Text

Response to an Apparent Violation in NRC Inspection Report 05000263/2013008 (EA-13-096)
	ML13233A068
Person / Time
Site:	Monticello
Issue date:	07/11/2013
From:	Schimmel M; Northern States Power Co, Xcel Energy
To:	; Document Control Desk, NRC/RGN-III
References
	EA-13-096, L-MT-13-062 IR-13-008
	Download: ML13233A068 (99)
	v • d • e

XcelEnergy Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 July 11, 2013 L-MT-13-062 EA-13-096 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22 Response to an Apparent Violation in NRC Inspection Report 05000263/2013008 (EA-1 3-096)

References:

1) Letter from Nuclear Regulatory Commission (NRC) to Mr. Mark A.

Schimmel, "Monticello Nuclear Generating Plant, NRC Inspection Report 05000263/2013008; Preliminary Yellow Finding," dated June 11, 2013 (Accession Number ML13162A776)

2) Letter from Northern States Power - Minnesota to NRC, "Notification of Intention Regarding NRC Inspection Report 05000263/2013008 (EA-13-096)," dated June 19, 2013 By the above referenced letter dated June 11, 2013, the NRC transmitted Inspection Report 05000263/2013008 for the Monticello Nuclear Generating Plant (MNGP). In the inspection report, the NRC identified one finding and apparent violation with a preliminary significance of Yellow for MNGP. In the referenced letter, the NRC stated that the site had failed to maintain a flood procedure, A.6, "Acts of Nature", such that it could support the timely implementation of flood protection activities within the 12 day timeframe credited in the design basis, as stated in the updated safety analysis report (USAR).

Northern States Power Company - Minnesota (NSPM) reviewed the apparent violation and, pursuant to the provisions of the choice letter, prepared a written response to the apparent violation. NSPM agrees that the failure to maintain a flood plan to protect the site from external flooding events is a violation of Technical Specification 5.4.1.a.

This letter submits additional information for the NRC's consideration in its final determination of the significance of the apparent violation. The enclosures address the following:

e-7A

Document Control Desk L-MT-1 3-062 Page 2 of 5 Response to Apparent Violation (Enclosure 1)

NSPM agrees that the failure to maintain an adequate flood plan to protect the site from external flooding events is a violation of Technical Specification 5.4.1 .a. NSPM is taking this failure to protect the site from external flooding very seriously and has used it to reinforce NSPM's policy and commitment to safety as a top priority in our Emergency Response plans, response to acts of nature, and effective corporate governance and oversight. The site and nuclear fleet are taking corrective actions to ensure protection of the radiological health and safety of the public in the event of an external flooding worst case scenario. A summary of the corrective actions to resolve the performance deficiency is presented in Enclosure 1.

As part of those actions, NSPM is performing cultural assessments focusing on decision making, effective communication, and closure follow-through not only at the site levels, but across the nuclear fleet to maximize learning from this situation.

Probabilistic Risk Analysis (Enclosure 2)

NSPM developed additional information providing further insight into the probability of a Probable Maximum Flood (PMF) at the Monticello site for the NRC's consideration. The report provides probabilistic risk analyses to support a best-estimate assessment of the significance of this finding as well as bounding analyses to support final significance determination prior to corrective actions taken by the site. The best-estimate analysis incorporated the assumptions necessary to support the assessment of a finding related to an external flooding event. Results are shown in the table below:

Nominal, Best Sensitivity 1: Sensitivity 2:

Estimate Bounding Flood SPAR-H HRA Freauencv Probabilities CDF 1.04E-06 3.1 OE-06 1.83E-06 ACDP 8.92E-07 2.66E-06 1.57E-06 The full results of the event tree quantification are summarized in Enclosure 2.

Monticello Nuclear Generating Plant Flood Protection Analysis (Enclosure 3)

The postulated PMF for the MNGP is compared to other site Mississippi river conditions in the table below. The PMF is not an instantaneous event, but rather a slowly developing evolution that allows for plant staff to monitor, predict, prepare, and implement appropriate actions to provide the required flood protection. Since actions have been taken to procure the bin wall and levee materials, performance of a reasonable simulation demonstrated that the levee and bin wall system can now be installed within the available time as defined in the licensing basis.

Document Control Desk L-MT-1 3-062 Page 3 of 5 Normal and Flooded River Flow Rates and Water Elevations Mississippi River Flow Rate (cfs) Water Elevation Condition (ft. msl)

Normal 4,600 905 Maximum Recorded 51,000 916 (1965) 1000 Year Flood -90,000 (1) 921 Probable Maximum 364,900 939.2 Flood 364,900_ 939.2 A report entitled "Monticello Flood Protection," was prepared for Monticello and addresses the aspects of flood protection for which MNGP was licensed and is included in Enclosure 3.

Annual Exceedance Probability (Enclosure 4)

Annual river exceedance probabilities based on annual peak flood estimates at the Monticello site were developed to support the probabilistic risk assessment. The probability of a PMF at the site was determined to be extremely low. provides a copy of the report entitled, "Annual Exceedance Probability Estimates for Mississippi River Stages at the Monticello Nuclear Generating Plant based on At-site Data for Spring and Summer Annual Peak Floods", June 28, 2013, developed by RAC Engineers & Economists.

Stakeholder Outreach (Enclosure 5)

NSPM hosted an open house to share information with its community neighbors on its operations and preparedness to handle potential emergencies and how it would respond to flooding, earthquakes and other unforeseen challenges.

The key message presented to visitors was that safety and security at the NSPM nuclear generating plants are top priorities for Xcel Energy. Further, that we understand the industry, NRC, and public's demand of higher safety standards and flood preparedness at the nation's nuclear power plants in the wake of events such as 9/11 and Fukushima Daiichi. The Monticello Flood Protection Strategy was identified and explained to demonstrate that the site is capable of withstanding a PMF and that the site is incorporating lessons learned from the industry to improve and assure protection methods.

Safety Culture Review NSPM agrees it missed an opportunity within its control to identify challenges to the implementation of the A.6 procedure, leading to the identified apparent violation. As such, NSPM assembled an expert panel to examine the behavioral and cultural aspects impacting decision making within the nuclear business unit. This activity was chartered

Document Control Desk L-MT-13-062 Page 4 of 5 as an immediate and interim measure preceding the extensive root cause evaluation that will be performed to identify the full magnitude of this issue, associated causes, and corrective actions to prevent recurrence.

This expert panel assembled to examine safety culture within its nuclear organization was comprised of five independent consultants and one Xcel representative. The team reported directly to the Vice President of Nuclear Operations Support. A phased approach is being utilized to examine the behavioral and cultural aspects impacting decision making within the nuclear business unit. Three phases are planned to examine this subject: Phase (1) is specifically focused on the Monticello flooding issue, Phase (2) more broadly examines Monticello issues and Phase (3) examines Prairie Island issues.

While the phases are specific to the individual sites, the scope includes developing an understanding of the corporate culture and influence beyond a site-centric examination of the behavioral and environmental influences. To date Phase (1) has been completed with scheduling of Phase (2) and (3) expected to commence and complete over the next few months. The initial phase identified improvement opportunities in the areas of decision making, leadership behaviors, and questioning attitude regarding the station's preparedness for a PMF. The results of this assessment have been insightful and will be applied across the nuclear fleet to ensure a healthy safety culture exists. Opportunities have been identified to strengthen fleet and Nuclear Oversight accountability for providing oversight to proactively detect performance gaps.

Interim actions are in place for the short term to focus on the areas for improvement, and longer term actions are in development.

Summary NSPM respectfully requests that the NRC consider the enclosed information in its final determination of the significance of the finding. Notwithstanding our assessment of the significance of the finding, NSPM clearly understands our performance shortcomings concerning flood protection for the entire spectrum of possible flooding events at the MNGP. Corrective actions have already been completed to address the NRC's identified performance deficiency. Additionally, we unequivocally acknowledge the need for overall performance improvement at MNGP. Actions are underway to ensure that the lessons learned from this finding are applied more broadly to overall performance.

Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.

Mark A. Schimmel Site Vice-President Monticello Nuclear Generating Plant Northern States Power Company-Minnesota

Document Control Desk L-MT-1 3-062 Page 5 of 5

Enclosures:

Enclosure 1 - Response to Apparent Violation Enclosure 2 - External Flooding Evaluation for Monticello Nuclear Generating Plant Enclosure 3 - Monticello Flood Protection Enclosure 4 - Annual Exceedance Probability Estimates Enclosure 5 - Stakeholder Outreach cc: Regional Administrator, Region III, USNRC Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC

Enclosure I Response to Apparent Violation EA-1 3-096 NRC Inspection Report 05000263/2013008 Monticello Nuclear Generating Plant 3 Pages Follow

Northern States Power Company - Minnesota Response to Preliminary Yellow Finding NRC Finding Summary The inspectors identified a preliminary Yellow finding with substantial safety significance and associated apparent violation (AV) of Technical Specification 5.4.1 for the licensee's failure to maintain a flood plan to protect the site from external flooding events. Specifically, the site failed to maintain flood Procedure A.6, "Acts of Nature," such that it could support the timely implementation of flood protection activities within the 12 day timeframe credited in the design basis as stated in the updated safety analysis report (USAR).

The inspectors determined that the licensee's failure to maintain an adequate flood plan consistent with the USAR was a performance deficiency, because it was the result of the failure to meet the requirements of TS 5.4.1 .a, "Procedures;" the cause was reasonably within the licensee's ability to foresee and correct; and should have been prevented. The inspectors screened the performance deficiency per Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, dated September 7, 2012, and determined that the issue was more than minor because it impacted the 'Protection Against External Factors' attribute of the Mitigating Systems Cornerstone and affected the cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, if the necessary flood actions cannot be completed in the time required, much of the station's accident mitigation equipment could be negatively impacted by flood waters.

NRC Baseline Significance Determination Process Review As part of the process, the Region III Senior Reactor Analyst (SRA) developed an event tree model to perform a bounding quantitative evaluation. The model presents an external flood event that exceeds grade level (930 ft. MSL) and requires implementation of Procedure A.6, "Acts of Nature" Section 5.0.

NSPM Response NSPM agrees that a performance deficiency exists. Procedure A.6, "Acts of Nature", at the time of the violation, did not provide sufficient guidance to execute mitigation strategies for a probable maximum flood (PMF) event. Adequate management oversight and engagement was not provided to ensure that the Monticello external flood mitigation procedure and strategies met expected industry standards and licensing basis requirements.

Actions have been completed to reduce the flood mitigation plan timeline by pre-staging equipment and materials required for bin-wall levee construction, improving the quality of the A.6 "Acts of Nature" procedure and pre-planning work orders necessary to carry out the A.6 actions.

Summary of Corrective Actions:

Acquired materials required for flood mitigation including, but not limited to:

" Hardware and components for construction of Bin-Wall

" Clay for levee construction (30000 cubic yards)

Rip-Rap stone for levee construction (1700 cubic yards)

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Sand for levee construction and filling sandbags (11000 cubic yards)

" Sand bagging machine (Capacity 1600 sand bags/hr)

Manual sand bag filling tools (25 on site)

" Gas Sump Pumps

Electric Sump Pumps

" Crushed concrete for alternate road access (2400 cubic yards)

Preventative maintenance plans are being developed for new flood mitigation equipment

Performance of reasonable simulation of major steps required by procedure A.6 "Acts of Nature" Section 5.0, including building of bin-wall sections, sandbagging, placement of various covers, and relocation of vital equipment.

Extensive procedure revisions to enhance feasibility of actions and reduce overall time required to execute the strategy.

Table top exercises of new revisions performed to ensure practicality.

Development of work orders to provide more detail for execution of steps within procedure A.6, "Acts of Nature" Section 5.

" The existing flood prediction surveillance was revised to occur on a monthly basis instead of yearly and contains provisions to continually monitor river predictions if certain conditions are met.

Meetings with the National Weather Service were held to develop more robust prediction capabilities and options.

Enhanced construction drawings of levee and bin-wall to provide more detail

Updated existing contracts and memorandums of understanding with vendors to assure equipment availability.

A modification is also in the design phase to install the base of the bin-wall on the west side of the Intake structure, simplify construction on the east side of the Intake Structure, and also update the steel plate design for protection of the Intake Structure.

Review of NRC Significance Determination NSPM has developed additional information providing new insight into the probability of a PMF at the Monticello site for your consideration. A report entitled "External Flooding Evaluation for Monticello Nuclear Generating Plant" was prepared by Hughes Associates, Inc., for NSPM. The report provides a best-estimate assessment of the significance of this finding. Two (2) sensitivities were performed to assess the bounding risk, addressing some of the uncertainty associated with this assessment.

The first sensitivity study provides the risk assessment if a bounding annual exceedance probability is assumed. As noted in the Hughes' report, the uncertainty associated with extreme flooding can be addressed by artificially restraining the AEP to a value of no less than 1E-05/year.

When this restraint is assessed, the ACDP is 2.66E-06. of this letter provides a copy of the report entitled, "Annual Exceedance Probability Estimates for Mississippi River Stages at the Monticello Nuclear Generating Plant based on At-site Data for Spring and Summer Annual Peak Floods", June 28, 2013, developed by RAC Engineers & Economists.

The second sensitivity address the different methodologies available for quantifying the Human Error Probability (HEP) associated with the manual operation of RCIC (Reactor Core Isolation Cooling) and HPV (Hard Pipe Vent). This sensitivity provides quantification using The SPAR-H Page 2 of 3

Human Reliability Analysis Method, NUREG/CR-6883. When the simplified SPAR-H methodology is used, the assessment results in a ACDP of 1.57E-06.

The result of the nominal, best-estimate assessment and the two (2) sensitivities performed are shown in the table, below, for ease of reference.

Sensitivity 1:

Bounding Flood Sensitivity 2: SPAR-H Nominal Best- Frequency HRA Probabilities Estimate CDF 1.04E-06 3.1OE-06 1.83E-06 ACDP 8.92E-07 2.66E-06 1.57E-06 of this letter provides a copy of a report entitled, Report Number 1SML16012.000-1, "External Flooding Evaluation for Monticello Nuclear Generating Plant," developed by Hughes Associates for consideration.

Page 3 of 3

Enclosure 2 Monticello Nuclear Generating Plant "External Flooding Evaluation for Monticello Nuclear Generating Plant" ISMLI16012.000-1 Hughes Associates 62 Pages Follow

.IHUGHES EASSOCIATES ENGINEERS CONSULTANTS SCIENTISTS External Flooding Evaluation for Monticello Nuclear Generating Plant 1SML16012.000-1 Prepared for:

Xcel Energy Project Number: 1SML16012.000 Project

Title:

Monticello External Flooding SDP Revision: 1 Name Date Preparer: Erin Collins/Paul Amico/Suzanne Loyd 7/8/2013

~ ~" ErinP.Collins 2013.07.082018:01 00-Reviewer: Pierre Macheret 7/8/2013

.M0 1-.U.

2"13007 P.",: o Review Method Design Review E] Alternate Calculation E]

Approved by: Francisco Joglar Francisco Joglar. ,.

N.l: "013.07- 0827*'0 I

7/8/2013

ISML-16012.000-1 Table of Contents TABLE OF CONTENTS

1.0 INTRODUCTION

.................................................................................................. 1

2.0 REFERENCES

................................................................................................. 2 3.0 METHODOLOGY and analysis ........................................................................ 4 3.1 Event Tree Analysis ............................................................................. 4 3.1.1 Evaluation of Flood Frequency ................................................... 4 3.1.2 Early W arning Probability ........................................................... 7 3.1.3 Protection of the Reactor Building ............................................... 7 3.1.4 Manual Local Operation of RCIC and the Hard Pipe Vent .......... 8

4.0 CONCLUSION

S ............................................................................................. 15 APPENDICES TABLE OF CONTENTS ................................................................... 16 Revision I Page ii

1SML16012.000-1 Introduction

1.0 INTRODUCTION

This analysis was developed to address the significance of a finding that was received by the Monticello Nuclear Generating Plant (MNGP) associated with External Flooding hazards. This SDP is summarized in NRC Letter EA-13-096 (Reference 5). The analysis quantifies the core damage frequency associated with a flood exceeding 930' at MNGP.

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ISML16012.000-1 References

2.0 REFERENCES

1. Annual Exceedance Probability Estimates for Mississippi River Stages at the Monticello Nuclear Generating Station based on At-Site Data for Spring and Summer Annual Peak Floods, David S. Bowles and Sanjay S. Chauhan, RAC Engineers and Economists, June 28, 2013.

2. Hydrologic Atlas of Minnesota, Division of Water, Department of Conservation, State of Minnesota, 1959.

3. US Geological Survey, Guidelines for Determining Flood Flow Frequency, Bulletin

17B, Hydrology Subcommittee, Interagency Committee on Water Data, Office of Water Data Coordination, 1982.

4. A Framework for Characterization of Extreme Floods for Dam Safety Risk Assessments, Robert E. Swain, David Bowles and Dean Ostenea, Proceedings of the 1998 USCOLD Annual Lecture, Buffalo, New York, August 1998.

5. NRC Letter EA-13-096,

Subject:

Monticello Nuclear Generating Plant, NRC Inspection Report 05000263/2013008; Preliminary Yellow Finding, United States Nuclear Regulatory Commission, Region III, 11 June 2013.

6. A Preliminary Approach to Human Reliability Analysis for External Events with a Focus on Seismic, EPRI 1025294, EPRI, December 2012.

7. Interim Staff Guidance for Performing the Integrated Assessment for External Flooding, Appendix C: Evaluation of Manual Actions, JLD-ISG-2012-05, Revision 0, U.S. NRC Japan Lessons-Learned Project Directorate, November 30, 2012.

8. Job Performance Measure JPM-A.8-05.01-001, Manual Operation of RCIC, Rev. 0, Task Number NL217.108 - Operation of RCIC without Electric Power, 3 timed exercises performed 17 June 2013.

9. Job Performance Measure JPM-A.8-05.08.001, Manually Open Containment Vent Lines, Rev. 0, Task Number FB008.007, 4 timed exercises performed 17 June 2013.

10. Hughes Associates Record of Correspondence, RCIC Manual Operation e-mails with Xcel Energy during June - July 2013, Hughes Associates, Baltimore, MD, 7 July 2013.

11. Hughes Associates Record of Correspondence, Hard Pipe Vent Manual Operation e-mails with Xcel Energy during June 2013, Hughes Associates, Baltimore, MD, 7 July 2013.

12. Monticello Procedures:

" 8900, OPERATION OF RCIC WITHOUT ELECTRIC POWER, Revision 2

" C.5-1 100, RPV CONTROL flowchart, Revision 11

" C.5-1200, PRIMARY CONTAINMENT CONTROL flowchart, Revision 16

" C.5-3505-A, Revision 10

A.6, ACTS OF NATURE, Revision 43

A.8-05.08, Manually Open Containment Vent Lines, Revision 1 Revision 2 Page 2

ISML16012.000-1 References 0 A.8-05.01, Manual Operation of RCIC, Revision 2

13. The EPRI HRA Calculator Software Users Manual, Version 4.21, EPRI, Palo Alto, CA, and Scientech, a Curtiss-Wright Flow Control company, Tukwila, WA.

14. NUREG/CR-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, (THERP) Swain, A.D. and Guttman, H.E., August 1983.

15. NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines, Draft Report for Public Review and Comment, November 2009.

16. NUREG- 1852, Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire, October 2007.

17. NUREG/CR-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, (THERP) Swain, A.D. and Guttman, H.E., August 1983.

18. Whaley, A.M, Kelly, D.L, Boring, R.L. and Galyean, W.J, "SPAR-H Step-by-Step Guidance", INL/EXT-10-18533, Revision 2, Idaho National Laboratory, Risk, Reliability, and NRC Programs Department, Idaho Falls, Idaho, May 2011.

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1SML16012.000-1 Methodology and Analysis 3.0 METHODOLOGY AND ANALYSIS 3.1 Event Tree Analysis Event trees were developed to calculate the core damage frequency (CDF) associated with External Floods. Event trees were developed for both a best estimate case as well as sensitivity cases. Floods were evaluated at three particular heights - 917' but less than 930', 930' but less than 935', and greater than or equal to 935'. The first flood height range (917' to 930') was evaluated for frequency but was not deemed feasible to cause core damage since much of the plant's critical safety equipment is not threatened unless flood levels exceed the 930' elevation.

The following sections describe the development of the flood frequency and conditional probabilities of the other events that may lead to core damage events.

Additionally, it is agreed that, as stated in the NRC letter EA- 13-96 (Reference 5), the evaluation of large early release frequency (LERF) risk is assumed to be no more significant than CDF-based risk; therefore, no evaluation of LERF was performed for this analysis.

3.1.1 Evaluation of Flood Frequency MNGP teamed with industry experts to evaluate the frequency of major floods in detail. Dr.

David Bowles was the lead author for the flood frequency analysis. To ensure that the flood frequency of exceedance analysis was sound, an independent review of the analysis was performed as part of this analysis (Reference 1). The review resulted in some recommendations for improvement that were incorporated by Dr. Bowles and his team prior to the development of the event trees that are included in this report. This section documents the results of that final flood frequency analysis.

