ML12349A328

Exelon'S Counter Affidavit Supporting Exelon'S Response Opposing Nrdcs Petition for Waiver of 10 C.F.R. Section 51.53(C)(3)(ii)(L)

ML12349A328
Person / Time
Site:	Limerick
Issue date:	12/14/2012
From:	Gabor J, Kelley E, Macleod D, Vanover D ERIN Engineering & Research
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 23891, 50-352-LR, 50-353-LR, ASLBP 12-916-04-LR-BD01
Download: ML12349A328 (29)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

)

In the Matter of )

) Docket Nos. 50-352-LR EXELON GENERATION COMPANY, LLC ) 50-353-LR

)

(Limerick Generating Station, Units 1 and 2) ) December 14, 2012

)

EXELONS COUNTER AFFIDAVIT SUPPORTING EXELONS RESPONSE OPPOSING NRDCS PETITION FOR WAIVER OF 10 C.F.R. § 51.53(C)(3)(ii)(L)

I. PERSONAL BACKGROUND INFORMATION I, Jeffrey R. Gabor (JG), state as follows:

1. (JG) I am Vice President of the Risk Management Group for ERIN Engineering and Research, Inc (ERIN Engineering). My qualifications are summarized in the attached curriculum vitae. Briefly, I have over 30 years of experience in nuclear power plant safety, including extensive experience in severe accident management and analysis pertaining particularly to Boiling Water Reactors (BWRs), such as the Limerick Nuclear Generating Station (Limerick). I hold a Bachelor of Science degree in Nuclear Engineering and a Master of Science degree in Mechanical Engineering from the University of Cincinnati, Ohio.

2. (JG) With respect to my experience with Severe Accident Mitigation Alternatives (SAMAs), I have supported over half of all U.S. nuclear plant license renewal SAMA analyses to date and am otherwise extremely involved with the nuclear industry. I was a primary author of NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) 1

Guidance Document (Nov. 2005), which the NRC Staff recommends as guidance for license renewal applicants who are required to prepare a SAMA analysis as part of their application.1 I also am experienced in developing SAMA files for computer model analysis and related personnel training.

3. (JG) I also have extensive experience in Probabilistic Risk Assessments (PRA). I was the lead technical analyst for severe accident response on numerous BWR PRAs, including Millstone Unit 1, Duane Arnold, Pilgrim, Nine Mile Point Units 1 and 2, Fermi, Vermont Yankee, Cofrentes (Spain), and Browns Ferry. I was a principal author of the BWR Modular Accident Analysis Program (MAAP), a computer code that simulates reactor accidents for PRA applications and which has a BWR-specific version.

I am a member of the Mitigating Systems Performance Index PRA Quality Task Group, and have made numerous technical presentations to the U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee on Reactor Safeguards, as well as the U.S. Department of Energy.

I, Donald E. MacLeod (DM), state as follows:

4. (DM) I am an expert in PRA with extensive experience in SAMA analysis. I have over fifteen years of experience with ERIN Engineering, specialize in Probabilistic Risk Assessment, and hold a Bachelor of Science degree in Nuclear Engineering from Rensselaer Polytechnic Institute.

5. (DM) My experience in SAMA analyses includes holding the role of lead analyst performing SAMA analyses for many U.S. nuclear plants, and co-developing several others. Specifically, I was lead analyst performing SAMA analyses for Three Mile 1

LR-ISG-2006-03, Final License Renewal Interim Staff Guidance LR-ISG-2006-03: Staff Guidance for Preparing Severe Accident Mitigation Alternatives Analyses at 1 (Aug. 2, 2007).

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Island, Shearon Harris, Wolf Creek, V.C. Summer, Brunswick, H.B. Robinson, Monticello, Palisades, Susquehanna, South Texas Project, and Palo Verde. I co-developed SAMA analyses for Peach Bottom, Salem Generating Station, Hope Creek, Diablo Canyon, and Crystal River. Particular to Limerick, I was the lead analyst in developing an update of the plants Human Reliability Analysis.

I, Donald E. Vanover (DV), state as follows:

6. (DV) I am an expert in PRA with extensive experience in developing and updating PRA models for several BWR and PWR reactors. My qualifications are summarized in the attached curriculum vitae. Briefly, I have over 25 years of experience in nuclear power plant safety, including extensive experience with all aspects of the Limerick PRA models.

While with ERIN Engineering since 1995, I have been involved in numerous applications of PRA models to meet current regulatory requirements, and also in support of license amendment requests to the NRC. I hold a Bachelor of Science degree and a Master of Science degree in Mechanical Engineering from the University of Delaware.

7. (DV) My experience in PRA has also led to the development of several industry guidance documents published by EPRI including guidance for the treatment of PRA model uncertainty. My experience in SAMA analysis includes being a principal contributor to the SAMA analyses performed for Peach Bottom, Susquehanna, and Vogtle.

I, Eugene Kelly (EK), state as follows:

8. (EK) I am an expert in licensing and design basis, with extensive experience in power plant operation and testing, engineering and design, and licensing. I have over 38 years of nuclear power plant experience, including 13 years at Limerick, with specialized expertise in engineering programs and testing. I also have 17 years of regulatory and 3

licensing experience with the NRC, including holding the position of Senior Resident Inspector at Limerick. I hold a Bachelor of Science degree in Physics from Villanova University and a Masters of Science in Mechanical Engineering from the University of Pennsylvania.

9. (EK) My experience in engineering programs includes managing the Engineering Programs branch at Limerick, chairing the INPO Programs excellence working group, and serving as the technical manager responsible for the Limerick License Renewal Application. Specifically, I was responsible for all tests and inspections at Limerick associated with engineering programs, including service water cooling systems and buried piping. I also piloted a risk-informed surveillance test frequency program that was licensed for Limerick and serves as the basis for an industry-wide surveillance test initiative that uses PRA and risk techniques and insights to create test programs for a wide variety of systems and components. As the technical lead for the Limerick license renewal project, I was responsible for the development of all 45 aging management programs including those for Open Cycle Cooling Water and Buried Piping and Tanks.

10. (EK) In my prior position as the NRC Region I manager of the Engineering Systems Branch, I was responsible for inspections of over 30 nuclear plants throughout the Northeast United States including team inspections of Generic Letter 89-13 service water testing programs. This NRC branch was also responsible for the oversight of PRA techniques and applications at over 30 nuclear plants.

11. (EK) My technical training and experience includes specialized expertise in heat transfer, fluid dynamics, risk management and safety analysis.

