# Appendix — Porter County Chapter of the Izaak Walton League of America, Inc. v. Nuclear Regulatory Commission > Briefs, arguments, decisions, and more. URL: https://www.frixlaw.com/law-library/documents/brief%3Amicro_IA40385004_0520%3A2 ## Record - **Collection:** Supreme Court brief - **Document type:** Appendix - **Published:** January 1, 1976 - **Citation:** 429 U.S. 945 ## Text FILEQOD APPENDIX AUG 27 1976 J meme. Rovax,se_cuERK In Tue Supreme Court of the Anited States Ocroser Term, 1976 No. 7676-291 PORTER COUNTY CHAPTER OF THE IZAAK WALTON LEAGUE OF AMERICA, INC., et al.; and THE PEOPLE OF THE STATE OF ILLINOIS, Ex Rel. WILLIAM J. SCOTT, Attorney General of the State of Illinois; and THE CITY OF GARY, INDIANA, Petitioners, Vs. THE NUCLEAR REGULATORY COMMISSION, THE UNITED STATES OF AMERICA, et al, Respondents. On Petition for a Writ of Certiorari to the United States Court of Appeals for the Seventh Circuit Printed by Authority of the State of Illinois SS ee TABLE OF CONTENTS Judgment of the Court of Appeals, April 15, 1976 Order Upon Denial of Petitions for Rehearing and Sug- gestions for Rehearing En Banc, June 1, 1976 —.......... 14 Judgment of the Supreme Court of the United States, November 11, 1975 —...........- Se ate TES EN 15 Judgment of the Court of Appeals, April 1, 1975 __..... 22 Order Upon Denial of Petitions for Rehearing, May 28, TID: -scasscinicideniietbiiisinsiindslthsteasateinenmvemansnenniannenstansietmetennens 62 Order Upon Motion for Stay Pending Judicial Review, I taal caalinas 63 Initial Decision of the Atomie Safety and Licensing ae semmmacsnmensuanees 65 Decision of the Atomic Safety and Licensing Appeal ee 6 Se 187 In Tue Supreme Court of the United States Ocroser Term, 1976 No. 76- PORTER COUNTY CHAPTER OF THE IZAAK WALTON LEAGUE OF AMERICA, INC., et al.; and THE PEOPLE OF THE STATE OF ILLINOIS, Ex Rel. WILLIAM J. SCOTT, Attorney General of the State of Illinois; and THE CITY OF GARY, INDIANA, Petitioners, vs. THE NUCLEAR REGULATORY COMMISSION, THE UNITED STATES OF AMERICA, et al, Respondents. On Petition jor a Writ of Certiorari to the United States Court of Appeals for the Seventh Circuit = Per Curiam Opinion On Remand from the Supreme Court UNITED STATES COURT OF APPEALS For the Seventh Circuit April 13, 1976. Before Hon. Tuomas E. Fairncump, Chief Judge Hon. Rosert A. Sprecner, Circuit Judge Hon. Pump W. Tone, Circuit Judge No. 74-1751 PORTER COUNTY CHAPTER OF THE IZAAK WALTON LEAGUE OF AMERICA, INC.; CONCERNED CITIZENS AGAINST BAILLY NUCLEAR SITE; BUSINESSMEN FOR THE PUBLIC INTEREST, INC.; JAMES E. NEWMAN, MILDRED WARNER and GEORGE HANKS, Petiti THE PEOPLE OF THE STATE OF ILLINOIS ex rel. WILLIAM J. SCOTT, Attorney General of the State of Illinois, Petitioners-I ntervenors, vs. THE ATOMIC ENERGY COMMISSION and THE UNITED STATES OF AMERICA, Respondents, and NORTHERN INDIANA PUBLIC SERVICE COMPANY, Petitioners-Intervenors. Petition for Review of Final Orders of the Atomic Energy Commission This cause came on to be heard on the transcript of the record from the Atomic Energy Commission, and was ar- gued by counsel. On consideration whereof, it is ordered and adjudged by this Court that the decision of the Atomic Energy Commis- sion entered in this cause on August 29, 1974, be, and the same is hereby Enforced, in accordance with the opinion of this Court filed this date. aufjun 3n Ginited States Court of Appeals For the Seventh Circuit No. 74-1751 Porter Cousxty Cuaprer Or Tae Izaak Warton Leacuve Or America, Inc.; Concernep Citizens Acatnst BalLiy Nectear Srre: Bustvessmexs For Tar Pvsuic I STeResT, Ixc.; James E. Newman; Mirprep Warner axnp GEORGE Hanks, : Petitioners, and Tre Peorte Or Tre State Or Icio ex rel. WituiaM J. Scorr, Attorney General of the State of Illinois, Petitioners-Intervenors, v. Tre Atomic Exercy Com™isston axyp Tar Usrrep States Or AMERICA, Respondents, and Nortrers Ixpiaxa Pratic Service Compayy, Petitioners-Intervenors, Petition for Review of Final Orders of the Atomic Fnergy Commission Ox Remaxp From Tue Supreme Covert Decrrep Apr 13, 1976 Before: Farrcump, Chief Judge, Sprecner and Tove, Circuit Judges. = Per Curiam. This proceeding to review an Atomic En- ergy Commission order granting a permit to construct a nuclear power plant was remanded to us for further pro- ceedings in Northern Indiana Public Service Co. v. Porter County Chapter of the Izaak Walton League of America, 423 US. ......., 46 L. Ed. 2d 156 (1975). Our jurisdiction is derived from 42 U.S.C. § 2239(b) and 28 U.S.C. § 2342(4). We hold that the ageney’s order is valid and deny the petition for review. The history of these proceedings, which is recounted in more detail in this court's earher decision, Porter County Chapter of the Izaak Walton Leaque of America vy. Atomtc Energy Commission, 515 F. 2d 513 (7th Cir. 1975), may be summarized as follows: In 1970 Northern Indiana Pub- lie Service Company (NIPSCO) filed with AEC an ap- plication for a construction permit and operating license for a nuclear facility to be constructed on the company’s Bailly site, on the southern shore of Lake Michigan, in Porter County, Indiana. The application was opposed by various organizations and individuals, petitioners herein, who were permitted to intervene. Hearings were held by the Commission’s Atomie Safety and Licensing Board (ASLB) in the fail of 1972 and throughout 1973. ASLB initially authorized issuance of the construction permit on April 5, 1974. RAT-74-4, 557. The decision was appealed hy petitioners to the Commission’s Atomic Safety and Licensing Appeal Board (ASLAB), which affirmed. RAI- 74-8, 244 (August 29, 1974). That decision subsequently became the final decision of AEC. See 10 C.F.R. 44 2.770, 2.785(a). After the filing of the petition for review in this court, the State of Illinois and the City of Gary, Indiana, were permitted to intervene on the side of petitioners. By a divided panel, this court set aside the order on the ground that issuance of the permit was in violation of an AEC regulation setting a minimum distance between the site for a nuclear facility and the nearest boundary of a densely populated center of over 25,000 residents. 515 M24 513 (1975). On writ of certiorari, the Supreme Court reversed the judgment and remanded the case for con- sideration of arguments in opposition to the permit not decided in our previous opinion, 423 U.S. ........ , 46 L. Ed. 2d 156 (1975). —_ I. The argument we address first is that, despite Con- gress’ grant of exclusive licensing authority to AEC,’ that agency lacks jurisdiction to approve the site, because the plant will encroach on adjacent federal lands administered by the Department of Interior. These lands, consisting of some 8300 acres known as the Indiana Dunes National Lakeshore, abut the eastern boundary of the 107-acre tract on which NIPSCO proposes to build the nuclear plant. See 515 BF. 2d at 517-518. The Department’s environmental concerns with the Bailly site were noted in our earlier opinion. 515 F. 2d at 526-527. Relying on United States v. Alford, 274 U.S. 264 (1927), Camfield v. United States, 167 U.S. 518 (1897), United States v. Cappaert, 508 F. 2d 315 (9th Cir. 1974), cert. granted, 95 S.Ct. 2654 (1975), and analogous cases in which the United States has been held to be entitled to equitable relief against uses of privately owned land which detrimentally affect adjoining govern- ment property, petitioners contend that AEC approval of of the site is barred. The Department of Interior has not advanced such a contention. Most of the cases relied upon by petitioners are readily distinguishable because they involve relief granted pur- suant to specific statutory prohibitions which are not ap- plicable here. But even assuming the Department of In- terior, by exercising its statutory power of supervision over the National Lakeshore, cf. United States v. Cap- paert, supra, could obtain injunctive relief against threat- ened irreparable injury to lands within its custody from sources outside, we could not justify interruption of the AEC licensing process when the Department has not sought such relief. While AEC’s authority to issue licenses and construction permits should be reconciled to the fullest extent possible with the interest of the Department of In- 1The Atomic —, AE Act of -_ w amended and in force at the time of the ‘agency 2 USC. §§2011-2282 authorized the Commission tq. issue oy of A the construction and utilization of an device which™ a a oe Ge a nuclear fuels. 42 U.S.C. §§ 2133, 2235. With the exception of certain activities carried on at the direction of the President or under AEC supervision, the Act made it unlawful for anyone to use oe Ube § nuclear materials without obtaining a license issued by AEC. 42 eae As noted in this court’s — a 515 bg 5 at 516 has recently been further with the Nuclear Regulatory Commission. 42 U.S.C. §§ —5— terior in protecting the National Lakeshore, nothing in the authorizing legislation for either agency suggests that the Department’s views on the advisability of an adjoir- ing nuclear facility are to be controlling. AEC has con- sidered the Department’s comments and has given atten- tion to the environmental effects of the Bailly plant upon the National Lakeshore, as the law requires. It has con- eluded that with the restrictions and controls it will im- pose as conditions to approval of the construction per- mit, including two monitoring programs which will detect changes in chemical and water level composition of the dunes area, the environmental impact of construction and operation of the nuclear plant will not be substantial’ enough to require disapproval of the site. This was a determination the ageney had authority to make. In a related argument intervenor State of Illinois con- tends that extension of the exclusion area and low-popula- tion zone, see 515 F. 2d at 520, surrounding the site into the National Lakeshore “is incompatible with Interior’s mandate to ‘preserve ... the Indiana dunes’... .” The exclusion area, however, will not extend into the Lakeshore area. As for the boundaries of the low-population zone, they are simply one factor to be considered in evaluating various sites, as noted by ASLAB in its final decision: “[T jhe creation of a ‘low population zone’ does not impose — and is not meant to impose — restrictions on the use of land within the perimeters of the zone. Concern that the Bailly low population zone might im- pinge on the Interior Department’s management pre- rogatives in the Indiana Dunes National Lakeshore is thus a non sequitur.”” RAT-748, 261. The jurisdictional arguments are therefore without merit. Il. In view of the Supreme Court’s decision, little is left of petitioner’s related argument that the order is not in 2In addition to ecological factors, both ASLB (RAI-7 600-601 and ASLAB (RAI-74-8, 258-259, 261, ay jconsidered_ visual the National Lakeshore, but intrusion” a outweighed by the necessity of ee -~q 4- reas ~~ compliance with AEC’s siting regulations, which provide, see 515 F. 2d 519-520, for three population buffer zones around the nuclear facility. 10 C.F.R. Part 100. Petition- ers’ remaining contention relating to the siting regula- tions is that the evidence is insufficient to support the find- ing that “the distance from the reactor to the nearest boundary of a densely populated center containing more than ahont 25,000 residents,” 100 C.F.R. § 100.3(e), is two miles. Once it has been determined, as the Supreme Court did, that political. boundaries are not controlling, the record and pertinent census data adequately support the ageney’s finding.’ Intervenor State of Illinois argues in this connection that AI°C’s issuance of the construction permit to NIPSCO is inconsistent with the agency’s recent rejection of another utility’s application for a permit at a site having a lower siting population factor than Bailly, on Newbold Island, New Jersey. So long as AEC’s regulations are complied with, as they are here, population density is only one factor in site evaluation. The record of the New Jersey case is not before us, and an intelligent comparison could not be made without the full records for both applications. Cf. American Meat Institute v. Environmental Protection Agency, 526 I. 2d 442, 466 (7th Cir. 1975). IIT. Petitioners also argue that, without regard to the dis- tance between the proposed reactor and the boundary of the nearest densely populated center with over 25,000 resi- dents, AEC failed to give sufficient weight to the density of population surrounding the Bailly site. Specifically, * The agency has also given consideration to the persons who, although not residents, may be near the site for other reasons. These include employees of the nearby plant of Bethlehem Steel Corporation, which was a party before the agency, and transient visitors to the National Lakeshore area near the reactor site. AEC has determined that radiation realistic possibility, which may be suffered by these employees and visitors and the estimated 7.3 million who live within a 50-mile radius of the site will be insignificant. points out that the possibility ere accident is so remote that nuclear plants t the country have been located on other sites ha low-population zones include federal lands. It also notes that the t of ——- == petitioners seek to apply an AEC staff working paper, dated April 17, 1973, and released April 9, 1974, shortly after ASLB’s initial decision was issued, and the Nuclear Regulatory Commission’s' Regulatory Guide 4.7 (revision 1), dated November 1975. Petitioners stress the following passage, which was added in the final version of the Regu- latory Guide: “If the population density, including weighted tran- sient population, projected at the time of initial opera- tion of a nuclear power station exceds 500 persons per square mile averaged over any radial distance up to 30 miles . . . special attention should be given to the consideration of alternative sites with the lower population densities.” Petitioners calculate there to be 806 persons per square mile within a 30-mile radius of the Bailly site. The working paper and a draft version of the Regula- tory Guide, which is similar to the final version now be- fore us, were said in our prior opinion to be relevant in determining “whether, on balance, the extraordinary re- sult of halting or delaying the development of needed en- ergy is warranted by” what the majority then perceived to be a violation of the AEC’s own regulations. 515 F. 2d at 528. The court did not, however, hold that AEC’s failure to apply the working paper and Regulatory Guide criteria was an independent ground for reaching that result. Unlike regulations, both the working paper and the Regulatory Guide are advisory rather than obligatory.° The working paper by its terms applies to applications docketed after January 1, 1974 (NIPSCO’s application was filed in 1970).