ML20091F451
| ML20091F451 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 09/16/1983 |
| From: | Brunner J CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML17198A223 | List:
|
| References | |
| CON-BOX-06, CON-BOX-6, FOIA-84-96 NUDOCS 8406020185 | |
| Download: ML20091F451 (3) | |
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Atomic Safety & Licensing Board U S Nuclear Regulatory Commission.
Washington, DC 20555 Dear ASLB Members Attached hereto is a memorandum of a conversation between J E Brunner and Mr John Donnell regarding the alleged Board Order Violation. Note the date of m'f conversation with Mr Denne11. I have sent a copy of this memorandum to the Region III investigater handling the investigation of the alleged Board Order Violation.
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Interna; REGARI:ING 3 CAP.D ORDER VIOLATION Ccr: esp. den:e W m*='MQ CC MIMiller E
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JWCook SEF 121983 BkMarguglio g.'DiB.3 PROET EHfgERT On September 6, 1983, I reached John Donnell by tel-ephone at the offices of Science Application Incorporated (SAI) in Las Vegas. After identifying cyself as an attorney for Consumers Power C.ompany, I told Donnell that we were conducting an investigation in regard to certain work perfor=ed on the deep-Q electrical duct bank at the Midland jobsite when he was e= ployed there.
I pointed out that the 13C was investigating the~same =atter.
Donnell stated that h3C had contacted hi= twice on the subject. He gave the h7C a written statement, although his me=ory of relevant facts was dim.
Donnell felt that he knew nothing of interest to the hTC.
Donnell could not recall the locatien of the deep-Q duct bank crossing of the freezewall or the work performed at that location. He did not remember talking to anyone on site about this work. According to Donnell, everyone on site knew, in essence, about the April 30, 1982 Board Order, which Donnell understood as requiring Landsman's advance approval of any drilling activities. Donnell did.not know of any intentional violations of this requirement. He suggested that I talk to Bob Sevo, who attended all meetings with h7C. Donnell was not aware of any person, -including Sevo, who knew that the 13C had prohibited excavation under the deep-Q duct bank.
Donnell offered a few spozitaneous comments not relevant to my questions. He pointed out that he departed the site on goed terms with the Company. He said that 3kN had pro =ised him a recon =endation, but added EkM "did him no favors."
I did not question him on this. He also said that he had centacted 23C to renew an e..pley=en application, which the h'RC people apparently mis' interpreted as a discrimination co= plaint.
I thanked Mi ::nnell for speaking to me.
1 Eased en the above discussion, I decided net to pursue a trip to Las Vegas to interview Mr I:nnell in person, pending release cf the IGC's inves-iga:ica i
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OM/0L SERVICE LIST Mr Frank J Xelley, Esq*
Atomic Safety & Licensing Attorney General of the Appeal Board State of Michigan U S Nuclear Regulatory Commission Ms Carole Steinberg, Esq Washington, DC 20555 Assistant Attorney General Environmental Protection Division Mr C R Stephens (3) 720 Law Building Chief, Docketing & Services Lansing, MI 48913 U S Nuclear Regulatory Commission Office of the Secretary Washington, DC 20555 Mr Myron M Cherry, Esq Suite 3700 Ms Mary Sinclair Three First National Plaza 5711 Summerset Street Chicago, IL 60602 Midland, MI 48640 Mr Wendell H Marshall Mr William D Paton, Esq RFD 10 Counsel for the NRC Staff Midland, MI 48640 U S Nuclear Regulatory Commission Washington, DC 20555 Mr Charles Bechhoefer, Esq Atomic Safety & Licensing Atomic Safety & Licensing Board Panel Board Panel U S Nuclear Regulatory Commission U S Nuclear Regulatory Commission i
Washington, DC 20555 Washington, DC 20555 Dr Frederick P Cowan Ms Barbara Stamiris 6152 N Verde Trail 5795 North River Road Apt B-125 Rt 3 Boca Raton, FL 33433 Freeland, MI 48623 Mr Fred Williams Mr Jerry Harbour Isham, Lincoln & Beale Atomic Safety & Licensing 1120 Connecticut Avenue, NW, Suite 325 Board Panel Washington, DC 20036
.U S Nuclear Regulatory Commission Washington, DC 20555 Mr James E Brunner, Esq Mr M I Miller, Esq Consumers Power Company Isham, Lincoln & Beale 212 West Michigan Avenue Three First National Plaza Jackson, MI 49201 52nd Floor Chicago, 11 60602 Mr D F Judd Mr John Demeester, Esq 4
Babcock & Wilcox Dow Chemical Building PO Box 1260 Michigan Division j
Lynchburg, VA 24505 Midland, MI 48640 Mr Steve Gadler, Esq Ms Lynne Bernabei 2120 Carter Avenue Government Accountability Project St Paul, MN 55108 1901 Q Street, NW Washington, DC 20009 i
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aquary 17, 19 In the Matter of i
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Docket Nos. 50-329-OM CONSUMERS POWER COMPANY
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50-330-OM
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50-329-OL (Midland Plant, Units 1
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50-330-OL and 2)
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Charles Bechhoefer, Esq.
Dr. Jerry Harbour Atomic Safety & Licensing Atomic Safety & Licensing Board-Panel soard Panel U.S. Nuclear Regulatory Com-U.S. Nuclear Regulatory Com-mission mission Washington, D. C.
20555 Washington, D. C.
20555 Dr. Frederick P. Cowan 6152 N. Verde Trail Apt. B-125 Boca Raton, Florida 33433
Dear Administrative Judges:
Enclosed is the first installment of Applicant's prepared testimony for the evidentiary hearings beginning February 14.
Pursuant to agreement of the parties and my conversation with Chief Judge Bechhoeffer on Friday, a second installment will be filed on January 24 dealing with Sinclair contention 3 (water hammer) and certain remaining subissues of Stamiris contention 4.
My understanding is that the Staff has been granted an extra week, until February 7, to file i
testimony with respect to the corresponding issues.
One of Applicant's witnesses on Sinclair contention 4 (steam generator corrosion due to pond water chemistry), Cris Hillman, may not be able to be present in Midland during the week of February 14 due to longstanding vacation plans.
Rather than asking Mr. Hillman to cancel his vacation, we have arranged for Mr. Bill Beckman, who as chemistry / Health Physics Superin-tendent at Midland Plant is Mr. Hillman's boss, to stand in for
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Administrative Judges January 17, 1983 Page 2 Mr. Hillman if necessary.
Mr. Beckman's resume is included with Applicant's testimony.
Also included in this package is Applicant's exhibit 29R, which is a redrafted figure showing the locations agreed upon by Applicant and the NRC Staff for Diesel Generator Building crack monitoring, together with an explanation of the frequency and acceptance criteria for such monitoring.
This material is provided with the Board's permission merely for the sake of clarity; no substantive changes in the drawing i
or the explanation of the monitoring program have been made.
(See Tr. 11068-72).
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One of the Attorneys for Consumers Power Company cc Service List Isham, Lincoln & Beale 3 First National Plaza Chicago, Illinois 60602 (312) 558-7500 l
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TESTIMONY OF CRIS HII.I. MAN AND TERRY POSTLEWAIT 83 r " 1 9 p,,, 3 3 ON SINCI. AIR CONTENTION 4 My name is Cris Hillman. I an employed by Consumers Power Company as the Plant Chemical Engineer at the Midland Nuclear Plant. In this job, my responsibilities include management and development of an integrated plant chemistry program. I have a BS in Chemical Engineering from Michigan State University and five years experience in chemical engineering in support of nuclear power plant operations at Palisades and Midland. Further information is contained in my resume in Attachement A.
I believe as a result of this I
training and experience, I as qualified to address'Sinclair Contention 4.
t My name is Terry Postlewait. I an employed by Consumers Power Company as a Staff Engineer in the Design Production Department, Mechanical / Nuclear Section of the Midland Nuclear Plant Project at the Company headquarters in Jackson'. In this job, my responsibilities include such activities as the rev.
if mechanical systems design and support of licensing i
l activities. As required, I interface with Bechtel and Consumers Power Company J
personnel in the resolution of design and construction problems. I have a BS in Mechanical Engineering from the University of Toledo and ten years experience in the design, installation and operation of mechanical equipment.
Eight of these years have been with Consumers Power Company, sesociated with fossil and nuclear power plants. I have been on the Midland Pro' ject specifically since April of 1980. Further information is contained in my 0
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I believe as a result of this training and experience, I am qualified to address Sinclair Contention 4.
Sinclair Contention 4 states:
"The degradation of steam tube integrity due to corrosion induced wastage, cracking, reduction in tube diameter, and vibration induced cracks is a serious unresolved safety problem at the Midland Nuclear Plant. It is admitted that the chemistry of the cooling water is critical to prevention of steam tube failure, (NUREG-0886). However, the fact that these plants depend on cooling water from the cooling pond increases the likelihood of corrosion and poor water chemistry because the DEIS states that the plant dewatering system will first be discharged to the cooling pond. (DEIS at 5-2).
That means that mary wastes, including radioactive materials-from leaks and spills on the reactor site, can enter the cooling pond and disrupt the chemistry of the pond. Therefore, due to 4
this contribution of an undetermined amount and quality of ground dewatering inflows to the cooling pond, the NRC's bland assurance that corrosion is unlikely due to the lack of sodium thiosulfate, is j
unsatisfactory.
(NRC Response to Interrogatory 9.J.)
