ML20127L313

Initial Decision.Director of Regulation Authorized to Issue Provisional CP to NSP

ML20127L313
Person / Time
Site:	Monticello
Issue date:	06/19/1967
From:	Murphy A Atomic Safety and Licensing Board Panel
To:	NORTHERN STATES POWER CO.
References
NUDOCS 9211230490
Download: ML20127L313 (10)

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UNITED STATES OF AMERICA I . ATOMIC E'NERGY . COMMISSION In the Matter of )

I NORTHERN STATES PC'GR CUMPANY ) Docket No. 50-263 Monticello Nuclear Generating Plant ) .

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INITIAL DECISION l f Preliminary Statement g 1 1. This proceeding involves the application of Northern 2

I State's Power, Company, a Minnesota corporation (Applicant), dated i

! August.1, 1966, and eight amendments thereto, dated September 8, 1966,

!, November 18, 1966, December 29, 1966,- January 10, 1967, January 21, 1967, March 3,1967, March 16,1967, and March 28,1967, (hereinafter i collectively referred to as "the application") for a provisional con-struction permit to construct a boiling water reactor, designed to l operate at 1674 megawatts (thermal), to be located at the Monticello

~ Plant of Northern States Power Company in Wright County, Minnesota.

( 2. The application was reviewed by the Regulatory Staff i

(Staff) of the Atomic Energy Commission-(Commission) which concluded

! that the proposed facility can be constructed at the proposed site without undue risk ,to the health and safety -of the public (Staff

$ Safety Evaluation, pp. 23-24). The application was reviewed by the Advisory Committee 'on ' Reactor Safeguards' which submitted reports l [datedMay 11, 1966 and April 13, 1967. These reports are included in i

the record as Appendices C-1 and C-2 of the Staff Safety Evaluation, i I Nj -f -1  %

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3. On April 21, 1967, the Commission issued a " Notice of The.

Hearing on Application for Provisional Construction Permit."

notice of hearing was published on April 22, 1967, in 32 Fed. Reg.

It contained the requi-6304 and othewise given as provided by law.

site findings proposed to be made by the' Director of Regulation to f

support the proposed issuance of the provisional construction permit.

3 4. Pursuant to the notice of hearing-and in accordance with the requirements of the Atomic Energy Act of-1954, as amended, (Act) a hearing was held before an Atomic Safety and Licensing Board (the Board) in the, Wright County Courthouse, Buffalo, Minnesota on May 25-26, 1967,* to cbnsider whether a provisional construction permit should be issued. The parties to the proceeding were the Applicant and the Staff.

Pursuant to the provisions of 10 CFR Section 2.715(c) of the Commission's

" Rules of Practice" limited appearancos woro entered by Mr. G. B. Sea-borne, representing Dr. Robert N. Barr, Secretary of the Minnesota Water I Pollution Control Commission; Mr. Eric Wykes, representing Dr. Barr in his capacity as Secretary and Executive Office of the Minnesota Depart-ment of Health; Mr. P. J. Murray and Sidney A. Fre11sen, representing the Honorable Jarle Leirfallom, Commissioner of Conservation in the Minnesota State Conservation Department; Mr. Leo J. Barthel, Chairman, Wright County Board of Commissioners; and Mr. Charles Erlandson, Mayor l

of the City of Montice31o. A limited appearance was also made by Mr.

t John Pogors a resident of Hopkins, Minnesota, on behalf of himself, the Citizens Committee to Preserve the Homeowners from a Nucicar I

• A_ pre-hearing conference was held at the same place on May 12, 1967.

A transcript of the pre-hearing conference was made and is included in the record of this proceeding.

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Catastrophe at Monticello, Minnosota, and Clear Air, Clear Water -

Unlimited. There were no intervenors.*

Findines of Fact

5. The proposed facility will be owned by the Applicant a privately owned electric utility corporation with assets in excess of $800 million, it plans to finance the cost of construction of i

the proposed facility in the same manner as it finances other plants, namely, through the internal generation of funds in the ordinary course of business and the issuance and sale of securities.

6 ., The Applicant has gained experience in the construc-tion and operation of nuclear powered generating stations as a result of its construction and operation of the Pathfinder Plant. General Electric Company, which will-design and construct the entire Plant,

, has had considerable experience in the design and construction of nuclear projects.

• A petition to intervene, dated May 10, 1967, was filed by Mr. Pegors on his own behalf and for Clear Air, Clear Water - Unlimited, "a local citizens' group concerned with matters of public health and safety."

The petition to intervene was denied by the Board by Order dated May 18, 1967, for failure to conform with the requirements of Section 2.714 of the Commission's " Rules of Practice," in that it did not set forth the interest of the petitioner, how that interest would be' affected by Commission action, and the contentions of petitioner. The May 18 Order granted permission to Mr. Pegors and. Clear Air, Clear Water - ,

, Unlimited to make limited appearances. The denial'of intervention was without prejudice to a new application complying _ with the Rules of Practice. By a series of letters . dated May 19 and May 20, Mr. Pegors '

sought to cure the defects in the petition and askad for reconsidera-

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tion of the Board's order. However, at the hearing, and before the Board ruled on the request for. reconsideration, Mr. P.egors withdrew -

his petition to intervene and entered a limited appearance. In addition to making a statement, Mr. Pegors requested the Board to ask the -

witnesses a number of questions, which with some exceptions,'_were asked j by the Board and responded to by witnesses.

