ML20085N466
| ML20085N466 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/05/1983 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| GL-83-28, NUDOCS 8311110049 | |
| Download: ML20085N466 (54) | |
Text
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BALTI MO RE GAS AND ELECTRIC CHARLES CENTER P.O. BOX 1475 BALTIMORE, MARYLAND 21203
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November 5,1983 Sumpty Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION: Darrell G. Eisenhut, Director Division of Licensing
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2, Docket Nos. 50-317 and 50-318 Generic Letter 83-28; Required Actions Based on Generic Implications of Salem ATWS Events
Reference:
(a)
Letter from D. G. Eistnhut to All Licensees dated July 8,1983; Same subject (b)
Letter from Mr. A. E. Lundvall, Jr., to Mr. Darrell G.
Eisenhut dated September 7, 1983, (60 day response to G.L. 83-28)
(c)
Letter from Mr. 3. R. Miller to Mr. A. E. Lundvall, J r.,
dated October 25, 1983, Clarification of Required Actions Based on Generic Implications of Salem ATWS Gentlemen:
The enclosures and accompaning attachments provided herein constitute our reply to the 120 day response requirements of Reference (a). Reference (b) forwarded a request for an extension to the 120 day re ;ponse period along with a description of some of the existing and intermediate actions we have taken to aadress the immediate concerns of Reference (a). In accordance with Reference (c) we are providmg the enclosed reply which includes a schedule for addressing the items in Reference (a) that are not covered in this transmittal. Enclosure 1 provius a tabular assessment of the status of our response to each item contained in Reference (a). Our proposed schedule for providing the completed responses takes into account previously scheduled manpower commitments to support other licensing activities and outage schedules.
We believe that the actions we have taken to date, and that are described in Reference (b), provide adequate justification to support the proposed schedule for responding to the remaining items not covered in this transmittal.
I B311110049 B31105 PDR ADOCK 05000317 l
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e D. G. Eisenhut November 5,1983 Should you have further questions regarding this reply, please do not hesitate to contact us.
Very truly yours,
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AEL/ LOW /BSM/sjb Enclosures cc:
- 3. A. Biddison, Esquire G. F. Trowbridge, Esquire D. H. Jaffe, NRC R. E. Architzel, NRC STATE OF MARYLAND :
TO WIT:
CITY OF BALTIMORE :
Arthur E. Lundvall, Jr., being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge,information, and belief; and.that he was authorized to provide the response on behalf of said Corporation.
WITNESS my Hand and Notarial Seal:
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ENCLOSURE 1 i
RESPONSE TO NRC GENERIC LETTER 83-28 ITEM NO.
DESCRIPTION STATUS / PROPOSED RESPONSE DATE 1.1.1 Criteria f or acceptability Complete, see Enclosure 2, Attachment 1, of restart Action F 1.1.2 Responsibilities and Complete, see Enclosure 2, Attachment I authority of review personnel 1.1.3 Qualifications and Training Complete, see Enclosure 2 of responsible personnel 1.1.4 Sources of plant informa-Complete, see Enclosure 2, Attachment 1, tion Action B, Attachments 2, 3,4, and 5 1.1.5 Methods and Criteria for Complete, see Enclosure 2, Attachment 1, comparing event information 5 and 6 1.1.6 Criteria for independent Complete, see Enclosure 2, Attachment 1, assessment and preservation Action G and item 1.2 of evidence 1.1.7 Systematic Safety Assess-Complete, see Enclosure 2, Attachment 1, ment procedures Actions A, B, C, D, E (1, 2, 3), G, F and 1.2.1 Capability for assessing Complete, see Enclosure 2 sequence of events 1.2.2 Capability for assessing Complete, see Enclosure 2 time-history of analog variables 1.2.3 Other data and information Complete, see Enclosure 2, Attachment I sources 1.2.4 Schedule for changes to Complete, see Enclosure 2 data and information capability 2.1.1 Equipment classification Complete, see Enclosure 3 for Reactor Protective System 2.1.2 Vendor Interface for Partial response, schedule to be provided Reactor Protective System following the results of NUTAC recommendations, see Enclosure 3 2.2.1.1 Equipment classification Complete, see Enclosure 3 for all safety-related systems
ENCLOSURE 1 RESPONSE TO NRC GENERIC LETTER 83-28 ITEM NO.
DESCRIPTION STATUS / PROPOSED RESPONSE DATE 2.2.1.2 Equipment classification Complete, see Enclosure 3 information handling system 2.2.1.3 Description of the infor-Complete, see Enclosure 3 mation handling system imple-mentation 2.2.1.4 Description of the manage-Complete, see Enclosure 3, and ment controls for the response to item 2.2.1.2 information handling system 2.2.1.5 Demonstrate design verifi-Complete, see Enclosure 3 cation and qualification testing is specified for safety-related components 2.2.1.6 Inclusion of broader class Partial response, schedule to be of systems for 2.2.1.5 provided following the results of AIF Subcommittee recommendations, see 2.2.2 Vendor Interface Program (See response and schedule for item 2.1.2) for all safety-related components 3.1.1 Results of test and maint-Incomplete, schedule to be provided by enance procedures and T.S.
February 29,1984 reviews for operability testing of RTS 3.1.2 Results of vendor and Complete, response for RTS Breakers, engineering recommendations see Enclosure 4 reviews of RTS 3.1.3 Identif y tests, procedures Incomplete, schedule to be provided and T.S. that degrade by February 29,1984 safety for the RTS 3.2.1 Results of test and maint-Incomplete, schedule to be provided enance procedures and T.S.
by February 29,1984 reviews for operability testing of all safety-related equipment 3.2.2 Results of vendor and Incomplete, schedule to be provided engineering recommendations by February 29,1984 reviews of all safety-related equipment
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o ENCLOSURE 1 RESPONSE TO NRC GENERIC LETTER 83-28 li'EM NO.
DESCRIPTION STATUS / PROPOSED RESPONSE DATE 3.2.3 Identify test, procedures &
Incomplete, schedule to be provided T.S. that degrade safety of by February 29,1984 safety-related equipment 4.1.1 Verify vendor-recommended Incomplete, schedule to be provided modifications implemented by February 29,1984 for the RTS 4.1.2 Verify written evaluations Incomplete, schedule to be provided f or vendor-recommended by February 29,1984 modifications not implemented for the RTS 4.2.1 Description of PM & STP Complete, see Enclosure 5 programs for periodic maintenance of the RTS 4.2.2 Description of parameter Partial response, see Enclosure 5, trending program for fore-response to item 3.1.2, schedule to be casting degradation of RTS provided by February 29,1984 4.2.3 Description of life cycle Partial response, schedule to be provided testing of RTS breakers by February 29,1984 4.2.4 Description of periodic Incomplete, schedule to be provided replacement program for RTS by February 29,1984 components 4.3 Description of automatic Not Applicable to Calvert Cliffs actuation of shunt trip for Westinghouse and B&W plants.
4.4.
Description of improvements No Applicable to Calvert Cliffs in maintenance and test proce-dures for B&W Plants 4.5.1 Description of on-line Incomplete, schedule to be provided functional testing of RTS by February 29,1984 4.5.2 Description of justifica-Incomplete, schedule to t,e provided tion for not making on-line by February 29,1984 functional testing modifications 4.5.3 Description of review of Incomplete, schedule to be provided on-line functional testing by February 29,1984 intervals to ensure high RTS availability a
ENCLOSURE 2 RESPONSE TO GENERIC LETTER 83-28 1.1 POST-TRIP REVIEW (Program Description and Procedures)
Consistent with Baltimore Gas and Electric's (BG&E) commitment to safe nuclear power plant operations, Calvert Cliffs Nuclear Power Plant (CCNPP) has procedures for analyzing unscheduled reactor shutdowns in order to make a determination that the plant can be restarted safely. These procedures ensure a comprehensive, systematic and independent evaluation of all reactor trip events.
They provide for initial post-trip and independent reviews, event reporting procedures, and post-trip critiques. These procedures, discussed below in the individual responses, are integrated into an overall program for analyzing unscheduled reactor shutdowns to achieve a complete understanding of the cause of the event and the resultant safety system response.
Item 1.1 of Generic Letter 83-28 requests a description of the Post-Trip Review Program. The CCNPP program for analyzing and evaluating unscheduled reactor shutdowns is summarized in Attachment 1 in terms of:
o Program actions o Personnel involved in each action o Authority / responsibility of involved personnel, and o Criteria, including data needs, governing program actions.
In requesting a description of the Post-Trip Review Program, item 1.1 of the generic letter provides that certain specific information be addressed. Since this information is substantially contained in the overall program description provided in Attachment 1,
we have identified in Attachment I the specific staff information request to which a program item is responsive. This further serves the purpose of providing the information response in the context of the overall program. For ease of reference, a summary is provided below showing where in or otherwise the response to a specific NRC information request in Item 1.1 of the generic letter is located.
1.1.1 NRC Request - Describe the criteria for determining the acceptability of restart.
Response - Restart criteria are provided as part of the overall program in, Action F (Review for Restart), Criteria column.
1.1.2 NRC Request - Describe the responsibilities and authorities of pesonnel who will perform the review and analysis of these events.
Response - See Columns 2 and 3 of Attachment I for personnel involved in post-trip reviews and their authority / responsibility for program actions.
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- ENCLOSURE 2 1.1.3 NRC Request - Describe the necessary qualifications and training for the responsible personnel.
Response - This information is not part of Attachment 1, but is provided below.