In fact, due to the contribution of groundwater, the NRC is not fully aware of the likely constituetts of the cooling pond, and the findings required by 10 CFR SS 50.57(a)(3)(i) and 50.57(a)(6) cannot be made."
I.
Introduction Sinclair Contention 4 is based on the faulty premise that the control of cooling (pond) water chemistry is critical to the prevention of steam generator tube failure (inaccurately deduced from
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misinterpretation of NUREG 0886). The fact is, NUREG 0886 refers to the control of secondary water chemistry, not cooling water (pond) chemistry, as being a key to minimizing steam generator tube degradation. As can be seen from the simplified sketch of the Midland i
Plant thermal cycle shown in Attachment C, cooling water from the cooling pond does not flow through the secondary system or the steam generators during normal operation. Regardless of what the cooling water chemistry may be, administrative procedures and design features of l
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3 the plant are adequate to meet B&W specified secondary water chemistry limits and thus, steam generator tube corrosion will be minimized.
The water normally used in the secondary system is supplied by the Plant Makeup Demineralizer System and does not come, directly or indirectly, from the cooling pond. There are only two circumstances under which cooling water from the cooling pond can be introduced into the secondary system and come into contact with the steam generators.
The first is leakage through the condenser. The second is emergency operation of the. auxiliary _itedwater system: taking water fra= th= paad These are discussed below.
II. Description of How Coolina (Pond) Water Can Enter the Secondary Cycle A.
Condenser Leaks Attachment C includes a simplified schematic cross section of a typical condenser. Each Midland Unit utilizes such a condenser to condense the turbine exhaust steam for reuse in the secondary cycle. The condenser is basically a shell and tube type heat exchanger with the turbine exhaust steam being condensed on the shell side and the cooling water being pumped through the tubes.
The cooling water flowpath is physically separated from that of the steam by means of the tubes and the tubesheets. Each tube-to-tubesheet joint is made watertight by mechanically expanding the tube into the corresponding' hole in the tubesheet.
Cooling water can find its way into the secondary cycle via the condenser if the tube-to-tubesheet joint or the tube itself I
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should develop j leak. Should this occur, the solution is to reroll the tube-to6tubesheet joint or to insert a mechanical plug in each end of the tube. These techniques are standard in the industry and are quite effective.
In-leakage of cooling water to the secondary system is norcally indicated by on-line-instrumentation. The sample point used to detect a suspected condenser leak is located in the condensate pump discharge piping as noted in Attachment C.
The
. condensate pump discharge sample is continuously monitored for cation conductivity, pH, dissolded oxygen, sodium and silica. A
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cooling water-to-secondary leak will normally be indicated by an increase in sodium and/or cation conductivity level. Grab samples for cation conductivity and sodium are also taken from the same sample point and analyzed in the laboratory once per shift. This serves as a check of the on-line instrumentation and is used to i
either coniirm or reject the suspected condenser leak.
i In addition to the methods and practices described above, each condenser is equipped with a hotwell* sampling system. The i
f system collects condensation and/or leakage from each tubesheet.
l Samples are thou collected and analyzed in the laboratory on an as-needed basis. This system is useful in identifying which i
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- The term "hotwell" is standard in the condenser industry and denotes the bottom section of the condenser in which the condensed steam (condensate)is collected prior to being pumped through the domineralizers to the steam generators.
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1 Procedures are now in place that require management I
personnel be informed of off-normal chemistry conditions which includes suspected condenser leaks. Additionally, Technical Specification 16.6.8.4, Ites C, Subsection vii has been proposed
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which concerns cooling water-to-secondary leakage. This proposed Technical Specification states: "When condenser in-leakage is confirmed, the leak shall be repaired, plugged or isolated within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />."
Once a leaking portion of'the condenser is identified, that half of the condenser can be isolated by shutting off one 1
l cooling water loop. The specific leaking element (eg, tube or tube-to-tubesheet joint)can then be further identified by means of portable leak detection equipment, and the leak corrected by plugging or rerolling the tube-to-tubesheet joint.
B.
Auxiliary Feedwater System The second path for potentially introducing cooling pond water into the secondary system is by way of the Auxiliary Feedwater TAW) System. The AW System, initiated by an Auxiliary Feedwater Actuation Signal (AFWAS), is used to supply water for shutdown to
,j the secondary side of the steam generators in the event main feedwater is lost. During Plant operation, the AFW pumps are i
aligned to take suction from the condensate storage tank, in the event AFW is needed. Should an AFWAS occur, condensate from the condensate storage tank would be pamped to the secondary side of the steam generators. In the unlikely event that this supply of water l
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'.to the'AW pomp suctica should be lost for a specified period of s.
time and an AFWAS be present,= suction is automatically transfereed i
to the service water system, which is a source of cooling pond water. In either of these cases, AW is used only to shut the plant 4
down, not to continue operating. The frequency of occurrence of the coincidental set of circumstances leading to the use of cooling pond water as a source of AW is extremely low. The plant can be safely shut down in this scenario of using cooling pond water in the steam I
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- 4 generator; prior to restarting the unit, necessary actions, inspections, etc,would be performed to ensure integrity of the steam j
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referredtoas" condensate,"isamixdureofcondensatefromthe I
.t secondary system and desineralized makeup water from the Plant i
Makeup Demineralizer System and is essentially the same quality as i
' secondary system water. The Plant hhkehp Demineralizer System takes j
s its normal suction from the City of Midland, water system but can use
- Dow Chemical Company's desineralized water as a backup source. In S
either case, the source water is processed th ergh the plant makeup i
desineralizers to assure proper qualitl prL to being puniped to the condensate ~ storage tank. The< cool.5,yer ng a source for s -
domineralized or makeup water.
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i III. Description of __ Secondary Water Chemistrv Control 3
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j The secondary system of the plant is. a wateri system which Loes s
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l maximum limits of these and other parameters are specified by B&W. The limits to be met by CP Co are in all cases the same or more stringent
, than the B&W recommendation.
The secondary systems of both Unit I and Unit 2 are equipped with full-flow deep bed condensate demineralizers. There are six condensate demineralizers in the system, five in service and one in standby. The demineralizers remove cationic and anionic impurities that exist in the condensate. The demineralizers are located downstream of the condensate pumps and have adequate capacity to purify the feedwater prior to its introduction to the steam generators. Thus, impurities which may enter the secondary system via a condenser leak are adequately removed prior to the condensate being sent to the steam generators.
Conclusions Cooling pond water quality itself is not relevant to steam generator tube corrosion. This is because the cooling pond water does not flow through the steam generators under normal operating conditions. In the event that condenser in-leakage were to occur, the Midland Plant's design features and i
procedures are adequate to meet the B&W minimum standards for secondary system water quality and thus, minimize steam generator tube corrosion.
In the unlikely event that cociing pond water were 'o be introduced into the steam generators by means of emergency actuation of the AFW system, the plant can be safely shut down. Appropriate actions would be taken prior l-to restarting the unit to ensure the integrity.of the steam generators.
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Thus, there is reasonable assurance that corrosion of the steam generator tubes does not depend on cooling pond water chemistry; therefore, the NRC can make the findings required by 10 CFR 50.57(a)(3)(i) and 50.57(a)(6).
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ATTACHMINT A Res..e of Cris Hillman Education and Training 1977 Bachelor of Science in Chemical Engineering: Michigan State University 197'9 AICHE Today Series; Industrial Water Conditioning - 1.4 CEU credits Nuclear Steam Sepply Technical Training, Eatcock & Wilcox -
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Mechanalysis Ccurse (Vibration Testing), IRD - 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 19S0 Principles of Supervision, Consumers Power Company - 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> 1981 Radiochemistry for S-pervisors, Babcock & Wilcox - 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> Ht:an Aspects of Management, Conseners Pcwer Company -
20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 19S2 Westinghcuse PVR Chemistry Course, Westinghouse - 5 weeks Ex;erience
'1977-1978 Graduate Engineer, Palisades Nuclear Plact, Consumers Power Comp ay.
Verhed on various projects during refueling outage.
1978 Associate Ingineer, Palisades Nuclear Plant. Worked for Plant Chemical Engineer (Palisades). Duties included support of demineralizer cperations and operator training on makeup dem'1neralizer.
1978-1980 Associate Ingineer, Midland Project Testing. Responsible for development and implementation of acceptance tcst procedures.
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1960 Associate Engineer, B C Cobb Steam Plant. Trouble shoot and increase reliability of new denineralizer system.
Associate Engineer, Midland Project Testing. Responsible for overall preparation and cocrdination of startup chemistry program including the writing of a startup chemistry manual.
Responsible for startup and operation of plant makeup demineralizer system.
1990-Present Chemical Engineer, Midland Nuclear Plant. Respo'nsible for directing plant chemistry staff activities such as startup chemistry support, budget preparation, chemistry procedure preparation and review, and technical support for chemistry related activities.
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ATTACF.ENT B Resume of Terry Postlevait Fnsition Staff' Engineer, Midland Nuclear Plant P.-oject, Design Productics repartment, Mechanical / Nuclear Section Etacation University of Tolado, 1970, BSME i
Eggerience
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Apr-? resent Jcised the Midland Project as a Senior Engineer in nhe Design 1980 Production Department. Duties include assisassee to Safety &.
Eicensing Departmett, review of design of me:hanical systens, resolutica of p:etlees, comnenting en develornent of specifi-cations, procedures and drawings and as-needed assistance to engineering and field personnel.
