown panels found additional labeling problems.
Circuit breaker control power is supplied to the shutdown boards from a normal and alternate (emergency)
supply taken from 250 volt direct current battery boards.
A spring return to the neutral position transfer switch is used to change the source of the control power.
The labeling of the switch may be incorrect and additional labeling was found penciled onto the breaker cabinet.
The label read EMERGENCY-NORMAL, but penciled above EMERGENCY was NORM and above NORMAL was ALT.
In the 3A room also penciled next to the transfer switch was that the label tag was wrong.
A tour of the unit one and two rooms found differences in labeling among the units.
The inspector discussed these. problems with plant management on August 28, 1986.
This problem is similar to a previous item which lead to the discovery of the incorrect wiring of control power to a 4160 volt shutdown board (IE Report 85-57 and Licensee Event Report 259/85056).
The inspector took a plant superintendent on a tour of these rooms on August 28.
Testing of the boards will be necessary to verify the correct electrical connections.
The switch design and labels may be inadequate for plant operations.
Resolution of the switch wiring, design, and labeling will remain unresolved until the licensee's evaluation is completed (259/260/296/86-28-04).
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 radiological controls were implemented as required.
Maintenance requests were reviewed to determine status of outstanding jobs and to assure that priority was assigned to safety-related equipment main-tenance which might affect plant safety.
The inspectors observed the below listed maintenance activities during. this report period:
a.
Replacement of fire hydrant no.
15 in the switchyard.
b.
Diesel generator fuel oil tank cleaning.
c.
Control Room Emergency Ventilation System Thermal overload trouble-shooting activities.
d.
Removal, repair and reinstallation of Reactor Protection System Motor-Generator (RPS MG) sets.
No violations or deviations were identified in this paragrap.
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 requi red frequency.
EECM Flow-Rate Test The inspector observed the performance of Surveillance Instruction SI 4.5.C. 1.(4),
Emergency Equipment Cooling Mater (EECM) System Annual Flow Rate Test conducted on August 22, 1986, in the Unit 3 control room.
This test was terminated due to numerous problems.
The inspector requested to be informed when the test was repeated, The test was performed on August 23, 1986, without difficulty, but the inspector was not informed of the repeat performance.
This item will remain unresolved until the inspector observes satisfactory performance'f the test conducted in accordance with the approved procedure.
This test fulfills Technical Specification requirement 4.5.C. l.b.
Annually each residual heat removal service water (RHRSW)
pump is required to be flow rate tested.
Each pump must pump at least 4500 gpm through its normally assigned flow path.
There are
RHRSM pumps at the site, and four of the
pumps are categorized EECM pumps.
The test on August 22, 1986, was conducted for the 'C3'ECW pump.
The test for EECM pumps requires running t5e pump under test only, while opening additional valves as necessary to bring the flow up to 4500 gpm.
After the 'C3'ump was started, the flow was 3800 gpm.
The first valve in the procedure to open was the cross connect valve to the reactor building closed cooling water (RBCCM) system for Unit 1.
No flow increase was observed after attempting to open the valve.
The valve position indication was intermediate showing both a red and green light, Next, the cross connect to Unit 2 was opened and the flow increased to 4200 gpm.
Again, both a
red and green valve position light was observed and several minutes were required for the flow to increase.
At this point a decisios was made by the operator to reopen the Unit 1 valve.
While waiting for the valve to open, two EECW pumps (B3 and 03) were observed to have started unexpectedly.
These pumps had been started from the common Units 1 and 2 control room for a diesel generator test.
At this point the Unit 3 operator terminated the EECW flow rate test.
Precaution and limitation 3.4 of SI 4.5.C. 1.(4) requires the unit operators to establish communications with each other whenever changes to the system are made.
This precaution was not observed and the unexpected start of two EECW pumps occurred while a flow rate test was in progress on another pum The EECM cross connect valves to RBCCM are described in Final Safety Analysis Report Section 10. 10. 3.
The valves are hydraulically-operated back-pressure valves using EECM header pressure and water as the hydraulic medium.
Normally the valves are shut but upon loss of raw cooling water pressure (15 lbs.
at the heat exchanger),
the valves open.
However, with normal raw cooling water pressure the valves are difficult to open due to the decreased differential pressure across the valves.
This was the probable cause for the intermediate position indication observed during the test.
However, the test procedure makes no reference to this peculiarity and the operator conducting the test was unaware of these system quirks.
The inspector reviewed the surveillance instruction (SI) cover sheet for recording of any delays or-problems during the test.
The remarks indicated that the SI was stopped on August 22, due to conflict with another SI.
No mention was made of the difficulty in opening the valves.
It was indicated that no revision was required to correct any problems with the procedure.
Also, the completed SI was a combination of data sheets from the initial and repeat tests.
A complete record of the initial test was lost due to not retaining all of the data sheets.
Unless a person witnessed the actual test there would be no way to distinguish or delineate system deficiencies.
Due to the difficulty of opening-the cross-connect valves, the inspec=-
tors questioned whether the valves were ever operationally tested.
