ML19322E553
| ML19322E553 | |
| Person / Time | |
|---|---|
| Site: | Bailly |
| Issue date: | 02/20/1980 |
| From: | Scott W, Sekuler S ILLINOIS, STATE OF |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 8003280345 | |
| Download: ML19322E553 (23) | |
Text
e i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket No. 50-367
)
NORTHERN INDIANA PUBLIC ) (Construction Permit Extension)
SERVICE COMPANY )
)
(Bailly Generating Station, ) February 20, 1980 Nuclear 1) )
SUPPLEMENTAL PETITION OF THE STATE OF ILLINOIS The People of the State of Illinois (Petitioner) by WILLIAM J. SCOTT, Attorney General for the State of Illinois, in accord with the Board's Order of February 8, 1980 hereby submits its supplemental petition and contentions. These contentions are cumulative in nature and do not supplant any previous filings by the State of Illinois. The State of Illinois Request for Hearing filed March 5, 1979, the Joint Supplement to Requests for Hearing, filed June 29, 1979, the Amendment to Requests for Hearing, filed October 3, 1979, and the State of Illinois Petition for Leave to Intervene filed December 20, 1979 are all incorporated by reference herein, and constitute, in part, issues and contentions to be raised at the hearing to be held on this matter.
The brief period of time between notice of the pre-hearing l
conference and the date for filing this supplement did not allow for preparation of all contentions Petitioner believes to be relevant l
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4 v -r and material to this proceeding. Therefore in addition to incorpor-ating by reference herein all previous filings which may allege contentions or issues, Petitioner reserves the right to augment and amend its Contentions. As Petitioner has not had sufficient time to prepare memoranda of law in support of its contentions and addressing other issues pertinent to this proceeding, Petitioner requests the Board to provide either a briefing schedule or an
- opportunity for oral argument at the pre-hearing conference. ,
J CONTENTIONS
- 1. No Construction Permit Er. tension mr.ybe granted until an Environment-al Impact Statement for this Proceeding has been performed.
7 Issuance of a construction permit extension under 42 U.S.C.
S2235 constitutes a major federal action, which action will lead to
} events which will have a significant effect on the human environment surrounding the construction, therefore a new EIS is required by the National Environmental Policy Act (NEPA) 42 U.S.C. S4321 et seg.
The EIS which previously considered the Bailly site is insufficient for the construction permit extension because:
I (1) it does not consider the effects on the environment that will occur due to an extension of the construction period.
(2) it does not account for the effects of changes in design and construction which have made since 1974.
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(3) it does not consider the changes in cost /
benefit ratio which have come into effect because the Northern Indiana Public Service Company (NIPSCO) failed to meet the deadline for construction of the Bailly plant; Specifically an EIS must be written which will take into account the changes in economic and environmental costs due to the longer period of construction as balanced against the shorter period of energy production, decrease in energy demand, the construction of additional electrical plants during the past five years, including the Kankakee station, and other public interest factors which will be modified by the proposed extension.
(4) it does not apply current NRC criteria to the cost benefit and alternative analyses. For instance new regulations and policy on siting and emergency /
evacuation plans have not been considered. See NUREG 0625, Regulatory Guides 4, 7 and 1.97.
(5) it does not conform to the guidelines for EIS's adopted by the Council on Environmental Quality, in July, 1979.
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- 2. NIPSCO has failed to show that the various events which caused delay in completion of the'Bailly plant were not within the control of the permit holder [10 CFR 550.55 (b) ] . NIPSCO alleges completion of construction has been delayed beyond the latest completion date due to the delay of four months in issue of the permit; the stay issued by the United States Court of Appeals for the 7th Circuit on October 16, 1974 plus 2 months remobilization; the 2 months necessary to construct slurry wall which was not included in the original construction schedule; and the fact that construction has been " effectively stopped" since September 28, 1977 pending completion of NRC Staff's review of the new method devised for installation of the foundation for the facility. (NIPSCO letter of February 7, 1979.)
