IR 05000259/1990032
| ML18033B566 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/15/1990 |
| From: | Hunt M, Merriweather N, Mark Miller NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18033B565 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 50-259-90-32, 50-260-90-32, 50-296-90-32, NUDOCS 9012040199 | |
| Download: ML18033B566 (19) | |
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UiNITEO STATES NUCLEAR REGULATORY COMMISSION
REGION II
10'I MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/90-32, 50-260/90-32, and 50-296/90-32 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 1, '2, and
Inspection C
te
ctober 22-26, 1990 Inspector
ri ather
. Hunt, cting hief nt Systems Section Engineering Branch Division of Reactor Safety Date Signed l~ />
e>
ate cygne rs <J ate Signe SUMMARY Scope:
This special, announced inspection was conducted in the areas of the licensee's conformance to Regulatory Guide (RG) 1.97, Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following on Accident and previously identifi.ed open items(s).
Results:
In the areas inspected, violations or deviations were not identified.
The licensee has developed an adequate program of plant instrumentation to meet the intent of Regulatory Guide
[RGj 1.97, Revision 3,
the NRC Safety Evaluation Reports, and their submittals with the exception of one recent submittal.
This submittal dated October 15, 1990 was sent to the NRC for review and approval.
It is further discussed in Paragraph 2.
The licensee has prov'ided adequate controls and planning to ensure that the RG 1.97 5'012040l5'5 901115'DR ADOCK LI5000259 F Eo-
program, as committed to by TYA to the NRC, will be fully implemented prior to Unit 2 restart.
The inspectors determined that the majority of the RG 1.97 program has been implemented with the exception of completing several plant modifications and performing the instrument surveillance and calibration requirements.'Stren the Engineering calculations are being prepared for all RG 1.97 variables and instruments to ensure that the complete design and installation, meets all commitments and requirements.
This, is further discussed in Paragraph 2.
Weaknesses A minor concern not related to RG 1.97 was identified with the plant modification program.
The licensee was not fully implementing instructions for lifting and tagging electrical leads.
This is further discussed in Paragraph REPORT DETAILS Persons Contacted Licensee Employees=
- M. Bajestani, Electrical/ISC Systems Supervisor
- S.
D. Brown, Modification Supervisor
- P. Carier, Site Licensing Manager
- J. Caron, Operations Manager
- T. L. Chandler, I&C Engineering
- J. Daniel, Site Programs
- W. H. Deen, Mechanical Engineering
- J.
C. Delockery, Task Manager Engineering
- J. Glass, Seismic Engineering
- R.
W. Johnson, Modifications Manager
- N. C. Kayanar, Vice President On Site
- L. W. Myers, Plant Manager
- R. R. Reeves, ISC Specialist
- E. E. Ridgell, Licensing
- J. Rupert, Project Engineer
- P. Salas, Compliance/Licensing Supervisor
- J. C. Traymor, Project Nanger H. Weber, Site Engineering and Modifications Manager
- 0. V. Zeringve, Site Director Other licensee employees contacted during this inspection included engineers, operators, technicians, and administrative personnel.
NRC Resident Inspectors
- W. C. Bearden
- D. R. Carpenter
- E. F. Christnot
- K. Ivey
- Attended exit interview Inspection of Licensee's Implementation of Multiplant Action Item A-17:
Instrumentation for Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident (Regulatory Guide 1.97)
(25587)
Criterion 13,
"Instrumentation and Control," of Appendix A to
CFR Part 50 includes a requirement that instrumentation be provided to monitor variables and systems over their anticipated ranges for accident conditions as appropriate to ensure adequate safety.
Regulatory Guide 1.97 describes a method acceptable to the NRC staff for complying with the Commission's regulations to provide instrumentation to monitor plant variables and systems during and following an acciden The licensee responded to RG 1.97 (NUREG 0737, Supplement 1) in letters dated April 30, 1984; May 7, 1985; August 23, 1988; and October 15, 1990.
The NRC Safety Evaluation Reports (SERs)
for RG 1.97 were issued January 23, 1985; June 23, 1988; January 19, 1989; and February 8, 1990.
