[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