° The Regulatory Guide, having been issued after the entry of the order under review, is like- *See note 1, supra. 5The working paper, in the words of ASLAB, “neither represents nor purports to present Atomic Commission policy respecting nuclear power plant sites.” RAI-74-8, 255. Similarly, the Guide’s title page notes that tory Guides are 3 for regulations, and compliance with them is not required. ® Because the working paper was inapplicable by because it ultimately became available to petitioners, even if it within the scope of petitioners’ discovery request, —3— wise of doubtful applicability. Cf. Douglas, J., concurring, 423 US. at ........ , 46 L. Ed. 2d at 157. Even if these docu- ments were applicable, they would not require setting aside the agency order. Both documents stress the importance of considering less-densely populated sites and the use of additional safety measures, where appropriate, if alterna- tive siting is not available. We conclude below that proper consideration was given to other sites, and the safety measures taken here are not the subject of attack. IV. Finally, petitioners argue that AEC has failed in several respects to comply with the requirements of the National Environmental Policy Act (NEPA), 42 U.S.C. §§ 4321-4347. Adopted in late 1969, NEPA requires, in section 102(2)(C), 42 U.S.C. § 4332(2)(C), a detailed statement of environmen- tal impact to be included in all federal agency proposals “significantly affecting the quality of the human environ- ment.” AEC regulations were amended in 1971 to provide a procedure for insuring compliance with NEPA. At the time relevant to these proceedings, the regulations required all applicants for nuclear plants to submit a detailed environ- mental report for AEC staff anaylsis. 50 C.F.R. Part 50, App. D, A(1) (1974).". NIPSCO filed its original report in 1971 and twice amended it in 1972. The staff then drafted its own preliminary statement of environmental considera- tions, which included an assessment of adverse enviroa- mental effects, alternatives to the proposed action, antici- pated costs in relation to net benefits, and various other environmental matters. The preliminary statement was then transmitted to interested parties,® for comment, after which the staff completed a final environmental impact 750 CFR. Part 50, D was replaced by 530 CFR. Part 51 in 1974. See 39 F.R. 26279 (July 18, 1974). proximity to the Bailly site, intervenor [Illinois alleges that it was an “affected State” under 50 C.F.R. Part 50, App. D, A(6), was therefore entitled to be sent a of the preliminary ~~ byte 4, ST K- or agencies, however including governor, the Illinois Commission on Atomic Energy, and the Environmental Protection Division of the Attorney General's —- oo a dutig LBP-74-19, supra at 561, 563 (Findings 14, ); (2) Low Population Zone: 2413 meters or 1.5 miles. Id. at 561, 563, 565 (Findings 14, 19, 24); (3) Population Center Distance: 3219 meters or 2 miles. /d. at 565 (Finding 24). As stated earlier, Findings (1) and (2) are based on a dosage calculations but (3) is simply 1% times (2). In accordance with these findings, the application of the 10 C.F.R. 100 regulations would require that no “densely populated center containing more than about 25.000 #esidents” be located at or nearer than two miles from the reactor. The City of Portage, Indiana, had a population of 19,- 127 in 1970 but the ASLB found that “f{u]pon question- fit ing concerning the City of Portage, the Staff acknowledged the expected population of Portage by 1980 would exceed 25,000.” NJPSCO, LBP-74-19, supra at 565 (Finding 24). Without so finding, ASLB referred to evidence that “the nearest boundary [of Portage] was 1.1 miles from the reactor site.” /d. The briefs of both the government and NIPSCO accept the fact that the political or municipal or corporate boundary of Portage is approximately one mile from the reactor.® Inasmuch as the site evaluation factors are promulgated to insure the safety of the adjoining pop- ulation,’ “[a]t the very least, consideration should. . . be given to the best available estimates as to what will be the population in the general vicinity of the reactor dur- ing the early vears of actual operation.”* The evidence * In discussing Joint Intervenor’s contentions, the govern- ment states that “the political boundary of the City of Portage . . . Was only one mile away,” (Br. at 35} and NIPSCO states that “[Portage’s] nearest municipal boundary is within 1.1 miles of the site.” (Br. at 24). * 10 C.F.R. § 100.10 provides that “the site location and the engineered features included as safeguards against the haz- ardous consequences of an accident, should one occur, shouid insure a low risk of public exposure.” * Southern California Edison Co.. ALAB-248, RAI-74-12, 957, 960 (Dec. 24, 1974), where the ASLAB also said in ~ ey to the San Onofre Nuclear Generating Station, Units 2 and 3: Part 100 does not specify the precise time at which the number of persons within a potential population center is to be measured. But, the purpose of the Part 100 criteria being to ascertain whether a particular site is suitable for reactor operation, it would make little sense to look only at the size of the population as of the time of the construction permit proceeding (several years prior to reactor opera- tion). [T]here is a strong likelihood that, sometime during the first few years of reactor operation (if not before), San Clemente will have a population of “more than about 25,- 000.” Therefore, we see no reason why San Clemente should not be treated as the closest population center, and we accordingly do so. Id. at 959-60. See also Long Island Lighting Co. (Shoreham ge A Power Station), ALAB-156, RAI-73-10, 831, 848 (Oct. ; 3) —s— here is uncontroverted that the population of Portage will grow to 25,000 between 1976 and 1977 and to about 29,000 in 1980, the early years of expected operation of this reactor. The ASLB and eventually the ASLAB accepted the Regulatory Staffs argument that the political boundary of Portage was to be ignored and instead some amorphous, flexible and movable center or centroid of population was to be considered. The fatal error in this alleged logic is that AEC itself has defined “population center distance” as the distance from the reactor “to the nearest boundary of a densely populated center con- taining more than about 25,000 residents.” 10 C.F.R. § 100.3(c) (emphasis added). NIPSCO’s own witness con- ceded that “the centroid itself doesn’t have a boundary; it is a point.” Section 100.11(aX3) also adds: In applying this guide, dc consideration should be given to the population distribution within the pop- ulation center. But giving such due consideration to population dis- tribution does not eliminate the need to establish a boun- dary. Recently the ASLAB in another case found that the city limits of San Clemente, with an estimated future population likely to exceed 25,000, was within the population center distance and te a recalculation of radiation dosage distances. We agree with AEC’s ASLAB in that case, that a densely populated center may extend beyond the political or corporate limits of the 25,000 population unit, but there is neither reason * Southern California Edison Co. (San Onofre Station), ALAB-248, RAI-74-12, 957, 957-61 (Dec. 24, 1974). In a foot- note, the ASLAB added: The “nearest boundary of a densely populated center” will not necessarily coincide with city or other political linftts, for densely populated areas may extend beyond such limits. In this case, a portion of the dividing line between {Orange and San Diego} counties does also represent the extreme southeastern limits of San Clemente. Id. at 960 n.7 (emphasis added). —36— nor sound safety policy to cut down the boundaries of that unit and make some hopeless attempt to construct imaginary boundaries. The ASLB in this case recognized the fallacy of sub- stituting centroids for boundaries" and found as follows: To initially determine the heavily populated portion of the population center, the Staff used official Geological Survey (USGS) maps ... on which densely populated areas are depicted in pink. NIPSCO, LBP-74-19, supra at 565 (Finding 25). The map relied upon for the “pink” sections (thus hav- ing boundaries) is the United States Department of the Interior geological survey of the Portage quadrangle. The map is “revised from aerial photographs taken 1967” and “field checked 1968.” It must be kept in mind that the principal portion of the hearings in this case took place in 1973, at least five years later. The map purports to show individual dwellings as dots but the “pink,” or what the map refers to as “red tint,” simply indicates areas where individual dwellings are not shown. The map itself makes no representations or even references to population. The map clearly indicates the “corporate boundary of Portage” as being about one mile from the reactor. In the critical determination of the “Population Center Distance,” it would be ludicrous to attempt to reduce the political boundaries of the population center by such HL The Board said that it “was particularly concerned with intervenors claim that the Applicant had erred in choosing Gary as the nearest population center, that indeed Portage would be the proper choice and that the nearest boundary was 1.1 miles from the reactor site.” NJP’SCO, LBP-74-19, supra at 565 (Finding 24). Member Walter H. Jordan of ine ASLB said during the evidentiary hearings (Oct. 11, 1972 Tr. 613): [Miy reading of Part 100 does not say distance to the center; it says distance to the population center, but it does not mean the center of the population center in mak- ing that calculation. It is the environs of that population center, I believe. vague and indefinite references. Presumably, the car- tographers could have tinted any part of the Portage quadrangle pink inasmuch as_ they indicated no — which they used in selecting the areas so tinted. Here the corporate boundary of Portage was about one mile from the reactor and the evidence showed that the boundary of census enumeration district 719 of Portage as well as the boundary of the enumeration district representing the entire City of Portage were both less than a mile from the reactor. Since both the ASLB and ASLAB found that the boundary of a densely populated center containing more than about 25,000 residents must be at least two miles away from the reactor, their con- = that a construction permit be granted cannot stand. This conclusion does not take into consideration the additional fact that the boundaries of each of the com- munities of Dune Acres, Porter and Burns Harbor are also less than two miles from the reactor and that their total population, when included with that of Portage and unincorporated county areas within two miles, of course, exceeds 25,000. Nor does this conclusion take into consideration the fact, discussed in greater detail later, that the master plan for the Indiana Dunes National Lakeshore con- templates that “[a]ctive use of all of the facilities provid- ed for in this plan, both in the State Park and in the Lakeshore, will permit a maximum visitation of about 87,000 persons a day.” (Joint Intervenors’ Ex. 23). Presumably, a large portion of such visitors would or could come within tne two-mile limit of the third buffer zone, but such a conclusion is not necessary because the AEC siting regulations have been viola in this case on the basis of Portage alone. It is also important to consider that Bethlehem Steel Corporation's 7,500 employees are located at about one mile from the Bailly site. IV In view of the vast consequences of shutting off or delaying a potential source of considerable energy in these times of energy crisis, together with the effect of such an occurrence upon the economic, financial and in- dustrial well-being and development of northern In- diana, obviously we cannot finally act without giving very serious consideration to every possible factor which may conceivably bear upon the problem. The factors in favor of granting the permit are so per- vasive and weighty that they are impossible to ignore. Northern Indiana, as the entire country, has a need for additional generating power, employment and economic assistance. Witness after witness made so-called “limited appearance statements,” appearing on their own behalf or on behalf of a group or organization or utility or union, requesting that a construction permit be issued in order to provide additional energy, jobs or stimuli to the economy. It is pertinent to note, however, that almost every witness in one fashion or another conditioned his request upon either his hope or confidence that AEC would adequately protect the public safety and, in some cases, also protect the environment. In fact these circumstances militating in favor of the permit are so overwhelming that perhaps the AEC ina completely well-intentioned and good faith effort to ac- complish ends which everyone seeks to accomplish has tended to become somewhat lax in assuring that the means employed thereby function in as safe a manner as possible so that one day we do not come to regret the proliferation of nuclear power. Judge Oakes of the Second Circuit Court of Appeals recently noted in a dissent his concern “that the AEC is charged with the dual duty of passing on licenses on the one hand but promoting the use of atomic power on the other.” Morningside Renewal Council, Inc. v. AEC, 482 —39— Kovd Bad. 240-4) (2d Cir. 1973), cert. denied, 417 U.S. GQ) (LOTA).! l’'resumably, this observation has been largely in- validated by the Energy Reorganization Act of 1974 which purports to divide the promotional and research aspects of atomic energy, to be administered by the Energy Research and Development Administration (ER- 1A). from the licensing and regulatory functions, to be administered by the Nuclear Regulatory Commission (NRC). Nevertheless, the orders which we are presently reviewing occurred under the old AEC procedure and the merging of the diverse functions thereunder may ex- plain to a certain extent some of the factors which we next consider. A AEC appears to have given no direct consideration to the clustering of nuclear power eng around the southern end of Lake Michigan within relatively short distances from the density of population of metropolitan Chicago. Although there appears to be no likelihood of a chain- reaction effect which would lead an accident from one plant to another, recent developments indicate that plants built within a contemporary time period tend to ‘' “AEC is supposed to stimulate the use of atomic ener and at the same time enforce the conditions of use imposed by its own regulations. There is sound basis for the argument that the two roles are incompatible.” C. ALLARDICE & E. TRAPNELL, THE ATOMIC ENERGY COMMISSION 123 (1974). See also Carolina Environmental Study Group v. AEC, slip op. No. 73-1869 at 9 (D.C. Cir., Jan, 21, 1975), where the court said: “The possibility of such a conflict of interests is obvious.” For case and commentator documentation of “whether the regulatory agency is unduly oriented toward the interests of the industry it is designed to regulate, rather than the public interest it is designed to protect, see Mr. Justice Douglas’ dis- sent in Sierra Club v. Morton, 405 U.S. 727, 745-48 (1972). develop structural deficiencies at or about the same time.'