ML042860314

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Engineering Report, Risk Impact Assessment of Extending Containment Type a Test Interval, Attachment 4 to Bvy 04-77, Technical Specification Proposed Change No. 268 - One-time Integrated Leak Rate Test (ILRT) Interval Extension
ML042860314
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 09/24/2004
From: Favara J, Palionis M
Entergy Nuclear Northeast, Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation
References
BVY 04-77 VY-RPT-04-00002, Rev 1
Download: ML042860314 (145)
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ATTACHMENT 4 TO BVY 04-77 TECHNICAL SPECIFICATION PROPOSED CHANGE NO. 268 ONE-TIME INTEGRATED LEAK RATE TEST (ILRT) INTERVAL EXTENSION RISK IMPACT ASSESSMENT OF EXTENDING CONTAINMENT TYPE A TEST INTERVAL ENTERGY NUCLEAR OPERATIONS, INC.

VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271

I Engineering Report No. VY-RPT-04-00002 Rev. 1 Page 1 Of 70 AM, 7trgyT ENTERGY NUCLEAR NORTHEAST Engineering Report Cover Sheet Engineering Report

Title:

Risk Impact Assessment of Extending Containment Type A Test Interval Engineering Report Type:

New El Revision E Cancelled El Superceded El .

Applicable Site(s)

IP1 El IP2 0l IP3 LI JAF El PNPS [E VY Z Quality-Related: E Yes 0 No Prepared by: MarkiEPalionis I Favara Date: 09-24-04 Resosble Er Verified/

Reviewed by: John Bo Date:. os-24 Design Vefier/Reviester--

Approved by: Clem Yeh Date: 09-24-04 Supervisor Multiple Site Review Site Design Verifier/Reviewer Supervisor Date

RECORD OF REVISIONS Engineering Report No: VY-RPT-04-00002 Page 2 of 70 Revision No. Description of Change Reason For Change 0 Original report NA 1 Report revised to reflect changes in VY PSA model VY04R1 PSA base model. _

+

4-1- I

TABLE OF CONTENTS Description Page Executive Summary 6 Nomenclature 14 Definitions 16

1. Introduction 18 1.1 Purpose 18 1.2 Background 18
2. Evaluation 21 2.1 Method of Analysis 21 2.2 Assumptions 22 2.3 Data and Design Criteria 23 2.4 Internal Events Impact 25 2.4.1 Quantify Baseline Accident Classes Frequencies (Step 1) 25 2.4.2 Containment Leakage Rates (Step 2) 30 2.4.3 Baseline Population Dose Estimate (Step 3) 30 2.4.4 Baseline Population Dose Rate Estimate (Step 4) 33 2.4.5 Change in Probability of Detectable Leakage (Step 5) 34 2.4.6 Population Dose Rate for New ILRT Interval (Step 6) 38 2.4.7 Change in Population Dose Rate Due to New ILRT Interval (Step 7) 41 2.4.8 Change in LERF Due to New ILRT Interval (Step 8) 42 2.4.9 Impact on Conditional Containment Failure Probability (Step 9) 44 2.5 External Events Impact 45 2.6 Containment Liner Corrosion Risk Impact - 46
3. Summary of Results 56 3.1 Internal Events Impact 56 3.2 External Events Impact 57 3.3 Containment Liner Corrosion Risk Impact 57
4. Conclusions 63 4.1 Internal Events Impact 63 4.2 External Events Impact 63 4.3 Containment Liner Corrosion Risk Impact 64 5 References 69

TABLE OF CONTENTS (continued)

Appendix A External Event Assessment During an Extension of the ILRT Interval Appendix B Risk Impact of Containment Liner Corrosion During an Extension of the ILRT Interval Attachment A Vermont Yankee Risk Impact of Containment Liner Corrosion During an Extension of the ILRT Interval Results LIST OF FIGURES Page Figure 2-1 Acceptance Guidelines for Large Early Release Frequency 48

LIST OF TABLES Page Table ES-1 Internal Events Quantitative Results as a Function of ILRT Interval 10 Table ES-2 Internal and External Events Quantitative Results as a Function of ILRT Interval 11 Table ES-3 Liner Corrosion Impact Quantitative Results as a Function of ILRT Interval 12 Table 2-1 Internal Core Damage Frequency Contributions by Plant Damage States 49 Table 2-2 Summary of Vermont Yankee PSA LEVEL 2 Release Categories 50 Table 2-3 Summary of Vermont Yankee PSA LEVEL 2 Containment Failures 50 Table 2-4 VY Population Doses for Each PSA LEVEL 2 Release Category 51 Table 2-5 VY PSA Level 2 Release Frequencies As A Function of EPRI Accident Class 52 Table 2-6 VY Population Dose Estimates As A Function of EPRI Accident Class within 50-Mile Radius 53 Table 2-7 VY Dose Rate Estimates as a Function of EPRI Accident Class For Population within 50-Miles (Base Line 3 per 10 year ILRT) 54 Table 2-8 EPRI Accident Class Frequency as a Function of ILRT Interval 55 Table 2-9 Baseline Dose Rate Estimates By EPRI Accident Class for Population Within 50-Mile 55 Table 3-1 Summary of Risk Impact on Extending Type A ILRT Test Frequency

- Internal Events 59 Table 3-2 Summary of Risk Impact on Extending Type A ILRT Test Frequency

- Effect of Internal and External Events Risk on VY ILRT Risk Assessment 60 Table 3-3 Summary of Risk Impact on Extending Type A ILRT Test Frequency

- Impact of Containment Steel Liner Corrosion on VY ILRT Intervals 61 Table 3-4 Containment Steel Liner Corrosion Sensitivity Cases 62 Table 4-1 Quantitative Results as a Function of ILRT Interval - Internal Events 65 Table 4-2 Quantitative Results as a Function of ILRT Interval - Internal and External Events 66 Table 4-3 Quantitative Results as a Function of ILRT Interval - Liner Corrosion Impact 67

EXECUTIVE

SUMMARY

Revisions to 10CFR 50, Appendix J allow individual plants to extend Type A surveillance testing requirements from three-in-ten years to at least once per ten years. The revised Type A test frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage was less than normal containment leakage or 1.OLa. Vermont Yankee (VY) selected the revised requirements as its testing program [21]. VY's current ten-year Type A test frequency is due to be performed during the upcoming refueling outage 25 (RFO-25), scheduled for October 2005. Prior to the performance of that test, however, VY is seeking an extension of the test interval to fifteen years. A substantial cost savings will be realized and unnecessary personnel radiation exposure will be avoided by deferring the Type A test for an additional five years.

Cost savings have been estimated for this outage at approximately $300,000, which includes labor and equipment, but does not include critical path outage time needed to perform the test. In addition, this initiative directly supports site goals related to capacity factor and World Association of Nuclear Operators (WANO) performance by shortening the planned outage duration for RFO-25.

An evaluation was performed to assess the risk impact of extending the current containment Type A Integrated Leak Rate Test (ILRT) interval. In performing the risk assessment evaluation, the guidelines of NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J", the methodology used in EPRI TR-104285, ""Risk Assessment of Revised Containment Leak Rate Testing Intervals," and the NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-informed Decisions On Plant-Specific Changes to the Licensing Basis', were used.

The assessment also followed the guidance and additional information distributed by NEI in November 2001 to their Administrative Points of Contact regarding risk assessment evaluation of one-time extensions of containment ILRT intervals and the approach outlined in the Indian Point Unit Three Nuclear Power Plant ILRT extension submittal.

The risk assessment evaluation uses the current VY PRA internal events model that includes a Level 21 analysis of core damage scenarios and subsequent containment response resulting in various fission product release categories (including no release). The release category end states from the VY Level 2 model are used together with the population dose information to provide an estimate of the person-rem dose per reactor year associated with various scenarios. The change in plant risk is then evaluated based on the potential change in population dose rate (person-rem/ry), change in Large Early Release Frequency (LERF), and the change in conditional containment failure probability (CCFP).

The risk assessment evaluation examined VY's PRA plant specific accident sequences in which the containment integrity remains intact or the containment is impaired. Specifically, the following were considered:

  • Core damage sequences in which the containment remains intact initially and in the long term (EPRI Class 1 sequences).
  • Core damage sequences in which containment integrity is impaired due to a pre-existing isolation failure of plant components associated with Type A integrated leak rate testing. For example, containment liner breach. (EPRI Class 3 sequences).

Level 2 - the evaluation of containment response to severe accident challenges and quantification of the mechanisms, amounts, and probabilities of subsequent radioactive material releases from the containment.

  • Core damage sequences in which containment integrity is impaired due to pre-existing 'failure-to-seal' failure of plant components associated with either a Type B or Type C local leak rate testing (EPRI Classes 4 and 5 sequences).
  • Core damage sequences involving containment isolation failures due to failures-to-close of large containment isolation valves initiated by support system failures, or random or common cause valve failures (EPRI Class 2 sequences) and containment isolation failures of pathways left 'opened' following a plant post-maintenance test, or valve failing to close following a valve stroke test (EPRI Class 6 sequences).
  • Core damage sequences involving containment failure induced by severe accident phenomena (EPRI Class 7 sequences) or containment bypassed (EPRI Class 8 sequences).

The steps taken to perform this risk assessment evaluation are as follows:

1) Quantify the baseline risk in terms of frequency per reactor year for each of the eight containment release scenario types identified in the EPRI report.
2) Determine the containment leakage rates for applicable cases, 3a and 3b.
3) Develop the baseline population dose (person-rem) for the applicable EPRI classes.
4) Determine the population dose rate; also know as population dose risk (person-rem/Ry) by multiplying the dose calculated in step (3) by the associated frequency calculated in step (1).
5) Determine the change in probability of leakage detectable only by ILRT, and associated frequency for the new surveillance intervals of interest (Classes 3a and 3b).
6) Determine the population dose rate for the new surveillance intervals of interest.
7) Evaluate the risk impact (in terms of population dose rate and percentile change in population dose rate) for the interval extension cases.
8) Evaluate the risk impact in terms of LERF.
9) Evaluate the change in conditional containment failure probability.

The risk assessment evaluation of the one time ILRT extension is characterized by the following risk metrics: (as used in previously approved ILRT test interval extensions:

  • The potential change in population dose rate (person-rem/ry)
  • The change in conditional containment failure probability (CCFP).

The impact of these risk metrics associated with extending the Type A ILRT interval, are presented in Table ES-1.

The conclusion of the plant internal events risk associated with extending the Type A ILRT interval from ten to fifteen years is as follows.

1) The increase in risk on the total integrated plant risk as measured by person-rem/ry increases for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.004% (0.0005 person-rem/ry). This value can be considered to be a negligible increase in risk.
2) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 104 /ry and increases in LERF below 10' 7 /ry. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 1-in-10 years to 1-in-15 years is 2.87 x 10'9/ry. Since Regulatory Guide 1.174 defines very small changes in LERF as below 10'7 /ry, increasing the ILRT interval at VY from the currently allowed one-in-ten years to one-in-fifteen years is non-risk significant from a risk perspective.
3) The change in conditional containment failure probability (CCFP) is calculated to demonstrate the impact on 'defense-in-depth'. For the current ten-year ILRT interval, sequences involving no containment failure or small releases contribute 12.01 % to the overall plant risk. Alternatively stated, the contribution of sequences involving containment failure for the ten-year interval is 87.99%. These numbers are consisted with those documented in the VY PRA. For the proposed fifteen-year interval, the contribution of sequences involving containment failure increased to 88.05%. Therefore, ACCFP10.15 is found to be 0.06%. This signifies a very small increase and represents a negligible change in the Vermont Yankee containment defense-in-depth.

In addition to the internal events risk assessment evaluation, the impact associated with extending the Type A test frequency interval is further examined by considering external event hazard or potential containment liner corrosion. The purpose for these additional evaluations is to assess whether there are any unique insights or important quantitative information associated with the explicit consideration of external event hazard or containment liner corrosion in the risk assessment results.

The external event hazards or potential containment liner corrosion evaluation was found not to impact any of the above conclusions. The results from these cases are presented in Tables ES-2 and ES-3 respectively and summarized below.

Considerations of the combined internal events and external event hazards assessment during an extension of the ILRT Interval yielded the following conclusions:

1) Based on conservative methodologies in estimating the combined core damage frequency for internal events, seismic events, and fires events, the increase in LERF from extending the VY ILRT frequency from 1-in-1 0 years to 1-in-15 years is 3.71 x 10' 8/ry. This value falls within the 10' 7/ry criterion of Region IlIl, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines in NRC Regulatory Guide 1.174 [5]. Therefore, increasing the ILRT interval at VY from the currently allowed 1-in-10 years to 1-in-15 years is non-risk significant from a risk perspective.
2) The combined internal and external events increase in risk on the total integrated plant risk as measured by person-rem/ry increases for those accident sequences influenced by Type A testing, given the change from a 1-in-1 0 years test interval to a 1-in-1 5 years test interval, is found to be 0.004% (0.007 person-rem/ry). This value can be considered to be a negligible increase in risk.
3) The change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-15 years is 0.05%. A change in ACCFP of less than 1% is insignificant from a risk perspective.
4) Other salient results are summarized in Table ES-2. The key results to this risk assessment are those for the 10-year interval (current VY ILRT interval) and the 15-year interval (proposed change).

Recently, the NRC issued a series of Requests for Additional Information (RAls) in response to the one-time relief requests for the ILRT surveillance interval submitted by various licensees. The RAls requested a risk analysis on the potential increase in risk due to drywell/torus liner leakage, caused by age-related degradation mechanisms.

The risk analysis utilizes the referenced Calvert Cliffs Nuclear Power Plant assessment [20] to estimate the risk impact from containment liner corrosion during an extension of the ILRT interval. Consistent with the Calvert Cliffs analysis, the following issues were addressed:

  • Differences between the containment basemat and the drywell and torus liner
  • The historical drywelltorus steel shell flaw likelihood due to concealed corrosion
  • The impact of aging
  • The corrosion leakage dependency on containment pressure
  • The likelihood that visual inspections will be effective at detecting a flaw Considerations of risk impact of containment liner corrosion during an extension of the ILRT Interval yielded the following conclusions:
1) The impact of including age-adjusted corrosion effects in the ILRT assessment has minimal impact on plant risk and is therefore acceptable.
2) The change in LERF, taking into consideration the likelihood of a containment liner flaw due to age-adjusted corrosion is non-risk significant from a risk perspective. Specifically, extending the interval to 15 years from the current 10 years requirement is estimated to be about 3.19 x 10 9/ry. This is below the Regulatory Guide 1.174 [5] acceptance criteria threshold of 107 /ry.
3) The age-adjusted corrosion impact in dose increase is estimated to be 5.0 x 104 person-rem/ry or 0.0042% from the baseline ILRT 10 year's interval.
4) The age-adjusted corrosion impact on the conditional containment failure probability increase is estimated to be 0.06%.
5) A series of parametric sensitivity studies regarding potential age related corrosion effects on the containment steel liner also demonstrated minimal impact on plant risk.
6) Other salient results are summarized in Table ES-3.

--=Entergy REPORT No. VY-RPT-04-00002 I Revision 1 IPage I 10 I 01 l 70 l Table ES-1 Internal Events Quantitative Results as a Function of ILRT Interval Quantitative Results as a Function of ILRT Interval Current Proposed (1-per-tD year ILRT) (1-per-15 year ILRT)

EPIDose Accident Population Dose Accident Population Dose Clas Category Description (Person-Rem eny Rate (Person- Frequency Rate (Person-CasWithin 50 (preqeny) Rem I fly Within Fpreqeny) Rem I fly Within

_ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ m iles)(') pr y 50 miles) (p r ' )50 m iles) 1 No Containment Failure (2) 1.30 x 103 5.43 x 10' 7.06 x 10 4 5.12 x 10 ' 6.65 x 10,4 2 Containment Isolation System Failure 2.73 x 106 2.14 x 109 5.85 x 10-3 2.14 x 10.9 5.85 x 10i3 3a Small Pre-Existing Failures (2), (3) 1.30 x 104 5.73 x 10 8 7.46 x 10-4 8.60 x 10.8 1.12 x 103 3b Large Pre-Existing Failures 2 ) 3 ) 4.55 x 104 5.73 x l0o9 2.61 x 10 4 8.60 x 10 9 3.91 x 10'5 4 Type B Failures (LLRT) _--- --- ... ...

5 Type C Failures (LLRT) ---.--- ... _---

6 Other Containment Isolation System Failure ..----

7a Containment Failure Due to Severe Accident (a)(4) 2.73 x 106 3.22 x 10'7 8.79 x 101 3.22 x 10 ' 8.79 x 10.1 7b Containment Failure Due to Severe Accident (b)(4) 2.73 x 106 3.85 x 10.6 1.05 x 10+ 3.85 x 10-6 1.05 x 10+'

7c Containment Failure Due to Severe Accident (c)(4) 3.53 x 106 9.71 x 1O9 3.43 x 10o2 9.71 x 1O9 3.43 x 1 2 o 7d Containment Failure Due to Severe Accident (d)(4) 2.73 x 106 1.63 x 10 7 4.46 x 10.1 1.63 x 10-7 4.46 x 10' 8 Containment Bypass Accidents 2.96 x 106 5.32 x 108 1.57 x 10 1 5.32 x 10-8 1.57 x 10w' TOTALS: 5.00 x 106 129 011 5.00 x 10.6 12.0215 Increase in Dose Rate Increase in LERF Increase in CCFP (%)

Entergy REPORT No. VY-RPT-04-00002 I Revision 1 1 Page I 11 nt 70f Table ES-2 Internal and External Events Quantitative Results as a Function of ILRT Interval Quantitative Results as a Function of ILRT Interval Current Proposed (1-per-10 year ILRT) (1-per- 5 year ILRT)

EPIDose Accident Population Dose Accident Population Dose ClsE CtgryDsrito (Person-Rem RateI (Person- FrRate Ct DesrptionWithin 50 Frequency Rem Ry Within Frequency Rem / (Person-fly Within

_______________________ miles)(') (per ry) 50 miles) (per ry) 50 miles) 1 No Containment Failure (2) 1.30 x 103 7.40 x 104 9.62 x IO-' 6.99 x 104 9.08 x 10-3 2 Containment Isolation System Failure 8 2.73 x 106 2.90 x 10.8 7.93 x 10.2 2.90 x 10 7.93 x 10.2 3a Small Pre-Existing Failures (2)(3) 4 3 1.30 x 10 7.42 x 10-7 9.64 x 10- 1.11 x IO, 1.45 x 10.2 3b 4 Large Pre-Existing Failures (2).(3) 4.55 x 10 7.42 x 10.8 3.38 x IO 3 1.11 X10'7 5.06 x 10'3 4 Type B Failures (LLRT) . .. _.._..

5 Type C Failures (LLRT) .. .. .. _.._..

6 Other Containment Isolation System Failure 7a Containment Failure Due to Severe Accident (a)(4) 2.73 x 106 4.36 x 10.6 1.19 x 10' 4.36 x 10.6 1.19 x 10' 7b Containment Failure Due to Severe Accident (b) (4) 2.73 x 106 5.21 x IO" 1.42 x 102 5.21 x 10 5 1.42 x 1 2 7c Containment Failure Due to Severe Accident (c)(4) 3.53 x 106 1.31 x 10-7 4.64 x 10 1 1.31 x 10 7 4.64 x 10.1 7d Containment Failure Due to Severe Accident (d)(4) 2.73 x 106 2.21 x 10.6 6.06 x 100 2.21 x 10.6 6.06 x 100 8 Containment Bypass Accidents 2.96 x 106 7.20 x 10'7 2.13 x 100 7.20 x 10'7 2.13 x 100 5

TOTAL r77 x10'5 6 162 7R11 A677 x I nV 162 7R71 0.06 .

Increase in Dose Rate 0.0065 Increase in LERF Increase in CCFP (%)

Enterngy REPORT No. VY-RPT-04-00002 Revision I lPage l 12 l Of l 70 I Table ES-3 Liner Corrosion Impact Quantitative Results as a Function of ILRT Interval Quantitative Results as a Function of ILRT Interval Current Proposed (1-per-10 year ILRT) (1.per-15 year ILRT)

Dose Population Dose Population Dose (Person-Rem Accident Rate (Person- Accident Rate (Person-EPRI Within 50 Frequency Rem / Ry Within Frequency Rem / Ry Within Class Category Description miles)() l (per ry) 50 miles) (per ry) 50 miles) 1 No Containment Failure1 2) 1.30 x 103 5.43 x 10-7 7.06 x 10-4 5.11 x 10- 7 6.65 x 10-4 Containment Isolation System Failure 2.73 x 106 2.14 x I0- 5.85 x 10- 2.14 x 1i09 5.85 x 10-3a Small Pre-Existing Failures (2)(3) 1.30 x 104 5.73 x 1o08 7.46 x 104 8.60 x 1o-8 1.12 x 10-3 3b Large Pre-Existing Failures (2).(3) 4.55 x 104 5.97 x 109 2.72 x 10'4 9.16 x 10 9 4.17 x 10-4 4 Type B Failures (LLRT) 0.0 0.0 0.0 0.0 5 Type C Failures (LLRT) 0.0 0.0 0.0 0.0 6 Other Containment Isolation System Failure 0.0 0.0 0.0 0.0 7a Containment Failure Due to Severe Accident (a)(4) 2.73 x 106 3.22 x 1i07 8.79 x 10-1 3.22 x 10-7 8.79 x 10.1 7b Containment Failure Due to Severe Accident (b)(4) 2.73 x 106 3.85 x 1O06 1.05 x 101 3.85 x 10.6 1.05 x 101 7c Containment Failure Due to Severe Accident (c)(') 3.53 x 106 9.71 x 10 9 3.43 x 1O.2 9.71 x 10 9 3.43 x 10.2 8a Containment Failure Due to Severe Accident (d) (4) 2.73 x 106 1.63 x 10-7 4.46 x 10" 1.63 x 10 -7 4.46 x 101 8b Containment Bypass Accidents 2.96 x 106 5.32 x 10.8 1.57 x 10.1 5.32 x 10.8 1.57 x 10.1 TOTALS: 5.00 x 10-6 12.0211 5.00 x 1O.6 12.0216 Increase in Dose Rate Increase in LERF Increase in CCFP (%)

Notes to Tables ES-1, Es-2, and ES-3:

1) The population dose associated with the Technical Specification Leakage is based on the analysis provided in Reference [24].
2) Only EPRI classes 1, 3a, and 3b are affected by ILRT (Type A) interval changes.
3) Dose estimates for EPRI Class 3a and 3b, per the NEI Interim Guidance, are calculated as 10 times EPRI Class 1 dose and 35 times EPRI Class 1 dose, respectively.
4) EPRI Class 7, containment failure due to severe accident, was subdivided into three subgroups based on VY Level 2 containment failure modes for dose allocation purposes. Note that this EPRI class is not affected by ILRT interval changes.

Nomenclature APB Accident Progression Bin ATWS Anticipated Transient Without Scram CAPB Collapsed Accident Progression Bin CCIs Core-Concrete Interactions CCFP Conditional Containment Failure Probability CD Core Damage CDF Core Damage Frequency CDFM Conservative Deterministic Failure Margin CET Containment Event Tree CF Containment Failure DCH Direct Containment Heating DW Drywell EPRI Electrical Power Research Institute EVNTRE Event Progress Analysis Code PGA Peak Ground Acceleration HCLPF High Confidence Low Probability of Failure ILRT Integrated Leak Rate Testing IPE Individual Plant Examination IPEEE Individual Plant Examination for External Events ISLOCA Interface System Loss of Coolant Accident IP3 Indian Point Unit Three Nuclear Power Plant LERF Large Early Release Frequency LLRT Local Leak Rate Testing LOCA Loss of Coolant Accident NEI Nuclear Energy Institute

LE nteroyj REPORT No. VY-RPT-04-00002 Revision 1 Page l15 Of 70 Nomenclature (continued)

NRC United States Nuclear Regulatory Commission MFCR Mean Factional Contribution to Risk PDS Plant Damage State PRA Probabilistic Risk Analysis PSA Probabilistic Safety Assessment RAI Request for Additional Information RCS Reactor Coolant System RPV Reactor Pressure Vessel RF Refueling Outage SCDF Seismic Core Damage Frequency SMA Seismic Margin Assessment TS Technical Specifications VY Vermont Yankee Nuclear Power Plant WANO World Association of Nuclear Operations WW Wetwell

Entergyl REPORT No. VY-RPT-04-00002 Revision 1 l Page l 16 l Of I 7 l Definitions Accidentsequence - a representation in terms of an initiating event followed by a combination of system, function and operator failures or successes, of an accident that can lead to undesired consequences, with a specified end state (e.g., core damage or large early release). An accident sequence may contain many unique variations of events that are similar.

Core damage - uncovery and heatup of the reactor core to the point at which prolonged oxidation and severe fuel damage is anticipated and involving enough of the core to cause a significant release.

Core damage frequency- expected number of core damage events per unit of time.

End State - is the set of conditions at the end of an event sequence that characterizes the impact of the sequence on the plant or the environment. End states typically include: success states, core damage sequences, plant damage states for Level 1 sequences, and release categories for Level 2 sequences.

Event tree - a quantifiable, logical network that begins with an initiating event or condition and progresses through a series of branches that represent expected system or operator performance that either succeeds or fails and arrives at either a successful or failed end state.

InitiatingEvent - An initiating event is any event that perturbs the steady state operation of the plant, if operating, or the steady state operation of the decay heat removal systems during shutdown operations such that a transient is initiated in the plant. Initiating events trigger sequences of events that challenge the plant control and safety systems.

ISLOCA - a LOCA when a breach occurs in a system that interfaces with the RCS, where isolation between the breached system and the RCS fails. An ISLOCA is usually characterized by the over-pressurization of a low-pressure system when subjected to RCS pressure and can result in containment bypass Large early release - the rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of off-site emergency response and protective actions.

Large early release frequency - expected number of large early releases per unit of time.

Level 1- identification and quantification of the sequences of events leading to the onset of core damage.

Level 2 - evaluation of containment response to severe accident challenges and quantification of the mechanisms, amounts, and probabilities of subsequent radioactive material releases from the containment.

LOCAOC - a LOCA Outside of Containment (LOCAOC) is a breach occurs in a system that interfaces with the RCS and bypasses containment, with the potential to impact systems needed for mitigation of such events. LOCAOC includes breaks in high pressure piping, without the high-to-low pressure interface characteristic of ISLOCA.

Plant damage state - Plant damage states are collections of accident sequence end states according to plant conditions at the onset of severe core damage. The plant conditions considered are those that determine the capability of the containment to cope with a severe core damage accident. The plant damage states represent the interface between the Level 1 and Level 2 analyses.

~EiEntergy REPORT No. VY-RPT-04-00002 Revision 1 I Page 17 l Of l 70 Probability- is a numerical measure of a state of knowledge, a degree of belief, or a state of confidence about the outcome of an event.

Probabilistic risk assessment - a qualitative and quantitative assessment of the risk associated with plant operation and maintenance that is measured in terms of frequency of occurrence of risk metrics, such as core damage or a radioactive material release and its effects on the health of the public (also referred to as a probabilistic safety assessment, PSA).

Release category- radiological source term for a given accident sequence that consists of the release fractions for various radionuclide groups (presented as fractions of initial core inventory), and the timing, elevation, and energy of release. The factors addressed in the definition of the release categories include the response of the containment structure, timing, and mode of containment failure; timing, magnitude, and mix of any releases of radioactive material; thermal energy of release; and key factors affecting deposition and filtration of radionuclides. Release categories can be considered the end states of the Level 2 portion of a PSA.

Risk- encompasses what can happen (scenario), its likelihood (probability), and its level of damage (consequences).

Risk metrics - the quantitative value, obtained from a PRA analysis, used to evaluate the results of an application (e.g., CDF or LERF).

Severe accident - an accident that involves extensive core damage and fission product release into the reactor vessel and containment, with potential release to the environment.

Split Fraction - a unitless parameter (i.e., probability) used in quantifying an event tree. It represents the fraction of the time that each possible outcome, or branch, of a particular top event may be expected to occur. Split fractions are, in general, conditional on precursor events. At any branch point, the sum of all the split fractions representing possible outcomes should be unity. (Popular usage equates split fractions with the failure probability at any branch [a node] in the event tree.)

Vessel Breach - a failure of the reactor vessel occurring during core melt (e.g., at a penetration or due to thermal attack of the vessel bottom head or wall by molten core debris).

L'ntergy REPORT No. VY-RPT-04-00002 Revision 1 l Page l 18 l Of l 70 SECTION 1 INTRODUCTION 1.1 Purpose The purpose of this report is to provide supplemental information to support the proposed Vermont Yankee (VY) Technical Specification change of implementing a one-time extension of the containment Type A integrated leak rate test (ILRT) interval from ten years to fifteen years.

The risk assessment follows the guidelines from NEI 94-01 Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J" [1], the methodology used in EPRI TR-104285 'Risk Assessment of Revised Containment Leak Rate Testing Intervals" [2], NEI's 'Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals" [3], NEI 's "Additional Information for ILRT Extensions" (41, and the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a request for a change in a plant's licensing basis as outlined in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis' [5].

In addition, the results and findings from the VY Probabilistic Safety Analysis update [6] are used for this risk assessment report.

1.2 Background In October 26, 1995, the Nuclear Regulatory Commission (NRC) revised 10 CFR 50, Appendix J. The revision to Appendix J provided a performance based option, Option B "Performance-Based Requirements", for leakage-rate testing of light-water-cooled containments.

Under Option B, the Integrated Leak Rate Testing (ILRT) Type A surveillance testing requirements was extended from three-in-ten years to at least once per ten years. The revised Type A test frequency is based on an acceptable performance history defined as two consecutive'leriodic Type A tests at least 24 months apart in which the calculated performance leakage is less than the maximum allowable containment leakage limit of 1.OLa.

In accordance with the revised containment leakage-rate testing for Appendix J, VY selected the requirements under Option B as its testing program. VY's current ten-year Type A test is due to be performed during refueling outage twenty-five (RFO-25, scheduled for October 2005). However, VY seeks a one-time exemption based on the substantial cost savings of $300,000.00 from extending the test from the RF025 schedule to RFO27. In addition, this initiative directly supports site goals related to capacity factor and World Association of Nuclear Operators (WANO) performance by shortening planned outage duration for RF025.

The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, which was established in 1995 during development of the performance-based Option B to Appendix J (13.

Section 11.0 of NEI 94-01 states that NUREG-1 493, "Performance-Based Containment Leak-Test Program," [7] dated September 1995, provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J.

l - E'ntergy REPORT No. VY-RPT-04-00002 Revision 1 l Page I 19 I Of l 70 The NUREG-1493 [71 report examined the impact of containment leakage on public health and safety associated with a range of extended leakage rate test intervals. The NUREG analyzed both Boiling Water Reactors (Peach Bottom and Grand Gulf) and Pressurized Water Reactors (Surry, Sequoyah, and Zion).

For Peach Bottom, (a comparable Boiling Water Reactor plant to VY's), it was found that increasing the containment leak rates several orders of magnitude over the design basis (0.5 percent per day to 50 percent per day), results in a negligible increase in total population exposure. Therefore, extending the ILRT interval does not result in any significant increase in risk.

To supplement the NRC's rulemaking basis, NEI undertook another similar study. The results of that study are documented in EPRI research project report TR-1 04285 [2]. The EPRI Methodology [2] used a simplified risk model--PRA containment event trees (CETs). These CETs provide a risk framework for evaluating the effect of containment isolation failures affected by leakage testing requirements. The complexity of the CET models however is not necessary to evaluate the impact of containment isolation system failures. Therefore, a simplified risk model was developed to distinguish between those accident sequences that are affected by the status of the containment isolation system versus those that are a direct function of severe accident phenomena. The simplified risk model allowed for a smaller number of CET scenarios to be evaluated to determine the baseline risk as well as subsequent analysis to quantify risk effects of extending test intervals. The methodology regrouped core damage accident sequences reported in PRAs reviewed in the study into eight classifications to permit the appropriate delineation among containment isolation failure and containment failure due severe accident phenomena. The eight EPRI accident classes in the simplified model are:

1) Containment remains intact initially and in the long term. The release of fission products (and accident consequences) is determined by the maximum allowable containment leakage.
2) Core damage accident sequences in which containment integrity is impaired due independent (or random) containment isolation failures that include those accident s sequences in which the containment isolation system function fails during the accident progression (i.e., failures-to-close of large containment isolation valves initiated by support system failures, or random or common cause valve failures).
3) Core damage sequences in which containment integrity is impaired due to a pre-existing isolation failure of plant components associated with Type A integrated leak rate testing. For example, containment liner breach.
4) Core damage sequences in which containment integrity is impaired due to an independent (or random) pre-existing isolation failure-to-seal of plant components associated with Type B integrated leak rate testing. These are the Type B-tested components that have isolated but exhibit excessive leakage.
5) Core damage sequences in which containment integrity is impaired due to an independent (or random) pre-existing isolation failure-to-seal of plant components associated with Type C integrated leak rate testing.
6) Core damage sequences in which containment integrity is impaired due to containment isolation failures that include those leak paths not identified by containment leak rate tests. The type of failures considered under this Class includes those valves left open or valves that did not properly seal following test or maintenance activities.
7) Core damage sequences involving containment failure induced by severe accident phenomena.

Changes in ILRTs or LLRTs requirements do not impact these accidents.

8) Core damage sequences in which the containment is bypassed (either as an initial condition or induced by accident phenomena). Changes in ILRTs or LLRTs requirements do not impact these accidents.

Based on the improved methodology, NEI issued in November 2001 enhanced guidance Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals" (3], and "Additional Information for ILRT Extensions," [4] that builds on the EPRI TR-1 04285 [2], IP3 [8] and Crystal River submittal [9]

methodology and is intended to provide for more consistent submittals to the NRC.

The VY evaluation assesses the change in the predicted population dose rate associated with the interval extension. The assessment also evaluated the risk increase resulting from extending the ILRT interval in terms of Large Early Release Frequency (LERF), and the impact on Conditional Containment Failure Probability (CCFP). Regulatory Guide 1.174 (5] provides guidance for using PRA in risk-informed decisions for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 [5] defines very small changes in the risk acceptance guidelines as increases in Core Damage Frequency (CDF) of less than 1 per reactor year and increases in LERF of less than 10-7 per o6 reactor year. Since the Type A test does not impact CDF, the only relevant criterion is the change in LERF. Regulatory Guide 1.174 [5] also encourages the use of risk analysis techniques to help ensure and demonstrate that key risk metrics such as defense-in-depth philosophy, are satisfied. Based on that, the increase in the CCFP, which helps to ensure that the defense-in-depth philosophy is maintained, was evaluated.

