ML20085J042

Draft Technical Assessment of Hold-Up and Retention of MSIV Leakag

ML20085J042
Person / Time
Issue date:	10/20/2021
From:	Shilp Vasavada NRC/NRR/DRA/APLC
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TECHNICAL ASSESSMENT OF HOLD-UP AND RETENTION OF MAIN STEAM ISOLATION VALVE LEAKAGE WITHIN THE MAIN STEAM LINES AND MAIN CONDENSER Executive Summary This assessment provides the technical basis for the low risk of gross failure of the so-called alternate pathway, shown in Figure 1, to the condenser at seismic accelerations at or below a plants design basis safe shutdown earthquake (SSE). The assessment supports a streamlined approach to demonstrating the seismic capacity of SSCs in the alternate pathway compared to the approach in Regulatory Guide (RG) 1.183, Revision 0.

The deterministic dose calculation of the Main Steam Isolation Valve leakage pathway typically credits only safety-related or seismic Category I structures, systems, and components (SSCs) to mitigate the radiological consequences and to estimate conservative doses. The main steam line piping and the main condenser provide large holdup volumes that reduce the release of fission products that may leak by the MSIVs. Power conversion systems, including the main steam piping downstream of the outboard MSIV and the main condenser, typically are not safety-related or seismic Category I, although the steam piping has been designed to appropriate standards and is subject to some quality assurance measures. As such, these deterministic analyses assume those SSCs are unavailable and all or most of the MSIV leakage travels directly to the atmosphere beyond the outboard MSIV. The NRC staff has previously approved alternative methods for showing compliance with the regulation for the MSIV leakage pathway. In 1999, the NRC staff approved credit for the so-called alternate pathway through the main steam drain lines and the condenser using the approach discussed in the NEDC-31858P, Revision 2, BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, dated September 1993, subject to the limitations in its safety evaluation (SE) dated March 3, 1999 (ADAMS Accession Number ML010640286). The credit for the alternate pathway in the dose calculations is provided through a decontamination factor that changes the results of the calculations. The alternate pathway, especially the condenser, provides a large hold-up volume for fission products. This results in a delay of fission product release, an increased potential for fission product deposition, and, therefore, a reduction in the calculated dose. Note that formal credit for the hold-up in the condenser in the deterministic dose calculations consistent with accepted regulatory positions assumes that the condenser is open (i.e., a fixed amount of leakage, specified in RG 1.183, Revision 0, leaves the condenser).

Over the last two decades, the staff has obtained knowledge and operating experience on power conversion systems, as well as the seismic capacity and risk at nuclear power plants.

Using that knowledge, along with experience gained from regulatory reviews, the staff completed this assessment to evaluate the seismic capacity of the SSCs in the alternate pathway during seismic events.

This assessment includes consideration of the likelihood and consequence, in the form of an undesirable outcome, of fission product transport through the alternate pathway rather than a direct release to the environment as is usually postulated in the deterministic dose calculations.

In other words, the assessment addresses the risk of fission products not transporting through the alternate pathway. This assessment uses engineering information, such as operations and design knowledge, as well as probabilistic and risk information on the seismic capacity (i.e., the ability of an SSC to withstand acceleration induced by a seismic event) of the SSCs in the realistic transport pathway to determine the risk of the unavailability of the SSCs in the power conversion system pathway for fission product hold-up and retention. The assessment approach is depicted in Figure 2 and discussed further in the Assessment section below.

