ML20203G517
| ML20203G517 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/14/1986 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20203G523 | List: |
| References | |
| TAC-57619, NUDOCS 8608010244 | |
| Download: ML20203G517 (22) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION MISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION DOCKET NO. 50-416 NOTICE OF CONSIDERATION OF ISSUANCE OF A'4ENDMENT TO FACILITY OPERATING LICENSE AND PROPOSED N0 SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION AND OPPORTUNITY FOR HEARING The U. S. Nuclear Regulatory Commission (the Commission) is considering issuance of an amendnent to Facility Operating License No. NPF-29, issued to Mississippi Power & Light Company, Middle South Energy, Inc., and South Mississippi Electric Power Association (the licensees), for operation of the Grand Gulf Nuclear Station (GGNS), Unit 1, located in Claiborne County, Mississippi.
The GGNS Unit 1 is a boiling water reactor with a Mark III containment.
The spent fuel pool is located in the auxiliary building, similar to spent fuel pool arrangements for pressurized water reactors. Above the GGNS reactor, and within the containment, there is an upper containment pool with racks for holding new fuel to be placed in the reactor and spent fuel removed from the reactor during refueling; however, before reactor startup after refueling, all spent fuel is transferred to the spent fuel pool for storage.
The amendment would revise Section 5.6 " Fuel Storage" of the Technical Specifications to allow increased upper containment pool capacity and increased spent fuel storage capacity. This increased capacity would be obtained by replacing the fuel racks in the upper containment pool and in the spent fuel
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storage pool with high density fuel racks. The center-to-center distance geo!%su*e8%A6 P
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I between fuel assemblies would be changed from 12 inches to 6.26 inches. This reracking would increase the upper containment pool capacity from 170 to 800 fuel assemblies in order to hold a complete core unloading, if necessary, and increase the spent fuel pool storage capacity from 1270 to 4348 fuel assentlies.
However, the number of fuel assemblies to be stored in the spent fuel pool would be limited by Technical Specifications to 2324. The Technical Specifi-cations would be changed to limit the spent fuel pool water temperature to 140*F rather than 150 F and require plant shutdown rather than a special report if the limiting temperature is exceeded. These changes were requested in the' licensee's application for amendment dated May 6, 1985 as amended by letters dated July 29, August 15, August 30, September 11, September 12, November 1, and December 18, 1985 and March 14, March 15, June 5 and June 9, 1986.
This public notice supersedes a previous notice published in the Federal Register on September 13, 1985 (50 FR 37451). The previous notice was based on the licensee's initial application for amendment dated May 6, 1985. The initial application and associated notice considered storage of 4348 fuel assemblies in the spent fuel pool, the physical limit of the high density spent fuel racks. However, during its safety review of the application, the staff noted that the licensee's analysis of spent fuel pool cooling was based on the use of both spent fuel pool cooling loops or one spent fuel pool cooling loop and one residual heat removal loop to maintain pool water temperature less than 140 F.
The staff requested that the provisions for cooling of t,he spent fuel pool meet criteria in the Standard Review Plan (NUREG-0800) by considering a single failure in the spent fuel pool cooling system and use of the residual heat removal system only when the plant is
. in cold s'hutdown.
In response, the licensee revised its application by proposing to change the Technical Specifications to limit the number of stored spent fuel assemblies to 2324 and to limit the spent fuel water temperature to 140 F instead of 150 F so that the Standard Review Plan criteria would be met. This notice is based on the revised application, from that initially noticed which results in additional safety margin for cooling of the spent fuel in the spent fuel pool and the upper containment pool. Appropriate changes to the initial notice have been incorporated in this notice regarding the cooling of spent fuel.
Before issuance of the proposed license amendment, the Commission will have made findings required by the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations.
The Commission has made a proposed determination that the amendment request involves no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the prob-ability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. Our evaluation of the reracking of the spent fuel pool and of the upper containment pool are considered separately.
A.