NIPSCO denies responsibility for or control over any of these events.
The delay of four months in issuance of permit was in no way sufficient to account for NIPSCO's inability to construct 99%
of its proposed plan by the final construction date and cannot therefore be considered "in large measure responsible" for the delay.
NIPSCO was responsible for the circumstances underlying the construction delay at least in the following respects:
7-1 f 5-A. The delay occasioned by the need to construct a a slurry wall was due to NIPSCO's inadequate assessment of the environmental effects of its construction plan. Because of NIPSCO's insufficient consideration at the design stage of the proper techniques to avoid consequences of dewatering, the slurry wall was not included in the original construct-ion schedule. The company later found it necessary to reconsider its original design and supplement it with a slurry wall. The delay incurred by the construct-ion of the slurry wall was therefore within the Appli-cant's control.
B. The delay in construction since September 28, 1977 was within the control of the Applicant. The administrative stay imposed by the NRC was made necessary because the company found, after construction began, that the original foundation design was not feasible. Thus the delay was clearly due to inadequate research and design of the foundation by NIPSCO and its contractors.
The following factors, over which NIPSCO exercised control should be considered in rendering a decision!
(i) the stratigraphic analysis prepared for NIPSCO was deficient in that it did not correctly identify the various geological strata through which the proposed pilings would have to penetrate.
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(ii) NIPSCO and its contractors failed to perform adequate testing, by borings or other means, prior to construction to identify the problems at the pre-construction stage including the inability to drive the piles to bedrock.
(iii) NIPSCO and its contractors did not devise a proper construction program which would have made it possible to utilize the original design.
(iv) The administrative stay was a result of the NIPSCO's own departure from its original foundation design and was therefore in control of NIPSCO. The new "short pilings" design devised for NIPSCO by Sargent & Lundy and Dames & Moore required thorough analysis by the NRC. The events which necessitated the adoption of a new foundation design in themselves indicated to the agency that a rigorous analysis of the untried technology being proposed in this novel approach to the construction of a nuclear plant was necessary. The onus of responsibility for the delay incurred by a thorough analysis of the short pilings proposal is not on the reviewing agency, but on NIFSCO.
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- 3. Reasons for delay alone do not constitute sufficient " good i
cause" to allow a Board to extend the construction permit.
To fully meet its burden to show good cause NIPSCO must show first that it was not in large measure responsible for the circumstances leading to the delay, 10 C.F.R.
S50.55;and second that the reasons assigned by NIPSCO for the extension do not give rise to health and safety or environmental issues, litigation of which is necessary to protect the interes':s of intervenors or public interest.
NIPSCO has not shown that the extension of the construction permit will not endanger the public health, safety or environment.
The statement in NIPSCO's February 7, 1979 letter to Harold R. Denton, Director, Office of Nuclear Reactor Regulation of the NRC that the amendment to the construction permit has no safety or environmental significant is incorrect. Extension of the construction permit will lead to considerable environmental damage due to the increased period of construction during and after which the dewatering will continue.
Contrary to early estimates that dewatering would continue for a maximum of 24 months during construction, it is now estimated that extended construction at the Bailly site will necessitate dewatering during the entire period of construction, and for several months
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afterward. (Experience at River Bend and at r orso, Italy, substan-tiates this fact.) Dewatering for the period estimated will cause irreparable injury.
A. The additional length of construction time will result in a longer period of construction site dewatering which will adversely affect water flow patterns in the Indiana Dunes National Lakeshore (IDNL) by altering the water levels, and water characteristics in such a manner as to cause the elimination of those species of flora which presently exist in the wet land system, including the " Bog Indicator" species which contribute to the significance of the Indiana Dunes. The existence of such species i
are directly dependent on the aquatic environnent.
