ML20154E024
| ML20154E024 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/30/1998 |
| From: | Collins S NRC (Affiliation Not Assigned) |
| To: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| Shared Package | |
| ML20154E014 | List: |
| References | |
| NUDOCS 9810080019 | |
| Download: ML20154E024 (12) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF )
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CONSUMERS ENERGY COMPANY ) Docket No. 50-155
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BIG ROCK POINT NUCLEAR PLANT )
EXEMPTION 1.
Consumers Energs vny (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 R sgulatory Commission (the Commission '
or NRC) now or hereaf ter 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 4
or place or retain fuelin the reactor vessel.
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Section 50.54(q) of Title 10 of the Code of FederalRegulations (10 CFR 50.54(q))
requires power reactor licensees to follow and maintain in effect emergency plans that 981oosoo19 980930 PDR ADOCK 05000155 F PDR ,
o!l10020014
2 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 q 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 I 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, respectively, of the Emergency !
I Planning Zones that are to be considered in emergency plans, and Appendix E to 10 CFR i Pa t 50 identifies the information that must be included in emergency plans.
Ill.
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 l
submitted and requested approval of its proposed BRP Defueled Emergency Plan (DEP), j 1
which was written on the basis of NRC staff approval of the proposed exemption request.
The exemption would a' low Consumers to discontinue certain aspects of offsite emergency planning and reduce the scope of onsite emergency planning.
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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, Juli 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 i-offsite requirements of 10 CFR 50.47 and Apper. dix E to 10 CFR Part 50 are addressed in 1
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. l l
The licensee stated that special circumstances exist at BRP because the plant is l l
permanently shutdown and defueled and the radiological source term at the site is reduced i
from that associated with reactor power operation. With the reactor power plant permanently shutdown and defueled, the design-basis accidents and transients postulated i
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4 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 l
temperature associated with reactor operation no longer exists. Additionally, due to the i
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 l planning when it is no longer warranted would result in undue financial hardship to BRP, its owners, and their ratepayers. l 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
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an emergency situation and that these events are bounded by the consideration.s described j l
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 opent fuel pin gap that contribute to the dose consequences of releases from operating reactors have decayed to ne:gligible 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 i
! consequences of a release from a low-level radioactive waste (LLRW) are bounded by l 1
- accidents involving spent fuel. l
- By letter dated November 20,1997, Consumers subrnitted its evaluation demonstrating the conclusion that 6 fire involving radioactive resin being stored at the !
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facility and gamma radiation resulting from a complete draindown of the SFP would not i exceed the EPA PAGs at the site area boundary. The resin fire is considered a bounding l LLRW accident at the site. This fire would involve the ion exchange resin used to process j 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 I
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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 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 love 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 l Spectra for Seismic Design of Nuclear Power Plants," value of 0.12g zero-period horizontal accelera@n. Thr 3FP structure has a floor and walls of reinforced concrete that vary in e s , - -
7 thickness from 3 few 6 inches to 6 feet 9 inches with a 3/16-inch stainless steelliner. To add to the robustness of this design, the seismicity of the SFP makeup water supply was designed to.0.129 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 subsequent ly reevaluated by
, Consumers to 0.12g. Geologic investigations at the site and throughout the Lake Michigan 2
basin, as described in the FHSR, have not found any indication of fault movement in the l l
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-besis event with a very low probability of occurrence.
Despite the robust design of the SFP, Consumers postulated a non-mechan'ctic loss of all water from the SFP and determined that the resulting gamma radiation from the srent 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 i
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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 soent fuel would be sufficient to maintain the spent fuel clad temperature below 565 *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 t provided in the Consumers calculation due to the continuing redection of decay neat in the spent fuel. In addition, the staff evaluated a bounding scenario where the active fuelis 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 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> for the hottest location in the highest power fuel assembly to reach 900 *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 ;
4 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 hea: transfer insulator. The staff considers this scenario to be bounding for any loss of inventory scenario since any other scenario would have some iieat removal from the assembly thereby resulting in a longer heat up time. The staff determined that in view of i
the low likelihood of the bounding scenario and the time elapsed since the shutdown of the
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9 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 levelinstrumentation, 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 l level can also be interpreted from installed temperature and radiation detection instrumentation. SFP weter level testoration 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 1
J portable and fully tested skid-mounted pump; tho staff considers the skid-mounted pump
, as a last-resort makeup water source providing defense-in-depth. Each source of water i 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 *F
. (66 *C) and maintain adequate SFP water inventory taking into consideration evaporation at 150 'F (66 *C). As described in the FHSR, the installed makeup water supply and fire water systems are designed to seismic Class 1 requirements.
I 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 42 water level below 20 feet above the top of the spent fuel assemblies. On the basis of the i
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10 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 gradualloss 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 finel shutdown following months of reactor operation, the spent fuel decay heat (assuming a fully off-loaded reactor core) was approximately 3.7E6 Btu /hr. On December 5,1997, with a decay heat rate of I l
0.7E6 Btu /hr and no SFP cooling, the licensee determined that it would tc,ke 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for
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the SFP to heat up to 150 F (66 *C) from an initial temperature of 80 *F (27 *C) . Since i
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 *F (66 *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 curront plant l l
status. The staft finds that the postulated dose to the general public from any reasonably l 1
conceivable accident would not exceed EPA PAGs and, for the bounding accident, the .
1 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
I1 warranted at BRP, and an exemption from some of the onsite and offsita 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 I Emergency Plans;" A.4; B; C; D.1 and 3; E.9.a and d; and F.1, 2, and 2.e. Further, this l
exemption covers all of the standards of 10 CFR 50.47(b)(10) and the requirements of l 1
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 j the basis of its review, the NRC staff finds that the postulated dose to the general public l
from any reasonably conceivable accident would not exceed EPA PAGs and, for the bounding accident, the .ongth of time available provides confidence that mitigative actions and, if necessary, offsito 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 l
bounded by the analyses presented therein and the considerations and assessments in the NRC's " Final Generic EnvironmentalImpact 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
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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 potentialimpact of a radiological emergency on the general public. Thus, the underlying purpose of the regulations will not j 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 l risk to public health and safety, and is consistent with the common defense and security.
4 Further, special circumstances are present as stated in 1G CFR 50.12(a)(2)(ii). Pursuant to l l
10 CFR 51.32, the Commission has determined that the granting of this exemption will riot have a significant effect on the quality of the human environment (63 FR 50930).
i This exemption is effective upon issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
ta ue . o ins, Director
' Office of Nuclear Reactor Regulation Dated at Rockville, Maryland
- this 30th day of September 1998 d