ML20134L567

Licensee Motion for Summary Disposition.* Petitioner Contention Should Be Dismissed

ML20134L567
Person / Time
Site:	Oyster Creek
Issue date:	11/15/1996
From:	Lewis E GENERAL PUBLIC UTILITIES CORP., SHAW, PITTMAN, POTTS & TROWBRIDGE
To:	Atomic Safety and Licensing Board Panel
Shared Package
ML20134L571	List: ML20134L567 ML20134L579 ML20134L585
References
CON-#496-18065 96-717-02-OLA, 96-717-2-OLA, OLA, NUDOCS 9611210084
Download: ML20134L567 (127)

Text

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UNITED STATES OF AMERICA 96 NOV 19 P3 :30 NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOAl }FICE Of SFCPEJARY JuCKETINS a SERvlCE BRANCH in the Matter of

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Docket No. 50-219-OLA GPU NUCLEAR CORPORATION

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(Tech. Spec. 5.3.1.B)

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(Oyster Creek Nuclear Generating Station)

)

ASLBP No. 96-717-02-OLA LICENSEE'S MOTION FOR'

SUMMARY

DISPOSITION 1.

(ODUCTION GPU Nuclear Corporation ("GPUN" or " Licensee") submits this motion for summary dis-position pursuant to the Atomic Safety And Licensing Board's (" Board") October 25,1996 4

Memorandum and Order Ruling on Intervention Petition of Nuclear Information and Resource Service, Oyster Creek Nuclear Watch, and Citizens Awareness Network (" Petitioners").E Con-sistent with that Memorandum and Order, the motion addresses the sole legal issue remaining in this proceeding. A Statement of Material Facts as to Which There is no Genuine Dispute, an Af-fidavit of John C. Fornicola ("Fornicola Aff."), and exhibits are attached in support.

The gravamen of Petitioners' position is that Technical Specification 5.3.1.B of the Oyster Creek license is a vital control prohibiting the movement of heavy loads in the Cask Drop l

. ASLB Memorandum and Order, Ruling on Intervention Petition, LBP-96-23, October 25,1996 (hereinafter E

LBP-96-23).

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Protection System ("CDPS") and, as such, cannot be changed as a matter oflaw. For the reasons l~

' discussed in this motion, this position is legally untenable because Technical Specification 1

5.3.1.B only applies to heavy loads moved over stored fuel in the spent fuel storage racks and is i

t i

no legal impediment to the movement of heavy loads over spent fuel in the CDPS.

l

.11.

BACKGROUND l

i The underlying action of the proceeding is the Licensee's request to moaify a technical l

- specification for the Oyster Creek Nuclear Generating Station (" Oyster Creek"). On April 15, I

1996, GPUN requested that Technical Specification 5.3.1.B be modified to add a second sub-l i

part to read as follows:

1. Loads greater tiian the weight of one fuel assembly shall not be l

moved over stored irradiated fuel in the spent fuel storage facility, except as noted in 5.3.1.B.2.

2. The shield plug and associated lifting hardware may be moved over irradiated fuel assemblies that are in a dry shielded canister l

within the transfer cask in the cask drop protection system.

Fornicola Aff.13 and Exh. 2 (GPUN Corporation, Oyster Creek Nuclear Generating Station,

- Technical Specification Change Request No. 244 (Apr.15,1996)). Pursuant to a final no signifi-cant hazards consideration determination, the NRC issued this amendment on November 7,1996.

U.S. Nuclear Regulatory Commission, Issuance of Amendment Re: Handling Heavy Loads Over Irradiated Fuel (Nov. 7,1996), attached as Exhibit A hereto. Petitioners challenge this change.

A Prior to the amendment, Technical Specification 53.1.B read " Loads greater than the weight of one fuel as-sembly shall not be moved over stored irradiated fuel in the spent fuel storage facility.",

I l

1 The legal issue in the proceeding established by the Board is from Petitioner's contention.

j The Board has summarized Basis C as follows:

The NRC's fundamental regulatory defense-in-depth principle is implemented in NUREG-0612 " Control of Heavy Loads at Nuclear Power Plants," which is the equivalent of a regulatory guide. Be-cause OCNGS does not employ a single failure proof crane for shield plug movement, consistent with NUREG-0612 guidelines as described in enclosure 1 to NRC Generic Letter 85-11 (June 28, 1985), GPUN must rely on analyzed safe load paths and restricted load limits for movement of heavy loads to " assure, to the extent practical" that heavy loads are not carried over or near irradiated fuel. Although GPUN claims in its safety evaluation regarding the proposed technical specification change that a shield plug drop ac-cident is not credible because of GPUN administrative controls (e.g., rail stops), operator training, and inspections concerning dry-storage related spent fuel movements, this does not adequately ad-dress human error or mechanical / electrical failure issues. Rather, the most effective way to avoid such failures is to restrict both human-directed activity and prohibit the movement of heavy loads as is done with current [ prior to the November 7,1996 amend-ment] Technical Specification 5.3.1.B. As such, consistent with j

the agency's NUREG-0612 defense-in-depth guidance, the [ pre-]

existing provision cannot be revised as the licensee has requested.

1 LBP-96-23 at 12.

Regarding Basis C, the Board states that " petitioners seek to establish the ' single fuel as-sembly' weight limitation in [ pre-] existing Technical Specification 5.3.1.B reflects an agency judgment about the particular measures that are necessary for compliance with the purported regulatory guidance in NUREG-0612 as it is asserted to implement the ' defense-in-depth' princi-j ple." 11 at 40. Petitioners assert that "this weight limitation [in Technical Specification 5.3.1.B]

l is a vital control meant to remove the potential that human error or any mechanical / electrical failure could cause damage to irradiated fuel." E at 40 (citing Petitioners' statements in the Pre-Hearing Conference, Tr. at 68). In summary, Petitioners assert that "[b]ecause of the importance of this limitation.., this technical snecification cannot be changed." id. at 40-41 (emphasis added).

i The Board identifies two factors that provide " sufficient reason to conclude Basis C es-l tablish[es] a material disputed issue oflaw that should be considered further." E at 41. The first thetor is that, based on the fact that Licensee's CDPS has been in place for some time 3-the Licensee and the NRC staff"had some notion that GPUN at some point could be in a position to place an object heavier that a fuel assembly over fuel assemblies being packaged for removal and storage," but "[n]onetheless, the [ pre-] existing technical specification with its specific ' fuel as-sembly' weight limitation seemingly was adopted for OCNGS after NUREG-0612 was issued with its 'to the extent practicable' language."1 1 at 41, fjling U.S. Nuclear Regulatory Com-mission, NUREG-0612, Control of Heavy Loads at Nuclear Power Plants (July 1980)(hereinaf-ter NUREG-0612), at 3-9 (Table 3.2-1),5-2. The implication is that both the Licensee and the NRC staff realized that the original Technical Specification 5.3.1.B war adop>.ed to prohibit the Licensee from ever packaging spent fuel into casks for removal from Oyster Creek.

L Fornicola Aff.,5 8. The Cask Drop Protection System ("CDPS") was approved by the Atomic Energy Com-mission in 1973 and subsequently installed at Oyster Creek.

i Although the assumption is easy to make, Technical Specification 5.3.1.B w as actually adopted in 1977, more than three y ears prior to NUREG-0612, in response to an NRC staff request related to changing out the spent fuel storage racks at Oyster Creek, as is discussed inft.a. S.c.g Fornicola Aff.,16.

_. - - -. ~.

The second factor identified by the Board as providing sufficient reason to establish a ma-i j

terial disputed issue oflaw is that the licensee and NRC staff"have asserted that NUREG-0612 is simply ' guidance' that contains no regulatory mandate," while at the same time "there are any number of references to NUREG-0612 ' requirements' in the Licensee and agency documents pro-vided to (the Board]." LBP-96-23 at 42 (citing the Pre-Hearing Conference, Tr. at 99-101, as i

well as the Certificate of Compliance for the NUHOMS system). This, the Board states, raises a

" legitimate question about the regulatory significance of[NUREG-0612] and its 'to the extent

.I l

practical' language." Id. at 42. Although it does not raise any question about the adequacy'of -

i

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GPUN's load handling training or procedures (11 at 41 n.19), Petitioners' Basis C, as summarized i

e by the Board, taken together with the two above factors, poses a " matter oflegal interpretation l

that merits further scrutiny." Id.

)

The legal issue in Basis C, as summarized by the Board, is the only remaining issue in l

this proceeding as all of the Petitioners' other contentions and bases were dismissed as inade-1 quate to establish a material dispute warranting further inquiry.11 at 36,39-40.

111.

STATEMENT OF ISSUE 3

The disputed issue oflaw established by the Board within this proceeding can be summa-rized as follows:

A. What is the Regulatory Significance of NUREG-0612 and is Technical Specification 5.3.1.B Required Pursuant to NUREG-0612.

B. May Technical Specification 5.3.1.B be Changed to Allow the Movement of Heavy Loads over Spent Fuel in the Cask Drop Pro-tection System.

Regarding Part A, GPUN maintains that the true regulatory significance of NUREG-0612 is only through the generic letters issued by the NRC requesting licensee implementation of se-lected parts of NUREG-0612. Further, the adoption of a technical specification comparable to 5.3.1.B was not one of the selected parts of NUREG-0612 which licensees were requested to im-plement by these generic letters.

Regarding Part B, GPUN maintains that the change to Technical Specification 5.3.1.B l

was permissible for several reasons. First, the technical specification, which was originally is-J sued before NUREG-0612, has always applied only to stored spent fuel in the fuel storage area and not to spent fuel in the CDPS being packaged for movement out of the reactor building. The recent amendment is merely a clarification of this intended scope and meaning. Further, even if 1

the technical specification were interpreted in light of the subsequent NUREG-0612 recommen-dations (including the interim recommendation that was not adopted in the generic letters), its scope and meaning would be unaltered. The interim technical specification recommended by NUREG-0612 also applies only to spent fuel stored in racks.

Accordingly, there is no legal restriction to the amendment to Technical Specification 5.3.1.B at issue in this proceeding. The Board should therefore decide the legal issue consistent with the Licensee's position and terminate this proceeding.

i IV.

ARGUMENT l

A.

NU LEG-0612 REFLECTS RECOMMENDATIONS OF AN NRC STAFF TASK GROUP BUT IS NOT A REGULATION. WHILE CERTAIN OF THOSE REC-OMMENDATIONS WERE ADOPTED BY GENERIC LETTERS,THE RECO>l-l MENDATION FOR A TECHNICAL SPECIFICATION WAS NOT.

1.

NUREG-0612 Reflects Only Recommendations, and the Recommen-dations which are of Regulatory Significance are those Adopted in NRC Generic Letters Issued to Licensees Pursuant to section 103 the Atomic Energy Act ("AEA"), the NRC is authorized to grant and modify commercial licenses for nuclear power plants " subject to such conditions as the Commission may by rule or regulation establish to effectuate the purposes of this Act." AEA Q 103,42 U.S.C. { 2133 (1994). The stated purposes oflicensing under the AEA are to maintain activities "in accord witti the common defense and security and (to) provide adequate protection to the health and safety of the public." AEA 182,42 U.S.C. { 2232 (1994).

Pursuant to the AEA, the NRC has established requirements specific to the movement of heavy loads at nuclear power plants in General Design Criteria established in 10 C.F.R. s 50.34 and 10 C.F.R. Part 50, Appendix A. Compliance with the General Design Criteria " provide [s]

reasonable assurance that the facility can be operated without undue risk to the health and safety of the public." 10 C.F.R. Part 50, App. A (introduction). General Design Criterion 61," Fuel Storage and Handling and Radioactivity Control," establishes the regulatory requirements for heavy load movements as follows:

The fuel storage and handling, radioactive waste, and other sys-tems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions.

l These systems shall be designed (1) with a capability to permit l

1 l

appropriate periodic inspection and testing of components impor-tant to safety, (2) with suitable shielding for radiation protection.

(3) with appropriate containment, confmement, and filtering sys-tems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat l

and other residual heat removal, and (5) to prevent significant re-duction in fuel storage coolant inventory under accident conditions.

E at Criterion 61.

The NRC issues NUREGs and Regulatory Guides as advisory guidance providing ap-proaches licensees can follow to meet legal requirements. Curators of the University of Mis-10.Un, CLI-95-8,41 N.R.C. 386,397 (1995). "[I]t,is well established.. that NUREGs and Regulatory Guides, by their very nature, serve merely as guidance and cannot prescribe require-ments." Curators of the University of Missouri, CL1-95-1,41 N.R.C. 71,98 (1995). Although conformance with the guidance will likely result in compliance with underlying regulations, non-conformance with the guidance does not equate to noncompliance with the regulations. R NUREG-0612 was initiated to " systematically examine stafflicensing criteria'and the adequacy of measures in efTect at operating plants, and to recommend necessary changes to as-sure the safe handling of heavy loads...." NUREG-0612 at 1-1. NUREG-0612 develops " rec-ommendations on actions that should be taken" based on an NRC task group's evaluation of NRC

" licensing criteria and the adequacy of measures in effect at operating plants." E At no time does the document maintain or assert that NUREG-0612 itself establishes requirements that li-censees must comply with. NUREG-0612 characterizes itself as "a summary of those recont-mended actions that should be taken to resolve the concern over the handling of heavy loads near irradiated fuel, or safety related equipment." E at 6-1 (emphasis added).

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The NRC adopted certain recommendations from NUREG-0612 through generic letters l

L to all licensees. The first generic letter, dated December 22.1980, requested all licensees to im-f-

ls

' plement selected interim actions, listed in Enclosure 2 of the letter, provide information showing l

i how their facility satisfied the guidelines of NUREG-0612, and demonstrate implementation of l

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the criteria of NUREG-0612 @ 5.1.1 over six months and NUREG-0612 s 5.1.2-5.1.5 over 9 months. U.S. Nuclear Regulatory Commission, Unnumbered Generic Letter on Control of Heavy l

Loads (Dec. 22,1980), attached as Exhibit B hereto, at 2.2 The second generic letter,85-11.

stated that only the criteria of NUREG-0612 { 5.1,.1 must be implemented (referred to as " Phase 1"), and rescinded the prior request to implement NUREG-0612 5.1.2-5.1.5 (referred to as t

" Phase 11"). U.S. Nuclear Regulatory Commission, Generic Letter 85-11, " Completion of Phase 11 of ' Control of Heavy Loads at Nuclear Power Plants' NUREG-0612" (June 28,1985), attached as Exhibit D hereto. Thus, it was only through these generic letters that licensees were requested l

to comply with any of the guidelines of NUREG-0612, and the generic letters only request im-l piementation of selected recommendations from NUREG-0612, not all of the " Recommended j

Guidelines" in the report.

2.

The Generic Letters Do Not Require the Implementation of a Techni-cal Specification Comparable to 5.3.1.B One of the recommendations made in NUREG-0612 was an " Interim Protection" meas-ure that:

2-This letter does not have a Generic Letter identifier, but subsequent NRC documentation on the subject refers to it as "the December 22,1980 generic letter on ' Control of Heavy Loads'." Set U.S. Nuclear Regulatory Commis-sion, Control of Heavy Loads (Phase 1)- NUREG-0612 - Oyster Creek Nuclear Generating Station (June 21.

1983), attached as Exhibit C hereto, at 1.

. f 1

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(1) Licenses for all operating reactors not having a single-failure-proof overhead crane in the fuel storage pool area should be re-vised to include a specification comparable to Standard Technical Specification 3.9.7, " Crane Travel - Spent Fuel Storage Pool Building" for PWR's and Standard Technical Specification 3.9.6.2,

" Crane Travel," for BWR's, to prohibit handling of heavy loads over fuel in the storage pool until implementation of measures which satisfy the guidelines of Section 5.1 (see Table 3.2-1).

NUREG-0612 at 5-18. This is the recommendation that is comparable to Technical Specification 5.3.1.B which was already in place at Oyster Creek. Five other " Interim Protection" recommen-dations were also developed in NUREG-0612. E at 5-18 to 5-19.

The generic letters only requested licensee implementation of the last five " Interim Pro-tection" recommendations, not " Interim Protection" recommendation (1) -- the Technical Specifi-cation guideline. The NRC requested licensees to " implement the interim actions described in i [to the De.:. 22,1980 Generic Letter]..." Exh. B at 2. The " interim actions" in to this generic letter are the five " Interim Protection" measures numbered (2) through (6). E at Encl. 2. Comnare NUREG-0612 at 5-18. Interim Protection recommenda-tion (1) of NUREG-0612, calling for a Technical Specification comparable to Oyster Creek Technical Specification 5.3.1.B., is not included in the December 22,1980 generic letter and is not made a requirement in any other NRC document.

e B.

TECIINICAL SPECIFICATION 5.3.1.8 MAY BE CilANGED TO ALLOW Tile MOVEMENT OF llEAVY LOADS OVER SPENT FUEL IN TIIE CASK DROP PROTECTION SYSTEM WHERE TIIE MEANING AND IIISTORY OF TIIE TECIINICAL SPECIFICATION SIIOW IT IIAS ALWAYS APPLIED ONLY TO STORED SPENT FUEL IN TIIE FUEL STORAGE AREA As discussed below, the recent change to Technical Specification 5.3.1.B explicitly al-lowing the movement of heavy loads over spent fuel in the CDPS merely clarifies the original in-tent and meaning of the technical specification. Technical Specification 5.3.1.B has always applied, by both wording and intent, to " stored" spent fuel. GPUN applied for the recent techni-cal specification change, at the suggestion of the NRC staff and out of an abundance of caution.

only to make this meaning more explicit. Fornicola Aff., $ 23. Thus, Technical Specification 5.3.1.B is not, and has never been a vital control intended to prevent movement of heavy loads i

over the CDPS, and the change is in fact nothing more than a non-substantive clarification.

In the same vein, the original Technical Specification 5.3.1.B was not issued after or in response to NUREG-0612. Rather, it was issued three years earlier (la, in 1977) to address the potential for load drops over fuel storage racks that were being modified. In any event, even if this technical specification were interpreted in a manner consistent with the recommendations of NUREG-0612, its intended scope would not change, because the technical specification recom-mended by NUREG-0612 applies only to fuel in storage racks. Thus, none of the recommenda-tions of NUREG-0612 would prevent the wording change proposed by GPUN.

~..

1.

The Plain Meaning of Technical Specification 5.3.1.B Indicates it Ap-plies Only to Stored Irradiated Fuelin the Fuel Storage Facility Under its literal terms? Technical Specification 5.3.1.B is and has always been a prohi-bition against moving certain loads over fuel that is " stored" in a " storage" facility. As originally adopted and prior to the November 7,1996 amendment, Technical Specification 5.3.1.B stated:

i 5.3.1.B. Loads greater than the weight of one fuel assembly shall l

not be moved over stored irradiated fuel in the spent fuel sloInge facility. (emphasis added).

In ordinary usage, " stored" means placed gr left in a location, (as a warehouse) for preser-vation or later use? A storage facility would thus be a location where things are placed and left for later use. This meaning is fully consistent with spent fuel that is placed in the storage racks of that part of the spent fuel pool set aside to leave spent fuel in for future use. On the other hand, this meaning is inconsistent with spent fuel transferred to the Cask Drop Protection System for packaging into a cask for immediate movement out of the building. If" stored" means any-thing, it means something that is not in the midst of a packaging operation where it is being pre-pared for transport out of the facility.

The requirement uses the specific terms " stored irradiated fuel" and " spent fuel storage I

facility" rather than the broader terms " fuel" and " spent fuel pool." The term " irradiated" is used.

to differentiate it from new or " fresh" fuel which is stored in a different location. The use of the Textual provisions of requirements should be interpreted using the plain meaning of the terms. Sss Estate of l

Cow art v. Nicklos Drilline Co. 505 U.S. 469, i12 S. Ct. 2589. 2594 (1992)(applying the canon to interpretation of a statute),

I Webster's Ninth New Colleeiate Dictionary 1162 (1987).

l j :

term " stored" irradiated fuel instead of simply " irradiated fuel" indicates a specific intent to limit the requirement to " stored" fuel, and not to any other fuel, including that being " packaged" for transport.L The use of the term " storage" facility indicates that the " packaging" area -- the CDPS

-- is not included. This careful wording shows that the textual meaning of Technical Specifica-tion 5.3.1.B has always been limited in scope to fuel stored in the storage racks area of the spent fuel pool, and that the scope has never included fuel being packaged in the CDPS.

This plain meaning is consistent with prior NRC Staffinterpretation. Prior to the adop-tion of the original Technical Specification 5.3.1.B 224 spent fuel assemblies were packaged in transportation casks and shipped to the Nuclear Fuel Services' reprocessing plant in West Valley, New York. Fomicola Aff., $$ 6 and Exh. 3 (U.S. Nuclear Regulatory Commission, Issuance of Amendment No. 22 to Provisional Operating License No. DPR-16 for the Oyster Creek Nuclear Generating Station (Mar. 30,1977)) at page 3 of enviromnental impact appraisal. In loading the transportation cask, a heavy shielded lid was lowered over the spent fuel assemblies in the cask within the CDPS. Fornicola Aff., $ 20. In 1984 and 1985, some seven years after the adoption of Technical Specification 5.3.1.B and four years after the publication of NUREG-0612, the same 224 spent fuel assemblies were returned to Oyster Creek from West Valley.2. Unloading the spent fuel from the transportation cask in the CDPS involved lifting the shielded transporta-tion cask lid, which is heavier than one fuel assembly, over spent fuel assemblies in the L

The expression of one thing in a requirement indicates the exclusion of others: expressia unius est exclusio al-terius. Ses O'Melveny & Ms ers v. FDIC. 512 U.S. $79, i 14 S. Ct. 2048,2054 (1994)(applying this canon to inter-pretation of a statute).

