ML20052F944
| ML20052F944 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/12/1982 |
| From: | Emch R Office of Nuclear Reactor Regulation |
| To: | O'NEILL, J. |
| Shared Package | |
| ML20052F939 | List: |
| References | |
| NUDOCS 8205140177 | |
| Download: ML20052F944 (75) | |
Text
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05/12/82 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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Docket No. 50-155 CONSUMERS POWER COMPANY (Big Rock Point Plant)
)
NRC STAFF RESPONSES TO O'NEILL INTERR0GATORIES DATED 4/2/82 ON REWORDED O'NEILL CONTENTION II.C Interrogatory (1)
Interrogatory (1) refers to question IV 9 which states: " Answer IV 1-7 for any inspections done by the NRC or their consultants." Ques-tions IV 1-7 are as follows:
IV.
Inspection 1.
How often is the crane inspected? Is there a schedule?
2.
Who inspects it? What are this persons qualifications and training for this job?
3.
Exactly what is inspected?
4.
Exactly what is looked for?
5.
What equipment is used in inspection?
6.
Is damage from incremental seismic activity sought? This activity would include the conti,1uous blasting at the Medusa Cement qua"ry nearby. See deposition of David Blanchard of 1
l Consumers' Sig Rock plant, January 12, 1982, p. 11.
7.
Is radiation fatigue to the crane looked for?
a.
How much radiation is the crane exposed to on an annual basis?
b.
What is the highest amount of annual exposure of radiation to the crane?
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c.
What evidence of radiation damage to the crane has been discovered?
d.
Could radiation damage in any way make the crane more susceptible to seismic vulnerability? Any history of this?
Response
As p&rt of the inspection program of the Office of Inspection and Enforcement, the NRC audits the Licensee's records to ensure that the Licensee has properly conducted crane inspections. These inspections are conducted according to Manual Chapter 60710 of the I&E Inspection itodules by the resident inspector or technical inspector from the Regional Office.
I consulted Mr. G. Wright, NRC's senior resident inspector at Big Rock Point, and Mr. F. Clemenson of the NRR staff in preparation of this response. Mr. Clemenson will present testimony regarding the crane on O'Neill Contention II.C. Manual Chapter 60710 of the I&E Inspection Modules is a pertinent document to this interrogatory.
Interrogatory (2)
" Detail inspection standards for topic raised in IV 9."
Response
See response to Interrogatory (1).
Interrogatory (3)
Interrogatory (3) refers to question VII.1.c and d.
VII 1.
Answer all of the questions in Interrogatory V, 1-7, for analyses performed by:
c.
Outside consultants of the NRC,
d.
The NRC.
i.
What changes in equipment, procedure, modification to equipment, procedures, reporting, inspections, maintenance, etc. were required by the NRC? Detail.
ii. Have these (VII, 1.d.1) been performed?
Detail.
iii. Answer (VII,1,d,1) for changes or modifi-cations recommended but not required.
Have these been performed? Detail.
iv. Detail any required or recommended changes that have been delayed or deferred.
v.
Detail any required or recommended changes that are contested by the licensee.
V.
1.
Supply all analyses of the crane, the casks, the fire water system piping and the possibility of the pool being breached and the water made up that have been performed by W.J. Hall.
2.
Supply all of his correspondence to licensee concerning the above.
3.
What are the specific recommendations he made to correct any identified problems?
4.
Which of these have been performed as recommended?
5.
Which of these will be performed as recommended in the future?
6.
What corrections have been made that differ from Hall's recommendations? Why?
7.
Upon which of Hall's recommendations have no actions been taken? Explain.
Response
Attached is a letter from Mr. Crutchfield of the NRC to Mr. Hoffman of Consumers Power Co. on September 29, 1981 regarding the seismic design of Big Rock Point. Mr. Hall's report on Big Rock Point is enclosed in that letter. Most of the subquestions are addressed by that letter. The overall
issue of seismic plant design is being addressed by the Systematic Evalu-ation Program review of Big Rock Point. All of Mr. Hall's recommendations are being. factored into that review. Some of Mr. Hall's recommendations such as anchoring the control room cabinets have been completed; however, all of his recommendations have not been fully addressed or implemented to date.
Also, under NRC contract, Mr. I. Sargent and D. Vito of WESTEC Services, Inc., are reviewing the overall issue of the handling of heavy loads at Big Rock Point. During a plant visit in March 1982, a draft working copy of a Technical Evaluation Report, TER-C5257-440 (TER) (copy attached) prepared by WESTEC was given to Mr. D. Blanchard of Consumers Power Co. Mr. Blanchard was asked to supply missing informa-tion and check the draft for accuracy in its description of the Big Rock Point program. Any recommendations contained in that report are presently informal and have not recessarily been endorsed by the NRC. All such recommendations will be factored into the ongoing NRC review of heavy load handling.
I consulted G. Wright, K. Herring, W. Russell, and F. Clemenson of the NRC staff and D. Vito and I. Sargent of WESTEC Services, Inc.,
in preparation of this response. Herring, Clemenson, and Sargent will present testimony regarding the crane on O'Neill Contention II.C.
The WESTEC TER and Mr. Crutchfield's 9/29/81 letter are pertinent documents to this interrogatory.
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Interrogatory (4)
" Reg 6rding VIII, what type of seismic monitoring is required?"
Response
No seismic monitoring is required by the NRC at Big Rock Point.
I consulted G. Wright and K. Herring of the NRC staff.
Interrogatory (6)
Interrogatory (6) refers to question XI.7.c. and d. which state:
XI. Operation of the crane 7.
Concerning all aspects of crane operation and crane controls:
c.
supply technical specifications regarding:
1.
operation of crane; ii. crane controls; iii. any monitors used; iv. any other related instruments, d.
Have any of the above technical specifications been violated? Give details, including fre-quency of violations.
Response
Technical Specification 7.4(a) states:
" Detailed written procedures shall be available prior to each refueling operation." This Tech. Spec.
does not address crane operation explicitly, however, crane operation is a part of the refueling operation. Actually, no Tech. Spec. directly addresses crane operation.
Big Rock Point has not been cited for violations of the Tech. Spec.
related to crane operation by the NRC since January 1979. This cutoff
^
date was agreed to by Mr. O'Neill for the purposes of bounding the infor-mation search.
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I I consulted with G. Wright, Big Rock Point senior resident inspector.
The Big Rock Point Technical Specifications are a pertinent document to this intarrogatory.
Interrogatory (7)
Interrogatory (7) refers to question XI 11 and also asks, "are any of these controls, monitors and wiring and related equipment required to be seismically qualified?" XI 11 states:
Are the following seismically qualified:
a.
crane controls?
b monitors?
c.
wiring for each of the above?
d.
Other related equipment necessary to crane operation?
Response
The crane controls, monitors, wiring, and related equipment are not required to be seismically qualified. However, the crane itself must be properly supported such that it would not collapse or fall during a seismic event.
I consulted K. Herring and F. Clemenson of the NRC staff. Mr. Herring will present testimony on this subject in response to O'Neill Contention II.C.
Interrogatory (8)
Interrogatory (8) refers to question XI.18.b which states:
Detail all of the actions the crane operator is trained to perform in response to events postulated in numbers 16 and 17.
b.
Detail technical specifications concerning these accidents.
Response
Therb are no Technical Specifications specifically concerning operation' of the crane during an emergency or accident.
I consulted G. Wright. The Big Rock Point Technical Specifications are a pertinent document to this interrogatory.
Interrogatory (9)
Interrogatory (9) refers to question XII.8.f which states:
Emergency use of the buried fire main has been disallowed by the NRC. What will replace this system?
f.
Will the entire replacement make-up system be seismically qualified?
Responsi Mr. O'Neill agreed to rephrase the question as follows, "Will the entire spent fuel pool make-up system be required to be seismically qualified." The necessity for the new make-up system to be seismically qualified has not been determined by the NRC. This issue will be covered in the SEP review of the overall seismic qualification of the Big Rock Point plant systems.
It is important to note that emergency use of buried fire main has not been disallowed by the NRC, as Mr. O'Neill states in XII.8.
I consulted K. Herring of the Systematic Evaluation Program Branch.
Interrogatory (10)
Inte'rrogatory (10) refers to question XIII.8.c. which states:
Supply any analyses of the accident postulated XIII performed by-c.
Include reconinendations.
In the case of the NRC, be sure to include mandated modifications, including those i.
implemented; ii. planned but not yet implemented, along with date of scheduled completion; iii. disputed.
Response
NRC analyses of fuel assembly and cask drops at Big Rock Point are presented in two letters:
1.
Goller, NRC, to Sewell, CPC, dated February 6,1976 (copy attached).
2.
Crutchfield, NRC, to Hoffman, CPC, dated May 15, 1981.
The second letter transmitted the SER for the Big Rock Point spent fuel pool expansion.
I consulted J. Donohew, W. Pasedag, M. Wohl, and W. Paulson of the NRC staff. The 1976 and 1981 letters are pertinent documents to this interrogatory. Mr. Wohl will present testimony on this subject under O'Neill Contention II.E.4. and F. Clemenson will present testimony on this subject under O'Neill Contention II.C.
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Interrogatory (11)
Interrogatory (11) refers to question XVII which states:
' Provide a transcript of the April 1 meeting between Consumers Power personnel and Ken Herring of the NRC, concerning seismic issues.
a.
Supply all documents discussed at that meeting, b.
Supply all documents generated as a result of that
- meeting, c.
Supply all documents relied upon but not discussed.
Response
There was no transcript taken at the April 1 meeting referred to in this interrogatory. Meeting minutes were prepared and sent to everyone on the Big Rock Point service list including Mr. O'Neill. The following other documents were discussed at the meeting or generated as a result of the meeting:
1.
Seismic Safety Margin Evaluation Big Rock Point, D'Appolonia, August 1981, revision 1.
2.
Seismic Analysis of the 75-Ton Containment Crane, EQE Inc., April 1982.
3.
Parametric Study Soil Structure Interaction, D' Appolonia, April 1982.
4.
Letter, Thomas Nelson, Lawrence Livermore National Laboratory, to William Russell, NRC, dated January 28, 1982, transmitting seismic review of Big Rock Point containment shell structure (copy attached).
5.
Letter, David Hoffman, CPC, to Dennis Ziemann, NRC, dated April 23, 1979, transmitting fuel pool expan-sion application.
6.
Derivation of Floor Response Reactor Building, D',Appolonia, June 1978.
I consulted K. Herring nf the NRC staff.
10 -
Interrogat$ry "Grandfathering" Detall those NRC standards, test standards, models for postulated events, implementation schedules and other matters of regulation have been modified, relaxed or dropped for the following Big Rock systems:
a.
seismic qualifications of the crane within containment; b.
threading on the fire water system piping and its seismic qualification; c.
breach of spent fuel pool scenarios; d.
models for and calculations surrounding cask drop accidents:
Describe the reasoning behind each.
The Board has raise the question of adequate NRC review. What con-clusions and reports of Licensee or their consultants have not been independently calculated by the NRC? Explain in depth. Include those models and programs of analysis that have not been thoroughly reviewed.
Response
Mr. O'Neill was informed that the NRC will not respond to the second part of "Grandfathering."
No NRC standards, models, regulations or schedules have been relaxed or dropped for the Big Rock Point systems presented in the first part of "Grandfathering."
I consulted with F. Clemenson, K. Herring, M. Wohl and W. Paulson of the NRC staff.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0ftSSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of CONSUMERS POWER COMPANY Docket br 50-155 (BigRockPointPlant)
)
(Spent Fuel Pool Modification)
AFFIDAVIT OF RICHARD L. EMCH, JR.
