ML16277A340

NRC Inspection Report 05000275/2016010 and 05000323/2016010; Preliminary White Finding

ML16277A340
Person / Time
Site:	Diablo Canyon
Issue date:	10/03/2016
From:	Troy Pruett NRC/RGN-IV/DRP
To:	Halpin E Pacific Gas & Electric Co
Jeremy Groom
References
EA-16-168 IR 2016010
Download: ML16277A340 (30)

See also: IR 05000275/2016010

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E. LAMAR BLVD.

ARLINGTON, TX 76011-4511

October 3, 2016

EA-16-168

Mr. Edward D. Halpin

Senior Vice President

and Chief Nuclear Officer

Pacific Gas and Electric Company

Diablo Canyon Power Plant

P.O. Box 56, Mail Code 104/6

Avila Beach, CA 93424

SUBJECT:

DIABLO CANYON POWER PLANT - NRC INSPECTION REPORT

05000275/2016010 AND 05000323/2016010; PRELIMINARY WHITE FINDING

Dear Mr. Halpin:

On September 12, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an

inspection at your Diablo Canyon Power Plant. On the same date, the NRC inspectors

discussed the results of this inspection with you and other members of your staff. Inspectors

documented the results of this inspection in the enclosed inspection report.

The enclosed inspection report discusses a finding that has preliminarily been determined to be

of low to moderate safety significance (White) that may require additional NRC inspections,

regulatory actions, and oversight. As described in Section 4OA2 of this report, the finding is

associated with an apparent violation of Technical Specification 5.4.1.a, Procedures, for the

failure to develop adequate instructions for the installation of external limit switches on motor-

operated valves. Specifically, Pacific Gas and Electric (PG&E) failed to provide adequate

maintenance instructions for ensuring that these limit switches were operated within the vendor

established overtravel settings. Consequently, the external limit switch for valve RHR-2-8700B,

Unit 2 residual heat removal pump 2-2 suction from the refueling water storage tank, was

installed such that the limit switch was operated beyond the overtravel setting resulting in a

sheared internal roll pin causing the limit switch to fail. The failure of this limit switch resulted in

failure of an input into the open permissive input logic for valve SI-2-8982B, Unit 2 train B

residual heat removal suction from the containment recirculation sump. PG&E restored valve

RHR-2-8700B to operable and replaced affected components, including the limit switch. PG&E

also initiated corrective actions to develop more detailed and appropriate instructions for

installing Namco' Snap Lock position switches.

This finding was assessed based on the best available information using the applicable

Significance Determination Process (SDP). The basis for the NRCs preliminary significance

determination is described in the enclosed report. The NRC performed a detailed risk

evaluation and determined the total resulting incremental conditional core damage

probability for internal and external initiators. Considering the failure mechanism was

E. Halpin

- 2 -

introduced during Refueling Outage 2R17 maintenance in February 2013, and the limit switch

was last successfully tested on October 22, 2014, the NRC evaluated the issue for the period

from October 22, 2014, until the limit switch failure became apparent on May 16, 2016. This

analysis resulted in a preliminary estimate of core damage frequency of 7.6E-06/year,

corresponding to a finding of low to moderate risk significance (White). The NRC will inform you

in writing when the final significance has been determined.

The finding is also an apparent violation of NRC requirements and is being considered for

escalated enforcement action in accordance with the Enforcement Policy, which can be found

on the NRCs Web site at http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html.

In accordance with NRC Inspection Manual Chapter 0609, we intend to complete our evaluation

using the best available information and issue our final determination of safety significance

within 90 days of the date of this letter. The significance determination process encourages an

open dialogue between the NRC staff and the licensee; however, the dialogue should not

impact the timeliness of the staffs final determination.

Before we make a final decision on this matter, we are providing you with an opportunity to

(1) attend a Regulatory Conference where you can present to the NRC your perspective on the

facts and assumptions the NRC used to arrive at the finding and assess its significance, or

(2) submit your position on the finding to the NRC in writing. If you request a Regulatory

Conference, it should be held within 40 days of the receipt of this letter, and we encourage you

to submit supporting documentation at least one week prior to the conference in an effort to

make the conference more efficient and effective. The focus of the Regulatory Conference is to

discuss the significance of the finding and not necessarily the root cause or corrective actions

associated with the finding. If a Regulatory Conference is held, it will be open for public

observation. If you decide to submit only a written response, such submittal should be sent to

the NRC within 40 days of your receipt of this letter. If you decline to request a Regulatory

Conference or to submit a written response, you relinquish your right to appeal the final SDP

determination, in that by not doing either, you fail to meet the appeal requirements stated in the

Prerequisite and Limitation sections of Attachment 2 of NRC Inspection Manual Chapter 0609.

Please contact Jeremy Groom at (817) 200-1148 and in writing within 10 days from the issue

date of this letter to notify the NRC of your intentions. If we have not heard from you within

10 days, we will continue with our significance determination and enforcement decision. The

final resolution of this matter will be conveyed in separate correspondence.

Because the NRC has not made a final determination in this matter, no Notice of Violation is

being issued for this inspection finding at this time. In addition, please be advised that the

number and characterization of the apparent violation described in the enclosed inspection

report may change as a result of further NRC review.

E. Halpin

- 3 -

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its

enclosure will be made available electronically for public inspection in the NRC Public Document

Room and in the NRCs Agencywide Documents Access and Management System (ADAMS),

accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html.

Sincerely,

/RA/

Troy W. Pruett, Director

Division of Reactor Projects

Docket Nos. 50-275 and 50-323

License Nos. DPR-80 and DPR-82

Enclosure:

Inspection Report 05000275/2016010 and

05000323/2016010

w/ Attachments:

1. Supplemental Information

2. Significance Determination

cc w/ enclosure: Electronic Distribution

SUNSI Review

By: JRG

ADAMS

Yes No

Non-Sensitive

Sensitive

Publicly Available

Non-Publicly Available

Keyword:

NRC-002

OFFICE

SRI:DRP/A

RI:DRP/A

SPE:DRP/A

DRS:SRA

TL:ACES

D:DRS

RC:ORA

NAME

CNewport

JReynoso

RAlexander

RDeese

MHay

AVegel

KFuller

SIGNATURE

/RA/

/RA/

/RA/

/RA/

/RA/

/RA/

/RA/

DATE

09/14/16

09/14/16

09/08/16

09/09/16

09/19/16

09/27/16

09/22/16

OFFICE

BC:DRP/A

D:DRP

NAME

JGroom

TPruett

SIGNATURE

/RA/

/RA/

DATE

09/26/16

10/3/16

Letter to Edward D. Halpin from Troy W. Pruett dated October 3, 2016

SUBJECT:

DIABLO CANYON POWER PLANT - NRC FOCUSED BASELINE INSPECTION

REPORT 05000275/2016010 AND 05000323/2016010; PRELIMINARY WHITE

FINDING

DISTRIBUTION:

Regional Administrator (Kriss.Kennedy@nrc.gov)

Deputy Regional Administrator (Scott.Morris@nrc.gov)

DRP Director (Troy.Pruett@nrc.gov)

DRP Deputy Director (Ryan.Lantz@nrc.gov)

DRS Director (Anton.Vegel@nrc.gov)

DRS Deputy Director (Jeff.Clark@nrc.gov)

Senior Resident Inspector (Christopher.Newport@nrc.gov)

Resident Inspector (John.Reynoso@nrc.gov)

Administrative Assistant (Madeleine.Arel-Davis@nrc.gov)

