ML20206K634

Affidavit of J Weldy & E Keegan Re Utah Contention C (Dose Limits).* Upon Rev of SAR to Reflect Applicant Revised Dose Analysis,No Basis for Utah Contention C Necessary.With Certificate of Svc

ML20206K634
Person / Time
Site:	07200022
Issue date:	05/11/1999
From:	Keegan E, Weldy J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS), NRC OFFICE OF THE GENERAL COUNSEL (OGC), SOUTHWEST RESEARCH INSTITUTE
To:
Shared Package
ML20206K619	List: ML20206K615 ML20206K634
References
ISFSI, NUDOCS 9905130222
Download: ML20206K634 (17)

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q, May 11,1999 UNITED STATES OF AMERICA I NUCLEAR REGULATORY COMMISSION

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PEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of . )

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PRIVATE FUEL STORAGE, L.L.C. ) Docket No. 72-22-ISFSI

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(Independent Spent Fuel )~

Storage Installation) ) ,

I AFFIDAVIT OF JAMES WELDY AND ELAINE KEEGAN CONCERNING UTAH CONTENTION C (DOSE LIMITS)

- James Weldy (JW) and Elaine Keegan (EK), having first been duly sworn, do hereby state as follows:

1(a). (EK) My name is Elaine Keegan. I am employed as a Health Physicist in the Technical Review Directorate, Spent Fuel Project Office, Office of Nuclear Materials Safety and Safeguards, U.S. Nuclear Regulatory Commission (NRC), in j Washington, D.C. A statement of my professional qualifications is attached hereto. ,

1(b). (JW) My name is James Weldy. I am employed as a Research Engmeer I l

at the Center for Nuclear Waste Regulatory Analyses (CNWRA), which is a division of

, the Southwest Research Institute (SWRI), in San Antonio, Texas. I am providing JA l affidavit under a technical assistance contract between the NRC Staff and the SWRI. A statement of my professional qualifications is attached hereto.

2. This Affidavit is prepared in response to the " Applicant's Motion for Summary Disposition of Utah Contention C," filed on April 21,1999 by Private Fuel i 9905130222 990511 Y PDR ADOCK 07200022 3 C POR s

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Storage L.L.C. (" Applicant" or "PFS"), along with the Affidavit of William Hennessy, i

dated April 21,1999. )

3.  : Utah Contention C, entitled " Failure to Demonstrate Compliance with

.'NRC Dose Limits," states as follows:

CONTENTION: The' Applicant has failed to demonstrate a reasonable assurance that the dose limits specified in 10 CFR i

~ 72.106(b) can and will be complied with in that:

-1. License Application makes selective and inappropriate use of data from NUREG-1536 for the fission product release fraction.

2. License Application makes selective and inappropriate use of data from SAND 80-2124 for the respirable particulate  ;

fraction. I 1

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3. The . dose analysis in the License Application only considers dose due solely to' inhalation of the passing

. cloud. Direct radiation and ingestion of food and water {

are not considered in the analysis.

4. Utah Contention C refers to the Applicant's calculation of the dose that would be received by a member of the public in the event of a loss-of-confinement accident l at the Private Fuel Storage Facility ("PFSF"), as presented in the Applicant's Safety Analysis Report ("SAR") that' was filed with its application of June 20,1997.

5. In the basis statements for Utah Contention C, the State of Utah asserted that the ~ Applicant's dose analysis in i 8.2.7.2 of the SAR did not provide an adequate evaluation'of the dose consequences'of a loss-of-confinement accident in that it "makes

. selective and inappropriate use of data sources regarding doses, and fails to take imponant dose contributors into account" (Utah Contentions, at 18).- Specifically, the State assened o

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(a) that the Applicant incorrectly assumed (in the table on SAR p. 8.2-37), that the fraction j

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- of Cs-134, Cs-137, and Sr-90 that will be released into the canister is 2.3 E-5, based on NUREG-1536, Standard Review Plan for Dry Ca* Storane Systems (January 1997) (Id. i I

at 19), (b) that the Applicant's dose analysis inappropriately relied upon a Sandia National l 1