SECTION 2 EVALUATION 2.1 Method of Analysis The VY risk assessment analysis uses the approach outlined in EPRI's TR-1 04285 [2], NEI's Interim Guidance [3], NEI 's "Additional Information for ILRT Extensions" [4], and the NRC regulatory guidance on the use of PRA findings and risk insights in support of a request for a change in a plant's licensing basis as outlined in Regulatory Guide 1.174 [5].

The EPRI TR-104285 methodology [2] involves a quantitative evaluation on the change in public risk of the affect of extending the ILRT and Local Leak Rate Test (LLRT) intervals. The EPRI TR-104285 study combined IPE Level 22 models with NUREG-1 150 "Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants" [10] Level 33 population dose models to perform the analysis. This study also used the approach of NUREG-1493 [7] in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

EPRI TR-104285 used a simplified containment event tree (CET) to provide a risk-based framework for evaluating the effects of containment isolation failures impacted by Appendix J penetrations testing requirements. The CET regrouped core damage accident sequences into eight accident classes of containment response. These eight accident classes are:

1) Containment intact and isolated
2) Containment isolation failures due to support system or active failures
3) Type A (ILRT) related containment isolation failures
4) Type B (LLRT) related containment isolation failures
5) Type C (LLRT) related containment isolation failures
6) Other penetration related containment isolation failures
7) Containment failure due to core damage accident phenomena
8) Containment bypass These eight accident classes allow the isolation failures modes and type of penetration analyzed to be correlated directly with Types A, B, and C test relaxation benefits. Each of the eight classes was categorized according to certain release characterization to determine the baseline incremental risk.

Vermont Yankee's risk assessment analysis uses the approach outlined in the NEI's Interim Guidance [3]

and NEI 's 'Additional Information for ILRT Extensions" [4]. The nine steps of the methodology are:

1) Quantify the baseline risk in terms of frequency per reactor year for each of the eight containment release scenario types identified in the EPRI report.
2) Determine the containment leakage rates for applicable cases, 3a and 3b.
3) Develop the baseline population dose (person-rem) for the applicable EPRI classes.

2 Level 2 - the evaluation of containment response to severe accident challenges and quantification of the mechanisms, amounts, and probabilities of subsequent radioactive material releases from the containment.

3 Level 3 -A measure of containment failure sequences leading to public health effects and their frequencies.

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4) Determine the population dose rate; also know as population dose risk (person-rem/ry) by multiplying the dose calculated in step (3) by the associated frequency calculated in step (1).
5) Determine the change in probability of leakage detectable only by ILRT, and associated frequency for the new surveillance intervals of interest (Classes 3a and 3b). Note that with increases in the ILRT surveillance interval, the size of the postulated leak path and the associated leakage rate are assumed not to change, however the probability of leakage detectable only by ILRT does increase.
6) Determine the population dose rate for the new surveillance intervals of interest.
7) Evaluate the risk impact (in terms of population dose rate and percentile change in population dose rate) for the interval extension cases.
8) Evaluate the risk impact in terms of LERF.
9) Evaluate the change in conditional containment failure probability.

The latest VY PSA Level 14 and Level 2 [6] were used to evaluate the change in population dose rate (person-rem/ry), change in Large Early Release Frequency (LERF), and the change in conditional containment failure probability. In order to assess the impact on offsite dose, a radiological analysis was performed to estimate the VY offsite dose.

The first seven steps of the methodology calculate the change in dose. The eighth step in the interim methodology calculates the change in LERF and compares it to the guidelines in Regulatory Guide 1.174

[5]. Because the change in ILRT test interval does not impact the CDF, the relevant criterion is LERF. The final step of NEI's interim methodology calculates the change in containment failure probability given the change of ILRT test interval from once-per-10 years to once-per-15 years.

2.2 Assumptions

1) The surveillance frequency for Type A testing in NEI 94-01 [1] is at least once per ten years based on an acceptable performance history. Based on the consecutive successful ILRTs performed in the early 1990's, the current ILRT interval for VY is once per ten years [17].
2) The VY Level 1 and Level 2 internal events PSA models provide representative results for the analysis [6].
3) Radionuclide release categories defined in this report are consistent with the EPRI TR-104285 methodology. [21
4) The EPRI methodology concluded that Severe Accident Phenomena and Bypass Classes accident sequences (e.g., drywell liner melt-through, ATWS, ISLOCA, and LOCAOC) contribution to population dose is unchanged by the proposed ILRT extension. These Classes are included for comparison purposes. As such, no changes introduced by this analysis will alter this conclusion.
5) The reliability of containment isolation valves to close in response to a containment isolation signal is not impacted by the change in ILRT frequency.

4Level 1 - Identification and quantification of the sequences of events leading to the onset of core damage.

ELtergyf REPORT No. VY-RPT-04-00002 lRevision 1 Page l 2 lO

6) The maximum containment leakage for Class 1 sequences is 1La [2]. (La is the Technical Specification maximum allowable containment leakage rate).
7) The maximum containment leakage for Class 3a sequences per the NEI Interim Guidance [3]

and previously approved methodology [8, 9] is 1OLa.

8) The maximum containment leakage for Class 3b sequences per the NEI Interim Guidance [3]

and previously approved methodology [8, 9] is 35La.

9) Class 3b release is categorized as LERF, based on the previously approved methodology [8, 9]

and NEI's interim methodology [3].

10) Containment leak rates greater than 2La but less than 35La indicate an impaired containment.

The leak rate is considered 'small' per the NEI Interim Guidance [3] and previously approved methodology [2, 8, and 9]. Furthermore, these releases have a break opening of greater than 0.5-inch but less than 2-inch diameter [8, 9].

11) Containment leak rates greater than 35La indicates a containment breach. This leak rate is considered 'large' per the NEI Interim Guidance [3] and previously approved methodology [8, 9].
12) Containment leak rates less than 2La indicates an intact containment. This leak rate is considered as 'negligible' per the NEI Interim Guidance [3] and previously approved methodology [8, 9].
13) EPRI accident Class 2 (Large Containment Isolation Failures) potential releases can be consider similar to a release associated with early drywell failure at high reactor pressure vessel (RPV) pressure.
14) Because EPRI Class 8 sequences are containment bypass sequences, potential releases are directly to the environment. Therefore, the containment structure will not impact the release magnitude.
15) An evaluation of the risk impact of the ILRT on shutdown risk is addressed using the generic results from EPRI TR-104285 [2] as augmented by NEI Interim Guidance [3, 4].

2.3 Data and Design Criteria

1) The VY Level 1 and 2 PSA updated model used as input to this analysis reflects the as built, as-operated plant. [6]
2) The CDF value, as reported in the VY PSA update is 5.03 x 104 /ry. [6]
3) The VY Level 2 PSA model is used to calculate the release frequencies for the accidents evaluated in this assessment. Table 2-1 summarizes the VY Level 1 PSA internal point-estimate frequency results by core damage plant damage states (PDS). (See Attachment A)
4) Table 2-2 summarizes the pertinent VY Level 2 PSA results in terms of release modes. The total release frequency is 4.42 x 104 /ry. Table 2-3 summarizes the correlation of the VY Level 2 PSA results for containment failure accident progression bins in terms of release magnitudes.

Enteray REPORT No. VY-RPT-04-00002 l Revision 1 l Page 24 I Of l 70

5) The random large containment isolation failure probability, from the VY PSA [6] is 4.26 104. This value was estimated using Top Event fault tree IS, split fraction ISILRT. Failure-to-close of large

(>2-inch diameter) containment isolation valves as well as failure of the containment isolation signal are included in this model. No credit is given for operator action to manually initiate containment isolation.

6) The conditional failure probability of having a small pre-existing containment leak is 0.027. This value is based on work performed in the IP3 ILRT submittal [8] and the NEI Interim Guidance [3].

From the IP3 submittal, the probability that a liner leak will be small made use of the data presented in NUREG-1493 [7]. The data reported in NUREG-1493 found that 23 of 144 tests had allowable leak rates in excess of 1.0La. However, of these 23 'failures' only 4 were found by an Type A ILRT, the others were found by Type B and C testing or errors in test alignments.

Therefore, the number of failures considered for 'small releases' are 4-of-1 44. Recent data collected by NEI and documented in the NEI Interim Guidance [3] found that an additional 38 ILRT have been performed since 1/1/95, with only one failure occurring. This indicates a failure probability of 5/182 (0.027) for a type A ILRT.

7) The conditional failure probability of having a large pre-existing containment leak is 0.0027. This value is derived from the NEI Interim Guidance [3]. It's based on the Jeffreys non-informative prior distributions for zero failures. The formula is as follows:

Number of Failures +1/2 Failure Probability =

Number of Tests + 1 The number of large failures is zero, so the probability is 0.5/183=0.0027.

8) The Vermont Yankee reactor power level used in this report is the power uprated adjusted value of 1912 MWth [13] (i.e., 120% of 1593 MWth).
9) The VY technical specification maximum allowed containment leakage is 1.5 volume%/day [14].
10) The VY containment free volume used in this report is 264,000 ft3 (drywell free volume of approximately 134,200 ft 3 and wetwell free volume of approximately 173,900 ft3 ) [15].

5 Applicaion of the Jeffreys non-informative prior is one of a number of statistical analysis approaches to estimating probabilities when no failures have been experienced. The approach was used inNUREG-1 150 and more recently in NUREG/CR-5750.

NUREG/CR-5750 is now the preferred source of initiating event data, which also Involves rare event approximations. The selected approach is more conservative than many other statistical approaches.

2.4 Internal Events Impact This section provides a step-by-step summary of the NEI guidance [3] as applied to the VY ILRT interval extension risk assessment. Each subsection addresses a step in the NEI guideline [3].

2.4.1 Quantify Baseline Accident Classes Frequencies (Step 1)

This step involves the quantification of the baseline frequencies for each of the EPRI TR-1 04285 accident classes [2].

Frequency of EPRI Class 1 Sequences. This group consists of all core damage accident progression sequences in which the containment remains isolated and intact (or containment leakage at or below maximum allowable Technical Specification leakage).

Consistent with NEI Interim Guidance [3], the frequency per reactor year for these sequences is calculated by subtracting the frequencies of EPRI Classes 3a and 3b from the sum of all severe accident progression sequence frequencies in which the containment is isolated and intact:

CLASS_1_FREQUENCY = NCF - CLASS_3a_FREQUENCY - CLASS_3bFREQUENCY Where:

CLASS_1_FREQUENCY = frequency of EPRI Class 1 given a 3-in-1 0 years ILRT interval NCF = frequency in which containment leakage is at or below maximum allowable Technical Specification leakage

= 6.06 x 10 7ry [Table 2-2]

CLASS_3aFREQUENCY = frequency of small pre-existing containment liner leakage

= 1.72 x 108/ry [See write-up below]

CLASS_3b_FREQUENCY = frequency of large pre-existing containment liner leakage

= 1.72 x 10-?/ry [See write-up below]

Therefore:

CLASS_1_FREQUENCY = 6.06 x 107 - 1.72 x 1c8 - 1.72 x 10 9 CLASS_1_FREQUENCY = 5.88 x 10-7/ry Frequency of EPRI Class 2 Sequences. This group consists of all core damage accident progression bins in which the containment isolation system function fails during the accident progression. These sequences are dominated by failure-to-close of large (>2-inch diameter) containment isolation valves [6].

The frequency per reactor year for these sequences is determined as follows:

CLASS_2_FREQUENCY = PROB large.c

CLASS_2_FREQUENCY = frequency of EPRI Class 2 given a 3-in-10 years ILRT interval PROB c arges = random large containment isolation failure probability (i.e. large valves)

= 4.26 x 104 [Section 2.3, input#5]

CDF = VY PSA Li core damage frequency = 5.03 x 10'6/ry [Section 2.3, input #2]

Therefore:

CLASS_2_FREQUENCY = [4.26 x 104] * [5.03 x 1016/ry]

CLASS_2_FREQUENCY = 2.14 x 10'/ry Frequency of EPRI Class 3a Sequences. This group consists of all core damage accident progression bins for which a small pre-existing leakage in the containment structure (i.e. containment liner) exists.

This type of failure is identifiable only from an ILRT and therefore, affected by a change in ILRT testing frequency.

Consistent with NEI Interim Guidance [4], the frequency per reactor year for this category is calculated as the pre-existing leakage probability multiplied by the residual CDF determined as the total CDF minus the CDF for those individual sequences that either may already (independently) cause a LERF:

CLASS_3aFREQUENCY = PROBdass 3a * [CDFBASE - CDFINDEP]

Where:

CLASS_3a_FREQUENCY = frequency of EPRI Class 3a given a 3-in-10 years ILRT interval PROBd,,s 3a = probability of small pre-existing containment liner leakage

= 0.027 [Section 2.3, input#6]

CDFBASE = VY PSA Li core damage frequency = 5.03 x 1086/ry [Section 2.3, input #2]

CDFINDEP = CDF for those individual sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences:

  • EPRI Class 2 = 2.14 x 10'9/ry
  • EPRI Class 7a = 3.22 x 10'7/ry
  • EPRI Class 7b = 3.85 x 10 4 /ry
  • EPRI Class 7c = 9.71 x 10'9/ry
  • EPRI Class 8a = 1.63 x 10'7/ry
  • EPRI Class 8b = 5.32 x 10'8/ry

= 4.40 x 1046/ry Therefore, CLASS_3a_FREQUENCY = 0.027 * [5.03 x 1046/ry - 4.40 x 10'6 /ry]

CLASS_3aFREQUENCY = 1.72 x 10'8 /ry

Frequency of EPRI Class 3b Sequences. This group consists of all core damage accident progression bins for which a large pre-existing leakage in the containment structure (i.e. containment liner) exists. This type of failure is identifiable only from an ILRT and therefore, affected by a change in ILRT testing frequency.

Consistent with NEI Interim Guidance [3], the frequency per reactor year for this category is calculated as:

CLASS_3b_FREQUENCY = PROBda.3b * [CDF3ASE - CDFINDEPI Where:

CLASS_3b_FREQUENCY = frequency of EPRI Class 3b given a 3-in-10 years ILRT interval PROBdS_3b = probability of large pre-existing containment liner leakage

= 0.0027 [Section 2.3, input #7]

CDFBASE = VY PSA Li core damage frequency = 5.03 x 10 6/ry [Section 2.3, input #2]

CDFINDEP = CDF for those individual sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences:

  • EPRI Class 2 = 2.14 x 10' 9/ry
  • EPRI Class 7a = 3.22 x 10'7/ry
  • EPRI Class 7b = 3.85 x 10I/ry
  • EPRI Class 7c = 9.71 x 10'9/ry
  • EPRI Class 8a = 1.63 x 10'7/ry
  • EPRI Class 8b = 5.32 x 10'8/ry

= 4.40 x 10' 6/ry Therefore, CLASS_3bFREQUENCY = 0.0027 * [5.03 x 10 4 /ry - 4.40 x 10 6/ry]

CLASS_3b_FREQUENCY = 1.72 x 10'9/ry Frequency of EPRI Class 4 Sequences. This group consists of all core damage accident progression sequences in which the containment isolation system function fails due to a pre-existing failure-to-seal of Type B test component(s). Consistent with NEI Interim Guidance [3], because these failures are detected by Type B tests and not by the Type A ILRT, this group is not evaluated further in this analysis.

FrequencV of EPRI Class 5 Sequences. This group consists of all core damage accident progression sequences in which the containment isolation system function fails due to a pre-existing failure-to-seal of Type C test component(s). Consistent with NEI Interim Guidance [3], because these failures are detected by Type C tests, this group is not evaluated any further.

Entergy REPORT No. VY-RPT-04-00002 Revision 1 lPage l 28 lOf l 70 Frequencv of EPRI Class 6 Sequences. This group consists of all core damage accident sequences in which the containment isolation function is failed due to 'other' pre-existing failure modes (e.g., pathways left open or misalignment of containment isolation vales following a test/maintenance evolution).

Consistent with NEI Interim Guidance [3], because these failures are detected by Type B or C tests, this group is not evaluated any further.

Frequency of EPRI Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (i.e. liner melt-through).

Consistent with NEI Interim Guidance [3], the frequency per reactor year for this class is based on the plant Level 2 PSA results [6].

The EPRI Class 7 is sub-divided in this report to reflect the sub-division of the VY Level 2 severe accident phenomena results into those threats posed by energetic phenomena and those related to direct contact of core material with the steel drywell shell. The following sub-classes are defined:

  • Class 7a: failure of overall integrity of the containment as a result of postulated energetic phenomena

[25], including:

1. Direct containment heating. This phenomena is caused by dispersal of core debris in the containment atmosphere during RPV failures from high pressure.
2. Missiles generated from structural debris.
3. In-vessel or ex-vessel steam explosions.
4. Vapor suppression failure. This could result from high pressure RPV failures (e.g., due to stuck-open drywell-to-wetwell vacuum breakers, or due to containment flooding which covers the vacuum breakers with water). Vapor suppression failure could also result from SRV discharge to the suppression pool while at elevated pool water temperature or low pool water level.
5. Recriticality following melt/relocation of core material.

Containment failures resulting from these phenomena are assumed to be large, and to be located at the drywell head. Choice of this location is conservative, since the release from containment will be directly to the refueling floor, bypassing the lower areas of the reactor building.

  • Class 7b: failure of the containment shell as a result of direct contact with core material [25].

The Mark I drywell steel shell provides the primary containment boundary. The drywell steel shell interfaces directly with the concrete floor of the drywell. The drywell floor is also the location where a substantial fraction of the core debris may be deposited if core damage cannot be arrested in-vessel and the RPV is subsequently breached. Without substantial water injection to the containment during the core melt progression, the steel shell may fail from extremely high local temperature due to the direct contact/proximity of core material. Top Event Si failure implies that a large hole (assumed to be about 2 ft2 ) is opened in the drywell shell at the elevation where it contact the pedestal floor.

The probability of shell meltthrough by molten core material depends on the following factors:

1. The mode and timing of RPV breach.
2. The temperature and quantity of debris released from the RPV.
3. Whether water is present in the pedestal region of the drywell at the time of RPV failure (and afterwards), and how effective this water is in cooling the debris and the drywell shell.
  • Class 7c: failure of the containment resulting from loss of containment heat removal capability [251.

Some form of containment heat removal is assumed to be required in order to achieve a safe, stable state. The Vermont Yankee Mark I containment system is provided with significant heat removal capacity and heat management capabilities. The management of heat in the containment prior to, during, and following a severe core damage event directly affects containment response. The Vermont Yankee containment heat capacity can be classified as both active and passive. The passive capacities include the suppression pool and the containment structure. The active heat management capabilities include the RHR system and venting.

Based on the VY Level 2 PSA results summarized in Reference [6], the frequency of Category 7a is 3.22 x 10 7/ry.

Based on the VY Level 2 PSA results summarized in Reference [6], the frequency of Category 7b is 3.85 x 1 0 6 /ry.

Based on the VY Level 2 PSA results summarized in Reference [6], the frequency of Category 7c is 9.71 x 10'9/ry.

Frequency of EPRI Class 8 Sequences. This group consists of all core damage accident progression bins in which the accident is initiated by a containment bypass scenario. Consistent with NEI Interim Guidance [3], the frequency per reactor year for this class is based on the plant Level 2 PSA results [6].

The EPRI Class 8 is sub-divided in this report to reflect the sub-division into those created by ATWS with high power oscillations, and those due to Interfacing Systems LOCAs. The following sub-classes are defined:

  • Class 8a: Containment bypass due to ATWS with high power oscillations (i.e., Accident Class IVA and IVL events). It is assumed that containment failure will be induced by hydrodynamic loads, and that the failure location will be located in the wetwell.
  • Class 8b: Containment bypass due to Interfacing Systems LOCAs (i.e., Accident Class V events).

Class V sequences where containment bypass and core melt occur are assumed to result in a High-RB release. Note that no credit for Reactor Building retention is taken for these sequences.

Based on the VY Level 2 PSA results summarized in Reference [6], the frequency of Category 8a is 1.63 x 107/ry.

Based on the VY Level 2 PSA results summarized in Reference [6], the frequency of Category 8b is 5.32 x 10 8 /ry.

Note: for this class the maximum release is not based on the maximum allowable containment leakage, because the releases are released directly to the environment. Therefore, the containment structure will not impact the release magnitude.

The EPRI TR-1 04285 Class frequencies that result in radionuclide releases to the public are derived in accordance with NEI Interim Guidance [3]. The EPRI TR-104285 class accident sequence frequency results are summarized in Table 2-5.

2.4.2 Containment Leakage Rates (Step 2)

This step defines the containment leakage rates for EPRI accident Classes 3a and 3b. As defined in Step 1, accident Class 3a and 3b are plant accidents with pre-existing containment leakage pathways (designated as 'small' and large") that are identifiable only when performing a Type A ILRT.

The NEI Interim Guidance 13] recommends containment leakage rates of 1OLa and 35La for accident Classes 3a and 3b, respectively. These values are consistent with previous ILRT frequency extension submittal applications [8]. La is the plant Technical Specification maximum allowable containment leak rate; for VY La is 0.8% of containment air weight per day (per VY Technical Specification).

By definition, and per the NEI Interim Guidance [3] and previously approved methodology [8] the containment leakage rate for Class 1 (i.e., accidents with containment leakage at or below maximum allowable Technical Specification leakage) is 1 La.

2.4.3 Baseline Population Dose Estimate (Step 3)

This step estimates the baseline population dose (person-rem) for each of the EPRI TR-1 04285 accident classes [3]. The NEI Interim Guidance [4] recommends two options for calculating population dose for the EPRI accident classes:

  • Use of NUREG-i150 dose calculations [10]
  • Use of plant-specific dose calculations Because the Vermont Yankee has a Level 3 PSA [24] and associated plant-specific dose, this risk assessment uses plant specific dose results.

The Vermont Yankee PSA offsite consequences are calculated by the MACCS2 consequence model

[24]. The principal phenomena analyzed are atmospheric transport of radionuclides, mitigative actions (i.e., evacuation, condemnation of contaminated crops and milk) based on dose projection, dose accumulation by a number of pathways, including food and water ingestion and economic costs. Input for the Level 3 analysis includes Vermont Yankee core radionuclide inventory, source terms from the Level 2 (containment performance analysis) model, site metrological data, projected population distribution (within 50-mile radius) for the year 2025, emergency response evacuation modeling and economic data.

The Vermont Yankee consequence analysis looks at the source term for fourteen collapsed release categories (Table 2-4). These categories represent the source term for each of the one-hundred-seventy-three different containment release modes associated with endstates of the Vermont Yankee containment event tree (Section 4.5 of Reference 6).

The MACCS2 code was used to estimate the consequences in terms of population dose within 50-miles and offsite economic cost. The Vermont Yankee Level 3 PSA MACCS2 population dose results are presented in Table 2-6. (Use of dose results for the 50-mile radius around the plant, as a figure of merit in this risk evaluation is consistent with NUREG-1 150 [10], past ILRT [8 &14] frequency extension submittals, and the NEI Interim Guidance. [4 & 5]).

The Vermont Yankee populations dose information presented in Table 2-6 when combined with the preceding information on the EPRI TR-1 04285 class accident sequence frequency results (Table 2-5),

provides the basis for the assignment of population dose for each EPRI accident category.

Population Dose for EPRI Class 1.

The dose for the 'no containment failure" EPRI class 1 sequences is based on PSA Level 2 Release Category"No Containment Failure (NCF)" (Table 2-4). Therefore, CLASS_1_DOSE = 1.30 x 101 person-sv *100 person-rem [Table 2-6]

1person-sv

= 1.30 x 103 person-rem Population Dose for EPRI Class 2.

The 50-miles population dose for the EPRI accident Class 2 (Large Containment Isolation Failures, failure-to-close) is based on the Vermont Yankee PSA Level 2 Release Category "EARLY HIGH" (Table 2-4), which represents the bounding release category for this class. Therefore, CLASS_2_DOSE = 2.73 x 10 4 person-sv

  • 100 person-rem [Table 2-6]

1person-sv

= 2.73 x 1o6 person-rem Population Dose for EPRI Class 3.

The 50-miles population dose for the EPRI accident Class 3a (Small Isolation Failures-Liner breach) and accident Class 3b (Large Isolation Failures-Liner breach), per the NEI Interim Guidance [4], are taken as factors of 10La and 35La [4, 8], respectively, times the population dose of EPRI accident Class 1.

Therefore, CLASS_3a-DOSE = 10

  • CLASS1 _DOSE CLASS_3bDOSE = 35
  • CLASS_1_DOSE CLASS_3aDOSE = 10
  • 1.30 x 103 person-rem CLASS_3b_DOSE = 35
  • 1.30 x 10 3 person-rem CLASS_3aDOSE = -1.30x lo' person-rem CLASS_3bDOSE = 4.55 x 104 person-rem Population Dose for EPRI Classes 4, 5 &6.

Per the NEI Interim Guidance [4], EPRI accident Classes 4 (Small Isolation Failure - failure-to-seal, Type B test), 5 (Small Isolation Failure - failure-to-seal, Type C test), and 6 (Containment Isolation Failures, dependent failures, personnel errors) are not affected by ILRT frequency and are not analyzed as part of this risk assessment. Therefore no selections of population does estimates are made for these accident classes.

Population Dose for EPRI Class 7a.

The 50-miles population dose for the EPRI accident Class 7a (failure of the containment as a result of postulated energetic phenomena) is based on the Vermont Yankee collapsed PSA Level 2 Release Category "EARLY HIGH" (Table 2-4) , which represents the bounding release category for this class.

Therefore,

CLASS_7aDOSE = 2.73 x 10 4 person-sv

  • 100 person-rem [Table 2-6]

1person-sv

= 2.73 x 106 person-rem Population Dose for EPRI Class 7b.

The 50-miles population dose for the EPRI accident Class 7b (failure of the containment as a result of the threat posed by direct contact with core material) is based on the Vermont Yankee PSA Level 2 Release Category 'EARLY HIGH" (Table 2-4) as the bounding release category for this class. Therefore, CLASS_7bDOSE = 2.73 x 10 4 person-sv

  • 100 person-rem [Table 2-6]

1person-sv

= 2.73 x 10 person-rem Population Dose for EPRI Class 7c.

The 50-miles population dose for the EPRI accident Class 7c (failure of the containment as a result of the loss of containment heat removal capability) is based on the Vermont Yankee PSA Level 2 Release Category 'INTERMEDIATE HIGH" (Table 2-4) , which represents the bounding release category for this class. Therefore, CLASS_7cDOSE = 3.53 x 10 4 person-sv

  • 100 person-rem [Table 2-6]

1person-sv

= 3.53 x 106 person-rem Population Dose for EPRI Class 8a.

The 50-miles population dose for the EPRI accident Class 8a (containment bypass due to ATWS with high power oscillations) is based on the Vermont Yankee PSA Level 2 Release Category 'EARLY HIGH" (Table 2-4), which represents the bounding release category for this class. Therefore, ---

CLASS_8a_DOSE = 2.73 x 104 person-sv *100 person-rem [Table 2-6]

1person-sv

= 2.73 x 106 person-rem Population Dose for EPRI Class 8b.

The 50-miles population dose for the EPRI accident Class 8b (containment bypass due to Interfacing Systems LOCAs) is based on the Vermont Yankee PSA Level 2 Release Category "Containment Bypass (V-Seq)" (Table 2-4) , which represents the bounding release category for this class. Therefore, CLASS_8bDOSE = 2.96 x 104 person-sv

  • 100 person-rem [Table 2-6]

1person-sv

= 2.96 x 10 person-rem

Using the preceding information, the population dose for the 50-mile radius surrounding the Vermont Yankee is summarized in Table 2-6. (Note: the use of dose results for the 50-mile radius around the plant as a 'figure of merit' in the risk evaluation is consistent with past ILRT frequency extension submittals, and the NEI Interim Guidance [4]).

2.4.4 Baseline Population Dose Rate Estimate (Step 4)

This step calculates the baseline does rates for each of the eight EPRI's accident classes. The calculation is performed by multiplying the dose calculated in Step 3 (Table 2-6) by the associated frequency calculated in Step 1 (Table 2-5). Since the conditional containment pre-existing leakage probabilities for EPRI accident classes' 3a and 3b are based on a 3-per-10 year ILRT frequency, the calculated baseline results reflect a 3-per-10 year ILRT surveillance frequency.

CLASS_1`_DOSERATE = CLASS_1_DOSE

  • CLASS_1_FREQUENCY CLASS_2_DOSERATE = CLASS_2_DOSE
  • CLASS_2_FREQUENCY CLASS_3aDOSERATE = CLASS_3aDOSE
  • CLASS_3a_FREQUENCY CLASS_3bDOS ERATE = CLASS_3b_DOSE
  • CLASS_3b_FREQUENCY CLASS_7a-DOSERATE = CLASS_7aDOSE
  • CLASS_7aFREQUENCY CLASS_7bDOSERATE = CLASS_7bDOSE
  • CLASS_7bFREQUENCY CLASS-7c-DOSERATE = CLASS_7cDOSE
  • CLASS-7c FREQUENCY CLASS_8a-DOSERATE = CLASS_8a_DOSE
  • CLASS_8aFREQUENCY CLASS_8b_DOSERATE = CLASS_8bDOSE
  • CLASS_8bFREQUENCY Where:

CLASS_1_DOSERATE = EPRI accident Class 1 dose rate given a 3-in-10 years ILRT interval CLASS_2_DOS ERATE = EPRI accident Class 2 dose rate given a 3-in-i 0 years ILRT interval CLASS_3aDOSERATE = EPRI accident Class 3a dose rate given a 3-in-10 years ILRT interval CLASS_3b_DOSERATE = EPRI accident Class 3b dose rate given a 3-in-10 years ILRT interval CLASS_7aDOSERATE = EPRI accident Class 7a dose rate given a 3-in-10 years ILRT interval CLASS_7bDOSERATE = EPRI accident Class 7b dose rate given a 3-in-10 years ILRT interval CLASS_7cDOSERATE = EPRI accident Class 7c dose rate given a 3-in-10 years ILRT interval CLASSBaDOSERATE = EPRI accident Class 8a dose rate given a 3-in-10 years ILRT interval CLASS_8b_DOSERATE = EPRI accident Class 8b dose rate given a 3-in-10 years ILRT interval CLASS_1_DOSE = EPRI accident Class 1 dose = 1.30 x lo, (person-rem) [Table 2-6]

CLASS_2_DOSE = EPRI accident Class 2 dose = 2.73 x10 6 (person-rem) [Table 2-6]

CLASS_3aDOSE = EPRI accident Class 3a dose = 1.30 X10 4 (person-rem) [Table 2-6]

CLASS_3bDOSE = EPRI accident Class 3b dose = 4.55 x10 4 (person-rem) [Table 2-6]

CLASS_7a_DOSE = EPRI accident Class 7a dose = 2.73 x10 6 (person-rem) [Table 2-61 CLASS_7b_DOSE = EPRI accident Class 7b dose = 2.73 x10 6 (person-rem) [Table 2-6]

CLASS_7cDOSE = EPRI accident Class 7c dose = 3.53 x10 (person-rem) [Table 2-6]

CLASS_8a_DOSE = EPRI accident Class 8a dose = 2.73 x10 6 (person-rem) [Table 2-6]

CLASS_8b_DOSE = EPRI accident Class 8b dose = 2.96 X10 6 (person-rem) [Table 2-6]

CLASS_1_FREQUENCY = frequency of EPRI accident Class 1 given a 3-in-10 years ILRT interval

= 5.88 x 10 /ry [Table 2-5]

CLASS_2_FREQUENCY = frequency of EPRI accident Class 2 given a 3-in-10 years ILRT interval

= 2.14 x 10- 9/ry [Table 2-5]

CLASS_3a_FREQUENCY = frequency of EPRI accident Class 3a given a 3-in-10 years ILRT interval

= 1.72 x 10o8/ry [Table 2-5]

CLASS_3bFREQUENCY = frequency of EPRI accident Class 3b given a 3-in-10 years ILRT interval

= 1.72 x 109/ry [Table 2-5]

CLASS_7aFREQUENCY = frequency of EPRI accident Class 7a given a 3-in-10 years ILRT interval

= 3.22 x 10O 7/ry [Table 2-5]

CLASS_7bFREQUENCY = frequency of EPRI accident Class 7b given a 3-in-10 years ILRT interval

= 3.85 x 104 /ry [Table 2-5]

CLASS_7c_FREQUENCY = frequency of EPRI accident Class 7c given a 3-in-10 years ILRT interval

= 9.71 x 10O9/ry [Table 2-5]

CLASS_8a_FREQUENCY = frequency of EPRI accident Class 8a given a 3-in-1 0 years ILRT interval

= 1.63 x 10 7/ry [Table 2-5]

CLASS_8b_FREQUENCY = frequency of EPRI accident Class 8b given a 3-in-10 years ILRT interval

= 5.32 x 104 /ry [Table 2-51 Therefore, CLASS_1_DOSERATE = 1.30 x10 3 5.88 x 10-7 = 7.64 x 1 O (person-rem/ry)

CLASS_2_DOSERATE = 2.73 x10 6

  • 2.14 x 109 = 5.85 x10 3 (person-rem/ry)

CLASS_3aDOSERATE = 1.30 x10 4 1.72 x 1O.8 = 2.24 x10 - (person-rem/ry)

CLASS_3b_DOSERATE = 4.55 x104 1.72 x 1 0.9 = 7.83 x 10.5 (person-rem/ry)

CLASS-7aDOSERATE = 2.73 x10 6 3.22 x 10O7 = 8.79 x10 1 (person-rem/ry)

CLASS-7b-DOSERATE = 2.73 x106 3.85 x 104 = 1.05 x10 .' (person-rem/ry)

CLASS_7cDOSERATE = 3.53 x10 6 9.71 x 1iO9 = 3.43 x 10 2 (person-rem/ry)

CLASS_8a_DOS ERATE = 2.73 x10 6

  • 1.63 x I0' = 4.46 x1IO1 (person-rem/ry)

CLASS_8bDOSERATE = 2.96 x106

  • 5.32x 108 = 1.57 x10 1 (person-rem/ry)

Table 2-7-summarizes the resulting baseline population dose rates by EPRI accident class., -

2.4.5 Change in Probability of Detectable Leakage (Step 5)

This step calculates the change in probability of leakage detectable only by ILRT, and associated frequency for the new surveillance intervals of interest. Note that with increases in the ILRT surveillance interval, the size of the postulated leak path and the associated leakage rate are assumed not to change, however the probability of leakage detectable only by ILRT does increase.

