[Federal Register Volume 63, Number 176 (Friday, September 11, 1998)]
[Notices]
[Pages 48768-48770]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-24461]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-309]


Maine Yankee Atomic Power Company, Maine Yankee Atomic Power 
Station; Exemption

I

    Maine Yankee Atomic Power Company (MYAPCo or the licensee) is the 
holder of Facility Operating License No. DPR-36, which authorizes 
possession of Maine Yankee Atomic Power Station (Maine Yankee). The 
license provides, among other things, that the facility is subject to 
all rules, regulations, and orders of the U.S. Nuclear Regulatory 
Commission (NRC or the Commission) now or hereafter in effect. The 
facility is a pressurized-water reactor (PWR) located on the licensee's 
site in Lincoln County, Maine. On August 7, 1997, the licensee 
submitted written certifications to the Commission that it had decided 
to permanently cease operations at Maine Yankee and that all fuel had 
been permanently removed from the reactor. In accordance with 10 CFR 
50.82(a)(2), upon docketing of the certifications contained in the 
letter of August 7, 1997, the facility operating license no longer 
authorizes MYAPCo to operate the reactor or to place fuel in the 
reactor vessel.

II

    Section 50.54(q) of Title 10 of the Code of Federal Regulations (10 
CFR 50.54(q)) requires power reactor licensees to follow and maintain 
in effect emergency plans that meet the standards of 10 CFR 50.47(b) 
and the requirements of Appendix E to 10 CFR Part 50.
    Pursuant to 10 CFR 50.12(a), the Commission may, upon application 
by any interested person or upon its own initiative, grant exemptions 
from the requirements of the regulations that are (1) authorized by 
law, will not present an undue risk to public health and safety, and 
are consistent with the common defense and security and (2) present 
special circumstances. Special circumstances exist when application of 
the regulation in the particular circumstance would not serve the 
underlying purpose of the rule or is not necessary to achieve the 
underlying purpose of the rule (10 CFR 50.12(a)(2)(ii)). The underlying 
purpose of Section 50.54(q) is to ensure licensees follow and maintain 
in effect emergency plans that provide reasonable assurance that 
adequate protective measures can and will be taken in the event of an 
emergency at a nuclear reactor. Sections 50.47(b) and (c) outline the 
planning standards and size of Emergency Planning Zones, respectively, 
that are to be considered in emergency plans and Appendix E to 10 CFR 
Part 50 identifies the information that must be included in emergency 
plans.

