ML20135F441
| ML20135F441 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/06/1996 |
| From: | Decamp W, Gunter P AFFILIATION NOT ASSIGNED, NUCLEAR INFORMATION & RESOURCE SERVICE |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| CON-#496-18101 96-717-02-OLA, 96-717-2-OLA, OLA, NUDOCS 9612130059 | |
| Download: ML20135F441 (20) | |
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/CO/ j DOCKETED USNRC December 6,1996 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION N DEC 10 m :03 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD OF SUPIN '
00Chtd66@ '
In the Matter of )
GPU Nuclear Corporation
) Docket 50-219-OLA Oyster Creek Nuclear Generating Station ) (Tech. Spec. 5.3.1.B)
) ASLB No. 96-717-02-0LA j 4
PETITIONER'S OPPOSITION TO GPUN MOTION FOR
SUMMARY
DISPOSITION L INTRODUCTION Nuc. lear Information and Resource Service (NIRS) and Oyster Creek Nuclear Watch (OCNW), hereafter referred to as the Petitioners, submit this reply as opposition to GPU Nuclear j Corporation's (GPUN) Motion for Summary Disposition dated November 15,1996 pursuant to the Atomic Safety and Licensing Board (ASLB a the Board) October '25,1996 Memorandum and Order Ruling on Intervention Petition ofNuclear Information and Resource Service, Oyster Creek Nuclear Watch, and Citizens Awareness Network as it regards GPUN's application to j change Technical Specifications governing movement of heavy loads over stored irradiated fuel.
The Petitioners argue that NUREG-0612 " Control of Heavy Loads at Nuclear Power i Plants" as the equivalent of a regulatory guide establishes clear requirements "to the extent l practical" which govern the activity (movement of the Shield Plug) as proposed by Oyster Creek Nuclear Generating Station (OCNGS) in its license amendment application to modify Technical Specification 5.3.1.B. The regulatory guidance provided in NUREG-0612 specifies the licensee to adhere to either analyzed safe load paths and load restrictions or a single failure proof crane.
The Petitioners argue that it is within the interpretation ofNUREG-0612 terminology "to the extent practical" that OCNGS in its application to modify its current technical specification with regard to the prohibition on loads in excess of a single fuel assembly (800 lbs.) could comply with NUREG-0612 guidelines for boiling water reactors with the installation of a crane that satisfies the singie-failure-proofguidelines of Section.5.1.6. [ S_eg ATTACHMENT A]
9612130059 961206 D gDR ADOCK 05000219 PDR S 30
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!, The Petitioners therefore request that the ASLB deny the GPUN Motion for Summary l Disposition. i V i IL I.
BACKGROUND j On April 15,1996, GPUN submitted an application to the Nuclear Regulatory i
Commission (NRC)in a request to amend the Updated Final Safety Analysis Report for Oyster !
! Creek Nuclear Generating Station (OCNGS) to alter Technical Specification 5.3.1.B and add a l second subpart which has been identified by 'the licensee as:
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! 1. Loads greater than the weight of one fuel assembly shall not be moved over stored
- irradiated fuel in the spent fuel storage facility, except as noted in 5.3.1.B.2.
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{ 2. The shield plug and associated lifting hardware may be moved over irradiated fuel assemblies that are in a dry shielded canister within the transfer cask drop protection l system.
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In response to the license amendment application as published in the Federal Register (61 Fed. Reg. 20848, May 08,1996), the Petitioners, including a third petitioner Citizen Awareness Network (CAN), challenged the proposed change in Technical Specification 5.3.1.B by petition to !
request a hearing and intervention status as submitted to the US NRC Office of the Secretary June 06,1996. The Petitioners argued that the proposed technical specification change presented (1) a significant increase in accident probabilities; (2) an accident not previously identified in the licensee's Safety Analysis Report for OCNGS; (3) a significant reduction in operating boiling l 1
water reactor safety margins. Both GPUN and NRC challenged the Petitioners' request for intervention on issues of standing and lack of supporting basis and specificity of the contention.
The petitioners filed a supplementalintervention petition on July 18,1996 further l identifying their contention that GPUN through application of the license amendment failed to provide defense-in-depth against the risks of a heavy load drop onto irradiated fuel and failed to l satisfy NRC regulatory guidance as provided in NUREG-0612 " Control of Heavy Loads At ;
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Nuclear Power Plants" pertaining to defense-in-depth risk management to assure that a heavy load drop does not impact or encroach on irradiated fuel which would adversely affect public health and safety.
