ML17265A554
| ML17265A554 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/25/1999 |
| From: | NRC (Affiliation Not Assigned) |
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| Shared Package | |
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| References | |
| NUDOCS 9903020242 | |
| Download: ML17265A554 (22) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATIONBYTHE OFFICE OF NUCLEAR REACTOR REGULATION RE VEST TO APPLY LEAK-BEFORE-BREAKSTATUS TO PORTIONS OF THE RESIDUAL HEAT REMOVALSYSTEM PIPING ATTHE R. E. GINNA NUCLEAR POWER PLANT ROCHESTER GAS AND ELECTRIC COMPANY DOCKET NO. 50-244
1.0 INTRODUCTION
By fetter dated November 11, 1997, Rochester Gas and Electric Company (RG&E), the licensee for the R. E. Ginna Nuclear Power Plant (Ginna), requested that the NRC review and approve their application to remove consideration of the dynamic effects of postulated ruptures of portions of the Ginna residual heat removal (RHR) system piping from the facility's licensing basis.
RG8E's submittal was based on an application of Title 10 of the Code of Federal Regulations Part 50, Appendix A, General Design Criteria 4, which states:
However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
For the purposes of this demonstration, RG&E submitted a leak-before-break (LBB) analysis prepared by Structural Integrity Associates (SIA) for the subject portions of the RHR system piping. LBB evaluations developed using the analysis methodology contained in NUREG-1061, Volume 3, "Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks," (Reference
- 1) and/or Draft Standard Review Plan (DSRP) Section 3.6.3 have been previously approved by the Commission as demonstration of an extremely low probability of piping system rupture.
2.0 REGULATORYRE UIREMENTS AND STAFF POSITIONS Nuclear power plant licensees have, in general, been required to consider the dynamic effects which could result from the rupture of sections of high energy piping (fluid systems that during normal plant operations are at a maximum operating temperature in excess of 200 'F and/or a maximum operating pressure in excess of 275 psig). This requirement has been formally included in 10 CFR Part 50, Appendix A, General Design Criteria 4 which states, "Structures, systems, and components important to safety....shall be appropriately protected against 9903020242 990225 PDR ADOCK 05000244 P,'
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dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit." For facilities such as Ginna,'which were licensed prior to the advent of the General Design Criteria, these requirements were included as part of plant-specific licensing reviews.
A condition at Ginna was identified during a Design Inspection Team audit in which sections of dosed loop piping near the unit's American Society for Mechanical Engineers (ASME) Code Class 1 RHR system piping would not be adequately protected from a failure of RHR piping. This finding was documented by the NRC via letter dated September 14, 1997 (Reference 2), and was designed as Inspector Follow-up Item 97-201-03.
Per the discussion in Section 1.0 above, RG&E has addressed this problem by performing an LBB evaluation of the subject RHR piping. The philosophy of "leak-before-break" behavior for high energy piping systems was developed by the NRC in the early 1980s, used in certain evaluations stemming from Unresolved Safety Issue A-2, "Asymmetric Blowdown Loads on PWR Primary Systems,"
and was subsequently expanded for application toward resolving issues regarding defined dynamic effects from high energy piping system ruptures.
The methodology developed by the NRC for performing LBB analyses was thoroughly detailed in NUREG-1061, Volume 3 and summarized in Draft Standard Review Plan Section 3.6.3, "Leak-Before-Break Evaluation Procedures," which was published for public comment in August 1987.
3.0 LICENSEE'S DETERMINATION The following discussion contains information supplied by RG&E in its November 11, 1997 (Reference 3), fonvarding letter to the NRC and the attachments to that letter. These attachments included the report prepared by SIA for RG&E: SIR-97-077, Rev. 0, "Leak-Before-Break Evaluation of Portions of the Residual Heat Removal (RHR) System at R. E.
Ginna Nuclear Power Station.". It also includes information provided in the licensee's response (Reference 4), dated August 6, 1998, to the NRC staff's Request for Additional Information,.
dated May 18, 1998. Additional information regarding the facility's leakage detection system was also provided to the staff and is addressed in Section 4.4.
