RA-19-0134, Duke Energy Response to NRC Request for Additional Information (RAI) Related to Oconee License Amendment Request 2018-05

Duke Energy Response to NRC Request for Additional Information (RAI) Related to Oconee License Amendment Request 2018-05

ML19066A316
Person / Time
Site:	Oconee
Issue date:	03/07/2019
From:	Burchfield J Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-19-0134
Download: ML19066A316 (8)

Text

J. Ed Burchfield, Jr.

Vice President Oconee Nuclear Station Duke Energy ON01VP l 7800 Rochester Hwy Seneca, SC 29672 o: 864.873.3478 f: 864.873.5791 Ed.Burchfield@duke-energy.com RA-19-0134 March 7, 2019 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Duke Energy Carolinas, LLC (Duke Energy)

Oconee Nuclear Station (ONS), Units 1, 2, and 3 Docket Numbers 50-269, 50-270, 50-287 Renewed License Numbers DPR-38, DPR-47, DPR-55

Subject:

Duke Energy Response to NRC Request for Additional Information (RAI) Related to Oconee License Amendment Request 2018-05

References:

1. Duke Energy Letter to USNRC, License Amendment Request Proposing a Revised Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods that Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 11; License Amendment Request No. 2018-05, dated November 1, 2018 (ML18318A320).

2. NRC Email, A. Klett (NRC) to A. Zaremba (Duke Energy), NRC Request for Additional Information for Oconee LAR 2018-05 (L-2018-LLA-0300), dated February 19, 2019.

By letter dated November 1, 2018, Duke Energy submitted a License Amendment Request (LAR) to the U.S. Nuclear Regulatory Commission (NRC) proposing to revise the facility as described in the Oconee Updated Final Safety Analysis Report with respect to the fission gas gap release fractions calculated for high burnup fuel rods. By email dated February 19, 2019 (Reference 2), the NRC requested additional information associated with the Reference 1 LAR. The NRC request for additional information (RAI) and the Duke Energy responses are provided in the enclosure to this letter.

The responses to the RAIs do not affect the conclusions of the No Significant Hazards Consideration provided in the Reference 1 LAR.

RA-19-0134 March 7, 2019 This letter contains no new or revised commitments. Should you have any questions regarding this submittal, please contact Mr. Art Zaremba, Fleet Nuclear Licensing Manager, at (980) 373-2062.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March 7, 2019.

Sincerely,

~[)~/

J. Ed Burchfield, Jr.

Vice President Oconee Nuclear Station

Enclosure:

Duke Energy Response to NRC Request for Additional Information (RAI)

Page2

RA-19-0134 March 7, 2019 Cc (w/enclosure):

Ms. Catherine Haney, Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Ms. Audrey Klett, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8G9A 11555 Rockville Pike Rockville, Maryland 20852 Mr. Eddy Crowe NRC Senior Resident Inspector Oconee Nuclear Station Page 3

Enclosure to Duke Energy Letter RA-19-0134 Page 1 of 3 Duke Energy Response to NRC Request for Additional Information (RAI)

By application dated November 1, 2018 (Agencywide Documents Access and Management System Accession Numbers ML18318A320), Duke Energy Carolinas, LLC (the licensee) proposed changes to the licensing basis for Oconee Nuclear Station, Units 1, 2, and 3 (Oconee). The licensee proposed to revise the facility as described in the updated final safety analysis report (UFSAR) to provide gap release fractions for high-burnup fuel rods (i.e., greater than 54 gigawatt days per metric ton of uranium (GWD/MTU)) that exceed the 6.3 kilowatt per foot (kW/ft) linear heat generation rate (LHGR) limit detailed in Table 3 of Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000.

During the U.S. Nuclear Regulatory Commission (NRC) staffs review the NRC staff determined that more information was needed to complete the review. The staff emailed a draft request for additional information (RAI) to the licensee on February 11, 2019. Based on an email from Mr. Chris Wasik of the licensees staff dated February 12, 2019, the staff determined that it needed to clarify the RAI. The email correspondence is in ADAMS at Accession No. ML19050A220. As discussed with Mr. Chris Wasik on February 14, 2019, the staff is requesting the licensee to respond to this request within 30 days of receipt.

Regulatory Analysis Basis Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.67, Accident source term, Paragraph (b)(2) states, in part, that the NRC may issue the amendment only if the applicants analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) TEDE.

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.

RG 1.183 provides the methodology for analyzing the radiological consequences of several design basis accidents to show compliance with 10 CFR 50.67.

Enclosure to Duke Energy Letter RA-19-0134 Page 2 of 3 10 CFR 50.36, Technical specifications, in part, requires that the technical specifications (TSs) be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto and includes items in following categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notifications; and (8) written reports.

Technical Basis for RAI-1 In its application, the licensee proposed gap release fractions for high-burnup fuel rods (i.e.,

greater than 54 GWD/MTU) that exceed the 6.3 kW/ft LHGR limit in Footnote 11 of Table 3, Non-LOCA [loss of coolant accident] Fraction of Fission Product Inventory in Gap, in RG 1.183.

