ML19347E575
| ML19347E575 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 05/07/1981 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19347E574 | List: |
| References | |
| DD-81-05, DD-81-5, NUDOCS 8105130035 | |
| Download: ML19347E575 (18) | |
Text
DD-81 5 h)
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION HAROLD R. DENTON, DIRECTOR In the Matter of
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Docket Nos. STN 50-454
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STN 50-455 COMMONWEALTH EDIS0N COMPANY
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(Byron Station, Units 1 and 2)
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DIRECTOR'S DECISION UNDER 10 CFR 2.206 By letter dated November 21, 1980, the Rockford League of Women Voters (the " League")
transmitted a petition pursuant to 10 CFR $2.206(a) requesting that the Director of Nuclear Reactor Regulation initiate a proceeding pursuant to 10 CFR 52.202 to:
(a) Modify ths construction permits issued to Comonwealth Edison Company for the Byron Nuclear Power Station, Unit 1 and 2, so that construction may not proceed without resolution of all outstanding safety problems presently applicable to the Byron Station; (b) Suspend or revoke the authority earlier granted to Commonwealth Edison Company to ;onstruct the Byron Station until such time as Commonwealth Edison has formulated an acceptable and realistic plan for resolving all outstanding safety problems; (c) Revoke the construction permits issued to Commonwealth Edison Company for the Byron Station, if the outstanding safety problems cannot be resolved prior to the completion of construction due to Edison's financial condition or for other reasons; and i
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(d) Pending full hearings and determinations on these requests, immediately halt further construction of the Byron Station.
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By my letter of December 22, 1980 to the League, I acknowledged receipt of the League's request of November 21, 1980 and denied request (d) for an immediate halt to further construction of the Byron Station. Pursuant to 10 CFR 2.206, this decision is my response to the League's remaining requests (a), (b) and (c).
In developing this response, I have also considered a supplemental, though unsigned, affidavit which the League submitted by letter dated January 27, 1981.
Discussion The construction permits for the Byron facility were issued by the Office of Nuclear Reactor Regulation on December 31, 1975. Consequently, the Byron facility is 4
currently being constructed pursuant to valid construction permits.
The League's 2/
petition raises principally questions concerning the adequacy of plant design.-
In the construction permit review, the staff primarily reviews design criteria and the plant's preliminary design.
Information regarding the detailed design of the plant is not required for the issuance of a construction permit. Detailed design features of the plant are generally developed after issuance of the con-struction permit and are evaluated and approved during the course of the staff's review of the operating license application.
In the interim, a licensee pursues
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Notice of recoipt of the petition was published in the Federal Register on December 31,1980 (45 FR 86584). Counsel for Commonwealth Edison submitted comments on the petition in a letter of February 13, 1981.
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The Byron facility is currently undergoing an operating license review.
Notice of receipt of application and availability of opportunity for a hearing was published in the Federal Register on December 15, 1978 (43 FR 58659). An operating license hearing will be held for the Byron facility and the trague has been admitted as an intervenor in that pro-
- ceeding, j
- construction work under a construction permit at its own risk pending approval of the final design of the plant.
Investment made by a utility in constructing a facility is not a proper factor when considering the issuance of an operating license. PRDC v. International Union of Electrical, Radio and Machine Workers, 367 U.S. 396, 415 (1961). Prior to receiving an operating license, the applicant will be required to do anything necessary to ensure safe operation of the plant.
The ccst or difficulty associated with implementing actions needed to ensure safety are not relevant considerations to this agency. Public Service Company of New Hampsbire, et al (Seabrook Station, Units 1 and 2), ALAB-623,12 NRC 670, 677-78 (Dec. 9, 1980). The safety standards which an applicant must meet at the operating license stage are unconditional. Consecuently, the League's concern that continued construction of the Byron facility might bias safety decisions at the operating license stage is not well founded given the NRC's responsibility and the unconditional safety standards that this agency applies.
See Porter County Chapter of the Izaak Walton League, Inc. v. NRC, 606 F.2d 1363,1369-70 (D.C.
Cir. 1979).
Reactor licensing is a two-stage process, the first stage being the consideration of issuance of a construction permit and the second being the consideration of an operating license for a facility. This two-stage process has been sanctioned by the Supreme Court in the PRDC case, and the courts have recognized that the institution of proceedings prior to the operating license stage is not mandated whenever an unresolved safety question is raised after issuance of a construction pe rmi t.
