ML16363A429

Final Significance Determination of a White Finding, Notice of Violation, and Follow-up Assessment Letter; NRC Inspection Report 05000275/2016010 and 05000323/2016010

ML16363A429
Person / Time
Site:	Diablo Canyon
Issue date:	12/28/2016
From:	Kennedy K Region 4 Administrator
To:	Halpin E Pacific Gas & Electric Co
Jeremy Groom
References
EA-16-168 IR 2016010
Download: ML16363A429 (15)

See also: IR 05000275/2016010

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E. LAMAR BLVD.

ARLINGTON, TX 76011-4511

December 28, 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 - FINAL SIGNIFICANCE

DETERMINATION OF A WHITE FINDING, NOTICE OF VIOLATION,

AND FOLLOW-UP ASSESSMENT LETTER; NRC INSPECTION

REPORT 05000275/2016010 AND 05000323/2016010

Dear Mr. Halpin:

This letter provides you the final significance determination of the preliminary White finding

identified in the Diablo Canyon Power Plant - NRC Inspection Report 05000275/2016010

and 05000323/2016010; Preliminary White Finding (Agencywide Documents Access and

Management System (ADAMS) Accession No. ML16277A340), dated October 3, 2016. The

finding is associated with the May 16, 2016, failure of the Unit 2 residual heat removal pump 2-2

suction valve (SI-2-8982B) from the containment recirculation sump to open from the main

control room. The NRC has determined the finding is of low-to-moderate safety significance

(White).

At your request, the NRC held a regulatory conference on November 15, 2016, to further

discuss your views on this finding. The meeting summary of this regulatory conference is

available at ADAMS Accession No. ML16336A765 and a copy of your presentation is available

at ADAMS Accession No. ML16335A439. In your presentation, you described several changes

to the probabilistic risk modeling of the failure of valve SI-2-8982B, including changes to the

common cause alpha factors and several assumptions related to medium break loss-of-coolant

accidents. Your staff also provided their perspectives on a variety of recovery methods

available to open valve SI-2-8982B, thereby, restoring the flow path from the containment sump

to the reactor core through residual heat removal pump 2-2.

Based on your staffs evaluation of these factors and the probability of success of these

recovery actions, your staff concluded that the change in core damage frequency was less than

the Green/White threshold of 1E-6 per year. As a result, you concluded that the inspection

finding should be characterized as very low safety significance (Green).

E. Halpin -2-

We have concluded that our preliminary significance determination change in core damage

frequency result of 7.6E-6 per year represents the upper range of the increase in core damage

frequency associated with the performance deficiency. Based on the information provided by

your staff at the regulatory conference, the NRC adjusted a number of assumptions used in the

preliminary significance determination. Specifically, the NRC lowered the common cause alpha

factors and adjusted several assumptions related to medium break loss-of-coolant

accidents. The NRC also performed a variety of human error probability calculations to

determine the likelihood of recovering the functionality of valve SI-2-8982B. The results of these

calculations, which removed much of the conservativism from the assumptions used in the

preliminary risk assessment, predicted a high likelihood of success (96.4 percent success) for

recovering valve SI-2-8982B.

Using these assumptions, the NRC concluded the lower range of increase in core damage

frequency associated with the performance deficiency to be 1.3E-6 per year. Because the

NRCs calculated lower and upper estimations of the increase in core damage frequency of the

performance deficiency were both greater than 1.0E-6 per year but less than 1.0E-5 per year,

the NRC determined the finding was of low-to-moderate safety significance (White). Our

evaluation of the risk significance of the finding is provided in the attachment to this letter.

You have 30 calendar days from the date of this letter to appeal the staffs determination of

significance for the identified White finding. Such appeals will be considered to have merit only

if they meet the criteria given in NRC Inspection Manual Chapter 0609, Attachment 2. An

appeal must be sent in writing to the Regional Administrator, Region IV, 1600 E. Lamar Blvd.,

Arlington, TX 76011.

The NRC has also determined that the failure to develop adequate instructions for the

installation of external limit switches on motor-operated valves is a violation of Technical Specification 5.4.1.a, Procedures, as cited in the enclosed Notice of Violation (NOV). In

accordance with the NRC's Enforcement Policy, the NOV is considered an escalated

enforcement action because it is associated with a White finding for Unit 2.

