ML040280609

Erin Report C467030603-5572, Dresden Risk Assessment to Support ILRT (Type a) Interval Extension Request

ML040280609
Person / Time
Site:	Dresden
Issue date:	09/30/2003
From:	Andersen V, Burns E, Kirchner R, Schlenger-Faber B, Teagarden G Engineering & Research
To:	Office of Nuclear Reactor Regulation
References
C467030603-5572
Download: ML040280609 (108)

Text

E1N Engineering and Research, Inc.

an SKF Go Conwny DRESDEN RISK ASSESSMENT TO SUPPORT ILRT (TYPE A)

INTERVAL EXTENSION REQUEST ERIN Report No. C467030603-5572 Principal Contributors V.M. Andersen E.T. Burns B.J. Schlenger-Faber G.A. Teagarden R.F. Kirchner Prepared for Exelon Nuclear Dresden SEPTEMBER 2003

Risk Impact Assessment of Extending Dresden ILRTInterval TABLE OF CONTENTS Section PaQe EXECUTIVE

SUMMARY

ii

1.0 INTRODUCTION

.1-1 1.1 Purpose.1-1 1.2 Background.1-1 1.3 Criteria.1-4 2.0 METHODOLOGY.2-1 2.1 General Resources Available.2-1 2.2 NEI Interim Guidance.2-6 2.3 Ground Rules.2-8 2.4 Plant-Specific Inputs.2-9 3.0 ANALYSIS.3-1 3.1 Baseline Accident Category Frequencies (Step 1).3-1 3.2 Containment Leakage Rates (Step 2).3-11 3.3 Baseline Population Dose Rate Estimates (Steps 3-4).3-12 3.4 Impact of Proposed ILRT Interval (Steps 5-9).3-20 3.5 Sensitivities.3-27 4.0 RESULTS

SUMMARY

.4-1

5.0 CONCLUSION

S.5-1 5.1 Quantitative Conclusions.5-1 5.2 Risk Trade-Off.5-2 5.3 External Events Impact.5-3 5.4 Previous Assessments.

5-4

6.0 REFERENCES

.6-1 Appendix A EXTERNAL EVENT ASSESSMENT Appendix B CONTAINMENT CORROSION SENSITIVITY Appendix C SENSITIVITY FOR LONG TERM STATION BLACKOUT i

C467030603-5572-12117/03

Risk Inpact Assessment ofExtending Dresden ILRTInterval EXECUTIVE

SUMMARY

The risk impact of a one-time extension of the Dresden integrated leak rate test (ILRT) interval from the currently approved 10 years to 15 years is evaluated. The results demonstrate that a change in the ILRT test interval from 10 years to 15 years represents a "very small" impact on risk, as defined by Regulatory Guide 1.174.

The Dresden ILRT risk assessment uses Dresden specific information to calculate the changes to the risk profile due to changes to the ILRT interval. The evaluation approach for the assessment of the risk is based on EPRI-TR-1 04285, NEI Interim Guidance (dated November 2001), and previous ILRT risk assessment submittals. The full power internal events PRA model for Dresden is used as the primary basis of the assessment. External events are addressed by sensitivity evaluations. The ex-plant consequences are based on plant specific values calculated as part of the Severe Accident Mitigation Alternatives (SAMA) evaluation for the Dresden License Renewal Application.

The Dresden accident sequence frequencies and doses are assigned to an equivalent EPRI category for consideration of the effects of ILRT interval changes. Three of the EPRI categories are affected by ILRT interval changes (1, 3a, and 3b). Table ES-1 compares the results of various risk measures for the current 10-year ILRT interval with the proposed 15-year ILRT interval.

Three risk measures are evaluated using the Dresden 2002A internal events PSA model to characterize the reduction in ILRT frequency from 1-per-10 years to 1-per-15 years:

ii ii

~~~~~~~C467030603-557

Risk Impact Assessment of Extending Dresden ILRT Interval Risk Measures Change in Large Early Release Frequency (LERF)

Change in conditional containment failure probability (CCFP)

Change in population dose rate (person-rem/yr)

Risk Increase 7.65E-9/yr 0.4%

Negligible(1)

The first risk measure change is considered by Reg. Guide 1.174 as a very small" impact on risk. The other two risk measure changes do not have criteria in Reg. Guide 1.174, but based on past ILRT interval extension requests these changes are also considered to represent "very small" impacts on risk.

Additionally, several sensitivity cases are evaluated and documented in this analysis.

These sensitivity cases demonstrate the following:

Inclusion of long-term station blackout scenarios in the EPRI categories 3a and 3b frequencies increases the risk measures a negligible amount and does not change the conclusion of this report.

LERF is not significantly impacted by the potential for containment leakage due to age-related degradation in non-inspectable areas; the ALERF remains within Region IlIl as a very small" risk change.

The inclusion of external events increases LERF by approximately an order of magnitude; however, the ALERF remains within Region IlIl as a "very small" risk change.

(

There is no calculated change in the total population dose rate unless more than four significant digits are utilized. See Sensitivity Calculation in Section 3.5.3.

iii C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval Table ES-1 QUANTITATIVE RESULTS AS A FUNCTION OF ILRT INTERVAL Quantitative Results as a Function of ILRT Interval Current Proposed (10-year ILRT Interval)

(15-year ILRT Interval)

Dose Population Dose Population Dose (Person-Rem Accident Rate (Person-Accident Rate (Person-EPRI Within 50-Frequency Rem/Year Within Frequency Rem/ Year Within Category Category Description miles)(t )

(per year) 50-miles)

(per year) 50-miles) 1 No Containment Failure(2) 2.08E+3 9.52E-7 1.98E-3 8.68E-7 1.80E-3 2

Containment Isolation System Failure 2.22E+7 4.67E-9 1.04E-1 4.67E-9 1.04E-1 3a Small Pre-Existing Failures (2) (3) 2.08E+4 1.53E-7 3.18E-3 2.29E-7 4.77E-3 3b Large Pre-Existing Failures(2 )' (3) 7.28E+4 1.53E-8 1.11 E-3 2.29E-8 1.67E-3 4

Type B Failures (LLRT)(5)

N/A N/A N/A N/A N/A 5

Type C Failures (LLRT)( 5)

N/A N/A N/A N/A N/A 6

Other Containment Isolation System Failure N/A N/A N/A N/A N/A 7

Containment Failure Due to Severe Accident 1.33E+7 7.65E-7 1.02E+1 7.65E-7 1.02E+1 8

Containment Bypass Accidents 2.79E+7 1.74E-9 4.85E-2 1.74E-9 4.85E-2 TOTALS:

1.89E-6 10.3 1.89E-6 10.3 Increase in Dose Rate Increase in LERF Increase in CCFP (%)

iv C467030603-5572-12/17/03

Risk Impact Assessment ofExtending Dresden ILRTInterval Notes to Table ES-1:

(1)

The population dose associated with the release categories is based on plant specific dose calculations (MACCS).

(2)

Only EPRI categories 1, 3a, and 3b are affected by ILRT (Type A) interval changes. The quoted values are based upon a containment leakage rate of 0.5%/day.

The Technical Specification leakage rate may increase as a result of the use of the Alternate Source Term (AST) methodology.

Categories 1, 3a and 3b would increase. A sensitivity evaluation of the T.S. leakage change confirms that the change in population dose rate remains negligible even for the higher containment leak rate. (See Section 3.5.3.)

(3)

Dose estimates for categories 3a and 3b, per the NEI Interim Guidance, are calculated as 10xCategory I dose and 35xCategory 1 dose, respectively.

(4)

There is no calculated change in the total population dose rate unless more than four significant digits are utilized. See Section 3.5.3 for sensitivity evaluation of the population dose.

(5)

An LLRT is a local leak rate test which is not affected by the ILRT frequency.

v C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval Section 1 INTRODUCTION 1.1 PURPOSE The purpose of this analysis is to provide an assessment of the risk associated with implementing a one-time extension of the Dresden containment Type A integrated leak rate test (ILRT) interval from ten years to fifteen years. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages.

The risk assessment follows the guidelines from NEI 94-01 [1], the methodology used in EPRI TR-104285 [2], the NEI Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals [3], NEI Additional Information for ILRT Extensions [21], and the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PSA) 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 [4].

1.2 BACKGROUND

Revisions to 10CFR50, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing requirements from three-in-ten years to at least once per ten years. The revised Type A 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 is less than normal containment leakage of 1.01La (allowable leakage).

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

Section 11.0 of NEI 94-01 states that NUREG-1493 [5],

"Performance-Based Containment Leak Test Program," September 1995, provides the technical basis to support rulemaking to revise leakage rate testing requirements 1-1 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRT Interval contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRC's rulemaking basis, NEI undertook a similar study. The results of that study are documented in Electric Power Research Institute (EPRI) Research Project Report TR-1 04285.

The NRC report, Performance Based Leak Test Program, NUREG-1493 [5], analyzed the effects of containment leakage on the health and safety of the public and the benefits realized from the containment leak rate testing. In that analysis, it was determined for a comparable BWR plant that increasing the containment leak rate from the nominal 0.5 percent per day to 5 percent per day leads to a barely perceptible increase in total population exposure, and increasing the leak rate to 50 percent per day increases the total population exposure by less than 1 percent.

Consequently, extending the ILRT interval should not lead to any substantial increase in risk. The current analysis is being performed to confirm these conclusions based on Dresden specific models and available data.

Earlier ILRT frequency extension submittals have used the EPRI TR-104285 methodology to perform the risk assessment. In November and December 2001, NEI issued enhanced guidance (hereafter referred to as the NEI Interim Guidance) that builds on the EPRI TR-104285 methodology and is intended to provide for more consistent submittals to the NRC. [3,21] The NEI Interim Guidance was developed for NEI by EPRI using personnel who also developed the EPRI TR-104285 methodology. This Dresden ILRT interval extension risk assessment employs the NEI Interim Guidance methodology.

The NEI methodology utilizes pre-existing ILRT-detectable leakage probabilities reflective of a 3-per-10 year ILRT frequency.

This 3-per-10 year frequency is utilized in this assessment and is termed the baseline case.

Since the latter half of the 1990's, the Dresden plant has been operating under a 1-per-10 year ILRT testing frequency consistent with the performance based Option B of 10 CFR Part 50, Appendix J [16].

1 -2 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval This 1-per-10 year frequency is referred to as the "current" frequency in this assessment.

The 1-per-15 year frequency is referred to as the "proposed" frequency in this assessment. The risk impacts of primary interest in this study are those associated with increasing the ILRT test interval from the current 1-per-10 year frequency to the 1-per-15 year frequency.

The NEI methodology directs that long term station blackout (SBO) scenarios not be included in pertinent EPRI category frequency calculations in the evaluation of the ILRT frequency extension assessment.

The Nuclear Regulatory Commission (NRC) has consistently asked licensees via Requests for Additional Information (RAls) to provide an assessment of the impact on risk results if long-term station blackout sequences were retained in selected EPRI categories. This ILRT extension assessment follows the NEI methodology pertaining to station blackouts, however, a sensitivity case is conducted to demonstrate that retaining SBO sequences does not change the conclusions of the overall assessment.

It should be noted that, in addition to ILRT tests, containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section Xl. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants.

Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E), require licensees to conduct visual inspections of the accessible areas of the interior of the containment 3 times every 10 years. These requirements will not be changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency. Type C tests are also not affected by the Type A test frequency change.

1-3 C467030603-5572-1 2/17103

Risk Impact Assessment of Extending Dresden ILRT Interval In response to previous ILRT extension request submittals, the NRC has consistently requested licensees to perform a quantitative assessment of the impact on LERF due to age-related degradation of non-inspectable areas of the containment.

Therefore, a quantitative assessment using the same approach used by other industry plants (e.g.,

Calvert Cliffs) is included as Appendix B to this ILRT extension evaluation.

This sensitivity case demonstrates that age-related degradation of non-inspectable areas of the containment does not change the conclusions of the overall assessment.

1.3 CRITERIA Based on previously approved ILRT extension requests, Dresden uses the following risk measures to characterize the change in risk associated with the one time ILRT extension:

Change in Large Early Release Frequency (LERF)

Change in conditional containment failure probability Change in population dose rate (person-rem/yr)

Consistent with the NEI Interim Guidance, the acceptance guidelines in Regulatory Guide 1.174 [4] are used to assess the acceptability of this one-time extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J.

RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in core damage frequency (CDF) less than 106 per reactor year and increases in large early release frequency (LERF) less than 10-7 per reactor year. Because the Type A test does not impact the at-power CDF(l), the relevant criterion is the change in LERF.

This approach is consistent with previous BWR ILRT submittals using the NEI Guidance and (1

It is noted that catastrophic containment failure would result in impacts on the mitigating systems, and therefore affect CDF. However, the containment failure sizes identified here that are prevented by ILRT are not sufficient to challenge the mitigating systems. Conversely, there is some probability that the leakage from containment could act as a beneficial influence by creating a self-relieving vent that 1-4 1-4 0~~~~~~~467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval approved by the NRC. RG 1.174 also discusses defense-in-depth and encourages the use of risk analysis techniques to show that key principles, such as the defense-in-depth philosophy, are met.

Therefore, the increase in the conditional containment failure probability, which helps to ensure that the defense-in-depth philosophy is maintained, will also be calculated. Figure 1-1 shows the acceptance guidelines for ALERF.

In addition, based on the precedent of other ILRT extension requests [6], the total annual risk (person-rem/yr population dose rate) and the conditional containment failure probability are examined to demonstrate the relative change in risk. (No threshold has been established for these parameter changes in RG 1.174.)

removes decay heat and prevents containment overpressure conditions. This latter 'benefit" is not credited with reducing the CDF in the Dresden ILRT assessment presented in this report.

1-5 C467030603-5572-1 2/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval

-J 10-6 REGION 11lE 1 0-7 10 10-5 104 LERF-*

Figure 2 Acceptance Guidelines* for Large Early Release Frequency (LERF)

• 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 decision-making, 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.

1-6 C467030603-5572-12/17/03

Risk ImpactAssessment ofExtending Dresden ILRTInterval Section 2 METHODOLOGY This section provides the following methodology related items:

A brief summary of available resource documents to support the methodology The NEI Interim Guidance for the analysis approach to be used The assumptions used in the evaluation The plant specific inputs required The following subsections address these items.

2.1 GENERAL RESOURCES AVAILABLE This section summarizes the general resources available for the ILRT internal extension risk assessment. Various industry studies on containment leakage risk assessment are briefly summarized here:

1)

NUREG/CR-3539 [10]

2)

NUREG/CR-4220 [11]

3)

NUREG-1273 [12]

4)

NUREG/CR-4330 [13]

5)

EPRI TR-1 05189 [8]

6)

NUREG-1493 [5]

7)

EPRI TR-104285 [2]

8)

NEI Interim Guidance [3, 21]

9)

NUREG-1150 [14] and NUREG/CR-4551 [9]

2-1 C467030603-5572-12/17/03

Risk Impact Assessment ofExtending Dresden ILRT Interval The first study is applicable because it provides one basis for the threshold that could be used in the Level 2 PSA for the size of containment leakage that is considered significant and to be included in the model. The second study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident.

The third study is applicable because it is a subsequent study to NUREG/CR-4220 that undertook a more extensive evaluation of the same database. The fourth study provides an assessment of the impact of different containment leakage rates on plant risk. The fifth study provides an assessment of the impact on shutdown risk from ILRT test interval extension. The sixth study is the NRC's cost-benefit analysis of various alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The seventh and eighth studies are EPRI studies of the impact of extending ILRT and LLRT test intervals on at-power public risk.

The ninth study provides consequence evaluations that can be used as surrogate results when plant specific characteristics are not available.

NUREG/CR-3539 [101 Oak Ridge National Laboratory (ORNL) documented a study of the impact of containment leak rates on public risk in NUREG/CR-3539. This study uses information from WASH-1400 [15] as the basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rates on LWR accident risks is relatively small.

NUREG/CR-4220 [111 NUREG/CR-4220 is a study performed by Pacific Northwest Laboratories (PNL) for the NRC in 1985. The study reviewed over two thousand LERs, ILRT reports and other related records to calculate the unavailability of containment due to leakage. The study calculated unavailabilities for Technical Specification leakages and "large" leakages.

2-2 C467030603-5572-12/17103

Risk Impact Assessment of Extending Dresden ILRTInterval NUREG/CR-4220 assessed the large" containment leak probability to be in the range of 1 E-3 to 1 E-2, with 5E-3 identified as the point estimate based on 4 events in 740 reactor years and conservatively assuming a one-year duration for each event.

NUREG-1273 [121 A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of the NUREG/CR-4220 database.

This assessment noted that about one-third of the reported events were leakages that were immediately detected and corrected.

In addition, this study noted that local leak rate tests can detect "essentially all potential degradations" of the containment isolation system.

NUREG/CR4330 [131 NUREG/CR-4330 is a study that examined the risk impacts associated with increasing the allowable containment leakage rates. The details of this report have no direct impact on the modeling approach of the ILRT test interval extension, as NUREG/CR-4330 focuses on leakage rate and the ILRT test interval extension study focuses on the frequency of testing intervals. However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 and other similar containment leakage risk studies:

"...the effect of containment leakage on overall accident risk is small since risk is dominated by accident sequences that result in failure or bypass of containment."

EPRI TR-105189 [81 The EPRI study TR-105189 is useful to the ILRT test interval extension risk assessment because this EPRI study provides insight regarding the impact of containment testing on shutdown risk. This study performed a quantitative evaluation 2-3 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval (using the EPRI ORAM software) for two reference plants (a BWRI4 and a PWR) of the impact of extending ILRT and LLRT test intervals on shutdown risk.

The result of the study concluded that a small but measurable safety benefit is realized from extending the test intervals. For the BWR, the benefit from extending the ILRT frequency from 3 per 10 years to 1 per 10 years was calculated to be a reduction of approximately 1 E-7/yr in the shutdown core damage frequency. This risk reduction is due to the following issues:

Reduced opportunity for draindown events Reduced time spent in configurations with impaired mitigating systems The study identified 7 shutdown incidents (out of 463 reviewed) that were caused by ILRT or LLRT activities. Two of the 7 incidents were RCS draindown events caused by ILRT/LLRT activities. The other 5 events involved loss of RHR and/or SDC due to ILRT/LLRT activities. This information was used in the EPRI study to estimate the safety benefit from reductions in testing frequencies. This represents a valuable insight into the improvement in the safety due to extending the ILRT test interval.

NUREG-1493 [51 NUREG-1493 is the NRC's cost-benefit analysis for proposed alternatives to reduce containment leakage testing intervals and/or relax allowable leakage rates. The NRC conclusions are consistent with other similar containment leakage risk studies:

Reduction in ILRT frequency from 3 per 10 years to 1 per 20 years results in an imperceptible" increase in risk.

Increasing containment leak rates several orders of magnitude over the design basis would minimally impact (0.2 - 1.0%) population risk.

Given the insensitivity of risk to the containment leak rate and the small fraction of leak paths detected solely by Type A testing, increasing the interval between integrated leak rate tests is possible with minimal impact on public risk.

2-4 C467030603-5572-1217103

Risk Impact Assessmnent of Extending Dresden ILRT Interval EPRI TR-1 04285 21 Extending the risk assessment impact beyond shutdown (the earlier EPRI TR-105189 study), the EPRI TR-104285 study is a quantitative evaluation of the impact of extending Integrated Leak Rate Test (ILRT) and (Local Leak Rate Test) LLRT test intervals on public risk due to at-power risk contributors. This study combined IPE Level 2 models with NUREG-1150 Level 3 population dose models to perform the analysis. The study also used the approach of NUREG-1493 in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

EPRI TR-1 04285 used a simplified Containment Event Tree to subdivide representative core damage sequences into eight (8) categories of containment response to a core damage accident:

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 Consistent with the other containment leakage risk assessment studies, this study concluded:

"These study results show that the proposed CLRT [containment leak rate tests] frequency changes would have a minimal safety impact. The change in risk determined by the analyses is small in both absolute and relative terms. For example, for the PWR analyzed, the change is about 0.02 person-rem per year.."

2-5 2-5

~~~~~~467030603-5572-1

Risk ImpactAssessment of Extending Dresden ILRTInterval NEI Interim Guidance [3, 211 NEI "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions of Containment Integrated Leakage Rate Test Surveillance Intervals"

[3] has been developed to provide utilities with revised guidance regarding licensing submittals. Additional information from NEI on the "Interim Guidance" was supplied in Reference [213.

