AECM-86-0011, Responds to 850809 Request for Addl Info Re 850506 Application for Amend to License NPF-29,allowing for Installation of High Density Spent Fuel Racks

Responds to 850809 Request for Addl Info Re 850506 Application for Amend to License NPF-29,allowing for Installation of High Density Spent Fuel Racks

ML20140E484
Person / Time
Site:	Grand Gulf
Issue date:	03/15/1986
From:	Kingsley O MISSISSIPPI POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AECM-86-0011, AECM-86-11, TAC-57619, NUDOCS 8603280001
Download: ML20140E484 (8)

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j MISSISSIPPI POWER & LIGHT COMPANY Helping Build Mississippi RIRalinilddB P. O. B O X 164 0, J A C K S O N, MIS SIS SIP PI 39215-1640 C. D. KINGSLEY, J R. March 15, 1986 YlCE PREllDENT NUCLEAR OPERATtONS U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20555 Attention: Mr. Harold R. Denton, Director

SUBJECT:

Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 High Density Spent Fuel Racks -

Response to Auxiliary Systems Branch Questions AECM-86/0011 By letter dated May 6, 1985 (AECM-85/0143) Mississippi Power 6 Light (MP&L) requested an amendment to License NPF-29 to allow for the installation of high density spent fuel racks at Grand Gulf Nuclear Station (GCNS) Unit 1.

By letter dated August 9, 1985 the NRC Staff's Auxiliary Systems Branch requested additional information regarding the amendment request.

The attachment to this letter is MP&L's response to the Staff's request.

If there are any additional questions, please contact this office.

Yours truly, Q.- .

• M ODK:dmm Attachment cc: See Next Page 8603280001 860315 PDR ADOCK 05000416 p PDR Y\ \

J14AECM85112502 - 1 Member Middle South Utilities System

AECM-86/0011

. :Page 2 cc: -Mr. T. H. Cloninger-(w/a)

Mr. R. .B. McGehee (w/a)

Mr..N. S..Reynolds (w/a)

Mr,. H. L. Thomas (w/o)

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Mr. R. C.-Butcher (w/a)

Mr. James M. Taylor, Director (w/a)

Office of Inspection & Enforcement  ;

'U. S. Nuclear Regulatory Commission Washington, D. C. 20555.

Ihr. J. Nelson Grace. Regional Administrator (w/a)

U. S. Nuclear Regulatory Commission Region II 101 Marietta St.,'N.'W., Suite 2900 Atlanta,-Georgia 30323 I

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Attrchssnt to AECM-86/0011 AUXILIARY SYSTEMS BRANCH REQUEST FOR ADDITIONAL INFORMATION GRAND GULF UNIT 1 SPENT FUEL POOL CAPACITY EXPANSION QUESTION 1 Operating License Condition 2.C.20 states, in part, that "No irradiated fuel may be stored in the Unit 1_ Spent Fuel Pool prior to completion of modifications to the Standby Service Water (SSW) system and verification that the design flow can be achieved to all SSW system components....

Until the SSW system is modified, the spent' fuel pool cooler shall be isolated from the SSW system by locked closed valves. The position of' these valves shall be verified every 31 days until the design flowrate for (the) SSW system is demonstrated." Provide a discussion, P& ids, and a schedule of completion for the modifications to 1) satisfy License

, Condition 2.C.20, and 2) remove the additional heat load imposed by the proposed increased storage capacity of the spent fuel pool.

NOTE: The increased heat used'for sizing the SSW system should conform to the results of Question 2 below.

RESPONSE

1.1 For the purpose of background information, the significant modifications to the Standby Service Water System (SSW) to satisfy License Condition 2.C. (20) consist of replacing the SSW pump impellers and motors to increase the flow rate to meet original design requirements. Following completion of the modifications, testing will be conducted to verify adequate system flow rates in the loss of power /LOCA system configuration. After the testing, the administrative 1y controlled minimum basin level of'107' can be returned to the minimum level of 84'-6", based on NPSH requirements.

