ML20081J304
| ML20081J304 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/04/1983 |
| From: | William Jones OMAHA PUBLIC POWER DISTRICT |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| GL-83-28, LIC-83-267, NUDOCS 8311080428 | |
| Download: ML20081J304 (79) | |
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l Omaha Pubilt Power District 1623 Harney Omaha. Nebraska 68102 402/536 4000 November 4, 1983 LIC-83-267 l l
l Mr. Darrell G. Eisenhut, Director U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Licensing Washington, D.C. 20555
Reference:
Docket No. 50-285
Dear Mr. Eisenhut:
Requircd Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)
Please find attached the Omaha Public Power District's response to Generic Letter 83-28, dated July '). 1983. This information in-cludes the current status of conformance with the positions con-l tained in Generic Letter 83-28. Schedules are also provided for l the identified improvements.
I l Sincerely, l
kk It~ h W. C. Jones Division Manager Production Operations WCJ/DJM:jmm Attachment cc: LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, D.C. 20036 Mr. E. G. Tourigny, Project Manager Mr. L. A. Yandell, Senior Resident Inspector 8311080428 831104 bO DR ADOCK 05000285 PDR y
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i 45 5124 Employment with Equal Opportunity Male female i-__,____--._- .
pr UNITE 6 STKTES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )
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Omaha Public Power District ) Docket No. 50-285 (Fort Calhoun Station, )
Unit No. 1) )
. . . . . being duly sworn, hereby deposes and says that he is Section Manager - Technical Services of Omaha Public Power District; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached response to issues identified in Generic Letter 83-28 dated July 8, 1983; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, infor-mation and belief.
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R. L. Jaworski Section Manager Technical Services STATE OF NEBRASKA)
) as COUNTY OF DOUGLAS)
Subscribed and sworn to before me, a Notary Fublic in and for the State of Nebraska on this d<a day of November, 1983.
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OMAHA PUBLIC POWER DISTRICT RESPONSE TO GENERIC LETTER 83-28 REQUIRED ACTIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATWS EVENTS fm
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NOVEMBER 4, 1983 0
p-p TABLE OF CONTENTS b
Page NRC Item 1.1 Post-Trip Review 1 NRC Item 1.2 Post-Trip Review - Data and 4 Information Capability NRC Item 2.1 Equipment Classification and 8 Vendor Interface (RPS)
NRC Item 2.2 Equipment Classification and 10 Vendor Interface NRC Item 3.1 Post-Maintenance Testing 16 NRC Item 3.2 Post-Maintenance Testing 18 i NRC Item 4.1 Reactor Trip System Reliability 20 (Vendor Related Modifications)
NRC Item 4.2 Reactor Trip System Reliability 21 (Preventative Maintenance and Surveillance Program for Reactor
() Trip Breakers) 23 NRC Item 4.3 NRC Item 4.4 24 NRC Item 4.5 Reactor Trip System Reliability 25 (System Functional Testing)
Operating Procedure OP-1 Appendix A Standing Order 0-2 Appendix B Standing Order 0-5 Appendix C Plant Outage Report (Form FC-96) Appendix D Introduction and User's Guide From Interim Appendix E Electrical CQE List Generating Station Engineering Mechanical, Appendix F Structural and Electrical CQE Evaluation of Fort Calhoun Nuclear Station O
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n NRC Item 1.1 POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROCEDURE)
U NRC Position Licensees and applicants shall describe their program for ensuring that unscheduled reactor shutdowns are analyzed and that a detemination is made that the plant can be restarted safely. A report describing the program for review and analysis of such unscheduled reactor shutdowns should include, as a minimum:
- 1. The criteria for determining the acceptability of restart.
- 2. The responsibilities and authorities of personnel who will perform the review and analysis of these events.
- 3. The necessary qualifications and training for the responsible personnel.
- 4. The sources of plant infomation necessary to conduct the review and analysis. The sources of information should include the measures and equipment that provide the necessary detail and type of information to reconstruct the event accurately anJ in sufficient detail for proper understanding.
(See Action 1.2).
- 5. The methods and criteria for comparing the event information with known or expected plant behavior (e.g., that safety-related equipment operates as required by the Technical Specifications or other performance specifications related to C) l V the safety function).
- 6. The criteria for determining the need for independent assess-ment of an event (e.g., a case in which the cause of the event
! cannot be poeitively identified, a competent group such as the
! Plant Opern;ons Review Committi:c, will be consulted prior to authorizing restart) and guidelines on the preservation of physical evidence (both hardware and software) to support independent analysis of the event.
! 7. Items 1 through 6 above are considered to be the basis for the establishment of a systematic method to assess unscheduled reactor shutdowns. The systematic safety assessment proce-dures compiled from the above items, which are to be used in conducting the evaluation, should be in the report.
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pa District Response The Trip Recovery portion of the Fort Calhoun Station Operating Procedure -
OP-1, dated July 13,1983, (Appendix A) requires the following:
- 1. Verification that the Reactor Protective System (RPS) properly initiated an automatic reactor trip;
- 2. Detemination if a manual trip preceded an automatic trip or followed an automatic trip or was not initiated;
- 3. Detemination of which RPS channel caused the trip;
- 4. Detemination of the cause of the RPS activation;
- 5. Attachment of a copy of the " Sequence of Events" log to the Trip Recovery procedure; and
- 6. ierification that the cause of the reactor trip is corrected.
The Supervisor-0perations reviews and approves with his signature the above items.
Standing Order 0-2 (Appendix B) on the Requirements for Criticality -
requires that permission be obtained from the Manager-Fort Calhoun Station prior to taking the reactor critical. The Supervisor -
(g',) Operations may act as an alternate. Operating Procedure OP-1, " Master Checklist for Start-up or Trip Recovery" provides the guidance by which the station is restarted.
After the Plant % nager gives pemission and final approval, the Shift Supervisor authorizes the reactor to be taken critical.
The Plant Manager as per Technical Specification 5.1.1 shall be responsi-ble for overall facility operation. The Supervisor - Operations is responsible specifically for plant operations. Shift Supervisor responsi-bilities are described in the attached copy of Standing Order 0-5 (Appen-l dix C).
ANSI 18.1-1971 requires that the Plant Manaaer have ten years of responsi-ble power plant experience of which three years shall be nuclear power plant experience. An SR0 license is not required, but the Plant Manager shall have the background required to sit for an examination.
ANSI 18.1-1971 requires the Supervisor-Operations to have eight years of responsible power plant experience of which three years shall be nuclear power plant experience. An SR0 license is required.
The Shift Supervisor must have power plant experience similar to the Supervisor-Operations and must also hold a current SR0 license. The Fort Calhoun Station Updated Safety Analysis Report provides the necessary The f) qualifications and pertinent information for essential plant staff.
v specific personnel infomation is accurate as of January 22, 1983.
Specific personnel have changed in some instances since that time, but the qualification requirements are still applicable. ,
O Several sources of plant information are necessary and available to the V Operations staff to conduct post-trip review and analysis. Instrumen-tation, either local or in the control room, together with direct operator observation aid in the plant transient analysis. The major source of information is the plant computer system which has four major functions to aid in post-trip analysis:
- 1. Sequence of Events;
- 2. Post Trip Review;
- 3. Alarm Typewriter;
- 4. Trend Typewriter.
Other very important sources of information are the analog trend recorders for various parameters.
All licensed Operations personnel, either assigned to shift or super-visory duties, have years of nuclear power plant experience. The fact that these personnel are R0 or SRO qualified gives proof of competence.
These individuals' experience and training also combine to give them the necessary ability, by directly observing recorded data and instrumenta-tion, to interpret plant behavior and recognize abnormal plant para-meters.
As stated above, the Plant Manager authorizes restart of the reactor after each trip. The Trip Recovery portion of Operating Praccdurc 0?-l fm also specifies the criteria for restart and requires preservation of the
() " Sequence of Events" log to support independent analysis of the event.
The Supervisor-0perations is responsible for completing a forced outage report (FC-96) fom ( Appendix D) which gives an explanation of circum-stances surrounding the outage or reactor trip and subsequent recovery.
An accumulation of all event-related data, reactivity work sheets, plots, procedures, computer print-outs, etc. are filed for every outage / trip /
start-up.
As requested by Iten 1.1.7, the following appendices are attached to this report:
A. Operating Procedure OP-1 B. Standing Order 0-2 C. Standing Order 0-5 D. FC-96 (Forced Outage Report Form)
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NRC Item 1.2 POST-TRIP REVIEW - DATA AND INFORMATION CAPABILITY NRC Position Licensees and applicants shall have or have planned a capability to record, recall and display data and information to permit diagnosing the causes of unscheduled reactor shutdowns prior to restart and for ascertaining the proper functioning of safety-related equipment.
Adequate da'.a and information shall be provided to correctly diagnose the
' cause of u'. scheduled reactor shutdowns and the proper functioning of safety-re'ated equipment during these events using systematic safety assessment procedures (Action 1.1). The data and information shall be displayed in a fonn that permits ease of assimilation and analysis by persons trained in the use of systematic safety assessment procedures.
A report shall be prepared which describes and justifies the adequacy of equipment for diagnosing an unscheduled reactor shutdown. The report shall describe as a minimun:
- 1. Capability for assessing sequence of events (on-off indications)
- 1. Brief description of equipment (e.g., plant computer, dedicated computer, strip chart)
- 2. Parameters monitored
- 3. Time discrimination between events
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- 4. Format for displaying data and information
- 5. Capability for retention of data and information.
- 2. Capability for assessing the time history of analog variables needed to determine the cause of unscheduled reactor shutdowns, and the functioning of safety-related equipment.
- 1. Brief description of equipment (e.g., plant computer, dedicated computer, strip charts)
- 2. Parameters monitored, sampling rate, and basis for selecting parameters and sampling rate
- 3. Duration of time history (minutes before trip and minutes
' aftertrip)
- 4. Format for displaying data including scale (readability) of time histories
- 5. Capability for retention of data, information and physical evidence (both hardware and software)
- 3. 0ther data and infonnation provided to assess the cause of unscheduled reactor shutdowns.
- 4. Schedule for any planned changes to existing data and information capability.
District Response General The District currently has the capability to record, retrieve, and display data and information to permit diagnosing the causes of unscheduled reactor shutdowns and for ascertaining the proper functioning of safety-related equipment. The District is also installing an Emergency Response Facilities 1 (ERF) computer system which will enhance this capability.
Descriptions of the existing and improved systems are provided i
along with an implementation schedule for the improved system.
EXISTING SYSTEM The sequence-of-events is monitored and recorded by the PRODAC-250 plant computer using digital inputs from the Reactor Protective System (RPS) and trip contactors. The parameters monitored for O se9eemce-or-eve #ts recoroi 9 4#c'eee: eecw reector Protective channel (A, B, C and D) of reactor power level, rate of change of power, reactor coolant flow, water level steam generator tio.1, water level steam generator No. 2, steam pressure steam generator No. 1, steam pressure steam generator No. 2, high pressurizer pressure, thenaal margin / low r essure, loss of load, high contain-nient pressure, and axial powei distribution. The status of the RPS trip contactors and manual trip contactors is also monitored. The sequence-of-events record provides a time discrimination of one cycle (1/60 of a second) between events. The record is printed by a Tally printer. The printout has a header and lists each event and the number of cycles after the initial event when the event occurred. The printed record is retained as part of post-trip review (see response to Item 1.1) and the control room records.
The PRODAC-250 is powered by either of the DC vital buses at Fort Calhoun Station.
The analog variables are recorded on strip charts. The reactor power level (as measured by the excore detectors) pressurizer pressure, containment pressure, steam generator level, steam genera-tor pressure, and wide range cold leg temperature are monitored on a continuous basis. The reactor coolant system hot leg and cold leg temperatures may be monitored using the analog trend feature of the plant computer and one of the analog trend strip chart record-The ers. The temperatures are sampled once every ten seconds.
parameters recorded on the dedicated scrip charts are recorded
A continuously. The typical speed of the strip chart recorders is U three inches per hour and the parameter is typically recorded over its full instrument range. These records do allow a determination of the value of a parameter at the time of trip. The strip charts are retained as part of the control room records. The strip charts are powered by either of the DC instrument buses.
The PRODAC-250 plant computer can provide additional data through digital trend logs and the alarm printer log. User specified parameters may be placed on a digital trend log with a user specified time interval. If a parameter exceeds the alarm setpoint this infonnation will be recorded on the alarm printer 109 There are no plans to change the existing system because it will be superceded by the improved system.
IMPROVED SYSTEM The sequence of events will be monitored and recorded by the ERF computer system using digital inputs from the RPS and trip con-tactors. The ERF computer system consists of Data Acquisition System (DAS) computers running in parallel, ERF computers running in parallel, and associated multiplexors, displays, and interfaces.
The DAS computers are MODCOMP 7821 machinu, and the ERF computers are MODCOMP 7870 machines.
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(# The parameters to be monitored by the ERF computer system for sequence-ofevents recording are the same as those for the existing system. The ERF computer system provides a time discrimination of 0.1 second between events for sequence-of-events recording. The sequence-of-events will be automatically printed on a high speed line printer. The report is titled with each event and the corre-sponding time of occurrence being printed on a separate line. The oldest event is printed first, the newest last. The printed record i will be retained as part of the control room record. In addition, the sequence-of-events log is automatically stored on disk for subsequence operator-requested logging and/or tape archival. The ERF computer syste:a is powered by an uninterruptible power supply as part of the Technical Support Center (TSC) power supply system.
The equip;nent used for assessing the time history of analog vari-ables is the same as that used to monitor the sequence-of-events with the improved system.
The selection of analog variables to be monitored by this system has not been completed. The system has the capability to monitor and record 600 analog points and 250 digital points in a mode that can be used for post-trip review. The parameters will be selected on the basis that they are an input to the RPS and/or Engineered Safety Features (ESF) systems or that they are a variable to be monitored to assure that safety functiors are being maintained in accordance with the Combustion Engineering Owners Group (CE0G)
) Emergency Procedure Guidelines.
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i P _ The system has the capability to record the value of each point for
- i two hours at 30-second intervals prior to the trip, for three
- minutes after the trip at one-second intervals, for 30 minutes after the trip at 10-second intervals, and for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the trip at one-minute resolution. The recording of data may be terminated by the operator 30 minutes following the trip.
The data may be displayed using any output function of the ERF computer. ERF data display functions include 1) logs which display point values, either on CRT's or line printers, as a function of time; 2) CRT strip chart emulators with a variable time interval; and 3) CRT x-y plots with user-defined axes. The CRT output may be placed on hard copy via a video copier. The ERF computer system has the capability to manipulate the data such that it can be displayed in a readable format.