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. 1.aboratories repon concerning transportation accidents (SAND 80-2124, Trananonation

' Accident Scenarios for Commercial Soent Fuel (1981)), to suppon its release fraction assumption that 90 % of the volatiles (Co-60, Sr-90, I-129, Ru-106, Cs-134 and Cs-137) released from the spent fuel to the canister will not escape the canister (Id., citing SAR at 8.2-38); (c) that the Applicant's dose analysis inappropriately relied upon the Sandia report for its assumption that only 5% of the release fraction of Co-60 and Sr-90 will be '

respirable (Id. at 20, citing SAR at 8.2-39); and (d) that the Applicant's dose analysis failed to take into account the dose contributed by pathways other than' inhalation of the j i

passing cloud, such as direct radiation from cesium deposited on the ground, and ingestion of food and water or incidental soil ingestion, in violation of 10 CFR f 72.24(m) (Id. l 1

at 21, citing SAR at 8.2-39). j

6. In its motion for sununary disposition of Utah Contention C, PFS assens that the bases for the contention have been eliminated and that the contention is therefore no longer valid. In support of this assertion, PFS states that it has revised the challenged ponions of its accident dose analysis, in response to the Staff's RAIs, and that its revised calculation was performed in r.ccordance with ISG-5. In panicular, PFS states that part 1 of the contention is no longer valid because its revised dose calculation no longer makes use of the fission product release fractions contained in NUREG-1536 or the assumptions 1 1

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in SAND 80-2124 about the fraction of particulates or volatile fission products that would

' be released by the fuel but retained in the canister. Second, PFS states that part 2 of the contention is no longer valid because its revised dose calculation no longer makes use of the respirable. particulate fraction contained in SAND 80-2124. Third, PFS states that part 3 of the contention is no longer valid because its revised dose calculation takes into

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account all applicable environmental pathways to which a member of the public may be exposed both during passage of the contaminated plume and following deposition of contaminated material on the ground. See Applicant's Motion at 17-18; Affidavit of William Hennessy at 3-4.

7. We have reviewed the Applicant's revised accident dose calculation, which PFS submitted to the NRC in its February 1999 response to the Staff's RAls. On the basis of our review, we believe that the Applicant's revised dose analysis satisfactorily addresses each of the concerns raised by this . contention. Further, we are satisfied that the Applicant's revised accident dose analysis appropriately follows the guidance in ISG-5, and

't hat its resulting dose estimates satisfy the regulatory requirements set forth in 10 C.F.R.

. art 72. ~ Accordingly, we are satisfied that upon revision of the SAR to reflect the Applicant's revised dose analysis, the license application wil.1 satisfy the Commission's l 1

.egulatory requirements pertaining to the analysis of offsite dose consequences of a loss-of- l l

confinement accident. The bases for these conclusions are as follows.

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8. Pursuant to 10 C.F.R. i 72.106(b) (as revised in October 1998, to be l consistent with 10 C.F.R. Part 20 dose calculational methodology (63 Fed. Reg. 54559)), l l

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--4 an applicant for an independent spent fuel storage installation (ISFSI) must establish a controlled area such that:

Any individual located on or beyond the nearest boundary of the controlled area may not receive from any design basis accident the more limiting of a total effective dose equivalent of 0.05 Sv (5 rem), or the sum of the deep-dose equivalent and the committed dose equivalent to any

. individual organ or tissue (other than the sens of the eye) of 0.5 Sv (50 rem). The lens dose. equivalent shall not exceed 0.15 Sv (15 rem) and the shallow dose equivalent to skin or to any extremity shall not exceed 0.5 Sv (50 rem). . . .

Also, as set forth in 10 C.F.R. I 72.24(m), an applicant's SAR is required to contain:

An analysis of the potential dose equivalent or committed dose equivalent to an individual outside the controlled area from accidents or natural phenomena events that result in the' release of radioactive material to the environment or direct radiation from the ISFSI . . . . The calculations of individual dose equivalent or committed dose equivalent must be performed for direct exposure,-

inhalation, and ingestion occurring as a result of the postulated design basis event.