A nine step process is defined which includes changes in the following areas of the previous EPRI guidance [2]:

Impact of extending surveillance intervals on dose Method used to calculate the frequencies of leakages detectable only by ILRTs Provisions for using NUREG-1 150 dose calculations to support the population dose determination.

This NEI Guidance is used in the Dresden ILRT analysis.

NUREG-1150 [141 and NUREG/CR 4551 [91 NUREG-1150 and the technical basis, NUREG/CR-4551, provide an ex-plant consequence analysis for a spectrum of accidents including a severe accident with the containment remaining intact (i.e., Technical Specification leakage).

The ex-plant consequences from NUREG-1 150 have been used as surrogate results by other BWRs for ILRT evaluation. Dresden has plant specific dose values available developed as part of the Dresden License Renewal Application.

2.2 NEI INTERIM GUIDANCE The Dresden risk assessment analysis uses the approach outlined in the NEI Interim Guidance. [3,21] The nine steps of the methodology are:

2-6 C467030603-5572-12117103

Risk Impact Assessment of Extending Dresden ILRTInterval

1.

Quantify the baseline (nominal three year ILRT interval) frequency per reactor year for the EPRI accident categories of interest. Note that EPRI categories 4, 5, and 6 are not affected by changes in ILRT test frequency.

2.

Determine the containment leakage rates for EPRI categories 1, 3a and 3b.

3.

Develop the baseline population dose (person-rem) for the applicable EPRI categories.

4.

Determine the population dose rate (person-rem/year) 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.

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 first seven steps of the methodology calculate the change in dose rate. The change in dose rate is the principal basis upon which the Type A ILRT interval extension was previously granted and is a reasonable basis for evaluating additional extensions. The eighth step in the interim methodology calculates the change in LERF and compares it to the guidelines in Regulatory Guide 1.174. Because there is no change in CDF, the change in LERF forms the quantitative basis for a risk informed decision per current NRC practice, namely Regulatory Guide 1.174. The ninth and final step of the NEI interim methodology calculates the change in containment failure probability. The NRC has previously accepted similar calculations (Ref. [7], referred to as conditional containment failure probability, CCFP) as the basis for showing that the proposed change is consistent 2-7 C4670306035572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval with the defense in depth philosophy. As such this last step suffices as the remaining basis for a risk informed decision per Regulatory Guide 1.174.

2.3 GROUND RULES The following ground rules are used in the analysis:

The Dresden Unit 2 Level 1 and Level 2 internal events PSA model provides representative results for the analysis. Due to the similarity of Units 2 and 3, the results of this Unit 2 assessment apply to Unit 3 as well.

It is appropriate to use the Dresden internal events PSA model as a gauge to effectively describe the risk change attributable to the ILRT extension. It is reasonable to assume that the impact from the ILRT extension (with respect to percent increases in population dose rate) from external events can be included in the calculations by sensitivity studies.

An evaluation of the risk impact of the ILRT on shutdown risk is addressed using the generic results from EPRI TR-105189 [8] as augmented by NEI Interim Guidance. [3, 21]

Radionuclide release categories are defined consistent with the EPRI TR-1 04285 methodology. [2]

The ex-plant consequence in terms of population dose results for the containment failures modeled in the PSA are based on plant specific values calculated as part of the Severe Accident Mitigation Alternatives (SAMA) evaluation for the Dresden License Renewal Application.

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.

Per the NEI Interim Guidance [3], the representative containment leakage for EPRI Category 1 sequences is 1 La (La is the Technical Specification maximum allowable containment leakage rate).

Per the NEI Interim Guidance [3], the representative containment leakage for EPRI Category 3a sequences is 10 La.

Per the NEI Interim Guidance [3], the representative containment leakage for EPRI Category 3b sequences is 35 La.

EPRI Category 3b is conservatively categorized as LERF based on the previously approved methodology [3] despite the fact that the 2-8 C467030603-5572-1 2117/03

Risk Impact Assessment of Extending Dresden ILRTInterval radionuclide release magnitude is significantly less than a large release.

The impact on population doses from Interfacing System LOCAs is not altered by the proposed ILRT extension.

The ISLOCA contribution to risk is accounted for in the EPRI methodology as a separate entry.

The containment isolation valve test frequency is not altered.

Therefore, the reduction in ILRT frequency does not impact the reliability of containment isolation valves to close in response to a containment isolation signal.

2.4 PLANT SPECIFIC INPUTS The inputs to the risk assessment include the following:

Past Dresden ILRT results to demonstrate the adequacy of the administrative and hardware issues.

Dresden source term and ex-plant consequence analysis (MACCS results) for population dose from the Severe Accident Mitigation Alternatives (SAMA) evaluation.

Dresden Unit 2 PSA Model 2002A, Rev. 0 (Level I & 2) 2.4.1 Prior Dresden ILRT Results The surveillance frequency for Type A testing in NEI 94-01 is at least once per ten years based on an acceptable performance history (i.e., two consecutive periodic Type A tests at least 24 months apart where the calculated performance leakage rate was less than 1.0 La) and consideration of the performance factors in NEI 94-01, Section 11.3. Based on the consecutive successful ILRTs performed in the 1990s, the current ILRT intervals for Dresden Units 2 and 3 are once per ten years. [16]

2.4.2 Ex-Plant Consequences As part of the License Renewal Application [19] for Dresden, ex-plant consequence assessments have been performed utilizing MAAP 4.0.4 to provide the source terms and 2-9 C4670306035572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval MACCS to provide the ex-plant population dose for the Dresden specific site. Table 2-1 reproduces the Dresden plant specific dose rate data from Table 4-4 of Appendix E of Reference 19.

The plant consequence analysis for Dresden is calculated for the 50-mile radial area.

The population surrounding the site was estimated for the year 2031 in the SAMA evaluation. [19] Population projections were determined using a geographic information system (GIS), U.S. Census block-group level population data for 2000 allocated to each sector based on the area fraction of the census block-groups in each sector, and population growth rates estimates for each county. The projected county growth rates were weighted by the fraction of each county in the 50-mile radius. The calculated growth rate of 1.408 from 2000 to 2031 was applied uniformly to all sectors. The total year 2031 population in the region was estimated as 9,967,934.

Table 2-2 reproduces the Dresden population estimate data from Table F-3 of Appendix E [19] and is provided for information. The population data of Table 2-2 is incorporated in the Dresden plant specific dose rate results of Table 2-1.

2.4.3 PSA Model Inputs The Dresden specific information used to perform this ILRT interval extension risk assessment includes the following:

Dresden Unit 2 Level 1 PSA Model 2002A, Rev. 0 Dresden Unit 2 Level 2 PSA Model 2002A, Rev. 0 2-10 C467030603-5572-1 2/17/03

Risk ImpactAssessment of Extending Dresden ILRT Interval Table 2-1 DRESDEN POPULATION DOSE RATE(')

MACCS Dominant MAAP Consequence Release Dose Frequency Dose Rate Run Category Categoryn2 (Person-Rem)

(/yr.)

(person-rem/yr.)

DR0024 L2-1 H/E 2.22E+07 3.01 E-07 6.682E+00 DR0040 L2-2 H/I 1.86E+07 1.48E-08 2.753E-01 DR0034 L24 M/E 1.21 E+07 1.09E-07 1.319E+00 DR0031 L2-5 M/l 5.44E+06 2.79E-07 1.518E+00 DR0028 L2-7 L or LUE 1.17E+07 3.29E-09 3.849E-02 DR0042 L2-8 L or LII or LUL 6.07E+06 5.78E-08 3.508E-01 DR0039 L2-9 Class V 2.79E+07 1.74E-09 4.855E-02 DR0043 L2-10 Intact 2.08E+03(3) 1.12E-06 2.330E-03 Frequency Weighted Totals 1.89E-06 10.23 (1) Data developed as part of SAMA evaluation for the Dresden License Renewal Application. Data assembled from Table 4-4 and Table 4-5 of Appendix E of Reference 19.

(2) These accident release category assignments are consistent with those made in the Dresden SAMA evaluation.

(3) Based on a 0.5%/day Containment Leak Rate.

2-11 2-11

~~~~~~~~~~~C467030603-557

Risk Impact Assessment of Extending Dresden ILRTInterval TABLE 2-2 ESTIMATED POPULATION DISTRIBUTION WITHIN A 50-MILE RADIUS, YEAR 2031 Sector 0-10 miles 10-20 miles 20-30 miles 30-40 miles 40-50 miles 50-mile total N

3,323 14,507 276,244 185,131 262,010 741,215 NNE 15,831 96,745 393,265 622,510 836,085 1,964,436 NE 11,110 166,697 139,921 886,086 2,996,261 4,200,075 ENE 3,146 54,327 236,184 800,958 783,427 1,878,042 E

3,398 9,246 38,013 177,215 288,290 516,162 ESE 893 1,753 17,321 14,360 22,838 57,165 SE 7,726 3,596 89,530 18,553 5,971 125,376 SSE 3,434 5,570 5,232 5,720 9,326 29,282 S

9,300 6,028 2,213 2,409 5,365 25,315 SSW 5,161 3,350 8,882 12,304 16,073 45,770 SW 2,392 2,421 1,962 4,274 6,849 17,898 WSW 2,981 3,468 2,756 29,516 5,498 44,219 W

14,208 4,295 24,645 19,927 49,364 112,439 WNW 2,065 2,385 9,784 6,034 14,900 35,168 NW 1,554 3,245 22,617 4,874 6,720 39,010 NNW 1,690 10,026 25,430 12,753 86,463 136,362 Total 88,202 387,649 1,293,989 2,802,614 5,395,430 9,967,934 2-12 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTIhteri'al 2.4.3.1 Dresden Unit 2 PSA The Dresden Unit 2 Level 1 and 2 PSA (2002A) used as input to this analysis represents the as-built, as-operated plant. The 2002A PSA model is the latest Dresden model('),

includes internal flooding, and includes detailed Level 2 sequences (i.e., both LERF and non-LERF contributors). The Dresden 2002A model is developed in CAFTA.

The total core damage frequency (CDF) as reported in the Dresden Level 2 Notebook is 1.89E-6/yr at a truncation of 1E-10/yr [18]. Table 2-3 summarizes the Dresden Level I PSA frequency results by core damage accident class.

Table 2-4 summarizes the Dresden Level 2 PSA results for containment failure.

The Dresden Level 2 PSA is used to calculate the release frequencies for the accidents evaluated in this assessment.

The Level 2 PSA is also developed in CAFTA.

Table 2-5 summarizes the pertinent Dresden Level 2 PSA results in terms of release categories. [18].

The total release frequency is 7.66E-7/yr, with a total CDF of 1.89E-6/yr. The No Release" frequency (i.e., containment leakage within Technical Specifications) for Dresden is 1.1 2E-6/yr. [18]

) Subsequent to the completion and documentation of the Dresden Level 2 PSA [18], minor changes were made to the Level I PSA such that the current Level 1 PSA model is identified as 2002A, Rev. 3. The changes associated with Revisions 1, 2, and 3 were reviewed and were judged to have a negligible impact to the ILRT Risk Assessment. The changes included minor modifications to a few basic event probabilities, correcting an omission in the mutually exclusive file, correcting file date references, and development of a simplified LERF model for use with on-line maintenance.

2-13 2-13

~~~~~~C467030603-5572-12

Risk Impact Assessment of Extending Dresden ILRT Interval 2.4.3.2 Dresden Unit 3 No substantive differences exist between the Dresden Unit 2 and Unit 3 that are judged to affect the conclusions of the PSA. As such, no separate PSA quantification is conducted for Unit 3. Since the Dresden PSA is judged applicable to both Unit 2 and Unit 3, the ILRT interval extension evaluation based upon the Dresden PSA is considered applicable to both Unit 2 and Unit 3.

2-14 C467030603-5572-1 2117103

Risk Impact Assessment of Extending Dresden ILRT Interval Table 2-3 CORE DAMAGE FREQUENCY CONTRIBUTIONS BY ACCIDENT CLASS [18](2)

Core Dama e

% of Contributing Accident Class Frequency")

CDF Transients Class IA/IE Transients - Core Melt with Vessel at High Pressure 1.06E-6 56.1%

Class IC ATWS with Loss of Injection 1.79E-8 0.9%

Class ID Transients - Core Melt with Vessel at Low Pressure 2.56E-8 1.4%

Class II Core Melts After Containment Failure Because of Loss 8.1 5E-8 4.3%

of DHR Capability SBO Class IBE Station Black Out - Early 3.OOE-7 15.9%

Class IBL Station Black Out - Late 1.08E-7 5.7%

LOCAs Class 3B Small or Medium LOCA - Core Melt with Vessel at 1.53E-8 0.8%

High Pressure Class 3C Medium or Large LOCA - Core Melt with Vessel at 8.20E-8 4.3%

Low Pressure Class 3D LOCA-Core Melt, RPV Breach, and Containment 1.18E-8 0.6%

Failure Nearly coincident due to Vapor Suppression Failure Class V Interfacing System LOCA 1.74E-9 0.1%

ATWS Class IV ATWS - Containment Fails Before Core Damage 1.86E-7 9.8%

Total 1.89E-6 100%

(I) All frequencies in events per reactor year.

(2)

Source: Table 6.6-2 of Volume 1 of the Dresden Level 2 PSA (DR-PSA-015) 2-15 C467030603-5572-12/17/03

Risk ImpactAssessment of ExtendingDresden ILRTInterval Table 2-4

SUMMARY

OF DRESDEN LEVEL 2 PSA RESULTS [18](2)

Core Damage Frequency End State (per year)(1)

Percent Containment Intact (Tech Spec leakage) 1.12E-6 59%

Containment Failure 7.66E-7 41%

(All other release categories)

Total 1.89E-6 I

100%

(1)

All frequencies in events per reactor year.

(2)

Source: Table 7.2-2 of Volume 1 of the Dresden Level 2 PSA (DR-PSA-015) 2-16 C467030603-5572-12/17/03

Risk Impact Assessment ofExtending Dresden ILRTInterval Table 2-5

SUMMARY

OF DRESDEN PSA LEVEL 2 RESULTS [18](2)

Release Category Frequencyz') (per year)

H/E - High Early (LERF) 3.03E-7 MIE - Medium Early 1.09E-7 UE - Low Early 3.29E-9 LUE - Low Low Early 0

H/I - High Intermediate 5.26E-9 M/l - Medium Intermediate 2.79E-7 LI - Low Intermediate 3.50E-8 LIJI - Low Low Intermediate 2.45E-9 H/L - High Late 9.49E-9 M/L - Medium Late 0

L/L - Low Late 1.68E-8 LUL - Low Low Late 3.27E-9 Total Release Frequency 7.66E-7 Core Damage Frequency 1.89E-6 (1) All frequencies in events per reactor year.

(2)

Source: Table 7.2-2 of Volume 1 of the Dresden Level 2 PSA (DR-PSA-015) 2-17 C467030603-5572-1 2/17103

Risk Impact Assessment of Extending Dresden ILRT Interval Section 3 ANALYSIS This section provides a step-by-step summary of the NEI guidance as applied to the Dresden ILRT interval extension risk assessment. Each subsection addresses a step or group of steps in the NEI guideline.

