ML20154E037
| ML20154E037 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/30/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20154E014 | List: |
| References | |
| NUDOCS 9810080022 | |
| Download: ML20154E037 (14) | |
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@un p *, UNITED STATES g j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066-0001
'+9 . . . . . ,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF A REQUEST FOR EXEMPTION FROM REQUIREMENTS OF 10 CFR 50.54(a) FOR EMERGENCY PLANNING CONSUMERS ENERGY COMPANY BIG ROCK POINT NUCLEAR PLANT DOCKET NO. 50-155
1.0 INTRODUCTION
On August 30, .997, the Consumers Energy Company (the licensee) shut down the Big i Rock Point (BRP) Nuclear Plant and on September 20,1997, completed the removal of fuel from the ceactor vessel By letter dated September 19,1997, as supplemented by letters !
dated October 29, and November 20,1997, and March 2, April 29, July 30, and j August 28,1998, Consumers Energy Company (the licensee) submitted a request for exemption from certain sections of 10 CFR 50.47 and Appendix E to 10 CFR Part 50. l
2.0 BACKGROUND
The NRC may grant exemptions from the requirenents of its regulations which, pursuant to 10 CFR 50.12(a), (1) are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security and (2)
, present special circumstances. Section 50.12(a)(2) of 10 CFR Part 50 identifies some of these special circumstances to be where 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, compliance would result in undue hardship or other costs that are significantly in excess of those incurred by others similarly situated, and circumstances exist that were not considered when the regulation was adopted.
In the licensee's submittals the following special circumstances were presented. BRP shut !
down on August 30,1997, and is now in a permanently shutdown and defueled condition.
l With the plant in a permanently shutdown and defueled condition, the applicable design-basis accidents are limited to a fuel handling incident and spent fuel cask drop. The licensee has also evaluated other radiological events such as a loss of spent fuel pool (SFP) water level and cooling, fire involving radioactive resin, and the decommissioning events p
Enclosure 2 9810080022 980930 PDR ADOCK 05000155 F PDR quCOtOD%
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summarized in the NRC's " Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities" (NUREG-0586). 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 April 6,1998, a beyond-design-basis event involving fuel damage (caused by a loss of SFP water and a subsequert over heating of the stored fuel) and the release of radioactive materials sufficient to exceed EPA PAGs at the site boundary is not credible. Once a radiological release warranting prompt offsite response is no longer possible, offsite response capability, including offsite emergency plans, would no longer be necessary. The licensee also stated, during a public meeting held at NPC Headquarters on August 13,1998, that
- requiring BRP to comply with the requiremems for offsite emergency planning when it is no
} longer warranted would result in undue financial hardship to BRP, its owners, and their ratepayers.
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3.0 DISCUSSION AND STAFF EVALUATION 3.1 Soent Fuel Storaoe 4
3.1.1 Seismicity 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 1 shield the spent fuel. Specifically, in the Final Hazards Summary Report (FHSR), the i licensee states 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 Pown 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 a cceleration; these reactor building structures were subsequently reevaluated by Consumers to 0.12g. Also, a fully tested skid-mounted pump can be used by the licensee as a last-resort makeup source providing deferse-in-depth capability. 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 accelerativa 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 uarthquake have significant margin in them, it is expected that an SSC built to withstand the hypothetical design-basis earthquake will actually be able ta withstand a larger earthquake. Thus, the loss of coolant from the BRP SFP, which
3 partially or completely uncovers the fuel, is a beyond-design-basis event with a very low probability of occurrence.