According to NUREG-1 493 [2] and per the NEI Interim Guidance [4], the calculation of the change in the probability of a pre-existing ILRT-detectable containment leakage is based on the relationship that relaxation of the ILRT interval results in increasing the average time that a pre-existing leak would exist undetected. Specifically, the relaxation of the Type A ILRT interval from 3-in-10 years to 1-in-10 years will increase the average time that a leak detectable only by an ILRT goes undetected from 18 to 60 months6 ,

a factor of 3.333 increase (60/18). Therefore, the change in probability of leakage due to the ILRT

~'En tergy REPORT No. VY-RPT-04-00002 I Revision 1 lPage l 35 l Of l 70 l interval extension is calculated by applying a multiplier factor determined by the ratio of the average times of undetection for the two ILRT interval cases.

From Section 2.3 "Input and Design Criteria", the calculated pre-existing ILRT detectable leakage probabilities based on 3 in-10 years ILRT frequency is 0.027 for small pre-existing leakage (EPRI accident class 3a) and 0.0027 for large pre-existing leakage (EPRI accident class 3b).

Since February 1998, the Vermont Yankee plant has been operating under a 1-in-1 0 years ILRT testing frequency consistent with the performance-based Option B of 10 CFR Part 50, Appendix J [14]. As a result, the baseline leakage probabilities, (which are based on a 3-in-10 years ILRT frequency) must be revised to reflect the current 1-in-10 years VY ILRT testing frequency. This is performed as follows:

PROBass_3a_10 = PROBdass_3a * [ SURTEST 1 o]

PROBdass-3b-10 = PROBdass_3b * [ SURTESTlo]

18 Where:

PROBcIass-3a1o = probability of small pre-existing containment liner leakage given a 1-in-1 0 years ILRT frequency.

PROBdassa_3 = probability of small pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.027 [Section 2.3, input#8]

PROBd1San3b = probability of large pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.0027 [Section 2.3, input #9]

SURTEST 10 = surveillance interval of interestxnonths/2 = 10 years*12months/2 = 60 months year Therefore, PROBdass _a_10 = 0.027 * [ 60 = 0.09 18 PROBdass_3bi1o = 0.0027 [ 60 = 0.009 18 I

Similarly, the pre-existing ILRT detectable leakage probabilities for the i-in-15 years ILRT frequency being analyzed by VY are calculated as follows:

PROBdass 3a_15 = PROBdass_3a SURTEST 1 5 18

PROBdasS3b15 = PROBdas3b [ SURTEST1 5 ]

L 18 Where:

PROBciass 3 ai15 = probability of small pre-existing containment liner leakage given a 1-in-15 years ILRT frequency.

PROBdOss_3a = probability of small pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.027 [Section 2.3, input#6]

PROBdass_3b = probability of large pre-existing containment liner leakage given a 3-in-1 0 years ILRT frequency = 0.0027 [Section 2.3, input #7]

SURTEST1S = surveillance interval of interest, months/2 = 15 years*12months/2 = 90 months year Therefore, PROBdass 3 a 15 = 0.027 * [ 90 ] 0.135 18 PROBdass_3b 15 = 0.0027 *[ 90 ] = 0.0135 18 Given the above revised leakage probabilities, the frequencies of the EPRI accident classes calculated in Step 1, also needs to be revised to reflect the increase change in leakage probabilities.

As previously stated, Type A tests impact only Class 1 and Class 3 sequences. Therefore, EPRI accident Class 1 frequency changes are calculated similar to Step 1, and the rest of EPRI's Classes; 2, 7 and 8 remain the same.

Revised Frequency of EPRI Class 3a Sequences. Consistent with NEI Interim Guidance [4], the frequency per reactor year for this category is calculated as:

CLASS_3aFREQUENCY 10 = PROBdass- 3a lo * [CDFBASE - CDFINDEP]

CLASS_3aFREQUENCY1s = PROBciass_3aai 5 * [CDFBASE - CDFINDEP]

Where:

CLASS_3aFREQUENCY ,o = frequency of small pre-existing containment liner leakage given a 1-in-10 years ILRT interval CLASS_3a_FREQUENCY 15 = frequency of small pre-existing containment liner leakage given a 1-in-15 years ILRT interval PROBd=s3a 10 = probability of small pre-existing containment liner leakage given a 1-in-10 years ILRT frequency = 0.09 [See above write-up]

-AM EEEntergyj REPORT No. VY-RPT-04-00002 Revision 1I Page 7 Of I70 PROBdass 3aljs = probability of small pre-existing containment liner leakage given a 1-in-15 years ILRT frequency = 0.135 [See above write-up]

CDFBASE = VY PSA Li core damage frequency = 5.03 x 106/ry [See step 1 write-up]

CDFINDEP = CDF for those individual sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences

= 4.40x 106 /ry [See step 1 write-up]

Therefore, CLASS_3aFREQUENCY 10 = 0.09 * [5.03 x 106/ry -4.40 x 104 /ry]

= 8 5.73 x 10 /ry CLASS_3aFREQUENCY 15 = 0.135 * [5.03 x 104/ry- 4.40 x 106/ry]

= 8.60 x 10 8/ry FrequencV of EPRI Class 3b Sequences. Consistent with NEI Interim Guidance [4], the frequency per reactor year for this category is calculated as:

CLASS_3b FREQUENCY 1 0 = PROBdass Sb 10 [CDFEAsE - CDFINDEPI CLASS_3bFREQUENCY 15 = PROBdass 3b 15 * [CFOASE - CDFINDEP]

Where:

CLASS_3b_FREQUENCY 1 ,o = frequency of small pre-existing containment liner leakage given a 1-in-10 years ILRT interval CLASS_3b_FREQUENCY 15 = frequency of small pre-existing containment liner leakage given a 1-in-15 years ILRT interval PROBdasSO3bo = probability of small pre-existing containment liner leakage given a 1-in-10 years ILRT frequency = 0.009 [See above write-up]

PROBdass 3b js = probability of small pre-existing containment liner leakage given a 1-in-15 years ILRT frequency = 0.01 35 [See above write-up]

CDFBASE = VY PSA Li core damage frequency = 5.03 x 106/ry [See step 1 write-up]

CDFINDEP = CDF for those individual sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences

= 4.40 x 106 /ry [See step 1 write-up]

Therefore, CLASS_3bFREQUENCYI0 = 0.009 * [5.03 x 106/ry- 4.40 x 106 /ry]

= 5.73 x 1 9/ry CLASS_3bFREQUENCYI 5 = 0.0135 * [5.03 x 10 6/ry -4.40 x 106 /ry]

= 8.60 x 1091ry

2.4.6 Population Dose Rate for New ILRT Interval (Step 6)

This step, per the NEI Interim Guidance [3], calculates the population dose rate for the new surveillance intervals of interest by multiplying the population dose (Table 2-6) by the frequency for each of the eight EPRI's accident classes (Tables 2-5 & 2-8). In addition, sum the accident class dose rates to obtain the total dose rate.

Per the NEI Interim Guidance [4], EPRI accident Classes 4 (Small Isolation Failure - failure-to-seal, Type B test), 5 (Small Isolation Failure - failure-to-seal, Type C test), and 6 (Containment Isolation Failures, dependent failures, personnel errors) are not affected by ILRT frequency and are not analyzed as part of this risk assessment. Therefore no selections of population dose estimates are made for these accident classes.

The calculation for a 1-in-10 years ILRT interval is as follows:

CLASS_1_DOSERATE-lo = CLASS_1_DOSE

  • CLASS_1_FREQUENCY 10 CLASS_2_DOSERATE10 = CLASS_2_DOSE
  • CLASS_2_FREQUENCY 10 CLASS_3a_DOSERATE.10 = CLASS_3aDOSE
  • CLASS_3a_FREQUENCYlo CLASS_3b-DOSERATE-10 = CLASS_3b_DOSE
  • CLASS_3bFREQUENCY 10 CLASS - 7aDOSERATE.10 = CLASS_7a_DOSE
  • CLASS_7aFREQUENCY 10 CLASS-7bDOSEFATE-10 = CLASS_7bDOSE
  • CLASS_7bFREQUENCY 10 CLASS_7cDOS ERATE10O = CLASS_7c_DOSE
  • CLASS_7cFREQUENCY 10 CLASS_8a_DOS ERATE-10 = CLASS_8a_DOSE
  • CLASSBa_FREQUENCY 10 CLASS-8b-DOSERATE-10 = CLASS_8bDOSE
  • CLASS_8bFREQUENCY 10 Where:

CLASS_1_DOSERATE.10 = EPRI accident Class 1 dose rate given a 1-in-10 years ILRT interval CLASS - 2 DOSERATE.10 = EPRI accident Class 2 dose rate given a 1-in-10 years ILRT interval CLASS_3aDOSERATE-10 = EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval CLASS_3b_DOSERATE-1o = EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval CLASS_7aDOSERATE-10 = EPRI accident Class 7a dose rate given a 1-in-10 years ILRT interval CLASS_7b_DOSERATE-1o = EPRI accident Class 7b dose rate given a 1-in-10 years ILRT interval CLASS_7c_DOSERATE-10 = EPRI accident Class 7c dose rate given a 1-in-10 years ILRT interval CLASS_8a_DOSERATE.10 = EPRI accident Class 8a dose rate given a 1-in-10 years ILRT interval CLASS_8b_DOSERATE.10 = EPRI accident Class 8b dose rate given a 1-in-10 years ILRT interval CLASS_1_DOSE = EPRI accident Class 1 dose = 1.30 x 103 (person-rem) [Table 2-6]

CLASS_2_DOSE = EPRI accident Class 2 dose = 2.73 x 106 (person-rem) [Table 2-6]

CLASS_3a_DOSE = EPRI accident Class 3a dose = 1.30 x 104 (person-rem) [Table 2-6]

CLASS_3b_DOSE = EPRI accident Class 3b dose = 4.55 x 104 (person-rem) [Table 2-6]

CLASS_7aDOSE = EPRI accident Class 7a dose = 2.73 x 106 (person-rem) [Table 2-6]

CLASS_7b_DOSE = EPRI accident Class 7b dose = 2.73 x 1o6 (person-rem) [Table 2-6]

CLASS_7cDOSE = EPRI accident Class 7c dose = 3.53 x 10o (person-rem) [Table 2-6]

CLASS_8a_DOSE = EPRI accident Class 8a dose = 2.73 x 106 (person-rem) [Table 2-6]

CLASS_8bDOSE = EPRI accident Class 8b dose = 2.96 x 106 (person-rem) [Table 2-6]

CLASS_1_FREQUENC Ylo = frequency of EPRI accident Class 1 given a 1-in-10 years ILRT Interval = 5.43 x 10 7/ry [Table 2-8]

CLASS_2_FREQUENC Yl 0 = frequency of EPRI accident Class 2 given a 3-in-10 years ILRT Interval = 2.14 x I0'9/ry [Table 2-5]

CLASS_3aFREQUENC Y10 = frequency of EPRI accident Class 3a given a 1-in-10 years ILRT Interval = 5.73 x 10'3/ry [Table 2-8]

CLASS_3b_FREQUENC Y10 = frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval = 5.73 x 10'9/ry [Table 2-8]

CLASS_7aFREQUENC Y10 = frequency of EPRI accident Class 7a given a 3-in-10 years ILRT Interval = 3.22 x 10' 7/ry [Table 2-5]

CLASS_7bFREQUENC Y10 = frequency of EPRI accident Class 7b given a 3-in-10 years ILRT Interval = 3.85 x 10 /ry [Table 2-5]

CLASS_7cFREQUENC Y10 = frequency of EPRI accident Class 7c given a 3-in-10 years ILRT Interval = 9.71 x 10'9/ry [Table 2-51 CLASS_8aFREQUENC Y10 = frequency of EPRI accident Class 8a given a 3-in-10 years ILRT Interval = 1.63 x 107/ry [Table 2-5]

CLASS_8bFREQUENC Y10 = frequency of EPRI accident Class 8b given a 3-in-1 0 years ILRT Interval = 5.32 x 10' 8/ry [Table 2-5]

Therefore, CLASS_1_DOSERATE.10 = 1.30 x103'

  • 5.43 x 10' 7.06 x 10' (person-rem/ry)

CLASS_2_DOS ERATE-10 = 2.73 x10 6

  • 2.14 x 10' 5.85 x 10'3 (person-rem/ry)

CLASS_3aDOSERATE-.10 = 1.30 x104

  • 5.73 x 1O-8 7.46 x 104 (person-rem/ry)

CLASS_3b_DOSERATE-1o = 4.55 x10 4

  • 5.73 x IO' 2.61 x 104 (person-rem/ry)

CLASS_7a_DOSERATE.10 = 2.73 x10 6 3.22 x 10'7 8.79 x 101' (person-rem/ry)

CLASS_7bDOSERATE-1o 6 3.85 x 10 1.05 x 10+1 (person-rem/ry)

CLASS-7c-DOSERATE-10 = 3.53 x106 9.71 x I0'9 3.43 x 10.2 (person-rem/ry)

CLASS_8a_DOSERATE.10 = 2.73 x10 6 1.63 x 10'7 4.46 x 10.' (person-rem/ry)

CLASS_8bDOSERATE.1o = 2.96 x106 5.32 x 10" 1.57 x 101' (person-rem/ry)

The calculation for a 1-in-15 years ILRT interval is as follows for the:

CLASS_1_DOSERATE.15 = CLASS_1_DOSE

  • CLASS_1_FREQUENCY 1 5 CLASS_2_DOSERATE.15 = CLASS_2_DOSE
  • CLASS_2_FREQUENCY 1 5 CLASS_3a_DOSERATE-15 = CLASS_3a_DOSE
  • CLASS_3aFREQUENCY 15 CLASS_3bDOSERATE.15 = CLASS_3b_DOSE
  • CLASS_3b_FREQUENCY 1 5 CLASS-7aDOSERATE.15 = CLASS_7aDOSE
  • CLASS_7aFREQUENCY 15 CLASS_7bDOSERATE-15 = CLASS_7bDOSE
  • CLASS_7bFREQUENCY 15 CLASS-7C-DOS ERATE-15 = CLASS_7cDOSE
  • CLASS_7cFREQUENCY 1 5 CLASS_8a_DOSERATE-15 = CLASS_8aDOSE
  • CLASS_8a_FREQUENCY 15 CLASS_8bDOSERATE.15 = CLASS_8b_DOSE
  • CLASS_8b_FREQUENCY1s Where:

CLASS_1_DOSERATE.15 = EPRI accident Class 1 dose rate given a 1-in-15 years ILRT interval CLASS_2-DOSERATE-15 = EPRI accident Class 2 dose rate given a 1-in-15 years ILRT interval

- Entergy REPORT No. VY-RPT-04-00002 I Revision 1 lPage l 40 l Of 70 l I

CLASS_3aDOSERATE-15 = EPRI accident Class 3a dose rate given a 1-in-15 years ILRT interval CLASS_3b_DOSERATE.15 = EPRI accident Class 3b dose rate given a 1-in-15 years ILRT interval CLASS-7a-DOS ERATE-15 = EPRI accident Class 7a dose rate given a 1-in-15 years ILRT interval CLASS_7bDOS ERATE-15 = EPRI accident Class 7b dose rate given a 1-in-15 years ILRT interval CLASS_7cDOSERATE-15 = EPRI accident Class 7c dose rate given a 1-in-15 years ILRT interval CLASS-8a-DOSERATE.15 = EPRI accident Class 8a dose rate given a 1-in-15 years ILRT interval CLASS_8b_DOSERATE-15 = EPRI accident Class 8b dose rate given a 1-in-15 years ILRT interval CLASS_1_DOSE = EPRI accident Class 1 dose = 1.30 x10 3 (person-rem) [Table 2-6]

CLASS_2_DOSE = EPRI accident Class 2 dose = 2.73 x106 (person-rem) [Table 2-6]

CLASS_3a_DOSE = EPRI accident Class 3a dose = 1.30 x104 (person-rem) [Table 2-6]

CLASS_3b_DOSE = EPRI accident Class 3b dose = 4.55 x104 (person-rem) [Table 2-6]

CLASS_7a_DOSE = EPRI accident Class 7a dose = 2.73 x106 (person-rem) [Table 2-6]

CLASS_7b_DOSE = EPRI accident Class 7b dose = 2.73 x106 (person-rem) [Table 2-6]

CLASS_7cDOSE = EPRI accident Class 7c dose = 3.53 x106 (person-rem) [Table 2-6]

CLASS_8a_DOSE = EPRI accident Class 8a dose = 2.73 x106 (person-rem) [Table 2-6]

CLASS_8bDOSE = EPRI accident Class 8b dose = 2.96 x106 (person-rem) [Table 2-6]

CLASS_1_FREQUENC Y15 = frequency of EPRI accident Class 1 given a 1-in-15 years ILRT Interval = 5.12 x 10' 7 /ry [Table 2-8]

CLASS_2_FREQUENC Y15 = frequency of EPRI accident Class 2 given a 3-in-10 years ILRT Interval = 2.14 x 10' 9/ry [Table 2-5]

CLASS_3a_FREQUENC Y15 = frequency of EPRI accident Class 3a given a 1-in-15 years ILRT Interval = 8.60 x 10'3/ry [Table 2-8]

CLASS_3b_FREQUENC Y15 = frequency of EPRI accident Class 3b given a 1-in-15 years ILRT Interval = 8.60 x 1i09/ry [Table 2-8]

CLASS_7aFREQUENC Y15 = frequency of EPRI accident Class 7a given a 3-in-10 years ILRT Interval = 3.22 x 10'7/ry [Table 2-5]

CLASS_7bFREQUENC Y1 5 = frequency of EPRI accident Class 7b given a 3-in-10 years ILRT Interval = 3.85 x 106/ry [Table 2-5]

CLASS_7c_FREQUENC Y1 5

= frequency of EPRI accident Class 7c given a 3-in-10 years ILRT Interval = 9.71 x 10'/ry [Table 2-5]

CLASS_8aFREQUENC Y15 frequency of EPRI accident Class 8a given a 3-in-10 years ILRT Interval = 1.63 x 10'7/ry [Table 2-5]

CLASS_8bFREQUENC Y15 = frequency of EPRI accident Class 8b given a 3-in-10 years ILRT Interval = 5.32 x 10 8 /ry [Table 2-5]

Therefore, CLASS_1_DOSERATE.15 = 1.30 x 103 5.12 x 10'7 6.65 x 10'4 (person-rem/ry)

CLASS_2_DOSERATE.15 = 2.73 x 106 2.14 x 10'9 5.85 x 10'3 (person-rem/ry)

CLASS-3a-DOSERATE-15 = 1.30 x 104 .

5.73 x 10.8 1.12 x 10'3 (person-rem/ry)

CLASS-3b-DOSERATE.15 = 4.55 x104 5.73 x 10'9 3.91 x 10i4 (person-rem/ry)

CLASS_7aDOSERATE.1,5 = 2.73 x 106 3.22 x 10'7 8.79 x 10' (person-rem/ry)

CLASS_7bDOS ERATE-15 = 2.73 x 106 3.85 x 104 1.05 x 10+1 (person-rem/ry)

A Entergy REPORT No. VY-RPT-04-00002 I Revision 1 lPage l 41 l Of 70 l CLASS-7cDOSERATE-15 = 3.53 x 106 9.71 x 109 = 3.43 x 102 (person-rem/ry)

CLASS_8aDOSERATE.15 = 2.73x10 6 1.63 x 10' 7 = 4.46 x 10' (person-rem/ry)

CLASS_8bDOSERATE.15 = 2.96x10 6 t 5.32 x 108 = 1.57 x 101t (person-rem/ry)

The dose rates per EPRI accident class as a function of ILRT interval are summarized in Table 2-9.

2.4.7 Change in Population Dose Rate Due to New ILRT Interval (Step 7)

This step, per the NEI Interim Guidance [4] calculates the percentage of the total dose rate attributable to EPRI accident Classes 3a and 3b (those accident classes affected by change in ILRT surveillance interval) and the change in this result dose rate from the base dose rate attributable to changes in ILRT surveillance interval.

Based on the results summarized in Table 2-10, for the current VY Station 1-inl 0 years ILRT interval, the percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b is calculated as follows:

PERCTDjo = percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given a 1-in-10 years ILRT interval, which is calculated using the following equation:

PERCTD ' [CLASS 3a DOSEATE +CLASS

-10 -3b-DO21007

- L TOTAL DOSE- JE1 CLASS_-3aDOSERATE-10 = EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= 7.46 x 104 [Table 2-9]

CLASS_3b_DOSERATE.10 = EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= 2.61 x 104 [Table 2-9]

TOTALDOSERATE-10 = Total dose rate for all EPRI's classes given a 1-in-10 years ILRT interval

= 12.0211 [Table 2-9]

Therefore, PER_CTD 0 = [6*1021-1

  • 100% = 0.0084%

0 [ 12.0211J The percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b based on the propose 1-in-1 5 years ILRT interval is calculated as follows:

PERCTD15 = percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given a 1-in-15 years ILRT interval, which is calculated using the following equation:

PER _CTD 5 = [CLASS 3a _ DOSERATE5 + CLASS _ 3b _ DOSER 4 TEp.I 100%

L TOTAL _ DOSERATE-15 CLASS_3a_DOSERATE.10 = EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= 1.12x104 [Table 2-9]

CLASS_3b_DOS ERATE-1o = EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= 3.91 x 104 [Table 2-9]

TOTAL-DOSERATE.1O = Total dose rate for all EPRI's classes given a 1-in-10 years ILRT interval

= 12.0215 [Table 2-9]

Therefore, PER CT0 1.12x10-3+3.91X104* 10 =0.0126%

- L 12.0215 j Based on the above results, the changes from the 1-in-10 years to 1-in-15 years dose rate is as follows:

INCREASE TOTAL-DOSEMTE 15-TOTAL D E* 100%

E0 115 TOTAL_ DOSE4L

- E I Where:

INCREASE1- 15 = percent change from 1-in-10 years ILRT interval to 1-in-15 years ILRT interval TOTAL-DOSERATE.IS= ' Total dose rate for all EPRI's classes given a 1-in-15 years ILRT interval

= 12.0215 (person-rem/ry) [Table 2-9]

TOTALDOSERATE-1o= Total dose rate for all EPRI's classes given a 1-in-10 years ILRT interval

= 12.0211 (person-rem/ry) [Table 2-9]

Therefore, INCREASE 101- 5 = [12.021512.021 1]

  • 100% = 0.0038%

2.4.8 Change in LERF Due to New ILRT Interval (Step 8)

This step, per the NEI Interim Guidance [4] calculates the change in the large early release frequency with extending the ILRT interval from 1-in-10 years to 1-in15-years.

The risk impact associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from containment could in fact result in a large release due to failure to detect a pre-existing leak during the relaxation period. For this evaluation only accident Class 3 sequences have the potential to result in large releases if a pre-existing leak were present. Class 1 sequences are not considered as potential large release pathways because for these sequences the containment remains intact. Therefore, the containment leak rate is expected to be small (less than 2La). A larger leak rate would imply an impaired containment, such as classes 2, 3, 6 and 7.

Late releases are excluded regardless of the size of the leak because late releases are, by definition, not a LERF event. At the same time, sequences in the Vermont Yankee PSA [6], which result in large releases (e.g., large isolation valve failures), are not impacted because a LERF will occur regardless of

Entergy REPORT No. VY-RPT-04-00002 I Revision 1 lPage l 43 l Of l 70 l the presence of a pre-existing leak. Therefore, the frequency of accident Class 3b sequences (Table 2-8) is used as the LERF for Vermont Yankee.

The affect on the LERF risk measure due to the proposed ILRT interval extension is calculated as follows:

ALERF10.1 5 = CLASS_3b_FREQUENCY 1 5 - CLASS_3bFREQUENCY 1 o Where:

ALERFIO.15 = the change in LERF from 1-in-10 years ILRT interval to 1-in-15 years ILRT interval CLASS_3bFREQUENC Y15 = frequency of EPRI accident Class 3b given a 1-in-15 years ILRT Interval = 8.60 x 109/ry [Table 2-8]

CLASS_3bFREQUENC Yl 0 = frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval = 5.73 x 1o9 /ry [Table 2-8]

Therefore, ALERF,0.15 = 8.60x l0'9/ry - 5.73 x 109/ry ALERF1 0.15 = 2.87 x 109/ry Regulatory Guide 1.174 [5] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 [5] defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10'6 /ry and increases in LERF below 10' 7 /ry. Since the ILRT does not impact CDF, the relevant risk metric is LERF.

This ALERF of 2.87 x 10'9 /ry falls into Region IlIl, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines in NRC Regulatory Guide 1.174 [5]. Therefore, because Regulatory Guide 1.174

[5] defines very small changes in LERF as below 10' 7/ry, increasing the ILRT interval at Vermont Yankee from the currently allowed 1-in-10 years to 1-in-15 years is non-risk significant from a risk perspective.

It should be noted that if the risk increase is measured from the original 3-in-1 0-year interval, the increase in LERF is as follows:

ALERF3 15 = CLASS_3b_FREQUENCY 15 - CLASS_3b_FREQUENCY 3 Where:

ALERF3 15 = the change in LERF from 3-in-10 years ILRT interval to 1-in-15 years ILRT interval CLASS_3bFREQUENC Y15 = frequency of EPRI accident Class 3b given a 1-in-15 years ILRT Interval = 8.60 x 10' 9/ry [Table 2-8]

CLASS_3b_FREQUENC Y3 = frequency of EPRI accident Class 3b given a 3-in-10 years ILRT Interval = 1.72 x 10' 9/ry [Table 2-8]

Therefore, ALERF3_15 = 8.60 x 10'9/ry - 1.72 x 10'9/ry ALERF3_15 = 6.88 x 10' 9/ry Similar to the ALERF10.15 result, the ALERF3.15 is also non-risk significant from a risk perspective.

En tergy REPORT No. VY-RPT-04-00002 l Revision 1lPage l O 70 1 2.4.9 Impact on Conditional Containment Failure Probability (Step 9)

This step, per the NEI Interim Guidance [4] calculates the change in conditional containment failure probability (CCFP). The CCFP risk metric ensures and shows that the proposed change in ILRT interval is consistent with the defense-in-depth philosophy describe in Regulatory Guide 1.174 [5]7.

In this calculation, the change in CCFP tracts the impact of the ILRT on both early (LERF) and late radionuclide releases. Based on the NEI Interim Guidance [4], CCFP consists of all those accident sequences resulting in a radionuclide release other than the intact containment state for EPRI accident Class 1, and small failures state for EPRI accident Class 3a. In addition, the CCFP is conditional given a severe core damage accident. The change in CCFP is calculated by the following equation:

CCFP =I -Intact-ContainmentlDFrequenc L Total CDFr - e OR

{[ICLASS - 1 FREQ+ CLASS 3a _ FREQ *1J CDFx,,SE JJ For the 1-in-10 years ILRT interval:

CLASS - I - FREQ10 + CLASS _ 3a _ FREQ l 100%

CCFP 0 = I CLASS CDFBASE ii Where:

CCFPI0 = conditional containment failure probability given 1-in-1 0 years ILRT interval CDFBASE = VY PSA Li core damage frequency = 5.03 x 10 4 /ry [Section 2.3, input #2]

CLASS_1_FREQ,0 = frequency of EPRI accident Class 1 given a 1-in-10 years ILRT Interval

= 5.43 x 107/ry [Table 2-8]

CLASS_3a_FREQO0 = frequencyof EPRI accident Class 3a given a 1-in-10 years ILRT Interval

= 5.73 x 104/ry [Table 2-8]

Therefore, rF5.43x10-7+5.73xI04-1 CCFP1 0 = I- I5

  • I 00% = 87.99%

IL 5.03x10-6 7 The defense-in-depth philosophy is maintained as areasonable balance among prevention of core damage, containment failure and consequence mitigation.

-~Entergy REPORT No. VY-RPT-04-00002 I Revision 1 IPage l 45 l Of I 70 l For the 1-in-15 years ILRT interval:

CCF,51- CAS - -FREQ 15 + CLASS_-3a_*00 5 1 {[c CP CDF8..tSE ]}

Where:

CCFP15 = conditional containment failure probability given 1-in-15 years ILRT interval CDFBASE = VY PSA Li core damage frequency = 5.03 x 106 /ry [Section 2.3, input #2]

CLASS_1_FREQ 15 = frequency of EPRI accident Class 1 given a 1-in-15 years ILRT Interval

= 5.12 x 107 1ry [Table 2-8]

CLASS_3aFREQ1 5 = frequency of EPRI accident Class 3a given a 1-in-15 years ILRT Interval

= 8.60 x 108 /ry [Table 2-8]

Therefore, CCFP55 {.12 x 10 + 8.60 x10

  • I00% = 88.05%

The change in the conditional containment failure probability from 1-in-1 0 years to 1-in-15 years is calculated as follows:

ACCFPo.15 = CCFPj 5 - CCFPo ACCFPIO,15 = 88.05% - 87.99%

ACCFPi0.is = 0.06%

This change in CCFP of less than 1% is insignificant from a risk perspective.

2.5 External Events Impact In response to Generic Letter 88-20, Supplement 4 116], VY submitted an Individual Plant Examination of External Events (IPEEE) in June 1996 [17]. The IPEEE was a review of external hazard risk (i.e., seismic, fires, high winds, external flooding, etc) to identify potential plant vulnerabilities and to understand severe accident risks. The results of the VY IPEEE are therefore used in this risk assessment to provide a comparison of the effect of external hazards when extending the current 1-in-10 years to 1-in-15 years Type A ILRT interval.

The VY IPEEE submittal [17] examined a spectrum of external events hazards based on acceptable screening methods (NRC seismic margin [19], the EPRI Fire Induced Vulnerability Evaluation Methodology (FIVE) [26], etc.). These screening methods use varying levels of conservatism; therefore, it is not practical to incorporate realistic quantitative risk assessments of all external event hazards into the ILRT extension assessment at this time. As a result, external events hazards are evaluated as a sensitivity case to demonstrate that the conclusions of the internal events analysis would not be changed if external events hazards were considered.

Ent gy REPORT No. Y-RPT-04-00002 Revision 1 l Page 46 Ofl The impact of external events on this ILRT risk assessment is summarized in this section (refer to Appendix A for further details).

The purpose of the external events evaluation is to determine whether there are any unique insights or important quantitative information that explicitly impact the risk assessment results when considering only internal events.

The quantitative consideration of external hazards is discussed in more detail in Appendix A of this report.

As can be seen from Appendix A, if the external hazard risk results of the VY IPEEE are included in this assessment (i.e., in addition to internal events), the change in LERF associated with the increase in ILRT interval from 10 years to 15 years is estimated to be 3.71 x 10 8/ry. This delta-LERF falls within the Region Ill, Very Small Change in Risk (Figure 2-1), of NRC Regulatory Guide 1.174 [5].

Other salient results from Appendix A, found the combined internal and external events increase in risk for those accident sequences influenced by Type A testing, compared with the total integrated plant risk, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, to be 0.004% or 0.007 person-rem/ry. In addition, the change in the combined internal and external events conditional containment failure probability from 1-in-1 0 years to 1-in-15 years is 0.05%. A change in CCFP of less than 1% is insignificant from a risk perspective.

Therefore, incorporating external event accident sequence results into this analysis does not change the conclusion of intemal events only risk assessment (i.e., increasing the VY ILRT interval from 10 to 15 years is an acceptable plant change from a risk perspective). This result is expected, because the proposed ILRT interval extension impacts plant risk in a very specific and limited way.

2.6 Containment Liner Corrosion Risk Impact Recently, the NRC issued a series of Requests for Additional Information (RAls) in response to the one-time relief requests for the ILRT surveillance interval submitted by various licensees. One of the RAls related to the risk assessment performed in this report is provided below.

Request for Additional Information:

Inspections of reinforcedandsteel containments at some facilities (e.g., North Anna, Brunswick D.C.

Cook, and Oyster Creek) have indicated degradation from the uninspectable (embedded) side of the steel shell and liner of primary containments. The major uninspectable areas of the Mark I containment are the vertical portion of the drywell shell and part of the shell sandwiched between the drywell floor and the basemat. Please discuss what programs are used to monitor their conditions.

Also, address how potential leakage due to age-related degradation from these uninspectable areas are factored into the risk assessment in support of the requested interval extension.

The impact of the risk assessment portion of the above RAls is summarized in this section (refer to Appendix B for further details).

The containment liner corrosion analysis utilizes the referenced Calvert Cliffs Nuclear Power Plant assessment [20] to estimate the likelihood and risk-implication of degradation-induced leakage occurring and going undetected in visual examinations during the extended test interval. It should be noted that the Calvert Cliffs analysis was performed for a concrete cylinder and dome containment with a steel liner whereas VY has a free standing steel containment building. Both sites do, however, have a concrete basemat with a steel liner.

Consistent with the Calvert Cliffs analysis, the following issues are addressed:

  • Differences between the containment basemat and the drywell and torus liner
  • The historical drywelltorus steel shell flaw likelihood due to concealed corrosion
  • The impact of aging
  • The corrosion leakage dependency on containment pressure
  • The likelihood that visual inspections will be effective at detecting a flaw Consistent with Calvert Cliffs analysis [20], the following six steps are performed:
1) Determine the historical liner flaw likelihood.
2) Determine aged adjusted liner flaw likelihood.
3) Determine the increase in flaw likelihood between 3, 10 and 15 years.
4) Determine the likelihood of containment breach given liner flaw.
5) Determine the visual inspection detection failure.
6) Determine the likelihood of non-detected containment leakage.

In additions to these steps, the following three additional steps are added to evaluate risk-implication of containment liner corrosion:

7) Evaluate the risk impact in terms of population dose rate and percentile change for the interval cases.
8) Evaluate the risk impact in terms of LERF.
9) Evaluate the change in conditional containment failure probability.