III

    By letter dated November 6, 1997, the licensee requested exemptions 
from certain requirements of 10 CFR 50.54(q), 10 CFR 50.47(b) and (c), 
and Appendix E to Part 50; the licensee also made available a draft 
copy of the Maine Yankee Defueled Emergency Plan (DEP) to assist the 
staff in its review of the exemption request. The exemptions would 
allow Maine Yankee to discontinue certain aspects of offsite planning 
and reduce the scope of onsite emergency planning. The licensee stated 
that the remaining requirements of 10 CFR 50.54(q), 10 CFR 50.47(b) and 
(c), and Appendix E to Part 50 will be addressed in the DEP. The 
licensee plans to implement the DEP without NRC review and approval. 
Under the provisions of Sec. 50.54(q), when a change to an emergency 
plan is made, the staff evaluates that change against the bases for 
commitments made in the plan to determine whether there is a decrease 
in effectiveness. It is not a decrease in effectiveness if the 
reduction in the commitment is commensurate with a reduction in the 
bases for that commitment. In this instance, the staff has determined 
that there has been a reduction in the bases that require offsite 
emergency planning. The revised DEP will be reviewed by the NRC after 
implementation. By letter dated March 25, 1998, the licensee submitted 
the Emergency Action Levels that it proposes to use with the Defueled 
Emergency Plan. By letter dated June 29, 1998, the licensee submitted 
additional information that revised the exemption request. By letters 
dated January 20, May 15, and June 18, 1998, MYAPCo submitted the 
results of an assessment of the Maine Yankee spent fuel heatup in the 
absence of water in the spent fuel pool. By letters dated July 9 and 
August 5, 1998, the licensee provided the results of radiological 
analyses applicable to Maine Yankee in the permanently shutdown 
condition.
    The licensee stated that special circumstances are present at Maine 
Yankee because (1) application of the regulation in the particular 
circumstances would not serve the underlying purpose of the rule or is 
not necessary to achieve the underlying purpose of the rule, (2) 
compliance would result in undue hardship or other costs that are 
significantly in excess of those contemplated when the regulation was 
adopted, or are significantly in excess of those incurred by others in 
similar circumstances, and (3) there is a material circumstance 
present, that was not considered when the regulation was adopted, for 
which it would be in the public interest to grant an exemption.
    With the plant in a permanently shutdown and defueled condition, 
the applicable design-basis accidents are limited to a fuel handling 
incident, spent fuel cask drop, and radioactive liquid waste system 
leak and failure. The calculated maximum offsite dose from these 
postulated releases is less than the U.S. Environmental Protection 
Agency (EPA) Protective Action Guides (PAGs). The licensee also 
estimated that, by March 1998, a beyond-design-basis event, involving 
fuel damage (caused by a loss of spent fuel pool water and a subsequent 
overheating of the stored fuel) and the release of radioactive 
materials sufficient to exceed EPA PAGs at the site boundary is not 
credible.
    Revision 14 to the Maine Yankee Defueled Safety Analysis Report 
(DSAR) includes revised analyses of postulated accidents at Maine 
Yankee in its permanently shutdown status. Chapter 5 of the DSAR 
describes the radiological consequences of accidents that could release 
radioactive materials and the consequences of a spent fuel pool

[[Page 48769]]

