ML20205S007
| ML20205S007 | |
| Person / Time | |
|---|---|
| Site: | 07200022 |
| Issue date: | 04/21/1999 |
| From: | Hennessy W AFFILIATION NOT ASSIGNED, STONE & WEBSTER, INC. |
| To: | |
| Shared Package | |
| ML20205S000 | List: |
| References | |
| ISFSI, NUDOCS 9904260047 | |
| Download: ML20205S007 (19) | |
Text
e UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the AmW Safety and Licensing Board in the Master of
)
)
PRIVATE FUEL STORAGE L.L.C.
)
Docket No. 72-22
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(Private Feel Storage Facility)
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AFFIDAVIT OF WrrMW HENNESSY CITY OF CHARLOTTE
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STATE OF NORTH CAROLINA
)
William Hannessy, being duly sworn, states as follows:
1.
I am the Assistant Project Manater and Lead Licensing #ade with Stone &
Webseer Engineering Corporation (Stone & Wehner) for the Private Fuel Storage Facility (PFSF). Stone & Websteris the architeet engineer for the PFSF. My professional and educational ewe is sunmarized in the curriculum vitae attached as Exhibit I to af5 davit.
In my capacity as Stone & Webster Mearw Project Manager and Lead 2.
Licensing Engineer for the PFSF,I oversaw the prepmemian and am knowledgeable of the radiation dose calculations prepared on behalf of Private Fuel Storner, L.L.C. (PFS) for the licensing of the PFSF by the Nuclear Regulatory Commission (NRC). I am also familiar with Contendon Utah C tained by the State of Utah in the NRC licensing heariag for the PFSF.
Based on then existing NRC Staff guidance, the PFS initial License Application j
3.
dated. tune 20,1997, analyzed the does consequences for apostulated loss ofcannaemans accident assuming a hypothetical, non wh=alstic breach of a canister suoring spem fuel at th 9904260047 990421 PDR ADOCK 07200022 C
l PFSF. The fission pmduct telease fractions from NUREG-1536 were used in perfonning this calculation. It was further assumed, based on infhetnat on ka:n Sandla National l.aboratories Report SAND 80-2124, "Transponstion-Accident Scenarios tbr Commercia1 Spent Fu percent of the perdeulate and volatile fluion products of various radionuclides (Co.60, 129, Ru-106, Cs-134, and Cs-137) rel***d from spem the fbel assemblies would be held up within the breached canister and would not escape to the atmosphere. PFSF Safety Analysis Report (SAR), Section 8.2.7.2. It was also assumed, based on information from SAND that only $ percent of the isotopes Co 60 and Sr-90 were of a size ihm would be respirable b person. SAR at Section 8.2.7.3. On thus basis,PFS calculated the total effective dose resulting from this hyW4 canister breach to an individua1 assumed to be located at the closest point on the boundary of the Owner Comnolled Area fmm inhalation of radionucl the pluene and exposure m direer neuerion nom the plume (submersion). Secondary environmemal pathways, such as direct va
- To contaminated soil, inhalation and ingestion of cantaminated soil, and ingestion of milk and beef fkom anhaals that have graamd on grass /fbdder grown in carnaminated soil, were not included in this calcuission because concibution from such perhways was believed to be negligible.
In its second round Raquests for Additions! !nformanon (RAI) dared December 4.
10,1998, the NRC requested PFS in RA17-1 to analyze the dose consequences fo loss of confinement accidentin accordance with the most recent Staff guidance provided i Interiza Staff Guidance 5 (ISG 5) (October 6,1998) which provides for the calculation of radiation doses resulting Rom canister leakage, with the leak rate based in part on the weld helium leak test acceptance criteria. PFS performed the analysis of the canister leak accident in accordance with ISO-5 for hypothetical accident condidons, and submitted of this analysis (which consdamas the new PFSF !!% basis fbr accident dose con in its response to RAI 7-1. PFS's response to RAI 7-1, filed under cover lener 10,1999, is ansched as Exluidt 2 to this af5 davit. A copy of PFS's responses to th round RAIs, including PFS's response to RAI 7-1, was sent to the State of Utah via o mail on February li,1009. A copy of the calculations and other backup to PFS's response including the backup caleM4 fbr PFS's response to RA17-1, were sent to the Sise businesa day delivery on February 13,1999.
2
a 5.
