ML20149K478
| ML20149K478 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/28/2020 |
| From: | Christopher Hunter NRC/RES/DRA/PRB |
| To: | |
| Littlejohn J (301) 415-0428 | |
| References | |
| LER 1989-005-00 | |
| Download: ML20149K478 (7) | |
Text
B-245 ACCIDENT SEQUENCE PRECURSOR PROGRAM EVENT ANALYSIS LER No: 338/89-005 Event
Description:
Reactor trip due to MFW regulator valve closure, steam generator tube leak, and RHR suction valve failure Date of Event: February 25, 1989 Plant: North Anna 1 Summary A reactor trip occurred due to a steam-flow-to-feedwater-flow mismatch caused by a main feedwater valve failing closed. During the recovery from the trip, a 60- to 70-gpm steam generator (SG) tube leak was detected that was the result of a failed hot leg tube plug.
Following cooldown for the leaking SG tube, the residual heat removal (RHR) system could not be placed in service because the suction isolation valve failed to remain open.
The conditional probability of core damage associated with this event is estimated to be 1.9 x 10-4. The relative significance of this event compared with other potential transients at North Anna is shown below.
LER 338/89-0051 1E-7 1E-6 1E-5 1E-4 I E-3 IE-2 II I I . I I Tripj Trip1
~ ~LOFWi
+1AWL L.precursor cutoff
~~36Oh AFW IL~oE L30hE Event Description At 1407 h on February 25, 1989, Unit 1 automatically tripped from 76% power. The initiating signal for the reactor trip was "C" SG steam-flow-greater-than-feedwater-flow mismatch coincident with a low SG level. The steam-flow-greater-than-feedwater-flow mismatch was caused by the closure of the "C" main feedwater regulating valve on the loss of control air. The valve lost control air because of the fatigue failure of the instrument air supply line around the fitting on the valve positioner.
B-246 Following the trip, indications of primary to secondary leakage were detected (later determined to be 74 gpm). SG "C" was identified as the source of the leak. Emergency boration, isolation of SG "C" AFW tubine-driven pump steam lines, and RCS cooldown and depressurization were initiated.
During the recovery from the tube leak, the residual heat removal system failed because the RHR suction isolation valve (1-RH-MOV-1701) failed to remain open. Once full open indication was received, the valve immediately stroked closed. The RHR system could not be placed in service for approximately 3 h and 10 min. The cause for the RHR suction isolation valve failing to remain open during cooldown was due to the failure of the high-pressure autoclosure relay. This failure generated a close signal when the valve reached the full open position.
ASP Modeling Assumptions and Approach This event was modeled as a steam generator tube leak with failure of the residual heat removal system. The probability of failing to locally recover RHR was assumed to be 0.12. This value is smaller than is usually assumed for local recovery in the ASP Program; however, the time period for recovery is long. The probability of a large tube rupture (300-500 gpm), given the observed 60- to 70-gpm leak, was assumed to be 0. 1.
Potential core damage sequences associated with a tube rupture were modeled using the following event tree.
In these sequences, auxiliary feedwater (AFW) or main feedwater (MEW) is assumed required for initial core cooling, and high-pressure injection (HPD) is required for makeup of inventory lost through the break. Unavailability of HPI or AFW and MEW is assumed to result in core damage (sequences 103, 106, and 107). Because of RCS flow out of the break, bleed and feed cooling is not considered a viable core cooling method.
If secondary-side cooling and BPI are successful, then isolation of the affected SG and reduction of RCS pressure below the SG relief valve setpoint are assumed to mitigate the event. In this case, flow out of the break is stopped, and secondary-side cooling using the remaining SGs provides for core cooling. Failure to reduce RCS pressure below the SG relief setpoint results in continued loss of RCS inventory. Since this inventory is not collected in the containment sump, no capability exists for high-pressure recirculation once RWST is depleted, and core damage results (sequences 102 and 105). In the event of successful reduction of RCS pressure below the SG relief valve setpoint, flow from the break is terminated. In this case, if the ruptured SG cannot be isolated, the event tree
B-247 recognizes the possibility that core cooling can still be maintained provided that the plant is placed in cold shutdown prior to depletion of RWST inventory. Failure to isolate the impacted SIG and place the reactor in cold shutdown prior to RWST depletion is assumed to lead to core damage (sequences 10 1 and 104). Failure to trip (sequence 108) results in an ATWS sequence and is not developed further.
SGTR RT AFW MFW HPI f RCS Depressurized IRuptured SG RC nCl Shutdown End Seq.
[ Below SG RV I solated SetpointDelto Pirto RWST State No.