The qualifications and training of personnel responsible for the review, analysis, and restart authorization are identified in the FSAR, Sections 12.1 and 12.2. The minimum qualifications and training for all positions comply, respectively, with Section 6.3, " Facility Staff Qualifications", and Section 6.4, " Training", of the CCNPP Technical Specifications. The qualifications and training are referenced to the appropriate sections of ANSI N18.1-1971,
" Selection and Training of Nuclear Power Plant Personnel". In most instances, the education and experience of CCNPP personnel surpasses the required qualifications.
In addition to the above, the CCNPP training staff has provided specific training during its 1983 requalification program to operations and technical personnel regarding:
Implementation of CCI-ill A, Post-Trip Review Requirements, Plant computer sequence of events capabilities, Plant computer post-trip review capabilities, and an Overview of the Salem ATWS events.
1.1.4 NRC Request - Describe the sources of plant information necessary to conduct the review and analyses.
Response - The sources of plant information needed for review and analysis purposes are identified in Attachment 1, Action B (Initial Review), Criteria column. Reference is made to Attachment 2 which identifies parameters that could indicate a condition leading to a plant trip; Attachment 3 whicn identifies parameters that are part of the Post-Trip Review output; Attachment 4 which identifies parameters that are obtained from the Technical Support Center; and Attachment 5 which identifies the analysis to be performed on various plant parameters and the sources of data for these parameters.
_3 ENCLOSURE 2 1.1.5 NRC Request - Describe the methods and criteria for comparing the event information with known or expected plant behavior.
Response - As specified in Attachment 1, the post-trip review is performed by two groups within CCNPP. One review is performed by Operations personnel; the other, independent of the first, is performed by personnel in the Operational Licensing and Safety (OL&S) Unit. Both reviews are directed to evaluating not only the cause of the event, but the manner in which the plant, particularly safety systems, responded to the event.
Guidelines for the Operations review are contained in Calvert Cliffs Instruction (CCI)-lll, " Post-Trip Review Requirements", which provides a checklist for evaluating the response of plant systems. The checklist, along with the sources of necessary information, is provided in.
The checklist provides criteria for assessing event information against known or expected plant behavior. In general, the checklist provides that the following be verified:
The response of a specified parameter is consistent with the response of one or more other specified parameters.
The response of specified parameters are within specified operating levels.
The response of specified parameters are in accordance with the CCNPP Technical Specifications.
The trend of specified parameters is consistent with the event.
The actuation of safety systems is consistent with the response of specified plant parameters.
The concurrent, independent review performed by OL&S Unit personnel 4
is outlined in Attachment 6. The attachment outlines the type of logic and criteria used in the review to compare event information with expected / postulated plant behavior.
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' ENCLOSURE 2 1.1.6 NRC Request - Describe the criteria for determining the need for independent assessment of an event and guidelines on the preservation of physical evidence to support independent analysis of the event.
Response -It is specified in Attachment 1, Action G (Independent Review of Trip), Criteria column, that in accordance with the requirements of CCI-127, "Calvert Cliffs Event Reports", an independent review is made of all unscheduled reactor trips. Preservation of physical evidence to support independent analysis of the event is consistant with the data requirements specified in Attachment 1.
Data acquisition / retention capabilities are discussed in the response to items 1.2.1 and 1.2.2.
In general, data is acquired automatically following a trip, or is available af ter the trip upon manual request.
1.1.7 NRC Request - Describe the systematic safety assessment procedures which are to be used in conducting the evaluation of unscheduled reactor shutdowns.
Response - The systematic safety assessment procedures which are used in conducting the evaluation of unscheduled reactor shutdowns are provided in Attachment 1 and the other referenced attachments. The procedures provide for the following elements in the review process.
Data Collection - (Addressed in Attachment 1, Actions B, C, D and G).
Trip Investigation - (Operations review addressed in Attachment 1, Actions A, B and E(1,2)); independent review by OL&S Unit addressed in Attachment 1, Action G, and Attachment 6).
Trip Investigation Review - (Addressed in Attachment 1, Action E(3) and G).
Restart Decision -(Addressed in Attachment 1, Action F)
Identification of Followup Actions - (Addressed in Attachment 1, Action G).
1.2 POST-TRIP REVIEW (Data and Information Capability)
The CCNPP Post-Trip Review Program incorporates equipment and methods that provide the capability to record, recall and display data and information to permit diagnosing proper functioning of safety-related equipment.
The NRC request for information in this area addresses the capability for as essing sequence of events and the time-history of analog variables.
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, ENCLOSURE 2 1.2.1 C==hility for Assessing Sequence of Events (Ordf Indications)
NRC Request: Describe the capability for assessing sequence of events.
Information should include the following:
o Brief description of equipment, o Parameters monitored, o Time discrimination between events, o Format for displaying data and information, o Capability for retention of data and information, and o Power sources.
Response: The plant computer has two functions which provide methods by which sequence of events can be determined and displayed. These functions are Sequence of Events and Alarming.
o Sequence of Events - The plant computer receives approximately 90 digital inputs, see Attachment 2, which could indicate a condition leading to a plant trip if their status changed to the " alarm" state.
Whenever a change to abnormal. status for one of these points is detected, a message indicating a plant trip is immediately displayed on the third line of the control room CRT. An "SE" indicator is provided to identify that this is the first event in a sequence that could lead to a plant trip and helps to differentiate this message from the alarms displayed on the CRT.
Detection of this event also causes the computer to begin recording subsequent events, until either 40 events have.been recorded or one i
minute has elapsed, whichever occurs first. This is repeated for the next 40 events, or one minute limit. After one minute the record of this sequence of events is automatically typed on the utility printer. As on the CRT, the time of the first event is recorded to the nearest second.
I Subsequent events are listed in order, with the time given in milliseconds af ter the initial event. This program has a two millisecond resolution.-
o Alarming - Each analog sensor that is tied to the computer is scanned periodically at one of seven frequencies ranging from one second to one minute. Most of these sensors have high and/or low limits assigned to them which are compared to the current value -
each time the input is scanned..When the input signal is found to be less than the low limit or greater than the high limits, a message indicating this condition is immediately displayed on the CRT. To distinguish this message from prior alarm messages still displayed, the point's identification will be flashed until the ALARM AKC button is pressed.
Also, the newest alarm message is always displayed as. the top line in.the CRT alarm. section.
For documentation purposes, the - message is automatically and simultaneously printed in expanded type on the alarm printer.
6-ENCLOSURE 2 Two features in the computer alarm system allow points to be continuously tracked that, for example, have exceeded the high limit setting and whose value has steadily increased. First, the value portion in each CRT alarm message is being updated at five-second intervals, so that the value displayed is always the current value.
Second, many points have incremental limits associated with them that may be set at any value above the high limit or below the low limit. The point that is continuing to increase in value would thus eventually exceed tne incremental setting and each time reflash the alarm message, pointing to the operator the existence of this condition.
To prevent repetitive alarm / clear messages in the situation where a sensor's value is cycling about its limit setting, a deadband percentage is applied to each point's limit.
There is space for simultaneously displaying 14 alarms on the CRT, with the newest messages appearing on the top. If more than 14 alarms are present, the oldest are pushed off the bottom of the CRT into an overflow storage area.
All points in this area may be retrieved through the OFF-NORMAL REVIEW function. As alarm conditions appearing on the CRT are cleared, messages from the overflow area will be returned to the bottom line of the CRT.
The above has described the alarm handling as applied to analog inputs. There are also many on/off, or digital inputs for which one position will generate an alarm message and the other a " clear" condition. While having no provision for deadbands or incremental limits, they are otherwise handled just as the analog inputs on the CRT and printer.
All essential plant computer equipment necessary for the continuous monitoring and recording of required plant parameters are provided from an uninterruptable power supply (UPS) with battery backup. In the event of UPS failure, essential computer loads will be automatically transferred to an alternate power source.
1.2.2 Cmhility for Assessing the Time-History of Analog Variables NRC Request: Relative to assessing the time-history of analog variables provide inf ormation regarding:
o Brief description of equipment, o Parameters monitored, sampling rate, and basis for selecting parameters and sampling rate, o Duration of time history (minutes before trip and minutes af ter trip)
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o Format for displaying data, o Capability for retention of data, information, and physical evidence,
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. ENCLOSURE 2 o Power source.
Response: The plant computer has four functions that provide methods by which time-history information can be retrieved and displayed by the computer. These functions are Post-Trip Review, Typed Trend, Recorder Trend and Logging.
In addition, the Technical Support Center (TSC) provides analog data which does not pass through the plant computer.
These sources of time-history information are described below.
o Post-Trip Review - The Post-Trip Review program periodically records a number of preselected input variables.
In this data acquisition phase,17 variables are collected at two-second intervals over a 28-second period, and 48 variables are collected at 30-second intervals over a five-minute period. The variables are stored on disc in two circular pre-trip buffers with newest data replacing oldest data. When a plant trip occurs, the time is recorded, pre-trip data collection stops in the first buffer (which stores data for a period of five minutes prior to a plant trip), and post-trip data is acquired for five minutes following the trip. The second buffer continues to collect pre-trip data until a second trip occurs. When post-trip data i
collection is complete, a message notifies the operator. Printout can be requested manually at anytime, or will occur automatically if a second trip occurs.
Parameters monitored for Post-Trip Review are identified in. The sampling rate of each parameter is as noted in.
o Typed Trend - This function permits the operator to print the value of a point, or group of points, at selectable time intervals ranging from 30 seconds to 99 minutes. Up to 35 groups of points may be assigned. Each group may contain from 1 to 17 points.
Groups 1-30 have points assigned to them that are equipment or system-related, so that the behavior of an individual system may be observed by activating the trend of the appropriate group. Groups 31-35 are retained for special trending. Group Numbers and the Associated Equipment or System are provided in Attachment 7.
