ML24073A234
| ML24073A234 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/06/2024 |
| From: | Ferneau K Indiana Michigan Power Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| AEP-NRC-2024-03 | |
| Download: ML24073A234 (1) | |
Text
Attachment 3 of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS
Full Power Internal Events F&O Status SRs F&O Description Disposition The Large Early Release Frequency (LERF) analysis Pressure and temperature induced uses NUREG/CR-6595 type evaluations for some SGTR events are modeled as LE-D5 portions of the LERF evaluation based on conservative progressing directly to LERF. This 2-19 LE-C1-C5 assessments. These portions of the LERF analysis do ensures an over-estimation of the (2015 LE-C9-C13 not meet Capability Category II of the standard. One significance of this assumption.
Full Open LE-E2 example of these conservatisms is assuming Steam Scope) Generator Tube Rupture (SGTR) is a containment MET: CC-I bypass event without considering success of secondary side isolation. However, these conservatisms generally do not impact the ability to perform PSA applications using LERF.
Most of the notebooks indicate that interviews with There is not expected to be a knowledgeable plant personnel were conducted to significant deviation between what is confirm that the systems analysis adequately reflected modeled in the PRA and actual the as-built, as-operated plant and that plant-specific plant condition such that there 4-4 data was appropriately collected where required. would be a substantive impact on (2015 SY-A2: MET However, a record of such interviews was not provided numerical model results. Some Full PR SY-A4: CC-I as part of the notebook documentation. walkdowns and interviews have Scope) SY-C2: MET Plant walkdowns are discussed in a generic walkdown been performed and did not identify document created in June 1991. There is no record of any necessary modeling changes, recent system walkdowns conducted with the same outcome is expected for knowledgeable plant personnel. Even if the system those systems that still need configuration has not changed during that time, there walkdowns and interviews should be a confirmatory walkdown to document that. performed.
Based on a discussion with Cook Nuclear Plant (CNP) Recent outage durations have been Probabilistic Risk Assessment (PRA) Engineer, the long due to work related to CNP model conservatively models the opposite unit's replacement of baffle bolts in the 6-19 outage unavailability by assuming 45 days outages reactor vessel, and therefore a (2015 Open DA-C13 with train unavailability equal to an equivalent portion value informed by recent Operating Full of the outage (e.g., a 2 train system would assume one Experience would result in an Scope) train unavailability is 22.5 days). overestimation of the risk associated with outage windows. The current estimate of 45-day outages bounds recent outage experience and is of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS
F&O Status SRs F&O Description Disposition therefore either not expected to impact model results, or result in an overestimation.
All containment failures caused by hydrogen The current implementation of the combustion are assumed to contribute to LERF in modeling is conservative and will report 01V015-RPT-01. There is no discussion in the result in an overestimation in the report that provides a basis for this approach to risk significance of sequences that scenario assignment based on containment failure could potentially be screened out location considerations. PRA-NB-LER, Revision O also based on containment failure does not relate the assignment of LERF scenarios to location. However, an improvement containment failure location. The latter document of this modeling would not result in a 2-4 identifies the most likely containment failure location significant improvement in the (2017 from the containment capacity report (Stevenson overall realism of the model results.
Hydrogen Open LE-D3: CC-I report) and includes a historical discussion that FSPR) provides an argument against assigning scenarios that involve that failure location to LERF. This indicates that there is some uncertainty about the release size from this most likely failure location which would constitute an effect of failure location on event classification that is not discussed.
However, this SR is considered met at CC I because the failure location was assessed in a conservative manner.
Attachment 3 of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS
Fire PRA F&O Status SRs F&O Description Disposition Observation: Parametric uncertainties of applied hot short The impact of probabilities have not been incorporated into the model. this Finding is limited to small CFA2-Finding & Observation (F&O) Closure Notes: portion of the 01 Partially CF-The CF and UNC notebooks were reviewed and confirmed that numerical parametric uncertainty (2010 Open A2 uncertainties are documented. However, several inconsistencies were identified analysis, and Full Met between the documentation and the values used in the model. Given that the majority thus does not Scope) of CF uncertainties have been correctly applied CF-A2 is now considered Met. impact the overall technical quality of the Fire PRA.
FQ-D1 FQ-Some of the Internal Events LE SRs were classified as CC-I due to conservative See disposition 02 D1 modeling. Therefore, the Fire LERF should also be limited to CC-I as appropriate for for FPIE F&O 2-(2022 Open PRM-applications. Revise the internal events PRA to meet CC-II for relevant SRs and 19.
Focused B2 implement changes to FPRA.
Scope)
Section 4.3 of the PP report says no spatial separation was credited as a partition Review of the element. This statement was the basis for the CC-I assessment in the original peer PRA review. There is a disconnect between Section 4.3 and Section 3.1 and 3.2 that needs implementation to be rectified. of fire modeling concluded that PP-B3-Additionally, Section 3.1 discussed the subdivision of the yard into sub-compartments the issue 01 PP-based on spatial separation, but these sub-compartments do not become separate described is a (2022 Open B3 listed fire zones and are not separated in PRA-NB-FIRE-IGN. No explanation is given documentation Focused IGN-for this in R1900-0041-0001. IF the intent is to separate them for fire modeling only ( as disconnect, and Scope) A7 suggested by Table 6-1 of PRA-NB-FIRE-IGN), this should be stated with the PP therefore its analysis and reiterated in the PAU Table in Attachment 1 with a note for clarity. resolution will Revise language in Section 4.3 regarding the use of spatial separation. Clarify not impact treatment of the yard sub-compartments by adding additional discussion to Section 3.1 numerical model with a clarifying note in Attachment 1. Alternatively, carry the sub-compartments results or risk forward as separate "fire zones" into Attachment 1 and the IGN consistent with the sub-insights.
divisions of the other fire compartments.