1 Fornicola Aff.,5 21.

transportation cask.8 This history reflects the NRC staffs and Licensee's understanding that Technical Specification 5.3.1.B did not apply to spent fuel packaging operations in the CDPS.

-This plain meaning is also consistent with prior NRC Staffinterpretation for other facili-ties. The sam ~e situation occurred in 1992 at the Consumers Power Company (" Consumers")

Palisades Nuclear Plant (" Palisades")in preparation to move spent fuel out of the facility and into dry storage. Palisades Technical Specification 3.21.2 prohibits the movement of heavy loads "over fuel stored in the main pool zone." Fornicola Aff.122 and Exh. 9 (Consumers

^

Power Company, Response to NRC Bulletin 96-02: Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety Related Equipment (May 16,1996)) at page 2 of attachment (emphasis added). This technical specification is comparable to Oyster Creek Technical Specification 5.3.1.B (as it read prior to the November 7,1996 amendment) because it specifically addresses heavy load movements over stored fuel. Palisades' dry storage activities required shield covers and other heavy loads to be moved over spent fuel in the transfer cask in the packaging area of the pool. E at pages 2-3 of attachment. Based on several conversations between Consumers and the NRC staff, it was determined that spent fuel in the transfer cask was not subject to the requirements of Technical Specifica. n 3.21.2 because the technical specifica-tion applied only to stored fuel and fuel in the transfer cask is not being stored but is rather in transit. E at page 3 of attachment. The NRC staffs conclusion on the non-applicability of Pali-sades Technical Specification 3.21.2 to spent fuel in a transfer cask in the cask packaging area is 2

Fornicola Aff.,121.

i consistent with the non-applicability of Oyster Creek Technical Specification 5.3.1.B to spent fuel in the transport cask in the CDPS.

This understanding is also reflected in the NRC's Safety Evaluation supporting the No-vember 7.1996 amendment to Technical Specification 5.3.1.B. The NRC staff stated:

[t]he current [ technical specification] is ambiguous regarding this movement because the DSC, at that point, contains irradiated fuel, and the weight of the shield plug and lifting yoke is greater than the stored weight of one fuel assembly. However, the fuel in the DSC is not " stored" in the pool and the prohibition against move-ment of a load heavier than an assembly plus its lifting gear refers to " stored" fuel GPU has sought to resolve the ambiguity by modifying the [ technical specification] to clarify that the shield plug may be moved onto the DSC after the DSC has been loaded with irradiated fuel.

Exh. A at page 1 of Safety Evaluation. This again indicates that Technical Specification 5.3.1.B refers only to " stored" fuel and spent fuel in the transfer cask within the CDPS is not " stored" fuel.

The current change to Technical Specification 5.3.1.B was not sought because of any change in meaning or intent. Rather in light of the current regulatory climate in which both the NRC and licensees are particularly sensitive to the need for a well-defined and understood licens-ing basis, it was decided that an amendment clarifying the technical specification would be desir-able. The purpose of the technical specification change is merely to make it more explicit. j

4 2.

NRC Staff and Licensee Documents Related to Technical Specifica-tion 5.3.1.B Establish that its Intent and Scope Is Spent Fuel Stored in Storage Racks and Not Spent Fuel Being Packaged for Transport in the CDPS While the meaning of the original Technical Specification 5.3.1.B is clear on its face, the regulatory history of the provision confirms that it was always intended to apply only to spent fuel stored in the storage racks.E Technical Specification 5.3.1.B was originally adopted for Oyster Creek as part of an amendment which increased the spent fuel pool storage capacity from 840 to 1800 fuel assemblies by replacing existing fuel storage racks with new closer-spaced stor-age racks. Fornicola Aff.,5 6 and Exh. 3 at cover' letter and page 2 of enclosure 3 (safety evaluation).

The initial amendment request for the spent fuel pool expansion clearly differentiated the area of the spent fuel storage racks as an area separate and distinct from the CDPS. Fornicola Aff.,5 7 and Exh. 4 (Jersey Central Power & Light Company, Request for Amendment to Provi-sional Operating License No. DPR-16 -- Technical Specification Change Request No. 44 and Fa-cility Description and Safety Analysis Report Amendment No. 78 (Mar. I8,1976)) at pages 10.0-16,10.0-17 of FDSAR amendment. This clear differentiation between the spent fuel stor-age area and the CDPS is continued throughout subsequent correspondence between the NRC staff and the Licensee on the amendment request. Although Technical Specification 5.3.1.B was not part of the initial application, (see 11 at pages 6-7 of the introductory material), in the process 11 Where the text of a requirement is considered ambiguous, the documents related to it should be considered to determine its meaning. Su Wisconsin Pub. Intervenor v. Mortier. 501 U.S. 59/,610 n.4 (199I)(applying the canon to use legislative history in interpretation of a statute).

l of reviewing the amendment request the NRC staff asked questions about the structural integrity of the new spent fuel storage racks, which eventually resulted in Technical Specification 5.3.1.B.

l Fornicola Aff.,510 and Exh. 5 (U.S. Nuclear Regulatory Commission, Request for Additional Information, Oyster Creek Nuclear Generating Station Spent Fuel Pool - Increased Storage Ca-J pacity (June 24,1976)) at pages 1-6.

Thus, question 39 from the NRC staff asked the Licensee to:

Provide (1) the number cf bundles that could be struck by a cask j

fall or tip, including effects of any superstructure on the cask; (2) a conservative analysis of fission product release from fuel bundles potentially subject to impact assuming that the most recently off-loaded fuel is in the impact area; (3) a realistic (best estimate) ra-diological analysis of a cask fall or tip; and (4) any technical speci-fications proposed on the decay time required prior to loading storage positions within the zone which co tid be struck by a cask fall or tip.

11 at 5. Licensee's response to Questicn 39 was that "[a] cask drop accident on or near stored

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fuel assemblies is not anticipated since the Oyster Creek spent fuel pool is equipped with a cask drop protection system (CDPS)," and the cask "will not be moved over the fuel storage area at any time." Fornicola Aff., $ 11 and Exh. 6 (Jersey Central Power & Light Company, Supple-ment No. I to Facility Description and Safety Analysis Report Amendment No. 78 (Aug. I1, 1976)) at page 39-1 of the FDSAR amendment. This again clearly differentiates the CDPS from the location of" stored fuel assemblies" and the " fuel storage area."

Question 40 from the NRC staff asked the Licensee to:

d l

Discuss the overhead cask handling system from the points of view of(l) yoke and/or cable failure, and (2) braking devices, their ca-pacity and effect on the ability of the handling system to withstand possible sudden decelerations induced by rapid braking following a loss of power to the system. Discuss all typical loads that may be carried near or over the spent fuel pool.

Fornicola Aff., j 12 and Exh. 5 at pages 5-6. Licensee responded to Question 40 that "[s]ince the

]

cask will not be moved over the fuel storage area, a yoke and/or cable failure is not expected to have any effect on stored assemblies." Fornicola Aff.,i 12 and Exh. 6 at page 40-1 of the j

i FDSAR amendment. This too differentiates the CDPS and the safe load path followed by the 1

cask from " fuel storage area" and " stored assemblies." It is important to note, as Licensee has previously identified, that the shield plug will follow the same safe load path as the cask. Forni-i cola AtT., j 3 and Exh. 2 at unnumbered page 4. Licensee's response to Question 40 also added that "[d]uring normal operation loads over the spent fuel pool will be limited to spent fuel assem-blies, weighing approximately 700 lbs." Fornicola Aff.,i 13 and Exh. 6 at page 40-1 of the FDSAR amendment. As a follow-up, the NRC staff requested the Licensee to "[p]ropose Tech-nical Specifications to limit the weight ofloads moved over stored fuel in the spent fuel pool."

Fornicola Aff.,i 14 and Exh. 7 (Jersey Central Power & Light Company, Revision No. I to Technical Specification Change Request No. 44 and Addendum No. I to Supplement No.1 Amendment No. 78 of the Facility Description and Safety Analysis Report (Nov. 30,1976)) at page 40-2 of the FDSAR amendment. Licensee responded with "a proposed Technical Specifi-cation Change to limit the maximum weight ofloads moved over the stored fuel in the spent fuel

-l8-

l' pool.. " 11 The proposed technical specification was labeled 5.3.1.D,2 but is otherwise iden-tical to what became Technical Specification 5.3.1.B and read:

Loads greater than the weight of one fuel assembly shall not be moved over stored irradiated fuel in the spent fuel storage facility.

Fornicola Aff., $ 14 and Exh. 7 at 5.3.1.D. The NRC staff request, Licensee response, and new technical specification all include storage-related terms including " stored fuel," " stored irradiated fuel," and " fuel storage facility." This usage is consistent with the terms discussed earlier, in-cluding " fuel storage area" and " stored assemblies,," that are used throughout the documentation of the amendment request.

The NRC issued the amendment to replace the existing fuel storage racks with higher ca-pacity spent fuel storage racks on March 30,1977, with the technical specification on loads over stored fuel adopted as proposed. Fornicola Aff.,515 and Exh. 3 at unnumbered page 1. This technical specification, now labeled 5.3.1.B was adopted over three years prior to the initial pub-i lication of NUREG-0612.2 Four new technical specifications were added as a result of this amendment,5.3.1.B - 5.3.1.E. Fomicola Aff.,5 2

14 and Exh. 7 at unnumbered page 7. Three of the four were subsquently moved to another section of the techni-

)

cal speci6caricas, leaving the subject technical specification to be renumbered 5.3.1.B.

j u

Fornicola Aff.,5 6 and Exh. 3 at unnumbered page 6 ofintroductory materials. The letter granted Amend-ment 22 and established the contents of Technical Specification 5.3.1.B. The technical speci6 cation was adopted as part of Amendment 22 in March 1977; NUREG-0612 was not published until July 1980. Table 3.2-1 of l

NUREG-0612 mistakenly indicated that Oyster Creek did not have a such a technical specification (one comparable j

to standard technical specification 3.9.6.2). This oversight regarding the Oyster Creek technical specification may have resulted from the fact that the Technical Specification 5.3.1.B is in Section 5, and not in Section 3, of the Oys-ter Creek technical specifications. l l

l*

In issuing the amendment to change the spent fuel storage racks the NRC specifically

' stated that the amendment will:

l continue to accommodate one fuel assembly shipping cask for off-site shipping of spent fuel assemblies from the Oyster Creek spent fuel pool when offsite spent fuel shipment is resumed at some in-definite future date....

l Fornicola Aff., j 17 and Exh. 3 at unnumbered page 1 ofintroductory materials. This NRC state-ment shows that both the stafTand the Ucensee understood that adopting Technical Specification l

5.3.1.B. which is part of the subjec* amendment, does not prohibit future " shipping of spent fuel l

assemblies from the Oyster Creek spent fuel pool." Id. This again indicates that the packaging l

l area for loading shipping casks is separate and distinct from the spent fuel storage area. This is consistent with the questions and responses related to Amendment 22, discussed SMpJ[a, that dif-l ferentiate the CDPS from the location of" stored fuel assemblies" and the "fue! storage area."

l Technical Specification 5.3.1.B thus has always allowed spent fuel to be packaged in the l

CDPS for transport out of the facility. The NRC statTand Licensee documentation related to adopting Technical Specification 5.3.1.B show it applies specifically to " stored fuel" in the " fuel storage area" and it does not apply to the CDPS. where the shield plug would be handled.

3.

The Standard Technical Specification Recommended by NUREG-0612 is Specifically for Stored Spent Fuel in Storage Racks and Does Not Mean Spent Fuel Heing Packaged for Transport l

l While Technical Specification 5.3.1.B was not issued in response to NUREG-0612, even i

l if the technical specification were interpreted in a manner consistent with the recommendations l

l,

of NUREG-0612. its meaning would not change. The technical specification recommended by NUREG-0612 is also limited to spent fuel stored in racks.

The interim protection measure specifically recommended by NUREG-0612 was imple-mentation of"a specification comparable to.. Standard Technical Specification 3.9.6.2. ' Crane Travel,' for BWR's." NUREG-0612 at 5-18.E This Standard Technical Specification states that:

Loads in excess of(2500) pounds shall be prohibited from travel over fuel assemblies in the scent fuel storage nool racks. (empha-sis added)

Fornicola Aff., j 19 and Exh. 8 (U.S. Nuclear Regulatory Commission, NUREG-0123, Standard Technical Snecifications for General Electric Boiling Water Reactors (Rev. 21979)) at page 3/4 9-9 (emphasis added). The " Applicability" of this Standard Technical Specification is specifi-cally limited to arcar of the spent fuel pool "with fuel assemblies in the spent fuel storage nool l

racks " li(empha,is added). Spent fuel storage pool racks are a set of specific structural ele-ments arranged in tne spent fuel pool to hold spent fuel until such time as it is transferred to the packaging area for transport out of the facility. The storage racks are different from, and inde-pendent of the CDPS that is used for packaging spent fuel in a transfer cask. Fornicola Aff.,5 8.

The Standard Technical Specification used as the model for interim protection recommendation (1) of NUREG-0612 thus applies only to heavy loads over spent fuel in storage racks, and not to 4

spent fuel being packaged for transport in the CDPS.

This technical specification is now Standard Technical Specification 3.9.7. Ett Fornicola Aff.,519 and Exh.

H 8 at page 3/4 9-9.

t While this interim recommendation was not adopted in the initial December 22,1980 Ge-neric Letter. subsequent reviews considered by the NRC staff considered the prior recommenda-tion and determined that Oyster Creek's Tecimical Specification 5.3.1.B (as it existed prior to the recent amendment)is comparable to the standard technical specificationF The comparability of j

Technical Specification 5.3.1.B again shows that the provision allows spent fuel to be packaged in the CDPS for transport out of the facility. The Standard Technical Specification that forms the j

basis for the recommendation of NUREG-0612 applies only to heavy loads over spent fuel pool storage racks and not to spent fuel in the CDPS.,

i 4.

Interpreting Technical Specification 5.3.1.B to Prohibit Loading Casks in the CDPS Leads to the Absurd Result that Spent Fuel Can Never Be Removed from the Facility The text of a requirement should not be interpreted to reach an absurd result. See Umted States v. Wilson, 503 U.S. 329, i12 S. Ct.1351,1354 (1992)(applying the canon to interpretation of a statute). Petitioners' Basis C, as summarized by the Board, asserts that the Technical Specification is a prohibition on the movement of heavy loads that "cannot be revised as the licensee has requested." LBP-96-23,1upla note 1, at 12. If the technical specitication could not have been revised as requested, and has the meaning petitioners attribute to it, the Licensee would be precluded from using a shield plug over irradiated fuel in the transfer cask.

A Franklin Research Center, Technical Evaluation Reoort - Control of Heavy Loads (June 10,1983), attached to Exh. C, at 22-23. Sig ahn U.S. Nuclear Regulatory Commission, NUREG-1382, Safety Evaluation Renort related to the Full-Term Oneratine License for Ovster Creek Nuclear Generatine Statica, attached as Exhibit E hereto, at 9-4 (1991).

Without shielding, the spent fuel could never be removed from the facility? Interpreting the text to require spent fuel to remain in the reactor building indefinitely is an absurd result that should not be attributed to Technical Specification 5.3.1.B.

The "to the extent practical" language of NUREG-0612 avoids this absurd result.

NUREG-0612 included *.his proviso in its summarized defense-in-depth approach as follows:

(2) Define safe load travel paths through procedures and operator training so that to the extent practical heavy loads avoid being car-ried over or near irradiated fuel or safe shutdown equipment.

NUREG-0612 at 5-2. The NRC Generic Letter 85-11 also includes this proviso and illustrates its application. Exh. D, Encl. I at 1,4. Closing out " Phase II" of the heavy loads program, this Ge-neric Letter states:

[M] 3st of the risk appears to be associated with carrying heavy loads over or in a location where spent fuel could be damaged.

The single most important example ot'this concerns loads handled over the open reactor vessel during refueling (such as the reactor vessel head). However, as previously mentioned, this is limited to the extent practical and where necessary, is performed with a spe-cifically implemented program in conformance with the Phase I guidelines.

Many regulations directly require the use of shielding for spent fuel moved out of the spent fuel pool, includ-ing 10 C.F.R. ss 71.47 (external radiation standards for transport packages); 72.104 (criteria for direct radiation from Independent Spent Fuel Storage Installation ("lSFSI") operations); 72.106 (dose limits for individuals at boundary of the ISFSI controlled area); 72.126 (design requirements for ISFSI radiological protection); 72.128 (shielding requirements for spent fuel storage systems); and 72.236 (shielding requirements for spent fuel storage cask approval).10 C.F.R. (( 71.47,72.104,72.106,72.126,72.128,72.236 (1996). Without the shield plug in place, the transfer cask would violate these requirements.

If the requirements for shielding were not applied, Petitioners' interpretation would lead to an equally absurd position that the shield plug could be raised over spent fuel in the dry shielded canister (DSC) once it has been re-moved from the spent fuel pool, but not raised over the DSC while it is in the spent fuel pool. Thus. the only effect of Petitioners' interpretation would be to prohibit a licensee from taking advantage of the shielding effect of the wa-ter in the spent fuel pool.,

l

LL at 4. Handling of the reactor vessel head over spent fuel in the open reactor vessel must be done to remove the head and to reinstall the head. There is no alternative. Similarly, handling of the shield plug over spent fuel in the transfer cask must be done to install shielding in the cask.

Again, there is no alternative. Just as the reactor vessel head movement is enabled under the "to the extent practical" proviso, the shield plug movement must also be enabled.

The meaning of Technical Specification 5.3.1.8 should not be interpreted to reach the ab-i surd result that the spent fuel can never be removed from the facility.

l V.

CONCLUSION i

For all of the above stated reasons, the Board should fmd that there is no legal restriction j

to changing Technical Specification 5.3.1.B as the Licensee has requested, and thus the Petition-ers' contention should be dismissed.

Respectfully submitted, Q

w Ernest L. Blake, Jr.

David R. Lewis SHAW, PITTMAN, POTTS & TROWBRIDGE 2300 N Street, N.W.-

Washington, D.C. 20037-1128 (202) 663-8084 Counsel for Licensee Dated: November 15,1996 1 l l

l

November 15,1996 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE TIIE ATOMIC SAFETY AND LICENSING BOARD In the Matter of

)

)

Docket No. 50-219-OLA GPU NUCLEAR CORPORATION

)

(Tech. Spec. 5.3.1.B)

)

l (Oyster Creek Nuclear Generating Station)

)

ASLBP No. 96-717-02-OLA I

LICENSEE'S MOTION FOR

SUMMARY

DISPOSITION EXHIBITS A-E l

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k i

377958-01 / DocsDC1 l

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I x

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Exhibit A-p.y\\

j UNITED STATES j

NUCLEAR REGULATORY COMMISSION g

g WASHINGTON, D.C. 20006-0001 g %,**,,* f November 7, 1996

]

Mr. Michael B. Roche

~

Vice President and Director GPU Nuclear Corporaticn Oyster Creek Nuclear Generating Station P.O. Box 388 Forked River, NJ 08731 i

SbdJECT:

. ISSUANCE OF AMENDMENT RE: HANDLING HEAVY LOADS OVER IRRADIATED FUEL (TAC NO. M95233)

Dear Mr. Roche:

l The Commission has issued the enclosed Amendment No.187 to Facility Operating License No. DPR-16 for the Oyster Creek Nuclear Generating Station, in response to your application dated April 15, 1996.

f.

The amendment revises Specification 5.3.1.8 to allow the shield plug and the l

associated lifting hardware to be moved over irradiated fuel assemblies that l

are in a dry shielded canister within the transfer cask in the cask drop protection system.

L

. A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Reaister notice.

i Sincerely, I

(

a d 8. Eaton, Senior Project Manager Project Directorate I-2 l

Division of Reactor Projects - I/II l

Office of Nuclear Reactor Regulation j

Docket No. 50-219

Enclosures:

1.

Amendment No.187 to DPR-16 2.

Safety Evaluation cc w/encls:

See next page

!t

\\

i

M.'Roche Oyster Creek Nuclear GPU Nuclear Corporation Generating Station cc:-

- 1 Ernest L.'Blake, Jr.,' Esquire Shaw, Pittman, Potts & Trowbridge 2300 N Street, NW.

Washington, DC 20037 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road

-King of Prussia, PA 19406 BWR Licensing Manager GPU Nuclear Corporation j

1 Upper Pond Road Parsippany, NJ 07054 Mayor

' Lacey Township L

818 West Lacey Road Forked River, NJ 08731 Licensing Manager i

1 Oyster Creek Nuclear Generating Station l

Mail Stop:

Site' Emergency Bldg.

P.O. Box 388 Forked River, NJ 08731 Resident Inspector l

c/o U.S. Nuclear Regulatory Commission f.

