SER Re Util 831105,840229 & 860404 Responses to Generic Ltr 83-28,Item 1.2, Post-Trip Review Data & Info Capabilities. Responses AcceptableML20197H226 |
Person / Time |
---|
Site: |
Byron |
---|
Issue date: |
05/05/1986 |
---|
From: |
NRC |
---|
To: |
|
---|
Shared Package |
---|
ML20197F122 |
List: |
---|
References |
---|
GL-83-28, NUDOCS 8605190093 |
Download: ML20197H226 (10) |
|
Similar Documents at Byron |
---|
Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217N3711999-10-13013 October 1999 Safety Evaluation Supporting Amends 111,111,104 & 104 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20211A8131999-08-10010 August 1999 Safety Evaluation Supporting Amends 110 & 110 to Licenses NPF-37 & NPF-66,respectively ML17191B4101999-07-27027 July 1999 Safety Evaluation Accepting Alternative to Licenses NPF-72, NPF-77,NPF-37,NPF-66,NPF-19,NPF-25,NPF-11,NPF-18,DPR-29 & DPR-30,respectively ML20207B6481999-05-25025 May 1999 SER Accepting Revised SGTR Analysis for Byron & Braidwood Stations.Revised Analysis Was Submitted to Support SG Replacement at Unit 1 of Each Station ML20206G6031999-05-0303 May 1999 Safety Evaluation Supporting Amends 108,108,101 & 101 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20206B5531999-04-23023 April 1999 Safety Evaluation Supporting Amends 107,107,100 & 100 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20205B5091999-03-26026 March 1999 SER Accepting Relief Requests 12R-24,Rev 0 & 12R-34,Rev 0, Related to Second 10-year Interval Inservice Insp for Byron Station,Units 1 & 2 ML20205C5101999-03-21021 March 1999 Revised Safety Evaluation Supporting Improved TS Amends Issued by NRC on 981222 to FOLs NPF-37,NPF-66,NPF-72 & NPF-77.Revised Pages Include Editorial Corrections ML20204G3831999-03-19019 March 1999 Safety Evaluation Accepting Second 10-yr Interval ISI Request for Relief 12R-11 ML20154N1641998-10-15015 October 1998 Safety Evaluation Supporting Amends 105,105,97 & 97 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20154F5071998-10-0606 October 1998 Safety Evaluation Supporting Amends 104,104,96 & 96 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20238F6551998-08-28028 August 1998 SE Authorizing Licensee Request for Relief NR-20,Rev 1 & NR-25,Rev 0 Re Relief from Examination Requirement of Applicable ASME BPV Code,Section XI for First ISI Interval Exams ML20248E0361998-05-26026 May 1998 Safety Evaluation Supporting Amends 103,103,93 & 93 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20216C1341998-05-0808 May 1998 Safety Evaluation Supporting Amends 102 & 102 to Licenses NPF-37 & NPF-66,respectively ML20217K7171998-04-20020 April 1998 Safety Evaluation Accepting Requests for Relief NR-22,NR-23 & NR-24 for First 10-yr Insp Interval ML20216F4921998-03-11011 March 1998 Correction to Safety Evaluation Re Revised SG Tube Rapture Analysis ML20212H1851998-03-0606 March 1998 SE Approving Temporary Use of Current Procedure for Containment Repair & Replacement Activities Instead of Requirements in Amended 10CFR50.55a Rule ML20203A2821998-02-0303 February 1998 Safety Evaluation Supporting Amends 101,101,92 & 92 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20199E7191998-01-29029 January 1998 Safety Evaluation Supporting Amends 99,99,90 & 90 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20199G2591998-01-28028 January 1998 Safety Evaluation Rept Accepting Revised SG Tube Rupture Analysis ML20199J0371998-01-23023 January 1998 Safety Evaluation Supporting Amends 98,98,89 & 89 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20199F3751998-01-22022 January 1998 Safety Evaluation Supporting Amends 97,97,88 & 88 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20199H0031998-01-21021 January 1998 SER Accepting Pressure Temp Limits Rept & Methodology for Relocation of Reactor Coolant Sys pressure-temp Limit Curves & Low Temp Overpressure Protection Sys Limits for Byron Station,Units 1 & 2 & Braidwood Station,Units 1 & 2 ML20199A7011998-01-16016 January 1998 SER Approving Exemption from Requirements of 10CFR50.60 