SER Re Util 831104 Response to Generic Ltr 83-28,Item 1.2, Post-Trip Review (Data & Info Capability). Licensee post-trip Review Data & Info Capabilities AcceptableML20127N313 |
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Farley |
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Issue date: |
06/12/1985 |
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NRC |
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Shared Package |
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ML20127N308 |
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References |
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GL-83-28, NUDOCS 8507010526 |
Download: ML20127N313 (10) |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20210T2161999-08-0606 August 1999 Draft SE Supporting Proposed Conversion of Current TS to ITS for Plant ML20206G7471999-05-0404 May 1999 Safety Evaluation Accepting Corrective Actions Taken by SNC to Ensure That Valves Perform Intended Safety Functions & Concluding That SNC Adequately Addressed Requested Actions in GL 95-07 ML20199D8611999-01-12012 January 1999 SER Accepting Relief Request for Inservice Insp Program for Plant,Units 1 & 2 ML20195E2281998-11-16016 November 1998 Safety Evaluation Authorizing Relief Request for Second 10-year ISI Program Relief Request 56 for Plant,Unit 1 ML20155E0271998-10-29029 October 1998 SER Approving & Denying in Part Inservice Testing Program Relief Requests for Plant.Relief Requests Q1P16-RR-V-3 & Q2P16-RR-V Denied Since Requests Do Not Meet Size Requirement of GL 89-04 ML20154B6121998-10-0101 October 1998 Safety Evaluation Granting Second 10-year ISI Requests for Relief RR-13 & RR-49 Through RR-55 for Jm Farley NPP Unit 1 ML20237C5471998-08-20020 August 1998 Suppl to SE Re Amends 137 & 129 to Licenses NPF-2 & NPF-8, Respectively.Se Being Supplemented to Incorporate Clarifications/Changes & Revise Commitment for Insp of SG U-bends in Rows 1 & 2 for Unit 2 Only ML20236U6141998-07-23023 July 1998 Safety Evaluation Authorizing Use of Alternative Alloy 690 Welds (Inco 52 & 152) as Substitute for Other Weld Metal ML20236R8671998-07-0909 July 1998 Safety Evaluation Concluding That Southern Nuclear Operating Co USI A-46 Implementation Program Has Met Purpose & Intent of Criteria in GIP-2 & Staff SSER-2 on GIP-2 for Resolution of USI A-46 ML20217D2591998-04-21021 April 1998 Safety Evaluation Accepting Licensee Proposed Alternative Re Augmented Exam of Reactor Vessel Shell Welds for Plant ML20217H3191998-03-31031 March 1998 Safety Evaluation Accepting Proposed Changes to Plant Matl Surveillance Programs ML20217D4081998-03-24024 March 1998 Safety Evaluation Accepting Proposed Changes to Maintain Calibration Info Required by ANSI N45.2.4-1972 ML20216H6731998-03-17017 March 1998 SER Accepting Quality Assurance Program Description Change for Joseph M Farley Nuclear Plant,Units 1 & 2 ML20199B0371998-01-22022 January 1998 SER Accepting Request for Relief (RR-27) for Plant,Units 1 & 2 from Certain Provisions of Section XI to ASME Boiler & Pressure Vessel Code.Relief Will Remove Insulation on ASME Code Class 1 Sys During Inservice Insp ML20198R5221997-10-29029 October 1997 Safety Evaluation Supporting Amends 132 & 124 to Licenses NPF-02 & NPF-08,respectively ML20216G9521997-09-0404 September 1997 Safety Evaluation Authorizing Request for Relief for IEEE 279-1971,Section 4.7.3 Requirements Concerning Steam Generator Water Level Control ML20236N3331997-08-21021 August 1997 SER Re Request for Interpretation of EDG TS 4.8.1.1.2.e for Farley Nuclear Plant,Units 1 & 2 ML20137E2951997-03-24024 March 1997 Safety Evaluation Supporting Amends 125 & 119 to Licenses NPF-2 & NPF-8,respectively ML20137B4371997-03-20020 March 1997 SER Accepting Request for Relief for 120-month Update of Facility Inservice Insp & Inservice Testing Programs & Code Addition & Addenda of Asme/Ansi Parts 6 & 10 ML20135E4811997-03-0404 March 1997 Safety Evaluation Accepting Implementation of 10CFR50.55a Requirements Related to Repair & Replacement Activities for Containment at Plant ML20056H1341993-08-23023 August 1993 Safety Evaluation Accepting Licensee 921217 Response to NRC 920917 SE Re Inservice Testing Program Relief Request ML20062D7001990-11-0909 November 1990 Safety Evaluation Supporting Util 881123 & 900917 Responses to Generic Ltr 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Matls & Its Effect on Plant Operations. Submittals Acceptable.Beltline Welds Discussed ML20245A8601989-06-13013 June 1989 Safety Evaluation Supporting Util 831104 & 850422 Responses to Generic Ltr 83-28,Item 4.5.3, Reactor Trip Sys Reliability for All Domestic Operating Reactors ML20195D5391988-10-31031 October 1988 Safety Evaluation Supporting ATWS Rule,10CFR50.62 ML20154C9651988-05-12012 May 1988 Safety Evaluation Re Flaw Indications in Reactor Pressure Vessel ML20147E2621987-11-16016 November 1987 Corrected Page 2 of Safety Evaluation Re Amends 74 & 66 to Licenses NPF-2 & NPF-8,respectively,deleting Ref to Quarterly Surveillance Testing on Staggered Test Basis ML20235K4441987-07-0808 July 1987 Safety Evaluation Supporting Granting Licensee Relief from Volumetric Exam of Steam Generator Primary Side Noozles Inside Radiused Sections ML20212E2241987-02-27027 February 1987 Safety Evaluation Accepting Util 831104 Response to Item 4.5.2 of Generic Ltr 83-28 Re on-line Functional Testing of Reactor Trip Sys,Including Independent Testing of Diverse Trip Features of Reactor Trip Breakers ML20212F5101987-01-0707 January 1987 Safety Evaluation Accepting Licensee 831104 Response to Generic Ltr 83-28,Item 2.1 (Part 1), Equipment Classification (Reactor Trip Sys Components) ML20211D5341987-01-0707 January 1987 Safety Evaluation Re Rev 1 to EGG-EA-6794, Conformance to Reg Guide 1.97,Joseph M Farley Nuclear Plant,Units 1 & 2 & Licensee Submittals.Response Acceptable ML20207C2671986-12-15015 December 1986 Safety Evaluation Accepting Licensee Responses to Generic Ltr 83-28,Item 2.1 (Part 2) & Item 2.2.2 Re Vendor Interface Programs for Reactor Trip Sys & All Other Site safety- Related Components ML20214Q1891986-11-17017 November 1986 Safety Evaluation Granting Relief Re Inservice Evaluation Criteria for Disposition of Linear Indication in Reactor Coolant pipe-to-safe End Weld on Cold Leg Pipe of Loop C ML20211H9811986-06-19019 June 1986 Safety Evaluation Supporting Util Request for Relief from Inservice Testing/Insp Requirements Re pressure-retaining Valve Body Welds & Internal Pressure Boundary Surfaces of Valves Exceeding 4 Inches Nominal Pipe Size ML20198C7851986-05-16016 May 1986 Safety Evaluation Concluding That Util Pressurized Thermal Shock Screening Criteria for Reactor Pressure Vessels Complies w/10CFR50.61 ML20140C9901986-03-19019 March 1986 Suppl 1 to Safety Evaluation Supporting Util 851114 Response to Generic Ltr 83-28,Item 3.2.2 Re Test & Maint Procedures ML20136H6821985-12-27027 December 1985 Safety Evaluation Granting Relief from Certain Inservice Testing/Insp Requirements Re Reactor Vessel Flange Ligaments,Reactor Coolant Pump Casing Internal Surfaces & Flange Bolts ML20136C4251985-11-12012 November 1985 Safety Evaluation Accepting Util 831104 & 850215 Responses to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing Requirements in Existing Tech Specs for Reactor Trip Sys Components ML20209J1941985-10-24024 October 1985 SER Accepting Licensee 831104 & 850422 Responses to Items 4.2.1 & 4.2.2 of Generic Ltr 83-28 Concerning Preventative Maint Program & Trending Parameters for DS-416 Type Reactor Trip Breakers,Respectively ML20135H3891985-09-12012 September 1985 Safety Evaluation Re Compliance W/License Condition 2.C.(12)(b),requiring Provisions to Assure That safety-grade Backup Means of RCS Depressurization Meets Requirements of Rev 1 to Branch Technical Position Rsb 5-1.Addl Info Needed ML20209G9691985-09-10010 September 1985 Safety Evaluation Re Generic Ltr 83-28,Items 3.1.1,3.1.2, 3.2.1,3.2.2,4.1 & 4.5.1.Addl Info Required for Item 3.2.2 Re Check of Vendor & Engineering Recommendations for Testing & Maint ML20127N3131985-06-12012 June 1985 SER Re Util 831104 Response to Generic Ltr 83-28,Item 1.2, Post-Trip Review (Data & Info Capability). Licensee post-trip Review Data & Info Capabilities Acceptable ML20129D5451985-05-21021 May 1985 SER Re Util 831104 Response to Generic Ltr 83-28,Item 1.1 Re post-trip Review Program & Procedures.Program & Procedures Acceptable Subj to Implementation of Listed Recommendations 1999-08-06