* There are presently eight nuclear power plants within about 75 miles of downtown Chicago!’ and six more are in the planning stage.'* The Bailly Generating Station, Nuclear-1, as this plant has been designated, would be the fifteenth plant clustered around metropolitan Chicago. The AEC siting regulations as they pertain to the pop- ulation center distance, discussed supra in Part III, provide in part: Where very large cities are involved, a greater dis- tance may be necessary because of total integrated population dose consideration. 10 C.F.R. § 100.11(a)(3). This distance referred to, in the circumstances of the present case, is, of course, two miles, but the regulations indicate AEC’s awareness that consideration must be given to “very large cities” such as Chicago. In what the ASLAB called “the draft of an internal staff working paper pertaining to allowable levels of ‘* In September 1974, the Dresden 2 plant at Morris, Illinois, developed a leak in a 4-inch bypass pipe in its emergency core cooling system. Inspections revealed cracks or preliminary in- dications of cracks in the same pipe in eight boiling water reactors in the United States aad | nod Recently, the NRC closed 28 plants for inspection when five small cracks were discovered in 10-inch cooling pipes at Dresden 2, N.Y. Times Feb. 6, 1975 at 24, col. 1. | ‘t About 30 miles away are Zion 1 and Zion 2 at Zion, Illinois (completed in 1972); about 50 miles away are Dresden 1 (1960), Dresden 2 (1970) and Dresden % (1972) at Morris, I]- linois, and Donald C. Cook 1 (1973) and Donald C. Cook 2 (1974) at Bridgman, Michigan; somewhat further is Palisades (1971) at South Haven, Michigan. ALLERDICE & TRAPNELL supra note 11, at 224-25. : '' About 50 miles away will be Braidwood 1 (1980) and Braidwood 2 (1981) at Braidwood, Illinois; somewhat further will be LaSalle 1 (1977) and LaSalle 2 (1978) at Seneca, II- linois, and Byron 1 (1979) and Byron 2 (1980) at Byron, II- linois. ALLERDICE & TRAPNELL, supra note 11, at 224-25. oe. —— population density around nuclear plant sites” (NJPSCO, ALAB-224, supra at 252), AEC's staff noted as of April 17, 1973: A long-standing policy of the Atomic Energy Commission'® has encouraged siting of nuclear plants away from densely populated areas until ad- ditional operating experience has been obtained.'* When the AEC made the staff working paper public on April 9, 1974, the AEC press release accompanying the paper virtually repeated one year later that: Under a policy established by the AEC in 1962, the siting of nuclear power plants at some distance from densely populated areas has not been en- couraged until additional operating experience has been obtained. AEC Press Release, T-160 at 2 (April 9, 1974). In Power Reactor Development Co. v. International Union of Electrical Workers, 367 U.S. 396 (1961), the Supreme Court reversed the court of appeals which had set aside an AEC order issuing a construction permit, saying: The Court of Appeals put forward as an alter- native basis for its decision the holding that under the law the Commission may not authorize the con- struction of a reactor near a large population center 135 miles from the center of Detroit and 30 miles from the center of Toledo] without “compelling reasons” for doing so,. . . 280 F.2d at 651-652, and that no such reasons had been found by the AEC in this case. .. . [T]he position is without merit. The > The staff footnote at this point referred to Statement of Consideration, Reactor Site Criteria, 10 C.F.R. Part 100, 27 Fed. Reg. 3509 (1962). 6 AEC REGULATORY STAFF WORKING PAPER, POPULATION DISTRIBUTION AROUND NUCLEAR POWER PLANT SITES, Appen- dix A (April 17, 1973, released April 9, 1974). The staff paper also observed: “There has been no reason to take the ad- ditional incremental risk, however small, of incurring doses to a large metropolitan pe ey as a result of any accident in the nuclear facility when other suitable sites, less densely pop- ulated, remain available.” /d. at 1. —_“ statute and regulations say nothing about “compell- ing reasons.” Of course Congress (and the Commis- sion, too, for that matter) had the problem of safety uppermost in mind, and of course that problem is most acute when a reactor, potentially dangerous, is located near a large city. But the Commission found reasonable assurance, for present purposes, that the reactor could be safely operated at the present loca- tion, and that is enough to satisfy the requirements of law. Id. at 414." In the Power Reactor case, the majority had disposed of “the fears of nuclear disaster” by pointing out that the issuance of the construction permit was only the first step, and that the applicant would be required to satisfy safety-of-operations requirements prior to obtaining an operating permit. In 1962, however, Congress amended the Atomic Energy Act of 1954 to eliminate the two- hearing requirement in effect at the time of the decision of Power Reactor, The mandatory hearing is now re- quired only prior to the issuance of the construction per- mit. Union of Concerned Scientists « ALC, 499 F.2d 1069, 1074-80 (D.C. Cir. 1974). It is now unquestionably true as Mr. Justice Douglas said in his dissent in Power Reactor that “the time when the issue of ‘safety’ must be resolved is before the Commission issues a construction permit.” 367 U.S. at 419. The Senate Report accompanying the Energy Reorganization Act of 1974, included the following observations:'" Most commercial reactors now use a low-enriched uranium fuel that is not suitable for manufacture of eee In a dissent concurred in by Mr. Justice Black, Mr. Justice Douglas said: The construction given the Act by the Commission (and to- day approved) is, with all deference, a light-hearted ap- — to the most aoe, the most deadly, the most dangerous process that man has ever conceived. 367 Us. at 419. * §. Rep. No. 93-980, 93d Cong., 2d Sess., U.S. Cop, CONG. & ADMIN. NEWS 4868, 4869 n.1 (1974). — = bombs. They all produce plutonium as a by-product, however, which after reprocessing is potentially ex- plosive and suitable for bombs. ... At present, there are about 1 million pounds of plutonium and enriched uranium authorized in the licensed sector, about half of it of weapons grade. Plutonium, in ad- dition to its potential explosiveness, is one of the most toxic substances known to man. One thirty- millionth of an ounce, less than a pollen grain will cause cancer if inhaled or swallowed.'’ Despite these factors, the AEC gave no direct con- sideration to the effect of the existence of from eight to fifteen nuclear plants clustered around the Chicago metropolitan population. B Not only did the AEC disregard what at least some of its expert staff deemed important population § con- siderations, but it also disregarded its own past course of conduct in approving nuclear power sites. The 1973 Working Paper prepared by the AEC Regulatory Staff (supra, note 16) analyzed all sites which up to that time (April 1973) had been given con- sideration for licensing, and listed each one together with both its maximum Site Population Factor (SPF) “ In 1974, according to NRC, the nuclear power industry ex- «rienced more than 1.400 “abnormal” events, four of which vad a “directly significant” bearing on nuclear safety and radiation control. Chicago Sun-Times, Feb. 14, 1975 at 18, col. 1. “ 1A] weighting of the incremental populations around a reactor at the annular distances of 1, 2, 3, 4, 5, 10, 20, 30, 40 and 50 miles in comparison to a hypothetical site hav- ing a uniform population distribution of 1,000 people per square mile. The weighting factors applied to the populations at the various annular distances are inversely proportional to the distance from the source. The inverse weighting is in con- sonance with the increased atmospheric dilution with dis- tance for an assumed release of radioactivity emanating from a reactor. According to the weighting, a given pop- ulation close to the site would be considered to present a higher risk than the same population farther away. The weighting used is the distance (d) raised to the -1.5 power. STAFF WORKING PAPER, supra note 16 at B-9. aufifipes and the SPF at 50 miles. Sixty-eight proposed and ac- tual sites are listed and Bailly, which is 30 miles from downtown Chicago, has the sixth highest SPF at 50 miles. The five sites with a higher SPF at that distance according to rank were: (1) Ravenswood, (2) Indian Point. (3) Newbold Island, (4) Limerick and (5) Zion. Ravenswood was withdrawn as a_ proposed site because of population considerations.“! In October 1973, the AEC informed the applicant utility that Newbold Island was not suitable because of population density. Because of the high population density at Limerick, the AEC staff issued special safety evaluations. With regard to high population density sites the staff stated: Indian Point and Zion are the highest population density sites which have received construction per- mits. STAFF WORKING PAPER, supra note 16 at B-5. At the present time the staff uses an informal guideline that sites with a population density greater than that of an envelope of the Zion, New- bold Island and Indian Point site populations are not considered acceptable. Id. at 1. The Bailly site is closer to downtown Chicago than the Zion site. Furthermore, if the granting of a construction permit to Bailly was allowed to stand, Bailly would achieve the dubious distinction of being nestled within one of the top few most dense population areas of all 68 sites proposed ‘! Another site, Burlington, which had a lower population figure at 50 miles but greater overal! than Bailly, was also withdrawn because of population considerations. —5— thus far in the United States for siting nuclear power plants. ° y. ¥ | - = the adequacy of design in accounting for tornadic forces (Contention 57). Other radiological contentions concerning the design of specific systems also involve certain physical characteristics of the site, including hydrological charac- teristics considered in connection with the contention re- lating to contamination of water supplies (Contention 55). 19. The principal thrust of these contentions related primarily to Joint Intervenors’ assertions that the Bailly reactor is ‘‘novel and unproven.’’ If so, they contend that the site criteria in Part 100 should be applied on the basis of TID 14844 guidelines without giving any consideration to the effect of engineered safeguards systems in assessing site suitability. In the absence of consideration of the effect of engineered safeguards systems, the Joint Interve- nors assert that Part 100 would require either that Bailly N-1 Station have an exclusion distance of 4,650 feet, a low population zone radius of 16.8 miles, and a popula- tion center distance of thirty (30) miles or, conversely, that the power level of a facility to be located at the Bailly site be restricted to 33 MWt. 20. In assessing whether the site complied with the criteria set forth in Part 100, the Applicant and the Regu- latory Staff utilized the methods described in Safety Guide 3 which takes into consideration the effect of various engineered safety features in reducing releases in the event of accidents assumed for purposes of Part 100 computa- tions. Both the Regulatory Staff and the Applicant con- cluded that the calculated doses at the boundary of the 188 meter exclusion area and the 1.5 mile LPZ would comply with the criteria of Part 100, although the Staff’s estimates were more conservative than those of the Ap- plicant (S.E.R., Sections 2.1 and 10.1-10.7; PSAR, Vol. 1, Sec. 14).’ * Appl. Ex. 1E. alin 21. After a careful analysis of the record in this pro- ceeding, we cannot accept the assessments made by the Joint Intervenors that the proposed Bailly Generating Station is novel and unproven. The Bailly design is fune- tionally the same as other BWR facilities, both larger ani smaller than Bailly which are in operation and under con- struction (Butler, following Tr. 5278; Tr. 5280-82; Davis, following Tr. 1807; S.E.R., pp. 17-21). Moreover, the basic design characteristics for BWR’s have been demonstrated and proved out over many years of operation of a number of progressively larger facilities, since the development work at the Vallecitos Boiling Water Reactors as the pro- totype for the first commercial BWR power facilitiy at Dresden Unit 1 (Davis, following Tr. 1807), and have been successfully proven in the operation of much larger BWR units such as Quad Cities, Millstone, Oyster Creek, Dres- den and Monticello (Butler, following Tr. 5278). Differ- ences between Bailly and these other facilities relate prin- cipally to difference in physical details such as size and configuration and in the use of improved engineered safe- ty systems in the Bailly design (Tr. 5279-5286). The de- sign variables have been conservatively estimated to assure that the design will provide adequate safety protection (Tr. 5294-5299, 5309-5329). We do not find significant Joint Intervenors’ reliance on the stipulated fact that the emergency core cooling system (ECCS) of the proposed Bailly facility has never been experimentally tested under full scale LOCA condi- tions. All that is necessary is that the Bailly ECCS meet the Commission’s Interim Acceptance Criteria. Consumers Power Company (Midland Units 1 and 2), ALAB-123, RAI-73-5 331, at 341-2 (May 18, 1973). The evidentiary record reveals that the Bailly ECCS fully complies with the Interim Acceptance Criteria (S.E.R., p. 79 et seq.; Butler and Minners, following Tr. 5278, pp. 1-5; Tr. 5306- 5308, 5321-5327). =~ = The record fully supports a finding that the proposed Bailly reactor design is not unproven as a prototype, nor is it novel in design and, accordingly, appropriate credit must be given to its engineered safeguards in determining compliance with 10 CFR Part 100. 22. During the proceeding, the Joint Intervenors have asserted that the construction permit cannot issue unless the complete designs for the various systems and compo- nents have been prepared and submitted. The Appeal Board decision in the Consumers Power Company pro- ceeding, ALAB-123, RAI-73-5 331 at p. 355 (May 18, 1973), deals with this issue and rules otherwise. See Power Re- actor Development Co. v. International Union of Electri- cal Radio and Machine Workers, AFL-CIO, et al., 367 U.S. 396, 6L ed 2d 924(1961), where the Supreme Court held that Section 50.35 of the Commission’s Regulations permits the Commission to defer a definite safety finding until operation is actually licensed, and clearly suggests that the final design can be left to that stage of the licens- ing process (at pp. 400-408). 