Attachment 3 of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS
Seismic PRA F&O Status SRs F&O Description Disposition
- 1) -Only a single method was considered to Seismic PRA (SPRA) results are not evaluate the liquefaction triggering potential, expected to be impacted. l&M (2014) liquefaction susceptibility, and post liquefaction initially performed a liquefaction volumetric strains. However, in F&O 20-7, Item 2, triggering (using Youd et al., 2001) more than one method was requested to conduct and settlement (using Tokimatsu and the liquefaction hazard evaluation as "the choice of Seed, 1987) analyses using the RLE any single method does not address the epistemic and obtained comparable results. l&M uncertainty in the field (which is the underlying considers that Figure 6-7 shows motivation of a recent National Academy study and liquefaction at some boreholes for 1E report)". 6 motions, but shows no lateral continuity of the liquefiable boreholes.
1-1 2)- Lateral spreading hazard at the site does not Based on this information, l&M has (2018 PR SHA-I1: MET address the evaluation of this potential hazard. concluded that the site can be Full SHA-I2: MET Lateral spreading can occur in slope gradients as screened out for site-wide lateral Scope) flat as 0.5 percent(%) (without a free face) spreading.
(See NA report). Additionally, Figures 6-7 and 6-9 shows that there is continuous layer of potentially liquefiable soils (in direction towards the lake) on borings B120, B124, B133, B142, and B141 between elevations of about 560 and 555 ft.
Therefore, the potential of lateral spreading and/or flow slides at the site should be evaluated.
3)- Provide a full reference to all citations included in the report.
1)- Include additional justification on why V/H ratios SPRA results are not expected to be should be used instead of vertical GMPEs in report impacted as this F&O has been 20-3 DC COOK-PR-02, Section 7.1 (e.g., inconsistency technically resolved.
(2018 DO SHA-J2: MET of controlling earthquakes between horizontal and Full vertical spectra if vertical GMPEs were used).
Scope)
- 2) - Perform a thorough editorial review of the reference citations and list of references. of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS
F&O Status SRs F&O Description Disposition
While the cracking assessment for the Containment SPRA results are not expected to be Building (CB) and TB/SH has been resolved the impacted. The studies performed in cracking assessment for AB has not been fully 15C4313-RPT-003 "Summary of resolved. Several changes were made to the AB Building Response Analysis for the structural model in 15C4313-CAL-010, "Response Cook Nuclear Plant (CNP) Unit 1 &
Analysis of Auxiliary Building," Revision 2, in Unit 2 SPRA," Attachment E, show response to other SFR F&Os. The updated AB that while there may be some model was used in the cracking assessment with cracking, it is not widespread at the un-cracked section properties. The SPRA team RLE-level. Additionally, l&M performed cracking assessment at earthquake engineering judgement is that with the levels corresponding to 0.5*RLE and 1.0*RLE. studies performed, cracking in the 2-1 Figures 1 through 8 in Attachment E present the structure will decrease the stiffness (2019 0 SFR-B3: MET shear stress contour plots on isometric views of the and increase the damping. These two FSPR) AB model showing the exterior walls. The stress effects tend to affect the structural contour plots only suggest that the building is overly response in opposite ways. Finally, stressed in certain regions. For a complex structure many significant contributors have low such as the AB, this is not sufficient to conclude fragilities for which consideration of a that cracking will or will not occur in the building cracked model would be non-especially under dynamic loads. The SPRA conservative.
development team has not assessed or documented the cracking assessment for the AB interior walls in a way that resolves the concern identified in the initial F&O issued by the peer review team.
Perform a sensitivity study to address items SPRA results are not expected to be determined to be risk significant based on F-V impacted. l&M position is that 22-2 importance greater-than or equal-to 0.005. additional studies for risk items not (2018 DO SFR-E3: CCII considered by risk significant as Full ( defined in the SPRA quantification Scope) notebook) will not change risk insights. of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS
Perform a sensitivity study to address items SPRA results are not expected to be determined to be risk significant based on F-V impacted. l&M position is that 22-5 importance. sensitivity studies documented in the (2018 SPRA quantification notebook Full DO SFR-E2: CCII envelope any small fragility changes Scope) that may be discovered by the additional sensitivity recommended here and will not change risk insights.
SPRA team has used the ASCE 4-16, SPRA results are not expected to be Section 3.7.2 dynamic coupling criteria for single-impacted. The l&M position is that the point attachment to show that the current CB simplified method used to modeling approach and response are realistic. demonstrate that the CB modelling While the modeling approach use probably does simplifications have no impact on the 28-2 not have an effect on overall response of the response in 15C4313-RPT-003 (2018 PR SFR-B3: MET structure but that conclusion has not been Attachment B is sufficient to address Full demonstrated adequately. the F&O. The close-out team Scope requested more detailed studies be performed to close the F&O, however the team stated that they believe the conclusion will most likely not change as a result. of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS
Appropriate damping was used for cracked and un-SPRA results are not expected to be cracked building sections in the building response impacted. The position of l&M is that sensitivity studies following the current industry and the conclusion provided in 15C4313-standard ASCE 4-16. The sensitivity practice RPT-003, Attachment E, is sufficient studies are documented in Attachments B and F of to justify the use of un-cracked 15C4313-RPT-003, respectively, for Containment damping for the AB model. See F&O 28-4 Building and Turbine Building/Screen House. 2-1 for further information.