3.1 Identification of Anal zed Pi in and Pi in Material Pro erties RG&E's submittal identified and analyzed the following sections of high energy piping for LBB behavior verification. RG&E addressed the ASME Code Class 1 portion of the RHR system from its connection to the reactor coolant system hot leg to motor-operated valve 700 as shown in Figure 1-1 (Attachment 1). RG&E addressed the ASME Code Class 1 portion of the RHR system from its connection to the reactor coolant system cold leg to motor-operated valve 721, as shown in Figures 1-2 (Attachment 2) and 1-3 (Attachment 3) ~
The RHR system piping was identified as having the following material components.
The piping and fittings of the Ginna RHR system were manufactured from wrought ASME specification SA-376 Type 316 stainless steel (SS). The welds in this system were identified as having been fabricated from SS using either submerged arc welding (SAW) or shielded metal arc welding (SMAW) processes.
For the material properties used in the RHR system LBB evaluations, RG&E/SIA used stress-strain representations forwrought Type 316 SS based on a consideration of the ASME Code minimum values and Ramberg-Osgood parameters based on EPRI Report NP-5531 (Reference 5). J-resistance (J-R) curve data was obtained for SS SMAW materials from generic characterizations in the EPRI Ductile Fracture Handbook (Reference 6). These data were used together in the initial analysis submitted on November 11, 1998.
RG& E/SIA subsequently submitted additional stress-strain representations for wrought Type 316 SS and for SMAWweld material along with J-R curve data forwrought Type 316 SS in their, August 6,1998, RAI response.
This additional information was used for sensitivity studies in the RAI response.
Archival samples and/or test data specific to the Ginna materials were not available.
3.2 General As ects of the Licensee's LBB Anal sis The analyses provided by RG&E/SIA sought to address the following four principal areas which were consistent with the criteria established for LBB analysis acceptability in NUREG-1061, Vol. 3 and/or DSRP Section 3.6.3:
4.
Demonstrate that the subject piping is a candidate for LBB analysis by showing that the piping is not particularly susceptible to active degradation mechanisms or atypical loading events.
Establish the critical through-wall flaw size under which analyzed locations would be expected to fail under normal operation (NOP) plus safe-shutdown earthquake (SSE) loading conditions.
Establish the leakage behavior of smaller through-wall flaws under NOP loads alone for each location.
Evaluate the margin between the critical through-wall flaw size and an appropriate leakage through-wall flaw size and the stability of the through-wall leakage flaw under loading conditions of K2 * (NOP+SSE) loads.
3.3 Evaluation of Residual Heat Removal S stem Pi in The analysis of the RHR system piping that was submitted to the staff as an attachment to the November 11, 1997, letter was prepared for the licensee by SIA as report number SIR-97-077, Rev. 0. This Section summarizes the results of the RG&E/SIA results for the four subject areas noted in Section 3.2 above.
Initially,the RG&E submittal addressed the issue of potential piping degradation mechanisms and atypical loading conditions.
Per the discussion of the limitations of LBB analyses in NUREG-1061, Volume 3, the LBB approach should not be considered when operating experience has indicated particular susceptibility to failure from the effects of corrosion, water hammer, or fatigue. RG&E's submittal concluded that pressurized-water reactor RHR system piping like that at Ginna has not been shown to'be particularly susceptible to the effects of water hammer, intergranular stress corrosion cracking, or flow-assisted corrosion.
RG&E/SIA included a fatigue analysis which indicated that the circumferential growth of postulated surface flaws (which were in excess of the size allowed following preservice inspection by the ASME Code) due to cyclic stresses would not be significant based on analysis of node 910 in the RHR cold leg since this location has the maximum thermal stress range.
Any significant through-wall growth of larger surface flaws without circumferential extension was concluded to be within the technical basis of LBB behavior.
Next, the RGB E/SIA analysis evaluated the RHR system piping by developing the applied stresses under NOP plus SSE loading from the facility's piping stress reports (References 7 and 8) and determining the critical through-wall flaw size for various locations along the piping.
In the determination of the applied stresses, the analysis included the tensile stress resulting from the internal pressure, and the bending stress resulting from deadweight, thermal expansion, and SSE loads.
In the load combination, the deadweight, thermal expansion, pressure, and SSE stresses were added absolutely for the critical flaw size determination; white the deadweight, thermal expansion, and pressure stresses were summed absolutely for the leakage flaw size determination.