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II. PURPOSE AND SCOPE

12. (All) We have reviewed the Declaration of Christopher J. Weaver, Ph.D., on Behalf of the Natural Resources Defense Council [NRDC] in Support of Motion for Waiver submitted on behalf of NRDC, and the other arguments NRDC makes in its Petition, By Way of Motion, For Waiver of 10 C.F.R. § 51.53(C)(3)(ii)(L) As Applied to Application for Renewal of Licenses for Limerick Units 1 and 2 (Nov. 21, 2012) (Waiver Petition). We offer our statements in this Counter Affidavit to support Exelons response opposing NRDCs Waiver Petition.

III. TECHNICAL ARGUMENT A. There are no major design changes or major plant modifications among the SAMAs that NRDC identified.

13. (DM, JG) NRDC identifies approximately 50 SAMAs in Paragraph 11 of the Weaver Declaration that its expert identified as cost-beneficial or potentially cost-beneficial at other BWRs. Paragraph 10 of the Weaver Declaration states:

Of the SAMA analyses I surveyed for BWRs, on average four cost-beneficial or potentially cost-beneficial SAMAs were found for each site, with a maximum of 11 cost-beneficial or potentially cost-beneficial SAMAs. Browns Ferry, Nine Mile Point and Peach Bottom had no cost-beneficial or potentially cost-beneficial SAMA candidates identified. Whether any of these cost-beneficial mitigation alternatives would be cost-beneficial at Limerick has not been determined, or even considered, in Exelons Environmental Report.

14. (DM) The Weaver Declaration characterizes these SAMAs only as cost-beneficial or potentially-cost-beneficial. It does not categorize them by whether they are procedural and programmatic in nature, minor design or hardware modifications, or major design or hardware modifications. The reason I use these categories is because the June 1996 rulemaking for 10 C.F.R. § 51.53(c)(3)(ii)(L) states that future SAMA analyses may 5

identify cost-beneficial improvements but that they generally would be procedural and programmatic fixes with any hardware changes being only minor in nature and few in number.2 In my opinion, procedural and programmatic fixes include changes to written operating procedures, staffing requirements, and personnel training.

15. (DM) The June 1996 rulemaking, on the same page, also states that: The Commission believes it unlikely that any site-specific consideration of severe accident mitigation alternatives for license renewal will identify major plant design changes or modifications that will prove to be cost-beneficial for reducing severe accident frequency or consequences.

16. (DM) I have reviewed the approximately 50 SAMAs in Paragraphs 11 of the Weaver Declaration of cost-beneficial or potentially cost-beneficial at other BWRs. All of these SAMAs are either: (a) procedural and programmatic fixes, such as changes to written procedures and operator training, or (b) minor design or hardware changes, such as adding portable equipment, cross-ties of existing systems, or adding cables. The minor design or hardware changes for any of those BWRs are few in number; the largest number of potentially cost-beneficial minor plant changes for any of the sites identified is five.

17. (DM) None of the 50 SAMAs identified in Paragraph 11 of the Weaver Declaration is a major design or plant modification. While there is not a commonly used definition for a major modification in the context of a SAMA analysis, I evaluated the SAMAs in paragraphs 11 of the Weaver Declaration assuming that a major modification is a plant change that results in the permanent installation of a new structure, system, or a 2

See Environmental Review for Renewal of Nuclear Power Plant Operating Licenses, 61 Fed. Reg. 28,467, 28,481 (June 5, 1996).

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redundant train of an existing system that changes the footprint of the facility. I have prepared a column labeled Major Modification? in Table A which explains why each of the SAMAs that NRDC identifies is not a major modification. Table A is attached to this Counter Affidavit. Using the same definition for major modification,

18. (DM) I also reviewed the SAMAs identified in paragraph 12 of the Weaver Declaration and determined that neither of those SAMAs are major modifications. The enhancement related to the RCIC control capabilities consists of the replacement of valves, control logic, and the addition of a low capacity generator to an existing system.

The proposed process to measure changes in safety related pipe wall thickness is a programmatic change.

B. The approximately 50 SAMAs listed in the Weaver Declaration are not new and significant for Limerick.

19. (DV) NRDC does not explain how the approximately 50 SAMAs listed in the Weaver Declaration, if implemented at Limerick, would present a seriously different picture of the environmental impact of plant operation. NRDC argues on page 7 of its Waiver Petition that to be significant, new information must present a seriously different picture of the environmental impact of the proposed project from what was previously envisioned. I am not aware of the NRC previously quantifying significance. However, I have reviewed of PRA standards and relevant guidance documents to develop a basis for what is significant in the SAMA context.

20. (DV) For the reasons stated below, I have selected a 50% reduction in the maximum averted cost-risk (MACR) as the threshold for what may be significant.

21. (DV) There are a few notable documents that provide numerical criteria that may be applied to determine the threshold for significance. The first one is the American Society 7

of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard3 which includes the following definition of a significant basic event.

significant basic event: a basic event that contributes significantly to the computed risks for a specific hazard group. For internal events, this includes any basic event that has an FV [Fussell-Vesely] importance greater than 0.005 or a RAW [Risk Achievement Worth] importance greater than 2.

Similar numerical criteria also appear in NUMARC 93-014, which includes the following guidance.

An SSC would probably be considered risk significant if its Risk Reduction Worth exceeds 0.5 percent of the overall Core Damage Frequency (Risk Reduction Worth >1.005).

[]

An SSC [structure, system or component] would probably be considered risk significant if its Risk Achievement Worth shows at least a doubling of the overall Core Damage Frequency and should be provided to the expert panel as an input in risk determination.

Finally, NEI 00-045 provides detailed guidance on categorizing structures, systems and components for licensees that choose to adopt 10 CFR § 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors. In the discussion of using risk analyses for SSC categorization, the following guidance is provided.

The risk importance process uses two standard PRA importance measures, risk achievement worth (RAW) and Fussell-Vesely (F-V), as screening tools to identify candidate safety-significant SSCs.

The criteria chosen for safety significance using these importance measures are based on previously accepted values for similar applications.

[]

3 Addenda to RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009, February 2009.

4 NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Rev. 2, April 1996.

5 NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, July 2005.

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The importance measure criteria used to identify candidate safety significance are:

• Sum of F-V for all basic events modeling the SSC of interest, including common cause events

> 0.005

• Maximum of component basic event RAW values > 2

22. (DV) In summary, an F-V value > 0.005 and a RAW value > 2 are well established indicators of PRA significance. This can be extended to apply to not just internal events core damage frequency (CDF) and large early release frequency (LERF), but to external events CDF and LERF, and other integrated key output figures of merit. In the context of license renewal, the accepted key output figure of merit for decision making is potential averted cost risk.