3.1 BASELINE ACCIDENT CATEGORY FREQUENCIES (STEP 1)

The first step of the NEI Interim Guidance is to quantify the baseline frequencies for each of the EPRI TR-104285 accident categories. This portion of the analysis is performed using the Dresden Level 1 and Level 2 PSA results. The results for each EPRI category are described below.

Tables 2-3, 2-4 and 2-5 of Section 2 compiled from the Dresden PSA [18] are used as the inputs to the accident frequency assessment.

Frequency of EPRI Categorv 1 This group consists of all core damage accident sequences in which the containment is initially isolated and remains intact throughout the accident (i.e., containment leakage at or below maximum allowable Technical Specification leakage). The ILRT methodology artificially divides this category among the Tech Spec leakage case (Category 1) and two other categories that are used to simulate possible changes due to reduced ILRT frequencies (i.e., Categories 3a and 3b; see below for their definition). Per NEI Interim Guidance, the frequency per year for this category is calculated by subtracting the frequencies of EPRI Categories 3a and 3b (see below) from the sum of all severe accident sequence frequencies in which the containment is initially isolated and remains intact (i.e., accident sequences classified as "OK" in the Dresden Level 2 PSA).

3-1 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval As discussed previously in Section 2.4, the frequency of the Dresden Level 2 PSA "OK" or No Release" accident bin is 1.12E-6/yr. As described below, the frequencies of EPRI Categories 3a and 3b are 4.59E-8/yr and 4.59E-9/yr, respectively. Therefore, the frequency of EPRI Category 1 is calculated as 1.12E-6/yr - 4.59E-8/yr - 4.59E-9/yr =

1.07E-6/yr for the assumed 3-year ILRT interval for which the ILRT leakage data has been collected.

Frequency of EPRI Category 2 This group consists of all core damage accident sequences in which the containment isolation system function fails during the accident progression (e.g., due to failures-to-close of large containment isolation valves initiated by support system failures, or random or common cause valve failures).

The frequency of this EPRI category is estimated by multiplying the conditional probability of containment isolation failure from the Dresden Level 2 PSA by the portion of the severe accident sequences (CDF) that would be challenged. The sequences that have containment isolation already failed are Class II, Class IIID, Class IV, and Class V.

Therefore, the EPRI Category 2 CDF does not include Dresden Level 1 Class II, Class IIID, Class IV, or Class V accident sequences. The following values are used for this calculation:

Containment Isolation System failure probability = 2.90E-3 [18] (1)

Total CDF = 1.89E-6/yr [18]

Class II sequences = 8.15E-8/yr [18]

Class IIID sequences = 1.18E-8/yr [18]

Class IV sequences= 1.86E-7/yr [18]

Class V sequences = 1.74E-9/yr [18]

1 Containment isolation system failure probability based on nodal quantification of event node S2 for loss of offsite AC or DC Division I and 11 (7.90E-3) minus the pre-existing containment failure probability basic event (5.OOE-3). Pre-existing containment failures are evaluated in other EPRI categories.

3-2 3-2

~~~~~~C467030603-5572-1 2/17/03

Risk Inpact Assessment of Extending Dresden ILRTInterval The frequency per year for this category is calculated as follows:

Frequency 2 = (containment isolation failure probability) x (CDF - CDF of Class II - CDF of Class IID - CDF of Class IV -

CDF of Class V)

Frequency 2 = (2.90E-3) X (1.89E-6/yr - 8.15E-8/yr - 1.1 8E-8/yr - 1.86E-7/yr -

1.74E-9/yr)

Frequency 2 = 4.67E-9/yr Note that pre-existing isolation failures are included in Category 6.

The frequency of EPRI Category 2 is 4.67E-9/yr.

Frequencv of EPRI Category 3a This group consists of all core damage accident sequences in which the containment is failed due to a pre-existing small" leak in the containment structure that would be identifiable only from an ILRT (and thus affected by ILRT testing frequency).

Consistent with NEI Interim Guidance [21], the frequency per year for this category is calculated as:

Frequency 3a

=

(3a conditional failure probability) x (CDF - CDF with independent LERF - CDF that cannot cause LERF)

The 3a conditional failure probability (2.7E-2) value is the conditional probability of having a pre-existing small" containment leak that is detectable only by ILRTs. This value is derived in Reference [3] and is based on data collected by NEI from 91 plants.

This value is also assumed reflective of ILRT testing frequencies of 3 tests in 10 years.

The pre-existing leakage probability is multiplied by the residual core damage frequency (CDF) determined as the total CDF minus the CDF for those individual sequences that either may already (independently) cause a LERF or could never cause a LERF due to the delay time of the release (i.e., non-early). As discussed previously in Section 2.4.2, 3-3 3-3

~~~~~~C467030603-5572-12

Risk Impact Assessment of Extending Dresden ILRTInterval the Dresden total core damage frequency is 1.89E-6/yr. Of this total CDF, the following core damage accidents involve either LERF directly (containment bypass) or will never result in LERF:

Long Term Station Blackout (SBO) scenarios (Class IBL) = 1.08E-7/yr [18]

Loss of Containment Heat Removal accidents (Class I): 8.15E-8/yr

[18]

Containment Bypass accidents (Class V): 1.74E-9/yr [18]

Therefore, the frequency of EPRI Category 3a is calculated as (2.70E-02) x (1.89E-6/yr -

1.08E-7/yr - 8.15E-8/yr - 1.74E-9/yr) = 4.59E-8/yr.

Frequencv of EPRI Category 3b This group consists of all core damage accident sequences in which the containment is failed due to a pre-existing large" leak in the containment structure that would be identifiable only from an ILRT (and thus affected by ILRT testing frequency). Similar to Category 3a, the frequency per year for this category is calculated as:

Frequency 3b

=

(3b conditional failure probability) x (CDF - CDF with independent LERF - CDF that cannot cause LERF)

The 3b failure probability (2.7E-3) value is the conditional probability of having a pre-existing large' containment leak that is detectable only by ILRTs. This value is derived in Reference [3] and is based on data collected by NEI from 91 plants. This value is also assumed reflective of ILRT testing frequencies of 3 tests in 10 years.

Therefore, similar to EPRI Category 3a, the frequency of Category 3b is calculated as (2.70E-03) x (1.89E-6/yr-1.08E-7/yr-8.15E-8/yr-1.74E-9/yr) = 4.59E-9/yr.

Frequency of EPRI Category 4 3-4 3-4

~~~~~~C467030603-5572-12

Risk Impact Assessment of Extending Dresden ILRTInterval This group consists of all core damage accident sequences in which the containment isolation function is failed due to a pre-existing failure-to-seal of Type B component(s) that would not be identifiable by an ILRT. Per NEI Interim Guidance, because this category of failures is only detected by Type B tests and not by the Type A ILRT, this group is not evaluated further in this analysis.

Frequency of EPRI Category 5 This group consists of all core damage accident sequences in which the containment isolation function is failed due to a pre-existing failure-to-seal of Type C component(s) that would not be identifiable by an ILRT. Per NEI Interim Guidance, because this category of failures is only detected by Type C tests and not by the Type A ILRT, this group is not evaluated further in this analysis.

Frequency of EPRI Category 6 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 valves that did not properly seal following test or maintenance activities) that would not be identifiable by containment leak rate tests.

Per NEI Interim Guidance, because this category of failures is not impacted by leak rate tests, this group is not evaluated further in this analysis.

3-5 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRT Interval Frequency of EPRI Category 7 This group consists of all core damage accident progression bins in which containment failure is induced by severe accident phenomena (e.g., overpressure).

Other severe accidents such as intact containment leakage and containment bypass are accounted for in other EPRI categories. Per NEI Interim Guidance, the frequency per year for this category is based on the plant PSA results.

For this analysis, the associated radionuclide releases are based on the plant specific MACCS results summarized earlier in Table 2-1.

EPRI Category 7 is divided into six sub-categories according to the MACCS consequence categories as shown in Table 3-1. MACCS consequence categories L2-9 and L2-10 are not included in the EPRI Category 7 frequency calculation since these two consequence categories reflect containment bypass and intact containment scenarios.

3-6 C467030603-5572-12/17103

Risk Impact Assessment of Extending Dresden ILRTInterval Table 3-1 EPRI CATEGORY 7 FREQUENCY EPRI MACCS MACCS Category Category Frequency Category 7a L2-1 3.01 E-7/yr Category 7b L2-2 1.48E-8/yr Category 7c L2-4 1.09E-7/yr Category 7d L2-5 2.79E-7/yr Category 7e L2-7 3.29E-9/yr Category 7f L2-8 5.78E-8/yr Total 7.65E-7/yr 3-7 C467030603-5572-12/17/03

Risk ImpactAssessment of Extending Dresden ILRTInterval Frequency of EPRI Cateqory 8 This group consists of all core damage accident progression bins in which the accident is initiated by a containment bypass scenario (i.e., Break Outside Containment LOCA or Interfacing Systems LOCA, ISLOCA). The frequency of Category 8 is the total frequency of the Dresden Level 1 PSA containment bypass scenarios (Class V).

Based on the Dresden Level 1 PSA results summarized earlier in Table 2-3, the frequency of Category 8 is 1.74E-9/yr.

Summary of Frequencies of EPRI Categories In summary, per the NEI Interim Guidance, the accident sequence frequencies that can lead to radionuclide releases to the public have been derived for accident categories defined in EPRI TR-104285.

The accident sequence frequency results by EPRI category are summarized in Table 3-2.

3-8 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval Table 3-2

SUMMARY

OF DRESDEN BASELINE RELEASE FREQUENCIES AS A FUNCTION OF EPRI CATEGORY EPRI ll l

Frequency Estimation l Frequency Category Category Description Methodology (1/yr) l No Containment Failure: Accident Per NEI Interim Guidance:

1.07E-6 sequences in which the containment I

remains intact and is initially isolated.

ITotal Dresden K"[releasen Only affected by I LRT leak testing category frequency- [Frequencyl frequency due to the incorporation of EPRI Categores 3a and 3b]

categories 3a and 3b.

1.12E-6/yr-4.59E-8/yr -

4.59E-9yr = 1.07E-6/yr l

2 Containment Isolation System Failure:

[Dresden containment isolation 4.67E-9 Accident sequences in which the failure probability] X [Total CDF -

containment isolation system function fails CDF of Class II - CDF of Class IIID -

during the accident progression (e.g., due CDF of Class IV - CDF of Class V]

to failures-to-close of large containment isolation valves initiated by support

[2.90E-3J X [1.89E-6/yr -

system failures, or random or common 8.15E-8/yr - 1.18E-8/yr -

cause failures). Not affected by ILRT leak

.86E-7yr-1.74E-9/yr] =

testing frequency.

4.67E-9yr 3a Small Pre-Existing Failures: Accident Per NEI Interim Guidance:

4.59E-8 sequences in which the containment is failed due to a pre-existing small leak in

[Dresden CDF for accidents not the containment structure or liner that involving containment would be identifiable only from an ILRT failure/bypass] x [2.7E-2]

(and thus affected by ILRT testing

[1.89E-6/yr-1.08E-7/yr-frequency).

8.15E-8/yr - 1.74E-9/yrJ x

[2.70E-02] = 4.59E-8/yr 3b Large Pre-Existing Failures: Accident Per NEI Interim Guidance:

4.59E-9 sequences in which the containment is failed due to a pre-existing large leak in

[Dresden CDF for accidents not the containment structure or liner that involving containment would be identifiable only from an ILRT failure/bypass]x[2.7E-3]

(and thus affected by ILRT testing

[1.89E-6/yr-1.08E-7Iyr-frequency).

8.15E-Byr - 1.74E-9/yrJ x

[2.70E-03] = 4.59E-9/yr 4

Tvpe B Failures: Accident sequences in Per NEI Interim Guidance:

N/A which the containment is failed due to a N/A pre-existing failure-to-seal of Type B components that would not be identifiable (not affected by ILRT frequency) from a ILRT (and thus not affected by ILRT testing frequency).

3-9 C467030603-5572-1 2/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval Table 3-2

SUMMARY

OF DRESDEN BASELINE RELEASE FREQUENCIES AS A FUNCTION OF EPRI CATEGORY EPRI l

Frequency Estimation Frequency Category ll Category Description Methodology (1/yr) 5 Tvpe C Failures: Accident sequences in Per NEI Interim Guidance:

NIA which the containment is failed due to a N/A pre-existing failure-to-seal of Type C components that would not be identifiable (not affected by ILRT frequency) from a ILRT (and thus not affected by ILRT testing frequency).

6 Other Containment Isolation System Per NEI Interim Guidance:

N/A Failure: Accident sequences in which the I

containment isolation system function fails N/A due to other" pre-existing failure modes (not affected by ILRT frequency) not identifiable by leak rate tests (e.g.,

pathways left open or valves that did not properly seal following test or maintenance activities). Not affected by ILRT leak testing frequency.

7 Containment Failure Due to Severe MACCS Consequence Category 7.65E-7 Accident:

frequencies. See Table 3-1 Vessel breach occurs and both the 7a 3.01 E-7/yr containment and the drywell have failed 7b 1.48E-8/yr either before or at the time of vessel I

breach. The containment sprays do not 7c 1.09E-7/yr operate before or at the time of vessel 7d 2.79E-7/yr breach.

7e 3.29E-9/yr 7f 5.78E-8/yr 7

7.65E-7/yr 8

Containment Bypass Accidents: Accident Total Dresden Containment 1.74E-9 sequences in which the containment is Bypass release frequency]

bypassed. Such accidents are initiated by LOCAs outside containment (i.e., Break Outside Containment LOCA, or Interfacing Systems LOCA). Not affected by ILRT leak testing frequency.

TOTAL:

1.89E-6 3-10 3-10

~~~~~~467030603-5572-1

Risk Impact Assessment of Extending Dresden ILRTInterval 3.2 CONTAINMENT LEAKAGE RATES (STEP 2)

The second step of the NEI Interim Guidance is to define the containment leakage rates for EPRI Categories 3a and 3b. As discussed earlier, EPRI Categories 3a and 3b are accidents with pre-existing containment leakage pathways ("small" and "large",

respectively) that would only be identifiable from an ILRT.

The NEI Interim Guidance recommends containment leakage rates of 10La and 35La for Categories 3a and 3b, respectively.

The NEI Interim Guidance describes these two recommended containment leakage rates as "conservative". These values are consistent with previous ILRT frequency extension submittal applications. La is the plant Technical Specification maximum allowable containment leak rate; for Dresden La is 1.6% of containment air weight per day (per Dresden Technical Specifications)(').

The NEI recommended values of 10La and 35La are used as is in this analysis to characterize the containment leakage rates for Categories 3a and 3b.

By definition, the containment leakage rate for Category (i.e., accidents with containment leakage at or below maximum allowable Technical Specification leakage) is 1.0La.

(1) Per Reference [20]. Exelon submitted a request to the NRC to Increase the maximum allowable containment leak rate, La. to 3.0%

for Dresden. This ILRT risk assessment utilizes a value of 0.5%. An increase In L. would have the following impacts on the three risk measures utilized In this analysis:

LERF:

Per the NEI methodology (see Section 3.4.4), Large Early Release Frequency (LERF) Is determined by the change In frequency of EPRI Category 3b. The frequency of Category 3b for 'arge' pre-existing containment leaks is independent of the containment leakage rate (see Section 3.1). Therefore, the proposed change to La will have no impact on the LERF results of this analysis.

CCFP:

Per the NEI methodology (see Section 3A.5), the Conditional Containment Failure Probability (CCFP) is determined by the change in freauencv of EPRI Category 1 accidents (containment Intact with maximum allowable containment leakage) and EPRI Category 3b accidents Csmall' pre-existing containment leaks). The frequency of Categories I and 3a are independent of containment leakage rate (see Section 3.1). Therefore the proposed change to La will have no impact on the CCFP results of this analysis.

Population Dose Rate: Per the NEI methodology (see Section 3.4). the population dose rate is a function of the maximum allowable containment leakage rate through incorporation of L In the MACCS plant specific evaluation. Table 3-7 of this analysis demonstrates that the total dose rate (10.3 person-rem/yr) is driven by EPRI Categories 7 2. and 8. which are effectively independent of La due to significant containment failure size in lieu of a leakage failure. The dose rates for EPRI Categories 1. 3a, and 3b. which are dependent on L. are negligible In comparison and remain negligible even if La was Increased to 3.0%/day. Therefore, the proposed change to La will have a negligible Impact on the population dose rate results of this analysis. See Section 3.5.3 for a sensitivity calculation.

3-11 C467030603-5572-1 2/17/03

Risk Impact Assessment ofExtending Dresden ILRTInterval 3.3 BASELINE POPULATION DOSE RATE ESTIMATES (STEPS 3-4)

The third and fourth steps of the NEI Interim Guidance are to estimate the baseline population dose (person-rem) for each EPRI category and to calculate the dose rate (person-rem/year) by multiplying the category frequencies by the estimated dose.

3.3.1 Population Dose Estimates (Step 3)

The NEI Interim Guidance recommends two options for calculating population dose for the EPRI categories:

Use of NUREG-1150 dose calculations Use of plant-specific dose calculations The NUREG-1150 [14] dose calculations were used in the EPRI TR-104285 study, as discussed previously in Section 2.1. The use of generic dose information for NUREG-1150 is recommended by NEI to make the ILRT risk assessment methodology more readily usable for plants that do not have a Level 3 PSA.

Dresden has recently completed the development of plant specific dose calculations as part of the License Renewal Application [19] process. These results, summarized earlier in Table 2-1, are therefore used in this ILRT risk assessment.

The plant specific dose results were developed using the MACCS2 code. The input parameters given with the MACCS2 "Sample Problem A," which included the NUREG-1150 food model (NRC 1989b) formed the basis for the analysis. These generic values were supplemented with parameters specific to the Dresden site and the surrounding area.

