ML20151P195

Affidavit of Kw Holtzclaw Re Five Proposed Contentions Covering Single Loop Operation of Recirculation Sys Contained in 851212 Motion to Reopen Record.Statement of Prof Qualifications & Certificate of Svc Encl

ML20151P195
Person / Time
Site:	Perry
Issue date:	12/30/1985
From:	Holtzclaw K GENERAL ELECTRIC CO.
To:
Shared Package
ML20151P165	List: ML20151P161 ML20151P195
References
OL, NUDOCS 8601030324
Download: ML20151P195 (21)

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December 30, 1985 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

< Before the Atomic Safety and Licensino Appeal Board In the Matter of )

)

THE CLEVELAND ELECTRIC ) Docket Nos. 50-440 ILLUMINATING COMPANY, ET AL. ) 50-441

)

(Perry Nuclear Power Plant, )

Units 1 and 2) )

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-AFFIDAVIT OF KEVIN W. HOLTZCLAW State of California )

ss:

County of Santa Clara )

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Kevin W. Holtzclaw, being duly sworn, deposes and says as follows:

1. I,IKevin W. Holtzclaw, am a principal licensing engi-l neer in the General Electric ("GE") Safety and Licensing Opera-tion of the Nuclear Energy Business Operation. My business address is 175 Curtner Avenue, San Jose, CA 95125.

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2. A statement of my professional qualifications is attached to this Affidavit as Exhibit "A". I have spent nearly 17 years as an engineer in the nuclear power industry, and have B601030324 851230 PDR ADOCK 05000440 0 PDR

worked for GE since 1969. I have spent over 12 years in GE

! Boiling Water Reactor ("BWR") fuel and core design and analy-sis. During this time I have performed numerous thermal and thermal-hydraulic evaluations for steady-state and anticipated transient conditions. As part of my work, I have evaluated thermal and thermal-hydraulic analyses of single loop operation

(" SLO") of the recirculation system of BWRs.

3. In this Affidavit I respond to the five proposed con-tentions (Contentions "B-1" through "B-5") covering single loop operation of the recirculation system of the Perry Nuclear Power Plant, contained in the Motion to Reopen the Record and to Submit New Contentions (December 12, 1985) (" Motion") of Ohio. Citizens For Responsible Energy ("OCRE"). For the reasons discussed in this Affidavit, the contentions are without tech-nical merit. I am familiar with the technical information and references included in Appendix 15F (PNPP Single Loop Operation Analysis) to Amendment 22 to the Perry Final Safety Analysis Report ("FSAR"), dated November 20, 1985 (" Appendix 15F"),

which OCRE's motion addresses. (Information identical to that contained in} Appendix 15F had been earlier transmitted to NRC by letter from Cleveland Electric Illuminating Company dated October 28, 1985.) I am also familiar with other portions of the Perry FSAR which deal with issues OCRE has raised, as dis-cussed in this Affidavit. I have personal knowledge of the

! matters set forth in this Affidavit and believe them to be true and correct.

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4. Under the current Perry Nuclear Power Plant Unit 1 Technical Specifications (" Technical Specifications"), the Perry Plant is permitted to operate with a single reactor coolant system recirculation loop for a limited period of time (a maximum of 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />). See Technical Specifications, S 3/4.4.1.1.a. Under the current Technical Specifications, in the event that one recirculation loop is not in operation, in-mediate action must be taken to reduce power from the level the plant would be at in SLO (as high as 70%) to a lev '. as high as 53%, within two hours. The Technical Specifications then per-mit the operators to remain in SLO at power levels up to 53%

before putting the unit in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. These provisions have been included in the Perry Tech-nical Specifications since they were first docketed in mid-1984.

5. If approved by the NRC, Appendix 15.F of the FSAR would permit extended SLO of Perry up to 70% of rated thermal power. Seventy percent represents the approximate maximum power limit capability of Perry's BWR-6 reactor under SLO. GE has performed extensive generic BWR analyses, and Perry-specific analyses, which support the conclusions in Appendix 15F. These analyses demonstrate that Perry can safely operate with a single recirculation loop. See Appendix 15.F, pp.

15.F.1 15.F.1-2; 15.F.8 15.F.8-2 (References).

