ML20078R714
| ML20078R714 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/08/1983 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-09, REF-GTECI-SY, TASK-A-09, TASK-A-9, TASK-OR 5211-83-330, GL-83-28, NUDOCS 8311150266 | |
| Download: ML20078R714 (36) | |
Text
_
GPU Nuclear Corporation U Nuclear
- rges:48o 8
Middletown, Pennsylvania 17057 717 944-7621 TELEX 84-2386 Writer"s Direct Dial Number:
November 8, 1983 5211-83-330 Office of Nuclear Reactor Regulation Attn:
D. G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Sir:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Required Actions Based on Generic Implications of Salem /ATWS Events Ref:
1.
GPUN Letter - H. D. Hukill to D. G. Eisenhut, dated 9/8/83 2.
GPUN Letter - H. D. Hukill to D. G. Eisenhet, dated 10/7/83 This letter and enclosed documents our response to Generic Letter 83-28 which requested the status of TMI-l conformance as well as our plans and schedules for any needed improvements to conform with the positions of the Generic Letter.
As you know, the B&W reactor trip system desiga used at TMI-1 is less vulnerable to the type of failure that occured in the Salem Event. However, since the reactor trip system is fundamental to reactor safety, we agree that special care and attention should be given to assure its reliability. To this end we have been working within our compnay, as well as with external organi-zations, to respond to the concerns raised by the Salem Event.
In March 1983, shortly after the Salem Event, we established an internal Task Force to make an assessment of the information and lessons learned from the various ongoing reviews of the Salem incident. This assessment included the areas listed below:
Quality Classification Operation Equipment & Maintenance Data
- Testing Maintenance
- Trending Training 8311150266 831108 for/
PDR ADOCK 05000289 P
- l GPU Nuclear Corporation is a subsidiary of the General Nblic Utilities Corporation
Mr. D. G. Eisenhut 5211-83-330 The Task Force's charter also included instructions to make recommenda-tions regarding areas where improvements were needed. Furthermore, the Task
-Force was to act as a review body to periodically examine and determine the effectiveness of these improvements.
Upon receipt of the Generic Letter we began to work more closely with the B&W Owners Group ATWS Committee.
This Committee was first fomed in February 1979 in response to NUREG 0460, Vol. 3.
Since that time and until the Salem event, the ATWS Committee was an active participant in industry groups providing constructive comments on the proposed ATWS rule.
The B&WOG ATWS Committee's.first involvement in the Salem event was the RRG/m C meeting in March 1983. The Committee presented the effects of an ATWS event on the B&W NSSS and demonstrated that the B&W design is less vulnerable to a Salem event than other NSSS designs. This continued interaction has progressed to the point where during the NRC/0G meeting of August 26, 1983, the E C Staff conditionally accepted the B&WOG concept for a diverse SCRAM System.
In April 1983, the ATWS Committee assumed the responsibility for the B&WOG efforts on Salem and was charged with the overall coordination of Salem issues with the various B&WOG committees. Since that time, the ATWS Committee has made presentations and met with the NRC Staff in April and May 1983 and on June 30 and August 26, 1983. Verbal contact with the EC has been made by the Comittee on at least a weekly basis in aodition to the contacts of Jim Taylor (B&W) and the B&WOG Steering and Executive Committees. The Committee has attempted to define generic efforts that can be undertaken on each of the Generic Letter (83-28) items. This generic program is consistent with direc-tions from various NRC Staff members who have strongly encouraged a generic approach to resolve the concerns in the Generic Letter. This generic program is further described in the B&W Owners Group Program submitted to the NRC on 11/5/83 by Mr. J. T. Enos of Arkansas Power & Light, Chairman of the B&WOG ATWS Committee.
In addition to the above activities, GPUN is an active member of NUTAC formed to address Section 2.2.2 of the Generic Letter. The formation of the NUTAC was conveyed to the EC by letter dated September 15, 1983 from E. P.
Griffin, NUTAC ChaiIman.
The mission of the NUTAC is to develop a workable approach to vendor interface problems that will contribute to the safety and reliability of nuclear plants. Although participation in the NUTAC will result in a submittal of an industry program plan (Ref. 2) in May 1984, the current status of GPUN's conformance with the position in Section 2.2.2 is discussed in the enclosure.
Based on our overall evaluation of GPUN's programs, we have determined that TMI-l currently complies in many important respects with the requirements of Generic Letter 83-28.
The enclosure provides our detailed response to each
l Mr. D.'G. Eisenhut.' 5211-83-330 of the items in the Generic Letter. Where we have identified the few areas that warrant improvement, we have outlined programs to attain compliance. We have also indicated those areas participate in generic activities in either the B&W Owners Group or NUTAC.
Sincerely, C
V H.
. Hukill Vice President - TMI-1 cc:
J. Van Vliet R. Conte Sworn and Subscribed to Before me this 8th day of November, 1983.
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6 Enclosure GPUN RESPCNSE TO GEPERIC LETTER 83-28 INTRODUCTION The following describes the objectives and approach, used in formulating a response to Generic Letter 83-28. For each item of the Generic Letter, this report details the specific GPUN actions taken to determine conformance.
The report also identifies the appropriate organization (B&W Owners Group, NUTAC or GPUN)- that will address plans and schedules for any needed improvements for conformance with the NRC positions.
o Objectives The principal objective of this program was to develop a response to Generic Letter 83-28 that contains the following:
Current status of compliance with the regulatory positions contained in the Generic Letter.
Plans and schedules for improvements identified as needed to improve plant safety and operability.
Responses in this enclosure show either full compliance with NRC positions or reasonable progress toward meeting the NRC positions to allow-safe operations while upgrades to the existing procedures are made, o
Approach In assessing the response to post-trip review, equipment classification /
vendor interface and post-maintenance testing, GPUN reviewed current operating, administrative and quality assurance procedures satisfied the intent of NRC positions. Planning for improvements in each of these areas considered the degree of compliance, ongoing programs and recommendations of the internal Task Force.
This overall program relies on the B&W Owners Group and NUTAC activities as much as possible for guidance. The B&W Owners Group activities are discussed in detail in the B&WOG program submitted to the NRC on 11/5/83
.by J. T. Enos of Arkansas Power.& Light, Chairman of the B&WOG ATWS Committee.
PU Nuclear Response to the Items of Generic Letter 83-28 1.1 Post Trip Review (Program Description and Procedures)
This report describes the reactor trip review program for GPUNC's Three Mile Island Unit 1 Nuclear Station. The report describes the program's essential elements and basic requirements, with specific details to be incorporated into THI-1 plant procedures and Technical Functions (TF) procedure EP-029,
" Analysis of PlN Plant Transients." This report addresses items 1.1 and 1.2 identified in NRC Generic Letter 83-28.
The status of actions required to implement this program and a schedule for completion of required actions is provided in the implementation section of this report.
Purpose The purpose of the program is to establish a systematic method of conducting the technical review and analyses of TMI-l plant performance associated with reactor trips in order to:
- 1. Determine the immediate and root cause(s) of the trip.
- 2. Identify unexpected, abnormal responses to the trip by plant systems, equipment, and personnel.
