ML20083B880

Forwards Response to NRC 910528 Request for Addl Info Re Reracking of Spent Fuel Pool.Holtec Rept HI-90477, Heat Loss to Ambient from Spent Fuel Pool:Correlation of Theory W/Experiments Also Encl.Rept Withheld (Ref 10CFR2.790)

ML20083B880
Person / Time
Site:	Three Mile Island
Issue date:	09/18/1991
From:	Broughton T GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19302F176	List: ML20083B880 ML20083B885
References
C311-91-2075, NUDOCS 9109250241
Download: ML20083B880 (12)

Text

,

y g GPU Nuclear Corporation 4 Post Office Don 480 Houte 441 South

%ddittown, Pennsylvania 170s7 0191 717 944 7(321 Tflf A 84 2366 Wr'ter's Direct Dial Numt>et-(717) 948-8005 September 18, 1991 C311-91-2075 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. UPR-50 Docket No. 50-289 l

! Response to Request for Additional Information Regarding Spent fuel Pool Rerack l

Enclosed is the GPU Nuclear response to a request for additional information contained in the NRC's letter dated May 28, 1991, regarding reracking of the spent fuel pool.

In accordance with 10 CfR 2.790, an affidavit certifying the proprietary nature of the enclosed Holtec Report HI-90477 is provided, it is requested that this document be withheld from public disclosure for the reasons delineated in the affidavit.

If any additional information is required, please advise, f, Sincerely, I

hk T. G. Br Qr..-

hton Vice President and Director, THI-l DJD/amk i>

ytylj

Enclosures:

1) IMI-l Response to NRC Request for Additional Information pga Regarding Spent fuel Pool Reracking cg pg 2) Affidavit Certifying Proprietary Nature of HI-90477 y'd cei< e- 3) HI-90477, Holtec Report, " Heat loss to the Ambient from gg ' " ' ' d 'l Spent Fuel Pool: Correlation of Theory with Experiments" "I

[f o cc: Thl-1 Senior Project Manager c4 Region 1 Administrator EEc_ TM1 Senior Resident inspector GPU Nuclear Ccapo@an es a wted ary of Gener# ot%: Ubutes Co*v ranon

[]a Of. ;

C311-91-2075 Page 1 UKL911!RL1 1MI-l Response to NRC Request for Additional Information 8tgardina Spent fuel Po.nl Rerath

1. ThermalMdnttLit Analyses

a. Document the validity of the equation (eq., 5-2 on page 5-9) which is used to determine the heat removal rate by the heat exchanger, Q, .

Show how P, the temperature effectiveness, was determined.

Rengn_ig The valiriity of eq. (5-2) can be readily verified as follows:

The heat duty of a heat exchanger can be expressed as:

0 , - W. C, (t, - t,) (1) where:

W, - coolant flow rate, lb/hr C, = coolant het.t capacity, Blu/lb *F t, coolant outlet temperature, *f t, - coolant inint temperature, 'F If we define the heat exchanger temperature effectiveness as p - id (2)

T - t, where T - fuel pool water inlet temperature, then eq. (1) can be [

written as -

Qo, - W, C, p (T - t.)

This is a standard definition of temperature effectiveness as used in

" Standards of Tubular Exchanger Manufacturers Asso:iation", 7th

. Edittoa, ifMA, 1988.

b. Show how Q (t), the heat generated by the reconi.ly discharged fuel, is handled in your calculation of bulk pool coolant temperature, i.e., either as a uniformly distributed heat load over the time of unloading, or as an added heat load each time a spent fuel assembly is loaded into the pool.

l C3'l-91-2075 Page 2 RtRDRE 1he heat generated by the recently discharged fuel is handled as an added heat load each time a spent fuel assembly is loaded into the pool. The discharge to the pool is assumed to be at the rate of six assemolies per hour, which is an upper bound rate for the rate of actual fuel transfer,

c. The guidelines of Section Ill.l.h of Standard Review plan 9.1.3,

" Spent fuel Pool Cooling and Cleanup Services," specify that decay heat generation for normal maximum and abnormal maximum heat loads consider the off-loading of refueling load or full core of spent fuel assemblies to be complete after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay. Justify any deviation from this time.