The quantitative consideration of the containment liner corrosion analysis is discussed in more detail in Appendix B of this report. As can be seen from Appendix B, including corrosion effects in the ILRT assessment would not alter the conclusions from the original internal events analysis. That is, the change in LERF from extending the interval to 15 years from the current 10-year requirement is estimated to be 3.19 x 10'9/ry. This value is below the NRC Regulatory Guide 1.174 [5] of 10'7 /ry. Therefore, because Regulatory Guide 1.174 [5] defines very small changes in LERF as below 10'7/ry, increasing the ILRT interval at VY from the currently allowed 1-in-10 years to 1-in-15 years and taking into consideration the likelihood of a containment liner flaw due to corrosion is non-risk significant from a risk perspective.

Additionally, the dose increase is estimated to be 5.00 x 104 person-rem/ry or 0.0042%, and the conditional containment failure probability increase is estimated to be 0.06%. Both of these increases are also considered to be small. As a result, the ILRT interval extension is considered to have a minimal impact on plant risk (including age-adjusted corrosion impacts), and is therefore acceptable.

In addition, a series of parametric sensitivity studies (discussed in more detail in Appendix B of this report) regarding the potential age related corrosion effects on the containment steel liner also predict that even with conservative assumptions, the conclusions from the original internal events analysis would not change.

~En terry REPORT No. VY-RPT-04-00002 I Revision 1 lPage l 48 l Of I 70 I Figure 2-1 Acceptance Guidelinesa for Large Early Release Frequency [5]

I cc 10-

..-Region i Region i1 10-15 LERF -

e The analysis will be subject to increased technical review and management attention as indicated by the darkness of the shading of the figure. In the context of the integrated decisionmaking, the boundaries between regions should not be interpreted as being definitive; the numerical values associated with defining the regions in the figure are to be interpreted as indicative values only.

-:Entergy REPORT No. VY-RPT-04-00002 I Revision 1 lPage 49 101 Table 2-1 Internal Core Damage Frequency Contributions by Plant Damage States Bin Bin Description l Frequency_l Total CDF ID Transient sequences with loss of all injection. Core damage occurs 1.4219 104 28.25%

with the reactor at low pressure.

Transient sequences with loss of all high pressure injection and IA failure to depressurize. Core damage occurs with the reactor at 1.1321 x 104 22.50%

high pressure.

IBL 'Late'SBO. Core cooling is maintained by HPCI/RCIC until 8.3971 x 10'7 16.69%

batteries deplete. _ _ _ _ _ _ _

IIA Transient sequence with loss of all containment heat removal. Core 4.3501 x -7 8.64%

damage is caused by containment failure.

IBE 'Early' SBO sequences. Core damage occurs due to early failure of 3.1907 x 10'7 6.34%

HPCI and RCIC.

1111B Small or Medium LOCA sequences for which the reactor cannot be 2.0785 x 10O 4.13%

depressurized prior to core damage occurring.

Transient sequences where the main condenser and RHR fail; and IIV the towus vent opens for containment pressure relief. Core damage 1.8050 x 10" 3.59%

occurs when ECCS systems fail NPSH, due to failure to reclose the vent.

IIC LOCA sequences with loss of injection. Core damage occurs with 1.3739 x 10'7 2.73%

the reactor at low pressure. _ _ _ _ _ _ _ _ _ _ _ _ _

IVA ATWS sequences where core damage is caused by containment 1.1218 x 10 7 2.23%

failure.

IED 'Early SBO sequences caused by failure of DC-1 and DC-2. 5.6127x 10.8 1.12%

V Containment Bypass sequences. (Interfacing systems LOCA and 5.3176 x 1.06%

LOCA outside of containment.) 5.16__0___

IVL ATWS sequences where core damage occurs due to overpressure 5.2938 x 10-8 1.05%

failure of the Reactor Coolant System. _ ._

IlL Loss of containment heat removal with RPV breach but no initial 4.7995 x 10." 0.95%

core damage; core damage after containment failure.

IC ATWS sequences where core damage is caused by loss of 1.6074 x 10.8 0.32%

injection during level/power control.

IEC Transient sequences with delayed loss of dc power due to failure of 9.8624 x 10 9 0.20%

_____ _____ battery chargers. _ _ _ _ _ _ _

LOCA sequences where core damage is caused by containment IIID failure. Containment fails due to failure of vapor suppression 6.3348 x 10-9 0.13%

(stuck-open vacuum breaker.)

IIIA RPV ruptures due to failure of all over-pressure protection systems 4.3631 x 10-9 0.09%

(SO=F). _ _ _ _ I__

TOTAL 5.0326 x 1046 100%

- Entergy REPORT No. VY-RPT-04-00002 I Revision 1 IPagel 50 Of 70I Table 2-2 Summary of Vermont Yankee PSA LEVEL 2 Release Categories [6]

Release Category Frequency  % of

__ __ _ __ _ _ __ _ _ (ry) Total CDF Early High 1.51 x 10 4 T 31.01 Nomenclature:

Intermediate High 4.37 x 109l 0.09 Late High 6.53 x 10-7 12.99 Timing:

4 Early Medium 2.09 x 10 41.57 Late: greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Intermediate Medium 2.43 x 10- 9 0.05 Intermediate: 6 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Early: Less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Late Medium 1.83 x 10.8 0.36 Early Low 7.81 x 10.8 1.55 Magnitude:

Intermediate Low 0.o0 x 10° < 0.01 NCF: Little to no release.

Low-Low: Less than 0.1% Iodine.

Late Low 0.00 x 100 < 0.01 Low: 0. I to I% Cs Iodine.

Early Low-Low 4.04 x 10-9 0.08 Medium: I to 10O Cs Iodine.

High: Greater than 10% Cs Iodide Intermediate Low-Low 1.07 x 10.8 0.21 Late Low-Low 0.00 x 10° < 0.01 Containment Bypass:

Containment Bypass (V-Seq) 5.32 x 10.8 1.08 V-Seq: ISLOCA & LOCAOC accident sequences; these No Containment Failure (NCF) 6.06 x 10' 12.07 are treated as Early High releases.

TOTAL 5.03 x 106 100.00 Table 2-3 Summary of VY PSA LEVEL 2 Containment Failures [6]

Containment End State Frequency  % Of Containment_

E n d _Sta t e (Iry) Total CDF 7

No Containment Failure 6.06 x 10 12.1 Containment Isolation Failure9 2.55 x 108 0.5 7

Containment Failure - Energetic Phenomena 3.22 x 10 6.4 Containment Failure - Contact w/core material 3.85 x 104 76.5 Containment Failure - Loss of Heat Removal Capability 9.71 x 109 0.2 Bypass (ATWS/ISLOCA/LOCAOC) 2.16 x 10-7 4.3 9 VY containment isolation failures include large containment Isolation failures (failure-to-close), small Isolation failures (liner breach), and large isolation failures (liner breach).

I Total l 5.03 x 104 l 100.0 I Table 2-4 VY Population Doses for Each PSA LEVEL 2 Release Category Release Category Population Dose (person-sv)

No Containment Failure (NCF) 1.30 x 10+1 Early High 2.73 x 10+4 Intermediate High 3.53 x 10e Late High 2.69 x 10" Early Medium 1.47 x 10"4 Intermediate Medium 1.02 x 10+4 Late Medium 1.13 x 10+4 Early Low 2.11 x 10+3 Intermediate Low 2.71 x 10+3 Late Low Early Low-Low 6.18 x 10+2 Intermediate Low-Low 2.80 x 10+2 Late Low-Low Containment Bypass (V-Seq) 2.96 x 10"4

LEn terby REPORT No. VY-RPT-04-00002 I Revision 1 Page l 52 l Of l 70 l Table 2-5 VY PSA Level 2 Release Frequencies As A Function of EPRI Accident Class EPRI Accident Class Description Frequency  % Of 1 No Containment Failure 5.88 x 107 11.75%

2 Large Containment Isolation Failures (Failure-to- 2.14 x 1O9 0.04%

close) 3a Small Isolation Failures (Liner breach) 1.72 x 108 0.34%

3b Large Isolation Failures (Liner Breach) 1.72 x 109 0.03%

4 Small isolation failure - failure-to-seal (Type B test) 5 Small isolation failure - failure-to-seal (Type C test) --- _---

Containment Isolation Failures (dependent failures, 6 personnel errors) 7a Severe Accident Phenomena Induced Containment 3.22 x 107 6.44%

Failure - energetic phenomena ._I Severe Accident Phenomena Induced Containment 7b Failure - direct contact with core material 3.85 x 10. 76.87%

7c Severe Accident Phenomena Induced Containment Failure 9.71 x 1 9 0.19%

- containment failures due to loss of heat removal 9 8a Containment Bypassed (ATWS) 1.63 x 10 7 3.26%

8b Containment Bypassed (ISLOCA, LOCAOC) 5.32 x IO' 1.06%

TOTAL 5.00 x lo, 100.00%

Entergy REPORT No. VY-RPT-04-00002 I Revision 1 Page 1 531 Of1 7 1 Table 2-6 VY Population Dose Estimates As A Function of EPRI Accident Class within 50-Mile Radius EPRI AcietCasDsrpinPerson-Rem Cls AcciClass lass Description within 50 miles 1 No Containment Failure 1.30 x 1 l 2 Large Containment Isolation Failures (Failure-to-close) 2.73 x 106 Sa Small Isolation Failures (Liner breach) 1.30 x 104 3b Large Isolation Failures (Liner Breach) 4.55 x 104 4 Small isolation failure - failure-to-seal (Type B test) _

5 Small isolation failure - failure-to-seal (Type C test) 6 Containment Isolation Failures (dependent failures, personnel errors) 7a Severe Accident Phenomena Induced Containment Failure - energetic 2.73 x 1o6 phenomena 7b Severe Accident Phenomena Induced Containment Failure - direct contact 2.73 x 1O6 with core material 7c Severe Accident Phenomena Induced Containment Failure - containment 3.53 1O6 failures due to loss of heat removal 8a Containment Bypassed (ATWS) 2.73 x 106 8b Containment Bypassed (ISLOCA, LOCAOC) 2.96 x 1O6

Table 2-7 VY Dose Rate Estimates as a Function of EPRI Accident Class For Population within 50-Miles (Base Line 3 per 10 year ILRT)

Person- Dose Rate EPRI Class Accident Class Description Rem Frequency (Person-within 50 (Iry) Rern/ry)

_ __ __ __ _ _ __ __ _ _ _ _ _ _ _ _ _ _ _ m ile s _ _ _ _ _ _ _ _ _ _ _ _

1 No Containment Failure 1.30 x 103 5.88 x 107 7.64 x 10'4 2 Large Containment Isolation 2.73 x 105 2.14 x 10 9 5.85 x 10' 3 Failures (Failure-to-close) 3a Small Isolation Failures (Liner 1.30 x 10 4 1.72 x 10.8 2.24 x 10'4 breach) 3b Large Isolation Failures (Liner 4.55 x 10 4 1.72 x 7.83 x 1O-

___________ Breach) 45x1 1.x10 78310 4 Small isolation failure - failure-to-seal (Type B test) 5 Small isolation failure - failure-to-seal (Type C test)

Containment Isolation Failures 6 (dependent failures, personnel errors)

Severe Accident Phenomena 7a Induced Containment Failure - 2.73 x 105 3.22 x 10-7 8.79 x 10 '

energetic phenomena Severe Accident Phenomena 7b Induced Containment Failure - 2.73 x 106 3.85 x 104 1.05 x 10+'

direct contact with core material Severe Accident Phenomena 7c Induced Containment Failure - 3.53 x 10 9.71 x 10' 3.43 x 10.2 heat removal 8a Containment Bypassed (ATWS) 2.73 x 106 1.63 x 1 ' 4.46 x 1o0l 8b Containment Bypassed (ISLOCA, 2.9 x 6IO' 5.32 x IO' 1.57 x IO-'

LOCAOC)5.30x10 T A 1.20x10o' TOTALS 5.00 x10, 1.20 x1IO+

Table 2-8 EPRI Accident Class Frequency as a Function of ILRT Interval Table 2-9 Baseline Dose Rate Estimates By EPRI Accident Class for Population Within 50-Mile Dose Rate as a Function of ILRT Interval EPRI (Person-Rem / ry)

Class Accident Class Description Baseline Current Proposed 3-per-10 years 1-per-10 years 1-per-15 years ILRT ILRT ILRT 1 No Containment Failure 7.64 x 104 7.06 x 104 6.65 x 10'4 2 Large Containment Isolation Failures 5.85 x 104 5.85 x 10 4 5.85 x 103 (Failure-to-close) 3a Small Isolation Failures (Liner breach) 2.24 x 10-4 7.46 x 10O4 1.12 x 104 3b Large Isolation Failures (Liner Breach) 7.83 x 105 2.61 x 104 3.91 x 1 5 4 Small isolation failure - failure-to-seal (Type B test) 5 Small isolation failure - failure-to-seal (Type C test) 6 Containment Isolation Failures (dependent failures, personnel errors) 7a Severe Accident Phenomena Induced 8.79 x 10 l 8.79 x 101 8.79 x 10 Containment Failure - energetic phenomena Severe Accident Phenomena Induced 7b Containment Failure - direct contact with core 1.05 x 10' 1.05 x 10t' 1.05 x 10+

material Severe Accident Phenomena Induced 7c Containment Failure - containment failures 3.43 x 102 3.43 x 10.2 3.43 x 10-2 due to loss of heat removal 8a Containment Bypassed (ATWS) 4.46 x 10l 4.46 x 10 l 4.46 x 10.1 8b Containment Bypassed (ISLOCA, LOCAOC) 1.57 x 10.1 1.57 x 101 1.57 x 10 l TOTALS 12.0204 12.0211 12.0215

wEn tergy REPORT No. VY-RPT-04-00002 Revision 1 l Page l56 l Of l 70 SECTION 3

SUMMARY

OF RESULTS 3.1 Internal Events Impact An evaluation was performed to assess the risk impact of extending the current containment Type A Integrated Leak Rate Test (ILRT) interval. In performing the risk assessment evaluation, the guidance and additional information distributed by NEI in November 2001 to their Administrative Points of Contact

[3,4] regarding risk assessment evaluation of one-time extensions of containment ILRT intervals and the approach outlined in the Indian Point Unit Three Nuclear Power Plant ILRT [8, 9] extension submittal were used. The assessment also followed previous work as outline in NEI 94-01 [1], the methodology used in EPRI TR-104285 [2], and the NRC Regulatory Guide 1.174 [5).

These results demonstrate a very small impact on risk associated with the one time extension of the ILRT test interval to 15 years. The following is a brief summary of some of the key aspects of the ILRT test interval extension risk analysis:

1) The baseline (3-in-1 0 years) risk contribution (person-rem) associated with containment leakage affected by the ILRT and represented by Classes 3a and 3b accident scenarios is 0.003% of the total risk.
2) When the ILRT interval is 1-in-1 0 years, the risk contribution of leakage (person-rem) represented by Classes 3a and 3b accident scenarios increases to 0.008% of the total risk.
3) When the ILRT interval is 1-in-15 years, the risk contribution of leakage represented by Classes 3a and 3b accident scenarios increases to 0.013% of the total risk.
4) The increase in risk on the total integrated plant risk as measured by person-rem/reactor year increases for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.004% (0.0005 person-rem/ry).

This value can be considered to be a negligible increase in risk.

5) The risk increase in LERF from reducing the ILRT test frequency from the current once-per-I 0 years to once-per-15 years is 2.87 x 10 9/ry. This is determined to be very small using the acceptance guidelines of Regulatory Guide 1.174.
6) The risk increase in LERF from the original 3-in-10 years test frequency; to once-per-15 years is 6.88 x 10- 9/ry. This is also found to be "very small" using the acceptance guidelines in Regulatory Guide 1.174.
7) The change in CCFP of 0.06% is deemed to be insignificant and reflects sufficient defense-in-depth.
8) Other salient results are summarized in Table 3-1. The key results to this risk assessment are those for the 10-year interval (current VY ILRT interval) and the 15-year interval (proposed change). The 3-in-1 0 year ILRT is a baseline starting point for this risk assessment given that the pre-existing containment leakage probabilities (estimated based on industry experience - refer to Section 1.2) are reflective of the 3-per-1 0 year ILRT testing.

Entergy REPORT No. VY-RPT-04-00002 l Revision 1 Page 57 l l Of I 7l 3.2 External Events Impact This analysis provides an evaluation of external events hazards (seismic, fires, high winds, external flooding, etc) impacts within the framework of the ILRT interval extension risk assessment. Similar to the internal events analysis, the combined impact of internal and external events confirms that the impact (due to the proposed ILRT extension) on the external hazard portion of the VY plant risk profile is comparable to that shown for internal events. It is deemed that the calculated risk increase for both internal and external hazards would remain 'small".

These results demonstrate a small impact on risk associated with the one time extension of the ILRT test interval to 15 years. The following is a brief summary of some of the key aspects of the ILRT test interval extension risk analysis for the combined internal and external events analysis:

1) The baseline (3-in-1 0 years) risk contribution (person-rem) associated with containment leakage affected by the ILRT and represented by Classes 3a and 3b accident scenarios is 0.003% of the total risk.
2) When the ILRT interval is 1-in-1 0 years, the risk contribution of leakage (person-rem) represented by Classes 3a and 3b accident scenarios increases to 0.008% of the total risk.
3) When the ILRT interval is 1-in-15 years, the risk contribution of leakage represented by Classes 3a and 3b accident scenarios increases to 0.012% of the total risk.
4) The combined internal and external events increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-1 0 years test interval to a 1-in-15 years test interval, is found to be 0.004% (0.007 person-rem/ry). This value can be considered to be a negligible increase in risk.
5) The combined internal and external events risk increase in LERF from reducing the ILRT test frequency from the current once-per-10 years to once-per-15 years is 3.71 x 10 8/ry. This is less than
10. /ry, which is within the bounds of Region l1l, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines of Regulatory Guide 1.174.
6) The combined internal and external events change in CCFP of 0.06% is deemed to be insignificant and reflects sufficient defense-in-depth.
7) Other salient results are summarized in Table 3-2.

3.3 Containment Liner Corrosion Risk Impact This analysis provides a sensitivity evaluation of considering potential corrosion impacts within the framework of the ILRT interval extension risk assessment. The analysis confirms that the ILRT interval extension has a minimal impact on plant risk. Additionally, a series of parametric sensitivity studies regarding the potential age related corrosion effects on the steel shell also indicate that even with very conservative assumptions, the conclusions from the original analysis would not change. That is, the ILRT interval extension is judged to have a minimal impact on plant risk and is therefore acceptable.

1) The baseline (3-in-1 years) risk contribution (person-rem) associated with containment leakage affected by the ILRT and represented by Classes 3a and 3b accident scenarios is 0.0027% of the total risk.

En ergy REPORT No. VY-RPT-04-00002 lRevision 1I Page l I Of l 70

2) When the ILRT interval is 1-in-10 years, the risk contribution of leakage (person-rem) represented by Classes 3a and 3b accident scenarios increases to 0.0085% of the total risk.
3) When the ILRT interval is 1-in-1 5 years, the risk contribution of leakage represented by Classes 3a and 3b accident scenarios increases to 0.0128% of the total risk.
4) The age-adjusted corrosion impact on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-1 0 years test interval to a 1-in-i5 years test interval, is found to be 0.0042% (0.0005 person-rem/ry). This value can be considered to be a negligible increase in risk.
5) The age-adjusted corrosion impact risk increase in LERF from reducing the ILRT test frequency from the current once-per-10 years to once-per-15 years is 3.19 x 10' 91ry. This is determined to be below the 1C'/ry criterion of Region l1l, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines of Regulatory Guide 1.174.
6) This age-adjusted corrosion impact change in CCFP of 0.06% is deemed to be insignificant and reflects sufficient defense-in-depth.
7) Other results (taken from Appendix B) of the updated ILRT assessment including the potential impact from non-detected containment leakage scenarios assuming that 100% of the leakages result in EPRI Class 3b are show in Table 3-3.

Additional sensitivity cases were also developed to gain an understanding of the containment liner corrosion sensitivity to various key parameters. The sensitivity cases are as follows:

  • Sensitivity Case 1 - Flaw rate doubles every 2 years
  • Sensitivity Case 2 - Flaw rate doubles every 10 years
  • Sensitivity Case 3 - 5% Visual inspection failures
  • Sensitivity Case 4 - 15% Visual inspection failures
  • Sensitivity Case 5 - Containment breach base point 10 times lower
  • Sensitivity Case 6 - Containment breach base point 10 times higher
  • Sensitivity Case 7 - Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5% visual inspection failures and 10% EPRI accident Class 3b are LERF (Lower bound)
  • Sensitivity Case 8 - Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15% visual inspection failures and 100% EPRI accident Class 3b are LERF (upper bound)

The results of the containment liner corrosion sensitivities cases, taken from Appendix B are summarized in Table 3-4.

Ea

- En fergy REPORT No. VY-RPT-04-00002 I Revision 1 IPage 1 59 1 Of l 70 l Table 3-1 Summary of Risk Impact on Extending Type A ILRT Test Frequency - Internal Events Base Case Extend to Extend to 3 Years 10 Years 15 Years EPRI CDF Pe-m Per-Rem CDF Pe-m Per-Rem CDF Per-RemPe-m Class (Per ) P-Rem (Per ry) (Per ry) (Per ry) (Per ry) (Per ry) 1 5.88x10'7 1.30x10 3 7.64x10'4 5.43xlO7 1.30x10 3 7.06x 10' 4 5.12x10'7 1.30x103 6.65x10'4 2 2.14x10'9 2.73x10 5.85x10'3 2.14xl1O' 2.73x106 5.85x10'3 2.14x10'9 2.73x106 5.85x104 3a 1.72 x 10o8 1.30 x 104 2.24 x 10' 4 5.73 x 104 1.30 x 104 7.46 x 10'4 8.60 x 104 1.30 x104 1.12 x 104 3b 1.72 x 10'9 4.55 x 104 7.83 x 10' 5 5.73 x 104 4.55 x 104 2.61 x 10'4 8.60 x 10'9 4.55 x104 3.91 x 10'5 4 0.00 0.00 0.00 5 0.00 0.00 0.00 6 0.00 0.00 - 0.00 7a 3.22 x 10' 7 2.73 x 106 8.79 x 10't 3.22 x 10'7 2.73 x 106 8.79 x 10 ' 3.22 x 10' 7 2.73 x106 8.79 x 10-'

7b 3.85 x 1o-6 2.73 x 1O6 1.05 x 10+1 3.85 x 104 2.73 x 1O6 1.05 x 10+' 3.85 x 10o 2.73 x106 1.05 x 10+1 7c 9.71 x 10' 9 3.53 x 1 o 3.43 x1I 2 9.71 x 10' 3.53 x 106 3.43 x 10 2 9.71 x 10'9 3.53 x 106 3.43 x 10o 7d 1.63 x 10'7 2.73 x 1O 4.46 x 10-1 1.63 x 1O- 7 2.73 x 106 4.46 x 10 1 1.63 x 10'7 2.73 x 106 4.46 x 1O-8 5.32 x 10o-2.96 x 106 1.57 x 101 5.32 x 104 2.96 x 106 1.57 x 10' 5.32 x 10-8 2.96 x 1O 1.57 x 10 Total 5.00 x 10 __ 12.0204 5.00 x 1046 - 12.0211 5.00 x 10 _ 12.0215 ILRT Dose Rate 3.02 x 10 1.01 x lo, 1.51 x 10'3 from 3a and 3b3

% Of Total 0.0025% 0.0084% 0.0126%

Delta Dose Rate from 3a and 3b 5.03 x 10J (10 to 15 vr)l LERF from 3b 1.72 x 10'9 5.73 x 10' 8.60 x 10'9 Delta LERF 2.87 xI' (10 to 15 yr)

CCFP % 87.91% 87.99% 88.05%

Delta CCFP % 0.06%

(10 to 15 yr)

"Entergy REPORT No. VY-RPT-04-00002 I Revision 1 IPage 1 60 j Of l 70 l Table 3-2 Summary of Risk Impact on Extending Type A ILRT Test Frequency - Effect of Internal and External Events Risk on VY ILRT Risk Assessment Base Case Extend to Extend to 3 Years 10 Years 15 Years EPRI CDF Per-Rem Per-Rem CDF Per-Rem Per-Rem CDF Per-Rem Per-Rem Class (Per ry) (Per ry) (Per ry) PerRem_(Per ry) (Per ry) (Per ry) 3 1 7.96 x 10 4 1.30 x 10 1.03 x 10 2 7.40 x 10 4 1.30 x 10 3 9.62 x 1033 6.99 x106 1.30 x 10 3 9.08 x 1048 2 2.90 x 10o8 2.73 x 1O 7.93 x 1o- 2 2.90 x 1o-8 2.73 x 106 7.93 x 10 2 2.90 x 104 2.73 x 106 7.93 x 10o2 3a 2.33 x 10 7 1.30 x 104 3.03 x 104 7.42 x 10 7 1.30 x 104 9.64 x 10-3 1.11 x 104 1.30 x 104 1.45 x 10 2 3b 2.33 x 108 4.55 x 104 1.06 x 10' 3 7.42 x 1o-8 4.55 x 104 3.38 x 10- 3 1.11 x 107 4.55 x 104 5.06 x 1043 4 0.00 - 0.00 0.00 5 - . 0.00 .. 0.00 0.00 6 . 0.00 0.00 0.00 7a 4.36 x 106 2.73 x 106 1.19 x 101 4.36 x 10 2.73 x 10 1.19 x 101 4.36 x 104 2.73 x 106 1.19 x 10' 7b 5.21 x 10'5 2.73 x 10 1.42 x 102 5.21 x 10' 2.73 x 10 1.42 x 10' 5.21 x 10- 2.73 x 10 1.42 x 10 7c 1.31 x 10 7 3.53 x 10 4.64 x10O- 1.31 x 10- 3.53 x 106 4.64 x 10 1.31 x 10-7 3.53 x 106 4.64 x 10 7d 2.21 x 10 6 2.73 x 10 6.06 x 100 2.21 x 10 O 2.73 x 106 6.06 x 10° 2.21 x 1O4 2.73 x 106 6.06 x 10 8 7.20 x 10'7 2.96 x 106 2.13 x 10° 7.20 x 10- 7 2.96 x 10 6 2.13 x 10° 7.20 x 10-7 2.96 x 106 2.13 x 10 Total 6.77 x 1o05 -- 162.7729 6.77 x 10'5 - 162.7811 6.77 x 10.5 - 162.7871 ILRT Dose Rate 4.09 x 1 1.30 x 10.2 1.95 x 10.2 from 3a and 3b

% Of Total 0.0025% 0.0080% 0.0120%

Delta Dose Rate from 3a and 3b 6.51 x 10'3 (10 to 15 yr)

LERF from 3b 2.33 x 10o 7.42 x1Io 8 1.11 10'7 Delta LERF 3.71 x 1O."

(10 to 15 yr)

CCFP % 87.91% 87.99% 88.04%

Delta CCFP % 0.05%

(10 to 15 yr)

AEn tergy REPORT No. VY-RPT-04-00002 I Revision 11 Page l 61 lOf I 70 I Table 3-3 Summary of Risk Impact on Extending Type A ILRT Test Frequency - Impact of Containment Steel Liner Corrosion on VY ILRT Intervals Base Case Extend to Extend to 3 Years 10 Years 15 Years EPRI CDF Per-Rem Per-Rem CDF Per-Rem CDF Per-Rem Per-Rem Class (Per Ry) (Per Ry) (Per Ry) Per-Rem (Per Ry) (Per Ry) Per-Rem (Per Ry) 1 5.88 x 10'7 1.30 x 103 7.64 x 10' 4 5.43 x 10'7 1.30 x10 3 7.06 x 10'4 5.11 x 10'7 1.30 x 103 6.65 x 10' 4 2 2.14 x 10'9 2.73 x 106 5.85 x 10' 3 2.14 x 10' 9 2.73 x10 6 5.85 x 103 2.14 x 10' 9 2.73 x 106 5.85 x 10' 3 3a 1.72 x 10o 1.30 x 10 4 2.24 x 10' 4 5.73 x 10 8 1.30 x10 4 7.46 x 10'4 8.60 x 10 8 1.30 x 10 4 1.12 x 10' 3 3b 1.76 x 10'9 4.55 x 104 8.01 x 10- 5 5.97 x 10' 9 4.55 x10 4 2.72 x 10'4 9.16 x 10' 9 4.55 x 10 4 4.17 x 10'4 4 0.0 N/A 0.0 0.0 N/A 0.0 0.0 N/A 0.0 5 0.0 N/A 0.0 0.0 N/A 0.0 0.0 N/A 0.0 6 0.0 N/A 0.0 0.0 N/A 0.0 0.0 N/A 0.0 7a 3.22x 107 2.73x 106 8.79 x 10 1 3.22x 10 7 2.73x10 6 8.79x 10o 3.22x 107 2.73 x 1o6 8.79 x1I 7b 3.85 x 104 2.73 x 106 1.05 x 10' 3.85 x 104 2.73 x 105 1.05 x101 3.85 x 104 2.73 x 106 1.05 x 101 7c 9.71 x 104 3.53 x 106 3.43 x 10.2 9.71 x 10-9 3.53 x 106 3.43 x 10-2 9.71 x 1 9 3.53 x 106 3.43 x 10-2 8a 1.63x107 2.73x1 6 4.46x10 1 1.63x10 7 2.73x1 o6 4.46x101 1.63x10 7 2.73x1 6 4.46x10' 8b 5.32 x 104 2.96 x 1 6 1.57 x 10 1 5.32 x 104 2.96 x 106 1.57 x 10 o1 5.32 x 104 2.96 x 1o6 1.57 x 10 Total 5.00x104 12.0204 5.00x10 4 12.0211 5.00x10 4 12.0216 ILRT Dose Rate 3.04 x 104 1.02 x 1o3 1.54 x 103 from 3a and 3b (+1.86 x 104) (+1.08 x 1j'5) (+2.53 x 10')

0.0025% 0.0085% 0.0128%

% Of Total (+0.00002%) (+0.0001%) (+0.0002%)

Delta Dose Rate from 3a and 3b 5.00 x 104 (10 to 15 yr) _ (+0.0014%)

1.76x10 5.97x104 9.16x10 9 LERF from 3b (+4.09 x 10.1) (+2.38 x 1010)' (+5.56 x 10 1 Delta LERF 3.19 x lo" (10 to 15 yr) (+3.18x10 0 CUP ° 87.98% 88.07% 88.13%

P (+0.0008%) (+0.0167%) (+0.0111 %)

Delta CCFP % (10 0.06%

to 15 yr) 0.06%

_ __ __ _ __ __ _ __ __ _ __ _ _ __ _ _ __ __ __ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ ( 0.00 6 3 % )

Denotes increase from original values presented in Section 2.4, Steps 7, 8, and 9 of this report.

Table 3-4 Containment Steel Liner Corrosion Sensitivity Cases Visual LERF LERF Total LERF Drywelll Inspection Likelihood LERF Increase Increase Increase Age Torus & Non- Flaw is Increase From From From ILRT (Step 2) Breach Visual LERF (EPRI From Corrosion Corrosion Extension (Step 4) Flaws Class 3b) Corrol0 easio (1-in-10 (i to 15 (10 to 15 (Step 4) (Step 5) 3-ICoyrs)

C lass years) years) years)

Base Case Base Case Base Case Base Case Base Case Base Case Doubles 0.8017%Iiner Base Case Base Case 4.09 x 101' 2.38 x 1010 5.56 x 10-10 3.19 x 10 9 every 5 yrs 0.0802%floor 10% 100%

Doubles Base Base Base 1.17x10O' 1 1.99x 10` 0 1.15 x 10' 9 3.82 x 10 9 every 2yrs __ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Doubles Base Base Base 6.08 x 1011 8.33 x 10.11 1.08 x 10.10 2.89x 109 every_1 0_yrs __ _ _ _ _ _

Base Base 5% Base 3.91 x 10' 1 2.28 x 10 0 5.32 x 10 °3179 x10 9 Base Base 15% Base 4.27x 10-1 2.49 x 10° 5.81 x 10- 3.20x 10 4

Base 0.01603%floorl Base Base 8.18 x 10-" 4.76 x 10" 1.1 ix 101" 2.93 x 109 Base 04004%floor' 5 Base Base 2.04 x 10.'° 1.19 x 109 2.78 x 109 4.46 x 10.

Lower Bound Doubles {0.1 6033%liner 1 4 1I 2I0 2I 1 I1 every 10 yrs l0.01603%floor 41 5% l 10% 1.16x10 '2 5.03x10 9.21 x10 l 2.87x 10 Upper Bound Doubles 04.0042% liner l 15 i l I ever 2 yrs 10.4004%floor 1 5 15% 100% 6.11 x 10'" 1.04 x 109 6.02 x 109 7.85 x 10O 10 Base point 10 times lower than base case of 0.0001 at 20 psia.

Base point 10 times higher than base case of 0.01 at 20 psia.

Entergy REPORT No. VY-RPT-04-00002 Revision 1 l Page l Of l 70 l SECTION 4 CONCLUSIONS 4.1 Internal Events Impact A risk assessment of the impact of changing VY Integrated Leak Rate Test (ILRT) interval from the currently approved 1-in-10 year interval to a one-time extension to 1-in-15 years has been performed.

Based on the above results, the following are main conclusions regarding the assessment of the plant risk associated with extending the Type A ILRT test frequency from ten-years to fifteen years:

1. Regulatory Guide 1.174 [5] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 [5] defines very small changes in risk as resulting in increases of CDF below 10 4 /ry and increases in LERF below 10-7/ry. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 1-in-10 years to 1-in-15 years is 2.87 x 10-9 /ry. Since Regulatory Guide 1.174 [5] defines very small changes in LERF as below 10 7 ry, increasing the ILRT interval at VY from the currently allowed one-in-ten years to one-in-fifteen years is non-risk significant from a risk perspective.
2. The increase in risk on the total integrated plant risk as measured by person-rem/reactor year increases for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.004% (0.0005 person-rem/ry).

This value can be considered to be a negligible increase in risk.

3. The change in conditional containment failure probability (CCFP) is calculated to demonstrate the impact on 'defense-in-depth'. The ACCFPj 1 s5 is found to be 0.06%. This signifies a very small increase and represents a negligible change in the VY containment defense-in-depth.

Table 4-1 summarizes the above conclusions.

4.2 External Events Impact Based on the results from Appendix A, "External Event Assessment During an Extension of the ILRT Interval," the following are main conclusions regarding the assessment of the plant risk associated with extending the Type A ILRT test frequency from ten-years to fifteen years:

1. Based on conservative methodologies in estimating the core damage frequency for internal events, seismic events, and fires events, the ALERFcOMBINED1O15 of 3.71 x 10-8/ry from extending the VY ILRT frequency from 1-in-1 0 years to 1-in-15 years is slightly above the 10 7 /ry criterion of Region Ill, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines in NRC Regulatory Guide 1.174 [5].