You are required to respond to this letter and should follow the instructions specified in the

enclosed NOV when preparing your response. If you have additional information that you

believe the NRC should consider, you may provide it in your response to the NOV. The NRC's

review of your response to the NOV will also determine whether further enforcement actions are

necessary to ensure compliance with regulatory requirements.

As a result of our review of Diablo Canyon Power Plants performance, including this White

finding, we have assessed the performance of Diablo Canyon Power Plant, Unit 2, to be in the

Regulatory Response column of the NRCs Action Matrix, effective the third quarter of 2016.

Therefore, we plan to conduct a supplemental inspection using Inspection Procedure 95001,

Supplemental Inspection Response to Action Matrix Column 2 Inputs, when your staff has

notified us of your readiness for this inspection. This inspection procedure is conducted to

provide assurance that the root cause and contributing causes of risk significant performance

issues are understood, the extent of condition and the extent of cause are identified, and the

corrective actions are sufficient to prevent recurrence.

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 ADAMS, accessible from the NRC Web site at

http://www.nrc.gov/reading-rm/adams.html.

E. Halpin -3-

To the extent possible, your response should not include any personal privacy, proprietary, or

safeguards information so that it can be made available to the Public without redaction.

Sincerely,

/RA/

Kriss M. Kennedy

Regional Administrator

Docket Nos. 50-275 and 50-323

License Nos. DPR-80 and DPR-82

Enclosure:

Notice of Violation w/Attachment:

Final Significance Determination

cc: Electronic Distribution via Listserv

for Diablo Canyon Power Plant, Units 1 and 2

ML16363A429

SUNSI Review ADAMS Non-Sensitive Publicly Available Keyword:

By: JRG Yes No Sensitive Non-Publicly Available

OFFICE RI:DRP/A SRI:DRP/A BC:DRP/A DRS:SRA TL:ACES RC:ORA OE

NAME JReynoso CNewport JGroom RDeese MHay KFuller GFigueroa

SIGNATURE /RA/Email /RA/Email /RA/ /RA/ /RA/ /RA/ /RA/ E

DATE 12/13/16 12/13/16 12/13/16 12/15/16 12/15/16 12/13/16 12/19/16

OFFICE D:DRS D:DRP RA

NAME AVegel TPruett KKennedy

SIGNATURE /RA/ /RA/ /RA/

DATE 12/13/16 12/13/16 12/28/16

NOTICE OF VIOLATION

Pacific Gas and Electric Company Docket No. 50-323

Diablo Canyon Power Plant License No. DPR-82

EA-16-168

During an NRC inspection conducted between May 16 and September 12, 2016, a violation of

NRC requirements was identified. In accordance with the NRCs Enforcement Policy, the

violation is listed below:

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.

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, Procedure MP E-53.10R, Augmented

Stem Lubrication for Limitorque Operated Valves, Revision 4, used to perform

maintenance on safety-related equipment, failed to provide instructions to establish and

check the travel of external switches installed on motor-operated valves are 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 on May 16, 2016. As a consequence of this inadequate maintenance

procedure issue, the licensee also violated Unit 2 Technical Specification 3.5.2, ECCS -

Operating, because train B of the emergency core cooling system was determined to be

inoperable for greater than the technical specification allowed outage time of 14 days.

This violation is associated with a White significance determination process finding.

Pursuant to the provisions of 10 CFR 2.201, Pacific Gas and Electric Company is hereby

required to submit a written statement or explanation to the U.S. Nuclear Regulatory

Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with a copy to the

Regional Administrator, Region IV; and a copy to the NRC resident inspector at the facility that

is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of

Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation;

EA-16-168" and should include for each violation: (1) the reason for the violation, or, if

contested, the basis for disputing the violation or severity level; (2) the corrective steps that

have been taken and the results achieved; (3) the corrective steps that will be taken; and (4) the

date when full compliance will be achieved.

Your response may reference or include previous docketed correspondence, if the

correspondence adequately addresses the required response. If an adequate reply is not

received within the time specified in this Notice, an order or a Demand for Information may be

issued as to why the license should not be modified, suspended, or revoked, or why such other

action as may be proper should not be taken. Where good cause is shown, consideration will

be given to extending the response time.

