ML18018B450
| ML18018B450 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 11/07/1983 |
| From: | Cutter A CAROLINA POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| GL-83-28, LAP-83-516, NUDOCS 8311110159 | |
| Download: ML18018B450 (84) | |
Text
HARRIS NUCLEAR PROJECT RESPONSE TO GENERIC LETTER 83-28 "REQUIRED ACTIONS BASED ON GENERIC IMPLICATIONS OP SALEM ATWS EVENTS" NOVEMBER 4, 1983 R3llll015q ~
(8434DCWccc)
I I'
p P
Page 1 of 19 1.1 POST-TRIP REVIEW (PROGRA>j DESCRIPTION AND PROCEDURE NRC POSITION Licensees and applicants shall describe their program for ensuring that unscheduled reactor shutdowns are analyzed and that a determination is made that the plant can be restarted safely. A report describing the program for review and analy-sis of such unscheduled reactor shutdowns should include, as a minimum:
- 1. The criteria for determining the acceptability of restarts
RESPONSE
SHNPP has under development a post-trip review procedure for implementation by the on-shift operating personnel for un-scheduled shutdowns (Appendix A). This procedure includes evaluation of plant conditions prior to the trip, first out annunciator, strip chart recorders on the main control board, equipment failures or transient conditions which occurred during the transient, and a review, if available, of the computer log of the event. The general criteria for determin-ing the acceptability of a restart will be (1) the cause of the event has been identified, (2) appropriate corrective action is taken to reduce the recurrence of a trip, and (3) engineered safeguards responded in a manner consistent with the Technical Specifications. If the cause of the trip cannot be determined by the shift staff, the Post-Trip Review Report will be for-warded to the Plant Nuclear Safety Committee (PNSC) for review and recommendations. In this case, the determination of whether a unit can be restarted is made by the General Plant Hanager.
NRC POSITION
- 2. The responsibilities and authorities of personnel who will perform the review and analysis of these events.
RESPONSE
The responsibilities and authorities of personnel who will perform the review and unscheduled trips are specifically delineated in the SHNPP post-trip review procedure, Section 3.0.
The Shift Foreman is responsible for completing the post-trip review report and for determining if the cause of the trip is known and corrected. The Shift Foreman has authority to restart if the cause of the trip has been determined to be not due to equipment or instrum'entation malfunction and has been corrected.
WKLB04
Page 2 of 19
- b. The Shift Technical Advisor will assist the Shift Foreman in the completion of the post-trip review report and must concur that the cause of the trip is known and corrected prior to restart.
The Operating Supervisor and the Hanager - Operations will review the post-trip review report. These reviews may be subsequent to restart as appropriate. The Operating Supervisor and/or the Hanager - Operations, as available, will be consulted and grant approval for restart for any trip for which the cause is known to be due to malfunction of equipment or instrumentation and has been corrected.
- d. The SHNPP Plant Nuclear Safety Committee (PNSC) will review and make recommendations on those unscheduled trips for which the cause is not known or has not been corrected.
- e. The Plant General Hanager will approve restart, based on the recommendations 'of the PNSC, for those unscheduled trips for which the cause has not been identified or corrected.
NRC POSITION
- 3. The necessary qualifications and training for the respon-sible personnel.
RESPONSE
The training and qualifications for personnel at the SHNPP are addressed in FSAR sections 13.1.2 and 13.2. As a minimum, the Shift Foreman, the Operating Supervisor, and the Hanager-Operations will hold NRC SRO licenses. The members of the SHNPP PNSC are identified in the draft Technical Specifications previously submitted to the NRC ~
NRC POSITION The sources of plant information 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 and in sufficient detail for proper understanding. (See Action 1.2)
RESPONSE
The essential sources of information which must be evaluated during the post-trip review process are identified in the post-trip procedure under development (Attachments 6.2 through 6.5 of Appendix A). The essential sources include safety-related instrumentation displays and recorders. This instru-mentation is identified and described in Section 7.0 of the FSAR. Other sources of information which may be available VKIB04
Page 3 of 19 include the Emergency Response Facility Information System (ERFIS), the plant computer (a subpart of ERFIS), the Safety Parameter Display System (SPDS), eyewitness accounts, and logs/records maintained at SHNPP.
The ERFIS has been the subject of submittals to the NRC in response to NUREG 0737, Supplement No. l. A brief description of the ERFIS is presented zn Section 1.2.1 below.
NRC POSITION
- 5. The methods and criteria for comparing the event informa-tion with known or expected plant behavior (e.g., that safety-related equipment operates as required by the Technical Specifications or other performance specifica-tions related to the safety function).
RESPONSE
The post-trip review procedure includes checklists (Attachments 6.2 and 6.3 of Appendix A) to review performance of engineered safeguards, such as the safety injection system and the reactor trip system, during the trip. The performance of the engi-neered safeguards will be compared with the required response as delineated in the SHNPP Technical Specifications.
NRC POSITION
- 6. A) The criteria for determining the need for independent assessment of an event (e.g., a case in which the cause of the event cannot be positively identified, a competent group such as the Plant Operations Review Committee, will be consulted prior to authorizing restart) and B) guide-lines on the preservation of physical evidence (both hardware and software) to support independent analysis of the event.
RESPONSE
6.A The completed forms from the post-trip review will be reviewed by at least two plant personnel with current SRO licenses. If this on-shift review cannot positively identify the cause of the trip and appropriate corrective actions, the matter will be referred to SHNPP operations management and the PNSC for evaluation prior to restart.
The decision to restart will be made by the Plant General Hanager with recommendations from the PNSC.
6.B The post trip review report and other applicable software data will be maintained as a permanent record of the event. Guidelines for preservation of physical evidence will be incorporated in the post-trip review proc'edure.
VKLB04
Page 4 of 19 NRC POSITION
- 7. Items 1 through 6 above are considered to be the basis for the establishment of a systematic method to assess unsched-uled reactor shutdowns. The systematic safety assessment procedures compiled from the above items which are to be used in conducting the evaluation, should be in the report.
RESPONSE
NRC positions 1 through 6 were used as references when the SHNPP Post Trip/Safeguards Review procedure was developed.
Section 4 of the post trip review report requires the Shift Foreman to perform a thorough systematic evaluation of the event and the plants response.
1.2 POST-TRIP REVIEV DATA AND INFORHATION CAPABILITY NRC POSITION 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
- 4) Format for displaying data and information
- 5) Capability for retention of data and information
RESPONSE
The sequence of events during a trip can be determined using the "first out" panel for the reactor trip system; strip chart recorders in the main control room; and the plant computer, if available.
The plant parameters which are recorded on strip charts are indicated in Section 7.5 of the SHNPP FSAR. Refer to the following tables for specific information on the noted systems:
Table 7.5.1 Containment Spray and Cooling System Table 7.5.1 Auxiliary Feedwater System Table 7.5.1 Control Room Emergency Filtration System WKLB04
Page 5 of 19 7.5.1 Table Table Table 7 '.
7.5.1-10 RAB Emergency Exhaust System 1 Reactor Coolant System
- Containment System Table 7.5.1 Hain Steam System Table 7.5.1 Refueling Mater Storage Tank Table 7.5.1 Component Cooling Mater. System Table 7.5.1 Nuclear Instrumentation The plant computer is an integral part of the SHNPP Emergency Response Information System. The ERFIS system has been described in CP&L's response to NUREG-0737 Supplement 1. The plant computer has the capability of monitoring 1500 inputs. The list of specific parameters to be recorded has not been finalized, but will be avail-able by June 1984. The time discrimination between events averages one milli-second; the maximum discrimination is two milli-seconds. The output of the plant comput: er can be displayed on control room CRT's or printers; a second CRT and printer is provided as a backup in the control room. Data for each input is stored for twelve hours.
This data can be manually transferred to magnetic tape.
The power sources for the computer consist of non-class IE AC power and a 30-minute battery powered backup.
NRC POSITION 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 after trip)
- 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)
RESPONSE
Refer to the response to item 1.2.1.
MKLB04
Page 6 of 19 NRC POSITION
- 3. Other data and information provided to assess the cause of unscheduled reactor shutdowns.
RESPONSE
Refer to discussion under 1.1.6.B and 1.2.1.
NRC POSITION
- 4. Schedule for any planned changes to existing data and information capability.
RESPONSE
The systems and hardware identified above are being designed to reflect NRC's post-THI recommendations on such systems.
Additions to these systems are not contemplated at the present time.