Therefore, increasing the ILRT interval at VY from the currently allowed 1-in-10 years to 1-in-15 years is non-risk significant from a risk perspective.

2. The combined internal and external events increase in risk on the total integrated plant risk as measured by person-rem/reactor year increases for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.004% (0.007 person-rem/ry). This value can be considered to be a negligible increase in risk.
3. The change in the combined internal and external events conditional containment failure probability from 1-in-1 0 years to 1-in-1 5 years is 0.05%. A change in ACCFP of less than 1% is insignificant from a risk perspective.

Table 4-2 summarizes the above conclusions.

4.3 Containment Liner Corrosion Risk Impact Based on the results from Appendix B, 'Risk Impact of Containment Liner Corrosion During an Extension of the ILRT Interval," the following are main conclusions regarding the assessment of the plant risk associated with extending the Type A ILRT test frequency from ten-years to fifteen years:

1. The impact of including age-adjusted corrosion effects in the ILRT assessment has minimal impact on plant risk and is therefore acceptable.
2. The change in LERF, taking into consideration the likelihood of a containment liner flaw due to age-adjusted corrosion is non-risk significant from a risk perspective. Specifically, extending the interval to 15 years from the current 10 years requirement is estimated to be about 3.19 x 10 9/ry. This is below the Regulatory Guide 1.174 [5] acceptance criteria threshold of 10- 7/ry.
3. The age-adjusted corrosion impact in dose increase is estimated to be 5.0 x 1Q4 person-rem/ry or 0.0042% from the baseline ILRT 10 year's interval.
4. The age-adjusted corrosion impact on the conditional containment failure probability increase is estimated to be 0.6%.
5. A series of parametric sensitivity studies regarding potential age related corrosion effects on the containment steel liner also demonstrated minimal impact on plant risk.

Table 4-3 summarizes the above conclusions.

rh

'Entergy REPORT No. VY-RPT-04-00002 I Revision 1 lPage l 65 l Of I 70 I Table 4-1 Quantitative Results as a Function of ILRT Interval - Internal Events Quantitative Results as a Function of ILRT Interval Current Proposed (1-per-10 year ILRT) (1-per-15 year ILRT)

Dose Aciet Population Dose Aciet Population Dose EPRI (Person-Rem Aciet Rate (Person- Aciet Rate (Person-Class Category Description Within 50 Frequency Rem I Ry Within Frequency Rem / Ry Within

___Category____________ 1 l_ miles)( ) (per ry) 50 miles) (per ry) 50 miles) 1 No Containment Failure(2) 1.30 x 103 5.43 x 10' 7 7.06 x 10-4 5.12 x 10-' 6.65 x 104 2 Containment Isolation System Failure 2.73 x 10 l 2.14 x 10'9 5.85 x 10-3 2.14 x 10 9 5.85 x 10-3 4 8 3a Small Pre-Existing Failures (2). (3) 1.30 x 10 5.73 x 1048 7.46 x 104 8.60 x 10 1.12 x 10'3 3b Large Pre-Existing Failures (2).(3) 4.55 x 104 l 5.73 x 109 2.61 x 104 8.60 x 10 9 3.91 x 105 4 Type B Failures (LLRT) .. 0.00 0.00 0.00 0.00 S Type C Failures (LLRT) l 0.00 0.00 0.00 0.00 6 Other Containment Isolation System Failure .. 0.00 0.00 0.00 0.00 7

7a Containment Failure Due to Severe Accident (a)(4) 2.73 x 1o6 3.22 x 107 8.79 x 1O" 3.22 x 10i 8.79 x 101 6 1 7b Containment Failure Due to Severe Accident (b)(4) 2.73 x 106 3.85 x 10- 1.05 x 10+ 3.85 x 10.6 1.05 x 10+1 7c Containment Failure Due to Severe Accident (c) (4) 3.53 x I 0o 9.71 x 109 3.43 x 10.2 9.71 x 10 9 3.43 x 10.2 7d Containment Failure Due to Severe Accident (d)(4) 2.73 x 106 1.63 x 10 7 4.46 x 10.1 1.63 x 10.7 4.46 x 101 8 Containment Bypass Accidents 2.96 x 106 5.32 x 108 1.57 x 101 5.32 x 1048 1.57 x 10.1 TOTALS: 5.00 x 10.6 12.0211 5.00 x i04 12.0215 Increase In Dose Rate O00004 Increase in LERF Increase in CCFP (%)

~Entergy REPORT No. VY-RPT-04-00002 I Revision IIPagel 66 l Ofl 70 l i

Table 4-2 Quantitative Results as a Function of ILRT Interval - Internal and External Events Quantitative Results as a Function of ILRT Interval Current Proposed (1-per-10 year ILRT) (1-per-15 year ILRT)

EPRI (PesoRem Accident Population Dose Accident Population Dose Category Descnption (Person-Rem Frequency Rate (Person- FrRauency Class aeoyecpinWithin 50 Fpreqeny) Rem / Ry Within Fpreqeny) RemteI (Person-Ry Within

__ __ __ __ __ __ _ _ _ _ _ _ _ _ _ _ miles)") pe y 50 miles) (p rr)50 miles) 1 No Containment Failure 1.30 x 7.40 x 10-6 3 4 (2) 103 9.62 x 10' 6.99 x 10 9.08 x 10' 3 2 Containment Isolation System Failure 2.73 x 106 2.90 x 10.8 7.93 x 10.2 2.90 x 10 8 7.93 x 10.2 3a Small Pre-Existing Failures (2)t (3) 1.30 x 104 7.42 x 10 7 9.64 x 10'3 1.11 X 104 1.45 x 10-2 4

3b Large Pre-Existing Failures (2).(3) 4.55 x 10 7.42 x 10. 3.38 x 10-3 1.11 X 10' 7

5.06 x 10' 3

4 Type B Failures (LLRT) 0.00 0.00 0.00 0.00 5 Type C Failures (LLRT) 0.00 0.00 0.00 0.00 6 Other Containment Isolation System Failure 0.00 0.00 0.00 0.00 4

7a Containment Failure Due to Severe Accident (a)") 2.73 x 106 4.36 x 10.6 1.19 x 10 4.36 x 10"6 1.19 x 10' 7b Containment Failure Due to Severe Accident (b) (4) 5 2.73 x 106 5.21 x 10 5 1.42 x 102 5.21 x 10 1.42 x 102 7 7 7c Containment Failure Due to Severe Accident (c)(4) 3.53 x 106 1.31 x 10 4.64 x 10. 1.31 x 10 4.64 x 1O" 7d Containment Failure Due to Severe Accident (d) (4) 2.73 x 106 2.21 x 10O6 6.06 x 100 2.21 x 10." 6.06 x 100 8 Containment Bypass Accidents 2.96 x 106 7.20 x 10 ' 2.13 x 100 7.20 x 10 ' 2.13 x 100 TOTALS: 6.77 x 10-' 162.7811 6.77 x 105 162.7871 Increase in Dose Rate Increase in LERF Increase in CCFP (%)

Table 4-3 Quantitative Results as a Function of ILRT Interval - Liner Corrosion Impact Quantitative Results as a Function of ILRT Interval Current Proposed (1-per-1 Dyear ILRT) (1-per-15 year ILRT)

Dose Population Dose Population Dose (Person-Rem Accident Rate (Person- Accident Rate (Person-EPRI Within 50 Frequency Rem / Ry Within Frequency Rem / Ry Within Class Category Description miles)(') (per ry) 50 miles) (per ry) 50 miles) 1 No Containment Failure(2) 1.30x 103 5.43 x l0'7 7.40 x 104 5.11 x 10'7 6.65 x 104 3

Containment Isolation System Failure 2.73 x 106 2.14 x 10'9 5.65 x 10- 2.14 x 10'9 5.85 x 10-3 4

3a Small Pre-Existing Failures (2). (3) 1.30 x 10 5.73 x 10-8 7.73 x 10 4 8.60 x 10o8 1.12 x 10-3 3b Large Pre-Existing Failures (2),(3) 4.55 x 104 5.97 x 10 9 2.82 x 10 4 9.16 x 10-9 4.17 x 10-4 4 Type B Failures (LLRT) .. 0.0 0.00 0.0 0.0 5 Type C Failures (LLRT) .. 0.0 0.00 0.0 0.0 6 Other Containment Isolation System Failure .. 0.0 0.00 0.0 0.0 7

7a Containment Failure Due to Severe Accident (a)(4) 2.73 x 106 3.22 x 10 7 8.91 x 101 3.22 x 10- 8.79 x 10 7b Containment Failure Due to Severe Accident (b) (4) 2.73 x 106 3.85 x 10-6 9.99 3.85 x 1O- 6 1.05 x 101 7c Containment Failure Due to Severe Accident (c) (4) 3.53 x 108 9.71 x 10 9 3.48 x 10.2 9.71 x 10 9 3.43 x 10.2 8a Containment Failure Due to Severe Accident (d)t41 2.73 x 106 1.63 x 4.16 x 10.1 1.63 x 7 10 4.46 x 10l 8b Containment Bypass Accidents 2.96 x 106 5.32 x 10.8 1.57 x 101 5.32 x 10.8 1.57 x 101 TOTALS: 5.00 x 10" 12.0211 5.00 x 10T 12.0216 Increase in Dose Rate Increase in LERF Increase in CCFP (%)

Notes to Tables 15.16, and 17:

1) The population dose associated with the Technical Specification Leakage is based on the analysis provided in Reference [24].
2) Only EPRI accident classes 1, 3a, and 3b are affected by ILRT (Type A) interval changes.
3) Dose estimates for EPRI Class 3a and 3b, per the NEI Interim Guidance, are calculated as 10 times EPRI Class 1 dose and 35 times EPRI Class 1 dose, respectively.
4) EPRI Class 7, containment failure due to severe accident, was subdivided into four subgroups based on VY Level 2 containment failure modes for dose allocation purposes. Note that this EPRI class is not affected by ILRT interval changes.

LEntergy REPORT No. VY-RPT-04-00002 Revision 1 lPage l69 l Of l 70 SECTION 5 REFERENCES (1) Nuclear Energy Institute, 'Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J", NEI 94-01, July 26,1995, Revision 0.

(2) Electric Power Research Institute, "Risk Assessment of Revised Containment Leak Rate Testing Intervals", EPRI TR-104285, August 1994.

(3) Letter from A. Petrangelo (NEI) to NEI Administrative Points of Contact, "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leak Rate Test Surveillance Intervals", November 13, 2001.

(4) Letter from A. Petrangelo (NEI) to NEI Administrative Points of Contact, "One-Time Extension of Containment Integrated Leak Rate Test Interval - Additional Information", November 30, 2001.

(5) U.S. Nuclear Regulatory Commission, "An Approach for Using Probabilistic Risk Assessment In Risk-informed Decisions On Plant-Specific Changes to the Licensing Basis", Regulatory Guide 1.174, November2002, Revision 1.

(6) Entergy Nuclear Northeast, 'Vermont Yankee PSA Model Update (VY04R1)", Engineering Report Number, VY-RPT-04-00004, September 24, 2004 Revision 0.

(7) U.S. Nuclear Regulatory Commission, 'Performance-Based Containment Leak-Test Program",

NUREG-1493, September 1995.

(8) Entergy Nuclear Northeast, "Risk Impact Assessment of Extending Containment Type A Test Interval", Calculation Number, IP3-CALC-VC-03357, January 3, 2001 Revision 0.

(9) Florida Power - Progress Energy, Crystal River Nuclear Plant Letter of June 20, 2001,

'Supplemental Risk Informed Information in Support of License Amendment Request No. 267".

(10) U.S. Nuclear Regulatory Commission, "Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants, NUREG-i150", December 1990.

(11) Payne et al., "Evaluation of Severe Accident Risks: Peach Bottom, Unit 2," NUREG/CR-4551, SAND86-1309, Volume 4, Rev 1, Part 1, December 1990.

(12) Entergy Nuclear Northeast, "Risk Impact Assessment of Extending Containment Type A Test Interval", Calculation Number, JAF-CALC-PC-04544, June 2, 2003 Revision 0.

(13) Entergy Nuclear Northeast, "Vermont Yankee Nuclear Power Station Updated Final Safety Analysis," Chapter One, Revision 17.

(14) Safety Evaluation by the Office of Nuclear Reactor Regulation,

Subject:

TS Amendment 152 (Appendix J Option B), dated February 26, 1998 (NVY 98-24).

(15) Vermont Yankee Nuclear Power Station Updated Final Safety Analysis, Table 5.2.1, Revision 17.

Enfte8rgy REPORT'No. VY-RPT-04-00002 Revision 1 I Page I 70 I Of l 70 (16) United States Nuclear Regulatory Commission, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10CFR 50.54(f)," Generic Letter 88-20, Supplement 4, June 28,1991.

(17) Vermont Yankee Nuclear Power Station Individual Plant Examination for External Events (IPEEE)

Report - Response to Generic Letter 88-20, Supplement 4, June 30, 1998.

(18) United States Nuclear Regulatory Commission, 'Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities,"

NUREG-1407, June 1991.

(19) Electric Power Research Institute, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," EPRI NP-6041-SL, Revision 1, August 1991.

(20) Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Letter from Mr. C. H. Cruse (Calvert Cliffs Nuclear Power Plant) to NRC Document Control Desk, March 27, 2002.

(21) Entergy Nuclear Northeast, 'Vermont Yankee Nuclear Power Station Primary Containment Leakage Rate Testing Program," (VY Program Procedure PP-7006, Revision 7s), November 18, 2003.

(22) R.P Kennedy, "Overview of Methods for Seismic PRA and Margin Analysis Including Recent Innovations", Proceedings of the OECD-NEA Workshop on Seismic Risk, Tokyo, Japan, August 1999.

(23) Electric Power Research Institute, "Probabilistic Seismic Hazard Evaluation at Nuclear Plant Sites in the Central and Eastern United States: Resolution of the Charleston Issue," EPRI NP-6395-D, April 1989.

(24) Vermont Yankee Calculation VYC-2371, Revision 0, "MACCS2 Model for VYNPS" (25) ERIN Engineering and Research, Inc., Report C1 209201-731-121793, 'Vermont Yankee Individual Plant Examination (IPE) Level 2 - Tier 2", December 1993 (26) EPRI TR 100370, "Fire Induced Vulnerability Evaluation (FIVE)," April 1992.

(27) Letter, VYNPC to USNRC, BVY 99-162, "Response to Request for Additional Information Concerning VY-IPEEE", December 1999.

(28) Vermont Yankee Nuclear Power Station, "Reactor Containment Building Integrated Leakage Rate Test," July 1995.

(29) Letter, VYNPC to USNRC, BVY 93-139, 'Vermont Yankee Response to Generic Letter 88-20:

Report on the Individual Plant Examination for Severe Accident Vulnerabilities - 10CFR50-54(f)",

December 1993

Appendix A External Event Assessment During an Extension of the ILRT Interval

Table of Contents Page No.

A1.0 Introduction A-2 A2.0 VY IPEEE Seismic Analysis A-2 A2.1 Seismic Analysis Methodology Selection A-2 A2.2 Seismic Analysis Assumptions A-2 A2.3 Seismic Analysis Input A-3 A2.4 Seismic Analysis Method of Analysis A-4 A3.0 VY IPEEE Fire Analysis A-6 A3.1 Fire Analysis Methodology Selection A-6 A3.2 Fire Analysis Method of Analysis A-6 A4.0 Other External Hazards A-8 A5.0 Effect of External Events Hazard Risk on ILRT Risk Assessment A-8 A5.1 Effect of External Events Hazard Risk on ILRT Risk Assessment Assumptions A-8 A5.2 Effect of External Events Hazard Risk on ILRT Risk Assessment Method of Analysis A-8 A5.3 Effect of External Events Hazard Risk on ILRT Risk Assessment Input A-9 A6. Conclusions A-1 8 Tables Page No.

Table A Vermont Yankee's EPRI Site Seismic Hazard Curve A-3 Table A Vermont Yankee's IPEEE Fire Events - Core Damage Frequency Results A-6 Table A Vermont Yankee's IPEEE Estimated Level 2 Release Frequencies As a Function of EPRI Accident Class A-9 Table A Effect of External Events Hazard Risk on VY ILRT Risk Assessment A-1 9 Table A Effect of Internal and External Events Risk on VY ILRT Risk Assessment A-20

A1.0 Introduction This appendix discusses the risk-implication associated with external hazards in support of the VY Integrated Leak Rate Testing (ILRT) interval extension risk assessment.

In response to Generic Letter 88-20, Supplement 4 [16], VY submitted an Individual Plant Examination of External Events (IPEEE) in June 1998 [17]. The IPEEE was a review of external hazard risk (i.e., seismic, fires, high winds, external flooding, etc) to identify potential plant vulnerabilities and to understand severe accident risks. The results of the VY IPEEE are therefore used in this risk assessment to provide a comparison of the effect of external hazards when extending the current 1-in-10 years to 1-in-15 years Type A ILRT interval.

A2.0 VY IPEEE Seismic Analysis A2.1 Seismic Analysis Methodology Selection The VY plant has been designed to accommodate a safe-shutdown earthquake (SSE) with 0.14g peak ground acceleration (PGA). The seismic analysis performed in the IPEEE study is intended to act as a performance check on the design, estimating seismic capacity beyond the SSE.

The seismic analysis methodology implemented for VY satisfied the NRC requirements for performing a seismic IPEEE as presented in Generic Letter 88-20, Supplement 4 [19]. The methodology comprises an NRC seismic margin assessment (SMA) following the guidance of NUREG-1407 "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," [18] and EPRI NP-6041 "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," [19] and a containment performance analysis. A seismic margin can be expressed in terms of the earthquake motion level that compromises plant safety--the seismic margin assessment determines whether there is high confidence that the plant can survive a given earthquake. No core damage frequency sequences were quantified as part of the IPEEE seismic risk analysis.

Seismic events were evaluated using the Seismic Margins Analysis (SMA) method. The SMA methodology uses a deterministic approach to identify the weakest components in terms of a High Confidence Low Probability of Failure (HCLPF) peak ground acceleration.

The conclusions of the VY IPEEE seismic risk analysis are as follows:

  • For Vermont Yankee, the SMA identified that the lowest HCLPF components in the selected primary and alternate safe shutdown paths are the Condensate Storage Tank (CST) with a HCLPF of .25g and the Diesel Fuel Oil Storage Tank (FOST) with a HCLPF of .29g. These values, although below the 0.3g review level earthquake, represent significant margin to the design basis 0.14g earthquake.

The HCLPF for all other components in the safe shutdown paths meet or exceed the 0.3g review level earthquake upon resolution of A-46 outliers.

A2.2 Seismic Analysis Assumltions

1) The Simplified Hybrid Method as presented in OECD-NEA Workshop on Seismic Risk, 'Overview of Methods for Seismic PRA and Margin Analysis Including Recent Innovations" [22] is used to approximate the VY seismic-induced core damage frequency based on the seismic margin analysis results found in the VY IPEEE submittal [17].

APPENDIX A A2.3 Seismic Analysis Input

1) Based upon the results from the SMA performed forthe Vermont Yankee IPEEE, the individual HCLPF values for the following components will be used to determine the risk impact of seismic events [17]:

Component Description Component HCLPF Fuel Oil Storage Tank (FOST) 0.29g PGA Condensate Storage Tank (CST) 0.25g PGA

2) The 10% NEP standard normal variable is -1.282. This value is derived from the OECD-NEA Workshop on Seismic Risk [22].
3) The 1% NEP standardized normal variable is -2.326. This value is derived from the OECD-NEA Workshop on Seismic Risk [22].
4) The Simplified Hybrid method presented in OECD-NEA Workshop on Seismic Risk [22] recommends a variable factor 3 equal to 0.3 to estimate the plant damage seismic risk.
5) The seismic hazard curve for the VY site, based upon EPRI NP-6395-D, "Probabilistic Seismic Hazard Evaluation at Nuclear Plant Sites in the Central and Eastern United States: Resolution of the Charleston Issue" [23], is summarized in tabular form in Table A-1.

Table A-1 Vermont Yankee's EPRI Site Seismic Hazard Curve Acceleration Frequency of Exceedance (/yr)

(g) 15% 50% 85% Mean 0.05 2.04E-04 7.06E-04 2.04E-03 1.29E-03 0.08 9.15E-05 3.31 E-04 l .04E-03 6.74E-04 0.15 1.99E-05 7.68E-05 2.92E-04 1.96E-04 0.26 4.87E-06 2.32E-05 1.01 E-04 6.91 E-05 0.31 2.71E-06 1.40E-05 6.62E-05 4.58E-05 0.41 9.84E-07 6.06E-06 3.22E-05 2.29E-05 0.51 3.9913-07 3.01E-06 1.72E-05 1.28E-05 0.66 1.20E-07 1.19E-06 7.57E-06 6.15E-06 0.82 4.27E-08 5.30E-07 3.88E-06 3.32E-06 1.02 1.14E-08 2.01 E-07 1.7813-06 1.64E-06

A2.4 Seismic Analysis Method of Analysis Although quantitative risk information is not directly available from the VY SMA IPEEE analysis, a paper presented by Robert P. Kennedy [22] provides a method called the Simplified Hybrid Method for obtaining a seismic-induced CDF estimate based on results of an SMA analysis. This methodology requires only the plant HCLPF to estimate the seismic CDF. The approach entails the following steps:

Step 1A - Determine the HCLPF seismic capacity CHCLPF from the SMA analysis.

Step 2A - Estimate the 10% conditional probability of failure capacity C10O..

Step 3A - Determine hazard exceedance frequency H10%that corresponds to CIO% from the hazard curve.

Step 4A - Determine seismic accident type risk PF.

Step 5A - Determine the seismic core damage frequency (SCDF).

Step 1A - Determine the HCLPF seismic capacity CHCLPF from the SMA analysis The VY seismic analysis found two components with values above the review level earthquake of 0.3g; fuel oil storage tank and condensate storage tank. The respective HCLPF values are:

CHCLPF-FOST = 0.29g CHCLPF-CST = 0.25g However, the above values do not directly consider the effect of random failures on seismic risk because it uses the HCLPF Maximum/Minimum method to approximate the seismic accident type fragility. Per the OECD-NEA Workshop on Seismic Risk [22] methodology a HCLPF reduction factor is applied to each of the above seismic component to account for non-seismic failures and human errors.

Based on an examination of the seismic IPEEE results a HCLPF reduction factor of 0.7 was selected. As a result, the revised seismic accident type HCLPF values are:

CHCLPF-FOST = 0.29g / 0.7 = 0.41 g CHCLPF-CST = 0.25g / 0.7 = 0.36g Step 2A - Estimate the 10% conditional probability of failure capacity CI0%

Per the work presented in OECD-NEA Workshop on Seismic Risk [???], the 10% conditional probability of failure capacity is calculated as follows:

C I01/ = F3 CHCLPF Where:

C10% = 10% conditional probability of failure capacity CHCLPF = seismic component type HCLPF capacity by the CDFM (Conservative Deterministic Failure Margin) method F3 = e(NEPIO%NEPI%)3 Where:

NEP10% = 10% NEP standard normal variable is = -1.282 [Input #2]

NEP1% = 1% NEP standard normal variable is = -2.326 [Input #3]

F3 = variable factor = 0.3 [Input #4]

Therefore, F3 = [(1282)-(2.326)1 03 = 1.37 and C1-O.FOST = 1.37

  • CHCLPF-FOST = 1.37
  • 0.41 = 0.56g PGA C0I.CST = 1.37
  • CHCLPF-CST = 1.37
  • 0.36 = 0.49g PGA Step 3A - Determine hazard exceedance frequency HI that corresponds to Ct; from the hazard curve The seismic hazard curve for the VY site, as presented in Table A-1 is used to determine the hazard exceedance frequency H10%that corresponds to C1O%. These are as follows:

Acceleration Mean Frequency of Seismic Accident Type (q) Exceedance (Ury) 0.56 1.06E-6 Hlo%FosT 0.49 1.28E-5 Step 4A - Determine seismic accident type risk PF Per the OECD-NEA Workshop on Seismic Risk [22], the seismic accident type risk is calculated as follows:

PF-FOST = 0-5

  • H1OFOST = 0.5
  • 1.06E-6 = 5.3E-7/ry PF-CST = 0.5
  • 1.28E-5 = 6.4E-6/ry Because any seismic event results in a loss-of-offsite power, a subsequent failure of the fuel oil storage tank results in a seismic-induced station blackout. Therefore, PF-FOST represents the seismic-induced station blackout frequency of 5.3E-7/ry and is assigned a core damage accident bin of S-IBL ('Late' SBO.

Core cooling is maintained by HPCI/RCIC until batteries deplete).

Similarly; the seismic failure of the condensate storage tank (PF-CST) implies successful emergency diesel generators operation following the occurrence of the seismic-induced loss-of-offsite power. Therefore, this implies a seismic-induced core damage progression resulting in the loss of reactor pressure vessel injection, and is assigned into the core damage accident class 'Transient sequences with loss of all high-pressure injection and failure to depressurize. Core damage occurs with the reactor at high-pressure' (S-IA).

Step 5A - Determine the seismic core damage frequency (SCDF)

The step involves the summation of the individual seismic accident types frequencies.

SCDF = PF-FOST + PF-CST SCDF = 5.3E-7Iry + 6.4E-6/ry SCDF = 6.93E-6/ry This information is used in Section A5.0 of this appendix to provide insight into the impact of external hazard risk on the conclusions of this ILRT risk assessment.

A3.0 VY IPEEE Fire Analysis A3.1 Fire Analysis Methodology Selection The Fire analysis performed for the VY IPEEE submittal [17] used the EPRI Fire Induced Vulnerability Evaluation Methodology (FIVE) [26]. The fire PRA analysis entailed the identification of critical areas of vulnerability, the calculation of fire initiation frequencies, the identification of fire-induced initiating events and their impact on systems, the disabling of critical safety functions, and potential fire-induced containment failure. Based on this examination, the Core Damage Frequency from internal fires is estimated to be 5.58 x 10-5/ry [27J.

The VY IPEEE concluded that fire-induced CDF sequences are dominated by the following fire events.

  • Cable Vault fire involving redundant divisions of equipment, which require Control Room evacuation and implementation of Alternate Shutdown methods. Automatic or manual suppression is not credited for these fire events based on fire damage calculations using the FIVE method.

o Control Room fire with failure of manual suppression, where Control Room evacuation and implementation of Alternate Shutdown methods are required.

a Switchgear Room fire with failure of automatic suppression. Based on fire damage calculations using the FIVE method, an unsuppressed Switchgear Room fire could result in a loss of normal power, failure of an emergency bus, and degradation of the Vernon Tie. Station power relies on one Emergency Diesel Generator (EDG) and its associated emergency bus.

A3.2 Fire Analysis Method of Analysis The VY IPEEE submittal [17,27] for the fire induced core damage scenarios and the associated frequency results were reviewed in support of this assessment. The result is judged to be conservative because of limited data and conservative fire propagation and mitigate assumptions. The CDF results for all the compartments, which were quantitatively evaluated, are provided in Table A-2.

This information is used in Section A5.0 of this appendix to provide insight into the impact of external hazard risk on the conclusions of this ILRT risk assessment.

Table A-2 Vermont Yankee's IPEEE Fire Events - Core Damage Frequency Results Firee Building/ Description of Fire Event(s) Fire Area Compartment CDF (Iry)

Reactor RBNEC Northeast ECCS Comer Room, El. 213' and 232' 3.80E-09 Building RBHP HPCI Room, El. 213' 9.OOE-09 RBRCL Lower RCIC Comer Room, El. 213' at NW Comer 6.70E-08 RBRCU Upper RCIC Comer Room, El. 232' at NW Comer 4.50E-08 RBSEC Southeast ECCS Comer Room, El. 213' and 232' 1.OOE-08 RBSWC Southwest CRD Comer Room, El. 213' and 232' (treated as part of RB4) See RB4 RBI Torus Room, El. 213', Zone RB1 (north) 1.30E-07 RB2 Torus Room, El. 213', Zone RB2 (south) 7.40E-07 RB3 Reactor Building, El. 252', Zone RB3 (north), Self-ignited Cable Fire 5.10E-06 Reactor Building, El. 252', Zone RB3 (north), In-situ MCC Fire

Reactor Building, El. 252', Zone RB3 (north), transient lube oil spill Reactor Building, El. 252', Zone RB3 (north), transient/in-situ Class A trash fire Reactor Building, El. 252', Zone RB4 (south), Self-ignited Cable Fire Reactor Building, El. 252', Zone RB4 (south), CRD Repair Room Fire RB4 Reactor Building, El. 252', Zone RB4 (south), In-situ MCC Fire 3.30E-06 Reactor Building, El. 252', Zone RB4 (south), transient lube oil spill Reactor Building, El. 252', Zone RB4 (south), transient/in-situ Class A trash fire RBMG Reactor Building, El. 280', Recirc. MG Set Fire 3.40E-07 RB5 Reactor Building, El. 280', Zone RB5 (north) 7.30E-07 RB6 Reactor Building, El. 280', Zone RB6 (south) 3.50E-07 RB303 Reactor Building, El. 303' 4.90E-07 RB31 8 Reactor Building, El. 318' 1.90E-08 RB345 Reactor Building, El. 345' 1.50E-09 RBSZ-S1 Reactor Building, El. 252', Separation Zone Div. SI Trays 6.50E-07 RBSZ-S2 Reactor Building, El. 252', Separation Zone Div. S2 Trays 6.50E-07

  • + $

West Switchgear Room at El. 248', Bus 1/8 Fire SGW West Switchgear Room at El. 248', Bus 3 Fire 9.OOE-06 West Switchgear Room at El. 248', T-8 Transformer Fire East Switchgear Room at El. 248', Bus 2/9 Fire SGE East Switchgear Room at El. 248', Bus 4 Fire 7.OOE-06 Control West Switchgear Room at El. 248', T-9 Transformer Fire Building Cable Vault, El. 262', Division S1 Panel Fire Affecting Division S2 Cable Trays CV Cable Vault, El. 262', Division S2 Panel Fire Affecting Division S1 Cable 1.50E-05 Trays Cable Vault, El. 262', Self-ignited Cable Fire CVBT Cable Vault Battery Room, El. 262' 3.20E-06 CR Control Room, El. 272' 5.70E-06 DGA Emergency Diesel Generator Room A 4.50E-07 Turbine DGB Emergency Diesel Generator Room B 4.60E-07 Building TURB Turbine Building, All General Areas 1.1 OE-06 WMACH Machine Shop and Stores Warehouse - South Turbine Building Intake & INTCW Circulating Water Pump Room Fire - Intake Structure 1.60E-09 Discharge INTSW Service Water Pump Room Fire - Intake Structure 3.1 OE-07 Structure DISCH Discharge Structure Fire 9.40E-10 RADW FRADW Radwaste Building Fire Radwaste 5.20E-08 RWC FRWC Radwaste Corridor Fire AOG Advanced Off Gas Building Fire 1.40E-07 DGOP EDG Fuel Oil Storage Tank and Transfer Pump House Fire 1.20E-08 Misc. FOB Office Building - North End of Turbine Building Structures RHOUSE Relay and Metering House - 345 kV Switchyard 4.OOE-07 MTFRM Main/Aux. Transformer Fire W/Propagation to Turbine Building 6.80E-08 STFRM Startup Transformer Fire W/Propagation to Turbine Building 2.80E-07 TOTAL 5.58E-05

A4.0 Other External Hazards The VY IPEEE submittal [17], in addition to the internal fires and seismic events, examined a number of other external hazards:

  • External Flooding
  • Transportation and Nearby Facility Accidents
  • Aircraft Hazard
  • Severe Weather & Lightning No risks to the plant occasioned by high winds and tornadoes, external floods, ice, and hazardous chemical, transportation and nearby facility incidents were identified that might lead to core damage with a predicted frequency in excess of 104/year. Therefore, these other external event hazards are not included in this appendix and are expected not to impact the conclusions of this ILRT risk assessment.

A5.0 Effect of External Events Hazard Risk on ILRT Risk Assessment A5.1 Effect of External Hazard Risk on ILRT Risk Assessment Assumptions Because both the seismic margin analysis and fire PRA did not report any seismic-induced accident progression releases, for the purpose of this report the EPRI accident classes for External Events will be based on percent contribution for the accident class frequencies for Internal Events as presented in Table 2-3.

Both the seismic margin analysis and the fire PRA are dominated by non-recoverable accident sequences that would be expected to result in large early release. Specifically, non-recoverable station blackout accident sequences, and loss of high-pressure injection systems with failure to depressurize. From the VY PSA Level 2 analysis for internal events, these class of accidents (i.e.: ID, IA, IED, IBE and IBL) dominate the occurrence of a large early release.

Per the NEI Guidance Document [4], Enclosure 1, Discussion of Conservatisms in Quantitative Guidance for Delta LERF Impact," specific accident sequences that independently cause a LERF or could never cause a LERF, are to be removed from Class 3b LERF evaluation. Therefore, for the external events impact on the ILRT risk assessment, the evaluation of LERF is performed by multiplying the Class 3b probability by only that portion of core damage frequency that is impacted by Type A ILRT.

A5.2 Effect of External Events Hazard Risk on ILRT Risk Assessment Input

1) Based on the examination in Sections A2.0 through A4.0, the VY external event initiated CDF is approximately 5.58 x 10 5/ry (internal fires) + 6.93 x 104 /ry (seismic) = 6.27 x 105/ry.
2) Based on Section A5.1, the following Level 2 release frequencies for external events is based upon the distribution determined for the EPRI Class frequencies of internal events as shown in Table 2-3.