The frequencies of Mississippi River floods at MNGP for flood heights of 917', 930' and 935' were estimated using stream-flow data from 1970 - 2012. This estimation is described in "Annual Exceedance Probability Estimates for Mississippi River Stages at the MNGP based on At-Site Data for Spring and Summer Annual Peak Floods" (Reference 1).

Separate flood frequency relationships were developed for spring and summer annual peak floods based on the Hydrologic Atlas of Minnesota (Reference 2) and an examination of flow records.

a) Spring annual peak floods generally peaked in the period March to May, but if it was clear from examination of the hydrograph that a snow melt flood event peaked in June then that peak was used.

b) Summer annual peak floods generally peaked in June to October, but flood peaks in June associated with snow melts were excluded since they were assigned to spring floods.

Frequencies were estimated based on extrapolation of flood frequency relationships developed from at-site flow data. The mean daily annual peak stages were obtained for spring and summer annual peak floods. The daily annual peak stages were converted to daily annual peak flows for spring and summer floods using a combined rating curve described in Reference 1. The Expected Moments Algorithm (PeakfqSA) was applied using methodology described in the current draft of the upcoming revision to USGS Bulletin 17B (Reference 3). The EMA software provided annual exceedance probability (AEP) estimates down to 1 in 10,000/yr. Annual peak discharge was plotted on a Log scale and AEPs on a z-variate scale (corresponding to a Normal Revision 2 Page 4

ISML16012.000-1 Methodology and Analysis probability distribution) to allow AEPs lower than 1 in 10,000/yr to be estimated by linear extrapolation.

The resulting median estimates for spring and summer floods at 917', 930' and 935' river heights were as follows:

Table 3-1 Median Flood Frequency Estimates at MNGP Site (Ref. 1)

Elevation 917' Elevation 930' Elevation 935' Spring Floods 6.3E-3/yr <lE-9/yr <1E-9/yr Summer Floods 1.7E-3/yr <1E-9/yr <1E-9/yr A full family of flood hazard curves is provided in the Bowles report. There is a substantial amount of aleatory uncertainty in the development of these curves, resulting in very wide confidence bounds as can be seen in the figures and tables in the report, especially for the flood levels of interest. For example, the elevation 917' 95th percentile estimate is 3.4E-2/yr and the 5th percentile estimate is 5.9E-7/yr for spring floods. This is indicative of the limited data available in the 1970 - 2012 year period. In addition, not all potential flood influences were seen in the 42 years of experience. Moreover, there is additional epistemic uncertainty due to simplifications used in the modeling process, For example, regional flood impacts were not considered, nor were the potential effect of ice blockage or changes in flood protective features such as dikes along the river. To address these as well as other potential flood contributors, the mean hazard for the purposes of quantification will be represented by the 84th percentile, which is the median plus one standard deviation for a lognormal form distribution. This is a common practice to provide margin to account for the uncertainties.

As a further consideration, there is a general consensus that the practical limit on AEP extrapolation is no better than 1.OE-5/yr no matter how much information, including information on paleofloods, is available (see for example, "A Framework for Characterization of Extreme Floods for Dam Safety Risk Assessments" (Ref. 4). It is believed that all of these factors would tend to increase the estimated frequency of floodings. Some of these uncertainties could be addressed in a more detailed analysis, such as a Monte Carlo rainfall-runoff approach, but others, such as the potential effect of ice blockage, would remain. In the case of this evaluation, there additional factors will be addressed by way of a sensitivity study, discussed at the end of this section.

The 84th percentile is the median plus one standard deviation, which is a common "margin value" that addresses uncertainty without being overly conservative. The margin is to cover the wide statistical uncertainty in the distribution plus the modeling uncertainty that comes from some of the issues we discussed, such as not considering ice blockage or other downstream flow bottlenecks, only considering site data, and not actually modeling the flows and performing a simulation. This results in the following flood frequency estimates:

The results presented in the Bowles report support the use of the 84th %-tile as a reasonable conservative, but not overly conservative, representation of the mean. A manually generated approximation of the mean (described in Ref. 1 - the code used cannot itself generate a mean) is shown to be in the range of the 70th percentile, plus or minus about 10 percentile. This is the result of the aleatory uncertainty. The use of the 84th percentile provides some additional Revision 2 Page 5

1SM L16012.000-1 Methodology and Analysis margin to account for epistemic uncertainty and give confidence that the "true mean" is not likely to exceed this value.

Treating the 84th percentile as the mean, the process of developing the initiating event frequencies is straightforward. It is typical in external hazard PRA to create initiating events by discretizing the hazard curve. Because for external flooding there are a clear series of flood levels of concern, the selection of the initiating events is clear. For MNGP, the levels of concern are 917', 930', and 935', with the exceedance of each level causing the same impact on plant systems until the flood exceeds the next level of concern. Therefore the initiating events can be defined as follows:

" IE1 - level >917' and <930'

" IE2 - level >930' and <935'

" IE3 - level >935' Using the 84th percentile values to represent the mean, we get exceedance probabilities for each level of concern as follows:

Table 3-2 84th Percentile Exceedance Frequencies (Used as Means) (Ref. 1)

Elevation 917' Elevation 930' Elevation 935' Spring Floods. 2.0E-02/yr 7.5E-08/yr <1E-09/yr Summer Floods 8.8E-03/yr 8.8E-06/yr 9.7E-07/yr Total 2.9E-021yr 8.9E-061yr 9.7E-071yr The values for spring and summer can be summed because they are each calculated on a per-calendar-year basis and because we are using these values to represent the "true mean" of the distribution, albeit conservatively.

Because these values are exceedance (i.e., the frequency that a flood exceeds a specified level),

the way to determine the frequency of a flood in a given range is to subtract the frequency of exceedance of the upper flood in the range from the frequency of exceedance of the lower flood in the range.

One event tree was developed to account for all floods that were greater than 930' (including those greater than 935'). The frequency of exceedance for the 930' floods was used and a conditional probability was applied to account for the floods within the 930' to 935' range and floods greater than 935'. The frequency shown as IE2 in Table 3-3 was used in the event tree.

The conditional probability that the flood was indeed greater than 935' was then applied in a subsequent branch to determine the CDF for sequences specific to floods greater than 935' (value of 0.109). Table 3-3 below shows the initiating event frequencies used for this assessment.

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1SML16012.000-1 Methodology and Analysis I SMLI 6012.000-1 Methodology and Analysis Table 3-3 Best Estimate Initiating Event Frequencies Initiating Event Initiating Event Frequency 1

IEl 2.9E-02/yr 2

IE2 8.90E-06/yr Not deemed feasible for core damage sequences. Not evaluated for this analysis.

2This frequency Is represented in the event tree along with a conditional probability that the flood will be >935' to evaluate the correct range of flood heights.

As discussed earlier, there is the question of the practical limit of extrapolation of flooding data.

Therefore, in addition to the initiating event frequencies presented above, a sensitivity analysis was performed limiting the flood frequency to no less than 1.OE-5/yr, consistent with the consensus reached in Ref. 4 concerning the limit of credible extrapolation for annual flood exceedance probability.

Table 3-4 Sensitivity Case Initiating Event Frequencies Initiating Event Initiating Event Frequency IE1 2.9E-02/yr IE22 2.OE-05/yr

'Not deemed feasible for core damage sequences. Not evaluated for this analysis.

2This frequency is represented In the event tree along with a conditional probability that the flood will be >935' to evaluate the correct range of flood heights.

It could be argued that the limit should apply across the 930' events (i.e., that the hazard curve goes flat at 1E-5/yr) and that it is sufficient to ignore the 935' event (because F(IE2930<=1v1<935) =

fexceed(930) - fcxceed(935) = 0) and simply evaluate the 930' flood at a frequency of lE-5/yr. Since this is a sensitivity case and not the best estimate, it was decided to assign 2E-5/yr to IE2 for purposes of providing the sensitivity insights and use a conditional probability of 0.5 to account for floods that are >935'.

In the event trees (shown in Appendix A), the flood frequencies are represented by the headings "EXTERNAL FLOOD >930"' and "<935.' for IE2 events. In addition, event trees used for sensitivities have similar headings.

3.1.2 Early Warning Probability Although it was considered qualitatively, there was no basis determined for giving credit to the potential for early warning of a flood >930'. The event tree model shows a failure probability of 1.0 for this node.

3.1.3 Protection of the Reactor Building In the event of a major flood, such as those evaluated in this analysis, MNGP plans to build a bin wall barrier that will protect the site from the high flood waters. If the bin wall levee is successful, the safety equipment needed to prevent core damage will be protected, providing defense in depth for each required critical safety function. Simple flood protection measures may be taken to protect the reactor building from floodwater, even if the bin wall levee fails.

These measures to protect the reactor building do not need to be in place until the flood height approaches the 935' elevation. A conservative value of 0.11 is assigned, consistent with the ES-13-096 NRC letter (Reference 5), to the failure probability of protecting the reactor building.

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1SML16012.000-1 Methodology and Analysis 3.1.4 Manual Local Operation of RCIC and the Hard Pipe Vent MNGP requested Hughes Associates to conduct a detailed human reliability analysis (HRA) to estimate human error probabilities for operator manual actions to prevent core damage through the use of RCIC as a high pressure injection source and the Hard Pipe Vent (HPV) to remove decay heat from containment. This detailed HRA was based upon thermal/hydraulics analyses, battery depletion calculations, several Job Performance Measure (JPM) exercises for the specific procedure A8.05.01 and A8.05.01 actions (References 8 and 9) considering flood and Station Blackout (SBO) conditions, and discussions with Operations and PRA staff (References 10 and 11). This section documents the initial conditions assumed, the analytical process, and the results of the detailed HRA. A comparison between the detailed analysis results and those obtained using SPAR-H is also provided as a sensitivity evaluation.

3.1.4.1 Initial Conditions for the Operator Manual Actions Operations staff at MNGP provided the following information on the conditions that would be evolving leading up to the need for the postulated operator manual actions evaluated in this HRA.

The A.6 Procedure, "Acts of Nature" (Reference 12) that addresses External Flooding directs de-energizing the 115 KV, 230 KV and 345 KV substations for flood levels in excess of the 930' elevation. In the case where the flood level is above 930' this would lead to a loss of offsite power and reliance on the EDGs. Since the normal long term fuel oil storage (Tank T-44) for the EDG's is not evaluated to survive flood levels above 932' elevation, and alternate fuel oil makeup methods are not pre-prescribed to support the EDG's, long term EDG operation is not easily defensible utilizing existing procedures. Additionally, there are several penetrations in the Plant Administration Building that would make positive flood proofing of the building difficult, leaving three of the four station batteries vulnerable to flooding.

The flooding engineer provides daily updates to the station on high river water levels including potentials to rise above any trigger points from the A.6 procedure. At this point, heightened awareness of the potential for flooding is implemented.

When river level exceeds 921 feet an evaluation of EALs would be performed. If visible damage has occurred due to flood water rising greater than 921 feet, then an Alert per EAL HA1.6 would be declared.

Prior to the river reaching these levels, operators would be walking down the procedures for alternate methods to vent primary containment and operate RCIC remotely. This would involve staging of equipment in the torus area to open the Hard Pipe Vent and verification that equipment is properly staged to operate RCIC remotely.

As water level reached the 930' elevations, Operations would prepare for isolation of off-site power and loading essential loads onto the emergency diesel generators as needed to conserve fuel. Only a single EDG is required for shutdown cooling and inventory makeup. Operators would be in the EDG rooms, intake and other critical areas ensuring no water intrusion and would be pumping water out of the room as needed. Also the battery rooms in the PAB would be of concern due to the high flood levels. At this point operators would be briefed to be ready to operate RCIC without electrical power as necessary.

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1SML16012.000-1 Methodology and Analysis The portable diesel fire pumps would also be staged at higher locations with hoses staged through higher elevations of the buildings to support alternate RPV makeup.

Operators would be dispatched with I&C technicians to the 962' elevation of the reactor building to install the temporary level indication per procedure A.8-05.01. This would allow for alternate level indication to be available in the event that the batteries are lost.

In the response scenario postulated for the performance of the operator manual actions evaluated in this HRA, EDGs and batteries are not available. Shutdown cooling, HPCI, and RCIC are not available from normal electrical means. RCIC is available for manual operation.

Operators would have temporary level indication set up in the reactor building. Pressure indication is available in the direct area of the level transmitters. The building is dark and most likely there is water in the basement of the reactor building. Additional portable lights are available to assist with lighting and boots staged for higher water. The operators would utilize procedure A.8-05.01 to un-latch the governor from the remote servo linkage and throttle steam flow to RCIC to start the turbine rolling while coordinating with operators monitoring water level and reactor pressure. Upon reaching the high end of the level band the operators would throttle closed the steam admission valve and await direction to re-start RCIC. Local operation of RCIC is demonstrated each refueling outage during the over speed test. Operation of a coupled turbine run is less complex because the turbine is easier to control with a load.

3.1.4.2 Detailed HRA The human error probabilities (HEPs) for manual local operation of RCIC and the Hard Pipe Vent during an extreme flooding and SBO event have been developed in detail utilizing the EPRI HRA Calculator (Reference 13). Event RCICSBOFLOOD (Fail to manually operate RCIC during SBO and extreme flooding conditions), and event HPVSBOFLOOD (Fail to operate the HPV using N2 bottles to provide containment heat removal during SBO/Flood) have values of 9.3E-02 and 1.3 E-02 respectively, for a combined value of 1.06E-01.

The HRA Calculator reports for these events, which provide a detailed basis for these HEP estimates, are presented in Appendix A. Section A. 1 documents RCICSBOFLOOD and Section A.2 documents HPVSBOFLOOD.

The cognitive portion of these HEPs was developed using the Cause Based Decision Tree Method (CBDTM) (Reference 13) and the execution portion utilized the Technique for Human Error Rate Prediction (THERP) (Reference 14). These methods are commonly used in internal events PRA, have been cited in the EPRI/NRC-RES Fire HRA Guidelines, NUREG-1921 (Reference 15) and applied in many fire PRAs, and are also cited in the following reference:

A Preliminary Approach to Human Reliability Analysis for External Events with a Focus on Seismic, EPRI 1025294, EPRI, December 2012.

NUREG-1921 and THERP are also listed as references to the following document:

Interim Staff Guidance for Performing the Integrated Assessment for External Flooding, Appendix C: Evaluation of Manual Actions, JLD-ISG-2012-05, Revision 0, U.S. NRC Japan Lessons-Learned Project Directorate, November 30, 2012.

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ISML16012.000-1 Methodology and Analysis Appendix C of the ISG on External Flooding concentrates primarily on demonstrating the feasibility and reliability of the manual actions consistent with NUREG-1852 (Reference 16),

"Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire."

One of the key variables in determining feasibility of operator manual actions is timing. As stated in the ISG, "For an action to be feasible, the time available must be greater than the time required when using bounding values that account for estimation uncertainty and human performance variability." In order to assess this feasibility, the following process is recommended in section C.3.2.4 Calculate Time Margin:

"The licensee should calculate the time margin available for the action using the values for time available and time required that have been developed for the analysis."

The time margin formula provided is:

Time Margin = [(Tsw-Tdelay)--(Tcog+Texe))

(Tcog.Texe) X 100%

The terms of the equation are defined as follows:

Tdelay = time delay, or the duration of time it takes for the cue to become available that indicates that the action will be necessary (assumes that action will not be taken in the absence of a cue);

Tsw = the time window within which the action must be performed to achieve its objective; Tcog = cognition time, consisting of detection, diagnosis, and decision-making; and Texe = execution time including travel, collection of tools, donning of PPE, and manipulation of relevant equipment.

The HEP calculations performed for the RCIC and HPV manual actions involved the estimation of the time parameters cited in the Time Margin formula. These times are based upon thermal/hydraulics analyses, battery depletion calculations, several Job Performance Measure (JPM) exercises for the specific proceduralized actions and considering flood and SBO conditions, and discussions with Operations and PRA staff.

Using this information, the time margin for these two events is calculated as:

Table 3-5 Time Margin for HFEs Times (in minutes) RCIC HPV Tsw 478.2 900 Tdelay 345 330 Tcog 10 10 Texe 80 45 Time Margin 48% 936%

The HEP timing estimates therefore meet the criteria that "using the calculation under C.3.2.4, the margin must be a positive percent value for an action to be deemed feasible."

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1SML16012.000-1 Methodology and Analysis The External Flooding ISG also recommends that estimates of time available and time required should account for sources of uncertainty and human performance variability. The timing estimates shown above are already believed to be conservative. For example, the Tdelay of 345 min for the RCIC event is based on 5.75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months until RCIC battery depletion. This can be considered conservative since, in reality, it is likely that an action would be taken before waiting for battery depletion. For the Texe values, the highest observed time in the JPM trials was used for the RCIC case and further time was added for transit time for actions in various locations. For the HPV case, the Texe value was also based on JPM trial data and the results above show that uncertainties are covered by a significant time margin.

In addition to feasibility, the reliability of the actions was evaluated through the detailed HRA Calculator analysis used to quantify HEPs.

Section C4 of the ISG says that for an action to be deemed reliable, "sufficient margin should exist between the time available for the action and the time required to complete it. This margin should account for: (1) limitations of the analysis (e.g., failure to identify factors that may delay or complicate performance of the manual action); and (2) the potential for workload, time pressure and stress conditions to create a non-negligible likelihood for errors in task completion... A simplified alternative criterion for determining if the margin is adequate to deem an action as reliable is to establish that the margin is not less than 100%. Such a margin may be justified when recovery from an error in performing the action could be accomplished by restarting the task from the beginning."

As shown above, the time margin for the HPV manual action is greater than 100% and the time margin for the RCIC manual action is estimated at around 50%. The evaluation of PRA and Operations staff in performing the JPMs for these actions was that there would be 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months or more to perform the procedures, allowing several opportunities to troubleshoot and/or re-perform steps if necessary. In addition, based on the HEP quantification, the combined probability for the RCIC and HPV actions is estimated to have a success rate of 89%. A sensitivity study was performed using SPAR-H estimates to evaluate the effect of reliability uncertainty.

Regarding the evaluation of factors such as workload and stress that could complicate task performance, performance Shaping Factors (PSFs) were considered for feasibility and addressed in the assessment of reliability of the operator manual actions.

The following table discusses each of the PSF categories cited in Appendix C of the ISG, the issues related to the RCIC and HPV manual actions and how they were addressed in the qualitative and quantitative analysis of these human failure events.

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1SML16012.000-1 Methodology and Analysis Table 3-6 Performance Shaping Factors for Human Failure Events Performance Shaping Factors Specific Considerations for RCIC and HPV Events under SBO and (PSFs) Flooding Conditions Cues Cues not only were provided by the Emergency Operating Procedure flowcharts for RPV Level and Containment Pressure and supporting procedure, but would be expected to be provided by the Emergency Response Organization staff and STA. It is expected that daily meetings would be held to assess the plant situation considering the long term effects of flooding and SBO and plans would be made to implement the RCIC and HPV actions. The decision to perform these actions would therefore be made by ERO and STA and the Shift Supervisor would give the direction to operations staff.

Indications Due to the Station Blackout, impacts to the normal set of indications were expected and these effects were implemented in the CBDTM module of the HRA Calculator consistent with the methods recommended in NUREG-1921, Appendix B as well as in the Execution PSFs and Stress module. In addition, the "Degree of Clarity of Cues & Indications" was degraded from "Very Good" to "Poor".

Complexity of the Required Action The response is considered to be Complex due to the flooding and SBO impacts to lighting and accessibility. The procedure steps of the actions that are required and that were evaluated in the timed Job Performance Measures (JPMs) specifically performed for these tasks are listed in the Execution Unrecovered module of HRA Calculator.

Special Equipment Flashlights, headlamps and boots were considered necessary by Training when the JPMs were performed for these tasks, and are reflected in the timing estimates for Tm and in the Execution PSFs Special Requirements by indicating that Tools, Parts and Clothing were required and available.

Human-System Interfaces The interfaces with the system will be similar, but with impacts due to the SBO and flooding in terms of indications and lighting. For the RCIC task, a hand-held reactor level monitor is installed and used by I&C technicians to monitor the parameter. These steps of the procedure were specifically included in the Execution portion of the RCIC action quantification. The HPV task involves the use of air cylinders and these steps were also included in the Execution portion of the quantification.

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ISMLI16012.000-1 Methodology and Analysis Table 3-6 Performance Shaping Factors for Human Failure Events Performance Shaping Factors Specific Considerations for RCIC and HPV Events under SBO and (PSFs) Flooding Conditions Procedures Multiple procedures provide for operation of the RCIC System without the availability of AC or DC power. They address the lighting and ventilation limitations that accompany a SBO, and provide for alternate means of monitoring reactor water level/pressure, controlling turbine speed, assuring adequate water source is available and accommodating the condensate from the turbine condenser.