1 3.1.2 SFP Water Level and Makeuo Water l i
Consumers has 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 l instrumentation, surge tank sight tank, and local level observation. The SFP level l 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 I temperature and radiation detection instrumentation. SFP water level restoration can be j accomplished by treated radioactive waste or demineralize water through the SFP cooling i system and by the installed makeup line. The emergency water sources ara fire water and !
water from Lake Michigan via a portable and fully tested skid-mounted pump. 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 *F <
(66 *C) and maintain adequate SFP inventory taking into consideration SFP water )
evaporation at 150 F (66 C). l The SFP has been and continues to be leaktight with no measurable loss of water detected l by the leak-detection system. There is no SFP drain and a concrete weir prevents any l piping failure from draining the SFP water level below 20 feet above the top of the spent I fuel assemblies. Tne licensee has previously installed an anti-siphon device in the SFP cooling system inlet piping to the SFP to preclude SFP draining if this inlet piping breaks. )
1 The SFP cooling system suction piping is connected to the SFP surge tank, fuel pool filter I tank, and other pipes. Therefore, a pipe failure in this portion of the system will allow SFP water level to decrease to the weir, thereby maintaining approximately 20 feet of water over the fuel. The SFP makeup water piping discharges above the highest capable water level and cannot creatc a siphon.
3.2 Radioloaical Conseauences I With the plant permanently shutdown and defueled, the design-basis accidents and transients postulated to occur during reactor operation are no longer possible. Specifically, the potential for a forced release of a large radiological source term to the environment, due to the high pressure and temperature associated with reactor operation, is no longer possible. Additiona!!y, due to the radioactive decay of short-lived isotopes, there is a continuing reduction in the potential radiological source term since the BRP shutdown on August 30,1997.
Revision 7 to the BRP FHSR includes revised analyses of postulated accidents at BRP in its permanently shutdown and defueled status. Chapters 9 and 15 of the FHSR describe the l radiological consequences of accidents that could release radioactive materials to the environment. In particular, these chapters, in part, describe a loss of SFP cooling and level, cask drop, and fuel assembly drop. The staff reviewed these FHSR analyses and the accidents and events evaluated in licensee letters dated February 27,1995, November 20,
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' l 1997, and April 29,1998, to determine whether the radiologicalimpact of these events would require an offsite emergency plan.
I 3.2.1 Heavv Load Droos onto Soent Fuel ;
The licensee considered a spectrum of potential accidents involving mechanical damage to spent fuelincluding:
e Single bundle damage during fuel handling e Damage to 22% of the core due to a drop of a heavy load into the pool e Damage to all bundles from the last core offload e Damage to all bundles in the pool due to a heavy load drop event The licensee's analysis demonstrated tha 1 ie offsite dose at the site bour.dary dropped
- below EPA PAG vtlues of 1 rem total effective dose equivalent (TEDE) and 5 rem to the thyroid 63 days following final reactor shutdown on August 30,1997. This analysis did not give credit for the additional time and associated radioactivo decay of radioisotopes from November 1997 (63 days after the plant permanently ceased operation) to the present, containment isolation, and used an atmospheric dispersion coefficient for a ground level release in accordance with Regulatory Guide 1.25, " Assumptions Used for Evaluating th<, Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," (Safety Guide 25).
3.2.2 Loss of SFP Coolina I
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 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the SFP to heat up to 150 'F (66 'C) from an initial temperature of 80 *F (27 *C) . Since this determination, the decay heat rate has decreased by a factor of two to approximately 0.3E6 Btu /hr. Further, the evaporation
- ate of SFP water at 150 *F is approximately 11 gpm, well within the 30 gpm capacity of the SFP makeup water supplies.
3.2.3 Lonaf SFP Water Level The licensee analyzed the potential for a loss of SFP water due to drainage from the pool.
The SFP is a concrete structure which is lined with a 3/16-inch stainless steel plate. The SFP utilizes an anti-siphoning makeup line and a weir discharge system to maintain approximately 20 feet of water over the fuel. Further, Consumers stated that the potential for accidental draining of the SFP that could cause the zirconium cladding to react with steam was evaluated as part of the Spent Fuel Pool Expansion Hearings in the early 1980s.
The licensee stated that this evaluation showed that accidental draining of the fuel pool
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was unlikely; however, to provide added assurance, Consumers installed an additional makeup source for the fuel pool. The licensee also stated that as part of the hearings, the NRC evaluated the licensee's andlycis and concluded that the reliability of the makeup system had been established based on the single faihim criterion of Appendix A to 10 CFR Part 50 and sound engineering practice.