23. Intervenors offered into evidence ‘‘the ACRS generic letter’’ of December 19, i972 (J.1. Ex. 27-A through 27-G) as ‘‘listing outstanding and unresolved concerns on the part of the ACRS’’ with respect to the matters de- scribed in the letter as unresolved. Intervenors asserted that such items should be resolved before issuance of a construction permit for Bailly. The Commission’s response, dated February 5, 1973 to the ACRS, manifested the Commission’s intent of continuing research in order to solve those items described as unresolved on a generic basis and to assure that such items were, in the meantime, resolved on a case by case basis (Tr. 7745). The Regula- tory Staff’s witness Dr. Butler, who was called with respect to this contention as Intervenors’ witness, testified that applicable items on the ACRS unresolved list have been specifically considered in the review of the Bailly = = reactor by the Regulatory Staff and the ACRS. All such items have been satisfactorily resolved with respect to the Bailly reactor (Tr. 7741). ACRS communications relating to generic matters have been carefully considered by the Appeal Board in a num- ber of proceedings. These cases conclude that specific items which are the subject of ACRS scrutiny on a generic basis may be satisfactorily dealt with in a particular pro- ceeding. In Long Island Lighting Company (Shoreham Nu- clear Power Station), ALAB-156 RAI-73-10 844 at 935, the Appeal Board stated: ‘*. . . The lack of resolution of certain items on a generic basis .. . does not mean that the item has not been satisfactorily dealt with in connection with a par- ticular reactor.’’ *° The record is undisputed that the Applicant and the Regulatory Staff have satisfied the questions raised by the ACRS in its December 18, 1972 letter. 2+. Intervenors contended, and presented considera- ble testimony in support thereof, that the Applicant and Staff had relied on out-of-date maps in determining popu- lation center distances. The Board was particularly con- cerned with intervenors claim that the Applicant had erred in choosing Gary as the nearest population center, that indeed Portage would be the proper choice and that the nearest boundary was 1.1 miles from the reactor site (H. Read following Tr. 7391). Board questions directed to the Applicant and Staff revealed that the Applicant in the PSAR and the Staff in its S.E.R. considered the popula- tion center distance for purposes of determining compli- ance with 10 CFR 100 to be the distance to the city of Gary which was stated to be 5.5 miles from the site (Tr. 6375). *° See also Wisconsin Electric Power Company (Point Beach 2), ALAB-137, RAI-73-7 at 507-8. —77— Upon questioning concerning the City of Portage, the Staff acknowledged the expected population of Portage by 1980 would exceed 25,000, but for reasons given in the next paragraph, the Staff stated the correct population center distance of Portage was 4.5 miles (Tr. 6375). This dis- tance was acceptable since the population center distance exceeded two miles, which is the minimum distance to the boundary of a population center required by 10 CFR Part 100.11(a)(3), for an LPZ of 1.5 miles. 25. The Regulatory Staff indicated that the distance it considered was the distance to the nearest heavily popu- lated portion of the population center, not the distance to the political boundary which was used by Intervenors. To initially determine the heavily populated portion of the population center, the Staff used official U. S. Geological Survey (USGS) maps (Reg. Staff’s Ex. 7) on which densely populated areas are depicted in pink. These maps show individual dwellings (Tr. 6376). Applicant’s evidence based on population centroids of census enumeration dis- tricts substantiates the position that the pink areas on the USGS maps generally define the boundaries of dense pop- ulation within the City of Portage (Brown, following Tr. 10451; Appl. Ex. 31). The evidence also demonstrates that dense population growth is unlikely within two miles of the site (Tr. 6387-89; Appl. Ex. 32; Tr. 10685-99; Snyder, following Tr. 10482). 25. Although there is some argument as to whether the political boundary should be used in determining the pop- ulation center distance rather than the distance to that portion of the population center at which the dense popu- lation starts, the Board notes that 10 CFR Part 100.11(a) (3) specifically provides that ‘‘In applying this guide [that the population center distance be at least one and one third times the distance to the LPZ boundary], due con- sideration should be given to population distribution with- in the population center.’’ The Board concludes that giving =- = due consideration to the low population density of those areas closer than 4.5 miles to the Bailly reactor, princi- pally industrial areas of the City of Portage, the provi- sions of Part 100 with respect to population center distance are satisfied. 26. The Board concludes that the Bailly facility is not ‘‘novel in design and or unproven as a prototype,’’ tha! there is adequate evidence of compliance with the Interim Acceptance Criteria and an adequate description of the principal criteria and design bases of the various safety systems and components. The Board also finds that the Bailly site was properly evaluated using appropriate safe- ty guides and satisfies the requirements of 10 CFR Part 100. C. Additional Technical Design Information To Be Supplied 97. The evidence indicates that further technical or design information will be required from the Applicant prior to operation of the Bailly nuclear unit. Information and data that will be required to be provided in the final safety analysis report, and which can be left for such later consideration, include: a. pipe whip restraints or high energy fluid lines (Branch, p. R-58-1, following Tr. 2416; Tr. 1482-84, 1490-1492, 1500-02, 2418-19; S.E.R., pp. 38-39; Burch, following Tr. 2416; Appl. Ex. 1C) b. anticipated transients without seram (Appl. Ex. 1L; Tr. 2014, 10215; Reg. Staff’s Ex. 11, pp. 7, 74; S.E.R., pp. 134-135); e. post-LOCA hydrogen control (Appl. Ex. 1C, pp. 5.2-16.4, 5.2.3.9-4 (behind the red tab); Wattelet and Davis, following Tr. 557, pp. 2-5 and 2-6, S.E.R., pp. 85-86; Tr. 770, 771-772) ; —79— ; d. fuel densification (Appl. Ex. 27, pp. 4-5; Tr. 2536-37 ; Safety Evaluation Supplement No. 2, follow- ing Tr. 9826) ; e. REVAB system (Rockwell, following Tr. 10636, p. 8; Tr. 1634-35, 1676-77, 1688-97, 10680-84; Appl. Ex. IF, p. B.6-4; S.E.R., pp. 37-38, 95, 133); f. main steam line isolation valve sealing system Benaroya, following Tr. 1568, p. 2; Tr. 1600-06, 2435, 7772-74: S.E.R., pp. 42-42); g. rod sequence control system (S.E.R., pp. 116- 120; Tr. 2536-39) ; h. recireulation and flow control valves (S.E.R., Sections 14.4 and 14.6); and i. ECCS (S.E.R., Section 6.1). D. Research And Development Required 28. The record does not disclose any evidence suggest- ing that a research and development program, in the lab- oratory sense, is required to resolve safety issues in this proceeding. As previously discussed in paragraph 27, above, technical design data will be required to be sup- plied at the time of the application for an operating li- * cense. E. Applicant's Financial Qualifications To Design And Construct The Proposed Facility . 29. Northern Indiana Public Service Company, an In- diana corporation, is a public utility company engaged in supplying electrical energy and natural gas to the public in the State of Indiana (Petersen, Tr. following p. 546, pp. 1-2). The company operates in 30 counties in the northern part of Indiana, serving an area of about 12,000 square miles, with an estimated population of 2,080,000. —sgo— In the five-year period ended December 31, 1971, the company experienced a growth in the annual gross electric reveues of approximately 29%, and an increase of 28% in the kilowatt hour sales during the same five-year period. Electric revenues for the year 1971 totaled $132,135,833 (Petersen, pp. 1-2 and 1-3). Moody's Investor Service and Standard and Poor’s have both assigned a AA rating to the NIPSCO’s first mortgage, and Standard and Poor’s have rated the Applicant’s preferred stock AA (Petersen, Tr. following 546, pp. 1-7 and 1-8). The weight of the evidence strongly supports a finding that the Applicant is financially qualified to design and construct the proposed Bailly facility. The Joint Inter- venors failed to offer any substantial, probative evidence to the contrary. F. Applicant’s Technical Qualifications 30. The Applicant has had considerable experience in the design construction and operation of numerous fossil- fired power plants. Its principa. contractors, General Elec- tric Company and Sargent and Lundy, have extensive ex- perience in the design and construction of nuclear power facilities (S.E.R., following Tr. 741, p. 126). The Applicant has taken steps to assure that individuals with relevant prior experience are assigned by such contractors to the Bailly nuclear project (Lyle, following Tr. 582). The pro- posed organization of NIPSCO’s staff is modeled on the organization of project groups which have successfully managed the design and construction of nuclear power plants (Coe, following Tr. 584). The Regulatory Staff has concluded that the Applicant is technically qualified to design and construct the Bailly facility (S.E.R., pp. 126, 138). 31. The Joint Intervenors allege in Contention 21 that the Applicant has had no experience in the nuclear field and therefore the Commission cannot properly issue a construction permit because NIPSCO is without the pre- requisite experience. The Joint Intervenors did not direct the Board to any legal precedent in support of their con- tention. Our independent search reveals a number of cases sug- gesting that prior nuclear experience is not required. Al- though these decisions do not have the finality of either a Commission or Appeal Board decision, nevertheless, we find them persuasive. In the Matter of Northern States Power Company, Docket No. 50-130, Intermediate Deei- sion, 1 AEC Rep. 314 at 323 (April 21, 1960); In the Mat- ter of Niagara Mohawk Power Corporation. Docket No. 50- 220, Initial Decision, 3 AEC Rep. 5 at 14. The approach suggested by the Joint Intervenors would bar substantially all public utilities from utilizing nuclear power, since this field is relatively new with few organiza- tions knowledgeable in the intracies 6f nuclear power. We perceive no reason why experienced electrical utility oper- ators should not be permitted to take advantage of such expertise in developing another source for production of electrical energy, provided that the Applicant develops a schedule of training for its own employees to enable the Applicant to properly operate a nuclear plant. 32. Accordingly, we reject the Joint Intervenors’ Con- tention 21. We find that NIPSCO is technically qualified to design, construct and operate a nuclear power plant. G. Common Defense And Security 33. We conclude that permiting NIPSCO to construet the proposed Bailly nuclear facility will not be inimical to the common defense and security of the nation. All of Applicant’s directors and principal officers are citizens and residents of the United States (Petersen, fol- lowing Tr. 546, pp. 1-2; Appl. Ex. 1A, p. 2). All activities of the Bailly facility will be conducted within the jurisdic- tion of the United States. No diversion of special nuclear material for military purposes is involved (S.E.R., p. 135). Applicant is not owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government. H. Health And Safety Of The Public Since Joint Intervenors’ health and safety contentions are extensive and pervade the entire ambit of issues deal- ing with the health and safety of the public, we believe that our resolution of these contentions will be determin- ative of whether or not construction of the plant will be detrimental to the well-being of the public. Therefore, we will now serutinize the matters raised by these contentions as we deem appropriate. 1. Emergency Plans 34. The Applicant has presented in Section 13.10 of the PSAR, as amended by Amendment 16 (Appl. Ex. 1E) a description of the emergency plans for use in the event of an accident involving the release of radioactive mate- rial to the environment. Joint Intervenors allege in Con- tention 19 that the emergency plan has not been sufficiently developed. Bethlehem also expressed concern regarding the Applicant’s emergency plan insofar as it affected the Burns Harbor Plant of Bethlehem Steel. While Bethlehem was admitted as a party with regard to this concern (Order Permitting Intervention dated May 15, 1972), Bethlehem indicated on the record that it would prefer to work out the details of an emergency plan with NIPSCO separately from this proceeding if possible (Tr. 692). Nonetheless, the Safety Evaluation Report carefully assesses the emer- geney planning considerations for the Bethlehem facility. 35. Plans for the evacuation of the Bethlehem employees from the LPZ have been developed which would reduce the number of employees within the LPZ from a daylight maximum of 3,000 to 1,500 within two hours. Within the — —s3— next six hours, the work force would be reduced to an estimated 259. Twelve hours after the evacuation notice has been given, there will be approximately 168 Bethlehem employees within the LPZ, who comprise the emergency work force. Detailed procedures for the protection of these employees will be available when the FSAR is submitted (Dunn, following Tr. 640, p. 4-8; Appl. Ex. 1E, § 13.10.10). The dose calculation for these emplovees are well within the levels set forth in the Commission’s guidelines for siting of facilities in 10 CFR Part 100 (S.E.R., Supp-1, p. 4). In addition, the exclusion area has been restricted to an area totally within the boundaries of the Applicant’s property at the site and is, therefore, under the control of the Applicant (Appl. Ex. 1B, p. 2.2-25). The LPZ contains no schools or hospitals (Appl. Ex. 1B, p. 2.2-35). Arrange- ments have been made with Federal and State govern- mental agencies having emergency responsibilities (Dunn, following Tr. 640, p. 4-2). Local hospitals have been con- tacted and have expressed their intent to cooperate in the development of offsite medical treatment of injured per- sonnel (Dunn, following Tr. 640, p. 4-2; Tr. 652-53, 863- 64). As to the National Lakeshore, the Assistant Secretary of the Interior gave his assurance that the Department would cooperate in the development of an appropriate evacuation plan if Bailly were licensed (Tr. 7586). We are not persuaded by the indication that the Assist- ant Seeretary of the Interior would find it objectionable if the Department was required to establish a methodology for evacuation beyond the scope of its normal duties of the National Park Service (Tr. 7562). The issue is whether or not Interior would cooperate. The Joint Intervenors have not introduced any conflicting testimony which specifically rules out such cooperation. Moreover, the evidence fully supports the proposition that a satisfactory emergency plan can be developed for = the National Lakeshore ( Maletskos, following Tr. 642, pp. »-6; Tr. 669-71). In view of the foregoing, we sce no merit in Contention 19. Applicant’s emergency plan satisfies the requirements of Appendix FE to 10 CFR Part 50. 