4 i The ASLB scheduled a prehearing conference on August 7,1996 where the parties argued the qualifications of the Petitioners standing and the legitimacy of the Petitioners' contentions.
l The ASLB Memorandum and Order served on October 25,1996 granted intervenor status l to Petitioners NIRS and OCNW, and denied formal standing to CAN but allowed CAN to submit comments as a friend of the court. The ASLB determined that there was a reasonable basis to believe that, as a consequence of a shield plug drop accident, those individuals with NIRS and ,
OCNW living in proximity to Oyster Creek could sustain injury to their health and safety which is
! protected by the Atomic Energy Act. The ASLB in its Ruling on the Intervention Petition ,
l identified several bases to the Petitioners' contention alleging that GPUN cannot revise the j technical specification as ' requested because-i (A) GPUN is legally required to establish and maintain safety limits governing activities '
potentially affecting fuel rod cladding and fuel pool liner integrity as specified under Technical ,
Specification 5.3.1.B; l
i (B) Potentially degraded fuel assemblies in the OCNGS irradiated fuel pool could be introduced into the Dry Shielded Cask as an unanalyzed threat in the event of a shield plug drop; l
(C) NUREG-0612 Control of Heavy Loads at Nuclear Power Plants" is the equivalent of a
! regulatory guide establishing the load limitations, safe load paths and/or the use of single failure proof cranes as criterion for movement of heavy loads over stored irradiami fuel. Absent the
' single-faihire-proof-crane, GPUN must rely on analyzed safe load pths and restricted load limits
{ "to assure, to the extent practical," as specified in NUREG-0612 that heavy loads are not carried i over or near stored irradiated fuel.
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. The ASLB accepted that Basis C to the Petitioners' contention satisfied the requisite
- specificity in 10 CFR 2.714(b)(2) and adequacy of the supporting bases in Section 2.174(b)(2)(ii) to establish a material factual dispute that warranted further inquiry. While the GPUN and NRC staff countered that the Petitioners interpretation of the significance and meaning of NUREG-0612 was misplaced, the Board identified a number of factors to support its finding which included-
- 1) GPUN had for some time intended to be in the position to place an object heavier than a single fuel assembly over stored fuel assemblies yet adopted the specdic fuel assembly weight limitation after NUREG-0612 was issued with its "to the extent practical" language.
- 2) GPUN and NRC staff have asserted that NUREG-0612 contains no regulatory mandate but should be viewed as merely " guidance", numerous references have been cited which identify NUREG-0612 " requirements."
The Board detennined that when these factors are combined with the Petitioners'
. challenge to the exact meaning of the NUREG-0612 "to the extent practical" terminology there is sufficient information to pose a matter oflegal interpretation.
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Pursuant to Section 2.714(b) of title 10 of the Code ofFederal Regulations, the Board
' determined that the Petitioners contention might be resolved on summary disposition without discovery and established a schedule ordering GPUN to file by November 15,1996 with an opportunity for the Petitioners, StarTand Amicus Curiae reply by December 6,1996.
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Subsequent to the Board's Memorandum and Order, the NRC staffissued the requested l
amendment on November 7,1996 pursuant to a " finding ofno significant hazards consideration i determination."
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' Ill. STATEMENT OFISSUES j The Board has established that the disputed issues oflaw within the Oyster Creeli
! proceeding are:
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- A.) What is the regulatory significance ofNUREG-0612 and its explicit terminology _"to the extent practical" as it applies to Technical Specification 5.3.1.B.-
4 p B.) May Technical Specification 5.3.1.B be changed to allow the movement of heavy loads over irradiated fuel in the Cask Drop Protection System (CDPS).
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The Petitioners maintain that sufficient documentation is provided to identify
- NUREG-0612 as the equivalent to a regulatory guide governing the control of heavy loads at -
- l. nuclear power plants. The Petitioners reassert that NUREG-0612 regulatory guidance is to be implemented as part of the " defense-in-depth" philosophy established to protect the public's health l and safety. The Petitioners maintain that al! changes and modifications made to the licensee's l, f technical specifications are required to be in compliance with NRC ' regulatory guides.