I, Richard L. Emch, Jr., being duly sworn, do depose and state:
1.
I am the Project Manager for the Big Rock Point Plant in the Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission, Washington, D.C. 20555.
2.
The answers to the O'Neill Interrogatories on Contention. II.C. were prepared by me.
I hereby certify that the answers are true andcorrect to the best of my knowledge, information and belief.
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Richard L. Emch, Jr.
f Subscribed and sworn to before j
me this 12th day of May,1982.
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NUCLEAR REGULATORY COMMISSION a
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September 29, 1981
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Docket No. 50-155 L505 09-073.
Mr. David P. Hoffman Nuclear Licensing Administrator Consumers Power Company 1945 W. Parnall Road Jackson, Michigan 49201 i
Dear Mr. Hoffman:
SUBJECT:
SEP TOPIC III-6 SE!SMIC DESIGN CONSIDERATIONS BIG ROCK POINT NUCLEAR POWER STATION In accordance with 10 CFR 50.54(f) of the Comission's Regulations, our letter to you dated August 4,1980 requested that you submit plans and proceed with a seismic reevaluation program for Big Rock Point facility and that you provide justification for your conclusion that continued operation is justified in the interim until the seismic reevaluation and any necessary upgrading, as results from this reevaluation, is completed. The staff has completed the review of the information supporting continued operation contained in your letters dated February 23, April 25,1979; February 13, March 31, October 10, 1980; and July 27, 1981 and the meeting summaries dated August 7,1979 and June 22, 1981.
Furthemore, the staff and its consultant (Prof.
W.J. Hall of University of Illinois) visited the site to evaluate the seismic resistance of the facility.
As a result of this review, the staff has concluded that continued operation of the Big Rock Point Nuclear Power Plant is justified under the following conditions:
(1) results of seismic analysis are submitted for NRC review on l
l the schedule specified in your July 27, 1981 letter; and I
(2) in case of any modifications shown to be necessary as a result of the seismic analysis which are not implemented by January 1,1983, the schedule for implementation and additional justification for continued operation over the period of this implementation are to be submitted and will be reviewed on a case by case basis.
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Enclosed is our safety evaluation report.
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Dennis M. Crutchfield, C ef Operating Reactors Branch #5 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
See next page l
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ec Mr. P aul A. Perry, Secretary.
U. S. Environmental Protection Consumers Pcwer Cogany Agency 212 West Michigan Avenue Federal Activities Branch Jackson, Michigan 49201 Region V Office ATTN:
EIS COORDINATOR Judd L. Bacon, Esquire 230 South Dearborn Street Consumers Power Co@any Chicago, Illinois 60604 212 West Michigan Avenue Jackson, Michigan 49201 Hertert Grossman, Esq., Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission Joseph Gallo, Esquire Isham, Lincoln & Beale Washington, D. C.
20555 1120 Connecticut Avenue Dr. Oscar H. Paris Room 325 Washington, D. C.
20036 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission Peter W. Steketee. Esquire Washington, D. C.
20555 505 Peoples Building Grand Rapids, Michigan 49503 Mr. Frederick J. Shon Atomic Safety and Licensing Board Alan S. Rosenthal, Esq., Chairman U. 5. Nuclear Regulatory Comission Atomic Safety & Licensing Appeal Board Washington, D. C.
20555 U. S. Nucitar Regulatory Comission Washington,* 0. C.
20555 Big Rock Point Nuclear Power Plant ATTN: Mr. C. J. Hartman Mr. John O'Neill, II Plant Superintendent Route 2, Box 44 Charlevoix, Michigan 49720 Maple City, Michigan 49664 Christa-Maria Charlevoix Public Library Route 2, Box 108C Charlevoix, Michigan 49720 107 Clinton Street Charlevoix, Michigan William J. Scanlon, Esquire 2034 Pauline Boulevard Chairman l
County Board of Supervisors Ann Arcor, Michigan 48103 Charlevoix County Charlevoix, Micnigan 49720 Resident Inspector Big Rock Point Plant Office of the Governor (2) c/o U.S. NRC
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Room 1 - Capitol Building RR #3, Sox 600 Lansing, Michigan 48913 Charlevoix, Michigan 49720 Mr. Jim E. Mills Herbert Semel Council for Christa Maria, et al.
Route 2, Box 108C Charlevoix, Michigan 49720 Urban Law Institute Antioch School of Law 2633 16th Street, NW Thomas S. Moore Washington, D. C.
20460 Atomic Safety & Licensing Appeal Board U. S. Nuclear Re911atory Comission Washington, D. C.
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BASES FOR CONTINUD OPERATION BIG ROCK POINT NUCLEAR POWER STATION INTRODUCTION In accorTance with the Commission Regulation 10 CFR 50.54(f), a letter was issued on August 4,1980 to Consumers Power Company requesting the licensee to provide justification for continued operation until their seismic reevalua-tion of their facility is complete.
In response to this letter, the licensee submitted its basis for continued operation on October 10, 1980. On January 9,1981, a summary of seismic reanalysis completed to date for the plant facilities was submitted.
In this summary report, safety margin of plant structures and syster.s (a total of 15 structures, systems and subsystems was included) was demonstrated. More recently (June 19, 1981), additional information for justification for continued operation was provided. On June 30,1981, the staff and its consultant, Professor W. J. Hall of University of Illinois, made a visit to the plant to discuss the seismic capability of structures, systems and components at the Big Rock Point Plant with Consumerr Power representatives. The staff's evaluation of the basis for continued cperation follows.
Seismic Hazard Consideration The staff, in its letter dated August 4,1980, directed the licensee to conduct the seismic reevaluation of Big Rock Point Nuclear Power Plant using the site specific spectrum (0.119 peak ground acceleration) as the free field ground. notion. The adequacy of this site specific spectrum was confirmed by the staff through a letter dated June 17, 1981. This ground motion is equivalent to an earthquake with return period between 1,000 years and 10,000 years.
Seismic Resistance of Structures, Systems and Components In response to NRC letters dated January 15, 1979 and August 4,1980, the lit.ensee completed a limited seismic reanalysis on Big Rock Point plant facility (15 itens were analyzed including most of safety related struc-tures as well as some systems). A summary of these results was submitted to the staff on January 9,1981 as part of justification for continued operation. Regulatory Guide 1.60 Spectrum scaled to 0.12g peak ground acceleration was used for input ground motion. This spectrum completely enveloped the site specific spectrum recommended by the staff, i.e., it is more conservative than the site specific spectrum. From the preliminary review of the summary report and other information received from the licensee (letters dated February 23, April 25 of 1979, February 13 March 31, October 10 of 1980, July 27,1981 and meeting summaries dated August 7,1979 and June 22,1981), the significant findings are highlighted below:
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The criteria and analytical procedures used for the structures in the reanalysis are generally acceptable to the staff.
. The r5sults presented by the licensee indicated that all items ident-ified were adequate for the postulated earthouake (0.12g Regulatory Guide 1.60 Spectrum) with the following exceptions:
(1) steel bracing and steel column bases of service building complex were found to be overstressed; (2) allowable stresses would be exceeded at junction of 24" recircu-1ation pipe and 4" cross-connecting loop and undesirable dis-placements were found at this junction.
The inadequacies identified above are considered insignificant because:
(a) conservatism does exist in the original design with respect to ductility, damping, actual material strength, etc..and (b) conserva-tive seismic input (in comparison with the staff recommended site specific spectrum) was used in this analysis.
In addition, the staff and 'its consultant (Prof. W. J. Hall of Univ. of Ill.) visited the plant site to evaluate the seismic resistance of the facility. The report of staff's consultant regarding the basis for continued operation of the facility is attached (Enclosure 1). One area, the threaded piping of the fire protection system, was identified as a ~
possible weak link in the facility by the consultant. The fire protection system is the primary source of water for the primary and backup core spray systems as well as the backup supply for the shell side of the emergency condenser. The licensee in a letter dated July 27, 1 981 sta ted that a redunoant source of water for the core spray system is provided by the Post Incident System. This piping system is completely welded and connected to the yard loop piping that has been seismically analyzed. As far as the plants capa.:ity to make up water to the pimary coolant system and the emergency condenser, alternate sources (water can be supplied througn a welded pipe system from the OMW-storage tank or from the domes-tic water system onsite with the water sources being the domestic water accumulator, the well water storage tank via the domestic water pump, or ultimately the deep well pump, if the DMW tank becomes depleted) do exist to provide a reasonable degree of redundancy for the removal of decay heat.
The consultant recomended that continued operation 'in the interim should be permitted.
Since early 1979, the folowing additional seismic issues have been addres-sed, resolved or are being resolved under the SEP seismic review:
In response to NRC "Achorage and Support of Safety-Related Electrical Equignent" issues dated January 1,1980 and July 28, 1980, a total of 52 items was inspected by the licensee and its consultants (CPCo letters dated February 13, 1980 and March 31, 1980). The necessary modifications of equipnent anchorage and support identified during field walk-down were installed based on the results of analysis performed to a 0.129 R. G.
1.60 Spectrum input (CPCo letters dated October 10, 1980 and January 22 and March 26 of 1981).
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. The. licensee, in response to NRC letters dated August 4,1980 and April 24, 1981, has initiated a seismic reevaluation program that is schedul-ed.for completion by the end of 1982.
CONCLUSION Based on the results provided to date from the analyses of the plant structures and systems / subsystems, the proper anchorage and support of safety related electrical components, the alternate cooling water sup-plies, and the inherent capacity of the remaining plant structures, systems and components as well as the low seismic hazard (NRC June 17, 1981 letter) associated with the Big Rock Point site, the staff concludes that the continued operation of Big Rock Point Nuclear Power Plant during the seismic reevaluation of the facility and the implementation of any modification shown to be necessary as a result of seismic reanalysis is justified under the following conditions:
(1) results of seismic analysis are submitted for NRC review on the schedule specified in the licensee's July 27, 1981 letter, and (2) in case of any modifications shown to be necessary as a result of the seismic analysis which are not implemented by January 1, 1983, the schedule for implementation and additional justification for continued operation over the period of this implementation are to be submitted and will be reviewed on a case by case basis.
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July 22, 1981 SM 81-194 Mr. P. Y. Chen Systematic Evaluation Prooram Branch Division of Licensino Office of Nuclear Reactor Reo.
U.S. Nuclear Regulatory Commissjon Washington, D.C.
20555
Dear P. Y.:
I have enclosed W. J. Hall's final letter regarding continued operation of the Big Rock Point Plant and EG&G's review of the Lacrosse program plan. Note that the Lacrosse review should be used in conjunction with the previous submittals and hichlights those items which have still not been adecuately addressed in the program plan.
Sincerely, A
Thomas A. Nelson Structural Mechanics Group Nuclear Test Engineerino Division TAN /mg 0007m Enclosures i
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July 15,1981 Mr. T. A. Nelson L-90 Lawrence Livermore Laboratory P. O. Box 808 Livermore, CA 94550 Re: Big Rock Point Nuclear Power Plant -- Docket No. 50-155 LLL Agreement 1523501
Dear Mr. Nelson:
Comments arising from my review of the Big Rock Point Nuclear Power Plant and in particular pertaining to its ibility to accommodate seismic effects follow.
Over the past several months I have received the following material for re-view pertaining to this case.
Material originating from Consumers Power Company 1.
Letter of March 31, 1980 (8 pages) -- Re: Anchorage and Support of Safety-Related Electrical Equipment 2.