for Byron Station,Units 1 & 2 & Braidwood Station,Units 1 & 2 ML20199C1401998-01-16016 January 1998 SER Accepting Request to Integrate Reactor Vessel Weld Metal Surveillance Program for Byron,Units 1 & 2 & Braidwood,Units 1 & 2 Per 10CFR50 ML20199C1721998-01-15015 January 1998 Safety Evaluation Supporting Amends 96,96,87 & 87 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20199C1231998-01-13013 January 1998 Safety Evaluation Granting Second 10-yr Inservice Insp Program Plan Relief Request ML20203D4081997-12-0404 December 1997 Safety Evaluation Supporting Amends 94,94,86 & 86 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20198H3211997-12-0303 December 1997 Safety Evaluation Re Licensee Submittal of IPE for Plant, Units 1 & 2,in Response to GL 88-02, IPE for Severe Accident Vulnerabilities ML20202E5781997-11-25025 November 1997 Safety Evaluation Supporting Amends 93 & 93 to Licenses NPF-37 & NPF-66,respectively ML20211L2151997-10-0303 October 1997 Safety Evaluation Supporting Licensee Relief Request,Per 10CFR50.55a(a)(3)(i) ML20217C1681997-09-22022 September 1997 Safety Evaluation Accepting Request for Relief from ASME Code,Section Iii,Div 2 for Repair of Damaged Concrete Reinforcement Steel ML20210K4151997-08-13013 August 1997 Safety Evaluation Supporting Amends 92,92,85 & 85 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20210K7091997-08-13013 August 1997 Safety Evaluation Supporting Amends 91 & 84 to Licenses NPF-66 & NPF-77,respectively ML20149B7671997-07-10010 July 1997 Safety Evaluation Supporting Amends 91,90,84 & 83 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20141L9321997-05-29029 May 1997 Safety Evaluation Accepting Use of ASME Boiler & Pressure Vessel Code,Section Ix,Code Cases 2142-1 & 2143-1 for Reactor Coolant Sys for Plants ML20138K0351997-05-0606 May 1997 Safety Evaluation Supporting Amends 89,89,81 & 81 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20137V5161997-04-15015 April 1997 Safety Evaluation Supporting Amends 87,87,79 & 79 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20137L5061997-04-0202 April 1997 Safety Evaluation Supporting Amends 86,86,78 & 78 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20134K8141997-02-12012 February 1997 Safety Evaluation Supporting Amends 85,85,77 & 77 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20134L7811996-11-18018 November 1996 Safety Evaluation Granting Listed Relief Requests,Per 10CFR50.55a(f)(6)(i) Based on Impracticalities in Design of Valves That Limit IST in Traditional Manner Using Position Indicating Lights ML20129F9101996-10-25025 October 1996 Safety Evaluation Accepting Request to Apply LBB Analyses to Eliminate Large Primary Loop Pipe Rupture from Structural Design Basis for Plant,Units 1 & 2 ML20113C2341996-06-26026 June 1996 Safety Evaluation Supporting Amends 84 & 76 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20101R3941996-04-12012 April 1996 Safety Evaluation Supporting Amends 83,83,72 & 75 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20101R5521996-04-10010 April 1996 Safety Evaluation Supporting Amends 82,82,74 & 74 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20101Q0961996-04-0404 April 1996 Safety Evaluation Supporting Amends 81,81,73 & 73 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20149L2241996-02-16016 February 1996 Safety Evaluation Supporting Amends 71 & 79 to Licenses NPF-37,NPF-66,NPF-72,NPF-77,respectively ML20094L1241995-11-0909 November 1995 Safety Evaluation Supporting Amends 77,77,69 & 69 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20094E5751995-11-0202 November 1995 Safety Evaluation Supporting Amends 76,76,86 & 86 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20085N2121995-06-22022 June 1995 Safety Evaluation Supporting Amends 72,72,63 & 63 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively 1999-08-10