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. IV. ARGUMENT The Petitioners have reviewed GPUN's Motion For Summary Disposition and its supporting documentation which includes the Affidavit ofMr. John C.Fornicola with Exhibits and the Licensee's Statement ofMaterial Facts As To Which There Is No Genuine Dispute. The j Petitioners offer the following arguments in rebuttal to the GPUN submittal. I i
l The Licensee has argued that NUREG-0612 was issued three years after the Oyster Creek adoption of the Technical Specification 5.3.1.B. circa 1977. The original Technical ;
Specification reads " Loads greater than the weight of one fuel assembly shall not be moved over stored irradiated fuel in the spent fuel storage facility." 1
[See Licensee's Motion at page 2, footnote 2.] l l
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The Petitioners assert that the engineering design basis for the original Oyster Creek Technical Specification is the analysis of a postulated fuel handling accident.. In this analysis, a single fuel assembly is postulated to drop from its maximum height over irradiated fuel assemblies .j
- so as to estimate the radiological releases from the irradiated fuel damaged by the dropped )
assembly. This analysis was performed to satisfy regulations including 10 CFR 100 for offsite .
dose consequences.
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. . .l The Petitioners argue that the intent of the original Oyster Creek Technical Specification was to ensure that the reactor was operated within its engineering design basis by prohibiting any load heavier than a single fuel assembly being carried over irradiated fuel. This intent of technical specification's prohibition to the movement ofloads greater than one fuel assembly over irradiated fuel is indisputably exempli 6ed by the use of the explicit terminology "shall not". Compliance with '
this Technical Specification would provide reasonable assurance that the consequences of a fuel handling accident would be bound by the analysis and therefore within 10 CFR 100 limits.
The Petitioners argue that the subsequent issuance ofNUREG-0612 was never intended to lessen or alter the design basis for the fuel handhng accident, but rather to provide regulatory guidance to the licensees for activities such as replacing low density spent fuel storage racks with high density spent fuel storage racks. Since the racks and other equipment weigh more than a single fuel assembly and could potentially impact irradiated fuel if dropped, NUREG-0612 specified two permissible options [ Sgg ATTACHMENT B] for boiling water reactors:
- 1) Safe load pathways that precluded transporting heavy loads over irradiated fuels;
- 2) Single-Failure-Proof Cranes The Petitioners argue that the fuel handling accident engineering design basis would be satisfied if either criterion was met. Oyster Creek's original Technical Specification 5.3.1.B essentially met the first criterion by establishing a prohibition on movement ofloads greater than a single fuel assembly over stored irradiated fuel. The first criterion ofNUREG-0612 therefore
7 does not alter either the frequency or consequences of the analyzed fuel handling accident.
Likewise, the second criterion would also satisfy the fuel handling accident's engineering design basis in that it renders the probability of dropping the heavy load onto stored irradiated fuel incredible.
The Petitioners argue that with Oyster Creek's departure from the movement of heavy load restriction as established under the original Technical Specification 5.3.1.B the licensee is no longer bound by the fuel handling accident's engineering design basis without the benefit of the second single-failure-proof crane criterion as specified by NUREG-0612.
The Petitioners argue that the "to the extent practical" terminology in NUREG-0612 encompasses compliance with either of the specified criterion for control of heavy loads in boiling water reactors. Therefore, the Oyster Creek license amendment to modify weight restrictions on loads to be moved needs to also consider upgrading the reactor building crane and associated lifting devices used for handling heavy loads to satisfy the single-failure-proof guidelines of Section.5.1.6 ofNUREG-0612 as a matter of satisfying the "to the extent practical" qualifier.
GPUN has argued that modifications to Technical Specification 5.3.1.B can be justified to allow the movement of heavy loads over irradiated fuel in the cask drop protection system. The Petitioners concur with the licensee's conclusion. Modifications to Technical Specification 5.3.1.B can bejustified provided that either one of the two criterion in NUREG-0612 are satisfied. Given that the shield plug and lifting apparatus weight more than a single fuel assembly, the second criterion ofNUREG-0612 is applicable. GPUN could revise Technical Specification 5.3.1.B to allow the movement of heavy loads in excess of a single fuel assembly over stored irradiated fuel by using a single-failure-proof crane and satisfy the "to the extent practical" qualifier explicit in NUREG-0612. If GPUN were to pursue this course of action, there would be no basis for further :
contention of the license amendment to Technical Specification 5.3.1.B. i l
l" The Petitioners maintain that the NRC's fundamental regulatory defense-in-depth principle is implemented in NUREG-0612 as the equivalent of a regulatory guide. The Board has already j
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8 pointed out during the prehearing conference, there are a number ofreferences to NUREG-0612-
" requirements" in the licensees and agency documents. Sec Tr. at 99-101. Additionally, the reference to NUREG-0612 " requirements" is explicit in the Cenificate of Compliance For Dry Spent Fuel Storage Casks. [ See Attachment C]
While the licensee has argued in its Motion for Summary Disposition that NUREG-0612 reflects only recommendations, the licensee's submitted exhibits reflect the staffs use of the terms
" guidelines" and " requirements" interchangeably in reference to NUREG-0612 and compliance with the NRC's fundamental regulatory defense-in-depth principle.