Enclosure

If you contest this enforcement action, you should also provide a copy of your response, with

the basis for your denial, to the Director, Office of Enforcement, United States Nuclear

Regulatory Commission, Washington, DC 20555-0001.

Because your response will be made available electronically for public inspection in the NRCs

Public Document Room or from the NRCs document system (ADAMS), accessible from the

NRC website at http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should not

include any personal privacy, proprietary, or safeguards information so that it can be made

available to the public without redaction.

If personal privacy or proprietary information is necessary to provide an acceptable response,

then please provide a bracketed copy of your response that identifies the information that

should be protected and a redacted copy of your response that deletes such information. If you

request withholding of such material, you must specifically identify the portions of your response

that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g.,

explain why the disclosure of information will create an unwarranted invasion of personal

privacy or provide the information required by 10 CFR 2.390(b) to support a request for

withholding confidential commercial or financial information). If safeguards information is

necessary to provide an acceptable response, please provide the level of protection described

in 10 CFR 73.21.

Dated this 28th day of December 2016

-2-

Diablo Canyon Power Plant, Unit 2, Failure of Valve SI-2-8982B to Open

Final Significance Determination

During the regulatory conference held on November 15, 2016, your staff described their

assessment of the significance of the finding. Specifically, your staff discussed differences

between the NRC's preliminary significance determination and the Diablo Canyon Power

Plant risk assessment. Based on the information provided at the regulatory conference, the

NRC determined that the preliminary significance determination represented the upper

range of the increase in core damage probability associated with the performance

deficiency.

The NRC utilized the information provided by the licensee at the regulatory conference to

estimate the lower range of the increase in core damage frequency associated with the

performance deficiency. The following elements of the risk evaluation were evaluated and

are discussed below.

1. Your staff provided updated motor-operated valve common cause alpha factors.

Based on this updated information, the NRC, when evaluating the lower range in the

increase in risk, reduced the value of the alpha-2 common cause factor in the SPAR

model from 1.92E-2 to 1.77E-2. The alpha-1 factor was also adjusted accordingly.

2. Your staff presented an updated containment analysis that demonstrated medium break

loss-of-coolant accidents (MLOCAs), with pipe breaks less than 4.5 inches in diameter,

would not result in sufficient containment pressure to start the containment spray pumps.

The NRC reviewed the updated containment analysis and agreed that for MLOCAs less

than 4.5 inches in diameter, the containment spray pumps likely would not start.

Consequently, the analyst adjusted the non-recoverable MLOCAs from 3.5 to 4.5 inches

when evaluating the lower range of the increase in core damage frequency for the final

significance determination. The analyst performed a sensitivity analysis for smaller and

larger non-recoverable MLOCAs ranging from 4.0 to 5.0 inches and concluded these

changes had a minor effect on the outcome of the detailed risk evaluation.

This application of less frequent, non-recoverable, loss-of-coolant accidents (LOCAs) led

the NRC to ensure all non-recoverable LOCAs were accounted for in their detailed risk

evaluation. Large break LOCAs (LOCAs with greater than 6-inch pipe breaks) were

truncated in the preliminary evaluation because they comprised less than 2 percent of

the estimate of total increase in core damage frequency. With the information provided

by the licensee, large break LOCAs came to comprise a larger and significant portion of

the estimate of total increase in core damage frequency and their contribution was

accounted for in the final detailed risk evaluation. The estimate of the increase in core

damage frequency from large break LOCAs was 1.4E-7 per year. The NRC discussed

the increased contribution from large break LOCAs at the regulatory conference.

3. Your staff presented a different methodology for determining the frequency of MLOCA

for differing break sizes by using NUREG-1829, Estimating Loss-of-Coolant Accident

(LOCA) Frequencies through the Elicitation Process, April 2008.

When evaluating the lower range of the increase in core damage frequency for the final

significance determination, the NRC updated the method of deconstructing MLOCAs

using a logarithmic-linear function method.

