NRC-15-0065, Response to NRC Request for Additional Information Regarding the Fermi 2 Expedited Seismic Evaluation Process Report
| ML15148A432 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/28/2015 |
| From: | Kaminskas V DTE Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC-15-0065 | |
| Download: ML15148A432 (10) | |
Text
Vito A. Kaminskas Site Vice President DTE Energy Company 6400 N. Dixie Highway, Newport, MI 48166 Tel: 734.586.6515 Fax: 734.586.4172 Email: kaminskasv@dteenergy.com DTE Energy-May 28, 2015 10 CFR 50.54(f)
NRC-15-0065 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001
References:
- 1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43
- 2) DTE Electric Company Letter to the NRC, "Fermi 2 Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," NRC-14-0074, dated December 9, 2014 (Accession No. ML14345A469)
- 3) Email from Stephen Wyman (NRC) to Kevin Burke (DTE Electric Company),
Subject:
Fermi ESEP Clarification Questions, dated April 30, 2015
Subject:
Response to NRC Request for Additional Information Regarding the Fermi 2 Expedited Seismic Evaluation Process Report In Reference 2, DTE Electric Company (DTE) submitted the Fermi 2 Expedited Seismic Evaluation Process Report, in response to NRC "Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident." In Reference 3, the NRC staff requested additional information. The enclosure of this letter provides DTE's response to the NRC staff request.
This letter contains no new regulatory commitments.
Should you have any questions or require additional information, please contact Mr.
Christopher Robinson, Licensing Manager at (734) 586-5076.
USNRC NRC-15-0065 Page 2 I declare under penalty of perjury that the foregoing is true and correct.
Executed on May XX, 2015 Vito A. Kaminskas Site Vice President
Enclosure:
Response to Request for Additional Information cc: NRC Project Manager NRC Resident Office Reactor Projects Chief, Branch 5, Region III Regional Administrator, Region III Michigan Public Service Commission Regulated Energy Division (kindschl1michigan.gov)
Enclosure to NRC-15-0065 Response to Request for Additional Information
Enclosure to NRC-15-0065 Page 1 Response to Request for Additional Information By email dated April 30, 2015, the U.S. Nuclear Regulatory Commission (NRC) requested that DTE Electric Company (DTE) provide additional information regarding the Expedited Seismic Evaluation Process (ESEP) Report for Fermi 2.
RAI-1
Section 6.2 of the Fermi ESEP Report describes the Expedited Seismic Equipment List (ESEL) component screening process using tables in Electric Power Research Institute (EPRI) NP-6041 "A Methodology for Assessment of Nuclear Power Plant Seismic Margin." These tables are applicable to components located up to 40 feet (ft) above grade. The ESEP report is silent concerning screening or high confidence of low probability of failure (HCLPF) calculations for components at elevations beyond 40 ft above grade. Therefore, describe how ESEL components located at elevations beyond 40 ft above grade either were screened out or had their HCLPF capacities calculated, including specific references for the applicable guidance utilized.
RESPONSE
There are 72 components on the ESEL identified in Attachment C of the Fermi 2 ESEP Report that were screened by using Table 2-4 of EPRI NP-6041. Grade elevation at Fermi 2 is 583 feet, therefore, the 40 feet above grade criteria recommended to use Table 2-4 for screening applies for equipment below elevation (EL) 623 feet. All of the ESEL components screened using Table 2-4 are located at or below this elevation.
There are 56 ESEL components at elevations greater than 40 feet above grade, these are addressed as follows:
" Nine were screened on the basis of their inclusion in the Individual Plant Examination of External Events (IPEEE) (see Section 6.2 of the Report).
" 29 are child devices that were given the fragility value of their parent component. This is by the Rule of the Box.
- The remaining 18 HCLPF's were calculated using the Conservative Deterministic Failure Margin (CDFM) Method described in Section 6.4 of the ESEP Report.
