[Federal Register Volume 63, Number 194 (Wednesday, October 7, 1998)]
[Notices]
[Pages 53940-53943]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-26852]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-155]


Consumers Energy Company; Big Rock Point Nuclear Plant; Exemption

I

    Consumers Energy Company (Consumers or the licensee) is the holder 
of Facility Operating License No. DPR-6, which authorizes possession of 
the Big Rock Point Nuclear Plant (BRP). The license provides, among 
other things, that the facility is subject to all the rules, 
regulations, and orders of the U.S. Nuclear Regulatory Commission (the 
Commission or NRC) now or hereafter in effect. The facility consists of 
a boiling-water reactor (BWR) located on the licensee's site in 
Charlevoix County, Michigan. The licensee submitted written 
certification to the Commission on June 26, 1997, that it had decided 
to permanently cease operations at BRP and on September 23, 1997, that 
all fuel had been permanently removed from the reactor vessel. In 
accordance with 10 CFR 50.82(a)(2), upon docketing of the 
certifications contained in the letters of June 26 and September 23, 
1997, the facility operating license no longer authorizes Consumers to 
operate the reactor or place or retain 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 Section 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 exemption 
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 circumstances 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 that adequate protective 
measures can and will be taken in the event of a radiological emergency 
at a nuclear reactor. Sections 50.47(b) and (c) outline the planning 
standards and size,

[[Page 53941]]

