ML20054C621

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Forwards Amend 3 to 811218 Spent Fuel Pool Mod in Response to 820304 Questions
ML20054C621
Person / Time
Site: Farley Southern Nuclear icon.png
Issue date: 04/05/1982
From: Clayton F
ALABAMA POWER CO.
To:
Office of Nuclear Reactor Regulation
References
TAC-47523, NUDOCS 8204210437
Download: ML20054C621 (12)


Text

f Miihng Address 9

Af arwna Power Company 600 thrth 18th Street Po',t Of fice [lon 2041

[hrmingham, Alabama 35291 1

T e le pt'one 205 783 0081 F. L. Clayton, Jr.

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Alabama Power n-wun, u ae w>.m April 5, 1982 Docket No. 50-364 g,j

\\r Director, Nuclear Reactor Regulation

.8 U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Attention:

Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Unit 2

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Spent Fuel Pool Modification Amendment 3 Gentlemen:

In response to your letter of March 4,1982, enclosed is Amendment No. 3 to Alabama Power Company's Spent Fuel Pool Modification submitted on December 18, 1981 and amended in February 1982 and March 1982.

This amendment responds to the questions identified in Enclosures 2, 3 and 4 of your March 4,1982 letter.

In addition, page III-7 was revised to incorporate the response to Question #5 of Enclosure 4.

If you have any questions, please advise.

Yours very truly, j-. 3

'qj/ /p/g, Aw g,

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3 F. L. Clayton,<r.

FLCJ r/ J AR :j c -D7 Enclosure cc:

Mr. R. A. Thomas SWORNTOANDjUBSCRIBE EF.0 E Mr. G. F. Trowbridge ME THIS /A.

DAY OF Mr. J. P. O'Reilly 1982.

Mr. E. A. Reeves Mr. W. H. Bradford

. /v W

/ /

Notafy Public My commission expires:

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/O-22-75 8204210437 820405 DR ADOCK 05000364 PDR J

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FARLEY NUCLEAR PLANT UNIT 2 SPENT FUEL POOL MODIFICATION AMENDMENT 3 REVISION INSERTION INSTRUCTIONS April 5, 1982 Section Page Instruction' III III-7 Replace Tab Response to Add NRC Questions Q

Q-ll Add Q

Q-12 Add Q

Q-13 Add Q

Q-14 Add Q

Q-15 Add Q

Q-16 Add Q

Q-17 Add Q

Q-18 Add Q

Q-19 Add t

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to the spent fuel pool 10 days after the emergency shutdown.

3.

105'E component cooling water temperature.

Results:

No. of Total Spent Fuel Cooling Trains Heat Load Pool Bulk Operation (los Stu /hr)

Temeerature (*F) 1 30.384 158 2

30.384 131 The computer code BPCOL is used to analyse the natural circulation cooling of the spent fuel in the event of a loss of all external means of cooling for the spent fuel pool.

BPOOL is a proprietary program of NAI.

The code is based on the assumption that boiling takes place near the top of the fuel channel.

BPOOL evaluates the saturation properties of the coolant on the basis of the static pressure at the top of the storage racks.

These properties include water density, temperature, and steam density.-

The steam is assumed to separate and flow out of the pool.

The water at the saturation temperature corresponding to the pressure at the top of the racks flows downward to the inlet of the storage racks.

The static pressure at this location is higher than the pressure at the top of the storage racks and as a result the fluid is subcooled as it enters the fuel assembly.

The fluid becomes less dense as it passes up the fuel channel.

Near the top of the fuel channel the fluid reaches saturation conditions and net boiling occurs.

The computer cock, BPOOL, assumes a loss of all external means of cooling, but it should be noted that the Earley spent fuel pool cooling system is redundant and single failure-proof.

Under normal conditions, voiding between fuel assemblies is highly unlikely because these spaces are not sealed to keep out water.

Holes are provided at the top and bottom of each inner can to permit a definitive flowpath for circulation of water in these spaces.

III.l.5.(4)

Potential Fuel and Rack Handling Accidents The high-density poison racks are of a free-standing design, utilizing bottom support pads, resting on the floor of the spent fuel pool.

The installation of the high-density racks I

will include removal of the existing 13-in. center storage racks.

The high-density racks will be installed wet since there is spent fuel in the storage pool.

l The following is a sequence of events for installing the high-density poison racks.

Phase I Install and test a temporary crane for handling the existing racks and the high-density racks.

The AMEND 3 4/82 l

l III-7 l

ENCLOSURE-2 OF NRC LETTER DATED MARCH 4, 1982 NRC Question 1:

In section III.I. 2 (2) of your fuel pool modification report you state that the size of the load that can be handled over the spent fuel pool when fuel is in the pool is limited to 3,000 lb by Farley Unit 2 Technical Specifications.

In Section IV. (1).b of the report you state that the spent fuel racks are designed to withstand a fuel bundle drop from 42 in. under various conditions and a 9-in. gate drop.

Provide the following information to assure your Technical Specifications are sufficient to cover all cases since your Technical Specifications do not impose a height restriction.

a.

