ML20216C909
| ML20216C909 | |
| Person / Time | |
|---|---|
| Issue date: | 05/06/1998 |
| From: | Wetzel B NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9805190435 | |
| Download: ML20216C909 (36) | |
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NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20666-0001
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liay 6, 1998 MEETING PARTICIPANTS: NUCLEAR ENERGY INSTITUTE (NEI) AND ELECTRIC POWER RESEARCH INSTITUTE (EPRI)
SUBJECT:
MEETING WITH NEl AND EPRI TO DISCUSS EPRI REPORT, TR-108812, " RESPONSE OF ISOLATED PIPING TO THERMALLY INDUCED OVERPRESSURIZATION DURING A LOSS OF COOLANT ACCIDENT" On March 25,1998, members of the NRC staff met with representatives from NEl and EPRI to discuss EPRI report, TR-108812, " Response of Isolated Piping to Thermally Induced Overpressurization During a Loss of Coolant Accident." The list of attendees is contained in. The presentation slides used by EPRI are contained in Enclosure 2.
NEl submitted TR-108812 to the NRC on January 15,1998, and requested a meeting with the staff to discuss the contents of the report. The report was developed to provide technical support for a proposed American Society of Mechanical Engineers (ASME) Code Case (N-584) addressing the thermal overpressurizat:on of isolated sections of piping issue in Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Cesign-Basis Accident Conditions." The staff provided its comments on the report in a letter to NEl dated February 23,1998 (Enclosure 3). EPRI addressed the NRC staffs comments during its presentation at the meeting.
The staff indicated that several of the comments in the February 23,1998, letter were based on the acceptance criteria for strain limits proposed in EPRI TR-108812. The staff considers the
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proposed use of a limit based on 70% of effective strain at fracture to be less conservative than
/
the current ASME Code Appendix F criteria. The staff noted that EPRI tested only a limited j
number of pipe specimens to assess the fracture strain. These tests were simple, straight pipe
/
geometries with no additionalloadings. The staff is concerned with uncertainties that may exist in estimating the fracture strains of actual plant components from a limited number of laboratory tests. The staff also noted that a previous proposal to incorporate similar strain criteria into y
Appendix F had not been adopted.
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On April 7,1998, a followup conference call was held between the NRC staff, EPRI, and NEl.
The staff reiterated its concerns regarding the uncertainty in the effective strain limit at fracture for actual plant components. In particular, the staff is concemed with the proposed strain limit for carbon steel components. The staff pointed out that the EPRI Slide 5 in the meeting handouts demonstrates a significant sensitivity to additional axial stresses at a hoop strain of 5%.
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. If you have a1y questions regarding the meeting or the staff's review of EPRI's technical report, please call me at (301) 415-1355.
0 Beth A. Wetzel, Senior roject Manager Project Directorate 111-1 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation
Enclosures:
As stated (3) cc w/encis See next page DISTRIBUTION:
Central File PUBLIC PD# 3-1 Reading B. Wetzel OGC ACRS E-Mall S. Collins /F. Miraglia (SJC1/FJM)
B. Boger (BAB2)
E. Adensam (EGA1)
C. Carpenter (CAC)
C. Jamerson (CAJ1)
T. Martin (SLM3)
R. Wessman K. Manoly K. Wichman J. Fair G. Hammer J. Tatum M. Hartzman B. McCabe (BCM)
M. Tschiltz (MDT)
T. Hiltz (TGH)
D. Lange (DJL)
C.Hehl L. Plisco G. Grant T. Gwynn K. Perkins D. Screnci (DPS)
K. Clark (KMC2)
J. Strasma (RJS2)
B. Henderson (BWH)
M. Hammond (MFH2)
OPA DOCUMENT NAME: G:\\WPDOCS\\GL9606\\MTGSUM.498 To receive a copy of this document, indicate in the box C= Copy w/o attachment / enclosure E= Copy with 1
attachment / enclosure N = No copy l
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OFFICIAL RECORD COPY
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NEl and EPRI cc:
David J. Modeen Director, Engineering Nuclear Generation Division 1776 i Street, NW Suite 400 Washington, DC 20006-3708 Dr. Avtar Singh 3412 Hillview Avenue Palo Alto, CA 94303 l
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LIST OF ATTENDEES NAME AFFILIATION l
R. Wessman NRC K. Manoly NRC K. Wichman NRC J. Fair NRC G. Hammer NRC B. Wetzel NRC J. Tatum NRC M. Hartzman NRC E. Forrest Lls H. Tang EPRI K. Cozens NEl A. Singh EPRI E. Wais EPRI Consultant S. Gosselin EPRI i
ENCLOSURE 1
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UNITED STATES l
A NUCLEAR REGULATORY COMMISSION U
\\*****/[
l WASHINGTON, D.C. 30e06 4001 February 23, 1998 Mr. David J. Modeon Director, Engineering Nuclear Generation Division Nuclear Energy Institute 1776 i Street, NW, Suite 400 l
Washington, D.C. 20006 l
l
SUBJECT:
REVIEW OF EPRI TECHNICAL REPORT TR-108812,
- RESPONSE OF ISOLATED PIPING TO THERMALLY INDUCED OVERPRESSURIZATION DURING A LOSS OF COOLANT ACCIDENT (TAC No. MA0695)
The Nuclear Energy institute (NEI) submitted Electric Power Research Institute (EPRI) report TR-108812, " Response of isolated Piping to Thermally Induced Overpressurization During a Loss of Coolant Accident," to the NRC on January 15,1998, for staff review. This report was developed to provide technical support for a proposed American Society of Mechanical l
Engineers (ASME) Code Case addressing the thermal overpressurization of isolated sections of piping issue in NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions." NEl requested a meeting with the staff to discuss the content of TR 108812.