Values for all points being trended are normally printed at the selected time intervals on the utility printer, except Groups 29 and 30 which are printed on the incore printer, along with the time of printing. All actions initiated by the operator pertaining to this function are displayed on the CRT and documented on the utility printer.
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- i ENCLOSURE 2 o Recorder Trend - This function allows the operator to trend a j
maximum of four values simultaneously on two 2-pen strip chart recorders.
Any addressable point (except cor.stants) may be I
trended. The pen selection, recorder offset and recorder range are chosen by the operator.
The bottom four iines of the CRT continuously display the point identity, range and offset for each pen that is currently trending.
o Logging - Each hour a summary of the last 60 minutes of operating history is printed automatically on the logging printer. At midnight a summary of the previous day is also printed.
Data printed on the log sheet consists of averaged quantities, accumulations or peak values. Each quantity on the daily log is usually reported in the same way as its 24 hourly constituents.
Whenever one of the sensors used in the hourly log is out of service or is deemed unreliable, the value entered on the log is preceded by an asterisk.
o Technical Support Center (TSC) Information - The TSC consists of a computer system capable of monitoring various power plant parameters (see Attachment 4 for a list of these parameters for CCNPP) for the purpose of providing real time and historical displays and reports to assist in unit shutdown assessment. The TSC consists of a Master Control Console (CRT/ keyboard, six 3-pen trend recorders, three high speed line printers and a CDC/ System status panel and one magnetic tape storage unit) and three work stations (CRT/ key board, one 3-pen trend recorder, and one slow speed printer). Through the work station consoles, one is able to access any of the data being monitored by the systems and to direct the display of this data on its CRT, its line printer, or its trend recorder.
The TSC instrumentation consists of existing. plant process instrumentation loops and loop power supplies from which isolating 4
transmitters drive the respective analog signals to be fed to a data acquisition and display system which has read outs in the TSC.
Incore neutron detectors and incore thermocouple signals to the TSC are obtained directly from the plant computer. Where practical, existing safety-related loops are used as the signal source. All essential TSC computer _ equipment necessary for the continuous monitoring and recording of required plant parameters are powered from a UPS with battery backup. -In the event of UPS failure, essential computer loads will be automatically transferred to an
_ alternate power source.
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_9 ENCLOSURE 2 The TSC data base is generated from data acquired from incore thermocouples and incore detectors every 30 seconds, and from all other instrumentation every second. The TSC system to display this data is divided into two major functions. They are pre / post trip log, and trend data display. The pre / post trip log provides data for plant conditions three hours pre-trip and one hour post-trip. The trend data display is used to present the three hour pre-trip file and real-time data base of the system TSC data is acquired upon manual request following each unscheduled reactor trip. This data is available for request until the TSC computer is reset manually to record information from a second trip.
1.2.3 NRC Request - Identify other data and information provided to assess the cause of unscheduled reactor shutdowns.
Response - Data and information sources are those previously identified in, or described in the response to Items 1.2.1 and 1.2.2.
1.2.4 NRC Request - Provide the schedule for any planned changes to existing data and information capability.
Response - No changes existing data and information capabilities that would significantly affect post-trip review are presently planned.
ENCLOSURE 3 RESPONSE TO GENERIC LETTER 83-28 2.1 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (RTS Components) 2.1.1 Equipment Classification NRC Request - Confirm that all components whose functioning 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.
Response - We have confirmed that components whose functioning 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.
This confirmation is based on a review of the following:
Q-List - (The list of safety-related items). The Q-List specifies that the Reactor Protective System (RPS) is safety-related. (See response to item 2.2.1.2 for description of classification system).
Maintenance Requests (MRs) - (The implementing document for all corrective maintenance, and plant modifications).
A complete review of all Unit I and 2 MRs written for the RPS was made to confirm that all MRs written for components whose functioning is required to trip the reactor have been consistently identified as safety-related.
The review confirmed that all MRs for such components were classified safety-related.
Facility Change Requests (FCRs) - (Documented proposal to perform any change that affects the fit, form, or function of a structure, system, or component of the facility). A review of all Unit I and 2 FCRs written for the RPS was made to confirm that all FCRs written for components whose functioning is required to trip the reactor have been consistently identified as safety-related. The review confirmed that all FCRs for such components were classified safety-related.
Preventive Maintenance Card (PM Card) - (Details individual preventive. maintenance actions and the manner in which each action is to be completed). A review of all Unit I and 2 PM Cards for the RPS was made to confirm that all ' PM Cards for components whose functioning is required to trip the reactor are l
identified as safety-related. The review confirmed that all PM Cards for such components are identified safety-related.
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o ENCLOSURE 3 Stock Spare Parts Index - (Index of all parts routinely stocked for use in CCNPP). The stock spare parts index provides descriptive information on each part including its safety classification. The safety classification, determined from the Q-List, is the basis for future purchases to replace parts drawn from stock. A thorough review of the index confirmed that all components identified in the index whose functioning is required to trip the reactor are ioentified as safety-related. A further spot check of purchase orders (for two of the last four years) verified that the proper safety classifications were incorporated into the parts requisitions.
2.1.2 Vendor Interface NRC Request - For RTS components, describe your program to 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 interface established.
Where vendors cannot 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 on reactor trip system components to assure that requisite control of the applicable instructions for maintenance work are provided.
Response -BG&E recognizes the importance of an effective vendor interf ace program. We have been actively involved in the INPO Nuclear Utility Task Action Committee (NUTAC) that was formed on September 1,1983, to address the vendor interface concerns identified by the NRC in Generic Letter 83-28. The NUTAC is working towards the objective of developing a practical approach to respond to the NRC's interests in a vendor interface program. The NUTAC is currently considering the use of INPO's NPRDS and SEE-IN programs as one option for improving the dissemination of vendor information regarding the operation, maintenance, repair, and replacement of safety-related components.
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- ENCLOSURE 3 We are presently reviewing and evaluating our procedures for ensuring proper receipt, control and use of vendor information for RPS and other safety-related components.
We intend to continue to monitor and participate in the NUTAC effort, and as appropriate, to evaluate other independent vendor interface programs. We plan to provide the NRC with our plans for ensuring an effective vendor information program by February 29,1984.
2.2 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (Program for all Safety Related Components) 2.2.1 Equipment Classification 2.2.1.1 NRC Request - Provide the criteria for identifying components as safety-related within systems currently classified as safety-related.
Response - Safety related components are defined as those components necessary to ensure:
(1)
The integrity of the reactor coolant pressure
- boundary, (2)
The capability to shutdown the reactor and maintain it in a safe condition, (3)
The capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures to individuals in excess of exposures specified in 10 CFR 20, or (4)
Items that the Q-List Committee considers should receive the same level of quality assurance coverage as those for items 1-3 above.
Implementing criteria provide that:
Items that appear on the Q-List are safety-related When an item is required for a system or component that appears on the Q-List, the item must be classified safety-related unless documentation can i
show that the Electric Engineering Department I
(EED) has classified it non safety-related.
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I Items shall be treated as safety-related when there is any doubt about their classification.
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. ENCLOSURE 3 2.2.1.2 NRC Request - Provide 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.
Response - The information handling system used by CCNPP to identify safety-related components is the Q-List and its l
attachments (i.e., the " red-lined" P&ID drawings, Q-List Interpretations, and the list of Class IE equipment). The Q-List is used in conjunction with the implementing criteria provided in the response to Item 2.2.1.1 to identify safety-related components for systems identified in the Q-List.
The Q-List provides a list of those items in the plant which meet one of the criteria of " safety-related" specified in the response to Item 2.2.1.1.
The following areas / equipment categories of the plant contain, in part or in total, safety-related items.
Structures
- Containment Structure,
- Containment Interior,
- Auxiliary Building,
- Intake Structure,
- Enclosure for Auxiliary Feedwater Pumps,
- Foundations / Supports for Seismic I Components,
- Concrete Block Work,
- Tank Enclosures, and
- Turbine Building.
Systems
- Engineered Safety Features Systems,
- Auxiliary Systems,
- Fuel and Reactor Component Handling Equipment /
- Fuel,
- Containment Ventilation Systems,
- Auxiliary Building Ventilation Systems,
- Waste Processing & Radiation Protection Systems,
' Steam & Power Conversion System, and
- Electrical Systems.
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-3 ENCLOSURE 3 Instrumentation and Control
- Control Boards (Partial),
- Systems and Included Instruments / Controls /
Indicators.
Electrical Wire, Cable and Circuits
- General, Power Circuits, and
- Instrumentation and Control Circuits.
Pipe Systems & Included Pipe, Valves, Fittings, &
Misc.
The Instrument Index provides a list of components for CCNPP that includes sensing elements, transmitters, indicators, control devices, recorders, power supplies, and control valves.
It identifies which of these components are classified safety-related.
The Instrument Index was developed by Bechtel Corporation for CCNPP and is referenced by the Q-List.
The " red-lined" P&ID drawings identify safety-related sections of piping systems for Unit I and can be used as guidance for Unit 2.
The red-lines were developed by BG&E personnel based on information contained in the primary Q-List. The red-lines themselves do not provide new information, but rather provide the Q-List information in a convenient manner for identifying safety-related piping systems and components.
The Q-List does not include every individual part of every safety-related component, it is, therefore, sometimes necessary to interpret it.
The Q-List Interpretations is a compilation of these individual classification actions.
The Class IE equipment list, which is an attachment to the Q-List identifies which equipment in Units 1 and 2 must be environmentally qualified.
