ML19344A404
| ML19344A404 | |
| Person / Time | |
|---|---|
| Site: | Bailly |
| Issue date: | 08/15/1980 |
| From: | Lynch M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19344A403 | List: |
| References | |
| ISSUANCES-CPA, NUDOCS 8008200197 | |
| Download: ML19344A404 (11) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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NORTHERN INDIANA PUBLIC
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Docket No. 50-367 SERVICE COMPANY
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(Construction Permit Extension)
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(Bailly Generating Station,
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e Nuclear-1)
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AFFIDAVIT OF MAURICE D. LYNCH I, Maurice D. Lynch ~, being duly sworn, state as follows:
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I am employed by the U. S. Nuclear Regulatory Commission as a Project Manager 1
in the Division of Licensing, Office of Nuclear Reactor Regulation.
I have been the project manager assigned to the Bailly Generating Station since 1973. A copy of my professional qualifications are attached.
In its August 7,1980 Order Following Special Prehearing Conference, the Licensing
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Board propounded several questions to the Staff regarding the Applicant's pile installation plans. The Staff response to these questions follows.
Question 1 i
Are the Permittee's plans with regard to the pilings advanced to the stage where they would be considered at a construction permit L.
proceeding?
If not, what remains further to be done to bring them to that stage?
Answer Yes.
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' Question 2 When does the Staff estimate it will complete its analysis of the short pilings proposal?
Answer On or about September 15, 1980.
Question 3 What are the [ practical or other] reasons if any, why it would be preferable to defer the short pilings proposal to the operating license proceeding, rather than hear it at this proceeding before further ccostruction comences?
1 Answer The fundamental, practical, engineering reason not to conduct a hearing on the short pilc,roposal at this time is the inherent site-specific nature of any foundation pile design. An ergineered structure above ground can be built to specific, well controlled tolerances and the nature of the design is amenable to analysis. The physical placement and the analysis of piles is subject to l
greater variability because the performance of the piles is highly dependent
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upon the in situ soil properties which are variable, especially in glacially f
deposited soils such those found at the Bailly site.
In recognition of this site-specific character of a pile foundation, good engineering practice requires r
r conservative design procedures.
The Staff finds that conservative design procedures have been adopted by the Applicant and that there is reasonable assurance that the pile foundation can be installed to satisfy the design criteria for these piles. This finding l
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will be fully explained in the forthcoming Staff evaluation of the pile installation proposal.
One of the basic tenets of conservative design for pile foundations is that an installed pile foundation is not accepted until after installation data and the pile load test program are evaluated. The Staff is following this particular procedure for the Bailly pile foundation.
While the " wave equation"* method was used to establish the driving specifica-tions (e.g., the blows per foot at final penetration) for the required load capacity, the load capability of the installed piles must be verified in keeping with good engineering practice as discussed above.
In this instance, good practice requires load testing of representative production piles to determine the ultimate load capability in compression, in tension and under lateral load.
These ultimate load capabilities will then be used to verify f
that the required factors of safety are achieved. This verification procedure was followed in the indicator pile program conducted at the Bailly site in 1978 which constituted a limited, representative sampling of the soil-pile conditions.
Pile load tests in the indicator pile program clearly demonstrated that an ultimate load capability significantly in excess of that required under the most extreme loading conditions (i.e., the Safe Shutdown Earthquake in combination with the design basis accident) could be readily achieved. These results are extremely encoura ging.
It should be recognized, however, that the indicator pile program represents only about three percent of all the safety-related piles. The indicator pile program load tests did not encompass all possible soil conditions which would i
- An example of the application of this method is provided in "The Mechanics of Pile-Soil Interaction In Cohesionless Soils," Holloway, Clough and Vesic, Decemb,er 1975.
- be encountered at the Bailly site. These load tests did establish that the piles can achieve the required load capability. The Staff is establishing a 1
t comprehensive pile load verification testing program for the production piles.
In this program, at least five types of pile tests will be required to encompass all significant soil-pile conditions at the Bailly site. This program will be fully explained and described in the forthcoming Staff evaluation of the pile installation proposal.