I have reviewed the State's bases underlying the three subparts of Contention Utah C as well as the contention itself PFS's new calculation of radiadon doses described in PFS's response to RAI 7-1 addresses the issues raised in each of the three subparts of Utah C.
t 6.
In subpart 1 of Utah C, the State claims that PFS's original calculation *W f
selective and inappropriate use of data fhun NUREG-1536 for the fission product release fraction." PFS's new calculation, however, no longer uses the fission product release tacdon from NUREO 1536, but now relies on NUREO 1617 for the release & action in accordance with ISO-5. Nor does PFS's new calcularion rely on NUREG-1536 in any other respect. In its bases i
to subpart 1 of Utah C, State specifically takes issue with the assumpdon used in PFS's original calculation that 90 percent of the volatile fission products assumed to be released from the spent fbel would plateout or holdup in the canister and therefore could not escape 'mto the enviromnent. In PFS's new calculation, however, "[n]o credit was taken for holdup (plasecut, f
deposition, etc.) of particulates or volatile fission produca released Aom the thel inside the canister." Response to RAI 7-1, at 2. Rather, the c61=% conservatively assurnes that 100%
of these radionuclides assumed to be reimased from the spent fuel rods are available for release f
from the canister.
7.
In subpart 2 of Utah C, the State takes issue with the assumpdon used in PFS's original calculation that only 5 percent of the isotopes Co-60 and Sr-90 :eleased Acm the spent fuel will be of tespirable size. PFS's new calculation no lancer uses, however, the assumption contained in SAND 80 2124 that only five percent of the isotopes Co-60 and Sr 90 released Aom the spent fbel will be respirable, nor does is rely on SAND 802124 in any other respect. Rasher, in PFS's new calculation %e resp!rable taction of the material released for all radianuclides is assumed to be 1.0," or 100 percent. R==aam to RAI 7-1, at 2.
8.
In subpart 3 of Utah C, the State takes issue with PFS's original calculadon because it did not consider dose pathways Aom direct radiation and ingestion of food and water.
PFS's new calculation, however does calculate the potential radiation doses Aom applicable environmental pathways fbliowing the deposition of radioactive maserial in the plame froen an accident, in addition to doses toen inhaladon and direct shine Dom the passing plume. Response to RAI 7-1, at 3-4. The new calculation includes direct exposure in contarninated ground, 3
r inhalmion of resuspended radioactive muerini, insemica of milk and beef follow ingemion of soil." Response to RAI 71, at 4. Water was not included as an pathway because such a perhway would need to involve surface drinkin significance. As stated in Section 2.5.1 of the PFSF Environmental Rep are no public or private surface drinking water supplies in the PFSF vic l ffluems there is no potable surface water supply that could be subject to nonnal or acci ham the facility."
/
h William Hennessy swmo before me this [l _ day of April IM.
W"
(
Notary Public My Commission expires _
481EllulhaegheskEM 4
TOTA A cy:;
i Exhibit 1
e Curriculum Vitae of William Hennessy Mr. Hennessy has 30 years of experience in nuclear project management and engineering. Currently, he is Assistant Project Manager for the Private Fuel Storage Facility (PFSF) project where his duties include providing the day-to-day project direction for engineering, radiological analysis, licensing, estimating, contract administration, controlling project costs, and scope change control. Concurrently, he is the Licensing Manager for Stone & Webster's Independent Spent Fuel Storage Installation (ISFSI) projects.
In addition to PFSF, other ISFSI projects include Maine Yankee and Indian Point-2. Responsibilities encompass developing the licensing strategy and preparation of the required documentation for NRC, state, and local license / permit applications. NRC licensing activities include ensuring that all the required reports, studies, investigations, and calculations are prepared in accordance with the conditions of 10CFR72 and working with the NRC in resolving comments developed during their evaluation of these documents.
Previously, he was the Lead Nuclear Safety / Environmental Engineer for the design of the Actinide Packaging & Storage Facility Project for the DOE Savannah River Site, having responsibility for the following areas: radiological shielding and safety analysis, criticality, environmental effluent monitoring and permitting, radwaste management, health physics lab / operation, emergency planning, design process hazards review, functional classification of SSCs, radiation monitoring equipment, and industrial safety.
Mr. Hennessy also managed and performed nuclear safety analyses for ongoing tasks for the DOE Rocky Flats site, eg, performing a radiological risk assessment for plutonium residue process streams, managing the preparation of the Conceptual Safety Report for the Interim Storage Vault, performing radiological dose analyses for wet combustible, and related safety assessments.