OK OK CD 101 CD 102 CID 103 OK OK 00 104 CD 105 CD 106 00 107 ATWNS108 The following conditional branch probabilities were used in the analysis:
Branch System N-onrecover Operator Failure STGR 1.0 RT 2.8 x 10-4. 0.12*
AFW 3.8 x104 0.26*
MEFW 0.2* 0.34*
BPI 3.0 x 10-4* 0.84*
Ruptured SG Isolated 1.0) x 10-2 1.0 RCS Depressurized Below SG RV Setpoint 1.0 x 10-2 4 1.0 4.0 x 10-RCS in Cold Shutdown Prior to RWST Depletion 1.0 x 10-2 4 1.0 4.0 x 10-
- Values are consistent with probability values used in other ASP calculations.
B-248 Analysis Results The conditional fýrobability of severe core damage estimated for this event is 1.9 x 10O4.
The dominant sequence for this event (highlighted on the following event tree), involves successful AFW, HPI, and RCS depressurization following the tube rupture, with subsequent failure to isolate the break and place the RCS in cold shutdown prior to RWST depletion.
B-249 ROS RCS InCold End Seq.
STGR RTAW MWI Depressurized Ruptured Below SG RV IsolatedSG Shutdown Prior to RWST state No.
Setpoint Depletion OK OK CD 101 CID 102 OD 103 OK OK CD 104 CD 105 CD 106 CD 107 ATWS 108 Dominant core damage sequence for LER 338/89-005
B-250 CONDITIONAL CORE DAMAGE PROBABILITY CALCULATIONS Event Identifier: 338/89-005 Event
Description:
Reactor trip, SO tube leak and failed RHiN system Event Date: 02/25/89 Plant: PMR SGTR INITIATING EVENT NON-RECOVERABLE INITIATING EVENT PROBABILITIES SGTR 1. OE-0l SEQUENCE CONDITIONAL PROBABILITY SUMS End State/Initiator Probability CD SGTR 1. 9E-04 Total 1 . 9E-04 AIMS SGTR 3.4E-06 Total 3.4E-06 SEQUENCE CONDITIONAL PROBABILITIES (PROBABILITY ORDER)
Sequence End State Prob N Rec**
101 SGTR -rt. -afw -hpi -rcs.depr<sg.rv.setpoint ruptured.sg.isol R CD 1. 2E-04 1 .2E-02 CS.COLD.PRIOR.TO.RMST.DEPL 102 SGTR -rt -afw -hpi rcs.depr<sg.rv.setpoint CO 4.1lE-05 1. OE-Ol 103 SGTR -rt -afw hpi CO 2.5SE-O5 8.4E-02 108 SGTR rt ATMS 3.4E-06 1 .2E-02
-* non-recovery credit for edited case SEQUENCE CONDITIONAL PROBABILITIES (SEQUENCE ORDER)
Sequence End State Prob N Nectt 101 SGTR -rt -afw -hpi -rcs.depr<sq.rv.setpoint ruptured.sg.isol R CD 1.2E-04 1. 2E-02 CS. COLD. PRIOR.TO. RMST .DEPL 102 SOIR -rt -afw -hpi rcs.depr<sg.rv.setpoint CD 4. lE-OS S.OE-01 103 SOIR -rt -afw hpi CD 2.5SE-05 8 .4E-02 108 SGTR rt AIMS 3.4E-06 1 .2E-02
- non-recovery credit for edited case SEQUENCE MODEL: c:\asp\1989\PMRSGTR.CMP BRANCN MODEL: c:\asp\1989\PMRSGTR.NEW PROBABILITY FILE: c:\asp\1989\PMR BSLl.PRO No Recovery Limit BRANCH FREQUENCIES/PROBABILITIES Branch System Non-Recc 'v Opr Fail SGTR 5.OE-03 > 5.OE-03 l.OE+00 > l.OE-01 Branch Model: INITOR Initiator Freq: S.O0E-03 rt 2 . BE-04 1.2E-01 a fw 3.8E-04 2. 6E-01 mfw 2. OE-01 3.4E-01 hpi 3. OE-04 8.4E-01 Event Identifier: 338/89-005
B-251 ruptured. sq.isol 1.05-02 1. 05+00 rcs .depr<sq. rv. setpoint 1.05-05 1. 05+00 4 .0E-04 ECS .COLD.PRIOR.TO.RWST.DEPL 1.05-02 > 1.05+00 1.05+00 > 1.2E-01 4 .0E-04 Branch Model: l.OF.l+opr Train 1 Cond Prob: 1.05-02 > Failed
- branch model file
- forced Minarick 06-14-1990 08: 53: 24 Event Identifier: 338/89-005