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7. The site of the proposed Monticello Plant consists of 1325 acres located partially in Sherburne County (on the eat ; bank of the Mississippi River) and partially in Wright County (on the west bank of the-River). The Plant will be located in Wright County. The site is about 22 miles southeast of St. Cloud (1960 population 33,815) and 30 miles northwest of Minneapolis -(1960 population of metropolitan area, including St. Paul, approximately 1,400,000). The nearest resi-The dence is approximately 3000 feet from the proposed Plant location.

' area surrounding the site is rural with only a few small villages within 15 miles of'the site.

8. The design of the major systems and containment structure, which bear significantly on the acceptability of the Plant under the site criteria guidelines identified in 10 CFR Part 100, have been analyzed and evaluated by the Applicant and the Staff at a power level of 1674 MWt.

The thermal and hydraulic characteristics of the reactor core were ana-lyzed and evaluated at 1469 MWt, the proposed initial power level.

To assess the suitability of the plant under the criteria of

' 10 CFR Part 100, the Applicant and Staff have each anelyzed extreme I

accident conditions and the resulting off-site doses from airborne fission

- products. The doses estimated by the Staff, from its calculations, .are greater than those estimated by the Applicant by factors approaching one i ,

million (TR' 103-9) .

In answer to questions by the Board concerning these discrepan-l cies, the Staff stated that "---- our analysis is directed to trying to determine a maximum upper limit of potential consequences - " (TR 87-17),

whereas a witness for the Applicant stated that "---- althobgh they (the i

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1 applicant's estimates) are not extremely conservative numbers, they are ,

valid numbers" _ (TR 89-23) . The difference between the St aff and Appli-I cant estimates seems to be due to differences in th. degree of reliance ,

l upon engineered safeguards to reduce the amount of fission products that I31

could be released. For example, in determining the amount of I that would be released in a rod-drop accident, the applicant assumed that the s l

iodine concentration would' be reduced by a factor of4 10 by wat n b w 3

tion, whereas the Staff, in interpreting the same basic test data, ass'umed 5

a factor of 10, Th'e Board is left with the impression that ,there' are. either 4

significant technical differences between the safety _ evaluations of the Staff and Applicant, or-significant differences.in their interpretations of the necessary degree of improbability- for, reactor accidents appropriate to site analysis under the criteria of 10 CFR Part 100, or both. The Board review would be' assisted if rationalization of such important differences preceded the Staff's recommendation of the, grant of a pro- .

Visional construction permit. ,

, The Board, mon its own review of the data presented in this hearing, concludes that extreme reactor accidents of the type analyzed .

[ by the Applicant and Staff are of sufficiently low probability and result-

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l ing off-site doses would fall within the 10 CFR Part 100 guidelines.

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! 9. The proposed facility is a single-cycle, for'ced circulation, boiling water reactor producing steam for ' direct use in the steam turbine.

The reactor will.be fueled with 'slightly_ enriched uranium dioxide pollets sealed in Zircaloy-2 fuel-rods. ' Reactivity control will-be provided by

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movable control rods, temporary control curtains, and variable recircula.

tion flow.

The Monticello reactor is one of a series of similar reactors I

including Quad Cities Units I and II, Dresden Units II and III and Millstone. Although of different power levels all of these reactors l

. have approximately the same average power density and all are being manufactured by the General Electric Company. Because of the similarity l

in the reactors the Staff's Safety Evaluation, "in response to the Commission's desire to shorten and simplify the hearing and decisional process," (Staff Safety Evaluation, p.3) deals only with those features of the propose'd reactor which differ from its predecessors in the series.

This practice, while it commendably shortens the record and tends to focus discussion on matters of greatest importance does, in j the view of the Board, create some difficulty. One aspect of the diffi-culty is.that the practice of dealing only with new features presupposes a familiarity with earlier' reactors in the series dich,'in fact, the

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i Board members may not have.

A more serious- aspect of the difficulty arises from the requireinent that the record of the proceeding contain sufficient infor-i j mation to support the findings proposed to be made by the Director of i'

Regulation and the issuance of the provisional construction permit.

Whether the record is sufficient of course turns on the question of the Board's function in these cases. Although one can foresee a time when reactor technology is sufficiently standardized and sufficient operating experience has been accumulated so that the jobs ~of the Staff and the i

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Board will be reduced to passing on new features,that point has not yct been reached. The new features in these cases are not departures from established standards but from other reactors ,whose "old" features remain in many cases untested. h'hile the fact that the Staff, the ACRS and previous Boards have approved issuance.of construction permits in earlier cases is comforting, it does not mean (at least in the B,oard's view) that the safety review can be restricted to the new features.*

For example, many of the engineered safeguards proposed in this appli-cation are yet to be worked out in detailed design, confirmed by test and placed f.n operation.- Decisions on similar safeguards in the case of previotis reactors were necessarily based, in part,. upon expectations of test programs yet to be carried out.