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217P0761999-10-0606 October 1999 Non-proprietary, Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs ML20217G0361999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20212E7451999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Hcgs,Unit 1.With Summary of Changes,Tests & Experiments Implemented During Aug 1999.With ML20216E4941999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Jmfnp.With ML20210T2161999-08-0606 August 1999 Draft SE Supporting Proposed Conversion of Current TS to ITS for Plant ML20211B2011999-08-0404 August 1999 Informs Commission About Results of NRC Staff Review of Kaowool Fire Barriers at Farley Nuclear Plant,Units 1 & 2 & Staff Plans to Address Technical Issues with Kaowool & FP-60 Barriers ML20210R6031999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20196J3791999-06-30030 June 1999 Safety Evaluation of TR WCAP-14750, RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop Pwrs. Rept Acceptable ML20209G0661999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With L-99-267, Monthly Operating Repts for June 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With L-99-023, Monthly Operating Repts for May 1999 for Jfnp Units 1 & 2. with1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Jfnp Units 1 & 2. with ML20206G7471999-05-0404 May 1999 Safety Evaluation Accepting Corrective Actions Taken by SNC to Ensure That Valves Perform Intended Safety Functions & Concluding That SNC Adequately Addressed Requested Actions in GL 95-07 L-99-020, Monthly Operating Repts for Apr 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20206C9461999-04-30030 April 1999 1:Final Cycle 16 Freespan ODSCC Operational Assessment L-99-161, Monthly Operating Repts for Mar 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20205N0961999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20204D7271999-03-15015 March 1999 ISI Refueling 15,Interval 2,Period 3,Outage 3 for Jm Farley Nuclear Generating Plant,Unit 1 ML20207M6421999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20203A2651999-01-31031 January 1999 Monthly Operating Repts for Jan 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20199D8611999-01-12012 January 1999 SER Accepting Relief Request for Inservice Insp Program for Plant,Units 1 & 2 ML20199E6591998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20206C8081998-12-31031 December 1998 Alabama Power 1998 Annual Rept ML20198K4091998-12-18018 December 1998 COLR for Jm Farley,Unit 1 Cycle 16 ML20198B2561998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20195E2281998-11-16016 November 1998 Safety Evaluation Authorizing Relief Request for Second 10-year ISI Program Relief Request 56 for Plant,Unit 1 ML20195C9681998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20155E0271998-10-29029 October 1998 SER Approving & Denying in Part Inservice Testing Program Relief Requests for Plant.Relief Requests Q1P16-RR-V-3 & Q2P16-RR-V Denied Since Requests Do Not Meet Size Requirement of GL 89-04 ML20154B6121998-10-0101 October 1998 Safety Evaluation Granting Second 10-year ISI Requests for Relief RR-13 & RR-49 Through RR-55 for Jm Farley NPP Unit 1 ML20151V8341998-09-30030 September 1998 Non-proprietary Rev 2 to NSA-SSO-96-525, Jm Farley Nuclear Plant Safety Analysis IR Neutron Flux Reactor Trip Setpoint Change ML20154H6001998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20154H0121998-09-30030 September 1998 Submittal-Only Screening Review of Farley Nuclear Plant IPEEE (Seismic