ML13305A831
| ML13305A831 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/13/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML13305A830 | List: |
| References | |
| GL-83-28, NUDOCS 8601030154 | |
| Download: ML13305A831 (19) | |
Text
- nclosure 3 SAFETY EVALUATION REPORT FOR GENERIC LETTER 83-28, ITEM 1.2 - POST-TRIP REVIEW (DATA AND INFORMATION CAPABILITY)
SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 DOCKET NOS.:
50-361, 50-362 I.
INTRODUCTION On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant (SNPP) failed to open upon an automatic reactor trip signal from the reactor protection system. This incident occurred during the plant start-up and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal.
The failure of the circuit breakers has been determined to be related to the sticking of the under voltage trip attachment. On February 22, 1983, during start-up of SNPP, Unit 1, an automatic trip signal occurred as the result of steam generator low-low level. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.
Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO) directed the staff to investigate and report on the generic implications of these occurrences. The results of the staff's inquiry into these incidents are reported in NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Commission requested (by Generic Letter 83-28 dated July 8, 1983) all licensees of operating reactors, applicants for an 8601030154 851213 PDR ADOCK 05000361 P
..PDR
-2 operating license, and holders of construction permits to respold to certain generic concerns. These concerns are categorized into four areas: (1)
Post-Trip Review, (2) Equipment Classification and Vendor Interface, (3)
Post-Maintenance Testing, and (4) Reactor Trip System Reliability Improvements.
The first action item, Post-Trip Review, consists of Action Item 1.1, "Program Description and Procedure" and Action Item 1.2, "Data and Information Capability." This safety evaluation report (SER) addresses Action Item 1.2 only.
II.
REVIEW GUIDELINES The following review guidelines were developed after initial evaluation of the various utility responses to Item 1.2 of Generic Letter 83-28 and incorporate the best features of these submittals. As such, these review guidelines in effect represent a "good practices" approach to post-trip review. We have reviewed the licensee's response to Item 1.2 against these guidelines:
A. The equipment that provides the digital sequence of events (SOE) record and the analog time history records of an unscheduled shutdown should provide a reliable source of the necessary information to be used in the post-trip review. Each plant variable which is necessary to determine the cause and progression of the events following a plant trip should be monitored by at least one recorder (such as a sequence-of-events recorder or a plant process computer) for digital parameters; and strip
3 charts, a plant process computer or analog recorder for analog (time history) variables. Performance characteristics guidelines for SOE and time history recorders are as follows:
o Each sequence of events recorder should be capable of detecting and recording the sequence of events with a sufficient time discrimination capability to ensure that the time responses associated with each monitored safety-related system can be ascertained, and that a determination can be made as to whether the time response is within acceptable limits based-on FSAR Chapter 15 Accident Analyses. The recommended guidelines for the SOE time discrimination is approximately 100 milliseconds. If current SOE recorders do not have this time discrimination capability the licensee should show that the current time discrimination capability is sufficient for an adequate reconstruction of the course of the reactor trip and post-trip events. As a minimum this should include the ability to adequately reconstruct the transient and accident scenarios presented in Chapter 15 of the plant FSAR.
O Each analog time history data recorder should have a sample interval small enough so that the incident can be accurately reconstructed following a reactor trip. As a minimum, the licensee should be able to reconstruct the course of the transient and accident sequences evaluated in the accident analysis of
-4 Chapter 15 of the plant FSAR. The recommended guideline for the sample interval is 10 seconds. If the time history equipment does not meet this guideline, the licensee should show that the time history capability is sufficient to accurately reconstruct the transient and accident sequences presented in Chapter 15 of the FSAR. To support the post-trip analysis of the cause of the trip and the proper functioning of involved safety related equipment, each analog time history data recorder should be capable of updating and retaining information from approximately five minutes prior to the trip until at least ten minutes after the trip.
o All equipment used to record sequence of events and time history information should be powered from a reliable and non-interruptible power source. The power source used need not be Class IE.
B. The sequence of events and time history recording equipment should monitor sufficient digital and analog parameters, respectively, to assure that the course of the reactor trip and post-trip events can be reconstructed. The parameters monitored should provide sufficient information to determine the root cause of the unscheduled shutdown, the progression of the reactor trip, and the response of the plant parameters and protection and safety systems to the unschediled shutdowns. Specifically, all input parameters associated with reactor trips, safety injections and other safety-related systems as well as output parameters sufficient to record the proper functioning of these
-5 systems should be recorded for use in the post-trip reviews The parameters deemed necessary, as a minimum, to perform a port-trip review that would determine if the plant remained within its safety limit design envelope are presented in Table 1. They were selected on the basis of staff engineering judgment following a complete evaluation of utility submittals. If the licensee's SOE recorders and time history recorders do not monitor all of the parameters suggested in these tables the licensee should show that the existing set of monitored parameters are sufficient to establish that the plant remained within the design envelope for the accident conditions analyzed in Chapter 15 of the plant FSAR.