The NRC SER dated February 8, 1990 concluded that the licensee's design is acceptable with respect to conformance to RG 1.97, Revision 3, with the exception of the neutron flux variable.
The.existing neutron flux
'nstrumentation is only acceptable for interim operation.
a 0 The licensee submittal dated October 15, 1990 provides a response to the NRC SER dated February 8, 1990 to address four previously submitted deviations to RG 1.97.
These four deviations are (1) Neutron Flux, (2) Reactor Coolant System Pressure, (3)
RHR Heat Exchanger Outlet Temperature, and (4) Primary Containment Isolation Valve Position Indication.
This submittal also identified two new deviations to RG 1.97 for which NRC review and approval is required.
These two new deviations are for (1) Primary Containment Isolation Valve (PCIV) Position For The RHR Shutdown Cooling Valves and (2)
PCIV Position Indication For The TIP Ball Valves.
The two new deviations were not addressed as part of this inspection although they were reviewed.
This inspection assessed the licensee's RG 1.97 instrumentation program using (1) the design and qualification criteria described in RG 1.97, Revision 3; (2)
the EGKG Technical Evaluation Report No. EGG-EA-6873 dated August 1989, Conformance to Regulatory Guide 1.97, Browns Ferry Units 1,
and 3;
(3) the licensee's submittal as discussed previously; and (4)
CFR Part 50.
A random sample of 14 variables from the licensee's submittal was selected to evaluated the licensee's program.
The variables selected were classified as Category
and 2 which have the most stringent design requirements of all RG 1.97 instruments.
The instruments examined and the results achieved are discussed in the paragraphs and tables below.
b.
Category 1 and 2 Instruments The instrumentation listed in the following Tables was examined to verify that the design and qualification criteria for RG 1.97,'he SERs and licensee commitments had been satisfied.
The instrumentation was inspected by reviewing drawings; procedures, data sheets and other documentation; and performing walkdowns for visual observation of selected installed equipment including control room indicators and recorders.
The following areas were inspected:
( 1)
Equipment Qualification - The EQ Master Equipment List, Q-list, ISC List, and instrument drawings were reviewed for confirmation that the licensee had addressed environmental and seismic qualification requirement (2)
Redundancy
~
Walkdowns were performed to verify by visual observation that selected instruments were installed as specified and that separation requirements were met.
In addition, wiring drawings for all listed Category I instrumentation were reviewed to verify redundancy and channel separation.
Power Sources
- Wiring drawings were reviewed to verify the instrumentation is energized from a safety-related power source if applicable.
Display and Recording - Walkdowns were performed to verify by visual observation that the specified display and recording instrumentation were installed.
Wiring drawings were reviewed to verify there was at least one recorder in a redundant channel and two indicators, one per division (channel) for each measured variable where required.
Range - Walkdowns were performed to verify the actual range of the indicator/recorders was as specified in RG 1.97 or as stated in the licensee's submittal.
Review of calibration procedures verified sensitivity and overlapping requirements of RG 1.97 for instruments measuring the same variable.
Interfaces
-
The wiring drawings, ISC List, and g-list were reviewed to verify that safety-related isolation devices were used when required to isolated the circuits from nonsafety systems.
Direct Measurement
- Wiring drawings were reviewed to verify that the parameters are directly measured by the sensors.
Service, Testing, and Calibration - The maintenance program for performing calibrations and surveillances was reviewed and discussed with the licensee.
Calibration and surveillance procedures, the latest data sheets, and the calibration schedule for each instrument were reviewed to verify the instruments have a valid calibration or will be calibrated prior to plant startup.