Portion) ML20197C8991998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20237C5471998-08-20020 August 1998 Suppl to SE Re Amends 137 & 129 to Licenses NPF-2 & NPF-8, Respectively.Se Being Supplemented to Incorporate Clarifications/Changes & Revise Commitment for Insp of SG U-bends in Rows 1 & 2 for Unit 2 Only ML20236Y1121998-07-31031 July 1998 Voltage-Based Repair Criteria 90-Day Rept ML20237B1891998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Jm Farley Nuclear Plant,Units 1 & 2 ML20237A2181998-07-24024 July 1998 Jm Farley Unit 2 ISI Rept Interval 2,Period 3 Outage 1, Refueling Outage 12 ML20236U6141998-07-23023 July 1998 Safety Evaluation Authorizing Use of Alternative Alloy 690 Welds (Inco 52 & 152) as Substitute for Other Weld Metal ML20236R8671998-07-0909 July 1998 Safety Evaluation Concluding That Southern Nuclear Operating Co USI A-46 Implementation Program Has Met Purpose & Intent of Criteria in GIP-2 & Staff SSER-2 on GIP-2 for Resolution of USI A-46 ML20236M5981998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Jm Farley Nuclear Plant,Units 1 & 2 ML20154H0461998-06-30030 June 1998 Technical Evaluation Rept on Review of Farley Nuclear Plant IPEEE Submittal on High Winds,Flood & Other External Events (Hfo) ML20248M3121998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Jm Farley Nuclear Plant,Units 1 & 2 ML20247F3631998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Jm Farley Nuclear Plant,Units 1 & 2 ML20217D2591998-04-21021 April 1998 Safety Evaluation Accepting Licensee Proposed Alternative Re Augmented Exam of Reactor Vessel Shell Welds for Plant ML20247E8851998-03-31031 March 1998 FNP Unit 2 Cycle 13 Colr ML20217H3191998-03-31031 March 1998 Safety Evaluation Accepting Proposed Changes to Plant Matl Surveillance Programs ML20216D5941998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Joseph M Farley Nuclear Plant,Units 1 & 2 ML20217D4081998-03-24024 March 1998 Safety Evaluation Accepting Proposed Changes to Maintain Calibration Info Required by ANSI N45.2.4-1972 ML20216H6731998-03-17017 March 1998 SER Accepting Quality Assurance Program Description Change for Joseph M Farley Nuclear Plant,Units 1 & 2 ML20216J6851998-03-16016 March 1998 Revised Pages 58 & 59 to Fnp,Units 1 & 2,Power Uprate Project BOP Licensing Rept ML20216D9811998-02-28028 February 1998 Monthly Operating Repts for Feb 1998 for Jm Farley Nuclear Plant,Units 1 & 2 1999-09-30
[Table view] |
Text
. _ _ _ _ _ _ _ _ _ _ _ _ __
ENCLOSURE 1 SAFETY EVALUATION REPORT FOR -
GENERIC LETTER 83-28 ITEM 1.2 - POST-TRIP REVIEW {-.
(DATAANDINFORMATIONCAPABILITY)
JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS.: 50-348, 50-364 I. INTRODUCTION l On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system. This incident occurred during the l 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 i
under voltage trip attachment. Prior to this incident, on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was generated based on steam generator low-low level during plant start-up. 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 (ED0), directed the staff to investigate and report on the generic implications of these occurrences at Unit I of the Salem Nuclear Power Plant. The results of the staff's inquiry into the generic implications of the Salem unit incidents are reported in NUREG-1000, " Generic Implications of the ATWS Events at the Salen Nuclear Power Plant." As a result of this investigation, the Comission (NRC) requested (by Generic Letter 83-28 dated July 8, 1983) all licensees of operating reactors, applicants for an operating license, and holders of 0507010526 850612 PDR ADOCK 0D000348 p PDR
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. 4 construcdon pemits to respond to certain generic concerns. ~Thhe concerns are categorize, ,into four areas: (1) Post-Trip Review, (2) Equipment Classification and Vendor s Interface, (3) Port-Maintenance Testing, and
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(4) Reactor Trip Shstem Reliabilf t Vrprovements.
