ML17017A437

NRC Response to Questions Involving Diablo Canyon Power Plant - Final Significance Determination of a White Finding; NRC Inspection Report 05000275/2016010 and 05000323/2016010

ML17017A437
Person / Time
Site:	Diablo Canyon
Issue date:	01/17/2017
From:	Jeremy Groom NRC/RGN-IV/DRP/RPB-A
To:	Halpin E Pacific Gas & Electric Co
Groom J
References
ML16363A429, ML17010A093 IR 2016010
Download: ML17017A437 (4)

See also: IR 05000275/2016010

Text

January 17, 2017

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:

NRC RESPONSE TO QUESTIONS INVOLVING DIABLO CANYON POWER

PLANT - FINAL SIGNIFICANCE DETERMINATION OF A WHITE FINDING;

NRC INSPECTION REPORT 05000275/2016010 AND 05000323/2016010

Dear Mr. Halpin:

This letter provides you the NRCs response to eight questions, submitted in writing by Pacific

Gas and Electric (PG&E) Company on January 4, 2017, (Agency wide Documents Access and

Management System (ADAMS) Accession No. ML17010A093). The question and answers are

related to the NRCs assumptions used in the final significance determination documented in

NRC inspection report 05000275/2016010 and 05000323/2016010, Diablo Canyon Power

Plant - Final Significance Determination of a White Finding, Notice of Violation and Follow-up

Assessment Letter, dated December 28, 2016 (ADAMS Accession No. ML16363A429). The

NRC provided verbal answers to these questions to Messrs. Hossein Hamzehee and Nathan

Barber of your staff, during a teleconference on January 9, 2017.

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.

Sincerely,

/RA Ryan Alexander Acting for/

Jeremy R. Groom, Chief

Project Branch A

Division of Reactor Projects

Docket Nos. 50-275 and 50-323

License Nos. DPR-80 and DPR-82

Enclosure: NRC Response to PG&E Questions

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E. LAMAR BLVD.

ARLINGTON, TX 76011-4511

ML17017A437

SUNSI Review

By: JRG/dll

ADAMS

Yes No

Non-Sensitive

Sensitive

Publicly Available

Non-Publicly Available

Keyword:

NRC-002

OFFICE

BC:DRP/A

DRS:SRA

D:DRS

D:DRP

BC:DRP/A

NAME

JGroom

RDeese

AVegel

TPruett

JGroom

SIGNATURE

/RA/

/RA/

/RA Jeff Clark

Acting for/

/RA/

/RA Ryan Alexander

Acting for/

DATE

1/11/2017

1/13/17

1/11/17

1/13/17

1/17/17

1

Enclosure

NRC RESPONSE TO PG&E QUESTIONS DATED JANUARY 4, 2017

The following eight questions were submitted in writing by PG&E on January 4, 2017, regarding

NRC inspection report 05000275/2016010 and 05000323/2016010, Diablo Canyon Power

Plant- Final Significance Determination of a White Finding, Notice of Violation and Follow-up

Assessment Letter, dated December 28, 2016 (ADAMS Accession No. ML16363A429).

The NRC provided the following verbal answers to these questions during a teleconference

with the licensee on January 9, 2017.

Question 1. Element #2 (Page A-1): The estimate of increase in core damage frequency

from large break LOCAs was 1.4E-7 per year. Is the 1.4E-7 value a delta-CDF or does it

represent the total contribution from large break loss-of-coolant accidents?

NRC Response to Question 1: The 1.4E-7 per year used for large break loss-of-coolant

accidents is a delta-CDF (core damage frequency) value.

Question 2. Element #5 (Page A-3): Following the termination of emergency core cooling

system flow which occurs at refueling water storage tank (RWST) Level=4%, did you

consider normal makeup to the reactor coolant system (RCS) from normal charging as

discussed within our response to request for information (RFI)-011?

NRC Response to Question 2: Yes. As discussed in the final significance determination,

Element 9, Page A-7, the NRC did consider normal makeup to the RCS from normal charging

when evaluating additional technical support center and emergency response organization

(TSC/ERO) directed recoveries.

Question 3. Element #7 (Page A-4): We note that the NRC has used emergency operating

procedure (EOP) E-1.3, Revision 15, in the response to Element #8. The response to RFI-

010 attached Revision 15 with its change documentation to illustrate when and why the

caution statement, discussed in the RFI, was added to the EOP. At DCPP, the current

revisions of EOP E-1.3 are Revision 31 (Unit 1) and Revision 22 (Unit 2), both effective

1/5/16. Did you use EOP E-1.3, Rev. 15 in your analysis for Element #7?

NRC Response to Question 3: Yes. However, the NRC reviewed both EOP E-1.3, Transfer to

Cold Leg Recirculation, Revision 15 and Revision 22 to inform our understanding of what

valves are manipulated prior to reaching the point at which operators enter emergency

contingency action procedure ECA 1.1, Loss of Emergency Coolant Recirculation. We also

compared this information with the most recent version of the Diablo Canyon Final Safety

Analysis Report to understand the sequence of valve manipulations performed in the first few

minutes of the transition to emergency core cooling system cold leg recirculation.

Question 4. Element #7 (Page A-4): What is the time duration used in your analysis to

perform local manual operation of 8982B?

NRC Response to Question 4: The NRC assumed 30 minutes for diagnosis, followed by

approximately 103 minutes of manual action to access and open valve SI-2-8982B through use

of the manual hand wheel.

2

Enclosure

Question 5. Element #8 (Page A-6): What EPRI study is referred to on page A-6?

NRC Response to Question 5: The Electric Power Research Institute (EPRI) study referenced

on page A-6 is EPRI TR-106563, Volume 1, Application Guide for Motor-Operated Valves in

Nuclear Power Plants, Revision 1. This report includes a summary of motor operated valve

incidents based on nuclear power plant experience documented in EPRI NP-6660D,

Application Guide for Motor-Operated Valves in Nuclear Power Plants, Appendix F.

Question 6. Element #8 (Page A-5): What is the time delay from RWST level at 33% to

initiate the electrical recovery?

NRC Response to Question 6: The NRC assumed the licensee would initiate the electrical

recovery option approximately 46 minutes after RWST level reached 33 percent. The NRC

assumed that the electrical recovery option would be completed approximately 209 minutes

after RWST level reached 33 percent.

Question 7. Element #9 (Pages A-6 and A-7): Please clarify whether the electrical and

mechanical recovery actions are performed in parallel or in series.

NRC Response to Question 7: The NRC assumed that the TSC/ERO could pursue both the

electrical and mechanical recovery actions in parallel but that because of the format of some

station procedures including those used to manually operate motor operated valves, certain

portions of the recovery actions were sequential.

Question 8. Element #9 (Page A-7): What actions are included in the effective failure

probability of 5.0E-1? Does this include actions to prolong the time to core damage or

actions to open 8982B?

NRC Response to Question 8: As discussed in the final significance determination, Element 9,

Page A-7, the actions included in the effective failure probability of 5.0E-1 include additional

TSC/ERO directed recoveries involving the electrical jumper method, additional electrical

contactor attempts, and/or additional mechanical operations.

As discussed in the final significance determination, Element 9, Page A-7, the NRC did account

for the added time to core damage through additional RCS/RWST makeup strategies when

evaluating the human error probability.