Procedure 8900 (Operation of RCIC without Electric Power)

Procedure C.4-L; Part F (Response to Security Threats; Initiate Injection to the RPV with RCIC)

Procedure A.8-05.01 (Manual Operation of RCIC)

Procedure B.08.09-05.H.4 (Condensate Storage System - Filling Condensate Storage Tanks from Alternate Source)

- Procedure A.8-05.05 (Makeup to CST)

Any of the following procedures provide for operation of the HPV without the need for normal support systems including electric power and pneumatic supplies. All necessary equipment to perform this function is specifically manufactured and pre-staged to allow opening the HPV valves.

- Procedure B.04.01-05.H.2 (Primary Containment System Operation -

Alternate N2 Supply for Operating AO-4539 and AO-4540)

- Procedure C.5-3505; Part A (Venting Primary Containment; Vent Through the Hard Pipe Vent)

- Procedure A.8-05.08 (Manually Open Containment Vent Lines)

The relevant procedures were reviewed, cues and operator action steps were itemized in the quantification and were evaluated using multiple iterations of the timed JPMs.

Training and Experience The RCIC and HPV capabilities are included in various aspects of periodic operations training. Training materials and mockup training devices related to these activities include:

" JPM - B.02.03-005 (Reset RCIC Overspeed Trip)

"Training mockup of the RCIC Turbine Trip Throttle Valve (MO-2080)

"Lesson Plan MT-ILT-EOP-002L (RPV Control)

"Lesson Plan MT-NLO-12C-002L (Emergency Operating Procedures Overview)

Lesson Plan MT-NLO-EOP-001 L (EOPs for NLOs (Turbine Building))

Lesson Plan MT-OPS-FB-004L (Level 4 - Extensive Damage Mitigation Guidelines)

- Lesson Plan M-8107L-003 (RCIC)

Additionally, control of RCIC, using the RCIC MO-2080 Turbine Trip Throttle Valve is actually performed during the startup from each refueling outage, in the performance of procedure 1056 (RCICI Turbine Overspeed Trip Test).

Operator training, procedure adequacy, and equipment readiness regarding use of these mitigation measures has been reviewed by the NRC as part of the B.5.b and Fukushima Flex response inspections and been determined to be acceptable.

ML11235A897 (NRC Fire/B.5.b Inspection Report; 8/23/11)

ML111320400 (NRC Temporary Instruction 2515/183; 5/13/11).

Perceived Workload, Pressure Under the quantification Execution PSFs, Workload has been assessed and Stress as High and PSFs of Negative, so the overall stress is assessed as High.

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ISMLI16012.000-1 Methodology and Analysis Table 3-6 Performance Shaping Factors for Human Failure Events Performance Shaping Factors Specific Considerations for RCIC and HPV Events under SBO and (PSFs) Flooding Conditions Environmental Factors The environment would be consistent with SBO (hot, dark, damp) and some Reactor Building flooding requiring boots. These were the conditions evaluated during the timed JPMs and are addressed in the quantification under Execution PSFs for Lighting, Heat/Humidity and Atmosphere.

Special Fitness Issues No special fitness issues were identified although the performance of multiple trials of the JPMs is considered to address the variability in personnel fitness.

Staffing Operations staffing to perform the procedures would be optimal (several operators assigned as desired to each procedure). Discussions were held to evaluate whether there would be dependencies in staffing between the RCIC and HPV actions and it was considered that due to the different timeframes, that the actions and staffing would be separate.

Communications Under SBO conditions the use of radio and walkie talkies would be expected to allow communications to be maintained between the Main Control Room and the I&C Technicians and the staff performing the key actions.

Accessibility The Equipment Accessibility is evaluated as "With Difficulty" due to reactor building lighting and flooding issues. The quantification Execution PSFs indicates this and these issues were addressed during the JPM timing sessions.

A sensitivity analysis was also performed by developing human error probabilities (HEPs) for manual local operation of RCIC and the Hard Pipe Vent during an extreme flooding and SBO event utilizing the SPAR-H module of the EPRI HRA Calculator and recommended practices from the Idaho National Laboratory step-by-step SPAR-H guidance (Reference 17). Using SPAR-H, event RCICSBOFLOOD (Fail to manually operate RCIC during SBO and extreme flooding conditions), and event HPVSBO FLOOD (Fail to operate the HPV using N2 bottles to provide containment heat removal during SBO/Flood) have values of 1.4E-01 and 5.5E-02 respectively, for a combined value of 1.95E-01.

The HRA calculator reports associated with these HEPs can be found in sections A.3 and A.4 of Appendix A.

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ISML16012.000-1 Conclusions

4.0 CONCLUSION

S The results of the analysis are shown in Table 4-1. This table includes the results of both the best estimate quantification and the sensitivities that were performed in support of this analysis.

The ACDF represents the difference between the best-estimate CDF and the baseline CDF. The baseline CDF is the CDF without the performance deficiency. The best-estimate CDF is the result of the event tree calculations shown in Appendix A. The exposure time represents the period of time the plant was exposed to the performance deficiency. This time period was from February 29, 2012 to February, 15, 2013 resulting in an exposure time of 352 days or 0.964 years. The failure to build construct a levee is assumed to be 0.11 (consistent with the probability assumed by the NRC). Therefore, the baseline CDF is calculated by multiplying the best-estimate CDF by 0.11. ACDF is equal to the best-estimate minus the baseline, thus:

= best-estimate - 0.11 x best-estimate

= (1-0.11) x best-estimate

= 0.89 x best-estimate.

ACDP = exposure x ACDF = 0.964 x 0.89 x best-estimate.

Table 4-1 Results of Event Tree Quantification Sensitivity 1: Sensitivity 2:

NRC Case Nominal, Bounding Frequency Flood SPAR-H HRA Probabilities CD Seq 1 8.41 E-07 1.06E-06 1.55E-06 CD Seq 2 9.15E-08 9.43E-07 1.68E-07 CD Seq 3 1.07E-07 1.10E-06 1.07E-07 CDF 4.20E-05 1.04E-06 3.10E-06 1.83E-06 ACDF 3.60E-05 8.92E-07 2.66E-06 1.57E-06 Significance Yellow Green White White The results of this analysis show that by a best-estimate analysis, the significance of the flooding event is Green. Two sensitivity studies performed to evaluate the effects of sources of uncertainty show that the significance of the flooding could be characterized as low to moderate safety or security significance or 'White' using bounding assumptions.

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I SML16012.000-1 Appendices APPENDICES TABLE OF CONTENTS A. APPENDIX A - HRA CALCULATOR REPORTS ........................................ A-1 A.1. RCICSBOFLOOD, Fail to manually operate RCIC during SBO and extrem e flooding conditions ............................................................... A-1 A.2. HPVSBOFLOOD, Fail to operate the HPV using N2 bottles to provide containment heat removal during SBO/Flood ....................................... A-19 A.3. RCICSBOFLOOD, Fail to manually operate RCIC during SBO and extreme flooding conditions (SPAR-H) ................................................. A-31 A.4. HPVSBOFLOOD, Fail to operate the HPV using N2 bottles to provide containment heat removal during SBO/Flood (SPAR-H) ...................... A-36 B. APPENDIX B - EVENT TREES ................................................................... B-1 Page 16 Revision 22 Page 16

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS A. APPENDIX A - HRA CALCULATOR REPORTS The following sections document the HRA Calculator reports that support this analysis.

A.1. RCIC_SBOFLOOD, Fail to manually operate RCIC during SBO and extreme flooding conditions Basic Event Summary

.%Plant Data File: ':FiledSize' File Dated. IR1ecord. Date Monticello Ext 913408 07/02/13 07/02/13 Flooding SDP HRAJune 2013.HRA

':Nam ;-Date Analyst Erin P. Collins, Hughes 07/02/2013 Associates Reviewer John Spaargaren & Pierre 07/02/2013 Macheret, Hug hes Associates Table 41: RCIC_SBOFLOOD

SUMMARY

______________ ~HEP,-Sumay __

Pcog Pexe Total HEP Error Factor Method CBDTM THERP CBDTM + THERP Without Recovery 2.9e-02 5.2e-01 With Recovery 9.2e-04 9.2e-02 9.3e-02 5 Initial Cue:

RPV Water Level Below 9 in.

Recovery Cue:

Before RPV level drops to -149 inches Cue Comments:

The cue for action is that the TSC and the Emergency Director have determined that RCIC operation is needed. The Control Room Supervisor (CRS) directs operator to initiate RCIC and inject into the RPV using procedure A.8-05.01, Manual Operation of RCIC, Part A, Placing RCIC in Service.

Due to the SBO, it is assumed that there will be multiple impacts to indications, so the degree of clarity has been set at "Poor".

Deqree of Clarity of Cues & Indications:

Poor Procedures:

Cognitive: C.5-1 100 (RPV CONTROL flowchart (Monticello)) Revision: 11 Execution: A.8-05.01 (Manual Operation of RCIC) Revision: 2 Other: A.6 (ACTS OF NATURE (Monticello)) Revision: 43 Other: 8900 (OPERATION OF RCIC WITHOUT ELECTRIC POWER (Monticello)) Revision: 2 Page A-I Revision 22 Page A-1

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Cognitive Procedure:

Step: LEVEL Instruction: Restore and maintain RPV water level 9 to 48 in. using Preferred Injection Systems Procedure and Training Notes:

Three JPM trials were performed emulating the specific external flooding conditions of this scenario on 18 June 2013. Observations were factored into this analysis.

Procedure A.8-05.01 - 2.0 ENTRY CONDITIONS This strategy is entered when one or more of the following conditions are met:

o As directed from A.8-01.01 (Extensive Damage Mitigation Strategy Overview) o As directed from A.8-03.01 (Initial Response Actions) o As directed from 5790-110-01 (Monticello Emergency Management Guideline) - presume this is the relevant case since ERF is staffed and will make the call on implementing the procedure Training:

Classroom, Frequency: 0.5 per year Simulator, Frequency: 0.5 per year JPM Procedure:

JPM-A.8-05-01-001 (Manual Operation of RCIC) Revision: 0 Identification and Definition:

1. Letter to Region III SRA write-up, 8 April 2013 Section G - HEP Associated with Protecting Plant Buildings from 930' Elevation Flood For the case where the site is not protected by a ring levee, but individual buildings / equipment are protected by flood barriers as called out in the A.6 (Acts of Nature) procedure, several redundant options remain available for protecting critical safety functions related to injecting water to the reactor, maintaining desired reactor pressure, and removing decay heat from containment.

The A.6 Procedure [Acts of Nature that addresses External Flooding] directs de-energizing the 115 KV, 230 KV and 345 KV substations for flood levels in excess of the 930' elevation. In the case where the flood level is above 930' this would lead to a loss of offsite power and reliance on the EDGs. Since the normal long term fuel oil storage (Tank T-44) for the EDG's is not evaluated to survive flood levels above 932' elevation, and alternate fuel oil makeup methods are not pre-prescribed to support the EDG's, long term EDG operation is not easily defensible. Additionally, there are several penetrations in the Plant Administration Building that would make positive flood proofing of the building difficult, leaving three of the four station batteries vulnerable to flooding.

2. PRA-MT-SY-RCIC, Reactor Core Isolation Cooling System Notebook, Revision 3.0, December 2012 Table 3 - IE_LOOP (Loss Of Offsite Power Initiating Event) Impact on RCIC System: A loss-of-offsite power does not affect RCIC, provided AC power remains available to the Division 1 battery chargers. A loss of Feedwater would result, and RCIC operation would be automatically actuated when Low-Low Level in the RPV is reached.

Key Assumptions:

It is assumed that there will be sufficient water and fuel supply for the equipment needed in this scenario despite the flooding conditions and considering the long term nature of the flood, which may take as many as 12 days to recede (from the Monticello Design Basis documentation).

This means that CST level is assumed to be maintained via a step in this HEP and is not quantified separately.

Another key assumption is that because the staffing needed for this event is separate from that used for the Hard Pipe Vent human failure event, and the timeframe for HPV is much longer, no dependency between these events was evaluated (in other words, the timing and staffing were considered as totally separate events).

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1SMIL160112.000-11 Appendix A - HRA CALCULATOR REPORTS ISM LI 6012.000-1 Appendix A HRA CALCULATOR REPORTS JPM-A.8-05.01-001 (Manual Operation of RCIC) Rev. 0 INITIAL CONDITIONS:

o Extreme flooding has led to a Station Blackout that has existed at Monticello for the last 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

o Div. 1 250 VDC battery system has been depleted and is not available.

o The plant was in Shutdown Cooling until the station blackout and has since been slowly repressurizing due to heating up.

o Current RPV pressure is 75 psig-and slowly rising.

o Current RPV water level is -40" and very slowly lowering.

o The TSC and the Emergency Director have determined that RCIC operation is needed.

o HPCI is inoperable.

o RCIC suction is from the CSTs.

o RCIC operation is required to maintain RPV level above TAF.

o Approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ago Radiation Protection reported -2" of water on the Rx Bldg basement floor.

o A second operator will be performing Part B, Set Up and Monitor of Rx Vessel with Fluke 707.

o A third operator will be maintaining CST level using A.8-05.05, Makeup to the CST, and verifying valve status in the steam chase.

INITIATING CUES (IF APPLICABLE):

o The CRS directs you to initiate RCIC and inject into the RPV using A.8-05.01, Manual Operation of RCIC, Part A, Placing RCIC in Service.

o Inform the CRS when RCIC is in service with discharge pressure at least 76 psig greater than reactor pressure..

RCIC local manual operation Job Performance Measure entry condition assumptions, Documented in and excerpted from Hughes Associates Record of Correspondence, RCIC Manual Operation e-mails with Xcel Energy during June - July 2013, Hughes Associates, Baltimore, MD, 7 July 2013:

Conditions anticipated following a SBO resulting from an external flooding event:

o ERO has been manned for the past several days, with these procedures predicted and planned to be implemented ahead of time o Plant is in cold shutdown condition (mode 4) o I&C would perform the reactor level monitoring portion of the RCIC procedure o Operations staffing to perform the procedures would be optimal (several operators assigned as desired to each procedure) o Environment would be consistent with SBO (hot, dark, damp) o There would be 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months to perform the procedures, allowing several opportunities to troubleshoot and/or re-perform steps if necessary o The ERO would place maximum priority on maximizing chances of successful performance of these procedures Operator Interview Insights:

Documented in and excerpted from Hughes Associates Record of Correspondence, RCIC Manual Operation e-mails with Xcel Energy during June - July 2013, Hughes Associates, Baltimore, MD, 7 July 2013:

From Xcel Operations: The series of events would progress during a flooding event such that the need for the alternate instrumentation would be known before the flood completely resulted in a station blackout with loss of DC. The alternate instrumentation (Fluke) would be connected to the selected locations. If this instrumentation did not agree with the actual permanent instrumentation (i.e. prior to failure),

assistance would be obtained to figure out the discrepancy. It is a relatively simple solution to get the temporary instrumentation to work by either pulling a fuse or lifting a lead. This action would be required Revision 2 Page A-3

I SMLI16012,000-1 Appendix A - HRA CALCULATOR REPORTS for the control room, cable spreading room and EFT locations (again not proceduralized). The other option would be to use the transmitter in the reactor building to get the readings which does not require the lifted lead or fuse pulled.

The other issue that is of concern is the need to density compensate the fluke readings to get an actual level. The indicated level can be drastically different from the actual level depending on the calibration conditions of the instrument (hot or cold calibration conditions) and the actual pressure/temperature at the time of the reading. None of this information is contained in the A.8 procedures. There are density compensation tables in the B.1.1 operations manual figures section six and they are also posted in the control room. It would take additional action for the on-shift team/technical staff to put this all together to determine what actual level was from the readings that came off the fluke.

Ma*nnnwer el FD "nniramanta CrewMe*mbern

=~

!.n cluded.

Total Available Reuired Execuitionfor PNotes

_________________i Reactor operators Yes 2 1 Plant operators Yes 2 0 Mechanics Yes 2 0 Electricians Yes 2 0 I&C Technicians Yes 2 0 Health Physics Technicians Yes 2 0 Chemistry Technicians Yes 1 0 Execution Performance Shapina Factors:

Environment: Lighting Portable Heat/Humidity Hot / Humid Radiation Background Atmosphere Steam (although steam will not be present, this PSF was used to indicate an off-nominal condition, such as would be present for flood and SBO)

Special Requirements: Tools Required Adequate Available Parts Required Adequate Clothing Required Adequate Complexity of Response: Cognitive Complex Execution Complex Equipment Accessibility Main Control Room Accessible (Cognitive):

Equipment Accessibility Reactor Building With Difficulty (Execution):

Stress: High Plant Response As Expected: Yes Workload: High I Performance Shaping Factors: Negative Revision 2 Page A-4

ISML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Performance Shaping Factor Notes:

The response is considered to be Complex due to the flooding and SBO impacts to lighting and accessibility. Flashlights, headlamps and boots were considered necessary by Training when the JPMs were performed for these tasks.

The Equipment Accessibility is evaluated as With Difficulty due to Rx building lighting and flooding issues.

Despite preparations and training, the flooding scenario is considered to be a high stress situation.

Key Assumptions (see that section) regarding the conditions provided to Training for performing the JPM for this task said that the "Environment would be consistent with SBO (hot, dark, damp)". The Training insights from the JPM performance (see Operator Interview Insights) stated that the operators recommended to "Stage additional flashlights and headlamps in RCIC room", so it is clear that portable lighting is used.

Page A-5 Revision 22 Page A-5

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Timing:

TS 7.97 Hours Tdly5.75 Hours T/ 10.00 Minutes TM 80.00 Minutes Irreversible Cue DamageState t=o Timingi Analysis:

TO = Station Blackout Tsw = Time from Station Blackout to the time by which RCIC must be restored.

Per Monticello MAAP Calculations, case "SBOCase3-R1", 27 June 2013:

Time to TAF = 7.17 hrs Time to -149" = 7.2 hrs Time to 1800 F = 7.97 hrs Damage is assumed to occur if the temperature exceeds 1800 F or 7.97 hrs, so this was used for Tsw as the time by which RCIC restoration is required.

Tdelay = PRA battery calc (PRA-CALC-1 1-002) indicates that there are 5.75 hrs until RCIC battery depletion. Also, the RCIC Water Flow (column BC) of the d41 tabs in the "SBOCase3-R1" MAAP analysis spreadsheet shows that RCIC injection stops at approximately the same time (5.74 hrs), so the MAAP runs agree with the calc. This Tdelay can be considered somewhat conservative, since in reality, it is likely that an action would be taken before waiting for battery depletion.

T1/2 = The cue for action is that the TSC and the Emergency Response Director have determined that RCIC operation is needed. Daily planning meetings will have been held to discuss actions to be taken as soon as the diesels are lost, so the 10 minutes is simply an estimate of the meeting time between TSC and ERF personnel to make the actual decision to manually operate RCIC. The Control Room Supervisor (CRS) directs operator to initiate RCIC and inject into the RPV using procedure A.8-05.01, Manual Operation of RCIC, Part A, Placing RCIC in Service.

Tm = Results of RCIC local manual operation Job Performance Measure performed 18 June 2013. The procedure was performed three times, taking 49 minutes, 37 minutes and 50 minutes to complete for an average time of 45 minutes. 50 minutes was used as the conservative value for JPM performance.

The JPM did not include the performance of Part B for installation and use of the Fluke level monitoring device; this was estimated to require 30 minutes, so the total time for Tm was estimated as 50 min + 30 min = 80 min.

Time available for cognition and recovery: 53.20 Minutes Time available for recovery: 43.20 Minutes SPAR-H Available time (cognitive): 53.20 Minutes Page A-6 Revision 22 Page A-6

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS I SMLI 6012.000-1 Appendix A - HRA CALCULATOR REPORTS SPAR-H Available time (execution) ratio: 1.54 Minimum level of dependence for recovery: LD Page A-7 Revision 22 Revision Page A-7

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS I SMII 6012.000-1 Appendix A HRA CALCULATOR REPORTS Cognitive Unrecovered RCICSBOFLOOD Table 42: RCICSBOFLOOD COGNITIVE UNRECOVERED Pc Failure Mechanism;. Birainch 'HEP Pca: Availability of Information d 1.5e-03 PCb: Failure of Attention m 1.5e-02 Pcc: Misread/miscommunicate data e 3.0e-03 Pcd: Information misleading b 3.0e-03 Pce: Skip a step in procedure 9 6.0e-03 Pcf: Misinterpret instruction a neg.

Pcg: Misinterpret decision logic k neg.

PCh: Deliberate violation a neg.

Sum of Pca through PCh = Initial Pc = 2.9e-02 Notes:

Normal RPV water level indication is not available and must be monitored with a hand held device installed by I&C. This is clearly proceduralized in Part B of A.8-05.01.

Presumed that SBO causes issues with normal alarms and indications so pc-a through -d were adjusted consistent with insights from EPRI 1025294, A Preliminary Approach to Human Reliability Analysis for External Events with a Focus on Seismic, October 2012.