Previously, he supported the Department of Energy at the Savannah River Site on several contracts in perfonning engineering / safety analysis reviews and developing engineering assessments. For the DOE-SR Reactors contract as Deputy Project Manager, he managed the day-to-day operations of the project, including providing technical support services fbr the SRS reactors and associated site facilities and programs. 'Ihe emphasis was in upgrading the reactor facility conduct of operations to be comparable with the NRC licensed nuclear facilities. He also was the Engineering Manager for the Aiken Office and provided support for S&W's High Level Waste Project in performing engineering / safety analysis reviews.
He was responsible for the technical and administrative direction of the Nuclear Technology and Licensing Division for the Stone & Webster Cherry Hill, NJ office. Areas of responsibility included thermal. hydraulic and radiological safety analyses, safety system design, safety analysis report preparation, NRC licensing, radiological shielding, probabilistic risk assessment, radwaste, and emergency planning. In addition, he has managed safety analyses studies, hazard and operability studies, fault tree analysis, and emergency planning work for the chemical and petro chemical industries.
Prior to that, he was responsible for supervising and performing safety analysis, licensing, and engineering design for the commercial nuclear power industry. His formal education includes a Master of Science in Physics from Lynchburg College and a Bachelor of Science in Mechanical Engineering from Drexel University. He is a registered Professional Engineer in the States of Pennsylvania and South Carolina, and has numerous publications dealing with nuclear safety analyses.
Exhibit 2 I
I l
t
SAFETY ANALYSIS REPORT CHAPTER 7-RADIATION PROTECTION 7-1 Revise the calculation of the impacts of the accident using the release fractions and methodology contained in Interim Staff Guidance-5 (ISG-5),
Accident Dose Calculations (Nuclear Regulatory Commission,1998) to show compliance with the accident dose limits in 10 CFR 72.106(b).
The calculation in the SAR has been conducted inappropriately. The use of a respirable fraction of 5% for the release of Co-60 is not appropriate.
The SAR cites Table XX cf SAND 80-2124 to justify the use of this fraction. However, page 39 of this document indicates that this fraction was measured fcr particulates released from the interior of the fuel via a burst-rupture me aanism. The majority of the source of Co-60 from the spent fuel would be from the CRUD on the exterior of the fuel assemblies.
The licensee's calculation of accident impacts in the SAR does not follow the most recent staff guidance on calculating the consequences of a postulated loss-of-confinement event. The current staff guidance on this calculation is published by the Spent Fuel Project Office as Interim Staff Guidance - 5 (ISG-5) (Nuclear Regulatory Commission,1998).
RESPONSE
The calculation of the impacts (individual doses) resulting from the hypothetical canister leakage accident for the PFSF has been revised in accordance with Interim Staff Guidance-5 (ISO-5) to show compliance with the accident dose limits in 10 CFR 72.106 (b).
For the hypothetical accident case, the calculated releases are based on leakage of the canister design that could lead to the largest release of radioactive material. Both TranStor and HI-STORM canisters were evaluated. It was determined that maximum dose rates are associated with postulated leakage of a TranStor canister containing 61 design basis BWR fuel assemblies. While the HI-STORM MPC-68 canister contains seven more BWR fuel assemblies, the calculated leak rate for the MPC-68 under bounding accident conditions of temperature and pressure (1.58 E-5 cc/sec, presented in Section 7.3.3.1 of the HI-STAR TSAR, Holtec Report No. HI-941184,NRC Docket No. 72-1008) is less than that assumed in this analysis for the TranStor canister. A leak rate of 1.0 E-4 cc/sec was used for the TranStor canister under the hypothetical accident conditions. Sufficiently low leak rate test criteria will be established for TranStor canisters to be stored at the PFSF to assure that the leak rate from these canisters will not exceed 1.0 E-4 cc/sec under hypothetical accident conditions.
