ML14066A410

Correction to Safety Evaluation Supporting Amendment No. 196 Measurement Uncertainty Recapture Power Uprate

ML14066A410
Person / Time
Site:	Fermi
Issue date:	03/14/2014
From:	Thomas Wengert Plant Licensing Branch III
To:	Plona J DTE Electric Company
Wengert T
References
TAC MF0650
Download: ML14066A410 (6)

Text

.. UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 14, 2014 Mr. Joseph H. Plena Senior Vice President and Chief Nuclear Officer DTE Electric Company Fermi 2-210 NOC 6400 North Dixie Highway Newport, Ml 48166

SUBJECT:

FERMI2- CORRECTION TO SAFETY EVALUATION SUPPORTING AMENDMENT NO. 196 RE: MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE (TAC NO. MF0650)

Dear Mr. Plena:

By letter dated February 10, 2014 (Agencywide Document Access and Management System Accession No. ML13364A131 ), the Nuclear Regulatory Commission (NRC) issued Amendment No. 196 to Facility Operating License No. NPF-43 for Fermi, Unit 2. The amendment consisted of changes to the technical specifications (TSs) in response to your application dated February 7, 2013 (ADAMS Accession No. ML13043A659), as supplemented by letters dated March 8, 2013 (ADAMS Accession No. ML13070A197), April 5, 2013 (ADAMS Accession No. ML13095A456), June 7, 2013 (ADAMS Accession No. ML13161A080), July 15, 2013 (ADAMS Accession No. ML13197A121), and September 27, 2013 (ADAMS Accession No. ML13273A464).

The amendment revised the Operating License and Technical Specifications to implement an increase of approximately 1.64 percent in rated thermal power from the current licensed thermal power of.3430 megawatts thermal (MWt) to 3486 MWt. The changes are based on increased feedwater flow measurement accuracy, which will be achieved by utilizing Cameron International (formerly Caldon) CheckPius TM Leading Edge Flow Meter (LEFM) ultrasonic flow measurement instrumentation.

Subsequent to issuing Amendment No. 196, two errors were identified in the supporting safety evaluation (SE). Your staff pointed out that, in two locations on page 52 of the SE, ~effective full power years (EFPY) should be 20.1 EFPY. The NRC staff concurs that this was an error.

In addition, the NRC staff identified an error in the technical evaluation section of the SE (Section 3.11.3.2.1.2, pages 65 and 66) concerning the adoption of Technical Specification Task Force (TSTF) 493, Revision 4.- The staff noted that the discussion of the limiting trip setpoint (LTSP) and the nominal trip setpoint (NTSP) was not correct. To correct and further clarify this discussion, several changes have been made to Section 3.11.3.2.1.2.

Corrected versions of pages 52, 65, and 66 of the safety evaluation are enclosed.

J. Plona Please contact me at 301-415-4037 if you have any questions.

Sincerely, Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1.

Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-341

Enclosure:

Corrected pages 52, 65, and 66 of the safety evaluation cc w/encl: Distribution via ListServ

ML053120186) to the Fermi 2 operating license, approving the implementation of these curves in the Fermi 2 TS for 24 EFPY and 32 EFPY.

The NRC staff noted that the currentTS PIT limits for Fermi 2 are based on the analysis of the limiting RPV beltline shell material. Excluding the minimum temperature criteria (which were correctly established based on the RT NDT for the RPV flange region per Table 1 of 10 CFR Part 50, Appendix G), the PIT limits were generated using the 24 EFPY and 32 EFPY ART values for the limiting beltline shell material at the one-quarter RPV wall thickness (1/4T) and three-quarters RPV wall thickness (3/4T) locations, as specified in the ASME Code,Section XI, Appendix G. Additionally, as documented in NED0-33133, the development of these curves also took into consideration the.non-beltline regions of the RPV, the upper RPV region, and the RPV lower head. For the non-beltline regions of the RPV, the licensee generated PIT limits for the bounding structural discontinuities based on the limiting stress concentration effects associated with these components. The PIT limits for the non-beltline components were found to be less restrictive than those for the limiting beltline shell material.