The additional period of construction time is sufficient to cause some of the rare species, particularly the
" Bog indicator" plants to disappear from the dunes ecosystem because of the change in the aquatic environment over the longer period of construction time. The construction site dewatering results in a a cone of depression in the water table and causes the ground water to flow toward the construction site excavation, a condition that exists as long as ground water is removed from the excavation. The result is a dryir.g effect in the wet land system l
around the site ano extending into the wet lands of
the Indiana Dunes National Lakeshore for distances of over a mile. A deficiency of water, a surplus in water characteristics, an increase in the rate or direction of water flow, dilution of the existing water, and/or replacement of the water which differs in characteristics from the water at any specific location within the Indiana Dunes National Lakeshore all will result in changing or otherwise adversely affecting the Indiana Dunes National Lakeshore.
B. Recent proposals to reintroduce replacement water from ,
alternative sources will not prevent the adverse impact on the Indiana Dunes National Lakeshore because:
- 1. The point (or points) of introduction of replacement water have not been shown to be capable of maintain-ing natural water levels within all areas of the Indiana Dunes National Lakeshore while maintaining dry excavation to tae depth needed for the additional period of construction time.
- 2. The replacement water levels are not keyed to the natural water table levels, or the natural inter-dunal pond and wet land levels, includir g measures to maintain the natural seasonal variations and yearly variations, within all areas of the Indiana Dunes National Lakeshore. Variations of water l l
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levels must also be timed to coincide with the timing of natural seasonal cycles, a condition which is not considered in the proposed water replacement program and which will be exacerbated by the longer construction period.
- 3. The replacement water will not have the same characteristics as the water removed. The natural water has a low flow rate and a long term turnover, during which time the water is conditioned by natural processes, such as organic decay. Further, the water characteristics vary considerably from one location to another, and from strata to strata, or from near surface to greater depths. The water replacement program does not match these characteristics with specific locality. -
C. The removal and replacement of ground water whatever the type or source of water, will result in an increased rate of ground water movement, whicn will dilute and replace the existing water without permitting the normal, slow, natural conditioning from organic process.
The additional period of tTme will permit a greater dilution and/or replacement of the natural water.
D. Recent studies by the U.S.G.S. has indicated that an underlying strata which previously was assumed to act as a barrier to limit the effect of site dewatering, i
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diminishes in thickness to the point of disappearance, or at least to the point of being ineffectual. The result is that construction site dewatering will drain an aquifer not previously considered by the ASLB and which has direct connection with the wet lands of the Indiana Dunes National Lakeshore considerably farther to the east than previously assumed including the waters of Cowles mog. The additional period of construction time will result in a longer period in which such waters are subjected to the artificially induced alterations which will result in an immediate and severe adverse impact upon such areas of the National Lakeshore.
E. Dewatering can alter subsoil structure in such a way that it can change the reaction to seismic occurrences.
Additionally the load bearing capacity of the plant and its foundation will be affected as will the ef fects of a core melt in a situation of altered water tables.
These changes must be taken into consideration as an effect of the construction permit extension on the Safety Evaluation of the plant.
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- 4. The documentation for the proposed construction permit extension is insufficient in that there is no analysis of the risks and consequences of Class 9 accidents which could occur at the Bailly plant. Class 9 accidents should be analyzed as part of the necessary Environmental Impact Statement (See Contention
- 1) and as part of the health and safety evaluation of the Staff (SER) and Board.
It is now acknowledged by the NRC that class 9 accidents can occur. (NRC Staff Response to Board Question No. 4 Regarding the Occurrence of a Class 9 Accident at Three Mile Island, In the Matter of Public Service Electric and Gas Company (Salem Nuclear Generating Station, Unit 1) Docket No. 50-272, August 24, 1979.) As the possibility of such an occurrence must be calculated, as must be probable consequences of various scopes, the NRC has recognized that Class 9 accident analysis is appropriate (Proposed Annex to Appendix D, Part 50 36 Fed Reg 22851, now denominated as NRC Regulatory Guide A.2; " Preparation of Environmental Reports for Nuclear Power Stations", NUREG 0099, July 1976). The Council on Environmental Quality (CEQ) is currently considering the amendment of its guide-lines to require consideration of Class 9 accidents in EIS preparation No construction permit extension should be granted to NIPSCO until adequate consideration of Class 9 accidents has been made in the EIS and the SER to be prepared by the NRC staff.
