ML15044A502
| ML15044A502 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  (DPR-026, DPR-064) |
| Issue date: | 02/12/2015 |
| From: | Hopenfeld J Riverkeeper |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, RAS 27239 | |
| Download: ML15044A502 (10) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD
x In re:
Docket Nos. 50-247-LR; 50-286-LR License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 Entergy Nuclear Indian Point 3, LLC, and Entergy Nuclear Operations, Inc.
February 12, 2015
x Riverkeeper, Inc. provisionally designates the attached Declaration of Dr. Joram Hopenfeld dated February 12, 2015 as containing Confidential Proprietary Information Subject to Nondisclosure Agreement REDACTED, PUBLIC VERSION
UNITED STATES NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD
x In re:
Docket Nos. 50-247-LR; 50-286-LR License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 Entergy Nuclear Indian Point 3, LLC, and Entergy Nuclear Operations, Inc.
February 12, 2015
x DECLARATION OF DR. JORAM HOPENFELD Joram Hopenfeld, hereby declares under penalty of perjury that the following is true and correct:
- 1.
I have been retained by Riverkeeper, Inc. (Riverkeeper) as an expert witness in proceedings concerning the application by Entergy Nuclear Operations, Inc. (Entergy) for the renewal of two separate operating licenses for the nuclear power generating facilities located at Indian Point on the east bank of the Hudson River in the Village of Buchanan, Westchester County, New York, for an additional 20 years beyond the expiration of their 40-year operating licenses.
- 2.
I submit this declaration in support of Riverkeeper and the State of New Yorks additional bases for previously admitted joint contention NYS-38/RK-TC-5, concerning Entergys failure to demonstrate that it has an adequate aging management programs (AMP) for various reactor components, including one to manage the effects of metal fatigue on reactor vessel internal (RVI) plant components during the proposed periods of extended operation of Indian Point Units 2 and 3.
- 3.
My professional and educational qualifications are described in the curriculum vitae appended hereto as Attachment 1. Briefly summarized, I am an expert in the field relating to nuclear power plant aging management. I am a mechanical engineer and hold a doctorate in mechanical engineering. I have 45 years of professional experience in the fields of thermal-hydraulics, material/environment interaction instrumentation, design, project management, and nuclear safety regulation, including 18 years in the employ of the U.S. Nuclear Regulatory Commission (NRC).
- 4.
My extensive professional experience has afforded me with knowledge and expertise regarding the material degradation phenomenon known as metal fatigue, that is, the fatigue or cyclic stress of metal parts due to repeated stresses during plant operation. Of note, I was a technical consultant and expert witness for the New England Coalition in the Vermont
Hopenfeld Declaration February 12, 2015 2
Yankee license renewal proceeding, where I testified at an adjudicatory hearing concerning metal fatigue.
- 5.
I reviewed the April 30, 2007 License Renewal Application (LRA) submitted by Entergy to renew the operating licenses for Indian Point Units 2 and 3, and assisted Riverkeeper with the preparation of Contention TC-1, which challenged Entergys aging management plan for addressing metal fatigue at Indian Point during the proposed period of extended operation.
- 6.
I reviewed Entergys January 22, 2008 amendment to its original LRA, in which Entergy purported to provide additional information regarding its aging management program for addressing metal fatigue, and assisted Riverkeeper with the preparation of an amended contention (Riverkeeper Contention TC1-A) to articulate the ongoing deficiencies with Entergys plan to deal with metal fatigue.
- 7.
I reviewed Entergys August 10, 2010 Notification of Entergys Submittal Regarding Completion of Commitment 33 for Indian Point Units 2 and 3, NL-10-082, as well as Entergys revised metal fatigue evaluations dated June 2010, and assisted Riverkeeper with the preparation of a new and amended consolidated contention (New York State 26-B/Riverkeeper TC-1B), which articulated various deficiencies with Entergys revised analysis, as well as Entergys ongoing failure to demonstrate that the affects of metal fatigue would be adequately managed at the Indian Point facilities during the proposed period of extended operation.
- 8.
I reviewed NRC Staffs Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Numbers 2 and 3, License Renewal Application, dated February 10, 2011, Entergys response thereto (Response to Request for Additional Information (RAI), Aging Management Programs, Indian Point Nuclear Generating Unit Nos. 2 & 3, Docket Nos. 50-247 and 50-286, License Nos. DPR-26 and DPR-64, dated March 28, 2011), and NRC Staffs Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3, Supplement 1 dated August 2011 (SER Supplement 1), all of which implicated and discussed metal fatigue, and assisted Riverkeeper with the preparation of joint contention NYS-38/RK-TC-5, concerning, among other things, Entergys failure to demonstrate that it has an adequate program for managing the aging effects of metal fatigue at Indian Point during the proposed period of extended operation.