In June of 1932, preccted to Staff Engineer and involved with additional duties such as coordinating Design Production Department assistacee to field activities involving completion / turnover of syst us, pi..ici-pation in pregram to control design changes and assista..we to Safety & Licensing Oepartnent in devele;ing and aininistering the Spatial E stens Interactics Program and in efforts to ob-f
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tain NF0ES Permit and finalize the Midland Plant I:virennental Statement.
Aug - Ap r Consumers Pever Company 1977 1980 Senior Encincer - Responsible for reviewing dr. sign of new, and modification to existing, power plants. Pricarily invcived with review of r.echanical design of Midland Units 1 & 2.
In Decer.ber 1977, also assured the duties of Project Engineer for ecepletics of D E Karn Plant, Units 3 & 4 and codification work on Karn Unit 4 cold reheat piping to selve excessive noise and vibration problem.
Feb - Aug Alyeska Pipeline Service Company 1975 1977 Rotatint Equi; rent Engineer - Assigned to the tanker loading 1
terminal for the Trans-Alaskan pipeline._ Duties included reviewing sechanical design, mord.toring the contractor to ensure proper installation and alignment of rotating equipment, startup problem assistance and field design changes.
1971 - 1975 Consumers Power Company Associate & General Engineer - Duties included review-cf mechacical design for new nuclear and fossil generating units including specifications, bids and bid evaluatiens, QA requirements, witnessing performance tests a:d preoperational test reviews. Was primarily associated with review of e.uchanical design of D E Karn Plant 3 & 4 i
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2 1965 - 1971 Surface Combustion Division, Midland Ross Corporation Technician..nd Design Engineer - Worked 1965 - 1970 as mechanical lab technician i- (avelopnent of steel and glass pelletizing processes, industrial heat treating furnaces and burners. Upon obtaining cy 33.'2 degree in 1970, worked as design engineer responsible for design and drafting on heat treating furnaces.
1965 Johns-Manville T.sborate-v Technician - Worked in R&D sectica developing fiberglass products and related manufacturi:g equipmen.
1964 - 1965 Stauffer Chemical Company Chemical Laboratorv Technician Performed qualitative analyses for on-line chenical processes.
1963 - 1964 Toledo-Beaver Tools E_,cuipment 0 aerator - Operated latbes, drill presses and mills in the production of pipe cu'. ting and threading tools.
Additional Training June 1977 Attended training on nachinery vibration detection, analysis and balancing given to IRD 5ech.nalysis, Inc
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June - Dec Atter.ded five courses on P:sgrar:natic Qaality Assurance
,,1978 1982 Training for the Midland Project June 1978 Attended Technical Seminar on machinery vibration given by tiechanical Technolog, Inc May 1979 Attended Seminar ou piping design given by Teledyne Engineering Jan 1981 Attended one week training course on the B&*=* simulator J
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RESUME OF WILLIAM BECK!!AN l' ? g-v:,.,,..,, ~,,.,,
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Education and Training
~~.V 1967:
Bachelor of Science in Chemistry:
Fairfield University, Fairfield, Connecticut
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Naval Officer Candidate School, Newport, Mo'de! 19 fl0 &
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Island U.S. Navy - 6 months Naval Nuclear Power School, Bainbridge, Maryland.-.
1972:
6 monthe Nuclear Power Training Unit, SIC Prototype, j
Windsor, Connecticut U.S. Navy - 6 months 1973-1975:
Various Service Schools 1976:
QA Indoctrination Workshop, Consumers Power Company - 3 days
~~
1977:
Radiochemistry for Supervisors, Babcock & Wilcox -
-10 days Principles of Supervision, Consumers Power
)
Company - 5 days 39 1978:
Busan Aspects of Management, Consumers Power Company - 5 days s
i 1979:
Principles of Data Processir g and FORTRAN Programming, Delta College - 4 credits Public Presentation Skills Course, Central Michigan University - 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> Principles of Leadership Effectiveness, Consumers Power Company - 3 days 1980:
I Techniques of Supervision, Consumers Power company - 5 days Managerial Economics, Consumers Power Company -
~
5 days i
1981:
Annual Short course on Radiation Protection, University of Michigan - 10 days selected Topics in Reactor Health Physics, HP society, 1981 summer school, University of Kentucky - 5 days Effective Management, Consumers Power Company -
4 days Revision 39 j
13A.2-12 11/81 l
i
, - ~..-
^
nwLanhU Led-t khK Experience
)
1967-1971:
Senior Sales Engineer, Stero Polymers, Goodyear Chemical Division, Goodyear Tire and Rubber
? -
1971-1972:
Attended various Service Schools, U.S. Navy 1972-1973:
Qualified Engineering Officer of the Watch on Sic reactor plant, Windsor, Connecticut 39 1973-1975:
Engineering Officer of the Watch, responsible for watch supervision and overall safe operation of SSW reactor plant - 32 months 1973-1974:
Main Propulsion Assistant and Radiologi~ cal Controls Officer, USS John C. Calhoun SSBN 630B.
Duties included supervision of nuclear plant mechanical maintenance radiation protection, chemistry, and radiochemistry - 24 months 1974-1975:
i Electrical Officer and Sonar Officer, USS John C.
Calhoun SSBN 630B. Duties included supervision of
~
nuclear plant electrical maintenance - 6 months l43 1975-1976:
General Engineer, Chemistry Department, Palisades Nuclear Plant (CE PWR), Consumers Power Company.
j Duties included design and implementation of plant modifications and ' preparation of chemistry j
3 procedures - 9 months i
1976:
General Office, Palisades Nuclear Plant, coordinated and scheduled refueling outage -
6 months 39 1976-1977:
General Engineer, Midland Nuclear Plant.
Duties i
included design reviews of plant systems with i
4 respect to chemistry, environmental, and radiological considerations 1977-1979:
Chemical Engineer, Midland Nuclear Plant.
Responsible for development of the i
Chemistry / Radiochemistry Department.
Duties expanded for 9 months in 1977 to include i
supervision of the environmental and radiation protection programs i
l l
1980-Present: Chemistry / Health Physics Superintendent, Midland Nuclear Plant 143 l39 i
f Revision 43 13A.2-13 4/82 I
I
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'N NUCLEAR REGULATORY COMMISSION,p CO UNITED STATES OF AMERICA....
BEFORE THE
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i ATOMIC SAFETY AND LICENSING BOARD it19 P053 DocketNos.Sh3WOM
,o In the Matter of
)
50-330 OM CONSUMERS POWER COMPANY
)
(Midland Plant, Units 1 & 2))
Docket Nos. 50-329 OL 50-330 OL 4
TESTIMONY OF DR. RICHARD D. WOODS ON BEHALF OF THE APPLICANT REGARDING SEISMIC SHAKEDONN SETTLEMENT AT THE MIDLAND SITE (EXCEPT DIESEL GENERATOR BUILDING)
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STATE OF MICHIGAN COUNTY OF WASHTENAW UNITED STATES OF AMERICA
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NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
Docket Nos. 50-329 OM
)
50-330 CONSUMERS POWER COMPANY
)
(Midland Plant, Units 1 & 2))
Docket Nos. 50-329 OL 50-329 OL AFFIDAVIT OF RICHARD D. WOODS Richard D. Woods being duly sworn, deposes and says that he is the author of " Testimony of Richard D. Woods concerning Seismic Shakedown Settlement at the Midland Site except Diesel Generator Building,' and that such testimony is true and accurate to the best of his knowledge and belief.
AA As RICHARD D. WOODS A
Sworn and Subscribed Before Me this d Day of 1,a u.
, 1983 i
o im. t 5 d Notary Public Washtenaw County, Michigan l
Mar.T 'A. smess, 7. : e.ry, ma,
.~
warrmaw coven. excazain i
at centsszon so.eas io-u.se
, My commission Expires t
~
1.0 ' BIOGRAPHICAL INFORMATION This is the testimony of Dr. Richard D. Woods.
My.
detailed resume is attached.
The following is a summary of that resume.
I received a Bachelor of Science degree in Civil Engineering from Notre Dame University in 1957 and a Master of Science degree from the same school in 1962.
I worked for the Air Force Weapons Center, Albuquerque, New Mexico, on the design of blast resistant 1
underground structures for one year and taught in the Civil Engineering Department at Michigan Technological University ~for one year before. going to the University of Michigan for a Ph.D. in Civil Engineering, which I received in 1967.
Since then I have been on the faculty of the Department of Civil Engi~neering at the University of Michigan, advancing to full Professor in 1976.
My research interests have been in the field of soil dynamics and earthquake engineering.
I have done part-time consulting in the fields of soil dynamics, earthquake engineering, structural vibrations, and general foundation i
engi,neering.
My clients have included Bechtel, Corning Glass Works, Rockwell. International, Eaton Corporation, TAMS, General Motors, Honeywell Inc., Woodward-Clyde Consultants, and Nucien (Nuclear Brazil).
I have directed I
research associated with dynamic soil properties and foundation vibrations.
I am a principal.in the foundation i
i consulting firm of Stoll, Evans, Woods, and Assoicates, Ann Arbor, Michigan and am a member of ASCE, ASEE, ASTM, and SSA.
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2.0 INTRODUCTION
My testimony is concerned with the evaluation of the potential for shakedown settlement of loose sands in the plant area at Midland (except the Diesel Generator Building).
The shakedown settlement was evaluated using a method based on blow count and results of an experimental study on the behavior of sands under seismic loading by Silver and Seud (1969).
