Pacific Gas and Electric Company It)

April18, 2018

PG&E Letter DCL-18-032

James M. Welsch Vice President Nuclear Generation and Chief Nuclear Officer

Diablo Canyon Power Plant P.O. Box 56 Avila Beach, CA 93424

805.545.3242 E·Mail: JMWl @pge.com

U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

10 CFR 50.59

Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Summary Report of 10 CFR 50.59. "Changes. Tests. and Experiments." for the Period of January 1. 2016. through December 31. 2017

Dear Commissioners and Staff:

Pursuant to 10 CFR 50.59, "Changes, tests, and experiments," Pacific Gas and Electric Company (PG&E) is providing the 10 CFR 50.59 Summary Report for Diablo Canyon Power Plant, Units 1 and 2, for the period of January 1, 2016, through December 31, 2017. In accordance with 10 CFR 50.59(d)(2), the Enclosure provides a brief description and summary of each evaluation for the changes, tests, and experiments performed during this period.

The Plant Staff Review Committee has reviewed all of the referenced evaluations.

PG&E makes no new or revised regulatory commitments (as defined by NEI 99-04) in this submittal.

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

Sincerely,

~.4-1 ul.tv----., ~es-M. Welsch

dqmg/6192/50846136 Enclosure cc: Diablo Distribution cc/enc: Kriss M. Kennedy, NRC Region IV Administrator

Christopher W. Newport, NRC Senior Resident Inspector Balwant K. Singal, NRR Senior Project Manager

A member of the STARS Alliance

Callaway • Diablo Canyon • Palo Verde • Wolf Creek

Diablo Canyon Power Plant, Units 1 and 2

Enclosure PG&E Letter DCL-18-032

SUMMARY REPORT OF 10 CFR 50.59, "CHANGES, TESTS, AND EXPERIMENTS," for the Period

January 1, 2016, through December 31, 2017

Summaries of the following 10 CFR 50.59 Evaluations (tracked as "Licensing Basis Impact Evaluations"- LBIE) are provided in this enclosure:

LBIE No. Subject Units 2016-001 Containment Subcompartment Analyses 1&2 2016-004* Changes to the WESTDYN Code for Reactor Coolant Loop 1&2

Piping Analyses 2016-005* Westinghouse (WEC) APOLLO Code Implementation 1&2 2016-006* Use of PAD 4.0 Model/Code (WCAP-15063) 1&2 2016-009* Control Rod Insertion after Cold Leg Loss of Coolant Accident 1&2

(LOCA), WCAP-16231-P 2016-012* Structural Finite Element Analyses Using SAP2000 1&2 2016-013 Open Phase Protection 1 2017-004 Hydrogen Recombiner Electrical Penetration Protection 2 2017-005 Hydrogen Recombiner Electrical Penetration Protection 1 2017-019 Open Class I Breaks 1&2 2017-023 Gas Decay Tank Rupture 1&2 2017-024 Vital Area Mission Doses 1&2 2017-025 Volume Control Tank Rupture 1&2 2017-026 Liquid Holdup Tank Rupture 1&2 2017-027 Open Phase Protection 2 2017-030 Revised Alternate Source Term Filter Doses 1&2 2017-032 Assessing Radiation Environments Used for Shielding 1&2

* Activity represents a re-evaluation (e.g., LBIEs performed by the Licensing Basis Verification Project) of a legacy item to verify consistency with current 1 0 CFR 50.59 standards.

2016-001 Containment Subcompartment Analyses

Description: Review of the impact of the revised containment subcompartment analyses on the containment internal structures as a result of the revised subcompartment peak pressure and differential pressure across the subcompartment structural elements.

Summary: The changes are only to the analysis inputs and postulated line breaks and are not a change to any structure, system, or component (SSG). The effect of the change on the subcompartment analysis was such that the previous analysis was no longer bounding. As such, a 50.59 evaluation was required. The revised structural integrity analyses demonstrate that the containment internal structures remain within

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their structural acceptance criteria and their integrity is maintained. The evaluation determined that prior Nuclear Regulatory Commission (NRC) approval was not required.

