ML20085N688
| ML20085N688 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/07/1983 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| GL-83-28, NUDOCS 8311110164 | |
| Download: ML20085N688 (33) | |
Text
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TENNESSEE VALLEY AUTHORITY CH ATTANOOG A, TENNESSEE 37401 400 Chestnut Street Tower II November 7, 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Denton:
In the Matter of the
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Docket Nos. 50-259 Tennessee Valley Authority
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50-260 50-296 Enclosed is our response to Generic Letter 83-28 dated July 8, 1983, subject, " Required Actions Based on Generic Implications of Salem ATWS Events," for the Browns Ferry Nuclear Plant.
Based on the enclosed information continued operation of Browns Ferry is justified and the licenses for Browns Ferry Nuclear Plant units 1, 2, and 3 should not be modified, suspended, or revoked.
Very truly yours, TENNESSEE VALLEY AUTHORITY 8311110164 831107 PDR ADOCK 05000259 P
PDR b
L. M. Mills, M nager Nuclear Licensing Subscribed gd sworn to j/ M,
fore me this 7 day of ///
1983 no M. h at i 051 APRM High-High Flux F 052 IRM Upscale. Trip on Level A 053 IRM Upscale Trip on Level C 054 IRM Upscale Trip on Level E 055 IRM Upscale Trip on Level G j
056 IRM Upscale Trip on Level B 057 IRM Upscale Trip on Level D 058 IRM Upscale Trip on Level F 059 IRM Upscale Trip on Level H 060 RPS Loss of M-G Set A 061 RPS Loss of M-G Set B 062 RFPT A Tripped-Punp Brg Oil Pressure Low 063 RFPT B Tripped-Pump Brg Oil Pressure Low f
064 RFPT C Tripped-Pump Brg Oil Pressure Low 065 RFPT A Tripped-Low NPSH 066 RFPT B Tripped-Low NPSH 067 RFPT C Tripped-Low NPSH 068 RFPT A Tripped Thrust Brg Excess Wear 069 RFPT B Tripped Thrust Brg Excess Wear 070 RFPT C Tripped Thrust Brg Excess Wear 071 RFPT A Tripped-Turb BRG Oil Pressure Low 072 RFPT B Tripped-Turb BRG Oil Pressure Low 073 RFPT C Tripped-Turb BRG Oil Pressure Low 074 RFPT A Tripped 075 RFPT B Tripped 076 RFPT C Trirped 077 Turbine Tripped-Condenser VAC Low' 078 Turbine tripped-Moisture SEP or Tank Level High' 079 Turbine Tripped-Stator Cooling Water System Failure 080 Turbine Tripped-Loss of Main Shaft Oil Pump' 081 Turbine Tripped-Low BRG Oil Tank Level' 082 Turbine Tripped-Vibration High' 083 Turbine Tripped-Backup overspeed Trip' 084 Turbine Tripped-Loss of Hydraulic Trip Pressure' 085 Turbine Tripped-Loss of Hydraulic Control Fluid Pressure
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089 Turbine Tripped-Loss of 24V DC Power *
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090 Turbine Tripped-Loss of 120V AC To EHC Control i
091 Main & RFP Turbs Tripped-Reactor High Water Level',..
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092 Turbine Tripped-Generator Overcurrent (151GX) 093 Turbine Tripped-Generator Negative Phase (146X) i 094 Turbine Tripped-Generator Neutral Overvoltage (159GNX) l 095 Turbine Tripped-Generator Differential (187GX) 096 Turbine Tripped-Main or USS Trans Failure (186C) 097 Turbine Tripped-Gen Backup-Gen Bkr Failure - Reverse Power Trans FDR Diff.