Attachment

Through the process of reviewing NUREG-1829, the NRC identified that initiating event

frequencies for LOCAs in the NRC SPAR model were derived from the 25-year fleet

average operations life when NUREG-1829 was published in 2008. The NRC averaged

the 25-year and 40-year LOCA frequencies to obtain frequencies with a more accurate

reflection of fleet average operations. The NRC discussed the fact that the LOCA

initiating event frequencies in the SPAR model were dated at the regulatory conference.

The application of these methods resulted in a reduction in the break frequency for

MLOCAs between 4.5 to 6 inches to 5.21E-6 per year.

4. Your staff presented information related to strategies for reactor coolant system

cooldown, throttling of the emergency core cooling system (ECCS) flow to the reactor

core and refilling of the refueling water storage tank (RWST). These strategies would be

employed to increase the time available for the various recovery methods.

When evaluating the lower range in the increase in risk, the NRC evaluated these

strategies to determine the impact on the time available to recover valve SI-2-8982B.

These strategies do not directly provide for successful recovery of valve SI-2-8982B, but

instead slow the drain rate of the RWST, allowing additional time to implement the

electrical and mechanical recovery options.

Throttling of ECCS flow is directed by Emergency Operating Procedure (EOP)

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

Revision 21, Step 18. This procedure directs operators to stop all but one ECCS

centrifugal charging pump, provided the reactor coolant system is at least 70°F

subcooled. This action could occur at various times during the reactor coolant system

cooldown and results in a reduction in ECCS flow to approximately 400-500 gallons per

minute (gpm).

Refilling of the RWST is directed in Step 7.a of Procedure ECA 1.1. This procedure step

directs operators to ECA 1.1, Appendix M, RWST Makeup, and provides two methods

for adding inventory to the RWST. Makeup from the spent fuel pool is the preferred

method. Your staff provided analysis that demonstrated the ability to add approximately

41,700 gallons of spent fuel pool water inventory to the RWST at a rate of 250 gpm. The

liquid hold-up tanks could also be used for makeup but only after recirculation, sampling,

and evaluation by the technical support center (TSC) staff.

The NRC reviewed these strategies and determined they would have a high likelihood of

success (97.8 percent) because they are procedurally driven, high stress, and have

mostly nominal performance shaping factors (PSFs).

While the actions to reduce the drain rate on the RWST through adding inventory and/or

reducing ECCS flow are not procedurally directed until directed by Procedure ECA 1.1,

the NRC considered that for smaller LOCAs, full staffing of the TSC would likely have

occurred prior to the swap over from the RWST to the containment recirculation sump.

With this additional technical expertise available, the NRC assumed that the reduced

drain rate on the RWST would provide additional time for recovery actions and factored

this time into the individual analysis of each method. A sensitivity analysis was

performed that showed changes in the failure rate to refill the RWST or throttle ECCS

had a negligible effect on the outcome of the detailed risk evaluation.

A-2

5. Your staff presented a timeline that, following cessation of ECCS injection flow,

demonstrated a peak core temperature of 1800°F (i.e., onset of core damage) occurring

at 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Experts from the NRCs Office of Nuclear Reactor Research reviewed your Modular

Accident Analysis Program (MAAP) thermal-hydraulic analysis and found the timing for

core damage to be acceptable for the conditions of the analysis. The original analyses

assumed the time to core damage after termination of injection was approximately

1.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, consistent with data from NUREG-2187, Confirmatory Thermal-Hydraulic

Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk

Models - Byron Unit 1. The NRC considered the additional time to core damage when

evaluating the lower range of the increase in core damage frequency for the final

significance determination.

6. Your staff presented an additional recovery method not originally recognized during

development of the preliminary risk assessment. This new recovery method involved

the use of a maintenance procedure to install electrical jumpers to bypass the interlock

that prevented opening of valve SI-2-8982B.

The NRC reviewed the additional recovery method involving the use of electrical

jumpers to bypass the interlock that prevented opening of valve SI-2-8982B. The

analyst used SPAR-H to determine the feasibility of the proposed recovery action. The

analyst identified several impediments to implementing this recovery method.

Specifically, the method would have to be diagnosed with low experience and

training, was moderately complex, and would be performed under high stress.

