DCL-15-047, License Amendment Request 15-02 Revision of Updated Final Safety Analysis Report for Beacon Power Distribution Monitoring System Methodology and Technical Specification 5.6.5. Core Operating Limits Report (Colr), For...

From kanterella
Jump to navigation Jump to search

License Amendment Request 15-02 Revision of Updated Final Safety Analysis Report for Beacon Power Distribution Monitoring System Methodology and Technical Specification 5.6.5. Core Operating Limits Report (Colr), For...
ML15107A333
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 04/16/2015
From: Allen B
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-15-047, TAC ME7803, TAC ME7804
Download: ML15107A333 (38)
v • d • e


Text

Pacific Gas and Electric Company Barry S. Allen Diablo Canyon Power Plant Vice President, Nuclear Services Mail Code 104/6 P. 0. Box 56 Avila Beach, CA 93424 805.545.4888 Internal: 691.4888 April16, 2015 Fax: 805.545.6445 PG&E Letter DCL-15-04 7 U.S. Nuclear Regulatory Commission 10 CFR 50.90 ATTN: Document Control Desk Washington, D.C. 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 License Amendment Request 15-02 Revision of Updated Final Safety Analysis Report for BEACON Power Distribution Monitoring System Methodology and Technical Specification 5.6.5. "Core Operating Limits Report (COLR)." for the PARAGON and NEXUS Core Design Methods

References:

1. NRC Letter, "Diablo Canyon Power Plant, Unit Nos. 1 And 2 - Issuance of Amendments Re: Change To Final Safety Analysis Report Update To Allow .

The Use of BEACQN Core Monitoring and Operations Support System (TAC NOS. ME7803 AND ME7804)," dated January 9, 2013

Dear Commissioners and Staff:

Pursuant to 10 CFR 50.90, Pacific Gas and Electric Company (PG&E) hereby requests approval of the enclosed proposed amendment to Facility Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP) respectively.

The enclosed license amendment request (LAR) proposes to revise the Best Estimate Analyzer for the Core Operations-Nuclear (BEACON) Power Distribution Monitoring System methodology described in the DCPP Units 1 and 2 Updated Final Safety Analysis Report (UFSAR) Section 4.3.2.2, "Power Distribution," to the method described in WCAP-12472-P-A, Addendum 4, "BEACON Core Monitoring and Operation Support System."

In addition, the enclosed LAR proposes to revise Technical Specification (TS) 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," Section b to replace A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

Document Control Desk PG&E Letter DCL-15-047 April16, 2015 Page 2 WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," with NRC-approved WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and NRC-approved WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."

The Enclosure provides a detailed description and technical evaluation of the proposed changes, including PG&E's determination that the proposed changes involve no significant hazards. provides the marked-up pages of the proposed TS 5.6.5.b change. provides the retyped proposed TS pages. Attachment 3 provides the marked-up pages to the DCPP UFSAR Section 4.3.2.2 for information only. provides marked-up TS Bases page changes for information only.

PG&E requests approval of this LAR no later than April 15, 2016. PG&E requests the license amendments be made effective upon NRC issuance, to be implemented for DCPP Unit 2 prior to MODE 4 at the start of Unit 2 Cycle 20 and for DCPP Unit 1 prior to MODE 4 at the start of Unit 1 Cycle 21.

PG&E makes no regulatory commitments (as defined by NEI 99-04) in this letter.

This letter includes no revisions to existing regulatory commitments.

In accordance with site administrative procedures and the Quality Assurance Program, the proposed amendment has been reviewed by the Plant Staff Review Committee.

Pursuant to 10 CFR 50.91, PG&E is sending a copy of this proposed amendment to the California Department of Public Health.

If you have any questions or require additional information, please contact Mr. Philippe Soenen at 805-545-6984.

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

Executed on April 16, 2015.

Sincerely, Jfelj _5, Aa-Barry S. Allen Vice President, Nuclear Services A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

Document Control Desk PG&E Letter DCL-15-047 April16, 2015 Page 3 kjse/4328 Enclosure cc: Diablo Distribution cc/enc: Marc L. Dapas, NRC Region IV Thomas R. Hipschman, NRC Senior Resident Inspector Siva P. Lingam, NRR Project Manager Gonzalo L. Perez, Branch Chief, California Dept of Public Health A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

Enclosure PG&E Letter DCL-15-047 Evaluation of the Proposed Change License Amendment Request 15-02 Revision of Updated Final Safety Analysis Report for BEACON Power Distribution Monitoring System Methodology and Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," for the PARAGON and NEXUS Core Design Methods

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION
3. TECHNICAL EVALUATION
4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3. Significant Hazards Consideration 4.3 Conclusions
5. ENVIRONMENTAL CONSIDERATION
6. REFERENCES ATTACHMENTS:
1. Technical Specification Pages Markups
2. Technical Specification Pages Retyped
3. UFSAR Pages Markups (For Information Only)
4. TS Bases Pages Markups (For Information Only)

Enclosure PG&E Letter DCL-15-047 EVALUATION

1.

SUMMARY

DESCRIPTION This letter is a request to amend the Facility Operating Licenses DPR-80 and DPR-82 for Units 1 and 2, respectively of the Diablo Canyon Power Plant (DCPP) as follows:

BEACON Methodology The proposed change involves a change to the Best Estimate Analyzer for the Core Operations-Nuclear (BEACON) methodology described in the Updated Final Safety Analysis Report (UFSAR), Section 4.3.2.2, "Power Distribution." In addition, the proposed change would revise the references to this methodology in Technical Specification (TS) Bases Sections 3.1.7, "Rod Position Indication;"

3.2.1, "Heat Flux Hot Channel Factor (F 0 (Z))," 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor(F~H);" 3.2.4, "Quadrant Power Tilt Ratio (QPTR);" and 3.3.1, "Reactor Trip System (RTS) Instrumentation."

The proposed change would update the current references in the UFSAR and TS Bases from the BEACON Power Distribution Monitoring System (PDMS) methodology contained in WCAP-12472-P-A, Addendum 1-A, "BEACON Core Monitoring and Operations Support System," (Reference 1) to the Westinghouse BEACON PDMS methodology contained in WCAP-12472-P-A, Addendum 4, Revision 0, "BEACON Core Monitoring and Operation Support System" (Reference 2).

PARAGON and NEXUS Methodology The proposed change would also revise TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," Section b to replace WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores,"

(Reference 3) with NRC-approved WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," (Reference 4) and NRC-approved WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology" (Reference 5). The reference to the topical reports (TR)

WCAP-16045-P-A and WCAP-16045-P-A, Addendum 1-A in TS 5.6.5 is without date and revision number consistent with Amendments No. 195 and 196 for DCPP Units 1 and 2 (Reference 16), respectively, that adopted TSTF-363, "Revise Topical Report References in ITS 5.6.5, COLR."

