ML20210A891
| ML20210A891 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Diablo Canyon |
| Issue date: | 01/06/1986 |
| From: | Norris J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML082840462 | List:
|
| References | |
| FOIA-86-197 NUDOCS 8601160593 | |
| Download: ML20210A891 (20) | |
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7590-01 UNITED STATES NUCLEAR REGULATORY COPNISSION PACIFIC GAS AND ELECTRIC COMPANY DOCKET NOS. 50-275 AND 50-323 NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENTS TO i
FACILITY OPERATING LICENSES DPR-80 AND DPR-82 I
j FOR DIA8LO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2, RESPECTIVELY, AND PROPOSED NO SIGNIFICANT HAZARDS 1
CONSIDERATION DETERMINATION AND OPPORTUNITY FOR HEARING The U.S. Nuclear Regulatory Connission (the Connission) is coasidering issuance of amendments to Facility Operating License Nos. DPR-80 tnd DPR-82, issued to Pacific Gas and Electric Company (PG&E or the licensee), for I
operation of the Diablo Canyon Nuclear Power Plant, Units 1 and 2 respectively, located in San Luis Obispo County, California.
The licensee requested the amendments, including associated changes in the combined Technical Specifications for Units 1 and 2, in License Amendment Request LAR'85-13, dat'ed October 30, 1985. The licensee's evaluation of proposed plant modifications and operations in support of the amendment requests are contained in the Reracking Report submitted to the NRC by letter dated September 19, 1985.
The amendments would authorize the licensee to increase the Unit I and Unit 2 spent fuel pool storage capacity from 270 to 1324 storage locations for each unit. The proposed expansion is to be achieved by reracking the spent fuel pools with a combination of poisoned racks and nonpoisoned racks in a two-region arrangement._
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Region 1 consists of three rack modules containing 290 poisoned storage locations which are designed to acconinodate 1.5 reactor cores of high enrichment nuclear spent fuel. The racks are designed to store Westinghouse 17 X 17 fuel arrays with an initial enrichment of 4.5 weight percent U-235.
The region consists of fuel racks with a nominal center-to-center cell spacing of 10.93 inches. The spent fuel rack design for Region 1 is based upon the comonly accepted physics principle of a " neutron flux trap" with the use of neutron absorber materials. The poison material to be used is Boraflex.
Region 2 consists of 13 modules of nonpoisoned spent fuel racks with a nominal center-to-center cell spacing of 10.93 inches. These modules consist of 1034 storage locations and 10 miscellaneous cells. Region 2 is designed for fuel with an initial enrichment of 4.5 weight percent U-235 with a minimum i
burnup of 34.5 Wd/kgU. The spent fuel rcck design is based on the criticality acceptance criteria specified in Revision 2 of Regulatory Guide 1.13, which allows credit for reactivity depletion in spent fuel.
(Previously, the physics criteria for fuel stored in the spent fuel pool were defined by the maximum unfrradiated initial enrichment of the fuel.)
The spent fuel racks in both regions are fabricated from 304L stainless steel that is 0.08 and.0.09 inches thick in Regions 1 and 2, respectively.
Each rack module is a free-standing module that satisfies the seismic design requirements without mechanical dependence on neighboring modules or fuel pool walls for support. The rack nudules are classified as Seismic Category I equipment. Racks of similar design have been licensed for other nuclear facilities. The use of two diverse regions is not unique, and two region spent fuel pools have been previously approved by the Comission.
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Both regions of the spent fuel pool have been designed to store fuel assemblies in a safe, coolable, suberitical configuration with.the effective neutron multiplication factor, k,ff, less than or equal to 0.95.
The racks have been designed and will be provided by the Joseph Oat Corporation (Dat), Camden,NewJersey. The licensee has transmitted the supporting report (the "Reracking Report") on the design and analysis for the reracking of the spent fuel pools for the Diablo Canyon Nuclear Power Plant, j
Units 1 and 2, to the NRC by letter DCL-85-306, dated September 19, 1985. Oat i
racks of this type have been licensed most recently by the NRC for use at the V. C. Sumer Plant of South Carolina Electric and Gas Company, i
The amendments for Unit 1 and Unit 2 would include the following specific changes to the combined Technical Specifications for Units 1 and 2.