2.1 E UIPHENT CLASSIFICATION AND VENDOR INTERFACE (REACTOR TRIP SYSTEH COHPONENTS NRC POSITION Licenses 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 maintenance, work orders, and parts replacement. In addition, for these components, licensees and applicants shall establish, implement, and maintain a continu-ing program to ensure that vendor information is complete, current, and controlled throughout the life of the plant, and appropriately referenced or incorporated in plant instructions and procedures. Vendors of these components should be contacted and an interface established. i<here vendors cannot be identi-fied, have gone out of business, or will not supply the informa-tion, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, to assure reactor trip system reliability. The vendor interface program shall include periodic communication with vendors to assure that all applicable information has been received. The program should use a system of positive feedback with vendors for mailings containing technical information. This could be accomplished by licensee acknowledgement for receipt of techni-cal mailings. The program shall also define the interface and division of responsibilities among the licensee and the nuclear and nonnuclear 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.
WKLB04
Page 7 of 19
RESPONSE
Confirmin Com onents as Safet -Related To aid operators and craft personnel to more easily identify safety related components, SHNPP is in the process of compiling the current list of safety-related components into a definitive Q-List document. Existing Ebasco (A/E) and Vestinghouse (NSSS) design documents and drawings list this information. As the Q-List is completed, the data will be reviewed to reconfirm that all components whose functioning is required to trip the reactor are identified as safety-related. This is scheduled to be complete by December, 1984. The Q-List will be used by operations and maintenance personnel while preparing final plant procedures and other safety-related activities.
B. Documentin Safet -Related Status Q-List procedures and control instruction will be con-tained in the Plant Operating Hanual. Plant procedures will ensure components whose functioning is required to trip the reactor are identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, including maintenance work orders and parts replacement, prior to plant operation.
As the Q-List identified in A above is prepared listing safety-related components, this information will be entered into the SHNPP Equipment Data Base System (EDBS),
which serves as the computerized listing of all plant equipment. Appendix B gives an overview of the data elements contained in this system. The EDBS will be used as the data base for ordering spare parts, providing schedule input for maintenance/testing activities, provid-ing input to maintenance work requests for use by the planner/analyst, and providing the material history records. Data entry is strictly controlled by appropriate plant procedures to preclude nonapproved changes or entry.
Vhen components are identified as safety-related, this information will be automatically available when work functions are planned.
This centralized system therefore provides a summary of each component's quality related designations, reference listing of related documentation such as procedures or technical manuals, spare parts information, and material history records, thus closely integrating all plant activities associated with that item. The format of the EDBS system has been developed and implemented at SHNPP, with data entry in progress. The system will be fully operational by June, 1985.
WKLB04
'l Page 8 of 19 C. Vendor Interface The Westinghouse Electric Corporation designed and provided the major Reactor Protection System components at SHNPP.
Westinghouse currently utilizes a Technical Bulletins System that documents recommended changes in equipment and procedures. These bulletins also provide information concerning unique operating conditions and experience at other PWR plants. All distributions of Westinghouse safety-related Technical Bulletins are now accompanied by a return receipt. The return receipts are pre-addressed to Westinghouse for recording Bulletins transmitted and their status. Technical Bulletins for which receipt is not acknowledged within a reasonable time are retransmit-ted. On a periodic basis, a list of current Technical Bulletins is issued by Westinghouse. This provides a positive feedback mechanism to ensure applicable Westing-house technical information has been received.
SHNPP is currently performing a review to verify that all Westinghouse Technical Bulletins issued to SHNPP have been received and implemented as appropriate. This review is currently scheduled to be completed by June 1985. This process ensures that the Westinghouse technical informa-tion is appropriately reviewed and incorporated, as applicable, in the plant instructions and procedures prior to and following initial plant operation.
The vendor interface program and plant procedures ensure that, adequate controls are established for the Reactor Protection System components in determining that the vendor information is complete, current, and controlled prior to plant operation and throughout, the life of the plant, and appropriately referenced and incorporated into the plant instructions and procedures.
For safety-related equipment requiring environmental qualification, a program is in progress to systematically review vendor data packages to identify unique features that need to be incorporated into the spare parts and maintenance activities. This will be completed by Decem-ber, 1984.
WKLB04
Page 9 of 19 2.2 E UIPHENT CLASSIFICATION AND VENDOR INTERFACE PROGRAHS FOR ALL SAFETY-RELATED COHPONENTS NRC POSITION Licensees and applicants shall submit, for staff review, a description of their programs for safety-related>'quipment classification and vendor interface as described below:
For equipment classification, licensees and applicants shall describe their program for ensuring that all compo-nents of safety-related systems necessary for accomplish-ing required safety functions are identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, including maintenance, work orders and replace-ment parts. This description shall include:
- 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.
- 2) A description of the information handling system used to identify safety-related components (e.g., compu-terized equipment list) and the methods used for its development and validation.
- 3) A description of the process by which station person-nel use this information handling system to determine that an activity is safety-related and what proce-dures for maintenance, surveillance, parts replace-ment and other activities defined in the introduction to 10CFR50, Appendix B, apply to safety-related components.
- 4) A description of the management controls utilized to verify that the procedures for preparation, valida-tion, and routine utilization of the information handling system have been followed.
Safety-related structures, systems, and components are those that are relied upon to remain functional during and following design basis events to ensure: (1) the integrity of the reactor coolant boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, and (3) the capability to prevent or mitigate the consequences of accidents that could result in potential off-site exposures comparable to the guidelines of 10CFR Part 100.
VKLB04
Page 10 of 19
- 5) A demonstration that appropriate design verification and qualification testing is specified for procure-ment of safety-related components. The specifica-tions shall include qualification testing for expected safety service conditions and provide support, for the licensees'eceipt of testing documentation to support the limits of life recommended by the supplier.
- 6) Licensees and applicants need only to submit for staff review the equipment classification program for safety-related components. Although not, required to be submitted for staff review, your equipment classi-fication program should also include the broader class of structures, systems, and components impor-tant to safety required by GDC-1 (defined in 10CFR Part 50, Appendix A, "General Design Criteria, Introduction" ).
For vendor interface, licensees and applicants shall establish, implement, and maintain a continuing program to ensure that vendor information for safety-related compo-nents is complete, current, and controlled throughout the life of their plants, and appropriately referenced or.
incorporated in plant instructions and procedures.
Vendors of safety-related equipment should be contacted and an interface established. 'Where vendors cannot be identified, have gone out of business, or will not supply information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, to assure reliability commensurate with its safety function (GDC-1). The program shall be closely coupled with action 2.2.1 above (equipment qualification).
The program shall include periodic communication with vendors to assure that all applicable information has been received. The program should use a system of positive feedback with vendors for mailings containing technical information. This could be accomplished by licensee acknowledgment for receipt of technical mailings. It shall also define the interface and division of respon-sibilities among the licensee and the nuclear and nonnu-clear 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.
RESPONSE
Equipment classification for safety-related equipment has been accomplished using the guidelines established in ANSI 18.2 for fluid system components and IEEE standards for electrical components. This listing by system is provided in Table 3.2.1-1 of the FSAR. Additional documents 1<KLB04
Page 11 of 19 provide the reference to the component level; including the valve list, line list, and instrument indices.
As described in our response to 2.1, SHNPP is preparing a computerized listing of all plant equipment, both safety-related and nonsafety-related to better aid the operators and craft personnel.
The data base provides sufficient information for all personnel to understand and apply the proper safety designations to all equipment in maintenance and opera-tions related activities. It will describe if the equip-ment is safety-related, what procurement requirements apply to individual parts, whether the equipment is environmentally qualified, which documents such as techni-cal manuals need to be referenced, and provides the storage of its material history. Data entry and manipu-lation is controlled by plant procedure 0-Tiki-101 which will require appropriate levels of approval prior to data entry.
Safety-related equipment requiring environmental qualifi-cation has been tested during initial purchase by the A/E, as described in FSAR Section 3.10 and 3.11. The documen-tation for the test results is being reviewed to verify that the suitability of the equipment for its expected environment, has been demonstrated and that unique aspects necessary to be incorporated into appropriate maintenance procedures have been identified. This w'ill be accom-plished by June, 1985.
The EDBS has been established to allow items to be also classified as "important to safety." SHNPP has not, however, committed to implementing a program for important to safety items. The results of the AIF working group addressing this issue are being followed; the results of which will be used as a basis for establishing a SHNPP approach.