Page A-8 Revision 22 Page A-8

I SM L16012.000-11 Appendix A - HRA CALCULATOR REPORTS ISM LI 6012.000-1 Appendix A - NRA CALCULATOR REPORTS pca: Avalabilty of infonnation u Aal in

,cation CR Inicalion WaIingAlternate 'l'kaining on CR Accurate W Procedure Infelators (a) neg.

(b) neg.

(c) neg.

3.-e-03 - (d) 1.5e-03 11.0e+00 Yes 1.0e-01 (e) 5.0e-02 1.0e+00 (f) 5.0e-01 1.0e400 (g) I.0e+O0 1LOO1M MCR indications may not be accurate due to the Station Blackout, however, either procedural or informal crew information on alternate indications and training should provide operator input to decision-making.

pcb: Falum of attention Low vs. M Check vs. Monitor Front vs. Back Alaued vmNot Workload Pan el Alnned ek O(a) IO neg.

.e*i Back I 0(b) .5e-04 3.0e-0 (c) 3.0e-03 1.8e+.00 Front 15.0e-02 (d) I.5e-04 io.00MoI (e) 3.0e-03 3e-03 Back 15.0e-02 (f) 3.0e-04

1. Cb ie 3.0e-03 (g) 6.Oe-03

2. Cb ice Front 15.0e-02 ((h)neg.

Check O.e00 e4 (i)neg.

o.oeo Back 5.0e,.2 (i) 7.Se-04 lh 3.0e-03 (k) 15e-02 Front -(I)7.e-04 Monito r O.O+- WL -- (m) I.5e-02 3.e35 I50e-02 (n) 1.5e-03 3.0e-3 I (o) 3.Oe-02 1.0e4.0 Page 9 Revision 22 Page 9

ISML16012.000-1 Appendix A - HRA CALCULATOR REPORTS ISM LI 6012.000-1 Appendix A HRA CALCULATOR REPORTS pcc nicate data Ihucatom Easy to Good&Bad Inicator Foanal LocateI CoImunKicaUons S O.e 4.0 0 (a) neg.

S3.0e-033.e-03 0.Oe+00 I (c) I.Oe-03 YeIot.0e-0 A(d) 4.0e-03 3.0e-03 No - - -o(e) 3.0e03 o.oe*Coo (6e3 3F)03 6.e-03 L3.OeO3 3Oe-03 (g) 4.0e-03 1.0e-03 3.Oe-03(h) ( 7.Oe-03 3.0e-03 For this scenario, normal reactor water level indication is not available and a hand-held level monitor will be jumpered in. Although procedural guidance on reading the monitor is clear, "not Easy" is selected to reflect the additional challenges to the task posed by this alternative source for level indication.

pcd: Infoonation mismeming M Cues as Stated mring of Specicr ,Rang I General Ta-anug I ~ ~Ifferences II o.oesoo (a) neg.

No---- -- -- ----- --- --- ----- --- ----- --- (b) 3.e-03 1.0e+O0 (c) 1.0e-02 1.0e-02 MCR indications may not be accurate due to the Station Blackout so cues may not be as stated in procedures.

pce: Sip a step in procedure Obqgiousv& Singl vs. NfUN*l Graphicaly Placekeqing Ad (a) 3.0e-03 3.3"1 (b) 3.0e-03 o.Oe+O 13.0e-03 ()e-02 (c) 3.0e-03 130e0eO0I (e) 2.0e-03

(h) 1.3e-02 (g) .oe-o3 1.0e-02 1.0e-4D 1 01) t.Oe -01 Page 10 Revision 22 Page 10

ISML16012.000-1 Appendix A - HRA CALCULATOR REPORTS ISM LI 6012.000-1 Appendix A HRA CALCULATOR REPORTS pd' Misinterpret instruction Standard or AN Required Training on Step Ambiguous wording Information I

--- ------ ------- ------- ------- (a)

(b) neg.

3.0e-03 Yes 3.0e-02 1.0e4.00 (:c)3.0e-02 NO1.0e,4 (d) 3.0e-03 0.0e+00 (e) 3.0e-02 3.0e-02 1(f) 6.0e-03 3.0e-02 (g) 6.0e-02 pcg: Misinterprt decision ogic "NOr Statement 'ANUD or 'Or Both "AND" & practiced Scenario Statement "OW 3-le0M (a) 1.6e-02 3.0e.02 I(b) 4.9e-02 1.2e-02 (c) 6.0e-03 0o.oe+oo l(d) 1.9-02 (e) 2.0e-03 O.Oe.+O0 (t) 6.0e-03 Yes3~e-. (g) 1.0e-02 NO 3.0e..02 -(h)

.0e4.00 3.1e-02 1.0e-03 (I) 3.0e-04 0.0e+00 O.OeO

--- Fi) l .0e44) 1.0e-03 3.----- (k) neg.

O.Oe.F (1)neg-pch: Delbe ate Wiolation BEleinAdequacy Advere Reasonable Polcy of of Instruction Consequenceff Alternatives "Vembatin"

-.---- ------- - -- - ------- ------- --------- - - (a)neg.

Ye15.0e- (b) 5.oe-o1 NO.e00 I.OeO (c) I.Oe-Oo 1.0e+00 I O.Oe+O0 (d)neg.

0.0e44)0 (e) neg.

Page 11 Revision 22 Page11

1SML-16012.000-1 Appendix A - HRA CALCULATOR REPORTS I SMLI 6012.000-1 Appendix A HRA CALCULATOR REPORTS Cognitive Recovery RCICSBOFLOOD Table 43: RCICSBOFLOOD COGNITIVE RECOVERY Initial HEP ',

U- , . a_ E - Final Final W

O~t*LU COOr "(LU > S0) Value S o n- -

ý.'PCa: 1.5e-03 X - 5.0e-01 7.5e-04 Pcb: 1.5e-02 X - X X MD 3.8e-03 5.7e-05 Pcc:: 3.0e-03 - - X X - 1.0e-02 3.0e-05 Pcd! 3.0e-03 - X X X - 5.0e-03 1.5e-05 Pce: 6.0e-03 X X X MD 1. le-02 6.6e-05 Pof, Pcf* nell. - 1.0e+00 Pc-:. neg.- - 1.0e+00 Pch: : ne . - 1.0e+00 Sumof P*,athrbugh Pch = Inita Pc-= 9.2e-04 Notes:

Due to long timeframe and severity of scenario, STA and Emergency Response Facility will be available.

Operations staffing to perform the procedures was assessed by Xcel as optimal (several operators assigned as desired to each procedure) so Extra Crew was credited.

Used Moderate Dependency due to high stress.

Page 12 Revision 22 Revision Page 12

ISML16012.000-1 Appendix A - HRA CALCULATOR REPORTS ISM LI 6012.000-1 Appendix A HRA CALCULATOR REPORTS Execution Unrecovered RCICSBOFLOOD Table 44: RCICSBOFLOOD EXECUTION UNRECOVERED Procedure: A.8-05.01, Manual Operation of RCIC . Comment. Stress- Over Ride%

,.Factor Step No. instrUction/Comment Error .. THERP - HEP.

.. . .. .* . . ... YPe.. 1.Table .. I .*Item

__._ *. _.

In RCIC Room, verify open valves MO-2106 and MO-2096 Depress declutch lever and turn handwheel.

EOM 20-7b 2 1.3e-03 5 A.8-05.01, Step 8 Location: Reactor Building EOC 20-13 1 1.3E-3 Total Step HEP 1.3e-02 Remove the pin securing the slip link to the governor lever to prevent interference from the hydraulic governor (See Attachment 1) 5 A.8-05.01, Step 9 Location: Reactor Building EOM 20-7b 2 1.3e-03 EOC 99 1 1.OE-2 Total Step HEP 5.7e-02 Uncap and throttle open RCIC-27 condenser cooling water starting YS A.8-05.01, Step 20 6082 drain. Close when steam is present. 5

-22 Location: Reactor Building EOM 20-7b 2 1.5e-03_5

-_22_EOC 20-12 5 1.3E-3 Total Step HEP 1.3e-02 Throttle MO-2080 and control as necessary A.8-05.01 Steps Location: Reactor Building EOM 20-7b 2 1.3e-03 5 17,18,25,26 EOC 20-12 5 1.3E-3 Total Step HEP 1.3e-02 At D31, Access Control - SET UP for Monitoring RX Vessel level by A.8-05-01, Step 2, opening circuits and kick out to Part B M EOM 20-7b 3 1.3e-03_5 at D31 Location: Reactor Building atD31_EOC 20-12 3 1.3E-3 Total Step HEP 1.3e-02 At D100, 1st Floor EFT - SET UP for Monitoring RX Vessel level by A.8-05-01, At Step 2, ST opening Location:circuits EFT and kick out to Part B EOM 20-7b 3 1.3e-03 At EFT EOC 20-12 3 1.3E-3 Total Step HEP 1.3e-02 SET UP for Monitoring RX Vessel level by selecting level instrument and PART B SET UP AND MONITOR OF RX VESSEL attaching Fluke 707 monitoring device - Control Room WITH FLUKE 707 Part B, Steps 29 & 29. SELECT a level instrument from list below, (preference should be given to accessibility and 5 30P-MCR environmental conditions such as radiation, temperature, lighting, etc.):

In Control Room; Revision 2 Page A-1 3

ISML16012.000-1 Appendix A - HRA CALCULATOR REPORTS SýProcedure:'A.8.45.01; Manual Operationof.RCIC Comment Stress. Over Ride Factor Step No. i. . " . .i .. : . Instruction/Commetn:t-:

.... . . -" ' i , " : i Errore*i :+ Tabley. THERP

.iT* ; I;.

, tem . .. HEP .I. .

LI-2-3-85A, Reactor Vessel Water Level, Panel C-03, Term Strip TT-62 to TT-61.

LI-2-3-85B, Reactor Vessel Water Level, Panel C-03, Term Strip PP-70 to PP-71.

LI-2-3-91A, Fuel Zone, Panel C-03, Term Strip TT-59 to TT-58.30.

IfLevel Indicator selected for use, Then ATTACH one lead from Fluke 707 to each terminal point listed.

Location: Main Control Room EOM 20-7b 3 1.3e-03 EOC 20-12 13 1.3E-2 Total Step HEP 7.2e-02 SET UP for Monitoring RX Vessel level by selecting level instrument and PART B SET UP AND MONITOR OF RX VESSEL attaching Fluke 707 monitoring device - Cable Spreading Room WITH FLUKE 707

29. SELECT a level instrument from list below, (preference should be given to accessibility and environmental conditions such as radiation, temperature, lighting, etc.):

In Cable Spreading Room; Part B, Steps 29.& LI-2-3-91A, Fuel Zone, Panel C-18, Term Strip BB- 5 30 - CSR 57 to BB-59.

IfLevel Indicator selected for use, Then ATTACH one lead from Fluke 707 to each terminal point listed.

Location: Cable Spreading Room EOM 20-7b 3 1.3e-03 EOC 20-12 13 1.3E-2 Total Step HEP 7.2e-02 SET UP for Monitoring RX Vessel level by selecting level instrument and PART B SET UP AND MONITOR OF RX VESSEL attaching Fluke 707 monitoring device - EFT WITH FLUKE 707

29. SELECT a level instrument from list below, (preference should be given to accessibility and environmental conditions such as radiation, temperature, lighting, etc.):

At 3rd Floor EFT, ASDS Panel; LI-2-3-86, Reactor Flooding Level, Panel C-292, Part B, Steps 29 & Term Strip (EFT 3) HH-4 to HH-5. 5 30 - EFT LI-2-3-91 B, Fuel Zone, Panel C-292, Term Strip (EFT 3) HH-1 to HH-2.

IfLevel Indicator selected for use, Then ATTACH one lead from Fluke 707 to each terminal point listed.

Location: EFT EOM 20-7b 3 1.3e-03 EOC 20-12 13 1.3E-2 I Revision 2 Page A-14

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS I SMLI 6012.000-1 Appendix A HRA CALCULATOR REPORTS Procedure: A.8-05.01, ManUal Operation of R(

Step No. instruction/Cot SET UP for Monitoring RX Vessel level by selecting level instrument and PART B SET UP AND MONITOR OF RX VESSEL attaching Fluke 707 monitoring device - Reactor Building WITH FLUKE 707

29. SELECT a level instrument from list below, (preference should be given to accessibility and environmental conditions such as radiation, temperature, lighting, etc.):

At 962', Rx Bldg; LT-2-3-85A, Reactor Vessel Water Level, Panel C-56.

LT-2-3-85B, Reactor Vessel Water Level, Panel C-55.

LT-2-3-61, Reactor Flooding Level, Panel C-55.

At 935', Rx Bldg; LT-2-3-112A, Fuel Zone, Rx Bldg 935' West, Panel Part B, Steps 29 & C-122. 5 31 - RxB LT-2-3-112B, Fuel Zone, Rx Bldg 935' East, Panel C-121.

31. If level transmitter selected for use, Then PERFORM the following:

a. REMOVE the cover (see Attachment 5 & 6).

b. LIFT positive lead from its conductor (see Attachment 7).

c. Slightly ENGAGE screw threads into transmitter.

d. CLAMP Fluke 707 leads in series with lifted positive wire and positive terminal point on transmitter.

Location: Reactor Building EOM 20-7b 3 1.3e-03 EOC 1 20-12 1 13 1 1.3E-2 t --

Total Step HEP 7.2e-02 MONITOR OF RX VESSEL WITH FLUKE 707 (Conducted separately by DETERMINE Rx water level by performing the I&C) following: (see Attachment 8 - diagram of FLUKE 707 CALIBRATOR pointing out buttons and displays)

a. PRESS green button to START Fluke 707.

b. PRESS MODE button until display reads MEASURE mA and Loop Power.

c. USE Attachment 9 to obtain vessel level. [read Part B, Step 32 table of RCIC - mA VS. RPV WATER LEVEL]

33. If connected to LT-2-3-61 or LI-2-3-86, Then OPERATE RCIC turbine to maintain steady indication as close to 8 mA as possible.

34. If NOT connected to LT-2-3-61 or LI-2-3-86, Then OPERATE RCIC turbine to maintain steady indication as close to 20 mA as possible.

Page A-15 Revision 22 Page A-1 5

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Procedure:-A.8-05.01, Manual Operation of RCIC Comment-w 'Stress

. Over Ride, Factor Step INo. Instrudtion/comnment. Error - ~ THERP :F HEP

. Typeý. Table,. Item Location: Reactor Building EOM 20-7b 3 1.3e-03 EOC 20-11 1 1.3E-3 Total Step HEP 1.3e-02 Interpret Fluke monitor readings using density compensation tables in E-mail from Xcel Operations to B.1.1 operations manual, section 6 (posted in Control Room) Xcel PRA Manager, 2 July 2013 The indicated level can be drastically different from the actual level depending on the calibration conditions of the instrument (hot or cold calibration conditions) and the actual pressure/temperature at the time of the reading. None of this information is contained in the A.8 procedures. There are density 5 Part B, Step 32c compensation tables in the B.1.1 operations manual figures section six and they are also posted in the control room. It would take additional action for the on-shift team/technical staff to put this all together to determine what actual level was from the readings that came off the fluke.

Location: Main Control Room EOM I 20-7b 3 1.3e-03 EOC 120-10 10 1.3E-2 Total Step HEP 7.2e-02 Maintain CST level and venfy valve status in the steam chase I Location: Reactor Building EOM 20-7b 4 4.3e-03 5 EOC 20-12 5 1.3E-3 Total Step HEP 2.8e-02 Feedback from I&C Level Monitoring i 5 Recovery 1 Location: Reactor Building EOM 99 1 1.0e-02 Total Step HEP 5.0e-02 Feedback from Control Room 1 5 Recovery 2 Location: Main Control Room EOM 20-7b .3 1 1.3e-03 Total Step HEP 6.5e-03 Revision 2 Page A-1 6

1SML-16012,000-1 Appendix A - HRA CALCULATOR REPORTS I SMLI 6012.000-1 Appendix A HRA CALCULATOR REPORTS Execution Recovery RCICSBOFLOOD Table 4-5: RCICSBOFLOOD EXECUTION RECOVERY Action HEP (Crit) HEP (Rec). Dep. Cond. HEP Total for Critical Step No. Recovery Step No. De. (Recl Stan, A-8-05.01, Step 8 In RCIC Room, verify open valves MO-2106 and MO-2096 1.3e-02 7.3e-04 Recovery 2 Feedback from Control Room 6.5e-03 LD 5.6e-02 A-8-05.01, Step 20 - Uncap and throttle open RCIC-27 condenser cooling water 1.3e-02 7.3e-04 22 starting YS 6082 drain. Close when steam is present.

Recovery 2 Feedback from Control Room 6.5e-03 LD 5.6e-02 A.8-05.01 Steps 17, Throttle MO-2080 and control as necessary 18,25,26 13e02 1.4e04 Recovery 1 Feedback from I&C Level Monitoring 5.0e-02 MD 1.9e-01 Recovery 2 Feedback from Control Room 6.5e-03 LD 5.6e-02 A.8-05-01, Step 2, At D31, Access Control - SET UP for Monitoring RX Vessel 1.3e-02 7.3e-04 at D31 level by opening circuits and kick out to Part B Recovery 2 Feedback from Control Room 6.5e-03 LD 5.6e-02 A.8-05-01, Step 2, At D100, 1st Floor EFT - SET UP for Monitoring RX Vessel level 1.3e-02 7.3e-04 At EFT by opening circuits and kick out to Part B Recovery 2 Feedback from Control Room 6.5e-03 LD 5.6e-02 Part B, Steps 29 & SET UP for Monitoring RX Vessel level by selecting level 30 - MCR instrument and attaching Fluke 707 monitoring device - 7.2e-02 1.1e-02 Control Room Recovery 2 Feedback from Control Room 6.5e-03 MD 1.5e-01 Part B, Steps 29 & SET UP for Monitoring RX Vessel level by selecting level 30 - CSR Instrument and attaching Fluke 707 monitoring device - Cable 7.2e-02 4.0e-03 Spreading Room__

Recovery 2 Feedback from Control Room 6.5e-03 LD 5.6e-02 Part B, Steps 29 & SET UP for Monitoring RX Vessel level by selecting level 7.2e-02 4.0e-03 30 - EFT instrument and attaching Fluke 707 monitoring device - EFT Recovery 2 Feedback from Control Room 6.5e-03 LD 5.6e-02 Part B, Steps 29 & SET UP for Monitoring RX Vessel level by selecting level 31 - RxB Instrument and attaching Fluke 707 monitoring device - 7.2e-02 4.0e-03 Reactor Building Recovery 2 Feedback from Control Room 6.5e-03 LD 5.6e-02 Part B, Step 32 MONITOR OF RX VESSEL WITH FLUKE 707 (Conducted 1.3e-02 7.3e-04 separately by I&C)

Recovery 2 Feedback from Control Room 6.5e-03 LD 5.6e-02 Part B, Step 32c Interpret Fluke monitor readings using density compensation tables In B.1.1 operations manual, section 6 (posted In Control 7.2e-02 7.0e-03 I Room)

Recovery I Feedback from I&C Level Monitoring 5.0e-02 LD 9.8e-02 Page A-17 Revision 22 Page A-17

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS I SMLI 6012.000-1 Appendix A HRA CALCULATOR REPORTS Critical St ep No. Recovery Step No. .ý.Action. HEP (Crit) HEP (Rec) Dep. Cond. HEP :

(Rec) sTotapfor

step A.8-05.05 Maintain CST level and verify valve status in the steam chase 2.8e-02 1.6e-03 Recovery 2 Feedback from Control Room 6.5e-03 LD 5.6e-02 A.8-05.01, Step 9 Remove the pin securing the slip link to the governor lever to prevent interference from the hydraulic governor (See 5.7e-02 5.7e-02 Attachment 1)

Total Unrecovered:. 5.2e-01 Total Recovered: 9.2e-02 Page A-18 Revision 2 Page A-1 8

ISML16012.000-1 Appendix A - HRA CALCULATOR REPORTS A.2. HPV_SBO_FLOOD, Fail to operate the HPV using N2 bottles to provide containment heat removal during SBO/Flood Basic Event Summary Plant DataFile ... File;Size. File Datei ."Re&cordDate Monticello Ext 901120 06/28/13 06/28/13 Flooding SDP HRA June 2013.HRA John Spaargaren & Pierre Macheret, Hughes Associates Table 46: HPVSBOFLOOD

SUMMARY

_*__"___._."_ __.___ '._____HEP Sumrnhiaty :. ..: .*. . . , . -. -.. ..

Pcog Pexe Total HEP Error Factor Method CBDTM THERP CBDTM + THERP Without Recovery 3.le-02 2.3e-01 With Recovery 6.1e-04 1.3e-02 1.3e-02 5 Initial Cue:

Drywell pressure above 2 psig Cue Comments:

The cue for action is that the TSC has recommended venting the DW by using the Hard Pipe Vent using procedure A.8-05.08, Manually Open Containment Vent Lines.

