ML20135E805

Forwards Response to NRC Bulletin 96-002, Movement of Heavy Loads Over Spent Fuel,Over Fuel in Reactor Core or Over Safety-Related Equipment

ML20135E805
Person / Time
Site:	Prairie Island
Issue date:	12/06/1996
From:	Wadley M NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
IEB-96-002, IEB-96-2, NUDOCS 9612120072
Download: ML20135E805 (9)

Text

• Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch, Minnesota 55089 December 6,1996 NRC Bulletin 96-02 l

U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Response to NRC Bulletin 96-02 Movement of Heavy Loads over Spent Fuel, over Fuel in the Reactor Core, or over Safety-Related Equipment NRC Bulletin 96-02 requested a review of our plans and capabilities for handling heavy loads within 30 days. On May 13,1996, we submitted our response. In that response we said: l To date, we have performed the requested review to a level of detail to allow us to provide a response to the request for information with a reasonable level of assurance of its accuracy. Since the review has not been completed to the level l of detail that we deem prudent (due to the resources necessary to complete such a review and the short reporting schedule specified in the bulletin), we will /

continue the review of our heavy loads handling against the regulatory guidance and ourlicensing basis and communicate tho results of the review .

l We have completed our review and attached are the results of that review. The attachment provides a complete report, it incorporates the results of both the initial

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review and the completed review. The report stands alone; it replaces, in its entirety, the report of May 13,1996.

Please note that there is a discussion of a lifting rig used for moving temporary spent fuel storage racks. This rig was procured prior to the issuance on December 22,1980 i 9612120072 961206 l PDR ADOCK 05000282 G PDR l

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USNRC NORTHERN STATES POWER COMPANY December 6,1996 Page 2 l

l of the untitled Generic Letter on Heavy Loads. It, therefore, does not meet all the

! current criteria for special lifting devices. However, there is a discussion justifying its continued use and a request for notification if you desire further communication regarding our intent to continue its use. Any use would not occur prior to January 1998.

In this letter Nuclear Regulatory Commission commitments are indicated as the italicized statements in the attachment.

Please contact Jack Leveille (612-388-1121, Ext. 4662) if you have any questions related to this letter.

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Michael D Wadley Plant Manager Prairie Island Nuclear Generating Plant c: Regional Administrator- Region Ill, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg Attachments:

1. Affidavit

2. Response to NRC Bulletin 96-02 t

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1 UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY l i

~ PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NO. 50-282 '

50-306 I

BULLETIN 96-02, MOVEMENT OF HEAVY LOADS OVER SPENT FUEL, OVER FUEL IN THE REACTOR CORE, OR OVER SAFETY-RELATED EQUIPMENT Northern States Power Company, a Minnesota corporation, with this letter is submitting information requested by NRC Bulletin 96-02.

This letter contains no restricted or other defense information.

NORTHERN STATES POWER COMPANY BY / .-

Michael D Wadley Plant Manager j

Prairie Island Nuclear Generating Plant l l

Onthis day of 12AM l  ! bbefore me a notary public in and for said County, personally appefared Michael D Wadley, Plant Manager, Prairie Island Nuclear Generating Plant; and being first duty sworn acknowledged that he is authorized to execute this document on behalf of Northem States Power Company, that he knows the ogntents thereof, and that to the best of his knowledge, information, and belief the statements mad in it are true and that it is not interposed for delay.

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l RESPONSE TO NRC BULLETIN 96-02 f l

Bulletin 30-day Reauired Resoonse:  ;

For licensees planning to implement activities involving the handling of heavy  ;

loads over spent fuel, fuel _ in the reactor core, or safety-related equipment within the next 2 years from the date of this bulletin, provide the following:

. A report, within 30 days of the date of this bulletin, that addresses the licensee's review of its plans and capabilities to handle heavy loads while the reactor is at power (in all modes other than cold shutdown, refueling, and defueled) in accordance with existing regulatory guidelines. The report i should also indicate whether the activities are within the licensing basis and should include, if necessary, a schedule for submission of a license amendment request. Additionally, the report should indicate whether changes to Technical Specification will be required.

PRAIRIE ISt.AND REPORT: ,

l Our Configuration Management group prepared a " Heavy Loads" Design Basis i Document (HL DBD) document in 1990 that evaluated the design and licensing criteria against the plant configuration and administrative processes. Discrepancies found  !

during this process were dispositioned in accordance with the Configuration l Management procedures. The initial DBD was verified in depth (in 1990). The l objectives of the verificetion were to verify that: (1) design bases information pertaining I i

to the Heavy Loads Program was accurately reflected in Revision 0 of the HL DBD, (2) l the implementation of the HL Program conforms to NRC requirements and NSP l

commitments, and (3) sufficient programmatic controls are in place to ensure continual compliance with regulatory requirements and NSP commitments.