The maximum ground acceleration was assumed to be 0.19g and 10 cycles of shearing strain J
i reversal were considered.
On the basis of my analysis and the proposed-remedial-measuree-being t:%;n in --the plant--
area, I have concluded that there is reasonable assurance
[
that the plant area including piping and duct banks will not suffer excessive settlement due to seismic shakedown.
3.0 DISCUSSION When earthquake excitation is a part of the design loads for a construction site, the potential for shakedown must be evaluated.
Shakedown settlement is a phenomenon by l
which loose, clean cohesionless soils densify due to ground shaking.
Soils of this type which have been i
deposited in a loose condition tend to undergo a redistribution of particle packing when shaken until a condition of minimum potential energy is achieved.
The
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redistribution of particle packing causes a reduction in j
the bulk volume of the soil, thereby causing a potential i
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for settlement of the ground surface and..ructures built on the sdrface or buried in the soil mass.
~
9 Whether or not a specific sand formation will undergo shakedown settlement is dependent upon chaiacteristics of the soil and factors associated with the earthquake which causes shaking.
Among the soil characteristics which influence the shakedown behavior are:
grain size distribution, grain shape and relative density.
Uniform grain size, rounded, loose sands are most susceptible to settlement due to shaking.
Sands with wider grain size distributions and with more angular individual grains are less susceptible to shakedown settlement.
Sands with high initial relative densities are less susceptible to
[
shakedown settlement than sands with low initial relative I
densities.
Characteristics of the earthquake which influence the potential for and magnitude of shakedown settlement are the maximum ground acceleration and the number of cycles of shearing strain.
Pockets of sand which have a potential for shakedown settlement exist at several locations at the Midland site.
Some areas occur under or near Category I j
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structures while others are distributed throughout the plant area where pipelines and duct banks are buried.
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,f s 4.0 EVALUATION OF SHAKEDOWN SETTLEMENT i
Silver and Seed (1969) published the results of an experimental study of the settlement of dry sand subject
~
to seismic loading conditions.
rhe results of th'is study are appropriate for a conservative evaluation of shakedown potential because sand in the dry state is most susceptible to shakedown settlement.
If some moisture 4
occurs in the sand, apparent cohesion is present and this reduces the potential for shakedown.
If sufficient water 7
is present in the soil, the danger becomes that of liquefaction potential not shakedown and liquefaction l
potential has been addressed in other testimony.
j To make use of the Silver and Seed (1969) study, the shear i
stress in the sand pocket under investigation due to the SSE is estimated from an equation based on Se'ed and Idriss (1971):
r = 0.65 amax o (1) y I
g in which:
T = shear stress max = maximum acceleration associated with earthquake a
i g = acceleration of gravity av = total. vertical stress 1
4
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Then a trial shear modulus is selected based on an assumed shearing strain and relative density from the curves proposed by Seed and Idriss (1970), Figure 1.
l The shear strain is then computed from:
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Y -
T (2) r i
G in which:
Y = shear strain T
shear stress
=
4 G = shear modulus i
succesive corrections are applied to the trial G until the shear strain for which G was selected and the shear strain from eq (2) are in reasonable agreement.
The relative density of the sand pocket is estimated from standard penetrition blow counts.
Using relative density and the magnitude of shear strain for which agreement was found l
above, the vertical strain due to 10 cycles of loading is l
l estimated from the Silver and Seed curves, Figure 2.
l i
Shear strain from Figure 2 is then multiplied by the l
thickness of the deposit to obtain an estimate of the 4
shakedown settlement due to one-dimensional shaking.
This settlement is multipled by three to obtain a conservative i
estimate of three dimensional shaking as suggested by Pyke l
et al (1975).
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The settlement of all pockets occuring along any vertical profile and below any category I structure conduit or pipe are summed up to estimate the local shakedown settlement.
5.0 RESULTS OF SHAKEDOWN SETTLEMENT INVESTIGATION Sands for which there is a potential for shakedown 4
settlement occur in only five areas for this testimony (one additional area, the diesel generator building, is covered by separate testimony).
These areas are shown on l
Figure 3 and are categorized as :
Borated water storage tank arear-railreed bey-asea..ef the s'3xiliery build 4ne 3
diesel fuel storage tankas underground piping areas; and retaining walls area.
l 5.1 Borated Water Storage Tanks No potential exists for shakedown settlement under the I
borated water storage tanks because the soil under these tanks is clay.
Furthermore, the sand within the ring foundation has been compacted to a relative density greater than 80% for which no significant shakedown i
settlement will occur.
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r' 5.2 Railroad Bay Three borings in the railroad bay area of the auxiliary building show pockets of sand.
The maximum settlement due to shakedown was estimated to be 0.25 inch.
The maximum differential settlement also would be aoout 0.25 inch because some portions of the same building are founded on till which will not settle due to shakedown.
l 5.3 Diesel Fuel Storage Tanks j
i One boring in the diesel fuel storage tank area showed f
pockets of sand.
The maximum shakedown settlement which would occur based on that boring amounts to about 0.10 i
inch, and relative to a point which does not settle at all 5
amounts to a differential settlement of the same i
magnitude.
These shakedown settlements present no hazard i
l to the diesel fuel storage tanks.
i l
5.4 Underground Piping and Conduits
- i An inspection of the botings throughout the regions where underground piping and duct banks are buried shows that i
i the worst situation, i.e. thickest sand deposits, occur 1
near the SWPS.
Remedial measures are planned for this I
area which call for removal of loose material to elevation 610 and replacement with suitable material.
The potential l
for shakedown settlement below elevation 610 near the SWPS l
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is small because of limited thickness of loose sand.
Category I piping and duct banks in other parts of the site have been evaluated for shakedown settlement by i
studying the boring logs near and under these, lines.
By comparison with areas for which shakedown settlement was computed for other structures, it was concluded that the areas under piping and duct banks will experience shakedown settlements of no more than 1/4 inch.
5.5 Retaining Walls Category I retaining walls are located northeast of the
)
SWPS.
Two other non-category I retaining walls are located south of the CWIS.
Foundations for these retaining walls are located at elevations 595 and 611.
Only loose sand below foundation levels are of concern for i
shakedown settlement and twelve borings in the region of the retaining walls showed that there is no loose sand 1
under the retaining wall foundations.
Shakedown settlement for these structures will be negligible.
l l
6.0
SUMMARY
AND CONCLUSION f
Limited pockets of loose natural sand and loose fill sand exist in the plant area and under the railroad bay of the 1
auxiliary building.
The potential for and magnitude of earthquake shakedown settlement of these sands has been
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T evaluated.
An earthquake with a maximum acceleration of 0.199 and 10 cycles of shear strain has been used in this evaluation.
In some areas near the SWPS remedial measures will eliminate the potential for shakedown settlement.
For loose sand pockets in other areas, the magnitude of shakedown settlement has been estimated and found to be 1/4 inch or less.
4 For an SSE of.129 the shakedown settlement would be about 50% of that reported here. -' ~^
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7.0 REF ERENCES j
- 1) Pyke, R; Chan, C.K. and Seed, H.B.
(1974), " Settlement and Liquefaction of Sands Under Multi-Directional Shaking," Report No. EERC 74-2, Earthquake Engineering Research Center, University of California, Berkeley.
February.
- 2) Seed, H.B. and Idriss, I.M.
(1970), " Soil Modulus and Damping Factors f or Dynamic Response Analysis,'
Earthquake Engineering Research Center, University of California, Berkeley, December.
- 3) Seed, H.B. and Idriss, I.M.
(1971), ' Simplified Procedure for Evaluating Soil Liquefaction Potential,"
Journal of the Soil Mechanics and Foundation Division, Proceedings ASCE, Vol. 95, No. SM9, Sept., pp.
1249-1272.
- 4) Silver, M.L. and Seed, H.B.
(1969), "The Behavior of Sands Under seismic Loading Conditions," Report No.
EERC 69-16, Earthquake Engineering Research Center, University of California, Berkely, December.,
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RICHARD D. WOODS, Ph.D., P.E.
Professor of Civil Enginea %.
University of Michigan e
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I RfSUMf RICHARD D. WOODS, Ph.D., P.E.
Professor of Civil Engineering University of Michigan l.
I August, 1980 Home 700 Mt. Pleasant Ann Arbor, MI 48103 (313) 769-4352 l
Office 2322 G. G. Brown Lab University of Michigan i
Ann Arbor, MI 48109 (313) 764-4303 i -
i l
PERSONAL DATA l
Age:
45, born U.S. citizen Physical:
Height 6'; weight 220 lb i
Health:
Excellent i
Military:
U.S. Marines Married:
Wife, Dixie Lee (Davis) i j
Daughter, Kathleen Ann, age 2 3 Daughter, Cecilia Marie, age 15 Daughter, Karen Teresa, age 12 EDUCATION I
High School, J. W. Sexton, Lansing, Michigan, 1953
{
3.5. Civil Engineering, University of Notre Dame, 1957 4
M.S. Civil Engineering, University of Notre Dame, 1962 4
Introductory (non-degree) Course, ASEE-AEC Basic Institute in Nuclear Engineering, North Carolina t
State College, 1964 i
l Ph.D. Civil Engineering, University of Michigan, 1967 i
l i
l
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Richird D. Wecas'. Jh.D., P.E.
