ML20154C182
| ML20154C182 | |
| Person / Time | |
|---|---|
| Issue date: | 10/01/1998 |
| From: | Wetzel B NRC (Affiliation Not Assigned) |
| To: | Modeen D NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT & |
| References | |
| GL-96-06, GL-96-6, TAC-MA0695, TAC-MA695, NUDOCS 9810060182 | |
| Download: ML20154C182 (5) | |
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October 1,1998 Mr. D::vid J. Mods:n Director, Engineering Nuclear Generation Division Nuclear Energy Institute 1776 i Street, NW, Suite 400 Washington, D.C. 20006
SUBJECT:
REVIEW OF EPRI TECHNICAL REPORT TR-108812, " RESPONSE OF ISOLATED PlPING TO THERMALLY INDUCED OVERPRESSURIZATION DURING A LOSS OF COOLANT ACCIDENT (TAC NO. MA0695)
The Nuclear Energy Institute (NEI) submitted Electric Power Research Instittte (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 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." The staff provided comments on the report to NEl in a letter dated February 23,1998. The staff met with representatives from NEl and EPRI to discuss these comments in a meeting on March 25,1998.
Subsequently, NEl transmitted EPRI's written response dated June 10,1998, to the staff's comments as an attachment to a June 15,1998 letter.
The EPRI response has not resolved all of the staff's concerns with TR-108812. The staff does not believe that the EPRI testing provides a sufficient technical basis to support the strain limits proposed in the September 3,1998, draft of ASME Code Case N-584. The staff's specific comments regarding NEl's June 15,1998 letter are enclosed. If you have any questions regarding this issue, please contact me at (301) 415-1355.
Sincerely, ORIGINAL SIGNED BY:
Beth A. Wetzel, Senior Project Manager e ate M 9810060182 981001 0I"..
PDR REVGP ERONUMRC Division of Reactor Projects-lll/IV PDR Office of Nuclear Reactor Regulation
Enclosure:
As stated I
cc w/ encl: Mr. Kurt Cozens Project Manager Nuclear Generation Division
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Mr. David J. Modeen Director, Engineering Nuclear Generation Division Nuclear Energy Institute
- 1776 i Street, NW, Suite 400 Washington, D.C. 20006
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) l l
The Nuclear Energy Institute (NEI) submitted Electric Power Research Institute (EPRI) report TR-108812 " Response of isolated Piping to Thermally Induced Overpress'urization 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 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." The staff provided comments l
on the report to NEl in a letter dated February 23,1998. The staff met with representatives i from NEl and EPRI to discuss these comments in a meeting on March 25,1998.
Subsequently, NEl transmitted EPRl's written response dated June 10,1998, to the staff's comments as an attachment to a June 15,1998 letter.
The EPRI response has not resolved all of the staff's concems with TR-108812. The staff does not believe that the EPRI testing provides a sufficient technical basis to support the strain limits proposed in the September 3,1998, draft of ASME Code Case N-584. The staff's specific comments regarding NEl's June 15,1998 letter are enclosed. If you have any questions regarding this issue, please contact me at (301) 415-1355.
Sincerely, hhv h, Beth A. Wetzel, Sen r Project Manager Project Directorate 111-1 Division of Reactor Projects-lll/IV i Office of Nuclear Reactor Regulation I
Enclosure:
As stated cc w/ encl: Mr. Kurt Cozens Project Manager Nuclear Generation Division Nuclear Energy Institute 1776 i Street, NW, Suite 400 !
Washington, D.C. 20006
. j Staff Assessment of EPRI June 10,1998, Response to NRC Comments Regarding EPRI Report TR-108812 1.a. The staff indicated that the EPRI testing did not address the impact of other design l 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 l loads due to seismic events.
In response to the staff comment, EPRI performed an assessment of the impact of a sustained axial stress on calculated hoop strain in a pressurized pipe segment. The results are shown in Figures 1 and 2 of the EPRI response. On the basis of this assessment, EPRI concluded that the contribution of sustained loads on predicted hoop l strain is negligible. The staff does not agree with this conclusion. Review of Figure 1 l indicates that at a hoop strain of 4%, the addition of a 10 ksi axialload would increase I the predicted hoop strain to 5.5%. This is a significant increase. The Code-allowable l stress limit for sustained loads and the Code-allowable stress limit for thermal expansion loads are both greater than 10 ksi. Consequently, the staff believes that bending stresses on the order 10 ksi or more due to the combination of deadweight and thermal 4 expansion loads are possible.
With regard to dynamic loads such as seismic events, EPRI argued that the concurrent ,
combination of seismic and loss-of-coolant accident (LOCA) loads is highly unlikely.
However, the staff concern was with the case where LOCA and seismic loads would have to be combined to meet a licensing-basis commitment. The EPRI response did not address this issue.
1.b,c. The staff indicated that the impact of local attachments on predicted strains has not been assessed. Many of the piping runs of concern contain test connections. The staff also questioned the applicability of the test results to pipe runs containing fittings such as elbows and tees.
In its response, EPRI stated that, although the local attachment will increase the localized peak strains due to the stresc concentrations caused by the discontinuity, the local discontinuity does not affect the general plastic membrane strain. However, no test data was presented in support of the argument. The staff is still concerned that attachments and other fittings could experience local membrane strains (not necessarily peak strains) which are greater than the average hoop membrane strain in the pipe.
The staff believes that further testing and/or analysis is needed to address this concern.
1.d. The staff indicated that the testing did not address the impact of potential flaws in the piping on the predicted hoop strain at fracture.