Letter of October 10, 1980 (7 pages) -- Re: Response to Staff Letter dated August 4,1980 - Proposed Seismic Evaluation Program and Basis for Continued Interim Operation 3.
Letter of January 9, 1981 (20 pages) -- Re: Preliminary Seismic Safety Margin Evaluation 4.
Letter of March 26, 1981 (4 pages) -- Re: Anchorage and Support of Safety-Related Electrical Equipment 19, 1981 (3 pages) -- Re: SEP Topic III-6, Seismic Letter of June 5.
Design - Proposed Progress and Justification for Continued Operation 6.
Excerpt pp. 53-55, copy from BRP risk analysis 7.. Excerpt pp. VI-113 to VI-174, copy from BRP risk analysis e
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2 Material originating from U. S. Nuclear Regulatory Comission
- 1.. Letter of August 4,1980(3 pages plus Attachments 1 and 2)
- 2. -Letter of April 24, 1981 (5 pages) -- SEP Topic III-6, Seismic De-
' sign Considerations Big Rock Point 3.
Letter of June 8, 1981 (28 pages) -- Site Specific Ground Response Spectra for SEP Plants Located in the Eastern United States 4.
Letter of June 22, 1981 (11 pages) -- Summary of Meeting Held with Consumers Power Company to Discuss Se'ismic Design Considerations (SEP Topic III-6) for the Big Rock Point Plant 5.
Plot (undated) -- 1 page, illustrating USNRC Site Specific Spectrum and the 0.12 g anchored REG. Guide 1.60 Spectra (84.1 percentile) employed by applicant for analysis 6.
Copy of Report entitled " Derivation of Floor Responses -- Reactor Building," prepared by D'Appolonia Consulting Engineers, Inc., June 1978, 62 p.
On June 30, 1981, in conjunction with T. Cheng and W. Paulson a d W. T.
Russell, I made a site visit to the Big Rock Point site and participated in technical discussions and a plant inspection.
On the basis of my review of the seismic portions of the risk analysis studies made available to me, and reflecting on the brief discussions at the time of the site visit wherein some of the uncertainties ano gaps'in the analysis were identified, I cannot recommend employing such an approach as a sole basis for continued operations. Such studies, when they encompass rig-orous total system perfonnance and the interactions therein, can be helpful as a basis for forming an opinion as to the adequacy of expected performance under various conditions of system disturbance. My brief review of the seismic portion of the risk analysis studies suggests that such an overall comprehensive treatment does not currently exist in the present case.
As one might surmise from my foregoing statements, and irrespective of whether or not the level of earthquake hazard is perceived to be low based on recent I
l recorded seismic history, I believe reasonably demonstrated adequacy of system j
resistance to earthquakes is necessary.
In view of the recent seismic quiescence of the region in which the plant is located, and on the basis of the recent USNRC/ TERA s.ite specific studies, spectra anchored at 0.11 to 0.12 g horizontal ground acceleration appear acceptable in this particular case. Although I appreciate the bases upon which the USNRC site specific spectra were generated, I do wish to note that f
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Dr. Newmark (prior to his death) and I expressed concern verbally that in some cases the amplified regions (acceleration and velocity) were low compared to Standard Reg. Guide 1.60 or NUREG CR/0098 spectra which we In this case it is my understanding that normally-would recommend for use.
the licensee and his consultants have employed Reg. Guide 1.60 spectra (84.1 -
percentile) anchored at 0.12 g horizontal ZPA for safety related structures and the reactor coolant loop; such spectra do contain reasonable accelera-tion and velocity amplifications and I concur with their use.
Even so, when reviewing the physical resistance of critical safety systems in such cases as this, namely older plants, I-recommend particular attention be paid to the margins that may be present to resist overloading from seismic However it is only fair to note that in the case of anticipated low effects.
4 seismic activity, as in this case, the loading contribution from seismic effects is normally only a small fraction of the total stressing at critical locations, especially when compared to allowables.
I subscribe fully to the content of the April 24, 1981 USNRC letter and shall not repeat the contents of that letter herein.
The site visit reveals that much of the equipment has been reviewed for adequacy of anchorage and sup-port, which is comforting, but much remains to be done (as for example, the walk-through alone, which involves limited inspection at test, indicated a need for anchoring the control room cabinets, anchoring cranes,. ~ anchoring i
l fire extinguishing equipment, and anchoring some batteries); comments made on the tour suggested that some portions of the eouipment have not been exam-ined as yet.
In any event, it is my recommendation that this program of up-grading be pursued rigorously, systematically and promptly.
Obviously I believe the total system integrity at the reactor coolant pressure boundary should be examined carefully as soon as possible on a documented system by system basis.
In this connection I am concerned that the fire water system witn its standard threaded pipe which is relied upon to provide post-incidence emergency water injection from the intake well, may not possess the desired inherent resistance. This system may not possess the resist-ance to seismic excitation that is believed to exist, and I strongly suggest that an upgraded system be developed and installed in the very near future, with some cegree of recundancy as to water sources, water paths and pumping capacity.
It appears to me that such upgrading can be done at minimal ex-pense, but care must be exercised that the system is anchored to sound structural support systems, i.e. not walls which can fail or near walls which could affect the system perfortnance. Alternatively it may be necessary to strengthen some walls.
In conclusion, the system as it currently exists may not be as inherently If the licensee resistant to seismic excitation as believed by the licensee.
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Sincerely yours, (VIhd W. J. Hall nJi.:efh cc:
W. T. Russell, USNRC e
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TER-C5257-440 CONTENTS Section Title Page 1
INTRODUC'l I ON.
1 1.1 Purpose of Review I
1.2 Generic Background.
1 1.3 Plant-Specific Background 2
EVALUATION AND RECO.9:ENDATIONS 2.1 General Guidelines 2.2 Interim Protection Measures.
3 CC:.0L'JDING SU$t*,RY 3.1 Ger.e ral Pr ovisions for Icad Handling 3.2 Inter ia, Frotection l'ez sures.
3.3 Sar.ary.
4 REF4FINCES
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This Technical Evaluation Report was prepared by Franklin Rescarch Center under a contract with the U.S. Nuclear Regulatory Con. mission (Office of Nacicar Reactor Regulation, Division of.Op'erating Reactors) for technical asristarice in rapport of NHC operating reactor licensing actions.
The technical' evaluation was conducted in accordance with criteria established by the.NRC.
Mr.
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I tG'BODUCTION 1.' l -FURPOSE OF REVIEW This technical evaluation report documents the Franklin Research Center (ThC) review of general load handling policy and procedures at the Consun.ers Power Company's Big Rock Point Nuclear Plant. Tnis evaluation was performed with the follo+ing objectives:
o to assess conformance to the general load handling guidelines of-NUREG-0612, " Control-of Heavy Ioads at'Naclear Power Plants" [1],
Section 5.1.1 o to asscss conformance to the interim protection measures of NUFIG-0612, Section 5.3.
1.2 GENERIC EAChGHOUND Gsneric Technical A tivity Task A-36 was established by the U.S. Nuclear 7,-g.:: atory Cri ssion (!ac) staff.t.o systematically examine stcff licensing criteria and the a?.eq>acy of n.easares in effect at operating nuclear power plan s to arsure the safe _ handling of heavy 1 cads and to reco=.end neces ary
- cha:.;cs to these measu:es.
Tnis activity was in.itiated by a letter issued by the N <C st af f cn May 17, 1978 12) to all power reactor licensees, requesting it,fc:a.3 tion concer ning the control of heavy loads near spent fuel.
The r es ults of Ta sk A-36 wer e r eported in fr.' FIG-0612, " Control of Heavy I.neds at IUclear Po-er Plants."
The staff's conclusion from this evaluation was t hat existing heasures' to cont iol the handling of heavy loads at operating plcr.ts, al tf.cugh pr oviding protect ion f rom certain patential problems, do not adequately cover the major causes of load handling accidents and should be upgraded.
In order to upgrade n.casures for the control cf heavy loads, the staff developed a series of guidelines designed to achieve a two-phase objective using an accepted approach or protection philescphy. The first tortion of the objective, achieved ~through a set of general guidelines identified in NUREG-0612, Article 5.1.1, is to ensure that all load handling systems at f.%
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TEE-C5357-440 nuclear,cwer plants are decisned and operated r_ h that their probability of failure is. uniformly small and appropriate for the critical tasks in which they are employed.
The second portion of the staf f 's objective, achieved
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to ensure that, for load handling systems in areas where their failure might result in significant consequences, either (1) fe atures are provided, in i
addition to those required for all load handling systems, to ensure that the potential for a load drop is extremely small (e.g., a single-f ailure-proof crcne) or (2) conservative evaluations of load handling accidents indicate tha,t the potential consequences of any load drop are acceptably small.
Acceptability of accident consequences is quantified in NUREG-0612 into'four i
accident analysis evaluation criteria.
i The apprcach used to ocvelop the staff guidelines for minimiz2ng the potential for a Icad drop w:s based on defense in Jepth and is sur.marized as follcas:
1.
prcvide sufficient operator training, handling system design, Icad handling instructions, and eqaipment inspection to assure reliable cperation of the handling system 2.
define safe lead t:4 vel paths through p c.:edures and operator training so that, to the extent practicil, heavy 1 cads ar e not carried cver or ncar irradiated fuel or rafe shutdewn equipment i
provide mechanical. tops or electrical interlocks to prevent movement j
3.
u of-heavy Icids crer irraciated fuel or in proximity to equipment atsociated with redundant shutdown paths.
Staff guidelines resulting from the forescing are tabulated in Section 5 ef RJ AG-0612.
Section 6 of NUREG-0612 recommended that a program be initiated i
to encare that these guidelines are implemented at' operating plants.
1.3 PLANT-SPECIFIC BACKGstrJND On December 22, 1980, the NRC issued a letter 13] to Consumers Power Company, the Licensee for the Big Rock Point Naclear Plant, requesting that the Licensee review provinions for handling and control of heavy loads, evaluate these provisions with respect to the guidelines of NUREG-0612, and
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On June 10, 1981, Consumers Pc.er provided the initial response [4] to this request.
Additional ir.forn.ation E s provided by the Licensee on July 1, 1981 [5] and Septer.ber 23, 1981 [6].
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EVALUATION AND M,CCW'.E'OATIONS FRC's evaluation of Icad hhndling at the Big Rock Point is divided into two categorie.
These categories deal separately with the general guidelines 4
of t.JhEG-0612 Article 5.1.1 and the recommended interim measures of Article 5.3 or their equivalents from NUs.EG-0612.
Applicable guidelines are referenced in cach category.
FRC's conclusion and reco cendations are provided in the serrary for each guideline.
2.1 GENERAL GOIDELINEs The NRO has established seven general guidelines which must be :r.et in order to provide the defense-in-depth approach for the handling of heavy 1 cads.
These guidelines consist of the following criteria from Section 5.1.1 of NUAEG-0612:
o G.:idel:.ne 1 - Eafe Icad raths o Guideline 2 - Lead Handling Frocedures o Guideline 3 - Crane Operator Training o Guideline 4 - Sp+cial Lif ting Devices o Gaideline 5 - Lif ting Devices (Not Speci' ally Designed) o Guideline 6 - Cranes (In:pection, Testing, and Paintc.ance) o Gaide]i..e 7 - C:ane Le.eign.
These reven guidelines cheald be satisfied for all overhead handling cystems and pro;:4.r.s in order to handle heavy Icads in the vicinity of t he
- t. tor vecsel, near ep.nt fuel in the spent fu l pool, or in other areas
.chere a 1cco or op ray em ge saf e chutdown syster.s.