As one example, Exhibit B in support of Licensee's Motion For Summary Disoosition.
the letter issued by NF C swTon December 22,1980 under 10 CFR 50.54(f) Enclosure 3 entitled
" Request For Additional Information On Control OfHeavy Loads" states:
_ "Regardless of the approach selected, the general guidelines of NUREG-0612, Section 5.1.1, should be satisfied to provide maximum practical defense-in-depth."
[See Licensee's Motion for Summary Disoosition Exhibit B, Enclosure 3 at Page 1]
"2.1 GENERAL REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS NUREG-0612, Section 5.1.1, identifies several general guidelines related to the design and operation of overhead load-handling systems in the areas where spent fuel is stored, in the vicinity of the reactor core, and in the areas where a load drop could result in damage to equipment required for safe shutdown or decay heat removal."
[See Licensee's Motion for Summary Discosition Exhibit B, Enclosure 3 at Page 2]
"2.2 GENERAL REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE REACTOR BUILDING NUREG-0612, Section 5.1.4, provides general guidelir.es concerning the design and operation ofload-handling systems in the vicinity of spent fuel in the reactor vessel or in storage."
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[See Licensee's Motion for Surnmmy Disposition Exhibit B, Enclosure 3 at Page 4]
The Petitioners have interpreted the NRC staffs interchangeable use of the terms ;
" requirements" and " guidelines" in reference to NUREG-0612 as intended to be more than
" recommendations" without regulatory mandare as argued by the licensee The Petitioners -
reassert their contention that the defense-in-depth principle is the overall fundamental regulatory
, mandate embodied in NUREG-M12. Therefore, the licensee in order to comply with this regulatory mandate must seek to incorporate provisions in their license amendment which maximize rather than diminish defense-in-depth at Oyster Creek.-
Additionally, the Licensee has argued that Technical Specification 5.3.1.B has always applied specifically to irradiated fuel stored in storage racks and has never applied to the Cask l
Drop Protection System. [ See Affidavit of John C. Fornicola. Page 2, Paragraph 5] The licensee argues that the CDPS packaging area is both physically distinct and separate from the irradiated fuel storage area.
l The Petitioners argue this to be the licensee's attempt at legalistic semantics. The issue at hand clearly pertains to movement of heavy loads over " stored irradiated fuel" as speci6ed in the Technical Specification as such movement would apply to all stored irradiated fuel whether it is in the storage rack or in an open NUHOMs 52 B dry shielded cask setting in the irradiated fuel pool.
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Similarly, the licensee seeks to employ legalistic semantics with the terms " stored fuel,"
" stored irradiated fuel," and " fuel storage facility" to effectively misrepresent the NRC staffs wording in its Safety Evaluation Report to characterize the Cask Drop Protection Sy stem as a j separate and distinct entity from the irradiated fuel storage area. [ See Affidavit ofJohn C.
Fomicola, Exhibit 3] As stated in the affidavit, The NRC's SER indicated that Oyster Creek will
" continue to accommodate one fuel assembly shipping cask for offsite shipping of spent fuel ;
assemblies from the Oyster Creek spent fuel pool when offsite fuel shipment is resumed at some indefinite future date " [ Emphasis added]
[ See Affidavit of John C. Fornicola. Page 7, Paragraph 17]
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l 10 l The Petitioners argue that GPUN is seeking this amendmant to permit onsite storage of-irradiated fuel, not offsite shipment. The basis of the NRC's approval of the Cask Drop Protection System was to explicitly support offsite shipment. Therefore, the Petitioners argue that GPUN's intentions are not consistent with the NRC's understanding for the use of the CDPS.