Verify that the load drops identified in Section IV. (1).b of your report do not have a higher possible kinetic energy than that assumed in your accident analyses used in Meveloping the Technical Specifications.

b.

Verify that when considering a load drop the weight of the handling fixtures was used in your analyses.

c.

For loads lighter than one fuel assembly verify that their lifting height will not result in a higher kinetic energy if dropped than the maximum used in your accident analyses.

Also specify how the height of lighter loads will be limited to an acceptable elevation.

APC Response:

a.

The Parley Unit 2 Technical Specification, Section 3.9.7.1, prohibits loads in excess of 3,000 lb from travel over fuel assemblies in the storage pools.

This load bounds the spent fuel assembly with control rod and handling fixture (1992 lb).

The fuel assembly drop is analyzed at a drop height of 42 in., which bounds the actual drop of 39.5 in.

The drop height is limited by a high level limit switch on the lift in combination with a long handling tool.

Administrative controls prohibit a gate (3600 lb) from being carried over spent fuel.

However, the racks are designed to withstand a gate drop from 9 in., which bounds the actual drop of 6 in.

The drop height is limited by a physical limitation in lifting capability.

The fuel assembly drop assuming a load of 3,000 lb at 42 in.

is the worst impact load condition, but is very conservative considering an actual drop of 1992 lb at 39.5 in.

Therefore, the load drops identified in Section IV. (1) b of the report do not have a higher kinetic energy than that assumed in the accident analysis.

Q-11 AMEND 3 4/82 I

b.

The loads associated with a fuel assembly with control rod (1650 lb) and handling fixture (376 lb) are bounded by the 3000 lb load specification, which was used in the drop accident analysis.

The handling fixture is not used with the 3600-lb gate.

The gate is designed with a clevis at top center, which the crane hook fits into for lifting and' relocating.

c.

As discussed in response to Question 1(a), the analyzed fuel assembly drop load of 3000 lb at 42 in. represents the worst fuel rack impact load condition; i.e.,

highest kinetic energy; of all loads that could be moved over the spent fuel pool.

The kinetic energy of loads lighter than one fuel assembly will be less than that calculated for the analyzed fuel assembly drop load descr3 bed above even if the lighter loads are released from the maximum possible elevation.

Therefore, limitations c:. lift height of these lighter loads are not required.

G AMEND 3 4/82 Q-12

ENCLOSURE-3 OF NRC LETTER DATED MARCH 4, 1982 NRC Question 281-1:

The February 1, 1982 Amendment request does not indicate any proposed modification of the spent fuel pool cooling and cleanup system (SFPCCS) in conjunction with the installation of high density poison spent fuel storage racks.

Describe what changes, if any, will be made to the SFPCCS to maintain the level of pool water purity with respect to visual clarity and activated corrosion and fission product buildup the same as for the original spent fuel storage capacity.

Assume that the number of defective fuel assemblies increases in proportion to the increased spent fuel storage capacity.

If no changes to the SFPCCS are to be made, indicate how the same level of pool water purity will be maintained.

APC Response:

No modifications to the spent fuel pool cooling and cleanup system will be made as a result of the installation of high density poison spent fuel storage racks.

The Farley spent fuel pool cleanup system is described in FSAR subsection 9.1.3.

The system's demineralizers and filters are designed to provide adequate purification to permit unrestricted access for plant personnel to the spent fuel storage area and maintain optical clarity of the spent fuel pool water.

The optical clarity of the spent fuel pool water surface is maintained by use of the system's skimmers, strainer, and skimmer filter.

Operating plant experience has indicated that crud release occurs immediately after refueling with the fuel pool water returning to normal conditions within a few days.

The increase in spent fuel assembly capacity of the spent fuel pool will not affect the optical clarity of the spent fuel pool water.

NRC Question 281-2:

Describe the samples and instrument readings and their frequency of measurement that will be performed to monitor the spent fuel pool water purity and need for demineralizer resin and filter replacement.

State the chemical and radiochemical limits to be used in monitoring the spent fuel pool water and initiating corrective action.

Provide the basis for establishing these limits.

Your response should consider variables such as:

boron, gross gamma and iodine activity, demineralizer and or filter differential pressure, demineralizer decontamination factor pH, and crud level.

APC Response:

A maximum radiation level and minimum decontamination factor (DF) will be established by procedure for the spent fuel pool (SFP) demineralizer.

Accordingly the demineralizer resins may be ex-changed at a more frequent interval than would be required for the original rack configuration.

The SFP filters may also be exchanged Q-13 AMEND 3 4/82

at a more frequent rate according to aP limitations in table Q-2.

The SFPCCS monitoring parameters, sampling frequency, limits to be maintained and their bases are shown in table Q-2.

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Q-14 AMEND 3 4/82

TABLE Q-2 (SHEET 1 OF 2)

PURITY MONITORING A.

Samples Parameter Frequency Limit Basis Boron Weekly 2000-3000 ppm The limitations on re-activity conditions during refueling ensure that:

1. The reactor will remain suberitical during core alternations, and 2.

a uniform boron concentration is main-tained for reactivity control in the water volume having direct access to the reactor vessel.