- i ENCLOSURE 3 The BG&E Q"ality Assurance Procedure (QAP)-28,
" Control of items Covered by the Quality Assurance Program",
is the governing document for the establishment and maintenance of a Q-List for CCNPP.
It specifies that a Q-List Committee be created to, among other things:
Specify items, activities, and services to be included in the initial Q-List, and approve revisions to the list.
4 Provide guidelines for:
- evaluating safety-related items, activities, and services to determine if any part or portion may be designated non-safety related, 1
and
- specifying the safety-related characteristics of an item, activity, or service designated safety-related.
QAP-28 specifies that the Q-List Committee consist of at least one permanent representative from each of the four units in the EED's Nuclear Generation Engineering Section (NGES) and four other groups as follows:
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Plant Engineering Unit Electrical Engineering Unit Instrumentation and Control Engineering Unit Nuclear Licensing and Analysis Unit Production Maintenance Department Nuclear Power Department Operations Quality Assurance Section Engineering Quality Assurance Unit The QAP-28 requirement _ to develop and maintain a Q-List is implemented by Electric Engineering Department Procedure (EEDP)-4, " Establishment and Control of the List of Safety-Related Items".
EEDP-4 provides requirements for EED activities relating to preparation, approval, issue, revision, and interpretation of the Q-List for CCNPP. The EED Nuclear Generation Engineering Section is responsible for revisions / interpretations to the Q-List and for:
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. ENCLOSURE 3 Ensuring that revisions / interpretations of the Q-List are based on guidelines established by the Q-List Committee (An independent review of all interpretations is conducted by the Q-List Committee Chairman).
Maintaining a
log of revisions and interpretations.
Documenting the basis for each interpretation.
Interpretations, as noted, are the responsiblity of the NGES. However, Q-List interpretations must be approved by the Q-List Committee before classifications become effective. Any member of the Q-List Committee who dissents on an issue and considers the contemplated action a safety-hazard may refer the matter to the Off-Site Safety and Review Committee (OSSRC) for review.
Minutes are kept of all Q-List Committee meetings. These minutes document changes to the Q-List and to guidelines for interpreting the Q-List.
Minutes are provided to the following individuals.
Chairman of the OSSRC, All members / alternates of the Q-List Committee, Plant Superintendent of CCNPP, Supervising Engineer, Nuclear Generation Engineering, Nuclear Project
- Manager, Project Management Department, General Supervisor, Operations Quality Assurance, Manager, Electric Engineering Department,
- Manager, Production Maintenance Department, Manager, Purchasing and Stores Department, Manager, Quality. Assurance Department, and Manager, Nuclear Power Department.
- ENCLOSURE 3 The OSSRC reviews the minutes of the Q-List Committee meetings to determine if amendments to the Q-List or to the guidelines for its interpretation constitute an Unreviewed Safety Question. If an OSSRC review indicates that an amendment to the Q-List is an Unreviewed Safety Question or would constitute a safety hazard, the Vice President-Supply Division would be notified immediately.
2.2.1.3 NRC Request - Provide 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 to 10 CFR 50, Appendix B, apply to safety-related components.
Response
The control of safety-related activities is provided by QAPs which are implemented through specific department procedures. In all cases, the procedures specify that safety-related structures,
- systems, components, materials, or services are as specified in the Q-List.
QAP-15 " Changes, Tests, and Experiments" provides controls that ensure that changes, tests, and experiments at CCNPP conform to the requirements of 10 CFR 50.59. The procedure specifies that an FCR be initiated and approved before the implementation of any change, test, or experiment to structures, systems, and components. The FCR documents the proposed activity and ensures that a detailed engineering review, including safety analysis is prepared for proposed changes to safety-related items or other items to the extent that they are described in the FSAR. QAP-15 provides for initiation of an FCR by Nuclear Power, Production Maintenance, Electric Engineering or Project Management Department personnel.
QAP-15 requires that an item's safety classification be specified on the FCR. It provides the implementing criteria specified in the response to Item 2.2.1.1 for classifying safety-related components. For uniformity and accuracy, the safety classification is specified by the department's FCR Coordinator.
QAP-14, " Plant Maintenance", defines the responsibilities and procedures associated with preventive maintenance, corrective maintenance, and modifications performed on safety-related structures, systems, or components at CCNPP.
MRs described in CCI-200, " Maintenance Requests" are used to:
_9-ENCLOSURE 3 Initiate corrective maintenance, Implement FCRs per procedures specified in CCI-126,
" Administrative Control of Facility Change Requests",
Initiate certain preventive maintenance in accordance with CCI-211, "Calvert Cliffs Preventive Maintenance Program", and Implement Plant Maintenance in accordance with CCI-201, "CCNPP Maintenance Procedures".
In preparing the MR form, the safety classification of the item under consideration must be specified. CCI-200 specifies that the List of Safety-Related Components (Q-List) shall be consulted to determine if the item requiring a maintenance action is safety-related. If there is any doubt as to its classification, it must be classified safety-related. A copy of the MR is forwarded to the Quality Control Unit to, among other things, verify the correct safety classification of the equipment.
The CCNPP Quality Assurance Program applies to all structures, system, components, and activities that have been designated safety-related.
Implementing QAPs ensure that the criteria of 10 CFR 50, Appendix B, will be met throughout the operations phase of the plant.
QAP's specify interdepartmental relationships and departmental responsibilities as they relate to particular activities, regulatory requirements, and BG&E commitments.
The procedures (i.e., QAP's) that apply to each of the activities in the introduction to 10 CFR 50, Appendix B, are specifi,ed in the matrix provided in Attachment 8.
Department or lower-level procedures exist to implement each of the QAPs.
2.2.1.4 NRC Request - Describe the management controls utilized to verify that the procedures for preparation, validation and routine utilization of the information handling system have been followed.
Response - The Q-List was described in the response to item 2.2.1.2. Management controls to verify that the procedures for preparation, validation and routine utilization of the information handling system have been followed are described below in terms of Q-List preparation and Q-List application.
~ ENCLOSURE 3 o Q-List Preparation - Management controls to ensure uniform and accurate development of the Q-List are as follows:
Q-List interpretations are based upon a uniform set of guidelines established by the Q-List Committee. These interpretations are prepared by EED, receive an independent review by the Q-List Committee Chairman, and receive final review / approval for consistency with the guidelines by the entire Committee before becoming effective.
EED must document the basis for each interpretation and maintain a log of revisions and interpretations.
Any member of the Q-List Committee who dissents on an issue and considers the contemplated action a safety hazard may refer the matter to the OSSRC for review.
The OSSRC reviews the minutes of the Q-List Committee meetings to determine if amendments to the Q-List or to the guidelines for its interpretations constitute an Unreviewed Safety Question.
o Q-List Application - Management controls to ensure consistent application of the Q-List are described in terms of its use for preparation of MRs and FCRs, and procuring spare parts.
Maintenance Requests - MRs were previously c.tscussed in the response to items 2.1 and 2.2.1.3.
Their preparation is described in plant administrative procedure CCI-200, " Maintenance Requests".
The MR requires that the safety classification of the item being worked on be specified. CCI-200 requires that the Q-List be consulted for the item's safety classification. It further specifies that if there is any doubt as to the item's classification, it must be classified safety related.
Administrative controls provide that the classification, as well as other parts of the MR, be reviewed / approved by the Maintenance Foreman and the Quality Control Unit, j
11 ENCLOSURE 3 Facility Change Requests - FCRs were previously discussed in the response to Items 2.1 and 2.2.. 3.
Their preparation is described in QAP-15, " Changes, Tests, and Experiments". The FCR requires that the safety classification of the item be specified.
QAP-15 specifies that the classification be determined by the originating department's FCR Coordinator.
It further specifies that items be treated as safety-related when there is any doubt about their classification.
Administrative controls provide that the classification, as weil as other parts of the FCR, be reviewed / approved by the Plant Superintendent for FCRs originating in the Nuclear Power Department or Production Maintenance Department; by the Supervising
- Engineer, Nuclear Generation Engineering Section and the Plant Superintendent for FCRs originating in EED; and by the Nuclear Project Manager and Plant Superintendent for FCRs originating in the Project Management Department.
Purchasing of spare parts - The controlling form for adding spare parts to company stock requires specifying whether the item being ordered is or is not safety-related. The determination is made from the stock spare parts index for which parts were originally classified from the Q-List.
~A new item to company stock would be classified directly from the Q-List.
Verification of the proper classification is ensured by review / approval of each stock order for NPD by the Supervisor-Test Equipment, General Supervisor-Electrical and
- Controls, Plant Superintendent, and NPD Manager.
4 2.2.1.5 NRC Request - Provide 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 and provide support for the licensees receipt of testing documentation to support the limits of life recommended by the supplier.
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ENCLOSURE 3 l
Response - QAP-2, " Procurement and Storage", provides that controls for procurement and storage for safety-related items be specified in the BG&E Procurement and Storage Manual. The Manual specifies that safety-related items or services shall be controlled by a specification package that meets the following requirements:
o The EED using guidelines established by the Q-List Committee, shall establish and document the following in the specification package:
The safety-related function of the item or the safety-related function that could be affected by the service.
The critical characteristics that should be controlled to ensure that the item or service will fulfill its safety-related function, or for safety-related items requiring equipment qualification, to ensure that they will operate under specified environmental conditions.
The activities that should be controlled to ensure that the item has the required characteristics.
The documentation required to establish that the foregoing requirements have been met.
(This specification requirement provides that necessary design verifications and qualification tesing be performed and documented to ensure component acceptability).