098 Turbine Tripped-USS Trans 1A Diff (1875XA) 099 Turbine Tripped-Main Breaker Failure (150BF or 186BF) i
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100 Loss of +30V EHC 101 Loss of -22V EHC 102 No EHC 250V DC input power 103 Power Load Unbalance 1
104 Intercept 71v Fast Closure 105 EHC Cab High Temp.
l 106 Malfunction Bus Energized 107 Main Trans Differential 108 Main Trans Sudden Pressure i
109 Gen Bkr (PC8214) Tripped (Note 3) 110 Turbine Tripped-Gen Loss of Field (140X) 111 Steam Line High Flow A 112 Steam Line High Flow B 113 Steam Line High Flow C 114 Steam Line High Flow D 115 Reactor Low, Low Water Level A 116 Reactor Low, Low Water Level B 117 Reactor Low, Lov Water Level C 118 Reactor Low, Low Water Level D 119 Main Steam Line Low Pressure A 120 Main Steam Line Low Pressure B 121 Main Steam Line Low Pressure C 122 Main Steam Line Low Pressure D 123 Turbine Tripped-USS Trans 1B Diff (187SBX) j 124 Scram Disch Air HDR Press Low Channel B j
Notes:
1.
Points 1 - 45, Reactor Protection System Trips.
2.
Points 46 - 50, Power Range Monitoring.
3 Points 51 - 60, Start-up Range Monitoring.
4.
Points 61 - 99, RFP & Main Turbine Trips.
5.
Points 100 - 1QS, Main Turbine.
6.' Points 106
' Indicated points which also activate the first-out annunciator on Pnl 9-7.
7.
PCB 214 becomes PCB 224 and PCB 234 for units 2 and 3 respectively.
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Table III 1
Instrument Monitored i
Identification Parameter FR/PdR-68-50 Total Core Flow / Core Press Prop FR-68-5 Recire Pump A & B Discharge Flow IRM,'APRM (4 channels - 2 charts)
LR-3-53/FR-3-78 Rx Vessel Level /RFW Flow PR-3-53/FR-46-5 Rx Vessel Press / Total Steam Flow FR-1-81/PR-3-59 Turbine Steam Flow /Rx Press FR-2-29 Condensate Flow From Pump Discharge Tit-1-1 Tail Pipe Temperature Computer postmortem Log XR-57-57 PRS Generator Load XR-47-16 Main Turbine Speed / Valve Position TR-47-164 Turbine Bearing Metal Temperature Sequential Events Tape TR-68-2 Recire Pump A & B Temperature FR-71-36 HPCI & RCIC Flow OD7-OP2 Computer scan for 02 scrams (control rod position) 1 l
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i 2.1 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (REACTOR TRIP SYSTEM COMPONENTS)
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Position Licensees and applicants shall confirm that all components whose function-ing is required to trip the reactor are identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, including maintenance, work orders, and parts replacement. In addition, for these components, licensees and applicants shall establish, implement and maintain a continuing program to ensure that vendor information is complete, current and controlled throughout the life of the plant, and appropriately referenced or incorporated in plant instructions and procedures. Vendors of these components should be contacted and an inter-face established.
Where vendors can not be identified, have gone out of business, or will not supply the information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement,.and repair, to compensate for the lack of vendor backup, to assure reactor trip system reliability.
The vendor interface program shall include periodic communication with vendors to assure that all applicable information has been received.
The program should use a system of positive feedback with vendors for mailings containing technical information.
This could be accomplished by licensee acknowledgement for receipt of technical mailings.
The program shall also define the interface and division of responsibilities among the licensees and the nuclear and nonnuclear divisions of their vendors that provide service un reactor trip system components to assure that requisite control of and applicable instructions for maintenance work are provided.
Response
Presently TVA's Division of Nuclear Power (NUC PR), identifies all components whose functioning is required to trip the reactor as safety-related.
These components which include the reactor protection system, the solid state protection system, and all other components whose function is defined as safety-related are now outlined in TVA's Operational Quality Assurance manual as critical systems, structures, or components (CSSC) which is a corporate document. Each individual plant has incorporated the applicable portions of this document into their prccedures. In addition, TVA's corporate procedures require all maintenance or modification activities to be documented prior to l
performing the work. This documentation is then reviewed by the appropriate plant organizations to ensure that it is properly identi-
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fled as CSSC or non-CSSC ar.1 to ensure that the applicable procedures and quality requirements for the idenitified work will be adhered to.