- 3. Assess the impact of identified abnormalities on nuclear safety, equipment reliability, system performance, and availability.
- 4. Develop corrective actions to prevent the recurrence of the trip and mitigate abnormal responses.
- 5. Document observed plant behavior for use in subsequent evaluations.
- 6. Satisfy reporting requirements.
Scope The GPUN Reactor Trip Review Program implemented at TMI-1 applies to every reactor trip, planned and unplanned. However, planned reactor trips need not undergo all phases of the review if response is normal. The scope of the information reviewed under the program is sufficient to accomplish its objec-tives and includes data on plant system behavior, actuation and sequence of operation of equipment, records of operator actions, and plant activities affecting the event. The program prescribes activities that are performed immediately following a trip, prior to restart, and continue through a sub-sequent in-depth evaluation that supports preparation of internal and external reports. The program also outlines the criteria for determining the approval and concurrence levels for plant restart.
The TMI-l Reactor Trip review program is consistent with the B&W Owners Group (B&WOG) Position adopted by the B&WOG Transient Assessment Committee. This position is outlined in the B&WOG Response to Generic Letter 83-28.
Roles and Responsibilities Several groups, including plant and Technical Functions staff participate in the reactor trip review program. The responsibility and authority of each participant has been clearly defined below.
Plant Operations is responsible for operating the plant. Under the program, the Shift Supervisor (SS) is responsible for notifying plant management. The Shift Technical Advisor (STA) is the Technical Functions' contact with the operating plant. The STA-is responsible for notifying Technical Functions (TF) personnel per TF Engineering Standard ES-005, "STA Duties and Responsi-bilities.". The Operating Crew, along with the STA, are responsible for diag-nosing and controlling the event and thus will have firsthand knowledge of the event. This information is to be promptly documented to help ensure that a l
complete record of the event is obtained.
The Plant Analysis Section at the operating site is responsible for performing the post trip review. The Plant Analysis Section reports to the Director of Systems Engineering, and thus will provide an independent assessment of the
. plant's behavior and the acceptability of restarting. The post-trip review must be completed and documented prior to restart.
Various Technical Functions Organizations, especially the Safety Analysis and Engineering & Design groups, will provide analytical and technical support as required.
The Plant Engineering and Operations and Maintenance Departments will deter-mine the root cause(s) of the event, specify corrective actions, and implement corrective actions.
Plant Management is responsible for determining when and how the unit is to be restarted.
Technical Functions is responsible for concurring with Restart Plans and cor-rective actions.
In addition, the Independent On-Site Review Group may conduct supplementary evaluations.
Training and Qualification Training and qualification for key responsible personnel is specified below:
Shift Supervisor - The SS holds a Senior Reactor Operator License and is
-trained through the Operator Requalification Training Program described in FSAR Update Section 12.2.4.2.
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-h.-PlantReview'Grotp'-PRGmembers_arequalifiedtotherequirementsofTMI-1 W
-procedure AP-1034, " Plant Review Group," which includes requirements of
. ANSI /ANS 3.1 Section-4. +1r a
a!
N TMI-l Division Personnel - TMI-l Division personnel (such as Operations &
s,
- '1.
Maintenance and Plant Engineering personnel)-participating in the reactor trip i VP review process will be qualified to the' specific position description.
c Shift Technical Advisor - The STA is qualifiec to TF procedure. TAP-005, "STA Selection & Qualification," and trained to-the. requirements of TMI-1-Training Department Program document, " Shift Technical Advisor Training Program."
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.Specifically. included is training in' safety analysis and transient and.
I' y;
- accident analysis.
~
3,15nt Analysis Section TMI-1 -' Plant Analysis Section personnel are qualified
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to the specific position description, which includes experience in safety
. analysis and transient and accident analysis,c
-t
' Technical, Function Division Personnel - Technical. Functions Division personnel-participating in the review process will be' qualified to the specific position
' description.
1 s-kINad$,ition,manyofthe.abovearequalifiedandtr$1nedasResponsible
-t hTechnical, Reviewers,(RTRs)~andfor,IndependentSafetyReviewers(ISRs).
Program Phases
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.The reactor trip review program consists of fourEdistinct phases:-
7
- 1. - Post-trip review l
-2. ' Restart decision
- 3.. Independent review 4.
Subsequent evaluation i
E'very reactor trip will be subjected to a post-trip review and restart deci-F sion. Planned react'or trips, where no abnormalities have been identified, need not proceed to the subsequent evaluation phase. The major elements of e
.each of these phases is described below.
l, l6 t
- 1; Post-Trip Review 1
--, - t
,f The post-trip review is performed immediately following the trip and completed.
J prior to restart.-
The purpose of the post-trip review is to:-
1.
Determine the cause(s) of the trip.
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. Identifyother-than-ehetedpe'rfor'manceofplantsystemsandequipment.
2.
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3.
Assess the impact of identified abnormal performance on safe operation.
4.
Ensure continued availability of information and data pertaining to the event.
The scope of the post-trip review has been established to ensure that abnormal performance in important systems will be identified. Guidelines and criteria, which define the range of e.npected system response, are used in the process.
The major elements of the post-trip review, and the responsible lead organi-zation are:
1.
Operations & Maintenance will determine the cause(s) of the trip.
2.
Plant Analysis will determine the reactor trip sequence.
j 1
3.
Plant Analysis will review the pre-and post-trip behavior of key j
parameters that reflect overc11 plant performance and will identify abnormal performance of impor*. ant systems.
4.
Plant Analysis will review the performance lof important systems and equipment, both safety and control,' to identify other-thanexpected response to the trip.
f 5.
The Plant Staff and Technical Functions will ioentify corrective actions that must be completed prior to restart.
Pre-determined checklists will be used to conduct the review specified by items 2, 3, and 4 above. The sources of information necessary to conduct the review and analysis are detailed in Item 1.2.
- 2. Restart Decision Prior to restarting the unit, Operations and Technical Functions must ercure that:
1.
The cause(s) of the trip (RPS Trip function and initiating event) are
'l known or have been investigated to the fullest extent possible. This means that when the root cause is unknown, attempts have been made to locate and duplicate the cause through troubleshooting and testing and i
appropriate calibration and maintenance checks have been conducted.
2.
The plant's transient response was as expected for the type of event, and either did not identify any problems that impact the ability of the unit to be safely restarted and operated or that the problems have been corrected.
AnyproblemswithequipmentsubhecttoTechSpecLCOrequirementsare 3.
corrected as required.
4.
The corrective actions identified during the post-trip review as being required prior to restart, are implemented.
7 1
0
ThE decision to restart will te made by the TMI-l Management with additional review and concurrence based on criteria specified in GPUN procedures.
t
- 3. Independent Review s
Under certain conditions, further independent review must be performed prior to restart to ensure that all questions regarding the ability to safely restart and operate the plant are resolved. Criteria have been established as
.to when an independent review is required. They are as follows:
1.
If the immediate (RPS trip function) and root cause(s) of the trip cannot be determined, or 1
2.
Plant post-trip response is abnormal, or 3.
If any unresolved safety issues exists, or 4.
