ML120100057

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Swp Presentation for Meeting January 11, 2012 at 9am
ML120100057
Person / Time
Site: Palisades Entergy icon.png
Issue date: 01/11/2012
From:
Entergy Nuclear Operations
To:
NRC/RGN-III/DRMA
References
Download: ML120100057 (42)
v • d • e


Text

ENTERGY NUCLEAR Entergy Nuclear Operations

ENTERGY NUCLEAR 2

Tony Vitale Site Vice President Entergy - Palisades

ENTERGY NUCLEAR 3

  • Introduction Tony Vitale
  • Objectives Tony Vitale
  • Apparent Violation Tony Vitale
  • Root Cause Alan Blind
  • Key Learnings Alan Blind
  • Key Corrective Actions Alan Blind
  • Common Cause Alan Blind
  • Significance Brian Brogan
  • Conclusions Tony Vitale

ENTERGY NUCLEAR 4

  • Discuss apparent violations and Entergys acceptance of the finding
  • Review the causal analysis of the event
  • Review corrective actions
  • Provide additional insights into the safety significance of the finding
  • Address common cause failure probability

ENTERGY NUCLEAR 5

Complex Technical Issue

  • Thorough investigation and cause analysis
  • Robust corrective actions
  • Different material properties between pumps
  • P-7A and P-7B had lower susceptibility
  • Material toughness not recognized as critical

ENTERGY NUCLEAR 6

  • Entergy concurs with the apparent violations.
  • Entergy has additional information for consideration regarding the safety significance of the finding.

ENTERGY NUCLEAR 7

Palisades Performance Recovery Plan

  • Plan Area

- Major divisions of culture or processes in need of improvement

- Five major plan areas including

  • condition problem statement
  • future condition vision statement

- Sixth plan area is Communications

  • Plan Elements
  • Plan Actions
  • Plan Metrics

ENTERGY NUCLEAR 8

Plan Areas Leadership Effectiveness Safety Culture Corrective Action Program Equipment Reliability Refueling Outages Communication Plan

ENTERGY NUCLEAR 9

Alan Blind Engineering Director Entergy - Palisades

ENTERGY NUCLEAR 10 Scope Elements Pump Operation Maintenance Procedures and Practices Organizational and Programmatic Factors Metallurgical Analysis

ENTERGY NUCLEAR 11

  • Root Causes:
  • In 2006, the ASTM specification selected lacked specificity to ensure all critical material testing requirements for use in the service water operating environment (RC2)

(O&P)

  • The 2009 and 2011 line shaft coupling failures were due to IGSCC (RC1)

(Metallurgical)

ENTERGY NUCLEAR 12

  • Contributing Causes:
  • Increased susceptibility to IGSCC caused by tempering embrittlement (CC1) (Metallurgical)
  • Insufficient use of qualified metallurgical expertise (CC2) (Criterion XVI) (O&P)
  • Ineffective use of operating experience (CC3)

(O&P)

ENTERGY NUCLEAR 13

1. 2009 - Installing ASTM conforming material; other possible factors were not investigated (O&P)
2. Use of all available resources including operating experience and third party reviews (O&P)

ENTERGY NUCLEAR 14 Service Water Pump Teams 2006/2007 Modification 2009 RCE 2010 Operating Experience Review 2011 RCE /

Modification (Criterion XVI)

(Criterion III)

Failure Analysis:

Lucius Pitkin, Inc X

Pumps:

Mancini Consulting Services X

Organizational Factors:

Seastate Group X

Technical Review:

Structural Integrity X

X X

Palisades Engineering X

X X

X Palisades Maintenance X

X Palisades Training X

Entergy Fleet Challenge X

X OEM:

Hydro Aire X

Supplier RCE X

ENTERGY NUCLEAR 15

  • New coupling material installed on all SWPs (RC1, CC1)

(Metallurgical)

- ASTM A564 type 630 SS Condition H1150 commonly referred to as 17-4PH (RC1, CC1)

- Mechanical testing requirements include:

  • hardness
  • toughness For additional assurance, an effectiveness review will be completed on the P-7C couplings by removal and inspection.