draindown event. The staff reviewed the licensee's analyses, as 
modified in licensee submittals dated July 9 and August 5, 1998, to 
determine whether the radiological impact of these events would require 
an offsite emergency plan.
    Decontamination of systems during decommissioning and dismantlement 
operations will generate significant quantities of radioactive waste in 
the form of contaminated demineralizer resins. The licensee has 
postulated a bounding accident for the release of radioactivity: the 
dropping of a highly loaded spent resin liner within the low-level-
waste storage building (LLWSB), resulting in the liner failure and a 
release of a fraction of its radioactive materials in an airborne 
cloud. The analysis indicates that an individual at the exclusion area 
boundary (EAB) could receive up to 0.11 rem total effective dose 
equivalent (TEDE) from this event.
    The licensee stated that this event was considered to have higher 
offsite consequences than the mishandling of resin during resin liner 
filling and dewatering operations since these activities are performed 
in containment. Hold-up and confinement of radioactive materials in a 
containment that is isolated would significantly decrease the potential 
for offsite release. In addition, the licensee committed in the DSAR to 
establish administrative controls to ensure that calculated offsite 
doses from potential decommissioning accidents do not exceed those 
calculated for a spent resin cask drop accident.
    The licensee did not postulate a fire concurrent with the resin 
mishandling event owing to the low flammability of the resin itself and 
the absence of flammable material in the LLWSB. However, the analysis 
did assume that 1.0 percent of the radioactivity in the liner became 
airborne during the event. This assumption is the same fraction of 
material expected to be released by a fire, and is consistent with the 
release fractions listed in Schedule C to 10 CFR 30.72 for mixed 
fission and corrosion products. The calculational methods and 
assumptions used in this analysis are acceptable to the staff.
    Wet storage of spent fuel possesses inherently large safety margins 
because of the simplicity and robustness of the spent fuel pool design. 
The design basis includes the ability to withstand an earthquake and to 
retain sufficient water to adequately cool and shield the stored spent 
fuel. Specifically, in the DSAR, the licensee states that the spent 
fuel pool structure is designed to Seismic Class I requirements and is 
capable of performing its intended safety function under the licensee's 
design-basis hypothetical earthquake with a 0.1-g peak ground 
acceleration. The pool has 6-foot reinforced-concrete walls and floor 
with a \1/4\-inch steel liner. To add to the robustness of the design, 
the pool is founded on bedrock and is embedded 12.5 feet below grade 
level, which is at the 20 foot, 1 inch elevation. Since the analyses 
used in designing the capability of structures, systems, and components 
(SSCs) to perform their safety function under a hypothetical earthquake 
have significant margin in them, it is expected that an SSC built to 
withstand the hypothetical design-basis earthquake actually will be 
able to withstand a larger earthquake. Thus, the loss of coolant from 
the Maine Yankee spent fuel pool, which partially or completely 
uncovers the fuel, is a beyond-design-basis event with a very low 
probability of occurrence.
    In a letter dated May 15, 1998, the licensee submitted analyses for 
a complete loss of inventory and several partial loss-of-inventory 
events within the spent fuel pool. That analysis showed that a partial 
draindown was more severe than a complete draindown for the licensee's 
plant. For this case, only 5.5 feet of the active fuel is covered by 
water. The licensee calculated that it would take 30 hours for the 
cladding to heat up to 827  deg.C. However, the staff reviewed the 
calculations and determined that the bounding scenario would be one 
with the active fuel totally uncovered and water blocking the assembly 
lower inlet so that no natural circulation flowpath exists. The staff 
calculated that, for this case, as of August 1, 1998, it would take 
approximately 10 hours for the hottest location in the highest power 
assembly to reach 900  deg.C. The heatup time was calculated assuming 
an adiabatic heatup of a fuel rod and using conservative decay heat 
assumptions. An adiabatic heatup is defined as one in which all heat 
generated is retained in the system, with no heat loss to the 
surroundings. This definition corresponds to a physical situation in 
which the spent fuel pool water is lost, no cooling mechanism is 
available, and the fuel is surrounded by a perfect insulator. The staff 
considers that this scenario would be bounding for any loss-of-
inventory scenario since any other scenario would have some heat 
removal from the assembly and a longer heatup time. Consequently, the 
staff determined that, in view of the low likelihood of the bounding 
scenario, and the time elapsed since the shutdown of the facility, 
there would be sufficient time for mitigative actions and, if 
necessary, offsite protective measures to be initiated after a 
postulated loss of water and before a postulated release of 
radioactivity resulting from spent fuel overheating.
    In the event that spent fuel pool water inventory is lost more 
gradually through the method discussed above or through some other 
means, such as a siphon or liner leak, plant personnel have various 
methods for detecting the loss of inventory. The staff reviewed these 
methods, which include indicators to alert and assist in identifying 
any loss of coolant inventory. The design includes a low coolant level 
indicator and an area radiation monitor, both of which alarm in the 
control room. Although not credited for accident mitigation, these 
alarms provide methods to alert the operators to a loss-of-inventory 
event. In the DSAR, the licensee also states that there are several 
sources of makeup water to the spent fuel pool. Among these sources are 
the normal sources of makeup water from the refueling water storage 
tank, demineralizer water from the primary water storage tank, 
emergency sources from the fire water system, and potable water from 
the town of Wiscasset water supply system. On the basis of indicators 
and alarms available to plant personnel and the availability of makeup 
sources to restore a gradual loss of coolant, the staff finds it 
reasonable to expect that fuel uncovery as a result of a gradual loss 
of coolant scenario is highly unlikely.
    Although the event is unlikely, the licensee evaluated the dose 
consequences of both partial and complete spent fuel pool draindown. 
Water and the concrete pool structure provide radiation shielding on 
the sides of the pool. However, water alone accounts for most of the 
shielding above the spent fuel. A loss of shielding above the fuel 
could increase the radiation levels at the exclusion area boundary 
(EAB) due to the scattering of gamma rays streaming up out of the pool. 
The licensee postulated a partial pool draindown event resulting from a 
break in the pool cooling system piping, concurrent with a failure of 
the associated anti-syphon device. The licensee assumed that additional 
pool water was lost through pool boiling for the following four days 
before effective corrective actions could be taken to reestablish 
adequate pool water level. The licensee calculated that the dose rate 
was 0.00076 rem per hour at the EAB. In addition the licensee 
calculated the postulated offsite dose rates in the event of a complete 
draindown of the spent fuel pool (a beyond-design-basis event). 
Assuming only one year of