Currently, SHNPP is phasing in a vendor interface program to provide adequate assurance that vendor inf'ormation for safety-related components, which is significant to safety, is appropriately incorporated in the plant instructions and procedures. This vendor interface program will include:
- 1) Procedural processing of vendor recommendations Processing of vendor recommendations will be in accordance with SHNPP procedures. Vendor recommenda-tions concerning plant equipment will be forwarded to plant engineering for evaluation and implementation, as appropriate, in plant instructions and procedures.
The vendor recommendations will also be forwarded to 4'KLBOA
Page 12 of 19 the On-Site Nuclear Safety Committee and included in the Operating Experience Feedback Program, if appropriate.
- 2) Procedural control of vendor technical manuals SHNPP maintenance activities will be performed in accordance with approved plant procedures. The vendor technical manuals are used as a source of reference material in preparing these procedures.
This procedural control provides for review and incorporation of technical manual information into the plant maintenance procedures. Thus, the vendor input is evaluated before use and integrated with site specific conditions and experience. In some cases where the equipment is complex and the vendor manuals are suitable for direct use, the particular section of the manual that provides instructions for accomplishing the desired activity will be referenced or incorporated as part of the approved procedure.
In such cases, the SHNPP procedure will require that the technical manual be plant approved.
SHNPP is in the process of developing a procedure for the review of safety-related vendor technical manuals in the preparation of the plant operating and mainte-nance instructions and procedures. This procedure is scheduled to be completed by Narch, 1984.
- 3) Control of the vendors manuals This will be accomplished by incorporating the vendor technical manuals in the document control process as controlled documents.
The vendor interface program with the SHNPP Nuclear Steam Supply System (NSSS) vendor is described in response to 2.1.
SHNPP additionally utilizes vendor information from other industry sources, such as INPO NPRDS and Notepad, and NOi1IS. Significant industry-wide events are identified and recommended action provided in INPO Significant Operating Reports (SOERs). Signifi-cant events (without recommendations) are provided in INPO Significant Event Reports.
Additionally, the NRC issues notification of safety-related concerns and regulatory requirements. IE Bulletins, Notices, and Circulars provide current information concerning component or design discrep-ancies. These notifications are further supported by the provision of Title 10, Chapter 1, CFR Part 21, which requires "that the directors and responsible WKLB04
Page 16 of 19 DS-416 UVTA replacement with modified shaft widened grooves for the retaining ring was issued. Refer to NS-EPR-2753, E. P.
Rahe of Westinghouse to R. C. DeYoung of'RC. As indicated in the letter of April 21, Westinghouse has committed to its utilities to replace UVTAs on DS-416 reactor trip switchgear supplied by Westinghouse for its Nuclear Steam Supply System so that, 1) the new attachments have modified (widened) grooves to accommodate the new retaining rings, 2) manufacturing drawings have been revised and quality control procedures modified so that critical design dimensions are maintained during manufac-ture, and 3) a field installation procedure will be provided for proper alignment and interface of the attachment with the breaker trip shaft. These replacement devices will be marked with a serial numbering system.
These modifications will be completed by December, 1984.
4.2 REACTOR TRIP SYSTEH RELIABILITY (PREVENTATIVE HAINTENANCE AND SURVEILLANCE PROGRAH FOR REACTOR TRIP BREAKERS NRC POSITION Licensees and applicants shall describe their preventative maintenance and surveillance program to ensure reliable reactor trip breaker operation. The program shall include the following:
- 1. A planned program of periodic maintenance, including lubrication, housekeeping, and other items recommended by the equipment supplier.
- 2. Trending of parameters affecting operation and measured during testing to forecast degradation of operability.
- 3. Life testing of the breakers (including the trip attach-ments) on an acceptable sample size.
- 4. Periodic replacement of breakers or components consistent with demonstrated life cycles.
RESPONSE
- 1. Based on the latest results of the life cycle testing program, and the recommendation(s) of Westinghouse, SHNPP will implement a comprehensive planned maintenance program covering the reactor trip breakers. This will be completed by December, 1984.
- 2. Incorporated with the planned maintenance program will be a review of the parameters necessary to be trended. These parameters will then be included under the overall plant trending program, which will be in operation by December, 1984.
WKLB04
Page 17 of 19 Life cycle testing of the shunt trip attachment and the undervoltage trip attachment of the reactor trip switch-gear is being conducted by Westinghouse for the Westing" house Owners Group. This program is aimed toward estab-lishing the service life of these devices, and substanti-ating periodic test requirements with proper maintenance.
The results of this program will be factored into mainte-nance, replacement, and qualification programs. The test program is scheduled for completion by December, 1984.
- 4. The need for periodic replacement of breakers or compo-nents will be determined based on the results of life cycle testing and incorporated into the planned mainte-nance system by June, 1985.
4.3 REACTOR TRIP SYSTEH RELIABILITY (AUTOi~fATIC ACTUATION OF SHUNT TRIP ATTACHHENT FOR WESTINGHOUSE AND B&V PLANTS NRC POSITION Vestinghouse and B&V reactors shall be modified by providing automatic reactor trip system actuation of the breaker shunt trip attachments. The shunt trip attachment shall be con-sidered safety related (Class IE).
RESPONSE
SHNPP is working closely with the Westinghouse Owners Group (WOG) in addressing the actions relating to Reactor Trip Syst: em reliability, particularly with respect to the requirement for automatic actuation of the shunt trip attachment and the on-line surveillance requirement. A detailed generic design package for incorporation of an automatic shunt trip feature into various Vestinghouse Reactor Protection Systems has been developed under VOG sponsorship. The complete generic design package of the automatic shunt trip modification was submitted to NRC on June 14, 1983, J. J. Sheppard, Chairman of VOG by letter OG-101.
The generic design package of the automatic shunt trip modi-fication contains a design basis, functional requirements, conceptual design, and assessment of conformance to safety criteria. The design of the system includes hard-wired and component installation provisions for on-line surveillance testing that independently verifies by manual means the opera-
.bility of the UVTA and the automatic shunt trip.
The NRC issued a favorable Safety Evaluation Report on the generic design on August 10, 1983 (letter from D. Eisenhut to J. J. Sheppard). The SER lists plant-specific information required for individual plant modifications, which will be provided by December, 1984. SHNPP has committed to installing a class IE automatic shunt trip attachment.
VKLB04
Page 18 of 19 E*
- 4.5 REACTOR TRIP SYSTEH RELIABILITY SYSTEH FUNCTIOiVAL TESTING NRC POSITION On-line functional testing of the reactor trip system, includ-ing independent testing of the diverse trip features, shall be performed on all plants.
The diverse trip features to be tested include the breaker undervoltage and shunt trip features on Westinghouse, B&V (see Action 4.3 above) and CE plants; the circuitry used for power interruption with the silicon controlled recti-fiers on B&V plants (see Action 4.4 above); and the scram pilot valve and backup scram valves (including all initi-ating circuitry) on GE plants.
Plants not currently designed to permit periodic on-line testing shall justify not making modifications to permit such testing. Alternatives to on-line testing proposed by licensees will be considered where special circumstances exist and where the objective of high reliability can be met in another way.
- 3. Existing intervals for 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
- 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 licensees currently not performing on-line testing shall be justified by informa-tion on the sensitivity of reactor trip system availability to parameters such as the test intervals, component failure rates, and common mode failure rates.
RESPONSE
- 1. As noted above in Section 4.3, the generic design package for the automatic shunt trip furnished to WOG for submittal to the NRC included an installation for on-line surveil-lance testing of the UVTA and automatic shunt trip that provided independent verification of each attachment. The existing generic RTS automatic trip is by UVTA only and WKLB04
Page 19 of 19 manual trip by either UVTA or shunt trip. Although the existing generic system does not include installed pro-visions for independent verification of the two trips, a procedure for performing independent verification during shutdown is feasible and has been recommended by Westing-house in Westinghouse Technical Bulletin NSD-TB-83-03, dated Harch 24, 1983. The procedure is intended only to provide general guidance from which SHNPP can develop its own plant-specific procedure. This bulletin addresses the concerns discussed in NRC IE Circular 81-12 and I.E.
Bulletin 83-01. It is noted that the procedure presented in this Tech Bulletin is not performed at power, since actuation of a manual trip on the generic Westinghouse reactor trip system would trip the reactor.
SHNPP will have on-line testing capability.
- 3. The Vestinghouse Owners Group in January, 1983, submitted WCAP-10271 to the NRC for review. VCAP-10271, "Evaluation of Surveillance Frequencies and out of Service Times for the Reactor Protection Instrumentation System" documents an evaluation of the impact on RPS unavailability of current and extended surveillance intervals.