The stresses from these load combinations are given in Table 1 (Attachment 4) for the RHR hot leg and cold leg piping.
For the purposes of LBB analyses, the critical flaw size can be defined as the longest preexisting through-wall flaw which could exist without growing unstably to double-ended pipe rupture under NOP plus SSE stresses.
The analysis performed by SIA was based on the J-integral/Tearing. Modulus (elastic-plastic fracture mechanics) approach to flaw stability which is applicable for the materials of most interest in this analysis.
Formally, piping failure is predicted when the applied J exceeds J,c (the material property value at which crack growth initiates) and the rate of increase of the applied J with crack extension (dJ/da) exceeds the rate of increase of the piping material's J-R curve with crack extension (d(J-R)/da).
The analysis in'SIR-97-077, Rev. 0 calculated the critical flaw size by using SIA's pc-CRACK' code. To do this, SIA first assumed that the stresses applied at an analyzed location were ail tensile stresses and determined a critical flaw size (a,) under these conditions. Then it was assumed that the stresses applied were all bending stresses and determined a critical flaw size (a,) under such conditions. A linear interpolation was then performed between the two results as:
a, = a,*(v,/(o,+ o,))+ a,*(cr,/(o, + a,))
where v, and a, were the bending and tensile components, respectively, of the overall stress and a, was the combined critical flaw size. The critical flaw size based on the consideration of NOP + SSE loadings for each RHR piping location was then given as in Column 2 of Table 2 (Attachment 5). This critical flaw size would therefore be two times the acceptable leakage flaw size at the given location based on the safety factor of two required to exist between the leakage and critical flaws using this definition. The acceptable leakage flaw size based on this criteria is given in Column 3 of Table 2.
NUREG-1061, Vol. 3 however, also requires that the acceptable leakage flaw size be demonstrated to be stable under loads which are equivalent to K2 times the NOP plus SSE loads.
In response to the NRC staff's RAI and using the methodology outlined above, RG&E/SIA determined what size flawwould be stable under such conditions.
RG8 E/SIA provided the information which is shown as Column 4 of Table 2.
In addressing this criteria, because of the margin of v'2 on the loads, no additional margin need be applied on the critical-to-leakage flaw size ratio and the prescribed critical and leakage flaw sizes under this criteria are considered to be equivalent.
Therefore, the controlling leakage flaw size is the minimum value when Column 3 and Column 4 are compared for each location. When performing the leakage analysis, the location at which the minimum leakage was obtained for the controlling leakage flaw size under NOP loads would then be the bounding location.
Having established the acceptable leakage flaw size from applying the appropriate factor of safety to the critical flaw size, the RG&E/SIA analysis then determined the leakage behavior of the'postulated leakage flaw. The leakage analysis performed by SIA was based on the use of the pc-LEAKcomputer code developed by SIA for calculating single or two-phase flowthrough cracks in light-v/ater reactor piping. By inputting the piping cross-section description, material property characteristics, and normal operating loads (or stresses) for each node into the program, the RG&E/SIA analysis determined that the acceptable 5.48 inch leakage flaw at hot feg node 680 provided the minimum amount of leakage, 4.7 gallons per minute (gpm), while the acceptable 5.75 inch leakage flaw at cold leg node 920 provided that leg's minimum leakage, 13.5 gpm.
The RG&E/SIA analysis concluded that these leakage rates were detectable since the iristailed Ginna leakage detection system was capable of detecting 1 gpm of leakage (consistent with NRC Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," guidance).
Therefore, RG&E concluded that the LBB behavior of this line had been demonstrated.
1 4.0 STAFF EVALUATION Based on the information provided by the licensee regarding the materials comprising the Ginna RHR system piping and the loads under NOP and SSE conditions, the staff independently assessed the compliance of these systems with the LBB criteria established in NUREG-1061, Vol. 3. The staff has concluded that the analysis submitted by the licensee, including the additional information supplied in response to the staff's RAI and information supporting a Ginna leakage detection system capability of less than 1 gpm, was sufficient to demonstrate that LBB behavior would be expected from the subject piping. The following sections willfocus on the differences between the details of the staff's analysis, conducted per NUREG-1061, Vol. 3, and the licensees.