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- 5. NIPSCO's inability to control the design and construction of the Bailly plan draws into question its ability to construct a nuclear plant which will meet specifications and federal regulations. The NRC Special Inquiry Group Report (Rogovin Report) recommended greater consideration in NRC construction permit and licensing proceedings of whether a utility.was technically competent to construct and operate a nuclear reactor (Three Mile Island, Volume I, Report to the Commissioners and to the Public, NRC Special Inquiry Group, Mitchell, Rogovin, Director, Recommendation No. 7, January, 1980; See also Id. Recommendation No. 3). In determining whether good cau'se exists to issue a construction permit extension the Board must consider whether, given its prior history, NIPSCO r can competently build and operate a safe reactor. The Board must also consider whether once permitted the utility and its contract-ors can construct the plant " consistently with the design plans" (See Rogovin at 141). Unless NIPSCO and its contractors prove that they are technically competent the Board cannot find good cause to issue a construction permit extension.
- 6. The application for Construction Permit Extension is inadequate-in that there is no updated discussion of the Bailly site.
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Among other siting questions such as proximity of the proposed plant to centers of large population, and effects of the construction on the ecology of the site, the Board should require a showing that the now acknowledged " unfavorable physical characteristics of the site" will be compensated for by the
" appropriate and adequate compensating engineering safeguards" 10 CFR S100.10 (d) . This is especially important as part of the showing of " good cause". Given NIPSCO's prior inability to adequately design engineering techniques that (1) compensate for inadequacies of the site such as the difficult geologic configu-rations and (2) do not create additional hazards or problems on the site e.g. dewatering complications, and soil changes attribut-able to pile driving. No construction permit extension should be issued by the Board until NIPSCO shows that that siting of the Bailly reactor will be in accord with 10 C.F.R. Part 100, Regulatory Guide 4.7, NUREG 0625, NUREG 0499 and any other NRC policies and regulations that control siting criteria.
- 7. The NIPSCO application is inadequate in that it fails to address or consider health and safety issues that affect the interests of the petitioners and the public. Several of the causes or the delay in the completion of construction in and of themselves seriously cast doubt on NIPSCO's ability to construct a safe facility. Before it can legitimately find " good cause" for an
extension of i.he construction permit the Board must review all such issues. In the Matter of Indiana and Michigan Electric Company and Indiana and Michigan Power Company, (Donald C. Cook Nuclear Plant) , ALAB-129 (AEC 414, 420 (1973). Issues to be considered include:
A. The modification in foundation design.
How will short pilings which will not be secured
. in bedrock alter the plant's tolerance of design basis seismic events; how will the change lead to differential settlement between class I and non-class I structures and how will this differential movement affect the various interconnected systems of the reactor; how will the change alter load bearing capacity; what other changes will occur due to the modifications in the foundation design?
B. The changes in soil structure which have occurred due to the tests and borings which have been conducted in relation to the proposed modification in the pilings desigz. (e.g. liquifaction) and the effects thereof.
C. Mark II containment. This type of containment has been listed by the ACRS as one of its Category A activities for resolution (See Joint Petition for Hearing, Paragraph 3).
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The containment has been under review for the past two years and at present the Bailly Primary Contain-ment System does not adequately meet the requirements of 10 C.F.R. Part 50, App. A. Criteria 4, 16, 50 and
- 51. NIPSCO's application gives no assurance that the use of the containment will be safe and in the public interest.