Pcg3 2 s
e, ORGANI ATIONff 1
American Society of Civil Engineers
,American Society for Testing and Materials Aper!can Society for Engineering Education
, chi Epsilon Society of the Sigma Xi seismological Society of America AWARD Collingwood Prize of American Society of Civil Engineers, 1969 EMPLOYMNT (Full Time) 1976 to Prefessor, Civi1=Zngineering, University of Michigan.
i Present Doursas e,sught Basic Soil Mechanics, Field Sampling and Laboratory Testing of Soils, Foundation Engineer-4 ing, Soil cynamics, Civil Engineering Dynamics 3
Measurements, Plane Surveying, Statics and Strength of Materials, Reinforced Concrete.
Research performed:
see separate paragraph below.
1971 Associate Professor, Civil Engineering, University to of Michigan.
Courses taught:
Included above.
1976 1967 Assistant Professor, Civil Engineering, University l
to of Michigan.
Courses taught:
Included above.
1971 l'
1965 Graduate Student, University of Michigan, supported 1.
to on NSF Traineeship.
1967 e
1964 Instructor, Civil Engineering, Michigan Techno-logical University, Houghton, Michigan.
Courses
' taught:
Included abave.
1963
. Pro $ect Engineer (GS-11), Air Force Weapons Labora-tory, Kirtland, AFB, Albuquerque, N.M.
Supervised contracts which were directed at determining engineering properties of soila under dynamic loads.
1960 Graducte student, University of Notre Dame, teaching to assistantshap, taught surveying camp.
1962 1957 Lieutenant, U.s. Marine Corps, Camp Pendleton.
i to California.
Six months as platoon leader, movable 1960 bridge company.
Remainder of service as hydraulic engineering officer preparing evidence for water rights litigation.
3-f
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Richard D. Woods, sh.D., P.E.
Pcgo 3 i
EMPLOYMENT (Short Courses and Special Appointments) i Fuero Fellow, University of Florida.
On sabbatical 1976 leave from University of Michigan.
Investigating use of static cone penetrometer with built-in pore pressure transducer to predict liquifaction*
i potential of sands.
t l
1974 Invited Author for Chapter on soil Dynamics for U.S. Army Corps of Engineers Soils Manual, with F. E. Richart.
1973 Invited Lecturer, Woodward-Clyde Consultants i
Symposium, Berkeley.
Topic:
" Seismic Methods to j
Measure Shear Wave velocity of Soils and Rock."
1973 Taucht Extension Courses (evening), " Applications 1972 of Soil Mechanics to Foundation Engineering,"
1 2-10 week lecture series for Commonwealth Associates, 1
Jackson, Michigan.
5 1972 Visiting Professor, Institute for soil and Rock Mechanics, University of Karlsruhe, Germany.
Taught Soil Dynamics and helped establish soil dynamics l'aboratory.
Research on propagation of Rayleigh Waves in region of obstacles.
l 1971 Visiting Professor, Indian Institute of Technology, i
Kanpur, India.
Helped establish basic soil dynamics laboratory and ' field measurements capability.
4 i
1971 Invited Lecturer, Earthquake Engineering Seminar, l
University of Massachusetts, sponsored by National J
Science Foundation.
Lectures on basic vibrations, wave propagation and dynamic soil properties.
l 1970 C:1 airman and Princicah Lecturer, two 2-day 1969 s sort courses, "Behav:,or of Soils for the Con-l struction Industry, Continuing Engineering Education Program, College of Engineering, Uni-versity of Michigan.
i 1968 Co-Chairman and Lecturer, Two-week short course,
" Vibration of Soils and Foundations," Continuing l
i Engineering Education program, College of Engineer-l ing, University of Michigan.
Lectures on basic vibrations, wave propagation and field and labora-tory measurements.
4
+
s
. oods, Ph.D., P.E.
Page 4 Richard D.
W 9
RESEARCH At University of Michigan Holocraphie Interferometry - Investigation of basic l
~ wave propagation and surface wave propagation in
. region of' barriers.,
Response of Pilk Fou'ndation's to Dynamic Loads -
with F. 2. Richart.
Dynamic Properties of Soils Laboratory and field measurement of compression and shear wave Velocity
~
and shear modul's of soils at both low and high u
amplitudes.
Isolation of Earthwaves by~ Barriers - Study of effectiveness of trenches and dy'lindrical holes at screening waves.
Dutch Static Cone Penetrometer - Study of use of penetrometer for identificatlon of scils.
AtMichicanEachnologicalUniverskty 1
Mechanics of Slide Dams - Investigation of creation o,f nams by blasting material'from canyon walls.
At Notre Dame University Preliminarv Design of Dvnamic Direct Shear Device i
CONSULTING EXPERIENCE Areas of Consulting vibration Measur'ements" 'on machines, in soil,'on structures Measurement of Dynamic Soil Properties, in lab and i
in field Stability of Soil, Masses (Resar've" Mihing tailings delta)
Analysis and' Design of foundations "for dynamic loads Site Investigations with Dutchi cene penetrometer Blasting Damage Evaluations Blasting Code Drafting Seismic Site Investigations Principal Clients Bechtel Power Corporation, Ann Arbor, Michigan Attorney General, State of Michigan (Reserve Mining Case) i f
.i
~ >.. -
Richard D. Woods,
.h.D.,
P.E.
Pago 5 CONSULTING EXPERIENCE--Continued Giffels,and Associates, Detroit, Michigan Smith, Hinchman and Grylls, Detroit, Michigan City of Rockwood, Michigan City of Ann Arbor, Michigan Honeywell Corporation, Minneapolis, Minnesota Woodward-Clyde Consultants, Orange, California, Oakland, California and Philadelphia, Pennsylvania Halpert, Neyer Associates, Farmington, Michigan U. W. Stoll and Associates, Ann Arbor, Michigan Eaton Brake Division, Detroit, Michigan Tippetts-Abbett-McCarthy-Stratton, New York (Tarbela Dam)
Site Engineers, Inc., Cherry Hill and Montclair, New Jersey Corning Glass Works, Corning, N.Y. and three other plants PUBLICATIONS AND REPORTS Woods, R. D.
(1963), " Preliminary Design of Dynamic-Static Direct Shear Apparatus for Soils and Annotated Bibliographies of Soil Dynamics and Cratering,"
Air Force Weapons Laboratory, RTD-TDR-63-3050.
- Woods, R.
D., Reddy, P. D. and Young, G. A.
(1964), " Study of the Mechanics of Slide Dams with Distorted Models, Progress Report," Contract 74-0030, Sandia Corporation, Albuquerque.
Woods, R. D.
and Richart, F.
E.,
Jr. (1967), " Screening of Elastic Surface Waves by Trenches," Proceeding 4 Symposium on Wave Propagation and Dynamic Properties of Earth Materiais, Albuquerque, N.M.,
August.
Woods, R. D. (1968), " Screening of Surface Waves in Soils,"
J. SMFD, Proc. ASCE, vol. 94, SM 4, July, pp.
951-979.
Richart, F. E., Jr., Hall, J. R., Jr., and Woods,
R. D.
(1970), Vibration 4 of Soiis and Foundation 4, Prentice-Hall, 414 pp.
Afifi, S. S. and Woods, R. D.
(1971), "Long-Term Pressure Effects on Shear Modulus of Soils," J. SMFD, Proc.
ASCE, vol. 97, SM 10, Oct., pp. 1445-1460.
t
Richard D. Woods, Ph.D.,
Page 6 PUBLICATIONS AND REPORTS--Continued Stokoe, K. H. and Woods, R. D.
(1972), "In Situ Shear Wave Velocity by Cross-Hole Method," J. SMFC, Proc. ASCE, Vol. 98, SM 5, May, pp. 443-460.
Woods'., R. D. and Sagesser, R.
(1973), " Holographic Inter-ferometry in Soil Dynamics," Proceedings of the Eighth international Conference on Soit Mechanics and Foundation Engineering, Moscow, August, Vol. 1, Part 2, pp. 481-486.
- Woods, R.
D.,
Barnett, N.
E.,
and Sagesser, R.
(1974),
" Holography--A New Tool for Soil Dynamics,"
J. GTD, Proc. ASCE, Vol. 100, No. GTil, Nov.,
pp. 1231-1247.
Anderson, D. G. and Woods, R. D.
(1975), " Comparison of Field and Laboratory Shear Moduli," Proceeding 4 of Conf. on in Sita Measurement of Soit Properties, Raleigh, North Carolina, Vol. 1, June, pp. 69-92.
Anderson, D. G. and Woods, R. D. (1976), " Time-Dependent Increase in Shear Modulus of Clay," J. GTD, Proc.
ASCE, Vol. 102, No. GTS, May.
Woods, R. D.
(1976), " Foundation Dynamics," Applied Mechanics Reviews,. Proc. ASME, Sept.
Woods, R. D.
(1977), " Parameters Affecting Dynamic Elastic Properties of Soils," Proceedings of the International Symposium on Dynamical Methods in Soil and Rock Mech-anics, Karlsruhe (F.R. Germany), September, Sponsored by NATO Scientific Affairs Division and the Institute of Soil Mechanics and Rock Mechanics, University of Karlsruhe.
Woods, R. D.
(1977), " Lumped Parameter Models for Dynamics Footing Response, Karlspuhe (as above).
Woods, R. D.
(1977), " Holographic Interferometry to Study Se.ismic Wave Isolation," Karlsruhe (as above).
Woods, R.D.