Porter County Chapter, supra, 605 F.2d at 1367, 1369.
In the words of the Court of Appeals for the Districtof Columbia Circuit, " permitting continued
s construction of the plant despite unresolved safety questions does not of itself pose any danger to the public health and safety".
Id. at 1369.
Thus, in the absence of extraordinary circumstances, I will not upset the Commission's usual two-stage licensing process by ina tituting a proceeding in response to a 10 CFR 2.206 petition to consider issues that are properly within the scope of the operating license review.
Public Service Company of Indiana (Marble Hill Nuclear Generating Station, Units 1 & 2), DD-79-21, 10 NRC 717, 720 (1979). This approach is in accord with the basic principle established by the Commission ~ that 10 CFR 2.206 is not to be used as a device to avoid a forum in which issues "more logically should be presented". Consolidated Edison Company (Indian Point, Units 1-3), CLI-75-8, 2 NRC 173,1977 (1975).
The League has raised many of the same issues in its petition under 10 CFR 2.206 as it has posed as contentions before the Licensing Board.
See LBP 30, 12 NRC 638 (Dec. 1980).
In this context, issues such as those raised Jy the League concerning detailed design features of the plant do not normally warrant suspension of construction, because such issues are resolved during the operating license review. As each, of the remaining portions of the League's request which I am considering in this decision, namely subparagraphs (a), (b) and (c), essentially request consideration now of ouestions which are more properly examined at the operating license stage, the petition may be denied on that ground alone. Nevertheless, I have examined each of the six safety areas which are addressed in the petition to determine if there are any exceptional circumstances which would warrant consideration of any agency action at this time.
Systems Interaction (Issue 1)
The League's concern is that the " systems interaction issue" may not receive adequate
consideration in connection with the Byron facility. The League in its petition identifies the systems interaction issue as dealing with related problems of accident and safety analysis. These include the lack of systems interaction analysis, the lack of multiple or "comon-cause" failure analysis, and the tendency of the " single-failure criterion" to exclude a laroe number of potential accident causing events.
Sandia National Laboratories, under contract assistance to the NRC staff, began the first of an intended two-phase program in May 1978 to evaluate whether present review procedures and safety criteria ensure an acceptable level of redundancy and independence for systems required for plant safety.
This study was organized to provide an independent investigation of safety functions, and systems required to perfonn these functions, in order to assess the adequacy of current review procedures.
It was conducted by evaluating the potential for undesirable interactions between and among systems.
Phase I of the work was completed with the issuance of a report by Sandia National Laboratories, " Final Report - Phase I Systens Interaction Methodology Applications Program," NUREG/CR-1321, SAND 80-0384,( April 1980). The results "9t report generally support continued use of NRC review procedures so long as sg.em interaction assessments are not restricted to the safety systems identified in the Standard Review Plan.
A follow-on program for FY '81 and FY '82 was developed by the staff with technical support from Battelle and Brookhaven Laboratories to prepare interim regulatory gudance for systems interaction evaluation in Light Water Reactor (LWR) plants.
It is further intended to test this guidance during FY '82 on several pilot LWR's, for purposes of identifying and evaluating significant systems interactions.
The results of this effort will be reflected in final regulatory guidance. Systems interactions that are determined to be adverse as a result of this effort will be evaluated for other plants not included in the pilot program.
In summary, the issue of systems interactions has been and is still under current study by the staff. Our contractor's findings indicate that with prudent con-sideration of interaction of systems identified as important to safety with non safety systems, the Standard Review Plan (SRP), which is being used to conduct the Byron review, ensures a substantive review of the safety aspects of systems interactions. To the extent continued investigations by the staff identify a need for modifications to the SRP or to the Byron facility, such modifications will be applied to Byron as needed and when identified.
Steam Generator Tube Integrity (Issue 2)
Another concern of the league is that technical issues related to the integrity of steam generator tubes may not receive adequate treatment in connection with the Byron facility. The technical issues center around whether or not the Westinghouse steam generator tubes have the capability to maintain their integrity during normal operation and postulated accident conditioni..
In addition, the League is concerned that requirements for increased steam generator tube inspections and repairs nay result in significant increases in occupational exposures to workers.
Corrosion resulting in steam generator tube wall thinning has been observed in several Westinghouse and Combustion Engineering plants for a number of years. Major changes in the design of the secondary water treatment process essentially eliminated this form of degradation. Another major corrosion-related phenomenon has also been observed in a number of plants in recent years, resulting from a buildup of
support plate corrosion products in the annulus between the tubes and the support plates. This buildup eventually results in reduction of the cross sectional area of the openings in the tubes, ca". led " denting", and deformatinn of the tube support plates.
