ML19350D741
| ML19350D741 | |
| Person / Time | |
|---|---|
| Issue date: | 04/17/1981 |
| From: | NRC - LOFT SPECIAL REVIEW GROUP |
| To: | |
| References | |
| NUREG-0758, NUREG-0758-ERR, NUREG-758, NUREG-758-ERR, NUDOCS 8105190026 | |
| Download: ML19350D741 (14) | |
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,d April 17, 1981 ERRATA SilEET gis f-{$ {}N'x FOR NUREG-0758 6 A. $ NR 2 e magi $: Ag.%s.,cfmyEN""[d m- ! IP REPORT OF THE LOFT SPECIAL REVIEW GR ,, p v Published: February 1981 (v U.S. Nuclear Regulatory Commission Please correct your copy of the report with the revised attached pages. Delete Page Replace With 7-6 7-6 7-7 7-7 7-9 7-9 7-9a None (new page) 7-10 7-10 7-10a None (new page) Please note that the " List of Tables" found on page ix does not need to be corrected. DIVISION OF TECilNICAL INFORMATION AND DOCUMENT CONTROL 01051900W
Table 7.1 Risk-Significant 5:qu;nces - PWRs Release Plant Ca tegory* Sequence Probability
- Description Surry**
2 TMLB'-6 2x10-6 Station blackout with loss of AFWS 2 TMLB'-y 7x10'7 Station blackrut with loss of AFWS 2 V. 4x10-6 Interfacing-systems LOCA 3 S C-6 2x10-6 Small LOCA, sprays fail, no water for 2 ECC/sp;ay recirculation Sequoyah*** 2 S HF-y 5x10-6 Small LOCA, ice condenser drains blocked 2 2 V 5x10-6 Interfacing-systems LOCA 3 $ H-y 2x10-5 Small LOCA, ECCS fails in recirculation 2 3 S HF-6,y 3x10-6 See S HF y j 2 3 TML-y 3x10-6 Total loss of feedwater Oconee*** 2 V 7x10-5 Interfacing systems LOCA 2 T KMUO'-y 4x10-6 ATWS, no HPI, sprays 2 3 T MQ-CH-y 7x10-6 Sf.ack-open PORV, no ECC recirculation, 2 sprays 3 S CH-y 6x10-6 Small LOCA, no ECC recirculation, sprays 3 Calvert 2 TMLOO'-6 lx10~4 Total loss of feedwater, no containment E5Fs Cliffs *** 2 TMQ-FH-y lx10-5 Stuck-open PORV, no ECC, spray recirculation 2 TMLOO'-y 1x10-5 See TMLOO'-6 3 TML-y 2x10-3 Total loss of feedwater 3 TMQ-D-y 1x10-4 Stuck-open PORV, ECC injection fails 3 TMLQ-y 7x10-5 Total loss of feedwater, stuck open P02V 3 TMQ-H-y 6x10-5 Stuck-open PORV, ECCS fails in recirculation 3 TKML y 5x10-5 ATWS with loss of all feedwater r
- Key to release categories is listed on page 7-8.
- cSee key - page 7-9 N See key - page 7-9a 7-6
Tablo 7.2 Risk-Significant Sequenc s - BWRs Release Pla:.t Ca tegory* Sequence Probability, Description Peach '2 TW-y' .3x10-6 Long-tem loss of decay heat removal, Bottom **' ECCS fails on containment failure 3 TW-y 1x10-5 See TW-y' 3 TC-y lx10-5 ATWS Grand 2 TPI-6 8x10~4 Stuck open relief valve, RHR failure Gulf ***- 2 TQW-6 3x10-5 Long term loss of heat removal 3 TC-y 5x10-5 ATWS CK y to release categories is listed on page 7-8.
