ML20073R968

Proposed Tech Specs Revising Section 5.3.1, Fuel Assemblies, Per GL 90-02,suppl 1

ML20073R968
Person / Time
Site:	Sequoyah
Issue date:	02/07/1994
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20073R967	List: ML20073R962 ML20073R968
References
GL-90-02, GL-90-2, NUDOCS 9402160203
Download: ML20073R968 (9)

Text

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.' ENCLOSURE 1 i i

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} PROPOSED TECHNICAL SPECIFICATION CHANGE 1

l SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 )

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5.3 REACTOR CORE FUEL ASSEMBLIES  :

5.3.1 The ctor al [contai AI3 fue assemb s W h4 ach fu 1 assembly ntaini 264 f e1 rods ad wit ircalo 4. E fuel - d sha have a minal ctive el leng of 144 'nches.

he in' ial cor loadi g39 l shall iave iaximum nrichm of 3.1 eight ercent -235. J load el sha) be milar physic design t the ip tial c e loadip and all have ajaxiq' .. enric ient of weight rcent F-235. g333 SE( fMrr I CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All

. control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM

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DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of i the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

. c. For a temperature of 650 F, except for the pressurizer which is 680 F.

VOLUME

, 5.4.2 The total water and steam volume of the reactor coolant system is 12,612 + 100 cubic feet at a nominal T avg f 525 F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

SEQUOYAH - UNIT 1 5-4 Amendmentt{o,.045, 144 August 1, 4,9

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FUEL ASSI21BLIIS  ;

5.3.1' The reactor shall contain 193 fuel assemblies. Each assembly

[ shall consist of a matrix of zircaloy clad fuel rods with'an initial-i composition of natural or slightly enriched uranium dioxide as fuel

• j material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved' applications i of fuel rod configurations, may be used. Fuel assemblies shall be limited to those' fuel. designs that have been analyzed-with applicable NRC

staff-approved codes and methods, and shown by tests or analyses to ,

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comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be-placed l

in nonlimiting core regions.

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! .0ESIG'N FEATURES I

i 5. 3 REACTOR CORE

{ FUEL ASSEMBLIES 5.3.1 T e re or c ha conta' 93 ei ass lies 1th each f l assem cop aini 64 f rods lad Zi -

c uel shal} av a nominal ac ve fu len f134 inch . Th niti i core R37l

lo ing hall ave a ximum,enric . t of .15 w ght cent 35. eldd f 1 s all simil in p esign/ o th / nitia core ading

hall have sical'f maximuin enric ment 5.0 ,eight percen -235. R125

) CONTROL R0D ASSEMBLIES l

j 5.3.2 The reactor core shall contain 53 full length and no part length

! control rod assemblies. The full length control rod assemblies shall contain i a nominal 142 inches of absorber material. The nominal values of absorber i

material shall be 80 percent silver, 15 percent indium and 5 percent cadmium.

All control rods shall be clad with stainless steel tusing.

9 5.4 REACTOR COOLANT SYSTEM

! DESIGN PRESSURE AND TEMPERATURE i

j 5.4.1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 <

i of the FSAR, with allowance for normal degradation pursuant.to the l applicable Surveillance Requirements, i

l b. For a pressure of 2485 psig, and f

l c. For a temperature of 650 F, except for the pressurizer which is j

680 F.

VOLUME i

i 5 4.2 The total water and steam volume of the reactor coolant system is 12,612 ! 100 cubic feet at a nominal T avg f 525'F.

l 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

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FUEL ASSEMBLIES  ;

t 5.3.1 The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of zircaloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel  :

material. Limited substitutions of zirconium alloy or stainless steel l

filler rods for fuel rods, in accordance with NRC-approved applications-of fuel rod configurations, may be used. Fuel assemblies shall be-limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests.or analyses to comply with all fuel safety design bases. A limited number of lead test' assemblies that have not completed representative testing may be placed in nonlimiting core regions.

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i ENCLOSURE 2 l

PROPOSED TECHNICAL SPECIFICATION CHANGE l -

SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 l

(TVA-SQN-TS-93 19)

DESCRIPTION AND JUSTIFICATION FOR '

! THE SUBSTITUTION OF FILLER RODS FOR l

FUEL RODS AS SPECIFIED IN GENERIC LETTER 90-02, SUPPLEMENT 1 f t

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Deacription of Change TVA proposes to modify the Sequoyah Nuclear Plant (SQN) Units 1 and 2 technical specifications (TSs) to revise Section 5.3.1, " Fuel Assemblies," in accordance with the guidance of Generic Letter 90-02, Supplement 1. This change will permit the substitution of zirconium alloy or stainless steel filler rods for fuel rods in fuel assemblies. i l

l Reason for Changs l This change will allow the substituti n of filler rods for fuel rods in fuel assemblies. The change is desired to permit the timely removal of fuel rods that are found to be leaking or are determined to be the probable source of future leaks.

Justification for Change The requirements for fuel assemblies specify the quantity of fuel

! assemblies and the number of fuel rods per assembly. Flexibility to deviate f rom the number of fuel rods is desirable to permit timely ,

l removal of fuel rods that are found to be leaking during a refueling  !

outage or are determined to be the probable sources of future leakage. j This improvement in SQN's fuel performance program will provide for l reductions in future occupational radiation exposure and plant radiological releases.

As stated in Generic Letter 90-02, Supplement 1, the substitution of .

j filler rods for fuel rods is acceptable when justified by an NRC-approved  !

methodology that has been explicitly approved in a written safety evaluation, or a plant-specific TS basis. The NRC-approved methodology will demonstrate that existing design limits and safety analyses criteria are met in advance of the next operating cycle.

LiYironmental_ Impact Evaluation The proposed change request does not involve an unreviewed environmental question because operation of SQN Units 1 and 2 in accordance with this change would not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by NRC's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board.

2. Result in a significant change in effluents or power levels.

3. Result in matters not previously reviewed in the licensing basis for SQN that may have a significant environmental impact.

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Enclosure 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PIANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-93-19)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION

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l Significant Hazards Evaluation-l l t TVA has evaluated the proposed technical specification change and has j determined that it does not represent a significant hazards consideration  ;

based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah j Nuclear Plant (SQN) in accordance with the proposed amendment.will not:  ;

1. Involve a significant increase in the probability or consequences of '

an accident previously evaluated.

i The substitution of filler rods will be justified using NRC-approved methodology. This methodology will demonstrate'that the existing. -

design limits and safety analyses criteria are met. .Therefore, the l proposed change does not increase the consequences of.an accident. l previously analyzed.

2. Creat the possibility of a new or different kind of' accident from  !

i any aviously analyzed.

Th- ,coposed change involves the substitution of filler rods for fuel  !

rods. This substitution requires the utilization of NRC-approved-  !

methodology. This methodology will ensure that the specific analyses i will not cause any new or-different kind of accident from that. j previously analyzed.

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3. Involve a significant reduction in a margin of. safety. l i

The substitution of filler rods for fuel rods would result in less active fuel in the core. Therefore, the amounts of radiological effluents that may be released offsite would.not increase. The NRC-approved methodology by which any reanalyses would be performed a'. ready accounts for the affects on grid strength or the mass, stiffness, and fundamental frequency of the fuel assembly during seismic and loss-of-cooling accident conditions. Thus, the margin of l

safety is not reduced when substituting filler' rods for' fuel _ rods.

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