(1) Specification 3/4.9.13 and associated Bases (restrictions on location of spent fuel prior to shipping cask operations near the spent fuel pool) would be revised to establish an exclusion zone in the vicinity of the cask pit in each spent fuel pool that prohibits movement of the shipping cask near the spent fuel pool with spent fuel in the exclusion zone. The present requirement prohibits shipping cask movement near the spent fuel pool u,nless spent fuel located in' existing Racks 5 and 6 has decayed for at least 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> since shutdown.
(2) Spec 1'fication 3/4.9.14 and associated Bases (a new specification limiting Region 2 storage and establishing boron concentration requirements) t l
would include (1) restrictions on the combination of initial enrichment and i
i cumulative burnup for spent fuel assemblies stored in Region 2, and (2) a
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requirement to maintain the boron concentration of the spent fuel pool greater f
than or equal to 2000 ppm.
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(3) Design Feature 5.6.1.1 (spent fuel storage rack design requirements to prevent criticality) would be revised to allow use of borated water to maintain k,ff lessthanorequalto0.95,deletethequantitativemeasureof uncertainty allowance, and reduce the allowable center-to-center distance between fuel assemblies to a nominal 10.93 inches. The present requirements specify use of unborated water to maintain k,ff less than or equal to 0.95, include a 2.6% k/k uncertainty allowance, and specify a nominal 21-inch center-to-center distance between fuel assemblies.
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I Design Feature 5.6.3 (capacity of spent fuel storage pool) would be l
revised to limit the storage capacity to no more than 1324 fuel assemblies, compared with the present limit of 270 fuel assen611es.
The Comission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exits by providing certain examples of amendments that are considered likely, and not likely, to involve significant hazards considerations. These examples were published in the FEDERAL REGISTER on April 6, 1983 (48 FR 14870). Spent fuel pool reracking was specifically excluded from the list of examples considered likely to involve a significant hazards consideration. Pending further study of this matter, the Comission is making a finding on the question of no significant hazards consideration such as this on a case-by-case basis, giving full consideration to the technical circumstance of the case using the standardsof10CFR50.92(48FR14869).
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 capacity-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 analyzed.
(2) Does increasing the spent fuel storage capacity, create the possibility of a new or different kind of accident from any accident previously analyzed? With respect to the Diablo Canyon Nuclear Power Plant.
Units 1 and 2, 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
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have been found to be conservatively bounded by the evaluations cited in the safety evaluation reports supporting each amendment. (3) Does increasing the i
. spent fuel pool storage capacity significantly reduce a margin of safety? The staff has not identified significant reductions in safety margins due to increasing the storage capacity of spent fuel pools. The expansion may result in a minor increase in pool temperature, but this heat load increase is generally well within the design limitations of the installed cooling systems.
In some cases it may be necessary to increase the heat removal capacity by relatively minor changes in the cooling system, e.g. by increasing a pu,mp capacity. 5utinall'casesthetemperatureofthepoolwillremainbelow 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 (k,ff). However, after extensive study, the staff detemined in 1976 that as long as the maximum
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I neutron multiplication 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 techniques utilized to l
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d calculate k,ff have been bench-marked against experimental data and are considered very reliable. Reracking to allow a closer spacing between fuel assemblies can be done by proven technologies.
In sumary, replacing existing racks with a design which allows closer spacing between stored spent fuel assemblies is considered not likely to involve significant hazards considerations if several conditions are met.
First, no unproven technology is utilized in either the construction process or in the anaytical techniques necessary to justify the expansion. Second, the K,ff of the pool is maintained less than or equal 'to 0.95.
Reracking to allow closer spacing satisfies these criteria.
l The licensee's submittal of October 30, 1985 included a discussion of the proposed action with respect to the issue of no significant hazards consider-ation. This discussion has been reviewed and the Comission finds it acceptable. Pertinent portions of the licensee's discussion, addressing each of the three standards, is provided herein.
The licensee's evaluation references specific sections of the reracking report included in the submittal dated September 19, 1985. The analysis of the proposed reracking was accomplished using applicable portions of cu.rrently acceptable codes and s'tandards as specified in Section 3.2 of the reracking report. The results of the licensee's analysis in relation to the three significant hazards consideration standards are described below.
A.
First Standard Involve a significant increase in the probability or consequences of an accident previously evaluated.
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The licensee's safety analysis of the proposed reracking (LAR 85-13, October 30,1985) has been accomplished using current NRC Staff accepted Codes and Standards applicable to the Diablo Canyon Power Plant. The results of the
9-safety analysis demonstrate that the proposal meets the acceptance criteria set forth in these standards.