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o EED, together with the Engineering Quality Assurance
- Unit, shall develop and specify i
procurement requirements that address, among other things, the following topics to ensure that the item or service meets requirements documented above:
Technical Requirements (includes the referencing of industry codes and standards, and regulatory requirements),
l Quality Assurance Program Requirements (includes the specification of specific qualification and testing requirements), and Documentation Requirements (includes certifications).
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ENCLOSURE 3 1
o For replacement items, require the supplier to certify that replacements for items' clasdfied j
safety-related/ environmentally qualified have not i
been subject to a design change that could affect their environinental qualification, or supply test data that establishes their environmental 4
qualification.
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Implementation of the above requirements 'ncludes i
among other things, technical assessment on a case-by-case basis of appropriate design verification and qualification testing needs for each item being procured.
These needs are addressed in the 4
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specification by reference to relevant industry codes j
and standards (e.g., ASME Boiler and Pressure Vessei Code for applicable mechanical components, IEEE '
s Standards for relevant electrical components), NRC requirements, or through direct specification of
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appropriate test conditions and methods.
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With regard to specifications that address s the environmental qualification of electrical equipment,'a subgroup within EED called " Electrical Equipment Qualification" has been assigned responsibility for the -
s following actions.
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Identify items for which environmental qualification must be documented,
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Prepare an environmental qualification section for each specification includes requirements for,
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temperature, pressure, humidity ranges, s
- radiation, chemical exposures, l
- seismic response, and y
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- aging characteristics.
specification of i
Review
- and approve. acceptability s
Specifications - written for procuring seismic Category -I equipment Identify the appropriate seismic qualification methods to -be used in verifying that such - equipment can perform its required function. This specification, normally in v 1
the form of a comprehensive Appendix, also;
' includes:
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ENCLOSURE 3
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Seismic design requirements, q
Environmental exposure prior to seismic i
- testing, H
. ' Acceptable qualification methods,
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Test acceptance criteria, m
k Seismic qualification conditions, and
- ,' s Documentation / certification requirements.
s BG&E specifications for procuring safety-related i
equipment also include a provision which typically requires that:-
The supplier cebify, in writi g, that the 1
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equipment satisfied fun'ctional test requirements and acceptable \\;;iteria (i.e.,
5 mechanical, physical, pressure, procf, leak, load, hydraulic / pneumatic, hydrostatic, and h
electrical).
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4 The supplier provide certified physical and chemical reports that show actual values ~of I
chemical and physical properties of materials ordered.
The supplier provide certified mill test l
reports for all pressure retaining materials.
3 Functional, performance, environmental, electrical, mechanical, physical, operative, proof, pressure, leak, or other tests performed t'o. satisfy requirements 'of this specification be documented in a report s
I provided to BGdE. This report shallinclude s
as a minimum:
Association of the test with the,
purchase order.
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- Identificatioh ' of product. or item tested. N -/ > 3 i
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Specification,
- drawing, standard, s.
and/or procedure (including revision s
number) used for the test. :
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- ENCLOSURE 3
- Acceptance limits of the test, if applicable.
Number of units tested.
- Method and equipment used for the test.
- Actual test
- results, including acceptance and number and nature of failures (if any).
- Name of test technicians and testing agency.
Signatures of individuals performing and observing the tests. and dates of these activities.
2.2.1.6 NRC Request - 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 ot 7tructures, systems, and components important to safety re.suired by GDC-1 (defined in 10 CFR 50, Appendix A, " Genera Design Criteria, Introduction").
Response - With respect to the equipment classification l
program for non safety-related (NSR) structures, systems and components, we are a member of both the AIF Subcommittee on Safety Classification and the Utility Safety Classification Group and are seeking a generic resolution to the NRC Staff's concerns through the efforts of these groups.
We take exception to the Staff's backfitting of the Nr t "important to safety" such that it tends to invoke tN (..eral Design Criteria for equipment heretofore cmh ae as NSR. We support the belief that NSR equipmait n 3:.my designed, installed and maintained in a manner whicn e.uures its reliability to perform its function.
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- s ENCLOSURE 3 2.2.2 Vendor Interface NRC Request - Describe your program to 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 established.
Where vendors cannot be identified, have gone out of business, or will not supply information, the licensee or applicant shall assure that suffF.ient 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 acknowledgment for receipt of technical mailings. It shall also define the nuclear and nonnuclear divisions of their vendors that provide service on safety-related equipment to assure that requisite centrol of and applicable instructions for maintenance work on safety-related equipment are provided.
Response - See response to item 2.1.2 1
ENCLOSURE 4 RESPONSE TO NRC GENERIC LETTER 83-28 3.1 POST MAINTENANCE TESTING (RTS Components) l 3.1.2 NRC Request - Confirm the results of any reviews of vendor and engineering recommendations to ensure that appropriate test i
guidance is included in the test and maintenance procedures or the Technical Specifications.
i Response -
We have confirmed that vendor and engineering recommendations have been appropriately incorporated into test guidance and maintenance procedures or the Technical Specifcations for the Reactor Trip Breakers.
This confirmation is based on a review of the following:
o Test and Maintenance Procedures -
In May 1983, General Electric Co. (GE) published a supplement to GE Service Advice 175. This supplement clarified GE's recommended maintenance intervals and modified earlier recommendations for making operational adjustments to their undervoltage trip devices.
The recommendations of the service advice letter have been incorporated into the appropriate test and maintenance procedures at Calvert Cliffs.
j In June,1983, Combustion Engineering, Inc. (CE) published CE Info-Bulletin 83-07 which clarified maintenance details and specified surveillance testing guidelines for trip breakers. The guidelines covered:
settmg up base line data, deciding on an initial test interval, using time response testing to trend breaker problems, and
- modifying. test and maintenance intervals based on trending results.
The recommendations of this Info-Bulletin have been incorporated into the appropriate test and maintenance -
procedures at CCNPP.
o Engineering recommendations
,We have reviewed engineering recommendations that have resulted in modifications to the RTS to verify that all modifications performed have been treated as safety-related.
The Enclosure 3 response to Item 2.1.1 provides a description of the results of our review.
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ENCLOSURE 4 o Surveillance Testing / Technical Specificationjs The current 4
surveillance test requirements of the CCNPP Technical Specifications have been compared to NUREG 0212, CE Standard Technical Specifications. Based on this review, specific changes to the surveillance requirements section of our current Technical Specifications are deemed necessary.
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ENCLOSURE 5 i
RESPONSE TO NRC GENERIC LETTER 83-28 4
4.2 REACTOR TRIP SYSTEM RELIABILITY (Preventative Maintenance and Surveillance Program for Reactor Trip Breakers) 4 4.2.1 NRC Request - Describe the preventative maintenance and surveillance program to ensure a planned program of periodic maintenance, including lubrication, housekeeping and other items recommended by the equipment supplier have been incorporated.
Response - We have confirmed that equipment supplier recommendations have been incorporated into our preventive maintenance and surviellance testing programs with regard to periodic maintenance, lubrication, housekeeping and other items.
This confirmation is based on a review of the following:
o Preventive Maintenance Program - Our current program, which is performed on a quarterly (PM No.1(2)-58-E-Q-1) and annual (PM No.
1(2)-58-E-A-1) basis, consists of the following major elements:
j PM No.1(2)-58-E-Q-1)
- verification that trip shaft torque is in accordance with GE Service Advice 175 and record "as found" and "as lef t" data, and
- lubrication of both the trip shaft bearing and latch roller bearing in accordance with GE Service Advice 175.
PM No.1(2)-58-E-A-1
- performance of an "as found" response time test for both the undervoltage and shunt trip devices, verification of the "as found" pickup and dropout settings on the unvoltage relay in accordance with G.E. Service Advice 175,
- verification of the positive trip _ adjustment on the undervoltage trip device armature in accordance with G.E. Service Advice 175, buffer paddles are' verified rigidly fastened to. the trip j
shaf ts and "as found" and "as lef t" data is recorded, i
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~ ENCLOSURE 5 latch engagement adjustment is verified and "as found" 4
and "as lef t" data is recorded,
- undevoltage trip device to rivet clearance is verified per G.E. Service Advice 175,
- inspection for cleanliness, pitting and mechanical wear of main contacts (stationary and moving), disconnects, arc chutes, breaker cubicles and bus compartment,
- lubrication of breaker components, and
- calibration of the undervoltage relay in accordance with G.E. Service Advice 175.
The annual PM also incorporates the steps provided in the quarterly PM.
o Surveillance Test Program - Our current program, which is performed on a monthly basis, consists of an independent 1
functional test of the shunt trip device and an independent response time test of the undervoltage trip device on all RTS breakers receiving automatic RPS trip signals. The results of any surveillance tests which are unacceptable are reviewed by the POSRC for the purpose of identifying any test results that may indicate unreviewed safety questions. We currently record the results of the "as found" response time data on the undervoltage trip devices for trending purposes.
The intent of this data trending is to provide plant maintenance and engineering personnel with from which they can make qualitative engineering l
judgements regarding replacement / repair of any suspect breakers.
4.2.3 NRC Request - Describe the program developed to perform life cycle testing of the RTS breakers (including the trip attachments).
Response - A life cycle testing program does not currently exist at our facility.
We are evaluating the merits of such a program in light of alternative methods 'which may be equally effective in identifying trip breaker _ failure modes and/or failure rate. In so doing, we have been discussing this issue with industry consultants and owners groups. We do not have any details about the program at this time. We will provide our response to this item before February 29,-1984.