SPENT FUEL P0OL The technical evaluation of whether or not an increased spent fuel pool storage capacity involves significant hazards considerations is centered on three standards:
(1) does increasing the spent fuel pool storage capacity
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significantly increase the probability or consequences of accidents previously evaluated? Reracking to allow closer spacing of fuel assemblies does not significantly increase the probability or consequences of accidents previously evaluated; (2) does increasing the spent fuel pool storage capacity create the possibility of a new or different kind of accident from any accident previously evaluated? With respect to Grand Gulf Nuclear Station (GGNS) Unit 1, the staff has not identified any new categories or types of accidents as a result of reracking to allow closer spacing for the fuel assemblies. The proposed reracking does not create the possibility of a new or different kind of accident previously evaluated for the spent fuel pool.
In all reracking reviews completed to date, all credible accidents postulated have been found to be conservatively bounded by the evaluations cited in the Safety Evaluation Report (SERs) supporting each anendment; and (3) does increasing the spent fuel pool storage capacity significantly reduce a margin of safety? The staff has not identified signifi-cant reductions in safety margins due to increasing the storage capacity of the spent fuel pool. The expansion results in an increased heat load, but this heat load is kept well within the design limitations of the installed cooling systems by Technical Specifications which would limit the number of fuel assemblies which can be stored. The full capacity of the high density racks may be utilized by increasing the heat removal capacity of the cooling system, e.g. by increasing heat exchanger or pumping capacity, but such utilization would require another license amendment. The time to boiling in the event of failure of spent fuel cooling systems would be increased by Technical Specifi-cations to 1imit the spent fuel pool water temperature to 140 F rather than the presently specified 150*F.
In all cases, the temperature of the pool will
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remain below design values. The small increase in the total amount of fission products in the pool is not a significant factor in accident considerations.
The increased storage capacity may result in an increase in the pool reactivity as measured by the neutron multiplication factor (Keff). However, after extensive study, the staff determined in 1976 that as long as the maximum neutron multi-plication factor was less than or equal to 0.95, then any change in the' pool reactivity would not significantly reduce a margin of safety regardless of the storage capacity of the pool. The licensee has indicated that the K would eff not exceed 0.95.
The techniques utilized to calculate K have been bench-eff marked against experimental data and are considered very reliable.
Reracking to allow a closer spacing between fuel assemblies can be done by proven tech-nologies.
In summary, replacing existing racks with a design which allows closer spacing between stored spent fuel assemblies is considered not likely to involve significant hazards consideration if two conditions are met.
First, no new technology or unproven technology may be utilized in either the construction process or in the analytical techniques necessary to justify the expansion.
Second, the K of the pool must be maintained less than or equal to 0.95.
eff Reracking to allow closer spacing satisfies these conditions.
The licensee's submittal included a discussion of the proposed action with respect to the issue of no significant hazards consideration. This discussion has been reviewed and the Commission finds it acceptable.
Pertinent portions of the licensee's discussion, addressing each of the three standards, is provided herein.
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n The licensee has stated that its analysis of the proposed reracking was accomplished using currently acceptable codes and standards and conforms to staff guidance of April 1978. The results of the licensee's analysis in relation to the three standards is as folows:
First Standard Involve a significant increase in the probability or consequences of an accident previously evaluated.
In the course of the analysis the licensee identified the following potential accident scenarios:
1.
A spent fuel assembly drop in the spent fuel pool.
2.
Loss of spent fuel pool cooling system flow.
3.
A seismic event.
4.
A spent fuel cask drop.
The probability of any of the four accidents is not affected by the racks themselves; thus reracking cannot increase the probability of these accidents.
The installation of the high density racks will be completed prior to the storage of any spent fuel in the present racks; thus, a construction accident involving spent fuel is not possible. Accordingly, the proposed rerack will not involve a significant increase in the probability of an accident previously evaluated.
The consequences of (1) a spent fuel assembly drop in the spent fuel pool are discussed in the licensee's submittal.
For this accident condition, the criticality acceptance criterion is not violated. The radiological consequences of a fuel assembly drop are not changed from the previous analysis. The results of the staff's evaluation were transmitted to the licensee in September 1981.