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217N3711999-10-13013 October 1999 Safety Evaluation Supporting Amends 111,111,104 & 104 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20217H5221999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Byron Station, Units 1 & 2.With ML20217P6351999-09-29029 September 1999 Non-proprietary Rev 6 to HI-982083, Licensing Rept for Spent Fuel Rack Installation at Byron & Braidwood Nuclear Stations ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20212B9261999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Byron Station,Units 1 & 2.With 05000454/LER-1998-005, :on 980307,reactor Manually Tripped.Caused by Indeterminate Rod Sequencing Problem.Halted Rod Insertion & Exercised Bank Overlap Thumbwheel Switch to Clean Contacts. with1999-08-20020 August 1999
- on 980307,reactor Manually Tripped.Caused by Indeterminate Rod Sequencing Problem.Halted Rod Insertion & Exercised Bank Overlap Thumbwheel Switch to Clean Contacts. with
ML20211A8131999-08-10010 August 1999 Safety Evaluation Supporting Amends 110 & 110 to Licenses NPF-37 & NPF-66,respectively ML20210R3431999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Byron Station, Units 1 & 2.With ML17191B4101999-07-27027 July 1999 Safety Evaluation Accepting Alternative to Licenses NPF-72, NPF-77,NPF-37,NPF-66,NPF-19,NPF-25,NPF-11,NPF-18,DPR-29 & DPR-30,respectively ML20210E2251999-07-21021 July 1999 B1R09 ISI Summary Rept Spring 1999 Outage, 980309-990424 ML20209G1751999-07-0808 July 1999 SG Eddy Current Insp Rept,Cycle 9 Refueling Outage (B1R09) ML20207H7851999-06-30030 June 1999 Rev 0 to WCAP-15183, Commonwealth Edison Co Byron,Unit 1 Surveillance Program Credibility Evaluation ML20207H8071999-06-30030 June 1999 Rev 0 to WCAP-15178, Byron Unit 2 Heatup & Cooldowm Limit Curves for Normal Operations ML20207H7771999-06-30030 June 1999 Rev 0 to WCAP-15177, Evaluation of Pressurized Thermal Shock for Byron,Unit 2 ML20207H7941999-06-30030 June 1999 Rev 0 to WCAP-15180, Commonwealth Edison Co Byron,Unit 2 Surveillance Program Credibility Evaluation ML20209H3711999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Byron Station, Units 1 & 2.With ML20207H7561999-06-28028 June 1999 Pressure Temp Limits Rept (Ptlr) ML20207H7621999-06-28028 June 1999 Pressure Temp Limits Rept (Ptlr) 05000454/LER-1999-003-02, :on 990513,automatic Reactor Occurred Due to Human Error During Surveillance Procedure.Appropriate Mgt Action Taken with Instrument Maint Involved & Sp Bsir 3.1.6-200 Revised.With1999-06-11011 June 1999
- on 990513,automatic Reactor Occurred Due to Human Error During Surveillance Procedure.Appropriate Mgt Action Taken with Instrument Maint Involved & Sp Bsir 3.1.6-200 Revised.With
05000454/LER-1999-002-03, :on 990509,missed TS Surveillance Was Noted. Caused by Error in Design Package.Changed Monthly Surveillance Procedure for Containment Isolation Valves1999-06-0808 June 1999
- on 990509,missed TS Surveillance Was Noted. Caused by Error in Design Package.Changed Monthly Surveillance Procedure for Containment Isolation Valves
ML20195J8001999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Byron Station,Units 1 & 2.With ML20207B6481999-05-25025 May 1999 SER Accepting Revised SGTR Analysis for Byron & Braidwood Stations.Revised Analysis Was Submitted to Support SG Replacement at Unit 1 of Each Station 05000454/LER-1999-001-05, :on 990422,depressing Both Feedwater Isolation Reset Pushbuttons Leads to LCO 3.0.3 Entry.Caused by Implementation Error in Procedure Review for Conversion to Its.Review of Procedures Will Be Conducted1999-05-21021 May 1999
- on 990422,depressing Both Feedwater Isolation Reset Pushbuttons Leads to LCO 3.0.3 Entry.Caused by Implementation Error in Procedure Review for Conversion to Its.Review of Procedures Will Be Conducted