2016-004* Changes to the WESTDYN Code for Reactor Coolant Loop Piping Analyses

Description: The activity consisted of changes to the WESTDYN code from its NRC-approved version (WESTDYN7, FIXFM3, and WESDYN2 codes) to its current version (WESTDYN, Version 7.2).

Summary: The revised version of WESTDYN (version 7.2), incorporating FIXFM3 and WESDYN2, yields results that are essentially the same as the NRC-approved version. Therefore, this change was not a departure from a method of evaluation described in the Updated Final Safety Analysis Report (UFSAR). The evaluation determined that prior NRC approval was not required.

2016-005 Westinghouse (WEC) APOLLO Code Implementation

Description: Implementation of the Westinghouse (WEC) APOLLO code (WCAP-13524-P-A, Revision 1-A) for performing one-dimensional, two group diffusion calculations in support of the fuel nuclear design. APOLLO is used in Diablo Canyon Power Plant (DCPP) reactor physics analyses as presented in the Unit 1 and Unit 2 Core Physics Reports. However, the existing method of evaluations for the fuel nuclear design as described in the UFSAR Chapter 4 does not include the APOLLO code. Therefore, the change to the use of APOLLO for DCPP is considered an alternate methodology.

Summary: The evaluation concluded the activity did not result in a departure from a method of evaluation described in the UFSAR because the alternative method has been approved by the NRC for the intended application. The criteria relating to the use/application of an alternative method of evaluation have been satisfied. The evaluation determined that prior NRC approval was not required.

2016-006 Use of PAD 4.0 Model/Code (WCAP-15063)

Description: Use of the WEC PAD 4.0 Model/Code (WCAP-15063-P-A, Revision 1 ). The PAD 4.0 is a best estimate fuel rod performance model used for both fuel rod performance analysis and safety analysis inputs. PAD 4.0 has been used by WEC since 2001 in DCPP rod performance analyses as documented in the Unit 1 and Unit 2 Final Reload Evaluation and Core Operating Limits Reports. However, the existing method of evaluations for the fuel rod performance analysis

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and design model as described in the UFSAR does not include PAD 4.0. Therefore, the change to the use of PAD 4.0 for DCPP is considered a new methodology.

Summary: The evaluation concluded the activity did not result in a departure from a method of evaluation described in the UFSAR because the alternative method has been approved by the NRC for the intended application. The criteria relating to the use/application of an alternative method of evaluation have been satisfied. The evaluation determined that prior NRC approval was not required.

2016-009* Control Rod Insertion after Cold Leg Loss of Coolant Accident (LOCA), WCAP-16231-P

Description: WCAP-16231-P, Revision 0 addressed the crediting of rod insertion for hot leg switchover (HLSO) after a cold leg loss of coolant accident (LOCA). The impact of the Nuclear Steam Supply System Structural and Licensing Review revised analysis was incorporated into WCAP-16231-P, Revision1. The UFSAR did not discuss the control rod assembly and driveline evaluation that was performed in WCAP-16231-P, Revision 0. WCAP-16231-P supports crediting the insertion of the control rods for criticality control during HLSO. As such, various sections of the UFSAR have been updated. Additionally, there was a change in evaluation methodology used to confirm the core will remain subcritical during hot leg switchover following a LOCA.

Summary: The change adversely affected how an UFSAR described design function is controlled. Specifically this change involved crediting control rod insertion to provide the negative reactivity to assure core subcriticality during emergency core cooling system switchover to hot leg recirculation under combined LOCA and seismic conditions. The evaluation concluded the change in the method of controlling did not require prior NRC approval. The evaluation also concluded that the revised method has been approved by the NRC for the intended application and that the criteria relating to the use/application of an alternative method of evaluation have been satisfied. The evaluation determined that prior NRC approval was not required.