Branch Chief, DRP/A (Jeremy.Groom@nrc.gov)

Senior Project Engineer, DRP/A (Ryan.Alexander@nrc.gov)

Project Engineer, DRP/A (Matthew.Kirk@nrc.gov)

Project Engineer, DRP/A (Thomas.Sullivan@nrc.gov)

Public Affairs Officer (Victor.Dricks@nrc.gov)

Project Manager (Balwant.Singal@nrc.gov)

Team Leader, DRS/TSS (Thomas.Hipschman@nrc.gov)

RITS Coordinator (Marisa.Herrera@nrc.gov)

ACES (R4Enforcement.Resource@nrc.gov)

Regional Counsel (Karla.Fuller@nrc.gov)

Congressional Affairs Officer (Jenny.Weil@nrc.gov)

RIV Congressional Affairs Officer (Angel.Moreno@nrc.gov)

RIV/ETA: OEDO (Jeremy.Bowen@nrc.gov)

RIV RSLO (Bill.Maier@nrc.gov)

ROPreports.Resource@nrc.gov

ROPassessment.Resource@nrc.gov

Enclosure

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket:

05000275; 05000323

License:

DPR-80; DPR-82

Report:

05000275/2016010; 05000323/2016010

Licensee:

Pacific Gas and Electric Company

Facility:

Diablo Canyon Power Plant, Units 1 and 2

Location:

7 1/2 miles NW of Avila Beach

Avila Beach, CA

Dates:

May 16 through September 12, 2016

Inspectors: C. Newport, Senior Resident Inspector

J. Reynoso, Acting Senior Resident Inspector

T. Sullivan, Project Engineer

R. Deese, Senior Reactor Analyst

Approved

By:

Troy W. Pruett, Director

Division of Reactor Projects

- 2 -

SUMMARY

IR 05000275/2016010, 05000323/2016010; 05/16/2016 - 09/12/2016; Diablo Canyon Power

Plant; Problem Identification and Resolution

The inspection activities described in this report were performed between May 16 and

September 12, 2016, by the resident inspectors at Diablo Canyon Power Plant and inspectors

from the NRCs Region IV office. The inspectors identified a preliminary White finding

associated with an apparent violation of NRC requirements. The significance of inspection

findings is indicated by their color (Green, White, Yellow, or Red), which is determined using

Inspection Manual Chapter 0609, Significance Determination Process, issued April 29, 2015.

Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, Aspects

within the Cross-Cutting Areas, issued December 4, 2014. Violations of NRC requirements are

dispositioned in accordance with the NRC Enforcement Policy. The NRCs program for

overseeing the safe operation of commercial nuclear power reactors is described in

NUREG-1649, Reactor Oversight Process.

Cornerstone: Mitigating Systems

Preliminary White. The inspectors identified a preliminary White finding associated with

an apparent violation of Technical Specification 5.4.1.a, Procedures, for the licensees

failure to develop adequate instructions for the installation, adjustment, and testing of

Namco' Model EA170 snap lock limit switches. Specifically, the licensee failed to provide

site-specific instructions for limiting the travel of these external limit switches when installed

on safety-related motor operated valves. Consequently, the lever switch actuator for valve

RHR-2-8700B, residual heat removal pump 2-2 suction from the refueling water storage

tank, was installed such that the limit switch was operated repeatedly in an over-travel

condition resulting in a sheared internal roll pin that ultimately caused the limit switch to fail.

Following identification of this issue, the licensee replaced the limit switch for valve

RHR-2-8700B and implemented actions to modify maintenance procedures for installing,

calibrating, and testing motor-operated valve external limit switches. The licensee entered

this issue into their corrective action program as Notification 50852345.

The performance deficiency is more than minor, and therefore a finding, because it is

associated with the procedure quality attribute of the Mitigating Systems cornerstone and

adversely affected the cornerstone objective to ensure the availability, reliability, and

capability of systems that respond to initiating events to prevent undesirable consequences

(i.e., core damage). Specifically, maintenance procedure MP E-53.10R, Augmented Stem

Lubrication for Limitorque Operated Valves, used to perform limit switch adjustments on the

Unit 2 valve RHR-2-8700B, did not provide adequate acceptance criteria to prevent

overtravel of the limit switch actuating lever. This resulted in a subsequent failure of the limit

switch, preventing the open permissive signal for valve SI-2-8982B, residual heat removal

pump 2-2 suction from the containment recirculation sump, used during the emergency core

cooling system (ECCS) recirculation mode. The inspectors evaluated the finding using the

Attachment 0609.04, "Initial Characterization of Findings," worksheet to Inspection Manual

Chapter (IMC) 0609, Significance Determination Process, issued June 19, 2012. The

attachment instructs the inspectors to utilize IMC 0609, Appendix A, Significance

Determination Process (SDP) for Findings At-Power, issued June 19, 2012. In accordance

with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems

Screening Questions, the inspectors determined that the finding required a detailed risk

evaluation because it represented an actual loss of function of the train B ECCS for greater

- 3 -

than its technical specification allowed outage time. A senior reactor analyst performed a

detailed risk evaluation in accordance with IMC 0609, Appendix A, Section 6.0, Detailed

Risk Evaluation. The calculated increase in core damage frequency was dominated by

small and medium loss of coolant accident initiators with failures of the opposite train of

ECCS or related support systems. The analyst did not evaluate the large early release

frequency because this performance deficiency would not have challenged the containment.

The NRC preliminarily determined that the increase in core damage frequency for internal

and external initiators was 7.6E-06/year, a finding of low to moderate risk significance

(White). The inspector did not identify a cross-cutting aspect with this finding because it was

not reflective of current performance. The inadequate procedure was developed in 2011

and did not reflect the licensees current performance related to procedure development.

(Section 4OA2)

- 4 -

REPORT DETAILS

4.

OTHER ACTIVITIES

Cornerstone: Mitigating Systems

4OA2 Problem Identification and Resolution (71152)

Annual Follow-up of Selected Issues

a. Inspection Scope

On May 16, 2016, during performance of surveillance procedure PEP V-7B, "Test of

ECCS Valve Interlocks," Revision 9, valve SI-2-8982B, Unit 2 residual heat removal

(RHR) pump 2-2 suction from the containment recirculation sump, failed to open from

the main control room. Subsequent review determined that external limit switch,

POS-648, for valve RHR-2-8700B, RHR 2-2 suction from the refueling water storage

tank (RWST), was in a failed position. The failure of this limit switch prevented the open

permissive signal for valve SI-2-8982B. Investigation by the licensee concluded that

limit switch POS-648 failed due to a sheared internal roll pin.

The inspectors assessed the licensees problem identification threshold, cause analyses,

and verified that corrective actions were commensurate with the significance of the

issue, appropriately prioritized and that these actions were adequate to correct the

condition. The inspectors also reviewed the licensees use of operating experience and

their incorporation of vendor guidance into site-specific maintenance procedures.

These activities constituted completion of one annual follow-up sample as defined in

Inspection Procedure 71152.

b. Findings

Failure to Establish Adequate Work Instructions for Installation of Namco' Snap Lock

Limit Switches

Introduction. The inspectors identified a preliminary White finding associated with an

apparent violation of Technical Specification 5.4.1.a, Procedures, for the licensees

failure to develop adequate instructions for the installation, adjustment and testing of

Namco' Model EA170 snap lock limit switches. Specifically, the licensee failed to

provide site-specific instructions for limiting the travel of these external limit switches

when installed on safety-related motor operated valves. Consequently, the lever switch

actuator for valve RHR-2-8700B was installed such that the limit switch was operated

repeatedly in an over-travel condition resulting in a sheared internal roll pin that

ultimately caused the limit switch to fail.