ML20195B2591999-05-19019 May 1999 Rev 66a to CE-1-A,consisting of Proposed Changes to QAP for Dnps,Qcs,Znps,Lcs,Byron & Braidwood Stations ML20206G6031999-05-0303 May 1999 Safety Evaluation Supporting Amends 108,108,101 & 101 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20206R6991999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Byron Station Units 1 & 2.With ML20195C7961999-04-28028 April 1999 Rev 2 to NFM9800233, Byron Station Unit 2 COLR for Cycle 8A (BY2C8A) M980023, Rev 2 to NFM9800233, Byron Station Unit 2 COLR for Cycle 8A (BY2C8A)1999-04-28028 April 1999 Rev 2 to NFM9800233, Byron Station Unit 2 COLR for Cycle 8A (BY2C8A) ML20206B5531999-04-23023 April 1999 Safety Evaluation Supporting Amends 107,107,100 & 100 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively ML20205N9241999-03-31031 March 1999 Analysis of Capsule X from CE Byron Unit 2 Rv Radiation Surveillance Program ML20205P7001999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Byron Station,Units 1 & 2.With ML20205B5091999-03-26026 March 1999 SER Accepting Relief Requests 12R-24,Rev 0 & 12R-34,Rev 0, Related to Second 10-year Interval Inservice Insp for Byron Station,Units 1 & 2 ML20205C5101999-03-21021 March 1999 Revised Safety Evaluation Supporting Improved TS Amends Issued by NRC on 981222 to FOLs NPF-37,NPF-66,NPF-72 & NPF-77.Revised Pages Include Editorial Corrections ML20204G3831999-03-19019 March 1999 Safety Evaluation Accepting Second 10-yr Interval ISI Request for Relief 12R-11 M990004, Rev 0 to NFM9900043, Byron Unit 1,Cycle 10 COLR in ITS Format W(Z) Function1999-03-17017 March 1999 Rev 0 to NFM9900043, Byron Unit 1,Cycle 10 COLR in ITS Format W(Z) Function ML20206A8831999-03-17017 March 1999 Rev 0 to NFM9900043, Byron Unit 1,Cycle 10 COLR in ITS Format W(Z) Function ML20207M9231999-03-12012 March 1999 Amended Part 21 Rept Re Cooper-Bessemer Ksv EDG Power Piston Failure.Total of 198 or More Pistons Have Been Measured at Seven Different Sites.All Potentially Defective Pistons Have Been Removed from Svc Based on Encl Results ML20204H9941999-03-0303 March 1999 Non-proprietary Rev 4 to HI-982083, Licensing Rept for Spent Fuel Rack Installation at Byron & Braidwood Nuclear Stations ML20204C7671999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Byron Station,Units 1 & 2.With ML20199C9721999-01-31031 January 1999 Non-proprietary Version of Rev 1 to WCAP-15123, Analysis of Capsule W from Commonwealth Edison Co Byron Unit 1 Reactor Vessel Radiation Surveillance Program ML20205M7061998-12-31031 December 1998 Unicom Corp 1998 Summary Annual Rept. with ML20202F6021998-12-31031 December 1998 Cycle 9 COLR in ITS Format & W(Z) Function ML20196K6731998-12-31031 December 1998 10CFR50.59 Summary Rept for 1998 ML20199G8271998-12-31031 December 1998 Rev 1 Comm Ed Byron Nuclear Power Station,Unit 1 Cycle 9 Startup Rept ML20199E6371998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Byron Station,Units 1 & 2.With ML20202F6181998-12-31031 December 1998 Cycle 8 COLR in ITS Format & W(Z) Function ML20206B4001998-12-31031 December 1998 Annual & 30-Day Rept of ECCS Evaluation Model Changes & Errors for Byron & Braidwood Stations 05000454/LER-1998-020-01, :on 981118,analysis Error in Boron Dilution Protection Sys Code Identified.Caused by Personnel Error. Replacement of Bdps Function with Volume Control Tank Level Monitoring & Operator Actions Evaluated1998-12-17017 December 1998
- on 981118,analysis Error in Boron Dilution Protection Sys Code Identified.Caused by Personnel Error. Replacement of Bdps Function with Volume Control Tank Level Monitoring & Operator Actions Evaluated
ML20207H8011998-11-30030 November 1998 Rev 0 to WCAP-15124, Byron Unit 1 Heatup & Cooldown Limit Curves for Normal Operation ML20198D1501998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Byron Nuclear Power Station,Units 1 & 2.With 1999-09-30
[Table view] |
Text
. _
e ENCLOSURE 1 SAFETY EVALUATION REPORT FOR GENERIC LETTER 83-28, ITEM 1.2 - POST-TRIP REVIEW (DATA AND INFORMATION CAPABILITY)
BYRON STATION DOCKET NOS.: 50-454/455 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 th'e 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 undervoltage 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 (ED01 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 re-ported in NUREG-1000. " Generic Implications of ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Comission requested (by i