Site-specific data included population distribution, economic parameters, and agricultural production. Plant-specific release data included the time-nuclide distribution of releases, release frequencies, and release locations. The behavior of the population during a release (evacuation parameters) was based on plant and site-specific set points (i.e., declaration of a General Emergency) and the emergency planning zone (EPZ) evacuation table. These data were used in combination with site-specific meteorology to 3-12 C467030603-5572-12117/03

Risk Impact Assessment of Extending Dresden ILRTInterval simulate the probability distribution of impact risks (exposure and economic) to the surrounding (within 50 miles) population from the large early release accident sequences.

Population Dose by MACCS Category for Dresden Table 3-3 summarizes the Dresden population dose results at 50 miles for each MACCS category.

Population Dose by EPRI Categorv for Dresden Using the preceding information, the population dose as a function of EPRI category for the 50-mile radius surrounding Dresden is summarized in Table 3-5.

The following discussion provides the basis for the assignment of population dose for each EPRI category.

Note that all population doses are derived from the plant specific dose calculations.

The dose for EPRI category #1 (core damage accident with isolated and intact containment, i.e., no containment failure) is based on MACCS Consequence Category L2-1 0, the MACCS category most reflective of an intact containment.

The dose for EPRI Category 2 for core damage accidents with containment isolation failures is based on MACCS Consequence Category L2-1. This assignment is based on assuming that the containment isolation failure of EPRI Category 2 occurs in the drywell as an unscrubbed release. MACCS Consequence Category L2-1 results in the highest dose of all the Dresden "containment failure" categories (which is indicative of a containment failure with an unscrubbed release) other than category L2-9 for breaks outside containment.

No separate assignment of MACCS Consequence Categories is made for EPRI Categories 3a and 3b.

Instead, per the NEI Interim Guidance, the doses for EPRI 3-13 C467030603-5572-1 2/17103

Risk Impact Assessmnent of Extending Dresden ILRT Interval Categories #3a and #3b are taken as factors of 10 and 35, respectively, times the population dose of EPRI Category 1.

As EPRI Categories 4, 5, and 6 are not affected by ILRT frequency and not analyzed as part of this risk assessment (per NEI Interim Guidance), no assignment of MACCS results is made for these categories.

The dose for EPRI Category 7 is based on the development of a weighted average person-rem dose representative of the EPRI Category 7 subcategories 7a - 7f.

This weighted average approach is acceptable since the total frequency and dose associated with EPRI Category 7 does not change as part of the ILRT extension.

Table 3-4 summarizes the dose for subcategories 7a - 7f and the weighted average Category 7 dose.

The dose for the containment bypass category, EPRI Category 8, is based on MACCS Consequence Category L2-9.

Category L2-9 results in the highest dose of all the MACCS categories and reflects containment bypass scenarios.

3.3.2 Baseline Population Dose Rate Estimates (Step 4)

The baseline dose rates per EPRI accident category are calculated by multiplying the population dose estimates from Table 3-5 by the frequencies summarized in Table 3-2.

The resulting baseline population dose rates by EPRI category are summarized in Table 3-6.

As the conditional containment pre-existing leakage probabilities for EPRI Categories 3a and 3b are reflective of a 3-per-10 year ILRT frequency (refer to Section 3.1), the baseline results shown in Table 3-6 are indicative of a 3-per-10 year ILRT surveillance frequency.

3-14 C467030603-5572-12/17/03

Risk Impact Assessment ofExtending Dresden ILRT Interval Table 3-3 DRESDEN POPULATION DOSE (MACCS)(1 )

MACCS Frequency Population Dose Category (per year) J (Person-Rem)

L2-1 3.01 E-07 2.22E+07 L2-2 1.48E-08 1.86E+07 12-4 1.09E-07 1.21 E+07 L2-5 2.79E-07 5.44E+06 L2-7 3.29E-09 1.17E+07 L2-8 5.78E-08 6.07E+06 L2-9 1.74E-09 2.79E+07 L2-10 1.12E-06 2.08E+03 2)

Total 1.89E-06 This data is a summary of Table 2-1 presented earlier in the report.

Person-rem within 50 miles is estimated using a 0.5%/day Containment Leak Rate.

If the Alternate Source Term (AST) is adopted for Dresden, the containment technical specification leakage, population doses, and population dose rates for EPRI Categories 1, 3a, 3b are calculated to increase by a factor of six (6). Section 3.5.3 shows the impact of the higher containment leakage rate.

3-15 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval Table 3-4 DRESDEN EPRI CATEGORY 7 POPULATION DOSE RATE MACCS Release Dresden Population Population Dose Risk EPRI Category Frequency Dose (50 miles)

(50 miles)

(MACCS) per yea Person-Rem Person-Rem/yr(2 )

7a (L2-1) 3.01 E-07 2.22E+07 6.68E+00 7b (L2-2) 1.48E-08 1.86E+07 2.75E-01 7c (L2-4) 1.09E-07 1.21 E+07 1.32E+00 7d (L2-5) 2.79E-07 5.44E+06 1.52E+00 7e (L2-7) 3.29E-09 1.17E+07 3.85E-02 7f (L2-8) 5.78E-08 6.07E+06 3.51 E-01 Category 7 Total 7.65E-7 I

1.33E+7(3) 1.02E+1 Notes:

(1)

Table 3-3 (2)

Obtained by multiplying the release frequency (Column 2) by the population dose (Column 3)

(3)

Weighted average population dose for Category 7 obtained by dividing the total population dose risk (Column 4) by the total release frequency (Column 2).

3-16 C467030603-5572-12/1703

Risk Impact Assessment of Extending Dresden ILRT Interval Table 3-5 DRESDEN POPULATION DOSE ESTIMATES AS A FUNCTION OF EPRI CATEGORY FOR POPULATION WITHIN 50-MILE RADIUS Dresden EPRI Person-Rem Category Category Description Within 50 miles l

No Containment Failure(1) 2.08E+03(8 )

2 Containment Isolation System Failure(2) 2.22E+07 3a Small Pre-Existing Failures(3) 2.08E+04(8 )

3b Large Pre-Existing Failure(4) 7.28E+04()

4 Type B Failures (LLRT)(5) n/a 5

Type C Failures (LLRT)(5) n/a 6

Other Containment Isolation System Failure(5) n/a 7

Containment Failure Due to Severe Accident(6) 1.33E+07 8

Containment Bypass Accidents(7) 2.79E+07 Notes:

(1)

Based on Release Category L2-10 of Table 3-3 (2)

Based on Release Category L2-1 of Table 3-3 (3)

Factor of 10 times EPRI Category 1 (4)

Factor of 35 times EPRI Category 1 (5)

Not analyzed since not affected by ILRT frequency (6)

Weighted average of subcategories 7a-7f of Table 3-4 (7)

Based on Release Category L2-9 of Table 3-3 (8)

Person-rem within 50 miles is estimated using a 0.5%/day Containment Leak Rate.

If the Alternate Source Term (AST) is adopted for Dresden, the containment technical specification leakage, population doses, and population dose rates for EPRI Categories 1, 3a, 3b are calculated to increase by a factor of six (6). Section 3.5.3 shows the impact of the higher containment leakage rate.

3-17 C467030603-5572-12117/03

Risk Impact Assessment ofExtending Dresden ILRT Interval Table 3-6 DRESDEN DOSE RATE ESTIMATES AS A FUNCTION OF EPRI CATEGORY FOR POPULATION WITHIN 50 MILES (Base Line 3-per-10 year ILRT)

Person-Dose Rem Baseline Rate EPRI Within 50 Frequenc (erson-l Category Category Description miles(5)

(per y)

Rem/yr)l I

No Containment Failure(')

2.08E+03f 8) 1.07E-06 2.22E-03 2

Containment Isolation System Failure(2) 2.22E+07 4.67E-09 1.04E-01 3a Small Pre-Existing Failures 3) 2.08E+04(8) 4.59E-08 9.54E-04 3b Large Pre-Existing Failures(3) 7.28E+04(8) 4.59E-09 3.34E-04 4

Type B Failures (LLRT) n/a n/a n/a 5

Type C Failures (LLRT) n/a n/a n/a 6

Other Containment Isolation System Failure n/a n/a n/a 7

Containment Failure Due to Severe Accident(4) 1.33E+07 7.65E-07 1.02E+01 8

Containment Bypass Accidents(5) 2.79E+07 1.74E-09 4.85E-02 Total I

1.89E-06 1.03E+01_]

3-18 3-18

~~~~~~~C467030603-557

Risk Impact Assessment of Extending Dresden ILRTInterval Notes to Table 3-6 (1)

The release for this EPRI category is based on L2-1 0 from Table 3-3.

(2)

EPRI Category 2 (Containment Isolation failures) may include drywell isolation failures.

Therefore, the release associated with this category is based on the release associated with L2-1 from Table 3-3.

(3)

Dose estimates for categories 3a and 3b, per the NEI Interim Guidance, are calculated as lOxCategory 1 dose and 35xCategory 1 dose, respectively.

(4)

Dose estimate for category 7 is the weighted average of subcategories 7a-7f of Table 3-4.

(5)

EPRI Category 8 sequences involve containment bypass failures; as a result, the person-rem dose is not based on normal containment leakage. The releases for this category are assumed to result in a direct path to the environment, and as such, are based on L2-9 from Table 3-3.

(6)

Table 3-5.

(7)

Table 3-2.

(8)

Person-rem within 50 miles is estimated using a 0.5%/day Containment Leak Rate.

If the Alternate Source Term (AST) is adopted for Dresden, the containment technical specification leakage, population doses, and population dose rates for EPRI Categories 1, 3a, 3b are calculated to increase by a factor of six (6). Section 3.5.3 shows the impact of the higher containment leakage rate.

3-19 C467030603-5S72-1 2/17/03

Risk Impact Assessment ofExtending Dresden ILRT Interval 3.4 IMPACT OF PROPOSED ILRT INTERVAL (STEPS 5-9)

Steps 5 through 9 of the NEI Interim Guidance assess the impact on plant risk due to the new ILRT surveillance interval in the following ways:

Determine change in probability of detectable leakage (Step 5)

Determine population dose rate for new ILRT interval (Step 6)

Determine change in dose rate due to new ILRT interval (Step 7)

Determine change in LERF risk measure due to new ILRT interval (Step 8)

Determine change in CCFP due to new ILRT interval (Step 9) 3.4.1 Change in Probability of Detectable Leakage (Step 5)

Step 5 of the NEI Interim Guidance is the calculation of the change in probability of leakage detectable only by ILRT (and associated re-calculation of the frequencies of the impacted EPRI categories). Note that with increases in the ILRT surveillance interval, the size of the postulated leak path and the associated leakage rates are assumed not to change; however, the probability of pre-existing leakage detectable only by ILRT does increase.

Per the NEI Interim Guidance, 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. Using the standby failure rate statistical model, the average time that a pre-existing containment leak would exist undetected is one-half the surveillance interval.

For example, if the ILRT frequency is 1-per-10 years, then the average time that a leak would be undetected is 60 months (surveillance interval of 120 months divided by 2). The impact on the leakage probability due to the ILRT interval extension is then calculated by applying a multiplier determined by the ratio of the average times that leakage would go undetected for the two ILRT interval cases.

3-20 C4670306035572-12117/03

Risk Impact Assessment of Extending Dresden ILRTInterval As discussed earlier in Section 3.1, the conditional probability of a pre-existing ILRT-detectable containment leakage is divided into two categories.

The calculated pre-existing ILRT-detectable leakage probabilities are reflective of a 3-per-10 year ILRT frequency and are as follows:

'Small" pre-existing leakage (EPRI Category 3a): 2.70E-2

'Large" pre-existing leakage (EPRI Category 3b): 2.70E-3 Since 1996, the Dresden plant has been operating under a 1-per-10 year ILRT testing frequency consistent with the performance-based Option B of 10 CFR Part 50, Appendix J. [16] The baseline(') leakage probabilities first need to be adjusted to reflect the current 1 -per-10 year Dresden ILRT testing frequency, as follows:

Small": 2.70E-2 x [(120 months/2) / (36 months/2)] = 9.OOE-2 Large": 2.70E-3 x [(120 months/2) / (36 months/2)] = 9.00E-3 Note that a nominal 36 month interval (i.e., as opposed to 40 months, 120/3) is used in the above adjustment calculation to reflect the 3-per-10 year ILRT frequency. This is consistent with the NEI Interim Guidance.

Similarly, the pre-existing ILRT-detectable leakage probabilities for the 1-per-15 year ILRT frequency currently being pursued by Dresden (and the subject of this risk assessment) are calculated as follows:

Small": 9.OOE-2 x [(180 months/2) / (120 months/2)] = 1.35E-1 Large" : 9.00E-3 x [(180 months/2) / (120 months/2)] = 1.35E-2 Given the above adjusted leakage probabilities, the impacted frequencies of the EPRI categories are summarized below (refer to Table 3-2 for details regarding frequency calculations for the individual EPRI categories):

3-21 3-21

~~~~~~467030603-5572-1

Risk Impact Assessment of Extending Dresden ILRT Interval EPRI Category Frequency as a Function of ILRT Interval EPRI Baseline Current Proposed Category (3-per-10 year ILRT)

(1-per-10 year ILRT)

(1-per-15 year ILRT) 1 1.07E-6 9.52E-7 8.68E-7 3a 4.59E-8 1.53E-7 2.29E-7 3b 4.59E-9 I

.53E-8 2.29E-8 Note that, per the definition of the EPRI categories, only the frequencies of Categories 1, 3a, and 3b are impacted by changes in ILRT testing frequencies.

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

The dose rates per EPRI accident category as a function of ILRT interval are summarized in Table 3-7.

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

As can be seen from the dose rate results summarized in Table 3-7, the calculated total dose rate increase is imperceptible utilizing 3 significant digits. Using extended significant digits results in a change in population dose rate of 0.002 person-rem/yr from the current Dresden 10-year ILRT interval to the proposed 15-year interval.

Per the NEI Interim Guidance, the change in percentage contribution to total dose rate attributable to EPRI Categories 3a and 3b is also investigated here.

Using the results summarized in Table 3-7, for the current Dresden 1-per-10 year ILRT interval, the percentage contribution to total dose rate from Categories 3a and 3b is shown to be very minor:

[(3.18E-3 + 1.11E-3)/ 10.3] x 100 = 0.04%

(1) The baseline case uses data characteristic of the 3-per-10 year ILRT frequency of testing.

3-22 3-22 0~~~~~~~467030603-5572.12/17/03

Risk Impact Assessment of ExtendingDresden ILRTInterval For the proposed 1-per-15 year LRT interval, the percentage contribution to total dose rate from Categories 3a and 3b increases slightly but remains very minor:

[(4.77E-3 + 1.67E-3) / 10.3] x 100 = 0.06%

3-23 C467030603-5572-1 2/17/03

Risk Impact Assessment of Extending Dresden ILRT Interval Table 3-7 BASELINE DOSE RATE ESTIMATES BY EPRI ACCIDENT CATEGORY FOR POPULATION WITHIN 50 MILES Dose Rate as a Function of ILRT Interval (Person-RemIYr)

Baseline Current Proposed EPRI (3-per-10 (1 -per-10 (1 -per-15 Category Category Description year ILRT) year ILRT) year ILRT) l No Containment Failure 2.22E-3(1) 1.98E-3(1) 1.80E-3(')

2 Containment isolation System 1.04E-1 1.04E-1 1.04E-1 Failure 3a Small Pre-Existing Failures 9.54E-4(')

3.18E-3(')

4.77E-3()l 3b Large Pre-Existing Failures 3.34E-4(1) 1.11 E-3')

1.67E-3(')

4 Type B Failures (LLRT)

N/A N/A N/A 5

Type C Failures (LLRT)

N/A N/A N/A 6

Other Containment Isolation N/A N/A N/A System Failure 7

Containment Failure Due to 10.2 10.2 10.2 Severe Accident 8

Containment Bypass Accidents 4.85E-2 4.85E-2 4.85E-2 TOTAL 10.3 10.3 10.3 (1)

See Section 3.5.3 for discussion of the impact of higher Technical Specification containment leak rates.

3-24 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRT Interval 3.4.4 Change in LERF Due to New ILRT Interval (Step 8)

The risk increase associated with extending the ILRT interval involves the potential that a core damage event that normally would not result in a radionuclide release from an intact containment could in fact result in a release due to the increase in probability of failure to detect a pre-existing leak. Per the NEI Interim Guidance, only Category 3b sequences have the potential to result in large releases if a pre-existing leak were present. As such, the change in LERF (Large Early Release Frequency) is determined by the change in the frequency of Category 3b.

Category 1 accidents are not considered as potential large release pathways because the containment remains intact. Therefore, the containment leak rate is expected to be small.

Similarly, Category 3a is a usmall" pre-existing leak. Other accident categories such as 2, 6, 7, and 8 could result in large releases but these are not affected by the change in ILRT interval.

Late releases are excluded regardless of the size of the leak because late releases are, by definition, not LERF contributors.

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

delta LERF = (Frequency of EPRI Category 3b for 15 year ILRT interval) -

(Frequency of EPRI Category 3b for 10 year ILRT interval)

= 2.29E-8/yr-1.53E-8/yr

= 7.65E-9/yr This delta LERF of 7.65E-9/yr falls into Region ll, Very Small Change in Risk, of the acceptance guidelines in NRC Regulatory Guide 1.174. Therefore, increasing the ILRT interval at Dresden from the currently allowed 10 years to 15 years represents a very small change in risk, and is an acceptable plant change from a risk-informed perspective.

3-25 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRT Interval 3.4.5 Impact on Conditional Containment Failure Probability (Step 9)

Another parameter that the NRC Guidance in Reg. Guide 1.174 states can provide input into the decision-making process is the consideration of change in the conditional containment failure probability (CCFP). The change in CCFP is indicative of the effect of the ILRT on all radionuclide releases, not just LERF. The conditional containment failure probability (CCFP) can be calculated from the risk calculations performed in this analysis.