2 Seismic Characteristics 36. Intervenors contend that Bailly N-1 fails to comply with Criterion 2 of the General Design Criteria in that the seismie design of the facility is insufficient to withstand the impact of an earthquake such as occurred in 1971 in the San Fernando Valley, and that the geological data relied upon to establish seismic design are inadequate. Intervenors called two qualified geologists to present testimony (see McKee, followi:.g Tr. 8265, and MeCammon, following Tr. 8258). They did not claim that an earthquake as severe as the San Fernando Valley quake of 1971 was a reasonable possibility in the Bailly region. The thrust of their testimony was rather that at this time it is inap- propriate because of the paucity of firm information to set limits for a design basis earthquake at Bailly site. Al- though this amounts to a revision of the contention, the Board felt the issue was of such importance that we should hear any evidence that might support such a con- clusion. 37. The Bailly site is located on the edge of the Michi- gan basin near the Kankakee Arch. The geology of this Region is characterized by a sequence of about 4,000 feet of gently dipping Paleozoic age sedimentary rocks (270 to 600 million years old) overlying the Precambrian age basement complex (more than 600 million vears old), which is composed of igneous and methamorphie rocks. Obser- vation of these Pleistocene deposits, which characterize eastern Illinois, western Michigan, and northern Indiana, has produced no evidence of active faults. Those faults —&85— indentified in northeastern Illinois and northern Indiana are classified as late Paleozoic to early Cretaceous (270 to 700 million years ago). There has been no evidence of movement within the past for well over 35,000 years (Leeds, Allen, Perry and Chauvin, following Tr. 2063; Tr. 2117-22, 6610, 6631-36, 6647-54). 38. The acceleration to be provided for in a design basis earthquake is, in large measure, determined by the distance to an active fault. Consequently, the Applicant’s and Regulatory Staff’s testimony to the effect that there were no active faults near Bailly was of erucial impor- tance. (Leeds et al., follows Tr. 2063: Tr. 2114-16, 2124- 32, 2154-60, 2063, 6630-32). Joint Intervenors’ witness Me- Kee (testimony following Tr. 8265) had studied the geol- ogy of the region extensively. She identified faults in the Chicago Area; however, she did not demonstrate that the faults were active. Indeed, the faults she identified were taken into consideration in defining the design basis earth- quake (Leeds, following Tr. 9768). Witness McKee further testified that even if all faults were inactive at present, they might be triggered by injee- tion of waste into underground strata. She cited man-made earthquakes in the Denver Region that resulted from deep- well disposal operations. This was countered by Appli- eant’s witness (Winar, following Tr. 9775, pp. 1-4) who pointed out that the geologic conditions in the Denver Re- gion are entirely different from that in Northern Indiana. Further, the quakes were very small compared to the de- sign basis earthquake. There have been no similar pheno- mena in the Chicago Region. 39. The evidence discloses that the calculated ground motion at the site would be well below that which is pro- posed for the design basis earthquake (Tr. 9777-78, 9814- 15). All Class 1 structures at Bailly are being designed to meet an earthquake of intensity VII on the Modified Mer- ealli Seale of 1931. We share the view of Applicant’s wit- —36— ness Leeds that thix design criteria represents appropriate conservatism (Leeds, «f ol. following Tr. 2063, pp. R-56-7 and -8). The acecleration selected by the Applicant is more than twice the level of what could reasonably be postulated in accordance with AE criteria and represents a conserv- ative estimate based on pertinent carthquake experience (Tr. 2096, 2101-2, 6646-48). The safe shutdown earthquak« assumes 20° of ¢ for the horizontal component and 13.34% of g for the vertical component. The comparable figures for the operating basis earthquake are 10% an 6.67% (Tr. 2136-38). 40. Our consideration and the testimony in this pro- ceeding with respect to seismology and geology reflect the proposed Appendix A to 10 CFR Part 100 which had been available for guidance since its publication by the Com- mission on November 25, 1971 (36 F.R. 22601). On Novem- ber 13, 1973, the Commission adopted Appendix A, in final form. to become effective December 13, 1973 (38 F.R. 31279). 41. In its adoption of the final Appendix A, the Com- mission described all of the significant differences between the proposed and the adopted versions (38 PLR. 31279-81). Manv of the revisions simply reflect refinements in ver- hiage and editorial improvements (¢.¢.. the term ‘*active fault’? has been renamed ‘‘capable fault,’’ but the defini- tion and application of the terms are essentially identical in both versions of Appendix A), while others pertain to matters not pertinent to the Bailly site (¢.g., consideration is now required ‘‘of the loading effects of dams or reser- voirs in the determination of soil stability’’; there are no dams or reservoirs in the vicinity of Bailly). All pertinent substantive revisions are satisfied by the seismie and geologic information submitted. For example, the Com- mission identified three major changes at 38 F.R. 31279. With regard to the first major change, the Applicant has defined the Safe Shutdown Earthquake and Operating Ba- sis Earthquake in terms of geology and seismology, as well as in terms of the effect of these earthquakes on plant structures (Appl. Ex. 1B, 6 2.5.8 and Appl. fx. 1E ¢ 12.2). With regard to the second major change, the Applicant’s witnesses on seismology have satisfactorily addressed the matter of three-dimensional investigation, including infor- mation obtained in offshore geologic surveys in Lake Michi- yan (Leeds, et al., following Tr. 2063, pp. R-56-5 and -6; Tr. 6635-36). With regard to the third major change, the Applicant has selected an Operating Basis Earthquake and applied same to safety-related portions of the Bailly nuclear facility (Appl. Ex. 1B, § 2.5.3.5.1; Appl. Ex. 1K, $$ 12.1, 12.2). 42. We conclude that Applicant has complied with Appendix A to 10 CFR Part 100 both as originally pro- posed and as finally adopted by the Commission. 3. Tornado Design 43. Intervenors contend that the design of the Appli- cant’s plant fails to comply with Criterion 2 of the Com- mission’s General Design Criteria with specific reference to tornado design. Specifically, the Joint Intervenors con- tend that inappropriate meteorological data were used to determine the likelihood of tornadoes at the plant site; that in the event of a severe tornado the spent fuel pool would be exposed and the water would be sucked out of the pool leading to a loss of coolant and shielding and subsequently leading to a release of radioactivity that would endanger the public health and safety. 44. The design basis tornado for this facility consists of the assumption of the simultaneous application of a 300 MPH rotational wind velocity, a 60 MPH translational wind velocity and a 3 psia pressure drop in 3 seconds (Holtzscher and Williams, following Tr. 2175; Butler and Markee, following Tr. 5830; S.E.R., p. 64; Appl. Ex. 1C.2.3- 7 and 1K, p. 12.2-3). There have been no reported tornadoes —88-- that reached or exceeded the values selected. In fact, the highest value reported in the GE topical report ‘*Tornado Protection for the Spent Fuel Storage Pool (APED 5696) ’’ (Appl. Ex. 7) was 230 MPH rotational, with an average translational speed of about 35 MPH (Tr. 2193-94, 2196-97, 2199, 2202, 2237-38, 2265-67). The Applicant’s expert wit- ness, a co-author of APED 5696, concluded that the tornado values selected are conservative (Tr. 2193, 2202-03). The Staff indicated that the selected conditions are the worst that can be reasonably encountered at any site (Butler and Markee, following Tr. 5830), and that the probability of such a tornado occurring at the Bailly site was, while low, was nevertheless possible (Tr. 5858). Both the Staff and the Applicant concluded that the facility complied with the requirements of Criteria 2 of the General Design Cri- teria (S.E.R., pp. 64, 131; Holtzscher and Williams, fol- lowing Tr. 2175). Intervenors offered no evidence on this part of the contention. 45. With respect to that part of the Joint Intervenors’ contention that in the event of a tornado, the spent fuel pool would be exposed, the Applicant testified that a maxi- mum of only about 11.4 feet of water could be removed and, moreover, ‘‘over 13 feet of water would still cover the fuel racks, which is more than adequate to cool the spent fuel.’’ (Holtzscher and Williams, following Tr. 2175, p. R-57). The Staff’s witness, Dr. Butler, provided testi- mony that was in substantial agreement with the Appli- eant’s testimony with respect to the amount of water that could be sucked out of the spent fuel pool by a tornado, however, remote the likelihood of that event may be (Tr. 5835). He concluded that the water remaining in the pool will be sufficient to provide necessary cooling and shield- ing of the fuel (Tr. 5836). Dispersal of the water removed from the spent fuel pool would not endanger public health and safety (Tr. 5838-5840, 5850-5851, 5853-5856). 46. It is also possible that tornadie forces might dam- age and remove the metal siding panels located on the top ee Ee ee ee oe — so © ym a —g9— portion of the reactor building above the surface of the spent fuel pool. Damage such as this will not affect the safe shutdown and maintenance of the safe shutdown condition for the reactor (Holtzscher and Williams, following Tr. 2175; Tr. 2206-2209, 5845-5850). 47. The Board finds that the tornado design criteria for the Bailly facility are conservative and that, with re- spect to protection against tornadoes, the proposed Bailly nuclear facility design basis complies with Criterion 2 of the General Design Criteria. 4. Post-LOCA Hydrogen Control 48. Intervenors contend that the engineered safeguard systems of the Bailly nuclear facility are insufficient to avoid the generation of quantities of hydrogen within the containment after a loss-of-coolant accident that would require immediate venting of the hydrogen in order to avoid an explosive rupture of the containment. 49. There are two primary potential sources of hydro- gen in a boiling water reactor following a_ postulated LOCA: (1) The metal-water reaction of zircaloy cladding with coolant during the post-LOCA fuel cladding tempera- ture transient; (2) Radiolytic decomposition of the emer- gency core cooling water exposed to the decay energy of fission products (Wattelet and Davis, following Tr. 557, p. 2-1). The concentration of hydrogen in the containment at- mosphere following postulated accidents must be controlled to satisfy the requirements that containment integrity is maintained. The Applicant has committed to provide neces- sary systems to meet the requirements of Safety Guide 7, which describes a method of implementing this criterion which is accepted by the Regulatory Staff (S.E.R., p. 85; Wattelet and Davis, following Tr. 557, pp. 2-5). —90— 50. Testimony of the Regulatory Staff and Applicant demonstrate that the provisions of Safety Guide 7 are con- servative not only as to design requirements but also as to the assumptions which must be used to establish the design requirements (Tr. 771). For example, the Regu- latory Staff witness testified that the G value of .5 mole- cules/100 ev is the highest value the Siaff has ever noted in any test conducted. Usually the amount of hydrogen generated by radiolytic decompositions is only a fraction of this value. The metal water reaction of 5% is higher than that expected or seen in tests (Tr. 773), and, in faet, is higher than that computed to assure compliance with the Interim Acceptance Criteria for the ECCS system. Com- pliance with the provisions of Safety Guide 7 adequately controls the concentration of hydrogen following a LOCA and will assure that containment integrity is maintained. The Applicant has committed itself to comply with the requirements of Safety Guide 7 in the design of contain- ment protection for the facility. 51. The Board finds that the design criteria for the facility, in accordance with the provisions of the Safety Guide 7, is adequate to assure that hydrogen generation after a LOCA will not require a venting of the hydrogen to avoid an explosive rupture of the containment and provides for adequate control of hydrogen concentration in containment following a loss of coolant accident to assure that containment integrity is maintained. 5. Post-LOCA Radioiodine Control 52. Joint Intervenors contend that engineered safe- guard systems in the Bailly reactor for controlling radio- iodine in the event ou. an aecident do not assure that the levels of radioiodine that would be released can meet the guidelines set forth in TID-14844 and the limits set forth in 10 CFR Part 100. ee —— he eqeeeeery - = 53. Analyses of the offsite radioiodine doses in the event of a design basis accident were made by Applicant and the Regulatory Staff on the basis of the conservative ussumptions of iodine release from the core in Safety Guide 3 (Jason and Dougherty, following Tr. 606; S.E.R., p. 113). In order to meet the requirements of 10 CFR Part 100 using these assumptions, the Applicant has proposed un engineered safeguard system for the purpose of con- trolling the release of radioactive iodine following a loss- of-coolant accident (Jason and Dougherty, following Tr. 606). This system includes the primary containment sys- tem which limits the rate at which iodine can escape into the reactor building. In addition, the atmosphere in the reactor building is circulated through the Standby Gas Treatment System, consisting of two redundant gas proc- ess systems, each of which has the capacity to process 100% of the required containment volume without the aid of the other (S.E.R., pp. 86-87; PSAR, See. 5.3). The Applicant proposes to use a new type of iodine filter de- sign in the SGTS for Bailly. The iodine filters for Bailly would have two &-inch deep beds of impregnated activated carbon, in series and would be a gasketless, weld seam con- struction. The deeper beds increase iodine residence time and improve removal effeiency. The gasketless design pre- vents leakage from bypassing the charcoal beds (Jason and Dougherty follows Tr. 606; Tr. 828). Applicant has indicated that the filter efficiency of each filter will be not less than 99%, as compared to a maximum removal effi- ciency of about 95% for earlier style filters (S.E.R., p. 87). The Staff has evaluated the proposed design and concludes that it is acceptable for the efficiencies proposed (S.E.R., p. 87). This is supported by testimony of Staff witness Za- vodowski (Tr. 816-842) and test data (Appl. Ex. 3A, 3B, 3C, 4; Staff Ex. 2 and 3). o4. The Board finds that the engineered safety system proposed for the Bailly facility will adequately control radioiodine releases to assure that such releases in the —92— event of a design basis accident at the facility will not ex- ceed the guidelines set forth in 10 CFR Part 100. The Joint Intervenors’ contention is not supported by the weight of the evidence. 