C. The information gathered by the sequence of events and time history recorders should be stored in a manner that will allow for data retrieval and analysis. The data may be retained in either hardcopy, (e.g., computer printout, strip chart record), or in an accessible memory (e.g., magnetic disc or tape). This information should be presented in a readable and meaningful format, taking into consideration good human factors practices such as those outlined in NUREG-0700.
D. Retention of data from all unscheduled shutdowns provides a valuable reference source for the determination of the acceptability of the plant vital parameter and equipment response to subsequent unscheduled shutdowns.
Information gathered during the post-trip review is to be
-6 retained for the life of the plant for post-trip review com arisons of subsequent events.
III.
EVALUATION AND CONCLUSION By letter dated November 29, 1983, Southern California Edison Company provided information regarding its post-trip review program data and information capabilities for San Onofre Nuclear Generating Stations (SONGS),
Units 2 and 3. We have evaluated the licensee's submittal against the review guidelines described in Section II. Deviations from the Guidelines of Section II were discussed with representatives of the licensee by telephone on August 16, 1985. A brief description of the licensee's responses and the staff's evaluation of the response against each of the review guidelines follows:
A. The licensee has described the performance characteristics of the equipment used to record the sequence of events and time history data needed for post-trip review. Based on our review of the licensee's submittal, we find that the sequence of events recorder characteristics conform to the guidelines described in Section II A, and are acceptable.
During our telephone conversation the licensee stated that the time history data are recorded continuously on either their Critical Function Monitoring System (CFMS) or strip chart recorders. The CFMS and strip chart recorders are powered by non-interruptible sources. The licensee stated that the duration of time history is from 5 minutes before the
-7 trip to 10 minutes after the trip. Based on this information, we find that the time history recorder characteristics conform to the guidelines described in Section II A, and are acceptable.
B. The licensee has established and identified the parameters to be monitored and recorded for post-trip review. Based on our review, we find that the parameters selected by the licensee include all of those identified in Table 1 and conform to the guidelines described in Section II B and are, therefore, acceptable. It should be noted that SONGS, Units 2 and 3 do not have PORVs. Therefore, PORV Position is not a relevant parameter for these units.
The licensee does not record all of the sequence of events and time history parameters in the specific manner recommended in Table 1.
However, based upon information provided by the licensee during our telephone discussion, we find that the licensee has alternative data sources for those parameters not recorded on the sequence of events recorders and time history recorders. These include: (1) a computer Plant Monitoring System (PMS), (2) the Critical Function Monitoring System (CFMS) with hardcopy capability, powered by a non-interruptible source, and (3) strip chart recorders with non-interruptible power supplies. On the basis of this information we find that thq licensee's selection of parameters meets the intent of the guidelines described in Section II B and is, therefore, acceptable.
-8 C. During our telephone discussion, the licensee described thebmeans for storage and retrieval of the information gathered by the sequence of events and time history recorders, and for the presentation of this information for post-trip review and analysis. The licensee stated that time, parameter name and any related set point information are provided.
Based on this discussion, we find that this information will be presented in a readable and meaningful format, and that the storage, retrieval.and presentation conform to the guidelines of Section II C.
D. The licensee's submittal and information provided by the licensee during our telephone conversation indicate that the data and information used during post-trip reviews is retained in an accessible manner for the life of the plant. Based on this information, we find that the licensee's program for data retention conforms to the guidelines of Section II D, and is acceptable.
Based on our review of the licensee's submittal and our telephone conversation with the licensee, we conclude that the licensee's post-trip review data and information capabilities for San Onofre Nuclear Generating Stations Units 2 and 3 are acceptable.
-9 TABLE 1 PWR PARAMETER LIST SOE Time History a
Recorder Recorder Parameter/Signal (1) x Reactor Trip (1) x Safety Injection x
Containment Isolation (1) x Turbine Trip x
Control Rod Position (1) x x
Neutron Flux, Power x
x Containment Pressure (2)
Containment Radiation x
Containment Sump Level (1) x x
Primary System Pressure (1) x x
Primary System Temperature (1) x Pressurizer Level (1) x Reactor Coolant Pump Status (1) x x
Primary System Flow (3)
Safety Inj.; Flow, Pump/Valve Status x
MSIV Position x
x Steam Generator Pressure (1) x x
Steam Generator Level (1) x x
Feedwater Flow (1) x x
Steam Flow
01 10 SOE Time History Recorder Recorder Parameter/Signal (3)
Auxiliary Feedwater System: Flow, Pump/Valve Status x
AC and DC System Status (Bus Voltage) x Diesel Generator Status (Start/Stop, On/Off) x PORV Position (1) Trip parameters (2) Parameter may be monitored by either an SOE or time history recorder.