Initial procedure entered on high drywell pressure is C.5-1200. The DW/Torus Pressure leg directs operators to C.5-3505. For the limiting PRA case, it is assumed that normal and alternate nitrogen and power via Y-80 is not available and operators must therefore use A.8-05.08 to install pre-staged nitrogen bottles directly to the inboard and outboard HPV isolation valves to open them.

Due to the SBO, it is assumed that there will be multiple impacts to indications, so the degree of clarity has been set at "Poor".

Degree of Clarity of Cues & Indications:

Poor Procedures:

Cognitive: C.5-1200 (PRIMARY CONTAINMENT CONTROL flowchart (Monticello)) Revision: 16 Execution: A.8-05.08 (Manually Open Containment Vent Lines) Revision: 1 Other: A.6 (ACTS OF NATURE (Monticello)) Revision: 43 Other: C.5-3505-A 0 Revision: 10 Cognitive Procedure:

Step: DW/TORUS PRESSURE Page A-19 Revision 22 Page A-1 9

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Instruction: BEFORE DW pressure reaches Fig. D, DW Pressure Limit (56 psig) vent to stay below Fig. D, DW Pressure Limit per C.5-3505 Procedure and Training Notes:

Three JPM trials were performed emulating the specific external flooding conditions of this scenario on 18 June 2013. Observations were factored into this analysis.

Training:

Classroom, Frequency: 0.5 per year Simulator, Frequency: 0.5 per year JPM Procedure:

JPM-A.8-05.08-001 (Manually Open Containment Vent Lines) Revision: 0 Identification and Definition:

This HFE is for the external flooding model for venting prior to core damage.

1. Initial Conditions: SBO due to external flooding.

2. Initiating Events: External flooding causes station blackout.

3. Accident sequence (preceding functional failures and successes):

No containment venting or heat removal Need to vent to maintain containment integrity prior to ultimate containment pressure for RCIC injection

4. Preceding operator error or success in sequence: None.

5. Operator action success criterion: Align pre-staged alternate nitrogen bottles directly to AO-4539 and AO-4540 (located on the torus catwalk) to open the hard pipe vent.

6: Consequence of failure: Failure to vent containment leads to containment failure. Any subsequent release will likely be through an unscrubbed and unfiltered release path.

Key Assumptions:

JPM A.8-05.08-001, Rev. 0, Manually Open Containment Vent Lines INITIAL CONDITIONS:

o Extreme flooding has led to a Station Blackout that has. existed at Monticello for the last 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

o Div. 1 and Div 2 250 VDC battery systems have been depleted and are not available.

o The plant was in Shutdown Cooling until the station blackout and has since been slowly repressurizing due to heating up.

o Current RPV pressure is 75 psig-and slowly rising.

o H SRV has failed to reseat and indication of the tailpipe vacuum breaker sticking open have led to a Drywell pressure of 45 psig and rising about 1 psig every 30 minutes.

o Efforts to align the diesel fire pump to DW sprays have been unsuccessful due to the flooding.

o The TSC has recommended venting the DW by using the Hard Pipe Vent using procedure A.8-05.08, Manually Open Containment Vent Lines INITIATING CUES (IF APPLICABLE):

o The CRS directs operator to initiate DW venting through the Hard Pipe Vent lAW procedure A.8-05.08, Manually Open Containment Vent Lines, Parts A and B.

Revision 2 Page A-20

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Hard Pipe Vent local manual operation Job Performance Measure entry condition assumptions, Information excerpted from Hughes Associates Record of Correspondence, Hard Pipe Vent Manual Operation e-mails with Xcel Energy during June 2013, Hughes Associates, Baltimore, MD, 7 July 2013:

Conditions anticipated following a SBO resulting from an external flooding event:

o ERO has been manned for the past several days, with these procedures predicted and planned to be implemented ahead of time o Plant is in cold shutdown condition (mode 4) o Operations staffing to perform the procedures would be optimal (several operators assigned as desired to each procedure) o Environment would be consistent with SBO (hot, dark, damp) o There would be more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to perform the procedures, allowing several opportunities to troubleshoot and/or re-perform steps if necessary o The ERO would place maximum priority on maximizing chances of successful performance of these procedures Operator Interview Insights:

The JPM that was completed by Xcel with 2 ROs and 2 NLOs was JPM A.8-05.08-001, Rev. 0, Manually Open Containment Vent Lines. The average time to complete the JPM was 30 minutes given the information that the N2 bottles were staged on the CRD catwalk. There were no problems or issues that required any of the operators to stop and get clarifying information, it was identified that removing fittings was not the best idea it would be better if the lines had tee's installed where the caps could be removed and the appropriate lines connected. This way the capped connection could be labeled to further minimize connecting to the wrong fitting. The operators stated that strips of non-skid should be placed on the areas around the site for safety reasons. They also mentioned using the LED headlights versus flashlights to allow both hands to be free.

Manpower Requirements: ... _....._.____"_

Ci~Wei~ier hclu~de~d:: TotaAIIVailale 4KRqiiiredfr 'Noe

____________ Ex'c~itip'"in _ _ _ __ _

Reactor operators Yes 2 1 Plant operators Yes 2 0 Mechanics Yes 2 0 Electricians Yes 2 0 I&C Technicians Yes 2 0 Health Physics Technicians Yes 2 0 Chemistry Technicians Yes 1 0 Execution Performance Shapina Factors:

Environment: Lighting Portable Heat/Humidity Hot / Humid Radiation Background Atmosphere Steam (although steam will not be present, this PSF was used to indicate an off-normal condition, such as would be present for flood and SBO)

Special Requirements: Tools Required Adequate Available Parts Required I I_Adequate Revision 2 Page A-21

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Clothing Required Adequate Complexity of Response: Cognitive Complex Execution Complex Equipment Accessibility Main Control Room Accessible (Cognitive):

Equipment Accessibility Reactor Building With Difficulty (Execution):

Stress: High Plant Response As Expected: Yes Workload: High I Performance Shaping Factors: Negative Performance Shaping Factor Notes:

The response is considered to be Complex due to the flooding and SBO impacts to lighting and accessibility. Flashlights, headlamps and boots were considered necessary by Training when the JPMs were performed for these tasks.

The Equipment Accessibility is evaluated as With Difficulty due to Rx building lighting and flooding issues.

Key Assumptions (see that section) regarding the conditions provided to Training for performing the JPM for this task said that the "Environment would be consistent with SBO (hot, dark, damp)". The Training insights from the JPM performance stated that the operators recommended the use of "LED headlights versus flashlights", so it is clear that portable lighting is used.

Despite preparations and training, the flooding scenario is considered to be a high stress situation.

The steps identified as Critical in JPM-A.8-05.08-001 were used for the Execution quantification.

Timing:

T SW 15.00 Hours Tdelay 5.50 Hours T1/2 10.00 Minutes TM 45.00 M Minutes~ I1 Irreversible Cue DamageState I I

-I.

t=0 Timing Analysis:

TO = Station Blackout.

Tsw = Per MAAP run Rcic-dgl 3-cts-ABS performed in support of an external flooding SDP, containment pressure reaches 56 psig at 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months following a SBO (flooding >930'). Core temperature reaches 1800 degrees F at 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months due to CST depletion and no transfer of RCIC to the torus. This is conservative timing as refilling of the CST is very likely.

Td = 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months - Based on an interview conducted in a prior analysis with a senior Shift Manager, the order to begin the procedure to manually operate the hard pipe vent would be given at approximately 27 psig containment pressure. This is due to the step in C.5-1200 (DW/Torus Pressure leg) that says if you cannot restore and maintain drywell pressure within Figure 0 (27psig for 0 ft torus level), then maintain drywell pressure less than Figure D (56 psig).

The 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is based on MAAP run Rcic-dg13-cts-ABS as the time when drywell pressure reaches 42 psia (27 psig) [Worksheet d43-1, column AC Drywell Pressure]

T1/2 = According to the initial conditions assumed by Training for the Job Performance Measure performed for this task, the ERO will have been manned for the past several days, with these procedures Revision 2 Page A-22

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS predicted and planned to be implemented ahead of time. Daily planning meetings will have been held to discuss actions to be taken, so the 10 minutes is simply an estimate of the meeting time between TSC and ERF personnel to make the actual decision to vent the DW by using the Hard Pipe Vent. The control room supervisor (CRS) will then direct operators to initiate the process.

Tm = Results of HPV local manual operation Job Performance Measure A.8-05.08-001 performed 18 June 2013. The procedure was performed four times, taking an average of 30 minutes. Additional 15 minutes for C.5-3505-A steps 3 and 4.

Time available for cognition and recovery: 525.00 Minutes Time available for recovery: 515.00 Minutes SPAR-H Available time (cognitive): 525.00 Minutes SPAR-H Available time (execution) ratio: 12.44 Minimum level of dependence for recovery: ZD Revision 2 Page A-23

1SMLII16012.000-1 Appendix A - HRA CALCULATOR REPORTS Cognitive Unrecovered HPVSBOFLOOD Table 47: HPVSBOFLOOD COGNITIVE UNRECOVERED

Pc Failure .Mecdhanism . .. -". 1 . 1 ."

11 .Branch : . HEP.::..

Pca: Availability of Information d 1.5e-03 PCb: Failure of Attention m 1.5e-02 Pcc: Misread/miscommunicate data e 3.0e-03 Pcd: Information misleading b 3.0e-03 Pce: Skip a step in procedure e 2.0e-03 Pcf: Misinterpret instruction a neg.

Pcg: Misinterpret decision logic c 6.0e-03 PCh: Deliberate violation a neg.

Sum of Pca through PCh = Initial Pc = 3.1e-02 Notes:

Presumed that SBO causes issues with normal alarms and indications so pc-a through -d were adjusted consistent with insights from EPRI 1025294, A Preliminary Approach to Human Reliability Analysis for Extemal Events with a Focus on Seismic, October 2012.

Page A-24 Revision 22 Revision Page A-24

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS pca: Avlabhty of infonnation Indiation ail in CR Indication bminglAlftmate Training on CR Accurate in Procedure Indcators I.0e-01 (a) neg.

.Oe.00 1.0e+00 (b) neg.

O.0e4.00 I .oe-o (c) neg-e0(d) 7- 1-pe-03 5.0e-0 Yes (e) 5.0e-02 No1.e+00 1.0e+00 f) 5.oe-o0 1.0e-+O (g) .0"WO MCR indications may not be accurate due to the Station Blackout, however, either procedural or informal crew information on alternate indications and training should provide operator input to decision-making.

pcrb: Failure of attention Low vs. Hi Check vs. Monitor Front vs. Back Aarmed vs.Not Wokdload Panel Alarmed Evout(a) neg-O.Oe0 Back .0(b) 1.5e-4 LOW 3.0e-03 1.0e+00 (c) 3.0e-03 1.0e+00 Front 5e-2(d) 5.0e-02O 1.-%-04 Monitor I0-0e+00 l(e) 3.0e-03 3.0e-03 Back 5.0e-02 (f) 3.0e-04

1. Cl* Nice 3.0e-03 1(g) 6.0e-03

2. c*m ic Fro,,t 15.,105(b0n-02 (h eg-Check O.Oe+0 18 0(i) neg.

0.0e*WBack (.0e.so 5.0e-02e-04 WIgh 3.0e-03 1Ae(k) 1.e-02 Front ()7.5e-04 Monitor - Oe 0- -- (m) I.5e-02 3 -Back 50e-02 (n) 1.5e-03 3.Oe-03 I1.0e0 (o) 3.0e-02 Page A-25 Revision 22 Page A-25

I SM L16012.000-1 Appendix A - HRA CALCULATOR REPORTS ISM LI 6012.000-1 Appendix A HRA CALCULATOR REPORTS pcc: Misreadftniscmmnnicate data Inhicators Easy to Good--ad Indicator Fom-- -

Locate cminunicafions I O.Oe.H) (a) neg.

O.Oe.)O 3i 3 (b) 3.0e-03

,O-Oe+0 (c) 1.0e-03 0(g) 4.0e-03 o.oe-4oo (g) 3.Oe-03 1.0e-03 (h) 4.0e-03 13.0e-03 3.0e-03 3.oe-o3 (h) 7.Oe-O3 pcd: Infonmtion fidNilng M Cuesas Stated Warnng IISpecoTc Tbn,,ig e Tining Dire o.oe+0o (a) meg.

No-------------------- - ------------- - ------- (b)3Je-03

.0-L-02 (c) 1.0e-02 1..eOe0 1..od-o (d) 1.Oe-O1 1.Oe+OO (e) 1.0e" 0 MCR indications may not be accurate due to the Station Blackout so cues may not be as stated in procedures.

pce: Sli a step in procedure Obvious Ms Eing V& s MUPleawaf Mlle cekeqing A&d 13.Oe-O3(a) 1.0e-03 3.3e-01 1.0e.-02 (b) 3.Oe-03 O.e.O0 3(c) 3.0e-03

.Oe.001e-0 (d) 1.8e-02

-(e) 4-e-03 3.3e-o------.----

Yes3.0e-03 13.e-3 (g) 6.0e-03 1.Oe4-OI (h) I.3e-02 1..0e-02 1.0e-01 (i) t.0e -0t Page A-26 Revision 22 Page A-26

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS I SMLI 6012.000-1 Appendix A HRA CALCULATOR REPORTS pf: Misinterpret instruction Standard or AM Required 7haning on Step Ambiguous wonling Information

--- ------- ------- --------- --- (a) neg.

(b) 3.oe-03 3.0e-02 (c) 3.0e-02 NO .0e4H (d) 3.0e-03 o.e-,.o I(e) 3.0e-02 3.0e-02 1.0e-01 (f) 6.Oe-03 3.0e..02 I 0 (g) 6.0e-02 pcg: Mi§sinrmpt decision ogic NMor statemmut -ir*N or -ow- Both AND' & IPracticed Scenail Statement "OFr 3.3e41 (a) 1.Ge-02 3.0e-02 (b) 4.9--02 1.2-02 -3(c) 6.0e-03

.Oe. 33.ie-Ol-(d) .9e-02 6.0e-03 (1.0e.9.42 3.3e-1 (e) 2.0e-03 O.Oe-Oe-0 1.00,6 0O (f) 6.0e-03 Yes OO +

(g) t.Oe-02 No 3.0e-02 -- (h) 3.1e-02 1.0e30.-0 I.0e-03 30"1 3.0e-0M 13.3e-01 (k) neg.

O.OedO l- (1) neg.

pch: Deliberate violation Belef in Adequacy Adverse Reasonable Poky of of Instructon Consequence if Alternatis "'Verbatic" oew------- - -- - ------ - ------ - -------- - ------ - ----- (a) neg.

Yes 5.0eM (b) 5.0"-01 oI. (c) 1.0e4W I.Oe4O110e00 O.Oe,.O0 (d)neg.

Io.oe.,o (e) neg.

Page A-27 Revision 22 Page A-27

1SML16012.000-1 REPORTS Appendix A - HRA CALCULATOR REPORTS I SMLI 6012.000-1 Appendix A HRA CALCULATOR Cognitive Recovery HPVSBOFLOOD Table 48: HPVSBOFLOOD COGNITIVE RECOVERY

< (" a) I- LL '.D Final InitiaIIl.

HE Z >a)* W 1 CO - '5: C

-r-.O W*

- nOa Lu O) Value0 Value Pc. 1.5e-03 X X - 2.5e-01 3.8e-04 Pcb: 1.5e-02 X X X MD 3.8e-03 5.7e-05 Pc': 3.0e-03 X X MD 2.1e-02 6.3e-05 PCd: 3.0e-03 X X X MD 7.3e-03 2.2e-05

<PCe: 2.0e-03 X X X MD 1.0e-02 2.0e-05

, :; neg. - 1.0e+00 P. 6.e-03 X X X MD 1.1 e-02 6.6e-05 I--neg. - 1.0e+00 SuoP P 6 *4It dt 6.1e-04 Notes:

Due to long timeframe and severity of scenario, STA and Emergency Response Facility will be available.

Operations staffing to perform the procedures was assessed by Xcel as optimal (several operators assigned as desired to each procedure) so Extra Crew was credited.

Used Moderate Dependency due to high stress.

Page A-28 Revision 22 Page A-28

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS ISM LI 6012.000-1 Appendix A HRA CALCULATOR REPORTS Execution Unrecovered HPVSBOFLOOD Table 49: HPVSBOFLOOD EXECUTION UNRECOVERED ProcedurYe Containm ent . , C:A8-5.08,eManuallyOpen CvehtLines omment Stress Over Ride jnstructionlComnent,-,

I IStep.No.* Error' - ,. HTHERP Factor.,

HEP Type' Table Item - .. ___- _

Connect and apply pressure from AH-1 cylinder to rupture Rupture Disk PSD-4543 A.8-05.08, Step 4 Location: Reactor Building EOM 20-7b 2 1.3e-03 EOC 20-12 5 1.3E-3 Total Step HEP 1.3e-02 Connect and adjust AH-1 regulator to less than 100 psig and slowly open AH-1 discharge valve to open valve AO-4539 5 A.8-05.08, Step 5 Location: Reactor Building EOM 20-7b 4 4.3e-03 EOC 20-13 5 1.3E-2 Total Step HEP 8.7e-02 Connect and adjust AH-2 regulator to less than 100 psig and slowly open AH-2 discharge valve to fully open valve AO-4540 5 A.8-05.08, Step 6 Location: Reactor Building EOM 20-7b 4 4.3e-03 EOC 20-13 5 1.3E-2 I Total Step HEP 8.7e-02 Open and Close the HPV isolation valves as directed by shift supervisor C.5-3505 Part A, Location: Reactor Building EOM 20-7b 1 4.3e-04 5 Step 3 1 EOC 20-13 2 3.8E-3 I Total Step HEP 2.1e-02 Monitor Containment Pressure and Radiation Levels in the Hard Pipe C.5-3505 Part A Vent. 5 Step 4 Location: Reactor Building EOM 20-7b 1 4.3e-04 Step 4 EOC 20-10 1 3.8E-3 Total Step HEP 2.1e-02 Feedback from Control Room 5 Recovery Location: Main Control Room EOM 20-7b 3 1.3e-03 I Total Step HEP 6.5e-03 Page A-29 Revision 22 Page A-29

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Execution Recovery HPV_SBOFLOOD Table 4-10: HPVSBOFLOOD EXECUTION RECOVERY c it.Recovery Step No.. -i HEP (crit)i HEP'Rec .Dep. P"Cond.

.Totalffor.

~Criticalý 'te k o .

.... ___.... ... _(Rec) Step A.8-05.08, Step 4 Connect and apply pressure from AH-1 cylinder to rupture 1.3e-02 7.3e-04 Rupture Disk PSD-4543 Recovery Feedback from Control Room 6.5e-03 LD 5.6e-02 A.8-05.08, Step 5 Connect and adjust AH-1 regulator to less than 100 psig and slowly open AH-1 discharge valve to open valve AO-4539 8.7e-02 4.9e-03 Recovery Feedback from Control Room 6.5e-03 LD 5.6e-02 A.8-05.08, Step 6 Connect and adjust AH-2 regulator to less than 100 pslg and 8.7e-02 4.9e-03 slowly open AH-2 discharge valve to fully open valve AO-4540 Recovery Feedback from Control Room 6.5e-03 LD 5.6e-02 C.5-3505 Part A, Open and Close the HPV isolation valves as directed by shift 2.1 e-02 1.2e-03 Step 3 supervisor Recovery Feedback from Control Room 6.5e-03 LD 5.6e-02 C.5-3505 Part A, Monitor Containment Pressure and Radiation Levels in the 2.1 e-02 1.2e-03 Step 4 1 Hard Pipe Vent.

I Recovery Feedback from Control Room 6.5e-03 LD 5.6e-02

-* .: Unrecoered:-

.TotalI.- . . .. 2.3e1-2ovIered: .Total R 1.3e2 Page A-30 Revision 2 Page A-30

ISML16012.000-1 Appendix A - HRA CALCULATOR REPORTS A.3. RCICSBOFLOOD, Fail to manually operate RCIC during SBO and extreme flooding conditions (SPAR-H)

Basic Event Summary

'Planlt;:".. i. Data :File .*.!:?'*~:*i:-*FileSize

-i:;*~* ;..'* File;*i!

Datie !:?i-;..* ;:: Rebo d *Date ::".

Monticello Ext 909312 07/02/13 07/02/13 Flooding SDP HRAJune 2013_SPAR H quant for sensitivity.HRA Table 11: RCIC_SBOFLOOD

SUMMARY

[Anal, i Resu!ts:.-I4 Cognitive Execution

[Failor bt-*b.l.t* 3.2e.. 9.1 e-02 1.4e-01 Plant:

Monticello Initiating Event:

External Flood + SBO Basic Event Context:

The flooding engineer provides daily updates to the station on high river water levels including potentials to rise above any A.6 trigger points. At this point, heightened awareness of the potential for flooding is implemented.

When river level exceeds 921 feet an evaluation of EALs would be performed. If visible damage has occurred due to flood water rising greater than 921 feet, then an Alert per EAL HA1.6 would be declared.