In addition, the licensee has reviewed NRC Staff safety evaluations for prior spent fuel pool rerackings involving spent fuel pool rack replacements to ensure that there are no identified concerns not fully addressed. The licensee has identified no such concerns.
As part of its evaluation, the licensee identified the following potential abnomal conditions and accident scenarios:
(1) spent fuel assembly drop in spent fuel pool (2) loss of spent fuel pool cooling (3) seismic event (4) tornado-generated missiles (5) spent fuel shipping cask dron (6) criticality accident (7) installation accident The first five cases are initiated either by external events, such as a seismic event, or by failure of an engineered system, such as dropping a fuel assembly. The probability of occurrence of any of the first five abnomal considerations and accidents is not affected by the racks themselves; thus reracking c'annot incrIase the probability of these conditions and accidents.
Spent Fuel Assembly Drop - Reracking does not affect the probability of this event. The consequences of a spent fuel assemt,1y drop in the spent fuel pool are discussed in Section 4.0 of the Reracking Report.
It is conserva-t tively assumed that, because of the energy associated with an assembly droo, the equivalent number of rods in one fuel assembly would be ruptured, reoard-less of whether they are in the dropped essembly or in one or more impacted assemblies. Therefore, the fission product inventory available for release due to this accident will be unchanged as a result of reracking.
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- this accident condition, the maximum effective neutron multiplication factor, k,ff, is less than 0.95.
The conclusions on the radiological consequences of a spent fuel assembly drop as presented in the Diablo Canyon FSAR (Chapter 15) remain valid, and offsite radiological dose consequences are well within 10 CFR 100 limits. Thus, the consequences of this type of accident will not be increased from those previously by evaluated in the FSAR.
Loss of Spent Fuel Pool Cooling - Reracking does not affect the probability of this event. The consequences of a loss of spent fuel pool cooling have been evaluated and are described in Section 5.0 of the Reracking Report. Because of the radioactive decay of the bulk of the stored spent fuel, there will be a negligible incremental increase in heat load in the pool as a result of rereacking. Also because of this decay, there will be only a negligible increase in offsite doses which would continue to be well within 10 CFR 100 limits.
In addition, if the loss of pool cooling occurred, there would be sufficient time to restore the cooling system or establish makeup water flow. A conservative analysis was perfonned to determine offsite radiological doses associated with a postulated spent fuel pool boiling event.
The assumptions used to v.lculate the heat generation and evaporation rates and the offsite doses'.,for loss of cooling to the spent fuel pool are discussed in Section 7.0 of the Reracking Report.
For the foregoing reasons, the consequences of this type of accident will not be significantly increased from those previously evaluated in the FSAR for loss of spent fuel pool cooling.
Seismic Event - Reracking does not affect the probability of this event.
The consequences of seismic events have been evaluated and are described in Section 6.0 of Reracking Report. The racks were evaluated against the appropriate codes and standards described in Section 3.0 of the Reracking Report. The racks are designed to Seismic Category I requirements. The
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-S-analysis methodology and techniques and acceptance criteria are the same as those used in reracking applications by other licensees which have been approved by the NRC staff. The results of the Diablo Canyon analysis show that the proposed racks and fuel pool structure meet the structural acceptance criteria applicable to the Diablo Canyon Nuclear Power Plant.
Because of the i
decay of fission products discussed above, the consequences of seismic events with increased fuel storage in the pool will not significantly increase from those previously evaluated in the FSAR.
Tornado-Generated ;11ssiles - Reracking does not affect the probability of this event. The consequences of tornado-generated missile impacts have been analyzed and are descri' ed in the FSAR, which concludes that the spent fuel o
storage pools and associated racks have adequate protection against tornado forces and tornado-generated missiles. The rack design does not affect the evaluation provided in the FSAR. Because of radioactive decay, discussed above, the consequences of tornado-generated missiles will not be significantly increased frov. those previously evaluated.