P.O. Box 445 L

Forked River, NJ 08731 l

l Kent To'sch, Chief l

I New Jersey Department of l

Environmental Protection l

Bureau of Nuclear Engineering l

CN 415 Trenton, NJ 08625 l

I l

lO 1

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f* **cu 3"

"'i UNITED STATES i

i j

NUCLEAR REGULATORY COMMISSION f

WASHINGTON, D.C. 2066H201

O \\..... /

J GPU NUCLEAR CORPORATION ANQ JERSEY CENTRAL POWER & LIGHT COMPANY l

DOCKET NO. 50-219 i

OYSTER CREEK NUCLEAR GENERATING STATION 4

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.187 j

License No. OPR-16 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by GPU Nuclear Corporation, et-al.

j (the licensee) dated April 15, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set ferth in 10 CFR Chapter I; ll B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable t.ssurance (1).that the activities authorized i

by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be j

conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common l

defense and security or to the health and safety of the public; and i

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable i

requirements have been satisfied.

l i

1 e

l!O a

-r

4 E

A4

,au a

.. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-16 is hereby amended to read as follows:

(2)

Technical Soecifications The Technical Specifications contained in Appendices A and B, as l

revised through Amendment No.187, are hereby incorporated in the license.

GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMISSION i

il. '

T. Stolz, Dire or oject Directorate I-2 O

Division of Reactor Projects - I/II j

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: November 7, 1996 i

O

1 ATTACHMENT TO LICENSE AMENDMENT NO.187 FACILITY OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert 5.3-1 5.3-1 5.3-2 5.3-2 O

P 5.3 AUXILIARY EOUIPMENT O

5.3.1 Fuel Storage A.

De fuel storage facilities are designed and shall be maintained with a K-effective equivalent to less than or equal to 0.95 including all calculational uncertainties.

I g

i B.

1.

Imds 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.

2.

The shield plug and the associated lifting hardware may be moved over irradiated fuel assemblies that are in a dry shielded canister within the transfer cask in the cask drop protection system.

C.

De spent fuel shipping cask shall not be lifted more thar. six inches above the top plate of the cask drop protection system. Vertical limit switchr.s shall be operable to assure the six l

inch vertical limit is met when the cask is above the top plate of the cask drop protection l

system.

D.

The temperature of the water in the spent fuel storage pool, measured at or near the surface.

shall not exceed 125 F.

I E.

De maximum amount of spent fuel assemblies stored in the spent fuel storage pool shall be 2645.

BASIS De specification of a K-effective less than or equal to 0.95 in fuel storage facilities assures an ample margin from criticality. This limit applies to unirradiated fuel in both the dry storage vault and the spent fuel racks as well as irradiated fuel in the spent fuel racks. Criticality analyses were performed on the poison racks to ensure that a K-effective of 0.95 would not be exceeded. The analyses took credit for burnable poisons in the fuel and included r.1anufacturing tolerances and uncertainties as described in Section 9.1 of the FSAR. Calculational uncertainties described in 5.3.1. A are explicitly dermed in FSAR Section 9.1.2.3.9. Any fuel stored in the fuel storage facilities shall be bounded by the analyses in these reference documents.

The effects of a dropped fuel bundle onto stored fuel in the spent fuel storage facility has been amlyzed.

His analysis shows that the fuel bundle drop would not cause doses resulting from ruptured fuel pins that exceed 10 CFR 100 limits (1,2,3) and that dropped waste cans will not damage the pool liner.

Administrative controls over crane movements, which include mechanical rail stops, serve to prevent travel of the crane outside the analyzed load path over the cask drop protection system. A safety factor greater j

than 10 with respect to ultimate strength, and redundant shield plug lift cables provide adequate margin for

' the shield plug lift. Dese featuro, combined with operator training and required inspections, contribute to

{

the determination that dropping the shield plug onto a loaded dry shielded canister in the spent fuel pool is

(

not a credible event.

r OYSTER CREEK 5.3-1 Amendment No.: 77,79,77,J71,179,187

The elevation limitation of the spent fuel shipping cask to no more than 6 inches above the top plate of the cask drop protection system prevents loss of the pool integrity resulting from postulated drop accidents.

An analysis of the effects of a 100-ton cask drop from 6 inches has been done (4) which showed that the pool structure is capable of sustaining the loads imposed during such a drop. Limit switches on the crane restrict the elevation of the cask to less than or equal to 6 inches when it is above the top plate.

Detailed structural analysis of the spent fuel pool was performed using loads resulting from the dead weight i

of the structural elements, the building lands. hydrostatic loads from the pool water, the weight of fuel and racks stored in the pool seismic loads. loads due to thermal gradients in the pool floor and the walls, and dynamic load from the cask drop accident. Thermal gradients result in two loading conditions; normal operating and the accident conditions with the loss of spent fuel pool cooling. For the normal condition.

the containment air temperature was assumed to vary between 65*F and 110 F while the pool water temperature varied between 85 F and 125'F. The most severe loading from the normal operating thermal gradient results with containment air temperatmes at 65'F and the water temperature at _125"F. Air temperature measurements made during all phases of plant operation in the shutdown heat exchanger room, which is directly beneath pan of the spent fuel pool floor slab. show that 65'F is the appropriate minimum air temperature. The sper.t fuel pool water temperature will alarm control room before the water temperature reaches 120*F.

Results of the structural analysis show that the pool structure is structurally adequate for the loadings associated with the normal operation and the condition resulting from the postulated cask drop accident (5)

(6). The floor framing was also found to be capable of withstanding the steady state thermal gradient conditions with the pool water temperature at 150 F without exceeding ACI Code requirements. The walls are also capable of operation at a steady state condition with the pool water temperature at 140 F (5).

Since the cooled fuel pool water retums at the bottom of the pool and the heated water is removed from the surface, the average of the surface temperature and the fuel pool cooling return water is an appropriate j

estimate of the average bulk temperature: alternately the pool surface temperature could be conservatively i

used.

References 1.

Amendment No. 78 to FDSAR (Section 7) 2.

Supplement No. I to Amendment No. 78 to the FDSAR (Question 12) 3.

Supplement No. I to Amendment 78 of the FDSAR (Question 40) 4.

Supplement No. I to Amendment 68 of the FDSAR 5.

Revision No. I to Addendum 2 to Supplement No. I to Amendment No. 78 of FDSAR (Questions 5 and 10) 6.

FDSAR Amendment No. 79 7.

Deleted 4

I OYSTER CREEK 5.3-2 Amendment No. W. Y77,187 i

i

pa acog ye UNITED STATES l

j NUCLEAR REGULATORY COMMISSION g

f WASHINGTON, D.C. 206d&0001

• * * * *,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.187 TO FACILITY OPERATING LICENSE NO. DPR-16

)

i GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219

1.0 INTRODUCTION

By letter dated April 15, 1996, GPU Nuclear Corporation (GPU, the licensee) submitted a request for changes to the Oyster Creek Nuclear Generating Station (OCNGS) Technical Specifications (TS).

The requested changes would revise TS pages 5.3-1 and 5.3-2 to pern:it loads in excess of the current TS limits to be moved over a cask loaded with fuel assemblies in the spent fuel storage facility.

By letter of August 23, 1996, the licensee supplemented its request O1 with an analysis of criticality potential and of the radiological consewnces of a hypothetical drop of the shield plug.

The supplement did not change the staff's conclusions in its proposed no significant hazards consideration determination (May 8, 1996, 61 FR 20849).

2.0 BACKGROUND

At the Oyster Creek plant site, the process of transferring spent fuel assemblies from the spent fuel storage facility to the Independent Spent Fuel Storage Installation (ISFSI) includes placing a dry shielded canister (DSC) within a transfer cask into the cask drop protection system (CDPS) located inside the spent fuel storage facility. The CDPS protects the spent fuel pool and the irradiated fuel stored in racks in the spent fuel pool in the event the cask is dropped.

This movement does not involve the handling of a heavy load over irradiated fuel. The DSC is then loaded with spent fuel assemblies.

Before the DSC and the transfer cask in which it is contained can be removed from the spent fuel storage facility, the DSC shield plug must be lowered into the CDPS and placed in position on top of the DSC to serve as a radiological shield. The current TS is ambiguous regarding this movement because the DSC, at that point, contains irradiated fuel, and the weight of the shield plug and lifting yoke is greater than the weight of one fuel assembly.

However, the fuel in the DSC is not " stored" in the pool and the prohibition against movement of a load heavier than an assembly plus its lifting gear refers to

" stored" fuel. GPU has sought to resolve the ambiguity by modifying the TS to clarify that the shield plug may be c.oved onto the DSC after the DSC has been loaded with irradiated fuel. The proposed TS change would facilitate the

l off-load of spent fuel to Oyster Creek's ISFSI by permitting the licensee to lower the DSC shield plug into the CDPS and place it in position on top of the DSC after the DSC has been loaded with irradiated fuel. This movement will not involve the handling of a heavy load over irradiated fuel in the storage racks.

3.0 EVALUATION Section 5.3.1, Fuel Storage, reads as follows:

B.

Loads greater than the weight of one fuel assembly shall not be moved over stored irradiated fuel in the spent fuel storage facility.

In order to implement the changes described in Section 2.0 above, the licensee l

proposes to change the TS as follows:

l B.

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.

1 l

2.

The shield plug and the associated lifting hardware may be moved over irradiated fuel assemblies that are in a dry shielded canister within the transfer cask in the cask drop protection O

system.

O As indicated above, this section would enable the licensee to lift the DSC shield plug and associated lifting hardware over irradiated fuel assemblies in the DSC within the transfer cask in the CDPS.

j l

In addition to the proposed change to the TS, the licensee has updated the TS Basis to state that

" Administrative controls over crane movements, which include mechanical rail stops, serve to prevent travel of the crane outside the analyzed load path over the cask drop protection system. A safety factor greater than 10 with respect to ultimate strength, and redundant shield plug lift cables provide adequate margin for the shield plug lift.

These features, combined with operator training and required inspections, contribute to the determination that dropping the shield plug onto a loaded dry shielded canister in the spent fuel pool is not a credible event."

The NRC staff has completed its review of the proposed change, the reason for i

the change, and the safety analysis provided by the licensee. This NRC staff l

review and evaluation is limited to the specific issue of placing the DSC shield plug (a heavy load) in position on top of the DSC after the DSC has i

been loaded with irradiated fuel.

This review does not address the movement of other heavy loads. The staff has considered the guidance of NUREG-0612,

" Control of Heavy Loads at Nuclear Power Plants," and NUREG-0554, " Single-l

i l

3 i

i Failere-Proof Cranes for Nuclear Power Plants," and other guidance such as j

ANSI B30.9, " Slings," and ANSI B30.2, " Overhead and Gantry Cranes (Top Running

]

Bridge, Single or Multiple Girder Top Running Hoist)."

According to information provided by the licensee, the reactor building (RB) i crane has a main hoist capacity of 100 tons. The actual safety factors of the main crane for its 100-ton rated load are: cables 6.5:1; main hoist gearing 5.2:1; and main hoist brakes 120% capacity.

These safety factors are within the guidelines established in NUREG 0612. These safety factors are with respect to ultimate strength. As a result, when moving the shield plug and the lifting yoke with a combined weight of approximately 7 tons, a safety 4

l factor greater than 14 with respect to the 100-ton rated capacity of the RB crane will be provided, and greater than 70 with respect to the ultimate strength.

For the lifting yoke, a safety factor greater than 26 will be provided, based on the lifting yoke's 105-ton rated capacity. The least i

conservative safety factor is that for the shield plug lift cables. The f

safety factor is 11:1, based on the ultimate load of 22,800 pounds. The i

shield plug lift cables are redundant and each of the four has sufficient capacity to support the total weight of the 8000 pound shield plug.

The licensee has modified the RB crane to enhance its performance and reliability by improving the instrumentation and controls and has developed an i

error-free plan that includes a dedicated management team and a dedicated crew 4

who will be trained and on-shift. The plan also includes detailed operating instructions and procedures.

In its April 15, 1996, application the licensee committed to a special crane inspection that will be performed prior to each l

dry fuel storage campaign; the main hoist coupling, shaft, and hook will be examined by NDE [ncndestructive examination) prior to each campaign. The licensee has also stated that personnel training, crane inspections, testing,

.i and maintenance will be in accordance with ANSI B30.2.

Based en the considerations discussed above, the NRC staff concludes that the design features and modifications of the crane, the licensee's error-free plan and commitments, and the significant factors of safety described in the licensee's request for changes to the TS makes a drop of the shield plug extremely unlikely to the point of not being credible.

This proposed TS amendment specifically addresses the issue of placing the shield plug (a radiological shield for the dry shielded canister) on the DSC.

i Even though the event is not credible, the staff evaluated the potential radiological consequences that could result from a hypothetical drop of a shield plug that lands in a random position on top of the DSC resulting in 4

damage to the spent fuel within the DSC.

By letter dated August 23, 1996, GPU Nuclear provided an analysis of the radiological consequences of dropping the shield plug after fully loading the dry storage canister with spent fuel.

Sixteen fuel assemblies are damaged such that all of the gaseous radioactive materials in the fuel pin gaps is released into the secondary containment. This radioactivity is assumed to immediately mix with the air volume of the reactor building and be exhausted O

to the environment through the plant stack by the standby gas treatment system (SBGT).

The staff used the TACT 5 computer code to calculate the resulting i

4 radiation doses at the exclusion area boundary (EAB) and low population zone (LPZ) as defined in 10 CFR Part 100. The following assumptions and input parameters were used:

(a)

All 16 fuel assemblies (1/35 of the core) were exposed to the maximum neutron flux for three operational cycles. Therefore, a peaking factor of 1.5 was applied (consistent with the guidance in Regulatory Guide 1.25) to the 1930 MWt full power level for each assembly.

(b)

The free volume of the secondary containment of 1,800,000 ft was taken 3

from Table 6.2-11 of the Oyster Creek UFSAR.

l (c)

No credit was taken for scrubbing of activity by the fuel pool water.

(d)

Charcoal filter in the SBGT system credited with removing.90% of the radioactive iodine species.

l (e)

Consistent with the guidance in Regulatory Guide 1.25, the fraction of l

the fuel's radioactivity in the fuel pin gap (i.e., available for release from the damaged fuel) was assumed to be 10% of the radioactive l

l iodines and 30% of the noble gases.

(f)

The affected fuel had 10 years of decay in the fuel pool before loading into the cask. (For comparison, a second calculation assuming only 1 year of decay was performed.)

For the case where the fuel had decayed for 10 years, virtually the only gaseous radioisotope remaining in the fuel gap is the noble gas Kr-85.

Therefore, as would be expected, the TACT 5 code calculated zero thyroid dose at the EAB and LPZ.

The 2-hour whole-body whole-bodydoseattheLPZwere4.12X10',doseattheEABangthe30-day rem and 1.62 X 10' rem, respectively. As noted above, a case was run with only 1-year of radioactive decay for the spent fuel.

Although the TACT 5 code calculated some residual

-Iodine-131 in the source term, Kr-85 still dominated the resulting dose such that zero thyroid dose was calculated at doses were 7.36 X 10 rem and 2.90 X 10',the EAB and LPZ. The whole-body rem, for the EAB and LPZ respectively.

The siting criteria in 10 CFR Part 100 specify that the doses resulting from a spectrum of accidents not exceed 300 rem to the thyroid or 25 rem to the whole body for individuals at the EAB and LPZ boundaries, respectively. As implemented in NRC staff policy for the acceptable consequences of a fuel handling accident in Section 15.7.4 " Radiological Consequences of Fuel Handling Accidents" in NUREG-0800 " Standard Review Plan,"

resulting doses do not exceed 25% of the Part 100 criteria. The doses calculated by the staff for the postulated accident are well within (6 orders of magnitude below) the acceptance criteria in Section 15.7.4 of NUREG-0800.

Accidental criticality caused by the dropping of the shield plug onto the DSC is not a credible event not only because of the multiple protections against i

i dropping the plug but also because of the design specifications for the DSC.

l On the basis of the analysis presented in the NUHOMS SAR and independent i

confirmatory calculations performed by the staff, the staff concluded in the NUHOMS SER that the standardized NUHOMS-52B design and proposed operating

l 5

procedures are adequate to maintain the system in a subcritical configuration and to prevent a nuclear criticality accident and therefore satisfy 10 CFR 72.124 and 10 CFR 72.236(c), subject to the key factors assumed by the vendor in the analysis, specifically:

1) criticality safety calculations presented in the SAR and independent confirmatory calculations performed by the staff showing that criticality safety is ensured for a maximum initial U-235 fuel enrichment of 4.0 wt%, which was determined for the design basis GE-2 7x7 fuel assembly; and 2) the criticality safety analysis assuming a minimum boron density of 0.75 wt% boron in the borated stainless steel absorber plates. The key factors and assumptions used by the vendor in the criticality safety analysis are as follows: 1) maximum fuel enrichment of fuel assemblies stored in the standardized NUHOMS-52B system of 4.0 wt% U-235; 2) minimum of 0.75 wt%

boron loading in the neutron absorber plates; and 3) altered mechanical configuration of the array of fuel assemblies resulting from an accident not credible.

i In addition to the analyses provided by the NUHOMS vendor for the NUHOMS SAR and the NRC staff confirmatory calculations, GPU has provided an analysis for a configuration specifically applicable for Oyster Creek.

The analysis used the widely used industry standard Monte-Carlo code KENO-Va (developed by ORNL

[0ak Ridge National Laboratory]), and standard auxiliary codes and data to provide cross section information.

These provide an acceptable methodology to j

examine criticality aspects of relevant configurations.

GPU validated its use of this methodology by comparison calculations from the cask safety analysis report calculations.

With a full load of 52 fuel assemblies in the cask the hypothetical drop of the shield plug, based on the geometry of the system, would not be expected to affect more than 16 assemblies.

Expected damage would be some crushing of the upper part of the fuel assemblies, in the area of the upper end reflector region of the fuel, and result in little change in reactivity. GPU, however, has analyzed a configuration in which all 52 assemblies are moved together to form a tight, cylindrical bundle to maximize the reactivity increase.

The boron / stainless steel blocks are assumed to remain between the assemblies, but the compression lowers their effectiveness by removing the flux trap water gaps initially present. The normal fuel assembly configuration is maintained since it is near maximum reactivity for the materials involved. The fuel enrichment used was 2.63 wt% U-235 with no burnable poison and no burnup assumed. The 2.63 value bounds the fuel enrichments to be used for dry storage. The burnup provides a considerable conservatism since the actual burnup would average over 23 GWD/MT, which would offer little if any potential for forming a critical configuration. The result of this calculation was a k(eff) value of 0.957 at a 95/95 probability / confidence level, considering the uncertainties associated with KENO-Va and the canister design. This provides a reasonable demonstration that there is little probability of a criticality event from rearrangement caused by a shield plug drop.

j 4.0 SIGNIFICANT HAZARDS CONSIDERATIONS COMENTS i

The licensee's request for amendment was noticed in the FEDERAL REGISTER on 1 -

May 8, 1996 (61 FR 20849).

In the notice, the staff made a proposed determination of no significant hazards consideration and offered an 4

O opportunity for hearing. On June G,1996, Nuclear Information and Resource Service (NIRS), Oyster Creek Nuclear Watch (OCNW), and Citizens Awareness Network (CAN) jointly filed a request for hearing and petition to intervene.

Included in the hearing request were comments on the proposed no significant hazards consideration determination.

Petitioners allege that the proposed amendment (1) represents a significant increase in the probability of an accident, (2) creates the possibility of an accident not previously identified in the Safety Analysis Report and, (3) constitutes a significant reduction in the margin of safety.

The staff's response to these comments follows.

5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The Commission's regulations in 10 CFR 50.92 include three standards used by the NRC staff to arrive at a determination regarding whether a request for amendment involves no significant hazards considerations. The regulation states that the Commission may make such a final determination if operation of a facility in accordance with the proposed amendment would not (1) involve a significant increase _in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The following staff evaluation in relation to the three standards. demonstrates that the proposed TS amendment to place the DSC shield plug in position on top O

of the DSC to serve as a radiological shield does not involve a significant hazards consideration.

First Standard

" Involve a significant increase in the probability or consequences of an accident previously evaluated."

In accordance with the information provided by the licensee, the reactor building (RB) crane has a main hoist capacity of 100 tons. The actual safety factors of the main crane for its 100-ton rated load are: cables 6.5:1, main hoist gearing 5.2:1, and main hoist brakes 120% capacity. These safety factors are with respect to ultimate strength. As a result, when moving the shield plug and the lifting yoke with a combined weight of approximately 7 tons, a safety factor greater than 14 with respect to the 100-ton rated capacity of the RB crane will be provided, and greater than 70 with respect to the ultimate strength.

For the lifting yoke, a safety factor greater than 26 will be provided, based on the lifting yoke's 105-ton rated capacity. The least conservative safety factor is that for the shield plug lift cables. The safety factor is 11:1, based on the ultimate load of 22,800 lbs. The shield plug lift cables are redundant and each of the four has sufficient capacity to support the total weight of the 8000-pound shield plug.

The licensee has modified the RB crane to enhance its performance and i

reliability by improving the instrumentation and controls, and has developed i

an error-free plan that includes a dedicated management team and a dedicated crew, who will be trained and on shift along with detailed operating

?

instructions and procedures. The licensee has committed to a special crane

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inspection that will be performed prior to each dry fuel storage campaign; the 1

main hoist coupling, shaft, and hook will be examined by NDE prior to each campaign. The licensee has also stated that personnel training, and crane inspections, testing, and maintenance will be in accordance with ANSI B30.2.