The radionuclide inventory for the TranStor BWR canister was based on 61 design basis BWR fuel assemblies (GE 8X8) with a burnup of 40,000 mwd /MTU,6 years cooling time, PFSF Safety RAI No. 2, SAR 7-1 Page 1 of 4
T and 2.95% enrichment. His is conservative, since fuel with these characteristics is too " hot" for shipment to the PFSF as indicated in Figure 5.0-1 of the TranStor Shipping Cask SAR (SNC-95-71SAR, Docket No. 71-9268), as well as the technical specifications in Chapter 12 of the Hi-STORM S AR. This ensures that the inventory used in the calculation exceeds that of fuel authorized for storage at the PFSF. The inventory of isotopes other than "Co was calculated with the SAS2H and ORIGEN-S modules of the SCALE 4.3 system. Isotopes that contribute greater than 0.1% to the total curie inventory for the fuel assembly and iodine were considered.
2 The "Co inventory was determined using the 140 pCi/cm cmd surface activity for PWR 2
rods and the 1254 pCi/cm crud surface activity for BWR rods provided in NUREG/CR-2 2
6487, multiplied by the surface area per assembly (3 E5 cm and i E5 cm for PWR and BWR rods respectively), also provided in NUREG/CR-6487. The "Co source terms were then decay corrected to account for the cooling time, using the half life of"Co.
The activity released is estimated as the product of: 1) the estimated activity per fuel assembly,2) the number of fuel assemblies contained in one canister,3) the fraction of the canister volume released per second,4) the release fraction (by radionuclide group), and 5) the accident duration. He hypothetical accident duration is assumed to be 30 days. Items 1 and 2 were provided by the cask vendors. Item 3 was calculated by dividing the canister release rate under accident conditions by the canister free gas volume, which was also provided by the cask vendors. Items 4 and 5 are based on the NRC regulatory guidance provided in NUREG/CR-6487 and ISG-5. No credit was taken for holdup (plateout, depositon, etc) of paniculates or volatile fission products released from the fuel inside the canister.
The primary approach used in this analysis is to estimate inhalation committed effective dose equivalents for the airbome pathway, since it has been noted by the NRC that, for accident conditions, for all materials of greatest imerest for fuel cycle and other radioactive material liceases, the dose from the inhalation pathway will dominate the (overall) dose (NUREG-1140). The approach of conducting inhalation dose estimates is also consistent with guidance provided in NUREGICR-6410 (Nuclear Fuel Cycle Facility Accident Analysis Handbook, l
1998) and in DOE-HDBK-3010-94 (Airbome Release Fractions / Rates and Respirable 1
Fractionsfor Nonreactor Nuclear Facilities,1994). However, as a verilication of this approach, three additional calculations are performed: 1) an estimate of the doses from submersion in the plume following the accident (to calculate the TEDE from the release),2) l an estimate of thyroid doses from radioiodine in the plume, and 3) an estimate of the doses from environmental pathways following deposition of material from the plume using the I
RESRAD computer program.
The revised accident analysis evaluates the potential mhalation dose to an individual located at two distances downwind: at 500 m (representing the nearest distance from a canister to the OCA fence), and at 3,219 m (representing the location of the nearest resident). For these revised calculations, the respirable fraction of the material released for all radionuclides is assumed to be 1.0. Inhalation committed effective dose equivalent factors and extemal dose conversion factors for submersion in air for the radionuclides that are greater than 0.1% of the PFSF Safety RAI No. 2, SAR 7-1 Page 2 of 4
activity present in the fuel (plus radioiodine) were obtained from the EPA Federal Guidance Report Nos. I1 (1988)and 12 (1993). No correction is made for the amount of time the wind blows in a given direction over the 30-day release period. The accide:tt X/Q values were i
estimated using a wind speed of 1m/s and atmospheric stability Class F, which is consistent with the guidance in ISG-5. The resulting values are: 1.94E-3 s/m'at 500 m downwind,and 9.42E-5 s/m' at 3,219 m downwind.
1 The release and dose estimates for plume passage were conducted usmg simple calculations and spreadsheet software. A printout of the calculations is shown in Tables 1 through 8 (attached) for downwind distances of 500 m and 3,219 m. For 500 m downwind, Table 1 l
shows the resulting inhalation CEDE as 74.7 mrem /y and Table 2 shows the effective dose from extemal exposure during submersion in the plume as 0.153 mrem. The resulting TEDE at 500 m downwind is 74.9 mrem /y, as shown in Table 3. The estimated dose to thyroid from I-129 at 500 m downwind is 0.0234 mrem, as shown in Table 4. For 3,219 m downwind, Table 5 shows the inhalation CEDE as 3.63 mrem /y, and Table 6 shows the effective dose from extemal exposure during submersion in the plume as 0.00743 mrem.