The licensee's analysis of the reracked design indicates a dropped fuel assembly
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would not penetrate through the base plate or distort the racks so they would not perform their safety function. Thus, the consequences of this type accident will not be significantly increased from previously evaluated spent fuel assembly drops, and have been found acceptable by the NRC.
The consequences of (2) loss of spent fuel pool cooling system flow have been evaluated for the existing spent fuel pool cooling system design as des-cribed in the GGNS FSAR and in the licensee's June 5, 1986, submittal. There are two spent fuel pool cooling system pump and heat exchanger trains.
One train will be operating and the other train will be maintained in an operable condition per Technical Specifications in the event of a failure in the cooling system. The number of stored spent fuel assemblies is limited so that the pool temperature can be maintained below 140 F with one spent fuel pool cooling system train and without use of the RHR for supplemental cooling after restart of the reactor following the refueling outage. The service water system that transports heat from the spent fuel pool cooling system to the ultimate heat sink is being upgraded in accordance with License Condition 2.C.(20).
In its June 5, 1986, submittal, the licensee stated that it will propose changes to the cooling cap-acity for the storage of a larger number of fuel assemblies up to the physical limit of the high density fuel racks. The Technical Specifications will limit the number of stored spent fuel assemblies until the cooling capacity is I
increased. The structural integrity of the spent fuel pool will be maintained and no new means of losing cooling water or flow have been identified. Thus, the consequences of this type accident will not be significantly increased from previously evaluated loss of cooling system flow accidents.
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The ' consequences of (3) a seismic event have been evaluated. The new racks will be designed and fabricated to satisfy the NRC staff accepted design criteria. The racks are designed to Seismic Category I criteria. The racks are neither anchored to the pool floor nor are they attached to the pool side walls.
The racks are structurally adequate to resist normal and accident load combina-tions. The racks are designed so that the floor loading from the racks filled with spent fuel assemblies does not exceed the structural capacity of the auxiliary building. Therefore, the integrity of the pool will be maintained and no new means of losing cooling water or flow have been identified. Thus, the consequences of a seismic event will not be significantly increased from pre-vicusly evaluated events.
The consequences of (4) a spent fuel cask drop accident are unchanged by the requested modification. The spent fuel cask handling crane rails do not extend over the spent fuel pool and the crane is designed to be single failure proof in accordance with the requirements of NUREG-0554 to preclude a drop on safety related equipment.
In addition, the crane meets the guidelines of NUREG-0612. Accordingly, the consequences of a cask drop accident are not signifi-cantly increased from previously evaluated events.
Therefore, it is concluded that the proposed amendment to rerack the spent fuel pool will not involve a significant increase in the probability or conse-quences of an accident previously evaluated.
Second Standard Create the possibility of new or different kind of accident from any acci-dent previously evaluated.
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The proposed reracking was evaluated by the licensee in accordance with the guidance of the NRC position paper entitled, "0T Position for Review and Accept-ance of Spent Fuel Storage and Handling Applications," NRC Regulatory Guides, NRC Standard Review Plans, and Industry Codes and Standards as listed in the licensee's submittal.
In addition, several previous NRC SERs for rerack applica-tions similar to this proposal have been reviewed. Neither the licensee nor the NRC staff could identify a credible mechanism for breaching the structural integrity of the spent fuel pool which could result in loss of cooling water such that cooling flow could not be maintained. As a result of this evaluation and these reviews, it is concluded that the proposed reracking does not, in any way, create the possibility of a new or different kind of accident from any accident previously evaluated for the GGNS spent fuel storage racks.
Third Standard Involve a significant reduction in a margin 'of safety.
The NRC staff safety evaluation review process has established that the issue of margin of safety, when applied to a reracking modification, will need to address the following areas:
1.
Nuclear criticality considerations.
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2.
Thermal-hydraulic considerations.
3.
Mechanical, material and structural considerations.
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The established acceptance criteria for criticality is that the neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions. This margin of safety has been adhered to in the criticality analysis methods for the new rack design as discussed in the licensee's submittal. The methods to be used in the criti-cality analysis conform with the applicable portions of the codes, standards, and specifications listed in the submittal.