(1978),
" Measurement of Dynamic Soil Properties,"
Proceedings of the ASCE Geotechnical Engineering Division Specialty Conference, EARTHQUAKE ENGINEERING AND SOIL DYNAMICS, June 19-21, Pasadena, CA., Vol. 1, pp 91-178.
- Richart, F.E.,
Jr., and R. D. Woods (1978), "Foundstions for Auto Shredders."
Presented at the 1978 Fall Convention, American Concrete Institute, Houston, Oct. 29-Nov. 3.
- Allen, N.F.,
- Richart, F.E., Jr., and Woods, R.D.
t1980), " Fluid Wave Propagation in Saturated and Nearly Saturated Sands."
Journal gf Geotechnical Eneineering _Divisien, ASCE, -
Vol. 106, No. GT 3, March, pp 235-254.
__ :_i
aAunnra u Woodo, Ph.D.
p Page 7 PUBLICATIONS Continued Woods, R.D. and Partos, A (1981),
\\
Improvement by Crosshole Testing," 2J22" Control of Soil Tenth M. lat. Csg. g.th
. M 3h1 and Found. M.s Inter. J,gg.,g..gg M pp. 79 M 96, June.
, Stockholm, Sweden, Vol. 3, Woods, R.D..and Henke, R.
(1981), " Seismic Techniques No. GT 10, Oct.in the I,aboratory," g,. STE yrgq. JyjgE
,, Vol. 107, Partos,A.. Woods, R.D. and Welsh, J. (1982), " Soil Modification for Relocating Die Forging Operation "
International Symposium on Grouting Engineering, New Orleans, Feb.
in Geotechnical Richart, F.E. Jr., and Woods, R.D.
conference gn Jg Q Dynamicsfor Auto Shredders" P (1982), "Foundaticns serine, Southam,p, ton England, W Eartheuake Encin-pp.811-824.
July 13-15, Vol. 2, e
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a MIDLAND NUCLEAR PLANT DIESEL GENERATOR BUILDING AREAS FOR CRACK WIDTH MONITORING DURING THE OPERATION OF THE PLANT At the locations indicated by sketch, Applicant Exhibit No 29R, the applicant will monitor widths of cracks along a horizontal line as shown.
This horizontal line will be picked as close as possible to the existing horizontal reinforcing bar in the near face of the wall.
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Frequency of Monitoring Once every year during operation of the power plant for five years and at five year intervals thereafter.
2.
Acceptance criteria I
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Alert Limits:
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Any one of the crack monitored reaching a crack width of 50 mils.
(0.050 inches) 2.
In 10 feet of gage length, the summation of the increase in crack vidths reaching 150 mils. (0.150 inches) b.
Action Limits:
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Any one of the crack monitored reaching 60 mils.
(0.060 inches) i 2.
In 10 feet of gage length, the summation of increase in crack vidths reaching 200 mils. (0.200 inches) l In measuring the widths of the cracks there will be a tolerance limit of (plus) +5 mils.
(+0.005 inches)
The defin'ition of Alert and Action Limits are similar to those defined in Staff'sTestimony, SSER 2 Page 2 - 48 under Article No 2.5.4.6.1.2 and testimony of Dr Steve Poulos during SWPS hearing.
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NUCLEAR REGULATORY COMMISS, ION N
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
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Docket Nos. 50-329-OM CONSUMERS POWER COMPANY
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50-330-OM
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50-329-OL (Midland Plant, Units 1
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50-330-OL and 2)
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TESTIMONY OF B. JOHN GARRICK AND LOUIS S. GIBSON CONCERNING THE ANALYSIS OF CLASS 9 ACCIDENTS IN THE MIDLAND FES This is the testimony of B. John Garrick and Louis S. Gi' son.
Dr. Garrick is a principal of Pickard, Lowe and Garrick, Inc. and has been retained by consumers Power Company as a consultant in the area of risk analysis for the Midland facility.
Dr. Garrick's resume outlining i
his professional and educational qualifications is to this testimony.
Mr. Gibson is employed by Consumers Power Company as the Section Head of the Nuclear Safety and Analysis Section of the Midland Project Safety and Licensing Depart-ment.
He is responsible for the nuclear safety analysis for the Midland Plant including accident analysis, transient analysis and probabilistic risk assessment.
Mr. Gibson's resume outlining his professional and educational qualifi-cations is Attachment 2 to this testimony.
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l INTRODUCTION AND SCOPE OF TESTIMONY The purpose of this testimony is to address Mary Sinclair's contention 13 (formerly revised new contention 3).
This contention asserts that the Midland FES is inadequate because the NRC staff utilized, in assessing the potential environmental risk associated with severe accidents, the results of a rebaselined Reactor Safety Study ("RSS") analysis performed by the NRC rather than the preliminary results of NUREG/CR-2497 " Precursors to Po*.ential Severe Core Damage Accidents:-1969-1979" published by the Nuclear Regulatory Com-mission (the " Precursor Study").
The assessment of the potential environmental risks resulting from accidents contained in section 5.9.4 of the Midland FES is utilized to satisfy the requirements of the National Environmental Policy Act of 1969 (NEPA) and NRC commission policy (45 FR 40101) for a reasoned consideration of the environmental risks (impacts) attributable to accidents.
The results of the FES probabilistic assessment of severe accidents found in section 5.9.4.5(2) are utilized in section 5.9.'4.6 as one of three justificatiens for the conclusion that the potential environmental impacts from accidents at Midland are small and that there are no special or unique radio-logical circumstances about the Midland site and environs that would warrant special mitigating features for the Midland Plant.
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. The other two reasons are (a) "the fact that considerable experience has been gained with the operation of similar facilities without significant degradation of the environment."
(b) "the fact that in order to obtain'a license to operate the Midland facility, Consumers Power must comply with the applicable Commission regulations and requirements."
The results of the environmental risk analysis are thereaf ter used as an input to the summary cost benefit analysis presented in chapter 6 of the FES.
The balance of this testimony is an evaluation of whether the conclusions regarding the potential environmental impacts from accidents at Midland as stated in FES sections 5.9.4.6 and 6.4.3 remain valid in light of the information contained in the Precursor Study.
DESCRIPTION OF PROBABILIETIC RISK ASSESSMENT AND ITS USE IN NUCLEAR POkTR PLANT ACCIDENT ANALYSIS.
Probabilistic risk assessment (PRA), as this term is used in this testimony, is the quantification of the frequency of occurrence of different levels of damage from accidents in-volving nuclear power plants.
In principle, probabilistic risk analysis admits into consideration any scenario that can be conceived involving any number of simultaneous failures and physical processes.
Component failure rate distributions are determined.
Equipment reliability is calculated.
Physical phenomena are evaluated.
Finally, the probability of any accident with any number of safety system failures can be q
_4 calculated along with the associated release of radioactive material.
Probabilistic risk asses,sment is a logical approach to the evaluation of radioactive hazards from nuclear reactors.
It attempts to quantify the likelihood of each possible release and, thereby, provides a means for comparing the risk of nuclear reactors with other public hazards and for identifying those components of the nuclear reactor system which most strongly affect that risk.
The NRC Staff utilized elements of PRA in Chapter 5 of the FES as one basis for evaluating the environmental effects of severe accidents.
Interest in probabilistic safety analysis has evolved since the 1960s.
Analysis techniques were borrowed from statisticians and reliability engineers and developed into detailed tools for predicting failure probabilities for large, complex power plant systems. In 1972, the U.S.
Atomic Energy Commission undertook the Reactor Safety Study under Professor N. C. Rasmussen of MIT.
This mammoth project i
(70 man-years and $4 million) took 3 years to complete and was definitely a turning point in the way we think about nuclear safety.
It was clearly the most thorough in-t vestigation of. reactor safety ever conducted and produced an enormous body of technical work that will influence safety analysis and understanding for ye'ars to come.
It calculated the risk probabilities and consequences from the operation of 100 current design light water reactors in the United States.
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. The complete report demonstrated that it is possible to methodically analyze results for policymakers and analysts alike.
The finished document formed a' basis for thorough discussions of risk methodology; i.e., a center for criticism, review and improvement.
Though it is a seminal work, initially the Reactor Safety Study was widely criticized.
Between release of a draft report in August, 1974, and the final version in October, 1975, comments were received from 87 organizations and in-dividuals representing government, industry, "public-interest" groups, and universities.
Many of these comments had a signifi-cant impact on the final report.
Congressional hearings on the Reactor Safety Study were held in 1976, " seeking from experts with disparate views assessments of the validity of the Reactor Safety Study's conclusions and the usefulness of the study as an aid to policy-making.
Congress w.inted te know how the study has increased our understanding of nuclear safety and how it might be improved upon."
The study's directors and authors, representatives of many of the groups who had commented on the study, and other knowledgeable and interested people testified.
Many criticized various aspects of the study, particularly its estimates of the s
uncertainty in specific results, but most reaffirmed that it was a valuable contribution to the understanding of nuclear risks.
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, The most complete and even-handed ' review of the Reactor Safety Study was conducted by the Risk Assessment Review Group chaired by Professor H. W. Lewis of the University of California, Santa Barbara.
The group was organized by the NRC on July 1,1977, at the request of Congressman Morris K. Udall, Chairman of the Committee on Interior and Insular Affairs, who had held hearings on the Reactor Safety Study.
It was chartered to evaluate the achievements and limitations of that study, the peer comments on the study, and the state-of-the-art of risk assessment methodology and to advise the NRC on the use of such methods in the regulatory Process.