This phenomenon has led to other problems, including stress corrosion cracking, leaks at the tube / support plate intersactions, and cracking of U-bend sections of tubes which were highly stresseo because of 3/
support plate deformation. Task A-3 in NUREG-0410 has been organized to provide resolution of the problem of tube degradation due to denting in Westinghouse steam generators.
The quNtion of steam generator tube integrity is an issue which will be reviewed by my staff for Byron.
In the recently licensed Sequoyah facility, which uses Westinghouse steam generators, the staff closely examined this very q';estion. They determined that specific measures adopted there, such as changes to steam generator design features, providing for a secondary water chemistry control and monitoring program, adding condensate demineralization features and the careful selection of condenser tubing materials, would minimize the onset of steam generator tube 4/
r eobl ems.-
In addition, inservice inspection provisions a'd Technical Specifications requirements for actions to be taken in the event that steam generator tube leakage occurs during plant operation provided a sufficient basis for a conclusion that Sequoyah Units 1 and 2 could be safely operated prior to the ultimate resolution of the steam generat-tube integrity issue.
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NUREG-0410 Appendix F, Task A-3 We: tinghouse Steam Generator Tube Integrity.
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NUREG-0011 Supplement No.1, SER Related to Operation of Sequoyah Nuclear Plant, i'
Units 1 and 2, February 1980, pages C.9 and C.10.
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_g-Task A-3 is expected to result in improvements in our current requirements for inservice inspection of steam generator tubes.
These improvemerts will include a better statistical basis for establishing inservice inspection program require-ments and consideration of the cost / benefit of increased inspections.
The staff has not yet completed its review of the steam generator design for the Byron facility. However, they will review design and associated operating provisions, inservice inspection provisions and Technical Specifications requirements.
As for Sequoyah, the staff will determine whether there is a sufficient basis for a conclusion that Byron Station Units 1 and 2 can be operated prior to the ultimate resolutic9 of the generic issue.
In summary, the issue of steam generator tube inteprity has been handled on a case-by-case basis in th2 operating license review and is still under study by the staff. Also, Westinghouse has introduced improvements in its steam cenerator designs. Some of these improvements have been incorporated in facilities for which operating licenses have been issued, and improvements introduced in the Byron steam generators will be reviewed. The petition on this issue raises 4
no new concerns that are not already being pursued by the staff in Operating License reviews.
Equipment Oualif t:ation and Deterioration (Issue 3)
The League's concern here is whether there is sufficient assurance that safety-related equipment for the Byron fat ';11ty will function in the manner intended when subjected i
to environmental service co'iitions under anti;ipated normal, abnormal and accident 5/
conditions. As a result of wo'k completed under Task A-24, NUREG-0410,- the
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NUREG-0410 Appendix F, Task A-24 Environmental Oualification of Safety Related Electrical Equipment.
staff has issued its proposed resolution of this issue for reactors under licensing review in its report, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," NUREG-0588, dated December 1979. Also the Division of Operating Reactors issued " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines) in November 1979. On May 23, 1980, the Commission issued a Memorandum and Order that is applicable to the Byron facility.
In that Memorandum and Order the Commission stated its findinas that the current NRC requirements in the environmental cualification area, including compliance with the staff's position on equipment qualification (NUREG-0588), provided reasonable assurance that the public health and safety would be adequately protected.
Petition for Emerpency and Remedial Action, CLI-80-21, 11 NRC 707 (1980). The Commission stated:
"Furthermore, pursuant to Section 161(b) of the Atomic Energy Act and based on the record in this proceeding the Commission is ordering today that these two documents (NUREG-0588 and D0R Guidelines) form the requirements which licensees and applicants must meet in order to satisfy those aspects of 10 CFR 50, Appendix A, General Design Criteria (GDC)-4 whica relate to environmental qualification of safety-related electrical equi pment". This Memorandum and Order finalized the framework within which the issue will be reviewed. Given the absence of any new significant information on this issue in the petition, this issue is properly left for examination at the operatinu license review for the Byron facility.