The resulting TEDE at 3,219 m downwind is 3.64 mrem /y, as shown in Table 7. He estimated dose to thyroid from I-129 at 3,219 m downwind is 0.00114 mrem, as shown in Table 8. The radionuclide that contributes the largest amount to the TEDEs in the radionuclide mixture is Co-60 at both downwind distances. Both of the estimated TEDEs are a small fraction of the 0.05 Sv (5 rem) accident limit imposed by 10 CFR 72.106 (b) (i.e.,75 mrem /y is about 1.5% of the 5 rem limit). In addition, the estimated thytoid doses are a small fraction of the 0.5 Sv (50 rem) individual organ limit from 10 CFR 72.1%(b)(i.e.,0.0234 mrem is a very small fraction of the 50 rem limit). Because of the small doses that result from the accidental releases, and because these doses are a small fraction of the regulatory limit, it is obvious that doses to the eye, skin, extremities, and intemal organs would not exceed their respective limits of 0.15 Sv (15 rem) and 0.5 Sv (50 rem).
As an evaluation of the potential doses from environmental pathways following deposition of material in the plume, a pathway analysis using the RESRAD computer program was next conducted. The first step of this evaluation was to estimate the amount ofmaterial deposited on the ground from the plume. This estimate was made assuming that the effluent concentration in a given sector is uniform across the sector at a given distance, as described in Regulatory Guide 1.111 (1977). Using a straight-line trajectory model, this approach requires that the relative deposition rate should be divided by the arc length of the sector at the given downwind distance being considered to estimate deposition. The values of relative deposition (m) were obtained from Figure 6 of Regulatory Guide 1.111, with resulting values of 8.0 E-5 mi' at 500 m, and 2.3E-5 m at 3,219 m, downwind. As shown in Tables I and 5, the deposition estimates were made for each of the radionuclides in the source term.
2 Dese values,in units of pCi/m, were next modified to units of pCi/g to match the input requirements of the RESRAD computer program, by assuming a soil density of 1.5 E+6 g/m' and an effective soil depth ofI cm.
He exposure scenario considered in the RESRAD analysis includes direct exposure to contaminated ground, inhalation of resuspended radioactive material, ingestion of milk and beef following grazing, and ingestion of soil. This scenario is considered to be a conservative PFSF Safety RAI No. 2, SAR 7-1 Page 3 of 4
e representation of the land use conditions and environment of the land surrounding the PFSF.
Since the 500 m downwind location is considered to be along the OCA fence line, it is not j
possible for an individual to continuously occupy this location. Therefore, for purposes of calculation, an exposure duration of 2,000 h/y is assumed at 500 m downwind. Although natural vegetation at the facility is quite sparse,it is conservatively assumed that the RESRAD default values for fodder intake are met both for the dairy and beef cattle. Default values for human consumption shown in RESRAD for air, milk, beef, and soil were 3
assumed. The default values include inhalation of 1,918 m of air (over 2,000 h/y) with a 3
mass loading factor for air of 2.0E-4 g/m, ingestion of 92 Uy of milk, ingestion of 63 kg/y of beef, and ingestion of 36.5 g/y of soil. The same scenario is evaluated at a downwind distance of 3,219 m, except that continuous exposure (8760 h/y) is assumed since this is the location of the nearest resident. The resulting TEDEs for these accident cases were: 2.67 mrem /y at 500 m downwind, and 0.522 mrem /y at 3,219 m downwind. Both of these doses are quite small compared to the 0.05 Sv (5 rem) accident limit imposed by 10 CFR 72.106(b). The dominant exposure pathway is extemal exposure to contaminated land and the radionuclide with the largest contribution to the dose is Co-60. From this analysis,it is concluded that these doses are sufficFemly small compared to the inhalation TEDEs from plume passage (about 4% at 500 m and about 14% at 3,219 m) that they canjustifiably be ignored in the accident analysis.
Finally, the doses presented here are likely overestunates of the doses that would potentially result from the estimated airborne releases over a 30-day period since this analysis assumes that the wind blows in a constant direction for 30 days. Variation of wind direction over the release period would reduce the magnitude of the estimated doses downwind.
References:
- DOE-HDBK-3010-94. December 1994. Airborne Release Fractions / Rates and Respirable FractionsforNonreactorNuclear Facilities. U.S. Department of Energy, Washington, D.C.