The VCAP considers common mode failure, operator error, reduced redundancy during testing and equipment also considers correlative effects on plant bypass'CAP-10271 operation and safety including the manpower expenditure associated with surveillance, the number of inadvertent trips which occur during testing and the distraction from plant monitoring on the part of the control room operator and shift supervisor associated with testing. Supplement 1 to VCAP-10271 which will be submitted to the NRC in September 1983 is an extension of the evaluation and provides a discussion of component wear out caused by testing. The NRC review of WCAP-10271 to date has resulted in a request for additional information the NRC felt necessary to complete the review. Information that will be submitted to the NRC in response to that request will include an overall evaluation of the impact on plant safety of RPS surveillance, a discussion of the uncertainty of failure rates and common mode failure and more detail concerning the impact of surveillance intervals on RPS unavailability. VCAP-10271, Supplement 1, and the infor-mation provided to the NRC in defense of VCAP-10271 provides in a comprehensive from the information requested by item 4.5.3. The conclusion of VCAP-10271 and Supple-ment, 1 is that although RPS unavailability is increased less frequent testing of RPS components is warranted and will result in an improvement in overall plant safety and equipment reliability.
VKLB04
Page 13 of 19 officers of organizations that construct, own, operate, or supply components of a facility or activity that is licensed or otherwise regulated by the Nuclear Regulatory Commission inform the Commis-sion if they obtain information reasonably indicating that such facility, activity, or basic component fails to comply with regulatory requirements relating to substantial safety hazards or that such facility, activity, or basic component contains a defect which could create a substantial safety hazard." This effectively provides an additional safeguard for early identification of safety-related component failure or design discrepancy.
Plant surveillance testing, equipment repair and replacement procedures, and the quality assurance programs additionally provide assurance of safety-related equipment reliability.
CPGL is supporting the INPO Nuclear Utility Task Action Committee (NUTAC) on Generic Letter 83-28, Section 2.2.2. NUTAC is currently formulating the recommendations for an industry-wide vendor information program for safety-related equipment.
CPGL believes this program will provide a practical industry-vide approach to assuring safety-related equipment reliability.
In summary, the SHNPP vendor interface program in conjunction with plant procedures, surveillance testing, equipment repair and replacement, and the quality assurance program provides adequate assurance of safety-related component reliability.
3.1 POST-HAINTENANCE TESTING (REACTOR TRIP SYSTEH COHPONENTS NRC POSITION The following actions are applicable to post-maintenance testing:
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 performing its safety functions before being returned to service.
- 2. Licensees 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 Specifica-tions, where required.
MKLB04
Page 15 of 19 that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications where required.
- 3. Licensees and applicants shall identify, if applicable, any post-maintenance test requirements in existing Techni-cal Specifications which are perceived to degrade rather than enhance safety. Appropriate changes to these test requirements, with supporting justification, shall be submitted for staff approval.
RESPONSE
While developing the maintenance procedures for safety-related equipment, SHNPP shall incorporate a review to ensure that post-maintenance operability testing is required to be conducted and that testing demonstrates the equipment is capable of performing its safety function prior to being returned to service. Vendor recommendations from technical manuals and technical bulletins, plus any technical specification require-ments, will be incorporated during the development of plant procedures. Any work performed on safety-related equipment will be done using approved plant procedures. Portions of technical manuals may be excerpted or referenced at the time a procedure is approved. Any standardized technical specifica-tions that are perceived to degrade rather than enhance safety will be identified during the submission of SHNPP specific technical specifications and justification provided for any deviations. Use of various surveillance test procedures will provide a documented basis of test results. This program will be implemented by June, 1985.
4.1 REACTOR TRIP SYSTEH RELIABILITY (VENDOR-RELATED HODIFICATIONS 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 Vestinghouse in NCD-Elec-18 for the DB-50 breakers and a Hazch 31, 1983, letter for the DS-416 breakers shall be implemented or a justification for not implementing shall be made available. Hodifications not previously made shall be incorporated or a written evalua-tion shall be provided.
RESPONSE
The Harch 31, 1983, letter refers to NS-EPR-2744 E. P. Rahe of Westinghouse to R. C. DeYoung of NRC. This letter applies to DS-416 UVTA's only, and addresses UVTA dimensional variations.
Subsequent to this letter, on April 21, 1983, a letter requiring MKLB04
Page 14 of 19 Licensees and applicants shall identify, if applicable, any post-maintenance test requirements in existing Techni-cal Specifications which can be demonstrated to degrade rather than enhance safety. Appropriate changes to these test, requirements, with supporting justification, shall be submitted for staff approval. (Note that action 4.5 discusses on-line system functional testing.)
RESPONSE
Verification that, SHNPP post-maintenance operability testing of safety-related components in the reactor trip systems is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety function before being returned to service will be performed during development of the procedures. Applicable vendor and engineer-ing recommendations will be incorporated during development of the procedures. This will be completed by June, 1985.
'4'estinghouse, under contract to the Westinghouse Owners Group, is conducting a compilation of all existing maintenance infor-mation regarding 4'estinghouse switchgear, including lessons learned in the post-Salem interval. This effort will be completed by the end of 1983. Any new or improved maintenance requirements resulting from cyclic test programs, such as from
'4'OG Program "Cyclic Life and Class IE Qualification of the DB and DS Circuit Breaker Shunt Trip Attachments for Reactor Trip Switchgear; Cycle Life Testing of DS-UVTA" will be reviewed and incorporated into appropriate SHNPP maintenance procedures.
The post-maintenance test procedures will be reviewed to demonstrate that they will not degrade the performance of the equipment and its safety functions before being returned to service. Testing requirements will be incorporated into the Technical Specifications. This will be completed by June, 1985.
3.2 POST-NAINTENANCE TESTING ALL OTHER SAFETY-RELATED COHPONENTS NRC POSITION The following actions are applicable to post-maintenance testing:
Licensees and applicants shall submit a report docu'menting the extending of test and maintenance procedures and Technical Specifications review to assure that post-maintenance operability testing of all safety-related equipment is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.
- 2. Licensees and applicants shall submit the results of their check of vendor and engineering recomm'endations to ensure MKLB04
APR008 APPENDIX A 0-OMM-04-Rev. 0
APR008 CAROLINA POWER 6 LIGHT COiHPAiNY SHEARON HARRIS NUCLEAR POMER PLANT UNIT 0 PLANT OPERATING HANUAL VOLUHE 3 PART 1 PROCEDURE TYPE: OPERATIONS HANAGEHENT HANUAL (OHH) ,
NUHBER: 0-OiHH-04 TITLE: POST TRIP/SAFEGUARDS REVIEV REVISION 0 APPROVED:
Signature Date TITLE:
( P(8Fi )
0-OHH-04-Rev. 0 Page 1 of 21
APR008 TABLE OF CONTENTS SECTION TITLE 1.0 Purpose 2.0 References 3.0 Responsibilities 3.1 Shift Foreman 3.2 Shift Technical Advisor 3.3 Operating Supervisor 3.4 Hanager - Operations 3.5 Plant General Hanager 3.6 Plant Nuclear Safety Committee 3.7 Plant Nuclear Safety Committee Chairman 4.0 Definitions/Abbreviations 4.1 Definitions 4.2 Abbreviations 5.0 Procedure - Post Trip/Safeguards Review Report 5.1 Data Collection 5.2 Initial Plant Conditions 5.3 Cause of Trip/Safeguards Actuation 5.4 Evaluation of Trip/Safeguards Actuation 5.5 Corrective Actions 5.6 Signatures 5.7 Post Trip/Safeguards Start-up Authorization 5.8 Scheduled Review 6.0 ATTACHHENTS 6.1 Post Trip/Safeguards Review Report 6.2 Table 1 - Reactor Trip Automatic Action Verification 6.3 Table 2 - Safeguards Automatic Action Verification 6.4 Table 3 - Strip Chart Data 6.5 Table 4 - Sequence of Events Summary 0-OHH-04-Rev. 0 Page 2 of 21
APR008 LIST OF EFFECTIVE PAGES PAGE REVISION 1 thru 0-0191-04-Rev. 0 Page 3 of 21
APR008 1.0 PURPOSE The purpose of this procedure is to establish the require-ments to perform and the methodology for performing a formal post event review for all reactor trips or safe-guards actuations. This procedure establishes the requirements and criteria that must be met prior to the start-up of the unit following a reactor trip or safeguards event.
2.0 REFERENCES
2.1 NUREG-1000 Generic Implications of ASS Event at Salem Nuclear Power Plant - Volume 1, Section 1.1 2.2 Procedures Administration Hanual 3.0 RESPONSIBILITIES 3.1 Shift Foreman 3.1.1 Ensure the Post Trip/Safeguards Review Report is complete.