The Liccnsee's verifica-tion of the extent to which these guidelines have been satisfied and FRC's evaluation of this verification are contained in the succeeding paragraphs.
2.1.1 NUREG-0612 Heavy load 0.'et head Handling Systems Summary of Licensee Statements and Conclusions a.
The Licensee stated that the follo.<ing overhead handling systems at Big Rock Point are subject to the general guidelines of NUREG-0612:
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TER-C5257-440 Reactor Crane Reactor Auxiliary Holst Reactor Depressurization System Holst Clean-up Demineralizer Hoist 3
W T.ergency Condenser Beam Ooi Turbin(~ Crane Additionally, portable gantries capable of lif ting several tons are located in the reactor building laydown area and the emergency condenser level. A jib crane is located over the reactor vessel during refueling operations and is rated at 500 lbs. A 500 lb winch also located on the bridge over the fuel pool and a 2000 lb winch is mounted on the 24 ton fuel transfer cask.
The Licensee has excluded the following load handling systems because equipment required for safe shutdown is not affected by a failure of close 1 cad handling system:
Decontamination Room Hoist Equipment Lock Crane Screen House Trolley Machine Snop Trollies CRD Hoist and Trolley D.
PHC Evaluation FRC concurs with the Licensee's determination of NUREG-0616 applicability with the exception of the Screen House Trolley.
The intent of NUREG-0612 is to reduce the probability of damage to any system required for plant shutdown or decay heat removal and spent fuel as a result of a load handling sy: tem failure.
FRC review of the Licensee's submittal indicates that a failure of-the Screen House Trolley could affect the circulating water and service water system, ability to provide cooling water for the main condenser,jcore! Spray ared !Postl Accident l Heat l Removal, automatic reactor depressurization and component cooling heat removal.
Furthermore, FRC consides the exclusion of the Equipment Lock Crane from
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NUREG-0613 applicability acceptable. However, the Licensee should provide justification to verify that the equipment lock crane operates in the vicinity
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of the loading dock only under conditions when containment integrity is not required.
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FRC Conclusions and Recommendations Big Rock Point substantially complies with hTREG-0612 concerning load.
handling system applicability to the general guidelines included in Section 5.1.1.
In order to fully comply, the Licensee should:
1.
Modify the heavy load handling program at Big Rock Point to include the Screen House Trolley.
2.
Verify that the mechanical blocks on the Decontamination Hoist are permanently installed, physical limitations to hoist travel.
3.
Verify that the Equipment Iock Crane operates in the vicinity of the loading dock only under conditions when containment integrity is not required.
2.1.2 Saf e Load Paths, [ Guideline 1, hTREG-0612, Article 5.1.l(1))
"Saf e load paths should be defined for the movement of heavy loads to minimize the potential for heavy loads, if dropped, to impact irradiated fuel in the reactor vessel and in the spent
- fuel pool, or to impact safe shutdown equipment.
The path should follow, to the extent practical, structural floor member s, beams, etc., such that if the load is dropped, the structure is more likely to withstand the impact.
These load p:ths should be defined in procedures, shown on equipment layout drawings, and clearly marked on the floor in the area where the load is to be handled.
Deviations from defined load paths should require written alternative procedures approved by the plant safety review committee."
a.
Su:*. mary of Licensee Statements and Conclusions _
The Licensee has indicated that the safe load paths inside of containment
_ _ have been selected to minimize the potential for damage to fuel and/or safe shutdown equipment should a heavy load be dropped in these areas.
Additionally, the Licensee has stated that restricted areas are designated both in containment and in the turbine building.
Lifts within restricted areas are procedurally controlled to limit the height of a heavy load lif t
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Deviations from safe load paths and restricted area limitations will be made only upon review by the Plant Review Committee.
b.
FRC Evalgetion FRC concurs with the Licensee actions concerning the establishment of safe load paths. 1nowever, the Licensee has made no statement concerning the,
marking of safe load paths to provide visual aids to operations and supervisors during load handling evaluations.
c.
FRC Conclusions and Recommendations Big Rock Point substantially complies with Guideline 1 of NUREG-0612 with the exception of marking of safe load paths.
The Licensee should provide information concerning the marking of safe lead paths and restricted areas in accordance with Article 5.1.l(1) of NUREG-0612.
2.1.3 Load Handling Procedures [ Guideline 2, NUREG-0612, Article 5.1.l(2) ]
" Procedures should be developed to cover load handling operations for heavy loads that are or could be handled over or in proximity to irradiated fuel or safe shutdown equipment., At a minimum, procedures should cover handling of those loads listed'in Table 3-1 of NUREG-0612.
These procedures should include:
identification of required equipment; inspections and acceptance criteria required before movement of load; the steps arid proper sequence to be followed in handling the Icad; defining the safe path; and other special precautiens."
a.
Summary of Licenst e Statements and Conclusions Tne Licensee has stated that a general operating procedure has been developed controlling loads in the containment building and the handling of other loads not addressed by specific handling procedures.
b.
FRC Evaluation FRC concurs that the Licensee has addressed the scope of the loads requiring procedural control.
liowever, insufficient information has been
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- 2. Inspection and acceptance criteria required before movement of the load e
- 3. Steps and proper sequence to be followed in handling the load.
'4.
Safe load path
- 5. Special precautions.
c.
FRC Conclusion Big Rock Point partial'ly complies with Guideline 2 of NUREG-0612.
In addition to fully comply with the criteria for load handling procedures, the Licensee should verify that procedures contain sufficient detail to adequately define safe load handling (i.e., identification of equipment, inspection and acceptar.ce criteria, sequence of steps, safe load paths, and precautions).
2.1.4 Crane Operator Training [ Guideline 3, NUREG-0613, Article 5.1.1(3)1
" Crane operators shoald be trained, qualifi'ed and conduct themselves in eccordance with Chapter 2-3 of ANSI B30.2-1976, ' Overhead and Gantry Cranes' [7)."
Sorsary of Licensee Staten.ents and Conclusions a.
Tne Licensee has stated that the Big Rock Point crane operator training, qualif2 cation and conduct has been reviewed and found in compliance with the ANSI B30.2 requirements with one exception:
visual examinations.
Big Rock Point crane operators presently are qualified in visual activity to ANSI stendards comparable to their normal job duties whether they are an auxiliary operator of a mechanical repairnan. While the Licensee intends to upgrade those standards as necessary to a.eet crane operator qualification of ANSI B30.2, it is concluded that the present visual testing suffices for the
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FRC Evaluation FRC review of the Licensee submittal, indicates that Big Rock Point complies with" Article 5.1.l(3),of NUREG-0612 based on the Licensee's certificatioT of compliance to ANSI B30.2 and a commitment to upgrade visual
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c.
FHC Conclusions and Recommendations The Big Rock Point complies with Guideline 3 of NUREG-0612.
2.1.5 Special Lifting Devices [ Guideline 4, NUREG-0613, Article 5.1.l(4)]
"Special lif ting devices should satisfy the guidelines of ANSI N14.6-1978,
' Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials' [8).
This standard should apply to all special lifting devices which carry heavy Icads in areas as defined above.
For operating plants certain inspections and load tests may be, accepted in lieu of certain material requirenents in the standard.
In addition, the stress design factor stated in Section 3.2.1.1 of ANSI N14.6 should be based on the combined maximam static and dynamic loacs that could be imparted on the handling device based on characteristics of the crane which will be used. This is in lieu of the guideline in Section 3.2.1.1 of ANSI N14.6 which bases the stress design factor on only the weight (static load) of the load and of the intervening components of the special h'andling device."
a.
Summary of Licensee Statements and Conclusions The Licensee has stated that Big Rock Point has only two of the lif ting yoke used for the fuel shipping cask, cobalt and T-II cask.
Both of these yokes are used for the fuel shipping cask which will not be used in the forseeable future.
The remaining yokes accompany their respective casks and are not controlled by Big Rock Point or tested to the requirements of ANSI N14.6-1978.
Procedures will be revised to provide for visual inspection of the yokes prior to their use to meet in part the ANSI N14.6 inspection criterion.
b.
FRC Evaluation FRC considers the handling of special lif ting devices unsatisf actory relative to Article 5.1.1. (4) of NUREG-0612.
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TER-C5257-440 these devices, ANSI N14.6-1978 should apply to all special lifting devices which carry heavy load in proximity to or over safe shutdown equipment or irradiated f*uel in the spent fuel pool, in the reactor building, and in other plant areas-et Big Rock Point.
The Licensee should evaluate or require subcontractors to evaluate special lif ting utilized at Big Rock Point in accordance with NUREG-0612, Guideline 4.
. Evaluation and review of ANSI N14.6-1978 by FRC has identified several areas which the Licensee should consider:
Section 3 (Design) Although this section deals with specifications for 5,u.pu.c new handling devices, FRC considers certain information to be of1[mportance.
Consequently, the Licensee should evaluate special lifting devices while considering the following design criteria:
Section 3.1:
a.
limitations on the use of the lifting devices (3.1.1) b.
identification of critical ccmponents and definition of critical characteristics (3.1.2) c.
signed stress analyses which denionstrate appropriate margins
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of safety (3.1. 3) d.
indication of permissible repair procedures (3.1.4)
Section 3.2:
use of stress design' factors of 3 for minimum yield strength e.
and 5 for ultimate' strength (3.2.1) b.
similar stress design factors for load bearing pins, links, and adapters (3.2.4) c.
slings used comply with ANSI B30.9-1971 (3.2.5) d.
subjecting materials to dead weight testing or Charpy impact testing (3.2.6)
Section 3.3:
a, consideration of problems related to possible lamellar tearing (3.3.1) b, design shall assure even distribution of the load (3. 3. 4) c.
retainers fitted for load carrying components which may l
become inadvertently disengaged (3. 3. 5) d.
verification that remote actuating mechanisms securely engage or disengage (3. 3. 6)
Section 4 (Fabrication ) Although most standard criteria concerning p,
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Section 4.1:
a.
verify selection and use of material (4.1. 3) b.
compliance with fabrication practice ( 4.1. 4)
- c. ~ qualification of welders, procedures, and operators ~ (4.1.5) d._, provisions for a quality assurance program (4.1.6) e.* provisions for identification and certification of equipment
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( 4.1. 7) f.-
verification that materials or services are produced under appropriate controls and qualifications (4.1.9)
Section 5 (Acceptance Testing, Maintenance and Assurance of Continued Compliance):
Section 5.1:
a.
implementation of a periodic testing schedule and a system to indicate the date of expiration (5.1.3) b.
provisions for establishing operating procedures (5.1. 4) c.
identification of subassemblies which may be exchanged (5.1.5) d.
suitable markings ( 5.1. 6) e.
maintaining a full record of history (5.1.7) f.
conditions for removal from service ( 5.1. 8)
Section 5.2:
Icad test to 150% and appropriate inspections prior to initial a.
use (5.2.1) b.
qualification of replacement parts (5.2.2)
Section 5.3:
satisfying annual load test or insp'ection requirements (5.3.1) a.
b.
testing following major..aintenance (5.3.2) c.
testing after application of substantial stresses (5.3.4) d.
inspections by operating. (5.3.6) and non-operating or maintenance personnel (5.3.7).
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c.
FRC Conclusions and Recom.mendations Tne Big Rock Point does not comply with Guideline 4 of NUREG-0612.
In order to satisf actorily comply, the Licensee should conduct a point-by-point
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Additionally, the Licensee should provide verification that ANSI N14.6-1978 re,quirements are imposed on subcontractors providing temporary special lif ting devices at Big Rock Point.
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TER-C5257-440 2.1.6 Lifting Devices [ Guideline 5, NUREG-0613, Article 5.1.l(5)]
" Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9-1971, ' Slings' [9).