1

Enclosure PG&E Letter DCL-15-04 7

2. DETAILED DESCRIPTION BEACON Methodology Updating the BEACON PDMS methodology referenced in the UFSAR from the methodology contained in WCAP-12472-P-A, Addendum 1-A, "BEACON Core Monitoring and Operations Support System," to the Westinghouse BEACON PDMS methodology contained in WCAP-12472-P-A, Addendum 4, Revision 0, "BEACON Core Monitoring and Operation Support System" would result in the following changes:

UFSAR Section 4.3.2.2, "Power Distribution":

Replace WCAP-12472-P-A, Addendum 1-A, listed as Reference 33 in UFSAR Section 4.3, with WCAP-12472-P-A, Addendum 4, Revision 0, September 2012.

It is noted that the UFSAR text in Section 4.3.2.2 only refers to the BEACON methodology contained in UFSAR Section 4.3 Reference 33 (when discussing the PDMS) and does not include the BEACON WCAP number or title. Therefore, no change is required to the text of the UFSAR Section.

TS Bases 3.1.7, "Rod Position Indication":

Replace WCAP-124 72-P-A, Addendum 1-A, listed as Reference 5 in the Bases, with WCAP-12472-P-A, Addendum 4, Revision 0, September 2012.

It is noted that the TS 3.1.7, 3.2.1, 3.2.2, 3.2.4, and 3.3.1 Bases text only refers to the TS Bases reference number (when discussing an Operable PDMS) and does not include the BEACON WCAP number or title. Therefore, no change is required to the text of the TS Bases.

TS Bases 3.2.1, "Heat Flux Hot Channel Factor":

Replace WCAP-12472-P-A, Addendum 1-A, listed as Reference 4 in the Bases with WCAP-12472-P-A, Addendum 4, Revision 0, September 2012.

TS Bases 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor":

Replace WCAP-12472-P-A, Addendum 1-A, listed as Reference 5 in the Bases, with WCAP-12472-P-A, Addendum 4, Revision 0, September 2012.

TS Bases 3.2.4, "Quadrant Power Tilt Ratio (QPTR)":

Replace WCAP-12472-P-A, Addendum 1-A, listed as Reference 5 in the Bases, with WCAP-12472-P-A, Addendum 4, Revision 0, September 2012.

2

Enclosure PG&E Letter DCL-15-047 TS Bases 3.3.1, "Reactor Trip System (RTS) Instrumentation":

Replace WCAP-12472-P-A, Addendum 1-A, listed as Reference 33 in the Bases, with WCAP-12472-P-A, Addendum 4, Revision 0, September 2012.

The proposed UFSAR changes are noted on the marked-up UFSAR pages provided in Attachment 3 to the Enclosure for information only.

The proposed TS Bases changes are noted on the marked-up TS Bases pages provided in Attachment 4 for information only.

PARAGON and NEXUS Methodology The TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," Section b change will replace WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," with WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."

TS 5.6.5 Section b lists the analytical methods used to determine the core operating limits. The proposed change would result in the following revisions to the list of analytical methods:

Current TS 5.6.5, Section b, Item 10 in the list of analytical methods, WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," would be deleted and replaced by WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON."

In addition, new Item 11, WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology," would be added to the list of analytical methods in TS 5.6.5 Section b.

The proposed TS changes are contained in the marked-up TS pages provided in Attachment 1 to the Enclosure. The proposed retyped TS pages are provided in Attachment 2 to the Enclosure.

3. TECHNICAL EVALUATION BEACON Methodology

System Description

DCPP employs two methods for performing core power distribution calculations.

Either BEACON or the Movable lncore Detector System (MIDS) can be used to 3

Enclosure PG&E Letter DCL-15-04 7 satisfy core power peaking TS Surveillance Requirements. The BEACON system also provides an on-line monitoring of the reactor core using current plant instrumentation. Excore neutron detectors and core exit thermocouples are used with a three dimensional (3-D) calculated power distribution on a nearly continuous basis. The generation of 3-D power distribution involves periodic calibration of BEACON using the MIDS, a 3-D nodal simulation of the core, and frequent processing of excore neutron detector and core exit thermocouple readings. Power distributions can be compared to TS limits for surveillance purposes.

The BEACON system also allows processing of MIDS flux map data. Results can be compared to TS limits for surveillance purposes. BEACON can provide core depletion and fuel isotopic distribution information by calculating 3-D power distributions and associated neutron flux distributions. BEACON can be used to provide core reactivity calculations such as estimated critical conditions and shutdown margin calculations in order to meet TS requirements and can provide load follow simulations.

Current Licensing Basis and Current Licensing Basis Acceptance Criteria In March 2004, the NRC approved the License Amendments for DCPP Units 1 and 2 to use BEACON PDMS (Reference 6). These amendments included revision of applicable TS and included the original BEACON WCAP-124 72-P-A (Reference 7) as a methodology referenced in the TS Bases.

In January 2013, the NRC approved the License Amendments for DCPP Units 1 and 2 to revise UFSAR Section 4.3.2.2, "Power Distribution," to allow the use of the Westinghouse BEACON Power Distribution Monitoring System methodology (Reference 8) as described in WCAP-12472-P-A, Addendum 1-A, "BEACON Core Monitoring and Operations Support System," (Reference 1).

Therefore, the current DCPP licensing basis for BEACON methodology is the original BEACON WCAP-12472-P-A as approved for DCPP by the NRC in Reference 6 and WCAP-12472-P-A, Addendum 1-A, which was approved for DCPP by the NRC in Reference 8.

The NRC approval of the original BEACON WCAP-12472-P-A allowed the use of the system at DCPP Units 1 and 2 as described above in the System Description section.

The significant improvement (for DCPP Units 1 and 2) to the BEACON system introduced in WCAP-124 72-P-A, Addendum 1-A was the use of a three dimensional advanced nodal code (ANC) neutronic model (specifically the PHOENIX-P/ANC methodology (Reference 3)) that provided increased accuracy in 3-D power distribution calculations. The other improvement addressed in WCAP-12472-P-A, Addendum 1-A, (i.e., regarding the use of Fixed lncore 4

Enclosure PG&E Letter DCL-15-04 7 Detectors (FIDs)), is not applicable to DCPP because DCPP is designed with a MIDS.

Compliance with the NRC licensing conditions associated with adopting the original BEACON WCAP and Addendum 1-A of that WCAP was addressed in the DCPP License Amendment Request (LAR) 12-01 dated January 5, 2012 (Reference 9), and approved by the NRC in Reference 8. The change proposed in this LAR (the adoption ofWCAP-12472-P-A, Addendum 4, Revision 0) does not impact DCPP compliance with the NRC safety evaluation conditions associated with the approval of the original BEACON WCAP-12472 and WCAP-12472-P-A, Addendum 1-A.