- Further, as set forth in 10 C.F.R. f 72.126(d), an applicant is required, inter alia, to

, submit analyses of design basis accidents which "show that releases to the general environment will be within the exposure limits given in f 72.106."

9. The NRC Staff has issued various guidance deraments concerning the proper methodology for calculating offsite doses for design basis events. Certain guidance is contained,- for example, in NUREG-1567, Standard Review Plan for Soent Fuel Dry Storage, at-15-32 (Draft. September 1998). More recently (and subsequent to the Applicant's submittal of its SAR), the Staff issued further guidance on the proper

. methodology to be utilized in calculating offsite doses resulting from a loss-of-confinement accident, as set forth in Interim Staff Guidance-5 (ISG-5), entitled " Accident Dose Calculations" (September 28, 1998). ISG-5 recommends the use of release fractions contained in NUREG-1617, Standard Review Plan for Transoortation Packanes for Soent Nuclear Fuel (DRAFT, . March 1998), Table 4-1. In addition, ISG-5 describes an acceptable method to account for radionuclides that are released into the cask volume but

- do not escape 'the cask volume based on the leakage rate of air out of a small hole in the confinement boundary. The technical bases for these release fractions (pertaining to the release of gases, volatiles and particulate from the fuel to the cask interior) and calculation methodology are described in NUREG/CR-6487, Containment Analysis for Tvoe B Packanes to Transoort Various Contents (November 1996). In contrast to previous Staff -

guidance, ISO-5 does not assume that the confinement boundary will be breached (non-mechanistic failure). This is consistent with stmetural analysis which demonstrates that the confinement integrity is maintained during normal, off-normal, and accident conditions.

Also, ISG-5 recommends the use of larger values for the concentration of " CRUD". on BWR fuel,-and consideration of a more comprehensive array of gaseous, volatile, and particulate radionuclides in the calculation. ISG-5 does not include any mitigation of the radioactive source term available for release from the cask interior to the environment --

i.e.. no credit is given for plateout, particle size, etc. This provides a bounding condition for the analysis.

10. In its _ initial SAR, the Applicant utilized the release fractions from

' NUREG 1536 Table 7.1, to estimate the quantity of radioactive material that is released

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from the fuel into the cask cavity during a loss-of-confinement accident. The Applicant then reduced this release quantity by (a) the fraction of volatile and particulate material that plates out on the interior of the cask and is not released to the environment, and (b) the fraction of Sr-90 and Cc-60 that are not respirable and cannot contribute to the inhalation

- dose, which the applicant obtained from SAND 80-2124.-

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11. In section 8.2.7 of its initial SAR, the Applicant included a calculation of the consequences from a postulated loss-of-confinement accident, which considered the committed effective dose equivalent (CEDE) from the inhalation of the passing cloud; the l SAR did not calculate the doses received by members of the public from other pathways, such as from direct exposure and ingestion, as required by 10 C.F.R. I 72.24(m).

12. The Applicant's accident dose analysis was the subject of two separate Requests for Additional Information (RAls) transmitted by the Staff to PFS. On April 1, 1998, the Staff requested additional information concerning the Applicant's accident dose calculations, including the basis for its assumption of a respirable fraction of 5 percent for  !

1 Co-60 and Sr-90, and its consideration of an inhalation pathway only (see RAIs 8-4, 8-5, )

1 and 8-8, dated April 1,1998). Subsequently, the Applicant's dose analysis was further-addressed in an RAI transmitted by the Staff to PFS on December 10,1998. In particular, RAls 7-1 and 8-4 of this second round of RAls requested that the Applicant revise its dose i calculations to correct its assumptions for the respirable fraction of Co-60 released in an accident; that the Applicant follow the latest NRC guidance on calculating offsite doses for i

'a loss-of-confinement accident, as set forth in ISG-5; and that the Applicant justify its failure to model pathways other than the inhalation pathway. Responses to the Staff's first Ir I

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. and second round RAls concerning these matters were submitted by PFS on May 19,1998, and February 10,1999, respectively. PFS submitted a partial revision to Chapter 8 of its SAR on May 22,1998; and we understand that PFS has indicated it will submit a further SAR revision.in May 1999 that will incorporate its revised accident dose analysis,

discussed below.