Table A-3 Vermont Yankee's IPEEE Estimated Level 2 Release Frequencies As A Function of EPRI Accident Class EPRI Class l IfExtenal Accident Class Description Events Frequency

______f j (ry) 1 No Containment Failure 7.37 x 1006 2 Large Containment Isolation Failures (Failure-to- 2.69 x 1O' close) -Il 3a Small Isolation Failures (Liner breach) 2.16 x 10O7 3b Large Isolation Failures (Liner Breach) 2.16 x 10' 4 Small isolation failure - failure-to-seal (Type B test) 5 Small isolation failure - failure-to-seal (Type C test) 6 Containment Isolation Failures (dependent failures, personnel errors) l _l 7a Severe Accident Phenomena Induced Containment 4.04 x 1046 Failure - energetic phenomena 7b Severe Accident Phenomena Induced Containment 82 s Failure - direct contact with core material 4.8 x 1 Severe Accident Phenomena Induced Containment 7c Failure - containment failures due to loss of heat 1.22 x 10" removal 8a Containment Bypassed (ATWS) . 2.05 x 1o l 8b Containment Bypassed (ISLOCA, LOCAOC) 6.67 x 107 TOTAL 6.27 x 1045 A5.3 Effect of External Events Hazard Risk on ILRT Risk Assessment Method of Analysis The VY IPEEE external events risk information presented in Sections A2, A3 and A4 is used to calculate the following, in accordance with the NEI Interim Guidance [3]:

1) Evaluate the risk impact for the New Surveillance Intervals of Interest
2) Evaluate the external hazard risk impact in terms of LERF
3) Evaluate the external hazard change in conditional containment failure probability
1) Evaluate the risk impact for the New Surveillance Intervals of Interest.

This step calculates the percentage of the total dose rate attributable to EPRI accident Classes 3a and 3b (those accident classes affected by change in ILRT surveillance interval) and the change in this result dose rate from the base dose rate attributable to changes in ILRT surveillance interval.

The change in population dose rate is calculated as outline in Step 7 (section 2.4.7, page 46 of 80) of this report. The results of this calculations is presented below as follows:

For 3-in-1 0 years (internal fires and seismic event),

EPRI Person-Rem Frequency Dose Rate Class within 50 miles (Iry) (Person-Rem/ry) 1 1.30 x 10 3 7.37 x 10 o6 9.58 x 10'3 2 2.73 x 106 2.69 x 1048 7.34 x 10-2 3a 1.30 x 10 4 2.16 x 104 7 2.81 x 103 3b 4.55 x 10 4 2.16 x 104m9.82 x 10'4 4 --- -

5 ---

6 ---

7a 2.73 x 106 4.04 x 10o6 1.10 X lo, 7b 2.73 x 10 6 4.82 x 104 5 1.32 x 102 7c 3.53 x 10 1.22 x 1047 4.30 x 101 8a 2.73 x 10 6 2.05 x 10 o6 5.59 x10 0 8b 2.96 x 10 6

6.67 x 104 7 1.97 x 100 TOTALS 6.27 x 1045 1.507525 x 102 For 1-in-i 0 years (internal fires and seismic event),

EPRI Person-Rem Frequency Dose Rate Class within 50 miles (Vry) (Person-Rem/ry) 1 1.30 x 10 3 6.85 x 10o6 8.91 x 10'3 2 2.73 x 106 2.69 x 1048 7.34 x 10-2 3a 1.30 x 104 6.85 x 10 7 8.90 x 10-3 3b 4.55 x 10 4 6.85 x 1 08 3.11 x 10'3 4 --- --

5 --- --

6 --- -----

7a 2.73 x 10 4.04 x 10 ' 1.10 x lo' 7b 2.73 x 106 4.82 x 10°0 5 1.32 x 1 2 7c 3.53 x 1o6 1.22 x 10' 4.30 x 101 '

8a 2.73 x 10 6 2.05 x 1 -06 5.59 x 100 8b 2.96 x 10 6.67 x 10'° 1.97 x 100 TOTALS 6.27 x 10.05 1.507600 x 102 For 1-in-15 years (internal fires and seismic event),

EPRI Person-Rem Frequency Dose Rate Class within 50 miles (/ry) (Person-Remlry) 1 1.30 x 103 6.48 x 100 6 8.42 x 103 2 2.73 x 106 2.69 x 104 7.34 x 102 3a 1.30 x 104 1.03 x 1O' 1.33 x 10 2 o

4 3b 4.55 x 10 1.03 x 104 4.67 x 104 4 ---

5 --- -

6 --- -

7a 2.73 x 1 o` 4.04 x 1O 1.10 X lo' 7b 2.73 x 10o 4.82 x 100 1.32 x 1 2 7c 3.53 x 106 1.22 x 1007 4.30 x 10-'

8a 2.73 x 106 2.05 x 1O 5.59 x 100 8b 2.96 x 106 6.67 x 10 7 1.97 x 100 TOTALS 6.27 x 10°S 1.507655 x 102 Based on the results summarized above and those presented in Table 2-9 (see page 57 of 80), for the current VY 1-inlO years ILRT interval, the percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b is calculated as follows:

PERCTDCOIB-IO =

L

-3a - DOSET 11 810 + CLAS -3

~TOTAL_ DOSEcOjIBIO DOSECO, 11 B-10 1*

100%

Where:

PER-CTDcoMB-Io = combined internal and external events percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given an 1-in-10 years ILRT interval CLASS_3aDOSE COMB-10 = combined internal and external events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= CLASS_3a_DOSE INTERNAL-10 + CLASS_3a-DOSE EXTERNAL-10 CLASS_3bDOSE COMB-10 = combined internal and external events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= CLASS_3b_DOSE INTERNAL-10 + CLASS_3LDOSE EXTERNAL -10 CLASS_3aDOSE INTERNAL-I 0 = internal events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval = 7.46 x 10 4 /ry [Table 2-9]

CLASS_3bDOSE INTERNAL-I = internal events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval = 2.61 x 104 /ry [Table 2-9]

CLASS_3aDOSE EXTERNAL- 10 = external events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval = 8.90 x 103 person-rem/ry [See 1-in-10 years table above]

CLASS_3b_DOSE EXTERNAL 1a = external events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval = 3.11 x 103 person-rem/ry [See 1-in-10 years table above]

TOTALDOSECOMBI10 = Total combined internal and external events dose rate for all EPRl's Classes given a 1-in-10 years ILRT interval

= TOTALDOSEINTERNAL-1o+ TOTALDOSEExTERNAMlO TOTALDOSE INTERNAL-10 = Total intemal events dose rate for all EPRI's Classes given a 1--

in-1 0 years ILRT interval = 12.0211 (person-rem/ry) [Table 2-9]

TOTALDOSE EXTERNAL-10 = Total external events dose rate for all EPRI's Classes given a 1-in-1 0 years ILRT interval = 150.7600 (person-rem/ry) [See 1-in-10 years table above]

Therefore, PER - CT 0 _(.46X1+8.90x10-)+(2.61x10 4+3.11x13)1*0%

_~- Cl=(12.0211 + 150.7600) 0 PER_CTD COMB-10 = 0.0080%

The percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b based on the proposed 1-in-15 years ILRT interval is calculated as follows:

PER-CTDCOMB-5 = CLASS 3a - D SEc-oC.B-5+CLSS -3bD

- CO. ]

  • 100%

Where:

PERCTDcoMB.15 = combined internal and external events percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given an 1-in-i5 years ILRT interval CLASS_3a_DOSE COMB-15 = combined internal and external events EPRI accident Class 3a dose rate given a 1-in-15 years ILRT interval

= CLASS_3a_DOSE INTERNAL-15 + CLASS_3aDOSE EXTERNAL-15 CLASS_3bDOSE COMB-15 = combined internal and external events EPRI accident Class 3b dose rate given a 1-in-15 years ILRT interval

= CLASS_3bDOSE INTERNAL-15 + CLASSSbDOSEEXTERNAL-15 CLASS_3aDOSE INTERNAL-15 = internal events EPRI accident Class 3a dose rate given a 1-in-15 years ILRT interval = 1.12 x 10 3 person-rem/ry [Table 2-9]

CLASS_3b_DOSE INTERNAL-15 = internal events EPRI accident Class 3b dose rate given a 1-in-i5 years ILRT interval = 3.91 x 1e person-rem/ry [Table 2-9]

CLASS_3a_DOSE EXTERNAL -15 = external events EPRI accident Class 3a dose rate given a 1-in-i5 years ILRT interval = 1.33 x 10.2 person-rem/ry [See 1-in-i5 years table above]

CLASS_3b_DOSE EXTERNAL 15 = external events EPRI accident Class 3b dose rate given a 1-in-i5 years ILRT interval = 4.67 x 10&3 person-rem/ry [See 1-in-i5 years table above]

TOTALDOSECOMBsls = Total combined internal and external events dose rate for all EPRI's Classes given a 1-in-i5 years ILRT interval

= TOTAL-DOSEINTERNAL-15 + TOTALDOSEEXTERNAJ.15 TOTALDOSE INTERNAL-15 = Total internal events dose rate for all EPRI's Classes given a i-in-15 years ILRT interval = 12.0215 (person-rem/ry) [Table 2-9]

TOTALDOSE EXTERNAL-15 = Total external events dose rate for all EPRI's Classes given a 1-in-i5 years ILRT interval = 150.7655 (person-rem/ry) [See 1-in-10 years table above]

Therefore,

5 (I.12 x I 0O'+1.33x1 0)+(3.91x 1 +4.67x1OT )1

..-ECTDCOMB-15[I121~ (12.0215+150.7655)J PERCTDcoMBE15 = 0.0120%

Based on the above results, the combined internal and external events changes from the 1-in-10 years to 1-in-15 years dose rate is as follows:

TOTAL -DOSEcoiffB.1 5 +TOTAL-DOSEcoAIB.lo1I

  • 100%

INCRCTDCOMIB-1-15 =

T TOTAL_-DSEco,11B-10 J Where:

INCR CTDcom_ 10.15 = combined internal and external events percent change from 1-in-10 years ILRT interval to 1-in-15 years ILRT interval TOTAL DOSEcoMB.1s = Total combined internal and external events dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval

= TOTAL-DOSEINTERNAL.15 + TOTALDOSEEXTERNAI-15 TOTALDOSEcoMB.Io = Total combined internal and external events dose rate for all EPRl's Classes given a 1-in-10 years ILRT interval

= TOTAL-DOSEINTERNAL.1O + TOTAL-DOSEEXTERNAI-1O TOTAL-DOSE INTERNAL-15 = Total internal events dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval = 12.0215 (person-rem/ry) [Table 2-9]

TOTAL-DOSE ExTERNAL-1s = Total external events dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval = 150.7655 (person-rem/ry) [See 1-in-10 years table above]

TOTALDOSE INTERNAL.10 = Total internal events dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval = 12.0211 (person-rem/ry) [Table 2-9]

TOTAL-DOSE EXTERNAL.10 = Total external events dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval = 150.7600 (person-rem/ry) [See 1-in-10 years table above]

Therefore, INCR- CTD CTCOIB-D-15 [

=(12.0215 +150.7655) - (12.021 1+150.7600)1*100%

(12.0211 +150.7600) J INCRCTDcoMs_10-1s = 0.004%

The above increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.004%. This value can be considered to be a negligible increase in risk.

Evaluate the External Events Hazard Risk Impact in Terms of LERF This step, per the NEI Interim Guidance [3] calculates the change in the large early release frequency with extending the ILRT interval from 1-in-10 years to 1-inl 5-years.

The combined internal and external events affect on the LERF risk measure due to the proposed ILRT interval extension is calculated as follows:

ALERFcOMBINED1o-15 = CLASS-3bcoMBINED15 - CLASS-3b COMBINEDIO Where:

ALERFCOMBINED1o-15 = the combined internal and external events change in LERF from 1-in-10 years ILRT interval to 1-in-15 years ILRT interval CLASS...3bCOMBINED15 = the combined internal and external frequency of EPRI accident Class 3b given a 1-in-15 years ILRT Interval 3

= CLASS- bINTERNAL.1S + CLASSJ3bEXTERNAL-15 CLASS-3bNTERNAL-15S = internal events frequency of EPRI accident Class 3b given a 1-in-15 years ILRT Interval = 8.60 x 10'9/ry [Table 2-8]

CLASS-3bEXTERNAL-15 = external events frequency of EPRI accident Class 3b given a 1-in-15 years ILRT Interval = 1.03 x ile 7/ry [See 1-in-15 years table above]

CLASS- 3 bCoMBINEDIO = the combined internal and external frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval 3 3

= CLASS- bjNTERNAL-1o + CLASS- bEXTERNAL.10 3

CLASS- biNTERNAL.10 = internal events frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval = 5.73 x 10'9/ry [Table 2-8]

3 CLASS- bEXTERNAL.10 = external events frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval = 6.85 x 10 8 /ry [See 1-in-10 years table above]

Therefore, ALERFCOMBINEDIO.15 = (8.60 x 10'9 + 1.03 x 10' 7) - (5.73 x 10' 9 + 6.85 x 1048/ry)

ALERFCOMBINED1O.15 = 3.71 x 1048/ry The risk acceptance criteria of Regulatory Guide 1.174 as previously discussed in Section 2.4.8, Step 8 of this report, is used here to assess the ILRT interval extension. Regulatory Guide 1.174, "An Approach for Using PRA in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" [5], provides NRC recommendations for using risk information in support of applications requesting changes to the license basis of the plant.

The ALERFCOMBINEDIO.1S of 3.71 x 108/ry from extending the VY LRT frequency from 1-in-10 years to 1-in-15 years falls into Region l1l, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines in NRC Regulatory Guide 1.174 [5]. Therefore, because Regulatory Guide 1.174 [5] defines very small

APPENDIX A changes in LERF as below 10i7/ry, increasing the ILRT interval at VY from the currently allowed 1-in-10 years to 1-in-15 years is non-risk significant from a risk perspective.

Evaluate the External Events Hazard Change in Conditional Containment Failure Probability This step calculates the change in conditional containment failure probability (CCFP).

Similar to Step 9 (Section 2.4.9) of this report, the change in CCFP tracts the impact of the ILRT on both early (LERF) and late radionuclide releases. Therefore, CCFP consists of all those accident sequences resulting in a radionuclide release other that the intact containment state for EPRI accident Class 1, and small failure states for EPRI accident Class 3a. In additional, the CCFP is conditional given a severe core damage accident. The change in CCFP is calculated by the following equation:

CLASS I FREQ + CLASS 3a FREQ 1 C

= 1- [CLSSCDAsE 3a REQ *,

For the combined internal and external events 1-in-1 0 years ILRT interval:

CCFPCO- [ CL -COAI- + CLASS - 0 MA iO 100%

- [CLASS CDFBASE-COMIB i Where:

CCFPCOMB-1O = combined internal and external events conditional containment failure probability given 1-in-1 0 years ILRT interval CLASS-1 cOMB-10 = combined internal and external events frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

= CLASS.1 INTERNAL-10 + CLASS_1 EXTERNAL-10 CLASS-1 INTERNAL-10 = internal events frequency of EPRI accident Class 1given a 1-in-10 years ILRT interval = 5.43 x 10i7/ry [Table 2-8]

CLASS.1 EXTERNAL-10 = external events frequency of EPRI accident Class 1given a 1-in-10 years ILRT interval = 6.85 x 10l/ry [See 1-in-10 years table above]

CLASS_3a coMEB10 = combined internal and external events frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

= CLASS- 3 aINTERNAL.10 + CLASS-3a EXTERNAL-10 CLASS .. 3aINTERNAL-10 = internal events frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval = 5.73 x 10O/ry [Table 2-8]

CLASS_ 3aEXTERNAL.10 = external events frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval = 6.85 x 10.7/ry [See 1-in-10 years table above]

CDFB3ASE-COMB = VY combined internal events and external events CDF

= 5.03 x 1016/ry [Section 5, input#2] + 6.27 x 10 5/ry [Section A5.3, input#1]

= 6.78 x 10 5/ry Therefore, I -'+ .85xO0-) + (5.73 x10-8 + 6.85 x1T)lI (5.3 ~

CCPCO,1B-10 I [ +6.78x10- 5 JJ CCFPCOMBINED-10 = 87.99%

For the combined internal and external events 1-in-15 years ILRT interval:

CFPCO.B-5 = {I-[ AS - ICOMB-I5 + CLASS - M}oMB-11 100%

CDFBxASECMco

  • Where:

CCFPcoMs1s = combined internal and external events conditional containment failure probability given 1-in-1 5 years ILRT interval CLASS.1 coMB-15 = combined internal and external events frequency of EPRI accident Class 1 given a 1-in-15 years ILRT interval

= CLASS-1 INTERNAL-15 + CLASS.1 EXTERNAL-15 CLASS-1INTERNAL-15 = internal events frequency of EPRI accident Class 1given a 1-in-10 years ILRT interval = 5.12 x 7/ry[Table 2-8]

CLASS-1EXTERNAL-15 = external events frequency of EPRI accident Class igiven a 1-in-10 years ILRT interval = 6.48 x 1O 6/ry [See 1-in-10 years table above]

CLASSj3a cOM&-15 = combined internal and external events frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

= CLASS_ 3aNTERNAL1.5 + CLASS 3a EXTERNAL-15 CLASS_3aINTERNAL.1S = internal events frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval = 8.60 x 10' 8/ry [Table 2-8]

CLASS3aEXTERNAL15 = external events frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval = 1.03 x 10 /ry [See 1-in-10 years table above]

CDFBASE.COMB = VY combined internal events and external events CDF

= 5.03 x 10 6/ry [Section 5, input#2] + 6.27 x 10 5/ry [Section A5.3, input#1]

= 6.78 x 10 5/ry

Therefore,

= *(5.12107 +6.48x10 +)+(8.60x10 8 + 1.03xIT06 +)l) 0%

CCFPcO,j1BI1O {i..[5.-

I 1~ 6.78 XI0-' JJ 0 CCFPCOMBINED.15 = 88.04%

Therefore, the change in the combined internal and external events conditional containment failure probability from 1-in-1 0 years to 1-in-15 years is:

ACCFPCOMBINED1O-15 CCFP COMBINEDIS - CCFPCOMBINEDIO ACCFP COMBINED1o-15 88.04% - 87.99%

ACCFP COMBINED1o-1s 0.05%

This change in CCFP of less than 1% is insignificant from a risk perspective.

The effects of external hazard risk on ILRT risk are shown in Table A-4. The combined internal and external events effect on the ILRT risk is shown in Table A-5. This Table combines the results of Table 2-9 with the results depicted in Table A-4.

A6.0 Conclusions This appendix discusses the risk-implication associated with external hazards in support of the VY Integrated Leak Rate Testing (ILRT) interval extension risk assessment. The following conclusions are derived from this evaluation

1. The ALERFCOMBINEDIO 1s of 3.71 x 1 0-8/ry from extending the VY ILRT frequency from 1-in-10 years to 1-in-15 years falls within Region ll, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines in NRC Regulatory Guide 1.174 [5]. Therefore, increasing the ILRT interval at VY from the currently allowed 1-in-10 years to 1-in-15 years is non-risk significant from a risk perspective.
2. The combined internal and external events increase in risk for those accident sequences influenced by Type A testing, compared with the total integrated plant risk, given the change from a 1-in-1 0 years test interval to a 1-in-15 years test interval, is found to be 0.004% (0.007 person-rem/ry). This value can be considered to be a negligible increase in risk.
3. The change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-15 years is 0.05%. A change in CCFP of less than 1% is insignificant from a risk perspective.

Table A-4 Effect of External Events Hazard Risk on VY ILRT Risk Assessment Dose Rate as a Function of ILRT Interval (Person-Rem / ry)

CEPas Accident Class Description Baseline Current Proposed 3-per-10 years 1-per-10 years 1-per-15 years ILRT ILRT ILRT 1 No Containment Failure 9.58 x 10 3 8.91 x 104 8.42 x 1-O3 2 Large Containment Isolation Failures 734x10*2 734x10' 734x IO" (Failure-to-close) 3a Small Isolation Failures (Liner breach) 2.81 x 10'3 8.90 x 14O 1.33 x 10.2 3b Large Isolation Failures (Liner Breach) 9.82 x 10'4 3.11 x 104- 4.67 x iO-,

4 Small isolation failure - failure-to-seal (Type B test)

Small isolation failure - failure-to-seal (Type C test) 6 Containment Isolation Failures (dependent failures, personnel errors) 7a Severe Accident Phenomena Induced 1.10x101 1.10x1lo 1.l0Xlo, Containment Failure - energetic phenomena Severe Accident Phenomena Induced 7b Containment Failure - direct contact with core 1.32 x 102 1.32 x 102 1.32 x 102 material Severe Accident Phenomena Induced 7c Containment Failure - containment failures 4.30 x 10 4.30 x 101 4.30 x 10 '

due to loss of heat removal 8a Containment Bypassed (ATWS) 5.59 x 100 5.59 x 100 5.59 x 100 8b Containment Bypassed (ISLOCA, LOCAOC) 1.97 x 100 1.97 x 100 1.97 x 100 TOTALS 150.7525 150.7600 150.7655

Table A-5 Effect of Internal and External Events Risk on VY ILRT Risk Assessment Dose Rate as a Function of ILRT Interval (Person-Rem / ry)

CEPas Accident Class Description Baseline Current Proposed 3-per-10 years 1-per-10 years 1-per-15 years ILRT ILRT ILRT 1 No Containment Failure 1.03 x 102 9.62 x 104 9.08 x 10i3 2 Large Containment Isolation Failures 7.93 x 10.2 7.93 x 102 7.93 x 102 (Failure-to-close) 3a Small Isolation Failures (Liner breach) 3.03 x 10'3 9.64 x 104' 1.45 x 10-2 3

3b Large Isolation Failures (Liner Breach) 1.06 x 104 3.38 x 104 5.06 x 10-Small isolation failure - failure-to-seal (Type B test)

Small isolation failure - failure-to-seal (Type C test) 6 Containment Isolation Failures (dependent failures, personnel errors) 7a Severe Accident Phenomena Induced 1.19 x1o 1.19 x101 1.19 x10' Containment Failure - energetic phenomena ._x_._x_._x Severe Accident Phenomena Induced 7b Containment Failure - direct contact with core 1.42 x 102 1.42 x 1O' 1.42 x 102 material Severe Accident Phenomena Induced 7c Containment Failure - containment failures 4.64 x 10.1 4.64 x 10 1 4.64 x 10 '

due to loss of heat removal 8a Containment Bypassed (ATWS) 6.06 x 100 6.06 x 100 6.06 x 100 8b Containment Bypassed (ISLOCA, LOCAOC) 2.13 x 100 2.13 x 100 2.13 x 100 TOTALS 162.7729 162.7811 162.7871

Appendix B Risk Impact of Containment Liner Corrosion During an Extension of the ILRT Interval

Table of Contents Page No.

B1.0 Introduction B-2 B2.0 Method of Analysis B-2 B3.0 Assumptions B-3 B4.0 Input B-3 B5.0 Steel Shell Corrosion Analysis B-4 B6.0 Steel Shell Corrosion Sensitivity B-20 B7.0 Conclusions B-20 Tables Page No.

Table B Flaw Failure Rate as a Function of Time B-21 Table B Flaw Failure Rate as a Function of Test Interval B-21 Table B Vermont Yankee Containment Failure Probability Given Containment Liner Flaw B-22 Table B Vermont Yankee Containment Liner Corrosion Base Case B-23 Table B Impact of Containment Steel Liner Corrosion on Vermont Yankee ILRT Intervals B-24 Table B Containment Steel Liner Corrosion Sensitivity Cases B-25 Figures Page No.

Figure B Vermont Yankee Containment Failure Probability Given Containment Liner Flaw B-22

B1.0 Introduction Inspections of reinforced and steel containments at some facilities (e.g., North Anna, Brunswick D.C. Cook, and Oyster Creek) have indicated degradation from the inaccessible side of the steel shell and liner of primary containments. The major inaccessible areas of the Mark I containment are the vertical portion of the drywell shell and part of the shell located between the drywell floor and the basemat. As a result of these inaccessible areas, a potential increase in risk due to liner leakage, caused by age-related degradation mechanisms may occur when extending the current 1-in-1 0 years to 1-in-1 5 years Type A Integrated Leak Rate Testing (ILRT) interval.

Therefore, this appendix evaluates the likelihood and risk-implication associated with containment liner corrosion going undetected in visual examinations during the proposed extension of the ILRT interval.

B2.0 Method of Analysis The analysis utilizes the referenced Calvert Cliffs Nuclear Power Plant assessment [20] to estimate the risk impact from containment liner corrosion during an extension of the ILRT interval.

Consistent with the Calvert Cliffs analysis, the following issues are addressed:

  • Differences between the containment basemat and the drywell and torus liner
  • The historical drywell/torus steel shell flaw likelihood due to concealed corrosion
  • The impact of aging
  • The corrosion leakage dependency on containment pressure
  • The likelihood that visual inspections will be effective at detecting a flaw The method of analysis determines the total likelihood of non-detected containment leakage given a change in the likelihood that a flaw exists (i.e., increase in flaw likelihood due to the ILRT extension), that the flaw is not detected and that flaw results in a breach.

Consistent with Calvert Cliffs analysis [20], the following six steps are performed:

1) Determine the historical liner flaw likelihood.
2) Determine aged adjusted liner flaw likelihood.
3) Determine the increase in flaw likelihood between 3, 10 and 15 years.
4) Determine the likelihood of containment breach given liner flaw.
5) Determine the visual inspection detection failure.
6) Determine the likelihood of non-detected containment leakage.

In additions to these steps, the following three additional steps are added to evaluate the risk-implication of containment liner corrosion:

7) Evaluate the risk impact in terms of population dose rate and percentile change for the interval cases.
8) Evaluate the risk impact in terms of LERF.
9) Evaluate the change in conditional containment failure probability.

B3.0 Assumptions

1) Consistent with the Calvert Cliffs methodology [20], a half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures.
2) Consistent with the Calvert Cliffs methodology [20], the leakage potential via the drywell floor (due to crack formation) is considered less likely than other sections of the containment structure.
3) Consistent with the Calvert Cliffs methodology [20], the likelihood of the containment atmosphere reaching the outside atmosphere given a liner flaw exists was estimated as a function of the pressure inside the containment.
4) Consistent with the Calvert Cliffs methodology [20], the containment liner flaw likelihood doubles every five years. This is based solely on judgment and is included in this analysis to address the increase likelihood of corrosion as the containment liner ages.
5) Consistent with the Calvert Cliffs methodology [20], the probability of a concurrent containment breach given a flaw in the containment liner is depicted as an exponential function.
6) Consistent with the Calvert Cliffs methodology [20], a 0.05 (5%) visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 0.10 (10%) is used12 .
7) Consistent with the Calvert Cliffs methodology [20], 1.0 (100%) visual inspection detection failure likelihood given the flaw is located in an inaccessible area of either the drywell or torus.
8) Consistent with the Calvert Cliffs methodology [20], all non-detectable containment failures are considered to result in large early releases.

B4.0 Input

1) The containment liner failure rate is based on two industry events:
1. On September 22, 1999, North Anna Unit 2 experienced through-wall corrosion of the metal liner.

The corrosion appeared to have been initiated from a piece of lumber imbedded in the concrete behind the liner plate.

2. On April 27, 1999, inspection at Brunswick 2 discovered two through-wall holes and pitting in the drywell shell. The through-wall condition was believed to have originated from the coated (visible) side.
2) The number of steel-lined containments is 70 [20].

12 Note: to date, all liner corrosion events have been detected through visual inspection.

3) The exposure time in detecting a containment flaw is 5.5 years. This is consistent with the Calvert Cliffs methodology [20] and reflects the time period since 10CFR 50.55a starting requiring visual inspection. This is deemed conservative, since the exposure time period is bounding as no additional failures have been identified in the nuclear industry since March 2002 and no failures were identified prior to September 1996 (the date when 10CFR 50.55a was implemented).
4) Consistent with the Calvert Cliffs methodology [20], leakage through the drywell floor is 10 times less likely than through other sections of the containment structure.
5) The probability of a concurrent containment breach given a flaw in the containment liner is depicted as an exponential function. This curve is used to interpolate the containment failure probability at the pressure at which the ILRT is to be performed for the accessible and inaccessible areas of containment. Consistent with the Calvert Cliffs methodology, the lower bound limit was assigned a failure probability of 0.1% at a pressure of 20 psia and the upper bound was assigned a failure probability of 100% at the ultimate containment failure pressure of 154.7 psia [29].

B5.0 Steel Shell Corrosion Analysis Step 1B - Determine the Historical Liner Flaw Likelihood.

This step calculates historical liner flaw likelihood consistent wit the Calvert Cliffs methodology [20]. This value, for Vermont Yankee consists of the accessible potion of the drywell and torus, the inaccessible portion of the drywell and submergence area of the torus, and the inaccessible area of the drywell floor.

The accessible portion of the drywell and torus liner flaw likelihood is determined as follows:

AHLFDT = NFAILa / (NPLANTS

  • TEXPO)

The inaccessible portion of the drywell and submergence area of the torus liner flaw likelihood is determined as follows:

IAHLFDT = NFAILa / (NPLANTS

  • TEXPO)

The inaccessible area of the drywell floor IAHLFDF = NFAILa / (NPLANTS

  • TEXPO)

Where:

AHLFDT = accessible portion of the drywell and torus liner flaw IAHLFDT = inaccessible portion of the drywell and submergence area of the torus liner flaw likelihood IAHLFDF = inaccessible area of the drywell floor liner flaw NFAILa = number of industry events due to liner corrosion = 2 [Section B4.0, Input #1]

NFAILja = number of industry events due basemat corrosion = 0.5 [Section B3.0, Input #1]

NPLANTS = number of steel-lined containments = 70 [Section B4.0, Input #2]

TEXPO = time exposure since issuing of 10CFR50.55a = 5.5 years[Section B4.0, Input #3]

Therefore,

AHLFDT = 2 / (70 - 5.5) = 5.19 x 103 /yr IAHLFDT = 2 / (70

  • 5.5) = 5.19x 103/yr IAHLFDF = 0.5 I (70
  • 5.5) = 1.30 x 10 3/yr The above results are documented in Table B-4.

Step 2B - Determine Aged Adiusted Liner Flaw Likelihood.

Per the Calvert Cliffs methodology [20], the aged adjustment liner flaw likelihood is calculated for a 15-year interval given that the failure rate doubles every 5 years (Section B3.0, assumption #4) or increases 14.9 % per year. In addition, the average for the 5th to 1 1h year was set to the historical failure calculated in Step I B.

The results, based on an iterative process that satisfies the above conditions are presented in Table B-1.

Step 3B - Determine the increase in flaw likelihood between 3,10 and 15 years 13 This step calculates the increase in flaw likelihood at 3-in-10 years interval (or 1-in-3 years), 1-in-10 years interval, and 1-in-15 years interval, per the Calvert Cliffs methodology [20]. The results of Step 2B are use to generate these values as follows:

Accessible portion of the drywell and torus, ADTFLAW3Io = I ADTFRATEii i=1,3 ADTFLAW1 .10 = I ADTFRATEii i=1,10 ADTFLAW1 .1s = v ADTFRATEii i=1,15 13 Note: the Calvert Cliffs analysis presents the delta between 3 and 15 years of 8.70/%to utilize in the estimation of the delta-LERF value. For this analysis, however, the values are calculated based on the 3-in-10 years, 1-in-10 years, and 1-in-15 years intervals consistent with the evaluation in this calculation, and then the delta-LERF values are determined from there.

Inaccessible portion of the drywell and submergence area of the torus, IDTFLAW 3.10 = Z IDTFAE i=1,3 IDTFLAW 1.10 = Y IDTFRTu i=1,10 IDTFLAW 1 .15 = Z IDTFRATEi i=1,15 Inaccessible area of the drywell floor DFFLAW3Io = E DFFRATEH i=1,3 DFFLAWI. 10 = Y: DFFRATEii i=1,10 DFFLAWI.15 = X DFFRATEi i=1,15 Where:

ADTFLAW3 10 = increase in flaw likelihood at 3-in-10 years test interval given accessible portion of the drywell and torus ADTFLAW 1 .10 = increase in flaw likelihood at 1-in-10 years test interval given accessible portion of the drywell and torus ADTFLAWI. 15 = increase in flaw likelihood at 1-in-15 years test interval given accessible portion of the drywell and torus IDTFLAW310 = increase in flaw likelihood at 3-in-10 years test interval given inaccessible portion of the drywell and submergence area of the torus IDTFLAW1.10 = increase in flaw likelihood at 1-in-10 years test interval given inaccessible portion of the drywell and submergence area of the torus IDTFLAWl.15 = increase in flaw likelihood at 1-in-15 years test interval given inaccessible portion of the drywell and submergence area of the torus DFFLAW3 1O = increase in flaw likelihood at 3-in-10 years test interval given inaccessible area of the drywell floor DFFLAW1 .10 = increase in flaw likelihood at 1-in-10 years test interval given inaccessible area of the drywell floor DFFLAW*1I5 = increase in flaw likelihood at i-in-15 years test interval given inaccessible area of the drywell floor ADTFRATU = aged adjusted liner flaw likelihood, given accessible portion of the drywell and torus (Table B-1)

IDTFRATEiH = aged adjusted liner flaw likelihood, given inaccessible portion of the drywell and submergence area of the torus (Table B-1)

DFFFRATEHi = aged adjusted liner flaw likelihood, given inaccessible area of the drywell floor (Table B-1)

Therefore, ADTFLAW3I0 = 0.71%, ADTFLAW 1.10 = 4.14%, ADTFLAW 1 .1S = 9.68%

IDTFLAW3.j0 = 0.71%, IDTFLAWI.10 = 4.14%, IDTFLAW1 .15 = 9.68%

DFFLAW3 1o = 0.18%, DFFLAW,.Io = 1.04%, DFFLAWI. 15 = 2.42%

The above results are documented in Table B-2.

Step 4B - Determine the Likelihood of Containment Breach Given Liner Flaw.

The likelihood of a breach in containment given a liner flaw is based on the Calvert Cliffs methodology

[20] with a Vermont Yankee specific value for the upper-end pressure failure (100% likelihood) taken from Section 4.4 of the PSA [29]. A containment pressure of 154.7 psia corresponds with the 100% probability of failure. The lower-end pressure failure (0.1% likelihood) is set at 20 psia, consistent with Calvert Cliffs

[20]. Per the Calvert Cliffs methodology [20], the containment failure probability (FP) versus containment pressure (P) is assumed to be an equation of the form:

FP (P) = b

  • em p Where:

FP (P) = containment failure probability given containment liner breach m = slope of the containment failure probability b = intercept of the containment failure probability p = containment pressure, psia The two anchor points of 0.1% at 20 psia and 100% at 154.7 psia provide sufficient information to solve for the slope m, and the intercept b, as follows:

Slope m, m = LN (FP(100%) - LN (0.1%) / (Upper Pressure- Lower Pressure) m = LN (1.0) - LN (0.001) I (154.7-20) m = 5.13 x 102 Intercept b, b = FP(100%) / P b = 1 / es13X102-154.7 b = 3.58x 10 The Vermont Yankee April 26,1995 ILRT used a test pressure of 45.9 psig (or 60.6 psia) [28]. Based on this pressure the likelihood of containment breach in the liner is:

FP (60.6 psia) = 3.58 x 104 *e5.'3 x10e2 '60.6 FP (60.6 psia) = 0.00802 or 0.802%

For the Drywell floor, the failure probability is set to one-tenth of the failure probability for Drywell walls, or 0.0802%. (See Section B3.0, Assumption #4 and Section B4.0, Input #2).