The proposed change to WCAP-12472-P-A, Addendum 4, for DCPP Units 1 and 2 would provide a further refinement to the BEACON PDMS methodology currently in use at DCPP Units 1 and 2 and confirm the continued use of the associated NRC-approved Westinghouse methodologies. The purpose of Addendum 4 to WCAP-12472-P-A is the following:

1) Provide the information needed to review and approve the updated thermocouple uncertainty analysis process that will be applied in the BEACON on-line core monitoring system.
2) Affirm the continued use of the NRC-approved Westinghouse design model methodology, currently PHOENIX-P/ANC, PARAGON/ANC, and NEXUS/ANC, in the BEACON system.
3) Establish that the uncertainties applied to power distribution monitoring using FIDs continue to be valid using higher order polynomial fits of the detector measurement variability and the fraction of inoperable detectors.

Note that item 3 above is not applicable to DCPP Unit 1 or 2, since the DCPP design utilizes a MIDS and does not use FIDs.

Regarding Item 2 above, DCPP Units 1 and 2 currently use the NRC-approved PHOENIX-P/ANC methodology (Reference 3) in the BEACON system. As discussed above, the PHOENIX-P/ANC methodology was approved by the NRC in Addendum 1-A to WCAP-12472-P-A. However, in the NRC Final Safety Evaluation (Reference 10) for Addendum 4 to WCAP-124 72-P-A (discussed below), the PARAGON/ANC (Reference 4), and NEXUS/ANC (Reference 5) methodologies were all found acceptable for use in the BEACON system.

Therefore, upon approval of this proposed change, the NRC-approved PARAGON/ANC and NEXUS/ANC methodologies will be used by DCPP Units 1 and 2 in the BEACON System in place of the older PHOENIX-P/ANC methodology.

5

Enclosure PG&E Letter DCL-15-047 In addition, one of the purposes of WCAP-12472-P-A, Addendum 4 (as described in Item 1 above) was to update the thermocouple uncertainty analysis process. A significant component in the BEACON PDMS is the methodology to apply uncertainties to the BEACON measured powers. The measured power uncertainty methodology for BEACON is described in Section 5 of WCAP-12472-P-A for plants using MIDs. A component of the measurement uncertainty is the variability of the thermocouple calibration factors.

The uncertainty methodology described in Section 4 ofWCAP-12472-P-A is based on the average thermocouple deviation at hot full power, being determined from the past performance of the thermocouples. This approach results in some limitations in determining accurate thermocouple uncertainties. When evaluating thermocouple performance from the previous cycle, the typical changes to the hardware from a refueling outage to operate in the current cycle are not considered. These changes include disconnection of the thermocouple electrical connectors, repair of a thermocouple, or repair of damage to the detectors. Any of these possibilities can lead to a change in the thermocouple response signal and statistical behavior which can cause inaccuracies in the BEACON on-line monitoring of the power distribution. While the overall behavior and characteristics of the thermocouple set would remain applicable, some individual thermocouples may have changed behavior.

To address these issues the updated thermocouple evaluation process uses thermocouple temperature and power data from the current cycle collected during the initial startup power ascension following the refueling. The analysis of the thermocouple mixing factors is performed as described in Section 4 of WCAP-124 72-P-A. Any planned or unplanned changes to the characteristics of the thermocouple behavior are in place and measured during the initial power ascension.

The updated BEACON method of analyzing the thermocouple mixing factor data is unchanged from the licensed method described in the NRC-approved TR WCAP-124 72-P-A. What has changed in the update is the use of current plant/cycle thermocouple data in the analysis to generate a plant/cycle specific power dependent thermocouple deviation function that replaces the function defined in Equation 4-4 ofWCAP-12472-P-A.

In the Final Safety Evaluation (Reference 10) associated with Addendum 4 to WCAP-12472-P-A the NRC concluded:

"The NRC staff has reviewed the Westinghouse submittal TR WCAP-12472-P/WCAP-12472-NP, Addendum 4, Revision 0, and found the updated thermocouple methodology, the use of approved Westinghouse design model methodology, and the use of higher order polynomial fits for FlO uncertainties provided in the TR acceptable. The basis for 6

Enclosure PG&E Letter DCL-15-047 acceptance is due to the provided qualitative and quantitative technical material contained in the TR."

The NRC Final Safety Evaluation for Addendum 4 to WCAP-12472-P-A contained no additional license conditions.

System Safety Analysis Basis The BEACON PDMS methodology contained in Westinghouse TR WCAP-12472-P-A, Addendum 4, Revision 0, that is proposed to be included in the UFSAR and TS Bases, has already been approved by the NRC.

The BEACON system is not used to control the performance of any plant equipment. No changes are being made to the UFSAR accident analyses, and no TS limits or surveillance frequencies are being changed. The proposed change to the UFSAR would allow the use of NRC-approved updated Westinghouse methodologies (PARAGON/ANC and NEXUS/ANC) in the BEACON system.

  • The proposed change would also allow the NRC-approved updated process for collecting thermocouple data described in WCAP-12472-P-A, Addendum 4, Revision 0, to be used for the BEACON system. These changes would result in more accurate and reliable operation of the BEACON PDMS without impacting the assumptions and limits of the safety analyses described in the DCPP UFSAR.

The proposed change is limited to changing the UFSAR reference for the BEACON PDMS, from WCAP-12472-P-A, Addendum 1-A, to WCAP-12472-P-A, Addendum 4, Revision 0, in the UFSAR and TS Bases.

System Summary/Conclusion The proposed change to revise the UFSAR reference for the BEACON PDMS, from WCAP-12472-P-A, Addendum 1-A, to WCAP-12472-P-A, Addendum 4, Revision 0, would allow the use of updated NRC-approved methodologies (PARAGON/ANC and NEXUS/ANC) to be used for the BEACON PDMS and would update the process for collecting thermocouple data as described above.

The allowance to use the updated methodologies will allow the latest NRC-approved analytical methods to be used in the BEACON PDMS. The updated method for collecting thermocouple data will ensure the most accurate data is used in the BEACON PDMS for each thermocouple. Therefore, the proposed change will not adversely affect the operation or methodology of the BEACON PDMS as currently discussed in the DCPP UFSAR.

PARAGON and NEXUS Methodology The second change proposed in this LAR is the revision of TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)" Section b to replace WCAP-11596-P-A, 7

Enclosure PG&E Letter DCL-15-047 "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," with WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."

The methodology in TS 5.6.5 Section b is used for determining the boron concentration to be included in the COLR for use in TS 3.9.1, "Boron Concentration."

The addition of the analytical methods to TS 5.6.5 by TR number and title, without date, is consistent with Amendments No. 195 and 196 for DCPP Units 1 and 2 (Reference 16), respectively, that added the reference to WCAP-11596-P-A and revised the referenced TRs to be by title and number only.

Amendments No. 195 and 196 adopted TSTF-363, "Revise Topical Report References in ITS 5.6.5, COLR," and the NRC concluded in the safety evaluation that the proposed change to only list the NRC-approved methodology by TR number and title is acceptable. Additionally, in a letter from the NRC to the Technical Specification Task Force (Reference 17) the NRC indicated that the NRC Staff does not intend to backfit licensees that have Travelers TSTF-363, TSTF-408, or TSTF-419 already in their TSs.