- 9.
I have now reviewed NRC Staffs Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3, Supplement 2 dated November 2014 (SSER2), as well as various Entergy documents, including, though not limited to, Entergys February 2012 Revised Reactor Vessel Internals Program and Inspection Plan (NL-12-037), and subsequent communications between Entergy and NRC Staff that developed and amended that plan. After a review of these documents, for the reasons explained more fully below, it is my professional opinion that Entergy has, to date, failed to demonstrate that the effects of metal fatigue on RVI components will be adequately managed at the Indian Point facilities during the proposed period of extended operation.
Hopenfeld Declaration February 12, 2015 3
- 10. It is my understanding that Entergy intends to rely on its Fatigue Monitoring Program (FMP) to manage the effects of fatigue on RVI components at Indian Point during the period of extended operation, and that in order to account for the effects of the reactor coolant environment on the fatigue of RVI components as required under the ASME Code Section III, Subsections NG-2160 and NG-3121, Entergy has committed, in regulatory commitment 49, to recalculating CUF values for RVI components to include reactor coolant environmental effects.
According to Entergys implementation schedule for regulatory commitment 49, Entergy completed such calculations with respect to Indian Point Unit 2 before September 28, 2013, and will complete such calculations with respect to Indian Point Unit 3 before December 12, 2015.
See, e.g., SSER2 at 3-51 to 3-52.
- 11. Based on my review of the SSER2 and NUREG guidance documents, as well as my extensive professional experience, it is apparent that Entergys commitment to recalculate the CUF values of RVI components to include reactor coolant environment effects involves a flawed methodology that fails to accurately and fully account for environmental effects, and, thus, assure that fatigue of such components will be adequately managed during the period of extended operation.
- 12. Entergys commitment 49 involves calculating CUF values of the limiting RVI components to include reactor coolant environmental effects using factors, known as Fen, that are provided in NUREG/CR-5704 and NUREG/CR-6909. See SSER2 at 3-52. However, Entergys reliance on the Fen equations provided in these guidance documents results in a flawed and non-conservative analysis, for several reasons.
- 13. Entergys reliance on NUREG/CR-5704 and NUREG/CR-6909 means that the CUFen calculations conducted thus far and those that will be conducted in the future have and will continue to exclude the synergistic effects of neutron irradiation/radiation on metal fatigue of RVI components, thus resulting in non-conservative fatigue life predictions. In particular, the Fen equations contained in NUREG/CR-5704 and NUREG/CR-6909 were derived from tests on smooth specimens in the absence of neutron irradiation/radiation effects and under controlled water chemistry. Thus, Fen for the fatigue of reactor vessel internals must be corrected to encompass and account for effects of radiation. In fact, Dr. O.K. Chopra, the author of NUREG/CR-5704 and NUREG/CR-6909, has specifically stated that the effects of radiation flux, or fluence, were not included in the Argonne National Laboratory (ANL) studies contained in NUREG/CR-6909, and that therefore the user of the Fen equations contained therein must account for the effects of fluence on the Fen and resulting CUFen. See Transcript of Advisory Committee on Reactor Safeguards Subcommittee on Materials, Metallurgy and Reactor Fuels (Dec. 6, 2006), ADAMS Accession No. ML12335A532, at 46-7 (Official Hearing Exhibit RIV000037). However, Entergy has simply used or committed to use the non-conservative equations contained in NUREG/CR-5704 and NUREG/CR-6909, and thus, has failed to, or will fail to, without any justification, include or account for the effects of radiation in their calculation of CUFen for RVI components.1 1 There is currently a draft revision to NUREG/CR-6909 (March 2014) out for public comment (Revision 1). As a draft, this report ostensibly does not apply to this proceeding. In any event, the revised report, which memorializes the results of an extensive review that was conducted of the literature on the effects of radiation on fatigue, demonstrates that the Fen as specified by the NUREG/CR-6909 and NUREG/CR-5704 equations include
Hopenfeld Declaration February 12, 2015 5
TC-1 and NYS-38/RK-TC-5: the use of these equations fails to adequately account for various critical parameters affecting fatigue. For example, the use of the Fen equations from NUREG/CR-5704 and NUREG/CR-6909 improperly accounts for variation in dissolved oxygen (DO) during thermal transients. To correct for this deficiency NUREG/CR-5704 and NUREG/CR-6909 specify that maximum DO and temperature values should be used during the transients for conservative calculations. NUREG/CR-6909 provides additional guidance that has not been observed in Entergys fatigue evaluations to date. The use of the Fen equations provided in NUREG/CR-5704 and NUREG/CR-6909 per se, without incorporating the specified instructions attached to these equations, results in significant underestimation of CUFen values, especially for carbon steel, and does not assure that the effects of fatigue will be adequately managed during the PEO.