The principal changes to the system P&ID are as follows:

1) Addition of system pressure relief valves (Trains A and B),

2) Replacement of_the butterfly valves in the SSW recirculation lines with globe valves (Trains A and B),

3) Addition of two secondary relief valves to the SSW system to protect the instrument air compressors (Train B only),

4) Increased size of the inlet and outlet piping to the SSW pump motor bearing oil coolers from 1/2 inch to 3/4 inch and the line orifice to 0.16 inch (Trains A and B).

These. modifications to the-P&ID do not impact the' cooling characteristics of the SSW system.or the Fuel Pool Cooling and Cleanup and RHR systems.

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.Attachsint to

'AECM-86/0011.

The.SSW 'B' train modification work was completed during a plant-outage ending December 10,'1985; the SSW 'A' train work will be i completed'during Refueling Outage 01 (RF01) scheduled to begin September 1, 1986.

1.2 lThe thermal / hydraulic analysis for' fuel storage utilizing high density storage racks does.not take credit for any SSW system modifications txt increase cooling water flowrate to the fuel pool cooling or RHR heat exchangers above the original design flowrate values. No SSW system modifications are being made as a result of fuel storage' system modifications.

QUESTION 2, The calculated spent' fuel decay heat loads identified in the submittal did not follow the guidelines of the Standard. Review Plan (NUREG-0800) Section 9.1.3 and Branch Technical Position ASB 9-2. The heat loads provided in the submittal are not conservative with respect to those calculated by using the aforementioned references and therefore the licensee's i calculations are not acceptable. Provide the results of revised calculations which'use the aforementioned references.

i RESPONSE.

The calculated heat loads identified in the submittal were-based on the ASB 9-2, Rev. 1. Revision 2 of this document, printed in July 1981, significantly changedothe heat rate terms. Revised results for the heat load and pool bulk temperature utilizing ASB 9-2, Revision 2 are provided in conjunction with'the response to NRC Question #3 below.

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QUESTION 3 Specify the overall heat transfer rate (BTU /hr-f t 2 *F) for the ' spent fuel pool cooling system's heat exchanger. If the heat transfer rate is similar to that of the RHR heat exchanger, then the fuel pool cooling system should be re-evaluated to assure that the spent fuel pool water temperature does not exceed 140*F for normal heat load conditions and the

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single failure of one spent fuel' pool cooling train. Reliance on the RHR

, system for normal refueling heat loads is not acceptable.

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RESPONSE

The overall heat transfer coefficient for the spent fuel pool heat exchanger is 277.7 BTU /hr-ft 2 *F and for the RHR the coefficient is 210 BTU /hr-ft 2 _.F. There are two spent fuel pool coolers and.two'RHR heat-exchangers available for cooling purposes. The overall heat rate for the 4 RHR heat exchangers is greater-than that.of the fuel pool coolers due to l the RHR heat exchanger's larger transfer surface area, transfer J14AECM85112502 S

Attach ;nt to AECM-86/0011 coefficient, and shellside flow rate. All units along with the associated piping and other appurtenances are seismic category I. The flow schematic used in the thermal hydraulic analysis is shown in Figure 5.1.1 of the Licensing Report.

Either RHR Train A or B can be lined up to service the Upper' Containment Pool (UCP) and the Auxiliary Building Spent Fuel Pool (SFP) to supplement cooling from the fuel pool cooling system. For the analyses described here and in the Licensing Report, only one train was assumed to be used for supplemental cooling. 'An RHR train when in the supplemental cooling mode delivers cooling water flow to locations other than the -UCP and SFP,

[e.g., the reactor cavity (FSAR 9.'l-30)]. For this reason only a portion of the total RHR flow was conservatively assumed to be available for these analyses. That portion is 2550 gpm (approximately.35% of the total RHR flow of 7450 gpm). This is described in Section 5.1.1 of the' Licensing Report. When an RHR train is assumed to be used in the' supplemental cooling mode, available flow from that train is distributed between the UCP and SFP (Cases 1, 2 and 3 below). The RHR train can also be directed exclusively to the UCP or the.SFP (Case 4 below).