The printed record will be retained as part of the control roan
! records. The data is stored on disk for subsequent operator-requested logging and/or tape archival. The ERF computer system is powered by an uninterruptible power supply which is part of the TSC
] power supply system. The strip chart recordings that are displayed as part of the existing system will provide additional data.
The ERF computer is scheduled to be operational after the Cycle 9 refueling, currently scheduled for the fall of 1985. Inputs necessary to fulfill the analog variable post-trip review function i
() are scheduled to be installed prior to or during the Cycle 10 refueling presently scheduled for fall of 1986.
However, if additional inputs are identified during a systematic verification of data required to perfonn a post-trip review, the
! system may not be fully operational until the refueling following ,
Cycle 11, which is currently scheduled for the fall of 1987. The inputs necessary for the sequence-of-events recording by the ERF computer will not initially be input to the ERF computer. These inputs will be transferred from the P-250 computer during the j refueling following either Cycle 11 or Cycle 12, currently scheduled for the fall of 1987 or 1988, respectively.
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NRC Item 2.1 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (REACTOR TRIP V SYSTEM COMPONENTS)
NRC Position Licensees and applicants shall 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 mainte-nance, work orders, and parts replacement. In addition, for these compo-nents, licensees and applicants shall establish, implement and maintain a continuing program to ensure that vendor information is complete, current and controlled throughout the life of the plant, and appropriately refer-enced 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 infor-mation 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 non-nuclear divisions of their vendors that provide service on reactor trip system components to assure that requisite control of and applicable instructions for maintenance work are provided.
District Response For the District's purposes, safety-related is defined as: those struc-tures, systems, components or items whose satisfactory performance is required to prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. The Omaha Public Power District uses the term " Critical Quality Element" (CQE) to denote such items.
The District currently utilizes the " Interim Electrical CQE List" to identify components of the Reactor Protective System (RPS) as safety-related. The District believes this list accurately identifies all l components of the Reactor Protective System as safety-related. Addi-tional assurance of the completeness of this list will be accomplished in !
conjunction with implementation of the NPRDS system and a new computer- l ized maintenance system. As part of the system implementation, a list of j Reactor Protective System components is being generated. This list will be compared with the Interim Electrical CQE List. The confirmation pro-cess will be completed by October 1,1984.
Additionally, the District plans to review applicable procedures and surveillance tests to insure that these documents adequately control activities regarding the Reactor Protective System components. The i
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' review will be done in conjunction with the work required by 2.2.1.3.
This work will be completed by September 1,1984.
i Vendor Information Status of Existing System The District does not currently pursue all vendors of its CQE equipment to determine if we hold the latest revision of technical manuals or if any modifications have been recommended. The District has relied on those vendors (i.e. the NSSS vendor) and others who have a modification program and the NRC's Notice, Circular, and Bulletin system to obtain hardware data. The District also participates in NUS's NOMIS program for interaction with other utilities.
Additional Work and Schedule Vendor information and manual updates for the reactor protective system will be accomplished and the guidelines of NRC positions 2.1 and 2.2 will be considered in this process. The District plans to complete both items as one task with higher priority given to the Reactor Protective System.
The District is an active participant in the INP0 "NUTAC on Generic Letter 83-28, Section 2.2.2". Through this program, a workable solution to the vendor interface problem is being pursued. The sch'eduled output date for this NUTAC is February 1,1984. Based upon review of the NUTAC document, the District will determine to what extent the findings of this NUTAC can be incorporated into District practices for both the RPS and other safety-related syster.is.
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q NRC Item 2.2 EQUIPt1ENT CLASSIFICATION AND VENDOR INTERFACE (PROGRAF 4S FOR ALL SAFETY-RELATED C0ftP0NENTS)
V NRC Position Licensees and applicants shall submit, for staff review, a description of their programs for safety-related equipment classification and vendor interface as described below:
2.2.1 For equipment classification, licensees and applicants shall describe their program for ensuring that all components of safety-related systems necessary for accomplishing required safety functions are identified as safety-related on docu-ments, procedures, and information handling systems used in the plant to control safety-related activities, including maintenance, work orders and replacement parts. This descrip-tion shall include:
NRC Item 2.2.1.1 The criteria for identifying components as safety-related within systems
! currently classified as safety-related. This shall not be interpreted to require changes in safety classification at the systems level.
District Response Appendix E and Appendix F provide the criteria and supporting information for establishing safety-related electrical and mechanical equipment and structures respectively.
hs NRC Item 2.2.1.2 A description of the information handling system used to identify safety-related components (e.g. , computerized equipment list) and the methods used for its development and validation.
District Response The Omaha Public Power District's present methods for identifying safety-related components involve the proper utilization and application of the five documents listed below. A description of the methods for develop-ment and validation of each is also provided.
- 1) The Interim Electrical CQE (Critical Quality Element) List.
Appendix F is an excerpt from this document and contains the methods for development of the Inter m Electrical CQE List. This i
list was validated originally as part of the initial effort to compile the list.
- 2) The Station Piping and Instrumentation Diagrams. Appendix G, an excerpt from the ilechanical CQE List, provides the methods and ,
criteria used for categorizing mechanical and structural items as safety-related.
(3 3) The Station Structural Drawings. Use of these drawings in conjunc-
'V tion with the guidelines of Appendix G, provides the methodology for detemining the safety-related status of structural components.
- 4) The Fort Calhoun Station Unit No.1 Technical Specifications.
These Technical Specifications have been approved by the NRC. They provide a valuable source of information in classifying systems as safety-related.
- 5) The Fort Calhoun Station Unit No.1 Updated Safety Analysis Report (USAR). The USAR serves as a valuable reference source and support document to the Fort Calhoun Unit No.1 Technical Specifications.
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This document was approved by the NRC.
Items 1, 2, and 3 are updated as the need arises. This process is part of the District's Generating Station Engineering Procedure A-9, " Document Control".
The Fort Calhoun Station Unit NO.1 Technical Specifications were devel-oped and validated as part of the original FSAR. Revision to the Tech-nical Specifications requires the approval of the Plant Review Committee and concurrence of the District's Safety Audit and Review Committee before any change to the Technical Specifications can be submitted to the NRC. The Fort Calhoun Station Unit No.1 Updated Safety Analysis Report (USAR) was prepared as an update to the Fort Calhoun Station Final Safety Analysis (FSAR). Additionally, in accordance with 10 CFR 50.71(e), this p)
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document is updated annually. These updates also must be reviewed by the Plant Review Committee, the Safety Audit and Review Committee, and the Licensing Department prior to issuance.
The District is in the process of implementing a program to provide The same computerized maintenance control and equipment history.
criteria as identified in Appendix E and Appendix F will be used to identify the safety-related (CQE) equipment for this system.
An independent validation of the equipment identified as safety-r21ated (CQE) in the coaputerized system will be made by the District's Genera-ting Station Engineering Section. This validation is scheduled to be completed by October 1,1984.
NRC Item 2.2.1.3 A description of the process by which station personnel use this infoma-i tion 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.
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District Response pd Omaha Public Power District personnel utilize the documents discussed in Item 2.2.1.2 as required by procedures, Station Standing Orders, Fort Calhoun Station Operating Manual, Quality Assurance Department Manual, Purchasing Manual, Generating Station Engineering Manual, and Technical Services Manual . Collectively, these documents define programs, record handling systems, administrative controls and procedures to permit District personnel to perform necessary plant functions and maintain a high level of quality at all times. Included in these functions are maintenance, preventive maintenance, testing, modifications, purchasing, records, requirements, audits, equipment storage, reviews and apprcvals.
Collectively, these processes contain controls to ensure that safety-related equipment is identified as such and handled in an appropriate manner.
Fort Calhoun Station Standing Order G-17 " Maintenance Order" requires that a determination be made of whether or not the maintenance is to be done on a safety-related system or component. If a determination is made that an activity involves a safety-related component, station procedures require implementation of appropriate actions to assure that station quality standards are maintained. Specifically, quality assurance and/or quality control must be involved, and operability requirements per appli-cable Technical Specifications must be met and approved by the Shift Supe rvi sor. Any applicable Technical Specifications must also be O referenced. The craftsman must designate either the purchase order number or requisition on stores number for any safety- related part or material used.
The Preventive Maintenance Program is controlled by Standing Order M-2.
A specially denoted preventive maintenance card is used to identify work which must be done without exception and includes safety-related items.
The Plant Review Committee must approve any changes to Standing Order M-2.
Surveillance Testing requirements are controlled by Station Standing Order G-23. The Surveillance Testing program is used to meet the surveil-lance requirements of the Fort Calhoun Station Unit No.1 Technical Speci-fications. This program requires that safety-related equipment be tested on a periodic basis to verify its operational readiness. If a failure occurs during testing, a maintenance order is generated and a retest will be perfomed as required. All surveillance tests are to be performed in accordance with written procedures approved by the Plant Review Commit-tee. Surveillance tests are reviewed in accordance with plant proce-dures.
Station Modifications are controlled by Standing Order G-21. This Standing Order defines how a modification can be made at the Fort Calhoun Station. It assigns responsibility and defines format and controls.
This Standing Order interfaces with the Technical Services Manual and the GSE Manual.
a p As part of the design, safety-related components are identified. Safety-V related installations are controlled at the station by a SRDC0 (Safety Related Design Change Order). Standing Order G-21 also covers document update and training.
The identification of safety-related equipment for the purposes of parts replacement must be made on the appropriate purchasing documents. Addi-tional guidance is provided in the District Purchasing Manual. The Purchasing lianual requires identification of safety-related (CQE) items and the necessary quality data must be present. The documents are reviewed by QA to insure the necessary quality information is included.
Receipt and storage of safety-related equipment is controlled by station Standing Orders G-22, G-24 and G-25.
Additional Tasks and Schedule To insure established controls are adequate and do not present any ambiguities, the District plans to review the Standing Orders and the other discussed documents and make any changes should a need be identifi ed. Also, the District will review Preventive Maintenance procedures and make any required changes (i.e. make these procedures Maintenance Procedures covered by the Maintenance Order System). This review effort will be completed by October 1,1984.
NRC ltem 2.2.1.4 A description of the management controls utilized to verify that the procedures for preparation, validation and routine utilization of the infonnation handling system have been followed.
District Response llanagement controls to verify the proper preparation, validation, and use of the CQE list are on two levels. These two levels are (1) direct man-agement interaction with the day-to-day procedures and (2) independent audits to verify compliance with the various District-governing docu-me nts.
As can be seen in section 2.2.1.3, the four areas of maintenance, surveil-lance testing, station modification, and purchasing are adequately con-trolled. Each of these areas has included in the governing procedures (also discussed in 2.2.1.4) required involvement of District supervisory New and management personnel in the review cycle to insure compliance.
surveillance test procedures are reviewcd by the PRC and approved by the Plant Manager. Modification Requests require both Generating Station Engineering (GSE) management and plant staff review. Purchasing requires quality review and management approval .
Independent audits serve to reinforce the management controls. Audits pV are performed by QA, the SARC (Safety Audit and Review Committee), INPO, American Nuclear Insurers and the NRC. These provide management with information to judge compliance with controlling documents and proper application of these documents.
NRC Item 2.2.1.5 A demonstration that appropriate design verification and qualification testing is specified for procurement of safety-related components. The specifications shall include qualification testing for expected safety service conditions and provide support for the licensees' receipt of testing documentation to support the limits of life recommended by the supplier.
District Response The District has defined requirements for purchasing in the Purchasing Manual. The individual initiating the purchase order is responsible for identifying the quality (qualification requirements) data necessary.
These purchasing documents are reviewed by QA and appropriate supervisory and management personnel . Appropriate specifications are included with the purchasing document (s).
For electrical equipment located in a harsh enviroment the District is O complying with 10CFR50.49 by the guidelines outlined in Standing Order G-17A. As part of this work, the District will also implement a quali-fled life program by December 1,1983 for harsh environment electrical equipment.
NRC Item 2.2.2 For vendor interface, licensees and applicants shall esta-blish, implement and maintain a continuing program to ensure that vendor infomation 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 equip-ment 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 sufficient attention is paid to equipment mainte- ,
nance, replacement, and repair, to compensate for the lack of vendor backup, to assure reliability commensurate with its safety function (GCC-1). The program shall be closely coupled with action 2.2.1 above (equipment qualificaticn). The pro-gram 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 tech-n)
% nical mailings. It shall also define the interface and divi-sion of responsibilities among the licensee and the nuclear and non-nuclear divisions of their vendors that provide service on safety-related equipment to assure that requisite control of and applicable instructions for maintenance work on safety-related equipment are provided.
District Response The District does not currently pursue all vendors of its CQE equipment to determine if we hold the latest revision of technical manuals or if any modifications have been recommended. The District has relied on those vendors (i.e. the NSSS vendor) and others who haya a modification program and the NRC's Notice, Circular, and Bulletin systen! to obtain hardware data. The District also participates in NUS's NOMIS program, INP0's SEE-IN, and IMP 0's NOTEPAD for interaction with other utilities.
The District is an active participant in the INP0 "NUTAC on Generic Letter 83-28, Section 2.2.2" Through this program, a workable solution to the vendor interface problem is being sought. The scheduled output date for this NUTAC is February 1,1984. As was previously stated, at that time the District will determine to what extent the findings of the NUTAC can be incorporated into District practices.
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p NRC Item 3.1 POST-MAINTENANCE TESTING (REACTOR TRIP SYSTEM COMPONENTS)
U NRC Position The following actions are applicable to post-maintenance testing:
3.1.1 Licensees and applicants shall submit the results of their review of test and maintenance procedures and Technical Specifications to assure that post-maintenance operability testing of safety-related components in the reactor trip system is required to be conducted and that the testing demonstrates that the equipment is capable of perfonning its safety functions before being returned to service.
District Response The plant maintenance procedures on the reactor trip system are approved by the Plant Review Committee and contain provisions to test that portion of the reactor trip system on which the maintenance was perfonned prior to returning it to service. The testing demonstrates that the component can perfonn its required function. For special procedures written to perform maintenance on the Reactor Protective System, Standing Order G-19 requires that the component be tested as deemed necessary by the Techni-cal Supervisor to verify proper operability before it is declared fully operatio nal . References cited in Technical Specification 5.8.1 also require post-maintenance testi ng.
3.1.2 Licensees and applicants shall submit the results of their O
V check of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or Technical Specifications, where required.
District Response Results of the District's check of vendor and engineering recommendations resulted in the writing of two new maintenance procedures and a surveil-lance test.
Maintenance procedure MP-RPS-Breaker includes manufacturer recommen-dations for preventive maintenance on the clutch power supply circuit breakers. The circuit breakers will also be cycled several times on a periodic basis as recommended by the manufacturer by use of surveillance test ST-RPS-10. Maintenance procedure MP-RPS-M-Contactor was written for the "M" contactors which are in the reactor protective system clutch power supply circuit.