RAI-2
Section 5.2 of the Fermi ESEP report states:
"Subsequent equipment HCLPF calculations and fragility evaluations are based on the conservative deterministic failure margin (CDFM) approach. In accordance with EPRI 1019200
[10] "Seismic Fragility Application Guide Update," the seismic analyses are performed using
[best estimate] BE structure stiffness, mass and damping characteristics, and the BE subsurface shear wave velocity (Vs) profile compatible with the expected seismic shear strains. The
Enclosure to NRC-15-0065 Page 2 resulting [in-structure response spectra] ISRS approximately represent the 84th percentile response suitable for use in the CDFM calculations."
Section 4 of the "Seismic Evaluation Guidance, Augmented Approach" (EPRI 3002000704) allows the development of ISRS calculated from new soil-structure interaction (SSI) models.
The guidance document indicates that: EPRI 1025287 "Screening, Prioritization, and Implementation Details" (SPID) and the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) Probabilistic Risk Assessment (PRA) Standard give guidance on acceptable methods to compute both the ground motion response spectrum (GMRS) and the associated ISRS. Table 6-5 in the SPID document, under the SFR-C6 entry, indicates that ASME/ANS PRA Standard (Addendums A and B) requires consideration of the variation of soil properties (Vs profile). Also, the SFR-C5 entry indicates that if the median-centered response analysis is performed, the evaluation should estimate the median response (i.e., structural loads and ISRS) and variability in the response using established methods.
Based on EPRI 1019200, which was referenced by the Fermi ESEP, parameter variation should be incorporated into SSI analyses in order to characterize the uncertainty in the SSI demands.
EPRI 1019200 indicates that the SSI analyses in American Society of Civil Engineers (ASCE) 4-98 "Seismic Analysis of Safety-Related Nuclear Structures and Commentary" be followed, which require that SSI evaluations include lower bound and upper bound soil profiles to account for parameter variation in SSI. EPRI 1019200 also indicates that for the structural model, the best estimate (median) and uncertainty variation in the frequency should be considered.
Therefore, please describe how parameter variation is incorporated into the SSI analyses for the structural model and subsurface while using only the best estimate (BE) structure stiffness, mass and damping characteristics, and the BE subsurface Vs profile. Related to the above discussion, if only the BE is used for the structural model and soil profile, explain how the ISRS would approximately represent the 84th percentile response, as stated in the ESEP report.
RESPONSE
The reactor building (RB) and auxiliary building (AB) are supported on the site bed rock which is characterized by a shear wave velocity (Vs) in excess of 5000 feet per second (fps). Because the SSI frequencies for the range of rock Vs are similar to the fixed base structure frequency, SSI effects are not expected to be significant. In fact, according to the SPID, the Fermi 2 structures could be evaluated as fixed base. Nevertheless, SSI effects were accounted for in the seismic evaluations performed in support of the seismic probabilistic risk assessment (SPRA). The SSI analysis is based on the best estimate shear wave velocity profile so that the calculated response is based on median stiffness and damping properties of the SSI system.
The effects of variability in the stiffness of the supporting rock medium were evaluated in accordance with the procedure of EPRI TR-103959 "Methodology for Developing Seismic Fragilities." This evaluation demonstrated that the difference in the lower and upper bound SSI frequencies (assuming a coefficient of variance (Cv) of 0.5) relative to the best estimate is less
Enclosure to NRC-15-0065 Page 3 than 5%. Therefore a single best estimate analysis is considered to be sufficient for use in the CDFM analysis.
RAI-3
Section 5.2 of the ESEP report indicates that in performing the seismic SSI analysis to obtain the in-structure response spectra (ISRS), the SSI analysis assumes that the structures are surface founded at their respective foundation levels. Clarify whether this means that the effects of embedments (i.e., connection of the subsurface material to the structures) are neglected. If so, then explain why embedments were not considered, since they may affect the ISRS (i.e., change in magnitude and shift in frequencies of the spectral peaks).
RESPONSE
The effects of embedment are not included in the soil structure interaction analysis for the RB or AB for the following primary reasons:
- 1. Because of adjacent buildings, the below grade walls of the RB/AB are not fully in contact with the soil.
- 2. Much of the embedment effects, if any, are extracted from the fill placed above the site bedrock. The stiffness contributed by the fill is not significant relative to the horizontal stiffness of the underlying bedrock.
