ML16165A280

From kanterella
Revision as of 06:35, 30 June 2018 by StriderTol (talk | contribs) (Created page by program invented by StriderTol)
Jump to navigation Jump to search
ASP Analysis - Reject - Diablo Canyon Both Trains of Residual Heat Removal Inoperable Due to Circumferential Crack on a Socket Weld (LER 275/2015-001)
ML16165A280
Person / Time
Site: Diablo Canyon, 05000375  Pacific Gas & Electric icon.png
Issue date: 06/09/2016
From: David Aird, Pfefferkorn C M, Tetter K M
NRC/RES/DRA/PRB
To:
Pfefferkorn C M
References
Download: ML16165A280 (7)
v • d • e


Text

Final ASP Program Analysis - Reject Accident Sequence Precursor Program - Office of Nuclear Regulatory Research Diablo Canyon Power Plant, Unit 1 Both Trains of Residual Heat Removal Inoperable Due to Circumferential Crack on a Socket Weld Event Date: 12/31/2014 LERs: 275-2015-001 IR: 50-275/2015-004 CCDP= 2x10-8 Plant Type: Westinghouse 4-Loop Pressurized Water Reactor with Large Dry Containment Plant Operating Mode (Reactor Power Level):

Mode 3 (0 Percent Reactor Power, Operating Temperature and Pressure)

Analyst:

Candace Pfefferkorn Reviewer:

Keith Tetter Contributors:

David Aird BC Approved Date:

6/9/2016 EVENT DETAILS Event Description.

On December 31, 2014, both trains of residual heat removal (RHR) at Diablo Canyon Power Plant, Unit 1, were declared inoperable. The declaration was made after plant personnel performing a walk-down identified boric acid accumulation/through-wall seepage (30 drops per minute) originating from a circumferential crack on the socket weld connection from the RHR pump common discharge header to relief valve (RV) RHR-1-RV-8708.

Relief valve RHR-1-RV-8708 is inside containment and protects RHR discharge piping to reactor coolant system (RCS) hot legs 1 and 2 from exceeding its design pressure rating. The condition was entered at 11:05 am and was exited at 10:56 pm when the plant entered Mode 4.

Immediate corrective actions included a repair of the cracked socket weld and installation of a pipe support on RV-8708. In addition, per technical specifications, the associated containment penetration flow path was isolated.

Cause. The root cause of the condition was determined to be containment fan cooler unit (CFCU) vibration that induced a resonant condition in the RHR piping generating stresses above the material endurance limit of the socket weld. Subsequent corrective actions included the replacement of the previously repaired socket weld and one additional socket weld on the RV discharge pipe, the relocation of the previously installed pipe supports, and correction of the condition that caused the CFCU vibrations. Notably, a similar condition (circumferential crack on the socket weld RHR connection RV RHR-1-RV-8708), was documented in LER 275-2013-005 for Diablo Canyon Nuclear Power Plant, Unit 1. In that case, the condition was attributed to low-stress, high-cycle fatigue caused by system vibration.

MODELING Basis for ASP Analysis/SDP Results. The ASP Program uses Significance Determination Process (SDP) results for degraded conditions when available and applicable. The ASP Program performs independent analyses for initiati ng events. ASP analyses of initiating events account for all failures/degraded conditions and unavailabilities (e.g., equipment out for test/maintenance) that occurred during the event, regardless of licensee performance.

1

1 ASP analyses also account for any degraded condition(s) that were identified after the initiating event occurred if the failure/degradation exposure period(s) overlapped the initiating event date.

LER 275-2015-001 2 In Inspection Report 50-275/15-04 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16035A481), the inspecto rs documented their review of the events described in License Event Report (LER) 275-2015-001 and 275-2013-005. The inspectors documented a green self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI "Corrective Action," for the licensee's failure to identify the cause and take corrective action to prevent recurrence of a significant condition adverse to quality impacting both trains of the Unit 1 safety-related residual heat removal (RHR) system. Specifically, the licensee failed to identify a definitive cause and implement corrective actions to prevent recurrent failures of the socket weld for RV RHR-1-RV-8708 for both trains of the RHR system.

While the issue associated with LER 275-2015-001 was addressed through the SDP in Inspection Report 50-275/15-04, as of the writing of this report, the LER has not been documented as closed in an inspection report.

An independent ASP analysis was performed for this condition assessment because this event has not yet been directly closed to date.

Analysis Type. The Diablo Canyon Unit 1 & 2 Standardized Plant Analysis Risk (SPAR) model, Revision 8.23, created May 2014, was used for this event analysis. The event was modeled as a condition assessment.