3.1.3 Licensees and applicants shall identify, if applicable, any post-maintenance test requirements in existing Technical Specifications which can be demonstrated +o degrade rather '
thal enhance safety. Appropriate changes to these test requirements, with supporting justification, shall be sub-mitted for staff approval.
District Response No post-maintenance test requirements in the Technical Specifications have been found which degrade rather than enhance safety.
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NRC Item 3.2 POST-MAINTENANCE TESTING (ALL OTHER SAFETY-RELATED (n/ EQUIPMENT)
NRC Position The folicwing actions are applicable to post-maintenance testing:
3.2.1 Licensees and applicants shall submit a report documenting the extending of test and maintenance procedures and Technical Specifications review to assure that post-maintenance opera-bility testing of all safety-related equipment is required to be conducted and that the testing demonstrates that the equip-ment is capable of perfonaing its safety functions before being returned to service.
District Response The present system at the Fort Calhoun Station for controlling post-maintenance operability testing is described in Plant Standing Orders.
This insures the system requirements of the Technical Specifications are translated into the necessary controls, documents, testing, and mainte-nance for safe plant operation.
In general, post-maintenance testing will use a surveillance test or portion of a surveillance test to demonstrate operability. There are,
,, however, cases when, because of the nature of the breakdown and plant status, a surveillance test cannot be performed. In these cases the
() District will generate test procedures to verify equipment operability using the Standing Orders as guidance documents.
l The District believes that the present system at the Fort Calhoun Station adequately controls post-maintenance testing. To insure full compliance, the District plans to review the plant Standing Orders to insure that the requirements for post-maintenance testing are clearly defined. Secondly, the District will review all repetitive maintenance procedures to insure requirements for post-maintenance testing are defined. Thirdly, the District plans to review all preventive maintenance procedures to insure proper control, procedure classification and required testing is defined.
The District also plans to review all special procedures.
! This effort will be part of the review to be completed October 1,1984 as part of 2.2.1.3.
3.2.2 Licensee and applicants shall submit the results of their check of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications where l requi red.
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District Response The District presently has Standing Order G-36 which requires periodic review of operating manual procedures. This Standing Order will be reviewed to insure all necessary procedures are included and will also be updated to include vendor and engineering information in the required review.
Schedule Standing Order G-36 will be reviewed and updated as appropriate by September 1, 1984. The initial procedure review will be done in parallel with the vendor information effort established in the response to items 2.1 and 2.2. As was previously noted, this effort may incorporate the programs developed by the INP0 "IlVTAC on Generic Letter 83-28, Section 2.2.2." and will be scheduled accordingly.
3.2.3 Licensees and applicants shall identify, if applicable, any post- maintenance test requirements in existing Technical Specifications which are perceived to degrade rather than enhance safety. Appropriate changes to these test require-ments with the supporting justification, shall be submitted for staff approval .
District Response
'.*e have identified no existing Technical Specification post-maintenance testing requirements which degrade rather than enhance safety.
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NRC Item 4.1 REACTOR TRIP SYSTEM RELIABILITY (VENDOR-RELATED MODIFICATIONS)
O NRC Position All vendor-recommended reactor trip breaker modifications shall be reviewed to verify that either: 1) each modification has, in fact, been implemented; or 2) a written evaluation of the technical reasons for not implementing a modification exists.
For example, the modifications recommended by Westinghouse in NCD-Elec-18 for the DB-50 breakers and a March 31, 1983, letter for the DS-416 breakers shall be implemented or a justification for not implementing shall be made available. Modifications not previously made shall be incorporated or a written evaluation shall be provided.
District Response The District has requested that the vendors investigate the equipment history and indicate to the District any recommended modification. The District will then evaluate the recommendations and make modifications if required. The schedule will be provided to the NRC when vendor informa-tion becomes available.
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NRC Item 4.2 REACTOR TRIP SYSTEM RELIABILITY (PREVENTATIVE MAINTENANCE AND SURVEILLANCE PROGRAM FOR REACTOR TRIP BREAKERS)
Description of the District's Present Program The District has in place at the Fort Calhoun Station both a preventive maintenance and surveillance test program to insure proper operation of the Reactor Protective System breaker (contactor-CEDM clutch power inter-rupt system) (reactor trip system). Automatic trip is accomplished by interruption of the control power to the operate coil of four Allen Bradley flodel 702 contactors which in turn interrupt power to the CEDM clutches. Manual trip is accomplished by either de-energizing the same A-B contactors or actuation of an undervoltage trip on the Westinghouse JA molded case circuit breaker (in series with two of the A-B contactors) via a diverse manual trip pushbutton, located on the Reactor Protective System Panel.
The preventive maintenance is based on manufacturer's recommendations for both the contactor and breaker. The two procedures are controlled by plant standing orders (G-17 and M-2) and are scheduled to be performed on a periodic basis (each refueling outage).
The surveillance test program at the Fort Calhoun Station tests both the manual and automatic trip functions. The automatic trip function is tested under ST-RPS-11. The test, performed on a monthly frequency, cycles the A-B contactors one at a time on-line. The manual trip circuit test, ST-RPS-10 is perfonned prior to each startup, if not done the previous week, (cycling of the breakers is done prior to a startup after refueli ng) .
NRC Position Licensees and applicants shall describe their preventative maintenance and surveillance program to ensure reliable reactor trip breaker opera-tion. The program shall include the following:
4.2.1 A planned program of periodic maintenance, including lubri-cation, housekeeping, and other items recommended by the equipment supplier.
District Response f
l The District currently has a periodic maintenance program in place (including the latest vendor maintenance information) as described in the preceding " Description of Present Program" section. The procedures include housekeeping, lubrication (or directions not to lubricate), and other vendor recommendations. The District plans no changes to this systen unless vendor information or experience dictate otherwise.
I 4.2.2 Trending of parameters affecting operation and measured during testing to forecast degradation of operability.
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i District Response The District presently has an equipment history file (Standing Order t1-3) which allows review of maintenance on equipment. Also, the District is presently in the process of upgrading its maintenance program, operating history and beginning participation in the ftPRDS program.
It should be noted that the District is not trending parameters on the
- RTS switchgear. The manufacturers do not recommend any detailed testing or data measurement to insure operability. This will be discussed in more detail in the District's response to positions 4.2.3 and 4.2.4.
4.2.3 Life testing of the breakers (including the trip attachments) on an acceptable sample size.
4.2.4 Periodic replacement of breakers or components consistent with demonstrated life cycles.
District Response Item 4.2.3 and 4.2.4 deal with the establishment and maintenance of a qualified life for the Reactor Protective System switchgear (Allen Bradley Contactors, Westinghouse fiolded Case Circuit Breakers).
The Reactor Protective System design for Fort Calhoun Station is substan-tially different from the design of the other reactor protective systems Q
D which have experienced reliability problems. The demonstrated relia-bility coupled with a comprehensive preventive maintenance program, a surveillance testing program, and the environment in which the equipment is located provide a high level of assurance that the Fort Calhoun Station Reactor Protective System will continue to be reliable.
The District has contacted the switchgear vendors in an attempt to obtain any reliability, aging, mean time between failure, or maximum operating cycles information. If information is obtained, appropriate changes will be implemented. The District plans no further action on items 4.2.3 and 4.2.4.
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NRC Item 4.3 District's Response J
This item is not applicable to the Fort Calhoun Station.
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NRC Item 4.4 District's Response This item is not applicable to the Fort Calhoun Station.
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NRC Item 4.5 REACTOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING)
NRC Position On-line functional testing of the reactor trip system, including independent testing of the diverse trip features, shall be perfomed on all plants.
4.5.1 The diverse trip features to be tested include the breaker under-voltage and shunt trip features on Westinghouse, B&W (see Action 4.3 above) and CE plants; the circuitry used for power interruption with the silicon controlled rectifiers on B&W plants (see Action 4.4 above); and the scram pilot valve and backup scram valves (including all initiating circuitry) on GE plants.
District Response The Fort Calhoun Station Reactor Protective System automatic trip does not have a diverse trip feature. The contactor coil is de-energized on reactor trip and the contactors drop out. The contactors are tested one at a time, monthly on-line.
The manual trips are accomplished either by de-energizing the contactors or dropping out an undervoltage trip device (via a diverse manual trip) on the breakers. These are tested each startup (unless tested within the preceeding five (5) days). The District believes this is an adequate O surveillance program.
4.5.2 Plants not currently designed to permit periodic on-line testing shall justify not making modifications to permit such testi ng. Alternatives to on-line testing proposed by licen-sees will be considered where special circumstances exist and where the objective of high reliability can be met in another way.
District Response The Fort Calhoun Station is capable of on-line testing of the Reactor Protective System; no further action is required.
4.5.3 Existing intervals for on-line functional testing required by Technical Specifications shall be reviewed to determine that the intervals are consistent with achieving high reactor trip system availability when accounting for considerations such as:
- 1. uncertainties in component failure rates
- 2. uncertainty in common mode failure rates
- 3. reduced redundancy during testing i
- 4. operator errors during testing
- 5. component " wear-out" caused by the testing Licensees currently not performing periodic on-line testing shall determine appropriate test intervals as described above.
Changes to existing required intervals for on-line testing as well as the intervals to be determined by the licensee 3 currently not perfonning on-line testing shall be justified by informatior, on the sensitivity of the reactor trip system availability to parameters such as the test intervals, component failure rates, and common mode failure rates.
District Response The Omaha Public Power District believes the existing intervals for on-line functional testing of the reactor protective system are appropri ate. During ten years of experience using existing test procedures, one instance of trip has occurred. This occurrence was during testing in which the reactor was tripped due to faulty test switches. These switches were replaced.
The plant has had ten (10) years of reliable operation. This experience coupled with the existing maintenance and surveillance programs provide a high level of assurance for the Fort Calhoun Station reactor protective system.
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,' - Appendix A OP-1-1 T Fort Calhoun Station Unit No.1 OPERATING PROCEDlRE OP-1 Master Checklist for Start-Up or Trip Recoverv I. General This procedure provides a listing of the items to be completed prior to:
- a. Plant Start-Up - Procedure IV. A.
- b. Trip Recovery - Procedure IV.B.
II. Initial Conditions None III. Precautions
- 2. Trip Recovery precautions are outlined in OP-7.
IV. Procedure A. Plant Start-Up Procedure NOTE: Unless directed otherwise by attached approved Form FC-84 all check-list items must be completed for each plant start-up.
NOTE: Items marked with an
- require double verification /
perfomance per Standing Order 0-37. .
l l I. Start-up No. following:
- 2. Prior to exceeding a RCS temperature of 210*F. complete the following:
Date Primary Plant Tech. Spec. Ref. Time Initial
- a. Raw Water System per 01-RW CL-A 2.3, 2.4
- b. Demineralized Water per
' -) 01-DW-4-CL-A fa JUL13 M F0/78 Rl8 7-13-83
OP-1-2 IV. Procedure (Continued)
O Tech. Spec. Date Primary Plant Reference Time Initial
- c. Compressed Air per DI-CA-1-CL-B1 NA
- d. Component Cooling Water pe r 01-CC-1-CL-A 2.3, 2.4
- e. Nitrogen Gas System per 01-NG-1-CL-A NA
- f. Auxiliary Building Venti-lation per 01-VA ~ CL-A, B 2. 9 9 Containment Ventilation per 01-VA-1-CL-A, B 2.4
- h. Chemical and Volume -
Control System per DI-CH-1-CL-A 2.2
- i. Chemical and Volume Control System per 0 I-CH-2-CL-A ta
- j. Chemical and Volume Control System per 01 -CH-5-CL-A 2.2
- k. Waste Disposal Liquid per 01-WDL-1 1L-B Thru OI-WDL-1-CL-J 2.9, 3.12
- 1. Waste Disposal liquid per 01-WDG-1-CL-B 2.9, 3.12 m .' Sampling Systems per 01-SL-1-CL-A ?M
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- n. Reactor Coolant System per I-RC-2B-CL-A, B. 2.1
- o. Containment Isolation Val ves per 01-CO-5-CL-A 2. 6 N$ g W"E.I bold [Ej JUL 131983 FC/78 R18 7-13-83
OP-1-3 IV. Procedure (Continued)
DATE
- 3. Secondary Plant TIME INITIAL
- a. Demineralized Water per 01-DW-3-CL-A
- b. Potable Water per 01-PW-1-CL-A
- c. Vacuum Priming per 01-VP-1-CL-A
- d. Circulating Water per 01-CW-2-CL-M
- e. Bearing Water System per 01-BW-1-CL-A
- f. Chemical Feed per 01-CF-1-CL-A
- g. Fire Protection per 01-FP-6 (Weekly Check)
- h. Secondary Sampling 01-SS-1-CL-A-
- 1. Turbine Generator Oil System per 01-ST-4-CL-A, B and F
- j. Condenser Evacuation System per
- 01-C E-1-CL-A
- k. Auxiliary Steam System per 0I-AS-1-CL-A
- 1. Main Steam System per 01-MS-2-CL-A REMARKS:
! Checked by Shift Supervisor Signature Date
- .m 7, q ,q p.g
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JUL 13133 FC/78 R18 /-13-83
OP-1-4 IV. Procedure (Continued)
O 4. Prior to exceeding a RCS temperature of 300*F.,
complete the following:
Tech. Spec. DATE Primary Plant Reference TIME INITIAL
- a. 4160KV EE-1-CL-B 2.7 .,
- b. 480V EE-2-CL-A 2. 7
- c. 125V D.C. EE-3-CL-A 2.7
- d. 125V A.C. - OI-EE-4-CL-A 2. 7
- e. ST-ESF-6, F.1, for both diesel generators (if ST-ESF-6, F.1 or F.2, not 01 perfomed previous week) 3.1/3.7
__D2
- f Diesel Generators as per 01-0G1-1-CL-A thru E and D1 01-DG2-1-CL-A thru E perfom before e. above) 2. 7 02
- g. Auxiliary Feedwater System per 01-FW-4-CL-A 2.5
, h. Prior to exceeding 300*F.
l verify that FW-6 can provide Aux. Feedwater to the steam generators via the nomal flow path (through HCV-1103 -
and HCV-1104). 3. 9 Secondary Plant None REMARKS:
Checked by Shift Supervisor Signature Date
- 5. When RCS pressure exceeds 600 psia and after a steam bubble has been fomed in the pressurizer, complete the following:
Tech. Spec. DATE
,- Primary Plant Reference TIME INITIAL sn 'r.;. .. *a. C7ntainment Spray per
-CS- W -A 2. 4 JUL 13 b3 FC/78 . R18 7-13-83 l
OP-1-5 IV. Procedure (Continued)
- 5. (Continued)
Tech. Spec. DATE Primary Plant Reference TIME INITIAL
- b. Safety Injection per 01-SI-1-CL-A, B, C 2. 3
- c. Engineered Safeguards Controls per 01-ES CL-A 2.3/2.4/2.14 Secondary Plant 1 None RENARKS:
Checked by Shift Supervisor Signature Date
- 6. Prior M Criticality complete the following:
Tech. Spec. DATE Primary Plant Reference TIME INITIAL
- a. ST-RPS-2 Section F.2 (if not per-l formed previous week) 3.1
- b. ST-RPS-9 (if not per-fonned previous week) 3.1
- c. ST-RPS-10 (if not per-fonned previous week) 3.1
- d. Reactor Protective System per 01-RPS-1.