Furthermore, NUC PR requires that all procurement documents be identi.
t ficd as pertaining to CSSC or non-CSSC equipment. These procurement documents are reviewed by plant organizations or division central office organizations (depending on their point of origination) to ensure they are properly identified and contain the appropriate and l
required quality controls and specifications. Depending on the i
quality grouping, as outlined in division procedures that the procure-ment documents come under, many of them are also reviewed by other division central office organizations to further ensure that they meet all requirements.
e In view of the present division and plant procedures pertaining to safety-related equipment identification, and information handling systems used to control safety-related activities, we believe TVA is in compliance with the NRC staffs position.
TVA's NUC prs vendor interface program is presently centered around the division's operating experience review (OER) program which was developed to ensure that vendor and other related information would be handled from a systematic approach to continually inform the plants and other cognizant organizations of revisions, modifications, or deficiencies in plant equipment or procedures. The vendor interface program hinges around the original nuclear steam supply system (NSSS) supplier who supplied all reactor trip system components.
Any infor-mation supplied by the NSSS vendor to the division corporate office is acknowledged upon receipt by the division and forwarded to the OER organization and entered in the system. This information is then forwarded to the cognizant organizations and plants for review, comments, recommendation, or incorporation into plant activities.
The review, comments, or recommendations are documented and returned to the operating experience review group (OERG)'normally within 30 days as presently required by the division procedures.
Any recom-mendations are forwarded to the plant for incorporation in plant activities or resolution.
This information is tracked and documented by the OERG during the entire process until it has been incorporated or resolved.
This documentation is then stored for the life of the plant for further reference.
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s' Conclusion As previously stated, we believe that TVA's programs properly identify the reactor trip system and related components as safety related.
We also believe that TVA adequately controls activities such as maintenance, modi-fiention, and procurement on reactor trip system components.
In addition, we belic?e that TVA's operating experience review program has established a comprehensive vendor interface Trogram and ensures that vendor activities are reviewed and incorporcted as necessary for the reactor trip system.
In conclusion, we believe TVA's program is in compliance with NRC position and recommendations as stated in Generic Letter 83-28 M.
t 2.2 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (PROGRAMS FOR ALL SAFETY-RELATED COMPONENTS)
Position Licensees and applicants shall submit, for staff review, a description of their programs for safety-related equipment classification and vendor interface as described below:
2.2.1 For equipment classification, licensees and applicants shall describe their program for ensuring that all components of safety-related systems necessary for accomplishing required safety functions are identified as safety-related on documents, procedures, and information including maintenance, work orders, and replacement parts. This cescription shall include:
2.2.1.1 The criteria for identifying components as safety-related within systems currently classified as safety-related. This shall not be interpreted to require changes in safety classification at the systems level.
Response
During design and construction, the equipment classifications were identified in various design output documents such as drawings and construction project specifications.
This classification was to be identified if the items fell under the requirements of a quality assurance program, not necessarily if it was safety-related.
NUC PR expanded this concept by establishing a Critical Structures Systems and Components (CSSC) list for each nuclear plant. The CSSC defines the scope of applicability of TVA's QA program on operating plants. All activities that could affect CSSC equipment are performed in accordance with QA program requirements.
The present criteria, which are a part of our Operational Quality Assurance Manual (00AM), are used for inclusion of items on the CSSC list are as follows.
I.
General Criteria A.
Those items that are necessary to ensure 1.
{hegntggrityofthereactorcoolantpressure 2.
The capability to shut down the reactor and maintain it in a safe condition 3
The capability to prevent or mitigate the consequences of an incident which could result in potential offsite exposures comparable to those specified in 10 CFR Part 100
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Those items which the CSSC Subcommitte consider should receive the same level of quality assurance coverage as those listed in the general criteria above.
II.
Specific Guidelines for Inclusion of Items on the CSSC List Specific systems, structures, or components should be added to the CSSC list if they perform any of the following safety-related functions.
A.
Maintains core reactivity control under emergency conditions including those covered by anticipated transients without scram (scram mechanisms).
B.
Instruments and controls which are essential for emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal or are otherwise required for preventing significant release of radioactive material to the environment. Instrumentation and controls that perform an essential secondary function shall be considered safety-related if they are designed primarily to accomplish one of the above functions or where their failure would prevent accomplishing one of the above functions.
This includes those instruments and controls that are designed as safety-related and:
1.
Automatically keep the reactor operating within safe region by shutting down the reactor whenever the limits of the region are approached (reactor trip signal instrumentation) 2.