23. (DV) When averted cost risks are analyzed, the F-V importance measure would be highly dependent on the assumed reliability of the system once it is installed. This is illustrated in Figure 1 which shows an example of how the F-V value changes with assumed failure probability values given a case where a 50% reduction in the measured parameter is estimated assuming perfect reliability. In this example, a 0.005 F-V value would be obtained when the failure probability is ~0.005. This failure probability represents a system or component that is 99.5% reliable, which is fairly representative of many components modeled in typical PRA analyses.

24. (DV) On the other hand, as the reliability of the system increases (i.e., as the likelihood of system failure decreases), the RAW importance measure would asymptotically approach a RAW of 2 if 50% of the measured parameter can be averted. This is illustrated in Figure 2, which shows an example of how the RAW changes with assumed failure probability values when a 50% reduction in the measured parameter is estimated 9

assuming perfect reliability. Therefore, a correlation to a RAW > 2 as the acceptance threshold for significance is established and a 50% reduction in the MACR is chosen for the significance threshold.

25. (DV) In other words, the threshold I have just described would be equivalent to a highly reliable system leading to doubling the cost risk when it is taken out of service for maintenance. This correlates to a well-established threshold for determining risk significance in the PRA applications discussed above.

0.10 0.09 0.08 0.07 0.06 F-V Value 0.05 0.04 0.03 0.02 Better Reliability 0.01 0.00 1.00E+00 1.00E-01 1.00E-02 1.00E-03 1.00E-04 1.00E-05 Failure Probability Figure 1 F-V AS A FUNCTION OF FAILURE PROBABILITY 10

2.50 2.00 1.50 RAW Value 1.00 0.50 Better Reliability 0.00 1.00E+00 1.00E-01 1.00E-02 1.00E-03 1.00E-04 1.00E-05 Failure Probability Figure 2 RAW AS A FUNCTION OF FAILURE PROBABILITY

26. (DV) In summary, a RAW value of 2 is selected as the key acceptance criterion for determining the significance of a potential plant enhancement in this context. This is consistent with a well-established threshold for determining risk significance in various PRA applications, and correlates well with a 50% reduction in the MACR for a SAMA that is implemented. Therefore, a 50% reduction in the MACR is chosen for the significance threshold.

27. (DM) Based on the definition the definition of significance presented above, implementation of a SAMA would have to result in at least a 50% reduction in a plants MACR to be considered significant. Yet each of the SAMAs, identified in Paragraph 11 of the Weaver Declaration that NRDC claims are relevant to the review of new and significant information in the Limerick Environmental Report (ER), falls into one of four categories:

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1. Inapplicable to Limerick. For example, a SAMA in this category may apply only to containments or reactors of designs that are different than Limericks BWR Mark II design (Table A accompanying this Counter Affidavit shows these SAMAs as yellow-shaded). (On a related note, NRDC argues on page 27 of the Waiver Petition that Exelon must consider an example of an accident that is caused by operating the plant with auxiliary feed pumps that are closed for maintenance. This particular issue is not relevant to Limerick because the equipment identified in the example is specific to PWRs. Auxiliary feedwater pumps, which are part of a PWRs secondary side heat removal function, provide inventory makeup to the steam generators and are not used in BWR Mark II plants); or

2. Applicable to Limerick, but Exelon already has incorporated it at Limerick (these SAMAs are green-shaded on Table A); or

3. Applicable to Limerick and, although not incorporated at Limerick, Exelon has implemented a functional equivalent at Limerick (these SAMAs are orange-shaded on Table A); or

4. Applicable to Limerick and, although not incorporated at Limerick, the SAMA reduces the MACR in the plant of origin by less than 50% (these SAMAs are unshaded on Table A).

28. (DM) Given that none of the SAMAs are considered to be significant in the plants for which they were determined to be cost-beneficial, there is no evidence to suggest that the SAMAs would be significant for Limerick.

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29. (EK) Paragraph 12 of the Weaver Declaration also argues that Exelon be required to consider, as a SAMA, a specific method of inspecting buried, safety-related piping to identify aging related degradation. However, Limerick already performs a functionally equivalent inspection process as part of its Aging Management Program implemented under NUREG-1801, Generic Aging Lessons Learned (GALL) Report (Rev. 2) (Dec.

2010).

30. (DM) While the proposed process may be possible, paragraph 12 of the Weaver Declaration does not suggest that replacing the existing program with the proposed process would result in a reduction in risk for Limerick. Rather, NRDC only suggests that use of the alternative process may be less costly than excavating safety-related pipes.

C. Contention 1-E, as previously admitted by the Board, would have required Exelon to analyze SAMA candidates that are not unique to Limerick. The SAMAs specifically identified for consideration by Exelon in Paragraphs 11 and 12 of the Weaver Declaration are, by definition, not unique to Limerick.

31. (JG) Page 15 of the Waiver Petition states that NRDC Satisfies The Criteria For A Waiver Of 10 C.F.R. § 51.53(c)(3)(ii)(L) With Respect To Contention 1E As Admitted By The ASLB . . . . But Contention 1-E, as previously admitted by the Board, would have required Exelon to analyze SAMA candidates that are not unique to Limerick. The SAMAs specifically identified for consideration by Exelon in Paragraph 11 of the Weaver Declaration are, by definition, not unique to Limerick because they all were originally developed for other sites, not for Limerick. In fact, many of NRDCs proffered SAMAs come directly from the generic industry BWR SAMA list included in Table 13 of NEI 05-01 (which I helped author), or are functional equivalents of SAMAs listed in NEI 05-01. The SAMAs in paragraph 12 of the Weaver Declaration address issues that 13

are common to many BWRs in the industry and are similarly not unique to Limerick.

NEI-05-01 can be found at http://pbadupws.nrc.gov/docs/ML0605/ML060530203.pdf D. MACCS2 is used for Level 3 PRA.

32. (DM) Page 21 of the Waiver Petition argues that Exelon should be required to use updated analytical tools to perform an analysis of SAMA candidates. The only such tool that NRDC identifies by name is the MELCOR Accident Consequence Code Systems 2 (MACCS2). This is the code used for BWRs and PWRs that takes the amount of radionuclide released from the plant in the case of a severe accident (as determined from the Level 2 PRA) and computes the dose to the public and any land contamination impacts. Accordingly, MACCS2 is used in Level 3 PRA, not Level 1 or 2.