6. None of OCRE's proposed contentions on SLO provides a l- credible, technically-justified basis to question the analyses and conclusions set forth in Appendix 15F, which amply supports the conclusion that SLO is safe for Perry. Thus, none of l

OCRE's contentions raises a significant safety issue. I will address each of OCRE's contentions, and the arguments in sup-port of the contentions, in the following paragraphs of this Affidavit.

CONTENTION B-1

7. Contention B-1 of OCRE's motion states: ,

Applicants should analyze the progression and consequences of an anticipated tran-sient without scram ("ATWS") initiated by the inadvertent startup of the idle recir-culation loop when operating at 70% of rated thermal power with single loop opera-l tion. The analysis should demonstrate that l this event will meet the safety criteria L

outlined in Section 15C.3 of the FSAR.

Motion at 2-3. The contention is without basis for the follow-

! ing reasons.

8. As stated in S 15.F.3.1 of Appendix 15F: "Inadver-tent restart of the idle recirculation pump has been analyzed

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.in the FSAR and is still applicable for single-loop operation."

, Appendix 15F, p. 15.F.3-2.

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! 9. The idle recirculation loop startup transient is dis-cussed in 5 15.4.4 of the Perry FSAR. The transient analyzed in 5 15.4.4 begins with reactor power at 54%. See Figure 15.4-1 (neutron flux curve at time zero). For this transient, the analysis demonstrates that scram is not required as a pro-tective action. The transient response shows that no damage occurs to the fuel barrier and that the Minimum Critical Power Ratio ("MCPR") (the ratio of critical fuel bundle power to operating fuel bundle power, used to monitor the potential onset of boiling transition which could result in fuel damage) remains above (i.e., within) the applicable safety limit. See S 15.4.4.3.3. Therefore, ATWS analyses are irrelevant at power levels up to 54%.

10. If the idle recirculation loop startup transient is initiated at a thermal power' level above 54%, conservative li-censing basis calculations predict that Average Power Range Monitor ("APRM") flux scram would occur. FSAR S 15.4.4.2.3.

Therefore, ATWS evaluations are relevant for this transient at power levels above 54%.

11. OCRE states at page 3 that the FSAR does not analyze an ATWS event initiated by the inadvertent startup of an idle recirculation loop. This ATWS event is not analyzed in the FSAR because.it is less severe than events which are analyzed there. For the same reason, this event is not included in NEDE-25518 (referenced by OCRE at page 3), the Perry-specific analysis of ATWS events performed by GE.

12. The selection of ATWS events to be analyzed in NEDE-25518 and in the FSAR was based on prior GE generic O

evaluations, including Volumes I and II of NEDO-24222 (Assess-ment of BWR Mitigation of ATWS (NUREG 0460 Alternate No. 3))

(1981). Volume II of NEDO-24222 concluded, for a BWR-6 such as Perry, that the idle recirculation loop startup event was less severe than the recirculation flow controller failure with increasing flow event, which in turn was less severe than the main steam isolation valve closure or turbine trip ATWS cases.

NEDO-24222, vol. II, SS 3.3.12, 3.3.13. Perry-specific analy-ses of these four transients show the same relative severity based on neutron flux and heat flux predictions, with the idle recirculation loop startup event (FSAR S 15.4.4) and recircula-tion flow controller failure (FSAR S 15.4.5) being much less severe than the turbine trip (FSAR S 15.2.3) and main steam isolation valve events (FSAR S 15.2.4).

13. The idle recirculation loop startup event discussed in NEDO-24222 and FSAR S 15.4.4 was based on power levels of less than 70%. However, the event at a 70% power level is less

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severe than at the levels assumed in NEDO-24222 and FSAR J

S 15.4.4.

The reason why the event starting at a higher power level is less severe is that the power rise from beginning to

peak is less for the 70% case. The relative core flow change that would be caused by the idle pump startup at 70% power is smaller because the reactor is already operating at a higher core flow than it would be at lower power levels. The evalua-tions in NEDO-24222 are bounding (and were chosen because they

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were bounding), and therefore the idle recirculation loop startup event at 70% power need not be analyzed for ATWS.