Prior to river levels reaching these levels, operators would be walking down the A.8 procedures for alternate methods to vent primary containment and operate RCIC remotely. This would involve staging of equipment in the torus area to open the Hard Pipe Vent and verification that equipment is properly staged to operate RCIC remotely.

EDGs and batteries are not available. Shutdown cooling, HPCI, and RCIC are not available from normal electrical means. RCIC is available for manual operation.

Operators have temporary level indication setup in the reactor building. Pressure indication is available in the direct area of the level transmitters. The building is dark and most likely water in the basement of the reactor building. Additional portable lights are available to assist with lighting and boots staged for higher water. The operators would utilize A.8-05.01 to un-latch the governor from the remote servo linkage and throttle steam flow to RCIC to start the turbine rolling while coordinating with operators monitoring water level and reactor pressure. Upon reaching the high end of the level band the operators would throttle closed the steam admission valve and await direction to re-start RCIC. Local operation of RCIC is demonstrated each refueling outage during the over speed test. Operation of a coupled turbine run is less complex because the turbine is easier to control with a load.

Revision 2 Page A-31

ISML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Timing:

T 7.97 Hours 1

Tdelay 5.75 Hours i T1/2 10.00 Minutes TM 80.00 Minutes Irreversible Cue DamageState t=o Timing Analysis: TO = Station Blackout Tsw = Time from Station Blackout to the time by which RCIC must be restored.

Per Monticello MAAP Calculations, case "SBOCase3-RI", 27 June 2013:

Time to TAF = 7.17 hrs Time to -149" = 7.2 hrs Time to 1800 F = 7.97 hrs Damage is assumed to occur if the temperature exceeds 1800 F or 7.97 hrs, so this was used as the time by which RCIC restoration is required.

Tdelay = PRA battery calc (PRA-CALC-1 1-002) indicates that there are 5.75 hrs until RCIC battery depletion. Also, the RCIC Water Flow (column BC) of the d41 tabs in the "SBOCase3-RI" MAAP analysis spreadsheet shows that RCIC injection stops at approximately the same time (5.74 hrs), so the MAAP runs agree with the calc. This Tdelay can be considered somewhat conservative, since in reality, it is likely that an action would be taken before waiting for battery depletion.

T1/2 = The cue for action is that the TSC and the Emergency Response Director have determined that RCIC operation is needed. Daily planning meetings will have been held to discuss actions to be taken as soon as the diesels are lost, so the 10 minutes is simply an estimate of the meeting time between TSC and ERF personnel to make the actual decision to manually operate RCIC. The Control Room Supervisor (CRS) directs operator to initiate RCIC and inject into the RPV using procedure A.8-05.01, Manual Operation of RCIC, Part A, Placing RCIC in Service.

Tm = Results of RCIC local manual operation Job Performance Measure performed 18 June 2013. The procedure was performed three times, taking 49 minutes, 37 minutes and 50 minutes to complete for an average time of 45 minutes. 50 minutes was used as the conservative value for JPM performance.

The JPM did not include the performance of Part B for installation and use of the Fluke level monitoring device; this was estimated to require 30 minutes, so the total time for Tm was estimated as 50 min + 30 min = 80 min.

Time available for recovery: 43.20 Minutes SPAR-H Available time (cognitive): 53.20 Minutes SPAR-H Available time (execution) ratio: 1.54 Minimum level of dependence for recovery: LD Page A-32 Revision 22 Page A-32

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS ISMLI6OI2.000.1 Appendix A HRA CALCULATOR REPORTS PART I. DIAGNOSIS PSFss PSF ,ve;s9

7 %fMultlper~or<.

Available Time Inadequate Time P(failure) = 1.0 (recommended choice Barely adequate time (~ 2/3 x nominal) 10 based on timing Nominal time 1 information in bold) Extra time (between 1 and 2 x nominal 0.1 and > 30 min)

Expansive time (> 2 x nominal and > 30 X 0.01 min)

Insufficient Information I Stress Extreme 5 High X 2 Nominal 1 Insufficient Information I Complexity Highly complex 5 Moderately complex X 2 Nominal 1 Obvious diagnosis 0.1 Insufficient Information 1 Experience/Training Low 10 Nominal X 1 High 0.5 Insufficient Information 1 Procedures Not available 50 Incomplete 20 Available, but poor 5 Nominal X 1 Diagnostic/symptom oriented 0.5 Insufficient Information 1 ErgonomicslHMI Missing/MisleadingJ 50 Poor 10 Nominal X I Good 0.5 Insufficient Information 1 Fitness for Duty Unfit P(failure) = 1.0 Degraded Fitness 5 Nominal X 1 Insufficient Information 1 Work Processes Poor 2 Nominal 1 Good X 0.8 Insufficient Information 1 Page A-33 Revision 22 Revision Page A-33

ISM LI6012.000-1ApndxA-HACLUTORERS Appendix A - HRA CALCULATOR REPORTS Diagnosis HEP:

3.2e-04 PART I1.ACTION PSFs PSF IUVeis. ~ Multi 11ýr for.

__________________ griagn sis Available Time Inadequate Time P(failure) = 1.0 (recommended choice Time available is - the time required 10 based on timing Nominal time X 1 information in bold) Time available >= 5x the time required 0.1 Time available >= 50x the time required 0.01 Insufficient Information 1 Stress/Stressors Extreme 5 High X 2 Nominal 1 Insufficient Information 1 Complexity Highly complex 5 Moderately complex X 2 Nominal 1 Insufficient Information 1 Experience/Training Low 3 Nominal X 1 High 0.5 Insufficient Information 1 Procedures Not available 50 Incomplete 20 Available, but poor X 5 Nominal 1 Insufficient Information 1 Ergonomics/HMI Missing/Misleading 50 Poor X 10 Nominal 1 Good 0.5 Insufficient Information 1 Fitness for Duty Unfit P(failure) = 1.0 Degraded Fitness 5 Nominal X 1 Insufficient Information 1 Work Processes Poor 5 Nominal 1 Good X 0.5 Insufficient Information 0.5 Page A-34 Revision 22 Page A-34

ISML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Action Probability:

9.1e-02 [Adjustment applied: 1.0e-3

1.0e+02 / (1.0e-3 * (1.0e+02 - 1) + 1)]

PART Ill. DEPENDENCY In~ mC.. Im caeI

- mhwamc cwi Task Failure WITHOUT Formal Dependence:

9.1le-02 Task Failure WITH Formal Dependence:

1.4e-01 Page A-35 Revision 22 Page A-35

ISML16012.000-1 Appendix A - HRA CALCULATOR REPORTS A.4. HPVSBOFLOOD, Fail to operate the HPV using N2 bottles to provide containment heat removal during SBO/Flood (SPAR-H)

Basic Event Summary

.:Plant , :Datale FileSize"..

FileDate : Rerd Date Monticello Ext 901120 06/28/13 06/28/13 Flooding SDP HRAJune 2013-SPAR H quant for 1sensitivity.HRA John Spaargaren & Pierre Macheret, Hughes Associates Table 412: HPVSBOFLOOD

SUMMARY

AnalyssResults-,* Cognitive Execution Fe! rN r 1b6 1i.ii 3.2e-04 5.0e-03 Total :HEP. I 5.5e-02 Plant:

Monticello Initiating Event:

External Flood + SBO Basic Event Context:

The flooding engineer provides daily updates to the station on high river water levels including potentials to rise above any A.6 trigger points. At this point, heightened awareness of the potential for flooding is implemented.

When river level exceeds 921 feet an evaluation of EALs would be performed. If visible damage has occurred due to flood water rising greater than 921 feet, then an Alert per EAL HA1.6 would be declared.

Prior to river levels reaching these levels, operators would be walking down the A.8 procedures for alternate methods to vent primary containment and operate RCIC remotely. This would involve staging of equipment in the torus area to open the Hard Pipe Vent and verification that equipment is properly staged to operate RCIC remotely.

EDGs and batteries are not available. Shutdown cooling, HPCI, and RCIC are not available from normal electrical means. RCIC is available for manual operation.

Page A-36 Revision 22 Page A-36

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS Timina:

TS 15.00 Hours Tdelay 5.50 Hours T1/2 10.00 Minutes TM 45.00 Minutes Ireversible Cue DamageState t=o Analysis: TO = Station Blackout.

Tsw = Per MAAP run Rcic-dg13-cts-ABS performed in support of an external flooding SDP, containment pressure reaches 56 psig at 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months following a SBO (flooding >930'). Core temperature reaches 1800 degrees F at 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months due to CST depletion and no transfer of RCIC to the torus. This is conservative timing as refilling of the CST is very likely.

Td = 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months - Based on an interview conducted in a prior analysis with a senior Shift Manager, the order to begin the procedure to manually operate the hard pipe vent would be given at approximately 27 psig containment pressure. This is due to the step in C.5-1200 (DW/Torus Pressure leg) that says if you cannot restore and maintain drywell pressure within Figure 0 (27psig for 0 ft torus level), then maintain drywell pressure less than Figure D (56 psig).

The 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is based on MAAP run Rcic-dg13-cts-ABS as the time when drywell pressure reaches 42 psia (27 psig) [Worksheet d43-1, column AC Drywell Pressure]

T1/2 = According to the initial conditions assumed by Training for the Job Performance Measure performed for this task, the ERO will have been manned for the past several days, with these procedures predicted and planned to be implemented ahead of time. Daily planning meetings will have been held to discuss actions to be taken, so the 10 minutes is simply an estimate of the meeting time between TSC and ERF personnel to make the actual decision to vent the DW by using the Hard Pipe Vent. The control room supervisor (CRS) will then direct operators to initiate the process.

Tm = Results of HPV local manual operation Job Performance Measure A.8-05.08-001 performed 18 June 2013. The procedure was performed four times, taking an average of 30 minutes. Additional 15 minutes for C.5-3505-A steps 3 and 4.

Time available for recovery: 515.00 Minutes SPAR-H Available time (cognitive): 525.00 Minutes SPAR-H Available time (execution) ratio: 12.44 Minimum level of dependence for recovery: ZD Page A-37 Revision 22 Page A-37

1SML16012.000-1 Appendix A - HRA CALCULATOR REPORTS PART I. DIAGNOSIS P

PSFs,6V , vels u-ti'for AfPSF

- Dagposis, Available Time Inadequate Time P(failure) = 1.0 (recommended choice Barely adequate time (~ 2/3 x nominal) 10 based on timing Nominal time 1 information in bold) Extra time (between 1 and 2 x nominal 0.1 and > 30 min)

Expansive time (> 2 x nominal and > 30 X 0.01 min)

Insufficient Information Stress Extreme 5 High X 2 Nominal 1 Insufficient Information 1 Complexity Highly complex 5 Moderately complex X 2 Nominal 1 Obvious diagnosis 0.1 Insufficient Information 1 Experience/Training Low 10 Nominal X 1 High 0.5 Insufficient Information 1 Procedures Not available 50 Incomplete 20 Available, but poor 5 Nominal X 1 Diagnostic/symptom oriented 0.5 Insufficient Information 1 Ergonomics/HMI Missing/Misleading 50 Poor 10 Nominal X 1 Good 0.5 Insufficient Information 1 Fitness for Duty Unfit P(failure) = 1.0 Degraded Fitness 5 Nominal X 1 Insufficient Information 1 Work Processes Poor 2 Nominal 1 Good X 0.8 Insufficient Information 1 Page A-38 Revision 22 Page A-38

1 SML-16012.000-I Appendix A - HRA CALCULATOR REPORTS I SMLI 6012.000-1 Appendix A - HRA CALCULATOR REPORTS Diagnosis HEP:

3.2e-04 PART II. ACTION

. PSFs,, PSFLevels' s Multiplier for S~ Diagflosis Available Time Inadequate Time P(failure) = 1.0 (recommended choice Time available is - the time required 10 based on timing Nominal time X 1 information in bold) Time available >= 5x the time required 0.1 Time available >= 50x the time required 0.01 Insufficient Information 1 Stress/Stressors Extreme 5 High X 2 Nominal 1 Insufficient Information 1 Complexity Highly complex 5 Moderately complex 2 Nominal X I Insufficient Information 1 Experience/Training Low 3 Nominal 1 High X 0.5 Insufficient Information 1 Procedures Not available 50 Incomplete 20 Available, but poor 5 Nominal X 1 Insufficient Information 1 Ergonomics/HMI Missing/Misleading 50 Poor X 10 Nominal 1 Good 0.5 Insufficient Information I Fitness for Duty Unfit P =failure) 1.0 Degraded Fitness 5 Nominal X 1 Insufficient Information 1 Work Processes Poor 5 Nominal 1 Good X 0.5 Insufficient Information 0.5 Action Probability:

5.0e-03 Page A-39 Revision 22 Page A-39

ISMLI16012.000-1 Appendix A - HRA CALCULATOR REPORTS PART II1. DEPENDENCY C,=, TD I E I I C ,

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5.3e-03 Task Failure WITH Formal Dependence:

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ISML16012.000-1 Appendix B - EVENT TREES B. APPENDIX B - EVENT TREES Page B-I Revision 2 Page B-1

EXTERNAL FLOOD >930' < 935' EARLY WARNING REACTOR BUILDING PROTECTED RCIC/RPV &HARD PIPE VENT Prob Name SUCCESSFUL EARLY WARNING O.00E+00 DK O.OOE+00 FLOOD <935' RCIC SUCCESS 0.891 7.09E-06 OK 0.894

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IMonticello, Flood SDP 930-935.eta 17/3/2013 1 Page 1 IMonticello Flood SDP 930-935.eta 7/3/2013 Page 1

EXTERNAL FLOOD >930' < 935' EARLY WARNING REACTOR BUILDING PROTECTED RCIC/RPV & HARD PIPE VENT Prob I Name SUCCESSFUL EARLY WARNING O.OOE+00 inflflF4flf :)K FLOOD <935' RCIC SUCCESS 0.5 a; mai--fnr, ::)K 0.894

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[2.00E-05] SUCCESSFUL EARLY WARNING

)K O.00E÷00 RCIC SUCCESS FLOOD > 935' 7.96E-06 :)K REACTOR BLDG PROTECTED 0.894

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Monticello Flood SDP 930-935 Sens Freq.eta 17/3/2013 1 Page 1

EXTERNAL FLOOD >930' < 935' EARLY WARNING REACTOR BUILDING PROTECTED RCIC/RPV & HARD PIPE VENT Prob Name SUCCESSFUL EARLY WARNING D.O0E+00 :)K FLOOD <935' 0.891 I1 O.OOE+00 RCIC SUCCESS 5.38E-06 :)K 0.805

[1] I 1.55E-06 ,D Seq 1 EXTERNAL FLOOD >930' [0.195]

[8.90E-06] SUCCESSFUL EARLY WARNING Ll=tJI.l* Jt I

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)K 0.OOE+00 RCIC SUCCESS FLOOD > 935' 5.95E-07 :)K REACTOR BLDG PROTECTED 0.805

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[0.11]

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'D Seq 3 Monticello Flood SDP 930-935 Sens SPAR-H.eta 7/3/2013 Page 1

Enclosure 3 Monticello Nuclear Generating Plant "Monticello Flood Protection" 11 Pages Follow

Monticello Flood Protection 1.0 PURPOSE The purpose of this document is to evaluate the flood protection provided at Monticello Nuclear Generating Plant (MNGP).

Nuclear power plants are designed to meet robust design criteria, referred to as General Design Criteria (GDC); which are now codified as part of NRC regulations in 10 CFR Part 50. The GDC have existed in various forms prior to being codified in part 50 and plant commitments to meet the GDC (or pre-existing requirements) depend on the age of the plant.

MNGP was designed before the publishing of the 70 General Design Criteria (GDC) for Nuclear Power Plant Construction Permits proposed by the Atomic Energy Commission (AEC) for public comment in July 1967, and constructed prior to the 1971 publication of the 10 CFR 50, Appendix A, GDC. As such, MNGP was not licensed to 10 CFR 50 Appendix A, GDC. The MNGP USAR, Section 1.2, lists the Principal Design Criteria (PDC) for the design, construction and operation of the plant. MNGP USAR Appendix E provides a plant comparative evaluation to the 70 proposed AEC design criteria. It was concluded in the USAR that the plant conforms to the intent of the GDC. A listing of the PDC and AEC GDC (by number and title) pertaining to external flooding is provided below:

PDC 1.2.1 .c "General Criteria" "The design of those components which are important to the safety of the plant includes allowances for the appropriate environmental phenomena at the site.

Those components important to safety and required to operate during accident conditions are designed to operate in the post accident environment."

AEC Criterion 2 - Performance Standards (Category A)

"Those systems and components of reactor facilities which are essential to prevention of accidents which could affect the public health and safety or to mitigation to their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established shall reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and surrounding area and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design."

This evaluation addresses the following aspects of flood protection that are provided for the MNGP to meet AEC Criterion 2:

Flood Analyses - this discussion describes the site location, hydrology and determination of the maximum predicted flood water elevations and timing.

Page 1 of 11

Flood Mitigation Strategy - this discussion describes the aspects provided to preclude the design bases flood from adversely impacting the site. This protection is provided by structural design and procedural actions.

Flood Protection Implementation - this discussion describes the actions taken at the site to provide reasonable assurance that flood protection strategy can be effectively implemented in a design bases flood scenario.

2.0 FLOOD ANALYSIS This section describes the site location and hydrology, and a summary of current design basis flood elevations. Information in this section is based on information in the MNGP Updated Safety Analysis Report (USAR) (Reference 1); specific sections are identified below.

2.1 Site Location and Description The plant is located within the city limits of Monticello, Minnesota on the right (west) bank of the Mississippi River. The topography of the MNGP site is characterized by relatively level bluffs which rise sharply above the river. Three distinct bluffs exist at the plant site at elevations 920, 930, and 940 ft. above msl. The finished plant grade is approximately 930 ft. msl. The plant grade surrounding Class I and Class II structures housing Class I equipment varies between 935 ft. msl and 930 ft. msl. The site description and topography is described in detail in the MNGP USAR, Section 2.2.

Hydrology The Mississippi River is the major hydrologic feature for the site. The river poses the significant flooding source for the site. Table 1, below, summarizes normal and flooded river flow rates and water elevations.

Table 1 Normal and Flooded River Flow Rates and Water Elevations Mississippi River Flow Rate (cfs) Water Elevation Condition (ft. msl)

Normal 4,600 905 Maximum Recorded (1965) 51,000 916 1000 Year Flood -90,000 (1) 921 Probable Maximum 364,900 939.2 Flood 364,900 _ 939.2 (1) Estimated using USAR Appendix G, Exhibit 8, for a water elevation of 921 ft.

Normal river level at the MNGP site is about 905 ft. msl at a distance 1.5 miles upstream, the normal river elevation is about 910 ft. msl and at an equal distance downstream, the river is at 900 ft. msl. The following flow statistics are estimated for the Mississippi River at the MNGP site:

Page 2 of 11

Average Flow - 4,600 cubic feet per second (cfs)

Minimum Flow - 240 cfs Maximum Flow - 51,000 cfs The maximum reported high water level at the MNGP site was about 916 ft. msl which was recorded during the spring flood of 1965 with an estimated river flow of 51,000 cfs.

The results of flood frequency study for the 1000 year flood estimated a peak stage of 921 ft. msl (USAR Section 2.4) 2.2 Design Basis Flood Hazard The following flood scenarios are evaluated as part of the MNGP licensing basis [USAR Appendix G]:

Flooding in Streams and Rivers

Flooding due to Downstream Ice Dam Build-Up A summary of the results as described in Reference 2 for each of these flooding scenarios is provided below; specific sections from Reference 2 are identified with the associated discussion.

2.2.1 Flooding in Streams and Rivers The probable maximum discharge was determined to be 364,900 cfs and a corresponding peak stage of elevation 939.2 ft. msl. The flood would result from meteorological conditions which could occur in the spring and would reach maximum river level in about 12 days. It was estimated the flood stage would remain above elevation 930.0 ft. msl for approximately 11 days.

The most critical sequence of events leading to a major flood would be to have an unusually heavy spring snowfall and low temperatures after a period of intermittent warm spells and sub-freezing temperatures has formed an impervious ground surface and then a period of extremely high temperatures followed by a major storm. The snowmelt and rainfall excesses were then routed to the plant site by computer modeling. A stage discharge rating curve was then constructed. The probable maximum discharge was determined to be 364,900 cfs with a corresponding peak stage elevation of 939.2 ft. msl from the discharge rating curve.

A probable maximum summer storm over the project area was also studied in detail and the resulting flood at the project site determined. Although the summer storm was much larger than the spring storm, the initial retention rate of zero for spring conditions, and the snowmelt contribution to runoff, resulted in the spring storm producing the more critical flood.

Key Assumptions Used to Determine Design Basis Flood Hazard The PMF evaluation for the spring storm conservatively maximizes the potential snow cover and precipitation. A limiting temperature sequence that results in an impervious ground surface due to subfreezing temperatures is assumed. This is followed by extreme high temperatures, and a subsequent major spring storm. The snowmelt and Page 3 of 11

rainfall maximizes the runoff to the river basin. This sequence of events is postulated to produce a PMF. Additional details regarding key assumptions used in the analyses are described in USAR Appendix G.