respectively, of the Emergency Planning Zones 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 September 19, 1997, the licensee requested 
exemption from certain requirements in 10 CFR 50.47(b) and Appendix E 
to 10 CFR Part 50. The licensee also submitted and requested approval 
of its proposed BRP Defueled Emergency Plan (DEP), which was written on 
the basis of NRC staff approval of the proposed exemption request. The 
exemption would allow Consumers to discontinue certain aspects of 
offsite emergency planning and reduce the scope of onsite emergency 
planning.
    Under the provisions of Section 50.54(q), a licensee may make 
changes to emergency plans without Commission approval only if the 
changes do not decrease the effectiveness of the plans and if the 
plans, as changed, continue to meet the standards of Section 50.47(b) 
and the requirements of Appendix E to 10 CFR Part 50. When the licensee 
determines that such a change may reduce the effectiveness of the 
emergency plans, the NRC staff evaluates that change against the bases 
for commitments made in the plan to determine whether there is a 
decreased effectiveness. It is not a decrease in effectiveness if the 
reduction in the commitment is commensurate with a reduction in the 
basis 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 basis for this determination is, in part, that 
the permanently shutdown and defueled condition of the BRP facility 
represents a substantially reduced risk to public health and safety.
    The NRC reviewed the proposed BRP DEP as submitted, supplemented, 
and modified by the letters dated September 19, October 29, and 
November 20, 1997, and March 2, April 29, July 30, and August 28, 1998, 
during its review of the licensee's exemption request. The requirements 
of 10 CFR 50.54(q) and the remaining onsite and offsite requirements of 
10 CFR 50.47 and Appendix E to 10 CFR Part 50 are addressed in the BRP 
DEP. Consumers intends to implement the BRP DEP following NRC staff 
review and approval, as stated by the licensee in its application dated 
September 19, 1997.
    The licensee stated that special circumstances exist at BRP because 
the plant is permanently shutdown and defueled and the radiological 
source term at the site is reduced from that associated with reactor 
power operation. With the reactor power plant permanently shutdown and 
defueled, the design-basis accidents and transients postulated to occur 
during reactor operation are no longer possible. In particular, the 
potential for a release of a large radiological source term to the 
environment from the high pressure and temperature associated with 
reactor operation no longer exists. Additionally, due to the 
radioactive decay of short-lived isotopes, there is a continuing 
reduction in the potential radiological source term following the BRP 
plant shutdown on August 30, 1997. Further, the licensee also stated, 
during a public meeting held at NRC Headquarters on August 13, 1998, 
that requiring Consumers to comply with the requirements for offsite 
emergency planning when it is no longer warranted would result in undue 
financial hardship to BRP, its owners, and their ratepayers.
    With the plant in a permanently shutdown and defueled condition, 
Consumers has stated that following 68 days post-shutdown (November 5, 
1997) there are no remaining design-basis accidents at BRP that would 
result in offsite doses exceeding the U.S. Environmental Protection 
Agency (EPA) Protective Action Guides (PAGs). The accidents and 
transients evaluated by Consumers are described in Chapters 9 and 15 of 
the BRP Final Hazards Summary Report (FHSR), Revision 6, and included 
the evaluation of gap activity from the spent fuel that is postulated 
to be released to the environment as a result of fuel handling 
incidents and heavy load drops on spent fuel.
    Subsequently, on February 12, 1998, Consumers submitted Revision 7 
to its FHSR, which included revised analyses of postulated accidents at 
BRP in its permanently shutdown and defueled status. In Revision 7, 
Consumers reevaluated the accidents described in Revision 6 to the 
FHSR. Consumers also evaluated other postulated radiological events to 
gain further assurance that decommissioning activities would not result 
in unacceptable levels of risk of effects on public health from 
radiation exposure in an emergency situation and that these events are 
bounded by the considerations described in the NRC's ``Final Generic 
Environmental Impact Statement on Decommissioning of Nuclear 
Facilities'' (NUREG-0586). In particular, these other radiological 
events included but were not limited to the evaluation of (1) fire 
involving radioactive ion exchange resin; (2) gamma radiation due to a 
loss of spent fuel pool (SFP) water level; and, (3) self-sustaining 
oxidation of spent fuel zirconium cladding. With the exception of 
krypton-85, the noble gas and volatile radioactive nuclides residing 
within the spent fuel pin gap that contribute to the dose consequences 
of releases from operating reactors have decayed to negligible amounts. 
Further, the source term from low-level radioactive waste (including 
ion exchange resins) temporarily stored at the site is much lower than 
that of the spent fuel. Additionally, the licensee has demonstrated 
that the potential dose consequences of a release from a low-level 
radioactive waste (LLRW) are bounded by accidents involving spent fuel.
    By letter dated November 20, 1997, Consumers submitted its 
evaluation demonstrating the conclusion that a fire involving 
radioactive resin being stored at the facility and gamma radiation 
resulting from a complete draindown of the SFP would not exceed the EPA 
PAGs at the site area boundary. The resin fire is considered a bounding 
LLRW accident at the site. This fire would involve the ion exchange 
resin used to process wastes resulting from the reactor coolant system 
chemical decontamination that was performed at the BRP facility in 
December 1997. As a postulated scenario, Consumers estimated that the 
fire consumed resin containing 300 curies, which correlates to the 
amount of radioactive material that Consumers estimated will be 
retained in the resins from chemical decontamination. Consumers 
calculated that this event would result in a total effective dose 
equivalent (TEDE) and a thyroid committed dose equivalent (CDE) well 
below EPA PAGs. The staff reviewed the licensee's calculations and 
methodologies and found them to be acceptable. To provide further 
assurance that fires involving LLRW do not result in offsite doses 
exceeding EPA PAGs, the NRC staff assessed the current LLRW situation 
at BRP. The licensee informed the staff that as of July 28, 1998, five 
high-integrity containers (HICs) of radioactive resin are being stored 
in the LLRW storage building located on the BRP site. These HICs are 
loaded with approximately 100-150 curies of radioactive material from 
various reactor operating and decommissioning activities and are stored 
inside a corrugated metal building utilizing a separate concrete vault 
for each HIC. Manual fire protection and industrial area personnel 
access controls are associated with this building. Further, the 
licensee

[[Page 53942]]