The two " base" cases from the Licensing Report were re-analyzed (Table 5.1.1B) along with two additional. cases. The re-analyzed cases and their associated results are described below:

(1) Normal batch discharge into the UCP 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> after reactor shutdown at the rate of 4 assemblies per hour followed by transfer to the SFP beginning 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later at a rate of 4 assemblies per hour. Two spent fuel pool coolers are in operation. One RHR train is assumed to providing supplemental cooling as described above for 30 days following shutdown.

(2) Full core discharge into the UCP'110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> after reactor shutdown at the rate of 4 assemblies per hour followed by transfer to the SFP beginning 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later at a rate of 4-assemblies per hour. Two spent fuel pool coolers are assumed to be.in service along with one RHR train providing flow as described above to both the UCP and SFP. For this case the RHR train is assumed to be in service.

(3) In addition to the above two analyses, MP&L also considered the case where case (1) above is modified to stipulate that one out of two spent fuel pool coolers is not available from the beginning of the fuel transfer process. A single RHR train is assumed to be in service providing flow to both the UCP and SFP as described above.

Table 3.1 below gives the pertinent output data of . interest for the first three cases which were analyzed to evaluate the SFP l temperature response.

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Attechisnt to AECM-86/0011 Tab 1e' 3.1 SFP and UCP Response - Key Parameters SFP SFP SFP MAX LOCAL UCP.

CASE MAX BULK TIME (HRS) . POOL WATER MAX BULK NO. TEMPERATURE TO BOIL

• TEMPERATURE TEMPERATURE (1) 126.3 15.6 150.9 108.2 (2) '145.1 7.1. 173.9 121.0 (3) 115.8 13.1 149.7 112.8

• (Assuming Loss of All Coolers & Heat Exchangers)

The above results- show that the SFP maximum bulk temperatures are below.

140' and 150* for a normal batch discharge and full core offload, respectively.

(4) An additional analysis was subsequently performed to evaluate the

'UCP temperature response under a full core discharge scenario with no transfer out to the.SFP. The full core discharge commenced ~at 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> after reactor shutdown at a transfer rate of I assembly per hour. One RHR train is assumed to be in service. In this case RHR return flow is assumed to be from the UCP, through the fuel pool gate into the reactor cavity and vessel and into the recirculation-In this case the RHR train is isolated from the SFP and is

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system.

directed to the UCP. For this analysis only a' portion (approximately

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25%) of RHR flow was assume'd to service ~the fuel storage area.- The analysis resulted in a maximum UCP bulk temperature of 146.4*.

These analyses' support operations that are consistent with the assumptions ~ ,

described above and in the cases presented and analyzed in the Licensing Report (Table 5.1.1A). It should be noted that there are additional' combinationsaof valving arrangements and scenarios with the pool gate installed or removed-which could physically be achieved, thus altering the flow path used in these licensing analyses.

For example, the Licensing Report described-in Assumption 6 (p. 5-5)

I assumes that return flow for the RHR system to be through the open UCP _ .

fuel pool gate, the reactor vessel and. recirculation system. Closure of the UCP fuel pool gate'in combination with an alteration of specified fuel assembly transfer rates would. require additional evaluation beyond that presently here or in the Licensing Report.

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Attschx:nt.to AECM-86/0011-Admir.istrative controls and appropriate system operating procedures will be intplemented as a result of these and any future . analyses to ensure that the maximum SFP bulk-temperature is maintained below 140* for a normal discharge and below 150* for a full core d.ischarge. The adoption of any flow path, flow rate, or spent fuel assembly transfer rates different from that described here or in the Licensing Report would require furthar analysis, evaluation per 10CFR50.59,.and appropriate administrative controls and procedures prior to implementation.

