ML23216A187

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Submittal of Relief Request RR-A25 for Extended License Period
ML23216A187
Person / Time
Site: Fermi DTE Energy icon.png
Issue date: 08/04/2023
From: Peter Dietrich
DTE Energy Company
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
NRC-23-0028
Download: ML23216A187 (1)
v • d • e


Text

Peter Dietrich Senior Vice President and Chief Nuclear Officer DTE Energy Company 6400 N. Dixie Highway, Newport, MI 48166 Tel: 734.586.6515 Email: peter.dietrich@dteenergy.com August 4, 2023 10 CFR 50.55a NRC-23-0028 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Fermi 2 Power Plant NRC Docket No. 50-341 NRC License No. NPF-43

Subject:

Submittal of Relief Request RR-A25 for Extended License Period

References:

1) NRC letter dated March 30, 2000, Fermi 2 - Relief Requests for the Second 10-Year Interval Inservice Inspection (ISI) Nondestructive Examination (NDE)

Program (Accession No. ML003697623)

2) DTE Electric Company Letter NRC-14-0028 to NRC, Fermi 2 License Renewal Application, dated April 24, 2014 (Accession No. ML14121A554)
3) DTE Electric Company Letter NRC-15-0022 to NRC, Response to NRC Request for Additional Information for the Review of Fermi 2 License Renewal Application - Set 12 Questions 4.2.2-1 and 4.2.2-2 and Set 15 Question 4.2.6-1, dated January 30, 2015 (Accession No. ML15030A350)
4) EPRI Proprietary Report TR-105697, BWR Vessel and Internals Project BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," dated September 1995, (Accession No. ML032200246).

In Reference 1, the NRC granted relief request RR-A25, allowing elimination of the inspection of the reactor circumferential welds through the initial operating license term. In Reference 2 and 3, DTE provided the basis for continuing the elimination of the inspection of the reactor circumferential welds through the period of extended operation; however, the request for relief would be submitted separately.

Pursuant to 10 CFR 50.55a, Codes and Standards, paragraph (z)(1), DTE Electric Company (DTE) hereby requests NRC approval for the elimination of full ultrasonic examination of the Reactor Pressure Vessel (RPV) circumferential shell welds for the period of extended operation.

The justification for this deferral is provided in the attached request for relief (RR-A25) and is further supported by Reference 4. Examination of longitudinal, RPV shell welds will be

USNRC NRC-23-0028 Page 2 completed as scheduled. A portion of the circumferential welds will be volumetrically examined at their points of intersection with the longitudinal welds.

Should you have any questions or require additional information, please contact Mr. Eric W. Frank ofmy staff at (734) 586-4772.

Peter T. Dietrich Site Vice President and Chief Nuclear Officer

Enclosure:

RR-A25 -Alternative to Reactor Vessel Circumferential Weld Examinations cc:

NRC Project Manager NRC Resident Office Regional Administrator, Region III M. Wilson - Authorized Nuclear Inservice Inspector (ANII)

D. Stenrose - Chief Inspector Michigan Department of Licensing and Regulatory Affairs (LARA)

Bureau of Construction Codes and Fire Safety - Boiler Division

Enclosure to NRC-23-0028 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 RR-A25 Alternative to Reactor Vessel Circumferential Weld Examinations to 10 CFR 50.55a Request Number NRC-23-0028 RR-A25 Page 1 of 7 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Components Affected ASME Code Class:

Code Class 1 Examination Category: B-A Item Number:

B1.11

==

Description:==

Circumferential Pressure Retaining Shell Welds in Reactor Vessel Weld Numbers:

4-308A, 4-308B, 1-313, and 9-307

2. Applicable Code Edition and Addenda

The Fermi 2 Inservice Inspection (ISI) Program is in its fourth ten-year interval. The applicable code is American Society of Mechanical Engineers (ASME)Section XI, Division 1, 2013 Edition. The applicable code will be updated in future intervals in accordance with 10 CFR 50.55a(g)(4)(ii).

3. Applicable Code Requirement

ASME Section XI, Division 1, 2013 Edition Table IWB-2500-1 (B-A) requires volumetric examination of essentially 100% of the weld length for all RPV circumferential pressure-retaining shell welds.