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It should be noted that there are a number of other verification steps in the f,
production pile placement procedures which will establish the most effective evidence that the safety-related piles will satisfy their primary design criteria established at the construction permit stage; namely, that they be high capacity, non-displacement piles. Areas which were disturbed by predrilling and jetting activities will receive special treatment and verification according to a plan j
proposed by NIPSCO. As noted before, the acceptability of the piles installed
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in these areas can only be determined after the piles are installed.
t, Another item which will receive special attention is the heave phenomenon which i
0 causes upward displacement of densely-grouped piles when adjacent piles are a
driven. This phenomenon was observed in the indicator pile program and verification steps have been proposed by the Applicant which will resolve any problems with heave.
However, the field data obtained during the placement 1
of the producticn piles will be the final determinant in establishing that the required driving specifications successfully account for the heave phenceenon.
-- Any hearing on this matter prior to completion of the verification procedures would likely culminate in a dispute over the efficacy of the proposed installation specifications. Such dispute is incapable of final settlement prior to an evaluation of the field data obtained from the physical placement of the production piles.
At that time, there is no strictly engineering reason why evidentiary consideration of the adequacy of the pile foundation must await a later point in time.
i In summary, analytical calculations, engineering experience with pile foundations, the 1978 indicator pile program, conservative pile load estimates and the pile installation specifications proposed by the Applicant combine to provide reasonable assurance that the fundamental pile design criteria will be met. However, the acceptability of the actual pile installation will be determined by the Staff only after all the verification steps in the production pile program have been performed.
If the piles, as installed, are unacceptable, further construction will not be pennitted unless and until any resultant problems are completely 4
resolved.
Question 4 What are the reasons, if any, why the Board should or should not be reasonably assured, without hearing that issue in this pro-ceeding, that all safety questions arising from the proposal to use short pilings will be resolved before the latest date mentioned in the request for the extension?
Answer-The Applicant has requested until December 1, 1987 to complete construction.
Production pile installation can commence imediately upon issuance of a favorable
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Staff safety evaluation. As already noted, this evaluation is expected to issue on or about September 15, 1980. Pile installation should take about six to nine months to accomplish, depending upon the weather.
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The Staff believes that there are numerous reasons which should reasonably i
assure the Board that all safety questions arising from the Applicant's present pile proposal will be resolved before the latest completion date in the Applicant's request for an extension. Basically, these reasons fall into two principal categories which augment each other:
(1) conservative design specifications; and (2) administrative procedures developed to ensure the proper implementation of the design specifications.
1.
Design Specifications The conservative design specifications include:
The design value of the acceleration for the Safe Shutdown a.
Earthquake (0.20 g) is derived from a conservative evaluation of the geology / seismology of the region; b.
The proposed seismic model of the Bailly facility is conservative in character; e.g., the control motion representing the seismic f
load is imposed at the foundation level; The design loads on the piles are conservatively calculated; c.
e.g., the seismic load is assumed to occur coincident with the design accident; a
d.
The Staff has required the Applicant to adopt conservative factors of safety for the piles; and The Staff has required the Applicant to adopt a conseryc.cive e.
method of ensuring that those areas whose soil density was m
a.
decreased by preconstruction activities, will be redensified l
to a value comparable to the original in situ density.
I All of the conservative design specifications have been adopted by the Applicant.
It is our position that additional design specifications are unnecessary. Therefore, any safety questions related to design specifications are considered resolved.
l 2.
Administrative Procedures The administrative procedures established by the Staff to ensure the proper implementation of the conservative design specifications are:
The Staff has reviewed the Applicant's proposed quality a.
assurance / quality control procedures to ensure that the detailed field procedures are consistent with the conservative des'ign specifications; b.
The Staff will have an on-site representative to monitor all significant installation and verification procedures on a continuing basis.
The Staff is requiring the Applicant to conduct a comprehensive c.
pile load test program to measure all significant soil-pile conditions at the site; d.
The Staff is requiring the Applicant to verify that the i
redensification program discussed above in item 1 (e) is i
successfully completed; and e.
The Staff has retained consultants with considerable stature s
and specialized expertise to review and evaluate the Applicant's proposal.
3.
Additional Reasons a.