Description. On May 16, 2016, the licensee performed surveillance procedure

PEP V-7B, "Test of ECCS Valve Interlocks," Revision 9, to test various interlock and

permissive circuits for the emergency core cooling system (ECCS). One interlock test

involved valve circuitry needed to transfer the RHR pump suction from the RWST to the

containment recirculation sump during the ECCS recirculation mode. During a loss of

coolant accident, operators would implement ECCS recirculation by closing the RWST to

RHR suction valves, valves RHR-8700A and RHR-8700B, and opening the containment

- 5 -

recirculation sump suction valves, SI-8982A and SI-8982B. The ECCS system design

includes an interlock, tested during procedure PEP V-7B, to ensure that operators can

only open containment sump suction valves if the respective RWST suction valve is

closed.

During performance of procedure PEP V-7B, Step 12.14.2, valve SI-2-8982B, RHR

pump 2-2 suction from the containment recirculation sump, failed to open from the main

control room. Licensee troubleshooting determined that external limit switch, POS-648,

for valve RHR-2-8700B, RHR pump 2-2 suction from the RWST, was in a failed position.

The failure of this limit switch, caused by a sheared internal roll pin, prevented the open

permissive signal for valve SI-2-8982B. Since limit switch POS-648 failed during a

planned refueling outage with Diablo Canyon Unit 2 shutdown, no technical specification

entries were necessary. The licensee replaced limit switch POS-648 under Work Order 60090383 on May 18, 2016, prior to exiting the planned refueling outage. The licensee

entered this issue into their corrective action program as Notification 50852345.

The inspectors reviewed the work history for valve RHR-2-8700B and limit switch

POS-648. During refueling outage 2R17 completed on February 21, 2013, the licensee

implemented Work Order 64014195 to replace the Limitorque actuator stem nut for valve

RHR-2-8700B and completed maintenance procedure E-53.10R, Augmented Stem

Lubrication for Limitorque Operated Valves, Revision 4. This maintenance included

removal and replacement of limit switch POS-648 and its actuating lever. The

inspectors noted that procedure MP E-53.10R included instructions for re-installing the

stem mounted position switches and checks for proper operation. Specifically,

procedure MP E-53.10R, Step 7.9.2(h), included instructions to Check switches are

properly operating by listening for an audible click from switch when valve is cycled

OPEN and CLOSED.

The inspectors noted that the licensee successfully tested POS-648 as part of

post-maintenance testing for Work Order 64014195 and again on October 22, 2014,

when procedure PEP V-7B was last performed. The licensee cycles valve

RHR-2-8700B quarterly as part of the inservice testing (IST) program; however, the

quarterly IST does not test the interlock provided by limit switch POS-648. As such, the

inspectors concluded that POS-648 failed sometime between the last successful

performance of surveillance procedure PEP V-7B on October 22, 2014, and the failure of

valve SI-2-8982B to open on May 16, 2016.

Limit switch POS-648 is a Namco' Model EA170 snap lock position switch, designed to

snap over when actuated and includes a hard stop. The inspectors reviewed applicable

maintenance, design, and testing instructions provided by the limit switch vendor. Within

the publically available vendor documents, the inspectors identified the following

precaution relative to the design, installation, and operation of Namco' Snap-Lock Limit

switches:

Operating mechanisms for limit switches MUST BE so designed

that, under any operating or emergency conditions, the limit switch

is not operated beyond its overtravel limit position.

The vendor guidance also directed switch owners to the specific bulletin for the switch

overtravel specifications. The inspectors reviewed the switch bulletin for Namco'

Model EA170-35100 snap lock limit switches, the same model used for POS-648. The

inspectors noted that the switch specifications included a recommended travel

- 6 -

of 7 degrees based on a required trip of 6.5 degrees, and a maximum overtravel of

36 degrees. The inspectors reviewed as-found photos of POS-648 following the

May 16, 2016, failure and noted that the switch actuating arm position was at a nearly

45-degree angle relative to the normal position indicating that the position switch had

exceeded the overtravel specification.

The inspectors determined that when POS-648 was re-installed following maintenance

on February 21, 2013, the licensee did not set the switch and actuating arm correctly in

accordance with the vendor recommendations to ensure that the overtravel specification

was not exceeded. By operating the switch beyond the overtravel specification, valve

force was applied to the limit switch lever and internal roll pin after reaching a hard stop.

The repeated overloading of the lever roll pin eventually led to the failure of POS-648.

While the instructions in procedure MP E-53.10R, Step 7.9.2.(h), to check for proper

operation by listening for an audible click, would verify the limit switch changed state, the

inspectors determined this procedure step was inadequate to prevent overtravel of the

externally mounted limit switch. Specifically, the inspectors determined that the

procedure lacked specificity because it only ensured that the trip and reset of the switch

occurs as the valve is exercised but did not provide adequate instructions to ensure the

switch overtravel specification was not exceeded.

The inspectors interviewed licensee personnel responsible for determining the cause of

the failure of POS-648. During that interview, the licensee shared conclusions regarding

the cause of the failure of POS-648 that corresponded with the independent conclusions

developed by the inspectors. In particular, the licensee determined that the

maintenance instructions in procedure MP E-53.10R to listen for an audible click were

insufficient to prevent over-ranging of the position switch lever. The licensee performed

an extent-of-condition review of other motor operated valve (MOV) external limit

switches that provide control or logic functions but would not provide an audible alarm or

other indication if in a failed state. The licensee identified fifteen other limit switches that

could be susceptible to the failure mechanism experienced on limit switch POS-648.

The licensee walked down these switches on June 1, 2016, and identified no other

similar switch installation problems. Notification 50852345 included corrective action

CA 1, due March 20, 2017, to revise procedure MP E-53.10R to include detailed

instructions for setting the travel of externally mounted limit switches.

Analysis. The inspectors determined the failure to establish adequate adjustment

criteria for maintenance procedure MP E-53.10R was a performance deficiency. The

performance deficiency is more than minor, and therefore a finding, because it is

associated with the procedure quality attribute of the Mitigating Systems cornerstone,

and adversely affected the cornerstone objective to ensure the availability, reliability, and

capability of systems that respond to initiating events to prevent undesirable

consequences (i.e., core damage). Specifically, procedure MP E-53.10R, used by the

licensee to perform limit switch adjustments on the Unit 2 valve RHR-2-8700B, did not

provide adequate acceptance criteria to prevent overtravel of the actuating lever. This

resulted in a subsequent failure of the limit switch, preventing the open permissive signal

for valve SI-2-8982B, residual heat removal pump 2-2 suction from the containment

recirculation sump, used during the ECCS recirculation mode. The inspectors evaluated

the finding using the Attachment 0609.04, "Initial Characterization of Findings,"

worksheet to Inspection Manual Chapter (IMC) 0609, Significance Determination

Process, issued June 19, 2012. The attachment instructs the inspectors to utilize

- 7 -

IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-

Power, issued June 19, 2012. In accordance with NRC Inspection Manual

Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the

inspectors determined that the finding required a detailed risk evaluation because it

represented an actual loss of function of the train B ECCS for greater than its technical

specification allowed outage time. A senior reactor analyst performed a detailed risk

evaluation in accordance with IMC 0609, Appendix A, Section 6.0, Detailed Risk

Evaluation.