Equipment Identification - Walkdowns were performed to verify that Types A, B and C instruments designated as Categories I and 2 were specifically identified with a
common designation on the control panel Variable TABLE I CATEGORY I INSTRUMENTS Instruments Reactor Coolant Level Reactor Pressure LT-003-52 LIS-003-52 LI-003-52 LT-003-62A LIS-003-62A LI-003-62B LT-003-58B LIS-003-58B LI-003-58A LT-003-58D LIS-003-58D LI-003-58B SPDS PT-003-74A PIS-003-74A P I-003-74A PT-003-74B PIS-003-74B PI-003-74B SPDS TABLE
CATEGORY I INSTRUMENTS Variables Containment Hydrogen Concentration Primary Containment Area Radiation High Range Instruments AN-76-37 H2I-76-37 H2R-76-104 AN-76-39 H2I-76-39 H2R-76-94 SPDS RE-90-272C RM-90-272C RR-90-272CD RE-90-273C RM-90-273C RR-90-273-CD
TABLE 1 CATEGORY
INSTRUMENTS Variable Primary Containment Isolation Valves Position Instruments FCV-73-2,3,26,27,30,31 FCV-74-47,48,53,57,58,60, 61,67,71,72,74,75 FCV-76-17, 18, 19,24 Suppression Pool Temperature TE-64-161-ATH, J, K,L TM-64-161-ATH,J, K,L TI-64-161 TR-64-161 TE-64-162-ATH,J,K,L TM-64-162-ATH,J,K,L TI-64-162 TR-64-162 SPDS Variable Drywell Atmosphere Temperature Drywell Pressure Suppression Pool Water Level TABLE 1 CATETORY
INSTRUMENTS Instruments TE-64-52A TM-64-52AA TI-64-52 TE-64-52C TM-64-52CA XR-64-50 SPDS PT-64-160A P I-64-160A PX-64-160A XR-64-50 PT-64-67B P I-64-67B PX-64-67B XR-64-59 SPDS LT-64-159A LI-64-159A LT-64-159 8 XR-64-159 SPDS
'P
Variable TABLE 2 CATEGORY 2 INSTRUMENTS Instruments Primary Safety Relief Relief Valve Position High Pressure Coolant Injection (HPCI)
Flow Cooling Water Temperature To Engineered Safety Features
[ESF]
XE-1-41-4,5,18, 19,22,23,30, 31,34,41,42,179,180 XI-1-41-4,5,18,19,22,23,30, 31,34,41,42,179,180 FT-73-33 FIC-73-33 TE-27-144A TM-27-144A TI-27-144A Variables TABLE 2 CATEGORY 2 INSTRUMENTS Instruments Status of Standby Power Amperes Voltage VARS Emergency Ventelation Damper Positions Diesel Generators A,B,C,D,3A,3B,3C,3D, Ammeters Voltmeters Varmeters FCO-64-5
[ASB]
FCO-64-6
[AdmlB]
FCO-64-13
[ASB]
FCO-64-14
[ASB]
[AIlB] Dual Unit
Discussion and Conclusion The licensee was well prepared for the inspection and extremely cooperative in providing assistance to the inspectors.
All documentation, drawings, calculation sheets, and calibration procedures for Categories I and 2 instrumentation were immediately available.
Knowledgeable engineers were assigned to provide assistance, answer questions, and assist in plant walkdowns.
The inspectors concluded that Browns Ferry has an adequate program to meet the intent of RG 1.97, the NRC safety evaluation reports, and their submittals.
The licensee has developed adequate controls and planning to ensure their RG 1.97 commitments to the NRC are fully implemented prior to Unit 2 startup.
Overall, the inspectors consider the RG 1.97 program satisfactory, and the engineering calculation packages to be exceptional.
However, several minor concerns were identified with the drawings and electrical lifted
'leads during plant modifications.
A brief discussion of the areas reviewed, the strengths, the weaknesses, and any licensee planned actions are summarized below.
Calibration Since Browns Ferry has not been operational for several years, the inspectors reviewed the calibration program for the RG 1.97 instrumentation and electrical equipment.
Surveillance procedures and calibration procedures were reviewed to ensure all instruments in each loop were included.
In addition, the electrical calibration procedures for the diesel generator power status and the motor operated isolation valves was examined.
Both of these electrical programs were very good.
The inspectors also reviewed the calibration program to ensure all instrumentation was scheduled prior to Unit 2 startup.
Overall, the calibration program for RG 1.97 instrumentation was found to be adequate, and all instrumentation and electrical equipment was either calibrated or scheduled to be calibrated.
Loop Drawings The inspector reviewed both electrical wiring drawings and instrumentation loop drawings.
Primarily the purpose for wiring drawings is for use during construction.
In many instances several sheets of wiring drawings are required for one instrument loop.