The NRC reviewed the licensees human reliability analysis for the electrical jumper

recovery method, which estimated the failure probability of the human action

of 8.7E-2. The NRC concluded that procedures were poor in both diagnosis and

action. The task was also subject to high dependency to the electrical recovery

because the same crew would be used, the tasks would be close in time, and no

additional cues would be present. Procedures for use of the jumper method were

not referenced in the EOPs and were subject to TSC staff action to develop and

prepare for use by referencing a maintenance procedure during the event. The

NRC estimated the failure probability of the jumper method to be 5.7E-1. The NRC

included credit for the electrical jumper method with TSC directed recoveries,

discussed in Item 9, when determining the lower range of the increase in core damage

frequency for the final significance determination.

7. Your staff presented information related to the Time Available, Experience and

Training, Procedures, and Ergonomics PSFs for recovery by local, manual operation

of the valve (also referred to as the mechanical recovery option). In particular, your staff

presented information that the Time Available PSF for action should be characterized

as Extra Time Available, and the Experience and Training and Ergonomics PSFs

should be characterized as Nominal. The NRC reviewed the licensees suggested

enhancements to the PSFs and the assumptions used in the preliminary detailed risk

evaluation.

For the Time Available PSF, the analyst determined the initial attempt to mechanically

open valve SI-2-8982B would occur prior to any TSC action to refill the RWST or throttle

ECCS flow. This is because these actions are directed by EOP emergency contingency

A-3

action procedures, which are implemented after the failure to open valve SI-2-8982B and

valve SI-2-8982A during implementation of Procedure EOP E-1.3.

The timeline presented at the regulatory conference was derived from licensee

calculation MAAP16-03, Diablo Canyon Power Plant Calculation - Loss of Recirculation

Function, Revision 0, and assumed ECCS injection was reduced to only one train of

charging injection 12 minutes after the RWST reached 33 percent level. The NRC

concluded that the licensees emergency operating procedures would not call for this

action until numerous steps had been completed after reaching the 33 percent level in

the RWST. The NRC identified that a sensitivity in calculation MAAP16-03, which

reduced ECCS injection to one charging train at 45 minutes after RWST level reached

33 percent, was more reflective of how the plant would be operated, recognizing it could

take longer than 45 minutes. This sensitivity included RWST make-up and shortened

the timeline by approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This aided the NRC in concluding that the time

available was shorter than the timeline presented at the regulatory conference and

would be less than five times that required for the human performance basic events.

Additionally, the NRC considered the guidance in Section 3.1, Available Time, of

document INL/EXT-10-18533, SPAR-H Step-by-Step Guidance, Revision 2, to assign

the available time PSF for action as nominal and the remainder of the time was assigned

to the diagnosis part of the event. As such, the analyst determined that the Time

Available PSF should be characterized as Nominal for mechanically opening

valve SI-2-8982B.

For the Experience and Training PSF, the licensee noted that operators are trained on

operations of similar motor-operated valves elsewhere in the plant. The NRC

considered local, manual valve operation in the recirculation chamber, while dressed in

protection clothing, as unique when compared to other motor-operated valves. This

uniqueness, combined with the few past operations of the valve in this manner, and

industry operating experience, resulted in the analyst concluding the Experience and

Training PSF to be low.

Your staff discussed the procedures used for opening the recirculation valve chamber

guard. The NRC noted that in calculation SDP16-05, SPAR Evaluation for

8982B/8700B Interlock Failure, Revision 0, the licensee assumed the Procedures PSF

to be available, but poor. For the Procedures PSF, the analyst determined that

information needed to complete the mechanical recovery method was not contained in

standard operating procedures.

In particular, your staff stated that there is not an existing emergency procedure to open

the valve SI-2-8982B chamber guard and that during the postulated event, existing

outage related work instructions would be used to develop the emergency instructions to

open the chamber guard to allow for the mechanical recovery method. The analyst

determined that this lack of guidance more closely aligned with the definition of an

incomplete procedure rather than a procedure that is available but poor.

For the Ergonomics PSF, the NRC concluded that the information supplied by your

staff and the design of the plant supports correct performance, but does not enhance

performance or make tasks easier to carry out than typically expected. As such, the

analyst determined that the Ergonomics PSF is more appropriately characterized as

Nominal.