Small and medium loss of coolant accident initiators with failures of the opposite train of

ECCS or related support systems dominated the calculated increase in core damage

frequency. The analyst did not evaluate the large early release frequency because this

performance deficiency would not have challenged the containment. The NRC

preliminarily determined that the increase in core damage frequency for internal and

external initiators was 7.6E-06/year, in the low to moderate risk significance range

(White). The results of the detailed risk evaluation are included in Attachment 2 of this

report.

The inspector did not identify a cross-cutting aspect with this finding because it was not

reflective of current performance. The inadequate procedure was developed in 2011

and did not reflect the licensees current performance related to procedure development.

Enforcement. Technical Specification 5.4.1.a, Procedures, requires, in part, that

written procedures shall be established, implemented, and maintained covering the

applicable procedures recommended in Appendix A of Regulatory Guide 1.33,

Revision 2. Section 9.a of Appendix A of Regulatory Guide 1.33, Revision 2, requires in

part, that maintenance that can affect the performance of safety-related equipment

should be properly preplanned and performed in accordance with written procedures,

documented instructions, or drawings appropriate to the circumstances. On

December 5, 2011, the licensee established procedure MP E-53.10R, Augmented Stem

Lubrication for Limitorque Operated Valves, Revision 4, to perform maintenance on

safety-related equipment including motor operated valves and their external limit

switches. Contrary to the above, on December 5, 2011, the licensee failed to establish

written procedures for performing maintenance on safety-related equipment which were

appropriate to the circumstances. Specifically, the procedure only checked that motor

operated valve external limit switches changed position during valve exercise but did not

provide instructions to establish and check the travel of these switches within vendor

established criteria. Consequently, the limit switch for valve RHR-2-8700B was installed

such that it was operated repeatedly beyond overtravel tolerances resulting in its failure.

The licensee entered this issue into their corrective action program as Notification

50852345 and initiated action to replace the failed limit switch. The licensee also

initiated corrective actions to change maintenance procedure MP E-53.10R to ensure

adequate acceptance criteria for limit switch travel were included, and performed an

extent of condition for all other MOV stem mounted position switch interlocks circuits. As

a consequence of this failed limit switch, the licensee was also in violation of Unit 2

Technical Specification 3.5.2, ECCS - Operating, because train B of the ECCS was

determined to be inoperable for greater than the technical specification allowed outage

time of 14 days, and the licensee failed to take actions required of the limiting condition

of operation. Because this finding has been preliminarily determined to be of greater

than very low safety significance (i.e., greater than Green), it is being characterized as

- 8 -

an apparent violation. AV 05000323/2016010-01, Failure to Establish Adequate Work

Instructions for Installation of Namco' Snap Lock Limit Switches

4OA6 Meetings, Including Exit

Exit Meeting Summary

On September 13, 2016, the inspectors presented the inspection results to Mr. E. Halpin, Senior

Vice President and Chief Nuclear Officer, and other members of the licensee staff. The licensee

acknowledged the issues presented. The licensee confirmed that any proprietary information

reviewed by the inspectors had been returned or destroyed.

Attachment 1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

T. Baldwin, Director, Nuclear Site Services

D. Evans, Director, Security & Emergency Services

L. Fusco, Manager, Mechanical Engineering

P. Gerfen, Station Director

M. Ginn, Manager, Emergency Planning

E. Halpin, Sr. Vice President, Chief Nuclear Officer Generation

H. Hamzehee, Manager, Regulatory Services

A. Heffner, NRC Interface, Regulatory Services

L. Hopson, Director Maintenance Services

T. Irving, Manager, Radiation Protection

K. Johnston, Director of Operations

M. McCoy, NRC Interface, Regulatory Services

J. Morris, Supervisor, Nuclear Regulatory Services

C. Murry, Director Nuclear Work Management

J. Nimick, Senior Director Nuclear Services

P. Nugent, Director, Quality Verification

A. Peck, Director, Nuclear Engineering

A. Warwick, Supervisor, Emergency Planning

J. Welsch, Site Vice President

R. West, Manager, System Engineering

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened 05000323/2016010-01 AV

Failure to Establish Adequate Work Instructions for Installation of

Namco' Snap Lock Limit Switches (Section 4OA2)

Section 4OA2: Problem Identification and Resolution

Procedures

Number

Title

Revision

PEP V-7B

Test of ECCS Valve Interlocks

8

MP E-53.10R

Augmented Stem Lubrication For Limitorque Operated

Valves

4-7

OP O-22

Emergency Operation of Motor Operated Valves

6

E-0

Reactor Trip or Safety Injection

35

EOP E-1.3

Transfer to Cold Leg Recirculation

22

A1-2

MP E-53.10A1

Low Impact External Inspections of Limitorque Motor

Operators

1

Notifications

50852066

50852180

50852345

50861001

Drawings

Number

Title

Revision

441239

Unit 2, Single Line Meter and Relay Diagram 480V System

Bus Section 2H

48

441310

Unit 2, Schematic Diagram Residual Heat Removal Motor

Operated Valves

31

441317

Unit 2, Schematic Diagram Safety Injection System Motor

Operated Valves

19

500628

Unit 2, Electrical Diagram of connections, Elevation 115-140

foot, Area H

26

507610

Unit 2, Arrangement of Electrical Equipment at Elevation

100, Area H

16

Work Orders

64014195

Miscellaneous

Number

Title

Revision

Calculation

SDP16-02

SI-2-8982B Failure to Open During PEP V-7B in 2R19 due

to Damaged Closed Position Switch for 8700B

0

Attachment 2

Significance Determination

Significance Determination Basis:

(a)

Screening Logic

Minor Question: In accordance with NRC Inspection Manual Chapter 0612,

Appendix B, Issue Screening, the finding was determined to be more than minor

because it was associated with the procedure quality attribute of the Mitigating

Systems Cornerstone, and affected the associated cornerstone objective to ensure

availability, reliability, and capability of systems that respond to initiating events to

prevent undesirable consequences. Specifically, the performance deficiency

associated with the inadequate maintenance procedure resulted in inadequate

criteria to ensure limit switch adjustments did not result in overtravel of the actuating

lever for valve RHR-2-8700B. This resulted in a subsequent failure of limit switch

POS-648, affecting the availability of the ECCS because this limit switch provides the

open permissive signal for valve SI-2-8982B, the containment sump suction for the

RHR system.

Initial Characterization: Using Manual Chapter 0609, Attachment 4, Initial

Characterization of Findings, the inspectors determined that the finding could be

evaluated using the significance determination process. In accordance with Table 3,

SDP Appendix Router, the inspectors determined that the subject finding should be

processed through Appendix A, The Significance Determination Process (SDP) for

Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated

July 1, 2012.

Issue Screening: In accordance with NRC Inspection Manual Chapter 0609,

Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors

determined that the finding required a detailed risk evaluation because it represented

an actual loss of function of the Unit 2 train B ECCS for greater than its technical

specification allowed outage time (i.e., 14 days). A senior reactor analyst performed

a detailed risk evaluation in accordance with IMC 0609, Appendix A, Section 6.0,

Detailed Risk Evaluation.

Results: The detailed risk evaluation result is an increase in core damage frequency

from the performance deficiency of 7.6E-6/year, characterizing the significance of the

finding to be of low to moderate safety significance. This estimate used best

available information and estimated the increase in core damage frequency to

be 7.1E-6/year from internal events and 5.4E-7/year from external events.