4.1 Identification of Anal zed Pi in and Pi in Material Pro erties The staff examined the list of materials identified for RHR system piping and concluded that the materials of primary interest for the LBB analysis would be the SS welds because of their susceptibility to thermal aging.
However, in evaluating the fracture behavior of the SS welds, the stress-strain properties of the surrounding wrought SS piping would also be used, as addressed below. NUREG-1061, Vol. 3 specifies particular aspects which should be considered when developing materials property data for LBB analyses.
First, data from the testing of the plant-specific piping materials is preferred.
However, in the absence of such data, more generic data from the testing of samples having the same material specification may be used.
More specifically, it was noted in Appendix A of NUREG-1061 that "Material resistance to ductile crack extension should be based on a reasonable lower-bound estimate of the material's J-resistance curve," while section 5.2 of NUREG-1061 stated that the materials, data should include, "appropriate toughness and tensile data, long-term effects such as thermal aging and other limitations."
Given the above, the staff did not concur with the RG8 E/SIA methodology for establishing the J-R curve properties of the SS weld materials.
SIA justified the use of the information from EPRI Report NP-6301-D (Reference 6) by noting that it had been used as the basis for the flaw evaluation criteria in ASME Section XI though the information did not account for the effects of thermal aging.
It is the staffs position that an LBB analysis is significantly different from a flaw evaluation and that the thermal aging of SS weld materials must be explicitly addressed.
An additional study from Argonne National Laboratory (NUREG/CR-6248, ANL-95/7)(Reference 9) was the staffs reference for this information and the staffs ch'aracterization of the J-R curve is given in Table 3.
The mean minus one standard deviation lower bound J-R curve used by the staff was actually developed by Wilkowski and Ghadiali at Battelle Columbus Laboratory as a fitto unaged SS weld data, but the conclusions of NUREG/CR-6248, ANS-95/47 (Reference 9) noted that there was little observed change in the fracture toughness behavior with thermal aging for those welds that began with inferior fracture toughness properties.
The J-R curve used by the staff was more conservative than that used by the licensee.
The stress-strain properties of aged SS weld material for this evaluation are also given in Table 3 (Attachment 6). The stress-strain properties provided by the licensee in their RAI response for the aged SS weld material was consistent the properties assumed by the staff.
4.2 General As ects of the Staff's LBB Anal sis The staffs analysis was performed in accordance with the guidance provided in NUREG-1061, Vol. 3. Based on the information submitted by the licensee, the staff determined the critical flaw size at the bounding location for each piping system using the codes compiled in the NRC's Pipe Fracture Encyclopedia (Reference 10). For the purposes of the staff's evaluation, the critical location was defined by those locations at which materials with low postulated fracture toughness existed in combination with high ratios of SSE-to-NOP stresses.
This was because high SSE stresses tend to reduce the allowable critical flaw size while low NOP stresses increase the size of the leakage flaw. When evaluating pipe welds, the staff used the LBB.ENG3 code developed by Battelle (NUREG/CR-6235, BMI-2179)(Reference 11) for that express purpose.
The LBB.ENG3 methodology is significantly different from the other codes in the Pipe Fracture Encyclopedia (Reference 10) and from the licensee's analysis in that LBB.ENG3 explicitly accounts for the differences in the stress-strain properties of the weld and an adjoining base material when determining the effective energy release from the structure with crack extension.
The same criteria as discussed in Sections 3.3 and 3.4 with regard to the applied J exceeding the material J,c and the applied dJ/da exceeding the material's d(J-R)/da were used to identify the critical crack size.
The staff then compared the critical flaw at the bounding location to the leakage flaw which provided 10 gpm of leakage under NOP conditions to determine whether the margin of 2 defined in NUREG-1061, Vol. 3 was achieved.
The leakage flaw size calculation was carried out using the PICEP (Pipe Crack Evaluation Program, Revision 1 analytic code) (Reference 12). The 10 gprn value was defined by noting that the compliance of the Ginna containment leakage detection system with the position in Regulatory Guide 1.45 indicates that this system would be able to detect a 1 gpm leak in the course of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and a factor of 10 is applied to
'his 1 gpm detection capability to account for thermohydraulic uncertainties in calculating the leakage through small cracks.
The stability of the leakage flaw under foadings a factor of K2
II greater than the combination of SSE+NOP loads was subsequently evaluated to check the final acceptance criteria of NUREG-1061, Vol. 3.