Initiate actuation of one or more of the engineered safety features in order to prevent or mitigate damage to the core and water coolant system components and ensure containment integrity'(engineered safety features activation system instrumentation) 3 Provide protective interlocks to prevent an U-operator error which could lead to incidents or events representing limiting plant design cases (permissive and interlock circuits).
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4.
Indicators and recorders and associated channels which are essential to:
a.
Perform manual safety functions and to perform postaccident monitoring following a reactor trip due to any condition up to and including the design limiting fault (containment pressure indicators).
b.
Maintain the plant in a hot shutdown condition or to proceed to a cold shutdown condition while meeting the limits of the plant's technical specification (system pressure monitor).
c.
Monitor conditions in the reactor core, reactor coolant systems, mainsteam and feedwater systems and containment (auxiliary feedwater flow monitor).
C.
Provides a barrier for containing reactor coolant within the reactor coolant pressure boundary (reactor coolant piping, valves, and fittings).
D.
Cools the reactor core under emergency conditions (residual core heat removal systems).
E.
Maintains fuel clad integrity (fuel clad, core power monitoring systems).
F.
Provides power, control, logic, indication, and protection to systems or components to enable them to accomplish their safety function (diesel generators, vital ac and de power).
G.
Supports or houses equipment that performs a safety function or protects that safety-related equipment from potential natural phenomena, equipment failure, and manmade hazards (seismic class I containment and structures, fire protection systems).
H.
Maintains specified environment (e.g., temperature, pressurg humidity, radiation) as required in vital areas to maintain equipment operability and personnel M.
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access (control room habitability systems).
I.
Supplies cooling water for the purpose of heat removal from the systems and components which provide a safety function (essential component cooling and service water systems).
. J.
Contains radioactive waste such that its failure could result in the release of radioactive waste to the offsite environments in violation of criteria A.3 (low-level radioactive waste discharge isolation valves).
K.
Controls fuel storage to prevent inadvertent criticality (fuel storage racks).
L.
Ensures adequate cooling for irradiated fuel in spent fuel storage (spent fuel cooling system).
M.
Minimizes the probability of dropping objects on stored fuel (overhead crane).
N.
Maintains primary containment as required by the FSAR to meet General Design Criteria 54, 55, 56, and 57 (containment penetrations and associated isolation and boundary valves).
O.
Doors and hatches which serve one or more of the following functions for safety-related equipment and areas:
(1) pressure confinement, (2) leakage confinement, (3) missile protection, (4) pipe whip and jet impinge =ent barrier, (5) equipment rupture flood protection, (6) natural flood protection, or (7) fire protection The items in parentheses are examples of items which would be considered as applicable to the listed guidelines and therefore eligible for inclusion on the CSSC list. These guidslines are continually reviewed and updated by the CSSC Review Committee to include changes j
in NRC requirements and plant design and safety criteria as they occur.
III. The CSSC list is supplemented by TVA EN DES identified Class lE equip =ent and requirements.
2.2.1.2 A description of the information handling system used to identify safety-related components (e.g., computerized equipment list) and the methods used for its development and validation.
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. Response The overall development and maintenance of the CSSC list is the responsibility of the CSSC committee. The CSSC committee is a review group comprised of multidisciplined nuclear-experienced engineers and quality assurance representa tives. The various technical branches under the oversight of the CSSC committee developed the initial CSSC list and evaluate all changes which are reviewed and approved by the CSSC committee.
The CSSC list is issued and controlled manually as part of the OQAM.
2.2.1 3 A description of the process by which station personnel use this information handling system to determine that an activity is. safety-related and what procedures for maintenance, surveillance, parts replacement, and other activities defined in the introduction of 10 CFR 50, Appendix B, apply to safety-related components.
Response
Plant activities that could affect equipment on the CSSC list are prescribed by instructions appropriate to the circumstances.
These instructions are prepared, reviewed, and approved in accordance with section 6.0 of the plant's technical specifications and the plant QA program.
2.2.1.4 A description of the management controls utilized to verify that the procedures for preparation, validation and routine utilization of the information handling system have been followed.
Response
After licensing, the inplant Quality Engineering Section routinely and independently verifies that the plant instructions appropriately utilize the CSSC list and meet the plant's quality assurance requirements.