4. Reason for Request

DTE requests an alternative in accordance with 10 CFR 50.55a(z)(1) on the basis that this alternative provides an acceptable level of quality and safety. The alternative is relief from circumferential weld examinations required by the Code from the current fourth ISI interval through the period of extended operation (PEO). Fermi 2 will continue to perform volumetric examinations of essentially 100 percent of the RPV longitudinal shell welds and a portion of the circumferential welds will be volumetrically examined at their points of intersection with the longitudinal welds.

As documented in NUREG-2210 (Reference 1), Section 4.2.5 Reactor Vessel Circumferential Weld Examination Relief, the Nuclear Regulatory Commission (NRC) has concluded that the evaluation for the Time Limiting Aging Analysis (TLAA) for Fermi 2s RPV circumferential welds is acceptable because 1) the 52 effective full power years (EFPY) conditional failure probability will remain bounded by the NRC analysis in the staffs safety evaluation report (SER) dated July 28, 1998 (Reference 2), and 2) the applicant will be using procedures and training to limit cold overpressure events during the period of extended operation. The NRC also stated

to 10 CFR 50.55a Request Number NRC-23-0028 RR-A25 Page 2 of 7 that the analysis is consistent with the evaluation criteria in the staffs SER for Boiling Water Reactor (BWR) Vessel and Internals Project (BWRVIP) report BWRVIP-05 (Reference 3); however, the applicant is still required to request relief from RPV circumferential weld examination requirements for ISI intervals over the extended period of operation, in accordance with 10 CFR 50.55a.

5. Proposed Alternative and Basis for Use

The proposed alternative is the elimination of the RPV circumferential weld examinations, except for the intersection at the axial welds. Report BWRVIP-05 (Reference 3) provides the technical basis for eliminating inspections of RPV circumferential shell welds. This report was transmitted to the NRC in September 1995 and in their July 28, 1998 final safety evaluation (SE) (Reference 2) the NRC staff concluded the failure frequency for RPV circumferential welds is sufficiently low to justify elimination of inservice inspection.

On November 10, 1998, the NRC issued Generic Letter (GL) 98-05 (Reference 4) informing BWR licensees that they may request permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV circumferential welds, Code Category B-A, Item Number B1.11 by demonstrating compliance with the following criteria:

1. At the expiration of their license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staffs July 28, 1998, safety evaluation (Reference 2), and
2. Licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the staffs July 28, 1998, safety evaluation (Reference 2).

Criterion 1 - At the expiration of their license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staffs July 28, 1998, safety evaluation.

DTE Energy Response The NRC evaluation of the BWRVIP-05 report included a Probabilistic Fracture Mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the PFM analysis are:

1. The neutron fluence used was estimated to be the end-of-license mean fluence;
2. The chemistry values are mean values based on vessel types; and
3. The potential for beyond design basis events is considered

to 10 CFR 50.55a Request Number NRC-23-0028 RR-A25 Page 3 of 7 For Fermis RPV, the single circumferential weld joint located between the lower shell course and the lower-intermediate shell course (1-313) is the limiting circumferential weld. For plants such as Fermi 2, with RPVs fabricated by Combustion Engineering (CE), the mean end-of-license neutron fluence used in the NRC PFM analysis is 0.4E+19 n/cm2 at 64 effective full power years (EFPY),

as shown in Table 2.6-5 of Reference 2. The peak surface fluence value for the limiting circumferential weld at the expiration of Fermi 2s extended operating license (estimated at 52 EFPY) is 5.74E+17 n/cm2. Thus, the Fermi 2 RPV weld 1-313 fluence is bounded by the fluence assumed for the corresponding weld in the NRC PFM analysis for 64 EFPY.

Table 1 of this Enclosure presents a comparison of the chemistry related data for Fermi 2s RPV and the values assumed in the NRCs safety evaluation for the BWRVIP-05 report (Reference 2). The calculated mean embrittlement shift in RTNDT (RTNDT) for the limiting circumferential weld at Fermi 2 (weld 1-313) at the end of the extended license (52 EFPY) is 59 °F. By comparison, Table 2.6-5 of Reference 2 indicates an allowable embrittlement shift of 128.5 °F at 64 EFPY for CE fabricated vessels.