14. NEDO-10349, referenced by OCRE at page 3 of the Mo-tion, is a 1971 analysis based on earlier GE BWR models. How-ever, *;he conclusion in NEDO-10349, that the idle recirculation loop startup transient is less severe than recirculation flow controller malfunction - increasing flow ATWS, which OCRE cites in its motion, is entirely consistent with GE's later transient analyses for BWR-6 models such as Perry, as discussed above.

15. For these reasons, further ATWS analyses based on recirculation loop startup are not required to demonstrate the safety of SLO at Perry.

CONTENTION B-2

16. Contention B-2 of OCRE's motion states:

Applicants have not demonstrated that the seizure of the operating recirculation pump when operating up to 70% of- rated thermal power with a single loop will not exceed fuel safety limits, assuming scram func-tions, and that ATWS initiated by this event will meet the safety criteria of FSAR Section 15C.3.

Motion at 3-4. The contention is without basis for the follow-ing reasons.

17. As indicated in S 15.F.3.1 of Appendix 15F: "The one recirculation pump seizure accident has been reviewed for sin-gle loop operation. Results show that this accident poses no threats to thermal (i.e., fuel safety] limits." As part of

l this review, GE specifically analyzed the recirculation pump l seizure accident for thermal power levels up to 70% power in SLO at Perry. The results of GE's analysis show that no safety limits were exceeded.

18. OCRE's first argument is that steam blanketing of the fuel rods would occur if a recirculation pump seizure accident occurred in SLO. GE's analysis shows that steam blanketing of the fuel rods (i.e., formation of a steam layer around the fuel rods sufficient to prevent liquid from cooling the fuel cladding) would not occur in the unlikely event of recircula-tion pump seizure. I have reviewed page 32 of Richard Webb's "The Accident Hazards of Nuclear Power Plants" (1976), cited by >

OCRE at page 4 of its motion. The cited portion of Mr. Webb's book does not refer in any way to SLO. Based on my extensive experience in fuel and core design and analysis, I know of no technical basis with respect to BWR's for Mr. Webb's unsupported assertion that steam blanketing could occur and threaten fuel rod breakup following a hypothetical instanta-ncous coolant pump seizure.

19. OCRE's second argument (pages 4-5) is that a pump seizure in SLO would put the reactor into natural circulation, that neutron flux oscillations might then occur, and that an ATWS might then take place which would prevent operators from inserting control rods to suppress the oscillations. GE's analysis of the pump seizure accident during SLO for Perry,

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discussed in 1 17, demonstrated that no scram is required fol-loving such an accident. The scram at 3.4 seconds into the pump seizure accident (cited by OCRE at p. 4) only occurs dur-inj normal two loop operation and would not occur in SLO.

Thus, contrary to;OCRE's contention, ATWS is not a concern for the SLO pump seizure accident postulated by OCRE.

20. Aside from the fact that ATWS is irrelevant to an SLO pump seizure accident, OCRE's argument fails to recognize the difference between the scram function and manual control rod insertion capability. The neutron flux oscillations postulated by OCRE are suppressed by manual control rod insertion, not by, the scram function. There is no technical justification or basis to postulate loss of all control rod insertion capability in addition to postulating the highly improbable pump seizure accident, as OCRE's contention would require. Control rod in-sertion capability is assured at Perry through a number .o1 plant systems which are designed to meet single-failure, ,

safety-grade requirements. Thus, even if a pump seizure acci-dent were to occur, the control rods could be inserted which would suppreds any resulting oscillations.

21. GE analyses (including recommended operational guidelines adopted for Perry) also demonstrate that potential neutron flux oscillations would not result in Perry fuel design limits being exceeded in SLO. NEDE-22277, " Compliance of Gen-eral Electric Boiling Water Reactor Fuel Designs to Licensing

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Criteria" (October 1984) and App. 15F, S 15.F.4. The NRC ac-cepted the GE analysis for referencing on April 24, 1985 (sub-sequent to the March 1984 Board Notification relied upon by OCRE) (see App. ISF, ref. 15.F.8-6).

22. For these reasons, further analyses of pump seizure during SLO for Perry are not necessary to assure that the safe-ty criteria of FSAR Section 15C.3 would be met.