Evaluation of Potential Accidents and Corrective Measures (Issna 4)
The League's concern here is that the consecuences of serious accidents, including those labelled " Class 9", may not receive adequate treatment in connection with the operating license review of the Byron facility. Specifically, the League calls for the development of liquid pathway interdiction systems for the Byron
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Station using models employed in a 1980 draf t Sandia study.~ This request is made by the League based on their understanding that the Sandia authors calculate large doses "in the case of plants with hydrogeologic features similar to the Byron site (i.e., high grounawater table and permeable rock)".
The League contends that "the potential radiation dose is approximately 2 to 7
5 x 10 person-rem (with uncertainties of an order of magnitude), which current studies tr"aslate into several thousand probable deaths from a major accident."
The draf t Sandia report, January 1980, was developed as a step in a study performed by the NRC. As a draft report, it has been available for information but has not in any way been adopted by the staff or the Comraission as representing a basis for regulatory decisions. Its availability for almost a year prior to issuance of the Sumer DES did not result in the staff's reliance on it in the Sumer analysis and the staff will not rely on it in its review of Byron. The staff will continue to review that draft report and subsequent reports by Sandia, as well as pertinent information from any other source, to determine whether the staff's approach in environmental reviews needs to be modified.
As I understand the League's petition, the League is asking that I initiate separate proceedings to consider Class 9 accidents, because the League does not believe that the environmental review for the Byron operating license will be subject to the Commission's new interim policy on consideration of severe accidents. A separate proceeding is unwarranted, since the Byron environmetnal review is indeed subject to the new policy.
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~ Sandia Study (Draft) for U.S.N.R.C., "Effect of Liquid Pathways on Consequences of Core Melt Accidents", January 1980.
In past NEPA reviews the staff's treatment of environmental consequences of postulated accidents has been guided by the Commission's proposed Annex to Appendix D of 10 CFR Part 50 (the " Annex") that was published for comment on December 1, 1971. Consideration of Accidents in Implementation of the National Environmental Policy Act of 1969, 36 FR 22851 (1971). The Commission issued a Statement of Interin Policy on June 13, 1980 in which it announced the withdrawal 7/
of the proposed Annex to Appendix D of 10 CFR Part 50.
It also announced its position that its Environmental Impact Statements shall include considerations of the site-specific environmental impacts attributable to accident seauences that lead to release of radioactive materials, including secuences that can result in inadequate cooling of the reactor fuel and to melting of the reactor In this regard, attention shall be given both to the probability of core.
occurrence of such releases and to environmental consequences of such releases.
Under the Commission's guidance, " releases refer to radiation and/or radioactive materials entering environmental exposure pathways, including air, water, and ground water." 45 FR at 40103 (emphasis added).
Although the environmental review for the operating license stage has not been started, the staff will follow the guidance of the interim policy statement of June 13, 1980.
In accordance with the interim policy statement, the staff will undertake a more extensive analysis of severe accidents in the environmental review of the Byron operating licenses.
Such analysis must include consideration of releases to air and liquid pathways, as the League requests.
It should be noted that the League, as a party to the operating license proceeding, has raised contentions on Class 9 accidents before the Licensing Board. See LBP-80, 7/
huclear Power Plant Accident Considerations Under the National Environmental Policy Act,0f 1969, 45 Fed. P.eg. 40101 (June 13, 1980).
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12 NRC 683, 692 (1880).
With regard to the liouid pathway analysis for the Byron facility, it is expected that implementation of the new policy will result in the application of a method of 8/
analysis similar to that used for the Virgil C. Summer Nuclear Station.- The approach in the Sunner evaluation was one of determining whether or not the Sumer site liquid pathway consecuences would be unioue when compared to land-based sites 9/
considered in the " Liquid Pathway Generic Study" (LPGS).--
This approach is conservative and provides bounding calculations. The LPGS was completed by the staff using realistic values of site parameters throughout, and the staff did not take into account mitigative measures. The actual method used in applying this approach to the Summer Plant consisted of a direct scaling of the LPGS population doses based on the relative values of key parameters characterizing the LPGS "small river" site and the Summer site. The individual and population doses for the liquid pathway in the LPGS ranged from a substantial fraction to a small fraction of those that can arise from the airborne pathways. The staff conclusion was that the Summer liquid pathway contribution to the population dose had been demonstrated to be the same order of magnitude as that predicted for the LPGS river site, which represents a " typical" river site. The staff then noted that there would be ample time to take measures which could be taken to minimize the impact of the liquid pathway.