The Office of Quality Assurance performs audits of the central office activities and plant activities to verify that the QA requirements are met.
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2.2.1.5 A demonstration that appropriate design verification and qualification testing is specified for procurement of safety-related components. The specifications shall include qualification testing for expected safety service conditions und provide support for the licensees' receipt of testing documentation to 1
support the limits of life recommended by the supplier.
Response
Predefined specification for various components and materials have been prepared by varioua technical branches for items such as ASME code valve parts, pump parts and materials, and class lE equipment. Also, when original specifications cannot be verified, the technical branches prepare specifications that are used in the procurement process. In addition, TVA prepares and utilizes substitution guides for standardized industry items such as bearings, V belts, capacitors and resistors.
The quality assurance program requires that for items that have storage and shelf life, the vendor furnish such information.
The quality assurance program requires that the original design specification or the TVA originated specifications, supplemented by class lE requirements, are used in procurement of CSSC components. All CSSC procurements are reviewed independently by a quality assurance or quality engineering group.
All items are receipt inspected to ensure that the required contract documentation and requirements are met.
2.2.1.6 Licensees and applicants need only to submit for staff review the equipment classification program for safety-related components. Although not required to be submitted for staff review, your equipment classification program should also include the broader class of structures, systems, and components important to safety required by GDC-1 (defined in 10 CFR Part 50, Appendix A, " General Design Criteria, Introduction").
Response--None Required 5
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-1 2.
For vendor interface, licensees and applicants shall establish, implement and maintain a continuing program to ensure that vendor information for safety-related components is complete, current and controlled throughout the life of their plants, and appropriately referenced or incorporated in plant instructions and procedures.,
Vendors of safety-related equipment should be contacted and an interface establi shed. Where vendors cannot be identified, have gone out of j
business, or will not supply information, the licensee or applicant '
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shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, to assure reliability commensurate with its safety function (GDC-1).
The program shall be closely coupled with action 2.2.1 above (equipment qualification). The program shall include periodic communication with vendors to assure that all applicable information has been received.
The program should use a system of positive feedback with vendors for mailings containing technical' information. This could be accomplished by licensee i
acknowledgment for receipt of technical mailings.
It shall also define l
the interface and division of responsibilities among the licensee and the nuclear and nonnuclear divisions of their vendors that provide service on safety-related equipment to assure that requisite control of and applicable instructions for maintenance work on safety-related equipnent are provided.
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Response
TVA is actively participating in the NUTAC associated with NBC Generic letter 83-28, section 2.2.2.
The results of NUTAC are expected to be available for approval during February 1984. Upon receipt of the NUTAC recommendations, TVA will evaluate and provide a plan for implementation.
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3.1 Post-Maintenance Testing (Reactor Trip System Components)
Position The following actions are applicable to post-maintenance testing.
4 1.
Licensees and applicants shall submit the results of their review of test and maintenance procedures and technical specifications to assure that post-maintenance operability testing of safety-related components in the reactor trip system is required to be con 4ucted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.
Response
The Division of Nuclear Power (NUC PR) Operations Quality Assurance Manual (0QAM) requires that the plant's maintenance instructions contain instructions which will ensure the following.
Maintenance Requests (MRs) shall be reviewed by the responsible supervisor or his designated representative and the plant's Quality Engineering (QE) staff before maintenance can be performed on any critical system, structure, or component (CSSC) equipment.
After maintenance is completed on any items on the CSSC list and before the item is returned to service, testing shall be performed to verify its operability.
The following are the guidelines and instructions to ensure the above is accomplished.
Technical specifications for Browns Ferry list the limiting conditions and surveillance requirements for unit operability for the reactor protection system. Browns Ferry's Standard Practices require that post-maintenance testing be performed. Standard Practice BF13.1,
" Format for Maintenance Instructions," establishes the format for maintenance instructions and provides for post-maintenance testing of affected components to establish operability. Standard Practice BF6.1, addresses the required post-maintenance testing of affected components for operability. BF6.1 is the governing document for performing maintenance and is supported by the document listed in the referenced section. Standard Practice BF7.5, " Maintenance Request and Tracking," establishes the instructions and responsibility for identifying, requesting, and authorizing maintenance. BF7 5 requires that all plant personnel report the need for corrective maintenance on E-plant equipment or systems by the use of HR form TVA 6436.