The licensee has argued that in 1984-1985 (after the issuance of Technical Specification 5.3.1.B) that 224 irradiated fuel assemblies previously shipped to West Valley were transponed back to Oyster Creek in TN-9 transportation casks which were opened in the CDPS and unloaded which included the movement of the heavy shielded cask lid over the spent fuel in the transportation cask. [See Affidavit of John C. Fornicola. Page 9 Paragraph 21)
The Petitioners argue that it is illogical for the licensee to cite this practice which was not
- in compliance with the Updated Final Safety Analysis Repon as outlined in Technical l Specification 5.3.1.B as a basis tojustify the moddication of the same technical specification.
1 Oyster Creek having violated Technical Specification 5.3.1.B in 1984-1985 is insufficient grounds j' to justify and permit future violations. If this were a legitimate argument on the pan of the j licensee, if the " understanding of the meaning of Technical Specification 5.3.1.B is consistent with
! past practice" as contended by Mr. Fornicola as only' applicable to irradiated fuel in the storage
{ racks,' then why was it necessary for the licensee to submit a license amendment to clarify this i Technical Specification? .
I Tlke Petitioners argue that this previous operation was in fact not the product of i permitted activity, but rather that this activity, which was in noncompliance with the Oyster Creek j Updated Final Safety Repon and NUREG-0612, went unchallenged by either the NRC or the affected public.
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V. CONCLUSION For all of the above stated reasons, the Petitioners argue that the Board should deny the ;
licensee's Motion for Summary Disposition.
l Respectfully Submitted, i l
Paul Gunter, Director Reactor Watchdog Project '
Nuclear Information and Resource Service 142416th Street NW Suite 404 '
Washington, DC 20036 i 202/328-0002 and Fax 202-462-2183 i Email >pgunter@igc. ape.org< '
William decamp, Jr.
l Oyster Creek Nuclear Watch l PO Box 243 Island Heights, NJ 08732 908/714-0334 and 201-376-6639 !
l Email >102115.3501@compuserve.com<
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1~9 00CKETED USHRC December 6,1996 UNITED STATES OF AMERICA 96 DEC 10 All:04 NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING @OKRIiF SECRETARY.
DOCrit i,ag & atRVICL BRAi4CH In the Matter of )
)
General Public Utility Nuclear Corporation ) Docket 50-219 OLA
) Tech. Spec. 5.3.1.B
) Movement ofHeavy Loads Over Ovster Creek Nuclear Generatinn Station ) Irradiated Fuel CERTIFICATE OF SERVICE I hereby certify that copies of "The Petitioners Opposition To GPUN Motion For Summary Disposition" has been faxed (*) and served by U.S. Mail, first class on this date December 6,1996: -
G. Paul Bollwerk, III, Chair
- Office of the Secretary (3)*
Administrative Judge U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Washington, DC 20555 U.S. Nuclear Regulatory Commission Attention: Docketing and Service Branch Washington, DC 20555 Dr. Peter S. Lam
- Dr. Charles N. Kelber*
Administrative Judge Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 Adjudicatory File (2) Ernest Blake, Jr. and David Lewis
- Atomic Safety and Licensing Board Shaw, Pittman, Potts, & Trowbridge U.S. Nuclear Regulatory Commission 2300 N. Street NW Washington, DC 20555 . Washington, DC 20037 Ann P. Hodgdon, Esq.(1)* Atomic Safety and Licensing Board Panel Office of General Counsel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 Office ofCommission Appellate Adjudication U.S. Nuclear Regulatory Commission Washington, DC 20555
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Michael Laggert GPU Nuclear Corporation j
, 1 Upper Pond Road Parsippany, NJ 07054 .
I William decamp, Jr. J Oyster Creek Nuclear Watch i P.O Box 243 i Island Height, NJ 08732 Deborah Katz <
Citizens Awareness Network j P.O Box 83 Shelburne Falls, MA 01370 s
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Paul Gunter Nuclear Information and Resource Service
l ATTACTm1ENT A i
1 NUREG-0612 l
Control of Heavy Loads at Nuclear Power Plants 1
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, Resolution of Generic Technical Activity A-36 I
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_.__ ___ ..______ _ _. . l Manuscript Completed: January 1980 Date Punished: July 1980 i H. George, Task Manager Division of Operating Reactors Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20665 p" "%..,
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Treatment System if facility technical specifications require its operation I j during periods when the load being analyzed would be handled. The analysis should also conform to the guidelines of Appendix A.