Based on the above discussion, the staff concludes that when considering the qualitative analysis of the safety factors and RB crane enhancements, the event is so unlikely as to be non-credible.

Second Standard

" Create the possibility of a new or different kind of accident from any accident previously evaluated."

1 The accident to consider with respect to the proposed TS amendment is dropping a shield plug (a shield plug is a heavy load for Oyster Creek) that lands in a random position on top of the DSC, damaging the fuel within the DSC.

l l

As discussed above, under the first standard, an accident resulting from a i

plug drop is not a credible event and, therefore, does not create the i

possibility of a new or different kind of accident from any accident l

previously evaluated.

Third Standard

' Involve a significant reduction in a margin of safety."

The staff agrees with the licensee's conclusion that dropping the DSC shield plug onto a loaded DSC and damaging the spent fuel assemblies therein is not a credible event.

The staff finds that the proposed amendment does not involve a significant hazards consideration.

6.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendment. The State official had no comments.

7.0 ENVIROMENTAL CONSIDERATION i

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no 1

significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the l

amendment involves no significant hazards consideration (61 FR 20849).

_ - _ - _ ~..

8 In Section 5.0 of this safety evaluation the Commission has made a final no significant hazards consideration determination with respect to this amendment. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amencuent.

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation i< the proposed manner, (2) such activities will be conducted in compliance w the Commission's regulation a

and (3) the issuance of the amendment will nct be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Harold Walker, SPLB Howard J. Richings, SRXB Roger L. Pedersen, PERB Date: November 7, 1996 OV a

i i

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. ; ' * * ' " %,S.

Exhibit B UNITf D STATES

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• 1, NUCLEAR REGULATORY COMMIS$10N

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nasmac'oa. o erosst q

I Oece.ber 22. 1980 9

FF.300'd31*'Y5 4

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TO ALL-LICE l:5EES OF CeEPATI'iG PLANTS Af;D we=m"wN APPLICA!!TS FCR ODEPATING LICENSES A!:0

7. E HOLDERS OF CONSTRUCTION PERMITS

• 6-Gentlemen:

Subject:

Control of Heavy Loads i

In January 1978, the l:RC oublished NUREG-0410 entitled, "NRC Progran 'or the Resolution of Generic Issues Related to *;uclear Power Plants -

l Report to Congress." As part of this procra, the Task Action Plan for i

Unresoived Safety !ssue Task Nc. A-36, " Control of Heavy Loads Near Spent Fuel." was issued.

Pe have comoleted our review of lead handling ocerations at nuclear oower plants. A report describing the results of this review has been issued as f.tREG-0612. " Control of Heavy Loads at 'luelear Power 81 ants -

Resolution of TAP A-36."

This eeport contains several reconnendations j

to te implemented by all licensees and applicants to ensure the safe i

handling of heavy loads.

The purpose of this letter is to request that you review your controls for the handling of heavy loads to determine the extent to which the y

guidelines of Enclosure 1 are c esently satisfied at your facility, and to identify the changes and : odifications that would be required in order to fully satisfy these guidelines.

To excedite your coeoliance with this request, we have enclosed the following:

f.tREG-0612. " Control of Feavy Loads at *:uclear 8 ewer Plants" (Enclosure 1).

Staff Position - Inte"in Actions fur Control of Heavy Loads (Enclosure 2).

Request for Additional Infor ation on Control of Heavy Leads (Enclosure 3).

%ith the exceptionusf licensees 'or Indian 8oint 2 and 3. Zion 1 and 2 and Threa File Island 1 (These vere previously sent a letter)

I 310 319 L 32 i

_ December 22, 1960 l

l O

You are recuested to imolement the interir actions described in Enclosuro 2 as soon as possible but no later than 90 days from the date of this le:ter.

In order to enable the NRC to detemine whether operating licenses should be modified (10 CFR 50.5a(f)) operating reactor licensees are recuesteri to provide the following:

1.

Submit a report documenting the results of your review and the i

required changes and edifications. This report should include the infomation identified in Sections 2.1 through 2.a of Enclosure 3. on how the guidelines of NUREG-0612 will be satisfied. This report should be submitted in two parts according to the following schedule:

Subr'it the Section 2.1 information within six morAhs from the date of this letter.

Submit the Sections 2.2, 2.3 and 2.4 information wi'.hin nine months.

2.

Furnish confirmation within six months that ime'eTentation of those changes and modifications you find are ":sssary will comence as soon as possible without waiting

taff review, so that all such changes, beyond the above inurim actions, will be comoleted within two years of submittal of Section 2.4 for the above report.

2.

Furnish justification within six months for any changes or modifications that would be required to fully satisfy the guidelines of Enclosure 1 which you beldeve are not necessary.

De :-iteria in NUREG-0612 are also appli:able to applicants for operating licenses. Such soolicants are exoected to provide the information reouested by item 1 above and to meet the same schedule of implementation as Meicated in 2 above. Any item for which the implementation date is crio* to the expected date of issuance of an operating license will be censicered to be a prerequisite to obtaining that license.

For any date that cannot be met. furnish a proposed revised date, justi'ication for the delay, and any planned compensating safety actions durir.g the' interim.

l

!O l

3 3

j This reouest for infomation was a:creved by GA0 under e blanket 0

-learance number R0072 which exoires November 30, 1983. Coments on burden and duplication iaay be directed to the U.S. General Accounting Office, Regulatory Reports Review, Room 5106, 441 G Street, N.W., Washington. 0.C.

20548.

Sincerely,

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. IG1 h

Trre E*isenhut, ' Director Division Liceising

Enclosures:

1.

NUREG-C612 2.

Staff Position 3.

Regt'est for Additional Information ec: w/o nclosure (1)

Service List O

O

D.' OS LRE 2 STAFF 30$!*:CN -

INTERIM aC!DNS FOR CONTRCL GF -EA7f 53 5 (1) Safe load paths should be defined per the guidelines of Section 5.1.1(1) (See Enclosure 1);

(2) p mcedures should be develoced and implemented per the guidelines of Section 5.1.1(2) (See Enclosure 1);

l

( 3) Crane operators should be trained, cualified and conduct themselves

er tne guidelines of Section 5.1.1(3) (Set inclosure 1);

(.1)

Cranes should be inszected, tested, and maintained in accordance with the guidelines of Section 5.1.1(5) (See Enclosure 1); and (5)

In addition to the above, special attention should be given to procedures, equipment, and personnel for the handling of heavs icads over the core, such as vessel internals or vessel inspection 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 sufficient detail is provided and that instructions are clear and conciset (2) visual inspections of load bearing components of cranes, slings, and special lifting devices to identify (flaws or deficiencies that could lead to failure of the component; 3) accropriate repair and reolacement of cefective components; and (a) verify that the crane oDerators 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.

I l

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a REQ;!ST '08 100:~10N1L 1srequa ds os C0'.*40'. Of H!1VV (*A* S 1.

INTRO}J"IION Verification by the licensee that the risk associated with lead-handling failures at nuclear power plants is extremely lev vill require a systematie e alua-tion of all lead-handling syntans at each site. The fo11 ewing specific inferr.ation 1

requests have been organized to suppor: such a systematic approach, and provide a i

basis for the staf f's review ef the licensee's evaluation. Additionally, they have been organized to address separately the two hazards requiring investigation (i.e..

radiclogical censequences ef da= ape to fuel and unavailability censequen:es of da. age to certain systems).

The felieving general inferr.ation is provided to assis:

in this evalue: ion and reduce :he need for clarification as to the inter.: and expec:-

ed results of this inquiry.

Kisk reduction can be desenstrated by either of tve approaches:

The likelihood of failure is =ade extremely lov throuch eek:-eed s.

I handling-system design features (N".*1IG 0612. Section 3.1.6).

b.

The consequences of a f ailure can be shr.m :o be accentable (NURI. 0622. See: ten 5.J. Cri:eria I-IV).

Regardless ef the at; roach selec:ed, the general guidelines !

N7AEG 0612. Section $.1.1. shculd be satisfied to provide =4ximu=

l oractical defense-in-denth.

2.

Evaluations concernine radiological consequences er criticality safety, where used, can rely en ei:bar the adeption of generic analyses reported in NURIG 0612 requiring only verfication : hat these generic essi:=ptions are valid for a specific site, or employ a si:e-specific analysis.

3.

Syste=s required fer safe snu:do.m and centinued decay heat removal are at:e-stecific a-d are not, t5erefore identified in this.eques:.

'ndividusi elants sh:vid censider svstens and com;cnen:n identified in Fe;ula:nry Guide 1.28 Pesi:fon C.1 (except tnese systers er portier.s ef systers :Ma: are required solely fer (a) erarsenev cere c:elia.r

[

(b) pest-accident c:ntainrent heat re-t.41. er (c; cos;-accident centainment a:n:s:here cleanups, fer evaluation and rece;r.1:e :..2:

he a ;r:ach taken 1-this resoect is si-ilar te t's: ide-rified ;n Fe;ula: cry Outde 1.19. ?:si:1:n C.1.

The fact that a lead-Faadli p syste-nJv be rrever.:ef ft - :nerating Juring :lant cenditiens re-l quirin; : e actual er seten:ial use of sent of'trese syste p.

1s rec-P00R01GIM.

i centred in this requ2st for information.

4 The scope of this systematic review should include all heavy loads carried in areas where the potential for net.-

ce:plian:e with the acceptance criteria 6.' REG 0612, Section 5.1) exists. A sum =arv of typica) loads to be considered has been provided in WRIG 0612. Tame 3.1-1.

It is re ognized that sooe cranes will carry additional miscellaneous leads, so=e of which are not identifiable 1

in detail in advance.

In such cases an evaluation or analysis demenstrating the acceptability of the handling of a range of loads should be p avided.

5.

At some sites loads which must 'ae evaluated will include licensed shipping casks provided for the transportation of irradiated. fuel, solidified radicactive waste, spent resins.

4 or other byproduct material. D

• ne, under 10CTR71 is not evidence that lif ting devices for these shipping casks meet the criteria specified in Nt* RIG n612 Sections 5.1.1(4), 5.1 l'5), 5.1.6(1), or 5.1.6(3), as appropriate, and thus does ro eliminate the need to provide appropriate infer =ation concerning these devices. A tabulation (Attachmer. 5) is i

prov4ded to indicate multiple-site use of these shipping casks.

~he results of the licensee's evaluation, as reported in respense to this request, sheuid previde infermation st,fficient for the staff to conduct an in-dependent review to determine that the intent of this effort (i.e., the uniform reduction of the pctential hazard from load-handling-syste:s f ailures) has been satisfied.

l 2.

IfFORMAT'Ti REOUE! ED FROM THE LICE'4SEE 2.1 GE'4EDAL REOUIREMENTS FOR OVERHEAD HANDLING SYSTEMS W REG 0612 Section 5.1.1, identifies several general guidelines related to the design and operation of ovarhead load handling systems in the areas where spent fuel is stored, in the vicinity of the reactor core, and in ether areas of the plant where a load drop could result in damage to equipment required for safe i

shutdown or decay hett removal.

Information provided in resoonse to this section s'- t li i ff.ti'-

the extent of notentially hatardous lead-handling cparations at a site and the extent of conf:r=ance to appropriate Ioad-handling ;cidance.

erert the results of your review of plant arrar aments to identifv all everhead haniling systems free wh..n a load j

drop sav reruit in da age to any system required for ;1 ant j

shutde.m. er decay heat removal (taking n c r edi.t for a y l

"O Pl10R ORS NAL

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inte-locrs, te: 9 ical s;ec:f::sti:.s. :; erat:ng ;ro:edures, or detailed structural analysis).

2.

Justify the exclusion of any everhead handling syste from the above catescry by verifying that there is sufficient physical separation fr== any load-1:;act point and any

. safety-related cc ;onent to ;ereit a deter =ination by inspec-tion that no heavy load drop can result in damage to any system or component required for plant shutdown or decay heat removal.

3.

'Jith respect to the design and operation of heavy-load-handling systems in the reactor building and those lead-handling systems. identified in 2.1-1, above, provida your evaluation concerning ec=pliance with the guidelines of N'.* RIG 0612. Sectica 5.1.1.

he following specific informa-j tion should be 1icluded in your reply:

a.

Drawings or sketches sufficient to clearly identify the location of safe lead paths. spent j

fuel, and safety-related equipment.

b.

A discussion cf measures taken to ensure that load-handling operations remain within safe load 1

paths. including procedures, if any, for deviation from these paths.

1 c.

A tabulation of heavy. loads to be handled by each crane which includes the load identification. load weight. its designated lifting device, and verifi-cation that the handling of such load is governed by a wrii..- orocedure containing. as a minimus.

lO the informatic identified in NUREG 0612. Section l

5.1.1(2).

l d.

Verifi6ation that lifting devices idencified in 2.1.

3-c. above, comply with the requirements of ANSI N14 l

6-1978. or ANSI 330.9-1971 as appropriate. yor lift-l ing devices where these standards, as supplemented by NUREG 0612. Section 5.1.1(4) or 5.1.1(5), are not met, describe any ;roposed alternatives and deson-l strate their equivalency in terna of load-handling l

reliability.

i e.

Verification that ANSI B30.2-1976. Chapter 2-2. has been invoked with respect to crane inspection, testing, and maintenance. 'ahere any exceptien is taken to this standard. sufficient information should be provided to j

j demonstrate the equivalesey of proposed alternatives.

f.

Verificatien that crane design coeplies with the guide-lines of CFAA Specification 70 and Chapter 2-1 of ANSI B30.2-1976, including the demonstration of acuivalency of actual design require =ents for instances where specific compliance with these standards is net previded.

i i l.

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g.

IXCerti ns. t f an'. taken te AS$1 53G.2-1)*t with l

retre:t te c; erat: train:ng. qualificatien. ano cenda:t.

2.2 S?ICITIC RIO'.*!.EMINT3 TIR OVE:3EA3 KAN; LING SYSTEMS OFIRA!!NO IN THE REACTJR B'JILDING NURIG 0610. Secticn 1.1.a. Trevides guidel nes concerning the design and operation of load-handling syste=s in the vicinity of spent fuel in the reacter vessel or in storage.

Infermatica provided in respense tc this section should deocnstrate that adequate measures have been taken to ensure that, in this area, either the likelthood cf a lead drop which might da= age spent f.el is extremely s=411. or that ene estimated censequences of such a drop will not exceed the li=its set by the evaluation criteria of N"RIO 0612 j

Section 5.1. Criteria I through III.

1.

Identify by na=e. type, caea:1:y. and equip =ent designs:cr.

any cranes :hysically capable (i.e.

ignoring interlockr.

moveable =e:hanical s: ops, er operating procedures) of carrying Icads over spent fuel in the storage pool or in the reactor vessel.

i 2.

Jue:ify _ the exclusion of any cranes in this area f rer *.. o above category by verifying that they are incapable of carrying heavy loads or are ps.rmanently prevented from movement of heavy leads over stored fuel or into any location where, following any failure, such load =ay drep

'nto *.he reactor vessel or spent fuel stcrage pool.

(\\%/

3.

Identify any cranes listed in 2.2-1. above.'which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be carried and the basis for this evaluatien (i.e.. complete corr 11ance with NURIG 061". Section 5.1.6 or partial ce=-

pliance supplemented by suitable alternative or additional design features). For each crane so evaluated provide the load-handling-sys tem (i.e.. crane-Icad-combination) informa-i tion specified in Attachment 1.

4.

For crares identified in 2.2-1. above, net categorized accord-ing to 2.2-3. demonstrate that the criteria of NUF.IC Ct12.

Section 5.1. are satisfied. Corp 11ance with criterien IV will be demonstrated in rescense to Section 2.4 of this request.

• • ith res;ect to Criteria I through III. previde a discussten

=

of your evaluation of crane o;eration in the Reactor Building and your determinatien et cerpliance. This res;cnse should include the folleving inf:rmatien for each crane:

a.

Where reliance is placed on the installatica and use 4

I I

tit:a1 irterlocks or mechanical stops. Indicate tS circ = stances ua er which these protective devi:es p

can re temeved or bypassed and the administrative ;ro-

..:u re : in'.'nied te ensure precer authorization of such

.i'*ter.

Dascuu any related er proposed technical spec-if1=:ations concerning the bypass of such interlocks.

b.

k?.cre reliance is placed on the operation of the Stand-by Gas Treat =ent System, discuss present and/or croposed technical specifications and administrative er physical centrols provided to ensure that these assumptions re-main valid.

l k*here reliance is placed on other site-specific con-c.

siderations (e.g., refueling secuencing), previde present i

or proposed technical specifications, and discuss adminis-l trative or physical controls provided to ensure the valid-l ity of such considerations.

d.

Analyses perfarned to demonstrate compliance with Criteria I threigh 1:1 should conform to the guidelines of Nt* REG 0612 l

A:pendix A.

Justify any exception taken to these guidelines.

l and provide the specific information requestid in Attachment 2.

)

t

3. o r I., as ap;repriate, for each analysis perfermed.

l l

2.3 SpECITIC REQ'.* REF.INTS FOR OVERHEAD IL\\NDLING SYSTEMS OPERATING IN PLANT l

AAEAS CONTAINING EQUIPMENT REQUIRID y0R REACTOR SHUTDOWN. DECAY HEAT RIP.0\\*AL. GR SPENT TUEL POOL COOLING

. p St* REC 0610, Section 5.1.5. provides guideJines concerning the design C

and operation of load-handling systems in the vicinity of equipment or com-penents required for safe reactor shutdown and decay heat removal.

Infor=a-tion provided in response to this section should be sufficient to demenstrate l

that adequate seasures have been taken to ensure that in these areas, either the likelihood of a load drop which might prevent safe reactor shutdoun or l

prehibit continued decay heat removal is extremely small, or that damege to such etuipment from load drops will be limited in order not to result in the loss of thesa safety-related functions. Cranes which must be evaluated in this section have been previously identified in your response to 2.1-1, and I

their loads in your res;ense to 2.1-3-c.

1.

Identify. any cranes listed in 2.1-1. above, which you have i

evaluated as having sufficient design fe:tures to -ske the l

likelihood of a lead drop extremely e.asil for all loads to be carried and the basis for thi:, evaluation (i.e.

ce=;1ete l

Fliance with.WREG 0612. Section 51.6 or partial ce:-

l r11ance supplemented by suitable alternative or additienal i

design features). F ?r each crane so evaluated. previde the lead-handline-system (i.e., erane-load-combination) infrrma-tion s;ecified in Attachment 1.

, p J

l

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l

l 2.

For any cranes identified in 2.1-1 net. designated as single-failure-preef in 2.3-1. a centrehensive hazard evaluation should be previded which includes the folleving in:ormation:

I a.

The ;resentatien 19 a :strix format of all heavy I \\

icads and petential 1:;act areas v5ere danare

\\ss sight occur to safety-related equip =en*..

Heavy loads identification should include designation and weitbt or cross-reference to ir. formation pro-vided in 2.1-3-c.

Impact areas sheuld be identi-fied by censtruction rones and elevaticas or by some other method such that the i= pact area can be located on.the plant general arrangement drawings. Figure 1 provides a typical matrix.

b.

For each interaction identified, indicate which of the load and impact area combinatiens can be eliminated because of separation and redundancy i

of saf et:-related equipment mechanical steps and/or electrical interlocks, or othat site-specific considerations. Elimination on the basis ef the aforecentiened consideration should be supplemented by the following specific informa-tion:

(1) For load / target combinations elizinated because of separation and redundancy ef-safety-related equipment discuss the basis for determining that load drops vill not affect continued system upera-tion (i.e., the ability of the system to perform its safety-related function).

l - (O,)

(2) *4here mechanical stops or electrical interlocks are to be provided, presant details showing the areas where crane travel will be prohibited. Additional-ly. provide a discussion concerning the procedures that are to be 'ssed for authorising the bypassing of interlocks

)

or removable stops, for verifying that j

interlocks are functional prior to crane use, and for verifying that interlocks are restored to operability after opera-l tiens which require bypassing have been j

completed.

l (3) Where load / target combinations are elim-insted on the basis of other, site-spec-ific censiderations (e.g., saintenance seque..ing) provide present and/cr pro-

-osed technical specifications and dis-cuss administrative procedures or physi-cal constraints invoked to ensure t'.e validity of such consideraticas.

-t.

k a

t

m___.___

i 5

c.

For interactions not elicinated bv the analysis of

'.3-2-5 above, identifv any handling sy stems for i

specific loads which yen '. ave evaluated as having i (

of a load drid extremely small and the basis for sufficient desian fe atur-= to ake the livv11 hood this evaluation (i.e., cocplete compliance with j

NUREC 0612. Section 5.1.6. or partial co=pliance l

supplemented by suitab?e alternative or addition-al design features). For each so evaluated, pro-l vide the load-handling-system (i.e.. crane-load-combination) information specified in Attachment 1.

i d.

For interactions not eliminated in 2.3-2-b or 2.3-

]

2-c. above. demonstrate using appropriate analysis that damage would not oreclude operation of suffi-4 cient equipment to allow the systes to perform its safety function folleving a load drop (KL'RIC 063 2 l

Section 5.1. Criterion IV).