The pile load capacity determined as part of the indicator pile program indicates that the ultimate pile load capacity of between 500 and 600 tons is significantly larger than the maximum conservative design load of about 220 tons.
b.
On the basis of its independent review of the Applicant's present pile proposal, the Advisory Committee on Reactor Safeguards (f.CRS) concluded, in part, that:
"[t]here wi)! be no difference in the safety of the 1
[Bailly] fac+1ity depending on whether longer or shorter piliNs are used [if certain referenced
- . itters are tre.ited as proposed].*
f All administrative procedures which are a prerequisite for pile driving have been established.
In sumary, the Staff believes that the above should provide reasonable assurance y.
to the Board that all safety questions involving the Applicant's present pile installation proposal are resolved.
The Staff will review and evaluate the entire pile placement program upon its completion and would expect that this review can be accomplished within three months. Any problems identified at that time should be capable of resolution well in advance of the latest completion date.
If such problems cannot be resolved, the Staff will take the necessary steps to prevent further construction.
- Letter f rom Max W. Carbon to then Chairman Hendrie, dated July 16, 1979. A copy of this letter is attached. The ACRS is kept informed on a continuing basis of all aspects of our review.
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_enav(28l Maurice D.
Subscribed and sworn to before me i
this 6th day of
//1980.
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' 'f ~ Notafy Puy ic I
My Comission expires:
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M. DAVID LYNCH PROFESSIONAL QUALIFICATIONS DIVISION OF LICENSING U. S. NUCLEAR REGULATORY COMMISSION My nane is M. David Lynch.
I am employ' : as a Project Manager, Licensing Branch No.1, Division of Licensing, U. S. Nuclear Regulatory Commission, Washington, D. C.
The Division of Licensing has t'1e respono bility of evaluating the environmental and radiological safety aspects of nuclear power plants for construction permits, operating licenses and operating nuclear power plants.
I attended the ' ooper Union School of Engineering and received a B. S.
C Degree in Civil Engineering in 1952.
I have also attended the Graduate School of Engineering at the University of California, Los Angeles, from 1958 to 1961, 3
l where I did extensive course work in physics, nuclear engineering, including j
reactor design and analysis, and in the field of engineering mechanics.
I have a total of 28 years of engineering experience after my B. S.
The last 22 years of my engineering experience have been in design, analysis and j
regulation of nuclear power reactors, as well as in the performance testing and analysis of operating nuclear reactors.
I have performed work in the fields of f
engineering mechanics, reactor physics, tnd systems analysis.
I joined Lockheed Aircraft Cerporation in 1952, where I was employed as a structures engineer responsible for the design and analysis of complex aircraft structures.
I was also responsible for the analysis of structural modifications to production line aircraft.
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In 1958, I joined Atomics International, a division of Rockwell International, where I was responsible for the structural analysis of major components of both liquid metal cooled and organically cooled reactors.
I joined the Marquardt Corporation in IMO where I performed the design and analysis of nuclear reacto'r cores for airborne, airbreathing nuclear randets.
Additionally, I was responsible for the design and analysis of various radioisotopic propulsion concepts.
In 1965, I joined the Knolls Atomic Power Laboratory, a facility operated for the then AEC by General Electric.
I was responsible for the nuclear reactor core analysis of the DIG reactor, a surface ship application of nuclear propulsion in the Naval Reactors program. My work included reactivity calculations of the reactor core, power distribution calculations, as well as the analysis of various core characteristics including rod sensitivities and temperature coefficients.
I was totally responsible for the special applications of the DIG reactor.
I joined the NRC (formerly the AEC) in 1972 and was assigned to the then Division of Project Management.
I was responsible for the licensing review of the Perry Nuclear Power Plant, Units 1 and 2, the Douglas Point Nuclear Generating Station, Units 1 and 2, the Vogtle facility, and Nine Mile Point. Unit 2.
I have also assisted in the review of the Limerick and Grand Gulf nuclear power plants and am presently responsible for the licensing review of the Washington Nuclear Plant, Unit 2 and the Bailly plant.