(perform after a., b.,
and c. above) 1.3/2.15
- e. ST-CEA-1 Section F.1 (perfonn prior to each startup if not per-formed within prevtous fl43Qf %f? 2 3m.%I.7d E
. .f 3 months) v/,!$ ll.? 3.1 JUL 1319g3 R18 7-13-83 FC/78
,- ,.- ,..w, ,------.--------% = - - - , . - - - - -
-p .- - - --, . _ - - . _ _ - - - _,_._-_.m.-,-- ,,,-,y,-,, ---- ,,,,,, , f, ,--,.,---w,,,,,,-_eg,y,
)
l OP-1-6 j IV. PROCE0tRE (Continued) TECH SPEC DATE O 6. (Continued)
REFERENCE TIME INITIAL
- f. Hydrogen Gas System per 01-HG-1-CL-A NA
- g. Containment checks per 01-CO-1-CL-A (to be done last) NA NOTE: Appropriate sections of C0-1-CL-A necessary for containment closure behind the biological shield prior to diluting within 1% of critical boron concen-tration.
- h. Reactor coolant system per 01 -RC-28-CL-D.
- i. ST-esc-1(F.2) (If not com- Table 3-2 pleted within the previous Item 1 31 days.)
- j. ST-CV-1(F.1) (If refueling or Cold Shutdown is longer than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and not per-fonned in past 9 mos.) 3.3(3)
- k. ST-CV-2(F.1) (If refueling or Cold Shutdown is longer than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and not per-fomed in past 9 mos.) 3.3(3)
Secondary Plant
- a. ST-FW-3 (F.2) (If not com- Table 3-2 pleted within the previous Item 22 31 days.)
RbiARKS:
Checked by Shift Supervisor (p.3,.qp
, .; ., .9 n r ;;%
,; Signature Date
?" j" iUM,. ;',ji UL t 3 ISS
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FC/78
OP-1-7 l IV. Procedura (Continued)
- 7. Prior to Power Operations complete the following: )
Primary Plant DATE TIME INITIAL !
- a. Final containment clesure per CO-1-CL-A. !
l Secondary Plant
- a. Main Feedwater per 01-FW-2-CL-A Secondary Plant l b. Vents and Drains per 01-VD-1-CL-A l
l c. Turbine Generator Stator Cooling System per 01-ST-7-CL-A
- d. Turbine Generator EHC System per l
01-ST-12-CL-A l l
- e. Turbine Generator Hydrogen Gas l System per 01-ST-6-CL-A l CAUTION: This checklist aligns valves l for normal operation in hydrogen gas; i prior to perfonning this checxlist, all I
air must have been purged from gener-
' ator (if required) per 01-ST-6.
1 REMARKS:
Checked by Shift Supervisor signature Date B. Trip Recovery Procedure NOTE: Not required for a startup fo110 wing a refueling outage or an outage following a normal shutdown.
DATE SHIFT SUPV.
TIME INITIAL Verify that the RPS properly initiated an automatic reactor trip.
S D 1.
- 2. Manual trip: preceded auto trip /followed O- JUL13133 auto trip /not initiated.
FC/78 R18 7-13-83
- ,,,,y . ,,, , . _. - . _ ,, __ ,,,
r ,-,----,.._--__,7
OP-1-8 IV . Procedure (Continued)
DATE SHIFT SUPV.
O TIME INITIAL ,
B. 3. RPS channel causing trip:
- 4. Cause of RPS activation:
l
- 5. Copy of the " Sequence of Events" log attached to Section IV.B.
- 6. Cause of trip corrected.
REMARKS:
O Supervi sor-Operations Date/ Time /
SignatQre l
ISSUED O =um 1
l FC/78 R18 7-13-83 l
OP-1-9 Fort Calhoun Station Unit No.1 FC-84 OP-1 CHECK-LIST ITEMS NOT REQUIRED FOR PLANT START-UP O The following OP-1 check-list item (s) are not required for plant Start-Up No. .
CHECK-LIST ITEM (S) Reason (s) Check-List Not Required Initiated By Signature ~0 ate ~
Supervi so r-Technical Supervi so r-Operations Approved:
Supervisor-Maintenance Manager-Ft. Calhoun Station Supervisor-Chem / Rad. Protection .
Authorized:
Supervisor-I and C and Elec. Field Maint. By Plant Review Committee Reactor Engineer Plant Engineer l
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Shif t Supv. 'Date/ Time
- Initials o!SSUED JUL 131983 FC/78 -
R18 7-13-83 i
Appendix B j A-0-2-1
(]
G 'y FORT CALHOUN STATION UNIT NO.1 STANDING ORDER NO. 0-2 Subj ect: Requirements for Criticality Procedure:
h/
1.0 PURPOSE
1.1 To define the minimum requirements prior to making the reactor critical.
27 2.0 RESPONSIBILITY:
2.1 It is the responsibility of the Shift Supervisor on duty to ensure these minimum requirements are satisfied prior to making the reactor critical.
3.0 REQUIREMENTS
Prior to making the reactor critical, the following requirements must be satisfied.
(S 3.1 The requirements of Technical Specification 2.10.1 are satisfied.
(4' 3.2 All four reactor coolant pumps shall be in operation.
- 3.3 The requirements of Technical Specification l 2.1.3 are satisfied.
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3.4 Both pressurizer safety valves are operable vith their lift settings adjusted to ensure i valve opening between 2500 psia and 25h5 l psia.
l l 3.5 The requirements of Technical Specification l
%er 2.2 are satisfied.
3.6 The requirements of Technical Specification l 2.3 are satisfied.
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3.7 The requirements of Technical Specification l
. 2.4 are satisfied.
3.8 The requirements of Technical Specification l 2.6 are satisfied.
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3.9 The requirements of Technical Specification tv 2.12 are satisfied.
OCT 231975 R2 10-23-75
A 2-2 3.10 The requirements of Technical Specification 2.13 are satisfied.
3.11 The requirements of Technical Specification 2.lb are satisfied.
3.12 The requirements of Technical Specification ,, 2.15 are satisfied.
3.13 The requirements of Technical Specification 1.3 are satisfied.
4.0 APPROVAL
-v 4.1 Prior to making the reactor critical, permission must be obtained fran the Manager - Fort Calhoun Station. The Operations Supervisor may act as an alternate.
50 PROCEDURE:
The applicable plant Operating Prccedures and Operating Instructions must be followed while making the reactor critical.
5.1 Changes in these Operating Procedures and j) Operating Instructions must be in accordance
'"'*' with Technical Specification 5.8 and Standing Order No. 0-30.
4 ISSUED CH:T 231975 R2 10-23-75
,. Appendix C I
A-O-5-1
)
Ov FORT CALHOU'i STATION UNIT :10.1 STANDING ORDER NO. 0-5 l
SUBJECT:
Shift Supervisors Duties PROCEDURE: ,
1.0 The Shift Supervisors duties and responsibilities a e as defined below.
1.1 Supervises the safe, efficient and continuous 4
operation of the Fort Calhoun Station during assigned shifts in accordance with established procedures and authorizing licenses. Maintains an integrated perspective of all plant operations including all plant para =eters. During normal plant operations as well as emergency or abnor=al operating situations the Shift Supervisor will '
) use all available resources to aid in the opera-tional decision process.
4 1.2 During abnormal operating conditions, the Shift Supervisor will not leave the control room complex.
Relief of the Shift Supervisor can only be by f-1 (g) another Senior Reactor Operator licensed individual.
Senior licensed plant staff vill advise the Shift i Supervisor during abnormal operating situations and not assume any of the co=cand function until plant conditions have stabilized and a proper ,
- turnover of authority has been' completed.
1.3 During routine operation, the Shift Supervisor may be absent frcs the Control Room complex if .
another licensed individual is designated as the lead control room operater. This will normally be the reactor operator.
l.h Supervises the activities of the operating personnel assigned to his shift to ensure that the plant and all related generating and auxi-
- liary equipment is operated in a safe, efficient and continuous manner within the Technical Specifications anc in accordance with established
}
vritten procedures, equipment specifications and proper nuclear engineering practices. Is fa=iliar with the capabilities and operation of the cceputer installatio n.
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A-o-5-2 O
(m / 1.5 Supervises the licensed activities of licensed operators in order to guarantee that the plant is operated in a safe manner.
1.6 Under the direction of the Supervisor-Operations, ~
is responsible for the proper training of subordinates in accordance with established operating and safety procedures and . development of proper safety practices in the efficient performance of their duties.
1.7 Supervises the documentation of all cperating activities which includes the preparation of operating logs and records required.
1.8 Responsible fcr all releases of gases and liquids in accordance with Federal and State regulations to insure that the radiation release per=it has not been exceeded and that the release is within the limits as specified. Responsible fbr the supervision of fuel covements. Supervises other periodic waste disposal preparations.
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- Appendix D i
- A-R-9-2 FC-96 m
Fort Calhoun Station Unit No. 1 PLANT OUTAGE REPORT Number Generator Breakers Open:
Date Time Generator Synchronized:
Date Time Duration of Outage:
Hours Reactor Condition before Initiation of Shutdown: ([) Hot Standby
([)PowerOperation Power Level Before Initiation of Shutdevn:
O Forced / Scheduled / Maintenance Outage (circle one) caused by what system and ca.ior component if equipment has malfunctioned:
l Method of Shutdown:
i l Conditions During Outage:
Corrective Action taken to prevent repetition (if applicable):
Reference Licensee Event Report (s) (LER) attributable to the Outage ___
(if any):
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i Os Supervisor-Operations Manager-Fort Calhoun Station R5 2-9-78
_ _ _ _ _ . _ . _ _ _ __ _ _ . _ . _ _ . _ _ _ . _._ . _ _ . _ . _ . . _ . _ . _ . ~ . . _ . . _ ,
Appendix E INTRODUCTION & USERS GUIDE E . .
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1.0 GENERAL -
~i-
, .y 4: #-
() Due to the increasing complexity of Regulatory requirements throughout . -
the Nuclear industry, combined with recently updated standards, a need t,
has arisen to clarify and define the electrical, control and instrumenta'-
tion requirements at Fort Calhoun. In order to provide the required ,-1
'DO
, c '. - -
.and clarification andCritical a list.of, the definition it became Quality Elements necessary (CQE) in the to prepare areas ofcriteria elec ,.' for, M trical control and. instrumentation. This document, and its associated" , '.' - {'],
attachment comprises that list. * '
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2.0 ORGANIZATION ** ' ' '
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The CQE list is comprised of several separate but interrelated documents [#F" which taken as a whole, form the entire list. " 'W A ,' % 'd 'I '
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2.1 The first of these is a Design ~ Criteria. 'The design criteria ~ sets ll ~
forth the philosophy and guidelines which determine the i n , T".
classification of a component as CQE or NON-CQE. In addition' it " ' l '.
~ ~
- defines the classification of certain large groups of Electrical . gg
' ,* ' . " ' ' Components (such as 4160V Switchgear) in order to minimize,the
. ? .'
numbers of components which require specific identification or listing. ., .
7l [^ f.6'. "
.... . 7 Furthermore, the design criteria contains a guide for the ~"
procurement of replacement components and new components for use at f- Fort Calhoun. * -
c]
Q ' *
'. 7 2.2 The second portion of this document is the listing of CQE components. This list identifies the field mounted devices and .
associated control pancIs which are CQE. It also contains a - - -
listing of major subcomponents (e.g. , breakers, starters, etc.) in, ' '
~
equipment such as switchgear and motor control centers. -
. 2, 1, . 3 J
g ..
2.3 The third portion of the list comprises a set of drawings which identify the panel mounted CQE components. In order to keep this ;,
' portion of the task manageable the means of identifying the .~
l components was selected as the panel bills of material. This has '
I kept the numbers of drawings associated with the CQE list within "
useable limits _(attachment 1}. _,
2.4 The list has been arranged in an alphabetical sequence by system.
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A system list is attached to the front of the CQE list and serves l as a table of contents to the list, as well as a system reference l point (see section 3). -
3.0 USERS GUIDE In order to properly utilize the CQE list, the first step which should be ,
taken is to read the design criteria. A familarity with the criteria can save a great deal of time when attempting to define a components CQE status. In addition the Design Criteria contains the definitions
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o)
C neccessary to properly evaluate the electrical systems and components i associated with the Fort Calhoun 4160V & 480V switchgear and 480V motor '
control centers. .
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MTERIM With the abow considerations un mind the following a'pproaches should be . .
followed to effectively utilize the CQE list. . .
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(Jl 3.1 ELECTRICAL COMPONENTS
' 3.1 1 Motors, valve operators and field mounted devic _es - Determine the system associated with the equipment drive by the motor and refer .
to the CQE list. If the device is listed it is CQE (ex. raw water
~
' . ' .; ', .. . pumpAC-10Amotor-referencesystemRWandref'rtotheRWlist,}'y e .,
.y where AC-10A will be found). , . . - . . *
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, 3.1.2 Electrical power components - Determine the voltage type (e.g., % $;.y..y
_. , AC or DC), the redundant power group (e.g. , division A/C or B/D) j,' .p.;.pc and refer to the CQE list. (Ex. Diesel gen D1 - AC voltage, and ,, . g,; p.a
('J,g -
,. feeds electrical power divisions A and C reference system "elec.
pwr.-div. A/C" and refer to the CQE list where diesel generator D1 m E ,..
E ' '}.n,~ '
- 'will be found.
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3.1.3 Relays and control devices associated with electrical power ze, ,
3.,
components - These devices are either identified on the control -f.@'g,.s.
.. panel bill of material for the panel on which they are mounted or 4:
discussed in the Design Criteria and identified in that document. n@i,% ,.;;Q,7..".;. . .
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3.2 Control components -
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. 3.2.1 Level switches, pressure suitches, pushbutton stations and other '?J.R.l -
field mounted devices - Determine the system associated with the * ' ' ' .
m control device and refer to the CQE List. If the device is listed, . . . .
i.'*-
l} it is CQE. (ex. PCS-230, CH-1C lube oil press switch - reference
\ system CVCS and refer to the list where this device will be found).*2:. 2'
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3.2.2 Panel or control board mounted' control devices ; ,..,,;...