.( ,
e <
q The fft st action iten, Post-Trip Review, consists of Action Item 1.1, I
s
" Program Description and Procedure" and Action Item 1.2, " Data and Inforw tion Cahability." This safety evaluation report (SER) addresses Action Item 1.2 only. t
- 1 II. REVIEW GintELDlES
, , s ',
The following review guidelines were developed after initial evaluation of the various utDity responses to Item 1.2 of Generic- Letter 83-28 and incor), orate the beit features of these sLbnittals. As such, these review guidelines in effect represent a " goo 6 practices" approach to post-trip review. Wihave reviewed the licensee's response to Item 1.2 against these
. guidelines': '
s, A. The equipment that provides the digital sequence of events (SOE) record e, <
and the analog time history records of <an unscheduled shutdown should 3
y provide ,a reliable source of the necessary information to ' ne* used in the l
post-trip review. Each plant, variable (Sich is necessary to determine I theecause and progression of the events following a plant trip should be
. monitored by at least one recorder (such as a slen:re. nce-of-events
, t .
l; recorder or 3, plant process computer) for digital parameters; and~ strip
^
3
%*f' 't i l' \
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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:
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 recomended 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.
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
e 4-Chapter 15 of the plant FSAR. The recomended guidel_ine 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 equipnent, each analog time history data recorder should be capable of updating and retaining infomation from approximately five minutes prior to the trip until at least ten minutes after the trip.
All equipment used to record sequence of events and time history infomation should be powered from a reliable and non-interruptible power source. The power source used need not be safety related.
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 infomation 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 unscheduled 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
systems should be recorded for use in the post-trip review. The parameters deemed necessary, as a minimum, to perform a pos.t-trip review that woul'd 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 infomation should be presented in a readable and meaningful fomat, 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
retained for the life of the plant for post-trip review co parisons of subsequent events.
III. EVALUATION AND CONCLUSION By letter dated November 4,1983, Alabama Power Company provided information regarding its post-trip review program data and information capabilities for Joseph M. Farley Nuclear Plant Units I and 2. We have evaluated the licensee's submittal a; gainst the review guidelines described in Section II.
Licensee deviations from the Guidelines of Section II were reviewed with the licensee by telephone on May 23 and 29, 1985. A brief description of the
-4 licensee's responses and the staff's evaluation of the response against each
. %^
of the review guidelines is provided below: ,
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, we find that the sequence of events recorder characteristics conform to the guidelines described in Section II A, and are acceptable. The time history recorder characteristics conform to these guidelines, except for pre-and post-trip recording durations. During our telephone calls with the licensee, the licensee stated that time history recorder duration is supplemented with strip chart recorders powered by non-interruptible power sources. Based upon this information, we find that the licensee meets the intent of the guidelines described in Section II A.
l B. The licensee has established and identified the parameters o be monitored and recorded for post-trip review. Based on our. review of the licensee's submittal and on information obtained during our telephone reviews, we find that the parameters selected by the licensee include all but one of those identified in Table 1. The licensee stated that while PORV position is not recorded on a sequence of event recorder, there are indirect indications in the control room which, when combined with special procedures, provide the desired infomation. The staff finds this alternative acceptable.
The licensee does not record all of the sequence of events and time history parameters in the specific manner reconinended in Table 1.
However, based upon information provided by the licensee during our telephone reviews, we find that the licensee has alternative data sources for thoses parameters not recorded on the sequence of events recorders and time history recorders. These include: (1)analarm typewriter, (2) the SPDS with hardcopy capability, and (3) strip chart recorders with non-interruptible power supplies. Consequently, we find that the licensee's selection of parameters meet the intent of the guidelines described in Section II B and is, therefore, acceptable.
C. The licensee has described the means for storage and retrieval of the information gathered by the sequence of events and time history recorders, and for the presentation of this infonnation for post-trip review and analysis. Based on our review, we find that this information
will be presented in a readable and meaningful format, and that the storage, retrieval and presentation conform to the guidelin.es of Section II C.
D.
The licensee's submittal indicates that the data and information used -
during post-trip reviews will be retained in an accessible manner for the life of the plant. Based on our review, 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, we conclude that the licensee's post-trip review data and information capabilities for Joseph M. Farley Nuclear Plant Units I and 2 are acceptable. '
Principal Contributor:
J. J. Kramer 9
9-
~
TABLE 1 PWR PARAMETER LIST -
SOE Time History 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
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) 4 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.
1
- I
, . . _ . . . . . _ _ _ _ _ _ _ _ _ _ . - .