For each analysis so conducted, the following information should be

~

provided:

5 (1) An indication of whether or not, for the specific lead baing investigated, the overhead crane-handling systes is designed and constructed such that the heinting system will retain its load in the event of seismic accelerations equivalent to those of a safe shutdown earthquake (SSE).

4 g

(2) The baats for any exceptions taken to the 4

l analytical guidelines of NUREG 0612. Ap-pendix A.

d i

1 (3) The information requested in Attachment 4 1

s

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i 4

3 i

l 4

I 4

1 3

1 d

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4 4

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NOTES To FIGURE 1 Note 1:

Indicate by symbols the safety-related equipment. The licensee should provide a list consistent with the clarification provided in 1.2-3.

Note 2: Hazard Eliminatien Categories Crane travel for this area / load combination prohibited a.

by electrical interlocks or mechanical stops.

b.

System redundancy and separation precludes loss of

{

capability of system to perform its safety-related j

function following this lead drop in this area.

1 c.

Site-specific consideratioas eliminate the need to con-sider load / equipment combination.

l l

d.

Likelihood of handling system failure for this load is j

extremely small (i.e. section 5.1.6 N" REC 0612 satis-fled).

e.

Analysis demonstrates that crane failure and load drop will not damage safety-related equipment.

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Attachnent (1)

SINGLE-FAILURE PR00r WANDLING SYSTEMS O

1.

Provide the name of the manufacturer and the design-rated load (DRL). If the maximum critical load (MCL), as defined in NL* REG 0554, is not the same as the DRL, provide this capacity.

2.

Provide a detailed. valuation of the overhead handling system with respect j

to the features of design, fabrication, inspection, testing, and operation as delineated in NUREC 0554 and suppleaentes' by the identified alternatives specified in NUREG 0612, Appendix C.

This evaluation must include a point-by-point comparison for ecch section of NUREG 0$$4 If the alternatives of NUREG 0612 Appendix C, are used for certain applications in lieu of complying with the reconsendation of NUREG 0554, this should be explicitly stated.

If an alternative to any of those contained in NUREG 0554 or NUREG l

0612 Appendix C, is proposed, details sust be provided on the proposed alternativetodemonstrateitsequivalency.1/

3.

With respecc to the seismic analysis employed to demonstrate that the over-head handling systes can retain the load during a seismic event equal to a safe shutdown earthquake, provide a description of the method of analysis.

l the assumptfons used, and the mathematical model evaluated in the analysis.

it

.ne desc'iption of assumptions should include the basis for selection of trolley and load position.

4 Provide an evaluation of the lifting devices for each single-failure-proof l

handling system with respect to the guidelines of NUREG 0612, Section 5.1.6 l

5.

Provide an evaluation of the interfacing lift points with respect to the 1

l guidelines of NUREC 0612. Section 5.1.6.

1/ If the crane in question has previously been approved by the staff as satisfying NUREG 0554. Reg. Guide 1.104, o: Part 8 to STP-ASBD-1, please reference the date of the staff's safety evaluation report or approval letter in lieu of providing the information requested by item 2.

I lO

Attata. eet (2)

AN*LYS!S CF RAO*0 LOG;;AL DELE'5ES The folleving inf:r=ation should be provided for an analysis conducted to demonstrate co=pliance with criterion I of NURI" 0612. Section 5.1.

1.

INITIAL CONDITIONS /ASSUMPTICNS Identify the tire after shutdown the number of fuel a.

avsemblies damaged. and the assumed duration of radio-

{

Icgical release associated with each accident ansivred.

b.

NUREG 0612. Table 2.1-2. provides the assumptions used to arrive at generic conclusions concerning radiological dose consequences. To rely on the radiological dose aralysis of NUREr. 0612. the licensen should verify that these assumstions are conservative uith regard to the plant / site evaluated.

If the assumptions are not cen-servative for the spe:ific plant. or if a ocre site-specific analysis ts required the licensee should identify plant-specific assumptions used in place cf those tabuisted.

Identify and provide the basis (e.g.. USNRC Regulatory Guide 1.25) for any assumptions emploved in site-specific analyses not identified in NUREG 0612. Table 2.1-2.

d.

Dese calculations based on the termination or sitigation of radiological releases should be supported by inferma.

tion sufficient tn demonstrate both that the time delay assumed is conservative and that the system provided to accomplish such termination or sitigation will perform its safety function upon demand (i.e.. tne systec meets the_ criteria for an Engineered Safety Feature). Specific information so provioed should include the following:

(1) Details concerning the location of accident sensors, parameters monitored and the values of these parameters at which a safety signal will be initiated, systee response time (including valve-operation time), and the i

total time required to automatically shift from normal operation to isolation or filtra-tion following an accident.

(2) A descriptien of the inst *urentation and con-trols associated with the Engineered Safety Teature which includes infermation sufficient te de.enstrate that the rettiirecent s (Section A) of IEEE 279-1971. " Criteria for protection Systems for Nuclear Fever Generating Stations."

are satisfied.

1 2-1 L

l l

(3) A description of any Engineered safety Testure filter syster which includes infor-mation sufficient to dementtrate compliance with the guidelines of USSRC Regulatory i

Guide 1.50. " Design. Testing. and Maintenance Criteria for Engineered Saf ety Testure Atmos-phere Cleanup System Air Filtration and Abscrption Units of Light-Water-Cooled Nuclear Power Plants."

(4) A discussien of any initial conditions (e.g.. manual valves ic:ked shut, containment airlocks or equipment hatches shut) necessary to ensure that releases will be terminated or sitigated upon Engineered Safety Teature actuatien and the measures employed (i.e.. Tech-nical Specification and administrative controls) to ensure that these initial conditions are satisfied and that Engineered Safety Teature systems are operable prior to the load lif t.

l

(

2.

METHOD OF AFALYSIS Discuss the method of analysis used to desonstract that post-accident dose will be well within 10CFkl00 limits.

In presenting methodelegy used in determining the radiological consequences, the following information should be provided.

a.

A description of the mathematical er physical model g-g g

employed.

b.

An identification and summary of any computer program used in this analysis.

c.

The consiaeration of uncertainties in calculational notheds, equipment perfor=ance, instrueertation response characteristics. or other indeteruinate effects taken into account in the evaluation of the results.

3.

CONCLUSION Provide an evaluation cesparint the results of the analysis to Criterion I of NURIC 0612. Section 5.1.

If the pestulated heavy-load-drep accident t

analyzed bounds other ;cstulated heavy-lead drops, a list of these bounded hcavy loads sheuld be previded.

i s

- ~

~.

3 Attachment (3) l I

^

I CRITICALITY ANALYSIS i

i "he following information should be provided for analysis coneucted to demon-strate compliance with criterion II of NUREG 0612. Section 5.1 s

i

)

1.

INITIAL CONDITIONS /AS$UMPTIONS L

The conclusions of NUREG 0612. Ser. tion 2.2, are based on a particular.

acdel fuel assembly. If a licensee uses the results of Section 2.2

{

rather than performing an independent neutronics analysis, the assusp-

]

tions should be verified to be compatible with plant-specific design.

Forany analysis conducted, the following assumptions should be provided as a minimust i

j a.

Wetar/U02 1""* '" I'

)

b.

The boron concentration for the refueling water and spent-fuel pool

}

c.

The amount of neutron poison in the fuel 1

d.

Fuel enrichment e.

The reactivity insertion value due to crushing of j

the core

}

f.

The k value allowed by technical specifications j

fort $$gcore during refueling l

j 2.

METN0D OF ANALYSIS Provide the method of analysis used to demonstrate that accidental dropping of a heavy load does not result in a configuration of the fuel j

such that k,gg is larger than 0. H.

N discussion of the method of analysis should include the following information:

s.

Identification of the computer codes employed b.

A discussion of allowances or compensation for calculation and physical uncertainties y

3.

CONCLU510N j

provide an evaluation cosparing the results of the analysis to Criterian II

{

of NUREG 0612. Section 5.1.

If the postulated heavy-load-drop accident 3-1

,iO l

4

bounds other postulated heavv-load drops, a list of these bounded heavy loads should be provided.

O 2-2 i

Attac ment (4) i ANALYSIS OF PLANT ST8UCiuttES l

The following info mation should be provided for analyses conducted to demon-strate co=pliance with Criteria III and IV of NVREC 0612. Section 5.1.

1.

INITIA1. CONDITIONS / ASSUMPTIONS Discuss the assu=ptions used in the analysis, including:

{

a.

Weight of heavy load b.

I= pact area of load c.

Drop height d.

Drop location e.

Assu=ptions regarding credit taken in the analysis for the action of i= pact 11=iters f.

Thickness of walls or floor slabs impacted 3

Assumptions regarding drag forces caused by the environment h.

Load combinations considered 1.

Material properties of steel and concrete

('

2.

METE 03 OT ANALYSIS Provide the =ethod of analysis used to demonstrate that sufficiar.t load-carrying capabiligv exists within the wall (s) or floor slab's1.

Identify any cc puter codes e= ployed, and provide a description of their capabilities.

If test data was employed, provide it and describe its applicability.

3.

CONCLUSIGN Provide an evaluation comparing the results cf this analysis with Criteria III and IV of NUREC 0612. Section 5.1.

Whste safe-shutdown equipnent has a ceiling or wall separating it from an overhead handling syste=, pr: vide an evaluation to dewnstrate that postulated 1 cad drops do not penetrate the ceiling or cause seccadary missiles that could prevent a safe-shutt.vn s;. s t e from performing its safety function.

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D6ibit C g,,., 'c.j NUCLEAR REGULATORY COMMISSION UNITED STATES i

t WASHINGTON. D. c. 20585

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%, "...a-June 21,1983 2

Docket No. 50-219 L505-83-06-045 i

Mr. P. B. Fiedl er Vice President and Director Oyster Creek Nuclear Generating Station f

Post Office Box 388 l

Forked River, New Jersey 08731 1

SUBJECT:

CONTROL OF HEAVY LOADS (PHASE 1) - NUREG-0612 -

OYSTER CREEK NUCLEAR GENERATING STATION

]

j Enclosed is a copy of our Safety Evaluation (SE) for Oyster Creek, which was developed based on your response to the December 22, 1980 generic letter on " Control of Heavy Loads." This SE was prepared i

after receivtng the Technical Evaluation Report (TER) prepared by the i

Franklin Research Center (FRC). We concur with the findings contained j

in the TER and conclude that the guidelines in NUREG-0612, Sections 5.1.1. and 5.3 have been satisfied and, therefore, conclude that Phase 1 for Oyster Creek is acceptable. The TER is attached to our SE.

Phase II of NUREG-0612 will be the subject of future correspondence.

l The issuance of this letter and the enclosed SE completes nur action or. Phase 1 of thisMtem.

i Sincerely.

^

9 2d M./ &

h Dennis M. Crutchfield, Chief f

Operating Reactors Branch #5 Division of Licensing i

Enclosures:

l Safety Evaluation and the Technical Evaluation Report Attached thereto ec w/ enclosures:

See next page j

b i:

!yk \\ f.I M b d N l

j i

_ _ _ _._.__ _.. _ _ _ _.._ __._.~.. _ __._.

I

. June 21, 1983 Mr. P. B. Fiedler cc G. F. Trowbridge, Esquire Resident Inspector Shaw, Pittman, Potts and Trowbridge e/o U. S. NRC 1800 M Street, N. W.

Post Office Box 445 Washington, D. C.

20036 Forked River, New Jersey 08731 J. B. Liebennan, Esquire Comissioner l

Berlack, Israelt, & Lieberman New Jersey Department of Energy.

26 Broa&tay 101 Connerce Street New York, New York 10004 Newark, New Jersey 07102 Dr. Thomas E. Murley, Frank Cosolito, Acting Chief l

Regional Administrator Bureau of Radiation Protection Nuclear Regulatory Comission, Region I Department of Environmental f

631 Park Avenue Protection King of Prussia, Pennsylvania 19406 380 Scotch Road Trenton, New Jersey 08628 j

J. Knubel 1

BWR Licensing Manager GPU Nuclear l

100 Interplace Partway l

Parsippany, New Jersey 07054 j

i Deputy Attorney General State of New Jersey l

Department of Law and Public Safety l

36 West State Street - CN 112 Trenton, New Jersey 08625 i

l Mayor Lacey Township 818 Lacey Road i

Forked River, New Jersey 08731 U. S. Environmental Protection i

Agency Region II Office l

ATTN: Regional Radiation Representative l

26 Federal Plaza l

New York, New York 10007 l

Licensing Supervisor 1

Oyster Creek Nuclear Generating Station i

Post Office Box 388 Forked River, New Jersey 08731 l

O s

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UNITED STATES 4

NUCLEAR REGULATORY COMMISSION

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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION DOCKET N0. 50-219 CONTROL OF HEAVY LOADS - PHASE 1 i

r

1.0 INTRODUCTION

As a result of Generic Task A-35. " Control of Heavy Loads Near Spent Fuel," NUREG-0612. " Control of Heavy Loads at Nuclear Plants" was developed.

Eollowing the. issuance of NUREG-0612 a generic letter, dated December 22, 1980, was sent to all operating plants, applicants l

for operating licenses and holder of construction permits requesting that responses be' prepared to indicate the degree of compliance with the guidelines of NUREG-0612. The responses were to be made in two stages.

The first response (Phase 1) was to identify the load handling equipment within the scope of NUREG-0612 and to describe the associated

~

general load handling operations such as safe load paths, procedures, operator training, special and general purpose lifting devices the maintenance, testing and repair of equipment and the handling equipment specifications.

The second response (Phase 2) was intencled to show i

that either single-failure-proof handling equipment was not needed or that single-failure-proof equipment had been provided.

This safety evaluation contains the staff's evaluation of Phase 1.

An evaluation of Phase 2 will be the subject of future correspondence.

2.0 EVALUATION AND CONCLUSION By letter dated December 22, 1980 the General Public Utilities Nuclear Corporation, the licensee for Oyster Creek was. requested to review their provisions for handling and control of heavy loads at Oyster Creek to determine the extent to which the guidelines of NUREG-0612 are presently satisfied and to discuss and commit to mutually agreeable changes and modifications that would be required in order to fully satisfy these 4

i cuidelines. The staff and its consultant, the Franklin Research Center l

IFRC), have reviewed the General Public Utilities Nuclear Corporation submittals for Oyster Creek.

As a result of its review FRC has issued a Technical Evaluation Report (TER). The staff has reviewed the TER and concurs with its findings that the guidelines in.NUREG-0612 Sections 5.1.1 and 5.3 have been satisfied. The staff, therefore, concludes that Phase 1 for Oyster Creek is acceptable.

. _ _ _ _... ~ _ _... _. - _.. _ _ _ _ _ _... _.... _ _ _ _. _ _.. _. _..... _ _ _. _ _ _

l 2

i 3.0 ACKNOWLEDGEMENT

. The following NRC employee was the principal contributor to this SE:

A. Singh.

Attached: Technical Evaluation Report, dated June 10,'1983.

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Date:

June 21,1983 I

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eCHNICAL EVALUATION Rer ORT 3

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g CONTROL OF HEAVY LOADS (c.10)

JERSEY CENTRAL POWER AND LIGHT COMPANY 0YSTER CREEK NUCLEAR POWER PLANT NRC OOCKET NO. 50-219 FRC PROJECT C5506 NRCTAC NO. 47128 FRC ASSIGNMENT 13 i

NRC CONTRACT NO. NRC-03-81130 FRCTASK 377 l

l l

1 Prepared by j

Franklin Research Center Author:

C. Bomberger N. Ahmed i

20th and Race Streets I

Philadelphig, PA.19103 FRC Group Leader:

I. E. Sargent I

b' Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer:

A. Singh T. Chan j

June 10, 1983 This report was prepared as an account of work sponsored by an agency of the United States Govemment. Neither the United States Govemment nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, appa-ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.

Prepared by:

Reviewed by:

Approved by:

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Principal Autdor M up Lalder Department D/ rect 8r 1~##~O G/#M8 Date: L 19 ft5 Date:

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CONTENTS Section Title Pace 1

l 1

INTRODUCTION.

1.1 Purpose of Review 1

r 1.2 Gen 3ric Background.

1

-1.3 Plac.t-Specific Background 2

4 2

EVALUATION 2.1 General Guidelines.

4 2.2 Interim Protection Measures.

22 25 3

CONOLUSION 3.1 General Provisions for Load Handling 25 25

{

}

3.2 Interim Protection Hessures.

4 REFERENCES 27 l

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4 I

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i TER-C5 50 6-37 7 FOR2HORD

,t Tnis Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions.

Se I

tschnical evaluation was conducted in accordance with criteria established by the NRC.

l l

Mr. C. R. Bomberger and Mr. I. H. Sargent contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.

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TIR-05 50 6-37 7 7

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INTRODICTION l

l 1.1 PURPOSE OF REVIIW This technical evaluation report documents an independent review of general load handling policy and procedures at the Jersey Central Power &

Light Company (JCP&L)/ General Pub'lic Utilities' (GPU) Oyster Creek Nuclear Power Plant.

This evaluation was performed with the following objectives:

to assess conformance t.o the general load handling guidelines of o

NUREG-0 612, " Control of Beavy Loads at Nuclear Power Plants" (1),

Section 5.1.1 to assess conformance to the interim protection measurais of o

~

NUREG-0 612, Section 5.3.

1.2 GENERIC BACKGROUND Generic Techn,ical. Activity Task A-36 was established by the U.S. Nuclear l

R;gulatory Commission (NRC) staff to systematically examine staf f licensing criteria and the adequacy of measures in effect at operating nuclear power f

plants to ensure the safe handling of heavy loads and to recommend necessary This activity was initiated by a letter issued by changes in these measures.

to all power reactor licensees, requesting the NRC staf f on May 17; 1978 [2]

information concerning the control of heavy loads near spent fuel.

The results of Task A-36 were reported in NUREG-0 612, " Control of Beavy i

Loads at Nuclear Power Plants."

The staff's conclusion from this evaluation was that existing measures to control the handling of heavy loads at operating I

plants, although providing protection from certain potential problems, do not edequately cover the major causes of load handling accidents and should be upgraded.

In order to upgrade measures for the control of heavy loads, the staff developed a series of guidelines designed to achieve a two-part objective j

using an accepted approach or protection philosophy. The first portion of the objective, achieved through a set of general guidelines identified in i

I NUREG-0612, Section 5.1.1, is to ensure that all load handling systems at

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'.".':.' F;anklin Res.earch Center l

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sclear power plants are designed and operated so that their probability of j

l f ailure is uniformly small and appropriate for the critical tasks in which i

th2y ar e employod, me second portion of the staff 's objective, achieved is through guidelines identified in NUREG-0612, Sections 5.1.2 through 5.1.5, for load handling systems in areas where their f ailure might 4

1 j

to ensure that, l

result in significant consequences, either (1) f eatures are provided, in i

addition to those required for all load handling systems, to ensure that the potential f or a load drop is extremely small (e.g., a single-f ailure-proof 1

l conservative evaluations of load-handling accidents indicat'e I

cr an e) or (2) that the potential consequences of any load drop are acceptably small.

l Ac=eptamility of accident consequences is quantified in NUREG-0612 into four 4

accident analysis evaluation criteria.

f A defense-in-depth approach was used to develop the staff guidelines to j

ensure that all load handling systems are designed and operated so that their The intent of this guideline f

f probability of f aiJure is appropriately small.

is to ensure that licensees of all nuclear power plants perform the following:

define safe load travel paths through procedures and operator training

.i l

o to the extent practical, heavy loads are not carried over or J

so that, j

j near irradiated fuel or safe shutdown equipment provide sufficient operator training, handling system design, load

]

h o

handling instructions, and equipment inspection to ensure reliable operation of the handling system.

l Staff guidelines resulting from the foregoing are tabulated in Section 5 Section 6 of NUREG-0 612 recommended that a program be of NUREG-0 612.

initiated to ensure that these guidelines are implemented at operating plants.

l J

1.3 PLANI-SPECITIC BACKGROUND On December 22, 1980, the N1C issued a letter *[3] to JCP&L, the Licensee for the Oyster Creen plant, requesting that the Licensee review provisions for handling and control of heavy loads at the Oyster Creek plant, evaluate these provisions with respect to the guidelines of NUREG-0612, and provide certain additional information to be used for an independent determination of

)

conformance to these guidelines.

On September 22, 1981, JCPEL provided the i

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i TTA-C 5 50 6-3 77 nitial response [4] to this request.

A draf t tecnnical evaluation report

(;r.E) was prepared based on this information and was informally tr ansmitted t.o

be Licensee f or review and consnent.

On July 9,1982, a telephone conf erence call was conducted with the representatives of NRC, FRC, and JCP&L to discuss unresolvec issues.

As a result of this call, additional information was I

provided by the Licensee on February 18, 1983 [5] and on May 27, 1983 [6]

which has been incorporated into this final technical evaluation.

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1 TEF.-C5506-377 i

2.

EVALUATION l

i This section presents a poi.nt-by-point evaluation of load handling j

f provisions at Oyster Creek Nuclear Power Plant with respect to NRC staff cuidelines provided in NUREG-0612.

Separate subsections are provided for both l

the general guidelines of NUEG-0612, Section 5.1.1 and the interim measures 4

of NUREG-0612, Section 5.3.

In each case, the guideline or interim measure is prssented, Licensee-provided information is summarized and evaluated, and a, conclusion as to the extent of compliance, including recommended additional action where appropriate, is presented.