33. (DV) Exelon has used PRAs to evaluate severe accidents since the 1989 SAMDA analysis performed for Limerick. The Limerick PRA model has been periodically updated to reflect as-built and as-operated condition of the plant. The PRA models include input from the Individual Plant Examination (IPE) and individual plant examination for externally initiated events (IPEEE) evaluations.

34. (JG) I assisted Exelon with preparing the ER to support Limericks license renewal application to the NRC. I assisted Exelon with reviewing the current Limerick PRA model to identify any new information relative to the quantification of risk (measured in core damage events per year) in comparison to information provided in 1989 in the Supplemental Final Environmental Statement Related to the Operation of Limerick Generating Station, Units 1 and 2 (NUREG-0974). The process included a review of the NRCs Supplement to NUREG-0974 itself, the June 1989 Limerick PRA Update, and the Limerick PRA model and updates subsequent to the publication of the Supplement to 14

NUREG-0974 in 1989. These are discussed in the ER at pages 5-4 to 5-6. We specifically discussed these PRAs in the Limerick ER because they informed our decision-making about what needed to be included in response to NRC requirements in 10 CFR Part 51.

I, Jeffrey R. Gabor, declare under penalty of perjury that the foregoing information attributed to me as indicated by my initials at the start of the paragraph is true and correct to the best of my knowledge, information, and belief.

Executed in Accord with 10 C.F.R. § 2.304(d)

Jeffrey R. Gabor Vice President, Safety and Reliability ERIN Engineering and Research, Inc.

158 West Gay Street West Chester, PA 19380 (610) 431-8260 Executed on December 14, 2012 I, Donald E. MacLeod, declare under penalty of perjury that the foregoing information attributed to me as indicated by my initials at the start of the paragraph is true and correct, to the best of my knowledge, information, and belief.

Executed in Accord with 10 C.F.R. § 2.304(d)

Donald E. MacLeod Consultant I, Probabilistic Safety Assessment and Reliability ERIN Engineering and Research, Inc.

158 West Gay Street West Chester, PA 19380 (610) 431-8260 Executed on December 14, 2012 15

I, Donald E. Vanover, declare under penalty of perjury that the foregoing information attributed to me as indicated by my initials at the start of the paragraph is true and correct to the best of my knowledge, information, and belief.

Executed in Accord with 10 C.F.R. § 2.304(d)

Donald E. Vanover Vice President ERIN Engineering and Research, Inc.

158 West Gay Street West Chester, PA 19380 (610) 431-8260 Executed on December 14, 2012 I, Eugene Kelly, declare under penalty of perjury that the foregoing information attributed to me as indicated by my initials at the start of the paragraph is true and correct, to the best of my knowledge, information, and belief.

Executed in Accord with 10 C.F.R. § 2.304(d)

Eugene Kelley Senior Project Manager, License Renewal Exelon 200 Exelon Way Kennett Square, PA PA 19348 (610) 765-5554 Executed on December 14, 2012 16

Table A to Exelon's Counter Affidavit Supporting Exelon's Response Opposing NRDC's Petition for Waiver of 10 C.F.R. § 51.53(C)(3)(ii)(L)

List of Titles of SAMAs Found to be Cost-Beneficial or Potentially Cost-Beneficial Base Case Implemented at Limerick? Base Case Averted Percent Item from the Weaver Declaration (Paragraph Maximum Averted Comments Major Modification?

Cost Risk [Note 2] Change Nuclear Power Plant # 11) Cost Risk [Note 1]

Unit 1? (Y/N) Unit 2? (Y/N)

Limerick already has a portable AC generator No. Portable equipment and procedure Y Y and DC rectifier to open 3 SRVs without normal changes.

1 Portable DC generator station DC power.

No. Installation of different instrumentation or N N $9,588,000 $267,916 2.8% an alternate set of instrumentation.

2 Diverse EDG HVAC logic No. Temporary cable alignments.

N N $9,588,000 $1,566,562 16.3%

3 Provide alternate feeds to panels supplied only by DC bus 2A-1 Provide an alternate means of supplying the instrument air Limerick already has procedures to support this No. Portable equipment and procedure Y Y 4 header capability. changes.

No. Procedure change.

N N $9,588,000 $463,930 4.8%

5 Proceduralize battery charger high voltage shutdown circuit inhibit Existing Limerick procedures provide the ability No. Portable equipment and procedure Y Y to cross-tie existing fuel oil transfer pumps to changes.

6 Portable EDG fuel oil transfer pump non-dedicated EDGs Limerick already has procedures to support this No. Procedure change and potentially the Y Y capability. installation/modification of a pipe connection for Brunswick 7 Use fire water as a backup for containment spray some plants.

No. Change in actuation instrumentation type N N $1,886,578 $72,565 3.8% or an alternate set of instrumentation, or a 1 Reduce CCFs between EDG-3 and EDG1/2 change in fuel vendor.

Improve the fire resistance of cables to the containment vent No. Cables of a new type or installation of valve N N $1,886,578 $333,703 17.7%

2 cable wrap.

No. Cables of a new type or installation of Improve the fire resistance of cables to transformer E-TR-S N N $1,886,578 $75,446 4.0%

Columbia 3 cable wrap.

See Brunswick #1 No. Portable equipment and procedure Y Y 1 Portable generator for DC power to supply the individual panels changes.

Existing Limerick procedures provide this No. Procedure change.

capability (defeat all RCIC isolation signals Y Y including RCIC turbine exhaust diaphragm high Revise procedure to allow bypass of RCIC turbine exhaust pressure at 10 psig).

2 pressure trip LGS currently trains once every four years. No. Training update.

N N $1,053,957 $52,605 5.0%

3 Improve training on alternate injection via FPS No. Procedure change.

N N $1,053,957 $199,969 19.0%

Revise procedures to allow manual alignment of the fire water 4 system to RHR heat exchangers No. Procedure change.

Proceduralize the ability to cross-connect the circulating water N N $1,053,957 $177,788 16.9%

5 pumps and the service water going to the TEC heat exchangers The Limerick PRA indicates this SAMA would No. Procedure change and potentially the have a low benefit for the site. The importance installation/modification of a pipe connection for measures indicate the Feedwater/Condensate some plants.

system contributes to less than 0.5% of the internal events CDF and are negligible contributors to Seismic and Fire risk; this SAMA N N $1,053,957 $431,725 41.0% would yield a very small averted cost-risk.