A-4

The NRC reassessed the overall human error probability based on the changes to the

PSFs discussed above. When evaluating the lower range of the increase in core

damage frequency for the final significance determination, the NRC lowered the

mechanical recovery failure probability from 5.8E-1 (42.0 percent success) to 1.1E-1

(89.0 percent success).

8. Your staff presented information related to the Time Available and Procedures PSFs

related to electrical recovery option using the motor contactors. In particular, your staff

presented information that the Time Available PSF should be characterized as

Expansive Time Available and the Procedures PSF should be characterized as

Available but Poor. The NRC reviewed the licensees suggested enhancements to the

PSFs and the assumptions used in the preliminary detailed risk evaluation.

For the Time Available PSF, the analyst determined that initiation of the electrical

recovery method occurs after initiation of the mechanical recovery method. The cause

of this delay is due to the structure of the EOPs. In particular, Procedure EOP-1.3,

Transfer to Cold Leg Recirculation, Revision 15, response not obtained for Step 6.b.2,

first directs operators to manually or locally open valve SI-2-8982B with assistance from

mechanical maintenance at the 64-foot residual heat removal penetration. The NRC

determined that because of standard procedural use and adherence rules, the manual,

mechanical opening of Valve SI-2-8982B would occur first.

Following the inability to open valve SI-2-8982B with assistance from mechanical

maintenance at the 64-foot residual heat removal penetration, the licensee would

progress through Procedure EOP-1.3, Step 8. In scenarios involving the inability to

establish cold leg recirculation using the train A ECCS components, the licensee would

then be directed to EOP ECA 1.1, Loss of Emergency Coolant Recirculation,

Revision 21. Step 2 of ECA 1.1 instructs operators to try to restore emergency coolant

recirculation equipment by several means. The NRC determined that this procedural

step is the first guidance that directs plant operators to attempt the electrical recovery

method. Specifically, Step 2.d has operators check power available to valves required

for recirculation swap over and refers to an appendix with valve power supplies.

The NRC assumed that this would delay the initiation of the electrical recovery to the

point where the time available would be less than five times the time required. This

timing led to the NRC assigning the PSF with nominal time.

Further, the NRC considered the guidance in Section 3.1, Available Time, of

document INL/EXT-10-18533, SPAR-H Step-by-Step Guidance, Revision 2, to assign

the available time PSF for action as nominal and the remainder of the time was assigned

to the diagnosis part of the event. This determination of nominal time for action was

different than originally characterized in the inspection report for this issue, where it was

characterized as extra time. Through re-analysis of the suggested changes to the PSFs,

the NRC determined that the PSF is better characterized as nominal time.

For the Procedures PSF, the analyst determined that information needed to complete

the electrical contactor recovery method is not contained in standard operations

department procedures. In particular, Procedure O-22 required plant personnel to refer

to other documents, including complex electrical drawings and schematics, to select the

correct contactors. The NRC determined that the PSF for Procedures was the less risk

significant available but poor even though the guidance in Procedure O-22 closely

aligned with the definition of an incomplete procedure.

A-5

For the Complexity PSF, the licensee proposed moderate complexity because there

was little ambiguity in operating the valve contactor. The NRC assigned the PSF for

Complexity with the less risk significant moderate despite having some aspects which

align with highly complex.

The NRC reassessed the overall human error probability based on the changes to the

PSFs discussed above. When evaluating the lower range of the increase in core

damage frequency for the final significance determination, the NRC concluded that the

electrical recovery yielded a failure probability of 3.8E-1 (62.0 percent success), the

same value used in the preliminary risk evaluation.

The licensee presented information at the regulatory conference that operation of the

wrong contactor was less likely to damage the motor than originally thought. The NRC

originally assumed that operation of the closed contactor would damage the motor for

the motor-operated valve based on data from the Electrical Power Research Institute

study.

The licensee provided information that a blue indicating light at the cabinet where the

valve was being operated would illuminate between 7 - 11 seconds to alert operators

that an electrical overload condition existed. The licensee stated that operators would

then cease operating the contactor. The licensee presented motor performance curve

data that showed that no motor damage would occur due to the motor operating at

locked rotor amperage for up to 10 seconds.