ENTERGY NUCLEAR 16 To address the 2006 modification issue:

2007 implemented EN-HU-104, Technical Task and Rigor, requires an Independent Technical Review for complex, high risk modifications (CC2) (Criterion III) 2007 implemented EN-DC-115, Engineering Change Development, requires a review of operating experience (CC3)

(Criterion III) 2012 implemented EN-MS-S-037-L, Requirements and Expectations for Material Change Design Changes, identifies requirements and expectations for material changes affecting installed plant equipment (RC1, RC2, CC1, CC2, CC3) (Criterion III)

ENTERGY NUCLEAR 17 Common Cause Susceptibility Analysis

ENTERGY NUCLEAR 18 Common Cause Susceptibility Analysis

  • SW Pumps (P-7A, B, C)

- Pumps are two stage vertical shaft with 350 HP motor, each rated 8000 gpm at 140 ft of TDH.

- Nos. 1-4 continuously submerged in lake water

- Nos. 5-7 experience wet/

dry cycles

- No. 8, near motor, is dry

ENTERGY NUCLEAR 19 Common Cause Susceptibility Analysis

ENTERGY NUCLEAR 20 Common Cause Susceptibility Analysis 2011 service water pump (SWP) P-7C failed coupling: cracks originated at thread roots and propagated to the outer diameter Slanted fracture of remaining ligament is evidence of an overload event Fracture Surface of 2011 Failed Coupling No. 6 in P-7C

ENTERGY NUCLEAR 21 Common Cause Susceptibility Analysis

  • Failed couplings were all located in wet/dry region
  • P-7A - No indications found
  • P7B - 40 days minimum from as removed condition to failure

ENTERGY NUCLEAR 22 Common Cause Susceptibility Analysis

  • 2006 - ASTM A582 Type 416 SS Design Conformance. Hardness range for intermediate temper 24 to 32 HRC

- 2006 - 2009 First P-7C Failure

  • Failed coupling hardness ranged from 34.8 to 37.1 HRC

- 2009 - 2011 Second P-7CFailure

  • Failed coupling hardness ranged from 24.0 to 33.6 HRC

ENTERGY NUCLEAR 23 Common Cause Susceptibility Analysis

  • Additional 2011 Testing:
  • Tensile Testing
  • The yield strength and elongation were found to be in the expected range for the specified intermediate temper condition
  • Charpy V-Notch Impact Energy
  • Low absorbed energy found: indicates low fracture toughness; which correlates with increased SCC susceptibility

ENTERGY NUCLEAR 24

  • SCC is a failure process that requires each of the following:
1. Susceptible Material
2. Corrosive Environment
3. Tensile Stress

ENTERGY NUCLEAR 25 Common Cause Susceptibility Analysis

1. Susceptible Material
  • P-7C most susceptible
  • Type 416 SS can be more or less susceptible to SCC depending on heat treatment
  • Couplings installed on P-7C SWP in 2009, were tempered in the range of 1025°F -

1090°F to achieve the specified hardness (28-32 HRC).

  • Tempering in critical range made the P-7C couplings less tough and more susceptible to SCC ASM Metals Handbook, 8th Ed., Vol. 2

ENTERGY NUCLEAR 26 Common Cause Susceptibility Analysis

2. Corrosive environment
  • Palisades intake water is chlorinated
  • Chlorine was present on fracture surfaces

ENTERGY NUCLEAR 27

3. Tensile Stress

ENTERGY NUCLEAR 28 Common Cause Susceptibility Analysis Stress Corrosion Cracking Common Cause Evaluation Summary SWP 7A SWP 7B SWP 7C Susceptibility:

More Single Temper X

Double Temper X

X Critical Temper Range X

Corrosive Environment X

X X

Tensile Stress X

X X

Wet / Dry Environment X

X X

2009 Nominal Run Time To Failure 2,414 hr 2011 Nominal Run Time To Failure 14,155 hr Nominal Run Time To Replacement 16,259 hrs 9,073 hrs