Create ability for emergency connection of existing new water 6 sources to feed water and condensate systems Operator procedure revisions to provide additional space cooling No. Portable equipment and procedure N N $1,053,957 $33,160 3.1%

7 to the EDG room via the use of portable equipment changes.

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Table A to Exelon's Counter Affidavit Supporting Exelon's Response Opposing NRDC's Petition for Waiver of 10 C.F.R. § 51.53(C)(3)(ii)(L)

List of Titles of SAMAs Found to be Cost-Beneficial or Potentially Cost-Beneficial Base Case Implemented at Limerick? Base Case Averted Percent Item from the Weaver Declaration (Paragraph Maximum Averted Comments Major Modification?

Cost Risk [Note 2] Change Nuclear Power Plant # 11) Cost Risk [Note 1]

Unit 1? (Y/N) Unit 2? (Y/N)

Provide an alternate means of supplying the instrument air See Brunswick #4 No. Portable equipment and procedure Y Y 8 header changes.

Existing Limerick procedures provide this No. Procedure change.

Proceduralize the use of a fire pumper truck to pressurize the fire Y Y capability (use of fire truck to supplement ring 9 water system header).

No. Programmatic change.

N N $1,053,957 $394,444 37.4%

10 Generation Risk Assessment implementation into plant activities No. Procedure change.

Modify procedures to allow use of the RHRSW system without a N N $1,053,957 $110,566 10.5%

Cooper 11 SWBP Existing Limerick procedures provide this No. Procedure change and potentially the Y Y capability (cross-tie ESW / RHRSW) installation/modification of a pipe connection for 1 Provide an alternate source of water for the RHRSW/ESW pit some plants.

Increase the reliability of the low pressure ECCS RPV low No. Procedure change and update of the pressure permissive circuitry. Install manual bypass of low N N $2,261,022 $276,000 12.2% control panels to include a bypass switch.

Duane Arnold 2 pressure permissive Procedural change to cross-tie open cycle cooling system to See Brunswick #7 No. Procedure change.

Y Y 1 enhance containment spray system Existing Limerick procedures provide this No. Procedure change.

Enhance procedures to refill CST from demineralized water or Y Y capability (Fire Water supply to CST via hoses).

2 service water system Increase operator training for alternating operation of the low No. Training update.

pressure ECCS pumps (LPCI and LPCS) for loss of SSW N N $821,403 $40,452 6.8%

Grand Gulf 3 scenarios Enhanced DC Power Availability (provide cables from DG-13, the See Brunswick #1 No. Temporary cable alignments.

security diesel, or another source to directly power division II Y Y 1 250V battery chargers or other required loads)

Enhance Alternate Injection Reliability (include the RHRSW and LGS uses hoses, hose connections, and No. Programmatic change.

N N $8,642,000 $687,044 8.0%

2 FSW valves in the maintenance testing program) manual valves. Maintenance is N/A Existing Limerick procedures provide this No. Procedure change.

capability (use of fire truck to supplement ring Y Y header). In addition, procedures exist for using Additional Diesel Fire Pump for FSW system (proceduralize the a portable pump for Alternate RPV injection.

use of a fire truck to pressurize and provide flow to the fire main 3 for RPV injection)

Refill CST (develop emergency procedures and ensure viability of See Grand Gulf #2 No. Procedure change.

Y Y 4 refilling the CSTs with FSW)

No. This SAMA involves changing the swing direction of a door and the inclusion of an N N $8,642,000 $1,899,615 22.0% interlock to open a door on high water level.

5 Divert Water from Turbine Building 931-foot elevation Existing Limerick procedures provide this No. Procedure change.

Y Y capability.

Monticello 6 Manual RCIC Operation Existing Limerick procedures provide this No. Procedure change with potential Y Y capability. interlock/equipment modifications on existing busses.

1 Allow 4160 VAC bus IC and ID cross-tie Not applicable to the Limerick design. Limerick No. Modification to existing sensors.

N N does not have an isolation condenser.

2 Provide an alternate method for IC shell level determination Portable DC battery charger to preserve IC and EMRV operability No. Portable equipment and procedure N N $4,462,000 $674,000 15.1%

3 along with adequate instrumentation changes.

No. Installation of cable wrap, new cables, or N N $4,462,000 $333,000 7.5%

4 Reduce fire impact in dominant fire areas fire barriers.

5 Operator Training N N $4,462,000 NA No. Training update.

Not applicable to the Limerick design. Limerick No. Reinforcement of an existing structure.

does not have combustion turbines, but the N N $4,462,000 $747,000 16.7% SAMA could potentially be extrapolated to address the EDG building.

6 Protect Combustion Turbines Oyster Creek 7 Upgrade Fire Pump House structural integrity N N $4,462,000 $438,000 9.8% No. Reinforcement of an existing structure.

2

Table A to Exelon's Counter Affidavit Supporting Exelon's Response Opposing NRDC's Petition for Waiver of 10 C.F.R. § 51.53(C)(3)(ii)(L)

List of Titles of SAMAs Found to be Cost-Beneficial or Potentially Cost-Beneficial Base Case Implemented at Limerick? Base Case Averted Percent Item from the Weaver Declaration (Paragraph Maximum Averted Comments Major Modification?

Cost Risk [Note 2] Change Nuclear Power Plant # 11) Cost Risk [Note 1]

Unit 1? (Y/N) Unit 2? (Y/N) 1 Enhance procedures to make use of AC bus cross-ties Y Y See Oyster Creek #1 No. Procedure change.

LGS cross-ties exist for 4 kV AC emergency No. Procedure change.

buses only.

N N $914,294 $19,761 2.2%

2 Enhance procedures to make use of DC bus cross-ties No. Additional fuses added to an existing electrical panel.

N N $914,294 $36,773 4.0%

3 Provide redundant DC power supplies to DTV valves No. Procedure change.

Proceduralize use of the diesel fire pump hydro turbine in the N N $914,294 $29,213 3.2%

4 event of EDG A failure or unavailability Proceduralize the operator action to feed B1 loads via B3 when No. Procedure change.

A5 is unavailable post trip; similarly, feed B2 loads via B4 when N N $914,294 $31,799 3.5%

Pilgrim 5 A6 is unavailable post-trip See Oyster Creek #1 No. Procedure change with potential Improve cross-tie capability between 4kV AC emergency buses Y Y interlock/equipment modifications on existing 1 (A-D, B-C) busses.