4.3 Evaluation of the Ginna Residual Heat Removal S stem Pi in Based on the licensee's results and the loadings supplied by the licensee, the staff concluded that the locations which would be expected to be limitingfor the RHR piping evaluation would be node 680 (for the RHR hot leg) and node 910 (for the RHR cold leg). Of these two locations, the analysis for the SS weld at node 680 bounded all of the RHR piping locations.
Since the weld at node 680 existed between two sections of wrought 316 SS piping the LBB.ENG3 code was used to evaluate the impact of the base material stress-strain properties on each side of the weld. Four different wrought 316 SS stress-strain property representations were used: two of which assumed "typical" yield strength (YS) and ultimate tensile strength (UTS) properties for the material (YS = 25 ksi, UTS = 75 ksi) and differed only in their Ramberg-Osgood parameters (a = 6.9, n = 4.8 or a = 5.8, n=3.6), one submitted by the licensee in their August 6, 1998, RAI response (YS = 29.6 ksi, UTS = 86.6 ksi [corrected], a =
12.0, n = 4.8) and one which used ASME Code minimum strength values at the system's operating temperature and the Ramberg-Osgood parameters provided in the licensee's original submittal (YS = 18.8 ksi, UTS = 71.8 ksi, a = 0.776, n = 3.81). These values resulted in critical fiaw sizes of 9.2, 9.2, 9.5, and 11.0 inches, respectively.
Based on these results, the staff concluded that the appropriate critical flaw size determined by assuming 316 SS properties would be an'average of the three non-Code minimum calculations, 9.3 inches.
The staff then used the PICEP code to evaluate two interpretations of the leakage flaw size for node 680.
Using the surface roughness value that the staff has used in previous LBB evaluations of c = 0.003 inch, the staff determined that 10 gpm of leakage would be expected from a 6.8 inch'through-walt flaw. Therefore, the factor of safety between. the length of critical and leakage size flaws using this approach would be 1.36. Alternatively, ifthe margin on the critical-to-leakage flaw size was fixed at 2, the amount of leakage from the flaw which was one-half the size of the critical flaw size (i.e. 4.65 inches) was determined to be 2.75 gpm.
If, as stated in the original RG&E submittal, the sensitivity of the Ginna leakage detection system was taken to be 1 gpm, then the margin on the leakage would be only 2.75 instead of the factor of 10 required in the guidance of NUREG-1061, Vol. 3.
Although, in previous LBB evaluations, the staff has concluded that margins of-slightly less than 2 on the critical-to-leakage flaw size are acceptable provided that a full margin of 10 is maintained on the leakage uncertainty, deviations of this magnitude are not acceptable.
The NUREG-1061, Vol. 3 guidance does provide, however, for licensees demonstrating leakage detection capabilities of less than 1 gpm.
RG8 E subsequently provided information to the staff to support the conclusion that less than 1 gpm of leakage could be detected by the Ginna containment monitoring system and this information has been evaluated by the staff below.
4.4 Evaluation of Ginna Containment Leaka e Detection S stem Radioactivity detection systems at the R. E. Ginna Nuclear Power Plant are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to reactor coolant system (RCS) leakage.
Section 5.2.5 of the Ginna Final Safety Analysis Report states that the containment air particulate monitor, R-11, is the most sensitive
instrument available for detection of RCS leakage in containment.
Assuming a complete dispersion of leaking radioactive solids consistent with very little or no,fuel cladding leakage, R-11, is capable of detecting leaks as small as approximately 0.013 gpm within 20 minutes.
Even ifonly 10% of the particulate activity is actually dispersed, a leakage rate on the order of 0.13 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is well within the detectable range of R-11, which is much lower than the minimum leakage detection requirement of 0.471 gpm at Ginna. The containment gaseous monitor, R-12, is much less sensitive, but can detect a leak of 2.0 to 10.0 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and is considered to be a backup to the particulate monitor. The containment sump level can measure approximately a 2.0 gpm leak within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Operability of these monitors is addressed in Technical Specification (TS) 3.4.15. Alternative means also exist to monitor RCS leakage inside containment, which include humidity detectors, air temperature and pressure monitoring, and condensate flowrate from the air coolers.