These limi-tations are consistent with the initial con-ditions assumed for the boron dilution incident in the accident analyses.

pH Weekly 4.0-4.9 To be consistent with the boron concentration.

Cl-Weekly 50.15 ppm Consistent with the acceptable limit for stainless steel systems.

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F Weekly 50.15 ppm Consistent with the acceptable limit for stainless steel systems.

Demineralizer DF Monthly ")

I}

I 210 Iodine specific DF to ensure the removal of 99 percent of the assumed 10-percent iodine gap activity released from the rupture of an irradi-ated fuel assembly.

B.

Instrument Parameter Frequency Limit Basis 1

Area Monitor RE-5 Continuous 15 mrem /hr N/A i

Filter AP Once per 20 psid Manufacturer's 8 hr recommendation.

Q-15 AMEND 3 4/82

TABLE Q-2 (SHEET 2 OF 2)

(a)

This will be done daily during forced oxygenation clean up prior to refueling.

(b)

If iodine inventory in the spent fuel pool is low; i.e.,

I-131

-4 s 1 x 10 pCi/g'; the demineralizer may be allowed to stay in service longer.

This may be at the discretion of the C&HP supervisor or C&HP section supervisor.

4 I

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l Q-16 AMEND 3 4/82

ENCLOSURE-4 OF NRC LETTER DATED MARCH 4, 1982 NRC Question 1:

What is the calculated nominal effective multiplication factor for the high density storage racks in their design configuration?

APC Response:

The nominal reactivity of the new Farley high density spent fuel pool (HDSFP) is determined using the Monte Carlo Code KENO-IV to be 0.9217*0.0044 with the 95-percent confidence interval ranging from 0.9127 to 0.9305.

The results have been adjusted to include the effects due to the presence of U-234 and Inconel spacer grids.

NRC Question 2:

What are the values of the calculated bias, calculational uncertainty, and mechanical uncertainty which are applied to the nominal effective multiplication factor?

APC Response:

The following summarizes the reactivity inventory of the Farley HDSFP.

Nominal K 0.9217*0.0044 eff Calculational bias

+0.0027

' Mechanical uncertainty (worst case geometry) +0.0159 Final K 0.9403*0.0044 eff c>

95-percent confidence interval 0.9315-0.9491 NRC Question 3:

What value of water temperature yields the maximum reactivity?

What value was used in the calculations?

APC Response:

A temperature reactivity sensitivity study covering the range from 68 F to 212 F shows that the effective multiplication factog has the highest value at the lowest reasonable pool temperature (68 F).

The nominal effective multipfication factor reported in response to Question 1 is for the 68 F condition.

Q-17 AMEND 3 4/82

NRC Question 4:

What organization performed the criticality calculations for the high density storage racks?

Has the organization benchmarked the versions of the KENO-IV and PDQ-7 codes which are operating on their computer system against critical experiments or does the reference to benchmarking refer to some other organization's efforts on a different computer system?

APC Response:

The criticality calculations are performed by Nuclear Associates International (NAI), a consulting service of the Control Data Corporation.

Both the KENO-IV and PDQ-7 code packages have been installed and in use on the CDC CYBER systems for a number of years.

The AMPX/ KENO-IV model, which is the basic code package used in the calculation of the effective multiplication factor of the infinite rack array, has been benchmarked by NAI against earlier ORNL critical experiments measured by Dr.

E.

Johnson, at La Crosse BWR measured cold critical;'and the more recent critical experiments conducted by S.

R.

Bierman, et al, at the Batelle Pacific Northwest Laboratory as reported in PNL-2438.

Although the PDQ-7 model is mainly used to provide relative reactivity results for the several sensitivity studies conducted for the Farley analysis, it is also used to calculate the infinite array reactivity of the nominal configuration in part to cross check and verify the results of the Monte Carlo KENO-IV calculations mentioned above.

The PDQ-7 model has been in use at NAI for core physics and fuel management work since the late 1960's when the code was first released for industrial use.

This model has been used to calculate cold critical measured at a dozen pressurized water reactor and boiling water reactor power plants in the United States.

Furthermore, it has also been used in calculating the La Crosse BWR cold critical and the ORNL/ Johnson critical experiments.

NRC Question 5:

The thermal hydraulic analysis of the cooling of the spent fuel states that voiding between fuel assenblies is not possible because these spaces contain poison plates.

This is not obvious from figure II-4.

1s there no coolant between fuel assemblies?

APC Response:

Under normal conditions, voiding between fuel assemblies is highly unlikely because these spaces are not sealed to keep out water.

Holes are provided at the top and bottom of each inner can to permit a definitive flowpath for circulation of water in these spaces.

AMEND 3 4/82 Q-18

O Voiding in the space between fuel channels can occur only when there is boiling in the fuel cavity itself, as might occur if all external means of cooling were lost.

Under these conditions, the voids in the volume between the fuel channels are always less than those in the fuel cavities and would not increase the reactivity of the

~

storage arrangement.

When the fuel cavities have no voids, the temperature is low enough to preclude voiding in the space between fuel channels.

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Q-19 AMEND 3 4/82