Consumers reevaluated the accidental draining of the SFP scenario described above and stated its conclusions in the September 19 and November 20,1997, letters to the staff.
The licensee initially presented a position that the zirconium cladding fire was not a credible accident based on, in part, a qualitative argument that !ncluded reference to the hearings conducted in the 1980s. In particular, Consumers stated that accidental draining of the SFP and zirconium cladding fire were already reviewed by the Commission during the SFP Expansion Hearings in the early 1980s as not being credible. Further, Consumers described some differences between the BRP plant and the representative plant in NUREG/CR-6451, "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants."
Subsequently, 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.
The staff reviewed the licensee's actual SFP conditions and concluded that they appropriately characterized its conditions. Consumers determined that as of April 6,1998, air cooling of the spent fuel would be sufficient to maintain spent fuel cladding temperature below 565 *C, the temperature Consumers used as being below which zirconium fuel cladding would not be subject to clad failure. Further, the staff notes that additional margin is provided in the Consumers' calculation due to the continuing reduction in decay heat in the spent fuel.
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 staft 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 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 fuelis 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 tne 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 from spelt fuel overheating.
6 In addition to the evaluation of incipient fuel damage resulting from the heat up of spent fuel due to the loss of SFP water level, the licensee evaluated the radiological consequences from gamma rays emanating from the uncovered fuct rods. The licensee calculated the offsite radiologicalimpact of this postulated scenario with the computer code Microskyshine. The fuel was modeled as a thin planar source to minimize the self-shielding effects of the fuel. An assumed complete draindown of the SFP results in a dose rate of 0.046 mrem per hour at the c'losest site boundary. In the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assumed to be needed to reestablish fuel pool water level for this event, an individual at the site area boundary would receive 1.1 mrem, well within the EPA's early-phase PAG of 1 rem TEDE.
Microskyshine is a commercially available computer code, widely used in the radiation shielding industry, and is acceptable to the staff.
3.2.4 Nonfuel-Related Decommissionina Accidents The licensee evaluated potential nonfuel-related decommissioning accidents including explosions and fires, loss of contamination control, waste transportation accidents, external events, and natural phenomena. The licensee's analysis demonstrated that all postulated decommissioning accidents for the BRP are bounded by the results described in the NRC's " Final Generic EnvironmentalImpact Statement on Decommissioning of Nuclear Facilities" (NUREG-0586) which found that radiation doses to the maximally exposed individual from an accidental airborne release of radioactive material during decorr.missioning were low.
Decontamination of systems during decommissioning and dismantlement operations is expected to generate radioactive waste in the form of contaminated demineralizer resins.
The worse-case event that would provide a motive force for the release and transport of airborne radioactive material offsite is a fire in a fully loaded resin liner Resins are collected and de-watered in liners onsite prior to transporting offsite for disposal. Using a release fraction of 1% and current design-basis meteorology, the licensee's calculations showed that a fire in a resin liner loadeo with a maximum of 147 curies would result in a maximum offsite dose of 96 mrem TEDE and 175 mrem thyroid committed dose equivalent (CDE). Further, the licensee hypothetically assumed that all radioactive material from chemical decontamination was loaded into one HIC, as a worst-caso scenario. This hypothetical situation using approximately 300 curies would result in an offsite dose of 195 mrem TEDE and 357 mrem thyroid CDE. The release fractions used in these analyses are consistent with the release fractions listed in Schedule C of 10 CFR 30.72, for mixed fission and corrosion products. The calculational methods and assumptions used in this analysis are acceptable to the staff.
The NRC staff also assessed the current low-level radioactive waste (LLRW) situation at BRP. An NRC inspector determined that as of July 28,1998, five 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 rac'ioactive 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 i
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7 a fire protection program for its onsite facilities and continually assesses combustible loading to minimize fire potential and consequences.