SPAR Model Modifications. Fault tree modifications and basic event additions were made within the BFN Unit 3 SPAR model as described below:

  • A new house event, titled LPI-RV-8708 (Relief Valve 8708 Breaks) was created to model a break resulting from the circumferential crack on the socket weld connection to RV RHR-1-RV-8708. The normal state of this valve is closed and therefore LPI-RV-8708 was set to FALSE.

Low Pressure Injection) fault tree was modified to include the basic house event LPI-RV-8708 (Relief Valve 8708 Breaks) under the LPI (

Low Pressure Injection

) "OR" gate.

  • The LPR (Low Pressure Recirculation) fault tree was modified to include the basic house event LPI-RV-8708 (Relief Valve 8708 Breaks) under the LPR-SYS-F (Failure of LPI Systems During Recirculation) "OR" gate.

LER 275-2015-001 3 Key Modeling Assumptions. The following modeling assumptions were determined to be significant to the modeling of this event:

  • While the RHR system was declared i noperable in accordance with Technical Specifications, the Diablo Canyon SPAR Report states that:

Success [of RHR] implies the RCS pressure and temperature are within the requirements to allow the RCS hot leg (to LPI pump suction) suction valves to be opened and provide a suction source to the LPI pumps. The dedicated LPI pump train heat exchangers will slowly cool down the reactor. This system requires an operator action to open the RCS hot leg valves which provide the suction source for the pumps. The success criterion is one-of-two LPI pumps providing sufficient flow through their respective heat exchangers Therefore, for this analysis, nominal probabilities for RHR basic events were used which include LPI hot leg suction valves, pumps, and heat exchangers, as well as the operator action to open RCS hot leg valves. Furthermore, most initiating event sequences that reference the RHR top event (fault tree) s ubsequently reference either low or high pressure recirculation (LPR or HPR faul t trees, respectively) implicating RHR in providing heat exchange and suction for LPI, LPR and HPR.

  • To ensure conservative bounding risk results, it is assumed that the subject circumferential crack on the socket weld connection to RV RHR-1-RV-8708 resulted in a full pipe rupture. The pipe rupture is assumed to be of sufficient size such that all RHR water is diverted from both cold leg LPI (or LPR) trains through the break until it is isolated manually by operators.
  • The time duration of the condition is assumed to be one hour. In an emergency core cooling system mode valve alignment, two motor operated valves are open. These values would have to be closed from the control room during an incident to isolate a break in the hot leg injection line. Based on review of abnormal operating procedures for malfunction of RHR, an upper bound estimate of one hour was ascribed to this action.
  • The subject pipe rupture, even if isolated, would result in a loss of low pressure hot leg injection during recirculation. Hot leg recirculation typically occurs approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after low pressure cold leg recirculation is initiated. Therefore, failure of hot leg injection during recirculation was not modeled in this analysis.
  • At the time of the event Diablo Canyon Power Plant, Unit 1, was in Mode 3 (Hot Standby) with no components reported as being out of service for test and maintenance. However, this analysis assumes a full break in the hot leg injection line resulting from the circumferential crack on the socket weld connection to RV RHR-1-RV-8708. Thus, basic

event LPI-RV-8708 was set to TRUE for this event analysis

  • All other safety systems responded as designed.

LER 275-2015-001 4 ANALYSIS RESULTS CCDP/Rejection Basis.

The CDP for this analysis is 1.8 x10

-8 which is below the ASP Program CDP threshold of 1x10

-6. Therefore, this event is not a precursor and is screened out of the ASP Program.

Dominant Sequence. The dominant accident sequence is Medium Loss of Coolant Accident (MLOCA) Sequence 02 (CDP = 1.7x10

-8) that contributes approximately 98% of the total internal events CCDP. Figure 1 in Appendix B illustrates this sequence. The cut sets/sequences that contribute to the top 95% and/or at least 1% of the total internal events CCDP are provided in Appendix A.

The events and important component/system failures in MLOCA Sequence 02 are:

  • Offsite electrical power succeeds,
  • High pressure injection succeeds
  • Low pressure recirculation fails.

REFERENCES

1. Diablo Canyon Power Plant, "LER 275-2015-001 Both Trains of Residual Heat Removal Inoperable Due to Circumferential Crack on a Socket Weld," dated February 11,

2016 (ML16042A470).

2. Diablo Canyon Power Plant, "LER 275-2015-001 Both Trains of Residual Heat Removal Inoperable Due to Circumferential Crack on a Socket Weld," dated March 2, 2015

(ML15061A548).

3. Diablo Canyon Power Plant, "LER 275-2013-005 Both Trains of Residual Heat Removal Inoperable Due to Circumferential Crack on a Socket Weld," dated August 22,

2013 (ML13235A101).

