ML18142A061

Submits Addl Info for Open Items Noted in Draft Technical Evaluation Rept Re Reactor Containment Polar crane,motor- Driven Platform & Hoists & 6 & 10-ton Monorail Sys,In Response to 840614 Ltr & 840726-27 Site Visit

ML18142A061
Person / Time
Site:	Surry
Issue date:	09/28/1984
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Eisenhut D Office of Nuclear Reactor Regulation
References
REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR 366, NUDOCS 8410030337
Download: ML18142A061 (4)

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e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261

w. L. STEWART VICE PRESIDENT September 28, 1984 NUCLEAR OPERATIONS Mr. Harold R. Denton, Director Serial No. 366 Office of Nuclear Reactor Regulation E & C: TLG:baj/2001N Attn: Mr. D. G. Eisenhut, Director Docket Nos. 50-280 Division of Licensing 50-281 U. S. Nuclear Regulatory Commission License Nos. DPR-32 Washington, D. C. 20555 DPR-37 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT NOS. 1 AND 2 NUREG 0612 - CONTROL OF HEAVY LOADS PHASE II DRAFT TECHNICAL EVALUATION REPORT {TER)

The following is being provided in response to your letter dated June 14, 1984 and the site visit and meeting held at Surry Power Station on July 26 and 27, 1984.

As requested, the site visit was arranged and walkdowns performed. As a result of the site visit four (4) of the TER items required additional information to be provided. The four open items were as follows:

I) 2.2.2 Reactor Containment Polar Crane II) 2.3.3 Motor-Driven Platform and Hoists III) 2.4.8 10-Ton Monorail System IV) 2.4.9 6-Ton Monorail System 2.2.2 REACTOR CONTAINMENT POLAR CRANE The remaining open items for this section include supplying a summary of the Westinghouse analysis on the effects of the reactor vessel head drop onto the reactor and a summary of the Stone and Webster Engineering analyses of the drop effects upon the reactor vessel supports. Further explanation and clarification of the nine month report statement that 11 Postulated load drops into the reactor vessel shall not be addressed 11 was requested. Also requested was clarification of the intent of Technical Specification 3.10 with respect to administratively controlling operations of the reactor containment cranes in the handling of heavy loads over the reactor vessel with fuel in the vessel.

e VIRGINIA. ELECTRIC AND POWER COMPANY TO Harold R. Denton The Reactor Vessel Head Drop Analysis was performed by Westinghouse to address the effects of the 11 Worst Case 11 drop scenario, thereby enveloping postulated heavy load drops onto the reactor. The results of the analysis showed that the reactor vessel nozzle stresses are well within allowable values. The maximum principal stress was found to be 14,040 psi with an allowable value of 84,000 psi.

The support loadings resulting from such a drop were then used to analyze the effects upon the containment mat, reactor vessel supports and the neutron shield tank. The final results of this effort showed that the compressive strength of the containment mat provides adequate shear capacity to withstand the loading which would result from such a drop. The reactor vessel supports were found to be adequately designed to withstand such loading without failure. The neutron shield tank was determined to be capable of withstanding the maximum calculated head drop load.

In conclusion, the reactor vessel head drop analysis showed that even if the 11 Worst-Case 11 drop occurred, the recommended guidelines and criteria of NUREG-0612, Section 5.1 would be satisfied.