(b)

Detailed Risk Evaluation:

(1)

Assumptions

Exposure time. The exposure time was 286 days. The licensee last successfully

tested valve SI-2-8982B and the interlock associated with POS-648 on

October 22, 2014. Valve SI-2-8982B failed to open 572 days later on May 16,

2016. Since the inception of the failure of the limit switch after the last operation

could not be determined, the analyst used a t/2 approached and assumed the

exposure time to be half of 572 days, or 286 days. Repair time was not added

A2-2

because the deficiency was discovered and returned to a functional status during

an outage when the valve was not needed.

Recovery. Overall recovery was assumed to have a failure probability of 2.4E-1

for small break LOCAs and smaller medium break LOCAs (MLOCAs); 3.4E-2 for

seal LOCAs; and 1.0 for larger MLOCAs. Two methods of recovery were

available - (1) local manual valve operation, and (2) electrical bypassing of the

interlock through manual contactor operation. The derivation of these recoveries

is covered in the Internal Events section of this evaluation.

Common cause. The increased potential for common cause failure of Valve

SI-2-8982A, the same valve on the redundant train, was considered applicable.

The analyst was unaware of any programmatic licensee action to defend against

common cause failure; therefore, the analyst set the failure of valve SI-2-8982B

to TRUE in the SPAR model. This increased the probability of common cause

failure of Valve SI-2-8982A from 3.6E-5 to 3.8E-2.

The analyst also considered the remaining valves installed on Units 1 and 2 with

externally mounted limit switches that receive the same maintenance as the

valve that is the subject of the performance deficiency. For Unit 1, the analyst

determined that the issue would be of very low safety significance since there

was not an actual failure of a component.

For Unit 2, the remaining valves would not result in a significant increase in risk

because the external limit switches are either 1) only associated with an

annunciator function, 2) only associated with an equipment interlock function that

is not used in an accident scenario or, 3) only associated with an equipment

interlock function needed for long-term containment pressure control.

Operating history. The analyst assumed the plant operated at power or at

shutdown conditions above those that necessitated operation of the RHR system

for decay heat removal during the entire exposure time. This allowed the analyst

to use the at-power SPAR model for the entire exposure time.

(2)

Internal Events

Background / Introduction. The results of the probabilistic risk assessment (PRA)

tool showed that the performance deficiency affected two initiators - small break

loss of coolant accidents (SLOCA) and MLOCA. These events are characterized

by reactor coolant leaking from the reactor coolant system, which would act to

lower inventory and pressure of the reactor coolant system. In response to the

loss of coolant and system pressure, a safety injection actuation signal actuates

to start ECCS pumps. These pumps include both RHR pumps, both safety

injection pumps, and both charging pumps. These pumps take suction from the

refueling water storage tank, pump water into the reactor coolant system, which

in turn leaks out of the break and into the containment where it collects in the

containment recirculation sump. When the refueling water storage tank level

reaches 33 percent level, operators secure the RHR pumps and perform valve

manipulations to swap the suction of the emergency core cooling pumps from the

refueling water storage tank to the containment recirculation sump. Valve

SI-2-8982B is the first valve in the flowpath leading from the containment sump.

The inability to open valve SI-2-8982B renders train B of core cooling inoperable

A2-3

during the recirculation phase of LOCAs. The licensee would have options to

recover and open the valve, which are discussed in this evaluation. The licensee

would also have the redundant train A flowpath available to successfully cool the

reactor core if valve SI-2-8982B were unrecoverable. PRA demonstrates that the

dominant core damage sequences involve failures of the train A flowpath and the

inability to recover valve SI-2-8982B.

Small Break Loss of Coolant Accidents

For the purposes of this evaluation, SLOCA include pipe breaks up to 2 inches,

catastrophic reactor coolant pump seal failures (seal LOCAs), and seal LOCAs

caused by losses of cooling to the reactor coolant pump seals (brought about by

loss of power to cooling for the seals).

SLOCA comprises 26.0 percent of the increase in core damage frequency. The

results are driven by the failure of valve SI-2-8982B, failures of train A flowpath

for recirculation sump flow, and the ability or inability to operate valve SI-2-8982B

by alternative means.

The primary contributor of failures of the train A flowpath is attributed to an

increased probability of common cause failure of its sump valve SI-2-8982A.

Because valve SI-2-8982B failed and valve SI-2-8982A is subject to the same

environment, maintenance, testing, etc., valve SI-2-8982A is exposed to an

increased probability of failure. The common cause failure of SI-2-8982A

comprises approximately two-thirds of the increase in core damage frequency

from SLOCAs. The remainder of the increase in core damage frequency comes

from power failures to components in the train A flowpath, valve failures in the

flowpath, and pumps failures in the flowpath.

Recovery of valve SI-2-8982B through alternative means is also a contributor.

These alternative means include either electrical operation by use of the motor

contactors or manually by accessing the valve and operating the handwheel on

the valve.

Recovery of Valve SI-2-8982B

Recovery actions to open valve SI-2-8982B are available by two alternate

means, either electrically by use of the motor contactors, or manually by

accessing the valve and operating the handwheel on the valve. In developing

their assessment of the success probability of recovering valve SI-2-8982B, the

licensee interviewed operators who indicated that both recoveries would be

pursued in parallel.

1. Electrical operation of Valve SI-2-8982A by use of motor contactors. This

recovery option takes advantage of the ability to bypass the interlock circuitry,

which is the subject of the performance deficiency, preventing valve SI-2-8982B

from opening. Manual operation of the electrical contactors provided line power

directly to the motor operator for valve SI-2-8982B. Operation of the electrical

contactors could be successful if properly performed, but inspectors found

several impediments to absolute success.

A2-4

The first potential impediment was the adequacy of procedural guidance used for

the electrical operation recovery option. The direction to pursue recovery paths

to open valve SI-2-8982B is contained in Emergency Operating Procedure (EOP)

Emergency Contingency Action (ECA) 1.1, Loss of Emergency Coolant

Recirculation, Revision 21. Step 2 of ECA 1.1 instructs operators to restore

emergency coolant recirculation equipment by several means. Step 2.d has

operators check power available to valves required for recirculation swap over

and refer to an appendix with valve power supplies. The performance deficiency

would not result in a loss of the valves main power supply. Instead, the

performance deficiency would result in the main-line contacts being held open by

the control circuit for valve SI-2-8982B. The analyst considered this an

impediment to recovery because the procedure did not explicitly call out actions

for a loss of control power to the motor operator. The analyst concluded from the

licensees analysis that operator experience would guide them to use Step 2.d as

the best fit for troubleshooting and take the steps response not obtained action

to locally operate the valves as required. The analyst judged that local

operations are at the location of the valve, not in the electrical cabinet located

away from the valve, and that this action to locally operate the valve did not

specifically address use of the electrical contactors. Again, the analyst

determined, based on interviews and discussions with the licensee, that operator

experience and training could employ this as an option even though it is not

explicitly called for in the emergency procedure.

The licensee established procedure O-22, Emergency Operation of Motor

Operated Valves, Revision 6 to operate motor operated valves through use of

the motor contactor. Procedure O-22 requires phone communication between

the control board operator in the control room and the operator in the field at the

cabinet when operating the valve. Inspectors toured the licensees training

facilities used to instruct operators on how to locally operate contactors. The

inspectors noted that the electrical cabinet used to train operators used a

Telemecanique brand contactor, different from the Westinghouse Cutler Hammer

brand contactors installed in the cabinet for valve SI-2-8982B. The different

contactors have different operating methods. To operate the Telemecanique

contactors, operators insert non-conducting rods above and below the contactor

of interest. To operate the Westinghouse contactors, operators depress a gray

plastic armature position indicator.