System Description

TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," requires that core operating limits be established prior to each reload cycle or prior to any remaining portion of a reload cycle and documented in the COLR. TS 5.6.5 Section a lists the TS limits included in the COLR. The specific TS limit listed in Section a of TS 5.6.5 affected by the proposed change is Item 8, "Refueling Boron Concentration limits in TS 3.9.1."

The Limiting Condition of Operation (LCO) forTS 3.9.1 ,"Boron Concentration,"

states:

"Boron concentrations of all filled portions of the Reactor Coolant System, the refueling canal, and the refueling cavity, that have direct access to the reactor vessel, shall be maintained within the limit specified in the COLR."

The limit on the boron concentrations of the filled portions of the Reactor Coolant System, the refueling canal, and the refueling cavity, that have direct access to the reactor vessel during refueling ensures that the reactor remains subcritical during Mode 6. Refueling boron concentration is the soluble boron concentration in the coolant in each of these volumes having direct access to the reactor core during refueling.

The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative sample of the coolant in each of the volumes. The refueling boron concentration limit is specified in the COLR. The 8

Enclosure PG&E Letter DCL-15-04 7 refueling boron concentration is sufficient to maintain shutdown margin with the most adverse conditions of fuel assembly and control rod position allowed by plant procedures. The boron concentration that is maintained in Mode 6 is sufficient to maintain keff less than or equal to 0.95 with the most reactive rod control assembly completely removed from its fuel assembly.

Current Licensing Basis and Current Licensing Basis Acceptance Criteria The guidance in NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," dated October 3, 1988, indicates that it is acceptable for licensees to control reactor physics parameter limits by specifying an NRC-approved calculation methodology. These parameter limits may be removed from the TS and placed in a cycle specific COLR that is required to be submitted to the NRC every operating cycle or each time it is revised.

Consistent with the guidance in GL 88-16, DCPP Unit 1 and 2 TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," requires the following:

  • An NRC-approved methodology be used to determine the core operating limits listed in Section a of TS 5.6.5,
  • The specific NRC-approved methodologies used to determine the core operating limits are required to be listed Section b of TS 5.6.5, and
  • The COLR, including any midcycle revisions or supplements, be provided upon issuance for each reload cycle to the NRC is a requirement of Section d of TS 5.6.5.

The current NRC-approved analytical method associated with determining the boron concentration limit forTS 3.9.1 is Item 10 in DCPP Unit 1 and 2 TS 5.6.5 Section b. Item 10 of TS 5.6.5 Section b is WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," (Reference 3), In the conclusions section of the NRC Final Safety Evaluation (Reference 11) for WCAP-11596-P-A (dated May 1988) the NRC states, in part:

"The review has concluded that the qualification process has covered an acceptable range of comparisons to measured experimental or operating reactor data (or in a few cases to other calculations) and that these comparisons demonstrate that the calculations have acceptably small deviations from the measured data, and that the systems have the capability of analyzing relevant PWR lattices and neutronic design problems equal to or superior to presently approved methodologies in these areas. The PHOENIX-P lattice physics methods and the PHOENIX-9

Enclosure PG&E Letter DCL-15-04 7 P/ANC nodal analysis systems described in this report are thus acceptable for use in PWR design analysis."

PHOENIX-P is the neutron transport code traditionally used to provide cross section data as input to the Advanced Nodal Code (ANC). The PARAGON computer code is a standalone neutron transport code based on collision probability techniques, and it is approved for use as a standalone lattice physics code and as cross section generation tool for core simulators, such as ANC, for uranium-fuel pressurized water reactors (PWRs). ANC is a core simulator code system, which performs calculations based on nuclear data supplied by a code such as PARAGON or PHOENIX-P. The PARAGON nuclear data methodology was developed as a direct replacement to PHOENIX-P.

WCAP-16045-P-A, Revision 0, confirms the qualifications of the PARAGON code both as a standalone transport code and as a substitute for the PHOENIX-P code, the code currently used in the DCPP design, as a nuclear data source for nodal codes. As part of the qualification process, WCAP-16045-P-A, Revision 0, includes a comparison of PARAGON predicted values to measured data from several plants. Benchmarking has shown that results from the PARAGON/ANC code package are essentially the same as those obtained from the current PHOENIX-P/ANC system. WCAP-16045-P-A, Revision 0, concludes that the application of PARAGON nuclear data methodology does not result in any undesirable changes in predicted fuel performance or safety analysis results.

The NRC Final Safety Evaluation (Reference 12) for the PARAGON nuclear data methodology (dated March 2004) states, in part:

" ... The staff considers the new PARAGON code to be well qualified as a stand-alone code replacement for the PHOENIX-P lattice code, wherever the PHOENIX-P code is used in NRC-approved methodologies."

WCAP-16045-P-A, Addendum 1-A, Revision 0, "Qualification of the NEXUS Nuclear Data Methodology," is an improvement to the PARAGON computer code. The NEXUS methodology is a re-parameterization of the PARAGON nuclear data output and a new reconstruction approach within the ANC core simulator code to simplify the use of this code system for design use. NEXUS has been implemented in the PARAGON/ANC code system for design use.

Specifically, the NEXUS methodology has been implemented in the parameterization of PARAGON cross sections for input to ANC and also in ANC to reconstruct those cross sections at specific nodal conditions. Since the NEXUS methodology provides a linkage between PARAGON and ANC, establishing a new code system, while still using PARAGON, both WCAP-16045-P-A and WCAP-16045-P-A, Addendum 1-A (without date and revision number) are proposed to be added toTS 5.6.5 Section b.

10

Enclosure PG&E Letter DCL-15-047 WCAP-16045-P-A, Addendum 1-A, Revision 0, verifies the accuracy of NEXUS for cross section representation. As part of this TR, different assembly types were calculated using NEXUS, which include the following: both Westinghouse and Combustion Engineering assembly types; U02 fuel; and integral fuel burnable absorber (IFBA); wet annular burnable absorber (WABA); and Gd 20 3 burnable absorbers. The k-infinity results from these calculations were compared directly to PARAGON k-infinity results at corresponding conditions.

The comparisons demonstrated that the NEXUS cross sections are accurate over the range of temperatures, boron concentrations, and power levels expected to be encountered in PWR core calculations.

The NRC Final Safety Evaluation (Reference 13) for the NEXUS nuclear data methodology (dated February 2007) states, in part:

"The NRC staff has reviewed the TR submitted by Westinghouse and determined that the NEXUS/ANC code system is adequate to replace the PARAGON/ANC code system wherever the latter is used in NRC-approved methodologies. The NRC staff, furthermore, has determined that NEXUS/ANC is qualified as a stand-alone code system so long as its use is limited by the provisions listed in Section 4.0 of this safety evaluation."

The provision listed in Section 4.0 of the NRC Final Safety Evaluation states:

"The NEXUS/ANC code system is limited to uranium-fueled, PWR applications as the only plant data assessments presented were for uranium-fueled, PWRs. While Westinghouse has provided comparisons of the relative performance of PARAGON/ANC and NEXUS/ANC for calculations with MOX fueled, PWR fuel assemblies, the PARAGON/ANC code system was not approved for this purpose. In the absence of actual plant data, NEXUS/ANC has not been approved for MOX applications."