Identification of the problem, corrective measures to be taken, and any post-maintenance testing required shall be documented on the MR.
BF7.6 identifies those responsible for completing the section of the MR addressing post-maintenance testing and how to address this item.
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. BF7.6 requires that any pre / post-maintenance testing or QC holdpoints required be identified on the MR form. MRs for CSSC/ safety-related equipment are reviewed and approved by the QE before performance of work unless the work is scheduled preventative maintenance using previously approved instructions. Following completion of the work, the responsible maintenance section representative shall review the MR to determine that the work was acceptable. This representative shall verify that all post-maintenance test requirements were met. The MR is then reviewed by the field QS section to determine completeness as QA records.
2.
Licensees and applicants shall submit the results of their check of vendor and engineering reccamendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the technical specifications, where required.
Response
The Division Procedure Manual (DPM) N72A39 (Area Plan PM-0600) implements a program to mest the requirements of NUREG-0737 which deals with the review of operating experience reports (OERs).
It establishes a system to ensure the review of OERs, to document their applicability to TVA plants, to provide required written responses, and to ensure proper disposition of all applicable items. The Reactor Engineering Branch (REB) is the responsible organization in the Division of Nuclear Power for coordinating this review, disseminating the information to the appropriate organizations that could be affected by this experience, and ensuring that the recommendations.
made by the reviewing organizations are addressed by the plant. A review of General Electric Company's (GE) Service Instruction Letters (SIL) has been conducted and it had been determined that the plant has been sent all appliable SILs with recommended action when required.
BF2.14 " Review of Plant Instructions," implements the periodic review of plant procedures and instructions. BF21.17 " Review, Reporting, and
' Feedback of Operating Experience Items," implements at the plant level the TVA program for review of OERs within the nuclear industry.
3 Licensees and applicants shall identify, if applicable, any post-maintenance requirements in existing technical specifi-cations which can be demonstrated to degrade rather than enhance safety. Appropriate changes to these test requirements, with supporting justification, shall be submitted for staff approval.
(Note that action 4.5 discusses online system functional testing.)
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Response
We are not aware.of any program currently being done by GE specifically to determine if current testing requirements degrade the j
reliability of.the equipment. We do not have any information that indicates any existing required testing degrades rather than enhances safety.
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32 Post-Maintenance Testing ( All Other Safety-Related Components)
Action 1.
Licensees and applicants shall submit a report documenting the extending of test and maintenance procedures and technical specifications review to ensure that post-maintenance operability testing of all safety-related equipment is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.
Response
The Division of Nuclear Power Operational Quality Assurance Manual (0QAM) requires that maintenance instructions shall contain measures to cover the following.
"Upon completion of maintenance on any item of the CSSC list and before release for service, appropriate testing shall be performed to verify operational acceptability.
Functional tests or industrial standard tests may be used for this purpose."
The OQAM also requires review of the maintenance request (MR) by the responsible cection and the Field Quality Engineering (CE) Section before performance of maintenance on CSSC equipment.
Standardized guidelines which include the following are provided for preparation / review of Fms.
1.
Specify appropriate post-maintenance testing and, where applicable, reference the proper plant instruction.
2.
Consider compliance with plant technical specifications.
Specifically:
a.
Will removal of equipment from service for this maintenance violate any limiting conditions for operations?
b.
Are adequate post-maintenance tests (sis) specified to ensure the equipment's readiness for operation?
3 Provide for return of equipment to normal status as required.
The MR requires that the section responsbile for the performance of the post-maintenance test and also the operations section shall sign E.
to concur that the post-maintenance test was performed and the equipment is ready for return to service.
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i Based upon our review, the NUC PR program does require post-i maintenance testing to demonstrate operability before safety-related components are returned to service.
These requirements are implemented at each plant through plant specific instructions.
j Browns Ferry Nuclear Plant i
Standard Practice BF6.1, " Performance of Maintenance," is the governing document for maintenance activities. This standard practice l
is supported by those instructions listed in the reference section of j
the procedure. Standard Practice BF13.1, " Format for Maintenance Instructions," provides that the testing section of'the maintenance j
procedure address the required post-maintenance testing of affected j
components for operability. Standard Practice BF7 6, " Maintenance
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Request and Tracking," identifies those responsible for completing the portion of the MR addressing post-maintenance testing, provides instructions on how to address this item, and provides guidance for the QA review of the MR before the performance of the work.