The NRC concluded that due to the proximity of the overload light illumination to the time

that the motor could sustain damage, that 75 percent of the time the motor would sustain

damage or burnup such that the motor for the motor operator would be rendered

unavailable. The 75 percent value was derived from judgement of the nominal

illumination time, time for recognition by the operator, uncertainty of the operators

knowledge of the meaning of the light, reaction time by the operator, and the EPRI study

of valve motors being damaged after just 12 seconds of locked rotor amp operations.

This motor unavailability would eliminate further electrical recovery attempts.

9. Your staff presented information that all recovery options, including the mechanical

opening of the valve and electrical opening of the valve by use of the motor start

contactors, would be pursued in parallel.

The NRC, when developing the final significance determination, reviewed the licensees

EOP and ECA procedures to determine the exact sequence of actions expected

following the failure of valve SI-2-8982B. Following any LOCA, operators transfer the

ECCS to cold leg recirculation following depletion of the RWST, as directed in

Procedure EOP-1.3, Transfer to Cold Leg Recirculation, Revision 15. For the core

damage sequence of concerns related to the performance deficiency that affected the

ability to open valve SI-2-8982B, the NRC noted the following important sequence of

actions expected:

EOP-1.3, Step 6.b.2, open valve SI-2-8982B to place residual heat removal

train B in cold-leg recirculation.

EOP-1.3, response not obtained for Step 6.b.2, locally open valve SI-2-8982B.

The NRC assumed this is the mechanical recovery option, discussed in Item 7

above.

A-6

EOP-1.3, Step 8, place residual heat removal train A in cold-leg recirculation.

EOP-1.3, response not obtained for Step 8.b, Go to ECA-1.1, Loss of

Emergency Coolant Recirculation.

ECA-1.1, Step 2.d, try to restore emergency coolant recirculation equipment by

locally operating valves as required. The NRC assumed this is the electrical

contactor recovery option, discussed in Item 8 above.

ECA-1.1, Step 7, add makeup to the RWST as necessary.

ECA-1.1, Step 10, initiate reactor coolant system cooldown to cold shutdown.

ECA-1.1, Steps 15-18, throttle ECCS flow to minimum required to remove decay

heat.

Following successful refilling of the RWST and throttling of ECCS, the licensee would

have a number of recovery options available through TSC directed recoveries. These

recoveries could include additional mechanical recovery attempts and if the motor

operator was not damaged, electrical recovery attempts through use of the motor

contactors or the electrical jumper method described at the regulatory conference. Extra

time could also be used to prolong the time available to restore ECCS recirculation

through strategies, such as, makeup to the RWST from the boric acid blender, initiation

of normal charging from the volume control tank or refilling of the RWST from the liquid

holdup tanks.

There is uncertainty associated with the likelihood of these recoveries because they

involve diagnostic troubleshooting and assessment by the TSC staff and actions that are

not, in some cases, procedurally driven. The NRC determined that TSC directed

recoveries are subject to high dependency because the same crew would be used, the

tasks would be close in time, and no additional cues would be present. As such, the

NRC evaluated the composite likelihood of failure of these actions and determined they

have an effective failure probability of 5.0E-1, using the SPAR-H guidance for high

dependency.

Based on the above, the NRC found that the proposed recovery actions are more

reflective of sequentially directed actions rather than parallel actions. The NRC

considered the continuous action nature of ECA-1.1, Step 2, which allows the TSC to

pursue multiple methods to recover ECCS recirculation following the initial failure of

valve SI-2-8982B and the inability to recover the valve by local manual operation.

Using insights from the above sequence, the NRC evaluated the availability of three

potential recovery methods combined with the failure probabilities of throttling ECCS

flow, refilling the RWST, and potentially damaging the motor during the electrical

recovery method. When evaluating the lower range of the increase in core damage

frequency for the final significance determination, the NRC approximated the overall

recovery probability by multiplying the failure probabilities of the mechanical and

electrical options and reduced the effective recovery failure probability from 2.4E-1

(76.0 percent success) to 3.6E-2 (96.4 percent success).