4. U.S. Nuclear Regulatory Commission, "Diablo Canyon Plant - NRC Integrated Inspection Report 05000275/2015004," dated February 4, 2016 (ML16035A481).
5. Nuclear Regulatory Commission, "Diablo Canyon Plant - NRC Integrated Inspection Report 05000275/2014002," dated April 23, 2014 (ML14113A527).

LER 275-2015-001 A-1 Appendix A: SAPHIRE 8 Worksheet Summary of Conditional Event Changes Event Description Cond Value Nominal Value LPI-RV-8708 Relief Valve 8708 Breaks True False Event Tree Dominant Results Only items contributing at least 1.0% to the total CCDP are displayed.

EVENT TREE CCDP CDP CDP DESCRIPTION MLOCA 1.71E-8 3.85E-11 1.71E-8 Diablo Canyon 1 & 2 Medium Loss of Coolant Accident LLOCA 2.85E-10 6.18E-13 2.85E-10 Diablo Canyon 1 & 2 Large Loss of Coolant Accident Total 1.90E-8 1.48E-9 1.75E-8 Dominant Sequence Results Only items contributing at least 1.0% to the total CCDP are displayed.

EVENT TREE SEQUENCE CCDP CDP CDP DESCRIPTION MLOCA 02 1.71E-8 3.75E-11 1.71E-8 /RPS, /OEP, /HPI, /AFW, /SSC1, LPR LLOCA 3 2.85E-10 2.59E-14 2.85E-10 /OEP, /ACC, LPI Total 1.90E-8 1.48E-9 1.75E-8 Referenced Fault Trees Fault Tree Description LPI LOW PRESSURE INJECTION LPR LOW PRESSURE RECIRCULATION Cut Set Report - MLOCA 02 Only items contributing at least 1% to the total are displayed.

  1. PROB/FREQ TOTAL% CUT SET Total 1.50E-4 100 Displaying 1 Cut Sets. (1 Original) 1 1.50E-4 100.00 IE-MLOCA,<TRUE>

Cut Set Report - LLOCA 3 Only items contributing at least 1% to the total are displayed.

  1. PROB/FREQ TOTAL% CUT SET 1.65E-7 100 Displaying 176 Cut Sets. (176 Original) 1 1.02E-7 61.80 IE-LOOPSC,ACP-CRB-CF-OPSD3 2 4.75E-8 28.84 IE-LOOPSC,RSW-STR-CF-ALL 3 2.60E-9 1.58 IE-LOOPSC,ACP-XHE-XM-EPSXT,EPS-XHE-XL-NR04H,OEP-XHE-XL-NR04HSC,OPR-XHE-XR-CASLT Referenced Events Event Description Probability IE-LLOCA LARGE LOCA 2.50E-6 IE-MLOCA MEDIUM LOCA 1.50E-4 LER 275-2015-001 B-1 Appendix B: Key Event Tree Figure 1: Diablo Canyon MLOCA Event Tree (Sequence 02 Bolded)

LER 275-2015-001 C-1 Appendix C: SPAR Model Modifications Figure 2: Change to LPI Fault Tree (Partial)

Figure 3: Change to LPR Fault Tree (Partial)