,13. The Applicant's revised dose analysis, set forth in its February 10,1999 response to'RAls and its February 1999 revision of its SAR, appropriately takes into  ;

I account the considerations set forth in ISG-5, with respect to the respirable release fractions of radionuclides and mitigation factors such as plateout and deposition. The Applicant's revised dose analysis conservatively assumes that 100% of the released radioactive material is respirable. The revised dose analysis bases the release quantity of radioactive material from the free volume inside the cask to the exterior of the cask on the

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volume of air that can leak through a very small diameter hole assumed to exist in the containment boundary under accident conditions, consistent with ISG-5. This methodology

'is in contrast to the Applicant's original accident dose calculation in that it does not rely i

on a constant fraction of mass released from the fuel that escapes containment to account  !

l for mitigation factors such as plateout and deposition of material within the breached cask.

14. - 'In light .of. the ' Applicant's revised dose analysis, Part 1 of Utah i i

' Contention C, which asserted that the license application made selective and inappropriate l

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use of data' from NUREG-1536 for the fission product release fraction, is no longer applicable, because (1) the accident dose calculation no longer utilizes data from j NUREG-1536 for the fission product release fraction; (2) the accident dose calculation no l

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. . longer utilizes data from SAND 80-2124 for the fission product release fraction; and (3) the accident dose calculation follows a single NRC guidance document (ISG-5) and does not

make selective and inappropriate use ofdata from any source. Accordingly, Pan 1 of Utah

- Contention C is no longer valid.

15. Similarly, the Applicant's revised dose calculation no longer takes credit for any reduction in dose due to the size distribution of the released particulates. The original dose calculation utilized data from SAND 80L 2124 to suppon its assumption'that only five percent of the isotopes Co-60 and Sr-90 released from the fuel assemblies will be respirable by a human. The revised dose calculation assumes that all paniculates matter released from the cask will be respirable. Therefore, Part 2 of Utah Contention C, which assened that the License Application made selective and inappropriate use of data from

. SAND 80-2124 for the respirable paniculate fraction, is no longer applicable because

_(1) the accident dose calculation no longer utilizes data from SAND 80-2124 for the respirable paniculate fraction; and (2) the accident dose calculation follows a single NRC guidance document (ISG-5) and therefore does not make selective and inappropriate use of data from any source. Accordingly, Pan 2 of Utah Contention C is no longer valid.

16. The Applicant's revised dose analysis also addresses the concerns raised in pan 3 of the contention, with respect to dose pathways. In its revised dose armlysis, the applicant has included an assessment of the dose delivered to members of the public following the deposition on ti,e ground of radioactive material in the plume from a loss-of-confinement accident. This is in accordance with the requirements of 10 CFR 72.24(m),

which requires that calculations of individual dose equivalent or committed dose equivalent

be performed for direct exposure, inhalation, and ingestion occurring as a result of postulated design basis events. The revised dose calculation assesses the dose received by a receptor from the direct exposure to contaminated ground, inhalation of resuspended radioactive material, ingestion of milk and beef following grazing of contaminated plants, and inadvertent ingestion of soil contaminated with radioactive material deposited on the ground. Additionally, the revised dose calculation determines the dose received from the external exposure to the contaminated plume as it passes the receptor. While the revised analysis omits the surface water and groundwater pathways, this is not inappropriate, based on the Applicant's determination, described in its Environmental Report, that there am no public or private surface drinking water supplies in the PFSF vicinity and there are no wells used for drinking water located near the boundary of the controlled area of the ISFSI, which is the location at which a member of the public could receive the greatest dose from the accident.