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10. Your staff identified that additional risk benefit could be gained through recovery of

valve SI-2-8982A, the opposite train valve that is subjected to the same maintenance as

valve SI-2-8982B and, therefore, could be susceptible to failure due to the same

incorrectly set external limit switch. Your staff did not specifically quantify recovery of

valve SI-2-8982A, but instead included a qualitative sensitivity that additional risk credit

could be gained through recovery of this valve.

The analyst followed the NRCs guidance for crediting recovery in cutsets involving a

common cause failure basic event contained in Section 5.0, Common Cause Failure

Modeling, of Volume 1, Internal Events, of the Risk Assessment of Operational

Events. This guidance prescribes that for cutsets that involve the common cause failure

basic event that was impacted by the observed single failure, the potential for recovery

should consider only the failure mechanism of the observed failure. As a result, the

NRC did not consider any recovery credit in cutsets containing a common cause failure

of valve SI-2-8982A.

11. Your staff presented application of updated recovery actions to the increase in core

damage frequency from external events included in the preliminary risk assessment.

Your analysis represented an increase in core damage frequency of 4.56E-8 per year.

Similarly, the NRC applied the recoveries, discussed in Items 4-9 above, to external

events when evaluating the lower range of the increase in core damage frequency for

the final significance determination. When including these recoveries, the NRC

estimated the increase in core damage frequency, from external events, as 5.0E-8 per

year.

In summary, we concluded that our preliminary risk assessment of 7.6E-6 per year represented

the upper range of the increase in core damage frequency associated with the performance

deficiency. Based on the information provided by your staff at the regulatory conference, the

NRC adjusted a number of assumptions used in the preliminary risk assessment to determine

the lower range of the increase in risk associated with the performance deficiency. Specifically,

the NRC adjusted the common cause alpha factors, the initiating events frequency for various

MLOCAs scenarios, and the assumption relative to the actuation of containment spray for a

MLOCA. The NRC also performed a variety of human error probability calculations to

determine the likelihood of recovering valve SI-2-8982B. Notably, the NRC adjusted the PSFs

for the mechanical and electrical recovery methods.

For the mechanical recovery method, the NRC applied the less risk significant Ergonomics -

Nominal PSF. The NRC continued to apply the Procedures - Incomplete PSF because, as

stated by your staff at the regulatory conference, instructions would need to be developed to

open the recirculation guard chamber and access valve SI-2-8982 during a LOCA event. The

NRC also continued to apply the Available Time - Nominal PSF for this recovery method.

Specifically, while your staff provided detailed information related to strategies to refill the RWST

and throttle ECCS flow, the NRC concluded the initial attempt to mechanically open

valve SI-2-8982B would occur prior to any procedurally driven action to refill the RWST or

throttle ECCS.

The NRC also considered information presented at the regulatory conference and concluded

that your staffs timeline represented recovery under ideal conditions. Using assumptions based

of the sequence of steps outlined in the EOPs and recognizing the uncertainty that

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accompanies complex reactor events, the NRC concluded the available time to be less than five

times the time required for the human performance basic events.

For the electrical recovery methods, the NRC applied the less risk significant Procedures and

Complexity PSF. The NRC also applied the more risk significant Available Time - Nominal

PSF for this recovery method. Specifically, the NRC staff reviewed the Diablo Canyon Power

Plant EOPs and identified that application of the electrical recovery method would not occur

first, as assumed in the preliminary risk assessment. Instead, the NRC staff concluded that the

EOP procedure structure would direct this action after the mechanical recovery method failed

but before action is taken to refill the RWST and throttle ECCS flow. Recognizing the

uncertainty that accompanies complex reactor events, the NRC concluded the available time to

be less than five times the time required for the human performance basic events.

The results of these calculations, which removed much of the conservativism from the

assumptions used in the preliminary risk assessment, predicted a high likelihood of success

(96.4 percent success) for recovering valve SI-2-8982B. Using these assumptions, the NRC

concluded the lower range of increase in core damage frequency associated with the

performance deficiency to be 1.3E-6 per year. Because the NRCs calculated lower and upper

estimations of increase in core damage frequency of the performance deficiency were both

greater than 1.0E-6 but less than 1.0E-5, the NRC determined the finding was of low to

moderate safety significance (White).

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