DCPP Unit 1 and Unit 2 TS 4.2.1, "Fuel Assemblies," states, in part:

"The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 2) as fuel material."

Therefore, the provision of Section 4.0 of the NRC Final Safety Evaluation (discussed above) is met by the requirement in the DCPP Units 1 and 2 TS 4.2.1 that the fuel must be composed of U02. A separate licensing action would be required in order for DCPP Units 1 or 2 to use mixed-oxide (MOX) fuel.

Therefore, NEXUS/ANC is acceptable for use at DCPP.

11

Enclosure PG&E Letter DCL-15-047 WCAP-16045-P-A, Addendum 1-A, Revision 0, includes a change to the cross section representation used in the overall nuclear design method. However, the proposed change to the list of NRC-approved analytical methods listed in TS 5.6.5.b involves only the method by which the boron concentration required by TS 3.9.1, "Boron Concentration" is determined. The requirements and Bases forTS 3.9.1 are discussed above in the System Description section for this change.

Therefore, the use of WCAP-16045-P-A, Addendum 1-A, Revision 0, for DCPP does not affect the inputs or method(s) for ensuring core subcriticality, both short and long-term post-Loss of Coolant Accident (LOCA), thereby precluding the potential for return to power following a large break LOCA. Since neither the post-LOCA boron source concentration nor heat generation are impacted by the use of Addendum 1-A, the current emergency operating procedure timing for boric acid precipitation and the action time for switching to simultaneous injection will continue to remain valid. As such, core design specific parameters that are verified each cycle to be conservative with respect to the LOCA inputs and the refueling boron concentration, will continue to be calculated using the NRC-approved methods for those parameters.

Later versions of the Westinghouse ANC code require cross section data, which is generated using the PARAGON neutron transport code. The NEXUS/ANC system is a version of the PARAGON/ANC system in that all nuclear data is based on PARAGON and only the methods of representing this data in ANC have been changed from the version of PARAGON/ANC described in WCAP-16045-P-A, Revision 0. Therefore, this combined PARAGON-NEXUS methodology will be used as a replacement to the PHOENIX-P methodology in TS 5.6.5 Section b. for determining the boron concentration required by TS 3.9.1.

System Safety Analysis Basis The proposed change involves a change in the methodology specified in TS 5.6.5.b for determining the boron concentration to be included in the COLR for use in TS 3.9.1, "Boron Concentration." TS 3.9.1 is applicable in Mode 6 (i.e.,

during refueling operations). During refueling operations, the reactivity condition of the core is consistent with the initial conditions assumed for the boron dilution accident in the accident analysis and is conservative for Mode 6. The boron concentration limit specified in the COLR is based on the core reactivity at the beginning of each fuel cycle (the end of refueling) and includes an uncertainty allowance.

The required boron concentration and the plant refueling procedures that verify the correct fuel loading plan (including core mapping) ensure that the keff of the core will remain :::;; 0.95 during the refueling operation. Hence, at least a 5 percent Llk/k margin of safety is established during refueling .

12

Enclosure PG&E Letter DCL-15-047 Future changes to the values of the operating limits in the COLR are controlled by the 10 CFR 50.59 process, may only be developed using NRC-approved methodologies, and must remain consistent with all applicable plant safety analysis limits addressed in the DCPP UFSAR. As such, the consequences of the design basis accidents will continue to be calculated using NRC accepted methodologies. As such, the proposed change to use the NRC-approved PARAGON-NEXUS methodology will not impact the assumptions used in the safety analyses or the safety analyses acceptance criteria.

System Summary/Conclusion The proposed change to TS 5.6.5 Section b would allow the use of the NRC-approved combined PARAGON-NEXUS methodology. The use of these NRC-approved and updated (from the existing PHOENIX-P) Westinghouse methodologies will continue to assure the boron concentration determined for TS 3.9.1 remains conservative and retains the required margin of safety (i.e., at least a 5 percent Llk/k).

No other COLR parameters would be affected by this change. The other methodologies used to determine operating limits referenced in the COLR remain applicable with the use of the PARAGON-NEXUS methodology for the refueling boron concentration limit.

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria Section 182a of the Atomic Energy Act requires that applicants for nuclear power plant operating licenses include TS as part of the facility operating license. The Commission's regulatory requirements related to the content of TS are set forth in 10 CFR 50.36. Pursuant to 10 CFR 50.36, TS are required to include items in the five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls.

However, the rule does not specify the particular requirements to be included in a plant's TS. Under 10 CFR 50.36(c)(2)(ii), a limiting condition for operation must be included in TS for any item meeting one of the following four criteria:

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

13

Enclosure PG&E Letter DCL-15-04 7 Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Those items that do not fall within or satisfy any of the above criteria do not need to be included in Section 3 of the TS.

With regard to the first proposed change to update the BEACON PDMS methodology referenced in the UFSAR from the methodology contained in WCAP-12472-P-A, Addendum 1-A, "BEACON Core Monitoring and Operations Support System," to the Westinghouse BEACON PDMS methodology contained in WCAP-124 72-P-A, Addendum 4, Revision 0, "BEACON Core Monitoring and Operation Support System" the PDMS instrumentation does not meet any of the criteria of 10 CFR 50.36(c)(2)(ii) for inclusion in the TS. PG&E has included the PDMS instrumentation requirements in equipment control guidelines (ECGs). The ECGs are plant-specific administrative controls, similar to TS controls, but ECGs are controlled by PG&E in accordance with 10 CFR 50.59.

The second proposed change would revise TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," Section b to replace WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," with WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology." The proposed updated and NRC-approved methodologies would be included in the Administrative Controls section of the TS and would be used to determine a core operating limit. The use of the proposed NRC-approved methodologies would continue to assure that the plant is operated in a safe manner. As such, the proposed change would be consistent with the Administrative Controls requirement of 10 CFR 50.36(c)(5).

14

Enclosure PG&E Letter DCL-15-047 Furthermore, with regard to the second proposed change, the guidance in NRC Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications (Generic Letter 88-16)," indicates that it is acceptable to control reactor physics parameter limits by specifying the calculation methodology. The generic letter indicates that such parameter limits may be removed from TS and placed in a cycle-specific COLR. The COLR is defined in Section 1.1 of the TS and the reporting requirements in TS 5.6.5 require that a COLR be submitted to the NRC each operating cycle, or each time the COLR is revised. The generic letter also recommended that the TS include a list of references for NRC-approved methodologies that are used to generate the cycle-specific core operating limits. TS 5.6.5 Section b identifies the NRC-approved analytical methods used to determine the core operating limits for DCPP Units 1 and 2. The guidance in the generic letter continues to be met since the proposed change will continue to specify the NRC-approved methodologies used to determine the core operating limits.