In response to the staff comment, EPRI performed an evaluation of impact of potential flaws on the calculated hoop membrane strain at failure for carbon steel pipe. The results of the evaluation indicate that only relatively deep circumferential cracks impact j the calculated hoop membrane failure strain. However, the EPRI evaluation indicates ENCLOSURE l
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2 that long axial cracks can have a significant impact on the calculated hoop membrane failure strain. Figure 8 of the EPRI response shows the hoop membrane strain at the calculated failure pressure for long axial cracks in the carbon steel pipe. Figure 8 indicates that the membrane hoop strain at failure would be less than 4% for long axial cracks greater than 15% through the thickness. The EPRI evaluation indicates that there is little tolerance for long axial cracks in carbon steel piping at a hoop membrane strain of 4%. The staff believes that the strain acceptance criteria must have adequate l margin to accommodate potential material variations, additional loads not accounted for in the evaluation, and potential flaws that may exist in the piping system.
l 2. The staff indicated that the actual test data for the tensile specimens should have been
! provided in the EPRI report. In addition, the staff indicated that variability of the ultimate l stress and ultimate strain values for the carbon and stainless steel materials listed in l
Section 2.2 of the EPRI report should have been discussed.
l In response to the staff comment, EPRI provided the data from the tensile tests. In addition, EPRI provided test data from the literature on stainless and carbon steel materials. EPRI concluded that the material properties in its test program are consistent with the comparable material properties published in the literature. The staff was
! concerned that a very limited amount of test data was used to assess margins associated the strain limits in the proposed ASME code case. As discussed in item 1.d, the staff believes that the potential variation of material properties should be considered in establishing the acceptance criteria. For example, the ASME Code Appendix F inelastic criteria allow the use of material strain curves developed from test data.
l However, the criteria also require the results be adjusted to account for code minimum j properties. This accounts for potential material property variations of installed i components. The staff believes that material property variations should be considered in establishing strain limits.
i 3. The staff indicated that additional detailed measurements of the pipe specimens that were burst tested should have been provided in the report. In addition, the staff indicated that the method used to calculate the hoop strain values at burst should have been provided.
EPd responded that the initial dimensions were nominal and that post burst diametrical measurements were deemed to be of little value. The staff was interested in the detailed pipe diameter measurements to determine whether the hoop strain in the specimen remained uniform during the burst test. This is an important consideration if the strain in the pipe is compared to the volumetric expancion of the fluid. A nonuniform strain distribution along the length of the pipe would result in a greater maximum hoop strain for a given volumetric expansion of the fluid. In discussions with the staff, EPRI indicated that the radial deformation of the pipe was uniform in the creas away from the
- structural discontinuities in the test specimens. The staff believes that detailed diameter
. measurements along the pipe would have provided useful information in assessing this issue.
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EPRI also indicated that the hoop strain values at burst were based on the displacement measurements obtained up to the point of rupture. This provided the information that the staff requested.
- 4. The staff questioned the assertion that the loading addressed by GL 96-06 is an " energy l controlled condition." The term " energy controlled condition" was used to justify an l
acceptance criteria proposed in EPRI Technical Report NP-1921, " Rationale for a l Standard on the Requalification of Nuclear Class 1 Pressure-Boundary Components."
i The acceptance criteria are based on the concept that an allowable limit of 70% of the l
failure strain for an " energy controlled condition" is equivalent to the ASME Code Appendix F criteria of 70% of the ultimate stress for a load controllad condition.
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The EPRI response contains a discussion of the failure strain due to a load controlled test verses the failure strain due to a deformation controlled test. This discussion and the discussion in item 6 indicate that a larger failure strain would have been measured in a deformation controlled test. The staff does not agree with the inference that the pipe specimens would have failed at a higher hoop strain value if tested with a GL 96-06 loading. Thre staff believes that test data is needed to substantiate the EPRI claim. 4 i
- 5. The staff indicated that the ASME Section til Special Working Group on Faulted Conditions considered the criteria proposed in EPRI Technical Report NP-1921 !
approximately 10 years ago. The NRC staff representative voted negative on the proposal. These criteria were never adopted by Section 111 for incorporation into Appendix F of the code. It appears that the technicalissues regarding the criteria were never resolved by the special working group.
EPRI responded that it was not aware of any staff concerns with EPRI Report NP-1921.
It is the staff's understanding that the Appendix F working group discussed the criteria in NP-1921 for use with dynamic impact loads. Whereas the use of energy considerations for evaluating impact loads is a common engineering design practice, extrapolation of criteria based on this concept to hoop membrane stress caused by internal pressure is highly questionable. The staff's technical concern regarding NP-1921 involved equation 5-24. This equation is based on a theoretical failure criteria developed by F. A.
McClintock. The theory had not been compared to test data to verify its applicability to materials used in nuclear power plants. In addition, the staff did not agree with the concept of using 70% of the failure strain as an acceptance criterion.
- 6. The staff indicated that the measured hoop burst strain for the carbon steel specimen reported in Table 2-5 of EPRI Report TR-108812 is less than 9%. The staff also indicated that the strain computed using the underlying theory presented in NP-1921 does not appear to ccrrelate very well with test results for hoop strain.
The EPRI response discusses the concept of load-controlled verses energy-controlled loading. EPRI concludes that the measured burst strain is consistent with the theory for l
a load-controlled failure. However, the discussion infers that a greater strain would have ,
been measured for an " energy-controlled loading." As discussed in item 4 above, the staff believes that the concept that the pipe would have failed at a greater strain if it were subjected to a GL 96-06-type loading condition needs to be confirmed by testing.
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