Based on the above equation, containment liner breach and drywell floor intermediate values for FP are calculated and presented in Table B-3 and Figure B-1.

Step 5B - Determine the visual inspection detection failure.

This step examines the visual inspection detection failure likelihood for Vermont Yankee. The three areas of interest are the accessible portion of the drywell and torus, the inaccessible portion of the drywell and submergence area of the torus, and the inaccessible portion of the drywell floor.

The visual inspection detection failure likelihood for the accessible area of the drywell and torus (100%

inside and outside of drywell head, 100% drywell liner inside, 100% torus outside area, and 100% torus inside area above waterline is set to 10%, consistent with the Calvert Cliffs analysis [20]. This represents a 5% (0.05) failure to identify a visual flaw and 5% (0.05) likelihood that the flaw is not visible.

The inaccessible portion of the drywell (virtually 0% drywell liner outside because it is encased in concrete), and submergence area of the torus is assigned a 100% (1.0) visual detection failure likelihood.

This is bounding, as the submerged area of the Torus may be examined.

Because the liner under the Drywell floor cannot be visually inspected, a visual detection failure likelihood of 100 % (1.0) is assigned, consistent with the Calvert Cliffs method.

The above results are documented in Table B-4.

Step 6B - Determine the likelihood of non-detected containment leakage Per the Calvert Cliffs methodology [20], the likelihood of a non-detected containment leakage is calculated by multiplying the results of Steps 3B, 4B, and 5B. This yields the following:

Accessible portion of the drywell and torus, ADTLEAK 3Io = ADTFLAW3Io

  • ADTFPILRT
  • ADTVISUAL ADTLEAKM.,o = ADTFLAWI.j 0
  • ADTFPILRT
  • ADTVISUAL ADTLEAKI.Is = ADTFLAWI.15
  • ADTFPILRT
  • ADTVISUAL Where:

ADTLEAK, 10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and accessible portion of the drywell and torus ADTLEAKI-10 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and accessible portion of the drywell and torus ADTLEAKI .15 = likelihood of non-detected containment leakage, given 1-in-15 years test interval and accessible portion of the drywell and torus ADTFLAW3Io = increase in flaw likelihood at 3-in-1 0 years test interval given accessible portion of the drywell and torus = 0.71% (0.0071) [Table B-21 ADTFLAW1.10 = increase in flaw likelihood at 1-in-10 years test interval given accessible portion of the drywell and torus = 4.14% (0.0414) [Table B-2]

ADTFLAWI. 15 = increase in flaw likelihood at 1-in-1 5 years test interval given accessible portion of the drywell and torus = 9.68% (0.0968) [Table B-2]

ADTFPILRT = likelihood of containment breach at ILRT test pressure (60.6 psia) given liner flaw and accessible portion of the drywell and torus = 0.00802 (0.802%) [Step 4B]

ADTVISUAL = visual inspection detection failure accessible portion of the drywell and torus = 0.1 (10%) [Step 5B]

Therefore, ADTLEAK3.j 0 = 0.0071

  • 0.00802
  • 0.1 = 5.694 x 10.6 (0.00057%)

ADTLEAKI.j 0 = 0.0414

  • 0.00802
  • 0.1 = 3.320 x 10 5 (0.00332%)

ADTLEAK1 .15 = 0.0968

  • 0.00802
  • 0.1 = 7.763 x 10 5 (0.00775%)

Inaccessible portion of the drywell and submergence area of the torus, IDTLEAK 3 _,0 = IDTFLAW3.10

  • ADTFPILRT
  • IDTVISUAL IDTLEAK 1.10 = IDTFLAW 1.10
  • ADTFPILRT
  • IDTVISUAL IDTLEAKI. 15 = IDTFLAWI. 15
  • ADTFPILRT IDTVISUAL Where:

IDTLEAK3 -10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and inaccessible portion of the drywell and submergence area of the torus IDTLEAK,-10 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and inaccessible portion of the drywell and submergence area of the torus IDTLEAK 1.15 = likelihood of non-detected containment leakage, given 1-in-15 years test interval and inaccessible portion of the drywell and submergence area of the torus IDTFLAW3.Io = increase in flaw likelihood at 3-in-10 years test interval given inaccessible portion of the drywell and submergence area of the torus = 0.71% (0.0071) [Table B-2]

IDTFLAW1.10 = increase in flaw likelihood at 1-in-10 years test interval given inaccessible portion of the drywell and submergence area of the torus = 4.14% (0.0414) [Table B-2]

IDTFLAW1. 15 = increase in flaw likelihood at 1-in-15 years test interval given inaccessible portion of the drywell and submergence area of the torus = 9.68% (0.0968) [Table B-2]

ADTFPILRT = likelihood of containment breach at ILRT test pressure (60.6 psia) given liner flaw and inaccessible portion of the drywell and submergence area of the torus = 0.00802 (0.802%) [Step 4B]

IDTVISUAL = visual inspection detection failure inaccessible portion of the drywell and submergence area of the torus = 1.0 (100%) [Step 5B]

Therefore, IDTLEAK 3 10 = 0.0071

  • 0.00802
  • 1.0 = 5.694 x 10'5 (0.0057%)

IDTLEAK 1 .10 = 0.0414

  • 0.00802
  • 1.0 = 3.320 x 1O4 (0.0332%)

IDTLEAK 1 .15 = 0.0968

  • 0.00802
  • 1.0 = 7.763 x 10' (0.0776%)

Inaccessible portion of the drywell floor, DFLEAK.,Io = DFTFLAW 3 .1 0 - DFTFPILRT

  • DFTVISUAL DFTLEAKI. 1 0 = DFTFLAWI.o0
  • DFTFPILRT
  • DFTVISUAL DFTLEAK 1 .15 = DFTFLAWI. 15
  • DFTFPILST
  • DFTVISUAL Where:

DFLEAK 3 10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and inaccessible portion of the drywell floor DFLEAK1.10 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and inaccessible portion of the drywell floor DFLEAKI. 15 = likelihood of non-detected containment leakage, given 1-in-15 years test interval and inaccessible portion of the drywell floor DFFLAW3.1 0 = increase in flaw likelihood at 3-in-1 0 years test interval given inaccessible portion of the drywell floor = 0.18% (0.0018) [Table B-2]

DFFLAW1 .10 = increase in flaw likelihood at 1-in-10 years test interval given inaccessible portion of the drywell floor = 1.04% (0.0104) [Table B-2]

DFFLAWI. 15 = increase in flaw likelihood at 1-in-1 5 years test interval given inaccessible portion of the drywell floor = 2.42% (0.0242) [Table B-2]

DFTFPILRT = likelihood of containment breach at ILRT test pressure (60.6 psia) given liner flaw and inaccessible portion of the drywell floor = 0.000802 (0.0802%) [Step 4B1]

DFVISUAL = visual inspection detection failure inaccessible portion of the drywell floor

= 1.0 (100%) [Step 5B1]

Therefore, DFTLEAK3 10 = 0.0018

  • 0.000802
  • 1.0 = 1.444 x 10 4 (0.00014%)

DFTLEAKI. 10 = 0.0104

  • 0.000802
  • 1.0 = 8.341 x 104 (0.00083%)

DFTLEAKI. 1 , = 0.0242

  • 0.000802
  • 1.0 = 1.941 x 105 (0.00194%)

Total Likelihood of Non-Detected Containment Leakage due to Corrosion is, TOTAL 3. 10 = ADTLEAK3.10 + IDTLEAK3.10 + DFTLEAK3 10 TOTAL 1.10 = ADTLEAK 1 .10 + IDTLEAK1 .10 + DFTLEAKI. 10 TOTALl. 15 = ADTLEAKI. 15 + IDTLEAK1 . 15 + DFTLEAK1 .15

Where:

TOTAL3 10 = total likelihood of non-detected containment leakage due to corrosion, given 3-in-1 0 years test interval TOTAL 1 .1 o = total likelihood of non-detected containment leakage due to corrosion, given 1-in-10 years test TOTAL 1.15 = total likelihood of non-detected containment leakage due to corrosion, given 1-in-1 5 years test interval ADTLEAK 3.10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and accessible portion of the drywell and torus ADTLEAK 1.10 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and accessible portion of the drywell and torus ADTLEAKI. 15 = likelihood of non-detected containment leakage, given 1-in-15 years test interval and accessible portion of the drywell and torus IDTLEAK3 .1 0 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and inaccessible portion of the drywell and submergence area of the torus IDTLEAK1 .10 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and inaccessible portion of the drywell and submergence area of the torus IDTLEAK1. 15 = likelihood of non-detected containment leakage, given 1-in-15 years test interval and inaccessible portion of the drywell and submergence area of the torus DFLEAK3-,o = likelihood of non-detected containment leakage, given 3-in-10 years test interval and inaccessible portion of the drywell floor DFLEAK1 .10 = likelihood of non-detected containment leakage, given 1-in-1 0 years test interval and inaccessible portion of the drywell floor DFLEAK1 15s = likelihood of non-detected containment leakage, given 1-in-15 years test interval and inaccessible portion of the drywell floor Therefore, TOTAL3-10 0.00057% + 0.00570% + 0.00014% = 0.00641 %

TOTAL1.10 0.00332% + 0.03322% + 0.00083% = 0.03737%

TOTAL1 .15 0.00776% + 0.07763% + 0.00194% = 0.08733%

The above results are documented in Table B-4.

Step 7B - Evaluate the Risk Impact in Terms of Population Dose Rate and Percentile Change for the Interval Cases.

This step calculates the change in population dose rate for EPRI accident Class 3b (all non-detectable containment failures are considered to result in large early releases), the change in percentage of the total dose rate attributable to liner corrosion and the change in this result dose rate from the base dose rate attributable to changes in ILRT surveillance interval.

The change in population dose rate is calculated as outline in Section 2.4.7 (Step 7), of this risk assessment (see page 43 of 73).

Increase to EPRI class 3b frequencies LINERCLASS_3bFREQ31 o = (PROBdCs 3b 3.jD + LINERCLASS_3B_LINCREASE3.1o) x (CDF - CDFINDEP)

LINERCLASS_3b_FREQ 1 .lo = (PROBdcj. 3blj.jo + LINER_CLASS_3B_INCREASE 1 .1 o) x (CDF - CDFINDEP)

LINERCLASS_3bFREQ1 .15 = (PROBd&ssl3bo3.10 + LINER_CLASS_3B_INCREASE 1.15 ) x (CDF - CDFINDEP)

Where:

L1NER_CLASSjb_FREQ3 ,o = frequency of EPRI Class 3b due to liner corrosion failure given a 3-in-10 years ILRT interval LINER_CLASS_3b_FREQ1 .Io = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-10 years ILRT interval LINER_CLASS_3b_FREQj.j 5 = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-15 years ILRT interval PROBdass-3bl3-10 = probability of large pre-existing containment liner leakage

= 0.0027 [Section 2.3, input #7]

PROBciass_3bl1.io = probability of large pre-existing containment liner leakage

= 0.0090 [Section 2.4.5 Step 5, page 37 of 73]

PROBdass-3b 1-15 = probability of large pre-existing containment liner leakage

= 0.0135 [Section 2.4.5 Step 5, page 38 of 73]

CDFINDEP = CDF for those individual sequences that independently cause a LERF, (identified as EPRI Class 2, Class 7a and 7b, and Class 8 sequences) and loss of containment heat removal sequences (EPRI Class 7c)

= 4.40 x 1076 /ry [Section 2.4.1 Step 1, page 28 of 73]

CDF = Vermont Yankee PSA core damage frequency

= 5.00 x 1004 /ry [Section 2.3, input #2]

LINERCLASS_3B_INCREASE 3 1 *0 = TOTAL 3.10 x EPRICLASS_3BFRACTION LINERCLASS_3B_INCREASE 1 .10 = TOTAL1.10 x EPRICLASS_3BFRACTION LINERCLASS_3B_INCREASE 1.1 5 = TOTAL1.15 X EPRICLASS_3BFRACTION Where:

LINERCLASS_3B_INCREASE 3 .10 = liner corrosion increase in EPRI class 3b given 3-in-10 years test interval LINERCLASS_3B_INCREASE 1.10 = liner corrosion increase in EPRI class 3b given 1-in-10 years test interval LINER_CLASS_3B_INCREASE1 .1 5 = liner corrosion increase in EPRI class 3b given 1-in-15 years test interval TOTAL-3 10 = total likelihood of non-detected containment leakage due to corrosion, given 3-in-10 years test interval

= 0.00641% [see above calculation and Table B-4]

TOTAL1 .10 = total likelihood of non-detected containment leakage due to corrosion, given 3-in-10 years test interval

= 0.03737% [see above calculation and Table B-4]

TOTAL1.15 = total likelihood of non-detected containment leakage due to corrosion, given 3-in-10 years test interval

= 0.08733% [see above calculation and Table B-4]

EPRI CLASS 3B3FRACTION = fraction of containment failures due to liner corrosion and considered to result in large early releases.

= 100% [Assumpiton#8]

Therefore:

LINERCLASS_3BINCREASE 3. 10 = 0.00642% x 1.0 = 0.00642%

LINERCLASS_3B_INCREASE 1.10 = 0.03737% x 1.0 = 0.03737%

LINERCLASS_3B_INCREASE1 .15 = 0.08729% x 1.0 = 0.08729%

Therefore:

LINER_CLASS_3bFREQ 3.10 = (0.0027 + 0.00642%) x (5.00 x 104/ry - 4.40 x 106 /ry)

LINER_CLASS_3b_FREQ 3. 10 = 1.76 x 10i9/ry LINERCLASS_3b.FREQ1.10 = (0.0090 + 0.03737%) x (5.00 x 104 /ry - 4.40 x 10 6/ry)

LINER_CLASS_3b_FREQO.j 0 = 5.97 x 10-9ry LINERCLASS_3bFREQ1 .1 5 = (0.01 35 + 0.08729%) x (5.00 x 10 6/ry - 4.40 x 10.6/ry)

LINERCLASS_3bFREQ1 . 15 = 9.16 x 109/ry

Increase to EPRI class 1 frequencies LINERCLASS-1 FREQ03 10 = NCF - CLASS_3aFREQUENCY - LINER_CLASS_3b_FREQ03 10 LINERCLASS-1 FREQ.10 = NCF- CLASS_3a_FREQUENCYIQ - LINERCLASS_3bFREQ1 .1 0 LINERCLASS-1 FREQ1.15 = NCF- CLASS_3a_FREQUENCY 15 - LINER_CLASS.3b_FREQ1 . 15 Where:

LINER_CLASS_1_ FREQ3 10 = frequency of EPRI Class 1 given a 3-in-1 0 years ILRT interval LINERCLASS_1_ FREQ 1.10 = frequency of EPRI Class 1 given a 1-in-10 years ILRT interval LINERCLASS_1. FREQ 1.15 = frequency of EPRI Class 1 given a 1-in-15 years ILRT interval CLASS 3a_ FREQUENCY = frequency of small pre-existing containment liner leakage

= 1.72 x 10 8/ry [Section 2.4.1 Step 1, page 28 of 70]

CLASS_3aFREQUENCY 1 0 = frequency of small pre-existing containment liner leakage given a 1-in-10 years ILRT interval

= 5.73 x 104 /ry [Section 2.4.5, Step 5 page 39 of 70]

CLASS_3a_FREQUENCY 1 5 = frequency of small pre-existing containment liner leakage given a 1-in-15 years ILRT interval

= 8.60 x 104/ry [Section 2.4.5, Step 5 page 39 of 70]

LINERCLASS_3b.FREQ 3.10 = frequency of EPRI Class 3b due to liner corrosion failure given a 3-in-10 years ILRT interval

= 1.76 x 109/ry [Above write-up, page B-1 4 of B-26]

LINERCLASS_3b-FREQ 1.1 0 = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-10 years ILRT interval

= 5.97 x 109/ry [Above write-up, page B-1 4 of B-26]

LINERCLASS_3b-FREQ1 -15 = frequency of EPRI Class 3b due to liner corrosion failure given a 3-in-15 years ILRT interval

= 9.16 x 10.9/ry [Above write-up, page B-1 4 of B-26]

NCF = frequency in which containment leakage is at or below maximum allowable Technical Specification Leakage

= 6.07 x 107/ry [Table 2-2]

Therefore:

LINERCLASS_1_ FREQ3.10 = 6.07 x 10 7 /ry -1.72 x 10' 8/ry - 1.76 x 10' 9/ry LINERCLASS_1_ FREQ3.jo = 5.8 x 107/ry LINERCLASS_1_ FREQ 1.1o = 6.07 x 10i7/ry -5.73 x 10O'/ry -5.97 x 10' 9/ry

= 5.43 x 10 7/ry LINERCLASS_1_ FREQ 1.10 LINERCLASS_1_ FREQ 1.15 = 6.07 x 10.7/ry -8.60 x 10O'/ry -9.i6 x 10' 9/ry LINERCLASS_1_ FREQ 1.15 = 5.11 x 10 7/ry

The results of other pertinent calculations, are presented below as follows:

For 3-in-10 years, EPRI Class Person-rem Frequency/Ry Person-rem/Rv 1 1.30x103 5.88x10'7 7.64x104 2 2.73 x 106 2.14 x 1i0 Corrosion Addition 5.85 x 10'3 3a 1.30 x 104 1.72 x 109 2.24 x 10O 3b 4.55 x 104 1.76 x 10' 9 4.09 x 10'" 8.01 x 10'5 4 N/A 0.0 0.0 5 N/A 0.0 0.0 6 N/A 0.0 0.0 7a 2.73 x 1O6 3.22 x 10'7 8.79 x 10' 7b 2.73 x 105 3.85 x 10o 1.05 x 10 1 7c 3.53 x 1O6 9.71 x 10' 3.43 x 10-2 8a 2.73 x 1O6 1.63 x 10'7 4.46 x 10'1 8b 2.96 x 1O6 5.32 x 109 1.57 x 101' Tota1 -12.0204 ILRT Dose Rate from 3a and 3b = 2.24 x 104 + 8.01 x 10'5 = 3.04 x 10-4 person-rem/ry

%Of Total = 100 * [2.24x104 + 8.01 x105 ] / 12.0204 = 0.0025%

LERFfrom3b = 1.76x10'9 CCFP%LINER310 = 1 - [(5.88 x 10'7 + 1.72 x 1o 8) / 5.03 x 10.61 = 87.97%

For 1-in-10 years.

EPRI Class Person-rem Frequency/RV Person-rem/Rv 1 1.30 x 103 5.43 x 10'7 7.06 x 10-4 2 2.73 x 106 2.14 x 10- Corrosion Addition 5.85 x 10-3 3a 1.30 x 104 5.73 x 10 4 7.46 x 10-4 3b 4.55 x 104 5.97 x 10'9 2.38 x 10.10 2.72 x 104 4 N/A 0.0 0.0 5 N/A 0.0 0.0 6 N/A 0.0 0.0 7a 2.73 x 106 3.22 x 10-7 8.79 x 101' 7b 2.73 x 106 3.85 x 10' 1.05 x 10' 7c 3.53 x 106 9.71 x I0' 3.43 x 10-2 8a 2.73 x 106 1.63 x 107 4.46 x 101' 8b 2.96 x 106 5.32 x 109 1.57 x 10.1 Total' ', I- . - .- . 1 5.00x10'; 12.0211 ILRT Dose Rate from 3a and 3b = 7.46 x 104 + 2.72 x 10'4 = 1.02 x 10'3 person-rem/ry

%Of Total = 100 * [7.46x104 + 2.72x10' 4 ] I 12.0211 = 0.0085%

LERF from 3b = 5.97 x 10' 9/ry CCFP%LINER1.10 = 1 - [(5.43x 10'7 + 5.73x 10'8) / 5.03 x 10'61 = 88.07%

For 1-in-15 Vears, EPRI Class Person-rem FrequencvlRv Person-rem/Ry 1 1.30 x 103 5.11 x10 7 6.65 x 10'4 2 2.73 x 106 2.14 x 10-9 Corrosion Addition 5.85 x 10'3 3a 1.30x 104 8.60x 1O'3 1.12x 10'3 3b 4.55 x 104 9.16 x 10- 9 5.77 x 10° 4.17 x 10'4 4 N/A 0.0 0.0 5 N/A 0.0 0.0 6 N/A 0.0 0.0 7a 2.73x 106 3.22 x10 7 8.79x 10 7b 2.73 x 1O6 3.85 x 1o0 1.05 x 1O' 7c 3.53 x 1O 9.71 x 10'9 3.43 x 10-2 8a 2.73x 106 1.63 x 107 4.46 x 10.1 8b 2.96 x 10 5.32 x 108 1.57 x 10 t 5.00 x10; <o.. 12.0216 ILRT Dose Rate from 3a and 3b = 1.12x 10'3 + 4.17 x 104 = 1.54 x 10'3 person-rem/ry 3

%Of Total = 100 * [1.16x10, + 4.32x10I / 12.0216 = 0.0128%

LERFfrom3b = 9.16x i0'/ry CCFP%LINER1.s1= 1 - [(5.11 xli' 7 + 8.60x1098) / 5.03 x 1061 = 88.13%

Based on the above results, the changes from the 1-in-10 years to 1-in-15 years dose rate is as follows:

INCREASELINER10-15 = [ TOT- DOSEATE.LINER15 - TOT- DOSERATE.LINER1O]

  • 100 TOT- DOSERATE.LINERIO Where:

INCREASELINER10. 15 = percent change from 1-in-10 years ILRT interval to 1-in-15 years ILRT interval TOT- DOSE RATE.LINER1s = Total dose rate forall EPRI's Classes given a 1-in-15 years ILRT interval

= 12.0216 (person-rem/ry) [See for 1-in-i5 years table above]

TOT- DOSE RATE.LINER10 = Total dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval

= 12.0211 (person-rem/ry) [See for 1-in-i 0 years table above]

Therefore, INCREASELINER1G01s = [ 12.0216 - 12.0211 ]

  • 100 = 0.0042%

12.0211 The above increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.0042%. This value can be considered to be a negligible increase in risk.

Step 8B - Evaluate the risk impact in terms of LERF This step calculates the change in the large early release frequency with extending the ILRT intervals from 1-in-10 years to 1-in1 5-years given the inclusion of a postulated liner corrosion flaw failure.

The affect on the LERF risk measure due to liner corrosion flaw is calculated as follows:

ALERFLNERIOI15 = LlNER_CLASS_3b_FREQ 1 .j 5 - LINERCLASS_3bFREQ 1 .lo Where:

ALERFLNERlGO15 = the change in LERF from 1-in-10 years ILRT interval to 1-in-15 years ILRT interval LlNER_CLASS_3bFREQ1 .j5 = frequency of EPRI accident Class 3b given a i-in-15 years ILRT interval = 9.16 x 10'9/ry [Step 7B]

LlNER_CLASS_3bFREQ1 .10 = frequency of EPRI accident Class 3b given a 1-in-10 years ILRT interval = 5.97 x 10'9/ry [Step 7B]

Therefore, ALERFLNER10 15 = 9.16 x10-9 - 5.97 xiQ 0 9 ALERFLNERIO.15 = 3.19xIcY 9/ry Based on this result, the inclusion of corrosion effects in the ILRT assessment would not change the previous conclusions of this calculation (See Sections 3 and 4). That is, the change in LERF from extending the interval to 15 years from the current 10 years requirement is estimated to be about 3.19 x I0'9/ry. This value is below the NRC Regulatory Guide 1.174 [5] of 10'7/yr. Therefore, because Regulatory Guide 1.174 [5] defines very small changes in LERF as below 10'7/yr, increasing the ILRT interval at Vermont Yankee from the currently allowed 1-in-1 0 years to 1-in-15 years and taking into consideration the likelihood of a containment liner flaw due to corrosion is non-risk significant from a risk perspective.

Similarly, the change in LERF from the original 3-in-1 0-year interval is calculated as follows:

ALERFLNER3-15 = LINERCLASS_3bFREO1.15 - LINERCLASS_3bYFREQ 3 10 t Where:

,ALERFLNEA13.15 = the change in LERF from 3-in-10 years ILRT interval to i-in-15 years ILRT interval LINERCLASS_3bFREQ 1.15 = frequency of EPRI accident Class 3b given a 1-in-15 years ILRT interval = 9.16 x 10'9/ry [Step 7B]

LINERCLASS_3b_FREQ 3 .1 0 = frequency of EPRI accident Class 3b given a 1-in-l1 years ILRT interval = 1.76 x 10'9/ry [Step 7B]

Therefore, ALERFLNER3-15 = 9.16x1O09 - 1.76 x 109 ALERFLNER3-15 = 7.40x10 9 /ry Similar to the ALERFLNER1o.15 result, the ALERFLNER315 is also non-risk significant from a risk perspective.

Step 9B - Evaluate the change in conditional containment failure probability This step calculates the change in conditional containment failure probability (CCFP). Similar to Section 2.4.9 Step 9 of this risk assessment, the change in CUEP tracts the impact of the ILRT on both early (LERF) and late radionuclide releases. Therefore, CCFP consists of all those accident sequences resulting in a radionuclide release other that the intact containment state for EPRI accident Class 1, and small failures state for EPRI accident Class 3a. In additional, the CCFP is conditional given a severe core damage accident. Therefore, the change in the conditional containment failure probability from 1-in-1 0 years to 1-in-15 years is:

ACCFPLINERIO 15 = CCFPLINER1-15 - CCFPLINERI-10 Where:

ACCFPLINER1O-15 = the change in conditional containment failure probability from 1-in-10 years to 1-in-15 years given non-detected containment leakage CCFPLINERI-10 = conditional containment failure probability given 1-in-10 years ILRT interval and potential non-detected containment leakage = 88.07% [Step 78]

CCFPLINERI-15 = conditional containment failure probability given 1-in-15 years ILRT interval and potential non-detected containment leakage = 88.13% [Step 7B]

Therefore, ACCFP LINERIO-15 = 88.13% - 88.07%

ACCFP LINERIO-15 = 0.06%

This change in ACCFP LINERlO-15 of less than 1% is insignificant from a risk perspective.

The results of Steps 7B, 8B, and 9B of the updated ILRT assessment including the potential impact from non-detected containment leakage scenarios assuming that 100% of the leakages result in EPRI Class 3b are show in Table B-5.

B6.0 Steel Shell Corrosion Sensitivity Additional sensitivity cases were also developed to gain an understanding of the sensitivity of this analysis to the various key parameters. The sensitivity cases are as follows:

  • Sensitivity Case 1 - Flaw rate doubles every 2 years
  • Sensitivity Case 2 - Flaw rate doubles every 10 years
  • Sensitivity Case 3 - 5% Visual inspection failures
  • Sensitivity Case 4 - 15% Visual inspection failures
  • Sensitivity Case 5 - Containment breach base point 10 times lower
  • Sensitivity Case 6 - Containment breach base point 10 times higher
  • Sensitivity Case 7 - Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5% visual inspection failures and 10% EPRI accident Class 3b are LERF (Lower bound)
  • Sensitivity Case 8 - Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15% visual inspection failures and 100% EPRI accident Class 3b are LERF (upper bound)

The above sensitivities cases used the calculational methodology presented in Steps 2B to 9B. These steps were developed in an EXCEL spreadsheet. They are reproduced in Attachment A.

These results are summarized in Table B-6.

B7.0 Conclusions This appendix provides a sensitivity evaluation of considering potential containment liner corrosion impacts within the structure of the ILRT interval extension risk assessment. The evaluation yields the following conclusions:

1. The impact of including age-adjusted corrosion effects in the ILRT assessment has minimal impact on plant risk and is therefore acceptable.
2. The change in LERF, taking into consideration the likelihood of a containment liner flaw due to age-adjusted corrosion is non-risk significant from a risk perspective. Specifically, extending the interval to 15 years from the current 10 years requirement is estimated to be about 3.19 x 10' 9/ry. This is below the Regulatory Guide 1.174 [5] acceptance criteria threshold of 10 7/yr.
3. The age-adjusted corrosion impact in dose increase is estimated to be 5.00 x 104 person-rem/ry or 0.0042% from the baseline ILRT 10 year's interval.
4. The age-adjusted corrosion impact on the conditional containment failure probability increase is estimated to be 0.06%.
5. A series of parametric sensitivity studies regarding potential age related corrosion effects on the containment steel liner also demonstrated minimal impact on plant risk.

Table B-1 Flaw Failure Rate as a Function of Time Accessible Area Inaccessible Area Drywell Floor Year Drywell and Torus Drywell and Torus Failure Rate Success Rate Failure Rate Success Rate Failure Rate Success Rate 0 1.79 x 10`3 9.98 x 10 ' 1.79 x 103 9.98 x 10l 4.46 x 10 4 1.00 1 2.05 x 10'3 9.98 x 10-' 2.05 x 10 9.98 x lo-, 5.13 x 104 9.99 x lo-,

2 2.36 x 10'3 9.98 x 10'l 2.36 x 10 9 .9 8 x 1 o-, 5.89 x 104 9.99 x 1o I 3 2.71 x 10' 9.97 x 10o- 2.71 x 10' 9.97 x 10o- 6.77 x 10`4 9.99 x 10 4 3.11 x 10'3 9.97 x 10'1 3.11 x 10'3 9.97 x 101- 7.78 x 1O04 9.99 x1lo-5 3.57 x 10 3 9.96 x 101 3.57 x 10'3 9.96 x 101 8.94 x 104 9.99 x 10 6 4.11 x 1o 9.96 x 10' 4.11 x 10 9.96 x 10 1.03 x 10 9.99 x10 7 4.72 x 10 9.95 x1 0 ' 4.72 x 10`3 9.95 x 105 1.1o8 x 1 03 9.99 x 10'l 8 5.42 x 103 9.95 xx1o-0l 5.42 x 10'3 9.95 x15o -l 1.36 x 10'3 9.99 x 10l 9 6.23 x 10' 9.94 x 1 0o' 6 .2 3 x 1 0 ' 9.94 x 1 0' 1.56 x 1i0'3 9.98 x 10-10 7.16 x 1 9.93 x 10-1 7.16 x 10'3 9.93 x 10 1.79 x 10'3 9.98 x 10 11 8.23 x 10 9.92 x 101 8.23 x 10'3 9.92 x 1o 2.06 x 10'3 9.98 x 10-'

12 9.45 x 1' 9.91 x 10l 9.45 x 10'3 9.91 x 10 ' 2.36 x 104 9.98 x 10lo 13 1.09 x 10 Z 9.89 x lo' 1.09 x 1o5Z 9.89 x 10 2.71 x 10- 9.97 x Io-'

14 1.25 x 10 z 9.88 x 10 1 1.25 x 10 z 9.88 x 101 3.12 x 10 9.97 x 10 15 1.43 x 10 9.86 x 10-l 1.43 x 10 9.86 x10o-T 3.58 x 10 9.96 x 10-Table B-2 Flaw Failure Rate as a Function of Test Interval Accessible Area Inaccessible Area Drywell Floor Years Drywell and Torus Drywell and Torus Failure Rate Success Rate Failure Rate Success Rate Failure Rate Success Rate 3-in-10 0.71% 9.93 x 10o1 0.71% 9.93 x 10.1 0.18% 9.98 x 10 1-in-10 4.14% 9.59 x 10' 4.14% 9.59 x 10' 1.04% 9.90 x1 1-in-15 9.68% 9.03 x 10" 9.68% 9.03 x 10 2.42% 9.76 x1

Table B-3 Vermont Yankee Containment Failure Probability Given Containment Liner Flaw Pressure (psia) l Containment Liner Drywell Floor Failure Probability Failure Probability 0 0.0004 0.00004 10 0.0006 0.0001 15 0.0008 0.0001 20 0.0010 0.0001 30 0.0017 0.0002 40 0.0028 0.0003 50 0.0047 0.0005 60 0.0078 0.0008 70 0.0130 0.0013 80 0.0217 0.0022 90 0.0362 0.0036 95 0.0468 0.0047 100 0.0605 0.0061 110 0.1011 0.0101 120 0.1688 0.0169 130 0.2820 0.0282 140 0.4710 0.0471 150 0.7866 0.0787 155 1.0 0.1 Figure B-1 Vermont Yankee Containment Failure Probability Given Containment Liner Flaw 1.00 0.90 0.80 Containment Uner 0.70 t 0.60

.0 0.