4.2 Precedent For the proposed change that would revise TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," Section b to replace WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," with WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology," there are two precedents.

a) On July 17, 2013, the NRC issued Amendment No. 191 to Renewed Facility Operating License No. NPF-2 and Amendment No. 187 to Renewed Facility Operating License No. NPF-8 (Reference 14) for the Joseph M. Farley Nuclear Plant, Units 1 and 2, respectively. This amendment adds a reference to WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology," to TS Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)."

b) On August 7, 2014, the NRC issued Amendment No. 209 to Renewed Facility Operating License No. NPF-42 (Reference 15) for the Wolf Creek Generating Station. The amendment revises TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," to replace the methodology of TR WCAP-11596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," with WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and WCAP-16045-P-A, 15

Enclosure PG&E Letter DCL-15-047 Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology," to determine core operating limits. The reference to the TRs WCAP-16045-P-A and WCAP-16045-P-A, Addendum 1-A in TS 5.6.5 was without date and revision number.

4.3 Significant Hazards Consideration PG&E has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change would revise the Updated Final Safety Analysis Report (UFSAR) to allow the use of the BEACON code methodology contained in the NRC-approved WCAP-12472-P-A, Addendum 4, Revision 0, instead of the BEACON methodology contained in NRC-approved WCAP-12472-P-A, Addendum 1-A. In addition, the proposed change would revise Technical Specification (TS) 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," Section b to replace WCAP-11596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," with NRC-approved WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and NRC-approved WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology," in the list of NRC-approved analytical limits used to determine core operating limits.

Specifically the limit for refueling boron concentration (i.e., the shutdown margin) required by TS 3.9.1, "Boron Concentration."

The changes to the BEACON system and TS 5.6.5 core operating limits methodologies, which this license amendment proposes, are improvements over the current methodologies in use at the Diablo Canyon Power Plant (DCPP). The NRC staff reviewed and approved these methodologies and concluded that these analytical methods are acceptable as a replacement for the current analytical methods. Thus the BEACON system operation to perform power distribution calculations and the core operating limits determined using the proposed analytical methods will continue to assure that the plant operates in a safe manner and, thus, the proposed changes do not involve an increase in the probability of an accident.

16

Enclosure PG&E Letter DCL-15-047 The BEACON system power distribution calculations and the core operating limits determined by use of the proposed new methodologies will not increase the reactor power level or the core fission product inventory, and will not change any transport assumptions or the shutdown margin requirements of the TS. In addition, the proposed changes will not alter any accident analyses assumptions discussed in the UFSAR. As such, the DCPP will continue to operate within the power distribution limits and shutdown margins required by the plant TS and within the assumptions of the safety analyses described in the UFSAR. As such, the proposed changes do not involve a significant increase in the consequences of an accident.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different accident from any accident previously evaluated?

Response: No.

The proposed change involves the use of new and NRC-approved

  • methodologies used by the BEACON System to perform core power distribution calculations and in TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," to determine core operating limits (i.e., refueling boron concentration or shutdown margin requirement).

The proposed change provides revised analytical methods for the BEACON system and determining core operating limit for refueling boron concentration, and does not change any system functions or maintenance activities. The change does not involve physical alteration of the plant, that is, no new or different type of equipment will be installed. The change does not alter assumptions made in the safety analyses and continues to assure the plant is operated within safe limits. This change does not create new failure modes or mechanisms that are not identifiable during testing, and no new accident precursors are generated.

The BEACON system is not used to control the performance of any plant equipment. The BEACON system core power distribution calculations and core operating limits developed using the new methodologies will be determined using NRC-approved methodologies, and will remain consistent with all applicable plant safety analysis limits addressed in the DCPP UFSAR and the shutdown margin requirements of the TS. As such, use of the new BEACON and COLR methodologies will not cause a new or different accident.

17

Enclosure PG&E Letter DCL-15-047 Therefore, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated.

The proposed changes do not physically alter safety-related systems, nor does it affect the way in which safety related systems perform their functions. The setpoints at which protective actions are initiated are not altered by the proposed changes. Therefore, sufficient equipment remains available to actuate upon demand for the purpose of mitigating an analyzed event. The proposed methodology changes are an improvement that will allow more accurate modeling of core performance and determination of the required refueling boron concentration. The NRC has reviewed and approved these methodologies for their intended use in lieu of the current methodologies; thus, the margin of safety is not reduced due to this change.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, PG&E concludes that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5. ENVIRONMENTAL CONSIDERATION PG&E has evaluated the proposed amendment and has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed 18

Enclosure PG&E Letter DCL-15-047 amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6. REFERENCES
1. WCAP-:-12472-P-A, Addendum 1-A, "BEACON Core Monitoring and Operations Support System," January 2000
2. WCAP-12472-P-A, Addendum 4, Revision 0, "BEACON Core Monitoring and Operation Support System," September 2012
3. WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988
4. WCAP-16045-P-A, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004
5. WCAP-16045-P-A, Addendum 1-A, Revision 0, "Qualification of the NEXUS Nuclear Data Methodology," August 2007
6. NRC Issuance of License Amendment Nos. 164 and 166 to Facility Operating License Numbers 80 and 82 for Diablo Canyon Power Plant, Units 1 and 2, respectively, March 31, 2004
7. WCAP-12472-P-A, "BEACON, Core Monitoring and Operations Support System," August 1994
8. NRC Issuance of License Amendment Nos. 214 and 216 to Facility Operating License Numbers 80 and 82 for Diablo Canyon Power Plant, Units 1 and 2, respectively, January 9, 2013
9. PG&E Letter DCL-12-002, License Amendment Request 12-01, "Revision to the Updated Final Safety Analysis Report Section 4.3.2.2,

'Power Distribution;' and Technical Specification Bases 3.1.7, 'Rod Position Indication;' 3.2.1, 'Heat Flux Hot Channel Factor (F 0 (Z)),'

3.2.2, 'Nuclear Enthalpy Rise Hot Channel Factor' (F~H); 3.2.4,

'Quadrant Power Tilt Ratio' (QPTR); and 3.3.1, 'Reactor Trip System Instrumentation (RTS),"' January 5, 2012

10. Letter from S. Bahadur (US NRC) to J. A. Gresham (Westinghouse),

"Final Safety Evaluation for Westinghouse Electric Company (Westinghouse) Topical Report WCAP-12472-P/WCAP-12472-NP, Addendum 4, 'BEACON' Core Monitoring and Operation Support System, Addendum 4,' (TAC No. ME5240)," dated August 9, 2012

11. Letter from Ashok C. Thadani (USNRC) toW. J. Johnson (Westinghouse), "Acceptance of Referencing of The Westinghouse Topical Report WCAP-11596, 'Qualification Of The Phoenix-P/ANC 19

Enclosure PG&E Letter DCL-15-047 Nuclear Design System for Pressurized Water Reactor Cores,"'

May 17, 1988

12. Letter from H. N. Berkow (NRC) to J. A. Gresham (Westinghouse),

"Final Safety Evaluation for Westinghouse Topical Report, WCAP-16045-P, Revision 0, 'Qualification of the Two-Dimensional Transport Code PARAGON', (TAC No. MB8040)," March 18, 2004