The staff requested that RG&E provide data based on past experience to demonstrate the capability of the containment air particulate monitor to detect such small leakages.
In a letter dated September 16, 1998, the licensee discussed a leakage detection event that occurred on August 19, 1998, which involved R-11. The event began when operations personnel noted an upward trend on their R-11 response.
Background readings for R-11 had varied from 90 cpm to 120 cpm, but within a few hours, the count rate increased and stabilized at 150 cpm to 200 cpm. The licensee's analysis of a containment air sample showed radioisotopes of the types Na-24, Mo-99, and l-133, with a total concentration of 2.0E-11 uCI/cc. The operators later pinpointed the leak at a vent connection on the letdown piping with a 2" nominal diameter.
The leak was estimated to be between 0.05 gpm and 0.10 gpm. The licensee stated that there have been other instances in the past where the R-11 detector has indicated higher count rates and has triggered a search using walkdowns or noting liquid inventory changes of the RCS make-up system.
The capability of these systems to detect RCS leakage is influenced by several factors including the containment free volume and detector location. RG&E stated that the capability to detect a low leakage of 0.013 gpm for the R-11 detector is attributed to Ginna's relatively small containment volume of approximately 970,000 cubic feet, effective recirculation of air inside the containment, and use of a second generation R-11 detector that was installed in 1986.
The staff also questioned RG&E about the availability of the R-11 monitor based on past experience.
As documented in a letter dated December 7, 1998, RG&E reviewed the inoperable equipment control records for R-11 and found the percentage of time that R-11 was operable for a given year, as follows: 1995 - 97%, 1996 - 98%, 1997 - 92%, and 1998 (to December 7, 1998) - 97%. Ginna's current TSs require an RCS water inventory balance every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> under normal conditions or every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> under certain conditions when containment leakage detection system components are inoperable.
Based on the importance of early RCS leakage detection to the leak-before-break methodology, Ginna has committed to increase the frequency of the RCS water inventory balance when R-11 in unavailable.
The licensee plans to add a requirement to the Ginna Technical Requirements Manual (TRM) to require either an RCS water inventory balance or analyses of containment atmosphere grab samples once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter when R-11 is unavailable while in Modes 1, 2, 3, and 4. This requirement was scheduled to be in place by December 31, 1998.
Also, the licensee committed to submit a license amendment request (LAR) to revise Ginna's improved TSs to require either an RCS water inventory balance or analyses of containment atmosphere grab samples once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter when monitor
R-11 is unavailable while in Modes 1, 2, 3, and 4. This LAR is scheduled for submittal prior to May 1, 1999.
Based on the data from the leakage detection event that occurred on August 19, 1998, the relatively small containment free volume, and the effective recirculation of air inside the containment, the staff concludes that the containment air particulate monitor at Ginna is capable of detecting fess than 1 gpm within an hour as recommended in Regulatory Guide 1.45 and for the purposes of the leak-before-break (LBB) evaluation, the licensee has demonstrated their ability to detect less than 0.25 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The staff accepts that the acceptable leakage flaw size may therefore be defined (to maintain the margin of 10 on leakage) as the flawwhich provides 2.5 gpm of leakage under normal operating conditions.
This results in the leakage size flaw for hot leg node 680 being slightly less than 4.65 inches and slightly less than one-half of the size of the critical flaw. The 4.65 inches flaw was shown to be stable under loadings a factor of K2 greater than the combination of SSE+NOP loads.
Therefore, both LBB criteria were demonstrated for the bounding location.
5.0 CONCLUSION
Based on the information and analysis supplied by the licensee, the staff was able to independently assess the LBB status of the analyzed portions of the Ginna RHR system piping. The staff has concluded that, because the appropriate margins on leakage and crack size have been met given the Ginna leakage detection system capability of 0.25 gpm, it has been demonstrated that these sections of piping willexhibit LBB behavior.
Furthermore, the licensee should be permitted to credit this conclusion for eliminating the dynamic effects associated with the postulated rupture of these sections of piping from the Ginna facility licensing basis, consistent with the provisions of 10 CFR Part 50, Appendix A, General Design Criteria 4.