EUNE 1he decay heat load calculations uerformed for the 1M1-1 Spent fuel Pool Cooling System for each of tle five (5) scenarios described in Section 5.4 and Table 5.4.3 of HI-89407 assume 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of thermal decay in the reactor, lhis is consistent with the intent of Section Ill.l.h of Standard Review Plan 9.1.3, and with the original design basis for the 1MI-l Spent fuel Cooling System as described in fSAR Section 9.4.

d. Show how you obtained the value(s) you used for the diffusion coefficient, h,(0), in calculating M (the mass evaporation rate).

Also explain the term " humidity ratio" when calculating H.

EMPE1H Response to this item is contained in (f) below,

e. Show how you derived the ambient temperatures (t ) used to determine the temperature difference (0) in equation 5-3 for the heat lost from the pool, 0, , due to evaporation and heat transfer to the air.

Justify any , assumptions used to derive t, and O.

Ets m a Response to this item is contained in (f) t'elow,

f. Calculation of Q,, involves two competing processes, i.e., diffusion of pool water into the air above the pool and heat transfer from the pool surface to the surroundings. In equation 5-3, both appear to involve the total pool surface area. However, if diffusion were to I occur across the entire pool surface, the water vapor formed would l

tend to inhibit heat transfer to the surroundings. Therefore, justify use of this equation or provide suitable corrections, in the equation, and in the values obtained when using it.

Page 3 l hm%R 1he htat transfer mechanisms of evaporation and convection are occurring concurrently over the entire pool surface. Convective heat loss also occurs at a reduced efficiency. The actual convective heat transfer coefficient is determined by experiment.

The methodology for quantifying the evaporation heat loss component is described in the enclosed Holtec Report entitled " Heat loss to the Ambient from Spent fuel Pool: Correlation of Theory with Experiments", Report H1-90477, which is being forwarded to the Commission on a proprietary basis. This document explains the analytical underpinnings of the calculation procedure and the pool experiments conducted to confirm their validity.

g. Discuss the determination of the initial value,1,,, of the pool water temperature and the validity of the equation used to evaluate it (shown on page 5-11 of the licensing report).

Bawns Prior to the beginning of discharge of fresh spent fuel, the heat generation rate in the pool is represented by the term PC0f4S. The steady state temperature of the pool water,1,,, is determined by equating PC0 tis to the heat removal rate of the cooler, which referring to eq. (5-2) is W C,p (T,, - t,)

Therefore, l., - 1D10 . , ,

W. C, p Where:

t, = coolant inlet temperature W,C, - thermal flow rate of coolant

h. Show how you used Holtec's validated numerical integration code, "Onepool" to solve equation 5.1, C dt/dz - Q - Q,,, which you used to t

determine the spent fuel pool bulk coolant temperature. Account for all the terms used and justify any assumptions made, accordingly,

C311-91-2075 Page 4 Runne ONEP00L utilizes a standard algorithm for integrating the single order differential eq. (5-1) on page 5-8 of the licensing report.

This program has been Q. A. validated by running typical bulk pool temperature problems, lhe validation manual is maintained under lloitec International's Q.A. system.

i. Show how the QA validated numerii;al quadrature code was used to determine the time-to-ball shown on Table 5.5.2 of the licensing report. Show the minimum times to boil in the event no makeup water is available in cases 1 (maximum normal heat load) and 4 (maximum abnormal heat load).

EUMnic The equation for time-to-boil is a first order differential equation .

c presented on page 5-12 of the licensing report. This equation can be integrated using any standard numerical quadrature subroutine, lloltoc utilizes Runge-Kutta subroutine DEIORK listed in the text

" fortran Scientific Subroutine Library," Wiley (NY).