4 5.1.5 Other Areas In other plant areas, loads may be handled which, if dropped in a certain location, may damage safe shutdown equipment. Although this is not a concern at all plants, loads that may damage safe shutdown equipment at some plants l i
include the spent fuel shipping cask, turbine generator parts in the turbine I j building, and plant equipment such as pumps, motors, valves, heat exchangers, j and switchgear. Some of these loads may be less than the weight of a fuel 4
assembly wit.h its handling tool, but say be sufficient to damage safe shutdown equipment.
j (1) If safe shutdown equipment are beneath or directly adjacent to a potential
, travel load path of overhead handling systems, (i.e. , a path not restricted j
by limits of crane travel or by mechanical stops or electrical interlocks)
- one of the following should be satisfied in addition to satisfying the
- general guidelines of Section 5.1.1
(a) The crane and associated lifting devices should conform to the single-failure-proof guidelines of Section 5.1.6 of this report; I OR (b) If the load drop could fi6 air the operation of equipment or cabling associated with redundant or dual safe shutdown paths, mechanical stops or electrical interlocks should be provided to prevent movement of loads in proximity to these redundant or dual safe shutdown equipment (In this case credit should not be taken for intervening floors unless justified by analysis).
OR (c) The effects of load drop Thave been analyzed and the results indicate that damage to safe shutdown equipment would not preclude operation of sufficient equipment to achieve safe shutdown. Analyses should conform to the guidelines of Appendix A, as applicable.
4 (2) Where the safe shutdown equipment has'a ceiling separating it from an overhead handling system, an alternative to Section 5.1.5(1) above would be to show by analysis that the largest postulated load handled by the handling system would not penetrate the ceiling or cause spalling that could cause failure of the safe shutdown equipment.
l 5.1.6 Single-Failure-Proof Handling Systems For certain areas, to meet the guidelines of Sections 5.1.2, 5.1.3, 5.1.4, or 5.1.5, the alternative of upgrading the crane and lifting devices may be .
chosen. The purpose of the upgrading is to improve the reliability of the l handling system through increased factors of safety and through redundancy or duality in certain active comon%r.t!. NUREG-0554, " Single-failure-Proof Cranes for Nuclear Power Plants," provices guidance for design, fabrication, installation, and testing of new cranes that are of a high reliability design.
For operating plants, Appendix C to this reDort, " Modification of Existing ,
Cranes," provides guidelines on implementation of NUREG-0554 for operat.ing plants and plants under construction.
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Section 5.1.1 of this report provides certain guidance on slings and special handling devices. Where the alternative is chosen of upgrading the handling
! system to be " single-failure proof", then steps beyond the general guidelines of Section 5.1.1 should be taken. - -
Therefore, the following additional guidelines shoult' be met where the alterna-i tive of upgrading handling system reliability is chesen:
l (1) Lifting bswices:
(a) Special lifting devices that are used fo'* heavy loads in the area where the crane is ti be upgraded should meet ANSI N14.6 1978,
" Standard For Special I.ifting Devicer for Shipping Containers Weighing l 10,000 Pounds (4500 kg) or More For Nuclear Materials," as specified
, in Section 5.1.1(4) of t.51s report except that the handling device ,
) should also comply with Section 6 of ANSI N14.6-1978. If only a '!
single lifting' device is provided instead of dual devices, the !
special lifting device should have twice the design safety factor as a required to satisfy the guidelines of Section 5.1.1(4). However,
{' loads that have been evaluat~i and shown to satisfy the evaluation .
- criteria of Section 5.1 need .iut have lifting devices that also l
comply with Section 6 of ANSI 6.4.6. 3 l (b) 1.ifting devices that are not specially designed and that are used
, for handling heavy loads in the area where the crane is to be upgraded j should meet ANSI B30.9 - 1971, " Slings" as'specified in Section 5.1.1(5) i of this report, except that one of the following should also be j satisfied unless the effects of a drop of the particular load have j been analyzed and shown to satisfy the evaluation criteria of
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Section 5.1: ,
l (i) Provide dual or redundant slings or lifting devices such that a j single component failure or malfunction in the sling will not result in uncontrolled lowering of the load; j 0,,R_
(11) In selecting the proper sling, the load used should be twice what is called for in meeting Section 5.1.1(5) of this report.
I (2) New cranes should be designed to meet NUREG-0554, " Single-Failure-Proof j Cranes for Nuclear Power Plants." For operating plants or plants under f construction, the crane should be upgraded in accordance with the imple-l mentation guidelines of Appendix C of this report.