3.1.2 Provide central direction for the investigation of the event.
3.1.3 Authorizes restart if the cause of the trip has been determined to be not due to equipment or instrumentation malfunction and has been corrected.
- 3. 1.4 Hay authorize the withdrawal of the shutdown banks provided his estimated critical position is avoided by at least 1000 PCH, as determined by the use of current approved rod worth and boron worth curves.
3.2 Shift Technical Advisor 3.2.1 Assist and advise the Shift Foreman in the completion of the Post Trip/Safeguards Review Report and must, concur that the cause of the trip is known and corrected prior to restart.
3.3 0 eratin Su ervisor (or Designated Alternate) 3.3.1 Reviews the Post, Trip/Safeguards Review Report.
3.3.2 Authorizes restart, if the Hanager - Operations is unavailable, for any trip for which the cause is known to be due to malfunction of equipment or instrumentation and has been corrected.
3.4 Hang er - 0 erations (or Designated Alternate) 3.4.1 Reviews Post Trip/Safeguards Review Reports 0-OMM-04-Rev. 0 Page 4 of 21
APR008 3.0 RESPONSIBILITIES (Cont'd) 3.4.2 Authorizes restart, for any trip for which the cause is known to be due to malfunction of equipment or instrumen-tation and has been corrected.
3.5 Plant General Hang er (or Designated Alternate) 3.5. 1 Reviews Post Trip/Safeguards Review Report.
3.5.2 Approves restart, based on the PNSC recommendations, for those trips for which the cause has not been identified or corrected.
3.6 Plant Nuclear Safet Committee (PNSC) 3.6.1 Reviews the Post Trip/Safeguards Review Report.
3.6.2 Reviews the Post Trip/Safeguards Review Report prior to restart, if applicable.
3.6.3 Recommends restart, for those trips for which the cause has not been identified or corrected.
3.7 PNSC Chairman 3.7.1 Approves the Post Trip/Safeguards Review Reports 3.7.2 Recommends restart, if appropriate.
4.0 DEFINITIONS ABBREVIATIONS Definitions 4.1.1 Reactor Trip - a reactor trip occurs anytime the reactor trip breakers open that causes one or more full length control rods to be inserted, except planned trips during performance of an approved procedure.
4.1.2 Safeguards Actuation - any manual or automatic safeguards signal that actuates or should have actuated the safe-guards sequencers, except planned actuations during performance of an approved procedure.
4.2 Abbreviations 4.2.1 PNSC - Plant Nuclear Safety Committee 5.0 PROCEDURE - POST TRIP SAFEGUARDS REVIEW'EPORT The Post Trip/Safeguards Review Report, (Attachment 6.1) will be completed by an individual or team qualified to assess the event. Individuals qualified to assess the event include licensed personnel, qualified Shift 0-OHH-04-Rev. 0 Page 5 of 21
APR008 5.0 PROCEDURE - POST TRIP SAFEGUARDS REVIEW REPORT (Cont'd)
Technical Advisors or other experienced personnel assigned by Plant Hanagement. The Post Trip/Safeguards Review Report does not take priority over any actions required to place the Plant in a safe condition.
5.1 Data Collection The purpose of the Data Collection section of the form is to obtain data as quickly as possible to evaluate the event.
Table 1, Attachment 6.2, is for verification of certain automatic reactor trip actions.
Table 2, Attachment 6.3, is to be completed only in the event a safeguards actuation occurs.
Table 3, Attachment 6.4, is for recording strip chart data. Data to be recorded includes the value of the variable immediately prior to the event, the maximum and minimum value of the variable during the event, and the value of the variable after the event has stabilized.
Each strip chart is to be labeled with the start and stop time of the event, date, and initials. Additional strip chart data should be added to the form if the variable is relevant to the event.
Table 4, Attachment 6.5, is to be completed only if the sequence of events printout from the computer is available. Table 4 is used to verify the proper operation of the reactor trip breakers using the sequence of events data.
Event summaries are to be obtained from individuals who were present when the event occurred and who were actively involved in the cause and mitigation of the event. Each individual involved in the event may prepare a written summary of his involvement in the event, sign it and forward it to the Shift Foreman. Also, the individuals involved may as a group or groups discuss the event. The individual leading the discussion will be responsible for preparing a summary of the discussion, signing it, and forwarding it to the Shift Foreman.
All event summaries should be obtained prior to any individual, who was involved in the event, leaving the Plant site.
Any other physical components that may be relevant to the trip event should be forwarded with the post trip review report if feasible (i.e., blown fuses, small failed components, photographs of larger components if possible).
0-OHM-04-Rev. 0 Page 6 of 21
APR008 5.2 Initial Plant Conditions The intent of this section is to obtain initial Plant conditions immediately prior to the event.
The sum of the shutdown bank trip and the control bank trip equal the total reactor trip number.
5.3 Cause of Tri Safe uards Actuation This section provides for a concise summary of the event summaries obtained. This section is completed without the benefit of the sequence of events printout in order to obtain a single concise summary of the event based on observations.
5.4 Evaluation of Tri Safe uards Actuation The purpose of this section is to analyze the information obtained from data collections to ensure the reactor protection/safeguards equipment and first out annunciator panel operated properly.
5.4.1 Sequence of Events - This section provides for the comparison of the first annunciator and sequence of events printout, if available, to identify any discrepancies.
5.4.2 Additional First Out Annunciators - This section provides for the evaluation of the proper operation of the First Out Annunciator Panel and for the identification of .other potential equipment problems.
5.4.3 Strip Chart. Data - This section provides for the evalua-tion of trend data from Table 3, Attachment 6.4, against expected responses based on operator experience and training.
5.4.4 Detail Summary of Event - This section provides for the detailed summary of the event based upon all of the available information.
Cause of Event - This section provides for establishing the cause of the event and written justification.
5.4.6 Halfunctions - This section provides for establishing the cause of any protection/safeguards equipment malfunctions or first out annunciator malfunctions.
5.4.7 Other - This section provides for listing other equipment malfunctions which, if it had functioned properly, would have helped mitigate the event.
0-OHH-04-Rev. 0 Page 7 of 21
APR008 5.5 Corrective Actions This section provides for the documentation of the correc-tive action taken as a result of the event. Any long term improvements identified should be submitted to Plant management utilizing the Plant Improvement Form.
5.6 ~Si natures This section provides a list of all individuals involved with the Post Trip/Safeguards Review Report and review by two current SRO licensed individuals.
5.7 Post Tri Safe uards Start-u Authorization The Shift Foreman may authorize restart of the unit if the cause of the event has been clearly identified and cor-rected and was not due to equipment or instrumentation malfunction(s). The approval of the Operating Supervisor or Hanager - Operations is required if the cause of the event has been clearly identified and corrected and was due to equipment or instrumentation malfunction(s). If the cause of the event has not been clearly identified or there are questions concerning the proper performance of protection/safeguards equipment or systems during the event, the Post Trip/Safeguards Review Report will be reviewed by the PNSC. The PNSC shall make recommendations on the restart of the unit. Upon completion of any additional corrective actions and the review thereof, the PNSC Chairman may recommend the restart of the Unit. The Plant General Hanager may authorize the restart of the unit based on the PNSC's recommendations.
This section also provides for the closeout of the Post Trip/Safeguards Review Report process if a restart is not planned.
5.8 Scheduled Review This section defines the normal review of the Post Trip/-
Safeguards Review Report.
6.0 ATTACHHENTS 6.1 Post Trip/Safeguards Review Report 6.2 Table 1 - Reactor Trip Automatic Action Verification 6.3 Table 2 - Safeguards Automatic Action Verification 6.4 Table 3 - Strip Chart Data 6.5 Table 4 - Sequence of Events Summary 0-OHH-04-Rev. 0 Page 8 of 21
ATTACHHENT 6. 1 PAGE 1 OF 7 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO. 1 POST TRIP SAFEGUARDS REVIEW REPORT 1.0 ., DATA COLLECTION Complete Attachment, 6.2, Table 1, Reactor Trip Automatic Actions Verification.
1.2 Complete Attachment 6.3, Table 2, Safeguards Automatic Actions Verification, if applicable.
1.3 Complete Attachment 6.4, Table 3, Strip Chart Data.
1.4 Complete Attachment 6.5, Table 4, Sequence of Events Summary, if sequence of events printout is available.
1.5 Attach event summaries from people involved in the event or a summary of discussion with people involved in the event.