No. Portable equipment and procedure Procure spare 480V AC portable station generator N N $538,000 $132,799 24.7%

Susquehanna 2 changes.

Shield injection system electrical equipment from potential water No. Barrier over or around existing equipment.

spray N N $523,269 $26,000 5.0%

1 Improve operator action: defeat low reactor pressure interlocks No. Procedure and/or training enhancements.

to open LPCI or core spray injection valves during transients with stuck open SRVs or LOCAs in which random failures prevent all N N $523,269 $142,000 27.1%

2 low pressure valves from opening No. Procedure change and update of the control panels to include a bypass switch.

N N $523,269 $142,000 27.1%

Install a bypass switch to bypass the low reactor pressure Vermont Yankee 3 interlocks of LPCI or core spray injection valve Note 1: This information was derived from originating site's Environmental Report. In some cases, it was necessary to calculate a MACR value based on the documented external events multiplier to ensure consistency with the averted cost-risk values provided in the Environmental Report.

Note 2: Obtained from the originating site's Environmental Report.

Not Applicable to the LGS Design Implemented Functionality addresses by other means at LGS.

3

JEFF GABOR WORK EXPERIENCE

SUMMARY

Mr. Gabor, Vice President of the Risk Management Group for Vice President ERIN Engineering and Research, Inc., is a Mechanical Engineer with considerable experience in the field of nuclear plant Safety and Reliability thermal-hydraulic and severe accident analysis. He has over 30 years of experience in Nuclear Power Plant Safety. Mr.

Gabor is the Manager of ERINs West Chester, PA office.

AREAS OF EXPERTISE WORK EXPERIENCE

• Plant Thermal-Hydraulic Mr. Gabor has been involved in several Level 2 PSA updates and Response continues to support a variety of severe accident and thermal-hydraulic analysis activities at numerous utilities. The following are some of his recent activities:

• Severe Accident Analysis

• Analysis on mitigation of radiological releases in severe accidents

• Level 2 PRA (EPRI 1026539).

• Update of Technical Basis Report.

• Severe Accident Management

• Accident Reconstruction using MAAP5 for Fukushima Dai-ichi Units 1,2 and 3 - EPRI Fukushima Technical Evaluation.

• Plant Modeling

• Severe Accident Mitigation Alternative (SAMA) analysis in support of license renewal for over half of all plant submittals.

• Thermal-Hydraulic and

• Member of MSPI PRA Quality Task Group.

Severe Accident Training

• MAAP Users Group Steering Committee Chairman.

• Integrated Leak Rate Testing Extension for Fermi and Farley.

• MAAP4 Implementation and Analysis for entire Exelon fleet.

• Technical Support in the Risk Evaluation of the Shearon Harris EDUCATION Spent Fuel Pool Expansion before the ASLB.

• Salem PSA Update Project.

M.S., Mechanical Engineering, University of

• Containment response analysis to support EQ evaluation at Cincinnati, Ohio Cooper Nuclear station. Work included detailed thermal-lag analysis of key components.

• Level 2 PSA updates at Palo Verde, Vogtle, Palisades, Quad Cities, B.S., Nuclear Engineering, LaSalle, Vermont Yankee, Brunswick, and Browns Ferry.

University of Cincinnati, Ohio

• Development of Technical Support Guidelines at Clinton Power Station, Duane Arnold Energy Center, WNP2 and Fermi.

• Lead technical analyst for implementation of MAAP4 at Nine Mile Point Unit 2, River Bend, Cooper, and WNP2.

• Manager and lead technical analyst for MAAP4 analysis in support of EdF PWR Level 2 PSA.

• Severe Accident Training at DAEC, WNP2, and Cofrentes.

• Lead Thermal-hydraulic analyst in support of Quad Cities PSSA.

As an associate with Dames & Moore, Mr. Gabor was responsible for JEFF GABOR resource development, strategic planning, and technical oversight for all nuclear activities carried out in the Westmont, Illinois office. He Page 2 worked with nuclear utilities in addressing issues related to plant thermal-hydraulic response.

• Lead technical analyst for the Garona Level 2 PRA for Nuclenor (Spain).

• Lead technical analyst for the Cofrentes Level 2 PRA for Iberdrola (Spain).

• Technical support to the Consumers Power Big Rock Point Nuclear Plant on issues related to plant thermal-hydraulic response, severe accident analysis, and equipment qualification.

• Development of MAAP4 parameter file and severe accident training for Cooper Nuclear Station.

• Implementation of BWROG Technical Support Guidelines for the Cofrentes Nuclear Station.

• Development and Implementation of Remote Monitoring for soil vapor extraction remediation system.

• Technical analyst for severe accident investigations in support of certification of advanced light water reactor designs.

Development of computer simulation tools and presentations to USNRC and the ACRS concerning severe accident behavior. This work was performed under contract with the Department of Energy.

• Member of a GE design review committee for the evaluation of the impact of Nobel Metal Chemical Addition on the containment atmospheric monitoring systems.

Vice President and Co-founder of Gabor, Kenton & Associates, Inc.

• Technical support for Level 2 PRA on the General Electric Advanced Boiling Water Reactors. Included numerous technical presentations to USNRC, ACRS, and USDOE.

• Lead technical analyst for severe accident response on a number of BWR Level 2 PRAs:

- Millstone Unit 1

- Duane Arnold

- Pilgrim

- Nine Mile Point Units 1 and 2

- Fermi

- Vermont Yankee

- Cofrentes

- Browns Ferry

• Cooper

• Lead BWR analyst for EPRI sponsored MAAP 3.0B Thermal-Hydraulic Qualification Study.

• Independent Review of MAAP 3.0B and MAAP 4 maintenance activities.

• Developed and managed Gabor, Kenton, & Associates Quality Assurance Program.

• Provided technical support for the containment vent evaluation of JEFF GABOR the Cofrentes and Garona plants (Spain). Performed vent sizing calculations and on-site radiation dose assessment.

Page 3 Manager of Plant Analysis and Special Projects for Fauske &

Associates, Inc.

• Severe accident evaluations of Grand Gulf and Peach Bottom in support of IDCOR.

• Pilgrim Safety Enhancement Program.

• BWR Owners Group Mark I Evaluation.

• Caorso Severe Accident Analysis.

• Mark I shell melt-through analysis and experiments.

• Swedish Reactor Accident Mitigation Analysis (RAMA).

• Shoreham PRA.

• Vermont Yankee 60 Day Study.

• Empire State Electric Energy Research Corporation LWR Code Comparison.