D. The Construction permit for Bailly was granted in 1974.
In 1980 the plant is le"s than 1% complete. Assuming construction was completed by the 1987 date requested by NIPSCO in the August 31, 1979 letter,the plant design,other than the foundation will be more than 15 years old when the NRC begins the operating license review. The design would probably deviate considerably from the then-current Standard Review
, Plan's acceptance criteria, making the staff's review more difficult and more time-consuming.
For example, the ASME Code to which the reactor vessel was purchased would have changed considerably.
The design of the Bailly Nuclear Plant, being at least 10 years old at this time, and facing several years of delay including additional years of construction, is being outdistanced by a rapidly changing technical 4
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and licensing environment. Therefore, it is important that the design be re-reviewed in such areas as the ability to meet current regulatory guides, the pressure vessel code criteria, the accident analyses in the PSAR, the resolution of safety issues applicable to the plant and implementation of the numerous recommendations resulting from TMI. The result should be a revised SER and a " documentation of deviations." In keeping with the River Bend decision and because of the extensive delays experienced and forecast in the Bailly construct-ion, this re-review should be completed prior to recommencing construction.
- 8. NIPSCO's application for a construction permit extension is inadequate in that it does not address known safety problems that relate to tr.e design, construction and operation of the Bailly Nuclear plant. The recent events at Three Mile Island have demonstrated that unforseen and highly inprobable accidents can occur. Post accident review by the NRC and independent groups l
have revealed that many heretofore accepted licensing and safety evaluation practices have not been sufficient to guarantee that a nuclear plant will be safely constructed. To give the public l
greater assurance that the Bailly plant will be sefe, if construct-l i
ed, this Board must consider (1) those features of the planned
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construction which could be related to a TMI type of accident and (2) generic safety problems which may lead to future accidental occurences of any sort. Although it might not be possible to analyze every credible accident scenario caused by "a sequence of successive failures more severe than those postulated on the design basis of the plant", (Mattson, supra in re class 9 accidents), the Board can at least rule on whether NIPSCO has adequately considered potential risks and consequences based on present knowledge of the hazards of certain aspects of the reactor design. The time for review is now.
" Licensing procedures should foster early and meaningful resolution of safety issues before major financial commitments in construction can occur." (emphasis added) Report of the Kemeny Commission, Recommendation A-10, at 65.
In reviewing potential safety problems the Board should take into consideration current regulatory guides, ACRS recommendations, the Kemeny and Rogovin reports and other recent documentation in regard to safety problems. The safety review should include issuance by the NRC of an updated Safety Evaluation Report on the construction permit extension.
The Bailly PSAR should be updated to include analysis of an accident such as that which occurred at Threo Mile Island in terms of the consequences.of such an accident. This analysis should consider at least:
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- 1. Generic unresolved technical issues which apply to the Bailly BWR. Several of these issues listed in the 1977 ACRS list of 28 unresolved generic issues include non-random multiple failures, behavior of reactor fuels under abnormal conditions, and stress corrosion cracking in BWR piping (See NUREG 0531).
- 2. Safety issues described in NUREG 0510, issued January 1979. Among the list of " Category A technical Ut tivities requiring prompt resolution" identified by the NRC in NUREG 0510 are at least four that relate to Bailly: Item A-8-Mark II containment; Item A-9-ATWS; Item A-10-BWR nozzle cracks; and Item A-34-Post accident monitoring design (See also Reg. Guide 1.97).
- 3. Safety issues resulting from the TMI-2 Accident.
Several reports on the TMI-2 accident have made substantive recommendations for improvement of reactor safety; some of which would be pertinent.
The Lessons learned Task Force generated a list of 23 recommended areas of improvement which have been listed in NUREG 0560. Several of the recommend-i ations apply to the Bailly construction such as: l l
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the need.for positive and direct indication of cri-tical valves; improved simulators and expanded training; improved design features to aid in operator response; improved general design criteria (GDC) for anticipated transients; revised standard review regarding transients and improved lay out of control rooms.