If compliance with licensing requirements is in question.
The independer.: review will be performed by a group of knowledgeable indivi-duals, such as RTRs or ISRs, designated by the Director - O&M and Director -
Systems Engineering. Results will be reported to the Director - O&M and Director - Systems Engineering.
- 4. Subsequent Evaluation Every unplanned reactor trip'will be subjected to a follow-up, in depth evaluation by Plant Analysis or TMI-1. In addition, planned reactor trips which show abnormalities in plant response will also receive further evaluation. The purpose of the subsequent evaluation is to ensure that all aspects of the events are fully investigated, evaluated, and documented.
The subsequent evaluation takes the knowledge. gained from the post-trip review and expands upon it in areas of identified abnormal response.
It ensures that the more subtle aspect of system performance, even though they did not signi-ficantly affect the plant response, are evaluated and needed corrective action identified.' This report need not be completed before restart. The scope of the subsequent evaluation is prescribed to ensure that all reporting require-
. ments can se met.
1 Implementation and Status The reactor trip. review program w'ill be implemented as described above at TMI-1 via plant procedure AP-1044, with supporting guidance contained in TF Procedure EP-029. Document Distribution and Control Center (DD&CC) will maintain records of all reviews and evaluations.
The above described program meats the requirements of Generic Letter 83-28 Action Items 1.1 and 1.2, and adheres to the B&W Owners Grotp Position adopted by the B&WOG Transient Assessment Committee.
The status of actions required to fully implement the TMI-1 Reactor Trip Review Program is included below. Where items require additional work to conplete implementation, a schedule is provided.
1)
Criteria for determining the acceptability of restart.
Action Required: The four elements discussed under " Restart Decision" and criteria for additional review and concurrence will be included into a THI-l procedure.
Completion Date: June 30, 1984 2)
Include responsibilities and authorities of personnel who will perform the post-trip review and analysis of the event.
Action Required: Program guidance described in " Roles and Responsibi-lities" section will be incorporated into a TMI-1 Procedure Similar guidance will be incorporated into TF Procedure EP-029 Completion Date: June 30, 1984 3)
Incorporate qualification and training requirements for responsible personnel.
Action Required: Qualification levels (e.g., SS, PRG member) will be incorporated into a plant procedure Plant Analysis incorporate qualification levels into TF EP-029.
Completion Date: June 30, 1984 Training Requirements are contained in Training Department Program Documents, " Senior Reactor Operator Training Program" and " Shift Technical Advisor Training Program" 4)
Describe sources of information necessary to conduct the review, analysis, and reconstruction of the event. See Item 1.2.
5)
Incorporate methods and criteria for comparing the event information with known or expected behavior.
Action Recpired: Add an Appendix to EP-029 with criteria for comparing the event information with known or expected behavior.
Completion Date: June 30, 1984 6)
Include criteria for determining the need for independent assessment.
Action Required: The 4 items from the " Independent Review" section above will be included into a plant procedure.
Completion Date: June 30, 1984
1.2 Post Trip Review (Data and Information Capability)
This section describes the data and information capability that will be used in support of the TMI-l Reactor Trip Review Program.
The TMI-l Plant Computer System (PCS) is a prime source of data used in support of the Reactor Trip Review Program. The PCS consist of a Bailey 855 and ModComp Classic Computer Systems. The Bailey system is fed from an inverter source with backup through a static inverter switch to an AC dis-tribution panel and is nonclass 1E. The ModComp system is fed from a dedi-cated AC distribution panel connected to an Engineered Safeguard Motor Control bcs which is fed from a station diese) generator. The ModComp system is nonclass lE. See Figure 1.
Capability for Assessing Sequence of Events Sequence of events monitoring (SEM) is provided by two separate computer systems consisting of a Bailey 855 computer and SEN Hardware and a ModComp Classic Computer and Computer Products, Inc. (CPI) SEM Hardware, such that both systems are redundant and may detect an on-off SEN state indication external to the comper s"" ems.
The Bailey system scans 128 contact states to detect the sequence in which changes of state occur. Changes of state occurring at least one millisecond apart are stored in exact sequence of occurrence.
As each occurrence is processed, a time tag is appended to the corresponding event. Storage is provided for the occurrence of 64 SEMs. The occurrence of an SEN will initiate an indication to operations by the flashing of an annun-clator light on the operators console where an operator may request a printout on a printer of the occurred SEMs.
The ModComp Classic system is basically the same as the Bailey with the excep-tion that the ModConp system is interrupt driven, and it is not necessary for operator intervention to retrieve SEM data. SEM's are automatically printed on a printer as they occur. Also an additional capability is provided to store on a history file up to eighteen SEM records where a record is defined as a series of consecutive sequence of events where a time interval of 30 seconds has not occurred between any two events since the beginning of the first event. These records are available for display on a color monitor for analysis.
The manufacture specifications for the Bailey system is a one millisecond between time tag events with a 4 millisecond resolution. The ModComp system is presently in the completion stage of installations and tests indicate a one millisecond resolution between time tag events with a 5 millisecond resolution.
Data is displayed in the following format:
A926:14:453 52032 Generator Ground TRIP i
Vital Power
- Cont. Rm.
- CPI Mod
- CRT Display:
- Multiplexer:
Comp
- Computer
- Printer
- Comp. Rm.
- Input
- Cont. Rm.
- Multiplexer:
855
- Recorder (Sel):
- Bailey
- Computer Printer
- Cont. Rm.
Non IE Pwr Batt. Backup COW UTER DISPLAY SYSTEM Fig. 1
Hard copies -in the form of printed output or video-copied output are available for all SEM events.
Typical parameters (128 total) monitored include:
R)S Channel Trip Status RPS RC High Pressure Trip RPS RC Low Pressure Trip RPS RC Press-Tep Trip RPS Overpower Trip RPS Power /Inbalance/ Flow Trip RPS Power / Pumps Trip RPS RC High Tmperature Trip RPS Reactor Building Pressure Trip RPS Turbine Tripped RPS FW Pumps Tripped RPS Power Supply Fault RPS Manual Trip ES Actuation - H'I ES Actuation - LPI ES Actuation - RB Isolation RC Pump Trip Status EFW System Auto Start Actuated FW Pump Trip Status (Selected)
Turbine Trip Status (Selected)
Generator Output Breaker Status (Selected)
Main Transformer Fault (Selected)
Auxiliary Transformer Fault (Selected)
Diesel Generator Output Breaker Status 4KV ES Bus Fault (Selected)
Capability for Assessing Analog Variable Time History Post Trip Review monitoring is provided by two separate computer systems as in the Sequence of Events.
The Bailey system provides a post-trip / memory review function which on trip indication will store analog data for 25 minutes before the trip and five minutes after the trip. Data is provided at 1 minute intervals for the first 20 minutes and at 15 second intervals for the 5 minutes before and the 5 minutes after the trip. Retrieval of the post trip / memory review is initiated upon operator request. A printed output format is produced.