':nese conclusions are summarized in I

Table 2.1.

g

)

I 2.1 GENERAL GUIDELINES The N C has established seven general guidelines which must be met in order to provide the defense-in-depth approach for the handling of heavy These guidelines consist of the following criteria from section 5.1.1 loads.

g k/ of NUREG-0 612:

Guideline 1 - Safe Load Paths o

Guideline 2 - Icad Handling Procedures o

Guideline 3 - Crane Operator Training o

o Guideline 4 - Special Lifting Devices l

Guideline 5 - Lifting Devices (Not Specially Designed) o

\\

Guideline 6 - Cranes (Inspection, Testing, and Maintenance) f o

j Guideline 7 - Crane De sign.

o

[

These seven guidelines should be satisfied for,all overhead handling systems that handle heavy loads in the vicinity of the reactor vessel, near 5

spent fuel in the spent" fuel pool, or in other areas where a load drop may The Licensee's verification of the extent to damage saf e shutdown systems.

which these guidelines have been satisfied and the evaluation of this verification are contained in Sections 2.1.1 through 2.1.8 of this report.

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TER-C5506-377 3,1.1 NUREG-0612, Eeavv Loyds Overhead Handline Systems a.

Summarv of Licensee Statements and Conclusions The Licensee has evaluated the load handling systems at the Oyster Creek plant and concluded that the following load handling systems are subject to NUEG-0612 :

t e Reactor building crane L

o Recirculation pump monorail spent fuel pool jib cranes.

o l

The Licensee has also identified other load handling devices that have been excluded frcm satisfying the criteria of the general guidelines of NUREC,-0612 due to physical separation from safe shutdown equipment or irradiated fuel; these devices include:

t o Machine shop monorail i

o Turbine building crane o Equipmena bandling monorail (outside CRD rebuild i

room at 75-ft elevation) o rilter and demineralizer monorail g

g e Equipment handling monorail (adjacent to reactor building equipment hatch at 95-f t elevation)

Batch bay crane o

.o CRD rebuild room monorail o Railroad bay monorail o Jib crane (located 23 f t from reactor building equipment batch) o Maintenance building crane o Radwaste building crane.

i A second 1-ton jib crane is located adjacent to the reactor building

)

j equipment hatchway and has been excluded from NUREG-0612 guidelines due to separation from the torus by the railroad bay floor.

The crane is used to lift small equipment, crates, and tools to various elevations in the reactor building.

A conservative analysis shows that a heavy load drop by this crane will not result in perforation or scabbing of this floor to damage the equipment located below it.

Tne intake gantry crane has been excluded from NUREG-0612 applicability due to removal from service.

If at some time in the future this crane is placed back into service, an evaluation will be performed to ensure that NUREG-0612 criteria are satisfied.

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I TIA-C5 50 6-377 1

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h The three refueling platform auxiliary hoists have been derated from 2

their current rating of 1000 lb to 750 lb so that heavy loads cannot be handled by these load handling systems.

This derating would not affect the 1

l lif ts that they were originally intended to service.

i The drywell air lock monorail has been excluded from NUREG-0612 due to l

the fact that it handles the air lock a few inches off the floor and there is no safe shutdown equipment in close proximity to the airlock.

A load drop i

will not affect safe shutdown capability based on ene evaluation of this handling system.

l

b., Evaluation i

The Licensee's conclusions regarding the applicability of general I

guidelines are is consistent with the intent of NUP3G-0612.

4 i

i c.

Conclusion ahd Recon:mendations The Oyster Creek. plant conplies with NUREG-0612 regarding applicability of heavy load overhead handling systems.

I i

2 '. l. 2 Safe Imad Paths- (Guideline 1, NUREG-0612, Section 5.1.1(1))

J

" Safe load paths should be defined for the movement of heavy loads to j

r.inimize 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 members, beams, etc., such that if the load is dropped, the structure is more likely to withstand the impact.

These load paths 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 r'eview committee."

a.

Summarv of Licensee Statements and Conclusions The Licensee has addressed the handling of heavy loads by defining four safety class designations.

Each heavy load is assigned one or more safety classes.

The safe load path / procedural requirements corresponding to the assigned safety class have been added to the appropriate plant operating or

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TF.R-C5 50 6-377 l

f i

l maintenance procedures.

When more than one safety class assignment is made for a particular load, the safe load path / procedural requirements of.au 4

safety class assignments are included in the procedures.

Safety class definitions and their respective handling requirements are listed in Table 1, and loads contained in each safety class are listed in Table 2.

These safety l

classes, by procedure, limit lif't height and time over areas of concern for j

the most critical loads (Safety Class 1), define areas over which loads shall l

not be carried (Safety Class 2), or define safe load paths that follow, trthe l

cxtent practical, structural floor members, using the min 4mtm practical lif t neight (Safety. Class 3).

j

.All loads designated as safety class 3 shall have specified load paths 3

chown on drawings and attached to load handling procedures.

In addition, a signalman win be used to ensure that the load is carried along its designated load path. The signalman with the job supervisor win walk down the designated load path prior to load movement to ensure that there are no obstructicas that could affect the ability of the crane operator to follow the designated path.

1 For the reactor building crane load block, shipping casks, fuel channel crates, and new fuel containers, the Licensee stated that the primary concern 1s the potential for dropping these loads the full length of the equipment hatch located in the :.eu ?.heast quadrant.

For these lifts, the crane will be R.

1 oriented so that the crane boist is directly over the main structural members for the. track bay floor when moving these loads up or down the equipment hatch, in order to assure maximum available resistance to impact in the event i

of a load drop.

In addition, the Licensee added that safe load paths will be defined for movement of shipping casks on the refueling floor prior to their including definition of load paths in specifi,c procedures covering

use, movement to and from the equipment hatch, spent fuel pool, and cask washdown These load paths will be defined by establishing boundaries around the area.

floor area over which the cask may travel, will be shown on a drawing included in the procedure, and will be marked temporarily using tape on the refueling

(

I floor. Within these boundaries, move height win not exceed 6 inches above_

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I m -C5506-377 able 1.

Lead Safety Classes and Safe Load Path Actions l

l Heavv Bandline situations Safe Load Path / Procedural Actions Recuired l

Safety Class 1:

Load must be carried Procedurally limit time and height load l

l directly over spent fuel, the reactor is carried over the area of concern; vessel, or safe shutdown equipment define laydown area, show on drawings (i.e., there are no intervening included in the procedure the prescribed I

l structures such as floors).

laydown area.

Procedures will be reviewed with crane operators andr signalmen prior to lifts over an cpen I

reactor vessel.

i l

Safety Class 2:

Load could be carried-Procedurally limit time and height that directly over spent fuel, the reactor load is carried over area of concern; I

vesse1, or safe shutdown equipment, define laydown area, show on drawings i.e., load can be handled during the attached to procedure the prescribed safe l

time when spent fuel or the reactor load path and laydown area.

j vessel is exposed or safe shutdown l

cquipment is required to be operable and there are no physical means (such as interlocks or hechanical stops) available to restrict load movement I

I over these objects.

l L Safety Class 3:

Load could be carried Define safe load paths that follow, to over spent f uel or safe shutdown the extent practical, structural floor aquipment, but the fuel.or equipment members.

Define laydown areas. Limit is not directly exposed to the load ~

load travel height to the minimum height

drop, i.e.,

intervening structures practical.

Icad paths and laydown areas cuch as floors provide some protec-shown on drawings attached to procedures.

tion.

Safety Class 4:

Load cannot be carried No safe load path or special procedural i

over spent f uel or over saf a shutdown actions required.

I squipment when such equipment is required to be operable, i.e., design or operational limitations prohibit j

movement.

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TER-C550 6-377 f.

Table 2.

Heavy Load Safety Classificatich Safaty Additional Classification Heavy Load Saf etv Clast a.s 1

Drywell head 3

Reactor vessel head 3

Steam dryer 3

Steam separator 3

2 Fuel pool gates Spent fuel casks 3

l Fuel transfer shield Equipment storage pool shield plugs 3

Dryer / separator sling assembly Fuel storage pool shield plugs 3

Bead strongback l

Stud tensioner assembly l

3 Reactor vessel head insulation Plant, equipment N6w fuel and shipping containers l

Cavity shield plugs l

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i TER-C5506-377 O

(or snali obstructions) and movement will follow structural members 21oor j re the extent practical.

l l

With regard to the recirculation pump monorail and the spent fuel pool

! jib cranes, the Licensee stated that safe load paths are limited by the physical capabilities of the equipment.

Operating procedures shall be developed, however, that will caution operators not to carry loads over or in f tho vicinity of spent fuel or safety-related equipment unless absolutely

} ~ncesseary and, if so, to limit the height and duration of the lifts.

t j

Each heavy load lift will be controlled by a designated individual who f ~ will be responsible for enforcing procedural requirements.

Deviations from those procedures and load paths require a revision to procedures or a I Temporary Procedure Change, either of which must be reviewed and approved by t' the Plant Operations Review Committee and the resident manager, r

i 4

j

.b.

Evaluation a

The Licensee's method..of identifying safety classes and differentiating o'rolative safety significance of the identified loads is consistent with j NUEEG-0612 guidelines.

~

As noted by the Licensee for Class 1 and 2 loads, the most direct route l

to the laydown area is most likely to be an acceptable load path.

Other precautions taken by the Licensee (defining laydown areas and incorporating' 1j - drcwings into plant procedures) are adequate to meet the intent of Guideline 1.

Identification of specific loads paths for Class 2 and 3 designatud loads and incorporation of these paths in the controlling load handling procedures j

. maats. the requirements of this guideline. The use of a knowledgeable signniman is a reasonable alternative which provide's the crane operator with i

visun1' aids to ensure that load movement adheres to the established load paths.

In addition, the handling of load path and procedure deviations meets i

tho intent of Guideline 1 because the authority to approve deviations is vented in the plant operations and review committee and the resident manager.

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TIR-C5506-377 Q Conclusion The cyster Creek plant complies with Guideline 1 based on the imple'menta-tien of actions preposed by the Licensee.

2.1.3 Load Bandline Procedures, (Guideline 2, NUREG-0612, Section 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 e

should cover handfing of those loads listed in Table 3-1 of NUREG-0612.

These procedures shculd include:- identification of required equipment; inspections and acceptance criteria required before movement of load; the steps and proper sequence to be followed in handling the load; defining the safe path; and other special precautions."

a.

Sun::narv of Licensee Statements and Conclusions The Licensee han. indicated that the following lif ting procedures are used t the oyster Creek' plant:

205.0 -

Reactor-refueling 701.1.001 - Reactor vessel head removal and replacement 701.1.002 - Reactor vessel steam dryer and steam separator removal and replacement 701.1.003 - Reactor vessel insulation removal and replacement 704.1.002 - Drywell head removal and replacement 756.1.002 - Fuel transfer shield installation and removal t

756.1.003 - Shield plugs removal and replacement 756.1.004 - Fuel pool gates removal and installation.

The Licensee has stated that all lifting procedures have been revised to satisfy the requirements of Section 5.1.l(2) of NUR2G-0612 including:

~

1.

description of the safety concern in the handling of heavy loads

~

with tne reactor building bridge crane 2.

defined safe load paths i

l 3.

precautions 4.

prerequisites 5.

identification of proper handling equipment I

I 6.

training and qualification requirements for crane operators

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TIR-Cf506-377

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verification that required detailed inspections have been performed

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7.

E.

sling selection criteria 9.

required crane inspection by operator prior to load handling i

10.

supervision of work involving a heavy load lif t by a designated job l

supervisor 11.

critical steps in order to perform the lif t.

In addition, the Licensee has indicated that new procedures are being dsveloped for the following load handling devices:

l o reactor building bridge crane l

recirculation pump monorail and hoist f

o spent fuel pool jib cranes a

spent fuel cask operation will be governed by a new procedure each o

time with special lif ting requirements applicable to that particular j

i cask.

b.

Evaluation l'

.he implementation-of procedural controls on load handling at the oyster Oreek plant meets the intent of Guideline 2 of NUREG-0612 based on the Licensee's description of oyster Creek plant lifting procedures.

l c.

Cenelusion

• he Oyster Creek plant complies with Guideline 2 of NUREG-0612.

2.1. 4 Crane Coerator Trainino (Guideline 3, NUREG-0612, Section 5.1.l(311 "Orane operarers should be trained, qualified, and conduct themselves in accordance with Chapter 2-3 of ANSI B30.2-lS76, ' Overhead and Gantry I

Cranes' [7). "

cf Licensee Statements and Conclusions a.

S u==arv l

(

The Licensee has stated that the curr.ent practices for qualification and l

craining of crane operators essentially cover the provisions of ANSI

(

) E30.2-1976, Chapter 2-3.

However, these practices are not currently in the form of an approved procedure. Portions of the training are performed by the 14

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i I

TIR-05506-377

.aintenance supervisc: and other portiens are performed by the plan training staff.

A new procedure with qualification records has been developed and i=plemented in order to formalize the program for crane operator qualification for the reactor building and spent fuel pool jib cranes.

The new procedure i

requires that operators be familiar with appropriate handling system operating procedures and pass a practical' operating examination with the handling system.

The Licensee has taken exception to ANSI B30.2-1976 with respect to e section 2-3.1.7, " conduct of operators, Part F."

The standard requires that "bef ore leaving he crane unattended, the operator shall land any attached load, place the controllers in the 'off' position, and open the main line device of the specific crane." However, during reactor disassembly at the oyster Creek plant, it is necessary to keep the steam separator covered with water during handling to maintain exposure levels as low as practicable.

Consequently, the separator is raised incrementally, and then lef t suspended until the water 2.evel rises sufficiently to allow additional raising of the separator.

The separator may stay suspended at one level as long as 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> while flooding is proceeding.

During these periods when the separator is left suspended, the crane operator may leave the cab until recalled.

However, prior to leaving the crane, the operator places the controller in.the."off" position and opens the main line device.

b.

Eialuation Crane operator training at the oyster Creek plant is considered acceptable based on the Licensee's verification that the program meets the provisions of ANSI E3 0.2-1976 and that a new procedure has been developed to formalize the The Licensee's exception to Chapter 2-3, Section 2.-3.1.7 concerning program.

leaving the crane unatitended 'while loaded is reasonable based upon the specified manner in which the crane is secured. However, it should be noted that this practice appears to be in violation of Title 29 CFR 1910.179. (N).

(3). (X) (OSHA) and thus should be evaluated by the Licensee unless such deviation has been previously approved..

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TIE-0550 6-3 7 7 e.

1usien and Rece:mendation The Oyster Creek plant co= plies with Guideline 3 of NUREG-0612 concerning crane operator training.

-2.1.5 soeeial Liftine Devices (Guideline 4, NURIG-0612, Section 5.1. l (4 )]

"Special lif ting devices should satisfy the guidelines of ANSI

- N14. 6-197 8, ' Standard for Special Lif ting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials' (8). ~

This standard should apply to all special lif ting devices which carry" beavy loads in areas as defined above.

For operating plants, certain inspections and load tests may be accepted in lieu of certain material requirements in the standard. In addition, the stress design f actor stated in Section 3.2.1.1 of ANSI N14.6 should be based on the combined

, naximum static and dynamic loads that could be imparted on the handling device based on characteristics of the crane which will be used.

This is stress design f actor en only the weight (static load) of the load and of the intervening cocponents of the special handling device [NUREG-0612, Guideline 5.1.1(4)). "

)

a.

Su=,ary of Licensee statements and conclusions The Licensee has indicated that there are six handling devices made up for special applications and currently used in handling heavy loads:

1.

dryer / separator sling 2.

head strongback 3.

cask yokes and slings 4.

fuel transfer shield slings 5.

cavity shield plug lifting beam 6.

equipment storage pool plug lifting beam.

The comparison of these special lif ting devices to ANSI N14.6-1978 was limited to Sections 3.2 and 5 of the standard.

The Licensee's review indicated the following exceptions to ANSI N14.6-1978:

1.

Sections 3.1 (Designer 's Responsibilities), 3.3 (Design Considera-tions), 4.1 (Fabricator 's Responsibilities), 4.2 (Inspector 's l

Responsibilities), and 4.3_ (Fabricator 's Considerations) are difficult to apply in retrospect. However, information on drawings indicates that sound engineering practices were placed on the O

fabricator and the inspector for the purpose of ensuring that the designer's intent was accomplished. i A

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2.

Sections 1.0 (Scope), 2.0 (Definitiens), 3.4 (Design Considerations i

to Minimize Decontamination Effects in Special Lif ting Device Use),

4 3.5 (Castings), and 3.6 (Lubricants) are not pertinent to load l

handling reliability.

j l

l 3.

Section 6, Special Lif ting Devices for Critical Loads, is not j

applicable at the oyster Creek plant because none of the loads lifted by these devices has been identified to be a critical load.

k 4.

Plant procedures do not specify a visual inspection by maintenance or l

other non-operating personnel at intervals of 3 months or less asr

}

required by Jection 5.3.7 of ANSI N14.6-1978.

Procedures have,been i

revised so that these devices are inspected by a qualified personnel 1

prior to each usage and so that a thorough testing and nondestructive i

examination is performed prior to each refueling.

Based on the i

controlled storage between periods of usage, dedicated single usage, and complete inspection schedule, the equivalency of Section 5.3.7 is i

demonstrated.

I j

5.

Section 5.3.3 of ANSI N14.6-1978 requires that special lifting 4

devices be load tested according to Section 5.2.1 to 150% of maximum load following any incident in which any load-bearing component may have beeh subjected to stresses substantially in excess of those for which it was qualified by previous testing, or following an incident O

that may have caused permanent distortion of load-bearing parts.

Since distortion may already have occurred or since defects may have already developed due to the overstressed condition, it seems more prudent and practical to perform the dimensional examinations for deformation and the NDE for defects to determine whether the device is still acceptable for use rather than subject the device to 150%

load testing.

If defects or deformation are detected, then the device shall be repaired or modified and tested.to 150% load followed by examination for defects or deformation.

During the Licensee's review of special lifting devices against Sections 3.2 and 5 of ANSI N14.6-1978, the following results were obtained:

1.

The drver/senarator sling design exceeds the criteria in ANSI E30.9 and ANSI N14.6.

The lifting device has been load tested to a weight well in excess of 150% of the rated load.

In addition, a preventive maintenance procedure has been developed for inspection of this lifting device in accordance with ANSI 130.9 and ANSI N14.6.

2Property "ANSI code" (as page type) with input value "ANSI N14.6.</br></br>2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..

The head stroneback drawings are available showing dimens'ional and material requirements and types,of welds to be used for each weldment.

However, information on stress analyses that may have been performed, design safety factors used, load tests performed, or O

processes and standards used in fabrication were not available.

Accordingly, the Licensee performed a stress analysis and design' 17-e.....

I 1-1 3

TIR-C5506-377 i

evalut. tion to demonstrate the adequacy of the design.

As a result of this evaluation, the head strongback was found not in full cenpliance

[

with ANSI specified f actors of safety against bending in th'e lif ting arms although stresses were within AISC allowables.

Modifications j

are being made to the lif ting arms to bring the head strongback into compliance with ANSI N14.6.

Following these modifications, the I

device will be load tested in accordance with Section 5.3.2 of ANSI i

N14. 6-19 7 8.

In addition, a preventive maintenance procedure including visual and NDE examination and inspections prior to each refueling has been developed to comply with ANSI N14.6 criteria. '

f 3.

Ter casks (including NAC-1) having unique special lifing devices or 4

yokes, the lifting devices are the property of the cask owner.

I

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Accordingly, procedures have been revised to require that a l

certification be obtained from the cask owner, prior to handling the

[

cask on-site, that verifies the cask lif ting device or yoke design

]

satisfies the criteria of ANSI N14.6, Section 3.2, and that the l

device has been inspected and maintained in accordance with ANSI N14.6, Section 5.0.

l 4.

The fuel transfer shield sline is used for the shield and the GE200 i.

cask.

The design of the sling assembly was compared to ANSI B30.9 and found to exceed the criteria in this standard.

In addition, a l

.new preventive maintenance procedure that complies with ANSI B30.9 f

criteria. requires inspections of the slings prior to each refueling.

5.

The cavity shield olue and ecuionent storace cool clue lifting beams have insufficient documentation to evaluate the beams against the i

criteria of ANSI N1.4.6.

Therefore, the Licensee performed a stress l

analysis and design evaluation of these lif ting beams.

As a result j

of this evaulation, these beams were found not to comply with ANSI N14.6 for f actors of safety against bending.

These beams are being modified to bring them in compliance with ANSI N14.6.

I Following these modifications, the devices will be load tested in I

accordance with Section 5.3.2 of ANSI N14.6-1978.

A preventive maintenance program that includes examination and inspection to l

satisfy ANSI N14.6 has been developed.

i A new lif ting device for the core spray sparger will be evaluated against t

i the design criteria of ANSI N14.6 when the design of the sparger and strongback 4

l are finalized.

b.

Ivaluation i

The Oyster Creek plant satisfies the criteria of ANSI N14.6-1978 Section 2.2 (Design Criteria) for the dryer / separator sling and the fuel transfer shield i

sling based upon verification by the Licensee that the design meets or exceeds 3

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i Tn-C5506-I77 tne criteria in ANSI N14.6 and/cr ANSI E30.9.