2.0 INITIAL PLANT CONDITIONS 2.1 Date of Event Time of Event Shutdown Bank Trip No.:
Control Bank Trip No.:
Total Reactor Trip No.:
2.2 Personnel on Dut Shift Foreman:
Senior Control Operator:
Control Operators:
Shift Technical Advisor:
2.3 Plant Conditions 2.3.1 Reactor Power ;o TAVE F Net NWe 0-ONN-04-Rev. 0 Page 9 of 21
ATTACHHENT 6.1 PAGE 2 OF 7 2.3 Plant Conditions (Cont'd) 2.3.2 List evolutions in progress on production or LCO equipment immediately prior to the event (i.e., surveillance testing, trouble shooting, maintenance, unit start-up activities, unit shutdown activities, or other activities which could have contributed to the event).
2.3.3 List any equipment that was inoperable which could have contributed to the event (i.e., major production equipment, LCO equipment, or instrumentation and controls equipment).
3.0 CAUSE OF TRIP SAFEGUARDS ACTUATION 3.1 Reactor Tri Actuation Record First Out Annunciator Briefly describe the cause of actuation based on review of event, summaries:
3.2 Safe uards Actuation If safeguards actuated, briefly describe cause of actuation based on review of event summaries:
0-OHH-04-Rev. 0 Page 10 of 21
I ATTACHHENT 6.1 PAGE 3 OF 7 4.0 EVALUATION OF TRIP SAFEGUARDS ACTUATION 4.1 Se uence of Events Does the sequence of events printout (if available) agree with the First Out Annunciator? YES NO If no, explain discrepancies.
4.2 Additional First Out Annunciators 4.2.1 Briefly describe any additional First Out Annunciators identified:
4.2.2 List and explain any other First, Out Annunciators which should have annunciated but did not:
4.3 Stri Chart Data List and explain any unexpected strip chart responses from Attachment 6.4, Table 3.
4.4 Detail Summar of Events Explain in detail the events which led to the actuation of the Trip/Safeguards and the actions performed to place the Plant in a stable condition (reflect the event summaries and the sequence of events printout, if available).
0-OHH-04-Rev. 0 Page 11 of 21
ATTACH>1ENT 6.1 PAGE 4 OF 7 4.0 EVALUATION OF TRIP SAFEGUARDS ACTUATION (Cont'd) 4.5 Describe the cause of the event and justification:
4.6 Describe the cause of any malfunction(s) of the protection/safeguards systems or First Out Annunciators:
4.7 List equipment other than protection/safeguards equipment which failed to function properly during the event which, if it had functioned properly, would have helped mitigate the event.
5.0 CORRECTIVE ACTIONS 5.1 Describe actions taken to correct cause of the event:
0-0>H1-04-Rev. 0 Page 12 of 21
ATTACHHENT 6.1 PAGE 5 OF 7 5.0 CORRECTIVE ACTIONS (Cont'd) 5.2 Describe corrective actions for any protection/safeguards equipment which failed to function properly during the event:
5.3 Describe any corrective actions for any First Out Annunciators which should have annunciated but did not:
5.4 List items identified in Step 4.7 which are repaired prior to restart:
5.5 Describe any corrective action for equipment which will not be repaired prior to restart and justification:
6.0 POST TRIP SAFEGUARDS REVIEW< REPORT PREPARATION List the individuals assisting in the preparation of the report:
Reviewed by: Date Time SRO Licensed: / Date Time Signature Title Shift Foreman Date Time Signature Title 0-OHH-04-Rev. 0 Page 13 of 21
.o ATTACHHENT 6.1 PAGE 6 OF 7 7.0 POST TRIP SAFEGUARDS START-UP AUTHORIZATION 7.1 I have reviewed the Post Trip/Safeguards Review Report.
The cause of the event has been clearly identified, has been corrected, and was not due to equipment or instrumen-tation malfunction(s). All Protection/Safeguards equip-ment or systems functioned as designed.
Start-up Approved By:
Date Time Shift Foreman Concurrence By:
Date Time Shift Technical Advisor 7.2 I have reviewed the Post Trip/Safeguards Review Report.
The cause of the event has been clearly identified has been corrected, and was due to equipment or instrumenta-tion malfunction(s). All Protection/Safeguards equipment or systems functioned as designed.
Start-up Approved By:
Date Time Signature Operating Supervisor or Hanager - Operations Concurrence By:
Time Shift Technical Advisor 0-01'1M-04-Rev. 0 Page 14 of 21
ATTACHHENT 6.1 PAGE 7 OF 7 7.0 POST TRIP SAFEGUARDS START-UP AUTHORIZATION (Cont'd) 7.3 If the cause of the event has not been clearly identified, the cause of the event has not been corrected, or there are questions concerning the proper performance of pro-tection/safeguards equipment or systems during the event, PNSC review required prior to restart.
Post Trip/Safeguards Review Report Submitted to PNSC Date Time Shift Foreman Additional corrective actions identified by PNSC:
Additional corrective actions completed if necessary.
PNSC has reviewed and recommends start-up.
Start-up-up Recommended By:
Date Time PNSC Chairman Start-up Approved By:
Date T3.me Plant General Hanager, 8.0 ScjiEDULED REVIEW 8.1 Send "For Information Only" copy of the Post Trip/-
Safeguards Review Report to Regulatory Compliance.
8.2 Send "For Information Only" copy of the Post Trip/-
Safeguards Review Report to ONSITE Nuclear Safety.
8.3 The Post Trip/Safeguards Review Report is to be reviewed by the PNSC at its next monthly meeting or sooner if deemed necessary by the Plant General Hanager.
Reviewed By: Date Operating Supervisor Reviewed By: Date Hanager - Operations Approved By:* Date PNSC Chairman 0-OHH-04-Rev. 0 Page 15 of 21
ATTACHMENT 6.2 PAGE 1 OF 1 TABLE 1 REACTOR TRIP AUTOMATIC ACTION VERIFICATION Automatic Manual Mould Function Function Not AUTOMATIC ACTION Occurred Required Function Comments
- 1. Reactor Trip Breaker A Trip
- 2. Reactor Trip Breaker B Trip
- 3. Reactor Trip Bypass Breaker A Trip
- 4. Reactor Trip Bypass Breaker B Trip
- 5. Turbine Trip (All Turbine Valves Shut)
- 6. 'eedwate'r Regulator Valves close when Tavg decreases to 564 F ANY OF THE ABOVE ITEMS MAY BE MARKED N/A IF IT VAS IN THE POST TRIP POSITION PRIOR TO THE EVENT.
- 1. Record First Out Annunciator:
- 2. Record Additional First Out Annunciators Received:
Signature Date Time 0-OMM-04-Rev. 0 PAGE 16 OF 21
ATTACHMENT 6.3 PAGE $ 07 2 TABLE 2 SAFEGUARDS AUTOMATIC ACTION VERIFICATION Automatic Manual Would Function Function Not SAFEGUARDS EQUIPMENT Occurred Required Function Comments
- 1. Charging/SI Pump lA-SA
- 2. Charging/SI Pump 1B-SB
- 3. Charging/SI Pump 1C-SAB
- 6. Emergency Service Water Pump 1A-SA
- 7. Emergency Service Water Pump 1B-SB
.8.. Emergency Service Water..
Booster Pump '1A-"SA
- 9. Emergency Service Water Booster Pump 1B-SB
- 10. Containment Fan AH-1
- 11. Containment Fan AH-2
- 12. Containment Fan AH-3
- 13. Containment Fan AH-4
- 14. Auxiliary Feedwater Pump A
- 15. Auxiliary Feedwater Pump B
- 19. C.R. Vent. Isolation 0-OHM-04-Rev. 0 PAGE 17 OF 21
ATTACHMENT 6.3 PAGE 2 QP' .
TABLE 2 SAFEGUARDS AUTOMATIC ACTION VERIFICATION (Cont'd)
CONTAINMENT SPRAY EQUIPMENT
- 1. Containment Spray Pump 1A-SA
- 2. Containment Spray Pump 1B-SB
- 3. Steam Line Isolation (A,B,C MSIV's Closed)
LIST ANY NECESSARY PINK STATUS LIGHTS THAT DID NOT ILLUMINATE.