Spent Fuel Shipping Cask Drop - Reracking does not adversely affect the probability of this event. Based on an analysis of the worst case spent fuel shipping cask tipping accident, the licensee'has proposed amended Technical Specifications which would preclude spent fuel shipping cask movement near the spent fuel pool. This provides an exclusion zone to protect the spent fuel which would reduce the probability of the event. To determine the consequences of a shipping cask drop, the licensee has evaluated which spent fuel storage racks could be impacted in the event of a cask drop. This was done to determine an adequate space where storage of spent fuel wo~uld not be allowed during shipping cask movement near the spent fuel pool. Because of radioactive decay, discussed earlier, the consequences of dropping a spent
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-10 fuel shipping cask will not significantly increase from those previously evaluated in the FSAR.
Criticality Accident - A discussion of the potential for criticality accidents is discussed in Sections 4.0 of the Reracking Report. Postulated events that could potentially involve accidental criticality were examined by the licensee.
It was concluded that the limiting value for criticality (k,ff of 0.95) would not be exceeded. With the inclusion of administrative controls as required in the amended Technical Specifications to (1) maintain the boron concentration in the fuel pools at a minimum of 2000 ppm, and (2) to limit storage of spent fuel assemblies in Region 2 of the spent fuel storage racks based on initial enrichment and cumulative exposure, none of the postulated events would result in a criticality accident. Therefore, the probability and consequences of a criticality accident are not significantly increased from those previously evaluated in the FSAR.
Installation Accident - Timely approval of the license amendment will pemit reracking in a dry, empty pool which would preclude the consideration
' of events which would have radiological consequences. Worker radiation exposure would be less than that which would be received if reracking were delayed beyond the fir'st refueling outage when spent fuel assemblies would be present in the pool during reracking.
With respect to wet rack installation with spent fuel in a pool, the Technical Specifications prohibit handling of loads in excess of 2500 pounds above fuel stored in the fuel pools, and strict administrative controls are in place which preclu6e the movement of racks or other heavy loads above stored
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spent fuel. The consequences of damage from a rack drop to the pool floor or liner would be similar to the previously analyzed cask drop accident.
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Both the old and the new racks fall into the category of heavy loads as defined in NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants."
- Installation will be performed consistent with the licensee's previous responses to the NUREG guidelines.
In SSER-27 and SSER-31, dated July 1984 and April 1985, respectively, the NRC staff concluded that the Diablo Canyon program for control of heavy loads.was in compliance with the requirements of NUREG-0612.
In sumary, therefore, it is concluded that the proposed amendments to rerack the spent fuel pools for Units 1 and 2 will not involve a significant increase in the probability and consequence of an accident previously evaluated.
B.
Second Standard Create the possibility of a new or different kind of any accident previously evaluated The licensee has evaluated potential accidents associated with the proposed reracking in accordance with the design bases specified in the Diablo Canyon FSAR, the guidance contained in NRC position paper "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications",
appropriate *NRC Regulatory Guides and Standard Review Plans, and appropriate industry Codes and Standards as listed in the Reracking Report.
No unproven technology will be utilized either in the construction
- process or in the analytical techniques necessary to justify the planned fuel storage expansion. The basic reracking technology in this instance has been developed and demonstrated in numerous applications for a fuel pool capacity i
increase which have already received NRC staff approval.
The change to a two-region spent fuel pool requires the performance of additional evaluatians to ensure that the criticality criterion is maintained.
These include the evaluation of the limiting criticality condition, i.e.,
dropping or misplacement of an unirradiated fuel assembly of 4.5 weight percent enrichment into a Region 2 storage cell or outside and adjacent to a Region 2 rack module. The evaluation for this case shows that when the boron concentration meets the proposed Technical Specifications requirement, the criticality criterion is satisfied. Although this change does create the requirement to address additional aspects of a previously analyzed accident, it does not create the possibility of a previously unanalyzed accident.
The licensee concludes that the proposed reracking does not create the possibility of a new or different kind of accident from any accident previously evaluated for the Diablo Canyon spent fuel storage facilities.
C.
Third Standard Involve a significant reduction in a margin of safety The issue of " margin of safety", when applied to a reracking modification, includes the following considerations:
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nuclear criticality considerations b.
thermal-hydraulic considerations c.
mechanical, material, and structural considerations The margin of safety that has been established for nuclear criticality considerations is that the effective neutron multiplication factor (k,ff) in the spent fuel pool is to be less than or equal to 0.95, including all reasonable uncertainties and under all postulated conditions. The criticality analysis for the proposed modification is described in the licensee's Reracking Report. The. NRC Staff determined in 1976 that as long as the maximum value of the effective neutron multiplication factor, k,ff, was less than or equal to 0.95, then any change in pool reactivity. would not significantly reduce the margin of safety, regardless of the storage capacity
of the pool. The methods used in the criticality analysis for the reracking for Diablo Canyon Nuclear Power Plant, Units 1 and 2 conform to the applicable portions of Codes. Standards, and specifications listed in Section 4.0 of the Reracking Report, including ANSI N210-1976, " Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations"; ANSI N16.9-1975,
" Validation of Calculation Methods for Nuclear Criticaility Safety"; the NRC 9uidance, "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" (April 1978), as modified (January 1976); and Regulatory Guide 1.13. " Spent Fuel Facility Design Basis," proposed Revision 2.