0.50 E

2= 0.40 U.0.2 0.30 0.20 0.10 0.00 0 20 40 60 80 100 120 140 160 Containment Pressure, psia

Table B-4 Vermont Yankee Containment Liner Corrosion Base Case Accessible Area Inaccessible Area Drywell Floor Step Description Drywell and Torus Drywell and Torus .-

1 Historical Steel Shell Flaw Likelihood 5.19 x 104 5.19 x 10' 3 1.30 x 10'3 2 Age Adjusted Steel Shell Flaw Year Failure Year Failure Year Failure Rate Likelihood Rate Rate 1 2.05 x 10'3 1 2.05 x 10'3 1 4.46 x 104 5-15 5.19 x 103 5-15 5.19 x 10' 3 5-15 1.30 x 103 15 1.43 x 102 15 1.43 x 10,2 15 3.58 x 104 3 Increase in Flaw Likelihood at 0.71% (3-to-10 years) 0.71% (3-to-10 years) 0.18% (3-to-10 years) 3,10, and 15 years 4.14% (1-to-10 years) 4.14% (1-to-10 years) 1.04% (1-to-10 years) 9.68% (1-to-1i years) 9.68% (1-to-1i years) 2.42% (1-to-1; years) 4 Likelihood of Breach in Pressure Likelihood Pressure Likelihood Pressure Likelihood Containment Given Steel (psia) of Breach (psia) of Breach (psia) of Breach Shell Flaw 20 0.0010 20 0.0010 20 0.0001 60.6 (ILRT) 0.0080 60.6 (ILRT) 0.0080 60.6 (ILRT) 0.0008 130 0.2820 130 0.2820 130 0.0282 140 0.4710 140 0.4710 140 0.0471 155 0.9658 155 0.9658 155 0.0966 5 Visual Inspection Detection 0.1 (10%) 1.0 (100%) 1.0 (100%)

Failure Likelihood 6 Likelihood of Non-Detected 0.00057% (3-to-10 years) 0.00570% (3-to-10 years) 0.00014% (3-to-10 years)

Containment Leakage (Steps 0.00332% (1-to-10 years) 0.03320% (1-to-10 years) 0.00083% (1-to-10 years) 3

  • 4- 5) 0.00776% (1-to-15 years) 0.07759% (1-to-15 years) 0.00194% (1-to-15 years)

Total Likelihood of Non-Detected 0.00642% (3-to-1 0 years) --

Containment Leakage 0.03737% (1-to-10 years) 0.08729% (1 -to-15 years)

Table B-5 Impact of Containment Steel Liner Corrosion on Vermont Yankee ILRT Intervals Base Case Extend to Extend to 3 Years 10 Years 15 Years EPRI CDF Per-Rem Per-Rem CDF Per-Rem Per-Rem CDF Per-Rem Per-Rem Class (Per Ry) (Per Ry) (Per Ry) (Per Ry) (Per Ry) (Per Ry) 1 5.88x 107 1.30x 103 7.64x 104 5.43x107 1.30x 103 7.06x10,4 5.11 x 107 1.30x 103 6.65x 104 2 2.14 x 10-9 2.73 x 106 5.85 x 103 2.14 x 109 2.73 x 106 5.85 x 10 3 2.14 x 109 2.73 x 106 5.85 x 103 3a 1.72 x 108 1.30 x 104 2.24 x 104 5.73 x 104 1.30 x 104 7.46 x10 4 8.60 x 104 1.30 x 104 1.12 x 103 3b 1.76 x 109 4.55 x 10 4 8.01 x 105 5.97 x 109 4.55 x 104 2.72 x 104 9.16 x 109 4.55 x 104 4.17 x 104 4 0.0 N/A 0.0 0.0 N/A 0.0 0.0 N/A 0.0 5 0.0 N/A 0.0 0.0 N/A 0.0 0.0 N/A 0.0 6 0.0 N/A 0.0 0.0 N/A 0.0 0.0 N/A 0.0 7b 3.22x 10 7 2.73x1 o6 8.79x10' 3.22x 107 2.73x1 o6 8.79x10' 3.22x 107 2.73x1 o 6 8.79x10' 7b 3.85 x 104 2.73 x 106 1.05 x 101 3.85 x 104 2.73 x 106 1.05 x 10' 3.85 x 104 2.73 x 106 1.05 x 101 7c 9.71 x 109 3.53 x 106 3.43 x 10-2 9.71 x 109 3.53 x 106 3.43 x10.2 9.71 x 10-9 3.53 x 106 3.43 x 102 8a 1.63 x 107 2.73 x 106 4.46 x 10' 1.63 x 107 2.73 x 106 4.46 x 10-' 1.63 x 107 2.73 x 106 4.46 x 10 '

8b 5.32 x 10o- 2.96 x 106 1.57 x 10 ' 5.32 x 104 2.96 x 106 1.57 x 10-1 5.32 x 104 2.96 x 106 1.57 x 10-1 Total 5.00x 104 12.0204 5.00x 104 12.0211 5.00x 104 12.0216 ILRT Dose Rate 3.04 x 104 1.02 x 103 1.54 x 103 from 3a and 3b (+1.86 x 104) (+1.08 x 10,5) (+2.53 x 105)

% Of Total 0.0025% 0.0085% 0.0128%

(+0.00002%) (+0.0001 %) (+0.0002%)

Delta Dose Rate 5.00 x 104 from 3a and 3b (+0.0014%)

(10 to 15 yr)

LERFfrom3b 1.76x 109 5.97x 109 9.16x 109

(+4.09 x 10") (+2.38 x 10.1o) (+5.56 x 10.10)

Delta LERF 3.19 x 10-9 (10 to 15 yr) (+3.18 x Io-"°)

CCFP % 87.98% 88.07% 88.13%

(+0.0008%) (+0.0167%) (+0.0111 %)

Delta CCFP % 0.06%

(10 to 15 yr) (+0.0063%)

' Denotes increase from original values presented in Section 2 of this report.

Table B-6 Containment Steel Liner Corrosion Sensitivity Cases Visual Lieiod LERF LERF LEAF Total LERF Drywell/ Inspection Likelihood Increase Increase Increase Increase Age Torus & Non- Flaw is From From From From ILRT (Step 2) Breach Visual (EPRI Corrosion Corrosion Corrosion Extension (Step 4) Flaws Class 3b) (3-in-10 (1-in-10 (1 to 15 (10 to 15 (Step 5) years) years) years) years)

Base Case Base Case Base Base Case Base Case Base Case Base Case Base Case Doubles 0.8017%Iiner Case 100% 4.09 x 10-" 2.38 x 10 .o 5.56 x 10.'0 3.19 x 10-9 every 5 yrs 0.0802%floor 10%

Doubles Base Base Base 1.17 x 10-" 1.99 x 101 1.15 x 10-9 3.82 x 10-9 every 2 yrs Doubles Base Base Base 6.08 x 10.11 8.33 x 10." 1.08 x 10.10 2.89 x 109 every 10 yrs_

Base Base 5% Base 3.91 x 10.1 2.28 x 10.'° 5.32 x 10.10 3179 x 10-9 Base Base 15% Base 4.27 x 1011 2.49 x 10.10 5.81 x 1010 3.20 x 10-9 9

Base 0.16033%Iiner 1 4 Base Base 8.18 x 1o-12 4.76 x 10. I 1.1 lx 10.10 2.93 x 10 0.01603%floor14 Base 4.0042% liner1 5 Base Base 2.04 x 10.'° 1.19 x 10 9 2.78 x 109 4.46 x 10 9 0.4004%floor' 5 ___

Lower Bound Doubles 10.1 6033%liner14 1 1 1 every 10 yrs J0.01603%floor' l 5% l 10% l 1.16 x 10. 5.03 x 10. 9.21 x 10 2.87 x 10-Upper Bound 5

Doubles 14.0042% liner' 1 1 6 l l every 2 yrs 10.4004%floor15 15 0% 1611x10' 10 0I60 - .5x1-14Base point 10 times lower than base case of 0.0001 at 20 psia.

5 Base point 10 times higher than base case of 0.01 at 20 psia.

Attachment A VY Risk Impact of Containment Liner Corrosion During an Extension of the ILRT Interval Results

Table of Contents Page No.

A1.0 Introduction A2 A2.0 Sensitivity Case 1 - Flaw rate doubles every 2 years A3 A3.0 Sensitivity Case 2 - Flaw rate doubles every 10 years A6 A4.0 Sensitivity Case 3 - 5% Visual inspection failures A9 A5.0 Sensitivity Case 4 - 15% Visual inspection failures A12 A6.0 Sensitivity Case 5 - Containment breach base point 10 times lower A15 A7.0 Sensitivity Case 6 - Containment breach base point 10 times higher A18 A8.0 Sensitivity Case 7 - Lower bound A21 A9.0 Sensitivity Case 8 - Upper bound) A24

A1.0 Introduction This attachment presents the results of the Vermont Yankee risk impact of containment liner corrosion during an extension of the ILRT interval. Seven sensitivity cases were examined. These are:

  • Sensitivity Case 1 - Flaw rate doubles every 2 years
  • Sensitivity Case 2 - Flaw rate doubles every 10 years
  • Sensitivity Case 3 - 5% Visual inspection failures
  • Sensitivity Case 4 - 15% Visual inspection failures
  • Sensitivity Case 5 - Containment breach base point 10 times lower
  • Sensitivity Case 6 - Containment breach base point 10 times higher
  • Sensitivity Case 7 - Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5% visual inspection failures and 10% EPRI accident Class 3b are LERF (Lower bound)
  • Sensitivity Case 8 - Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15% visual inspection failures and 100% EPRI accident Class 3b are LERF (upper bound)

The EXCEL spreadsheet results are presented in the following sections.

A2.0 Sensitivity Case 1 - Flaw Rate Doubles Every 2 Years 3-in-10 years From Estimated Change Inaccessible DrywellUTorus DW/Torus Drywell Floor 1 to 3 years 0.20% 0.20% 0.05%

1 to 10 years 3.46% 3.46% 0.86%

1 to 15 years 20.07% 20.07% 5.02%

Other Assumptions:

Containment Breach 0.8017% 0.8017% 0.0802%

Visual Inspection Failures 10.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies DrywellUTorus DWlTorus Drywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00016% 0.00163% 0.00004% 0.00184%

0.00184%

Release type VY Dose CDF . Case Dose Person-rem Frequency/ry 3 in 10 yrs Person-remlry 1 1.30E+03 5.88E-07 7.64E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 1.72E-08 0.00E+00 2.24E-04 3b 4.55E+04 1.73E-09 1.17E-11 7.88E-05 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.OOE+00 0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 Total ;-Ttl; .204 ;5.00E-06 ~.> :1-H2.0204  :;.

Risk Contribution: 0.003%

From 3a and 3b: 3.02E-04 3b LERF: 1.73E-09 CCFP: 87.98%

1-in-10 vears Increases to 3a and 3b Frequencies Drywell/Torus Inaccessible DW/TorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.00000%

0.0028% 0.0277% 0.0007% 0.03117%

0.03117%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 10 yrs Person-rem/ry 1 1.30E+03 5.43E-07 7.06E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 5.73E-08 O.OOE+00 7.46E-04 3b 4.55E+04 5.93E-09 1.99E-10 2.70E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 Total 5.00E-06, ,,;  ; , ,'2021f; Risk Contribution: 0.01%

From 3a and 3b: 1.02E-03 3b LERF: 5.93E-09 CCFP: 88.07%

1-in-15 years Increases to 3a and 3b FrequenciesDrywelllTorusinaccessible DW/TorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.00000%

0.0161% 0.1609% 0.0040% 0.18099%

0.18099%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 15 yrs Person-remlry 1 1.30E+03 5.11 E-07 6.64E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 8.60E-08 O.OOE+00 1.12E-03 3b 4.55E+04 9.76E-09 1.15E-09 4.44E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 Total  : ': 5.00E-06"' '12.0216 Risk Contribution: 0.01%

From 3a and 3b: 1.56E-03 3b LERF: 9.76E-09 CCFP: 88.14%

Other Pertinent Risk Metrics:

10to 15 Increase (Person-rem/ry)- -  ; 5.04E-04 3to 15 Increase (Person-remiry): 1.16E-03 10to 15 Delta-LERF ; -. -3.82E-09 3 to 15 Delta-LERF: 8.02E-09 10 to 15 Delta-CCFP.;: 0.08%

3 to 15 Delta-CCFP: 0.16%

3 to 15 Delta-LERF from Corrosion: 8.02E-09 10 to 15 Delta-LERF from Corrosion: 3.82E-09 Increase in LERF (ILRT 3-to-15 years) 1.08E-08

A3.0 Sensitivity Case 2 - Flaw Rate Doubles Every 10 Years 3-in-10 years From Estimated Change Inaccessible Drywell/Torus DWlTorus Drywell Floor 1 to 3 years 1.06% 1.06% 0.26%

I to 10 years 4.58% 1.06% 1.15%

1 to 15 years 8.38% 1.06% 2.10%

Other Assumptions:

Containment Breach 0.8017% 0.8017% 0.0802%

Visual Inspection Failures 10.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Frequencies Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00085% 0.00848% 0.00021% 0.00955%

0.00955%

Release tyype VY Dose CDF Case Dose Person-rem Frequency/ry 3inlOyrs PC erson-rem/ry 1 1.30E+03 5.87E-07 7.64E-04 2 2.73E+06 2.14E-09 C1:rrosion Addition 5.85E-03 3a 1.30E+04 1.72E-08 0.OOE+00 2.24E-04 3b 4.55E+04 1.78E-09 6.08E-1 1 8.1 OE-05 4 N/A 0.OOE+00 0.OOE+00 5 N/A O.OOE+00 0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 Total 5.OOE-06 '12.0204 - - -

Risk Contribution: 0.003%

From 3a and 3b: 3.05E-04 3b LERF: 1.78E-09 CCFP: 87.98%

1-in-10 vears Increases to 3a and 3b Frequencies DrywelllTorusinaccessible DWI1orusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.00000%

0.0037% 0.0085% 0.0009% 0.01308%

0.01308%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 10 yrs Person-remlry 1 1.30E+03 5.43E-07 7.06E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 5.73E-08 O.OOE+00 7.46E-04 3b 4.55E+04 5.82E-09 8.33E-1 1 2.65E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01

_ Tota -..... -5.00E-06

-06 20211 Risk Contribution: 0.01%

From 3a and 3b: 1.01 E-03 3b LERF: 5.82E-09 CCFP: 88.06%

1-in-15 vears Increases to 3a and 3b Frequencies Drywell/Torus Inaccessible DW/Torus Drywell Floor Total 0.0000% 0.0000% 0.0000% 0.00000%

0.0067% 0.0085% 0.0017% 0.01689%

0.01689%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 15 yrs Person-rem/ry 1 1.30E+03 5.12E-07 6.65E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 8.60E-08 O.OOE+00 1.12E-03 3b 4.55E+04 8.71 E-09 1.08E-10 3.96E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 Total 5.00E-06 : 12.0215 Risk Contribution: 0.01%

From 3a and 3b: 1.51 E-03 3b LERF: 8.71 E-09 CCFP: 88.1 2%

Other Pertinent Risk Metrics:

10 to 15 Increase (Person-rem/ry): 4.63E-04 3 to 15 Increase (Person-rem/ry): 1.11 E-03 10 to 15 Delta-LERF ';-'2.89E-09 3to 15 Delta-LERF: 6.93E-09 10 to 15 Delta-CCFP:-' 0.06%

3 to 15 Delta-CCFP: 0.14%

3 to 15 Delta-LERF from Corrosion: 6.93E-09 10 to 15 Delta-LERF from Corrosion: 2.89E-09 Increase in LERF (ILRT 3-to-15 years) 1.99E-09

A4.0 Sensitivity Case 3 - 5% Visual Inspection Failures 3-in-10v ears From Estimated Change Inaccessible DrywelUTorus DWJTorus Drywell Floor 1 to 3 years 0.71% 0.71% 0.18%

1 to 10 years 4.14% 4.14% 1.04%

1 to 15 years 9.68% 9.68% 2.42%

Other Assumptions:

Containment Breach 0.8017% 0.8017% 0.0802%

Visual Inspection Failures 5.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies DrywellTorus DWlTorus Drywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0003% 0.0057% 0.0001% 0.0061%

0.0061%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 3 in 10 yrs Person-rem/ry 1 1.30E+03 5.88E-07 7.64E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 1.72E-08 0.OOE+00 2.24E-04 3b 4.55E+04 1.76E-09 3.91 E-1 1 8.01 E-05 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.OOE+00 0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01

~ otal - 500E*6 2 020 0;-2:0^

Risk Contribution: 0.003%

From 3a and 3b: 3.04E-04 3b LERF: 1.76E-09 CCFP: 87.98%

1-in-10 years Increases to 3a and 3b Frequencies DrywelUTorus Inaccessible DW/rorus Drywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0017% 0.0332% 0.0008% 0.0357%

0.0357%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 10 yrs Person-rem/ry 1 1.30E+03 5.43E-07 7.06E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 5.73E-08 O.OOE+00 7.46E-04 3b 4.55E+04 5.96E-09 2.28E-10 2.71 E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01

,'Total", - 5.00E-06 '; 12.02117 Risk Contribution: 0.01%

From 3a and 3b: 1.02E-03 3b LERF: 5.96E-09 CCFP: 88.07%

1-in-15 Vears Increases to 3a and 3b Frequencies DrywellTorusinaccessible DW/TorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0039% 0.0776% 0.0019% 0.0834%

0.0834%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 15 yrs Person-rem/ry 1 1.30E+03 5.11E-07 6.65E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 8.60E-08 O.OOE+00 1.12E-03 3b 4.55E+04 9.13E-09 5.32E-10 4.16E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 *3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 Total 5.00E-06 - 12.0216' Risk Contribution: 0.01%

From 3a and 3b: 1.53E-03 3b LERF: 9.13E-09 CCFP: 88.13%

Other Pertinent Risk Metrics:

10 to 15 Increase (Person-rem/ry): - - 4.76E-04 3 to 15 Increase (Person-rem/ry): 1.13E-03 10to 15 Delta-LERF: 3.17E-09 3 to 15 Delta-LERF: 7.37E-09 10 to 15 Delta-CCFP ;0.06%

3 to 15 Delta-CCFP: 0.15%

3 to 15 Delta-LERF from Corrosion: 7.37E-09 10 to 15 Delta-LERF from Corrosion: 3.17E-09 Increase in LERF (ILRT 3-to-15 years) 6.30E-09

A5.0 Sensitivity Case 4 - 15% Visual Inspection Failures 3-in-10 years From Estimated Change Inaccessible DrywelUTorus DW/rorus Drywell Floor 1 to 3 years 0.71% 0.71% 0.18%

1 to 10 years 4.14% 4.14% 1.04%

1 to 15 years 9.68% 9.68% 2.42%

Other Assumptions:

Containment Breach 0.8017% 0.8017% 0.0802%

Visual Inspection Failures 15.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies Drywell/Torus DWlTorus Drywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0009% 0.0057% 0.0001% 0.0067%

0.0067%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 3 in 10 yrs Person-rem/ry 1 1.30E+03 5.88E-07 7.64E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 1.72E-08 0.OOE+00 2.24E-04 3b 4.55E+04 1.76E-09 4.27E-1 1 8.02E-05 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.OOE+00 0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01

-Total 5.00E-06  : 1

i2.0204 Risk Contribution: 0.003%

From 3a and 3b: 3.04E-04 3b LERF: 1.76E-09 CCFP: 87.98%

1-in-10 vears Increases to 3a and 3b Frequencies DrywellTorusInaccessible DW/TorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0050% 0.0332% 0.0008% 0.0390%

0.0390%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 10 yrs Person-rern/ry 1 1.30E+03 5.43E-07 7.06E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 5.73E-08 0.00E+00 7.46E-04 3b 4.55E+04 5.98E-09 2.49E-10 2.72E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 r '.Tota- 1 20211 '

Risk Contribution: 0.01%

From 3a and 3b: 1.02E-03 3b LERF: 5.98E-09 CCFP: 88.07%

1-in-15 years Increases to 3a and 3b Frequencies DrywelUTorus Inaccessible DW/TorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0116% 0.0776% 0.0019% 0.0912%

0.0912%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 15 yrs Person-remlry 1 1.30E+03 5.11 E-07 6.65E-04 2 2.73E+06 2.14E-09 Corrosion Additi on 5.85E-03 3a 1.30E+04 8.60E-08 O.OOE+00 1.12E-03 3b 4.55E+04 9.18E-09 5.81 E-10 4.1 8E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01

Total 5.00E '.1 .. . .~

12.0216 Risk Contribut ion: 0.01%

From 3a and 3b: 1.54E-03 3b LERF: 9.18E-09 CC,FP: 88.13%

Other Pertinent Risk Metrics:

10 to 15 Increase (Person-rem/ry) . 4.77E-04 3 to 15 Increase (Person-rem/ry): 1.13E-03 10 to 15 Delta-LERF:. , , 3.20E-09 3 to 15 Delta-LERF: 7.42E-09 10 to 15 Delta-CCFP: - - V.

0.06%

3to 15 Delta-CCFP: 0.15%

3 to 15 Delta-LERF from Corrosion: 7.42E-09 10 to 15 Delta-LERF from Corrosion: 3.20E-09 Increase in LERF (ILRT 3-to-15 years) 6.89E-09

A6.0 Sensitivity Case 5 - Containment Breach Base Point 10 Times Lower 3-in-10 years From Estimated Change Inaccessible DrywellUTorus DWJTorus Drywell Floor 1 to 3 years 0.71% 0.71% 0.18%

1 to 10 years 4.14% 4.14% 1.04%

1 to 15 years 9.68% 9.68% 2.42%

Other Assumptions:

Containment Breach 0.1603% 0.1603% 0.0160%

Visual Inspection Failures 10.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies Drywell/Torus DW/Torus Drywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00011% 0.00114% 0.00003% 0.00128%

0.00128%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 3 in 10 yrs Person-rem/ry 1 5.88E-07 7.64E-04 5.88E-07 2 2.14E-09 Corrosion Addition 5.85E-03 2.14E-09 3a 1.72E-08 0.00E+00 2.24E-04 1.72E-08 3b 1.73E-09 8.18E-12 7.87E-05 1.73E-09 4 0.OOE+00 0.OOE+00 0.OOE+00 5 0.OOE+00 0.OOE+00 0.OOE+00 6 0.OOE+00 0.OOE+00 0.OOE+00 7a 3.22E-07 8.79E-01 3.22E-07 7b 3.85E-06 1.05E+01 3.85E-06 7c 9.71 E-09 3.43E-02 9.71 E-09 8a 1.63E-07 4.46E-01 1.63E-07 8b 5.32E-08 1.57E-01 5.32E-08

Total : 500E06 .. - ;120204 _ 5.00E-06 Risk Contribution: 0.0025%

From 3a and 3b: 3.02E-04 3b LERF: 1.73E-09 CCFP: 87.98%

1-in-10 vears Increases to 3a and 3b Frequencies Drywel/Torus Inaccessible DWlTorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00066% 0.00664% 0.00017% 0.00747%

0.00747%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 10 yrs Person-rem/ry 1 5.43E-07 7.06E-04 5.43E-07 2 2.14E-09 Corrosion Addition 5.85E-03 2.14E-09 3a 5.73E-08 O.OOE+00 7.46E-04 5.73E-08 3b 5.78E-09 4.76E-1 1 2.63E-04 5.78E-09 4 O.OOE+00 O.OOE+00 O.OOE+00 5 O.OOE+00 O.OOE+00 O.OOE+00 6 O.OOE+00 O.OOE+00 0.OOE+00 7a 3.22E-07 8.79E-01 3.22E-07 7b 3.85E-06 1.05E+01 3.85E-06 7c 9.71 E-09 3.43E-02 9.71 E-09 8a 1.63E-07 4.46E-01 1.63E-07 8b 5.32E-08 1.57E-01 5.32E-08 Total 5.0 E-06 :2.02ii ^ 5.00E06 Risk Contribution: 0.0084%

From 3a and 3b: 1.01 E-03 3b LERF: 5.78E-09 CCFP: 88.06%

1-in-15 years Increases to 3a and 3b Frequencies DrywellTorusinaccessible DWlTorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00155% 0.01552% 0.00039% 0.01746%

0.01746%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 15 yrs Person-rem/ry 1 5.12E-07 6.65E-04 5.12E-07 2 2.14E-09 Corrosion Addition 5.85E-03 2.14E-09 3a 8.60E-08 O.OOE+00 1.12E-03 8.60E-08 3b 8.71 E-09 1.11 E-10 3.96E-04 8.71 E-09 4 O.OOE+00 O.OOE+00 O.OOE+00 5 O.OOE+00 O.OOE+00 O.OOE+00 6 O.OOE+00 O.OOE+00 O.OOE+00 7a 3.22E-07 8.79E-01 3.22E-07 7b 3.85E-06 1.05E+01 3.85E-06 7c 9.71 E-09 3.43E-02 9.71 E-09 8a 1.63E-07 4.46E-01 1.63E-07 8b 5.32E-08 1.57E-01 5.32E-08

-Total 5.00E 12.0215 5.00E-06 Risk Contribution: 0.0126%

From 3a and 3b: 1.51 E-03 3b LERF: 8.71 E-09 CCFP: 88.12%

Other Pertinent Risk Metrics:

10 to 15 Increase (Person-rem/ry).- 4.65E-04 3 to 15 Increase (Person-rem/ry): 1.11E-03 10 to 15 Delta-LERF: 2.93E-09 3 to 15 Delta-LERF: 6.98E-09 10to 15 Delta-CCFP! ,- 0.058%

3to 15 Delta-CCFP: 0.14%

3 to 15 Delta-LERF from Corrosion: 6.98E-09 10 to 15 Delta-LERF from Corrosion: 2.93E-09 Increase in LERF (ILRT 3-to-15 years) 1.32E-09

A7.0 Sensitivity Case 6 - Containment Breach Base Point 10 Times Higher 3-in-10 years From Estimated Change Inaccessible DrywelUTorus DWlTorus Drywell Floor 1 to 3 years 0.71% 0.71% 0.18%

1 to 10 years 4.14% 4.14% 1.04%

1 to 15 years 9.68% 9.68% 2.42%

Other Assumptions:

Containment Breach 4.0042% 4.0042% 0.4004%

Visual Inspection Failures 10.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies Drywell/Torus DW/Torus Drywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00285% 0.02849% 0.00071% 0.03205%

0.03205%

Release tjype VY Dose CDF Case Dose Person-rem Frequency/ry 3 in 10 yrs Person-rem/ry 1 1.30E+03 5.87E-07 7.64E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 1.72E-08 0.00E+00 2.24E-04 3b 4.55E+04 1.92E-09 2.04E-10 8.76E-05 4 NIA 0.OOE+00 0.00E+00 5 N/A 0.OOE+00 0.00E+00 6 N/A 0.OOE+00 0.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01

'- Total . ,5.00E-06 :12.0204 Risk Contribution: 0.0028%

From 3a and 3b: 3.23E-04 3b LERF: 2.OOE-09 CCFP: 87.12%

1-in-10 years Increases to 3a and 3b Frequencies DrywelITorusInaccessible DW/TorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.01659% 0.16593% 0.00415% 0.18667%

0.18667%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 10 yrs Person-remlry 1 1.30E+03 5.42E-07 7.05E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 5.73E-08 O.OOE+00 7.46E-04 3b 4.55E+04 6.92E-09 1.19E-09 3.15E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01

, oal::^5.OOE-0  ;: 12.021 Risk Contribution: 0.0088%

From 3a and 3b: 1.06E-03 3b LERF: 6.92E-09 CCFP: 88.09%

1-in-15 years Increases to 3a and 3b Frequencies DrywelrTorusInaccessible DWlTorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.03876% 0.38756% 0.00969% 0.43600%

0.43600%

Release type VY Dose CDF Case . Dose Person-rem Frequency/ry 1 in 15 yrs Person-rem/ry 1 1.30E+03 5.09E-07 6.62E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 8.60E-08 O.OOE+00 1.12E-03 3b 4.55E+04 1.14E-08 2.78E-09 5.18E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A 0.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 Total - I5.OOE-06 12.0217.-

Risk Contribution 0.0136%

From 3a and 3b : 1.64E-03 3bLERF 1.1 4E-08 CCFP ': 88.18%

Other Pertinent Risk Metrics:

10 to 15 Increase (Person-rem/ry)-' -5.32E-04 3 to 15 Increase (Person-rem/ry): 1.22E-03 10 to 15 Delta-LERF- -4.46E-09 3 to 15 Delta-LERF: 9.46E-09 10 to 15 Delta-CCFP:., ' 0.089%

3 to 15 Delta-CCFP: 0.19%

3 to 15 Delta-LERF from Corrosion: 9.46E-09 10 to 15 Delta-LERF from Corrosion: 4.46E-09 Increase in LERF (ILRT 3-to-1 5 years) 3.29E-08

A8.0 Sensitivity Case 7 - Lower bound (Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5% visual inspection failures and 10% EPRI accident Class 3b are LERF) 3-in-10 years From Estimated Change Inaccessible DrywelltTorus DW/Torus Drywell Floor 1 to 3 years 1.06% 1.06% 0.26%

1 to 10 years 4.58% 4.58% 1.15%

1 to 15 years 8.38% 8.38% 2.10%

Other Assumptions:

Containment Breach 0.1603% 0.1603% 0.0160%

Visual Inspection Failures 5.0% 100.0% 100.0%

EPRI Class 3a Fraction 90.0% 90.0% 90.0%

EPRI Class 3b Fraction 10.0% 10.0% 10.0%

Increases to 3a and 3b Inaccessible Frequencies DrywelltTorus DWtTorus Drywell Floor Total 0.00008% 0.00153% 0.00004% 0.00164%

0.00001% 0.00017% 0.00000% 0.00018%

0.00182%

Release type VY Dose CDF Case Dose Person-rem Frequencylry 3 in 10 yrs Person-rem/ry 1 1.30E+03 5.88E-07 7.64E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 1.72E-08 1.05E-1 1 2.24E-04 3b 4.55E+04 1.72E-09 1.16E-12 7.83E-05 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.OOE+00 '0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01

, Total 5.00E-06 ~ ' ': - 12;0204'

Risk Contribution: 0.0025%

From 3a and 3b: 3.02E-04 3b LERF: 1.72E-09 CCFP: 87.98%

1-in-lO yea rs Increases to 3a and 3b Frequencies DrywelllToruslInaccessible DWJTorus Drywell Floor Total 0.00033% 0.00661% 0.00017% 0.00711%

0.00004% 0.00073% 0.00002% 0.00079%

0.00790%

Release type VY Dose CDF Case Dose Person-rem Frequencylry 1 in 10 yrs Person-rem/ry 1 1.30E+03 5.43 E-07 7.06 E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 5.74E-08 4.53E-1 1 7.46E-04 3b 4.55E+04 5.74E-09 5.03 E-1 2 2.61 E-04 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.OOE+00 0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a. 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 -~5.32E-08 -1 .571E-01 V o5.0E-06.' i2.02 1.

Risk Contribution: 0.0084%

From 3a and 3b: 1.01 E-03 3b LERF: 5.74E-09 CCFP: 88.06%

1-in-i5 years Increases to 3a and 3b Frequencies DrywellYrorus Inaccessible DW/TorusDrywell Floor Total 0.00060% 0.01210% 0.00030% 0.01301%

0.00007% 0.00134% 0.00003% 0.00145%

0.01445%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 15 yrs Person-rernlry 1 1.30E+03 5.1 2E-07 6.65E-04 2 2.73E+06 2.1 4E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 8.61 E-08 8.29E-1 1 1.12E-03 3b 4.55E+04 8.61 E-09 9.21 E-1 2 3.92E-04 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.OOE+00 0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 I Total ..:5.00OE-06 12.0215 -

Risk Contribution: 0.0126%

From 3a and 3b: 1.51 E-03 3b LERF: 8.61 E-09 CCFP: 88.12%

Other Pertinent Risk Metrics:

10 to 15 Increase (Person-rem/ry)- -. 4.63E-04 3 to 15 Increase (Person-rem/ry): 1.11 E-03 10 to 15 Delta-LERF: - 2.87E-09 3 to 15 Delta-LERF: 6.89E-09 10 to 15 DeIta-CCFP:!7-- 0.057 %

3 to 15 Delta-CCFP: 0.14%

3 to 15 Delta-LERF from Corrosion: 6.89E-09 10 to 15 Delta-LERF from Corrosion: 2.87E-09 Increase in LERF (ILRT 3-to-15 years) 1.22E-09

A9.0 Sensitivity Case 8 - Upper Bound (Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15% visual inspection failures and 100% EPRI accident Class 3b are LERF) 3-in-10 vears From Estimated Change Inaccessible DrywelUTorus DW/Torus Drywell Floor 1 to3years 0.20% 0.20% 0.05%

1 to10 years 3.46% 3.46% 0.86%

1 to 15 years 20.07% 20.07% 5.02%

Other Assumptions:

Containment Breach 4.0042% 4.0042% 0.4004%

Visual Inspection Failures 15.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies Drywell/Torus DW/Torus Drywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00122% 0.00816% 0.00020% 0.00959%

0.00959%

Release type VY Dose CDF Case Dose Person-rem Frequencylry 3 in 10 yrs Person-remlry 1 1.30E+03 5.87E-07 7.64E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 1.72E-08 0.OOE+00 2.24E-04 3b 4.55E+04 1.78E-09 6.11 E-11 8.11 E-05 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.00E+00 0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 Total - 5.OOE-06 i  : 12.0204 Risk Contribution: 0.0025%

From 3a and 3b: 3.05E-04 3b LERF: 1.78E-09 CCFP: 87.98%

1-in-10 years Increases to 3a and 3b Frequencies Drywell/TorusInaccessible DW/TorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.02076% 0.13837% 0.00346% 0.16259%

0.16259%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 10 yrs Person-rem/ry 1 1.30E+03 5.42E-07 7.05E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 5.73E-08 O.OOE+00 7.46E-04 3b 4.55E+04 6.77E-09 1.04E-09 3.08E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 7d 2.73E+06 1.63E-07 4.46E-01 8 2.96E+06 5.32E-08 1.57E-01

- Total -5.00E 06 .: 2.021i Risk Contribution: 0.0088%

From 3a and 3b: 1.05E-03 3b LERF: 6.77E-09 CCFP: 88.08%

1-in-15 years Increases to 3a and 3b Frequencies DrywelirorusInaccessible DW/TorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.12053% 0.80354% 0.02009% 0.94416%

0.94416%

Release type VY Dose CDF Case Dose Person-rem Frequency/ry 1 in 15 yrs Person-rern/ry 1 1.30E+03 5.06E-07 6.58E-04 2 2.73E+06 2.14E-09 Corrosion Addition 5.85E-03 3a 1.30E+04 8.60E-08 O.OOE+00 1.12E-03 3b 4.55E+04 1.46E-08 6.02E-09 6.65E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 2.73E+06 3.22E-07 8.79E-01 7b 2.73E+06 3.85E-06 1.05E+01 7c 3.53E+06 9.71 E-09 3.43E-02 8a 2.73E+06 1.63E-07 4.46E-01 8b 2.96E+06 5.32E-08 1.57E-01 Total '- : '5.OOE-06 12.0218 Risk Contribution: 0.0148%

From 3a and 3b: 1.78E-03 3b LERF: 1.46E-08 CCFP: 88.24%

Other Pertinent Risk Metrics:

10 to 15 Increase (Person-rem/ry)- - 6.82E-04 3 to 15 Increase (Person-rem/ry): 1.37E-03 10 to 15 Delta-LERF:-. 7.85E-09 3 to 15 Delta-LERF: 1.28 E-08 10 to 15 Delta-CCFP - 0.1 56%

3 to 15 Delta-CCFP: 0.26%

3 to 15 Delta-LERF from Corrosion: 1.28E-08 10 to 15 Delta-LERF from Corrosion: 7.85E-09 Increase in LERF (ILRT 3-to-15 years) 5.62E-08

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