The analyst concluded that the difference in layout and methods of operating

contactors between the training electrical cabinet and the plant electrical cabinet

would present challenges to successful operation of the contactor. Also, during a

walkdown with the licensee electricians, the inspectors noted that the electrical

cabinet for valve SI-2-8982B housed both the open and close contactors.

However, these contactors are not labelled such that an operator could tell which

contactor was the open contactor.

The inspectors noted that Procedure O-22, Attachment 2, provided a typical

cabinet layout for motor-operated valves in the plant. This diagram showed the

open contactor located above the close contactor. During the walkdown, the

inspectors asked electrical personnel if the orientation illustrated in

Procedure O-22, Attachment 2, was the same orientation for the cabinet for valve

SI-2-8982B. After approximately 6 minutes of inspecting the cabinet with the

A2-5

electrical schematic diagram, the three electrical personnel determined that the

orientation was opposite of that illustrated in Procedure O-22, because the close

contactor was located above the open contactor. The analyst considered these

aspects to be additional impediments to successful operation of the valve.

The inspectors noted that prior to Step 6.11, the step instructing operators to

locate the appropriate contactor, Procedure O-22 included a boxed Note that

read, Those contactors that cant be clearly identified may require assistance

from engineering or maintenance for positive identification.

The analyst concluded that Procedure O-22, Attachment 2 that provided a typical

cabinet layout for motor operated valves, created a likelihood that some

operators would consider the valve SI-2-8982B contactor orientation typical and

not heed this note. The analyst also considered that to follow the note, additional

time is required to have an engineer or electrician report to the cabinet, obtain

the proper electrical print, and trace the cabinet wiring to ascertain which

contactor was the open contactor and which contactor was the close contactor.

This additional time affects the time available to open the valve using the

electrical contractor and adversely influences the success rate of this action. The

analyst also noted that operation of the contactors would require a screwdriver to

defeat the door latch breaker trip and the operator would have to be dressed in

an arc flash suit which the operator would have to obtain prior to this action.

The consequences of operating the incorrect contactor are potentially severe. If

the licensee personnel operated the close contactor thinking they were opening

the valve, the valve motor would drive the valve in the close direction with all of

the motor-operated valve protective features bypassed. Because the valve is

already closed, the motor would be in a stall condition and motor current would

be at or near locked rotor amperage. The potential consequences of this

mis-operation could include motor damage or burnout.

The analyst included these factors in the human reliability analyses performed

using the SPAR-H method.

With the two methods performed in parallel (i.e., electrical contactors and manual

valve manipulation methods), the inspectors concluded that the electrical

contactor method would be ready for attempted use first. The assumed timing

was:

Action

Time

(minutes)

Total time

(hh:mm)

Briefing the operation

15

00:15

Gather tools, dress out in arc flash suit, report to

breaker, open cabinet

20

00:35

Recognize no labelling, summons electrician to

cabinet

10

00:45

Obtain electrical print

10

00:55

A2-6

Operate contactor (valve)

5

01:00

When added to the 10 minutes assumed to attempt swap over to recirculation

and 30 minutes assumed to troubleshoot the issue, diagnose indications, and

decide on a course of action, the analyst estimated a total time to success of

approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 40 minutes.

The analyst used these points to obtain the following human reliability analysis:

Electrical Recovery - Diagnosis (=1E-2)

Time Available

Extra

0.1

The 1:40 hour time to diagnose and

perform gives extra time when

compared to the licensees estimate

of 2:35 hour to deplete the RWST

(applying both diagnosis and

action). The time from a depleted

RWST until occurrence of core

damage was also considered.

Stress

High

2

The level of stress would be higher

than the nominal level due to

unexpected alarms being present

and consequences that could

threaten plant safety.

Complexity

Nominal

1

No event information is available to

warrant a change in this diagnosis

performance shaping factor (PSF)

from Nominal.

Experience/Training Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

Procedures

Incomplete

20

Task instructions are absent to

guide the operator to the

appropriate electrical contactor

operation

Ergonomics

Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

Fitness For Duty

Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

Work Processes

Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

Result = 4E-2 = 0.1 x 2 x 1 x 1 x 20 x 1 x 1 x 1 x 1E-2

A2-7

Electrical Recovery - Action (=1E-3)

Time Available

Extra

0.1

The 1:40 hour time to diagnose and

perform gives extra time when

compared to the licensees estimate

of 2:35 hour to deplete the RWST

(applying both diagnosis and

action). The time from a depleted

RWST until occurrence of core

damage was also considered.

Stress

High

2

The level of stress would be higher

than the nominal level due to

unexpected alarms being present

and consequences that could

threaten plant safety.

Complexity

Highly

5

The evolution involved equipment

line-up that involved defeated

interlocks on valves, a highly

complex task.

Experience/Training Low

3

Different contactors were present in

the cabinet than were trained on

during operator training.

Procedures

Incomplete

20

The procedure provided operators

with a generic orientation of the

contactors which did not match the

in-plant configuration. The note for

operators to seek assistance is not

explicit, stating that the situation ..

may require assistance..

Ergonomics

Poor

10

The contactors in the panel are not

labelled causing poor human-

machine interface.

Fitness For Duty

Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

Work Processes

Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

PSF = 0.1 x 2 x 5 x 3 x 20 x 10 x 1 x 1 = 600

Result = 3.8E-1 = 1E-3 x 600 / [1E-3 x (600 - 1)] + 1

Combining diagnosis and action (4.0E-2 + 3.8E-1) yielded a failure probability of

4.2E-1.

2. Manual operation of Valve SI-2-8982A by handwheel. This recovery action

involves operators utilizing the handwheel to open valve SI-2-8982B. The

analyst considered the diagnosis to employ this option to be similar to the

decision for electrical contactor operation, except Procedure ECA 1.1 was

appropriate in directing local manual valve operations. Also the analyst

concluded the assumption of 10 minutes to attempt swap over to recirculation

and 30 minutes to troubleshoot the issue, diagnose indications, and decide on a

A2-8

course of action that was appropriate for diagnosis of this action. The inspectors

considered that the local manual valve operation path would present operators

with the decision to incur more dose, face uncertain environmental and

radiological factors at the valve, the potential to introduce a containment bypass

flowpath, and the manual handwheel option requires more time than the

electrical contactor option. In their analysis, the licensee considered this local

manual valve operation as the sole credited recovery option. However, for the

previously stated reasons, the analyst concluded this option would be employed

after the electrical contactor option.

The inspectors noted several attributes of this action made it more complex. The

valve is located adjacent to the containment in a special chamber. The chamber

has an enclosed environment that may become radioactively contaminated

following a LOCA. The licensee would need to implement actions to sample the

environment for suitable breathing to prevent a radioactive intake. Alternatively,

an operator would have to don protective clothing to prevent contaminating

himself, don a respirator, climb a ladder to enter the chamber, and operate the

valve. Any leakage from this valve (e.g., packing leakage) could serve to

pressurize this chamber and require additional protective clothing to prevent

contamination. To access the valve inside of the chamber, the licensee needs to

remove 32 nuts, which act to secure the chamber. This additional time affects

the time available to open the valve and adversely influences the success rate of

this action.