13. Letter from H. K. Nieh (NRC) to J. A. Gresham (Westinghouse), "Final Safety Evaluation for Westinghouse Electric Company (Westinghouse)

Topical Report (TR) WCAP-16045-P-A, Addendum 1, 'Qualification of the NEXUS Nuclear Data Methodology' (TAC No. MC9606)," February 23,2007

14. NRC letter from E. Brown to C. R. Pierce, "Joseph M. Farley Nuclear Plant, Units 1 and 2, Issuance of Amendments Regarding Changes to Nuclear Methodology References (TAC NOS. ME9244 and ME9245)(NL-12-1226)," July 17, 2013
15. NRC Letter from Carl F. Lyon to Adam C. Heflin, "Wolf Creek Generating Station- Issuance of Amendment Re: Replace A Methodology In Technical Specification 5.6.5, 'Core Operating Limits Report (COLR)' (TAC No. MF2790)," August 7, 2014
16. NRC Issuance of License Amendment Nos. 195 and 196 to Facility Operating License Numbers 80 and 82 for Diablo Canyon Power Plant, Units 1 and 2, respectively, April17, 2007
17. NRC letter from J. R. Jolicoeur to TSTF, "Implementation Of Travelers TSTF-363, Revision 0, 'Revise Topical Report References ITS 5.6.5 ,

COLR [Core Operating Limits Report],' TSTF-408, Revision 1,

'Relocation of LTOP [Low Temperature Overpressure Protection]

Enable Temperature and PORV [Power Operated Relief Valve] Lift Setting to the PTLR [Pressure-temperature Limits Report], AND TSTF-419,' Revision 0, 'Revise PTLR Definition and References in ISTS [Improved Standard Technical Specification] 5.6.6, RCS [Reactor Coolant System] PTLR,"' August 4, 2011 20

Enclosure Attachment 1 PG&E Letter DCL-15-047 Technical Specification Markups

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1. WCAP-1 0216-P-A, Relaxation of Constant Axial Offset Control Fa Surveillance Technical Specification, (Westinghouse Proprietary),
2. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, (Westinghouse Proprietary),
3. WCAP-8385, Power Distribution Control and Load Following Procedures, (Westinghouse Proprietary),
4. WCAP-1 0054-P-A, Westinghouse Small Break LOCA ECCS Evaluation Model Using the NOTRUMP Code, August 1985 (Westinghouse Proprietary),
5. WCAP-1 0054-P-A, Addendum 2, Revision 1, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI WCAP-16045-P-A, Condensation Model, July 1997 (Westinghouse Proprietary),

"Qualification of the 6. WCAP-12945-P-A, Westinghouse Code Qualification Document for Two-Dimensional Best-Estimate Loss of Coolant Analysis, June 1996 (Westinghouse Transport Code Proprietary),

PARAGON," and 7. WCAP-12945-P-A, Addendum 1-A, Revision 0, "Method for Satisfying 10 CFR 50.46 Reanalysis Requirements for Best Estimate LOCA Evaluation Models," December 2004. (Westinghouse Proprietary) (Unit 1 Only),

WCAP-16009-P-A, Revision 0, Realistic Large-Break LOCA Evaluation 11 . WCAP-16045-P-A, Methodology Using the Automated Statistical Treatment of Uncertainty Addendum 1-A, Method (ASTRUM), January 2005. (Westinghouse Proprietary)

"Qualification of the (Unit 2 Only),

NEXUS Nuclear Data WCAP-8567-P-A, "Improved Thermal Design Procedure," a+lG Methodology."

\NCAR 11596 P A, "Qualification of the PHOENIX P/ANC Nuclear Design System for Pressurized VVater Reactor Cores."

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Move 5.6.5.c Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient and 5.6.5.d to analysis limits, and accident analysis limits) of the safety analysis are met.

Page 5.0-21

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-20 Unit 1 - Amendment No. ~,~,W+,+%,19g ' ....EJ I r1 Unit 2 - Amendment No. ~,+de,.:t-9a,+W,~ T'-

5.6.5 CORE OPERATING LIMITS Reporting Requirements REPORT (COLR) 5.6 (continued) 5.6 Reporting Requirements (continued) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, Low Temperature Overpressure Insert 5.6.5.c and Protection (L TOP) arming, and PORV lift settings as well as heatup and 5.6.5.d from Page cooldown rates shall be established and documented in the PTLR for the 5.0-20 following:
1. Specification 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and
2. Specification 3.4.12, "Low Temperature Overpressure Protection (L TOP)

System."

(continued)

DIABLO CANYON - UNITS 1 & 2 Rev 29 Page 22 of 27 5.0-21 Unit1 -Amendment No. +3§, +98 Unit 2 -Amendment No.~' +99

f. I rJ j"\-

Tab_5!0u3r29.DOC 0320.1506

Enclosure Attachment 2 PG&E Letter DCL-15-047 Technical Specification Retyped Pages Remove Page Insert Page 5.0-20 5.0-20 5.0-21 5.0-21

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1. WCAP-10216-P-A, Relaxation of Constant Axial Offset Control F0 Surveillance Technical Specification, (Westinghouse Proprietary),
2. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, (Westinghouse Proprietary),
3. WCAP-8385, Power Distribution Control and Load Following Procedures, (Westinghouse Proprietary),
4. WCAP-1 0054-P-A, Westinghouse Small Break LOCA ECCS Evaluation Model Using the NOTRUMP Code, August 1985 (Westinghouse Proprietary),
5. WCAP-1 0054-P-A, Addendum 2, Revision 1, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI Condensation Model, July 1997 (Westinghouse Proprietary),
6. WCAP-12945-P-A, Westinghouse Code Qualification Document for Best-Estimate Loss of Coolant Analysis, June 1996 (Westinghouse Proprietary),
7. WCAP-12945-P-A, Addendum 1-A, Revision 0, "Method for Satisfying 10 CFR 50.46 Reanalysis Requirements for Best Estimate LOCA Evaluation Models," December 2004. (Westinghouse Proprietary) (Unit 1 Only),
8. WCAP-16009-P-A, Revision 0, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005. (Westinghouse Proprietary)

(Unit 2 Only),

9. WCAP-8567 -P-A, "Improved Thermal Design Procedure,"
10. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and
11. WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-20 Unit 1 -Amendment No. 4-Ja,4-Je,4-9-:t-,4-%,4W, Unit 2 -Amendment No. 4-Ja,4-Je,~,4-00,4-99,

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provid~d upon issuance for each reload cycle to the NRC.

5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, Low Temperature Overpressure Protection (L TOP) arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
1. Specification 3.4.3, "RCS Pressure and Temperature (PIT) Limits," and
2. Specification 3.4.12, "Low Temperature Overpressure Protection (L TOP)

System."