The NRC staff has reviewed the EPRI report TR-108812. Our comments on the report are enclosed. We are transmitting these comments to you in anticipation of a proposed meeting on the subject report, which is tentatively set for March 25,1998. A formal meeting notice will be issued shortly. During the staffs review of TR-108812, the staff identified several pertinent items that are not addressed, items where additional information is needed, and items that require further clarification and technical validation. These items should be resolved prior to the staffs consideration of the EPRI report in support of the proposed ASME Code Case.
l l
Sincerely, u
A 2
Both A. Wetzel, Senio roject Manager Project Directorate ill-1 Division of Reactor Projects-lll/IV I
Office of Nuclear Reactor Regulation l
l
Enclosure:
As stated 1
ENCLOSURE 3 L
l
~
I Comments Regarding EPRI Report TR-108812 1.
Section 1.2 of the report describes its purpose. The report presents the results of EPRl's Phase 1 Generic Letter (GL) 96-06 Testing Pogram. The report indicates that EPRI believes the results of the testing presented in the report will provide a technical basis for the acceptance of proposed ASME strain limits. However, the testing involved only three simple pipe geometries. The testing does not address the following issues:
a.
The impact of other design loads on the predicted strains. These other design loads may be sustained loads due to deadweight or suppressed thermal expansion of the pipe run, or they may be dynamic loads due to seismic events.
b.
The impact of local attachments on predicted strains has not been assessed. Many of the piping runs of concem contain test connections.
c.
The applicability of the test results to pipe runs containing fittings such as elbows and tees.
d.
The impact of potential flaws in the piping on the predicted hoop strain at failure 2.
Section 2.3 of the report provides the ultimate stress and ultimate strain values for each pipe material heat. These are the average values of four tensile specimens tested for each pipe material heat. The actual test values should have been provided in the report. Section 2.2 of the report contains a list of materials obtained from a plant survey. In order to assess the applicability of the test results to components in the plants, the potential range of the ultimate stress and ultimate strain values for the carbon and stainless steel materials listed in Section 2.2 of the report should be discussed.
3.
Section 2.6 of the report describes the hydrostatic burst tests on the pipe specimens. The following information on these test specimens was not provided in the report:
a.
The initial dimensions of the pipe specimens that were burst tested. The initial diameter and thickness should have been measured at severallocations on the pipe prior to the burst test.
b.
The final dimensions of the pipe specimens after the burst tests. The final diameter and thickness should have been measured at the same locations on the pipe after the burst test.
c.
The method used to calculate the burst hoop strain values reported in Table 2 5.
4.
Section 4.1 of the report introduces the concept that the loading addressed by GL 96-06 is an " energy controlled condition." The term " energy contro!!ad condition" needs to be clearly defined. Since any pressurized system will contain a finite amount of intomal energy, is the concept applicable to intomat pressure in general? If not, then there must be some definitive criteria to differentiate between " load controlled" and " energy controlled" pressure conditions.
Enclosure
I l
i 6
2 5.
Section 4.2 of the report specifies strain criteria taken from EPRI technical report NP-1921,
" Rationale for a Standard on the Requalification of Nuclear Class 1 Pressuro-Boundary Components." In EPRI technical report NP-1921, the strain criteria was recommended for
" energy-controlled" events. NP-1921 does not define an " energy-controlled" event. The i
ASME Section lll Special Working Group on Faulted Conditions considered the criteria proposed in EPRI technical report NP-1921 approximately 10 years ago. The NRC staif representative voted negative on the proposal. This criteria was never adopted by the Code for incorporation into Appendix F. It appears that the technical issues regarding this criteria were never resolved by the special working group.
6.
Section 5.3.4 of EPRI technical report NP-1921 contains the theoretical basis for the proposed strain criteria. The theory predicts that a deformation instability in a pressurhed cylinder occurs at a hoop strain that is considerably higher than the strain predicted at lud instability. However, the burst hoop strain for the cart >on steel component reported in Table 2-5 of TR-108812 is considerably lower than the theory would predict. For example, Equation 5-16 in NP-1921 predicts a hoop Mrain of over 28% for the cart >on steel specimen before occurrence of deformation instability. In Table 4-3 of TR 108812, a hoop strain of over 16% is predicted prior to ductile tearing. However, the measured hoop burst strain reported in Table 2 5 of TR-108812 is less than 9%. The theory does not appear to correlate very well with test results for hoop strain. This discrepancy between the underlying theory and the actual test results needs to be resolved.
e