Typical parameters (26 total) monitored include:
RPS Power Range Level Generator Megawatts RC Inlet Tenp (m)
RC Outlet Temp (m)
RC Loop Pressure (m)
RC Pressure (WR)
RC Pressurizer Level
.OTSG Full Range Level 1
l l
.0TSG Feedwater Flow i
OTSG Steam Pressure OTSG Steam Temperature FW Pump Speed ~
FW Pug Discharge Pressure Main Condenser Pressure (3rd stage)
>The ModCo m system provides a capability.for a post-trip review through a Transient. Monitoring. System.- The Transient Monitoring System (TMS) is a special purpose computerized data acquisition and display system. The system is designed.specifically for recording selected plant parameters on a con-tinuous basis at a relatively rapid rate and only retains data which is perti-nent to a normal variation. The normal variation is detected automatically at specified time intervals by' comparing selected plant parameters to a reference value. Data retrieval capabilities are provided for all the plant parameters (current data or previously stored data) in the form of graphical data or formatted printer listings. The capability to retrieve information ime-diately following plant transient aids in analyzing and possible isolating and resolving the' problems more expediently.
The Transient Monitor System (TMS) monitors 112 analog and 112 digital irputs on a continuous basis at approximately 0.5 second intervals. Selected para-meters in the following systems are monitored:
Reactor Protection System-l Integrated Control System Control Rod Drive System Nuclear Instrumentation System-Reactor Coolant System Non-Nuclear Instrumentation-Main Steam System
- Turbine-Generator System Turbine Bypass System Steam Generating System
.Feedwater System
. Intervals for display of data are operator-selectable from 0.5 seconds to several hours.
Othe'r Data and Information Sources
-Other data and'information sources used to assess the cause of unscheduled reactor shutdowns are listed below:
Control Panel Strip Charts
' Operator Logs
-Operator Interviews
. Computer Alarm Printouts Computer Utility Printer Output CRT Video-Copied Data
Implementation The Plant Computer system serves as a prime source for plant post trip evaluation. Plant Computer System parameters are selected to be consistent with B&WOG Transient Assessment Guidelines. Specific guidelines to conduct the post-trip review will be provided for cases where the computer system may be degraded or unavailable.
Action Required:-
Check list will be incorporated into TMI-l procedures to ensure assembly of complete data package Completion Date:
June 30, 1984
l 2.1 Equipment Classification and Vendor Interface (Reactor Trip System Components)
The Reactor Trip System components fall within the scope of those systems which are identified as Important To Safety (ITS) per GPUNC Technical Functions Standard ES-Oll (QCL). Specifically, these are the RPS and the trip portion of the CRD control system, respectively.
In addition we have reviewed (See B&WOG response) a listing of components identified by B&W as related to the Reactor Trip System.
This vendor listing is available on-site for review. Owners Group activities are now' underway to contact the vendors of the specific component in an attempt to obtain the latest applicable vendor technical information.
At THI-1, ES-Oll is recognized as the document which provides the guidance for quality classification of systems. Plant surveillance and maintenance procedures and work requests are quality classified on the basis of ES-Oll. Requisitions for replacement components are quality classified in accordance with Standard ES-011. Minor ancillary com-ponents associated with a reactor trip system such as structural hard-ware, nuts, bolts, conduit and junction boxes may be obtainable only as commercial grade. However, the purchase requisition would be identified as Important to Safety and receive the associated administrative controls.
Vendor information literature, which is received is reviewed by the appropriate engineering section where it is quality classified in accordance with ES-Oll prior to it being distributed for implementation.
A review of a variety of documents such as plant surveillance proce-dures, maintenance procedures and work requests associated with reactor trip components indicates that a quality classification of ITS or NSR (Nuclear Safety Related) is identified on each document. Either of these classifications places the document or component within the scope of the GPUNC Operational Quality Assurance Plan which is the governing document to assure compliance with 10CFR50 Appendix B.
With regard to the vendor interface program dealing specifically with the reactor trip system, GPU Nuclear is actively participating in the B&WOG program addressing this item. Details of this program / activity are discussed in the B&WOG submittal of 11/5/83.
In conclusion, it has been determined that we are actively approaching conformance in meeting Section 2.1 of NRC Generic Letter 83-28 by assuring that all our documents, procedures and information handling systems associated with reactor trip components are properly admini-stered.
2.2 LEquipment Classification and Vendor Interface (Programs for All-Safety-Related Components) i 2.2.1-The following describes the PUN program for ensuring that proper classification is made for all components and procedures'used to control safety related activities. This description includes details from the various procedures in affect as well as an overall view of the implementation of-the program.
l The criteria ~for identifying components as safety related within systems is described in TUN Tech. Functions Procedure EP-011 - Quality Classification List. The quality classification of systems, structures and major components are listed in Tech. Functions Standard ES-011 -
i Methodology and Content of GPLN Quality Classification List.
These procedures provide the means for maintaining the quality classification of the QCL. In addition, PUN Engineering & Design
-(E&D) maintains a published list identifying the E&D Section Manager and the responsible engineer for each primary system component or structure identified for TMI-1. The E&D uses a plant equipment i
identification. system that incorporates a numerical identification coupled with a verbal description of the referenced system, component or structure.
2.2.1.1 Our overall assessment for Item 2.2.1 of the Generic Letter is that we have procedures in place that meet the intent of the stated NRC positions. The following are the salient excerpts and, summaries from t
those procedures described above that provide the basis for GPUN's conformance to the Generic Letter:
l EP-Oll - Quality Classification List ES-011 - Methodology and Control of PUN Quality Classification List i
Purpose & Scope I
EP-Oll establishes the method for using the Quality Classification List-(QCL) to assign quality classifications.to PUN station structures, systems, components, and parts. It also assigns responsibilities for interpreting and maintaining ES-011.
The standard, ES-011 contains a QCL for TMI-l which specifies the quality classification of the major components, systems, and i
structures.. The_se classifications must be used directly by all GPUN personnel to specify the quality classification epplicable toactivities associated with these components, systems, and structures.
i l
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., -,. = - _. -. _ _ -. _ - - _ _.
ES-Oll also delineates the methodology to be used to develop the QCL. This methodology is used by engineers to determine the quality classification for a new minor item. It should be emphasized that use of the graded approach delineated in the.
Operational Quality Assurance Plan allows both the engineer and the QCL Interpreter to determine and specify only those attributes which are important to safety thus requiring QA coverage. The scope of QA coverage can thus be narrowed accordingly; however, to determine if a minor item can be assigned a lower quality classi-fication than that listed for the associated major item, a func-tional applicability approach is used by the QCL Interpreter. For example, the pressure boundary of a pump may be required to remain intact to achieve a function "Important to Safety," but the pump may not be required to operate to achieve this function.
In this case, the pump motor with its power supply and control circuits may be assigned a lower quality classification.
Important to Safety (ITS)--The "Important to Safety" class includes the following:
(1) Those items classified as " Nuclear Safety Related." To the extent that'this standard addresses " Nuclear Safety Related," it is in the same context as the previous classi-fication system and analagous to " Safety Grade" as used in the TMI-l FSAR and Technical Specification.
(2) Those items required (under all conditions) to achieve
-cold shutdown.
(The previous classification system was based upon achieving hot standby.)