Based on the assessment summarized in this document, the staff concludes that the risk of the unavailability of SSCs in the alternate pathway for fission product hold-up and retention is low, including at seismic accelerations corresponding to a plants SSE. In addition, conservatisms in this assessment maintain safety margins. Therefore, the staff finds that the main steam lines and condenser provide reliable fission product hold-up volumes with the following considerations satisfied:

Design and fabrication of the steam system to appropriate standards is consistent with assumptions (i.e., no outliers)

The site specific seismic ground motion response spectrum is considered The as-built steam system is verified to match design to an extent commensurate with importance to safety of the steam system Figure 1. Schematic illustration of the alternate pathway (reproduced from Assessment of BWR Main Steam Line Release Consequences, ADAMS Accession No. ML062920249)

Figure 2. Assessment Approach Assessment The staff evaluated the likelihood that the main steam and power conversion systems would serve to effectively mitigate the dose consequences of MSIV leakage. The staff used engineering insights as well as probabilistic and risk information related to seismic events as part of its evaluation.

Engineering Insights The high-pressure and high-temperature design of the SSCs in the alternate pathway assure margin in material strength to accommodate seismic loads under the low pressure and temperature conditions that would exist based on the postulated post-accident conditions used for MSIV leakage dose calculations. Post-accident conditions would also support condensation of water vapor in the gases leaking by the MSIVs, which would enhance the ability of the main steam system to retain the fission products. The NRC staff approved credit for the so-called alternate pathway through the main steam drain lines to the main condenser using the approach discussed in the NEDC-31858P, Revision 2, BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, September 1993 subject to the limitations in its safety evaluation (SE) dated March 3, 1999 (ADAMS Accession Number ML010640286). The limitations imposed by this safety evaluation primarily relate to measures that ensured the seismic robustness of the piping for this function.

In this assessment, the staff evaluated the design of the main steam piping downstream of the second MSIV by surveying BWR plants to identify design standards and quality classification applicable to that piping In the plants with BWR 3 and BWR 4 designs, this piping is typically designed to the American Society of Mechanical Engineers (ASME) Standard B31.1.0, Power Piping, or equivalent, and constructed to augmented quality standards in the areas of material certification, testing, and non-destructive examination. In plants with BWR 5 and BWR 6 design, this piping is typically seismically qualified, designed to ASME Boiler and Pressure Vessel Code,Section III, Class 2 standards for nuclear piping, and treated as safety-related.

Therefore, the design standards provide confidence regarding the robustness of the main steam piping in the power conversion system.

Reliability of SSCs in the Alternate Transport Pathway Under Seismic Events The probability of failure (and consequently, reliability) of an SSC under seismic demand is represented by the fragility of the SSC. Seismic fragility is expressed in terms of multiples of gravitational acceleration (e.g., 0.5g) and, unless otherwise noted, expressed in relation to (or anchored to) the peak ground acceleration (PGA), which corresponds to a frequency of 100 Hertz (Hz). A commonly used measure of the seismic fragility of an SSC is the median fragility value. The higher the median seismic fragility of an SSC, the lower the failure probability (or higher the reliability) of that SSC under seismic demand.

The SSCs in the alternate pathway include the MSL piping, the bypass and drain piping, and the main condenser. Several of these SSCs are non-safety related. As noted in the 1999 SE, requiring the non-seismically analyzed portions of the main steam system piping and components to meet seismic Category I requirements would be impractical because the modifications required to upgrade the system to those requirements would be very costly.

Multiple and diverse sources, including recently developed seismic probabilistic risk assessments (SPRAs) have demonstrated that welded and bolted piping, such as MSLs, bypass and drain piping, have high median fragility1. The sources either use or compile the results of analytical methods (e.g., conservative deterministic failure margin and separation of variables) as well as earthquake experience for the fragility determination of various SSCs.

Consideration of failure modes is inherent in the fragility determination process because the fragility of an SSC is dependent on the failure modes that a fragility analyst and plant systems analyst, in conjunction, consider to be limiting to the functionality of the SSC.

The above-mentioned sources document the high median seismic fragility of welded and bolted piping ranging from 1g to greater than 5g (anchored to PGA), with most of the data clustering around 2g. As examples, NUREG/CR-4334 provides median seismic fragilities of 2.5g for main steam piping, 2.2g for balance-of-plant piping, and 1.6g for reactor coolant system (RCS) piping.