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ATTACHMENT 1 OVERVIEW OF POST-TRIP REVIEW PROCEDURES I
ACTION PERSONNEL AUTHORITY /RESPONSIBILITYI CRITERIA A. Initiation of Shif t Supervisor Following all reactor trips, and Trip review shall be conducted Review prior to reactor startup, shall in accordance with CCI-lll initiate a thorough trip review.
Shall select a reviewer Reviewer normally a Senior License Holder B.
Initial Review Senior License Holder Shall gather data /information and (Response to GL 83-28, Item perform a thorough trip review 1.1.4; See also Attachment 5 with the intent of:
for sources of plant Identifying the root cause(s) information) of the trip.
As a minimum, trip review Ensuring and understanding of shall include review of the the sequence of events following sources of info.:
immediately before and af ter Review of the transient with the trip.
personnel involved using Ensuring that the trends of key available Control Room parameters are understood.
instrumentation and Post incident Critique Reports Review of Sequence of Events Printout; See Att. #2 Review of Post-trip Review Printout; See Att. #3 Review of Alarm Typewriter Printout Review of Technical Support Center Printout; See Att. #4 I Provides Personnel Authorities / Responsibilities in response to GL 83-28 Item 1.1.2
ATTACHMENT 1 OVERVIEW OF POST-TRIP REVIEW PROCEDURES ACTION PERSONNEL AUTHORITY / RESPONSIBILITY CRITERIA C. Post incident Shif t Supervisor Should hold a critique during or Post incident critique shall be Critique Report (or designee) immediately af ter the shif t so held in accordance with GS-O that personnel involved can Standing Instruction 82-1, " Post Everyone with signi-compare and attempt to establish incident Critique".
ficant involvement and document the facts. Pertinent facts in incident (includ-should be written down and collected by ing non-operations the Shif t Supervisor, personnel if appro-priate).
As an alternative to the critique the Shif t Supervisor may direct all personnel with significant involvement in the event to write a narrative describing their observations and actions.
D.
First Post-Trip GS-O Relate present understanding of Meeting to be held within Meeting GS-Maintenance and causes/results of event.
approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Modifications identif y, as appropriate, plant following trip.
modifications / repairs required for restart based upon preliminary information E.
Complete / Submit Senior License Holder
- 1. Complete Post-Trip Review Post-Trip Review checklist Post-Trip checklist shall include:
Review Check-Event date list and Data Event time Trend Analysis Affected Unit Personnel interviewed Sequence of Events reviewed (consistant with operator /
techn:cian observations)
Post-Trip Review printout reviewed Technicial Support center printout reviewed
ATTACHMENT 1 OVERVIEW OF POST-TRIP REVIEW PROCEDURES ACTION PERSONNEL AUTHORITY / RESPONSIBILITY CRITERIA Alarm typewriter printout reviewed (consistency of alarms with events)
Verification of First Post-Trip meeting List of equipment out-of-service which would prevent unit from being returned to service.
List of technical specifica-tions or administrative requirements which would prevent unit from being returned to service.
E.
(Cont'd)
Senior License
- 2. Complete Post-Trip Review Data Trend Analysis shall Complete /Sebmit Holder Data Trend Analysis include review of parameters Post-Trip Review associated with:
Checklist and Data Trend Analysis Reactor Power Pressurizer Pressure Pressurizer Level RCS Flow SuYbolehbar*g!n" Monitor ASI Incore Temperatures Reactor Trip Circuit Breakers Boronometers Feed Pump Discharge Pressure Main Feed Flow
e ATTACHMENT 1 OVERVIEW OF POST-TRIP REVIEW PROCEDURES ACTION PERSONNEL AUTHORITY / RESPONSIBILITY CRITERIA Main Turbine First Stage Pressure Steam Generator Outlet Temperature Condenser Pressure Steam Genertor Pressure 3 team Generator Levels Containment Temperature Containment Pressure Analytical criteria for parameter evaluation and source of parameter data provided in Attachment 2 E.
(Cont'd)
Senior License Holder 3.
Submit Review checklist and POSRC review required Complete /
Shift Supervisor /GS-O Data Trend Analysis to Shif t if any of the information Submit Post-Supervisor /GS-O for review and sources specified in Action E Trip Review of determination of whether above are unavailable.
Checklist and POSRC review is required prior Data Trend Analysis to restart of the unit.
F.
Review for Shif t Supervisor Upon completion of his review, (Response to GL 83-23, Item Restart GS-O the findings should be presented 1.1.1) verbally by the Shif t Supervisor Restart may proceed (with the to the GS-O (designee or Plant permission of the GS-0, or the Superintendent).
Plant Superintendent)if the GS-O and the Shif t Supervisor agree on:
ATTACHMENT 1 OVERVIEW OF POST-TRIP REVIEW PROCEDURES ACTION PERSONNEL AUTHORITY / RESPONSIBILITY CRITERIA The root cause(s) of the trip The sequence of events All alarms printed on the alarm typewriter printout The trends of key parameters if agreement cannot be reached, or if any information sources specified in Action G are unavailable, then POSRC and Plant Superintendent approval must be ontained prior to reactor startup.
G. Independent OL&S Gathe. data, perform independent Response to GL 33-28, Item Review of Trip review of the trip and present 1.1.6)
Data results to the POSRC within 14 Independent Review of all days of the trip.
unscheduled trips shall be made in accordance with CCI-127. Review and presentation to the POSRC need not be completed prior to reactor startup. Written report should be completed as soon as possible, but not necessarily within 14 days from trip.
Report shall summarize results of independent
- review, including:
Brief description of occurrence Identification of root cause(s)
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ATTACHMENT 1 OVERVIEW OF POST-TRIP REVIEW PROCEDURES ACTION PERSONNEL AUTHORITY / RESPONSIBILITY CRITERIA Sequence and consequence of event (including design, hardware, procedural and/or operational deficiencies).
Immediate corrective actions Previous f ailure data Recommended corrective actions i
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ATTACHMENT 2 SEQUENCE OF EVENTS INPUT PARAMETERS
- INPUT PARAMETERS INPUT PARAMETERS CH A PRIMARY COOLANT FLOW LO LOW BEARING OIL PRESSURE CH B PRIMARY COOLANT FLOW LO THRUST BEARING GEAR CH C PRIMARY COOLANT FLOW LO EXHAUST HOOD HIGH TEMP CH D PRIMARY COOLANT FLOW LO MOISTURE SEPARATOR HI LVL CH A STEAM GEN LOW LEVEL LOSS OF CONDENSER VACUUM CH B STEAM GEN LOW LEVEL AUTO STOP FLUID LOW PRESS CH C STEAM GEN LOW LEVEL TURBINE OVERSPEED CH D STEAM GEN LOW LEVEL AUTOSTOP TRIP SOLENOID CH A STEAM GEN LOW PRESS MANUAL TURBINE TRIP P.B.
CH B STEAM GEN LOW PRESS TURBINE LOAD REJECT CH C STEAM GEN LOW PRESS HYDRAULIC FLUID LOW PRESS CH D STEAM GEN LOW PRESS 21A FW HEATER VERY HIGH LVL CH A THERM MARGIN LOW PRESS 21B FW HEATER VERY HIGH LVL CH B THERM MARGIN LOW PRESS 21C FW HEATER VERY HIGH LVL CH C THERM MARGIN LOW PRESS 22A FW HEATER VERY HIGH LVL CH D THERM MARGIN LOW PRESS 22B FW HEATER VERY HIGH LVL CH A TURBINE LOSS OF LOAD 22C FW HEATER VERY HIGH LVL CH B TURBINE LOSS OF LOAD 21 & 22 SG FW PUMP SUCT PRESS CH C TURBINE LOSS JF LOAD 21 SG FW PUMP DISCH PRESS CH D TURBINE LOSS OF LOAD 22 SG FW PUMP DISCH PRESS CH A CONTAINMENT PRESSURE HI TRANS 21386/P21 OR 388/T21 CH B CONTAINMENT PPESSURE Hi TRANS 22 386/P22 OR 388/T22 CH C CONTAINMENT PRESSURE HI GEN DIFF OR GND 386/G-21 CH D CONTAINMENT PRESSURE HI GEN 388/G-28 OR 386/G-2C REACTOR TRIP BREAKER 1 GENERATOR FIELD BREAKER REACTOR TRIP BREAKER 2 UNIT DIFF OR GEN GND 386/U2 REACTOR TRIP BREAKER 3 U TR 534KV LEAD DIFF 386/TH2 REACTOR TRIP BREAKER 4 REACTOR TRIP BREAKER 5 REACTOR TRIP BREAKER 6
- Unit 2 inputs, similar to Unit 1.
ATTACHMENT 2 SEQUENCE OF EVENTS INPUT PARAMETERS INPUT PARAMETERS REACTOR TRIP BREAKER 7 REACTOR TRIP BREAKER 8 MANUAL TRIP MANUAL TRIP MANUAL TRIP CH A PRESSURIZER HI PRESSURE CH B PRESSURIZER HI PRESSURE CH C PRESSURIZER HI PRESSURE CH D PRESSURIZER HI PRESSURE CH A REACTOR HI POWER LEVEL CH B REACTOR Hi POWER LEVEL CH C REACTOR HI POWER LEVEL CH D REACTOR HI POWER LEVEL CH A RATE OF POWER CHANGE CH B RATE OF POWER CHANGE CH C RATE OF POWER CHANGE CH D RATE OF POWER CHANGE TRANS 2 586/512P/ST28 386/P2 RCP BUS FDR BREAKER 252-2201 RCP BUS FDR BREAKER 252-2104 13KV BUS 21 FDR BKR 252-2104 HYDRAULIC FLUID LOW LEVEL AUTO STOP RELAY LOAD UNBAL DEMOD RELAY EHC 21 STEAM GENERATOR HI LEVEL 22 STEAM GENERATOR HI LEVEL AXIAL PWR TRIP CH A AXIAL PWR TRIP CH B AXIAL PWR TRIP CH C AXIAL PWR TRIP CH D REACTOR TRIP BUS LOW VOLTAGE
I ATTACHMENT 3 POST-TRIP REVIEW INPUTS
- i DATA COLLECTION FREQUENCY INPUT STD. 30-SEC.