(2018 SFR-B3: MET Appropriate damping is also used for AB response Full PR analysis model documented in 15C4313-CAL-010 Scope) Revision 2.
However, the focused scope peer review F&O 2-1 would require to reassess the cracking assessment of AB and appropriate damping should be used if cracking is assessed to be of significance.
The SPRA development team added an argument SPRA results are not expected to be that due to the way that fragilities were developed, impacted. The sensitivity studies including the application of uncertainty with respect performed in 15C4313-RPT-003 to frequency was sufficient to allow no variation in between un-cracked and cracked structural properties. The variation in frequency is properties show that structural 28-11 intended to reflect uncertainty in the value of the variability has a minor impact on (2018 calculated frequency. The variation in structural response compared to the soil Full 0 SFR-B4: MET properties is intended to reflect uncertainty in those property variability. l&M will review the Scope) properties. Both effects must be considered when small number of impacted risk-developing fragilities. significant components on a case-by-case basis, adjusting the FROI by an additional +/- 15% to ensure structural variability is captured in the fragility calculations. of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS
The gap in PSD as described in the F&O should be SPRA results are not expected to be addressed per latest fragility guidance document. If impacted. l&M position is that there it is confirmed that there is a gap in PSD at FRO! of are not significant gaps in energy near structure, then it is recommended to perform a frequencies that are important to risk-sensitivity study to assess the impact of the gap in significant fragilities. The PSDs as energy. The SPRA development team can perform presented were developed using a 28-13 this by comparing the PSD functions of the five-logarithmic frequency interpolation (2018 time histories that were generated by resolution of which tends to emphasize magnitude Full PR SFR-B4: MET F&O 28-09 to the PSD function of the artificial time variation at low frequencies. A review Scope) history, or the development team can integrate the of the non-interpolated PSDs and PSD function to show that a smooth curve is PSDs developed using a linear generated. frequency interpolation supports the determination that the gaps identified in the F&O are not significant.
The documentation needs to be further SPRA results are not expected to be updated to provide a basis for not considering impacted, as this F&O has been SSSI effects. Subsequent to the closure review, technically resolved. Additional additional documentation was added to the quantitative justification added 28-19 calculations. However, the closure review team to Section 4.4 of 15C4313-(2018 does not consider this additional documentation to RPT-003 is adequate in showing that Full SFR-F2: MET be sufficient to address the concern originally SSSI effects do not control over RLE Scope) DO identified. demand for applicable components.
Also note that components associated in this documentation item are not risk significant. of Enclosure 2: DISPOSITION AND RESOLUTION OF OPEN PEER REVIEW FINDINGS AND SELF-ASSESSMENT OPEN ITEMS
Resolved with Open Documentation In the SPRA SPRA results are not expected to be Model Quantification Notebook,Section 8.2.2, impacted as this F&O has been Revision 1, the cutset review included a statement technically resolved.
on non-significant cutsets - samples are covered 25-7 by the examination of G1 and G2 bins. G1 and G2 (2018 bins contain relatively fewer seismic-induced Full DO SPR-E3: CCII failures and the cutsets have features more like the Scope) internal events PRA. A recommendation is made to expand the review to other ground motion bins so that model logic related specifically to the SPRA can be confirmed to be appropriate and as intended.
Some of the supporting internal events LE SRs No impact to SPRA results -
were met at CC-I only; therefore, this SR is met for The LERF modeling is built upon the CC-I only. internal events LERF model and is essentially unchanged.
25-9 The SPRA LERF model includes (2018 seismic-specific aspects such as Full 0 SPR-E6: CCI unique containment failure Scope) probabilities. Whereas there are some internal events LE supporting SRs that meet both CC-I and CC-II, the majority of the SRs are met at CC-I. Therefore, this SR is considered to be met at CC-I only.
Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
A calculation PRA-EXT-HAZ-SCRN was developed for external Hazards screening. See Table 4-4 below.
Table 4-4: Hazard Dispositions 50.69 Hazard Screening Disposition for 10 CFR 50.69 Criterion Aircraft Impacts PS2 Per Updated Final Safety Analysis Report (UFSAR) Section 2.1.4 [6.3], there are two airports within a 15-mile vicinity of the DC Cook Nuclear Plant (CNP): Southwest Michigan Regional Airport located approximately 12 miles North East (NE) of the plant on the NE edge of Benton Harbor and Andrews University Airport located approximately 10 miles East of the plant near Berrien Springs.
For airports beyond this 15-mile radius, the orientation of runways and normal flight patterns are not in the direction of the plant or the normal glidepath heights are not within the plant vicinity so that aircraft utilizing the facilities of these airports would not normally fly over the plant site.
Southwest Michigan Regional Airport, data from 2017 shows approximately 18,350 operations (take-offs or landings) [6.10]. Due to the north-easterly location of the airport and the orientation of the runways, normal glide paths would not approach the vicinity of the plant.
. Andrews University Airport/Airpark has one runway. Due to the easterly location of the airport and the orientation of the runway, normal glide paths would not approach the vicinity of the plant. For 2016, which is the most current data for this airport, there were approximately 7,300 operations
[6.10].
The annual movements are below the critical number at which a probability analysis for aircraft accidents would be required according to Regulatory Guide 1. 70 [6.11]. Therefore, the probability of aircraft crashing into the site is considered to be remote, and airplane crashes need not be considered for design basis events.