QUALITY CLASSIFICATION SYSTEM IMPORTANT TO SAFETY
- (NOTE 1) :
NOT ITS PRESENT NUCLEAR SAFETY RELATED : ADD. ITEMS TO ADD. NRC : BENEFITS : NO SAFETY SYSTEM
- ACHIEVE COLD REQUIRE
- REACTOR : SIGNIFI
- SHUTDOWN MENTS
- SHUTDOWN : CANCE (NOTE 2)
OQA PLAN COVERAGE
- (NOTE 1) : NO QA r
Notes:
1.
Coverage under the OQA Plan is subject to engineering determination on a case-by-case basis for the specific activity being performed.
Activities to be covered by the OQA Plan are then specified.
2.
Additonal requirements are based upon:
Reg. Guide 1.143; 10 CFR 71, Appendix E; BTP ?.SB 9.5-1; Reg. Guide 1.29 and 10 CFR 73.55.
Benefits Reactor Shutdown (BRS): The " Benefits Reactor Shutdown" class consists of those structures, systems, and components which are normally use to perform the functions required to achieve cold shutdown. Normally implies only normal operations and does not include abnormal, emergency, casualty, or accident conditions. For example, the Atmoshperic Dump System is classified "Important to Safety." However, the Turbine Bypass System normally performs the heat rejection function and is therefore classified " Benefits Reactor Shutdown."
For items classified as " Benefits Reactor Shutdown", the usefulness of applying attributes of the OQA Plan to all or a portion of the activities affecting these items is evaluated on a case-by-case basis. To the extent determined by the Technical Functions Division to be appropriate, the activities will be identified as "Important to Safety" so that OQA Plan coverage is obtained.
Applicability This standard and procedure apply to the development and maintenance of the Quality Classification List for TMI-1. The classifications identified in the QCL shall be used to establish quality assurance coverage for all plant activities.
The methodology and-terminology used in the preparation of the Quality Classification List is to be employed as the basis for further classification at the subsystem, component, and part level.
In all cases, those classifications are documented but at the current time not computerized. Computerization is under study.
The Director of Plant Engineering maintains the master list of QCL Linterpretation decisions. This list is controlled to ensure that all additions or changes are available for reference in making subsequent
-interpretations.
The Director of Engineering & Design revises ES-011 and its appendices to maintain it current with plant configuration and all applicable regulatory requirements.
2.2.1.2 The description and methodology used to develop the QCL is contained in ES-011 which is available on-site for review.
2.2.1.3 ES-011 in conjunction with R.G. 133, Rev. 2 App. A provides the basis for determination of activities which are important to safety.
2.2.1.4 Management Controls for important to safety activities are described in the Operation Quality Assurance Plan for TMI-l (e.g., inspection, monitoring, special processes and auditing).
2.2.1.5 Tech Fmetions procedure TAP-ll " Purchase Requisitions" specifies the need to obtain design verification and qualification testing informa-tion for safety related equipment. EP-003 Vendor Document Review establishes methods for review, comment and acceptance if these vendor documents.
These procedures are available on-site for review.
2.2.1.6 See 2.2.1.1 for details.
While the above indicated status reflects current activities, the conclusion of the Salem Task Force was that GPUN needs a Component Parts Quality Classification List, which is easier to use than the present method of classification. Classification methods described above is time consuming and requires a repetitive evaluation of safety functions for TMI-l systems com-ponents.
Personnel who must deal with classification of component parts on a day to day basis need a simpler system. A separate task force is studying the problem and will make detailed recommendations by March 31, 1984.
Vendor Interface GPUN is an active participant in the industry NUTAC. GPUN will integrate recommendations and guidelines developed by the NUTAC into its internal programs (described below).
The overall GPUN program is comprised of several specific elements that integrate within and between the several GPUN Divisions, the comprehensive technical and administrative coverage of vendor information. These elements include approved procedures in place for the technical review and control of equipment manuals and vendor documents required to support plant operations.
These procedures include EP-021 - Technical Manuals and EP-003 - Vendor Document Review and EP-017 - Review of Operating Experience (ccurrently under revision).
These procedures set forth the requirements for technical review and updating of Technical Manuals prepared or acquired by G)UN. The procedures also establish methods for review, comment and acceptance of vendor documents.
The Engineering Data & Configuration Control (ED&CC) section numbers receives, records, distributes, coordinates technical reviews and approvals, and flies
all vendor documentation provided to GPUN. ED&CC also maintains a vendor Document Status Log and a Configuration Control List of these vendor docu-ments.
The site Technical Library and Document Distribution and Control
-Center (DD&CC) serves as a central repository for technical manuals and vendor information. A description of the procedure follows:
A vendor is defined as a firm which manufactures items at an offsite facility or supplies items to GPUN and operates under the requirements of their own quality assurance program.
Vendor Documents are normally equipment related technical documents such as drawings, procedures, reports, manuals, specifications, etc., which are prepared by a vendor to be used in the design, construction, maintenance, or operation of any GPUN stations, systems, or components. All documents received by TUN from a vendor shall have been approved by vendor in accor-dance with his internal procedures.
Documents such as drawings, procedures, reports, manuals, or specifications, pertaining to purchased items, which are provided gratis or as a part of a Contract of Purchase Order are processed under this procedure.
Requests for revisions to Vendor or GPUN Manuals shall be accomplished by Design Change Notice (DCN). Revisions received directly from Vendors shall follow the same process as original submittal.
The Engineering & Design Department of P UN will identify and list specific Three Mile Island Unit plant procedures for technical review by the respon-sible engineering discipline to ensure that vendor information is appro-priately referenced or incorporated. See our response for item 3.2.2.
The description for this work effort is as follows:
1.
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 Three Mile Island Unit 1 and appropriately referenced or incorporated in plant instructions and procedures.
2.
Initial tasks are to (1) conduct a technical review of all vendor manuals for safety-related components per EP-021, (2) conduct a technical review of plant procedures for safety-related components /
systems vs. vendor manuals. The specific categories of procedures for review are preventive maintenance, corrective maintenance and surveillance.
3.
Long term task is to provide an on-going program to assure continued accuracy and completeness of vendor information and plant procedures.
I-4.. Due to the extent and range of this review program, an initial estimate fo~
1 technical manual review will require approximately 2600 man '
Estimated completion date: January 31, 1987.
5.
Comparable e.gineering efforts are required to review plant maintenance, preventive mraintenance and surveillance proce'res.
Estimated completion date: January 31, 1990.
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3.1 Post Maintenance Testing (Reactor Trip System Components)
_3.1.1 GPUN identified seventeen (17) test and maintenance procedures and Tech Specs requiring review to assure that post maintenance opera-bility testing of safety related components in the Reactor Trip System is required. The review also examined whether the testing demon-strated the equipment is capable of performing its safety function before being returned to service.
The results of the review indicated two procedures requiring revision to meet the requirement. Procedure changes have been submitted and the changes will be in effect by February 1, 1984. Any additional guidance provided by the B&WOG will be factored into additional procedure reviews if needed.
The following table lists the procedures reviewed and the resultant changes:
PROCEDURES REVIEWED Procedure No.