INCLUDES 2-SEC.
BORONOMETER X
TOTAL COOLANT FLOW X
X 11 STEAM GEN FEEDWATER FLOW X
X 12 STEAM GEN FEEDWATER FLOW X
X CH X PRESSURIZER LEVEL X
X CH Y PRESSURIZER LEVEL X
X CH A 11 STEAM GEN LEVEL X
X CH A 12 STEAM GEN LEVEL X
X CH A REACTOR HI POWER LEVEL X
X PWR RNG SAFETY CH A UPR DETR.
X X
X X
AVG INTERNAL ASI X
CH A INTERNAL ASI X
CH B INTERNAL ASI X
CH C INTERNAL ASI X
CH D INTERNAL ASI X
CH X PRESSURIZER PRESSURE X
X CH Y PRESSURIZER PRESSURE X
X CH A II STEAM GEN PRESSURE X
X CH A 12 STEAM GEN PRESSURE X
X CH A CONTAINMENT PRESSURE X
TURB BEARING LUBE OIL PRESSURE X
X TURB FIRST STAGE PRESSURE X,
X 11 CONDENSER PRESSURE X
12 CONDENSER PRESSURE X
13 CONDENSER PRESSURE X
11 MAIN FW PUMP DISCH PRESS X
X CH A TERM MARGIN SETPOINT X
X IN-CORE ASSY N-ll OUT TEMP X
IN-CORE ASSY R-6 OUT TEMP X
- Unit 1 inputs, identical to Unit 2
l ATTACHMENT 3 l
POST-TRIP REVIEW INPUTS l
L DATA COLLECTION FREQUENCY INPUTS STD. 30-SEC.
INCLUDES 2-SEC.
IN-CORE ASSYB-15 OUT TEMP X
IN-CORE ASSY G-13 OUT TEMP X
IN-CORE ASSY V-15 OUT TEMP X
11 STEAM GEN STM OUT TEMP X
11 MOISTURE SEP. OUTLET TEMP X
IN-CORE ASSY V-19 OUT TEMP X
IN-CORE ASSY C-18 OUT TEMP X
IN-CORE ASSY D-3 OUT TEMP X
IN-CORE ASSY W-4 OUT TEMP X
LOOP 11 T HOT X
LOOP llA T COLD X
LOOP 12 T HOT X
LOOP 12A T COLD X
T AVG FR REAC REG SYS 1 X
T AVG FR REAC REG SYS 2 X
T REF FR REAC REG SYS 1 X
T REF FR REAC REG SYS 2 X
12 STEAM GEN STM OUTLET TEMP X
ATTACHMENT 4 TECHNICAL SUPPORT CENTER INPUTS UNIT SYSTEM PROCESS PARAMETER 1
2_
CONTAINMENT Containment 11 Pressure Indication X
Containment Dome Temp X
X Containment 21 Pressure Indication X
Hydrogen Concentration X
X Containment Sump Water Level X
X EMERGENCY CORE COOLING HPSI Flow to Loop ll A X
HPSI Flow to Loop llB X
HPSI Flow to Loop 12A X
HPSI Flow to Loop 12B X
LPSI Flow to Loop 11A X
LPSI Flow to Loop llB X
LPSI Flow to Loop 12A X
LPSI Flow to Loop 12B X
Containment Spray Header 11 Flow X
Containment Spray Header 12 Flow X
LPSI Flow Control X
HPSI Flow to Loop 21B X
HPSI Flow to Loop 21A X
HPSI Flow to Loop 22B X
HPSI Flow to Loop 22A X
LPSI Flow to Loop 21B X
LPSI Flow to Loop 21 A X
LPSI Flow to Loop 22B X
LPSI Flow to Loop 22A X
Containment Spray Header 21 Flow X
Containment Spray Header 22 Flow X
Charging Pumps Discharge flow X
X Salt Water Pumps Discharge Header Pressure X
X Salt Water Pumps Pressure X
X Component Cooling Pump 11 Discharge Pressure X
Component Cooling Pump 12 Discharge Pressure X
Shutdown Heat-Exchanger 11 Outlet Temp.
X Shutdown Heat-Exchanger 12 Outlet Temp.
X Refuel Water Tank 11 Level X
Service Water Header 11 Pressure X
Component Cooling Pump 21 Discharge Pressure X
ATTACHMENT 4 TECHNICAL SUPPORT CENTER INPUTS UNIT SYSTEM PROCESS PARAMETER
_1_
2_
Component Cooling Pump 22 Discharge Pressure X
Shutdown Heat-Exchanger 21 Outlet Temp.
X Shutdown Heat-Exchanger 22 Outlet Temp.
X Refuel Water Tank 21 Level X
Service Water Header 21 Pressure X
& MAKE-UP Aux. Feedwater Flow Steam Generator 11 X
Aux. Feecwater Flow Steam Generator 12 X
Condensate itorage Tank 12 Level X
Aux. Feedwater Flow Steam Generator 21 X
Aux. Feedwater Flow Steam Generator 22 X
Feedwater Flow to Steam Generator 11 X
Feedwater Flow to Steam Generator 12 X
Feedwater Flow to Steam Generator 21 X
Feedwater Flow to Steam Generator 22 X
Condensate Storage Tank 11 Level X
Condensate Storage Tank 12 Level X
MAIN STEAM Steam Generator Level 11 X
Steam Generator Level 12 X
Steam Generator 11 Pressure X
Steam Generator 12 Pressure X
Steam Generator Level 21 X
Steam Generator Level 22 X
Steam Generator 21 Pressure X
Steam Generator 22 Pressure X
NUCLEAR INSTRUMENTATION
% Power (INTERMEDIATE)
X X
Q-Power X
X Thermal Power X
X
% Power Source Range X
X In-Core Flux Detectors X
X In-Core Thermocouples X
X RADIATION MONITORING Waste Processing Area Radiation Monitor X
X Liquid Waste Discharge Radiation X
X j
Containment High Range (East)
X X
ATTACHMENT 4 TECHNICAL SUPPORT CENTER INPUTS UNIT SYSTEM PROCESS PARAMETER 1
2_
Containment High Range (West)
X X
Noble Gas Main Vent Low Range X
X Noble das Main Vent Mid Range X
X Noble Gas Main Vent High Range X
X Main Vent Flow X
X Condensate Vacuum Pump Discharge Radiation X
X Condensate Vacuum Pump Flow Rate X
X REACTOR COOLANT SYSTEM RCS Flow Loop 11B X
RCS Flow Loop 12B X
Subcooled Margin Loop 11 X
Subcooled Margin Loop 12 X
Pressurizer Level Hot X
X Pressurizer Level Cold X
X Pressurizer Pressure X
X RCS Hot Leg Temp Loop 11 X
RCS Hot Leg Temp Loop 12 X
RCS Cold Leg Temp Loop 11A X
RCS Cold Leg Temp Loop 11B X
RCS Cold Leg Temp Loop 12A X
RCS Flow Loop 21B X
RCS Flow Loop 22B X
Subcooled Margin Loop 21 X
Subcooled Margin Loop 22 X
RCS Hot Leg Temp Loop 21 X
RCS Hot Leg Temp Loop 22 X
RCS Cold Leg Temp Loop 21 A X
RCS Cold Leg Temp Loop 21B X
RCS Cold Leg Temp Loop 22A X
RCS Cold Leg Temp Loop 22B X
l l
t i
ATTACHMENT 5 POST-TRIP REVIEW DATA TREND ANALYSIS AND DATA SOURCES DATA PARAMETER SOURCE
- ANALYSIS REACTOR POWER PTR,SOE 1.
If reactor power increases to 104.5%, verify reactor trip.
TSC 2.
Verify parameter trend is consistent with the event.
PRESSURIZER PRESSURE PTR,SOE,AT 1.
If RCS pressure decreases to TM/LP SETPT, verify low pressure trip.
PTR, SOE, AT 2.
If RCS pressure increases to 2400 PSIA, verify PORV's open.
CRI 3.
Verify PORV's reseat.
PTR 4.
If RCS pressure increases to 2500 PSIA, verify Safeties open.
CRI 5.
Verify Safeties reseat.
PTR 6.
Verify technical specification safety limit of 2750 PSIA is not exceeded.
PTR, AT 7.
If RCS pressure decreases to 1740
- PSIA, verify SIAS actuation.
PTR, TSC 8.
If pressure decreases to 1400 PSIG, verify appropriate HPSI flow.
PTR, TSC 9.
Verify parameter trend is consistent with the event.
PTR, TSC 10.
Verify pressure.E 2225 PSIA prior to trip (T.S. 3.2.5).
PRESSURIZER LEVEL PTR,TSC 1.
Verif y parameter trend is consistent with the event.
If RCS flow 4 95%, verify reactor trip.
PTR, TSC 2.
Verify parameter trend is consistent with the event.
- PTR - Post-Trip Review Printout Sequence of Events Printout 50E Technical Support Center Printout TSC AT
- Alarm Typewriter Printout CRI - Control Room Instrumentation
i ATTACHMENT 5 POST-TRIP REVIEW DATA TREND ANALYSIS AND DATA SOURCES DATA PARAMETER SOURCE ANALYSIS T
PTR,TSC 1.