• Managed and participated in MAAP and severe accident phenomenology training courses for nuclear industry along with numerous presentations to the NRC and ACRS on severe accident phenomenology.

• Designated Westinghouse Expert Engineer in severe accident thermal hydraulic transient analysis.

System Engineer for Cincinnati Gas and Electric Company

• Implemented post-TMI design changes.

• Conducted pre-operational testing program.

• Completed RETRAN training program.

Resident Student Associate at Argonne National Laboratory

• Experimentation on the transition from film boiling to nuclear boiling on a flat plate.

• Computer analysis of LMFBR core design.

Mr. Gabor's engineering experience was gained through employment with the following companies:

• ERIN Engineering and Research, Inc.

• Dames & Moore.

• Gabor, Kenton & Associates, Inc.

• Fauske & Associates, Inc.

• Cincinnati Gas and Electric Company.

• Argonne National Laboratory.

DONALD MACLEOD WORK EXPERIENCE

SUMMARY

Mr. MacLeod has over fifteen years of experience with ERIN Consultant I Engineering specializing in Probabilistic Safety Assessment.

Probabilistic Safety Assessment His specific experience includes Human Reliability Analysis for and Reliability full power and shutdown conditions, Severe Accident Mitigation Alternative Analysis, and systems analysis.

AREAS OF EXPERTISE WORK EXPERIENCE

• Human Reliability Analysis Mr. MacLeod holds a Bachelor of Science degree in Nuclear Engineering from Rensselaer Polytechnic Institute. Relevant work

• On-Line Maintenance experience includes the following:

• Served as the lead analyst in the development of the revised

• Probabilistic Safety Peach Bottom, Limerick, Salem, Susquehanna, Quad Cities and Assessment Dresden post-initiator HRAs as part of PSA enhancement and support projects.

• Shutdown Safety

• Performed the Fire HRA for Farley Nuclear Plant.

• Performed External Flooding HRA for TMI and Internal Flooding

• Fault Tree Analysis HRA for Byron/Braidwood.

• Event Tree Analysis

• Performed pre-initiator analyses for RNP, Palisades, SSES, Byron/Braidwood, and St. Lucie.

• SAMA Analysis

• Co-developed the Human Reliability Analysis for the Lungmen PRA. This project required detailed analysis of plant functions, design, procedures, the PRA model, and interface with system engineers for the Lungmen plant.

EDUCATION

• Performed shutdown HRAs for the following sites; LaSalle, Duane Arnold, Quad Cities, Zion, Dresden, Cooper, Peach Bottom, and B.S., Nuclear Engineering, Fermi 2. These analyses included an assessment of the clarity Rensselaer Polytechnic and completeness of the procedures, quantification of HEPs, and identification of any potential procedural improvements.

Institute

• Updated and revised the NMP-2 data analysis as part of the PRA update program. Compiled a plant specific failure database, recalculated component failure probabilities and system maintenance unavailabilities, and requantified the common cause SECURITY CLEARANCE failure analysis using INEL-94/0064.

• Served as a member of the Columbia Generating Station PRA U.S. Citizen certification team.

• Served as the lead analyst in the performance of the Severe Accident Mitigation Alternative (SAMA) analyses for TMI, HNP, Wolf Creek, V.C. Summer, Brunswick, H.B. Robinson, Monticello, Palisades, Susquehanna, STP, and Palo Verde. Co-developed the Peach Bottom, SGS, Hope Creek, DCPP, and Crystal River SAMA analyses.

• Participated in the development of Probabilistic Shutdown Safety Assessments (PSSAs) for the LaSalle, Duane Arnold, Quad Cities, Zion, Dresden, Cooper, Peach Bottom, and Fermi 2 plants using the Outage Risk Assessment and Management (ORAM) code.

Assisted in the adaptation and application of the EPRI Cause-Based method for use in PSSA Human Reliability Analysis.

• Conducted extensive model quantifications for several U.S.

utilities, using both the NUPRA and CAFTA software packages.

In summary, Mr. MacLeod is a PRA expert in the following areas:

DONALD MACLEOD

• Data Synthesis Page 2

• Data Evaluation

• Fault Tree Development

• Quantification

• HRA

• Shutdown Model Development

• On-Line Maintenance Model Quantification

• SAMA Analysis

DONALD VANOVER WORK EXPERIENCE

SUMMARY

Mr. Vanover is a technical manager with over twenty years of Vice President experience serving the commercial nuclear power industry.

While at ERIN Engineering and Research, Inc., Mr. Vanover has completed a variety of tasks including numerous at-power PRA model updates, analyses, applications, and relief requests, as well as Probabilistic Shutdown Safety Assessments for both BWR and PWR Plants. Additionally, he has acted as project manager for several Software Development projects. Before joining ERIN Engineering, Mr.

Vanover was extensively involved in the completion of numerous Individual Plant Examinations (IPEs) both in the U.S. and in Spain. Prior to that, he assisted with the testing, development, and benchmarking of many of the models contained in the Modular Accident Analysis Program (MAAP).

AREAS OF EXPERTISE WORK EXPERIENCE Mr. Vanover is presently a Vice President at ERIN Engineering and

• Probabilistic Risk Research, Inc. in the Philadelphia area office. He has extensive Assessment knowledge of systems and procedures for both BWR and PWR reactors and has the ability to apply his hands-on and theoretical

• PRA Applications background to understand and analyze system response. Since joining ERIN Engineering in 1995, Mr. Vanover has contributed to the successful completion of several PRA-related projects, including major

• Configuration Risk updates to the Peach Bottom and Limerick PRA models, and the Management Modeling completion of several shutdown PRA models for both BWR and PWR plants. He has also performed seismic and fire probabilistic risk

• Thermal-Hydraulic analyses for Peach Bottom and Limerick, and has been extensively Analysis involved in the use of the Peach Bottom and Limerick PRA models to support on-line maintenance activities. Mr. Vanover has also been a

• Project Management major contributor and project manager on several other PRA application projects.

While with Gabor, Kenton, and Associates, Inc., Mr. Vanover performed, documented, and supplied technology transfer of engineering and thermal-hydraulic analysis for numerous nuclear EDUCATION utilities for their Individual Plant Examination submittals to the NRC (or equivalent). He also served as an instructor for severe accident M.S., Mechanical phenomenology and Modular Accident Analysis Program (MAAP)

Engineering, University of training courses both in the U.S. and abroad.