The head strongback, cavity-shield plug lif ting beam, and the equipment storage pool plug lifting beam will cc= ply after the proposed modifications and load tests have been completed.

j The Licensee's response that Subsections 3.4, 3.5, and 3.6, Section 4, and Section 6 of ANSI N14.6-lS78 are not applicable or pertinent is consistent with the desired intent of this guideline.

The Licensee's response that design evaluations have been performed for all lif ting devices and, where not 4

in compliance, will be modified to satisfy criteria of ANSI N14.6-lS71 is also consistent with the requirements of this guideline.

The preventive maintenance program that includes inspection by qualified perronnel and nondestructive examination prior to use appears to address the need for continuing compliance testing set forth in Section 5 of ANSI N14.6.

The Licensee's decision to require visual inspection by nonoperating or maintenance personnel prior to each use is in keeping with ANSI N14.6-1978 requir ements.

In addition, lead tests to be performed for the head strongback i

and lifting beams for the cavity shield plug and the equipment storage pool plug satisfy the guideline requirements, as does the Licensee requirement that cask owners ccmply with Section 5.0 of ANSI N14.6-lS78.

No load test is needed for the fuel transfer shield sling since it is only subject to compliance with ANSI E30.9-1971.

c.

Conclusion and Recommendations The Oyster Creek plant complies with Guideline 4.

2.1.6 Lif tine Devices (Not Soecially Desiened) (Guideline 5, NURIG-0612 Section 5.1.l(5))

" Lifting devices that are not specially designed should be installed and used in accc: dance with the guidelines of ANSI B30.9-lS717 ' Slings'

[9).

However, 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 icad.

Where 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."

Om r....

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CIA--C 5 5 0 6-3 7 7 a.

Su=.ary of Licensee Statenents anc Conclusions The Licensee has stated that, to ensure that appropriate slings are selected for use in handling miscellaneous loads and ti.at slings are properly maintained, the following program changes have been made:

1.

Load handling procedures require the use of ANSI B30.9 criteria for sling selection and rigging techniques.

2.

A new preventive maintenance procedures has been developed for ennual inspection of slings.

3.

Load handling procedures require a visual inspection of slings for damage prior to making a lift.

4.

A tagging procedure has been developed for slings to P atify sling rating, application, last examination, and expiration date of examination.

Based on an analysis performed, dynamic loading on slings associated with the reactor buildng -crane were found to be approximately 3% of the static load.

This 34 increase in loading is insignificant and may be disregarded.

b.

,Evalua tion Sling installation and usage at the oyster Creek plant complies with NUREG-0612, Section 5.1.1(5).

On the basis of information provided by the Licensee, dynamic loads are a reasonably small percentage of the. overall static lead and may be disregarded in rating the slings.

c.

Conclusion The Oyster Creek plant complies with Guideline 5 of NUREG-0612.

2.1.7 -Cranes (Insoection, Testine, and Maintenance) (Guideline 6, NUREG-0612, Section 5.1.1(6))

"The crane should be inspected, tested and maintained in accordance with' l

Chapter 2-2 of ANSI E30.2-1976, ' Overhead and Gantry Cranes,' with the exception that tests and inspections should be performed prior to.use when it'is not practical to meet the frequencies of ANSI B30.2 for periodic inspection and test, or where frequency of crane use is less 5

than the specified inspection and test frequency (e.g., the polar crane

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..._ _ _ _. _ _ = _ _ _._ ___.- _ _ _ __... _ _ _ _ _ _

TIR-01506-377 a

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v inside a PWR containment ray only me used every 12 to 18 months during 4

refueling operations and is generally not accessible during power operatien.

ANSI B30.2, however, calls for certain inspections to be l

performed daily or monthly.

For such cranes having limited usage, th e inspections, tests, and maintenance should be performed prior to their j

use)."

2 a.

Sumarv of Licensee Statements and Conclusions i

{

The Licensee has stated that new procedures for inspection, testing and y

maintenance of the recirculation pump monorail, spent fuel pool jib crene, and l

reacter building crane are being developed.

In addition, provisions have been i_

included in the new crane operation procedures, to include appropriate operator inspections prior to load movement.

With these revisions and additions, the a

f procedures will satisfy the criteria in ANSI B30.2-1976, Chapter 2-2 without exception.

b.

Evaluation a Upon i=plementation, the oyster Creek plant inspection procedures will be consistent with section 5.1.l(6) of NUREG-0612.

c. ' conclusion The oyster Creek plant complies with Guideline 6 of NUREG-0612.

2.1.8 Crane Desien [ Guideline 7, NUREG-0612,~ Section 5.1.1(7)1 "The 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 CMAA-70 [10), ' Specifications for Electric overhead Travelling Cranes.'

An alternative to a specification in ANSI B30.2 or CMAA-70 may be a,ccepted in lieu of specific compliance if the intent of i

the specification is satisfied."

a.

Sumarv of Licensee Statements and Conclusions The Licensee has stated that the reactor building crane was designed and f abricated by Whiting Corporation to the specifications in EOCI-61 [10),

" Specifications for Electric overhead Traveling Cranes-1961" and in accordance -

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.,, C + 3 = L 4 - r e e..; = C * *tt *'

TIA-C5506-377 with additional requirements specified by the architect-engineer.

The Licensee performed a review of the original specifications versus CMAA-70 (1975) and ANSI 330.2-1976.

The results of this detailed point-by-point

]

comparison were submitted in Reference 5.

b.

Evaluation e

The reactor building crane at the oyster Creek plant substantially, j

ecmplies with the criteria specified in Guideline 7 because the original procurement specification was based on ECCI-61.

In addition, for those t

l criteria in CMAA-70 noted to be more restrictive than requirements of ECCI-61, t

the Licensee has demonstrated compliance with CHAA-70 or provided reasonable I

assurance that the existing design meets the intent of the CMAA criteria.

i j

c.

Conclusion

{

The oysted Creek plant complies with Guideline 7.

j 2.2 INTERIM PROTECTION MEASURES i

The NRC has established six interim protection measures to be implemented 1

l at operating nuclear power plants to provide reasonable assurance that no heavy.

f loads will be handled over the spent fuel pool and that measures exist to i

reduce the potential for accidental load drops to impact on fuel in the core j

or spent fuel pool.

Four of the six interim measures of the report consist of Guideline 1, Safe Load Paths; Guideline 2, Load Handling Procedures; Guideline 3, Crane operator Trainings and Guideline 6, Cranes (Inspection, Testing, and j

Mainte nanc e).

The two remaining interim measures cover the following criteria:

l 1.

Heavy load technical specifications 2.

Special review for heavy loads handled over the core.

i Licensee implementation and evaluation of these interim protection

]

measures are contained in the succeeding paragraphs of this section.

1 O

i l

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n.....

_ ~ _ _ _. _ _.. _.. _ _. _. _. _ _ _. -

i i

TER-05506-377 2.2.1 Tecnnical Soecificaciens (Interi= Protection Measure 1, gcgrc.-0612.

g Section 5.3 (11) i

" Licenses for all operating reactors not having a single-f ailure-proof overhead crane in the fuel storage pool area should be revised to include a specification comparable to Standard Technical Specification 3.9.7,

' Crane Travel - Spent Fuel Storage Building,' for PWR's and Standard 3

Technical Specification 3.9.6.2, ' Crane Travel,' for BWR's, to prohibit handling of heavy loads 'over fuel in the storage pool until implementa-tion of measures which satisfy the guidelines of Section 5.1 [of l

NUREG-0612)."

1 l

a.

Summarv of Licensee Statements and Conclusions A review of the Oyster Creek Technical Specifications indicates that 1

i Section 5.3.l(d) prohibits the movement of loads greater than the weight of one 1

l fuel assembly over irradiated fuel in the fuel pool.

i l

l b.

Evaluation and Conclusions 1

The oyster Creek plant complies with Interim Protection Measure 1.

([])

i 2.2.2 Administrative controls (Interim Protection Measures 2, 3, 4, and 5, NUREG-0612, Sectionc 5.3 (2)-5.3 (5))

4

" Procedural or administrative measures [inchaing safe load paths, load l

handling procedures, crane operator trainirg, and crane inspection)...

can be accomplished in a short time periot and need not be delayed for i

completion _ of evaluations and modificatier.s to satisfy the guidelines of Section 5.1 [of NUREG-0612)."

l l

i Summarv of Licensee Statements and Conclusions a.

j Summaries of Licensee statements and conclusions are contained in 1

l

' discussions of the respective general guidelines in Sections 2.1. 2, 2.1. 3,

2.1.4, and 2.1.7.

b b.

Evaluations, Conclusione and Recommend 2tions b

Evaluations, 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.

I 4

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.... FrankJin Research Center Ao a.ew.- m.

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5 d

TIR-C5506-377 j

2.2.3 Soecial Review for Heavv Loads Bandled Over the Core [Interi:. Protectien Measure 6, NURIG-0612, Section 5.3 (6))

j

"...special attention should be given to procedures, equipment, and

{

personnel for the handling of heavy loads over the core, such as vessel internals or vessel inspection tools.

This special review should include the following for these loads:

(1) review of procedures for installation l

of rigging or lifting devices and movement of the load to assure that sufficient detail is provided and that instructions are clear and concise: (2) visual inspections of load bearing components of cranes, clings, and special lifting devices to identify flaws or deficiencies.

i i that could lead to f ailure of the component; (3) appropriate repa"u and

,I replacement of defective components; and (4) verify that the crane operators have been properly trained and are familiar with specific l

procedures used in handling these loads, e.g., hand signals, conduct of 4

operation, and content.of procedures."

i a.

Summarv cf Licensee Statements and Conclusions With regard to the implementation of interim actions, the Licensee has j

stated that the required changes to procedures have been developed and are 1

j currently being reviewed and approved.

Full implementation of the approved procedures will be. effected prior to the next refueling outage.

i j

b.

Evaluation The Licensee has adequately addressed the requirement for a review of all load handling procedures.

In light of responses to Guidelines 2 and 3, it is apparent that procedures for handling loads over the core and ^ operator training have been reviewed and upgraded as appropriate.

In addition, design of cranes at the oyster C:eek plant and programs for selection and use of slings have been reviewed and found to comply with NUREG-0612.

c.

Conclusion The oyster Creek plant complies with Interim Protection Heasure 6 based upon Licensee-provided information.

M _N

-24 I

2...' Frankjin Research Center a cm e n. %.a.-

.___._.._.~_

TEP.-05 5 0 6-3 7 7 i

l 3.

CONO:,USION 0.is summary is provided to consolidate the results of the evaluation contained in Section 2 concerning individual NE staff guidelines into an overall evaluation of heavy load handling at oyster Creek Nuclear Power l

Plant.

overall conclusions and recc=nended Licensee actions, where L

apprcpriate, are provided with respect to both general provisions for load handling (NUREG-0612, Section 5.1.1) and completion of the staff recc=:endations for interim protection (NUREG-0612, Section 5.3 ).

l l

3.1 GIhT.?AL PROVISIONS FOR LOAD BANDLING rhe Nr staff has established seven guidelines concerning provisions for handling heavy loads in the area of the reactor vessel, near stored spent fuel, or in other areas where an accidental load drop could damage equipment requirec fer safe shutdown or decay heat removal.

The intent of these I

guidelines is wofold.

A plant conforming to these guidelines will have develeped and implemented, through procedures and operator training, safe load l

l travel paths such that, to the maximum extent practical, heavy loads are not l

carr:ed over or near irradiated fuel or safe shutdown equipment. A plant conforming to these guidelines will also have provided sufficient operator training, handling system design, load handling instructions, and equipment inspe=tien to ensure reliable operation of the handling system.

As detailed i

in Section 2, it has been found that load handling operations at oyster Creek l

Nuclear Power plant can be expected to be conducted in a highly reliable r.anner consistent with the staff's objectives as expressed in these guidelines.

' 3.2 hT R.IM PRCTECTION MEASURES The h7C staf f has established certain measdres (NUREG-0612, Section 5.3)

~

that snocid be initiated to provide reasonable assurance that handling of heasy loads will be performed in a safe manner until final implementation of 4

the general guidelines of NUREG-0612, Section 5.1 is complete.

Specified measures include the implementation of a technical specification to prohibit

.he handling of heavy loads over fuel in the storage pool; compliance with ds

-2 5-

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.... Fran ;.:n Research Center

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m.mn.

J

':IE-C5 506-3 77

~

Guidelines 1, 2, 3, and 6 of NUREG-0612, Section 5.1.1; a review of load handling procedures and operator training; and a visual inspection, program, i

including component repair or replacement as necessary of cranes, slings, and special lif ting devices to eliminate deficiencies that could lead to component failure.

The evaluation of information provided by the Licensee indicates 4

that the Oyster Creek plant complies with the staff's measures for interim protection.

'l i

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-2 6-2.i ranklin Research' Center F

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"EF-05506-377 O

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1.

" Control of Heavy Loads at Nuclear Power Plants" NIC, July 1980 NUREG-0612 2.

V. Stello, Jr. (NRC)

Letter to all Licensees Sub3ect:

Request for Additional Information on Control of Heavy Loads Near Spent Fuel May 17, 1978 3.

NRC i

Letter to Jersey Central Power and Light Corpany (JCPEL)

Subject:

Request for Review of Heavy Load Bandling at Oyster Creek Nuclear Power Station December 22, 1980 4.

J. T. Carroll (JCP&L)

{

Letter to D. G. Eisenhut (NBC)

Subject:

Control of Heavy Loads September 22, 1981 5.

P. B. Fiedler (GPU)

O Letter to D. G. Eisenhut (NRC) i

Subject:

Control of Beavy Loads February 18, 1983 6.

Telephone conference call involving J. Lombardo and A. Singh (NBC), C.

Bomberger (WSI/TRC), and GP6 representatives l

May 27,1983

(

7.

American National Standards Institute

" Overhead and Gantry Cranes

• ANSI B30.2-1976 8.

American National Standards Institute

" Standard for Special Lifting Deivces for Shipping Containers Weighing i

10,000 Pounds (4500 kg) or More for Nuclear Materials" l

ANSI N14.6-1978 9.

American National Standards Institute

" Slings" ANSI B30.9-1971 10.

Crane Manufacturers Association of America,1975

" Specifications for Electric overhead Traveling Cranes" CMAA-70 t

A ;.d. Franklin lesearch Center a o a.# n n.a

Exhibit D l

{

df ue.,,%,

umTEp STATE i

8 NUCLEAR REGULATORY COMMISSION WASHINGTON. o. C. 20665 i

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WN 2 61985 i

i TO ALL LICENSEES FOR OPERATING REACTOR $

Gentlemen:

i SUSJECT:

COMPLETION OF fHASE II 0F " CONTROL 0F HEAVY LOADS AT NUCLEAR i

POWER PLANT 5" N1MtEG-0612. (GENERIC LETTER 85-11 )

i On December 22, 1980, NRC issued a generic letter (unnumbered) which was j

supplemented February 3, 1931 (Generic Letter 81-07) regarding NUREG-0612 j

" Control of Heavy Loads at Wuclear Power Plants".. This letter requested l

that you implement certain interie actions and provide the NRC infomation l

related to heavy loeds at your facilities. Your submittals were requested In two parts; a six'Jonth response (Phase !) and a nine month response l

(Phase II).

)

All licensees have completed the requirement to perform a review and submit a Phase I and a Phase !! report. Based on the improvements in heavy loads l

handling obtai m d from implementation of NUREG-0612 (Phase !), further action is not required to reduce the risks associated with the handling of heavy j

loads (See enclosed NUREG-0612 Phase II). Therefore, a detailed Phase !!

review of heavy loads is not necessary and Phase II is considered completed.

i However, while not a requirement, we encourage the implementation of any actions you identified in Phase II regarding the handling of heavy loads j

that you consider appropriate.

For each plant which has a license condition requiring cosettaents acceptable to the NRC regarding Phase II, an application for license amendment may be subeltted to the NRC to delete the license condition citing l

this letter as the basis.

If you have any questions, contact your Project Manager or Don Neighbors (301) 492-4837.

l Sincerely.

. Thomp ector 01 ton of Licensing I

i

Enclosure:

i As Stated 1

$506@$

i 8506270216 i

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LO NUREG-0612. " CONTROL 0F HEAVY LOADS AT f

NUCLEAR POWER ptANT5" RESOLUTION OF PHASE !!

t 1

i Generic Technical Activity A-36 was established to systematically examine the i

l staff's licensing criteria, adequacy of measures in effect at operating plants and recommend necessary changes to assure the safe handling of heavy loads.

j The task involved review of licensee information, evaluation of historical

[

data, performance of accident analyses and criticality calculations.

l-development of guidelines for operating plants, and review of licensing criteria. The review indicated that the major causes of load handling l

accidents include operator errors, rigging failures, lack of adequate inspec-tion and inadequate procedures. The results of the review culminated in the l

1ssuance of NUREG-0612. " Control of Heavy Loads at Nuclear Power plants" in j-July 1980 NUREG-0612 described a resolution of Task A-36.

NUREG-0612 presents an overall philosophy that provides a defense-in-depth approach for controlling the handling of heavy loads. The approach is directed l

to preventing load drops. The following sumerizes this defense-in-depth l

spgtoach:

j

1. Assure that there is a well designed handling system.

I i

2.

provide sufficient operator training. Ioad handling instructions, and equipment inspection to assure reliable operation of the handling system.

I 3.

Define safe load travel paths and procedures and operator training to l

assure to the entent practical that heavy loads are not carried over or j

near irradiated fuel or safe shutdown equipment.

j 4.

provide mechanical stops or electrical interlocks to prevent movement of heavy loads over irradiated fuel or in proximity to equipment associated with redundant shutdown paths.

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5.

Where mechanical stops or electrical interlocks cannot be provided.

(

provide a single-failure-proof crane or perform load drop analyses j

to demonstrate that unacceptable consequences will not result.

i l

By Generic Letters dated December 22. Ig80, and February 3. Ig81 (Generic i

Letter 81-07), all utilities were requested to evaluate their plants against-the guidance of NUAEG-0612 and to provide their submittals in two parts; Phases j

I(sixmonthrespense)andPhase!!(ninemonthresponse).PhaseIresponses were to address Section 5.1.1 of NUREG-0612 which covers the following areas:

4 1.

Definition of safe load paths l

2.

Development of load handling procedures j

3.

Periodic inspection and testing of cranes j

4.

Qualifications, training and specified conduct of operators-l 5.

Special lifting devices should satisfy the guidelines of AN51 N14.6 6.

6.

Lifting devices that are not specially designed should be installed and l

used in accordance with the guidelines of AN51 830.g

)'

7.

Design of cranes to AN51 830.2 or CMAA-70 Phase !! responses were to address Sections 5.1.2 thru 5.1.6 of NUREG-0612 1

which cover the need for electrical interlocks / mechanical stops, or f

alternatively, single-failure-proof cranes or load drop analyses in the spent l

fuel pool area (PWR). contairment building (PWR). reactor building (BWR) other l

areas and the specific guidelines for single-failure-proof' hand 1tng systems.

i' We have completed our review of the utilities' submittals for Phase I for nearly all operating reactors. Only one plant still remains.o be reviewed.

o l

During our review we verified that the seven guidelines listed above were i

providing the desired level of safety indicated in NUREG-Ca12. By way of l

the utilities' responses to the criteria of NUREG-0612. Section 5.1.1 i

and through discussions with our consultants based on their experiences from the reviews, we have concluded that the Phase I guidelines have provided i

an increased awareness by the utilities of the importance of heavy load i

handling.'

Our review has indicated that satisfaction of the Phase I guidelines assures i

that the potential for a load drop is extremely small. We have notep

!O i

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l j

improvements in heavy load handling procedures and training and crane and j

handling tool inspection and testing. These changes have been geared to I.

limiting the handling of heavy loads over safety-related equipment and spent i

fuel to the extent practical, but where this can not be avoidedpo j

accomplishing it with the operational and other features of the1frocram imolemented in Phase I.

We therefore conclude that the guidelines of Phase I are adequately j

providing the inten'ded level of protection against load drop accidents.

To date we have received Phase !! submittals from all licensees. We i

interpret Phase II of NUREG-0612 as an enhancement to Phase !.

Thus, prior j

to undertaking a review of the utilities' Phase !! respotse for all of the l

operating reactors, and as a test of the adequacy of the Phase I program, j

we drecided to undertake a pilot program with a limited number of plants.

l The findings from the pilot program would then provide a basis for a j

decision on whether to proceed with the review of the Phase !! submittals for all operating reactors, to reduce the scope of the review, or to l

totally eliminate the review.

l The pilot program involved the review of operating reactors at 12 sites, a

[

totalof20 reactors (eightBWRsand12PWRs). Of the 20 reactors. 5 BWRs j

(Browns Ferry 1. 2 and 3 and Peach Botton 2 and 3) have single-failure-proof l

cranes for all heavy load lifts. " Single-failury-proof

• is used to mean a l

crane which meets the guidelines of N'JREG-0554. " Single-Failure-Proof Cranes l

for Nuclear Power Plants." Three BWR units (Dresden 2 and 3 and Big Rock j

Point) have taken credit for a combination of single-failure-proof cranes in l

some plant areas and load drop analyses in others. Five PWR reactors (Millstone 2. Prairie Island 1 and 2. and Surry 1 and 2) have utilized the load drop analysis approach. One plant (Kewaunee) has taken credit for a cambination of electrical interlocks in some plant areas and load drop analyses l

in others. The remaining sin reactors (Davis Besse ladian Point 2. Arkansas 1 l

and 2 and Calvert Cliffs 1 and 2) chose to take credit for a combination of administrative controls, procedures and Technical Specification restrictions in conjunction with some typeof load drop analysis. This approach does not meet I

the criteria of Sections 5.1.2 to 5.1.6 of NUREG-0612. Rather it is an

{

duplification of the quidelines of the Phase ! effort, reflecting Section 5.1.1

]

of NUREG-0612.