(3) Acceptable recorder options are; (a) system flow recorded on an SOE recorder, (b) system flow recorded on a time history recorder, or (c) equipment status recorded on an SOE recorder.
ENCLOSURE 2 SALP EVALUATION SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 36 DOCKET NOS.:
50-361, 50-362 GENERIC LETTER 83-28, ITEM 1.2, POST-TRIP REVIEW A. Functional Areas: Licensing Activities - Generic Letter 83-28.
Item 1.2, Post-Trip Review Data and Information Capability
- 1. Management involvement in assuring quality Based on our review of the licensee's response to Generic Letter 83-28, we find that the licensee has an effective capability for the collection, storage and retrieval of data needed to assess unscheduled reactor trips.
Rating: Category 2.
- 2. Approach to resolution of technical issues from a safety standpoint.
Rating: N/A
- 3. Responsive to NRC initiatives Based on our review, we find that the licensee is responsive to NRC initiatives.
Rating: Category 2.
- 4. Staffing Rating: N/A
- 5. Reporting and analysis of reportable events Rating: N/A
- 6. Training and qualification effectiveness Rating: N/A
- 7. Overall Rating for Licensing Activity Functional Areas:
Category 2 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION GENERIC IMPLICATIONS OF SALEM ATWS EVENT GENERIC LETTER 83-28, ITEMS 3.1.1 AND 3.1.2 SAN ONOFRE NUCLEAR GENERATING STATION -
UNITS 2 AND 3 DOCKET NOS.
50-361 AND 362 I.
INTRODUCTION On February 25, 1983, during startup of the Salem Unit 1 plant, both circuit breakers in the Reactor Trip System failed to open automatically upon receipt of a valid trip signal. As a result of that event, 2the NRC's Office -of Inspection and Enforcement issued IE Bulletin 83-01 which described the event and requested specified prompt corrective and preventive actions by licensees. As the cause and ramifications of the event were more clearly developed, the NRC's Office of Nuclear Reactor Regulation issued on July 8, 1983, Generic Letter 83-28, "Required Actions Based on Generic Implications of Salem ATWS Events."
This letter addressed issues related to reactor trip system reliability and general management capability. The letter was sent to all licensees of operating reactors, applicants for operating licenses and holders of construction permits.
One of the areas of reactor trip system reliability considered in Generic Letter 83-28 (GL 83-28), is that of post-maintenance testing of reactor trip system components. This is identified in GL 83-28 as Items 3.1.1 and 3.1.2. This evaluation addresses the acceptability of the response to these items provided by the Southern California Edison Company on behalf of itself and the San Diego Gas and Electric Company, the City of Riverside and the City of Anaheim, California (the licensees), for San Onofre Nuclear Generating Station, Units 2 and 3 (the facilities).
II. EVALUATION Items 3.1.1 and 3.1.2 of GL 83-28 state as follows:
"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 Technical Specifications, where required."
g*
2 By letter dated November 29, 1983, the licensees responded to a number of GL 83-28 items, including Items 3.1.1 and 3.1.2.
In response to questions raised by the staff, additional information was provided by letter dated October 2, 1985.
Regarding Item 3.1.1, the licensees' letter of November 29, 1983, indicated very specific post-maintenance testing was required for reactor trip 'breakers. For other components in the reactor trip system, however, the response indicated the maintenance work order was subject to an interdisciplinary review which specified the minimum operability testing required. The response also indicated a more definitive program for specifying post-maintenance testing was being developed.
In response to a request by the staff, additional information on this subject was provided by the licensees' letter of October 2, 1985.
This response indicated the 'more definitive program' mentioned in the November 29, 1983 response had been formally implemented as the Post-Maintenance Retest Program (PMRP). According to the licensees, all safety-related equipment in the reactor trip system is tested following.
maintenance activity per the PMRP. The licensees also state that, as a minimum, this post-maintenance testing demonstrates the equipment operates per design documents and is capable of performing its safety function.