.g,--
- u. .:..
. For devices such as contro'1 switches and relays associated Mith .', ,
. valve controls etc. refer to the bill of material drawing for the ,
panel that the' component is mounted on. The CQE designation will_
' be marked on the drawing. .
" -2
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INSTRUMENTATION
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Sr. ,
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For instrumentation the device or component will be found using the panel '
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, method outlined above for control devices. . ,\ M ~
' 4.0
SUMMARY
The electrical CQE List is a composite of several documents. Access to a given device can be achieved by several methods including Design -
Criteria, Drawings and the list itself.
I Examples have been included to allow the user easier access to any given device.
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Provisions have been included to allow for selection of replacement components and updating of the list. .'
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SECTION II . ..
DESIGN CRITERIA --
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CRITERIA FOR PREPARATION AND MAINTENANCE OF ELECTRICAL CQE LIST p(/
FORT CALHOUN STATION OMAHA PUBLIC POWER DISTRICT T D1M g 11iEyE
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1.0 Purpose ,
-u- .
1.1 The purpose of these criteria is to provide a document for.the
.- preparation and maintenance of a list of electrical controls and
. - ' . instrumentation components that are Critical Quality. Elements (CQE)
, .- for the Fort Calhoun Nuclear Facility. In addition,:. guidelines will
- 7 tbe established to assist in the repair / replacement of;these CQE components and subassemblies.
1.2 This does not cover the installa~ tion requirements.
2.0 Basic' Criteria -
2.1 The following referenced documents form the bases'for 1:
. establishment of these criteria:
5.- . ... - , , , , ,.
A. Fort Calhoun FSAR , . y ;, .
B. Fort Calhoun QA Manual .;.c h- ,
C. Fort Calhoun P&ID's . .. - $"M,,i D. Fort Calhoun elementary diagrams, loop diagrams & logic diagrams E. IEEE Standards & Guides 279, 308, 323, 344, 379, 384, 420" 3.0 General .._ , ,._
3.1' A CQ2 component (class IE)* for the Fort Calhoun' Stati< . ,c d. fined as follows: " Critical Quality Elements are defined as thone strr o rm ,
.. systems, components or items whose satisfactory performance is re-
. .. quired to prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public."
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- ' For the electrical CQE list the above definition has been applied to those electrical, control and instrumentation components that
- function directly, or in support of systems that are required to meet _
".- the above definition. ,
3 3.2 Components required to support systems are identified by means of a system analysis. The system analysis is to be performed as follows:
A. 1. The Fort Calhoun P&ID's will be revicwed and those systems or portions of systems required for mitigation of postulated accidents will be identified. A list of these systems will be included with the CQE list. (See j section 3, CQE List). ,
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- Reference to IEEE Standards and term Class IE is not intended to imply that I all replacements are required to be in accordance with the latest standards.
Criteria for replacement components is defined in Paragraph 5.0 of this d+u-p/
x ment.
- 2. Each of these systems will be evaluated to' IRrMifytiM se ' .
electrical, control or instrumentation components ,
required to support operation of the system.
[V)' . -
- 3. These components will be Tirted in the CQE list.
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. ;[~ . ,.,' ,
B. Panel, switchgear, motor control center or control board
.' ~
' mounted items required for aupport of these cornprinents will be analyzed and identifie~d. Panel and control board. mounted g,,fy,'.(~y;f" e ", .' . , ., -, . . components , are identified o.i the bill of material for the - -
- ,%. .'d.' E .* . , ' . . . ' ' Panel or board. Switchgear and motor control center components are discussed further in the specific criteria
.4GI c:. .g( @$nC ':'f,sections ,/ ' ;-k di...
of this document. ,
n.'y * . . '
,y.J / * ., 3.3 ;In addition to a systems analysis, components are evaluated to ,
25- ensure that failure of a device will not lead to an event "i.,'S',.[s.
',] . ' ',.;;'/h;[ ,' detrimental to the health and safety of the public.
W '. &, . j. d: ]%, ,c/ f - . ~ . - ~ .* - -
.J. J.J.c 3.4 ,,
Replacement criteria for CQE corronents will be developed as a
.- 4 .?."., specific criteria. (See sectio.. c. 0;
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4.0,5kcific Criteria - Electrical, .Contro .t tv -
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4.1 Electrical components -' '
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4.1.1 The' electrical system for Fort C.ilhoun consists of'a group of redundant distribution systems designed to supply normal and emergency power to the facility. Tbc major components which make up the distribution system are as follovn:
A. Station service transformern 4160 Volt switchgear & breakers B.
C. Diesel generators for emergency power -
D. 480 Volt switchgear and breakers
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E. 480 Volt motor control centers F. 120V AC & 125V DC distribution panels
' * [ .; G.
. 125V DC batteries ,
H. Battery chargers ,
. .. . J. 125V DC to 120V AC inverters .
,i K. Miscellaneous power distribution panels- (e.g. pressurizer
. heater control panel)
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L. -Cable ..
/ .. ' M. Cable trays and conduit
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N. JMotors and motor operators n
4 I
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.:. - INTERIM The classification of each of these items and a CQE breakdown of *
- the individual components is handled by any one of three methods: -
.- u. - '
Identific5tionofthecomponentontheCQElist O A. .- -
V ..:
B. Identifying the components as CQE on bill of material
' drawings. - ,,,
u F'
.C.
.. ,' . Defining the component as a portion of the criteria. -
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4.1.2 The criteria which applies to each major component and its J/ . % @,^Y ?
f1 - %"( ,, -... . associated subcomponents identified above is given in the
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.c, ... ;. following' list. - 2
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.v:' W %. . ..,j,,s r. . . p d-f*2 .' / G~ * '.n' ?. ,q.-f A. Station Service Transformer
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The transformers (TIA-1, 2, 3 and 4) serve ~to supply power '.' e ' .;i 9.y;i O ' .
~..
to the Fort Calhoun Station during start-up and normal *@iWp; te '
- '" 1[ ~,I operation. In addition,'during a design basis event, the[ N - M .;
.t . ',; . W .. safeguards systems are powered from the station service l^ .; s. g L,i .
.T..g ;[,,, transformers, with the station's emergency diesels in a ~.. ,', f, . / [ ,,
4.,w-. -
..',. standby mode (idling). '. - .- 't W ~ ~ .
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. - Should a loss of offsite power. occur, the bus undervoltage s. s.*.,. s.
relays will operate to disconnect the 4160V switchgear, fiom ,. q,.v,.
.,c, .. . , , > . c . :-
. offsite power and initiate full diesel generation via , - ,'( ., ; - [. ,
loao-shed and sequencing. : ** .g. *:
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Since the station service transformers serve no purpose -&
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. . ., ,m; eff;.; w:. ,ge. -
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during this event., they are classified non-CQE.
f3 @
~.. M d- B. 4160V Switchgear, relays, and associated breakers
. - A7>[n..-2.Yi 2,
. . .,K.- - .
There are four 4160V switchgear buses at Fort Calhoun, IA1, l-(,f.
1A2, 1A3 and 1A4. In the event of a loss of offsite power, < +F, - .
buses IA1 and 1A2 are tripped. Accordingly buses IA1 and, , p'.'
a IA2, and their associted breakers are non-CQE. -pW- .. . - W 'Q", *' .
,, . . Y m. .
l', - Busesi1A3 and 1A4 are connected to diesel generators D1 and ".
D2 reupectively. These diesels constitute the source of g' , s . 3 :C.
3
.s emergency power in the event of a loss of offsite power. .g.d. *
. . : ~
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In addition, both engineered safeguards and normal loads are , ,
t -
-fed from these 4160V buses. . ... .
~. .. .
' 'In order to accomplish a transfer to emergency diesel power
- several steps are necessary. -
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- 1. Trip 4160 volt uteu ers unciuding e . .
safeguards if a design basis accident is present). .,
- 2. Bring the diesels to full speed and voltage. ~ '
- 3. Using the sequencers, reclose engineered safeguards breakers in a timed sequence. -
. Because of this sequence of events the IA3 and 1A4 4160V " ,_,
, ,,,3 . buses and breakers. require some special provisions with . ;.. d. ,
g, ,
regard to CQE designations. ,
. - . .; .,g,j(;,.
.~~. .i .. ,'. '
.:q.;.),%%,'0 to.: } ,
These are: , . -
.. ,7 ...:,.c.-
L $.. .". ) .O + -
. .1. All breakers directly connected to buses IA3 and 1A4, ^Ig:, d
.,[C[,Z
, including their close and trip circuits are considered ,
CQE.
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- 2. Breakers feeding safeguard loads are listed in'the'CQE'. ~,. u.
list to facilitate cable requirement. (Cables connected U~.T". .
, to these breakers are CQE-E prefixed). -
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... e. .. r .c. Ee
' ~ '
. . . '3 . Breakers feeding non-safeguar'd loads are 'not 1iste'd in th ' '
.-) . . ...
CQE list but are CQE because they provide isolation
- between .. .
the safety bus and non-CQE load. Cables connected to these 3..,
breakers can be non-CQE provided they can meet flame test .
requirements (qualified per IEEE 383) and separation re- 2 quirement.
~
For Fort Calhoun these cables have been pro- .
vided with prefix A, B, C, or D and were purchased as CQE -
. cables.
4 . ,
y ', . . . . .
- 4. Undervoltage relay schemes which initiate and maintain .
the non-CQE breakers in the tripped position are CQE. -
( " 5.The 4160V switchgear structural components and buses ..
1 - - are considered CQE. -
- m. .",,
' Diesel Generators
=
C. ,
4 .
The diesel components will be identified as CQE on the .' , , f'.
list and on the diesel panel bills of materials. Diesel ' ;. -
control schemes will be reviewed to determine which ..
[ ,..- Q..W< :
components are required for diesel operation. These will be identified either on the CQE list or on a panel bill .
of material.
D. 480V Switchgear, Relays, a'nd Associated Breakers l The same reasoning applies to the 480V switchgear as was
! applied to the 4160V switchgear. Since both non-CQE and i CQE breakers are fed from the same 480V buses, it is essential to diesel operation to load shed and sequence the emergency loads back onto the buses. In addition the
, non-CQE breakers are the primary isolation point between l the CQE and non-CQE system.
~
x ,.
.. . r.
% . f?cfer. IEEE 384 -1977 I
. . . .. . . . . - -..-J -~ -
- lGERIM Since this is the case, the following criteria apply: . .
- 1. All breakers directly connected to buses lA3 and 1A4',
including their close and trip circuits are considered CQE. _
. 2. Breakers feeding safeguard loads are listed in the CQE '
list to facilitate cable requirement. (Cables connected to these breakers are CQE-E prefixed). ... ,
,. '3. Breakers feeding non-safeguard loads are not listed in the CQE list but are CQE because they provide isolation
- between
.. the safety bus and non-CQE load. Cables connected to'these breakers can be non-CQE provided they can meet flame test requirements (qualified per IEEE 383) and separation re- .
quirement. For Fort Calhoun these cables have been pro-vided with prefix A, B, C, or D and were purchased as CQE cables.
- 4. Load shed circuits for the 480V breakers are CQE. ,L,
'5. The 480V switchgear structural components and buses '
../
. . . , are considered C,QE. ,
E. 480 Volt Motor Control Centers and Starters Again, as with the case of 4160V and 480V switchgear, .
both CQE and non-CQE loads are mixed on the same buses. ' .
Accordingly, 480V starte.rs and breakers feeding O- engineered safeguards and other CQE loads are listed on the CQE list. - .
Motor starters and breakers feeding non-CQE loads are
. also designated CQE since they serve as the isolation device between a CQE bus and a non-CQE load.
In addition, non-CQE loads fed from motor starters are l
load shed to eliminate these loads from the diesels (i.e.
. they drop out on undervoltage and do not restart).
. Accordingly, the following criteria apply to 480V motor control center motor starters and breakers:
- 1. All motor starters and breakers feeding CQE loads l
will be listed in the CQE list to facilitate l determination of CQE cable requirements. Cables I connected to these starters are provided with an *
'E' prefix.
- 2. Motor starters and breakers located on CQE motor control centers feeding non-CQE loads are also designated CQE since they serve as primary isolation l
and load shed devices. However these starters are I not individually listed. Cables connected to these starters are provided with A, B, C or D prefix.
ab Ja 384-im
~ ~
- 3. Motor control centers, taent.ified on the CQE list,'
. INTERIM '
7:
are classified as CQE including enclosures and '2.j -
O buses. All other motor control centers are non-CQE N c. ' -
including starters and breakers since these MCC's are '
b load shed by the feeder breaker located at the 480V .l 1 5;.$."' ,
.' switchgear. >
,.-s. Y,3 J. 9; 't... M; ,. h
.- c. ; ..q.~ ;js- :
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,F. '.' 120VAC & 125VDC Distribution Panels
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All 120VAC & 125VDC distribution panels listed in the CQE .j. "jifl;.;
" . ,;, - list are considered CQE including all of their breakers. , .y4.j .f.
,,;,,7, * , Again, only breakers feeding CQE loads are listed in the "My.h:3
.CQE list. Breakers feeding non-CQE loads are ' considered, Gr!!p,.
, CQE since they serve as the isolation device between,CQE 7.'.5/JR 7 and non-CQE systems. -" ' . l's F
,... a- ~!' E
- w.. , , . - ,;, q s.s ..w. Q::.y'z..
.;,y;' . r :- .' 4.. :
.. 7.
G. 125VDC Batteries -
. n. u.;:: .
125VDC batteries,' feeding 125VDC buses 1 and 2 are CQE, .,
including battery racks and fuses. .
a 3... .c ,. l . . .y,$i;. ..
/ G,.
' . .. ~.u' . .. w .
,- 3 H.".M ;. Battery Chargers :. - c, . .. .
. s -t'l' O d Y. .s..:'N.h.',N.db$E.'^7dl.hh These are considered CQE in their entirety. ,, /,W r V . :., '. ;
J.- 125VDC to 120VAC Inverters < .- . i ' e,.
.: f..!,.d y 1, All four inverters (A to D) are CQE since they supply 9.; -
power to vital 120V AC instrument buses.
K. Miscellaneous Power Distribution Panels '
, ~~CQE panels that fall under this category will be . . -
~
identified in the CQE list. .
j py .
Y" '
Cable'
.,,. . , L. - -
4
Power, control and instrumentation cable is considered .
.s CQE if it serves the following: .
, . . , . g, -
- 1. feeds a CQE load -
.. ' . ~ nti;.
- 2. is part of a CQE control circuit M 2'
-3. is part of a CQE instrument loop s.
.s'-'
.s...