Additionally, the Individual Plant Examination of External Events (IPEEE) [6.12] checked against the requirements of the US Nuclear Regulatory Commission (NRC) Standard Review Plan. The first requirement is met and was previously explained. The second requirement is the plant is at least five (5) statute miles from the edge of military training routes, including low level training routes. According to the IPEEE, this requirement is satisfied as the nearest military training route, VR 1640, is approximately 50 miles from the site. The third requirement is the plant is at least two (2) statute miles beyond the nearest edge of a federal airway, holding pattern, or approach pattern.
This requirement is not met as a low altitude flight path V526 is within the two (2)-mile limit and Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
50.69 Hazard Screening Disposition for 1 0 CFR 50.69 Criterion required further analysis. The high-altitude flight path is J584 and is at least five (5) miles from the site at its closest approach and does not require further analysis. The IPEEE calculated the probability per year of an aircraft crashing into the plant for flight pattern V526 for 1990 (with data from 1988) and it was less than the limit of 1E-7 and thus precluded further analysis. The IPEEE concluded the contribution to plant risk is insignificant relative to other initiating events.
Based on this review, the aircraft impact hazard can be considered to be negligible.
Avalanche C3 The location of CNP precludes the possibility of an avalanche.
Based on this review, the avalanche impact hazard can be considered to be negligible.
Biological Events c5 The only biological event that may credibly affect CNP is zebra mussel blockage of circulating water system intakes. According to the IPEEE [612], effects of mussel buildup are continuously monitored, and the plant would have sufficient warning if conditions warranted shutdown.
Additionally, according to UFSAR Section 2.6.4 (6.31. biocides supplemented by mechanical cleaning and design changes including strainers, filters, screens, and chemical delivery systems, work to protect plant systems. A zebra mussel monitoring program utilizing side-stream and artificial substrate monitors, along with diver and heat exchanger inspections, is used to evaluate the effectiveness of chemical and physical control measures.
Based on this review, the Biological Event impact hazard can be considered to be negligible.
Coastal Erosion c5 Per UFSAR Section 2.3.3 [6.3], shoreline erosion is not evident at the site.
The long-time periods required to produce sufficient coastal erosion to endanger the plant would provide sufficient time for plant shutdown, and, therefore, no further analysis was performed.
Based on this review, the Coastal Erosion impact hazard can be considered to be negligible.
Drought c5 Drought is a slowly developing hazard allowing time for orderly plant reductions, including shutdowns. According to the IPEEE [6.12], the depth of the intake cribs at the Cook site (about 10 feet below the record low lake level) precludes further analysis.
Based on this review, the drought impact hazard can be considered to be negligible.
Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
External Flooding 1 The evaluation of the impact of the external flooding hazard at the site was updated as a result of the NRC's post-Fukushima 50.54(f) Request for Information. The station's FHRR was submitted to NRC for review on March 6, 2015 [6.6). The results indicate all flood-causing mechanisms, except LIP, are bounded by the Current Licensing Basis and do not pose a challenge to the plant.
Modifications to prevent critical plant equipment from being adversely impacted by flood water intrusion from a Beyond Design Basis LIP event were performed. Specific details are discussed in Section 4.2.1 of PRA-EXT-HAZ-SCRN.
Consistent with Figure 5-6 in NEI 00-04 [6.9], an evaluation was performed for the screening of the
. external flooding mechanism LIP to determine if there are any components that participate in screened scenarios and whose failure would result in an unscreened scenario. There are several components whose failure to be in their normal position (e.g., doors in the closed position) or function appropriately (e.g., roof scuppers shedding water to not allow water to pool) during a LIP event to limit the ingress of water to the Auxiliary and Turbine Buildings would result in an unscreened scenario. These components should be categorized as HSS in accordance with NRC approved guidance.
With credit taken for these components during a postulated LIP, external flooding mechanisms are screened as not impacting 10 CFR 50.69 categorization.
The components are identified in Table 4-3 of PRA-EXT-HAZ-SCRN.
Based on this analysis of the external flooding hazard for CNP, the hazard has negligible impact.
Extreme winds C1,PS4 Based on information in the UFSAR [6.3], the plant design for wind pressure and the low frequency and Tornadoes of design tornadoes, a demonstrably conservative estimate of Core Damage Frequency (CDF) associated with high wind hazard (other than wind generated missiles) is much less than 1E-6/yr.
Details are provided in Section 4.1 of PRA-EXT-HAZ-SCRN.
Section 1.4.7 of the UFSAR [6.3] documents that a limited number of SSCs located near openings/penetrations in Seismic Category I structures or located outside of such structures have been evaluated and do not require additional physical tornado missile protection features. These Safety System Components have been evaluated with respect to the overall risk resulting from tornado-generated missiles upon potential off-site dose consequences exceeding the guidelines of Regulatory Guide 1.183 and 10 CFR 50.67; the acceptance criterion is being less than 1.0E-06 per reactor-year. TORMIS determines the probability of tornado generated missiles striking targets, Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
which may include, but are not limited to, walls and roofs of buildings, penetrations of Seismic Category I structures, and exposed portions of systems/components. The probability is calculated by simulating many tornado strike events at the site. This results in a calculated probability per unit area of striking any target. Then, the exposed surface area of each component is factored in to determine the probability of striking each item. The TORMIS analysis for CNP is documented in SD-990930-004 Revision 12 [6.4].
The TORMIS analysis for CNP is updated for this assessment with a bounding analysis which incorporates the CNP FPIE PRA MOR [6.2] and updates the target (e.g., component) to basic event mapping.