Methodology Used to Develop Design Basis Flood Hazard The predicted flood discharge flow and PMF level at the MNGP site was defined using Department of the Army, Office of the Chief of Engineers, the U.S. Army Corps of Engineers, Engineer Circular No. 1110-2-27, Enclosure 2, "Policies and Procedures Pertaining to Determination of Spillway Capacities and Freeboard Allowances for Dams,"

dated August 1, 1966 (Reference 2).

The PMF at the MNGP site was determined by transposing an actual critical spring storm to the drainage basin and maximizing the precipitation for potential moisture.

Potential snow cover and a critical temperature sequence were developed for determining snowmelt contribution to flood runoff.

The study area was divided into four major sub-basins and synthetic unit hydrographs were developed for each, using Snyder's method, which is derived from the various physical basin characteristics. Unit hydrograph peaks were also increased by 25 percent and basin lag decreased by one-sixth, in accordance with standard Corps of Engineer practice.

Snowmelt and rainfall excesses were applied to unit hydrographs and the resulting hydrographs determined for each sub-basin. Sub-basin hydrographs were then routed to the project site by computer program using the modified Wilson method. Travel times for flood routing were taken from Corps of Engineers recorded travel times for large floods.

Base flow was determined from long-term records of stream flow for nearby stations.

Base flow was then added to the total of the routed flood hydrographs.

The stage-discharge curve at the MNGP Site was extended above the range of historical experience by means of hydraulic computations based on the river channel downstream.

This was done by a series of backwater computations based on a range of discharges.

Backwater computations were made using water surface elevations and their corresponding discharges as determined from the rating curve downstream from Monticello. Using the discharges and the resulting water surface elevations, a stage discharge curve was constructed for the site.

Results The detailed analysis results are presented in USAR Appendix G. To summarize, the analysis predicts a probable maximum discharge of 364,900 cfs and a corresponding peak stage of elevation 939.2 ft. msl. The flood would reach maximum river level in about 12 days after the beginning of high temperatures, and it was estimated the flood stage would remain above elevation 930.0 ft. msl for approximately 11 days.

It is noted that the 12 day time period is for the river elevation to reach the peak level.

Other important levels are the elevation of the Intake Structure (919 ft.) and Plant Grade (930 ft.). Based on USAR Appendix G Exhibits 8 and 9, water elevation of 919 ft. could be exceeded at about the fourth day and water elevation of 930 ft. could be exceeded at the eighth day.

Page 4 of 11

2.2.2 Floods due to Ice Dam Build-Up Flooding due to backwater, usually caused by ice jams, was considered. USAR, Appendix G, Chapter II, Page G.2-5 states that two types of flooding occur in the basin --

open-water flooding and backwater flooding. Flooding while open-water conditions prevail is caused by runoff producing rains, or by melting snow, or by a combination of the two. Flooding because of backwater is usually caused by ice jams. The most serious flooding throughout the basin has been associated with excessive snowmelt and rainfall.

Thus, the open-water flooding was considered to be more limiting that the backwater flooding, and was analyzed in detail in the USAR.

3.0 FLOOD MITIGATION STRATEGY Flood protection features and flood mitigation procedures are described below. The PMF event is applicable to all modes of operation (i.e., power operation, startup, hot shutdown, cold shutdown, and refueling). Flood Protection requirements necessary to prevent external flooding or flood damage to Class I Structures or Class II structures housing Class I equipment, are identified in USAR Section 12.2.1.7.1. Flood protection features utilized at MNGP in the event of a PMF include both incorporated (installed) and temporary active and passive barriers. MNGP does not rely upon any flood protection features external to the immediate plant area as part of the current licensing basis that protect safety related systems, structures and components from inundation and static/dynamic effects of external floods.

Incorporated engineered passive or active flood protection features are features that are permanently installed in the plant that protect safety related systems, structures, and components from inundation and static/dynamic effects of external flooding. Examples include external walls and penetration seals that are permanently incorporated into a plant structure.

Temporary passive or active flood protection features at MNGP include portable pumps, sandbags, plastic sheeting, steel plates, levees, etc., that protect safety related systems, structures and components from the effects of external flooding.These features are temporary in nature, i.e., they are installed prior to design basis external flood levels attaining specific levels.

The following Class I and II structures are protected from flooding up to 939.2 ft. msl:

1. Reactor Building (including High Pressure Coolant Injection (HPCI) structure)

2. Turbine Building

3. Intake Structure (including access tunnel)

4. Off-gas Stack and Compressed Gas Storage Building

5. Radwaste Building

6. Diesel Generator Building

7. Plant Control and Cable Spreading Structure

8. Emergency Filtration Train (EFT) Building

9. Diesel Fuel Oil Pump House

10. Diesel Oil Storage Tank Page 5 of 11

Flood preparations at the site begin with a flood surveillance procedure (Reference 6).

During the time period of interest the surveillance was initiated by procedure annually in the late winter. The procedure is currently performed monthly for river level predictions and an annual performance includes inventory and inspection in addition to the river level prediction. The purpose of this procedure is to determine if the potential for plant flood exist prior to and during the spring flooding season to ensure adequate steps are taken to protect the plant ifthe potential for flooding exists. The actions taken in Reference 6 are summarized as follows:

Based on the nature of the design basis flood (heavy snow pack, thawing/freezing cycle, coupled with heavy rain) the flood scenario is slow developing and flood levels are generally predictable. Reference 6 determines the potential for flooding based on forecast information from the National Weather Service and river level monitoring. Procedure A.6 (Reference 7), Note to Step 5.2.1, indicates that the National Weather Service Flow Exceedance Probability Forecast on internet http://www.crh.noaa.qov is used to forecast river elevations. The information for the St. Cloud and Anoka measurement stations is provided on a weekly basis in terms of the probability that the river flow will exceed a given flow rate. The prediction information at the website is for the next 90 days based on current conditions. A flow discharge curve in Reference 7 is used to determine predicted river water elevation based on the predicted flow rate. Given the conditions that precede the PMF; i.e., snowpack with thawing and refreezing, it is reasonable to expect that the responsible individuals at the plant (engineering, operations, management) would be keenly aware of the need to monitor river water elevations for predicted flood conditions. Increased monitoring and use of the predictive National Weather Service tools would increase the time available to implement flood protective actions.

" Flood preparation measures are taken as part of Reference 6 to ensure that flood protection materials such as sandbags, steel plates, covers and gaskets, and plugs are available. Contact information for vendors that would be used as part of flood preparation activities are confirmed to still be valid. This contact information includes vendors that would be involved with construction of the bin wall and earthen levee. These actions are implemented even if flood conditions are not predicted. A memorandum of understanding is in place with VeitVeit & Company, a local construction firm, to provide construction related services in the event of a site emergency, and would cover activities such as construction of the earthen levee.

" In the event that the potential for flood conditions, dump trucks and excavators are ensured to be available for installation of the levee, and a detailed flood plan is developed.

MNGP Procedure A.6 (Reference 7), "Acts of Nature," (Part 5 -.External Flooding) stipulates the actions to be taken in the event flood waters are predicted to exceed elevation 918 ft. Revision 41 through Revision 45 of Procedure A.6 (Reference 7) were in effect during the period of time from February 29, 2012 through February 15, 2013.

Revision 41 was issued on February 28, 2012 and Revision 45 was issued on February 14, 2013.

Page 6 of 11

The following summarize the actions in A.6 based on the different predicted flood water elevations.

Step 5.2.8, river level is predicted to exceed elevation 918 ft. Notification of Unusual Event is declared. Actions are taken to protect equipment such as the discharge structure substation.

Step 5.2.9, river level is predicted to exceed elevation 919 ft. Actions are taken to protect the Intake Structure from flooding. As noted above the Intake Structure is at elevation 919 feet.

Step 5.2.10, river level is predicted to exceed elevation 921 ft. An Alert is declared and the plant is shutdown and cooled down to cold shutdown conditions. Actions are taken to ensure a supply of service water is available.

Step 5.2.11, river level is predicted to exceed elevation 930 ft. The bin walls and earthen levee are built. Steel plates are installed on the outside roof areas of the Intake Structure. Yard drains and other paths that could result in a water pathway that bypasses the levee are closed. An alternate access route to the plant is provided from higher ground in the event that the normal access road is flooded.

The levee is designed to provide flood protection up to a river elevation of 941 ft.

Backup flood protection to the levee can be provided by closing up the various buildings using steel plates, installing sand bags, etc. It is noted that the levee is identified in the procedure as the preferred option but, per the procedure, the backup flood protection can be used in lieu of constructing the levee. This is discussed in more detail below.

" The remaining Steps 5.2.12 and 5.2.13 provide additional backup flood protection for predicted river elevations above 930 feet. These are backup flood protection measures to the levee.

As described in A.6, Step 5.2.11, Note 2, the preferred flood protection measure is construction of a levee around the plant. The decision to use the levee as the preferred flood protection is based on a recommendation from the US Army Corps of Engineers (USACE), letter dated November 8, 2001. This USACE letter is referred to in the Bases discussion for Part 5 of Reference 7. However, Reference 7 includes an option for providing flood protection in lieu of construction of the levee. This optional flood protection means involves installing barriers (steel plates, etc.), sandbags, and sealing penetrations. Resource loaded schedules developed in support of the A.6 procedure demonstrate that the activities were achievable in the time required. Recent simulations and demonstrations confirm the construction time for the bin wall, steel plate installation and sand bagging.

As shown on Figure 13.10 of Reference 7, construction of the levee includes construction of a bin wall to the immediate east and west of the Intake Structure. The bin wall was added as part of Revision 41 to A.6 on February 28, 2012. Prior to Revision 41, the levee was made entirely of earthen material. The decision to use the bin wall was based on an analysis performed by Short Elliot Hendrickson, Inc., (SEH) (Reference 8).

As part of this same change, the configuration of the levee was modified from a ring levee entirely around the plant to a horseshoe design that ties into areas of the site that Page 7 of 11

are above the peak PMF water elevation. The recommendation to use the bin wall was made as part of Reference 8 after considering various options for the tie to the Intake Structure. Reference 8 included the following recommendations:

Secure a borrow source of levee fill within 15 minutes of the site or purchase and store on site.

Purchase bin wall materials, assemble in modules to reduce installation time frame, and store on site.

The deficiencies identified in Reference 5 have subsequently been addressed. In addition to the noted deficiencies, other areas were also identified for improvement to the plant and procedures. All of these areas for improvement were entered into the plant corrective action system.

Additional actions have been implemented to further improve the flood protection at the site. These additional actions are summarized below:

Bin wall materials have been procured and are now stored on site. The procurement of the bin walls took approximately eight weeks; however, this was treated as a normal procurement. The bin walls were supplied by Contech Engineered Solutions. Based on discussions with Contech Engineered Solutions it is estimated that the bin wall sections could be provided in approximately 14 days in an emergency situation. As discussed above, in the event that the bin walls cannot be constructed due to unavailability of materials, flood protection could still be provided as stipulated in the procedure using the sandbag and flood barrier option. This option is independent of the levee and bin walls.

Levee materials have been procured and are now stored on site. Levee materials were delivered to the site within four 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months shifts.

External flood surveillance procedure, 1478, (Reference 6) has been improved to increase the frequency of river level monitoring during potential flooding conditions. The additional river level monitoring ensures timely plant preparation for a potential flood. The increased river level monitoring serves to provide earlier warning of predicted flood levels and increases available time to implement protective actions.

Procedure A.6 (Reference 7) has been revised to improve the procedure clarity, remove unnecessary steps, and ensure completeness of protective actions.

" Detailed work instructions have been developed to implement actions in A.6.

The work instructions provide the technical detail necessary to implement the required action. Pre-staging the work instructions prior to the event reduces the required time frames to implement the required actions in A.6.

Monticello conducted a self-assessment of the site flood protection response to take an additional critical review. The self assessment was performed by a team of Xcel and contract professionals experienced in areas of flood protection.

Page 8 of 11

Specific areas for improvement were identified during the self assessment and were entered in the corrective action program and are being actively addressed.

4.0 FLOOD MITIGATION STRATEGY - FURTHER DEMONSTRATIONS Table top walkthroughs of procedure A.6 have been performed to demonstrate feasibility of performance of the required actions. A detailed schedule is developed for the actions in A.6 using input from the site departments who would execute the actions. The schedule shows actions to be performed, time frames, and sequencing, and demonstrates that the actions can be completed within the available time period.

Detailed work instructions have been developed to implement the actions in A.6. Pre-staging of the work instructions reduces the overall time to perform the tasks by removing the time associated with work planning, identifies that materials that may be needed to accomplish the work, and identifies any potential interferences or impediments to completing the required task ahead of time.

As described in Section 3.0, above, the materials to construct the bin wall sections and the earthen levee have been procured and are stored on site. As previously discussed, a memorandum of understanding (MOU) is in place with VeitVeit & Company, a local construction firm, to provide equipment and services for construction of the bin wall and earthen levee.

Reasonable simulation of construction of several of the actions believed to be more time consuming was performed in order to demonstrate that the actions could be performed within the available time frame. Specific actions examined were construction and filling of the bin walls, construction of the steel plates around the roof of the Intake Structure, and filling of sandbags. The results from these reasonable simulations are summarized below.

Construction of bin walls. For the reasonable simulation, approximately 5% of the total bin wall sections were constructed and filled. The simulation was contracted to Veit to add realism per our MOU and exercise the mobilization of personnel.

The reasonable simulation took 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months for one crew to fully construct and fill.

Based on using six crews, available per our MOU, to construct and fill the bin walls during implementation of procedure A.6, this would indicate that the entire bin wall sections could be fully constructed within 1.4 days. Accounting for issues such as excavation, inclement weather, security concerns, coordination, total construction time of four days is reasonable.

Steel plates around roof of Intake Structure. As part of procedure A.6 steel plates are attached to the wall of the Intake Structure with anchors and the seams between the plates welded to form part of the flood protection barrier. For the reasonable simulation, approximately 20% of the plates were installed on a mock-up. The reasonable simulation took 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and 11 minutes. Based on one welder all of the plates could be installed within 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months . Using two welders would reduce this time to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Additional welders would reduce this time even more. This time period is much less than the available time and provides margin for working in inclement weather conditions.

Page 9 of 11

Sandbagging. Per procedure A.6, approximately 100,000 sandbags are filled.

Sandbags are used in several steps in A.6 to seal openings, provide backup flood protection. Sandbags are critical in the event that the sandbag and flood barrier option were implemented in A.6 in lieu of the levee option. Reasonable simulation indicates that 600 sandbags can be filled per hour using one machine and 8 people. Go-Baggers are a manual bagging apparatus with which an individual can fill 55 bags an hour. With one machine and 20 people working around the clock, the 100,000 sandbags can be filled in 2 1/2 days. Reasonable simulation also showed that a steel double door can be sandbagged by five personnel in 34 minutes. Furthermore, it was shown that laying lumber and sandbagging 1 EDG room can be accomplished by 14 personnel in four hours.

In the three cases discussed above, the reasonable simulation concluded that the required actions can be accomplished within the available time frame.

5.0 CONCLUSION

S The following conclusions are drawn from the above discussion:

The postulated flood scenario for the MNGP is considered to be very conservative. The methodology employed provides conservative results. This can be seen from the comparison of river flow rates and water elevations in Table 1 in Section 2.1, above.

The flood is a relatively slow developing evolution that allows time for plant staff to monitor, predict and implement appropriate actions to provide the required flood protection.

The flood mitigation procedure clearly identifies actions for plant staff to implement to provide the required flood protection.

In the event that the levee were not able to be constructed due to not having the bin wall materials available, the procedure provides an optional approach to implement flood protection without relying on the levee and bin wall system.

Table top review of the steps to implement this optional means of flood protection demonstrated that the protection could be provided within the available time.

Subsequently actions have been taken to procure the bin wall and levee materials. Reasonable simulation has demonstrated that the levee and bin wall system can be installed well within the available time.

Page 10 of 11

6.0 REFERENCES

1. Monticello Nuclear Generating Plant, Updated Safety Analysis Report (USAR), Revision 29.

2. Department of the Army, Office of the Chief of Engineers, the U.S. Army Corps of Engineers, Engineer Circular No. 1110-2-27, Enclosure 2, "Policies and Procedures Pertaining to Determination of Spillway Capacities and Freeboard Allowances for Dams,"

dated August 1, 1966.

3. NRC Letter to Licensees, dated March 12,2012, "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1,2.3, and 9.3 of the Near Term Task Force Review of Insights from the Fukushima Daiichi Accident" (ADAMS Accession No. ML12053A340).

4. NEI 12-07, Revision O-A, "Guidelines for Performing Verification Walkdowns of Plant Flood Protection Features," dated May 2012 (ADAMS Accession No. ML12173A215).

5. Xcel Energy Letter L-MT-1 2-097, "MNGP Final Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated November 27, 2012.

6. Procedure 1478, "External Flood Surveillance," Revision 7. [Revision 7 is the procedure revision currently in effect. During 2012, Revisions 4 through 6 was in effect and the procedure was titled "Annual Flood Surveillance."]

7. Procedure A.6, "Acts of Nature," Revision 46. [Revision 46 is the procedure revision currently in effect. When used, previous revision numbers are identified in the text.]

8. Xcel Contract No. 38398, SEH No. MONNE 117980, "Monticello Nuclear Generation Plant, External Flooding Plan Update: Alternative Analysis and Final Design Report," dated January 5, 2012.

Page 11 of 11

Enclosure 4 Monticello Nuclear Generating Plant "Annual Exceedance Probability Estimates for Mississippi River Stages at the Monticello Nuclear Generating Plant based on At-site Data for Spring and Summer Annual Peak Floods" 12 Pages Follow

Annual Exceedance Probability Estimates for Mississippi River Stages at the Monticello Nuclear Generating Plant based on At-site Data for Spring and Summer Annual Peak Floods David S. Bowles and Sanjay S. Chauhan RAC Engineers & Economists June 28, 2013

Purpose:

To estimate the annual exceedance probabilities (AEPs) for Mississippi River Stages 917, 930 and 935 ft.

NGVD 29 at the Monticello Nuclear Generating Plant (MNGP) using at-site data for spring and summer floods.

These estimates are intended to improve on the previous annual peak flood estimates that were submitted on April 8, 2013. The previous estimates were was based on an at-site flood frequency curve constructed using a) a conservatively assigned AEP to the Harza spring PMF, and b) at-site flood frequency estimates obtained from a drainage-area weighted interpolation between provisional USGS annual peak flood frequency estimates for the upstream and downstream Mississippi River gages at St.

Cloud and Elk River, respectively.

Given more time, we recommend that a Monte Carlo rainfall-runoff approach should be used to develop estimates of extreme flood frequencies to make use of regional precipitation data and a more physically-based transformation of rainfall to runoff, including snow melt and explicit consideration of uncertainties.

Available Information:

Observed mean dailyflows at the following locations:

1) Station Number 05270700 Mississippi River at St. Cloud, MN with a period of record from 1989 to 2012.

2) Station Number 05275500 Mississippi River at Elk River, MN with a period of record from 1916 to 1969.

3) Monticello Nuclear Generating Plant (at-site) with a period of record from 1970 to 2012.

Floodfrequency estimates of annual peak discharges:

4) Provisional 2013 USGS annual peak discharge flood frequency analyses with estimates of AEPs ranging from 1 in 1.005 to 1 in 500 based on maximum daily flow rates for the annual peak flows:

1

a. Station Number 05270700 Mississippi River at St. Cloud, MN with drainage area of 13,320 sq. miles.

b. Station Number 05275500 Mississippi River at Elk River, MN with drainage area of 14,500 sq. miles.

Probablemaximum flood (PMF) peak discharge and stage estimates:

5) Harza 1969 (spring) PMF peak discharge and river stage at the Monticello Nuclear Generating Plant - 1912 Datum.

6) Bechtel 20121 spring and summer PMF peak discharges and river stages at the Monticello Nuclear Generating Plant - NAVD88 Datum.

Dischargerating relationships:

7) Harza 1969 relationship between river stage (1912 Datum) and river discharge (cubic feet per second, cfs) over the range 26,000 to 437,000 cfs.

8) Ops manual equation between river discharge (cubic feet per second, cfs) up to 4,000 cfs and river stage (NGVD 29 Datum): Q = 122(Stage - 901)2.2.

Datum conversionsfor river stages:

9) NGVD 29 Datum = 1912 Datum -0.36 ft.

10) NGVD 29 Datum = NAVD 88 Datum - 0.4 ft.

Procedure:

The following two approaches were examined for developing improved at-site estimates of annual exceedance probabilities (AEPs) for the spring and summer annual floods in the Mississippi River at the MNGP:

1) Drainage-area weighted interpolation of flood frequency estimates for the upstream and downstream USGS gages: Similar to the April 8, 2013 approach, an at-site flood frequency curve was obtained from a drainage-area weighted interpolation between flood frequency estimates for the upstream and downstream Mississippi River gages at St. Cloud and Elk River, respectively. However, this revised approach was conducted separately for spring and summer annual peak floods and it did not conservatively assign an AEP to the Harza PMF as was done in the April 8, 2013 approach. Instead the flood frequency curve was extrapolated to extreme floods thus providing estimate of the AEPs for the spring and summer PMF peak flow estimates and for the three elevations of interest.

