DCL-20-072, Response to Additional NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System

From kanterella
(Redirected from ML20233B187)
Jump to navigation Jump to search

Response to Additional NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System
ML20233B187
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 08/20/2020
From: Gerfen P
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-20-072
Download: ML20233B187 (6)
v • d • e


Text

Paula Gerfen Site Vice President Diablo Canyon Power Plant Mail code 104/6/605 P.O. Box 56 Avila Beach, CA 93424 805.545.4596 Internal: 691.4596 Fax: 805.545.4234 A member of the STARS Alliance Callaway

  • Diablo Canyon
  • Palo Verde
  • Wolf Creek August 20, 2020 PG&E Letter DCL-20-072 U.S. Nuclear Regulatory Commission 10 CFR 50.91 ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to Additional NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System

References:

1. PG&E Letter DCL-20-066, License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System, dated August 12, 2020, ADAMS Accession No. ML20225A303
2. PG&E Letter DCL-20-068, Response to NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System, dated August 16, 2020, ADAMS Accession No. ML20229A016
3. PG&E Letter DCL-20-069, Response to Additional NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System, dated August 18, 2020, ADAMS Accession No. ML20231A838
4. E-mail from NRC Senior Project Manager, Samson Lee, Diablo Canyon additional request for additional information: Exigent License Amendment Request for Application to provide a new Technical Specification 3.7.5, Auxiliary Feedwater System, Condition G (EPID: L-2020-LLA-0176), dated August 20, 2020

Dear Commissioners and Staff:

In Reference 1, Pacific Gas and Electric Company (PG&E) submitted an exigent license amendment request to revise Technical Specification 3.7.5, Auxiliary m

PacHic Gas and Electric Company*

Document Control Desk PG&E Letter DCL-20-072 August 20, 2020 Page 2 A member of the STARS Alliance Callaway

  • Diablo Canyon
  • Palo Verde
  • Wolf Creek Feedwater System. In References 2 and 3, PG&E responded to an NRC Staff request for additional information (RAI). In Reference 4, the NRC Staff provided an additional RAI via an e-mail, dated August 20, 2020. The Enclosure to this letter provides the PG&E responses to the Reference 4 RAI.

This letter does not contain new regulatory commitments (as defined by NEI 99-04).

If you have any questions or require additional information, please contact Mr. James Morris at (805) 545-4720.

I state under penalty of perjury that the foregoing is true and correct.

Executed on August 20, 2020.

Sincerely, Paula Gerfen Site Vice President kjse/51084143-16 Enclosure cc:

Diablo Distribution cc/enc: Samson S. Lee, NRR Senior Project Manager Scott A. Morris, NRC Region IV Administrator Christopher W. Newport, NRC Senior Resident Inspector Gonzalo L. Perez, Branch Chief, California Department of Public Health

Enclosure PG&E Letter DCL-20-072 1

PG&E Response to Additional NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System

References:

1. PG&E Letter DCL-20-066, License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System, dated August 12, 2020, ADAMS Accession No. ML20225A303
2. PG&E Letter DCL-20-068, Response to NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System, dated August 16, 2020, ADAMS Accession No. ML20229A016
3..PG&E Letter DCL-20-069, Response to Additional NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System, dated August 18, 2020, ADAMS Accession No. ML20231A838
4. E-mail from NRC Senior Project Manager, Samson Lee, Diablo Canyon additional request for additional information: Exigent License Amendment Request for Application to provide a new Technical Specification 3.7.5, Auxiliary Feedwater System, Condition G (EPID:

L-2020-LLA-0176), dated August 20, 2020 NRC Request for Additional Information (RAI)

Background:

When in proposed Condition G in Technical Specification (TS) 3.7.5, steam generator (SG) 1-2 will be isolated from AFW, but the SG 1-2 will continue to be used for power generation. The accident analysis for Steam Generator Tube Rupture (SGTR) is addressed in Updated Final Safety Analysis Report (UFSAR) Section 15.4.3. An important consideration in the accident consequence analysis is to identify the SG with the ruptured SG and isolate that SG. The analysis states that in the event of an SGTR, the plant must diagnose the SGTR and perform the required recovery actions to stabilize the plant and terminate the primary to secondary leakage. The major operator actions include identification and isolation of the ruptured SG, cool down and depressurization of the RCS to restore inventory, and termination of safety injection to stop primary to secondary leakage. The specific actions for detection are described in Item (1), which partially states: The AFW flow will begin to refill the SGs, distributing equal flow to the SGs. Since primary to secondary leakage adds additional liquid inventory to the ruptured SG, the water level will return to narrow range earlier in that SG and will continue to increase more rapidly. This response, as indicated by the SG water level instrumentation, provides confirmation of an SGTR event and also identifies the ruptured SG.

Enclosure PG&E Letter DCL-20-072 2

RAI:

When in proposed Condition G of TS 3.7.5, if a SGTR were to occur in the isolated SG, provide a discussion of the impact on the SGTR accident analysis in UFSAR Section 15.4.3, with emphasis on how operator actions to detect and isolate the SG may change.

PG&E Response:

SGTR Accident Analysis for a SGTR in SG 1-2 The Diablo Canyon Power Plant (DCPP) SGTR analyses are performed to address both thermal hydraulic concerns and offsite and control room dose concerns. A key input to the analyses is the timeframe for operators to identify and isolate the ruptured SG.