Such measures might include slurry walls and well-point dewatering systems to isolate the radioactive contaminants at the source. These are conventional means for controlling and limiting groundwater movenent used in civil engineering and earthwork construction.
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NUREG-0534 Supplement, Draft Environmental Statement related to the operation of Virgil C. Summer Nuclear Station, Unit No.1, Docket No. 50-395, hovember 1980.
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NUREG-0440, Liquid Pathway Generic Study, February 1978.
3-In connection with its request for analysis of Cless 9 accidents, the League expresses concern that systems interaction evaluation and modifications and an Interim Reliability Evaluation Program (IREP) are not being implemented for the Byron Station. As explained in the NRC Action Plan, the IREP is a pilot program which utilizes a few typical plants to determine whether changes 10/
need to be made in the review of each plant. -
There is no pressing
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reason why Byron should be selected in lieu of another plant as one of the typical plants for the pilot program.
The Commission has not mandated that the IREP be applied to plants like Byron which are under review for operating licenses. The Action Plan describes a gradual implementation of IREP and studies for operating reactors, but IREP is not part of the requirements for new operating licenses outlined in NUREG-0737, Clarification of TMI Action Plan Recuirements.
In this regard:
"the Commission has concluded that the list of TMI-related requirements for new operating licenses found in NUREG-0737 can provide a basis for responding to the TMI-2 accident.
The Connission has decided that current operating license applications should be measured by the NRC Staff against the regulations, as augmented by these requirements.
In general, the remaining items of the Action Plan should be addressed through the normal process for development and adoption of new requirements rather than through inmediate imposition on pending applicants." Further Commission Guidance for Power Reactor Operating Licenses - Revised Statement of Policy, 45 FR 85236, 85238 (Dec. 8, 1980).
In summary, the staff is reviewing the issue of potential accidents and potential need for liquid pathway interdiction, both in a generic sense and on a case specific basis for Byron. The staff will conduct its review of Byron in accordance
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NRC Action Plan Developed as a Result of the THI-2 Accident, NUREG-0660, Vol. 1, at pp. II.C.-2 to II.C.5 (May 1980).
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" with the Commission's Statement of Interim Policy of June 13, 1980.
If new facts from the Sandia study, the systems interaction study, the IREP program l
or from other sources are shown to pose questions as to the validity of the current approach, an assessment will be made and, as appropriate, the review requirements modified and applied to Byron.
Conformance to Current Regulatory Practices (Issue 5)
The League's concern here is that the staff assessment of the Byron facility at the operating license review will consist of less than a complete review of plant design against all current Regulatory Guides and safety standards. Also, the League is concerned with construction quality assurance and control based upon its perception of the findings of the Office of Inspection and Enforcement's Inspection Reports.
In regard to compliance with Regulatory Guides, the League cites exceptions to Regulatory Guides taken by the applicant in the Byron Final Safety Analysis Report (FSAR) as demonstration of non-compliance with the identified Regulatory Guides. However, as Regulatory Guides are not regulations but merely one means of satisfying a regulatory requirement, conformance with Regulatory Guides is not a prerequisite to the issuance of any Commission license. The applicant, by submitting the FSAR with these exceptions, must demonstrate that the Commission's regulations will be met in the appropriate area by means other than exact conformance with the applicable Regulatory Guides.
Such proposed alternative
,means of complying with the Commission's regulations must be determined to be acceptable prior to a decision on issuance of an operating licenses for the Byron f acility.
In no event will ny staff recommend issuance of an operating license with respect to the Byron facility unless all safety requirements are satisfactorily met.
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In summary, the staff routinely reviews exceptions to Regulatory Guide positions to determine that the applicable regulations of the Commission are met by other means.
A second League concern relative to conformance to current regulatory practices is a concern as to whether the Byron Ouality Assurance /0uality control (0A/0C) programs are satisfactorily carried out in practice.
Issues related to the implementation of 0A/0C programs have been reviewed by the Office of Inspection and Enforcement (I&E). That Office has conducted an assessment of the Byron 0A/0C program relative to the concerns raised by the League. That assessment was provided in a memorandum to the Commissioners on March 9,1981 in response to a request by the Commission following its review of my partial denial on December 22, 1980 of the League's petition. Copies of this memorandum were served on counsel for the League and Commonwealth Edison. A copy of that memoraridum is attached to this denial.
In summary, the asstssment by the Office of Inspection and Enforcement indicates that, from mid-1979 to date, the Byron 0A/QC program generally has been effective and, in the staff's judgment, the information presented in the League's petition does not support the League's allegation that the Byron 0A/0C program was not effective.