CONTENTION B-3

23. Contention B-3 of OCRE's motion states:

Applicants have not demonstrated that the traversing incore probe ("TIP") noise un-certainty-values reported in the FSAR Sec-tion 15.F.2.2 are applicable to single loop operation up to 70% of rated thermal power; consequently, the minimum critical power ratio ("MCPR") may not be determined in a conservative fashion.

Motion at 5. The contention is without basis for the following reasons.

24. The TIP noise value, referenced in this contention, is a random variation of the measured reading of the reactor local power level. TIP reading uncertainty values are used in determining she MCPR safety limit. See FEAR SS 15.F.2. As re-flected in FSAR S 15.F.2.2, rather than relying on theoreti-cally derived TIP noise values, GE was able to use actual noise values from an SLO test performed at an operating BWR, to de-

!s termine the TIP noise uncertainty for the Perry SLO MCPR fuel cladding integrity safety limit analysis.

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25. There would be no significant dif ference in TIP noise seen at 59.3% thermal power, as in the test referenced in 5 15.F.2.2, and TIP noise values that would be expected for SLO up to 70% of rated power for SLO at Perry. This is be'cause the major contributors to the TIP noise, as determined in previous GE generic BWR analyses, are geometric mislocation of the TIP detector and neighboring fuel channels with respect to their nominal design positions, and the random neutron, electronic and boiling noise in the reactor. See NEDO-20340,>GE Licensing Topical Report, " Process Computer Performance Evaluation Accu-racy" (June 1974) (referenced in FSAR S '4.3.5). In fact, the percent uncertainty of TIP noise will decrease as power in-creases.

26. For these reasons, the TIP noise uncertainty given in FSAR S 15.F.2.2, and the MCPR safety limit analysis in S 15.F.2.2, are applicable and conservative for SLO at Perry up to 70% power. Therefore, contrary to.OCRE's contention, no further analysis of TIP noise uncertainty and the resulting MCPR compliance is necessary or appropriate.

v CONTENTION B-4

27. Contention B-4 of OCRE's motion states:

Applicants' Technical Specifications for single loop operation up to 70% of rated thermal power should include limits on the core plate pressure drop.

Motion at 5. The contention is without basis for the following reasons.

28. The core plate pressure drop measures differential pressure between the core inlet and outlet flow conditions.

The general information that can be obtained by measuring core plate pressure drop is provided in a more direct and useful form by measuring core power and flow. Appropriate proposed Perry Technical Specification limits for SLO, which will in-clude core power and flow, will be submitted to the NRC based on the results of start-up tests which are part of the power ascension program. For this reason, no core plate pressure drop limits are needed in the Perry Technical Specifications for SLO.

29. The current technical specifications for the Cooper Nuclear Station, the plant referenced by OCRE at pages 5-6 of its motion, do include a limitation on fluctuation in the core plate differential pressure. This fluctuation involves small variations in the measured pressure. Since the Perry Technical Specifications will adequately monitor and control core power and flow during SLO through appropriate Technical Specification limits, there is no need to include separate limits on core plate pressure drop fluctations.

30. Contrary to OCRE's assertions at pages 5-6 of the Mo-tion, Technical Specification limits on core plate pressure drop (or on core plate pressure drop fluctation), would not permit better regulation of core flow. As noted in prior GE analyses, approved by the NRC, there is negligible flow assymetry (flux tilt) during SLO. See NEDO-10722A, " Core Flow Distribution in a General Electric Boiling Water Reactor as Measured in Quad Cities Unit 1" (August 1976). Since power and flow are not tilted due to SLO, Technical Specification limits on core plate pressure drop, or on core plate pressure drop fluctuation, would not lead to less variable within-core coolant flow, or yield more even cross-core pc er, as OCRE as-serts at page 5 of its motion.

31. For these reasons, contrary to OCRE's contention, there is no need to include in the Perry Technical Specifica-tions SLO limits on core plate pressure drop (or on core plate pressure drop fluctuation), and such limits have not been in-cluded in the Technical Specifications.

CONTENTION B-5

32. Contention B-5 of OCRE's motion states:

Applicants have not demonstrated that sin-gle loop operation up to 70% of rated ther-mal power will not aggravate the strong variability in flow rate along the fuel channel seen in fast BWR transients, or that this phenomenon has been conserva-tively accounted for in analyses of fast transients.

Motion at 6. The contention is without basis for the following reasons.