Howeverr in selecting the proper sling, the load used should be.the sum of the static and maximum dynamic load.
The rating identified on the sling should be in terms of the ' static load' which produces the maximum static and dynamic load. k~here this restricts slings' to use on only certain cranes, the slings should be clearly marked as to the cranes with which they may be used."
a.
Summary of Licensee Statements and Conclusions The Licensee has stated that slings used in the handling of heavy loads are inspected periodically and prior to use to comply with the ANSI B30.9-1971 requirements, b.
FRC Evaluation The Licensee has not provide suf ficient information for FRC to evaluate cc.Tpliance with Guideline 5.
Insufficient information is available to verify the following:
1.
slings are installed and used in accordance with ANSI B30.9-1971 2.
Ioad rating of the sling is based upon maximum static and dynamic load 3.
slings are marked with the, static load which produces the maximum static and dynamic Icad 4.
slings which are restricted in use to certain cranes are clearly tr.arked to so indicate.
c.
FBC Conclusions and Recomn.endations The Big Rock Point does not comply with Guideline 5.
In order to comply, the Licensee should perform the following:
1.
verify that slings are installed and used in accordance with ANSI B30.9-1971.
2Property "ANSI code" (as page type) with input value "ANSI B30.9-1971.</br></br>2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
verify that the load used in selecting and marking the proper sling is based upon the sum of the maximum static and maximum dynamic loads.
3.
verify that slings which are restricted in use to certain cranes are clearly marked as to the cranes with which they may be used. bfrankhn Res,earch Center
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TER-C5257-440 2.1.7 Cranes (Insoection, Testing, and Maintenance) [ Guideline 6, NUREG-0612, Ar ticle~ 5.1. l (6) ],
"The crane should be inspected, tested, and maintained in accordance with
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'inside a Pn'R containment may only be used every 12 to 18 months during refueling operations, and is generally not accessible during power operation.
ANSI B30.2, however, calls for certain inspections to be performed daily or monthly.
For such cranes having limited usage, the inspections, test, and maintenance should be performed prior to their use)."
Summary of Licensee Statements and Conclusions a.
Tree Licensee has stated that crane inspection, testing, and maintenance have been conducted in the past in accordance with MIOSRA standards which are comparable to the ANSI B30.2-1976 Chapter 2-2 except for the inspection intervals. Consegaently, the sphere semi-gantry crane inspection intervals have becn revised to meet the monthly and yearly requirements.
Additionally, both the turbine building and Icading dock cranes are inspected on quarterly and yearly intervals due to their infrequent use.
Inspections prior to use are conducted on all cranes in accordance with the standard.
Crane testing per the requirements of the ANSI standard have not been required ~as they only apply to new, reinstalled, altered, extensively repaired or nodified cranes.
The Licensee has stated that Big Rock Point cranes do not fall within these categories.
Crane maintenance as required by ANSI B30.2 has been included as part of the inspection-program.
Although not specifically covered under the ANSI standard, the remaining hoists and lifting devices used in the handling of heavy loads will be reviewed and the applicable inspection, testing, and maintenance requirements will be invoked on them as well. A[ franklin Research Center bJb a tw
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FRC Evaluation FRC comparison of MIOSHA standards Part 18 (" Overhead and Gantry Cranes")
with ANSI B30.2-1976 indicates that the crane inspection, testing and maintenance 5equirementsarecomparable. Furthermore, FRC concurs with the r.
Licensee's decisions concerning inspection intervals.
With reference to the rated load test criteria of ANSI B30.2-1976,~ FRC has been provided insufficient information to evaluate the load test criteria.
The Licensee should provide crane acceptance testing data for review.
Big Rock Point does comply with the maintenance criteria of Guideline 6 of NUREG-0612 based on the Licensee's certification of compliance. With the maintenance cri.eria of ANSI B30.2-1976.
c.
FRC Conclusion Big Rock Point substantially complies with Guideline 6 of NUREG-0612.
In order to fully comply, the Licensee should:
1.
Provide initial load test data on all cranes subject to the general guidelines of NUREG-0612.
2.1.8 Crane resion (Guideline 7, NUREG-0612, Article 5.1.l(7))
"Tr.e crane should be designed to meet the applicable criteria and guidelines of Chapter 2-1 of ANSI B30.2-1976, ' Overhead and Gantry Cranes,' and of C'iAA-70, ' Specifications for Electric Overhead Traveling Cranes' (10).
An alternative to a specification in ANSI B30.2 or CMAA-70 may be accepted in lieu of specific compliance if the intent of the specification is satisfied."
a.
_ Summary of Licensee Statements and Conclusions The Licensee has stated that verification of the crane design will be reviewed by the crane vendors.
However, due to workload priorities, the vendor has indicated that review of the crane design will not commence prior to Mid-October 1981."
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TER-C5257-440 b.
FRC Evaluation FRC is unable to evaluate Big Rock Point compliance to Guideline 7 of 1
NUPEG-061J due to lack of suf ficient inforn.ation.
S 2.2 INTERIM PROTECTION MEASURES The NRC has established six interim protection measures to be imple'mente'd at op'etating nuclear power plants to provide reasonable assurance that no heavy loads will be handled over the spent fuel pool and that measures exist to reduce the potential for accidental load drops to impact on fuel in the core or spent fuel pool.
Four of the six interim measures of the report consist of general Guideline 1, Safe Load Paths; Guideline 2, Ioad Handling Procedures; Guideline 3, Crane Operator Training; and Guideline 6, Cranes (Inspection, Testing, and Maintenance).
The two remaining interim measures-cover the following criteria:
=
1.
Heavy load technical specifications 2.
Special review for heavy loads handled over the core.
Licensee implementation and evaluation of these interim protection measures as contained in the succeeding paragraphs of this section.
2.2.1 Interim Protection Measure 1 - Technical Specifications
" Licenses for all operating reactors not having a single-failure-proof overhead crane in the fuel st6 rage pool area should be revised to include a specification comparable to StanJard Technical Specification 3.9.7,
' Crane Travel - Spent Fuel Storage Pool Building,' for PWR's and Standard Technical Speci ficat ion 3.9. 6.2,
' Crane Travel,' for BWR's, to prohibit handling of heavy loads over fuel in the storage pool until implementation of nieasures which satisfy the guidelines of Section 5.1."
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a.
FRC-Evahta tion-The Licensee has stated that the interim actions have been completed based on their understanding of the direction, except for specific
.y ~ wt-requirements associated with crane spector visual examinations.
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TER-C5257-440 b.
FFC Evaluation Further investigation by the FRC to clarify the above response revealed
.s.
that Big Rock, Point refueling procedures require the use of a 24 ton fufl transfer casil to transfer spent fuel between the reactor vessel and the S
o storage pool. Therefore, the movement of this heavy load over the fuel storage pool is necessary in order to refuel.
Consequently, Big Rock Point has no technical specifications prohibiting the handling of heavy loads ever the fuel storage pool.
However, the Licensee has added a safety sling between the crane and the cask to prevent the dropping of the load.
This device has not been evaluated to determine if it meets the intent of this single-failure re C 4.
proof crane Acy a g
c.
FRC Conclusions and Recommendations ss ff va n o
Available information is suf-Ho-ient to determine if Big Rock Point cc plies with Interim Protection Measure 1.
Due to the nature of the 6RP refueling process, it appears necessary that heavy loads be moved over the spent fuel pool.
The BRP fuel transfer cask load handling system should be evaluated to establish whether it meets the intent of the single-failure proof crane specified by the NRC staff.
2.2.2 Interim Protection Measures 2, 3, 4,
and 5 - Administrative Controls
" Procedural or administrative measures [ including safe load paths, load handling procedures, crane operator trainihg, and crane inspection)...
can be accomplished in a short time period and need not be delayed for comp]etion of evaluations and modifications to satisfy the guidelines of Section 5.1 of [NUREG-0612)."
a.
Summary of Licensee Statements and Conclusions 1
Summaries of Licensee statements and conclusions are contained in oiscussions of the respective general guidelines in Sections 2.1.2, 2.1.3, i
2.1.4, and 2.1.7. respectively.
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FRC Evaluations, Conclusions, and Recommendations FRC's ev,aluations, conclusions, and. recommendations are contained in discussions of the respective general guidelines in Sections 2.1.2, 2.1.3, 2.1.4, and 2.1.7.
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2.2.3 Interim Protection Measure 6 - Special Reviews for Heavy Loads Over the Core "Special attention should be given to procedures, equipment, and personnel for the handling of heavy 3 cads over the core, such as vessel intarnals or vessel inspect ion tools.
This special review should include the following for these loads:
(1) review of procedures for installation of rigging or lifting devices and movement of the load to assure that suffic~ient detail is provided and that instructions are clear and concise; (2) visual inspections of Icad bearing components of cranes, slings, and special lifting devices to identify flaws or deficiencies that could lead to failure of the component; (3) appropriate repair and replacement of defective components; and (4) verify that the crane operators have been properly trained and are familiar with specific procedures used in handling these loads, e.g., hand signals, conduct of operations, and content of procedures."
Sammary of Licensee Statements and Conclusions a.
Tne Licensee has stated that the interim action have becn completed based on their understanding of the direction, except for the specific requirements associated with crane operator visual examination.
b.
FRC Evaluation Big Rock Point complies with Interim Protection Measure 6 of NUREG-0612 based on the Licensee's certification of compliance.
However, the Licensee should insure that appropriate documentation is available to document this one-time detailed review of procedures, equipment, and personnel training involved with the handling of heavy loads over the core.
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CONCLUDING SUv.".ARY This summary is provided t,o consolidate the conclusions and recommenda-tions of Secflon 2 and to document FRC's overall evaluation of the handling of -fg A m t, &cr f%ab heavy loads at the for4ey Nuclear Plant.
It is divided into two sections dealing with general provisians for lead handling at nuclear power plants (NUREG-0612, Ar ticle 5.1.1) and the staff recommendations for interim protection, pending complete implementation of the guidelines of NUREG-0612 (NUREG-0612, Article 5.3).
In each case, recommendations for additional Licensee action, and additional NRC staff action where appropriate, are provided.
3.1 GENERAL PROVISIONS FOR IDAD HANDLING The NRC staff has established seven guidelines concerning provisions for handling heavy 1 cads in the area of the reactor vessel, near stored spent fuel, or in other areas where an accidental load drop could damage safe shutdewn systems. Compliance with these guidelines is necessary to ensu:e that load handling system design, administrative controls, and operator training and qualification are such that the possibility of a load drop is very small for the critical functions performed by cranes at nuclear power plants.
These 99, 6 c,cA_ a:. s t guidelines are partially satisfied at the (Earley Nuclear Plant.
This conclusion is presented in tabular form as Table 3.1.
Specific recommenda-tions for achieving full compliance with these guidelines are provided as follows:
Guideline Recommendation
(
l a.
Verify that safe load paths and restricted areas are adequately I
marked to provide visual references for both operations and supervisory personnel.
Verify that procedures contain sufficient detail to adequately 2
a.
ib$ino safe load handling (i.e., identification of equipment, l
inspection and acceptance criteria, sequence of steps, safe load paths, and precautions.
3 (Big Rock Point complies with the guidelines.)
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,,1 p Capacity sete Baed Crane operator special Lifting Crane. Test Technleal Specie!
gearr inada (tonal Patm e Proceduree __ Tretnine Devices 818nes and Inspection Crone Dealgg Specificattone attention
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l malgat intette Interia os Guideline 1 Gesideline 3 Guideline 3 Culdeline 4 Guideline 5 Cuide11s.e 6 Guideline 1 steasure 1 sesseure 6 Capacity Safe Imad Crane Operatoc Special L11 ting Crane - Test Technical Special i
as u v y tuMs (tone).