2016-012* Structural Finite Element Analyses Using SAP2000

Description: Use of an alternate finite element analysis method (SAP2000 computer code) for structural analyses of PG&E Design Class I and II structures and framing at DCPP. SAP2000 was used for the structural analysis of the Unit 1 polar crane and the Unit 2 annulus structure framing associated with the installation of the replacement recirculation sump strainer.

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Summary:

2016-013 2017-027

Description:

Summary:

2017-004 2017-005

Enclosure PG&E Letter DCL-18-032

The evaluation concluded the activity did not result in a departure from a method of evaluation described in the UFSAR, because it has been demonstrated that the use of SAP2000 corresponds to an "alternative method" that has been approved by the NRC for the intended application. The evaluation determined that prior NRC approval was not required.

Open Phase Protection for Unit 1 Open Phase Protection for Unit 2

Installation of two-channel PSStech open phase protection (OPP) devices on the high voltage side of both the 230/12 kV start-up transformers and on the high voltage side of both the 500/25 kV generator step-up transformers, at the solidly grounded neutral point of the Y connected windings. These devices will automatically detect if the associated transformer is not supplied with 3-phase power and, upon detection, will alarm in the control room and trip the transformer when the trip functions are enabled.

The OPP devices are normally configured in a 2-out-of-2 coincidence logic to protect against inadvertent trips. No failures can cause the firmware portion of the devices to initiate a trip of both offsite power circuits. The 500 kV OPP devices and the 230 kV OPP devices have different settings, so an error in setting would not affect both offsite power sources. The 500 kV OPP devices are interlocked with the main generator motor-operated disconnect switch, such that its trip function is inhibited when the plant is online. The evaluation determined that prior NRC approval was not required.

Hydrogen Recombiner Electrical Penetration Protection for Unit 2 Hydrogen Recombiner Electrical Penetration Protection for Unit 1

Description: Replacement of the existing fuse and addition of a new redundant fuse in series for the hydrogen recombiner power supply panel.

Summary: The addition of a redundant fuse in the hydrogen recombiner power supply panel will increase the likelihood of power failure (by virtue of increasing the number of components). However, the addition of a redundant fuse will not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. The UFSAR Section 15.5.17.2.9 describes that the LOCA dose analysis does not credit the hydrogen recombiner function. The added fuse for this activity does not alter the fundamental operational characteristics of the hydrogen recombiners such that these analysis assumptions continue to be bounding for any hydrogen recombiner

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failure. The evaluation determined that prior NRC approval was not required.

2017-019 Open Class I Breaks

Description: Revision of UFSAR Section 3.2.2.1.1, "PG&E Design Class 1," to allow on a case-by-case basis, a normally open valve or no valve at all at the point where systems transition from PG&E Design Class I components to a lower classification.

Summary: UFSAR Section 3.2.2.1.1 describes that a PG&E Design Class I boundary includes those portions of the system required to accomplish the specified PG&E Design Class I safety function and connected piping up to and including the first valve (including a safety or relief valve) that is either normally closed or capable of automatic closure when the PG&E Design Class I safety function is required. The analyses performed, for the specific cases where a valve would be normally open or where no valve existed, supported that the design functions are not impacted. However, the fundamental method of controlling the design functions is adverse and screened into a 10 CFR 50.59 evaluation. The evaluation determined that prior NRC approval was not required.

2017-023 Gas Decay Tank Rupture

Description: The dose analyses associated with a rupture of a gas decay tank (GOT) were revised to demonstrate the radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exclusion area (EAB) for two hours and located at any point on the outer boundary of the low population zone (LPZ) are below 0.5 rem whole body and 3 rem thyroid.

Summary: The evaluation determined that the revised GOT rupture analysis, based on updated inputs and conservative assumptions, results in a decrease to the whole body dose and no impact on the thyroid dose and, therefore, does not cause more than a minimal increase in dose consequences. The evaluation determined that prior NRC approval was not required.