6. Pressure Vessel Integrity—Major Weld Cracking 55. Intervenors originally contended that due to weld cracking problems in the fabrication of other boiling water reactor vessels, special considerations existed which re- quired analyzing the failure of the reactor pressure vessel for the Bailly facility. After considerable deveiopment of this contention in the prehearing process, the Board ad- mitted this contention but stated: ‘‘In approving this specific contention, the Board in- tended that evidence thereon would be restricted sole- ly to the question of major weld cracking and their effects on a finding of reasonable assurance that the Bailly pressure vessel will be constructed in conform- ity with the applicable regulations. Moreover, we also intended that in regard to this contention, no evidence would be received on the environmental effects of a pressure vessel failure.’’ "’ 56. The reactor pressure vessel for the Bailly nuclear generating station is being manufactured by Combustion Engineering at its facility in Chattanooga, Tennessee (Car- lisle and Vassar, following Tr. 1884). Both Applicant and Staff witnesses testified that the Bailly reactor pressure vessel is being manufactured in accordance with Section III of the ASME Boiler and Pressure Vessel Code (Car- lisle and Vassar, following Tr. 1884 and S.E.R., p. 30). There was extensive testimony concerning the welding and examination and testing requirements of the ASME Code. "Ruling on Applicant's Petition for Reconsideration of Board’s Action on Contentions at Prehearing Conference of March 9, 1973’’ (dated April 3, 1973, p. 2). ND Mi a ie eae a —93— These vessels are also subject to ASME Section XI in- spection requirements. For this vessel preservice baseline examination will employ ultrasonic techniques (Carlisle and Vassar, following Tr. 1884, p. R-49-2). 57. With respect to welding performed on the pressure vessel by the Combustion Engineering Company, the de- tails of the welding procedures have been reviewed over a period of several years, by a number of AEC inspectors. These procedures have been found to be in conformance with Code requirements (Tr. 5074), and there has been no evidence of rejectable weld defects in the work per- formed. Applicant also presented evidence to the same effect (Carlisle and Vassar, following Tr. 1884). 58. In support of their contention, the Joint Interve- nors called Lawrence Bieker, a licensed structural engi- neer (Bieker, following Tr. 8584, p. 1). He has many years of experience in pressure vessel welding, but not with the large, thick-walled reactor or pressure vessels (Tr. 8586- 8589). Mr. Bieker’s testimony, and Joint Intervenors’ Ex- hibits 5, 6, and 28, pointed out that there had been prob- lems in reactor pressure vessel fabrication, chiefly with the Hatch vessel. No evidence was presented that would indi- eate that the Bailly vessel would encounter similar prob- lems. Mr. Bieker, at Tr. 8607, discussed the difficulties with certain small weld grooves, but the Applicant later testi- fied that the weld groove configuration described by the witness would not be involved in the fabrication of Bailly N-1 (Carlisle, following Tr. 10001, p. 2). Mr. Bieker claimed that welds in the Bailly pressure vessel could be repaired without the approval of independent quality con- trol personnel (Tr. 8594, p. 3). Applicant’s witness Car- lisle testified that after a weld is repaired, it is subject to the same quality assurance examination as any other weld in the pressure vessel (Tr. 10003-05). Intervenors’ witness Bieker also testified concerning an alleged design —94— defect due to stub tubes in the bottom head assembly being welded without subsequent machining to eliminate distor- tion (Bieker, following Tr. 8584, p. 2). He admitted that the problem would be eliminated if the machining were done after the welding (Tr. 8619-20). Applicant’s witness Car- lisle testified that, in fact, the machining will be done sub- sequent to the welding (Carlisle, following Tr. 10001, ). 3; Tr. 10009-13). 59. Joint Intervenors’ Exhibit 7 indicates a manufac- turing deviation in the welding of the Bailly vessel, name- ly, inadequate preheat. The Applicant testified that on sev- eral occasions involving a small percentage of the welding performed on the Bailly vessel, the preheat fell below the 300° required by General Electric and that this was re- ported in accordance with established procedures (Car- lisle and Vassar, following Tr. 1884, pp. R-49-4 and -5; Tr. 1947-56). This was not a violation of ASME Section IL, but, nevertheless, a special ultrasonic and radiograph ex- amination of the subject area will be made (Carlisle and Vassar, following Tr. 1884, pp. R-49-4 and -5). 60. The Board finds that there is reasonable assurance that the Bailly vessel will be completed in accordance with the ASME Code and will be free of major welding de- fects, and that there is no evidence of any special consid- eration concerning the integrity of the Bailly pressure ves- sel. 7. Anticipated Transients Without Scram (ATWS)— Common Mode Failures And The REVAB System 61. The Joint Intervenors contend that as a result of a common mode failure, the reactor may fail to scram when challenged. Further, the failure to scram on some transients that can be anticipated within the life of the power plant can lead to unacceptable consequences. It is the position of the Applicant and Regulatory Staff that the probability of failure to scram is very small. In addi- —95— tion, they claim that the Bailly facility ean survive an ATWS event since a trip of the recireulation pumps is provided as a back-up to the seram. Protection against such common mode failures is pro- vided by functional diversity in the use of different plant variables; physical diversity in the separation of safety related instrumentation and components; quality assur- ance in the review of design and testing of instrument com- ponents, and operational administrative measures includ- ing checking of maintenance or repair work, and accessi- bility of instrument settings and calibration controls (Appl. Ex. 8; Tr. 2594-2723, 4772-4812). The Regulatory Staff concluded that Applicant’s design against common mode failures is acceptable (Tr. 6362) in view of the additional protection provided by recirculation pump trip. 62. After carefully reviewing the evidence, we con- clude the position of the Joint Intervenors on common mode failures has not been sustained. 63. As part of Contention 51, the Joint Intervenors assert that the Applicant’s use of the REVAB system would impose duties on the relief valves beyond their duty to protect the primary system’s integrity. There are 10 safety/relief valves on the main steam lines inside the primary containment to provide the pro- tection against primary system over-pressure. The opera- tion of these valves with the REVAB system permits the reactor to accommodate a loss of load without reactor scram (Benaroya, Tr. 1628-1658; Davis, following Tr. 2617; S.E.R., p. 37). The added function of the REVAB to these valves would not adversely affect the safety fune- tion of these valves (Benaroya, Tr. 1629-1632; Davis, fol- lowing Tr. 2617). The Regulatory Staff concluded that Bailly’s REVAB system was acceptable (Tr. 1634-1635; Tr. 1676-1677). —96— Intervenors’ witness was concerned that a sustained continued release through the REVAB system might raise suppression pool temperature such that the ECCS would not adequately cool the core in the event of a LOCA (Jackson, following Tr. 9340). This area was previously pursued by Board inquiry (Tr. 1682-1684). The Staff indicated that the operating technical specifications would include a limitation on pool temperature, presently con- sidered at about 125°F, which would require shutdown of the facility (Tr. 1688); thus, assuring that a failure to close REVAB valves would not affect the ability of the ECCS to perform in the event of a LOCA (Tr. 1688). 64. As to ATWS, the Staff has recently summarized their position in Staff Exhibit 11 (WASH-1270). They conclude that nuclear plants must either provide diverse protection systems which make failure to seram much less likely, or provision must be made for mitigating the con- sequences of a failure to scram on anticipated transients. The proposed Bailly design includes provision for trip- ping the recirculation pumps in the ATWS event (Davis, following Tr. 2001, p. R-50-1), as a means of complying with WASH-1270. The signal for tripping the recircula- tion pumps will be completely separate from the scram signals (Appl. Ex. 1L; S.E.R., p. 135). Recirculation pump trip and the attendant reduction in reactivity due to bub- ble boiling restoration in the core should enable the reactor to survive the transient long enough for the liquid from poison to be injected into the core, thereby reducing the reactivity below zero and shutting down the plant. 65. Summurizing, the Board is satisfied that adequate provisions will be incorporated in the design of the Bailly nuclear facility to assure adequate protection against the effects of common mode failures and anticipated transients without scram. 8. Decay Heat And Containment Pressure 66. The crux of the Joint Intervenors’ contention is that, ‘‘ ... the maximum pressure peak in the dry well and pressure suppression chamber will exceed the design limit . . .’’ because the Applicant has failed to inelude all the available energy sources in its analysis of the peak containment pressure. They cite as their example the fact that decay heat was computed by use of ANS 5.1 which ignores certain key factors in computing the decay heat component of the total energy. 67. There are four principal sources of energy that contribute to peak containment pressure following a LOCA and decay heat is the least of the four (Tr. 5868, 5891-5902, 5907-5908). The fraction of total energy contributed by decay heat is very small. It is so small that any uncer- tainty as to its quantity would have a negligible effect in determining the maximum pressure in the containment following a loss-of-coolant accident (Tr. 2283, 5869). The evidence also showed that the peak pressure in the dry well is governed more by (1) the rate of coolant blowdown [the rate of energy removal from the reactor vessel], and (2) the rate of energy transferred from the dry well to the wet well (Tr. 5893). 68. Both the Regulatory Staff and the Applicant pre- sented evidence that the calculated containment peak pres- sure is 37 psig, whereas the design pressure is 45 psig (S.E.R., p. 53; Dunn and Davis, following Tr. 2282, p. R- 52-2). This design margin is adequately conservative (S.E.R,. p. 53, adopted by Dr. Butler as his testimony, Tr. 5881). 69. The thrust of the testimony of M. R. Jackson, Joint Intervenors’ witness, is that uncertainty exists in the cal- culations for primary containment peak pressure during a major LOCA because the calculations were hased on models and were unsupported by experimental verification (Jack- son, following Tr. 9340, pp. 6-7). Mr. Jackson’s conclusion —9g— was based on his assertions that: (1) the steam, which will flow from the dry well into the wet well via the vent pipes following a LOCA, will not be completely condensed, anc (2) his reliance on Rockwell's testimony that there ha- been no full-scale prototype blowdown test of the Bailly containment and suppression chamber (Tr. 10668). The witness was not aware of experimental data having been taken of simulated LOCA conditions with an 18-inch diam- eter break venting fluid at the rate of 43,000 pounds per hour and had no knowledge of tests which used a 24-inch diameter downcomer into the suppression pool. He agreed that if there were information showing the same volume flow rate through a 24-inch diameter downcomer it would well completely eliminate any scale up factor (Tr. 9357). In rebuttal, Dr. Butler of the Regulatory Staff testified that experimental data taken on a full seale, 1/48 segment of the Humboldt Bay containment and 1/112 segment of the Bodega Bay containment, served as the basis for con- cluding that there will be complete condensation of the steam which passes through the vent pipes into the sup- pression pool water (Butler, following Tr. 10253, p. 1). Further experimental data is contained in Board Exhibit 6, non-proprietary portions of GE Topical Report, NEDE- 19182, dated May 1970, selected by Joint Intervenors from a document supplied by the Staff, and received without ob- jection. Mr. Jackson also asserted that Fauske data re- vealed that Applicant's blowdown flow model was non- conservative (Jackson, following Tr. 9340). Applicant’s evidence rebutted this assertion showing that the model used, in fact, aceurately predicts the Fauske data (Rock- well, following Tr. 10636). 70. In view of the foregoing, the Board concludes that total energy sources have been appropriately taken into account in calculating the peak pressure following a LOCA and that the containment design pressure contains an ade- quate margin of safety. —99— 9. In-Core Instrumentation 71. The uncontroverted evidence is that the system provides sufficiently rapid measurement of the local power densities so that if safety related actions need to be taken either by the automatic systems or the manual system, such action can be taken in time to prevent any of the full dam- age limits being exceeded (Tr. 5116). Furthermore, the power oscillations are adequately suppressed (Tr. 5118- 19, 5129). Intervenors presented no contradictory evidence on this contention. The Board concludes that the in-core monitoring system is satisfactory and complies with the i General Design Criteria. a oe 10. Restraints Against Pipe Whipping ; 72. Intervenors contend that the design of Bailly N-1 fails to comply with Criterion 4 of the General Design Criteria because adequate restraint against pipe whipping of the primary cooling pipe has not been provided. 