- (3) Interfacing lift points such as lifting lugs or cask trunions should also
- meet one of the following for heavy loads handled in the area where the 4 crane is to be upgraded unless the effects of a drop of the particular l load have been evaluated and shown to satisfy the evaluation criteria of Section 5.1
l (a) Provide redundancy or duality such.that a single lift point failure
- will not result in uncontrolled lowering of the load; lift points i' should have a des Qn safety factor with respect to ultimate strength of five (5) times the maximum combined concurrent static and dynamic load after taking the single lift. point failure.
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ATTACRMENT B l
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t NUREG-0612 9
Control of Heavy Loads at-Nuclear Power Plants I
Resolution of Generic Technical Activity A-36 l
Manuscript Completed: January 1980 Date Published: July 1980 H. George. Teek Meneger Division of Operating Reactors Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 p." ~s., ,
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(2) Rapid containment isolation is provided'with prompt automatic actuation on high radiation so that postulated releases are within limits of evaluation Criterion I of Section 5.1 taking into account delay times in detection and actuation; and analyses have been performed to show that evaluation criteria II, III, and IV of Section 5.1 are satisfied for postulated load
, drops in this area. These analyses should conform to the guidelines of l Appendix A.
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(3) The effects of drops of heavy Icads should be nelyzed and shown to satisfy the evaluatica criteria of Section 5.1. Loads analyzed should-include the following: reactor vessel head; upper vessel internals; vessel inspection platform; cask for damaged fuel; irradiated sample cask; reactor coolat pump; crane load block; and any other heavy loads brought over or near the reactor vessel or other equipment required for continued decay heat removal and maintaining shutdown. In this analysis, credit may be taken for containment isolation if such is provided; however analyses should establish adequate detection and isolation time. Addi-tionally, the ana~ysis should conform to the guidelines of Appendix A.
5.1.4 Reactor Building - 8WR The reactor building in 8WRs typically contains the reactor vessel and spent fuel pool, as well as varicus safety-related equips *ent.
The reactor building ove-head crane may be used in many day-to-day operations such as moving various shielded shipping casks or handling plant equipment related to maintenance or modification activities. The crane is also used during refueling operations for removal and reinstallation of shield plugs, drywell head, reactor vessel head, steam dryers and separators, and refueling canal plugs and gates. The crane would also be used subsequent to refueling for handling of the spent fuel shipping cask. This cask may be lifted as high as 100 feet (30 m) above the grade elevation at which the cask is brought into the reactor building. Additionally the overhead crane's load block may be moved over fuel in the reactor or over the spent fuel pool when handling smaller loads or no load at all. Due to the weight of the load block alone, this should also be considered as a heavy load.
To assure that the evaluation criteria of Section 5.1 are satisfied one of the l' following should be met in addition to satisfying the general guidelines of Section 5.1.1:
(1) The react 3r building crane, and associated lif ting devices used for i handling the above heavy loads, should satisfy the single-failure-proof l guidelines of Section 5.1.6 of this report.
OR (2) The effects of heavy load drois in the reactor building should be analyzed to show that the evaluattan criteria of Section 5.1 are satisfied. The loads analyzed should include: shield plugs, drywe'll head, reactor .
vessel head; steam dryers and separators; refueling canal plugs and gates; shielded spent fuel shipping casks; vessel inspection platfors:
and any ether heavy loads that may be brought over or near safe shutdown ;
eoulpment as well as fuel in the reactor vessel or the spent fuel pool. ;
i Credit may be taken in this analysis for opernt.f on of the Standby Gas F-S
l ATTACHMENT C
-i i
e
- Certificate of Compliance
! FOR DRY SPENT FUEL STORAGE l
CASKS -
l 10 CFR PART 72
- 1. a. CERTIFICATE NUMBER: 1004
- b. REVISION NUMBER: 0 t"'*4 O 8"""
- c. PACKAGE IDENTIFIJATIONapp: 41SA/h1004 e] Q?g
/
- d. PAGE NUMBER: 1 'yf5 P PAGES: 4 jf
- e. TOTAL NUM
- 2. PR is certificate is issued to certify that the cask ntents, des fed in item 5 below, meet the applicable safety standardsssa'tjorth in Title %LO, code of Federal Regulations, Part 72," Licensing Requirements for '
ndepishdent Storage of Spent Nuclear Fuel and High-Levet";Ridioadfive i.*
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~ ~
e nt f asaf f sis report okthe r
- 3. f*TNISCERTIFICAT t dardi ,, S-528 sk design, Mo UHOMS-ZfP
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) EAN$I ATION
- a. P RARED BY (Name '"""'"
y . vr i ICATION Inc. - s, I .k.a *N VECT W 7echnol _
Pact.f Servif ,l ; Pic{,ff "f(schpue , Inc. %
6203,jglNuclea the ("'""
San Ignac e ' Su ,AjT '
Sfa
$hp!y(ntrysisRe]l retzed NUHOMS ontalModuQg San Jose CA 9 ,
f"i Sto T for ! atedNucigga Fuel" e
Y M g '/d -
1 8
_- BER)72-1004
- /")
i V u / #K4 V
- 4. CONDITIpRS This certificate upo fbj rements f rt 72, as applicab condition $illingthehg s specifi eTow.