For these reasons, the limiting circumferential weld at Fermi 2 is less brittle than the corresponding weld in the NRCs PFM case study and is therefore bounded by the Staffs limiting conditional failure probability for CE circumferential welds.

Thus, Criterion 1 of GL 98-05 is met.

TABLE 1 Fermi 2 RPV Circumferential Weld 1-313 Information NRC Limiting Plant-Specific Analysis at 64 EFPY (Circ Welds)a NRC Limiting Plant-Specific Analysis at 64 EFPY (Circ Welds)b Fermi 2 at 52 EFPY (Circ Weld 1-313)

Inside diameter neutron fluence (1E+19 n/cm2) 0.4 0.4 0.0574 Initial (unirradiated) reference temperature (RTNDT) 0 0

-50 °F Weld chemistry factor (CF) 151.7 172.2 236 Weld copper content (%)

0.13 0.183 0.23 Weld nickel content (%)

0.71 0.704 1.00 Increase in reference temperature without margin (RTNDT) 113.2 °F 128.5 °F 59 °F Mean adjusted reference temperature (Mean ART =

RTNDT + RTNDT) 113.2 °F 128.5 °F 9 °F Probability of a failure event P(F/E) NRC 1.99E-04 4.38E-4

<1.99E-04c

to 10 CFR 50.55a Request Number NRC-23-0028 RR-A25 Page 4 of 7 a) Chemistry information reported in BWRVIP-05.

b) Chemistry information reported in Combustion Engineering Owners Group (CEOG) report.

c) The mean ART values of reactor vessel circumferential welds govern the conditional probability of failure for the welds. Consistent with the provisions of GL 98-05, the Fermi analysis confirmed that the mean ART value of the circumferential welds is significantly less than the NRC accepted mean ART value of the limiting case circumferential welds for Combustion Engineering (CE) reactor vessels. Therefore, the conditional probability of failure for Fermi would be lower than the NRC accepted value for CE reactor vessels.

Criterion 2 - Licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the staffs July 28, 1998, safety evaluation.

DTE Energy Response In GL 98-05, the NRC stated that beyond design-basis events occurring during plant shutdown could lead to Low Temperature Over-Pressure (LTOP) events that could challenge RPV integrity. The BWRVIP assessment indication that the major contribution to LTOP events frequency results from unmitigated injections from Condensate or Control Rod Drive (CRD) Systems and a failure to properly realign the reactor water cleanup system following a reactor trip at low temperatures. For a BWR to experience such an event would require several operator errors. Although no LTOP events have occurred at a domestic BWR, the NRC identified several events that could be considered precursors to such an event and cited one actual LTOP event that occurred at a foreign BWR. Fermi 2 has in place procedures and Technical Specifications which monitor and control reactor pressure, temperature, and water inventory during all aspects of cold shutdown which would minimize the likelihood of a Low Temperature Over-Pressurization (LTOP) event from occurring. Additionally, these procedures are reinforced through operator training.

The Pressure Test procedures which are used at Fermi 2, have sufficient procedural guidance to prevent a cold, over-pressurization event. Pressure testing is performed at the conclusion of each outage. The system leakage tests include requirements for operations management to perform a "pre-job briefing" with all essential personnel. This briefing details the anticipated testing evolution with special emphasis on: conservative decision making, plant safety awareness, lessons learned from similar in-house or industry operating experiences, the importance of open communications, and finally, the process in which the test would be aborted if plant systems responded in an adverse manner.

Vessel temperature and pressure are required to be monitored throughout these tests to ensure compliance with the Technical Specification pressure-temperature curve.

to 10 CFR 50.55a Request Number NRC-23-0028 RR-A25 Page 5 of 7 Additionally, to ensure a controlled, deliberate pressure increase, the rate of pressure increase is administratively limited throughout the performance of the test. If the pressurization rate exceeds this limit, direction is provided to remove the CRD pumps which are used for pressurization, from service.

With regard to inadvertent system injection resulting in an LTOP condition, the high pressure makeup systems (High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems, as well as the normal feedwater supply (via the Reactor Feedwater Pumps) at Fermi 2 are all steam driven. During reactor cold shutdown conditions, no reactor steam is available for the operation of these systems. Therefore, it is not possible for these systems to contribute to an overpressure event while the unit is in cold shutdown.