' The processing and packaging of the chemical decontamination resins are no different now (during decommissioning) than they were during plant operation. Further, the storage and transportation of resins, the procedures for these processes, and the engineering controls associated with the SSCs used to process resin are effectively the same now as they were during reactor operation. Regarding the training and qualification of personnel conducting radioactive resin handling, Consumers has augmented its radiation protection staff with contractors; however, the licensee still maintains responsibility for titis site activity and implements controls to ensure that personnel are appropriately trained. The processes and personnel utilized by the licensee to handle resin and other LLRW are functionally equivalent now to when BRP was operating.
The staff has determined from review of the licensee's analysis that the postulated dose to the general public from any reasonably conceivable accident would not exceed the EPA PAG levels and, for the bounding accideret, the length of time available provides confidence that mitigative actions and, if necessary, offsite protective measures for the public could be taken, without preplanning. Therefore, the staff concludes that it is acceptable to exempt the licensee from those emergency preparedness requirements for responding to events involving the release of radioactive material that would result in offsite doses in excess of the EPA PAGs.
3.3 Evaluation of Emeraencv Plan Exemotions Under 10 CFR 50.54(q), a licensee authorized to possess and operate a nuclear power reactor shall follow and maintain in effect both onsite and offsite emergency plans that meet the standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR Part 50. The licensee .equested exemption from some of the standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR Part 50 that relate to response to events involving the release of radioactive material that would result in offsite dose in excess of the EPA PAGs. The staff's evaluation of each of the requested exemptions is described below.
The licensee requested exemption from that part of 10 CFR 50.47(b)(3) regarding the arrangement to accommodate State and local staff at a near-site emergency operation facility. The emergency plan would continue to maintain arrangements for requesting and using assistance resources from other organizations.
The staff found this change acceptable.
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The licensee requested exeniption from that part of 10 CFR 50.47(b)('4) regarding the need for State and local plans to rely on information provided by the licensee for offsite response measures.
The staff found this change acceptable.
The licensee requested exemption from that part of 10 CFR 50.47(b)(5) regarding the need for establishing procedures and means for notification of the public within the emergency planning zone (EPZ).
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. The staff found this change acceptable.
l The licensee requested exemption from that part of 10 CFR 50.47(b)(6) regarding the need ,
for the provision of prompt communications with the public. i The staff found this change acceptable.
The licensee requested exemption from that part of 10 CFR 50.47(b)(7) regarding the need for providing information to the public on a periodic basis concerning initial actions in an emergency and a physicallocation for the dissemination of information to the news media, l
The staff found these changes acceptable.
10 CFR 50.47(b)(R1 The licensee requested exemption from that part of 10 CFR 30.47(b)(0) regarding the requirement to have the capability to assess and monitor specifically "offsite" consequences of radiological emergency conditions.
L The licensee's exemption request is based on the conclusion that no design-basis accident or credible beyond-design-basis accident can result in radioactive material releases that exceed EPA FAGS at the site boundary The bounding accident assessed by the staff was a low likelihood event (loss of all SFP water cooling and adiabatic heat up of the spent fuel, Section 3.3.3 of this safety evaluation) that resulted in postulated doses in excess of EPA
> PAGs. However, the length of time available provides confidence that mitigative actions
. and,'if necessary, offsite measures for the public could be taken without preplanning. In Section 10.2, " Assessment Methods," of the licensee's proposed Defueled Emergency Plan (DEP), the licensee stcted the following:
9 Radiological release assessments are petformed using measured radiological and meteorological data. Dose assessment graphs are initially used by the Site Emergency Director or other trained emergency response personnel to determine the industrial area dose rate.
Additional assessments of potential radiation dose to plant personnel from direct radiation or potential exposure from various other sources will be performed as appropriate.
The licensee has committed to revise this section of its DEP to specify that it will have the capability to determine the potentialimpact of a radiological emergency to the general public. This information would be used to determine whether offsite measures for the ge.3erni public would be appropriate. The staff found this chhnge acceptable on the basis of the licensee's commitment to maintain the capability to determine the potentialimpact of a radiological emergency on the general public.