The primary concern of the UFSAR Section 15.4.3 SGTR analysis is the overfill of the ruptured SG which could result in water relief through the safety valves and damage to the 10 percent atmospheric dump valves. One of the priorities associated with the identification and isolation of the ruptured SG is halting auxiliary feedwater (AFW) flow to prevent an overfill condition in the ruptured SG. With AFW pre-isolated to SG 1-2, the AFW contribution to overfill is eliminated.

The UFSAR Section 15.5.20 SGTR thermal hydraulic dose input analysis focuses on maximizing the reactor coolant system (RCS) mass release to the environment from the ruptured SG to conservatively estimate the offsite and control room dose. In the SGTR thermal hydraulic dose input analysis, the primary contribution to the atmospheric release occurs before AFW is initiated following a reactor trip, and after AFW is isolated when the 10 percent atmospheric dump valve on the ruptured SG is assumed to fail open (limiting single failure). The contribution of mass released to the atmosphere during the approximate 10-minute assumed run time of the AFW pumps is less than 5 percent of the overall mass released during the event. Unit 1 is currently operating below 0.1 percent of the TS allowed RCS and secondary activity levels; therefore, the resulting accident dose would be well within the total effective dose equivalent regulatory limits..

In the UFSAR Section 15.4.3 described design basis SGTR analysis, the time to identify and isolate the ruptured SG is conservatively assumed to occur when the ruptured SG narrow range level is greater than the midpoint of just-on-span and 50 percent narrow range span. Although the AFW contribution to the SG 1-2 would be isolated, the break flow would continue to fill the SG alerting operations to the abnormal condition and not impede diagnosis of the SGTR and SG isolation, as described below.

SGTR Accident Analysis for a SGTR in SG 1-2 Conclusion In conclusion, the pre-isolation of AFW to SG-1-2 while in proposed TS Condition G, taking into consideration the low operational RCS activity, is bounded by the SGTR accident analyses.

Enclosure PG&E Letter DCL-20-072 3

Operator Response for a SGTR in SG 1-2 The operator response for DCPP Unit 1 to a postulated SGTR accident while in proposed TS 3.7.5 Condition G, including identification and isolation of the ruptured SG, is described below.

Identification of Ruptured SG Identification of the ruptured SG is accomplished by the following methods:

1.

An upward trend or spike on Radiation Monitors (RMs) RM-71 through RM-74 (RM-72 is associated with SG 1-2), even though the RMs are conservatively not credited for identification of the ruptured SG in the UFSAR Section 15.4.3 accident analyses.

a.

This capability is not impacted by AFW flow isolation to SG 1-2.

2.

An unexpected rise in SG level.

a.

Since it is known that AFW is isolated to SG 1-2, any rise in level would be considered unexpected. Also, in order to be in Unit 1 Emergency Operating Procedure (EOP) E-3, "Steam Generator Tube Rupture," a Safety Injection signal would have occurred, and the ruptured SG would be filling at approximately 300 gallons per minute (gpm) after 10 minutes considering the break flow flashing fraction. At initiation of the SGTR and upon entry into the initial EOP E-0 "Reactor Trip or Safety Injection," AFW flows are throttled to 435 gpm total to minimize the cooldown while maintaining a heat sink, which would be approximately 150 gpm to SGs 1-1, 1-3, and 1-4. Therefore, SG 1-2 would be filling at approximately twice the rate as the intact SGs for the UFSAR Section 15.4.3 assumed severance of one entire SG tube.

b.

Therefore, detection capability to identify the ruptured SG continues to exist with AFW isolated to SG 1-2.

3.

If the ruptured SG is still not able to be identified, then per the guidance of EOP E-3, the SGs are sampled by the Chemistry personnel to assist in rupture SG identification.

a.

The ability to sample is not impacted by AFW isolation.

Isolation of Ruptured SG and Remaining Actions Following identification of the ruptured SG, operators are instructed to isolate flow from the affected SG in accordance with EOP E-3 Appendix FF as follows:

  • Auto Setpoint on impacted 10 percent steam system dump valve is raised to control at 1040 pounds per square inch gage (psig) instead of 1020 psig.

Enclosure PG&E Letter DCL-20-072 4

  • Turbine-Driven AFW steam supply valves are isolated as applicable
  • Associated SG blowdown and sample isolation valves are closed
  • Ruptured SG is verified isolated from remaining SGs SG isolation would be complete as soon as the operator completes Appendix FF in accordance with step 3 of EOP E-3.

EOP E-3 step 3 directs operators to Appendix FF, and then in parallel, the Shift Foreman continues with step 4 of the procedure. EOP E-3 Step 4 directs the operators to check if SG level greater than 15 percent Narrow Range (NR) level; and if it is not, to maintain SG feed flow. Once the SG reaches 15 percent NR level, the procedure directs the operator to isolate feed flow. In the scenario with AFW isolated to SG 1-2, feed flow will already be isolated. Operators would not make any attempts to establish feed flow outside of the procedural guidance and would continue through the EOP E-3 procedure to cooldown and depressurize the ruptured SG, and isolate safety injection.

Operator Response for a SGTR in SG 1-2 Conclusion In conclusion, the isolation of AFW to SG 1-2 does not have any adverse impact on the operational response to a SGTR.

Retrieved from "https://kanterella.com/w/index.php?title=DCL-20-072,_Response_to_Additional_NRC_Request_for_Additional_Information_Regarding_License_Amendment_Request_20-01,_Exigent_Request_for_Revision_to_Technical_Specification_3.7.5,_Auxiliary_Feedwater_System&oldid=3957065"