In this assessment, based on the NEI Interim Guidance, CCFP is defined such that containment failure includes all radionuclide release end states other than the intact state (EPRI Category 1) and small failures (EPRI Category 3a). The conditional part of the definition is conditional given a severe accident (i.e., core damage).

Consequently, the change in CCFP can be calculated by the following equation:

CCFP

= [1 - (Intact Containment Frequency / Total CDF)] x 100%, or

= [1 - ((#1 Frequency + #3a Frequency) / CDF)] x 100%

For the 1 0-year interval:

CCFPo = [1 - ((9.52E-7 + 1.53E-7) /1.89E-6)] x 100%

= 41.5%

For a 15-year interval:

CCFP15 = [1 - ((8.68E-7 + 2.29E-7) / 1.89E-6)] x 100%

= 41.9%

Therefore, the change in the conditional containment failure probability is:

A CCFP = CCFP15 - CCFPo = 0.4%

3-26 C467030603-5572-1 2/17103

Risk Impact Assessment of Extending Dresden ILRTInterval This change in CCFP of less than 1% is insignificant from a risk perspective.

3.5 SENSITIVITIES It is observed that the NRC, via Requests for Additional Information (RAls), has previously requested additional quantitative assessments regarding age related degradation of non-inspectable areas of the containment and assumptions regarding long term station blackout sequences.

In addition, the Dresden use of the Alternate Source Term methodology may result in an increase in the Containment Technical Specification leak rate to 3.0%/day.

This section summarizes the results of sensitivities performed to address these issues.

A sensitivity assessment of external event impacts is discussed in Section 5 of this report.

3.5.1 Containment Degradation Sensitivity Inspections of some reinforced and steel containments (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 containment.

In response to previous ILRT extension request submittals, the NRC has consistently requested licensees to perform a quantitative assessment of the impact on LERF due to age-related degradation of non-inspectable areas of the containment.

Therefore, a quantitative assessment using the same approach used by other industry plants (e.g., Calvert Cliffs) is performed as a sensitivity case to this ILRT extension evaluation. Appendix B provides the analysis details.

The results of the sensitivity case indicate that the increase in LERF from the 10-year ILRT interval to the 15-year ILRT interval is 9.06E-9/year, compared with 7.65E-9/yr without corrosion effects. This is still well below the Regulatory Guide 1.174 acceptance criterion threshold for very small" changes in risk of I.OE-7/yr. This confirms that the 3-27 C467030603-5572-12117/03

Risk Impact Assessment of Extending Dresden ILRTInterval proposed interval extension is acceptable from a

risk-informed perspective.

Additionally, the dose rate increase is negligible compared with the total of 10.3 person-rem/yr. The increase in the CCFP is determined to be insignificant (42.1% for the 15-year interval case versus 41.6% for the 1 0-year interval case). The results demonstrate that including corrosion effects in the ILRT assessment do not alter the conclusions from the original analysis.

3.5.2 Long Term Station Blackout Sensitivity The NRC has previously asked licensees, via RAls, to provide technical justification for the assumption that long term station blackout (SBO) scenarios do not contribute to LERF, and to provide an assessment of the impact on risk results if long-term station blackout sequences are retained in selected EPRI categories.

Appendix C provides the technical justification for not including SBO scenarios in the LERF assessment and also provides a sensitivity case to demonstrate that placing SBO sequences in the LERF category does not change the conclusions of the overall ILRT assessment.

The frequency of long term SBO core damage sequences in the Dresden PRA is 1.08E-7/yr. [18] The subject sensitivity case in Appendix C repeats the calculations of the ILRT assessment performed in Section 3, with the exception that the long term SBO sequences are retained in the EPRI Category 3a and 3b frequency calculations.

Retaining the SBO sequences in the EPRI Categories 3a and 3b frequency calculations results in a LERF increase of 8.13E-9/yr for the change from the current 10-year ILRT interval to the 15-year interval. This represents an additional LERF increase of 4.8E-10/yr (a 6% increase) over the best estimate ILRT increase in LERF of 7.65E-9/yr based on the full power internal events PRA. Including the long term SBO contribution, however, still results in a LERF increase below the NRC Regulatory Guide 1.174 Region IlIl criterion of 3-28 3-28

~~~~~~467030603-557212

Risk Impact Assessment of Extending Dresden ILRTInterval 1.OE-7/yr indicating a "very small" risk change. The increase in the population dose rate remains negligible, the same as in the baseline analysis. The increase in the conditional containment failure probability (CCFP) is determined to be unchanged (0.4% with SBO sequences included).

The sensitivity case demonstrates that even if long term SBO scenarios are included in the EPRI Category 3a and 3b frequencies, the conclusion of the risk assessment does not change; that is, the Dresden ILRT interval extension to 15 years has a minimal impact on plant risk.

3.5.3 Containment Technical Specification Leak Rate The implementation of the Alternate Source Term (AST) Methodology to Dresden may result in the change in the Technical Specification containment leakage rate to 3%/day.

This change in Technical Specification containment leakage rate will affect the dose rates for EPRI Categories 1, 3a, and 3b, which are all based on multiples of the containment leakage rate.

The estimates of dose rates in this report are based on a containment leak rate of 0.5%/day incorporated in the MACCS plant specific evaluation. The implementation of the AST would increase these dose rates by a factor of six (6).

The following tables summarize the change in population dose rate as a function of assumed containment leakage rate.

Population Dose Rate for 0.5%/day T.S. Leakage Rate ILRT Interval Category 15 year 10 year A

(Person-(Person-(Person-rem/yr) remlyr) remlyr) 3-29 C467030603-5572-12/17103

Risk Impact Assessment of Extending Dresden ILRTInterval Population Dose Rate for 0.5%/day T.S. Leakage Rate ILRT Interval Category 15 year 10 year A

(Person-(Person-(Person-remlyr) rem/yr) remlyr) 1 1.80E-3 1.98E-3

-0.18E-3 3a 4.77E-3 3.18E-3 1.59E-3 3b 1.67E-3 1.11E-3 0.56E-3 Total l 1.97E-3 Population Dose Rate for 3.0%/day T.S. Leakage Rate ILRT Interval Category 15 year 10 year A

(Person-(Person-(Person-rem/yr) rem/yr) rem/yr) 1 1.08E-2 1.19E-2

-1.1E-3 3a 2.86E-2 1.91 E-2 9.5E-3 3b 1.OOE-2 6.66E-3 3.34E-3 F

Total 1.17E-2 The dose rate change of 1.17E-2 person-rem/yr associated with a 3% T.S. leakage rate is negligible compared to the total population dose rate of 10.3 person-rem/yr.

In summary, per reference [20], Exelon submitted a request to the NRC to increase the maximum allowable containment leak rate, La, to 3.0%/day for Dresden. This ILRT risk assessment utilizes a value of 0.5%/day.

An increase in La would have the following impacts on the three risk measures utilized in this analysis:

LERF:

Per the NEI methodology (see Section 3.4.4), Large Early Release Frequency (LERF) is determined by the change in frequency of EPRI 3-30 C467030603-5572-1 2/17/03

Risk Impact Assessment of Extending Dresden ILRTIhterval Category 3b. The frequency of Category 3b for large" pre-existing containment leaks is independent of the containment leakage rate (see Section 3.1). Therefore, the proposed change to La will have no impact on the LERF results of this analysis.

CCFP: Per the NEI methodology (see Section 3.4.5), the Conditional Containment Failure Probability (CCFP) is determined by the change in frequency of EPRI Category 1 accidents (containment intact with maximum allowable containment leakage) and EPRI Category 3b accidents ("small" pre-existing containment leaks). The frequency of Categories 1 and 3a are independent of containment leakage rate (see Section 3.1). Therefore the proposed change to La will have no impact on the CCFP results of this analysis.

Population Dose Rate:

Per the NEI methodology (see Section 3.4), the population dose rate is a function of the maximum allowable containment leakage rate through incorporation of La in the MACCS plant specific evaluation. Table 3-7 of this analysis demonstrates that the total dose rate (10.3 person-rem/yr) is driven by EPRI Categories 7, 2, and 8, which are effectively independent of La due to significant containment failure size in lieu of a leakage failure.

The dose rates for EPRI Categories 1, 3a, and 3b, which are dependent on La, are negligible in comparison and remain negligible even if La was increased to 3.0%/day.

Therefore, the proposed change to La will have a negligible impact on the population dose rate results of this analysis.

Specifically, the total population dose rate of 10.3 person rem/year remains unchanged at 10.3 person rem/year as the Technical Specification containment leak rate varies by a factor of six (6).

3.5.4 Conclusions The sensitivity cases evaluated demonstrate:

LERF is not significantly impacted by the potential for containment leakage due to age-related degradation in non-inspectable areas.

3-31 3-31

~~~~~~467030603-5572-1

Risk Impact Assessment ofExtending Dresden ILRT Interval Inclusion of long-term station blackout scenarios in the EPRI Category 3a and 3b frequencies does not change the conclusion of this report.

The total population dose of 10.3 person rem/year remains unchanged at 10.3 person rem/year as the Technical Specification containment leak rate varies by a factor of six (6).

3-32 C467030603-5572-12/17103

Risk Impact.Assessment of Extending Dresden ILRTInterval Section 4 RESULTS

SUMMARY

The application of the approach based on NEI Interim Guidance [3, 21], EPRI-TR-104285

[2] and previous risk assessment submittals on this subject [6, 18, 20] have led to the quantitative results summarized in this section. The Dresden full power internal events Probabilistic Risk Assessment (PRA) is used in the quantification.

The results demonstrate a very small impact on risk associated with the one time extension of the ILRT test interval to 15 years.

The analysis performed examined Dresden specific accident sequences in which the containment remains intact or the containment is impaired. The accidents are analyzed and the results are displayed according to the eight (8) EPRI accident categories defined in Reference [2]:

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 The quantitative results are summarized in Table 4-1. The key results to this risk assessment are those for the 10-year interval (current Dresden condition) and the 15-year interval (proposed change).

The 3-per-10 year ILRT frequency is also presented for reference. It is a baseline starting point for this risk assessment given that the pre-existing containment leakage probabilities (estimated based on industry 4-1 C467030603-5572-12/17/03

Risk Impact Assessment ofExtending Dresden ILRT Interval experience - refer to Section 3.1) are reflective of the 3-per-10 year ILRT testing frequency.

The following is a brief summary of some of the key aspects of the ILRT test interval extension risk analysis:

Increasing the current 10-year ILRT interval to 15 years results in an negligible increase in total population dose rate.

The increase in the LERF risk measure is also insignificant, a 7.65E-9/yr increase. This LERF increase is categorized as a very small" increase per NRC Reg. Guide 1.174.

Likewise, the conditional containment failure probability (CCFP) increases by 0.4%. This is judged to be a negligible change in the CCFP.

4-2 C467030603-5572-12/17103

Risk Impact Assessment of Extending Dresden ILRT Interval Table 4-1 QUANTITATIVE RESULTS AS A FUNCTION OF ILRT INTERVAL Quantitative Results as a Function of ILRT Interval Baseline Current Proposed (3-per-10 rear ILRT)

(1-per-10 ear ILRT)

(1-per-15 year ILRT)

Dose Accident Population Accident Population Accident Population EPR (PeDos-em Accidentcy Dose Rate Accidenty Dose Rate Accidentcy Dose Rate EPRI (Person-Rem Frequency i (Person-RemNear Frequency (Person-Rem/Year Frequency (Person-RemNear Category

- Within 50 miles)

(per year)

Within 50 miles)

(per year)

Within 50 miles)

(per year)

Within 50 miles) 1 2.08E+3 1.07E-6 2.22E-3 9.52E-7 1.98E-3 8.68E-7 1.80E-3 2

2.22E+7 4.67E-9 1.04E-1 4.67E-9 1.04E-1 4.67E-9 1.04E-1 3a 2.08E+4 4.59E-8 9.54E-4 1.53E-7 3.18E-3 2.29E-7 4.77E-3 3b 7.28E+4 4.59E-9 3.34E-4 1.53E-8 1.11 E-3 2.29E-8 1.67E-3 4

N/A N/A N/A N/A N/A N/A N/A 5

N/A N/A N/A N/A N/A N/A N/A 6

N/A N/A N/A N/A N/A N/A N/A 7

1.33E+7 7.65E-7 1.02E+1 7.65E-7 1.02E+1 7.65E-7 1.02E+1 8

2.79E+7 1.74E-9 4.85E-2 1.74E-9 4.85E-2 1.74E-9 4.85E-2 TOTALS:

Increase in Dose Rate (1) 1.89E-6 10.3 1.89E-6 10.3 1.89E-6 10.3 Increase in LERF (2)

Increase in CCFP (%)(3) 4-3 4-3

~~~~~~~~~~~~~C467030603-5572-12117/03

Risk Inipact Assessment of Extending Dresden ILRTInterval Notes to Table 4-1:

(1)

The increase in dose rate (person-rem/year) is with respect to the results for the preceding ILRT interval, as presented in the table. For example, the increase in dose rate for the proposed 15-year ILRT is calculated as: total dose rate for 15-year ILRT, minus total dose rate for 10-year ILRT. For each case, the dose rate increase is insignificant.

(2)

The increase in Large Early Release Frequency (LERF) is with respect to the results for the preceding ILRT interval, as presented in the table. As discussed in Section 3.4.4 of the report, the change in LERF is determined by the change in the accident frequency of EPRI Category 3b. For example, the increase in LERF for the proposed 15-year ILRT interval over that for the 10-year ILRT interval is calculated as: the 3b frequency for the 15-year ILRT interval, 2.29E-8/yr, minus the 3b frequency for the 10-year ILRT interval, 1.53E-8/yr, which equals 7.65E-9/yr.

(3)

As discussed in Section 3.4.5, the conditional containment failure probability (CCFP) is calculated as:

CCFP =

[1 - ((Category #1 Frequency + Category #3a Frequency) / CDF)] x 100%

(4)

There is no calculated change in the total population dose rate unless more than four significant digits are utilized. See Section 3.5.3 for sensitivity calculation to establish this negligible change.

4-4 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval Section 5 CONCLUSIONS 5.1 QUANTITATIVE CONCLUSIONS The conclusions from the risk assessment of the one time ILRT extension can be characterized by the risk measures used in previously approved ILRT test interval extensions. These include:

Change in LERF Change in conditional containment failure probability Change in population dose rate 5.1.1 LERF Based on the results from Sections 3 and 4, the main conclusion regarding the impact on plant risk associated with extending the Type A ILRT test interval from ten years to fifteen years is:

Reg. Guide 1.174 [4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis.

Reg. Guide 1.174 defines very small changes in risk as resulting in increases of CDF below 106/yr and increases in LERF below 10-7/yr. 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 10 years to 15 years (using the change in the EPRI Category 3b frequency per the NEI Interim Guidance) is 7.65E-9/yr. Guidance in Reg. Guide 1.174 defines very small changes in LERF as below 107/yr. Therefore, increasing the Dresden ILRT interval from 10 to 15 years results in a very small change in risk, and is an acceptable plant change from a risk perspective.

5-1 C467030603-5572-12117/03

Risk Impact Assessment ofExtending Dresden ILRT Interval 5.1.2 CCFP The change in conditional containment failure probability (CCFP) is also calculated as an additional risk measure to demonstrate the impact on defense-in-depth. The ACCFP is found to be very small (0.4% increase) and represents a negligible change in the Dresden defense-in-depth.

5.1.3 Population Dose Rate The change in population dose rate is also reported consistent with previously approved ILRT interval extension requests. The change in population dose rate from the current 1-per-10 year ILRT frequency to 1-per-15 year frequency is insignificant. This is confirmed by a sensitivity case presented in Section 3.5.3.

5.2 RISK TRADE-OFF The performance of an ILRT introduces risk. An EPRI study of operating experience events associated with the performance of ILRTs has indicated that there are real risk impacts associated with the setup and performance of the ILRT during shutdown operation [8]. While these risks have not been quantified for Dresden, it is judged that there is a positive (yet unquantified) safety benefit associated with the avoidance of frequent ILRTs.

The safety benefits relate to the avoidance of plant conditions and alignments associated with the ILRT which place the plant in a less safe condition leading to events related to reactor drain down or loss of shutdown cooling.

Therefore, while the focus of this evaluation has been on the negative aspects, or increased risk, associated with the ILRT extension, there are, in fact, positive safety benefits associated with reducing the risk contribution from shutdown risk configurations.

5-2 C4670306035572-12117103

Risk Impact Assessment of Extending Dresden ILRT~hterval 5.3 EXTERNAL EVENTS IMPACT External hazards were evaluated in the Dresden Individual Plant Examination of External Events (IPEEE) Submittal in response to the NRC IPEEE Program (Generic Letter 88-20 Supplement 4). The IPEEE Program was a one-time review of external hazard risk to identify potential plant vulnerabilities and to understand severe accident risks. Although the external event hazards in the Dresden IPEEE were evaluated to varying levels of conservatism, the results of the Dresden IPEEE are nonetheless used in this risk assessment to provide a conservative comparison of the impact of external hazards on the conclusions of this ILRT interval extension risk assessment.

The proposed ILRT interval extension impacts plant risk in a limited way. Specifically, the probability of a pre-existing containment leak being the initial containment failure mode given a core damage accident is potentially higher when the ILRT interval is extended.

This impact is manifested in the plant risk profile in a similar manner for both internal events and external events.

The spectrum of external hazards has been evaluated in the Dresden IPEEE by screening methods with varying levels of conservatism. Therefore, it is not possible at this time to incorporate a realistic quantitative risk assessment of all external event hazards into the ILRT extension assessment. As a result, external events have been evaluated as a sensitivity case to show that the conclusions of this analysis would not be altered if external events were explicitly considered.

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 Dresden 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 will be 9.07E-8/yr. This delta LERF falls below the Region IlIl boundary of

<1E-7/yr and, therefore, is within the NRC RG 1.174 Region IlIl ("Very Small Changes in risk).

5-3 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval Therefore, incorporating external event accident sequence results into this analysis does not change the conclusion of this risk assessment (i.e., increasing the Dresden ILRT interval from 10 to 15 years is an acceptable plant change from a risk perspective).

5.4 PREVIOUS ASSESSMENTS The NRC in NUREG-1493 [5] has previously concluded that:

Reducing the frequency of Type A tests (ILRTs) from three per 10 years to one per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.

Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk.

The findings for Dresden confirm the above general findings on a plant specific basis when considering the following: (1) Dresden severe accident risk profile, (2) the Dresden containment failure modes, and (3) the local population surrounding the Dresden site.

5-4 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRT Interval Section 6 REFERENCES

[1]

Nuclear Energy Institute, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J. NEI 94-01, July 1995.

[2]

Electric Power Research Institute, Risk Impact 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]

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, Revision 1, November 2002.

[5]

Performance-Based Containment Leak-Test Program, NUREG-1493, September 1995.

[6]

Letter from R.J. Barrett (Entergy) to U.S. Nuclear Regulatory Commission, IPN 007, dated January 18, 2001.

[7]

United States Nuclear Regulatory Commission, Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MB01 78), April 17, 2001.

[8]

ERIN Engineering and Research, Shutdown Risk Impact Assessment for Extended Containment Leakage Testing Intervals Utilizing ORAM"M EPRI TR-105189, Final Report, May 1995.

[9]

Sandia National Laboratories, Evaluation of Severe Accident Risks: Peach Bottom, Unit 2, Main Report NUREG/CR-4551, SAND86-1309, Volume 4, Revision 1, Part 1, December 1990.

[10]

Oak Ride National Laboratory, Impact of Containment Building Leakage on LWR Accident Risk, NUREG/CR-3539, ORNL/TM-8964, April 1984.

[11]

Pacific Northwest Laboratory, Reliability Analysis of Containment Isolation Systems, NUREG/CR-4220, PNL-5432, June 1985.

6-1 6-1

~~~~~~467030603-5572-1 2/17103

Risk Impact Assessment ofExtending Dresden ILRTInterval

[12]

U.S. Nuclear Regulatory Commission, Technical Findings and Regulatory Analysis for Generic Safety Issue I.E.4.3 'Containment Integrity Check', NUREG-1273, April 1988.