The computer programs, data libraries, and benchmarking data used in the evaluction have been used in previous spent fuel reracking applications by other licensees and have been reviewed and approved by the NRC. The licensee has performed criticality analyses for the Diablo Canyon Nuclear Power Plant,
- Units 1 and 2 reracking assuming operation of the spent fuel storage facilities consistent with the proposed Technical Specifications. The results of these analyses indicate that k,ff is less than 0.95 at a 95/95 probability / confidence level under all postulated conditions, including a margin for uncertainties in reactivity calculations and mechanical tolerances.
Thus, in meeting the acceptance criteria for. criticality, the proposed reracking does not involve a significant reduction in the margin of safety for nuclear criticality.
From a thermal-hydraulic consideration, the areas of concern when evaluating whether there is a significant reduction in margin of safety are:
(1) maximum fuel cladding temperature, and (2) the increase in spent fuel pool water temperature. The licensee's thermal-hydraulic evaluation is-described in Section 5.0 of the Reracking Report. The new storage configuration will result in an increase in the maximum heat load in the spent fuel pool. The l
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maximum spent fuel pool temperature will not exceed 150*F for a partial core discharge and 175'F for a full core discharge. Nonetheless, the fuel cladding temperatures under all conditions are sufficiently low to preclude structural failure. Thus, it is concluded that there is no significant reduction in the margin.of safety with respect to thermal-hydraulic considerations or spent fuel cooling considerations.
The mechanical, material, and structural considerations of the propcsed rack replacement are also analyzed in the licensee's Reracking Report. The~
racks are designed in accordance with applicable NRC Regulatory Guides, Standard Review Plans, position papers, and appropriate industry Codes and Standards, as well as to Seismic Category I requirements. The materials utilized are compatible with the spent fuel pool and the spent fuel assemblies. The conclusion of the analysis is that the margin of safety is not significantly reduced by the proposed reracking.
The main function of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a stable configuration through all nonnal and abnonnal loadings, such as an earthquake, and under accident conditions.
Nuclear criticality, thermal-hydraulic, material, and structural considerations of the ' proposed new racks are described in the licensee's Reracking Report. The neutron poison and rack materials are compatible with materials used for the spent fuel pool liner and the spent fuel assemblies.
The rack structural considerations address adequate margins of safety of critical items during seismic motion and the racks are seismically qualified.
Thus, the margins of safety are not significantly reduced by the proposed expansion of pool storage capacity.
The licensee's request to expand the Diablo Canyon Nuclear Power Plant, Units 1 and 2 spent fuel storage pool capacities satisfies the following
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(1) the storage expansion method consists of replacing existing racks with a design which allows closer spacing between stored spent fuel assemblies; (2) the storage expansion method does not involve rod consolidation or double tiering; (3) the k,ff of the pool is maintained less than or equal 0.95; and (4) no unproven technology is utilized in either the construction process or the analytical techniques necessary to.iustify the expansion.
On the basis of the foregoing discussion of the elements of 10 CFR 50.92 and because the proposed reracking technology has been well developed and demonstrated, the Commission proposes to determine that operation of the facility in accordance with the proposed amendment does not involve a significant hazards consideration.
The Comission is seeking public comments on this proposed detemination.
Any comments received within 30 days after the date of publication of this notice will be considered in making any final detemination. The Comission will not nomally make a final detemination unless it receives a reouest for a hearing.
Coments should be addressed to the Rules and Procedures Branch, Division of Rules an'd Records,k.0ffice of Administration, U.S. Nuclear Regulatory Comission, Washington, D. C.
20555.
By February 12,1986 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. Request for a hearing and petitions for leave to intervene shall be filed in accordance with the Comission's Rules of Practice for Domestic Licensing Proceedinos" in 10 CFR Part 2.