The results of the bounding analysis, presented in Table 4-2 of PRA-EXT-HAZ-SCRN, indicate this hazard screens because the CDF is below 1.0E-06 per year (PS4). Based on this analysis of the high winds/ tornadoes hazard for CNP, the hazard has negligible impact.
Fog C4 Fog can increase the frequency of occurrence of accidents or events. Fog is implicitly included in data for transportation accidents and, if freezing fog, included in the weather-related loss of offsite power initiating event in the internal events PRA.
Based on this review, the Fog impact hazard can be considered to be negligible.
Forest Fire / C1 External/Forest fires have the potential to cause a grid-related or switchyard loss of offsite power External Fire event or cause the control room to become uninhabitable. The IPEEE [6.12] screened out these two scenarios from the analysis due to low likelihood and small probabilities of adverse effects.
The loss of offsite power initiator is included in the internal events PRA. The control room habitability scenario is not considered a problem because the resultant smoke will not cause an equipment failure or a reactor trip and because the control room personnel would be notified almost immediately of a major fire by the security patrols. Additionally, control room ventilation would be placed in recirculation (isolation) mode.
Based on this review, the External/Forest Fire impact hazard can be considered to be negligible.
Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
Frost C1 The principal effects of such events would be to cause a loss of offsite power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for CNP.
Hail C1 The principal effects of such events would be to cause a loss of offsite power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for CNP High Summer C1 The principal effects of such events would result in elevated lake temperatures, which are Temperature monitored by station personnel. Should the ultimate heat sink (UHS) temperature exceed the DC Cook Technical Specification (TS) 3.7.9 temperature limit [6.13], an orderly shutdown would be initiated in accordance with appropriate TS required actions. Additionally, if the containment temperature (due to inadequate cooling because of high UHS temperature) exceeds the DC Cook TS 3.6.5 limit (6.13], then an orderly shutdown would be initiated.
Based on this review, the High Summer Temperature impact hazard can be considered to be negligible.
High Tide / High C4 According to UFSAR Section 2.6.2.3 [6.31. the plant is flood protected from the maximum (monthly Lake Level mean) high lake water level; however, a design basis seiche occurring when the lake is at its maximum recorded level will cause flooding in the Turbine Building Screen-house. Safety-related components located in the Turbine Building Screen-house have been evaluated for the condition and flood sensitive components (associated with ESW System) have been protected. These flood sensitive components had their terminations in their respective local terminal boxes reworked to bring them above a postulated seiche flood level, which were performed via EC-46977 [6.14], EC-46978 [6.15], EC-46979 [6.16], and EC-46980 [6.17]. Therefore, protection has been provided for safety-related equipment from flooding, waves, ice storms and other lake-related hazards.
See also "External Flooding."
Hurricane C3 The location of CNP precludes the possibility of a hurricane.
Based on this review, the Hurricane impact hazard can be considered to be negligible.
Ice Cover C1, C4, C5 The principal effects of such events would be to cause a loss of offsite power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for CNP(C1).
The potential hazard of ice clogging the intake is prevented by the De-icing System. This system opens Unit 1 and 2 motor operated valves 1-WMO-16 and 2-WMO-26, respectively, which are in the 591' elevation of the Screen House and provides circulating water discharge from the Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
condensers to the de-icing tunnel to prevent ice buildup in the forebay [6.18] (C5). Therefore, these components (1-WMO-16 and 2-WMO-26) credited for mitigating ice clogging the intake and which are pertinent to this hazard screening will be treated as HSS if categorized, in accordance with the guidance provided in NEI 00-04 Figure 5-6 [6.9]. The IDP will be informed of the basis for ice clogging the intake (Ice Cover) screening during their reviews of categorization results.
See also "External Flooding" (C4).
Based on this review, the Ice Cover impact hazard can be considered to be small.
Industrial or C3 According to UFSAR Section 2.1.2 [6.3], there are no military installations, missile sites, or Military Facility industrial facilities located beyond the DC Cook Nuclear Plant Site boundaries at which an accident Accident might cause interaction with the plant so as to affect public health and safety.
Based on this review, the Industrial or Military Facility Accident impact hazard can be considered to be negligible.
Internal Fire N/A The CNP Internal Fire PRA includes evaluation of risk from internal fire events.
Internal Flooding N/A The CNP Internal Events PRA includes evaluation of risk from internal flooding events.
Landslide C3 According to the IPEEE [6.12], the topography is such that a landslide is not possible.
Based on this review, the Landslide impact hazard can be considered to be negligible.
Lightning C1 Lightning strikes are not uncommon in nuclear plant experience and can result in losses of offsite power. Loss of offsite power events are incorporated into the CNP internal events model through the incorporation of generic data [6.19].
Based on this review, the Lightning impact hazard can be considered to be negligible.
Low lake or river C4 Included under drought.
water level Low winter C4,C5 According to the IPEEE [6.12], the likelihood of the lake freezing solid to a depth of the intake cribs temperature is insignificantly small. Also, there would be ample warning time for the plant to shutdown with respect to freezing of the heat sink. A procedure is used to increase the forebay temperature when signs of localized freezing are observed; this is covered under the "Ice Cover" hazard (C4).
The procedure used to increase the forebay temperature (and level) with Circulating Water Deicing system in service is 1-OHP-4021-057-002, "Placing In/Removing from Service Circulating Water Deicing System" [6.20]. Other procedures that provide guidance during severe weather, such as snow and ice storms (which occur during low winter temperature), are 12-OHP-4022-001-010, Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
"Severe Weather" [6.21], PMP-5055-001-001, "Winterization/Summerization" [6.22], and PMP-5055-SWM-001, "Severe Weather Guidelines" [6.23).