Section 3.2 of the SAR states that the RPV water level instrument nozzles are bounded by the current TS PIT limit curves only up to 20.1 EFPY and that this issue affects the operation of Fermi 2, regardless of the MUR implementation. Section 3.2 of the SAR also indicates that resolution of this issue is being pursued in a separate LAR. The NRC staff noted that the analyses documented in NED0-33133 did not address the RPV beltline water level instrument nozzles for the generation of the 24 EFPY and 32 EFPY TS PIT limit curves. 10 CFR Part 50, Appendix G requires that PIT limits be developed by considering all beltline and non-beltline ferritic RCPB components, in particular RPV nozzles, penetrations, and other discontinuities that exhibit higher stresses than the RPV beltline shell region, and which could result in more restrictive PIT limits, even if the RT NDT for these components is not as high as that of the limiting RPV beltline shell material. Therefore, the PIT limit curves must bound the water level instrument nozzles in order for the curves to remain in compliance with 10 CFR Part 50, Appendix G beyond 20.1 EFPY.

By letter dated December 21, 2012 (ADAMS Accession No. ML13004A134), the licensee submitted an LAR to implement a PIT limits report (PTLR) in accordance with criteria established in GL 96-03, "Relocation of the Pressure-Temperature Limit Curves and Low Temperature Overpressure Protection System Limits." This LAR addresses the necessary PIT limit curve revision for ensuring that the curves bound the water level instrument nozzles, as referenced in Section 3.2 of the MUR SAR. The proposed PTLR contains new PIT limit curves for 24 EFPY and 32 EFPY. The NRC staff's review of the new PIT limit curves confirmed that these curves are bounding for all ferritic RCPB components, including the water level instrument nozzles. Additionally, the new PIT curves continue to remain bounding for the non-beltline regions of the RPV .since there is no significant neutron embrittlement affecting the component-specific limits for these non-beltline components. Forth~ bounding RPV components, which include the limiting beltline shell material and the water level instrument nozzles, the proposed 32 EFPY PIT limit curves were generated using 32 EFPY ART values that correspond with those listed in Table 3-1 of the SAR. Therefore, since the curves are calculated based on neutron fluenceprojections and corresponding ART values for 115 percent of CLTP, which bounds plant operation for the proposed MUR (1 01.64 percent of CLTP), the NRC staff finds that the PIT limit curves included in the proposed PTLR would be acceptable for MUR conditions. The staff's detailed review and findings regarding the acceptability of the proposed PTLR and PIT limit curves, relative to criteria of GL 96-03 and the requirements of 10 CFR Part 50, Appendix G, are documented in the staff's SE associated with Amendment No. 195, dated February 4, 2014 (ADAMS Accession No. ML133468067).

Corrected by letter dated March 14, 2014

The licensee proposed adding the following two notes toTS Table 3.3.1.1-1, "Reactor Protection System Instrumentation," Function 2.b, Average Power Range Monitors Simulated Thermal Power- Upscale trip function:

Noted: If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

Note e: The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field setting) to confirm channel performance. The NTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Technical Requirements Manual.

3.11.3.2.1.1 Regulatory Evaluation The Commission's regulatory requirements related to the content of the TS are contained in 10 CFR 50.36. The regulation requires, in part, that the TS include limiting safety systems settings.

Section 50.36(c)(1 )(ii)(A) states, in part:

Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions.

Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor.