The Commission during its review under 10 CFR 2.206 of my partial denial had noted the subsequent issuance by the Director, Region III, of an Immediate Action Letter confirming suspension of work on electrical cable installation. The assessment by I&E was that the limited scope of that problem made limited
stop-work the most appropriate technique to achieve corrective action and that there was no necessity for a total suspension of work at the Byron site. When the memorandum was forwarded to the Commission in March, the licensee had already met its connitments under the Immediate Action Letter and that constraint had been lifted.
In summary, the I&E staff is monitoring the implementation of the Byron QA/0C program for construction and is taking enforcement action appropriate to problems that occur. I hava no reason to believe that I&E will not continue to take actions appropriate to any future problems that may arise including suspension of construction, if warranted.
Open Generic Issues (Issue 6)
The League in its affidavit cites the foregoing five issues as "... example [s]
of a large number of such issues, all of which the NRC has identified (sometimes repeatedly) and a sizable portion of which NRC has repeatedly termed high-11/
priority matters.*
The League cites the November 1977 decision, Gulf States Utilities Co. (River Bend Station, Units 1 and 2, ALAB-444, 6 NRC 760, 774-75 (1977), by the NRC's Atomic Safety and Licensing Appeal Board as imposing an affirmative duty on the NRC staff to identify generic safety issues, and evaluate their impact on plant safety.
Although not noted by the League, the Atomic Safety and Licensing Appeal Board in Virginia Electric Power Company (North Anna Nuclear Power Station Unit 1), ALAB-491, 8 NRC 245, 247-50 (1978) later specified that staff conclusions are required with respect to unresolved generic issues for issuance of an operating license.
The staff routinely reviews each facility prior to the issuance of an operating license to evaluate IT/
-- Petition, page 64.
the safety significance of unresolved safety issues with respect to that facility.
These unresolved safety issues have already been determined generically to not pose imminent public health and safety concerns so as to prohibit continued operation or continuation of licensing actions.
In accordance with ALAB-491, the staff provides further explanation of the basis for licensing a particular facility in the absence of the long term generic resolution of these issues. The staff's findings are presented in the Safety Evaluation Report (SER) for each facility.
Such SER's have been developed for issuance of operating licenses to applicants for stations generally similar to Byron that use Westinghouse-designed nuclear reactors, e.g., North Anna Unit 2, Sequoyah Unit 1 and Salem Unit 2.
The League presents no arguments that Byron would be substantially different from facilities already reviewed. Therefore, this issue is one which may be left for review at the operating li ense stage.
Finally, the League also expressed concerns about " institutional disincentives to safety including a concern that assertion by the NRC staff of safety concerns, particularly those that may be controversial, is most unlikely to advance one's 12/
career and is far more likely to result in stigmatization and ' career paralysis.'"--
The Commission recognizes the potential problems in this area and has taken strong action to formulate a policy to preclude such institutional disincentives 13/
from affecting the NRC staff.--
One of the prime objectives of that policy is to provide sanctions against employees who take retaliatory actions with respect to differing professional opinions witnin the NRC staff.
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-- Petition, pages 64-65.
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-- NRC Manual Chapter 4125 " Differing Professional Opinions."
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Conclusion I have determined for the reasons set forth above that there exists no adequate basis for instituting a proceeding pursuant to 10 CFR 2.202 to inquire why the Byron Station, Units 1 and 2, construction permits should not be:
(a) Modified so that construction may not proceed without resolution of all outstandino safety problems applicable to the Byron Station;
'b) Suspended or revoked until s. Oh time as Commonwealth Edison has formulated an acceptable and realistic plan for resolving all outstanding safety problems; and (c) Revoked if the outstanding safety problems cannot be resolved prior to the completion of construction due to Edison's financial condition or for other reasons.
Accordingly, the Novembar 21, 1980 request of the Rockford League of Women Voters is denied.
A copy of this decision will be filed with the Secretary of the Commission for its review in accordance with 10 CFR 2.206(c) of the Comission's regulations.
In accordance with 10 CFR 2.206(c) of the Commission's Rules of Practice, this decision will constitute the final action of the Commission 25 days after the date of issuance, unless the Commission on its own motion institutes the review of this decision within that time.
AiWhl Har'old R. Denton, Director Office of Nuclear Reactor Regulation Dated agday of $7b ethesda, Maryland this g
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