1 The report, Bechtel 2012 spring and summer PMF peak discharges and river stages at the Monticello Nuclear Generating Plant - NAVD88 Datum, has been provided to NSPM. This study provides bounding estimates to site peak flood elevations applicable to the development of annual exceedance probabilities at the Monticello Nuclear Generating Plant.

2

2) Flood frequency analysis based on at-site flow data: A flood frequency analysis was conducted on the available at-site streamflow data for the period 1970 to 2012 with extrapolation to extreme floods. This provided estimate of the AEPs for the three elevations of interest and for the PMF peak flow estimates for spring and summer annual peak floods.

Both of the above approaches included estimating separate flood frequency relationships for spring and summer annual peak floods. Data for estimating these relationships were obtained using the following definitions of spring and summer floods based on discussions in the Hydrologic Atlas of Minnesota (State of Minnesota 1959) and examination of the flow records:

1) Spring annual peak floods generally peaked in the period March to May, but if it was clear from examination of the hydrograph that a snow melt flood event peaked in June then that peak was used.

2) Summer annual peak floods generally peaked in the June to early October period, but flood peaks occurring in June, which were clearly associated with snow melt events, were excluded as mentioned in 1). Since the recession limb of the annual snow melt hydrograph extends through the summer, the peak flow rates for summer floods, which are associated with convective storms, are dependent to some degree on the magnitude of flow on this recession limb at the time of the summer flood.

The two approaches are discussed in more detail below.

Approach 1): Drainage-areaweighted interpolationof flood frequency estimatesfor the upstream and downstream USGS gages The provisional USGS annual peak flow flood frequency estimates for the Mississippi River gages at St.

Cloud and Elk River were verified using USGS flood frequency software following Bulletin #17B flood frequency analysis procedures (USGS 1982). The following softwares were applied to the maximum daily annual peak streamflow data assembled by the USGS to verify their results:

1) PeakFQ: Bulletin #17B procedure based on method of moments parameter estimation for a Log Pearson Type 3 probability distribution (Flynn et al 2006)

2) PeakfqSA: The more efficient Expected Moments Algorithm (EMA) applied to the Bulletin #17B methodology (Cohn 2012)

Following the USGS provisional analysis, the Bulletin #17B (PeakFQ) software was applied to the St Cloud gage and the EMA (PeakfqSA) software was applied to the Elk River gage for the verification step.

Separate flood frequency analyses were then conducted for spring and summer annual peak flow data at the St. Cloud and Elk River USGS gages. The maximum daily annual peak streamflow data were assembled by the USGS and provided with their provisional flood frequency analyses. These data comprised a mixture of spring and summer floods. These data were separated into spring and summer floods and mean daily peak flow data were obtained from USGS flow records at both gages for those cases that were not covered by the maximum daily annual peak streamflow data assembled by the 3

USGS. An additional year (2012) of data was added for the St Cloud gage. Maximum daily annual peak flows were estimated from mean daily annual peak flows for those data not included in the USGS provisional analyses using regression relationships established between maximum daily annual peak flow data and mean daily annual peak flow data for spring and summer flows for each gage.

The PeakfqSA software containing improved EMA parameter estimation (Stedinger 2013) was applied to both spring and summer flood data for both gages. No outliers were identified. The EMA software provided AEP estimates from I in 1.0001 to I in 10,000. Estimates of annual peak discharge on the Mississippi River at the MNGP were then obtained based on linear interpolations between various frequency (AEP) estimates developed for the St Cloud and Elk River gages as a function of drainage area with the drainage area at the MNGP being 14,071 sq. miles. This interpolation the procedure is the same as developed for the April 2013 AEP estimates. The at-site annual peak discharge estimates were converted to at-site river stages using a combination of the Harza and Ops Manual equation rating curves shown in Figure 1.

Examination of the relationship between flood estimates for various AEPs and drainage area shown in Figure 2 showed an inconsistent relationship that was increasing or decreasing with drainage area. As a result we have not relied on these estimates in favor of using the flood frequency estimates obtained from analysis of at-site data in the second approach.

947 942 937

932 S927 0J _______

.5922 S917 0J __________

912 907 1 1 _________

9021 100 1,000 10,000 100,000 1,000,000 Discharge In cfs

- Ops Man Eqn, Q= 122(Stage-901)^2.2 - Harza Discharge Rating -- Transition - . Final curve Figure 1. Combined Harza and Ops Manual Equation stage-discharge rating curve 4

Interpolation of at-site quantiles 80,000

-TF

0.995 70,000 U- -0.99

-0.95 60,000

- 0.9 50,000 .-- 0.6667

ý0.5 0 40,000 --*-0.4292

-o-0.2 E

. 30,000 U

-0.04

-0.02 20,000

-.-- 0.01 0.005 10,000 0.002

-MNGS 13,000 13,500 14,000 14,600 I 5,000 Drainage Area (sq. miles)

Figure 2. Relationships between flood estimates for various AEPs and drainage area for Approach 1 Approach 2): Floodfrequency analysis based on at-site flow data The second approach is based on extrapolation of flood frequency relationships developed from at-site flow data. The mean daily annual peak stages were obtained for spring and summer annual peak floods following the process summarized above. Since only single observations have been recorded for each day it was not possible to obtain maximum daily annual peak stages. The daily annual peak stages were converted to daily annual peak flows for spring and summer floods using the combined rating curve shown in Figure 1. The EMA (PeakfqSA) software was applied to estimate the flood frequency relationships for spring and summer annual peak floods. No outliers were identified for the spring season, but one low outlier (2,416 cuffs) was identified for the summer season using the Multiple Grubbs-Beck Test) low outlier identification method. The EMA software provided AEP estimates for the range 1 in 1.0001 to 1 in 10,000.

Annual Exceedance Probability Estimates:

Figures 3 and 4 show the resulting flood frequency estimates for the spring and summer annual peak floods obtained from the second approach. The annual peak discharge is plotted on a Log scale and AEPs are plotted on a z-variate scale (corresponding to a Normal probability distribution). In addition to 5

Monticello NGS - Spring 1,000,000 B el 2012 PMF estim te

-a--T9.

an---

--- nbrPes-Tmae - - . - - .,e.

0l vafion 93S El,-atlon 930 100,000 . .. '"'*

... ""h...

=-- -- ~ ~... :...........

10,000

v

- *Annual Exce dance Probability 1E I 1E-1 1E-2 E-3 1E-4 iE-5 1E-6 1E-7 1E-8 1 9 1,000 . I I I" I 1- 1

-4.0 -3.0 -2.0 -1.0 0.0 1.0 2.0 3.0 4.0 5.0 6.0 z-variate Figure 3. Spring Annual Peak Flood Frequency (approximate mean shown by black dashed line) 6

Monticello NGS - Summer 1,000,000 N

100,000 0

2 4) 4)

4) a.

10,000 1,000

-4.0 -310 -2.0 -1.0 0.0 1.0 2.0 3.0 4.0 5.0 6.0 z-variate Figure 4. Summer Annual Peak Flood Frequency (approximate mean shown by black dashed line) 7

2 providing median ( 5 0 th percentile) and approximate mean estimates, the 20% (4 0th and 6 0 th percentiles), 40% ( 3 0 th and 7 0 th percentiles), 60% (2 0 th and 80th percentiles), 80% ( 1 0 th and 9 0 th percentiles), and 90% (5 th and 9 5 th percentiles) confidence interval estimates are provided with linear extrapolation to smaller AEPs beyond 1 in 10,000. The site elevations of 917, 930 and 935 are shown by horizontal lines based on the NGVD 29 datum matching the datum used on the site drawings. Also the Harza and Bechtel PMF estimates are shown by horizontal lines corresponding to peak elevations for these events.

It is noted above that only single daily river stage observations have been recorded at site and therefore it was not possible to obtain maximum daily annual peak stages. Since the difference between mean daily and maximum daily peak stages decreases with smaller AEPs, it is likely that the flood frequency relationships are slightly steeper than shown. This effect would tend to make AEP estimates for extreme flows slightly conservative (i.e. slightly larger) than of this effect were removed.

Tables 1 and 2 contain numerical AEP estimates for spring and summer annual peak floods, respectively, for river stages 917, 930 and 935 ft. NGVD 29 for the median ( 5 0th percentile) and approximate mean, and for various confidence percentiles.

Table 1. Spring Annual Peak Flood Frequency Estimates Elevation 917 9 5 th I ge 1 8 4th 1 0 th I 7 0 th I 60 I Median 5 0e 5 th Approx Mean AEP Estimates 3.4E-02 2.5E-02 2.0E-02 1.7E-02 1.2E-02 8.9E-03 6.3E-03 5.9E-07 7.9E-03 i in Tyears 29 40 51 59 82 112 158 1,680,000 127 Elevation 930 95t' 90 84th 80' 70 60h Median 50 5t Approx Mean AEP Estimates 7.OE-06 7.6E-07 7.5E-08 1.7E-08 < 1E-9 < 1E-9 < 1E-9 < 1E-9 < 1E-9 1inTyears 143,0001 1,320,0001 13,400,0001 60,500,0001 >1E+9 I >1E+9 [ >1E+9 I >1E+9 [ >1E+9 I _Elevation 935 F-9 90o 84th 806 70& 60' Median 506' 51 Approx Mean AEP Estimates 4.3E-07 2.OE-08 < 1E-9 < 1E-9 < 1E-9 < 1E-9 < 1E-9 < 1E-9 < 1E-9 1 in T years 2,350,000 51,100,000 >1E+9 >1E+9 >1E+9 >1E+9 >1E+9 >1E+9 >1E+9 Table 2. Summer Annual Peak Flood Frequency Estimates

___Elevation 917 9Sth 90t 84th 80th 70 th 60d Median 50th 5th Approx Mean AEP Estimates 1.9E-02 1.3E-02 8.8E-03 7.2E-03 4.4E-03 2.8E-03 1.7E-03 8.5E-07 3.OE-03 1 in Tyears 52 79 113 140 225 358 590 1,180,000 328

_ _Elevation 930

_95_ go 9 0 84th 80eh 7 0 th 6 0e Median 506h 5 th Approx Mean AEP Estimates 2.1E-04 4.3E-05 8.8E-06 3.1E-06 2.4E-07 1.5E-08 < 1E-9 < 1E-9 1.6E-06 1in Tyears 4,7201 23,4001 11,00 324,000 4,160,000 65,100,000 >1E+9 >1E+9 641,000 I-_ Elevation 935 95h 90 o 841 80h 70 60'h Median 50' 5h Approx Mean AEP Estimates 6.1E-05 7.9E-06 9.7E-07 2.5E-07 8.9E-09 < 1E-9 < 1E-9 < 1E-9 2.OE-07 I in T years 16,500 127,000 1,030,000 4,000,000 112,000,000 >1E+9 >1E+9 >1E+9 1 4,930,000 2 The approximate mean estimates were obtained by weighting the various percentile estimates by their respective intervals of probability that each represents. For example, the 6 0 th percentile represents the interval between the mid-points of the 50 th - 6 0 th and 6 0 th - 7 0 th percetile intervals and hence is weighted by the difference between the percentiles associated with the mid-points of these two intervals, i.e. 0.65-0.55 = 0.10.

8

The following is a summary of the median estimates and ranges (Upper - 9 5th percentile and Lower - 5 th percentile) of the AEP estimates for the river stages of 917, 930 and 935 ft. NGVD 29 at the Monticello Nuclear Generating Plant for spring and summer annual peak floods and for the Harza and Bechtel PMF estimates. The April 8, 2013 spring estimates are shown in italics for comparison. The comparison shows that these estimate were conservative relative to those obtained using the second approach.

Spring Floods:

Elevation 917 ft. NGVD 29:

Upper (9 5 th): 3.4E-02 (1 in 29/year) 4.OE-02 (1 in 25/year)

" Median (5 0 th): 6.3E-03 (1 in 158 /year) 7.2E-03 (1 in 140/year)

Lower (5 th): 5.9E-07 (1 in 1,680,000 /year) 9.5E-04 (1 in 1,100/year)

Elevation 930 ft. NGVD 29:

Upper (95th): 7.OE-06 (1 in 143,000 /year) 1.6E-04 (1 in 6,300/year)

Median (5 0 th): < 1E-9 (1 in >1E+9 /year) 1.6E-05 (1 in 61,000/year)

Lower (5th): < 1E-9 (1 in >1E+9 /year) 2.2E-06 (1 in 460,000/year)

Elevation 935 ft. NGVD 29:

Upper (95th): 4.3E-07 (1 in 2,350,000 /year) 3.OE-05 (1 in 33,000/year)

Median (50th): <1E-9 (1 in >1E+9 /year) 3.1E-06 (1 in 330,000/year)

Lower (5th): <1E-9 (1 in >1E+9 /year) 3.4E-07 (1 in 2,900,000/year)

The Harza Spring PMF AEP estimates are shown below with the April 8, 2013 AEPs assigned to the Harza (spring) PMF shown in italics for comparison.

Upper (9 5th): 5.9E-08 (1 in 16,900,000) 1 in 10,000,000

Median (5 0 th): < 1E-9 (1 in >1E+9 /year) 1 in 1,000,000

" Lower (5 h): <1E-9 (1 in >1E+9 /year) 1 in 100,000 The estimates of AEPs assigned to the Harza PMF in the April 8, 2013 work are therefore confirmed to be conservative (i.e. larger than now estimated).

The AEP Bechtel spring PMF estimates are shown below:

" Upper ( 9 5 th): 1.8E-08 (1 in 54,500,000)

Median (5 0 th): < 1E-9 (1 in >1E+9 /year)

Lower (5 th): < 1E-9 (1 in >1E+9 /year) 9

Summer Floods:

Elevation 917 ft. NGVD 29:

Upper (9 5 th): 1.9.4E-02 (1 in 52 /year)

Median ( 5 0 th): 1.7E-03 (1 in 590 /year)

Lower (5 th): 8.5E-07 (1 in 1,180,000 /year)

Elevation 930 ft. NGVD 29:

Upper (9 5 th): 2.1E-04 (1 in 4,720 /year)

Median (5 0 th): < 1E-9 (1 in >1E+9 /year)

" Lower (5 th): <1E-9 (in >1E+9/year)

Elevation 935 ft. NGVD 29:

0 Upper (9 5 th): 6.1E-05 (1 in 16,500 /year) 0 Median (5 0th): < 1E-9 (1 in >1E+9 /year) 0 Lower (5 th): < 1E-9 (1 in >1E+9/year)

The AEP Bechtel summer PMF estimates are shown below:

Upper ( 9 5 th): 5.7E-04 (1 in 1,750/year)

Median (5 0th): 3.3E-08 (1 in 29,900,000 /year)

Lower (5 th): < 1E-9 (1 in >1E+9/year)

The second approach used to develop these revised AEP estimates is preferred to the initial approach used to develop our April 2013 estimates for the following reasons:

1) It separates the spring and summer flood events.

2) It relies on at-site date rather than a drainage-area weighted interpolation of AEP estimates at upstream and downstream Mississippi River USGS gages.

3) It does not rely on an assignment of an AEPtothe PMF.

A graphical comparison of the April 2013 and the current estimates is presented in Figure 5. It indicates that the April 2013 AEP estimates are likely overly conservative as a result of the assignments of the AEPs to the Harza PMF.

Figure 5 is similar to Figure 3 but includes the USNRC (2013) AEP estimates: 9.37E-05 for Elevation 930 and 2.72E-05 for Elevation 935 (based on 6.65E-05/year for Elevation 930-935). The NRC estimates exceed our current 9 5 th percentile estimates but are very similar to our April 2013 9 5 th percentile estimates, which were based on assigning an AEP of 1E-5 to the Harza PMF. According to USNRC (2013) their estimates are based on flood frequency estimates from the Monticello USAR and IPEEE but we are not clear about the origin of those estimates or the curve fitting approach that was used by the USNRC (2013). Therefore it is not possible to make a more informed comparison with our estimates; although it 10

Monticello NGS - Spring 1,000,000 100,000 z

9 9L aJ tU

.X 10,000 1,000

-4.0 -3.0 -2.0 -1.0 0.0 1.0 2.0 3.0 4.0 5.0 6.0 z-variate

- April 2013 Best Estimate - -April 2013 Upper Estimate - -April 2013 Lower Estimate

NRC Estimates - Approx Mean Figure 5. Comparison of the current Spring Annual Peak Flood Frequency with the April 2013 estimates and the NRC (2013) estimates 11

would appear that our estimates rely on more recent site-specific data than the NRC had available. In addition they use the Bulletin #17B flood frequency approach, which is the standard for flood frequency analysis in the US. We also used the improved EMA parameter estimation approach that will be included in the revision of the Bulletin #17B procedure that has been drafted.

However, we recommend that a Monte Carlo rainfall-runoff approach be considered to develop estimates of extreme flood frequencies with explicit consideration of uncertainties in the future. This approach can be expected to provide improved estimates based on the use of regional rainfall analysis, and a more physics-based representation of the rainfall-runoff (including snow melt) processes for extreme floods than is associated with the current extrapolation approach.

References Cohn, T. 2012. User Manual for Program PeakfqSA Flood-Frequency Analysis with the Expected Moments Algorithm DRAFT. September.

Flynn, K.M., Kirby, W.H., and Hummel, P.R., 2006, User's Manual for Program PeakFQ Annual Flood-Frequency Analysis Using Bulletin 17B Guidelines: U.S. Geological Survey, Techniques and Methods Book 4, Chapter B4; 42 pgs.

State of Minnesota. 1959. Hydrologic Atlas of Minnesota. Davison of Water, Department of Conservation, State of Minnesota.

Stedinger, J.R. V. Griffis, A. Veilleux, E. Martins, and T. Cohn. 2013. Extreme Flood Frequency Analysis:

Concepts, Philosophy and Strategies. Proceedings of the "Workshop on Probabilistic Flood Hazard Assessment (PFHA)" sponsored by the U.S. Nuclear Regulatory Commission's Offices of Nuclear Regulatory Research, Nuclear Reactor Regulation and New Reactors in cooperation with U.S. Department of Energy, Federal Energy Regulatory Commission, U.S. Army Corps of Engineers, Bureau of Reclamation and U.S. Geological Survey organized. Rockville, Maryland.

January 29 - 31.

USGS (US Geological Survey). 1982. Guidelines for Determining Flood Flow Frequency. Bulletin #17B, Hydrology Subcommittee, Interagency Advisory Committee on Water Data, Office of Water data Coordination.

USNRC (U.S. Nuclear Regulatory Commission). 2013. Monticello Nuclear Generating Plant, NRC Inspection Report 05000263/2013008; Preliminary Yellow Finding.

12

Enclosure 5 Monticello Nuclear Generating Plant "Stakeholder Outreach" I Page Follows

Stakeholder Outreach NSPM hosted an open house on Thursday, June 6, from 4 p.m.-8 p.m. to share information with its community neighbors on operations and preparedness to handle potential emergencies and how we would respond to flooding, earthquakes and other unforeseen challenges. The Site employed numerous methods to publicize the event:

personal, direct invitations to community leaders, a full page ad was purchased in weekly newspapers, a news release was distributed to local media and 14,000 postcards were mailed to neighbors in surrounding communities. The outreach event had full corporate support and the Xcel Energy Chairman, President and CEO, and the Chief Nuclear Officer attended, as well as numerous senior members of the corporate nuclear staff. The Monticello Site Vice President and Plant Manager were also joined by the site's senior leadership team at the event.

A total of 515 persons from Monticello and surrounding communities attended the event at the Monticello Training Center.

The key message presented to visitors was that safety and security at the NSPM nuclear generating plants are top priorities for Xcel Energy. Further, that we understand the NRC's increased scrutiny of safety and flood preparedness at the nation's nuclear power plants in the wake of events such as 9/11 and Fukushima Daiichi. The Monticello Flood Protection Strategy was identified and explained to demonstrate that the site is designed to withstand a hypothetical flood beyond anything reported in the Monticello area. The broad underlying key messages were reinforced and manifest in specific subject items such as: B.5.b Pump/Electrical Generator/Trailer, Portable Emergency Response Equipment, Backup Power Sources including description of backups to the backup (Battery Systems) and the continual focus on improving emergency preparedness capabilities.

Page 1 of 1

L-MT-13-062, Response to an Apparent Violation in NRC Inspection Report 05000263/2013008 (EA-13-096)

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References:

Enclosures:

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4.0 CONCLUSION

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L-MT-13-062, Response to an Apparent Violation in NRC Inspection Report 05000263/2013008 (EA-13-096)

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Enclosures:

Title:

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2.0 REFERENCES

4.0 CONCLUSION

1.0 INTRODUCTION

2.0 REFERENCES

Subject:

4.0 CONCLUSION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

5.0 CONCLUSION

6.0 REFERENCES

Purpose:

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