3.11.3.2.1.2 Technical Evaluation The APRM Simulated Thermal Power- Upscale trip setpoint documented in the Fermi 2 Technical Requirements Manual (TRM) Table TR 3.3.1.1-1 is the LSSS. It corresponds to the number calculated as the "NTSP" in the Fermi 2 GE methodology calculation, which TSTF-493 would recognize as the calculated limiting trip setpoint (LTSP). In the Fermi 2 setpoint methodology the NTSP is the limiting setting for the channel trip setpoint considering all credible instrument errors associated with the instrument channel. Therefore, the NTSP is the least conservative value (with an as-left tolerance) to which the channel must be reset at the conclusion of periodic testing to ensure that the analytical limit will not be exceeded during an anticipated operational occurrence or accident before the next periodic surveillance or calibration. It is impossible to set a physical instrument channel to an exact value, so a calibration tolerance is established around the NTSP. Therefore, the NTSP adjustment is considered successful if the as-left instrument setting is within the setting tolerance (i.e., a range of values around the NTSP). In this case of the APRM Simulated Thermal Power-Upscale trip setpoint, the NTSP satisfies the 10 CFR 50.36(c)(1 )(ii)(A) requirement that LSSSs be included in TSs. Additionally, to ensure proper use of the NTSP, the methodology for calculating the as-left and as-found tolerances is incorporated, by reference, in TSs surveillance Footnote (e) as Corrected by letter dated March 14, 2014

the TRM. The licensee-proposed Table 3.3.1.1-1 Footnote (e) correctly identifies the "NTSP" as the LSSS documented in the TRM.

The NRC staff reviewed Table 3.3.1.1-1 Footnote (d) and Footnote (e) and found they are consistent with the wording of the two notes in Option A of TSTF-493, Revision 4. Therefore, the NRC staff finds the addition of Table 3.3.1.1-1 Footnote (d) and Footnote (e) to be acceptable for meeting the requirements for LSSS under 10 CFR 50.36 for the APRM Simulated Thermal Power-Upscale trip function.

Based on the above discussion and the NRC staff's review of the licensee's LAR, the NRC staff concludes that the licensee provided sufficient justifications for the proposed TS changes. The NRC staff concludes that the licensee has followed the guidance in Items A through C in Section VIII of Attachment 1 to RIS 2002-03, and therefore has met the relevant regulatory requirements of 10 CFR Part 50, Appendix K.

3.11.4 *Instrumentation and Control Systems Conclusion The NRC staff reviewed the licensee's proposed plant-specific implementation of the Cameron LEFM CheckPius FW flow measurement device and the power uncertainty calculations. Based on its review of the licensee's LAR, uncertainty calculations, and referenced topical reports, the NRC staff finds that the licensee's proposed amendment is consistent with the approved Cameron Topical Report ER-80P and its supplement, Topical Report ER-157P, as well as the guidance of RIS 2002-03. The NRC staff also finds that the licensee adequately accounted for all instrumentation uncertainties in the reactor thermal power measurement uncertainty calculations. Therefore, the licensee's proposed amendment meets the relevant requirements of Appendix K to 10 CFR 50.

The NRC staff further concludes that the proposed TS changes meet the requirements of 10 CFR 50.36, the guidance of RG 1.1 05, Revision 3, and TSTF-493, Revision 4. On the basis of these considerations, the NRC staff finds the instrumentation and controls aspects of the proposed thermal power uprate acceptable.

3.12 Plant Systems 3.12.1 Regulatory Evaluation The NRC staff's review in the area of plant systems covers the impact of the proposed MUR power uprate on balance of plant piping, safety-related cooling water systems, ultimate neat sink, radioactive waste systems, and spent fuel pool (SFP) storage and cooling. The licensee evaluated the effect of the MUR on the plant systems. This evaluation is reflected in Enclosures 7 and 9 of the licensee's application dated February 7, 2013.

3.12.2 Technical Evaluation Balance of Plant Piping The principal construction codes for the Balance of Plant (BOP) systems are listed in the UFSAR Chapter 3, "Design of Structures, Components, Equipment, and Systems." Principal codes include ASME Boiler and Pressure Vessel Code Sections Ill and VIII and American National Standards Institute B31.1 Standard Code for Pressure, Power Piping. Section 3.5.2 of the Fermi Safety Analysis Report (NEDC-33578P) lists 24 BOP systems evaluated for TPO Corrected by letter dated March 14, 2014

J. Plona Please contact me at 301-415-4037 if you have any questions.

Sincerely,

/RAJ Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-341

Enclosure:

Corrected pages 52, 65, and 66 of the safety evaluation cc w/encl: Distribution via ListServ DISTRIBUTION:

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