Title 1302-5.1 Reactor Coolant Temperature Channels and Pressure /
Temperature Comparator 1302-5.2 RPS High and Low RC Pressure Channels Required Interval - Refueling Interval 1302-5.4 Reactor Coolant Flux Flow Comparator Required Interval
- Refueling Interval 1302-5.6 RPS Punp/ Flux Comparator and RCP Power Monitor Surveillance Calibration 1302.5.7 High Reactor Building Pressure Channel 1302.5.34 Reactor Trip on Loss of Feedwater/ Main Turbine Trip Required Interval - Refueling Interval 1303-4.1 Reactor Protection System Required Interval Monthly 1401-7 Reactor Coolant Temperature Detector Removal and Installation 1420-CRD-4 Troubleshoot CRD Breaker 1430-CRD-18 (Note 2)
Check _CRD Power Supplies for Faulty SCR's and Diodes and Gate Drive Units.
1430-RPS-1 Repair Linear Amplifier, RPS 1
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f1430-FS-2 Repair an FS Module 4;
E-36 (Electrical) CRD Trip Breaker Check 1303-11.1 (Note 1)
Control Rod Drop time IC-120-Reactor Protection Cabinet PM's IC-66
- Instrumentation System Preventive Maintenance E-11
- Control Rod Drive Resistance Check T.S. #4.1 Operational. Safety-Review 1.
PCR 1-EG-83-0140' submitted 10/10/83. Paragraph 6.4.5 added requiring FS Channel C Logic Crack to verify the Electronic Trip was not affected by connections during test.
2.
PCR l-MT-83-6507. submitted 10/6/83. Revised procedure to require
. independent verification of lead termination and retest after repair to assure trip function is not affected.
3.1.2..Pending completion of the vendor information now being developed under Item 2.1 the applicable procedures will be reviewed and updated to incorporate any ' additional guldenance received from the B&WOG program addressing this item.
3.1.3
.To date, GPUN has not identified any post maintenance test require-ments in existing Tech Specs which can be demonstrated to degrade rather than enhance safety.
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3.2 Post-Maintenance Testing (All Other Safety Related Components) 3.2.1 GPUN performed a review of applicable administrative procedures and Tech Specs to assure 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. Listed below are the existing administrative procedures and Tecn Specs by number and title and a short summary of the requirements of each of these procedrues which address the subject of post maintenance testing.
(1) AP 1001D Procedure Preparation of the procedure addresses maintenance procedures format and Content. It requires each procedure to include an Acceptance Criteria Section which addresses any checks, tests, etc. to be performed after the actual work is completed prior to returning the equipment to service. 0 of the procedure addresses surveillance procedure format and content. It requires the procedure to include detailed steps for removing a channel or system from service, performing the test and restoring the channel or system to service. The acceptance criteria is based upon Tech. Spec. requirements where applicable. It should be noted that for many post maintenance tests, the applicable surveil-lance procedure is run.
(2) AP 100lK Periodic Procedure Review of the procedures " Periodic Procedure Review Checklist" addresses procedure conformance to the current Technical Specifica-tion, follow-up testing and provisions for proper returning the system to service.
(3),AP 100lJ Techncial Specification Surveillance Testing Program One of the objectives of this procedure is to insure that documen-tation of test results is provided to permit determination of component operability. It also address the situation where if deletion of all or part of a test is necessary because of plant conditions, the deleted sections must be performed before the system is needed for plant operation. Shift Supervisor / Shift Foreman approval is required prior to return to service.
(4) AP 1027 Preventive Maintenance This procedure requires that for tasks to be performed on components listed in the QCL (ES-011), a standing preapproved preventive
maintenance (PM) procedure must be attached. It also requires that acceptance criteria and post maintenance testing requirements be included in each individual PM procedure.
(5)f4) 1407-1 Station Corrective Maintenance Procedure For corrective maintenance work which is performed without an approved procedure, the Job Ticket Form has a section which covers post main-tenance testing required and acceptance criteria to demonstrate that structures, systems and components will perform satisfactorily in service. When the component is repaired, the Maintenance Foreman approves the release for Operations Department system realignment amd operability testing. Tte Shift Foreman approves operability testing completed and that the retest met acceptance criteria.
(6) Technical Specifications Section 4.1 Operational Safety Review One can conclude that the above procedures provide adequate guidance and provide the direction to assure post maintenance operability testing is demonstrated. One must also bear in mind that post maintenance opera-bility testing is also highly dependent F che type and extent of the maintenance performed.
GPUN plans to review plant maintenance, PM and surveillance procedures and will review the post maintenance testing requirements for adequacy. See our response for item 2.2.
The Plant Staff will also review each main-tenance procedure as part of the bi-annual review program and will pay particular attention to the post maintenance testing section to assure operability.
3.2.2 Vendor manuals have been used in the past as guidance in the writing of maintenance procedures as well as a source of appropriate test guidance.
Vendor Service Bulletins, etc. were also reviewed when received for applicability and procedure changes were made to include the latest requirements.
Tech. Ftnctions is presently conducting a technical review of vendor manuals and will use these manuals as a reference document when they perform their review of plant maintenance, PM and surveillance proce-dures. Also, a formal program which is described below.s being estab-lished to better control all vendor information for safety-related components, which should help assure that the latest requirements are picked up in the applicable procedures. This will be an ongoing effort.
PROCEDURE FOR REVIEW CF VEtOOR INFORMATION
Background
During the last audit, IPPO suggested that GPUN establish a coor-dinated system to review all information received from vendors to
ensure no important items are missed. This requirement is a result of the Salem incident and the failure to imple;nent a maintenance item sent from a vendor. As a secondary item, this program will require an administrative system to track all the items and to document the resolution of the items.
Scope This program will collect, review and track all technical information received from vendors with the exception of technical manual changes which will be handled through EP-021. For example, this includes vendor bulletins, information letters, service letters, maintenance recommendations, etc.
It would specifically exclude vendor proposals or other project related correspondence.
Irtplementation This program will be primarily implemented by a change to EP-017
" Review of Operating Experience". All incoming infonnation will be reviewed by Plant Analysis personnel and then assigned to Technical Functions, Plant Engineering, Operations, or Maintenance for action, as appropriate.
These items will be tracked to completion in the Plant Analysis computer tracking system. The review of these items will be performed in the same manner and in the same depth as all other operating experience items. This means an attempt will be made to determine the applicability of the item to our plants. However, many items will be sent to the appropriate department for detenni-nation.
Items which during any part of the review are determined to be not applicable will receive at least two reviews by corrpetent technical perso';nel prior to final filing. Items effecting any design consideratiorct will be reviewed in Tech. Functions prior to sending to the plant.
Some other changes to procedures in both project engineering and at the plant will be required. There changes will be made to insure that vendors are required to send all this information to Plant Analysis Section when the equipment is procered. Also that information that is received by any means either at the plant or through the projects organization will be sent to Plant Analysis, so that one central organization will see all the vendor information. This new system is not intended to discourage other individuals from reviewing infor-mation they receive and making recommendations prior to sending the material to Plant Analysis. These types of review will be helpful and would speed the implementation of all this information.
