ML20197B100
| ML20197B100 | |
| Person / Time | |
|---|---|
| Issue date: | 02/23/1998 |
| 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 9803100120 | |
| Download: ML20197B100 (3) | |
Text
.___-_-_____-___- - _____ - . --
e February 23, 1998 Mr. David J. Modeon Director, Engineering Nuclear Generation Division Nuclear Energy Institute i 1776 l Street, NW, Suite 400 Washington, D.C. 20008
SUBJECT:
REVIEW OF EPRI TECHNICAL REPORT TR 108812, " RESPONSE OF ISOLATED FIPING TO THERMALLY INDUCED OVERPRESSURIZATION DURING A LOSS OF COOLANT ACCIDENT (TAG No. MA0695) I j
The Nuclear Energy institute (NEl) 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 l, s 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." NEl requested a me'eting~with l .i e staff to discuss the content of TR 108812. -
i 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 snould be resolved prior to the staffs -
+ consideration of the EPRI report in support of the proposed ASME Code case, l
Sincerely, Original signed by:
Beth A. Wetzel, Senior Project Manager Project Directorate 1111 Division of Reactor Projects lll/IV Office of Nuclear Reactor Regulation
Enclosure:
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,. Comments Regarding EPRI Report TR 108812
- 1. Section 1.2 of the report desertbes its purpose. The report presents the results of EPRI's Phase 1 Generic Letter (GL) 96-06 Testing Program. 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 vanes of tour tensile specimens tested for esin 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 rangs of the __
ultimate stress and ultimate stra'n 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 infonnation on these test specimens was not provided in the report:
- a. The initial dimensions of the ploe specimens that were burst tested. The initial diameter and thickness shou!d 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 b9en 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 controlled condition" needs to be clearty defined. Since any pressurized system will contain a finite amount of intemal energy, is the concept applicable to intomal pressure in generali If not, then there must be some definitive criteria to differentiate between "loact controlled
- and
- energy controlled' pressure conditions.
Enclosure
+ ?
i e
y 3 5. Section 4.2 of the report specifies strain artierta taken from EPRI technical report NP 1921,
' Rationale for a Standard on the Requalification of Nuclear Class 1 Pressure Boundary Components.' in EPRI technical report NP 1921, the strain critoria was recommended for
' energy controlled
- events. NP 1921 does not define an 'onergy-controlled" event. The ASME Section ill Special Working Group on Faulted Conditions considered the criteria oroposed in EPRI technical report NP 1921 approximately 10 years ago. The NRC staff representative voted negative on the proposal. This criteria was never adopted by the Code for incorporstion into Appendix F. It appears that the technicalissues regarding this criteria j were never resolved by the special working group.
j 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 pressurized cylinder occurs at a hoop strain that is considerably higher than the strain predicted at load Instability, However, the burst hoop strain for the cart >on steel component reported in Tabie 2 5 of TR 108812 is considerably lower than the theory would piedict. For example, Equation 516 in NP 1921 predicts a hoop strain of over 28% for the carbon steel specimen i 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 i'
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.
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