Generic Letter 83-28 dated July 8,1983) all licensees of operating reactors, applicants for an operating license, and holders of construction permits'to respond to certain generic concerns. These concerns are categorized into four (1) Post-Trip Review, (2) Equipment Classification and Vendor Interface, areas:
(3) Post-Maintenance Testing, and (4) Reactor Trip System Reliability Improve-ments.
9605190093 860505 PDR ADOCK 05000454 P
PDR N
L
The first action item, Post-Trip Review, consists of Action Item 1.1, " Program Description and Procedure" and Action Item 1.2, " Data and Infonnation 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 re-viewed the applicant's response to Item 1.2 against these guidelines:
i 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 charts, a plant process computer or analog recorder for analog (time history) i 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.
Each analog time history data recorder should have a sample o
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 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 functinning 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.
All equipment used to record sequence of events and time history o
infonnation should be powered from a reliable and non-interruptible power source. The power source used need not be Class 1E.
B.
The sequence of events and time history recording equipment should 3
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 t
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 para-meters deemed necessary, as a minimum, to perform a post-trip review that would determine if the plant remained within its safety limit design envelo'pe 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 licen-see should show that the existing set of monitored parameters ure suffi-cient to establish that the plant remained with.in 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., com-puter printout, strip chart record), or in an accessible memory (e.g.,
magnetic disc or tape). This information should be presented in a read-able 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 equipnent 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 comparisons of subsequent events.
III. EVALUATION AND CONCLUSION By letters dated February 29, 1984 and April 4, 1986, Commonwealth Edison Company provided information regarding its post-trip review program data and information capabilities for Byron Station. We have evaluated the licensee's submittals 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 February 7,1986. 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 ifcensee's,
submittals and our telephone conversation, we find that the sequence of events recorder and time history recorder characteristics conform to the guidelines described in Section II A, and are acceptable.
I
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 therefore are acceptable.
C.
The licensee describ,ed 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 information 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 guidelines of Section II C.
D.
The licensee's submittal of April 4,1986 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 this information, we find that the licensee's program for data retention conforms to the guidelines of Section II D, and is acceptable.
y
Based on our review of the licensee's submittals and our telephone conversa-tion with the licensee, we conclude that the licensee's post-trip review data and information capabilities for Byron Station are acceptable.
m
--n-r 4
w-
,c-c.e 4
,,.g,,,,,
,y,._
TABLE 1 PWR PARAMETER LIST SOE Time History Recorder Recorder Pa rameter/ 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 Containnent Pressure (2)
Containment Radiation x
Containnent 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 O
..n.
e '
SOE Time History Recorder Recorder Parameter / Signal (3)
Auxiliary Feedwater System: Flow, Pump / Valve Status AC and DC System Status (Bus Voltage) x x
Diesel Generator Status (Start /Stop, i
l_
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.
,