73. The Bailly nuclear facility is in the construction permit stage of licensing and final design details are not available. Since pipe routing has not been finalized, re- straints against pipe whipping cannot be finalized (Tr. 2416). Although the pipe whip protection has not yet been designed, criteria has been established in the form of iden- tification of piping systems for which pipe whip pro- tection must be provided, identification of the technique to be used for the particular piping systems (e.g., re- straints, check valve near vessel, separation), identifica- tion of the degree of whipping permitted (e.g., none, limited area), and a prohibition against damage to essential sys- tems and components (Appl. Ex. 1F; S.E.R., pp. 39, 40; Benaroya, following Tr. 1471, p. 1-2). In a letter dated December 21, 1972 (supplemented on January 15, 1973) the Staff provided detailed guidance to the Applicant on methods of analysis of pipe whip for high energy lines —100— outside containment (Peg. Staff Ex. 4). The Applicant has agreed to comply with these Design Criteria (Tr. 1540, 2425-2426). 74. A piping system designed to the Applicant’s pipe whip criteria will satisfy General Design Criterion 4, with respect to pipe whip. The Applicant must complete his design in accordance with those criteria and submit the completed designs for such protection for the Regulatory Staff’s review as a part of the final safety analysis report along with the design details of the other important safety and protection systems and components. The evidence shows that Applicant will be able to satisfy these require- ments (Branch, following Tr. 2416; Appl. Ex. 1C; Tr. 1490-1492). 75. The Joint Intervenors presented no evidence on this contention. The Board finds that the Applicant will be able to submit a detailed design of piping systems to meet the criteria to which it has committed itself, and that such criteria satisfy the requirements of General Design Criterion 4. 11. Reliability Of Containment Isolation Valves 76. Intervenors contend that the design of the Appli- eant’s plant fails to comply with Criterion 54 of the Gen- eral Design Criteria because there is no assurance that the containment isolation valves of the Applicant’s plant will be sufficiently reliable to effect isolation of the con- tainment in the event of an accident in view of the record of isolation valves sticking open on other operating reac- tors. 77. There have been some instances in which main steam line isolation valves have failed to close properly. These instances have been investigated and corrective ac- tions taken (Davis and Kuhlman, following Tr. 2433; Tr. 1610, 2439-46, 2449-50). As a result of past experience, the —101— Bailly valves were selected so as to provide extra assur- ance of satisfactory operation (Davis and Kuhlman, fol- lowing Tr. 2433). 78. Intervenors introduced a Staff memorandum (Joint Intervenors’ Ex. 1) concerning an incident at Oyster Creek involving the failure of a relief valve to close, causing primary system blowdown after a reactor scram. During this event, one of the main steam line isolation valves failed to close. Even in this instance, isolation of the reactor was accomplished by proper operation of the redundant main steam line isolation valve (Joint Intervenors’ Ex. 1; Tr. 1586-1588). 79. The Bailly facility is provided with redundant MSIV’s. The probability of two MSIV’s in series failing is close to very low. Although not impossible, such fail- ure has not, in fact, occurred (Tr. 1599, 1608). The elec- trical system for the valves in each line are redundant, and meet the requirements of IEEE-279. There are a num- ber of signals for valve closure generated through the iso- lation control system and the inner and outer valves are controlled by different systems (Tr. 1617-18, 2457-58, 2465- 67, 2480-81). To provide further protection against the possibility of leakage past both valves if both valves close but some leakage exists, or even if one MSIV failed to close, the Applicant will install a sealing system in con- nection with these valves. Although such system has not yet been designed, a number of methods are feasible. When the design of such a system is selected, it will be submitted to the Commission for review (Tr. 1600-06, 2435, 7772-74). 80. The Applicant has provided preliminary informa- tion with respect to inspection and testing of the MSIV’s (Appl. Ex. 1C, 1F), and the details of such a program will be presented in the FSAR (Davis and Kuhlman, following Tr. 2433, pp. R-59-2 and -3). The operating technical spec- ifications will impose requirements for testing, which would normally be done weekly (Benaroya, following Tr. 1568; —102— Davis and Kuhlman, following Tr. 2433, pp. R-59-2 and -3; Tr. 1607-08, 2437-39, 2447-49). 81. We conclude that the steps being taken by the Applicant including the utilization of redundant valves. the addition of a sealing system to the MSIV’s, the type- of valves selected, and the proposed test program, provide assurance that there will not be a failure of the isolation valves to close when required and to effect isolation of the containment and that the requirements of Criterion 54 will be satisfied. The contention is rejected. 12. Control Rods 82. Joint Intervenors have alleged in Contention 60 that the N-1 design fails to comply with General Design Criterion 25 in that the design of the control rods and control rod drive mechanisms is such that stuck rods, mis- aligned rods, and failure to scram are highly credible events which the Applicant has failed to analyze in depth. Cri- terion 25 deals with protective system requirements for reactivity control malfunctions. It requires that the pro- tection system be designed to assure that specified accept- able fuel design limits are not exceeded for any single malfunction of the reactivity control system. 83. The control rod system has been described in de- tail (Appl. Ex. 1B (PSAR), Sees. 3.4, 3.5). They are standard GE designs which are similar to those used in previous BWR reactors going back to Oyster Creek. Pos- sible malfunctions including continuous rod withdrawal and malfunctions due to stuck and misaligned rods were described along with designs and requirements to prevent such malfunction or to protect fuel limits in the event of malfunction (Davis, following Tr. 2535; Tr. 5191-5232). 84. Rod withdrawal of the maximum worth rod from any normal pattern would not result in damage, and with- drawal from abnormal patterns is prevented by rod block —103— monitors (Davis, following Tr. 2535, p. R-60-2). The poten- tial for stuck rods is minimized by observation of in-core monitor readings and position indicator switches, careful quality control testing, and weekly surveillance require- ments (Butler, following Tr. 5191). Tests have been run with misalignment more than twice the tolerance limits without significantly affecting scram performance (Davis, following Tr. 2535). 85. No evidence was supplied by the Joint Intervenors that ‘‘stuck rods, misaligned rods, and failure to scram are highly credible events.’’ The Board finds that the con- trol rod drive mechanisms will comply with Criterion 25. 86. Although the matter is not related to Criterion 25, the Board inquired into the design basis rod drop accident (Tr. 1159). The Staff reassessed this postulated accident as described in Supp. 1 to the Safety Evaluation Report (following Tr. 741), and now requires that systems be provided which will limit the maximum energy yield of such an accident to less than 280 calories per gram. The Applicant wil! install a system to provide the required as- surance, which will be similar to the hard-wired rod se- quence control system recently required for other GE plants at the operating license stage (S.E.R., Supp. 1; Tr. 2536-39). Such a system would adequately prevent or mitigate the consequence of the design basis rod drop acci- dent. The design details of the system and any related design modification can reasonably be implemented during final design of the Bailly facility. 13. Fuel Densification 87. The Joint Intervenors assert that operation of the proposed Bailly reactor will be endangered by densi- fication of the reactor fuel. The record supports the conclusion that uel densifica- tion would not adversely affect the ability of Bailly to operate with adequate safety (Appl. Ex. 27; S.E.R. Supp. —104— 2, following Tr. 9826, pp. 10 and 11; Tr. 9744). The Appli- cant has proposed limitations on the linear heat generation rate which comply with the Regulatory Staff's conclusions (Tr. 9744). 88. The Board finds that the public health and safety would not be endangered by densification of reactor fuel in the operation of the Bailly facility under the proposed limitations of the Regulatory Staff and the Applicant. 14. Ground Water Contamination 89. Joint Intervenors’ Contention 55 alleges that ra- dioactivity from normal, abnormal, or emergency opera- tions of Bailly N-1 may contaminate the ground water and ultimately the civilian and potable water supplies, in- cluding Lake Michigan, thereby endangering the health and safety of the local public. The contention has two as- pects: first, Joint Intervenors challenge the validity of the conclusions concerning ground water flow to Lake Mich- igan; second, the contention involves assessment of the impact of possible spills.. 90. On the basis of a number of studies, including borings at the site (Appl. Ex. 1B, 1E, 6), Applicant’s hy- drologist concluded that the ground water gradient is to the north toward Lake Michigan and that there is only the possibility of a small gradient developing to the west (Dougherty and Annambhotla, following Tr. 2321). Later investigations were introduced to confirm the conclusion that ground water movement at the site was northward (Appl. Ex. 29). The Staff hydrologist Mr. Hulman also testified that the flow pattern of the ground water is clearly in a northerly direction into Lake Michigan (Hulman, fol- lowing Tr. 1444, admitted at Tr. 5139). 91. Joint Intervenors’ witness Mr. Hilty testified that ground water might flow east into Cowles Bog, on the basis of some data collected off site and from observation of —105— surface water flows (Hilty, following Tr. 8658; Tr. 8663, 8705). He recognized that the data were limited, and he had made no determination that ground water at this loca- tion flowed in the same direction as surface water (Tr. 8663, 8671, 8686, 8702). He indicated his belief that there was not a good hydraulic connection between Lake Michi- gan and the site (Hilty, following Tr. 8658; Tr. 8770). This testimony was rebutted by evidence of a good hydrau- lic connection between Lake Michigan and the site, based on borings showing a continuous stratum of sand from the site to Lake Michigan (Annambhotla and Brissette, fol- lowing Tr. 10334; Tr. 10354, 10357-10359). 92. The record demonstrates sufficient investigation and evidence concerning ground water flows to establish that the ground water flow gradient is to the north toward Lake Michigan and to enable conservative estimates of ground water flow to be made (Dougherty and Annambhot- la, following Tr. 2321; Hulman, following Tr. 5140). 93. With respect to the other aspect of this contention, possible sources of liquid contaminants considered were: (1) service water, (2) cooling tower blowdown, (3) proc- essed liquid wastes, (4) makeup demineralizers, (5) sew- age treatment plant effluents, and (6) liquids resulting from radwaste accidents (Dougherty and Annambhotla, follow- ing Tr. 2321). The makeup demineralizers and the sewage treatment plant effluents are kept physically separate from potentially radioactive liquids and, therefore, cannot be- come sources of radioactivity (Dougherty and Annambhot- la, following Tr. 2321). The service water system and cool- ing tower blowdown and processed liquid wastes will be monitored carefully to assure that radioactivity contained in liquid effluents will not contain concentrations in excess of 10 CFR Part 20, and, in fact, will comply with the limitations set forth in Proposed Appendix I to 10 CFR Part 50 (Dougherty and Annambhotla, following Tr. 2321, p. R-55-3). —106— 94. The remaining source of possible liquid contami- nation is an accidental spill in the liquid radwaste system. Applicant’s analysis of a possible worst case spill showed that the concentrations of radioactive materials reaching the Lake would be a small percent of the values containe:! in 10 CFR Part 20, Appendix B (Dougherty and Annam- bhotla, followig Tr. 2321; Tr. 2350-2353, 2376-2406). The Staff undertook an independent assessment on the basis of more conservative worst case assumptions. Even under these assumptions the release of radioactivity into the lake of all isotopes would be less than the concentration values in 10 CFR Part 20, Appendix B, Table II, at the point of admixture with the Lake, without regard for dilution in the main body of the Lake (Lynch, following Tr. 10305; Butler, Tr. 10312). The postulated conditions would occur for only a few hours and are not averaged over an entire year, so the postulated exposures involved would be quite small (Tr. 10312), and the possibility of such a major spill of the contents of a large radwaste tank is very unlikely (Tr. 10313-10321). 95. The Board finds that there has been adequate in- vestigations of ground water to establish the direction of flow and conservative estimates of the amounts of such flows. The Board also finds that, based upon such infor- mation and the very conservative estimates of possible spills, there is reasonable assurance that public health and safety will not be endangered by contamination of civilian and potable sources. 15. Quality Assurance 96. Notwithstanding the fact that quality assurance in general was not raised as an issue by the Joint Intervenors, the Board in this proceeding has considered very carefully the Applicant’s quality assurance program for the purpose of determining whether it would be in public interest to grant the Applicant a construction permit. —107— 97. The Quality Assurance (QA) program for the Bailly facility involves the Applicant, General Electric (the NSSS supplier), Sargent & Lundy (the architect- engineer), and C. F. Braun (the construction manager). (Appl. Ex. 1F, p. D. 1-2; Lyle, following Tr. 582, p. 36; Carlisle, following Tr. 9590, p. 1). The areas of responsi- bility for each organization are identified in the PSAR (Appl. Ex. 1F, p. D. 2-6), but the Applicant maintains full responsibility for quality assurance for the Bailly facility (Carlisle, following Tr. 9590). The Applicant will have a QA staff of 8-10 persons assigned to the project. In addi- tion, the United States Testing Company serves as Appli- eant’s QA consultant (Lyle, following Tr. 582, pp. 39-40; Carlisle, following Tr. 9590). The organizational struc- ture is diseussed further belew. 