Based on this review, the Low Winter Temperature impact hazard can be considered to be negligible.
Meteorite/Satellite PS4 The frequency of a meteor or satellite strike is judged to be so low as to make the risk impact from strikes such events insignificant. This hazard also was reviewed as part of the IPEEE submittal [6.12) and screened based on low frequency of occurrence.
Based on this review, the Meteorite or Satellite impact hazard can be considered to be negligible.
Pipeline Accident C3 According to the IPEEE [6.12], DC Cook receives no hazardous materials via pipeline. Additionally, the "Evaluation of Offsite Sources of Toxic Gas" [6.24] was reviewed for accidents from pipelines carrying hazardous waste; none were identified.
Based on this review, the pipeline accident hazard can be considered to be negligible.
Release of PS1 The impact of releases of hazardous materials stored on-site was evaluated in the IPEEE submittal Chemicals from [6.12]. The IPEEE documents analyses that were performed to calculate the maximum control On-site Storage room concentrations given an accident involving on-site hazardous materials, which included ammonia (ammonium hydroxide) 29% solution, chlorine (gas), hydrazine (35% solution), and sulfuric acid. The analyses considered the consequences of a rupture of the single largest container and the largest container of the most concentrated solution of these chemicals in their respective locations, their dispersion, and subsequent build-up in the control room ventilation system. This analysis indicates the hazardous materials stored within the site boundaries of CNP present no threat to plant safety; however, the analyses remain only applicable for ammonia (ammonium hydroxide) 29% solution, chlorine (gas), and sulfuric acid. Analysis CA-91-04, which was used in the IPEEE to address a hydrazine spill, was superseded by CA-03-01 [627] because it did not address the then most recent setup of hydrazine storage.
CA-03-01 [6.27] reanalyzed the storage locations of hydrazine, including two 55-gallon drums on the east side of the Turbine Building and in the Auxiliary Chemical Feed Gallery and a 200-gallon tote stored in the west central section of the Turbine Building on the 591' elevation. CA-03-01 identified unacceptable conditions and that a postulated hydrazine spill event would be a challenge to control room habitability. Regulatory Guide 1. 78 established screening criterion for evaluation of control room habitability during a postulated hazardous chemical release. An evaluation is not needed when storage of the toxic chemical is within 0.3 miles from the control room if the storage is less than 100 lbs. of the chemical or partial vapor pressure of such chemical is less than 10 Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
mmHg. A new non-toxic chemical, carbohydrazide, was to replace hydrazine for water treatment.
However, since hydrazine is needed to be available for post-trip steam generator chemistry, hydrazine was to be stored in smaller containers (i.e., 55-gallon drums and in locations on the east side of the Turbine Building elevation 591). In addition, the partial pressure of hydrazine at the storage temperature exceeded 10 mm Hg. This new storage did not meet the requirements of Regulatory Guide 1. 78, and therefore, an analysis of control room habitability was needed and documented in CA-03-01 [6.27].
CA-03-01 was applicable to the post conversion conditions that used carbohydrazide for water treatment and that stored hydrazine in 55-gallon drums on the east side of Turbine Building elevation 591'. The postulated spill assumed a non-mechanistic failure of the worst-case container, a 55-gallon drum. The new storage location for the 55-gallon drums yielded favorable results with respect to control room habitability due to no freestanding fans in the vicinity of the drums. In addition, because of floor drains, the postulated spill event would be terminated with no operator actions required. Plant modifications were not required [6.27].
Airborne hydrazine is conservatively released from the nearest point on the Turbine Building to the control room intakes, and no credit is taken for Turbine Building roof ventilation units. The resulting 8-hour time-weighted average control room concentration is 0.79 ppm. This result is below the OSHA acceptance criterion of 1 ppm. Therefore, it is concluded that the control room remains habitable following a postulated accidental release of liquid hydrazine [6.27].
Based on this review, the Release of Chemicals in Onsite Storage impact hazard can be considered to be negligible.
River Diversion C3 The location of DC Cook along Lake Michigan precludes the possibility of a river diversion.
Based on this review, the River Diversion impact hazard can be considered to be negligible.
Sandstorm C3 According to the IPEEE [6.12]. a sandstorm hazard is not relevant for this region.
Based on this review, the Sandstorm hazard can be considered to be negligible.
Seiche C4 According to UFSAR Section 2.6.2.3 [6.3], the plant is flood protected from the maximum (monthly mean) high lake water level; however, a design basis seiche occurring when the lake is at its maximum recorded level will cause flooding in the Turbine Building Screen-house. Safety-related components located in the Turbine Building Screen-house have been evaluated for the condition and flood sensitive components (associated with ESW System) have been protected. These flood sensitive components had their terminations in their respective local terminal boxes reworked to bring them above a postulated seiche flood level, which were performed via EC-46977 [6.14], EC Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
46978 [6.15]. EC-46979 [6.16), and EC-46980 [6.17). Therefore, protection has been provided for safety-related equipment from flooding, waves, ice storms, and other lake related hazards.
See also "External Flooding."
Seismic Activity N/A The DC Cook Seismic PRA includes evaluation of risk from seismic events.
Snow C4,C5 This hazard is slow to develop (C5) and can be identified via monitoring and managed via normal plant processes. Potential flooding impacts covered under external flooding (C4).
Based on this review, the Snow impact hazard can be considered to be negligible.