The median fragility of motor operated valves, considering various failure modes including failure of the yolk, is also documented to be high with most of the data clustering around 2.5g.

The median fragility for pipe hangers is reported as 1.46g in NUREG/CR-4550.

Due to the high probability of occurrence of loss-of-offsite power (LOOP) during seismic events, the main condenser is not modeled in SPRAs. Therefore, documentation of the fragility of the main condenser is uncommon. However, the main condenser is a large box which, based on 1 These sources include NUREG/CR-4334, An Approach to the Quantification of Seismic Margins in Nuclear Power Plants, August 1985 (ADAMS Accession No. ML090500182); Electric Power Research Institute (EPRI) Report 30020000709, Seismic Probabilistic Risk Assessment Implementation Guide, December 2013 (publicly available from EPRIs website); Park, Y.J., Hofmayer, C. H., Chokshi N.C.,

Survey of seismic fragilities used in PRA studies of nuclear power plants, Reliability Engineering and System Safety, Vol. 62, pp. 185-195, 1998; and seismic PRAs submitted in response to the 10 CFR 50.54(f) letter based on post-Fukushima Near-Term Task Force (NTTF) recommendations (e.g., ADAMS Accession Nos. ML18240A065, ML18093A445, ML18271A109, ML19009A124). Note that EPRI Report 30020000709 has not been endorsed by the NRC. Citing it as a source of information for fragility data does not constitute an endorsement of the report.

earthquake experience, is expected to have high seismic capacity. The main condenser is usually a seismic Category II structure which would necessitate its anchorage being designed to avoid failure at the plants design basis seismic loads. In addition, the very large and heavy main condenser is anchored directly to the floor of the turbine building. The location, size, and weight of the main condenser adds to its capacity to withstand the seismic acceleration, especially at a plants SSE. The readily available information about seismic fragility relevant to main condenser is for the expansion joint for the circulating water piping connection to the condenser from EPRI report 30020000709 with a median seismic fragility of 0.4g (with randomness variability [r] of 0.22 and epistemic uncertainty [u] of 0.22).

For the purposes of this assessment, the 0.4g median seismic fragility, with the r and u of 0.22 are used to determine failure probability at a plants SSE. The intent of using these values is to use median fragility parameters that include the weakest link in the realistic pathway. The selected fragility parameters encompass various SSCs in the realistic pathway as well as their respective failure modes. The deterministic dose calculations assume a prescribed release amount of fission products from the condenser (i.e., the condenser is assumed to be open) and therefore, the use of the fragility parameters for the expansion joint represents a conservatism as compared to the seismic capacity of the remaining SSCs (such as piping and valves) in the realistic transport pathway.

The selected median fragility values would also address failure modes resulting from the collapse of the turbine building because the median fragility for turbine buildings (assuming non-safety related building construction) in the available information has lower bound of 0.5g. The selected median fragility values for this assessment result in a high confidence of low probability of failure (HCLPF; 95% confidence that failure probability is 5% or less) of approximately 0.2g.

For context, the review level earthquake for every nuclear power plant in the US was at least 0.3g during the Individual Plant Examination of External Events (IPEEE) effort. Further, the lowest median fragility that is repeatedly documented (in the cited source documents as well as recent SPRAs) is 0.3g (for ceramic insulators on offsite power lines). It is also worth noting that inclusion of the failure of the expansion joints represents a broader range of failure modes than previously considered for the realistic pathway.

SSEs for majority of plants, especially BWRs, fall within 0.12g and 0.25g PGA. Using the selected median fragility results in a failure probability ranging from of 0.08% to 5% at and below the range of SSEs2. Therefore, even under the selected fragility parameters, the failure probability of SSCs in the realistic pathway at a plants SSE would be low. The median fragility provides a conditional probability of failure given a seismic event and the consideration of the occurrence of a seismic event is discussed in Section 1.3 of this assessment.