The licensee requested exemption from the complete requirement of 10 CFR 50.47(b)(10) to, in part, develop protective actions for the plume exposure and ingestion pathway EPZs.
Tne staff found this change acceptable.
10 CFR Part 50. Accendix E. IV The licensee requested exemption from that part of 10 CFR Part 50, Appendix E, IV.
regarding the requirement to provide an analysis of the time required to evacuate and take other protective actions offsite.
The staff found this change acceptable.
10 CFR Part 50. Anoendix E. IV. A.3.
The licensee requested exemption from the complete requirement of 10 CFR Part 50, i Appendix E, IV, A.3. to, in part, describe licensee headquarters personnel who will be sent ;
to the plant site in an emergency.
The staff found this change acceptable.
10 CFR Part 50. Accendix E. IV. A.4.
The licensee requested exemption from that part of 10 CFR Part 50, Appendix E, IV, A.4.
regarding the elimination of the term "offsite" in relation to dose projections that will be performed.
The licensee's exemption request is based on the conclusion that no design-basis accident
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10 or crcdible beyond-design-basis accident can result in radioactive material releases that exceed EPA PAGs at the site boundary. The bounding accident assessed by the staff was a low likelihood event (loss of all SFP water cooling and adiabatic heat up of the spent fuel, Section 3.3.3 of this safety evaluation) that resulted in postulated doses in excess of EPA PAGs. However, the length of time available provides confidence that offsite measures for the public could be taken without preplanning. In Section 10.2, " Assessment Methods," of the licensee's proposed DEP, the licensee stated the following:
Radiological release assessments are performed using measured radiological and meteorological data. Dose assesament graphs are initially used by the Site Emergency Director or other trained emergency response personnel to determine the industrial area dose rate.
Additional assessments of potential radiation dose to plant personnel from direct radiation or potential exposure from various other sources will be performed as appropriate.
The licensee has committed to revise this section of its DEP to specify that it will have the l capability to determine the potentialimpact of a radiological emergency to the general '
public. This information would be used to determine whether offsite measures for the l general public would be appropriate. The staff found this change acceptable on the basis )
of the licensee's commitment to maintain the capability to determine the potentialimpact I of a radiological emergency on the general public.
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10 CFR Part 50. Anoendix E. IV, AL I l
i The licensee requested exemption from the complete requirement of 10 CFR Part 50, i Appendix E, IV, A.S. to, in part, identify individuals with special qualifications.
The staff found this change acceptable.
10 CFR Part 50. Anoendix E. IV. A.8.
The licensee requested exemption from the complete requirement of 10 CFR Part 50, Appendix E, IV, A.8. to, in part, identify State and local officials responsible for protective l l
actions.
The staff found this change acceptable.
10 CFR Part 50. Accendix E. IV. B.
The licensee requested exemption from that p art of 10 CFR Part 50, Appendix E, IV, B.
regarding the requirement to base emergency action levels on offsite monitoring results and censideration of protective measures outside the site boundary.
The staff found this change acceptable.
4 11 10 CFR Part 50. Anoendix E. IV. C.
The licensee requested exemption from that part of 10 CFR Part 50, Appendix E, IV, C.
regarding the requiren.ent to base emergency action levels on offsite radiological monitoring information and information from other sensors, such as containment pressure, that are no longer applicable or appropriate. The licensee requested to eliminate emergency classification above the Alert level.
The staff found these changes acceptable.
10 CFR Part 50. Accendix E. IV. D.1.
The licensee requested exemption from that part of 10 CFR Part 50, Appendix E, IV, D.1.
regarding the requirement for administrative and physical means for the prompt notification of the public of protective measures.
The staff found this change acceptable.
10 CFR Part 50. Accendix E. IV. D.2.
l The licensee requested exemption from the complete requirement of 10 CFR Part 50, Appendix E, IV, D.2. regarding, in part, yearly dissemination of emergency planning information to the public.
The staff found this change acceptable.
10 CFR Part 50. Anoendix E. IV. D.3.
The licensee requested exemption from that part of 10 CFR Part 50, Appendix E, IV, D.3.
regarding the demonstration that State / local officials have the capability to make public notification promptly and changed the time for notification of State and local agencies to within 30 minutes after declaring an emergency.