[13]

Pacific Northwest Laboratory, Review of Light Water Reactor Regulatory Requirements, NUREG/CR-4330, PNL-5809, Vol. 2, June 1986.

[14]

U.S. Nuclear Regulatory Commission, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, NUREG -1150, December 1990.

[15]

U.S. Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October 1975.

[16]

NRC letter to ComEd dated January 11, 1996 Issuing Technical Specification Amendment for Dresden Station to implement the requirements of 10 CFR 50, Appendix J, Option B for performance-based primary reactor containment leakage testing.

[17]

ERIN Engineering and Research, Inc., Identification of Risk Implications Due to 17% Power Uprate at Dresden, Doc. #C1340019-4309, December 2000.

[18]

"Dresden Detailed Level 2 Evaluation", DR PSA-015, Volumes 1 and 2 dated July, 2002.

[19]

Dresden License Renewal Application, forwarded by letter from Jeffrey Benjamin (Exelon) to NRC, dated January 3, 2003.

[20]

Letter from Keith R. Jury (Exelon) to the NRC, RS-02-174, dated October 10, 2002.

[21]

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

6-2 C467030603-5572-12117103

APPENDIX A EXTERNAL EVENTASSESSMENT

Risk Impact Assessment of Extending Dresden ILRTInterval Appendix A EXTERNAL EVENT ASSESSMENT This appendix discusses the external events assessment in support of the Dresden ILRT interval extension risk assessment.

External hazards were evaluated in the Dresden Individual Plant Examination of External Events (IPEEE) Submittal in response to the NRC IPEEE Program (Generic Letter 88-20 Supplement 4). The IPEEE Program was a one-time review of external hazard risk to identify potential plant vulnerabilities and to understand severe accident risks. Although the external event hazards in the Dresden IPEEE were evaluated to varying levels of conservatism, the results of the Dresden IPEEE are nonetheless used in this risk assessment to provide a conservative comparison of the impact of external hazards on the conclusions of this ILRT interval extension risk assessment.

A.1 DRESDEN IPEEE INTERNAL FIRES ANALYSIS The Dresden plant risk due to internal fires was updated in 1999 as part of the revised Dresden Individual Plant Examination of External Events (IPEEE) Submittal [A-7]. The EPRI FIVE Methodology [A-9] and Fire PRA Implementation Guide (FPRAIG) [A-10]

screening approaches and data were used to perform the study.

The Dresden Unit 3(1) CDF contribution due to internal fires in the unscreened fire areas was calculated at 2.97E-5/yr.

The breakdown of the Dresden Unit 3 fire risk profile provided in Figure 2-8 of Reference A-8 is as follows:

(1) Unit 3 value of 2.97E-5/yr is chosen because it is higher than the Unit 2 value of 1.69E-5Iyr. Differences between the Unit 2 and Unit 3 fire CDFs are primarily due to cable routing paths. Reference A-8 presents a more complete discussion of the Unit 2 and Unit 3 fire CDF differences.

A-i C467030603.5572-12/1 7103 A-1 C467030603-5572-12/17/03

Risk Impact Assessment ofExtending Dresden ILRTInterval Fire-induced loss of decay heat removal scenarios 44.2%

Fire-induced loss of inventory control scenarios 22.2%

(RPV at low pressure)

Fire-induced loss of inventory control scenarios 9.9%

(RPV at high pressure)

Other fire-induced scenarios (severe fires, etc.)

23.7%

This information is used in Section A.4 of this appendix to provide insight into the impact of external hazard risk on the conclusions of this ILRT risk assessment.

A.2 DRESDEN IPEEE SEISMIC ANALYSIS The Dresden seismic risk analysis was performed as part of the Individual Plant Examination of External Events (IPEEE).

Dresden performed a seismic margins assessment (SMA) following the guidance of NUREG-1407 and EPRI NP-6041.

The SMA is a deterministic evaluation process that does not calculate risk on a probabilistic basis.

No core damage frequency sequences were quantified as part of the IPEEE seismic risk evaluation.

Although probabilistic risk information is not directly available from the Dresden SMA IPEEE analysis, Reference [A-1] provides a method (called the Simplified Hybrid Method) for obtaining a seismic-induced CDF estimate based on results of an SMA analysis. Reference [A-1] has shown that only the plant HCLPF (High Confidence Low Probability of Failure) seismic capacity is needed in order to estimate the seismic CDF within a precision of approximately a factor of two. The approach is as follows:

Step 1: Determine the plant HCLPF seismic capacity CHCLPF from the SMA analysis A-2 C467030603-5572-12/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval SteD 2: Estimate the 10% conditional probability of failure capacity Co%

from:

Cl0% = F CCLPF Fa = e 10446 where 1.044 is the difference between the 10% NEP standard normal variable (-1.282) and the 1% NEP standardized normal variable (-2.326).

Experience gained from high quality seismic PSA studies indicates that the plant damage state fragility determined by rigorous convolution will tend to have 0c values in the range of 0.30 to 0.35 (the plant damage state Pc value is equal to or less than the c

values for the fragilities of the individual components that dominate the seismic risk).

As such, the Simplified Hybrid method recommends:

C1o% = 1. 4CHCLPF Step 3:

Determine hazard exceedance frequency H1 0% that corresponds to C10% from hazard curve.

Step 4:

Determine seismic risk PF (e.g., CDF) from:

PF = 0.5 H1o%

Using the Simplified Hybrid Method, an approximation of the Dresden seismic-induced CDF is performed here.

Step 1: If the SMA analysis screens out every component on the Seismic Safe Shutdown Paths at the Review Level Earthquake (RLE), the plant HCLPF is equal to the RLE. Otherwise, the plant HCLPF is determined by the lowest seismic capacity component in the seismic safe shutdown paths. Dresden falls into the latter category. The Dresden RLE specified in the NRC IPEEE program is 0.30g PGA. A number of equipment items were identified during the Dresden A-46 and seismic IPEEE analyses to have HCLPF capacities less than the 0.30g PGA RLE(1). These items were addressed either by plant improvements or other resolutions. Based on these findings and resolutions,

() Note that the finding of HCLPFs lower than the plant RLE is not an indication of any vulnerability, but is consistent with the IPEEE definitions of RLEs. The IPEEE RLEs were defined such that high capacity items would be screened and the lower capacity items would be identified.

A-3 C467030603-5572-12/17/03

Risk Impact Assessment ofExtending Dresden ILRTInterval the Dresden plant HCLPF was ultimately estimated to be at least 0.20g PGA.

[A-5]

Step 2:

Using the relationship recommended above, the plant 10% capacity point (C10%) is estimated as 1.4 x 0.20g PGA = 0.28g PGA.

Step 3: The seismic hazard curve for the Dresden site, based upon EPRI NP-6395-D, is summarized in tabular form in Table A-1. As can be seen from Table A-1, the seismic hazard frequency associated with the 10% capacity point (0.28g PGA) is approximately 8.5E-6/yr.

Step 4:

Using the relationship recommended above, the seismic-induced CDF is approximated as 0.50 x 8.5E-6/yr = 4.3E-6/yr.

The Simplified Hybrid Method only provides an overall seismic-induced CDF estimate and does not provide information as to the breakdown of seismic accident sequence types. A more rigorous analysis (e.g., a seismic PSA, or the Rigorous Hybrid Method referred to in Reference [A-1]) is required for such information. Such an analysis was not performed as part of this ILRT risk assessment.

However, a Rigorous Hybrid Method calculation was recently completed for another Exelon BWR plant (Limerick)

[A-2]. The results of that study (Case #2 of Reference [A-2]) are used here to provide a reasonable approximation of the relative risk profile due to seismic-induced accident sequences. They are as follows:

Seismic-induced loss of decay heat removal scenarios

-35%

Wide-spread failure of seismic safe shutdown SSCs

-20%

Seismic-induced ATWS scenarios

-15%

Other seismic-induced accidents (e.g., SBO, loss of coolant

-30%

makeup, etc.)

This information is used in Section A.4 of this appendix to provide quantitative insights into the impact of external hazard risk on the conclusions of this ILRT risk assessment.

AX4 A-4

~~~~~~46703D603-5572-1

Risk Impact Assessment ofExtending Dresden ILRTInterval A.3 OTHER EXTERNAL HAZARDS In addition to internal fires and seismic events, the Dresden IPEEE Submittal analyzed a variety of other external hazards:

High Winds/Tornadoes External Flooding Transportation and Nearby Facility Accidents Other External Hazards The Dresden IPEEE analysis of high winds, tornadoes, external floods, transportation accidents, nearby facility accidents, and other external hazards was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards.

Based upon this review, it was concluded that Dresden meets the applicable Standard Review Plan requirements and therefore has an acceptably low risk with respect to these hazards. As such, these hazards were determined in the Dresden IPEEE to be negligible contributors to overall plant risk.

Accordingly, these other external event hazards are not included explicitly in this appendix and are reasonably assumed not to impact the results or conclusions of the ILRT interval extension risk assessment.

A-5 C467030603-5572-12117/03

Risk Impact.Assessment of Extending Dresden ILRTInterval Table A-1 DRESDEN SITE SEISMIC HAZARD CURVE

- EPRI NP-6395-D(1 )

Peak Ground Acceleration EPRI Exceedance cm/s2 9

Frequency (1/yr, mean) 5 0.01 5.9E-3 50 0.05 2.8E-4 100 0.10 7.8E-5 250 0.25 9.5E-6 280 0.28 8.5E-6(2) 500 0.51 1.1 E-6 700 0.71 3.2E-7 1000 1.02 7.7E-8 (1) From Table 3-27 and Figure 3-79 of EPRI NP-6395-D, Appendix E.

(2)

Frequency for Dresden 10% plant capacity interpolated from EPRI data points.

A-6 C467030603-5572-1 2/17/03

Risk Impact Assessment of Extending Dresden ILRTInterval A.4 IMPACT OF EXTERNAL HAZARD RISK ON LERF The NEI Interim Guidance calculation of delta LERF performed in Section 3 of this report is re-performed here including, in addition to internal event information, the Dresden IPEEE external event risk information discussed in the previous sections.

Per the NEI Interim Guidance, the impact on the LERF risk measure due to the proposed ILRT interval extension is calculated as follows:

delta LERF = (Frequency of EPRI Category 3b for 1-per-15 year ILRT interval) -

(Frequency of EPRI Category 3b for 1-per-10 year ILRT interval)

As discussed in Section 3.1, the frequency per year for EPRI Category 3b is calculated as:

Frequency 3b = [3b conditional failure probability] x [CDF - (CDF with independent LERF + CDF that cannot cause LERF)]

Based on the previous discussion in Sections A.1 through A.3, the Dresden external event initiated CDF is approximately 2.97E-5/yr (internal fires) + 4.3E-6/yr (seismic) =

3.40E-5/yr. Of these external events, the following external event accident scenarios are excluded from the LERF (Category 3b) frequency calculation because they cannot result in a LERF release or they independently result in LERF:

Fire-induced loss of decay heat removal scenarios:

0.442 x 2.97E-5/yr = 1.31 E-5/yr Seismic-induced loss of decay heat removal scenarios:

0.35 x 4.3E-6/yr = 1.51E-6/yr Wide-spread failure of seismic safe shutdown SSCs:

0.20 x 4.3E-6/yr = 8.6E-7/yr A-7 A-7

~~~~~~467030603-5572-1

Risk ImpactAssessment of Extending Dresden ILRTInterval The loss of decay heat removal sequences are extremely long term events in which the RPV inventory is maintained for more than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

The containment conditions continue to degrade and ultimately result in the catastrophic failure of containment. The long time available for planning and implementing an evacuation is considered sufficient to preclude its inclusion as a LERF contributor consistent with the RG 1.174 definition.

As far as the widespread failure of seismic safe shutdown SSCs, this type of seismic event is considered to be a LERF regardless of the ILRT interval. Therefore, by the definition of Category 3b in the NEI methodology, there is no Category 3b contribution from these events that is affected by the ILRT interval.

Therefore, the baseline (3-per-10 year) frequency of category 3b due to external events is calculated as (2.70E-03) x [(3.40E-5/yr) - (1.31 E-5/yr + 1.51 E-6/yr + 8.6E-7/yr)] = 5.OOE-8/yr.

Using the relationship described in Section 3.4.1 for the impact on 3b frequency due to increases in the ILRT surveillance interval, the EPRI Category 3b frequency for the 10-year and 15-year ILRT intervals are calculated as 1.67E-7/yr and 2.50E-7/yr, respectively. Therefore, the change in the LERF risk measure due to extending the ILRT interval from 10 years to 15 years, including both internal and external hazard risk, is estimated as:

3b Frequency 3b Frequency (1-per-10 year ILRT)

(1-per-15 year ILRT)

LERF Increase External Events Contribution 1.67E-7/yr 2.50E-7/yr 8.30E-8/yr Internal Events Contribution 1.53E-8/yr 2.29E-8/yr 7.65E-9/yr Combined (Internal + External) 1.82E-7/yr 2.73E-7/yr 9.07E-8/yr Thus, the increase in LERF due to the external events contribution is estimated as 8.30E-8/yr. When the internal and external events contributions are summed, the total A LERF A-8 A-8

~~~~~~467030603-5572-1

Risk ImpactAssessment of Extending Dresden ILRTInterval can be obtained. The summed A LERF remains within Region 1II, i.e., the very small risk change region of RG 1.174.

A.5 COMPARISON TO RG 1.174 ACCEPTANCE GUIDELINES NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis", provides NRC recommendations for using risk information in support of applications requesting changes to the license basis of the plant. As discussed in Section 2 of this report, the risk acceptance criteria of RG 1.174 is used here to assess the ILRT interval extension.

The 9.07E-8/yr increase in LERF from reducing the Dresden ILRT frequency from 1-per-10 years to 1-per-15 years falls into Region IlIl ("Very Small Change" in risk) of the RG 1.174 acceptance guidelines. Per RG 1.174, when the calculated increase in LERF due to the proposed plant change is in the range of IE-7 to E-6 per reactor year (Region II, "Small Change" in risk), the risk assessment must also reasonably show that the total LERF from all hazards is less than 1 E-5/yr. Although not required in this case (since the delta LERF is less than 1E-7 and falls in Region ll), the total LERF from all hazards is calculated in this analysis for completeness.

Per the Dresden internal events Level 2 PRA (2002A), the Dresden LERF due to internal event accidents including internal flooding is 3.03E-7/yr [A-6]. The LERF due to external events is estimated here using the Dresden conditional LERF probabilities as a function of core damage accident type. As can be seen from Table A-2, the external events LERF is estimated at 2.94E-6/yr. Therefore, the total LERF for Dresden is estimated at 3.03E-7/yr + 2.94E-6/yr = 3.24E-6/yr, which is less than the RG 1.174 acceptance guideline of 1 E-5/yr for Region II ("Small Change" in risk).

A-9 A-9

~~~~~~~C467030603-557

Risk Impact Assessment of Extending Dresden ILRTInterval Table A-2 ESTIMATE OF DRESDEN LERF DUE TO EXTERNAL EVENTS Conditional LERF External Event Accident Type CDF Probability(:)

LERF Loss of decay heat removal 1.46E-5(4) 0 0

Seismic-induced ATWS 6.45E-7(5 )

5.06E-1 3.26E-7 Wide spread failure of seismic safe 8.60E-7 (6) 1.00P) 8.60E-7 shutdown SSCs Other scenarios (e.g., loss of coolant 1.79E-5) 9.79E-2(3) 1.75E-6 makeup, SBO, severe fire, etc.)

Totals 3.40E-5/yr N/A 2.94E-6/yr Notes:

(1) LERF conditional probabilities as a function of core damage accident type are taken from the Dresden Level 2 PRA (2002A) [A-6, Table 6.6-2]

(2)

The LERF conditional probability for ISLOCA sequences (1.00) used to model LERF for seismic accidents involving widespread failure of safe shutdown SSCs. This is judged reasonable.