Each subsequent review should have the benefit of an identifi-cation by Applicant and Staff of further information from related test programs, which may confirm earlier favorable expectations or raise questions of safety not previously anticipated. A portion of this hearing was devoted to such a summary by the Staff and Applicant. In response to questions during the hearing concerning the adequacy of the features of this. reactor under the AEC' general design criteria for nuclear power plants, the Staff provided information from their safety .

l evaluation of the Dresden III reactor. Also presented during the hearing i

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was a comparison between safety features of this plant and similar features last reviewed by members of this Board. The Board conclude ~s that the total record of this application is sufficient for a decision of the issues ft for consideration by this Board.

• In the case of Board decisions, unless specific, reference to novel features is made in the findings by the Board, there is no assurance that any par-ticular feature was scrutinized and found acceptable.

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10. The Monticello Plant will be the first reactor to utilize a reactor pressure vessel which is to be partially assembled at the site, although a significant portion of the vessel will be shop fabricated. This approach to fabrication of large reactor pressure i vessels was dictated by problems arising in the transportation of very large loads to this particular site. Although a' field erection pro-

. cedure has not yet been used for reactor pressure vessels, the tech >

nique has been used for large vessels in other industrial application..

Evidence and testimony were presented to support the conclusion that the field-assembled reactor vessel in the present instance can be fabricated to the same quality which would be expected from conven-tional shop-fabrication. The Plant design will permit in-service inspection of the reactor vessel and other primary system components.

11. The reactor primary system will be located within an independent primary containment system (consisting of two large steel pressure vessels: the drywell and the pressure suppression chamber) which -is designed (a) to accommodate the pressures and temperatures which could occur in the unlikely' event of a failure equivalent to a circumferential rupture of a recirculation line within the primary 7 g containment, and (b) to retain radioactive fission products which might f be released as a consequence of_ this and lesser accidents. Th'e pro-4 posed facility will also utilite an independent secondary containment system which consis'ts of the reactor building (which houses the primary containment system), the standby gas treatment system, and the 290-foot

! stack. The s.econdary containment system is designed to minimize the release of radioactive materials by providing controlled and filtered m-a

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e, release of effluents in the unlikely event of an accident. Reactor process and auxiliary lines which pass through the containment will be

, provided with isolation valves which close automatically under accident conditions.

12. Auxiliary cooling systems will be provided to prevent fuel clad melting for the full range of primary system pipe sizes which may be postulated to fail. These systems will be multiple and indepen-dent to provide maximum reliability and to assure that no single failure would preclude the continued operation of these independent safety features. -

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13. The proposed facility incorporates severn1 design features for which additional information will be provided to the Staff by the Applicant. These design features include the core spray system, the control rod worth minimizer, control rod vclocity liditor, the in-core neutron monitoring syt '.em, and the j et pumps. These features have been reviewad in prior proceedings in connection wita, and are also being included i:.the Dresden Units 2 and 3 (Dockets Nos. 50-237 and 50-249) and the Quad Cities Units 1 and 2 (Dockets Nos. 50-254 and 50-265) which are now under construction and are scheduled to commence regular opera-tion prior to the completion of the Monticello facility. In addition, the Applicant will . provide to the Staff a test program, and the resuits thereof, for the testing of the main steam line isolation valves under simulated accident conditions.

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14. The application contains a description of the site and I

the-basis for its suitability, the principal architectural and engineer-l- ing ' criteria for the design of the facility, a detailed description of i

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information, and the review of the application by the Staff has been adequate to support (1) the findings proposed to be made by the Director of R,cgulation, and (2) the issuance of the provisional construction permit, as proposed by the Director of Regulation.

17. Pursuant to the Act and the Commission's regulations, i

IT IS ORDERED THAT, the Director of Regulation is authorized to issue a H provisional construction permit to Northern States Power Company sub-stantially in the form of Appendix A to the " Notice of Hearing on Applica-tion for Provisional Construction Permit" in the captioned, matter.

, E IS FURTHER ORDEP.ED, in accordance with 10 CFR 2.764, good cause to the contrary not 6aving been shown by a party, that this Initial Decision shall be effective immediately upon issuance, and in the absence of any further order from the Commission, shall constitute the final decision of the Commission forty-five days after issuance,- subject to the review thereof and further decision by the Commission upon exception filed by any party pursuant to 10 CFR 2.762 or upon its own notion.

ATOMIC SAFETY AND LIf'ENSING BOARD b 5h b?- .0 s . .m b ,

Arthur W. Murphy, Chairman'

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m-Thomas H. Pigford ,g/

Dated at Germantown, Maryland t

this 19th day of June , 1967.

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