Soil Shrink-Swell C1 According to the IPEEE [6.12], the site-suitability evaluation and site development for the plant are designed to preclude the effects of this hazard.
Based on this review, the Soil Shrink-Swell impact hazard can be considered to be negligible.
Storm Surge C4 The location of DC Cook along Lake Michigan precludes the possibility of a sea level driven storm surge. Potential flooding impacts by water levels of Lake Michigan are covered under external flooding.
Based on this review, the Storm Surge impact hazard can be considered to be negligible. See also "External Flooding."
Toxic Gas C4,PS1 According to Reference 6.24,an analysis was performed for cases that involved 34 mobile truck sources, 16 rail car sources and 2 stationary sources at Reliable Disposal. The analyses included both burst and leak cases, except for the one stationary source that was the below ground diesel fuel tank. Scenarios were analyzed for the different cases/chemicals. Each chemical that burst or leaked was analyzed along with the amount available for release and the release location. The evaluation showed whether the release scenario resulted in the control room concentration remaining below the toxicity limit for the chemical or there being at least 2 minutes in the time to act calculation. All release scenarios do not challenge the control room habitability with the exception of three chemicals that require further scrutiny to assess control room habitability. These are the acrolein, bromine, and hydrogen fluoride mobile truck burst scenarios. For the acrolein burst, 22 of the 672 parametric cases run resulted in a time to act of less than 2 minutes, with the minimum time to act as 1.5 minutes. Each of these 22 cases occurred with a combination of wind speed and atmospheric stability class that are usually incongruous (i.e., high wind speed with relatively stable conditions) and only occur less than 0.25% of the time at the plant. For the bromine burst, 16 of the 672 parametric cases run resulted in a time to act of less than 2 minutes, with the minimum time to act as 1. 75 minutes. These 16 cases were for meteorological conditions that occur less than 0.01% of the time at CNP. Similarly, for the hydrogen fluoride burst, only 4 of the 672 cases resulted in a time to act of less than 2 minutes, with the minimum time to act as 1. 75 minutes.
Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
These 4 hydrogen fluoride cases were for meteorological conditions that occur less than 0.01 % of the time at the plant. Given the low number of cases with time to act less than 2 minutes and the infrequency with which the meteorological conditions used to generate those cases actually occur, it is reasonably surmised that neither acrolein, bromine, nor hydrogen fluoride mobile truck burst scenarios pose a challenge to control room habitability (PS1 ).
See also "Release of Chemicals from On-site Storage," "Pipeline Accident," and "Industrial or Military Facility Accident" (C4).
Based on this review, the Toxic Gas impact hazard can be considered to be negligible.
Transportation C1, C4 For shipping impact hazards, according to the IPEEE [6.12], due to the physical location of DC Accidents Cook buildings and structures, the only danger to the plant from ships/barges are from those that run-aground that collapse the circulating water intake cribs resulting in flow obstruction of all three circulating water system intake lines. This results in the shutdown of both units; however, essential service water system flow can be maintained to remove heat from the component cooling water system and other essential service water system loads by opening sluice gates (1-WMO-17, 2 WMO-27) between the discharge chambers and forebay, which allows water from the discharge chambers to enter the forebay to supply the essential service water pumps [6.18]. Therefore, no plant damage leading to core damage or radiological release is expected as a result of a shipping accident. Additionally, DC Cook receives no hazardous materials via ship or barge (C1).
Therefore, sluice gates (1-WMO-17, 2-WMO-27) between the discharge chambers and forebay credited for allowing water from the discharge chambers to enter the forebay to supply the essential service water pumps (in case there is flow obstruction of all three circulating water system intake lines) and which are pertinent to this hazard screening will be treated as HSS if categorized, in accordance with the guidance provided in NEI 00-04 Figure 5-6 [6.9]. The IDP will be informed of the basis for shipping impact (transportation accident) screening during their reviews of categorization results.
See also "Aircraft Impacts" (C4).
Based on this review, the Transportation Accident impact hazard can be considered to be negligible.
Tsunami C3 The location of DC Cook along Lake Michigan precludes the possibility of a tsunami.
Based on this review, the Tsunami impact hazard can be considered to be negligible.
Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
Turbine-generated C1 Per UFSAR Section 14.1.13.1 [6.3], the turbine-generators are safe and reliable and the chance of Missiles a failure during normal operation, which could endanger the reactor and associated Seismic Class I nuclear systems, is extremely small. Additionally, the chance of a turbine running away out of control to destruction is also extremely small. For the Unit 1 and Unit 2 Alstom low-pressure turbines, turbine missile probability analysis indicates the probability of the generation of a turbine missile (including turbine overspeed conditions) is below the NRC limit which would require missile analysis. Therefore, no additional missile analysis is required for the Unit 1 and Unit 2 Alstom low-pressure turbines.
Based on this review, the Turbine-Generated Missiles' impact hazard can be considered to be negligible.
Volcanic Activity C3 Not applicable to the site because of location (no active or dormant volcanoes located near plant site).
Based on this review, the Volcanic Activity impact hazard can be considered to be negligible.
Waves C4 Waves associated with external flooding are covered under that hazard.
Based on this review, the Waves impact hazard can be considered to be negligible.
PRA-EXT-HAZ-SCRN [26] References
6.1. Nuclear Energy Institute, "NEI 50.69 LAR Template," retrieved from https://www.nei.org.
6.2. Doc. No. PRA-NB-QU, Revision 5, "Internal Events Quantification Notebook," April 6, 2018.
6.3. Indiana and Michigan Power, D.C. Cook Nuclear Plant, Updated Final Safety Analysis Report, Revision 30.
6.4. Doc. No. SD-990930-004, Revision 12, "Probability of Tornado Missile Strike on Targets at D.C. Cook Nuclear Plant," May 22, 2019.