In meeting the acceptance criteria for criticality in the spent fuel pool, such that K,ff is always less than 0.95, including uncertainties of 95/95 probability confidence level, the proposed amendnent to rerack the spent fuel pool will not involve a significant reduction in the margin of safety for nuclear criticality.
In its initial submittal dated May 6, 1985, the licensee stated in its analysis of the reracked pool that conservative methods were used to calculate the maxinum fuel tenperature and the increase in temperature of the water in the spent fuel pool. The calculated maximum fuel cladding temperature of 166 F is substantially less than the temperature at which local boiling would be initiated and sustained (243 F). The calculated maximum water temperature of 140*F for a normal refueling operation and 148*F for an abnormal unloading of the complete core arc slightly higher than temperatures calculated for the pre-sent fuel racks; however, the licensee stated that temperatures for the new racks meet the NRC staff's acceptance criteria of 140*F for normal refueling and 150*F for an abnormal unloading. During its review of the application, the NRC staff noted that the licensee's analysis of spent fuel pool cooling was based on the use of both spent fuel pool coolers to maintain pool water tempera-ture below 140*F.
In response to the staff's requests, the licensee made
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analyses 'of spent fuel pool cooling based on the criterion in the Standard Review Plan (NUREG-0800) that the 140 F pool temperature must be maintained assuming a single failure and based on use of the residual heat removal system for supplemental cooling only when the plant is in cold shutdown. These cal-culations resulted in a calculated pool temperature of about 140 F at the end of 13 refueling outages. The licensee proposed in its June 5,1986, submittal to limit the number of fuel assemblies stored in the spent fuel pool to 2324 fuel assemblies, which is the number calculated to be discharged from 10 re-fueling outages.
In addition, the licensee proposed to change the fuel pool temperature limit in the Technical Specifications from 150 F to 140 F and to require a plant shutdown, if the pool temperature cannot be maintained below 140 F.
These changes result in additional safety margin for cooling the spent fuel from that initially proposed. Because the licensee's spent fuel pool cooling calculations are based on the Standard Review Plan criterion and the limiting number of stored fuel assemblies is based on 10 refueling outages instead of 13, the staff concludes that there is no significant reduction in the margin of safety for thermal hydraulic or spent fuel pool cooling concerns.
The main safety function of the spent fuel pool and the racks is to main-tain the spent fuel assemblies in a safe configuration through all normal and abnormal loadings, such as an earthquake, impact due to a spent fuel cask drop, drop of a spent fuel assembly, or drop of any other heavy object. The mechanical, material, and structural considerations of the proposed rerack are described in the licensee's submittal. The proposed racks are to be designed in accordance with applicable portions of the "NRC Position for Review and Acceptance of Spent
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Fuel Storage and Handling Applications," dated April 14, 1978, as modified January 18, 1979; and Standard Review Plan 3.8.4 The rack materials used are compatible with the spent fuel pool and the spent fuel assemblies. The struc-tural considerations of the new racks address margins of safety against tilting and sliding, including impact on each other or the pool walls, damage of spent fuel assemblies, and criticality concerns. As previously stated, neither the licensee nor the NRC staff cotid identify a credible mechanism for breaching the structural integrity of the spent fuel pool which could result in loss of cooling water such that cooling flow could not be maintained. Thus, the margins of safety are not significantly reduced by the proposed rerack.
B.
UPPER CONTAINMENT P0OL The technical evaluation of whether or not the reracking of the upper con-tainment pool involves significant hazards considerations is also based on the three standards in 10 CFR 50.92.
First Standard:
Involve a significant increase in the probability or consequences of an accident previously evaluated.
For the upper containment pool, the licensee identified the following potential accidents:
1.
A spent fuel assembly drop in the pool 2.
Loss of pool cooling system flow 3.
A seismic event 4.
Drop of a heavy load.