2017-024 Vital Area Mission Doses

Description: Revision of the PG&E analyses and the DCPP Units 1 & 2 Radiation · Shielding Review to assess the estimated dose to operators performing post-accident missions in vital areas outside containment following a LOCA. The change also included the use of an alternate method (the use of source term scaling factors) for assessing the

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estimated dose to operators performing post-accident missions in vital areas outside containment following a LOCA.

Summary: The operator dose for each of the post-LOCA vital missions remains well below the regulatory limit of 5 rem whole body as required by NUREG- 0737. The evaluation also determined that the change did not result in a departure from a method of evaluation described in the UFSAR. The use of source term scaling factors corresponds to an "alternative method" that has been approved by the NRC for the intended application. The criteria relating to the revised analyses have been satisfied. The evaluation determined that prior NRC approval was not required.

2017-025 Volume Control Tank Rupture

Description: The dose analysis associated with a rupture of a volume control tank (VCT) was revised to demonstrate that the radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the EAB for two hours and located at any point on the outer boundary of the LPZ are below 0.5 rem whole body and 3 rem thyroid.

Summary: The evaluation determined that the revised VCT rupture analysis, based on updated inputs and conservative assumptions, results in a decrease to the whole body dose. The incremental increase in thyroid · dose does not exceed 10 percent of the difference between the previous calculated value and the regulatory limit. Therefore, the revised VCT rupture analysis does not cause more than a minimal increase in dose consequences. The evaluation determined that prior NRC approval was not required.

2017-026 Liquid Holdup Tank Rupture

Description: The dose analyses associated with a rupture of a liquid holdup tank (LHUT) was revised to demonstrate the radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the EAB for two hours and located at any point on the outer boundary of the LPZ are below 0.5 rem whole body and 3 rem thyroid.

Summary: The evaluation determined that the revised LHUT rupture analysis, based on updated inputs and conservative assumptions, results in a decrease to the thyroid dose. The incremental increase in whole body dose does not exceed 10 percent of the difference between the previous calculated value and the regulatory limit. Therefore, the revised LHUT rupture analysis does not cause more than a minimal

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increase in dose consequences. The evaluation determined that prior NRC approval was not required.

2017-030 Revised Alternate Source Term Filter Doses

Description: Revisions of PG&E analyses for determining the post-accident contact radiation levels to environmentally qualified equipment adjacent to the auxiliary building ventilation system (ABVS) filters, the control room ventilation system filters, and the fuel handling building ventilation system (FHBVS) filters were completed. This revision includes determining radiation doses to adjacent environmentally qualified equipment from the ABVS filters and the FHBVS filters. The revised analyses also performed an assessment of the mass iodine loading and heat generation loading as result of the alternate source term analyses on the impact of meeting the Regulatory Guide 1.52 acceptance criteria.

Summary: The evaluation concluded that the impact on the radiation environment used for equipment qualification demonstrates no equipment important to safety required to operate post-accident is impacted such that it would result in more than a minimal increase in consequences from a malfunction. The evaluation determined that prior NRC approval was not required.

2017-032 Assessing Radiation Environments Used for Shielding

Description: Use of additional analyses performed for assessing the radiation environments used for shielding (normal power operation) and the post-accident doses used for equipment qualification (EQ) as required by various DCPP licensing documents. The change also included an alternate method of evaluation since it applied scaling factors to the original analyses results rather than revising the original analyses.

Summary: The additional analyses demonstrated that the existing radiation environment used for components required to operate post-accident were not affected (i.e. remain within the EQ radiation dose limits). With respect to the alternative method for assessing the radiation environments, the change from one method of evaluation to another under 10 CFR 50.59 is acceptable if the new method has been approved by the NRC for the intended application. This is confirmed by showing that the method is approved for the type of analysis being conducted and applicable terms, limitations, and conditions for its use are satisfied. The evaluation determined that prior NRC approval was not required.

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