ENTERGY NUCLEAR 29 Technical Conclusion

  • No common cause failure

-P-7A - highest inservice time, least susceptible to SCC

-P-7B - capable of meeting 30-day mission time

-P-7C failure - P-7A and P-7B continued to provide two operable service water pumps

ENTERGY NUCLEAR 30 Brian Brogan Sr. Staff Engineer Entergy - Palisades

ENTERGY NUCLEAR 31

  • Present the key input/assumption differences that affect the safety significance determination
  • Show differences between:
  • Entergy Calculation No: EA-PSA-SDP-P7C-11-06, SDP Assessment of P-7C Coupling Failures

ENTERGY NUCLEAR 32

  • Independent Failure Rate
  • Common Cause Failure Rate

Frequency

ENTERGY NUCLEAR 33

  • The service water pump failure-to-run basic event (BE) probability was updated
  • Entergy concurs with the value derived for this probability in the inspection report

ENTERGY NUCLEAR 34

  • Entergy common cause analysis includes:
  • Independent engineering analysis performed
  • Conservative statistical analysis of failure probability based on projected failure date from metallurgical analysis
  • Independent metallurgical analysis performed
  • No indications of cracking in the P-7A couplings

ENTERGY NUCLEAR 35

  • Specific initiating event model to evaluate the increase in the LOSW-IE due to pump failures
  • Conservative treatment of the common cause term in the initiating event model
  • Method consistent with ASME/ANS PRA Standards for Capability Category II

ENTERGY NUCLEAR 36

  • Failure of the two normally running pumps and failure or unavailability of the standby pump
  • Failure of the two normally running pumps during the time frame when the first pump is out of service (OOS) for repairs
  • The standby pump can fail to start or fail to continue running while both of the normally operating pumps are OOS for repairs LOSWIE= CCFR(S+ FR CCF +QMSP)+ 2IFR(FR IF)(S+ FR IF +QMSP)

F(LOSW-IE)/yr = 8766LOSWIEA

ENTERGY NUCLEAR 37 LOSWIE= CCFR(S+ FR CCF +QMSP)+ 2IFR(FR IF)(S+ FR IF +QMSP)

CCFR = FRFR Failure rate for common cause failures of the two normally running pumps S = Failure rate for failure of the standby pump to start on demand FR =

Common cause beta factor for failure to run of two normally operating pumps.

This factor is conservative as it accounts for all failure modes (not just the failure mode introduced by increasing the potential of IGSCC). The quantitative failure probability analysis supports the conservatism in this value.

FR = Failure rate for failure of the standby or operating pump to run IFR=(1-FR) FR Failure rate for independent failure to run for each normally running pump CCF = Mean time to repair of at least one pump after a common cause failure to run IF = Mean time to repair of a normally operating pump after an independent failure to run QMSP = Maintenance unavailability of a Standby pump while plant in operation

ENTERGY NUCLEAR 38 Time Period Inspection Report Increase in LOSW-IE Entergy Increase in LOSW-IE P-7C In Service 3.23 1.3 P-7C Out of Service 1590 30

ENTERGY NUCLEAR 39 Model Inspection Report CDF/yr Entergy CDF/yr Full Power Internal Events 4.7E-6 4.3E-7 Flooding (screened out) 1.0E-8 Fire 3.0E-7 7.0E-9 Seismic 3.5E-7 (not significant)

Total 5.4E-6 (White) 4.5E-7 (Green)

ENTERGY NUCLEAR 40

  • The P-7C failures were determined to be repeated independent failures of a single component
  • LOSW-IE is dominant impact on the results
  • Impact of this condition on service water as a mitigating system yields results consistent with very low risk

ENTERGY NUCLEAR 41 Tony Vitale Site Vice President Entergy - Palisades

ENTERGY NUCLEAR 42

  • Entergy concurs with the violations
  • A thorough review of the event was performed to identify all related causes
  • Entergy has developed effective corrective actions and root and contributing causes
  • Entergy has performed a rigorous analysis consistent with ASME/ANS PRA standards
  • Entergys determination of the safety significance is very low
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