The accuracy of the results has been confirmed by hand calculations.

The results presented in Table 5.5.2 are for G - Q_gpm (no make-up water flow).

Finally, we observe that the time-to-boil ica.lts are quite conservatively calculated in as much as boiling is assumed to occur with the entire water mass to be at 212 'f. In reality, the pool water near the bottom of the pool will be at approximately 260 'f (due to hydrostatic head) when the top of the pool is boiling. -

j. Report at what point (i.e., volume of water left .. Sol) spent fuel elements just start to be uncovered on figures 5.5.9 - 5.5.13.

EMD2B1R Table A-1 provides the time for the level of water to drop to the elevation of the tip of the racks subsequent to loss-of-all-forced-cooling. No make-up water is assumed to be available. Loss-of-cooling is assumed to occur at the instant when the pool water temperature is at its peak value. Results for all five cases are provided.

C311-91-2075 Page 5 lable A-1 lime Af ter loss-of-Cooling When Water Surface Drops to lop of Racks CutRVat>cr 11m_htL10 0) 1 257 2 263 3 305 4 135 5 141

k. In Table 5.4.1, " fuel Specific Power and pool Capacity Data." you define the "dimensionless decay power of "old" discharges" to be 0.1170. Explain this term and how it is used in the calculations, h3MA12 0.1176 is the ratio of the cumulative heat load of all previously stored fuel to the average specific power of one fuel assembly in the reactor. Thus, referring to Table 5.4.1, the cumulative decay heat from the inventory of the stored fuel is (0.1176) (60.3f 4 06) - 7.09 million Btu /hr. This heat load is the term PCONS which appears throughout section 5.8 of the licensing report.

1. What is the temperature rating of the ion exchanger in the spent fuel pool cleanup system? Are any other parts of cleanup system rated at any lower temperatures?

h120 Alt The cation demineralizers which remove fission products from the spent fuel pool water are designed for 120*f. lhe ion exchangers are downstream of the spent fuel pool cooler. Therefore, the maximum water temperature to which the resins will be exposed is less than 135 'f for the worst case abnormal condition (1 spent fuel poal cooler in operation) with fully loaded racks. This condition corresponds to Case 1 in Table 5.5.1 of L.icensing Report til 89407.

The resins are capable of withstanding 150 "f water for extended periods. The precoat filters which remove particulates from spent fuel pool water are designed for 200'f. The remainder of the liquid radwaste system is designed for a minimum of 1501 ,

m. Any proprietary documents which are furnished in response to questions above must be evaluated by the staff in order to determine whether the documents are entitled to be certified as proprietary.

Any documents which already have been so certified should be forwarded with the notice that they have previously been evaluated as proprietary with a citation attesting to the location of such certification, s

)- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _

in addition, if any of the methods you used in thermal / hydraulic I analysis of conditions in and around the 1MI-l spent fuel pool have been found acceptable by the staff, provide the appropriate citation (s) for those documents, also.

Re1PMit As referenced above in response to items 1.d. l.e, and 1.f Iloltec Report Hi-90477, *lleat loss to the Ambient from Spent iuel Pool:

Correlation of Theory with Experiment s" is forwarded for NRC review.

Holtec Report H1-90477 is considered proprietary and Enclosure 2 provides an affidavit certifying the proprietary nature, in accordance with 10 CfR 2.790.

The methods utilized in the thermal-hydraulic analysis of the 1Ml-1 pool are similar to those used in J. A. Fitzpatrick, Indian point Unit Two, Vogtle Unit 2, Millstone Unit One, Grand Gulf Unit One, Diablo Canyon Units 1 and 2. St. Lucie Unit One, Dyron Units 1 and 2, Braidwood Units 1 and 2, Pilgrim Urit One, Quad Cities Units 1 and 2, Enrico fermi Unit 2, Rancho Seco, Oyster Crack and others.