.* . I'
M. Cable Trays Conouit and Junction boxes c
Cable trays, conduit (including flexible conduit) and 1 '
junction boxes are CQE if they are used for routing of ..
')
(V ~
CQE cable. .The term CQE with respect to these devices
r means the following: .
. i - . .,
' , 1. They must be seismically supported. .' .J.* f
~ s ;,* ; - .,
- 2. Cablesmustbeinstalledinthecorrectdiv'isiSs ~ .'-L. b '
~
raceway .gg ,, . ~
, . a n n., ,
.; , ,. . . ,;j . ut.Wip., ':
- 3. 2 ,
Proper isolation between raceway maintained.
,. . . ? and ; '. t,g cable must be;Q For procurement purposes, conduit fittings'and' junction ~.. . M. h i:_!. .. .
boxes may be procured "off the shelf" using commerically -
available material, provided that the equipment is .ae;6 . -
installed seismically supported. . .? , , , , . . , '_
. . .'Es:' ..
. N.
Terminal Blocks and Splices
' '],kT' ' ', '
. . ,..:.. . m . .:. .
. . Any terminal blocks or splices us'ed in any CQE circuits , ,
are CQE. ~
- 3. .; ,
4.2 Control Components -
.g. g .
., . ; y t a . ,. ~
Control devices (limit. switches control switches, relays etc.) ' '
fl will be designated CQE if they perform, or are required to b- '
maintain their status, to achieve any of the following actions:
- a. If a device is a part of a control circuit for a CQE item
~'and is not isolated from the circuit through a (class IE) _
,'I. isolator (e.g. breaker, motor circuit, load shed circuit 3
~~
etc.) then it will be designated CQE.
'. A. -
.,. . . ' . ;3 .
.b. . If a device, through its failare during a design basis
,- event, can cause adverse operation or failure of a ,
circuit leading to an event which will affect the health -
and safety of the public, then that device will be designated CQE (e.g. a control switch or relay which could inadvertently close or open in a seismic event .
. leading to a radiation release).
- c. Any device that is required to initiate or maintain l
Engineered Safeguards Operations.
- d. Any device required for operation of Reactor Protection Systems. ,
- e. Any device which provides system status or component l status, if that system or component status is vital to .
l ^
operator action during a postulated event. (e.g. CQE valve limit switches and lights).
(G) -
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l . . . . . . - . . .s,.
L
5 4.3 Instrumentation ,
The following criteria will apply to the elements of the Fort Calhoun
(~') instrumentation systems. Instruments (level, pressure, temperature,
\/ flow transmitters, power supplies, indicators, etc.) will be designated CQE if they perform or are required to maintain their status to achieve
- any of the following actions: ,
'a . If a device is part of an instrument loop that initiates or
. performs a protective function related to the prevention or
, - mitigation of a design basis event than that device shall be CQE, unless it is suitably ' isolated from the loop. Suitable isolation is defined in IEEE-384. Considering the fact that Fort Calhoun predates IEEE 384, Precision Resisters provided in the current loops for converting current signal to voltage signal (e.g. inputs for the computer) are considered acceptable isolation means. However, the resister in this case will be classified as CQE. For all future modification isolation devices shall be in accordance with IEEE 384 and Reg Guide .
- b. Any instrument whose failure could cause a design basis event
- shall be CQE. ,
- c. Any instrumentation vital to ensure proper operator' action to prevent or mitigate the consequences of postulated accidents will be designated CQE.
4.4 CQE Justifications
/~T (m / After reviewing electrical, control and instrumentation components per the above criteria, the component and the reason for its CQE classification will be given in the CQE list.
To facilitate the listing of the reason for a CQE classification,
- the criteria of 4.1, 4.2, and 4.3 can be further defined into the
, .' following CQE justifications which will encompass all Electrical, Control or Instrument CQE components at Fort Calhoun.
The CQE justification for the CQE list are:
- 1. Safety signal input.
- 2. Required flow path follouing a Safety Signal.
- 3. Required to operate following Safety Signal.
- 4. Provides alternate / redundant flow path for a safety related function. -
- 5. Provides CQE component condition indication.
Sa. Provides CQE system condition indication.
I 6. CQE component positioner or operator.
- 7. Required for proper CQE component' operation.
7a. Provides indication of impending CQE component / system trouble.
- 8. Power supply to CQE components.
Ba. Provides centralized control / indication for CQE l components.
(V) .
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5.0 Replacement criteria lQ 'e
.. . , ).f . ~ s Electrical, control and instrument.ation components in use at Fort Calhoun .._.
(h,)- were selected and procured based on a set of regulations, standards and 'g' l e i;
.. criteria which, in some cases, have been revised since the initial con
. },4 struct. ion of the facility. Certain of these revisions are minor in nature .y; tr ,r . . '
and have little affect on components. In other cases new criteria or regu- .. .
?!f.h ' . , lations impose much stricter requirements. These have greatly increased ., .-
.,' MrM.i athe burden of personnel involved in replacement of components and subassemb '; 'y '
L
, Ms.j[/.eh'M5.b,'.;liesofcomponents. . -.
li b e .) >;.
h'., q y. <. .;N
- q. q . .; , 9 3. .
, .7., 3 fM N !!~y Procur'ement criteria'for equipment which is to replace existing CQE equipment ).h.
d : located in harsh environment or which is new CQE equipment must be reviewed N4'.i .
4
.'R: .. M ;' in light of NUREG-0588 and NRC Bulletins 79-01 and 79-01B and NRC's me'morandum S c.@21 and order CL1-80-21. In the responses,to Bulletin 79-OlB, the NRC required J.17 :
- ?/. . ' 1 ' ' "all plants to address equipment aging conditions which are requirements of 1 ;;A
' ^
q-c , LIEEE-323-1974.
, ' ' .t.g. '.".y. s . Q..
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- [,, Considering a possible future change in the NRC's attitude'towarde not y ..y..-G .
~
,. j ,'k. .- applying current standards to plants built prior to the effective date r '. of that standard,~the following procurement criteria are recommended. .;r;M,l . a.c. 3,@y r Qf,N Q5.l' q &-:- ? . .'. r -
or Damaged CQE4G'N,';Q.lg
' P[h' ' Replacement of. -'Wrn . . . - . wi., . , . Ju. ; v. .., . .~ :
Equipment DJ - '
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. . .. s. -G.
.A. Were a component requires replacemeni due to normal wear'or'9. ~
) V2.f,
' - [t'
.~ damages, performance of the component has been satisfactory, and , p an exact replacement is available, that replacement can be u' sed. ~ i jy+. .
- , ,, . . : ;.- ....~,,g.:
n g, , . Exception - Were that. component has been identified by either " ,
. the District or NRC as a component riot suitable for the appli
. -: .- cation and is to be r'eplaced at earliest convenience by a speci-
~ '
., fied replacement. , ,
~
.,..y
.'t.,. --'
, q e q,q .c . , l.
B. Were a component requires replacement due to abnormal wear ,g ;
caused either by service or environmental conditions it should 3 ; ,1; , ,
f' . ..'be replaced by a ne'w' type of component suitable for the' prevalent
, . f service and, environmental conditions. Unless there are sound -
reasons to the contrary the component shall be qualified in ' '
accordance with IEEE 323-1974 as amended by NUREG 0588. '
~. . . . .
C. Were a' component requires replacement due to normal wear, but f, f.
.'a'one to one direct replacement does not exist, the faulty T-
.. component should be replaced by:
'l . A component suitabic for th' cservice and environmental
conditions. Unless there are strong reasons to the contrary the replacement component shall be qualified as ,
Class IE per IEEE-323-1974 and NUREG 0588. ,
- 2. If the procurement of the replacement component to ?
IEEE-323-1974 is logistically impractical, a component can be used having been qualified to the standards that ..
existed at plant start. up. Basis of qualification could
) be the use of said replacement component in another piece G..
of CQE equipment and qualified to the same or harsher service and environmental conditions. . . .
u .
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4' . m ,'
'If the replacement is not on "one to e asis n an lysis . .'? '
', .' justifying the substitution should be forwarded to GSE for m :. .
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.. review. , .. . .-
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ty ' 5.2 New Equipment
...- . , . q, g . . , p . eg - - <
y.., -
, ., .. . .m . . , .
3.p , . < A. Were a component is to be purchased as part of an entirely ' ,.f;.,,.:n..u.t;.;}M .
ff-W
.. new system or an addition to an existing system which is to be classified as CQE, this component should be qualified'as. Class" f.,'
~
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p..:j./.p.' ... rf.' .
..IE per IEEE-323-1974 .and Reg Guide 1.89.
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. ry. (. : :. v .
.s 3:; %.4 3.y.'.v,4g. q, .g-~p Q.g :,v
- v. . m;p
_+ ..
- g. '
yep-QQ~~. ,
gi'M,2 ' j/f.f : -/. - B. Were a comp'onc'nt is required per part of,a" above, but.it i,s,gg.g;,g[, ,.;
either economically or logistica11y impractical to purchasega
~ ,
P :-
-l g}j *;w, . :q . .% i; ".
. , component qualified to IEEE-323-1974, thenthecomponentf$,j:phg]. y:.4, .f' 7,y['. ,6 , .. .,,
..should plant atbe purchased qualified to the standard existing at.,the. $ ,y*f;Q.
pc ,f ' , .
.,. start-up.
(Several~referenc.es have been made W.to;lg:[.y"W~ g:,*-
qualification to either-the 1974 or 1971 version of IEEE ".
.$.i?f- .
requires UM.?
- -Standard 323)." Th'e procurement *of CQE equipment also.'nt g ; ~- . ,
that proper documentation be supplied with the equipme ;.gfq;y, .y k . stating that it is qualified for its particular application.
Proper documentation should already exist at the plant forf. "[. l,' Q S "$.ii ,'.
- p. p.' , . . :.
[iQ equipment purchased as a:one to one replacement, or' identical.,jf.5-gg.i
,.,'- . .:, / to existing equipment but,to be used in.a different capacity. , . .,. p . .
.gy e . . .
-For equipment which is new to be themore. plantdifficult or thattohas been, p p$ 1,,7;d't. .
reclassified, documentation may obtain.,,
Every means should be pursued to acquire this documentation. 9 @ .:..d.:Z :o
... e :
.. , B.wkw+g:.w g l.~:.. - .>
C. GSE Specifications GSEE-0801, 0802, and 0803 may be used to q sy A O
v facilitate procurement process. These specifications outline T,'
the general requirements applicable to CQE components. However, . '
'the user should carefully review the specifications and specify "
l
. applicable service conditions. If in doubt consult GSE-Nuclear.
.. ,~ . , . . , .
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l 6.0 Summary .
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.A list of Critical Quality Electrical Equipment will be attached. <The.
_' ; ' J.; ;'
definition and criteria used to establish the CQE systems and -
. assemblies that make up those systems have been discussed. Criteria -
.'for procurement of replacement or new CQE equipment have been discussed 1 '. ' .^.,sM -
c., < ji
. and procurement criteria recommended. .
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- Appendix F DRAFT i ,
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GENERATING SIATION ENGINEERING
.. MECHANICAL, STRUCTURAL AND ELECTRICAL
~
CQE EVALUATION OF FORT CALHOUN NUCLEAR STATION !
1 i
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OMAHA PUBLIC P0kIR DISTRICT i
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Copy No. Revision 3 l Assigned To: Da t.e : April, 1982 i NUC3081/a/1 j i
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.. - ..-,-. .... - .. ..-_. _ _. - - - .. _ _ _ . _ _ , - ~ _ . . - _ _ __ _ . . _ - - - -
~
, OV.: 4A!.L F.IVISIO!; SHEET
.8 RALL SECTIONS CHIJ:GED/ APPROVID - GSE Rz.iISIO:: FUF20SE PREPARED SY
- SECTIOli MGR./DATE 0 All/For I=ple=entation egg Q6 M g
[ h hY
//T/7o Pages 13 and Appendix A/
1 revise CQE requirements. gg ()h f/'/2//d
.A p x A/ Flag change on c))$ W 2
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DRAFT (6
PART ONE MECHANICAL AND STRUCTURAL CQE LIST
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- =um DR.AM Table of Contents C Section Title Page I. INTRODUCTION 1 A. Code Development History (Mechanical) 1 B. Background and Present Code Application at Fort Calhoun Station (Mechanical) 1 C. Structural, Miscellaneous Mechanical, and Interface /
Accessory Items 4 D. Various Nuclear Component Code Comparisons 5 E. Conclusions and Recommendations 7 II. MECMNICAL AND , STRUCTURAL CQE REQUIPIME,NTS 8 A. Definition of CQE & Limited CQE, and Examples 8 B. Mechanical CQE 10
! 1. Steam and Water Retaining Systems and Components
- and General Requirements 10
- 2. Other Mechanical Systems 13 C. Structural CQE 15 III. REFEPINCES 16 Attachment I -
Listing of drawings showing mechanical CQE class boundary flag.
Attachment II -
NRC Regulatory Guide 1.29, Rev. 3 "Scismic Design Classification" September, 1978 Revision 3 k April, 19S2
DRAFT I. INTRODUCTION (A) Code Development History (Mechanical)
AS!E Section III existed as early as 1963. The AStE Section III code of 1963, 1965, and 1968 was a nuclear vessels code only. During this time period, USAS B31.1 (now ANSI B31.1) was used for the design of other pressure retaining components, e.g. valves and piping. In many cases, the methods and criteria of analysis and design specified by the early AStE Section III code were used in addition to B31.1 to design piping. Such is the case with the Fort Calhoun reactor coolant piping. Eventually, B31.1 developed specialized requirements for certain areas of piping and components to be used in nuclear power plants. These areas were given class divisions and a separate code and ANSI B31.7 was developed to handle them. B31.7 was initiated in 1967 and was changed and developed through the years. From about 1968 through 1971 it was incorporated into AStE Section III. The present AS?E Section III code not only covers nuclear vessels, but all pres-sure retaining, components that carry ste,am or water. This code also
, has rules for the supporting structure of these components. At the l present time, the AStE Section III code is in transition from a com-ponent concept to a system concept. i (B) Background and Present Code Application at Fort Calhoun Station (Mechanical)
Fort Calhoun Station came under the requirements of various codes during design, construction, and operation. Because of this situs-tion, it is inevitable that confusion will arise with respect to which code applies to a certain system or component at the present time.
AS!E Section XI, Article 7000 - Replacements, can clarify this.