- 18. In addition to uncertainties in the CUFen due to incorrect accounting for dissolved oxygen and radiation Entergys methodology also introduces another major uncertainty by once again apparently relating the CUFen to the CUF of record, i.e. CUFen = Fen x CUF (of record).
The CUF of record is based on calculations that were valid when the plants were initially designed because all components were presumably in pristine conditions. However, after 40 years of exposure to a hostile light water reactor (LWR) environment most of the components have undergone a change in geometry and surface structure due to erosion/corrosion, stress corrosion, swelling, pitting, and cavitation. Such changes are known to affect fatigue life because they introduce local discontinuities that introduce high local stress concentrations. Such stress concentrations are known to significantly reduce fatigue life. The ASME fatigue curves are based on average stresses only. At least 100 years of experience has been accumulated to show that sharp surface discontinuities introduce high local stress concentrations where cracks are initiated. The ASME code requires that the average stress of a component be multiplied by the appropriate stress intensity factor. For example, the fatigue life of a component with lathe-formed surface is lower by a factor of 10 than if that surface was superfinished. See S.
McKelvey & A. Fatemi, Effect of Forging Surface on Fatigue Behavoir of Steels: A Literature Review (University of Toledo) (citing and discussing P. Fluck, 1951, Influence of surface roughness on the fatigue life and scatter of test results of two steels, Proceedings of American Society for Testing and Materials, Vol. 51, pp. 584-592, Am. Soc. of Testing Materials, Philadelphia, PA). It is highly unlikely the stress concentration factors for many reactor components remain the same after 40 years. NUREG/CR-6909 also discusses the importance of surface finish. Effect of LWR Coolant Environment on the Fatigue Life of Reactor Materials, Final Report, NUREG/CR-6909, ANL-06/08 (February 2007), at § 4.1.6. Relying on the CUF of records without allowing for surface changes during service completely ignores the overwhelming importance of surface topography on fatigue life. It would be reasonable to multiply all CUFs of record with potential for surface change by a factor of 10 to account for real life effects of surface deterioration during service. Without such a correction there is no scientific basis for a claim that the CUFens are conservative.
- 19. In the absence of the appropriate consideration of radiation effects and other critical parameters, a CUFen value that is less than 1.0 does not necessarily indicate that fatigue issues will not arise during the PEO. Importantly, a CUFen of less than one does not necessarily
Hopenfeld Declaration February 12, 2015 8
analyses, evaluations of the uncertainties must be discussed and analyzed in terms of the change in core melt frequency.
- 24. In addition, in Commitment 49 Entergy proposed (and NRC Staff have agreed) that it need not complete the calculation of CUFen values for IP3 RVI components until December 2015. See SSER2, at A-15. The schedule for Entergys resolution of this issue extends beyond the time frame for the hearings in this ASLB proceeding and thus will not allow for a testing of the adequacy of the proposed resolution of these issues in this proceeding. That timeline likely will prevent Riverkeeper from playing any meaningful role in their development or resolution.
Entergys commitment to complete CUFen evaluations for RVI components in the future, without review by the public, the ASLB, and/or NRC Staff, is unacceptable. By merely making a future commitment, it is not possible to fully determine the adequacy of the calculated CUFen values and Entergys AMP for RVI components. The analysis must be performed before a determination is made about license renewal. NRC Staffs acceptance of Entergys commitment 49 in the SSER2 to conduct environmentally corrected metal fatigue evaluations for RVI components at some time in the future is not warranted or acceptable. By failing to undertake and complete the CUFen analysis now and not in the future, Entergy has failed to demonstrate that metal fatigue of RVI components will be adequately managed during the PEO.
- 25. Furthermore, Entergy has not yet developed inspection acceptance criteria for baffle former bolts in either IP2 or IP3. See SSER2, at 3-20. Instead, it has agreed to develop a technical justification including acceptance criteria for baffle former bolts sometime prior to the first round of anticipated inspections, which might not occur until 2019 for IP2 and 2021 for IP3 and would be after an evidentiary hearing in this proceeding. See SSER2, at 3-20. The schedule for Entergys resolution of this issue extends beyond the time frame for the hearings in this ASLB proceeding and thus will not allow for a testing of the adequacy of the proposed resolution of this issue in this proceeding. That timeline likely will prevent Riverkeeper from playing any meaningful role in their development or resolution.
- 26. In light of the foregoing, Entergy has failed to demonstrate that it has a program to monitor, manage, and correct metal fatigue related degradation sufficient to comply with 10 C.F.R. § 54.21(c), or the regulatory guidance of NUREG-1801, Generic Aging Lessons Learned (GALL) Report, or the MRP-227-A guidance developed by EPRI.