If a request
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- for a hearing or petition for leave to intervene is filed by the above date, the Comission or an Atomic Safety and Licensing Board, designated by the Comission or by the Chainnan 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 12.714, 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)thenature 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 proceeding; and (3) the possible effect of any order which may 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 pro-ceeding 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 prehearing 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 is required to file a supplement to
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the petition to intervene which must include a list of the content 1ons which are sought to be litigated in the matter, and the bases for each contention set forth with reasonable specificity, pursuant to 10 CFR 62.714(b).
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Contentions shall be limited to matters within the scope of the amendment under consideration. A petitioner who fails to file such a supplement which satisfies ~these requirements with respect to at least one contention will not be permitted to participate as a party.
The Comission hereby provides notice that this is a proceeding on an application for a license amendment falling within the scope of section 143 of the Nuclear Waste Policy Act of 1982 (NWPA), 42 U.S.C. I 10154. Under section 143 of the NWPA, the Comission, at the request of any party to the proceeding, is authorized to use hybrid hearing procedures with respect to "any matter which the Comission deter 1 nines to be in controversy among the parties." The hybrid procedures in section 143 provide for oral argument on matters in controversy, preceded by discovery under the Comission'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 Comission's rules implementing section 134 of the NWPA are fo,und in 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).
10 CFR 62.1101 g seg. 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.
(Ksoutlined above, the Comission's rules in 10 CFR Part 2, Subpar't G, and 52.714 in particular, continue to govern the filing of requests for a hearing or
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petitions to intervene, as well as the admission of contentions.) The presiding officer shall grant a timely request for oral argument. The l
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 l
and after providing the other parties an opportunity to respond to the untimely request.
If the presiding officer grants 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 i
time available for discovery and require that an oral argument be held to
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detemine whether any contentions must be resolved in an adjudicatory hearing.
If no party to the proceeding requests oral argument, or if all untimely
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requests 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 pemitted to intervene become parties to the proceeding and have the opportunity to participate fully in the conduct of any hearing which is held, including the opportunity to present evidence and cross-examine witnesses at such hearing.
If a hdaring is requested, the Comission will make a final determination on the issue of no significant hazards consideration. The final detemination will serve to decide when the hearing is held.
If the final detemination is that the amendment request involves no significant hazards consideration, the Comission 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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19 If the final determination is that the amendment involves a significant hazards consideration, any hearing held would take place before the issuance l
of any amendment.
Nomally, the Comission will not issue the amendment until the expira-tion 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 Comission may issue the license amendment before the expiration of the 30-day notice period, provided that its final detemination is that the amendment involves no significant hazards consideration. The final detemination will consider all public and State coments received. Should the Comission take this action, it will publish a notice of issuance and provide for opportunity for a hearing after issuance. The Comission 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 Comission, U.S. Nuclear Regulatory Comission, Washington, D.C.
20555, Attention: Docketing and Service Branch, or may be delivered to the Comission's Public Document Room,1717 H Street, N.W., Washington, D.C), by the above date.. here petitions are filed during W
the last ten (10) days of the notice period, it is requested that the petitioner promptly so infom the Comission 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 Steven A. Varga: petitioner's name and
$elephonenumber;datepetition~wasmailed;plantname;andpublicationdate and page nunter of the FEDERAL REGISTER notice. A copy of the petition should j
also be sent to the Executive Legal Director, U.S. Nuclear Regulatory I
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Commission, Washington, D.C.
20555, and to Philip A. Crane, Esq., Richard F.
Locke, Esq., Pacific Gas and Electric Company, P. O. Box 7442., San Francisco, California 94120 and Bruce Norton, Esq., Norton, Burke, Berry and French, P.O. Box 10569, Phoenix, Arizona 95064.
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 pre-siding Atomic Safety and Licensing Board, that the petition and/or request should be granted 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 (LAR 85-13, dated October 30,1985) that is available for public inspection at the Commission's Public Document Room,1717 H Street, N.W.,
Washington, D.C., and at the California Polytechnical State University Library, Government Documents and Maps Department, San Luis Obispo, California 93407.
Dated at Bethesda, Maryland, this 6th day of January,198bt FOR THE NUCLEAR REGULATORY COMMISSIO_N I
j$7 c43 -
v Jan A. Norris, Act ng Director Div(isionofPWRLicensing-A PW:
Project Directorate No. 3 d
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