Regarding Item 3.1.2, the licensees' response of November 29, 1983 states, vendor and engineering recommendations have been reviewed and incorporated in reactor trip system test and maintenance procedures. It is noted, however, this was not done in response to the request contained in GL 83-28, but rather as a result of reactor trip breaker operability problems identified at the facility about four months prior to issuance of GL 83-28. Upon discovery of these problems, the licensees undertook a comprehensive investigation in conjunction with the NSSS vendor (Combustion Engineering) and the reactor trip breaker manufacturer (General Electric). Part of this investigation was a review of vendor and engineering recommendations to assure their incorporation into reactor trip system test and maintenance procedures. This review was documented in a report titled, "Reactor Trip Breakers" transmitted to the NRC by licensee letter dated April 15, 1983. Based on review of this submittal, as documented in a Safety Evaluation Report transmitted by NRC letter dated May 2, 1983, the staff concluded there was reasonable assurance control of vendor information and technical manuals could and would be accomplished in an acceptable manner.
0 3
III. CONCLUSION Based on the licensees' confirmation that facility procedures require post-maintenance testing of safety-related components in the reactor trip system to demonstrate the equipment is capable of performing its safety function, we conclude the licensees have satisfactorily completed the actions requested by Item 3.1.1 of Generic Letter 83-28. Accordingly, this -item is closed.
Based on the corrective actions taken by the licensees with respect to control and use-of vendor information and the staff's conclusions regarding these actions, as noted above, we conclude the licensees have satisfactorily completed the equivalent of the actions requested by Item 3.1.2 of Generic Letter 83-28. Accordingly, this item is closed.
Principal Contributor:
G. Zwetzig
plosure 5 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION GENERIC IMPLICATIONS OF SALEM ATWS EVENT GENERIC LETTER 83-28, ITEMS 3.2.1 AND 3.2.2 SAN ONOFRE NUCLEAR GENERATING STATION - UNITS 2 AND 3 DOCKET NOS. 50-361 AND 362 I.
INTRODUCTION On February 25, 1983, during startup of the Salem Unit 1 plant, both circuit breakers in the Reactor Trip System failed to open automatically upon receipt of a valid trip signal. As a result of that event, the NRC's Office of Inspection and Enforcement issued IE Bulletin 83-01 which described the event and requested specified prompt corrective and preventive actions by licensees. As the cause and ramifications of the event were more clearly developed, the NRC's Office of Nuclear Reactor Regulation issued on July 8, 1983, Generic Letter 83-28, "Required Actions Based on Generic Implications of Salem ATWS Events."
This letter addressed issues related to reactor trip system reliability and general management capability. The letter was sent to all licensees of operating reactors, applicants for operating licenses and holders of construction permits.
One of the areas of reactor trip system reliability considered in Generic Letter 83-28 (GL 83-28), is that of post-maintenance testing of safety-related components other than those in the reactor trip system.
This is identified in GL 83-28 as Items 3.2.1 and 3.2.2.
This evaluation addresses the acceptability of the response to these items provided by the Southern California Edison Company on behalf of itself and the San Diego Gas and Electric Company, the City of Riverside and the City of Anaheim, California (the licensees), for San Onofre Nuclear Generating Station, Units 2 and 3 (the facilities).
II. EVALUATION Items 3.2.1 and 3.2.2 of GL 83-28 state as follows:
"1. Licensees and applicants shall submit a report documenting the extending of test and maintenance procedures and Technical Specifications 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 recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or Technical Specifications, where required."
2 By letter dated November 29, 1983, the licensees responded to a number of GL 83-28 items, including Items 3.2.1 and 3.2.2.
In response to questions raised by the staff, additional information was provided by letter dated October 2, 1985.
Regarding Item 3.2.1, the licensees' letter of November 29, 1983,.
indicated that "All maintenance orders involving safety-related, Technical Speci-fication related to ASME code (Section III and XI) related equipment..." were subject to an interdisciplinary review which specified the minimum operability testing required. The response also indicated a more definitive program for specifying post-maintenance testing was being developed.
In response to a request by the staff, additional information on this subject was provided by the licensees' letter of October 2, 1985. This response indicated the 'more definitive program' mentioned in the November 29, 1983 response had been formally implemented as the Post-Maintenance Retest Program (PMRP). According to the licensees, all safety-related equipment is tested following maintenance activity per the PHRP. The licensees also state that, as a minimum, this post-maintenance testing demonstrates the equipment operates per design documents and is capable of performing its safety function.
Regarding Item 3.2.2, the licensees' response of November 29, 1983 states, vendor and engineering recommendations have been reviewed and incorporated in reactor trip system test and maintenance procedures. It is noted, however, this was not done in response to the request contained in GL 83-28, but rather as a result of reactor trip breaker operability problems identified at the facility about four months prior to issuance of GL 83-28. Upon discovery of these problems, the licensees undertook a comprehensive investigation in conjunction with the NSSS vendor (Combustion Engineering) and the reactor trip breaker manufacturer (General Electric).