11 Postulated load drops into the Reactor Vessel shall not be addressed. 11 This statement was meant to imply that the probability of a heavy load drop into the reactor vessel is essentially non-existant. Administrative controls such as the Operation Procedures, Maintenance Procedures, the use of the safe load path drawings, and Technical Specification 3.10 restrict any movement of 11 Heavy Loads 11 over the reactor vessel when fuel is in the vessel. The only heavy loads lifted over the vessel when fuel is in place are those items which must be removed for refueling (Reactor Vessel Head and Upper Internals). The reactor vessel head drop analysis envelopes these items, therefore they need not be analyzed individually. In addition, as noted in Technical Specification 3.10, 11 the fuel handling accident has been analyzed based on the activity that could be released from fuel rod gaps of 204 rods of the highest power assembly with a 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay period following power operations at 2550 MWT for 23,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

The requirements detailed in Specification 3.10 provide assurance that refueling unit conditions conform to the operating conditions assumed in the accident analysis. 11 Technical Specification 3.10 specifies operating limitations during refueling operations to assure that no accident could occur during refueling operations that would affect public health and safety. The Technical Specification includes such limitations as: required equipment availability, water cover for shielding, minimum boron concentration with appropriate sampling requirements, communication requirements, subcritical time period, limiting weight of loads to be handled and interlocks used to insure safe handling of fuel assemblies.

Technical Specification Section 3.10B ensures that the administrative controls and operating limitations are met and provide defense-in-depth as stated, 11 If any one of the specified limiting conditions for refueling is not met, refueling of the reactor shall cease, work shall be initiated to correct the conditions so that the specified limit is met, and no operations which increase the reactivity of the core shall be made. 11

e VrnoINIA ELEcTRic AND PowER CoM:PANY To Harold R. Denton 2.2.3 MOTOR-DRIVEN PLATFORM AND HOISTS The remaining open item for this section is the movement of the transfer canal door over spent fuel and the use of administrative controls to satisfy the requirements of NUREG-0612.

As discussed during the site visit, Vepco is in the process of having a drop analysis of the transfer canal door onto the spent fuel racks performed. At this time the analysis is not complete. However, Vepco is scheduled to submit the results of the transfer canal door drop analysis to the NRC by the end of 1984. In addition, engineering design and fabrication of a storage rack to be located and installed on the east end of the fuel pool is in progress. The rack will allow the Unit 2 transfer canal door to be stored next to its location when not in use. With the addition of the new storage rack and an acceptable drop analysis, the reliance upon the administrative controls will no longer be required. The above work will be complete and in place prior to the next Unit 2 refueling outage.

2.4.8 10-TON MONORAIL SYSTEM The remaining open item for this section deals with the reliance upon administrative controls to impose a one (1) foot load-carrying height restriction in the Auxiliary Building on elevation 27ft. 6in.

The load drop analyses performed were based upon enveloping the maximum loads of the removable slabs (8.5 tons), conservative assumptions, and the one foot maximum load lift height. Stresses in the structure resulting from a drop of these defined conditions entered into the inelastic range, however, complete failure did not occur in the structure. The one (1) foot load-carrying height restriction shall be adhered to in handling of the 8.5 ton removable slabs. However, other loads which may be carried by the monorail system are of a magnitude considerably less than maximum load used for the analyses and have considerable flexibility in the load-carrying height. In the movement of the lesser loads, the one (1) foot height will be used as a guide and not a restriction.

2.4.9 6-TON MONORAIL SYSTEM The remaining open item for this section was the same as the open item for the 10-Ton Monorail System; reliance upon administrative controls to impose a one(l) foot load-carrying height restriction in the auxiliary building on elevation 13ft. Oin.

e VIRGINIA ELECTRIC AND POWER COMPANY TO Harold R. Denton The load drop analyses for this system were enveloped by the maximum removable slab loads of 8.5 tons used for the 10-ton Monorail System analyses. However, the actual maximum loads carried by this system are the removable slabs of 4.5 tons, which is much less than the maximum load used in the analyses. Therefore, there is considerable flexibility in the load-carrying height. In the movement of the lesser loads, the one (1) foot height will be used as a guide and not a restriction.

The information provided is to be used in finalizing the TER. If further clarification or information is required, please contact us.

aL\~

ry truly yours, W. L. Stewart cc: Mr. James P. 0 Reilly 1

Regional Administrator Region II Mr. D. J. Burke NRC Resident Inspector Surry Power Station