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a.
Conduct a point-by-point review of all special lif ting devices against the criteria of ANSI N14.6-1978
~
b.
Verify that ANSI N14.6-1978 requirements are imposed on subcontractors providing special lifting devices.
M, 5
a.
Verify that slings are installed and used in accordance with e
b.
Verify that the load used in selecting and marking the slings is based upon the sum of the maximum static and maximum dynamic loads.
c.
Verify that slings that are restricted in use to certain cranes are clearly marked as to the crane (s) with which they may be used.
Provideinitialloadtestdateonallcranessubjecttofhe 6
a.
general guidelines of NUREG-0612.
7 a.
Provide information concerning the cranes design criteria for those cranes subject to NURIG-0612 general guidelines.
Additionally, the Licensee should review load handling systems at BRP for NURIG-0612 applicability including the following actions:
a.
Modify the heavy load handling program at ERP to include the Screen House Trolley Verifythatthemechanicalblocksonthe)hecontamination/k!oistante b.
permanently installed physical limitations to hoist travel.
Verify that the dkuipment dock brane operates in the vicinity of the c.
3cading76$$5 only under conditions when containment integrity is not required.
3.2 INTERIM PROTECTION The NRC staff has established (NUREG-0612, Article 5.3) that certain measures should be initiated to provide reasonable assurance that handling of heavy loads will be performed in a safe manner until final implementation of the general guidelines of NUREG-0612, Article 5.1 is complete. Specified measures include the implementation of a technical specification to prohibit the handling of heavy loads over fuel in the storage pool; compliance with Guidelines 1, 2, 3, sand 6 of NURF.G-0612, Section 5.1.1; a review of load handling procedures and operator training; and a visual inspection program, including component repair or replacement as necessary of cranes, slings, and bd Frankhn,n, search Center Re a %..-.
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ER-C5257-440 r,pecial lif ting devices to eliminate deficiencies that could lead to component failure.
FRC's evaluation of information provided by the Licensee indicates
~
that the fo11owing actions are necessary to ensure that the staff's measures
.%)L for interim < protection;s,t. b iat Ea44+y Nuclear Plant are met M,
o Interim Measure Recommendation 1
,4 Establish the potential equivalency of the transfer ca'sk redundant support assembly to a single-failure-proof crane.
2, 3
_'7M Implement the recommendations of Guidelines 1 and 2 identified in Section 3.1 4
(Big Rock Point complies with this interim protection measure) 5 y4 Implement the recommendations of Guideline 6 identified in Section 3.1 6
(Big Rock Point complies with this interim protection measure).
3.3 S W.v.ARY NRC's general guidelines and interim protection measures of NUREG-0612 have been partially complied with at the Big Rock Point Naclear Plant.
Several programs have been installed which comply with NRC staff guidelines, including operator training and interim inspections.
In $rder for the Licensee to fully comply with NUREG-0612, Licensee action is required on the remaining general guidelines and interim actions.
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TER-C5257-440 4.
REFERENCES 1.
NUREG-0612 "ControE of Heavy Icads at Nuclear Power Plants" y
~
NRC, July 1980 2.
V. Stello, Jr. (NRC)
Letter to all licensees.
Subject:
Request for Additional Inforrnation on Control of Heavy Icads Near Spent Fuel NRC, 17 May 1978 C., 4 323 LcPc )
3.
NRC d. > => s tas hwx^-
Letter to aff. Mc:k1.Pbfot Request for Review of Heavy Load Handling at Big Rock Point
Subject:
H ent-22 December 1980 4.
G. C. Withrow (CPC)
Letter to D. M. Crutchfield (NRC)
Control of Heavy Loads - Big Rock Point
Subject:
10 June 1981 5.
D. P. Hoffman (CPC)
Letter to D. M. Crutchfield (NRC)
Control of Heavy Loads - Big Rock Point Sabject:
1 July 1981 C C /)
6.
T. C. Sordine Le t ter to D. M. Crutchfield (NRC)
Control of Heavy 1 cads - Big Rock Point Sabject:
u sc-b btss%t 7.
A':SI B30.2-1976
" Overhead and Gantry Cranes" MOSI N14.6-1978 B.
"Stancard for Lif ting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials" 9
9.
ANSI B30.ri-1971
" Slings" CMAA-70 OI N f
12.
" Specifications for Electric Overhead Traveling Cranes"
. 4bbhu l'ranklin Reseaich Center A r%,.. n. r..e M
'/
BNWo 60 UNITED STATES i
f y
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20065
-l j/
February 6,1976 a t t* e r, y g rre 1 c n Docket No. 50-155 Nt!Grq LKIN5;,g Consumers Power Company ATIN:
Mr. Ralph B. Sewell Nuclear Licensing Administrator 212 West Michigan Avenue Jackson, Michigan 49201 Gentlemen:
By letter dated February 4,1974, we requested that you provide an analysis and relevant information relating to possible damage in the event of a postulated spent fuel cask drop accident. The information was needed to study the consequences of the accident and to see if design or procedural changes would reduce the probability of such an 3
accident. Your 1ctter of July 1,1974, submitted the study.
Answers 4
to our questions of June 30, 1975, were provided in your letter dated August 25, 1975. You submitted additional information by letter on O
January 22, 1976.
Your analysis shows that with your present procedures and design a drop of either the fuel shipping or fuel transfer cask into the spent fuel pool could result in a leak larger than the pool makeup capability.
Q)
For this reason you proposed placing an energy absorbing impact pad in the pool. You also proposed modified cask handling and routing proce-dures to minimize the probability of a cask drop on spent fuel or safety related equipment.
In addition, you proposed crane modifications and 7
described crane testing which should improve crane reliability. You N
advised that the planned modifications will be impicmented within one year from the date of our approval. You also stated that the procedure This O
changes are being impicmented for the current fueling outage.
schedule is acceptable.
S 544715 imu?w=%L
C Consumers Power Company February 6,1976 We have evaluated your analysis and the additional information provided. -,
We have concluded that the proposed modifications and procedure changes,
. shen completed, will adequately minimize the probability and potential consequences of a cask drop accident. A copy of our related Safety Evaluation is enclosed.
Sincerely,
/
Karl R. Goller, Assistant Director for Operating Reactors Division of Operating Reactors
Enclosure:
Safety Evaluation cc w/ enclosure:
g See next page 5
O c
T N
O 544716
Consumers Power Company February 6,1976 cc w/ enclosure:
Mr. Paul A. Perry, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Peter W. Steketec, Esquire Freihofer, Cook, Hecht, Oosterhouse and De Boer Union Bank Building, Suite 950 Grand Rapids, Michigan 49502 George C. Freeman, Jr., Esquire llunton, Williams, Gay 6 Gibson 700 East Main Street Richmond, Virginia 23212 Charles F. Bayless C3 Of Counsel Consumers Power Company 212 West Michigan Avenue C)
Jackson, Michigan 49201 Charlevoix Public Library 107 Clinton Street Charlevoix, Michigan 49720 CD ON
%T C4 C3
+.
l 544717
d UNITED STATES j
NUCLE AR REGULATORY COMMISSION y
y j
WAsHiNcToN. o. c. zones e
s.,...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION.
SUPPORTING ACCEPTANCE OF CASK DROP ACCIDENT ANALYSIS CONSUMERS POWER COMPANY BIG ROCK POINT PLANT DOCKET NO. 50-155 INTRODUCTION By letter dated February 4,1974, we requested Consumers Power Company (CPCo) to provide an analysis and relevant information required to determine the potential consequences of a fuel cask drop caused by a W
system failure at the Big Rock Point Plant. Both the fuel transfer cask and the fuel shipping cask were to be considered in this analysis.
O We further requested that if the evaluation indicated changes were N
necessay to protect plant structures, systems, or components important to safety or to prevent damage to irrad.ated fuel, information on the O
plant modifications should be provided as well as the schedule for design, fabrication and installation of any modifications.
In response to our Februay 4,1974 letter, a study performed by Bechtel Corporation for CPCo was filed by CPCo on July 1,1974, with CD supplemental information filed on August 25, 1975, in response to our request dated June 30, 1975. Additional information was provided in CPCo's letter of January 22, 1976. This Safety Evaluation presents the results of our review of the information provided relative to the T
postulated fuel cask drop accidents at Big Rock Point.
DISCUSSION / EVALUATION O
Areas of particular concem which we asked to be included in the CPCo evaluation and which we have reviewed are as follows:
1.
Potential effect on spent fuel storage pool - The licensee was l
asked to determine whether dropping a cask over the storage pool could damage the pool floor to the extent that adequate makeup water capability could not be assured or resultant flooding could l
cause critical systems to become inoperable. As analyzed by CPCo 1
544718 m
k/
a free fall of the fuel shipping cask or the fuel transfer cask could cause a crack of the spent fuel pool bottom slab and the pool liner plate may be perforated.
Slab penetration is' not considered possible.
In such an event, the rate of pool drainage into the 585' - 6" level of containment could exceed the pool makeup rate. The proposed modification to prevent structural failure of the spent fuel pool base slab in this area is the installation of a cellular energy absorption assembly. We have reviewed the proposed modifi-cation and have determined that it constitutes an acceptable method for eliminating this potential failure.
CPCo stated in its letter of July 1,1974, that the proposed modifi-cations to the spent fuel pool would be completed within one year from the date of our concurrence with the proposed modifications.
We consider this schedule to be acceptable.
- -)
2.
Potential effect on spent fuel
'Ihe licensee was asked to determine O
whether a cask can be positioned over or near spent fuel so that a cask could drop on the spent fuel or be deflected onto spent fuel.
N As analyzed by CPCo in its July 1,1974 submittal, the potential exists for fuel damage from a direct drop of tha transfer cask or O
the shipping cask onto fuel assemblies located in the fuel pool "imilar damage could occur as the result of either storage racks.
of these casks if dropped to strike the edge of the spent fuel pool and then be deficcted onto the fuel racks.
CPCo proposed relocation of fuel racks and a different cask routing scheme to limit cask movement over the spent f uel pool. The fuel transfer cask is the only cask allowed over the fuel racks by new v
procedures. 'Ihese changes greatly reduce the probability of a cask drop onto the spent fuel. Although the possibility exists for the N
transfer cask dropping on the spent fuel stored in the pool, the Probability is very low. 17hTaadition, safetyis11nWdded to.thei itransfer caskf as. a result.of a previous cask drop-analysi's. willi O
' prevent _ dropping the cask if. a crane _, failure should occurJ Even so, we have calculated the potential magnitude of offsite radiation doses for the accident. We assumed that 28 fuel assemblies (1/3 of the core) in the spent fuel pool were damaged by a cask drop accident.
Based on the meteorological assumptions and the models given in Regulatory Guide 1.35, we concluded that the site boundary doses at Big Rock Point would be within the guidelines of 10 CFR Part 100 for a design basis cask drop accident.
544719
3 W}(a The site boundary dose a would be within acceptable limits with the containment ventilation system operating. CPCo advised
'in its January 22, 1975 letter that procedures would require operator action to shut the isolation valves on receipt of a containment area high radiation alarm in the control room.
However, to provide an additional level of safety, the need for operator action will be eliminated by installation of an automatic containe-nt purge system isolation feature. CPCo has agreed to complete these modifications within the same time frame as the modifications to the crane and installation of the impact pad.