SIGNATURE Date 'Zime 0-OMM-04-Rev. 0 PAGE 18 OF 21
ATTACHMENT 6.4 PAGE 10P 2.,
TABLE 3 STRIP CHART DATA Value Max Value Min Value Value After Immediately During During Event s STRIP CHART VARIABLE Prior Event Event Stabilized S/G A Narrow Range Level LR-478 S/G B Narrow Range Level LR-488 S/G C Narrow Range Level LR-498 Pressurizer Level LR-459 Pressurizer Pressure PR-444
- 6. T - Avg. TR-408
- 7. Loop 1 Hot Leg Temp. TR-413
'8.Loop 2 Hot'Leg Temp TR-413 Loop 3 Hot Leg Temp.
TR-413 10 Loop 1 Cold Leg Temp.
TR-410 Loop 2 Cold Leg Temp.
TR-410
- 12. Loop 3 Cold Leg Temp.
TR-4 10
- 13. Nuclear Power Range N-45
- 14. S/G A Wide Range Level LR-4 77
- 15. S/G B Wide Range Level LR-477
- 16. S/G C Wide Range Level LR-477
- 17. RCS Wide Range Loop'A Pressure PR-402
ATTACHMENT 6.4 PAGE 2 OF 2
~
TABLE 3 STRIP CHART DATA (Cont'd)
Label each strip chart lested above with the start and stop time of the event, date, and initials. ~
SIGNATURE Date Time List any unexpected strip chart responses S1GNATURE Date Time
~ ~ ~
I 0-OMM-04-Rev. 0 PAGE 20 OF 21
I 0
ATTACKKNT 6.5 PAGE 1 OF 1 TABLE 4 SE UENCE OF EVENTS
SUMMARY
TIME SEQUENCE OF EVENTS STARTED: HOUR MINUTES SECONDS INITIATING EVENT (0 Cycles) 1st Reactor Trip Signal Initiated at Cycles 1st Safeguards Initiating Event at Cycles Computer Delta Acceptable Address Cycle Time In Delta Time SUBSEQUENT ALARMS Id Time Cycles In Cycles
- 1. Reactor Trip Breaker A Trip 5 (I ater)
- 2. Reactor Trip Breaker B Trip 5(Later)
- 3. Reactor Trip Bypass Breaker A Trip 5(Later)
Reactor Trip Bypass Breaker B Trip 5(Later) 5.."TB .HYD..OIL LO (Turbi'ne Trip)
P.
' ' ' ' '" : " : "' " . 'S(La'ter') .
Reactor Manual Trip Breaker 1 Trip N/A N/A
- 7. Reactor Manual Trip Breaker 2 Trip N/A N/A Any items 1-5 above may be marked N/A if it was in the Post Trip position prior to the event. Items 6 & 7 should be completed only if a manual trip button was pushed subsequent to the first trip initiating signal.
The delta time in cycles is the time between the subsequent alarm and the 1st trip initiating event.
DELTA TIME = (Subsequent Alarm Time in Cycles) - (1st Trip Initiating Event in Cycles)
If the delta time does not meet the acceptance criteria, it will be evaluated in Section 4.0 of the Post Trip/Safeguards Review Report.
NOTE: 50 Cycles = 1 Second Attach the Sequence of Events printout and the Post Trip Review printout to .1.
SIGNATURE Date Time 0-OMM-04-Rev. 0 PAGE 21 OF 21
,~
0'
APPENDIX B VKLB04
~ ~
APPENDIX B 0-PLP-608 EQUIPHENT DATA BASE 1.0 Purpose The purpose of the Equipment Data Base System is to provide a consistent, centralized means of identifying each plant struc-ture, system, and component, and storing any information relat-ing to those structures, systems, and components so that it. is easily and conveniently retrievable.
2.0 References The Equipment Data Base System in itself is not specifically required by Technical Specifications, the FSAR, or any regula-tion; however, the information contained therein is required by many such commitments, as is an efficient method for storage and retrieval of such information.
3.0 Responsibilities Responsibilities for the creation, maintenance, and utilization of the Equipment Data Base System will be detailed in Section 6.0 (Later), Implementation, of this program.
4.0 Definitions CHHS - Corporate Haterials Hanagement System HHS - Haintenance Hanagement System NPRDS - Nuclear Plant Reliability Data System 5.0 General The Equipment Data Base is an organized table of equipment-related information. Vithin each plant, the table is organized first by unit, then by systems within each unit, by functional equipment designations within each system, and by specific occurrence of each functional equipment type. For instance, Cooling Tower Hakeup Pump 1X-NNS for unit 1 at SHNPP has a plant code of 03 for SHNPP, a unit code of '1, a system code of 4030 for the Cooling Towers Hakeup System, an equipment code of CBU for Cooling Tower Hakeup Pump, and a specific identifier of 1X-NNS The Equipment Data Base serves as the central repository of all equipment-related information utilized by various plant and corporate systems and groups. It stores all component informa-tion required by Haintenance Hanagement, such as component identifiers, applicability of various special programs, level of review required, safety class, and spare part availability.
It also can store the history of each maintenance activity, both 'corrective and periodic.
WKLB04
a The Quality classification of a component is determined by analyzing the component against various criteria. The Equip-ment Data Base allows the storage and recall of this analysis, and the analysis of each repair part related to the component.
Carolina Power and Light Company is committed to support the Nuclear Plant Reliability Data System (NPRDS). This system maintains a file of engineering and failure history data for selected safety-related components within a plant. The Equip-ment Data Base serves as the entry point and storage location for that information at CP&L.
Safety-related electrical components installed in a harsh environment must demonstrate their qualification to function in that environment. The Equipment Data Base provides storage for both the requirements and the demonstrated capabilities of these components.
Through these and other types of information, the Equipment Data Base serves as the common tie between many separate functions. This is illustrated by attachment l.
Copies of some of the displays and brief descriptions of the data displayed are found in attachment 2.
VKLB04
~
I -~
A lf )
MATERIALS Q-LIST MANAGEMENT 4~
0>
v
~o 0
0 p h
~+
0 0
EQUIPMENT.
~ t DATA BASE
\ ~h 4 q ~v
~O h t 4 0
0 h 0 1
MAINTENANCE NPRDS MANAGEMENT ATTAGHMfNT 1
)-F'LF'-598, Da ta Desc r i p t i ons Component identification Accounting !Charge ) information Worl< Req,uest Header information F'ending actions Equiprneni out-of-service data LCO status 7 Physical, l,ocation within the plant 8 Regulatory references Q Drawings which reference this system or component 30 Technical. Manuals which reference this system or component 4 l Specifications which reference this system or component F'rocedures which reference th i s system or component
$ 3 Special tool or equ i pment notes
$ 4 Design Rasis classifications jR In-Ser v i ce-inspec t i on app i cab i i ty L L 1b Environmental. C>>uaL i f i cat ion app icab i l i ty l.
17 NPRDS reporiabiLity
$ 8 importance to fire protection lo Criticality io power production capability Energy industry identification System classifications NPRDS identification codes
%4 Xn-service .date'Ae jL'.r cable>>'.code.'ia: sia'ada,r',d',".":. ':-.,;:-'-':;".":.:.":.'.':;:;.::",'-:.".:;;.'-,:;.,':";:!;:
Natu facturer and vendor i deist i f i cat i on rianu facturer 's model and ser i al, number Vicsnufacturer or vendor's component i dent i f iers Component physical. and or era t ing pa rarneter s HPRDS operat ing data code=
30 Qual ity classi f icat'ion just Qual i ty class if ication Review and approval documentation
>>>> Periodic Act i on Procedure i dent i f i ca t i on Per Iodlc Act ion performance respons lb i L l ty 34 Plant mode required for per iodic act ion performance 7+ Estimated t'rne requ i red Per iodic i ty, incLuding tolerance
'7 Or i gin of cornmi trnent io per form this particular per iodic ac< ion
+8 Last date per i odi c act i on was per formed for th i s component Range of dates -'or. 'next required .performance 4>>; ~
Mork request 4$ Time of fa i lure PLant status at time of fai Lure 4> F a r Lure desc r i pi i on narra t i ve Fa i Lure caus. narrat ive Corrective action narrative 4n Failure codes for trend anaLysis 47 I irensoe Event Resort ~ and date 48 trav r nla l env r r onlnent a t colnponent s oca i on L r
>>)0 En' ronrnenta L aa rarneters io wni cn component must be qua l. i f i ed KC>> ident i f i cat ion of actuaL component instal. l.ed Demon. irated caoab i i t>>>>'f ac4ua l. cornc onent insta Led L L
' EQUIPMENT DATA BASE ++w GENERAL INFORMATION PLANT UNIT SYSTEM EQUIPHKNT SPECIFIC ID TAG 03 3 4065 EZK SW-Bh SA 5 SW-I VALVE, ISOLATION DESCRIPTION ESW HDR A AUX RES INTAKE
.CHARGE TO H31 CAPITAL 321 HAINT 529 OPERATIONS 524 TAX CODE WA 2 Q-CLS VR ISI FIRE LCO RQD SKCURITY ENVIR SKIS NPRD CRIT RWP RQD OWP
~
N 0 ACTIONS PENDING: 0
...-.;::;:TYPE::=":,~'.":.. ',.':PESCR'IPTION."i.;. '..-.:. ~::,',::,'.:":".: .:.'..., .. i'.:
~ - ':.:.:,;.'.-EST HR:
MORE N
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%+%%%%WK+%%%%%%%%%%%%%%%%%%%%K%%%%%%%%%%%%%%%%%%%%
OUT OF SERVICE DATA: Qs CLEAR Y/N TINE OUT TIME DUE RACK DOCUHENTATION. LCO 4 LCO EXPIRATION Qs FUNCTION PLANT UNIT SYSTEH . EQUIP SPEC-ID CPL ~/INDEX TYPE
I I j
~ ~ ~
EQUIPMENT DATA BASE w+w ENVIRONMENTAL QUALIFICATION DATA PLANT UNIT SYSTEM EQUIPNEHT SPECIFIC ID 'TAG 03 1 4665 EZK SM-B1SA VALVEi ISOLATION DESCRIPTIOH ESM HDR A AUX RES INTAKE NORMAL EHV: TEMP DEG F PRESS PS)A REL HUH Q<8
).'AD ))R/HR QUALIFICATION REQUIREMENTS:
'I TENP DEG F PRESS PSIA REL HUN X AGING YR SUBMERGE
~
t; 'CHEN",SP)RAY:"CONTENT. ~<.::.". '.,,: .";;..::.','.'.