33. In the article on " Critical Power Ratio in BWR Tran-sient Analyses," cited by OCRE at page 6 of its motion, the authors correctly note that flow rate may vary strongly along the fuel channel during a fast transient. The article makes no distinction between fast transients during SLO and normal two loop operation, and does not suggest that this fuel channel flow rate phenomenon would be more severe during SLO fast tran-l sients. I know of no technical basis indicating that this phe-nomenon would be aggravated in a fast t'ransient initiated dur-ing SLO operation.

i l 34. The authors of the article cited by OCRE applied GE's methodologydordeterminingCriticalPowerRatio("CPR") (the l

l same methodology used in the FSAR ' Appendix 15F analysis), as l

l well as an iterative method which resulted in a lower (more conservative) CPR. The article concludes that for fast tran-sients, "the values of CPR calculated by both methods remain

within the required safety limits." The article thus does not indicate that GE's methodology is inappropriate for determining CPR in fast BWR transients. The NRC has reviewed and approved GE's CPR methodology, which is described in the GE licensing topical report listed as Reference 15.F.8-1, FSAR S 15.F.8, page 15.F.8.1.

35. Fast transients such as feedwater controller failure (maximum demand) and generator load rejection with bypass fail-ure, were specifically analyzed by GE in support of FSAR S 15.F.3. See FSAR 55 15.F.3.1.1, 15.F.3.1.2. These FSAR ref-erences-demonstrate that MCPR for both cases remains well with-in safety limits.

36. There is no technical basis for OCRE's statement at page 6 of its motion, that SLO at Perry would involve "non-uniform flow rates throughout the core." As noted in 1 30 above, and in the NRC-approved topical report referenced there-in, there would be negligible flow assymetry during SLO.

37. OCRE also suggests at page 6 of the Motion that the increased flow measurement uncertainty during SLO, as discussed in FSAR S 152F.2, might aggravate variability in flow rate along the fuel channel in fast BWR transients. OCRE suggests no technical basis for this statement, and there is none. In-creased flow measurement uncertainty during SLO is explicitly accounted for in the Appendix 15F analysis establishing core and fuel limits, as discussed in FSAR S 15.F.2. Core flow t

a measurement uncertainty is thus factored into GE's calculation of the CPR used in evaluating all of the transients considered in Appendix 15F. Increased core flow measurement uncertainty cannot affect, or " aggravate," channel flow variability during fast transients, as suggested by OCRE. Core flow measurement uncertainty is related only to initial core flow and not to channel flow variability during the transient.

38. For these reasons, there is no basis for OCRE's sug-gestion that SLO at Perry might aggravate the variability in flow rate along the fuel channel in fast BWR transients, or that Applicants have failed to conservatively account for this phenomenon in their SLO analysis of fast transients.

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39. For the reasons stated above, Contentions 3-1 through B-5 in OCRE's motion are without any credible technical basis. The contentions fail to call into question the adequacy of Applicants SLO analysis contained in Appendix 15F, and fail to raise any safety concerns.

Executed at San Jose, California, this M day of E (cef k _ , 198[.

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Kevin W. Holtzclaw Subscribed and sworn before me this - day of C4 M , 1980 .

OFFICIAL SC AL I .

._. t . ~ _ .ut PAULA F HUSSiiY ( QN 0*C 0" 9 NO""

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My ceram, expires JAN 13, 1969 )

NOTARY PUBLIC, STATE OF CALIFORNIA s

EXHIBIT A STATEMENT OF PROFESSIONAL QUALIFICATIONS KEVIN W. HOLTZCLAW Education: B.S. Mechanical Engineering (Nuclear Option), San Jose State University, M.S. Mechanical Engineering, University of California, Berkeley, General Electric Advanced Courses in Engineering Experience: March 1982 to Present: Principal Licensing Engineer, Program Manager of the GE Severe Accident Program February 1980 to March 1982: Senior Licensing Engineer, BWR Systems Licensing (GE)

June 1974 to February 1980: Technical Leader, Fuel Applications & Thermal Design (GE)

January 1971 to June 1974: Engineer, Fuel Applications and Thermal Design (GE)

July 1969 to January 1971: Program Engineer Fuel Performance &

Applications (GE)

June 1968 to January 1971: Engineer -

Nuclear Power Department (San Francisco Bay Naval Shipyard)

Licensing Experience: Approximately 17 years engineering experience in the nuclear plant power in-dustry. Since 1980, concentration on BWR I licensing issues. As a senior licensing engineer through 1982, responsible for defining and planning programs related to NRC degraded core rulemaking. Responsible for the safety and licensing program man-agement of the Limerick Probabilistic Risk Analysis. GE representative on AIF Indus-try Degraded Core Rulemaking Technical Ad-visory Group.