The buildings are effectively embedded into till and rock at a height of about 25 feet. Because of the relatively large velocity contrast, the material overlying the stiff rock and till is not considered to be effective in modifying the input motion, and the SSI response.
Based on the foundation dimensions of the building (209 ft x160 ft), the equivalent foundation radius is about 103 ft. The ratio of the effective embedment is thus 0.24. ASCE 4-98 states that if the ratio is less than 0.3, embedment effects are small.
RAI-4
Section 3.1.5 - Critical Instrumentation Indicators, in the ESEP report states:
"Critical indicators and recorders are typically physically located on panels/cabinets and are included as separate components; however, seismic evaluation of the instrument indication may be included in the panel/cabinet seismic evaluation (rule-of-the-box)."
Section 6.1 - Summary of the Methodologies Used, in the report states:
"Consolidate ESEL by "boxing" breakers and switches and other devices and instrumentation items into "parent" component in which they are housed, such as a motor control center (MCC)
Enclosure to NRC-15-0065 Page 4 or switchgear. This consolidation is performed so that devices and instruments could be evaluated for HCLPF capacities."
The information provided in both paragraphs is not clear. Please provide a more detailed description of both approaches, how they are different, when would each approach be applied, and examples for both approaches to show how the HCLPF values of the devices were determined, including consideration of cabinet amplification, if applicable. Also, describe whether any of these devices are sensitive to vibration as are relays and other devices with contacts, and if so, how they were evaluated. Lastly, if the qualification of the devices is based on the cabinet/panel they are housed in, which have been previously qualified as part of an equipment class ("parent" component), how is it known/confirmed that the parent component normally contains the particular device.
RESPONSE
With the exception of specific relays, all devices in the seismic equipment list (SEL) were "boxed" into the parent component in which they are housed. The CDFM method was used to develop HCLPF values of the parent components. The HCLPF values for the parent components were assigned to all the devices housed therein.
Specific relays identified in the ESEL were evaluated by comparing the appropriate capacity generic equipment ruggedness spectra (GERS) to in-cabinet demand. The in-cabinet demand is obtained by applying the cabinet amplification factors recommended in EPRI TR-103959 to the in structure response spectrum at the base of the cabinet in the frequency range at and above the cabinet fundamental frequency.
RAI-5
Section 6.4 of the ESEP report indicates that ESEL items not included in the Fermi SPRA evaluations were evaluated using the CDFM approach described in EPRI NP-6041 and EPRI 1019200. Section 6.1 of the ESEP report indicates that the on-going SPRA fragility calculations in support of the SPRA are being performed using the CDFM approach following the guidelines described in EPRI 6041, EPRI 1002988 "Seismic Fragility Application Guide", and EPRI 1019200 "Seismic Fragility Applications Guide Update." Section 6.1 of the report also indicates that the SPRA HCLPF values are used for the ESEP.
For the ESEL items relying on the on-going SPRA fragility calculations,, it is not clear what approach was utilized to obtain the HCLPF values reported in the ESEP report. Please verify that the HCLPF calculations were conducted using either of the two methods identified in the Augmented Approach guidance (i.e., no other method other than the two methods identified in the guidance was used). This should include a statement whether the HCLPF values were determined using only the CDFM approach directly or the HCLPF values were determined from fragility analyses where the median capacity and the randomness/uncertainty beta values are
Enclosure to NRC-15-0065 Page 5 calculated in a detailed manner for the various parameters. If another approach was used, please explain how the method is equivalent to the methods agreed upon the in the guidance.
RESPONSE
The fragility parameters used in the SPRA are developed using the CDFM method as described in EPRI 1019200 "Seismic Fragility Application Guide Update." The separation of variable approach, as described in EPRI 1002988 "Seismic Fragility Application Guide," has not been used for the reported HCLPF values.
RAI-6
Section 7.1 states that the licensee's "[Seismic Review Team] SRT developed confidence to address approximately 27 items on the basis of similar equipment and installation, supplemented by previous walkdowns and photographs. No items on the ESEL are identified as inaccessible."
