[Federal Register Volume 64, Number 54 (Monday, March 22, 1999)]
[Notices]
[Pages 13828-13829]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-6909]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 72-20]


Department of Energy, Idaho Operations Office; Issuance of 
Environmental Assessment and Finding of No Significant Impact Regarding 
the Proposed Exemption From Certain Regulatory Requirements of 10 CFR 
Part 72

    The U.S. Nuclear Regulatory Commission (NRC or Commission) is 
considering issuance of an exemption, pursuant to 10 CFR 72.7, from the 
requirements of 10 CFR 72.102(f)(1) to the U.S. Department of Energy, 
Idaho Operations Office (DOE-ID or applicant). Exemption from 10 CFR 
72.102(f)(1) would relieve DOE-ID from the requirements to use a design 
earthquake (DE) ground motion equivalent to that of a safe shutdown 
earthquake (SSE) for a nuclear power plant, as evaluated by the methods 
of Appendix A of Part 100 for its proposed Independent Spent Fuel 
Storage Installation (ISFSI). The proposed ISFSI is to be located at 
the Idaho National Engineering and Environmental Laboratory (INEEL), 
within the Idaho Nuclear Technology and Engineering Center (INTEC) site 
in Scoville, Idaho. The proposed ISFSI would store the spent nuclear 
fuel debris created as a result of the Three Mile Island Unit 2 (TMI-2) 
accident.

Environmental Assessment (EA)

Identification of Proposed Action

    The applicant is seeking Commission approval to construct and 
operate an ISFSI at INTEC. INTEC is an existing facility initially 
constructed to both store and reprocess spent fuel and high-level waste 
possessed by DOE. Pursuant to 10 CFR part 72, DOE-ID submitted an 
application, including a Safety Analysis Report (SAR), for the ISFSI, 
by letter dated October 31, 1996, as supplemented. NRC staff is 
currently performing a review of that application. On September 15, 
1997, DOE-ID requested an exemption from the requirement of 10 CFR 
72.102(f)(1) which states: ``For sites that have been evaluated under 
the criteria of appendix A of 10 CFR part 100, the design earthquake 
(DE) must be equivalent to the safe shutdown earthquake (SSE) for a 
nuclear power plant.'' In this context, ``DE'' and ``SSE'' refer to the 
design peak ground acceleration (PGA), with an appropriate response 
spectrum, caused by the largest credible earthquake. The most recent 
deterministic seismic hazard analysis for the ISFSI site, completed in 
accordance with appendix A of part 100, yields a DE of 0.56 g PGA. 
However, DOE-ID proposes a DE with a 0.36 g PGA as an adequately 
conservative seismic design for the ISFSI.
    The staff is considering granting the requested exemption from 10 
CFR 72.102(f)(1).