The changes and revisions to this procedure (EP-017) are anticipated to be implemented by April 1984.
3.2.3 GPJN to date has not identified any post maintenance test requirements in existing Tech Specs that are perceived to degrade rather than enhance all other safety related components.
4.1 Reactor Trip System Reliability (Vendor-Related Modificatons)
The reactor trip circuit breakers at TMI-l are General Electric, Type AK-2-25.
As documented in G.E. letter to B&W dated September 7,1963 (see the B&WOG Response - App. B) concerning Vendor Recommended Reactor Trip Breaker Modifications, G.E. does not recommend any modification to their subject circuit breakers. The letter suggests that the shunt trip should be actuated simultaneously with the U/V release mechanism for the reactor trip function. This facility is presently being provided and described in response to Section 4.3 of the Generic Letter.
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4.2 Reactor Trip System Reliability (Preventr.cive Maintenance and Surveillance Program for Reactor Trip Breakers) 4.2.1 Preventive Maintenance procedure PM E-36 contains the instructions for periodic maintenance, lubrication, cleaning, adjustments and trip time testing. A revision to the procedure is being prepared to incorporate the latest vendor recommendations in Operating Plant Service Bulletin 25-009 which contains a General Electric Company Service Advisory Supplement pertaining to the G.E. AK-type breakers that are in service as control rod drive trip breakers at TMI-1. General Electric provided this information to B&W in response to a request by B&W on behalf of the B&WOG Regulatory Response Group immediately following the Salem ATWS events of February 1983. The maintenance is performed on a 6 month interval based on the environment that the breaker is in and past experience. The revised procedure will be issued by February 1,1983.
Each breaker is functionally tested monthly (when the control rod system is energized) to assure that it trips when the undervoltage device is de-energized. This test is accomplished by Surveillance Procedure 1303-4.1 Reactor Protection System. The number of cycles for this monthly test has been reduced from 6 to 1 to reduce wear on the breaker. This was done at the recommendation of the vendor.
4.2.2-4 The B&WOG Response addresses those itens concerning Reactor Trip Breakers surveillance and reliability programs.
GPUN has committed to participating in the OG RTB Monitoring Program that will compile and analyze maintenance / surveillance data for G.E. AK '2 CRDS air circuit breakers.
n 4.3 Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment for B&W Plants)
Below is a system design description of the proposed modification which pluvides automatic trip of the CRDM power feeder circuit breakers through the breaker's shunt trip devices.
A detailed schedule for implementing the modification has not been completed because the engineering and design will be performed by an Architect / Engineer from whom a detailed schedule has not yet been obtained. However, a project end target date has been established as 9/30/84, which includes revision of the surveillance procedure to provide independent on line testing of the new diverse trip mechanism.
- If Tech Spec changes are. required, they will be submitted for review prior
- to implementation of this modification.
CROM Circuit Breaker (Automatic Actuation of Shunt Trip) l.
1.0 Purpose and Scope
The control rod drive mechanism (CRDM) power feeder circuit breakers must trip to initiate an emergency reactor control rod insertion for reactor shutdown. The reactor protection system (RPS) logic or a manual trip command de-energizes the circuit breakers' undervoltage (U/V) trip devices which trip their-associated. circuit breakers. Because these existing U/V trip devices have had a history of failing to trip their associated l
- circuit breakers, a diverse method to trip these breakers is provided by this modification using the shunt trip mechanism as a secondary means of tripping each CRDM power feeder circuit breaker.
2.0 Functions and Design Requirements 2.1 Functions Two 480VAC_ circuit breakers supply power to the CRDM power supplies.
Downstream of the DC Hold power supplies are four 125VDC circuit breakers which are the power feeders for the holding power to the safety rod groups. To initiate a reactor shutdown, these AC and DC circuit breakers are commanded to trip by either the RPS logic or a manual trip signal which de-energizes the circuit breakers' instantaneous U/V trip device.
Per this modification,'each circuit breaker's shunt trip coil will be used to back up the U/V trip device to trip the circuit breaker upon a reactor trip command.
The reactor trip command is in the form of a loss of voltage to each circuit breaker's U/V trip coil. The circuit breaker shunt trip mechanism must be energized at 125VDC to trip the circuit breaker. As such, a new undervoltage relay will be added for each RPS logic output
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channel to sense trip command voltage. The new undervoltage relay contacts will be used to energize the existing shunt trip coil from a separate 125VDC power soume when the uitdervottage relay drops out.
2.2 System Configuration A shunt tilp device is currently mounted as an integral part of each AC and DC CRDM power cimuit breaker. The 125VDC power to the shunt trip device coils will be fed through normally closed contacts of the new undervoltage relays.
For AC circuit breakers CBl units 10 and 11, their respective undervoltage relays will be mounted inside available cubicles above the corresponding circuit breakers. For AC circuit breaker unit 10, the RPS trip logic switches power which is derived from vital bus lA (red channel) to operate the breaker's U/V trip device coil. The sensing mechanism of the under-voltage relay for that breaker will be connected in parallel to the U/V trip device coil. The control power of the new relay will be taken from the "A" channel 125VDC source which will also feed that breaker's shunt trip coil. For AC circuit breaker unit 11, the RPS trip logic controls power which is derived from vital bus 1B (green channel) to operate that breaker's undervoltage trip device coil. The new undervoltage relay's sensing mechanism will be connected in parallel to the U/V trip device coil. This circuit breaker's shunt trip coil and control power to the new undervoltage relay will be fed from the "B" channel 125VDC source.
For the two pairs of DC circuit breakers, the two undervoltage relays will be mounted in available cabinet space adjacent to the CRDM DC circuit breaker stack in the relay room. The pair of circuit breakers: CBl and CB2, are tripped via their U/V trip device from the trip logic which con-l trols power from vital bus "C" (yellow channel). One new undervoltage relay will sense voltage in parallel with the two existing U/V trip coils for the breakers. The control power to the new undervoltage relay will be taken from the "C" (yellow) channel 125VDC source which will also feed the two corresponding shunt trip coils. The second pair of DC circuit breakers: CB3 and C84, are tripped via their U/V trip devices from the trip logic which controls power from vital bus "D" (blue) channel. The new undervoltage relay for this pair of breakers will sense voltage in parallel. with the two existing U/V trip coils.
The control power to the new unuervoltage relay and the power to the circuit breakers' shunt trip coils will be fed from the "O" (blue) channel.
2.3 Instrumentation and Control The new automatic shunt trip circuit shall meet the single failure, circuit isolation and testability requirements of IEEE 279 - 1971.
For each of the four CRDM trip channels, new status indicating lanps will be provided to indicate presence of voltage at the undervoltage relay sensing points and at the shunt trip coils. These status lamps will be mounted at the panels containing the new relays.
Amunciation of the loss of the 125VDC shunt tripping voltage will be provided. A control relay connected at each of the four tripping circuits to sense incoming 125VDC power will have its contacts wired to the main control room amunciator.