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The probability of any of these accidents is not affected by the racks them-selves; thus reracking cannot increase the probability of these accidents. The installation of the high density racks will be completed prior to unloading any fuel from the reactor; thus a construction accident involving spent fuel is not possible. Accordingly, the preposed rerack will not involve a significant increase in the probability of an accident previously evaluated.
The considerations of the structural damage of (1) a spent fuel assembly drop in the upper containment pool are the same as the considerations for a drop in the spent fuel pool because the same design is used for the pool liner and the cells in the racks in both pools. The offsite radiological consequences of a fuel assembly drop inside primary containment are much less than for a drop of a fuel assembly in the spent fuel pool, which is inside secondary containment.
Staff's evaluation of radiological consequences provided in the SER, September 1981, is not changed by the reracking. Accordingly, the consequences of this type accident will not be significantly increased by the reracking.
The consequences of less of upper containment pool cooling system flow have been evaluated. The cooling of spent fuel in the upper containment pon1 is accomplished by the spent fuel pool cooling system, supplemented by one train of the residual heat removal RHR system. Both the spent fuel pool cooling system and the RHR system have redundant pumps and heat enchangers, so that the inoper-ability of one component in the systems would be compensated by use of a redun-dant component.
Reracking does not affect this capability. The structural integrity of the upper containment pool will be maintained and no new means of losing cooli,ng water or flow have been identified. Thus, consequences of (2) loss of cooling flow will not be significantly increased from previously evaluated accidents of this type.
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The ' consideration of the consequences of (3) a seismic event is the same as the consideration for the spent fuel pool because the rack design is the same.
The upper containment pool rack modules are lighter than modules in the spent fuel pool (121 cells versus 304 cells). Therefore, the shear stress in the upper containment pool liner for a seismic event is bounded by the analysis for the spent fuel pool liner. The pool floor loading from the racks filled with spent fuel assemblies does not exceed the structural capacity of the floor. The portion of the upper containment pool which is used to store spent fuel during refueling is designed similar to the spent fuel pool to preclude drainage below a safe shielding level to assure no accidental loss of pool water. Therefore, the integrity of the pool will be maintained and no new means of losing cooling water or flow have been identified. Thus, the consequences of a seisnic event will not be significantly increased from previously evaluated events.
The consequences of (4) drop of a heavy load on spent fuel in the upper containment pool were considered by the licensee. The containment polar crane and critical components of the fuel handling system are designed to seismic category I requirements so that they will not fail in a manner which results in unacceptable fuel damage or damage to safety-related equipblent. Heavy load handling equipment inside containment meets the guidelines of NUREG-0612 " Control of Heavy Load at Nuclear Power Plants." The licensee has analyzed the con-sequences of a dropped fuel transfer canal gate on the fuel racks inside con-tainment. The analysis showed that there would be no gross buckling of fuel cells in the racks and consequently the geometry of the active fuel in a fuel assembly would be preserved. Accordingly, the consequences of dropped heavy load are not significantly increased from previously evaluated accident analysis.
Second St'andard:
Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed reracking in the upper containment pool was evaluated by the licensee in accordance with the guidance of the NRC position paper "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," NRC Regulatory Guides, NRC Standard Revicw Plans, and Industry Codes and Standards as listed in licensee's submittal.
In addition, the staff has made a preliminary review of the proposal. Neither the licensee nor the staff could identify a credible mechanism for breaching the structural integrity of the upper contain-ment fuel pool which could result in a loss of cooling water such that cooling flow could not be maintained. Accordingly, it is concluded that the proposed reracking does not create the possibility of a new or different kind of accident from any accident previously evaluated for the upper containment fuel pool racks.
Third Standard:
Involve a significant reduction in a margin of safety.
The consideration of the margin of safety of the proposed reracking modifi-cation for the upper containment pool addressed the same three areas that were found necessary to be addressed in reracking of the spent fuel pool:
1.
Nuclear criticality considerations 2.
Thermal-hydraulic considerations 3.
Mechanical, material and structural considerations.
As in the spent fuel pool, the neutron multiplication factor will be less than or equal to 0.95 including all uncertainties under all conditions. Methods used l
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to calculate criticality are the same as those used for the spent fuel pool.