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-21 Unit 1 -Amendment No. ~' W, Unit 2 -Amendment No. ~' 99,

Enclosure Attachment 3 PG&E Letter DCL-15-047 Updated Final Safety Analysis Report Markups (for information only)

DCPP UNITS 1 & 2 FSAR UPDATE

30. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 (Westinghouse Proprietary).
31. Kersting, P. J., et al., Assessment of Clad Flattening and Densification Power Spike Factor Elimination in Westinghouse Nuclear Fuel, WCAP-13589-A (Proprietary), March 1995, and WCAP-14297 -A (Non-Proprietary), March 1995.
32. WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994 (Westinghouse Proprietary).

33.

WCAP-12472-P-A, Addendum 4, Revision 0, BEACON Core Monitoring and Operations Support System, September 2012.

4.3-40 Revision 21 September 2013

Enclosure Attachment 4 PG&E Letter DCL-15-04 7 Technical Specification Bases Markups (for information only)

Rod Position Indication B 3.1.7 BASES (continued)

SURVEILLANCE SR 3.1.7.1 REQUIREMENTS Verification that the DRPI agrees with the demand position within 12 steps ensures that the DRPI is operating correctly. Verification at 24, 48, 120, and 228 steps withdrawn for the control and shutdown banks provides assurance that the DRPI is operating correctly over the full range of indication.

This surveillance is performed prior to reactor criticality after each removal of the reactor head, since there is potential for unnecessary plant transients if the SR were performed with the reactor at power.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 13.

2. FSAR, Chapter 15.
3. WCAP-1 0216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control and F0 Surveillance Technical Specification,"

February 1994.

4. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

I WCAP-12472-P-A, Addendum 4, Revision 0, "BEACON Core Monitoring and Operations Support System," September 2012.

DIABLO CANYON - UNITS 1 & 2

Fa(Z)

B 3.2.1 BASES SURVEILLANCE SR 3.2.1.2 (continued)

REQUIREMENTS prevent Fa(Z) from exceeding its limit for any significant period of time without detection. Performing the Surveillance in MODE 1 prior to exceeding 75o/o RTP or at a reduced power at any other time, and meeting the 100°/o RTP Fa(Z) limit, provides assurance that the Fa(Z) limit will be met when RTP is achieved, because peaking factors are generally decreased as power level is increased.

Fa(Z) is verified at power levels;:;: 20°/o RTP above the THERMAL POWER of its last verification, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions to ensure that Fa(Z) is within its limit at higher power levels.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50.46, 1974.

2. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.
3. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

'AlGAR 12472 P A, Addendum 1 A, "BEACON Core Monitoring and Operations Support System," January 2000.

WCAP-12472-P-A, Addendum 4, Revision 0, "BEACON Core Monitoring and Operations Support System," September 2012.

DIABLO CANYON - UNITS 1 & 2

F~H B 3.2.2 BASES SURVEILLANCE SR 3.2.2.1 (continued)

REQUIREMENTS After each refueling, F~H must be determined in MODE 1 prior to exceeding 75°/o RTP . This requirement ensures that F~H limits are met at the beginning of each fuel cycle. Performing this Surveillance in MODE 1 prior to exceeding 75°/o RTP, or at a reduced power level at any other time, and meeting the 100°/o RTP F~H limit, provides assurance that the F~H limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased .

The Surveillance Frequency is based on operating experience ,

equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. Regulatory Guide 1.77, Rev. 0, May 1974.

2. 10 CFR 50, Appendix A, GDC 26.
3. 10 CFR 50.46.
4. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

l l Operations Support System," January 2000.

WCAP-12472-P-A, Addendum 4, Revision 0, "BEACON Core Monitoring and Operations Support System ," September 2012.

DIABLO CANYON - UNITS 1 & 2

QPTR B 3.2.4 BASES SURVEILLANCE SR 3.2.4.2 (continued)

REQUIREMENTS With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

For purposes of monitoring the QPTR when one or more power range channels are inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is performed with a full incore flux map or two sets of four thimble locations with quarter core symmetry. The two sets of four symmetric thimbles is a set of eight unique detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.

The symmetric thimble flux map can be used to generate symmetric thimble "tilt." This can be compared to a reference symmetric thimble tilt, from the most recent full core flux map, to generate an incore QPTR. Therefore, incore QPTR can be used to confirm that QPTR is within limits.

With one NIS channel inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. To confirm that no change in tilt has actually occurred, which might cause the QPTR limit to be exceeded, the incore tilt result may be compared against previous tilt values either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surveillance should be within 2°/o of the tilt shown by the most recent power distribution measurement data.

REFERENCES 1. 10 CFR 50.46.

2. Regulatory Guide 1.77, Rev 0, May 1974.
3. 10 CFR 50, Appendix A, GDC 26.
4. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

Operations Support System," January 2000.

WCAP-12472-P-A, Addendum 4, Revision 0, "BEACON Core Monitoring and Operations Support System," September 2012.

DIABLO CANYON - UNITS 1 & 2

RTS Instrumentation B 3.3~ 1 BASES REFERENCES 17. WCAP-11 082, "Westinghouse Setpoint Methodology for (continued) Protection Systems, Diablo Canyon Units 1 and 2, 24 Month Fuel Cycle Evaluation and Replacement Steam Generator,"

September 2007.

18. NSP-1-20-13F Unit 1 "Turbine Auto Stop Low Oil Pressure."
19. NSP-2-20-13F Unit 2 "Turbine Auto Stop Low Oil Pressure."
20. J-11 0 "24 Month Fuel Cycle Allowable Value Determination I Documentation and ITDP Uncertainty Sensitivity."
21. IEEE Std. 338-1977.
22. License Amendment 61/60, May 23, 1991.
23. Westinghouse Technical Bulletin ESBU-TB-92-14-R1, "Decalibration Effects of Calorimetric Power Measurements on the NIS High Power Reactor Trip at Power Levels less than 70°/o RTP ," dated February 6, 1996.
24. DCPP NSSS Calculation N-212, Revision 1.
25. License Amendments 157/157, June 2, 2003.
26. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.
27. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," October 1998.
28. WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," October 1998.
29. WCAP-15376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," March 2003.
30. WCAP-11394-P-A, "Methodology For The Analysis of the Dropped Rod Event," January, 1990
31. License Amendments 205/206, April 29, 2009
32. WCAP-16769-P Revision 1, "Westinghouse SSPS Universal Logic Board Replacement Summary Report 6D30225G01/G02/G03/G04," July 2008.

and Operations Support System," January 2000.

DCL 85-161, "Unit 2 Technical Specifications- Additional Information," April 22, 1985 WCAP-12472-P-A, Addendum 4, Revision 0, "BEACON Core Monitoring and Operations Support System," September 2012.

DIABLO CANYON - UNITS 1 & 2

Retrieved from "https://kanterella.com/w/index.php?title=DCL-15-047,_License_Amendment_Request_15-02_Revision_of_Updated_Final_Safety_Analysis_Report_for_Beacon_Power_Distribution_Monitoring_System_Methodology_and_Technical_Specification_5.6.5._Core_Operating_Limits_Report_(Colr),_For...&oldid=2449052"