The non-LOCA gap fractions stated in Table 3 of RG 1.183 are applied to the non-LOCA accidents if fuel failure occurs during the accident. The following accidents at Oconee assume fuel failure: fuel handling accident, fuel cask handling accident, locked rotor accident, and control rod ejection accident. In its application, the licensee stated that no non-LOCA accidents that may result in departure from nucleate boiling are considered (e.g., locked rotor accident or rod ejection accident) because the fuel cycles for Oconee are designed so that no fuel rod predicted to enter departure from nucleate boiling will have been operated beyond the current limit in RG 1.183, Footnote 11 for maximum LHGR. Because of this, the NRC would be approving a change to the gap fractions for only the FHA. However, the application does not incorporate this new design requirement into the licensing basis as reflected in the UFSAR, nor does it place a requirement in Oconees TSs, such as Section 5.0, Design Features, or any other document controlled under 10 CFR 50.59, such as the Core Operating Limits Report (COLR).

The UFSAR currently states that the fuel cycle design ensure that none of these fuel pins experience DNB following any design basis accident. This sentence refers to the non-DNB fuel pins that exceed the rod power/burnup criteria of Footnote 11 in RG 1.183 and the new proposed gap fractions provided in the application, which apply to FHA only. The application does not explicitly state that the fuel cycle design ensures that no fuel rod predicted to experience DNB in any other non-LOCA accidents will have operated beyond the power/burnup criteria of Footnote 11 in Regulatory Guide 1.183 and that the gap fractions used in these non-LOCA accidents analyses remain those stated in Table 3 of RG 1.183. The current UFSAR only explains that the rods will not experience DNB following any DBA, but it does not clearly state that the gap fractions in the locked rotor and rod ejection accident remain those in Table 3 of RG 1.183 and that the exceeding the power/burnup criteria in Footnote 11 in RG 1.183 does not apply to these accidents.

Enclosure to Duke Energy Letter RA-19-0134 Page 3 of 3

RAI-1

The staff requests the licensee to either describe how it will incorporate the new design requirement into the Oconee licensing basis, as reflected in the UFSAR, TSs, or any other document controlled under 10 CFR 50.59 (such as the COLR); or provide the revised radiological consequence analyses for the other design basis accidents that assume fuel failure (such as locked rotor accident, control rod ejection accident, etc.)

that demonstrate that the regulatory limits will be met with the new proposed gap fractions for high-burnup fuel rods.

Duke Energy Response to RAI-1 Duke Energy will incorporate the new design requirement into the Oconee licensing basis via the attached proposed revision to the Updated Final Safety Analysis Report (UFSAR). The attached UFSAR mark-up (2 pages) supersedes the UFSAR Section 15.1.10 mark-up included in the November 1, 2018 license amendment request. The remainder of the originally proposed UFSAR changes are not affected by this RAI response.

Oconee Nuclear Station UFSAR Chapter 15

4. Many of the transient and accident analyses involve control rod movement. These analyses credit the normal withdrawal sequence, overlap, and rod speed, which are controlled by non-safety control systems.

5. For certain failures in the EFW System, credit is taken for realigning EFW flow through the non-safety MFW System.

6. Steaming of the steam generators with manual non-safety atmospheric dump valves is credited.

7. Deleted per 2003 update

8. The capability to remotely throttle certain valves is credited. Some of the controls required to remotely throttle these valves are not safety-grade.

9. Electrical bus voltage and frequency control are credited. These are controlled by non-safety components.

10. The Integrated Control System trips both main feedwater pumps on a high steam generator level indication. A high level indication may occur following a main steam line break due to the pressure drops that result from the blowdown of the steam generator. Tripping of the main feedwater pumps will be assumed to occur in the steam line break analysis only if the plant response is more limiting.

15.1.10 Environmental Consequences Calculation Methodology Environmental Consequences A summary of the offsite doses is presented in Table 15-16. A description of each accident analysis is given in the appropriate section.

Fission Product Inventories Inventory in the Core: Fission product inventories within the core are calculated based on the ORIGEN methodology (e.g., ORIGEN-ARP or SAS2H/ORIGEN-S of the SCALE computer code)(Section 15.1 , Ref. 27). The core inventories for the Maximum Hypothetical Accident are shown in Table 15-15.

Inventory in the Reactor Coolant: The quantity of fission products released to the reactor coolant during steady state operation is based on the use of escape rate coefficients (sec- 1 )

derived from experiments involving purposely defected fuel elements. (Section 15.1 ,

References 29, 30, .fil_, 32) These coefficients represent the fraction of the activity in the fuel (31 DEC 2017) 15.1 - 7

Insert A:

For non-DNB fuel pins that exceed the rod power/burnup criteria of Footnote 11 in RG 1.183, the gap fractions from RG 1.183 are increased by a factor of 4 for Kr-85, Cs-134, and Cs-137. The gap fractions for all other isotopes remain at their pertinent RG 1.183, Table 3 values (References 46 and 47). The fuel cycle design ensures that none of these fuel pins experience DNB following any design basis accident.

The fuel cycle design also ensures that no fuel rod predicted to experience DNB in any other non-LOCA accidents (e.g., locked rotor accident or rod ejection accident) will have operated beyond the power/burnup criteria of Footnote 11 in Regulatory Guide 1.183 and that the gap fractions used in these non-LOCA accident analyses remain those stated in Table 3 of RG 1.183.