Accordingly, the proposed reracking of the upper containment fuel pool will not involve a significant reduction in the margin of safety for nuclear criticality.
The licensee's analysis of maximun fuel and maximum pool temperature con-sidered the interconnection of the spent fuel pool and the upper containment pool during refueling and the use of the spent fuel pool cooling system supple-mented by the RHR system. The results of that analysis, which are described above for the spent fuel pool rerack considerations, show that NRC staff's acceptance criteria would be met. Accordingly, there is no significant reduc-tion in the margin of safety for thermal hydraulic or upper containment fuel pool cooling concerns.
The mechanical, material and structural considerations of the proposed rerack in the upper containment pool are the same as those described above for the spent fuel pool. Accordingly, the margins of safety for these considerations in the upper containment pool are not significantly reduced by the proposed rerack.
C.
SUMMARY
The licensee's request to expand GGNS spent fuel storage pool and upper containment pool capacities satisfies the following conditions: (1) The storage capacity expansion method consists of modifying a portion of the existing racks with a design which allows closer spacing between stored spent fuel assemblies; (2) the storage capacity expansion method does not involve rod consolidation or double tiering; (3) the K of the pools cre maintained less than or equal to eff 0.95; and (4) no new technology or unproven technology is utilized in either the 1
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V construction process or the analytical techniques necessary to justify the expansion. Consequently, the request does not involve significant hazards con-sideration in that it: (1) does not involve a significant increase in the probability or consequences of an accident previously evaluated, (2) does not create the possibility of a new or different kind of accident from any accident previously evaluated, and (3) does not involve a significant reduction in a margin of safety.
Accordingly the Commission proposes to determine that these changes do not involve a significant hazards consideration.
The Commission is seeking public comments on this proposed determination.
Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. The Commission will not normally make a final determination unless it receives a request for a hearing.
Con.ments should be addressed to the Rules and Records Branch, Division of Rules and Records, Office of Administration, U.S. Nuclear Regulatory Comission, Washington, D.C.
20555.
By
, the licensee may' file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written petition for leave to intervene.
Requests for a hearing and petitions for leave to intervene shall be filed in accordance with the Comission's " Rules of Practice for Domestic Licensing Proceedings" in 10 CFR Part 2.
If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or
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an Atomic ~ Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of hearing or an appropriate order.
As required by 10 CFR 92.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following factors: (1) the nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the pro-ceeding; and (3) the possible effect of any order which nay be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect (s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to fifteen (15) days prior to the first pre-hearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.
Not later than fifteen (15) days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter, and the bases for each contention set forth with reason-able specificity. Contentions shall be limited to matters within the scope of
p rmU the amendment under consideration. A petitioner who fails to file such a supple-ment which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party.
The Commission hereby provides notice that this is a proceeding on an application for a license amendment falling within the scope of section 134 of the Nuclear Waste Policy Act of 1982 (NWPA), 42 U.S.C. Q 10154. Under section 134 of the NWPA, the Commission, at the request of any party to the proceeding, is authorized to use hybrid hearing procedures with respect to "any matter which the Commission determines to be in controversy among the parties." The hybrid procedures in section 134 provide for oral argument on matters in controversy, preceded by discovery under the Commission's rules, and the designation, following argument, of only those factual issues that involve a genuine and substantial dispute, together with any remaining questions of law, to be resolved in an adjudicatory hearing. Actual adjudicatory hearings are to be held on only those issues found to meet the criteria of section 134 and set for hearing after oral argument, The Commission's rules implementing section 134 of the NWPA are found in r
l 10 CFR Part 2, subpart K, " Hybrid Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors" (published at 50 FR 41662 (October 15,1985). Under those rules, any party to the proceeding may invoke the hybrid hearing procedures by filing with the presiding officer a written request for oral argument under 10 CFR 2.1109.
To be timely, the request must be filed within ten (10) days of an order granting a request for hearing or petition to intervene.