Need for the Proposed Action

    The applicant is preparing to build and operate the TMI-2 ISFSI as 
described in its application and SAR, subject to approval of the 
pending licensing application. Specifically, DOE is concerned with 
designing low risk facilities, such as an ISFSI, to the requirements of 
10 CFR part 100, appendix A, as it would set precedent that appears to 
be unnecessary, technically inappropriate, and potentially unattainable 
throughout the DOE complex. The DOE-ID seismic hazard analysis meeting 
the requirement of 10 CFR 72.102(f)(1) yields a DE of 0.56 g PGA, with 
an appropriate response spectrum, for the ISFSI site. DOE-ID proposes a 
DE of 0.36 g PGA, with an appropriate response spectrum. DOE-ID 
justifies this value with a site-specific radiological risk analysis.
    In response to DOE's September 15, 1997, letter requesting this 
exemption, the staff prepared a safety evaluation report which was 
forwarded to the Commission as an attachment to SECY-98-071 (April 8, 
1998). In that paper, the staff recognized that although 10 CFR part 72 
does not currently allow PSHA e.g., ``risk-based,'' as an acceptable 
methodology for deriving a DE for an ISFSI, the PSHA results are being 
accepted by NRC in other licensing actions. The PSHA method is 
acceptable for nuclear power plants under the January 1997 revisions to 
10 CFR parts 50 and 100. Furthermore, NRC has accepted the PSHA method 
for the design and performance assessment for the proposed high-level 
waste repository at Yucca Mountain. On May 20, 1998, the Commission 
informed the staff that it did not object to the proposed exemption.
    A complete safety evaluation is available as part of SECY-98-071. 
In summary, it found that when 10 CFR part 72 was first promulgated in 
1980, ISFSIs were largely envisioned to be either spent fuel pools or 
single, massive dry storage structures. Given the potential accident 
scenarios, a DE equivalent to a nuclear power plant SSE seemed 
appropriate for these facilities. Furthermore, for ISFSIs to be located 
at a nuclear power plant, the DE value was readily available without 
additional site characterization work, save the geotechnical 
investigation at the specific ISFSI location. However, an ISFSI storing 
spent fuel in dry casks or canisters is inherently less hazardous and 
less vulnerable to earthquake-initiated accidents than an operating 
nuclear power plant. NRC recognized this in the initial part 72, 
``Statements of Consideration,'' and stated that the DE for cask and 
canister technology need not be as high as a nuclear power plant SSE: 
``For ISFSIs which do not involve massive structures, such as dry 
storage casks and canisters, the required design earthquake will be 
determined on a case-by-case basis until more experience is gained with 
licensing these types of units.'' The staff believes that this 
experience has been gained over the past 13 years of ISFSI operations.

Environmental Impacts of the Proposed Action

    The ``Final Environmental Impact Statement (FEIS) for the 
Construction and Operation of the TMI-2 Independent Spent Fuel Storage 
Installation,'' NUREG-1626 (March 1998), considered the potential 
environmental impacts of licensing this facility, including potential 
accidents during storage. A description of the potential accidents 
during storage is provided in Section 4.1.2.7.3 of NUREG-1626.
    An ISFSI is designed to mitigate the effects of design basis 
accidents that could occur during storage. Design basis accidents 
account for human-caused events and the most severe natural phenomena 
reported for the site and surrounding area. Postulated accidents 
analyzed for an ISFSI include tornado winds and tornado generated 
missiles, design basis earthquakes, design basis floods, accidental 
cask drops, lightning effects, fires, explosions, and other incidents.
    Special ISFSI design features include using nonflammable materials, 
providing a horizontal storage module with walls and a roof of 
structural steel and reinforced concrete (approximately 2.5 feet (0.76 
meter) thick) to house a dry-shielded steel canister, and a passive 
ventilation system. Considering the specific design requirements for 
each accident condition, the design of the ISFSI would prevent loss of

[[Page 13829]]

containment, shielding, or criticality control.
    The bounding consequences of a major seismic event at an ISFSI 
using the NUHOMS system technology are limited by a canister drop onto 
the concrete pad, although this would occur only at a ground motion 
well above the proposed 0.36 g PGA design value, as detailed in Section 
8.2.3.2 of the TMI-2 ISFSI SAR. The casks and canisters are designed to 
withstand such events with no release of radioactive material. The 
effects of a NUHOMS canister drop are analyzed in Section 8.2.5.2 of 
the SAR. In addition, analysis of beyond-design basis accidents leading 
to cask or canister rupture estimate off-site doses well below the 0.05 
Sv (5 rem) whole body dose limit of 10 CFR 72.106(b). In a letter dated 
July 19, 1996, DOE-ID presented a conservative analysis of off-site 
doses resulting from a beyond-design basis accident. In this 
hypothetical accident, for which neither DOE-ID nor the staff has 
identified a credible mechanism, both a NUHOMS dry shielded canister 
and one of the 12 inner core debris canisters are assumed to fail, 
allowing unmitigated dispersal of the contents. The calculated off-site 
dose from such an accident is 0.75 mSv (75 mrem), well below the 0.05 
Sv (5 rem) siting evaluation factor of 10 CFR 72.106(b).
    DOE-ID has completed both a Deterministic Seismic Hazard Analysis 
(DSHA) (Appendix A of Part 100) and PSHA (10 CFR 100.23) for the ISFSI 
site. The staff has evaluated these analyses and finds the resultant 
values acceptable: 0.56 g PGA for an SSE by the deterministic method 
and 0.30 g PGA mean ground motion with a 2000-year return period by the 
probabilistic method. The staff finds acceptable the risk-graded 
approach to seismic hazard characterization and design in DOE Standard 
1020, which is similar to the risk-graded approach of using the 2000-
year return period mean ground motion as the DE is adequately 
conservative. Moreover, the expected life span of the ISFSI, 20 years 
with the possibility of renewal, per 10 CFR 72.42, justifies use of 
this ground motion as the DE. The DE proposed by DOE-ID for the ISFSI, 
0.36 g PGA with an appropriate response spectrum exceeds the 0.30 g PGA 
value for the 2000-year return period mean ground motion. Therefore, 
the staff concludes that granting the requested exemption from 10 CFR 
72.102(f)(1) will maintain an adequate design margin for seismic events 
and will not be inimical to public health and safety.