The Lewis report concluded that the fault tree / event tree methodology was sound.
The authors of that report looked carefully into certain statistical questions and identified several areas where in their judgment there was a lack of mathematical rigor.
On the other hand, the Lewis group agreed that the lack of mathematical rigor would have little effect on actual calculations.
i The Lewis report was a competent technical review of the Reactor Safety Study.
It provided constructive suggentions th,at would enhance future PRAs.
The comments favorable to the Reactor Safety Study, particularly in regard to.the basic l
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While 1
the list of criticisms was substantial, the " bottom line" seemed to be that the methodology was' sound except in the area of the quantification of ti.3 uncertainty of the results.
The observation t
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It is important that neither the Lewis report nor the NRC disavowed the fault tree / event tree methodology; rather, both find the methodology sound and encourage its further use in the regulatory process.
Of course, the Reactor Safety Study, together with its thorough review, especially the Lewis review, provided an excellent background for further development of the PRA methcdology.
Contemporary full scope PRAs such as those performed on the specific Zion and Indian Point plants have further advanced the development of PRA techniques.
The result of these advancements is a much greater insight into the components of risk associated with the' plants investigated.
Perhaps the most important insight provided by the full scope risk studies is ' hat the risk associated with t
nuclear power plant operation is quantifiable and it appears to be small.
It is believed that the major breakthroughs in risk analysis methodology and the ability to identify contributors to risk have already occurred.
The idea now is to apply the methods and fine tune the process.
Most of the advances of the future should be in the quality of information supporting risk analysis, e.g., a better understanding of the physical processes following the onset of a damaged core.
Among the important observations from the major studies i
are the followings i
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Public risk from nuclear power is less than the financial risk to the owner / operator of the plant.
The evidence takes,the form of a demonstration that nuclear plants are much more able to cope with a damaged core or even a meltdown than had been perceived.
Thus, although the possibility of a melt-down remains an economic risk, the likelihood of a radioactive release threatening the public health is much smaller than had been thought previously.
Sequences contributing to risk vary depending on the figure of merit adopted.
Not only is there a 1
difference between contributors to core melt fre-i quency and to health risk, but there are even dif-
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ferences for different types of health risk.
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The risk studies have moved a long way toward dis-i pelling the " China Syndrome" scenario.
The evidence
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is very strong that basemat melt-through is not a i
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Full scope PRAs have indicated the importance of a probabilistic treatment of such external events as i
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In many cases, the external events are the major contributors to risk by virtue of the large uncer-t tainties associated with their frequencies.
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The emphasis in newer plants on safety system train i
independence and separation criteria is not necessarily favorable to risk reduction.
While such designs favorably impact such rare events as pipe ruptures, large fires, and extensive floodings, they make it more difficult to compensate for more frequently occurring failures. That is, the absence of crossties between systems denies access, for example, to alternate supplies of coolant water.
Such limitations often turn out to be more important to risk.
THE ANALYSIS OF SEVERE ACCIDENTS IN THE MIDLAND FINAL ENVIRONMENTAL STATEMENT The FES for Midland utilizes the RSS to analyze the probabilities of severe accidents.
The RSS has been "re-baselined" for this purpose, as described in Appendix E to the FES. Risk estimates for severe accidents, early fatalities and latent cancers are made in the FES, and the overall conclusion of that document is that the environmental-costs associated with radiological health impacts from severe accidents is "small".
The remainder of this testimony discusses whether studies subsequent to the RSS, particularly the Precursor Study referred to in the contention, invalidate this overall conclusion.
THE "PRECURSDR STUDY" The Precursor Study published by the U. S.
Nuclear Regulatory Commission was in direct response to questions raised by the previously mentioned Lewis group.
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It does this by sorting through the Licensee Event Reports (LERs) for the period 1969-1979 and identifying incidents which it calls " precursors for potential severe core damage accidents."
It then puts these precursors through a fairly elaborate event tree type calculation procedure and comes up with an SCD frequency of 1.7 to 4.5 x 10~3 Per year.
This has caused considerable stir, for this number is significantly higher than that calculated in many of the industry's probabilistic risk assessments.
In particular, the RSS calculated a point estimate core melt I
frequency of 6 x 10-5, a factor of almost 100 less.
The key qu'estion then is whether the Precursor Study number
+
invalidates the RSS and more generally the process of PRA.
To examine this, let us paraphrase the methodology and line of argument of the Precursor Study.
The essence of it boils down to the following:
Up through 1979 we have had 432 years of reactor operation and one SCD accident; namely TMI.
We have also had a number of "near misses";
e.g., Browns Ferry and Rancho Seco.
We assign each of these near misses a " severity factor,"
which we get from the event trees.
Adding these up we consider that the near misses all together are the equivalent i
of about one more SCD accident.
So we consider that the r
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events in 432 years which gives a frequency of 2/432 = 4.6 x 10-3, The way in which the Precursor Study handled the near misses can be subject to much criticism both, on an engineering modeling basis (i.e., the details of the event tree / severity factor work) and on the basis of statistical logic (e.g.,
whether near misses should be counted at all-).
We consider much of this criticism to be valid.
However, for our present purposes this is not a key issue.
What is pertinent is that as of 1979 we experienced one SCD, TMI, in 432 years, whereas, RSS calculated a core melt frequency of one in 20,000 years.
The question now is what does this experience tell us about the RSS, about PRA, and about nuclear safety in general.
Evaluating the RSS result, we would point out firstly that counting all free world power reactors, there is today about 1,500 plant years of experience.
Thus, our statistical evidence is now 1/1,500 rather than 1/432.
Secondly, the RSS result is for a different event than that analyzed in the Precursor Study, core melt rather than severe core damage.
Severe core damage is a superset of, and a much more frequent event than, core melt.
Thirdly, the RSS
' result is for one plant, Surrey, and cannot be used to quantify contributors to risk at any other specific plant.
1 1
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4 With respect to the PRA process itself, we would note that what a PRA is, basically, is a way of calculating the frequency of compound events from the frequencies of the
" elemental" events, which together make up the compound event.
The way in which this is done is pure logic and as such is incontestible.
Thus, PRA in general, in our view, is not called into question by the Precursor Study.
In any particular application of PRA, however, there can, of course, be errors d
in logic or arithmetic.
.With respect to this latter point, it is interesting to note that all three of the events identified as primary
}
contributors to the estimated frequency of severe core damage in the Precursor Study (Three Mile Island, Browns Ferry, and Rancho Seco) involved human errors of commission events, which j
are inherently difficult to anticipate and predict.
At I
Three Mile Island, cooling systems were mistakenly turned off by an operator; at Browns Ferry the initiating event was a fire accidentally caused by a technician; and at Rancho 4
Seco, a light bulb dropped into a circuit assembly caused the failure of nonnuclear instrumentation.
In particular,
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the Precursor Study finds that about 38 percent of all j
j significant precursors involved human error of1some kind.
Thus, although its calculations of frequency should not be i
regarded as gospel, the Precursor Study does do a valuable service in calling attention to the human element in the incidents of the past.
Recent policies of the Commission r
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and the industry have put much attention on the human aspect of recator safety.
Thus, assessment of risk today would include the substantial design and operational changes brought about by the three top-ranked events of the Precursor Study with a corresponding decrease in calculated risk.
- Indeed, the foreward to the precursor Study makes just that point.
CONCLUSION From the foregoing discussion, it is apparent that PRA is an analytical tool which is useful in a variety of I
contexts.
One use is that found in the FES, as a technique
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which will aid the NRC Staff in reaching a judgment as to 1
the environmental effects of operating the Midland facility.
Another use of PRA is to assist utility management in de-termining the contribution to risk of severe accident scenarios, keeping in mind that all NRC safety regulations have been satisfied.
For this latter purpose, a site specific PRA, rather than the generic use of the RSS by the NRC in the Midland FES, must be conducted.
Such a study is underway.
In contrast to the Midland site-specific PRA and other studies subsequent to the RSS, the Midland FES uses a
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generic risk study and point estimates, with a certain amount of "rebaselining", in order to draw the conclusion I
that the risk to the public, i.e. adverse radiological health effects, from potential accidents is "small". For I
the reasons given in the preceding sections, we believe i
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, that this conclusion is not invalidated by the Precursor Study and will be supported by the more definitive Midland PRA study now being conducted.
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In addition, it should be emphasized that the results of the FES risk analysis are but one of the three I
bases relied on by the Staff in its evaluation of the effects of accidents at Midland.
The other two bases for the Staff's conclusion are (a) "the fact that considerable experience has been gained with the operation of similar facilities without significant degradation of the environment."
(b) "the fact that in order to obtain a license to operate the Midland facility, Consumers Power must comply with the applicable Commission regulations and require-ments."
i Finally, it is worthy of ncte that the collective actions taken to improve the safety of nuclear power plants as a result of the investigations of the TMI accident and other events have reduced the risk of severe accidents t
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identified in both the RSS and the Precursor Study.
i Recently published PRAs that reflect the latest method-ology and modifications made in response to the TMI event demonstrate that the risk to the public has been I
reduced and is sharply lower than was previously believed.
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B. JOHN GARRICK EDUCATION Ph.D., Engineering, University of California, Los Angeles.1968.
M.S., Engineering, University of California, Los Angeles,1962.
B.S., Physics, Brigham Young University,1952.
U.S. Atomic Energy Commissicn Grant-in-Aid, Oak Ridge School of Reactor l
Technology, 1954-1955.
