Information Notice 1992-21, Spent Fuel Pool Reactivity Calculations

Spent Fuel Pool Reactivity Calculations
	ML031200466
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	03/24/1992
From:	Rossi C; Office of Nuclear Reactor Regulation
To:	;
References
	IN-92-021, NUDOCS 9203180053
	Download: ML031200466 (6)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 March 24, 1992 NRC INFORMATION NOTICE 92-21: SPENT FUEL POOL REACTIVITY CALCULATIONS

Addressees

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information

notice to alert addressees to potential errors in reactivity calculations for

spent fuel pools. It is expected that recipients will review the information

for applicability to their facilities and consider actions, as appropriate, to

avoid similar problems. However, suggestions contained in this information

notice are not NRC requirements; therefore, no specific action or written

response is required.

Description of Circumstances

On February 14, 1992, the NRC was notified by Northeast Utilities of a discrep- ancy between reactivity calculations performed for the Millstone, Unit 2, spent

fuel pool by ABB Combustion Engineering (CE) and the licensee's contractor

(Holtec). The licensee has indicated that the k calculated by Holtec was

approximately 5 percent higher than that previou¶ly calculated by CE.

The NRC has recently learned that Houston Lighting and Power (HLP) has identi- fied a discrepancy between the reactivity calculations performed for the South

Texas, Unit 1, spent fuel pool by Pickard, Lowe and Garrilck (PLG) and the

licensee's contractor (Westinghouse). The licensee has indicated that the k f

calculated by Westinghouse was approximately 2 to 2.5 percent higher than that

previously calculated by PLG.

Boraflex is utilized as a neutron absorber between spent fuel pool rack cells

in both the Millstone, Unit 2, and South Texas, Unit 1, spent fuel pools.

Discussion

The computer code analyses performed by CE to predict neutron transport for the

Millstone, Unit 2, spent fuel storage racks used the two-dimensional, discrete

ordinates code DOT. CEPAK was used to generate the neutron cross sections for

DOT. The computer code analyses performed by Holtec used KENO (Monte Carlo

method). The source of the discrepancy between the CE and Holtec calculations

has been attributed by CE to two approximations made in the generation of

eutron ros sections. First, a transport cross section was used by CE as

( 92318005 2_

I ,I

IN 92-21 March 24, 1992 an approximation for the total cross section. While this approximation is

valid for most materials, it is not valid for materials having large thermal

cross sections. Therefore, applying this approximation to regions containing a

strong neutron absorber (such as Boraflex) results in an overestimation of the

neutron absorption and a corresponding lower calculated k in that region.

Second, a geometric buckling term corresponding to a sparsly populated and

weakly absorbing (unpoisoned) array was utilized by CE as an approximation of

buckling in the highly absorbing configuration. This approximation, however, is not valid for the specific configuration found in the Millstone racks where

the assembly pitch is small and the fuel assembly is completely surrounded by a

strong neutron absorber. After these approximations were corrected, the

results of the CE analyses were in good agreement with Holtec's.

The original computer code analyses performed by PLG to predict neutron

transport for the South Texas, Unit 1, spent fuel storage racks used the

two-dimensional diffusion theory code PDQ. LEOPARD was used to generate the

cross sections for PDQ. Computer code analyses performed by Westinghouse

utilized KENO (Monte Carlo method). The lower value of k f calculated by

PLG has been attributed by HLP to the inaccuracies inherefltin using diffusion

theory to predict neutron attenuation through a thin region that strongly

absorbs neutrons (such as Boraflex).

Both the CE and PLG methodologies had been benchmarked against criticality

experiments-that have been reported-to-closely represent the characteristics

of the spent fuel storage racks. However, it should be noted that the number

of criticality experiments that included a strong neutron absorber (such as

Boraflex) was limited.

This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation (NRR) project manager..

/r~les E.Rossi, Drector

Division of Operational Events Assessment

Office of Nuclear Reactor Regulation

Technical contacts: Jack Ramsey, NRR

(301) 504-1167 Larry Kopp, NRR

(301) 504-2879 Attachment: List of Recently Issued NRC Information Notices

I

Attachment

IN 92-21 Match 24, 1992 v

Page 1lof I

LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

92-20 Inadequate Local Leak Rate 03/03/92 All holders of OLs or CPs

Testing for nuclear power reactors.

92-19 Misapplication of Potter & 03/02/92 All holders of OLs or CPs

Brumfield MDR Rotary Relays for nuclear power reactors.

92-18 Potential for Loss of Re- 02/28/92 All holders of OLs or CPs

mote Shutdown Capability for nuclear power reactors.

during A Control Room Fire

92-17 NRC Inspections of Pro- 02/26/92 All holders of OLs or CPs

grams being Developed at for nuclear power reactors.

Nuclear Power Plants in

Response to Generic

Letter 89-10

92-16 Loss of Flow from the 02/25/92 All holders of OLs or CPs

Residual Heat Removal for nuclear power reactors.

Pump during Refueling

Cavity Draindown

92-15 Failure of Primary System 02/24/92 All holders of OLs or CPs

Compression Fitting for nuclear power reactors.

92-14 Uranium Oxide Fires at Fuel 02/21/92 All fuel cycle and uranium

Cycle Facilities fuel research and development

licensees.

92-02, Relap5/Mod3 Computer Code 02/18/92 All holders of OLs or CPs

Supp. 1 Error Associated with the for nuclear power reactors.

Conservation of Energy

Equation

92-13 Inadequate Control Over 02/18/92 All holders of OLs or CPs

Vehicular Traffic at for nuclear power reactors.

Nuclear Power Plant Sites

OL = Operating License

CP = Construction Permit

IN 92-21 March 24, 1992 an approximation for the total cross section. While this approximation is

valid for most materials, it is not valid for materials having large thermal

cross sections. Therefore, applying this approximation to regions containing a

strong neutron absorber (such as Boraflex) results in an overestimation of the

neutron absorption and a corresponding lower calculated k in that region.

Second, a geometric buckling term corresponding to a spar!ety populated and