(3)

The LERF conditional probability for Class IA accidents (loss of coolant makeup with RPV at high pressure) is used to model the other miscellaneous sequence types. This is judged conservative.

(4)

Sum of fire-induced (1.31 E-5/yr) and seismic induced (1.51 E-6/yr) loss of decay heat removal scenarios (5) Seismic induced CDF (4.3E-6/yr)

• 0.15 (6)

Seismic induced CDF (4.3E-6/yr)

• 0.20

7) Seismic induced CDF (4.3E-6/yr)

• 0.30 + fire-induced CDF (2.97E-5/yr) * (0.222 +

0.099 + 0.237)

A-10 C467030603-5572-12117/03

Risk Impact Assessment of Extending Dresden ILRTInterval REFERENCES

[A-1]

Kennedy, R.P., Overview of Methods for Seismic PSA and Margin Analysis Including Recent Innovations", Proceedings of the OECD-NEA Workshop on Seismic Risk, Tokyo, Japan, August 1999. Available from OECD Nuclear Energy Agency, La Seine St.-Germain, 12 Boulevard des lies, 92130 Issy-les-Moulineaux, France.

[A-2]

ERIN Engineering and Structural Mechanics Consulting, Summary of Limerick Generating Station Seismic Margins Insights Evaluation, ERIN Report No.

C0467010033-4801, June 2002.

[A-3]

Electric Power Research Institute, Probabilistic Seismic Hazard Evaluations at Nuclear Plant Sites in the Central and Eastern United States: Resolution of the Charleston Earthquake Issue, NP-6395-D, April 1989.

[A-4]

Identification of Risk Implications due to 17% Power Uprate at Dresden, ERIN Report, Decembere 2000.

[A-5]

Dresden Letter from Preston Swafford to USNRC providing responses to NRC RAls on the Dresden IPEEE, March 30, 2000.

[A-6]

Dresden Detailed Level 2 PRA Evaluation, DRESDEN-PSA-015, Volume 1, June 2002.

[A-7]

Dresden IPEEE Submittal Report, Rev. 1, March 2000.

[A-8]

ERIN Engineering and Research, Dresden Fire IPEEE: Insights and Sensitivities, ERIN Report No. R134-99-02.R08, Rev. 1, April 2000.

[A-9] Professional Loss Control, Inc., Fire-induced Vulnerability Evaluation (FIVE)

Methodology Plant Screening Guide, EPRI TR-1 00370, Electric Power Research Institute, April 1992.

[A-10] W.J. Parkinson, et. al., Fire PRA Implementation Guide, EPRI TR-105928, Electric Power Research Institute, December 1995.

A-11 A-Il

~~~~~~46703603-5572-12

APPENDIX B CONTAINMENT DEGRADA TION SENSITIVITY

Sensitivity Calculation for Dresden ILRTRisk Assessment Appendix B CONTAINMENT CORROSION SENSITIVITY B.1 BACKGROUND Inspections of some reinforced and steel containments (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 containment.

In response to previous ILRT extension request submittals, the NRC has consistently requested licensees to perform a quantitative assessment of the impact on LERF due to age-related degradation of non-inspectable areas of the containment.

Therefore, a quantitative assessment using the same approach used by other industry plants (e.g., Calvert Cliffs) is included as Appendix B to this ILRT extension evaluation.

The analysis described in Sections 3 and 4 of the report was performed to evaluate the risk impact of extending the Integrated Leak Rate Test (ILRT) interval for the Dresden Nuclear Power Station.

That analysis was performed using the recommended approach developed by NEI for performing assessments of one-time extensions for containment ILRT surveillance intervals [B-1].

The results of that analysis are summarized in Table B-1, which is a copy of Table 4-1 from the main report.

The risk increase from extending the ILRT interval from the present 1-in-10 year requirement to 1-in-15 years is quantified by the increase in LERF (the CDF is not impacted by the ILRT interval). The NRC Regulatory Guide 1.174 [B-2] defines very small changes in risk as resulting in increases in LERF below 1.OE-7/yr. The Regulatory Guide also states that when the calculated increase in LERF is in the range of 1.OE-6/yr to 1.OE-7/yr, applications will be considered only if it can be reasonably shown that the total LERF is less than 1.OE-5/yr.

B-1 C467030603-5572-1 2/17/03

Sensitivity Calculation for Dresden ILRTRiskAssessnent For Dresden the increase in LERF from the 1-in-10 year interval to the 1-in-15 year interval was calculated in the main report to be 7.65E-9/yr, which is well below the very small change threshold. Also, the dose rate increase was determined to be negligible compared with the total 10.3 person-rem/yr. The increase in the containment failure probability (CCFP) was determined to be 0.4%, which is also judged to be insignificant.

B-2 c467030603-5572-1 2/17/03

Sensitivity Calctlation for Dresden ILRT Risk Assessment Table B-1 QUANTITATIVE RESULTS AS A FUNCTION OF ILRT INTERVAL EPRI Dose Baseline Current Proposed Category (Person-Rem (3-per-10 year ILRT)

(1y-per-10 ear ILRT)

(1-per-15 year ILRT)

Within 50 miles)

Accident Population Accident Population Accident Population Frequency Dose Rate Frequency Dose Rate Frequency Dose Rate (per year)

(Person-Rem/Year (per year)

(Person-Rem/Year (per year)

(Person-Rem/Year Within 50 miles)

Within 50 miles)

Within 50 miles) 1 2.08E+3 1.07E-6 2.22E-3 9.52E-7 1.98E-3 8.68E-7 1.80E-3 2

2.22E+7 4.67E-9 1.04E-1 4.67E-9 1.04E-1 4.67E-9 1.04E-1 3a 2.08E+4 4.59E-8 9.54E-4 1.53E-7 3.18E-3 2.29E-7 4.77E-3 3b 7.28E+4 4.59E-9 3.34E-4 1.53E-8 1.11E-3 2.29E-8 1.67E-3 4

n/a n/a n/a N/a n/a n/a n/a 5

n/a n/a n/a N/a n/a n/a n/a 6

n/a n/a n/a n/a 7_

1.33E+7 7

8 _

2.79E+7

]

I 7

1.02E+1 7.65E-7 j

1.02E+1 j

7.65E-7 1.02E+1 9

4.85E-2 I

1.74E-9 I

4.85E-2 1.74E-9 4.85E-2 TOTALS:

Increase in Dose Rate ~')

Increase in LERF 'i)

Increase in CCFP(%))

(1) The Increase In dose rate (person-rem/year) Is with respect to the results for the preceding ILRT interval, as presented In the table. For example, the increase In dose rate forthe proposed 1-per-15 ILRT is calculated as: total dose ratefor 1-per-15 year ILRT, minus total dose rate for 1-per-10 year ILRT. For each case, the dose rate Increase Is Insignificant.

(2) The Increase in Large Early Release Frequency (LERF) is with respect to the results for the preceding ILRT Interval, as presented In the table. As discussed in Section 3.4.4 of the report, the change in LERF is determined by the change In the accident frequency of EPRI Category 3b. For example, the Increase In LERF for the proposed 1-per-1 5 ILRT is calculated as: 3b frequency for 1-per-15 year ILRT, 2.29E-8/yr, minus 3b frequency for 1-per-10 year ILRT, 1.53E-8lyr, equals 7.65E-9/yr.

(3) As discussed in Section 3.4.5, the conditional containment failure probability (CCFP) is calculated as:

CCFP% = [1 - ((Category #1 Frequency + Category #3a Frequency) / CDF)] x 100%

(4) There is no calculated change In the total population dose rate unless more significant digits are utilized.

B-3 B-3

~~~~~~~~~~~~~C467030603-5572-12117/03

Sensitivity Calculation for Dresden ILRTRiskAssessment B.2 CORROSION ANALYSIS This containment corrosion sensitivity analysis utilizes the methods of the Calvert Cliffs liner corrosion analysis [B-3] to estimate the likelihood and risk-implications of degradation-induced leakage occurring undetected during the extended test interval.

The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, each with a steel liner. The Dresden containment is a pressure-suppression BWR/Mark I type with a steel shell in the drywell region, including the portion below the concrete drywell floor. The shell is surrounded by a concrete shield.

The following approach is used to determine the change in likelihood of detecting corrosion of the steel containment shell due to extending the ILRT. This likelihood is then used to determine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the following issues are addressed:

Differences between the containment floor and other regions of the containment The historical steel liner 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 B.2.1 ASSUMPTIONS The following assumptions are utilized in the sensitivity analysis:

A. The Oyster Creek incident is assumed to be applicable for Dresden for a concealed flaw in the shell in the floor. (See Table B-2, Step 1.) In the Calvert Cliffs analysis, no applicable events were identified and 0.5 failures were assumed. For Dresden it will be assumed that there has been one failure in industry experience for the floor area.

B-4 C467030603-5572-12117/03

Sensitivity Calculation for Dresden ILRTRisk Assessment B. The two corrosion events used to estimate the wall liner flaw probability in the Calvert Cliffs analysis are assumed to be applicable to the Dresden containment analysis. These events, one at North Anna Unit 2 and one at Brunswick Unit 2, were initiated from the non-visible (backside) portion of the containment liner.

C. For consistency with the Calvert Cliffs analysis, the estimated historical flaw probability is calculated using a 5.5-year data period. This reflects the span from September 1996 when 10 CFR 50.55a started requiring visual inspection to the time of the Calvert Cliffs analysis. Additional success data were not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date (and have been performed since the time frame of the Calvert Cliffs analysis), and there is no evidence that additional corrosion issues were identified. (See Table B-2, Step 1.)

D. Consistent with the Calvert Cliffs analysis, the corrosion-induced steel liner flaw likelihood is assumed to double every five years. This is based solely on judgment and is included in this analysis to address the increased likelihood of corrosion as the steel liner ages. (See Table B-2, Steps 2 and 3.) Sensitivity studies are included that address doubling this rate every ten years and every two years.

E. In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere given that a liner flaw exists was estimated (based on an assessment of the containment fragility curve versus the ILRT test pressure) as 1.1% for the containment walls and dome region and 0.11% (factor often less) for the basemat. For Dresden the containment failure probabilities are conservatively assumed to be 10% for the shell wall and 1% for the floor.

Sensitivity studies are included that increase and decrease the probabilities by an order of magnitude. (See Table B-2, Step 4.)

F. In the Calvert Cliffs analysis it is noted that approximately 85% of the interior wall surface is accessible for visual inspections. The Dresden interior wall surface accessible for visual inspections is estimated to be at least 95% (only minor attachment locations for equipment such as lighting is inaccessible). Therefore, consistent with the Calvert analysis, a 5%

visual inspection detection failure likelihood given the flaw is visible and a 5% likelihood of a non-detectable flaw are used. This results in a total undetected flaw probability of 10%, which is assumed in the base case B-5 c467030603-5572-12117103

Sensitivity Calculation for Dresden ILRTRiskAssessment analysis. (See Table B-2, Step 5.) Sensitivity studies are included that evaluate a total detection failure likelihood of 5% and 15%, respectively.

(See Table B4 for sensitivity studies.) Additionally, it should be noted that to date, all liner corrosion events have been detected through visual inspection and repaired.

G. Consistent with the Calvert analysis, all non-detectable containment failures are assumed to result in early releases. This approach avoids a detailed analysis of containment failure timing and operator recovery actions.

B-6 C467030603-5572-12117/03

Sensitivity Calcldation for Dresden ILRTRiskAssessnent B.2.2 ANALYSIS Table B-2 STEEL LINER CORROSION BASE CASE Step Description Containment Walls Containment Floor 1

Historical Steel Liner Flaw Industry Applicable Events:

Industry Applicable Events:

Likelihood 2

1 (North Anna and Brunswick (Oyster Creek event Failure Data: Containment location events assumed to be assumed applicable to specific (applicable wall events and applicable to Dresden)

Dresden) derived failure value is consistent with Calvert Cliffs analysis; one floor 2/(70

• 5.5) = 5.2E-3 1/(70

• 5.5) = 2.6E-3 event assumed applicable for Dresden whereas the Calvert Cliffs (Based on 70 units with (Based on 70 units with analysis assumed 0.5 failures).

liners over 5.5 years) liners over 5.5 years) 2 Age Adjusted Steel Liner Flaw Flaw Flaw Likelihood Year Likelihood Year Likelihood 0

1.8E-03 0

8.9E-04 During 15-year interval, assume 1

2.1E-03 I

1.0E-03 failure rate doubles at the end of 2

2.4E-03 2

1.2E-03 every five years (which equates to a 3

2.7E-03 3

1.4E-03 14.9% increase per year). The 4

3.1E-03 4

1.6E-03 average over the 5e' through 10t 5

3.6E-03 5

1.8E-03 year period is set equal to the 6

4.1E-03 6

2.1E-03 historical failure rate of Step 1 7

4.7E-03 7

2.4E-03 (consistent with Calvert Cliffs 8

5.4E-03 8

2.7E-03 analysis). These assumptions are 9

6.2E-03 9

3.1E-03 used to calculate the flaw likelihood 10 7.1E-03 10 3.6E-03 for each year (for a 15 year period) 11 8.2E-03 11 4.1 E-03 12 9.4E-03 12 4.7E-03 13 1.1 E-02 13 5.4E-03 14 1.2E-02 14 6.2E-03 15 1.4E-02 15 7.1 E-03 3

Flaw Likelihood at 3, 10, and 15 7.12E-3 (at 3 years) 3.56E-3 (at 3 years) years 4.14E-2 (at 10 years) 2.07E-2 (at 10 years) 9.66E-2 (at 15 years) 4.83E-2 (at 15 years)

This cumulative probability uses the age adjusted liner flaw likelihood of (Note that the Calvert Cliffs (Note that the Calvert Cliffs Step 2 (consistent with Calvert Cliffs analysis presents the delta analysis presents the delta analysis - See Table 6 of Reference between 3 and 15 years of between 3 and 15 years of

[B-3]). For example, the 7.12E-03 8.7% to utilize in the 2.2% to utilize in the (at 3 years) cumulative flaw estimation of the delta-LERF estimation of the delta-LERF likelihood is the sum of the year 1, value. For this analysis the value. For this analysis the year 2, and year 3 likelihoods of Step values are calculated based values are calculated based

2.

on the 3,10, and 15 year on the 3,10, and 15 year intervals.)

intervals.)

B-7 C467030603-5572-12/17/03

Sensitivity Calcidationfor Dresden 1LRTRiskAssessrnent Table B-2 STEEL LINER CORROSION BASE CASE Step Description Containment Walls Containment Floor 4

Likelihood of Breach In 10%

1%

Containment Given Steel Liner Flaw The wall failure probability of the containment is assumed to be 10%

(compared to 1.1% in the Calvert Cliffs analysis). The floor failure probability is assumed to be a factor of ten less, 1%, (compared to 0.11%

in the Calvert Cliffs analysis).

5 Visual Inspection Detection 10%

100%

Failure Likelihood 5% failure to identify visual Cannot be visually Utilize assumptions consistent with flaws plus 5% likelihood that inspected.

Calvert Cliffs analysis.

the flaw is not visible (not through-wall but could be detected by ILRT).

All events have been detected through visual inspection. 5% visible failure detection is a conservative assumption.

6 Likelihood of Non-Detected 7.12E-5 (at 3 years) 3.56E-5 (at 3 years)

Containment Leakage 7.12E-3

• 10%

• 10%

3.56E-3

• 1%

• 100%

(Steps 3

• 4

• 5) 4.14E-4 (at 10 years) 2.07E-4 (at 10 years) 4.14E-2

• 10%

• 10%

2.07E-2

• 1%

• 100%

9.66E-4 (at 15 years) 4.83E-4 (at 15 years) 9.66E-2

• 10%

• 10%

4.83E-2

• 1%

• 100%

B-8 C467030603-5572-12/17/03

Sensitivity Calculation for Dresden ILRTRiskAssessment Cumulative Likelihood of Non-Detected Containment Leakaqe Due to Corrosion The total likelihood of the corrosion-induced, non-detected containment leakage is the sum in Step 6 for the containment walls and the containment floor:

At 3 years: 7.12E-5 + 3.56E-5 = 1.07E-4 At 10 years: 4.14E-4 + 2.07E-4 = 6.21 E-4 At 15 years: 9.66E-4 + 4.83E-4 = 1.45E-3 Table B-3 summarizes the results of the revised ILRT assessment including the potential impact from non-detected corrosion-induced containment leakage scenarios, with the assumption that all of these scenarios result in EPRI Class 3b (i.e., LERF).

The impact of including the potential for corrosion-induced leakages compared to the original analysis results presented in the main report is noted in parentheses.

The factors calculated above are applied to those core damage accidents that are not already independently LERF or that could never result in LERF. For example, the 3-in-10 year case is calculated as follows:

Per Table B-1, the EPRI Class 3b frequency is 4.59E-9/yr.

As discussed in Section 3.1, the Dresden CDF associated with accidents that are not independently LERF or that could never result in LERF is 1.89E-6/yr - (1.08E-7/yr + 8.15E-8/yr + 1.74E-9/yr) = 1.70E-6/yr.

The increase in the base case Class 3b frequency due to the corrosion-induced concealed flaw issue is calculated as 1.70E-6/yr

• 1.07E-4 =

1.82E-10/yr, where 1.07E-4 was previously shown to be the cumulative likelihood of non-detected containment leakage due to corrosion at 3 years.

The 3-in-10-year Class 3b frequency including the corrosion-induced concealed flaw issue is then calculated as 4.59E-9Iyr + 1.82E-10/yr =

4.77E-9/yr.