We have reviewed the proposed relocation of fuel racks within the pool and the proposed restrictions on cask movements over the pool. Both proposals will reduce the probability of a cask drop on the spent fuel and are acceptable.
3.
Potential effect on critical systems and equipment - The licensee was asked to determine whether the casks were moved over systems y.
or equipment important to safety which could be damaged by a cask drop. They were asked to consider the capability of the floors to o
protect equipment or systems important to safety which are located below the floor.
CPCo's analysis showed that certain safety related N
systems might be damaged if their existing cask routing and pro-c dures were used and a cask dropped.
For these reasons they pro-O posed cask routing changes, craae modifications and procedural changes to minimize the accident potential.
We reviewed the licensae's July 1,1974 submittal and determined that additional information was necessary to complete our review.
O CPCo was requested to show that the safety margins for the spent y
fuel pool floor slabs, beams and protective structures for Category I equipment are of the sare order of magnitude as required for v
Category I structures outside of containment. We also requested more infomation relating to the energy absorbing impact pads pro-N posci for the floor area of the spent fuel pool.
O Our review of CPCo's July 1,1974 submittal and the answers provided by their letter of Augast 25, 1975, in response to our questions is complete. We find that the design criteria and design methods used are those of the Bechtel Topical Report " Design of Structures for Missile Impact" BC-TOP-9-A, Revision 2,1974 previously reviewed and approved by NRC. The design of the energy absorbing impact pad is acceptable.
Based on the considerations discussed above we conclude that the potential effect of cask drop accidents on critical systems and equipment is minimal and acceptable.
.544720
..o 4.
Integrity of handling equipment for the spent fuel cask - Re licensee was asked to evaluate-the existing crane facilities as
,they relate to fuel cask handling and movement. He CPCo's st:bmittal of July 1,1974, proposed modifications to upgrade the cranes with regard to brakes, limit switches, thermal over-loads and redundant controls. Rey proposed to rigidly control travel paths and handling procedures which would restrict cask and crane movements.
In addition, CPCo has a program of opera-tional tests and inspections of the cranes and associated handling equipment. His program will give added assurance that the fuel handling equipment is reliable.
We conclude that the proposed modifications and program of opera-tional tests and inspections provide adequate assurance of the integrity of the handling equipment.
CONCLUSION Based on our review of the analysis, the descriptive information pro-vided and CPCo's response to our questions, we have concluded that the provisions for preventing a postulated fuel transfer cask or spent o
fuel shipping cask accident at the Big Rock Point Plant are acceptabic N
when the proposed changer are completed.
O Date: February 6,1976 co O'
T N
O D
544721
i 4.'
g.'
Lawrence Liveimore National Laboratory A
~
January 28, 1982 SM 82-21 FIN A0436 Docket # 50-155 Mr.4 William T. Russell, Branch Chief Systematic Evaluation Program Branch Division of Licensing Office nf Nelear Reactor Reg.
wasiangton, D.C.
20555
Dear Bill:
We have finished the independent seismic analysis for the reactor building of the Big Rock Point Nuclear Power Plant, FIN A0436, Task 38. Attached are the results and conclusions of this analysis in draft format.
Should you or your staff have any comments, please call us as soon as possible since the final completion date for this task is only one month away (February 28, 1982).
Sincerely, Thomas A. Nelson Project Manager Structural Mechanics Group Nuclear Test Engineering Division TNumg 0286m Enclosure
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i SEISMIC REVIEW 0F THE BIG ROCK POINT NUCLEAR POWER PLANT, CONTAINMENT SHELL STRUCTURE AS PART OF THE SYSTEMATIC EVALUATION PROGRAM o
i 1
PREPARED BY:
3 C. Y. LIAW EG8G/ SAN RAMON OPERATIONS 1
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SU'f.ARY As part of the safety assessments of Big Rock Point Nuclear Power Plant facilities, the containment shell structure of Big Rock Point was analyzed using site-specific ground response spectra with peak ground accelerations of 0.11 g in the horizontal dire' tion and 0.07 g in the c
vertical direction.
The preliminary results of the seismic and dead-load
. Ianalyses are included in this report.
Since the intent of this analysis is a' to perform an independent evaluation, results were compared with the '
licensee's analyses.
Similiar conclusions were obtained which include the -
following:
Stresses in the shell structure are much lower than the allowabic tensile' stress of the steel.
The containment structure has sufficient factors of safety against overturning, sliding,- and twisting.
The problem of clastic stability of the shell is, as expected, the more critical factor under seismic consideration.
- However, a minimum factor of safety of 1M is available to resist the combined compressive stresses iiduced by seismic and dead loads.
2.
ANALYSIS OF THE CONTAINMENT SHELL STRUCTURE The objective of this analysis is to perform an independent seismic evaluation of the containment shell structure using site specific spectra of the Big Rock Point site.
The spectra were based on those recommended in Reference 7 with peak ground accelerations of 0.119 in the horizontal directibns and 0.07 g "in the vertical direction.
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In accordance with the intent,of the Systematic Evaluation Program-(SEP),'the structural review is not based on demonstrating compliance with specific criteria in the Standard Review Plan or Regulatory Guides, but rather the seismic resistance of the structure is' compared to the licensing criteria.
2.1 DESCRIPTION
OF STRUCTURE The containment shell structure, which is part of the reactor.
building, is a spherical steel shell,130 feet in diameter and having a thickness varying from 0.702. inch to 0.774 inch.
The containment shell encloses the reinforced-concrete intern.al structure which houses the nuclear steam supply system (NSSS), spent fuel storage pool and emergency
' condenser, as shown in Figure 1.
4 The lower portion of the steel containment is embedded in the concrete,
foundation of the internal structure.
The foundation has the shape of an inverted spherical segment approximately 7 feet. thick. To provide smooth transition of the supporting edge where the steel shell embedded into the concrete, foundation, an 8-foot-deep sand-filled cavity was constructed around the edge of the foundation.
The Big Rock Point site is situated ir a limestone area.
The soil foundation of the reactor building can be. idealized as being composed of a layer of very dense glacial till on top of several layers of limestone (Reference 2) l 2.2 MATliEMATICAL MODEL l
The major concern of this analysis is the structural integrity of the containment shell.
However, since the shell structure and the internal structure share a common foundation which is situated on soft soil, the -
soil structure interaction effect is an important consideration.
The shell and the massive internal structure cannot be separated into two independent structures.
Therefore, the modeling approach adopted in this analysis is
~
to include a simplified model of the internal structures in the same model l
withsa' de. tailed rep,resentation of the shell to calculate the shell's l --
- scismic and dead-load responses.
The internal structure was modeled in
' sufficient detail to include its i'teraction effects with the shell.
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The containment shell is an axisymmetric structure, but the internal l
structure is asymetric with an eccentric mass and stiffness distribution
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w' ich inay cause significant torsional responses under seismic loads.
It h
was therefore decided to model the coupled shell-internal structure as a three-dimensional structure without utilizing the benefit of the symmetric shell.-
In determining a proper representation of'the shell structure, we considered the fact that under seismic and dead loads, the stress in this
- - kind of shell is normall'y very low, well below the yield stress of the 4material.
Thus, compressive membrane stress under elastic stability criteria is a more' critical consideration.
Therefore, a three ' dimensional shell element mesh was constructed with elements sized to capture mainly membrane behavior.
A'il uniform shell thickness of 3/4 inch was assumed for the model used in the analysis. When the information about the exact shell thickness was available later, a new model incorporating these thicknesses was constructed and a confirmatory cigenvalue analysis was performed.
The discrepancy between the two models in terms of modal frequencies was less than 2%; therefore, ine original model based on 3/4-inch thickness was considered to be adequate.
4 As mentioned previously, the idea of modeling the internal st'ructure is to capture its interaction effects on the shell.
Lumped masses connected by beam elements are adequate for the purpose of including the mass and stiffness effects.
The properties of the internal structure are based on information available in Reference 2.
The containment shell is supported at the 584.5-foot elevation by a concrete foundation consisting of a rigid di'sk connected to the shell around the edge and to the internal structure at the center.
The rigid disk is supported by six springs to represent the three translational and three rot'ation'ai degrc'es of freedom of the concrete foundation resting on the so'il.- The spriog, constants' and the associated radiation damping
~
' coefficients were estimated based on an~ eta'stic half space assumption for the soil.
The damping values for horizontal and vertical directions were taken as 75 percent.of the theoretical values in keeping ~with the recommendations of the Senior Scismic Review Team (Refe'rence 8).
A
r-Also,'according to Reference 8, three different soil modulus conditions are to be considered in the structural malysis because of uncertainty in soil properties.
The suggested three conditions are:
- 1) 50 percent of the modulus corresponding to the best. estimate of the large-strain condition, 2) 90 percent of the modulus corresponding to the best estimat'e of the low strain condition, md' 3) a best-estimate shear modulus.
4 4
The subsurface conditions of the Big Rock Site may be idealized as being composed of approximately 30 to 40 feet of medium dense to very dense glacial till on top of several layers of limestme (Reference 2).
The l
foun datial of the reactor building is an inverted spherical concrete dome approximately 7 feet thick embedded in the, soil.
The glacial till has a 6
shear modulus of 14.2 x 10 psf, based on a shear wave velocity survey 6
(Reference 2).
The limestone layers have moduli varying from 79.6 x 10 psf 6
to 138.5 x 10 psf, between the 553-foot and 413-foot elevations.
The surf ace grade is at elevation 593 feet.
To properly represent the soil-structure interaction effect in the model, several approaches were considered.
The first approach used the spring constants and the associated radiation damping coefficients based on an elastic half-space assumption for the soil (Reference 3)'. An average shear modulus was used to represent the layer structure.
The shear modulus 0
for low strain was estimated to be 68.2 x 10 psf in this case.
For the l arge-strain conditial case ( a = 5 x 10-5), a reduction f actor of 82 percent was estimated for hard glacial till (Reference 10) which gave a 6
shear modulus of 55.8 x 10 psf.
The corresponding 50-percent best 8
6 estimate of the large-strain case had a r. hear modulus of 27.9 x 10 psf.
Trie ;.second approa'ch was based on the method suggested in Reference 11,
~ hich gave st'iffness ' coefficients for c~ircular footing embedded in a layer w
of soil al top of the bed ock. Th,e coefficients were given in. terms of.
j equivalent el'astic half-space solution.
The low-strain shear modulus for
~
6 the glacial till in this case was estimated to be 49.7 x 10 sf.
The' corresponding large-strain and 50-percent large-strain cases then have 6
6 shear moduli of 40.8 x 10 sf and 20.4 x 10 psf, respectively.
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From the above two approaches, set-shear moduli were selected. Their values and the corresponding spring caistants and damping ratios together with mass density aid Poisson's~ ratio are listed in Table 1.
Another approach, suggested in Reference 9, gave the spring constants.
6
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The horizontal spring constait was calculated 'to be 1.13 x 10 h/in. which 6 i is close to Case B, the rocking spring constant was 1.53 x 10 k "*/ rad which is close to Case C.
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Based on the above discussion on these different approaches, the values listed in Table 1 can be considered as a reasonable bound for the soil springs.
Therefore, three models were con ~structed using the listed values.
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The said cushion arouid the shell edge was modeled by equivalent
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elastic springs in the direction normal to the shell surf ace.
The equivalent spring constants were calculated from the data given in Figure Al-7 of Reference 2.
The structural damping ratios were assumed to be 2% aid 3% for the steel shell aid coicrete internal structure, respectively.
These values are those suggested in NUREG/CR-0098 (Reference 4) for welded assemblies and reinforced, concrete structures subjected to stresses below one-half the yield point.