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%g UNITED STATES
!e NUCLEAR REGULATOPY COMMISSION
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS a, f4 f
wA$MNGToN, o. C. 20555 8
July 16, 1979 Honorable Joseph M. Hendrie Chairman
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U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
BAILLY GENERATING STATION, NUCIIAR 1
Dear Dr. Hendrie:
During its 231st meeting, July 12-14, 1979, the Advisory Committee on Reactor Safeguards reviewed the design of the pile foundations for the Bailly Generating Station, Nuclear 1, being constructed by the Northern j
Indiana Public Service Company (NIPSCO).
h is matter was considered by an ACRS Subcommittee at a meeting held in Portage, Indiana, near the site, on July 9,1979.
During its review, the Committee had the benefit of discus-sions with representatives and consultants of NIPSCO and of the NRC Staff.
he Committee also had the benefit of the documents listed below and of statements received from members of the public.
In your letter dated June 8,1979, you made the following request:
"The Commission requests the Committee to identify and address the significance (if any) of the engineering and safety issues arising from use of the shorter pilings as i
f oppsed to the longer pilings.
In particular:
(1) is the use of shorter pilings a significant design change from the standpoint of engineering, and would it require significant alteration of other aspects of the design of the facility; (2) what differences, if any, would there be in the safety of the facility depending on whether longer or shorter pilings are used?"
The Committee heard reparts on the experience to date relating to the driving of piles at the site, including the exploratory driving of the longer piles to the till or rock, the extensive exploratory driving of the shorter piles into th,e interbedded sand and clay layer, and the various borings and pile load tests that have been made over the pasc few years.
2 e Committee also heard reports on analyses relating to the factors of safety to be provided against various loading combinations and to the expected settlements of the structures sup;nrted on piles.
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Honorable Joseph M. Hendrie July 16,1979
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ne Committee has identified only two potential safety issues arising frcan the use of the shorter piles as opposed to the longer piles, and has con-cluded that neither of these will have any effect on the safety of the facility if the procedures proposed by NIPSCO or required by the NRC Staff -
are followed.
Be first of these results from the fact that some of the exploratory longer piles were installed with the aid of high pressure water jets which resulted in disturbance of the soil (chiefly the sand) in the interbedded layer. mis disturbance is limited to only a snall portion of the foundation area at four locations.
Unless remedial measures are taken, the shorter piles driven in these areas might be deficient in load-bearing capacity.
NIPSCO has proposed the use of " compaction piles" in the areas of disturbed soil to densify the disturbed soil so that it will be able to provide support equivalent to that in the other areas.
h e NRC Staff believes that this procedure is acceptable, and the Committee agrees, subject to compliance with the following procedures:
1.
Exploration by borings or by penetration devices to determine the vertical and horizontal exte,nt of the disturbed areas.
I 2.
Compaction of the disturbed material by driving compaction piles.
3.
Verification by borings or by penetration devices that all of the disturbed soil has been compacted.
I 4.
Performing a compression load test on at least one production pile in each disturbed area to verify its y
load-carrying capacity and load-deformation charac-J teristics.
NIPSCO has agreed to the.ce procedures.
I n e second issue resulting from the use of the shorter piles is the potential settlement of the supported structures.
R a settlement after construction would have been expe;ted to be essentially zero for the longer pile foundation.
For the shorter piles, the settlement has been estimated by NIPSCO to be on the order of two inches.
Settlement of this magnitude is not fr. usual for a nuclear plant and would have no significance to safety.
2 e Ccamittee has recommended to the NRC Staff, however, that the method of calculating the settlement be reviewed to assure that it has been done conservatively.
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Honorable Joseph M. Hendrie July 16,1979
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In addition, NIPSCO has proposed a program to measure settlement at numerous locations on the structures during operation of the plant, and the NRC Staff has stated that such me'asurements will be required by the Technical Specifi-cations and that suitably conservative limits an permissible settlements will be established.
In view of these commitments, the Committee believes that potential settlements, even if greater than those now predicted, would not represent a hazard to the Fublic.
The NRC Staff is continuing its review of the foundation design, and the Cormittee believes that the remaining foundation-related issues, not related to the use of shorter piles, can be resolved by the Staff.
i In direct response to the questions raised by your request, the ACRS believes that:
1.