The staff found these changes acceptable.
10 CFR Part 50. Aooendix E. IV. E.8.
The licensee requested exemption from the complete requirement of 10 CFR Part 50, Appendix E, IV, E.8. to, in part, provide for a near-site emergency operations facility.
The staff found this change acceptable.
I 12 10 CFR Part 50. Accendix E. IV. E.9.a.
The licensee requested exemption from that part of 10 CFR Part 50, Appendix E, IV, E.9.a.
regaro;ng the elimination of the term "within the plume exposure pathway EPZ" from the description of communication systems with State / local governments.
The staff found this change acceptable.
10 CFR Part 50. Accendix E. IV E.9.c.
The licensee requested exemption from the complete requirement of 10 CFR Part 50, Appendix E, IV, E.9.c. for, in part, provision of commun: cations among the near-site emergency operation facility, the nuclear facility, State and local emergency operations centers, and field assessment teams.
The staff found this change acceptable.
10 CFR Part 50. Accendix E. IV. E.9.d.
The licensee requested exemption from that part of 10 CFR Part 50, Appendix E, IV, E.9.d.
regarding the requirements for provision of communications by the licensee from the onsite technical support center and near-site emergency operation facility to the NRC Headquarters and Regional Operations Centers. The licensee will maintain provision for these communications from the control room / monitoring station and changed the frequency of test of such communications to quarterly.
The staff found these changes acceptable.
10 CFR Pa t 50. Aopendix E. IV F.1.
The licensee requested exemption from that part of 10 CFR Part 50, Appendix E, IV, F.1.
to include local news media persons in the list of categories of emergency personnel that are provided periodic training.
The staff found this change acceptable.
10 CFR Part 50. Aooendix E. IV. F.2.
The licensee requested exemption from that part of 10 CFR Part 50, Appendix E, IV, F.2.
regarding the requirement to test the public notification system as part of emergency preparedness exercises.
The staff found this change acceptable.
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1 13 10 CFR Part 50. Aooendix E. IV. F.2.c.
The lice:'see requested exemption from the complete requirement of 10 CFR Part 50, Appendix E, IV, F.2.c. to, in part, conduct a biennial exercise with full participation of offsite authorities.
The staff found this change acceptable.
10 CFR Part 50. Accendix E. IV. F.2.d.
The licensee requested exemption from the complete requirement of 10 CFR Part 50, Appendix E, IV, F.2.d. regarding, in part, an ingestion pathway exercise.
The staff found this change acceptable.
10 CFR Part 50. Accendix E. IV. F.2.e.
The licensee requested exemption from that part of 10 CFR Part 50, Appendix E, IV, F.2.e.
to modify the language of this section by deleting the phrase " located within the plume exposure pathway EPZ" and add the word " contiguous" to clarify the governmental entities who would be permitted to participate in emergency preparedness drills.
The staff found the.se changes acceptable.
10 CFR Part 50 Aooendix E. IV F.2.f. .
The licensee requested exemption from the complete requirement of 10 CFR Part 50, Appendix E, IV, F.2.f. to, in part, eliminate the requirement to perform remedial exercises.
The staff found this change acceptable.
4.0 CONCLUSION
S The staff concludes that the licensee's request for an exemption from the requirements of 10 CFR 50.54(q) to follow and maintain in effect emergency plans that meet all the standards in 10 CFR 50.47(b) and all the requirements in Appendix E to 10 CFR Part 50 is acceptable in view of the 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 the EPA PAG dose levels and, for the bounding accident, the length of time available provides confidence that mitigative actions and, if necessary, offsite measures for the public could be taken without preplanning. 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
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14 to continue to maintain capabilities for dose assessment and personnel necessary to I determine the potentialimpact of a radiological emergency on the general public. !
Principal Contributors: Jim O'Brien Roger Pedersen ;
Diane Jackson Paul Harris
' Date: September 30, 1998 I: -
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