maintains a fire protection program for its onsite facilities and 
continually assesses combustible loading to minimize fire potential and 
consequences. Therefore, the staff finds that a fire involving more 
than one HIC has a very low probability of occurrence.
    Wet storage of spent fuel possesses inherently large safety margins 
because of the simplicity and robustness of the SFP design. The design 
basis includes the ability to withstand an earthquake and to retain 
sufficient water to adequately cool and shield the spent fuel. 
Specifically, the licensee states in the FHSR that the SFP 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 0.05g acceleration. This value was 
reevaluated by the licensee to a Regulatory Guide 1.60, ``Design 
Response Spectra for Seismic Design of Nuclear Power Plants,'' value of 
0.12g zero-period horizontal acceleration. The SFP structure has a 
floor and walls of reinforced concrete that vary in thickness from 3 
feet 6 inches to 6 feet 9 inches with a \3/16\-inch stainless steel 
liner. To add to the robustness of this design, the seismicity of the 
SFP makeup water supply was designed to 0.12g and the reactor building 
reinforced-concrete internal structure, support for the reactor 
enclosure plenum, and equipment were designed to withstand a 0.05g 
acceleration; these reactor building structures were subsequently 
reevaluated by Consumers to 0.12g. Geologic investigations at the site 
and throughout the Lake Michigan basin, as described in the FHSR, have 
not found any indication of fault movement in the recent geologic past. 
Further, as described in the FHSR, the materials beneath and around the 
seismic Class I structures are not likely to liquefy with a ground 
acceleration of 0.12g, and settlement of structures and stability of 
slopes at the BRP site during ground acceleration are not a safety 
concern. 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 will actually be able to withstand a larger 
earthquake. Thus, the loss of coolant from the BRP SFP, which partially 
or completely uncovers the fuel, is a beyond-design-basis event with a 
very low probability of occurrence.
    Despite the robust design of the SFP, Consumers postulated a non-
mechanistic loss of all water from the SFP and determined that the 
resulting gamma radiation from the spent fuel would not result in 
offsite exposures exceeding EPA PAGs, as documented in the licensee's 
November 20, 1997, letter to the staff. For this scenario, Consumers 
calculated an offsite dose of 1.10 mrem TEDE at the closest site area 
boundary, which is significantly below EPA PAGs. The NRC staff reviewed 
the licensee's calculational methods and assumptions supporting 
Consumers' gamma shine analysis and found them to be acceptable.
    In a letter dated April 29, 1998, Consumers submitted an analysis 
for a complete loss of water inventory in the SFP. The analysis was 
based on the actual spent fuel decay heat generation rates, actual 
spent fuel and SFP configuration and engineering assumptions including 
a pin peaking factor and no credit for forced-ventilation cooling. 
Consumers determined that as of April 6, 1998 (220 days after permanent 
reactor shutdown), air cooling of spent fuel would be sufficient to 
maintain the spent fuel clad temperature below 565  deg.C. The staff 
reviewed the licensee's actual SFP conditions and concluded that they 
appropriately characterized its conditions. Further, the staff notes 
that additional margin is provided in the Consumers calculation due to 
the continuing reduction of decay heat in the spent fuel. In addition, 
the staff evaluated a bounding scenario where the active fuel is 
totally uncovered and water is blocking the assembly lower inlet so 
that no natural circulation flow path exists. The staff calculated it 
would take approximately 14 hours for the hottest location in the 
highest power fuel assembly to reach 900  deg.C. The heat up time was 
calculated assuming an adiabatic heat up of a fuel rod and using 
conservative decay heat assumptions. An adiabatic heat up 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 
condition in which the SFP water is lost and the fuel is surrounded by 
a perfect heat transfer insulator. The staff considers this scenario to 
be bounding for any loss of inventory scenario since any other scenario 
would have some heat removal from the assembly thereby resulting in a 
longer heat up time. 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 measures after a postulated loss of 
water and before a postulated release of radioactive material occurs 
from spent fuel overheating.
    In the event that SFP water is lost gradually, plant personnel have 
various methods of detecting SFP water loss and restoring SFP water 
level. As described in the FHSR and licensee procedures, detection 
includes remote reading level instrumentation, surge tank sight tank, 
and local level observation. The SFP level instrumentation can be 
powered by a diesel generator in the event of a loss of offsite power. 
The staff also notes that gross SFP level can also be interpreted from 
installed temperature and radiation detection instrumentation. SFP 
water level restoration can be accomplished by treated radioactive 
waste or demineralizer water through the SFP cooling system and by the 
installed makeup line. The emergency water sources are fire water and 
water from Lake Michigan via a portable and fully tested skid-mounted 
pump; the staff considers the skid-mounted pump as a last-resort makeup 
water source providing defense-in-depth. Each source of water can 
supply at least 30 gallons per minute, which is the flow rate 
determined by the licensee to maintain the bulk pool water less than 
the design temperature of 150  deg.F (66  deg.C) and maintain adequate 
SFP water inventory taking into consideration evaporation at 150  deg.F 
(66  deg.C). As described in the FHSR, the installed makeup water 
supply and fire water systems are designed to seismic Class 1 
requirements.
    The SFP has been and continues to be leaktight with no measurable 
loss of water detected by the leak-detection system. There is no SFP 
drain and a concrete weir and siphon protection features prevent any 
piping failure from draining or siphoning the SFP water level below 20 
feet above the top of the spent fuel assemblies. On the basis of the 
installed instrumentation, operator tours of the SFP, the engineered 
features associated with the SFP SSCs, and the availability of the 
makeup water sources to restore a gradual loss of SFP water, the staff 
finds it highly unlikely to expect that the fuel will uncover as a 
result of a gradual loss of coolant scenario. In addition, Consumers 
evaluated the loss of spent fuel cooling and concluded that it does not 
represent a safety concern, in part, because spent fuel decay heat rate 
has markedly decreased since the final reactor shutdown. On August 30, 
1997, when the plant conducted its final shutdown following months of 
reactor operation, the spent fuel decay heat (assuming a fully off-
loaded reactor core) was