We use of shorter piling is not a significant design change from the standpoint of engineering.
2.
%e use of shorter piling would not require significant alteration of other aspects of the design of the facility.
3.
I tere will be no difference in the safety of the facility depending on whether longer or shorter pilings are used if t
the matters referred to above are treated as now proposed.
i Sincerely, E
Max W. Carbon Chairman h
References:
1.
Preliminary Safety Analysis Report on Bailly Generating Station, Nuclear 1.
2.
Design Analysis and Installation of Driven H-Piles Foundation, Report SL-3629, submitted on March 8, 1978.
3.
NIPSCO's Responses to NRC Staff Questions, submitted on July 14, 1978.
4.
Indicator Pile Program, submitted by NIPSCO to NRC on September 26, 1978.
5.
Sopplementary Information on Driven H-Pile Foundation, NIPSCO, December 4, 1978.
6.
Letter, D. 3. Vassallo, NRC, to H. P. Lyle, NIPSCO, June 28, 1979.
7.
Bailly Generating Station, Nuclear 1 Construction Permit, May 1,1974.
- 8. ~ Recuest by the Porter County Chapter of the Izaak Walton League of America, Inc., February 27, 1979.
9.
Ietter, E. M. Shorb, NIPSCO, to D. B. Vassallo, NRC, June 29, 1979.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
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In the Matter of NORTHERN INDIANA PUBLIC Docket No. 50-367
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SERVICE COMPANY (ConstructionPermitExtension)
(Bailly Generating Station,
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Nuclear-1)
)
1 CERTIFICATE OF SERVICE 1
I hereby certify that copies of "NRC STAFF RESPONSE TO BOARD QUESTIONS RE-GARDING THE SHORT P.ILINGS ISSUE" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class or, as indicated by an asterisk, inrough deposit in the Nuclear Regu-latory Commission's internal mail system, this 18th day of August,1980.
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- Herbert Grossman, Esq., Chairman Edward W. Osann, Jr., Esq.
Atomic Safety and Licensing Board Panel Suite 4600 U.S. Nuclear Regulatory Commission One IBM Plaza Washington, D.C.
20555 Chicago, Illinois 60611
- Dr. Richard F. Cole Robert L. Graham, Esq.
Atomic Safety and i.icensing Board Panel One IBM Plaza U.S. Nuclear Regulatory Commission 44th Floor p
Washington, D.C.
- 20555, Chicago, Illinois 60611
- Mr. Glenn 0. Bright George and Anna Grabowski Atomic Safety and Licensing Board Panel 7413 W. 136th Lane U.S. Nuclear Regulatody Commission Cedar Lake, Indiana 46303 Washington, D.C.
20555 Dr. George Schultz Kathleen H. Shea, Esq.
110 California Street Lowenstein, Newman, Reis, Axelrad Michigan City, Indiana 46360 and Toll i
1025 Connecticut Avenue, N.W.
Richard L. Robbins, Esq.
Washington, D.C.
20036 Lake Michigan Federation 53 West Jacksor. Boulevard Robert J. Vollen, Esq.
Chicago, Illinois 60604 C/o BPI 10S North Dearborn Street Chicago, Illinois 60602
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John Van Vranken, Esq., Chief
- Atomic Safety and Licensing l
Northern Region Board Panel
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t Environmental Control Division U.S. Nuclear Regulatory Comission
- i 188 West Randolph Street Washington, D.C.,20555 Chicago, Illinois 60601 i
- Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Clifford Mezo, Acting President
- Docketing and Service Section i.ocal 1010 Office of the Secretary 6
United Steelworkers of America U.S. Nuclear Regulatory Commission l
3703 Euclid Avenue Washington, D.C.
20555 j
East Chicago, Indiana 46312 William H. Eichhorn, Esq.
Eichhorn, Morrow & Eichhorn 5243 Hohman Avenue Hemond, Indiana 46320 Diane B. Cohn, Esq.
f Suite 700 t
2000 P Street, N.W.
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Washington, D.C.
20036 fyl ital u
l M S Steven C. Goldberd t
Counsel for NRC Staff t'
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