6.5. Doc. No. PRA-TORMIS-QNT-001, Revision O, "Risk Quantification to Support TORMIS Analysis for Cook Tornado Assessment," August 2, 2013.
6.6. Letter from J. P. Gebbie, Indiana Michigan Power Company (l&M), to the NRC, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2, Response to March 12, 2012, Request for Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
Information, Enclosure 2, 'Recommendation 2.1: Flooding,' Required Response 2, Hazard Reevaluation Report," dated March 6, 2015, AEP-NRC-2015-14, ADAMS Accession No.
6.7. Letter from J.P. Gebbie, Indiana Michigan Power Company (l&M), to the NRC, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2, Update to Interim Action Plan re. Flood Hazard Reevaluation," dated December 14, 2015, AEP-NRC-2015-116.
6.8. PMP-5091-FLD-001, Revision 9, "Flood Protection Program Implementation," 8/30/2021.
6.9. Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline,"
Revision 0, July 2005.
6.10. Federal Aviation Administration, "Airport Data & Contact Information, retrieved from https://www.faa.gov/airports/airport_ safety/airportdata_ 5010.
6.11. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," LWR Edition, Revision 3, November 1978.
6.12. American Electric Power Service Corporation, Donald C. Cook Nuclear Plant Units 1 and 2, "Individual Plant Examination of External Events Summary Report," April 1992.
6.13. Donald C. Cook Nuclear Plant, Unit 1/2 Technical Specifications, September 5, 2019.
6.14. EC-46977, "Flood Control Wiring Protection for Unit 1 A Train," Revision 0, 9/21/06.
6.15. EC-46978, "Flood Control Wiring Protection for Unit 1 B Train," Revision 0, 9/21/06.
6.16. EC-46979, "Flood Control Wiring Protection for Unit 2 A Train," Revision 0, 9/21/06.
6.17. EC-46980, "Flood Control Wiring Protection for Unit 2 B Train," Revision 0, 9/21/06.
6.18. OP-12-5119, "Flow Diagram Circulating Water, Priming System & Screen Wash Units No.
1 & 2, Revision 88, May 26, 2017.
6.19. Idaho National Laboratory, INL/EXT-18-45359, "Analysis of Loss of Offsite Power Events: 1987-2017," August 2018.
6.20. 1-OHP-4021-057-002, "Placing In/Removing from Service Circulating Water Deicing System," Revision 26, 1/20/2020.
6.21. 12-OHP-4022-001-010, "Severe Weather," Revision 21, 7/21/2017.
6.22. PMP-5055-001-001, "Winterization/Summerization,", Revision 32, 9/25/2019.
6.23. PMP-5055-SWM-001, "Severe Weather Guidelines," Revision 10, 7/27/2017.
6.24. Calculation MD-12-MSC-003-N, "Evaluation of Offsite Sources of Toxic Gas," Revision 5, Red Wolf Associates, December 10, 2019.
6.25. NUREG/CR-4461, "Tornado Climatology of the Contiguous United States, Revision 2, February 2007.
Attachment 4 of Enclosure 2: EXTERNAL HAZARDS SCREENING
6.26. EPRI 3002003107, "High Wind Risk Assessment Guidelines," June 2015.
6.27. CA-03-01, "Control Room Habitability Following Postulated Release of Hydrazine,"
Revision 3, 6/26/2017.
6.28. Calculation No. SD-991215-004, Revision 1, "Seismic Analysis of Tornado Missile Barrier for EDG Combustion Air Intake," 5/20/2000.
Attachment 5 of Enclosure 2: PROGRESSIVE SCREENING APPROACH FOR ADDRESSING EXTERNAL HAZARDS
Event Analysis Criterion Source Comments
Initial Preliminary c1. Event damage potential is < events for NUREG/CR-2300 and Screening which plant is designed. ASME/ANS Standard RA-Sa-2009 C2. Event has lower mean frequency and NUREG/CR-2300 and no worse consequences than other ASME/ANS Standard RA-Sa-2009 events analyzed.
C3. Event cannot occur close enough to NUREG/CR-2300 and the plant to affect it. ASME/ANS Standard RA-Sa-2009 C4. Event is included in the definition of NUREG/CR-2300 and Not used to screen.
another event. ASME/ANS Standard RA-Sa-2009 Used only to include within another event.
C5. Event develops slowly, allowing ASME/ANS Standard RA-Sa-2009 adequate time to eliminate or mitigate the threat.
Progressive PS1. Design basis hazard cannot cause a ASME/ANS Standard RA-Sa-2009 Screening core damage accident.
PS2. Design basis for the event meets the NUREG-1407 and criteria in the NRC 1975 Standard ASME/ANS Standard RA-Sa-2009 Review Plan (SRP).
PS3. Design basis event mean frequency NUREG-1407 as is < 1E-5/y and the mean conditional modified in core damage probability is < 0.1. ASME/ANS Standard RA-Sa-2009 PS4. Bounding mean CDF is < 1E-6/y. NUREG-1407 and ASME/ANS Standard RA-Sa-2009 Detailed PRA Screening not successful. Probabilistic Risk NUREG-1407 and Assessment (PRA) needs to meet ASME/ANS Standard RA-Sa-2009 requirements in the ASME/ANS PRA Standard.