However,
'a single loop drawing can usually detail the complete instrument loop and its logical operation.
These loop drawings are an ideal tool in operating plants for training, troubleshooting, and calibration.
Loop drawings at any plant are considered a significant strength for the maintenance and system engineering departments.
Initially the licensee had indicated to the inspectors that the loop drawings would no longer be maintained.
However, TVA management has
now taken the position that the need for loop drawings will be evaluated and will take action based on the results of the evaluation.
f.
Engineering Calculation The licensee has developed 'an engineering calculation work package for each instrument loop.
Each work package reviews the design and installation to ensure all components meet NRC regulatory requirements and licensee commitments.
All electrical and mechanical parameters are examined.
Setpoints and calibration calculations are reviewed.
Each package contains items and requirements for g-list, Eg, calibration procedures, list of cables,
'drawings, cable separation, instrument line separation and slope, existing deficiencies, control room action-plan, documentation, and the technical basis.
The inspectors considered the engineering calculation work packages to be very thorough and a strength to the RG 1.97 program.
g.
Plant Modification A minor concern not related to RG 1.97 was identified with lifted electrical wires.
The licensee was not fully implementing modification instructions for lifting and tagging electrical wires during performance of plant modifications.
Appendix C, Lifting and Relanding Cables/Wires in Plant Modification Instruction MAI-3.3 requires all wires to be tagged with the work instruction number and terminal block number prior to lifting.
This requirement is important when the work cannot be performed by the same craftsman during the same shift.
The licensee stated appropriate corrective action would be taken to ensure all lifted wires in closed panels would be properly tagged; MAI-3.3, Appendix C would be revised to reflect current practices; and the plant modification staff (electrical) will be trained.
The inspectors considered this corrective action as acceptable.
3.
Action on Previous Inspection Findings (92701)
(Closed) Part 21, Report 84-01, ELMA Power Supplies Units 1 and
A 10 CFR Part 21 Report was made by TVA on January 11, 1984, identifying numerous problems found in ELMA power supplies (part number 164C5261P004).
The problems found involved defective wiring, cold solder joints and overall poor qualify workmanship.
A previous NRC Inspection (84-14)
examined the immediate corrective action taken to inspect and repair sixteen units that were in service.
However this item remained open pending completion of the final corrective action to replace all ELMA Power Supplies in the Reactor Protection System (RPS)
and Emergency Core Cooling System (ECCS).
Since that time the licensee has initiated ECN Nos.
P0706 and P0707 to replace ELMA Power Supplies on all three unit The ECNs have been completed on Unit 2 and one is partially field complete on Units I and 3.
These ECNs are being tracked by the Regulatory Compliance Open Items tracking system and are required to be completed on Units I and 3 prior to restart from their respective outages.
Based on the above, the inspectors concluded that no further action is required by NRC on this issue.
Therefore, this item is considered closed for all
units, 5.
'cronyms and Initialisms AN CCW ERF SPDS Eg ESF FCO FCV FI FIC FT H2I H2R HPI I gtC LI LIS LT LPI PORVA NRC NRR PCIV PI PIS PT px gAg-List RG-RE RM RR SER SW TE TM TR TVA VAR Analyzer Component Cooling Water Emergency Response Facility Safety Parameter Display System Environmental qualification Engineered Safety Features Flow Control Operator Flow Control Valve Flow Indicator Flow Indicating Controller Flow Transmitter Hydrogen Indicator Hydrogen Recorder High Pressure Injection Instrumentation and Control Level Indicator Level Indicating Controller Level Transmitter Low Pressure Injection Power Operated Relief Valve Nuclear Regulatory Commission Nuclear Reactor Regulatory (NRC)
Primary Containment Isolation Valve Pressure Indicator Pressure Indicating Controller Pressure Transmitter Special Pressure Recorder guality Assurance Equipment gualification List Regulatory Guide Radiation Element Radiation Monitor Radiation Recorder Safety Evaluation Report Service Water Temperature Element Temperature Monitor Temperature Recorder Tennessee Valley Authority Volt-Ampere Reactive
XE XI XR Acoustic Valve Position Element (pickup)
Acoustic Valve Position Indicator Special Recorder