98. Applicant’s QA manual gives a detailed explana- tion of the manner of implementation of each of the Cri- teria of Appendix B to Part 50 (Appl. Ex. 26, Sees. 1-18). The remainder of the manual consists of Appendices A through D which give further detailed explanations of certain of the criteria such as organizational charts, re- sponsibility charts, essential components lists, and QA procedure lists. This manual requires safety essential com- ponents and services be furnished by qualified vendors and contractors. Each vendor or contractor must have a docu- mented quality program which is reviewed and approved by Applicant and must submit its procedures to Applicant for review. The program also requires vendors and con- tractors to establish with Applicant a notification program whereby Applicant is notified of specific QA activities so it may be represented at these activities. The manual ref- erences detail procedures for activities to be performed directly by NIPSCO (e.g., review of vendor’s QA manual). These procedures described explicit rules to be followed and documentation required in performing that activity. All audit and surveillance activities are performed in ac- cordance with a written plan. Any nonconformance iden- tified is documented and must be resolved to the satisfac- tion of Applicant’s QA supervisor (Carlisle, following Tr. 9590). 99. The Regulatory Staff has conducted a number of reviews of Applicant’s quality assurance program. Staff witnesses on this matter were the inspectors responsible who have conducted the on-site review of Applicant’s qual- ity assurance program and have audited implementation of the quality assurance program requirements for pro- curement activities for the Bailly facility (Vandel, fol- lowing Tr. 9649; Sutton, following Tr. 9652). During the course of these inspections, various deficiencies were iden- tified and brought to the attention of the Applicant, and have been corrected by Applicant, including one violation involving the lifting of certain components with a crane in the absence of written procedures for such activities. Satisfactory procedures have been provided by the Ap- plicant. 100. The Staff has concluded that the QA program for the Bailly station meets the requirements of 10 CFR Part 50, Appendix B and the commitments reflected in the PSAR (Tr. 9649, 9652). See Regulatory Staff’s ‘‘ Motion For Leave To Amend Staff’s Proposed Findings Of Fact And Conclusions Of Law Dated February 1, 1974,’’ filed February 14, 1974. See Tr. 5-6 of Oral Argument on Pro- posed Findings, February 19, 1974. 101. The Applicant’s organization provides for a QA staff which is separate from the nuclear staff and reports directly to top management (Lyle, following Tr. 582, p. 39; Coe, following Tr. 584, p. 7-4). The Applicant submit- ted an undated organizational chart more clearly establish- ing the independence of the QA organization of the Appli- cant (Appl. Ex. 28). The Board requested clarification of whether a certain unexplained relationship between the Executive Manager and the Manager of the Fossil Plant Engineering compromised the clear separation of reporting =i. lines of authority between the Bailly Nuclear Staff and the Bailly N-1 QA Staff. This limited nonreporting rela- tionship was satisfactorily explained and does not com- promise the clear separation of these organizations. The merger of reporting lines between the Nuclear Staff and the QA Staff is at top corporate management level, the Vice President or those acting in his absence. We conclude that the Applicant’s QA organization is provided with sufficient independence to comply with 10 CFR Part 50, Appendix B. Support for this finding is contained in Commonwealth Edison Company (LaSalle County Nuclear Station, Units 1 and 2), Commission Mem- orandum and Order, CLI-73-32, RAI-73-12 1072 (Decem- ber 7, 1973). 102. In view of the foregoing findings, the Board con- cludes that the issues raised during the health and safety phase of this proceeding pose no barrier to the granting of a construction permit to the Applicant. Issuance of such permit will not be inimical to the health and safety of the public. III. Frypincs on EnvironmMentaL Matters 103. The location of the proposed facility on the shore of Lake Michigan to the north and bounded on the east by the Indiana Dunes National Lakeshore is characterized © by its proximity to an environment of special qualities. The Indiana Dunes region includes an exceptional com- bination of sand dunes, marshes, swamps and bogs, white sand beaches, and diversified flora and fauna—a natural area difficult to equal anywhere in the Nation. That portion of the Indiana Dunes National Lakeshore adjacent to the Bailly site and consisting of some 1200 acres is one of the best remaining areas in the Indiana Dunes. Starting at the edge of Lake Michigan and proceeding inland, it con- tains a cross-section including a wide beach of largely —110— quartz sand, a low foredune covered by spar-e vegetation; higher dunes, some active and sparsely covered and others having trees and other vegetation, rising from 25 to almost 200 feet above Lake level in a complex of blowouts, ridges, valleys, and interdunal ponds. This cross-section exempli- fies terrestrial plant succession on dune sand as described by Cowles over 70 years ago. The dynamic nature of the Indiana Dunes has attracted scientists as an ideal outdoor laboratory to study the principles of ecological succession (Olmsted, Beecher, Klotz, Krekeler, following Tr. 7979, pp. 1-2). On the south of the dune complex, and to the north of the ancient shoreline of the Calumet stage of Lake Michi- gan (now the location of U.S. Highway 12), is a unique wetland area known as Cowles Bog which contains an undrained pond or depression filled with water and largely covered by a relatively thick mat of vegetation. The fact that the Department of Interior’s National Park Service has designated a sizeable portion of these unique natural settings abutting the proposed plant site as a Class IV Land namely ‘‘an outstanding natural area of special significance for public edification’? (Reed, following Tr. 7528, p. 5; J-I Ex. 23) has weighed heavily in the Board’s consideration of the potential impact of the proposed plant on its surroundings. 104. The effects of the proposed plant on the environ- ment can be conveniently considered as those involved in: (a) the impact of construction; (b) the impact from opera- tion; and (c) the impacts of decommissioning. A. Impact of Construction 1. General 105. The potential environmental impacts of construc- tion of proposed plant identified in the FES are as follows: (a) Noise and ground vibrations from pile driv- ing during the initial phases of construction which may —lli— temporarily disturb visitors and wild life in the nearby Indiana Dunes National Lakeshore (pp. IV-5, VII-1); (b) Dust from construction activities, possibly in- cluding blowing sand and cement aggregate (p. IV-5); (c) An inerease in automobile traffic during con- struction due to the travel of the construction crew to and from the site (pp. IV-6, VII-1); (d) A possible impact on local beach erosion and lake bottom disturbance if a temporary dock for de- livery of the reactor vessel is constructed (pp. IV-6, VII-1); (e) The impacts of surface water run-off (p. IV- 6); and (f) The impacts of dewatering (p. IV-6). 106. The Applicant has agreed to acopt measures to control dust from the concrete batch plant, as well as mea- sures to cover graded areas during construction (Reg. Staff Ex. 5, pp. IV-7, D-84 and D-85). 107. The impacts of construction are not expected to be greater than those normally attendant to large construc- tion projects (Reg. Staff Ex. 5, pp. IV-6, IV-7). In any event, they will be temporary and will not remain after the completion of construction (Reg. Staff Ex. 5, p. IV-7). There is a possibility of greater impact with respect to dewatering activities and surface water run-off, both of which we discuss below. 108. Assuming control measures are taken with respect to dewatering as discussed below, we find that the adverse environmental impacts of construction on the proposed plant will be small. 2. Surface Water Runoff 109. In Contention 63, Intervenors allege that Appli- cant’s ultimate site development due to construction and —112— operation of N-1 will result in inereased surface runoff and storm water disposal problems. ~The Applicant and the Regulatory Staff testified that the Bailly site is generally flat and the surface generally porous (Chauvin, following Tr. 4680, p. E-63-1; Fisher, following Tr. 6039, p. 1). Joint Intervenors concede that this is the case (Hilty, Olmsted, and Read, following Tr. 8759, pp. 14-15). This flat, porous surface has a capacity to accept storm water without ponding (Chauvin, follow- ing Tr. 4580, p. E-63-1; Fisher, following Tr. 6039, p. 1; Ililty, et al., following Tr. 8759, p. 15; Tr. 4700, 8785). No previous problem with surface runoff has been encount- ered (Tr. 4684-85, 4695), and none is expected even though construction of Bailly N-1 will reduce the porous surface available to accept the water (Fisher, following Tr. 6039, p. 1; Tr. 4681). During construction, while structures are open to the weather, storm water collecting in the structure, as well as water discharged from roof drains, will be discharged into the existing discharge flume for fossil Units 7 and 8, and ultimately to Lake Michigan (Chauvin, following Tr. 4680, p. E-63-1; Tr. 4692). Similarly, any increase in the quantity of surface water runoff from other sources could be controlled by collection and discharge through the dis- charge flume for Units 7 and 8 (Tr. 4694). In addition, it could be trucked from the site in tank trucks (Tr. 4694), or diverted by physical barriers in some manner (Fisher, following Tr. 6039, p. 1; Tr. 6060) so that it would not enter the National Lakeshore. Although Applicant had stated that no problem was expected with respect to in- creases in contaminants of surface water runoff due to construction activities, control of any such contaminants, if required, could be accomplished relatively easily by diversion to a collection point and separation of the con- taminants from the water for disposal (Fisher, following Tr. 6039, p. 2; Tr. 4688-90, 4693-94, 4698-99, 6060). —113— The Joint Intervenors’ direct testimony stated that the quantity of runoff during heavy rains would cause ‘sheet flow and washing . . . along unvegetated grades of 15 per- cent or more. ...’” (Hlilty et al., following Tr. 8759, p. 15). However, their witnesses were unable to identify any such grade on the Applicant’s site (Tr. 8774, 8805). The testi- mony also stated that airborne particulates will gradually cover the permeable surface of the site and increase the amount of runoff (Hilty ef al., following Tr. 8759, p. 16). However, no such effect has yet been observed even though the fossil generating plants at the site have been operating for some time (Tr. 4695-96). 110. We find that the allegations of the Joint Inter- venors’ Contention 63 that increased surface water runoff from the site will ‘‘pollute’’ the Lakeshore lack merit. The surface water runoff from the site due to construction and operation of the proposed plant will not have a meaningful impact on the Lakeshore either because of increased amount of water or because of increased amount of con- taminants in the water. In the event that surface water runoff problems arise unexpectedly, there are ample means to correct the situation. 3. Dewatering * 111. In Contention 62, the Joint Intervenors allege that during the life of the plant, the Applicant engage in a site dewatering program and that dewatering resulting from the construction of the proposed plant will result in permanent damage to portions of the Indiana Dunes Na- tional Lakeshore by lowering the ground water table and draining ponds and bogs in the adjacent portion of the National Lakeshore. '? Although Joint Intervenors apparently allege that dewatering an surface water runoff will have adverse effects during operation, as well as construction, we have chosen to discuss these two subjects under the heading ‘‘Impact of Construction’’ since any potentially vperational impacts are less significant. —114— Actually, the Applicant will not engage in any dewater- ing activities during the operation of the proposed plant because the use of membrane waterproofing and water- stops on the outside of basement walls will make it unnec- essary (Chauvin, foilowing Tr. 4634, p. E-62-2, Tr. 4638). The membrane waterproofing proposed by Applicant ap- pears to be sufficient for the purpose of preventing ground water from entering station buildings during the life of the proposed plant (Tr. 4639-40). Only negligible seepage, if any, of ground water into the building during operation is expected (Tr. 4668). Any seepage that is encountered will be dealt with by sump pumps and will not result in site dewatering (Tr. 4666). 112. Thus, the Board finds that there will not be any environmental effects due to dewatering during operation of the proposed plant. 113. During construction, dewatering of excavations for station buildings will be required (Tr. 4635; Reg. Staff Ex. 5, p. IV-6). The site will be dewatered to an elevation of approximately -4 feet (Annambhotla and Brissette, fol- lowing Tr. 10334, p. 5; Tr. 4650, 10369, 10383-84). Ground level water will be prevented from being lowered in any part of the excavation below the -4 foot elevation by plac- ing sheet piling which is essentially impervious to water around that part of the excavation below -4 foot eleva- tion and removing any seepage through the sheet piling by local pumping (Tr. 4650, 10369). The surrounding ground water table elevation will not be significantly af- fected by this local pumping (Tr. 10369). Applicant’s experts caleulated the potential drawdown effect due to dewatering and presented evidence to show that the estimates of drawdown resulting from the caleu- lations are conservatively high (Annambhotla and Bris- sette, following Tr. 10334, pp. 2, 5-6; Tr. 10374, 10386, 10388-89). In addition, ash pond seepage is likely to offset the drawdown to some extent (Annambhotla and Brissette, —115— following Tr. 10334, p. 6). Intervenors’ hydrologist also calculated the drawdown effects of dewatering (Tr. 8790, 8805, 8810-13). Applicant’s hydrologist reviewed these cal- culations and stated that, in his opinion, they contained erroneous assumptions as well as numerical errors (An- nambhotla and Brissette, following Tr. 10334, p. 4). Joint Intervenors’ calculations assume that the site will be de- watered to an elevation of -10 feet when the record clear- ly shows that Applicant will dewater only to the -4 foot elevation ( [Text truncated at 120,000 characters. The full text is on the page linked above.] --- Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/brief%3Amicro_IA40385004_0520%3A2. Public record. Not legal advice.