CONCLUSION The State of Illinois has not specifically addressed legal issues in this filing, however we believe each of the contentions listed herein meets the standard of 10 C.F.R. S' . 714 (b) , and that each presents a subject appropriate to be litigau at a construction permit extension hearing. The State of Illinois urges the Board, following the March 12, 1980 prehearing conference, to enter an order establishing (1) a briefing schedule for any legal issues still unresolved at that time; (2) a time period within which the parties should be required to meet informally and attempt to reach agreement 4
on contentions to be litigated in this proceeding; (3) a time period for the establishing of any formal discovery needing to be done; (4) a date for a second prehearing conference to resolve any differences among the parties concerning contentions to be litigated, or other matters, and to establish a hearing schedule.
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DATED: February 26, 1980 Respectfully submitted, WILLIAM J. SCOTT Attorney General State of Illinois BY:
SUSAN N. SEKULER Assistant Attorney General Environmental Control Division 188 West Randolph, Suite 2315 Chicago, Illinois 60601 (312) 793-2491 OF COUNSEL:
GEORGE WM. WOLFF Chief, Environmental Control Division JOHN VAN VRANKEN Chief, Northern Region Environmental Control Division DEAN HANSELL Assistant Attorney General Environmental Control Division 188 West Randolph, Suite 2315 Chicago, Illinois 60601 (312) 793-2491 9
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CERTIFICATE OF SERVICE I hereby certify.that on this 26th day of February, 1980 I served copies of the foregoing Supplemental Petition dated February 26, 1980, upon each of the persons named on the attached Service List, by causing copies to be deposited in the U.S. Mail, in envelopes properly addressed and sealed, first class postage prepaid.
SUSAN N. SEKULER 1
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SERVICE LIST l Herbert Grossman, Esq., Chairman George & Anna Grabowski ,
Atomic Safety & Licensing Board 7413 W. 136th Lane '
Panel Cedar Lake, Indiana 46303 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. George Schultz 110 California Street Dr. Richard F. Cole Michigan City, Indiana 46360 Atonic Safety and Licensing Board Panel Richard L. Robbins, Esq.
U.S. Nuclear Regulatory Commission Lake Michigan Federation Washington, D.C. 20555 53 W. Jackson Blvd.
4 Chicago, Illinois 60604 Mr. Glenn O. Bright Atomic Safety & Licensing Board Panel Mr. Mike Olszanski U.S. Nuclear Regulatory Commission Mr. Clifford Mezo Washington, D.C. 20555 United Steelworkers of America 3703 Euclid Avenue Maurice Axelrad, Esq. East Chicago, Indiana 46312 Kathleen H. Shea, Esq.
Lowenstein, Newman, Reis, Steven C. Goldberg, Esq.
Axelrad and Toll Office of the Executive Legal 1025 Connecticut Ave., N.W. Director Washington, D.C. 20036 U.S. Nuclear Regulatory Com.
Washington, D.C. 20555 1
William H. Eichhorn, Esq.
Eichhorn, Eichhorn & Link Stephen Laudig, Esq.
5243 Hohman Avenue 445 N. Pennsylvania Ave.,
Hammond, Indiana 46320 Indianapolis, Indiana 46204 Diane B. Cohn,Esq. Docketing & Service Section William P. Schultz, Esq. Office of the Secretary Suite 700 U.S. Nuclear Regulatory 2000 P Street, N.W. Commission Washington, D.C. 20555 Washington, D.C. 20555' Atomic Safety and Licensing Robert J. Vollen Board Panel James Geocaris U.S. Nuclear Regulatory Commission 109 North Dearborn Washington, D.C. 20555 S-1300 Chicaco, Illinois 60602
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Atomic Safety and Licensing Appeal Board Panel Edward W. Ossan, Jr.
U.S. Nuclear Regulatory Commission One IBM Plaza Washington, D.C. 20555 Suite 4600 Chicago, Illinois 60611 Robert Graham One IBM Plaza 44th Floor Chicago, Illinois 60611
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