Federal Guidance Report No.12. September 1993. ExternalExposure to Radionuclides in Air, Water, andSoil. U.S. Environmental Protection Agency, Washington, D.C.
Federal Guidance Report No. I 1. September 1988. Limiting Values ofRadionuclide intake and Air Concentration and Dose Conversion Factorsfor Inhalation, Submersion, andIngestion. U.S. Environmental Protection Agency, Washington, D.C.
NUREG-1140 (McGuire, S.). January 1988. A RegulatoryAnalysis on Emergency Preparednessfor Fuel Cycle and Other Radioactive Material Licensees, U.S. Nuciear Regulatory Commission, Washington D.C.
NUREGICR-6410. March 1998. Nuclear Fuel Cycle Facility Accident Analysis Handbook. Prepared for the U.S. Nuclear Regulatory Commission by Science Applications Intemational Corporation, Washmgton, D.C.
Regulatory Guide 1.111. July 1977. Methodsfor Estimating Atmospheric Transport and Dispersion ofGaseous Ejluents in Routine Releasesfrom Light-Water-CooledReactors.
U.S. Nuclear Regulatory Commission, Washington, D.C.
PFSF Safety RAI No. 2, SAR 7-1 Page 4 of 4
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w 00CKETED pc u p e' UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 9 APR 23 P c 22 OT.
Before the Atomic Safety and Licensing Board M.
ADM' F
In the Matter of
)
)
PRIVATE FUEL STORAGE L.L.C.
)
Docket No. 72-22
)
(Private Fuel Storage Facility)
)
CERTIFICATE OF SERVICE I hereby certify that copies of the Applicant's Motion for Summary Disposition Of Utah Contention C - Failure to Demonstrate Compliance with NRC Dose Limits, Statement Of Material Facts on Which No Genuine Dispute Exists and Affidavit of Wil-liam Hennessy were served on the persons listed below (unless otherwise noted) by e-mail with conforming copies by U.S. mail, first class, postage prepaid, this 21st day of April 1999.
G. Paul Bollwerk III, Esq., Chairman Ad-Dr. Jerry R. Kline ministrative Judge Administrative Judge Atomic Safety and Lice # sing Board Panel Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Washington, D.C. 20555-0001 e-mail: GPB@nrc. gov e-mail: JRK2@nrc. gov Dr. Peter S. Lam
- Susan F. Shankman Administrative Judge Deputy Director, Licensing & Inspection Atomic Safety and Licensing Board Panel Directorate, Spent Fuel Project Office U.S. Nuclear Regulatory Commission Office of Nuclear Material Safety &
Washington, D.C. 20555-0001 Safeguards e-mail: PSL@nrc. gov U.S. Nuclear Regulatory Commission Washington, D.C. 20555
,e l
i l
I Office of the Secretary
- Adjudicatory File l
U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Panel Washington, D.C. 20555-0001 U.S. Nuclear Regulatory Commission Attention: Rulemakings and Adjudications Washington, D.C. 20555-0001 Staff e-mail: hearingdocket@nrc. gov (Original and two copies)
Catherine L. Marco, Esq.
Denise Chancellor, Esq.
Sherwin E. Turk, Esq.
Assistant Attorney General Office of the General Counsel Utah Attorney General's Office Mail Stop O-15 B18 160 East 300 South,5* Floor U.S. Nuclear Regulatory Commission P.O. Box 140873 Washington, D.C. 20555 Salt Lake City, Utah 84114-0873 e-mail: pfscase@nrc. gov e-mail: dchancel@ state.UT.US John Paul Kennedy, Sr., Esq.
Joro Walker, Esq.
Confederated Tribes of the Goshute Land and Water Fund of the Rockies Reservation and David Pete 165 South M '. Suite 1 1385 Yale Avenue Salt Lake Cit UT 84111 Salt Lake City, Utah 84105 e-mail: joro61@inconnect.com e-mail: john @kennedys.org Diane Curran, Esq.
Danny Quintana, Esq.
Harmon, Curran, Spielberg &
Skull Valley Band of Goshute Indians Eisenberg, L.L.P.
Danny Quintana & Associates, P.C.
2001 S Street, N.W.
50 West Broadway, Fourth Floor Washington, D.C. 20009 Salt Lake City, Utah 84101 e-mail:DCurran.HCSE@zzapp.org e-mail: quintana @xmission.com By U.S. mail only 1
Paul A. Gaukler l
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