- 17.. The Applicant's revised dose analysis includes dose calculations for a receptor located at the PFSF site boundary and at a location representing the nearest actual residences to the facility using realistic estimates of exposure times for receptors located at both locations. Both locations showed that the dose following deposition of radioactive material in the soil was dominated by external exposure to Co-60. The Applicant's calculations also showed that a higher dose was received by an individual located at the PFSF fence than by individuals located at actual residences in the area.

18.- Based on our review of the Applicant's revised dose analysis, as set forth in its February 1999' response to the Staff's RAIs, we believe the pathways considered by l

the Applicant are appropriate and adequate to assess the dose that an individual located at the site boundary v ould receive from the passing cloud and following the deposition of radioactive material on the ground after a loss-of-confinement accident. Further, we agree with the Applicant's determination that an individual located at the site boundary would be the member of the public who would receive the largest dose from a loss-of-confinement accident.

19. In light of the Applicant's revised dose analysis, part 3 of Utah Contention C is no longer valid, in that (1) the revised dose calculation determines the dose j l

from' direct exposure to the maximally exposed member of the public from .the contaminated plume of airborne radioactive material; (2) the revised dose calculation determines the dose from all applicable pathways for the maximally exposed member of the public including direct exposure, inhalation, and ingestion pathways from the soil contaminated by radioactive material deposited by the plume; and (3) the ingestion of l l

contaminated water is not a credible pathway because the member of the public who would I

receive the largest dose from a loss-of-confinement accident is a hypothetical individual l located just outside the site boundary -- and there are no permanent residences, or public or private surface drinking water supplies or wells used for drinking water at the location of the maximally exposed individual.

20. - Based on our review of the Applicant's revised dose analysis, we are

- satisfied that the revised dose calculation was performed in accordance with applicable Staff guidance, contained in ISG-5, and that it satisfies applicable NRC requirements.

Specifically, the revised dose analysis meets the requirements of 10 CFR 72.24(m), by

performing calculations of individual dose equivalent for direct exposure, inhalation, and ingestion occurring as a result of a loss-of-confinement accident. Further, the revised dose calculation meets the requirements of 10 CFR 72.106(b), by demonstrating that any individual located on or beyond the nearest boundary of the controlled area will not receive from a loss-of-confinement accident a total effective dose equivalent of 0.05 Sv (5 rem).

Additionally, if the entire dose calculated by the Applicant was deposited in any single organ, the sum of the deep-dose equivalent and the committed dose equivalent to any )

individual organ or tissue other than the lens of the eye would not exceed 0.5 Sv (50 rem);

l the lens dose equivalent would not exceed 0.15 Sv (15 rem); and the shallow dose equivalent to skin or any extremity would not exceed 0.5 Sv (50 rem). )

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21. Based upon the above considerations, we have concluded that upon revision j l

of the SAR to reflect the Applicant's revised dose analysis, the license application will )

I satisfy the Commission's regulatory requirements pertaining to the analysis of offsite dose consequences of a loss-of-confinement accident. Further, upon revision of the SAR to reflect the Applicant's revised dose analysis, there is no basis for Utah Contention C.

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22. I hereby certify that the foregoing is true and correct to the best of my knowledge, information and belief.

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(-l^L41LI 6 eedal Elaine Keeggif )

l Subscribed and sworn to before me l E.

this lith day of May,1999. j D ,

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' Notary Public '

My commission expires: ) MT ,

M.. , f b d J L, J James Weldy /

f& t. 1 Subscribed and sworn to before me W TARY, i this 1Ith day of May,1999.