~
'~.'.:.<.".',",;,, RATE. -:..:.:-'BN.'.. DURATN).:: ").'IP.
CUM. RADIATIOH (RADS) 2HR 30DAYS . 'YR 40YRS SOURCE'-
QUALIFICATION OF INSTALLED EQUIPMENT:
MANU NOD S/H Qso TEMP DEG F PRESS PS A I REL HUM / AGING YR SUBMERGE CHEN SPRAY CONTENT RATE GPN DURATH NIH 51 CUN. RADIATIOH < RADS) 2HR 30DAYS 1YR 40YRS DATA SOURCE:
FUNCTION 208 PLANT 03 UNIT 1 SYSTEM 4065 EQUIP EZK SPEC-ID SM-B1SA CPL ~/INDEX TYPE
( I l
/
EQUIPMENT DATA RASE wow REFERENCE DATA PLANT UNIT SYSTEM EQUIPMEHT SPECIFIC ID TAG~
03 1 4065 EZK S4l-B 3 SA 1 S4l-1 VALVEi ISOLATION DESCRIPTION ES4l HDR A AUX RES INTAKE LOCATION 7 REG REF s MORE H HO DRA4lINGS e MORE V HO TECH MANUALS 0<0 NQRK N HO SPECIFICATIONS MORE H HO
'",: %%%%%9f%%%%%%Q%%%%gg%gg34k'%%%%$ &%3fkH&%%4'4W%%'%%g%%%%%%%%4f%%%+%3HE%%%+%%%%%%%%~'P%9fk%
PROCEDURES TYPE 4 REV DESCRIPTIOH I ~ ~
MORE N'HO SPECIAL TOOLS OR EQUIPMEHT MORE N HQ FUNCTION 202 PLANT 03 UNIT 1 SYSTEM 4065 EQUIP EZK SPEC-ID S4l-Bf SA CPL 4/I HDEX TYPE MSG
I -)
EQUIPHEHT DATABASE++wENGINEERING DATA PLANT UNIT SYSTEM EQUIPHEHT SPECIFiC ID TAG 03 3 40b5 EZK SM-BISA 5 SM-5 VALVE, ISOLATION DESCRXPTION ESM HDR A AUX RES INTAKE Q CLS- N Q GRP- SAFE CLS- Q<4 SEIS- ISI-QT FIRE-Q12 CPIT-Qs EI XS: SYSTKH COMPONENT FUNCTION 20 COHPONENT APPLICATION NPRD: SYS NODE ENVIRONMENT: INT 21 'XT IN SERVICE DATE / Q22 / APPLICABLE CODE OR STD Q22 HAHUF: CPL- NPRD- NDLO ~ SER~ 25 24 UENDOR,:;i:CPL'-;:::,:.", ';..:NPRD-2:.. VQ3DQR.. COMP/SYS:: ID',;:2::'" .. 'Qss, DATA CD DESC CD DESC CD DESC DSC DSC DSC OPS DATA OUT-OF-SERVICE- TEST TYPE FREQ INTERVAL HOURS CRIT HOURS FOR-TESTING DATA: CHECK Qs S-BY HOURS .FUNCT S-DH HOURS CALX'B FUNCTXON 203 PLANT 03 UNIT 1 SYSTEN 4065 EQUXP EZK SPEC-ID SM-91SA CPL 0/INDEX TYPE
EQUIPMENT DATABASE w+w QUALITY CLASSIFICATION ANALYSIS PLANT UNIT SYSTEM EQUIPMENT SPECIFIC ID TAG 03 $ 4065 EZK SM-91SA ASM-1 VALVE, ISOLATION DESCRIPTION ESW HDR A AUX RES INTAKE AFFECTED CRITERIA:
CLASS-4 DESCRIPTIOH OF INFLUENCE A/P
~ ' '9 I MORE Y/N N HO ACCIDENT IMPACT MORE Y/H H HO SSSUI TANT QUALITY CLASS Qeo ANALYST: TITLE: DATE:
REVIEWED; 0 TITLE:, DATE:
APPROVED: TITLE: DATE:
FUNCTION 204 PLANT 03 UNIT 1 SYSTEM 40h5 EQUIP EZK SPEC-ID SM-81SA CPL ~/INDEX TYPE
EQUIPMEHT DATA BASE %ww PERIODIC ACTION REQUIREMENTS PLANT UNIT SYSTEM EQUIPMEHT SPECIFIC ID TAG 03 ,1 4065 EZK SL4-B1SA 1SM-1 VALVE, ISOLATION DESCRIPTIOH ESM HDR k AUX RES INTAKE TYPE 4 DESCRIPT ION 0 RESP MOD EST HR FREQ
~33 34 +ag TOL /UNITS COMMIT T R H LAST ACTION DATE EARLY HFXT LATE NEXT 3 +sg 39
+%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%K%W
. 0.
sz 33 34 ss
~ ~
~ ~
~ 35 ' ~ ..
0
%%%%%%4%%%%%%%%%%%%%%%%K%%%%%%3f%%%%%%%%%%%%%%%%%%%%%%%%%%%++%%+++W+++%+<+++++++
~ ~" . '
as 34 SS ag
+
3 8
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%K%% %%%%%%+5 % % %+%%'%%%%%+'W 9
32 33 a4 36 +
3 0
%% %%% %%%%%% %%%%%%%%%% %%%%%%%%%%%%%% %%%%%%%%%%%% %%%%%%% %%%% %% %% %%% % K
~
%%%'4 %% '.f %%%
9
++
MORE Y/H H FUNCTION 205 PLANT 03 UNIT 1 SYSTEM 4065 EQUIP EZK SPEC-ID .SW-B1SA CPL ~/INDEX TYPE MSG ENTER ADDITIONAL DATA
K: ~
j ,lt ~
EQUIPMENT DATABASE>>>>>> FAILURE HISTORY PLANT UNIT SYSTEM EQUIPMEHT SPECIFIC ID TAG 83 f 4065 EZK SW-BfSA VALVE, ISOLATION DESCRIPTION ESM HDR A AUX RES INTAKE I
KVEHT DATA: CONTROL ~ EVENT START' EVENT EHD STATUS AT TIME OF FAILURE 041 0~
FAILURE DESCRIPTION:
Qm t
I CAUSE OF FAILURE:
0 MORE Y/N CORRECTIVE ACTION:
MORE Y/N PRIOR EVENT Y/H>>>>>>>>>>>>>>>>>> ADD EVENT Y/H FAILURE ANALYSIS: TYPE MODE . CAUSE LER P 4y EFFECT DETECTION ~48 ACTION TAKEH LER SUBMITTED FUNCTION 207 PLANT 03 UNIT f SYSTEM 4065 EQUIP EZK" SPEC-ID SL4-Rf SA CPL 4/INDEX TYPE