Operating Experience - Walkdown Results Post-earthquake walkdowns of nuclear power plants have also demonstrated high seismic capacity of balance-of-plant components. Examples include the walkdowns performed for the nuclear power plants at Kashiwazaki-Kariwa in Japan and North Anna Power Station in the US.

Both the plants experienced beyond design-basis earthquakes. The independent walkdown of Kashiwazaki-Kariwa performed by EPRI is documented in EPRI report 1016317, EPRI Independent Peer Review of the TEPCO Seismic Walkdown and Evaluation of the Kashiwazaki-2 The outcome at the fundamental frequency of various SSCs in the alternate pathway would be similar due to the use of the spectral ratios to scale the fragility from PGA to the frequency of interest.

Kariwa Nuclear Power Plants, January 20083. The results from the independent walkdown do not identify damage in the turbine building or piping connected to reinforced concrete including snubbers and pipe hangers.

Shortly following the 2011 Mineral, Virginia, earthquake, both the Unit 1 and Unit 2 reactors at North Anna tripped, and there was a loss of offsite power to the station. Subsequent analysis indicated that the spectral and peak ground accelerations for the operating basis and design basis earthquakes (OBE and DBE, respectively) were exceeded at certain frequencies for a short period of time (3 seconds). The technical evaluation by the Office of Nuclear Reactor Regulation related to the restart of North Anna after the occurrence the earthquake (ADAMS Accession No. ML11308B406) documents the licensees walkdowns and the NRC staffs review of SSCs to determine damage and loss of functionality.

The evaluation states that the licensee performed inspections of piping and pipe supports, including checking for snubber damage, leakage of hydraulic fluid and bent piston rods, damage at rigid supports to identify deformation of support structure, deformation of pipe due to impact to support structure, damage of expansion joints, damage or leakage of piping and branch lines and for damage to pipe at building joints and interfaces between buildings. The licensee visually inspected welds, flanges, attachment lugs, and couplings. The NRC staffs review agreed with the licensees basis for concluding that piping and pipe supports have not been damaged. The licensee also walkdown and inspected safety related balance-of-plant SSCs and did not find any loss of functionality; a conclusion that the NRC agreed with.

The Great Tohoku Earthquake of 2011 produced the highest recorded ground motions experienced by operating nuclear power reactors. The Onagawa site located to the Northeast of Sendai, Japan, was the site closest to the earthquake epicenter and experienced peak ground accelerations exceeding 0.5g. These accelerations exceeded the facility design basis at certain frequencies. Unit 1, a General Electric BWR 4 design plant constructed by Toshiba, and Unit 3, a General Electric BWR 5 constructed by Toshiba and Hitachi, were operating at full power at the time of the earthquake. As documented in an International Atomic Energy Agency (IAEA) assessment report entitled, IAEA Mission to Onagawa Nuclear Power Station to Examine the Performance of Systems, Structures and Components following The Great East Japanese Earthquake And Tsunami, 2012, the plants safely shutdown without incident following the earthquake. Little damage was noted in the turbine building affecting the power conversion system. The IAEA team identified damage to the main turbine bearing bolts (due to stretching) and to the ends of the low-pressure turbine blades due to wear from relative motion between the rotor and casing. No damage to the steam piping was noted. Section 7.4 of the IAEA report states [t]he systems supporting the balance of plant did not suffer damage including the turbine bypass and turbine stop valves since they operated after the earthquake.