2. 11sy1103tLLIldling

a. The method you have chosen for complying with the guidelines of NUREG-0512. " Control of Heavy Loads at Nuclear power Plants,* is inadequate. Compliance with the guidelines of NUREG-0612 requires either evaluating potential heavy load drops to satisfy the criteria of Section 5.1 of HUREG-0612 or showing that the potential for a load drop is extremely small by employing a single failure proof heavy load handling system which complies with the guidelines of Section 5.1,6, " Single-f ailure-Proof Handling Systems." Your submittal does not address either set of guidelines completely, provide a submittal which complies, adequately, with either set of guidelines or suitable (

alternatives. The demonstration that a potential load drop is of low probability by the use of statistical probability arguments is not an acceptable means of meeting those objectives.

Runme Compliance - 'h the objectives of NUREG-0612 follows the guidelines contained 1 icction 5 of that document. These Section 5 guidelines of NUREG-06D specify measures to " provide an adequate defense-in-depth for handling of heavy loads near sp nt fuel...", and cite four major causes of load handling accidents, namely

1. operator errors

11. rigging failure til, lack of adequate inspection iv, inadequate procedures

0311-91-?075 page 7 1ho 1Hi-1 rerack program will provide adequate measures to minimize the occurrence of potential load drop acciderts. A summary of the measures specifically planned to deal with the major causes of load handling accidents is provided below.

Operal9LerIors: 1Hl-1 plans to provide comprehensive training to the rack installation crew using videotapes of the actual lif t rigs built for the rerack project.

Ringinglaihite: 1he lift rig as described in the licensing report, is single failure proof, and complies with the provisions of ANS!

14.6.

LRC1.lf_AdCWAl0_ju$piKl190: Current procedures require that prior to operating the fuel llanding Building Crane a daily / monthly inspection is performed. If the crane was inspected 6 months ago or longer, 6 yearly inspection is performed. The daily / monthly inspection checks the functional operating mechanisms; electrical and hydraulic components and connections; the hoist hooks for visual cracks or deformations; the main and auxiliary holt.t ropes for wear, twisting, stretching, broken strands; and the gear box oil levels.

The yearly insaection checks the main switch contacts and connections; t1e drum rollers; contactors, relays and other electrical components: motors; lights and over-travel interlocks; structural integrity of rails, bridge, trolley and stops; bridge and trolley gear boxes, drive wheels, and line shaft; the main hoist drum gear box, mechanical and electrical load brake; and hooks and ropes.

IDidf 0VAlt_ErQIfiKC.1: Operating procedures will cover the entire gamut of operations pertaining to the rerack effort, such as mobilization, rack handling, upending, lifting, installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance. Procedures will address handling of old racks as well as "

the new racks. The series of procedures planned for TMl-1 rerack effort will be based on procedures successfully implemented in other similar rerack projects.

In addition to the above, a complete inspection of the fuel llandling Crane and relubrication of moving parts prior to the start of reracking is planned. Safe load paths are being developed and sacrificial impact shields have been designed to strengthen regions of the fuel llandling Building operating floor. The design of the impact shields has been certified using finite element methods.

Table A-2, below, provides a synopsis of the requirements delineated in NUREG-0612, and our intended compliance.

C311-91-2075 Page 8 Table A-2 lleavy Load llandling Compliance Matrix (NURI'G-061/)

Gr_iltr_iOD CcMlllALKe

1. Are safe load paths defined for the Yes

.novement of heavy loads to minimize the potential of impact, if dropped, on the load on irradiated fuel?

2. Will procediares be developed to cover: Yes identification of required equipment, inspection and acceptance criteria recuired before movement of load, steps ant proper sequence for handling the load, defining the safe load paths, and special precautions?