The cognizant engineer stated the valves have not been tested in the past but would be included in the pump and valve test program.
This program has not been approved yet and is required for Unit 2 startup.
FSAR Section 10. 10.3 states that these valves are designed to shut off flow to the equipment (RBCCM) on low header pressure in order to guarantee flow to the RHR and core spray environmental coolers, the RHR seal heat exchangers, the control building emergency cooling unit, and the diesel generators.
Addition of these valves to the test program will be an Inspector Followup Item (259/260/296/86-28-05).
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
'echnical 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 appropriate, were conducted for those reports indicated by an asterisk.
The following licensee event reports are closed;
LER No.
Date Event 8259/86-13 259/86-15
"259/86"08 4-22-86 4-29-86 2-12-86 Loss of Secondary Containment in the Refuel Zone During Ventilation Damper Maintenance Reactor Protection System Trip Due to Voltage Regulator Control Switch Being Accidentally Bumped Diesel Generator Air Start System Malfunction Due to Corrosion Products No violations or deviations were identified in this paragraph.
9.
Whole Body Count Procedures The inspector reviewed procedural requirements and implementation related to the mobile whole body count system used by the licensee to document the in vivo component of the internal dosimetry program.
Portions of the following procedures-.were verified to ascertain regulatory=
requirements and ANSI N 343/1978 recommendations were met:
(a)
QAB Audit Report No.
QSS-A-85-0010.related to Internal Radiation Control.
(b)
Health Physics Dosimetry Section Letter No.
6 for Browns Ferry which describes mobile whole body counter operation.
(c)
Radiological Health Dosimetry Section Procedure, OOS-3, for calibration of the mobile whole body counter system for annual calibrations in 1985 and 1986 (WBC-12).
(d)
Health Physics Dosimetry Section Instruction Letter No.
6 for Browns Ferry which describes the mobile whole body counter system operation.
(e)
Training qualification record for the whole body count operator on duty on August 5, 1986.
(ESTTSIL-4) - Engineering Aide Performance Verification Sheet.
.(f)
Radiological Control Instruction 8 for the Radiobioassay Program
'mplementation.
(g)
Required monthly resolution checks (July 8 August 1986) for the mobile WBC detector for the 662 KeV photopeak of CS-137.
(h)
Records of standard check source performance completed daily on the WB pl
The program overall was adequate in meeting the ANSI N 343 recommendations and regulatory requirements; however, several minor deficiencies were noted as listed below:
(1)
The mobile WBC may be run in a SCAN or FIXED mode.
The inspector observed the operation of the mobile unit in the fixed mode since the scan mode was mechanically broke.
The operational procedure, DSIL-6, did not adequately address the method the operator should use to use the fixed method.
DSIL-6 did not address usage of Table II (Stationary Mode Action Levels) in step 5.3 when the fixed mode is used for the normal count method.
(2)
Performance Verification Sheet -
E 8 TTSIL-4 did not adequately specify all required reading for adequate qualification performance in step II of Appendix I.
Specifically, the operational instruction, DSIL-6, was not required for knowledge demonstration in step II.
The licensee took action to correct the above listed deficiencies immediately.
The inspector observed the anti-claustrophobial device working adequately during his body count procedure.
There were no violations or deviations in this area.
t 10.
Status of TMI Action Items for Browns"Ferry The resident reviewed the status of TMI Action Items, mented as follows:
The status is docu-1)
Item I.C. 1.2.b and I.C. 1.3.b - Short-team Accident and Procedures Last letter:
TVA to NRC 3-26-86, TVA and NRC both have items to close, TVA has an action to do EOI training prior to startup.
2)
Item II.E.4. 1 - Dedicated Hydrogen Penetrations BFN has H2 penetrations.
TVA owes NRC a letter, NRC must evaluate response.
Last correspondence:
1-25-82, TVA to.NRC.
3)
Item II.F. 1. 1 and II.F. 1.2 - Accident Monitoring - Noble Gas TVA has action to install monitors before startup.
TVA to NRC letter:
11-28-84.
4)
Item II.F.1.3,4,5 - Accident Monitoring High-Range, Containment Pressure, Containment Water Level TVA is to install equipment prior to startup Last letters:
II.F. 1. 3-NRC letter 8-24-84.
II.F. 1.4-TVA letter 3-7-84.
II.F. l. 5-TVA letter 3-7-8 ps
5)
Item II.F.2 -
Instrumentation For Detection of Inadequate Core Cooling TVA action to re-route water level reference legs.
Last letter:
TVA letter of 3-12-86.
The modifications are to be complete prior to cycle 8 operation on each unit.
Last NRC letter is Generic Letter 84-23.
6)
Item II.K.3. 18 -
ADS Actuation TVA's current position is not to perform mods.
Last letter is TVA letter of 3-26-86.
NRC must respond to last letter.
7)
Item II.K. 3. 28 - qualification of ADS Accumulators TVA has action to install backup N2 system.
Installation'will be part of integrated schedule.