Part of this investigation was a review of vendor and engineering recommendations to assure their incorporation, as appropriate, into the test and maintenance procedures for all safety-related components. This review was documented in a report titled, "Reactor Trip Breakers" transmitted to the NRC by licensee letter dated April 15, 1983.
Based on review of this submittal, as documented in a Safety Evaluation Report transmitted by NRC letter dated May 2, 1983, the staff concluded there was reasonable assurance control of vendor information and technical manuals could and would be accomplished in an acceptable manner.
Regarding use of vendor information, the licensee also states "...all maintenance procedures incorporating safety-related components undergo a biennial review. During the biennial review, vendor and engineering recommendations are checked, assuring their incorporation into maintenance procedures."
III. CONCLUSION Based on the licensees' confirmation that facility procedures require post-maintenance testing of safety-related components to demonstrate the equipment is capable of performing its safety function, we conclude the
- licensees have satisfactorily completed the actions requested by Item 3.2.1 of Generic Letter 83-28. Accordingly, this item is closed.
Based on the corrective actions taken by the licensees with respect to control and use of vendor information and the staff's conclusions regarding these actions, and based on the licensees' biennial review of maintenance procedures for safety-related components, we conclude the licensees have satisfactorily completed the equivalent of the actions requested by Item 3.2.2 of Generic Letter 83-28. Accordingly, this item is closed.
Principal Contributor:
G. Zwetzig
- Enclosure 6 SAFETY EVALUATION BY NRC REGION V GENERIC IMPLICATIONS OF SALEM ATWS EVENT GENERIC LETTER 83-28, ITEM 4.1 SAN ONOFRE NUCLEAR GENERATING STATION - UNITS 2 AND 3 DOCKET NOS. 50-361 AND 362 I.
INTRODUCTION On February 25, 1983, during startup of the Salem Unit 1 plant, both circuit breakers in the Reactor Trip System failed to open automatically upon receipt of a valid trip signal. As a result of that event the NRC's Office of Inspection and Enforcement issued IE Bulletin 83-01 which described the event and requested specified prompt corrective and preventive actions by licensees. As the cause and ramifications of the event were more clearly developed, the NRC's Office of Nuclear Reactor Regulation issued on July 8, 1983, Generic Letter 83-28, "Required Actions Based on Generic Implications of Salem ATWS Events. This letter addressed issues related to reactor trip system reliability and general management capability. The letter was sent to all licensees of operating reactors, applicants for operating licenses and holders of construction permits.
One of the areas of reactor trip system reliability considered in Generic Letter 83-28 (GL 83-28), is that of vendor-recommended modifications.
This is identified in GL 83-28 as Item 4.1. This evaluation addresses the acceptability of the response to this item provided by the Southern California Edison Company (the licensee) for San Onofre Nuclear Generating Station, Units 2 and 3.
II. EVALUATION Item 4.1 of GL 83-28 states that "All vendor-recommended reactor trip 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."
This item of GL 83-28 also states that licensees should submit a statement confirming that this action has been implemented.
By letter dated November 29, 1983, the licensee responded to a number of GL 83-28 items, including Item 4.1.
In response to Item 4.1, the licensee stated that Units 2 and 3 utilize General Electric (GE) type AK-2 breakers (instead of the Westinghouse Type DB-50 breakers used at Salem). Nonetheless, the licensee stated that all vendor-recommended modifications had been reviewed and incorporated as appropriate. In addition, the licensee stated that recommendations for adjustment and maintenance of the undervoltage device and for preventive maintenance frequency, as set forth in NRC I&E Bulletin 79-09, had been incorporated into facility preventive maintenance procedure NPES-008.
-2 The licensee stated that additional informal vendor recommendations were provided by vendor representatives who were called to assist the licensee when reactor trip breaker difficulties were.experienced during,startup of Unit 2. The licensee adds that these recommendations have been implemented at Units 2 and 3 in the form of monthly surveillance testing and quarterly preventive maintenance of the breakers. The licensee also stated that based upon recent problems, the preventive maintenance frequency has been increased from quarterly to every two months.
III. CONCLUSION Based on the licensee's confirmation that the facility has implemented all vendor-recommended modifications existing at the,time of issuance of the generic letter, we conclude the licensee has satisfactorily completed for San Onofre Units 2 and 3, the actions prescribed by Item 4.1 of Generic Letter 83-28. Accordingly, this item is closed.
Principal Contributor:
G. Zwetzig