Since 1990, the DBD has been revised twice. These revisions have focused on the changes that have taken place since the initial issuance of the HL DBD (e.g., upgrading the auxiliary building crane for the handling of spent fuel storage casks) and have not l re-verified the details that were verified in depth in 1990. For the review requested by this bulletin, we have performed a verification in depth.

We have determined that, heavy loads are handled as described in the correspondence between the NRC and NSP in response to the December 22,1980 letter from Darrell l- Eisenhut, Director of Licensing, NRR, to all licensees of operating plants, titled " Control of Heavy Loads" with four notable differences.

! 1. The Prairie Island auxiliary building crane was upgraded to meet the single-failure-l proof criteria of Section 5.1.6 and Appendix C of NUREG-0612 in anticipation of l handling spent fuel dry storage casks. Prior to handling the first dry storage cask,

! submittals to the NRC in support of Technical Specification changes regarding the

Attachment 2 December 6,1996 Page 2 l 1

, use of the upgraded crane were made (see details starting on page 3). I 1

l j 2. A construction gantry crane can be temporarily installed for the purpose of l

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installing or removing temporary spent fuel storage racks in Pool #1. The heavy i load handling considerations utilizing this crane have been discussed in docketed correspondence with respect to a fuel pool re-rack modification. This  ;

correspondence occurred at roughly the same time period as that of the generic i

! heavy load issue discussions. Since, at the time, this crane was thought of as i temporary (just to be used during these modifications), it was -not addressed  !

during our responses to the generic issues. In fact, the temporary racks were j l removed once using this crane and utilizing the heavy loads handling process  ;

reviewed by the NRC for the re-rack project. Also, the racks were installed and removed once using the upgraded auxiliary building crane. During this evolution, ,

the rack movements were controlled by procedures which met heavy loads  !

handling generic requirements but the rack movements have not been  !

incorporated into the formal plant heavy loads handling program. - We do not  !

anticipate needing to use this crane. However, in the event that we will need to use the crane, we intend to meet the process control requirements of NUREG-0612 (we would have adequate lead time to make the necessary procedure i

changes). We > .uerate our commitment of May 13,1996: We willensure '

l compliance with NUREG-0612, Section 5.1 prior to the next use of the gantry crane within the spent fuelpool enclosure or submit conespondence detailing any  :

necessary exception andits basis.

Also related to the movement of temporary spent fuel storage racks is the rack Iifting rig. The lifting rig satisfies the criteria to be considered a special lifting device and, thus, according to the NUREG-0612 Section 5.1.1.(4) and 5.1.6(1)a,  !

should meet the requirements of ANSI N14.6-1978, including Section 6.0. A  !

substantial portion of the fixture's design and fabrication does meet this standard,  !

including a design margin of 10 to 1 which includes the crane dynamic load factor.

Nonetheless, there are some ANSI N14.6 criteria that are not met by the device.

These items are:

. The design output documents do not specify required repair procedures or post repair testing.

. Load bearing members have no associated material testing.

. No specifications relative to surface smoothness for decontamination purposes.

. No record of surface preparation before painting.

! . A 150% post fabrication load test was performed versus a 300% test.  ;

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Attachment 2 December 6,1996 Page 3 ,

Certain of these items can be addressed from a practical standpoint.  ;

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! . As yet there have been no repairs required on the rig. If there were to be, the r l

repair and testing of the device would be acceptable only if it met its original design criteria and testing.  !

. The two items associated with our ability to decontaminate the device are the surface smoothness requirements and surface preparation for painting. Our l experience with the device during the re-rack project and subsequent lifts [

l shows that it is "deconable." Thus, although these items are not specifically

, addressed in the design output documents, their intent is fulfilled in the ,

l device. Finally, these items do not affect the load handling capability of the  :

device.  ;

. The lifting rig was specified and designed before the un-numbered Generic Letter for heavy loads handling dated December 22,1980. Because of this timing, material testing and a 300% versus 150% load test were not included in the design specification. Further, there are grounds that these items can be exempted in the gase of this device. First, the temperature of the environment in which the rig is used is approximately 70*F in the auxiliary building and approximately 110*F in the pool. These temperatures are well above nil ductility transition temperatures of the materials. Second the rig l has had extensive use through the re-racking project and has successfully passed all NDE performed on the load bearing welds.

All the factors discussed above indicate that the design and fabrication of this special lifting device will provide a degree of load handling reliability equivalent to that expected from an initial design in accordance with ANSI N14.6-1978.

l It is our intent to utilize this lifting rig for temporary spent fuel storage rack movement within the spent fuel pool enclosure, as necessary, based on the above discussion. Please notify us if you require further communication on this matter.

' 3. In our response to the December 22,1980 undated Generic Letter on Heavy Loads, we stated that we would perform NDE on special lifting devices (reactor head lifting I device, reactor internals lifting device, and turbine rotor spreader assembly) at  ;

certain intervals. We have been unable to locate evidence that these examinations have been performed as committed. We have been unable to determine why these examinations have not been performed. However, these devices willbe examined as previously committedprior to theirnext use. In addition, these examination requirements have been incorporated into the plant's formal Inservice Inspection 4 programs for each Prairie Island unit to ensure the examinations are performed in the future as required.