In the case of low pressure system initiation, the Fermi 2 pressure-temperature limit curves for hydrostatic testing (Reference 5 Figure 1), permit pressures up to 312 psig at temperatures from 72 °F to 102 °F. Above 102 °F, the permissible pressure increases immediately to near 800 psig and increases rapidly with increasing temperature. The shutoff head for the Core Spray and Residual Heat Removal Pumps are both below 400 psig. Therefore, the potential for an over-pressurization event which would exceed the pressure-temperature limits, as the likelihood of an inadvertent actuation of these systems is very low.

Procedural control is also in place to respond to an unexpected or unexplained rise in reactor water level which could result from a spurious actuation of an injection system.

Actions specified in this procedure included preventing condensate pump injection, securing Emergency Core Cooling System (ECCS) injection, tripping CRD pumps, terminating other injection sources, and lowering RPV level via the Reactor Water Cleanup (RWCU) system, and the steam line drains.

In addition to procedural barriers, Licensed Operator Training is given which further reduces the possibility of the occurrence of LTOP events. During Initial Licensed Operator Training the following topics are covered: Brittle fracture and vessel thermal stress; Operational Transient (OT) procedures, including the OT on reactor high level; Technical Specifications training, including Section 3.4.10, "RCS Pressure/Temperature (P/T) Limits" and Simulator Training of plant heatup and cooldown including performance of surveillance tests which ensure pressure-temperature curve compliance.

In addition to the above, continuous review of industry operating plant experiences is conducted to ensure that the Fermi 2 procedures consider the impact of actual events, including potential LTOP events. Appropriate adjustments to the procedures and associated training are then implemented to preclude similar situations from occurring at Fermi 2.

Based on the above, the probability of a cold over-pressure transient is considered to be highly unlikely, and Criterion 2 of GL 98-05 is met.

to 10 CFR 50.55a Request Number NRC-23-0028 RR-A25 Page 6 of 7 Conclusion The BWRVIP-05 report provides the technical basis for eliminating inspection of RPV circumferential shell welds. The BWRVIP-05 report concludes that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. Based on an assessment of the materials in the limiting circumferential weld in the beltline of the Fermi 2 RPV, the conditional probability of RPV failure is less than or equal to that estimated in the NRC's analysis through the end of the PEO. Based on established operator training, practices and procedural controls, the frequency of an LTOP event at Fermi 2 is less than or equal to the frequency assumed in the NRC's July 28, 1998, safety evaluation.

Therefore, DTE Energy herein requests approval to implement this alternative examination methodology for Fermi 2 as allowed by the SE of BWRVIP-05 and proposes to modify the Fermi 2 ISI schedule to perform volumetric examinations of essentially 100 percent of the RPV axial shell welds and only the circumferential welds at the points of intersection with the RPV axial welds.

6. Duration of Proposed Relief Alternative The duration of this request is for the remainder of the Fermi 2 extended license ending on March 20, 2045.

7. Precedents

The NRC has authorized similar requests to adopt an alternative to the ASME Section XI, Table IWB-2500-1, Examination Category B-A, Item. No. B1.11 criteria for permanent relief from the volumetric examination of RPV circumferential shell welds for the PEO operation.

Similar relief requests have been granted to:

  • Browns Ferry Nuclear Plant, Unit 1 - Resubmittal of Request for Relief No. 1-ISI-27 for The Period of Extended Operation (EPID L-2018-LLR-0389), Accession No. ML19284C736, October 28, 2019.

to 10 CFR 50.55a Request Number NRC-23-0028 RR-A25 Page 7 of 7

8. References
1. NUREG-2210, Safety Evaluation Report Related to the License Renewal of Fermi 2, Docket Number 50-341, (Accession Number ML16356A234).
2. Letter, Gus C. Lainas (NRC) to Carl Terry, BWRVIP Chairman, "Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report, dated July 28, 1998.
3. EPRI Proprietary Report TR-105697, BWR Vessel and Internals Project BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," dated September 1995, (Accession No. ML032200246).
4. NRC Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds,"

November 10, 1998, (Accession No. ML03140368).

5. DTE Letter to NRC, Transmittal of Revision to the Pressure and Temperature Limits Report (PTLR), NRC-20-0009, dated June 8, 2020 (ML20160A461).
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