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i Action l
2.
Licensees and applicants shall submit the results of their check of vendor and engineering recommendations to ensure that any I
appropriate test guidance is included in the test and maintenance
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procedures or the technical specifications where required.
t Resconse I
TVA's philosophy has always been to utilize engineering judgment, operating experience, TVA policy, and industry experience in conjunction with vendor and engineering recommendations to ensure that I
any appropriate test guidance is included in the test and maintenance procedures or the technical specifications where required.
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This is supplemented by a program dealing with the review of operating l
experience reports. This program establishes a system to ensure the review of operating experience reports to document their applicability to TVA plants, to provide required written responses, and to ensure proper disposition of all applicable items.
Also, in order to comply with IE Bulletin 79-01B and NUREG-0588, clas's 1E electrical equipment is being reviewed for applicable maintenance instructions required to maintain the environmental qualification of the equipment.
This activity will be completed in accordance with the NRC ruling on environmental qualification.
In addition to the above, periodic review of procedures and i
l instructions is required by the OQAM to determine if changes are necessary or desirable. This review is conducted no less frequently
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than every two years by an individual knowledgeable in the area affected by the procedure / instruction.
o It is TVA's opinion that the above programs and philosophy provide sufficient checks and balances to provide reasonable assurance that vendor and engineering recommendations are incorporated as appropriate.
Specific information for Browns Ferry is given below.
Browns Ferry fluelear Plant The above philosophy has been utilized in establishing maintenance and test programs. Standard Practice BF2.14
" Review of Plant Instr.uctions," implements the periodic review of procedure and instructions.
Standard Practice BF21.17, " Review, Reporting, and Feedback of Operating Experience Items," implements at the plant level the TVA program for review of operating experience reports within the nuclear industry.
Action 3
Licensees and applicants shall identify, if applicable, any post-maintenance test requirements in existing technical specifications which are perceived to degrade rather than enhance safety.
Appropriate changes to these test requirements, with supporting justification, shall be submitted for staff approval.
Resoonse - All Plants It is not TVA's philosophy to propose changes in existing technical specifications which are perceived to degrade rather than enhance safety. If items are identified in the existing specifications which are perceived to degrade safety, appropriate changes will be submitted along with supporting justi#ication.
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O Browns Ferry Nuclear Plant 4.5 Reactor Trip System Reliability (System Functional Testing)
Position Online functional testing of the reactor trip system, including independent testing of the diverse trip features, nhall he performed on all plants.
Action 1.
The diverse trip features to.be tested include the breaker uv and shunt trip features on Westinghouse, B&W (see action 4.3 above) and Combustion Engineering, Inc., plants; the circuitry used for power interruption with the silicon controlled rectifiers on B&W plants (see action 4.4 above); and the scram pilot valve and backup scram valves (including all initiating circuitry) on GE plants.
2.
Plants not currently designed to permit periodic online testing shall justify not making modifications to permit such testing.
Alternatives to online testing proposed by licensees will be considered where special circumstances exist and where the objective of high reliability can be met in another way.
3 Existing intervals for online functional testing required by Technical Specifications shall be reviewed to determine that the intervals are consistent with achieving high reactor trip system availability when accounting for considerations such as:
1.
uncertainties in component failure rates 2.
uncertainty in common mode failure rates 3
reduced redundancy during testing 4.
operator errors during testing 5.
component " wear-out" caused by the testing Licensees currently not performing periodic online testing shall determine appropriate test intervals as described above. Changes to existing required intervals for online testing as well as the intervals to be determined by licensees currently not performing online testing shall be justified by information on the sensitiv-ity of reacter trip system availability to parameters such as the test intervals, component failure rates, and common _ failure rates..
g.,
Response to Action Items 1, 2, and 3 TVA is evaluating its test program and feasibility of modifications to i
allow online testing of backup scram valves. We will update the NRC on the status of this study by February 29, 1984.
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