[[Page 48770]]

radioactive decay and a site boundary distance of 610 meters, the 
complete draindown resulted in a postulated dose rate of 0.01 rem per 
hour. The licensee's calculated dose rate indicates it would take 4.1 
days for this event to exceed the EPA early-phase PAG of 1 rem.
    The staff concludes that the licensee's request for an exemption 
from certain requirements of 10 CFR 50.54(q), 10 CFR 50.47(b) and (c), 
and Appendix E to Part 50 is acceptable in view of the greatly reduced 
offsite radiological consequences associated with the current plant 
status. The staff finds that the postulated dose to the general public 
from any reasonably conceivable accident would not exceed EPA PAGs and, 
for the bounding accident, the length of time available gives 
confidence that offsite measures for the public could be taken without 
preplanning. The staff finds acceptable the licensee's commitment in 
the DSAR to establish administrative controls to ensure that calculated 
offsite doses from potential decommissioning accidents do not exceed 
those determined for a spent resin cask drop accident. Therefore, the 
staff concludes that the requirement that emergency plans meet all of 
the standards of 10 CFR 50.47(b) and all of the requirements of 
Appendix E to Part 50 is not now warranted at Maine Yankee and an 
exemption from the requirements for offsite emergency planning is 
acceptable.

IV

    The NRC staff has completed its review of the licensee's request 
for an exemption from the requirements of 10 CFR 50.47(c)(2) and from 
the requirements of 10 CFR 50.54(q), that emergency plans must meet all 
of the standards of 10 CFR 50.47(b) and all the requirements of 
Appendix E to 10 CFR part 50. The standards of 10 CFR 50.47(b) and the 
requirements of Appendix E to 10 CFR part 50 that remain in effect are 
listed in Attachment II to the licensee's letter dated June 29, 1998. 
On the basis of its review, the NRC staff finds that the postulated 
dose to the general public from any reasonably conceivable accident 
would not exceed EPA PAGs and, for the bounding accident, the length of 
time available provides confidence that offsite measures for the public 
could be taken without preplanning. The analyses submitted by the 
licensee are consistent with the commitment made in its DSAR, which 
stated that any decommissioning activities will be analyzed and 
administrative controls will be established to ensure that the 
calculated offsite doses do not exceed those determined for the spent 
resin cask drop accident. The staff finds the exemption from two 
requirements, 10 CFR 50.47(b)(9) and 10 CFR 50 Appendix E.IV.A.4, 
acceptable on the basis of the licensee's commitment to continue to 
maintain capabilities for dose assessment and personnel equivalent to 
those described in section 7.0 of the draft Defueled Emergency Plan 
provided in Attachment III to the licensee's letter dated November 6, 
1997. The information developed from the capability would be used to 
determine whether offsite measures for the general public would be 
appropriate. Maine Yankee will continue to maintain an onsite emergency 
preparedness organization capable of responding to the consequences of 
radiological events still possible at the site. Thus, the underlying 
purpose of the regulations will not be adversely affected by 
eliminating offsite emergency planning activities or reducing the scope 
of onsite emergency planning.
    For the foregoing reasons, the Commission has determined that, 
pursuant to 10 CFR 50.12, elimination of offsite emergency planning 
activities will not present an undue risk to public health and safety 
and is consistent with common defense and security. Further, special 
circumstances are present as stated in 10 CFR 50.12(a)(ii). Pursuant to 
10 CFR 51.32, the Commission has determined that this exemption will 
not have a significant effect on the quality of the human environment 
(63 FR 43968, August 17, 1998).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland this 3rd day of September 1998.

    For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 98-24461 Filed 9-10-98; 8:45 am]
BILLING CODE 7590-01-P