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e James Weldy Research Engineer Center for Nuclear Weste Regulatory Analyses B.S. In Nuclear Engineering, The University of Michigan,1995 M.Eng. In Radiological Health Engineering, The University of Michigan,1996 Mr. Weldy has experience in many areas of Radiological Health and Nuclear Engineering. He is skilled in environmental transport and pathway analysis, radiation dose calculation, risk assessment, radiation shielding .

analysis, criticality assessment, and detection and measurement of radiation. i l

During his practicum for his Master's degree at Argonne National Laboratory, he designed a Bonner sphere l system to measure the dose delivered to a person by a pulsed neutron source. This included writing a Monte 4 Carlo code to model the transpon of neutrons through the spheres and the temporal arrival of the neutrons at I the detector. 1 a

He is currently performing calculations and modeling to assess the performance of a proposed high-!evel j nuclear waste repository at Yucca Mountain, Nevada. His responsibilities at the CNWRA include conducting I research into dose assessments for the proposed high-level radioactive waste disposal site. He also provides l technical suppon to the Nuclear Regulatory Comrnission for projects including uranium recovery activities re'ated to in situ leach mining, evaluating reclamation plans and license amendment requests, and assessing interim storage of spent nuclear fuel including evalusting safety analysis repons and emironmental repons.

He has been responsible for the evaluation of radiation hazards to workers and members of the public from Independent Spent Fuel Storage Installations and has been involved in the licensing of these facilities. He has assisted in the preparation of emironmental impact statements and evaluations of the hazards associated with the wastes contained in the high-level waste tanks at the Hanford Site in Washington state. l l'

PROFESSIONAL CHRONOLOGY: Research Fellow, Argonne National Laboratory,1996; Engineer, Southwest Research Institute, Center for Nuclear Waste Regulatory Analyses,1997-1998; Research Engineer, Southwest Research Institute, Center for Nuclear Waste Regulatory Analyses,1998-present j MEMBERSHIPS: American Nuclear Society, Health Physics Society.

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4 Elaine Keegan Health Physicist Spent Fuel Project Office Office of Nuclear Materials Safety and Safeguards (NMSS)

U. S. Nuclear Regulatory Comtnission B.S. in Radiological Health Physics, University of Lowell,1978

' Ms. Keegan has experience in many areas of Radiological Health Physics. She is skilled in radiation shielding analysis, environmental tsansport and pathway analysis, radiation dose calculation, and emergency plannmg.

Ms. Keegan is currently performing shielding evaluations for the licensing of spent nuclear fuel transportation i and storage casks. Her work includes the evaluation ofaccident dose analyses, occupational dose calculations, -

and the adequacy of proposed radiation protection programs. She has provided input for the preparation of NUREG-1567, Standard Review Plan for Spent Fuel Dry Storage Facilities; NUREG-1536, Standard Retiew  !

Plan for Dry Cask Storage Systems; and NUREG-1617, Standard Review Plan for Transportation Packages l for Spent Nuclear Fuel. She has reviewed emergency plans for dry storage facilities. She has also prepared i several emironmental assessments for licensing actions invohing the storage of spent fuel.

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Ms. Keegan also has been project manager for the license renewal of a major fuel fabrication facility, which j included evaluating criticality, radiation protection, emergency planning, fire safety, and emironmental l

. protection. She has prepared numerous environmental assessments for the renewal of fuel cycle facilities, and she has assisted in the preparation of several environmental impact statements relating to the fuel fabrication

. facilities.

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Whils working at the Vermont Yankee nuclear power plant, Ms. Keegan was responsible for implementing j the facility's radiological emironmental monitoring program. She prepared the monthly, quarterly, and annual i dose calculations for de public from plant operations. She reviewed the data from the environmental monitoring program to ensure that plant operations were not negatively impacting the environment. She was responsible for revising and maintaining the emergency plan and preparing emergency drill scenarios. During emergency drills, she directed the off-site teams as to where and what types of radiation, air, and environmental samples to collect.

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' PROFESSIONAL CHRONOLOGY: Radiation Monitor, Reynolds Electric & Engmeering Co., Inc.,1979; Environmental Analyst, Reynolds Electric & Engineering Co., Inc.,1979-1980; Chemistry and Health Physics Technician, Vermont Yankee Nuclear Power Corpi,1980-1984; Environmental Coordinator, Vermont Yankee Nuclear Power Corp., 1984-1990; Environmental Scientist, Division of Fuel Cycle Safety and Safeguards, i Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,1990 1996; Health I Physicist, Spent Fuel Project Office, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear  !