Regarding the reliance'on.the RHR system for normal refueling heat loads, it is MP&L's position that RHR supplemental cooling may be utilized for a-period of time during refueling operations until the spent fuel pool heat load can be handled by the spent fuel pool cooling system alone. MP&L's-position is supported by discussion in the Grand Gulf FSAR,(Section 9.1.3) and endorsed by the NRC. staff in the Grand Gulf SER (Section 9.1.3).

QUESTION 4

' Verify'that FSAR Figure 9.1-26 (Amendment 52) is still applicable in that it.shows two heat exchangers and pumps whereas Figure 5.1.1 (of the licensing report) of the May 6th submittal indicates one heat exchanger.

RESPONSE

FSAR Figure 9.1-26 (Amendment 52) correctly shows two fuel pool cooling heat exchangers. Figure 5.1.1 of the Licensing Report is a schematic representation of the computer model used to model the RHR and Fuel Pool Cooling system flow paths and does not conflict with the referenced FSAR figure.

QUESTION 5 Section 5.1.1 of the submittal states that "The upper containment pool does not contain any fuel while the plant is operating." Verify that this means that no spent fuel will be in the upper containment pool in any-operating mode other than the refueling mode.

RESPONSE

MP&L's position is that' prior to return to a reactor restart following refueling operations all spent fuel will be removed from the UCP..

During refueling operations the plant may be in either Operational Condition 4 - Cold Shutdown Mode or Operational Condition 5 - Refueling Mode.

Further delineation of MP&L's position is presented in CGNS FSAR Gection 6.2.7.3.3 " Inadvertent Dump" and MP&L's letter to the NRC AECM-85/0289 dated September 12, 1985.

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QUESTION 6 With respect to a gate drop accident, it is not clear whether damage to

.any fuel bundles has been assumed. Since the fuel bundle handles extend above the top of the racks (Refer to submittal Figure 3.6), it'should be assumed that the gate will impact the top of the fuel bundles. Therefore, provide the results'of an analysis of'the consequences of damaging the maximum number of fuel bundles that can be hit with the gate's " smallest cross-sectional dimension" and also those hit by the gate rotating and landing on the side with the " largest cross-sectional dimension".

RESPONSE

MP&L has performed a preliminary evaluation of a postulated drop of the spent. fuel pool gate onto the high density racks to determine the potential for fuel damage. This preliminary analysis supports the conclusions stated in MP&L's final response to NUREG 0612 (MP&L letter AECM-82/149, dated May 4, 1982).

The 1982 heavy load assessment assumes that the gate initially falls vertically-and impacts 10 fuel bundles. Damage to fuel cannot be precluded for the initial straight drop impact. The gate then rotates and topples over and impacts at least 39 other fuel bundles. No fuel is damaged in this second impact since the energy absorbed by the 39 fuel bundles does not result in strains in excess of 1%.

With the original storage racks (7-inch pitch within racks, 12-inches between racks), the maximum number of fuel bundles struck in the initial impact-is 10. With the high density racks (6.25-inch pitch), the maximum number of fuel bundles struck in the initial impact is 11. Consequently, the number of failed rods and the resulting doses would be greater by a factor of 11/10 = 1.1. For the high density racks, the number of-fuel bundles struck in the second impact would be 48. Consequently, the energy per bundle would be less, anu fuel damage is not likely.

Since fuel damage in the initial impact cannot be precluded, administrative controls are required to prevent moving the gate over locations containing stored fuel.

Technical Specification 3.9.7 prohibits the travel of loads in excess of 1140 pounds over fuel assemblies in the spent fuel or upper containment pool storage racks. Since the high density racks provide storage locations near the gate and the gate storage position, some locations shall be vacated during gate movements.

A dropped gate could impact vacant locations in a partially loaded rack; however, the analysis in the Licensing Report demonstrates that the integrity of the rack is maintained; therefore, fuel damage could not result from structural failure of the rack.

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