3. Will crane operators be trained and qualified? Yes

4. Will special lifting devices meet the guidelines Yes ANSI 14.6-19787

5. Will non-custom lifting devices be installed Yes and used in accordance with ANSI B30.9-19717

6. Will the cranes be inspected and tested Yes prior to use in rcrack?

7. Docs the crane meet the intent of ANSI B30.2-1976 Yes and CMMA-707 The measures described in the original Itcensing submittal, in conjunction with those outlined above, fulfill the intent of the defense-in-depth approach required in NUREG-0612. The existing fuel Handling Duilding overhead crane and associated equipment was reviewed by NRC in NRC Safety Evaluation Report dated January 11, 1985, which concluded that the guidelines in NUREG-0612, Section 5.1.1 have been satisfied.

The above ap) roach is identical to the one utilized in the most recent reract project (Indian Point Unit 2), concluded in October, 1990.

C311-91-2015 Page 9

b. The discussion pro'<lded in Section !!!.C, entitled, " Installation Accident," of your fechnical Specification Change Request (TSCR) '

No. 201 states that the alternative to single failure proof crane design is satisfied but does not provide details so as to permit the staff to conclude that the 1MI-l fuel handling crane complies with the criteria of Apperdix C, " Modification of Existing Cranes." Provide the information necessary for staff review.

Response  !

The response to this question is contained in (a) above,

c. Further, in Section Ill.C of TSCR 201, you state that redundant special lifting devices will be used in movement of the storage racks. Such special lifting devices should be designed in accordance with the guidance of ANSI N14.6-1970. Show how these special lifting devices will conform to the requirements of the aforementioned standard or provide suitable justification for any deviation _from that guidance.

RuP3J112 The lifting device designed for handling and installation of the racks in the 1MI-l Pool A has redundancies in the lift legs, and lift eyes such that there are four independent load paths.

Failure of any one load bearing member would not lead to uncontrolled lowering of the load. The rig complies with all provisions of ANSI-14.6-1978, including compilance with the primary stress criteria, load testing at 150% of maximum lift load, and dye penetrant examination of critical welds.

The THI-l rig design is similar to the rigs used in the rerack of numerous other plants, such as Millstone Unit 1, Indian Point Unit Two, and J. A. Fitzpatrick,

d. In Section Ill.C of TSCR 201, you state that both technical specificatio" and administrative procedures will preclude the movement of a rack over any fuel. Show the minimum distances to be maintained between racks-and spent fuel, both horizontally and ,

vertically, during the removal and installation processes and how this separation will be accomplished.

Ef1RQDIR TMI-1 Technical Specification Section 3.11.6 imposes administrative limits on handling loads weighing in excess of 3,000 lbs. above the Spent fuel Pool operating floor (348 elev.)

to minimize the potential for heavy loads, if dropped, to impact irradiated fuel in the spent fuel pool. The safe load path shall 1

.._ _ __ _ ..__._._ _ __. _ .m_._ _ _ _ . _ _ _- _ _ _. _ ._ _ _

l C311-91-2075 )

page 10 l t

follow, to the extent practical, structural floor menibers, beams, etc. such that if the load is dropped, the structure is more likely to withstand the impact.

A fuel reshuffle scheme for fuel assemblies stored in the spent -

fuel pool is being developed which is predicated on the following  !

criterth: '

(1) No heavy load (rack or rig) with a potential to drop on an installed rack will have less than 40 inch lateral free zone clearance from active fuel. This clearance 1 will be ensured through safe load path procedures.

(2) Heavy loads are lifted in such a manner that the center  :

of gravity of the lif t loint is aligned with the center of gravity of the load acing lifted.

(3) Turnbuckles will be utilized to provide vertical adjustnient of the rack being lifted.  !

The above procedures and steps are similar to those employed in recent rerack crojects.

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UlfLQ1MBL2 P -iavit Certifying Proprietary Nature of HI-90477 t

y

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