Regulatory Commission,1996-present.

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• UNITED STATES OF AMERICA 00CKEIED NUCLEAR REGULATORY COMMISSION USHRC BEFORE THE ATOMIC SAFETY AND LICENSING BOARD '09 NAY 12 All :58 In the Matter of ) Om <e m

) RUi PRIVATE FUEL STORAGE L.L.C. ) Docket No. 72-22-ISMIUD , sp

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(Independent Spent )

Fuel Storage Installation) )

CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S RESPONSE TO APPLICANT'S MOTION FOR

SUMMARY

DISPOSITION OF UTAH CONTENTION C (DOSE LIMITS)" in the above captioned proceeding have been served on the following through deposit in the Nuclear Regulatory Commission's internal mail system, or by deposit in the United States mail, first class, as indicated by an asterisk, with copies by electronic mail as indicated, this lith day of May,1999:

G. - Paul Bollwerk. III, Chairman Atomic Safety and Licensing Board Panel Administrative Judge U.S. Nuclear Regulatory Commission ,

Atomic Safety and Licensing Board Washington, DC 20555 l U.S. Nuclear Regulatory Commission Washington DC 20555 Office of the Secretary (E-mail copy to GPB@NRC. GOV) ATTN: Rulemakings and Adjudications Staff U.S. Nuclear Regulatory Commission Dr. Jerry R. Kline Washington, DC 20555 Administrative Judge (E-mail copy to:

Atomic Safety and Licensing Board HEARINGDOCKET@NRC. GOV)

U.S. Nuclear Regulatory Commission

. Washington, DC 20555 Office of the Commission Appellate l (E-mail copy to JRK2@NRC. GOV) Adjudication i J

Mail Stop: 16-C-1 OWFN Dr. Peter S. Lam U.S. Nuclear Regulatory Commission Administrative Judge Washington, DC 20555 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission James M. Cutchin, V

. Washington, DC 20555 Atomic Safety and Licensing Board (E-mail copy to PSI 4NRC. GOV) U.S. Nuclear Regulatory Commission

- Washington, DC 20555 (by E-mail to JMC3@NRC. GOV) 1

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i Danny Quintana, Erg.* Diane Curran, Esq.*

Danny Quintana & Associates, P.C. Harmon, Curran, Spielberg 50 West Broadway, Fourth Floor & Eisenberg, L.L. P.

Salt Lake City, UT 84101 1726 M. Street N.W., Suite 600 (E-mail copy to Washington, D.C. 20036 quintana @Xmission.com) (E-mail copy to dcurran@harmoncurran.com)

Jay E. Silberg, Esq.*

l Ernest Blake, Esq.* John Paul Kennedy, Sr., Esq.*

Paul A. Gaukler, Esq.* 1385 Yale Ave.

SHAW, PITTMAN, POTTS & Salt Lake City, UT 84105 TROWBRIDGE (E-mail copy to john @kennedys.org) i 2300 N Street, N.W.

Washington, DC 20037-8007 Joro Walker, Esq.*

(E-mail copies to. jay _silberg, Land and Water Fund of the Rockies paul _gaukler, and ernest _blake 165 South Main St., Suite 1

@shawpittman.com) Salt Lake City, UT 84111 (E-mail copy to joro61@inconnect.com)

Denise Chancellor, Esq.*

Fred G. Nelson, Esq. Richard E. Condit, Esq.

Utah Attorney General's Office Land and Water Fund of the Rockies 160 East 300 South,5th Floor 2260 Baseline Road, Suite 200 P.O. Box 140873. Boulder, CO 80302 Salt Lake City, UT 84114-0873 (E-mail copy to rcondit@lawfund.org)

(E-mail copy to dchancel@Statt.UT.US)

Connie Nakahara, Esq.*

l Utah Dep't of Environmental Quality '

168 North 1950 West P. O. Box 144810 Salt Lake City, UT 84114-4810 (E-mail copy to enakahar@ state.UT.US)

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h tt) t v Sherwin E. Turk Counsel for NRC Staff i _