A test switch shall be provided for each of the four CRDM tripping cir-cuits which will permit bypassing of the undervoltage relay and shunt trip mechanism to allow testing of the breaker trip through the U/V trip coil independent of the new undervoltage relay / shunt trip device. The arrange-ment of the test switch shall also be such that the shunt tripping of the breakers may be tested independent of the existing U/V trip coils. Each test switch shall key lockable and mounted inside the panel which contains its corresponding undervoltage relay.
The control relay for the amunciation of loss of control power shall be arranged to sense the bypassing of the undervoltage relay and the shunt trip coil when in the test mode.
For the two AC circuit breakers, the existing source interruption circuit will be modified to eliminate the existing non class 1E relay from the shunt trip coil circuit. The dry contact logic for the source interrup-tion trip will be connected to the shunt trip coil in parallel with the tripping contacts of the undervoltage relay.
2.4 Separation Requirements The addition of the automatic shunt trip shall not degrade the reliability or integrity of channel independence of the existing reactor trip systems.
The new undervoltage relays will be mounted inside separate cubicles or shall be separated from one another by a steel or a fire rated barrier.
All new wiring for this modification shall be run in chamelized raceways respecting the chamel color of the circuit.
All panel internal wiring of different channels shall be separated by a minimun of 6 inches of free air space or protected by a fire retardant coating or a fire barrier tape.
2.5 Operational Requirements The reactor trip signal which actuates the existing CRDM circuit breaker U/V trip device will also actuate the new interposing undervoltage relay whose contacts will control the 125VDC power to the operating coil of the shunt trip device. The new undervoltage relay will be energized continu-ously during reactor operation via sensing of the RPS logic output vol-tage. The reactor trip command is the interruption of this logic output voltage.
The undervoltage relay sensing voltage will be a nominal 12DVAC and the relay control voltage will be 125VDC. To prevent a inadvertent reactor
trip upon loss of a DC power source, the undervoltage relay shall be such that it does not drop out upon loss of the 125VDC control power.
2.6 Structural Requirements The new undervoltage relays test switches and alarm relays shall be qualified to a seismic response spectrum which meets or exceeds the spectrum applicable for the 338' elevation of the control tower.
The new undervoltage relays, test swit hes, test lamps, alarm relays, the shunt trip device and associated wiring shall all be mounted to withstand the maximun SSE loading defined for the 338' elevation of the control tower. These components are all class lE devices and as such Bhall be protected against any possible damage from adjacent non seismic mounted equipment.
The undervoltage relays, tests switches, the shunt trip device, and the alarm control relays shall be qualified for Class IE application in accordance with IEEE 323, 1974 and shall also be seismic qualified in accordance with IEEE 344, 1975. Those new components per this modifica-tion which will be located in a mild controlled environment will not require envrionmental qualification.
2.7 Testing The tripping function using the undervoltage relay and shunt trip mechanism shall be incorporated into the existing plant surveillance procedure. The CRD circuit breaker testing shall include tripping of the circuit break via the U/V trip device and via the undervoltage relay / shunt trip mechanism independent of one another by using tha test isolation switch. While in the test mode, the corresponding trip circuit's voltage status indicating lights will be DFF indicating no voltage to the under-voltage relay and its associated shunt trip circuit. These status lamps will be located at the panels which will contain the U/V relays and test switches and will be near their respective DC and AC power feeder breakers on elevation 33' of the control tower. Also, in the test modo, an alarm on the main control room annuciator will alrert the operator that a trip function is being bypassed.
3.0 Quality Assurance This system is classified Nuclear Safety Related. The Operational QA Plan for GPUNC shall apply.
4.0 Human Factors The arrangement of the test swithces and indicating lamps will be reviewed by Human Factors prior to finalizing the design. Human Factors shall have final approval of component identification and annunciator legends.
F-l 4.4 Reactor Trip System Reliability The "CRDCS Trip Portion" is classified as Nuclear Safety Related by the TMI-l Quality Classification List. The trip portion includes the electronic trip and SCRs.
The trip function of the SCRs was not previously verified. A change to test procedure 1303-4.1-(Reactor Protection System) has been approved and was distributed.
The revised procedure confirms the trip function by verifying the reduction in current from the affected power supply.
The language of the TMI-l Tech Specs is broad enough to cover the SCR portion of the trip function. However, for clarity, an administrative Tech Spec change will be included with several other administrative changes at a future convenient time, i
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4.5 Reactor Trip System Reliability (System Functional Testing) 4.5.1 At T)C-1, on lirk. testing of 'the' reactor trip system is performed on a
/ monthly basis, per TMI-l Surveillance Procedure 1303-4.1.
Independent testing of the two presently installed diverse trip features is provided in the procedure, which includes the CRDM power feeder breakers' U/V trip device test and the CRD control system (CRDCS) SCR trip circuit test. The CRDCS SCR trip test guidelines, as provided by B&W, have been incorporated in Surveillance Procedure 1313-4.1 as of October 12, 1983.
The facility for tripping the CRDM power feeder circuit breakers via their shunt trip devices is scheduled to be installed by 9/30/84.
Surveillance Prccedure 1303-4.1 will be revised by 9/30/84. At that time, this procedure will include independent testing of the three diverse trip features: U/V trip of CROM power feeder breakers, shunt trip of the CRDM power feeder breakers and SCR tripping.
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4.5.2 TMI-1 is designed to permit periodic on-line testing.
4.5.3 The B&WOG addressed this' item _in the OG response to the Generic Letter.
GPUN is an active participant inethis program:and will review the results of this Owners Group activity to determine if the on-line functional testing required by Tech Specs is consistent with these results.
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Generic Letter 83-28 Estimated Schedule for GPUN (TMI-1) Activities
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Estimated Item Description _
Completion-Date 1,1-Implement TMI-1 Reactor Trip Program 6/30/84 /
Recommendation into Procedures Implement Training 6/30/84 1.2 Incorporate Data Checklists 6/30/84 into TMI-l Procedures 2.1 Equipment Classification (RTS)
Complete Vendor Interface.
B&WOG Activity 2.2 Equipment Classification (all Safety -
Complete Related Equipment)
Recommendations issued on a Component 3/31/84 Quality Classification List Technical Review of all Vendor Manuals 1/31/90 Review of Maintenance and Surveillance 1/31/87 Procedures 3.1 Review of GPUN Procedures (RTS)
Complete Revise Procedures (RTS) 2/1/84 Additional Cuidelines B&WOG Activity 3.2 Reveiw Program of Maintenance & Surveillance See 2.2 Procedures Technical Review of all Vendor Manuals See 2.2 Revise Procedures for Review of Operating Experience 3/31/64 4.1 Vendor Recommendations (RTB's)-
Complete 4.2 Revise Procedure due to Vendor Recommendation 2/1/84 Reactor Trip Breaker Reliability Program B&WOG Activity 4.3 Install Automatic Actuation of Shunt Trip 9/30/84 4.4 Revise Procedure for SCR Testing Complete' 4.5
. Revise Surveillance Procedure based on B&W Guidelines I
Complete Revise Surveillance Procedures for the Three Diverse Trip Functions 9/30/84 Program to Justify Online Test Intervals B&WOG Activity s
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