Verif y temperatures are HOT' COLD consistent with pressurizer level and pressure.
PTR,TSC 2.
Verify temperatures stabilize at approximately 532 F.
PTR,TSC 3.
Verify T prior to tr ph.w.as4548 F C
S 3.2.5).
PTR,TSC 4.
Verify cooldown rates were technical specification limits 4 (100 F/HR when TAVE >
250"F).
PTR,TSC 5.
Verify RCS pressure and temperature are consistent with MPT limits (T.S. Figure 3.4-2a (Unit 1) or 3.4-2b (Unit 2))
T PTR,TSC 1.
Verify consistent with AVG y
HOT COLD SUBCOOLED MARGIN MONITOR TSC, PTR 1.
Verify 30 subcooled margin.
ASI PTR,SOE 1.
If ASI approaches the RPS setpoint, verify reactor trip.
INCORE TEMPERATURES PTR 1.
Verify incore temperature is consisjent with THOT (within 25 F).
REACTOR TRIP CIRCUlT SOE 1.
Verify circuit breaker BREAKERS response times are in accordance with Table 3.3-2 of T.S. 3.3.1.1.
BORONOMETER PTR,TSC 1.
Verify parameter trend is consistent with the event.
p 1
ATTACHMENT 5 l
POST-TRIP REVIEW DATA TRtiND ANALYSIS AND DATA SOURCES DATA PARAMETER SOURCE ANALYSIS
(
FEED PUMP DISCHARGE PTR 1.
Verify parameter trend is l
PRESSURE consistent with the event.
MAIN FEED FLOW PTR,TSC 1.
Verify main feed flow is consistent with S/G level.
PTR,TSC 2.
Verify feed flow decreasing prior to reset.
TSC,PTR 3.
Verify parameter trend is consistent with the event.
MAIN TURBINE FIRST STAGE PTR 1.
Verify parameter trend is PRESSURE consistent with the event.
STEAM GENERATOR OUTLET PTR 1.
Verify S/G outlet temperature TEMP is consistent with S/G pressure.
CONDENSER PRESSURE PTR 1.
Verify parameter trend is consistent with the event.
STEAM GENERATOR PTR,SOE 1.
If S/G pressure decreases to PRESSURE 685 PSIA, verify reactor trip.
PTR,AT 2.
IF S/G pressure decreases to 653 PSIA (U-1), 703 PSIA (U-2J verify SGIS initiation.
PTR 3.
Verify turbine bypass valves operate correctly.
PTR 4.
If S/G pressure increases to 985
- PSIG, verify proper operation of the steam safeties.
PTR 5.
Verify S/G pressure does not l
exceed 1035 PSIG.
l l
STEAM GENERATOR LEVELS PTR,SOE 1.
If S/G level decreases to -50",
l verify reactor trip.
l PTR,SOE 2.
If S/G levelincreases to +50",
I verify main turbine trip.
l
?
ATTACHMENT 5 POST-TRIP REVIEW DATA TREND ANALYSIS AND DATA SOURCES DATA PARAMETER SOURCE ANALYSIS PTR,AT 3.
If S/G level decreases to
-170", verify AFAS actuation.
PTR,TSC 4.
Verify S/G level consistent with feed flow.
CONTAINMENT TEMPERATURE TSC 1.
Verify parameter trend is consistent with the event.
CONTAINMENT PRESSURE PTR,TSC 1.
If containment pressure increases to 2.4 PSIG, verify the reactor trip.
PTR,TSC,AT 2.
If containment pressure increases to 2.8 PSIG, verify CIS and SIAS.
PTR, TSC, AT 3.
If containment pressure increases to 4.25 PSIG, verify CSAS actuation.
.a-4
- e e
j Pr$ Flow I
NO Hr3 Yh/Tc I
Did 3,,ge,yggp Is JS/GPressure re Are List Farsecters Rf'S/ETFAS Oerter ik Event Sta11st Compare Key Any Tech Spee whleh eseeeded fiesgenJ As Accidents
~
'"o a design 6
Actual Trends Parameters W S Pressure
(
Loom At RIY/ESFA.
Safet.a Limits ESFAS or RPS upected YES Agminst Basis Event YE3 with
\\
Maxim e Valy...
Cetreinte os LSSS Paccaded.30 Oetpoints and
?
Expect *1 Trends EV t ressurizer urch g of Parameters /
axecedad (See rection Annunt Exceeled
/
Trends Or 9
2.0 3,
9/1 Levat r
Pirw t e r?
Tech rpac Y
Pm :t4r fewar YES NO
'to Evsluste tvant
'4ska Pecorsar.1s t l ov.
I'0tif7 meument enj h
,iry
\\)Imtim Wil I
Appropriate Investigate WAppropriate 3everity Age nst p
0n Whether New Levels or incons is t enci as.
Management FSAR Analysel Event Chould be h
Mung* wet Levels Aceident Anstyzed Cor sinmarit I ressi.4rtj Imme1intely 1
U4
__J Is Complete go,t,compg,g,3 ea
'lertry All Printouts
,sn Event nocument gew!*w CCI 111 Post-Trip
- Form to 100B0/
)i A tu%
(Cee CCI 111) gre Couna Pa ram *
'Cl 111 Checklist Revgev aske 14-day A
Trends ransistent YM mnalstant with Dat a rmi aai frr>m yea
^' hec k l i s
- Cons htent rFJ rerification presentation t
with specte1 expected trenas of Event e.vgey with )LtS Ferm ran 't mnitored equipment /
and Statements /
Itesialts 7
pirsmeters (other than Interviews 9
tay psrsmaters) y flo i
Leeun-nt and -'
NO Ito i r.cor.sistencies
_tiott fy Investlaste
' Appropriate t=wument Locument Menspament fossible/Frr-bable snd Le vels
- Causes
- Investigste inannafeter.c1.
d Of firEL!NC FOR INIT.TENDENT IOST-TEIP PFVIW ATTAePf4 MT 6 ee...
ATTACHMENT 7 TYPED TREND EQUIPMENT / SYSTEM GROUPINGS
- GROUP NUMBER ASSOCIATED EQUIPMENT OR SYSTEM 1
Main Turbine-Generator Bearing Temperatures 2
Main Turbine-Generator Thrust Bearing, Oil and Metal 3
Pressurizer 4
No.11 Steam Generator 5
No.12 Steam Generator 6
Primary Coolant 7
No. II A Reactor Coolant Pump 8
No.11B Reactor Coolant Pump 9
No.12A Reactor Coolant Pump 10 No.12B Reactor Coolant Pump 11 No.11 FP Turbine Bearing Temperatures 12 No.12 FP Turbine Bearing Temperatures 13 No.11 Aux. FP Turbine Bearing Temperatures 14 No.12 Aux. FP Turbine Bearing Temperatures 15 Nos.11-13 Circ. Water Pumps and Motors 16 Nos.14-16 Circ. Water Pumps and Motors 17 Condensate Pumps and Motors 18 Service Water Pumps 19 Heater Drain Pump Motors 20 Switchyard Volts, Vars, and Amps 21 RAD Waste System 22 Nos.11-12 Condensate Booster Pumps and Motors 23 Nos.13 Condensate Booster Pump and Motor 24 Condenser Vacuum Pumps 25 Component Cooling Pumps and Motors 26 In-core Thermocouples 27 Primary Plant Miscellaneous Parameters 28 Secondary Plant Miscellaneous Parameters 29 Nuclear Fuel Management Miscellaneous Parameters 30 Nuclear Fuel Management Miscellaneous Parameters
- Unit I groupings, similar to Unit 2.
ATTACHMENT 8 BG&E QUALITY ASSURANCE PROCEDURES (QAPs)
THAT APPLY TO ACTIVITIES SPECIFIED IN THE INTRODUCTION TO 10 CFR PART 50, APPENDIX B R E _13_ H H _16 E H H 2_0 E 2_2 2_3 y 26 27_ 28_ 31 M ACTIVITY I
2 3
4 5_
6 7_
9 3
Designing X
X X
X X X Purchasing X
X X
X X X Fabricating X
X X
Handling X
X X Shipping X X Storing X
X X
X Cl:aning X
X Erecting X
X X Installing X
X Inspecting X
X X
X X
X X
Testing X
X X
X X X
X X
X X
X X
X Operating X
X X
X X X X
X X
X X X X X Maintaining X
X X
X X
X X
X Repairing X
X X Rcfueling X X X
X X
X Modifying X
X X
X X
QA Program X
X X
QAP NO.*
TITLE QAP NO.*
TITLE QAP NO.*
TITLE 1
Procedure Control 15 Changes, Tests, & Experiments 27 Testing & Evaluating Materials 2
Procurement & Storage 16 Surveillance Testing 28 Control of Items Covered by 3
Radioactive Waste 17 Control & Calibration of the Quality Assurance Program 4
Reload Core Design & Procure-Measuring & Test Equipment 31 Administrative Controls for ment of Core Components 18 Radiation Safety & Chemistry Operating Nuclear Power Pits.
5 Non-Destructive Examination 19 Environmental Monitoring 35 Control of Technical Support 6
Welding 20 Training
& Plant Engineering 7
Records Management 21 Review & Audit of the Quality 9
Fire Protection Assurance Program 10 Security 22 In-Core Fuel Management 12 Drawing & Technical Manual 23 Out-of-Core Fuel Management Controls 25 Plant Operations 13 In-Service Inspection 26 Control of Conditions Adverse to
- QAP Nos. 8,11,24,29,30 14 Plant Maintenance Quality 32,33,34 do not exist.
-