It is recognized that site characteristics, location of SSCs, and operational practices are important factors in the plant response to an earthquake. Therefore, the information from walkdowns of nuclear power plants presented above is used in this assessment to provide insights on the seismic capability of SSCs in the realistic pathway rather than definitive conclusions about potential earthquake impacts. The insights from the walkdowns discussed above reveal the appreciable seismic capacity of SSCs in nuclear power plants, and the ability of both safety and non-safety related SSCs to remain functional during and after an SSE. Every operating nuclear power plant in the US has performed a walkdown focused on identifying 3 EPRI Report 1016317 has not been endorsed by the NRC. Citing it as a source of information for insights from post-earthquake walkdown does not constitute an endorsement of the report.

weaknesses in SSCs when exposed to seismic events (including beyond-design-basis seismic events) and several plants have performed an Expedited Seismic Evaluation Process (ESEP) review as part of post-Fukushima actions resulting from Near-Term Task Force (NTTF) recommendation 2.3. The ESEP reviews were performed to demonstrate seismic margin and expedite plant safety enhancements through evaluations and potential near-term modifications of certain core and containment cooling equipment while more comprehensive plant seismic risk evaluations are being performed.

The staff notes that material degradation due to aging can result in reduction in seismic capacity of SSCs. NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Revision 2 (ADAMS Accession No. ML103490041), provides the NRC staffs generic evaluation of the existing plant programs and documents the technical basis for determining where existing programs are adequate without modification and where existing programs should be augmented for the period of extended operation. The programs, with or without modification, are termed aging management programs (AMPs).Section VIII of NUREG-1801 discusses AMPs for Steam and Power Conversion System including separate discussions for main steam system (BWR),

extraction steam systems, condensate system, external surfaces of components and miscellaneous bolting, and common miscellaneous material/environment combinations.Section III.B2 of NUREG-1801 discusses supports for conduits and non-ASME piping and components including anchorage and supports and corresponding AMPs. Similarly,Section III.B1 discusses AMPs for supports for ASME piping and components. Therefore, material degradation due to aging in SSCs relevant to this assessment are addressed for licensees that currently have extended operating licenses or will apply for such license in the future.

In summary, based on the available information and using the fragility parameters that represent various SSCs in the realistic path and their failure modes, the probability of the unavailability of the realistic pathway at a plants SSE is low.

Occurrence Frequencies of Design Basis Seismic Events The median fragility assessment discussed in the previous section provides information about the failure probability of SSCs in the realistic pathway if an SSE were to occur. Using the plant-specific seismic hazard in conjunction with the median fragility parameters provides an indication of the frequency of occurrence of a radioactive release. Such an occurrence frequency can be determined by convolving the seismic hazard with the selected median fragility parameters. Such an approach assumes that every earthquake, even one at or below a plants SSE, results in core damage.

Every operating nuclear power plant in the US has performed a re-evaluation of the plant-specific seismic hazard using present day information as part of post-Fukushima actions resulting from NTTF recommendations. Therefore, generic or assumed hazard curves are unnecessary. Since the median fragility parameters are anchored to the PGA, the hazard curve of interest would be the mean PGA hazard curve (i.e., the mean hazard curve for 100 Hz frequency).

It is exhaustive and beyond the scope of this assessment to perform the convolution discussed above for every operating BWR (or a subset thereof). For the purposes of this assessment, the convolution was carried out for three BWRs with SSEs corresponding to 0.13g, 0.15g, and 0.24g (PGA). In each case, the convolution of the hazard and the selected median fragility parameters resulted in a cumulative occurrence frequency of failure of the SSCs in the realistic pathway of the order of magnitude of 1x10-6 considering even the entire hazard curve (i.e.,

beyond design basis earthquakes). The contribution from earthquakes at and below the SSE was less than 1x10-6 per year4. Therefore, even under the selected median fragility parameters and assumptions on accident initiation and progression, the risk of unavailability of the realistic pathway at a plants SSE is low. Even under the assumption that failure of the realistic pathway results in the releases going directly to the control room or the environment, the occurrence frequency of radiological releases to the control room and/or the public is low.

Uncertainty Evaluation As demonstrated in the previous sections, the assessment summarized in this document includes several conservatisms such as the use of the selected median fragility and consideration of an SSE in conjunction with the MHA. These conservatisms address uncertainties in the assessment by bounding the seismic failure probabilities of various SSCs and corresponding seismically-induced failure modes. It is worth noting that the calculation of the failure probability using the median fragility parameters includes consideration of uncertainty in that parameter.