The licensee estimated 90 minutes would be required to brief personnel, gather

tools, and open the manway. Next the licensee estimated 10 minutes to open

the valve. The analyst noted that according to licensee information, the valve

would take 468 turns of the handwheel to open the valve. Factoring in fatigue

from repetitive motion along with potentially cumbersome clothing in a hot

environment, 25 minutes (or one turn approximately every 3 seconds) would be

required. This makes the timeline as follows for execution:

Action

Time

(minutes)

Total time

(hh:mm)

Briefing the operation, gather tools, and open

manway

90

01:30

Operate valve

25

01:55

When added to the 10 minutes assumed to attempt swap over to recirculation

and 30 minutes assumed to troubleshoot the issue, diagnose indications, and

decide on a course of action, the total time to success was estimated to be

approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 35 minutes (2.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />).

A2-9

The analyst used these points to obtain the following human reliability analysis:

Mechanical Recovery - Diagnosis (=1E-2)

Time Available

Nominal

1

The 2:35 hour time to diagnose and

perform gives nominal time when

compared to the licensees estimate

of 2:35 hour to deplete the RWST

(applying both diagnosis and action).

Combined with the time from a

depleted RWST until occurrence of

core damage.

Stress

High

2

The level of stress would be higher

than the nominal level due to

unexpected alarms being present

and consequences that could

threaten plant safety.

Complexity

Moderate

2

Several variables are involved in

diagnoses including the knowledge

of introducing a potential

containment bypass path.

Experience/Training Nominal

1

Adequate amount of instruction to

perform.

Procedures

Nominal

1

Evaluated not to be a performance

driver.

Ergonomics

Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

Fitness For Duty

Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

Work Processes

Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

Result = 4.0E-2 = 1 x 2 x 2 x 1 x 1 x 1 x 1 x 1

Mechanical Recovery - Action (=1E-3)

Time Available

Nominal

1

The 2:35 hour time to diagnose and

perform gives nominal time when

compared to the licensees estimate

of 2:35 hour to deplete the RWST

(applying both diagnosis and

action). Combine with the time from

a depleted RWST until occurrence

of core damage.

Stress

High

2

The level of stress would be higher

than the nominal level due to

unexpected alarms being present

and consequences that could

threaten plant safety.

A2-10

Mechanical Recovery - Action (=1E-3)

Complexity

Nominal

1

Little ambiguity existed in what

needs to be performed

Experience/Training Low

3

The licensee was unable to provide

prior examples where the valve was

operated manually by operators.

Operators are not trained on manual

valve operations inside the

chamber.

Procedures

Incomplete

20

References for task instructions for

opening the chamber are absent.

Operators would have to refer to an

outage procedure for guidance on

opening the chamber.

Ergonomics

Poor

10

Poor human-machine interface is

present. Access to the valve

chamber requires a ladder. In

chamber, the operator would be

manipulating the valve, possibly in a

respirator and wearing protective

clothing. Operation of the valve

would be in a hot environment, with

awkward and tight clearances

relative to the chamber walls.

Fitness For Duty

Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

Work Processes

Nominal

1

No event information is available to

warrant a change in this diagnosis

PSF from Nominal.

PSF = 1 x 2 x 1 x 3 x 20 x 10 x 1 x 1 = 1200

Result = 5.4E-1 = 1E-3 x 1200 / [1E-3 x (1200 - 1)] + 1

Combining diagnosis and action (4.0E-2 + 5.4E-1) yielded a failure probability of

5.8E-1.

Net effect. The analyst assumed the licensee would always have and attempt the

electrical contactor option first. The SPAR-H analysis yielded a result that

58 percent (failure rate = 4.2E-1) of the time the licensee would successfully

open the valve via the electrical contactor method. The analyst then assumed

that failure to select the correct contactor to operate the valve would result in

damage to the valves electric motor, requiring the licensee to utilize the

mechanical recovery option with the failure rate derived by SPAR-H (5.8E-1) for

manual valve operations. This yielded an effective failure rate of 2.4E-1,

calculated as follows:

peff = pe x pm

peff = the effective human performance failure rate for both recoveries

pe = the failure rate by electrical contactor operation

pm = the failure rate by local manual valve operation

A2-11

Catastrophic Seal LOCA. The results of this group is similar to the SLOCA

group. The analyst combined the template events ZT-RCS-MDP-LK-BP1,

Reactor Coolant Pump Seal Stage 1 Integrity Fails (Binding/Popping Open),

and ZT-RCS-MDP-LK-BP2, Reactor Coolant Pump Seal Stage 2 Integrity Fails

(Binding/Popping Open), in the SPAR model to develop an initiating event

frequency for a catastrophic seal failure event of 2.5E-3/year. The analyst

obtained this failure probability from WCAP-15603, Westinghouse Owners

Group 2000 Reactor Coolant Pump Seal Leakage for Westinghouse Pressurized

Water Reactors. This value matches the initiating event frequency used by the

licensee in their model within 2 percent. The analyst then applied the conditional

core damage probability from a SLOCA to this initiating event frequency to

estimate the change in core damage frequency resulting from a catastrophic seal

failure with the performance deficiency present. The analyst considered that the

low leakage rate from a failed reactor coolant pump seal would provide extra time

for recovery via the electrical contactor and via the mechanical operation paths.

This changed the effective recovery from this initiator to 3.4E-2.

Induced Seal LOCA. These reactor coolant leaks result from a loss of cooling to

the reactor coolant pump seals. The dominant initiating events in SPAR which

lead to induced seal failure are grid related losses of offsite power (LOOP),

switchyard centered LOOPs, and transients. These events represent the

smallest contribution to increase in core damage frequency. The analyst

assumed a recovery of 3.4E-2, similar to the recovery of a catastrophic seal

LOCA.

Medium Break Loss of Coolant Accidents

In NRC probabilistic risk assessment analyses, MLOCAs are breaks from 2 to

6 inches in size. MLOCAs may or may not increase pressure high enough to

actuate the containment spray actuation signal, which occurs when pressure in

the containment building reaches approximately 22 psig. This actuation signal

would start the two containment spray pumps that combine to pump around

5000 gallons per minute from the RWST to the containment. This additional

draw of water from the RWST would lower the available time for operators to

take action to open valve SI-2-8982B by the alternative means and therefore

adversely influence the success rate of these actions. The analyst reviewed

Diablo Canyon PRA Calculation MAAP13-03, Diablo Canyon Power Plant

MAAP Success Criteria - Loss of Coolant Accident Definitions, Revision 0, to

determine at which break size would actuate the containment spray actuation

signal and start the containment spray pumps. In this calculation, a 2.9-inch

break produced an 18 pound per square inch pressure in the containment. The

analyst estimated that breaks above 3.5 inches would produce pressure in the

containment sufficient to start the containment spray pumps.

From this estimate, the analyst broke MLOCAs into two classes. The first class

consisted of breaks between 2 and 3.5 inches in size, not sufficient to start the

containment spray pumps. Based on this 1.5-inch range, the analyst estimated

simplistically that 37.5 percent of the MLOCAs would not cause starting of the

containment spray pumps. Conversely, 62.5 percent of MLOCAs were assumed

to start containment spray pumps. Once started, the analyst assumed that

A2-12

operators would leave the containment spray pumps running as required by the

emergency operating procedures.

The analyst split the initiating event frequency by this 37.5 - 62.5 percent split

and applied different recovery actions based on the differing times available. For

the 37.5 percent of MLOCAs which would not start the containment spray pumps,

recovery was similar to SLOCAs.