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l Attachment 2 l December 6,1996 j Page 4

4. In our response to the December 22,1980 undated Generic Letter on Heavy Loads, i we stated that no loads could physically be carried over the spent fuel pool enclosure with the auxiliary building crane because there was a lack of space to i accomplish this. This was in error; we surmise that this was written in our response l with the 125-ton hook in mind - in which case the statement is true. However, there l is a small hook on the crane also. Using the small hook, loads can be carried over the enclosure. This can be done within the guidance of NUREG-0612 and we are currently changing our procedures to allow loads to be carried over the enclosure i l within this guidance. We will not move any future loads over the enclosure until the necessaryprocedures are in place. l Licensing Basis  ;

i Our review has determined that we are in compliance with our current licensing basis  !

with the exceptions of two Updated Safety Analysis Report (USAR) mis-statements. t One mis-statement is of the plant configuration with respect to the clearance above the spent fuel pool enclosure roof and the auxiliary building crane, discussed in item 4 l above. The other error is the weight given for the spent fuel pool protective covers; the USAR states a weight of 3700 pounds per cover (which is the weight that the design  !

drawing designated) whereas the weight has been determined to be 4550 pounds. l' These mis-statements do not affect our compliance with NUREG-0612 and we anticipate correcting the errors through the 10CFR50.59 process.

Technical Soecifications in addition, our review has determined that no Technical Specification changes will be l l required.

l i l License Amendment Reauests i

Because the discrepancies are likely remedied through the 10 CFR 50.59 process and there are no Technical Specification changes required, no license amendment requests need to be submitted.

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Attachment 2 December 6,1996 Page 5 Details regarding the upgraded auxiliary building crane follow:

Prairie Island Cask Handling System Design l

In preparation for the dry cask storage project, the Prairie Island auxiliary building crane  ;

was upgraded in 1992 to meet the single-failure-proof criteria of Section 5.1.6 and  !

l Appendix C of NUREG-0612. The upgrade involved replacement the existing trolley l structure and hoist with an Ederer, Inc. designed trolley and hoist. The Ederer single-l failure-proof design is described in Revision 3, Amendment 3, to Generic Licensing

! topical Report EDR-1(P)-A, "Ederer Nuclear Safety-Related Extra Safety and l n + toring (X-SAM) Cranes," dated October 8,1982. .The report describes the design and testing of the single-failure-proof features which are included in Ederer's X-SAM cranes intended for handling spent fuel casks and other safety related loads in a j

nuclear power plant. NRC staff approval of the report was documented in a Safety Evaluation transmitted by NRC letter dated August 26,1983.

Licensing l

NSP submitted the crane upgrade for approval to the NRC via letter dated October 4, 1 1991, as supplemented by letter dated December 16,1991, i l The NRC approved the license amendment request and issued License Amendment i

Nos. 99 and 92 to Facility Operating License Nos. DPR-42 and DPR-60 on July 9, 1992.

As requested in the July 9,1992 approval, NSP submitted additional information to the NRC on January 25,1993. This information was reviewed and found acceptable by the NRC on May 3,1993.

In addition to the crane upgrade, the TN-40 cask and lifting system are designed to l meet the single-failure proof requirements of NUREG-0612. The lift beam is designed, i fabricated and tested in accordance with ANSI N14.6-1986. The cask trunnions are l designed with an increased safety factor, as required by section 5.1.6 (3) of NUREG-l 0612.

l The cask and lift beam design were reviewed as part of the dry cask storage Part 72 l

application. This cask design was approved and issued NRC license SNM-2506 on i October 19,1993. i

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Attachment 2 December 6,1996 l Page 6 1 The pre-operational load test of the cask trunnions was a subject of a separate NRC j SER, dated June 12,1995. l l

NRC Pre-Ooerational Insoection Cask and lift beam fabrication, as well as load handling procedures and practices were  !

reviewed by the NRC prior to initial cask loading during a special NRC inspection from  !

January 24 through May 11,1995. This inspection was documented in inspection l Report 50-282/95002(DRP); 50-306/95002; 72-10/95002(DRP), June 30,1995. This inspection concluded: ,

The licensee satisfied all the design and testing requirements specified in established industry standards for the control of heavy loads, such as dry cask storage.

Section 2 of this inspection report discussed the inspector's review of the Prairie island ,

cask handling system, and included the areas of auxiliary building crane modification, l post modification testing, load testing of the TN-40 lift beam, load testing of the aux!!iary )'

building crane, load testing of the cask trunnions, auxiliary building crane hook load test, cask transport vehicle route, special lifting device configuration, rated load of the auxiliary building crane, lid lift shank hook and rigging, NDE of the lid lifting bridle, and cask handling in the spent fuel pool. ,

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