[[Page 53943]]

approximately 3.7E6 Btu/hr. On December 5, 1997, with a decay heat rate 
of 0.7E6 Btu/hr and no SFP cooling, the licensee determined that it 
would take 72 hours for the SFP to heat up to 150  deg.F (66  deg.C) 
from an initial temperature of 80  deg.F (27  deg.C) . Since this 
determination, the decay heat rate has decreased by a factor of two to 
approximately 0.3E6 Btu/hr. Further, the evaporation rate of SFP water 
at 150  deg.F (66  deg.C) is approximately 11 gpm, well within the 30 
gpm capacity of the SFP makeup water supplies.
    The staff concludes that the licensee's request for an exemption 
from certain requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR 
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 
mitigative actions and, if necessary, offsite measures for the public 
could be taken without preplanning. Therefore, the staff concludes that 
the requirement in 10 CFR 50.54(q) that emergency plans meet all the 
requirements of 10 CFR 50.47(b) and all the requirements of Appendix E 
to 10 CFR Part 50 is not now warranted at BRP, and an exemption from 
some of the onsite and offsite emergency planning standards and 
requirements is acceptable.

IV

    The NRC staff has completed its review of the licensee's request 
for an exemption from the requirements of 10 CFR 50.54(q) that 
emergency plans must meet all of the standards of 10 CFR 50.47(b) and 
from the requirements of Appendix E to 10 CFR Part 50. This exemption 
includes partial exemption from the standards of 10 CFR 50.47(b)(3) 
through (7), and (9) and the requirements of 10 CFR Part 50, Appendix 
E, IV, ``Content of Emergency Plans;'' A.4; B; C; D.1 and 3; E.9.a and 
d; and F.1, 2, and 2.e. Further, this exemption covers all of the 
standards of 10 CFR 50.47(b)(10) and the requirements of 10 CFR Part 
50, Appendix E, IV, A.3, 5, and 8; D.2; E.8 and 9.c; and F.2.c, d, and 
f. 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 mitigative actions and, if 
necessary, offsite protective measures for the public could be taken 
without preplanning. The analyses submitted by the licensee are 
consistent with the statements made in its FHSR and proposed DEP, which 
state that any decommissioning activity will be bounded by the analyses 
presented therein and the considerations and assessments in the NRC's 
``Final Generic Environmental Impact Statement on Decommissioning of 
Nuclear Facilities'' (NUREG-0586). Consumers will continue to maintain 
and implement an onsite emergency preparedness organization capable of 
responding to and mitigating the consequences of radiological events 
still possible at the site and will continue to coordinate, as 
necessary, with offsite organizations to ensure effective emergency 
response to onsite situations, if needed. 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 
necessary to determine the potential impact of a radiological emergency 
on the general public. Thus, the underlying purpose of the regulations 
will not be adversely affected by eliminating offsite emergency 
planning activities and 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 undue risk to public health and safety, and 
is consistent with the common defense and security. Further, special 
circumstances are present as stated in 10 CFR 50.12(a)(2)(ii). Pursuant 
to 10 CFR 51.32, the Commission has determined that the granting of 
this exemption will not have a significant effect on the quality of the 
human environment (63 FR 50930).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland this 30th day of September 1998.

    For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 98-26852 Filed 10-6-98; 8:45 am]
BILLING CODE 7590-01-P