4iO i

O lt should also be noted that we have completed our review of phase !! for five operating license applicants. Of these, two (WMP-2 ar.d Ferai-2) have O

iai-<ii 4i'**-(cii

'<c Catawba 1 and 2) employ a combination of electrical interlocks, aghanical stops. limit switches and load drop analyses.

'..g In addition to the detailed reviews of the Phase !! reports in the pilot program and in connection with the five operating liceese applications, we have performed a sufficient review of all other Phase II reports to flag any outstanding plant-spedific concerns reported.

From our pilot program and OL Phase !! reviews, together with the above-mentioned reviews of the other Phase !! repoKs,we have concluded that the risks associated with damage to safe shutdown systems are relatively small because:

1.

nearly all load paths avoid this equipment 2.

most equipment is protected by an intervening floor 3.

of the general independence between crane failure probability and safety-related systans which has been observed 4.

redundancy of components We did not identify any outstanding plant specific safety concern associated with heavy loads handling.

Therefore, most of the risk appears to be associated with carrying heavy loads over or in a location where spent fuel could be damaged. The single most important example of this concerns loads handled over the open reactor vessel l

during refueling (such as the reactor vessel head). Mauever, as previously mentioned, this is limited to the extent practical and where necessary, is i

performed with a specifically implemented program in conformance with the phase I

I guidelines.

From the pilot program and OL reviews we noted that nine of the twenty j

reactors, all PWRs. do not have single-failure-proof cranes. To date, we have mot identified any WRs with single failurSproof cranes. Further since p

i j

electrical interlocks and mechanical stops are not possible for PWR polar

.m_.

l b

cranes, these reacters would be required to perfers costly detailed load f

drop analyses.

If satisfactory results could not be demonstrated from these analyses. NUREG-0612 would call for installation of a single-failure-proof crane.

4 l

Based on the above, since a single failure proof crane becomes 4 e only solution for satisfying the NUREG-0612 criteria. the cost / benefit should be examined. Because we are dealing primarily with pWRs. the

)

cost for modification of a polar crane to meet single failure criteria l

(NUREG-0554) guidelines) is approximately $30 million. This ine.ludes.

as the dominant cost element, the cost of the extended shutdown which is required in order to gain access to containment. On the benefit side.

given the improvements obtained from the Phase I implementation End the infomation obtained in the course of the pilot program and OL Phase 11 i

)

reviews, we cannot perceive a significant enough benefit in conversion j

to single-failure-proof polar cranes to warrant the high costs. (See j

Attachment I for a cost-benefit analysis.) We believe that the cost /

l benefit analysis in MtfREG-0612 is ao longer valid because of the l

benefits realized by Phase ! implement tion.

f We believe the above assessment is further borne out by the industry j

experience with handling of heavy loads over the years. precautions l

have been and are being taken such that no heavy load drop accidents I

affecting any fustures of the defense-in-depth against severe core-damage accidents have occurred.* This determination is also supported l

by the recomendation of our contractor for the pilot program reviews l

(Franklin Research Center) and our benefit-cost analysis suggesting i

that we accept other, less strigent but less costly seans for phase !!

j compliance as an alternative to the criteria of NUREG-0612 with respect

{

to conversion to single-failure-proof crancs.

)

Conclusion and Recomendation l

Based on the above, we believe the phase ! implementation has provided

{

sufficient protection such that the risk associated with potential heavy load b

i

• Thers have, however, been recent occurrences of lesser severity. (Seefor l

example. It information Notice No. 85-12: Recent Fuel Handling Events LER St and LER 34-006 SanOnofreI. Polar 84-015. Fort Calhoun 1. Lead Over the

_ O Crane Malfunctien). Accordinaly4mphasi of continued attention to the noth1 inthisseterminalionshouldbe regarded as a basis for afy te.

4 safe handlinq_pf_heny_lpa_ds.

i i

j

i 6

)

i drops is acceptably small. We further conclude that the objective identified i

in Section 5.1 of NUREG-0612 for providing " maximum practical defense in depth" j

is satisfied by the Phase I compliance, ed that the Phase !! analyses did not j

indicate %t need to require further generic action at this time.1 j

This conclusion has been confirmed by the results obtained fren'tNe Phase !!

oflot program and additional Phase !! reviews, which identified no residual 2

heavy loads handling concerns of sufficient significance to demand further j

generic action. All plants have examined their load handling practices against f

the reconnendations of Phase II and submitted the Phase 11 report. In this j

way, the utilities were required to identify any unexpected problems to the j

staff.

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f ATTACHMENT I LO

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SUMARY OF COST-SENEFIT ANALYSIS OF PWR POLAR CRANE CONVERSION T0 $1NGLE-FAILURE-PR00F FEATURES SCOPE The safety benefit of converting the polar crane in the containment of an operating or compieted or nearly completed PWR to single-failure proof features and the cost of the conversion were estimated and compared.

The safety benefit was estimated in terms of the resulting reduction in the risk of a severe accident, involving major radioactive material release, during the remaining plant life. The risk was expressed as the product of the accident probability and the population radiation dose from the release, should the accident occur.

The cost estimate included the cost of shutdown (or extension of a non-operating period) needed to accomplish the conversion.

i ACCIDENT FREQUENCY ESTIMATES Crane Failure Frecuency There were 32 crane LER events in the approximately 400 reactor years of U.S. power-reactor operation in the 10 year period July 1969 to July 1979 (NUREG-0612, p. 4-6).

None resulted in radioactive release. Of the 32 events,17 (i.e., just over half) were apparently due to hardware design or fabrication causes, the other 15 to human factors.

(Navy crane statistics, cited in NUREG-0612, for 40 load-drop or potential load-drop events in 1974-77 show 80% of the events to be due to human factors.)

O

.g.

O lt may be assumed, as a rough approximation, that Phase 1 of NRC's heavy-loads generic program is addressed to all the human factors causes and one-half of the hardware causes and succe?ds an reducing the affected part of the f ailure frequency to a quite small fraction of the frequency originally present.

Since human factors and hardware each contribute about one-half of the failures, approximately 3/4 of the total crane failures can l

be expected to be eliminated by the Phase I program.

Single-failure proof (SFP) cranes should substantially reduce the remaining 1/4 of the failure frequency, though those failures would not be eliminated altogether, since the SFP feature (as defined in NUREG-0554) does not protect against all types of possible failure (e.g., the bridge is not SFP and the SFP feature itself is subject to defeat by some types of human error). On the other hand, the SFP feature would make the cranes more " forgiving" of imperfections in the Phate I implementation.

Accordingly, one may reasonably assume that the SFP feature would have a net effect of eliminating 1/4 of the pre-Phase I failure frequency.

l D[a l

Frequency of Accidents Involving Radioactive Release Not all LER events involve radioactive release.

In over 600 reactor years of U.S. power-reactor operation to date (19823 there have, to our knowledge, been no radioactive releases due to load drops.

The 10 year period covered by the survey in NUREG-0612, which included 32 crane LER events, all without release, represents about 60% of all U.S. power-reactor operating time to date.

An assumption of a pre-fix frequency of some radioactive release once in 1,000 reactor years appears consistent with the LER-reflected failure experience, taken together with the absence of releases to date. With 1/4 of these releases averted by an SFP crane feature, the pertinent release frequency reduction would be 1 in 4,000 RY.

For the most part, these can be assumed to be minor releases due to limited fuel damage in the spent-fuel storage pool or in the reactor.

l n

LU 1

i

! ! O Frequency of Accidents Involvino Ma_for Releases For a load-drop event to cause a major accident, with major radioactive release, special circumstances need to be present -- circumstances that Phase I is intended to make much less likely to occur. A highly damaging heavy load drop, such as one that could destroy a core cooling feature through violation of -- or imperfections in -- Phase I provisions combined with other failures, should be unlikely, and very unlikely to lead to major release, because of back-up safety provisiens (e.g., independent additional core cooling provisions).

Review of typical load paths and associated crane-operation frequencies suggests that of all load drops in a typical PWR plant that could have radiological consequences, some 1/4 could involve equipment with a role in safe reactor shutdown, including primary-system piping.

If one assumes that there is typically a 1% probability that back-up revisions would also fail, then the pertinent major-release frequency is 1 in 1,600,000 reactor years.

Frequency Reduction Single-failure proof cranes may reasonably be expected to eliminate most, perhaps 90%, cf the residual load-drop probability after the Phase 1 J

improvements.

Thus,-the frequency reduction for major release is approximately 1 in 1,800,000 RY (90% of 1/1,600,000).

It should be noted that these estimates are sensitive to plant layout.

Plant-specific evaluations could, depending on case specifics, point to a much higher or lower major-release frequency estimate for a specific case.

For example, should layout of a specific plant be such that a particularly unfortunate load drop could destroy all means of core cooling or incapacitate the control room (possibilities suggested by the situations at Montecello and Arkansas Nuclear 1, respectively, before remedial actions were taken at those plants), the above generic analysis could be wide of the mark O

. _ _ _ ~.

. [

.V for such a plant.

The major-release accident frequency could well be an order of magnitude higher for such a plant (i.e., of the order of 1 in l

100,000 reactor years) -- or even higher, depending on plant and crane features, load paths, and operating practices.

CONSEQUENCES ESTIMATE Potential radiological consequences of load-drop accidents encompass a wide l

range of possibilities, depending on specific features of plant design, operating practices, and the nature and location of the specific load-drop event. We assume that some -- though very rough -- indication of the severity of the load-drop accident risks may be gained by using in these simplified calculations certain selected release categories described in WASH-1400 Appendix VI, pp. 2-1 to 2-4.

Category PWR 4 was selected for the l

major-release estimates for pressurized water reactors.

l In PWR 4 core cooling and containmant both fail.

Core melt occurs.

This release category is used to explore sonsequences of a load drop that O

incapacitates core cooling (during or promptly after reactor operation),

j with containment open.

The release estimates, stated as resulting public dose, based on representative generic estimates, for a hypothetical site with a projected Year 2000 mean U.S. power-reactor-site population density, developed by Battelle Pacific Northwest 8.aboratories (NUREG/CR-2800) is 2,700,000 j

person-rem.

COST ESTIMATE Costs of change-over to single-failure-proof cranes are subject to wide plant-specific variation, depending on the number of features of the specific cranes involved and other aspects of plant design and status.

l l

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Based on advice from the Auxiliary Systems Branch, 051, and lini'.ed vendor i

and utility contacts, we take the following estimates as representative (as of 1982, when the estimates were s.ade).

For future plants, the cost differential for original inclusion of SFP features is estimated at about $250,000 for PWRs (based on information from Ederer Crane Co.).

i At the pre-operating-license stage, with no startup delay, the costs --

including planning, engineering, hardware, installation, and testing -- are l

estimated at $2 million per plant. This is based on the Monticello L

experience (1 M in 1976, adjusted for inflation).

(The Monticello information was obtained from the licensee through the NRC resident l

inspector.)

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For operating PWRs the estimated costs are dominated by plant shutdown during modifications of the polar crane located inside the containment building.

(The shutdown may be an extension of a shutdown for refueling or other purposes.) The cost effect of a startup delay for a completed or nearly ceinpleted plant would be similar. With a 3-month shutdown and with shutdown costs taken as determined by the cost of replacement power at $300,000 per day, representative total change-over costs for operating PWRs are estimated at about $30 million.

RISK REDUCTION l

Based on the foregoing frequency and contequences estimates, the " expected I

value" of the risk subject to being affected by the possible Phase 11 SFP feature, i.e., the magnitude of release times the frequency of its b

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occurrence, integrated for the remaining plant life taken as 20 years, is as follows:

1 Major release risk = 20 x 2,700,000 = 30 person-rem / reactor 2,800,000 COST-BENEFIT RATIO I

The cost-benefit ratio indicated by the foregoing estimates is approximately

$1,000,000/ person-rem.

This estimate is subject to wide plant-to plant variation as well as large uncertainties in the underlying estimates of I

accident frequency and consequences. Nevertheless, it is possible to conclude with reasonable confidence that the benefit-cost ratio for the crane conversion would fail to meet a $1,000/ person-rem worthwhileness criterion by a large margin.

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Exhibit E 4

1 NUREG-1382

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j Safety Evaluation Report i

related to the full-term operating license for Oyster Creek Nuclear Generating Station Docket No. 50-219 l

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GPU Nuclear Corporation and Jersey Central Power & Light Company i

U.S. Nuclear Regulatory Commission Omce of Nuclear Reactor Regulation January 1991

,* ~ 5 N. )

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TABLE OF CONTENTS (Continued) 8 ELECTRIC POWER SYSTEMS.............................................

8-1 i

8.1 Introduction.................................................

8-1 8.2 Offsite Power System.........................................

8-1 l

8.2.1 Potential Equipment Failures Associated With

~ Degraded Grid Voltage (SEP Topic VIII-1.A)............

8-1 8.3 Onsite Power Systems.........................................

8-1

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8.3.1 Station Blackout (Generic Task A-44).................

8-1 l

8.3.2 Onsite Emergency Power Systems (Diesel Generator)

(SEP Topic VIII-2)....................................

8-3 l

8.3.2.1 Diesel Generator Annunciators................

8-3 8.3.2.2 Diesel Generator Trip Bypass.................

8-3 I

8.3.3 DC Power Systems (Onsite).............................

8-3 8.3.3.1 Station Battery Capacity Test Requirements (SEP Topic VIII-3.A).........................

8-3 8.3.3.2 DC Power System Bus Voltage Monitoring l

and Annunciation (SEP Topic VIII-3.B)........

8-4 t

lg 8.4 Electrical Penetrations of Reactor Containment (SEP

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Topic VIII-4)................................................

8-4 8.5 Appendix K - Electrical Instrumentation and Control Re-Reviews (SEP Topic VI-7.C.1)..............................

8-5 l

9 AUXILIARY SYSTEMS...................................................

9-1 9.1 Fuel Storage (SEP Topic IX-1)................................

9-1 l

9.1.1 Spent Fuel Pool Cooling System - Seismic Upgrade......

9-1 9.1.2 Control of Heavy Loads at Nuclear Power Plants (Generic Task A-36)-...................................

9-2 L

9.2 Water Systems (SEP Topic IX-3)...............................

9-4 l

9. 3 Ventilation Systems (SEP Topic IX-5).........................

9-4 9.3.1 Restoration of Ventilation............................

9-5 9.3.2 Core Spray and Containment Spray Pump Ventilation.....

9-5 9.3.3 Battery, Motor Generator, and Switchgear Room Ventilation...........................................

9-5 9.4 fire Protection..............................................

9-6

,O9 i

NUREG-1382 x

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The following modifications were made to upgrade the SFPCS in the reactor building:

i (1) Six new supports and nine replacement pipe supports were added.

(2) A 6-inch m ually operated gate valve was added where the original SFPCS i

connects with the augmented SFPCS.

(3) Seismic supports were added to the SFPCS head exchangers.

These modiffcations thus upgraded the original SFPCS from its non-seismic condition to a condition that would ensure that the pressure boundary would remain intact and functional.

In addition to the above modifications, several SFPCS valves were qualified for operability following a seismic event to ensure an isolated, seismically i

qualified cooling loop.

The modification ensures that the equipment, valves, piping, and supports contained in the cooling loop meet operability criteria l

following a seismic event and that the boundary will remain intact and functional.

9.1.2 Control of Heavy Loads at Nuclear Power Plants (Generic Task A-36)

All plants have overhead handling systems that are used to handle heavy loads in the area of the reactor vessel or spent fuel in the spent fuel pool.

Addi-tionally, loads may be handled in other areas where if they are accidentally dropped, they may damage safe shutdown systems.

Therefore, in accordance with NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants," dated July 1980 1

(Generic Task A-36), all plants should satisfy each of the following criteria for handling heavy loads that could be brought in proximity to or over safe i

shutdown equipment or irradiated fuel in the spent fuel pool area, in the reactor building, and in other plant areas.

(1) Safe 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 practicable, structural floor members, beams, etc., so that if the load is dropped, the structure is more likely to withstand the impact.

These load paths 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.

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 irra-i diated fuel or safe shutdown equipment.

At a minimum, procedures should cover handling of those loads listed in Table 3-1 of NUREG-0612.

These l

l procedures should include identification of required equipment, inspec-l tions and acceptance criteria required before the load is moved, the steps and proper sequence to be followed in handling the load, defining the safe load path, and other special precautions.

O NUREG-1382 9-2

y-(3) Crane operators should be trained and qualified and should conduct them-selves in accordance with Chapter 2-3 of American National Standards

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l Institute (ANSI) B30.2-1976, " Overhead and Gantry Cranes."

(4) Special lif ting devices should satisfy the guidelines of ANSI N14.6-1978, l

" Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials." This standard should_ apply to all special lifting devices that carry heavy loads in areas as defined above.

For operating plants certain inspections and loads tests may be accepted in lieu of certain material requirements in the standard.

In addition, the stress design factor stated in Sec-tion 3.2.1.1 of ANSI N14.6 should be based on the combined maximum static and dynamic loads that could be imparted on the handling device on the

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basis of the characteristics of the crane that will be used.* This is in lieu of the guideline in Section 3.2.1.1 of ANSI N14.6, which baset the stress design factor on only the weight (static load) of the load and of the intervening components of the special handling device.

(5)

Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9-1971, " Slings."

However, 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" that produces the maximum static and dynamic load.

Where 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.

(6) The crane should be inspected, tested, and maintained in accordance with O'

Chapter 2-2 of ANSI B30.2-1976, except that tests and inspections should be performed before use where it is not practicable to meet the frequencies of ANSI B30.2 for periodic inspection and test, or where the frequency of crane use is less than the specified inspection and test frequency.

(7) The crane should be designed to meet the applicable criteria and guidelines of Chapter 2-1 of ANSI B30.2-1976 and of CMAA-70, " Specifications for Electric Overhead Travelling Cranes" (Crane Manufacturers Association of America).

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 specifica-tion is satisfied.

i A plant conforming to these seven guidelines will have developed and implemented, through procedures and operator training, safe load travel paths so that, to the maximum extent practicable, heavy loads are not carried over or near irradiated fuel or safe shutdown equipment.

A plant conforming to these guidelines will also have provided sufficient operator training, handling-system design, load-handling instructions, and equipment inspection to ensure reliable operation of the handling system.

It has been found that load-handling operations at Oyster Creek can be expected to be conducted in a highly reliable manner consistent with the staff's objectives as expressed in these guidelines.

• For the purpose of selecting the proper sling, loads imposed by the safe shut-down earthquake need not be included in the dynamic loads imposed on the sling or lifting device.

NUREG-1382 9-3

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l NUREG-0612, Section 5.3, also lists certain measures that should be initiated i

to provide reasonable assurance that the handling of heavy loads will be per-O formed in a safe manner until final implementation of the general. guidelines i

of NUREG-0612 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, and 6 identified above; a review of load-handling procedures and operator training; and a visual inspec-tion program, including component repair or replacement as necessary of cranes, slings, and special lifting devices to eliminate deficiencies that could lead to component failure.

The evaluation of information provided by the licensee indicates that Oyster Creek complies with the staff's measures for interim protection.

By Generic Letter 85-11 dated June 81, 1985, the staff concluded that the Oyster Creek station along with other plants has provided sufficient protection so that the risk associated with potential heavy-load drops is acceptably small and that the objective identified in Section 5.1 of NUREG-0612 for providing

" maximum practical defense in depth" is satisfied.

9.2 Water Systems (SEP Topic IX-3)

Under SEP Topic IX-3, the staff reviewed the licensee's turbine building closed cooling water system, reactor building closed cooling water system, service water system, and emergency service water system to ensure that the systems have the capability to meet their design objectives and, in particular, to ensure the following:

(1) Systems are provided with adequate physical separation so that there are y

no adverse interactions among those systems under any mode of operation.

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(2) Sufficient cooling water inventory has been provided, or adequate provi-sions for makeup are available.

(3) Tank overflow cannot be released to the environment without monitoring and unless the level of radioactivity is within acceptable limits.

(4) Vital equipment necessary for achieving a controlled and safe shutdown is not flooded as a result of the failure of the main condenser circulating water system.

On the basis of its review of the station service and cooling water systems for Oyster Creek, the staff concluded that the essential system and function are the emergency service water. system for torus heat removal.

In a letter dated November 13, 1981, the staff determined that the design of the above system conforms with current regulatory guidelines and with GDC 44 regarding the capability and redundancy of the essential functions of the system.

9.3 Ventilation Systems (SEP Topic IX-5) 10 CFR Part 50 (GDC 4, 60, and 61), as implemented by SRP Sections 9.4.1, 9.4.2, l

9.4.3, 9.4.4, and 9.4.5, requires that the ventilation systems have the capabil-l ity to provide a safe environment for plant personnel and for engineered safety NUREG-1382 9-4 l

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