(As outlined above,
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the Commission's rules in 10 CFR Part 2, subpart G continue to govern the
filing of requests for a hearing or petitions to intervene, as well as the admission of contentions.) The presiding officer shall grant a timely request for oral argument. The presiding officer may grant an untimely request for oral argument only upon a showing of good cause by the requesting party for the failure to file on time and after providing the other parties an opportunity to respond to the untimely request.
If the presiding officer granti a request for oral argument, any hearing held on the application shall be conducted in accordance with the hybrid hearing procedures.
In essence, those procedures limit the time available for discovery and require that an oral argument be held to determine whether any contentions must be resolved in an adjudicatory hearing.
If no party to the proceeding makes a timely request for oral argument, and if all untimely requests, if any, for oral argument are denied, then the usual procedures in 10 CFR Part 2, subpart G apply.
Subject to the above requirements, and any limitations in the order granting leave to intervene, those permitted to intervene become parties to the proceeding, have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross-examine witnesses.
If a hearing is requested, the Commission will make a final detemination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held.
If the final determination is that the ene:far' equest involves no signifi-cant hazards consideration, the Commission may issue the amendment and make it effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment.
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I If the final determination is that the amendment involves a significant hazards consideration, any hearing held would take place before the issuance of[anyamendment.
Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards con-sideration. The final cetermination will consider all public and State comments l
received. Should the Commission take this action, it will publish a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently.
A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Conmission, Washington, D.C.
20555, Attention: Docketing and Service Branch, or may be delivered to the Commission's Public Document Room, 1717 H Stree'., N.W.
Washington, D. C., by the above date. Where petitions are filed during the last ten (10) days of the notice period, it is requested that the petitioner promptly so inform the Commission by a toll-free telephone call to Western Union at (800) 325-6000 (in Missouri (800) 342-6700). The Western Union operator should be given Datagram Identification Number 3737 and the following message addressed to Walter R. f5utler: petitioner's name and telephone number; date
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. 3 petition 'as mailed; plant name; and publication date and page number of this w
FEDERAL REGISTER notice. A copy of the petition should also be sent to the Deputy General Counsel, U.S. Nuclear Regulatory Commission, Washington, D.C.
l 20555, and to Nicholas S. Reynolds, Esquire, Bishop, Liberman, Cook, Purcell and Reynolds, 1200 17th Street, N.W., Washington, D. C.
20036, attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board designated to rule on the petition and/or request, that the petitioner has made a substantial showing of good cause for the granting of a late petition and/or request. That determination will be based upon a balancing of the factors specified in 10 CFR 2.714(a)(1)(1)-(v) and 2.714(d).
For further details with respect to this action, see the application for amendment which is available for public inspection at the Comission's Public Document Room, 1717 H Street, N.W., Washington, D. C.
20555, and at the Hinds Junior College, McLendon Library, Raymond, Mississippi 39154.
Dated at Bethesda, Maryland, this 14th day of July 1986.
FOR THE NUCLEAR REGULATORY COMMISSION Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensing
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. petition was mailed; plant name; and publication date and page number of this FEDERAL REGISTER notice. A copy of the petition should also be sent to the Deputy General Counsel, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, and to Nicholas S. Reynolds, Esquire, Bishop, Liberman, Cook, Purcell and Reynolds, 1200 17th Stree, N.W., Washington, D. C.
20036, attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board designated to rule on the petition and/or request, that the petitioner has made a substantial showing of good cause for the granting of a late petition and/or request.
t determination will be based upon a balancing of the factors specified in 10 CFR 2.714(a)(1)(1)-(v) and 2.714(d).
For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's Public Document Room, 1717 H Street, N.W., Washington, D. C.
20555, and at the Hinds Junior College, McLendon Library, Raymond, Mississippi 39154.
Dated at Bethesda, Maryland, this 14th day of July 1986.
FOR THE NUCLEAR REGULATORY COMMISSION Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensing
- Previously concurred:
PD#4/LA PD#4/PM OELD PD#4/D
- M0'Brien
- LKintner:lb
- MYoung
- WButler 07/08/86 07/08/86 07/10/86 07/14/86