Attachments:
- 1. RHR Hot and Cold Leg Stresses
- 2. RHR Hot and cold Leg Critical Flow Sizes
M. Mitchell V. Ordaz Date:
February 25, 1999
6.0 REFERENCES
1.
NUREG-1061, Volume 3, "Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks," November 1984.
2.
Richards, S.A. (USNRC) to Mecredy, R.C. (RG8E), "R.E. Ginna Nuclear Power Plant Design Inspection (NRC inspection Report No. 50.244/97-201)," September 14, 1997.
3.
Mecredy, R.C. (RG8 E) to USNRC Document Control Desk, "Rochester Gas 8 Electric Corporation's, 50-244/97-201-03 Inspection Report, R.E. Ginna Nuclear Power Plant, Docket No. 50-244," November 11, 1997.
4.
Mecredy, R.C. (RG&E) to USNRC Document Control Desk, "Response to Request for Additional Information (RAI) Related to Leak-Before-Break (TAC No. MA0389), R.E.
Ginna Nuclear Power Plant, Docket No. 50-244," August 6, 1998.
5.
EPRI Report NP-5531, "Evaluation of High-Energy Pipe Rupture Experiments," January 1988.
6.
EPRI Report No. NP-6301-D, "Ductile Fracture Handbook," June 1989.
7.
Westinghouse Stress Report, SDTAR-80-05-05, Rev. 1, RHR 2500, dated 3/4/81
~
8.
Westinghouse Stress Report SDTAR-80-05-26, Sl-200 dated 3/20/81.
9.
- Gavenda, D.J., et al., "Effects of Thermal Aging on Fracture Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds," NUREG/CR-6428, ANL-95/47.
10.
Pipe Fracture Encyclopedia, produced on CD-ROM by Battelle-Columbus Laboratory for the U.S. Nuclear Regulatory Commission, 1997.
11.
Brust, F.W., et. al., "Assessment of Short Through-Wail Circumferential Cracks in Pipes,"
NUREG/CR-6235, BMI-2179.
12.
ERPI Report NP-3596-SR, Revision 1, "PICEP: Pipe Crack Evaluation Program (Revision 1)," December 1987.
0
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Table 1: Ginna RHR Hot and Cold Leg Stresses Hot Leg Node 680 50 60 70 Cold Leg Node 8400 910 920 930 950 960 Normal Operation (NOP) Stresses:
Deadweight+ Thermal + Pressure 7.415 ksi 7.661 ksi 7.393 ksi 6.422 ksi 13.368 ksi 16.851 ksi 12.995 ksi 11.675 ksi 10.336 ksi 9.239 ksi NOP + Safe Shutdown Earthquake Stresses 17.811 ksi 16.628 ksi 16.774 ksi 14.420 ksi 15.419 ksi 19.153 ksi 16.629 ksi 14.655 ksi 12.659 ksi 11.188 ksi Attachment 1
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Table 2: Ginna RHR Hot and Cold Leg Critical Flaw Sizes Node 680 50 60 70 8400 910 920 930 950 960 Critical Flaw Size Based on NOP + SSE Stresses 10.967 inches 11.499 inches 11.432 inches 12.552 inches 12.065 inches 10.390 inches 11.498 inches 12.436 inches 13 456 inches 14.358 inches Acceptable Leakage Flaw Size based on*
NOP + SSE Stresses 5.483 inches 5.750 inches 5.716 inches 6.276 inches 6.032 inches 5.195 inches 5.749 inches 6.218 inches 6.728 inches 7.179 inches Acceptable Leakage Flaw Size Based on 1.414'NOP+SSE) stresses 7.950 inches 8.524 inches 8.453 inches 9.683 inches 9.156 inches 7.322 inches 8 524 inches'.555 inches 10.692 inches 11.601 inches
Table 3: Parameters used in Staff Evaluation of Ginna Aged SS Pipe Welds Parameter Young's Modulus Yield Strength Ultimate Tensile Strength Sigma-zero Epsilon-zero Ramberg-Osgood Alpha Ramberg-Osgood n
4c Value 25000 ksi 49.4 ksi 61.4 ksi 49A ksi 0.00197 9.0 9.8 73.4 KJ / m~
83.5 KJ / m'mm 0.643 Note: J = J~+ C(ha)" and a point-by-point representation was converted to English System units after the calculation was completed in metric units.
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