LPILOW PRESSURE INJECTIONLPI-TRNB-FFAILURE OF LPI TRAIN B TO COLD LEGS 3 & 4 ExtNYON 1 & 2 EAT REMOVAL AIN BLPI-TRNB-F1FAILURE OF COLD LEGS 3 & 4LPI-CLINJ-CL3-FNO FLOW TO COLD LEG 3 DURING LPILPI-RCS-CL3-F1FAILURE OF INJECTION TO RCS COLD LEG 3FalseHE-LOCAHOUSE EVENT: LOCA2.50E-01LOCA-CL3LOCA OCCURED IN COLD LEG 34.26E-08RCS-CKV-CF-8948ABCDCCF OF RCS COLD LEG DISCHARGE CHECK VALVES1.07E-05RCS-CKV-CC-8948CCOLD LEG INJECTION CKV 8948C FAILS TO OPEN1.07E-05LPI-CKV-CC-8818CCOLD LEG 3 INJECTION CHECK VALVE FROM LPI/RHR TRAINS FAIL4.26E-08LPI-CKV-CF-ALLCCF OF LPI/RHR COLD LEG INJECTION CHECK VAVLESLPI-CLINJ-CL4-FNO FLOW TO COLD LEG 4 DURING LPILPI-RCS-CL4-F1FAILURE OF INJECTION TO RCS COLD LEG 4FalseHE-LOCAHOUSE EVENT: LOCA2.50E-01LOCA-CL4LOCA OCCURED IN COLD LEG 44.26E-08RCS-CKV-CF-8948ABCDCCF OF RCS COLD LEG DISCHARGE CHECK VALVES1.07E-05RCS-CKV-CC-8948DCOLD LEG INJECTION CKV 8948D FAILS TO OPEN1.07E-05LPI-CKV-CC-8818DCOLD LEG 4 INJECTION CHECK VALVE FROM LPI/RHR TRAINS FAIL4.26E-08LPI-CKV-CF-ALLCCF OF LPI/RHR COLD LEG INJECTION CHECK VAVLES3.13E-06LPI-AOV-OC-HCV637RHR MDP 1-2 DISCHARGE CONTROL AOV 637 FAILS TO REMAIN OPEN8.13E-07LPI-MOV-OC-8809BLPI/RHR DISCHARGE MOV 8809B FAILS TO REMAIN OPEN9.63E-04LPI-MOV-CC-FCV641BFAILURE OF RHR TRAIN B MOV FCV-641B1.85E-05LPI-MOV-CF-FCV641ABCCF OF RHR TRAIN A/B MOVs FCV-641A/B4.37E-07HPI-TNK-FC-RWSTRWST FAILS TO PROVIDE WATER2.02E-06HPI-XVM-OC-XVM11RWST MANUAL VALVE MV 1-1 FAILS8.13E-07LPI-MOV-OC-8980LPI DISCHARGE MOV 8980 FROM RWST FAILS TO REMAIN OPEN1.07E-05LPI-CKV-CC-8981LPI DISCHARGE CHECK VALVE FROM RWST FAILS TO OPENFalseLPI-RV-8708Relief Valve 8708 BreaksLPRLOW PRESSURE RECIRCULATIONLPR-SYS-FFAILURE OF LPI SYSTEMS DURING RECIRCULATIONLPR-CLINJ-FFAILURE OF COLD LEG INJECTION DURING LPRLPR-CLINJA-FFAILURE OF FLOW FROM RHR/LPI TRAIN A DURING LPRLPR-CLINJ12-FFAILURE OF FLOW FROM RHR TRN A TO COLD LEGS 1 & 24.26E-08RCS-CKV-CF-8948ABCDCCF OF RCS COLD LEG DISCHARGE CHECK VALVES1.07E-05RCS-CKV-CC-8948ACOLD LEG INJECTION CKV 8948A FAILS TO OPEN1.07E-05LPI-CKV-CC-8818ACOLD LEG 1 INJECTION CHECK VALVE FROM LPI/RHR TRAINS FAIL4.26E-08LPI-CKV-CF-ALLCCF OF LPI/RHR COLD LEG INJECTION CHECK VAVLESLPR-CLINJ-CL2-FNO FLOW TO COLD LEG 2 DURING LPILPR-RCS-CL2-F1FAILURE OF INJECTION TO RCS COLD LEG 2FalseHE-LOCAHOUSE EVENT: LOCA2.50E-01LOCA-CL2LOCA OCCURED IN COLD LEG 24.26E-08RCS-CKV-CF-8948ABCDCCF OF RCS COLD LEG DISCHARGE CHECK VALVES1.07E-05RCS-CKV-CC-8948BCOLD LEG INJECTION CKV 8948B FAILS TO OPEN1.07E-05LPI-CKV-CC-8818BCOLD LEG 2 INJECTION CHECK VALVE FROM LPI/RHR TRAINS FAIL4.26E-08LPI-CKV-CF-ALLCCF OF LPI/RHR COLD LEG INJECTION CHECK VAVLES8.13E-07LPI-MOV-OC-8809ALPI/RHR DISCHARGE MOV 8809A FAILS TO REMAIN OPEN9.63E-04HPI-MOV-CC-8982ASUMP SUCTION MOV 8982A FAILS TO OPEN1.85E-05HPI-MOV-CF-8982ABCCF OF CONTAINMENT SUMP ISOL MOVs 8982A/B3.13E-06LPI-AOV-OC-HCV638RHR MDP 1-1 DISCHARGE CONTROL AOV 638 FAILS TO REMAIN OPEN9.63E-04LPI-MOV-OO-8700ARWST ISOLATION MOV 8700A FAILS TO CLOSE7.78E-06LPI-MOV-CF-8700ABCCF OF LPI/RHR RWST ISOLATION MOVs 8700A/B ExtLPR-CLINJBFAILURE OF FLOW FROM RHR/LPI TRAIN B DURING LPRFalseLPI-RV-8708Relief Valve 8708 Breaks2.00E-03LPI-XHE-XM-RECIRCOPERATOR FAILS TO INITIATE LPR OPERATION MODE

Retrieved from "https://kanterella.com/w/index.php?title=ML16165A280&oldid=764358"