Alternatives to the Proposed Action

    Since there are no significant environmental impacts associated 
with the proposed action, any alternatives with equal or greater 
environmental impact are not evaluated. The alternative to the proposed 
action would be to deny approval of the 10 CFR 72.102(f)(1) exemption 
and require that DOE design the facility to withstand the effects of a 
higher PGA. This alternative would have no significant environmental 
impact as well.

Agencies and Persons Consulted

    On March 1, 1999, Mr. Alan Merritt from the State of Idaho, INEEL 
Oversight Program, was contacted about the EA for the proposed action 
and had no concerns.

Finding of no Significant Impact

    The environmental impacts of the proposed action have been reviewed 
in accordance with the requirements set forth in 10 CFR part 51. Based 
upon the foregoing EA, the Commission finds that the proposed action of 
granting an exemption from 10 CFR 72.102(f)(1), given the absence of 
radiological consequences from any credible seismic event, will not 
significantly impact the quality of the human environment. Accordingly, 
the Commission has determined not to prepare an environmental impact 
statement for the proposed exemption.
    The staff finds acceptable the risk-graded approach to seismic 
hazard characterization and design in DOE Standard 1020, which is 
similar to the risk-graded approach to design basis events in 10 CFR 
part 60. Given the absence of radiological consequences from any 
credible seismic event, the staff finds that the DOE Standard 1020 
risk-graded approach of using the 2000-year return period mean ground 
motion as the DE is adequately conservative. Moreover, the expected 
life span of the ISFSI, 20 years with the possibility of renewal, per 
10 CFR 72.42, justifies use of this ground motion as the DE. The DE 
proposed by DOE-ID for the ISFSI, 0.36 g PGA with an appropriate 
response spectrum, exceeds the 0.30 g PGA value for the 2000-year 
return period mean ground motion. Therefore, the staff concludes that 
granting the requested exemption from 10 CFR 72.102(f)(1) will maintain 
an adequate design margin for seismic events and will not be inimical 
to public health and safety.
    This application was docketed under 10 CFR part 72, Docket 72-20. 
For further details with respect to this action, see the application 
for an ISFSI license dated October 31, 1996, and the request for 
exemption dated September 15, 1997, which is available for public 
inspection at the Commission's Public Document Room, 2120 L Street, NW, 
Washington, DC 20555 and the Local Public Document Room at the INEEL 
Technical Library, 1776 Science Center Drive, Idaho Falls, ID 83402.

    Dated at Rockville, Maryland, this 13th day of March 1999.

    For the Nuclear Regulatory Commission.
E. William Brach,
Director, Spent Fuel Project Office, Office of Nuclear Material Safety 
and Safeguards.
[FR Doc. 99-6909 Filed 3-19-99; 8:45 am]
BILLING CODE 7590-01-P