Delaware Mr. Vanover conducted thermal-hydraulic analyses for use in the industry-sponsored initiative for Steam Generator Alternate Repair B.S., Mechanical Criteria and provided station blackout analyses for the Salem plant in Engineering, University of support of PSE&G's response to NUMARC 87-00. He developed MAAP Delaware source code modifications for the Trillo and Zorita plants in Spain.

Additionally, he was responsible for evaluating available data to perform numerous comparisons and sensitivity runs with both the BWR and PWR versions of MAAP.

At Fauske & Associates, Inc., Mr. Vanover successfully headed the development of new models and organized an entire code to provide SECURITY CLEARANCE a severe accident analysis tool for CANDU reactors. Prior to that, he performed benchmark calculations for thermal-hydraulic transient U.S. Citizen analysis using the MAAP BWR and PWR codes.

Before attending graduate school, Mr. Vanover worked as a DONALD VANOVER Mechanical Maintenance Supervisor for Bethlehem Steel Corporation at the Sparrows Point, Maryland plant. There he Page 2 assigned daily tasks to millwrights, pipefitters, and welders, and was responsible for maintaining pumps, compressors, condensers, heaters, and turbines for an on-site power plant.

PROFESSIONAL ORGANIZATIONS American Nuclear Society

Eugene M. Kelly 200 Exelon Way Kennett Square, PA 19348 Email: gene.kelly@exeloncorp.com Qualifications:

Over 30 years experience in the commercial nuclear industry encompassing design, analysis, licensing, regulation and operation. In-depth working knowledge of NRC rules, licensing procedures, codes and standards, performance assessment and risk evaluation.

Exceptional organizational spokesperson in industry seminars, public meetings, and briefings of senior leadership. Extensive experience with the application of system engineering, and management of technical programs. Team leader for multi-disciplinary investigations involving complex problem solving. Recognized expert in application of 10CFR50.59 safety evaluation process, design basis assessment, license renewal concepts, and risk management initiatives.

Education:

1968-1972 BS Physics, Villanova University 1975-1979 MS Mechanical Engineering, University of Pennsylvania Employment:

2009- Present: Senior Project Manager, License Renewal, Exelon Technical lead responsible for project to re-license the Limerick Nuclear Generating Station. Supervise 10-person multi-disciplinary engineering team.

2007-2008: Senior Manager, Design, Limerick and Oyster Creek Stations, Exelon Manage interactions with and successful response to WANO Evaluation, champion margin management initiatives, oversee preparation and execution of refueling outage commitments, chair Quality Review Team and Design Subcommittee, and coordinate CDBI readiness and execution.

2002-2006: Manager, Engineering Programs, Exelon Nuclear, Limerick Station Oversee engineering programs including risk management, Maintenance Rule, fire protection, ISI and IST, reactor vessel internals, FAC and heat exchangers, thermal performance, leak rate testing, and valve reliability (MOV, AOV, Check Valve, MSIV).

Responsible for $6.6M annual budget. Chairman of INPO Working Group on Engineering Programs Excellence. Project responsibility for two innovative risk-informed industry pilot initiatives on PRA model quality and technical specification surveillance frequency extension (SFCP).

Resume

1999-2001 Manager, Electrical Plant Systems, Exelon Nuclear, Limerick Station Responsible for the performance of electrical systems including eight emergency diesel generators, 220 and 500 kV switchyards, a large DC battery distribution network, ventilation and fire protection, security systems and reactor protection instrumentation.

Coordinated preventive maintenance, special testing, failure casual analysis, vendor interface and modification improvements. Recruit, train and qualify new system managers and design engineers. Instituted process improvements in engineering work management. Chairman of Maintenance Rule Expert Panel and member of Plant Operating Review Committee.

1994-1998 Manager, Systems Engineering Branch, USNRC, Division of Reactor Safety, King of Prussia, PA Responsible for assessment of engineering programs at 20 nuclear reactor sites throughout the Northeast. Manage engineering projects and specialist inspectors in areas including motor operated valves, service water, in-service testing, core physics and mechanical systems. Special projects include complex team inspections (e.g. SSFI),

event follow up, design basis investigations and the Millstone Task Force. Agency spokesperson for inspection program. Developed risk-based approaches for inspection.

Recruit, train and qualify new inspectors.

1991-1994 Reactor Projects Chief, USNRC Managed field offices and supervised resident inspectors at eight sites including Millstone, Haddam Neck, Rowe, Pilgrim and Vermont Yankee. Project management included coordination of Congressional correspondence, enforcement actions and performance assessment reports. Organized and participated in high visibility public meetings and briefings of elected officials and NRC executive management.

1988-1990 Technical Support Staff Chief, USNRC Developed nationwide Master Inspection Planning System and core inspection program, including institution of budget analysis and new technical initiatives. Managed diagnostic teams, generic issue follow-up and integration of risk assessment techniques.

Streamlined senior management problem plant briefing process by creation of plant status reports. Coordinated industry conferences on regulatory topical areas.

1985-1988 Limerick Senior Resident Inspector, USNRC Managed NRC field office including two residents and a clerical aide. Supervised detailed inspections of design, test, maintenance, and event follow-up. Coordinated inspection oversight for startup and power ascension programs on one unit and completion of construction activities at the other. Primary author of Systematic Assessment Performance (SALP) Report for first commercial year of Limerick operation.

Resume

1982-1984 Reactor Engineer, USNRC Conducted inspections of construction, preoperational and startup testing at regional sites. Specialized training and qualification on General Electric, Westinghouse and Combustion Engineering plants. Special projects included engineering evaluations at all Yankee sites, and follow-up of employee concerns at Shoreham. Created unique NTOL assessment technique to support operating license decisions for five units.

1980-1982 Systems Engineer, Catalytic, Inc., Philadelphia, PA Developed plant modifications for clients including design specifications, detailed engineering and calculations, coordination of procurement, test and field installation.

1979-1980 Nuclear Engineer, GPU Nuclear Corporation, Middletown, PA Responsible for radioisotope analysis, shielding calculations, Krypton venting evaluations and containment sump water sampling at Three Mile Island site following the accident.

1974-1979 Safety Analysis Engineer, United Engineers and Constructors, Inc.

Advanced Engineering Department, Philadelphia, PA Prepared Safety Analysis Reports for six nuclear projects including HTGR and conventional designs. Performed thermal hydraulic studies, radiological dose and shielding calculations and system performance analyses. Developed a heat transfer model for an ultimate heat sink spray pond.

1973-1974 Physics, Mathematics Teacher, Monsignor Bonner HS, Upper Darby, PA Resume