AS!E Section XI essentially states that replacement of existing com-ponents, modification, or rerouting systems is controlled by the requirements of the edition of the construction code to which the original component or part was constructed. The use of later codes is acceptable as long as all the requirements of *.he original code are met or exceeded. In all cases, if a more restrictive code or require-l ment is used, the minimum code requirements should be stated as well as the code actually used.
l In all cases, it becomes apparent that the original design basis and codes should be known for each component and system. This can usually be determined by referring to the Fort Calhoun Station FSAR, the Bill of Materials List, the Fort Calhoun Station valve index, and associated contract documents. The codes and requirements established by these
- documents represent the minimum yet mandatory requirements for replace-1 ment parts and components. This remains true unless a subsequent agreement with the NRC or other regulatory or enforcement agency has been reached to alter the situation. Likewise, additional rules, laws, or regulations may be enacted to change the original agreement.
tO l
l
DRAFT Until the present AStE Section III component code was developed, items such as pumps and valves were not specifically covered.by a nuclear code. USAS B31.1 and B31.7 were generally applied as piping codes. However, for nuclear plants whose construction permits were issued prior to January 1, 1971, 10 CFR Part 50.55a requires the following criteria to be applied:
- 1. Piping
- b. Nondestructive examinations performed in accordance with -
the same codes for construction. 2
- 2. Pumps
- a. Construction for Class I, " Draft ASME Code for Pumps and Valves".
- b. Nondestru'ctive examinations perfo'rm'ed in accordance with ASA B31.1.
- 3. Valves
- a. Construction in accordance with ASA B31.1 and USAS B31.1.
- b. Nondestructive examinations in accordance with the same codes for construction. l@'
In order to apply codes which were, in many cases, written for piping caly, the architectural engineering firm and OPPD developed the specifications for pumps and valves incorporation through codes listed above. The local designations are Class A (safety class) and Class B (non-safety class). The specifications for these classes can be obtained by referring to the original contract documents. '
It is important to be aware that the ASME codes, or any other codes, are not by themselves considered law. Each individual state must adopt the code for it to become law. Also, local governraents may have additional or different requirements which must be known and adhered to in order to construct or alter a system. Even the NRC does not require nuclear components to be certified in accor-dance with the ASME Section III Code although for newer plants, they state that each component must meet standards equivalent to the code. For systems designed to the ASME Section III code or equivalent, it should be clearly understood that the NRC does not require the N stamp. By stating the requirements in the design specification and itemizing each requirement rather than stating, for example, an ASME Section III Class 2 valve with N-stamp is l
l
DRAFT required, an item can sometimes'be obtained at a lower cost. Also, by I
itemizing the requirements for a component, an exact record is produced which can be used in the future to replace the component if necessary.
This reduces the amount of work to research codes, etc., to find out what exact code requirements were at the time of the purchase. Codes are con-tinually changing, therefore, the better and more complete records that are kept, the easier the job will be in the long run.
The Fort Calhoun Station P & ID's have been drawn showing boundary flags which indicate boundaries of systems as described by 10 CFR 50.55(a) and NRC Regulatory Guide 1.26 (see Attachment II). These boundaries were set up using one of two standards. For those systems which have steam and water retaining components, the flags represent the equivalent AS?E Section III Class 1, 2, 3 or non-class boundaries. For systems that do not contain steam and water retaining components, such as the HVAC system, the AS?E Section III code does not necessarily apply. For these systems, a special " safety class" is used. This " safety class" closely represents the safety class 3 that appears in ANSI N18.2 although some safety class 2 specifications fall into
, the general definition.
The use of the AS!E Section III classes and this
" safety class" is illustrated in Table II. ** '
It must be understood that the majority of the systems were not originally built to the AS!E Section III code or the ANSI N18.2 code. For a newer plant, these boundaries would represent an actual difference in the quality requirements for components falling into one part of the boundary or the other. At Fort Calhoun Station, however, the boundary may or may not represent a quality difference. The purpose of setting these boundaries is
'S to indicate the relative importance or difference in function of components that are separated by a boundary. These boundaries also provide the guide-lines not only for determining the in-service inspection requirements for the system but also for the addition of new systems with regard to what class the new system will be built to. Also, if old systems are to be upgraded or undergo major modifications which will change their design function, the boundaries will supply guidelines as to what codes the system will be upgraded or modified to. If a one-for-one replacement of compon-ents is required, it must first be determined if the failure was due to normal wear and usage. If a failure was due to a deficiency in the speci-fication, then additional or different specifications must be used which may include upgrading to the level of AS!E Section III. Otherwise, using an exact replacement and adhering to the original construction code is all that is required. The rerouting of piping and the addition of valves into -
a system is allowed by ASME Section XI, Article 7000-Replacements to use the original construction code of that system.
l
(
.. DRAFT
, (C) Structural, Miscellaneous Mechanical and Interface / Accessory Items h Various documents were written specifically for identifying the various quality and code requirements for mechanical / piping systems in the nuclear industry. That situation has been discussed in the previously stated information. Even though these documents have been changed radically, and requirements are continually changing, the mechanical system quality re-quirements were essentially established early. Despite the fact that the requirements for mechanical systems are not completely clear, the require-ments for structural and various miscellaneous mechanical items is even less clear. This portion of the discussion will attempt to clarify the situation.
Items such as lubricants, seals, packing, gaskets and various other associ-ated items were never covered by the codes and requirements previously discussed. ASME Section III and other piping codes, in fact, specifically exclude such items. It is obvious, however, that if any of these items are used improperly, serious equipment damage or failure could result. It is apparent that some sort of rule or procedure should be established to handle these miscellaneous and interface types of items.
As already indicated, mechanical systems such as the HVAC are not covered by piping codes. The actual status of the HVAC systems at Fort Calhoun Station are difficult to determine. The containment HVAC and H2 purge are definitely engineered safeguards, and specific quality requirements apply.
When considering the NRC Regulatory Guide 1.52, the Control Room, the Spent C' Regulatory Guide 1.52 the containment HVAC is considered a " primary system" whereas the Centrol Room and the Spent Fuel /SI Pump Room HVAC would become
" secondary systems". All other HVAC systems in the plant would be consid-ered as non engineered safeguard systems as far as NRC guidelines are concerned. When reviewing the Fort Calhoun FSAR, it is apparent that the requirements that were applied are somewhat different than what is indi-cated by the NRC guidelines. Consider, also, that like many of the systems at Fort Calhoun Station, the HVAC system was designed and installed prior to the now existing NRC guidelines.
Structures, also have a confusing situation as to what is required from a quality standpoint. The containment, because of its obvious safety function is considered to be required to meet a high level of quality.
Because it not only acts as a structure but also as a pressure vessel, the l
requirements of ASME Section III apply. Also various NRC Regulatory Guides now exist which cover any aspects of quality that aren't discussed by ASME Section III. The requirements for other structures aren't as clearly identi-l fied. All structures are primarily support systems and therefore must be designed to withstand various loading combinations. The various design loading requirements are essentially the same for any support system. One design criteria that makes significant differentiation between various
, support requirements is the scismic design requirements. The seismic l requirements therefore have been used to identify, or at least assist in identification of the various quality requirements for structural systems.
tO l
" DRAFT At the time of design and construction of Fort Calhoun Station, the FSAR itemized various system, components and structures under a Seismic Class One requirement. Although the FSAR Seismic Class One definition closely resembles the Regulatory Guide 1.29 Seismic Category One definition, a review of various system design specifications, testing and documentation requirements reveal several subtle but important differences. It becomes apparent that at the time the FSAR seismic class one definition was written
, it was agsumed that if a system would not fail in a Maximum Hypothetical Earthquake it would remain functional. Since that time various codes and requirements have come about (such as ANSI-278.1-1975) that specifically talk about operability requirements, active components and maintaining a functional capability. In reality many of the systems classed by the FSAR as Seismic Class I'would fit more closely into paragraph C.2 of Regulatory Guide 1.29 after evaluating the systems original design, testing, and documentation requirements. Aside from the various mechanical systems discussed elsewhere, the only structure that comes under the Seismic Cate-gory One definition is the containment building. For systems or structures which have no active function or moving parts the requirements of C.2 are essentially the same,as C.I. ,
A copy of NRC Regulatory Guide 1.29, Rev. 3 is attached for ease of reference.
~
(D) Various Nuclear Component Code Comparisons To better understand the relationship of various codes and documents see Table I. A warning is given to be aware that these parallels between codes ,
cannot be taken as one-to-one comparisons. No two codes are exactly alike.
The parallels are shown only to give the individual a feel for what the
' boundary flags are trying to represent. Individuals are urged to refer to the documents referenced in Table I when they are seeking guidance on design or requirements for systems. Codes such as ANSI B31.1, B31.7, and ASME Section III apply only to pressure retaining components carrying water or steam. Systems such as HVAC and diesel systems do not necessarily come
- under the above codes and will have their own requirements. An attempt to identify individual systems and other basic code requirements has been attempted in a general sense. Special attention is given to systems like HVAC and the diesel systems. In many cases codes such as ANSI B31.1 or l ASME Sect. III will be used for systems containing other than steam or I water. This will usually be the result of being specified by other codes
- or requirements.
l l (1) The Maximum Hypothetical Earthquake (MHE) corresponds to what is now called the Safe Shutdown Earthquake (SSE).
l l
l l
1 l
TABLE I DRAFT Table Comparing Various Codes and Regulations and How They Relate to One Another**
ANSI B31.1 ASME Sect. III NA/CA NB NC ND NE NF NG 1977 1 2 3 EiC 1, 2, CS (Code Class) 3 & MC ANSI B31.7 I II III j ASME Sect. III-1963, 65 and 68 for pressure
- A C or C B vessels only ANSI-N18.2 1 2 3 2 Reg. Guide 1.26 ^
B G' ' D 10 CFR 50.55(a) ***
1 l
- ASME Section III - 1963, 65 and 68 was essentially a pressure vessel code and
[T Qj did not deal with components (pipes, valves, etc.). When retrofitting to a j mid 60's plant you must look at Class C vessels on a case by case basis
! because they may fit into different categories by today's standards.
- The above table is not intended to indicate that various codes may be equivalent but rather give the relationship of various code classes and terminology.
- Editions prior to 1971 of the ASME Boiler and Pressure Vessel Code,Section III, " Nuclear Power Plant Components", use the term Class A in lieu of Class I.
D
DRAFT (E) Conclusions and Recommendations A summary of the situation at Fort Calhoun Station is as follows:
- 1. The boundaries, in most cases, do not indicate ASFE Section III code class of a system since the original system was not built to the ASPE Section III code. These boundaries do indicate a relative difference in the function of components indicated by the different code classes.
- 2. If a component has been evaluated to fail for a reason other than a deficiency in the original design specification, the replacement of that component with one that meets the same specification is adequate unless there is a requirement by some other agreement or regulation to use a different code.
- 3. The rerouting of piping or the addition of new valves into a system shculd be done using the original design specifications that apply to that system unless there is a requirements by some other agrdement or regulation to tisd a different code.
- 4. The addition of a completely new system should be done using the codes that apply at the time of design and construction of that system. A major modification of a system may require the application of present codes.
- 5. The use of design techniques or requirements available in the ASME Section III code to design replacements or rerouting of piping to a system is allowed as long as all original require-ments are accounted for.
- 6. The itemization of all design requirements on a purchast specification or design specification is recommended rather than statements to the effect of " Supply one 6 inch 600 lb.
500 F weld end ASME Section III Class 2 gate valve".
- 7. A procedure or rule for handling items associated with systems or equipment that have safety significance, but are not covered by specific codes, etc. should be developed (e.g. lubricants, gaskets, etc.). This has been done in the following section.
- 8. Structural quality requirements should be itemized and a list developed. This has been done in the following section.
- 9. Systems not containing steam and water should be individually discussed since the code requirements are not ccmpletely clear. This has been done in the following section.
- 10. Drawings should be issued showing the quality class boun-daries. This has been done in Attachment I.
- 11. Rules or methods for using these drawings to establish OPPD quality requirements should be formulated. This has been done N, f') in the following section.
DRAFT II. ?!cchanical and Structural CQE Requirements This section incorporates the recommendations and conclusions reached in the previous section.
This section establishes the CQE lists for mechanical and structural systems and components at Fort Calhoun Station including directions and clarifications for its use.
(A) Definition of CQE & Limited CQE, and Examples Critical Quality Elements (CQE) are defined as those structures, systems, components or items whose satisfactory performance is required to prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public.
1 NOTE There are several items that, functionally, could very easily fit the above definition. Frora a practical standpoint, items such as gaskets, lubricants, packing and other consufnable types of items would be difficult, if not impossible, to procure under the strin-gent requirements set for CQE items. Likewise, many building structures such as the auxiliary building need to have the versa-i tility to have minor modifications made to it on a periodic basis.
Therefore these items serve functions which are essential to the integrity of various CQE items but may not be capable of practical procurement under the CQE requirements. They are generally associ-ated with CQE items.
Limited Critical Quality Elements (Limited CQE) are defined as
~
those structures, systems, components or items whose satisfactory
' performance is required to prevent or mitigate the failure of those structures, systems components or items identified as CQE.
f In order to clarify the categories of limited CQE and CQE the following examples and discussions are presented. This is not to be construed as represerting an unabridged list.
Examples:
- 1. The containment acts as a support for many safety related items. In thi:: respect it may be considered more in the category of Limited CQE. Iloweve r , it also acts as a pressure retaining barrier preventing the release of radioactive iso-topes in the event of an accident. In this respect it is performing an active function directly mitigating the conse-quences of an accident. It becomes obvious that the contain-ment is definitely a CQE structure and pressure vessel.
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iDRAFT
- 2. As a comparison with the first example, consider the Auxiliary Building. Many safety related systems, such as the safety injection system are located within this building. The build-ing serves to support these various systems and to protect them from environmental hazards such as tornadoes, floods, etc. It does not have an active function to directly mitigate an accident. The building although contributing to the ability of various systems to carry out safety related functions does not serve a direct safety related function such as the contain-ment performs. The Auxiliary Building is therefore in the category of Limited CQE.
- 3. Various valves are identifed by the Mechanical CQE list as being CQE. The codes that typically identify the quality requirements for valves specifically exclude gaskets, packing and other items that are often associated with valves. These items, in themselves, may not contribute directly to safety.
However, it is obvious that if incorrect accessories to valves are used, it may prevent the valves from carrying out their safety function'. These items will then be identified as-limited CQE.
This concludes the definitions and examples for CQE and Limited CQE. Specific guidelines for the identification and requirements of CQE and Limited CQE items will be given in the Mechanical CQE and Structural CQE portion of this document.
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- ATTACHMENT I :.:i.