- See key - page 7-10
- cSee key - page 7-10a 7-7
e KEY TO PWR ACCIDENT, SEQUENCE SYMBOLS (Surry plant) A = Intermeetate to large 1DCA. D - railure of electric power to F.srs. C' - railure to recover either onsite or ot'fsite electric power within about 1 to 3 hours following an initiating transient which is a loss of offsite AC power. C - railure of the containment spray injection system. 9 - Failure of the emergency core cooling injection system. F - railure of the containment spray recirculation eystem. C - railure of the containment heat removal system. O - Failure of the esergency core cooling recirculation system. K - railure of the reactor protection system. L - Failure of the secondary system steam relief valves and the auxiliary feedwater system. Q - railure of the secondary system steam relief valves and the power conversion system. 9 - railure of the primary system safety relief valves to reclose after opening. Q - Massive rupture of the reactor vessel. O - A small IDCA with an equivalent diameter of about 2 to 6 inches. g 5 - A small IDCA with as equivalent diameter of about 1/2 to 2 inches. 2 T - Transient event. V - LPIS check valve failure. o - Containment rupture due to a reactor vessel steam explosion. 8 - Containment failure resulting from inadequate isolation of containment openings and penetrations. Y - Containment failure due to hydrogen burning. 4 - Containment failure due to overpressure. C - Containment vessel melt-through. 0 - Failure of containment heat removal function. U - Failure of decay heat removal by high pressure makeup and bleed. 7-9 P00R E M
d Key _to_PWR Accident Sequence Symbols. ~ (Sequoyah,'Oconee,CalvertCliffsplants) Initiating Events,, Tj - Loss of Offsite Power Transient T Loss of Power Conversion System Transient Caused By Other Than a 2 loss of Offsite Power T3 - Reactor Shutdowns with the Power Conversion System Initially Available Sy - Intermediate LOCA (10"<D_<13.5") S2 - Small LOCA (4"<D<10") 53 - Small-Small 1.0CA (D<4") V - Interfacing Systems LOCA System Failures .(B )-Emergency Power System 3 C - Containment Spray Injection System D - Emergency Coolant Injection System F - Containment Spray Recirculation System H - Eme. 'ency Coolant Recirculation Systen K - Reactor Protection System L - Emergency Feedwater System M - Power Conversion System-(Normal Operation) 0 - Reactor Building Cooling System O'- Containment sprey System I Q ~ T:eclosure of Pressurizer Safety / Relief Vsives U - High Pressure Injection System Containment Failure Modes o - Yessel Steam Explosion 4 8 - Penetration Leakage Y - Overpressure Due to Hydrogen Burning 4 - Overpressure Due to Steam or Non Condensible Gases 4 c - Base Mat Melt Through 7-9a i
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l KEY 70 BWR ACCIDENT SEQUENCE SYMBOLS (Peach Bottom plant) - Ruptt,re of reactor coolant boundary with an equivalent diameter of greater than six inches. A 3 - railure of electric power to ISrs. C - Failure of the reactor protection system. D - ra11ure of vapor suppression. E - Failure of emergency core cooling injection. r - Failure of ernergency core cooling functionability. ~ G - Failure of containment isolation to limit leakage to less than 100 volume per cent per day. H - railure of core spray recirculation system. I - railure of low pressure recirculation system. J - Tallure of high pressure service water system. M - Failure of safety / relief valves to open. P - Tallure of safety / relief valves to reclose after opening. Q - Failure of normal feedwater system to provide core sake-up water. S - Small pipe break with an equivalent diameter of about 2*-6". 3 S - S8811 pipe break with an equivalent diameter of about 1/2*-2*. 2 ? - Transient event. U - Fai.ure of HPCI or RCIC to provide core make-up water. V - Failure of low pressure ECCS to provide core make-up water. t2 - Failure to remove residual core beat. C - Containment failure due to steam explosion in vessel. 8 - Containment failure due to steam explosion in containment. Y - Containr.4nt failure Bue to overpressure - release through reactor building. Y' - Containment failure due to overpressure - release direct to atsesphere. 6 - Containment isolation failure in drywell. c - Containment isolation failure in wetwell. C - Containment leakage greater than 2400 volume per cent per day. n - Reactor building isolation failure. 0 - Standby gas treatment system failure. 7-10
1 h._ Key to BWR-Accident nee Symbols (Grand Gulf plant) Ini tia ting' Events - T -- A loss of offsite power transient j 'T23 ~ Any ther transient which requires an emergency reactor shutdown S. - A small LOCA'(the break area is less than one square foot) ISL - An interfacing s, stem LOCA System or Component Failures C - Failure of the Reactor Protection System E - Failure of the Emergency Core Cooling System 'I - Failure of residual heat removal systems after i LOCA (including transient induced LOCAs) P.- Failure of a safety / relief valve to reseat Q - Failure of the Power Conversion System - U-- Failure of the High Pressure Core Spray and Reactor Core Isolation Cooling System V - Failure of the low pressure ECCS systens to provide core flow-W - Failure of the residual heat renoval systems after a transient Containment Failure Modes
- a - Containment failure due to a steam explosion y - Containment failure due to an averpressure caused by rapid hydrogen burning 6 - Containment failure due to an overpressure caused by gas generation
- Note: The symbols used for the Grand Gulf containment failure modes are somewhat different than those used in the RSS.
7-10a}}