Please clarify the following:
a) Since Section 7.1 is titled "Identification of ESEL Items Inaccessible for Walkdowns,"
and the last sentence quoted above indicates that all items are accessible, it is not clear if the 27 items identified are accessible and if so, why is a walkby or walkdown not being performed?
b) What are the timeframes of the assessments for the similar equipment and installation and how does the age of the component being walked down compare to the age of the 27 items not being walked down?
c) What are the dates of the various walkdowns (i.e., SPRA walkdowns, ESEP walkdowns, previous walkdowns) used to develop confidence for the applicable 27 items?
d) When were the photographs taken to support the basis of similar equipment and installation?
RESPONSE
a) Of the 291 components on the ESEL, 151 are independently mounted (i.e. not a device mounted on something else). Consistent with the ESEP requirements these 151 components should all be walked down.
Since the time of the ESEP submittal, an additional eight components were able to be documented. This leaves 19 components that have not been walked-down, instead of the 27 reported in the ESEP Report.
Of the remaining 19 components on the ESEL:
- Two were screened by engineering judgment as being seismically rugged. These are small, ruggedly mounted components at low elevation in the building. It is not expected that these will be affected by the GMRS earthquake. One is T50P402, a small electrical box that routes the wires from temperature probes from one side of the Torus wall to the other. This is essentially an electrical penetration. It is located high
Enclosure to NRC-15-0065 Page 6 on the Torus and is not located in an environment where significant environmental aging degradation mechanisms would occur. The other is E41N062B. This is a rugged instrument in the basement level of the RB. Review of photos of a similar component (E41N062D) show that the small component is ruggedly mounted directly to the wall and is in a mild environment.
- 15 items were screened on the basis of Table 2-4 from EPRI NP-6041. It is expected that these are all generally rugged and free of degradation.
Drawings show this rack is similar in construction to several other high safety significance (HSS) supported racks in the plant and is not in an environment where degradation is likely to occur. The rack is on the second floor of the reactor building on the outside of the primary containment shield wall, the environment in this location is mild. It was judged that the rack is adequate. The other is T5000F420B. While this valve was not visible, the T5000F420A valve was. It is expected that the valves will have similar construction and installation dates. It is judged that this valve is adequate based on the walkdown of the sister valve.
Finally, the majority of the components which were not visible during the walkdowns are valves associated with either the high pressure coolant injection (HPCI) or reactor core isolation cooling (RCIC). Both the HPCI and RCIC turbines were walked down during the SPRA walkdowns in 2013, but specific valves on or around their skid were not documented by specific plant identification system (PIS) number. The turbines were not accessible when the ESEP walkdown was conducted in February 2014. Based on the SPRA walkdown of the HPCI and RCIC turbines and the area walkdown of the HPCI room reported in the Near-term Task Force (NTTF) 2.3 Report, the SRT was confident that the turbines, as well as ancillary equipment were free of degradation.
Further walkdowns and data collection on these 19 components have occurred since the submittal of the ESEP Report and have confirmed the original assessments of these components.
b) As stated above, all but two of the inaccessible components were able to be screened by either Table 2-4, or judging that they are inherently rugged. Rack T49P400B is likely as old as other similar racks in the plant, none of the other racks showed degradation. Valve T500F420B is likely as old as the sister valve that was walked-down. No degradation was noted with the sister valve.
c) The NTTF 2.3 Walkdowns were conducted in August 2012. The SPRA walkdowns were conducted in March and September 2013. The ESEP walkdown was conducted in February 2014.
The IPEEE walkdowns were conducted in the 1990s. Notes of these walkdowns were used as reference, but IPEEE items were re-walked down since it has been about 20 years since
Enclosure to NRC-15-0065 Page 7 the IPEEE walkdowns were conducted. No walkdown photographs from the IPEEE were credited.
d) The photos provided were from 2.3 walkdowns or from the plant picture archives. The plant pictures were taken from the last time a piece of equipment was accessed, these pictures are taken for pieces of equipment that are not normally accessible because they are located in areas with higher dose rates, contamination, or in areas only accessible during refueling outages.
On the basis of the above, no components were designated as requiring walkdown in the future.