In addition, conservatisms exist in the postulated deterministic dose calculation approach. The Statement of Considerations accompanying the publication of 10 CFR 50.67 (64 FR 71990) clarify that the design basis accidents analyzed for dose calculations, are intentionally conservative in order to address uncertainties in accident progression, fission product transport, and atmospheric dispersion.

Recommendations The staffs safety evaluation dated March 3, 1999 (ADAMS Accession No. ML010640286),

which addresses NEDC-31858P, Revision 2, BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, dated September 1993, included the following limitations for use of the methodology:

1.

Individual licensees should provide a detailed description of the ALT drain path and the basis for its functional reliability, commensurate with its intended safety related function. The licensee should also describe their maintenance and testing program for the active components (such as valves) in the ALT path.

2.

Individual licensees should provide plant-specific information for piping design parameters (e.g., uniqueness of piping configurations, pipe span between supports, and diameter-to-thickness ratios for each pipe size), to demonstrate that they are enveloped by those associated with the earthquake experience database.

3.

Individual licensees should demonstrate that the plant condenser design falls within the bounds of design characteristics found in the earthquake experience database. This should include a review of as-built design documents and/or a walkdown to verify that the condenser has adequate anchorage.

4 The results continue to remain valid using the so-called simple average approach from the efforts related to Generic Issue (GI)-199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants. (ADAMS Accession No. ML100270598).

4.

Individual licensees should perform a plant-specific seismic evaluation for representative supports and anchorages associated with affected piping and the condenser.

5.

Individual licensees should confirm that the condenser will not fail due to seismic 11/1 type of interaction (e.g., structural failure of the turbine building and its internals).

6.

Individual licensees of plants whose FSARs or UFSARs reference Appendix A to 10 CFR Part 100 should perform a bounding seismic analysis for the ALT path piping.

Those licensees committed to Part 100 should discuss the basis for selecting a particular portion of the bypass/drain line for the bounding analysis.

7.

The methodology and criteria used for the analytical evaluations should be those which are in compliance with the design basis methodology and criteria, or those which are acceptable to the staff.

8.

The facility ground motion estimates shown in Figures 1 through 13 of this attachment have been reviewed and accepted by the staff for inclusion in BWROG's earthquake experience database. These 13 facility ground motion estimates may be used to verify the seismic adequacy of equipment in the alternative MSIV leakage pathway for plants referencing the BWROG's Topical Report, NEDC-31858P, Revision 2.

9.

At the present time, there is no standard, endorsed by NRC, that provides guidance for determining what constitutes an acceptable number of earthquake recordings and their magnitudes and for determining the required number of piping and equipment items, that should be referenced in the earthquake experience database when utilizing the BWROG methodology. Therefore, individual licensees are responsible for ensuring the sufficiency of the data to be submitted for staff review and determination. When a revision of the QME Standard that incorporates specific criteria for use of experience data in the qualification of mechanical equipment is endorsed by NRC, such criteria should be followed in future applications involving MSIV ALT pathway evaluations.

Licensees choosing to credit the main steam lines and alternative drain pathway and who do not use the guidance in the NRC staff safety evaluation establishing limitations on credit for the so-called alternate pathway through the main steam drain lines and the condenser using the approach discussed in the NEDC-31858P, Revision 2, BWROG Report for Increasing MSIV Leakage Limits and Elimination of Leakage Control Systems, dated September 1993, should provide information to support the staffs reasonable assurance determination that structures, systems, and components (SSCs) in the pathway will not undergo gross seismic failure. The assessment summarized in this document provides the basis for the staff conclusion that the risk of the unavailability of SSCs in the alternate pathway for fission product hold-up and retention is low, including at seismic accelerations corresponding to a plants SSE. In addition, conservatisms in this assessment maintain safety margins with minimal activities to verify the as-built configuration of these SSCs. Therefore, it is recommended that the following information is used to establish reasonable assurance that the main steam lines and condenser provide reliable fission product hold-up volumes and quantitative credit for the hold-up can be used in the dose consequence analysis:

The main steam line and attached piping should not experience a gross pressure boundary failure from seismic loading to provide reasonable assurance the leakage does not bypass the holdup volumes and deposition surfaces. Therefore, all licensees choosing this alternative should describe the code of record used for the main steam lines and the extent of quality assurance measures applied to the design, materials, and fabrication of the steam lines and attached piping. If the main condenser is credited as a holdup volume, the description should also include the alternate pathway identified to the main condenser. In addition, it is recommended that the following tiered information is provided for the site, as applicable:

o If the piping and valves in the alternate pathway have been subjected to dynamic seismic analysis for the as-built configuration to a code of record (e.g., ASME B31.1, Power Piping), and the magnitude of the seismic response spectrum for the analysis equals or exceeds the licensees safe shutdown earthquake (SSE),

a description of the dynamic analysis provides sufficient justification.

o If the SSCs in the alternate pathway have not been subjected to dynamic seismic analysis to a code of record (e.g., ASME B31.1, Power Piping) and if the peak spectral acceleration of the ground motion response spectra (GMRS) based on the licensees most recent site-specific probabilistic seismic hazard is less than 0.4g, the justification should include (1) discussion of seismic capacity and margin present in the relevant SSCs, including the condenser, based on their design code(s) of record, (2) insights from plant-specific seismic assessment performed as part of the Individual Plant Examination for External Events (IPEEE) for relevant SSCs, and (3) a walkdown of a sample of the relevant SSCs, including the condenser, performed by knowledgeable licensee staff to verify that they have been constructed as designed, and (4) confirmatory calculations for a sample of piping supports to verify that they provide acceptable flexibility at terminal ends of piping and major branch connections. The extent of the selected samples should be justified based on the plant-specific seismic hazard and quality assurance applied to design and fabrication. Details of the walkdown, including qualification of licensee staff performing them, should be retained in archival documentation.

o If the SSCs in the alternate pathway have not been subjected to dynamic seismic analysis to a code of record (e.g., ASME B31.1, Power Piping) and if the peak spectral acceleration of the GMRS based on the licensees most recent site-specific probabilistic seismic hazard is greater than 0.4g, the justification should include (1) discussion of seismic capacity and margin present in relevant SSCs, including the condenser, based on their design code(s) of record, (2) IPEEE insights described above plus plant-specific seismic walkdown information for the relevant SSCs, and (3) walkdown(s) of the SSCs in the alternate pathway, including the condenser, performed by knowledgeable licensee staff to ensure that items impacting the seismic capacity of relevant SSCs (e.g., loose or missing anchorage, degraded pipe supports), and (4) confirmatory calculations for a sample of piping supports to verify that they provide acceptable flexibility at terminal ends of piping and major branch connections. The extent of the selected samples should be justified based on the plant-specific seismic hazard and quality assurance applied to design and fabrication. Details of walkdown(s),

including qualification of licensee staff performing them, should be retained in archival documentation.

It is not the intent to have the licensee perform detailed fragility analysis as part of the justification, unless voluntarily performed. If fragility analysis is voluntarily performed, the analysis should use state-of-practice methods such as Conservative Deterministic Failure Margin (CDFM) and Separation-of-Variables (SoV).

Conclusions Based on the available information and assessments using assumptions about the seismic capacity that encompass the SSCs in the realistic pathway, there is high confidence that the main steam lines and SSCs in the alternate pathway will be available for fission product dilution, hold-up, and retention, especially at the seismic accelerations corresponding to a plants design basis SSE. Conservatisms in the assessment result in additional safety margin. It is recommended that tiered information needs are used to establish reasonable assurance that the main steam lines and condenser provide reliable fission product hold-up volumes and quantitative credit for the hold-up can be used in the dose consequence analysis.