Figure 2 shows the mathematical model for the coupled shell, intenial structure, and s. oil system.
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Table 1.
Properties of soil springs.
l CASE A
'B
'C 6
6 6
G (psf) 20.4x10 68.2x10 34.1x10 d
t SPRl!G CONSTRAINTS 6
6 6
VERTICAL (k/in.)
0.57x10 1.90x10 0.95x10 0
6 6
HORIZONTAL (k/in.)
. 0.40x10 1.35x10 0.68x10 11 11 1I ROCKING (k "*/ rad)
~1.16x10 3.86x10 1.93x10 i
II 11 11 TORSION (k "*/ rad) 1.27x10 4.25x10 2.12x10 DAMP!fG RATIOS (%)
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VERTICAL 57 57 57 H3RIZONTAL 33 33 33 ROCKING 11 11 11 TORSION 9
9 9
i lb.sec /in.4 2
50ll !4 ASS DENSITY:
2.18x10-POISSON'S RATIO:
0.45.
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tiathematical model for Big Rock Point containment structure.
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s 2.3 METHOD OF NiALYSIS The computer code used for the analysis is a version of SAPlV modified by Lawrence Livermore National Laboratory.
The seismic responses were computed by the response spectrum method. The undamped n atural frequencies and mode shapes of the soil structure system dere calculated, then the composite modal damping ratios of each mode were computed using the
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stiffness proportion damping method.
Site specific spectra based on t'he
' computed composite modal damping ratio were input in two horizontal and the 4
vertical directions. Table 2 lists the horizontal spectral values at 5%
damping for the site specific spectrum of the Big Rock Site. The spectral values at different frequencies for different damping values were calculated according to the formula given in Reference 7.
The vertical The spectral value was taken to be two-thirds of the horizontal value.
peak ground acceleration was 0.11 g for the horizontal directions and 0.07 g for the vertical direction.
The modal responses for each direction were combined by the square-root-of-sum-of-squares (SRSS) method. The total responses of the structure were obtained by combining the absolute values of responses to each input direction with combination factors of 1.0 md 0.4 for the two horizontal directions and 0.4 for the vertical direction as suggested in Reference 4.
The major horizontal direction (i.e., the one with the f actor 1.0) was input in each of the two perpendicular directions with respect to the structure, md the combination with higher response was used in the final results.
The above malytical procedure was repeated for the three cases with different soil shear modulus.
2.4 RESULTS
' Twenty.Jnodes,for each of the three. soil cise were extracted and included -in the seismic analyses.
The-modal frequencies and the composite
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modal' damping ratios of the first ten modes md the twentieth' mode are presented in Table 3.
Among the three models which have different soil shear modulus, the first five modes are very similar.
The first mode is an intemal structure mode, with the major response occurring at the steam
Table 2.
Horizontal site specific spectrum.
2 Pseudo Spectral Acceleration (cm/sec )
Period (5% Damping)
O.03 102.50 4
0.04 122.29 0.05 130.19 O.08 152.05 0.10 179-69 0.20 213.50 0.30 201.96 O.40 171.68 1.00 122.90 Table 3.-
Modal frequencies and composite damping ratios.
!ODEL 1 IODEL 2 K) DEL 3 Mode Freq.
Damping Freq.
Damping Freq.
Dampin g (Hz)
Ratio (%)
- (Hz)
Ratio
.(Hz)
Ratio 1
5.00 4
5.15 3'
5.08 3
2
.. 7.13
. 13 7.98 2
7.52 3
3 7.22 11 7.98 2
7.54-3 4
- 8. 85.
.6 11.30
'll 9.60 14
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- 11. 71- -..
11 9.85 14 12.51 20
.18.39 10
.15.48 41 6
15.91 17 20.07 3
18.34 8
7 8-16.48 20 21.83 31 19.06 11 9
16.56 12 '
24.47 16 20.00 17 10 19.58 5
25.26 17
- 20.36 20 20 32.07 3
32.14 6
32.12 5
t.
drum enclosure. The second aid third modes, which have nearly the same frequencies are containment shell modes in two orthogonal directions. The fourth aid fifth modes, also with very close frequencies, are combined shell and internal structure modes.
Several typical mode shapes of Model B are shown in Figures 3 through 7.
The sixth mode is 'a vertical mode and the seventh mode is a torsional
' mode of the shell structure.
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The maximum shell stresses due to seismic and dead loads are~ listed in Table 4.
The maximum stresses are quite uniform in the hoop direction.
In the meridional direction, they have typical distribution as shown in Figure 8 for Model B.
The maximum combined seismic and dead-load stresses in the shell are in the order of 650 psi for tension,1200 psi for compression, and 800 psi,
for shear. They are all very low compared to the material yield stress.
Therefore, the only critical condition for an Safe-Shutdown Earthquake event is either compressive or torsional shell buckling. There are very few generally accepted buckling criteria for a sphere under seismic loads.
Reference 5 gives a critical load on a thin truncated shell under axial load, F = 0.277 (2 Et cos2) 2 which is based on 170 tests aid will'give 95% confidence in at least 90% of-EMP
.the coces carrying more than this critical load..If applied to this shell, tht 8@b'rmula gives al equivalent uniform meridional compressive stress of
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Mode shape--first mode, Model B.
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Mode shape--second mode, Model B.
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tODEL Dead Stress (psi)
A B
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+341 4246-4 455 193
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Maximum membranc stresses in the steel shell.
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Stress distribution in the meridional direction, Model B.
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Reference 6 suggests a critical pressure for-sphere under vacuum 2
P=Pcl [0.14 + 3.2/ A ],1 > 2 A = [12(1 v )31/4 (R/t)1/22 sin 6/2 2
2 1/2 E(t/R)2 Pcl = 2/[3(1 v )3 t
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In this case, P = 4.55 psi which gives an allowable shell compressive stress.of 2364 psi.
The above two allowable compressive stresses (5660 psi-and 2364 psi) are for the case where the shell is under uniform compression. Under seismic excitation, only a portion of the shell will be subjected to compression.
These criteria, if applied to this analysis..are very conservative.
Therefore, the shell is considered to have a factor of safety at least 1.9 against buckling.
For torsional -buckling, Reference 5 gives a formula which is applicable to thin truncated conical shell.under torsion.
If-it is applied to this case, th~e critical shear stress is 4780 psi, which would result in a safety factor of 6.
Again this is a conservative estimate.
The overall stability of the shell structure is also of' concern during an SSE.
From the response spectrum analysis, the maximum overturning l
moment, torsional moment, and' sliding force can be obtained from the soil' spring forces.
From-the structural dead weight, the vertical seismic
. force, and an assumed friction coefficient of 0.45, the. resisting moment and forces can be calculated.
The resulting factor of safety from these
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calculatihns is 7 for " overturning and torsion, and 4 for sliding.
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3.
COMPARISON WITH LICENSEE'S ANALYSIS The major differences between the licensee's alalysis (Reference ' )
2 for the containment structure and this alalyses are listed in Table 5.
The licensee's alalysis used the R.G.1.60 spectra with higher peak ground and spectral accelerations than the site specific ' spectrum..
The licensee's model for the containment structure included two parts:
a~ 1) a concrete internal structure with soil springs and a single lumped mass for the shell, ald 2) a finite elemen.t shell model 'for the containment shell without the internal structure and soil springs. The analysis was performed in two steps. First, Model 1 was aialyzed by the time history method using an artificial earthquake generated from the R.G.1.60 spectrum, including soil structure interaction. Second, the floor response spectra developed from Model 1 at the shell base were input to Model 2, and a response spectrum analysis of Model 2 was performed. The responses of Model 2 were used as the basis for structural evaluation.
In estimating the soil. spring coefficients, the licensee's aialysis considered the embedment effect of the foundation and included some correction due to the frequency dependent characteristics of the soil springs. However, only one.
set of spring constants were used in the licensee's analysis.
The radiation damping values of soil were also reduced by a.f actor of one-half in the licensee's analysis.
In our alalysis, the theoretical damping values for traislational motions were reduced to 75 percent.
Both analyses predict frequencies of the dominant shell modes at about i
8 Hz.
The membrane seismic stresses from both analyses are in proportional
~
to their spectral values, but the licensee's analysis predic;ed slightly higher ' bending stress because of their more refined finite element model.
~~
However,Ithe critical condition of the shell is controlled by its elastic
~
stabilitywh.ich,depenps01 the shell membrale stress.
6 S
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r EG8G ANALYSIS LICENSEE'S ANALYSIS Site Specific Spectrum R.G'.1.60 Spectrum INPUT Peak Ground 0.11 g (H) 0.12 g (H)
Acceleration 0.07 g (H) 0.08 g (V)-
MODEL Structure ONE MODEL TWO MODELS Model 3-D shell elements 1.
Single lumped. mass for the shell, for the~shell.
Lumped masses-stick Lumped masses-stick 4
beams for the intemal, beams for the intemal and soil springs.
and soil springs.
2.
Refined shell model axisymetric shell.
MATERI AL DAMPIfG Steel 2%
4%
Con crete 3%
7%
S0ll SPRItG No embedmmt effect.
Embedment and frequency Variation of shear dependent ~ effects, damping
- modulus, reduction. No variation of shear modulus.
METHOD OF ANALYSIS Response spectrum Time history method analysis (RSA) for Model 1. 'RSA'for Model 2.
~
Table 5.
Differences between the EG5G analysis and the Licensee's analysis.
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- r-s REFERENCES 1.
"Fi'nal Hazard Summary Report of the Big Rock Point Nuclear Generating S'ation,", NRC. Docket 50155-2, Consumers Power Company, Jackson, t
2.
"Scismic Safety Margin Evaluation, Big Rock Point Nuclear Power Plant Facilities", Volume I & II, Consumers Power Company, Jackson, Michigan, 1978.
3.
Richart, Jr., F. E., Hall, Jr., J. R., and Wood, R. D., " Vibration of Soils and Foundations", Prentice Hall, Inc., Englewood Cliffs, N.
J.,
1970, pp. 347 & 382.
t 4.
Newmark, N. M., Hall, W. J., " Development of Criteria for Seismic Review of Selected'Nucicar Power Plants", U.S. NRC, NUREG/CR-0098, May, 1977.
5.
Roark, R. J., Young, W. C., " Formulas for Stress and Strain",
McGraw-Hill Book Co., pp. 557 & 558.
6.
Citerley, R..L.,
" Stability Criteria for Primary Me'tal Containment Vessel Under Static & Dynamic Loads", Anamet Laboratories, Inc., for General Electric NED0-21564.
7.
" Site Specific Ground Response Spectra for SEP Plants Located in the-Eastern United States", Letter of D. M. Crutchfield/,NRC to all SEP owners, dated June 8, 1981.
".SSRT Guidelines for SEP Soil-Structure Interaction Review", SSRT 8.
Tea'm,Decem5er'8i1980.
~
9.
Kaus.el.E., Whitman, R. V., Muray, J. P., and Elsabee, F., "The Spring Method for Embedded F'oundation", Nuclear Engineering and Design 48, 337-392, 1978.
1
,o<.
7 REFERENCES (cont'd)
- 10. Murphy, D.
J., Koutsoftas, D., Covey, J.
N.,
and Fischer, J. A.,
" Dynamic Properties of Hard Glacial Till", Earthquake Engineering and Soil'Pjnamics, June, 1978,-Proceedings of ASCE Geotechnical Speciality Conference.
' 11.
Johnson, G. R., Christiano, P., and Espstein, H.
I., " Stiffness
?
Coefficients.for. Embedded Footings", ASCE, J of Gcotechnical Engineering Division, August, 1975.
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