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The following drawings have been labeled with class boundary flags'as' des'c"r'ib'e'd :x- '*
in Section I, II and Table II. They are to be used to aid in In-Service-Inspections and consulted when adding, modifying or replacing y M a drawings should not be used *.o (see Section I). Although some items determine the design are classified as ASME codeSection of III anc'omponents existing 'comp the majority of items were built to different codes. For an exact specification, w
. The consult revision humber, consult Aided Retrieval System.
the FORT valve CAI.HOUNindex, latest revision of these drawings should be referred. Ifindoubtrega'rding7,$3 the P&ID in the control room materials or refer to computerlist, g..e '
Drawine No. Title
- Revision ifk5hi@
.. . . .,4 rt : w v w D-4078 Reactor Coolant Gas Vent System ._~3,..u$.ww,.h[$nW^Ijk.%
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Nw -
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E-23S66-210-110 Reactor Coolant System P&I ' - . %. -s3)?Ij J' Diagram .m' j'g"14 3 vsig$gg
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D-23866-210-111 Reactor Coolant Pump P&I . ' ,' . , s.:.lf,;@,ijj.)$g{
Diagram (14 :gh 7';.jft.ly'pg p E-23866-210-120 Chemical & Volume Control .. '.- . .' ;'::s:w f M. .i. s'D,7% ed@g&%
System P&I Diagram Sh.1/Sh.2 16/11+ Ms.y%.4:
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(h;m u..; p :
E-23S66-210-121 Chemical & Volume Control c A{QP Sh. 2 System P&I Diagram 14 . gjA '; i };W '
l E-23866-210-130 Safety Injection and Contain- " ' M;' l.; : ? , c;m:x
. $"sE92 Sh.1/Sh.2 ment Spray System P&I Diagram 16/16 . i[$/7 .
'.R - U*.&4:
11405-M-1 Containment Heating, Cooling .
, c.j.:.7; and Ventilation Flow Diagram 20 _, ,g . .y.y,,:. .
11405-M-5 Demineralized Water System '
M' Sh. 3 Flow Diagram '
18 " Mr[' '
11405-M-6 Waste Disposal System Flow . _ '% v ' .
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l Sh. I Diagram 19 W8
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11405-M-7 Waste Disposal System Flow ._ is5 . .
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! Sh. 2 Diagram 20
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11405-M-8 Waste Disposal System Flow Sh. 3 Diagram 19 . . l ,_ -
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Revision 3 A-I-I April 1932 g.,. , _ - , , _ , . - _ _ _ _
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DRAFT Drawine h*o. Title Revision 11405-M-10 Auxiliary Coolant, Component Sh. 1 Cooling System Flow Diagram 14 11405-M-40 Auxiliary Coolant, Component Sh. 2 Cooling System Flow Diagram 10 11405-M-11 Auxiliary Coolant, Spent Fuel
. Pool Cooling System Flow Diagram 11 11405-M-12 Primary Plant Sampling System Flow Diagram. 15 11405-M-13 Plant Air Flow Diagram 10 11405-M-42 Nitrogen & Hydrogen Gas Flow Diagram 20 11405-M-98 Waste Disposal System Flow Diagram 18' 11405-M-100 Raw Water Flow Diagram 17 i 11405-M-119 Auxiliary Coolant, Component Cooling System Control Element Drive Mechanism Flow Diagram 7 11405-M-252 Flow Diagram, Steam 18 1
11405-M-253 Flow Diagram, Steam Generator Feedwater and Blowdown 26 11405-M-254 Flow Diagram, Condensate 19 11405-M-262 Flow Diagram, Fuel Oil 11 Sh. I 11405-M-262 Flow Diagram, Turbine Oil 1 Sh. 2 -
B120F06001 Fuel Oil System Schematic (Schoomaker Co.) G All of these drawings have been thoroughly reviewed for correctness and completeness. Notification of minor errors will be issued as they are discovered and will be corrected on routine revision. Major errors are subject to immediate revision.
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Revisien 3 A-I-2 April 1952 1
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' Ng\i TI (B). Mechanical CQE
((, This portion of the document will identify the CQE mechanical items and establish guidelines for CQE requirements. Likewise the Limited CQE requirements for mechanical will be given. These may vary somewhat from structural requirements and are therefore given separately.
i Various codes, identifying quality requirements for steam and water retaining components, have existed for some time. Details of this situation were discussed more fully in Section I of this document.
These codes are used to establish the quality requirements of most of the Mechanical CQE items. Part 1 of the following discussion establishes the general guidelines for all Mechanical CQE systems which are also the specific guidelines for steam and water pressure rating systems. Part 2 discusses, in more detail, certain mechani-
- cal systems for which code / quality requirements have not been as clearly established.
i 1. Steam and Water Retaining Systems & Component Requirements, and General Requirements c
- a. The class boundaries have been established (keeping in mind the discussion in Section I) by the P & ID's listed in Attachment I. Pumps, valves, pipes and pressure vessels, etc. identified by the drawings as ASME Section III Class 1, 2 or 3 or ANSI N18.2 Safety Class (See Table II) shall be
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considered to be CQE.
The following procedure is recommended in analyzing the class boundaries on the P & ID's.
(1) The class boundaries are indicated by diamond shaped
" flags". Various classes, and ultimately CQE bounda-rics, are differentiated in Table II.
(2) Any component that is located within a class boundary, ASME Section III or ANSI N18.2 safety class, will be considered CQE. Any component that is located in a non-class area will be considered non-CQE. -
(3) For a transition from a higher class to a lower class or a class to a non-class, the boundary is ordinarily established at an isolable point such as a valve or a flange. The boundary valve or flange should meet the llIGIER class requirements. The ,
first weld and support after the valve or flange should also meet the llIGlER class requirements.
(4) Non-class instrumentation should be isolated at the NEAREST valve or other isolation point. Instruments are NOT isolation points.
n a gman N. Eb (5) The middle of a section of pipe is NOT an isolation point.
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(6) The tubing and shell of a heat exchanger can repre-sent two different classes or a class and non-class.
(7) All containment penetration piping shall be desig-nated with a MINIMUM of ASME Section III Class 2.
All sleeves of penetrations shall be designated as ASME Section III Subsection MC. Penetration boun-J daries shall be determined in accordance with 10 CFR 50 Appendix A Part 5.
- b. As previously discussed, certain items associated with various mechanical CQE items will have to meet, as a MINIMUM, Limited CQE Requirements. The following list contains examples of such items. (This list is for example only and may not be complete).
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(1) Valve Operators and Controllers (2) Position Indicators (3) Pump Impellers and Drives (4) Permanently Sealeo Fluid-Filled Tubing Systems
! (5) Consumable Items Such as Gaskets, Packing, and i
Lubricants, etc. (this does not include weld rod).
> lh (tj IMPORTANT NOTE
! An important point should be made concerning the above items, particularly with respect to valve operators, controllers, position indicators, pump impellers and i drives. The point is that in recent years (since 1975) i it has become clear that certain operability requirements must exist if a component is to carry out its safety
! function. In various cases some of the accessories in the above list would be identified by ANSI N278.1-1975 as
" active" components and therefore must meet CQE require-ments in order to satisfy operability requirements.
- c. " Limited CQE", when applied to an item, will require as a MINIMUM the following type of quality requirements.
(1) The item is materially compatible (where appli-cable) with the environment in which it will be I placed. (e.g. Packing to be used in systems of' high temperature, high pressure stainless steel should be free of chlorides). A statement by the manufacturer or catalog information or engineer's analysis is acceptable to confirm compatibility.
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0i b (2) Verification upon receipt that the item received (Ag"- is the item that was ordered. This may consist of checking the item identification number .
against the catalog number or making a dimen-sional check.
(3) Verification that the item identified for installation is, in fact, the item that is being installed.
(4), " Limited CQE" items do not have to be ordered from the Approved Vendors List. They do not have to have 10 CFR 21 applied to them. The item should be verified by Q.A. or Q.C. (either on the purchase order or at some point prior to installation) to meet the above listed require-ments. Verifications should be recorded and retained until the item is replaced.
TABLE II ,
SW1BOLS USED FOR CQE IDENTIFICATION IN P & ID'S E
( Symbol Indicates Class Solid Indicates Non-Class Example:
Indicates Transition From Non-Class to ASME Section III Class 2 Equivalent Symbol Equivalent Class 1 ASME Section III Class 1 2 ASME Section III Class 2 ,
3 ASME Section III Class 3 S Safety Class Defined by ANSI N18.2 Safety Classes 2 & 3 tO
DR r=,a r
- 2. Other Mechanical Systems h Because of the difficulty identifying the precise requirements of some systems, they are discussed individually in this portion of the report.
- a. HVAC Of the various HVAC systems existing at Fort Calhoun Station, only a few areas are claarly CQE type systems. The Containment HVAC and all the various penetrations through containment out to and including the isolation valves are considered to be CQE (Fans VA2A and VA2B are non-CQE). This is indicated on Drawing 11405-M-1. The H2 Purge system is also considered to be CQE, as part of the Containment HVAC.
As discussed by Section I, certain HVAC systems such as Control Room HVAC, Fuel Handling HVAC, and SI Pump Room HVAC systems are
, . considered by Regulatory Guide 1.52 to be secondary ESF (Engi-neered Safeguard Feature) systems. Because these systems have a safety significance but are not required to meet the same strin-gent requirements as the Containment HVAC, they will be required to meet, as a MINIMUM, Limited CQE Requirements. Because of the lack of clearly established guidelines dealing with these systems, it is further recommended that any major changes or alterations to these systems be thoroughly evaluated at that time as to CQE status. All other HVAC systems are non-CQE.
b The following table is intended to summarize the above discussion.
HVAC CQE REQUIREMENTS
SUMMARY
CQE Applicable Documents System Requirements (Minimum Guidelines)
Containment HVAC CQE Original Spec. or Reg. Guide 1.52 Fans VA2A & VA2B Limited CQE Original Spec. or Reg. Guide 1.52 SI Pump Room Limited CQE Original Spec. or Reg. Guide 1.52 Fuel Handling Area Limited CQE Original Spec. or Reg. Guide 1.52 Control Room Limited CQE Original Spec. or Reg. Guide 1.52 Remaining Aux- Non-CQE Original Spec. or Reg. Guide' 1.140 liary Building HVAC Controlled Access Remaining Aux- Non-CQE Original Spec. or Accepted iliarv Building Industry Practice HVAC Non Con-trolled Access
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- b. Fuel Oil System for Standby Diesel Generators e
The fuel oil systems for the standby emergency diesel generators shl1 be designated as ASME Section III Class 3 for purposes stated in Section I of this document. The basis for this decision is ANS-59.51/N195-76 Section 7.
The boundary extends from the 15,000 gallon storage tank (FO-1) up to and including all piping in the injection systems. Included in the boundaries are all pipes and pipe supports, valves, tanks (except concrete tanks),
pumps, strainers, filter housings, and all fittings ,
necessary in the system. Filter internals are classed as '
" Limited ~CQE". Fuel oil shall meet the requirements set I forth by Regulatory Guide 1.137, or as agreed to between ;
OPPD and the NRC.
- c. Plant Air System The only portion of the plant air,syst.em that is to be classified as CQE is the containment penetration boundary at penetration #74. This system is defined as a " closed system" by 10 CFR 50 Appendix A Part V and shall there-fore include only one automatic, locked closed, or remote manually operated valve outside containment for isola-tion. The CQE portion of the system will begin at HCV-1749 and terminate at the valve CA-555. The precise boundaries are at the appropriate welds on the valves.
For non steam and water pressure retaining piping such as the piping in the plant air system, containment penetra-tions are defined by ANSI N18.2 Section 2.3 as Safety Class 2. But since the penetration piping was designed and fabricated to a steam and water pressure retaining piping code (USAS B31.1), the penetration piping shall be designated as ASME Section III Class 2 rather than Safety Class S.
- d. N nd H Systems 2 2 .
Like the plant air system, the containment penetration points of the nitrogen and hydrogen systems are desig-nated as ASME Section III Class 2 and are therefore CQE.
Since the nitrogen system vents directly to containment atmosphere, the lin.e is defined as a Primary Contairment Isolation by 10 CFR 50 Appendix A Section 5 and contains one automatic isolation valve inside and one outside containment. The other class boundaries designated in Drawing 11405-M-42 are lines to isolation point valves from class tanks. The reactor containment hydrogen control system is included in the HVAC system.
DRAFT (C) Structural CQE As was discussed in Section I of this document, the definition of structural requirements is not always clearly determined and the seismic requirements have been clouded by the attempt to compare the FSAR Seismic Class One definition with the Seismic Category One (from Regulatory Guide 1.29) definition. Considering the information already presented, the following table has been developed to establish the CQE requirements.
CQE Structures Reg. Guide 1.29 Code Class Or Item Requirements Requirements
- 1. Supports for CQE Sect. C.1 Original Spec. or Piping Systems ASME Sect. III Subsect. NF
- 2. Containment Sect. C.1 Original Spec. or ASME Sect. III Subsect. MC and Reg. Guide 1.94 and
, 1.29, C.I.
- 3. Auxiliary Build- Sect. C.2 Contract 1237 Spec. and ing Crane Reg. Guide 1.104
- 4. Fueling Machine Sect. C.2 Original Spec.
- 5. Spent Fuel Handling Sect. C.2 Original Spec.
Machine
- 6. Containment HVAC Sect. C.1 Original Spec. or Supports Reg. Guide 1.52 & 1.29 C.1
- 7. Polar Crane Runway Sect. C.2 Original Spec or ASME Sect. III Subsect. MC Limited CQE Structures Ren. Guide 1.29 Code Class Or Item Requirements Requirements
- 1. Auxiliary Building Section C.2 Original Specifi-and Intake Structure cations or Reg.
Guide 1.142 and applicable portions of 1.94.
f
- 2. All Radiation Shield Section C.2 Original Specifi-Walls cation or Reg. Guide
( 1.142 and Applicable Portions of 1.94
, S.~ 7 Ud -L'I3 III. PERTINENT REFERENCES TO THIS DOCUMENTS
- 1. ANSI - B31.1
- 2. ANSI - B31.7
- 4. ANS N18.2
- 5. ANS-59.51/N195-76 Section 7
- 6. NRC Regulatory Guide 1.26
. 7. NRC Regulatory Guide 1.29
- 8. h7C Regulatory Guide 1.52
- 9. NRC Regulatory Guide 1.94
- 10. NRC Regulatory Guide 1.140
- 11. NRC Regulatory Guide 1.142
- 12. AS."E Section III 1963, 65 and 68 Code
- 13. ASME Section III 1977 Code
- 14. ASME Section XI, Article 7000-Replacements
- 15. 10 CFR 50.55(a) j 16. 10 CFR 50 Appendix A -
f 17. Fort Calhoun Station FSAR ,
l 18. Fort Calhoun Station Valve Index
! 19. Fort Calhoun Station Mechanical P&ID's l 20. Fort Calhoun Station Bill of Materials 4 21. All Contracts Pertinent to Fort Calhoun Station l 22. ERDA 76-21 i
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Revision 3 s April, 19S2
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