PROFESSIONAL EXPERIENCE General Summary A principal of Pickard, Lowe and Garrick, Inc. Consultant in reliability and availability, risk analysis, licensing and safety, management systems, and engineering. Pioneered early use of reliability and risk analysis technology in nuclear and fossil power plants. Served on several design review and safety committees and other task forces related to power plant design 'and operations.
Study director of numerous major risk studies of nuclear power plants including Oyster Creek. Zion, Indian Point, LaSalle, Pilgrim 1, Midland, Browns Ferry, Sequoyah, and Seabrook. Extensive experience with hearings and the general nuclear licensing process. Coordinator and principal lecturer for the annual UCLA short course on power plant reliability. Presented numerous seminars on risk and safety arialysis at such institutions as MIT, the University of California, and the United Kingdom's National Centre of Systems Reliability.
Served on several accreditation teams evaluating engineering curriculum at i
different universities. Organized and conducted numerous workshops and training programs on maintenance, reliability, and availability for EPRI, DOE, and many utilities. Author of over 90 papers and reports on reliability and risk, nuclear power, power plant siting, and energy technology.-
Adjunct Professor, University of California, Los Angeles; member of several institutional committees including the UCLA Radiation Committee, the Select Review Committee for the C) inch River Breeder Reactor, Design Review Board j
for the Midland Nuclear Power Plant, Direction and Control System Advisory
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Committee of the Governor's Emergency Task Force on Earthquake Preparedness, and Boston Edison's Nuclear Safety Review and Audit Committee. Peer reviewer of such national efforts as: (1) PRA Procedures Guides (2) NRC human relia-2 bility research project; and (3) NRC NREP Procedures Guide.
Chronological Summary l
1975-Present Principal, Pickard, I4we and Garrick, Inc.
4 1957-1975 Holmes a Narver, Inc.
i Key. Positions: Member of Board of Directors:
President, Nuclear & Systems Sciences Group Sr. V'ce President Vice President, Science & Technology, i
The Resource Sciences Corporation. Tulsa, Oklahoma (parent company).
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i B. JOHN GARRICK --- page 2 1955-1954 Physicist. Hazards Evaluation Branch, U.S. Atomic Energy Commission Washington, D.C.
1952-1954 Physicist, Phillips Petroleum Company, National Beactor Testing Station, Idaho.
MEMBERSHIPS. LICENSES. AND HONORS American Nuclear Society.
Fellow, Institute for the Advancement of Engineering.
New York Academy of Sciences.
Registered Professional Engineer, State of California.
Leaders in American Science (Eighth Edition).
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Senior Staff Engineer - Midland Project Education:
- Michigan State University, 1974, Masters in Business Administration
- U.S. Navy Training, Nuclear Engineering (12 months) 1964
- University of Notre Dame, 1963, BSEE Experience:
Consumers Power Company:
1980-1983 Section Head - Midland Project - Safety and Analysis Section.
Responsible for ensuring i
proper performance and evaluation of the i
nuclear safety analysis aspects of the plant design including accident analysis, transient analysis and probabilistic risk assessment.
Also responsible for development of conceptual design options for resolution of fluid mechanical and nuclear safety system issues.
1977-1980 l
Section Head - Reactor and Control Systems.
Responsible for design review for new power plant instrument and control systems and nuclear reactor and support systems.
Conduct detailed analysis and evaluation of certain critical areas of new plant design and modifi-cations to existing plants.
Supervise members of the section in carrying out the above section responsibilities.
1975-1977 Supervisory Engineer - Perform design review for new power plant systems and equipment in the mechanical and instrument and control areas.
Have conducted detailed design analysis of l
certain critical aspects of design of new plants and of modifications to existing plants.
Pro-vide advice and assistance to the department head in the area of Quality Assurance.
1972-1975 Administrator of Quality Assurance - Responsible i
for developing and implementing a program for operational quality assurance for the company's
+
central office and its two nuclear power plants.
Served as a member of the Off-Site Safety and Audit Review Board for two operating nuclear plants.
Participated in the review of design i
specifications for new nuclear plants.
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page 2 1970-1972 General Engineer - Supervised preoperational, hot functional and startup testing of a large PWR nuclear plant.
Qualified as Senior Reactor Operator by successfully completing examinations administered by the Atomic Energy Commission; supervised shift operations.
1967-1970 Lieutenant, U.S. Navy - Assigned to a prototype nuclear power plant which was utilized for testing and training of personnel.
Certified as senior reactor operator.
Responsible for electrical and instrumentation systems at the prototype, including supervision of forty technicians.
I later shared responsibility with my civilian counterpart for planning and scheduling the routine repair effort; partici-pated in a major refueling and overhaul of the facility.
I also served as a member of the Senior Operator Qualification Board.
1964-1967 Lieutenant, Junior Grade, U.S. Navy - Served as Reactor Control Division Officer, USS Long Beach; responsible for the maintenance, operating and testing of all the nuclear reactor in-strumentation and control systems; supervised twenty-five electronic technicians and partici-I pated in a major refueling, overhaul and testing of the nuclear plants.
I was qualified as senior watchstander on shift.
Additional Training 1973-1974 Graduate level courses - Thermodynamics, Systems Analysis, Michigan State University.
1981 One week training course on B&W Simulator 1981 One week CP Co. course on Effective Management i
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g' UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the matter of CONSUMERS POWER COhPANY Docket'Nos. 50-3290M (Midland Plant, Units 1 and 2) 50-3300M 50-3290L 50-3300L I, B. John Garrick, being first duly sworn, state that i
my accompanying testimony addressing Mary Sinclair's contention 13 (analysis of Class 9 Accidents in the FES) is true and correct to the best of my knowledge and belief.
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Subscribed and sworn to before me this /3,.cul day of January,1983.
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UNITED STATES OF AMERICA NUCLEAR RECULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In The Matter of
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Docket Nos. 50-329 ON
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50-330 ON CONSUMERS POWER COMPANY
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Decket Nos. 50-329 OL (Midland Plant. Units 1 and 2)
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50-330 OL I, Louis S. Gibson, being first duly sworn, state that my accompanying testimony addressing Mary Sinclair's Contention 13 (Analysis of Class 9 Accidents in the FES) is true and correct to the best of my knowledge and belief.
' Louis S. C'ibson 4
Subscribed and sworn to
'before me this /J* day of January,1983.
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DW Notary Public My commission esPires b f, /i[b2 t
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pp) COnrASPONDENCE
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L UNITED STATES OF' AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Hatter of
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Docket Nos. 50-329-OM CONSUMERS POWER COMPANY-(
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50-330-OM
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50-329-OL (Midland Plant, Units 1 sand 2))
50-330-OL CERTIFICATE OF SERVICE I, Philip P. Steptoe, one of the attorneys for Cons'.uners Power Company, hereby certify that copies of
" Testimony of Cris Hillman and Terry Postlewait on Sinclair contention 4", " Applicant's, Exhibit 29R", " Testimony of Richard D. Woods Concerning Seismic Snakedown Settlement at the Midland Site except Diesel Generator Building", "Appli-cant's Suggestion of Mootness or, in the Alternative, Motion to Close the Record with Respect to d'aterial False State-ment", and " Testimony of B. John Garrick and Louis S. Gibson Concerning the Analysis of Class 9 Accidents in the Midland FES" were served upcn all persons shown in the attached service list'by deposit in the United States mail, first class, this 17th day of. January, 1983.
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l SERVICE LIST Frank J. Kelley, Esq.
Steve Gadler, Attorney General of the 2120 Carter Avenue State of Michigan St. Paul, Minnesota 55108
.Carole Steinberg, Esq.
Assistant Attorney General Atomic Safety & Licensing Environmental Protection Div.
Appeal Panel 720 Law Building U.S. Nuclear Regulatory Comm.
Lansing, Michigan 48913 Washington, D.C. 20555 Cherry & Flynn Mr. C. R. Stephens Suite 3700 Chief, Docketing & Services 3 First National Plaza U.S. Nuclear Regulatory Comm.
Chicago, Illinois 60602 office of the Secretary j
Washington, D.C. 20555 Mr. Wendell H. Marshall 4625 S. Saginaw Rd.
Ms. Mary Sinclair Midland, Michigan 48640 5711 Summerset Street i
Charles Bechhoefer, Esq.
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Atomic Safety & Licensing William D. Paton, Esq.
Board Panel Counsel for the NRC Staff U.S. Nuclear Regulatory Comm.
U.S. Nuclear Regulatory Comm.
Washington, D.C. 20555 Washington, D.C. 20555 Dr. Frederick P. Cowan Atomic Safety & Licensing 6152 N. Verde Trail Board Panel Apt. B-125 i
U.S. Nuclear Regulatory Comm.
Boca R4 ton, Florida 33433 Washington, D.C. 20555 i
Lee L. Bishop Barbara Stamiris Harmon & Weiss 5795 North River Road 1725 I Street, NW H06 Route 3 i
Washington, D.C.
20006 Freeland, Michigan 48623 Mr. D. F. Judd Jerry Harbour i
Babcock & Wilcox Atomic Safety & Licensing P.O. Box 1260
' Board Panel Lynchburg, Virginia 24505 U.S. Nuclear Regulatory Comm.
Washington, D.C.
20555 j
James E. Brunner, Esq.
Consumers Power Company l
212 West Michigan Avenue Jackson, Michigan 49201 e
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