of 10 4F s v Effective Date: w/"A-
- FOR THE NUCLEAR REGULATORY COM ISSION 4F January 23, 1995 Expiration Date.
January 31, 2015 Charles J. Hadgfiney, Chie() -
Storage and Transport Systems Branch Division of Industrial and Medical Nuclear Safety, NMSS
4 I ,
j !'
4-
,i
! l
- b. Description . ,t
- .I ne Standardized NUHOMS System and its analyses and operations are described in the SAR (Docket 72-1004) identified previously. The Nuclear Regulatory Commission has reviewed the l' '
SAR in the Safety Evaluation Report identified previously.
3 he system which is being cernfied is described in Sections 1, 3, 4, 5, 6, 7 and 8 of the SAR . ; ,
[ and in the NRC's SER accompanying the SAR. (The system drawings, which reflect this L' description, are contained in Appendix E of the SAR.) The Standardized NUHOMS System is a .
horizontal canister system composed of a steel dry shielded canister (DSC), a reinforced concrete l horizontal storage module (HSM), and a transfer cask (TC). The welded DSC provides !
- confinement and criticality control for the storage and transfer of irradiated fuel. The concrete i!
l module provides radiation shielding while allowing cooling of the DSC and fuel by natural ' ' {
convection during storage. The TC is used for transferring the DSC from/to the Spent Fuel Pool Building to/from the HSM.
- 4 I!
4 The prir.cipal component subassemblies of the DSC are the shell with integral bottom cover plate !
and shield plug and ram / grapple ring, top shield plug, top cover plate, and basket assembly.
i ne shell length is fuel-specific. The internal basket assembly is composed of guide sleeves, j suppon rods, and spacer disks. This assembly is designed to hold 24 PWR fuel assemblies or ' !
52 BWR assemblies. It aids in the insertion of the fuel assemblies, enhances subcriticality i
' during loading operations, and provides structural suppon during a hypothetical drop accident.
l The DSC is designed to slide from the transfer cask into the HSM and back without undue ,
- galling, scratching, gouging, or other damage to the sliding surfaces. I-i l,!
' I The HSM is a reinforced concrete unit with penetrations located at the top and bottom of the side walls for air flow. The penetrations are protected from debris intrusions by wire mesh I i~ screens during storage operation. The DSC Support Structure, a structural steel frame with l rails, is installed within the HSM module to provide for sliding the DSC in and out of the HSM ,
and to suppon the DSC within the HSM. ! <
l The TC is designed and fabricated as a lifting device to meet NUREG-0612 and ANSI N14.6 :
requirements. It is used for transfer operations within the Spent Fuel Pool Building and for - I transfer operations to/from the HSM. The TC is a cylindrical vessel with a bottom end closure assembly and a bolted top cover plate. Two upper lifting trunnions are located near the top of the cask for downending/ uprighting and lifting of the cask in the Spent Fuel Pool Building. The e lower trunnions, located near the base of the cask, serve as the axis of rotation during ,
downending/ uprighting operations and as supports during transport to/from the Inclependent ;
- Spent Fuel Storage Installation (ISFSI). i
- l. With the exception of the TC, fuel transfer and auxiliary equipment necessary for ISFSI i operations are not included as a pan of the Standardized NUHOMS System to be reviewed for a ;
Cenificate of Compliance under 10 CFR Pt.rt 72, Subpan L. Such equipment may include, but !
is not limited to, special lifting devices, the transfer trailer, and the skid positioning system.
J l i 2
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