B-9 B-9

~~~~~~~C467030603-557

Sensitivity Calculatiofifor Dresden ILRTRisk Assessment Table B-3 DRESDEN ILRT CASES: BASE, I IN 10, AND 1 IN 15 YR EXTENSIONS (Including Age Adjusted Steel Liner Corrosion Likelihood) (')

EPRI Dose Base Case Extend to Extend to Category (Per-Rem) 3 in 10 Years I in 10 Years I in 15 Years Core Damage Dose Rate Core Damage Dose Rate Core Damage Dose Ratel Frequency (person-Frequency (person-Frequency (person-(Iyr)

Remlyr)

(Iyr)

Rem/yr)

(Iyr)

Remlyr) 1 2.08E+3 1.07E-6 2.22E-3 9.51E-7 1.97E-3 8.66E-7 1.80E-3 2

2.22E+7 4.67E-9 1.04E-1 4.67E-9 1.04E-1 4.67E-9 1.04E-1 3a 2.08E+4 4.59E-8 9.54E-4 1.53E-7 3.18E-3 2.29E-7 4.77E-3 3b 7.28E+4 4.77E-9 3.47E-4 1.63E-8 1.19E-3 2.54E-8 1.85E-3 7

1.33E+71 7.65E-71 1.02E+1 7.65E-7 1.02E+1 7.65E-7 1.02E+1 8

2.79E+7 1.74E-9 4.85E-2 1.74E-9 4.85E-2 1.74E-9 4.85E-2 Total 1.89E-6 10.33 1.89E-6 10.33 1.89E-6 10.34 Dose Rate from 3a 1.30E-3 4.37E-3 6.62E-3 and 3b (person-(+1.3E-5)

(+7.7E-5)

(+1.8E-4)

Remlyr)

I Increase in Total From 3-per-10 yr 2.82E-3 4.89E-3 Dose Rate

= 0.03%

= 0.05%

(person-Remlyr)

I

(+6.0E-5)_

(+1.6E-4)

From 1-per-10 yr 2.07E-3

= 0.02%

(+1.OE-4)

LERF from 3b 4.77E-9 1.63E-8 2.54E-8 (nyr)s

(+1.8E-10n

(+1.1E-9)

(+2.5E-9)

Increase in LERi From 3-per-10 yr 1.16E-8 2.06E-8 (Y)

(+8.7E-10)

(+2.3E-9)

From 1-per-10 yr 9.06E-9

_(+1.4E-9)

CCFP %

41.0%

41.6%

42.1%

(+0.01%)

(+0.05%)

(+0.12%)

Increase in CCFP From 3-per-10 yr 0.61%

1.09%

(+0.04%)

(+0.11%)

From 1-per-10 yr 0.48%

(+ 00 7% )

() The numbers in parenthesis represent the incremental change (compared to Table B-1) due to inclusion of the impact from the corrosion analysis.

B-1 0 C467030603-5572-12/17/03

Sensitivity Calculation for Dresden ILRTRisk Assessment Results Based on the results shown in Table B-3, it can be seen that including corrosion effects in the ILRT assessment does not alter the conclusions from the original analysis. The increase in LERF from the 1-in-10 year interval to the 1-in-15 year interval is 9.06E-9/year, compared with 7.65E-9/yr without corrosion effects. This is still well below the Regulatory Guide 1.174 [B-2] acceptance criterion threshold for very small changes in risk of.OE-7/yr. This confirms that the proposed interval extension is acceptable from a risk basis. Additionally, the dose rate increase is negligible compared to the total 10.3 person-rem/yr. The increase in the CCFP is determined to be insignificant (42.1% for the 1-in-15 year case versus 41.6% for the 1-in-10 year case).

B-1 1 C467030603-5572-1 2/17/03

Sensitivity Calculation for Dresden ILRTRiskAssessrnent B.3 SENSITIVITY STUDIES Sensitivity cases were also developed to gain an understanding of the sensitivity of this analysis to the various key parameters. The time for the flaw likelihood to double was adjusted from every five years to every two and every ten years. The failure probabilities for the shell wall and the floor were increased and decreased by an order of magnitude. The total detection failure likelihood was adjusted from 10% to 15% and 5%. The results of the sensitivity cases are summarized in Table B4. For all cases the increase in LERF (for 3-in-10 years to 1-in-15 years) due to corrosion is less than 1.OE-7/yr. The total increase in LERF is within the range of 1.85E-8/yr and 8.57E-8/yr.

B-12 B-I 2

~~~~~C467030603-5572-

Sensitivity Calculation for Dresden ILRTRisk Assessment Table B-4 DRESDEN STEEL LINER CORROSION SENSITIVITY CASES Increase in Class 3b Frequency Visual (LERF) for ILRT Extension From Age Containment Inspection &

3-in-10 to 1-in-15 years (Step 3)

Breach Non-Visual Qvr)

(Step 4)

Flaws (Step 5)

Total Increase Increase Due to Corrosion Base Case Base Case Base Case 2.06E-8 2.28E-9 (Doubles every (10% Walls, %

(10%)

5 yrs)

Floor)

Doubles every Base Base 2.34E-8 5.05E-9 2 yrs Doubles every Base Base 2.02E-8 I.86E-9 10 yrs Base Base 15%

2.14E-8 3.04E-9 Base Base 5%

1.99E-8 1.52E-9 Base 100% Walls, 10%

Base 4.1 E-8 2.28E-8 Floor Base 1% Walls, 0.1%

Base 1.86E-8 2.28E-10 Floor Lower Bound Doubles every 1% Walls, 0.1%

5%

1.85E-8 1.24E-10 10 yrs jFloor I__________ ______

Upper Bound Doubles every I 100% Walls, 10%

15%

l 8.57E-8 6.74E-8 2 yrs Floor10___________

B-13 C467030603-5572-12/17/03

Sensitivity Calculation for Dresden ILRTRiskAssessment B.4

SUMMARY

AND CONCLUSIONS This sensitivity analysis provides a quantitative assessment of the impact on risk of the potential for undetected steel liner corrosion due to an extension of the ILRT interval.

The increase in LERF due to extending the test interval from the present schedule of 1 in 10 years to 1 in 15 years is 9.06E-9/yr, of which 1.4E-9/yr is due to corrosion. This value is considerably less than the RG 1.174 very small" change criterion of 1.OE-7/yr.

This confirms that the proposed interval extension is acceptable from a risk basis.

Additionally, a series of parametric sensitivity studies regarding the potential age-related corrosion effects on the steel liner 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 public risk and is therefore acceptable.

B-14 C467030603-5572-12117/03

Risk Impact Assessment of Extending Dresden ILRT Interval REFERENCES

[B-1] Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Intervals, Developed for NEI by John M. Gisclon, EPRI Consultant, William Parkinson and Ken Canavan, Data Systems and Solutions, November 2001.

[B-2] 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, Revision 1, November 2002.

[B-3] 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.

B-1 5 c467030603-5572-12117/03

APPENDIX C SENSITI VIT Y FOR LONG TERM STA TION BLACKOUT

Risk Impact Assessment ofExtending Dresden ILRT Interval Appendix C SENSITIVITY FOR LONG TERM STATION BLACKOUT The definition of Large Early Release Frequency (LERF) requires the release to be both

'large" and early." The definition of early is tied to a release that occurs prior to effective protective actions for the public (e.g., evacuation). Station Blackout (SBO) events which result in radionuclide release at times after effective public protective actions have been implemented (sometimes referred to as long term SBO events) are not treated as LERF.

The Nuclear Regulatory Commission (NRC) has previously requested licensees to provide (as part of ILRT RAls) the technical justification for the assumption that long term station blackout scenarios do not contribute to LERF, and to provide an assessment of the impact on risk results if long-term station blackout sequences were retained in selected EPRI categories.

This appendix provides the technical justification for not including long term SBO scenarios in the LERF assessment and also provides a sensitivity case to demonstrate that retaining SBO sequences in the analysis does not change the conclusions of the overall ILRT assessment.

C.1 TECHNICAL JUSTIFICATION FOR EXCLUSION OF SBO SCENARIOS Typical of many industry PRAs, the Dresden PRA uses a radionuclide release categorization scheme comprised of two factors: release timing and release magnitude.

The Dresden long-term station blackout core damage accidents (Class IBL) are classified as non-LERF releases based on release timing rather than release magnitude (i.e., IBL core damage accidents have the potential to result in the entire spectrum of release magnitudes, including High magnitude releases; but, they cannot result in Early releases).

The following describes the timing issues of Dresden Class IBL scenarios.

Three timing categories are used, as follows:

C-1 C467030603-5572-12117/2003

Risk Impact Assessment of Extending Dresden ILRTInterval

1.

Early (E)

Less than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />

2.

Intermediate (I)

Greater than or equal to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, but less than or equal to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

3.

Late (L)

Greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The above accident release categories are based upon past experience concerning offsite accident response:

0-5 hours is conservatively assumed to include cases in which minimal offsite protective measures have been observed to be performed in non-nuclear accidents.

5-24 hours is a time frame in which much of the offsite nuclear plant protective measures can be assured to be accomplished.

>24 hours are times at which the offsite measures can be assumed to be effective and resources for repair and recovery can be implemented effectively.

The timing categories are relative to the declaration of the Dresden General Emergency Action Level (per Exelon Nuclear's Radiological Emergency Plan Annex For Dresden Station" [C-1]).

The Dresden Class IBL accident scenarios (i.e., long term station blackout) include only those sequences in which high pressure inventory control (HPCI or the isolation condenser) is available initially in the accident but subsequently fails without AC Power recovery. There are a spectrum of SBO events that can lead to core damage at various times within Class IBL. These times vary as shown in the following examples:

C-2 C467030603-5572-12117/2003

Risk Impact Assessment ofExtending Dresden ILRT Interval Description Time to Core Damage IC operates with shell makeup and with seal leakage (No HPCI)

IC operates without additional shell makeup (No HPCI)

IC fails, HPCI operates for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (RPV depressurized) RPV must depressurize to boil down.

IC operates until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when DC failure causes loss of IC. RPV repressurizes until DEOPs direct RPV depressurization. (Despite no DC available, RPV depressurization is assigned because it results in more rapid core damage event.)

13.2 Hrs Not a long term SBO event See case below 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> The shortest duration representative IBL sequence for Dresden is sequence LOOP-14 of the LOOP event tree. Sequence LOOP-14 proceeds as follows according to MAAP 4.0.4 Case DR009:

Event Time After Plant Trip Loss of Offsite Power initiating event 0

Failure of emergency AC power (EDGs) and 0

the SBO Diesel Generators HPCI/lsolation Condenser Initiation

-1 min.

Battery depletion 4 hrs.

Failure to blowdown (no DC power) 4 hrs.

Loss of HPCI/lsolation Condenser (no DC power)(')

4 hrs.

RPV Depressurization (Manual Action via HPCI Steam 5.5 hrs.

Liner or MS Drains)

Time to core damage (1800F)

-6 hrs.

Time to energetic containment failure (fastest, but low 7 to 8 hrs.

frequency, release scenario)

(1) With the failure of DC power, HPCI and the Isolation Condenser are potentially unavailable if local manual actions fail. This time line represents the most limiting of these late" SBO events.

C-3 C467030603-5572-1211712003

Risk Impact Assessment of Extending Dresden ILRTInterval As can be seen from the above scenario, the Dresden IBL accident class results in a radionuclide release no earlier than approximately 7 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the LOOP initiator.

The 7 to 8-hour release for the IBL core damage accident makes the conservative assumption that an early energetic containment failure mode (in-vessel steam explosion) occurs at about the time of molten core relocation to the lower head (a low probability containment failure mode for the IBL accident).

The Dresden Emergency Plan (Recognition Category MG1) directs declaration of a General Emergency (i.e., the emergency classification with associated directives for evacuation) for the following station blackout conditions [C-1]:

1. Loss of all AC power to ECCS bus 23-1 (33-1)

AND

2. Loss of all AC power to ECCS bus 24-1 (34-1).

AND

3. Any of the following:

Restoration of power to EITHER bus 23-1 (33-1) OR 24-1 (34-1) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is NOT likely.

OR Conditions are imminent that a loss of 2 Fission Product Barriers and Loss or Potential Loss of the 3rd (FG1) will occur prior to restoration of AC power to the Unit. (Imminent is defined as "mitigation actions have been ineffective and trended information indicates that the event or condition will occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.")

The loss of offsite and emergency power occurs at t=O for Class IBL sequences. The Dresden PRA assumes that the determination that AC power is not likely to be restored within the 4-hour time frame is made within the first hour into the accident. As such, a C-4

-C467030603-5572-12117/2003

Risk Impact Assessment of Extending Dresden ILRTInterval General Emergency is assumed declared at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> into the event. The evacuation process would be initiated within minutes after the declaration and is estimated to be completed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 45 minutes under worst assumed conditions based on site specific evacuation studies for weather and times of day variations [C-1]. The earliest release for the IBL scenario occurs at approximately 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> (over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after evacuation is expected to be completed). Therefore, the IBL core damage accident is not an Early release.

C.2 SENSITIVITY FOR LONG TERM SBO The NRC has previously asked licensees, via Requests for Additional Information (RAls),

to provide an assessment of the impact on risk results if long term station blackout sequences were retained in selected EPRI categories.

This appendix provides that information for use by the NRC in their deliberations.

The frequency of long term SBO core damage sequences in the Dresden PRA is 1.08E-7/yr. [C-2]

This sensitivity case repeats the calculations of the ILRT assessment performed in Section 3 of the main report, with the exception that the long term SBO sequences are retained in the EPRI Category 3b frequency calculations as potential LERF contributors.

Refer to Section 3.1 of the main report for the EPRI Category calculational methodology.

Retaining the SBO sequences in the EPRI Categories 3a and 3b frequency calculations results in the following new frequencies:

Category 3a

= 4.88E-8/yr Category 3b

= 4.88E-9/yr Category I

= 1.07E-6/yr (unchanged due to round-off) c-5 C467030603-5572-12/172003

Risk Impact Assessment of Extending Dresden ILRTInter'al The impact of the changes to these EPRI Category frequencies is shown in Table C-1.

The total increase in LERF using the full power internal events PRA due to the extension of the ILRT interval from 10 years to 15 years is determined to be 8.13E-9/yr when long term SBO scenarios are included in the EPRI Category 3a and 3b frequencies. This represents an additional LERF increase of 4.8E-10/yr (a 6% increase) over the best estimate ILRT increase in LERF of 7.65E-9/yr based on internal events. Including the long term SBO contribution, however, still results in a LERF increase below the NRC Regulatory Guide 1.174 criterion of 1.OE-7/yr for "very small" risk change. The population dose rate for the 10 year and 15 year ILRT intervals with long term SBO sequences included remains, the same as in the baseline analysis. The increase in the conditional containment failure probability (CCFP) is determined to remain unchanged (0.4% with SBO sequences included versus 0.4% in the baseline analysis).

The sensitivity case demonstrates that even if long term SBO scenarios are included in the EPRI Category 3a and 3b frequencies, the conclusion of the risk assessment does not change; that is, the Dresden ILRT interval extension to 15 years has a minimal impact on plant risk.

C-6 C-6

~~~~~~C467030603-5572-7/2003

Risk ipact Assessment of Extending Dresden ILRT Interval Table C-1 QUANTITATIVE RESULTS AS A FUNCTION OF ILRT INTERVAL

- Sensitivity Case to Include Long-Term SBO Contributions In Category 3a and 3b Frequencies -

Baseline Current Proposed (3-per-10 ear ILRT)

(1-per-10 ar ILRT)

(1-per-15 ear ILRT)

Dose Accident Population Accident Population Accident Population EPRI (Person-Rem Frequency Dose Rate Frequency Dose Rate Frequency Dose Rate Category Within 50 miles)

(peryear)

(Person-Remfear (peryear)

(Person-Rem/Year (per year)

(Person-Rem/Year Within 50 miles)

Within 50 miles)

Within 50 miles) 1 2.08E+03 1.07E-06 2.22E-03 9.41 E-07 1.96E-03 8.52E-07 1.77E-03 2

2.22E+07 4.67E-09 1.04E-01 4.67E-09 1.04E-01 4.67E-09 1.04E-01 3a 2.08E+04 4.88E-08 1.01 E-03 1.63E-07 3.38E-03 2.44E-07 5.07E-03 3b 7.28E+04 4.88E-09 3.55E-04 1.63E-08 1.18E-03 2.44E-08 1.78E-03 4

n/a n/a n/a n/a n/a n/a n/a 5

n/a n/a n/a n/a n/a n/a n/a 6

n/a n/a n/a n/a n/a n/a n/a 7

1.33E+07 7.65E-07 1.02E+01 7.65E-07 1.02E+01 7.65E-07 1.02E+01 8

2.79E+07 1.74E-09_

4.85E-02 1.74E-09 4.85E-02 1.74E-09 4.85E-02 TOTALS:

il 1.F-06R 10.3 i 1 8Q-06 10.3 I

1 A9E-06 10 3 Increase in Dose Rate (1)

Increase in LERF (2)

Increase in CCFPy%3 )

(1) The Increase in Dose Rate (person-rem/year) is with respect to the preceding ILRT interval, and is calculated by subtracting the Dose Rate totals.

(2) The Increase in LERF is with respect to the preceding ILRT interval, and is calculated by subtracting the EPRI Category 3b frequencies.

(3) The Increase in CCFP% (units in percentage points) is with respect to the preceding ILRT interval. The CCFP% is calculated as:

CCFP% = [1 - ((Category I Frequency + Category 3a Frequency) / CDF)] x 100 (4) There is no calculated change in the total population dose rate to four significant digits.

C-7 0-7

~~~~~~~~~~~C467030603-5572

Risk ImpactAssessment of Extending Dresden ILRTInterval REFERENCES

[C-1] Exelon Nuclear, EP-AA-1004, RadioloQical Emergency Plan Annex For Dresden Station, Rev. 17, August 2003.

[C-2]

Dresden Detailed Level 2 Evaluation," DR PSA-015, Volume I and 2, July 2002.

C-8 C467030603-5572-121 7/03