For the 62.5 percent that would actuate containment spray pumps, the analyst

assumed that the RWST would deplete quickly and not allow sufficient time for

recovery. Licensee estimates were that operators would only have around

30 minutes between RWST level of 33 percent and 4 percent. The 33 percent

level is the point where operators would be required to attempt to swap from

injection from the RWST to the containment recirculation sump. The 4 percent

level is the level at which procedures instruct operators to secure all emergency

core cooling pumps, thereby terminating any injection. That difference of

29 percent (33 - 4) would be depleted by the containment spray pumps in

approximately 30 minutes. Actions to operate the motor contactors or locally

manually operate the valve were far in excess of this timing, so the analyst

considered that recovery was not possible.

Summary of Internal Events

The table below summarizes the dominant initiators and their contribution to the

increase in core damage frequency. The overall results were an increase in core

damage frequency of 8.2E-6/year from internal events:

Contributor

Increase in Core Damage

Frequency

SLOCA

2.0E-6

Catastrophic Seal LOCA

1.4E-7

Induced Seal LOCA

1.1E-9

Smaller MLOCA

3.8E-8

Larger MLOCA

4.8E-6

Total

7.1E-6

(3)

External Events

The analyst estimated the increase in core damage frequency from all external

events to be 5.4E-7/year, using the individual estimates below.

Seismic. The analyst performed a seismic analysis using Revision 8.23 of the

SPAR model. This analysis used a baseline conditional core damage probability

representing a non-recoverable, switchyard-centered LOOP. The fragilities from

Table AA-2 of Volume 2, External Events, of the Risk Assessment of Operational

A2-13

Events Handbook were used. The increase in core damage frequency from

seismic events was estimated to be 3.2E-7/year.

High winds. The analyst assumed no risk from high winds due to the historically

low tornadic activity at Diablo Canyon.

Fire. The analyst used information from the licensees fire probabilistic risk

assessment model as the best available information to estimate the increase in

core damage frequency from fires. The licensee received their safety evaluation

for approval of application of NFPA 805 and is in transition to full compliance.

The analyst applied the licensees risk achievement worth value of 1.0452 to the

baseline core damage frequency of 1.70E-5/year to estimate the increase in core

damage frequency from fires to be 6.02E-7/year. Due to the low contribution

relative to the internal events estimation of increase in core damage frequency,

the analyst applied a generic recovery failure probability of 2.4E-1 derived from

the SPAR-H for SLOCAs and applied it to all fires. This resulted in an increase in

core damage frequency from fires of 2.2E-7/year.

(4)

Large Early Release Frequency

The analyst reviewed the dominant sequences and compared them to Manual

Chapter 0609, Appendix H, Containment Integrity Significance Determination

Process. The analyst performed a LERF screening to assess whether any of

the core damage sequences affected by the finding were potential LERF

contributors. The analyst determined that none of the sequences were

significant LERF contributors and the increase in LERF was considered to be

negligible.

(5)

Uncertainties

Analytical

The analyst reviewed the analysis uncertainty for the base case with no recovery

credit for the limited use model with basic event HPI MOV CC 8982B set to

TRUE. The analyst then extrapolated the results to estimate that approximately

75 percent of results from a Monte Carlo distribution resulted in an increase in

core damage frequency between 1.0E-6/year and 1.0E-5/year or less.

Qualitative Considerations

Competing priorities. The detailed risk evaluation only considered the recovery

activities for the failed valve SI-2-8982B. For the core damage sequences of

interest, other plant equipment would malfunction and attempts would be made

to recover them. For example, in a case where the pump on the opposite train of

the recirculation path was not working, operators would be challenged with

additional diagnosis of that problem as well as deciding which recirculation path

was more easily recoverable. This additional diagnosis would divert plant

resources from recovery of valve SI-2-8982B. These competing priorities for

recovery add uncertainty to the detailed risk evaluation performed and would

serve to make recovery more unlikely.

A2-14

Anecdotal information from a simulated recovery attempt. When the inspectors

walked through operation of the valve SI-2-8982B by use of the electrical

contactors with one engineer and two electricians, these individuals initially

indicated that they would operate the contactors as represented in

Procedure O-22. This operation would act to further close the valve, potentially

causing irreparable damage. When the inspectors pointed this out, the

individuals traced the wiring with the electrical drawing and corrected their

response on the proper contactor they would operate. This was done by

electrical personnel in a training environment. The uncertainty in how electrical

personnel, if summoned to assist, would respond was only considered as

success in the analyses. This information for recovery adds uncertainty to the

detailed risk evaluation performed and would serve to make recovery more

unlikely.

Temperature of the Recirculation Valve Chamber. The temperature of the

recirculation valve chamber at the time operators would be required to enter and

manipulate valve 8982B is unknown. If the temperature exceeded 130 degrees

Fahrenheit, local manual valve operation could likely be impossible. This lack of

information adds uncertainty and would serve to make recovery more unlikely.

(6)

Sensitivities

The analyst performed sensitivities runs showing the results for various scenarios

altering the influential assumptions:

Different assumptions of recovery of the valve: The analyst adjusted the

failure probability for various cases and compared them to the assumed

failure probability in the table below:

Failure Probability

of Recovery

Comment

Increase in Internal

Events CDF

1.1E-2

98.9% success in recovery

5.0E-6/year

4.0E-2

96% success in recovery

5.3E-6/year

1.0E-1

90% success in recovery

6.0E-6/year

2.4E-1

76% success in recovery

(assumed in analysis)

7.1E-6/year

5.0E-1

50% success in recovery

9.2E-6/year

No recovery

0% success in recovery

2.0E-5/year

The potential for common cause failure of Train A Valve 8982A is not affected

by the failure of Valve 8982B: The analyst estimated the increase of

removing the cutsets which contained the common cause failure of

valve 8982A. Result: Increase in CDF of 2.7E-6/year

A2-15

Consideration that valves 8982A and 8982B were tested in a staggered

scheme: The analyst assumed the valves were tested nine months apart

vice testing both during refueling outages. Result: Increase in CDF of

4.9E-6/year

Use of the licensees MLOCA frequency value combined with SPAR-H

nominal recovery: The analyst used the licensees lower initiating event

frequency value of 2.3E-5/year along with the SPAR-H nominal recovery

value of 1.1E-2. Result: Increase in CDF of 1.6E-6/year

(7)

Licensee Results

The licensee provided the analyst with their analysis. The estimated increase in

core damage frequency was 2.9E-5/year without recovery applied. This value

did not adjust for common cause failure of the train A valve (valve SI-2-8982A).

The analyst estimated that the SPAR model, when adjusted for catastrophic seal

LOCAs and removal of consideration of elevated common cause failure of the

train A valve, would estimate the increase in core damage frequency of

3.3E-5/year.

The licensee derived a failure probability for recovery with the local manual valve

operation of 1.2E-2. When the licensee applied this recovery to their model, they

estimated the increase in core damage frequency to be 7.5E-7/year. The analyst

considered that the value of 1.2E-2 for recovery was conservative in light of the

numerous adjustments needed to the performance shaping factors for less than

nominal conditions affecting the recoveries. SPAR-H uses a nominal failure

probability of 1.1E-2, which is near the licensees recovery value. The analyst

considered the application of SPAR-H to provide more realistic estimations of

failure probabilities.

(8)

Model Adjustments

Limited Use Model Version DCAN-RICK-2187 of the Diablo Canyon SPAR

Model, was used with SAPHIRE Version 8.1.4. This version incorporated

modifications to the model derived from the lessons learned from NUREG-2187,

Confirmatory Thermal-Hydraulic Analysis to Support Success Criteria in the

Standardized Plant Analysis Risk Models - Byron Unit 1, Revision 0. The

analyst used the default truncation of 1.0E-11.