Since March 1982, GE Program Manager of the GE Severe Accident Program. This has en-tailed managing the BWR/6 standard plant Probabilistic Risk Assessment and Severe Accident submittals relating to evaluations

o beyond current design bases. Continued as the GE representative on the Industry De-graded Core Rulemaking (IDCOR) Technical Advisory Group. Responsible engineer in the GE Safety and Licensing organization for the GE Fission Product Retention Pro-gram and Severe Accident Source Terms and for programs relating to hydrogen genera-tion and control. Numerous presentations to domestic and foreign regulatory groups and nuclear societies.

Additional Work Experience Engineer and technical leader in General Electric's Fuel Design Department from 1969 to 1980 responsible for performing Reload and Initial Core Fuel Thermal and Thermal-Hydraulic fuel design and safety analyses.

Principal responsibilities in development of thermal analysis methods, the design and licensing of 8x8 fuel and extended exposure fuel designs, and in defining acceptance criteria for fuel thermal-mechanical fuel integrity properties and capabilities. Me-chanical design engineer in the nuclear power department of the San Francisco Bay Naval Shipyard.

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December 30, 1985 D9. .

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION '86 JAll -2 P2 '08

' 5 *. ; a /o Before the Atomic Safety and Licensing Appeal Boa 5Syif;'

In the Matter of )

)

THE CLEVELAND ELECTRIC ) Docket Nos. 50-440 ILLUMINATING COMPANY, E][ AL. ) 50-441

)

(Perry Nuclear Power Plant, )

Units 1 and 2) )

CERTIFICATE OF SERVICE This is to certify that copies of the foregoing APPLICANTS' ANSWER TO OCRE MOTION TO REOPEN THE RECORD AND TO SUBMIT NEW CONTENTIONS were served by deposit in the United States Mail, #irst class, postage prepaid, this 30th day of December 1985, to all those on the attached Service List.

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Jay .S ilberg DATED: December 30, 1985 O

UNITED STATES OF AMERICA

?w NUCLEAR REGULATORY COMMISSION O

BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of )

)

THE CLEVELAND ELECTRIC ) Docket Nos. 50-440 ILLUMINATING COMPANY, ET AL. ) 50-441

)

(Perry Nuclear Power Plant, )

Units 1 and 2) )

SERVICE LIST Alan S. Rosenthal, Chairman Atomic Safety and Licensing Atomic Safety and Licensing Appeal Board Panel Appeal Board U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Washington, D. C. 20555 Dr. W. Reed Johnson Docketing and Service Section Atomic Safety and Licensing Office of the Secretary Appeal Board U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Washington, D. C. 20555 Mr. Howard A. Wilber Colleen Woodhead, Esquire Atomic Safety and Licensing Office of the Executive Legal Appeal Board Director U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Washington, D. C. 20555 James P. Gleason, Chairman Terry Lodge, Esquire 513 Gilmoure Drive Suite 105 Silver Spring, Maryland 20901 618 N. Michigan Street Toledo, Ohio 43624

. Jerry R. Kline Ms. Susan L. Hiatt Atomic Safety and Licensing Board 8275 Munson Avenue U.S. Nuclear Regulatory Commission Mentor, Ohio 44060 Washington, D.CJ 20555 Glenn O. Bright Donald T. Ezzone, Esquire Atomic Safety and Licensing Board Assistant Prosecuting Attorney U.S. Nuclear Regulatory Commission Lake County Administration Center Washington, D.C. 20555 105 Center Street .

Painesville, Ohio 44077 Atomic Safety and Licensing ~ Atomic Safety and Licensing Appeal Board Board Panel U.S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 John G. Cardinal, Esquire Prosecuting Attorney Ashtabula County Courthouse Jefferson, Ohio 44047

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